Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

Nuclear Engineering and Design 308 (2016) 142–153

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Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx Hyeong-Yeon Lee ⇑ Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, Republic of Korea

h i g h l i g h t s  Comparison of elevated temperature design (ETD) codes was made.  Material properties and evaluation procedures were compared.  Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study.  Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted.

a r t i c l e

i n f o

Article history: Received 5 March 2016 Received in revised form 6 August 2016 Accepted 11 August 2016 Available online 6 September 2016 JEL classification: F. Reactor Components

a b s t r a c t The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted. Ó 2016 Elsevier B.V. All rights reserved.

1. Introduction Generation IV(Gen IV) nuclear reactor systems such as a sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), very high temperature reactor (VHTR), and gas-cooled fast reactor (GFR) operate at high temperature of the creep regime. Gen IV reactor systems have improved features, including sustainability, economics, safety, and proliferation resistance, compared to those of Generation III+ reactors currently under construction or operation around the world. The two major elevated temperature design (ETD) codes that are currently being used for the design of such Gen IV reactors are ASME Section III Subsection NH (ASME Boiler, 2015c) and RCC-MRx (RCC-MRx, 2013a). The materials with the same material names in the two codes are not actually the same because chemical compositions as well as some materials properties are quite different. For instance, the ⇑ Fax: +82 42 868 8739. E-mail address: [email protected] http://dx.doi.org/10.1016/j.nucengdes.2016.08.024 0029-5493/Ó 2016 Elsevier B.V. All rights reserved.

material of Mod.9Cr-1Mo (ASME Grade 91, hereafter ‘Gr.91’) steel is one of the listed materials in both ETD codes, but the chemical composition of Gr.91 steel in ASME-NH is different from that of RCC-MRx. Therefore, Gr.91 steel of RCC-MRx is different from that of ASME-NH in terms of the chemical compositions and material properties. In addition, the procedures for determining inelastic strain and creep-fatigue damage are basically the same in terms of concept, but there are some significant differences in the detailed evaluation procedures on the inelastic strain and creep damage between the two codes. The evaluation results from the application of the two ETD codes to the SFR components showed that there are no negligible differences in inelastic strain or creep-fatigue damage (Lee et al., 2007, 2012a, 2013). A comparison of the two ETD codes is necessary, but to the best of the author’s knowledge, there has been no systematic comparison thus far of the two ETD codes in terms of their procurement specifications for materials or evaluation procedures. A comparison study was conducted for the material properties. In the case of the material properties, comparisons were made

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Nomenclature EFU ETD FE Gen IV

elastic follow-up elevated temperature design finite element Generation IV

GFR LFR SFR VHTR

from the chemical compositions to the creep and fatigue material properties. The target materials in the present study are austenitic stainless steel 316 and Gr.91 steel, which are the two major materials in a PGSFR (Abram et al., 2012) and other SFRs such as a PFBR (Kumar et al., 2014) and JSFR (Wakai et al., 2015). Comparisons were also made for the design evaluation procedures of the two ETD codes on the evaluation of the deformation-controlled limits and load-controlled limits. The two ETD codes have some technical issues that should be addressed so that the design evaluations can be more reliable. The current version of the design code requires validation in terms of some material properties and design procedures. The contents of Tome 6 (RCC-MRx Tome 6, 2013c) of RCC-MRx and the code case (ASME Code Case, 2011) of ASME require validations before they are upgraded as part of the regular design rules. In this study, the differences in the material properties and the evaluation procedures were quantified based on the properties provided in the codes and through design evaluations according to the two design codes.

2. Elevated temperature design rules Elevated temperature design rules are used for the design evaluation of Gen IV nuclear energy systems. The Gen IV reactors under development in Korea are the Prototype Generation IV sodiumcooled Fast Reactor (PGSFR) and Nuclear Hydrogen Development and Demonstration (Chang et al., 2007), as shown in Fig. 1. The Gen IV reactors under development in Europe are the sodiumcooled fast reactor ASTRID, the lead-cooled fast reactor of MYRRHA, and the gas-cooled fast reactor of ALLEGRO, the schematics of which are shown in Fig. 2. ASME-NH is the main ETD code for the PGSFR and NHDD, whereas the RCC-MRx code is that for ASTRID, MYRRHA, ALLEGRO,

gas-cooled fast reactor lead-cooled fast reactor sodium-cooled fast reactor very high temperature gas-cooled reactor

and for some components in the fusion experimental reactor ITER under construction in France. 2.1. ASME Section III Subsection NH The overall design evaluation procedures of ASME-NH are shown in Fig. 3. If the wall temperature from the heat transfer analysis is higher than 427 °C for austenitic stainless steel (304 and 316), or higher than 371 °C for 2.25Cr-1Mo steel and Mod.9Cr-1Mo steel, a design evaluation should be conducted according to ASME-NH. Otherwise, a design evaluation should be conducted according to the ASME Section III Subsection NB procedures, as shown in Fig. 3. In the evaluation procedures of ASME-NH, the design limits on the load-controlled stress limits and deformation-controlled limits should be satisfied. In deformation-controlled limit checking, three items of inelastic strain, creep-fatigue damage, and buckling limits should be satisfied, as shown in Fig. 3. The predecessors of ASME-NH were the ASME Code Case for 1331-5, which was used in the design and construction of the FFTF (Fast Flux Test Facility), and ASME Code Case N-47, which has been used for the design and evaluations of the CRBR (Crinch River Breeder Reactor) and PRISM (Power Reactor Innovative Small Module). Code Case N-47 was upgraded to ASME Subsection NH in 1995. The material of Mod.9Cr-1Mo steel has been newly added from the 2004 edition. For a design evaluation of high-temperature reactors in Gen IV reactor systems, ASME Section III Division 5 Subsection HB Subpart B (ASME Boiler, 2015e) was developed. ASME Division 5 Subsection HB Subpart B can cover not only fast reactor systems such as a sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and gas-cooled fast reactor (GFR) but also a very high temperature gas-cooled reactor (VHTR). The code cases of N-253 (Construction of Class 2 or 3 Components) and the code case of N-499 (Use of

Reactor System

Hot Gas Duct Power Conversion Unit

(a) PGSFR (SFR)

(b) NHDD (VHTR)

Fig. 1. Generation IV nuclear reactors under development in Korea.

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H.-Y. Lee / Nuclear Engineering and Design 308 (2016) 142–153

(a) ASTRID (SFR)

(b) MYRRHA (LFR)

(c) ALLEGRO (GFR)

Fig. 2. Generation IV nuclear reactors under development in EU.

Thermal Analyses

Metal Temperature T > 427oC for 304SS, 316SS, Alloy 800H T > 371oC for 2.25Cr-1Mo, 9Cr-1Mo-V

No

Yes ASME - NB

ASME - NH Appendix T (Non-mandatory)

Limits on Load-Controlled Stresses (Time Dependent / Independent)

Limits on Deformation Controlled Quantities

Inelastic Strain Limits on LoadControlled Stresses (Time Independent)

Fatigue

Plastic strain, Creep strain

Creep-Fatigue

Buckling

Damage (Interaction)

Instability

Fig. 3. Design evaluation flow in ASME Section III Subsection NH.

SA-533 & SA-508 for Limited Elevated Temp Service) have already been merged to ASME Division 5 (ASME Boiler, 2015e). The current 2015 edition of ASME-NH is known to have merged into ASME Section III Division 5 from the next edition. 2.2. RCC-MRx The overall design evaluation procedures of RCC-MRx are shown in Fig. 4. If the wall temperature from the heat transfer analysis is higher than 450 °C for austenitic stainless steel of 316L(N) and 316L, higher than 425 °C for austenitic stainless steel of 304 and 304L, or higher than 375 °C for Mod.9Cr-1Mo (Grade 91) steel and 2.25Cr-1Mo (Grade 22) steel, a design evaluation should be conducted according to the ‘high temperature rules’ of RCC-MRx. Otherwise, the ‘low temperature rules’ should be used, which are analogous to those contained in RCC-M for light water reactors.

In the evaluation procedures of RCC-MRx, the design limits on the load-controlled stress and deformation-controlled limits should be checked as in ASME-NH. In deformation-controlled limit checking, requirements on a fast fracture as well as the three items of inelastic strains, creep-fatigue damage, and buckling limits should be satisfied. It should be noted that the ASME code basically does not assume a defect in the design, whereas the RCC-MRx code takes a fast fracture into account at the design stage. The fast fracture requirements on the fracture toughness of JIC should be satisfied. 3. Comparison of elevated temperature design codes 3.1. Chemical compositions Chemical compositions of the registered materials in ETD codes are provided in ASME Section II Part A (hereafter, ASME II-A)

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Thermal Analyses

Metal Temperature T > 450oC for 316(L)N, 316L T > 425oC for 304, 304L T > 375oC for Gr.91, Gr.22

No

Yes RCC-M

RCC-MRx

Limits on Load-Controlled Stresses (Time Dependent/Independent)

Limits on Deformation Controlled Quantities

Inelastic Strain Limits on LoadControlled Stresses (Time Independent)

Plastic strain, Creep strain

Fatigue

Creep-Fatigue

Buckling

Damage (Interaction)

Instability

Fast Fracture

Notch Integrity J < JIC

Fig. 4. Design evaluation flow in RCC-MRx.

(ASME Boiler, 2015a) and RCC-MRx Tome 2 (RCC-MRx, 2013b). The chemical compositions of austenitic stainless steel 316 and Mod.9Cr-1Mo steel are compared in Tables 1 and 2, respectively. As shown in Table 1, 316 stainless steel (hereafter ‘316SS’), 316H, 316L(N), and 316L are included in the ASME code, whereas only 316L(N) and 316L are in RCC-MRx. In the ASME code, it should be noted that the 316SS in ASME-NH is different from that of ASME Section II Part A as shown in Table 1. It is important to note that 316SS in ASME-NH is actually closer to 316H rather than 316SS in ASME II-A in terms of the chemical compositions. A significant difference between Sections II and III can be found in the specified range of carbon content. The carbon content of 316SS in Section II is 0.08 wt.%, whereas that of Section III is 0.04–0.06 wt.% C. Therefore, 316SS in Section II will have an improved creep strength because carbon significantly increases the high-temperature mechanical strength. In addition, the chemical compositions of 316L(N) and 316L in RCC-MRx are stricter than those of 316L(N) and 316L in ASME II-A as shown in Table 1. In the case of Gr.91 steel, the chemical compositions of RCCMRx are also stricter than those in ASME II-A, as shown in Table 2. It should be noted that Gr.91 steel of RCC-MRx is not the same as that of ASME II-A, and with the introduction of stricter requirements in the chemical compositions, the intended materials in RCC-MRx are expected to have improved material properties with an increase in the manufacturing cost. RCC-MRx restricts the carbon range more tightly. It appears that increasing the lower limit of carbon from 0.06 to 0.08 will assure an improved creep strength, and decreasing the upper limit from

0.15 to 0.12 will increase the microstructure stability through the suppression of carbide coarsening and an undesirable phase formation. RCC-MRx aims to restrict more tightly the impurity content of phosphorus and sulfur. The segregation of tramp elements such as P and S to prior austenite grain boundaries may cause an embrittlement of Gr. 91. As for nickel, RCC-MRx restricts more strictly the Ni content. RCC-MRx is more concerned about a reduced radioactivity because Ni is likely to be a radioactive element through irradiation-induced transmutation. As for Nb, RCC-MRx specifies additionally the Nb content. The small addition of Nb results in an improvement of the creep strength and better microstructure stability owing to the fine precipitation of NbC. Therefore, it is expected that Gr.91 steel of RCC-MRx will have an improved creep strength with microstructure stability. 3.2. Material properties 3.2.1. Material strengths Comparisons of the material strengths between ASME codes and RCC-MRx were made for the materials of austenitic stainless steel 316 and Gr.91 steel. Here, the ASME codes indicate ASME Section II Part D (ASME Boiler, 2015b) (hereafter, ASME II-D) and Subsection NH because some elevated temperature material properties are provided in ASME-NH. The yield strengths, tensile strengths, and design stress intensities for Gr.91 steel are given in Figs. 5–7, respectively. In the actual design evaluation, the design stress intensity properties in Fig. 7 are generally used. It was shown that the properties of RCC-MRx

Table 1 Chemical compositions of austenitic stainless steel 316 in ASME code and RCC-MRx. Code/Test

Grade

C

Mn

P

S

Si

Cr

Ni

Mo

N

B

Other

ASME

316SS (Sec.III-NH) 316SS (Sec.II-part A) 316H (Sec.II-part A) 316LN (Sec.II-part A) 316L (Sec.II-part A)

0.04–0.06 0.08 0.04–0.10 0.03 0.03

1.0–2.0 2.0 2.0 2.0 2.0

<0.03 0.045 0.045 0.045 0.045

<0.02 0.03 0.03 0.03 0.03

0.6 1.0 1.0 0.75 0.75

17.0–18.0 16.0–18.0 16.0–18.0 16.0–18.0 16.0–18.0

11.0–12.5 10.0–14.0 10.0–14.0 10.0–14.0 10.0–14.0

2.5–3.0 2.0–3.0 2.0–3.0 2.0–3.0 2.0–3.0

0.04–0.07 ... ... 0.10–0.16 0.1

0.003 ... ... ... ...

Al 0.05

RCC-MRx

316LN 316L

0.03 0.03

1.6–2.0 2.0

0.03 0.03

0.015 0.015

0.5 1.0

17.0–18.0 16.5–18.5

12.0–12.5 10.5–13.0

2.3–2.7 2.5–3.0

0.06–0.08 60.11

... ...

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Table 2 Chemical compositions of Mod.9Cr-1Mo steel in ASME-NH and RCC-MRx. Grade

C

Mn

S

ASME

Gr.91

0.06–0.15

0.25–0.66

0.025

0.012

0.18–0.56

7.90–9.60

0.43

RCC-MRx

X10CrMoVNb9-1 (Gr.91)

0.080–0.120

0.30–0.60

60.020

60.005

0.20–0.50

8.00–9.50

60.20

are lower than those of the ASME code, which means that RCCMRx is more conservative in design stress intensity than the ASME codes for Gr.91 steel. In the case of austenitic stainless steel 316, the yield strengths, tensile strengths, and design stress intensities were the lowest in 316L steel of RCC-MRx, and were the highest for the 316SS of ASME code. This trend in austenitic stainless steel is the same as in Gr.91 steel, which means that the strength data of RCC-MRx are the lowest, as shown in Figs. 8–10. It should be noted that hightemperature material properties among the austenitic stainless steel 316 family in the ASME codes are available only for 316SS, whereas only 316L(N) and 316L stainless steel are available in RCC-MRx, which means a direct comparison is not possible between the ASME codes and RCC-MRx in the case of austenitic stainless steel 316. The modulus of elasticity in ASME II-D and RCC-MRx are similar in magnitude for Gr.91 steel and austenitic stainless steel 316, as shown in Figs. 11 and 12, respectively. The ASME code tends to provide lower values in modulus of elasticities at low temperatures, whereas higher values than RCC-MRx at high temperature for both materials are shown in Figs. 11 and 12.

3.2.2. Thermal expansion and thermal conductivity properties The thermal expansion coefficient of Gr.91 steel and austenitic stainless steel 316 for the ASME code and RCC-MRx are given in Figs. 13 and 14, respectively, for instantaneous and mean values. It is shown that as a whole the ASME values are higher than those of RCC-MRx. From the viewpoint of thermal stresses, the ASME code generally provides more conservative values than RCC-MRx because a higher thermal expansion coefficient will induce higher thermal stresses. In the case of thermal conductivity for the two materials, the ASME code provides higher values than RCC-MRx for Gr. 91 steel, while providing slightly lower values for austenitic stainless steel, as shown in Figs. 15 and 16. In the case of ASME II-D, thermal conductivity for Gr. 91 steel shows somewhat weird behavior because

Si

Cr

Ni

Mo

N

Nb

Other

0.80–1.10

0.025–0.080

...

Al. 0.02

0.85–1.05

0.03–0.07

0.06–0.10

Al.60.040

800

ASME_Gr.91 RCC-MRx_Gr.91

700

Tensile strength (MPa)

P

600

500

400

300

200 0

100

200

300

400

500

600

o

temperature ( C) Fig. 6. Comparison of tensile strength of Gr.91 steel.

200

150

ASME_Gr.91 RCC-MRx_Gr.91

Sm (MPa)

Code

100

50 0

500

100

200

300

400

500

600

700

o

temperature ( C) Fig. 7. Comparison of design stress intensity of Gr.91 steel.

Yield strength (MPa)

400

the value reaches a peak at 27.9 W/m°K over a temperature of 400– 500 °C and reduces from 550 °C. 300

ASME-NH_Gr.91 RCC-MRx_Gr.91 200

100 0

100

200

300

400

500 o

temperature ( C) Fig. 5. Comparison of yield strength of Gr.91 steel.

600

700

3.2.3. Fatigue strength Fatigue strength curves of Gr. 91 steel and austenitic stainless steel 316 in the ASME and RCC-MRx codes are compared in Figs. 17 and 18, respectively. In the case of Gr. 91 steel, ASME-NH provides less conservative values than those of RCC-MRx. ASME-NH provides fatigue life cycles of up to 108 cycles, whereas RCC-MRx provides life cycles of up to 106 cycles. In the case of austenitic stainless steel 316, however, RCC-MRx is more conservative at up to around 200 cycles, and ASME-NH is more conservative at fatigue cycles of higher than 200 cycles, as

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220

ASME_316SS RCC-MRx_316L(N) RCC-MRx_316L

modulus of elasticity (MPa)

Yield Strength (MPa)

200

150

100

ASME_Gr.91 RCC-MRx_Gr.91

210 200 190 180 170

0

100

200

300

400

500

600

160

700

0

100

200

300

400

500

o

Temperature ( C)

600

o

temperature ( C)

Fig. 8. Comparison of yield strength of austenitic stainless steel 316.

Fig. 11. Modulus of elasticity in Gr.91 steel.

ASME_316SS RCC-MRx_316L(N) RCC-MRx_316L UTS (MPa)

500

400

modulus of elasticity (MPa)

600

180

160

300 0

100

200

300

400

500

ASME_316SS RCC-MRx_316L,316L(N)

200

0

600

100

200

300

400

500

600

o

o

temperature ( C)

temperature ( C)

Fig. 12. Modulus of elasticity in austenitic stainless steel 316.

Fig. 9. Comparison of tensile strength of austenitic stainless steel 316.

15

150

Thermal expansion coefficient

ASME_316SS RCC-MRx_316L(N) RCC-MRx_316L Sm (MPa)

125

100

75 0

100

200

300

400

500

600

700

o

Temperature ( C) Fig. 10. Comparison of design stress intensity of austenitic stainless steel 316.

ASME_Gr.91(i) RCC-MRx _Gr.91(i) ASME_Gr.91(m) RCC-MRx _Gr.91(m)

14

13

12

11

10 0

100

200

300

400

500

o

temperature ( C) Fig. 13. Comparison of thermal expansion coefficient of Gr.91 steel.

600

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21

ASME-NH (Gr.91) RCC-MRx (Gr.91) strain range (%)

Thermal expansion coefficient

10

ASME_316SS_i RCC_316L,316L(N)_i ASME_316SS_m RCC_316L,316L(N)_m

18

1

15 0

100

200

300

400

500

600

o

temperature ( C)

0.1 1 10

10

2

10

3

10

4

10

5

10

6

10

7

10

8

cycle (N) Fig. 14. Comparison of thermal expansion coefficient of austenitic stainless steel 316.

Fig. 17. Comparison of fatigue strength of Gr.91 steel.

30

10

Strain range (%)

26

o

Conductivity (W/m K)

28

24

ASME_Gr.91 RCC-MRx_Gr.91

22 20

ASME-NH (316SS) RCC-MRx (316L(N)) 1

18 16 0

100

200

300

400

500

600

o

temperature ( C)

0.1 1 10

10

2

10

3

10

4

10

5

10

6

10

7

8

10

Cycle (N)

Fig. 15. Comparison of thermal conductivity of Gr.91 steel.

Fig. 18. Comparison of fatigue strength of austenitic stainless steel 316.

24

21

o

Conductivity (W/m K)

ASME_316SS RCC-MRx_316L,316L(N)

18

15

12 0

100

200

300

400

500

600

o

temperature ( C) Fig. 16. Comparison of thermal conductivity of austenitic stainless steel 316.

shown in Fig. 18. In addition, RCC-MRx provides fatigue strength data of up to 108 cycles in 316L(N), whereas ASME-NH provides 106 cycles. It is worth mentioning that the fatigue strengths of 316L(N) are the same with those of 316L in RCC-MRx.

3.2.4. Creep rupture strength Creep rupture strengths (CRS) of Gr. 91 steel in RCC-MRx are higher than those of ASME-NH at sustained stresses of higher than 240 MPa, as shown in Fig. 19. However, the RCC-MRx values are lower than those of ASME-NH at sustained stresses of lower than 240 MPa, and the discrepancies between the two codes increase as the stress levels decrease, as shown in Fig. 19. Although Gr.91 steel of ASME-NH is not the same as that of RCC-MRx, RCC-MRx is more conservative at a relatively low level of sustained stress. It is worth noting in CRS of ASME-NH that CRS for Alloy 800H is available at up to 500,000 h from the 2013 edition as shown in Table 3, which is equivalent to a 60-year lifetime with consideration of the refueling and maintenance period. The CRS for other materials in ASME-NH and heat resistant materials in RCC-MRx A3 is available for up to 300,000 h as shown in Table 3. In the case of austenitic stainless steel 316, the CRS of 316SS is higher than those of 316L and 316L(N) of RCC-MRx at sustained stress levels of higher than about 320 MPa but lower than those of 316L and 316L(N) at sustained stress levels lower than about 270 MPa, as shown in Fig. 20. Therefore, the CRS of the ASME-NH data depends on the materials and stress levels rather than simply

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300

400

Creep rupture strength (MPa)

Creep rupture strength (MPa)

350 250

ASME-NH (Gr.91) RCC-MRx (Gr.91)

200

150

100 1 10

10

2

10

3

10

4

10

5

10

300 250 200

ASME_316SS RCC_316L(N) RCC_316L

150

6

time (h)

100

1

10

100

1000

Fig. 19. Comparison of creep rupture strength of Gr.91 steel.

higher or lower than those of RCC-MRx values. It was shown that the creep rupture strengths of 316L(N) are quite higher than those of 316L in RCC-MRx.

10000 100000

time (h) Fig. 20. Comparison of creep rupture strength of austenitic stainless steel 316.

3.3. Design evaluation procedures In addition to the differences in the material properties, there are significant differences in some aspects of the design evaluation procedures of the two ETD codes. The design evaluation procedures of ASME-NH and RCC-MRx are basically the same as shown in Figs. 3 and 4. However, there are some significant differences between the two codes in the detailed procedures including the determination procedures of the total inelastic strain and creep-fatigue damage, as shown in Table 3. One of the greatest differences in the two codes is that isochronous curves are used in the ASME-NH, whereas the creep law is directly used in RCC-MRx for determination of the inelastic strain and creep strain, as shown in Table 3. The isochronous curves are the stress–strain relationships determined at a fixed time in the 3D space of the stress, strain and time, as shown in Fig. 21. The isochronous curves of 316SS and Gr.91 steel at 566 °C are compared in Fig. 22. It is interesting to note that Gr.91 steel relaxes significantly more than 316SS. The amount of relaxation from 180 MPa during 30 h for 316SS was 2.5%, whereas that for Gr.91 steel was 36.3%, which means the Gr.91 steel relaxes far more than 316SS, as shown in Fig. 22. In addition, it should be noted that isochronous curves are based on uni-axial and monotonic stress–strain relationship and are known to be very conservative. Elastic follow-up (EFU) is considered in ASME-NH implicitly in such a way that secondary stresses with elastic follow-up (i.e., pressure-induced membrane and bending stresses and thermal

σ

Fig. 21. Schematic of isochronous curves.

Table 3 Comparison of design rules in ASME-NH and RCC-MRx. ASME-NH Calculation of total strain range and creep damage Isochronous curves used. Elastic Follow-up Implicitly considered Strain limits 1% (membrane), 2% (bending), 5% (peak) Ratcheting rule Mod. Bree diagram (O’Donnel-Porowski) Peak terms in strain calculation Should be decomposed in elastic approach Creep strength data - 300,000 h (304SS, 316SS, Gr.91, Gr.22) - 500,000 h (Alloy 800H) Creep-fatigue damage envelope intersection point (Df, Dc) = (0.1, 0.01) (Gr.91 steel) Environmental effects No guideline except long-time service effect

RCC-MRx Creep laws directly used. Explicitly considered (q = 3, default) 1% (membrane), 2% (bending) Efficiency diagram method (pending for Gr.91) Decomposition not necessary because total stress intensity is used - 300,000 h (materials in RCC-MRx A3) (Df, Dc) = (0.3, 0.3) Explicitly considered (thermal aging & irradiation), but lack of data

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Fig. 22. Isochronous curves of 316SS and Gr.91 in ASME-NH and their comparison of stress relaxations from 180 MPa during 30 h.

induced membrane stresses) are classified as primary stresses, whereas EFU is explicitly taken into account with a default value of ‘3’ in RCC-MRx as shown in Table 3. In determination of inelastic strains according to ASME-NH based on elastic analysis, peak strains arising from geometric discontinuities should be excluded since the effects are considered in other steps. These decomposition procedures of all strain components are very complicated and time-consuming. In case of RCC-MRx on the determination of inelastic strains, those decompositions on peak stresses or peak strains are not necessary as shown in Table 3 because total stress ranges including peak stresses are directly used in calculation of inelastic strains. Therefore, RCC-MRx is more straightforward and reproducible regardless of the users in determination of inelastic strains than ASME-NH. The creep-fatigue damage envelope is significantly different between ASME-NH and RCC-MRx for Gr.91 steel, as shown in Fig. 23(a) and Table 3. The intersection point of ASME-NH is (0.1, 0.01), whereas that of RCC-MRx is (0.3, 0.3), which means that the area of allowable zone in ASME-NH is remarkably smaller than that of RCC-MRx. The creep-fatigue envelope of ASME-NH for Gr. 91 steel is generally known to be overly conservative, which is why ASME Code Case N-812 (ASME Code Case, 2011; Asayama et al., 2008, 2013) was developed for Gr. 91 steel. Based on quantitative evaluations on the conservatism in creep-fatigue damage evaluation procedures, an alternate creep-fatigue damage envelope for Gr. 91 steel was developed as ASME Code Case N-812. The creep-fatigue envelope of CC N-812 is shown in Fig. 23(b), and the intersection point is (0.3, 0.3), which can be applied under the two conditions that elastic analysis procedures in ASME-NH should be followed and isochronous curve should be used. ASME codes such as ASME-NH and ASME Section III Subsection NB (ASME Boiler, 2015d) generally do not take environmental effects into consideration, whereas ASME-NH considers exceptionally ‘long-time service effects’ with the introduction of strength reduction factors for the yield strength and tensile strength as shown in Table 3. However, a recent study (Lee et al., 2016) has shown that it is necessary to validate the guidelines on the longtime service effects in ASME-NH based on an investigation of the long-time effects with material tests. In the case of the RCC-MRx code, it has started to take the environmental effects into consideration since the 2012 Edition, although many data are still left blank.

4. Comparison of conservatisms Comparisons of the ETD codes were made based on the application of the two codes to the SFR components. Two components of a decay heat exchanger (DHX) made of Gr.91 steel and expansion tank made of austenitic stainless steel 316L were used for the quantification of the conservatisms in ASME-NH and RCC-MRx. In present study, elevated temperature design evaluations were conducted according to the elastic analysis procedures in ASMENH and RCC-MRx. In elastic analysis procedures of both codes, the inelastic behaviors from plasticity and creep are taken into account with the introduction of various adjustment factors such as multiaxial plasticity, Poisson ratio, and triaxiality factor etc. along with the Neuber’s rules, stress relaxation and elastic follow-up effect. Creep damage in ASME-NH was calculated using monotonic stress–strain relations in isochronous curves while in RCC-MRx calculated using mathematical creep equations directly. As for creep damage evaluation in RCC-MRx, Bailey–Norton creep law is used for primary creep while Norton’s creep law is used for secondary creep for Gr.91 steel and 316L stainless steel. 4.1. Application of the ETD codes to a heat exchanger of DHX 4.1.1. Design evaluation A schematic of the decay heat exchanger, which has been installed in STELLA (sodium test facility of sodium integral effect test loop for safety simulation and assessment)-1, is shown in Fig. 24(a) (Lee et al., 2012a). The design transients of the primary and secondary sides of the DHX are shown in Fig. 25(a) and (b), respectively. The number of creep-fatigue load cycles in Fig. 25 for DHX in STELLA-1 is 500. A heat transfer analysis for the design transients has been conducted and the temperature profile obtained from heat transfer analysis is shown in Fig. 24(b). In addition, thermal stress analysis results in Fig. 24(c) show that the maximum stress intensity under the thermal loads was 361.2 MPa. 4.1.2. Comparison of conservatisms The design evaluation results according to ASME-NH have shown that the total inelastic strain at nozzle number 3 in Fig. 24(d) was 0.139%, while that as per RCC-MRx was 0.098%. The final creep-fatigue damage calculated as per ASME-NH and RCC-MRx were almost negligible, as shown in Fig. 26.

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RCC-MRx ASME-NH

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ASME CC-N812 ASME-NH

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0.8

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Fatigue Damage, Df

(a) ASME-NH and RCC-MRx

(b) ASME CC-N-812 and ASME-NH

Fig. 23. Creep-fatigue damage envelopes for Gr.91 steel in ETD rules.

(a) DHX

(b) temperature

(c) stress

(d) four critical points

Fig. 24. Schematic of DHX, analysis results and critical locations in DHX (Gr.91 steel).

30 hr

30 hr

200 3.1hr

Sodium temperature (°C)

Sodium temperature (°C)

510

500 30 hr

200 3.0 hr

3.1hr

3.0 hr

time (hr)

time (hr)

(a) primary side

(b) secondary side

Fig. 25. Design transients of the decay heat exchanger.

It should be noted that the allowable region in the creep-fatigue damage envelope of ASME-NH is far narrower than that of RCC-MRx, as shown in Figs. 23 and 26. Because the intersection point for ASME-NH in the envelope is (0.1, 0.01), whereas that

for RCC-MRx is (0.3, 0.3) for Gr. 91 steel, the allowable region for creep damage for ASME-NH is 1/30 compared to that of RCC-MRx. The evaluation result of creep-fatigue damage, (Df, Dc), at nozzle number 3 in Fig. 24(d) according to ASME-NH was

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4.2. Application of the ETD codes to expansion tank 4.2.1. Design evaluation A schematic of the expansion tank in STELLA-1 is shown in Fig. 27(a) and the design transients are shown in Fig. 28(Lee et al., 2014). The number of creep-fatigue load cycles in Fig. 25 for expansion tank is 500. The actual material of the expansion tank is austenitic stainless steel 316L. However, the material was assumed to be austenitic stainless steel 316 when the ASME-NH procedure is taken into consideration. In the case of RCC-MRx, the material properties of 316L are available. Heat transfer analysis results are shown in Fig. 27(b), and stress intensity profiles in Fig. 27(c) under the thermal

transients show that the stress intensity levels with a maximum value of 152.04 MPa at the nozzle of N2 in Fig. 27(a) are relatively low.

Sodium temperature (°C)

(5.0  10 5, 2.05  10 2), while that according to RCC-MRx was (5.0  10 5, 2.3  10 4), the results of which are plotted in Fig. 26. Therefore, it was shown that ASME-NH was more conservative than RCC-MRx for the present Gr.91 steel heat exchanger of DHX. It should be noted that conservatism in ASME-NH was significantly high in creep damage for Gr.91 steel, although Gr.91 steels of ASME-NH and RCC-MRx are not the same material with different chemical compositions.

30 hr

200 3.1 hr

3.1 hr

time (hr) Fig. 28. Design transients of expansion tank.

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Bilinear Limit Exp.Tank_RCC-MRx Exp.Tank_ASME-NH

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0.2 0.2

0.0 0.0 0.0

0.2

0.4

0.6

0.8

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Fig. 29. Evaluation results of creep-fatigue damage for 316L SS tank.

N1 Nozzle

N2 Nozzle

N3 Nozzle

(a) Expansion tank

(b) temperature distribution

0.8

Fatigue Damage, Df

Fatigue Damage, Df Fig. 26. Evaluation results of creep-fatigue damage for Gr.91 steel DHX.

0.4

(c) stress distribution

Fig. 27. Schematic shape and analysis results of expansion tank (316SS/316L).

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4.2.2. Comparison of conservatisms Design evaluations were conducted similarly to DHX according to ASME-NH and RCC-MRx. The total inelastic strain for the expansion tank was 0.052%, whereas that for RCC-MRx was 0.148%. The creep-fatigue damage calculated as per ASME-NH and RCC-MRx were almost negligible, as shown in Fig. 29. The creep-fatigue damage, (Df, Dc), according to ASME-NH, was (5.0  10 5, 1.0  10 4), whereas that according to RCC-MRx was (7.9  10 4, 1.0  10 4), and the results are plotted in Fig. 28. The intersection point of the creep-fatigue damage envelope for austenitic stainless steel 316 in ASME-NH(316SS) and RCC-MRx(316L) is the same at (0.3, 0.3). Although a direct comparison of the two ETD codes for austenitic stainless steel 316 is not possible because 316SS and 316L are different materials, and creep-fatigue damage for the expansion tank of both materials is negligible, RCC-MRx was shown to be slightly more conservative in the total inelastic strain and fatigue damage for the present expansion tank. Previous studies on a comparison of the conservatism of the two ETD codes (Lee et al., 2007, 2012a,b, 2013) showed that ASME-NH was generally more conservative than RCC-MRx for austenitic stainless steel heat exchangers. Therefore, it was shown from the evaluation results of DHX and previous studies that the design rule of ASME-NH is shown to be generally more conservative than RCC-MRx. 5. Conclusions A comparison study on the elevated temperature design codes of ASME Section III Subsection NH and RCC-MRx was conducted from chemical compositions to the conservatism of the two codes based on application to high-temperature components in Gen IV reactor components. ETD code comparisons were also made in terms of the material properties and design evaluation procedures for the recent editions of the two codes. The target materials were austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the two major materials in a Gen IV sodium-cooled fast reactor. There were significant differences in the chemical compositions between the ASME codes and RCC-MRx. For Gr.91 steel, the chemical compositions in RCC-MRx were restricted more tightly and were expected to have a better creep strength. It should be noted that Gr.91 steel of ASME-NH is not the same as that of RCC-MRx. In the case of austenitic stainless steel 316, a direct comparison between the two codes is not possible because ASME-NH provides only for 316SS, whereas RCC-MRx provides only 316L and 316L(N). The material properties highlighted in the present comparison are material strengths of yield strength and tensile strength, modulus of elasticity, conductivity, thermal expansion coefficient, fatigue strength and creep rupture strengths. Based on the comparison, it was shown that there were significant differences in some material properties although many data were similar. In addition, the differences in the design evaluation procedures of the two ETD codes were also highlighted. There are some significant differences in the design evaluation procedures as well as between the two ETD codes. One of the important issues in ASME-NH is reducing conservatism for the Gr.91 steel components including creep-fatigue damage envelope with the intersection point of (0.1, 0.01), while a ratcheting rule in RCC-MRx should be

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supplemented for Gr.91 steel. Other technical issues to be addressed in the two codes were raised in this study. A comparison of conservatism showed that ASME-NH was as a whole more conservative, especially for Gr.91 steel components, than RCC-MRx. Acknowledgements This work was supported by the International Research & Development Program Foundation of the National Research Foundation of Korea (NRF) grant (2013K1A3A7A03078195) and by NRF grant (2012M2A8A2025635) funded by the Korea government (Ministry of Science, ICT and Future Planning (MSIP)). References Abram et al., 2012. Status of Fast Reactor Research and Technology Development, IAEA-TECDOC-1691. IAEA. Asayama, Tai, Jetter, Robert, 2008. An Overview of Creep-Fatigue Damage Evaluation Methods and an Alternative Approach. In: Proceedings of ASME PVP 2008, Paper # PVP2008-61820. Asayama Tai, Takaya, Shigeru, Nagae, Yuji, Ando, Masanori, Tsukimori, Kazuyuki, 2013. Creep-Fatigue Evaluation Methodologies and Related Issues for Japan Sodium Cooled Fast Reactor (JSFR). In: 6th International Conference on Creep, Fatigue and Creep-Fatigue Interaction, Procedia Engineering, vol. 55, pp. 309– 313. ASME, Code Case N-812, 2011. Alternate Creep-Fatigue Damage Envelope for 9Cr1Mo-V Steel. ASME Boiler and Pressure Vessel Code, 2015a. Section II Part A. ASME. ASME Boiler and Pressure Vessel Code, 2015b. Section II Part D. ASME. ASME Boiler and Pressure Vessel Code, 2015c. Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, Subsection NH, Class 1 Components in Elevated Temperature Service. ASME. ASME Boiler and Pressure Vessel Code, 2015d. Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, Subsection NB, Class 1 Components. ASME. ASME Boiler and Pressure Vessel Code, 2015e. Section III, Rules for Construction of Nuclear Power Plant Components, Division 5, High Temperature Reactors. ASME. Chang, J.-W., Kim, Y.-W., Lee, K.-Y., Lee, Y.-W., Lee, W.-J., Noh, J.-M., 2007. A study of nuclear hydrogen production demonstration plant. Nucl. Energy Technol. 39 (2), 111–122. Kumar, Prabhat, Pai, Aravinda, 2014. An overview of welding aspects and challenges during manufacture of intermediate heat exchangers for 500mwe prototype fast breeder reactor. Procedia Eng. 86, 173–183. Lee, H.-Y., Lee, S.-H., Kim, J.-B., Lee, J.-H., 2007. Creep–fatigue damage for a structure with dissimilar metal welds of modified 9Cr–1Mo steel and 316L stainless steel. Int. J. Fatigue 29, 1868–1879. Lee, H.-Y., Kim, J.-B., Park, H.-Y., 2012a. High temperature design and damage evaluation of Mod.9Cr-1Mo steel heat exchanger. J. Pressure Vessel Technol. Trans. ASME 134, 051101-1–051101-10. Lee, H.-Y., Kim, J.-B., Park, H.-Y., 2012b. Creep-fatigue damage evaluation of sodium to air heat exchanger in sodium test loop facility. Nucl. Eng. Des. 250, 308–315. Lee, H.-Y., Eoh, J.-H., Lee, Y.-B., 2013. High temperature design of finned-tube sodium-to-heat exchanger in a sodium test loop. Nucl. Eng. Des. 265, 833–840. Lee, H.-Y., Kim, J.-B., Park, H.-Y., 2014. Evaluation of creep-fatigue integrity for high temperature pressure vessel in a sodium test loop. J. Pressure Vessel Technol. Trans. Korean Soc. Mech. Eng. A 38 (8), 831–836. Lee, H.-Y., Lim, D.-W., Kim, W.-G., Kim, J.-B., 2016. Effects of long-time elevated temperature service on material strength and J–R curve for Grade 91 steel. Int. J. Fatigue, submitted, March. RCC-MRx, Section I Subsection B, 2013a. 1st Addendum 2013, Class 1 N1RX Reactor Components its Auxiliary Systems and Supports, 2012 ed. AFCEN, Dec. RCC-MRx, Section III Tome 2, 2013b. Materials, 1st Addendum 2013, 2012 ed. AFCEN. RCC-MRx, Section III Tome 6, 2013c. Probationary Phase Rules, 1st Addendum, 2012 ed. AFCEN. Wakai, Takashi., Machida, Hideo., Yoshida, Shinji., Yanagihara, Seiji., Suzuki, Ryosuke., Matsubara, Masaaki., et al., 2015. Development of an unstable failure analysis procedure considering change of compliance at a crack part of SFR pipes. Eng. Fail. Anal. 56, 484–500.