PNP 2014-054, Response to NRC Request for Additional Information - 10 CFR 50 Appendix G Equivalent Margin Analysis - Mf 2962

Response to NRC Request for Additional Information - 10 CFR 50 Appendix G Equivalent Margin Analysis - Mf 2962

ML14163A662
Person / Time
Site:	Palisades
Issue date:	06/12/2014
From:	Vitale A Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PNP 2014-054
Download: ML14163A662 (14)

Text

Entergy Nuclear Operations, Inc.

J?.- #rijca,

Palisades Nuclear Plant 1_il 1Lt7 27780 Blue Star Memorial Highway Covert, MI 49043-9530 Tel 269 764 2000 Anthony J. Vitale Site Vice President PNP 2014-054 June 12, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to NRC Request for Additional Information

- Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis

ME 2962 Palisades Nuclear Plant Docket No. 50-255 License No. DPR-20

REFERENCES:

1. Palisades Nuclear Plant, Application for Renewed Operating License, dated March 22, 2005 (ADAMS Accession No. ML050940446).

2.

Entergy Nuclear Operations, Inc. letter PNP 2013-028, Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margins Analysis, dated October 21, 2013 (ADAMS Accession No. ML13295A448).

3.

NRC email to Entergy Nuclear Operations, Inc., Request for Additional Information

- Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis

- MF2962, dated May 13, 2014 (ADAMS Accession No. ML14133A684).

Dear Sir or Madam:

In the Palisades Nuclear Plant (PNP) license renewal application (Reference 1),

Nuclear Management Company (NMC), the former license holder for PNP, committed to submit an equivalent margins analysis (EMA) for Nuclear Regulatory Commission (NRC) approval at least three years before any reactor vessel beltline material Charpy upper-shelf energy (USE) decreases to less than 50 ft-lb, in accordance with 10 CFR 50 Appendix G, Section IV, Fracture Toughness Requirements.

Entergy Nuclear Operations, Inc. submitted the required EMA in Reference 2.

~Entergy Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043*9530 PNP 2014-054 June 12,2014 Tel 269 764 2000 Anthony J. Vitale Site Vice President u.s. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to NRC Request for Additional Information - Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis -

MF 2962 Palisades Nuclear Plant Docket No. 50-255 License No. DPR-20

REFERENCES:

1. Palisades Nuclear Plant, Application for Renewed Operating License, dated March 22, 2005 (ADAMS Accession No. ML050940446).

2. Entergy Nuclear Operations, Inc. letter PNP 2013-028, Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margins Analysis, dated October 21, 2013 (ADAMS Accession No. ML13295A448).

3. NRC email to Entergy Nuclear Operations, Inc., Request for Additional Information - Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis - MF 2962, dated May 13, 2014 (ADAMS Accession No. ML14133A684).

Dear Sir or Madam:

In the Palisades Nuclear Plant (PNP) license renewal application (Reference 1),

Nuclear Management Company (NMC), the former license holder for PNP, committed to submit an equivalent margins analysis (EMA) for Nuclear Regulatory Commission (NRC) approval at least three years before any reactor vessel beltline material Charpy upper-shelf energy (USE) decreases to less than 50 ft-Ib, in accordance with 10 CFR 50 Appendix G, Section IV, "Fracture Toughness Requirements."

Entergy Nuclear Operations, Inc. submitted the required EMA in Reference 2.

PLP 2014-054 Page 2 of 2 In Reference 3, ENO received a request for additional information (RAI) concerning the EMA submittal.

The ENO response to RAI questions 1, 3, 4, 5, and 6 is provided in the attachment.

The response to RAI question 2 will be provided at a later date, as agreed upon during a RAI clarification phone call between ENO and the NRC on June 6, 2014.

This letter contains no new commitments and no revised commitments.

I declare under penalty of perjury that the foregoing is true and correct; executed on June 12, 2014.

Sincerely, ajv/jse

Attachment:

Response to NRC Request for Additional Information

- Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis

- ME 2962 cc:

Administrator, Region Ill, USNRC Project Manager, Palisades, USN RC Resident Inspector, Palisades, USN RC PLP 2014-054 Page 2 of 2 In Reference 3, ENO received a request for additional information (RAI) conceming the EMA submittal.

The ENO response to RAI questions 1, 3, 4, 5, and 6 is provided in the attachment.

The response to RAI question 2 will be provided at a later date, as agreed upon during a RAI clarification phone call between ENO and the NRC on June 6, 2014.

This letter contains no new commitments and no revised commitments.

I declare under penalty of perjury that the foregoing is true and correct; executed on June 12, 2014.

Sincerely, ajv/jse

Attachment:

Response to NRC Request for Additional Information - Palisades Nuclear Plant 10 CFR 50 Appendix G Equivalent Margin Analysis - MF 2962 cc:

Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION -

PALISADES NUCLEAR PLANT 10 CFR 50 APPENDIX G EQUIVALENT MARGIN ANALYSIS - MF 2962 A follow-up request for additional information (RAI) was received from the Nuclear Regulatory Commission (NRC), by electronic mail on May 13, 2014. The RAI requested that the response to the RAI be docketed within 30 days of receipt of the request.

The Entergy Nuclear Operations, Inc. (ENO) Palisades Nuclear Plant (PNP) response to the RAI is provided below.

NRC Request (May 13, 2014) 1.

The EMA is based on American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Section Xl, Appendix K as supplemented by Regulatory Guide (AG) 1.161 Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 Ft-Lb ASME Code Section Xl, Appendix K, Article K-4210 and RG 1.161 both include equations for calculating the stress intensity factor due to radial thermal gradients. In Section 5.1 of the EMA submittal, the licensee discusses through-wall thermal stress and states that typical through-wall stress and stress distribution during a heatup transient are shown in Figures 5-1 and 5-2. But Figures 5-1 and 5-2 of the EMA submittal do not show these stresses as discussed. Provide figures showing typical through-wall stress and stress distributions during a heatup transient to support the discussion in paragraph 5.1 of the EMA submittal.

ENO Response to RAI-1 Figures detailing typical heatup thermal axial stress and typical through-wall axial stress for the PNP reactor vessel used in the equivalent margins analysis (EMA) submittal are provided below.

1 of 12 ATTACHMENT RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION -

PALISADES NUCLEAR PLANT 10 CFR 50 APPENDIX G EQUIVALENT MARGIN ANAL VSIS - MF 2962 A follow-up request for additional information (RAI) was received from the Nuclear Regulatory Commission (NRC), by electronic mail on May 13, 2014. The RAI requested that the response to the RAI be docketed within 30 days of receipt of the request.

The Entergy Nuclear Operations, Inc. (ENO) Palisades Nuclear Plant (PNP) response to the RAI is provided below.

NRC Request (May 13, 2014)

1.

The EMA is based on American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, Appendix K as supplemented by Regulatory Guide (RG) 1. 161 "Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 Ft-Lb". ASME Code Section XI, Appendix K, Article K-4210 and RG 1.161 both include equations for calculating the stress intensity factor due to radial thermal gradients. In Section 5. 1 of the EMA submittal, the licensee discusses through-wall thermal stress and states that typical through-wall stress and stress distribution during a heatup transient are shown in Figures 5-1 and 5-2. But Figures 5-1 and 5.. 2 of the EMA submittal do not show these stresses as discussed. Provide figures showing typical through-wall stress and stress distributions during a heatup transient to support the discussion in paragraph 5. 1 of the EMA submittal.

ENO Response to RAI-1 Figures detailing typical heatup thermal axial stress and typical through-wall axial stress for the PNP reactor vessel used in the equivalent margins analysis (EMA) submittal are provided below.

1 of 12

10 Heatup Transient Thermal Stress - Sxx vs Time ID Midwall 5E 0

1 2

3 4

5 6

7 8

U, U,

I U,-5

[-

-10

-15 Time (hr)

Figure 5-1 (a)

PNP Typical Thermal Transient Axial Stress Profile Stress versus Time 2 of 12 10

~I _H~aJ~p Ir~nJ~J~!'t l11~r.!ltal $t!_~SS - S~ ~~~ 11m~

I I

I

' 10

. Midwall j

OD 5

---ir o

1

-10 1--- --

-15 Time (hr)

Figure 5-1 (a) - PNP Typical Thermal Transient Axial Stress Profile - Stress versus Time 2 of 12

Heatup Transient Thermal Stress - Sxx vs. x/t 10 5

0 U,.

Ll

-5

-10

-15 Figure 5-2(a) - PNP Typical Thermal Transient Axial Through-Wall Stress Distribution NRC Request (May 13, 2014) 2.

Section 5.1 states, Only circumferential base metal flaws are considered in this analysis, because only the weak orientation USE is projected to drop below 50 ft-lbs as described below. Please demonstrate that assuming a circumferential flaw in the base metal with the weak Charpy V-Notch (CVN) value in the EMA is more limiting than assuming an axial flaw in the base metal with the strong CVN value. Please note that the significantly greater applied J integral associated with the axial flaw may challenge the fundamental assumption in the EMA submittal.

ENO Response to RAI-2 Based on discussions between ENO and the NRC during a June 6, 2014 RAI conference call, the response to this RAI question will be provided at a later date, as agreed upon during the call.

0 0.1 0.2 0,3 06 0.7 08 09

---

H I Transient Time (sec)I 6300 12600 15700 30300 Distance Ratio x/t 3 of 12 Heatup Transient Thermal Stress - Sxx vs. x/t 10 r

5 o

-10

-15 r

I I

-~. ----r-l

---

~ ----r ------ ----

0!5 0.6 Or7 0.8 0.9 1

I

~- ----,:-----+-

__________ 0-.1 Trarisient Time (sec)1

-- 6300

-- 12600 15700 I

- - - 30300

--1-__

Distance Ratio x/t Figure 5-2(a) - PNP Typical Thermal Transient Axial Through-Wall Stress Distribution NRC Request (May 13, 2014)

2.

Section 5. 1 states, "0nly circumferential base metal flaws are considered in this analysis, because only the "weak" orientation USE is projected to drop below 50 ft-Ibs as described below." Please demonstrate that assuming a circumferential flaw in the base metal with the weak Charpy V-Notch (CVN) value in the EMA is more limiting than assuming an axial flaw in the base metal with the strong CVN value. Please note that the significantly greater applied J integral associated with the axial flaw may challenge the fundamental assumption in the EMA submittal.

ENO Response to RAI-2 Based on discussions between ENO and the NRC during a June 6, 2014 RAI conference call, the response to this RAI question will be provided at a later date, as agreed upon during the call.

3 of 12

NRC Request (May 13, 2014) 3.

The applied J-integral values for the circumferential flaws for all Level A and B service level conditions are shown in Figure 5-1, and the applied J-integral values for the circumferential flaws for Level C and D service level conditions are shown in Figure 5-2. Since Section 5.1 provides very limited information regarding the applied J-integral calculations, please confirm that the calculations underlying Figures 5-7 and 5-2 are based on the formulas in RG 1.761, REvaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 FT-LB. If not, please describe, in addition to your response to RAt-1, your plant-specific calculations to support their acceptance in this application.

ENO Response to RAI-3 Yes, the applied J-integral calculations underlying Figures 5-1 and 5-2 are based on formulas in RG 1.161.

NRC Request (May 13, 2014) 4.

Table 4-4 was presented but without being mentioned in Section 4 regarding how it was used in the EMA analysis. Therefore, please confirm that the calculated available margins presented in Table 5-3 for various time during coo/down are results, using the relevant J-R curves adjusted by the material margin factors of Table 4-4.

ENO Response to RAI-4 As discussed in Section 2.2 of WCAP-17651-NP, Palisades Nuclear Power Plant Reactor Vessel Equivalent Margins Analysis, Revision 0, RG 1.161 material margin factors (ME) in Table 4-4 were used for the J-R curves.

NRC Request (May 13, 2014)

5. Section 5.2 provides justification for using the high-toughness/low-sulfur model from RG 1.161 in the proposed EMA for the high-sulfur plates, and Section 5.3 provides the corresponding EMA results. When the high-sulfur model (e.g., for the 6T specimen) of NUREG/CR-5265, Size Effects on J-R Curves for A 302-B Plate, is used, please demonstrate that The updated safety factors (see Table 5-3), after adjusting for temperature, will still be greater than 1.15.

The updated applied J/J-R curves (see figures 5-8, 5-9, and 5-12), after adjusting for temperature, will still show that dJappjjeci/da <dJmaterja/da at Japplied = Jmateriai 4 of 12 NRC Request (May 13, 2014)

3.

The applied J-integral values for the circumferential flaws for all Level A and B service level conditions are shown in Figure 5-1, and the applied J-integral values for the circumferential flaws for Level C and D service level conditions are shown in Figure 5-2. Since Section 5. 1 provides very limited information regarding the applied J-integral calculations, please confirm that the calculations underlying Figures 5-1 and 5-2 are based on the formulas in RG 1.161, uEvaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 FT-LB." If not, please describe, in addition to your response to RAI-1, your plant-specific calculations to support their acceptance in this application.

ENO Response to RAI-3 Yes, the applied J-integral calculations underlying Figures 5-1 and 5-2 are based on formulas in RG 1.161.

NRC Request (May 13, 2014)

4. Table 4-4 was presented but without being mentioned in Section 4 regarding how it was used in the EMA analysis. Therefore, please confirm that the calculated available margins presented in Table 5-3 for various time during cooldown are results, using the relevant J-R curves adjusted by the material margin factors of Table 4-4.

ENO Response to RAI-4 As discussed in Section 2.2 of WCAP-17651-NP, "Palisades Nuclear Power Plant Reactor Vessel Equivalent Margins Analysis," Revision 0, RG 1.161 material margin factors (MF) in Table 4-4 were used for the J-R curves.

NRC Request (May 13, 2014)

5. Section 5.2 provides justification for using the high-toughness/low-sulfur model from RG 1.161 in the proposed EMA for the high-sulfur plates, and Section 5.3 provides the corresponding EMA results. When the high-sulfur model (e.g., for the 6Tspecimen) of NUREG/CR-5265, uSize Effects on J-R Curves for A 302-8 Plate," is used, please demonstrate that The updated safety factors (see Table 5-3), after adjusting for temperature, will still be greater than 1. 15.

The updated applied J/J-R curves (see figures 5-8, 5-9, and 5-12), after adjusting for temperature, will still show that dJappliecida < dJmaterialda at Japplied = Jmaterial.

4 of 12

If the above cannot be demonstrated, perform a sensitivity study, showing at what percentage of the proposed J-R cuive (e.g., 90%), your EMA calculation results will meet the criteria on both crack extension and stability.

6. Section 6 presents conclusions of this submittal. For Service Level C condition with 400°F/hr cooldown, it is concluded that, The equivalent margins analyses for the plate materials are acceptable and bounded by the conservative test data reported in NUREG/CR-5265 in all cases for the Level C transient. This conclusion was repeated later for Service Level D condition with 600°F/hr cooldown, with C in the quote replaced by D. Plot the relevant NUREG/CR 5265 6T data in Figure 5-12 and provide sufficientjustification to support your conclusions.

ENO Responses to RAI-5 and RAI-6 Updated Figures 5-9 and 5-12 with the V-50 plate data are provided below, along with added Tables 5-4 and 5-5 showing the Level C and D safety factors, respectively.

Table 5-6 was added, which demonstrates the available margins on pressure loading with the V-50 plate data, adjusted for temperature, with consideration of all service loadings, Level A, B, C and D. The minimum safety factor (SF) with consideration of the V-SO plate data and the PNP-specific J-applied values is 1.5, which is above the minimum required SF of 1.15 per RG 1.161.

Figure 5-8 from WCAP-1 7651-NP, Revision 0, along with the updated Figures 5-9 and 5-1 2 below, all demonstrate that at Japplied = Jmaterial, dJapplied/da < dJmateriai/da, is satisfied for all three cases (i.e., the slope of the Japplied is smaller than the Jmaterial at the point of intersection).

Therefore, as demonstrated below and in WCAP-1 7651-NP, the equivalent margins of safety per ASME Code Section Xl (References 4 and 5) are found to be acceptable for the PNP reactor vessel beltline and extended beltline regions with predicted Charpy upper-shelf energy levels falling below the 50 ft-lb 10 CFR 50, Appendix G requirements at end-of-license-extension.

Westinghouse discovered during the development of this RAI response that the Level C and D loading J-applied curves plotted in WCAP-1 7651-NP, Figure 5-12, were not the most limiting case. This error also propagated onto Figures 5-2 and 5-13 in the WCAP.

This has been updated in the attached figures as part of this RAI response. Note that the conclusions to the report, including the safety factor determination, are unchanged; only the figures were in error. This has been documented in the Westinghouse corrective action system, and will be corrected when the WCAP is revised to incorporate these RAI changes.

Lastly, note that the Level C and D margin tables (Tables 5-4 and 5-5 below) were originally omitted from WCAP-1 7651-NP because Service Level A and B, as discussed in Section 5.3 of WCAP-1 7651-NP, are the governing transients.

5 of 12 If the above cannot be demonstrated, perform a sensitivity study, showing at what percentage of the proposed J-R curve (e.g., 90%), your EMA calculation results will meet the criteria on both crack extension and stability.

6. Section 6 presents conclusions of this submittal. For Service Level C condition with 400°Flhr cooldown, it is concluded that, "The equivalent margins analyses for the plate materials are acceptable and bounded by the conservative test data reported in NUREGICR-5265 in all cases for the Level C transient." This conclusion was repeated later for Service Level D condition with 600°Flhr cooldown, with IIC" in the quote replaced by liD." Plot the relevant NUREGICR 5265 6T data in Figure 5-12 and provide sufficient justification to support your conclusions.

ENO Responses to RAI-5 and RAI-6 Updated Figures 5-9 and 5-12 with the V-50 plate data are provided below, along with added Tables 5-4 and 5-5 showing the Level C and D safety factors, respectively.

Table 5-6 was added, which demonstrates the available margins on pressure loading with the V-50 plate data, adjusted for temperature, with consideration of all service loadings, Level A, B, C and D. The minimum safety factor (SF) with consideration of the V-50 plate data and the PNP-specific J-applied values is 1.5, which is above the minimum required SF of 1.15 per RG 1.161.

Figure 5-8 from WCAP-17651-NP, Revision 0, along with the updated Figures 5-9 and 5-12 below, all demonstrate that at Japplied = Jmaterial, dJapplieJda < dJmateria~da, is satisfied for all three cases (i.e., the slope of the Japplied is smaller than the Jmaterial at the point of intersection).

Therefore, as demonstrated below and in WCAP-17651-NP, the equivalent margins of safety per ASME Code Section XI (References 4 and 5) are found to be acceptable for the PNP reactor vessel beltline and extended beltline regions with predicted Charpy upper-shelf energy levels falling below the 50 ft-Ib 10 CFR 50, Appendix G requirements at end-of-license-extension.

Westinghouse discovered during the development of this RAI response that the Level C and D loading J-applied curves plotted in WCAP-17651-NP, Figure 5-12, were not the most limiting case. This error also propagated onto Figures 5-2 and 5-13 in the WCAP.

This has been updated in the attached figures as part of this RAI response. Note that the conclusions to the report, including the safety factor determination, are unchanged; only the figures were in error. This has been documented in the Westinghouse corrective action system, and will be corrected when the WCAP is revised to incorporate these RAI changes.

Lastly, note that the Level C and D margin tables (Tables 5-4 and 5-5 below) were originally omitted from WCAP-17651-NP because Service Level A and B, as discussed in Section 5.3 of WCAP-17651-NP, are the governing transients.

5 of 12

Table 5-4 Available Margins on Pressure Load for Level C, 400°F/hr Cooldown Base Material Weld Material Circumferential Flaw Circumferential Flaw Jo.

1 Jo.

1 Time SF J-applied x SF material SF J-applied x SF material (sec)

(in-lb/in 2

)

(in-lb/in 2

)

(in-lb/in 2

)

(in-lb/in 2

)

0 5.8 682 682 5.2 511 510 1,197 8.6 885 839 7.1 613 613 4,122 8.9 1,535 1,534 7.1 1,050 1,050 Minimum SF 5.8 5.2 Table 5-5 Available Margins on Pressure Load for Level D, 600°F/hr Cooldown Base Material Weld Material Circumferential Flaw Circumferential Flaw Jo.

1 Jo.

1 Time SF J-applied x SF material SF J-applied x SF material (sec)

(in-lb/in 2

)

(in-lb/in 2

)

(in-lb/in 2

)

(in-lb/in 2

)

0 5.8 682 682 5.2 510 510 798 8.1 830 830 6.9 607 607 2,748 7.0 1,491 1,491 5.2 1,023 1,023 Minimum SF 5.8 5.2 6 of 12 Table 5-4 Available Margins on Pressure Load for Level C, 400°F/hr Cooldown Base Material Weld Material Circumferential Flaw Circumferential Flaw JO*1 JO*1 Time SF J-applied x SF material SF J-applied x SF material (sec)

(in-lb/in2)

(in-lb/in2)

(in-lb/in2)

(in-lb/in2) 0 5.8 682 682 5.2 511 510 1,197 8.6 885 839 7.1 613 613 4,122 8.9 1,535 1,534 7.1 1,050 1,050 Minimum SF 5.8 5.2 Table 5-5 Available Margins on Pressure Load for Level D, 600°F/hr Cooldown Base Material Weld Material Circumferential Flaw Circumferential Flaw JO*1 JO*1 Time SF J-applied x SF material SF J-applied x SF material (sec)

(in-lb/in2)

(in-lb/in2)

(in-lb/in2)

(in-lb/in2) 0 5.8 682 682 5.2 510 510 798 8.1 830 830 6.9 607 607 2,748 7.0 1,491 1,491 5.2 1,023 1,023 Minimum SF 5.8 5.2 6 of 12

Table 5-6: Available Margins on Pressure Load for All Transients, Levels A, B, C and D, with Consideration of V-50 Plate Data V-50 Plate V-50 Plate V-50 Plate Level A Circumferential Level C Circumferential Level D Circumferential and B J

01 J

01 Flaw Flaw Flaw material material J-applied J-applied J-applied material Time SF SF (in-Time (in-Time SF x SF (in-lb/in 2

)

SF xSF (sec)

(in-lb/in 2

)

lb/in 2

)

(sec)

(in-lb/in 2

)

lb/in 2

)

(sec)

(in-lb/in 2

)

0 1.8 397 397 0

3.4 397 397 0

3.4 397 397 2800 1.6 441 441 1,197 4.7 488 488 798 4.7 483 483 3600 1.6 459 459 4,122 1.9 893 893 2,748 1.5 868 868 5400 1.7 503 503 7200 1.8 554 554 9000 1.9 611 611 10800 18.0 675 675 I

Minimum

[ Minimum Minimum SF 1.6 SF 1.9 SF 1.5 7 of 12 Table 5-6: Available Margins on Pressure Load for All Transients, Levels A, B, C and 0, with Consideration of V-50 Plate Data Level A V-50 Plate V-50 Plate V-50 Plate and B Circumferential JO*1 LevelC Circumferential JO.1 Level 0 Circumferential Flaw material Flaw material Flaw JO*1 Time J-applied (in-Time J-applied (in-Time J-applied material (sec)

SF xSF Iblin2)

(sec)

SF xSF Iblin2}

(sec)

SF xSF (in-lblin2)

(in-lblin2)

(in-lblin2)

(in-lblin2) 0 1.8 397 397 0

3.4 397 397 0

3.4 397 397 2800 1.6 441 441 1,197 4.7 488 488 798 4.7 483 483 3600 1.6 459 459 4,122 1.9 893 893 2,748 1.5 868 868 5400 1.7 503 503 7200 1.8 554 554 9000 1.9 611 611 10800 18.0 675 675 Minimum 1.6 Minimum 1.9 Minimum 1.5 SF SF SF 7 of 12

Applied J-lntegral Curve - Circumferential Flaw, Level C & D, a = lIlOt, SF=1 400 Load LeveIC

Load LevelD 300 C\\4 100 0

0.8 0.9 1

1.1 1.2 1.3 1.4 Flaw Depth in Base Metal a (in)

WCAP-1 7651-NP, Revision 0, Updated Figure 5-2 with Corrected, Limiting, Level C and 0 Transients 8 of 12 f::j

< c:

.Q

~

e Ie)

~

J; Applied J-Integral Curve - Circumferential Flaw, Level C & D, a = 1/10t, SF=1 400

-- Load Level C

- - - Load level 0 300 200

,--~

100

~ ---------------------------------------------------------------

o 0.8 0.9 1.1 1.2 1.3 1.4 Flaw Depth in Base Metal a (in)

WCAP-17651-NP, Revision 0, Updated Figure 5-2 with Corrected, Limiting, Level C and D Transients 8 of 12

Circumferential Flaw Stability - Base Metal, P2.T5ksi IOOF/hr Cooldown, alt=114t, SFI.25 N

(=

I:1 C

0) t Japplied T=533F Japplied T=480F Japplied T46OF Japplied Th31 3F JR Base T=533F JR Base T=480F JR Base T=460F JR Base T=31 3F da=O.1 Line 6T V-50 plate data at I SOF A

6T V-50 plate data 1800 1600 1400 1200 1000 800 600 400 200 0

+

AAA A

A 2.1 2.2 I

2.3 A

A A

2.4 2.5 2.7 Flaw Depth a (in) 2.6 WCAP-17651-NP, Revision 0, Updated Figure 5-9 with V-50 Plate Data Included 9 of 12 N

.E :a

'"i c::

~

C)

CD..

C

'"i..,

Circumferential Flaw Stability - Base Metal, P=2.75ksi 100F/hr Cooldown, a/t=1/4t, SF=1.25 1800

-- Japplied T=533F

-- Japplied T=480F 1600 I-I __ Japplied T=460F

-- Japplied T=313F 1400 r I <=== JR.Base T:;S33F

-- JR Base T=480F JR Base T=460F 1200 ~ I -=====- JR Base T=313F 1000 I 800 600 400 l 200 o

2.1 I ~.= da = 0.1" Line tr 2.2

+,

0

~

~

A A

~

2.3 2.4 2.5 2.6 Flaw Depth a (in)

WCAP-17651-NP, Revision 0, Updated Figure 5-9 with V-50 Plate Data Included 9 of 12 2.7

Circumferential Flaw Stability - Base Metal, Level C & D, a = lIlOt, SFI

=

C)

=

) LOJ

=

Leve{DLoadT=135F

=JR Baset/1O4IDF JR Base tIlO 5XF JR BasetIlO6lOF

=da=0.1 6TV-50pIatec1ata 180F 6TV-50 pIe data 1800 1600 1400 1200 1000 800 600 400 200 0

0.8

-A-J A

A A

0.9 1

1.1 1.2 1.3 A

Flaw Depth in Base Metal a (in) 1.4 WCAP-1 7651-NP, Revision 0, Updated Figure 5-12 with V-50 Plate Data Included and Corrected, Limiting, Level C and D Transients 10 of 12 N"

< c,

.0

~

E 1:7) 1800 1600

~ I 1400

~ I 1200

~ I 1000 1-1 Circumferential Flaw Stability - Base Metal, Level C & 0, a = 1/10t, SF=1

-- Level C Lom T = 121F

- - - Level DLoooT= 135F

~

JR Base tl10 400F JR Base t/10 500F

==-JR Baset/10610F

<= <= da= 0.1D

• 6T V-50 pille data Ii 100F

~ 6TV-5O plme data adusted to rolF

~

800

i;

~

I 600

~

o o

j

~ A_ ~

- A - ~- - - - - - - - - i - - - - - - - - -~ - - - - - -

A A

400 200 o

0.8 0.9 1.1 1.2 1.3 1.4 Flaw Depth in Base Metal a (in)

WCAP-17651-NP, Revision 0, Updated Figure 5-12 with V-50 Plate Data Included and Corrected, Limiting, Level C and 0 Transients 10 of 12

C4

-,)

Circumferential Flaw Stability Weld Metal, Level C & D, a = lIlOt, SF=1 LeveIC Load 121F

Level D Load 135F

===JRWeld t/10 400F JRWeldfJ105O0F 0

===JRWeldfJ1O61OF 0

da0.1 1800 1600 1400 1200 1000 800 800 400 200 0

II 0.8 0.9 1

1.1 1.2 1.3 1.4 fl___.

WCAP-1 7651-NP, Revision 0, Updated Figure 5-13 with Corrected, Limiting, Level C and D Transients 11 of 12 Circumferential Flaw Stability. Weld Metal, Level C & 0, a = 1/10t, SF=1 1800 1600

--Level C Load 121F

- - - Level D Load 135F 1400

--JRWeld tl10 400F JR Weld tl10 500F 1200

=

JRWeldtl10610F c:::> =- da = O.1n N'

< c 1000 B

~

1!

~

800 s

oJ, 600 400

-- ------- ~ --------------------------------------.

200 0

~

________ ~

___ L_ _________ i 0.8 0.9 1.1 1.2 1.3 1.4 Flaw Depth in Base Metal a (in)

WCAP-17651-NP, Revision 0, Updated Figure 5-13 with Corrected, Limiting, Level C and 0 Transients 11 of 12

References 1.

Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less than 50 Ft-Lb, U. S. Nuclear Regulatory Commission, June 1995.

2.

Westinghouse Report WCAP-1 7651-NP, Revision 0, Palisades Nuclear Power Plant Reactor Vessel Equivalent Margins Analysis, February 2013 (ADAMS Accession No. ML13295A448).

3.

Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

4.

ASME Boiler and Pressure Vessel (B&PV) Code, Section Xl, Division 1, Appendix K, Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels, 2007 Edition up to and including 2008 Addenda.

5.

ASME B&PV Code, Section Xl, Division 1, Appendix G, Fracture Toughness Criteria for Protection Against Failure, 1998 Edition up to and including 2000 Addenda.

12 of 12 References

1.

Regulatory Guide 1.161, "Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less than 50 Ft-Lb," U. S. Nuclear Regulatory Commission, June 1995.

2.

Westinghouse Report WCAP-17651-NP, Revision 0, "Palisades Nuclear Power Plant Reactor Vessel Equivalent Margins Analysis," February 2013 (ADAMS Accession No. ML13295A448).

3.

Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

4.

ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Appendix K, "Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels,"

2007 Edition up to and including 2008 Addenda.

5.

ASME B&PV Code,Section XI, Division 1, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1998 Edition up to and including 2000 Addenda.

12 of 12