AEP-NRC-2015-94, Response to License Condition Regarding Analysis of Pressure and Temperature Curves for Ferritic Reactor Vessel Materials

Response to License Condition Regarding Analysis of Pressure and Temperature Curves for Ferritic Reactor Vessel Materials

ML15272A491
Person / Time
Site:	Cook
Issue date:	09/25/2015
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2015-94, TAC MF4280, TAC MF4281
Download: ML15272A491 (26)

Text

IND*IANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWEROne Cook Place Bridgrnan, MI 49106 A unit of American Electric Power IndianaMichiganPower.com September 25, 2015 AEP-NRC-201 5-94 10 CFR 50.90 Docket Nos. 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Response to License Condition Regarding Analysis of Pressure and Temperature Curves for Ferritic Reactor Vessel Materials

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, License Amendment Request Regarding a Change to the Reactor Coolant System Pressure and Temperature Limits," AEP-NRC-2014-24, dated April 9, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14101A367.

2. Email from T. A. Beltz, NRC, to H. L. Etheridge, I&M, 'Draft Requests for Additional Information Vessels & Internals Integrity Branch and Technical Specifications Branch of the Office of Nuclear Reactor Regulation Regarding the Donald C. Cook Nuclear Plant, Units 1 and 2, License Amendment Request to Change Reactor Coolant System Pressure and Temperature Limit Curves to Address Vacuum Fill Operations (TAC Nos. MF4280 and MF4281)," dated July 21, 2014, ADAMS Accession No. ML14217A325.

3. Letter from J. P. Gebbie, I&M, to U. S. NRC, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to a Request for Additional Information Regarding the License Amendment Request to Change the Reactor Coolant System Pressure and Temperature Limits,"

AEP-NRC-2014-63, dated August 15, 2014, ADAMS No. ML14230A677.

4. Letter from J. P. Gebbie, I&M, to U. S. NRC, "Donald C. Cook Nuclear Plant Units 1 and 2, Supplemental Response to a Request for Additional Information Regarding the License Amendment Request to Change the Reactor Coolant System Pressure and Temperature Limits," AEP-NRC-2014-79, dated September 25, 2014, ADAMS No. ML14273A258.

5. Letter from M. L. Chawla, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2, Issuance of Amendments to Change the Reactor Coolant System Pressure and Temperature Limits (TAC Nos. MF4280 and MF4281)," dated October 1, 2014, ADAMS Accession No. ML14259A549.

U. S. Nuclear Regulatory Commission AEP-NRC-201 5-94 Page 2 This letter provides Indiana Michigan Power Company's (I&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to the License Conditions established by Reference 5.

By Reference 1, l&M submitted a request to amend the Technical Specifications (TS) to CNP Units 1 and 2 Renewed Facility Operating Licenses DPR-58 and DPR-74. I&M proposed to change TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits," to address an issue regarding the applicability of Figure 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits -

Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY)"

and Figure 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)," during vacuum fill of the Reactor Coolant System.

By Reference 2, the U. S. Nuclear Regulatory Commission (NRC) transmitted a Request for Additional Information (RAI) regarding the proposed amendment. By Reference 3, I&M provided a response to Reference 2. In Reference 3, as part of the response to RAI-EVIB-1 in Reference 2, I&M proposed to add a License Condition to the Unit 1 and Unit 2 licenses. Subsequently, I&M submitted Reference 4, in which I&M proposed a clarification to the License Condition for the Unit 1 and Unit 2 Facility Operating Licenses DPR-58 and DPR-74 that was submitted in Reference 3.

In Reference 5, the NRC approved the amendment requested by Reference 1. Reference 5 also prescribed the License Condition for each unit that was proposed by Reference 4, as described below:

Unit 1:

"Operation with vacuum fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits," with corresponding revisions to Figure 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY)," and Figure 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits -

Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)," as approved in License Amendment No. 323 to Renewed Facility Operating License No. DPR-58. This includes an approved extension to -14.7 pounds per square inch gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the PIT curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No. 323, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials."

Unit 2:

"Operation with vacuum fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, "RCS Pressure and Temperature (PIT) Limits," with corresponding revisions to Figure 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY)," and Figure 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits -

Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)," as approved in License Amendment No. 306 to Renewed Facility Operating License No. DPR-74. This includes an

U. S. Nuclear Regulatory Commission AEP-NRC-201 5-94 Page 3 approved extension to -14.7 pounds per square in gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the P/T curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No. 306, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials." to this letter provides an affirmation statement. Enclosure 2 to this letter contains the Westinghouse Electric Company letter that provides l&M's response to the License Conditions established by Reference 5.

Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President TLC/ams

Enclosures:

1. Affirmation

2. Westinghouse Electric Company Letter MCOE-LTR-15-82, Rev.0, "D. C. Cook Units 1 and 2 Pressure-Temperature Limits License Amendment Request: NRC Request for Additional Information and License Condition Response"~

c: A. W. Dietrich, NRC, Washington, D.C.

J. T. King - MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures

Enclosure 1Ito AEP-NRC-2015-94 AFFI RMATI ON I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company.

Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THISLSDAY OF2,_¢i"* S.L, 2015 NotaryDANIELLE BURGOYNE Public. State of Michigan County of Berrien My Commission Expires 0O0.4-201 8 Not b Acting in the County of~v My Commission Expires t ŽA-t-'A.- - Z--'*b.

Enclosure 2 to AEP-NRC-2015-94 Westinghouse Electric Company Letter MCOE-LTR-1 5-82, Rev.0 "0. C. Cook Units 1 and 2 Pressure-Temperature Limits License Amendment Request:

NRC Request for Additional Information and License Condition Response"

Westinghouse Non-Proprietary Class 3 B Westinghouse To: Charlie E. Meyer Date: September 2, 2015 From: Anees Udyawar Your ref: N/A Tel: 412-374-4300 Our ref: MCOE-LTR- 15-82, Rev. 0 Fax: 724-940-8565

Subject:

D. C. Cook Units 1 and 2 Pressure-Temperature Limits License Amendment Request: NRC Request for Additional Information and License Condition Response

References:

1. NRC Email from Terry Beltz (NRC) to Helen Etheridge (AEP),

Subject:

"D. C. Cook Nuclear Plant, Units 1 and 2 - Draft Requests for Additional Information re: Change to RCS P-T Limit Curves to Support Vacuum Fill (TAC Nos. MF4280 and MF4281)," dated July 21, 2014. [NRC ADA MS Accession number ML14217A325]

2. U.S. NRC Correspondence,

Subject:

"Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Change the Reactor Coolant System Pressure and Temperature Limits RE: (TAC NOS. M1F4280 AND MF4281)," dated October 1, 2014. ENRC ADAMS Accession number ML14259A549]

3. U.S. NRC Regulatory Issue Summary 2014-11, "INFORMATION ON LICENSING APPLICATIONS FOR FRACTURE TOUGHNESS REQUIREMENTS FOR FERRITIC REACTOR COOLANT PRESSURE BOUYNDARY COMPONENTS," dated October 14, 2014.

Attachment A provides the Westinghouse response to the NRC Request for Additional Information (RAI)

(Reference 1) on the D. C. Cook Units 1 and 2 pressure-temperature (P-T) limit curves that are. contained in the D. C. Cook License Amendment Request (LAR) to revise the Technical Specifications (Reference 2). The guidance of NRC Regulatory Issue Summary 2014-11 (Reference 3) is satisfied in Attachmnent A of this letter.

Do not hesitate to contact the undersigned if you have any questions regarding the contents of this letter.

Authored by: ELECTRONICALLY APPROVED* ELECTRONICALLY APPROVED*

Alley M. Carolan Benjamin A. Rosier Piping Analysis and Fracture Mechanics Materials Center of Excellence ELECTRONICALLY APPROVED*

Anees Udyawar Piping Analysis and Fracture Mechanics Verified by: ELECTRONICALLY APPROVED*

Amy E. Freed Materials Center of Excellence Approved by: ELECTRONICALLY APPROVED* ELECTRONMCALLY APPROVED*

David B. Love, Acting Manager John L. McFadden, Manager Materials Center of Excellence Piping Analysis and Fracture Mechanics

• Electronically approved records are authenticated in the electronic document management system.

© 2015 Westinghoutse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 Page 1 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Attachment A D. C. Cook Units 1 and 2 Pressure-Temperature Limits LAR: NRC RAI and License Condition Response

©2015 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 Page 2 of 20.

MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015

Background

In a License Amendment Request (LAR) (Reference 1) regarding proposed Technical Specifications changes associated with the heatup and cooldown pressure-temperature (P-T) limit curves, Indiana Michigan Power Company submitted revised P-T limit curves for D. C. Cook Units 1 and 2 for 32 effective full-power years (EFPY). The purpose of the LAR is to account for vacuum refill in the P-T limit curves. The extension of the vertical axis to account for vacuum refill does not directly relate to the analysis contained herein. However, since the bAR P-T limit curves account for vacuum refill, the comparative figures in this report will also account for vacuum refill. The basis documents for the D. C.

Cook Units 1 and 2 LAR P-T limit curves are WCAP-15878, Rev. 0 (Reference 2) and WCAP-15047, Rev. 2 (Reference 3), respectively. The LAR included the same P-T limit curves from WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2, with the addition of allowable pressures to address vacuum refill. The P-T limit curves were developed using the methodology described in Topical Report WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 4).

For Westinghouse nuclear steam supply systems, the Topical Report WCAP-14040-NP-A, Revision 2 describes the methodology that is used to comply with the requirements of 10 CFR 50 Appendix G, "Fracture Toughness Requirements" (Reference 5). Since the reactor vessel (RV) materials surrounding the core region receive significant neutron fluence and undergo neutron embrittlement, the RV beltline region is considered to be the limiting reactor coolant system (RCS) component for P-T limits. The methodology in WCAP-14040-NP-A, Revision 2 only addresses the RV beltline region of the RCS as the most limiting for the P-T limits. The original Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) for this topical report states, "We find the report to be acceptable for referencing in the administrative controls section of technical specifications for license amendment applications to the extent specified and under the limitations delineated in the report and the associated NRC safety evaluation, which is enclosed. The safety evaluation defines the basis for acceptance of the report." The SE further states, "The staff finds the WCAP-14040 methodology consistent with Appendix G to Section III of the ASME Code and SRP Section 5.3.2." and "T is the metal temperature and RTNDT is the ART value of the limiting vessel material" confirming that the reactor vessel is the limiting component evaluated in the development of the P-T limits. Table 1 of the NRC SE provides requirements regarding the fluence methodology, surveillance capsule program requirements, LTOPS requirements, adjusted reference temperature (ART) calculation, and 10 CFR 50 Appendix G temperature requirements, which have all been addressed in WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2, consistent with the NRC SE.

Subsequent to submitting the LAR to revise the Technical Specifications associated with the heatup and cooldown P-T limit curves for D. C. Cook Units 1 and 2, Indiana Michigan Power Company received Requests for Additional Information (RAIs) related to the 32 EFPY P-T limits. The discussion in this letter report addresses the NRC RAT, and the resultant regulatory license condition to submit the additional analysis of P-T limits, for the extended beltline and non-beltline reactor vessel components and other ferritic reactor coolant pressure boundary (RCPB) components with regards to P-T limits for D. C.

Cook Units 1 and 2.

Westinghouse Non-Proprietary Class 3 Page 3 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 The text below is the NRC RAI taken directly from an NRC correspondence (Reference 13). The pages thereafter contain the response to the RAI.

VESSELS & INTERNALS INTEGRITY BRANCH (EVIB)

Background

Title 10O of the Code of Federal Regulations (10 CFR) Part 50, Appendix G, "Fracture Toughness Requirements," states, "this appendix specifies fracture toughness requirements for ferritic materials of pressure-retainingcomponents of the reactor coolant pressure boundary (RCPB) of light water nuclear power reactors to provide adequate margins of safety... " In addition, 10 CFR Part 50, Appendix G, ParagraphIV.A states that, "the pressure-retainingcomponents of the RCPB that are made of ferritic materials must meet the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), supplemented by the additional requirements set forth in

[paragraphIV.A. 2, "Pressure-Temperature(P/T) Limits and Minim umn Temperature Requirements ".1... "

Therefore, 10 CFR Part 50, Appendix G requires that P/T limits be developed for the entire RCPB, consisting offerritic RCPB materials in the reactor vessel (RV) beltline (neutronfluence > I x J017 n/cm 2, E > 1 MeV), as well asferritic RC'PB materials not in the RVbeltline (neutronfluence < 1 x ]017 n/cm 2 , E

>1I Me V).

RAI-E VIB-1 The P-T limit calculationsfor feriqtic RCPB components that are not RV beltline shell materials may define P/T curves that are more limiting than those calculatedfor the RV beltline shell materials due to the following factors:

1. RV nozzles, penetrations, and other discontinuities have complex geometries that may exhibit significantly higher stresses than those for the R V beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even ifthe reference temperature (RTNDT,)

for these components is not as high as that of RV beltline shell materials that have simpler geometries.

2. Ferritic RCPB components that are not part of the RV may have initial RTN1DT values, which may define a mnore restrictive lowest operating temperature in the P-T limits than those for the R Vbeltline shell materials.

Please describe how the P/T limit curves in Technical Specification Figures 3.4.3-1, "Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Heat Test Limit (Applicable for service period up to 32 effective full power years [EFPYJ)" and 3.4.3-2, "Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY)" submitted for CNP Units 1 and 2, and the methodology used to develop these curves, considered all RV materials (beltline and non-b eltline) consistent with the requirements of 10 CFR Part50, Appendix G.

Westinghouse Non-Proprietary Class 3 Page 4 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Response to RAI-EVIB-1 Reactor Vessel Beitline Components The RAI requires that materials exhibiting end of life neutron fluence _ 1 x 1017 n/cm2 (EB> 1 MeV) be considered in the P-T limit curves. The P-T limits in WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2 only considered the intermediate and lower shell materials, and did not account for all materials outside the traditional beitline that may reach fluence > 1 x 1017 n/cm 2 (E > 1 MeV) at 32 EFPY. To address RAI-EVIB-1, the projected neutron fluence values for the D. C. Cook Units 1 and 2 reactor pressure vessels were calculated in order to determine the extent of the neutron fituence threshold of 1 x 1017 n/cm 2 (EB> 1 MeV). The most recent fast neutron fluence (B > 1.0 MeV) projections applicable to the D. C.

Cook Units 1 and 2 reactor pressure vessels were previously generated in support of the stretch power uprate (SPU). Although the SPU was not implemented, the fluence calculations performed in support of the SPU used the U. S. NRC approved methodology described in Reference 6. This methodology follows the guidance of NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 (Reference 7). The SPU fluence values were justified with consideration of the recently completed fuel cycles and remain bounding through 32 EFPY.

Tables 1 and 2 summarize the fluence projections for the potential extended beltline materials that were not analyzed in WCAP-15878, Rev. 0 or WCAP-15047, Rev. 2. From Tables 1 and 2, the upper shell plates and longitudinal welds, as well as the upper to intermediate shell circumferential weld are projected to exceed the 1 x i0 1' n/cm2 fluence threshold at 32 EFPY. The nozzle forgings as well as the nozzle to upper shell welds remain below 1 x 1017 n/cm 2 through 32 EFPY. However, the nozzle comer regions must still be evaluated for P-T limit curves since they may exhibit significantly higher stresses than those for the RV beltline shell region. The lower shell to lower vessel head welds for D. C. Cook Units 1 and 2 have neutron fluence values less than 1 x 1017 n/cm 2 through 32 EFPY. Therefore, no further consideration or analysis is necessary to address RAI-EVIB-1 for materials below the lower shell region.

Table 1 D. C. Cook Unit 1 Calculated Neutron Fluence Projections on Reactor Vessel Extended Beltline Materials at 32 EFPY 32 EFPY 2Calculated Fluence Reactor Vessel Material (n/cm , E > 1.0 MeV)

Upper to Intermediate Shell Circumferential Weld 2.62E+17 Upper Shell Plates 2.62E+ 17 Upper Shell Longitudinal Welds I1.99E+17 Inlet Nozzles (lowest extent at the weld) 2.34E+16 Outlet Nozzles (lowest extent at the weld) 1.79E+16 Lower Shell to Lower Vessel Head Circumferential Weld 1.74E+15

Westinghouse Non-Proprietary Class 3 Page 5 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Table 2 D. C. Cook Unit 2 Calculated Neutron Fluence Projections on Reactor Vessel Extended Beltline Materials at 32 EFPY Reactor Vessel Material 32 EFPY (n/cm 2Calculated Fluence

, E > 1.0 MeV)

Upper to Intermediate Shell Circumferential Weld 1.71lE+17 Upper Shell Plates 1.71lE+17 Upper Shell Longitudinal Welds 1.07E+17 Inlet Nozzles (lowest extent at the weld) 2.69E+16 Outlet Nozzles (lowest extent at the weld)* 1.83E+16 Lower Shell to Lower Vessel Head Circumferential Weld 1.39E+15 As shown above in Tables I and 2, the upper shell plates, upper shell longitudinal welds, and upper to intermediate shell circumferential welds receive projected fluence values greater than 1 x 1017 rl/Cmr2 at 32 EFPY. Note that these materials are collectively considered to be the extended beltline region materials; whereas, the traditional beltline materials typically include the intermediate shell, lower shell and associated longitudinal and circumferential welds, which were documented in WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2. The reactor vessel extended beltline materials are analyzed in this section and compared to the traditional beltline materials already analyzed in WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2 for D. C. Cook Units 1 and 2, respectively.

The ART values for the D. C. Cook Units 1 and 2 extended beltline materials were calculated at 32 EFPY and were determined using the methodology contained in Regulatory Guide 1.99, Revision 2 (Reference 8). When possible, the as-measured nickel (Ni) and copper (Cu) weight-percent (wt%) values along with initial RTNDT values were obtained from the certified material test reports (CMTRs) for each of the extended beltline materials. In the absence of test data, generic values were used. The chemistry factor (CF) values were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 8) methodology. The Position 1.1 CF values were calculated using the weight-percent copper and nickel values along with Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

The ART values for the D. C. Cook Unit 1 extended beltline materials are documented in Tables 3 and 4 for the 1/4T and 3/4T locations, respectively.

The ART values for the D. C. Cook Unit 2 extended beltline materials are documented in Tables 5 and 6 for the 1/4T and 3/4T locations, respectively.

Westinghouse Non-Proprietary Class 3 Page 6 of 20 MCOE-LTR- 15-82, Rev. 0 Attachment A September 2, 2015 Table 3 1/4T ART Calculations for the D. C. Cook Unit 1 Reactor Vessel Extended Beitline Cylindrical Shell Materials at 32 EFPY(a)

RatrVseMaeil Wt. % Wt. % CF(C) 1/TFun/cmd 1/4T Initia ARTN*DT *I *fTA~e Margin 1/4T ART 0

(E+19 n/m, FF(d) RTD( ) (0 F) (0 F) (0 F) (0 F) (0 F) ectrVseMaeilCu Ni (°F) E > 1.0 MeV) (F Upper Shell Plate B-4405-1 0.14 0.46 93.7 0.016 0.1479 10 13.9 0 6.9 13.9 38 Upper Shell Plate B-4405-2 0.14 0.45 93.3 0.016 0.1479 26 13.8 0 6.9 13.8 54 Upper Shell Plate B-4405-3 *0.14 0.48 94.6 0.0 16 0.1479 30 14.0 0 7.0 14.0 58 Upper Shell Longitudinal Welds 0.21 0.873 208.7 0.012 0.1236 -56(O 25.8 17(0 12.9 42.7 12 1-442 A, B, and C UprtInemdaeSel 0.216 0.737 188.4 0.016 0.1479 -56t0 27.9 l7(° 13.9 44.0 16 Circumferential Weld 8-442 Notes for Table 3:

(a) The Regulatory Guide 1.99, Revision 2 methaodology was utilized in the calculation of the ART values.

(b) Initial RTr*T values were based on measured data, unless otherwise noted.

(c) CF values were calculated using the copper and nickel weight percent values and Regulatory Guide 1.99, Revision 2.

(d) l/4T fluence and 1/4T fluence factor (FF) values were calculated using Regulatory Guide 1.99, Revision 2.

(e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal *a 17°F for Position 1.1 and the weld *a 28°F for Position 1.1; however, aA need not exceed 0.5*ART~vDT.

(f) These initial RTNrDT values are generic per 10 CFR 50.61; therefore, ai 17°F.

Westinghouse Non-Proprietary Class 3 Page 7 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Table 4 3/4T ART Calculations for the D. C. Cook Unit 1 Reactor Vessel Extended Beltline Cylindrical Shell Materials at 32 EFPY('*)

3/4T Fluencetd) Initial Wt. % Wt. % CF~c) nc 2 3/4T Tb) ARTNDT 6T 1 trt Margin 3/4T ART CuENi (9F)FF(d) 0 (0 )( 0 F) (0 F) (0 F) (0 F) (0 F)

Reco eslMtra u N ) E > 1.0 MeV) (F Upper Shell Plate B-4405-1 0.14 0.46 93.7 .0.006 0.073 5 10 6.9 0 3.4 6.9 24 Upper Shell Plate B-4405-2 0.14 0.45 93.3 0.006 0.0735 26 6.9 0 3.4 6.9 40 Upper Shell Plate B-4405-3 0.14 0.48 94.6 0.006 0.0735 30 7.0 0 3.5 7.0 44 Upper Shell Longitudinal Welds 0.21 0.873 208.7 0.004 0.0600 -56(0 12.5 17(0 6.3 36.2 -7 1-442 A, B, and C Upper to Intermediate Shell 0.216 0.737 188.4 0.006 0.0735 -56(O 13.9 17(0 6.9 36.7 -5 Circumferential Weld 8-442 Notes for Table 4:

(a) The Regulatory Guide 1.99, Revision 2 methodology was utilized in the calculation of the ART values.

(b) Initial RTNDT values were based on measured data, unless otherwise noted.

(c) CF values were calculated using the copper and nickel weight percent values and Regulatory Guide 1.99, Revision 2.

(d) 3/4T fluence and 3/4T FE values were calculated using Regulatory Guide 1.99, Revision 2.

0 (e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal o*A =17°F for Position 1.1 and the weld A = 28°F for Position 1.1; however, UA need not exceed 0.5*ART~rT.

(f) These initial RTNDT values are generic per 10 CFR 50.61; therefore, a1i 17°F.

Westinghouse Non-Proprietary Class 3 Page 8 of 20 MCOE-LTR- 15-82, Rev. 0 Attachment A September 2, 2015 Table 5 1

1/4T ART Calculations for the D. C. Cook Unit 2 Reactor Vessel Extended Beitline Cylindrical Shell Materials at 32 EFPY("

1/4T Flnence(d) Initial 2

Reactor Vessel Material Cu Ni (0F)

(E+19 n/cm ,

E > 1.0 MeV) FF(d)

RTNDRTD RTD(F) (0F)

~(°F) (0F)

Mrgn

(°F) 1/T0 R

( F)

Upper Shell Plate 11-1 0.14 0.59 99.6 0.010 0.1116 011.1 0 5.6 11.1 22 Upper Shell Plate 11-2 0.12 0.57 82.4 0.010 0.1116 10 9.2 0 4.6 9.2 28 Upper Shell Plate 11-3 0.12 0.61 83.2 0.010 0.1116 20 9.3 0 4.6 9.3 39 Upe ogtdnl hl 0.35 1.0 272 0.006 0.0805 10(o 21.9 17tD 10.9 40.4 72 Welds Upper to Intenmediate Shell 0.056 0.956 76.4 0.010 0.1116 -35 8.5 0 4.3 8.5 -18 Circumferential Weld Notes for Table 5:

(a) The Regulatory Guide 1.99, Revision 2 methodology was utilized in the calculation of the ART values.

(b) Initial RTNDT values were based on measured data, unless otherwise noted.

(c) CF values were calculated using the copper and nickel weight percent values and Regulatory Guide 1.99, Revision 2.

(d) l/4T fluence and 1/4T FF values were calculated using Regulatory Guide 1.99, Revision 2.

(e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal 0 A = 17°F for Position 1.1 and the weld aAx 28°F for Position 1.1; however, aA need not exceed 0.5*ARTNDT.

(t) This initial RTNOT value is generic; therefore, a1 = 17°F.

Westinghouse Non-Proprietary Class 3 Page 9 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Table 6 3/4T ART Calculations for the D. C. Cook Unit 2 Reactor Vessel Extended Beltline Cylindrical Shell Materials at 32 EFPY(a)

Rectr ese Mteia W.% t.% F~) 3/4T Fluencecd) 2 Initial tAC W.% W.%

C()nc 3/4T R b) ARTNDT 6I Margin 3/4T ART CuN(°) E+>1.0 MeV) (0F) (0 F) (0F) (OF) (0F) (0F)

Upper Shell Plate 11-1 0.14 0.59 99.6 0.004 0.0535 05.3 0 2.7 5.3 11 Upper Shell Plate 11I-2 0.12 0.57 82.4 0.004 0.0535 10 4.4 0 2.2 4.4 19 Upper Shell Plate 11-3 0.12 0.61 83.2 0.004 0.0535 20 4.4 0 2.2 4.4 29 Upe ogtdnl hl 0.35 1.0 272 0.002 0.0370 10(° 10.1 17(0 5.0 35.5 56 Welds Upper to Intermediate Shell 0.056 0.956 76.4 0.004 0.0535 -35 4.1 0 2.0 .4.1 -27 Circumferential Weld Notes for Table 6:

(a) The Regulatory Guide 1.99, Revision 2 methodology was utilized in the calculation of the ART values.

(b) Initial RTNOT values were based on measured data, unless otherwise noted.

(c) CF values were calculated using the Copper and nickel weight percent values and Regulatory Guide 1.99, Revision 2.

(d) 3/4T fluence and 3/4T FF values were calculated using Regulatory Guide 1.99, Revision 2.

(e) Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal *a = 170 F for Position 1.1 and the weld a,. = 28°F for Position 1.1; however, aA need not exceed 0.5 *ARTN*T.

(f) This initial RTNOT value is generic; therefore, ai 17°F.

Westinghouse Non-Proprietary Class 3 Page 10 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Tables 7 and 8 present the comparison of the limiting D. C. Cook Units 1 and 2 32 EFPY ART values used in the development of the P-T limit curves with the limiting 32 EFPY ART values of the extended beltline materials calculated in Tables 3 through 6. The limiting ART values used in the development of the 32 EFPY P-T limit curves bound the 32 EFPY ART values of the extended beitline materials.

Therefore, in response to RAI-EVJB-1, the D. C. Cook Units 1 and 2 32 EFPY P-T limit curves are applicable for all reactor vessel beltline and extended beltline materials that are projected to receive a neutron fluence > 1 x 1017 n/cm 2 , E > 1 MeV.

Table 7 Comparison of the Limiting ART Values used in Development of the 32 EFPY D. C. Cook Unit 1 P-T Limit Curves to the Extended Beltline Cylindrical Shell Material Limiting ART Values Reactor Vessel Material Type of Flaw 1/1AT 34AR

(°F) (°F)

Existing 32 EFPY Curves per Table 8-S of WCAP-158 78 (Reference 2)

Intermediate to Lower Shell Circ. Weld Seams (Position 2.1) Circumferential 211 150 Intermediate and Lower Shell Axial Weld Seams Axial 199--

Lower Shell Plate B-4407-3 Axial --- 143 Extended Beitline Cylindrical Shell Materials (see Tables 3 and 4 herein,)

Upper to Intermediate Shell Circumferential Weld 8-442 Circumferential 16 -5 Upper Shell Plate B-4405-3 Axial 58 44

Westinghouse Non-Proprietary Class 3 Page 11 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Table 8 Comparison of the Limiting ART Values used in Development of the 32 EFPY D. C. Cook Unit 2 P-T Limit Curves to the Extended Beltline Cylindrical Shell Material Limiting ART Values Reactor Vessel Material I

Type of Flaw

] 1/TAT3TAR 0

( F) J (0F)

Existing 32 EFPY Curvesper Table 8-S of WCAP-15047, Revision 2 (Reference 3,)

Intermediate Shell Plate 10-1 Axial 3 200 169 Extended Beitline CylindricalShell Materials (see Tables 5 and 6 herein,)

Upper Shell Longitudinal Welds Axial 3 72 56

Westinghouse Non-Proprietary Class 3 Page 12 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Reactor Vessel Non-Beltline Components The LAR P-T limit curves were developed using the methodology contained in WCAP-14040-NP-A, Revision 2. However, WCAP-14040-NP-A, Revision 2 does not specifically consider the highly stressed reactor coolant loop inlet and outlet nozzles. The inside corner regions of these nozzles are the most highly stressed ferritic components outside the beitline region of the reactor vessel; therefore, these components are analyzed in this section.

Per Tables 1 and 2, the D. C. Cook Units 1 and 2 nozzles do not reach fluence greater than 1 x 1017 nlcm 2 (E > 1 MeV) at 32 EFPY. Therefore, evaluation of the nozzles does not need to consider embrittlement.

Nozzle Initial RTNDT Values For D. C. Cook Unit 1, the initial RTNDT values were determined for each of the reactor vessel inlet and outlet nozzle forging materials using the BWRVIP-173-A (Reference 9), Alternative Approach 2 methodology, contained in Appendix B of that report. For each of the nozzles, the material-specific Charpy V-Notch impact energy data from the CMTRs was fit using a hyperbolic tangent curve to detenrmine the transition temperatures at 35 ft-lb and 50 ft-lb as specified in the Alternative Approach 2 methodology. The 35 ft-lb and 50 ft-lb temperatures were then evaluated, per the Alternative Approach 2 methodology presented in BWRVIP-173-A, to determine the initial RTNDT values for the inlet and outlet nozzle materials for D. C. Cook Unit 1. It should be noted that the orientation of the Charpy V-Notch specimens for six of the eight nozzles was not clearly identified in the CMTRs; therefore, for conservatism, it was assumed that the forging specimens were oriented in the strong direction. The 50 ft-lb transition temperatures were increased by 30°F to provide conservative estimates for specimens oriented in the weak direction.

For D. C. Cook Unit 2, four of the eight nozzles had full data sets (including Charpy V-Notch tests as well as drop-weight tests) such that the initial RTNDT values were determined using ASME Code,Section III, NB-233 1 and the Branch Technical Position 5-3 method to account for uncertainty in the testing orientation (values reduced to 65%).

The remaining four D. C. Cook Unit 2 nozzles had limited data available. For these nozzles, only the summary of the drop-weight test results and the mean 30 ft-lb temperatures were available. However, these four nozzles were manufactured by the same supplier, in the same timeframe, and have similar or identical heat numbers to the four nozzles that do have full data sets. Using the limited information available, an approach was taken similar to that described in BWRVIP-173-A. For each of the four known nozzles, the mean 30 ft-lb temperature and the minimum Charpy data points were used to approximate the slope between the mean 30 ft-lb temperature and the minimum Charpy impact energy data point (> 50 ft-lb) at its respective test temperature. The most conservative result from the four known nozzles was then applied to the four unknown nozzles to approximate the minimum 50 ft-lb temperature.

Westinghouse Non-Proprietary Class 3 Page 13 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 This approach is similar to BWRVIP-173-A, Appendix B, Alternative Approach 2, Item 2, wherein 2°F per ft-lb is advised. However, the BWRVIP-173-A method is meant to operate on the lowest Charpy V-notch data point. Since the D. C. Cook 30 ft-lb values are based on a mean-fit curve, a plant-specific trend was established to account for the change from the 30 ft-lb mean-fit value to the 50 ft-lb minimum value, as required by ASME Code,Section III, NB-2331. The BWRVIP-173-A recommendation of an additional 30 0 F is not needed herein since the impact energy values have been reduced to 65% of the original values to account for the unknown orientation. The most conservative slope was calculated to be 3°F per ft-lb. This value was used for the four unknown nozzles to approximate the initial RTNDT values.

A summary of the limiting inlet and outlet nozzle initial RTNDT values for D. C. Cook Units 1 and 2 is presented in Table 9.

Table 9 Summary of the Limiting Initial RTNDT Values for the Inlet and Outlet Nozzles for D. C. Cook Units 1 and 2 D. C. Cook Unit No. Nozzle Limiting Initial RTNDT Value (°F)

Inlet Nozzle -7 Unit 1 Outlet Nozzle -61 Inlet Nozzle 0 Unit 2 Outlet Nozzle0

Westinghouse Non-Proprietary Class 3 Page 14 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Nozzle P-T Limits A calculation of the D. C. Cook Units 1 and 2 nozzle cooldown P-T limits was completed using the inlet and outlet nozzle initial RTNOT values. The stress intensity factor correlations used for the nozzle corners are consistent with the ASME PVP201 1-57015 (Reference 10) and Oak Ridge National Laboratory (ORNL) study, ORNL/TM-2010/246 (Reference 11). The methodology used included postulating an inside surface 1/4T nozzle corner flaw, along with calculating through-wall nozzle corner stresses for a cooldown rate of 100°0F/hour.

The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form:

o5= Ao+ A lx+A2 x2+A 3x3 where,

=through-wall stress distribution x = through-wall distance from inside surface A0 , A1 , A2, A3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below The stress intensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010/246. The stress intensity factor expression for a rounded corner is:

v K /-- [0.706A0+ 0.537 (-2-) A1 + 0.448 (-)A 2+ 0.393 (,-i)A ] 3 where, KI = stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius corner a = crack depth at the nozzle corner, for use with 1/4T (25% of the wall thickness)

The D. C. Cook Unit 1 nozzle P-T limit curve shown in Figure 1 and the D. C. Cook Unit 2 nozzle P-T limit curve shown in Figure 2 are based on the stress intensity factor expression discussed above; also shown in these figures are the traditional beltline P-T limits from Technical Specifications (TS) Figure 3.4.3-2 for D. C. Cook Units 1 and 2 (Reference 1). The basis documents for the D. C. Cook Units 1 and 2 TS P-T limit curves are WCAP-15878, Rev. 0 (Reference 2) and WCAP-15047, Rev. 2 (Reference 3), respectively. The TS figures included the same P-T limit curves from WCAP-15878, Rev.

0 and WCAP-15047, Rev. 2, with the addition of allowable pressures to address vacuum refill. Note that the figures show the most limiting P-T limit curve of the inlet and outlet nozzles for each unit. The nozzle P-T limits are provided for a cooldown rate of 100°F/hr, along with a steady-state curve.

Westinghouse Non-Proprietary Class 3 Page 15 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 It should be noted that an outside surface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is not provided, since it would be less limiting than the cooldown nozzle P-T limit curve in Figures 1 and 2 for an inside surface flaw.

Based on the results shown in Figures 1 and 2, it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits contained in the LAR (Reference 1), which are based on the technical evaluations contained in WCAP-15878, Rev. 0 and WCAP-15047, Rev. 2 for 32 EFPY, are still applicable for the beltline and non-beltline reactor vessel components of D. C. Cook Units 1land 2.

Westinghouse Non-Proprietary Class 3 Page 16 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 2500 2250 2000 1750 0, 1500 C,m 0/)

I1) 1250 U_

1000 4-I On Co 750 CU 0v 500 250

-250 ... ]I I . . .. , ,RCS

4. 7psi Vacuum s T.j1.. .. I. . .. . . . +I . .

0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (°F)

Figure 1: Comparison of D. C. Cook Unit 1 32 EFPY TS P-T Limits to Nozzle Limits

Westinghouse Non-Proprietary Class 3 Page 17 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 2500 2250 2000 1750 1500 L-

0. 1250 1000 I-o-

o U) o1-750 U,

500 250 0

-250 0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (0F)

Figure 2: Comparison of D. C. Cook Unit 2 32 EFPY TS P-T Limits to Nozzle Limits

Westinghouse Non-Proprietary Class 3 Page 18 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Other Ferritic Components in the Reactor Coolant Pressure Boundary 10 CFR Part 50, Appendix G (Reference 5), requires that all RCPB components meet the requirements of Section III of the ASME Code. The ferritic RCPB components that are not part of the RV consist of the replacement steam generators (RSG) and the original Westinghouse designed pressurizers. The D. C. Cook Units 1 and 2 pressurizers were constructed to the 1965 Edition Section III ASIME Code through the Winter 1966 Addenda. The D. C. Cook Unit 1 replacement steam generators were designed and evaluated to the 1989 Edition ASME Section III Code. For D. C. Cook Unit 2, the upper assembly steam generators were designed and evaluated to the 1968 Edition Section III ASME Code through Winter 1968 Addenda, while the replaced lower assemblies were designed and evaluated to the 1983 Edition Section III ASME Code through Summer 1984 Addenda. These components met all applicable requirements at the time of construction; therefore, no further consideration is necessary for these components with regards to P-T limits.

The lowest service temperature (LST) requirement of NB-2332(b) of ASME Section ITI is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 1/22 inches (Reference 12). Note that the D. C. Cook Units 1 and 2 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps or valves. Therefore, the LST requirements of NB-2332(b) are not applicable to the D. C. Cook Units 1 and 2 P-T limits.

Westinghouse Non-Proprietary Class 3 Page 19 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 Conclusions NRC RAI-EVIB-1 for D. C. Cook Units 1 and 2 was addressed to ensure that the LAR P-T limit curves for 32 EFPY (Reference 1) remain bounding with consideration of beltline and non-beitline reactor vessel components. The extended beltline materials that achieve fluence values > 1 x 1017 n/cm2 through 32 EFPY were defined, and ART values were calculated for 32 EFPY. Tables 7 and 8 compare the ART values used in the LAR P-T limit curves with the extended beltline ART values. Tables 7 and 8 demonstrate that the ART values used to develop the LAR P-T limit curves bound the extended beltline material ART values through 32 EFPY for D. C. Cook Units 1 and 2.

Even though neutron embrittlement does not need to be considered for the D. C. Cook Units 1 and 2 reactor vessel inlet and outlet nozzles through 32 EFPY, the inside corner regions of the reactor vessel nozzles are considered to be highly stressed. P-T limit curves were developed for the reactor vessel nozzles using the limiting initial RTNOT values. Figures 1 and 2 display the limiting nozzle P-T limit curves, and demonstrate that the LAR P-T limit curves bound the reactor vessel nozzle P-T limit curves through 32 EFPY.

Additionally, other ferritic components in the RCPB were considered. The D. C. Cook Units 1 and 2 steam generators and pressurizers met all applicable ASME Code Section III requirements at the time of construction; therefore, no further consideration is necessary for these components with regards to P-T limits.

Furthermore, the LST requirements of NB-2332(b) are not applicable to the P-T limits since the D. C. Cook Units 1 and 2 reactor coolant systems do not have ferritic materials in the Class 1 piping, pumps or valves.

In conclusion, the D. C. Cook Units 1 and 2 LAR P-T limit curves remain applicable for 32 EFPY and no further changes are necessary in response to NRC RAI-EVIB-1.

Westinghouse Non-Proprietary Class 3 Page 20 of 20 MCOE-LTR-15-82, Rev. 0 Attachment A September 2, 2015 References

1. U.S. NRC Correspondence,

Subject:

"Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Change the Reactor Coolant System Pressure and Temperature Limits RE: (TAC NOS. MF4280 AND MF4281)," dated October 1, 2014. [NRC ADAMS Accession number ML14259A549]

2. WCAP-15878, Revision 0, "D. C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation for 40 Years and 60 Years," December 2002.

3. WCAP-15047, Revision 2, "D.C. Cook Unit 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," May 2002.

4.. Westinghouse Report WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit. Curves,"

January 1996.

5. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.

6. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

7. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

8. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.

9. BWR VIP-1 73-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle ForgingMaterials. EPRI, Palo Alto, CA: 2011. 1022835.

10. ASMIE PVP201 1-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.

11. Oak Ridge National Laboratory Report, ORNL/TM-20 10/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles -

Revision 1," June 2012.

12. ASMLE B&PV Code Section III, Division I, NB-2332, "Material for Piping Pumps, and Valves, Excluding Bolting Material."

13. NRC Email from Terry Beltz (NRC) to Helen Etheridge (AEP),

Subject:

"D. C. Cook Nuclear Plant, Units 1 and 2 - Draft Requests for Additional Information re: Change to RCS P-T Limit Curves to Support Vacuum Fill (TAC Nos. M1F4280 and MF4281)," dated July 21, 2014. [NRC ADAMS Accession number ML14217A325]