NRC 2014-0016, Submittal of License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)- Response to Request for Additional Information
| ML14071A405 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/11/2014 |
| From: | Mccartney E Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2014-0016, TAC MF0532, TAC MF0533 | |
| Download: ML14071A405 (24) | |
Text
WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 NEXTera" ENERGY~
.~
March 11, 2014 NRC 2014-0016 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR)
Response to Request for Additional Information
References:
(1) NextEra Energy Point Beach, LLC letter to NRC, dated January 15, 2013, License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
(2) NextEra Energy Point Beach, LLC letter to NRC, dated March 1, 2013, Supplement to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (ML13063A292)
(3) NRC electronic mail to NextEra Energy Point Beach, LLC, dated July 11, 2013, Point Beach Nuclear Plant, Units 1 and 2- Draft Requests for Additional Information (EVIB) Supporting Review of LAR to Implement New PTLR (TAC Nos. MF0532 and MF0533)
(4) NextEra Energy Point Beach, LLC letter to NRC, dated September 12, 2013 License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) Response to Request for Additional Information (ML 13256AS064)
Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively (Reference 1). The proposed amendments would Enclosure 2 to this Jetter contains proprietary information.
Withhold from public disclosure under 10 CFR 2.390.
Upon removal of Enclosure 2, this letter is uncontrolled.
NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
Document Control Desk Page 2 revise the PBNP Technical Specifications (TS) to allow the use of two new methodologies; Framatome ANP Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNor of Linde 80 Weld Materials," and Westinghouse Owners Group (WOG) WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." The revision would add BAW-2308, Revisions 1-A and 2-A and WCAP-14040-A, Revision 4, as approved methodologies toTS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits. This request was supplemented with additional information in Reference (2).
Via Reference (3), the NRC determined additional information was required to enable the staff's continued review of this license amendment request. NextEra responded to questions 1 and 2 in Reference (4). NextEra committed to answering questions 3 and 4 of Reference (3) within 45 days of receiving the information from Westinghouse. NextEra received the needed information from Westinghouse on February 20, 2014. contains NextEra's response to questions 3 and 4 of Reference (3). Enclosure 2 contains information proprietary to EPRI in the response to question 4. This information is supported by an affidavit signed by EPRI, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses the considerations listed in Paragraph (b)(4) of Section 2.390 of the Commission's regulations. contains the EPRI authorization letter, dated February 27, 2014, "Request for Withholding of the following Proprietary Information Included in: Point Beach Units 1 and 2 Pressure-Temperature Limits License Amendment Request: NRC Request for Additional Information Response Westinghouse Reference Document MCOE-LTR-13-115, Rev 0," with accompanying affidavit. This affidavit is associated with the information in Enclosure 2. It is requested that the information which is EPRI proprietary be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting EPRI affidavit should reference this letter and should be addressed to Neil Wilmhurst, Vice President and Chief Nuclear Officer, Electric Power Research Institute, Inc., 1300 West W.T. Harris Blvd, Charlotte, NC, 28262-8550.
This letter completes the following commitment:
- NextEra Energy Point Beach, LLC will contract with Westinghouse to provide support in responding to NRC Requests for Additional Information (RAis) #3 and #4. This information will be provided to the NRC within 45 days of receipt from Westinghouse.
There are no new regulatory commitments and no changes to existing commitments.
Document Control Desk Page 3 The information contained in this letter does not alter the no significant hazards consideration contained in Reference (1) and continues to satisfy the criteria of 10 CFR 51.22 for categorical exclusion from the requirements of an environmental assessment.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 11, 2014.
Very truly yours, NextEra Energy Point Beach, LLC
~~~
Eric McCartney U Site Vice President Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC
ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NON-PROPRIETARY COPY The NRC staff determined that additional information was required (Reference 15 of this enclosure) to enable the continued review of the Point Beach Nuclear Plant (PBNP) License Amendment Request 252, Technical Specification 5.6.5, Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR) (References 1 and 2). The following information is provided by NextEra Energy Point Beach, LLC (NextEra) in response to the NRC staff's request.
Please provide essential parameters such as thermal diffusivity for the NRC staff to validate the difference between the RPV coolant temperature and the RPV metal temperature.
Additionally clarify how RPV metal temperature at the crack tip was derived.
NextEra Response The RV metal temperature at the crack tip is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (Reference 5). The time-dependent temperature solution utilized in both the heatup and cooldown analysis is based on the one-dimensional transient heat conduction equation:
ar = a [o T +taT]
2 at or rar 2
The following boundary conditions applied at the inner and outer radii of the reactor vessel:
At r = ri: q(ri,t) c h(Tc -T)
At r = ro: q(r0 ,t) = 0
- Where, ri = reactor vessel inner radius ro =reactor vessel outer radius a = thermal diffusivity T = local temperature r = radial location t =time q = boundary heat flux h = heat transfer coefficient between the coolant and the vessel wall Tc =coolant temperature Page 1 of 17
These equations are solved numerically to generate the position and time-dependent temperature distributions, T(r,t), for all heatup and cooldown rates of interest. The thermal diffusivity is 0.518 fethr at 70°F and 0.379 fethr at 550°F. The convective heat-transfer coefficient is set as a constant 7000 Btu/hr-ft2-°F.
The regulations in 10 CFR Part 50, Appendix G, Paragraph IV. A state that, "the pressure-retaining components of the reactor coolant pressure boundary [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, Section Ill], supplemented by the additional requirements set forth in [paragraph IV.A.2, "Pressure-Temperature (P-T) Limits and Minimum Temperature Requirements'1- .. " Therefore, 10 CFR Part 50, Appendix G requires that P- T limits be developed for the ferritic materials in the reactor vessel (RV) belt/ine (neutron f/uence ;;::::
1 x 10 17 nlcm 2, E > 1 MeV), as well as ferritic materials not in the RV belt/ine (neutron f/uence <
17 2 1 x 10 nlcm , E > 1 MeV). Further, 10 CFR Part 50, Appendix G requires that all RCPB components must meet the ASME Code, Section Ill requirements. The relevant ASME Code, Section Ill requirement that will affect the P- T limits is the lowest service temperature requirement for all RCPB components specified in Section Ill, NB-2332(b).
The P- T limit calculations for ferritic RCPB components that are not RV beltline shell materials may define P-T curves that are more limiting than those calculated for 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 RV beltline shell region. These higher stresses can potentially result in more restrictive P- T limits, even if the reference temperature (RTNoT) 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 RTNoT values, which may define a more restrictive lowest operating temperature in the P -T limits than those for the RV beltline shell materials.
Consequently, please describe how the P-T limit curves submitted for PBNP units and the methodology used to develop these curves, considered all RV materials (beltline and non-beltline) and the lowest service temperature of all ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G.
Your description should include the following:
- Confirm availability of material data (initial RTNoT and copper and nickel contents) for all non-beltline ferritic materials for all PBNP RPVs and demonstrate that none of them will become limiting under the 50 EFPY f/uence.
- Confirm that the lowest service temperatures (LSTs) for all ferritic RCPB components that are not part of the RV have been established for both Point Beach units, and the lowest temperature of 60 oF in the proposed P- T limits are higher than these LSTs.
Page 2 of 17
NextEra Response Reactor Vessel Beltline Components The reactor vessel beltline and extended beltline materials were previously analyzed in WCAP-16669-NP, Revision 1 (Reference 3) for P-T limits. As documented in Tables 3 and 4 of the Pressure Temperature Limits Report (PTLR) markups contained in Enclosure 2 of Reference 2, the nozzle belt forging materials and nozzle belt to intermediate shell circumferential welds for Point Beach Units 1 and 2 are projected to receive fluence values greater than 1 x 1017 n/cm 2 at 50 EFPY. Note that these materials are collectively considered to be the extended beltline region materials; whereas, the traditional beltline region materials include the intermediate shell, lower shell and the associated welds. It is noted that although the nozzle belt forgings and associated nozzle belt to intermediate shell circumferential welds 17 are projected to exceed the 1 x 10 n/cm 2 threshold, the reactor vessel inlet and outlet nozzle forgings as well as the inlet and outlet nozzle forging to nozzle belt welds remain below 1 x 1017 n/cm2 through 60 EFPY (See Reactor Vessel Non-Beltline Components Discussion below).
Reactor Vessel Non-Beltline Components WCAP-14040-A, Revision 4 does not consider the embrittlement of ferritic materials in the area adjacent to the beltline region, specifically the stressed inlet and outlet nozzles. The inside corner regions of these nozzles are the most highly stressed ferritic components outside the beltline region of the reactor vessel. The reactor vessel inlet and outlet nozzles also receive the most irradiation outside of the beltline and extended beltline regions. Due to these factors, only these components are analyzed in this section.
The ART values for the Point Beach Units 1 and 2 inlet and outlet nozzle corner regions were calculated at 60 EFPY for each reactor vessel inlet and outlet nozzle forging. These ART values were determined using the methodology contained in Regulatory Guide 1.99, Revision 2 (Reference 10) along with the inputs described below.
Nozzle Chemistry Data Nickel weight percent (wt%) values were obtained directly from the material-specific analyses documented in the respective Certified Material Test Report (CMTR) for each of the Point Beach Units 1 and 2 reactor vessel inlet and outlet nozzles. Similarly, the copper weight percent values were also obtained directly from the material-specific analyses documented in the CMTRs for each of the Point Beach Unit 1 reactor vessel inlet and outlet nozzles. The Point Beach Unit 2 CMTRs did not contain copper wt% values because at the time that the nozzles were manufactured, it was not required for SA-508, Class 2 low-alloy steel. Therefore, a best-estimate Cu wt% value of (( )) from Section 4 of the NRC-approved BWRVIP report, BWRVIP-173-A (Reference 11 ), was utilized for the Point Beach Unit 2 inlet and outlet nozzles.
A mean plus two standard deviations confidence limit was applied to the data in BWRVIP-173-A to determine a conservative Cu wt% value of (( )). The data in the BWRVIP report were tabulated from an industry-wide database of SA-508, Class 2 forging materials. Therefore, as stated above, the conservative best-estimate Cu wt% from the BWRVIP report of (( )) was assigned to the Point Beach Unit 2 inlet and outlet nozzles.
Page 3 of 17
The Chemistry Factor (CF) values used in this NRC RAI response were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 10) methodology. The Point Beach Unit 1 CF values were calculated using the material-specific weight percent copper and nickel values along with Table 2 of Regulatory Guide 1.99, Revision 2. The Point Beach Unit 2 CF values were calculated using the weight percent copper value of (( )) from BWRVIP-173-A and the material-specific weight percent nickel values along with Table 2 of Regulatory Guide 1.99, Revision 2.
Nozzle Initial RTNor Values The initial RTNoT values were determined for each of the Point Beach Units 1 and 2 reactor vessel inlet and outlet nozzle forging materials using the BWRVIP-173-A, Alternative Approach 2 methodology, contained in Appendix B of that report. For all eight of the Point Beach inlet and outlet nozzle materials, CVGraph Version 5.3 was utilized to plot the material-specific Charpy V-Notch impact energy data from the CMTRs to determine 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 RT NoT values for the inlet and outlet nozzle materials for Point Beach Units 1 and 2. It should be noted that the orientation of the Charpy V-Notch forging specimens 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 per the Alternative Approach 2 methodology in BWRVI P-173-A.
Nozzle Neutron Fluence Values The maximum 60 EFPY fast neutron (E > 1 MeV) exposure of the Point Beach Units 1 and 2 reactor pressure vessel inlet and outlet nozzle materials was calculated using an NRC approved methodology that follows the guidance and meets the requirements of Regulatory Guide 1.190 (Reference 7). The overall analytical methodology is described in References 5, 8, and 9. The NRC approvals are noted in References 5 and 9.
The Point Beach Units 1 and 2 maximum calculated neutron fluence projections at 60 EFPY for the inlet and outlet nozzle forging materials are summarized in Table 1. The inlet nozzles are projected to achieve a maximum fluence of 4.57 x 10 16 n/cm 2 and 4.60 x 10 16 n/cm 2 (E > 1 MeV) at the lowest extent of the nozzles at 60 EFPY for Units 1 and 2, respectively. Similarly, the outlet nozzles are projected to achieve a maximum fluence of 4.19 x 10 16 n/cm 2 and 4.22 x 10 16 n/cm 2 (E > 1 MeV) at the lowest extent of the nozzles at 60 EFPY for Units 1 and 2, respectively. Note that the fluence values used in the ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside corner of the nozzle, for conservatism.
Page 4 of 17
Table 1 Point Beach Units 1 and 2 Calculated Neutron Fluence Projections on the Reactor Vessel Nozzle Materials at 60 EFPY Fluence1a1 2
Reactor Vessel Material (n/cm , E > 1.0 MeV)
Inlet Nozzle Forgings 4.57E+16 Unit 1 Outlet Nozzle Forgings 4.19E+16 Inlet Nozzle Forgings 4.60E+16 Unit 2 Outlet Nozzle Forgings 4.22E+16 Note for Table 1:
(a) Fluence values at the lowest extent of the nozzles.
Nozzle ART Values The ART values for the nozzle corner regions were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 10) methodology and are documented in Tables 2 and 3 for Point Beach Units 1 and 2, respectively. The ART values were conservatively calculated using the maximum fluence at the lowest extent of the nozzle, rather than at the standard 1/4T (25% of the wall thickness) location. Note that the ART values for the inlet and outlet nozzle forgings were conservatively calculated at 60 EFPY, even though the applicability date of the proposed P-T limit curves decreased from 53 EFPY to 50 EFPY with consideration of the extended power uprate (EPU), as described in the NextEra Energy RAI 1 response in Reference 4. Since Point Beach Units 1 and 2 only have one set of traditional beltline curves that are used for both units, the limiting ART values for the inlet and outlet nozzle between the two units were then used for the 1/4T flaw evaluation at the nozzle corner region.
Page 5 of 17
Table 2 ART Calculations for the Point Beach Unit 1 Reactor Vessel Nozzle at 60 EFPY CF(bl Maximum RT NDT(U)(dl a/l(e)
Wt% Wt% FF<cl .dRT NOT a1 Margin ART Reactor Vessel Material Ni(a) Fluence<cl cu<a) (oF) 2 COF) (oF) (oF) (oF) (oF) (oF)
(n/cm , E > 1.0 MeV)
Inlet Nozzle ZT2670-1 0.12 0.67 84.1 4.57E+16 0.0627 -52 5.3 0 2.6 5.3 -41.5 Inlet Nozzle ZT2670-2 0.12 0.67 84.1 4.57E+16 0.0627 -45 5.3 0 2.6 5.3 -34.5 Outlet Nozzle BT2305 0.14 0.73 103.3 4.19E+16 0.0587 1 6.1 0 3.0 6.1 13.1 Outlet Nozzle ZT2630-1 0.12 0.75 85.3 4.19E+16 0.0587 -20 5.0 0 2.5 5.0 -10.0 Notes for Table 2:
(a) The Cu and Ni wt% values are material-specific values as documented in each respective material CMTR.
(b) CF values were calculated using the Cu and Ni wt% values and Table 2 of Regulatory Guide 1.99, Revision 2.
(c) Maximum fluence values at the lowest extent of the nozzles for the inlet and outlet forgings at 60 EFPY. Fluence Factor (FF) values were calculated using Regulatory Guide 1.99, Revision 2.
(d) RTNoT(U) values were determined using the Alternative Approach 2 methodology as described in Appendix 8 of BWRVIP-173-A.
(e) Per Regulatory Guide 1.99, Revision 2, the base metal nozzle forging materials CJa = 1rF for Position 1.1 without surveillance data. However, CJa need not exceed 0.5*~RTNDT*
Page 6 of 17
Table 3 ART Calculations for the Point Beach Unit 2 Reactor Vessel Nozzle at 60 EFPY CF(bl Maximum RT NDT(U) (dl aA(e)
Wt% Wt% FF(c) .dRTNoT a, Margin ART Reactor Vessel Material Ni(a) Fluence<cl cu<a} (oF) 2 (oF) COF) ("F) COF) (oF) (oF)
(n/cm , E > 1.0 MeV)
Inlet Nozzle 9-5414 (( )) 0.84 141.6 4.60E+16 0.0630 -15 8.9 0 4.5 8.9 2.8 Inlet Nozzle 9-5742 (( )) 0.85 141.8 4.60E+16 0.0630 -76 8.9 0 4.5 8.9 -58.1 Outlet Nozzle 9-5691 (( )) 0.77 140.0 4.22E+16 0.0590 -37 8.3 0 4.1 8.3 -20.5 Outlet Nozzle 9-5716 (( )) 0.82 141.3 4.22E+16 0.0590 -20 8.3 0 4.2 8.3 -3.3 Notes for Table 3:
(a) Cu wt% values are the best-estimate values for SA-508, Class 21ow-alloy steel as documented in BWRVIP-173-A. The Ni wt% values are material-specific values as documented in each respective material CMTR.
(b) CF values were calculated using the Cu and Ni wt% values and Table 2 of Regulatory Guide 1.99, Revision 2.
(c) Maximum fluence values at the lowest extent of the nozzles for the inlet and outlet forgings at 60 EFPY. Fluence Factor (FF) values were calculated using Regulatory Guide 1.99, Revision 2.
(d) RTNoT(UJ values were determined using the Alternative Approach 2 methodology as described in Appendix B of BWRVIP-173-A.
(e) Per Regulatory Guide 1.99, Revision 2, the base metal nozzle forging materials cra = 1rF for Position 1.1 without surveillance data. However, aa need not exceed 0.5*t.RTNOT.
Page 7 of 17
A summary of the limiting inlet and outlet nozzle ART values at Point Beach is presented in Table 4. Again, since Point Beach Units 1 and 2 only have one set of traditional beltline curves that are used for both units, the limiting ART values for the inlet and outlet nozzle between the two units were then used for the 1/4T flaw evaluation at the nozzle corner region.
Table 4 Summary of the Limiting ART Values for the Point Beach Inlet and Outlet Nozzle Materials Limiting Nozzle Material Limiting ART Value EFPY and ID Number (oF)
Unit 2 Inlet Nozzle 9-5414 2.8 60 Unit 1 Outlet Nozzle 13.1 BT2305 Nozzle P-T Limits A calculation of the Point Beach nozzle cooldown P-T limits was completed using the limiting inlet and outlet nozzle ART values at 60 EFPY to account for nozzle embrittlement. The stress intensity factor correlations used for the nozzle corners are consistent with the ASME PVP2011-57015 (Reference 12) and ORNL study, ORNL/TM-201 0/246 (Reference 13). 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°F/hour. The stresses used for the Point Beach Units 1 and 2 nozzle corners are based on a representative 2-loop Pressurized Water Reactor (PWR) nozzle analysis, which was based on a three-dimensional (3-D) finite element model (FEM).
The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form:
cr = Ao+ A1 x+A2x 2+A3x3 where, a =through-wall stress distribution x =through-wall distance from inside surface A 0 , A 1 , A 2 , A 3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below Page 8 of 17
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-201 0/246. The stress intensity factor expression for a rounded corner is:
where, K, = 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 Point Beach Units 1 and 2 inlet and outlet nozzle P-T limit curves are shown in Figures 1 and 2, respectively, based on the stress intensity factor expressions discussed above. The nozzle P-T limits are provided for a cooldown rate of -100°F/hr, along with a steady-state (SS) curve valid to 60 EFPY. Also shown in these figures is the traditional beltline P-T limit curve for cooldown and steady-state operation, which is valid through 50 EFPY (Reference 4). Again, note that the applicability date of the proposed P-T limit curves decreased from 53 EFPY to 50 EFPY with consideration of the extended power uprate (EPU), as described in the NextEra Energy RAI 1 response in Reference 4.
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. Similarly, the traditional beltline P-T limit curve for heatup is also not shown in Figures 1 and 2, since it is less limiting than the traditional beltline cool down P-T limit curve for comparison purposes to the developed nozzle curves.
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 provided in the LAR, which are based on Figures 5-7 and 5-8 ofWCAP-16669-NP, for 50 EFPY (Reference 4) are still applicable for the beltline and non-beltline reactor vessel components.
Page 9 of 17
Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 50 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wiKic) 2500 I I
,_ Inlet Nozzle Cool down 2250 Steady State I I I j'.... Inlet Nozzle Cool down 2000 ~~
-100 ("F/hr) 1750 Acceptable 6'
00 1500 I -
Operation e:,
J
~
=
~
1250 --
~""'
"0
~
=
..st 1000 u
~
Traditional Beltline Cooldown Rates
("F/hr)-
Steady State 750 -20 500 -- I
-40
-60
-100 250 --- --- 1----
0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 1: Comparison of Point Beach Beltline P-T Limits to Inlet Nozzle Limits*
- Note: The Point Beach beltline P-T limit curves were originally developed in WCAP-16669-NP, Revision 1 (Reference 3) for 53 EFPY; however, with consideration of the EPU, the P-T limits applicability date decreased to 50 EFPY as described in the NextEra Energy RAI 1 response in Reference 4.
Page 10 of 17
Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100"Fihr) Applicable for 50 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wiK1c) 2500 I I I
Outlet Nozzle r.- Cooldown 2250 Steady State I
2000 l Outlet Nozzle Cooldown
-100 ("F/hr) 1750 II Acceptable Operation G'
.... 1500 00 e::,
J Q,l
=
"'"' 1250 Q,l 1/
~
"0 Q,l
~
.;= 1000 r"
Traditional Beltline u Cooldown Rates
("F/hr)-
Steady State 750 -20
-40
,..,., -60
-100 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2: Comparison of Point Beach Beltline P-T Limits to Outlet Nozzle Limits*
- Note: The Point Beach beltline P-T limit curves were originally developed in WCAP-16669-NP, Revision 1 (Reference 3) for 53 EFPY; however, with consideration of the EPU, the P-T limits applicability date decreased to 50 EFPY as described in the NextEra Energy RAI 1 response in Reference 4.
Page 11 of 17
Other Ferritic Components in the Reactor Coolant Pressure Boundary The lowest service temperature (LST) requirement of NB-2332(b) is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 Y2 inches (Reference 14). Note that the Point Beach Units 1 and 2 reactor coolant systems do 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 Point Beach P-T limits.
The other ferritic RCPB components that are not part of the RV consist of the original Series 84 Pressurizers and the Replacement Steam Generators (RSG). The original Series 84 Point Beach Units 1 and 2 Pressurizers were constructed to the ASME Section 1111965 Code Edition with Summer 1966 Addendum and met all applicable requirements at the time of construction.
Since these components have not been replaced and have not undergone neutron embrittlement, further consideration of these components for P-T limits is not required. The Point Beach Unit 2 Delta 47 replacement steam generators (RSGs) have been evaluated based on the 1986 ASME Section Ill, Appendix G requirements. Since the Point Beach Unit 2 RSGs ferritic components meet the ASME Code, Section Ill Appendix G requirements for protection against non-ductile fracture analysis, no further consideration is necessary for these components with regards to P-T limits. A representative ASME Section Ill Appendix G evaluation was not performed for the Point Beach Unit 1 Model 44F RSGs; therefore, the discussion in the subsequent pages will demonstrate that the P-T limits for the Point Beach Unit 1 RSG components are less limiting than the P-T limit curves developed for the reactor vessel beltline in WCAP-16669-NP (Reference 3).
Unit 1 Replacement Steam Generator A representative ASME Section Ill Appendix G evaluation was not performed for the Point Beach Unit 1 RSGs. In order to demonstrate that the P-T limit curves in WCAP-16669-NP (Reference 3), as supplemented by Reference 4, for the Point Beach Unit 1 RV beltline region bounds the RSG ferritic components, two critical locations in the RSG were reviewed based on the ASME Section XI, Appendix G fracture mechanics analysis. The first location is the steam generator (SG) channel head to tube sheet region lower junction, and the second location is the primary nozzle knuckle region (nozzle corner). These two locations are highly stressed regions in the SG and contain discontinuities that should be considered for a non-ductile failure evaluation.
SG Tube Sheet to Channel Head Junction To determine if the SG tube sheet to channel head junction location is more limiting for P-T limits than the RV beltline, allowable pressures based on ASME Section XI, Appendix G are calculated at several temperature values and compared with that of the RV beltline region for Point Beach.
For the fracture mechanics analysis at the RSG tube sheet to the channel head junction, inside surface axial and circumferential flaws with aspect ratios of 6:1 are considered, per ASME Section XI, Appendix G. The cooldown transient (ramp down of 100°F/hr) is also considered, as it will produce high tensile stresses on the inside surface for pressure and thermal transients.
The limiting tensile stress components are chosen to determine the primary and secondary stress intensity factors, with the appropriate safety factors, and membrane and bending geometric factors consistent with the ASME Section XI, Appendix G evaluation.
Page 12 of 17
The initial RT NDT value is based on the design specification for the Point Beach Unit 1 RSG ferritic materials (base metals and welds); this value is conservatively set to 60°F for the tube sheet junction location.
The limiting P-T limit values for the SG tube sheet junction are plotted on Figure 3 and compared with the Point Beach RV beltline P-T limits for the cooldown transient. The SG tube sheet junction P-T limit values are based on a circumferential flaw, and are more limiting than the P-T limit values determined for an axial flaw. The data from Figure 3 demonstrates that the RV beltline P-T limit curves, from WCAP-16669-NP for Point Beach, bound the tube sheet to channel head junction P-T limit curve for Point Beach Unit 1.
Page 13 of 17
Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 50 EFPY (with Hafilium Removal and without Margins for Instrumentadon Errors) Using 1998 App. G Methodology (wiK1J 2500 ........ , ... .....
l Point Beach 2250 r Unit 1 I****
Steam Generator Tube Sheet 2000 Junction Cooldown-100eF/hr) 1750 .... I** . . . . . . -** -+
Acceptable
~
Operation
.... r ......... . . ..
/--
v Traditional
' Beltiine Cooldown Rates eF/hr)-
750
__..... Steady State
-20
-40
-60 500 -100 250 -** ~ . . . .. *- *****
o~~~~~~~~~~~~~~~~~~~~~~~~~~
0 50 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Moderator Temperature (Deg. F)
Figure 3: P-T Limit Curve Comparison of Tradition Beltline and the RSG Tube Sheet to Channel Head Junction for Point Beach Unit 1*
- Note: The Point Beach beltline P-T limit curves were originally developed in WCAP-16669-NP, Revision 1 (Reference 3) for 53 EFPY; however, with consideration of the EPU, the P-T limits applicability date decreased to 50 EFPY as described in the NextEra Energy RAI 1 response in Reference 4.
Page 14 of 17
Replacement Steam Generator Primary Nozzle Knuckle Region The Point Beach Unit 1 RSG primary nozzle knuckle region is also considered to be another potential limiting location for a non-ductile failure evaluation for an ASME Section XI, Appendix G evaluation due to the discontinuity at the nozzle corner region. Therefore, the RSG nozzle corner region was reviewed to determine its impact on the P-T limit curves as compared to the reactor pressure vessel P-T limit curves.
In order to demonstrate that the P-T limits for the Point Beach Unit 1 RSG primary nozzle knuckle regions are less limiting than the reactor vessel beltline, a comparison of the stresses and the component material properties was performed. In the response to the NRC RAI concerning the RV outlet nozzle corner region in the earlier section of this letter report, it was demonstrated that the P-T limits for the RV outlet nozzle corner regions are less limiting than the reactor vessel beltline P-T limits. For the RSG nozzle corner herein, it will be demonstrated that the RSG nozzle corner or knuckle region P-T limits are less limiting than the reactor vessel outlet nozzle corner for a 1/4T inside surface corner flaw using the stresses due to the cooldown transient.
Fracture Toughness Material Property Based on the plant-specific CMTR for the RSG primary nozzles, the RT NDT is +1 0°F. This maximum RT Nor is considered for evaluation as discussed below. Since the RSG nozzles experience negligible neutron irradiation, the maximum RT Nor of 1ooF for this material can be taken as the Adjusted Reference Temperature (ART). For the RV outlet nozzle corner P-T limit development, the limiting ART value for the RV outlet nozzle with the irradiation shift is 13.1 °F, as discussed in Table 4 of this report. As a result, for the entire cooldown transient, the fracture toughness K1c for the reactor vessel outlet nozzle corner regions are less (more limiting) than the RSG nozzles, since the RTNor for the reactor vessel outlet nozzles are higher than the RSG nozzles.
Primary and Secondary Stresses The Model 44F RSG has a larger inside knuckle radius than the reactor vessel outlet nozzle corners; therefore, the peak stresses for the Model 44F primary nozzle inside corner are less than the reactor vessel outlet nozzle corner.
The hoop pressure stresses due to a unit pressure stress of 1000 psi were compared to the RSG nozzle and the Point Beach Unit 1 reactor vessel outlet nozzle corners. Based on this comparison, the pressure stresses at the RSG nozzle corner region were less than the RV outlet nozzle corner regions from the inside surface up to 90% of the wall thickness. Therefore, it was concluded that the K1p (stress intensity factor due to pressure) at the RSG primary nozzle knuckle region is less than that of the RV outlet nozzles.
The Point Beach Unit 1 RSG primary nozzle wall thickness at the knuckle region is approximately 10.2 inches while the RV outlet nozzle corner through-wall thickness is approximately 18.1 inches. Therefore, the cooldown thermal transient hoop stresses for the RSG primary nozzles are also less than the RV outlet nozzles since the wall thickness of the RSG nozzles are smaller than the RV outlet nozzles at the location of interest. Therefore, it is Page 15 of 17
concluded that the K11 values for a postulated 1/4T flaw at the 10.2 inches RSG primary nozzles are also less than the RV outlet nozzle corner regions.
Pressure- Temperature Limit Discussion Since both the primary and secondary stress intensity factors for the RSG nozzles are less than the RV outlet nozzle corner regions, the total applied K1 for the RSG nozzles is also less than the RV outlet nozzles. Furthermore, the K1c allowable value for the RSG nozzle is greater than the RV outlet nozzles since the RTNor values for the RSG nozzles are less than the RV outlet nozzle. Therefore, it is concluded that the allowable pressures for the P-T limits for the RSG nozzle corner region are higher (less limiting) than the RV outlet nozzle corners. As a result, the RSG nozzle corners have P-T limit values that are less limiting than the RV beltline region for Point Beach Unit 1, since based on the earlier section of this letter report, the P-T limit curve for the RV outlet nozzle corners is bounded by the reactor vessel beltline region.
References
- 1. NextEra Energy Letter, "License Amendment Request 252 Technical Specification 5.6.5 Reactor Coolant System (RCS) Pressure and Temperature Limits Report {PTLR),"
January 15, 2013. (ADAMS Accession Number ML13016A028)
- 2. NextEra Energy Letter, "Supplement 2 to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report," April 18, 2013. (ADAMS Accession Number ML 13113AOOB)
- 3. Westinghouse Report WCAP-16669-NP, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," January 2009.
- 4. NextEra Energy Letter, "License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report Response to Request for Additional Information," September 12, 2013. (ADAMS Accession Number ML13256A064)
- 5. Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"
May 2004.
- 6. 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.
- 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. Westinghouse Report WCAP-15557, Revision 0, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology," August 2000.
- 9. Westinghouse Report WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluations of Light Water Reactor Dosimetry," May 2006.
- 10. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"
U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.
Page 16 of 17
- 11. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
- 12. ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.
- 13. Oak Ridge National Laboratory Report, ORNL/TM-201 0/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles- Revision 1,"June 2012.
- 14. ASME B&PV Code Section Ill, Division I, NB-2332, "Material for Piping Pumps, and Valves, Excluding Bolting Material," 2004 Edition.
- 15. NRC electronic mail to NextEra Energy Point Beach, LLC, dated July 11, 2013, Point Beach Nuclear Plant, Units 1 and 2- Draft Requests for Additional Information (EVIB) Supporting Review of LAR to Implement New PTLR (TAC Nos. MF0532 and MF0533)
Page 17 of 17
ENCLOSURE3 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ELECTRIC POWER RESEARCH INSITUTE REQUEST FOR WITHHOPLDING PROPRIETARY INFORMATION AND ACCOMPANYING AFFIDAVIT 3 pages follow
-=~1211
~~-
ELECTRIC POWER RESEARCH INSTITUTE NEIL WILMSHURST Vice President and Chief Nuclear Officer February 27, 2014 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Request for Withholding of the following Proprietary Information Included in:
Point Beach Units 1 and 2 Pressure-Temperature Limits License Amendment Request: NRC Request for Additional Information Response Westinghouse Reference Document MCOE-LTR-13-115, Rev. 0.
To Whom It May Concern:
This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified in the attached report. Proprietary and non-proprietary versions of the Response and the Affidavit in support of this request are enclosed.
EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the enclosed submittal to the NRC by NextEra Energy. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.
If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (650) 855-2271. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.
Sincero6 lJV Together ... Shaping the Future of Electricity 1300 West W.T. Harris Boulevard, CharloHe, NC 28262-8550 USA
- 704.595.2732
- Mobile 704.490.2653
- nwilmshurst@epri.com
e ~121l.-
ELECTRIC POWER RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In:
Point Beach Units 1 and 2 Pressure-Temperature Limits License Amendment Request: NRC Request for Additional Information Response Westinghouse Reference Document MCOE-LTR-13-115, Rev. 0.
I, Neil Wilmshurst, being duly sworn, depose and state as follows:
I am the Vice President and Chief Nuclear Officer at Electric Power Research Institute, Inc. whose principal office is located at 1300 West WT Harris Blvd, Charlotte, NC. ("EPRI") and I have been specifically delegated responsibility for the above-listed report that contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI.
EPRI Information is identified by double square brackets. ((This sentence is an example.)) Tables containing EPRI proprietary information are identified with double square brackets before and after the object. In each case, the superscript notation {E} refers to this affidavit as the basis for the proprietary determination.
EPRI requests that the Proprietary Information be withheld from the public on the following bases:
Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g., 10 C.F.R. § 2.390{a)(4)):
- a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.
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- c. The information sought to be withheld is considered to be proprietary for the following reasons. EPRI made a substantial economic investment to develop the Proprietary Information and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. If the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric and/or nuclear power industry, they would be able to use the Proprietary Information for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Proprietary Information.
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"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:
(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."
- e. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Proprietary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.
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I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of California.
Executed at 1300 W WT Harris Blvd being the premises and place of business of Electric Power Research Institute, Inc.
Dat~ -:-[- 'LJ -- 2cttr_
\Ljl&
Neil Wilmshurst (State of North Carolina)
(County of Mecklenburg)
Subscribed and sworn to (or affirmed) before me on this ~ 1~ay of ......s;;.~~~-* 20l!f. by
_11~1/~J,__./".o£)'-"'~:r.:..:..:..rec.q,k'.t@'l"'"'tJ4t~-------' proved to me on the basis of satisfac evidence to be the person(s) who appeared before me.
Signature QJ~ Jl- ~IJI#f (Seal)
My Commission Expires J-~ay of flrp&/ ,20_lj,