NRC 2013-0022, Supplement to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
| ML13063A292 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/01/2013 |
| From: | Meyer L Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2013-0022, TAC MF0532, TAC MF0533 | |
| Download: ML13063A292 (85) | |
Text
March 1,201 3 NRC 201 3-0022 10 CFR 50.90 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 Supplement to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant Svstem (RCS1 Pressure and Temperature Limits Report (PTLR)
References:
(1) NextEra Energy Point Beach, LLC letter to NRC, dated January 15,2012, License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (ML13016A028)
(2) NRC E-Mail to NextEra Energy Point Beach, dated 1 February 2013, Supplement to LAR re: P-T Limit Curves and LTOP Limits (TAC Nos.
MF0532 & MF0533)
In Reference (I), NextEra Energy Point Beach, LLC (NextEra) submitted a license amendment request to amend renewed Facility Operating License Nos. DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively. The proposed amendments would 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 FITNDT 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 to TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits.
In Reference (2), the NRC informed NextEra that supplemental information would be required in order for the LAR to meet acceptance review criteria. This letter provides the requested information. provides an engineering evaluation of the applicability of the PTLR using the requested methodologies. Enclosure 2 provides a mark-up of TRM 2.2, Pressure Temperature Limits Report. Enclosure 3 provides WCAP-16669-NP, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation."
NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
Document Control Desk Page 2 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.
Approval of the proposed amendment is requested by January 1, 2014. NextEra will implement the amendment within 180 days of Commission Approval.
This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.
The supplemental information to the LAR has been reviewed by the Plant Operations Review Committee.
In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 1,201 3.
Very truly yours, NextEra Eneryeoint
-- Beach, LLC P P C Y
Larry Meyer Site Vice President Enclosures cc:
Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW
ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
ENGINEERING EVALUATION OF THE APPLICABILITY OF THE PTLR USING THE REQUESTED METHODOLOGIES (MASTER CURVE)
- 1. PURPOSE AND SCOPE This engineering evaluation provides the basis for the mark-up of TRM 2.2, Pressure Temperature Limits Report (PTLR). The current PTLR curves in TRM 2.2 are applicable to 35.9 effective full power years (EFPY). This evaluation provides the basis for new PTLR applicable to 50 EFPY. The PTLR mark-up is based on WCAP-16669-NP which uses the two new methodologies, BAW-2308, Revision I
-A, and WCAP-14040-A, Revision 4, for determining RCS pressure and temperature (PT) limits with updated fluences from power uprate as identified in WCAP-16983-P.
The enclosed PTLR mark-up in Enclosure 2 is based on 53 EFPY fluence projections which assumed 10% power uprates occurred for both units in 2008. This engineering evaluation determines the effective period for these PTLR curves using fluence projections with 17% power uprates for both units in 201 1 (WCAP-16983-P).
- 2. BACKGROUND License Amendment Request 252 (Reference 5) proposes to amend the Point Beach Nuclear Plant (PBNP) Technical Specifications (TS) to allow the use of two new methodologies; Framatome ANP Topical Report BAW-2308, Revisions I
-A (Reference 2) and 2-A (Reference 3), "Initial FITNDT 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" (Reference 6). The revision would add BAW-2308, Revisions 1 -A and 2-A and WCAP-14040-A, Revision 4, as approved methodologies to TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits.
On January 31, 2013, a teleconference between the NRC and PBNP staffs was held to discuss the NextEra Energy Point Beach, LLC (NextEra) submittal. The submittal involves a significant change in PTLR methodologies and the NRC staff did not have sufficient information to evaluate use of the two new methodologies at PBNP. NextEra was informed by the NRC they would need to provide a marked-up PTLR that used the proposed new methodologies. Also, any supporting documents would need to be provided.
- 3. APPROACH Section 3 of WCAP-16669-NP describes the overall approach for the criteria for allowable pressure-temperature relationships. WCAP-16669-NP states the ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIc, for the metal temperature at that time. KI, is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code. The KIc curve is given by the following equation:
KIc = 33.2+20.734*e[O.O~(~-~~
- where, KI, = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTND~
Page 1 of 5
The following equation is used to calculate values of ART for each weld and plate or forging in the reactor vessel beltline.
ART = Initial RTNDT + ARTNDT + Margin ARTNDT is the mean value of the transition temperature shift, or change in ARTNDTI due to irradiation, and is calculated using the following equation:
The fluence value, f, is the end-of-life neutron fluence, in units of 10" n/cm2 (E > 1.0 MeV), for each vessel beltline material.
For each beltline material, the only variable that would affect the applicability of the PTLR curves is the end-of-life neutron fluence value. The fluence values for each beltline material in WCAP-16669-NP will be compared to corresponding fluence versus EFPY tables in WCAP-16983-P. As long as the fluence value in WCAP-16983-P is less than the fluence value in WCAP-16669-NP for a given material, the proposed PTLR curves will remain effective.
- 4. PROPOSED PTLR CURVES IN WCAP-16669-NP WCAP-16669-NP, "Point Beach Unit 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," was issued in December 2006. WCAP-16669-NP used fluence projections for 53 EFPY (End of Extended Life) based on the following assumptions:
Power uprates of 10% (1678 MWth) for both units would be implemented in 2008, and Hafnium absorbers in the peripheral fuel assembly locations would be removed from both units in 2008.
WCAP-16669-NP provides the methodology and results of the generation of heatup and cooldown PT limit curves for normal operation of the PBNP Units 1 and 2 reactor vessels.
These PT curves were generated based on the reactor vessel information and calculated fluences based on the plant operating conditions listed in Table 1-1 of WCAP-16669-NP.
Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the surface, 114T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the cladlbase metal interface.
The heatup and cooldown curves in WCAP-16669-NP were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the "axial flaw" and "circ flaw" methodologies of the 1998 ASME Code,Section XI through the 2000 Addenda was used, which makes use of the KIc methodology.
The ART values were obtained from AREVA calculation 32-901 9240-000, "ART Values for Point Beach Units 1 and 2." This document makes use of the NRC approved Topical Report BAW-2308, Revision 1 -A, "Initial FITNDT of Linde 80 Weld Materials." The fluence projections are based on the Ferret Code best-estimate values verified by Westinghouse in letter Page 2 of 5
WEP-06-13, dated February 14,2006, as approved for use by the NRC in WCAP-16083-NP-A, Revision 0.
The 2008 PT limit curves were generated for 43 and 53 EFPY using heatup rates of 60 and 1 OO°F/hr, cooldown rates of 0, 20, 40, 60 and 1 OO°F/hr, using uprated conditions for both units, and withlwithout hafnium removal. The curves were developed without margins for instrumentation errors. These curves can be found in Figures 5-1 through 5-8 of WCAP-16669-NP.
The PT limit curves for Case 3 in WCAP-16669-NP (1 0% power uprate in 2008,53 EFPY, and hafnium rods removed in 2008) shown in Table 1, were selected since this case has the highest fluence values.
Table 1 Plant Operating Conditions Reflecting Development of PT Limit Curves for 53 EFPY
- 5. EXTENDED POWER UPRATE CONDITIONS IN WCAP-16983-P WCAP-16983-P, Point Beach Units 1 and 2 Extended Power Uprate (EPU) Engineering Report, Section 5.1.2.4 provides results of the neutron fluence evaluations, including pressure vessel neutron exposure. The fluence projections are based on a 17% uprate (1 800 MWth) of both units in 201 1.
Hafnium Rods?
Removal October 2008 Removal April 2008 The fast neutron fluence (E > 1.0 MeV) values applicable to the cladlbase metal interface for the PBNP Units 1 and 2 pressure vessels are provided in Tables 5.1.2-1 and 5.1.2-2 of WCAP-16983-P. The neutron exposure tabulations include fuel-cycle-specific results for Unit 1 at the conclusion of Cycle 31 (29.7 EFPY) and for Unit 2 at the conclusion of Cycle 29 (29.1 EFPY). The tabulations also include projections for both units for operation through the new period of PT curve applicability based on an assumed core power distribution without hafnium absorbers in peripheral fuel assembly locations.
EFPY 53 53 Case 3
As presented, the data in Tables 5.1.2-1 and 5.1.2-2 of WCAP-16983-P represent the maximum exposure at the cladlbase metal intetface for each of the materials making up the beltline region of the reactor pressure vessel. The beltline region is defined as that portion of the pressure vessel that is anticipated to be exposed to a neutron fluence (E > 1.0 MeV) greater than 1.OE+17 n/cm2 over the service life of the unit. Pressure vessel materials not included in Tables 5.1.2-1 and 5.1.2-2 WCAP-16983-P are projected to be subjected to a neutron fluence < 1.OE+17 n/cm2 and are, therefore, not considered a part of the beltline region.
Page 3 of 5 Unit 1
2 Power (MWth) 151 8.5 startup to 2/3/2003 1540.0 2/3/2003 to 1012008 1678.0 1012008 to 53 EFPY 151 8.5 startup to 2/3/2003 1540.0 2/3/2003 to 412008 1678.0 412008 to 53 EFPY
The Low Temperature Overpressure Protection System (LTOPS) setpoint analysis for PBNP Units 1 and 2, January 2007, calculated the following LTOPS single setpoint applicable for 53 EFPY with and without hafnium rods for the entire LTOPS operating region:
LTOPS Single Setpoint = 420 psig (with RCP operating restriction, no more than one RCP in operation for RCS temperatures 5 180°F)
Section 4.3.4 of WCAP-16983-P concluded that for the EPU, the same LTOPS single setpoint remains acceptable with an arming temperature equal to 285OF.
- 6. COMPARISON OF PROJECTED FLUENCE VALUES The fluence values for each beltline material in WCAP-16669-NP were compared to corresponding fluence versus EFPY tables in WCAP-16983-P at 50 EFPY as shown in Table 2.
Table 2 Comparison of Projected Fluence for Point Beach Nuclear Plant Units 1 and 2 As illustrated in Table 2, the projected fluence values for all of the beltline materials are less than the fluence values used to generate the proposed PTLR curves. Thus, the PTLR curves remain conservative and valid through a 50 EFPY period of applicability.
Page 4 of 5 ID Fluence (xlOIY)
WCAP-16983-P 50 EFPY 0.36 4.79 4.35 0.36 3.13 3.1 3 4.25 2.94 0.49 4.76 4.57 0.49 4.34 ID Fluence ( ~ 1 0 ~ ~ )
WCAP-16669-NP 53 EFPY 0.36 4.90 4.55 0.36 3.1 9 3.1 9 4.43 3.05 0.50 5.05 4.90 0.50 4.65 Component Description Unit 1 Nozzle Belt Forging Intermediate Shell Plate Lower Shell Plate Nozzle Belt to Intermed.
Shell Circ Weld (1 00%)
lntermediate Shell Long Seam (ID 27%)
lntermediate Shell Long Seam (OD 73%)
Intermed. to Lower Shell Circ. Weld (1 00%)
Lower Shell Long Seam (1 00%)
Unit 2 Nozzle Belt Forging lntermediate Shell Forging Lower Shell Forging Nozzle Belt to Intermed.
Shell Circ Weld (1 00%)
Intermed. to Lower Shell Circ. Weld (1 00%)
Heat or HeatlLot 122P237 A981 1-1 C1423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 123V352 123V500 122W195 21 935 72442 (SA-l 484)
Reactor vessel total EFPY values were determined in December 2012 to be 33.35 EFPY for Unit 1 and 32.8 EFPY for Unit 2. Assuming the units operate 95% of the time in the future (5% for future refueling outages), the period of applicability for the PTLR curves would be:
(50 EFPY - 33.35 EFPY) / 0.95 = 17.5 years This additional 17.5 years would make the PTLR effective through the end of 2029.
- 7. CONCLUSION The proposed RCS Pressure/Temperature (P/T) limits for PBNP Units 1 and 2 Technical Specifications are valid and conservative through 50 Effective Full Power Years (EFPY)
(approximately through 2029).
- 8. REFERENCES
- 1. WCAP-16669-NP, "Point Beach Unit 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation", dated December 2006
- 2. Framatome ANP Topical Report BAW-2308, Revision I
-A, "Initial RTNDT of Linde 80 Weld Materials," approved August 2005
- 3. Framatome ANP Topical Report BAW-2308, Revision 2-A, "Initial RTNDT of Linde 80 Weld Materials," approved March 2008
- 4. WCAP-16983-P, "Point Beach Units 1 and 2 Extended Power Uprate (EPU) Engineering Report", dated September 2009
- 5. NextEra Energy Point Beach, LLC letter to NRC, dated January, 15 2012, License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (MLI 301 6A028)
- 6. 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", dated May 2004 Page 5 of 5
ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
MARK-UP OF TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 16 pages follow
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT Note: Applicability limits for pressure temperature limits are discussed in Section 2.0, "Operating Limits."
RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
This RCS Pressure and Temperature Limits Report (PTLR) for Point Beach Nuclear Plant Units 1 and 2 has been prepared in accordance with the requirements of Technical Specification 5.6.5.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC; specifically those described in NRC Safety Evaluations dated October 6, 2000, July 23, 2001,&
October 18,2007, and XXXXXXX..
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto (Ref 5.19). Based upon fluence values in Westinghouse t q. w x k - H W I WCAP-16983-P (Ref 5.15), this PTLR is effective for a5;9 X E F P Y (approximately Jwx~W-4 2029). @+&q The Technical Specifications addressed in this report are listed below:
1.I 3.4.3 Pressurenemperature (P-T) Limits 1.2 3.4.1 2 Low Temperature Overpressure Protection (LTOP) System 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. Changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.5.
These limits have been determined such that applicable limits of the safety analysis are met. Items that appear in capitalized type are defined in Technical Specification 1.l, "Definitions."
2.1 RCS Pressure and Temperature Limits (LC0 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are:
- a. A maximum heatup rate of 100°F in any one hour.
- b. A maximum cooldown rate of 100°F in any one hour.
- c. An average temperature change of 510°F per hour during inservice leak and hydrostatic testing operations.
2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively. (Ref 5.2)
POINT BEACH TRM REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 2.1.3 The minimum temperature for pressurization or bolt up, using the methodology, is 60°F, which when corrected for possible instrument uncertainties is a minimum indicated RCS temperature of 78°F (as read on the RCS cold leg meter) or 70°F using the hand-held, digital pyrometer.
2.2 Low Temperature Overpressure Protection Svstem Enable Temperature (LC0 3.4.6, 3.4.7, 3.4.10 and 3.4.12)
The enable temperature for the Low Temperature Overpressure Protection System is 285°F (includes instrument uncertainty for RCS T, wide range). (Ref 5.4) 2.3 Low Temperature Overpressure Protection Svstem Setpoints (LC0 3.4.12)
Pressurizer Power-Operated Relief Valve Lift Setting Limits The limiting trip setpoint (Ref 5.26) for the pressurizer power-operated relief valves (PORVs) is 5420 psig (includes instrument uncertainty).
The following operating restrictions ensure continued operability of the LTOP system:
2.3.1 RCP Operating Restriction - No more than one RCP in operation for RCS temperature < I 80°F. (Ref 5.20 to 5.24) 2.3.2 Charging Pumps - Limit the number of operating charging pumps to two when LTOP is in service. (Ref 5.20 to 5.24) 2.4 Criticalitv and Hvdrostatic Leak Test Limits 2.4.1 Criticality and hydrostatic leak test limits are shown on the RCS Pressure Temperature Limits for heatup, Figure 1. (Ref 5.2) 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedules for Units 1 and 2 are provided in Tables 1 and 2, respectively.
For the period of the renewed facility operating license, all capsules in the reactor vessel that are removed and tested shall meet the test procedures and reporting requirements of ASTM E 185-82. Any changes to the capsule withdrawal schedule, including spare capsules, shall be approved by the NRC prior to implementation. (Ref 5.16 and 5.17)
POINT BEACH TRM REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT The pressure vessel surveillance program is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the nil-ductility temperature, RTNDT, which is determined in accordance with ASTM E208. The empirical relationship between FITNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E l 85-82.
Surveillance specimens for the limiting materials for the PBNP reactor vessels are not included in the plant specific surveillance program. Therefore, the results of the examinations of these specimens do not meet the credibility criteria of Regulatory Guide 1.99, Revision 2, for PBNP Units 1 and 2.
During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5.5)( Ref 5.25) 4.0 SUPPLEMENTAL DATA INFORMATION The limiting RTpTs values for the PBNP limiting beltline materials at 3523 3 EFPY are:
Unit 1 - lntermediate to Lower Shell Circ Weld = ?%M 230.7";
Lower Shell Axial Weld = W 220.0" (Ref. 7 5.2. Table 4-10)
Unit 2 - lntermediate to Lower Shell Circ Weld = 2954 265.2";
lntermediate Shell Forging =
150.6"
((Ref. 7 5
2 Table 4-10)
POINT BEACH TRM REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT
5.0 REFERENCES
5.1 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"
Revision 4, Mav 2004 5.2 W 4 X,
WCAP-16669, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," Revision 1,-
January 2009 5.3 WEPCO Calculation Addendum No. 98-0156-00-A, Revision 0, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61 782, Point Beach Unit 1," 9/22/1999 5.4 Westinghouse Letter V!EP 98-25, "TrwsmiHal cf LTCDI S II
,-h I A,,,% Qn Low Temperature Overpressure Protection Svstem
{LTOPS) Setpoint Analvsis. Januarv 2007 5.5 PWR Owner Group Topical Report BAW-1543(NP), Revision 4, Supplement 6-A, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. MC9608), June 2007 5.6 BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 5.7 CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 c I qnrl '3 ED1 I D T 5.8 V I U l l U L. L I -
I I
V Deleted 5.9 ASME B&PVC Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division 1 "
5.10 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Exemption from the Requirements of 10CFR50.60 (TAC NOS. MA9680 and MA9681)," dated October 6,2000
- 5. 11 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure Temperature Limits Report (TAC Nos. MA8459 and MA8460)," dated July 23,2001
- 5. 12 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: The Conversion to Improved Technical Specifications (TAC Nos. MA71 86 and MA71 87)," dated August 8,2001 5.13 Deleted POINT BEACH TRM REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 5.14 NRC SE "Amendment Nos. 229/234 to Facility Operating Licenses DPR-24 and DPR-27, (approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)," dated October 18,2007
'IS y y
h L;;=G 2
'3 I=
r l D 11 Uprate (EPU) Enqineerinq Report 5.16 Renewed Facility Operating License DPR-24, Point Beach Nuclear Plant Unit 1 5.17 Renewed Facility Operating License DPR-27, Point Beach Nuclear Plant Unit 2 5.1 8 Deleted 5.1 9 Root Cause Evaluation 01092944, "Apparent Non-compliance with TS 5.6.5.c,"
Corrective Action to Prevent Recurrence (CATPR) 2 Root Cause (RC)2.
5.20 CL 4C, Low Temperature Overpressurization Protection Unit 1 5.21 CL 4C, Low Temperature Overpressurization Protection Unit 2 5.22 OP 3C, Hot Standby to Cold Shutdown 5.23 OP 48, Reactor Coolant Pump Operation 5.24 OP 1A, Cold Shutdown to Hot Standby 5.25 NextEra Point Beach Letter, "Reactor Vessel Surveillance Program Request to Change Reactor Vessel Surveillance Specimen Withdrawal Schedule," dated January 19,2010 5.26 Point Beach Nuclear Plan Design Guide DG-101, Instrument Setpoint Methodology POINT BEACH TRM REV. 8 MARK-UP
OINT BEACH NUCLEAR PLANT ECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT REPLACE FIGURE WITH FIGURE 5-7 OF WCAP-16669-NP Figure 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP Leak Test Limit \\
h m
m Q
V with RCP restrict~on Moderator Temperature ("F)
POINT BEACH TRM REV. 8 MARK-UP
MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 53 EFPY (Hafnium Removal):
1/4T, 220.0°F 3/4T, 184.6"F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-7 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIJ WCAP-16669-NP January 2009 Revision 1
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT REPLACE FIGURE WITH FIGURE 5-8 OF WCAP-16669-NP Figure 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN h
m LTOP Enable Temperature = 285OF Maximum LTOP Setpoint = 420 psig with RCP restriction Moderator Temperature ("F)
POINT BEACH TRM 2.2 - 7 REV. 8 MARK-UP
MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 53 EFPY (Hafnium Removal):
1/4T, 220.0°F 3/4T, 184.6"F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-8 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KI,)
WCAP-16669-NP January 2009 Revision 1
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 TABLE 1 (**)
POINT BEACH NUCLEAR PLANT UNlT 1 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.
Capsule Identification Letter V
S R
T P
N
- During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master lntegrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5. q Ref 5.25)
Approximate Removal Date*
September 1972 (actual)
December 1975 (actual)
October 1977 (actual)
March 1984 (actual)
April 1994 (actual)
Standby TABLE 2 (**)
POINT BEACH NUCLEAR PLANT UNlT 2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.
Capsule Identification Letter V
T R
S P
N A
- During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master lntegrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5.5)( Ref 5.25).
Approximate Removal Date*
November 1974 (actual)
March 1977 (actual)
April 1979 (actual)
October 1990 (actual)
June 1997 (actual)
Standby April 2022 POINT BEACH TRM REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT I RPV BELTLINE 50 EFPY VALUES(~)
Based on Westinghouse Report WXPS9X,WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2). b!&&MAe Vessel Manufacturer:
1 Babcock & Wilcox Plate and Weld Thickness (without cladding):
1 6.5", without clad '"
Footnotes:
Limiting material 3&%FwLtW zwMk~60
~lr,nr@
W 037 U 4 QS9 QA@
048
(
From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = fsufl x e-0.24X, where fsud is expressed in units of El 9 n/cm2, E>1 MeV, and x is the desire:
depth in inches into the vessel wall. For :
P Z
k W
- EFPY, K"\\ +-n~rv----
-,, 8 EEDV ZT FiueAce
( ~ 1 9 nlcm2) (')
04180.11 4451.52 W1.41 wm NIA Wo.99 w1.38
-0.95 fD)
Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 Component Description Nozzle Belt Forging Intermediate Shell Plate Lower Shell Plate Nozzle Belt to Intermed. Shell Circ Weld (100%)
lntermediate Shell Long Seam (ID 27%)
lntermediate Shell Long 'A' Seam (OD 73%)
Intermed. to Lower Shell Circ.
Weld (1 00%)
Lower Shell Long Seam 'A' (1 00%)
EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. y
?h2 3W-EFPY
%.a EFPY-*
114T Fluence (El 9 n/crn2) ('I W0.24
-3.32 29363.09 43470.24 4432.16 NIA
-3.00 4 4 4 z POINT BEACH TRM 2.2 - 9 REV. 8 MARK-UP 3523 EF?V -
F a 6. t ~ ~ ~
0253 4422 4420 QS3 4 4 4
%4 W
Heat or HeatRot 1 22P237 A981 1-1 (31423-1 8T1762 (SA-1426) 1 PO81 5 (SA-812) 1 PO661 (SA-775) 71 249 (SA-l 01 )
61 782 (SA-847)
Inside Surface Fluence (El 9 nlcrn2) 4x2s-338-3~244.55 CL250.36
-3.19 Mu 3a54.43 a333.05
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BELTLINE zm EFPY VALUES(~)
Based on Westinghouse Report kVWWsV6,WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2).
Vessel Manufacturer:
Plate and Weld Thickness (without cladding):
(
From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = f,,~ x e'0.24x, where fsurt is expressed in units of E l 9 n/cm2, Esl MeV, and x is the desired depth in in2hes into the vessel wall.
EFPY, =$A EM\\ f - n q*
Babcock & Wilcox and Combustion Engineering 6.5", without clad "'
(DJ Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970 EDV CEf 3MShawe wf6) 044 4434 4434-8;44 m
" EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.
C! EF?Y Footnotes:
Limiting Material
-CEf 314T Fluence (EI 9 n/cm2) (')
Wo.16 M1.57 44321.52 w0.16 w1.44 POINT BEACH TRM FEnV
.Imr
-a O m 422 4 2 2 om 4433 REV. 8 MARK-UP FCnV el (El 9 n/cm2) (')
G239.34 2293.42 2233.32 G230.34 2423.15 Component Description Nozzle Belt Forging Intermediate Shell Forging 'A' Lower Shell Forging Nozzle Belt to Intermed. Shell Circ Weld (1 00%)
Intermed. to Lower Shell Circ Weld (1 00%) 'A)
Heat or HeatILot 123V352 123V500 122W195 21 935 72442 (SA-1484)
Inside Surface Fluence (El 9 n/cm2) 0340.50 3485.05 33Q4.90 0340.50 m4.65
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 1/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT a
50 EFPY.(~)
Footnotee
-pp w, See Table 3
'" Credible Surveillance Data; see BAW-2325 for evaluation.
("
Non-credible surveillance data: see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure ARTNDT and predicted ARTNDT based on Table CF is less than 20 (56°F).
")
Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1,"
(Ref.5.3) utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.
(E)
Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTN~T
+ ARTNDT + Margin, where ARTNDT = Chemistry Factor x Fluence Factor, and Margin = 2(0: +
with cil defined as the standard deviation of the Initial RTNDT and 0, defined as the standard deviation of ARTNDT.
0
(' T&
Wilcox, September 1969.
(G) Deleted.
(*) EFPY value listed here is based on various reactor fuel management strategies and reactor power l e
v e
l s
i
?!?s 364 KEDV
- a POINT BEACH TRM 2.2 - 11 REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV i/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT M
5q EFPY.(')
1 3 K "nr, Unless otherwise noted, all ART input data obtained from I,
-,WCAP-16669.
Revision 1, 'Point B e
a c
i i
~
d 2
H H
C) EDI I
- L. w Component Description Intermediate Shell Forging Vessel Manufacturer:
Plate and Weld Thickness (without cladding):
Footnotes:
7a, See Table 4
(
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNDr and predicted ARTNDT based on Table CF is less than 20 (34°F)
Credible surveillance data; see BAW-2325 for evaluation.
Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore, conservative.
Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ART~JDT
+ Margin, where ARTNDT = Chemistry Factor x Fluence Factor, and Margin = 2(012 + 0t)0.5, with q defined as the standard deviation of the Initial FITNDT, and o~ defined as the standard deviation of ARTNDT.
APT - nn
?,I - i ~ n ~ c r
O
(
~
a i
R c
t o
~
V
~
d h
Nuclear Deleted.
Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 (Ref.5.7).
(IJ EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.
Babcock & Wilcox and Combustion Engineering 6.5", without cladfF' POINT BEACH TRM 7
REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 354 50 EFPY 'H' Footnotes:
See Table 3.
Credible Surveillance Data; see BAW-2325 for evaluation.
'Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured ARTNDT are predicted ARTNDT based on Table CF is less than 20 (56°F).
lD)
Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1,"
utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.
fE)
Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTND, = Chemistry Factor x Fluence Factor, and Margin = 2(0i2 + 0:)O5, with 01 defined as the standard deviation of the Initial RTNDT, and o,,
defined as the standard deviation of ARTNDT.
A DT - +n 1 7 7
, CIA - I.I 9.c r o.
m m
a
?
P l
~
u r
d
~
e e
l m
fG)
Deleted.
")
EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. @-c v?
5.2) i%&w6sb C? EFPV values. 3 3
3%
ccnv
-.. m POINT BEACH TRM 2.2 - 13 REV. 8 MARK-UP
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 359 5q EFPY (')
Unless otherwise noted, all ART input data obtained from
,I mWGAWWX,WCAP-l 666gL Revision 1, 'Point B e
a c
h h
Vessel Manufacturer:
( Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding):
1 6.5", without clad")
Footnotes:
See Table 4.
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNDT and predicted ARTNDT based on Table CF is less than 20 (56°F).
Credible surveillance data; see BAW-2325 for evaluation.
Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using su~eillance data, and therefore, conservative.
Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTNDT = Chemistry Factor x Fluence Factor, and Margin = 2(1$
+
with q defined as the standard deviation of the Initial RTNDT, and o~ defined as the standard deviation of ARTNDT. D 0
I Combustion Engineering, CE Book #4869, October 1970.
Deleted.
Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.
cf EFPY POINT BEACH TRM REV. 8 MARK-UP
ENCLOSURE 3 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
WCAP-16669-NP POINT BEACH UNITS 1 AND 2 HEATUP AND COOLDOWN LIMIT CURVEES FOR NORMAL OPERATION 59 pages follow
WCAP-16669-NP Revision 0 Westinghouse Non-Proprietary Class 3 December ZOO6 Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation
B.N. Burgos*
December 2006 Approved:
Elecsronicallv A ~ ~ r o v e d
- J.S. Carlson, Manager Primary Component and Asset Management
- Electronically approved records are authenticated in the Electronic Document Management System.
Westinghouse Electric Company LLC Energy Systems P.O. Box355 Pittsburgh, PA 15230-0355 02006 Westinghouse Electric Company LLC All Rights Reserved
This report has been technically reviewed and verified by:
Frank C. Gift: Elecrronicallv Auvroved*
- Electronically approved records are authenticated in the Electronic Document Management System.
Revision 0:
Original Issue IRIECORD OF REVISION i
WCAP-16669-NP ii I
PREFACE I
1 I I I
I I
i I
I TABLE OF CONTENTS LIST OF TABLES..
iv LIST OF FIGURES................................................
......vi EXECUTIVE
SUMMARY
.vii 1
INTRODUCTION 1
2 FRACTURE TOUGHNESS PROPERTIES
- 3 3
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERA-RELATIONSHIPS
.6 4
CALCULATION OF ADJUSTED REFERENCE TEMPERATURE
.. 10 5
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES
.22 6
REFERENCES......................
.48 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KI~)
WCAP-16669-NP iv LIST OF TABLES Table i -1 Table 2-1 Table 2-2 Table 2-3 Table 4-1 Table 4-2 Table 4-4 Table 4-4 Table 4-5 Table 4-6 Table 4-7 Table 4-8 Table 4-9 Plant Operating Conditions Reflecting Development of PT Limit Curves for 43 and 53 EFPY.................................
2 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Point Beach Units 1 and 2 Reactor Vessel Materials................................................... 4 Summary of the Initial RTNDT Values for the Point Beach Units 1 and 2 Closure Head and Vessel Flanges......................
5 Summary of the Point Beach Units 1 and 2 Reactor Vessel Beltline Material Chemistry Factors.,.................,,,,,,,.,,,.,,.,,,,,,..,,,....,,.....,,.....,......................,......,....,......,...5 Fluence (E>I.O MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafi-iium rods, through 53 EFPY.......................................... 12 Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hahiurn rods, through 43 EFPY..,........................,....,,......,. 13 Fluence (Dl.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel BeltIine Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 5 3 EFPY.................................................,..,,,.,...,,,.,.,,.,,..,,.,...,.,..,..,....., -14 Fluence (Dl.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY................................................................................................ 15 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY... 16 Adjusted Reference Temperature Evaluation for the Point Beach Unit I and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY.., 17 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY....................................................... 18 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY.....................,..,..,....,,,,....... 19 Summary of the Limiting ART Values (with Hafnium) Used in the Generation of the Point Beach Units 1 and 2 Heatup/Cooldown Curves...........................
2 1
LIST OF TABLES - (continued)
Table 4-10 Table 5-1 Table 5-2 Table 5-3 Table 5-4 Table 5-5 Table 5-6 Table 5-7 Table 5-8 Table A-1 Table A-2 Summary of the Limiting ART Values (without Hafnium) Used in the Generation of the Point Beach Units 1 and 2 HeatupICooldown Curves
.................. 21 43 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Hafnium, w/Klc, wFlange Notch and wlo Uncertainties for Instrumentation Errors)..................... 26 43 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (wmafnium, w/KIc, w/Flange Notch and wlo Uncertainties for Instrumentation Errors) 28 53 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Hafnium, WE,,
w/Flange Notch and wlo Uncertainties for Instrumentation Errors) 32 53 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (wIHafaium, w/Klc, w/Flange Notch and wlo Uncertainties for Instrumentation Errors) 34 43 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (wmafnium Removal, w/Klc, w/Flange Notch and w/o Uncertainties for Instrumentation Errors)..... 3 8 43 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Hafnium Removal, w/Klc, w/Flange Notch and w/o Uncertainties for Instrumentation EJTO~S) 40 53 EPPY Iieatup Curve Data Points Using 1998 App. G Methodology (wmafnium Removal, w/KIc, w/Flange Notch and wlo Uncertainties for Instrumentation Errors) 44 53 EFPY Cooldown Curve Data Points Using 1998 App, G Methodology (wmafnium Removal, w/KTC, w/Flange Notch and wlo Uncertainties for Instrumentation Errors) 46 K,,
Values for 100°F/hr Heatup Curve (w/o Margins for Instrument Errors)................. A-2 K,, Values for 1 OO°F/hr Cooldown Curve (wlo Margins for Instrument Errors)
A.3
1 i
WCAP-16669-NP vi LIST OF FIGURES Figure 5-1 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and I OO°F/hr) Applicable for 43 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wIK1,)
24 Figure 5-2 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 43 EFPY (with Hahiurn and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (we,)
25 Figure 5-3 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 1 OO°F/hr) Applicable for 53 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIc) 30 Figure 5-4 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°Fkr) Applicable for 53 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App, G Methodology (WAG,)......................................
3 1 Figure 5-5 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 43 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/K,,)
36 Figure 5-6 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown 1
Rates up to 1 OOOI;mt) Applicable for 43 EFPY (with Hafhium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIo).................. 37 Figure 5-7 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of I
60 and 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins I
for Instrumentation Errors) Using 1998 App. G Methodology (w/KI,).............................
42 I
Figure 5-8 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown I
Rates up to 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wK1,).................. 43
WCAP-16669-NP vii EXECUTIVE
SUMMARY
This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Point Beach Units 1 and 2 reactor vessels. The PT curves were generated based on the latest available reactor vessel information and updated calculated fluences.
The new Point Beach Unit 1 and 2 heatup and cooldown pressure-temperature limit curves were generated using adjusted reference temperature (ART) values that bound both units. The highest ART values from the two units were from the Unit 2 intermediate to lower shell girth weld, however the limiting materials are actually the intermediate shell axial welds from Unit 1, depending on the vessel thickness [% thickness (114T) or % thickness (314T) locations]. The axial welds become limiting over the girth weld through use of "circ-flaw" methodology from ASME Code Case N-588. This methodology is less restrictive than the standard "axial-flaw" methodology from the 1998 ASME Code,Section XI through 2000 Addenda. In addition to the use of Code Case N-588, the PT curves also made use of the KI, methodology detailed in ASME Code Case N-640. Bath ASME Code Case N-588 and N-640 were joined together under ASME Code Case N-641 and incorporated into the 1998 ASME Code,Section XI, through 2000 Addenda.
The PT limit curves were generated for 43 and 53 EFPY using heatup rates of 60 and 100°F/hr, cooldown rates of 0,20,40,60 and 100°F/hr, with uprated conditions for both units, and with/without hafnium removal in 2008. The curves were developed without margins for instrumentation errors. These curves can be found in Figures 5-1 through 5-8.
1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTND~,
and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least SO ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.
RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (lRTmT). The extent of the shift in RTNW is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" [Reference
- 11. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNnT
+ ARTNDT + margins for uncertainties) at the surface, 1/4T and 314T locations, where T is the thickness of the vessel at the beltline region measured fiom the cladhase metal interface.
The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 4 [Reference 21, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the "axial flaw" and "circ flaw" methodologies of the 1998 ASME Code,Section XI through the 2000 Addenda Beference 31 was used, which makes use of the KI, methodology.
The ART values were obtained from AREVA calculation 32-9019240-000, "ART Values for Point Beach Units 1 and 2" [Reference 41. This document makes use of the NRC approved Topical Report BAW-2308, Revision I-A, "Initial RTNDT of Linde 80 Weld Materials" [Reference 51. The fluence projections are based on the Ferret Code best-estimate values verified by Westinghouse in letter WEP 13, dated February 14,2006 [Reference 61 as approved for use by the NRC in WCAP-16083-NP-A, Revision 0 [Reference 71.
The purpose of this report is to present the calculations and the development of the Point Beach Units 1 and 2 heatup and cooldown curves for 43 and 53 EFPY for plant conditions that reflect a power uprate and withfwithout removal of hahiurn rods. The four cases analyzed for each unit are shown in Table 1-1.
This report documents the calculated ART values and the development of the PT limit curves for normal operation. The PT curves herein were generated without instrumentation errors. The PT curves include a hydrostatic leak test limit curve from 2485 psig to 2000 psig, along with the pressure-temperature limits for the vessel flange region per the requirements of 10 CFR Part 50, Appendix G [Reference 81,
TABLE 1-1 Plant Operating Conditions Reflecting Development of PT Limit Curves for 43 and 53 EFPY L
Case 1
2 3
4 unit I
2 1
2 1
2 1
2 Power w t h ). I<::-;,-
151 8.5 startup to 2/3/2003 1540.0 2/3/2003 to 10/2008 1678.0 10/2008 to 53 EFPY 1518.5 startup to 2/3/2003 1540.0 2/3/2003 to 412008 1678.0 412008 to 53 EFPY 1518.5 startup to 2/3/2003 1540.0 2/3/2003 to 1012008 1678.0 1012008 to 43 EFPY 15 18.5 startup to 2/3/2003 1540.0 2/3/2003 to 4/2008 1678.0 412008 to 43 EFPY 15 18.5 startup to 2/3/2003 1540.0 2/3/2003 to 10/2008 1678.0 10/2008 to 53 EFPY 1518.5 startup to 2/3/2003 1540.0 2/3/2003 to 4/2008 1678.0 412008 to 53 EFPY 151 8.5 startup to 2/3/2003 1540.0 2/3/2003 to 10/2008 1678.0 1012008 to 43 EFPY 15 18.5 startup to 2/3/2003 1540.0 2/3/2003 to 4/2008 1678.0 412008 to 43 EFPY
...EFPYI 53 53 43 43 53 53 43 43
. :.. ~ ~ f ~ i ~ ~ ! ~ ~ & i Yes Yes Yes Yes Removal October 2008 Removal April 2008 Removal October 2008 Removal April 2008
2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan [Reference 91. The beltline material properties of the Point Beach Units 1 and 2 reactor vessel are presented in Table 2-1. The unirradiated RTNm values for the closure head and vessel flange are documented ill Table 2-2.
The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 4 [Reference 21.
The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.I and 2.1. Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents, which are presented in Table 2-1. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The determination of CFs using Positions 1.1 and 2.1 are documented in Reference 4. Table 2-3 summarizes the Positions 1.1 and 2.1 CFs determined for the Point Beach Units 1 and 2 beltline materials.
It should be noted that in the calculations of Position 2.1 chemistry factors in Table 2-3, the ratio procedure described in Reference I was applied to account for chemistry differences between the vessel weld material and the surveillance weld material [Reference 41. No temperature adjustments are required for the Point Beach Units 1 and 2 data since it is being applied to their own plants.
TABLE 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial R T N ~ ~
Values for the Point Beach Units 1 and 2 Reactor Vessel Materials [Reference 41 Material Description..!.'.,
.l.
- ..it_....:.
Reactor Vebel.
5..,
Belt,jae Reg,on L*&iion
-;;i~iti&,;
.I.._.
.,?k =::.:...,
- i;
- ~.'RT;
- :..::L~'
rnT'".
- r.
- . li-T
.7L:,:.:!+-:
rz=:lr~,
A. -
- l..
- ;, 'i->i'::;:;
=:<
.,, ;>-:c,ke,&i&l,,,
j
=.-. Co.dpos4iOn j,.
. _ I 2.::
pd.i~t Beac,&.,Uuit;.l,,
- ~
- f.2:::;!:-c,
,:':;.,:, 1 ;;:.'.,:,.-;:j::-!..;'
,.7iL~..
. :Matl
- . $dent*.
..,:,: '.:yla.
- .: NtYo.;
Nozzle Belt Forging (NB)
Intermediate Shell Plate (IS)
Lower Shell Plate (LS)
NB to IS Circ. Weld (100%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
,.. Ni ::, : :
.. y+y..;
....... :=.:..-..;
i t
......::. : :.... 1: -.
122P237 12213237 SA-508 C1.2 0.1 1 0.82 50QF A9811-1 A981 1-1 SA-508 C1.2 0.20 0.06 1°F C1423-1 C1423-1 SA-508 Cl. 2 0.12 0.07 1 OF SA-1426 8T1762 Linde 80 Flux 0.19 0.57
-47'6°F SA-8 12 1 PO8 15 Linde 80 Flux 0.17 0.52
-47.6"F SA-775 1P0661 Linde 80 Flux 0.17 0.64
-47.6"F SA-1101 71249 Linde 80 Flux 0.23 0.59
-47.4OF SA-847 61782 Linde 80 Flux 0.23 0.52
-47.6"F
.:\\-.
..l-;.:.:
N,, mb&:
, :::y.::,:;;.
, :..... m.
e,,..
).',.
- 7:.:.=
'7&.7
'.,,.S
- +:-
-~oini EeacbUnit 2..
Nozzle Belt Forging (NB)
Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ, Weld (100%)
2.:
SA-508 C1. 2 SA-508 C1.2 SA-508C1.2 Linde 1092 Flux Linde 80 Flux 123V352 123V500 123W195 21935 SA-1484 123V352 123VSOO 123W195 21935 72442 0.1 1 0.09 0.05 0.18 0.26 0.73 0.70 0.72 0.70 0.60
=27-... ::I' 40°F 40°F 40°F
-56°F
-30°F
TABLE 2-2 Summary of the Initial RTNDT Values for the Point Beach Units 1 and 2 Closure Head and Vessel Flanges Notes:
(a)
Certified Material Test Report Initial RTNIJT values documented in References 12 and 13 for the replacement reactor vessel closure heads.
(b)
From WCAP-1 5 12 1, Revision 1 [Reference 101 Material Identiflcikion Initi&l:R~ihT (OF>:
Po'int Beach..Unit 1 TABLE 2-3 Summary of the Point Beach Units 1 and 2 Reactor Vessel Beltline Material Chemistry Factors
[Reference 41 Closure Head Flange Vessel Flange
-68'")
48@'
,; ~oint.~ea(h Unit.2,.
Beltline Materials Nozzle Belt Forging (NJ3)
Intermediate Shell (IS)
Lower Shell (LS)
NB to IS Circ. Weld (100%)
IS Long. Weld (ID 27%)
lntermediate to LS Circ. Weld (100%)
Nozzle Belt Forging (NB)
Closure Head Flange Vessel Flange Cg'.....
wt%?,
0.11 0.20 0.12 0.19 0.17 0.17 0.23 0.23 0.11
-60'")
6 0 ~ '
~
'. unit:"
1 1
1 1
1 1
1 1
2
.;.Material
.. -1dint.
122P237 A981l-1 (21423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 123V352
- ..... :.NF,:.'
. i
.$$"/d?.
- I 0.82 0.06 0.07 0,57 0.52 0.64 0.59 0.52 0.73 Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ. Weld (100%)
(a) Determined from surveillance data
'!{~fi~d$~i~:"
j'
,,Fictvr.' ::,
77.0 79.3(@
35.8'a) 167.0 167.0 167,O 167.6 167.0 76.0 2
2 2
2 58.0 42.8(a) 170.5 180.0 Notes; 123V500 123W195 21935 SA-1484 0.09 0.05 0.18 0.26 0.70 0.72 0.70 0.60
3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3 1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kr, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KI,, for the metal temperature at that time. KI, is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Reference 31.
The KIc curve is given by the following equation:
- where, KI,
=
reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K,, curve is based on the lower bound of static critical Kl values measured as a function of temperature on specimens of SA-533 Grade B Class-1, SA-508-1, SA-508-2, SA-508-3 steel, 3.2 METHODOLOGY FOR PRESSURE-TEMPERATTJRE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
- where, KI,),
=
stress intensity factor caused by membrane (pressure) stress stress intensity factor caused by the thermal gradients Krc K,, =
function of temperature relative to the RTN~T of the material C
=
2.0 for Level A and Level B service limits C
=
1.5 for hydrostatic and leak test conditions during which the reactor core is not critical
For membrane tension, the corresponding KI for the postulated defect is:
where, M, for an inside surface flaw is given by:
M,
=
0.926 for 2 1 4 5 3.464, M,,
=
3.21 for & > 3.464 Similarly, Mm for an outside surface flaw is given by:
M,
=
1.77 for fi < 2, M,
=
3.09 for fi > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.
For bending stress, the corresponding KI for the postulated defect is:
Klh = Mb
- Maximum Stress, where Mb is two-thirds of M, The maximum KI produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Ka = 0.953~10'~
x CR x I?.',
where CR is the cooldown rate in OFhr., or for a postulated outside surface defect, KI, = 0.753~10" x HU x ?.', where HU is the heatup rate in 'Fk.
The through-wall temperature difference associated with the maximum thermal K1 can be determined from Fig. G-22 14-1. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal ISI.
(a)
The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).
(b)
Alternatively, the KI for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a %-thickness inside surface defect using the relationship:
or similarly, KIT during heatup for a %-thickness outside surface defect using the relationship:
where the coefficients C,, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
and x is a variable that represents the radial distance fiom the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that Equations 3,4 and 5 were implemented in the OPERLM computer code, which is the program used to generate the pressure-temperature (PT) limit curves. NO other changes were made to the OPERLIM computer code with regard to PT calculation methodology. Therefore, the PT curve methodology is unchanged f?om that described in WCAP-14040-NP-A, Revision 4 "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"
[Reference 21 Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.
At any time during the heatup or cooldown transient, KI, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting fiom the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K,,, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of K,,
at the 114T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds ISlt, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 114T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the IS,, for the 114T crack during heatup is lower than the KI, for the 114T crack during steady-state conditions at the same coolant temperature, During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIc values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 114T flaw located at the 114T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken fiom the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches fiom the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
3.3 CLOSURE HEAD/VESSEL FLANGE REQUIREmNTS 10 CFR Part 50, Appendix G [Reference 81 addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNm by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3 107 psig for Point Beach Unit 1 and 3125 psig for Point Beach Unit 2), which is calculated to be 621 psig (the more limiting value) for both Units 1 and 2.
The limiting unirradiated R b T of 60aF occurs in the vessel flange of the Point Beach Unit 2 reactor vessel, so the minimum allowable temperature of this region is 180°F at pressures greater than 621 psig (without instrument uncertainties). This limit is shown in Figures 5-1 through 5-8 wherever applicable.
4 CALCULATION OF ADJUSTED REFElRlENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = Initial RTNDr + ARTNor + Margin (7)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section DI of the ASME Boiler and Pressure Vessel Code Deference 91. If measured values of initial RTND.r for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTmT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
To calculate ARTNDT at any depth (e.g., at 114T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
where x inches (vessel beltline thickness is 6.5 inches) is the depth into the vessel wall measured from the vessel cladfbase metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTmT at the specific depth.
The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections and the results are presented in Tables 4-1 through 4-4 [Reference 1 I]. The Ferret Code best estimate values used from Reference 11 are an exception to the methods presented in WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Meatup and Cooldown Limit Curves" Deference 21. Their use is permitted based on NRC approval documented in WCAP-16083-NP-A, Revision 0 [Reference 71 and has been verified by Westinghouse in letter WEP-06-13 [Reference 61. Tables 4-1 through 4-4 also provide a summary of the vessel fluence projections at the 1/4T and 3/4T locations. Tables 4-5 and 4-8 contain the 1/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the 43 and 53 EFPY ART values for all beltline materials in the Point Beach Units 1 and 2 reactor vessels.
Margin is calculated as, M = 2,/=
+ The standard deviation for the initial RTmT margin term (09 is O°F when the initial RTNI~I.
is a measured value and 17'F when a generic value is available. The standard deviation for the ARTmT margin term, oh, is 17OF for plates or forgings, and 8.5"F for plates or forgings when credible surveillance data is used. For welds, a* is equal to 28OF when surveillance capsule data is not used, and is 14'F (half the value) when credible surveillance capsule data is used. c r ~
need not exceed 0.5 times the mean value of ARTNDT.
Table 4-1 Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY [Reference 41 Beltline Materials
~ a t & i ~ l
- Ideat, 122P237 A9811-1 C1423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 123V352 123V500 123W195 21935 SA-1484 Unit.
53,,EmY,Fi$~
ce:,I &ir&$:,:j
$??; :,!,::
Nozzle Belt Forging (NB)
Intermediate Shell (IS)
Lower Shell (LS)
NB to IS Circ. Weld (100%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
Nozzle Belt Forging (NB)
Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ. Weld (100%)
Inner wetted::: :::
' surface 2,84E+18 4.86E+19 3.70E+19 2.84E+18 3.10E+19 3.10E+19 3.71E+19 2.60E+19 4.12E+18 4.66E+19 4.15E+19 4.12E+18 3.75E+19 1
1 1
I I
1 1
1 2
2 2
2 2
. :,.., : 1 1 4 ~
.'I:
,,.(et.625);;
1.92E+18 3,29E+19 2.51E+19 1.92E+18 2.10E+19 NIA 2.51E+19 1.76E+19 2.79E+18 3.16E+19 2.8lE+19 2.79E+l8 2,54E+19
.:; ;;$jj,y;;:.::?;::
- , 'Li<$ti,&sj;t
- ..,i&;875j;:::
8.81E+17 1.51E.t-19 1.15E+19 8.81E+17 NIA 9.62E+18 1.15E.tl9 8.07B+18 1.28E+18 1.44E+19 1.29E+19 1.28E+18 1.16E+19
Table 4-2 Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Reference 41
. i..
Beltline ~ ~ t & i i [ ~
,, unit:
Material'
. ' '.. Ident;. '
Nozzle Belt Forging (NJ3)
Intermediate Shell (IS)
Lower Shell (LS)
NB to IS Circ. Weld (1 00%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
Nozzle Belt Forging (NB)
Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ. Weld (100%)
122P237 A9811-1 (21423-1 SA-1426 SA-812 SA-775 SA-I 101 SA-847 123V352 123V500 123W195 21935 SA-1484 1
1 1
1 1
1 1
1 2
2 2
2 2
7.39E4-17 lm231E+19 9.65E+18 7.39E+17 NIA 7.82E+18 9.62Et18 6.67E+18 1.06E+18 1.18E+19 1.07E+19 1.06E+18 9.75B+18 2.38E+18 3.95E+19 3.11E+19 2,38E+18 2.52E+19 2.52E+19 3.10Et19 2.158+19 3.42E-t18 3.81E+19 3.45E+19 3,42E+18 3.14E+19 1.61E+18 2,67E+19 2.11E+19 1.61E+18 1.71E-l-19 NIA 2.10E+19 1.46E+19 2.32E+18 2.58E-tl9 2.34E-1-19 2.32E+18 2.13E+19 1
,(
~ ~ ~ : ~, ~ ~ ~ 3 ~ 4 ~ : j ~ ~ c ~ {
~ ~ : ~ ; : I ~ ~ c ~ @ $ ~ :,,
-7. \\:.:.,O;~;*/;.. :
<~',.{~,i8~5)~j'=,
-:..=. " "
- -._,.:==
..i7.-r -;.."
,'S:.?
xnn&,;i:i,
- .'.wad
.. f
?
. /
L..
,':. ';;.,::i/4.~::$.,!~>;:
., ',,:::I~.&~M&~:-~
.. -..., A,..:;:.f
':.' G&(&5);i;s
.( -...
Table 4-3 Fluence (D1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY [Reference 41
~[~.'j&J.y:z1~~~~~,
nl~~~.j.:I_'
- \\-:;-
A _ =..... _..:::
,, n : :, :
- i14T
,:. :.::..; 2/4~.:::;,;
Material
, wettid;:..
Lqcatibn
'L$&id9.'-:i
~eltline.~aterials' Unit
,%lent.:.
- .:~urh<e--:.:
1:..:(x=i.625)
- (&4;83~)l$
L 7.,.
.i Nozzle Belt Forging (NB) 1 122P237 3.58Ec18 2.42E+18 1.llE-l-18 Intermediate Shell (IS) 1 A9811-1 4.90E+19 3.32B+19 1.52E+19 Lower Shell (LS) 1 C1423-1 4.55E+19 3.08E+19 1.41E+19 -
NB to IS Circ. Weld (100%)
1 SA-1426 3.58E+18 2.42E-t.18 l.llE+18 IS Long. Weld (ID 27%)
1 SA-812 3.19E+19 2.16E+19 N/A IS Long. Weld (OD 73%)
1 SA-775 3,19E+19 N/A 9.90E+18 Intermediate to LS Circ. Weld (100%)
1 SA-1101 4.43E+19 3.00E.tl9 1.38E-kl9 LS Long. Weld (1 00%)
1 SA-847 3.0SE+19 2.07E+19 9.47E+18 Nozzle Belt Forging (NB) 2 123V352 S.O4E+18 3.41E+18 1.56E+18 Intermediate Shell Forging (IS) 2 123V500 5.05E+19 3.42E+19 1.57E+19 Lower Shell Forging (LS) 2 123W195 4.90E+19 3.32E+19 1.52E+19 NB to IS Circ. Weld (100%)
2 21935 5.04E+18 3.41E-kl8 1.$6E+18 Intermediate to LS Circ. Weld (100%)
2 SA-1484 4.65E+19 3.15E+19 1.44E+19
Table 4-4 Bluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EPPY [Reference 4)
Beltline ~ i f e f j i l ~.
Nozzle Belt Forging (NI3)
Intermediate Shell (IS) 1 A9811-1 3,97E+19 2.69E+19 1,23E+19 Lower Shell (LS) 1 C1423-1 3.59B+19 2.43E+19 1.1 lE+19 NI3 to IS Circ. Weld (100%)
1 SA-1426 2.80E+18 1.90E+18 8.69E+17 IS Long. Weld (ID 27%)
1 SA-812 2.57E+19 1,74E+19 NIA IS Long. Weld (OD 73%)
1 SA-775 2.57E3-19 NIA 7.98E+18 Intermediate to LS Circ. Weld (100%)
1 SA-1101 3.51E-t.19 2.38E+19 1.09Et19 LS Long. Weld (1 00%)
1 SA-847 2.42B+19 1.64E+19 7.51E+18 Nozzle Belt Forging (NB) 2 123V352 3.95E+lS 2,67E+18 1.23E+18 Intermediate Shell Forging (IS) 2 123V500 4.04E+19 2.74E+19 1.25E+19 Lower Shell Forging (LS) 2 123W195 3.88E+19 2.63E+19 1.20E+19 NB to IS Circ. Weld (100%)
2 21935 3.95E+l8 2.67E+18 1.23E+18 Intermediate to LS Circ. Weld (100%)
2 SA-1484 3.67E+19 2.49E+19 1.14E+19 43:@,~py'~m"@h&
,dchz.
, 1;. (,::,: :. :!.,.:
, 9,,. -
.n
\\.
Z Z L X. -=
- , ~ ~ i t - ; :,
1
' ater rial
. :.,-.'&nt.
122P237
- i,..:I
., w,,if ! ;:,,;,
' wetiei
, ':,:(siGfa&
' i.
2.80E+l8 lj4T ' :,.:
~ ~ c p i i o p
'.j
.: +1:'625)'
.(',-.
i-:,:
1.90E+18
- ',.,314~.
??::..:-
- .: :.J,dsiti+i:.:l
.....(r4.875).i,'::'
,..I. Y 8.69E+17
WCAP-I 6669-NP 16 Table 4-5 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EPPY [Reference 41
" - Determined from surveillance data.
Material Description :
hitial RTm Beat Number Reactor V-l Beltllne Region Loration...:.......
. I..
- Matl.
'1dent; 50 1
1
-47.6
-47.6 47.6
-47.4 47.6 40 40 40
-56
-30 Chemistry Factor
~ T N D D OF Point Beach Unit 1 Evalnation at 53 114T Nozzle Belt Forging (NB)
Intermediate Shell Plate (IS)
Lower Shell Plate (LS)
NB to IS Circ. Weld (1 00%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
Point Beach Unit 2 Evaluation Nozzle Belt Forging (NB)
Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Cuc. Weld (200%)
. Intermediate to LS Circ. Weld (100%)
77.0 79.3*
35.8*
167.0 167.0 167.0 167.6 167.0 76.0 58.0 42.8' 170.5 180.0 53 EFPY Fluence, dcmZ.
EFPY 314T Margin
- ART, OF Clad/Lob Alloy Se Interface -:
114T 122P237 A981 1-1 C1423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 at 53 114T.
3/4T 122P237 A981 1-1 C1423-1 ST1 762 1P0815 1 PO66 1 71249 61782 EFPY
, 314T 2.84Et18 4.86Et19 3.70E+19 2.84E+18 3.10Et19 3.10E+19 3.71E+19 2.60E+19 4.12Et18 4.66E+19 4.15E+19 4.12Et-18 3.75E+19
. :. I/&.. ::I.
1 -
314T. '
123V352 123V500 123W195 21935 SA-1484 1.92E+18 329E+19 2.51E+19 1.92E+18 2.10E+19 N/A 2.51E+19 1.76E+19 2.79E-t-18 3.16E+19 2.81E+19 2.79E+18 2.54E+19 34.0 56.4 56.4 65.7 65.7 N/A 61.7 65.7 123V352 f23V.500 123W195 21935 72442 127.1 161.4 102.0 111.6 218.8 N/A 223.4 211.1 123.6 149.6 111.6 120.7, 255.8 30.2 56.4 56.4 65.7 NIA 65.7 61.7 65.7 110.4 145.7 94.6 83.6 NIA 183.3 188.4 175.3 109.6 137.9 102.8 893 218.4 30.2 88.3 37.2 65.5 NIA 165.2 174.1 157.0 35.6 63.9 45.8 79.8 187.6 8.81E+17 1.51E+19 1.15E+19 8,81E+17 N/A 9.62E+l8 1.15E+19 8.07E+18 1.28E+18 1.45E+19 1.29E+19 1.29E+18 1.16E+19 34.0
' 34.0 17.0 65.5 60.8 43.1 104.0 44.6 93.5 200.7 N/A 209.1 192.9 49.6 75.6 54.6 111.2 225.0 34.0 34.0 17.0 65.5 60.8
Table 4-6 Adjusted Reference Temperature Evaluation for the Point Beach Unit I and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY peference 41
" - Determined from surveillance data.
Materid ~escri~tiorj.
Initial RT-50 1
1
-47.6
-47.6
-47.6
-47.4
-47.6 40 40 40
-5 6
-30 Reactor~ekel
{
~ e ~ t i i n e Regio~ Location 1 :
Point Beach Unit 1 Evaluation Nozzle Belt Forging (NB)
Intermediate Shell Plate (IS)
Lower Shell Plate (LS)
NB to IS Circ. Weld (100%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
Point Beach Unit 2 Evalnation Nozzle Belt Forging (NB)
Intermediate She11 Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ. Weld (1 00%)
ARTNIvn O F at 114T 40.0 100.2 43.1 86.7 191.6 NIA 201.5 184.4 46.0 Chemistry Factor 77.0 79.3*
35.8*
167.0 167.0 167.0 167.6 167.0 76.0 58.0 42.8*
170.5 180.0 Margin
'-~afl.
-. Idink!:
122P237 A981 1-1 C1423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 123V352 123V.500 123W195 21935 SA-1484 43.EFPY 3 / 4 ~
27.6 83.8 35.4 60.0 NIA 155.5 165.8 148.1 32.6 ART, O F
' :V4T 34.0 56.4 56.4 65.7 65.7 N/A 61.7 65.7 34.0 Heat
?umber 122P237 A9811-1 C1423-1 8TI 762 1P0815 1P0661 71249 61782 123V352 123V500 123W195 21935 72442 at 43 114T 124.0 157-6 100.5 104.8 209.7 N/A 215.8 202.5
,, i ; -
120.0 146.7 109.6 112.7 247.7 3/4T 27.6 56.4 56.4 65.7 NIA 65.7 61.7 65.7 32.6 34.0 17.0 65.5 60.8 43,EFPY Fluence, nlcm2 72.7 52.6 103.2 216.9 EFPY 314T 105.2 1412 92.8 78.1 NIA 173.6 180.1 166.4 105.2 134.7 100.6 82.6 209.5
, ~lri'&L&w'
-Alloy Steel
.Interface 2.38E+18 3.95E+19 3.11E+I9 2.388+18 2.52E+19 2.52E+3+39 3.10E+E9 2.15E+19 60.7 43.6 73.1 178.7 114T 1.61E+18 2.67E+I9 2.11E+19 1.61E+18 1.71E+19 N/A 2.10E+19 1.46E+19 34.0 17.0 65.5 60.8 3/4T.
7.39E+l7 1.23E+19 9.65E+18 7.40E+17 NiA 7.82E+18.
9.62E+18 6.67E+18 1.06E+18 l.l8E+19 1.07E+19 1.06E+I8 9.75E+18
-., j::.T,.., ;;
3.42E+18 3.81B+19 3.45E+19 3.42E+18 3.14E+19 2.32E+18 2.58E+19 2.34E+19 2.32E+18 2.13E+19
Table 4-7 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafninm removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY [Reference 41
.I Material D&iption 53 EFPY ~~uence,
&!cmL
~T-~'O$-'
Margin ART, OF Initial Chemistry at 53, EFPY at 53 EFPY R T m Factor Cladnow
. Reactor Vessel
.MaU Heat
', Beltline Region-Location Ident.
Number AUoy Steef 114T
.3/4T.
1/4T 3/4T 114T 3/4T 1/4T 3/4T
~nterfaee Point Beach Unit 1 Evaluation Nozzle Belt Forging (NB) 122P237 122f237 50 77.0 3.58E+3+18 2.42E+18 1 l.llE+18 47.4 33.7 34.0 33.7 131.4 117.4 Intermediate Shell Plate {IS)
A9811-1 A9811-1 I
79.3*
4.90E+19 3.32E+19 1.52E+19 104.1 88.5 56.4 56.4 161.5 145.9 Lower Shell Plate {LS)
C1423-I Cl423-1 1
35.8*
4.55E+19 3.09E+19 1.41Ei-19 46.4 39.2 56.4 56.4 103.8 96.6 M3 to IS Circ. Weld (1 00%)
SA-1426,
8T1762
-47.6 167.0 3.58E+18 2.42E+18 1.1 1E+18 1029 73.2 65.7 65.7 121.0 91.3 IS Long. Weld (ID 27%)
SA-812 1P0815
-47.6 167.0 3.19E+19 2.16E+19 NIA 201.9 N/A 65.7 NIA 220.0 NIA IS Long. Weld (OD 73%)
SA-775 1P0661
-47.6 167.0 3.19E+19 NIA 9.90E+18 NiA 166.5 N/A 65.7 NIA 184.6 Intermediate to LS Circ. Weld (100%)
SA-I 101 71249
-47.4 167.6 4.43E+19 3.OOE-tl9 1.38E+19 216.4 182.3 61.7 61.7 230.7 196.6 LS Long. Weid (100%)
SA-847 61782
-47.6 167.0 3.05E+19 2.07E+19 9.47E+18 199.9 164.5 65.7 65.7 218.1 182.7
..,+.
I Point Beach Unit 2 Evaluation j
?:..!:..:::.,?:..
Nozzle Belt Forging (NB) 123V352 123V352 40 76.0 5.04E+18 3.41E+3+18 1.56E+18 53.5 38.9 34.0 34.0 127.5 112.9 Inlrmediate Shell Forging (IS) 123V500 I23V500 40 58.0 5.05E+l9 3.42E+19 1.57E+19 76.6 653 34.0 34.0 150.6 139.2 Lower SheIl Forging (LS) 123W195 I23W195 40 42.8*
4.90E+19 3.32E+19 1.52Ei-19 56.2 47.8 17.0 17.0 1132 104.8 NB to IS Circ. Weld (100%)
21935 21935
-56 170.5 5.04E+18 3.41E+18 I.56E+18 120.0 873 65.5 65.5 129.5 96.8 Intermediate to LS Circ. Weld (100%)
SA-1484 72442
-30 180.0 4.65E+19 3.15E-i-19 1.44E+19 234.4 198.4 60.8 60.8 2652 229.2
- - Determined from sucveillance data.
WCAP-16669-NP 19 Table 4-8 Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY [Reference 41
- - Determined from suweillance data.
~at'ehal
~ e ~ ~ r i ~ f i ~ ~
~nitial RTm 50 1
1
-47.6
-47.6
-47.6
-47.4
-47.6 40 40 40
-56
-3 0 Reactor V-L.
Beltline Region L*tion-Point Beach Unit 1 Evaluation Nonle Belt Forging (NB)
Intermediate Shell Plate (IS)
Lower Shell Plate (LS)
NB to IS Circ. Weld (100%)
IS Long. Weld (ID 27%)
Intermediate to LS Circ. Weld (100%)
Point Beaeb Unit 2 Evaluation Nozzle Belt Forging (NB)
Intermediate Shell Forging (IS)
Lower Shell Forging (LS)
NB to IS Circ. Weld (100%)
Intermediate to LS Circ. Weld (100%)
~h&$ry
. Factor 77.0 79.3*
35.8*
167.0 167.0 167.0 167.6 167.0 76.0 58.0 42.8*
170.5 180.0
. Matl.
' Idkt 122P237 A9811-1 C1423-1 SA-1426 SA-812 SA-775 SA-1101 SA-847 Heat Number 122P237 A9811-I C1423-1 8T1762 1P0815 1 PO66 1 71249 61782 43.EF$;Y ~ l u e n c ~
n/cm2 123V352 ' 123V352 C1a-w
~ l l o y Steel hhrface 2.80E+18 3.97E+l9 3.59E+19 2.80E+18 2.57E+19 2.57E+l9 3.51E+19 2.42E+19 3.95E+18 4.04Et19 3.88E+19 3.95E+18 3.67E+19 123V500 123W195 21 935 SA-1484 mThn.n OF 123V500 123W195 2 1935 72442 U4T 1.9OE+I8 2.69E+19 2.43E+19 1.890E+18 1.74E+19 N/A 2.38E+19 1.64E+19
.,/,.,
2.67E+lS 2.74E+l9 2.63Ei19 267E+I8 2.49E+19 at L14T 42.9 100.2 44.4 93.0 192.4 N/A 206.7 189.7 48.7 73.5 53.9 109.3 224.1
- 314T
- - -.-.,.:l'.l..
8.69E+17 I.UE+19 1.1 lE+19 8.69E+17 NIA 7.98E+18 1.09E+19 7.51E-i-18 1.23E+18 1.25E+19 1.20E+19 1.23E+18 1.14E+19 43'EFPY 314T.
30.0 83.9 36.9 65.0 N/A 156.5 171.6 153.6 34.9 61.7 45.0 78.3 186.5 M a w
. 114T
.;'.I; 34.0 56.4 56.4 65.7 ART, OF 3/4T 30.0 56.4 56.4 65.7 at 114T:
126.9 157.6 101.8 111.1 210.5 N/A 221.0 207.8 122.7 147.5 110.9 118.8 254.9 43.EFPY 314~
110.0 141.3 94.3 83.1 N/A 174.6 185.9 171.9 108.9 135.7 1020 87.8 217.3 65.7 N/A 61.7 65.7
.,..,.....:. 1 34.0 34.0 17.0 65.5 60.8 N/A 65.7 61.7 65.7 34.0 34.0 17.0 65.5 60.8
Contained in Tables 4-9 and 4-10 is a summary of the limiting ART values used in the generation of the Point Beach Units 1 and 2 reactor vessel PT limit curves. The limiting materials for the "axial-flaw" methodology are Point Beach Unit 1 Intermediate Shell Axial Welds SA-812 (114T location) and SA-775 (3/4T location). The limiting material for the "circ-flaw" methodology is Point Beach Unit 2 Intermediate to Lower Shell Circumferential Weld SA-1484.
TABLE 4-10 Summary of the Limiting ART Values (without Hafnium) Used in the Generation of the Point Beach Units 1 and 2 Heatup/Cooldown Curves WCAP-16669-NP 2 1 TABLE 4-9 Summary of the Limiting ART Values (with Hafnium) Used in the Generation of the Point Beach Units 1 and 2 Heatup/Cooldown Curves
..,+C?
...::.'A
,.<, :.:;~.,',
'. ILWting
<cAxiai-F,!ajv
.......mT.2c.:
1- :.;:.=.
.2':ii:..
7.::-..:.---:.ir::l
.7 ::.:=x:,..:...
I
/..............
- '..
- . : J ~ ~ T, ~ J T J < ~ Z ; ; ~ ~
i
)..
)
\\
iL??...&
EFPY -
I Limiting 6 ' ~ / r e - ~ ~ k w ' ! ~. e ~
.,/
_.....I -
- .-.L17',:;:r;
..L'..-:
,7::=.....
pdi$ti:.~&&h:uiitj1:
I.::'..:.:..
I -.,: -::.
- ...
- ;',I>-..+.=::.
. J :=,,, 7+;,
.a.
. 1 / 4 ~ 0:
1,,
-:-I,
-:',, :374$::t0~)
173.6 183.3 209.7 218.8 180.1 188.4 43 53 Point!B@ach Uhit 2.
.)
7................;
.i 215.8 223.4 146.7 149.6 209.5 218.4 43 53 134.7 137.9 247.7 255.8
5 HEATUP AND COOLDOWN PIRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A, Revision 4.
Figure 5-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 43 EFPY with the "Flange-Notch" requirement and hafnium and using the "Axial-flaw" methodology. This curve was generated using the 1998 ASME Code Section XI, Appendix G. Figure 5-2 presents the limiting cooldown curve without margins for possible instrumentation errors using cooldown rates of 0,20,40,60 and 100°F/hr applicable for 43 EFPY with the "Flange-Notch" requirement and hafnium. Again, this curve was generated using the 1998 ASME Code Section XI, Appendix G. These PT limit curves bound those generated using the "Circ-flaw" methodology with the limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth weld.
Figure 5-3 presents the limiting heatup curve without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 53 EFPY with the "Flange-Notch" requirement and hafnium. This curve was generated using the 1998 ASME Code Section XI, Appendix G. Figure 5 4 presents the limiting cooldown curve without margins for possible instrumentation errors using cooldown rates of 0,20,40,60 and 100°F/hr applicable for 53 EFPY with the "Flange-Notch" requirement and hafnium. Again, this curve was generated using the 1998 ASME Code Section XI, Appendix G These PT limit curves bound those generated using the "Circ-flaw" methodology with the limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth weld.
Figure 5-5 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 43 EFPY with the "Flange-Notch" requirement and hafhiumremoval in 2008. This curve was generated using the 1998 ASME Code Section XI, Appendix G. Figure 5-6 presents the limiting cooldown curve without margins for possible instrumentation errors using cooldown rates of 0,20,40,60 and 100°F/hr applicable for 43 EFPY with the "Flange-Notch" requirement and hafnium removal in 2008. Again, this curve was generated using the 1998 ASME Code Section XI, Appendix G. These PT limit curves bound those generated using the "Circ-flaw" methodology with the limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth weld.
Figure 5-7 presents the limiting heatup curve without margins for possible instrumentation errors using heatup rates of 60 and 1 OO°F/hr applicable for 53 EFPY with the "Flange-Notch" requirement and hafnium removal in 2008. This curve was generated using the 1998 ASME Code Section XI, Appendix G. Figure 5-8 presents the limiting cooldown curve without margins for possible instrumentation errors using cooldown rates of 0,20,40,60 and 100°F/hr applicable for 53 EFPY with the "Flange-Notch" requirement and hafnium removal in 2008. Again, this curve was generated using the 1998 ASME Code Section XI, Appendix G. These PT limit curves bound those generated using the "Circ-flaw"
WCAP-16669-NP 23 methodology with the limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth
- weld, Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-8. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed below in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1 and 5-7 (heatup curves only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 ASME Code Section XI, Appendix G as follows:
- where, K,, is the stress intensity factor covered by membrane (pressure) stress, T is the minimum pennissible metal temperature, and RTNur is the metal reference nil-ductility temperature.
The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 8. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperatures for the in service hydrostatic leak tests for the Point Beach Units 1 and 2 reactor vessel at 43 EFPY (with hafnium) is 260°F, at 43 EFPY (without hafnium) is 26l0F, at 53 EFPY (with hafnium) is 26g°F, and at 53 EFPY (without hahium) is 270°F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature Iimit curve constitutes the limit for core operation for the reactor vessel.
Figures 5-1 through 5-8 define all of the above limits for ensuring prevention of non-ductile failure for the Point Beach Units 1 and 2 reactor vessel for 43 and 53 EFPY with the "Flange-Notch" requirement, without instrumentation uncertainties, and withlwjthout hafhium [Reference 81. The data points used for developing the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-8 are presented in Tables 5-1 through 5-8..
MATERIAL PROPERTY BASIS LIMITING MATERIAL,: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 43 EFPY (with Hafnium):
1/4T, 209.7"F 3/4T, 173.6"F Figure 5-1 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 43 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KlJ
WCAP-16669-NP 25 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMlTING ART VALUES AT 43 EFPY (with Hafnium):
1/4T, 209.7"F 3/4T, 173.6"F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-2 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 10O0F/hr) Applicable for 43 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/K13
TABLE 5-1 43 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Hafnium, wlKIc, wlPlange Notch and wlo Uncertainties for Instrumentation Errors)
280 2263 325 2417 280 325 285 2417 285 2303 330 2454 290 2454 Leak Test Limit Temperature (OF) (
241 1
260 1
I Pressure (psig)
- )
2000 1
2485 1
WCAP-16669-NP 28 TABLE 5-2 43 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Hafnium, w/KIc, w/Flange Notch and w/o Uncertainties for Instrumentation Errors)
MATERlAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LTMITING ART VALUES AT 53 EFPY (with Hafnium):
- 114T, 218.8%
3/4T, 183.3"F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-3 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 10O0F1hr) Applicable for 53 EFPY (with Hafnium and without Margins for Tastrumentation Errors) Using 1998 App. G Methodology (w/K,,)
MATERIAL PROPERTY BASIS LIMITTNG M A T E W : Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 53 EFPY (with Hafnium):
1/4T, 218.S°F 3/4T, 183.3OF 0
50 I00 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg, F)
Figure 5-4 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 53 EFPY (with Hafnium and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/K,,)
TABLE 5-3 53 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Hafnium, w/KIc, w/Flange Notch and wlo Uncertainties for Instrumentation Errors)
WCAP-I 6669-NP 33
TABLE 5-4 53 EFPY Cooldown Curve Data Points Using 1998 App. GMethodology (w/Hafniurn, w/Klc, w/Flange Notch and wlo Uncertainties for instrumentation Errors)
MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 43 EFPY (Hafnium Removal):
1/4T, 210.5"F 3/4T, 174.6OF 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-5 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 43 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIJ
UTERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 43 EFPY (Hafnium Removal):
1/4T, 210.5"F 3/4T, 174.6OF 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-6 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 43 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KrJ
TABLE 5-5 43 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (wIHafnium Removal, w/KIc, w/Ii'lange Notch and w/o Uncertainties for Instrumentation Errors) 6o"F'lhr Heatup.
Criticality
.... - '1Liit.
., ~ o Q Q F / ~ ~
Heatup,.. :.critica1ity::~imiti
-:..i T
P -
7
- p;;!:?;
.,...I
. '(OF)
(pig)
,:". "<"Q
'.y~g!g)jTZi 60 0
261 0
60 0
26 1 0
60 62 1 261 62 1 60 62 1 261 62 1 65 62 1 261 62 1 65 62 1 26 1 62 1 70 621 26 1 62 1 70 62 1 261 62 1 75 62 1 261 62 1 75 62 1 261 62 1
TABLE 5-6 43 EFPY Cooldown Curve Data Points Using 1998 App, G Methodology (wIHafnium Removal, w/Klc, w/Flange Notch and wlo Uncertainties for Instrumentation Errors)
WCAP-16669-NP 4 1
MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 53 EFPY (Hafnium Removal):
1/4T, 220.0°F 3/4T, 184.6'F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-7 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIc)
MATEFUAL PROPERTY BASIS LIMITING 1MATElUAL: Intermediate Shell Longitudinal Welds SA-812 (ID) and SA-775 (OD)
LIMITING ART VALUES AT 53 EFPY (Hahiurn Removal):
1/4T, 220.0°F 3/4T, 184.6OF 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 5-8 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/KIJ
TABLE 5-7 53 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w1Hafnium Removal, w/Klc, wmlange Notch and wlo Uncertainties for Instrumentation Errors) 60°~/bi Heatup Criiiiality Limit,.,'10O0P1hr Heatub,.
C!~ticality:L~f:';~:
T -.
P
'T.
'.p
- ..,. :T p
T,
. p:"
- r
(!I!)..
@sig).
cPq, '... (prig):.:
... (OF), : ; (psig), ;,... ~ ~ ( O. F ) ~ ~ ; ~ ;
.;@dg)>.;;
60 0
270 0,
60 0
270 0
60 62 1 270 62 1 60 62 1 270 62 1 65 62 1 270 62 1 65 62 1 270 62 1 70 62 1 270 62 1 70 62 1 270 62 1 75 62 1 270 62 1 75 62 1 270 62 1
6O0F/hr Heatup Criticality Limit
' 1 0 0 ~ ~ / h ~ ~ ~ e a t u ~
criticality, ~ i m i t P
T P
T T
P T
- p.
("I?)
(pig)
(OF);
(psi@
(~sig)
(OV..(~sig).
255 1519 300 1595 255 1505 300 1573 260 1595 305 1680 260 1573 305 1648 265 1680 310 1773 265 1648 3 10 1731 270 1773 315 1876 270 1731 315 1822 275 1876 320 1989 275 1822 320 1923
TABLE 5-8 53 EFPY Cooldown Curve Data Points Using 1998 App, G Methodology (w1Hafnium Removal, wMlc, wmlange Notch and w/o Uncertainties for Instrumentation Errors)
6 REFERENCES
- 1.
Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.
Nuclear Regulatory Commission, May 1988.
- 2.
WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et al., May 2004.
Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure." Dated December 1998, through 2000 Addendum.
AREVA Document 323-9019240-000, "ART Values for Point Beach Unit 1 and Unit 2," S.B.
Davidsaver, et al., dated 6120106.
AREVA Document 43-2308-02, K. K. Yoon, "Initial RTNDT of Linde 80 Weld Materials, " (BAW-2308, Revision 1-A), August 2005.
Westinghouse Letter to WEP, WEP-06-13, "Statistical Evaluation of Reactor Vessel Dosimetry -
Point Beach Units 1 and 2," Kerry B. Hanahan, dated February 14,2006.
WCAP-16083-NP-A, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," S.L. Anderson, May 2006.
Code of Federal Regulations, 10 CFR Part 50, Appendix C;; "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
"Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
WCAP-15 12 1, Revision 1, "Point Beach Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," J.H. Ledger, April 2001.
Westinghouse Letter LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations - Point Beach Units 1 and 2," S.L. Anderson, dated June 4,2004.
Certified Material Test Report MHI-NMC-3277PB1 for Point Beach Unit I, "CMTR for Closure Head", dated 2/14/05.
Certified Material Test Report MHI-NMC-I 838PB2 for Point Beach Unit 2, "CMTR for Closure Head", dated 7/2/04.
APPENDIX A Thermal Stress Intensity Factors (Kit)
The following pages contain the thermal stress intensity factors (KIJ for the maximum heatup and cooldown rates. The vessel radii to the 114T and 3/4T locations are as follows:
I 114T Radius = 67.78 1" 3/4T Radius = 71.03 1"
WCAP-16669-NP A-2 TABLE A-1 Klt Values for 100°F/hr Heatup Curve (wlo Margins for Instrument Errors)
Water Temp.
("'9 Vessel Temperature lor 1000F'hr Heatup (OQ 1I4T 'I'hermIS*ess intensity factor (m SQ, RTr IN,), '
vessel Te~peratuie 1 1 4 ~
L a o n for l()O~F/&;Heatirp (OQ
.. 3/4T
~ht$i%t~l
~t&$i.:{;:jl
~ ~ t e r i s i t ~ ~ ~ a c t & r ; ~ ~ ' : : ~, ~
~
1 SQ;{RT,::IN.)
.. :. _,,-..(, i
..;,,-,. '. :. :=I:/,
- ~
,' (.', :
TABLE A-2 Kit Values for 10O0F/hr Cooldown Curve (w/o Margins for Instrument Errors)