ML023610223

License Amendment Request, TS Changes to TS 3.3.2, TS 3.4.3, TS 3.4.12

ML023610223
Person / Time
Site:	McGuire, Mcguire
Issue date:	12/12/2002
From:	Jamil D Duke Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML023610223 (165)

Text

Po Power.

Duke Duke Power McGuire Nuclear Station 12700 Hagers Ferry Road A Dk, &ev rp-Huntersville, NC 28078-9340 (704)875-4000 D.M. Jamil (704) 875-5333 OFFICE Vice President,McGuire (704) 875-4809 FAX December 12, 2002 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request (LAR)

Technical Specifications (TS) Changes TS 3.3.2 - Engineered Safety Feature Actuation System (ESFAS) Instrumentation TS 3.4.3 - Reactor Coolant System (RCS) Pressure and Temperature (P-T) Limits TS 3.4.12 - Low Temperature Overpressure Protection (LTOP) System By letter dated February 14, 2001, as supplemented by letters dated April 4, 2001 and May 30, 2001, Duke submitted a LAR for changes to the subject TSs. By letter dated August 1, 2001, the NRC transmitted a Request for Additional Information (RAI) to Duke. By letter dated January 28, 2002, Duke withdrew that LAR due to pending NRC review of WCAP-15315, Revision 1. This WCAP proposed the elimination of Reactor Pressure Vessel (RPV) flange requirements of 10 CFR 50 Appendix G.

Pursuant to 10 CFR 50.90, Duke Energy (Duke) hereby resubmits the LAR to revise the following: TS Table 3.3.2-1 Footnote (c) is revised to correct an editorial error in the description of Steam Line Isolation on Steam Line Pressure Negative Rate - High blocking. TS 3.4.3 is revised to update the RCS P-T limits for use up to 34 Effective Full Power Years (EFPY). TS 3.4.12 is revised to update the LTOP limits for use up to 34 EFPY.

Pursuant to 10 CFR 50.60, Duke hereby requests an exemption from the requirements of 10 CFR 50 Appendix G to allow application of American Society of Mechanical Engineers (ASME) Code Case N-641 in the development of P-T and LTOP limits.

The difference between the February 14, 2001 LAR and this LAR includes: the application of the full 10 CFR 50 Appendix G RPV flange temperature requirement in the P-T limits, use of the fourth Diablo Canyon Unit 2 RPV capsule surveillance data for McGuire Unit 1, use of calculated neutron fluence projection instead of best estimate fluence projection, change in the initial RTN* of the RPV and closure head flanges (for Unit 1 only), and application of ASME Code Case N-641 in the I4QQP7 7

U.S. Nuclear Regulatory Commission December 12, 2002 determination of P-T limits and LTOP setpoints. The enclosed technical analysis includes the information requested by the RAI.

Duke concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

Duke requests approval of the LAR no later than May 30, 2003 to allow for implementation before Unit 1 P-T limits expire in Mid August of 2003 and before Unit 2 end of cycle 15 refueling outage starts in mid-September of 2003. A significant number of procedure changes is required to support implementation; therefore, Duke requests that once approved, the LAR shall be implemented within 90 days.

Upon NRC approval of this LAR, Duke commits to implement procedure changes to incorporate the maximum allowable heatup and cooldown rates for RCS pump operating conditions, as specified in , Tables 3 and 6. Revision to the Updated Final Safety Analysis Report (UFSAR) as a result of this LAR will be made in accordance with the requirements of 10 CFR 50.71(e).

The McGuire Plant Operations Review Committee and the Corporate Nuclear Safety Review Board have reviewed and approved this LAR.

Pursuant to 10 CFR 50.91, a copy of this LAR is being forwarded to the North Carolina Division of Radiation Protection.

If there are any questions regarding this LAR, please contact P.

T. Vu at (704) 875-4302.

Very truly yours, D. M. Jamil

Enclosures:

1. Notarized Affidavit

2. Licensee's Evaluation of the Proposed Changes

3. Exemption Request for use of ASME Code Case N-641

4. Westinghouse WCAP-15192, Revision 2

5. Westinghouse WCAP-15201, Revision 2

6. ASME Code Case N-641 2

U.S. Nuclear Regulatory Commission December 12, 2002 Attachments:

1. Proposed Technical Specification Changes (mark-up)

2. Changes to TS Bases Pages (mark-up)

3. Proposed Technical Specification Pages (retyped)

4. TS Bases Pages (retyped)

5. NUREG-1431 mark-up page in May 27, 1997 LAR cc:

L. A. Reyes NRC Region II Administrator Atlanta Federal Center 61 Forsyth ST., SW, Suite 23T85 Atlanta, GA 30303 S. M. Schaeffer NRC McGuire Senior Resident Inspector R. E. Martin NRC Senior Project Manager Office of Nuclear Reactor Regulation Mail Code 08G9 Washington, DC 20555-0001 B. 0. Hall Section Chief, Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645 3

bcc:

L. Vaughn (EClIX)

M. Arey (EC090)

L. Nicholson (ON03RC)

G. Gilbert (CNO1RC)

L. Kunka (MG05SE)

J. Gilreath (EC090)

K. Redmond (MG03A6)

T. Yadon (EC08G)

R. Hart (CN01RC)

NSRB ELL Master File 1.3.2.9 4

ENCLOSURE 1 NOTARIZED AFFIDAVIT

U.S. Nuclear Regulatory Commission Enclosure 1 December 12, 2002 AFFIDAVIT D. M. Jamil, being duly sworn, states that he is Vice President of Duke Energy Corporation; that he is authorized on the part of said corporation to sign and file with the Nuclear Regulatory Commission this amendment to the McGuire Nuclear Station Facility Operating Licenses Numbers NPF-9 and NPF-17 and Technical Specifications; and that all statements and matters set forth herein are true and correct to the best of his knowledge.

D. M. Jamil, Vice President Subscribed and sworn to me: DeCmbr /c, c Date Notary Public: t},. -4011 r. -"*/ -*I My Commission Expires: I

.CC.dZ )> C:Yef Date i.--. N -"

SEAL I

ENCLOSURE 2 LICENSEE'S EVALUATION OF THE PROPOSED CHANGES

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 LICENSEE'S EVALUATION

SUBJECT:

Proposed Changes to Technical Specifications (TS). TS 3.4.3, Reactor Coolant System (RCS) Pressure and Temperature (P-T) Limits, TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System, and TS 3.3.2, Engineered Safety Features Actuation System (ESFAS)

Instrumentation.

1. DESCRIPTION

2. PROPOSED CHANGE

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES I

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002

1.0 DESCRIPTION

This letter is a request to amend Operating Licenses Numbers NPF 9 and NPF-17 for McGuire Nuclear Station Units 1 and 2.

The proposed changes would revise TS 3.4.3 and TS 3.4.12 to update the RCS P-T and LTOP limits for use up to 34 EFPY, and to correct an editorial error in TS 3.3.2. The proposed changes are necessary because the current P-T and LTOP limits will expire in Mid-August of 2003 for Unit 1 and January of 2004 for Unit 2.

2.0 PROPOSED CHANGE

S TS Table 3.3.2-1 Footnote (c) is revised to specify that the Steam Line Isolation on Steam Line Pressure Negative Rate - High function may be blocked below P-lI when Steam Line Isolation on Steam Line Pressure - Low is not blocked. This footnote currently incorrectly specifies that this function may be blocked below P-lI when Safety Injection Steam Line Pressure - Low is not blocked.

TS 3.4.3 is revised to update the P-T limits for use up to a maximum of 34 EFPY for Units 1 and 2. The current P-T limits are effective only to 16 EFPY. Figure 3.4.3-1 currently has two heatup figures, one for each unit. Figure 3.4.3-2 currently has two cooldown figures, one for each unit. Figure 3.4.3-1 will be replaced with four heatup figures (Figures 3.4.3-1 through 3.4.3 4), two for each unit. Figure 3.4.3-2 will be replaced with two cooldown figures (Figures 3.4.3-5 and 3.4.3-6), one for each unit. Limiting condition for operation (LCO) 3.4.3 is revised to refer to these new figures.

TS 3.4.12, Required Actions A.2.2.1 and A.2.2.2 are revised to reflect the temperature and/or cooldown rate limits associated with the use of the residual heat removal (RHR) suction relief valve when any combination of two centrifugal charging and safety injection pumps are capable of injecting into the RCS. TS 3.4.12, Required Action F.1 is revised to reflect the temperature limit associated with the use of the RHR suction relief valve when the required action and associated completion time for one inoperable power operated relief valve (PORV) (Condition E) are not met.

Two new Required Actions (A.5.1 and A.5.2) are proposed for TS 3.4.12. The combination of these two new required actions, a RCS vent of greater than or equal to 2.75 square inches and two operable PORV, is proposed for the condition when any combination of two centrifugal charging and safety injection pumps are capable of injecting into the RCS.

In summary, the proposed amendments revise TS 3.4.3 and TS 3.4.12 to update P-T and LTOP limits for use up to a maximum of 34 EFPY, and to correct an editorial error in TS 3.3.2.

2

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 TS 3.3.2 Bases is revised to reflect the change for TS 3.3.2 described above. TS 3.4.3 Bases is revised to include a summary description of the new excore cavity dosimetry program at McGuire. The new program employs excore cavity dosimetry to determine the reactor vessel neutron fluence through calculation based fluence determination. The new program meets the requirements of 10 CFR 50 Appendix H and guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001."

TS 3.4.12 Bases is revised to reflect the changes for TS 3.4.12 described above, and to clarify that a secured open PORV means that it is physically secured or locked open to prevent it from being subject to active failure. This clarification is consistent with that for a RCS vent already described in the same Bases.

3.0 BACKGROUND

RCS P-T Limits RCS P-T limits are established to provide a margin to brittle failure of the reactor vessel and piping of the RCS pressure boundary. The vessel is the component most subject to brittle failure, and the P-T limits apply mainly to the vessel. 10 CFR 50 Appendix G requires the establishment of P-T limits for specific material fracture toughness requirements of the RCS pressure boundary materials. This rule requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of ASME Section XI Appendix G.

The neutron embrittlement effect on the material toughness is reflected by an increase in the nil ductility reference temperature (RTND 1 ) as exposure to neutron fluence increases. The actual shift in the RTND of the vessel material is established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E-185 and Appendix H of 10 CFR 50. The operating P-T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 Revision 2. The normal use of P-T limits is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable P-T curve to determine that operation is within the allowable region.

McGuire Units 1 and 2 current P-T limits are effective only up to 16 EFPY. The development of the proposed P-T limits relies, in part, on an alternate methodology allowed by ASME Code Case N 641. Westinghouse performed reactor vessel integrity assessments and generated new P-T limits for McGuire Units 1 and 2. The new P-T limits are effective up to 34 EFPY for both units.

Westinghouse Topical Reports WCAP-15192, Revision 2 and WCAP-3

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 15201, Revision 2 provide the details regarding the determination of the new P-T limits.

10 CFR 50 Appendix G Table 1 specifies that the metal temperature of the flange region must exceed the material unirradiated RTNDT by at least 120 °F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure, which is 621 psig for a typical PWR. This requirement was originally based on concerns about the fracture margin in the flange region.

During the boltup process, outside surface stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 120 OF and the pressure limitation of 20 percent of pre-service hydrostatic test pressure were developed using the KIA fracture toughness curve, in the mid 1970s, to ensure that appropriate margins would be maintained.

The requirements for P-T limits incorporate nine numbers of safety margins, one of which is the lower bound fracture toughness curve. There are two lower bound fracture toughness curves available in ASME Section XI, KIA, which is a lower bound on all static, dynamic and arrest fracture toughness, and Kic, which is a lower bound on static fracture toughness only. ASME Code Case N-641 permits the use of Kic fracture toughness curve instead of KIA fracture toughness curve in development of P-T limits. The other margins involved with the process remain unchanged.

The NRC has approved the use of ASME Code Case N-641 in LARs for several nuclear power stations. Below are the licensee application dates and NRC approval dates for these LARs:

Licensee NRC Application Approval Plant Name Date Date Point Beach 1/2 7/14/00 10/6/00 Turkey Point 3/4 7/7/00 10/24/00 North Anna 1/2 6/22/00 5/2/01 Arkansas Nuclear One 2 10/30/01 4/15/02 LTOP System The LTOP system controls RCS pressure at low temperatures so that integrity of the RCS pressure boundary is not compromised by violating the P-T limits of 10 CFR 50 Appendix G. The reactor vessel is the most limiting component within the RCS pressure boundary for demonstrating such protection. Technical Specifications for LTOP provide the maximum allowable actuation logic setpoints for the PORV. Technical Specifications for P-T limits provide the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the requirements of 10 CFR 50 Appendix G during the LTOP mode. The 4

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures. RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased. The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown. Exceeding the RCS P-T limits by a significant amount could cause brittle cracking of the reactor vessel. Once the LTOP system is enabled, no operator action is involved for the system to perform its intended pressure mitigation function. Each time the P-T limits are revised, the LTOP system must be re-evaluated to ensure its functional requirement can still be met. LTOP calculation provides LTOP setpoints and heatup and cooldown restrictions. LTOP setpoint determination is consistent with 10 CFR 50 Appendix G and ASME Code Case N-641.

Reference to Updated Final Safety Analysis Report UFSAR Section 5.2.2 describes the overpressurization protection for the RCS. Section 5.2.4 describes the requirements for RCS pressure boundary fracture toughness. Section 5.4.3.7 describes the requirements for RPV material surveillance program.

Prior Correspondence By letter dated February 14, 2001, as supplemented by letters dated April 4, 2001 and May 30, 2001, Duke submitted a LAR for changes to the subject TSs. By letter dated August 1, 2001, the NRC transmitted a Request for Additional Information (RAI) to Duke. By letter dated January 28, 2002, Duke withdrew that LAR due to pending NRC review of WCAP-15315, Revision 1. This WCAP proposed the elimination of Reactor Pressure Vessel (RPV) flange requirements of 10 CFR 50 Appendix G.

The difference between that LAR and this LAR includes: the application of the full 10 CFR 50 Appendix G RPV flange temperature requirement in the P-T limits, use of the fourth Diablo Canyon Unit 2 RPV capsule surveillance data for McGuire Unit 1, use of calculated neutron fluence projection instead of best estimate fluence projection, change in the initial RTNDT of the RPV and closure head flanges (for Unit 1 only), and application of ASME Code Case N-641 in the determination of P-T limits and LTOP setpoints. The technical analysis section below includes the information requested by the RAI.

Effect on Current P-T limits The effect of ASME Code Case N-641 and Diablo Canyon's RPV capsule surveillance data on McGuire's current P-T limits has been evaluated. The current P-T limits bound the proposed P-T limits for all analyzed heatup and cooldown conditions; 5

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 therefore, they are valid for continued operation of the plant to 16 EFPY. Duke's justification for this conclusion is documented in a site specific calculation (Reference 7).

4.0 TECHNICAL ANALYSIS

Determination of Adjusted RTN* (ART)

The 34 EFPY ART values at the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations for the beltline regions of the McGuire reactor vessels were calculated by Westinghouse. These calculations were in accordance with Regulatory Guide 1.99, Revision 2. Regulatory Guide 1.99, Revision 2 credibility criteria are applied by Westinghouse to determine the appropriate margin term. The calculations determined the ART for the various reactor vessel (RV) materials using Regulatory Guide 1.99, Revision 2, Regulatory Positions 1.1 and 2.1. The selected controlling values are those RV locations with the highest ART for 1/4T and 3/4T whether determined using Regulatory Position 1.1 or 2.1 methodology. Surveillance capsule weld data obtained from Diablo Canyon Unit 2 are used in the determination of the limiting ART value for McGuire Unit 1. The first of four weld data points from Diablo Canyon falls outside the scatter band of

+/- 28 'F. Regulatory Guide 1.99, Revision 2 treats this data as non-credible; however, the data is very predictable when analyzed using more recent embrittlement correlations, such as the recently approved ASTM E-900 correlation. Thus, the Diablo Canyon Unit 2 surveillance weld data described above are utilized in the determination of the limiting ART for McGuire Unit 1, with conservatism added by including a full margin of 56 'F.

The calculation of the ART values for the 1/4T and 3/4T locations at 34 EFPY is presented in Tables 8 and 9 of WCAP-15192, Revision 2 and WCAP-15201, Revision 2. The limiting ARTs used in the generation of the P-T curves are summarized below:

Summary of the Limiting ART Values Used in the Generation of the McGuire Unit I Heatup/Cooldown Curves Limiting Material I RG 1.99 R2 Method 'A T Limiting4RT TLimiting ART, T3/4 Lower Shell Longitudinal Position 2.1 but with full 202 146 Weld Seams 3-442A & C margin added Summary of the Limiting ART Values Used in the Generation of the McGuire Unit 2 Heatup/Cooldown Curves 6

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Determination of P-T Limits The proposed P-T limits are developed using the methods and criteria described in NRC-approved Topical Report WCAP-14040-NP A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", with exception of the following:

1) The fluence values are calculated fluence values (i.e., the values are derived to comply with Regulatory Guide 1.190), and are not the best estimate fluence values.

2) The KIc fracture toughness curve is used in place of the KIA fracture toughness curve (ASME Code Case N-641).

3) The 1996 version of Appendix G to ASME Section XI is used rather than the 1989 version.

The P-T limits and supporting technical analysis are included in WCAP-15192, Revision 2 and WCAP-15201, Revision 2.

The limiting unirradiated RTNDT of 10 OF for the Unit 1 RPV and closure head flanges results in the minimum allowable temperature of 130 °F for this region at pressures greater than 621 psig. It should be noted that the Charpy V-notch tests for Unit 1 RPV and closure head flanges were re-evaluated (Appendix A of Reference

1) using the Material Certifications from Bethlehem Steel Corporation. This re-evaluation concluded that the initial RTNET values for the RPV and closure head flanges should be 10 °F. The limiting unirradiated RTNT of 1 OF for Unit 2 flange region results in the minimum allowable temperature of 121 °F for this region at pressures greater than 621 psig. These limits are included in the P-T curves proposed in this LAR.

Determination of LTOP Limits In general, the methodology presented in Westinghouse technical report, "Pressure Mitigation Systems Transient Analysis Results,"

dated July 1977 and its supplement dated September 1977 is used since this is the current NRC approved licensing basis for McGuire's LTOP system. During Catawba Nuclear Station's most recent LAR for LTOP, the NRC required more advanced methodology be applied to the analysis of the heat input transient due to the recent change in steam generators. A similar analysis is performed for McGuire. The RETRAN computer code is used to model the response of the RCS to the LTOP heat input transient. The RCS model is described in Duke's NRC-approved Topical Report DPC NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology."

This topical report is listed in TS 5.6.5.

For operation at low temperatures, three bounding transients are analyzed which will cause an increase in internal vessel pressure. These are a mass input from a safety injection pump, a 7

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 mass input from a centrifugal charging pump, and a heat input from reactor coolant pump (RCP) startup with temperature asymmetry between the RCS and steam generators. These three transients are evaluated for 2 second PORV opening times to determine a peak pressure equal to setpoint plus overshoot. The pressure overshoot for a mass input transient is a function of the mass input rate, RCS volume, PORV stroke time and an additional factor that accounts for relief capacity asa function of set pressure.

The peak transient pressures are adjusted for static pressure effect, dynamic pressure effect and instrument uncertainty. Duke Energy has an engineering directive that outlines the requirements for performing instrument uncertainty and setpoint calculations. The primary purpose of this directive is to provide a consistent methodology based on standard industry practices for performing instrument uncertainty and setpoint calculations. The calculation methodology is consistent with the intent of ISA-67.04, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation." The methodology conservatively accounts for the typical uncertainty terms such as reference accuracy, drift, temperature effect, and calibration effects (which include measurement and test equipment uncertainty, calibration tolerance and resolution). The random-independent uncertainty terms are combined via the "square-root-sum-of-the-square" (SRSS) technique, whereas the random-dependent and bias uncertainty terms are combined via a combination of SRSS and/or algebraic techniques. The uncertainty calculation performed for the RCS resulted in +/- 12 OF (computer point), +/- 14 °F (chart recorder) and +/- 30 psid (LTOP pressure).

The peak transient pressures are compared to the Appendix G limits, and the heatup and cooldown restrictions are developed to keep the peak pressure for each transient below the Appendix G P-T limits. The PORV nominal setpoint and tolerance are included in the calculation of the peak transient pressures. The PORV actuation setpoint for LTOP is high enough to allow RCP startup and normal operations but low enough to assure adequate protection against system pressurization above the Appendix G limits.

The temperature at which the LTOP protection system must be placed in service is determined using the methodology of ASME Code Case N-641.

Additional requirements must be established for conditions where two centrifugal charging pumps, two safety injection pumps, or one centrifugal charging pump and one safety injection pump are capable of mass addition. These conditions are evaluated and all specific requirements are developed.

8

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 LTOP Limits - Unit 1 Tables 1 and 2 below summarize the Appendix G limits, and peak vessel pressures for the mass input and heat input transients, respectively. The peak transient pressure is the sum of the nominal PORV setpoint, pressure overshoot, dynamic pressure effect, static pressure effect, and instrument error. A 380 psig PORV nominal setpoint and 30 psid instrument error are used in the calculation of the peak transient pressures. The 385 psig PORV lift setpoint in TS 3.4.12 is the maximum allowable value.

The difference between this TS value and the nominal value is bounded by the 30 psid instrument error. This allowable value is verified to be adequate for LTOP protection by comparing the peak transient pressures to the Appendix G limits and establishing the maximum heatup and cooldown rates to ensure the peak transient pressures do not exceed the Appendix G limits. The 385 psig allowable value is adequate for LTOP protection provided the maximum heatup and cooldown rates in Table 3 are followed.

Table 1: Unit I Appendix G Limits RV Beitline Region RV Beline Region Pressure limit per WCAP -15192, Rev. 2 (psig)

Temperatur (

Actual Indicated (w I Steady state Cooldown @ Cooldown @ Cooldown @ Cooldown @ Heatup @

12°F Margin) 20 F/Hr 40 F/hr 60F/hr 100F/hr 100F/hr 60 72 0 0 0 0 0 0 60 72 621 579 528 476 367 579 65 77 621 582 530 478 370 579 70 82 621 584 533 481 373 579 75 87 621 587 536 484 376 579 80 92 621 590 539 488 380 579 85 97 621 594 543 492 385 579 90 102 621 598 547 496 390 579 95 107 621 602 552 501 396 579 100 112 621 607 557 507 403 579 105 117 621 613 563 513 410 579 110 122 621 619 570 520 419 579 115 127 621 621 577 528 428 579 120 132 621 621 585 537 439 580 125 137 621 621 594 547 450 583 130 142 621 621 604 557 464 587 135 147 706 661 615 570 478 593 140 152 717 672 628 583 495 600 145 157 728 685 '641 598 513 609 150 162 741 699 657 615 534 620 155 167 755 714 674 634 557 633 160 172 771 731 692 654 582 647 165 177 788 750 713 677 611 664 170 182 807 771 736 703 642 683 175 187 828 794 762 731 677 704 9

Commission Regulatory U.S. Nuclear U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Table 2: Unit 1 Peak Transient Pressures Transient Description Nominal Pressure Dynamic Static Instrument Peak PORV Overshoot Pressure Pressure Error (psi) Pressure Setpoint (psi) Effects (psi) Effects (psi) (psig)

(psig)

Safety Injection with....

0 RCP's in Operation 380 51.3 0.1 4.7 30 466.1 1-2 RCPs in Operation 380 51.3 15.9 4.7 30 481.9 3 RCP's in Operation 380 51.3 34.0 4.7 30 500.0 4 RCP's in Operation 380 51.3 58.9 4.7 30 524.9 Charging/LetdownMismatch with....

0 RCP's in Operation 380 45.6 0.1 4.7 30 460.4 1-2 RCP's in Operation 380 45.6 15.9 4.7 30 476.2 3 RCP's in Operation 380 45.6 34.0 4.7 30 494.3 4 RCP's in Operation 380 45.6 58.9 4.7 30 519.2 Heat Input (Primary- Secondary Temp Mismatch) with...

0 RCP's in Operation AND Tave of 60- WOOF From Reference 6 468.3 Tave to 100 - 180F 522.5 Tave to 180 - 250'F 514.9 1-4 RCP's in Operation Transientnot crediblewith an NC pump initiallyrunning.

Table 3: Unit 1 Heatup and Cooldown Restriction Number of RCP's RCS Temperature Range Running Without Margin OAC Points Chart Recorder Heat-up Limit 0 100 FO/ hr 0 to 4 > 60'F > 72°F > 74 F Cooldown Limit 0 > 150OF > 162 > 164 100 OF/hr 150-115 162-127 164-129 60 OF/hr 115-60 127-72 129-74 40 OF/hr 1-2 > 140OF > 152 > 154 100OF / hr 140-75 152-87 154-89 60 OF/ hr 75-60 87-72 89-74 40 OF/ hr 3 > 1450 F > 157 > 159 100 OF/hr 145-95 157-107 159-109 60 OF/hr 95-60 107-72 109-74 40 °F/hr 4 > 150OF > 162 > 164 100 F//hr 150-115 162-127 164-129 60 °F/hr 115-60 127-72 129-74 40 °F/hr 10

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 From the previous LTOP LAR (Reference 12), the technical justification for verifying that the enable temperature is adequate to provide overpressure protection states that the licensed basis for the enable temperature ensures that the Appendix G limits are protected by either the PORV setpoint or the pressurizer safety valves (PSV). The PSVs are required to relieve pressure between 2435 - 2559 psig. The corresponding Appendix G steady state pressure limit is 2,240 psig at 275 OF and 2,412 psig at 280 OF (Reference 1). Linear interpolation of the P-T limits for a nominal PSV lift setpoint of 2485 psig results in a corresponding temperature limit of 282 OF. Using this methodology, there is no interruption of overpressure protection if the enable temperature is set at 282 OF (294 OF including instrument error) or above.

The LTOP enable temperature Te is the temperature at or above which the safety relief valves provide adequate protection against non-ductile failure. ASME Code Case N-641 specifies that LTOP system shall be effective below the higher temperature determined in accordance with (1) a coolant temperature of 200F and (2) a coolant temperature corresponding to a reactor vessel metal temperature, for all vessel beltline materials, where Te is calculated on a plant specific basis for the axial or circumferential reference flaws using the following equation:

Te = RTND + 50 Ln [((F.Mm(pRi/t)) - 33.2)/20.734]

where, for McGuire Unit 1:

RTNDT = 202 -F F = 1.1, accumulation factor for safety relief valves Mm = the value of Mm determined in accordance with G-2214.1. As provided in WCAP-15192, Revision 2, Section 3.2, Mm =

0.926(Square Root of t) = 2.72 p =vessel design pressure, 2.485 ksi Ri= vessel inner radius, 86.5 in.

t = vessel wall thickness, 8.625 in.

Instrument error of 12 OF and a maximum temperature lag of 30.1

°F between the 1/4 T reactor vessel location and coolant temperature during 100 °F/hr heatup are added to the above equation Thus, Te for Unit 1 = 202 + 34.5 + 12 + 30.1 = 278.6 OF The current McGuire Unit 1 LTOP enable temperature of 300 'F bounds each of the methods discussed above, and is acceptable for operation up to 34 EFPY.

11

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 It is desirable to operate with one centrifugal charging pump and one safety injection pump in service for brief periods during plant heatup (i.e., for accumulator fill and check valve testing). This has shown to be acceptable provided the RHR suction relief valve is available to provide additional relief capacity.

The peak vessel pressure with the RHR suction relief at maximum relieving capacity is approximately 560 psig, which includes static and dynamic pressure effects. The capacity of the RHR suction relief valve (902 gpm) alone is adequate to relieve the full flow of either the centrifugal charging pump (runout limit =

560 gpm), the safety injection pump (runout limit = 675 gpm), two centrifugal charging pumps, or two safety injection pumps.

Therefore, RHR relief valve in conjunction with one assured PORV is adequate to relieve the combined flow of any combination of two centrifugal charging and safety injection pumps. The single failure criterion is satisfied in that any two of the three relief valves will be adequate for LTOP. The peak pressure considering the RHR relief valve (560 psig) is somewhat higher than the PORV's (524.9 psig for 2.0 second stroke time);

therefore, tighter restrictions must be placed on when the second pump may be made operational.

Additional restrictions for use of RHR suction relief valve are provided in TS 3.4.12 Required Actions A.2.2.1, A.2.2.2, and F.1:

1. From Table 1, the Appendix G limits for steady-state condition and 100 °F/hr heatup rate are 621 psig and 579 psig, respectively (at 60 OF). Therefore, the RHR suction relief valve is adequate for all steady-state and heatup conditions.

2. From Table 1, the Appendix G limit for a 100 °F/hr cooldown is 582 psig at 160 OF (172 OF - 174 OF with instrument uncertainties). Therefore, 100 °F/hr cooldown rate must not be used below 174 OF indicated temperature.

3. From Table 1, the Appendix G limit for a cooldown rate of 20

°F/hr is 579 psig at 60 OF (72 OF'- 74 OF with instrument uncertainties). Therefore, cooldown rates of 20 °F/hr or less must be used between 174 OF and 74 OF indicated temperature.

LTOP Limits - Unit 2 Tables 4 and 5 below summarize the Appendix G limits, and peak vessel pressures for the mass input and heat input transients for Unit 2, respectively. Similar to Unit 1, the 385 psig allowable value is adequate for LTOP protection provided the maximum heatup and cooldown rates in Table 6 are followed.

12

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Table 4: Unit 2 Appendix G Limits RV Belfinee Region RV Belthne Region Pressure Limit per WCAP -15201 (psig)

Temperature (*_)

Actual Indicated (w I Steady state Cooldown @ Cooldown @ Cooldown @ Cooldown @ Heatup @ Heatup @

120F Margin) 20 F/Hr 40 F/hr 60F/hr IOOF/hr 60F/hr 100F/hr 60 72 0 0 0 0 0 0 0 60 72 621 621 621 591 512 621 621 65 77 621 621 621 605 530 621 621 70 82 621 621 621 621 549 621 621 75 87 621 621 621 621 570 621 621 80 92 621 621 621 621 593 621 621 85 97 621 621 621 621 620 621 621 90 102 621 621 621 621 621 621 621 95 107 621 621 621 621 621 621 621 100 112 621 621 621 621 621 621 621 105 117 621 621 621 621 621 621 621 110 122 621 621 621 621 621 621 621 115 127 621 621 621 621 621 621 621 120 132 621 621 621 621 621 621 621 125 137 993 980 970 963 962 822 130 142 1034 1025 1019 1018 1 1004 846 Table 5: Unit 2 Peak Transient Pressures Safety Iniection with....

0 RCP's in Operation 380 51.3 0.1 4.7 30 466.1 1-2 RCP's in Operation 380 51.3 15.9 4.7 30 481.9 3 RCP's in Operation 380 51.3 34.0 4.7 30 500.0 4 RCP's in Operation 380 51.3 58.9 4.7 30 524.9 Charging/LetdownMismatch with....

0 RCP's in Operation 380 45.6 0.1 4.7 30 460.4 1-2 RCP's in Operation 380 45.6 15.9 4.7 30 476.2 3 RCP's in Operation 380 45.6 34.0 4.7 30 494.3 4 RCP's in Operation 380 45.6 58.9 4.7 30 519.2 Heat Input (Primary- Secondary Temp Mismatch) with...

0 RCP's in Operation AND Tave of 60 - 100°F From Reference 6 468.3 Tave to 100- 180'F 522.5 Tave to 180 - 250'F 514.9 1-4 RCP's in Operation Transientnot credible with an NCpump initially running. -

13

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Table 6: Unit 2 Heatup and Cooldown Restrictions Number of RCPs RCS Temperature Range Heat-up Limit Running Without Margin OAC Points Chart Recorder Max Rate 0 to 4 > 60°F > 72°F > 74°F 100°F/hr Number of RCP's RCS Temperature Range Cooldown Limit Running Without Margin OAC Points Chart Recorder Max Rate 0-3 > 60°F >72 >74 10 OF / hr 4 > 65°F > 77 >79 100°F/hr 65-60 77-72 79-74 60 F hr From the previous LTOP LAR (Reference 12), the technical justification for verifying that the enable temperature is adequate to provide overpressure protection states that the licensed basis for the enable temperature ensures that the Appendix G limits are protected by either the PORV setpoint or the pressurizer safety valves (PSV). The PSVs are required to relieve pressure between 2435 - 2559 psig. The corresponding Appendix G steady state pressure limit is 2188 psig at 195 OF and 2,355 psig at 200 OF (Reference 2). Linear interpolation of the P-T limits for a nominal PSV lift setpoint of 2485 psig results in a corresponding temperature limit of 204 OF. Using this methodology, there will be no interruption of overpressure protection if the enable temperature is set at 204 OF (216 OF including instrument error) or above.

Similar to Unit 1:

Te = RTNDy + 50 Ln [((F.Mm(pRi/t)) - 33.2)/20.734]

where, for McGuire Unit 2:

RTND = 123 OF F = 1.1, accumulation factor for safety relief valves Ms = the value of Ms determined in accordance with G-2214.1. As provided in WCAP-15201, Revision 2, Section 3.2, M, =

0.926(Square Root of t) = 2.69 p = vessel design pressure, 2.485 ksi Ri= vessel inner radius, 86.5 in.

t = vessel wall thickness, 8.465 in.

14

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Instrument error of 12 OF and a maximum temperature lag of 29.0

°F between the 1/4 T reactor vessel location and coolant temperature during 100 °F/hr heatup are added to the above equation Thus, Te for Unit 2 = 123 + 35.2 + 12 + 29.0 = 199.2 OF The current McGuire Unit 2 LTOP enable temperature of 300 OF bounds each of the methods discussed above, and is acceptable for operation up to 34 EFPY.

Similar to Unit 1, additional restrictions for use of RHR suction relief valve are provided in TS 3.4.12 Required Actions A.2.2.1, A.2.2.2, and F.1:

1. From Table 4, the Appendix G limit for steady-state condition and 100 °F/hr heatup rate is 621 psig (at 60 OF). Therefore, the RHR suction relief valve is adequate for all steady-state and heatup conditions.

2. From Table 4, the Appendix G limit for a 100 °F/hr cooldown is 570 psig at 75 OF (87 OF - 89 OF with instrument uncertainties). Therefore, 100 °F/hr cooldown rate must not be used below 89 OF indicated temperature.

3. From Table 4, the Appendix G limit for a cooldown rate of 60

°F/hr is 591 psig at 60 OF (72 OF - 74 OF with instrument uncertainties). Therefore, cooldown rates of 60 °F/hr or less must be used between 89 OF and 74 OF indicated temperature.

Two Centrifugal and Safety Injection Pumps Capable of Injecting into the RCS during LTOP Mode Two new Required Actions (A.5.1 and A.5.2) are proposed for TS 3.4.12. This LCO currently requires two operable PORVs (one for single failure concern) or a RCS vent of greater than or equal to 2.75 square inches for a centrifugal charging or safety injection pump capable of injecting into the RCS. The note in the LCO specifies that a PORV secured in the open position with its block valve opened and power removed is equivalent to a RCS vent of greater than or equal to 2.75 square inches. The combination of proposed Action A.5.1, a vent of greater than or equal to 2.75 square inches and proposed Action A.5.2, two operable PORVs is judged to be adequate to prevent RCS overpressurization when any combination of two centrifugal charging and safety injection pumps capable of injecting into the RCS. The effective mass input rate into the RCS from any combination of two centrifugal charging and safety injection pumps would be less than twice that of a single pump.

15

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 Steam Line Isolation on Steam Line Pressure Negative Rate - High The current TS 3.3.2, Function 4.d.(2), Footnote c specifies that this trip function may be blocked below P-Il when Safety Injection Steam Line Pressure - Low is not blocked. Footnote c should specify that this trip function may be blocked below P-lI when Steam Line Isolation on Steam Line Pressure - Low is not blocked. The Safety Injection on Steam Line Pressure - Low function was proposed to be deleted in TS change submittal dated October 6, 1997. The NRC approved this change by letter dated September 22, 1998 (Amendments 182/164). In the conversion from the former Technical Specifications to NUREG-1431-based Technical Specifications, the NUREG-1431 mark-up page was incorrectly marked to reflect the proposed deletion of Safety Injection on Steam Line Pressure - Low as proposed in the October 6, 1997 submittal. The NRC approved the NUREG-1431-based Technical Specifications for McGuire by letter dated September 30, 1998 (Amendments 184/166). Attachment 5 includes the NUREG-1431 mark up page showing this error.

Impact on UFSAR Accident Analyses The UFSAR Loss of Coolant Accident analysis does not assume a break in the reactor vessel. Compliance with 10 CFR 50 Appendix G, as modified by ASME Code Case N-641, continues to ensure that the integrity of the reactor vessel is maintained. Therefore, the proposed changes do not impact UFSAR accident analyses.

Summary of Technical Analysis The proposed changes to the P-T and LTOP limits satisfy the requirements of 10 CFR 50 Appendix G, Appendix H, and ASME Section XI Appendix G, as modified by Code Case N-641. The calculation of ART is consistent with the method in RG 1.99, Revision 2. The calculation of fluence values is consistent with the guidance in RG 1.190. The LTOP changes are performed in accordance with approved procedures under Duke QA program and are consistent with the method in ASME Code Case N-641. Duke concludes that the proposed changes conform to the underlying purpose of NRC regulations and maintain the safe operation of the station.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Duke has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

16

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to the reactor coolant system (RCS) pressure and temperature (P-T) limits and low temperature overpressure protection (LTOP) limits are developed utilizing the methodology of American Society of Mechanical Engineers (ASME)Section XI, Appendix G, in conjunction with the methodology of ASME Code Case N-641. Usage of these methodologies provides compliance with-the underlying intent of 10 CFR 50 Appendix G and provides operational limits established to prevent non-ductile failure of the reactor vessel. The Loss of Coolant Accident analysis and other accident analyses in the Updated Final Safety Analysis Report (UFSAR) do not assume failure of the reactor vessel. The P-T, and LTOP limits are not initiators or contributors to accident analyses addressed in the UFSAR. The proposed changes do not alter any assumption previously made in the radiological consequence evaluations nor affect the mitigation of the radiological consequences of an accident previously evaluated.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The changes to RCS P-T limits and LTOP limits are proposed to prevent non-ductile failure of the reactor vessel. The proposed changes do not modify the RCS pressure boundary, nor make any physical changes to the facility. The proposed changes do not introduce any new mode of system operation or failure mechanism.

Therefore, .the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes are developed utilizing the methodology of ASME Section XI, Appendix G, in conjunction with the methodology of ASME Code Case N-641. Usage of these methodologies provides compliance with the underlying intent of 10 CFR 50 Appendix G and provides operational limits established to prevent non-ductile failure of the reactor vessel. This Code case constitutes relaxation from the current requirements of 10 CFR 50 Appendix G.

The alternate methodology allowed by the Code case is based on 17

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 industry experience gained since the inception of the 10 CFR 50 Appendix G requirements and replaces some requirements that have now been determined to be excessively conservative. The more appropriate assumptions and provisions allowed by the Code case maintain a margin of safety that is consistent with the intent of 10 CFR 50 Appendix G. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Duke concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The NRC has established requirements in 10 CFR 50 to protect the integrity of the RCS pressure boundary in nuclear power plants.

The TS P-T limits for heatup, cooldown, inservice leak testing, and hydrostatic testing are established to define an acceptable operational region for prevention of non-ductile failure of the RCS pressure boundary, and to comply with the fracture toughness requirements of 10 CFR 50 Appendix G. These limits are calculated using the adjusted reference nil-ductility temperature, RTN*, corresponding to the limiting beltline region material of the RPV. Because of the neutron embrittlement effect on the RPV material toughness, the RTNDT increases as the reactor exposure to neutron fluence increases, and the P-T limits must be adjusted, as necessary, based on exposure evaluation.

The LTOP system, with appropriate relief capacity and setpoints, is designed to automatically prevent the RCS pressure from exceeding the P-T limits, and prevent the RPV from being exposed to conditions of fast propagating brittle fracture. Once the system is enabled, no operator action is involved for the LTOP system to perform its intended pressure mitigation function.

Whenever the P-T limits are revised, an evaluation is necessary to determine whether the LTOP enable temperature and setpoint remain acceptable.

The staff evaluates the P-T limit curves based on the following NRC regulations and guidance: Appendix G to 10 CFR 50; Generic Letter (GL) 88-11; GL 92-01, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2; and Standard Review Plan (SRP), Section 5.3.2. GL 88-11 advised licensees that the staff would use RG 1.99, Revision 2, to review P-T limit curves. RG 1.99, Revision 2, contains methodologies for determining the increase in transition temperature and decrease in upper-shelf energy resulting from neutron radiation. GL 92-01, Revision 1, requested that licensees submit their RPV data for their plants to the staff for review. GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data from other licensees that 18

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 could affect their RPV integrity evaluations. These data are used by the staff as the basis for the review of P-T limit curves. Appendix G to 10 CFR 50 requires that P-T limit curves for the RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the ASME Code, 1995 Edition through the 1996 Addenda.

SRP Section 5.3.2 provides an acceptable method of determining the P-T limit curves for ferritic material in the beltline of the RPV based on the linear elastic fracture mechanics methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor KI, which is a function of the stress state and flaw configuration. Appendix G to Section XI of the ASME Code requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 for hydrostatic testing.

Appendix G also requires a safety factor of 1.0 on stress intensities resulting from thermal loads for normal and transient operating conditions, as well as for hydrostatic testing. The methods of Appendix G postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress (i.e., of axial orientation). This flaw is postulated to have a depth that is equal to 1/4 of the RPV beltline thickness and a length equal to 1.5 times the RPV beltline thickness. The critical locations in the RPV beltline region for calculating heatup and cooldown P-T curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively. The methodology found in Appendix G to Section XI of the ASME Code requires that licensees determine the adjusted reference temperature (ART or adjusted RTNDr). The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTN*), the mean value of the adjustment in reference temperature caused by irradiation (Delta RTND), and a margin (M) term.

The Delta RTND is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The M term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The M term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the fluence, and the calculational procedures.

RG 1.99, Revision 2, describes the methodology to be used in calculating the M term.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the 19

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002 public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.22(b), an evaluation of the proposed amendment has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9). The proposed amendment meets the criteria for categorical exclusion if it does not involve the following:

1. A significant hazards consideration:

The proposed changes do not involve a significant hazards consideration. This conclusion is supported by the no significant hazards consideration evaluation above.

2. A significant change in the types or significant increase in the amounts of any effluent that may be released offsite:

The proposed changes provide operational limits established to prevent non-ductile failure of the reactor vessel. The changes do not modify the RCS pressure boundary, nor make any physical changes to the facility. The proposed changes do not adversely affect the integrity of the RCS such that its function in the control of radiological consequences is affected. Therefore, the proposed change does not involve a significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

3. A significant increase in individual or cumulative occupational radiation exposure:

In addition to the above, the proposed changes do not involve any new mode of system operation or failure mechanism. Therefore, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

20

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002

7.0 REFERENCES

1) WCAP-15192, Revision 2, McGuire Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, September 2002.

2) WCAP-15201, Revision 2, McGuire Unit 2 Heatup and Cooldown Curves for Normal Operation, September 2002.

3) WCAP-14040, Revision 3, Methodology Used to Develop Cold Overpressure Mitigation System Setpoints and RCS Heatup and Cooldown Unit Curves, April 2002.

4) MCC-1223.03-00-0005, McGuire Nuclear Station Unit 1 Pressurizer Power Operated Relief Valve Setpoint Determination for the LTOP System, November 05, 2002.

5) MCC-1223.03-00-0033, McGuire Nuclear Station Unit 2 Pressurizer Power Operated Relief Valve Setpoint Determination for the LTOP System, November 07, 2002.

6) MCC-1552.08-00-0305, Rev 0, McGuire Nuclear Station Replacement Steam Generator LTOP Heat Input Transient, January 28, 2000.

7) MCC-1201.01-00-0045, Revision 0, Evaluation of McGuire 16 EFPY Operating Curves, September 2002.

8) NRC letter to North Anna Power Station dated May 2, 2001, Issuance of Amendments 226 and 207. These amendments approve new P-T limits, LTOP system setpoints, and LTOP system effective temperature in the Technical Specifications. McGuire's submittal is very similar to this submittal.

9) NRC letter to Arkansas Nuclear One dated April 15, 2002, Issuance of Amendment 242. This amendment approves revised P-T limits for the reactor pressure vessel, and to approve additional Technical Specification restrictions for operation of the LTOP system. McGuire's submittal is similar to this submittal except LTOP effective temperature determination.

10) NRC letter to Point Beach dated October 6, 2002, Exemption from the Requirements of 10 CFR 50.60. This exemption allows the use of ASME Code Case N-641 in support of the licensee's application for revised P-T limits. McGuire's exemption is very similar to this exemption.

21

U.S. Nuclear Regulatory Commission Enclosure 2 December 12, 2002

11) NRC letter to Turkey Point dated October 24, 2000, Exemption from the Requirements of 10 CFR 50.60 and Appendix G. This exemption allows the licensee to apply the methodology of ASME Code Cases N-588 and N-641 as the basis for establishing the P-T limits specified in Technical Specifications and the cold overpressure mitigation system.

McGuire's exemption is similar to this exemption.

12) Duke letter to NRC dated March 29, 1995, McGuire Nuclear Station Units 1 and 2, Technical Specification 3.4.9.3 Amendment, Heatup and Cooldown Curves and LTOP.

22

ENCLOSURE 3 EXEMPTION[ REQUEST FOR USE OF ASME CODE CASE N-641

U.S. Nuclear Regulatory Commission Enclosure 3 December 12, 2002 Request for Exemption from 10 CFR 50 Appendix G Requirements Pursuant to 10 CFR 50.60, the following information provides the basis for the request for exemption from 10 CFR 50 Appendix G requirements, for use of ASME Code Case N-641. 10 CFR 50.60 states that alternates to the requirements in Appendix G of this part may be used when an exemption is granted by the Commission pursuant to 10 CFR 50.12. 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that:

1. The requested exemption is authorized by law:

No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternates to 10 CFR 50 Appendix G and Appendix H when an exemption is granted by the Commission under 10 CFR 50.12.

2. The requested exemption does not present an undue risk to the public health and safety:

The proposed P-T limits and LTOP limits rely on the requested exemption. The proposed P-T limits have been developed using the Kic fracture toughness curve shown on ASME XI, Appendix A, Figure A-2200-1, in lieu of the KIA fracture toughness curve of ASME XI, Appendix G, Figure G-2210-1. Margins that exist in the ASME XI, Appendix G P-T limit determination process are unaffected by this request.

Use of the KIc fracture toughness curve in the development of P-T and LTOP limits is more realistic than the assumption under the use of the KIA fracture toughness curve. The Kic fracture toughness curve models the slow heatup and cooldown process of a reactor coolant system, with the fastest rate allowed being 100 OF per hour. The rate of change of pressure and temperature is often constant, so the stress is essentially constant in this case. Both the heatup and cooldown and the pressure testing are essentially static processes. During development of ASME Code Case N-641 and the accompanying Appendix G Code change, the ASME Section XI Working Group on Operating Plant Criteria (WGOPC) performed assessments of the margins inherent to KIA using realistic heatup and cooldown curves. These assessments led to the conclusion that utilization of the KIA fracture toughness curve was excessively conservative and the Kjc fracture toughness curve provided adequate margin for protection from brittle fracture.

The KIA fracture toughness curve was codified in 1974. The initial KIA conservatism was necessary due to limited experience and knowledge of the fracture toughness of reactor pressure vessel materials over time. The conservatism also provided margin thought to be necessary to I

U.S. Nuclear Regulatory Commission Enclosure 3 December 12, 2002 cover uncertainties and a number of postulated but unquantified effects. Since 1974, additional knowledge has been gained from examination and testing of reactor pressure vessels that had been subject to the effects of neutron embrittlement in both an operating and test environment.

The KIA fracture toughness curve was based on 125 data points. The Kic fracture toughness curve is based on more than 1500 data points. The additional data has significantly reduced the uncertainties associated with embrittlement effects and reduced other uncertainties. The new information indicates the lower bound on fracture toughness provided by the KIA curve is extremely conservative and is well beyond the margin of safety required to protect the public health and safety from potential reactor pressure vessel failure.

3. The requested exemption will not endanger the common defense and security:

This request does not modify any physical plant architectural features, surveillance or alarm features.

Therefore, the common defense and security are not endangered by this exemption request.

4. Special circumstances are present which necessitate the request for an exemption from 10 CFR 50 Appendix G:

Compliance with the regulation would result in undue hardship or other cost that are significant. The RCS P-T operating window is defined by the P-T limit curve developed in accordance with the ASME Section XI Appendix G methodology and the minimum P-T curve for pump operation.

Continued operation of McGuire with these P-T curves without the relief provided by ASME Code Case N-641 would unnecessarily restrict the operating window that results from these operating P-T limits. This constitutes an unnecessary burden that can be alleviated by the application of ASME Code Case N-641 in the development of the proposed P-T curves and LTOP setpoints. Implementation of the proposed P-T curves and LTOP setpoints as allowed by ASME Code Case N-641 does not significantly reduce the margin of safety.

The exemption provides only temporary relief from the applicable regulation until the Code case is codified in the applicable regulation. The underlying purpose of the regulation will continue to be achieved. Application of ASME Code Case N-641 methodology satisfies the underlying requirement for: 1) The RCS pressure boundary be operated in a regime having sufficient margin to ensure, when stressed, the vessel boundary behaves in a non-brittle manner and the probability of a rapidly propagating fracture is minimized; and 2) P-T limits provide adequate margin in consideration 2

U.S. Nuclear Regulatory Commission Enclosure 3 December 12, 2002 of uncertainties in determining the effects of irradiation on material properties.

Conclusion for Exemption Acceptability:

Compliance with the requirements of 10 CFR 50 Appendix G would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety. ASME Code Case N 641 allows a reduction in the fracture toughness used by ASME Section XI Appendix G in the determination of RCS P-T limits.

This proposed alternative is acceptable because it reduces the excess conservatism in the current Appendix G. The safety margin that exists with the revised methodology is still very large.

Restrictions on allowable operating conditions and equipment operability requirements are established to ensure RCS pressure and temperature are within the heatup and cooldown P-T limits specified in Technical Specifications. Therefore, exemption from 10 CFR 50 Appendix G requirements for use of ASME Code Case N-641 in the development of P-T and LTOP limits should be granted.

3

ENCLOSURE 4 WESTINGHOUSE WCAP-15192, REVISION 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15192, Revision 2 McGuire Unit 1 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham SEPTEMBER 2002 Prepared by the Westinghouse Electric Company LLC for the Duke Energy AreJ. A. Gresham, Manager Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02002 Westinghouse Electric Company LLC All Rights Reserved

ii PREFACE This report has been technically reviewed and verified by:

J.H. Ledger RECORD OF REVISION Revision 0: Original Issue Revision 1: An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15192 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-15192 Rev. 0.

Note that only the heatup curves and associated data point tables have changed. The cooldown curves and data points remain valid and were not changed.

Revision 2: Revised PT Curves to account for the use of "Calculated" Neutron Fluence Projections and to included the full flange requirement from 10 CFR 50 Appendix G. Test and tables have been updated to account for the changes described above. In addition to these changes, the Initial RTNDT of the Closure Head Flange has been revised from 40TF to 10TF, while the initial RTNDT of the Vessel Flange has been revised from 29 0F to 10TF (See Appendix A). The credibility analysis for the weld heat 21935/12008 from Diablo Canyon Unit 2 has been updated due to an additional surveillance capsule data point (See Reference 7).

iii TABLE OF CONTENTS LIST OF TABLES .................................................................................................................................. iv LIST OF FIGURES ................................................................................................................................. v EXECUTIVE

SUMMARY

............................................................................................................. vi 1 INTRODUCTION ...................................................................................................................... 1 2 FRACTURE TOUGHNESS PROPERTIES .......................................................................... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .............. 8 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ..................................... 12 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .................. 20 6 REFERENCES ......................................................................................................................... 31 APPENDIXA INITIAL RTmT DETERMINATION FOR VESSEL FLANGES .............................. 32

iv LIST OF TABLES Table 1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTmirr Values for the McGuire Unit 1 Reactor Vessel Materials ....................................................... 3 Table 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules

@ McGuire Unit I and Diablo Canyon Unit 2 ............................................................ 5 Table 3 Calculation of Chemistry Factors using McGuire Unit 1 and Diablo Canyon Unit 2 Surveillance Capsule Data ......................................................................................... 6 Table 4 Summary of the McGuire Unit I Reactor Vessel Beltline Material Chemistry Factors ....... 7 Table 5 Summary of the Peak Pressure Vessel Neutron Fluence Values at the Clad/Base Metal Interface at 34 EFPY (n/cm 2, E > 1.0 MeV) ................................................... 13 Table 6 Summary of the Peak Pressure Vessel Neutron Fluence Values at 34 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) .................................. 14 Table 7 Summary of the Calculated Fluence Factors used for the Generation of the 34 EFPY Heatup and Cooldown Curves ................................................................... 15 Table 8 Calculation of the ART Values for the 1/4T Location @ 34 EFPY ............................. 17 Table 9 Calculation of the ART Values for the 3/4T Location @ 34 EFPY ............................. 18 Table 10 Summary of the Limiting ART Values Used in the Generation of the McGuire Unit 1 Heatup/Cooldown Curves ......................................................................................... 19 Table 11 34 EFPY, 60°F/hr Heatup Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 25 Table 12 34 EFPY, 80 and 100°F/hr Heatup Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 27 Table 13 34 EFPY Cooldown Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 29

V LIST OF FIGURES Figure 1 McGuire Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 60 0F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 ..................... 22 Figure 2 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 ...................... 23 Figure 3 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 ...................... 24

6 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 McGuire Unit 1 reactor vessel. The PT curves were generated based on the latest available reactor vessel information and updated calculated fluences.

The new McGuire Unit 1 heatup and cooldown pressure-temperature limit curves were generated using the "axial flaw" methodology of 1995 ASME Code,Section XI through the 1996 Addenda. In addition, the PT curves were developed using ASME Code Case N-641, which allows the use of the K1, methodology. The materials with the highest adjusted reference temperature (ART) were the lower shell longitudinal weld seams. The PT limit curves were generated for 34 EFPY using heatup rates of 60, 80 and 100F/hr and cooldown rates of 0, 20, 40, 60 and 100°F/hr. These curves can be found in Figures 5-1 through 5-3.

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 RTmjT 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 ART=ur, and adding a margin. The unirradiated RTmr is designated as the higher of either the drop weight nil ductility transition temperature (NDTI) or the temperature at which the material exhibits at least 50 fl-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.

RTmT 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 RTNTT (IRTmr). The extent of the shift in RTNDT 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.'"11 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base 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 212],

"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values (i.e. comply with Reg. Guide 1.190), not the best estimate fluence values. 2)

The K1, critical stress intensities are used in place of the KI, critical stress intensities. This methodology is taken from approved ASME Code Case N-64P1[3 (which covers Code Cases N-640 and N-588). 3) The 1996 Version of Appendix G to Section XIE41 will be used rather than the 1989 version.

The purpose of this report is to present the calculations and the development of the McGuire Unit 1 heatup and cooldown curves for 34 EFPY. 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 G15].

2 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 161. The beltline material properties of the McGuire Unit 1 reactor vessel are presented in Table 1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 1. Additionally, surveillance capsule data is available for four capsules (Capsules U, X, V and Y) already removed from the McGuire Unit 1 reactor vessel. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Table 3. These CF values are summarized in Table 4. It should be noted that in addition to McGuire Unit 1, surveillance weld data from Diablo Canyon Unit 2 was used in the determination of CF for the lower shell longitudinal weld seams of heat # 21935/12008. It should also be noted that Diablo Canyon Unit 2 operates at a lower temperature than McGuire Unit 1. Thus, for conservatism (i.e reduction in ARTNIDT), no temperature adjustments were made to the chemistry factor for weld heat # 21935/12008.

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 2.

3 TABLE I Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the McGuire Unit 1 Reactor Vessel Materials Matceria Decitohu() N() Iitial RT1 1 '

Closure Head Flange B5002 - 0.75 10°F Vessel Flange B4701 - 0.73 10°F Intermediate Shell Plate B5012-1 0.11 0.61 34 0 F Intermediate Shell Plate B5012-2 0.14 0.61 0°F Intermediate Shell Plate B5012-3 0.11 0.66 -130 F Lower Shell Plate B5013-1 0.14 0.58 0°F Lower Shell Plate B5013-2 0.10 0.51 30°F Lower Shell Plate B5013-3 0.10 0.55 150F Intermediate Shell Longitudinal Welds, 2-442A, B & C(b) 0.199 0.846 -50OF Lower Shell Longitudinal Welds, 3-442A, B & Ccb) 0.213 0.867 -50°F Circumferential Weld 9-442(b) 0.051 0.096 -70°F McGuire 1: Surveillance Program Weld Metal 0.198 0.874(c)

Diablo Canyon 2: Surveillance Program Weld Metal 0.219 0.867 Notes:

(a) The Initial RTNDT values for the plates and weld are based on measured data. The closure head flange and vessel flange values were revised to 10'F from 40OF and 29 0F, respectively. See Appendix A for this re evaluation of Initial RTNDT for the flanges.

(b) The intermediate shell longitudinal weld seams 2-442A, B and C were fabricated with weld wire heat numbers 20291 and 12008, Flux Type 1092 Lot Number 3854. The intermediate to lower shell circumferential weld seam 9-442 was fabricated with weld wire heat number 83640, Flux Type 0091 Lot Number 3490. The lower shell longitudinal weld seams 3-442A, B and C were fabricated with weld wire heat number 21935 and 12008, Flux Type 1092 Lot Number 3889. The McGuire Unit 1 surveillance weld metal was made with the same weld heat as the intermediate shell longitudinal weld seams, while the Diablo Canyon Unit 2 surveillance weld metal was made with the same weld wire heat as the lower shell longitudinal weld seams (Justification in WCAP-13949, Ref. 8) Per Reg. Guide 1.99, Revision 2, "weight percent copper" and "weight percent nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld."

(c) Value is the average of five measurements (see Tables 4-1 and 4-2 in WCAP-1499319 ).

4 The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Table 3 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases. Included in Table 2 are the Calculated fluence values for McGuire Unit 1, along with Diablo Canyon Unit 2. All capsule fluence values were determined using ENDF/B-VI cross-sections and followed the guidance in Regulatory Guide 1.190[I°1.

It should be noted that in the calculation of chemistry factor in Table 3, the McGuire Unit 1 and Diablo Canyon Unit 2 data had the ratio procedure applied. In both cases the ratio was slightly less than 1.0. As for Temperature adjustments are concerned, the McGuire Unit 1 data does not require any adjustments since it is being applied to its own plant. The Diablo Canyon surveillance data, on the other hand, should be adjusted for temperature. However, in this instance, it would cause a further reduction in ARTNDr (or Chemistry Factor) since Diablo Canyon Unit 2 operated at a lower temperature than McGuire (for the capsules already withdrawn from Diablo Canyon Unit 2). Thus, it was decided for conservatism, not to reduce the chemistry factor any further beyond what was done with the ratio procedure.

Credibility Surveillance capsule data exists from McGuire Unit 1 for the intermediate shell plate B5012-1 and the intermediate shell longitudinal welds (Heat # 20291/12008). Additionally, surveillance capsule weld data (Heat # 21935/12008) was obtained from Diablo Canyon Unit 2. This is the same heat as the McGuire Unit 1 lower shell longitudinal welds. The McGuire surveillance data has been deemed credible in a Technical Report (ATI-98-012-T005, Revision 1, Dated November 1998 by Tim Greisbach), which has been issue to the NRC through Duke Energy (From H.B. Baron on 1/7/99). The surveillance weld data from Diablo Canyon was evaluated for credibility under WCAP-154231t. In this report, it was determined that the first of four weld data points falls outside the scatter band of +/- 28 0F. Regulatory Guide 1.99, Revision 2 treats this data as non-credible, however, the data is very predictable when analyzed using more recent embrittlement correlations, such as the recently approved ASTM E-900 correlation. Thus, the Diablo Canyon Unit 2 surveillance weld data will be utilized for determination of the limiting adjusted reference temperature (ART), with additional conservatism using the full aA in the margin term.

5 TABLE 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules

@ McGuire Unit I and Diablo Canyon Unit 2 U 4.05 x 10" n/cm2 , (E > 1.0 MeV)

X 1.50 x 10'9 n/cm2 , (E > 1.0 MeV) 2 V 2.08 x 1019 n/&m , (E > 1.0 MeV)

Z(C) 2.38 x 1019 n/crn2 , (E > 1.0 MeV) 2 Y 2.86 x 1019 n/cn (E > 1.0 MeV)

U 3.38 x 10 t (E > 1.0 MeV)

X 9.19 x 10" (E > 1.0 MeV)

Y 1.55 x 10' 9 (E > 1.0 MeV)

V 2.41 x 10'9 (E > 1.0 MeV)

NOTES:

(a) Re-Calculated in WCAP-15253tl 2l.

(b) Per WCAP-15423t7 1.

(c) Dosimeters only.

6 TABLE 3 Calculation of Chemistry Factors using McGuire Unit I and Diablo Canyon Unit 2 Surveillance Capsule Data Mateija! Ca "

1ýFF

)apsl FFAR Intermediate Shell U 0.405 0.749 30.95 23.18 0.561 Plate B5012-1 X 1.50 1.11 33.51 37.20 1.23 V 2.08 1.20 81.01 97.21 1.44 (Longitudinal) Y 2.86 1.28 93.10 119.17 1.64 Intermediate Shell U 0.405 0.749 48.44 36.28 0.561 Plate B5012-1 X 1.50 1.11 60.69 67.37 1.23 V 2.08 1.20 74.60 89.52 1.44 (Transverse) Y 2.86 1.28 108.58 138.98 1.64 SUM: 608.91 9.742 CFB5012-1 = E(FF

• RTNT) + 7-(FF2) = (608.91) + (9.742) = 62.5°F Inter. Shell Longitudinal U 0.405 0.749 15 9 .7 1(d) 119.62 0.561 Welds 2-442A, B & C X 1.50 1.11 16 8 .9 8(d) 187.57 1.23 V 2.08 1.20 17 8 .3 5 (d) 214.02 1.44 Y 2.86 1.28 1 8 8 . 52 (d) 241.31 1.64 SUM: 762.52 4.871 0F CF W.d 2 -442 = E(Ff

• RTNT) + Z( FF 2) = (762.52) + (4.871) = 156.5 Lower Shell Long. U 0.338(c) 0.701 170.57(c) 119.57 0.491 Welds 3-442A, B & C X 0.919(c) 0.976 200.38(e) 195.57 0.953 (Using Diablo Canyon Y 1.55(c) 1.121 208.43(e) 233.65 1.257 Unit 2 Surv. Data) V 2.41(c) 1.237 221.33(e) 273.79 1.530 SUM: 822.58 4.231 CF W.d3 -442 = 7(FF

• RTMNrr) + 2:(F72) = (822.58) + (4.23 1) = 194.4°F Notes (a) f= Calculated fluence (x 10"' n/cm2, E > 1.0 MeV), See Table 2.

(b) FF = fluence factor = (028 - o l1agi*.

(c) ARTNDT values are the measured 30 fl-lb shift values.

(d) Ratio = 0.99 (201.3 - 204.2).

(e) Updated Per Capsule V, WCAP-15423. Ratio = 0.986 (208.2 + 211.2). For Conservatism, No Temperature Adjustment used.

7 TABLE 4 Summary of the McGuire Unit 1 Reactor Vessel Beltline Material

"",Material II II II Intermediate Shell Plate B5012-1 74.2 0 F 62.5 0F Intermediate Shell Plate B5012-2 100.3 0 F Intermediate Shell Plate B5012-3 74.9 0 F Lower Shell Plate B5013-1 99.1 0 F1 Lower Shell Plate B5013-2 65.01F Lower Shell Plate B5013-3 65.0 0F Inter. Shell Plate Long. Weld Seams 201.3 0 F 156.5 0 F 2-442A, B, C (Heat # 20291/12008)b))

Lower Shell Plate Long. Weld Seams 208.2 0 F 194.40FOa 3-442A, B, C (Heat # 21935/12008)(c)

Inter. to Lower Shell Plate Circ. Weld 37.5 0 F Sean 9-442 (Heat # 83640)

McGuire 1 Surveillance Program 204.2 0 F Weld Metalcb)

Diablo Canyon 2 Surveillance Program 211.2 0F Weld Metal(c)

Notes (a) This was determined using surveillance capsule data from Diablo Canyon Unit 2.

(b) The McGuire Unit 1 surveillance capsule weld material fabricated with the same weld wire heat as the intermediate shell longitudinal weld seams 2-442A, B, C (Heat # 20291/12008).

(q) The Diablo Canyon Unit 2 surveillance capsule weld material was fabricated with the same weld wire heat as the lower shell longitudinal weld seams 3-442A, B, C of the McGuire Unit 1 (Heat # 21935/12008).

8 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, 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 Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1'3" 41 of the ASME Appendix G to Section XI. The Kic curve is given by the following equation:

K1, = 3 3 .2 + 2 0.7 3 4 *e°2(T-RTr)] (1)

where, K1. = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KI curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Class1, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kim + Kit < Ki (2)

where, Kh = stress intensity factor caused by membrane (pressure) stress K = stress intensity factor caused by the thermal gradients Ki = function of temperature relative to the RTNDT 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

9 For membrane tension, the corresponding K, for the postulated defect is:

Kim = Mmx (plAit) (3) where, Mm for an inside surface flaw is given by:

Mm = 1.85 for ft < 2, Mm = 0.926 47 for 2* f47t_ 3A64, Mm = 3.21 for 7f > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm = 1.77 for 47 < 2, Mm = 0.89347 for 2*47*3.464, Mm = 3.09 for 47 > 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:

Kn, = Mb

• Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kit = 0.953x10"3 x CR x tf, where CR is the cooldown rate in 'F/hr., or for a postulated outside surface defect, Kit = 0.753x10" 3 x HU x t5, where HU is the heatup rate in IF/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1 .

(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)(1) and (2).

(b) Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

Kit = (1.0359Co + 0.6322C, + 0.4753C2 + 0.3855C3) *4f (4)

10 or similarly, Krr during heatup for a 'A-thickness outside surface defect using the relationship:

Kit = (1.043Co + 0.630C, + 0.481C2 + 0.401C 3)

• 4= (5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress chstribution at any specified time during the heatup or cooldown using the form:

o(x) = Co+Ci(x/a)+C2(x / a)2 + C3(xl a)3 (6) and x is a variable that represents the radial distance from 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 OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves'" 21 Section 2.6 (equations 2.6.24 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1 cis determined by the metal temperature at the tip of a postulated flaw at the l/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, 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 KI, at the l14T location for finite cooldown rates than for steady-state operation Furthermore, if conditions exist so that the increase in K1 , exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.

11 The above procedures are needed because there is no direct control on temperature at the l/4T 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 l/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 K1 t for the l1/4T crack during heatup is lower than the Kic for the 1/4T 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 KI, 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 l/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 l/4T flaw located at the l1/4T 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 from 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 from 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 HEADIVESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G151addresses 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 RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for McGuire Unit 1. The limiting unirradiated RTNrm of I 0F (See Appendix A) occurs in both the vessel flange and closure head flange of the McGuire Unit 1 reactor vessel, so the minimum allowable temperature of this region is 1301F at pressures greater than 621 psig. This limit is shown in Figures 6-1 and 6-2 wherever applicable.

12 4 CALCULATION OF ADJUSTED REFERENCE 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 RTNDT + ARTNDT + Margin (7)

Initial RT= is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Coder!]. If measured values of initial RTNDT 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.

ART=r is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

2 ARTNDT = CF

• f o8-o1logos) (8)

To calculate ARTTrir at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth fýdqAhx) = fffr..

• e (-0 24x) (9) where x inches (vessel beltline thickness is 8.63 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections and the results of the calculated peak fluence values at various azimuthal locations on the vessel clad/base metal interface are presented in Table 5. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented inWCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". Table 6 contains the I/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the McGuire Unit 1 reactor vessel.

13 TABLE 5 Summary of the Peak Pressure Vessel Neutron Fluence Values at the Clad/Base Metal Interface at 34 EFPY 2, E > 1.0 (n/crm MeV)

14 TABLE 6 Summary of the Peak Pressure Vessel Neutron Fluence Values at 34 EFPY used for the Calculation ofART Values (n/cm, E > 1.0 MeV)

Intermediate Shell Plate B5012-1 1.95 x 10'9 1.16 x 1019 0.412 x 10'9 Intermediate Shell Plate B5012-2 1.95 x 1019 1.16 x 1019 0.412 x 10'9 Intermediate Shell Plate B5012-3 1.95 x 10'9 1.16 x 1019 0.412 x 10'9 Lower Shell Plate B5013-1 1.95 x 10" 1.16 x 1019 0.412 x 1019 Lower Shell Plate B5013-2 1.95 x 109 1.16 x 1019 0.412 x 1019 Lower Shell Plate B5013-3 1.95 x 10'9 1.16 x 1019 0.412 x 1019 Intermediate Shell Longitudinal 1.21 x 10'9 0.721 x 10'9 0.256 x 10'9 Weld Seam 2-442A (00 Azimuth)

Intermediate Shell Longitudinal 1.73 x 1019 1.03 x 1019 0.366 x 10' Weld Seam 2-442B & C (1200 & 2400 Azimuth)

Intermediate to Lower Shell 1.95 x 10'9 1.16 x 10'9 0.412 x 1019 Cirumferential Weld Seam 9-442 Lower Shell Longitudinal Weld 1.73 x 10' 9 1.03 x 1019 0.366 x 1019 Seams 3-442A & C (600 & 3000 Azimuth)

Lower Shell Longitudinal Weld 1.21 x 1019 0.721 x 1019 0.256 x 10'9 Seam 3-442B (1800 Azimuth)

15 Contained in Table 7 is a summary of the fluence factor (FF) values used in the calculation of adjusted reference temperatures for the McGuire Unit 1 reactor vessel beltline materials for 34 EFPY TABLE 7 Summary of the Calculated Fluence Factors used for the Generation of the 34 EFPY Heatup and Cooldown Curves Matmil........~T ORT' Intermediate Shell Plate B5012-1 1.16 x 1019 1.04 0.412 x 10'9 0.75 Intermediate Shell Plate B5012-2 1.16 x 1019 1.04 0.412 x 10'9 0.75 Intermediate Shell Plate B5012-3 1.16 x 10'9 1.04 0.412 x 10'9 0.75 Lower Shell Plate B5013-1 1.16 x 10'9 1.04 0.412 x 10'9 0.75 Lower Shell Plate B5013-2 1.16 x 1019 1.04 0.412 x 10'9 0.75 Lower Shell Plate B5013-3 1.16 x 10'9 1.04 0.412 x 10'9 0.75 Intermediate Shell Longitudinal 0.721 x 1019 0.91 0.256 x 10'9 0.63 Weld Seam 2-442A (00 Azimuth)

Intermediate Shell Longitudinal 1.03 x 10'9 1.01 0.366 x 10'9 0.72 Weld Seams 2-442B & C (1200 & 24 0 °Azimuth)

Intermediate to Lower Shell 1.16 x 0'9 1.04 0.412 x 10'9 0.75 Cirumferential Weld Seam 9-442 Lower Shell Longitudinal Weld 1.03 x 10'9 1.01 0.366 x 10'9 0.72 Seams 3-442A & C (600 & 3000 Azimuth)

Lower Shell Longitudinal Weld 0.721 x 10'9 0.91 0.256 x 10'9 0.63 Seam 3-442B (1800 Azimuth)

Notes:

(a) Fluence Factor at the I4T vessel thickness location.

(b) Fluence Factor at the 3/4T vessel thickness location.

16 Margin is calculated as, M = 2 ac?, + aý . The standard deviation for the initial RTNDT margin term, is a, 00F when the initial RTNDT is a measured value, and 17'F when a generic value is available. The standard deviation for the ARTmNT margin term, ca, is 17'F for plates or forgings, and 8.5"F for plates or forgings when surveillance data is used. For welds, 0 A is equal to 28'F when surveillance capsule data is not used, and is 14MT (half the value) when credible surveillance capsule data is used oa need not exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 8 and 9 are the calculations of the 34 EFPYART values used for generation of the heatup and cooldown curves.

17 TABLE 8 Calculation of the ART Values for the 1/4T Location @ 34 EFPY

• .* ~, ~,~~ ,:*~ ~7 ~7&i ~ ~ ~ ~*,..~~ ý 'i* .:

.........-  : . =,:...... .... . 44at*

Mateia 2 CF F RTT "'AR1.)

RT~ M'gn ART

~~~~~~~~~~~~~~~

. ..... ...... .. , *~ ... ... .....

u-Intermediate Shell Plate B5012-1 Position 1.1 74.2 1.04 34 77.2 34 145 Position 2.1 62.5 1.04 34 65.0 17 116 Intermediate Shell Plate B5012-2 Position 1.1 100.3 1.04 0 104.3 34 138 Intermediate Shell Plate B5012 -3 Position 1.1 74.9 1.04 -13 77.9 34 99 Lower Shell Plate B5013-1 Position 1.1 99.1 1.04 0 103.1 34 137 Lower Shell Plate B5013-2 Position 1.1 65.0 1.04 30 67.6 34 132 Lower Shell Plate B5013 -3 Position 1.1 65.0 1.04 15 67.6 34 117 Intermediate Shell Longitudinal Position 1.1 201.3 0.91 -50 183.2 56 189 Weld Seam 2-442A (00 Azimuth) Position 2.1 156.5 0.91 -50 142.4 28 120 Intermediate Shell Longitudinal Position 1.1 201.3 1.01 -50 203.3 56 209 Weld Seam 2-442B & C (1200 & 2400 Azimuth) Position 2.1 156.5 1.01 -50 158.1 28 136 Intermediate to Lower Shell Position 1.1 37.5 1.04 -70 39.0 39.0 8 Circumferential Weld Seam 9-442 Lower Shell Longitudinal Position 1.1 208.2 1.01 -50 210.3 56 216 Weld Seams 3-442A & C (60*& 3000 Azimuth) Position 2.1 194.4 1.01 -50 196.3 56() 202 Lower Shell Longitudinal Position 1.1 208.2 0.91 -50 189.5 56 196 Weld Seam 3-442BI (1800 Azimuth) Position 2.1 194.4 0.91 -50 176.9 28 155 Notes:

(a) Initial RTNDT values are measured values (b) ART = Initial RTm)r + ARTNDT + Margin (OF)

(c) ARTmT = CF

• FF (d) Deemed Not Credible (Reference 7), thus the full coA was used in the Margin Term.

18 TABLE 9 Calculation of the ART Values for the 3/4T Location @ 34 EFPY

- -A f) J:(O (0tqd)

Intermediate Shell Plate B5012-1 Position 1.1 74.2 0.75 34 55.7 34 124 Position 2.1 62.5 0.75 34 46.9 17 98 Intermediate Shell Plate B5012-2 Position 1.1 100.3 0.75 0 75.2 34 109 Intermediate Shell Plate B5012 -3 Position 1.1 74.9 0.75 -13 56.2 34 77 Lower Shell Plate B5013-1 Position 1.1 99.1 0.75 0 74.3 34 108 Lower Shell Plate B5013-2 Position 1.1 65.0 0.75 30 48.8 34 113 Lower Shell Plate B5013 -3 Position 1.1 65.0 0.75 15 48.8 34 98 Intermediate Shell Longitudinal Position 1.1 201.3 0.63 -50 126.8 56 133 Weld Seam 2-442A (00 Azimuth) Position 2.1 156.5 0.63 -50 98.6 28 77 Intermediate Shell Longitudinal Position 1.1 201.3 0.72 -50 144.9 56 151 Weld Seam 2-442B & C (120° & 2400 Azimuth) Position 2.1 156.5 0.72 -50 112.7 28 91 Intermediate to Lower Shell Position 1.1 37.5 0.75 -70 28.1 28.1 -14 Circumferential Weld Seam 9-442 Lower Shell Longitudinal Position 1.1 208.2 0.72 -50 149.9 56 156 Weld Seams 3-442A & C (60°& 3000 Azimuth) Position 2.1 194.4 0.72 -50 140.0 56(d) 146 Lower Shell Longitudinal Position 1.1 208.2 0.63 -50 131.2 56 137 Weld Seam 3-442B (1800 Azimuth) Position 2.1 194.4 0.63 -50 122.5 28 101 Notes:

(a) Initial RThDT values are measured values.

(b) ART = Initial RTNDT + ARTNDT + Margin (IF)

(c) ARTmr = CF

• IF (d) Deemed Not Credible (Reference 7), thus the full 0c a was used in the Margin Term

19 The Lower Shell Longitudinal Welds (Seams 3-442A and C) are the limiting beltline material for all the PT limit curves to be generated. Contained in Table 10 is a summary of the limiting ARTs to be used in the generation of the McGuire Unit I reactor vessel PT limit curves. These limiting curves will be presented in Section 5.

TABLE 10 Summary of the Limiting ART Values Used in the Generation of the McGuire Unit 1 Heatup/Cooldown Curves In"uitigA Ti ARL~n 202 146

20 5 HEATUP AND COOLDOWN PRESSURE-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 3.0 and 4.0 of this report This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section 1 of this report.

Figures 1 and 2 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60, 80 and 100*F/hr applicable for the first 34 EFPY. These curves were generated using a combination ofthel996 ASME Code Section XI, Appendix G with the limiting ART values and the ASME Code Case N-641. Figure 3 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100 F/lhr applicable for 34 EFPY. Again, these curves were generated using a combination of the 1996 ASME Code Section XI, Appendix G with the limiting ART values and the ASME Code Case N-641. Allowable coribinations of temperature and pressure for specific temperature change rates are below and to the right of the limit line shown in Figures 1 through 3. 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 I and 2. 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 Code Case N-641131 (approved in February 1999) as follows:

1.5 K* < KI, where, K* is the stress intensity factor covered by membrane (pressure) stress, KI, = 33.2 + 20.734 elo °rRr)],

T is the minimum permissible metal temperature, and RTNDT 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 5. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 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 4.0 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 McGuire Unit 1 reactor vessel at 34 EFPY is 262TF. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve constitutes the limit for core operation for the reactor vessel.

21 Figures 1 through 3 define all of the above limits for ensuring prevention of nonductile failure for the McGuire Unit 1 reactor vessel for 34 EFPY. The data points used for the heatup and cooldown pressure temperature limit curves shown in Figures 1 through 3 are presented in Tables 11, 12 and 13.

22 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202-F 3/4T, 146-F 2500 2250 2000 1750 0

S1500 E 1250 D

=U 1000 Cu C,

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 1 McGuire Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 6 0 °F/hr)

Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

23 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202-F 314T, 146 0F 2500 2250 2000 1750 1500 A

IL

• 0 1250 Ul ca 1000 Cu t.,

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2 McGuire Unit I Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

24 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T,2020 F 3/4T, 146-F 2500 2250 2000 1750 1500 In In E 1250 00

'a 750 500 250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 3 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

25 TABLE 11 34 EFPY, 601F/hr Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

... 6ft0 P ... ' xCritiality Limit Le. k Tet Limit Tem.' Pre'st.m ~mp' remress.' )TeieWX Pes, 60 0 262 0 245 2000 60 621 262 621 262 2485 65 621 262 621 70 621 262 621 75 621 262 621 80 621 262 621 85 621 262 621 90 621 262 621 95 621 262 621 100 621 262 621 105 621 262 621 110 621 262 621 115 621 262 621 120 621 262 621 125 621 262 621 130 621 262 621 130 621 262 621 130 673 262 673 135 687 262 687 140 703 262 703 145 721 262 721 150 741 262 741 155 755 262 755 160 771 262 771 165 788 262 788 170 807 262 807 175 828 262 828 180 851 262 851 185 877 262 877 190 905 262 905 195 937 262 937 200 972 262 972 205 1010 262 1010 210 1052 262 1152

26 TABLE 11 - (Continued) 34 EFPY, 6 0 °F/hr Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 60 sFhr:,VitmityLiit LeakTes 'LimJf*It,57 1 _ress* Ttemp. -_: z".. Temp. .PresL, 215 1099 265 1208 220 1151 270 1271 225 1208 275 1323 230 1271 280 1380 235 1323 285 1443 240 1380 290 1513 245 1443 295 1589 250 1513 300 1674 255 1589 305 1767 260 1674 310 1870 265 1767 315 1983 270 1870 320 2109 275 1983 325 2247 280 2109 330 2399 285 2247 290 2399

27 TABLE 12 34 EFPY, 80 and 100°F/hr Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

Press.' S*T

'Temp. Press. T p. 'P es.** Tem.......... .. ........

60 0 262 0 60 0 262 0 245 2000 60 601 262 601 60 579 262 579 262 2485 65 601 262 601 65 579 262 579 70 601 262 602 70 579 262 579 75 601 262 603 75 579 262 579 80 601 262 605 80 579 262 579 85 601 262 607 85 579 262 579 90 601 262 610 90 579 262 579 95 601 262 614 95 579 262 579 100 601 262 616 100 579 262 579 105 601 262 621 105 579 262 579 110 602 262 621 110 579 262 579 115 605 262 621 115 579 262 579 120 610 262 621 120 580 262 580 125 616 262 621 125 583 262 583 130 621 262 624 130 587 262 587 130 621 262 634 135 593 262 593 130 624 262 645 140 600 262 600 135 634 262 658 145 609 262 609 140 645 262 673 150 620 262 620 145 658 262 691 155 633 262 633 150 673 262 710 160 647 262 647 155 691 262 732 165 664 262 664 160 710 262 757 170 683 262 683 165 732 262 784 175 704 262 704 170 757 262 814 180 728 262 728 175 784 262 848 185 755 262 755 180 814 262 886 190 784 262 784 185 848 262 927 195 818 262 818 190 886 262 972 200 854 262 854 195 927 262 1010 205 895 262 895 200 972 262 1052 210 941 262 941 205 1010 262 1099 215 991 262 991 210 1052 262 1151 220 1047 262 1047

28 TABLE 12 - (Continued) 34 EFPY, 80 and 100°F/hr Heatup Curve Data Points Using 1996 App. G &ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 80 0 F/b "Ofitic1Wi~ Limt. >:o0~ ~ "Uiic. Limit', ekTstLmt Tem..Pess

'emp. :Prez~'Teip~ ~Prs. 'Temp. treu.'lTemp.-R'riss.

215 1099 265 1208 225 1108 265 1108 220 1151 270 1271 230 1176 270 1176 225 1208 275 1328 235 1251 275 1251 230 1271 280 1380 240 1334 280 1334 235 1328 285 1437 245 1425 285 1425 240 1380 290 1500 250 1494 290 1494 245 1437 295 1570 255 1557 295 1557 250 1500 300 1647 260 1627 300 1627 255 1570 305 1731 265 1704 305 1704 260 1647 310 1824 270 1789 310 1789 265 1731 315 1927 275 1882 315 1882 270 1824 320 2040 280 1985 320 1985 275 1927 325 2165 285 2098 325 2098 280 2040 330 2303 290 2223 330 2223 285 2165 335 2455 295 2360 335 2360 290 2303 295 2455

29 TABLE13 34 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

IF2 XI'AO0j hFr.gi. .:h O-TO.40hr..0 '.600 /a*: -1adStteIOO 0

(! P ,(psig) T (T=P (pAiO) pg .T( %T(F

'P(u)'P (psig).

60 0 60 0 60 0 60 0 60 0 60 621 60 579 60 528 60 476 60 367 65 621 65 582 65 530 65 478 65 370 70 621 70 584 70 533 70 481 70 373 75 621 75 587 75 536 75 484 75 376 80 621 80 590 80 539 80 488 80 380 85 621 85 594 85 543 85 492 85 385 90 621 90 598 90 547 90 496 90 390 95 621 95 602 95 552 95 501 95 396 100 621 100 607 100 557 100 507 100 403 105 621 105 613 105 563 105 513 105 410 110 621 110 619 110 570 110 520 110 419 115 621 115 621 115 577 115 528 115 428 120 621 120 621 120 585 120 537 120 439 125 621 125 621 125 594 125 547 125 450 130 621 130 621 130 604 130 557 130 464 130 621 130 621 135 615 135 570 135 478 130 697 130 651 140 628 140 583 140 495 135 706 135 661 145 641 145 598 145 513 140 717 140 672 150 657 150 615 150 534 145 728 145 685 155 674 155 634 155 557 150 741 150 699 160 692 160 654 160 582 155 755 155 714 165 713 165 677 165 611 160 771 160 731 170 736 170 703 170 642 165 788 165 750 175 762 175 731 175 677 170 807 170 771 180 790 180 762 180 716 175 828 175 794 185 821 185 797 185 759 180 851 180 820 190 856 190 835 190 807 185 877 185 848 195 894 195 878 195 861 190 905 190 879 200 937 200 925 200 920 195 937 195 914 205 984 205- 978 200 972 200 952 210 1036 205 1010 205 995 210 1052 210 1042 215 1099 215 1093 220 1151

30 TABLE 13 - (Continued) 34 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 225 1208 230 1271 235 1341 240 1418 245 1503 250 1598 255 1702 260 1817 265 1944 270 2085 275 2240 280 2412

31 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 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

3. ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1", January 17, 2000.

[Sub Reference 1: ASME Code Case N-640, "AlternativeReference FractureToughnessfor Development ofP-T Limit Curvesfor Section X), Division I ", February26, 1999.]

[Sub Reference 2: ASME Boiler andPressureVessel Code, Case N-588, "Attenuationto Reference Flaw OrientationofAppendix Gfor CircumferentialWelds in Reactor Vessels", Section X1, Division 1, Approved December 12, 1997.]

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix (, "Fracture Toughness Criteria for Protection Against Failure." Dated December 1995, through 1996 Addendum.

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

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

6. "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.

7. WCAP-15423, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program," B Burgos, et. al., September 2000.

8. WCAP-13949, "Analysis of Capsule V Specimens and Dosimeters and analysis of Capsule Z Dosimeters From Duke Power Company McGuire Unit I Reactor Vessel Radiation Surveillance Program," E. Terek, et al., February 1994.

9. WCAP- 14993 Analysis of Capsule Y from The Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., December 1998.

10. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

11. 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."

12. WCAP-15253, "Duke Power Company Reactor Cavity Neutron Measurement Program for William B McGuire Unit 1 Cycle 12," J.D. Perock, et, al., July 1999.

32 APPENDIX A INITIAL RTNDT DETERMINATION FOR THE VESSEL FLANGES

34 2172 Ift.. 1.63 BETHLEHEM STEEL. CORPORATION

---Bethlehem ------PLANT-,AETAtURGICAt DEPARTMENT

-2..--

REPORT OF TESTS P. -

No.R 750- A PAS.ComnbustionL-Engineering__Inc._

-ffune_-28.__-_,,69 P-lrh ...r's Req.1s.t6 Dmt- I *s-rPtlo. B.B.C. Ord, 47-64621 V24/67 Closure Head Flange - Code B5002 3102-386 Contract 2167-NP Dwg: SC-2067-2-0 - Item 2 REQ.UIREMENTS C M P S S N C, v jMoj "AS1 Spec.A508-64 (Sl) Cde Case 1332 Y.S. (.2) T.n.I'. Elg I V-NC IB Icust, SPec.

Tangential 50,000 80,000 18.0 1 38.0 P3C9 (D)

I A30 ft.lbs.O CH1EMICAL COMPOSITION I D A39

____ N_ _ C Mn P S Si NI C, V Me CO 123X362

-Checx Analyses: .21 .65 .008 . .28 071 -,18 .01 .60 ,01 TI12_

T6 .20

.2. -73

.71 .01

.01 .010

.010 .29

.29 ..75l- ,..19 .0 66

.6-,1 ,011

,012 123X369VA La.-,d-io-T.l2 .505x2"

.. i.s. 67P500

(.T) T...,.

TEST R ESULTS 90,000 Elon 27.0 72.2 Br..kn "T6 "68,ooo 90,0. 72.7

-rV-Kotcl_2harays Broken +1.0"F Test specimens vere str as relieved at 1150F _ Ft.Lbs. Fibrous L. E.

for.40-hours --- hen fur ace cooled tD below 60001? T12 119 66 .084

. 7e.* T12 102 55 .073 Subscribed and Sworn t before me t a ) T12 116 66 .o077 230 day of Jue, 1969._ T6 82 61 .082

_ __T6_ _68 33 1.052

____ __ __ T6 152 10 .092 V James It Hanga I ____

Nota v"Public Yagnet c Partic le Insp .ction i a performed

,A--

MY commission Txpires: Jan. 29th, LYY3 3 s^4 perfoz r4 n+ -I^"

in ace ~dance -with csoe Nih*ompto u+/-em . & r..4.2.......... 4........

_ _ __ _ _ _ _ _ (Pro-d qethod)

I CERTIFY THE ABOVE RESULTS TO BE CORRECT AS CONTAINED IN THE RECORDS OF THE COMPANY.

SDrop Weight tet material was not cyclic tempered. ---- ---- E. -A. Reidý 1575 7F isELx Foe Z4 P~ours - Wa-rFr ~UArwovcjI*, ~2fr IZZO~ o-e OL~ ý&'"

35 1175 (10-68) BETHLEHEM STEEL CORPORATION REPORT OF TRANSITION TEMPERATURE SURVEY Report No .... 75,0

- - -e --- - -...........-- 969 CUSTOMER Coadbmstion Erngineering, Inc.

CUSTOMER ORDER NO BSC ORDER NO 47-6467a 3102-386 TYPE FORGING Clcsize Head Flange - 12,X36kVA -I I.S.T.T.- Minus 9a'F CODE TEMP. FT. LGS. FTELS FIBROUS LATERAL CODE TEMP. FT. LBS F LS  % FIBROUS LATERAL AEAEEXPANSION AVERAGE EXPANSION

-7 - -- 12r0F I 3 _ 0_ --tzoo __ _ _ _ _ _ __ _ __ _ __ _ _ _

r-2 54 0 150

• -- _--___---** 4o - . 4... ._ _ _ __ _ __,_ _ _ _ _ _ _

T12 7 33.7 0 .0033 _

T12 T_

Ti21

-1_ 4 7_5_

16 2-5

.oRo

.057 1 I [0 86 67.3 3(i .068 _

T6 -P-c'F 12 2 6 1 .o. ql ,

" _2 88 40 .C069 124 _11.3 64 .c£5 __I af , 150 *__.1 N C*q_

. I_

o00 , -.

"12_ l.-,,

148 .091 J L1200F 154_

-_ __o__

If ___

I-T12 -4*,~ ,*

___I____

_ _D68__ ___ -__ _

T6

____ ___6___ _156-.o ____ _10____ *1

____ ____ _____ _____ _____ _____ __j__ _______ ______

_ __ _ ___i - _ __ __ _ _ _ _ _ _ _ _ _

__ i _ _ _ _ I I _ _ _ _ _ _

t I -

- I TRANSITION CURVE ATTACHED.

36 272 (R.1 -.- BETHLEHEM STEEL CORPORATION

---Beth lehem _LANT-'AETAUURGICAt DEPARTMEWT REPORT OF TESTS 966 REPORT NO --

P*,,1 .. Comb o.. tion.E.ig*ne.er.ig, -Inc.......... . ..

D o t, " D escrip tion B .S C . Or e

? u rc h. .. B

• q..f 0itlon sReer 47 - 64620 I 4/24/67 Vessel Flange--Item 1-Code B4701 3102-383 Drawing: SC-2067 Rev.1 REQUIREMENTS P S S. Nn V Mo C- MJ v_ *S ...(_97., *,, ~ a. V.=I._C~uY*.[*$_pc ific t_

, +10-F P3C9D (Modified) 50000 80000 18.0 38.0 30 ft.lbs.avi.

Tangential I a fN1ICAI CflMPAS1TION -

I aS; I ra.j I 1.01.- //

I '-"-'

H-1. No. I C Mn

?.. 1 P

P J

II S S

lllll-I I J'*  %/ T 77 N;

l Ie I

I I C.r .60/

.l 1 Z2 W2 U 1 .1 _ _. o- - - l I . .,.* ,- I-r . " I

" I I I I I I Il I II I I L * *

• i,l  ! I  ! I I I C.neCi( Analyses T K 1 .6'

--- [-.1 0 .010

-73 2

-K,2# .3 5// .01.2' oi

." T6 6_2I1 0 _.009 .73 / .34 .. 02/ [ 61 .1 TEST RESULTS

,d.ntF.-tion No- I  ; I Y.S. (.2% k

1. T.-.i. /..r R.d /

1f2-2W201VA1 T12. 505x2" 1 67500 1~ 8 900011f 216.5 72

.0 I for V-Noch ys at +10°F Test specimens wbre tempred at 11251F.,

HT TRENIT (OF **100/,

FORGING) 587. 1.080

• .-, - h FTt- 12 Ft s ios .084 1220'F., held fo* 40 hou$s - Furnace ,Cooled F 1*  ; 112 15-7 587. .0-8

__/6**-___'j-I/

76 58% 1.077

__________ - ~ O76 112 527 .485 07 Drop Weight test material was not Cy lic 98i Tempered. Was g ven the same treatm nt as 10 I -agnetic Particle Inspec :ion was "I*_s_"' e-fo'*&M*, in o-rda*-ne-'-t-h--

Subscrib and s, orn to efore me thisaar

.1(Prodi Method aarp

-My coxxnissio n ex i res: 0 h1*. 1969 City of Bethleh__m-North_ pton Count_1_1_

RECORDS OF THE COMPANY.

I CERTIFY THr ,.COVE RESULTS TO BR CORRECT AS COWTAINED IN-1THZ REMA.RKS S-*A*Reid

37 BETHLEHEM STEEL CORPORATION 2172 (,. 1-651 e

Bethi eh eM -.. PLANT-MEIALLURGICAI

--- DEPARTMEM Supplement REPORT OF TESTS - -- - - RFPORT No 966-A__ _

.-.- r.-rrt-8/_2.o... _68 PUA Combustion Engineering, Inc.

P.F-h...... . . ....

ident f-tion No TR.~e I do-~

122-1201VAI sransition Temperature Survey I _

Temp. Ft.Lbs Av. i . p Ftb h_[__s -_

0F 0% .002 ÷20OF 110 _ 61%

55% _ .078

.073

-120 4 104 3 5-- --0%.4----- .000

.0- 92 .072 11% .017 80OF 59 ______ 100W. .094

-80oF 23

.005 153 1007. .092

-40°F 80 31%= .062 54 25%° .041 0 /00_

I.S.T.T. = Mins 69 F _-__j_

Transition Cure AttaZed. - I IN TilE RECOPD% OF THE COMPANY.

I CERTIFY THE ABOVE RESULTS TO BE CORRECT AS CONTAINED E.A.REID/,

This report issued as a supplement to test report - --

No. 966 dated July 25th, 1968. Please attach to - ---- MlALLUR

?*

S subject test report.

.e

ENCLOSURE 5 WESTINGHOUSE WCAP-15201, REVISION 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15201, Revision 2 McGuire Unit 2 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham SEPTEMBER 2002 Prepared by the Westinghouse Electric Company LLC for the Duke Energy Approved. 44 Gresham, Manager Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02002 Westinghouse Electric Company LLC All Rights Reserved

PREFACE This report has been technically reviewed and verified by:

J.H. LedgerJ A RECORD OF REVISION Revision 0: Original Issue Revision 1: An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15201 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-1 5201 Rev. 0.

Note that only the heatup curves and associated data point tables have changed, however the 100 Deg.F/Hr. Heatup Rate was not affected by this error for McGuire Unit 2. The cooldown curves and data points remain valid and were not changed.

Revision 2: Revised PT Curves to account for the use of "Calculated" Neutron Fluence Projections and to included the full flange requirement from 10 CFR 50 Appendix G. Text and tables have been updated to account for the changes described above. Title was changed to remove reference to Code Case N-640.

iii TABLE OF CONTENTS LIST OF TABLES ................................................... .............. .......................... iv LIST OF FIGURES ....................................................... v EXECUTIVE

SUMMARY

.................................................. vi I INTRODUCTION ................. . ................... ............ 1 2 FRACTURE TOUGHNESS PROPERTIES ........................................................................... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .............. 8 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .................................... 12 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .................. 20 6 REFEREN CES ......................................................................................................................... 28

iv LIST OF TABLES Table I Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT=rxr Values for the McGuire Unit 2 Reactor Vessel Materials ........................................................ 3 Table 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules

@ M cGuire Unit 2 ..................................................................................................... 5 Table 3 Calculation of Chemistry Factors using McGuire Unit 2 Surveillance Capsule Data ......... 6 Table 4 Summary of the McGuire Unit 2 Reactor Vessel Beltline Material Chemistry Factors ....... 7 Table 5 Summary of the Peak Pressure Vessel Neutron Fluence Values at the Clad/Base Metal Interface at 34 EFPY (n/em 2 , E > 1.0 MeV) ................................................... 13 Table 6 Summary of the Peak Pressure Vessel Neutron Fluence Values at 34 EFPY used for the Calculation of ART Values (n/cm2 , E > 1.0 MeV) .................................. 14 Table 7 Summary of the Calculated Fluence Factors used for the Generation of the 34 EFPY Heatup and Cooldown Curves .................................................................... 15 Table 8 Calculation of the ART Values for the 1/4T Location @ 34 EFPY ............................. 17 Table 9 Calculation of the ART Values for the 314T Location @ 34 EFPY ............................. 18 Table 10 Summary of the Limiting ART Values Used in the Generation of the McGuire Unit 2 Heatup/Cooldown Curves .......................................................................................... 19 Table 11 34 EFPY, 60*F/hr Heatup Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 25 Table 12 34 EFPY, 80 and 100F/hr Heatup Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 26 Table 13 34 EFPY Cooldown Curve Data Points Using 1996 App. G

& ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) .............. 27

V LIST OF FIGURES Figure 1 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 601F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology &ASME Code Case N-641 ...................... 22 Figure 2 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 80

& 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology &ASME Code Case N-641 ...................... 23 Figure 3 McGuire Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 0F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 ...................... 24

vi 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 McGuire Unit 2 reactor vessel. The PT curves were generated based on the latest available reactor vessel information and updated calculated fluences.

The new McGuire Unit 2 heatup and cooldown pressure-temperature limit curves were generated using the "axial flaw" methodology of 1995 ASME Code,Section XI through the 1996 Addenda. In addition, the PT curves were developed using ASME Code Case N-641, which allows the use of the K1, methodology. The material with the highest adjusted reference temperature (ART) was the lower shell forging 04. The PT limit curves were generated for 34 EFPY using heatup rates of 60, 80 and 100°F/hr and cooldown rates of 0, 20, 40, 60 and 100 0 F/hr. These curves can be found in Figures 5-1 through 5-3.

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTrDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTmr 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 ARTmNT, and adding a margin. The unirradiated RTmr 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 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 601F.

RTmT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT*NT 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 RTmT (IRTNDT). The extent of the shift in RTNDT 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.'" 11 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTmTrr + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base 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 212],

"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values (i.e. comply with Reg. Guide 1.190), not the best estimate fluence values. 2)

The K1t critical stress intensities are used in place of the Kia critical stress intensities. This methodology is taken from approved ASME Code Case N-6411 31 (which covers Code Cases N-640 and N-588). 3) The 1996 Version of Appendix G to Section X11 41 will be used rather than the 1989 version.

The purpose of this report is to present the calculations and the development of the McGuire Unit 2 heatup and cooldown curves for 34 EFPY. 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 G15).

2 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 []. The beltline material properties of the McGuire Unit 2 reactor vessel are presented in Table 1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 1. Additionally, surveillance capsule data is available for four capsules (Capsules V, X, U and W) already removed from the McGuire Unit 2 reactor vessel. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Table 3. These CF values are summarized in Table 4.

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 2.

3 TABLE 1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the McGuire Unit 2 Reactor Vessel Materials MtaDe's"!MptionT, ;T-z Closure Head Flange - -- 1F Vessel Flange - - -4 0F Intermediate Shell Forging 05 0.153 0.793 -40 F Lower Shell Forging 04 0.15 0.88 -30°F Intermediate to Lower Shell Girth Weld 0.039 0.724 -68 0 F McGuire Unit 2 Surveillance Weld 0'b 0.04 0.74 -

Notes" (a) The initial RTmr values for the plates and welds are based on measured data.

(b) The surveillance weld was made with the same weld wire and flux as the intermediate to lower shell girth weld (weld wire heat # 895075, Type Grau L.O. # LW320 Flux, Lot P46).

4 The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Table 3 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases. Included in Table 2 are the Calculated fluence values for McGuire Unit 2. All capsule fluence values were determined using ENDF/B VI cross-sections and followed the guidance in Regulatory Guide 1.1901103.

It should be noted that in the calculation of chemistry factor in Table 3, the ratio was not required since it was equal to 1.0. As for Temperature adjustments are concerned, the McGuire Unit 2 data does not require any adjustments since it is being applied to its own plant.

5 TABLE 2 Calculated Integrated Neutron Exposure of the Surveillance Capsules

@ McGuire Unit 2 V 3.23 x 10s n/cr 2 , (E > 1.0 MeV)

X 1.47 x 10' 9 n/cm2, (E > 1.0 MeV U 2.04 x 10'9 ncmr2 , (E > 1.0 MeV) yO,) 2.08 x 10'9 n/cm 2 , (E > 1.0 MeV)

Z(11) 2.41 x 10'9 nkrcm 2 , (E > 1.0 MeV)

W 3.07 x 10'9 n/cm 2 , (E > 1.0 MeV)

NOTES:

(a) Re-Calculated in WCAP-15334171.

(b) Dosimeters only.

6 TABLE 3 Calculation of Chemistry Factors using McGuire Unit 2 Surveillance Capsule Data MateiV Cam~ apue ....

RT~3/4 Intermediate Shell V 0.323 0.689 58 64 40.40 0.475 Forging 05 X 1.47 1.11 91.12 101.14 1.23 U 2.04 1.19 84.14 100.13 1.42 (Axial) W 3.07 1.30 130.33 169.43 1.69 Intermediate Shell V 0.323 0.689 68.97 47.52 0.475 Forging 05 X 1.47 1.11 98.28 109.09 1.23 U 2.04 1.19 91.18 108.50 1.42 (Tangential) W 3.07 1.30 102.03 132.64 1.69 SUM 808.85 9.63 CFFwoi g = *(FF

• RTmrr) + _(FF2) = (808.85) + (9.63) = 84.0F V 0.323 0689 38.51 26.53 0.475 Circumferential X 1.47 1.11 35.93 39.88 1.23 Weld Seam(d) U 2.04 1.19 23.81 28.33 1.42 W 3 07 1.30 43.76 56.89 1.69 SUM 151.63 4.815 CFS/p Wad = E(FF *RTND) 2

+ _(FF ) (151.63) + (4.815) 31.5F Notes:

(a) f= Calculated fluence (x 10'9 n/cn 2 , E > 1.0 MeV). See WCAP-15334t1 1.

(0 (N) FF = fluence factor-= f *oso o (c) ARTT values are measured, See WCAP-147991s], Appendix B.

(d) Per WCAP-15201111 , the ratio procedure was shown not to be applicable (i.e. Ratio = 1.0).

7 TABLE 4 Summary of the McGuire Unit 2 Reactor Vessel Beltline Material Chemistry Factors

"~~~~~~~.........

• ?*.........................*'"" ' ...." * ... "

• Intermediate Shell Forging 05 117 0 F 84.0°F Lower Shell Forging 04 115.8°1F (a)

Beitline Region Circ. Weld Metal 52.7°F 31 .5°F Notes:

(a) No surveillance material for forging 04, thus Position 2.1 does not apply.

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, K1 , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1 ,, for the metal temperature at that time. KI, is obtained from the reference fracture toughness curve, defined in Code Case N-64 1, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1"43&41 of the ASME Appendix G to Section XI. The KIc curve is given by the following equation:

KI, = 3 32 + 2 0.7 3 4 *e[°R2(OT-RTT)] (1)

where, Ki = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This Kic curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kr + Kit < Kic (2)

where, Kil = stress intensity factor caused by membrane (pressure) stress K1, = stress intensity factor caused by the thermal gradients K1 = function of temperature relative to the RTrr 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 K1 for the postulated defect is:

Ki = Mm x (pR / t) (3) where, M. for an inside surface flaw is given by:

M. = 1.85 for ." < 2, Mm = 0.926 f- for 2 < -,5* 3.464, Mm = 3.21 for l > 3.464 Similarly, M. for an outside surface flaw is given by:

Mm = 1.77 for f-" < 2, Mm = 0.893 F1 for 2* 4r" <53.464, M. = 3.09 for f-" > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding K, for the postulated defect is:

K1 , = Mb

• Maximum Stress, where Mb is two-thirds of Mm The maximum KI produced by radial thermal gradient for the postulated inside surface defect of G-2120 is KIt = 0.953x10-3 x CR x tfs, where CR is the cooldown rate in 0F/hr., or for a postulated outside surface defect, Ki, = 0.753x10-3 x HU x tf 5, where HU is the heatup rate in 'F/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1 .

(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)(1) and (2).

(b) Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

Kit = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3) *4* (4)

10 or similarly, Krr during heatup for a 1/4-thickness outside surface defect using the relationship:

Kit = (1.043Co + 0.63 0Ci + 0.481C2 + 0.40IC3) * (5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

o-(x) = Co+ Ci(x / a)+ C2(x /a) 2 + C3(x / a) 3 (6) and x is a variable that represents the radial distance from 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 OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves'4 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 from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, 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 KI, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.

11 The above procedures are needed because there is no direct control on temperature at the 1/4T 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 K1 , for the l/4T crack during heatup is lower than the K1 c for the 1/4T 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 KI, 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 l1/4T flaw located at the 1/4T 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 from 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 from 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 REQUIREMENTS 10 CFR Part 50, Appendix GI- 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 RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for McGuire Unit 2. The limiting unirradiated RTyrr of I°F occurs in the closure head flange of the McGuire Unit 2 reactor vessel, so the minimum allowable temperature of this region is 121'F at pressures greater than 621 psig. This limit is shown in Figures 1 and 2 wherever applicable.

12 4 CALCULATION OF ADJUSTED REFERENCE 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 RTNDT + ARTmrTr + Margin (7)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code191. If measured values of initial RTmNT 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.

ARTNmT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTNDT = CF * (28-0o10 log 0 (8)

To calculate ARTNrr at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

fid fsr*e c-0*)

= f..) (9) where x inches (vessel beltline thickness is 8.465 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections and the results of the calculated peak fluence values at various azimuthal locations on the vessel clad/base metal interface are presented in Table 5. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". Table 6 contains the 1/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the McGuire Unit-2 reactor vessel.

13 TABLE 5 Summary of the Peak Pressure Vessel Neutron Fluence Values at the Clad/Base Metal Interface at 34 EFPY (n/cm2, E > 1.0 MeV)

14 TABLE 6 Summary of the Peak Pressure Vessel Neutron Fluence Values at 34 EFPY used for the Calculation of ART Values (n/cm 2, E > 1.0 MeV)

IIurf 4TII..

Intermediate Shell Forging 05 1.85 x 109 1.11 x 1019 0.403 x 109 Lower Shell Forging 04 1.85 x 109 1.11 x I01 0.403 x 1019 Intermediate to Lower Shell 1.85 x 109 1.11 x 101 0.403 x 101 Cirumferential Weld

15 Contained in Table 7 is a summary of the fluence factor (FF) values used in the calculation of adjusted reference temperatures for the McGuire Unit 2 reactor vessel beltline materials for 34 EFPY.

TABLE 7 Summary of the Calculated Fluence Factors used for the Generation of the 34 EFPY Heatup and Cooldown Curves 3  :~

-Material - ~ 3 T "'t~~

-,i Intermediate Shell Forging 05 1.11 x 1019 1.03 0.403 x 10'9 0.75 Lower Shell Forging 04 1.11 x 1019 1.03 0.403 x 10'9 0.75 Intermediate to Lower Shell 1.11 x 1019 1.03 0.403 x 10'9 0.75 Cirumferential Weld Notes:

(a) Fluence Factor at the l1/4T vessel thickness location.

(b) Fluence Factor at the 3/4T vessel thickness location.

16 Margin is calculated as, M = 2 1?+ a . The standard deviation for the initial RTNDT margin term, is F, 0°F when the initial RTNDT is a measured value, and 17'F when a generic value is available. The standard deviation for the ARTNDT margin term, a,, is 17'F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, aA is equal to 28°F when surveillance capsule data is not used, and is 14°F (half the value) when credible surveillance capsule data is used. a, need not exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 8 and 9 are the calculations of the 34 EFPYART values used for generation of the heatup and cooldown curves.

17 TABLE 8 Calculation of the ART Values for the 1/4T Location @ 34 EFPY

.n, , . ( 1.;

Intermediate Shell Forging 05 Position 1.1 117.0 1.03 -4 120.5 34 1511 (Heat # 526840) Position 2.1 84.0 1.03 -4 86.5 17 100 Lower Shell Forging 04 Position 1.1 115.8 1.03 -30 119.3 34 123 (Heat # 411337/11) 1 Intermediate to Lower Shell Position 1.1 52.7 1.03 -68 54.3 54.3 41 Circumferential Weld Seam W05 Position 2.1 31.5 1.03 -68 32.4 28 -8 Notes (a) Initial RTmr values are measured values (b) ART = Initial RTT + ARTmr + Margin (0F)

(c) ARTNT = CF

• FF

18 TABLE 9 Calculation of the ART Values for the 3/4T Location @ 34 EFPY

++++++++++++++~~~~~~ +.......... ... + +

IRG .1-99 FF ART Intermediate Shell Forging 05 Position 1.1 117.0 0.750 -4 87.8 34 118 (Heat # 526840) Position 2.1 84.0 0.750 -4 63 17 76 Lower Shell Forging 04 Position 1.1 115.8 0.750 -30 86.9 34 91 (Heat # 411337/11) ____________________

Intermediate to Lower Shell Position 1.1 52.7 0.750 -68 39.5 39.5 11 Circumferential Weld Seam W05 Position 2.1 31.5 0.750 -68 23.6 23.6 -21 Notes:

(a) Initial RTmT values are measured values.

(b) ART =Initial RTNDT + ARTmT + Margin (TF)

(C) ARTmT = CF

• F

19 The Lower Shell Forging 04 is the limiting beltline material for all the PT limit curves to be generated.

Contained in Table 10 is a summary of the limiting ARTs to be used in the generation of the McGuire Unit 2 reactor vessel PT limit curves. These limiting curves will be presented in Section 5.

TABLE 10 Summary of the Limiting ART Values Used in the Generation of the McGuire Unit 2 Heatup/Cooldown Curves

20 5 HEATUP AND COOLDOWN PRESSURE-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 3.0 and 4.0 of this report.' This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section 1 of this report.

Figures 1 and 2 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60, 80 and 100*F/hr applicable for the first 34 EFPY. These curves were generated using a combination ofthel996 ASME Code Section XI, Appendix G with the limiting ART values and the ASME Code Case N-641. Figure 3 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20,40, 60 and 100°F/hr applicable for 34 EFPY. Again, these curves were generated using a combination ofthel996 ASME Code Section XI, Appendix G with the limiting ART values and the ASME Code Case N-641. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit line shown in Figures 1 through 3. 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 1 and 2. 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 Code Case N-641[31 (approved in February 1999) as follows:

1.5 Kim< Kio where, KfI is the stress intensity factor covered by membrane (pressure) stress, Kir = 33.2 + 20.734 e[°°02 rRr- r)],

T is the minimum permissible metal temperature, and RTNDT 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 5. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40TF higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4.0 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 McGuire Unit 2 reactor vessel at 34 EFPY is 184T. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve constitutes the limit for core operation for the reactor vessel.

21 Figures 1 through 3 define all of the above limits for ensuring prevention of nonductile failure for the McGuire Unit 2 reactor vessel for 34 EFPY. The data points used for the heatup and cooldown pressure temperature limit curves shown in Figures 1 through 3 are presented in Tables 11, 12 and 13.

22 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 1230F 3/4T, 91OF 2500 2250 2000 1750 S1500 S1250 3 1000 U

Lu C-750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 1 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60 0 F/hr)

Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

23 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123 0F 3/4T, 91°F 2500 rVersion:5 I Run 30646 Operim 2250 Leak Test Limit 2000 Unacceptable Operation 1750 U Heatup Rate C S1500 0Deg. FIHrr l S1250 1250 --.. Heatup Rate * -

I. 100 Deg. F/Hr I

1000 500 250 .T-..

0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

24 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMrITNG ART VALUES AT 34 EFPY: l/4T,1230 F 3/4T, 91°F 2500 2250 2000 1750 I 1500 2

S1250 Cc II

= 1000 0

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 3 McGuire Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1001F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

25 TABLE 11 34 EFPY, 60°F/hr Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 7;ti<16ftpIn. "i.t" Ie1 CItialt Tes Lim)it Teinp: Pes'.uTRp <rs. ep Pes 60 0 184 0 167 2000 60 621 184 621 184 2485 65 621 184 621 70 621 184 621 75 621 184 621 80 621 184 621 85 621 184 621 90 621 184 621 95 621 184 621 100 621 184 621 105 621 184 621 110 621 184 621 115 621 184 621 120 621 184 621 121 621 184 932 121 932 184 962 125 962 184 1004 130 1004 184 1051 135 1051 184 1103 140 1103 185 1161 145 1161 190 1226 150 1226 195 1296 155 1296 200 1353 160 1353 205 1416 165 1416 210 1486 170 1486 215 1562 175 1562 220 1646 180 1646 225 1739 185 1739 230 1842 190 1842 235 1955 195 1955 240 2080 200 2080 245 2218 205 2218 250 2370 210 2370 1

26 TABLE 12 34 EFPY, 80 and 100°F/hr Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

Temp "'rs.n~i& mp.l press Tcnp. nss Tefmp. 1'?- Preiss:" "ep Pý 60 0 184 0 60 0 184 0 167 2000 60 621 184 621 60 621 184 621 184 2485 65 621 184 621 65 621 184 621 1 70 621 184 621 70 621 184 621 75 621 184 621 75 621 184 621 80 621 184 621 80 621 184 621 85 621 184 621 85 621 184 621 90 621 184 621 90 621 184 621 95 621 184 621 95 621 184 621 100 621 184 621 100 621 184 621 105 621 184 621 105 621 184 621 110 621 184 621 110 621 184 621 115 621 184 621 115 621 184 621 120 621 184 621 120 621 184 621 121 621 184 859 121 621 184 805 121 859 184 881 121 805 184 822 125 881 184 913 125 822 184 846 130 913 184 950 130 846 184 874 135 950 184 991 135 874 184 906 140 991 185 1037 140 906 185 943 145 1037 190 1089 145 943 190 984 150 1089 195 1146 150 984 195 1030 155 1146 200 1210 155 1030 200 1081 160 1210 205 1281 160 1081 205 1139 165 1281 210 1360 165 1139 210 1203 170 1360 215 1447 170 1203 215 1273 175 1447 220 1544 175 1273 220 1352 180 1544 225 1650 180 1352 225 1439 185 1650 230 1769 185 1439 230 1536 190 1769 235 1899 190 1536 235 1642 195 1899 240 2017 195 1642 240 1760 200 2017 245 2143 200 1760 245 1890 205 2143 250 2282 205 1890 250 2034 210 2282 255 2435 210 2034 255 2193 215 2435 215 2193 260 2368 220 2368

27 TABLE 13 34 EFPY Cooldown Curve Data Points Using 1996 App. G &ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) fl

• T ~ 1 (psg) P(pu) T ~) ~jiug) ( A P (Osig).

60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 591 60 512 65 621 65 621 65 621 65 605 65 530 70 621 70 621 70 621 70 621 70 549 75 621 75 621 75 621 75 621 75 570 80 621 80 621 80 621 80 621 80 593 85 621 85 621 85 621 85 621 85 620 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 121 621 121 621 121 621 121 621 121 621 121 963 121 947 121 934 121 924 121 918 125 993 125 980 125 970 125 963 130 1034 130 1025 130 1019 130 1018 135 1080 135 1075 135 1074 140 1130 145 1185 150 1247 155 1315 160 1390 165 1473 170 1564 175 1665 180 1777 185 1901 190 2037 195 2188 200 2355

28 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 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

3. ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1", January 17, 2000.

[Sub Reference 1: ASME Code Case N-640, "AlternativeReference FractureToughnessfor Development ofP-T Limit Curvesfor Section X, Division 1 ",February26, 1999.]

[Sub Reference 2: ASME Boiler andPressureVessel Code, Case N-588, "'Attenuationto Reference Flaw OrientationofAppendix Gfor Circumferential Welds in Reactor Vessels"'Section XI, Division 1, Approved December 12, 1997.]

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix (; "Fracture Toughness Criteria for Protection Against Failure." Dated December 1995, through 1996 Addendum.

5. Code of Federal Regulations, 10 CFR Part 50, Appendix Q4 "Fracture Toughness Requirements,"

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

6. "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.

7. WCAP-15334, "Duke Power Company Reactor Cavity Neutron Measurement Program for William B McGuire Unit 2 Cycle 12,"A.H. Fero, November 1999.

8. WCAP-14799, "Analysis of Capsule W From The Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program," E. Terek, et. al., March 1997.

9. 1989 Section mH, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."

10. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

11. WCAP-15201, Revision 1, "McGuire Unit 2 Heatup and Cooldown Curves for Normal Operation Using Code Case N-640," J. H. Ledger, April 2001.

ENCLOSURE 6 ASME CODE CASE N-641

CASE N-641 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: January 17, 2000 See Numeric Index for expiration and any reaffirmation dates Case N-641 throughout the life of the component at each temperature Alternative Pressure-Temperature Relationship with KIm from G-2214.1, K1, from G-2214.3, and Kc and Low Temperature Overpressure Protection from Fig. G-2210-1.

System Requirements The allowable pressure at any temperature shall be Section XI, Division I determined as follows.

(a) For the startup condition, Inquiry: What alternatives to Appendix G-2215 may (1) consider postulated defects in accordance with be used for determination of pressure-temperature rela G-2120; tionships and low temperature overpressure protection (2) perform calculations for thermal stress intensity system effective temperatures and allowable pressures? factors due to the specified range of heat-up rates from G-2214.3; Reply: It is the opinion of the Comnmittee that, as (3) calculate the K1c toughness for all vessel beltline an alternative to Appendix G-2215, the following may materials from G-2212 using temperatures and RTNDT be used. values for the corresponding locations of interest; and (4) calculate the pressure as a function of coolant inlet temperature for each material and location. The

-1000 INTRODUCTION allowable pressure-temperature relationship is the mini mum pressure at any temperature determined from

-1100 Scope (a) the calculated steady-state (K,, = 0) results This Case presents alternative procedures for calculat for the 1/4 thickness inside surface postulated defects ing pressure-temperature relationships and low tempera using the equation:

ture overpressure protection (LTOP) system effective temperatures and allowable pressures. These procedures take into account alternative fracture toughness proper ties, circumferential and axial reference flaws, and plant specific LTOP effective temperature calculations.

-2215 Allowable Pressure (b) the calculated results from all vessel beltline

-2215.1 Pressure-Temperature Relationship. The materials for the heatup stress intensity factors using the equations below provide the basis for determination corresponding 1/4 thickness outside-surface postulated of the allowable pressure at any temperature at the defects and the equation:

depth of the postulated defect during Service Condi tions for which Level A and Level B Service Limits are specified. In addition to the conservatism of these assumptions, it is recommended that a factor of 2 be 2M,, /,

applied to the calculated K, values produced by pri mary stresses. In shell and head regions remote from discontinuities, the only significant loadings are: (I) (b) For the cooldown condition; general primary membrane stress due to pressure; and (1) consider postulated defects in accordance with (2) thermal stress due to thermal gradient through the G-2120; thickness during startup and shutdown. Therefore, the (2) perform calculations for thermal stress intensity requirement to be satisfied and from which the allow factors due to the specified range of cooldown rates able pressure for any assumed rate of temperature from G-2214.3, change can be determined is: (3) calculate the K1, toughness for all vessel beitline materials from G-2212 using temperatures and RTNvDT 2KM,, + Kir < KI, (1) values for the corresponding location of interest; and Illl SUPP. 8 - NC

CASE (continued)

N-641 CASES OF ASME BOILER AND PRESSURE VESSEL CODE (4) calculate the pressure as a function of coolant (2) a coolant temperaturel corresponding to a reac inlet temperature for each material and location using tor vessel metal temperature2 , for all vessel beltline the equation: materials, where T, is defined for inside axial surface flaws as RTNDT + 40'F, and T, is defined for inside circumferential surface flaws as RTNDT - 85°F; (3) a coolant temperature' corresponding to a reac tor vessel metal temperature 2, for all vessel beltline 2Mm R materials, where T, is calculated on a plant specific basis for the axial and circumferential reference flaws using the following equation:

The allowable pressure-temperature relationship is T, = RTvDT+ 50 In [((F. M, (pR, I t)) - 33.2) / 20.734]

the minimum pressure at any temperature, determined from all vessel beltline materials for the cooldown where stress intensity factors using the corresponding 1/4 thick ness inside-surface postulated defects. F=1.1, accumulation factor for safety relief valves

-2215.2 Low Temperature Overpressure Protection M,. =the value of M,, determined in accordance System. Plants having LTOP systems may use the with G-2214.1 following temperature and pressure conditions to pro p =vessel design pressure, ksi vide protection against failure during reactor startup R,= vessel inner radius, in and shutdown operation due to low temperature over t =vessel wall thickness, in.

pressure events that have been classified Service Level A or B. (b) LTOP System Allowable Pressure.LTOP systems (a) LTOP System Effective Temperature. The LTOP shall limit the maximum pressure in the vessel to 100%

system effective temperature T, is the temperature at of the pressure determined to satisfy Eq. (1) if K1, is or above which the safety relief valves provide adequate used for determination of allowable pressure, or 110%

protection against nonductile failure. LTOP systems of the pressure determined to satisfy Eq. (1) if Ki, is shall be effective below the higher temperature deter used (as an alternative to K1,) for determination of mined in accordance with (1) and (2) below. Alterna allowable pressure.

tively, LTOP systems shall be effective below the 2 The vessel metal temperature is the temperature at a distance one higher temperature determined in accordance with (1) fourth of the vessel section thickness from the clad-base-metal and (3) below. interface in the vessel beltline region RTNmt is the highest adjusted (1) a coolant temperature' of 2001F; reference temperature, for weld or base metal in the belline region, at a distance one-fourth of the vessel section thickness from the The coolant temperature is the reactor coolant inlet temperature clad-base-metal interface as determined in accordance with Regulatory Guide 1 99, Rev 2.

SUPP. 8 - NC 1112

w ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

ESFAS Instrumentation 3.3.2 Table 3.3 2-1 (page 3 of 6)

Engineered Safety Feature Actuation System Instnrmentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Steam Lne Isolation (contnued)

<s 120(d) psi 100(d) psi I (2) Negative 3 (b)(c) 3 per steam D SR 3.3.2-1 fine SR 3.3Z-.5 Rate - High SR 3.32.8 SR 332.9

5. Turbine Tnp and Feedwater Isolation 2 trahis SR 3.3.2.2 NA NA

a. Automatic SR 3.32.4 Actuation Logic and Actuation SR 3.32.6 Relays SR 3.32.1 < 85.6% 83.9%

b. SG Water Level 3 per SG SR 3.32.2 High High (P-14) SR 3.32.4 SR 3.3.2.5 SR 3.3.2.6 SR 3.32.8 SR 3.32-9

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

553F I

d. T-1-Low (e1 per loop j SR 3.3Z.1 >551*F SR 3.3.2.5 SR 3.32.8 coincident wit Reerto Function 8.a (ReactDr Trip. P-4) for al initiabon functions and Reactor Trip. P.4 requiements.

L.M SR 3.3.2.1 S.13 inches 12 inches I

e. Doghouse Water 1.2(0) 2 per pertrgn SR 3.3.2.7 LevelHigh High Doghos

6. AuxITy Feedwater 123 2 takins H SR 3.232 NA NA

a. Autmatlc Achinton Logi SR 3.324 SR 3.32.6 and Achgeion

_ 15% 16.7%

1,23 4 per SG D SR 3.32.1

b. SG Water Level SR 3.32.5 Low Low SR 3.3.2.8 SR 3.32.9 (continued)

(b) Except when all MSIVs are closed and do-acbvated. Iw er;)

(C) T

• it_Won automatically

-.Steam blocked above P-1I (Pressurizer Pressure) interlock and may be blocked below P-11v is not blocked.

(-- -C) Line Pressure-Low

{d -e utilizd in the rate/lag conlroller is > 50 seconds.

manual (e) Except when all MFIVs. MFCVs, and associated bypass valves are dosed and de-actvated or isolated by a cosed valve McGuire Units 1 and 2 3.3.2-12 Amendment Nos. q

• L5,.,4

RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PrT) Limits LCO 3.4.3 RCS pressure and RCS temperature shall be limited during RCS heatup and cooldown, criticality, and inservice leak and hydrostatic testing in accordance with:

a. A maximum heatup rate as specified in Figure 3.4.3-i_

b. A maximum cooldown rate as specified in

c. A maximum temperature change of < 1 0°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during inservice leak and hydrostatic testing operations above the heatup and cooldown limit curves.

Ar PPICAAI-TA al 3t APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE--------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 within limits.

shall be completed whenever this Condition AND is entered.

---

A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> acceptable for continued Requirements of LCO operation.

not met in MODE 1, 2, 3, or 4.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 with RCS 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> pressure < 500 psig.

(continued)

McGuire Units 1 and 2 3.4.3-1 Amendment Nos. E

-ii;- RCS P/T Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. --------- NOTE ----------- C.1 Initiate action to restore Immediately Required Action C.2 parameter(s) to within shall be completed limits.

whenever this Condition is entered. AND C.2 Determine RCS is Prior to entering Requirements of LCO acceptable for continued MODE 4 not met any time in other operation.

than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 -------------------- NOTE -----------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS heatup 30 minutes and cooldown rates are within limits.

McGuire Units 1 and 2 3.4.3-2 Amendment Nos. 184/166

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIALS; LOWER SHELL LOWER SHELL PLATE B5013-2 -

LIMITING ART AT 16 EFPY 1/4t, 149.5 3/4-t,102.0 ACCEPTAB

- OPE N CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC TEST TEMPERATURE (282 F)

FOR THE SERVICE PERIOD UP TO 16 EFPY 0 100 200 Figure 3.4.3-1 RCS Heatup Limitatdn-s (UNIT 1 ONLY)

(Without margins for Instru tation Errors)

NRC REG GUID .99, Rev. 2 Applicable f e first 16 EFPY Aepb-t-eJ VL.Vý 4wo 4-vný McGuire Units 1 and 2 Výý Amendment Nos.kiý

Z 22!

f f f I l 7 I e 7 e/'

MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202°F 3/4T, 146-F 2500 2250 2000 1750 L 1500 o2 QZ E 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 (j ýo--at rTermperature (Deg. F)

F c Figureo McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 6O0F/hr)

Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

f 7

1. 1.Z

/

4r / ýý=23 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202-F 3/4T, 146-F 2500 2250 2000 1750 co C 1500 9! 1250 1000 750 500 250 Criticality Limit based on Inservice hydrostatic test temperature (262 F) for the service period up to 34 EFPY 0

0 50 100 150 200 250 300 350 400 450 500 550

--- (Deg. F)Fature Figure1o McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIALS; LOWER SHELL FORGING1 0 LIMITING ART AT 16 EFPY /

%-t, 104 deg. F /

3/-t, 102 0 deg. F #

- -0 I II II LEAK TEST LIMI 2250 AI I II

- LL I L I I I lItII L .

i~~~~~~~~~

Iiill ii l lii -

2000 iI II I IL L l Y

• Il JI l l

• I ~ I I -

l I

Ii Hiii I Ii Il I-IH -l Il I I i l l

I 0 1750 I IMIT,-= -T I I (

S1500 V :::: EI I.

a) C EP UN I OPERATIONIi "1250 ONSPVCE I I Z YDOSTTI 0

I lie SOPSER ATIEHDOSTATI rnM S-"-+" --FOR THE SERVICE PERIOD UP

"*"+'+ -TO 16 EFPY 250  :  : : II II II Bellline Region Fluid Temperraturre (Dee*"

Figure 3.4.3-1 RCS HeatupLi i ons (Unit 20Only*.,

(Without margins for In mnainErrors)

NRC REG 1.99, Rev. 2 Aplc for the First 16 EFPY McGuire Units 1 and 2 V Amendment Nos.

MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123 0F 3/4T, 91OF 2500 loperlim Version 5 1 Run 30648 2250 LeakTest LmIt-2000 1- 1750 tL C

[ 1500

-j f 1250 q- 1:000

(* 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 M der q Temperature (Deg. F Figurek McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G, Methodology & ASME Code Case N-641

Z-I2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123'F 3/4T, 91'F 2500250 *I operfimn Version 51 Run-30646 2250 Le-akTest Limit 2000 Unacceptable 0Operation 1750

~g. 1500 0SDeg. FIHr C5 025 1000 0 eg I~

~ 750 2500 0 50 100 150 200 250 300 350 400 450 500 550 ode r Temperature (Deg. F)

Figure4 7 j McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

-- 100*F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

° RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIALS; LOWER SHELL LONGITUDINALWELDS 3-44:2A and LOWER SHELL PLATE B5013-2 LIMITING ART AT 16 EFPY 1

/-t, 149 5 deg. F 3/4-t, 1020 deg. F 2500" I y UNACCEPABLECEP 2250/OPERTION I i -ER)

• TABLE ATION 50040 F/R

.25 "5 .P/-O/Eor 210000 1ý

-- -0 FH oodw00n Rats u

_'0 Fiur 0FH 3..32 CSColdwnLiittin 20 30 I, 400 500 jI

-100ow Rates up to10dg /

eg. F)

(Without margins for Instrumentation Errors)

NRC REG GUIDE 1.99, Rev. 2 Applicable for the first 16 EFP 4q ~JC-- 4-re McGuire Units 1 and 2 (99) Amendment Nos. qEý)

/

I 4~

T2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T,202°F 3/41".146 0F 2500 2250 2000 J 1750

,1500

_ 1250 1000 7-I 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550

(-5 ggRDTemperature

• ~0 -' ID /D ,

(Deg. F)

. -"/.

T--r 3 . it 3 "* ,- "- l ,"--e, -,* "9&"-

Figurel&) McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIALS; LOWER SHELLFORGING 04 LIMITING ART AT 16 EFPY:

%-t, 104 deg. F 3/4-t, 73 deg. F 2500 2250 1750 CD 1500 CM L

C 1250 o=

C, 1000 0J 750

(.-

0 50 100 150 200 250 00 350 4C)0 450 Reactor Beltline R on Fulud Temperature (DEG. F)

Figure 3.4.3-2 Cooldown Limitations ohcut Margins for Instrumentation Errors)

Applicable NRC REG for GUIDE 1.99,16Rev.

the First EFPY 2

&fleltej 1A L--ý McGuire Units 1 and 2 Amendment Nos.OEY)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: l/4T,123°F 3/4T, 91-F 2500 [O*perm Version 5.1 Run3o646 2250 2000 2000

-- Unacceptable

-Fln-accrntinn d

1750 CL. 1500 I 1250 S1000 SJ750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 3der oTemperature (Deg. F)

Figure@j McGuire Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100OF/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641

F,--ý, 0-t-IT LTOP System 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of one centrifugal charging pump or one safety injection pump capable of injecting into the RCS and the accumulators isolated and either a or b below:

a. Two power operated relief valves (PORVs) with lift setting _ 385 psig or

b. The RCS depressurized and an RCS vent of > 2.75 square inches.

......------------------ NOTE -----------------....-------------------

A PORV secured in the open position may be used to meet the RCS vent requirement provided that its associated block valve is open and power removed.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is < 300'F, MODE 5, MODE 6 when the reactor vessel head is on.

S-- --------- \II J r - -- -.----.-. ----------.-.----.-. --.-.----.-.--

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in Specification 3.4.3.

McGuire Units 1 and 2 3.4.12-1 Amendment Nos. 184/166

LTOP System 3.4.12 ACTIONS

- -.-.-.-------------------.-. --. -.---.-.----------.-.---- .-NOT "E--------. -.- .---------------- -.

LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION J COMPLETION TIME A. Two centrifugal charging A.1 NOTE------ ---------

pumps capable of Two centrifugal charging injecting into the RCS. pumps may be capable of injecting into the RCS OR during pump swap operation for < 15 minutes.

One centrifugal charging pump and one safety Initiate action to verify a Immediately injection pump capable maximum of one centrifugal of injecting into the RCS. charging pump or one safety injection pump is OR capable of injecting into the RCS.

Two safety injection pumps capable of OR injecting into the RCS.

A.2.1 Verify RHR suction relief Immediately valve is OPERABLE and the suction isolation valves are open.

AND A.2.2.1 Verify RCS cold leg temperatureo.. Immediately OR >Lq? F (~y~

A.2.2.2Verify RCS cold le Immediately tem erature> 1070 and coo own te <_OF/hr/' < 0~ °F/*,. (&.*-z), or _

y>T 7 ý0 A(nud)A CO 4t OR (continued)

_____________________________ I _______________________________ I McGuire Units 1 and 2 3.4.12-2 Amendment Nos.9A)

LTOP System 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Verify two-PORVs secured Immediately open and associated block valves open and power removed.

OR A.4 Depressurize RCS and Immediately establish RCS vent of

> 4.5 square inches.

B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in Specification 3.4.3.

C. Required Action and C.1 Increase RCS cold leg 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion temperature to > 300°F.

Time of Condition B not met. OR C.2 Depressurize affected 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed by Specification 3.4.3.

/ D. One PORV inoperable in D.1 Restore PORV to 7 days MODE 4. OPERABLE status.

- V________________

ed) e:111_° A .5 2 JOerý'o -POPV, McGuire Units 1 and 2 3.4.12-3 Amendment Nos. 9ýD

LTOP System 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One PORV inoperable in E.1 Suspend all operations Immediately MODE 5 or 6. which could lead to a water solid pressurizer.

AND E.2 Restore PORV to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

F. Required Action and F.1. Verify RCS cold lea 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion temperaturi .

Time of Condition E not 1>174 Ofu/..t/) "Y met. AND F.2 Verify RHR suction relief 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> valve is OPERABLE and the suction isolation valves are open.

4 I-G. Two PORVs inoperable. I G.1 Depressurize RCS and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> establish RCS vent of OR > 2.75 square inches.

Required Action and associated Completion Time of Condition C, D, E, or F not met.

OR LTOP System inoperable for any reason other than Condition A, B, C, D, E, or F.

McGuire Units 1 and 2 3.4.12-4 Amendment Nos. 99)

ATTACHMENT 2 CHANGES TO TS BASES PAGES (mark-up)

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Manual and automatic initiation of steam line isolation must be OPERABLE in MODES 1, 2, and 3 when there is sufficient energy in the RCS and SGs to have an SLB or other accident. This could result in the release of significant quantities of energy and cause a cooldown of the primary system. The Steam Line Isolation Function is required in MODES 2 and 3 unless all MSIVs are closed and de-activated. In MODES 4, 5, and 6, there is insufficient energy in the RCS and SGs to experience an SLB or other accident releasing significant quantities of energy.

c. Steam Line Isolation-Containment Pressure-High High This Function actuates closure of the MSIVs in the event of a LOCA or an SLB inside containment to maintain three unfaulted SGs as a heat sink for the reactor, and to limit the mass and energy release to containment. The Containment Pressure - High High function is described in ESFAS Function 2.C.

Containment Pressure-High High must be OPERABLE in MODES 1, 2, and 3, when there is sufficient energy in the primary and secondary side to pressurize the containment following a pipe break. This would cause a significant increase in the containment pressure, thus allowing detection and closure of the MSIVs. The Steam Line Isolation Function remains OPERABLE in MODES 2 and 3 unless all MSIVs are closed and de-activated. In MODES 4, 5, and 6, there is not enough energy in the primary and secondary sides to pressurize the containment to the Containment Pressure High High setpoint.

d. Steam Line Isolation-Steam Line Pressure (1) Steam Line Pressure-Low Steam Line Pressure-Low provides closure of the MSIVs in the event of an SLB to maintain three unfaulted SGs as a heat sink for the reactor, and to limit the mass and energy release to containment.

This Function provides closure of the MSIVs in the event of a feed line break to ensure a supply of steam for the turbine driven AFW pump.

McGuire Units 1 and 2 B 3.3.2-16 Revision No.

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Steam Line Pressure-Low Function must be OPERABLE in MODES 1, 2, and 3 (above P-1 1), with any main steam valve open, when a secondary side break or stuck open valve could result in the rapid depressurization of the steam lines.

This signal may be manually blocked by the operator below the P-1 1 setpoint. Below P-1 1, an inside containment SLB will be terminated by automatic actuation via Containment Pressure-High High. Stuck valve transients and outside containment SLBs will be terminated by the Steam Line Pressure-Negative ate-High signal for Steam Line Isolation T

PUwf pes~- Low below P-11 when as been manually blocked. The Steam Line Isolation Fu ction is required in MODES 2 and 3 unless all MSIVs are closed and de-activated. This Function is not required to be OPERABLE in MODES 4, 5, and 6 because there is insufficient energy in the secondary side of the unit to have an accident.

(2) Steam Line Pressure-Negative Rate-High Steam Line Pressure-Negative Rate-High provides closure of the MSIVs for an SLB when less than the P-1 1 setpoint, to maintain at least one unfaulted SG as a heat sink for the reactor, and to limit the mass and energy release to containment. When the operator manually blocks the Steam Line Pressure-Low main steam isolation signal when less than the P-1 1 setpoint, the Steam Line Pressure-Negative Rate High signal is automatically enabled. Steam Line Pressure Negative Rate-High provides no input to any control functions.

Thus, three OPERABLE channels are sufficient to satisfy requirements with a two-out-of-three logic on each steam line.

Steam Line Pressure-Negative Rate-High must be OPERABLE in MODE 3 when less than the P-1 1 setpoint, when a secondary side break or stuck open valve could result in the rapid depressurization of the steam line(s). In MODES 1 and 2, and in MODE 3, when above the P-11 setpoint, this signal is automatically disabled and the Steam Line Pressure-Low signal is automatically enabled. The Steam Uine Isolation Function is required to be OPERABLE in McGuire Units 1 and 2 B 3.3.2-17 Revision No.

RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This Specification contains P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G (Ref. 2).

The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases.

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and McGuire Units 1 and 2 B 3.4.3-1 Revision No. 0

RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the fi;; Erecommendations of Regulatory Guide 1.99 (Ref. 5).

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limit curve includes the Reference 1 requirement that it be

>_ 40°F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

The ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)

SAFETY ANALYSES analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Although the P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36 (Ref. 7).

McGuire Units 1 and 2 B 3.4.3-2 Revision NO./

Insert 1:

A second program that employs excore cavity dosimetry to monitor the reactor vessel neutron fluence has been installed in both units. This program meets the requirements of 10 CFR 50 Appendix H (Ref. 4).

LTOP System B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the specified limit. The setpoints are derived by analyses that model the performance of the LTOP System, assuming the limiting LTOP transient of one centrifugal charging pump or one safety injection pump injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.

The PORV setpoints will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are (determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, 'RCS Pressure and Temperature (P/T) Limits," discuss these examinations.

The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

RCS Vent Performance With the RCS depressurized, analyses show a vent size of 2.75 square inches is capable of mitigating the allowed LTOP overpressure transient.

The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, one centrifugal charging pump or one safety injection pump OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance.

The RCS vent is passive and is not subject to active failure.

The LTOP System satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).

McGuire Units 1 and 2 B 3.4.12-5 Revision No. 30

LTOP System B 3.4.12 BASES LCO This LCO requires that the LTOP System is OPERABLE. The LTOP System is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO permits a maximum of one centrifugal charging pump or one safety injection pump capable of injecting into the RCS and requires all accumulator discharge isolation valves closed and immobilized when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in LCO 3.4.3.

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:

a. Two OPERABLE PORVs (NC-32B and NC-34A); or A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the specified limit and testing proves its automatic ability to open at this setpoint, and motive power is available to the valve and its control circuit.

b. A depressurized RCS and an RCS vent.

An RCS vent is OPERABLE when open with an area of

> 2.75 square inches.

]er4 K Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

APPLICABILITY This LCO is applicable in MODE 4 when any RCS cold leg temperature is

< 3000 F, in MODE 5, and in MODE 6 when the reactor vessel head is on.

The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 3000F. When the reactor vessel head is off, overpressurization cannot occur.

LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves,n requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above 3000F.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

McGuire Units 1 and 2 B 3.4.12-6 Revision No. 30

Insert 2:

The LCO is modified with a note that specifies that a PORV secured in the open position may be used to meet the RCS vent requirement provided that its associated block valve is open and power removed. With the PORV physically secured or locked in the open position with its associated block valve open and power removed, this vent path is passive and is not subject to active failure.

LTOP System B 3.4.12 BASES APPLICABILITY (continued)

The Applicability is modified by a Note stating that accumulator isolation is only required when the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves. This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

ACTIONS LCO 3.0.4 is not applicable for entry into LTOP operation.

A.1,* A.2.1, A.2.2.1 . A.2.2.2., A.3, A.4 ) ,si 1aýA.')

With two centrifugal charging pumps, safety injection pumps, or a combination of each, capable of injecting into the RCS, RCS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

Two pumps may be capable of injectinp into the RCS provided the RHR suction relief valve is.OPERABLE witi-theFC cold leg terr~erature >

(70; or> 10° and codown rate limi d to 20°F, or if *o PORVs ae)

• l se~cured open with theassociated bloc valves open ant'power removr, Sf/.*with an RCS ve,ltof A. sr-quare. in *s.-For cases -whFereno reactor" coolant pumps are in operation, RCS cold leg temperature limits are to be J".rV*

L 3./ met by monitoring of BOTH the WR Cold Leg temperatures and Residual Heat Removal Heat Exchanger discharge temperature. With both PORVS and block valves secured open, or with an RCS vent of 4.5 square inches, there are no credible single failures to limit the flow relief capacity. For the RHR relief valve to be OPERABLE, the RHR suction isolation valves must be open and the relief valve setpoint at 450 psig consistent with the safety analysis. The RHR suction relief valves are spring loaded, bellows type water relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves.

Required Action A.1 is modified by a Note that permits two centrifugal charging pumps capable of RCS injection for < 15 minutes to allow for pump swaps.

B.1, C.1, and C.2 An unisolated accumulator requires isolation within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This is only required when the accumulator pressure is at or more than the maximum McGuire Units 1 and 2 B 3.4.12-7 Revision No.

Insert 3:

1. RCS cold leg temperature > 174 OF (Unit 1), or

2. RCS cold leg temperature > 89 OF (Unit 2), or

3. RCS cold leg temperature > 74 OF and cooldown rate < 20 OF

/hr (Unit 1), or

4. RCS cold leg temperature > 74 °F and cooldown rate < 60 OF

/hr (Unit 2), or

5. two PORVs secured open with associated block valves open and power removed, or

6. a RCS vent of > 4.5 square inches, or

7. a RCS vent of > 2.75 square inches and two OPERABLE PORVs (the RCS vent shall not be one of the two OPERABLE PORVs).

LTOP System B 3.4.12 BASES ACTIONS (continue,d) )Ih* rU-4' r F.1 and F.2 Ž-- q u )

Ifthe Required Actions and associated Completion Times of Condition E are not met, then alternative actions are necessary to establish the required redundancy in relief capacity. This is accomplished by verifying that the RHR relief valve is OPERABLE and the RHR suction isolatio valves open and the RCS cold leg temperature 166For cases where no reactor coolant pumps are in operation, R cold leg temperature limits are to be met by monitoring of BOTH the WR Cold Leg temperatures and Residual Heat Removal Heat Exchanger discharge temperature. The -D 0 Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reflects the importance of restoring the required redundancy at lower RCS temperatures.

G.1 The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when:

a. Both required PORVs are inoperable; or

b. A Required Action and associated Completion Time of Condition C, D, E, or F is not met; or

c. The LTOP System is inoperable for any reason other than Condition A, B, C, D, E, or F.

The vent must be sized >2.75 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

SURVEILLANCE SR 3.4.12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all but one centrifugal charging pump or McGuire Units 1 and 2 B 3.4.12-9 Revision No.o

ATTACHMENT 3 PROPOSED TECHNI :CAL SPECIFICATION PAGES (retyped)

ESFAS Instrumentation 3.3.2 Table 33.2-1 (page 3 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Steam Line Isolation (continued)

(2) Negative 3 (b)(c) 3 per steam D SR 33.2.1 <_ 120(d) psi 10 0 (d) psi Rate - High line SR 3325 SR 3328 SR 3329

5. Turbine Trip and Feedwater Isolation

a. Automatic 2 trains I SR 3.322 NA NA Actuation Logic SR 3.324 and Actuation SR 3.32.6 Relays

b. SG Water Level 3 per SG J SR 3.32.1 < 856% 83.9%

High High (P-14) SR 332.2 SR 332.4 SR 332.5 SR 332.6 SR 332.8 SR 332.9 c Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements d Tavg-Low 1,2(e) 1 per loop J SR 3 3.2.1 > 551*F 553°F SR 33.2.5 SR 3328 coincident with Refer to Function 8.a (Reactor Trip, P-4) for all initiation functions and Reactor Trip, P-4 requirements.

e Doghouse Water 2 per train L,M SR 3.32 1 <13 inches 12 inches 1 ,2 (e)

Level-High High per SR 3.3 27 Doghouse 6 Auxiliary Feedwater a Automatic 1,2,3 2 trains H SR 332.2 NA NA Actuation Logic SR 332.4 and Actuation SR 332.6 Relays 1,2,3 4 per SG D SR 33.2.1 > 15% 16.7%

b SG Water Level Low Low SR 33.2.5 SR 33.2.8 SR 33.29 (continued)

(b) Except when all MSIVs are closed and de-activated (c) Trip function automatically blocked above P-1I (Pressurizer Pressure) interlock and may be blocked below P-11 when Steam Line Isolation on Steam Line Pressure-Low is not blocked (d) Time constant utilized in the rate/lag controller is > 50 seconds (e) Except when all MFIVs, MFCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

McGuire Units 1 and 2 3.3.2-12 Amendment Nos.

RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.3 RCS pressure and RCS temperature shall be limited during RCS heatup and cooldown, criticality, and inservice leak and hydrostatic testing in accordance with:

a. A maximum heatup rate as specified in Figure 3.4.3-1, Figure 3.4.3-2, Figure 3.4.3-3, or Figure 3.4.3-4;

b. A maximum cooldown rate as specified in Figure 3.4.3-5 or Figure 3.4.3-6; and

c. A maximum temperature change of < 1 0°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during inservice leak and hydrostatic testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE-------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 within limits.

shall be completed whenever this Condition AND is entered.

A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> acceptable for continued Requirements of LCO operation.

not met in MODE 1,2, 3, or4.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 with RCS 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> pressure < 500 psig.

(continued)

McGuire Units 1 and 2 3.4.3-1 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202°F 3/4T, 146°F 2500 2250 2000 1750 1500 75 1250 1

1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beitline Reaion Fluid Temperature (Dea. F)

Figure 3.4.3-1 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-3 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 1/4T, 202°F 314T, 146°F 2500 2250 2000 1750 500 W

LA

1. 250 C 000 750 C..

500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beltline Region Fluid Temperature (Deg. F)

Figure 3.4.3-2 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrunientation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-4 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123 0F 314T, 91OF 2500 2250 2000 1750 1500 r

1250 1000 0

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beltline Region Fluid Temperature (Deg. F)

Figure 3.4.3-3 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60'F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-5 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123 0F 3/4T, 91OF 2500 2250 2000 1750 ci:

1500 I

1250 C

1000 750 0

500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beltline Region Fluid Temperature (Deg. F)

Figure 3.4.3-4 McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 &

100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-6 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL LONGITUDINAL WELD LIMITING ART VALUES AT 34 EFPY: 114T, 202°F 3/4T, 146°F 2500 2250 2000 W: 1750 I

1500 0,

1250 1000 0=

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beltline Region Fluid Temperature (Deg. F)

Figure 3.4.3-5 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/lir) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-7 Amendment Nos.

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 34 EFPY: 1/4T, 123 0F 3/4T, 91OF 2500 2250 2000 1750 1500 r3 A3 1250

0. 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Reactor Beitline Region Fluid Temperature (Deg. F)

Figure 3.4.3-6 McGuire Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App.G Methodology & ASME Code Case N-641 McGuire Units 1 and 2 3.4.3-8 Amendment Nos.

LTOP System 3.4.12 ACTIONS I--...........................................................

LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Two centrifugal charging A.1 -------------------- NOTE-------

pumps capable of Two centrifugal charging injecting into the RCS. pumps may be capable of injecting into the RCS OR during pump swap operation for < 15 minutes.

One centrifugal charging pump and one safety Initiate action to verify a Immediately injection pump capable maximum of one centrifugal of injecting into the RCS. charging pump or one safety injection pump is OR capable of injecting into the RCS.

Two safety injection pumps capable of OR injecting into the RCS.

A.2.1 Verify RHR suction relief Immediately valve is OPERABLE and the suction isolation valves are open.

AND A.2.2.1 Verify RCS cold leg Immediately temperature > 174 0F (Unit

1) or > 89°F (Unit 2).

OR A.2.2.2Verify RCS cold leg Immediately temperature > 74 0F and cooldown rate < 20'F/ hr (Unit 1), or > 74°F and cooldown rate < 60°F/hr (Unit 2).

OR (continued)

______________________________ .1 McGuire Units 1 and 2 3.4.12-2 Amendment Nos.

LTOP System 3.4.12 ACTIflN5 (ontinued' CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Verify two PORVs secured Immediately open and associated block valves open and power removed.

OR A.4 Depressurize RCS and Immediately establish RCS vent of > 4.5 square inches.

OR A.5.1 Depressurize RCS and Immediately establish RCS vent of

> 2.75 square inches.

AND A.5.2 Verify two PORVs are Immediately OPERABLE.

B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in Specification 3.4.3.

C. Required Action and C.1 Increase RCS cold leg 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion temperature to > 300 0 F.

Time of Condition B not met. OR C.2 Depressurize affected 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed by Specification 3.4.3.

D. One PORV inoperable in D.1 Restore PORV to 7 days MODE 4. OPERABLE status.

(continued)

McGuire Units 1 and 2 3.4.12-3 Amendment Nos.

LTOP System 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One PORV inoperable in E.1 Suspend all operations Immediately MODE 5 or 6. which could lead to a water solid pressurizer.

AND E.2 Restore PORV to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

F. Required Action and F.1. Verify RCS cold leg 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion temperature > 174 0 F (Unit Time of Condition E not 1) or > 89°F (Unit 2).

met.

AND F.2 Verify RHR suction relief valve is OPERABLE and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the suction isolation valves are open.

G. Two PORVs inoperable. G.1 Depressurize RCS and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> establish RCS vent of OR > 2.75 square inches.

Required Action and associated Completion Time of Condition C, D, E, or F not met.

OR LTOP System inoperable for any reason other than Condition A, B, C, D, E, or F.

McGuire Units 1 and 2 3.4.12-4 Amendment Nos.

ATTACHMENT 4 y

TS BASES PAGES (retyped)

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Steam Line Pressure-Low Function must be OPERABLE in MODES 1, 2, and 3 (above P-1 1), with any main steam valve open, when a secondary side break or stuck open valve could result in the rapid depressurization of the steam lines.

This signal may be manually blocked by the operator below the P-1 1 setpoint. Below P-1 1, an inside containment SLB will be terminated by automatic actuation via Containment Pressure-High High. Stuck valve transients and outside containment SLBs will be terminated by the Steam Line Pressure-Negative Rate-High signal for Steam Line Isolation below P-1 1 when Steam Line Pressure - Low has been manually blocked. The Steam Line Isolation Function is required in MODES 2 and 3 unless all MSIVs are closed and de-activated. This Function is not required to be OPERABLE in MODES 4, 5, and 6 because there is insufficient energy in the secondary side of the unit to have an accident.

(2) Steam Line Pressure-Negative Rate-High Steam Line Pressure-Negative Rate-High provides closure of the MSIVs for an SLB when less than the P-1 1 setpoint, to maintain at least one unfaulted SG as a heat sink for the reactor, and to limit the mass and energy release to containment. When the operator manually blocks the Steam Line Pressure-Low main steam isolation signal when less than the P-1 1 setpoint, the Steam Line Pressure-Negative Rate High signal is automatically enabled. Steam Line Pressure Negative Rate-High provides no input to any control functions.

Thus, three OPERABLE channels are sufficient to satisfy requirements with a two-out-of-three logic on each steam line.

Steam Line Pressure-Negative Rate-High must be OPERABLE in MODE 3 when less than the P-1 1 setpoint, when a secondary side break or stuck open valve could result in the rapid depressurization of the steam line(s). In MODES 1 and 2, and in MODE 3, when above the P-11 setpoint, this signal is automatically disabled and the Steam Line Pressure-Low signal is automatically enabled. The Steam Line Isolation Function is required to be OPERABLE in McGuire Units 1 and 2 B 3.3.2-17 Revision No.

RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 5).

A second program that employs excore cavity dosimetry to monitor the reactor vessel neutron fluence has been installed in both units. This program meets the requirements of 10 CFR 50 Appendix H (Ref. 4).

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limit curve includes the Reference 1 requirement that it be

> 40°F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

The ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)

SAFETY ANALYSES analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Although the P/T limits are not derived from any DBA, the P/iT limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36 (Ref. 7).

McGuire Units 1 and 2 B 3.4.3-2 Revision No.

LTOP System B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the specified limit. The setpoints are derived by analyses that model the performance of the LTOP System, assuming the limiting LTOP transient of one centrifugal charging pump or one safety injection pump injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.

The PORV setpoints will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," discuss these examinations.

The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

RCS Vent Performance With the RCS depressurized, analyses show a vent size of 2.75 square inches is capable of mitigating the allowed LTOP overpressure transient.

The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, one centrifugal charging pump or one safety injection pump OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance.

The RCS vent is passive and is not subject to active failure.

The LTOP System satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).

McGuire Units 1 and 2 B 3.4.12-5 Revision No.

LTOP System B 3.4.12 BASES LCO This LCO requires that the LTOP System is OPERABLE. The LTOP System is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO permits a maximum of one centrifugal charging pump or one safety injection pump capable of injecting into the RCS and requires all accumulator discharge isolation valves closed and immobilized when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in LCO 3.4.3.

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:

a. Two OPERABLE PORVs (NC-32B and NC-34A); or A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the specified limit and testing proves its automatic ability to open at this setpoint, and motive power is available to the valve and its control circuit.

b. A depressurized RCS and an RCS vent.

An RCS vent is OPERABLE when open with an area of

> 2.75 square inches.

Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

The LCO is modified with a note that specifies that a PORV secured in the open position may be used to meet the RCS vent requirement provided that its associated block valve is open and power removed.

With the PORV physically secured or locked in the open position with its associated block valve open and power removed, this vent path is passive and is not subject to active failure.

APPLICABILITY This LCO is applicable in MODE 4 when any RCS cold leg temperature is

< 3000F, in MODE 5, and in MODE 6 when the reactor vessel head is on.

The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 300 0F. When the reactor vessel head is off, overpressurization cannot occur.

LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of McGuire Units 1 and 2 B63.4.12-6 Revision No.

LTOP System B 3.4.12 BASES APPLICABILITY (continued) the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above 300°F.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

The Applicability is modified by a Note stating that accumulator isolation is only required when the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves. This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

ACTIONS LCO 3.0.4 is not applicable for entry into LTOP operation.

A.1, A.2.1, A.2.2.1, A.2.2.2, A.3, A.4, A.5.1, and A.5.2 With two centrifugal charging pumps, safety injection pumps, or a combination of each, capable of injecting into the RCS, RCS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

Two pumps may be capable of injecting into the RCS provided the RHR suction relief valve is OPERABLE with:

1. RCS cold leg temperature > 174 0F (Unit 1), or

2. RCS cold leg temperature > 89 0F (Unit 2), or 0

3. RCS cold leg temperature > 74'F and cooldown rate < 2 °F/hr (Unit 1),

or

4. RCS cold leg temperature > 740F and cooldown rate < 60'F/hr (Unit 2),

or

5. two PORVs secured open with associated block valves open and power removed, or

6. a RCS vent of > 4.5 square inches, or

7. a RCS vent of > 2.75 square inches and two OPERABLE PORVs (the RCS vent shall not be one of the two OPERABLE PORVs).

For cases where no reactor coolant pumps are in operation, RCS cold leg temperature limits are to be met by monitoring of BOTH the WR Cold Leg temperatures and Residual Heat Removal Heat Exchanger discharge temperature. With both PORVS and block valves secured open, or with McGuire Units 1 and 2 B 3.4.12-7 Revision No.

LTOP System B 3.4.12 BASES ACTIONS (continued) active failure of the remaining valve path during this time period is very low.

E.1 and E.2 The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 8). Thus, with one of the two PORVs inoperable in MODE 5 or in MODE 6 with the head on, all operations which could lead to a water solid pressurizer must be suspended immediately and the Completion Time to restore two valves to OPERABLE status is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The Completion Time represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE PORV to protect against overpressure events.

F.1 and F.2 If the Required Actions and associated Completion Times of Condition E are not met, then alternative actions are necessary to establish the required redundancy in relief capacity. This is accomplished by verifying that the RHR relief valve is OPERABLE and the RHR suction isolation 0 0

valves open and the RCS cold leg temperature > 174 F (Unit 1) or > 89 F (Unit 2). For cases where no reactor coolant pumps are in operation, RCS cold leg temperature limits are to be met by monitoring of BOTH the WR Cold Leg temperatures and Residual Heat Removal Heat Exchanger discharge temperature. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reflects the importance of restoring the required redundancy at lower RCS temperatures.

G.1 The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when:

a. Both required PORVs are inoperable; or

b. A Required Action and associated Completion Time of Condition C, D, E, or F is not met; or B 3.4.12-9 Revision No.

McGuire Units 1 and 2

ATTACHMENT 5 NUREG-1431 MARK-UP PAGE IN MAY 27, 1997 LAR

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page I of 8)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SUIRVEI LLAN.CE ALLOWABLE FUINCTIONI CONDITIONS CHANNELS CONDITIONS REOUIREMENTS VALUE

,. Steam Line Isolation (continued)

c. Containment 0 Gý SR 3.3.2.1 Pressure 1( ý1ý F SR psig PSg*

-N SR 3.3.2.iko SR

d. Steam Line Pressure (1) Low 3 per O SR 3.3.2.1 k Stea SR Line SR 3.3.2.1s SR c9 (2) Negative 3 per a SR 3.3.2.1 s ) i Rate - Nigh steam Si 3.3.2.5 s%

i s tine SR 3.3.2.a*

SR 3.3.2.kjý Rev i ewer, N~ote: Unit spxpific japtementytions may ctan ~

oain twat W~* &pning on Setl nt Stu:

Above the P-11 (Pressurizer Pressure) interlock.

lin Ca con ro er are t, t2 sec Abov e P-12 (,-L Low) in dock.

Less an or equal to a ftancti defined as IP corresponding to 4"1% full steam flow below (202% Iad. Ai Seasing linearly from (441 full steam flow at (M0% load J114]% full steam flow at (100]% oaand I corresponding to (1141% f t steam flow above 100% load.

7,tess than or equal to a f tion defined as LI' correspoedi to (402% full steam flow between 01% and (202 load and then a AP' *ncrc.t ng linearly from (40]% stc-n f at [202% load to (1101% full z amflow at Except when a areclosed Rde-activatev

1. 04/07 WC O*-ýTS 3.3-35 'Rev 1, 04/07/95,"---

(A- &' . M jSo Lcfoa ' eLae4 o.