ML110100758

Pressure Temperature Limits Report Attachment

ML110100758
Person / Time
Site:	Davis Besse
Issue date:	01/10/2011
From:	Lashley P FirstEnergy Corp
To:	Michael Mahoney Plant Licensing Branch III
Mahoney, M NRR/DORL/LPLIII- 2 415-3867
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ML110100757	List: ML110100755 ML110100758
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Download: ML110100758 (9)

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32 EFPY PTLR Page 1 of 9

FIRSTENERGY NUCLEAR OPERATING COMPANY DAVIS-BESSE UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT FOR THE EARLIER OF 32 EFFECTIVE FULL POWER YEARS OR APRIL 22, 2017

Prepared by: Dennis Blakely ___________________

Reviewed by: Kevin Burnworth ____________________

Approved by: Kevin Zellers ______________________

32 EFPY PTLR Page 2 of 9

FirstEnergy Nuclear Operating Company Davis-Besse Unit 1 32 Effective Full Power Years Pressure and Temperature Limit Report

1.0 Introduction

This Pressure and Temperature Limit Report (PTLR) provides the information required by Davis-Besse Nuclear Power Station (DBNPS) Technical Specification 5.6.4 to ensure that the Reactor Coolant System Pressure Boundary is operated in accordance with its design.

The PTLR provides the RCS Operating Limits in Section 2.0, which satisfies Technical Specification 5.6.4.a. The Analytical Methods used to develop the limits, including determination of the vessel neutron fluence, are provided in Section 3.0, fulfilling Technical Specification 5.6.4.b. The information and formatting of Section 3 follows the guidance of Attachment 1 to Generic Letter 96-03. The PTLR requirements are provided in Section 4.0 of the report, fulfilling Technical Specification 5.6.4.c.

Revisions to the PTLR are to be submitted to the NRC after issuance.

2.0 RCS Pressure and Temperature Limits

a. The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines and ramp rates shown on Figures 1, 2, and 3 (Reference 5.7) during heatup, cooldown, criticality, and in-service leak and hydrostatic (ISLH) testing with:

1. A maximum heatup of 50°F in any one hour period, and

2. A maximum cooldown of 100°F in any one hour period with a cold leg temperature of > 270°F and a maximum cooldown of 50°F in any one hour period with a cold leg temperature of < 270°F.

b. During periods of low temperature operation (T avg <280 °F), Technical Specification 3.4.12 (Reference 5.3) provides additional requirements for RCS pressure and temperature limits. Those limits are maintained in the Technical Specifications because they are not determined using methods generically approved by the NRC.

32 EFPY PTLR Page 3 of 9 Figure 1: Composite Normal Heatup/Cooldown Limit - Hot Leg "A" Pressure Tap 020040060080010001200140016001800200022002400260050100150200250300350400450Pressure, psigHeatup/Cooldown LimitCriticality Limit Temperature, °F Notes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.

3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps. 5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.

6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.

7. Instrument error is not accounted for in these limits.

A B C D E F G J K H I N O P L M Q R U V S T 32 EFPY PTLR Page 4 of 9 Figure 2: Composite Normal Heatup/Cooldown Limit - Hot Leg "B" Pressure Tap 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 260050100150200250300350400450Pressure, psigHeatup/Cooldown LimitCriticality Limit Temperature, °FNotes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.

3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps. 5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.

6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.

7. Instrument error is not accounted for in these limits.

A B C D E F G J K N O P H I L M Q R U V S T 32 EFPY PTLR Page 5 of 9 Figure 3 Reactor Coolant System Pressure-Temperature Heatup and Cooldown Limits for In-Service Leak and Hydrostatic Tests 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 260050100150200250300350400Pressure, psig"A" Tap Press"B" Tap Press Temperature, °F Notes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.

3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.

4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps.

5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.

6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.

7. Instrument error is not accounted for in these limits.

A B C D E F 32 EFPY PTLR Page 6 of 9

3.0 Analytical

Methods 3.1 The limits provided in Section 2 and Figures 1, 2, and 3 are valid until the Reactor Vessel has accumulated 32 Effective Full Power Years (EFPY) of fast (E> 1 MeV) neutron fluence or April 22, 2017, whichever comes first.

3.2 The neutron fluence is calculated (Reference 5.12 with Reference 5.13) consistent with Regulatory Guide 1.190 using the NRC-approved methodology described in BAW-2241P-A (Reference 5.5). Table 1 provides the neutron fluence values used in the adjusted reference calculations. The listed fluence values are based on 52 EFPY of operation. The limits in Section 2 are administratively limited as described in Section 3.1 based on the current Operating License of Davis-Besse Nuclear Power Station.

3.3 The Davis-Besse Reactor Vessel Material Surveillance Program complies with the requirements of Appendix H to 10 CFR 50 and is described in BAW-1543A (Reference 5.6). This information was approved by the NRC in the SER of Amendment 199 (Reference 5.1). The specimen capsule withdrawal schedule is contained within the supplements of the topical report. All plant specific specimen capsules have been withdrawn from the reactor vessel. The ART values were not calculated using surveillance data (Reference 5.14) since it was determined to be non-credible.

3.4 Low Temperature Overpressure Protection (LTOP) limits are addressed in Section 2.b, above, and Technical Specification 3.4.12 (Reference 5.3). Reference 5.7, which meets the requirements of ASME Section XI, Appendix G, discusses the methods used to determine the temperature at which LTOP must be active. The pressure limit was determined using ASME Section XI, Appendix G (Reference 5.9), as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640.

3.5 Table

1 provides the Adjusted Reference Temperature (ART) for each reactor beltline material. The ART values were calculated in accordance with Regulatory Guide 1.99, Revision 2. For welds in the reactor beltline region, the initial RT NDT values used (in part) to determine ART was calculated using an alternate methodology described in the NRC-approved BAW-2308, Revisions 1-A and 2-A (Reference 5.10). As stated in the NRC Safety Evaluations for the BAW-2308 topical reports, an exemption request to use the alternate initial RT NDT values must be submitted to the NRC. The exemption was granted, and the limits and conditions for using the methodology were confirmed by the NRC to be satisfied in the SER of Amendment ??? (Reference 5.8).

3.6 The Pressure-Temperature (P/T) limits of Section 2 and Figures 1, 2, and 3 (with applicability as stated in 3.1) were generated consistent with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99, Revision 2, using the methods described in BAW-10046A (Reference 5.4) and ASME Section XI, 32 EFPY PTLR Page 7 of 9 Appendix G (Reference 5.9), as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640.

3.6.1 The NRC has reviewed the methods described in BAW-10046A (Reference 5.4) and approved the topical report by issuance of a Safety Evaluation Report (SER) dated April 30, 1986. Section 1.2 of BAW-10046A states that it is applicable to all current B&W nuclear steam systems. 3.6.2 ASME Code Cases N-640 and N-588 have been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008) and thus may be used per NRC Regulatory Issue Summary (RIS) 2004-04. Specific approval for application at DB NPS is included in Ref. 5.8.

3.7 The minimum temperature requirements of 10CFR50, Appendix G are included on Figures 1 and 2. Figure 3 provides the In-Service Leak and Hydrostatic (ISLH) Test Limits. These limits were calculated in accordance with the requirements of 10CFR50, Appendix G and ASME Code Section XI, Appendix G, 1995 Edition, with Addenda through 1996 and ASME Code Cases N-588 and N-640. 3.8 Davis-Besse has removed more than two surveillance capsules. The capsule test results have been evaluated and found to be non-credible (Reference 5.14). Consequently, ART calculations are not based on the surveillance data. The Adjusted ART - Predicted ART data scatter was less than 2, so the Regulatory Guide 1.99, Rev. 2 Chemistry Table values used in the ART calculations are conservative.

4.0 PTLR Requirements 4.1 The PTLR has been prepared in accordance with the requirements of Technical Specification 5.6.4 (see Reference 5.11). The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto. Davis-Besse will continue to meet the requirements of 10 CFR 50, Appendix G, and any changes to the Davis-Besse P/T limits will be generated in accordance with the NRC approved methodologies described in TS 5.6.4.

32 EFPY PTLR Page 8 of 9 Table1: Davis-Besse Nuclear Power Station Reactor Vessel Beltline Region Data (Applicable as noted in Section 3.1)

Reactor Vessel Location Material Identification Fluence @ 52 EFPY (Wetted Surface) (n/cm 2) (E> 1 MeV) ART @ 1/4 T (°F) @52 EFPY (Note1) ART @ 3/4 T (°F) @52 EFPY (Note1) Limiting Mat'l? (Yes/No) RT PTS (°F) (Note 2) Nozzle Belt Forging ADB 203 2.29E+18 74.8 64.8 No 81.2 Nozzle Belt to Upper Shell Weld (ID 9%) WF-232 2.29E+18 Note 3 Note 3 No 118.3 Nozzle Belt to Upper Shell Weld (OD 91%) WF-233 2.29E+18 100.4* 67.8* No Note 4 Upper Shell Forging AKJ 233 1.69E+19 71.8 57.3 No 79.4 Upper Shell to Lower Shell Weld WF-182-1 1.69E+19 156.2* 106.4* Yes 182.2* Lower Shell Forging BCC 241 1.70E+19 89.9 78.8 Yes 95.7 Note 1: Reported ART values are based on Regulatory Guide 1.99, Revision 2 (Ref. 5.15). P/T Limit calculation was based on a temperature value which is more conservative than the listed ART value. (Ref. 5.13) Note 2: Values from Ref. 5.16, which are based on the location specific clad to vessel interface fluence at 52 EFPY. Note 3: This weld material does not extend out to the 1/4T or 3/4T location. Note 4: This weld material is not present at the clad to vessel interface, so RT PTS does not apply to it.

• Based on the initial RT NDT provided in BAW-2308, Rev. 1A and 2A with NRC Safety Evaluation Reports (Ref. 5.8)

32 EFPY PTLR Page 9 of 9

5.1 Safety

Evaluation by the NRC Office of Nuclear Reactor Regulation Related to Amendment No. 199 to Facility Operating License No. NPF-3 Davis-Besse Nuclear Power Station, Unit No. 1, attached to correspondence dated July 20, 1995. 5.2 Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

5.3 Technical

Specification 3.4.12, "Low Temperature Overpressure Protection."

5.4 BAW-10046A, Revision 2 "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G."

5.5 BAW-2241P-A, "Fluence and Uncertainty Methodologies," dated April 1999.

5.6 BAW-1543A, "Master Integrated Reactor Vessel Material Surveillance Program."

5.7 ANP-2718, Revision 3, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station," dated August 2010. 5.8 Safety Evaluation by the NRC Office of Nuclear Reactor Regulation Related to License Amendment Request 08-034 to Facility Operating License No. NPF-3 Davis-Besse Nuclear Powe r Station, Unit No. 1.

5.9 ASME Code Section XI, Appendix G, as modified by the alternate rules provided in ASME Code Case N-640 and ASME Code Case N-588. ASME Code Cases N-640 and N-588 have been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008).

5.10 BAW-2308, Revision 1-A and Revision 2-A, "Initial RT NDT of Linde 80 Weld Materials," dated August 2005 (1-A) and March 2008 (2-A).

5.11 Calculation C-NSA-064.02-037, Revision 0, "Davis-Besse 52 EFPY PT Limits - Midland RV Closure Head," dated ????

5.12 AREVA Report 86-9015129-000, "DB1 - Cycles 13-15 Fluence Analysis Report," dated 4/21/2006.

5.13 AREVA Report 51-9123331-000, "Davis-Besse - EOL Fluence Reconciliation,"

dated 10/8/2009.

5.14 AREVA Document 32-9031157-000, "Davis-Besse Revised ART Values at 52 EFPY," dated 9/20/2006.

5.15 AREVA Document 32-9017744-003, "Davis-Besse ART Values at 52 EFPY,"

dated 10/29/2009.

5.16 AREVA Document 32-9123247-000, "RTPTS Values of Davis-Besse Unit 1 for 52 EFPY, Including Extended Beltline," dated 11/12/09.