L-14-289, Pressure and Temperature Limits Report. Revision 2

Pressure and Temperature Limits Report. Revision 2

ML14266A109
Person / Time
Site:	Davis Besse
Issue date:	09/22/2014
From:	Lieb R FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-14-289
Download: ML14266A109 (11)

Text

FENOC' ~

5501 North State Route 2 Oak Harbor, Ohio 43449 FirstEnergy Nuclear Operating Company Raymond A. Lieb 419-321-7676 Vice President, Nuclear Fax: 419-321-7582 September 22, 2014 L-14-289 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Pressure and Temperature Limits Report. Revision 2 In accordance with Technical Specification 5.6.4, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," FirstEnergy Nuclear Operating Company hereby submits Revision 2 of the PTLR for the Davis-Besse Nuclear Power Station, Unit No. 1, which was issued on September 15, 2014.

Revision 2 reissued the 32 Effective Full Power Years (EFPY) Pressure-Temperature limits to incorporate Revision 4 of ANP-2718, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station." Revision 4 of ANP-2718 combined the Heatup/Cooldown Curves into a single figure. This resulted in the re-numbering of the PTLR In-Service Leak and Hydrostatic Tests Figure from Figure 3 to Figure 2. No methodology changes occurred in this revision.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.

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Enclosure:

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, Revision 2 cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager Utility Radiological Safety Board

Enclosure L-14-289 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, Revision 2 (Nine Pages Follow)

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 Revision 2

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32 EFPY PTLR Rev. 2 Page 2 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

1. 0 Introduction This Pressure and Temperature Limits 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 (RCS) pressure boundary is operated in accordance with its design. The limits provided are valid to 32 Effective Full Power Years (EFPY) of operation or April 22, 2017, whichever occurs first.

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.

Revision 0 was the initial issue of the 32 EFPY PTLR after issuance of License Amendment 282, which authorized use of new methodologies.

Revision 1 is re-issuing the 32 EFPY Pressure-Temperature limits to include the limits for the Reactor Vessel Closure Head (RVCH) installed in October 2011 Cycle 17 Mid-cycle Outage. The limits associated with the RVCH obtained from the Midland nuclear power plant have been removed. No methodology changes occurred in this revision.

Revision 2 is re-issuing the 32 EFPY Pressure-Temperature limits to incorporate Revision 4 of ANP-2718, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station" (Reference 5.7).

Revision 4 of ANP-2718 combined the Heatup/Cooldown Curves into a single Figure.

This results in the re-numbering of the In-Service Leak and Hydrostatic Tests Figure to Figure 2. No methodology changes occurred in this revision.

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 and 2 (Reference 5. 7) during heatup, cool down, criticality, and in-service leak and hydrostatic (ISLH) testing with:

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

32 EFPY PTLR Rev. 2 Page 3 of 9

2. A maximum cooldown of 100°F in any one hour period with a cold leg temperature of 2: 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 oflow temperature operation (Tavg <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.

Intentionally Blank

32 EFPY PTLR Rev.2 Page 4 of 9 Figure 1: Composite Normal Heatup/Cooldown Limit - Both Hot Leg Pressure Taps

32 EFPY PTLR Rev.2 Page 5 of 9 Figure 2 Reactor Coolant System Pressure-Temperature Heatup and Cooldown Limits for In-Service Leak and Hydrostatic Tests 2600 I I I 1 Point Temp Press Point Temp Press A 70 871 c 140 1296 2400 75 876 D 145 1486 80 889 150 1583 2200 85 909 155 1685 90 933 E 160 1795 95 961 165 1859 2000 100 993 170 1924 105 1028 175 1997 1800 110 1066 115 1108 F 180 185 2078 2170 C) 120 1154 190 2270 Notes:

"Ci) 1600 125 1205 195 2382 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change c.. B 130 1261 G 200 2507 followed by an 18-minute hold.

a) s.;..

1400 1 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.

~ 1200 VJ

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. --1 ct 1000 4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the OHR system operating. The step temperature change is defined as RC temp minus the OHR return temps to 800 the reactor coolant system prior to stopping all pumps.

5. When the decay heat removal system (DH) is operating without any RC 600 pumps operating, indicated DH return temperature to the reactor vessel

- - - ISLH Limit, Both Taps shall be used.

f 400 6. The acceptable pressure and temperature combinations are below and to --i I

1 the right of the limit curve.

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

0 0 50 100 150 200 250 300 350 400 Temperature, °F

32 EFPY PTLR Rev. 2 Page 6 of 9 3.0 Analytical Methods 3 .1 The limits provided in Section 2 and Figures 1 and 2 are valid until the Reactor Vessel has accumulated 32 Effective Full Power Years (EFPY) of fast (E> 1 MeV) neutron fluence or April 22, 201 7, 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 temperature 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 CPR 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 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, as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640 (Reference 5.9).

3.5 Table 1 provides the Adjusted Reference Temperature (ART) for each reactor vessel 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 RTNDT values used (in part) to determine ART were calculated using an alternate methodology described in the NRC-approved BAW-2308, Revisions 1-A and 2-A (Reference 5.10). The NRC required licensees to obtain an exemption from 10 CPR 50.61and10 CPR 50, Appendix G to use the alternate initial RTNDT values provided inBAW-2308 Revisions 1-A and 2-A. The required exemption was granted by the NRC in Reference 5 .17. The NRC confirmed the limits and conditions for using the methodology were satisfied in the SER of Amendment 282 (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 CPR 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 Rev. 2 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 DBNPS is included in Ref. 5.8.

3.7 The minimum temperature requirements of 10 CFR 50, Appendix Gare 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 10 CFR 50, 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 Measured ~RTNDT - Predicted ~RTNDT data scatter was less than 2cr, 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 PIT limits will be generated in accordance with the NRC approved methodologies described in TS 5.6.4.

32 EFPY PTLR Rev.2 Page 8 of 9 Table1: Davis-Besse Nuclear Power Station Reactor Vessel Beltline Region Data (Applicable as noted in Section 3.1)

Fluence ART ART

@52EFPY @Y<IT @%T (Wetted Surface) (oF) (oF) Limiting RTPTS Reactor Vessel Material (n/cm2) @52 EFPY @52EFPY Mat'l? (oF)

Location Identification (E> 1 MeV) (Note 1) (Note 1) (Yes/No) (Note 2)

Nozzle Belt ADB 203 2.29E+l8 74.8 64.8 No 81.2 Forging Nozzle Belt to Upper Shell Weld WF-232 2.29E+18 Note3 Note3 No 157.9 (ID 9%)

Nozzle Belt to Upper Shell Weld WF-233 2.29E+18 100.4* 67.8* No Note4 (OD 91%)

Upper Shell AKJ 233 l.69E+ 19 71.8 57.3 No 79.4 Forging Upper Shell to Lower Shell WF-182-1 l.69E+ 19 156.2* 106.4* Yes 182.2*

Weld Lower Shell BCC 241 l.70E+ 19 89.9 78.8 Yes 95.7 Forging Note 1: Reported ART values are based on Regulatory Guide 1.99, Revision 2 (Ref. 5.15). PIT 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 )4T or %T 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 RTNDT provided in the NRC Safety Evaluation Reports to BAW-2308, Rev. 1-A and 2-A (Ref. 5.10).

32 EFPY PTLR Rev. 2 Page 9 of 9 5.0 References 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 CPR 50 Appendix G."

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

5. 6 BAW-154 3A, "Master Integrated Reactor Vessel Material Surveillance Pro gram."

5.7 ANP-2718, Revision 4, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station," dated December 2013.

5.8 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 282 to Facility Operating License No. NPF-3, FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station, Unit No. 1, (FENOC Ltr. Rl 1-030), dated 01/28/2011.

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 subsequently 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 RTNDT of Linde 80 Weld Materials," dated August 2005 (1-A) and March 2008 (2-A).

5.11 Calculation C-NSA-064.02-037, Revision 1, "Davis-Besse 52 EFPY PT Limits -

Chalon RV Closure Head," dated 9/23/2011.

5.12 AREVA Report 86-9015129-000, "DBl - 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 AREY A 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.

5.17 NRC Letter to FirstEnergy Nuclear Operating Company, "Davis-Besse Nuclear Power Station, Unit 1-Exemption from the Requirements of 10 CPR Part 50.61 and 10 CPR Part 50, Appendix G," (FENOC Ltr. Rl0-298) dated December 14, 2010.