L-11-085, Regarding the Submittal of the Unapproved Pressure and Temperature Limits Report

Regarding the Submittal of the Unapproved Pressure and Temperature Limits Report

ML110690027
Person / Time
Site:	Davis Besse
Issue date:	03/07/2011
From:	Byrd K FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-11-085, TAC ME1127, TAC ME1128
Download: ML110690027 (12)

Text

FENOC Davis-Besse Nuclear Power Station O5501 N. State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 March 7, 2011 L-1 1-085 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station Docket No. 50-346, License No. NPF-3 Submittal of the Unapproved Pressure and Temperature Limits Report (TAC Nos. ME1127 and ME1128)

By letter dated April 15, 2009 (Accession No. ML091130228), and supplemented by letters dated December 18, 2009 (Accession No. ML093570103), August 26, 2010 (Accession No. ML102440105) and October 8, 2010 (Accession No. ML102861221),

the FirstEnergy Nuclear Operating Company (FENOC) submitted to the Nuclear Regulatory Commission (NRC) a license amendment request and an exemption request for the Davis-Besse Nuclear Power Station (DBNPS). The proposed amendment would incorporate alternate methodologies for developing the DBNPS reactor pressure vessel pressure and temperature limit curves. The proposed exemption would exempt DBNPS from certain requirements contained in 10 CFR 50.61 and 10 CFR Part 50, Appendix G.

By electronic mail dated January 10, 2011 (Accession No. ML110100757), FENOC submitted, as requested by the NRC staff, an unapproved version of the DBNPS pressure and temperature limits report (PTLR) based upon the use of the methodologies requested in the proposed license amendment request and the exemption request.

Davis-Besse Nuclear Power Station L-1 1-085 Page 2 of 2 This submittal provides the unapproved PTLR for inclusion onto the DBNPS docket.

The unapproved version is being provided for administrative purposes only (for example, report format and type of information to be provided). The technical data included in the report is preliminary data and is subject to change. Submittal of the approved version, which may differ from the unapproved version, will be performed in accordance with Technical Specification 5.6.4, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)."

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) 761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 7,2011.

Sincerely, Kendall W.td Director, Site Performance Improvement

Attachment:

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 cc:

NRC Region III Administrator NRC Project Manager NRC Resident Inspector Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

Attachment L-1 1-085 FirstEnergy Nuclear Operating Company Davis-Besse Unit 1 Pressure and Temperature Limits Ieport for the Earlier of 32 Effective Full Power Years or April 22, 2017 (Nine Pages Follow)

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 NOT APPROVED Reviewed by: Kevin Bumworth NOT APPROVED Approved by: Kevin Zellers NOT APPROVED

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 (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.

32 EFPY PTLR Page 3 of 9 Figure 1: Composite Normal Heatup/Cooldown Limit - Hot Leg "A" Pressure Tap 2600 2400 2200 2000 1800

" 1600

- 1400

= 1200

, 1000

o. 800 600 400 200 0

Y

~

K.

Poini A

I B

IF G

I K

L M

_____N 0

P Q

R I

S I

V

• .Teffm 70 180 190 195 205 210 215 220 228 270 270 220 220 230 235 245 250 255 26b 268 310 310 Press 540 54W 957 1165 1832 2064 2190 2329 2467 2467 2500 0

540 957 1165 1832 2064 2190 2329 2467 2467 25(0 A

j rp 7-4

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II Notes:

1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 0F step change followed by an 18-minute hold.

Q-

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 0F 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.

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50 100 150 200 250 300 350 400 450 Temperature, OF

32 EFPY PTLR Page 4 of 9 Figure 2: Composite Normal Heatup/Cooldown Limit - Hot Leg "B" Pressure Tap K,

2600 2400 2200 2000 1800

"; 1600 a 1400

= 1200

= 1000 800 600 400 200 0

Point A

B C

D E

F G

H I

J K

L M

N 0

P Q

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V Tema 70 180 190 195 205 210 215 22O 228 270 270 220 220 230 235 245 250 255 260 268 310 310 Press 565 565 982 1190 1857 2064 2190 2329 2492 2492 2525 0

565 982 1190 1857 2064 2190 2329 2492 2492 2525 p

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4 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.

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-- Heatup/Cooldown Limit...

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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 2600 I

I I

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2400 2200 2000 1800

• ;;1600 1400 1200 l1000 800 600 400 200 0

"A" Tap "B" Tap Point Ternp Press Press A

70 540 565 B

160 540 565 C

175 1165 1190 D

190 2165 2190 E

195 2382 2382 F

200 2507 2507 Notes:

I. 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" Tap Press c "B" Tap Press -

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50 100 150 200 250 300 350 400 Temperature, OF

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-2241 P-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 RTNDT values used (in part) to determine ART was calculated using an alternate methodology described in the NRC-approved BAW-23 08, 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 RTNDT 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 DBNPS 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 2y, 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)

Fluence ART ART

@ 52 EFPY

@ 1/4 T

@ - T (Wetted Surface)

(OF)

(OF)

Limiting RTPTS Reactor Vessel Material (n/cm 2)

@52 EFPY

@52 EFPY Mat'l?

(OF)

Location Identification (E> 1 MeV)

(Notel)

(Notel)

(Yes/No)

(Note 2)

Nozzle Belt ADB 203 2.29E+ 18 74.8 64.8 No 81.2 Forging Nozzle Belt to Upper Shell Weld WF-232 2.29E+ 18 Note 3 Note 3 No 118.3 (ID 9%)

Nozzle Belt to Upper Shell Weld WF-233 2.29E+18 100.4*

67.8*

No Note 4 (OD 91%)

UpperShell AKJ 233 1.69E+19 71.8 57.3 No 79.4 Forging Upper Shell to Lower Shell WF-182-1 1.69E+19 156.2*

106.4*

Yes 182.2*

Weld I

LowerShell BCC 241 1.70E+19 89.9 78.8 Yes 95.7 Forging II_

I 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 1/4AT or Y4T location.

Note 4: This weld material is not present at the clad to vessel interface, so RTpTs does not apply to it.

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

32 EFPY PTLR 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 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 Power 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 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 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.