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Framatome Inc., Document ANP-2718NP, Revision 007, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station
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Site:	Davis Besse
Issue date:	05/01/2019
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To:	; FirstEnergy Nuclear Operating Co, Office of Nuclear Reactor Regulation
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ML22262A150	List: L-22-194, Submittal of Supplemental Information for the Reanalysis for Protection Against Low Temperature Reactor Coolant System Overpressure Events; ML22262A152;
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Enclosure A L-22-194 Framatome Inc. document ANP-2718NP-007, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station. Nonproprietary (57 Pages Follow)

Framatome Inc. ANP~2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company P ei Appendix G Pressure-Temperature Limits For52 EFPY for Davis-Besse Nuclear Power Station Framatome (formerly AREVA Inc.) Document No.

77-2718P-007 AP 0414-12, Revision 35, Section 4.3 Prepared for First Energy Nuclear Operating Company Framatome 3315 Old Forest Road Lynchburg, Virginia 24506-0935

Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company p

Nature of Changes Revision Description Date 000 Original Release March 2008 001 All July2008 002 All October 2008 Tabl~s 7-2 through 7-7 and Figures 7-1 through 7-6 were updated to reflect changes in their source documents and incorporate customer comments.

003 Section 6.2, addressed minor editorial comment August 2010 Corrected Title in section 7 .1.

Corrected text on page 7-1 as follows "As shown in +

Added criticality limit number table to Figure 7-3 through Figure 7-6.

004 This revision addresses the location correction factors December changes found for the B hot leg pressure tap as a 2013 result of WebCAP 2013-8190.

Section 1.0 and throughout Section 7.0: Replaced "current RV head (Midland)" with "Midland RV head";

replaced "future replacement RV head (Chalan)" with "Chalan RV head" Table 6-1, Sections 6.3, and 7.1: Updated location pressure correction factors for B hot leg pressure tap so that A and B correction factors are identical Section 7.0: Corrected a typographical error "summery" with "summary" Figure 7-1: Updated P"T curves by deleting B tap pressures Table 7-2: Deleted B tap pressures Section 7.3: Deleted Tables and Figures for B tap in Rev. 3 (Tables 7-5, 7-7, Figures 7-4, 7-6); updated Table and Figure numbers Figure 7-4: Corrected title "Nozzle" to "Head"

Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company Pa elfi Revision Description Date 005 This revision addresses the lowest service temperature July 2017 (LST) requirement.

Results for the "Midland RV Head" are removed throughout the document since the "Midland RV Head" is no longer in service.

Section 5.0: Added discussion of Lowest Service Temperature.

Section 6.1 & Table 6-1: Pressure Correction factor for RCS piping added.

Section 6.3: Updated to include the Pressure Correction factor for RCS piping. Consequently LTOP pressure limit value updated.

Section 7.1: The LTOP pressure limit value updated.

Table 7-3: Updated to include the LST requirement of the RCS piping.

Figure 7-2: Updated _to include the LST requirement of the RCS piping.

Table 7-5: Updated to include the LST requirement of the RCS piping.

Figure 7-4: Updated to include the LST requirement of the RCS piping.

Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company Pa eiv Revision Description Date 006 Complete revision. This revision establishes 52 EFPY February 2018 P-T limits based on DB-1 Technical Specification 5.6.4 and Amendments 33 and 43 to the DB-1 License Renewal Application.

Note that Revision 006 to 77-2718 contains proprietary information (e.g., beltline and extended beltline fluence, location correction factors, and heat transfer coefficients as a function of RCP combination). This data supports 60-year plant operation that is not Included In Revisions 000 through 005 of 77-2718.

Revisions 000 through 005 do not include proprietary information and as such there are no separate proprietary and non-proprietary versions. Revision 008 is the first in the series of 77-2718 document revisions to receive both a proprietary and non-proprietary version.

This report has been revised from an ANP report designation to a 77-type report design.ation since it is intended for submittal by Framatome to FENOC.

Note that all quotations from external reference sources are italicized.

Proprietary data is identified by bold brackets [ ].

007 Partial Revision. April 2019 Replaces pages i, Ii, iii, iv Page 3-3 (customer comment incorporation regarding closure head ffuence),

Page 4-5 (correct incorrect statement regarding 2 orders of magnitude), and Page 9-1 (listed one preparer and one reviewer as requested by customer).

Revision 006 combined with Revision 007 constitute a complete document.

77-2718NP-006 TABLE OF CONTENTS TABLE OF CONTENTS ......... ********************************************************************************** ..................... V LIST OF TABLES ......................................................................................................................... vi LIST OF FIGURES ..................................................................................................................... vii

1.0 INTRODUCTION

............................................................................................................1-1

2.0 BACKGROUND

.............................................................................................................2-1 3.0 52 EFPY P-T LIMIT REQUIREMENTS-TECHNICAL SPECIFICATIONS AND LRA .. 3-1 4.0 60-YEAR TRADITIONAL and EXTENDED BELTLINE-MATERIALS .......................... 4-1 4.1 RV BeHline (traditional beltline and extended beltllne) .........................................4-1 4.2 Adjusted Nil-Ductility Transition Reference Temperatures ...................................4-1 4.2.1 Material Properties-Reconciliation to Current Licensing Basis ............................ .4-2 4.3 UPPER SHELF ENERGY ....................................................................................4-4 4.3.1 Reconciliation of EMA for RV Inlet/Outlet Nozzles ...............................................4-5 5.0 DESIGN BASIS FOR PRESSURE/TEMPERATURE LIMITS ........................................5-1 5.1 Material Properties ...............................................................................................5-1 5.2 Postulated Flaws ..................................................................................................5-2 5.3 Upper ShelfToughness ........................................................................................ 5-2 5.4 Uncorrected Reactor Vessel Closure Head Limits ............................................... 5-2 5.5 Convection Film Coefficient. .................................................................................5-2 5.6 Reactor Coolant Transients ..................................................................................5-3 5.7 Adjusted Reference Temperatures ...................................................................... 5-4 5.8 LTOP Transient & LTOP P-T Limits .....................................................................5-4 6.0 TECHNICAL BASIS FOR PRESSURE/TEMPERATURE LIMITS ................................. 6-1 6.1 Fracture Toughness .............................................................................................6-2 6.2 Thermal Analysis and Thermal Stress Intensity Factor ........................................6-2 6.3 Unit Pressure Stress Intensity Factor for Reactor Vessel Beltline Region ........... 6-4 6.4 Unit Pressure Stress Intensity Factor for Reactor Vessel Nozzles....................... 6-4 7.0 PRESSURE CORRECTION ..........................................................................................7-1 7.1 Correction for Normal Heatup/Cooldown .............................................................7-1 7.2 Correction for ISLH ...............................................................................................7-2 8.0

SUMMARY

OF RESULTS .............................................................................................8-1 8.1 P-T limits for Low Temperature Overpressure Protection {LTOP) ..................... 8-19 8.1.1 ITS Figures 3.4.12-1 and Figure 3.4.12-2 ..........................................................8-21 8.2 P-T Limits for RV Shell Transition Regions ........................................................ 8-21 8.3 Lowest Service Temperature ............................................................................. 8-22 9.0 CERTIFICATION ............................................................................................................9-1

10.0 REFERENCES

.............................................................................................................10-1 V

A AREVA

77-271 BNP-006 LIST OF TABLES Table 4-1: Adjusted Reference Temperature Evaluation for the Davis-Besse Reactor Vessel Beltline Materials at the 1/4-Thickness and 3/4-Thickness Locations Applicable Through 52 EFPY ....................................................................................................................4-3 Table 4-2: Fast (E > 1 MeV) Fluence Comparison - LRA to 52 EFPY P-T Analysis (through Cycles 17 & 18) .......................................................................................................... 4-4 Table 5-1: Material Properties ...................................................................................................... 5-1 Table 5-2: RCS Temperatures, RCP Combinations, and HTCs for Heatup and Cooldown ................. 5-3 Table 5-3: Limiting ARrs for DB-1 Beltline Materials...................................................................... 5-4 Table 7-1: Location Pressure Correction Factors for Both A and B Pressure Taps ............................ 7-1 Table 7-2: RCS Piping Location Correction Factors ....................................................................... 7-2 Table 8-1: PTLR Basis Normal Heatup P-T Limits .......................................................................... 8-2 Table 8-2: PTLR Basis Criticality Limit P-T Limits .......................................................................... 8-5 Table 8-3: PTLR Basis Cooldown P-T Limits ................................................................................. 8-8 Table 8-4: Operational Constraints for Plant Heatup ....................................................................... 8-9 Table 8-5: Operational Constraints for Plant Cooldown ................................................................... 8-9 Table 8-6: PTLR Basis ISLH Heatup P-T Limits ........................................................................... 8-12 Table 8-7: PTLR Basis ISLH Cooldown P-T Limits ....................................................................... 8-15 Table 8-8: Reactor Coolant Temperatures for Use in Establishing an LTOP System Effective Temperature [22, TableS-2] ....................................................................................... 8-20 vi A

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77-2718NP-006 LIST OF FIGURES Figure 8-1: PTLR Nomial Heatup and Criticality Core P-T Limits .................................................... 8-10 Figure 8-2: PTLR Nom,al Cooldown P-T Limits.............................................................................8-11 Figure 8-3: PTLR ISLH Heatup P-T Limits .................................................................................... 8-18 Figure 8-4: PTLR ISLH Cooldown P-T Limits ................................................................................ 8-19 vii A

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77-271 BNP-006

1.0 INTRODUCTION

This report presents operational 52 EFPY Appendix G pressure-temperature (P-T) limits for the reactor vessel at Davis-Besse Unit 1 {DB-1). These limits are expressed in the fonn of a curve of allowable pressure versus temperature. In addition, the minimum temperature for core criticality is determined to satisfy the regulatory requirements of 10 CFR Part 50, Appendix G [1]. Uncorrected pressure-temperature limits are calculated for the reactor vessel belUine, inlet and outlet nozzles, and closure head locations for nonnal heatup, normal cooldown, and inservice leak and hydrostatic (ISLH) test conditions. Pressure correction factors were determined between the RCS hot leg pressure taps, Decay Heat Removal (DHR) System relief valve (DH-4849), and various other RCS locations. These correction factors were applied to the uncorrected location-based P-T limits to produce a unifonn set of location adjusted P-T and low-temperature overpressure protection (LTOP) limits. The DB-1 52 EFPY location adjusted Appendix G P-T and LTOP limits reported herein are prepared in accordance with the requirements specified in Section 5.6.4 of the DB-1 Technical Specifications [2], as supplemented by consideration of extended beltline materials as defined in the DB-1 License Renewal Application {LRA) [3], including Amendments 33 and 43

[4, 5], and the NRC SER of the DB-1 LRA, NUREG-2193 [6]. The DB-1 P-T limits are updated in accordance with the PTLR [7], which revises the operating limit date from April 22, 2017 to April 22, 2037.

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.A AREVA

77-271 BNP-006

2.0 BACKGROUND

Reactor vessels may continue to be operated only for that service period within which the requirements of Appendix G are satisfied. For the reactor vessel beltline materials, including welds, plates and forgings, the values of RTNOT and Charpy upper-shelf energy must account for the effects of neutron radiation, including the results of the surveillance program required by CFR Part 50, Appendix H [8]. The effects of neutron radiation must consider the radiation conditions (i.e., the fluence) at the deepest point on the crack front of the flaw assumed in the analysis.

Appendix G includes requirements for (1) reactor vessel Charpy upper-shelf energy, and (2) pressure-temperature limits and minimum temperature requirements. The 52 EFPY Charpy upper shelf energy requirements for DB-1 are addressed in Section 4.2.2 of the DB-1 LRA [3] and the associated NRC SER, NUREG-2193 [6], Section 4.2.2. The pressure-temperature limits and minimum temperature requirements at 52 EFPY are addressed in this ANP document.

A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components" [9). This method utilizes fracture mechanics concepts and the reference temperature for nil-ductility transition, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature (ASTM E208 [1 OD or the temperature that is 60 °F below that at which the material exhibits 50 ft-lbs and 35 mis lateral expansion. The RTNOT of a given material is used to index that material to a reference stress intensity factor curve ~ , Figure G-2210-1 in ASME Section XI, Appendix G. When a given material is indexed to the Kie curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Plant operating limits (i.e., pressure-temperature limits) can then be determined using these aUowable stress intensity factors.

The RTNOT of the reactor vessel materials, and in tum, the pressure-temperature limits of a reactor vessel, must be adjusted to account for the effects of irradiation. Neutron embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel are monitored by the surveillance program required by Appendix H to 10 CFR Part 50, consisting of periodic removal of surveillance capsules from an operating reactor and testing of reactor vessel material specimens obtained from the capsules. The increase in the Charpy V-notch 30 ft-lb temperab.lre is added to the unirradiated RTNOT to adjust It for neutron embrtttlement.

The DB-1 52 EFPY location adjusted P-T limits reported herein are prepared in accordance with the requirements specified in Section 5.6.4 of the DB-1 Improved Standard Technical Specifications [2] as supplemented by Amendments 33 and 43 to the OB-1 license renewal application and associated NRC SER of the DB-1 LRA.

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77-271 BNP-006 3.0 52 EFPY P-T LIMIT REQUIREMENTS-TECHNICAL SPECIFICATIONS AND LRA DS-1 Improved Standard Technical Specifications In accordance with the DB-1 Improved Standard Technical Specifications (2], Section 5.6.4, RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the Pressure and Temperature Limits Report {PTLR) for the following.

LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits."

"The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

1. BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," June 1986;

2. ASME Code Section XI, Appendix G, 1995 Edition with Addenda through 1996, as modified by the alternative procedul9s provided in ASME Code Case N-640 and ASME Code Case N-588; and

3. BAW-2308, Revision 1-A and Revision 2-A, "Initial RTNDT of Unde 80 Weld Materials,"

August 2005 and March 2008, respectively."

Pressure-temperature limits for DB-1 are developed herein in accordance with the requirements of 10 CFR Part 50, Appendix G [1), utilizing the analytical methods and flaw acceptance criteria of topical report BAW-10046A, Revision 2 [11] and the 2007 Edition with 2008 Addenda of ASME Code Section XI, Appendix G [9]. In accordance with NRC Regulatory Issue Summary {RIS) 2004-04 (dated April 5, 2004) [12], licensees may use the provisions of any edition and addenda of the ASME Code Section XI, Appendix G incorporated into 10 CFR 50.55a for RV P-T limit curve development, up to and including the most recently incorporated edition and addenda, without the need for an exemption (previously required when using the code cases). At present, the edition and addenda of ASME Section XI codified by 10 CFR 50.55a are the 2013 Edition and no addenda. ASME Code Cases N-588 and N-640 were incorporated Into ASME Section XI, Appendix G, 1998 Edition through 2000 Addenda [12) and all subsequent editions and addenda.

Davis ~ 1 License Renewal Application, Amendment 33 The DB-1 license renewal application, Amendment 33 [4], includes the following request from the NRC relative to development of 52 EFPY P-T limits. Describe how the P-T limit curves to be developed for use in the period of extended operation, and the methodology used to develop these curves, consider all RV materials (beltline and nonbeltline) and the lowest service temperature of all ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50 Appendix G.

Excerpts from the FENOC response are provided below.

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77-2718NP-006 NFirstEnergy Nuclear Operating Company (FENOC) used the methods described in Topical Report BA W-10046A, nMethods of Compliance with Fractum Toughness and Operational Requimments of 10 CFR 50 Appendix G," Revision 2 [Reference 1), to develop the pressure-temperature (P- 7J I/mils fbr Davis-Besse. For Babcock & Wilcox (B&W) nuclear steam systems, Topical Report BAW-10046A describes methods for compliance with the requirements of 10 CFR 50 Appendix G, "Fracture Toughness Requirements." This mport addresses all reactor coolant pressure bounda,y components (beltline and non-beltline). The NRC has reviewed the methods described in Top/cal Report BAW-10046A and approved the report by issuance of a Safety Evaluation Report (SER) dated April 30, 1986 [Reference 2). Additional detalls fbr development of the Davis-Besse pressure and temperature (P- 7J limits are provided below."

" ... Section 2 of BAW-10046A includes a list of all RCPB items used in the fabrication of B&W-designed plant components. Section 3 describes the material properties (including initial RTNDT and CvUSE (upper shelf energy)) of these femtic RCPB items. Considering all felTitic RCPB items and consideration of lowest service temperatures, Section 4, Pages 4-1 and 4-2 of BAW-10046A, concludes that the reactor vessel closure head region, the reactor vessel outlet nozzles, and the beltline region are the only portions of the RCPB that, at different stages of the vessel's design life, regulate the pressure temperature limitations fbr normal operation and inservlce pressure tests.

For beltline materials, AREVA Inc. (AREVA) has traditionally selected the mactor vessel (RV) material wtth the highest adjusted reference temperature as limiting for the evaluation of P-T limits.

The controlling beltline material fbr Davis-Besse has traditionally been the upper shell to lower shell circumferential seam weld WF-182-1. AREVA recently investigated the impact of structural discontinuity in the vicinity of the beltllne region on P-T limits for Davis-Besse by evaluating the upper transition weld WF-2321233, which is directly below a taper transmon. The evaluation concluded that the stress intensity factors for weld WF-2321233 exceed the stress intensity factors fbr weld WF-182-1; however, the material properties at weld WF-2321233 are significantly better than weld WF-182-1, and more than compensates fbr the higher stress intensity factors. Therefore, weld WF-182-1 continues to be the limiting beltline material relative to establishment of P-T limits fbr Davis-Besse.

As described in BAW-10046A, non RV beltline materials have always been considered when establishing P-T limits fbr Davis-Besse. Page 4-2 of BAW-10046A reports that the outlet nozzle of the reactor vessel is the largest nozzle In the Reactor Coolant System (RCS) and the inside comer of the nozzle is subjected to high local stresses produced by pressure. The RV outlet nozzle, due to consideration of loading conditions, is more limiting relative to stress than any of the Class 1 ferritic branch connections (e.g., hot leg surge nozzle) and the large bore RCS piping and the prima,y nozzles of the steam generators.

With regard to replacement of Class 1 ferritic RCPB items (e.g., RV closure head and any future replacement of RCPB components), ASME Ill NB-3211 (d) requires that protection against non-ductile fracture be provided by satisfying one of the following provisions:

1. performing an evaluation of service and test conditions by methods simHar to those contained in Appendix G; or

2. for piping, pump, and valve material thickness greater than 21/2 In. (64 mm) establishing a lowest service temperature that is not lower than RTNDT (NB-2331) + 100°F (56 *cJ; 3-2 A

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Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company Pa e 3-3

3. for piping, pump, and valve materiel thickness equal to or less than 21/2 in. (64 mm), the requirements of NB-2332(a) shall be met at or below the lowest se,vice temperature as established in the design specification.

Therefore, for replacement components, an ASME Section Ill, Appendix G, analysis is required to ensure that the new component is bounded by the ASME Section XI, Appendix G, analysis of the RV used to derive the P-T limits.*

The three areas of the reactor coolant pressure boundary addressed in this ANP report are the beltline shell region, the reactor coolant nozzles, and the closure head flange region. The beltline shell and nozzle regions (including upper and lower taper transition regions and nozzle belt forging regions) are analyzed specifically for DB-1 using the Kie reference fracture toughness Indexed by the reference nil-ductility temperature, which accounts for the effect of the change in fracture toughness of the RV shell and nozzle regions. The 52 EFPY fluence at the top of the RV Inlet and Outlet nozzle-to-shell welds reported in Reference [16, Table 3-12) is less than [ ]

n/cm2* Since the RV closure flange is above this location, the fluence at the closure head Is estimated to be less than 1.0E17 n/cm2 at 52 EFPY or 60 years, and is therefore not susceptible to reduction In fracture toughness. The closure head has a fracture toughness reference temperature, RTMDT, of [ ] 113]. The P-T limits for DB-1 are based on postulated flaws in the vessel and nozzles as discussed in Section 5.2 below.

Davis Besse-1 License Renewal Application, Amendment 43 The DB-1 license renewal application, Amendment 43 [5], Includes the following additional iequirements relative to the development of 52 EFPY P:-T limits for the period of extended operation relative to beltline and extended beltline materials.

" The current P- T limits, generated consistent with the r&quirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99 Revision 2~ are valid until 32 EFPY, or April 22, 2017, whichever occurs first. A revised pressure and temperature limits report (PTLR) will be submitted to the NRC, in accordance with Technical Specification 5.6.4, before Davis-Besse operates beyond 32 EFPY, or April 22, 2017, whichever occurs first, in accordance with the requirements of 10 CFR 50, Appendix G. The mvised P-T limits for the period of extended operation will be based on an evaluation of the effects of neutron embrittlement for the 60-year belt/ine materials, the stresses in the closure head region of the reactor vessel (subject to significant stresses due mechanical loads resulting from bolt preload) and the stresses in the reactor vessel outlet nozzles (largest nozzles in the RCS and the inside corners of the nozzles are subjected to high local stresses produced by pressure). The 60-year reactor vessel beltilne materials are those listed below, plus any otJ?er that could experience 52 EFPY inside surface f/uence greater than 1. 0E+17 n/cm2. "

"Nozzle Belt Forging (ADB 203)

Upper Shell Forging (AKJ 233)

Lower Shelf Forging (BCC 241)

Nozzle Belt Forging to Upper Shell Forging Circumferential Weld (Inside 9%) (WF-232) I (Outside 91 %) (WF-233)

Upper Shell Forging to Lower Shell Forging Circumferential Weld (WF-1 82-1)

Rflactor Vessel Inlet Nozzle Forgings (BSS 270)

Reactor Vessel Outlet Nozzle Forgings (ATS 239)

Dutchman Forging (122Y384VA1)

77-2718NP-006

Nozzle Belt Forging to Bottom of Reactor Vessel Inlet Nozzle Forging Weld (WF*233 I WF*232Y The commitment to revise the PTLR by April 22, 2017 was revised to April 22, 2037 through PTLR Revision 3 (FENOC letter L-16-229, dated July 28, 2016) [7].

LTOP The D-B LTOP system utilizes the Decay Heat Removal (DHR) System relief valve (D~849) with a lift setting of< 330 psig at RCS temperatures less than 280 °F (Table 1.1*1 and Section 3.4.12 of the Improved Technical Specifications [2D. Normal operation heatup and cooldown rates are considered for the development of LTOP P-T limits. To further support the development of low temperature overpressure protection (LTOP) system limits, temperature differences between the reactor coolant in the downcomer region and the 1/4 t wall location are detennined for the maximum heabJp rate transient. The 1/ 4 t wall location is defined as a point within the vessel wall that is located at a distance of one quarter of the vessel thickness from the cladding-base metal interface.

As specified in ASME Section XI, G-2215, the DB--1 LTOP system shall limit the maximum pressure in the vessel to 100% of the pressure determined to satisfy Eq. (1 ).

Improved Technical Specification Figure 3.4.12*1, RCS Pressure Versus Pressurizer Level Limit for Inoperable OHR System Relief Valve in MODE 4, and Figure 3.4.12-2, RCS Pressure Versus Pressurizer Level Limit for Inoperable OHR System Relief Valve in MODE 5 and MODE 6 when the reactor vessel head is on, provide operating restrictions in Modes 4 through 6 to pro~ against a postulated LTOP event (i.e., failed open MU valve) with an inoperable DHR System relief valve (single failure). The pressurizer level restrictions ensure that the P-T limits reported in the PTLR are not exceeded during the assumed [ ] {i.e., charging injection depletes the MU tank) of the LTOP event such that most limiting P-T HU/CD limit is not exceeded in [

l £14J.

Technical Specification Figures 3.4.12-1 and Figure 3.4.12-2 are based on corrected cooldown P*T limits at 10 EFPY [14, Figure 5]. This calculation was subsequently shown to be acceptable for 21 and 32 EFPY by FENOC. ITS Figures 3.4.12-1 and 3.4.12*2 will remain unchanged if the P*T limits at 52 EFPY for RCS temperatures less than 280 °F are less restrictive than those at 1O EFPY.

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77-2718NP-006 4.0 60-YEAR TRADITIONAL and EXTENDED BELTLINE-MATERIALS 4.1 RV Beltline (traditional beltline and extended beltllne)

In accordance with Section 4.2.1.3 of the DB-1 License Renewal Application, the DB-1 reactor vessel beltline materials (i.e., traditional RV beltline plus extended beltline} include all items with 52 EFPY fluence values greater than 1.0E+17 n/cm2* The fluence analysis methodology from BAW-2241 P-A [15] was used to calculate the fast neutron fluence (E > 1.0 MeV) of the reactor vessel welds and forgings. Fluence results were calculated for Cycles 17-18 operation using a computer model that extends from below the core to the vessel mating surface. This data was benchmarked against cavity dosimetry data for Cycles 17-18. Reactor vessel fluence is projected to end-of-life (52 EFPY) by combining the fluence at the point of interest with the product of the best-estimate fluence rate and the projected time. The fluence rate is from the Cycle 17 analysis. This assumption is acceptable as long as Mure core configurations are equivalent to the Cycle 17 core design with 2817 Mega-Watts of thermal power.

The fast neutron fluence (E > 1.0 MeV) calculation methodology described in BAW-2241P-A complies with the uncertainty requirements of the U.S. NRC Regulatory Guide (RG) 1.190 (i.e.,

uncertainty of the fluence must be 20% (1 o) or less) for traditional beltline materials. There is an absence of regulatory and industry guidance outside the traditional RV beltline (e.g., RV inlet/outlet nozzles and dutchman forging}. In addition, there is insufficient confidence outside the traditional RV beltllne for uncertainty quantification in accordance with RG 1.190, Position 1.4. Fluence values calculated using the NRC-approved BAW-2241 P-A best-estimate methodology, applied to the RV nozzle and dutchman regions adjacent to the traditional RV beltline, are conservatively modeled. Fluence values at traditional beltline locations are best-estimate with uncertainty less than 20% (1 o) assuming appropriate monitoring. A summary of all inner wetted surface fluence values of RV items that exceed 1E+17 nlcrrf at 52 EFPY for the OB-1 reactor vessel is provided in Table 4-1 [16]. The locations identified In Table 4-1 are consistent with the traditional and extended beltline locations identified in the Davis-Besse-1 license renewal application, Table 4.2-4, including Amendment 43 to the DB-1 LRA {Section 3.0 above).

4.2 Adjusted NII-Ductility Transition Reference Temperatures The adjusted reference temperature (ART) for the DB-1 reactor vessel was calculated using the guidelines ouUined in Regulatory Guide 1.99, Revision 2. Consistent with the DB-1 license renewal application, Section 4.2.4, neutron fluence values at the 1/4t and 3/4t locations for the RV welds that connect the nozzles to the nozzle belt forging and at the 12-inch thick section of the nozzle belt forging were obtained by adding the attenuation from both the inside and outside surface. Wetted surface fluence with no attenuation was conservatively assumed for calculation of ARTs at the postulated flaw location in the nozzles.

The 1/ 4t and 3'4t ART values [17] for the Davis-Besse reactor vessel beltllne materials applicable to 52 EFPY are listed in Table 4-1. The beltline material with the highest ART for the Davis-Besse reactor vessel is the upper shell to lower shell circumferential weld, WF-182-1, with ART values at 52 EFPY of 150.9 °Fat the 1/4t wall location and 101.2 °Fat the 3/ 4t wall location. For the inlet and outlet nozzle forgings at the postulated flaw locations, ARTs were calculated and are listed in Table 4-1.

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77-2718NP-006 4.2.1 Material Properties-Reconciliation to Current Licensing Basis The initial RT NOT of the reactor vessel materials reported in Table 4-1 are consistent with Table 4.2-4 of the DB-1 LRA, with the following exceptions [18].

Initial RT NT for welds WF-232, WF-233, and WF-182-1

The DB-1 LRA-initial RTNoT values for WF-232, WF-233, and WF-182-1 were obtained from BAW-23O8, Revision 1-A. BAW-23O8, Revision 2-A [30], is used for the 52 EFPY P-T limits reported herein (Table 4-1) .

The Davis Besse Technical Specifications permit use of BAW-23O8, Revisions -1A and -

2A for initial RT NOT and the corresponding 0 1 (ou) values. Therefore, use of BAW-23O8, Revision -2A is acceptable for beltline welds (WF-182-1, WF-232, and WF-233).

However, for the Inlet/Outlet Nozzle-to-Nozzle Belt welds and Lower Shell-to-Dutchman Forging circumferential welds, an exemption request would be required should FENOC elect to implement BAW-23O8, -1A or -2A, for any subsequent NRC transmittal, including subsequent license renewal.

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n-27111NP-006 Table 4-1: Adjulted Rmnmce Temperature Evaluatlon for Ille Davfa.Bnae Reeetor VNael Bellllne Materials at the Y.-

Thlcknen and 3/4-ThlckneH Locauona Applicable Through 62 EFPY 4-3 A

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77-2718NP-006 4.3 UPPER SHELF ENERGY The requirements in 10 CFR Part 50, Appendix G, are applicable for both RPV USE values and P-T limits. The Cycles 17-18 52 EFPY fluence projections used to develop the 52 EFPY P-T limits reported herein [16, Table 3-15] are compared to the 52 EFPY P-T limits reported in the DB-1 LRA, Table 4.2-1. As noted in Table 4-2 below, the updated Cycle 17-18 52 EFPY projections are less than those reported in the DB-1 LRA at all locations with the exception of the RV inlet and outlet nozzle welds. As such, the NRC findings reported in Section 4.2.2 of NUREG-2193 [6] relative to equivalent margins analyses remain applicable with the exception of the equivalent margins analysis of the RPV inlet and outlet nozzle to shell welds to 60-years

[19).

Table 4-2: Fast (E > 1 MeV) Fluence Comparison - LRA to 52 EFPY P-T Analysis (through Cycles 17 & 18)

Current *Llcensing Basis Cycle 17-18 Projection Material ID, Reactor Vessel Location 52 EFPY Fluence 52 EFPY Fluence Heat Number (nlcm 2) (n/cm2)

Forgings ADB203, Nozzle Belt Forging (NBF) 123Y317 2.29E+18 [ ]

AKJ233, Upper Shell Forging (USF) 123X244 1.69E+19 [ ]

BCC241, Lower Shell Forging (LSF) 5P4086 1.70E+19 [ ]

B$S270, A13315; BDF280, 207487:

Bottom of INF BWVV279, 1.17E:.17.. [ ]

2014n:

BWV\/278, 2014n ATS 239, 2V1520; Bottom of CNF ATS249, 2.30E+17** [ ]

2V1520 Dutchman Forging 122Y384VA1 2.33E+17** [ ]

Welds WF232, NBF to USF Circ. Weld 8T3914 2.29E+18 [ ]

WF 182-1, USF to LSF Circ. Weld 821T44 1.69E+19 [ ]

WF232, Bottom of INF Weld 8T3914 1.17E+17** [ ]

WF232, Bottom of ONF Weld 8T3914 2.30E+17** [ ]

WF232, LSF to Dutchman Cire. Weld 8T3914 2.33E+17 [ ]

" Fluence projections outside the traditional RPV beltline are conservative, best-estimate detennination.

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Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company Pa e4-5 4.3.1 Reconciliation of EMA for RV Inlet/Outlet Nozzles The reactor vessel inlet and outlet nozzle-to-shell welds were evaluated for low upper-shelf energy levels by linear elastic fracture mechanics analytical techniques to satisfy the requirements of Appendix K of ASME Code. The .analysis [20] Is based on an upper bound surface fluence of [ ] at 52 EFPY, calculated at the lowest elevation of the outlet nozzle-to-shell weld. Stresses were derived at the weld location considering the influence of the nozzle-to-shell geometric discontinuity and attached piping reactions for Level A, B, C, and D service loadings.

The analyses reported in Reference (20) were performed using the 2007 Edition with 2008 Addenda [9) of Section XI of the ASME Code, Appendix K. The current edition of ASME Section XI listed in 10 CFR 50.55a is the 2013 Edition [21]. With regard to Appendix K, there are no differences between the 2007 Edition with 2008 Addenda and the 2013 Edition of ASME Section XI, and hence the Reference [20] ASME Section XI, Appendix K analyses are equally applicable to the 2013 Edition of the ASME Code.

The material properties used in the Reference [20] analysis are based on ASME Section II, Part D, 2007 Edition with 2008 Addenda. The only change in the material properties listed in the 2013 Edition of ASME Section II, Part D, for the applicable properties is the coefficient of thermal expansion for stainless steel at 600°F; this value was changed from 9.8E-6 in/in/°F to 9.9E-6 in/inl°F. This does not impact the Levels A and B evaluation, and at the limiting time points in the Level C & D analysis, where cladding effects are included, the temperature of the cladding is well below 600°F, and thus this change doe*s not impact the low upper shelf toughness analysis reported in Reference [20]. ***

The reactor vessel nozzle-to-shell welds satisfy all acceptance criteria of ASME Code,Section XI, Article K-2000. For Level A and B service loadings, the applied J-integral of the material at 1.15 times the accumulation pressure, plus thermal loadings is less than the J-lntegral of the material at a ductile flaw extension of 0. 10*, by a margin of [ ] . The applied J-integral for Level C and D Service loadings is less than the required measure of J-integral resistance by a margin of [ ] . Furthermore, the criterion for ductile and stable flaw extensions is satisfied for all Level A, B, C and D service loadings.

77-2718NP-006 5.0 DESIGN BASIS FOR PRESSURE/TEMPERATURE LIMITS Essential geometric data and analytical parameters used in the preparation of DB-1 P-T limits are described below [22J.

5.1 Material Properties Table 5-1 describes the material properties used in the development of the P-T limits for the DB-1

[22, Table 4-5J.

Table 5-1: Material Properties 5-1 A

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77-2718NP-006 5.2 Postulated Flaws Two types of flaw are postulated: nozzle corner flaw and vessel wall flaw, where vessel wall flaw occurs on inside and outside surface of cylindrical shell and in two possible orientations, either axial or circumferential [22, Section 4.2).

Nozzle Flaw o a 3 inch deep comer flaw in the 12 inches nozzle belt section is postulated on the inside surface of the reactor vessel inlet, outlet, and core flood nozzles

Vessel Flaw o per G-2120 of ASME Code,Section IX , a longitudinal semi-elliptical surface flaw 1

/4 t (t = wall thickness) deep and % t long is postulated at both the inside and outside surfaces of the reactor vessel beltline and nozzle belt regions.

Comparison of the outlet, inlet and core flooding nozzle radii show that the outlet nozzle has the highest radius, therefore the hoop stress is maximum in the outlet nozzle. For this reason, the corner flaw at the inlet and core flooding nozzle comers are from the structural point of view bounded by the comer at the outlet nozzle.

5.3 Upper Shelf Toughness The upper shelf fracture toughness of the low alloy steel is limited to [ ] . For the nozzle forging, the Kie curve without a cut-off limit is used [22, Section 4.3].

5.4 Uncorrected Reactor Vessel Closure Head Limits Pressure-temperature limits for the reactor vessel head-to-flange closure region were derived for the replacement closure head region based on the Kie fracture toughness curve [23].

5.5 Convection Film Coefficient Effective convection heat transfer film coefficient at the cladding-to-base metal interface for times during heatup and cooldown when Reactor Coolant Pumps (RCPs) are in use are provided in Reference [18, Table 3-1]. The temperature ranges listed in Table 5-2 below are uncorrected RV inlet and RV outlet nozzle temperatures and are based on review of DB-1 Heatup (HU) and Cooldown (CD) procedures.

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77-2718NP-006 Table 5-2: RCS Temperatures, RCP Combinations, and HTCs for Heatup and Cooldown 5.6 Reactor Coolant Transients Both ramped and stepped transient definitions were modeled for nonnal operation heatup and cooldown. The limiting nonnal heatup and cooldown transients (as determined by the controlling P-T limits) are also used to simulate the reactor coolant transients used for inservice leak and hydrostatic (ISLH) pressure testing. Two cases of normal heatup transients from 70°F to 550°F as provided in reference [22, Section 4.5.1] were analyzed:

1. Heatup Case 1: heabJp rate 75°F/hour

a. Ramped: 75°F/hour

b. Stepped: 15°F steps with 12 minute hold times, steps equivcientto 75°F/hour

2. Heatup Case 2: heatup rate S0°F/hour

a. Ramped: 50°F/hour

b. Stepped: 15°F steps with 18 minute hold times, steps equivalent to 50°F/hour Two cases of normal cooldown transients from 550°F to 70°F as provided in reference [22, Section 4.5.2] were analyzed:

1. Cooldown Case 1: cooldown rate 100°F/hour followed by cooldown rate 50°F/hour

a. Ramped: 100°F/hour ramp from 550°F to 270"F followed by 50°F/hour ramp

b. Stepped: 15°F steps with 9 minute hold times from 550°F to 270°F foRowed by 1s°F steps with 18 minute hold times 5-3 A

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77-271 BNP-006

2. Cooldown Case 2: cooldown rate 100°F/hourfollowed by cooldown rate 25°F/hour

a. Ramped: 100°F/hour ramp from 550°F to 270°F followed by 25°F/hour ramp

b. Stepped: 15°F steps with 9 minute hold times from 550°F to 270°F followed by 15°F steps with 36 minute hold times 5.7 Adjusted Reference Temperatures As discussed in Section 4.0, limiting values of the adjusted reference temperatures (ARTs),

(ARTsi, were evaluated. The limiting ART values that were used for determining the P-T curves are listed in Table 5-3 for the 1/ 4t and 3/ 4t locations of the reactor vessel belUine at 52 EFPY [22, Table 4-3]; the applicable ART values listed in Table 4-1 are rounded up to full degrees (1 °F).

An ART of [ ] was used for the reactor vessel inlet nozzles and an RTNor of [ ) is for the reactor vessel outlet nozzles.

Table 6-3: Limiting ART's for DB-1 Beltline Materials Vessel Wall Limiting ART (°F) at Component Location Material 52 EFPY Beltline 1/4t SA-508 Cl. 2 [ ]

3/4t Lower Shell Forging SA-508 Cl. 2 [ ]

(BCC 241) 1/4t Upper Shell to \NF-182-1 [ ]

3/4t Lower Shell Circ. Weld \NF-182-1 [ ]

CWF-182-1}

5.8 LTOP Transient & LTOP P-T Limits The normal operation heatup and cooldown rates were considered in the development of the LTOP P-T limits. In accordance with ASME Section XI, Subarticle G-2215, the maximum allowable pressure in the RV (for development of LTOP systems) is limited to 100% of the Appendix G P-T limit.

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77-2718NP-006 6.0 TECHNICAL BASIS FOR PRESSURE/TEMPERATURE LIMITS Pressure-temperature limits are developed using AREVA NP computer code PTPC with the analytical inputs described in Section 5.4.3. The analytical approach is in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G [9].

Additional requirements are contained in Table 1 of Appendix G to Title 10, Code of Federal Regulations, Part 50 [1]. The analytical techniques used to calculate P-T limits are based on approved linear elastic fracture mechanics methodology described in topical report BAW-10046A (11]. The fundamental equation used to calculate the allowable pressure is where, Pallow = allowable pressure KIR = reference stress intensity factor ( K re )

Krr = thermal stress intensity factor KIP = unit pressure stress intensity factor (due to 1 psig)

SF = safety factor For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, inlet nozzle, outlet nozzle, and closure head. In the beltline region, flaws are postulated to be present at the 1'4t and 3'4t locations of the controlling material (shell forging, or circumferential weld),

as defined by the fluence adjusted RTNOT* The reactor vessel nozzle flaws are located at the inside juncture (comer) with the nozzle belt shell, and the closure head flaw is located near the outside juncture with the head flange. P-T limits for the beltline and nozzle regions are calculated using a safety factor of 2 for normal operation and 1.5 for ISLH operation. The P-T limit curves presented consist of the allowable pressures for the controlling beltline flaw, inlet and outlet nozzles, and closure head, as a function of fluid temperature. These curves have been "smoothed", as necessary, to eliminate irregularities associated with the startup of the first reactor coolant pump during heatup and the initiation of decay heat removal during cooldown. After the initial determination of the P-T limit curves, location specific curves were adjusted for sensor location. No instrument error correction has been applied. The final results include the determination of a minimum/lower bound P-T curve.

The criticality limit temperature is obtained by determining the maximum required ISLH test temperature at a pressure of 2500 psig (approximately 10% above the normal operating pressure).

The ISLH analysis considers the most limiting heatup and cooldown transients. The approach satisfies the requirement of Item 2.d in Table 1 of 10 CFR 50, Appendix G [1]. It requires the minimum temperature to be the larger of minimum permissible temperature for inservice system hydrostatic pressure test or the RTNOT of the closure flange material + 160 °F.

A Temperature differences between the reactor coolant and the 1/ 4t wall location are needed for the determination of the LTOP systems effective temperature per Article G-2215 of ASME Code 6-1 AREVA

77-271 BNP-006 Section XI, Appendix G. During normal heatup, the metal temperature is lower than the coolant temperature. The temperature differences between the 1'4t wall and coolant temperature are calculated for both the ramp and step heatup cases.

6.1 Fracture Toughness The fracture toughness of reactor vessel steels is expressed as a function of crack-tip temperature, T, indexed to the adjusted reference temperature of the material, RTNOT* Pressure/temperature limits developed in accordance to ASME Code,Section XI, Appendix G [9] utilize the crack initiation fracture toughness, Kie = 33.2 + 20.734 exp [ 0.02 ( T-RTNOT ) ]

The upper shelf fracture toughness of the low alloy steel is limited to [ ] . For the nozzle forging, the Kie curve without a cut-off limit Is used [22, Section 4.3]. The crack-tip temperature needed for these fracture toughness equations is obtained from the results of a transient thermal analysis, described below.

6.2 Thermal Analysis and Thermal Stress Intensity Factor Through-wall temperature distributions are determined by solving the one-dimensional transient axisymrnetric heat conduction equation, subject to the following boundary conditions:

at the inside surface, where r = Ri, at the outside surface, where r = Ro, a-r

-=O ar where, p= density 6-2 A

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77~2718NP-006 Cp= specific heat k= thermal conductivity T= temperature r= radial coordinate t= time h= convection heat transfer coefficient Tw= wall temperature Tb= bulk coolant temperature

~= inside radius of vessel Ro= outside radius of vessel The above equation is solved numerically using a finite difference technique to determine the temperature at [ ] points through the wall as a function of time for prescribed changes in the bulk fluid temperature, such as multi-rate ramp and step changes for heatup and cooldown transients.

Through-wall thermal stress distributions are determined by trapezoidal integration of the following expression:

Thermal hoop stresses:

2 2 O'e(r)=- Ea 1 ( r + R ; f.R r*

°Trdr+ Ji Trdr-Tr 2 ) [24, Equation (255)]

1-v 2r Ro2 -R;2 R; R; Expressing the thermal stress distributions by

<1(x) = Co + C1 (x/a) + C2 (xla)2 + C3 (xla)3 ,

where, x= is a dummy variable that represents the radial distance from the appropriate (i.e., inside or outside) surface, in.

a= the flaw depth, in.

The thermal stress intensity factors are defined by the following relationships:

For a 1/4 t inside surface flaw during cooldown, Kn = (1.0359 Co + 0.6322 C, + 0.4753 C2 + 0.3855 C3) .Jir;.

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77-2718NP-006 For a 1/ 4 t outside surface flaw during heatup, Kit = (1.043 Co + 0.630 C1 + 0.481 ~ + 0.401 Cs) ./rri.

6.3 Unit Pressure Stress Intensity Factor for Reactor Vessel Beltline Region The membrane stress intensity factor in the reactor vessel shell due to a unit pressure load is where

~ = vessel inner radius, in.

t= vessel wall thickness, in.

For a longitudinal 1/4 -thickness x 3/ 2 -thickness semi-elliptical surface flaw:

at the inside surface, Mm= 1.85 fort <4 in.

= 0.926 t for4sts 12 in.

= 3.21 fort> 12 in.

at the outside sulface, Mm = 1.77 fort< 4

= 0.893 t for4sts 12in.

= 3.09 fort> 12 in.

6.4 Unit Pressure Stress Intensity Factor for Reactor Vessel Nozzles Considering a nozzle as a hole in a shell, WRC Bulletin 175 [25] presents the following method for estimating stress intensity factors for a nozzle comer flaw:

Kim = o./rri. F(a/rn) where Ri = nozzle belt shell inner radius, in.

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77-271 BNP-006 t= nozzle belt shell wall thickness, in.

a= flaw depth, in.

rn = apparent radius of nozzle, in.

= n+ 0.29rc r1= inner radius of nozzle, in.

re= nozzle corner radius, in.

and F(a/rn) = 2.5 - 6.108(alrn) + 12(alrn)2 - 9.1664{alrn)3 6-5 A

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77-271 SNP-006 7.0 PRESSURE CORRECTION 7.1 Correction for Normal Heatup/Cooldown Pressure correction factors between the RCS hot leg pressure taps and various other RCS locations were determined using a hydraulic analysis code to calculate RCS flow rates and pressure losses. These correction factors, listed in Table 7-1 and Table 7-2 £26, Table 4-1 and Table 4-2), were applied to the uncorrected location-based P-T limits, as appropriate, to produce a uniform set of P-T limits which can be complied with by monitoring RCS pressure as indicated by the hot leg pressure taps.

Pressure location correction factors were determined between the hot leg RCS pressure taps and the following RCS locations:

-reactor vessel beltline

-reactor vessel closure head (RVCH)

-reactor vessel inlet and outlet (cold leg and hot leg) nozzles These location correction factors were applied to the uncorrected location-based P-T limits. This produces a uniform set of location adjusted P-T limits (considering both of the hot leg pressure taps).

Table 7-1: Location Pressure Correction Factors for Both A and B Pressure Taps 7-1 A

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77-2718NP-006 Table 7-2: RCS Piping Location Correction Factors 7.2 Correction for ISLH For ISLH conditions, the same correction factors used for the NOT limits will be applied. Therefore, the correction factors shown in Section 7.1 are used for pressure location adjustments for ISLH conditions.

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77-2718NP-006 8.0

SUMMARY

OF RESULTS The pressure temperature limit report (PTLR) P-T limits at 52 EFPY for normal heatup and criticality conditions, normal cooldown, and inservice leak hydrostatic conditions are provided in Figure 8-1 through Figure 8-4 [26). These P-T limits have been developed considering the operational conditions described in Section 5.0. Maintaining the reactor coolant system pressure below the upper limit of the pressure-temperature limit curves ensures protection against non-ductile failure. Acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the applicable P-T limit curves.

The 52 EFPY P-T limit curves have been adjusted based on pressure differential between point of system pressure measurement and the point in the reactor vessel that establishes the controlling unadjusted pressure limit. The P-T limit curves provided in Figure 8-1 through Figure 8-4 do not include margins for instrument error. The reactor is not permitted to be critical until the pressure-temperature combinations are, as a minimum, to the right of the criticality curve (Figure 8-1). The numerical values for the Pressure Temperature Limit Report (PTLR) P-T curves for normal heatup, normal cooldown and ISLH heatup provided in Figure 8-1 through Figure 8-3 are provided in Table 8-1 Hirough Table 8-3. The operational constraints for these curves are tabulated in Table 8-4 and Table 8-5 [18J.

These P-T curves also meet the pressure and temperature requirements for the reactor pressure vessel listed in Table 1 of 10CFR Part 50, Appendix G [1]. The P-T heatup limits for DB-1 are provided in Table 8-1. The criticality limit temperature, Table 8-2, is 226.0 °F. The P-T limits for normal cooldown for DB-1 are generated as the limiting allowable pressure at every calculated temperature as shown in Table 8-3. The P-T limits for ISLH heatup and cooldown are provided in Table 8-6 for ISLH heatup, Table 8-7 for ISLH cooldown, and presented graphically in Figure 8-3 and Figure 8-4.

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77-271 BNP-006 Table 8-1: PTLR Basis Normal Heatup P-T Limits Fluid Governing Adjusted Pressure Temperature HU Rate:75°F/hr HU Rate:SOOF/hr (Of) (psig) (psig) 70 519 519 75 519 519 80 519 519 85 519 519 90 519 519 95 519 519 100 519 519 105 519 519 110 519 519 115 519 519 120 519 519 125 519 519 130 519 519 135 519 519 140 519 519 145 519 519 150 519 519 150 704 758 155 739 802 160 778 850 165 820 902 170 866 960 175 917 1024 180 973 1095 185 1035 1174 190 1102 1260 195 1177 1355 200 1258 1460 205 1349 1576 210 1448 1705 215 1557 1846 220 1677 . 2002 225 1810 2175 230 1956 2365 235 2116 2575 8-2 A

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77-2718NP-006 Fluid Governing Adjusted Pressure Temperature HU Rate:75°F/hr HU Rate:S0°F/hr (Of) (psig) (psig) 240 2293 2806 245 2487 2960 250 2702 2960 255 2938 2960 260 2960 2960 265 2960 2960 270 2960 2960 275 2960 2960 280 2960 2960 285 2960 2960 290 2960 2960 295 2960 2960 300 2960 2960 305 2960 2960 310 2960 2960 315 2960 2960 320 2960 2960 325 2960 2960 330 2960 2960 335 2960 2960 340 2960 2960 345 2960 2960 350 2960 2960 355 2960 2960 360 2960 2960 365 2960 2960 370 2960 2960 375 2960 2960 380 2960 2960 385 2960 2960 390 2960 2960 395 2960 2960 400 2960 2960 405 2960 2960 410 2960 2960 415 2960 2960 420 2960 2960 425 2960 2960 430 2960 2960 8-3 A

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77-271 BNP-006 Fluid Governing Ad] usted Pressure Temperature HU Rate:7S°F/hr HU Rate:SOOF/hr (Of) (psig) (psig) 435 2960 2960 440 2960 2960 445 2960 2960 450 2960 2960 455 2960 2960 460 2960 2960 465 2960 2960 470 2960 2960 475 2960 2960 480 2960 2960 485 2960 2960 490 2960 2960 495 2960 2960 500 2960 2960 505 2960 2960 510 2960 2960 515 2960 2960 520 2960 2960 525 2960 2960 530 2960 2960 535 2960 2960 540 2960 2960 545 2960 2960 550 2960 2960 8-4 A

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77-2718NP-006 Table 8-2: PTLR Basis Criticality Limit P-T Limits (a) Criticality Limit Determination Criticality Umlt Temperature at 2500 pslg during ISLH Heatup at 75°F/hr Pressure Temperature (pslg) (Of) 2451 225 2645 230 Interpolated values:

2500 226 (b) Criticality limit P-T Limits Governing Fluid Adjusted Temp Pressure (OF) (pslg}

226 0 226 1048 230 1102 235 1177 240 1258 245 1349 250 1448 255 1557 260 1677 265 1810 270 1956 275 2116 280 2293 285 2487 290 2702 295 2938 300"' 2960

"'Table truncated at 300°F (fluid temp.),

since the pressure already surpasses the normal operating pressure.

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77-2718NP-006 Table 8-3: PTLR Basis Cooldown P-T Limits Governing Adjusted Pressure CD Rate: CD Rate:

1000F/hrup 100°F/hr up Fluid to 270°F, then to 270°F, then Temperature at SOOF/hr at 25°F/hr (Of) (psig) (pslg) 70 418 475 75 467 496 80 478 518 85 519 519 90 519 519 95 519 519 100 519 519 105 519 519 110 519 519 115 519 519 120 519 519 125 519 519 130 519 519 135 519 519 140 519 519 143 519 519 148 519 519 150 519 519 150 1054 1054 153 1098 1093 158 1184 1184 163 1277 1277 168 1378 1374 173 1494 1494 178 1620 1620 183 1757 1752 188 1913 1913 193 2083 2083 198 2269 2264 203 2438 2438 208 2438 2438 213 2438 2438 218 2438 2438 223 2438 2438 8-6 A

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77-2718NP-006 Governing Adjusted Pressure CD Rate: CD Rate:

l0OOF/hrup 100°F/hr up Fluid to 2700F, then to 270°F, then Temperature at50°F/hr at25°F/hr

(°F} (psig) (pslg) 228 2438 2438 233 2438 2438 238 2438 2438 243 2438 2438 248 2438 2438 253 2438 2438 258 2438 2438 263 2438 2438 259 2438 2438 254 2438 2438 249 2443 2443 270 2451 2451 275 2453 2453 280 2455 2455 285 2457 2457 290 2459 24S9 295 2461 2461 300 2463 2463 305 2466 2466 310 2469 2469 315 2470 2470 320 2473 2473 32S 2477 2477 330 2481 2481 335 2484 2484 340 2489 2489 345 2493 2493 350 2498 2498 355 2502 2502 360 2507 2507 365 2512 2512 370 2512 2512 375 2519 2519 380 2526 2526 385 2533 2533 390 2540 2540 8-7 A

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77-271 SNP-006 Governing Ad'usted Pressure CD Rate: CD Rate:

100°F/hr up 100°F/hrup Fluid to 2700F, then to 2700F, then Temperature atSOOF/hr at25°F/hr (Of) (pslg) (psig) 395 2548 2548 400 2556 2556 405 2564 2564 410 2572 2572 415 2581 2581 420 2590 2590 425 2600 2600 430 2610 2610 435 2620 2620 440 2635 2635 445 2641 2641 450 2652 2652 455 2664 2664 460 2676 2676 465 2689 2689 470 2702 2702 475 2716 2716 480 2730 2730 485 2745 2745 490 2760 2760 495 2776 2776 500 2792 2792 505 2809 2809 510 2827 2827 515 2845 2845 520 2863 2863 525 2882 2882 530 2901 2901 535 2919 2919 540 2937 2937 545 2952 2952 550 2960 2960 8-8 A

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77-271 SNP-006 Table 8-4: Operational Constraints for Plant Heatup CONSTRAINT RC TEMPERATURE HEATUP RATE ALLOWED PUMP

(*F) ("F/hour) COMBINATION RC Temperature T~70 Ramp at 75 °F/hour; or, 15°F steps with 12 minute hold times NA or Ramp at 50°F/hour; or, 15°F steps with 18 minute hold times 70.::_T<140 0/0 RC Pumps 140 <T<270 210, 0/2 270.::_T ,::440 2/1, 1/2 or 210, 0/2 440<T,::550 NA 212 or 2/1, 1/2 Table 8-5: Operational Constraints for Plant Cooldown CONSTRAINT RC TEMPERATURE ("F) COOLDOWN RATE ALLOWED PUMP

("F/hour) COMBINATION RC Temperature T~270 Ramp at 100°F/hour; or, 15°F steps with 9 minute hold times Ramp at 50°F/hour; or 1S"F 270>T~70 steps with 18 minute hold times NA or Ramp at 25°F/hour; or, 15°F steps with 36 minute hold times RC Pumps 70,::T<140 0/0 140.:: T < 270 2/0, 0/2 270,::T.::_440 2/1, 1/2 or 2/0, 0/2 440<T~550 NA 2/2 or2/1, 1/2 8-9 A

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77-271 SNP-006 Figure s..1: PTLR Normal Heatup and Criticality Core P-T Limits 2400 2200 2000 1800

!I 1600 1111

'iii a.

,; 1400

I Ill Ill 1200 I!!

D-1000 800 II ,I :'

600 -,-~-+---.:..-i~-----+--+---+---ii**--+-+---

l Ht*.:. .:. .:. 1 I I 400 I I !

I I I i 50 I:

100 I

200 _ ___.__~,*__._.....__ _ __,_i-~-------'------------1-~-'--~------'---------

150 200 250 300 Temperature, °F 8-10 A

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77-271 SNP-006 Figure 8-2: PTLR Normal Cooldown P-T Limits 2400 -i; ----i-~ . ! . .

! I i

-.!,- Tech SpecCun,e for Normal CD, lOO'F/hr upto 270"F, then SO'F/hr 2200

- a - Tech Spec curve tor Nonnal CD, l0O'f/hr upto 210"F, then 25"F/hr 2000 1800 1600

.!!I VI a.

...ti 1400 i

Ill 1200 CII k.

1000 800 600

. : i

~t*a-a~a-A-4MH!}-~!~Ma

/ I 400 200 +----'----'----'----'----+-----'---'---'---+-----'----'-------+-----'---'--........- ........_~

50 100 150 200 250 Temperature, °F 8-11 A

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77~2718NP-006 Tabl e 8-6 PTLR Bass i ISLH Heatup PT - LI mIts Governing Adjusted Pressure Fluid HU HU Temperature Rate:75°F/hr Rate:SOOF/hr

{°F) (psig) (psig) 70 519 519 75 519 519 80 519 519 85 519 519 90 519 519 95 519 519 100 519 519 105 519 519 110 519 519 115 519 519 120 519 519 125 519 519 130 519 519 135 . 519 519 140 519 519 145 519 519 150 519 519 150 976 1048 155 1023 1107 160 1075 1170 165 1131 1241 170 1193 1318 175 1261 1404 180 1335 1498 185 1417 1602 190 1507 1717 195 1606 1845 200 1716 1985 205 1836 2140 210 1968 2311 215 2114 2499 220 2274 2708 225 2451 2937 230 2645 3191 23S 2859 3471 240 3095 3779 8-12 A

AREVA

77-271 SNP-006 Governing Adjusted Pressure Fluid HU HU TemDerature Rate:75°F/hr Rate:SOOF/hr (Of) (psig) (psig) 245 3354 3991 250 3640 3991 255 3954 3991 260 3991 3991 265 3991 3991 270 3991 3991 275 3991 3991 280 3991 3991 285 3991 3991 290 3991 3991 295 3991 3991 300 3991 3991 305 3991 3991 310 3991 3991 315 3991 3991 320 3991 3991 325 3991 3991 330 3991 3991 335 3991 3991 340 3991 3991 345 3991 3991 350 3991 3991 355 3991 3991 360 3991 3991 365 3991 3991 370 3991 3991 375 3991 3991 380 3991 3991 385 3991 3991 390 3991 3991 395 3991 3991 400 3991 3991 405 3991 3991 410 3991 3991 415 3991 3991 420 3991 3991 425 3991 3991 430 3991 3991 8-13 A

AREVA

77-271 BNP-006 Governing Adjusted Pressure Fluid HU HU Temperature Rate:75°F/hr Rate:SD°F/hr (OF) (pslg) (pslg) 435 3991 3991 440 3991 3991 445 3991 3991 450 3991 3991 455 3991 3991 460 3991 3991 465 3991 3991 470 3991 3991 475 3991 3991 480 3991 3991 485 3991 3991 490 3991 3991 495 3991 3991 500 3991 3991 sos 3991 3991 510 3991 3991 515 3991 3991 520 3991 3991 525 3991 3991 530 3991 3991 535 3991 3991 540 3991 3991 545 3991 3991 550 3991 3991 8-14 A

AREVA

77-2718NP-006 Table 8-7: PTLR Basis ISLH Cooldown P-T Limits Govemlng Adjusted Pressure CO Rate: CD Rate:

100°F/hrup 100°F/hr up to Fluid to 270°F, then 270°F, then at Temperature at50°F/hr 2S°F/hr

(°FI (psig) (psig) 70 519 519 75 519 519 80 519 519 85 519 519 90 519 519 95 519 519 100 519 519 105 519 519 110 519 519 115 519 519 120 519 519 125 519 519 130 519 519 135 519 519 140 519 519 143 519 519 148 519 519 150 519 519 150 1443 1443 153 1501 1496 158 1616 1616 163 1741 1741 168 1875 1869 173 2030 2030 178 2198 2198 183 2381 2374 188 2588 2588 193 2814 2814 198 3064 3056 203 3288 3288 208 3288 3288 213 3288 3288 8-15 A

AREVA

77-2718NP-006 Governing Ad usted Pressure CD Rate: CD Rate:

100°F/hrup lOOoF/hr up to Fluid to 2700F, then 2700F, then at Temperature at 50°F/hr 25°F/hr (Df) (DSlg) (pslg) 218 3288 3288 223 3288 3288 228 3288 3288 233 3288 3288 238 3288 3288 243 3288 3288 248 3288 3288 253 3288 3288 258 3288 3288 263 3288 3288 259 3288 3288 254 3289 3289 249 3295 3295 270 3311 3311 275 3313 3313 280 3316 3316 285 3318 3318 290 3321 3321 295 3324 3324 300 3327 3327 305 3331 3331 310 3334 3334 315 3336 3336 320 3340 3340 325 3345 3345 330 3350 3350 335 3355 3355 340 3361 3361 345 3367 3367 350 3373 3373 355 3379 3379 360 3386 3386 365 3393 3393 370 3393 3393 375 3401 3401 380 3410 3410 8-16 A

AREVA

77-2718NP-006 Governing Adjusted Pressure CD Rate: CD Rate:

100°F/hrup 100oF/hr up to Fluid to 270°F, then 27D°F, then at Temperature at50°F/hr 25°F/hr (Of) (pslg) (psig) 385 3420 3420 390 3429 3429 395 3440 3440 400 3450 3450 405 3461 3461 410 3472 3472 415 3484 3484 420 3497 3497 425 3509 3509 430 3523 3523 435 3537 3537 440 3557 3557 445 3566 3566 450 3581 3581 455 3596 3596 460 3613 3613 465 3630 3630 470 3647 3647 475 3665 3665 480 3684 3684 485 3704 3704 490 3724 3724 495 3746 3746 500 3768 3768 505 3790 3790 510 3814 3814 515 3838 3838 520 3862 3862 525 3887 3887 530 3912 3912 535 3937 3937 540 3960 3960 545 3980 3980 550 3990 3990 8-17 A

AREVA

77-2718NP-006 Figure 8-3: PTLR ISLH Heatup P-T Limits 2750 i . --,-

1 i

2250

,~-+-----,----+---'---+----+-~

- -:- -ti~--

2000 I I

I

- --~- *<I i

+ * -i * ** *-----

-*---*c* - -

i 1750 I

-- -i

. i 1111 i

~ * ' I Cl, ; ' i I 1500 - --t--t---;---+--+----*- -- ~ - - + - - - - - t - - - + - - ----,l/f-~19---+--~-~--;--;--- .

,,; I '

i I

' ' i

i Ill l I r--1--- : !

e i -

Ill 1250 ~

a. I [ ' I .

1000 i !

i l 750 . I 500 -- r.el~. .

--J-_B,o""S'!!5~~:,..;i~,-,,..;...

iI

,,,,;.,s,i,-,s,~,a.,B,e!S~:i-__L;_ _ _ _,__-+_ _;__t'___1,____,

250 '

0 so 100 150 200 250 Temperature, °F 8-18 A

AREVA

77-271 BNP-006 Figure 8-4: PTLR ISLH Cooldown P-T Limits 2750

~~1--~1~----r*--n_ l _l--+----1---]-

1 2500 I

1 , : , i *

w, 100-F/hr upto ZlO"F, thtn SO"F/hr I '

i tu f l; 1 i

I

-*-+' . ,I '

-&**Tech Spec: P-T Umlu for ISU1 2250 - e - Tech Spec P-T Umlts for ISLH m, IOO"F/hr upto 270-F, then 2S*F/hr . +---i ~

---+-

I 1_____; _ ~ - ; - - - - - - ;

i /

, ' ' I

j' i ; I ! . ,

! :

I ' * : ' I * ;

7&_***-r **---+-*--** **--**t**---~---*--;-**- **-, -**-***

t' 2000 *-- .. ' -* - - --- ,* r, . .. --- --- - t - ... - ----*-- -- *--*;---

-**-r-** *--**+-***** ----

iI :: / la !, :. i,

-*t---*

II 1 1***--*, .-*** -*--*f*-----*-***-*--;

1750 .... 1 :.,a-* **** ...

11111 iii i J'! '. ' i. ,'

. ii i *. / ; '

1' Q.

1500 Cl1

~

~ 1250 + - - + -- ----+--+---,--+---"-----+--'1--~---,-----.---* - * - - ~--+--+-----1 f

Cl.

1000 I

' l 250 _ _i _-i-----li--+--+-~-----;...--+-...;__ _ _ _..;.............C.--!-- -J - - - ~ -~---1' - ~

I 0 -~

50

_ _.I__ -~iI

~!

' --t-~-~-----'-i--+---------+--~i--~--'-!

100 150 200

---t 250 Temperature, °F 8.1 P-T limits with Low Temperature Overpressure Protection (LTOP)

Development of LTOP Enable Temperature:

The P-T limits with LTOP are obtained by taking 100% of the controlling normal operation heatup/cooldown limit. To support the development of an LTOP system effective temperature, the temperature difference is calculated between the 1/ 4t location of the RV wall and the RCS fluid for the condition when the metal temperature is at [ ] [22, Section 6.0],

as required by Article G-2215 of Section XI [9] where [ ] is the highest adjusted reference temperature from Table 5-3 and 50 °Fis a margin term. Temperature differences between the reactor coolant and the vessel wall are reported in Table 8-8 [22, Table 6-2] for heatup conditions when the metal temperature is lower and significantly lags the coolant temperature. During 8-19 A

AREVA

77-2718NP-006 cooldown the coolant temperature is always lower than the metal temperature and therefore It is not limiting in the development of the LTOP system effective temperature.

The information in Table 8-8 [22, Table 6-2] is provided at the 1/..t wall location, relative to the inside surface, where t is the thickness of the base metal. The corresponding reactor coolant temperature during the heatup transient is 225.0 °F, resulting in a temperature difference of [

]. The minimum LTOP enable temperature is therefore 225.0 °F, plus any adjustments for instrument error [22]. Using a measurement uncertainty correction of 18°F, the corrected LTOP enable temperature is 243°F. Since 243°F is less than the existing LTOP enable temperature of 280°F, the existing value is acceptable with no changes.

Development of LTOP Set Point:

Table 8-8: Reactor Coolant Temperatures for Use In Establlshlng an LTOP System Effective Temperature [22, Table 6-2]

8-20 A

AREVA

77-2718NP-006 8.1.1 ITS Figures 3.4.12-1 and Figure 3.4.12-2 As discussed in Section 3.0, Improved Technical Specification (ITS) Figure 3.4.12-1, RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 4, and Figure 3.4.12-2, RCS Pressure Versus Pressurizer Level Limit for Inoperable OHR System Relief Valve in MODE 5 and MODE 6 when the reactor vessel head is on, provide operating restrictions in Modes 4 through 6 to protect against a postulated LTOP event (i.e., failed open MU valve) with an inoperable DHR System relief valve (single failure) [27]. The pressurizer level restrictions ensure that the P-T limits reported in the PTLR are not exceeded during the assumed [ ] of the LTOP event (I.e., charging injection depletes the MU tank), such that the most limiting P-T HU/CD limit is not exceeded in [ ].

Improved Technical Specification (ITS) Figure 3.4.12-1, "RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 4", and Figure 3.4.12-2, "RCS Pressure Versus Pressurizer Level Limit for Inoperable OHR System Relief Valve in MODE 5 and MODE 6 when the reactor vessel head is on", provide operating restrictions in Mode 4 through Mode 6 to protect against a postulated LTOP event (i.e., failed open MU valve) with an inoperable DHR System relief valve (single failure). The pressurizer level restrictions ensure that the P-T limits reported in the PTLR are not exceeded during the assumed [

] . ITS Figure 3.4.12-1 and Figure 3.4.12-2 are based on corrected cooldown P-T limits at 1O EFPY. Since the 1O EFPY pressure limits are more restrictive than the 52 EFPY pressure limits calculated herein, ITS Figures 3.4.12-1 and 3.4.12-2 remain unchanged for 52 EFPY.

8.2 P-T Limits for RV Shell Transition Regions 52 EFPY P-T limit curves were generated for the upper and lower thickness transition regions of the DB-1 Reactor Vessel to confirm that these regions, which contain structural discontinuities, do not result in P-T limits that are more limiting than the P-T limits reported In Figure 8-1 and Figure 8-2. [

) . This confirms the statements regarding transitions welds reported in the DB-1 LRA, Amendment 33 (See Section 3.0 above), and weld WF-182-1 continues to be the limiting beltline material relative to establishment of P-T limits for Davis-Besse.

8-21 A

AREVA

77-271 BNP-006 8.3 Lowest Service Temperature In accordance with BAW-10046A, Revision 2 [11 ], Section 4.1, the components of the RC system in a typical B&W power plant have been analyzed to determine the minimum required reactor coolant temperature for pressures of 626, 2250, and 3125 psig. The 626 psig pressure was selected because it is 1 psig above the pressure corresponding to 20% of the preoperational system hydrostatic test pressure. This is the maximum allowable pressure (625 psig) for a component when the reactor coolant temperature (or the volumetric average metal temperature) is below the lowest service temperature of the component. The components for which a lowest service temperature must be defined include the RC loop piping and the control rod drive mechanism (the CRDM is an appurtenance to the reactor vessel). The lowest service temperature of these components is 150 °F (based on RTNDT + 100F) for RCS ferritic piping and 100 °F for the CROM. The location correction factor for RCS piping is [ ] [26, Table 4-2], for RCS temperatures at 150F; therefore, the allowable RCS pressure at and below 150 Fis 625 psig minus [ ] psig, considering LST for RCS piping. The corrected pressure-temperature curves reported in Figures 8-1 and 8-2 consider the minimum permissible pressure by considering LST for RCS components (i.e., RCS piping is [

] and below) and location adjusted permissible pressures calculated in accordance with ASME Section XI, Appendix G.

The DB-1 52 EFPY P-T limits are in compliance with BAW-10046A relative to lowest service temperature for RCS components [29].

8-22 A

AREVA

Framatome Inc. ANP-2718NP Revision 007 Appendix G Pressure-Temperature Limits For 52 EFPY for Davis-Besse Nuclear Power Station - First Energy Nuclear Operating Company a e 9-9.0 CERTIFICATION Pressure/temperature limits for the DB-1 reactor vessel have been calculated to satisfy the requirements of 10 CFR Part 50, Appendix G using analytical methods and acceptance criteria of the ASME Boiler and Pressure Vessel Code,Section XI, AppendixG.

~~ngineer~

Plant Operations and License Renewal This report has been reviewed for technical content and accuracy.

Lead Reviewer (all Sections) ~~- 'l--/4'6'/1q 1 Ashok D. Nana, Advisory Engineer Date Component Analysis, Fracture & Materials Unit Verification of independent review This report is approved for release Beverly J. Watson Date Project Manager

77-271 SNP-006

10.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50 - Domestic Licensing of Production and Utilization Facilities, Appendix G - Fracture Toughness Requirements, Federal Register Vol. 60. No. 243, December 19, 1995.

2. *Davis Besse Unit 1 Improved Technical Specifications, Revision 330

3. DB-1 License Renewal Application Davis-Besse Nuclear Power Station-License Renewal Application and Ohio Coastal Management Program Consistency Certification.

(Accession No. ML102450565)

4. Letter from Allen B. S., FENOC: Reply to Supplemental Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640). (Accession No. ML12240A219)

5. Letter from Lieb R. A., FENOC: Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640). (Accession No. ML13156A388)

6. NUREG-2193, Volume 1, Volume 2, and Supplement 1, Adams accession numbers ML16104A207, ML16104A301, and ML16104A350

7.

Davis Besse Unit1 Pressure and Temperature Limits Report for Up To 32 EFPY, Revision 3, June 2016

8. Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, December 19, 1995.

9. American Society of Mechanical Engineers (ASME) Boller and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," and Section 11-D, "Section 11-D Properties (customary) Materials"2007 Edition with Addenda through 2008.

10. ASTM Standard E 208-81, "Standard Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials, Philadelphia, Pennsylvania.

11. AREVA Document BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Appendix G," by H. W. Behnke et al.

12. NRC Regulatory Issue Summary 2004-04: Use of Code Cases N-588, N-640, and N-641 in Developing Pressure-Temperature Operating Limits, ADAMS Accession Number ML040920323, dated April 5, 2004.

13. [

]

10-1

77-2718NP-006

14. [ ]

15. [

]

16. [

]

17. [ ]

18. [

]

19. [

]

20. AREVA document 77-2431-004, LOW UPPER-SHELF TOUGHNESS FRACTURE MECHANICS ANALYSIS OF REACTOR VESSEL OF DAVIS-BESSE FOR 32 EFFECTIVE FULL POWER YEARS and 52 EFFECTIVE FULL POWER YEARS.

21. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI. "Rules for lnservice Inspection of Nuclear Power Plant Components," and Section 11-D, "Section 11-D Properties (customary) Materials, " 2013 Edition

22. [ 1

23. [

]

24. Timoshenko, S.P., and Goodier, J.N., Theory of Elasticity, Third Edition, McGraw-Hill Book Company, 1970.

25. PVRC Ad Hoc Group on Toughness Requirements, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," Bulletin No. 175, Welding Research Council, August 1972.

26. [

1

27. [

]

28. [

]

10-2

77-271 SN P-006

29. [

]

30. AREVA Document 43-2308-04 (BAW-2308, Revision 2-A), "Initial RTNDT of Linde 80 Weld Materials,"

- References identified with an (*) are maintained within the FENOC Records System and are not retrievable from AREVA Records Management.

10-3

Enclosure B L-22-194 Affidavit Pursuant to 10 CFR 2.390 for 32-9271138-003 (3 Pages Follow)

AFFIDAVIT

1. My name is Gayle Elliott. I am Deputy Director, Licensing and Regulatory Affairs, for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in Calculation Summary Sheet 32-9271138-003, entitled "Updated Inputs to 52 EFPY P-T Operating Curves," dated August 2022 and referred to herein as "Document." Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatome's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(d) and 6(e) above.

7. In accordance with Framatome's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: August 17, 2022 g::~~

Digitally signed by ELLIOTT ELLIOTT Gayle 2022.08.171s:28:07

-04'00' Gayle Elliott

Enclosure C L-22-194 Framatome Inc. document 32-9344638-001 Updated inputs to 52 EFPY P-T Operating Curves. Nonproprietary (32 Pages Follow)

0402-01-F01 (Rev. 021, 03/12/2018)

PROPRIETARY CALCULATION

SUMMARY

SHEET (CSS)

Document No. 32 - 9344638 - 001 Safety Related: Yes No Title Updated Inputs to 52 EFPY P-T Operating Curves PURPOSE AND

SUMMARY

OF RESULTS:

The purpose of this calculation is to provide a Non-Proprietary version of Reference [ 1].

The purpose of Reference [1] is to generate inputs to the procedural NDT/LTOP limits using the location corrected NDT/LTOP PT curves from Reference [ 2] and the uncertainties in pressure and temperature measurement from Reference [ 3].

The results of Reference [1] are presented in Section 4.

The purpose of Revision 001 of this calculation is to update the revision level of Reference 1.

FRAMATOME INC. PROPRIETARY This document and any information contained herein is the property of Framatome Inc. (Framatome) and is to be considered proprietary and may not be reproduced or copied in whole or in part. This document shall not be furnished to others without the express written consent of Framatome and is not to be used in any way which is or may be detrimental to Framatome. This document and any copies that may have been made must be returned to Framatome upon request.

If the computer software used herein is not the latest version per the EASI list, THE DOCUMENT CONTAINS AP 0402-01 requires that justification be provided.

ASSUMPTIONS THAT SHALL BE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: VERIFIED PRIOR TO USE CODE/VERSION/REV CODE/VERSION/REV Yes No Page 1 of 32

0402-01-F01 (Rev. 021, 03/12/2018)

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Review Method: Design Review (Detailed Check)

Alternate Calculation Does this document establish design or technical requirements? YES NO Does this document contain Customer Required Format? YES NO Signature Block P/R/A/M Name and Title and Pages/Sections (printed or typed) Signature LP/LR Date Prepared/Reviewed/Approved Trevor Connell, TD CONNELL LP ALL Engineering Supervisor 8/19/2022 Steven Claunch, SL CLAUNCH LR ALL Advisory Engineer 8/19/2022 Darren Wood, DH WOOD A ALL Engineering Manager 8/19/2022 Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A);

LP/LR designates Lead Preparer (LP), Lead Reviewer (LR);

M designates Mentor (M)

In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use All or All except

___ in the pages/sections reviewed/approved. All or All except ___ means that the changes and the effect of the changes on the entire document have been prepared/reviewed/approved. It does not mean that the lead preparer/reviewer/approver has prepared/reviewed/approved all the pages of the document.

With Approver permission, calculations may be revised without using the latest CSS form. This deviation is permitted when expediency and/or cost are a factor. Approver shall add a comment in the right-most column that acknowledges and justifies this deviation.

Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)

Name Title (printed or typed) (printed or typed) Signature Date Comments Beverly Watson Project Manager BJ WATSON 8/19/2022 Page 2

0402-01-F01 (Rev. 021, 03/12/2018)

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Record of Revision Revision Pages/Sections/Paragraphs No. Changed Brief Description / Change Authorization 000 ALL Non-Proprietary version of 32-9271138-001.

001 Section 5.0 References Update the revision level of Reference 1 to 32-9271138-003.

Page 3

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table of Contents Page SIGNATURE BLOCK ............................................................................................................................. 2 RECORD OF REVISION ....................................................................................................................... 3 LIST OF TABLES .................................................................................................................................. 5 LIST OF FIGURES ................................................................................................................................ 6 1.0 PURPOSE .................................................................................................................................. 7 2.0 ANALYTICAL METHODOLOGY................................................................................................. 7 3.0 ASSUMPTIONS ......................................................................................................................... 8 3.1 Unverified Assumptions ................................................................................................... 8 3.2 Justified Assumptions ...................................................................................................... 8 4.0 CALCULATIONS ........................................................................................................................ 9 4.1 NDT Limit Curves ............................................................................................................ 9 4.2 LTOP Limit Curves ........................................................................................................ 22 4.3 Surge Line Limit ............................................................................................................ 23

5.0 REFERENCES

......................................................................................................................... 32 Page 4

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves List of Tables Page Table 2-1: Summary of Inputs ............................................................................................................... 7 Table 2-2: Nomenclature ...................................................................................................................... 8 Table 4-1: Normal Heatup (PI-RC2A4 or PI-RC2B4) .......................................................................... 10 Table 4-2: ISLH Heatup (PI-RC2A4 or PI-RC2B4) .............................................................................. 13 Table 4-3: Normal Cooldown (PI-RC2A4 or PI-RC2B4) ...................................................................... 16 Table 4-4: ISLH Cooldown (PI-RC2A4 or PI-RC2B4) .......................................................................... 19 Table 4-5: 10 EFPY and 52 EFPY P-T Limit Comparison ................................................................... 23 Table 4-6: Surge Line Limits ............................................................................................................... 23 Page 5

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves List of Figures Page Figure 4-1: Normal Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) ..................................... 24 Figure 4-2: Normal Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) ..................................... 25 Figure 4-3: ISLH Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) .......................................... 26 Figure 4-4: ISLH Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) ......................................... 27 Figure 4-5: Normal Cooldown NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) .............................................................................................................................. 28 Figure 4-6: Normal Cooldown NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) .............................................................................................................................. 29 Figure 4-7: ISLH Cooldown NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) .............................................................................................................................. 30 Figure 4-8: ISLH Cooldown NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4) .............................................................................................................................. 31 Page 6

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves 1.0 PURPOSE The purpose of this calculation is to provide a Non-Proprietary version of Reference [1]. The content is identical to Reference [1], which includes the original revision levels of the calculation references at the time the document was approved.

The purpose of Reference [1] is to generate inputs to the procedural NDT/LTOP limits using the location corrected NDT/LTOP PT curves from Reference [2] and the uncertainties in pressure and temperature measurement from Reference [3].

The results of Reference [1] are presented in Section 4.

2.0 ANALYTICAL METHODOLOGY The calculation is performed by applying the appropriate pressure and temperature measurement uncertainties from Reference [3] to the location corrected PT limits developed in Reference [2]. Table 2-1 presents a summary of the inputs used in this calculation.

Table 2-1: Summary of Inputs Parameter Value Source Wide Range Pressure [ ] Page 34 of Reference [3] and Measurement Uncertainty Read @ Page 12 of Reference [4]

Pressure Taps PI-RC2A4 or PI-RC2B4 Low Range Pressure Measurement [ ] Page 34 of Reference [3] and Uncertainty Read @ Pressure Tap Page 12 of Reference [4]

PI-RC2A6.

Temperature Measurement 18°F Page 34 of Reference [2]

Uncertainty Read @ Wide Range Temperature TIRC4A2 and TIRC4B2 DH-4849 Lift Pressure 330 psig Page 32 of Reference [3]

(Uncorrected)

LTOP Enable Temperature 280°F Page 32 of Reference [3]

(Uncorrected)

LTOP Allowable Pressure Limit [ ] Section 6.2 of Reference [2]

(RV @ 70°F, Uncorrected)

Minimum LTOP Enable 243°F Section 6.2 of Reference [2]

Temperature (Corrected)

Largest P Between Locations of [ ] Section 6.2 of Reference [2]

Interest in the RCS and DH-4849 when TRCS is below 140°F Technical Specification Basis P=f(T) Table 6-5 of Reference [2]

Normal HU PT Curves (Location Corrected)

Page 7

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Parameter Value Source Technical Specification Basis P=f(T) Table 6-7 of Reference [2]

Normal CD PT Curves (Location Corrected)

Technical Specification Basis P=f(T) Table 6-12 of Reference [2]

ISLH HU PT Curves (Location Corrected)

Technical Specification Basis P=f(T) Table 6-13 of Reference [2]

ISLH CD PT Curves (Location Corrected)

Maximum LCF from PI-RC2A6 to [ ] Case DB-D11.OUT, DH-4849 with Three or Fewer Table 4-15 of Reference [5]

RCPs Operational Table 2-2: Nomenclature Acronym Definition CD Cooldown DHR Decay Heat Removal HU Heatup ISLH In-Service Leak and Hydrostatic (Pressure Test)

LCF Location Correction Factor LTOP Low Temperature Over-Pressure NDT Nil-Ductility Transition PT Pressure-Temperature RCP Reactor Coolant Pump RCS Reactor Coolant System 3.0 ASSUMPTIONS The assumptions employed during the development of the inputs to the procedural NDT/LTOP limits calculated herein, if any, are described below.

3.1 Unverified Assumptions There are no unverified assumptions requiring verification prior to using the results of this calculation.

3.2 Justified Assumptions

1. The starting temperature for plant heatup and the ending temperature for plant cooldown are assumed to be 70oF. [

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves

2. [

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4.0 CALCULATIONS The development of the procedural NDT/LTOP limits inputs is presented below. Unless otherwise noted, all inputs (and sources) used herein are summarized in Table 2-1.

4.1 NDT Limit Curves The inputs to the procedural NDT limits are developed by applying the pressure and temperature measurement uncertainties to the location corrected values. [

] These calculations are presented in Table 4-1 through Table 4-4 for the normal HU, ISLH HU, normal CD, and ISLH CD, respectively. These data are also presented in Figure 4-1 through Figure 4-8. Per Justified Assumption 1, the HU/CD transients are considered to begin/end at 70oF for the transients with the temperature uncertainty applied.

Unmodified starting/ending temperatures were left as-is (i.e., @ 70oF).

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table 4-1: Normal Heatup (PI-RC2A4 or PI-RC2B4)

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Note: Refer to 77-2718NP-006/007 Table 8-6 for Non-Proprietary data.

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table 4-2: ISLH Heatup (PI-RC2A4 or PI-RC2B4)

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Note: Refer to 77-2718NP-006/007 Table 8-3 for Non-Proprietary data.

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table 4-3: Normal Cooldown (PI-RC2A4 or PI-RC2B4)

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Note: Refer to 77-2718NP-006/007 Table 8-3 for Non-Proprietary data.

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table 4-4: ISLH Cooldown (PI-RC2A4 or PI-RC2B4)

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Note: Refer to 77-2718NP-006/007 Table 8-7 for Non-Proprietary data.

[

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves 4.2 LTOP Limit Curves The LTOP input to the procedural curves is developed below, and is plotted with the NDT limits presented in Figure 4-1 through Figure 4-8. [

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[

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[

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Improved Technical Specification (ITS) Figure 3.4.12-1, RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 4, and Figure 3.4.12-2, RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 5 and MODE 6 when the reactor vessel head is on, provide operating restrictions in Mode 4 through Mode 6 to protect against a postulated LTOP event (i.e., failed open MU valve) with an inoperable DHR System relief valve (single failure). The pressurizer level restrictions ensure that the P-T limits reported in the PTLR are not exceeded during the assumed [ ] of the LTOP event [ ]. ITS Figure 3.4.121 and Figure 3.4.12-2 are based on corrected cooldown P-T limits at 10 EFPY presented in Figure 5 of Reference [8]. Table 4-5 presents these data alongside the corresponding 52 EFPY data for comparison [

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Table 4-5: 10 EFPY and 52 EFPY P-T Limit Comparison Since the 10 EFPY pressures in Reference [8] are more restrictive than the 52 EFPY pressure limits calculated in Reference [2] and presented in Table 4-3, ITS Figures 3.4.12-1 and 3.4.12-2 remain unchanged for 52 EFPY.

4.3 Surge Line Limit The surge line limits, which are taken from Figure 8-1 of Reference [9], are added for convenience to Figure 4-1 through Figure 4-8 and tabulated below. These data are not corrected herein for location or measurement uncertainty.

Table 4-6: Surge Line Limits Page 23

Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Figure 4-1: Normal Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4)

Note: ITS Figure 3.4.12-1 and Figure 3.4.12-2 control RCS pressure/temperature if DH-4849 is unavailable. Also, refer to 77-2718NP-006 for the Non-Proprietary Normal Heatup plot without uncertainty.

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Document No. 32-9344638-001 PROPRIETARY Updated Inputs to 52 EFPY P-T Operating Curves Figure 4-2: Normal Heatup NDT Limits, [ ] (PI-RC2A4 or PI-RC2B4)