Text

Ses, Unit 3, License Amendment Request NPF-38-250, Revision to Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years
	ML041620063
Person / Time
Site:	Waterford
Issue date:	10/22/2003
From:	Venable J; Entergy Nuclear Operations, Entergy Nuclear South
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC MC1156, W3F1-2003-0075
	Download: ML041620063 (68)
	v • d • e

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Mc. i5* ' KiMona, LA 70066 AEntergy Tel 504 739 6660 Fax 504 739 6678 jvenabltentergy.corn Joseph E. Venable Vice President, Operations Waterford 3 W3FI-2003-0075 October 22, 2003 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request NPF-38-250, Revision to Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for Waterford Steam Electric Station, Unit 3 (Waterford 3). The proposed change will modify Technical Specifications (TSs) 3.4.1.3, Reactor Coolant System, Hot Shutdown; 3.4.1.4, Reactor Coolant System, Cold Shutdown - Loops Filled; 3.4.8.1, Pressure/Temperature Limits; and TS 3.4.8.3, Overpressure Protection Systems.

The existing pressure/temperature (P/T) limits will be changed from 16 to 32 effective full power years (EFPY). In addition, the maximum heatup rate will be changed to 600F per hour (F/hr) and the maximum cooldown rate will be changed to 1000F/hr for all reactor coolant system (RCS) temperatures. For inservice hydrostatic pressure and leak testing the maximum heatup rate will be changed to 60 0 F/hr and the maximum cooldown rate will be changed to 100 0 F/hr.

The P/T limit curves were generated based on the latest available reactor vessel information and updated calculated fluences, including consideration of the measurement uncertainty recapture power uprate that started at the beginning of Cycle 12 and the extended power uprate that is planned at the start of Cycle 14.

The proposed change is based on the analysis of the second reactor vessel surveillance capsule which was removed from the core at 13.83 EFPY. The results of the capsule analysis are contained in Westinghouse report WCAP-16088-NP, "WaterfordUnit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation"(September 2003) which is provided as Enclosure 1 to this letter. In accordance with TS surveillance requirement (SR) 4.4.8.1.2, the revised specimen evaluation was used to develop new heatup/criticality, cooldown, and inservice hydrostatic test curves which are contained in the proposed TS Figures 3.4-2 and 3.4-

3. The proposed change is discussed in detail in Attachment 1.

10 CFR 50.60 requires that P/T limits be established for reactor pressure vessels (RPV) during normal operating and hydrostatic or leak rate testing conditions using the criteria of 10 CFR 50, f fdD (itl 4

W3Fl-2003-0075 Page 2 Appendices G and H. Appendix G of 10 CFR 50 specifies that the requirements for these limits are the ASME Section XI, Appendix G limits. The new P/T analysis credits the use of Code Case N-641, which covers Code Cases N-640 and N-588. These Code Cases were approved in Revision 13 dated June 2003 of Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability ASME Section Xl Division 1. Code Case N-641 permits the postulation of a circumferentially oriented flaw (in lieu of an axially oriented flaw) for the evaluation of the circumferential welds in RPV PIT limit curves. Code Case N-641 also permits the use of alternate reference fracture toughness data (Kx: fracture toughness curve instead of Kia fracture toughness curve) for reactor vessel materials in determining the PIT limits.

A similar P/T analysis was performed for the low temperature overpressure protection (LTOP) conditions. The proposed change to TS 3.4.8.3 will revise the enable temperature for the alignment of the shutdown cooling system suction line relief valves (SI-406A and SI-406B) that provide the LTOP function. The analysis was performed considering the measurement uncertainty recapture power uprate that commenced at the start of Cycle 12 and the extended power uprate planned to start at the beginning of Cycle 14.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal. The proposed change does not include any new commitments. The NRC. has approved similar Technical Specification changes for other plants.

Entergy requests approval of the proposed amendment by June 1, 2004. The existing P/T curves are acceptable to 16 EFPY, which is expected to be reached in August 2004. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Dana Millar at 601-368-5445.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 22, 2003.

Sincerely, X J. E. Venable Vice President, Operations Waterford Steam Electric Station, Unit 3 JEV/DM/cbh

W3FI-2003-0075 Page 3 Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Changes to Technical Specification Bases Pages - For Information Only

Enclosure:

Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operations (WCAP-1 6088-NP, Revision 1, September 2003) cc: Bruce Mallett, NRC Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop 0-7 D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 NRC Senior Resident Inspector Waterford NPS P.O. Box 822 Killona, LA 70066-0751 Louisiana DEQfSurveillance Division P.O. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers ATTN: Library Town Center Suite 300S 29' South Main Street West Hartford, CT 06107-2445

Attachment I W3FI -2003-0075 Analysis of Proposed Technical Specification Change

Attachment 1 to W3F1-2003-0075 Page 1 of 10

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3). The proposed change will establish new pressure/temperature (PIT) limits valid for 32 effective full power years (EFPY). A second reactor vessel specimen was removed and analyzed consistent with Technical Specification surveillance requirement 4.4.8.1.2 to establish limits for plant operations that are appropriate to ensure reactor vessel fracture toughness. As a result of the analyses, new heatup, cooldown, and inservice hydrostatic test curves were developed and new maximum heatup and cooldown rates were defined. In addition the low temperature overpressure protection (LTOP) conditions were reanalyzed, which resulted in a lower LTOP enable temperature. The analyses for the P/T limits and LTOP conditions were performed considering the measurement uncertainty recapture power uprate that started at the beginning of Cycle 12 and the extended power uprate that will commence at the start of Cycle 14. Westinghouse developed a revised fluence and fracture toughness analysis of the Waterford 3 reactor vessel. The summary report "Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation" (WCAP-1 6088-NP, Revision 1, September 2003) is included as Enclosure 1.

2.0 PROPOSED CHANGE

Changes to Technical Specifications (TS) 3.4.1.3, 3.4.1.4, 3.4.8.1, and 3.4.8.3 are proposed and described below.

TS 3.4.1.3, Mode 4 Reactor Coolant and/or Shutdown Cooling Loops and TS 3.4.1.4, Mode 5 Reactor Coolant and/or Shutdown Cooling Loon The ** note states in part 'A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold.leg temperatures less than or equal to 2720 F unless ... ' The reference to 2720 F will be changed to 200OF to be consistent with the new analysis associated with the LTOP enable temperature.

The associated TS Bases will also be revised to reflect the lower temperature.

TS 3.4.8.1, Pressure/Temperature Limits The maximum heatup and cooldown rates that are included in the limiting conditions for operation (LCO) items 'a' through uf` will be changed. The proposed change will include item na," which will specify a maximum heatup rate of 60 0F/hr and item ub" that will specify a maximum cooldown rate of 100°F/hr.

TS Figures 3.4-2 and 3.4-3 will be revised to reflect the new P/T limits valid to 32 EFPY. These curves do not contain instrument uncertainty. It is standard practice to include instrument uncertainty, as appropriate, in station operating procedures. The list of figures in the index will be revised to reflect the new figures.

The TS Bases for TS 3.4.8.1 will be modified to reflect the new PIT limits and analyses. Of particular note is the removal of the analytical results table (Table B 3/4 4-1). This level of fracture toughness detail is contained in the summary analysis of the capsule (Analysis of Capsule 2630 from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation to W3F11-2003-0075 Page 2 of 10 Surveillance Program, WCAP-16002-NP) that was provided to the NRC by letter dated March 28, 2003. This level of detail is excessive and is not consistent with the level of detail contained in the Bases of NUREG 1432, Revision 2, "Standard Technical Specifications Combustion Engineering Plants."

TS 3.4.8.3, Overpressure Protection Systems The applicability will be changed. Currently the TS is applicable in MODE 4 when the temperature of any RCS cold leg is less than or equal to 2720F, MODE 5, and MODE 6 when the head is on the reactor vessel and the RCS is not vented through a 5.6 square inch or larger vent. Based on the analyses, the LTOP enable temperature will be changed to 2000F. The proposed change will require LTOP to be enabled in MODE 4 when the temperature of any RCS cold leg is less than or equal to 2000F, MODE 5, and MODE 6 when the head is on the reactor vessel and the RCS is not vented through a 5.6 square inch or larger vent.

In addition, the # note will be deleted. The note was required to allow performance of the hydrostatic test to prove restoration of the structural integrity of any ASME Code Class 1 component (TS 3.4.9, Structural Integrity, Action a). With the reduction in the LTOP enable temperature, the note is no longer needed. Based on the proposed PIT curves, in order to satisfy TS 3.4.9, Action a, the structural integrity of the ASME Code Class 1 components must be restored prior to reaching 2600 F (based on the lowest service temperature of 190OF plus the TS Action requirement of 70 0F). If a hydrostatic test were required, the LTOP cannot be inservice. The lowest temperature at which a hydrostatic test can be performed is 1900 F; the LTOP can be isolated above 2000F, which allows a 600 F margin to perform the test prior to reaching the TS 3.4.9, Action ua" limit. Instrument uncertainty is not included in the enable temperature. Note: Amendment 189 (September 22, 2003) recently approved the relocation of TS 3.4.9 to the Technical Requirements Manual.

The associated TS Bases will be modified.

In summary, based on the analyses performed by Westinghouse, new P/T heatup and cooldown curves are proposed along with a maximum heatup rate of 600 F/hr, a maximum cooldown rate of 1000F/hr for any RCS temperature. In addition, changes will be made to a note associated with coolant loops that are required during Mode 4 and to the applicability of the overpressure protection system along with the deletion of a related note.

3.0 BACKGROUND

TS 3.4.1.3, Mode 4 Reactor Coolant and/or Shutdown Cooling Loops and TS 3.4.1.4, Mode 5 Reactor Coolant and/or Shutdown Cooling Loop The ** note was added with the approval of Operating License Amendment 106, dated May 8, 1995.

TS 3.4.8.1. Pressure/Temperature Limits The current fluence extrapolation was based on the 4.44 EFPY capsule work that was removed at the end of cycle 4. A summary report of the analysis of this capsule was provided to the NRC

Attachment 1 to W3FI-2003-0075 Page 3 of 10 by letter dated November 25,1992. Based on this analysis the TS P/T limit was originally approved for 15 EFPY in Operating License Amendment 106, dated May 8, 1995. The period of applicability for the current TS P/T limits is 16 EFPY which was approved in Operating License Amendment 160, dated April 24, 2000. The allowable RCS cooldown rate was relaxed with the approval of Operating License Amendment 177, dated January 8, 2002.

TS 3.4.8.3, Overpressure Protection Systems TS 3.4.8.3 defines the requirements for LTOP provided by the shutdown cooling (SDC) system relief valves and the applicable modes of plant operation. Operating License Amendment 5 revised the applicability in Mode 4 to allow a lower RCS temperature during inservice leak and hydrostatic testing without imposing the requirements of LTOP.

Operating License Amendment 72, dated April 17, 1992, in part, added the # note that states:

"260 0 F during inservice leak and hydrostatic testing with Reactor Coolant System temperature changes restricted in accordance with Specification 3.4.8.1g." The note was added to allow compliance with TS requirements to establish the integrity of all ASME Code Class 1, 2, and 3 components.

Operating License Amendment 106 revised the LTOP enable temperature to 2720F based on the results of the first capsule work that was performed at the end of cycle 4.

4.0 TECHNICAL ANALYSIS

The Waterford 3 reactor vessel is made up of six beltline plates and seven welds. These are identified as:

Intermediate Shell Plate M-1003-1, -2, -3

Lower Shell Plate M-1 004-1, -2, -3

Intermediate Shell Plate Longitudinal Weld Seams 101-124A, B & C

Lower Shell Plate Longitudinal Welds 101-142A, B & C

Intermediate to Lower Shell Plate Circumferential Weld Seam 101-171 Evaluation of the Second Waterford 3 Reactor Vessel Specimen 10 CFR 50, Appendix H, aReactor Vessel Material Surveillance Program Requirements, "defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data permit determination of the conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture throughout its service life.

Six surveillance capsules for monitoring the effects of neutron exposure on the Waterford 3 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The capsules were positioned in the reactor vessel between the core barrel and the vessel wall. Capsule 2630 orW263 was irradiated in the 263° position and

Attachment 1 to W3FI-2003-0075 Page 4 of 10 removed after 13.83 EFPY of plant operation. The capsule contained Charpy V-notch impact and tensile specimens made from reactor vessel lower shell course plate M-1 004-2, submerged arc weld metal identical to the beltline region girth weld seam and heat-affected-zone (HAZ) metal. Standard Reference Material from HSST-01MY Plate was included within capsule 2630 in addition to the reactor vessel materials. The number of specimens of each material contained in capsule 2630, the chemical compositions of the surveillance materials, and the location of the individual specimens within the capsule is contained in Westinghouse report WCAP-1 6002-NP, "Analysis of Capsule 263 0from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program" (March 2003) which was provided to the NRC by Entergy letter dated March 28, 2003.

10 CFR 50, Appendix G, 'Fracture Toughness Requirements," also requires a minimum initial Charpy upper shelf energy of no less than 75 foot pounds (ft-lbs) for the reactor vessel beltline materials unless it is demonstrated that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those required by ASME Section Xl, Appendix G. No action is required for a material that does not meet the initial 75 ft-lbs requirement provided that the irradiation embrittlenment does not cause the upper shelf energy to drop below 50 ft-lbs. The results of the capsule analysis for upper shelf energy to 32 EFPY confirm that the 50 ft-lbs minimum requirement was met.

Fluence Evaluation Process Fast neutron exposure parameters in terms of fast neutron fluence (E>1.OMeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule W263, withdrawn at the end of the eleventh fuel cycle, was performed. In addition, to provide an up-to-date data base applicable to the Waterford 3 reactor, the sensor set from the previously withdrawn capsule (W97) was re-analyzed using the current dosimetry evaluation methodology. The methodology used for the calculations and dosimetry evaluations follow the guidance and meet the requirements of Regulatory Guide 1.190, 'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

Changes to the Pressure/Temperature TSs for 32 EFPY 10 CFR 50, Appendix G, specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations. The fracture toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

RTNDT (defined as the greater of either the drop weight nil-ductility transition temperature or the temperature 600 F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate) of a given material is used to index that material to a reference stress intensity factor curve (KIa or K1. curve, as applicable). The Kiacurve appears in Appendix G of ASME Code Section Xl. In February 1999, ASME Code Case N-640, now N-641, was approved to permit the use of the K1, curve as given in Appendix A of ASME Code Section Xl in

- to W3F1-2003-0075 Page 5 of 10 lieu of the Kia curve. When a given material is indexed to the K1c curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Plant operating limits can then be determined using these allowable stress intensity factors.

A method for guarding against non-ductile fracture in reactor vessels is described in Appendix G to Section III of the ASME Code, 'Nuclear Power Plant Components" and Appendix G to Section Xl,

'Rules forlnservice Inspection". The application of Appendix G to the ASME Code is established in the requirements of 10 CFR 50, Appendix G. This method uses fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition (NDT) temperature (in accordance with ASTM E 208 "Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Feritic Steels") or the temperature that is 600 F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion as determined from the Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Ka or Kc1 curve), which appears in Appendix G of Section Xl of ASME Code. The Kla curve is a lower bound of dynamic, crack arrest results and static fracture toughness data obtained from several heats of pressure vessel steel.

The Kc1 curve is based on the lower bound of static critical Kevalues measures as a function of temperature on specimens of SA-533 Grade B Class 1,SA-508-A, SA-508-2, and SA-508-2 steel.

When a given material is indexed to the Kl or KI,curve, allowable stress intensity factors can be obtained for the material as a function of temperature. The operating limits can then be determined using these allowable stress intensity factors.

The RTNDT and, in tum, the operating limits, are adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel materials. The irradiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which surveillance capsules containing prepared specimens of the reactor vessel materials are periodically removed from the operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. The adjusted RTNDT is used to index the material to the KIR curve which, in turn, is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

Regulatory Guide 1.99, Revision 2, describes general procedures acceptable for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor. To date there have been two surveillance capsules removed from the Waterford 3 reactor vessel. To use these surveillance data sets, they must be shown to be credible. The Waterford 3 surveillance data satisfies the Regulatory Guide 1.99, Revision 2, credibility requirements. The surveillance data for W263 was analyzed using Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1.

Position 1.1 requires an evaluation of the adjusted reference temperature (ART) for each material in the beltline. The ART for the reactor vessel beltline region materials is calculated in accordance with Regulatory Guide 1.99, Revision 2. The ART is calculated by adding the initial

Attachment 1 to W3FI-2003-0075 Page 6 of 10 RTNDT, the predicted radiation-induced shift in RTNDT (ARTNoT), and a margin term to obtain conservative, upper-bound values of ART. The predicted radiation induced ARTNDT is calculated using the respective reactor vessel beltline materials copper and nickel contents (chemistry factor) and the neutron fluence applicable to 32 EFPY including an estimated increase in flux due to proposed power uprates. The AT and %T wall locations for each beltline material are determined by adding the minimum thickness of the cladding to the distance into the base metal at the 1/4T and 3/4T locations. The YAT and %T ART results for the Waterford 3 reactor vessel beltline region materials applicable to 32 EFPY are presented in the enclosed Westinghouse Report, WCAP-16088-NP, Tables 5-3 and 5-4, respectively. Based on these results, the controlling beltline material for the Waterford 3 reactor vessel is the lower shell plate M-1 004-2. The applicability of 32 EFPY is also consistent with the removal schedule for the next capsule at *32 EFPY (25.4 to 50.8 EFPY) as shown in Waterford 3 FSAR Table 5.3-10.

Position 2.1 requires that if there is evidence that the copper/nickel content of the surveillance specimen differs from that of the vessel, the measured values of ARTNDT should be adjusted by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance specimen. The surveillance data would be fitted to obtain the relationship of ARTNDT to fluence by calculating the chemistry factor for the best fit by ratioing each adjusted ARTNDT by its corresponding fluence factor. This position was used for the lower shell plate M-1 004-2 material and is reported in Tables 5-3 and 5-4 of the enclosed Westinghouse report.

The P/T curves provided in this proposed amendment are adjusted for sensor location but do not include instrument uncertainty. Protection against non-ductile failure is ensured by using these curves to limit the reactor coolant pressure. The P/T limits for normal heatup (including criticality core limits) at 32 EFPY are provided on the revised TS Figure 3.4-2. The criticality limit temperature is 2300 F above 554.1 psia and is 110F below 554.1 psia. The composite cooldown PIT limits are shown in revised TS Figure 3.4-3.

Based on the calculations performed, the vessel may be heated up or cooled down at the maximum analyzed rate, 1OO 0 F/hr, at temperatures above the bolt-up temperature and the minimum allowable temperature of the flange region (both subject to the pressure limit), and the lowest service temperature. The 100 0 F/hr rate is also applicable to the in-service hydrostatic pressure and leak test. The RCS heatup rate during normal operations and in-service hydrostatic pressure and leak testing will however be limited to 60°F1hr. This limit is due to stress limitations in the reactor coolant pump (RCP). As part of the loss of coolant accident support scheme, the RCP has a ring around the suction nozzle of the pump to which the support skirt is welded. Due to this design, the heatup rate must be limited to maintain acceptable thermal stresses.

The P/T curves were developed using Code Case N-641, "Altemative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements" (January 17, 2000). The Code Case was approved in Revision 13 dated June 2003 of Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability ASME Section Xl Division 1.

Changes to Suction Line Relief Valves TSs for 32 EFPY ASME Code Case N-641 presents alternative procedures for calculating pressure temperature relationships and low temperature overpressure protection (LTOP) system effective temperatures and allowable pressures. These procedures take into account alternative fracture to W3F11-2003-0075 Page 7 of 10 toughness properties, circumferential and axial reference flaws, and plant-specific LTOP effective temperature calculations. The Code requires that the LTOP system be effective at coolant temperatures less then 200OF or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature calculated below:

(1) Te = RTNDT + 40 + max (ATmetat), OF (2) Te = RTNDT + 50 In [((F*Mm(p R / t)) -33.2) / 20.734], OF where, Mm = 0.926(t) (112), for an inside surface flaw F = 1.1, accumulation factor for safety relief valves p = 2.500, vessel design pressure, ksi Ri = 87.125, vessel inner radius, in.

t = 8.625, vessel wall thickness, in.

RTNDT is the highest ART for the limiting beltline material at a distance one fourth of the vessel section thickness from the vessel inside surface, as determined by Regulatory Guide 1.99, Revision 2. The lower value of Te (equation I or 2 above) is compared to the 2000 F criterion, and the higher from that determines the LTOP system effective temperature.

The highest calculated Y4T ART for Waterford 3 reactor vessel beltline regions at 32 EFPY is 50 0F. The maximum ATmetai is 18.480F. The enable temperature associated with equation 1 is 108.418 0F, while the enable temperature associated with equation 2 is 85.704 0F. Based on the guidance for determining the LTOP system effective temperature contained in Regulatory Guide 1.99 Revision 2, the minimum required enable temperature for Waterford 3 is 2000 F for 32 EFPY. The proposed change revises the LCO applicability to reflect the new enable temperature.

The LTOP enable temperature was also included in the # note associated with TS 3.4.8.3 and the ** note associated with TSs 3.4.1.3 and 3.4.1.4. The changes to these notes and the associated basis were described earlier in this request.

Evaluation of Pressurized Thermal Shock A pressurized thermal shock (PTS) evaluation for the Waterford 3 reactor vessel beltline materials was performed in accordance with the screening criteria contained in 10 CFR 50.61,

'Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."

The results are shown in Section 7 of the enclosed Westinghouse report. The results demonstrate that the Waterford 3 reactor vessel beltline materials will not exceed the PTS screening criteria before 32 EFPY. The controlling beltline material for the Waterford 3 reactor vessel with respect to PTS is the lower shell plate M-1004-2, with a RTPTS of 53°F that is well below the PTS screening criterion of 270°F.

to W3Fl-2003-0075 Page 8 of 10

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

The P/T curves were developed using Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements" (January 17, 2000). Code Case N-641, similar to Code Case N-588 permits the postulation of a circumferentially oriented flaw (in lieu of an axially oriented flaw) for the evaluation of the circumferential welds in reactor pressure vessel (RPV) pressure temperature (P/T) limit curves.

Also, Code Case N-641 similar to Code Case N-640 permits the use of an alternate reference fracture toughness (K1, fracture toughness curve instead of Kia fracture toughness curve) for reactor vessel materials in determining the P/T limits. Since the pressure stresses on a circumferentially oriented flaw are lower than the pressure stresses on an axially oriented flaw by a factor of 2, postulating a circumferentially oriented flaw for the evaluation of the circumferential welds (as permitted by Code Case N-641) in establishing the PIT limits would be more realistic than the methodology currently endorsed by 10 CFR 50, Appendix G. These Code Cases were approved in Revision 13 dated June 2003 of Regulatory Guide 1.147, Inservice inspection Code Case Acceptability ASME Section Xl Division 1.

5.2 No Significant Hazards Consideration Entergy Operations, Inc. (Entergy) is proposing that the Waterford Steam Electric Station, Unit 3 (Waterford 3) Operating License be amended to modify the pressuretemperature (PIT) limits from 16 to 32 effective full power years (EFPY) as reflected in Waterford 3 Technical Specification 3.4.8.1. This includes changing the heatup/criticality, cooldown, and inservice hydrostatic pressure test operational curves as well as the maximum rate for heatup and cooldown during both normal and inservice hydrostatic testing. A vessel specimen was retrieved during the spring 2002 Waterford 3 refueling outage and tested in accordance with acceptable industry standards for determining the vessel nil ductility of the irradiated vessel plates. Analyses were performed in accordance with Regulatory Guide 1.99, Revision 2 using recent code cases for performing fracture toughness.

In addition, the enable temperature designating the low temperature overpressure (LTOP) operating region was evaluated. Based on the analyses performed using recent code cases, the applicability forTS 3.4.8.3 will be modified. Associated notes in TS 3.4.8.3, TS 3.4.1.3, and TS 3.4.1.4, which were based on the LTOP enable temperature, will also be changed.

The proposed change considered the increase in neutron fluence related to power uprates. A measurement uncertainty recapture power uprate commenced at the start of cycle 12 and an extended power uprate is planned to start at the beginning of cycle 14.

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

to W3F1-2003-0075 Page 9 of 10

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

Response: No.

The probability of occurrence of an accident previously evaluated for Waterford 3 is not altered by the proposed amendment to the TSs. The accidents currently analyzed in the Waterford 3 Final Safety Analysis Report (FSAR) remain the same considering the results of the proposed changes to the P/T limits and the LTOP enable temperature.

The new P/T and LTOP enable temperature limits were based on NRC accepted methodologies along with ASME Code alternatives. The proposed changes do not impact the integrity of the reactor coolant pressure boundary (RCPB) (i.e., there is no change to the operating pressure, materials, loadings, etc.). The proposed change does not affect the probability nor consequences of any design basis accident (DBA). The proposed P/T limit curves, maximum heatup and cooldown rates, and LTOP enable temperature are not considered to be an initiator or contributor to any accident currently evaluated in the Waterford 3 FSAR. The new limits ensure the long term integrity of the RCPB.

Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data permit determination of the conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture throughout its service life. During the spring 2002 Waterford 3 refueling outage a reactor vessel specimen capsule was withdrawn and analyzed to predict the fracture toughness requirements using projected neutron fluence calculations. 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.

The predicted radiation induced ARTNDT was calculated using the respective reactor vessel beltline materials copper and nickel contents and the neutron fluence applicable to 32 EFPY including an estimated increase in flux due to proposed power uprates. The RTNDT and, in turn, the operating limits for Waterford 3 were adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel materials.

Therefore, new operating limits will be established which are represented in the revised operating curves for heatup/criticality, cooldown, and inservice hydrostatic testing contained in the TSs.

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

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

Response: No.

The proposed changes to the P/T and LTOP enable temperature will not create a new accident scenario. The requirements to have P/T limits and LTOP protection are part of the licensing basis for Waterford 3. The approach used to develop the new P/T limits to W3F1-2003-0075 Page 10 of 10 and LTOP enable temperature meets NRC and ASME regulations and guidelines. The data analysis for the vessel specimen removed during the last Waterford 3 refueling outage confirms that the vessel materials are responding as predicted.

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

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

Response: No.

The existing P/T curves and LTOP enable temperature in the TSs are reaching their expiration period for the number of years at effective full power operation. The revision of the PIT limits and curves will ensure that Waterford 3 continues to operate within the operating margins allowed by 10 CFR 50.60 and the ASME Code. The material properties used in the analysis are based on results established through Westinghouse material reports for copper and nickel content. The application of ASME Code Case N-641 presents alternative procedures for calculating P/T and LTOP temperatures in lieu of that established for ASME Section Xl, Appendix G-2215. This Code alternative allows certain assumptions to be conservatively reduced. However, the procedures allowed by Code Case N-641 still provide significant conservatism and ensure an adequate margin of safety in the development of P/T operating and pressure test limits to prevent non-ductile fractures.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 1.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 W3FI -2003-0075 Proposed Technical Specification Changes (mark-up)

LIST OF FIGURES FIGURE PAGE 3.1-1 REQUIRED STORED BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION ............................................ 3/4 1-13 3.4-1 DELETED ............ , ....... i .. 314 4-27 rf fFlo 9 j D liV--rr3 t4SEf I . C -4 2, E;P 3.4-2 EACTO LANT SYSTEM PRESSURE - TEMPERATURE

......... I . .... 3144-30 3.4-3 EACTOR COOESSSRE -TEMPERATURE LIM ITS ........................................ 3/4 4-31 3.6-1 DELETED. 3/4 6-12 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST 314 7-26 5.1-1 EXCLUSION AREA .5-2 5.1-2 LOW POPULATION ZONE .5-3 5.1-3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS. 54 5.6-1 ALTERNATIVE CHECKERBOARD ARRANGEMENTS 5-6a 5.6-2 ACCEPTABLE BURNUP DOMAIN FOR UNRESTRICTED STORAGE OF SPENT FUEL INREGION 2.5-6b 5.6-3 ACCEPTABLE BURNUP DOMAIN FOR SPENT FUEL IN CHECKERBOARD ARRANGEMENT WITH FUEL OF 5% ENRICHMENT (OR LESS) AT 27 MWD/KgU .5-6c 6.2-1 DELETED ........... 6-3..........................

63 6.2-2 DELETED. 64 WATERFORD - UNIT.3 NXX AMENDMENT NO. 13,27, 102,181,488-

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loop(s)/train(s) listed below shall be OPERABLE and at least one reactor coolant and/or shutdown cooling loops shall be in operation.*

a. Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump,**

b. Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump,**

c. Shutdown Cooling Train A,

d. Shutdown Cooling Train B.

APPLICABILITY: MODE 4 ACTION:

a. With less than the above required reactor coolant andlor shutdown cooling loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With no reactor coolant or shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meeting SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2 and immediately initiate corrective action to return the required coolant loop to operation.

All reactor coolant pumps and shutdown cooling pumps (LPSI pumps) may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2, and (2)core outlet temperature is maintained at least 100F below saturation temperature.

- A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to -72200°F unless (1) the pressurizer water volume l is less than 900 cubic feet or (2) the secondary water temperature of each steam generator is less than 100IF above each of the Reactor Coolant System cold leg temperatures.

WATERFORD - UNIT 3 314 4-3 AMENDMENT NO. 406, 485,

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4 At least two of the loop(s)/trains listed below shall be OPERABLE and at least one reactor coolant and/or shutdown cooling loop shall be in operation.*

a. Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump*,

b. Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump**,

c. Shutdown Cooling Train A,

d. Shutdown Cooling Train B.

APPLICABILITY: MODE 5 with reactor coolant loops filled*".

ACTION:

a. With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE status or to restore the required level as soon as possible.

b. With no reactor coolant or shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2 and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.4.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be 250% of wide range indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.3 At least one reactor coolant loop or shutdown cooling train shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps and shutdown cooling pumps (LPSI pumps) may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2, and (2) core outlet temperature is maintained at least 100F below saturation temperature.

A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 2-7 2000F unless (1)the pressurizer water volume is less than 900 cubic feet or (2)the secondary water temperature of each steam generator is less than 1000F above each of the Reactor Coolant System cold leg temperatures.

WATERFORD -UNIT 3 3/4 4-5 AMENDMENT NO. 416,485,

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSUREITEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup rate of 34600 F per hour with Roactor Coolant Syctom cold log tempeFaturo less than 2000F.

b. A ma ximu heatup rate of 500 F per hour with Reactor Coolant System cold leg tempereatU eFthan 2002F and -lesthan or equal to 315 0F. I

c. A maximum heatup rate of 600 F per hour with Reactor Coolant System cold log teroneraturo geater than 3254F.

A maximum cooldown rate of 3010 0 °F per hour with Reactor Coolant SysteM Ib.Geld leg temperature less than 1100 F.

e. A maximum cOIGIW rwiate of 4000F-prhouF with Reaactor Coolant System nold leg temperature greaterthan or equal to 110 F.

A maximum; tmerat a e of less than or equal to 100 F in an) 1--hour period during insepGice hydrostatic and leak testing operations above the heatup and cooldown limit curvos.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tvg and pressure to less than 2000 F and 500 psia, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WATERFORD - UNIT 3 3/4 4-28 AMENDMENT NO. 47-7-,

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_ . _, ._ . -. - . _ U -a SE M6XT PAC-if FVO(Žt WATERFORD-UNIT 3 3/4 4 -30 AMENDMENT NO. 160

FIGURE 3.4-2 WATERFORD UNIT 3 HEATUP CURVE - 32 EFPY REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS (Curves do not include margins for instrument uncertainties)

WATERFORD - UNIT 3 3/4 4-30 AMENDMENT NO. 406, 460,

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-=~!SZ. PEAKY SHRFASE RLUFrNCE -2.29g itO81 tiA., .2 20 EFnY WATERFORD - UNIT 3 3/4 4-31 AMENDMENT NO. 40&T460,177 SE-F' Nek- PACiE F-GP-- WtSfP- I

U) 1750 l -

Unacceptable Acceptable 1500 Operation - Operation 1000 __ l _ Cooldown Li it and Inservice Hydrostatic w 750 5 Pressure and Leak _l a} Test, 100 deg. F/hr 500 250 l - . _ Boltup Temp. l 2060 deg. F _ _ _ _ _ _ _

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

FIGURE 3.4-3 WATERFORD UNIT 3 COOLDOWN CURVE- 32 EFPY REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS (Curves do not include margins for instrument uncertainties)

WATERFORD - UNIT 3 314 4-31 AMENDMENT NO. 44,41,77,

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.8.3 Two Shutdown Cooling (SDC) System suction line relief valves (Sl-406A and SI-406B) shall be OPERABLE with a lift setting of less than or equal to 430 psia.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 272200FX MODE 5, and MODE 6 when the head is on the reactor vessel and the RCS is not vented through a 5.6 square inch or larger vent.

ACTION:

a. With one SDC System suction line relief valve inoperable in MODE 4, restore the inoperable valve to OPERABLE status within 7 days, or depressurize and vent the RCS through at least a 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one SDC System suction line relief valve inoperable in MODES 5, or 6, either (1) restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) complete depressurization and venting of the RCS through at least a 5.6 square inch vent within a total of 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .

c. With both SDC System suction line relief valves inoperable, complete depressurization and venting of the RCS through at least a 5.6 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. In the event either the SDC System suction line relief valve(s) or the RCS vent(s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the SDC System suction line relief valve(s) or RCS vent(s) on the transient, and any corrective action necessary to prevent recurrence.

e. The provisions of Specification 3.0.4 are not applicable.

260 0F during inscrsice loak and hydrostatiG testing with Roactor Coolant System tomporaturo Ghanges4est4ctod in accordance with Spocification 3.4.8.1g.

WATERFORD -UNIT 3 3/4 4-34 AMENDMENT NO. 5,7-2.4-6,

Attachment 3 W3FI -2003-0075 Changes to Technical Specification Bases Pages For Information Only

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops or trains (either shutdown cooling or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling trains be OPERABLE.

(DRN 03-375, Ch. 19)

The operation of one reactor coolant pump or one shutdown cooling (low pressure safety injection) pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control. If no coolant loops are in operation during shutdown operations, suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1 or 3.1.1.2, as applicable, is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that which would be required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

((DRN 03-375, Ch. 19)

The restrictions on starting a reactor coolant pump in MODES 4 and 5, with one or more RCS cold legs less than or equal to 272200 0 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against ovepressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100 0F above each of the RCS cold leg temperatures.

WATERFORD - UNIT 3 B 3/4 4-1 Rev.'ised by NRC Lotter dated March 17, 1999 CHANGE NO. 19

REACTOR COOLANT SYSTEM As used in this specification, the term 'cold leg temperature' is intended to be representative of that entering the reactor vessel beltline. During periods with the reactor coolant pumps in operation, the TCOLD temperature indication meets this intent. However, during periods when the reactor coolant pumps are not in service, the TcOLD temperature indicator is in a stagnant segment of piping and the indication may not necessarily be indicative of that entering the reactor vessel beltline. During the condition when the reactor coolant pumps are operating, the lowest TCOLD of a loop with an operating reactor coolant pump is used to monitor the P-T limits. However, during periods when the shutdown cooling system is in operation and following coastdown of the last RCP, the shutdown cooling temperature is the 'cold leg temperature' used to monitor P-T limits.

The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 600 F per hour or cooldown rate of up to 1000 F per hour.

The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figures 3.4-2 and 3.4-3. The limitations on the Reactor Coolant System heatup aRd cooldGn rates a-e Isfuwthe restricted due to stress limitations in the Reactor Coolant Pump. As part of the LOCA support scheme, the Reactor Coolant Pump has a ring around the suction nozzle of the pump.

The support skirt is welded to the ring. Due to this design, the heatup and cooldown rates must be limited to maintain acceptable thermal stresses.

The reactor vessel materials have been tested to determine their initial RTNDTA-494e-.suUs of these test arc shown in Table B 311.1 4. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RTt 4DT. Therefore, an adjusted reference temperature, based upon the fluence, copper and nickel content of the material in question, can be predicted using FSAR Table 5.3-1 and the recommendations of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNDT at the end of the applicable service period, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82 and 10 CFR Part 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in FSAR Table 5.3-10. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the delta RTNOT determined from the surveillance capsule is different from the calculated delta RTNDT for the equivalent capsule radiation exposure.

WATERFORD - UNIT 3 B3/4 4-7 CHANGE NO. 1-AMENDMENT NO. 406

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REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The maximum RTNDT for all Reactor Coolant System pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 90 0 F. The Lowest Service Temperature limit line shown on Figures 3.4-2 and 3.4-3 is based upon this RTNOT since Article NB-2332 of Section IlIl of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 1001F for piping, pumps, and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia (as corrected for elevation-ans-iuxeRto-eror.

Instrument uncertainty is not included in the Figures 3.4-2 and 3.4-3.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of the shutdown cooling system relief valve or an RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 27-200 0F. Each shutdown cooling system relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100OF above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with injection into a water-solid RCS. The limiting transient includes simultaneous, inadvertent operation of three HPSI pumps, three charging pumps, and all pressurizer backup heaters in operation. Since SIAS starts only two HPSI pumps, a 20% margin is realized.

The restrictions on starting a reactor coolant pump in MODE 4 and with the reactor coolant loops filled in MODE 5, with one or more RCS cold legs less than or equal to 2722000 F, are provided in Specification 3.4.1.3 and 3.4.1.4 to prevent RCS pressure transients caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100OF above each of the RCS cold leg temperatures.

Maintaining the steam generator less than 100OF above each of the Reactor Coolant System cold leg temperatures (even with the RCS filled solid) or maintaining a large surge volume in the pressurizer ensures that this transient is less severe than the limiting transient considered above.

WATERFORD - UNIT 3 B 3/4 4-1 0 Amendment No. 7, 406,

Enclosure-I WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16088-NP, Revision I Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation Stephen T. Byrne Reactor Coolant System Design and Analysis T. Laubham Reactor Component Design and Analysis September 2003 Reviewer. 1A 9~

C.L. Hoffnann Reactor Coolant System Design and Analysis Approved:___ _____ __

Buce M. Hint n, Manager Reactor Coolant System Design and Analysis Record of Revision No. Date Pages Involved Original Issue July 2003 all I September 2003 Pages iii, 2-1 Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii TABLE OF CONTENTS LIST OF TABLES ............. v LIST OF FIGURES ............. vii I INTRODUCTION .1-1 2 FRACTURE TOUGHNESS PROPERTIES .2-3 RADIATION ANALYSIS AND NEUTRON DOSIMETRY .3-1 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS .4-1 4.1 OVERALL APPROACH .4-1 4.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPM ENT........................................................................................................... 4-1 4.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS .4-4 4.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS .4-5 4.5 BOLTUP TEMPERATURE REQUIREMENTS .4-5 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE5-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .. 6-1 7 EVALUATION OF SCREENING CRITERIA .. 7-1 8 REFERENCES .. 8-September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vV WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF TABLES Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Waterford Unit 3 Reactor Vessel Materials ........................................................... 2-2 Table 2-2 Summary of the Waterford Unit 3 Reactor Vessel Beltline Material Chemistry Factors.................................................................................................................................. 2-3 Table 3-1 Calculated Neutron Exposure of the Middle Shell Plates (M-1003-1, M-1003-2, and M-1003-3) ..... 3-2 Table 3-2 Calculated Neutron Exposure of the Lower Shell Plates (M-1004-l, M-1004-2, and M-1004-3) .3-2 Table 3-3 Calculated Neutron Exposure of the Middle Shell Longitudinal Welds .3-3 Table 3-4 Calculated Neutron Exposure of the Lower Shell Longitudinal Welds .3-3 Table 3-5 Calculated Neutron Exposure of the Middle Shell to Lower Shell Circumferential Weld (I01-171) .3 Table 5-1 Summary of the Vessel Surface, I/4T and 3/4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curve .5-3 Table 5-2 Summary of the Peak Calculated Fluence Factors used for the Generation of the 32 EFPY Heatup/Cooldown Curve .5-3 Table 5-3 Calculation of the ART Values for the I/4T Location at 32 EFPY. 54 Table 5-4 Calculation of the ART Values for the 3/4T Location at 32 EFPY .5-5 Table 5-5 Summary of the Limiting ART Values Used in the Generation of the WMaterford Unit 3 Heatup/Cooldown Curves .5-6 Table 6-1 32 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 . 6-5 Table 6-2 32 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 .6-6 Table 7-1 Calculation of the Waterford Unit 3 RTPTS Values for 32 EFPY .7-3 Table 7-2 Calculation of the Waterford Unit 3 RTpTs Values for 48 EPY. 7-4 Table 7-3 Calculation of the Waterford Unit 3 Upper Shelf Energy Values at Vessel /4T .7-5 September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 , vii vii WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES Figure 6-1 Waterford Unit 3 Heatup Curves Applicable to 32 EFPY (without uncertainty for instrumentation errors)......................................................................................................... 6-3 Figure 6-2 Waterford Unit 3 Cooldown Curves Applicable to 32 EFPY (without uncertainty for instrumentation errors)......................................................................................................... 64 September 2003

%%'CAP-62088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix ix WESTINGHOUSE NON-PROPRIETARY CLASS 3 EXECUTIVE

SUMMARY

This report describes the methodology and results for the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Waterford Unit 3 reactor vessel. These limits can be found in Figures 6-1 through 6-2. The PT limit curves were generated based on the latest available reactor vessel information and updated calculated fluences, including consideration of power uprate to 3441 MWt at the start of Cycle 12 and to 3716 MWt at the start of Cycle 14. The new Waterford Unit 3 heatup and cooldown pressure-temperature limit curves were generated in accordance with 10CFR50 Appendix G based on the 1995 ASME Code, Section Xl, through the 1996 Addenda. The PT curves were developed using ASME Code Case N-641, which allows the use of the K1 , methodology. The PT limit curves include consideration for the reactor vessel flange region per IOCFR50 Appendix G and the lowest service temperature per Section III, Article NB-2332 to the ASME B&PV Code. The adjusted reference temperature (ART) values were obtained using the methods of Regulatory Guide 1.99, Revision 2, for each beltline material. The material with the highest predicted adjusted reference temperature (ART) was the lower shell plate M-1004-2. The PT limit curves were generated for 32 EFPY considering heatup rates of 30, 50, and 60 'F/hr and cooldown rates of 0, 10, 30, and 100WF/hr. The PT limits were actually based on the minimum pressurization temperature and the lowest service temperature which bounded the curves based on the KI, methodology. The vessel may be heated or cooled at the maximum analyzed rate, 100WF /hr, at temperatures within the allowable regions shown in Figures 6-1 and 6-2.

In addition to the PT limits, the Waterford Unit 3 reactor vessel beltline materials were evaluated relative to the Pressurized Thermnal Shock (PTS) screening criteria from IOCFR50.61 and the upper shelf energy screening criteria from IOCFR50 Appendix G. All of the beltline materials meet the screening criteria with a substantial margin for 32 EFPY as well as for 48 EFPY. Values of RTPTs are given for each material in Tables 7-1 and 7-2. The highest values are 530 F and 550 F for lower shell plate M-1004-2 after 32 and 48 EFPY, respectively. The predicted upper shelf energy are given in Table 7-3. The lowest predicted upper shelf energy was 70 fl-lb after 48 EFPY demonstrating that the beltline materials will far exceed the 50 ft-lb screening criterion of 10CFR50, Appendix G These RTPTs and upper shelf energy projections demonstrate the low radiation sensitivity of the Waterford Unit 3 beltline materials.

September 2003 WCAr- 16088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 l -l 1-I WESTINGHOUSE NON-PROPRIETARY CLASS 3 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted reference temperature (ART) corresponding to the limiting reactor vessel beltline region material. The adjusted reference temperature of the limiting material in the beltline region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties (RTNDT), estimating the radiation-induced shift (ARTNDT), and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting (highest) RTNDT for the reactor vessel at any given time, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."111 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 2 f?],

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

The K1 , critical stress intensities arc used in place of the Kia critical stress intensities. This methodology is taken from approved ASME Code Case N-64l1 31 (which covers Code Cases N-640 and N-588). 3) The 1996 Version of Appendix G to Section X114 1was used rather than the 1989 version.

The purpose of this report is to present the calculations and the development of the Waterford Unit 3 heatup and cooldown curves for 32 EFPY. This report documents the calculated ART values and the development of the PT limit curves for normal operation per the requirements of 10 CFR Part 50, Appendix G15 1. The PT curves herein were generated without allowance for instrumentation errors.

In addition to PT-limit curves, Pressurized Thermal Shock (PTS) values are calculated for each beltline material in accordance with 10CFR50.6116 3. The upper shelf screening criteria of 10 CFR Part 50, Appendix G15] is evaluated using the methods of Regulatory Guide 1.99, Revision 21'1, for all beltline materials.

Introduction September 2003 WCAP-16088.doc-092903 Revision I

WNESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan I7k.The properties of the Waterford Unit 3 reactor vessel beltline material are presented in Table 2-1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 2-1. The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2'1", Positions 1.1 and 2.1. Position 1.1 uses Tables I and 2 from Reference I along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. Surveillance capsule data are available from two capsules (Capsules W97 and W263) already removed from the Waterford Unit 3 reactor vessel. The calculation of the CF values per Position 2.1 of Reference 1 is detailed in Reference 8.

These CF values are summarized in Table 2-2. (All capsule fluence values were determined using ENDF/B-VI cross-sections and followed the guidance in Regulatory Guide 1.19019O.)

Fracture Toughness Properties September 2003 WCAP-1 6088.doc-092903 Revision I

2-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Valucs for the Wlaterford Unit 3 Reactor Vessel Materials Material Description Cu (%) Ni(%) Initial RTNDT(

Closure Head Flange M-603-1 --- 20 0 F Vessel Flange M-1001-I --- 200 F Intermediate Shell Plate M-1003-1 0.02 0.71 -30 0 F Intermediate Shell Plate M-1003-2 0.02 0.67 -500 F Intermediate Shell Plate M-1003-3 0.02 0.70 42 0 F Lower Shell Plate M-1004-1 0.03 0.62 -15 0 F Lower Shell Plate M-1004-2(d) 0.03 0.58 220 F Lower Shell Plate M-1004-3 0.03 0.62 -10 0 F Intermediate Shell Longitudinal Welds, 0.02 0.96 -60 0 F 101-124 A, B & C(b)

Lower Shell Longitudinal Welds, 101-142A, B & C(b) 0.03 <0.2 -80 0 F Circumferential Weld 1 0 1 -17 1 (b,c) 0.05 0.16 -70 0 F Notes:

(a) The initial RTNDT values for the beitline plates and welds (and for the flange forgings) are based on measured data.

(b) The intermediate shell longitudinal weld seams 101-124A, B and C were fabricated with weld wire heats BOLA and HODA (E8018C3 electrodes). The intermediate to lower shell circumferential weld seam 101-171 was fabricated with weld wire heat number 88114, Flux Type 0091 Lot Number 0145. The lower shell longitudinal weld seams 101-142A, B and C were fabricated with weld wire heat number 83653, Flux Type 0091 Lot Number 3536.

(c) The Waterford Unit 3 surveillance weld metal was made with the same weld heat as the circumferential weld seam.

(d) The Waterford Unit 3 surveillance plate was from lowver shell plate M-1004-2.

Fracture Toughness Properties September 2003 WCAP-M 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 Table 2-2 Summary of the Waterford Unit 3 Reactor Vessel Beltline Material Chemistry Factors Material Chemistry Factor Position 1.1 Position 2.1 Intermediate Shell Plate M-1003-1 20 0 F Intermediate Shell Plate M-1003-2 200 F Intermediate Shell Plate M-1003-3 20 0 F Lower Shell Plate M-1004-1 20 0 F Lower Shell Plate M-1004-2 20`F 12.4 0F Lower Shell Plate M-1004-3 20 0 F Inter. Shell Plate Long. Weld Seams 270 F 101-124A, B, C Lower Shell Plate Long. WN'eld Seams 350 F 101-142A, B, C Intermediate to Lower Shell Plate 44.4 0 F 16.2 0 F Circumferential Weld Seam 101-171 Fracture Toughness Properties September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 RADIATION ANALYSIS AND NEUTRON DOSIMETRY The radiation analysis and dosimetry performed for the Waterford Unit 3 reactor is described in WCAP-16002181. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 McV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. In addition, neutron dosimetry sensor sets from the first two surveillance capsules withdrawn from the Waterford Unit 3 reactor were analyzed using current dosimetry evaluation methodology. The results of these dosimetry re-evaluations provided a validation of the plant specific neutron transport calculations. The validated calculations were then used to project future fluence accumulation through operating periods extending to 32 and 48 effective full power years (EFPY).

All of the calculations and dosimetry evaluations described in WCAP-16002183 were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." 9J The core power distributions used in the plant specific transport analysis for the Waterford Unit 3 reactor were taken from the appropriate fuel cycle design reports for Cycles I through 11. Fluence projections beyond the end of Cycle II were based on a 1.5% uprate (to 3441 MWt) at the start of Cycle 12 and a 8%

uprate (to 3716 MWt) at the start of Cycle 14. Projections for Cycle 14 and beyond were based on the assumption that future operation would continue to make use of low leakage fuel management and that a representative equilibrium spatial power distribution from the 8% uprate would be typical of future operating cycles. [Note: In the original analysis, the calculated fluence t81 assumed a 107% RCS flow rate. In a subsequent assessment 31 it was determined that the fluence values based on the 107% flow assumption are bounding for the 110% RCS flow case.]

The maximum calculated fast neutron fluence (E > 1.0 MeV) values for the Waterford Unit 3 pressure vessel are provided in Table 3-1. These data represent the maximum exposure at the pressure vessel clad/base metal interface at azimuthal angles of 0, 15, 30, and 45 degrees relative to the core cardinal axes. The data tabulation includes the plant specific calculated fluence at the end of cycle 11 (the last cycle completed at the Waterford Unit 3 plant) and projections for future operation to 32 and 48 EFPY.

Similar data applicable to the lower shell plates, the axial welds, and the intermediate shell to lower shell circumferential weld are given in Tables 3-2 through 3-4.

Radiation Analysis and Neutron Dosimetry September 2003 WCAP- I 6088.doc-092903 Revision I

3-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 3-1 Calculated Neutron Exposure of the Middle Shell Plates (MNI-1003-1, M-1003-2, and N1-1003-3)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [ncm2 Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 I 3.28E+07 3.28E+07 1.04 1.48E+18 9.25E+17 8.07E+17 6.37E+17 2 3.18E+07 6.46E+07 2.05 2.68E+18 1.64E+18 1.47E+18 1.12E+18 3 3.64E+07 1.01E+08 3.20 4.06E+18 2.43E+18 2.12E+18 1.61E+ 18 4 3.82E+07 1.39E+08 4.41 5.38E+18 3.22E+18 2.83E+18 2.16E+18 5 3.93E+07 1.79E+08 5.66 6.75E+18 4.02E+18 3.53E+18 2.70E+18 6 4.09E+07 2.19E+08 6.95 8.19E+18 4.70E+18 4.02E+18 3.15E+18 7 4.26E+07 2.62E+08 8.30 8.92E+ 18 5.24E+ 18 4.58E+18 3.60E+18 8 4.27E+07 3.05E+08 9.66 9.85E+18 5.81E+18 5.04E+18 4.01E+18 9 4.55E+07 3.50E+08 11.10 1.08E+19 6.42E+18 5.52E+18 4.38E+I18 10 4.43E+07 3.95E+08 12.50 1.16E+19 7.OOE+18 6.08E+ 18 4.86E+18 11 4.19E+07 4.36E1+08 13.83 1.23E+19 7.43E+18 6.48E+18 5.25E+18 12 (Pjt) 4.53E+07 4.82E+08 15.27 1.32E+19 7.99E+18 6.96E+18 5.66E+18 13 (Pjt) 4.23E+07 5.24E+08 16.61 1.40E+19 8.52E+18 7.40E+18 6.05E+18 Future 2.2 1IE+08 1.0 IE+09 32.00 2.48E+19 1.58E+19 1.42E+19 1.17E+19 Future 3.79E+08 1.51 E+09 48.00 3.60E+19 2.33E+19 2.12E+19 1.75E+19 Table 3-2 Calculated Neutron Exposure of the Lower Shell Plates (Ml-1004-1, M-1004-2, and M-1004-3)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation Length Time Time Cycle [EFPSJ [EFPS] [EFPM] 00 150 300 450 1 3.28E+07 3.28E+07 1.04 1.48E+18 9.23E+17 8.05E+17 6.35E+17 2 3.18E+07 6.46E+07 2.05 2.67E+ 18 1.63E+18 1.47E+ 18 1.12E+18 3 3.64E+07 1.0 IE+08 3.20 4.05E+18 2.42E+18 2.12E+18 1.611E+18 4 3.82E+07 1.39E+08 4.41 5.36E+18 3.21E+18 2.82E+ 18 2.15E+18 5 3.93E+07 1.79E+08 5.66 6.72E+18 4.01E+18 3.52E+18 2.69E+18 6 4.09E+07 2.19E+08 6.95 8.14E+18 4.68E+18 4.OOE+18 3.14E+18 7 4.26E+07 2.62E+08 8.30 8.87E+ 18 5.21E+18 4.56E+18 3.58E+18 8 4.27E+07 3.05E+08 9.66 9.80E+ 18 5.78E+18 5.01 E+ 18 3.99E+ 18 9 4.55E+07 3.50E+08 11.10 1.07E+19 6.39E+18 5.49E+18 4.36E+18 10 4.43E+07 3.95E+08 12.50 1.16E+19 6.97E+18 6.05E+18 4.84E+18 11 4.19E+07 4.36E+08 13.83 1.22E+ 19 7.40E+ 18 6.45E+18 5.23E+ 18 12 (Pjt) 4.53E+07 4.82E+08 15.27 1.31E+19 7.96E+18 6.93E+18 5.65E+18 13 (Pjt) 4.23E+07 5.24E+08 16.61 1.40E+1 9 8.49E+ 18 7.39E+ 18 6.03E+18 Future 2.21 E+08 1.OIE+09 32.00 2.47E+19 1.57E+19 1.41E+19 1.17E+19 Future 3.79E+08 1.51 E+09 48.00 3.59E+19 2.32E+19 2.12E+19 1.75E+19 Radiation Analysis and Neutron Dosimetry September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 Table 3-3 Calculated Neutron Exposure of the Middle Shell Longitudinal W\elds Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [ncm2']

Length Time Time Weld WVeld Weld Cycle [EFPS] [EFPS] [EFPY] 101-124A 101-124B 101-124C 1 3.28E+07 3.28E+07 1.04 1.48E+18 8.07E+17 8.07E+17 2 3.18E+07 6.46E+07 2.05 2.68E+18 1.47E+18 1.47E+18 3 3.64E+07 1.01 E+08 3.20 4.06E+18 2.12E+18 2.12E+18 4 3.82E+07 1.39E+08 4.41 5.38E+18 2.83E+18 2.83E+18 5 3.93E+07 1.79E+08 5.66 6.75E+18 3.53E+18 3.53E+18 6 4.09E+07 2.19E+08 6.95 8.19E+18 4.02E+ 18 4.02E+18 7 4.26E+07 2.6213+08 8.30 8.92E+ 18 4.58E+18 4.58E+18 8 4.27E+07 3.05E+08 9.66 9.8513+18 5.04E1+18 5.04E+18 9 4.55E+07 3.50E+08 11.10 1.08E+ 19 5.52E+18 5.52E+18 10 4.43E+07 3.95E+08 12.50 1.1613+19 6.08E+18 6.08E+18 II 4.19E+07 4.36E+08 13.83 1.23E+19 6.48E+18 6.48E+18 12 (Pjt) 4.53E+07 4.82E+08 15.27 1.32E+19 6.96E+18 6.96E+18 13 (Pjt) 4.23E+07 5.24E+08 16.61 1.40E1+19 7.40E+18 7.40E+18 Future 2.2 1E+08 1.0 IE+09 32.00 2.48E+19 1.42E+19 1.42E+19 Future 3.79E+08 1.51E+09 48.00 3.60E+19 2.12E+19 2.12E+19 Table 3-4 Calculated Neutron Exposure of the Lower Shell Longitudinal \'elds Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm2I Length Time Time Weld Weld Weld Cycle [EFPS] tEFPSJ fEFPY 101-142A 101-142B 101-142C 1 3.28E+07 3.28E+07 1.04 1.48E+18 8.05E1+17 8.05E+ 17 2 3.18E+07 6.46E+07 2.05 2.67E+18 1.47E+18 1.47E+18 3 3.64E+07 1.01IE+08 3.20 4.05E+18 2.12E+18 2.12E+18 4 3.82E+07 1.39E+08 4.41 5.366E+18 2.82E+18 2.82E+18 5 3.93E+07 1.79E+08 5.66 6.72E+18 3.52E+18 3.52E+18 6 4.09E+07 2.19E+08 6.95 8.14E+ 18 4.00E+18 4.00E+18 7 4.26E+07 2.62E+08 8.30 8.87E+18 4.56E+18 4.566E+18 8 4.27E+07 3.05E+08 9.66 9.80E+18 5.0]E+18 5.01E+18 9 4.55E+07 3.50E+08 11.10 1.07E+19 5.49E+ 18 5.49E+ 18 10 4.43E+07 3.95E+08 12.50 1.16E+19 6.05E+ 18 6.05E+18 11 4.19E+07 4.36E+08 13.83 1.22E+19 6.45E+18 6.45E+18 12 (Pjt) 4.53E+07 4.82E+08 15.27 1.31E+19 6.93E+ 18 6.93E+ 18 13 (Pjt) 4.23E+07 5.24E+08 16.61 1.40E+19 7.39E+18 7.39E+ 18 Future 2.21 E+08 1.01 E+09 32.00 2.4713+19 1.41E+19 1.41E+19 Future 3.79E+08 1.51 E+09 48.00 3.59E+19 2.12E+199 2.12E+19 Radiation Analysis and Neutron Dosimetry September 2003 WCAP-I 6088.doc-092903 Revision I

3-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 3-5 Calculated Neutron Exposure of the Middle Shell to Lower Shell Circumferential WNeld (101-171)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [nlcm 2J Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 30° 450 3.28E+07 3.28E+07 1.04 1.48E+18 9.23E+17 8.05E+17 6.35E+17 2 3.18E+07 6.46E+07 2.05 2.67E+18 1.63E+18 1.47E+18 1.12E+18 3 3.64E+07 1.01E+08 3.20 4.05E+18 2.42E+18 2.12E+18 1.61E+18 4 3.82E+07 1.39E+08 4.41 5.36E+18 3.211E+18 2.82E+18 2.15E+ 18 5 3.93E+07 1.79E+08 5.66 6.72E+18 4.01E+18 3.52E+18 2.69E+18 6 4.09E+07 2.19E+08 6.95 8.14E+ 18 4.68E+18 4.OOE+ 18 3.14E+ 18 7 4.26E+07 2.62E+08 8.30 8.87E+18 5.21E+18 4.56E+18 3.58E+18 8 4.27E+07 3.05E+08 9.66 9.80E+ 18 5.78E+18 5.01]E+18 3.99E+ 18 9 4.55E+07 3.50E+08 11.10 1.07E+ 19 6.39E+ 18 5.49E+ 18 4.36E+18 10 4.43E+07 3.95E+08 12.50 1.16E+19 6.97E+18 6.05E+18 4.84E+18 11 4.19E+07 4.36E+08 13.83 1.22E+19 7.40E+18 6.45E+18 5.23E+18 12 (Pjt) 4.53E+07 4.82E+08 15.27 1.31E+19 7.96E+18 6.93E+18 5.65E+ 18 13 (Pjt) 4.23E+07 5.24E+08 16.61 1.40E+19 8.49E+18 7.39E+18 6.03E+ 18 Future 2.21 E+08 1.0 IE+09 32.00 2.47F,+19 1.57E+19 1.41E+19 1.17E+19 Future 3.79E+08 1.51 E+09 48.00 3.59E+19 2.32E+19 2.12E+19 1.75E+19 Radiation Analysis and Neutron Dosimetry September 2003 WCAP-I 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 4.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kic, for the metal temperature at that time. K1, is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section XI"13' 4] of the ASME Code, Appendix G to Section XI. The K1 , curve is given by the following equation:

Ki. =332+20.734*e[02(T RTNDT)] (4-1)

where, Kc= reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT(ksi 4in.).

This Kic curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

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

C* Kim + Kit < K1, (4-2)

where, Kim = stress intensity factor caused by membrane (pressure) stress Kit = stress intensity factor caused by the thermal gradients K1, = lower bound fracture toughness (function of temperature relative to the RTNDT of the material)

C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Criteria for Allowable Pressure-Temperature Relationships September 2003 WCAP-1 6088.doc-092903 Revision I

4-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 For membrane tension, the corresponding K, for the postulated defect is:

Kaxo = M. x (pRi I t) (4-3) where, Mm for an inside surface flaw is given by:

Mm = 1.85 for t < 2, Mm = 0.926 4 for 2 < 4 < 3.464, Mm = 3.21 for r >3.464 Similarly, Mm for an outside surface flaw is given by:

Mm = 1.77 for 4 < 2, Mm = 0.8931f for 2*< I4 3.464, Mm = 3.09 for 4 > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

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

Klb = Mb

Maximum Stress, where Mb is two-thirds of Mm For the radial thermal gradient, the maximum K, produced by the gradient for the postulated inside surface defect of G-2120 is:

Ki, = 0.953x 10 3 x CR x t25 where CR is the cooldown rate in 'F/hr.

For the radial thermal gradient, the maximum K, produced by the gradient for a postulated outside surface defect is:

K 1, = 0.753x103 x HUx t2 ,5, where HU is the heatup rate in 'F/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-22 14-1 of Appendix G141. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal K1 .

(a) The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2) of Appendix dt4I.

(b) Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress Criteria for Allowable Pressure-Temperature Relationships September 2003 WCAP-16088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 4-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 distribution and at any specified time during cooldown for a l/4-thickness inside surface defect using the relationship:

li, = (1.0359Co+ 0.6322C,+ 0.4753C2+ 0.3855C3) *l; (44) or similarly, Krr during heatup for a l/4-thickness outside surface defect using the relationship:

ID, = (1.043Co + 0.630C, + 0.481C2 + 0.401C3)

4 (4-5) where the coefficients C0 , C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldownN using the form:

v(x) = Co + Ci(x / a) + C2(x / a)2 + C3(x / a)3 (4-6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note that equations 4-3, 4-4 and 4-5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" 12 1 Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K1,, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.

From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of K1 , at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that Criteria for Allowable Pressure-Temperature Relationships September 2003 WCAP-1 6088.doc-092903 Revision 1

4-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 the increase in K1. exceeds K1t, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1. for the 1/4T crack during heatup is lower than the K1, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KI, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curates for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup ratcs is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses that are tensile in nature and, therefore, tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

4.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10CFR Part 50, Appendix G[51 addresses the requirements for the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the unirradiated RTNDT of the material by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3125 psia), which is 625 psia for Waterford Unit 3.

Criteria for Allowable Pressure-Temperature Relationships September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 4.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS The lowest service temperature is the minimum allowable temperature at which pressure can exceed 20%

of the pre-service hydrostatic test pressure (3125 psia) or 0.20 times (1.25*Design Pressure). This temperature is defined by Paragraph NB-2332 of ASME Code Section III['21 as the most limiting RTNDT for the balance of the reactor coolant system components plus 1000 F. The balance of the reactor coolant system components includes consideration of the ferritic materials outside the reactor vessel beltline but within the primary system. (Consideration of the reactor vessel inlet and outlet nozzles is included in the lowest service temperature.)

4.5 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTNDT, in the closure flange region. This requirement is established in IOCFR Part 50, Appendix Gi5l.

Criteria for Allowable Pressure-Temperature Relationships September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + ARTNDT + Margin (5-1)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Codett 01. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

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

ARTNDT = CF

f(o. 2 s-0.0IogI (5-2)

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

fidepth x) = fs,

e (-.4X(5-3) where x is the depth in inches into the vessel wall measured from the inner wetted surface of the vessel at the location of the postulated defect (where vessel beltline thickness is 8.625 inches plus the minimum clad thickness of 0.125 inches). The resultant fluence is then used in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group projected the vessel fluence values as summarized in Section 3 of this report. [Note: In the original analysisl81 , the calculated fluence assumed a 107% RCS flow rate. In a subsequent assessment 13 1 it was determined that the fluence values based on the 107% flow assumption are bounding for the 110% RCS flow case.] The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". Tables 3-1 through 3-5 contain the calculated vessel fluences values (vessel clad base metal interface) at various azimuthal locations, including the peak values in the intermediate shell plates. Table 5-1 contains the peak values of neutron fluence at the inner wetted surface and the peak values of I/4T and 3/4T neutron fluence. Table 5-2 contains the peak values of 1/4T and 3/4T neutron fluence and the fluence factors that were calculated per Regulatory Guide 1.99, Revision 2. (Note that the peak fluence values used correspond to the intermediate shell course plates and welds. Thus it was a conservative assumption when applied to the lower shell plates and welds.)

Margin is calculated as:

M = 2 NGfs + ac ,

Calculation of Adjusted Reference Temperature September 2003 WCAP-I6088.doc-092903 Revision 1

5-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 The standard deviation for the initial RTNDT margin, c;, is 00 F when the initial RTNDT is a measured value, and 171F when a generic value is used. The standard deviation for the shift (ARTNDT), a,&, is 170 F for plates or forgings and 280 F for welds. c, need not exceed 0.5 times the mean value of ARTNDT. When credible surveillance data are used, cA is equal to half the value.

Contained in Tables 5-3 and 5-4 are the calculated ART values used for generation of the heatup and cooldown curves for 32 EFPY. The values of initial RTNDT, chemistry factor (CF), neutron fluence, and fluence factor (FF) are from Tables 2-1, 2-2, 5-1 and 5-2, respectively. The calculation of the ART values is documented in Reference 14.

Calculation of Adjusted Reference Temperature September 2003 WCAP-I 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 5-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 5-1 Summary of the Vessel Surface, 1/4T and 3J4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curve Material Inner Wetted Surface 1/4T 3/4T (n/cm2 ,E > 1.0 MeV) (n/cm2 ,E > 1.0 MeV) (n/cm2 ,E > 1.0 MeV)

Intermediate Shell Plates and 2.56 x 10'9 1.48 x 1095.25 x 10 Axial Weld (Peak Fluence)

Lower Shell Forging and Axial 2.54 x 1019 1.47 x 10'9 5.23 x 10" Weld, and Intermediate to Lower Shell Girth Weld Table 5-2 Summary of the Peak Calculated Fluence Factors used for the Generation of the 32 EFPY Heatup/Cooldown Curve Material 1/4T Fluence 1/4T FF 3/4T Fluence 3/4T FF 2

(n/cm ,E > 1.0 MeV) (n/cm 2

,E > 1.0 MeV)

Intermediate Shell Plates and 1.48 x 10'9 1.109 5.25 x 10t 0.820 Axial Weld (Peak Fluence)

Calculation of Adjusted Reference Temperature September 2003 WCAP-1 6088.doc-092903 Revision I

5-4 WESTINGHOUSE NON-PROPRIETARY CLASS CLASS 33 5-4 WESTINGHOUSE NON-PROPRIETARY Table 5-3 Calculation of the ART Values for the 1/4T Location at 32 EFFY Material RG 1.99 R2 CF FF IRTNDT(s) ARTNDT Margin(c) ART(b)

Method (OF) (IF) (OF) (OF) (OF)

Intermediate Shell Plate M-1003-1 Position 1.1 20 1.109 -30 22 22 14 Intermediate Shell Plate M-1003-2 Position 1.1 20 1.109 -50 22 22 -6 ne-3 Position 1.1 20------------

1.109- --------- 22------ 22.......2.

Intermediate Shell Plate M-1003-3 Position 1.1 20 1.109 -42 22 22 2 Lower Shell Plate M-1004-1 Position 1.1 20 1.109 -15 22 22 29 Lower Shell Plate M-1004-2 Position 2.1 12.4 1.109 22 14 14(d) 50 Lower Shell Plate M-1004-3 Position 1.1 20 1.109 -10 22 22 34 Intermediate Shell Axial Weld Position 1.1 27 1.109 -60 30 30 0 Seams 101-124 A,B,C Lower Shell Axial Weld Seams Position 1.1 35 1.109 -80 39 39 -2 101-142 A,B,C Intermediate to Lower Shell Girth Position 2.1 16.2 1.109 -70 18 18(d -34 Weld Seam 101-171 Notes:

(a) Initial RTNDT values are measured values.

(b) ART = Initial RTNDT + ARTNrijT + Margin (OF)

(c) qa does not have to exceed one-half the predicted shift.

(d) Surveillance data are credible, thus one-half cra may be used in determining the margin term.

Calculation of Adjusted Reference Temperature September 2003 WCAP-I 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-5 WESTINGHOUSE NON-PROPRIETARY CLASS 3 5.5 Table 54 Calculation of the ART Values for the 3/4T Location at 32 EFPY Material RG 1.99 R2 CF FF IRTNDT'S) ARTNDT Margin(c) ART(b)

Method (of;)F) (0F) (OF) ( 0oF)

Intermediate Shell Plate M-1003-1 Position 1.1 20 0.820 -30 16.4 16.4 3 Intermediate Shell Plate M-1003-2 Position 1.1 20 0.820 -50 16.4 16.4 -17

.nte.edite.hellPlae Poitio-1.

. . . . 20-.820

. -- 16.4. ....

16.4- . ---------

Intermediate Shell Plate M-1003-3 Position 1.1 20 0.820 -42 16.4 16.4 19 Lower Shell Plate M- 1004-1 Position 1.1 20 0.820 -15 16.4 16.4 18 Lower Shell Plate M-1004-2 Position 2.1 12.4 0.820 22 10.1 10.1(dl 42 Lower Shell Plate M-1004-3 Position 1.1 20 0.820 -10 16.4 16.4 23 Intermediate Shell Axial Weld Position 1.1 27 0.820 -60 22.1 22.1 -16 Seams 101-124 AB,C Lower Shell Axial Weld Seams Position 1.1 35 0.820 -80 28.7 28.7 -23 101-142 A,B,C Intermediate to Lower Shell Girth Position 2.1 16.2 0.820 -70 13.3 1 3 .3(d) 43 Weld Seam 101-171 Notes:

(a) Initial RTNDT values are measured values.

(b) ART = Initial RTNDT + ARTNDT + Margin (0F)

(c) q, does not have to exceed one-half the predicted shift.

(d) Surveillance data are credible, thus one-half q&may be used in determining the margin term.

Calculation of Adjusted Reference Temperature September 2003 WCAP-I 6088.doc-092903 Revision I

5-6 WESTINGHOUSE NON-PROPRIETARY CLASS3 5-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 The lower shell plate M-1004-2 has the highest predicted ART at 1/4AT and 3/4T at 32 EFPY. (Since the girth weld 101-171 has the lowest ART values, there is no benefit to be gained from using the less restrictive methodology of Code Case N-641.) Contained in Table 5-5 is a summary of the limiting ART values to be used in the generation of the Waterford Unit 3 reactor vessel heatup and cooldown curves.

These curves are presented in Section 6.

Table 5-5 Summary of the Limiting ART Values Used in the Generation of the Waterford Unit 3 Heatup/Cooldown Curves 4 T Limiting ART 'A T Limiting ART 50 ' 42 Calculation of Adjusted Reference Temperature September 2003 WCAP-I 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limits for normal heatup and cooldown and in-service hydrostatic pressure and leak test of the Waterford Unit 3 reactor coolant system at 32 EFPY have been calculatedly using the methods discussed in Sections 4.0 and 5.0 of this report. The PT limit curves were generated based on the latest available reactor vessel beltline region information. This includes updated neutron fluence values calculated considering power uprate to 3441 MWt at the start of Cycle 12 and to 3716 MWt at the start of Cycle 14. WCAP-14040-NP-A, Revision 2 presents the approved PT limit methodology that was used with the exception of those items discussed in Section 1 of this report.

Figure 6-1 presents the limits for heatup without margins for possible instrumentation uncertainties applicable for 32 EFPY. These limits were generated using the 1996 ASME Code Section XI, Appendix G with the limiting axial flaw. Heatup rates of 30, 50 and 60'F/hr were assumed in the evaluation.

Figure 6-2 presents the limits for cooldown without margins for possible instrumentation uncertainties applicable for 32 EFPY. These limits were generated using a combination of the 1996 ASME Code Section XI, Appendix G with the limiting axial flaw ART values, and the ASME Code Case N-641 with the limiting circumferential flaw ART values. Cooldown rates of 0 (steady state), 10, 30, and 100 0 F/hr were assumed in the evaluation. The heatup and cooldown limit curves were evaluated using the following limiting values from Section 5 of ART for lower shell plate M-1004-2 at 32 EFPY:

1/4 T location ART = 500 F 3

/ T location ART = 420 F The limiting pressures and temperatures for the different heatup or cooldown rates and for the other criteria described below produce the bounding heatup limits given in Figures 6-1 and 6-2. The other criteria include those that must be met before the reactor is made critical. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in the figures.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit lines shown in Figure 6-1. The criticality limits are defined by the minimum permissible temperature for the hydrostatic leak test, 1100 F, and by the lowest service temperature (230 0 F for heatup and 1900 F for cooldown) as required by Appendix G to 10 CFR Part 50. (The minimum permissible temperature for the hydrostatic leak test is shown in Table 6-1. It corresponds to the temperature at which the full leak test pressure, 2500 psia, is allowable.) The governing equation for the hydrostatic test is defined in Code Case N-64 1[3] (approved in February 1999) as follows:

1.5 Kim< Kic

where, Kim is the stress intensity factor covered by membrane (pressure) stress, Kic = 33.2 + 20.734 e[1002 (r- TNDT)],

Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-16088.doc-092903 Revision I

6-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Appendix G to 10 CFR Part 5015O. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum permissible temperature for the hydrostatic leak test, 110 0 F, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup calculated as described in Section 4.0 of this report. Because the lowest service temperature is more limiting than the pressure-temperature curves for heatup and cooldown, the minimum temperature for operation above 554.1 psia is defined by the lowest service temperature (2301F for heatup and 190'F for cooldown). For the in-service hydrostatic pressure and leak test, the minimum temperature is 60'F for pressures below 554.1 psia. It corresponds to the boltup temperature (discussed below). The minimum temperature above 554.1 psia for the in-service hydrostatic pressure and leak test is 190'F. Both temperatures correspond to the limits for the core non-critical.

(Note that the pressure and temperature values are given without margins for instrumentation errors.)

The minimum boltup temperature is 60'F as shown in Figure 6-1. The metal temperature of the closure flange region must exceed the unirradiated RTNDT of the material by at least 120'F for normal operation when the pressure exceeds 625 psia. The limiting unirradiated RTNDT of 20'F occurs in the vessel and closure head flange of the Waterford Unit 3 reactor vessel, and the minimum allowable temperature of this region is 1401F at pressures greater than 625 psia. The lowest service temperature is the minimum allowable temperature at which pressure can exceed 20% of the pre-service hydrostatic test pressure (3125 psia) or 0.20 times (1.25*Design Pressure). This corresponds to the most limiting RTNDT for the non-beltline reactor coolant system components plus 1000 F. The maximum RTNDT for the balance of the Waterford Unit 3 reactor coolant system components is 90°F. Therefore, the lowest service temperature is 190°F as shown in Figures 6-1 and 6-2. The 625 psia pressure limit is shown as 554.1 psia in Figures 6-1 and 6-2 to reflect the pressure correction factor of 70.9 psi (625-70.9 = 554.1 psia).

Figures 6-1 and 6-2 define all of the above limits for ensuring prevention of non-ductile failure for the Waterford Unit 3 reactor vessel for 32 EFPY. Figures 6-1 and 6-2 also define the Low Temperature Overpressure Protection (LTOP) alignment temperature1'4 1, 200°F, as a dashed vertical line. The data points used for the pressure and temperature limit heatup and cooldown curves shown in Figures 6-1 and 6-2 are presented in Tables 6-1 and 6-2. Note that the heatup and cooldown curves are defined by the pressure limit based on 20% of the preservice hydrostatic test pressure and by the lowest service temperature. Tables 6-1 and 6-2, therefore, give the same limit on pressure for each of the heatup and cooldown rates because the minimum pressurization temperature and the lowest service temperature are more restrictive than the limits based on the Appendix G K1 c curves. Based on the calculations performedl 4 j, the vessel may be heated up or cooled down at the maximum analyzed rate, 100lF /hr, at temperatures above the bolt-up temperature and the minimum allowable temperature of the flange region (both also subject to the pressure limit), and the lowest service temperature as shown in Figures 6-1 and 6-2. The 100°F /hr rate is also applicable to the in-service hydrostatic pressure and leak test.

Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 2500 1500 n 1500 -C l tCriticality Limit 1250 - tO_100 _i7 deg. F/hr limit u) 1 250 adIsrvc for Heatup in 1000 Non-Critical Limit F 6 for Heatup and Inservice ILi : \

Hydrostatic I Pressure and _ _ - i 750 ~~Leak Test,_ __ _ - _ _ _ _ _-

750 100 deg. Flhr limit 500. 110 deg. F-250 1 1-Boltup Temp.

60 deg. F 0 so 100 150 200 250 300 350 400 450 500 550 Cold-Leg Temperature (Deg. F)

Figure 6-1 Waterford Unit 3 Heatup Curves Applicable to 32 EFPY (without uncertainty for instrumentation errors)

Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-16088.doc-092903 Revision I

64 WESTINGHOUSE NON-PROPRIETARY CLASS 33 6-4 WESTINGHOUSE NON-PROPRIETARY CLASS 2500 .

FOperrim Verslon:5.1 Run:21343l 2250 _LTOP Alignment Temperature, 200 deg. F 2000 ,i -t _ _

Unacceptable Operationi f 1750ii_

0 O o Acceptable Lowest IOperation Service 1500 Temperature, ------

190 deg.F

,n I 1250 l -. _ ___

N4 i ooldown Limit and I000 nservlce -

Hydrostatic Pressure and Leak Test,_100 deg. Fhr 750 _ ---------

l 554.1 psla ,1i 500 __

Boltup fremp. 1 i l ~60 deg. Fl 250 , __

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

Figure 6-2 Waterford Unit 3 Cooldown Cunres Applicable to 32 EFTY (wvithout uncertainty for instrumentation errors)

Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-1 6088.doc-092903 Revision I

WVESTINGHOUSE NON-PROPRIETARY CLASS 3_ 6-5 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 Table 6-1 32 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (with pressure drop, without Uncertainties for Instrumentation Errors) 30°F/hr. Criticality 500 F/hr. Criticality 60°F/hr. Criticality T (°F) P T (°F) P T (F) P T (°F) P (PSIA) T (0 F) P (PSIA) T (0 F) P (PSIA)

(PSIA) (PSIA) (PSIA) -- -

60 15 110 15 60 15 110 15 60 15 110 15 60 554.1 110 554.1 60 554.1 110 554.1 60 554.1 110 554.1 65 554.1 110 554.1 65 554.1 110 554.1 65 554.1 110 554.1 70 554.1 115 554.1 70 554.1 115 554.1 70 554.1 115 554.1 75 554.1 120 554.1 75 554.1 120 554.1 75 554.1 120 554.1 80 554.1 125 554.1 80 554.1 125 554.1 80 554.1 125 554.1 85 554.1 130 554.1 85 554.1 130 554.1 85 554.1 130 554.1 90 554.1 135 554.1 90 554.1 135 554.1 90 554.1 135 554.1 95 554.1 140 554.1 95 554.1 140 554.1 95 554.1 140 554.1 100 554.1 145 554.1 100 554.1 145 554.1 100 554.1 145 554.1 105 554.1 150 554.1 105 554.1 150 554.1 105 554.1 150 554.1 110 554.1 155 554.1 110 554.1 155 554.1 110 554.1 155 554.1 115 554.1 160 554.1 115 554.1 160 554.1 115 554.1 160 554.1 120 554.1 165 554.1 120 554.1 165 554.1 120 554.1 165 554.1 125 554.1 170 554.1 125 554.1 170 554.1 125 554.1 170 554.1 130 554.1 175 554.1 130 554.1 175 554.1 130 554.1 175 554.1 135 554.1 180 554.1 135 554.1 180 554.1 135 554.1 180 554.1 140 554.1 185 554.1 140 554.1 185 554.1 140 554.1 185 554.1 145 554.1 190 554.1 145 554.1 190 554.1 145 554.1 190 554.1 150 554.1 195 554.1 150 554.1 195 554.1 150 554.1 195 554.1

-155- 554.1 -200 554.-I 155-- 554.1- --200-- -554.1 ---- 155 554.1 200- 554.1-160 554.1 205 554.1 160 554.1 205 554.1 160 554.1 205 554.1 165 554.1 210 554.1 165 554.1 210 554.1 165 554.1 210 554.1 170 554.1 215 554.1 170 554.1 215 554.1 170 554.1 215 554.1 175 554.1 220 554.1 175 554.1 220 554.1 175 554.1 220 554.1 180 554.1 225 554.1 180 554.1 225 554.1 180 554.1 225 554.1 185 554.1 230 554.1 185 554.1 230 554.1 185 554.1 230 554.1 190 554.1 230 2500 190 554.1 230 2500 190 554.1 230 2500 190 2500 - - 190 2500 .1. _ 190 2500 -- i- _:

0.- e --i-j. .. ; ~<HydroStatic

-;ti ; Leak Test.'- :t -~--',-.iP L(LF) 1 93 i 110 P(psia) 2015 l 2500 _

Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-1 6088.doc-092903 Revision I

6-6 WESTINGHOUSE NON-PROPRIETARY CLASS 33 6-6 WESTINGHOUSE NON-PROPRIETARY CLASS Table 6-2 32 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 (with pressure drop, svithout Uncertainties for Instrumentation Errors)

Steady State 10 0 F/hr 30F/hr. [ 100OF/hr.

T.(°F) P(PSIA) T (°F)l P(PSIA) Tj(F) P(PSLA) T(OF) P(PSIA) 60 15 60 15 60 15 60 15 60 554.1 60 554.1 60 554.1 60 554.1 65 554.1 65 554.1 65 554.1 65 554.1 70 554.1 70 554.1 70 554.1 70 554.1 75 554.1 75 554.1 75 554.1 75 554.1 80 554.1 80 554.1 80 554.1 80 554.1 85 554.1 85 554.1 85 554.1 85 554.1 90 554.1 90 554.1 90 554.1 90 554.1 95 554.1 95 554.1 95 554.1 95 554.1 100 554.1 100 554.1 100 554.1 100 554.1 105 554.1 105 554.1 105 554.1 lOS 554.1 110 554.1 110 554.1 110 554.1 110 554.1 115 554.1 115 554.1 115 554.1 115 554.1 120 554.1 120 554.1 120 554.1 120 554.1 125 554.1 125 554.1 125 554.1 125 554.1 130 554.1 130 554.1 130 554.1 130 554.1 135 554.1 135 554.1 135 554.1 135 554.1 140 554.1 140 554.1 140 554.1 140 554.1 145 554.1 145 554.1 145 554.1 145 554.1 150 554.1 150 554.1 150 554.1 iS0 554.1 155 554.1 155 554.1 155 554.1 155 554.1

'160 554.1 160 554.1 -160 554.1 160- 554.1 165 554.1 165 554.1 165 554.1 165 554.1 170 554.1 170 554.1 170 554.1 170 554.1 175 554.1 175 554.1 175 554.1 175 554.1 180 554.1 180 554.1 180 554.1 180 554.1 185 554.1 185 554.1 185 554.1 185 554.1 190 554.1 190 554.1 190 554.1 190 554.1 190 2500 190 2500 190 2500 190 2500 Heatup and Cooldown Pressure-Temperature Limit Curves September 2003 WCAP-16088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 EVALUATION OF SCREENING CRITERIA In this section, the Waterford Unit 3 reactor vessel beltline materials are evaluated relative to the Pressurized Thermal Shock (PTS) screening criteria of 10CFR50.6 1E6] and the upper shelf screening criteria of 10' CFR Part 50, Appendix G151. The PTS values are calculated in accordance with IOCFR50.6116]. The predicted upper shelf energy values are evaluated using the methods of Regulatory Guide 1.99, Revision 2 r1], for each beltline material. The calculation of the PTS and upper shelf energy values, as documented in Reference 15, represents power uprate conditions, including a 1.5% uprate (3441 MWt) at the start of Cycle 12 and a 8% uprate (3716 MWt) at the start of Cycle 14.

Best estimate copper (Cu) and nickel (Ni) weight percent values and the initial RTNDT values are the same as provided in Table 1. The determination of the chemistry factor values per Position 1.1 and 2.1 of Reference I is detailed in Reference 8 and summarized in Table 2. These data are repeated in Table 7-1 and 7-2. In these tables, the "Chemistry Factor Basis" refers to values from Table 1 and 2 of 10CFR50.6116] for weld and plate, respectively, and to "surveillance data" for the values derived in Reference 8. The neutron fluence is the peak value for the corresponding plate and weld for the two time periods given, 32 EFPY and 48 EFPY. (All fluence values were determined8 ly using ENDF/B-VI cross-sections and followed the guidance in Regulatory Guide 1.19019].)

RTmswas determined for each material in the beltline region is given by the following expression:

RTprs = Initial RTNDT + ARTpTs + Margin (7-1)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III-of the ASME Boiler and Pressure Vessel Code~t91 .- Measured values of initial RTNDT are available for each of the materials.

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

ARTm = CF

fo8-0-.0 logf) (7-2) where f is the vessel fluence at the clad-base metal interface given in units of 1019 n/cm2 . The appropriate fluence for each material is given in Tables 7-1 and 7-2.

Margin is determined based on the uncertainty in initial RTNDT and the uncertainty in the ARTp prediction. Margin is calculated as:

M = 2 jc4; + ai As stated in Note (a) in Tables 7-1 and 7-2 the initial RTNDT values are based on measured values and, therefore, a, is equal to 00F. The uncertainty in the ARTpTs prediction, (;, is 28 0 F for welds and 170F for plates. However, the value of 2 0A does not have to exceed ARTprs. (Note: The value of cA for the two materials for which credible surveillance data are available does not have to exceed 140 F for welds and Evaluation of Screening Criteria September 2003 WCAP-16088.doc-092903 Revision 1

7-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8.5 0F for base metal.) The values of ai, a&, and the total margin are given for each material in Tables 7-1 and 7-2. Total margin in these cases is 2GA or ARTm, whichever is smaller.

Values of RTm are given for each material in Tables 7-1 and 7-2. The highest values are 530 F and 550 F for lower shell plate M-1004-2 after 32 and 48 EFPY, respectively. All the projected values for the Waterford Unit 3 reactor vessel beltline materials are well below the Pressurized Thermal Shock (PTS) screening criteria of 2700 F for axial welds and plates, and 300'F for circumferential welds.

The predicted upper shelf energy values for each of the Waterford Unit 3 beltline material are evaluated in Table 7-3 in accordance with 10 CFR Part 50, Appendix G15 ] using Position 1.2 of Regulatory Guide 1.99, Revision 211. The predictions were based on the predicted fluencel 8 l at the vessel '/4T location after 32 and 48 EFPY. (Note: The peak beltline fluence at /4 T was used for all the beltline materials rather than taking credit for the relatively small variation in the axial fluence between the intermediate and lower shells.) The projected upper shelf energy values far exceed the 50 ft-lb screening criterion of I 0CFR50, Appendix G'5 1 after 32 and 48 EFPY. In addition, when the Waterford Unit 3 surveillance measurements were comparedr8] to projections based on Position 1.2 of Regulatory Guide 1.99, the measurements were significantly less than the projections. This is likely a result of the controls placed on both copper content and on other residual elements during the fabrication of the plates and welds. For the plate material, this may also result from using the 0.05% copper lower bound curve from Figure 2 of Regulatory Guide 1.99 to predict upper shelf energy decrease for plates containing 0.02 to 0.03% copper. It is likely that the projections in Table 7-3 will be conservative for these reasons.

The plate and weld data from the surveillance capsule analyses[ 8 ] were also evaluated using Position 2.2 of Regulatory Guide 1.99, Revision 2[']. The projections of upper shelf energy decrease were based on the upper bound of the measurements extended parallel to the lines in Figure 2 of Reference 1. The projected shelf decrease for plate M-1004-2 based on the measurements is 14% and 16% after 32 and 48 EFPY, respectively. The projected shelf decrease for weld 101-171 based on the measurements is 9.5% and 11.5% after 32 and 48 EFPY, respectively. The projected upper shelf energies are given below after 32 and 48 EFPY, respectively. The corresponding values based on Position 1.2 from Table 7-3 are also shown below. This comparison demonstrates the low radiation sensitivity of the Waterford Unit 3 beltline materials, and it adds confidence to the expectation that those materials will far exceed the 50 ft-lb screening criterion of 10CFR50, Appendix G151 after 32 and 48 EFPY.

Comparison of Upper Shelf Energy Decrease Predictions Material 32 EFPY USE 48 EFPY USE 32 EFPY USE 48 EFPY USE (ft-lbs) using (ft-lbs) using (ft-lbs) using (ft-lbs) using Position 1.2 Position 1.2 Position 2.2 Position 2.2 Plate M-1004-2 117 109 121 118 Weld 101-171 124 121 141 138 Evaluation of Screening Criteria September 2003 WCAP-16088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 g 7-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 Table 7-1 Calculation of the Waterford Unit 3 RTrrs Values for 32 EFPY Material Chemistry CF Fluence RTNDT°" ARTr~b) CT - A Margin RTm (T)

Factor Basis (0F) (xi Ohnlcm2) (OF) (0 F) (OjF) (OF) (IF)

-0 - --- - -

Intermediate Shell Platc M-1003-1 Table 2 20 2.48 -30 24.9 0 12.4 24.9 20 Intermediate Shell Plate M-1003-2 Table 2 20 2.48 -50 24.9 0 12.4 24.9 0 Intcrmcdiate Shell Plate M-1003-3 Table 2 20 2.48 -42 24.9 0 12.4 24.9 A 8 Lower Shell Plate M-1004-1 Table 2 20 2.47 -5 24.9 0 12.4 24.9 35 Lower Shell Plate M-1004-2 Surveillance 12.4 2.47 22 15.4 0 7.7 15.4 53 Data Lower Shell Plate M-1004-3 Table 2 20 2.47 -10 24.9 0 12.4 24.9 40 Intermediate Shell Longitudinal Weld Table 1 27 2.48 -60 33.6 0 16.8 33.6 7 SeamslOl-124 A,B,C Lower Shell Longitudinal Wcld Table 1 35 2.47 -80 43.5 0 21.8 43.5 7 Seams 10 1-142 A,B,C Intermediate to Lower Shell Girth Surveillance 16.2 2.47 -70 20.1 0 10.1 20.1 -30 Weld Seam 101-171 Data Notes:

(a) Initial reference temperature (RTNDT) values are measured. Thus, a,is equal to 00 F.

(b) ARTrs = CF

FF (c) RTm = Initial RTNDT + ARTprs + Margin (fF)

Evaluation of Screening Criteria September 2003 WCAP- 16088.doc-092903 Revisinn 1

74 WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 7-2 Calculation of the Waterford Unit 3 RTprs Values for 48 EFPY Material Chemistry CF Fluence RTNDT(') ART (A MarginTm Factor Basis (OF) (x1O9nlcm2) ( 0F) (0F) (OF) (OF) (`Ff)

Intermediate Shell Plate M-1003-1 Table 2 20 3.60 -30 26.7 0 13.3 26.7 23 Intermediate Shell Plate M-1003-2 Table 2 20 3.60 -50 26.7 0 13.3 26.7 3 Intermediate Shell Plate M-1003-3 Tablc 2 20 3.60 -42 26.7 0 13.3 26;7 11 Lower Shell Plate M-1004-1 Table 2 20 3.59 -15 26.6 0 13.3 26.6 38 Lower Shell Plate M-1004-2 Surveillance 12.4 3.59 22 16.5 0 8.3 16.5 55 Data Lower Shell Plate M-1004-3 Table 2 20 3.59 -10 26.6 0 13.3 26.6 43 Intermediate Shell Longitudinal Weld Table 1 27 3.60 -60 36.0 0 18.0 36.0 12 Seams 101-124 A,B,C Lower Shell Longitudinal Weld Table 1 35 3.59 -80 46.6 0 23.3 46.6 13 Seams 101-142A,B,C Intermediate to Lower Shell Girth Surveillance 16.2 3.59 -70 21.6 0 10.8 21.6 -27 Weld Seam 101-171 Data Notes:

a) Initial reference temperature (RTNDT) values are measured. Thus, a, isequal to 00F.

b) ARTm=CF*FF c) RTrr = Initial RTNDT + ARTm + Margin (fF)

Evaluation of Screening Criteria September 2003 WCAP-1 6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 Table 7-3 Calculation of the Waterford Unit 3 Upper Shelf Energy Values at Vessel 1/4T Material Copper Initial USE 32 EFPY 48 EFPY 32 EFPY 48 EFPY Content (%) (ft-tb) .(XlFluence 1 Fene1 2 .USE (ft-lb) USE (ft-lb) 0 'nlcm 2) .(Xl0 1'nl ) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Intermediate Shell Plate M-1003-1 0.02 94[3a 1.48 2.15 74 73 Intermediate Shell Plate M-1003-2 0.02 97 a1a 1.48 2.15 77 75 Intermediate Shell Plate M-1003-3 0.02 goCa1 1.48 2.15 71 70 Lower Shell Plate M-1004-1 0.03 106ta1 1.48 2.15 84 82 Lower Shell Plate M-1004-2 0.03 141 bI1.48 2.15 117 109 Lower Shell Plate M-1004-3 0.03 94Ca] 1.48 2.15 74 73 Intermediate Shell Longitudinal Weld Seam 0.02 1061c] 1.48 2.15 84. 82 101-124 A Intermediate Shell Longitudinal Weld 0.02 131C[l 1.48 2.15 104 101 SeamslOI-124 B,C Lower Shell Longitudinal Weld Seams 101- 0.03 129M 1.48 2.15 102 t00 142 A,B,C Intermediate to Lower Sheil Girth Weld 0.05 156(b 1.48 2.15 124 121 Seam 101-171 Notes:

a) Initial transverse orientation upper shelf energy based on 65 percent of longitudinal Charpy impact data, Reference 15.

b) Initial transverse orientation upper shelf energy based on surveillance plate and weld Charpy impact data (mean of all values at 100% shear, Reference 16).

c) Initial upper shelf energy from Reference 11.

Evaluation of Screening Criteria September 2003 WCAP- I6088.doc-092903 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-NP-A, Revision-2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves," J.D.

Andrachek, et. al., January 1996.

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

[Sub Reference 1: ASME Code Case N-640, "Alternative Reference FractureToughnessfor Development ofP-TLimit Curvesfor Section Xl, Division 1, "Section Xl, Division 1, Approved February26, 1999.]

[Sub Reference 2: ASAE Boiler andPressure Vessel Code, Case N-588, "Attenuation to Reference Flaw OrientationofAppendix Gfor Circumferential Welds in Reactor Vessels, "Section XI, Division 1, ApprovedDecember 12, 1997.]

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

5. Code of Federal Regulations, 10 CFR Part 50, Appendix G "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

6. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December l9, 1995.

7. "Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

8. WCAP-16002, Revision 0, "Analysis of Capsule 2630 from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program," March 2003.

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

10. 1989 Edition,Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Article NB-233 1, 'Material for Vessels."

11. C-MECH-93-074, "Waterford Unit 3 Reactor Vessel Beltline Welds; Initial Upper Shelf Energy," ABB Combustion Engineering Letter dated October 22, 1993.

12. 1986 Edition,Section III, Division I of the ASME Boiler and Pressure Vessel Code, Article NB-2332.

13. LTR-REA-03-92, "Evaluation of Waterford Unit 3 Reactor Pressure Fast Neutron Fluence Projections for Increased Reactor Coolant System Flow Rate Conditions," Westinghouse Letter dated June 26, 2003.

References

September 2003 WCAP-1 6088.doc-092903 Revision 1

8-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

14. CN-EMT-03-60, "Waterford Unit 3 Pressure-Temperature Limit Curves," Westinghouse Calculation Note dated July 2003.

15. CN-CI-03-38, "Waterford Unit 3 Power Uprate for Reactor Vessel Beltline Materials,"

Westinghouse Calculation Note dated June 6, 2003.

16. TR-C-MCS-002, "Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program, Louisiana Power & Light, Waterford Steam Electric Station Unit No.

3," Combustion Engineering Report dated October 1977.

References September 2003 WCAP-1 6088.doc.092903 Revision I

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W3F1-2003-0075, Ses, Unit 3, License Amendment Request NPF-38-250, Revision to Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years

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SUBJECT:

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1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

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W3F1-2003-0075, Ses, Unit 3, License Amendment Request NPF-38-250, Revision to Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years

Text

SUBJECT:

Dear Sir or Madam:

Enclosure:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

SUMMARY

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