NRC 2004-0085, License Renewal Application Revised Information

License Renewal Application Revised Information

ML042660308
Person / Time
Site:	Point Beach
Issue date:	09/10/2004
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2004-0085, TAC MC2099, TAC MC2100
Download: ML042660308 (42)

Text

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Committed to Nuclear Excelln Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC September 10, 2004 NRC 2004-0085 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant, Units I and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR-27 License Renewal Application Revised Information (TAC Nos. MC2099 and MC 2100)

Reference:

1) Letter from NMC to NRC dated February 25, 2004 (NRC 2004-0016)

2) Letter from NMC to NRC dated August 3, 2004 (NRC 2004-0079)

3) NRC Memorandum from Luis A. Reyes to Chairman Diaz, et al, dated May 27, 2004, "Pressurized Thermal Shock Analyses for Renewal of Certain Nuclear Power Plant Operating Licenses" In Reference 1, Nuclear Management Company, LLC (NMC), submitted the Point Beach Nuclear Plant (PBNP) Units 1 and 2 License Renewal Application (LRA).

In Reference 2, NMC withdrew an associated request for exemption to 10 CFR 50.61 and Appendices G and H to 10 CFR 50. As a result, LRA sections that were presupposed on approval of those exemptions required revising. The revised LRA sections, 4.1.2, Identification of Exemptions, 4.2.1, Reactor Vessel Pressurized Thermal Shock, 4.2.2, Reactor Vessel Upper Shelf Energy, 4.2.3, Reactor Vessel Pressure/Temperature Limits, Appendix A, Program and Time Limited Aging Analysis Descriptions, and B2.1.18, Reactor Vessel Surveillance Program, are hereby submitted for review.

This revised information takes credit for use of the 10 CFR 54.21(c)(1)(iii) option as discussed in Reference 3. Enclosure 1 contains the revised LRA sections and provides a demonstration that the effects of aging on the intended functions of each system, structure, and component will be adequately managed for the period of extended operation. As a result, we have elected to not extend the existing TLAA at this time. contains BAW-2467NP, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units 1 and 2 for Extended Life through 53 Effective Full Power Years", dated July 2004. BAW-2467NP is an analysis for satisfying the reactor vessel Charpy upper-shelf energy requirements, which is being submitted 6590 Nuclear Road

• Two Rivers, Wisconsin 54241 Telephone: 920.755.2321

Document Control Desk Page 2 for review and approval in accordance with the requirements of 10 CFR 50, Appendix G,Section IV.A.1.c. Please note that this is a non-proprietary version of the analysis. The proprietary version will be provided upon request.

Should you have any questions concerning this submittal, please contact Mr. James E. Knorr at (920) 755-6863.

Summary of Commitments Commitments made as part of this submittal are listed as follows:

1. PBNP will continue to implement the low-low leakage loading fuel management pattern to minimize the limiting weld fluence. In addition, PBNP will continue operation with Hafnium absorber assemblies in service until the resolution of the Unit 2 intermediate-to-lower shell girth weld PTS issue via an alternative analysis methodology.

2. Documentation of a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 (b) for the Unit 2 Reactor Pressure Vessel intermediate-to-lower shell girth weld will be completed within one -year of receipt of the extended license. Documentation within this time frame will support submittal of any required safety analysis at least three years prior to the time frame that RTPTS for Unit 2 is projected to exceed the screening criteria.

3. If acceptable PTS results cannot be provided prior to EOL with alternate analysis techniques, the PBNP flux reduction program will evaluate the feasibility and practicality of pursuing additional aggressive flux reduction measures prior to EOL, such as the insertion of part length shielded fuel assemblies.

I declare under penalty of perjury that the forgoing is true and correct. Executed on September 10, 2004.

Dennis L. Koehl Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosures cc:

Administrator, Region ll, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE I REVISED SECTIONS 4.1.2, 4.2.1, 4.2.2, 4.2.3, APPENDIXES A 15.2.18,15.4.1,15.5, AND B 2.1.18 TO THE POINT BEACH NUCLEAR PLANT, UNITS I AND 2 LICENSE RENEWAL APPLICATION Following are revised sections of the Point Beach Nuclear Plant License Renewal Application which reflect the use of 10 CFR 50.61 as a program to manage the effects of aging on reactor vessel integrity. These sections replace in total those in the February 25, 2004, License Renewal Application submittal with the same numbers.

Note that the references listed in the text of these sections are the same as those referenced in the original submittal with the addition of Reference 75.

Table 4.1-2 Time limited Aging Analyses Line number 1, column 5, in Table 4.1-2, Time Limited Aging Analyses (TLAA) is changed to; "(iii) effects of aging on the intended function will be adequately managed for the period of extended operation."

4.1.2 Identification of Exemptions The requirements of 10 CFR 54.21 (c) stipulate that the application for a renewed license should include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12, and that are in effect, based on time-limited aging analyses, as defined in 10 CFR 54.3.

Active 10 CFR 50.12 exemptions were reviewed to determine whether the exemption was based on a time-limited aging analysis. No TLAA related exemptions granted pursuant to 10 CFR 50.12 were identified.

4.2 Reactor Vessel Irradiation Embrittlement This group of time-limited aging analyses concerns the effect of irradiation embrittlement on the belt-line regions (adjacent to the reactor core) of the Point Beach Nuclear Plant Units 1 and 2 reactor vessels, and how this mechanism affects analyses that provide operating limits or address regulatory requirements. The calculations discussed in this section use predictions of the cumulative effects on the reactor vessels from irradiation embrittlement. The calculations are based on periodic assessment of the neutron fluence and resultant changes in the reactor vessel material fracture toughness.

The intermediate and lower shells, and welds that join them in the beltline region, of the reactor vessel are fabricated from low alloy steels. These ferritic steels exhibit a ductile-brittle transition that results in fracture toughness property changes as a function of both temperature and irradiation. The material property of particular importance in Page 1 of 40

assessing reactor vessel integrity is fracture toughness, which can be defined as the capability of a material to resist sudden failure caused by crack propagation. Fracture toughness is reduced by neutron irradiation. The measure of fracture toughness of the reactor vessel materials when the reactor vessel is above the brittle fracture/ductile failure transition temperature is referred to as upper-shelf energy. Upper-shelf energy is related to the ability of a material to resist ductile tearing. In addition, the temperature at which the brittle fracture/ductile failure transition occurs increases with increasing radiation. This shift in the transition temperature is referred to as the shift in reference nil ductility transition temperature (RTNDT).

The effect of embrittlement due to neutron bombardment is evaluated for reactor vessel temperatures throughout the range of normal operating values. Heatup and cooldown curves consider normal, relatively slow thermal transients. Pressurized thermal shock transients are characterized by a rapid and significant decrease in reactor coolant temperature with high pressure in the reactor vessel. The high reactor vessel thermal stresses, when combined with the pressure stresses, are assumed to initiate the propagation of a small flaw that is postulated to exist in the reactor vessel beltline.

Postulated high pressures could cause propagation of the flaw through the reactor vessel wall.

The first step in addressing the TLAAs associated with neutron embrittlement is the projection of the neutron fluence that the critical vessel locations experience. The Westinghouse Radiation Engineering and Analysis Group performed PBNP reactor vessel fluence projections. The evaluations used the ENDF/B-VI scattering cross-section data set. The calculated fluence projections were determined using methods consistent with Regulatory Guide 1.190, uCalculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

These fluence projections were based on historical operational data, and forecasted uprated (1678 MWt) power conditions using a low-low leakage fuel management pattern (L4P) without the presence of Hafnium power suppression absorber rods. The fluence projections performed in 2002 assumed that the units were uprated to 1678 MWt in 2002. The 2002 fluence projections were used as the input basis for the Upper Shelf Energy (USE) evaluation required by 10 CFR 50, Appendix G, and the RCS Pressure-Temperature (P-T) Operating Limits required by 10 CFR 50, Appendix G.

The fluence projections were revised in 2004 to account for actual unit operational history (including the 1.7 % mini-uprates performed in 2003), and full unit uprates to 1678 MWt in 2008. The 2004 fluence projections were used as the input basis for the Pressurized Thermal Shock evaluation required by 10 CFR 50.61.

The fluence projections were performed at uprated power conditions to allow for future unit operations at the increased power level.

These fluence projections are bounding and conservative. The analyses for both units have been performed with fluences projected at 53 EFPYs (Effective Full Power Years).

The EOEL (End of Extended License) EFPYs for Unit 1, assuming a 95% capacity Page 2 of 40

factor, is forecasted to be 51 EFPYs. The EOEL EFPYs for Unit 2, assuming a 95%

capacity factor, is forecasted to be 53 EFPYs.

The results of the calculated neutron fluence values at various locations on the Reactor Pressure Vessel (RPV) are presented in Table 4.2-1.

Table 4.2-1 Summary of the Calculated RPV Neutron Fluence Values at 53 EFPY (1019 nlcm2, E > 1.0 MeV)

Component 1 Fluence 1 Surface (i) 1/4T (ii) 3/4T (ii)

Description J Projection ]

I I

PBNP Unit 1: 53 FPY (End of License Extension )

Nozzle Belt 2002 0.42 0.28 0.13 Forging (122P237) 2004 0.38 Inter. Shell Plate 2002 5.26 3.56 1.63 (A9811-1) 2004 5.21 Lower Shell Plate 2002 4.79 (iii) 3.24 1.49 (C1423-1) 2004 4.83 Nozzle Belt to 2002 0.42 0.28 0.13 Intermediate Shell Circ.

2004 0.38 Weld (8T1 762)

Intermediate 2002 3.44 2.32 1.07 Shell Axial Weld - ID 27%

2004 3.39 (1 P081 5)

Intermediate 2002 3.44 2.32 1.07 Shell Axial Weld - OD 73%

2004 3.39 (1 P0661)

Intermediate to 2002 4.91 3.32 1.52 Low er__

Shell Circ. Weld 2004 4.71 (71249)

Lower Shell Axial 2002 3.37 2.28 1.05 Weld (61782) 2004 3.25 Page 3 of 40

Component l

Fluence 1 Surface (i) [

1/4T (ii) l 3/4T (ii)

Description j

Projection l

PBNP Unit 2: 53 FPY (End of License Extension Nozzle Belt 2002 0.55 0.37 0.17 Forging (123V352) 2004 0.53 Inter. Shell 2002 5.39 3.65 1.67 Forging__

(123V500) 2004 5.26 Lower Shell 2002 5.32 3.60 1.65 Forging (122W195) 2004 5.11 Nozzle Belt to 2002 0.55 0.37 0.17 Intermediate Shell Circ.

2004 0.53 Weld (21935)

Intermediate to 2002 5.09 3.46 1.58 Lower Shell Circ. Weld 2004 4.85 (72442)

i. These fluence values are the calculated fluence values considering power uprate (1678 MWt) without Hafnium suppression rods.

ii. Neutron attenuation per Reg. Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials (Draft ME 305-4, Proposed Revision 2, published 02/1986), Rev. 2."

iii. 4.79 is in error due to a calculational summary transpositional anomaly. The correct value is 5.06. There is no affect on the generation of the PT curves or the conclusions of the Upper Shelf Energy Equivalent Margins Analysis.

In addition to the plant specific neutron exposure calculations, dosimetry sets from three (3) in-vessel and twenty (20) ex-vessel sensor sets irradiated at Unit 1 and four (4) in-vessel and twenty (20) ex-vessel sensor sets irradiated at Unit 2 were also re-analyzed using dosimetry evaluation methodologies that follow the guidance provided in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." The results of these dosimetry re-evaluations were then used to validate the calculational models that were applied in the plant specific neutron transport analysis of the PBNP RPVs.

The welds in the reactor vessel are basically the same material as the parts being joined and may be considered to be included in the preceding discussions. The chemistry differences between weld metal and base metal affect the material properties that are degraded by embrittlement; therefore, the welds are evaluated separately when considering the aforementioned aging effect. The fracture toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance Page 4 of 40

with the NRC Standard Review Plan. The beltline material properties of the Point Beach Unit 1 and 2 reactor vessels are presented in Table 4.2-2. The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1. Position 1.1 uses the tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date.

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Table 4.2-2 Summary of the Best Estimate Cu and Ni Weight Percent, Initial RTNDT Values and Chemistry Factor values for PBNP Units I and 2 Reactor Vessel Materials Material Description wt.% Cu l wt.% Ni Initial CF I

RTNDT PBNP Unit I Nozzle Belt Forging 0.11 0.82 500F 770F (122P237)

Inter. Shell Plate 0.20 0.06 10F 880F (A981 1-1) 79.30F(i)

Lower Shell Plate 0.12 0.07 10F 55.30F (C1423-1) 35.80F(i)

Nozzle Belt to 0.19 0.57

-50F 152.40F Intermediate Shell Circ. Weld (8T1 762)

Intermediate Shell Axial 0.17 0.52

-50F 138.20F Weld - ID 27%

(1 P0815)

Intermediate Shell Axial 0.17 0.64

-50F 157.60F Weld - OD 73%

(1 P0661)

Intermediate to Lower 0.23 0.59 10F 167.60F Shell Girth Weld (71249)

Lower Shell Axial Weld 0.23 0.52

-50F 157.40F (61782) 163.3°F(i)

PBNP Unit 2 Nozzle Belt Forging 0.11 0.73 40°F 76°F (123V352)

Inter. Shell Forging 0.09 0.70 40°F 58°F (123V500)

Lower Shell Forging 0.05 0.72 40°F 31°F (122W195) 43-F (i)

Nozzle Belt to 0.18 0.70

-56°F(ii) 170°F Intermediate Shell Girth Weld (21935)

Intermediate to Lower 0.26 0.60

-5°F 180°F Shell Circ. Weld (72442)

i. Per Regulatory Guide 1.99, ii. Generic Value of RTNDT-Rev. 2, Position 2.1.

The calculated fluences and RPV material properties noted above were used in the TLAA calculations associated with RPV neutron embrittlement.

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In addition, changes in RPV material properties are verified through surveillance specimen irradiation and testing.

Westinghouse Electric Company developed the original surveillance program for the PBNP Units 1 and 2 RPVs. Although the original program was in accordance with ASTM E 185-66, subsequent testing has followed the latest version of ASTM E 185 that was approved by the NRC, through ASTM E 185-82. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-7513 for Unit 1 (Reference 57), and WCAP-7712 for Unit 2 (Reference 58). The original PBNP surveillance program consisted of six surveillance capsules in each Unit attached to the outside of the reactor vessel internals thermal shield. Each capsule contained mechanical specimens, dosimetry, and thermal monitors. The mechanical specimens were fabricated from material representative of the PBNP RPVs.

To date, four surveillance capsules have been removed and tested from each Unit's RPV. One of the standby capsules has also been removed from each Unit's RPV and is being stored at Point Beach. The final, originally installed standby capsule, remains in each PBNP RPV.

The surveillance materials in the capsules of PBNP Units I and 2, and other early plant specific Reactor Vessel Surveillance Programs (RVSPs) were not selected in accordance with ASTM E 185-82. Hence, the materials monitored by the RVSPs are not always the materials judged in 10 CFR 50 Appendix H, to most likely be the controlling beltline region materials with regard to irradiation embrittlement for the RPV for which the RVSP was designed. Consequently, the applicability of the data generated in the plant specific RVSP is limited. However, by combining the data developed from several RVSPs, it is possible to use data developed in a given RVSP for application at a different RPV, and also practical to develop a database to predict irradiation behavior of those welds for which there is no specific data. This does not preclude plant specific characterization should sufficient credible surveillance data become available.

Although the PBNP Units 1 and 2 specific surveillance program capsules contained mechanical specimens representative of the materials of the PBNP RPVs, the capsules did not contain materials representative of the PBNP RPVs limiting welds. Since the actual heat of the limiting weld metal for either of the PBNP Units RPV is not in the respective Unit's surveillance program, participation in the B&W Owners Group (B&WOG) Master Integrated Reactor Vessel Surveillance Program (MIRVP)

(Reference 59) allows access to irradiated surveillance data of the PBNP limiting RPV welds.

The MIRVP combines 16 separate plant specific RVSPs and provides for sharing of irradiation sites. It addresses requirements for acquiring irradiation data and the need to improve the quality and quantity of fracture toughness data to support operation of the participating plants.

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The MIRVP correlates data from both power reactor surveillance monitoring and test reactor research programs. The principal sources of information are the power reactor surveillance efforts; which consists of three parts. The first part is the continuation of the plant-specific RVSPs that monitor the irradiation damage to selected materials, as originally planned and licensed. These capsules contain samples of weld metal, plate, forging, and heat-affected zone (HAZ) material from the vessel beltline, and neutron dosimetry and thermal monitors. This part of the program will continue to monitor the long-term effects of neutron irradiation on the reactor vessel materials.

The second part of the program consists of a series of specially designed supplementary weld metal surveillance capsules (SUPCAPS) to study the effects of irradiation on a number of weld metals. The welds were selected because they were anticipated to be highly sensitive to irradiation damage because of their chemical composition and low initial Charpy upper shelf energies. These capsules differ from regular plant-specific RVSP capsules in that they include the necessary specimens to obtain fracture toughness properties of individual weld metals. The capsules are located in the same irradiation holder tubes as the regular plant-specific capsules at Crystal River-3 and Davis Besse.

The third part of the MIRVP consists of higher fluence supplementary weld metal surveillance capsules (HUPCAPS) to obtain irradiated weld metal data (primarily fracture toughness properties) to satisfy the requirements 10 CFR 50, Appendices G and H, and 10 CFR 50.61 for the current license and license renewal of the plants in the MIRVP. Additional objectives are to (1) provide a capsule of Westinghouse design for correlation of irradiation data in the Westinghouse neutronic environment with the B&W 177-FA environment; (2) provide irradiation of reconstituted specimens to accelerate data gathering; and, (3) provide definitive information on the annealing response of this family of materials.

The PBNP Unit I remaining original plant-specific RVSP capsule contains SA-1 263 weld material that is a surrogate for SA-1 585 and SA-1 650, but is not relevant to the PBNP Unit 1 RPV limiting weld materials. The limiting beltline welds for PBNP Unit 1 are SA-847 and SA-1 01. SA-847 is covered by surrogate materials SA-1 036 in Ginna, and SA-1 135 in SUPCAPS. The SA-1 101 material is in the Turkey Point Unit 3 RVSP and SA-1 094, a surrogate for SA-1 101, is in the Turkey Point Unit 4 RVSP. SA-1 263 benefits Surry Units 1 and 2 and Oconee Unit 1. However, it is covered in the SUPCAPS and HUPCAPS and no additional data is required for this weld material.

These MIRVP capsules contain several Charpy V-notch and compact fracture toughness specimens of the WF-847 & SA-1 101 weld material. EOEL data currently exists for the SA-847 surrogate material SA-1 036. Additionally, Capsule A2 will be removed at a target EOEL fluence of 3.7 x 1019 n/cm2. This capsule will be removed in approximately 2008 and will be used in EOEL evaluations of the PBNP Unit 1 SA-847 material.

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In addition, a new PBNP surveillance capsule has been installed in PBNP Unit 2 for the purpose of obtaining relevant fracture toughness data at the EOEL fluence. The new PBNP Unit 2 surveillance capsule contains surveillance specimens that will be used to determine the fracture toughness of the PBNP Unit I weld material SA-1 101. When removed and tested, this surveillance capsule will provide EOEL data for the SA-1 101 weld material. The supplemental surveillance capsule for PBNP Unit 2 was installed following Cycle 25. Details regarding the specific contents of the supplemental capsule may be found in WCAP-1 5856.

The PBNP Unit 2 remaining original plant-specific RVSP capsule contains WF-1 93 weld material that is a surrogate for WF-1 12 and WF-I 54, but is not relevant to the PBNP Unit 2 RPV limiting weld materials. The limiting beltline weld for PBNP Unit 2 is SA-1484. HUPCAP A3 provided data on SA-1484 weld material with a fluence of 1.7 x 1019 n/cm2. The WF-67 weld was produced using the same weld wire (heat 72442) as the SA-1484 weld and is well characterized in the SUPCAPS and HUPCAPS.

These MIRVP capsules contain several Charpy V-notch and compact fracture toughness specimens of the WF-67 weld material. Two of these capsules have a target fluence of 3.0 x 1019 n/cm2, which is approximately the projected EOL fluence for PBNP Unit 2. Capsule Al was scheduled for removal from Davis Besse in 2008. Capsule A4 in Crystal River 3 should be available at about this same time depending upon the actual operating schedule. The exact status for capsule Al will depend upon a revised operation schedule at Davis Besse. Capsule L2 in Davis Besse has a lower target fluence and thus has little relevance for the PBNP Unit 2 vessel. When any or all of these specimens are tested, the new results will be integrated with the existing data to further assess RPV integrity.

In addition, as stated previously, a new PBNP Unit 2 surveillance capsule has been installed in PBNP Unit 2 for the purpose of obtaining relevant fracture toughness data at the EOEL fluence. The new Unit 2 surveillance capsule contains surveillance specimens that will be used to determine the fracture toughness of the PBNP Unit 2 weld metal heat 72442, as well as, weld and plate materials from PBNP Unit 1 RPV and a weld for the Davis Besse RPV. The supplemental surveillance capsule for PBNP Unit 2 was installed following Cycle 25. Details regarding the specific contents of the supplemental capsule may be found in WCAP-15856 (Reference 60).

The target fluence for the PBNP Unit 2 supplemental surveillance capsule will correspond to the peak reactor vessel fluence at EOEL for the limiting weld metal.

Surveillance data obtained from this capsule will provide fracture toughness measurements for the limiting weld metal at EOEL fluence. The EOEL peak fluence estimate for the PBNP Unit 2 circumferential weld is 5.085 x 1019 n/cm and considers the affects of hafnium removal and power uprate. The resulting data will provide direct evidence to demonstrate adequate reactor vessel fracture toughness throughout the license renewal term.

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Based on the fluence lead factor for the capsule irradiation location, the supplemental surveillance capsule should be removed and tested at just over 38 EFPY (representing an EOEL fluence for the capsule materials).

The PBNP Units 1 and 2 specific reactor vessel surveillance program coupled with participation in the B&WOG MIRVP meets the requirements of 10 CFR 50 Appendix H.

There are three distinct time-limited aging analyses associated with Reactor Vessel Irradiation Embrittlement:

• Pressurized Thermal Shock evaluation required by 10 CFR 50.61.

• Upper Shelf Energy (USE) evaluation required by 10 CFR 50, Appendix G.

• RCS Pressure-Temperature (P-T) Operating Limits required by 10 CFR 50, Appendix G.

Each of these analyses is discussed separately below.

4.2.1 Reactor Vessel Pressurized Thermal Shock A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooking (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.

In 1985, the NRC issued a formal rule on PTS, 10 CFR 50.61. It established the screening criteria for pressurized water reactor (PWR) vessel embrittlement as measured by the reference temperature termed RTPTS. Screening criteria were set corresponding to EOL plant operation for beltline axial weld seams, forgings, and plates at 2700F, and at 3000F for beltline circumferential weld seams. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with these criteria through the end of plant operation.

The NRC amended its regulations for PWR plants to change the procedure for calculating radiation embrittlement RTpTS values. The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14, 1991, and later updated on December 19, 1995 with an effective date of July 29, 1996.

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These amendments made the procedure for calculating RTpTS values consistent with the method given in Regulatory Guide 1.99, Revision 2.

Pressurized thermal shock analyses were performed for the PBNP Unit 1 RPV by Westinghouse using the 2004 fluence projections. These analyses were performed at full uprated power conditions (1678 MWt), without Hafnium absorber rods, for a 60-year operating period. RTpTS values were calculated for the inside surface of the beltline region materials for the Unit 1 RPV using Charpy based fracture toughness evaluations in accordance with the methods of 10 CFR 50.61. The values are summarized in Table 4.2.1-1.

Table 4.2.1-1 Summary of Unit I Calculated RTpTs Values RPV Inside Surface, 53 EFPY, 1678 MWt, Without Hafnium, Charpy Based Methodology Component Description Fluence Factor ARTpTS(0F)

RTpTS(°F)

Nozzle Belt Forging 0.73 56.21 140 (122P237)

Inter. Shell Plate 1.41 124.08 189 (A9811-1) 111.81 (i)169 (i)

Lower Shell Plate 1.40 77.42 142 (C1423-1) 50.12 (i) 107 (i)

Nozzle Belt to 0.73 111.25 174 Intermediate Shell Circ.

Weld (8T1 762)

Intermediate Shell Axial 1.32 182.42 245 Weld - ID 27%

(1 P0815)

Intermediate to Lower 1.39 232.96 299 Shell Circ. Weld (71249)

Lower Shell Axial Weld 1.31 158.71 222 (61782)

__213.92 (i) 257 (i)

i. Per Regulatory Guide 1.99, Rev. 2, Position 2.1.

The RTpTS values for the Unit 1 Reactor Pressure Vessel beltline region materials at the EOEL were calculated to be lower than the applicable screening criteria values established in 10 CFR 50.61. The analyses associated with PTS for the Unit 1 RPV have been projected to the end of the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).

PTS analyses were performed for the PBNP Unit 2 RPV by Westinghouse using the 2004 fluence projections. These analyses were performed at full uprated power conditions (1678 MWt), without Hafnium absorber rods, for a 60-year operating period.

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RTPTS values were calculated for the inside surface of the beltline region materials for the Unit 2 RPV using Charpy based fracture toughness evaluations in accordance with the methods of 10 CFR 50.61. The values are summarized in Table 4.2.1-2.

Table 4.2.1-2 Summary of Unit 2 Calculated RTpTs Values RPV Inside Surface, 53 EFPY, 1678 MWt, Without Hafnium - Charpy Based Methodology Component Description Fluence Factor ARTpTS(0F)

RTPTs(0F)

Nozzle Belt Forging 0.82 62.32 136 (123V352)

Inter. Shell Forging 1.41 81.78 156 (123V500)

Lower Shell Forging 1.41 43.71 118 (122W195) 60.63 (i) 118 (i)

Nozzle Belt to 0.82 139.40 149 Intermediate Shell Circ.

Weld (21935)

Intermediate to Lower 1.40 252.00 316 Shell Circ. Weld (72442)

i. Per Regulatory Guide 1.99, Rev. 2, Position 2.1.

The RTpTS values for the Unit 2 Reactor Pressure Vessel beltline region materials at the end of the extended operating period were calculated to be lower than the applicable screening criteria values established in 10 CFR 50.61, with the exception of the intermediate to lower shell circumferential weld. The intermediate to lower shell circumferential weld is the limiting Unit 2 Reactor Pressure Vessel weld.

It should be noted that all the RTPTS values are lower than the screening criteria values established in 10 CFR 50.61 for the current license period.

As shown in the above Table, the EOEL fluence yields an RTpTS value of 316'F when using Charpy based methods for the limiting weld for a power uprate to 1678.0 MWt and removal of the Hafnium power suppression assemblies. The screening criteria established in 10 CFR 50.61 (3000F) will be exceeded for the limiting Unit 2 intermediate to lower shell girth weld at a neutron fluence of 3.31 X 1019 n/cm2. The 2004 fluence projections indicate that the limiting weld will experience this fluence at 38.1 EFPYs. Assuming a long-term capacity factor of 95 %, this fluence would be achieved late in 2017.

10 CFR 50.61 (b)(3) states 'For each pressurized water nuclear power reactor for which the value of RTpTs for any material in the beltline is projected to exceed the PTS screening criterion using the EOL fluence, the licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS Page 12 of 40

screening criterion set forth in paragraph (b)(2) of this section. The schedule for implementation of flux reduction measures may take into account the schedule for submittal and anticipated approval by the Director, Office of Nuclear Reactor Regulation, of detailed plant-specific analyses, submitted to demonstrate acceptable risk with RTPTs above the screening limit due to plant modifications, new information or new analysis techniques."

PBNP Flux Reduction Program The PBNP Reactor Vessel Surveillance Program will include a flux reduction program to manage the Unit 2 RPV intermediate-to-lower shell girth weld PTS issue for the period of extended operation in accordance with 10 CFR 50.61 (b)(3).

PBNP RPV embrittlement was recognized as a potential issue early in plant life. A flux reduction program was implemented for both PBNP Units. The initial flux reduction efforts incorporated a "low leakage loading pattern" (L3P) fuel management plan. The low leakage loading pattern fuel management plan was implemented in 1980 for both PBNP Units I & 2. Incorporation of the low leakage loading pattern fuel management plan provided a vessel flux reduction of approximately 25 to 30 percent.

PBNP RPV embrittlement was recognized as a potentially limiting issue in early plant life extension studies. Westinghouse performed a Reactor Vessel Flux Reduction Evaluation to identify the best means of reducing the rate of neutron embrittlement of the reactor vessels. Neutron flux reduction goals were established to maintain RPV weld properties within perceived regulatory limits through EOEL. A basic criteria imposed on the flux reduction program was that any flux reduction measures would not adversely affect plant reliability and capacity.

Fourteen (14) possible fuel management techniques, and lower internals redesign and replacement were considered. Redesign and replacement of the lower internals package was not found to be practical. Two (2) of the evaluated fuel management techniques met the neutron flux reduction goals. These 2 fuel management techniques included a low-low leakage loading pattern (L4P) with Hafnium neutron absorber assemblies in the guide tubes of peripheral fuel assemblies (12 locations per Unit), and a low-low leakage loading pattern with modified fuel assemblies, with stainless steel rods - in place of fuel rods, on the outboard side of peripheral fuel assemblies (12 locations per Unit). The technique consisting of a low-low leakage loading pattern with modified fuel assemblies was not considered practical, and was rejected based on cost, reduced core thermal design margins, increased possibility of fuel assembly damage, and an increased amount of material that needed to be disposed of as high-level nuclear waste.

Implementation of a flux reduction program was determined to be prudent, and the flux reduction technique of a low-low leakage loading pattern (L4P) with hafnium neutron absorber assemblies was selected. The transition to a low-low leakage loading pattem, Page 13 of 40

and installation of the hafnium neutron absorber assemblies was performed during the refueling outages in April 1989 for Unit 1 and October 1989 for Unit 2.

To verify the analytical flux reduction predications, ex-vessel neutron dosimetry sets were installed in the reactor cavity annulus.

The transition to a L4P plus Hafnium flux reduction program achieved an approximate flux reduction factor of 1.5 to 2, and was initially forecasted to achieve the goals for EOEL. Subsequently, incorporation of extended cycles (> 12 months) increased projected EOEL fluences through increased flux rates and an increase in unit capacity factor. In addition, changes occurred in the definition of limiting weld material properties and PTS calculational constraints. These issues resulted in the forecasted Unit 2 RPV limiting weld PTS value exceeding the acceptance criteria at EOEL.

The PTS screening criteria will not be met at EOEL for the limiting Unit 2 weld even if the power level is held at the current CLB level (1540 MWt), with the continued presence of Hafnium neutron absorber assemblies installed in the peripheral fuel assemblies.

Flux reduction measures that limit plant capacity (power level or capacity factor) are not considered reasonable or practical options. Maximizing power output is an important factor in ensuring that PBNP continues to provide a cost competitive product.

PBNP will continue to implement the low-low leakage loading fuel management pattern to minimize the limiting weld fluence. In addition, PBNP will continue operation with Hafnium absorber assemblies in service until the resolution of the Unit 2 intermediate-to-lower shell girth weld PTS issue via an alternate analysis methodology.

Continued funding of Hafnium neutron absorber assemblies over the long term is not reasonable since the flux reduction provided by these devices does not prevent the limiting weld PTS value from exceeding the acceptance criteria.

Improved Analysis Technique Option The current ASME Code reference toughness methodology is generally very conservative. The Master Curve method of using measured, small-specimen, fracture toughness properties to define a new indexing temperature, with a statistically derived lower bound tolerance curve, more accurately describes the fracture toughness behavior of RPV materials.

The Master Curve fracture toughness approach has precedence with the Nuclear Regulatory Commission (NRC) as evidenced in the Safety Evaluation (SE) issued for the Kewaunee Nuclear Power Plant approving the application of a Master Curve based methodology for the RPV (Reference 68).

Page 14 of 40

A PBNP Unit 2 specific RTpTS calculation was performed using the Master Curve methodology of BAW-2308, Revision 1, and is documented in Framatome ANP Calculation 32-5019743-01, UPBNP Unit 2 Power Uprate PTS Evaluation 53 EFPY,"

Revision 1, 08/19/2003 (Reference 74). Application of the Master Curve methodology for the limiting Unit 2 RPV girth weld demonstrates acceptable PTS values at EOEL.

Therefore, detailed plant-specific analyses can be submitted to demonstrate acceptable RTpTS below the screening limit by application of new analysis techniques.

Aggressive Flux Reduction Program Option Flux reduction strategies that required radical assembly design modifications were initially rejected during the original flux reduction program evaluations based on initial cost, increased fuel cycle costs, reduced core thermal design margins, increased possibility of fuel assembly damage, and an increased amount of material that needed to be disposed of as high-level nuclear waste.

One option that has the potential to achieve flux reduction factors approaching a factor of ten is the use of part length shielded fuel assemblies (PLSA) in the twelve peripheral fuel assembly locations on the core flats, i.e., along the core major axes. These are the locations where the Hafnium power suppression assemblies are currently located. This approach has been successfully implemented at the H. B. Robinson Plant at the onset of Cycle 10 in 1984.

Although the H. B. Robinson plant is a Westinghouse designed 3-loop reactor, the core configuration along the major axes is similar to the 2-loop configuration characteristic of Point Beach Unit 2. Both designs include twelve assemblies along the core flats with the main difference being the size of the individual assemblies. Further, for both reactor designs, the maximum pressure vessel fluence occurs along the major axes. Therefore, based on judgment, shielding effects similar to those observed at H. B. Robinson may be achievable at Point Beach. It should be noted that detailed feasibility studies of implementing this option at PBNP have not been performed.

In the design of the PLSA assemblies, the fuel pellets in a portion of the fuel assembly are replaced by stainless steel. This effectively removes the peripheral neutron source opposite the location of the limiting circumferential weld and, in addition, provides some shielding to prevent neutrons from the core interior from reaching the vessel.

In designing PLSA assemblies for potential application at Point Beach Unit 2, the specific axial location of the intermediate shell to lower shell circumferential weld must be taken into account. This location along with the desired degree of flux reduction will dictate the axial elevation of the center of the stainless steel pellet array, as well as, the total height of the steel region. Once a conceptual design of the PLSA assemblies is available, potential impacts on design, operation, and cost can be assessed.

Page 15 of 40

One potentially significant advantage of a flux reduction plan based on the insertion of the PLSA assemblies is that the large flux reduction achieved by this approach allows the actual implementation to be scheduled well into the future. Assuming 1678 MWt, without Hafnium, and a flux reduction factor of 8, it appears feasible that Unit 2 could meet the current PTS acceptance criteria at EOEL if PLSAs were incorporated at EOL.

Pursuit of PLSA assemblies at this time is not considered reasonable or practical in view of cost, loss of core design margins, schedule requirement, and the capability of providing acceptable PTS results with alternate analysis techniques.

If acceptable PTS results cannot be provided prior to EOL with alternate analysis techniques, the PBNP flux reduction program will evaluate the feasibility and practicality of pursuing additional aggressive flux reduction measures prior to EOL, such as the insertion of part length shielded fuel assemblies.

Safety Analysis Option 10 CFR 50.61 (b) (4) states "For each pressurized water nuclear power reactor for which the analysis required by paragraph (b)(3) of this section indicates that no reasonably practicable flux reduction program will prevent RTpTS from exceeding the PTS screening criterion using the EOL fluence, the licensee shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results, and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis must be submitted at least three years before RTpTS is projected to exceed the PTS screening criterion."

Preliminary Plant Life Extension studies were performed to determine the viability of technical acceptance of the PBNP reactor vessels throughout a 20-year license renewal term. Westinghouse performed a comprehensive scoping risk assessment for the PBNP Reactor Vessels that is documented in WCAP-1 1676, "Scoping Risk Assessment for the PBNP Units I and 2 Reactor Vessel Life Extension Study", 1987. The Westinghouse report evaluated PBNP for PTS based on NRC Regulatory Guide 1.154.

The analysis was performed without crediting any flux reduction measures. The report concluded that both PBNP Units 1 and 2 will be well within (by two orders of magnitude) the acceptance criteria of R.G. 1.154 throughout the 20-year license renewal term without any flux reduction measures being implemented. The fluence values used in the Westinghouse analysis conservatively envelope the current fluence projections.

Therefore, based on the results of the preliminary analysis, it is believed that a safety analysis in accordance with 10 CFR 50.61(b) (4) can be prepared which demonstrates acceptable risk from PTS for the license renewal term.

Page 16 of 40

Annealing Option 10 CFR 50.61 (b)(7) states "If the limiting RTpTs value of the plant is projected to exceed the screening criteria in paragraph (b)(2), or the criteria in paragraphs (b)(3) through (b)(6) of this section cannot be satisfied, the reactor vessel beltline may be given a thermal annealing treatment to recover the fracture toughness of the material, subject to the requirements of § 50.66. The reactor vessel may continue to be operated only for that service period within which the predicted fracture toughness of the vessel beltline materials satisfy the requirements of paragraphs (b)(2) through (b)(6) of this section, with RTpTS accounting for the effects of annealing and subsequent irradiation."

Thermal annealing of RPVs is a proven technology for recovering vessel material properties that have been degraded due to long-term exposure to neutron irradiation.

Previous anneals have been performed principally in Eastern Europe on RPVs designed by the former Soviet Union.

Although there is no experience with annealing an operating commercial U.S. RPV, the Marble Hill Annealing Demonstration Project demonstrated that an anneal of a commercially sized U.S. RPV is technically feasible utilizing existing equipment and procedures.

Thus, annealing the PBNP Unit 2 RPV is a potential method to manage aging degradation associated with the loss of fracture toughness.

Conclusion Maintenance of current power level, or continued operation with Hafnium suppression inserts in service is not necessary at this time since either the pursuit of an alternate fracture toughness evaluation methodology, pursuit of an aggressive flux reduction program in the future, pursuit of risk analysis, or pursuit of RPV annealing will provide technically acceptable methods of achieving EOEL with the PBNP RPVs.

The Reactor Vessel Surveillance Program will provide reasonable assurance that the Unit 2 RPV intermediate-to-lower shell girth weld PTS issue will be adequately managed for the period of extended operation in accordance with 10 CFR 50.61, per 10 CFR 54.21 (c)(1 )(iii).

4.2.2 Reactor Vessel Upper Shelf Energy The requirements on reactor vessel Charpy upper-shelf energy are included in 10 CFR 50, Appendix G. Specifically, 10 CFR 50, Appendix G requires licensees to submit an analysis at least 3 years prior to the time that the upper-shelf energy of any of the reactor vessel material is predicted to drop below 50 ft-lb., as measured by Charpy V-notch specimen testing. Limiting PBNP RPV weld materials fall below the Page 17 of 40

10 CFR 50, Appendix G, requirement of 50 ft-lb. Consequently, fracture mechanics evaluations were performed to demonstrate acceptable equivalent margins of safety against fracture.

The B&W Owners Group (B&WOG) performed equivalent margins analysis for B&W RPVs. The PBNP RPVs are included within the scope of the analyses. These analyses were performed assuming an original licensed power level, a low-low loading pattern with Hafnium, and EOL conditions. The analyses demonstrated acceptable equivalent margins of safety against fracture. The analyses are summarized in B&W Owners Reactor Vessel Working Group reports BAW-2178PA (Reference 63),

"Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Level C & D Service Loads," and BAW-2192PA (Reference 64), "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Level A & B Service Loads," both dated April 1994. The NRC staff reviewed and approved both of these reports for referencing in licensing applications in separate safety evaluations on March 29, 1994 (Reference 65 and Reference 66).

Additional equivalent margins analyses have been performed for the PBNP RPVs to address the uprated power condition of 1678 MWt, without Hafnium power suppression absorber rods installed, and at EOEL conditions. The 2002 fluence projections were used to define EOEL vessel fluences. These analyses used the same methodologies described in the above references. The analyses, performed by Framatome ANP, are summarized in BAW-2467P, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units 1 and 2 for Extended Life through 53 Effective Full Power Years", July 2004 (Reference 75).

The revised analysis addresses ASME Levels A, B, C, and D Service Loadings. PBNP specific transient information was reviewed and incorporated into the plant specific analyses. There are no Level C service load transients specified for PBNP. For conservatism, three Level D transients were evaluated. These included the Reactor Coolant Line Break (LOCA), the FSAR Steam Line Break, and the RPV Equipment Specification Steam Line Break transients. The LOCA transient is the most limiting transient.

For Levels A and B Service Loadings, the low upper-shelf toughness analysis is performed according to the acceptance criteria and evaluation procedures contained in Appendix K to Section Xl of the ASME Code. The evaluation also utilizes the acceptance criteria and evaluation procedures prescribed in Appendix K for Levels C and D Service Loadings. Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal and stress models and linear elastic fracture mechanics methodology of Framatome-ANP's PCRIT computer code to determine stress intensity factors for a worst case pressurized thermal shock transient.

The analysis shows that the ASME Code, Section Xl, Appendix K acceptance criteria have been satisfied for Levels A, B, C, and D Service Loadings.

Page 18 of 40

The limiting weld for the Upper Shelf Energy analysis is the SA-847 axial weld of the Point Beach Unit 1 RPV (at an uprated power condition without Hafnium power suppression assemblies).

The analysis for Levels A and B service loadings shows that with factors of safety of 1.15 on pressure and 1.0 on thermal loading, the applied J-integral (Jr) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (Jo.,). The ratio Jo., / Jr = 1.87 which is significantly greater than the required value of 1.0. The analysis for Levels A and B service loadings also shows that with a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable.

The EOEL lower bounding J-R values and all acceptance ratios for Levels A and B Service Loadings are summarized in Table 4.2.2-1.

Table 4.2.2-1 EOEL Lower Bounding J-R Values and Acceptance Ratios Levels A and B Service Loadings Unit Weld Weld Lower Acceptance Acceptance Number Orientation Bounding Criterion I Criterion 2 J0.r @1I4T J1 Jo.1IJ1 Jro.1/J (lblin)

(lb/in)

(lb/in)

Unit SA-1101 Circ.

608 98 6.20 113 5.38 1

SA-847 Long.

618 331 1.87 388 1.59 Unit SA-1484 Circ.

578 104 5.56 119 4.86 2

1__

The Unit 2 RPV intermediate-to-lower shell circumferential weld SA-1484 contains the minimum lower bounding J-R value at EOEL of 578 lb/in. The controlling weld is the Unit 1 RPV longitudinal weld SA-847. The minimum ratio of material J-R to applied J for acceptance criterion I and 2 at EOEL is 1.87 and 1.59 respectively. Since the values of the J-R ratios are greater than one, the acceptance criteria for the equivalent margins analysis have been met.

The analysis for Levels C and D service loadings shows that with a factor of safety of 1.0 on loading, flaw extensions are ductile and stable. The analysis for Levels C and D service loadings also shows that the flaw remains stable at much less than 75% of the vessel wall thickness. It has also been shown that the remaining ligament is sufficient to preclude tensile instability by a large margin.

The analysis associated with upper-shelf energy has been projected to the end of the period of extended operation in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).

Page 19 of 40

4.2.3 Reactor Vessel Pressure/Temperature Limits Atomic Energy Commission (AEC) General Design Criterion (GDC) 14 of 10 CFR 50, Appendix A, 'Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage or rapid failure and of gross rupture. Likewise, GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary be designed with sufficient margin to reasonably assure that when stressed by operation, maintenance, and testing conditions, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. GDC 32, Inspection of Reactor Coolant Pressure Boundary,"

requires an appropriate materials surveillance program for assessing the structural integrity of the reactor vessel's beltline region.

Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced 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 600F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactors life, RTNDT 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 NRC has published a method for predicting irradiation embrittlement in Regulatory Guide 1.99, Revision 2, uRadiation Embrittlement of Reactor Vessel Materials". 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 measured from the clad/base metal interface.

New heatup and cooldown pressure temperature (PT) limit curves have been developed for normal operation of PBNP Units 1 and 2 reactor pressure vessels through EOEL.

The PT curves were generated based on the 2002 fluence projections assuming power uprating (1678 MWt) and the removal of the Hafnium absorber rods.

Page 20 of 40

The heatup and cooldown curves were generated using the NRC approved methodology documented in WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 67) with the 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 Kic critical stress intensities are used in place of the Kia critical stress intensities. This methodology is taken from approved ASME Code Case N-641 (which covers Code Cases N-640 and N-588); 3) The 1996 Version of Appendix G to Section Xl will be used rather than the 1989 version; and 4) PT Curves were generated with the most limiting circumferential weld ART value in conjunction with Code Case N-588. These curves are bounded by the curves using the standard "axial" flaw methodology from ASME Code 1996 App. G with the ART from the intermediate or lower shell axial welds depending of the flaw location, 1/4T versus 3/4T.

The new Point Beach Unit I and 2 heatup and cooldown pressure-temperature limit curves were generated using adjusted reference temperature (ART) values that bound both units. The highest ART values from the two units were from the Unit I and Unit 2 intermediate to lower shell girth welds, however the limiting materials are actually the intermediate and lower shell axial welds from Unit 1, depending on the vessel thickness (1/4 T or 3/4 T location). The axial welds become limiting over the girth weld through use of mcirc-flaw" methodology from ASME Code Case N-588. This methodology is less restrictive than the standard "axial-flaw" methodology from the 1995 ASME Code,Section XI through the 1996 Addenda. In addition to the use of Code Case N-588, the PT curves also made use of ASME Code Case N-640, which allows the use of the Kic methodology. Both ASME Code Case N-588 and N-640 were joined together under ASME Code Case N-641.

The calculation of heat-up and cool-down curves requires ART values at the 1/4-thickness (1/4T) and 3/4-thickness (3/4T) through wall locations corresponding to the peak fluence for the girth (circumferential) weld. The attenuation of fluence through the wall of the RPV was determined using the method in Regulatory Guide 1.99, Rev. 2.

Contained in Table 4.2.3-1 is a summary of the limiting ARTs used in the generation of the PBNP Units 1 and 2 PT limit curves.

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Table 4.2.3-1 Summary of the Limiting ART Values Used in the Generation of the PBNP Units I and 2 Heatup/Cooldown Curves Limiting "Circ-Flaw" ART (I)

Limiting "Axial-Flaw" ART EFPY (OF) 1/4T(°F) l 3/4T (OF) 1/4T(°F) 3/4T (0 F)

PBNP Unit I (ii) 53 286 254 243 224 PBNP Unit 2 (iii) 53 301 267 1

152 140

i. PBNP Units 1 and 2 Limiting Circ. Flaw ART comes from the Intermediate to lower shell circumferential welds (Heat #s 71249 and 72442, respectively) ii.The "Axial-Flaw" ARTs for PBNP Unit 1 are from the lower shell axial welds (1/4T) and the intermediate shell axial welds (3/4T) iii.The 'Axial-Flaw" ARTs for PBNP Unit 2 are from the intermediate shell forging 123V500.

Limiting heatup curves were generated using heatup rates of 60 and 1 00F/hr for 53 EFPY (EOEL). These curves were generated using a combination of the 1996 ASME Code Section Xl, Appendix G with the limiting ART values from the Unit 1 intermediate and lower shell longitudinal welds and the ASME Code Case N-641. These heatup curves bound those generated using the 'Circ-flaw" methodology portion of ASME Code Case N-641 with the limiting circ-weld ART values from the Unit 1 or 2 intermediate to lower shell girth weld.

Limiting cooldown curves were generated using cooldown rates of 0, 20,40, 60 and 1 000F/hr for 53 EFPY (EOEL). Again, these curves were generated using a combination of the1996 ASME Code Section Xl, Appendix G with the limiting ART values from the Unit 1 intermediate and lower shell longitudinal welds and the ASME Code Case N-641. These cooldown curves bound those generated using the "Circ-flaw" methodology portion of ASME Code Case N-641 with the limiting circ-weld ART values from the Unit I or 2 intermediate to lower shell girth weld.

The PT curves include a hydrostatic leak test limit curve from 2485 psig to 2000 psig, along with the pressure-temperature limits for the vessel flange region per the requirements of 10 CFR 50, Appendix G. A copy of these PT limit curves was provided to the NRC in Reference 70.

In addition, maximum allowable low-temperature, overpressure protection system (LTOPS) power-operated relief valve (PORV) lift setpoints have been developed for 53 EFPY (EOEL), based on the P-T limits applicable to the period of operation.

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The analysis associated with reactor vessel pressure-temperature limit curves has been projected to the end of the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(ii).

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Appendix A Revisions The proposed FSAR Aging Management Program discussion as provided in Appendix A of the License Renewal Application is as follows.

15.2.18 Reactor Vessel Surveillance Program The Reactor Vessel Surveillance Program manages the aging effect reduction of fracture toughness due to neutron embrittlement of the low alloy steel reactor vessels. Monitoring methods will be in accordance with 10 CFR 50, Appendix H. This program includes (a) capsule insertion, withdrawal and materials testing/evaluation, (including upper shelf energy and RTNDT determinations), (b) fluence and uncertainty calculations, (c) monitoring of Effective Full Power Years (EFPY), (d) development of pressure-temperature limitations, (e) determination of low-temperature overpressure protection (LTOP) set points, and (f) implementation of a flux reduction program, and other options as necessary, allowed by 10 CFR 50.61 (b) for the Unit 2 intermediate-to-lower shell girth weld. The program ensures the reactor vessel materials (a) meet the fracture toughness requirements of 10 CFR 50, Appendix G, and (b) have adequate margins against brittle fracture caused by Pressurized Thermal Shock (PTS) in accordance with 10 CFR 50.61.

Section 15.4 EVALUATION OF TIME-LIMITED AGING ANALYSES As part of a License Renewal Application, 10 CFR 54.21 (c) requires that an evaluation of time-limited aging analyses (TLAAs) for the period of extended operation be provided. The following TLAAs have been identified and evaluated to meet this requirement. These discussions are numbered and inserted into the FSAR sections where these subjects are covered.

15.4.1 Reactor Vessel Irradiation Embrittlement PBNP Units 1 and 2 reactor vessels are described in Chapters 3.0 and 4.0.

Time-limited aging analyses (TLAAs) applicable to the reactor vessels are:

o Pressurized thermal shock o Upper-shelf energy o Pressure-temperature limits The Reactor Vessel Surveillance Program manages reactor vessel irradiation embrittlement utilizing subprograms to monitor, calculate, and evaluate the time-dependent parameters used in the aging analyses for pressurized Page 24 of 40

thermal shock, upper-shelf energy, and pressure-temperature limit curves to ensure continuing vessel integrity through the period of extended operation.

Reactor Vessel Pressurized Thermal Shock The requirements in 10 CFR 50.61 provide rules for protection against pressurized thermal shock events for pressurized water reactors. Licensees are required to perform an assessment of the projected values of the maximum nil ductility reference temperature (RTpTs) whenever a significant change occurs in projected values of RTpTS, or upon request for a change in the expiration date for the operation of the facility.

The calculated RTpTs values at the end of life extension for the PBNP Units 1 and 2 reactor vessels are less than the 10 CFR 50.61(b)(2) screening criteria of 2700F for intermediate and lower shells and 3000F for the circumferential welds, with the exception of the Unit 2 RPV intermediate-to-lower shell circumferential weld.

The EOEL fluence yields an RTpTS value of 3160F when using Charpy based methods for the limiting weld of the Unit 2 RPV. The screening criteria established in 10 CFR 50.61 (3000F) will be exceeded for the limiting Unit 2 intermediate to lower shell girth weld at a neutron fluence of 3.31 X 10'9 n/cm2. The 2004 fluence projections indicate that the limiting weld will experience this fluence at 38.1 EFPYs. Assuming a long-term capacity factor of 95 %, this fluence would be achieved late in 2017.

The PBNP Reactor Vessel Surveillance Program includes a flux reduction program to manage the Unit 2 RPV intermediate-to-lower shell girth weld PTS issue for the period of extended operation in accordance with 10 CFR 50.61 (b)(3).

The current PBNP flux reduction actions will not prevent the intermediate-to-lower shell girth weld from exceeding the PTS screening criteria at EOEL.

The PBNP Reactor Vessel Surveillance Program will include other options to manage Reactor Vessel Integrity per 10 CFR 50.61. These options will include consideration of an alternate fracture toughness evaluation methodology, pursuit of an aggressive flux reduction program in the future, pursuit of risk analysis, or pursuit of RPV annealing. Each of these options can provide technically acceptable methods of achieving EOEL with the PBNP RPVs.

The Reactor Vessel Surveillance Program will provide reasonable assurance that the Unit 2 RPV intermediate-to-lower shell girth weld PTS issue will be Page 25 of 40

adequately managed for the period of extended operation in accordance with 10 CFR 50.61, per 10 CFR 54.21(c)(1)(iii).

Use of the Master Curve methodology, extrapolated to EOEL fluence, shows that the RPV limiting weld metal meets PTS screening criteria out to EOEL and beyond. These projections will be confirmed by additional testing of weld heat 72442 from the B&W Owners Group MIRVP prior to reaching the EOL fluence at PBNP Unit 2. A supplemental surveillance program will be designed and implemented at PBNP Unit 2 that includes the limiting weld metal for future evaluation using the Master Curve methodology. The testing of this supplemental capsule at a fluence corresponding to EOEL will confirm the toughness condition for the PBNP Unit 2 RPV weld at about 38 EFPY, which is well before EOEL is reached.

The analysis associated with pressurized thermal shock has been projected to the end of the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).

Reactor Vessel Upper-Shelf Energy The requirements on reactor vessel Charpy upper-shelf energy are included in 10 CFR 50, Appendix G. Specifically, 10 CFR 50, Appendix G requires licensees to submit an analysis at least 3 years prior to the time that the upper-shelf energy of any reactor vessel material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

A fracture mechanics evaluation was performed in accordance with Appendix K of ASME Section Xl to demonstrate continued acceptable equivalent margins of safety against fracture through the end of life extension.

The analysis associated with upper-shelf energy has been projected to the end of the period of extended operation in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).

Reactor Vessel PressurefTemnerature Limits The requirements in 10 CFR 50, Appendix G, ensure that heatup and cooldown of the reactor pressure vessel are accomplished within established pressure-temperature limits. These limits specify the maximum allowable pressure as a function of reactor coolant temperature. As the reactor pressure vessel becomes embrittled and its fracture toughness is reduced, the allowable pressure is reduced.

Operation of the Reactor Coolant System is also limited by the net positive suction head curves for the reactor coolant pumps. These curves specify the minimum pressure required to operate the reactor coolant pumps. Therefore, Page 26 of 40

in order to heatup and cooldown, the reactor coolant temperature and pressure must be maintained within an operating window established between the Appendix G pressure-temperature limits and the reactor coolant pumps net positive suction head curves.

To address the period of extended operation, the end of license extension projected fluences, and the RPV material properties were used to determine the limiting materials, and calculate pressure-temperature limits for heatup and cooldown. The new Point Beach Unit I and 2 heatup and cooldown pressure-temperature limit curves were generated using adjusted reference temperature (ART) values that bound both units. The highest ART values from the two units were from the Unit 1 and Unit 2 intermediate-to-lower shell girth welds, however the limiting materials are actually the intermediate and lower shell axial welds from Unit 1, depending on the vessel thickness (1/4 T or 3/4 T location). The axial welds become limiting over the girth weld through use of "circ-flaw" methodology from ASME Code Case N-588. This methodology is less restrictive than the standard "axial-flaw' methodology from the 1995 ASME Code, Section Xl through the 1996 Addenda. In addition to the use of Code Case N-588, the PT curves also made use of ASME Code Case N-640, which allows the use of the K1c methodology. Both ASME Code Case N-588 and N-640 were joined together under ASME Code Case N-641.

The analysis associated with reactor vessel pressure-temperature limit curves has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

15.5 Exemptions The requirements of 10 CFR 54.21 (c) stipulate that the application for a renewed license should include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and that are based on time-limited aging analyses, as defined in 10 CFR 54.3. Each active 10 CFR 50.12 exemption has been reviewed to determine whether the exemption is based on a time-limited aging analysis. No existing TLAA related exemptions were identified.

Page 27 of 40

Appendix B Revisions The Reactor Vessel Surveillance Program as provided in Appendix B of the License Renewal Application is as follows.

B2.1.18 Reactor Vessel Surveillance Program Program Description The Reactor Vessel Surveillance Program manages the aging effect reduction of fracture toughness due to neutron embrittlement of the low alloy steel reactor vessels.

Monitoring methods will be in accordance with 10 CFR 50, Appendix H. This program includes (a) capsule insertion, withdrawal and materials testing/evaluation, (including upper shelf energy and RTNDT determinations), (b) fluence and uncertainty calculations, (c) monitoring of Effective Full Power Years (EFPY), (d) development of pressure-temperature limitations, (e) determination of low-temperature overpressure protection (LTOP) set points, and (f) implementation of a flux reduction program, and other options as necessary, allowed by 10 CFR 50.61 (b) for the Unit 2 intermediate-to-lower shell girth weld. The program ensures the reactor vessel materials (a) meet the fracture toughness requirements of 10 CFR 50, Appendix G, and (b) have adequate margins against brittle fracture caused by Pressurized Thermal Shock (PTS) in accordance with 10 CFR 50.61.

The Reactor Vessel Surveillance Program consists of six major subprograms:

• Surveillance Capsule Insertion, Withdrawal, and Evaluation,

• Fluence and Uncertainty Calculations,

• Monitoring of Effective Full Power Years (EFPY),

o Development of Pressure-Temperature Limit Curves, o Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP)

Setpoints, and o Implementation of a Flux Reduction Program and 10 CFR 50.61(b) Options for Unit 2 NUREG-1801 Consistency The Reactor Vessel Surveillance Program is an existing program that is consistent with, but includes exceptions to, NUREG-1801, "Generic Aging Lessons Leamed (GALL)

Report,"Section XI.M31, "Reactor Vessel Surveillance" (Reference 3). The Reactor Vessel Surveillance Program is also an existing program that consists of the appropriate ten elements described in Branch Technical Position RLSB-1, "Aging Management Review-Generic," which is included in Appendix A of NUREG-1800,

'Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

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Exceptions to NUREG-1801 See the following element discussion for elaboration on the exceptions to the NUREG-1 801 aging management program element assumptions:

• Acceptance Criteria Enhancements Enhancements to the Reactor Vessel Surveillance Program include changes to the FSAR and TRM to reflect the materials and withdrawal schedule of the new surveillance capsule and revisions to plant procedures to clarify organizational responsibilities, describe the plan/schedule for removal, testing and evaluation of surveillance capsules, and implement a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61(b) for the Unit 2 intermediate-to-lower shell girth weld.

These enhancements are required to satisfy the NUREG-1801 aging management program requirements. Details of the enhancements are included in the appropriate element descriptions below.

Implementation of a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 (b) for the Unit 2 RPV intermediate-to-lower shell girth weld will be completed within 1-year of receipt of the extended license. Implementation within this time frame will support submittal of any required safety analysis at least three years prior to the time that RTpTS for Unit 2 is projected to exceed the screening criteria. The other enhancements are scheduled for completion prior to the period of extended operation.

Aging Management Program Elements The key elements, which are used in the Reactor Vessel Surveillance Program, are described below. The results of an evaluation of each key element against NUREG-1801, 'Generic Aging Lessons Learned (GALL) Report,"Section XI.M31, "Reactor Vessel Surveillance," is provided below. An evaluation of each key element against the appropriate ten elements described in Branch Technical Position RLSB-1, "Aging Management Review-Generic," which is included in Appendix A of NUREG-1 800, 'Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," was also conducted.

Elements of the first three listed subprograms (i.e. Surveillance Capsule Insertion, Withdrawal and Evaluation, Fluence and Uncertainty Calculations, and Monitoring of Effective Full Power Years) are addressed by the NUREG-1 801 program. The subprograms for the Calculation and Monitoring of LTOP Setpoints, Development of Pressure-Temperature Limit Curves, and Implementation of a Flux Reduction Program and 10 CFR 50.61 (b) Options for Unit 2 are not addressed by the NUREG-1 801 program.

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Scope of Program The Reactor Vessel Surveillance Program consists of PBNP activities that manage the aging effects for components in the following systems and structures:

lReactor Vessel The Reactor Vessel Surveillance Program only applies to the PBNP-1 and PBNP-2 reactor pressure vessels.

Surveillance Capsule Insertion. Withdrawal, and Evaluation The program controls the development of surveillance capsule insertion and withdrawal schedules and capsule materials testing. Although the original surveillance capsules did not contain the most limiting material with respect to embrittlement, an additional surveillance capsule was installed in 2002 that contains the most limiting material. The surveillance program therefore meets the requirements of ASTM E 185-82.

The capsule installed in 2002 will be withdrawn during an outage at which it has accumulated a fluence equivalent to the 60-calendar year vessel fluence. Data from an integrated surveillance program that includes all PWRs with reactor vessels fabricated by B&W will also be used to predict embrittlement. Spare capsules remaining in both the PBNP-1 and PBNP-2 reactor vessels do not contain the most limiting materials and there are no current plans to withdraw these capsules.

The results of capsule materials testing, fluence analysis, and EFPY monitoring are used to predict the effects of neutron embrittlement through the end of extended life (EOEL). Prediction of the effects of radiation on reactor vessel beltline materials is in accordance with RG 1.99, Revision 2. Both the chemistry tables (RG 1.99, Revision 2, Position 1) and surveillance data (RG 1.99, Revision 2, Position 2) are used to project embrittlement. The limitations of RG 1.99, Revision 2, Position 1.3 are observed for material properties, temperature, material chemistry, and fluence.

The results of capsule tests, fluence analysis, and EFPY monitoring are also used to determine compliance with the PTS screening criteria of 10 CFR 50.61. A flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 (b) will be implemented if the value of RTpTs for any material in the beltline is projected to exceed the PTS screening criteria using the EOEL fluence.

Fluence and Uncertainty Calculations Calculations are performed for the PBNP-1 and PBNP-2 reactor vessels in accordance with RG 1.190, HCalculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." The results are used as an input to embrittlement predictions.

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Monitoring of Effective Full Power Years (EFPY)

EFPY monitoring is accomplished using operations data for the PBNP-1 and PBNP-2 reactors. The results are used to project the fluence corresponding to specific values of EFPY.

Development of Pressure-Temperature Limit Curves The Reactor Vessel Surveillance Program controls the development of pressure and temperature limit curves in accordance with 10 CFR 50, Appendix G requirements. The methods of ASME Section Xl, Appendix G are used to determine pressure and temperature limits. The fracture toughness used in calculating P-T limits is determined as a function of the difference in temperature from RTNDT. RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" is used to determine RTNDT.

ASME Code Case N-641 allows the use of the KIc curve, an alternate fracture toughness curve to the KIR curve, which is a modification to the acceptance criteria of ASME Section Xl, Appendix G.

Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP)

Setpoints The Reactor Vessel Surveillance Program requires the calculation of LTOP set points for the PBNP-1 and PBNP-2 reactor coolant systems. These set points ensure that an LTOP event will not increase the probability of brittle fracture of the reactor vessels.

LTOP set points include the maximum pressure allowed before the LTOP system actuates to relieve the pressure, and the temperature below which the LTOP system must be effective. These pressures and temperatures are determined using the method of ASME Section Xl, Appendix G or using an alternative method provided by ASME Code Case N-641.

Implementation of a Flux Reduction Program and 10 CFR 50.61 (b) Options for Unit 2 Because the RTpTs value of the Unit 2 RPV intermediate-to-lower shell girth weld is projected to exceed the PTS screening criteria prior to the EOEL, a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 (b) will be implemented on Unit 2. Ex-vessel neutron dosimetry sets, installed in the reactor cavity annulus, may be used to verify analytical flux reduction predications of the flux reduction program.

This element is consistent with the NUREG-1801 aging management program.

Preventive Actions This surveillance program determines neutron embrittlement for upper-shelf energy and pressure-temperature limits for 60 years in accordance with the RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

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Surveillance Capsule Insertion, Withdrawal, and Evaluation Surveillance Capsule Insertion, Withdrawal, and Evaluation do not constitute preventive actions.

Fluence and Uncertainty Calculations Fluence and uncertainty calculations do not constitute preventive actions.

Monitoring of Effective Full Power Years (EFPY)

EFPY monitoring is a monitoring activity, not a preventive action.

Development of Pressure-Temperature Limit Curves The development of, and operation within, P-T limit curves minimizes the probability of brittle fracture of the reactor vessel during normal operation.

Calculation and Monitorinq of Low Temperature Overpressure Protection (LTOP)

Setpoints The LTOP system with the actuation setpoints and operational restrictions established by the LTOP analysis, minimizes the probability of an LTOP event, and therefore, helps to minimize the probability of reactor vessel brittle fracture.

Implementation of a Flux Reduction Program and 10 CFR 50.61(b) Options for Unit 2 Because the RTpTs value of the Unit 2 RPV intermediate-to-lower shell girth weld is projected to exceed the PTS screening criteria prior to the EOEL, a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61(b) will be implemented. The flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 will ensure that the probability of brittle fracture of the reactor vessels during a PTS event is acceptably low.

This element is consistent with the NUREG-1 801 aging management program.

Parameters Monitored or Inspected Surveillance Capsule Insertion, Withdrawal, and Evaluation The program monitors the effects of neutron irradiation on the PBNP-1 and PBNP-2 reactor vessel beltline materials. Fracture toughness of beltline materials is indirectly monitored through measurement of the impact energy of Charpy V-Notch (CV) specimens, made from representative materials from the PBNP reactor vessels beltline regions. CV test results from capsules irradiated in other PWRs participating in an Page 32 of 40

integrated surveillance program are also used to aid in trending the change in material properties of the PBNP reactor vessels. Fracture toughness specimens to be irradiated in the PBNP-2 vessel and in the Master Integrated Reactor Vessel Surveillance Program (MIRVSP) will be withdrawn and tested. The surveillance capsules also contain neutron dosimetry that monitors the amount of neutron fluence received by the test specimens.

Fluence and Uncertainty Calculations This subprogram does not monitor, inspect, or test any parameters. Neutron fluence measurements acquired under the surveillance capsule insertion, withdrawal and testing subprogram are used to validate analytical models that determine the fluence received by the reactor vessel.

Monitoring of Effective Full Power Years (EFPY)

Effective Full Power Years (EFPY) are monitored and used to predict the fluence that the vessel will accumulate at some future time, which is then used to predict change in RTNDT and upper shelf energy (USE).

Development of Pressure-Temnerature Limit Curves No parameters are monitored or inspected under this subprogram.

Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP)

Setpoints LTOP system relief valve operation is monitored to determine whether an LTOP event could have occurred had the LTOP system been inoperable. Operation within the P-T limits is also monitored.

Implementation of a Flux Reduction Program and 10 CFR 50.61 (b) Options for Unit 2 Ex-vessel neutron dosimetry sets, installed in the reactor cavity annulus, may be used to verify analytical flux reduction predications of a flux reduction program.

This element is consistent with the NUREG-1801 aging management program.

Detection of Aging Effects Surveillance Capsule Insertion. Withdrawal, and Evaluation Aging effects are detected through testing of surveillance materials. CV tests are performed to determine the decrease in USE and increase in transition temperature RTNDT, for materials that closely match reactor vessel beltline materials.

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Fluence and Uncertainty Calculations This subprogram does not detect aging effects.

Monitoring of Effective Full Power Years (EFPY)

This subprogram does not detect aging effects.

Development of Pressure-Temperature Limit Curves This subprogram does not detect aging effects.

Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP)

Setpoints This subprogram does not detect aging effects.

Implementation of a Flux Reduction Program and 10 CFR 50.61(b) Options for Unit 2 This subprogram does not detect aging effects.

Enhancements will be made to revise the surveillance capsule withdrawal schedule contained in the PBNP FSAR and TRM to reflect the planned withdrawal of the new surveillance capsule that was installed in PBNP-2 during the 2002 refueling outage. A description of the materials included in this capsule, including fracture toughness specimens, must also be added to the PBNP FSAR. In addition, plant procedures will be modified as follows:

• Add a requirement that the reactor vessel engineer shall ensure that all withdrawn surveillance capsules not discarded as of August 31, 2000, are placed in storage for the purposes of future reconstitution and use, if necessary.

• Add a requirement that the reactor vessel engineer shall ensure that the number of EFPY accrued by PBNP-1 and PBNP-2 is updated by January 1 of each year.

• Add a requirement that the reactor vessel engineer shall ensure that the fluence and uncertainty calculations for PBNP-1 and PBNP-2 are updated periodically. The reactor vessel engineer should trend the rate of fluence accumulation versus EFPY.

Based on the updated projection of fluence versus EFPY, the reactor vessel engineer shall review the number of EFPY associated with the expiration of the current P-T limits to determine if this projected amount of EFPY remains valid.

• Add a requirement that a determination of the number of EFPY accumulated by January 1 of the current year shall be performed and documented annually.

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• Provide a description of the existing equivalent margins analyses for low USE that was performed for PBNP Unit-1 and PBNP Unit-2 with projected fluence values for EOEL (53 EFPY) assuming power uprate (1678 MWt) conditions without Hafnium power suppression assemblies installed.

• Specify that the methods of RG 1.99, Revision 2, are used to demonstrate compliance with the fracture toughness requirements of 10 CFR 50, Appendix G.

• Add a description of the methodology of fluence and uncertainty calculations.

• Describe the implementation of a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61 (b) for the Unit 2 intermediate-to-lower shell girth weld.

• Add a requirement to install neutron dosimetry if the last surveillance capsule in PBNP -2 is withdrawn prior to the 55th year of operation.

• Add a description of the plan/schedule for removal, testing and evaluation of surveillance capsules.

This element is consistent with the NUREG-1801 aging management program.

Monitoring and Trending Surveillance Capsule Insertion. Withdrawal, and Evaluation Monitoring of reactor vessel beltline fracture toughness is accomplished through testing of surveillance specimens from surveillance capsules that are periodically withdrawn from the vessels. Trending is accomplished through the RG 1.99, Revision 2 methods for projection of RTNDT and USE. Projection of the increase in RTNDT and the decrease in USE provides early indication if the fracture toughness properties of the PBNP reactor vessel beltline materials will fail to meet regulatory requirements. The RTpTs projection is compared to the PTS screening criteria of 270'F for plates, forgings, and axial welds, and 300*F for circumferential welds specified in 10 CFR 50.61. USE projections are compared against the requirement to maintain 50 ft-lbs or greater given by 10 CFR 50, Appendix G.

Fluence and Uncertainty Calculations Fluence measurements from capsules are trended to verify that actual fluence is adequately represented by fluence models and to project fluence for future dates. A surveillance capsule containing neutron dosimetry or some form of neutron dosimetry, will remain installed in the reactor vessels until at least the 55th year of operation.

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Monitoring of Effective Full Power Years (EFPY)

EFPY are monitored and trended to allow the EFPY for particular calendar dates, such as the end of the current and extended license periods, to be projected, and to establish deadlines for revising P-T curves that are valid only to a particular number of EFPY.

These projections will be extended to a number of EFPY corresponding to the end-of-extended-life (EOEL).

Development of Pressure-Temperature Limit Curves This subprogram does not perform trending, but relies on the trending of changes in material properties, fluence, and EFPY to set limits on P-T curve validity.

Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP)

Setpoints PBNP monitors and trends actuation of relief valves relied on for LTOP protection, to determine if the actuation is a reportable event. This subprogram relies on the trending of changes in material properties, fluence, and EFPY, to determine the inputs to calculations of LTOP set points.

Implementation of a Flux Reduction Program and 10 CFR 50.61(b) Options for Unit 2 Ex-vessel neutron dosimetry sets, installed in the reactor cavity annulus, may be used to verify analytical flux reduction predications of a flux reduction program.

This element is consistent with the NUREG-1801 aging management program.

Acceptance Criteria Surveillance Capsule Insertion, Withdrawal, and Evaluation The upper shelf energy of the most limiting material in the reactor vessel beltline must remain above 50 ft-lbs until the end-of-extended-life, using the methods of RG 1.99, Revision 2 with the PBNP specific and integrated surveillance program data as inputs or equivalent margin demonstrated. The RTpTs of the most limiting material in the reactor vessel beltline must not exceed the PTS screening criteria specified by 10 CFR 50.61 (2700F for plates, forgings, and axial welds, and 300*F for circumferential welds), unless it can be demonstrated by alternate means, as allowed by 10 CFR 50.61, that the probability of brittle fracture of the reactor vessel in a PTS event is acceptably low.

Fluence and Uncertainty Calculations These calculations do not have specific acceptance criteria.

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Monitoring of Effective Full Power Years (EFPY)

This activity does not have specific acceptance criteria. EFPY monitoring does affect the validity of the pressure-temperature limit curves, which are linked to a specific range of EFPY.

Development of Pressure-Temperature Limit Curves The acceptance criteria for P-T curves is that the flaw stability criteria of ASME Section Xl, Appendix G, are met for all normal operating conditions as required by 10 CFR 50, Appendix G. The acceptance criteria of Appendix G may be modified through application of ASME Code Case N-641, which allows the use of the Kic curve, an alternate fracture toughness curve to the KIR curve. Pressure-temperature curves are acceptable only through a specific value of EFPY that is based on a fluence projection for that number of EFPY.

Calculation and Monitoring of Low Temperature Overpressure Protection (LTOP) Setpoints LTOP set points are acceptable only through a specific value of EFPY that is based on a fluence projection for that number of EFPY.

Implementation of a Flux Reduction Program and 10 CFR 50.61 (b) Options for Unit 2 If no reasonably practical flux reduction program can be shown to prevent RTpTs from exceeding the PTS screening criteria prior to EOEL, other options allowed by 10 CFR 50.61(b) will be evaluated and implemented.

This element includes exceptions to the NUREG-1801 aging management program.

NUREG-1 801 does not provide for use of ASME Code Case N-641 fracture toughness curves when calculating P-T limit curves.

PBNP meets the intent of this NUREG-1801 aging management program.

Corrective Actions Corrective actions are implemented in accordance with the requirements of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and ANSI N18.7-1976, "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," as committed in Section 1.4 of the PBNP Final Safety Analysis Report (FSAR).

This element is consistent with the NUREG-1801 aging management program.

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Confirmation Process The confirmation process is part of the corrective action program, which is implemented in accordance with the requirements of 10 CFR 50, Appendix B and ANSI N18.7-1976, as committed in Section 1.4 of the PBNP FSAR.

This element is consistent with the NUREG-1801 aging management program.

Administrative Controls The Reactor Vessel Surveillance Program is implemented through various plant documents. These implementing documents are subject to administrative controls, including a formal review and approval process, in accordance with the requirements of 10 CFR 50, Appendix B and ANSI N18.7-1976, as committed in Section 1.4 of the PBNP FSAR.

This element is consistent with the NUREG-1801 aging management program.

Operating Experience PBNP-1 and PBNP-2 have generally operated successfully within their licensed P-T limits. New P-T curves are developed and issued, as required. An event involving the actuation of the LTOP system relief valves at PBNP-1 occurred on October 23, 1997.

The event was evaluated and the conclusion was that an over pressurization event would have occurred if the LTOP system had been inoperable. A report to the NRC was therefore required. However, the LTOP system functioned correctly, preventing the over pressurization. The calculation also took no credit for manual operator action that may have prevented the over pressurization.

PBNP-1 will continue to meet the requirements of 10 CFR 50, Appendix G and 10 CFR 50.61 through the end of extended life. RTpTS for the intermediate-to-lower shell girth weld in the PBNP-2 vessel is predicted to exceed the PTS screening criteria prior to the end of license extension and will be addressed through a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61(b) as part of the Reactor Vessel Surveillance Program.

The program has been modified to incorporate data from the B&W integrated surveillance program. A replacement surveillance capsule containing materials closely matching the limiting materials for both PBNP-1 and PBNP-2 has been installed in the PBNP-2 reactor vessel during the 2002 refueling outage. The selection of materials for this capsule reflects the evolution in the understanding of the variables that control embrittlement of reactor pressure vessel steels, which resulted in a reassessment of the identity of the limiting materials in the PBNP-1 and PBNP-2 vessels.

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Industry operating experience related to the Reactor Vessel Surveillance Program includes GL 92-01, Revision 1, "Reactor Vessel Structural Integrity," and Supplement 1 to GL 92-01, Revision 1, "Reactor Vessel Structural Integrity." PBNP's response to these documents has been incorporated into the Reactor Vessel Surveillance Program.

A review of NRC Inspection Reports, QA Audit/Surveillance Reports, and Self-Assessments since 1999 revealed no issues or findings that could impact the effectiveness of the Reactor Vessel Surveillance Program. The Second Quarter 2000 Engineering Audit assessed the reactor vessel integrity program, as controlled by plant procedures. The audit examined several activities including a calculation of the date at which the neutron fluence would exceed the limits of the current P-T curves, the progress of a submittal to the NRC of revised P-T curves, and a calculation of the LTOP applicability date. These activities were found to have been completed satisfactorily.

The auditors found that corrective actions related to previous Condition Reports had been completed and there were no new Condition Reports. The auditors therefore judged the program to be effective. As additional operating experience is obtained, lessons learned may be used to adjust this program.

This element is consistent with the NUREG-1 801 aging management program.

Conclusion The Reactor Vessel Surveillance Program provides reasonable assurance that the aging effects will be managed consistent with the current licensing basis for the period of extended operation. The Reactor Vessel Surveillance Program complies with the requirements of 10 CFR 50.60, 10 CFR 50.61, and 10 CFR 50, Appendices G and H.

The combination of the PBNP original surveillance program and the B&W Integrated Surveillance Program has been used to demonstrate that the reference temperature for the PBNP-1 reactor vessel limiting beltline materials will not exceed the PTS screening criteria of 10 CFR 50.61 prior to the end-of-extended life. Because the PBNP-2 reactor vessel intermediate-to-lower shell girth weld reference temperature is predicted to exceed the PTS screening criteria prior to the end of license extension, a flux reduction program and other options, as necessary, allowed by 10 CFR 50.61(b) will be included for Unit 2 as part of the Reactor Vessel Surveillance Program.

The upper shelf energy (USE) for the limiting materials of the PBNP-1 and PBNP-2 reactor vessels is projected to fall below 50 ft-lbs by the end of the current license.

PBNP has performed analyses that demonstrate equivalent margins against ductile fracture to those required by 10 CFR 50, Appendix G, through the end-of-extended-life.

To further refine the predictions of the material properties at the end-of-extended life (corresponding to 60 calendar years), an additional surveillance capsule containing materials that closely match the limiting materials in the reactor vessel beltline of both PBNP-1 and PBNP-2, has been installed in PBNP-2. This capsule will be withdrawn after it has received a fluence equivalent to the vessel fluence at 60 calendar years.

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References Reference 75: Framatome ANP, Inc., AREVA and Siemens Company Calculation, BAW-2467P, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units 1 and 2 for Extended Life through 53 Effective Full Power Years, July 2004 Page 40 of 40