NRC 2007-0049, Supplement to License Amendment Request 251; Technical Specification 5.6.5., Reactor Coolant System Pressure and Temperature Limits Report

Supplement to License Amendment Request 251; Technical Specification 5.6.5., Reactor Coolant System Pressure and Temperature Limits Report

ML071650095
Person / Time
Site:	Point Beach
Issue date:	06/13/2007
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LTR-REA-04-64, NRC 2007-0049
Download: ML071650095 (127)

Text

NMC>

Cornrnilted to Nuclear Exoellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC June 13,2007 NRC 2007-0049 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Supplement to License Amendment Reauest 251 ;

Technical Specification 5.6.5, Reactor Coolant System Pressure and Temperature Limits Report

References:

(1 )

NMC to NRC Letter Dated December 14,2006 (NRC 2006-0090),

License Amendment Request 251 (ADAMS Accession No. ML063470599)

Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC), submitted a proposed amendment to the Technical Specifications (TS) for Point Beach Nuclear Plant (PBNP),

Units 1 and 2 on December 14, 2006. The proposed amendment would revise TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)". The revision would add the FERRET Code as an approved methodology for determining reactor coolant system (RCS) pressure and temperature limits. The NRC approved the existing PTLR methodology for PBNP on July 23, 2001.

The FERRET Code was evaluated in Westinghouse Report WCAP-16083-NP, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry." WCAP-16083 was approved by the NRC for referencing in plant-specific license amendments via NRC Safety Evaluation dated January 10, 2006.

In response to a request for additional information raised by the NRC staff during review of this application, the following documents are enclosed in this supplement to enable continued staff review of the proposed amendment:

1. Enclosure 1 - Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations," Point Beach Units 1 and 2, S. L. Anderson, June 2004, Non-Proprietary Class 3.

2. Enclosures 2a & 2b - Operability Recommendation OPR000175, Revision 0, approved February 15, 2006, and Revision 1, approved June 11, 2007.

3. Enclosure 3 - TRM 2.2, Revision 1, marked up to show the proposed addition of References 5.1 3 and 5.14 for use of the FERRET methodology for determining RCS pressure and temperature limits.

6610 Nuclear Road Two Rivers, Wisconsin 54241-9516 Telephone: 920.755.2321

Document Control Desk Page 2 As stated in OPR000175 Revision 1, it has been concluded that the pressureltemperature limit curves and the low temperature overpressure protection (LTOP) limit settings for PBNP Units 1 and 2 remain the same. The current expiration date of these curves is June I, 2008, for Units 1 and 2. The expiration date of the curves is based on the limiting components in the Unit 1 reactor vessel.

PBNP Units I and 2 did not exceed operational limits specified in TRM 2.2 during the period since February 2004. Actual vessel fluence values are within the limits assumed in the calculation of the PTLR limits. Using the recently approved fluence calculation methodology (FERRET Code) and the most recent fluence data contained in Enclosure 1, the limiting Unit 1 vessel values will remain within limits until 29.5 EFPY.

Submittal of this additional information does not alter the technical analysis, regulatory analysis or conclusions supporting the proposed change as described in Reference (1). Accordingly, the No Significant Hazards Consideration and Environmental Considerations are not affected.

Revision of TRM 2.2, "Pressure Temperature Limits Report," is in progress. TRM 2.2 Revision 2 will undergo required internal reviews and approvals and be implemented within 15 days of NRC approval of this license amendment request and issuance of the associated safety evaluation.

This letter contains no new commitments or revisions to existing commitments.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Wisconsin Official.

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

Executed*

13,2007.

/ Dennis L. Koehl 1

Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosures cc:

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

ENCLOSURE 1 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 251 POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 WESTINGHOUSE REPORT LTR-REA-04-64 PRESSURE VESSEL NEUTRON EXPOSURE EVALUATIONS 92 Pages Follow

Westinghouse Non-Proprietary Class 3 LTR-REA-04-64 Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2 S. L.. Anderson Radiation Engineering and Analysis June 2004 WESTINGHOUSE ELECTRIC COMPANY LLC P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

1.0

Background

In the assessment of the state of embrittlement of light water reactor (LWR) pressure vessels, an accurate evaluation of the neutron exposure of each of the materials comprising the beltline region of the vessel is required. In Appendix G to 10 CFR SO['],

the beltline region is defined as "the region of the reactor vessel shell material (including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the reactor core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage". In Appendix H to 10 CFR SO['],

the threshold fast neutron fluence (E > 1.0 MeV) to be used to determine if a specific material is to be included in the beltline region is further defined as 1.OE+17 n/cm2.

Therefore, in order to encompass all areas of the reactor vessel anticipated to accrue a a fast neutron exposure greater than 1.OE+17 n/cm2, plant specific exposure assessments must include evaluations not only at locations of maximum exposure at the inner diameter of the vessel, but, also as a function of axial, azimuthal, and radial location throughout the vessel wall. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron ~luence"~~]

describes state-of-the-art calculation and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence applicable to these beltline locations.

This report describes neutron fluence assessments performed for the Point Beach Units 1 and 2 pressure vessel beltline regions based on the guidance specified in Regulatory Guide 1.190. In these assessments, fast neutron exposures expressed in terms of fast neutron fluence (E > 1.0 MeV) were established for each of the materials comprising the beltline region of the respective pressure vessels.

Exposures of the materials were determined on a plant and fuel cycle specific basis for the first 29 operating cycles at Point Beach Unit 1 and the first 27 operating cycles at Point Beach Unit 2. At the conclusion of fuel cycle 29 (on or about October 2005) Point Beach Unit 1 is anticipated to have accrued approximately 27.0 effective full power years (efpy) of operation. Likewise at the conclusion of fuel cycle 27 (on or about April 2005) Point Beach Unit 2 is expected to have accrued approximately 26.3 efpy.

Following completion of the plant specific exposure assessments encompassing the fuel cycles designed to date, projections of the future neutron exposure of the pressure vessel beltline materials extending to 54 efpy of operation were performed for both units.

In performing the fluence projections for future operation, several scenarios were assumed in order to provide a matrix of material exposures based on a series of fuel management and power uprate options.

The fuel management strategies employed during the operating lifetime of both Point Beach Unit 1 and Unit 2 can be conveniently subdivided into three general categories as follows:

Additionally, a mini uprate in operating core power from 1518.5 MWt to a new core power of 1540.0 MWt was accomplished at both units on February 3, 2003. This uprate took place during cycle 28 at Unit 1 and cycle 26 at unit 2. A further uprate to an operating power level of 1678.0 MWt is planned to occur at Unit 1 in October 2008 at the beginning of cycle 32 and at Unit 2 in April 2008 at the onset of cycle 30. The following fluence projection scenarios were considered based on these various fuel management strategies and and planned power uprates:

Unit 2 Cycles 1-5 Cycles 6-1 5 Cycles 16-27 Fuel Management Approach Out-In fuel management strategy with fresh fuel assemblies located on the core periphery.

Conventional low leakage fuel management strategy with burned fuel located on the core periphery.

Conventional low leakage fuel management strategy supplemented with the inclusion of part length hafnium power suppression rods located in the twelve peripheral fuel assemblies spanning the core cardinal axes.

Unit 1 Cycles 1-7 Cycles 8-1 6 Cycles 17-29 Case 1 - Unit 1 2 - Unit 1 3 - Unit 1 4 - Unit 1 5 - Unit 2 6 -Unit 2 7 - Unit 2 8 - Unit 2 Fuel Management Approach Low Leakage With Hafnium Rods Low Leakage With Hafnium Rods Low Leakage Without Hafnium Rods Low Leakage Without Hafnium Rods Low Leakage With Hafnium Rods Low Leakage With Hafnium Rods Low Leakage Without Hafnium Rods Low Leakage Without Hafnium Rods Core Power (MWt) 151 8.5 - startup to 02/03/2003 1540.0 - 02/03/2003 to 10/2008 1 678.0 - 1 0/2008 to 54 efpy 151 8.5 - startup to 02/03/2003 1540.0 - 02/03/2003 to 54efpy Same as Case 1 with removal of Hf rods in October 2008 Same as Case 2 with removal of Hf rods in October 2008 151 8.5 - startup to 02/03/2003 1540.0 - 02/03/2003 to 0412008 1 678.0 - 0412008 to 54 efpy 151 8.5 - startup to 02/03/2003 1540.0 - 02/03/2003 to 54 efpy Same as Case 5 with removal of Hf rods in April 2008 Same as Case 6 with removal of Hf rods in April 2008

In the above tabulation, there are four operating scenarios defined for each unit. These include operation with and without the presence of Hf power suppression rods and with and without the extended power uprate to 1678.0 MWt.

Based on these operating scenarios, neutron exposure projections applicable to each of the beltline materials were developed for both Point Beach Units 1 and 2. In performing the fluence projections for each of these scenarios the lengths of fuel cycles 30 and 31 at Point Beach Unit 1 were assumed to be the same as for cycle 29. Likewise, for Point Beach Unit 2, the lengths of fuel cycles 28 and 29 were assumed to be the same as cycle 27. These two fuel cycles at each Unit represent a period of operation at 1540.0 MWt following completion of the last design fuel cycle and the projected power uprates to 1678 MWt. Results of these plant specific calculations are provided in Section 2.0 of this report.

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 1.1 90. The results of these dosimetry re-evaluations were then used to validate the calculational models that were applied in the plant specific neutron transport analyses of the Point Beach Units. The dosimetry evaluation methodology and comparisons of plant specific calculations with measurement results are provided in Section 3.0 of this report.

The combination of the in-vessel and ex-vessel calculation to measurement comparisons represent a substantial data base for each unit that can be used to identify biases in the calculated results. In Section 4.0 of this report, these data bases are used to adjust the calculated fluence at the pressure vessel wall to produce exposure estimates with reduced uncertainties.

2.0 Neutron Transport Calculations 2.1 Method of Analysis In performing the fast neutron exposure evaluations for the Point Beach Units 1 and 2 reactors, plant specific forward transport calculations were carried out using the three-dimensional flux synthesis technique described in Section 1.3.4 of Regulatory Guide 1.190. For the fuel cycles using either out-in loading patterns or traditional low leakage loading patterns, the following single channel synthesis equation was employed:

where @(r,8,z) is the synthesized three-dimensional neutron flux distribution, @(r,0) is the transport solution in r,0 geometry, @(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and @(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,8 two-dimensional calculation.

For the fuel cycles employing part length hafnium power suppression rods in peripheral fuel assemblies, the axial power distributions in those assemblies differ significantly from the axial power distribution characteristic of the remainder of the reactor core. Therefore, for these fuel cycles the following multi-channel synthesis equation was used in the transport calculations:

where the subscript " A refers to the region of the reactor core that does not include hafnium power suppression rods and the subscript " B refers to the region of the core comprised of the peripheral assemblies located on the cardinal axes that contain the part length absorber rods.

For the Point Beach Units 1 and 2 analyses, all of the transport calculations were carried out using the DORT two-dimensional discrete ordinates code Version 3. 1 ~ ~ ~

and the BUGLE-96 cross-section libraryf4]. The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor application. In these analyses, anisotropic scattering was treated with a P, legendre expansion and the angular discretization was modeled with an Sls order of angular quadrature.

This calculational methodology has been submitted to the NRC and approved with no restrictions per the Final Safety Evaluation provided in Reference 59.

A plan view of the r,0 model of the Point Beach Units 1 and 2 reactor geometry at the core midplane is shown in Figure 2.1-1. Since the reactor exhibits octant symmetry only a 0' to 45' sector is depicted. In addition to the core, reactor internals, pressure vessel and primary biological shield, the model also included explicit representations of the surveillance capsules, the pressure vessel cladding, and the insulation located external to the pressure vessel.

From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structure in the analytical model is significant. Since the presence of the capsules and structure has a marked impact on the magnitude of the neutron flux as well as on the relative neutron and gamma ray spectra at dosimetry locations within the capsules, a meaningful evaluation of the radiation environment internal to the capsules can be made only when these perturbation effects are properly accounted for in the analysis.

In contrast to the relatively massive stainless steel and carbon steel structures associated with the internal surveillance capsules, the thin walled aluminum capsules used for the measurements in the reactor cavity were designed to minimize perturbations in the neutron flux and, thus, to provide free field data at the measurement locations. Therefore, explicit descriptions of these small capsules in the transport models was not required.

In developing the r,0 analytical model of the reactor geometry shown in Figure 2.1-1, nominal design dimensions were employed for the various structural components.

Likewise, water temperatures and, hence, coolant density in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc. The r,0 geometric mesh description of the reactor model shown in Figure 2.1-1 consisted of 148 radial by 105 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

A section view of the r,z model of the Point Beach Units 1 and 2 reactors is shown in Figure 2.1-2. The model extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation 5 inches below the active fuel to 17 inches above the active fuel. The axial extent of the model was chosen to permit the determination of the maximum exposure of vessel materials expected to experience a fast neutron fluence greater than 1.0e+17 n/cm2 (E > 1.0 MeV) over the service life of the respective units.

As in the case of the r,0 model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone.

The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of the reactor model shown in Figure 2.1-2 consisted of 127 radial by 155 axial intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was

achieved on a pointwise basis. The convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 127 radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a mesh-wise basis throughout the entire geometry.

The core power distributions used in the plant specific transport analysis for the Point Beach Units 1 and 2 reactors were taken from the appropriate fuel cycle design reports for the respective units (References 5 through 31, 60, and 61 for Unit 1 and 32 through 56, 62, and 63 for Unit 2). The data extracted from the design reports included fuel assembly specific initial enrichment, beginning of cycle burnup and end of cycle burnup.

Appropriate axial distributions were also extracted from the respective design reports.

In constructing the core source distributions from the fuel assembly specific enrichment and burnup data, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes; and from that fission split, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

Figure 2.1-1 Point beach Units 1 and 2 r,8 Reactor Geometry

Figure 2.1 -2 Point beach Units 1 and 2 r,z Reactor Geometry

2.2 Calculated Neutron Exposure of Point Beach Unit 1 Beltline Materials A schematic of the beltline region of the Point Beach Unit 1 reactor pressure vessel is shown in Figure 2.2-1. In the case of the Unit 1 pressure vessel, the beltline region is constructed of two (2) shell plates, two (2) longitudinal welds (one in the intermediate shell plate and one in the lower shell plate), two (2) circumferential welds (one joining the intermediate and lower shell plates and one joining the intermediate shell plate to the upper shell forging), and the upper shell forging itself. Each of these seven materials must be considered in the overall embrittlement assessments of the pressure vessel.

Pressure vessel materials not included in this compilation are not anticipated to experience a fast neutron (E > 1.0 MeV) exposure in excess of 1.OE+17 n/cm2 over the service life of the unit and, therefore, are not included in the definition of the beltline region.

The plant and fuel cycle specific calculated fast neutron (E > 1.0 MeV) fluence experienced by the materials comprising the beltline region of the Point Beach Unit 1 reactor are presented in Tables 2.2-1 through 2.2-6 for plant operation through the conclusion of the 29th fuel cycle. The data listed in Tables 2.2-1 through 2.2-6 provide a summary of the calculated fast neutron fluence (E > 1.0 MeV) for all Point Beach Unit 1 fuel cycles designed to date.

In Tables 2.2-1 through 2.2-6, cycle specific maximum neutron exposures at the pressure vessel cladlbase metal interface are given at azimuthal angles of 0°, IS0, 30°,

and 45' relative to the core cardinal axes. The data provided in Tables 2.2-1 through 2.2-6 were taken at the axial location of the maximum exposure experienced by each material based on the results of the three-dimensional synthesized neutron fluence evaluations.

As noted in Section 1.0 of this report, three general fuel management strategies have been employed during the course of the first 29 fuel cycles of operation at Point Beach Unit 1. These three approaches are summarized as follows:

Cycles 1 -7 8-1 6 17-29 Fuel Management Approach Out-In fuel management strategy with fresh fuel assemblies located on the core periphery.

Conventional low leakage fuel management strategy with burned fuel located on the core periphery.

Conventional low leakage fuel management strategy supplemented with the inclusion of part length hafnium power suppression rods located in the twelve peripheral fuel assemblies spanning the core cardinal axes.

For Point Beach Unit 1, the hafnium power suppression rods were 72 inches in length and were positioned in the bottom half of the active core region. Thus, the maximum flux reduction impact of the power suppression rods was experienced by the lower shell plate in the 0"-20" azimuthal span as well as by the lower shell longitudinal weld located on the 15" azimuth. Other materials comprising the beltline region of the Point Beach Unit 1 pressure vessel were impacted to a much lesser degree by the presence of the part length absorber rods.

Projections of calculated fast neutron fluence to be accrued beyond the end of cycle 29 were based on the four potential scenarios described in Section 1.0 of this report. The assumed core power level and core power distributions used in these projections are summarized as follows:

For operation with hafnium power suppression rods, the relative core power distribution averaged over the thirteen fuel cycles that have contained the suppression rods was assumed to be representative of future fuel cycle designs. For operation without the hafnium suppression rods, a relative power distribution averaged over the nine fuel cycles employing conventional low leakage (cycles 8-16) was assumed to be representative of future operation. In all cases, the transport calculations were normalized to the operating core power levels as noted in the preceding tabulation.

Case 1

2 3

4 Results of the fast neutron (E > 1.0 MeV) fluence projections for the Point Beach Unit 1 beltline materials are provided in Tables 2.2-7 through 2.2-18. In Tables 2.2-7 through 2.2-12, data are provided both with and without the presence of hafnium power suppression rods assuming continued operation at a power level of 1540.0 MWt until October 2008 followed by operation at the extended uprate power of 1678.0 MWt. In Tables 2.2-13 through 2.2-18, data are given both with and without the presence of the hafnium absorbers based on continued plant operation at 1540.0 MWt through 54 effective full power years.

Power Distribution Avg. of cycles 17-29 Avg. of cycles 17-29 Avg. of cycles 17-29 Avg. of cycles 17-29 Avg. of cycles 17-29 Avg. of cycles 6-1 8 Avg. of cycles 17-29 Avg. of cycles 6-18 Core Power (MWt) 1540.0 1678.0 1540.0 1540.0 1540.0 1678.0 1540.0 1540.0 Applicable Cycles Cycles 30-31 Cycles 32 - 54 efpy Cycles 30-31 Cycles 32 - 54 efpy Cycles 30-31 Cycles 32 - 54 efpy Cycles 30-31 Cycles 32 - 54 efpy

Figure 2.2-1 Locations of Pressure Vessel Beltline Materials Point Beach Unit 1

Table 2.2-1 Intermediate Shell to Lower Shell Circumferential Weld (SA-1101)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1 Fuel Cycle 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Cumulative Time

[ef~sl 4.69E+07 7.58E+07 1.1 4E+08 1.36E+08 1.61 E+08 1.87E+08 2.1 4E+08 2.34E+08 2.53E+08 2.74E+08 2.93E+08 3.23E+08 3.48E+08 3.75E+08 Cycle Time

[efps]

4.69E+07 2.89E+07 3.82E+07 2.20E+07 2.48E+07 2.57E+07 2.75E+07 2.03E+07 1.88E+07 2.04E+07 1.94E+07 3.04E+07 2.49E+07 2.66E+07 Neutron Fluence [n/cm2]

45 Deg.

6.98E+17 1.1 0E+18 1.65E+18 2.01 E+18 2.42E+18 2.84E+18 3.24E+18 3.53E+18 3.78E+18 4.06E+18 4.25E+18 4.61 E+18 4.91 E+18 5.23E+18 30 Deg.

8.07E+17 1.27E+18 1.92E+18 2.34E+18 2.81 E+18 3.29E+18 3.77E+18 4.08E+18 4.34E+18 4.63E+18 4.86E+18 5.28E+18 5.65E+18 6.03E+18 0 Deg.

2.00E+18 3.1 1 E+18 4.64E+18 5.62E+18 6.68E+18 7.71 E+18 8.83E+18 9.52E+18 1.O1 E+19 1.06E+19 1.1 2E+19 1.20E+19 1.27E+19 1.34E+19 15 deg.

1.20E+18 1.88E+18 2.82E+18 3.41 E+18 4.07E+18 4.71 E+18 5.39E+18 5.80E+18 6.1 5E+18 6.51 E+18 6.84E+18 7.41 E+18 7.89E+18 8.39E+18

Table 2.2-2 Lower Shell Plate ((2-1423)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1 Fuel Cycle 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Cumulative Time

[efps]

4.69E+07 7.58E+07 1.14E+08 1.36E+08 1.61 E+08 1.87E+08 2.14E+08 2.34E+08 2.53E+08 2.74E+08 2.93E+08 3.23E+08 3.48E+08 3.75E+08 Cycle Time

[efps]

4.69E+07 2.89E+07 3.82E+07 2.20E+07 2.48E+07 2.57E+07 2.75E+07 2.03E+07 1.88E+07 2.04E+07 1.94E+07 3.04E+07 2.49E+07 2.66E+07 Neutron Fluence [n/cm2]

45 Deg.

7.23E+17 1.13E+18 1.70E+18 2.07E+18 2.50E+18 2.94E+18 3.36E+18 3.66E+18 3.93E+18 4.22E+18 4.42E+18 4.80E+18 5.1 1 E+18 5.44E+18 0 Deg.

2.07E+18 3.21 E+18 4.78E+18 5.80E+18 6.90E+18 7.97E+18 9.14E+18 9.86E+18 1.05E+19 1.l OE+19 1.16E+19 1.25E+19 1.32E+19 1.39E+19 15 deg.

1.24E+18 1.94E+18 2.90E+18 3.52E+18 4.21 E+18 4.87E+18 5.58E+18 6.01 E+18 6.38E+18 6.76E+18 7.1 1 E+18 7.70E+18 8.20E+18 8.72E+18 30 Deg.

8.36E+17 1.32E+18 1.98E+18 2.42E+18 2.91 E+18 3.41 E+18 3.90E+18 4.22E+18 4.51 E+18 4.81 E+18 5.05E+18 5.50E+18 5.87E+18 6.27E+18

Table 2.2-3 Intermediate Shell Plate (A-981 1)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1

Table 2.2-4 lntermediate Shell (SA-7751812) and Lower Shell (SA-847) Longitudinal Welds Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1 Fuel Cycle 1

2 3

4 5

6 7

8 Neutron Fluence [n/cm2]

Cycle Time

[ef PSI 4.69E+07 2.89E+07 3.82E+07 2.20E+07 2.48E+07 2.57E+07 2.75E+07 2.03E+07 Intermediate Shell 1.21E+18 1.93E+18 2.82E+18 3.42E+18 4.07E+18 4.72E+18 5.41 E+18 5.83E+18 Cumulative Time

[efps]

4.69E+07 7.58E+07 1.14E+08 1.36E+08 1.61 E+08 1.87E+08 2.14E+08 2.34E+08 Lower Shell 1.24E+18 1.94E+18 2.90E+18 3.52E+18 4.21 E+18 4.87E+18 5.58E+18 6.01 E+18

Table 2.2-5 Intermediate Shell to Upper Shell Circumferential Weld (SA-1426)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1 Cumulative Time ref PSI 4.69E+07 7.58E+07 1.14E+08 1.36E+08 1.61 E+08 1.87E+08 2.1 4E+08 2.34E+08 2.53E+08 2.74E+08 2.93E+08 3.23E+08 3.48E+08 3.75E+08 Fuel Cycle 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Neutron Fluence [n/cm2]

Cycle Time

[ef~sl 4.69E+07 2.89E+07 3.82E+07 2.20E+07 2.48E+07 2.57E+07 2.75E+07 2.03E+07 1.88E+07 2.04E+07 1.94E+07 3.04E+07 2.49E+07 2.66E+07 45 Deg.

4.03E+16 8.72E+16 1.23E+17 1.47E+17 1.80E+17 2.10E+17 2.41 E+17 2.64E+17 2.89E+17 3.09E+17 3.26E+17 3.53E+17 3.80E+17 4.07E+17 0 Deg.

1.1 5E+17 2.46E+17 3.45E+17 4.09E+17 4.95E+17 5.71 E+17 6.56E+17 7.1 2E+17 7.66E+17 8.06E+17 8.55E+17 9.1 5E+17 9.77E+17 1.04E+18 15 deg.

6.93E+16 1.49E+17 2.10E+17 2.49E+17 3.02E+17 3.49E+17 4.01 E+17 4.35E+17

. 4.68E+17 4.95E+17 5.24E+17 5.65E+17 6.08E+17 6.51 E+17 30 Deg.

4.66E+16 1.O1 E+17 1.44E+17 1.71 E+17 2.09E+17 2.44E+17 2.80E+17 3.06E+17 3.31 E+17 3.53E+17 3.73E+17 4.04E+17 4.36E+17 4.70E+17

Table 2.2-6 Upper Shell Forging Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 1 Cumulative Time

[ef~sl 4.69E+07 7.58E+07 1.1 4E+08 1.36E+08 1.61 E+08 1.87E+08 2.14E+08 2.34E+08 2.53E+08 2.74E+08 2.93E+08 3.23E+08 3.48E+08 3.75E+08 Fuel Cycle 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Neutron Fluence [n/cm2]

Cycle Time

[ef~sl 4.69E+07 2.89E+07 3.82E+07 2.20E+07 2.48E+07 2.57E+07 2.75E+07 2.03E+07 1.88E+07 2.04E+07 1.94E+07 3.04E+07 2.49E+07 2.66E+07 45 Deg.

4.03E+16 8.72E+16 1.23E+17 1.47E+17 1.80E+17 2.10E+17 2.41 E+17 2.64E+l7 2.89E+17 3.09E+17 3.26E+17 3.53E+17 3.80E+17 4.07E+17 30 Deg.

4.66E+16 1.O1 E+17 1.44E+17 1.71 E+17 2.09E+17 2.44E+17 2.80E+17 3.06E+17 3.31 E+17 3.53E+17 3.73E+17 4.04E+17 4.36E+17 4.70E+17 0 Deg.

1.1 5E+17 2.46E+17 3.45E+17 4.09E+17 4.95E+17 5.71 E+17 6.56E+17 7.1 2E+17 7.66E+17 8.06E+17 8.55E+17 9.1 5E+17 9.77E+17 1.04E+18 15 deg.

6.93E+16 1.49E+17 2.1 0E+17 2.49E+17 3.02E+17 3.49E+17 4.01 E+17 4.35E+17 4.68E+17 4.95E+17 5.24E+17 5.65E+17 6.08E+17 6.51 E+17

Table 2.2-7 Intermediate Shell to Lower Shell Circumferential Weld (SA-1101)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-8 Lower Shell Plate (C-1423)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-9 Intermediate Shell Plate (A-981 1)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-1 0 Intermediate Shell (SA7751812) and Lower Shell (SA-847) Longitudinal Welds Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-1 1 Intermediate Shell to Upper Shell Circumferential Weld (SA-1426)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-1 2 Upper Shell Forging Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 2.2-1 3 Intermediate Shell to Lower Shell Circumferential Weld (SA-1101)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.2-1 4 Lower Shell Plate (C-1423)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.2-1 5 Intermediate Shell Plate (A-981 1)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.2-1 6 Intermediate Shell (SA7751812) and Lower Shell (SA-847) Longitudinal Welds Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.2-1 7 Intermediate Shell to Upper Shell Circumferential Weld (SA-1426)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.2-1 8 Upper Shell Forging Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

2.3 Calculated Neutron Exposure of Point Beach Unit 2 Beltline Materials A schematic of the beltline region of the Point Beach Unit 2 reactor pressure vessel is shown in Figure 2.3-1. In the case of the Unit 2 pressure vessel, the beltline region is constructed of two (2) ring forgings, two (2) circumferential welds (one joining the intermediate and lower shell forgings and one joining the intermediate shell forging to the upper shell forging), and the upper shell forging itself. Each of these five materials must be considered in the overall embrittlement assessments of the pressure vessel.

Pressure vessel materials not included in this compilation are not anticipated to experience a fast neutron (E > 1.0 MeV) exposure in excess of 1.OE+17 n/cm2 over the service life of the unit and, therefore, are not included in the definition of the beltline region.

The plant and fuel cycle specific calculated fast neutron (E > 1.0 MeV) fluence experienced by the materials comprising the beltline region of the Point Beach Unit 2 reactor are presented in Tables 2.3-1 through 2.3-5 for plant operation through the conclusion of the 27th fuel cycle. The data listed in Tables 2.3-1 through 2.3-5 provide a summary of the calculated fast neutron fluence (E > 1.0 MeV) for all Point Beach Unit 2 fuel cycles designed to date.

In Tables 2.3-1 through 2.3-5, cycle specific maximum neutron exposures at the pressure vessel cladlbase metal interface are given at azimuthal angles of 0°, 15', 30°,

and 45' relative to the core cardinal axes. The data provided in Tables 2.3-1 through 2.3-5 were taken at the axial location of the maximum exposure experienced by each material based on the results of the three-dimensional synthesized neutron fluence evaluations.

As noted in Section 1.0 of this report, three general fuel management strategies have been employed during the course of the first 27 fuel cycles of operation at Point Beach Unit 2. These three approaches are summarized as follows:

Cycles 1-5 6-1 5 16-25 Fuel Management Approach Out-In fuel management strategy with fresh fuel assemblies located on the core periphery.

Conventional low leakage fuel management strategy with burned fuel located on the core periphery.

Conventional low leakage fuel management strategy supplemented with the inclusion of part length hafnium power suppression rods located in the twelve peripheral fuel assemblies spanning the core cardinal axes.

For Point Beach Unit 2, the hafnium power suppression rods were 36 inches in length and were positioned axially such that the center of the suppression rods corresponded to the elevation of the intermediate shell to lower shell circumferential weld. Thus, the maximum flux reduction impact of the power suppression rods was experienced by the lower shell circumferential weld in the 0"-20" azimuthal. Other materials comprising the beltline region of the Point Beach Unit 2 pressure vessel were impacted to a much lesser degree by the presence of the part length absorber rods.

Projections of calculated fast neutron fluence to be accrued beyond the end of cycle 27 were based on the four potential scenarios described in Section 1.0 of this report. The assumed core power level and core power distributions used in these projections are summarized as follows:

For operation with hafnium power suppression rods, the relative core power distribution averaged over the twelve fuel cycles that have contained the suppression rods was assumed to be representative of future fuel cycle designs. For operation without the hafnium suppression rods, a relative power distribution averaged over the ten fuel cycles employing conventional low leakage (cycles 6-15) was assumed to be representative of future operation. In all cases, the transport calculations were normalized to the operating core power levels as noted in the preceding tabulation.

Case 5

6 7

8 Results of the fast neutron (E > 1.0 MeV) fluence projections for the Point Beach Unit 2 beltline materials are provided in Tables 2.3-6 through 2.3-15. In Tables 2.3-6 through 2.3-10, data are provided both with and without the presence of hafnium power suppression rods assuming continued operation at a power level of 1540.0 MWt until April 2008 followed by operation at the extended uprate power of 1678.0 MWt. In Tables 2.3-1 1 through 2.3-15, data are given both with and without the presence of hafnium absorbers based on continued operation at 1540.0 MWt through 54 effective full power years.

Applicable Cycles Cycles 28-29 Cycles 30 - 54 efpy Cycles 28-29 Cycles 30 - 54 efpy Cycles 28-29 Cycles 30 - 54 efpy Cycles 28-29 Cycles 30 - 54 efpy Power Distribution Avg. of cycles 16-27 Avg. of cycles 16-27 Avg. of cycles 16-27 Avg. of cycles 16-27 Avg. of cycles 16-27 Avg. of cycles 6-15 Avg. of cycles 16-27 Avg. of cycles 6-1 5 Core Power (MWt) 1540.0 1678.0 1540.0 1540.0 1540.0 1678.0 1540.0 1678.0

Figure 2.3-1 Locations of Pressure Vessel Beltline Materials Point Beach Unit 2 TOP VIEW z

Table 2.3-1 Intermediate Shell to Lower Shell Circumferential Weld (SA-1484)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 2

Table 2.3-2 Lower Shell Forging (1 22W 195)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 2 Fuel Cycle 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Cumulative Time

[ef~sl 4.80E+07 8.12E+07 1.09E+08 1.36E+08 1.64E+08 1.91 E+08 2.19E+08 2.47E+08 2.71 E+08 3.09E+08 3.36E+08 3.61 E+08 3.87E+08 4.14E+08 Cycle Time ref PSI 4.80E+07 3.31 E+07 2.75E+07 2.74E+07 2.79E+07 2.73E+07 2.82E+07 2.70E+07 2.50E+07 3.77E+07 2.68E+07 2.52E+07 2.55E+07 2.72E+07 Neutron Fluence [n/cm2]

45 Deg.

7.65E+17 1.33E+18 1.80E+18 2.22E+18 2.62E+18 3.04E+18 3.46E+18 3.85E+18 4.21 E+18 4.76E+18 5.1 1 E+18 5.41 E+18 5.70E+18 6.07E+18 30 Deg.

8.85E+17 1.53E+18 2.06E+18 2.54E+18 3.05E+18 3.56E+18 4.00E+18 4.42E+18 4.80E+18 5.38E+18 5.80E+18 6.17E+18 6.52E+18 6.92E+18 0 Deg.

2.1 5E+18 3.65E+18 4.87E+18 5.99E+18 7.1 5E+18 8.01 E+18 8.97E+18 9.95E+18 1.09E+19 1.21 E+19 1.28E+19 1.35E+19 1.42E+19 1.49E+19 15 deg.

1.30E+18 2.21 E+18 2.94E+18 3.62E+18 4.35E+18 4.93E+18 5.51 E+18 6.09E+18 6.63E+18 7.38E+18 7.90E+18 8.39E+18 8.85E+18 9.36E+18

Table 2.3-3 Intermediate Shell Forging (1 23V500)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 2

Table 2.3-4 Intermediate Shell to Upper Shell Circumferential Weld (21 935)

Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 2

Table 2.3-5 Upper Shell Forging Neutron Fluence (E > 1.0 MeV)

Point Beach Unit 2

Table 2.3-6 Intermediate Shell to Lower Shell Circumferential Weld (SA-1484)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 2.3-7 Lower Shell Forging (1 22W 195)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 2.3-8 Intermediate Shell Forging (1 23V500)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 2.3-9 Intermediate Shell to Upper Shell Circumferential Weld (21 935)

Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 2.3-10 Upper Shell Forging Calculated Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 2.3-1 1 Intermediate Shell to Lower Shell Circumferential Weld (SA-1484)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.3-1 2 Lower Shell Forging (1 22W 195)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.3-1 3 Intermediate Shell Forging (1 23V500)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.3-14 Intermediate Shell to Upper Shell Circumferential Weld (21 935)

Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 2.3-1 5 Upper Shell Forging Calculated Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy

3.0 Neutron Dosimetry Evaluations 3.1 Method of Analysis Evaluations of neutron sensor sets contained in the in-vessel and ex-vessel dosimetry capsules withdrawn to date from the Point Beach Units 1 and 2 reactors were evaluated using the current state of the art least squares methodology.

Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a best estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure parameters such as @(E > 1.0 MeV) along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to reactor dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, relates a set of measured reaction rates, Ri, to a single neutron spectrum, Qg, through the multigroup dosimeter reaction cross-section, og, each with an uncertainty 6. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the Point Beach Units 1 and 2 dosimetry, the FERRET codeL5'] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best estimate values of exposure parameters along with associated uncertainties.

The application of the least squares methodology requires the following input:

1 -

The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2 -

The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3 -

The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the current application, the calculated neutron spectrum at each measurement location was obtained from the results of plant specific neutron transport calculations described in Section 2.0 of this report. The spectrum at each sensor set location was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements. The sensor reaction rates were derived from the measured specific activities of each sensor set and the operating history of the respective reactors. The dosimetry reaction cross-sections were obtained from the SNLRML dosimetry cross-section librap].

In addition to the magnitude of the calculated neutron spectra, the measured sensor set reaction rates, and the dosimeter set reaction cross-sections, the least squares procedure requires uncertainty estimates for each of these input parameters. The following provides a summary of the uncertainties associated with the least squares evaluation of the Point Beach Units 1 and 2 dosimetry.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

These uncertainties are given at the lo level.

Reaction Cd3(n, a)Co60

~ e ~ ~ ( n,

p) ~n~~

~i~'(n, p)Co5'

~ ~ ~ ' ( n, f ) C s ' ~ ~

~ p ~ ~ ~ ( n, f ) ~ s ' ~ '

Co5'(n, -y)Co60 Dosimetry Cross-Section Uncertainties As noted above, the reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources. Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors is provided in Reference 58.

Uncertainty 5%

5%

5%

10%

10%

5%

For sensors included in the Point Beach Units 1 and 2 dosimetry sets, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Reaction

~ u ~ ~ ( n, a ) c o ~ ~

Fe"(n, p ) ~ n ~ ~

N i58(n, p ) ~ o ~ ~

u ~ ~ ~ ( ~, ~ ) F P

~p~~'(n,f)Fp co5'(n, y)co60 Calculated Neutron Spectrum Uncertainties While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Uncertainty 4.08-4.16%

3.05-3.11 %

4.49-4.56%

0.54-0.64%

10.32-1 0.97%

0.79-3.59%

M,.

=RZ +Rg *Fig. *Pgg.

where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg9 and R, specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

where The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short range correlations over a group range y (8 specifies the strength of the latter term). The value of 6 is 1.0 when g = g' and 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Point Beach Units 1 and calculated spectra was as follows:

Flux Normalization Uncertainty (R,)

15%

Flux Group Uncertainties (&, R,.)

(E > 0.0055 MeV)

(0.68 eV c E c 0.0055 MeV)

(E < 0.68 eV)

Short Range Correlation (8)

(E > 0.0055 MeV)

(0.68 eV c E c 0.0055 MeV)

(E c 0.68 eV)

Flux Group Correlation Range (y)

(E > 0.0055 MeV)

(0.68 eV < E < 0.0055 MeV)

(E < 0.68 eV)

These uncertainty assignments are consistent with an industry consensus uncertainty of 15-20% (lc) for the fast neutron portion of the spectrum and provide for a reasonable increase in the uncertainty for neutrons in the intermediate and thermal energy ranges.

3.2 Dosimetry Evaluations for Point Beach Unit 1 During the course of first 27 operating fuel cycles at Point Beach Unit 1, four in-vessel and 20 ex-vessel sensor sets were irradiated at the core midplane elevation and subsequently withdrawn for analysis. A summary of the locations and time of irradiation of each of these multiple foil sensor sets is provided in Table 3.2-1.

In this section, comparisons of the measurement results from each of the sensor set irradiations with corresponding analytical prediction at the measurement locations are presented. These comparisons are provided on two levels. In the first instance calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least squares evaluations of the sensor sets. In the second instance, calculated values of neutron exposure rates in terms of Q(E > 1.0 MeV) are compared with the best estimate exposure rates obtained from the least squares evaluation.

In Table 3.2-2, comparisons of measured to calculation (ME) ratios are listed for the threshold sensors contained in in-vessel capsules S, R, and T. Data from capsule V, the first capsule withdrawn from Point Beach Unit 1, was not included in the comparisons, since the original capsule report did not provide sufficient information to upgrade the data evaluation to current methodology and nuclear data. From Table 3.2-2, it is noted that for the individual threshold foils the average M/C ratio ranges from 0.94 to 1.16 with an overall average of 1.03 with an associated standard deviation of 8.1%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table 3.2-3 similar comparisons are provided for the twenty sensor sets withdrawn from the core midplane elevation in the reactor cavity. From Table 3.2-3, it is noted that for the individual threshold foils the average M/C ratio ranges from 0.84 to 0.95 with an overall average of 0.88 with an associated standard deviation of 5.6%. As in the case of the in-vessel comparisons, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Tables 3.2-4 and 3.2-5, best estimate to calculation (BEE) ratios for fast neutron flux (E > 1.0 MeV) resulting from the least squares evaluation of each dosimetry set are provided for the in-vessel and ex-vessel irradiations, respectively. For the in-vessel capsules the average BEIC ratio is seen to be 1.02 with an associated uncertainty of 6.7%. The corresponding average BEE ratio from the ex-vessel irradiations is 0.86 with an uncertainty of 10.6%.

Table 3.2-1 Location and Time of lrradiation for Sensor Sets Withdrawn from Point Beach Unit 1 Axial Elevation Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Azimuthal Location 13 Deg.

23 Deg.

13 Deg.

33 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

Capsule ID V

S R

T 17-1 17-2 17-3 17-4 18-1 18-2 18-3 18-4 19-1 1 9-2 19-3 19-4 22-1 22-2 22-3 22-4 24-1 24-2 24-3 24-4 Cycles of Irradiation 1

1-3 1-5 1-11 17 17 17 17 18 18 18 18 19 19 19 19 20-22 20-22 20-22 20-22 23-24 23-24 23-24 23-24 In-Vessel In-Vessel In-Vessel In-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel

Table 3.2-2 Comparison of Measure and Calculated Threshold Foil Reaction Rates In-Vessel Capsules - Point Beach Unit 1 Capsule v

S R

T Average

% std dev Reaction 63Cu (n, a) 54~e(n,~)

5 8 ~ i ( n, ~ )

238~(n,f) 237~p(n,f)

Linear Average M/C Ratio Average MIC 1.05 1.OI 0.94 1.OO 1.16 1.03 63 Cu(n,a) 1.O1 1.06 1.08 1.05 3.6

% Standard Deviation 3.6 3.1 12.8 8.0 5.5 8.1

'"u(n,r) 0.91 1.06 1.02 1.OO 8.0 2"Npo 1.11 1.15 1.24 1.16 5.5 54Fe(n,~)

0.97 1.04 1.O1 1.01 3.1 "Ni(n,~)

0.86 1.03 0.94 12.8

Table 3.2-3 Comparison of Measure and Calculated Threshold Foil Reaction Rates Ex-Vessel Capsules - Point Beach Unit 1 Capsule 17-1 17-2 17-3 17-4 18-1 18-2 18-3 18-4 19-1 19-2 19-3 19-4 22-1 22-2 22-3 22-4 24-1 24-2 24-3 24-4 Average

% std dev Reaction 6 3 ~ u (n, a)

="W~,P) 5 8 ~ i ( n, ~ )

'%(n,f) 237~p(n,f)

Linear Average 237N~(n,9 0.87 0.88 0.99 1.07 0.79 0.91 1.14 1.10 0.89 0.94 1.04 1.22 0.83 0.91 0.67 0.96 0.95 14.7 Average MIC 0.92 0.85 0.84 0.85 0.95 0.88

%u(n,a) 0.80 0.89 0.96 0.97 0.76 0.90 1.OO 1.OO 0.79 0.92 1.03 1.OO 0.80 0.89 0.98 1.03 0.75 0.90 0.97 1.01 0.92 10.1

% Standard Deviation 10.1 9.9 10.5 13.3 14.7 5.6 M/C Ratio 5 8 ~ i ( n, ~ )

0.73 0.79 0.86 0.89 0.72 0.83 0.95 0.95 0.72 0.81 0.91 0.92 0.73 0.81 0.92 0.95 0.69 0.84 0.90 0.94 0.84 10.5 54Fe(n,~)

0.75 0.80 0.87 0.88 0.73 0.84 0.92 0.94 0.73 0.84 0.92 0.93 0.74 0.79 0.93 0.93 0.71 0.84 0.95 0.94 0.85 9.9 238u(n99 0.68 0.77 0.83 0.89 0.69 0.82 0.88 0.92 0.71 0.83 1.OO 0.96 0.70 0.76 0.89 0.91 0.74 0.91 1.OO 1.07 0.85 13.3

Table 3.2-4 Comparison of Best Estimate and Calculated Neutron Flux (E > 1.0 MeV)

In-Vessel Capsules - Point Beach Unit 1

% Standard Deviation 7.0 7.0 7.O 6.7 Capsule v

S R

T Average BEIC 0.94 1.06 1.06 1.02

Table 3.2-5 Comparison of Best Estimate and Calculated Neutron Flux (E > 1.0 MeV)

Ex-Vessel Capsules - Point Beach Unit 1

% Standard Deviation 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 10.6 Reaction 17-1 17-2 17-3 17-4 18-1 18-2 18-3 18-4 19-1 19-2 19-3 19-4 22-1 22-2 22-3 22-4 24-1 24-2 24-3 24-4 Average BElC 0.75 0.80 0.87 0.91 0.73 0.84 0.95 0.96 0.75 0.84 0.94 0.98 0.74 0.81 0.91 0.94 0.70 0.85 0.94 0.96 0.86

3.3 Dosimetry Evaluations for Point Beach Unit 2 During the course of first 25 operating fuel cycles at Point Beach Unit 2, four in-vessel and 20 ex-vessel sensor sets were irradiated at the core midplane elevation and subsequently withdrawn for analysis. A summary of the locations and time of irradiation of each of these multiple foil sensor sets is provided in Table 3.3-1.

In this section, comparisons of the measurement results from each of the sensor set irradiations with corresponding analytical prediction at the measurement locations are presented. These comparisons are provided on two levels. In the first instance calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least squares evaluations of the sensor sets. In the second instance, calculated values of neutron exposure rates in terms of $(E > 1.0 MeV) are compared with the best estimate exposure rates obtained from the least squares evaluation.

In Table 3.3-2, comparisons of measured to calculation (MIC) ratios are listed for the threshold sensors contained in in-vessel capsules V, S, R, and T. From Table 3.2-2, it is noted that for the individual threshold foils the average MIC ratio ranges from 0.97 to 1.09 with an overall average of 1.01 with an associated standard deviation of 2.8%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table 3.3-3 similar comparisons are provided for the twenty sensor sets withdrawn from the core midplane elevation in the reactor cavity. From Table 3.3-3, it is noted that for the individual threshold foils the average M/C ratio ranges from 0.89 to 0.99 with an overall average of 0.93 with an associated standard deviation of 4.6%. As in the case of the in-vessel comparisons, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Tables 3.3-4 and 3.3-5, best estimate to calculation (BUC) ratios for fast neutron flux (E > 1.0 MeV) resulting from the least squares evaluation of each dosimetry set are provided for the in-vessel and ex-vessel irradiations, respectively. For the in-vessel capsules the average BUC ratio is seen to be 1.O1 with an associated uncertainty of 0.6%. The corresponding average BEIC ratio from the ex-vessel irradiations is 0.91 with an uncertainty of 3.3%.

Table 3.3-1 Location and Time of lrradiation for Sensor Sets Withdrawn from Point Beach Unit 2 Cycles of Irradiation 1

1-3 1 -5 1-16 15 15 15 15 16 16 16 16 17 17 17 17 18-20 18-20 18-20 18-20 21 -23 21 -23 21 -23 21 -23 Axial Elevation Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Core Midplane Azimuthal Location 13 Deg.

23 Deg.

13 Deg.

33 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

0 Deg.

15 Deg.

30 Deg.

45 Deg.

Capsule ID V

T R

S 15-1 15-2 15-3 15-4 16-1 16-2 16-3 16-4 17-1 17-2 17-3 1 7-4 20-1 20-2 20-3 20-4 23-1 23-2 23-3 23-4 I n-Vessel I n-Vessel I n-Vessel In-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel Ex-Vessel

Table 3.3-2 Comparison of Measure and Calculated Threshold Foil Reaction Rates In-Vessel Capsules - Point Beach Unit 2 Capsule V

T R

S Average

% std dev Reaction 63Cu(n,a)

"Fe(n,p) 5 8 ~ i ( n, ~ )

238~(n,f) 237 Np(n,f)

Linear Average MIC Ratio Average MIC 0.98 0.97 0.97 1.05 1.09 1.01

% Standard Deviation 6.1 4.0 4.3 5.8 2.8 5.7

'"u(n,9 1.11 0.98 1.03 1.09 1.05 5.8 W ~ i ( n, ~ )

0.98 1.OO 0.92 0.97 4.3

%u(n,a) 0.92 0.95 1.O1 1.05 0.98 6.1

'"Np(n19 t.05 1.1 1 1.12 1.09 1.09 2.8 54Fe(n,~)

0.92 1.OO 0.97 0.97 4.0

Table 3.3-3 Comparison of Measure and Calculated Threshold Foil Reaction Rates Ex-Vessel Capsules - Point Beach Unit 2 Capsule 15-1 15-2 15-3 15-4 16-1 16-2 16-3 16-4 17-1 17-2 17-3 17-4 20-1 20-2 20-3 20-4 23-1 23-2 23-3 23-4 Average

% std dev Reaction 63~u(n,a) 54~e(n,~)

5 8 ~ i ( n, ~ )

'"u(n,f) 237~p(n,f)

Linear Average

='Np(n99 0.94 1.03 1.03 1.09 0.97 0.98 1.O1 1.15 0.99 0.91 0.99 0.95 0.87 0.95 0.98 0.98 0.96 1.02 0.99 6.5 Average MIC 0.96 0.89 0.90 0.91 0.99 0.93 238u(n,9 0.86 0.94 0.91 0.96 0.88 0.97 0.92 0.94 0.81 0.89 0.92 0.88 0.85 0.92 0.89 0.89 0.90 0.95 0.96 0.94 0.91 4.5

% Standard Deviation 4.3 4.1 3.8 4.5 6.5 4.6 M/C Ratio "Ni(n,p) 0.88 0.95 0.91 0.95 0.85 0.89 0.93 0.96 0.87 0.91 0.91 0.93 0.87 0.88 0.89 0.89 0.89 0.90 0.84 0.85 0.90 3.8 "Cu(n,a) 0.91 1.O1 0.95 1.02 0.90 0.95 1.OO 1.02 0.93 0.99 0.96 1.O1 0.93 0.98 0.98 0.98 0.90 0.94 0.92 0.91 0.96 4.3 54Fe(n,~)

0.87 0.94 0.94 0.92 0.84 0.87 0.93 0.95 0.87 0.92 0.91 0.93 0.85 0.87 0.87 0.86 0.89 0.89 0.85 0.84 0.89 4.1

Table 3.3-4 Comparison of Best Estimate and Calculated Neutron Flux (E > 1.0 MeV)

In-Vessel Capsules - Point Beach Unit 2 Capsule V

S R

T Average BEIC 1.01 1.02 1.02 1.O1 1.O1

% Standard Deviation 7.0 7.0 7.0 7.0 0.6

Table 3.3-5 Comparison of Best Estimate and Calculated Neutron Flux (E > 1.0 MeV)

Ex-Vessel Capsules - Point Beach Unit 2

% Standard Deviation 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 7.0 3.3 Reaction 15-1 15-2 15-3 15-4 16-1 16-2 16-3 16-4 17-1 17-2 17-3 1 7-4 20-1 20-2 20-3 20-4 23-1 23-2 23-3 23-4 Average BE/C 0.88 0.95 0.94 0.96 0.87 0.92 0.94 0.98 0.87 0.90 0.93 0.91 0.87 0.92 0.92 0.92 0.89 0.91 0.89 0.90 0.91

Table 3.3-4 Summary of In-Vessel and Ex-Vessel Data Comparions

'%u(n,a) 5 4 ~ e ( n, ~ )

5 8 ~ i ( n, ~ )

238U(n,f) 237Np(n,f)

All Reactions

%u(n,a)

"Fe(n,p)

"Ni(n,p)

• "u(n,r) 237~p(n,f)

. All Reactions Ex-Vessel Data In-Vessel Data Average M/C 0.96 0.91 0.94 0.94 1.01 0.95 Combined Data Set Average MIC 0.98 0.97 0.97 1.05 1.09 1.O1

% std dev 5.6 7.2 8.6 8.4 8.2 3.4 Average WC 0.97 0.94 0.95 1.OO 1.05 0.98

% std dev 6.1 4.0 4.3 5.8 2.8 2.1

% std dev 4.2 4.1 4.8 5.0 4.2 2.0

4.0 Best Estimate Exposure Evaluations In this section the measurement results provided in Section 3.0 are combined with the results of the neutron transport calculations described in Section 2.0 to establish best estimates for the neutron exposure of the Point Beach Units 1 and 2 pressure vessel beltline materials. The best estimate exposures were determined from the following equation:

where: aBE

= The best estimate neutron exposure for the material of interest.

K

= The plant specific best estimate BEIC bias factor derived from the set of in-vessel and ex-vessel measurements available for each reactor.

Qc = The absolute calculated fast neutron exposure for the material of interest.

It should be noted that the approach described in this section has not been approved by the NRC staff. Specifically, the approval granted in Reference 59 for the calculational methodology described in Section 2.0 of this report allows the measurement results provided in Section 3.0 of this report to be used to validate, but not to modify the calculated results. A report describing the benchmarking of the least squares methodology as implemented in the FERRET code[641 will be submitted for approval in the near future.

In the development of the bias factor (K) for each of the respective reactors, the in-vessel and ex-vessel comparisons were treated as independent data sets and combined linearly to produce an overall value for K.

In the case of Point beach Unit 1 the in-vessel, ex-vessel, and combined BEIC comparisons for neutron flux (E > 1.0 MeV) may be summarized as follows:

The corresponding data comparisons applicable to Point Beach Unit 2 are as follows:

In-Vessel Ex-Vessel KI BEIC 1.02 0.86 0.94 I n-Vessel Ex-Vessel K2

% Uncertainty 6.7 10.6 6.1 BUC 1.O1 0.91 0.96

% Uncertainty 0.6 3.3 1.6

Resultant best estimate fluence projections for Point Beach Unit 1 are listed in Tables 4-1 through 4-12 for each of the beltline materials. Projections are provided for fuel management scenarios with and without hafnium power suppression rods and with and without the extended power uprate to 1678.0 MWt. Corresponding fluence projections for Point Beach Unit 2 are given in Tables 4-1 3 through 4-22.

Table 4-1 Intermediate Shell to Lower Shell Circumferential Weld (SA-1101)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-2 Lower Shell Plate (C-1423)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-3 Intermediate Shell Plate (A-981 1)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-4 Intermediate Shell (SA7751812) and Lower Shell (SA-847) Longitudinal Welds Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-5 Intermediate Shell to Upper Shell Circumferential Weld (SA-1426)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-6 Upper Shell Forging Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 1 - Case 1 and Case 3 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Power uprate to 1678.0 MWt was assumed to occur in October 2008 at 29.7 efpy of operation.

Table 4-7 Intermediate Shell to Lower Shell Circumferential Weld (SA-1101)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-8 Lower Shell Plate (C-1423)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-9 Intermediate Shell Plate (A-981 1)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-1 0 Intermediate Shell (SA7751812) and Lower Shell (SA-847) Longitudinal Welds Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Operating Time Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy Neutron Fluence [nlcmL]

Inter.

Lower

Table 4-1 1 Intermediate Shell to Upper Shell Circumferential Weld (SA-1426)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-1 2 Upper Shell Forging Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 1 - Case 2 and Case 4 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 29 (- October 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-1 3 Intermediate Shell to Lower Shell Circumferential Weld (SA-1484)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 4-1 4 Lower Shell Forging (1 22W 195)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Time" column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 4-1 5 Intermediate Shell Forging (1 23V500)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 4-1 6 Intermediate Shell to Upper Shell Circumferential Weld (21 935)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 4-1 7 Upper Shell Forging Adjusted Neutron Fluence (E > 1.0 MeV) Projections With Uprate Point Beach Unit 2 - Case 5 and Case 7 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Power uprate to 1678.0 MWt was assumed to occur in April 2008 at 29.1 efpy of operation.

Table 4-1 8 Intermediate Shell to Lower Shell Circumferential Weld (SA-1484)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-1 9 Lower Shell Forging (1 22W 195)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-20 Intermediate Shell Forging (1 23V500)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-2 1 Intermediate Shell to Upper Shell Circumferential Weld (21 935)

Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

Table 4-22 Upper Shell Forging Adjusted Neutron Fluence (E > 1.0 MeV) Projections Without Uprate Point Beach Unit 2 - Case 6 and Case 8 With Hafnium Suppression Rods Without Hafnium Suppression Rods Note: The first value listed in the "Operating Timen column is the estimated efpy at the end of cycle 27 (- April 2005).

Operation at 1540 MWt was assumed to occur from 02/03/2003 through 54 efpy.

References Code of Federal Regulations Title 10 Part 50, "Domestic Licensing of Production and Utilization Facilities".

Regulatory Guide 1.1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

RSlCC Computer Code Collection CCC-650, "DOORS 3.1, One-, Two-, and Three-Dimensional Discrete Ordinates NeutronIPhoton Transport Code System,"

August 1996.

RSlC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDFIB-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

L. R. Scherpereel, "Core Physics Characteristics of the Point Beach Nuclear Plant (Unit 1, Cyclel)," WCAP-7430, December 1969.

E. J. Piplica, "The Nuclear Design - Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 2," WCAP-8120, March 1973.

E. J. Piplica and J. P. Hawrylak, "The Nuclear Design - Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 3," WCAP-8325, May 1974.

D. J. Franks, "The Nuclear Design - Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 4," WCAP-8652, November 1975.

A. Saeed and K. A. Forcht, The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 5," WCAP-8120, September 1976.

A. Saeed and K. A. Forcht, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 6," WCAP-9131, August 1977.

A. Saeed and K. A. Forcht, The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 7,"

WCAP-9368, August 1978.

K. A. Forcht, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 8," WCAP-9548, August 1979.

M. L. Hubbard, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 9AlW WCAP-9841, January 1981.

P. W. Robertson and J. L. Stern, 'The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 10," WCAP-9974, October 1981.

R. D. Jones and J. L. Cole, The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 11," WCAP-10191, November 1982.

R. D. Jones and J. L. Cole, The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 12," WCAP-10497, March 1984.

17.

R. D. Jones, et al, 'The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 13," WCAP-10799, May 1985.

R. D. Jones, et al,, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 14," WCAP-11118, April 1986.

D. M. Chapman, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor -Cycle 15," WCAP-11487, May 1987.

D. M. Chapman, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 16," WCAP-11754, March 1988.

D. M. Chapman, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 17," WCAP-12194, March 1989.

D. M. Chapman, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 18," WCAP-12524, March 1990.

D. M. Chapman, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 19," WCAP-12903, March 1991.

T. P. Phelps, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 20," WCAP-13204, March 1992.

T. P. Phelps, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 21," WCAP-13653, March 1993.

T. P. Phelps and R. T. Smith, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 22," WCAP-13989, March 1994.

T. P. Phelps and R. T. Smith, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 23," WCAP-14294, March 1995.

R. T. Smith and M. A. Kotun, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 24," WCAP-14609, March 1996.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 25," WCAP-15045, June 1998.

M. A. Kotun, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 26," WCAP-15326, October 1999.

M. A. Kotun, et al, "The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 27," WCAP-15661, April 2001.

L. R. Scherpereel, "Core Physics Characteristics of the Point Beach Nuclear Plant (Unit 1, Cyclel)," WCAP-7430, December 1969.

J. P. Hawrylak, "Revised Cycle 2 Nuclear Design Characteristics for Point Beach Unit 2," WCAP-8418, Reviion 1, November 1974.

34.

J. P. Hawrylak and S. A. Antin, "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 3," WCAP-8759, March 1976.

J. P. Hawrylak and J. L. Dauberman, "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 4," WCAP-8934, February 1977.

J. P. Hawrylak and K. A. Forcht, "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 5," WCAP-9275, February 1978.

E. H. Pilzer and K. A. Forcht, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 6," WCAP-9493, April 1979.

W. J. Scherder and K. A. Forcht, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 7," WCAP-9667, February 1980.

R. T. Smith and J. L. Stern, 'The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 8," WCAP-9846, March 1981.

R. T. Smith and J. L. Stern, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 9," WCAP-10048, March 1982.

R. T. Smith and R. D. Jones, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle lo," WCAP-10278, March 1983.

W. A. Boyd, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 11 Rev. 1," WCAP-10583, Rev-1, August 1984.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 12 REV. 1," WCAP-10897, Rev. 1, November 1985.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 13," WCAP-11288, November 1986.

R. T. Smith, et at, The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 14," WCAP-11571, September 1987.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle t 5," WCAP-1 t 903, September 1988.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 16," WCAP-12362, September 1989.

R. T. Smith, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 17," WCAP-12683, September 1990.

J. A. Hoerner, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 18," WCAP-13063, September 1991.

J. A. Hoerner, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 19," WCAP-13467, September 1992.

B. R. Beebe, et al, 'The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 20," WCAP-13843, September 1993.

R. T. Smith and 6. R. Beebe, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 21,"

WCAP-14158, September 1994.

R. T. Smith, The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 22," WCAP-14458, September 1995.

R. T. Smith, The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 23," WCAP-14735, September 1996.

R. T. Smith, et al, The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 24," WCAP-15155, January 1999.

R. D. Erwin, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 25," WCAP-15593, October 2000.

RSlC Computer Code Collection PSR-145, "FERRET Least-Squares Solution to Nuclear Data and Reactor Physics Problems," January 1980.

RSlC Data Library Collection DLC-178, 'SNLRML Recommended Dosimetry Cross-Section Compendium, " July 1994.

USNRC 02/27/04 WOG - Final SE for Topical Report WCAP-14040, Rev. 3, "Method Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". (TAC. MB5754).

C. Guler, et al, The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 28," WCAP-15944, September 2002.

E. Lenzo, et al, 'The Nuclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor - Cycle 29," WCAP-16233-P, March 2004.

M. C. Adler, et al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 26," WCAP-15865, Rev 1, June 2002.

C. Guler, et a!, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 27," WCAP-16150, rev 1, November 2003.

S. L. Anderson, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," WCAP-16083-NP, April 2004.

ENCLOSURE 2a SUPPLEMENT TO LICENSE AMENDMENT REQUEST 251 POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 OPERABILITY RECOMMENDATION OPR000175, REVISION 0 7 Pages Follow

3F-1100 Rev 2 (FP-OP-OL-01)

SSC affected by condition: PBNP-1 Reactor Vessel I Identify the overal scope of the condition that calls OPEfUBILtTY into question.

PBNP Units 1 and 2 use a common set of Pressure-Temperature (P-T) Limit Curve and LTOP setpoints. These curves and setpoints are documented in Technical Requirements Manual (TRM) Section 2.2 'Pressure Temperature Limits Report" (Reference 1). The applicability of TRM 2.2, and its included curves and setpoints, are defmed by specific limits of neutron fluence.

The neutron fluence limits of TRM 2.2 are exceeded on several PBNP Unit 1 material locations using the most recent neutron exposure calculations (Reference 2).

The neutron fluence limits for PBNP Unit 2 have not been exceeded.

I The scope of this operability determination is to evaluate this condition as it relates to the operability of the PBNP Unit 1 reactor vessel.

Describe the specified safety, or safety support, function(s) of the SSC. Identify the Licensing Basis functions and performance requirements, including Technical Specifications. FSAR, NRC Commitments, or other appropriate information (reference SCOPE section 5.3).

Final Safety Analysis Rewrt The reactor vessel is part of the reactor coolant system. As stated in FSAR Section 4.1 "Reactor Coolant System," the reactor coolant pressure boundary shall be designed, fabricated, and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime.

FSAR Section 1.3.6 "Reactor Coolant Pressure Boundary" discusses that the pressure-temperature limits are determined in accordance with the methods of analysis and the margins of safety of Appendix G of ASME Code Section XI and are included in the Point Beach Pressure Temperature Limits Report (PTLR).

More specifically, the reactor vessel and the pressure-temperature limits must adhere to General Design Criteria 34 which states:

"The reactor coalant pressure boundary shaN be deslgned and operated to reduce to an acceptable level the probability of rapidIy propaga&g type failures. Cmsideration is given (a) lo the provisions for contrd over service temperature and irradiation effects which may require operational restrictions, (b) to ihe desgn and construction of the reactor pressure vessel in accordance with applicable codes, including those which establish requirements for absorption of energy within the elastic strain energy range and for absorptian of energy by plastic deformation and (c) to the design and conslruction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes. "

vessel must also adhere to General Design Criterion 36 'Reactor Coolant Pressure

"Reactor coolant pressure boundary components shaN have provisions for inspection, festrng, and surveillance of critical areas by appropriate means to assess the structural and leaktight integnty of the boundary components during their service lifetime. For the reador vessel, a material surveiiance program conforming with current applicable codes shall be provided. "

The reactor coolant pressure boundary is designed to reduce to an acceptable level the probability of a rapidly propagating type failure. The fracture toughness of the materials in the beltline region of the reactor vessel will decrease as a result of fast neutron irradiation induced ernbrittlement. The decrease in fracture toughness, which is a function of several factors, including accumulated fast neutron fluence, requires a corresponding increase in reference nil ductility temperature (RTNDT) in order to maintain strengthistress requirements. This change in material properties is factored into the operating procedures such that the reactor coofant system pressure is limited with respect to RCS temperature during plant heatup. cooldown, and normal operation. These limits are determined in accordance with the methods of analysis and the margins of safety of Appendix G of ASME Code Section XI and are included in the Point Beach PTLR. The Low Temperature Overpressure Protection System pmvides protection during low-temperature operations. All pressure containing components of the Reactor Cooiant System are designed, fabricated, inspected, and tested in conformance with the applicable codes at the time of order placement.

The operating pressure temperature limits for the reactor vessel are located in Technical Requirements Manual (TRM)

Section 2.2 'Pressure Temperature Limits R e w (Reference 1).

Technical S~ecifications Technical Specification 3.4.3 "RCS Pressure and Temperature (PIT) Limits state that RCS pressure, RCS temperature, and RCS heatup and cootdown rates shall be maintained within the limits specified in the PTLR.

Technical Specification 5.6.5 'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)" discusses reporting requirements of the PTLR and states that the NRC will review and approve analytical methods to determine the RCS pressure and temperature limits.

I NRC Commitments NRC Letter NPC 2001-01263 'Point Beach Nuclear Plant. Units 1 8 2. - Acce~tance of Methodology for Referencing Pressure Temperature ~imits Report" - his NRC document quires that PBNP will not utilize the FERRET code until staff approval is received, and FERRET code additions to the PRR will be submitted to the NRC. The document also states that the that Point Beach's pred~cted expiration of the PTLR was October 30,2003, for Unit 1, and October 1. 2008, for Unit 2.

NhlC Letter NRC 2003-0108 -Addresses the P-T Curves applicability discussed in NRC Letter 01 -002/1. Because actual plant operation is dflerent that that predicted it1 the PTLR. this letter informs the NRC that fluence values associated with 25.59 EFPY will not be reached on Unit 1 until February, 2004. PBNP staled that they would revise the PT curves prior to exceeding the 1

i.. :. -. _..

25 59 EFPY for Unil1.

SER 2005-0006 'Safety Evaluation Report (SER) with Open Items Related lo the License Renewal of Point Beach Nuclear Plant, Units 1 and 2, dated May 2,2005 condudes that PBNP's methods of fluence calculation adhere to the requirements of NRC RG 1.190.

'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials" specifies method used to predict the effect of neutron radiation on reactor vessel materials unless they can justify the use of different methods. This regulatory guide is referenced in the above SER NRC Reaulations 10CFRSO Appendix G 'Fracture Toughness Requirements" specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor codant pressure boundary to provide adequate margins of safety during any condition of normal 10CFR50 Appendix H "Reactor Vessel Material Surveillance Program Requirements" describes material surveillance program requirements to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from expure of these materials to neutron irradiation and the thermal environment.

10CFR50.61 'Fracture Toughness Requirements for Proiect~on Against Pressurized Thermal Shock Events" states lirnrt and requirement for calculations of End Of Life (EOL)

RTWT values.

Evaluate the effects of the condition, including potential failure modes, on the ability of the SSC to perform its specified safety, or safety support, function(s)

Exceed~ng limits of P-T Curve fast neutron tluence would decrease the predicted fracture toughness of the reactor vessel components. An unanalyzed decrease in fracture toughness, without corresponding changes to allowed RCS pressure and temperature, could jeopardize the ability of the reactor vessel to perfom it safety and support functions.

I Is the SSC in its present condition capable of performing its safety or safe9 suppMt function(s)?

Explain basis.

(Use engineering analysis or engineering judgment to determine whether the design safety function can be provided given the existence of the deficiency. When using engineering judgment, provide supporting information from sources such as field walkdowns, industry experience, proven systemlc~nponent performance under similar service conditions, etc.)

Some of the PBNP Unit 1 fluence values are in non-conformance with the accepted fluence calculation methodology stated in the PTLR. However, the PBNP Unit 1 Reactor Vessel is fulb capable of performing all of its required safety and safety support functions. As described below, using recently approved fluence catculation methodology (FERRET Code) and the most recent fluence calculations, demonstrate that PBNP Unit 1 hence limits are not exceeded. The Unit 1 vessel will remain within the best estimate fluence values until an EFPY of 32.2 is 1 reached.

I PBNP Calculation 2000-0001 (Reference 3) performed the calculation of the P-T curves and LTOP setpoints that are contained in TRM 2.2 (or the Pressure Temperature Limits Report i (PTLR)). This calculation used what is termed 'best-estimate" fluence calculations as a design input. "Best Estimate" refers to the method of correcting calculated neutron fluence calculations for plant specific biases due to plant-specific core geometry. A Westinghouse code termed FERRET performed this 'best estimate" calculat~on (References 4 & 5).

During the approval process of the PBNP PTLR (TRM 2.2) the NRC recognized that Westinghouse's FERRET fluence code had not received formal approval (Reference 6 % 7). In response to NRC M s, PBNP Calculation Addendum 2000-0001-01 was prepared to state that only calculated (vice best estimate" or FERRET code) neutron fluence values were to be used.

Additionally, the predicted applicable EFPY values of the curves were calculated (Reference 8) based on the calculated values of fluence to account for the conservative action of not using the FERRET code. The PTLR curve remained the same. An EFPY value based on the calculated values of fluence and an EFPY based on the best estimate fluence are both contained in the PTLR. We are committed to using the calculated value. The calculated values without the FERRET code result in Unit 1 exceeding the established allowed fluence for the PTtR.

NRC approval of the FERRET code was received by the Westinghouse Owners Group on January 10, 2006. The NRC SER (Reference 9) states that the FERRET code is acceptable to be referenced if the licensee's analyses comply with the uncertainty values approved by the SER. Westinghouse letter WEP 06-13 (Reference 10) documents that the uncertainty values used by the PBNP fluence caIculations (References 4 & 5) comply with the requirements of the NRC SER on FERRET code. Therefore, use d FERRET code fiuence values in the PBNP PTLR is acceptable for confirming operability. Use of the FERRET code is currently not part of our licensing basis and the condition is non conforming.

I The latest PBNP fluence calculations using the FERRET code adjustment (Reference 2) show that the fluence values are well within the required limits. The best estimate fluence values used I

in generating the PTLR are conservative and allow operation of the Unit 1 Reactor Vessel to 32.2 EFPY and the Unit 2 Reactor Vessel to 34 EFPY. The Consolidated Data Entry database 1 maintained by INPO reports tk January 2006 cumulalrve Life of Plan1 Reactor ~ r i t k a l hours as

! 255386.72 for Unrt 1 and 2491 11.82 for Unit 2. A conservative assumption is that all hours were

._--.14r>-.'

As such, the PBNP Unit 1 Reactor Vessel can support its specified safety, and safety support,

QF-t to0 Rev 2 (FP-OP-OL-01)

If the SSC is not fully capable (Full QuarIfiwtion) of performing ik safety or safety support function(s), then determine if Compensatory Measures are to be taken to restore OPERABWTY or ro enhance the czyubttity a f thc SSC.

(Describe the Compensatory Measures, basis for which the Compensatory Measures maintain OPERABILITY, implementation mechanism (procedure, temp mod. etc.). and under what conditions the Compensatwy Measures may be terminated.)

The PBNP Unit 1 Reactor Vessel Is fully capable of performing all of its required safety and safety support functions. No compensatory measures are required.

I if the SSC is not fully capable of performing its safety or safety support function(s), then provide an Aggregate Review of the condition. identify related Action Requests (CAP numbers).

The PBNP Unit 1 Reactor Vessel is fully capable of performing all of its required safety and safety support functions.

Equipment recommended to be:

C Operable

Operable, But Degraded

/Operable, But Nonconforming

. inoperable (Notify Shift Manager immediately) 0.

Responsible i

Date: Z ] t!j[&~xt:

t h Verifier:

DS~: 2 j : b ~ x t : 7314 Approval Recommendation Cognizant ngi wing Manager:

l D a t e : x & &

eLs!&z%

? P ib-('-

I Shift Manager

/ 7 7 7 ~ ~ !' 1;. ' -

./

QF-1100 Rev 2 (FP-OF-OL-01)

I identify references used. (Reference Name and Section (s))

I

1. PBNP Technical Requirements Manual Section 2.2 'Pressure Temperature Limits Report Unit 1 and 2," Revision 1

2. Westinghouse Electric Company Letter LTR-REA-04-64 "Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2," dated June 2004

3. PBNP Calculation 2000-0001 'RCS Pressure-Temperature Limits and LTOP Setpoints Applicable Through 32.2 EFPY - Unit 1 and 34.0 EFPY - Unit 2." dated March 6,2000

4. WCAP-12794, "Reactor Cavity Neutron Measurement Program for Point Beach Unit 1,"

Rev. 4. February 2000

5. WCAP-12795, 'Reactor Cavity Neutron Measurement Program for Point Beach Unit 2,"

Rev. 3, August 1995

6. NPL 2000-0510 'Response to Request for Additional Information; Supplement 1 to Technical Specifications Change Request 21 9 Adoption of Pressure and Temperature Limits Report and Revised P-T and LTOP Limits Point Beach Nuclear Plant, Units 1 and 2." dated November 20,2000.

7. CR 00-2655 'Validity of PIT Heatup and Cwldown Curves," dated 1/9/2001.

8. PBNP Cakulation 2000-0001-00-A 'Evaluation of P-T Limits and LTOP Applimbility Date." dated October 10, 2000

9. NRC Safety Evaluation 'Final Safety Evaluation for Westinghouse Owners Group Topical Report WCAP-16083-NP, Revision 0. 'Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry." Dated January 10.

2006

10. Westinghouse Electric Company Letter WEP 06-13 'Statistical Evaluation of Reactor Vessel Dosimetry - Point Beach Units 1 and 2." dated February 14.2006.

Continuation.

ENCLOSURE 2b SUPPLEMENT TO LICENSE AMENDMENT REQUEST 251 POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 OPERABILITY RECOMMENDATION OPR000175, REVISION 1 7 Pages Follow

QF-1100, Revision 3 (FP-OP-OL-01)

PagelOr7 PBNP Units 1 and 2 use a common set of Pressure-Temperature (P-T) Limit Curve and LTOP setpoints.

These curves and setpoints are documented in Technical Requirements Manual (TRM) Section 2.2 'Pressure Temperature Limits Report" (Reference 1). The applicability of TRM 2.2, and its induded curves and setpoints, are defined by speciiic limits of neutron fluence. The neutmn fluence limits of TRM 2.2 are exceeded on several PBNP Unit 1 material locations using the most recent neutron exposure calculations (Reference 2).

The neutron fluence limits for PBNP Unit 2 have not been exceeded.

The scope of this operability determination is to evaluate this condition as it relates to the operability of the PBNP Unit 1 reactor vessel. Revision 0 of the operability recommendation calculated expiration date of the PTLR based on the 32.2 EFPY limit stated for the PTLR. This new revision calculates expiration based on the fluence projection for the limiting component and is more conservative.

The condition was discovered during revision of the Pressure Temperature Limit Report and basis on February am, m.

Describe the specified safety function(s) of the SSC. Identify the Licensing Basis functions and performance requirements, including Technkal Specifications, FSAR, NRC CommHmeWs, or ather appropriate informaion (re-SCOPE d o n 5.3).

Final Safety Anatysis Report The reactor vessel is part of the reactor coolant system. As stated in FSAR Section 4.1 Weactor Coolant System,' the reactor coolant pressure boundary shall be designed, fabricated. and constructed so as to have an exceedingty low prubabili of gross rupture or significant uncontrolled leakage throughout its design lifetime.

FSAR Section 1.3.6 'Reactor Coolant Pressure Boundarf discusses that the pressure-temperature limits are determined in accordance with the methods of analysis and the margins of safely of Appendix G of ASME Code Section XI and are induded in the PBNP Pressure Temperature Limits Report (PTLR).

More specifically, the reactor vessel and the pressure-temperature limits must comply with General Design Criterion 34 which states:

The reactw coolant pressure boundary shall be designed and operated to reduce.to an acceptable level the probabilily of rapidly m a t i n g type failures. Corrsideration is given (a) to the provisions for control owx service temperature and irradiation effects which may require operational restrictions. @) to the design and mtructian of the reactcK pressure d in accordance with applicable codes, induding those w h i i establish requirements for absorption of energy within the elastic strain energy range and for absorption of energy by plastic deformation and (c) to the design and construction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes.'

The reactor vessel must also comply with General Design Criterion 36 'Reactor Coolant Pressure Boundary Surveillance' which states:

'Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming with current applicable d e s shall be provided.'

Form retained in accordance with record retention schedule identified in FP-G-RM-01

QF-1100, Revision 3 (FP-OP-1)

The reactor coolant pressure boundary is designed to reduce to an acxeptaMe level the probability of a rapidly propagating type failure. The fracture toughness of the materials in the beltline region of the reactor vessel will decrease as a result of fast neutron irradiation induced embrittkment. The decrease in fracture toughness, which is a function of several factors, induding accumulated fast neutron fluence, requires a corresponding increase in reference nil ductility temperature (RTm) in order to maintain strengthlstress requirements. This change in material properties is factored into the operating procedures such that the reactor coolant system pressure is limited with respect to RCS temperature during plant heatup, cooldown, and normal operation.

These limits are determined in accordance with the methods of analysis and the margins of safety of Appendix G of ASME Code Section XI and are induded in the PBNP PTLR. The Low Temperature Overpressure Protection System provides protection during low-temperature operations. All pressure containing components of the Reactor Coolant System are designed, fabricated, inspected, and tested in conformance with the applicable codes at the time of order placement The operating pressure temperature limits for the reactor vessel are located in Technical Requirements Manual (TRM)

Section 2.2 'Pressure Temperature Limits RepoK (Reference 1).

Technical Specifications Technical Specification 3.4.3 'RCS Pressure and Temperature (PIT) Limits-state that RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits speufied in the PTLR Technical Specitication 3.4.6, RCS Loops -

MODE 4,' and Technical Spedfication 3.4.7, WCS Loops -

MODE 5, Lwps Fill& require that no RCP shall be started with any RCS cdd leg temperature less than or equal to Low Temperature Overpressure Protection (LTOP) enabling temperature specified in the PTLR unless the secondary side water temperature of each steam generator (SG) is less than or equal to 50 T above each of the RCS d d leg temperatures.

Technical Specitication 3.4.10 'Pressurizer Safety Valved requires the plant to be in mode 4 with the RCS cold leg less than or equal to the LTOP enabiing temperature specrfied in the PTLR if required action and action time not met for one safety valve inoperable or if two pressurizer safety valves are inoperable.

Technical Speakation 3.4.12 l o w Temperature Overpressure Protection (LTOP) Systemg is required to be operable when in mode 4 with the RCS cold leg less than or equal b the LTOP enabling temperature specified in the PTLR. The speufication also requires accumulators to be depressurized or RCS system temperature raised if required action and action time not met to isdate affected accumulator.

Technical Specification 5.6.5 'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)" discusses repwting requirements of the PTLR and states that the analytical methods used to determine the RCS pressure and temperature limits &ail be those previously reviewed and approved by the NRC, specifically those desaibed in the NRC Letters dated October 6,2000 and July 23,2001.

NRC Commitments NRC Letter Point Beach Nudear Plant, Units 1 And 2 -Acceptance of Methodology For Referencing Pressure Temperature Limits Report (TAC NOS. Mf48459 AND MA8460) dated July 23,2001 (Point Beach reference NPC 2001-01263) - This NRC document requires that PBNP will not utilize the FERRET code until staff approval is received, and FERRET code additions to the PTLR will be submitted to the NRC. The document also states that the that Point Beach's predicted expiration of the PnR was October 30,2003, for Unit I, and October 1,2008, for Unit 2.

NMC Letter NRC 2003-0108 - Addresses the PT Curves applicability discussed in NRC Letter dated July 23, Form retained in accordance with record retention schedule identified in FP-G-RM-01

QF-1100. Revision 3 (m-0P-OL4f)

Page 3 d 7 CAP:

OPR:

000175 that they would revise the Pi? cum prior to exceeding the 25.59 EFPY for Unit 1.

NRC Safety Evaluation dated November 29,2002, 'Point Beach Nudear Plant, Units 1 And 2 - Issuance of Amendments RE: Measurement Uncertainty Recapture Power Uprate PAC Nos. MB4956 And MB4957) (Point Beach reference SER 2002-001 0).

Safety Evaluation (SE) with Open Items Related to the License Renewal of Point Beach Nuclear Plant, Units 1 and 2, dated May 2,2005 condudes that PBNP's methods of fluence calculation camply with the requirements of NRC RG 1.190, 'Calculational and Dosimetry M o d s for Determining Pressure Vessel Neutron Fluence.'

(Point Beach refereme SER 2035-0006)

NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials" specifies method used to predict the effect of neutron radiation on reactor vessel materials unless they can justify the use of diiecent methods. This regulatory guide is referenced in the above SE.

I NRC Regulations I

10CFR50 Appendix G 'Fracture Toughness Requirements-specifies fracture toughness requirements for fem'tic materials of pressure-retaining amponents of the reador coohnt pressure boundary to provide adequate margins of safety during any condition of normal operation.

10CFRSO Appendix H 'Reactor Vessel Material Surveillance Program Requirements' describes material surveillance program requirements to monitor ctianges in the fracture toughness properties of fenitic materials in the reactor vessel bertline region which result from exposure of these materials to neutron irradiation and the thermal environment I

10CFR50.61 'Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Eventsg states limit and requirement for calculations of End Of Life (EOL) RT, values.

I Evaluate the effects of the condition, Including potential failure modes, on the ability of the SSC to perform its specified safety function(s)

Exceeding limits of PT Curve fast neutron fluence would decrease the predicted fracture bughness of the reactor vessel components. An u n a w e d decrease in fracture toughness. without corresponding changes to allowed RCS pressure and temperature, could jeopardize the ability of the reador vessel to perform its safety and support fundkm.

(Use engineering analysis or engineering judgment to determine whether the specified safety function can be provided given the existence of the deficiency. When using engineering judgment, provide supporting information from sources such as fieki walkdowns, industry experience, proven systern/ampw&

performance under similar m-ce conditions, etc.)

Some of the PBNP Unit 1 PTLR fluence values have been exceeded using the accepted fiuence calculation methodology stated in the PTLR. However, the PBNP Unit 1 Reactor Vessel is fulty capable of performing all of its required safety and safety support functions. As desaibed below, using recently approved fluence calculation methodology (FERRET Code) and !he most recent fluence calcolations, demonstrate that PBNP Unit 1 fluence limits are not exceeded for limiting components. The Unit 1 vessel best estimate fluence values Form rebimed in accordance with record retention schedule identified in FP-G-RM-Ol

QF-1100. Revision 3 (FP-OP-OL-01)

Page 4 of 7 will remain within limits until an EFPY of 29.5 is reached.

PBNP Calculation 20004001 (Reference 3) performed the calculation of the P-T curves and LTOP setpoints that are contained in TRM 2.2 (or the Pressure Temperature Limits Report (PTLR)). This calculation used what is termed "best-estimate' fluence calculations as a design input 'Best Estimate' refers to the method of correcting calculated neutron fluence calculations for plant s p d c biases due to plant-specific core geometry.

A Westinghouse code termed FERRET performed this 'best estimateg calculation (References 4 8 5).

During the approval process of the PBNP PTLR in 2000, the NRC recognized that Westinghouse's FERRET fluence code had not received formal approval (Reference 6 8 7). In response to NRC RAls, PBNP Calculation Addendum 2000-0001-00-A (Reference 8) was prepared to state that only calculated (vice 'best estimate" or FERRET code) neutron fiuence values were to be used. Additionaily, the predicted applicable EFPY values of the curves were calculated (Reference 8) based on the calculated values of fluence to account for the conservative action of not using the FERRET code. The PTLR curve remained the same. An EFPY value based on the calculated values of fluence and an EFPY based on the best estimate fluence are both contained in the PTLR We are committed to using the calculated value. The calculated values without the FERRET code result in Unit 1 exceeding the established allowed fluence for the PTLR.

NRC approval of the FERRET code was received by the Westinghouse Owners Group on January 10.2006.

The NRC SE (Reference 9) states that the FERRET code is acceptable to be referenced if the licensee's analyses comply with the uncertainty values appr~ved by the SE. Westinghouse letter WEP 06-13 (Reference

10) documents that the uncertainty values used by the PBNP fluence cdculations (References 4 8 5) compty with the requirements of the NRC SE on FERRET code. Therefore, use of FERRET code fiuence values in the PBNP PTLR is acceptable for confirming operability. Use of the FERRET code is currently not part of our licensing basis and the condition is nonconforming.

The latest PBNP calculations using the FERRET code adjustment (Reference 2) show that the fluence values are within the fluence limits established in the PTLR calculation for limiting components (Reference 3).

Unit 1 -

As stated in the PTLR and calculation 2000-0001-00-A, the limiting Unit 1 component in the generation of the PT Curves is the Lower Shell Long Seam Weld (SA-847). The accumulated fluence limit (i.e.. inside surface fluence value in the PTLR) for this component is 1.55 El9 n l d. Table 4-10 of the current fluence calculation FERRET code adjusted best estimate (Reference 2) list the predicted best estimate fluence values for the Lower Shell Long Seam Weld at the present plant conditions (i.e., no full power uprate and with hafnium rods).

Interpolation of this table reveals that the fluence value for weld SA-847 will be reached at 29.5 EFPY as shown below.

Unit 2 -

As stated in the fTLR and calculation 2000MlOi-00-A, the limiting Unit 2 component in the generation of the PT-Curves is the Intermediate to Lower Circumferential Weld (SA-1484). The accumulated fluence limit (i.e., inside surface fluence value in the PTLR) for this cwnponent is 2.606 El9 n l d. Table 4-18 of the L i m i Component 32.2 EFW Inside 27.0 EFPY Withwt 29.7 EFW Wrthout ln@polation to 29.5 Description Surface Fluence Llpate Wilh Hafnii Uprate With W u n EFW W w

Uprate

[El9 n l d )

Rods (~19nlcm2)

~ a l s ( ~ ? 9 & ]

WlthHaftrirrmRods

-a 1

(Reference 2)

(Reference 2)

(El 9 nh2)

Lower Shell Longitudinal Seam Weld 1.55 1.45 1.56 1.55 61 782 (SA-847)

Form retained in accordance with record retention schedule identied in FP-G-RM-01

QF-1100. Revision 3 (FP-OP-OL-01)

Page 5 of 7 I

for the Intermediate to Lower ~ircumferentiai~eld at the present plant conditions (i.e.. no full power uprate and with hafnium rods). Intermlation of this table reveals that the fluence value for weld SA-1484 will be reached at I The adjusted fluence for some Unit 1 components is above or will be above that assumed in the PTLR calculation prior to 29.5 EFPY, however, these components are not limiting. The Unit 1 PT curves are based on the controlling material, which for Unit 1 is the Lower Shell Longitudinal Seam (61782 (SA-847)) per the PTLR and calculation 2000-0001 A.

As stated previously, PBNP uses mrnon PTLR and PT-limit curves. Therefore, the limiting value of 29.5 EFPY sets the applicability date for the PBNP PTLR. The accumulated EFPY value for PBNP Unit 1 on June 1,2007 was 28.41 EFPY. The difference between the current EFPY and limiting EFPY is approximately 1 year. Therefore, the existing PTLR, with the existing PT curves and LTOP sew~nts, is valid until at least June 2008. As such, the PBNP Unit 1 Reactor Vessel can support its specified safely, and safety support, functions until a stated expiration date of June 1,2008. PBNP will revise the PTLR curves based on approved methodology prior to expitation of this date.

Inteqmtation to 34.7 EFW Without Uprate with Hafnium ~ o d s (El9 n l d )

2.606 Form retained in accordance with record retention schedule identified in FPGRM-01 Limiting Component 34.0 EFPY Inside Description Surface Fluenca (El9 n/ad)

(Reference 1) 34.0 EFPY Without Uprate With Hafnium

~ o d s (El9 n/an2)

(Reference 2) 2.57 Intermediate to Lower Shell Circumferential Weld 72442 (SA-1484) 36.0 EFW Without Upmte with ~ a f n i Rods (El 9 nlcm2)

(Reference 2) 2.68 2.606

QF-1100. Revision 3 (FP-OP-OL41)

Page 6 of 7 modifications as the result of an OPR SHALL be reviewed in accordance with 10 CFR 50.59.

(Describe the Compensatory Measures, basis for which the Compensatory Measures maintain OPERABILITY, implementation mechanism (procedure, temp mod, etc.), and under what conditions the Compensatory Measures may be terminated.)

The PBNP Unit 1 Reactor Vessel is capable of performing all of its required safety and safety support functions using the best estimate adjusted fluence values. No compensatory measures are required.

The fluence values assumed in the generation of the PTLR curves have been exceeded based on NRC approved methodology for PBNP Unit I. The fluence values are not exceeded for the limiting component when the fluence is adjusted using the FERRET code. The FERRET code has been approved generically by the NRC. The FERRET code has not been approved for use by PBNP and the condition is non conforming.

A review of other operability recomdatians revealed no other pertinent issues.

Inoperable (Notify Shift Manager lrnmediatety)

[7 Outside the scope of the Operability Process Responsible Engineer:.-

P && dLc;a Date:1=/11\\2~4Ext L - G Y ~

Verifier.

&PYX I;& rbn/ f i y ~ f l ?

r w 4 oak:

+dm 7

~

3 Approval Recommendation Cognizant Engineering Manager:

Date: L/J 6 & P Date and Time:

Identify References Used (Reference Name and Section(s)):

1. PBNP Technical Requirements Manual Section 2.2 'Pressure Temperature Limits Report Unit 1 and 2,' Revision 1

2. Westinghouse Uectric Company Letter L T R - R E A M 'Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2,' dated June 2004

3.

PBNP Calculation 2000-0001 'RCS Pressure-Temperature Limits and LTOP Setpoints Applicable Through 32.2 EFPY - Unit 1 and 34.0 EFPY - Unit 2.' dated March 6.2000

4. WCAP-12794, 'Reactor Cavity Neutron Measurement Program for Point Beach Unit 1," Rev. 4, February 2000 Form retained in accordance with record retention schedule identified in FP-G-RMU1

OF-1 100. Revision 3 (FP-OP-OL-01)

Page 7 of 7 OPERABtW RECOMMENDATION

7. CR 00-2665 "Validity of P/l Heatup and Cooldown Curves,' dated 1/9/2001.

8. PBNP Calculation 2000-0001-00-A 'Evaluation of P-T Limits and LTOP Applicability Date,' dated October 10,2000 Form retained in accordance with record retention schedule identified in FP-G-RM-OI

ENCLOSURE 3 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 251 POINT BEACH NUCLEAR PLANT UNITS I AND 2 TRM 2.2, REVISION 1, MARKED UP PAGES 16 Pages Follow

TRM PRESSURE TEMPERATURE LIMITS REPORT UNIT 1 AND UNIT 2 REVISION 1 (Effective through 25.59 EFPY for Unit 1)*

(Effective through 30.51 EFPY for Unit 2)*

Note*: Applicability limits for pressure temperature Iirnits are discussed in Section 2.0, "Operating Limits."

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 1.O RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This RCS Pressure and Temperature Limits Report (PTLR) for Point Beach Nuclear Plant Units 1 and 2 has been prepared in accordance with the requirements of Technical Specification 5.6.5. Revisions to the PTLR shall be provided to the NRC upon issuance.

The Technical Specifications addressed in this report are listed below:

1.1 3.4.3 PressurefTernperature (P-T) Limits 1.2 3.4.1 2 Low Temperature Overpressure Protection (LTOP) System The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. All changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.5.

These limits have been determined such that all applicable limits of the safety analysis are met. All items that appear in capitalized type are defined in Technical Specification 1.1, "Definitions."

All EFPY values listed in this procedure are estimates based on reactor power of 15 18.5 MWt. Applicability of the operating limits are determined by accumulated fluence values listed in Tables 3 and 4. This report will be revised with new P-T limits prior to exceeding the associated fluence values.

2.1 RCS Pressure and Temperature Limits (LC0 3.4.3')

2.1.1 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°F in any one hour.

b. A maximum cooldown rate of 100°F in any one hour.

c. An average temperature change of I 1 0°F per hour during inservice leak and hydrostatic testing operatiom.

2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively (includes instrument uncertainty).

2.1.3 The minimum temperature for pressurization or bolt up, using the methodology, is 60°F, which when corrected for possible instrument uncertainties is a minimum indicated RCS temperature of 78°F (as read on the RCS cold leg meter) or 70°F using the hand-held, digital pyrometer.

POINT BEACH TRM

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 2.2 Low Temperature Overpressure Protection System Enable Temperature (LC0 3.4.6. 3.4.7, 3.4.10 and 3.4.12)

The enable temperature for the Low Temperature Overpressure Protection System is 270°F (includes instrument uncertainty for RCS T, wide range).

2.3 Low Temperature Ovemressure Protection System Setpoints (LC0 3.4.121 Pressurizer Power Operated Relief Valve Lift Setting Limits The lift setting for the pressurizer Power Operated Relief Valves (PORVs) is 1500 psig (includes instrument uncertainty).

3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedules for Units 1 and 2 are provided in Tables 1 and 2, respectively.

The pressure vessel surveillance program is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the nil-ductility temperature, RTN~T, which is determined in accordance with ASTM E208. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveilIance capsuIe removal scheduIe meets the requirements of ASTM E185-82.

Surveillance specimens for the limiting materials for the Point Beach reactor vessels are not included in the plant specific surveillance program. Therefore, the results of the examinations of these specimens do not meet the credibility criteria of USNRC Regulatory Guide 1.99, Rev. 2 for Point Beach Nuclear Plant, Units 1 and 2.

POINT BEACH TRM

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT

5.0 REFERENCES

5.1 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 2, January 1996 5.2 WCAP-12794, "Reactor Cavity Neutron Measurement Program for Point Beach Unit 1," Rev. 4, February 2000 5.3 WCAP-12795, "Reactor Cavity Neutron Measurement Program for Point Beach Unit 2," Rev. 3, August 1995 5.4 EPRI TR-107450, "P-T Calculator for Windows, Version 3.0," Revision 0, December 1998 5.5 Westinghouse Report, "Pressure Mitigating Systems Transient Analysis Results,"

July 1977 5.6 Westinghouse Report, "Supplement to the July 1977 Report, Pressure Mitigating Systems Transient Analysis Results," September 1977 5.7 Wisconsin Electric Calculation 2000-0001, Revision 0, RCS P-T Limits and LTOP Setpoints Applicable through 32.2 EFPY - Unit 1 and 34.0 EFPY - Unit 2 5.8 Wisconsin Electric Calculation 2000-000 1 A, Revision 0, Evaluation of P-T Limit and LTOP Applicability Date 5.9 ASME B&PVC Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division 1 "

5.10 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Exemption fiom the Requirements of 1 OCFR50.60 (TAC NOS. MA9680 and MA968 I)", dated October 6,2000

5. 1 1 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure Temperature Limits Report (TAC NOS.

MA8459 and MA8460)", dated July 23,2001

5. 12 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: The Conversion to Improved Technical Specifications (TAC NOS. MA7186 and MA7187)", dated August 8,2001 POINT BEACH TRM

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.1 The RTpTs values for the Point Beach Nuclear Plant limiting beltline materials is 278°F for Unit 1 and 291°F for Unit 2 at 32 EFPY.

4.2 Tables POINT BEACH TRM Table Number Table 1 Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Table 8 Table Description Point Beach Nuclear Plant, Unit 1 Reactor Vessel Surveillance Capsule Removal Schedule Point Beach Nuclear Plant, Unit 2 Reactor Vessel Surveillance Capsule Removal Schedule Point Beach Unit 1 RPV Beltline 25.59 EFPY Fluence Values Point Beach Unit 2 RPV Beltline 30.5 1 EFPY Fluence Values Point Beach Unit 1 RPV 114t Beltline Material Adjusted Reference Temperatures at 25.59 EFPY Point Beach Unit 2 RPV 114t Beltline Material Adjusted Reference Temperatures at 30.5 1 EFPY Point Beach Unit 1 RPV 3/4t Beltline Material Adjusted Reference Temperatures at 25.59 EFPY Point Beach Unit 2 RPV 314t Beltline Material Adjusted Reference Temperatures at 30.5 1 EFPY

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 5.13 NMC License Amendment Request 25 1, dated December 14,2006 (NRC 2006-0090), Technical Specification 5.6.5, Reactor Coolant System Pressure and Temperature Limits (Application for use of FERRET Code as approved methodology for determining RCS pressure and temperature limits) 5.14 NRC SE dated XX/XX/XX issuing Amendment Nos. XXIXX to Facility Operating Licenses DPR-24 and DPR-27, approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)

POINT BEACH TRM

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT FIGURE 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP Moderator Temperature PF) 2500 -

I I

I v:

LLAK TL8T LlMK UNACCEPTABLE OPERATION -ABOVE AND/OR TO LEFT OF CURVE8 ACCEPTABLE OPERATION -BELOW ANDlOR TO RlOlMT OF CURVE8 2000 --

LTOP 8lTPOlNT8 INCLUDE CORRECTION8 FOR IN8TRUMINT UNCERTAINTY LTOP ENABLE TEMPCRATURE

• POINT BEACH TRM 2.2 - 7 500 -

0 -

50 100 4 5 0 200 250 300 3M 400 MAXIMUM LTOP 8CTPOINT I

1 1

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT FIGURE 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN Moderator Temperature PF)

POINT BEACH TRM

\\

400 2500 -

2000 3

8 1600--

I I fopo-1 4

wo 0 T 50 I00 150 200 250 so0 360 I

UNACCiPrABLIE OPERATION -ABOVE ANDlOR TO LEFT OF CURVE ACCEPTABLE OPERATION - BELOW AND/OR TO RKIHT OF CURVE LTOP 8ElPOINTS INCLUDE CORRECTIONS FOR INSTRUMENT UNCERTAINTY r

COOLDOWN LIMIT - CORE NON- -

- LTOPENABLE CRITICAL UP TO 100*FMr TEMPERATURE r

1 MAXIMUM LTOP 8ET'POINT I

I

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 1 POINT BEACH NUCLEAR PLANT UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE

• The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

Capsule Identification Letter V

S R

T P

N TABLE 2 POINT BEACH NUCLEAR PLANT UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Approximate Removal Date*

September 1972 (actual)

December 1975 (actual)

October 1977 (actual)

March 1984 (actual)

April 1994 (actual)

Standby

• The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

Capsule Identification Letter V

T R

S P

N POINT BEACH TRM Approximate Removal Date*

November 1974 (actual)

March 1977 (actual)

April 1979 (actual)

October 1 990 (actual)

June 1997 (actual)

Standby

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT 1 RPV BELTLINE 25.59 EFPY'" $talc (32.2 EFPY $Bcst.Est,) VALUES'~)

Based on WCAP-12794, "Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit I," Rev. 4, February 2000. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 32 EFPY, due to changes in core design at certain points in the operating history of the unit. As intermediate input to hrther calculations, these values are not rounded in accordance with ASTM E29 (Ref. 1 I).

'"' Interpolation of neutron cxposurc (in units of El9 n/cm2, E>1 MeV) to a particular valuc of cffcctivc full powcr ycars (EFPY) is pcrformcd bascd on WCAP-12794, Revision 4. For examplc, for thc nozzlc bclt forging, hcat no. 122P237, Vessel Manufacturer:

I Babcock & Wilcox Plate and Weld Thickness (without cladding):

1 6.5", without clad @)

flucncc = 0.547 + l 0.796 - 0.547

) x (32.2 EFPY - 32.0 EFPY) = 0.550 El9 n/cm2 (8 EFPY - 32 EFPY)

From an insidc surface flucncc valuc (not including cladding), fluencc is attcnuatcd to a dcsircd thickness using equation (3) of Regulatory Guidc 1.99, Rcvision 2: f = f,,t x c-024x

, where f.,,r is expressed in units of El9 n/cm2. E>1 MeV, and x is thc dcsircd dcpth in inches into the vcssel wall. For cxamplc, for thc nozzlc bclt forging, hcat no. 122P237, at 32.2 EFPY, at a dcpth of 114 of thc 6.5" vcsscl wall (1.625"), f = 0.550 x c~0.2*l.62J1

= 0.3724 El9 n/cm2.

32.2 EFPY h e r t. ~ s t.

314T Fluence Factor "'

0.5322 0.9452 0.8993 0.5322 N/ A 0.8293 0.8993 0.7960 Component Description Nozzle Belt Forging Intermediate Shell Plate Lower Shell Plate Nozzle Belt to Intermed. Shell Circ Weld (100%)

Intermediate Shell Long Seam (ID 27%)

Intermediate Shell Long Seam (OD 73%)

Intermed. to Lower Shell Circ.

Weld (1 00%)

Lower Shell Long Seam (I 00%)

("

Thc dimcnsionlcss flucncc factor is calculated using the flucncc factor formula from cquation (2) of Rcfulato Guidc 1.99, Rcvision 2: ff = f0~2"00.'0'oBo

, whcrc f is the flucncc in units of El9 n/cm2. For cxamplc, the 32.2 EFPY INT flucncc factor for noulc belt forging, hcat no. I22P237. If= 0.3724(O.'"

"'w0,y24) = 0,7269.

(D' Instruction Manual, 132-lnch I.D. Reactor Prcssurc Vcsscl, Babcock & Wilcox, Scptcmbcr 1969 (Rcf. 12).

Pcr Wisconsin Elcctric Calculation 2000-001 A thc calculatcd flucncc for the critical material (SA-847) occurs at 25.59 EFPY vcrsus 32.2 EFPY I$B.,,.E~, bascd upon K=I$B~cr~.Em./4Calc=0.838.

32.2 EFPY b c a t. ~ s t.

314T Fluence (E 19 nlcmz) ("

0.1707 0.8225 0.6983 0.1707 N/ A 0.543 1 0.6983 0.48 1 1 (F'

EFPY value listcd hcrc is bascd on a reactor powcr of I5 18.5 MW,. Scc Section 2.0, "Operating Limits," for discussion of applicability dates.

Heat or HeatILot 122P237 A981 1-1 C1423-1 8T1762 (SA-1426) 1P08 (SA-8 12)

P066 (SA-775) 249 (SA-I 01 )

6 1782 (SA-847)

POINT BEACH TRM 2.2 - 10 32.2 EFPY h e s t. ~ s t.

Inside Surface Fluence (El9 n/cm3) "

0.550 2.65 2.25 0.550 1.75 1.75 2.25 1.55 32 EFPY h r g t. ~ e t.

Inside Surface Fluence (El9 nlcmz) 0.547 2.64 2.24 0.547 1.74 1.74 2.24 1.54 32.2 EFPY 6eest.~st.

114T Fluence (El9 nlcmz) 'B' 0.3724 1.794 1.523 0.3724 1.185 N/ A 1.523 1.049 32.2 EFPY heat.&st.

114T Fluence Factor (')

0.7269 1.160 1.116 0.7269 1.047 NIA 1.1 16 1.013

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BELTLINE 30.5 1 EFPY" (bcalc (34.0 EFPY 4 ~ ~ ~ t. ~ s t. )

VALUES'~)

Based on WCAP-12795, "Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2," Rev. 3, August 1995. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 32 EFPY, due to changes in core design at certain points in the operating history of the unit. As intermediate input to further calculations, these values are not rounded in accordance with ASTM E29 (Ref. 11).

Footnotes; Vessel Manufacturer:

Plate and Weld Thickness (without cladding):

"' Interpolation of neutron exposure (in units of El9 n/em2, E71 MeV) to a particular value of effective full power years (EFPY) is performed based on WCAP-12795, Revision 3. For example, for the nozzle bclt forging, heat no. 123~352, Babcock & Wilcox and Combustion Engineering 6.5", without clad ID) fluencc = 0.548 + ( 0.784 - 0.548

) x (34 EFPY - 32 EFPY) = 0.5775 El9 n/cm2 (48 EFPY - 32 EFPY)

From an inside surface fluence value (not including cladding), flucnce is attenuated to a desired thickncss using cquation (3) of Regulatory Guide 1.99, Revision 2: f = f,,f x e-0.24x, where f.,,r is expressed in Component Description Nozzle Belt Forging Intermediate Shell Forging Lower Shell Forging Nozzle Belt to Interned. Shell Circ Weld (1 00%)

Intermed. to Lower Shell Circ Weld (1 00%)

li units of El9 n/cmz, E>1 MeV, and x is the desireddepth in inches into the vessel wall. For cxamplc, for the nizzie belt forging, hcat no. -123~352, at 34.0 EFPY, at a depth of 114 of the 6.5" vessE~ wall (1.62S1'), f = 0.5775 x e-0.2q'.62" = 0.3910 El9 n/cm2.

(

The dimensionless fluence factor is calculated using the flucnce factor formula from equation (2) of Replato Guide 1.99, Revision 2: ff = $0.2R-0.10'ogo, where f is the fluence in units of El9 n/cm2. For cxamplc, the 34.0 EFPY 114T fluence factor for nozzle belt forging, heat no. 123V352, ff = 0.3910'0.2n'.lO1qO '

= 0.7399.

mi Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970 (Ref. 13).

'"' Per Wisconsin Electric Calculation 2000-001-00-A the calculated fluenee for the critical material (SA-1484) occurs at 30.51 EFPY versus 34.0 EFPY i$&,,.~.~based upon K=+B~or.Em./i$Cale=O.92l.

" EFPY value listed here is based on a reactor power of 15 18.5 MW1. See Section 2.0, "Operating Limits," for discussion of applicability dates.

Heat or HeatLot 3 23V352 123V500 122W195 2 1935 72442 (SA-1484)

POINT BEACH TRM 2.2 - 11 32 EFPY

+ ~ e s t. ~ s t.

Inside Surface Fluence (El 9 n/cm2) 0.548 3.01 2.52 0.548 2.49 34.0 EFPY 4Best.~st.

Inside Surface Fluence (El9 n/cm2) 'A' 0.5775 3.174 2.654 0.5775 2.606 34.0 EFPY

$Best.Est.

1/4T Fluence (El9 n/cm2) ")

0.3910 2.149 1.797 0.3910 1.764 34.0 EFPY

~Bert.Est.

1/4T Fluence Factor ")

0.7399 1.208 1.161 0.7399 1.156 34.0 EFPY

$Best.~st.

3/4T Fluence (El9 n/cm2) IB) 0.1792 0.985 1 0.8237 0.1792 0.8088 34.0 EFPY

~BestEst.

3/4T Fluence Factor '"

0.5435 0.9958 0.9456 0.5435 0.9405

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 114T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 25.59 EFPY") $talc (32.2 EFPY (b~est.Est.) (H)

Unless othcnvisc noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Prcssurc Vessel Integrity," May 1998 (Ref. 14). including thc most recent best-estimate chcrnistry values for welds, applying currcnt B&WOG mean-of-the-sources approach. All beltline materials arc included for comparison.

Faatnotcs; (A)

See Table 1 Vessel Manufacturer:

I Babcock & Wilcox Plate and Weld Thickness (without cladding):

( 6.5", without clado In)

Credible Surveillance Data; see BAW-2325 for evaluation.

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure ARTNDT and predicted ARTNw based on Table CF is less than 2 0 (56°F).

(D" Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Hcat 61782, Point Beach Unit 1," (Ref. 15) utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.

ARTNDT (OF) 55.97 102.08 9 1.99 61.71 161 Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNqT + ARTNDT + Margin, where ARTNDT

= Chemistry Factor x Fluence Factor, and Margin = 2(aI2 +

a d )'", with UI defined as the standard deviation of the Initial RTNDT and a&

defined as thc standard dcviatlon of ARTNDT. For example, for nozzle belt forging, heat no.122P237, ART = 50 + (77 x 0.7269) t 34 = 140°F. Calculated ART values are rounded to the noaxst "F in accordance with thc rounding-off method ofASTM Practice E29.

0 26.9 9,

26.9

!I Nozzle Belt to Intermed. Shell Circ Weld (100%)

Intermediate Shell Long Seam (ID 27%)

Intermediate Shell Long Seam (OD 73%)

Intermed. to Lower Shell Circ. Weld (1 00%)

Lower Shell Long Seam (100%)

I!

Component Description I Nozzle Belt Forging Intermediate Shell Plate

,I Lower Shell Plate (F'

Instruction Manual, 132-Inch I.D. Reactor Prcssurc Vessel, Babcock & Wilcox, September 1969.

=A 17 17 8.5 17 Margin (OF)

II I f 8T1762 (SA-1426)

-5 0.19 0.57 152.4 1 PO8 15 (SA-8 12)

-5 0.17 0.52 138.2 1 PO66 1 (SA-775) 7 1249 (SA-1 10 1) 6 1782 (SA-847)

I!

Heat or HeatJLot 122P237 A981 1-1

,I C1423-1 By inspection, these are the limiting material properties.

ART OF)(^)

Initial R T N ~ T (OF)

+50

+I 9,

+ 1

'" Per Wisconsin Electric Calculation 2000-001-00-A the calculated fluence for the critical material (SA-847) occurs at 25.59 EFPY versus 32.2 EFPY I+8e.t.Ea, based upon K=$B~crr.Ecr~./I+Calc=0.838.

34 1

140 63.64 1

167 56.42 1

149 63.64 1 1 2 6

%Cu 0.1 1 0.20 0.12

('I EFPY value listed here is based on a reactor power of 15 18.5 MWt. See Scction 2.0, "Operating Limits," for discussion of applicability dates 114T 32.2 EFPY best.~st.Fluence Factor(*)

0.7269 1.160

!I 1.116 POINT BEACH TRM 2.2 - 12 CF Method Table Table Surv. DatafnJ Table

%Ni 0.82 0.06 0.07 CF 77 88 79.3 55.3

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV 114T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 30.51 EFPY'~' $talc (34.0 EFPY +~cst.~st.)

(1)

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 (Ref. 14), including the most recent best-estimate chemistry values for welds, applying current B&EWOG mean-of-the-sources approach. All beltline materials are included for comparison.

Vessel Manufacturer:

I Babcoek & Wileox and Combustion Engineering Plate and Weld Thickness (without cladding):

1 6.5", without clad(F)

Footnotes.

(*I Scc Tablc 2.

("

Non-crcdiblc survcillancc data; see BAW-2325 for cvaluation. Tablc CF conscrvativc becausc diffcrcncc bctwccn mcasurcd ARTNDT and prcdictcd ARTNDT bascd on Tablc CF is less than 2a (34OF)

('I Crcdiblc survcillancc data; see BAW-2325 for cvaluation.

Margin (OF)

Non-crcdiblc survcillancc data; Tablc CF valuc based on best-estimate chemistry is higher than bcst fit calculatcd using survcillancc data, and thercfore, conscrvativc.

ART (OF)@)

Component Description I

Nozzle Belt Forging Intermediate Shell Forging Lower Shell Forging

$1

'"' Adjustcd rcfcrcncc tcmpcrature (ART) calculatcd pcr Regulatory Guide 1.99, Rcv. 2. ART = Initial RTND~

+ ARTNDT

+ Margin, where ARTNDT

= Chemistry Factor x Fluence Factor, and Margin = 2(a2 +

~d2)'.~, with UI defined as the standard deviation of thc Initial RTNDT, and a& dcfincd as the standard dcviation of ARTNDT. For cxamplc, for novlc belt forging, hcat no. 123V352, ART = 40 + (76 x 0.7399) +

34 = 130°F. Calculated ART valucs arc roundcd to thc ncarcst "F in accordancc with the rounding-off method of ASTM Practicc E29.

Nozzle Belt to Intermed. Shell Cire Weld 0.18 0.70 RTNOT (OF)

+40

+40

+40 I,

C F 76 58 31 42.8 170 Heat or HeatJLot 123V352 123V500 122W195 I 1

" Instruction Manual, Reactor Vesscl, Point Bcach Nuclear Plant Unit 2, Combustion Engineering, CE Book #4869, Octobcr 1970.

28 Intermed. to Lower Shell Circ. Weld (100%)

72442 (SA-1484)

-5 0.26 0.60 180 208.08 34 1

130 34 1 144(&)

34 1

110 17 1

107 (O'

By inspcction, these arc the limiting matcrial properties.

17 17 17 8.5 28

%Cu

.011 0.09 0.05 C F Method Table able'^)

Table Surv. DatafL' able'^)

19.7 ARTNDT (OF) 56.23 70.06 35.99 49.69 125.78 65.51

('" Table CF valuc bascd on best-estimate chcmistry data from CEOG Rcport "Bcst Estimatc Coppcr and Nickel Values in CE Fabricated Rcactor Vcsscl Wclds," CE NPSD-1039, Revision 2, Final Rcport, Junc 1997 (Rcf. 6).

%Ni 0.73 0.70 0.72 114T 34.0 EFPY Fluence FactorfA) 0.7399 1.208 1.161 It 0.7399 I

0 0

0 91 17 135 (o

Per Wisconsin Elcctric Calculation 2000-001 A thc calculated flucncc for the critical matcrial (SA-1484) occurs at 30.5 1 EFPY vcrsus 34.0 EFPY +&.r,~,L bascd upon K=+B~m.Em/+Calc=O.921 68.47 1 272'b)

'JJ EFPY valuc listed hcrc is bascd on a rcactor powcr of 15 18.5 MW,. Sce Scction 2.0, "Opcrating Limits," for discussion of applicability datcs.

POINT BEACH TRM

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 25.59 EFPY") $talc (32.2 EFPY $ ~ e s t. ~ s t. )

(H)

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All beltline materials are included for comparison.

IFootnoBs,

'A' See Table 1.

In)

Credible Surveillance Data; see BAW-2325 for evaluation.

Vessel Manufacturer:

I Babcock & Wilcox Plate and Weld Thickness (without cladding):

1 6.5", without clad(F)

L

-Nan-crcdible surveillance data; see BAW-2325 for evaluation. Table CF conscrvativc bccausc difference between ratio-adjusted measured L\\RTNm arc predicted ARTNDT based on Table CF is less than 2 a (56°F).

/"/

Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1," utilizing latcst time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.

ART (OF)@)

125 148 132 114 90 145 N/A 194fb7 217fb7 189 173 Ad.usted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTNDT = Chemistry Factor x Fluenee Factor, and Margin = 2(a12 +

I a,,

w~th at defined as the standard dcviation of the Initial R T N ~ ~,

and a~ defined as the standard deviation of ARTNDT. For example, for nozzle belt forging, heat no. 122P237, ART = 50 + (77 x 0.5322) +

34 = 125OF. Calculated ART values are rounded to the nearest O F in accordance with the rounding-off method of ASTM Practice E29.

'F' Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

17 17 8.5 17 8.5 28 N/A 28 28 28 14 lG' By inspection, these are the limiting material properties.

fl" Per Wisconsin Electric Calculation 2000-001 A the calculated flucnce for the critical material (SA-847) occurs at 25.59 EFPY versus 32.2 EFPY I$B,~,E~.

based upon K=I$B~crr.Em./I$Calc=0.838.

Margin (OF) 34 63.64 56.42 63.64 56.42 68.47 N/A 68.47 56 68.47 48.34 Component Description Nozzle Belt Forging Intermediate Shell Plate

,I Lower Shell Plate 80 Nozzle Belt to Interned. Shell Circ Weld

( 1 00%)

Intermediate Shell Long Seam (ID 27%)

Intermediate Shell Long Seam (OD 73%)

Interned. To Lower Shell Circ. Weld (1 00%)

Lower Shell Long Seam (100%)

I1 EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates POINT BEACH TRM 2.2 - 14 I

0 26.9 11 26.9 I,

19.7 19.7 19.7 0

19.7

%Cu 0.1 1 0.20 0.12 0.19 0.17 0.17 0.23 0.23 CF 77 88 79.3 55.3 35.8 152.4 138.2 157.6 167.6 157.4 163.3

%Ni 0.82 0.06 0.07 0.57 0.52 0.64 0.59 0.52 Heat or HeatILot 122P237 A981 1-1 I t C1423-1 II 8T1762 (SA-1426)

I PO8 I5 (SA-8 12) 1 PO66 1 (SA-775) 71 249 (SA-1 101) 6 1 782 (SA-847)

,I CF Method Table Table Surv. DatafnJ Table Surv. Datafn)

Table Table Table able'^

Table Surv. Datafu)

Initial RTNDT (OF)

+50

+1 11

+ 1 11

-5

-5

-5

+10

-5 I t 3/4T 32.2 EFPY O,,,.B,,.Fluence FactodA) 0.5322 0.9452 I1 0.8993 I,

0.5322 N/A 0.8293 0.8993 0.7960 11 ARTNDT (OF) 40.98 83.18 74.95 49.73 32.19 81.11 N/A 130.70 150.72 125.29 129.99

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 30.51 EFPY'~) $talc (34.0 EFPY $ ~ c s t. ~ s t. )

(1)

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All beltline materials are included for comparison. '

L Vessel Manufacturer:

( Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding):

] 6.5", without cladp)

L Ik&m?&s Scc Tablc 2.

in' Non-credible su~eillancc data; see BAW-2325 for cvaluation. Tablc CF conscrvativc bccause differcncc betwccn mcasurcd ARTNDT and predicted ARTNDT bascd on Tablc CF is lcss than 2 0 (56°F).

Component Description Nozzle Belt Forging Intermediate Shell Forgina Lower Shell Forging I

Nozzle Belt to Interned. Shell Circ Weld

( 1 00%)

Interned. to Lower Shell Circ. Weld (100%)

Crcdiblc survcillancc data; scc BAW-2325 for cvaluation.

%Cu 0.1 1 0.09 0.05 0.18

, 0.26.

Non-crcdiblc survcillancc data; Tablc CF valuc bascd on best-cstimatc chcmistry is higher than best fit calculated using survcillancc data, and thcrcforc, conscrvativc.

Heat or Heat/Lot 123V352 123V500 122W195

,I 21935

, 72442 (SA-1484)

("'

ATustcd rcfcrcncc tcmpcraturc (ART) calculatcd pcr Regulatory Guide 1.99, Rcv. 2. ART = Initial RTNDT + ARTNDT + Margin, whcre ARTNDT

= Chemistry Factor x Flucncc Factor, and Margin = 2(012 +

UA )"', with 01 defined as thc standard dcviation of thc Initial RTNDT, and CIA dcfincd as the standard dcviation of ARTNDT. For cxamplc, for nozzle bclt forging, heat no. 123V352, ART = 40 + (76 x 0.5435) +

34 = I 15°F. Calculated ART valucs arc roundcd to thc ncarcst OF in accordance with the rounding-off mcthod of ASTM Practicc E29.

Initial RTNDT (OF)

+40

+40

+40

$1

-56

-5 iF' Instruction Manual, Rcactor Vcsscl, Point Bcach Nuclcar Plant No. 2, Combustion Engineering, CE Book #4869, Octobcr 1970.

By inspection, these arc thc limiting matcrial propcrtics.

CF Method Table TablefnJ Table Surv. DatafLJ able(^

Table 'U)

%Ni 0.73 0.70 0.72 0.70 0.60 Table CF value bascd on best-estimate chemistry data from CEDG Rcport "Bcst Estimatc Coppcr and Nickcl Valucs in CE Fabricatcd Reactor Vcsscl Wclds," CE NPSD-1039, Rcvision 2, Final Rcport, June 1997 314T 32.2 EFPY I ~ I & ~. ~ # ~.

Fluence actor(*'

0.5435 0.9958 0.9456 I,

0.5435 0.9405 CF 76 58 31 42.8 170

. 180 Per Wisconsin Elcctric Calculation 2000-001-00-A thc calculatcd fluencc for thc critical matcrial (SA-1484) occurs at 30.51 EFPY versus 34.0 EFPY I$Best.~,L bascd upon K=I$B~or.Eor./+Calc=O.921.

N' EFPY valuc listed hcrc is bascd on a rcactor powcr of 15 18.5 MW,. See Scction 2.0, "Operating Limits," for discussion of applicability datcs.

17 17 17 8.5 28

, 28 POINT BEACH TRM ARTNDT (OF) 41.31 57.76 29.3 1 40.47 92.40 169.29 I

0 0

0 !,

17

, 19.7 Margin (OF) 34 34 34 17 65.51

, 68.47 ART OF)(^)

115 1 32Iw 103 97 102

, 233'"