Text

BAW-2467NP, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units I and 2 for Extended Life Through 53 Effective Full Power Years
	ML042660313
Person / Time
Site:	Point Beach
Issue date:	07/31/2004
From:	; Framatome ANP
To:	; Office of Nuclear Reactor Regulation, Nuclear Management Co
References
	77-2467NP-00, BAW-2467NP
	Download: ML042660313 (44)
	v • d • e

ENCLOSURE 2 BAW-2467NP LOW UPPER-SHELF TOUGHNESS FRACTURE MECHANICS ANALYSIS OF REACTOR VESSEL OF POINT BEACH UNITS I AND 2 FOR EXTENDED LIFE THROUGH 53 EFFECTIVE FULL POWER YEARS JULY 2004 Page 1 of 44

BAW-2467NP July 2004 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Point Beach Units I and 2 for Extended Life through 53 Effective Full Power Years AREVA Document No. 77-2467NP-OO (See Section 11 for document signatures.)

Prepared for Nuclear Management Company Prepared by Framatome ANP, Inc.

An AREVA and Siemens company 3315 Old Forest Road P. 0. Box 10935 Lynchburg, Virginia 24506-0935 Page 2 of 44

BAW-2467NP EXECUTIVE

SUMMARY

Nuclear Management Company is considering plant life extension, power uprate to 1678 MWt and removal of hafnium power suppression assemblies from the core for Point Beach Units I and 2. As a result of these changes. operating conditions including vessel temperatures and projected fluence values at 53 effective full power years (EFPY) of plant operation have changed. It must be ensured that these changes do not affect the plant adversely from a regulatory compliance point of view. One of the compliance Issues Is Appendix G to 10 CFR Part 50 where low upper-shelf toughness Is addressed. An equivalent margins assessment has to be made for material toughness when the upper-shelf Charpy energy level fats below 50 ft-lb. This report addresses this particular compliance Issue regarding low upper-shelf toughness only.

The Charpy upper-shelf value of reactor vessel beliline weld materials at Point Beach Units 1 and 2 may be less than 50 ft lb at 53 EFPY. In order to demonstrate that sufficient margins of safety against fracture remain to satisfy the requirements of Appendix G to 10 CFR Part 50, a low upper-shelf fracture mechanics analysis has been performed. The liniting welds In the beltline region have been evaluated for ASME Levels A, B, C, and D Service Loadings based on the evaluation acceptance criteria of the ASME Code, Section Xl, Appendix K.

The analysis presented In this report demonstrates that the limiting reactor vessel beltline weld at Point Beach Units I and 2 satisfies the ASME Code requirements of Appendix K for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels for the weld material at 53 EFPY.

Page 3 of 44

BAW-2467NP TABLE OF CONTENTS 1.0 Introduction......

1-1 2.0 Changes In Operating Condition Parameters.2-1 3.0 Material Properties and Reactor Vessel Design Data 3-1 3.1 J-Integral Resistance Model for Mn-Mo-NULinde 80 Welds.

.3-1 3.2 Reactor Vessel Design Data................................

3-1 3.3 Mechanical Properties for Weld Material...........................

3-1 3.3.1 hdal Weld SA-847...

3-2 3.3.2 Circunferential Weld SA-11011.........................

3-3 3.3.3 Circurnferential Weld SA-1484.........................

, 3-4 4.0 Analytical Methodology................

4-1 4.1 Procedure for Evaluating Levels A and B Service Loadings.

..................... 4-1 4.2 Procedure for Evaluating Levels C and D Service Loadings.

4-1 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations..

4-3 4.4 Effect of Cladding Material..............................

4-3 5.0 Applied Loads...............................

.. 5-1 5.1 Levels A and B Service Loadings 5-1 5.2 Levels C and D Service Loadings...........................

5-1 6.0 Evaluation for Levels A and B Service Loadings 6-1 7.0 Evaluation for Levels C and D Service Loadings

......................... 7-1 8.0 Summary of Results............

8-1 9.0 Concluslon...........

9-1 10.0 References.......

10-1 11.0 Certification..........................................................

11-1 12.0 Appendb A

.12-1 Page 4 of 44

BAW-2467NP LIST OF TABLES Table 2-1 Evaluation Conditions.................

................................................ 2-2 Table 3-1 Mechanical Properties for SA-847 Weld of Point Beach Unit I.............................. 3-2 Table 3-2 Mechanical Properties for SA-1 101 Weld of Point Beach Unit 1.................... I...... 3-3 Table 3-3 Mechanical Properties for SA-1484 Weld of Point Beach Unit 2............................ 3-4 Table 6-1 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition I - Uprated Power Conditions Without HaNuum Assemblies................. 6-2 Table 6-2 Material JIntegral Resistance for Levels A and B Service Loadings - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies.........

62........

2 Table 6-3 Material J-lntegrat Resistance for Levels A and B Service Loadings - Evatuation Condition 3 - Current Power Conditions With Hafnium Assemblies..........

............ 6-2 Table 6-4 Flaw Evaluation for Levels A end B Service Loadings - Evaluation Condition I -

Uprated Power Conditions Witout Hafnium Assemblies................

...................... 6-3 Table 6-5 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 2 -

Current Power Conditions Without Hafidum Assemblies....................................... 6-3 Table 6-6 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 3 -

Current Power Conditions With Hafnium Assemblies.....................................,,.6-3 Table 7-1 AIntegral vs. Flow Extension for Evaluation Condition I - SA-847......................... 7-6 Table 7-2 -Irntegral vs. Flaw Extension for Evaluation Condition 1 - SA-1 101....................... 7-7 Table 7-3 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-1484....................... 7-8 Table 7-4 Level D Service Loadings - Internal Pressure at Tensile Instability - SA-47.........

7-9 Page 5 of 44

BAW-2467NP LIST OF FIGURES 2-1 Reactor Vessel of Point Beach Unit......................................................

2-3 2-2 Reactor Vessel of Point Beach Unit 2......................................................

2-4 5-1 Level D transients - Reactor Coolant Temperature and Pressure vs. Time..........

52 6-1 J-Integral vs. Flaw Extension for Levels A and B Service Loading - Evaluation Condition 1 - Uprated Power Conditions Without Hafnium Assemblies - Weld SA-847.........

4....

4 7-1 14 vs. Crack Tip Temperature for Evaluation Condition 1 - SA-847............................. 7-2 7-2 1 vs. Crack Tip Temperature for Evaluation Condition 1 - SA-1 101........................... 7-3 7-3 4 vs. Crack Tip Temperature for Evaluation Condition 1 - SA-1484........................... 7-4 7-4 J-lntegral vs. Flaw Extension - All Welds......................................................

7-10 7-5 J-lntegral vs. Flaw Extension - SA-847......................................................

7-11 v

A AREVA Page 6 of 44

BAW-2467NP 1.0 Introduction Nuclear Management Company is considering plant life extension, power uprate to 1678 MWt and removal of hafnium power suppression assemblies from the core for Point Beach Units 1 and 2. This document assesses the effect of these proposed changes on the upper-shelf fracture toughness of the reactor vessels.

The B&W Owners Group (B&WOG) fracture toughness model was used In the low upper-shelf toughness fracture mechanics analyses of the reactor vessels of the B&WOG Reactor Vessel Working Group (RVWG) which includes the Point Beach Units I and 2 reactor vessels. The low upper-shelf toughness analysis for all reactor vessels of the B&WOG RVWG for Levels A & 8 Service Loadings was documented In BAW-2192PA [1]. An additional fracture mechanics analysis for Levels C & D Service Loadings was carried out for all these reactor vessels and docurnented in BAW-21 78PA [2]. Both these reports have been accepted by the NRC.

As a result of a subsequent power uprate, an additional low upper-shelf toughness analysis covering end-of-license and end-of-license renewal fluence values was performed for Point Beach Units 1 and 2 (31. For the current planned changes, the effect on the reactor vessel materials upper-shelf toughness Is assessed In this report.

Welds in the beltline region of all B&W Owners Group Reactor Vessel Working Group plants, Including the Point Beach Units 1 and 2 vessels, have been analyzed [1, 21 for 32 effective full power years (EFPY) of operation to demonstrate that these low upper-shelf energy materials would continue to satisfy federal requirements for license renewal. In Reference 3, the Point Beach vessels were analyzed up to their forecasted end-of-license extension periods at a partially uprated power level of 1650MWI with hafnium power suppression assemblies, and both vessels were shown to be acceptable. The purpose of the present analysis Is to perform a similar low upper-shelf toughness evaluation of the reactor vessel welds at the Point Beach plants for projected neutron fluences at 53 EFPY.

The present analysis addresses ASME Levels A. B. C, and D Service Loadings. For Levels A and B Service Loadings, the low upper-shelf toughness analysis Is performed according to the acceptance criteria and evaluation procedures contained In Appendix K to Section Xi of the ASME Code [4]. The evaluation also utilizes the acceptance criteria and evaluation procedures prescribed in Appendix K for Levels C and D Service Loadings. Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal and stress models and linear elastic fracture mechanics methodology of Framatome ANP's PCRIT computer code to determine stress intensity factors for a worst case pressurized thermal shock transient.

1-1 A

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BAW-2467NP 2.0 Changes In Operating Condition Parameters As a result of the planned updates to the Point Beach Units I and 2. there are increases In the projected end of life fluenoes for both the units. There are also changes In the plants' operating temperatures. These inputs were provided by the Nuclear Management Company and included as Appendix A and summarized in this section.

The analysis for current licensed rated power conditions (1540 MWt) gives a maximum cold leg temperature of 544.5-F. As a result of the power uprate to 1678 MWt, the maximum cold leg temperature Is reduced to 541.40F. The projected reactor vessel fluence values at 53 EFPY are provided In Table 2-1. For this analysis, three cases, termed Evaluation Conditions, are studied - uprated power conditions without hafnium assembles, current power conditions without hafnium assemblies, and current power conditions with hafnium assemblies. Fluenoe values for these three cases are reported only for the controlling welds Identified through review of the results reported In References 1, 2 and 3. Locations of the reactor vessel welds for Point Beach Units 1 and 2 are illustrated in Figures 2-1 and 2-2 respectively.

2-1 A

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BAW-2467NP Table 2-1 Evaluation Conditions Fluence (n/cm 2) at 53 EFPY EVALUATION EVALUATION EVALUATION CONDITION I CONDmON 2 CONDmON 3 Weld Cu Ni Uprated Power Current Power Cunent Power Weld Number (wt%)

(wt%)

Conditions Without Conditions Without Conditions With Plant Location (1]

(11

[51 151 Hafnium Assembfles Hafnium Assembles Hafnium Assemblies Cold Leg Temp:

Cold Leg Temp:

541.40F 544.6-F 544.5F Po lwer Shell SA-847 0.23 0.52 3.37E+19 3.12E+19 2.67E+19 Inter.

Shell/Lower SA-1t01 0.23 0.59 4.91E+19 4.52E+19 3.82E+19 Shell Circ, Inter.

PB-2 Shell/Lower SA-1484 0.26 0.60 5.09E+19 4.65E+19 3.79E+19 Shell Circ.

-D co CD CD 01 2-2 A

AR EVA

BAW-2467NP Figure 2-1 Reactor Vessel of Point Beach Unit I

~ -

~-Weld SA-1426 Weld SA-8 12 Inside 27%

L SA-775 Outside 73%

intermediate Shell (Plate) A981 1-1

~-

Weld SA-1101 Weld SA-847 Lower Shell (Plate) C1423-1 3987 Weld SA-110 2-3 A

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BAW-2467NP Figure 2-2 Reactor Vessel of Point Beach Unit 2 8.44 l 4

CE Weld 8.44M Ea Intermediate Shell (Forging) 123V500VA I 150 Q

Weld SA-1484 Lower Shell (Forgfng) 122W195VA1 2-4 A

A REVA Page 11 of 44

BAW-2467NP 3.0 Material Properties and Reactor Vessel Design Data An upper-shelf fracture toughness material model Is discussed below, as well as mechanical properties for the weld material and reactor vessel design data.

3.1

-Integral Resistance Model for Mn-Mo-NilUnde 80 Welds A model for the Jlntegral resistance versus crack extension curve (J-R curve) required to analyze low upper-shelf energy materials has been derived specifically for Mn-Mo-NiInide 80 weld materials. A previous analysis of the reactor vessels of B&W Owners Group RVWG 11 described the development of this toughness model from a large data base of fracture specimens.

A lower bound (-2S,) J-R curve is obtained by multiplying 1-Integrals from the mean J-R curve by 0.699 [11.

It was shown in a previous low upper-shelf toughness analysis performed for B&W Owners Group plants [6) that a typical lower bound J-R curve Is a conservative representation of toughness values for reactor vessel beltline materials, as required by Appendix K (4] for Levels A, B. and C Service Loadings, The best estimate representation of toughness required for Level D Service Loadings Is provided by the mean J-R curve [7].

3.2 Reactor Vessel Design Data Pertinent design data for upper-shelf flaw evaluations in the belttine region of the reactor vessel are provided below for Point Beach Units I and 2.

Design Pressure, PO,

= 2485 pslg [23 (use 2500 psig)

Inside radius, R,

= 66 in. 12)

Vessel thickness, I

= 6.5 In. [2J Nominal cladding thickness, t, a 0.1875 In. [21 3.3 Mechanical Properties for Weld Material Mechanical properties for the base and weld materials are presented hI Tables 3-1 through 3-3.

The reactor vessel base metal at Point Beach Unit I Is SA-302, Grade B low alloy steel, and at Point Beach Unit 2 Is SA-508, Grade 2, Class 1 low alloy steel [8]. Base metal properties are found In the ASME Code [9).

Weld metal tensile properties are taken from appropriate surveillance capsule data of each weld material.

The ASME transition region fracture toughness curve for Kf*, used to define the beginning of the upper-shelf toughness region, is indexed by the Initial RTNor of the weld material. Also, Polsson's ratio, v, is taken to be 0.3.

3-1 A

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BAW-2467NP 3.3.1 Axial Weld SA-847 Table 3-1 Mechanical Properties for SA-847 Weld of Point Beach Unit 1 Temp.

E Yield Strength (a.)

Ultimate Strength (o)*

et Material:

Base Base Weld Base Weld Base Metal Metal SA-847 metal SA-847 Metal Source:

Code Code Actual Code Actual Code

[Ref.]

19]

[9]

[10]

[9]

[10]

[9]

(OF)

(ksi)

(ksl)

(nAnrF) 100 29200 50.00 95.00 s0 99.8 7.06E-06 200 28500 47.60 89.60 80 99.8 7.25E-06 300 28000 46.10 86.01 80 99.8 7.43E-06 335 27790 45.74 85.10 80 97.6 7.48E-06 400 27400 45.10 84.77 80 99.8 7.58E-06 500 27000 44.50 84.26 80 99.8 7.70E-06 541.4 26751.6 44.16 84.04 80 99.8 7,75E-06 544.5 26733 44.14 84.03 80 99.8 7.76E-06 550 26700 44.11 84.00 80 99.8 7.77E-06 600 26400 43.80 83.74 80 99.8 7.83E-06 Note: The ultimate stregth values of the base and weld metals given here are not used in calculations Initial R7wr = -5.0eF [51 Margin = 48.30F [5 3-2 A

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BAW-2467NP 3.3.2 Circumferential Weld SA-1011 Table 3-2 Mechanical Properties for SA-1 101 Weld of Point Beach Unit 1 Temp.

E Yield Strength (cry)

Ultimate Strength (q,)*

a Material:

Base Base Weld Base Wetd Base Metal Metal SA-1101 Metal SA-1101 Metal Source:

Code Code Actual Code Actual Code

[Ref.]

[91

[9]

fill

[91

[11]

19]

(IF)

(ksi (ksl)

(ksl)

(ksi)

(ksi) iniinrF) 100 29200 50.00 93.66 80 105.10 7.06E-06 200 28500 47.50 92.20 80 104.90 7.25E-06 300 28000 46.10 90.74 80 104.70 7.43E-06 400 27400 45.10 89.29 80 104.50 7.58E-06 500 27000 44.50 87.83 80 104.30 7.70E-06 541A 26751.6 44.14 87.23 80 104.21 7.76E-06 544.6 26733 44.14 87.18 80 104.21 7.76E-06 550 26700 44.11 87.10 80 104.20 7.77E-06 600 26400 43.80 86.37 80 104.10 7.83E-06 Note: The ulUmate strength values of te base and weld metals given here are not used hi calculations Initial RTJHr 10.0 F [51 Margin = 56.0*F [5]

3-3 A

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BAW-2467NP 3.3.3 Circumferential Weld SA-1484 Table 3-3 Mechanical Properties for SA-1484 Weld of Point Beach Unit 2 Temp.

E Yield Strength (os)

Ultimate Strength (q.)'

a Material:

Base Base Weld Base Weld Base Metal Metal SA-1484 Metal SA-1484 Metal Source:

Code Code Actual Code Actual Code (Ref.]

91

[9]

(121

[91 (121 (9]

(OF)

(ksi)

(ksl)

(ksi)

(ksl)

(inAnrF) 100 27800 50.00 82.10 80 96.90 6.50E-06 200 27100 47.50 79.57 80 92.98 6.67E-06 300 26700 46.10 78.00 80 90.40 6.87E-06 400 26100 45.10 77.17 80 89.41 7.07E-06 450 25900 44.76 76.80 80 89.60 7.15E-06 500 25700 44.50 76A2 80 90.29 7.25E-06 541.4 25460 44.16 76.15 80 91.25 7.32E-06 544.5 25444 44.14 76.13 80 91.34 7.33E-06 580 25264 43.94 76.00 80 92.50 7.39E-06 600 25200 43.80 75.80 80 93.28 7.42E-06 Note: The ultinate sbtength values of the base and weld metals given here are not used in calculatuns Initial RTNOT = -5.0-F P5]

Margin = 68.5'F (5]

34 A

A RE VA Page 15of44

BAW-2467NP 4.0 Analytical Methodology Upper-shelf toughness is evaluated through use of fracture mechanics analytical methods that utilize the acceptance criteia and evaluation procedures of Section Xl, Appendix K [41, where applicable.

4.1 Procedure for Evaluating Levels A and B Service Loadings The applied Jinregral Is calculated per Appendix K. paragraph K-4210 [41. using an effective flaw depth to account for small scale yielding at the crack tip, and evaluated per K-4220 for upper-shelf toughness and per K-431 0 for flaw stability.

4.2 Procedure for Evaluating Levels C and D Service Loadings Levels C and D Service Loadings are evaluated using the one-dimensional, finite -lement, thermal and stress models and linear elastic fracture mechanics methodology of the PCRIT computer code to determine stress intensity factors. The beltline region welds identifiled In Section 3.3 am analyzed for all Level C and D transients. Two Level D transients are specified for the Point Beach Units. The original equipment specification Includes a Steam LUne Break (SLB) transient and a Reactor Coolant Line Break (LOCA) transient. The Point Beach FSAR contains a Steam Line Break (two loops In service) without Offslte Power transient [13].

The transients considered appear in Figure 5.1. Transients are assumed to hold steady at the end of their definitions, and are held constant until the thermal gradient through the shell has developed fully and begins to dissipate.

The evaluation is performed as follows:

(1)

For each transient described above, utilize PCRIT to calculate stress intensity factors for a serni-elliptical flaw of depth '/lo of the base metal wall thickness, as a function of time, due to internal pressure and radial thermal gradients with a factor of safety of 1.0 on loading.

The applied stress intensity factor, &,

calculated by PCRIT for each of these transients is compared to the It, limit of the weld. The transient that most closely approaches the Kk limit Is chosen as the limitfng transient, and the critical time In the 1miting transient occurs at the point where 14 most closely approaches the upper-shelf toughness curve.

(2)

At the critical transient time, develop a crack driving force diagram with the applied J-integral and J-R curves plotted as a function of flaw extension. The adequacy of the upper-shelf toughness Is evaluated by comparing the applied J-integral with the J-R curve at a flaw extension of 0.10 in.

Flaw stability Is assessed by examining the slopes of the applied J-integral and J-R curves at the points of intersection.

(3)

Verify that the extent of stable flaw extension Is no greater than 75% of the vessel wall thickness by deternining when the applied J.-ntegral curve Intersects the mean J-R curve.

4-1 A

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BAW-2467NP (4)

Verify that the remaining ligament Is not subject to tensile instability. The internal pressure p shall be less than Pi, where Pt Is the Internal pressure at tensile instability of the remaining ligament Equations for Pi are given below for the axial and circumferential flaws [14J. These equations first appear in the 2001 Edition of the ASME Section Xl code that Is cited.

(a) For an axial flaw, Pi =.,7o[(R I -)(A, A)l

[eqn. 1) where CO

[eqn. 2]

2 A = t(e +t)

[eqn. 3 gat Ieqn. 4]

and I

z surface length of crack, six times the depth, a Rm z mean radius of vessel This equation for Pi Includes the effect of pressure on the flaw face.

(b) For a circumferential flaw, Pi = 1.07aco

/ 1 - (A,

/A)]

[eqn. 5J where ao. A. ard A, are given by equations 2, 3 and 4, respectively.

This equation for Pi includes the effect of pressure on the flaw face. This equation Is valid for Internal pressures not exceeding the pressure at tensile instability caused by the applied hoop stress acting over the nominal wall thickness of the vessel. This validity limit on pressure for the circumferential flaw equation for Pi Is Pi s 1.07Go[ ]

[eqn. 6]

4-2 A

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SAW-2467NP 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations Upper-shelf fracture toughness Is determined through use of Charpy V-notch Impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level. Similarly, fracture toughness can be addressed in three different regions on the temperature scale. i.e. a lower-shelf toughness region, a transition region, and an upper-shelf toughness region. Fracture toughness of reactor vessel steel and associated weld metals are conservatively predicted by the ASME Initiation toughness curve, Kt, in the lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, Kj, is derived from the upper-shelf JAntegral resistance model described in Section 3.1. The upper-shelf toughness then becomes a function of fluence, copper content temperature, and fracture specimen size.

When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperature. Since te present analysis addresses the low upper-shelf toughness Issue, only the upper-shelf temperature range, which begins at the intersection of KJ, and the upper-shelf toughness curves, Kj,, Is considered.

4.4 Effect of Cladding Material The PCRIT code utilized in the flaw evaluations for Levels C and D Service Loadings does not consider stresses In the cladding when calculating stress Intensity factors for thermal loads. To account for this cladding effect, an additional stress intensity factor, Kmd, is calculated separately and added to the total stress intensity factor computed by PCRIT.

The contribution of cladding stresses to stress Intensity factor was examined previously [2j. In this low upper-shelf toughness analysis performed for B&W Owners Group Reactor Vessel Working Group plants, the Zion-1 WF-70 weld using thermal loads from the Turkey Point SLB was deternined to be the bounding case. The Zion-1 vessel was as thick as or thicker than any other vessel. The thicknesses of the reactor vessels for the both Point Beach units are 6.5' whereas the Zion vessel is 8.44g. The nominal cladding thickness Is 31160 for both vessels.

From a thermal stress perspective, It Is conservative to consider the thicker vessel. For the Zion vessel, the maximum value of KJw, at any time during the transient and for any flaw depth, was determined to be 9.0 ksiNin. This bounding value is therefore used as the stress Intensity factor for idwin this Point Beach low upper-shelf toughness analysis.

4-3 A

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BAW-2467NP 5.0 Applied Loads The Levels A and B Service Loadings required by Appendix K are an accumulation pressure (internal pressure load) and a cooldown rate (thermal load). Since Levels C and D Service Loadings are not specified by the Code, Levels C and D pressurized thermal shock events are reviewed and a worst case transient Is selected for use in flaw evaluations.

5.1 Levels A and B Service Loadings Per paragraph K-1300 of Appendix K [4], the accumulation pressure used for flaw evaluations should not exceed 1.1 times the design pressure. Using 2.5 ksi as the design pressure, the accumulation pressure is 2.75 kWl.

The cooldown rate Is also taken to be the maximum required by Appendix K, 100lFihour.

5.2 Levels C and D Service Loadings As discussed In Section 4.2, the SLB and LOCA transients are evaluated using the computer code PCRIT. Pressure and temperature ime histories for the two transients considered are shown in Figure 5-1.

5-1 A

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BAW-2467NP Figure 5-1 Level D transtients - Reactor Coolant Temperature and Pressure vs. Tirme 5-2 A

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BAW-2467NP 6.0 Evaluation for Levels A and B Service Loadings The material mean and lower bounding J-R values for Evaluation Conditions 1, 2 and 3 detailed in Table 2-1 are given in Tables 6-1 through 6-3, respectively. initial flaw depths equal to '14 of the vessel wall thickness are analyzed for Levels A and B Service Loadings following the procedure outlined In Section 4.1 and evaluated for acceptance based on values for the J-integral resistance of the materials from Section 3.3.

The results of the evaluation are presented in Table 6-4 through 6-6, where It Is seen that the minimum ratio of material J-integral resistance (Jo.,) to applied J-integral (Jo) Is 1.87 for the SA-847 axial weld for Evaluation Condition 1, uprated power conditions without hafnium power suppression assemblies. This ratio Is higher than the minimum acceptable value of 1.0. Also Included In Table 6-4 through 6-6 is the applied J-integral at (JO.,) with a safety factor on pressure of 1.25.

The flaw evaluation for the controlling weld (SA-847) and controlling Evaluation Condition (1) Is repeated by calculating applied J-ntegrals for various amounts of flaw extension with safety factors (on pressure) of 1.15 and 1.25. The results, along with mean and lwver bound J-R curves, are plotted In Figure 6-1. The requirement for ductile and stable crack growth is also demonstrated by Figure 6-1 since the slope of the applied J-integral curve for a safety factor of 1.25 is considerably less than the slope of the lower bound J-R curve at the point where the two curves Intersect 6-1 A

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BAW-2467NP Table 6-1 Material J-ntegral Resistance for Levels A and B Service Loadings - Evaluation Condition 1 - Uprated Power Conditions Without Hafnium Assemblies J-R at Aa = 0.1 in.

Cold Controlling Weld Fluence Lower Plant Leg Material Weld Cu x 1019 Mean Bound Temp.

ID Orientation Content (nlac2) at -2Se r(F)

(wto) at l.S.

at tl4 (btn)

(frin)

PB-1 641.4 SA-B47 L

023 33.70 2228 885 618 PB-1 541A SA-1101 C

0.23 49.10 32.46 870 608 PB-2 541A SA-1484 C

0.26 50.90 33.64 828 678 Table 6-2 Material J-ntegral Resistance for Levels A and B Service Loadings - Evaluation Condition 2 - Current Power Conditions Without Hafnium Assemblies J-Ratika=0.1 In. I Cold Controlling Weld Fluence Lower Plant Leg Materhl Weld Cu x 101' Mewn Bound Temp.

ID Orientation Content (rncm2) at -2Se (F)

(wt%)

atl.S.

attV4 (b)

(

PB-1 544.5 SA-847 L

0.23 31.15 20.59 885 618 PB-1 544.5 SA-1 101 C

023 45.20 29.88 870 608 PB-2 544.5 SA-1484 C

0.26 46.45 30.70 827 578 Table 6-3 Material J-Integral Resistance for Levels A and B Service Loadings - Evaluation Condition 3 - Current Power Conditions With Hafnium Assemblies J-R at a =0.1 In.

Cold Controlling Weld Fluence Lower Plant Leg material Weld Cu x loll Mean Bound Temp.

ID Orientation Content (nkm2) at -2Se

("F)

(wt%)

at I.S.

at14 OMn)

(l/)

PB-1 544.5 SA-847 L

023 26.65 17.62 891 623 P1-1 544.5 SA-1101 C

0.23 38.20 25.25 877 613 PB-2 544.5 SA-1484 C

0.26 37.85 25.02 836 585 6-2 A

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BAW-2467NP Table 6-4 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition I -

Uprated Power Conditions Without Hafrium Assemblies Lower Bounding SF = 1.15 SF = 125 Plant Weld Weld Jo, at V4 J1 Jo., IJI Ji Jo., MI Number Orientation (lbIn)

(lb/n)

(lbrin)

PB-I SA-847 L

618 331 1.87 388 1.59 PB-I SA-101 C

608 98 6.20 113 5.38 PE-2 SA-1484 C

578 104 5.56 119 4.86 Tabb 6-5 Flaw Evaluation for Levets A and B Service Loadings - Evaluation Condition 2 -

Current Power Conditions Without Hafnium Assemblies Lower Bounding SF

1.15 SF a 125 Plant Weld Weld Jo., at 4 J.

J. MIJ 1

J, j Jo./I/

Number Orientation Ob(ln)

(lblin)

(Ibfli)

PB-1 SA-847 L

618 331 1.87 388 1.59 PB-1 SA-1101 C

60B 98 6.20 113 5.38 P8-2 SA-1484 C

578 104 5.6 I

119 4.86 Table 6-6 Flaw Evaluation for Levels A and B Service Loadings - Evaluation Condition 3 -

Current Power Conditions With Hafnium Assemblies LowerBounding SFz1.15 SF=I125 Plant Weld Weld JI. at t/4 J1 J0.1 MI J1 Jo. IJ Number Orientation obftn)

(lbrmn)

(b/in)

PB-I SA-847 L

623 331 1.88 388 1.61 PB-I SA-11`1 C

613 98 6.26 113 6.42 P8-2 SA-1484 C

585 104 6.63 119 4.92 6-3 A

AREVA Page 23 of 44

BAW-2467NP Figure 6-1 J-Integral vs. Flaw Extension for Levels A & B Service Loadings - Evaluation Condition 1 - Uprated Power Conditions Without Hafnium Assemblies -Weld SA-847 800 750 650 600 c

I3 500 450 400 350 300 _.

0.00 0.05 0.10 0.15 0.2D Flaw Extenslon, Aa (in.)

0.25 64 A

ARE VA Page 24 of 44

BAW-2467NP 7.0 Evaluation for Levels C and 0 Service Loadings A flaw depth of 111o of the base metal wall thickness, plus the cladding thickness, is used to evaluate the Level 0 Service Loadings. The stress intensity factor 1K calculated by the PCRIT code Is the sumn of thermal, residual stress, deadweight, and pressure terrs. PCRIT Is run for each Level D transient. RTMT Is also calculated by PCRIT. Transition region toughness is obtaIned from the ASME Section Xi equation for crack Initiation [15].

Kk 33.2 + 2.806 exp[0.02(T-RTpot+ 100F)J

[eqn. 7]

where:

= transition region toughness, ksl~n T = crack Up temperature, OF Upper-shelf touhness is derived from the J-Integral resistance model of Section 3.1 for a flaw depth of '/Io of the waft thickness, a crack extension of 0.10 in., and fluence, as follows:

K=

0sE

[eqn. 8]

1000(1 _-V2) where

.KM uppershelf region toughness, ksiain J,= J-ntegral resistance at ea = 0.1 in.

Figure 7-1 through 7-3 shows the variation of applied stress intensity factor, K4. transition range toughness, K,&, and upper-shelf toughness, Ktc with temperature for the Evaluation Condition I described in Table 2-1 for the three welds. The markers on the K4 curve indicate points In time at which PCRIT soluflons are available. For all the three welds that were analyzed, the LOCA transient Is liiting since it most closely approaches the Kk limit of each weld. All subsequent analysis will pertain to this transient In the upper-shelf toughness range, the /4 curve Is closest to the lower bound KM curve at a particular time point Into the transient for each weld, as listed below:

We I

Time (mmi SA-847 2.40 SA-1011 1.50 SA-1484 1.30 For each weld, the time specified above is selected as the critical Urme in the transient at which to perform the flaw evaluation for Level D Service Loadings.

7-1 A

AREVA Page 25 of 44

BAW-2467NP Fgure 7-1 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-847 220 200 1I0 160 140 120 e

100 ED G0 4D 20 o _-

275 325 375 425 475 525 575 Crack Tip Temperature (OF) 7-2 A

AREVA Page 26 of 44

BAW-2467NP Figure 7-2 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-1 11 220 i

II-

- -Kic l

-*--KJc Mean I

KJc Lower Bound

! 1

-Upper Shef Lkt 180

'._iKl at a=1110 for 20103 FSAR SLB

..- _KI at a=10 fior ESPEC LOCA I

-_, -K i at a=10 for ESPEC SLB 160 140

_ 1Evaluation pointat

/

1.50 min. Into Itrnsent

100

}

g.oo 60 40 2 DUpper-Shelf Toughness Range 0

275 325 375 425 475 525 575 Crack Tp Temperature (OF)

A ARE VA Page 27 of 44

BAW-2467NP Figure 7-3 K, vs. Crack Tip Temperature for Evaluation Condition I - SA-1484 220 200 180 160 140 I

I

- - - Kic

--KJc Mean I

KJcLowerBound UpperShief Linmt i

-. + -Kat aV1I0 for 20O3 FSAR SLB I

_KIU at a-VI0 for ESPEC LOCA

~

I i

I Ki at azt/10 for ESPEC SLB Evaluation point at

,1.3 into wsient 120 C

15 100 80 60 40 4--.

4 i

I Upper-Sheff Tou 20 I I -

1z, 0

275 325 375 425 475 Crack Tip Temperature (OF) 625 575 7-4 A

AREVA Page 28 of 44

BAW-2467NP Applied i-integrals for the LOCA transient are calculated for each weld at the critical time points identified above for various flaw depths in Table 7-1. 7-2, and 7-3 using stress intensity factors from PCRIT and adding 9.0 ks1ln to account for cladding effects. Stress Intensity factors are converted to iI-ntegrals by the plain strain relationship, J,,Wd (a) = I1000 Kiw(a) (I_ V2 )[eqn.

9]

E Tables 7-1. 7-2, and 7-3 lists flaw extensions vs. applied J-Integrals.

As the Point Beach vessels are 6.5 In. thick, the Initial flaw depth of '/l of the wall thickness Is 0.65 in. Flaw extension from this flaw depth Is calculated by subtracting 0.65 In. from the built-in PCRIT flaw depths I the base metal. The results, along with mean J-R curve, are plotted In Figure 7-4.

This figure indicates that Weld SA-847 Is limiting as the ratio of the applied J-Integral to the material J-R curve Is less than the other two welds. Figure 7-5 Is a plot of the applied J-Integrals and the mean i-R curves for the three Evaluation Conditions from Table 2-1 for Weld SA-847. Evaluation Condition 1. uprated power conditions without hafnium power session assemblies, is the limiting case as the ratio of the mean J-R curves to applied i-ntegrals Is the minimum of the three Evaluation Conditions. The requirements for ductile and stable crack growth are demonstrated by Figure 7-5 since the slopes of the applied J-Integral curves are considerably less than the slopes mean J-R curves at the points of Intersection. The Level D Service Loading requirement that the extent of stable flaw extension be no greater than 75% of the vessel wall thickness Is easily satisfied since the applied J-Integral curves Intersects the mean i-R curves at flaw extensions that are only a small fraction of the wall thickness (less than 1%).

The last requirement Is that the Internal pressure p shall be less than P., the Internal pressure at tensile instability of the remaining EgamenL Table 7-4 gives the results of the calculations for Pi for flaw depths up to 1.365 Inches for Evaluation Condition 1. As the Internal pressure p Is less than Pa the remaining ligament is not subject to tensile instability.

7-5 A

AREVA Page 29 of 44

BAW-2467NP Table 7-1 J-Integral vs. Flaw Extension for Evahuation Condition 1 - SA-847 Time =

2.40 min E = 26751.6 ksa Crack tip at VI0 1=

6.5 In.

v =

0.3 W*ro0 a6 As Temp.

Kip.f Kdd VV4I JWWP (in.)

(in.)

(F)

(lb/in) 1 0.1625 24BA0 62.08 9.0 71.1 172 2

0.3250 274.0 83.65 9.0 92.7 292 3

0.4875 302.10 94.64 9.0 103.6 365 4

0.6500 0.0000 328.00 100.97 9.0 110.0 411 5

0.8125 0.1625 352.70 10424 9.0 113.2 436 6

0.9750 0.3250 375.90 105.82 9.0 114.8 448 7

1.1375 0.4875 397.70 106.12 9.0 115.1 451 8

100 0.6500 41790 105.76 9.0 114.8 448 9

1.4625 0.8125 436.50 104.86 9.0 113.9 441 10 1.6250 0.9750 453.60 103.22 9.0 112.2 428 12 1.9500 1.3000 483.10 98.74 9.0 107.7 395 14 2.2750 1.6250 507.00 93.05 9.0 102.1 354 16 2.W00 1.9500 525.80 88.28 9.0 97.3 322 18 2.9250 2.2760 540.10 82.87 9.0 91.9 287 20 3.2500 2.6000 550.70 77.27 9.0 86.3 253 22 3.5750 2.9250 558.40 71.71 9.0 80.7 222 24 3.9000 3.2500 563.90 6653 9.0 75.5 194 26 42250 3.5750 567.60 61.61 9.0 70.8 171 28 4.5500 3.9000 670.00 6720 9.0 66.2 149 30 4.8750 4.2250 571.60 62.58 9.0 61.6 129 32 5.2000 4.5500 572.60 48.13 9.0 57.1 111 Note:

ao Is the flaw depth In the base metal 7-6 A

A REVA Page 30 of 44

BAW-2467NP Table 7-2 J-Integral vs. Flaw Extension for Evaluation Condition I - SA-1 101 Tkne =

1.50 min E

26751.6 ksl Crack Up at t10 t=

6.5 hI.

v=

0.3 (arit)'40 a8 Aa TenMp.

Khum Kjdad Kma J.,

(hi)

(hi.)

(F) b(in) 1 0.1625 280.80 69.65 9.0 68.7 160 2

0.3250 31480 78.67 9.0 87.6 261 3

0.4875 346.70 86.65 9.0 95.7 311 4

0.6500 0.0000 376.30 90.22 9.0 99.2 335 S

0.8125 0.1625 403.60 91.26 9.0 100.3 342 6

0.9750 0.3250 428.40 90.74 9.0 99.7 338 7

1.1375 0.4875 450.60 89.06 9.0 98.1 327 8

1.3000 0.6500 470.60 86.71 9.0 95.7 312 9

1.4625 0.8125 488.00 83.66 9.0 92.7 292 10 1.5620 0.975D 503.10 80.42 9.0 89.4 272 12 1.9500 1.3000 527.20 72.98 9.0 82.0 229 14 22750 1.6250 544.30 65.06 9.0 74.1 187 16 2.6000 1.9500 555.90 57.27 9.0 66.3 149 13 2U250 22750 563.40 49.24 9.0 58.2 115 20 32500 2.6000 568.10 41.31 9.0 50.3 88 22 3.5750 2.9250 570.90 34.09 9.0 43.1 63 24 3.9000 3.2500 572.40 27.47 9.0 36.5 45 26 4.2250 3.5750 573.30 21.94 9.0 30.9 33 28 4.5530 3.9000 573.70 17.63 9.0 26.6 24 30 4.8750 4;2250 573.90 14.36 9.0 23A 19 32 5.2000 4.5500 574.00 11.69 9.0 20.6 14 Note:

a is fte flaw deth hi the base metal 7-7 A

AREVA Page 31 of 44

BAW-2467NP Table 7-3 J-lntegral vs. Flaw Extension for Evaluation Condition I - SA-1484 Time -

1.30 min E -

25459.9 ksi CrackUp atVI0

t.

6.5 in.

V 0.3 (a'"A)*40 a'+

Ma Temp.

K,,

Kjd Kww J.P (in.)

(in.)

(F)

Qbin) 1 0.1625 292.60 51.19 9.0 602 129 2

0.3250 328.30 67.16 9.0 76.2 207 3

0.4875 361.

73.97 9.0 83.0 248 4

0.6500 0.0000 392.10 76.91 9.0 85.9 264 5

0.8125 0.1625 419.80 77.72 9.0 86.7 269 6

09750 0.3250 444.70 77.16 9.0 86.2 265 7

1.1375 0.4875 466.60 75.59 9.0 84.6 256 8

1.3000 0.6500 485.80 73.43 9.0 82.4 243 9

1.4625 0.8125 502.50 70.67 9.0 79.7 227 10 1.6250 0.9760 616.40 67.71 9.0 76.7 210 12 1.A0 1.3000 538.10 61.07 9.0 70.1 175 14 22750 1.6250 5.60 54.04 9.0 63.0 142 16 2.6000 1.9500 861.80 47.18 9.0 86.2 113 18 2.9250 22750 567.40 40.21 9.0 49.2 87 20 3.2500 2.6000 570.60 33.42 9.0 42.4 64 22 3.5750 2.9250 572.40 27.38 9.0 36.4 47 24 3.8000 3.2500 573.30 21.99 9.0 31.0 34 26 4.2250 3.5750 573.80 17.69 9.0 26.7 25 28 4.5500 3.9000 574.00 1453 9.0 23.5 20 30 4.8750 42250 574.00 12.34 9.0 21.3 16 32

.2000 4.5500 574.10 10.58 9.0 19.6 14 Note:

a ilhefiawdepthlnthebasemetal 7-8 A

AREVA Page 32 of 44

BAW-2467NP Table 7-4 Level D Service Loadings - Internal Pressure at Tensile Instability - SA847 flawdeptha(n.)

Pi (kal) 0.0650 9.18 0.1300 9.16 0.1950 0.14 02600 9.12 0.3250 9.09 0.3900 9.06 0.4550 9.02 0.5200 8.98 0.5850 8.93 0.6500 8.88 0.7150 8.84 0.7800 8.78 0.8450 8.73 0.9100 8.68 0.9750 8.62 1.0400 8.56 1.1050 8.51 1.1700 8.45 1.2350 8.39 1.3000 8.32 1.3660 8.26 7-9 A

ARE VA Page 33 of 44

BAW-2467NP Figure 7-4. JIntegral vs. Flaw Extension - All Welds 1600 1400 -

1200 SA-847 Mean J-R Curve

-> -- SA1101 Mean J-R Curve 1000 SA-1484 Mean J-R Curve SA-847 J applied 6 SA-1101 Japplied SO A-1484 J appled jBoo 600 -

400 200 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Extension, Aa (In.)

7-10 A

ARE VA Page 34 of 44

BAW-2467NP Figure 7-5. JAlntegral vs. Flaw Extension -Weld SA-847

-W 1400.

1200 1000 J-R 1l A* J-R Irl J-R Irl J appli

-f-J appli J appli legral Uprated Without Hafnium egral Current Without Hafnium egral Current With Hafnbum ed Uprated Wihout Hahfium ed Current Without Hafnium ed Current With Hafnium 00 -1 600 -

400 200

=M= -

0 O I...

' ' '

I *

I i

0.00 0.05 0.10 0.15 020 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Extension, aa (in.)

7-11 A

ARE VA Page 35 of 44

BAW-2467NP 8.0 Summary of Results A low upper-shelf toughness fracture mechanics analysis has been performed to evaluate the reactor vessel welds at Point Beach Units I and 2 for projected low upper-shelf energy levels at 53 EFPY, considering Levels A, B. C, and D Service Loadings of the ASME Code.

Evidence that the ASME Code, Section Xl. Appendix K (41 acceptance cnteria have been satisfied for Levels A and B Service Loadings is provided by the following:

(1)

The limiting weld is the SA-847 axial weld of Point Beach Unit 1 in the uprated power condition without hafnium power suppression assembles.

Figure 6-1 shows that with factors of safety of 1.15 on pressure and 1.0 on thermal loading, the applied J-nlegral (J1) is less than the Jintegral of the material at a ductile flaw exdension of 0.10 In. (Jt 1). The ratio JQ.1/J1 = 1.87 which is significantly gater than the required value of 1.0.

(2)

Figure 6-1 shows that with e factor of safety of.1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-Integral curve Is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Evidence that the ASME Code, Section Xl. Appendix K [4j acceptance criteria have been satisfied for Level D Service Loadings Is provided by the following:

(1)

Figure 7-5 shows that with a factor of safety of 1.0 on loading, flaw extensions are ductile and stable since the slope of the applied J-ntegral curve Is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

(3)

Figure 7-5 shows that the flaw remains stable at much less than 75% of the vessel wall thickness. It has also been shown that the rernaining ligament is sufficient to preclude tensile instability by a large margin.

8-1 A

AREVA Page 36 of 44

BAW-2467NP 9.0 ConclusIon The limiting Point Beach Units 1 and 2 reactor vessel bettline weld (axial weld SA-847 of Unit 1) satisfies the acceptance criteria of Appendix K to Section Xl of the ASME Code 14] for projected low upper-shel Charpy impact energy levels at 53 effective full power years of plant operation for the three conditions evaluated: uprated power conditions (1678 MWt) without hafnium power suppression assemblies, current power conditions (1540 MWt) without hafnium power suppression assemblies, and current power conditions (1540 MWt) with hafnium power suppression assemblies.

A AREVA Page 37 of 44

BAW-2467NP 10.0 References

1. BAW-2192PA. Low Upoer-Shelf Touahness Fracture Mecharics Analsis of Reactor Vessels of B&W Owners Reactor Vessel Working GrouD For Level A & B Service Loads, April 1994.

2. BAW-2178PA, Low Uoper-Shelf Touahness Fracture Mechanics Analvsis o Reactor Vessels of B&W Owners Reactor Vessel Workina Group For Level C 8 D Service Loads.

April 1994.

3. BAW-2255, Effect of Power Uggrade on Low Upoer-Shelf Toughness Issue. May 1995.

4. ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with Addenda through 2000.

6. USNRC Reactor Vessel Integrity Database Version 2.0.1 (RVID).

6. BAW-2275, Low Upper-Shelf Toughness Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY, August 1998.

7. BAW-2312, Revision 1, Low Upper-Shelf Touahness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 end 4 for Extended Life through 48 Effective Full Power Years, December 2000.

8. BAW-21 50, Materials Information for Westinghouse-Designed Reactor Vessels Fabricated by B&W, December 1990.

9. ASME Boiler and Pressure Vessel Code,Section III, Appendices. 1989 Edition with no Addenda.

10. WCAP-13902, Analysis of Caosule S from the Rochester Gas and Electric Corporation R. E. Ginna Reactor Vessel Radiation Surveillance Proaram, December 1993.

11. WCAP-1 5916. Analysis of Capsule X from the Florida Power and Light Turkey Point 3 Reactor Vessel Radiation Surveillance Program, September 2002.

12. BAW-2254, Test Results of Cacsule CR3-LG2: B&W Owners Grouo - Master Integrated Reactor Vessel Survellance Program, October 1995.

13. Point Beach Nuclear Plant Units I and 2 Final Safety Analysis Report, June 2003.

14. ASME Boiler and Pressure Vessel Code, Appendix K, Section Xl, 2001 Edition.

15. EPRI NP-719SR, T.U. Marston, Flaw Evaluation Procedures: ASME Section Xl, Electric Power Research Institute, Palo Alto, California, August 1978.

10-1 A

A RE VA Page 38 of 44

BAW-2467NP 1 1.0 Certification This report is an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels at Point Beach.

I

~

1 aLI/e f

H. P. Gunawardane, Engineer IlIl Date Materials and Structural Analysis Unit This report has been revtewed and found to be an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels at Point Beach.

A. D. Nana7Wmcpal Engineer Date Materials and Structural Analysis Unit Verification of independent review.

This report is approved for release.

A. D. McKim, Manager Date Materials and Structural Analysis Unit 7Q P-4 ?/Uinto

. Austin, Project Development Manager Date 11-1 A

AREVA Page 39 of 44

BAW-2467NP 12.0 Appendix A The following pages contain Input Information from Nuclear Management Company.

12-1 A

AREVA Page 40 of 44

BAW-2467NP commiaDt a, Nto Ar ExcIOhN Paint Beach Nuclear Plant Operated by Nuclear Managernent Company. LLC NPL 2004-0139 Junc 29, 2004 Heshan Gunawardane AREVA / Framatomre ANP, Lac.

MS OF50 3315 Old Forest Road Lynchburg. VA 24501 Heshan:

This correspondence will serve to formally document the requcsted inputs for the PBNP Units I and 2 RPV Equivalent Margins Assessment that is being pcrformcd in accordance with ARE VA Proposal FANP-04-1067, April 2, 2004.

baplicable ASME Section Xl Code The PBNP IS1 Program is in the fourth ten-year interval, which began on July 1, 2002 for both PBNP-I and PBNP-2. The program is in accordance with the 1998 edition through 2000 addenda (9SAOO) of ASME Section XJ Code as modified by 10 CFR 50.55a and approved relief requests and code cases.

(Reference 1)

Fluence Proiections For the case of full uprated power condition (1678 MWt), without hafnium absorber asscmblies, for EOLE (53 EFPY) use the older calculated fluence projections contained in Section 2 of Reference 2.

This is requested for input consistency with the remaining RV embrittlement analyses.

For the cases of mini uprared power condition (1540 MWt), with and without hafnium absorber assemblics, for EOLE (53 EFPY) use the revised calculated fluence projections contained in Section 2 of Reference 3.

0590 Nuclear Road

Two Rivers, Wisconsin 54241 Telephone: 920.7552321 12-2 A

ARE VA Page 41 of 44

EBAW-2467NP NPL 2004-0139 June 29,2004 Page 2 Normal Hcatup and Cooldown Rates The PBNP RCS heatup and cooldown rates for normal operation are 100 degrees Fahrenheit per hour for both beattaps and cooldowns. (Reference 4)

Predicted Oferating Tempraturesa The analyses for curnt licensed rated power conditions (1540 MWt) include a range of full load T(avg)'s from 558.1 to 574 degrees Fahrenbeit. The resulting T(hot) and T(cold) ranges are 588.1 to 603.5, and 528 to 544.5 degrees Fahrenheit, respectively (Reference 5). PBNP currently uses a T(avg) program of 547 to 570 degrees Fahrenheit (no load to full load) (Reference 6), resulting in a T(hot) and T(cold) of approximately 597 and 542 degrees Fahrenheit, Yespectively (Refcrence 7).

The analyses for the 10.5 percent uprated power condition (1678 MWt) include a range of T(avg) from 558.6 to S73.4 degrees Fahrenheit. The resulting T(hot) and T(cold) ranges are 591.2 to 605.5. and 526 to 541.4 degrees Fahrenheit, respectively (Reference 8).

Transient Inforrnation The original component transients arc defined in each RPV design specification (References 9 and 10 for Units 1 and 2, respectivcly). A revised set of component design transients was generated to support steam generator replacement, a partial power uprate (8.7 percent), and license renewal (Reference I l).

The RPV transients were evaluated and characterized for the partial power uprated condition in Refrence 12. The RPV btansients were further evaluated and characterized for full uprated conditions in Reference 13.

In addition, Chapter 14 of the PBNP FSAR (Reference 14) has been provided via previous conespondence. Chapter 14 contains the PBNP safety analysis summaries. These transients should be reviewed for bounding conditions with respect to the component design transients.

Alicable ASME Section n and Im Code ASME Boiler and Pressure Vessel Code, Section H, 1989, no Addenda.

ASME Boiler and Pressure Vesscl Code, Section mH 1989, no Addenda.

A 12-3 AREVA Page 42 of 44

BAW-2467NP NPL 2004-0139 June 29, 2004 Page 3 Sincerely, Brad Fronnm PBNP License Renewal Nuclear Management Company James E. Knorr Manager of License Renewal PBNP Nuclear Management Company bms RefMe ces:

I. SER 2001-0010, 'Point Beach Nuclear Plant, Units I and 2 - Relief Requests RR 1-24 (Unit 1)

And RR-2-30 (Unit 2) Re: Use Of ASME Code Section XI, 1998 Edition With Addenda Through 2000 (TAC Nos. MB2230 And MB2231)", dated November 6.2001.

2. Westinghouse Letter Report, LTR-REA-02-23, "Pressure Vessel Neutron Exposure Evaluations, Point Beach Units I and 2, S. L Anderson, Radiation Engineering and Analysis, Febnrary 2002.

3. Westinghouse Letter Report, LTR-REA04-464, 'Pressure Vessel Neutron Exposure Evaluations, Point Beach Units 1 and 2, S. L Anderson, Radiation Engineering and Analysis, June 2004.

4. Point Beach Nuckear Plant Technical Requirements Manual Pressure Temperature Limits Report, Section 2.1, "1RCS Pressure and Temperature Limits (LCO 3A.3)", page 2.2-2, Revision 1, dated December 20,2002.

5. NMC Ltter, NRC 2002-0075, 'Responses to Requests for Additional Information, License Amendment Request 226, Measurement Uncertainty Recapture Power Uprate", August 29,2002.

6. Setpoint Document, STPT 5.1, "Primary Control Systems Rod Speed Control", Revision 7.

A 12-4 AREVA Page 43 of 44

BAW-2467NP NPL 2004-0139 June 29, 2004 Page 4

7. Internal PBNP email, Steve Barkhahn to Brad Fromm, dated 4/17/04.

8. Westinghouse, Power Uprate Project, PBNP Units I and 2, Volume 1 NSSS Engineering Report, and Volume 2 BOP Engineering Report, April 2002.

9. Section 4 of Westinghouse Equipment Specification G - 676243, 'Reactor Coolant System -

Reactor Vessel", Revision 0, 05/05/1966.

10. Section 4, and Figurcs 1 through 15 of Westinghousc Equipment Specification E-spec 677456, "Addendum to Equipment Specification 676413 Rev. 1, Reactor Coolant System - Reactor Vessel", Revision 2,07/06/1971.

I l. Appendix A of Westinghouse Design Specification, 414A83, 'Point Beach Nuclear Plants Units I and 2, replacement Reactor Vessel Closure Head (RRVCH)", Revision 0.

12. Appendix B of WCAP-14448, -Addendum to the Stress Reports for tie Point Beach Unit Nos. 1 and 2 Reactor Vessels (RSG/Uprating Evaluation), August 1995.

13. Section 5.1.4 of Westinghouse Report, 'Power Uprate Project, Point Beach Nuclear Plant, Units I and 2, NSSS Engineering Report", April 2002.

14. Chapter 14 of the PBNP Units 1 and 2 Fmal Safety Analysis Report, June, 2003.

Noles:

References 1, 4, 5, 6, 7, 8, and 14 document the sources of the information.

Rcfcrences 2, 3, 9, 10, 11, 12, and 13 are enclosed.

References 9, 10, 11, 12, and 13 are Westinghouse Proprietary and shall be treated in accordance with the associated Westinghouse Proprietary Agreement established between AREVA/Framatome-ANP, NMC, and Westinghouse in June 2004.

A 12-5 AR EVA Page 44 of 44

ML042660313

Text

SUMMARY

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