ML13151A060

From kanterella
Jump to navigation Jump to search
Enclosure C, WCAP-16527-NP, Analysis of Capsule X from Firstenergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program, Supplement 1, Revision 1, Dated September 2011
ML13151A060
Person / Time
Site: Beaver Valley
Issue date: 09/30/2011
From: Freed A
Westinghouse
To:
Office of Nuclear Reactor Regulation
References
L-13-036 WCAP-16527-NP, Suppl 1, Rev. 1
Download: ML13151A060 (36)
v • d • e


Text

Enclosure C L-13-036 WCAP-16527-NP, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

Supplement 1, Revision 1, dated September 2011 (35 Pages Follow)

V Westinghouse Non-Proprietary Class 3 WCAP-16527-NP Supplement 1 September 20O Revision 1 Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program Westinghouse 11

1.

it WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16527-NP Supplement 1 Revision 1 Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program A. E. Freed*

Aging Management & License Renewal Services September 2011 Reviewer:

E. J. Long*

Aging Management & License Renewal Services Approved:

M. G. Semmler*, Acting Manager Aging Management & License Renewal Services

  • Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2011 Westinghouse Electric Company LLC All Rights Reserved

4-WESTINGHOUSE NON-PROPRIETARY CLASS 3 ii RECORD OF REVISIONS Revision 0:

Original Issue Revision 1:

The purpose of this revision is to update the reactor vessel integrity evaluations contained in this document due to updated neutron fluence projections. Furthermore, due to updated surveillance capsule fluence values, the surveillance capsule credibility evaluation has also been updated since the Capsule X analysis and is included in Appendix A of this report.

Note that these reactor vessel integrity calculations, along with the credibility evaluation contained within this document, supersede the previous respective evaluations. Change bars were not used in this document to record the changes between Revisions 0 and I since this revision should be considered an entirely new document based on the nature of the updates.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3

°°° TABLE OF CONTENTS L IST O F TA B L E S.......................................................................................................................................

iv L IST O F FIG U R E S......................................................................................................................................

v EX EC U TIV E SU M M A RY..........................................................................................................................

vi I

IN TR O D U C T IO N........................................................................................................................

1-1 2

PRESSURIZED THERMAL SHOCK RULE..............................................................................

2-1 3

METHODOLOGY FOR CALCULATION OF RTprs AND USE...............................................

3-1 3.1 R T rs................................................................................................................................

3 -1 3.2 U S E..................................................................................................................................

3 -2 4

VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES........................................

4-1 5

NEUTRON FLUENCE VALUES................................................................................................

5-1 6

DETERMINATION OF RTprs AND USE VALUES FOR ALL BELTLINE AND EXTENDED BELTLINE REGION MATERIALS......................................................................

6-1 6.1 BVPS-2 RTprs CALCULATIONS FOR 54 EFPY..........................................................

6-1 6.2 BVPS-2 UPPER-SHELF ENERGY CALCULATIONS FOR 54 EFPY......................... 6-2 7

PRESSURE-TEMPERATURE LIMIT CURVES APPLICABILITY..........................................

7-1 8

SURVEILLANCE CAPSULE REMOVAL SCHEDULE............................................................

8-1 9

C O N C L U SIO N............................................................................................................................

9-1 10 R E FE R E N C E S...........................................................................................................................

10-1 APPENDIX A BEAVER VALLEY UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY EV A L U A T IO N............................................................................................................................

A -1 WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARYCLASS 3 iv LIST OF TABLES Table 4-1 BVPS-2 Reactor Vessel Beltline Material Properties......................................................

4-1 Table 4-2 BVPS-2 Reactor Vessel Extended Beltline Material Properties......................................

4-2 Table 5-1 Maximum Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for B V P S-2.......................................................................................................

5-1 Table 5-2 Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-2 for the Beltline and Extended Beltline Regions................................................

5-2 Table 5-3 Calculated Fluence (E > 1.0 MeV) at the Surveillance Capsule Locations for BVPS-2...

.... °..........................

°o°..... °........ °o..................................................° 5-3 Table 6-1 BVPS-2 Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1........................................................................

6-2 Table 6-2 BVPS-2 Extended Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, R evision 2, Position 1.1..........................................................................................

6-3 Table 6-3 RTpTs Values for BVPS-2 Beltline Region Materials at 54 EFPY...................................

6-4 Table 6-4 RTprs Values for BVPS-2 Extended Beltline Region Materials at 54 EFPY................... 6-5 Table 6-5 BVPS-2 Beltline Materials Projected USE Values at 54 EFPY.................

6-7 Table 6-6 BVPS-2 Extended Beltline Materials Projected USE Values at 54 EFPY....................... 6-8 Table 8-1 Recommended Surveillance Capsule Withdrawal Schedule for BVPS-2........................ 8-1 Table A-I Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit 2 Surveillance Capsule Data Only...............................................................

A-3 Table A-2 Best-Fit Evaluation for Beaver Valley Unit 2 Surveillance Materials Only........

A-4 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V

LIST OF FIGURES Figure 6-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in USE as a Function of Copper and Fluence for B V PS-2................................................................................................

6-10 WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi EXECUTIVE

SUMMARY

The purpose of this supplement is to determine the Reference Temperature for Pressurized Thermal Shock (RTpTs) values and Upper-Shelf Energy (USE) values for the Beaver Valley Unit 2 reactor vessel beltline and extended beltline materials. This analysis will be based upon the results of the latest surveillance capsule X evaluation and the implementation of the Extended Power Uprate program. These calculations are performed for End-of-License Extension (EOLE) at 54 Effective Full Power Years (EFPY).

Furthermore, as part of this analysis, the current pressure-temperature limits along with the surveillance capsule withdrawal schedule and credibility evaluation will be assessed based on the updated fluence results.

The limiting plate material in the Beaver Valley Unit 2 reactor vessel is the upper shell plate B9003-2 with a projected EOLE RTPTS value of 161.6'F for 54 EFPY. This value is below the screening criterion of 270'F for forgings/plates in 10 CFR 50 Part 61. The limiting weld material in the Beaver Valley Unit 2 reactor vessel is the upper shell longitudinal weld (heat BOHB (E-8018)) with an EOLE RTpTs value of 130.4°F. This RTpTs value is well below the screening criteria value of 270'F for axial welds at EOLE (54 EFPY). All of the beltline and extended beltline materials maintain USE above 50 ft-lbs at EOLE.

WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1

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

The predicted decrease in Upper Shelf Energy (USE) is determined as a function of fluence and copper content using either 1) Figure 2 of Regulatory Guide 1.99, Revision 2, Position 1.2, or 2) Surveillance program test results and Figure 2 of Regulatory Guide 1.99, Revision 2, Position 2.2 [Reference 1]. Both methods require the use of the 1/4T (1/4 vessel thickness) vessel fluence.

The purpose of this report is to determine the Reference Temperature for Pressurized Thermal Shock (RTpTs) and USE values for the Beaver Valley Power Station Unit 2 (BVPS-2) reactor vessel using the results of the surveillance Capsule X evaluation and the implementation of the Extended Power Uprate (EPU) Program. The results presented in this report are for End-of-License Extension (EOLE) at 54 Effective Full Power Years (EFPY). Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTpTs and USE. Section 4.0 provides the reactor vessel beltline and extended beltline region material properties for the BVPS-2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTPTs and USE calculations are presented in Section 6.0. The current pressure-temperature limit curve applicability is presented in Section 7.0. The recommended surveillance capsule withdrawal schedule is presented in Section 8.0.

The conclusion and references for the PTS and USE evaluations follow in Sections 9.0 and 10.0, respectively.

The surveillance capsule credibility analysis, based on the results of the surveillance Capsule X analysis and the updated capsule fluence values, is presented in Appendix A of this report.

WCAP-16527-NP Supplement I September 2011 Revision 1

A WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2

PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) amended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness requirements for reactor pressure vessels, including pressurized thermal shock requirements. The revised PTS Rule, 10 CFR Part 50.61, was published in the Federal Register on December 19, 1995, with an effective date of January 18, 1996

[Reference 2].

This amendment to the PTS Rule makes the following changes:

The rule incorporates the method for determining the reference temperature, RTNDT, including treatment of the unirradiated RTNDT value, the margin term, and the explicit definition of "credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 2

[Reference 1].

The rule is restructured to improve clarity, with the requirements section giving only the requirements for the value of the reference temperature for EOL fluence, RTPTS.

Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTpTs.

The PTS Rule requirements consist of the following:

For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTpTs, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material.

The assessment of RTpTs must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTprs for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material.

This assessment must be updated whenever there is significant change in projected values of RTpTs or upon the request for a change in the expiration date for operation of the facility.

Changes to RTpTs values are significant if either the previous value, the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant.

The RTprs screening criterion values for the beltline region are:

- 270'F for plates, forgings and axial weld materials

- 300'F for circumferential weld materials All available surveillance data must be considered in the evaluation.

All credible plant-specific surveillance data must also be used in the evaluation.

WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3

METHODOLOGY FOR CALCULATION OF RTPTs AND USE 3.1 RTPTS RTPTs must be calculated for each vessel beltline material using a fluence value, f, which is the EOL or EOLE fluence for the material. Equation 1 must be used to calculate values of RTNDT for each weld and plate or forging in the reactor vessel beltline.

RTNDT = RTNDT(U) + M + ARTNDT (1)

Where, RTNDT(M) =

Reference Temperature for a reactor vessel material in the pre-service or unirradiated condition M

=

Margin to be added to account for uncertainties in the values of RTNDT(I), copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2 M=2* o' 2 +=o-(2) au is the standard deviation for RTNDTM) cU

=

00F when RTNDTMJ is a measured value u

=

17'F when RTNDTMU) is a generic value Ga is the standard deviation for RTNDT For plates and forgings:

A

=

17'F when surveillance capsule data is not used (T

=

8.5'F when surveillance capsule data is used For welds:

U

=

28°F when surveillance capsule data is not used CA

=

147F when surveillance capsule data is used cA should not exceed one half of ARTNDT ARTNDT is the mean value of the transition temperature shift, or change in ARTNDT, due to irradiation, and must be calculated using Equation 3.

ARTNDT = (CF)

  • f(o.28-o.10logf)

(3)

"CF" ('F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Table 1 for welds and Table 2 for base metal (plates or forgings) of the PTS Rule. Surveillance data deemed credible must be used to determine a material-specific value of CF. A material-specific value of WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 CF is determined in Equation 5.

"f' is the calculated neutron fluence, in units of 1019 n/cm 2 (E > 1.0 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. The EOL or EOLE fluence is used in calculating RTPrs.

Equation 4 must be used for determining RTpTs using Equation 3 with EOL or EOLE fluence values for determining ARTprs.

RTPrs = RTNDT(U) + M + ARTPTs (4)

To verify that RTNDT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes, but is not limited to, the reactor vessel operating temperature and any related surveillance program results. Results from the plant-specific surveillance program must be integrated into the RTNDT estimate if the plant-specific surveillance data has been deemed credible.

A material-specific value of CF is determined from Equation 5.

CF =

-[Ai *f(0.

2 8-0.20ogi)

(]

C

= Z [fi(o.56_o.201ogfi)] I5 In Equation 5, "Ai" is the measured value of ARTNDT and "fi" is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of RTNDT must be adjusted for differences in copper and nickel content. This is done by multiplying them by the ratio of the chemistry factor for the vessel material to that of the surveillance weld.

3.2 USE Per Regulatory Guide 1.99, Revision 2, the Charpy V-notch USE is assumed to decrease as a function of fluence and copper content when surveillance data is not used, as indicated in Figure 2 of the guide.

Linear interpolation is permitted. In addition, if surveillance data is to be used, the decrease in USE may be obtained by plotting the reduced plant surveillance data on Figure 2 of the guide and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph. The USE can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials and/or the results of the capsules tested to date using Figure 2 in Regulatory Guide 1.99, Revision 2.

WCAP-16527-NP Supplement 1 September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4

VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the BVPS-2 vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is defined as "the region of the reactor vessel (shell material including welds, heat-affected zones and plates and forgings) that directly surrounds the effective height of the active 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 addition to the beltline regions, materials that exceed I x 1017 n/cm 2 (E>1.0 MeV) are subject to the guidelines provided in Appendix H of 10 CFR 50 [Reference 3]. In accordance with 10 CFR 50, Appendix H, any materials exceeding Ixl0 17 n/cm 2 (E>1.0 MeV) must be monitored to evaluate the changes in fracture toughness.

Reactor vessel materials not traditionally regarded as plant limiting because of low levels of neutron radiation must now be evaluated to determine the accumulated fluence at 54 EFPY.

Material property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the BVPS-2 surveillance capsule testing program [Reference 4]. The average copper and nickel values were calculated for each beltline and extended beltline region material using all of the available material chemistry information. A summary of the pertinent chemical and mechanical properties of the beltline and extended beltline region forgings/plates and weld material of the BVPS-2 reactor vessel is provided in Tables 4-1 and 4-2.

Table 4-1 BVPS-2 Reactor Vessel Beltline Material Propertiest a)

Material Heat Initial Initial USE Material Description ID Number Wt % Cu Wt % Ni RTNDT(b) ('F)

(ft-lbs)

Intermediate Shell Plate B9004-1 0.065 0.55 60 83 Intermediate Shell Plate B9004-2 0.06 0.57 40 79 Lower Shell Plate B9005-1 0.08 0.58 28 82 Lower Shell Plate B9005-2 0.07 0.57 33 78 Intermediate to Lower Shell 101-171 83642 0.046 0.086

-30 145 Girth Weld Intermediate Shell 101-124 83642 0.046 0.086

-30 145 Longitudinal Welds A&B Lower Shell Longitudinal 101-142 83642 0.046 0.086

-30 145 Welds A&B Surveillance Weld 83642 0.065 0.065 Notes:

a)

Materials information taken from WCAP-16527-NP [Reference 4] and WCAP-16528-NP [Reference 5].

b) All initial RTNDT values are measured values.

WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 Table 4-2 BVPS-2 Reactor Vessel Extended Beltline Material Properties(._

Material Heat Number Wt %

Wt %

Initial Initial USE Material Description ID (Lot Number)

Cu Ni RTNDT(b) (oF)

(ft-lbs)

B9003-1 A9406-1 0.13 0.60 50 98 Upper Shell B9003-2 B4431-2 0.12 0.60 60 80 B9003-3 A9406-2 0.13 0.60 50 98 51912 (3490)

0. 156(d) 0.0591d)

-50 96 101-122A 5

3536) 0.15 0.059M

-70 114 Upper Shell 101-122B EAIB 0.02 0.98 10 (Gen)(c) 118 Longitudinal Welds 101-122C IAGA 0.03 0.98

-30 160 BOHB 0.05 1.00 10 (Gen)(c) 97 BAOED 0.02 1.00

-50 150 Upper Shell to 4P5174 (1122) 0.09 1.0-0-0

-50 70 Intermediate Shell Girth 103-121 51922 (3489) 0.05 1.0 0(g)

-56 (Gen)(c) 102 Weld AAGC 0.03 0.98

-70 111 KOIB 0.03 0.97

-60 110 B9011-1 2V2436-01-002 0.11 0.85 60 10568.25)e)

Inlet Nozzles B9011-2 2V2437-02-001 0.13P) 0.88 60 (Gen)7')

116 (75.4 e)

B9011-3 2V2445-02-003

0. 1P) 0.84 70 122 (79.3) e 4P5174 (1122) 0.09 1.0019)

-50 70 LOHB 0.03 1.03

-60 137 HABJC 0.02 1.02

-70 162 BABBD 0.02 1.04

-70 142 105-121B FABGC 0.03 1.02

-80 119 105-12 EOBC 0.02 0.96

-60 127 FAAFC 0.07 1.04

-60 119 CCJC 0.02 0.99

-60 109 FAGB 0.02 1.06

-30 114 BAOED 0.02 1.00

-50 150 B9012-1 AV8080-2E9558 0.13")

0.72

-10 146 Outlet Nozzles B9012-2 AV8120-2E9560 0.13")

0.74

-10 128 B9012-3 AV8097-2E9559 0.13V) 0.70

-10 127 BABBD 0.02 1.04

-70 142 FAAFC 0.07 1.04

-60 119 107-121A HAAEC 0.03 1.03

-80 178 Outlet Nozzle Welds 107-121B HABJC 0.02 1.02

-70 162 107-121C HAGB 0.02 1.04

-40 194 GACJC 0.03 1.00

-80 113 1 _

JAHB 0.03 0.97

-40 149 Notes:

a)

All of the materials data is obtained from Combustion Engineering report MISC-PENG-ER-021 [Reference 6] except as noted.

b)

Initial RTNDT values are measured values, unless otherwise noted.

c)

The generic initial RTNDT values were determined in accordance with NUREG-0800 [Reference 8] and 10 CFR 50.61 [Reference 2].

d) Chemistry obtained from CE Report NPSD-1039, Revision 2 [Reference 7].

e)

Value in parenthesis is the 65% value per NUREG-0800 [Reference 8].

f)

The Cu wt% was not available from the CMTR so in accordance with Regulatory Guide 1.99, Rev. 2

[Reference 1], a standard deviation analysis (average + standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Nozzle Forgings (178 data points).

g)

Default Wt % Ni content per Regulatory Guide 1.99, Revision 2.

WCAP-16527-NP Supplement 1 September2011 WCAP-I 6527-NP Supplement I September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5

NEUTRON FLUENCE VALUES The maximum neutron exposures at the pressure vessel clad/base metal interface at azimuthal angles of 00, 15', 30', and 450 relative to the core major axes are presented in Table 5-1. The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the BVPS-2 reactor vessel are shown in Table 5-2 for the beltline and extended beltline materials. The calculated fast neutron fluence (E > 1.0 MeV) values at the radial and azimuthal center of the surveillance capsule positions, 170 and 200, are presented in Table 5-3. The fluence projections were determined using ENDF/B-VI cross sections and are based on the results of the Capsule X radiation analysis and comply with Reg. Guide 1.190 [Reference 9].

These fluence data tabulations include fuel-cycle-specific calculated neutron exposures at the end of Cycle 14 as well as future projections to the end of Cycle 15 and for several intervals extending to 60 EFPY.

Neutron exposure projections beyond the end of Cycle 15 were based on the spatial power distributions and associated plant characteristics of Cycles 14 and 15 in conjunction with an uprated core power level of 2900 MWt.

Table 5-1 Maximum Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-2 Neutron Fluence (E > 1.0 MeV)

Cumulative Irradiation

[n/cm 2]

Cycle Time [EFPY]

00 150 300 450 1

1.25 1.94 x 101 1.08 x 1018 7.99 x 1017 5.46 x 1017 2

2.26 3.00 x 1018 1.75 x 1018 1.30 x 1018 8.87 x 1017 3

3.50 4.33 x 10Is 2.64 x 1018 2.00 x 1018 1.39 x 1018 4

4.76 5.80 x 1018 3.56 x 10I 2.67 x 1018 1.82 x 1018 5

5.99 7.25 x 1018 4.44 x 1018 3.28 x 1018 2.23 x 1018 6

7.24 8.50 x 1018 5.27 x 10t8 3.99 x 10l 2.74 x 10ls 7

8.51 9.83 x 1018 6.11 x 1018 4.68 x 10l 3.26 x 1018 8

9.84 1.11 x 1019 6.94 x 1018 5.34 x 1018 3.74 x 101" 9

11.06 1.24 x 10' 9 7.67 x 1018 5.89 x 10I8 4.17 x 10Is 10 12.56 1.38 x 10'9 8.56 x 101 6.59 x 1018 4.67 x 1018 11 14.00 1.52 x 10' 9 9.45 x 1018 7.24 x 1018 5.11 x 1018 12 15.41 1.66 x 10' 9 1.03 x 1019 7.84 x 1018 5.55 x 10's 13 16.82 1.82 x 10' 9 1.11 x 1019 8.46 x 1018 5.98 x 1018 14 18.18 1.95 x 1019 1.19 x 1019 9.06 x 1018 6.40 x 1018 15 19.44 2.06 x 1019 1.26 x 10'9 9.60 x 1018 6.80 x 101" Future 32.00 3.21 x 10t9 1.97 x 10'9 1.51 x 1019 1.08 x 10' 9 Future 48.00 4.67 x 1019 2.88 x 10' 9 2.21 x 10' 9 1.58 x 10' 9 Future 54.00 5.21 x 1019 3.22 x 10 19 2.47 x 1019 1.77 x 1019 Future 60.00 5.76 x 1019 3.56 x 1019 2.73 x 10'9 1.96 x 1019 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-2 Table 5-2 Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-2 for the Beltline and Extended Beltline Regions Neutron Fluence (E > 1.0 MeV)

In/cm 2]

Material 11 11 11 1ý 11 Extended Beltline Materials Upper Shell Plates 2.97 x 1018 4.56 x 10'8 5.15 x 1018 5.75 x 1018 Upper to Intermediate Shell Girth W eld 2.97 x 10i8 4.56 x 1018 5.15 x 1018 5.75 x 101s Girth Weld In let Nozzle to Upper Shell 1.70 x 1017 2.63 x 1017 2.98 x 1017 3.33 x 1017 Weld - Lowest Extent OutletNozzletoUpper 8.60 x 1016 1.34 x 1017 1.51 x 1017 1.69 x 1017 Shell Weld - Lowest Extent Lower Shell to Lower Closure Head Weld(a) 4.11 x 10 6.19x 5

6.97x 10i5 7.74x 1015 Notes:

a)

Extended beltline materials are currently interpreted to be the reactor vessel materials that will be exposed to a neutron fluence greater than or equal to 1 x 1017 n/cm2 (E > 1.0 MeV) at EOLE. Since the fluence for the lower shell to closure head weld material is less than 1 x 1017 n/cm2, this material has been omitted from the calculations contained in this report.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 Table 5-3 Calculated Fluence (E > 1.0 MeV) at the Surveillance Capsule Locations for BVPS-2 Neutron Fluence (E > 1.0 MeV)

Cumulative in/cm 2]

Cycle Irradiation Time

[EFPY]

170 200 1

1.25 6.15 x 1018 5.25 x 1018 2

2.26 1.02 x 10' 9 8.74 x 1018 3

3.50 1.57 x 10' 9 1.37 x 10' 9 4

4.76 2.12 x 10'9 1.85 x 1019 5

5.99 2.64 x 10'9 2.29 x 1019 6

7.24 3.14 x 10'9 2.74 x 10'9 7

8.51 3.63 x 1019 3.17 x 1019 8

9.84 4.13 x 1019 3.61 x 10'9 9

11.06 4.55 x 10'9 3.98 x 1019 10 12.56 5.09 x 1019 4.45 x 10'9 I11 14.00 5.63 x 10'9 4.91 x 1019 12 15.41 6.11 x 10'9 5.32 x 1019 13 16.82 6.63 x 1019 5.76 x 1019 14 18.18 7.10x 1019 6.16x 1019 15 19.44 7.53 x 10' 9 6.53 x 10' 9 Future 32.00 1.18 x 1020 1.03 x 1020 Future 48.00 1.73 x 1020 1.50 x 1020 Future 54.00 1.94 x 1020 1.68 x 1020 Future 60.00 2.14 x 1020 1.85 x 1020 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6

DETERMINATION OF RTPTS AND USE VALUES FOR ALL BELTLINE AND EXTENDED BELTLINE REGION MATERIALS 6.1 BVPS-2 RTPTS CALCULATIONS FOR 54 EFPY Using the prescribed PTS Rule methodology, RTpTs values were generated for all beltline and extended beltline region materials of the BVPS-2 reactor vessel for fluence values at EOLE (54 EFPY).

Each plant shall assess the RTPTS values based on plant-specific surveillance capsule data. For BVPS-2, the related surveillance program results have been included in this PTS evaluation. Specifically, the BVPS-2 plant-specific surveillance capsule data for the intermediate shell (IS) plate B9004-2 and weld metal (heat 83642) is provided and applied as follows:

1) There have been four capsules removed from the BVPS-2 reactor vessel.
2) The data for the surveillance program plate material is deemed non-credible. The data was used with a aA margin of 17°F.
3) The data for the Unit 2 surveillance program weld material is deemed credible. The data was used with a aA margin of 14'F.
4) The surveillance capsule materials are representative of the actual vessel plate and all of the beltline welds (weld heat 83642).
5) The resulting RTPTS values for intermediate shell plate B9004-2 remains below the screening criteria at 54 EFPY based on Positions 1.1 and 2.1 of Regulatory Guide 1.99, Revision 2. The resulting RTpTs values for all other materials remains below the PTS Rule screening criteria at 54 EFPY.

Chemistry factor values for the BVPS-2 beltline region materials based on Position 1.1 and 2.1 from Regulatory Guide 1.99, Revision 2, are presented in Table 6-1. Additionally, chemistry factor values for the BVPS-2 extended beltline materials based on Position 1.1 and 2.1 from Regulatory Guide 1.99, Revision 2, are presented in Table 6-2. Tables 6-3 and 6-4 contain the RTPTS calculations for all beltline and extended beltline region materials at 54 EFPY.

WCAP-16527-NP Supplement 1 September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 6.2 BVPS-2 UPPER-SHELF ENERGY CALCULATIONS FOR 54 EFPY Surveillance data exists for plate B9004-2 and weld heat 83642 for BVPS-2. Each of the measured drops in USE for each of these material heats is plotted on Figure 2 of Regulatory Guide 1.99, Revision 2 with a horizontal line drawn parallel to the existing lines as the upper bound of all data. Figure 6-1 was used in the determination of the percent decrease in USE for the beltline and extended beltline materials. Tables 6-5 and 6-6 document the USE values for all of the materials at 54 EFPY. All of the beltline or extended beltline material USE values maintain 50 ft-lbs or greater at 54 EFPY.

Table 6-1 BVPS-2 Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material Heat Chemistry Factor Material Description ID Number (OF)

Position 1.1 Position 2.1 Intermediate Shell Plate B9004-1 40.5 Intermediate Shell Plate B9004-2 37 51.4 Lower Shell Plate B9005-1 51 Lower Shell Plate B9005-2 44 101-142' Intermediate to Lower Shell Girth Weld A14 83642 34.4 12.5 A&B Intermediate Shell Longitudinal Welds 101-124 83642 34,4 12.5 A&B Lower Shell Longitudinal Welds 101-171 83642 34.4 12.5 Surveillance Weld 83642 38 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-2 BVPS-2 Extended Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 Material Description Material ID Heat Number Position 1.1 Chemistry (Lot Number)

Factor (OF)

B9003-1 A9406-1 91.0 Upper Shell Plates B9003-2 B4431-2 83.0 B9003-3 A9406-2 91.0 51912 (3490) 73.71 51912 (3536) 73.71 Upper Shell 101-122AEAB2.

EAIB 27.0 Longitudinal 101-122B Welds 101-122C BOHB 68.0 BAOED 27.0 4P5174 (1122) 122.0 Upper to Intermediate 103-121 51922 (3489) 68.0 Shell Girth Weld AAGC 41.0 KOIB 41.0 B901 1-1 2V2436-01-002 77.0 Inlet Nozzles B9011-2 2V2437-02-001 96.0 B9011-3 2V2445-02-003 96.0 4P5174 (1122) 122.0 LOHB 41.0 HABJC 27.0 BABBD 27.0 105-121A FABGC 41.0 Inlet Nozzle Welds 105-121 B 105-121C EOBC 27.0 FAAFC 95.0 CCJC 27.0 FAGB 27.0 BAOED 27.0 B9012-1 AV8080-2E9558 94.0 Outlet Nozzles B9012-2 AV8120-2E9560 94.5 B9012-3 AV8097-2E9559 93.5 BABBD 27.0 FAAFC 95.0 107-121A HAAEC 41.0 Outlet Nozzle Welds 107-12 1B HABJC 27.0 107-121C HAGB 27.0 GACJC 41.0 JAHB 41.0 WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 Table 6-3 RTPrs Values for BVPS-2 Beitline Region Materials at 54 EFPY Surface Material Heat Neutron Fluence Chemistry Initial ART

(

A Margin RT Material Description ID Number Fluence

Factor, Factor RTNDT(b)

(O)

(

(F (O)

(O)

(NJ

(

1O'n/cm)

F08)

(OF)

(OF)

( 0F)

( 0F)

( 0F)

(°F)

( 0F)

Intermediate Shell Plate B9004-1 5.18 1.4092 40.5 60 57.1 0

17 34 151.1 Intermediate Shell Plate B9004-2 5.18 1.4092 37 40 52.1 0

17 34 126.1 UsingUsing non-credible surveillance data-1 5.18 1.4092 51.4 40 72.4 0

17 34 146.4 Lower Shell Plate B9005-1 5.21 1.4104 51 28 71.9 0

17 34 133.9 Lower Shell Plate B9005-2 5.21 1.4104 44 33 62.1 0

17 34 129.1 Intermediate to Lower Shell 101-171 83642 5.18 1.4092 34.4

-30 48.5 0

24.2 48.5 67.0 Girth Weld Using credible surveillance datat0 5.18 1.4092 12.5

-30 17.6 0

8.8 17.6 5.2 Intermediate Shell 101-124 83642 1.76 1.1554 34.4

-30 39.7 0

19.9 39.7 49.5 Longitudinal Welds I

A&B Using credible surveillance data(0 1.76 1.1554 12.5

-30 14.4 0

7.2 14.4

-1.1 Lower Shell Longitudinal 101-142 83642 1.77 1.1569 34.4

-30 39.8 0

19.9 39.8 49.6 Welds A&B Using credible surveillance data*0 1.77 1.1569 12.5

-30 14.5 0

7.2 14.5

-1.1 Notes:

a)

FF = fluence factor = f(O 2 8 -o.1 log(f))

b)

Initial RTNDT values are measured values.

c)

ARTprs = CF

d)

M-=2*(u_2 + *a2)1v2.

e)

RTPTS = Initial RTNDT + ARTPTS + Margin.

f)

The BVPS-2 surveillance weld metal is the same weld heat as the BVPS-2 beltline welds (heat 83642). The BVPS-2 surveillance weld data is credible (see Appendix A); therefore, the reduced Ga term of 14'F was utilized for BVPS-2 weld heat 83642.

The BVPS-2 surveillance plate material is representative of the BVPS-2 intermediate shell plate B9004-2. The surveillance plate material is non-credible (see Appendix A);

therefore, the higher GA term of 177F was utilized for BVPS-2 plate B9004-2.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 Table 6-4 RTp-rs Values for BVPS-2 Extended Beltline Region Materials at 54 EFPY Surface Material Huron Fluence Chemistry Initial Deripo Material ID Heat Number Neutron

Factor, Factor RTNDT(b)

ARTm(d) aY GA Margin(e)

RTMS 0

Description (Lot Number)

Fluence F*

(OF)

(OF)

(OF)

(OF)

(F (xl 019 n/cm 2)

FF(a)

(OF)

(OF)

(OF)

B9003-1 A9406-1 0.515 0.8147 91.0 50 74.1 0

17 34 158.1 Upper Shell Pper B9003-2 B4431-2 0.515 0.8147 83.0 60 67.6 0

17 34 161.6 Plates B9003-3 A9406-2 0.515 0.8147 91.0 50 74.1 0

17 34 158.1 51912 (3490) 0.515 0.8147 73.71

-50 60.1 0

28 56 66.1 51912 (3536) 0.515 0.8147 73.71

-70 60.1 0

28 56 46.1 UpperLShell 101-122A EAIB 0.515 0.8147 27.0 10(c) 22.0 17 11.0 40.5 72.5 Longitudinal 101-1 22B Welds 101-122C IAGA 0.515 0.8147 41.0

-30 33.4 0

16.7 33.4 36.8 BOHB 0.515 0.8147 68.0 10(c) 55.4 17 27.7 65.0 130.4 BAOED 0.515 0.8147 27.0

-50 22.0 0

11.0 22.0

-6.0 4P5174 0.515 0.8147 122.0

-50 99.4 0

28 56.0 105.4 Upper to 51922 0.515 0.8147 68.0

-56(c) 55.4 17 27.7 65.0 64.4 Intermediate 103-121 Shell Girth Weld AAGC 0.515 0.8147 41.0

-70 33.4 0

16.7 33.4

-3.2 KOIB 0.515 0.8147 41.0

-60 33.4 0

16.7 33.4 6.8 B9011-1 2V2436-01-002 0.0298 0.2188 77.0 60 16.8 0

8.4 16.8 93.7 Inlet Nozzles B9011-2 2V2437-02-001 0.0298 0.2188 96.0 60(c) 21.0 17 10.5 40.0 121.0 B9011-3 2V2445-02-003 0.0298 0.2188 96.0 70 21.0 0

10.5 21.0 112.0 4P5174 0.0298 0.2188 122.0

-50 26.7 0

13.3 26.7 3.4 LOHB 0.0298 0.2188 41.0

-60 9.0 0

4.5 9.0

-42.1 Inlet Nozzle 105-121A HABJC 0.0298 0.2188 27.0

-70 5.9 0

3.0 5.9

-58.2 105-121B Welds 105-121C BABBD 0.0298 0.2188 27.0

-70 5.9 0

3.0 5.9

-58.2 FABGC 0.0298 0.2188 41.0

-80 9.0 0

4.5 9.0

-62.1 EOBC 0.0298 0.2188 27.0

-60 5.9 0

3.0 5.9

-48.2 WCAP-16527-NP Supplement I September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 Table 6-4 RTpms Values for BVPS-2 Extended Beitline Region Materials at 54 EFPY Surface Fluence Chemistry Initial ATpTs(d) 0 U GA Marginse)

RT lrs 0

Material Heat Number Neutron

Factor, Factor RT4DT7b)

((e)

Description Material ID (Lot Number)

Fluence (OF)

(OF)

(OF)

(OF)

(OF)

(xlO' 9 n/cm2)

FOB)

(OF)

(OF)

FAAFC 0.0298 0.2188 95.0

-60 20.8 0

10.4 20.8

-18.4 CCJC 0.0298 0.2188 27.0

-60 5.9 0

3.0 5.9

-48.2 FAGB 0.0298 0.2188 27.0

-30 5.9 0

3.0 5.9

-18.2 BAOED 0.0298 0.2188 27.0

-50 5.9 0

3.0 5.9

-38.2 AV8080-B9012-1 E98 0.0151 0.1440 94.0

-10 13.5 0

6.8 13.5 17.1 2139558 Outlet Nozzles B9012-2 AV8120-0.0151 0.1440 94.5

-10 13.6 0

6.8 13.6 17.2 2E9560 AV8097-B9012-3 2E9559 0.0151 0.1440 93.5

-10 13.5 0

6.7 13.5 16.9 BABBD 0.0151 0.1440 27.0.

-70 3.9 0

1.9 3.9

-62.2 FAAFC 0.0151 0.1440 95.0

-60 13.7 0

6.8 13.7

-32.6 107-121A HAAEC 0.0151 0.1440 41.0

-80 5.9 0

3.0 5.9

-68.2 O ele 107-121B HABJC 0.0151 0.1440 27.0

-70 3.9 0

1.9 3.9

-62.2 Welds 107-121C HAGB 0.0151 0.1440 27.0

-40 3.9 0

1.9 3.9

-32.2 GACJC 0.0151 0.1440 41.0

-80 5.9 0

3.0 5.9

-68.2 JAHB 0.0151 0.1440 41.0

-40 5.9 0

3.0 5.9

-28.2 Notes:

a)

FF = fluence factor = f(O.28 -0.1 log (0) b)

Initial RTNDT values are measured values, unless otherwise noted.

c)

Initial RTNDT values are generic.

d)

ARTPTS = CF

e)

M=2*(_u2 +

A2)1/2.

f)

RTPTS Initial RTNDT + ARTPTS + Margin.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7 Table 6-5 BVPS-2 Beltline Materials Projected USE Values at 54 EFPY I/4T EOLE Initial Projected Projected Material Material Heat Wt %

USE EOLE Description ID Number Cu (lX019 n/cm2)

(ft-lbs)

Decrease USE (xl' 9n/m 2

(f-ls)

(M (ft-lbs)

Intermediate Shell B9004-1 0.065 3.229 83 25 62.3 Plate Intermediate Shell B9004-2 0.06 3.229 79 13(b) 68.7 Plate Lower Shell Plate B9005-1 0.08 3.248 82 25 61.5 Lower Shell Plate B9005-2 0.07 3.248 78 25 58.5 Intermediate to Lower Shell Girth 101-171 83642 0.046 3.229 145 5.8(c) 136.6 Weld Intermediate Shell 101-124 83642 0.046 1.097 145 4.6(c) 138.3 Longitudinal Welds A&B LowerShell 101-142 83642 0.046 1.103 145 4.6(c) 138.3 Longitudinal Welds A&B Notes:

a)

Unless otherwise noted, percent USE decreases are based on the closest Cu Wt. % chemistry line (rounding up) on Figure 2 of Regulatory Guide 1.99, Revision 2.

b)

Based on BVPS-2 surveillance plate results for B9004-2 [Reference 4].

c)

Based on BVPS-2 surveillance weld results for heat 83642 [Reference 4].

WCAP-16527-NP Supplement I September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-8 Table 6-6 BVPS-2 Extended Beltline Materials Projected USE Values at 54 EFPY 1/4T EOLE Initial Projected Projected Material Material Heat Number Wt %

Fluence USE USE EOLE USE Description ID (Lot Number)

Cu (X1019 n/cm 2)

(ft-lbs)

Decrease()

(ft-lbs)

B9003-1 A9406-1 0.13 0.321 98 18.5 79.9 Pper B9003-2 B4431-2 0.12 0.321 80 18.5 65.2 Plates B9003-3 A9406-2 0.13 0.321 98 18.5 79.9 51912 (3490) 0.156 0.321 96 26 71.0 51912(3536) 0.156 0.321 114 26 84.4 Upper Shell 101-122A EAIB 0.02 0.321 118 14.5 100.9 Longitudinal 101-122B Welds 101-122C IAGA 0.03 0.321 160 14.5 136.8 BOHB 0.05 0.321 97 14.5 82.9 BAOED 0.02 0.321 150 14.5 128.3 Upper to 4P5174 0.09 0.321 70 18.5 57.1 Intermediate 51922 0.05 0.321 102 14.5 87.2 103-12 1 Shell AAGC 0.03 0.321 111 14.5 94.9 Girth Weld KOIB 0.03 0.321 110 14.5 94.1 B9011-1 2V2436-01-002 0.11 0.019 68.25 9.5 61.8 Inlet Nozzles B9011-2 2V2437-02-001 0.13 0.019 75.4 9.5 68.2 B9011-3 2V2445-02-003 0.13 0.019 79.3 9.5 71.8 4P5174 0.09 0.019 70 9.5 63.4 LOHB 0.03 0.019 137 7.5 126.7 HABJC 0.02 0.019 162 7.5 149.9 BABBD 0.02 0.019 142 7.5 131.4 Inlet Nozzle 105-121A FABGC 0.03 0.019 119 7.5 110.1 105-121B Welds 105-121C EOBC 0.02 0.019 127 7.5 117.5 FAAFC 0.07 0.019 119 9.5 107.7 CCJC 0.02 0.019 109 7.5 100.8 FAGB 0.02 0.019 114 7.5 105.5 BAOED 0.02 0.019 150 7.5 138.8 B9012-1 AV8080-0.13 0.009 146 9.5 132.1 2E9558 Outlet AV8 120-Oztles B9012-2 2E9560 0.13 0.009 128 9.5 115.8 Nozzles 2E9560 AVS097-B9012-3 2E975 0.13 0.009 127 9.5 114.9 2E9559 BABBD 0.02 0.009 142 7.5 131.4 FAAFC 0.07 0.009 119 9.5 107.7 107-121A HAAEC 0.03 0.009 178 7.5 164.7 Outlet Nozzle 107-121B HABJC 0.02 0.009 162 7.5 149.9 Welds 107-121C HAGB 0.02 0.009 194 7.5 179.5 GACJC 0.03 0.009 113 7.5 104.5 JAHB 0.03 0.009 149 7.5 137.8 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-9 Notes:

a)

The lower line in Figure 2 of Regulatory Guide 1.99, Revision 2, was used as the bounding value when Wt % Cu values were below this limit for plates and/or welds.

b)

The minimum fluence value (2 x 1017 n/cm2) displayed on Figure 2 of Regulatory Guide 1.99, Revision 2 was conservatively used to determine the projected USE decrease values for the inlet/outlet nozzle forgings and welds.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-10 100.0 ESurveillance Material: Weld Heat# 83642 A Qiita,=illnn

ýKn fah~~trinl. IQ* 13Iýfft CI::

A_'ýl 0.

a.

10.0 Plate line Weld line 1.00E+20 1.0 -I-1.OOE+17 1.OOE+18 1.OOE+19 Neutron Fluence, n/cm2 (E > I MeV)

Figure 6-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in USE as a Function of Copper and Fluence for BVPS-2 WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7

PRESSURE-TEMPERATURE LIMIT CURVES APPLICABILITY Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT corresponding to the limiting material in the beltline region of the reactor pressure vessel. The most limiting RTNDT of the material in the core (beltline) region of the reactor pressure vessel is determined by using the unirradiated reactor pressure vessel material fracture toughness properties and estimating the irradiation-induced shift (ARTNDT).

RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. Using the adjusted reference temperature (ART) values, pressure-temperature (P-T) limit curves are determined in accordance with Appendix G to Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code

[Reference 10] as specified by CFR Part 50, Appendix G [Reference I I].

The BVPS-2 P-T limit curves for normal heatup and cooldown of the primary reactor coolant system were previously developed in WCAP-16528-NP, Revision 1 [Reference 5], for 22 EFPY, 30 EFPY, 40, and 54 EFPY. Currently, BVPS-2 is operating to their 22 EFPY curves. The existing 22 EFPY P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material.

The BVPS-2 22 EFPY P-T limit curves were developed by calculating ART values utilizing the maximum clad/base metal fluence across all beltline materials. The limiting ART values used in the development of the 22 EFPY P-T limit curves correspond to the intermediate shell plate B9004-1 (Position 1.1 - without using surveillance data).

Taking into account the updated fluence values, as shown in Section 5 of this report, as well as the updated Position 2.1 chemistry factor values in Section 6, the intermediate shell plate B9004-1 continues to be the limiting material for the current BVPS-2 P-T limit curves. Additionally, the BVPS-2 updated vessel and surveillance capsule fluence values do not reduce the existing 22 EFPY applicability term for which the P-T limit curves were originally developed; therefore, the existing P-T limit curves remain valid as documented in Reference 5 for BVPS-2.

WCAP-16527-NP Supplement 1 September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8

SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82

[Reference 12] and is recommended for future capsules to be withdrawn from the BVPS-2 reactor vessel.

Table 8-1 Recommended Surveillance Capsule Withdrawal Schedule for BVPS-2 Capsule Capsule Location Lead Factor(a)

Withdrawal EFPY(b)

Fluence, E10 e

U 3430 3.17 1.25 6.15 x 10" V

1070 3.64 5.99 2.64 x 10'9 W

1100 3.25 9.84 3.61 x 1019 X

2870 3.69 14.00 5.63 x 10'9 Y(C) 2900 3.17 (c) 6.53 x I1019(c)

Z(C) 3400 3.17 (c) 6.53 x 10'9(c)

A(d) 1070 3.58 3 2 (d) 3.40 x 1 0 19(d)

Notes:

a)

Updated in recent fluence analysis; see Section 5 of this report.

b)

Effective Full Power Years (EFPY) from plant startup.

c)

Either Capsule Y or Z is to be withdrawn in accordance with the recommendations in ASTM E 185-82 for a 60-year and 80-year license renewal.

Accumulated fluence value through EOC 15.

It is recommended that one of these capsules be pulled at a time when the capsule fluence exceeds one times the projected peak 80-year EOL fluence, but before the capsule fluence reaches two times the projected 60-year EOL fluence. Therefore, based on the current information, Capsule Y or Z should be withdrawn between 20.5 EFPY and 32.4 EFPY. In order to be consistent with NUREG-1929

[Reference 13], this capsule should be pulled at the refueling outage closest to 26.1 EFPY, which corresponds with a peak vessel fluence level of 8.5 x 10'9 n/cm2 (E > 1.0 MeV). The second capsule will remain in the vessel to provide fluence monitoring or future testing. The withdrawal of this capsule will be revisited at a later time.

d)

Supplemental Capsule A contains BVPS-1 initially unirradiated material specimens, as well as previously irradiated material specimens from BVPS-1, St. Lucie, and Fort Calhoun. Supplemental Capsule A was inserted in the vacant Capsule V location at the EOC 8. Accumulated fluence value through EOC 15. It is to be removed and tested when it reaches a fluence value equivalent to the peak 80-year vessel fluence for BVPS-1. This is expected to occur at approximately 32 EFPY.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9

CONCLUSION

" All of the beltline and extended beltline region materials in the BVPS-2 reactor vessel have EOLE RTPTS values well below the screening criteria values of 270'F for forgings/plates and 300'F for circumferential welds at EOLE (54 EFPY).

" All of the USE values for the beltline and extended beltline materials are greater than 50 ft-lbs at EOLE (54 EFPY).

The current P-T limit curves remain valid through 22 EFPY for BVPS-2.

  • Four surveillance capsules have been withdrawn and tested from BVPS-2. With the withdrawal of these four capsules, BVPS-2 has satisfied the current requirements for their 40-year license.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1 10 REFERENCES

1.

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

Nuclear Regulatory Commission, May 1988.

2.

Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

3.

Code of Federal Regulations, 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal Register, Volume 60, No. 243, dated December 19, 1995.

4.

WCAP-16527-NP, Revision 0, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," B.N. Burgos, et.

al., March 2006.

5.

WCAP-16528-NP, Revision 1, "Beaver Valley Power Station Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," N. R. Jurcevich, June 2008.

6.

Combustion Engineering Report MISC-PENG-ER-021, Revision 00, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the Beaver Valley Unit 2 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," J.P. Molkenthin, et. al., October 1995.

7.

"Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CEOG Report CE NPSD-1039, Revision 2, ABB Combustion Engineering, June 1997.

8.

"Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"

NUREG-0800, MTEB 5-2 and 5-3, June 1987.

9.

Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

10.

Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."

11.

Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements," U.S.

Nuclear Regulatory Commission, Washington D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

12.

ASTM El185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society of Testing and Materials, 1982.

13.

NUREG-1929, "Safety Evaluation Report Related to the License Renewal of Beaver Valley Power Station, Units I and 2 (Docket Nos. 50-334 and 50-412) FirstEnergy Nuclear Operating Company," October 2009 (ADAMS Accession Number ML093020276).

WCAP-16527-NP Supplement I September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A BEAVER VALLEY UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION INTRODUCTION Regulatory Guide 1.99, Revision 2 [Reference A-i ] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, there have been four surveillance capsules removed and tested from the BVPS-2 reactor vessel.

To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the BVPS-2 reactor vessel surveillance data in order to determine if that surveillance data is credible.

EVALUATION Criterion 1:

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," [Reference A-2] as follows:

"the region of the reactor vessel (shell material including welds, heat affected zones, and plates orforgings) that directly surrounds the effective height of the active 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."

The BVPS-2 reactor vessel beltline region consists of the following materials:

1.

Intermediate Shell Plates B9004-1 and B9004-2

2.

Lower Shell Plates B9005-1 and B9005-2

3.

Intermediate Shell Longitudinal Welds 101-124 A&B (Heat 83642)

4.

Intermediate to Lower Shell Girth Weld 101-171 (Heat 83642)

5.

Lower Shell Longitudinal Welds 101-142 A&B (Heat 83642)

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Per WCAP-9615, Revision 1 [Reference A-3], at the time the BVPS-2 surveillance program was being developed, selection of the surveillance material was based on an evaluation of initial toughness (characterized by the reference temperature, RTNDT, and upper-shelf energies), the predicted effect of chemical composition (residual copper and phosphorus) and neutron fluence on the toughness (RTNDT shift) during reactor operation. Intermediate shell plate B9004-2 was selected as the surveillance base metal since it had one of the highest adjusted EOL RTNDT of the four beltline region plates. Weld heat 83642 was selected because it is the same heat used in the fabrication of all of the longitudinal and girth beltline welds.

Based on the discussion above, Criterion 1 is met for the BVPS-2 surveillance program.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP-16527-NP, Revision 0

[Reference A-4].

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the BVPS-2 surveillance materials unambiguously.

Hence, Criterion 2 is met for the BVPS-2 surveillance orogram.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 287F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82 [Reference A-5].

The functional form of the least-squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28°F for the weld and less than 17'F for the plate.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [Reference A-6]. At this meeting the NRC presented five cases. Of the five cases, Case 1

("Surveillance data available from plant but no other source") most closely represents the situation for the BVPS-2 surveillance plate and weld material.

WCAP-16527-NP Supplement 1 September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 Following the NRC Case 1 guidelines, the BVPS-2 surveillance plate and weld metal (heat 83642) will be evaluated using the BVPS-2 data. This evaluation is contained in Table A-1. Note that when evaluating the credibility of the surveillance weld data, the measured ARTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only BVPS-2 data is being considered; therefore, no temperature adjustment is required.

Table A-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit 2 Surveillance Capsule Data Only Capsule f ARTNDT FF*ARTNDT Material Capsule (xlO' 9 n/cm2, E FF (OF)

(OF)

FF2

> 1.0 MeV)

U 0.615 0.864 24.0 20.73 0.746 Intermediate Shell V

2.64 1.260 56.0 70.54 1.587 Plate B9004-2 (Longitudinal)

W 3.61 1.334 71.0 94.68 1.778 X

5.63 1.425 98.0 139.65 2.031 U

0.615 0.864 17.7 15.29 0.746 Intermediate Shell V

2.64 1.260 46.1 58.07 1.587 Plate B9004-2 (Transverse)

W 3.61 1.334 63.4 84.55 1.778 X

5.63 1.425 104.1 148.34 2.031 SUM:

631.87 12.284 CFis Plate B9004-2 = (FF

  • ARTNDT)

I E(FF 2) = (631.87)

(12.284) = 51.4-F U

0.615 0.864 4.1 3.54 0.746 V

2.64 1.260 25.7 32.37 1.587 Beaver Valley Unit W

3.61 1.334 6.0 8.00 1.778 2 Weld Metal (Heat 83642) 5.63 1.425 22.9 32.63 2.031 SUM:

76.55 6.142 CF8 3642= E(FF

  • ARTNDT) + E(FF 2) = (76.55) + (6.142) = 12.5°F WCAP-16527-NP Supplement 1 September 2011 WCAP-16527-NP Supplement I September 2011 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-4 The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A-2.

Table A-2 Best-Fit Evaluation for Beaver Valley Unit 2 Surveillance Materials Only CF Capsule f Measured Predicted Scatter

<17 0F Material Capsule (Slopebt-nrt)

(xlO' 9 n/cm 2, FF ARTNDT ARTNDT ARTNDT (Base Metal)

(OF)

E > 1.0 MeV)

(OF)

(OF)

(OF)

<28*F (Weld)

U 51.4 0.615 0.864 24.0 44.4 20.4 No Intermediate Shell V

51.4 2.64 1.260 56.0 64.8 8.8 Yes Plate B9004-2 (Longitudinal)

W 51.4 3.61 1.334 71.0 68.6 2.4 Yes X

51.4 5.63 1.425 98.0 73.3 24.7 No U

51.4 0.615 0.864 17.7 44.4 26.7 No Intermediate Shell V

51.4 2.64 1.260 46.1 64.8 18.7 No Plate B9004-2 (Transverse)

W 51.4 3.61 1.334 63.4 68.6 5.2 Yes X

51.4 5.63 1.425 104.1 73.3 30.8 No U

12.5 0.615 0.864 4.1 10.8 6.7 Yes Beaver Valley Unit V

12.5 2.64 1.260 25.7 15.7 10.0 Yes 2 Weld Metal (Heatd83642)

W 12.5 3.61 1.334 6.0 16.6 10.6 Yes (Heat 83642)

X 12.5 5.63 1.425 22.9 17.8 5.1 Yes The scatter of ARTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 17°F for base metal. Table A-2 indicates that only three of the eight surveillance data points fall within the +/- la of 17°F scatter band for surveillance base metals; therefore, the intermediate shell plate B9004-2 data is deemed "non-credible" per the third criterion.

Note that the original Capsule X report [Reference A-4] deemed this material "credible" when considering an increased allowable scatter band; however, the reduced margin term that would be used in subsequent reactor vessel integrity evaluations is not needed for this BVPS-2 material. Therefore, the "non-credible" conclusion in this updated evaluation is considered a conservative approach and will be utilized for the BVPS-2 intermediate shell plate B9004-2.

The scatter of ARTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A-2 indicates that all four surveillance data points fall within the +/- la of 28°F scatter band for surveillance weld materials; therefore, the weld material (heat 83642) is deemed "credible" per the third criterion.

Note that although the intermediate shell plate B9004-2 did not meet Criterion 3, it may still be used in determining the upper-shelf energy decrease in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-5 Criterion 4:

The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25'F.

The BVPS-2 capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad. The location of the specimens with respect to the reactor vessel beltline ensures that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25F.

Hence, Criterion 4 is met for the BVPS-2 surveillance program.

Criterion 5:

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The BVPS-2 surveillance program does not contain correlation monitor material; therefore, this criterion is not applicable to the BVPS-2 surveillance program.

CONCLUSION:

Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, the BVPS-2 surveillance plate data is deemed non-credible, whereas the surveillance weld data is deemed credible.

REFERENCES A-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

A-2 10 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No.

243, December 19, 1995.

A-3 WCAP-9615, Revision 1, "Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," P. A. Peter, June 1995.

A-4 WCAP-16527-NP, Revision 0, "Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," B. N. Burgos et al., March 2006.

A-5 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society of Testing and Materials, 1982.

A-6 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RP V Integrity Issues, February 12, 1998.

WCAP-16527-NP Supplement 1 September 2011 Revision 1

Retrieved from "https://kanterella.com/w/index.php?title=ML13151A060&oldid=5236127"