ML082740205

From kanterella
Jump to navigation Jump to search
WCAP-15571, Supplement 1, Revision 1, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program
ML082740205
Person / Time
Site: Beaver Valley
Issue date: 04/30/2008
From: Gift F
Westinghouse, Westinghouse
To:
Office of Nuclear Reactor Regulation
References
L-08-289 WCAP-15571, Suppl. 1, Rev. 1
Download: ML082740205 (30)
v • d • e


Text

ENCLOSURE A Beaver Valley Power Station (BVPS), Unit Nos. 1 and 2 Letter L-08-289 WCAP-15571 Supplement 1, "Analysis of Capsule Y from First Energy Company Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program,"

Revision 1, April 2008

Westinghouse Non-Proprietary Class 3 WCAP-15571 Supplement 1 Revision 1 April 2008 Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program MeMonghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15571, Supplement I Revision 1 Analysis of Capsule Y from Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program Frank C. Gift Jr.

April 2008 Approved:

Electronically Approved*

Patricia C. Paesano, Manager Primary Component and Asset Management

  • Electronically approved records are authenticated in the Electronic Document Management System Westinghouse Electric Company LLC P. 0. Box 355 Pittsburgh, PA 15230-0355

@2008 Westinghouse Electric Company LLC All Rights Reserved

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PREFACE Revision I to this report has been technically reviewed and verified by:

Natalie R. Jurcevich:

Electronically Approved*

  • Electronically approved records are authenticated in the Electronic Document Management System RECORD OF REVISION Revision 0:

Original Issue Revision 1:

The purpose of this revision is to address CAPS Issue 08-059-M009, which resulted in corrections made to Pages 6-1 and 6-5. In addition, editorial changes were made, including an update to Reference 4 to refer to Revision I ofWCAP-15571.

WCAP-1 5571, Supplement I April 2008 WCAP-1 5571, Supplement 1 Apris 2008 Revision 1

TABLE OF CONTENTS LIS T O F TA B LE S..........................................................................................................................

iv LIS T O F F IG U R E S.......................................................................................................................

iv EX E C U T IV E S U M M A R Y..............................................................................................................

v

1.

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

1-1

2.

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

2-1

3.

METHODOLODY FOR CALCULATION OF RTpTs AND USE........................................

3-1

4.

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

4-1

5.

NEUTRON FLUENCE VALUES.............................................................. 5-1

6.

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

6-1

7.

C O N C LU S IO N...............................................................................................................

7-1

8.

R E FE R E N C E S..............................................................................................................

8-1 WCAP-15571, Supplement 1 April 2008 Revision I

iv LIST OF TABLES Table 1 BVPS-1 Reactor Vessel Beltline Material Properties.........................................

4-2 Table 2 BVPS-1 Reactor Vessel Extended Beltline Material Properties......................... 4-3 Table 3 Maximum Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base M etal Interface for BV PS-1................................................................................

5-2 Table 4 Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-1 for the Beltline and Extended B eltline R egions.................................................................................................

5-2 Table 5 Summary of the BVPS-1 Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1..................... 6-3 Table 6 BVPS-1 Extended Beltline Material Chemistry Factors.....................................

6-4 Table 7 RTPTS Values for BVPS-1 Beltline Region Materials at 54 EFPY....................... 6-5 Table 8 RTPTS Values for BVPS-1 Extended Beltline Region Materials at 54 EFPY....... 6-6 Table 9 BVPS-1 Beltline Materials Projected USE Values at 54 EFPY.......................... 6-7 Table 10 BVPS-1 Extended Beltline Materials Projected USE Values at 54 EFPY.......... 6-8 LIST OF FIGURES Figure 1 USE % Drop for BVPS-1 Beltline Materials for 54 EFPY...................................

6-9 Figure 2 USE % Drop for BVPS-1 Extended Beltline Materials for 54 EFPY................. 6-10 WCAP-1 5571, Supplement 1 April 2008 Revision 1

v 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 Power Station Unit 1 (BVPS-1) reactor vessel beltline and extended beltline materials. This analysis will be based upon the results of the latest surveillance capsule Y evaluation, sister plant surveillance data, and the implementation of the Extended Power Uprate (EPU) program.

These calculations are performed for End-Of-License-Extended (EOLE) at 54 Effective Full Power Years (EFPY).

The limiting plate material in the BVPS-1 beltline is the lower shell plate B6903-1 with a projected EOLE RTpTs value of 275.70F using the BVPS-1 surveillance capsule data for 54 EFPY (equivalent to a fluence of 6.09x1019 n/cm2 (E > 1.0 MeV)). This value is slightly above the screening criteria of 270°F for forgings/plates in 10 CFR 50 Part 61. The screening limit of 2701F for lower shell plate B6903-1 will be reached at a fluence level of 4.961x10 19 n/cm2 (E >

1.0 MeV), which is equivalent to 43.87 EFPY. The limiting weld material in the BVPS-1 reactor vessel beltline is the lower shell longitudinal weld (heat number 305414) with an EOLE RTpTs value of 243.20F using the Fort Calhoun surveillance capsule sister plant data. 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-1 5571, Supplement 1 Apris 2008 Revision 1

1-1 1

INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing 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 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 vessel fluence.

The purpose of this report is to determine the RTpTs and USE values for the BVPS-1 reactor vessel using the results of the surveillance Capsule Y evaluation, sister plant data, and the implementation of the EPU Program. The results presented in this report are for EOLE at 54 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-1 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 conclusion and references for the PTS and USE evaluations follow in Sections 7.0 and 8.0, respectively.

WCAP-15571, Supplement 1 April 2008 Revision 1

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 in total, and therefore makes binding by rule, 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 RTPTS 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.

WCAP-15571, Supplement 1 April 2008 Revision 1

2-2 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 RTpTs screening criterion values for the beltline region are:

270°F for plates, forgings and axial weld materials 300OF 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-15571, Supplement 1 April 2008 WCAP-15571, Supplement 1 April 2008 Revision 1

3-1 3

METHODOLOGY FOR CALCULATION OF RTPTs AND USE 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 RTNDTfor each weld and plate or forging in the reactor vessel beltline.

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

Where, RTNDT(U)

=

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(U),

copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2 M = 2

  • U+o2 (2) cyu is the standard deviation for RTNDT(U)
  • YU 0°F when RTNDT(U) is a measured value GU

=

171F when RTNDT(U) is a generic value cA is the standard deviation for RTNDT For plates and forgings:

G

=

170F when surveillance capsule data is not used GA

=

8.50F when surveillance capsule data is used For welds:

G

=

280F when surveillance capsule data is not used TA

=

14 0F when surveillance capsule data is used YA 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.

AR TNDT = (CF) *f(0. 28-0.10logf)

(3)

WCAP-15571, Supplement 1 April 2008 Revision 1

3-2 "CF" (OF) 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 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 RTPTS.

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

RTPTs RTNDT(U) + M + ARdTPTs (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

-Z[Ai

  • f(O.2 8 -O.O2ogfi) ]

CF Z,. [I f(.56-0.20,ogfi) I 5

In Equation 5, "A1" is the measured value of ARTNDT and "fr" 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.

WCAP-1 5571, Supplement 1 April 2008 Revision 1

3-3 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 as indicated in Figure 2 of the guide when surveillance data is not used. 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-15571, Supplement 1 April 2008 WCAP-15571, Supplement 1 Apri 12008 Revision 1

4-1 4

VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the PTS evaluation, a review of the latest plant-specific material properties for the BVPS-1 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 1 x 11017 n/cm2 (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 1x10 17 n/cm 2 (E>1.0 MeV) must be monitored to evaluate the changes in fracture toughness. Reactor vessel materials not traditionally thought of as being 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-1 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-1 reactor vessel is given in Tables 1 and 2.

WCAP-15571, Supplement 1 April 2008 Revision 1

4-2 Table I BVPS-1 Reactor Vessel Beltline Material Properties(a)

Intermediate Shell Plate B6607-1 0.14 0.62 43 94 Intermediate Shell Plate B6607-2 0.14 0.62 73 83 Lower Shell Plate B6903-1 0.21 0.54 27 83 Lower Shell Plate B7203-2 0.14 0.57 20 85 Intermediate to Lower Shell Girth Weld 11-714 (Heat 90136)

Intermediate Shell Longitudinal 0.28 0.63

-56 112 Weld 19-714 A&B (Heat 305424)

Lower Shell Longitudinal Weld 20-0.34 0.61 714 A&B (Heat 305414)

Surveillance Weld (Heat 305424) 0.26 0.61 Notes:

(a) Materials information taken from WCAP-16799-NP [Reference 5] and WCAP-15571 [Reference 4].

(b) The Initial RTNDT values are measured values for the plates while the weld values are generic.

WCAP-15571, Supplement 1 April 2008 Revision 1

4-3 Table 2 BVPS-1 Reactor Vessel Extended Beltline Material Properties(a)

W t W ti %',~i

  • v I

ni tin tal RTNoT, Ini......l Material Description ID Heat Number Cu Ni (F

S ID

~~

1/2 O~)~~

~155 Upper Shell Forging B6604 123V339VA1 0.1 2 (b) 0.68 40 (101)(c) 305414 (3951) 0.337) 0.609()

-56 (Gen)(e) 97 Upper Shell to 305414 (3958) 0.3 3 0.609(d)

-56 (Gen)(e) 97(t)

Intermediate Shell Girth 10-714 AOFJ 0.03 0.93 10 (Gen) ill Weld FOIJ 0.03 0.94 10 (Gen) 104 EODJ 0.02 1.04 10 (Gen) 156 HOCJ 0.02 0.93 10 (Gen) 160 B6608-3 95712-1 0.08 0.79 60 (Gen) 97 Inlet Nozzles B6608-1 95443-1 0.10 0.82 60 (Gen) 82.5 B6608-2 95460-1 0.10 0.82 60 (Gen) 94 EODJ 0.02 1.04 10(Gen) 156 FOIJ 0.03 0.94 10 (Gen) 104 1-717B HOCJ 0.02 0.93 10 (Gen) 160 Inlet Nozzle Welds 1-717D DBIJ 0.02 0.97 10 (Gen) 123 1-717F EOEJ 0.01 1.03 10 (Gen) 152 ICJJ 0.03 0.99 10 (Gen) 123 JACJ 0.04 0.97 10 (Gen) 116 B6605-1 95415-1 0.137g7 0.77 60 (Gen) 93 Outlet Nozzles B6605-2 95415-2 0.1 3(g) 0.77 60 (Gen) 112.5 B6605-3 95444-1 0.09 0.79 60 (Gen) 103 ICJJ 0.03 0.99 10 (Gen) 123 IOBJ 0.02 0.97 10 (Gen) 122 1 -71 7A Outlet Nozzle Welds 1-717C JACJ 0.04 0.97 10 (Gen) 116 1-717E HOCJ 0.02 0.93 10 (Gen) 160 EODJ 0.02 1.04 10 (Gen) 156 FOIJ 0.03 0.94 10 (Gen) 104 Notes:

(a) All of the materials data is obtained from Combustion Engineering report MISC-PENG-ER-022

[Reference 6] except as noted.

(b) The Cu wt% was not available from the CMTR so in accordance with Regulatory Guide 1.99, Rev. 2, a standard deviation analysis (average + standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Shell Forgings (55 data points).

(c) Value in parenthesis is the 65% value per Regulatory Guide 1.99, Revision 2.

(d) Chemistry obtained from CE Report NPSD-1 039, Revision 2 [Reference 7].

(e) The generic Initial RTNDT values were determined in accordance with NUREG-0800 [Reference 81 and the 10CFR50.61 [Reference 2].

(f) The USE for 1092 welds documented in CEN-622-A. [Reference 9].

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

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

WCAP-15571, Supplement 1 Apris 2008 Revision 1

5-1 5

NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the BVPS-1 reactor vessel are shown in Tables 3 and 4 for the beltline and extended beltline materials, respectively. These values were projected using ENDF/B-VI cross sections and are based on the results of the Capsule Y radiation analysis and comply with Reg. Guide 1.190 [Reference 10].

These fluence data tabulations include fuel cycle specific calculated neutron exposures at the end of the seventeenth fuel cycle (the last completed at BVPS-1) as well as future projections to the end of Cycle 18 (the current operating cycle) and for several intervals extending to 54 EFPY.

The calculations account for a core power uprate from 2689 MWt to 2900 MWt at the onset of Cycle 18.

Neutron exposure projections beyond the end of Cycle 17 were based on the spatial power distributions and associated plant characteristics of Cycle 18 in conjunction with an uprated core power level of 2900 MWt.

WCAP-15571, Supplement 1 Apri 12008 Revision 1

5-2 TABLE 3 Maximum Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-1 1 1 18 Future Future Future Future Future Future IZ1.!0 21.0 25.0 32.0 36.0 40.0 48.0 54.0 Z.LZ)_1-I W 2.40E+ 19 2.85E+19 3.63E+19 4.08E+19 4.53E+19 5.42E+1 9 6.09E+19

1. -1 (I-

-I ji 1.25E+19 1.47E+19 1.87E+19 2.09E+19 2.32E+19 2.77E+19 3.11E+19 7.29E+18 8.51 E+1 8 1.06E+19 1.19E+19 1.31E+19 1.55E+19 1.73E+19 4.. Ur- ÷1 4.98E+18 5.80E+18 7.24E+18 8.06E+18 8.87E+18 1.05E+19 1.17E+19 TABLE 4 Calculated Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for BVPS-1 for the Beltline and Extended Beltline Regions Lower Shell Lower Shell Longitudinal Welds Lower Shell to Intermediate Shell Weld Intermediate Shell Intermediate Shell Longitudinal Welds Intermediate Shell to Upper Shell Weld Upper Shell RCS Inlet Nozzle to Upper Shell Weld RCS Outlet Nozzle to Upper Shell Weld 2.25E+19 4.70E+18 2.24E+19 2.24E+19 4.67E+18 2.36E+18 2.36E+18

< 1.OOE+17

< 1.OOE+17 5.42E+ 19 1.05E+19 5.40E+19 5.39E+19 1.05E+19 6.56E+18 6.56E+18 2.85E+17 2.11E+17 6.09E+19 1.17E+19 6.07E+19 6.06E+ 19 1.17E+19 7.45E+18 7.45E+18 3.26E+17 2.41E+17 WCAP-1 5571, Supplement 1 April 2008 Revision 1

k6-1 6

DETERMINATION OF RTPTs and USE VALUES FOR ALL BELTLINE and EXTENDED BELTLINE REGION MATERIALS 6.1 BVPS-1 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-1 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-1, the related surveillance program results have been included in this PTS evaluation.

Specifically, the BVPS-1 plant-specific surveillance capsule data for the lower shell plate B6903-1 and weld metal (heat 305424) is provided and applied as follows:

1) There have been four capsules removed from the BVPS-1 reactor vessel.
2) The data for the BVPS-1 surveillance program plate material is deemed not credible. The data was used with a a margin of 171F.
3) The data for the BVPS-1 surveillance program weld material is deemed not credible. The data was used with a YA margin of 280F.
4) The surveillance capsule materials are representative of the actual vessel plate (B6903-1) and intermediate shell longitudinal weld metal (weld heat 305424).
5) The resulting RTPTS values for lower shell plate B6903-1 exceed 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 remain below the PTS Rule screening criteria at 54 EFPY.

Chemistry factor values for the BVPS-1 beltline region materials based on Position 1.1 and 2.1 from Regulatory Guide 1.99, Revision 2 are presented in Table 5. Additionally, chemistry factor values for the BVPS-1 extended beltline materials based on Position 1.1 of Regulatory Guide 1.99, Revision 2 WCAP-15571, Supplement 1 April 2008 Revision 1

6-2 are presented in Table 6. Tables 7 and 8 contain the RTpTs calculations for all beltline and extended beltline region materials at 54 EFPY, respectively.

6.2 BVPS-1 Upper Shelf Energy Calculations for 54 EFPY For BVPS-1, there exists surveillance data for plate B6903-1 and weld heat 305424. 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. Figures 1 and 2 were used in the determination of the % decrease in USE for the beltline and extended beltline materials. Tables 9 and 10 document the USE values for all of the materials at 54 EFPY. All of the beltline and extended beltline material USE values maintain 50 ft-lbs or greater at 54 EFPY.

WCAP-15571, Supplement 1 April 2008 Revision 1

6-3 Table 5 Summary of the BVPS-1 Beltline Material Chemistry Factor Values Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 i**

7*!* ~~~~~Ce mi~str *iF a cto r:;'ii *7k:

Material Description Position 1.1 Poiion 2.1 Intermediate Shell Plate B6607-1 100.5 0F Intermediate Shell Plate B6607-2 100.5 0F Lower Shell Plate B6903-1 147.2 0F 149.20F Lower Shell Plate B7203-2 98.7 0F Intermediate to Lower Shell Girth Weld 11-714 (Heat 90136)

Intermediate Shell Longitudinal Weld 19-714 A&B (Heat 305424)

Lower Shell Longitudinal Weld 20-714 A&B (Heat 305414)

WCAP-15571, Supplement 1 April 2008 Revision 1

6-4 Table 6 BVPS-1 Extended Beltline Material Chemistry Factors uppul,

r ulyllylI I-OVOOZV/ MI U.IL U. E 305414 (3951) 0.337 0.609 305414 (3958) 0.337 0

AOFJ 0.03

).609 0.93 209.11 209.11 41.0 41.0 Upper Shell to Intermediate Shell Girth Weld 10-714 FOIJ 0.03 U.U-+l EODJ 0.02 1 1.04 27.0 HOCJ 0.02 0.93 27.0 B6608-1 95443-1 0.10 0.82 67.0 B6608-2 95460-1 0.10 0.82 67.0 B6608-3 95712-1 0.08 0.79 51.0 EODJ 0.02 1.04 27.0 FOIJ 0.03 0.94 41.0 1-717B HOCJ 0.02 0.93 27.0 1-717D DBIJ 0.02 0.97 27.0 1-717F EOEJ 0.01 1.03 20.0 ICJJ 0.03 0.99 41.0 JACJ 0.04 0.97 54.0 B6605-1 95415-1 0.13 0.77 95.25 B6605-2 95415-2 0.13 0.77 95.25 B6605-3 95444-1 0.09 0.79 58.0 ICJJ 0.03 0.99 41.0 IOBJ 0.02 0.97 27.0 Outlet Nozzle Welds 1717C JACJ 0.04 0.97 54.0 1-717E HOCJ 0.02 0.93 27.0 EODJ 0.02 1.04 27.0 FOIJ 0.03 0.94 41.0 WCAP-15571, Supplement 1 April 2008 Revision 1

6-5 Table 7 RTPTS Values for BVPS-1 Beltline Region Materials at 54 EFPY 1/2 FluenceI Su*f Neutron*

Chemistry Initial '

(C)

(

S(e)"

Material RG

~Fluence

~

Fact(r, Factor RTDT~b agn i 1 Pos...

(1/2 xl 101 nlcm')(F

('F)~

F 0 )

()~

F Intermediate Shell Plate B6607-1 1.1 6.0600 1.4384 100.50 43.0 144.6 0.0 17.0 34.0 221.6 Intermediate Shell Plate R6607-2 1.1 6.0600 1.4384 100.50 73.0 144.6 0.0 17.0 34.0 251.6 1.1 6.0900 1.4392 147.20 27.0 211.9 0.0 17.0 34.0 272.9 Lower Shell Plate B6903-1 2.1 6.0900 1.4392 149.20 27.0 214.7 0.0 17.0 34.0 275.7 Lower Shell Plate B7203-2 1.1 6.0900 1.4392 98.70 20.0 142.1 0.0 17.0 34.0 196.1 Intermediate to Lower Shell Girth 1.1 6.0700 1.4386 124.30

-56.0 178.8 17.0 128.0 65.5 188.3 Weld 11 -714 (Heat 90136) 2.1 6.0700 1.4386 84.80

-56.0 122.0 17.0 14.0 (')

44.0 110.0 Intermediate Shell Longitudinal 1.1 1.1700 1.0438 191.70

-56.0 200.1 17.0 28.0 65.5 209.6 Weld 19-714 A&B (Heat 305424) 2.1 1.1700 1.0438 188.80

-56.0 197.1 17.0 28.0 65.5 206.6 Lower Shell Longitudinal Weld 1.1 1.1700 1.0438 210.50

-56.0 219.7 17.0 28.0

65.

229.2 20-714 A&B (Heat 305414) 2.1 1.1700 1.0438 223.9

-56.0 233.7 17.0 2801 65.5 243.2 NOTES:

(a) FF = fluence factor =

(0.28 -0.1 log (f))

(b) Initial RTNDT values are measured values with the exception of the vessel welds.

(c) ARTPTS = CF* FF.

(d)M = 2 *(ai2 +

A2) 1 2 (e) RTPTS = Initial RTNDT + ARTPTS + Margin.

(f) The St. Lucie Unit 1 surveillance weld metal is the same weld heat as the BVPS-1 intermediate to lower shell girth weld (heat # 90136). The St. Lucie Unit 1 surveillance weld data is credible (see Appendix D of Reference 4); therefore, the reduced oY term of 14°F was utilized for BVPS-1 weld heat #90136.

(g)The Ft. Calhoun surveillance weld metal is the same weld heat as the BVPS-1 lower shell longitudinal weld (heat # 305414). The Ft.

Calhoun surveillance weld data is not credible (see Appendix D of Reference 4); therefore, the higher a, term of 28°F was utilized for BVPS-1 weld heat #305414.

WCAP-15571, Supplement 1 April 2008 Revision 1

6-6 Table 8 RTPTS Values for BVPS-1 Extended Beltline Region Materials at 54 EFPY Upper to Inter Girth Weld FOIJ 1.1 0.7450 0.9174 41.00 10.0 37.6 17.0 18.8 50.7 98.3 Upper to Inter Girth Weld EODJ 1.1 0.7450 0.9174 27.00 10.0 24.8 17.0 12.4 42.1 76.8 Upperto Inter Girth Weld HOCJ 1.1 0.7450 0.9174 27.00 10.0 24.8 17.0 12.4 42.1 76.8 Inlet Nozzle 95443-1 1.1 0.0326 0.2305 67.00 60.0 15.4 17.0 7.7 37.3 112.8 Inlet Nozzle 95460-1 1.1 0.0326 0.2305 67.00 60.0 15.4 17.0 7.7 37.3 112.8 Inlet Nozzle 95712-1 1.1 0.0326 0.2305 51.00 60.0 11.8 17.0 5.9 36.0 107.7 Inlet Nozzle Weld (EODJ) 1.1 0.0326 0.2305 27.00 10.0 6.2 17.0 3.1 34.6 50.8 Inlet Nozzle Weld (FOIJ) 1.1 0.0326 0.2305 41.00 10.0 9.5 17.0 4.7 35.3 54.7 Inlet Nozzle Weld (HOCJ) 1.1 0.0326 0.2305 27.00 10.0 6.2 17.0 3.1 34.6 50.8 Inlet Nozzle Weld (DBIJ) 1.1 0.0326 0.2305 27.00 10.0 6.2 17.0 3.1 34.6 50.8 Inlet Nozzle Weld (EOEJ) 1.1 0.0326 0.2305 20.00 10.0 4.6 17.0 2.3 34.3 48.9 Inlet Nozzle Weld (ICJJ) 1.1 0.0326 0.2305 41.00 10.0 9.5 17.0 4.7 35.3 54.7 Inlet Nozzle Weld (JACJ) 1.1 0.0326 0.2305 54.00 10.0 12.4 17.0 6.2 36.2 58.7 Outlet Nozzle 95415-1 1.1 0.0241 0.1928 95.25 60.0 18.4 17.0 9.2 38.6 117.0 Outlet Nozzle 95415-2 1.1 0.0241 0.1928 95.25 60.0 18.4 17.0 9.2 38.6 117.0 Outlet Nozzle 95444-1 1.1 0.0241 0.1928 58.00 60.0 11.2 17.0 5.6 35.8 107.0 Outlet Nozzle Weld (ICJJ) 1.1 0.0241 0.1928 41.00 10.0 7.9 17.0 4.0 34.9 52.8 Outlet Nozzle Weld (IOBJ) 1.1 0.0241 0.1928 27.00 10.0 5.2 17.0 2.6 34.4 49.6 Outlet Nozzle Weld (JACJ) 1.1 0.0241 0.1928 54.00 10.0 10.4 17.0 5.2 35.6 56.0 Outlet Nozzle Weld (HOCJ) 1.1 0.0241 0.1928 27.00 10.0 5.2 17.0 2.6 34.4 49.6 Outlet Nozzle Weld (EODJ) 1.1 0.0241 0.1928 27.00 10.0 5.2 17.0 2.6 34.4 49.6 Outlet Nozzle Weld (FOIJ) 1.1 0.0241 0.1928 41.0 10.0 7.9 17.0 4.0 34.9 52.8 NOTES:

(a)

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

(b)

(c)

(d)

(e)

Initial RTNDT value for the upper shell forging is a measured value. All other values are generic.

ARTPTS = CF* FF.

M = 2 *((312

+ (*_2)1/2 RTPTS = Initial RTNDT + ARTPTS + Margin.

WCAP-1 5571, Supplement 1 April 2008 WCAP-15571, Supplement 1 April 2008 Revision 1

6-7 Table 9 BVPS-1 Beltline Materials Projected USE Values at 54 EFPY Intermediate Shell Plate B6607-1 0.14 3.7781 94 32 63.9 Intermediate Shell Plate B6607-2 0.14 3.7781 83 32 56.4 Lower Shell Plate B6903-1 0.21 3.7968 83 38(a) 51.5 Lower Shell Plate B7203-2 0.14 3.7968 85 32 57.8 Intermediate to Lower Shell Girth 0.27 3.7843 144 52 69.1 Weld 11-714 (Heat 90136)

Intermediate Shell Longitudinal 0.28 0.7294 112 2 8 (b) 80.6 Weld 19-714 A&B (Heat 305424)

Lower Shell Longitudinal Weld 0.34 0.7294

>100 41 59.0 20-714 A&B (Heat 305414)

NOTES:

(a)

Based on results from BVPS-1 Surveillance Plate B6903-1 [Reference 4].

(b)

Based on results from BVPS-1 Surveillance Weld Heat 305424 [Reference 41.

WCAP-15571, Supplement 1 April 2008 Revision 1

6-8 Table 10 BVPS-1 Extended Beltline Materials Projected USE Values at 54 EFPY 1/4TEE Initial USE Ue Projected Component Heat ~

Wt % u~

(a)

Fltuence Iiia S

EOLE USEA (10"n/c 2) ft-bs) ecrase (ft-lbs)

Upper Shell 127.9 Uprghel 123V339VA1 0.12 0.4645 155 (10 1)(b) 17.5 183.39(b)

Forging Shell 30541(3951 0.337 0.4645 97 38 360.1 305414 (3951) 0.337 0.4645 97 38 60.1 UprSel 305414 (3958) 0.337 0.4645 97 38 60.1 to AOFJ 0.03 0.4645 111 16 93.2 Inermediate FOIJ 0.03 0.4645 104 16 87.4 Weld EODJ 0.02 0.4645 156 16 131.0 HOCJ 0.02 0.4645 160 16 134.4 95443-1 0.10 0.0203 82.5 7.5 76.3 95460-1 0.10 0.0203 94 7.5 87.0 95712-1 0.08 0.0203 97 7.5 89.7 EODJ 0.02 0.0203 156 7.5 144.3 FOIJ 0.03 0.0203 104 7.5 96.2 Inlet Nozzle HOCJ 0.02 0.0203 160 7.5 148.0 Welds DBIJ 0.02 0.0203 123 7.5 113.8 EOEJ 0.01 0.0203 152 7.5 140.6 ICJJ 0.03 0.0203 123 7.5 113.8 JACJ 0.04 0.0203 116 7.5 107.3 Outlet 95415-1 0.13 0.0150 93 9

84.6 Nozzles 95415-2 0.13 0.0150 112.5 9

102.4 95444-1 0.09 0.0150 103 7.5 95.3 ICJJ 0.03 0.0150 123 7.5 113.8 IOBJ 0.02 0.0150 122 7.5 112.9 O ele JACJ 0.04 0.0150 116 7.5 107.3 HOCJ 0.02 0.0150 160 7.5 148.0 EODJ 0.02 0.0150 156 7.5 144.3 FOIJ 0.03 0.0150 104 7.5 96.2 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) Value in parenthesis is the 65% value.

WCAP-15571, Supplement I April 2008 Revision 1

C/)

CD 3

0

"'1

-n CD m

0

-9 0

"0 V

"o

,D Cap Cap Cap V

.U W

Cap Y

I I

I I I Surv Weld 305424 Plate B6903-1 w

at C

6 a, IV 2

4 6

a, 1""01 2

FL..UENCE, n/cm 2(E 1Me" C

' l/4T EOLR Fluence 7.3E18

-IGMLJE 2

c ein Shdf Energy as a* wi f C

Coe

,,a F Weld Surveillance Data 1/4T EOLR Fluence 3.61E19 Plata Surveillance Data

"o l "3

CI) 040 013 METAL WELDS "V

0.35 0.30-0.30-0.2 -

O@

0.20

  • 0.15-CO 00 2x CCD C

m

(

3 MT 1

ELOS N "2

,1 _1 0 1 7 4..

.6 a,: 1 0 18.

.2

. 4.

6 10 1 9

.i 02 0

1.50E17 and 2,03E17

.1*

114T EOLR Fluence rr 4 65E_18 to 00 000

7-1 7

CONCLUSION All of the beltline and extended beltline region materials in the BVPS-1 reactor vessel have EOLE RTPTS values below the screening criteria values of 270°F for forgings/plates and 300°F for circumferential welds at EOLE (54 EFPY) with the exception of lower shell plate B6903-1.

This plate has a 54 EFPY RTPTS value of 275.70°.

Based on the fluence information provided in Section 5, the PTS screening criteria of 270°F is reached at a fluence value of 4.961x1019 n/cm 2 (E>1.0 MeV). This fluence value of 4.961x10 19 n/cm2 (E>1.0 MeV) equates to 43.87 EFPY for BVPS-1.

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

WCAP-1 5571, Supplement I April 2008 WCAP-15571, Supplement 1 Aprii 2008 Revision 1

8-1 8

REFERENCES

1.

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

U.S. Nuclear Regulatory Commission, May 1988.

2.

10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", Federal Register, Volume 60, No. 143, 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-1 5571, Revision I "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," N.R. Jurcevich, April 2008.

5.

WCAP-1 6799-NP, Revision 1, "Beaver Valley Power Station Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," B.N. Burgos, June 2007.

6.

Combustion Engineering Report MISC-PENG-ER-022, Revision 00, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the Beaver Valley Unit 1 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," S.M. Schloss, 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.

"Generic Upper Shelf Values for Linde 1092, 124 and 0091 Reactor Vessel Welds," CEOG Report CEN-622-A, ABB Combustion Engineering, December 1996.

10.

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.

WCAP-15571, Supplement 1 April 2008 Revision 1

Retrieved from "https://kanterella.com/w/index.php?title=ML082740205&oldid=5325268"