ML20101R393

Responds to GL 92-01 Re NRC Verification of Compliance W/Licensing Basis Concerning Reactor Vessel Fracture Toughness & Matl Surveillance for RCPB

ML20101R393
Person / Time
Site:	Beaver Valley
Issue date:	07/08/1992
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NUDOCS 9207150345
Download: ML20101R393 (42)

Text

a Dugmsne Lkfit Company m;p-$1-f>hermpoort, PA 150 '7KO4 (41h 3PM25$

e JOHN C SIEMR vge Prestoent. Nuclear Group g,g y99y U.

S.

Nuclear Regulatory Commission Attn Document Control Desk Washington, DC 20555

Subject:

Deaver Valley Power Station, Unit No. 1 and No. 2 DV-1 Docket No. 50-334, Licence No. DPR-66 DV-2 Docket No. 50-412, Licenso No. HPF-73 Response To Generic Lotter 92-01 Enclosed t 9 the fcllowing two reports which provide the

1. formation requ sted by Generic Letter 92-01:

1.

Beaver Valley Unit 1,

Response

To Generic Letter 92-01, Reactor Vessel Structural Integrity, 10 CPR 50.54 (f) 2.

Deaver Valley Unit 2,

Response

To Generic Letter 92-01, Reactor Vessel Structural Ir.tegrity, 10 CFR 50.54 (f)

We understand this information is required for NRC verification of our compliance with our current licensing basis regarding reactor ves, 1 fracture toughness and materia 1 surveillance for the reactor coolant pressure boundary.

Cincerely, 1

/ dbr^

D.

Slober cc:

Mr.

L.

U.

Rossbach, Sr. Resident Ins ector Mr.

T.

T. Martin, NRC Region I Administrator Mr. A.

W.

DeAgazio, Project Manager Mr. M.

L.

Bowling (VEPCO) h e *

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92071503o5 920708

[fff PDR ADOCK 03000334 P

PDR i

, i' COMMONWEALTH OF PENNSYLVANIA)

)

SS COUNTY OF BEAVER

)

llkh On this

/

day of

, 1992,

/

[//d7, M'

/fdI/M a N tary P blic in and for said before me,

- Commonwealth and County, personally appeared J. D. Sieber, who being

- duly sworn, deposed, and said that -(1) he is Vice President - Nuclear

' of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of'said company, and (3) the statements set.'forth-in the Submittal are true and correct to the best of his knowledge, information and belief.

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s BEAVER VALLEY UNIT 1 RESPONSE TO GENERIO LETTER 92-01 REACTOR VESSEL STRUCTURAL INTEGRITY, 10 CFR 50.54(f)

WP1331:lD/062392

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TABLE OF CONTENTS Section Titte Page TABLE OF CONTENTS i

LIST OF TABLES ii LIST OF FIGURES ii 1.0 Introduction 1

2.0 Reactor Vessel Structural Integrity Required Information 2

p 3.0 Conclusions 21 4.0 References 24

(

WP1331:10/ 062392 i

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4 LIST OF TABLE 5 (ABLE TITLE PAG {

l Beaver-Valley Unit 1 End-of-License (32 EFPY) Upper Shelf 6

Energy Values 2

Beaver Valley Unit 1 Materials Certification Inform 2 tion 8

3-Beaver Valley Unit 1 Materials certification Information 9

4 Beaver Valley Unit 1 Materials Cartification Information 10 5

Beaver Valley Unit 1 Materials Certification Information 11 6

Beaver Valley Unit 1 Materials Certification Informatiot.

12 7

Beaver Valley Unit 1 Materials Certification Information 13 8-Beaver Valley. Unit 1 Materials Certification Information 14' 9

Beaver Valley Unit 1 Materials Certification Information 15 10 Beaver Valley Unit 1 Materials Certification Information 16 11 Beaver: Valley Unit 1 Materials Certification Information 17 12 Be..

Valley Unit 1 Materials Certification Information 18 13 Comparisen Beaver Valley Unit 1 Reactor Versel Surveillance 22 Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Rev. 2 Predictions

)

14 Beaver Valley Unit 1 Adjusted Reference Temperature Values 23

~

for12/16/91 and End-of-License (32 EFPY) 1 1

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- 1. 0 INTRCLUCTION On 6 March 1992 the U.S. Nuclear Regulatory Commission (NRC) issued Revision 1 of Generic letter (GL) 92-01 to obtain information needed to assess compliance with the requirements and commitments regarding reactor vessel structural integrity._ Section 50.60(a) of Title 10 of the Code of Federal Regulations (10 CFR 50.60(a)) requires that all light water nuclear power reactors, licensed by the NRC, meet fracture toughness requirements and have a material surveillance program for the-reactor pressure boundary.- These requirement 3 are set forth in Appendices G and H to 10 CFR Part 50.

If the requirements of

-Appendices G and H cannot be met, an exemption pursuant to 10 CFR 50.12 is required.

10 CFR 50.61 provides fracture toughness requirements for protecting pressurized water reactors again<t pressurized thermal shock events.

On a related topic, the U.S. Nuclear Regulatory Commission on 12-July-88 issued Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials'and its Impact on Plant Operations".

The l

purpose of this generic letter was to call attention to Revision 2 of 1

Regulatory Guide _1.99, " Radiation Embrittlement of Reactor Vessel

' Materials".

Each licensee was required to submit the results of a technical analysis relative to the implementation of this regulation.

This submittal addressed the-following:

1)

Recalculation of adjusted reference temperature values at the _1/4 and 3/4 reactor vessel wall thickness locations for all potentially limiting materials using Raguhtory Guide 1-.99 Revision 2.

p 2)

Determine the date of applicability of the P/T limits using Regulatory Guide Revision 2.

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If the use of the Revision 2 methodology resulted in a modification p

of the P/T limits contained in the Technical Specification, a proposed schedule was required to be sub.nitted that defined whatever action needed to be taken, including those required to address an expected restriction in operating flexibility.

Per GL 92-01, "This generic letter is part of a program to evaluate reactor vessel integrity and take regulatory actions, if needed, to ensure thct licensees and permit holders are complying with the requirements of 10 CtR 50.60 and 10 CFR 50.61, and are fulfilling commitments made in response to GL 88-11."

This report describes the methods used and the results obtained in evaluating Beaver Valley Unit I relative to GL 92-01.

1 2.0 REACTOR VFSSEL STRUCTURAL INTEGRITY REQUIRED INFORMATION Ouestion 1:

Compliance with Appendix H to 20 CFR 50:

Determine if the Reactor Vessel Irradiation Surveillance Program is in

- compliance with Appendix H to 10 CFR 50 by. determining which version of ASTM E-185_was used to develop the Reactor Vessel Irradiation Surveillance Program.

t if the program does not meet ASTM E-185-73, -79, or -82 and is not part of an j

approved " integrated surveillance program", describe the actions that have or I

will be taken to ensure that the Reactor Vessel Irradiation Surveillance Program complies with Appendix H to 10 CFR 50.

L

Response

l The Beaver Valley Unit 1 vessel was designed to the Winter 1968 Addenda-to Section III of the-ASME Code.

ASTM E185 73 was the standard'in place at the time the surveillance program was de:igned. The Beaver Valley Unit I surveillance program complies-with ASTM E185-73.

Testing of surveillance captules after July 26, 1983 has been performed in accordance with ASTM

[

- WP1331:lD/062392 2

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Standard ver.< ion E185 82.

Furthermora, the surveillance program design was approved during the FSAR licensing process and the present capsule testing program has been approved as ptrt of the UFSAR.

Therefore, it ir. determined that the surveillance program for Beaver Valley Unit 1 meets the requirements of Appendix H to 10 CFR Part 50 and that an exemption request is not considered necessary.

Question 2:

Compliance with Appendix G to 10 CFR 50:

Question 2a: Calculate the Charpy Upper Shelf Energy (USE) for all of the beltline materials of the Beaver Valley Unit I reactor vessel using the methods of Regulatory Guide 1.99, Revision 2 for December 16, 1991 and for tne end of the current license.

If the calculated USC of the limiting beltline weld or plate of the Beaver Valley Unit I reactor vessel is below 50 ft-lbs, describe the actions that hase or will be taken pursuant to Paragraphs

~

IV.A.1 or V.C of Appendix G in 10 CFR Part 50 to address this issue.

Response

Table 1 contains the December 16, 1991 and E0L Charpy upper shelf energy for Beaver Valley Unit 1 beltline materials. The calculated E0L Charpy upper shelf energy for all the beltline materials are predicted to be above the 50 ft-lb criteria.

WP1331:lD/062392 3

5

'w Table 1 Beaver Valley Unit 1 Calculated Upper-Shelf Eneroy (USE) Values Initial It/16/91 1/4" EOL(c)

USE USE F' uence E0L USE Beltline Material (ft-lb)

(ft-lbs) (x10L9 n/cm2)

(ft.lbs)

Intermediate Shell Plate, 86607-1 90 71.1 2.16 59,9 Intermediate Shell Plate, 86607-2 82.6 65.2 2.16 65.3

' Lower Shell Plate, B7203-2 83.5 66 2.16 60.5 Lower Shell Plate, 86903-1 80.0 59.4 2.16 52 Intermediate-Shell Longitudinal Weld ll2(a) 80.6 0.49 69 Seam, Heat 305424 Lower Shell Longitudinal Weld Seam, (b)

Heat 305414 Intermediate to Lower Shell (b)

Circumferential Weld, Heat 90136 4

(a)- -Based on surveillance weld.

(b)

Initial Upper Shelf Energy-values are not available.

-(c).

Fluence projections based on recent change in core design incorporating L4P.

1 WP1331 10/062392' 4

1 1

i Question 2b: If the Beaver Valley Unit ! reactor vessel was constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition, provide the following information:

a) All Charpy and drop weight test results for all unirradiated beltline materials, the unirradiated reference temperatures for each beltline material, and a description of the methods used for calculating these values b) A description of the heat treatment performed on all beltline and surveillance material.

c) The heat numbers for each beltline plate or forging and the weld and flux lot number used to fabricate each beltline material and surveillance material and weld.

d) A description of the chemical composition of each beltline and surveillance material, e) The heat number of the wire used for determining the weld chemical composition.

Response

The Beaver Valley Unit I reactor vessel was constructed to Section 111 of the ASME Code, 1968 Edition with Addenda through Winter 1968.

Thus, the Beaver Valley Unit I reactor vessel was constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition.

Tables 2 through 16 document the unirradiated data (Charpy and drop weight test results, reference temperature.

- upper shelf energy, heat treatment, heat numbers, flux lot number and chemical composition) for all beltline region and surveillance materials.

These values were developed using the current reactor pressure vessel material test requirements and acceptance standards at the time of fabrication.

WP1331:lD/062392 5

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Table 2 Beaver Valley Unit 1 Materials Certification Information Component:

Intermediate Shell, B6607-1, Heat No.:

C4381-1 Material Specification A533 C1 1 MILL Chemical Analysis [Il C

Mn P

S Si Ni

_ Mo Cu Al Co 0.23 1.40 0.015 0.016 0.25 0.62 0.55 0.14 0.010 Charpy Impact and Fracture Tests - Transverse [2]

Temp *F Ft lbs

% Shear Lat. Exp)

(inches

-40 14.5 10 8

-40 16 10 9

-40 13 10 6

0 19 10 12 0

22 15 14 0

23.5 15 15 50 32.5 30 27 50 35 30 28 50 44 34 32 110 57 60 44 110 62.5 65 50 110

.66 65 52 RT 58 50 42 RT-44 40 36 RT 40 40 33 160 80 85 63 160 90 100 71 160

-95 100 76 210 91 100 63 210 87 100 71 210-95 100 75 Temp. 'F-Drop Weights NDT RTNDT USE

-0 NF 10'F-43*F 90 ft-lb 0

NF

-10 F

-r0 F

liegt Treatment [1]

1600-1650'F, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

Brine quenched.

1200-1225'F, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.. Brine quenched.

1100;1150'F, 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

Furnace cooled.

WP1331:lD/062392 6

.=

t Table 3 Beaver Valley Unit 1 Materials Certification information i

Component:

Intermediate Shell, B6607-2 Heat No.:

C4381-2 i

Material Specification A533 Cl 1 MILL Chemical Analysis [3]

C Mn P

S Si N1 Mo Cu Al Co 0.23 1.40 0.015 0.016 0.25 0.62-0.55 0.14 0.010 Charpy Impact and Fracture Tests - Transversel43 l

Temp 'F Ft-Lbs

% Shear Lat. Exp.

(mils)

-40 13.5 10 6

-40 15.5 10 8

40 18 10 10 0

30 20 20 0

19 15 11 0

17 15 12 1

50 26.5 30 18 50 27 30 22 50 29 30 22 110 49 50 40 t

110 44 50 38 110 52 50 42 160-71 85 57 160-60 85 52 160 76 90 63 210 81 100 65 210

. 86 100 70 210 81 100 65 Temp.F Drop Waights' NDT RTNDT USE 10 2,F 10'F 73'F 82.6 ft lb 20 F

0 NF-0 NF Heat Treatment [3]

1600-1650*F, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. -Brine-quench.

-1200 1225'F, 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.- Brine quench.

1100-1150*F, 60, hours.- Furnace cooled.

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Table 4 Beaver Valley Unit 1 Materials Certification Information Component:

Lower Shell, 07203-2 Heat No.:

C6293-2 Material Specification A533 C1 1 Mill Chemical Analysis [5]

C Mn P

S Si Ni Mo tu Al Co O.19_

l.30 0.015 0.015 0.18 0.57 0.59 0.14 0.026 0.021 Charpy Impact and Fracture Tests - Tiansverse[6.1 l

Temp 'F-Ft Lbs

% Shear Mils Lat. Exp.

-40 26 10 17 13 10 9

40 12 10 6

0 27 30 23 0

20 30 16 0

19 30 17 40 34 35 30 40 37 35 32 40 54 40 42 110 79 75 64 110 66.5-70 54 110 72.5 80 58 160 86.5 100 71 160 81'.5 100 67 160 86 100 67 210-91 100 72 210

-75 100 62 210 81 100 67

. Temp. 'F Drop Weights-NDT RTNDT USE J20 F-

-20*F

.20*F 83.5 ft-lb

-10 2 NF

-0 NF Heat Treatment (5) 1550-1650*F,-4 hours. -Water quenched, 1200*F - 1250*F,.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Air cooled.

ll25'F - Il75aF, 40. hours.

Furnace cooled.~

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Table 5 Beaver Valley Unit 1 Materials Certification Information 9

[qm2onent:

Lower Shell, B6903-1 Heat No.: C6317-1 Material Specification A533 Cl 1 ljill Chemical Analysis [7]

C Mn P

S Si Ni Mo Cu Al Co 0.20 1.31 0.010 0.015 0.18 0.54 0.55 0.20 0.028 0.014 Charpy impact and Fracture Tests - Transverse [8]

Temo 'F Ft-lbs

% Shear Mils Lat. Exo.

-:.00 4.0 3

4

-:.00 d.5 3

2

-:,00 4.5 3

0

-50

',1. 0 5

7 6.0 3

2

-50

-50

,1. 0 5

10-40.0 29 31

,0 20.0 21

? ~.

,0 28.5 18 19 40 33.0 28 30 40 46.5 33 37 40-34.0 28 29 1:.0 63.5 51 54 L:.0 64.0 53 oO

.10

-77.0 57 63

.160-76.0 100 60 L60 82.0 100 67 L60 82.5 100 69 2:.0

-83.0 100.

70 2: 0 82.5 100 66 2: 0 82.0-100 69 Temo. 4F Droo Weichts NDT RTNDT U$1

-60 F

-50'F 27'F 80 ft-lb

-50 F

h

-40<

-2 NF

=

-20

'4 F

-0

'4 F

-Heat Treatment [7]

.550aF-1250*F' 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quene ed.

3 200* : +:,250 F 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Air coole.

.'125*

I-Ll75'F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled.

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Table 6 Beaver Valley Unit 1

. Materials Certification Information Component:

Intermediate Shell Seams19-714 A&B Heat No.: 305424 luxLinde10'92, lux Lot 388 84 Mod.

MILL Chemical Anal _ysisl93 C

Mn P

S Si Ni Mo Cu Al Co 0.13 1.46-0.013 0.010 0.18 0.64 0.52 0.30 Charpy impact and Fracture Tests (9)

Temo 'F Ft-Lbs

% Shear Mils Lat. Exo.

.0 82..

LO 87

.0

.92 Temp. 'F Drop' Weights NOT RTNDT USE

• -56*F Post Weld Heat Treatmentl9) ll25-ll75'i, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, Furnace cooled.

I i

• Estimated per NRC Regulatory Review Plan MTEB 5-2 L

WP1331:lD/0.62392_

10

s Table 7 Beaver Valley Unit 1 Materials Certification Information Component:

Intermediate Shell to Lower Shell Heat No.: 90136, Seam 11-714 Flux Lir.de 0091, Flux Lot 3977 B4 MILL Chemical Analysis [9}

C.

Mn P

S Si Ni Mo Cu Al Co l

[

0.11-1.17 0.013 0.010 0.17 0.49 0.30 Charpy Impact and Fracture Tests [9}

f Temp 'F Ft-Lbs

% Shear Mils Lat. Exp.

10 100 10-112 10 108-Temp. 'F-

~ Drop Weights NDT RTNDT USE t

• -56*F

-Eo,_s_t Weld Heat Treatment [9}

ll25-1175'F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, Furnace. cooled.

• Estimated; per NRC Regulatory Review Plan MTEB 5-2

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Table 8 Beaver Valley Unit 1 Materials Certification Information The following information was taken from the Weld Inspection Report prepared by Combustion Engineering, Inc. in March 1970.

j Component:

Lower Shell Seams20-714 A&B Heat No.: 305414, Flux Linde 1092, Flux Lot 394.7 84 Mod.

MILL Chemical Analysis (10]

C Mn P

S Si Ni Mo Cu Al Co 0.14 1.45 0.012 0.010 0.18 0.59 0.51 0.33 Charpy impact and Fracture Tests (10)

Temp *F Ft-Lbs

% Shear Mils Lat. Exp.

10 82 10 66

-10 80

-Temp. 'F Drop Weights NDT RTNDT USE

• -56'F Post Weld Heat' Treatment (10]

1125-1175'F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled.

• ! Estimated per NRC. Regulatory Review Plan MTEB 5-2

. WP1331:10/0623924 12.

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Table 9

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Beaver Valley Unit 1 Materials Certification Information Component:

Intermediate Shell to Lower Shell Heat No.: 90136, Seam 11-714 Flux Linde 0091, Flux Lot 3998 B=

MILL Chemical Analysis [Ill C

Mn P

S Si Ni Mo Cu Al to 0.11 1.16 0.013 0.010 0.16 0.50 0.37 Charpy impact and Fracture Tests [Ill Temp 'F Ft-Lbs

% Shear Hils Lat. Exp.

.10 110 10 116 10 107 l

Temp. *F-Orop Weights NDT RTNOT USE

• -56*F Post Weld Heat Treatment [ll)

Il25-ll75'F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, Furnace cooled.

i

• Estimated per NRC Regulatory. Review Plan MTEB 5-2 i

WP1331:lD/062392

-13

i Table 10 l

Beaver Valley Unit 1 Materials Certification Information Component:

Surveillance Plate (Vessel Lower Heat No.: C6317-1 Shell). B6903-1 HILL Chemical Analysis [12)

C Mn P

S Si Ni Mo Cu Al Co 0.20 1.31 0.010 0.015 0.180 0.54 0.55 0.20 0.028 0.014 Charny impact and Fracture Tests - Transversqll23 Temp 'F Ft-Lbs

% shear Lat. Exp. (inches)

-100-4.0 3

4

-100 2.5 3

2

-100 4.5 3

0

'100 5.0 3

0

-50 11.0 9

6

-50 6.0 3

2

-50 11.0 5

7

-50 11.0 5

5

-50 13.5 5

7 10 28.5 23 23 10 40.0 29 31 10 20.0 21 26 10 28.5 18 19 10 33.0 27 Is 40 31.0 27 27 40 33.0 28

' 30 40-46.5 33 37

.40 34.0 28 29 40 41.0 33 33 40 30 31 32-110 65.0

' 51-55 110-63.5 51 54 l-110 64.0 53 60 110 77.0 57 63 160-76.5 100 67 160 76.0-100 60 1.

160 82.0 100 67

.160 --

82.5

!OO 69 l

160 79.5 100 67 R

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,i' Temp 'F Ft Lbs

% Shear Lat. Exp. (inches) 210 83.0-100 80 210 82.5 100 66 210 62.0 100 69 210 75.0 100 65 Temp. 'F Drop Weights NDT RTNDT U SI'.

-50'F 27'F 80 f* lb HeatTreatment[12) ll50-1650*F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Water quenched.

- 1200-1250*F,.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Air cooled.

Il25-ll75'F i 25*F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, Furnace cooled.

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Table 11

":a. -- i Beaver Valley Unit I taterials Certification Information-i r

. Component:: Surveillance Weldment Heat No.: 305424, Flux Linde 1092, f>

Flux 3899, B4 Mod.

MILL Chemical Analysis [12]

I L

C Mn-

.P.

5 Si til Mo Cu Al Co

-0.110 1 370 0.018 0.006 t.

270 0.620 0.480 0.260 0.010 0.014 Charpy Impact and Fracture Tesn[12]

c.._

~ Temp 'F Ft-Lbs

% Shear

, Lat. Exp. (inches)

-150 4.0' 50' O.0

-150 2.5-40 0.0

-150-2.0 25 0.0

-60 37.0 35 28.0

-60'

-27.0 35 22.0 26.0 30

.22.0

-25

-88.0 85 68.0

-25' 77.0 70 53.0 4 75.0 70 19.0 0.-

80.0 75 57.0 s"

0 66.5' 50 47,0

's -

.0 88.0 75-60.0 a

0,c4 -

2100?

'-100.0 95-78.0 100; 108.5 99 81 0 cl00

-117.5 luu

-88.5 t

210'

'110.0 -

100 84.0 E # 'i^

1210

'103.5 100 82.0

^

210 122.0:

J100 l

93.0 g

~ Temp. 'F-FDrop-W6ights NDT RINDT l

USE i

-60*F-

-60'F 112 ft lb.--

Post Weld Heat Treetment[12].

.g 7-L1125-1175'F:'40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace. cooled.

.m s.-C - a g;

WP1331:lD/062392.

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Beaver Valley Unit 1 Materials Certification-Infermation

Component: :Surveillcnce Weld Heat-Affected Zone Material MILL Chemical Analysis C-Mn P

Si Ni

-Mo Cu Al Co Charpy Impact and Fracture Tests [12)

Temp *F Ft-Lbs

% Shear Lat. Exp. (inches)

-150

-5.0 20 0.0

-150-4.5 35 0.0

- 150 :-

7.0 20 0.0

-40 50.0 35 29.0

-40 50.5 50 31.0

-40 44.5-50 25.0 L102.0 70 69.0 t

-0 u

0-75.0 60-49.0

-0 85.0 65 55.0 90.0-75 58.0 20-91.0.

. 7 10 61.0 p

20' 96'.0 80 66.0

- b

-100 129.0 100 72.5-100-111.0-100 76.0

=1001 100.0 100 64.0-210~

-114.0.

100 66.0 1210' 138.5 100' 76.0-

'210 131.5 100,_

76.0 g-di Temp' 'F:

Drop Weights:

NDT RTNDT-~

U S E--

o

-40*F

-40aF 128 ft-lb i

Ic i

s i' -

t t

(WP1331:lD/062392' 17-LYi oh

The' nil-ductility transition temperature (NDTT) is defined as the maximum temperature at which a standard drop weight specimen breaks when tested according to the provisions specified in ASTM E-208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels".

The NDTT was determined ' r each material by dropweight tests (ASTM E-208) performed by Combustion Engineering.

The unirradiated RiNDT of thu altline region matarHls was established from the drop weight NDTT tests and the Charpy V-notch tests, using the guidance provided in NUREG-0800, Branch Technical Position, MTEB 5-2, " Fracture Toughness Dm;g emggt$", and gubartjcle NB-2300 of the ASME Boiler and Pressure iessel Code,Section III.

The specific criteria used for each of the Beaver Valley Unit 1 beltline plates and weldments are as follows:

• The NDTT temperature, as determined by drop weight tests (ASTM E-208) is the RTNDT if, at 60*F above the NDIT, at least 50 ft-lbs of energy and 35 mils lateral expansion are obtained in Charpy V-notch tests on transverse specimens.

Otherwise, the RTNDT is the temperature at which 50 ft-lbs and 35 mils latert.. expansion 're obtained on transverse Charpy specimens, minus 60*F.

These criteria were appiled in determining the initial RTNDT values for the surveillance plate and weldment.

If drop weight tests were not-performed, but full Charpy-V-notch curves were obtained, the NOTT for SA-533 Grade B, Class 1 plate and weld material may be assumed to be the higher of the 30 ft lb temperature, or 0*F.

The Charpy V-notch data for intermediate shells, heats B6607-1 and B6607-2, and luwer shells, heats B7203-2 and B6903-1 were from transverse specimens tested by Westinghouse.

The initial RTNDT values were the 30 f t-lb temperature from the full Charpy V-notch curves.

if measured values of RTNDT are not available, the generic mean values must e

be used:

0*F for welds made with Linde 80 flux, and -56af for welds made with Linde 0091,1092 and 124 and ARCOS B-5 weld fluxes, as per 10 CFR 50.61, " Fracture-Toughness Requirements for Protection Against Pressurized O

WP1331:lD/062392 18

{

4 i 4.t.i-7 --

Thermal Shock Events".

These criteria were used in establishing the initial RTNOT values -for the intermediate shell seams19-714 A and B, intermediate shell to lower shell seams11-714, and lower shell seams 20-

.714 A and B.

The unirradiated upper shelf energy for the beltliiie region plates, and weldments were determined from Charpy V-notch tests using transverse specimen data. The upper shelf energy is the average of the transverse Charpy energy values for-specimens exhibiting fully ductile behavior (i.e.100% shear).

The surveillance mat 6 rial charpy and tensne specimens received ~ heat treatments,-including stress relieving operations, equivalent to those given to the actual reactor vessel materials as required by Section III of the ASMr Boiler and-Pressure Vessel Code. Combustion Engineering supplied Westinghouse Electric Corporation with secticas of A533 Grade B, Class 1 plate used in the core region of the Beaver Valley Unit 'l reactor pressure vessel for use in the

-Reactor Vessel Radiation Surveillance Program.

The sections of material were removed from the 7 7/8-inch lower shell course of'the pressure vessel.

Combustion Engineering, Inc., also supplied a weldment made from sections of the intermediate;shell~ plates B6607-1 anj B6607-2 using weld wire p

- representative of that used in_ the original fabrication. The plates were L

-produced by:Lukens' Steel Co.

The heat treatment history of the pressure

vessel beltline-region material and surveillance materials are given in Tables

21through 12.

?.-

- WP1331ilD/062392 19 i-

ca;

' t.]

---Question 3: ' Generic Letter 88-ll Commitments:

Question 3a: How the embrittlement effects of operating at an irradiation temperature-(cold leg or recirculation suction temperature) below 525 degrees F were con::idered.

In particular licensees are reauested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy.

Responsg:.

-Beaver Valley; Unit 1;has not-operated at temperatures below 525'F, and therefore~, will have no impact on prediction of RTNDT or upper shelf energy.

Question 3b: How the Beaver Valley Unit 1 surveillance results on the A

predicted amcunt o" embrittlement were considered in GL 88-11.

Response

~ The Beaver Valley Unit I surveillance data are credible in all respects as, judged'by the-criteria-defined.in. Regulatory Guide l'.99, Revision 2:

In the Beaver Valley Unit I generic letter

.88-11 response, the surveillance re ults were considered in predicting.the ARTNDT-and margin values and calculating chemistry factors required-by Regulatory Guide 1.99 Revision..

The current operating limits incorporate. these values.

t

! 0uestion:3c:~ 'If a measured increase-in reference. temperature exceeds the mean-plus-two standard deviations predicted by Regulatory '

Guide ~ 1.99, Revision 2, or if, a measured decrease in upper shelf-energyLexceeds.the value predicted.using the: guidance in Paragraph'C.I.2 in Regulatory Guide O.99, Revision 2, the licensee is reques'.ed to report the inforniation and describe the effect of the surhillance results on the adjusted referenu-temperature and Charpy upper shelf energy for each beltline y

. WP1331:1D/0623924 20 r

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[

material as predicted for December 16, 1991, and for the end of its current license.

Response

Comparison of the Beaver Valley Unit I survr:' lance capsule data with predicted changes in the 30 ft-lb transition temperature and upper shelf energy using the methods of Regulatory Guide 1.99, Rev. 2 are provided in Tabia 13.

The measured percent decease in upper shelf energy are less than that predicted by the Regulatory Guide. The measu.ed transition temperature increase for-plate B6903-1 have exceeded the mean-plus-two standard deviation bound predicted by Regulatory Guide l.99, Revision 2.

The adjusted reference temperature (ART) for each beltline material are provided in Table 14.

The

-larger measured shift in RTNDT for lower shell plate B6903-1 is accounted for 3

-by the use of this data in determining the current operating limits.

3.0 CONCLUSION

S

-The following is a summary of the conclusions relative to GL 92-01:

The Beaver Valley Unit i surveillance programs meets the requirements of Appendix H to 10 CFR Part 50.

The Beaver Valley Unit I reactor vessel was construct'ed in accordance with the Winter 1968-Addenda of Section III of the AdME Code.

The projected:E0L upper shelf energy values for the Beaver Valley

-Unit 1 beltline materials are above the 50 f t-lbs criteria.

' Beaver Valley Unit 1-has= not operated at temperatures below 525*F.

4 WP1331:1D/062392 21 l

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3;

.;i %

.a f.

Table ~ 13

. Comparison of; Beaver Valley Unit 1 Reactor.' Vessel Surveillance Capsule Charpy Impact Test Results

.With Regulatory Guide 1.99 Revision 2 Predictions ARTNDT (30 ft-lb Increase).

A Use

-R.G. I'.99 Rev. 2.

R.G. Pred. +-

Heasured R.G. 1.99 Rev. 2 Measured Material-Capsule

'(*F) 2ar

(*F)

(%)

(%)

' Plate B6903 'V 94 128 130 22 14.9

-(longitudinal)

U 125 159 120 26 26.1 W

.140 174 150 29 14.9

. Plate B6903-1 V.

94-128~

140 22.

6.2 (transverse)

U 125-159 135 26 2.5 W --

140

-174 185 29 26.2 Weld Metal V

118' 174 150 29 21.4 U

' 158 ~

.214 155 35 25.9 W

176-

'232 185 38 30.4 HAZ Metal V

0 10.4 U

35 18.0 W

60 10.9 WP1331:10/062392 22

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~

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41

,,3...<

Table 14:

i:

'meaver valley Unit.1' adfasted saference Teacerature f" -

values for 12/16/91-and End-of-License (32 EFFT).

. I,0

~

CU 22 DtTEST

.12/16/31 32 EFFY Beltlias Material (Wet) (NT%)

'CF

'(*F) -

Mergin ART ( *F).

RET

h

j'

. Inter. = hall, 36507-1 3.14 6.62 104.$4.'

43' 34

~182 210 P, '

Inter. shall,936807-2.

'O.14^

'G.62

~108.50 73-34 212

,240-M ahall, T 6903-1,.

0.20 0.54

'141.86-

-27 34 209 243.

E 167.81 27 17*

219 266.

using'S/C data Lower shall,'37243-2..

0.14 9.57 98.65 20 34 157 185 Long.. wold,' 305424-

0.28 0.63-191.65

~56 66 121 188 191.33 44*

99' 166 t

5 usiner S/C dat a a

i Long.. weld, 3G5414.

6.35 0.61 213.45

-56 66 134 206 Circumferential' weld 8.29' O.07 132.90

--56 66 148 186 2 12).

• EAergia was @ M according to ne&

• logy, specified in Reg. Guide 1.33 Rev.

l 23

.p.-

l l.

~

T

a 3,,.

j s.

Tha Beaver Valley Unit I surveillance data are credible in all e-respects as judged by the criteria defined in Regulatory Guide 1.99, Revision 2.

The response to GL 88-11 considered the Beaver Valley Unit 1 surveillance results in predicting ARTNDT and calculating chemistry factors used in RTPTS calculations and current operating limits-incorporate these results.

The measured transition temperature increase for pla e 86903-1 have exceeded the mean-plus-two standard deviation bound predicted by Regulatory Guide 1.99, Revision 2.

The effect of this was taken into account when the PTS submittal and operating limits were generated.

4.0 REFERENCES

1.

Material Certification Report, Babcock & Wilcox Company, January 5, 1971. _(PlateB6607-1,heatC4381-1) 2.

Westinghouse. Laboratory Services EML No. A 1843, October 8, 1973.

3.

Materials Certification-Report, Babcock & Wilcox Company, Jaluary 5, 1971.

(Plate 86607-2, heat C4381-2)

~

4.

Westinghouse Laboratory Services,-EML No. A 1844, October 8, 1973.

5.

Material Certification report, Combustion Engineering, March 17, 1970,

.Jcb No. X-96253-001BS.

[

6 ',

. Westinghouse Laboratory Services,.EML No. A 1845, October 8, 1973.

7..

Materials Certification Report, Combustion Engineering, March 17, 1970,

' Job No. X-96253-003BS.

WP1331:lD/062392

'24

s 8.

. Westinghouse Laboratory Services, EML No. A 1019, March 29, 1972.

9.

Weld Inspection Report, Combustion Engineering, February 1971.

10.

Weld Inspection Report, Combustion Engineering,-March 1970.

11.

Weld Inspection Report, Cembustion Engineering, March 1971.

12.

WCAP-8457, "Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program Report, Westinghouse, October 1974.

-WP1321:lDg2392 25

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'g g

e BEAVER VALLEY UNIT 2 RESPONSE TO GENERIC LETTER 92-01 REACTOR VESSEL STRUCTURAL INTEGRITY, 10 CFR 50.5t.(f) 1 WP1334:10/062392-L!

_ _ _ _ -_-_ _ _ _- __-:_-_=__=--_----_-_--_---_--

--_-___ _ _ _ _ x

.i.

.s TABLE OF CONTENTS SECTION TITLE PAGE 1.0 INTPODUCTION 1

2.0 REACTOR VESSEL STRUCTURAL INTEGRITY REQUlttED 2

INFORMATION-

-3.0

-CONCLUSIONS 8

4.0 REFERENCES

10 V

b

}

WP1334:10/062392

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1.0 INTRODUCTION

On 6 March 1992 the U.S. Nuclear Regulatory Commission (NRC) issued

. Revision 1 of Genetic Letter (GL) 92-01 to obtain information needed to assess compliance with the requirements and commitments regarding reactor vessel structural integrity.

Section 50.60(a) of Title 10 of the Code of Federal Regulations (10 CFR 50.60(a)) requires that all light watcr nuclear power reactors, licensed by the NRC, meet fracture toughness requirements and have a material surveillance program for tne reactor pressure boundary.

These-requirements are set forth in Appendices G and H to 10 CFR Part 50.

If the requirements of Appendices G and H cannot be met, an exemption pursuant to 10 CFR 50.12 is reauired.- 10 CFP. 50.61 prov*oes fracture toughness requirements for

. protecting pressurized water reactors against oressurized thermal shock events.

On.a related topic, the U.S. Nuclear Regulatory Commission on 12-July-88 issued Generic Letter 88-11, "NRC Position on Radiation Embrittlement of

-Reactor Vesse' Materials and its Impact on Plant Operations".

The purpose of this generic letter was to call attention to Revision 2 of

- Regulatory Guide 1.99, " Radiation Embrittlcment of Reactor Vessel

-Materials".

Each. licensee was required to subndt the results of a technical. analysis relative tof the implementation of this regulation.

This submittal

/

-addressed the following:

1)-

Recalculation of ad,iusted _ reference temperature values at tne 1/4

~

and 3/4 reactc,r-vessel wall thickness locations for all potentially limitina materials using'hegulatory Guide 1.99 Revision 2.

2)

Determine the date'of applicability of the P/T limits using T

Regulatory Guide Revision 2.

WP1334ilD/062392 1

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,g,c 3)

If the use' of the Revision ? methodology resulted in a modification

.of the P/T limits contained in the Technical Specification, a

' proposed scheduie was required to be submitted that defined whatever action needed to be taken, including those required to address an expected restriction in operating flexibility.

4 l

-Par GL 92-01, "This gen eric letter is part of a program to evaluate reactor vessel integrity and take regulatory actions, if needed, to ensure that licensees and permit holders are complying with the requirements. of 10 CFR 50.60 and-10 CFR 50.61, and are fulfilling commitments made in response to GL 88-11."

This report describes the methods used and the results obtained in evaluating Belver Valley Unit 2 relative to GL.92-01.

2.0 REAC10R VESSEL STRUCTURAL INTEGRITY REQUIRED INF0P}1ATION LQuestion1:

Compliance with Appendix H to 10 CFR 50:

-Determire if the Reactor Vessel Irradiation Surveillance Program is in

- compliance with Appendix H to 10-CFR 50 by determining which version of ASTM E.-185 was used to develop the Reactor Vessel. Irradiation Surveillance Program.

y If the. program does not meet ASTM.E-185-73, -79, or -82 and is not part of an approved " integrated surveillance program", describe the actions that have or will -be' tthn to ensure that the Reactor Vessel Irradiation Surveillance Program. complies with Appendix H to 10 CFR 50.

g 3

Response

'The Ceaver Valley Unit.2 vessel was fabricated in accordance with the requirements of the ASME C6de,1971 Edition including Addendh through Summer

'1972. ASTM'E185-73:was the standard in place at the time the surveillance

. program was designed. The Beaver Valley Unit 2 surveillance program complies

~

with ASTM E185-73. Testing o# surveillance capsules after July 26, 1983 has 4

m-

.WP1334r10/062392; 2-a

-a (beenperformedinaccordancewithASTMStandardversionE185-82.

Furthermore, since the surveillance program design was approved during the FSAR licens'ing process, the capsule testing program has been approved as part of the plant Technical Specifications.

Therefore, it is determined that the surveillance progra'n for Beaver Valley Unit 2 meets the requirements of Appendix H to 10 CFR Part 50 and that an exemption request is not considered necessary.

Question 2:

Compliance with Appendix G to 10 CFR 50:

Question 24: Calculate the Charpy Upper Shelf Energy (USE) for all of the beltline materials of the Beaver Valley Unit 2 reactor vessel using the methods of Regulatory Guide 1.99, Revision 2 for the end of the current license.

If the calculated USE of the limiting beltline weld or plate of the Beaver Valley Unit 2 reactor vessel is below 50 ft-lbs, describe the actions that have or will-be taken persuant to Pare,raphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50 to addr.s this issue.

Responjie:

A'l vessel beltline materials for Beaver Valley Un't 2 are characterized in l

Table 1-in terms of initial upper-shelf' energies, < oprr chemistries, and an\\/;cipated fluence levels at -the 1/4. thickness pc.ition in the vessel.

The in.itial upper-shelf energy levels for the beltline materials were determined from the unirradiated Charpy specimen data.

Average copper values were

-determined from available chemistry data.

Sources of chemistry data included fabrication material.' certifications and weld qualifications, and surveillance

. capsule specimen evaluations. Also, available chemistry data from plants with the same heat of weld wire as Beaver Valley Unit 2 beltline region welds were factored into the calculated copper values.

Fluence at the 1/4 thickness

' position in'the vessel is' computed using the formula:

f.luencel/4 T = fsurf (e-0.24x) u WP133_4:lD/062392 3

• 1 wherc ! surf is the calculated value of fluence at the inside surface of the vessel, and x is the distance in inches from the inside surface (ignoring cladding) to the 1/4 thickness depth of the vessei wall.

The predicted percent decrease in upper-shelf energy as a function of copper content and fluence was established using Figure 2 from Regulatory Guide 1.99, Revish.. ;." M For materials that are estimated to be low in upper shelf energy (i.e., less than 50 foot-pounds) using the above prediction method, alternative methods of determining upper-shelf energy drop may be considered when available.

The calculcted end-of-life upper shelf energy values (see Table 1) for-the Beaver Valley Unit 2 beltline materials are above the 50 foot-pounds criteria.

i Question 2b: If the Beaver Valley Unit 2 reactor vessel was constracted to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition, provide the following information:

a) All_ Charpy and drop weight test results for all unirradiated beltline materials, the unirradiated reference temperatures for each beltline material, and a description of the methods used for calculating these values, b) A! description of the heat treatment performed on all beltline and surveillance n.aterial.

c) The heat numbers for each beltline plate or foiging and the weld and flux lot number used to fabricate each beltline material and surveillance material and weld.

3 d) A description of the chemical composition of each beltline 3

and surveillance material, e).The heat number of the wire used for determining the weld chemical composition.

WP1334:lD/062392 4

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,y,.

Table 1 Beaver Valley Unit 2 End-of-License (32 EFPY) Uppe-Shelf Energy (USE) 1/4T E0L(bXnitial Fluence USE E0L USE Beltline Material Cu (Wt. %) % Decrease (x 1019 n/cm2) (ft-lb) (ft-lb)

Intermediate Shall Plate, 0.07 26 3.87 83 61.4 B9N1-1 Intermediate Shell Plate, 0.07 26 3.87 75.5 55.9 89004-2 Lower Shell Plate, 89005-1 0.08 26 3.87 82 60.7 Lower Shell Plate, B9005-2 0.07 26 3.87 77.5 57.4 Longitudinal Weld Seams, 0.06(a) 22 1.16 144.5 112.7 Ileat 83642 Intermediate to Lower Shell 0.06(a) 26 3.87 144.5 106.9 Circumferential Weld, Heat 83642 (a)

All welds, including surveillance, fabricated from same heat of weld wire.

Previously reported value from chemistry on surveillance weld.

Value reported is average of vessel welds and surveillance weld chemistry data from Combustion Engineering records.

(b)

Calculated from neutron exposure projections previded in Capsule U.[2]

WP1334:1D/062492 5

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t

Response

'The Beaver Valley Unit 2 reactor vessel was fabricated in accordance E th the requiremen_ts of the ASME-Boiler and Pressure Vessel Code 1971 Edition including Addenda _thrtagh Summer 1972.

Question ~3:

Generic Letter 88-11 Commitments:

L 0uestion 3a:. How the embrittlement effects of operating at an irradiation temperature (cold leg or_ recirculation suction temperature)

below 525 degrees F were considered.

In particular licensees are requested to' describe consideration given to determining the

effect.of-lower irraaiation temperature on the reference

. temperature and on the Charpy upper shelf energy.

R

Response

Beaver Va'lley Unit-.2 has not operated. at t'emperatures below 525'F,_ and therefore, willLhave no: impact _ on prediction of RTNDT or Upper Shelf Energy.

' Question 3b: "uw the Beaver Valley Unit '2 surveillance-results on the

~

predicted amount-of embrittlement were considered in GL 88-11, 2

Response

LThelBeaver Valley Unit 2 surveillance data are credible in all respects as

_-judged;bylthe criteria ~ defined in Regulatory Guide 1.99, Revision 2.[1]

In

'the Beaver Valley _ Unit!2 Generic 1 Letter 88-11, the ' single set of avail _able -

surveillance results was not considered. in predicting the ARTNDT-and

' calculating. chemistry factors used inLRTPTS calculations.

~

10uestion 3c: lIf a measured increase in reference temperature exceeds the mean;plus-two standard deviations predicted by Regulatory 2 Guide 1.99, Revision 2,ior if a measured decrease in upper shelf L

-WP1334:lD/0623921

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J

, s.

i energy exceeds the value predicted using the guidance in Paragraph C.I.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and for the end of its current license.

Response

Comparison of the Beaver Valley Unit 2 surveillance capsule data with predicted changes in the 30 ft-lb transition tempa ature and upper shelf energy using the methods of Regulatory Guiae 1.99, Rev. 2 are provided in Table 2.

This measured percent. cease in upper shelf energy are less than that predicted by the Regulatory Guide.

The measured transition temperature increase for plate 89004-2 are bounded by the mean-plus-two standard deviation definedbyRegulatoryGuide1.997 Revision 2.

i I-WP1334:lD/062392 7

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"4.:(at TABEE~2

. COMPARISON 0F BEAVER VALLEY. UNIT 2~

! REACTOR VESSEL: SURVEILLANCE CAPSULE 0 HARPY IMPACT TEST RESULTS tWITH REGULATORY' GUIDE:1.99 REVISION;.2 PREDICTIONS'

'ARTNDT (30 ft-lb Increase)-

A Use R.G. 1.99 Rev. 2' R.G. Pred. +-

~ Measured.

.R.G. 1.99 Rev. 2

. Measured' Material:

Capsule

(*F) 2a:

~(*F)

(%)

(%)

Plate 89004-2

.U~

27:

61 15 17 0

(longitudinal

)

P1 ate 89004-2 U

-27 61 30 17 0

(transverse)~

Weld Metal U

' 37 -

93 25 19 4

b 9P1334:lD/062392 8

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<*I i

e ly 9

3.0 CONCLUSION

S The following is a summary of the conclusions relative to GL 92-01:

The Beaver Valley Unit 2 surveillance programs meets the requirements of Appendix H to 10 CFR Part 50.

The Beaver Valley Unit 2 reactor vessel was constructed in accordance with the requirements of the ASME Boiler and Pressure Vessel Code 1971 Edition including Addenda through Summer 1972.

The projected end-of-life upper shelf energy values for the Beaver Valley. Unit 2 beltline materials are above the 50 ft-lbs criteria.

Beaver Valley Unit 2 has not operated at temperatures below 525'F.

The Beaver Valley Unit 2 surveillance data are credible in all respects as judged by the criteria defined in Regulatory Guide 1.99, Revision 2.

The response to GL 88-11 did not consider the Beaver Valley Unit 2 surveillance results in predicting ARTNDT and calculating chemistry factors used in RTPTS-calculations.

The measured transition temperature increase for plate B9004-2 are bounded by the mean-plus-two standard deviation defined in Regulatory Guide 1.99, Revision 2.

4.0

' REFERENCES 1.

negulatory ' Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel ~ Materials," USNRC, May 1988.

WP1334:lD/062392 9

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4 2._

S. E. Yanichko, et. al, " Analysis of Capsule U from the Duquesne Light Company. Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-12406, Westinghouse, September 1989.

WP1334:lD/062392 Jo

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