ML18100A640

Forwards Suppl Response to GL 92-01,Rev 1, Rv Structural Integrity, Including Addl Pressure Vessel Matl Transverse Charpy V-notch Impact Data Located by Westinghouse for Intermediate Beltline Plates B2402-1,B2402-2 & B2402-3

ML18100A640
Person / Time
Site:	Salem
Issue date:	09/29/1993
From:	Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NLR-N93149, NUDOCS 9310070242
Download: ML18100A640 (30)

Text

Public Service Electric and Gas Company I

Stanley LaBruna Public Service Electric and Gas* company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Vice President - Nuclear Engineering SEP 2 9 1993 NLR-N93149 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

SUPPLEMENTAL RESPONSE TO GENERIC LETTER 92-01, REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY, 10CFR50.54(f)

SALEM GENERATING STATION UNIT NO. 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Generic Letter 92-01, Revision 1 requested PSE&G to submit information to enable the NRC to assess compliance with 10CFR50.60 and 10CFR50.61, and the fracture toughness and material surveillance requirements for the Reactor Coolant Pressure Boundary set forth in 10CFR50, Appendices G and H.

The Salem Unit 1 response to Generic Letter 92-01, Revision 1 was submitted to the NRC in Letters NLR-N92081 dated June 30, 1992 and NLR-N93125 dated August 4, 1993.

These submittals included all known longitudinal Charpy V-notch impact data for the six (6) pressure vessel beltline plates.

Since the date of these submittals, additional Salem Unit 1 pressure vessel material transverse Charpy V-notch impact data was located by Westinghouse for intermediate beltline plates B2402-1, B2402-2 and B2402~3.

The Salem Unit 1 response to the Generic Letter has therefore been revised to include this new Charpy V-notch data and is provided in Attachment 1.

At PSE&G's request, Westinghouse completed a search of their files and found no additional beltline plate Charpy V-notch impact data for Salem Unit 1.

(~~-- ___.....,_,....._

i 931'00702~9309'">9--o*

~DR. ADOCK 0~000272

• PDR.

Document Control Desk NLR-N93149 2

SEP 2 9 1993 Please do not hesitate to contact us if there are any questions regarding this submittal.

Attachment Affidavit Sincerely, c

Mr. T. T. Martin, Administrator - Region I

u. s. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J. C. Stone, Licensing Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr.

C~ S. Marschall (S09)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625

I. N-LR-N93149 e

e ATTACHMENT 1 SALEM UNIT 1 RESPONSE TO GENERIC LETTER 92-01, REVISION 1 REACTO~ VESSEL STRUCTURAL INTEGRITY SEPTEMBER 8 1 1993 REVISION 1

NLR-N93149 PSE&G has prepared the following information in response to the requests in Generic Letter 92-01, Revision 1 titled "REACTOR VESSEL STRUCTURAL INTEGRITY".

In the following text, the individual requests for information are stated in bold face type as written in GL 92-01, and each request is followed by the PSE&G response in regular (non-boldface) type.

1.

Certain addressees are requested to provide the following information regarding Appendix H to CFR Part 50:

Addressees who do not have a surveillance program meeting ASTM E 185-73, -79, or -82 and who do not have an integrated surveillance program approved by the NRC (see Enclosure 2),

are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.

Addressees who plan to revise the surveillance program to meet Appendix II to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC staff for review.

If the surveillance program is not to be revised to meet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part so under 10 CFR S0.60(b).

Response

ASTM E-185-70 was the standard in place at the time the surveillan9e program was designed.

The Salem Unit 1 surveillance program complies with ASTM E-185~70. Testing of surveillance capsules after July 26, 1983 has been performed in accordance with ASTM standard version E-185-82.

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

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

2.

certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part so:

a.

Addressees of plants for which the Charpy upper shelf energy is predicted to be less than so foot-pounds at the end of their licenses using the guidance in Paragraphs c.1.2 or c.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the 1

Revision 1

NLR-N93149

Response

Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or v.c of Appendix G to 10 CFR Part

50.

Table 1 contains the unirradiated, the December 16, 1991 and the EOL (August 13, 2016) Charpy upper shelf energy values for Salem Unit 1 beltline materials.

The December 16, 1991 and EOL values were calculated using Figure 2 of Regulatory Guide 1.99 Revision

2.

The calculated EOL Charpy upper shelf energy values for all the beltline materials which have known unirradiated USE values are predicted to be above the 50 ft-lb criteria.

b.

Addressees whose reactor vessels were constructed to an ASME Code earlier than the summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph II.A of 10 CFR Part so, Appendix G:

(1)

The results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of determining the unirradiated reference temperature from the Charpy and drop weight test; (2)

The heat treatment received by all beltline and surveillance materials; (3)

The heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate each beltline weld; (4)

The heat number for each surveillance plate or forqinq and the heat number of wire and flux lot number used to fabricate the surveillance weld; 2

Revision 1

NLR-N93149

Response

(S)

The chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillance material; and (6)

The beat number of the wire used for determining the weld metal chemical composition if different than Item (3) above.

The Salem Unit 1 reactor vessel was constructed to the 1965 Edition, through Winter 1965 Addenda of Section III of the ASME Code.

Thus, the Salem Unit 1 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 material test requirements and acceptance standards that were current at the time of reactor pressure vessel construction.

(Note that the chemical composition of the welds was determined from the weld wire heat numbers of the actual welds.)

The nil-ductility transition temperature (NOTT) 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 NOTT was determined for each beltline material by dropweight tests (ASTM E-208) performed by Combustion Engineering, except for welds 2-042A, 2-042B, 2-042C, 3-042A, 3-042B, 3-042C, and 9-042.

The unirradiated reference temperature (RTNDT) of the beltline region materials was established from the drop weight NOTT tests and the Charpy v-notch tests, using the guidance provided in NUREG-0800, Branch Technical Position, MTEB 5-2, "Fracture Toughness Requirements", and the ASME Boiler and Pressure Vessel Code,Section III.

The following three paragraphs summarize pertinent information from these two references, and the fourth following paragraph summarizes information from 10 CFR 50.61, "Fracture.Toughness Requirements for Protection Against.

Pressurized Thermal Shock".

3 Revision 1

1*

NLR-N93149 The NOTT temperature, as determined by drop weight tests (ASTM E-208) is the RTNOT if, at 60°F above the NOTT, at least 50 ft-lbs of energy and 35 mils lateral expansion are obtained in Charpy v-notch tests on transverse specimens.

Otherwise, the RTNOT is the temperature at which 50 ft-lbs and 35 mils lateral expansion are obtained on transverse Charpy specimens; minus 60°F.

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 asstimed to be the higher of the 30 ft-lb temperature, or 0°F.

If transverse Charpy V-notch specimens wer*e not tested, the temperature at which 50 ft-lbs and 35 mils lateral expansion would have been obtained on'transverse*specimens may be estimated by using 65% of the values from longitudinal specimens, or increasing the 50 ft-lb and 35 mil lateral expansion temperatures for longitudinal specimens by 20°F.

If measured values of RTNOT are not available, the generic mean values must be used:

0°F for welds made with Linde 80 flux, and -56°F 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 Thermal Shock Events".

The Charpy V-notch data for.the beltline region plates and surveillance test plates tested by Combustion Engineering were taken in the longitudinal direction.

The RTNOT values for each of these materials were determined by either (1) increasing the temperature at which 50 ft-lbs and 35 mils lateral expansion was obtained for longitudinal specimens by 20°F or (2) using 65%

values for the energy and lateral expansion data, in order to estimate the temperature for which 50 ft-lbs and 35 mils lateral expansion would have been obtained for transverse specimens.

These values were then reduced by 60°F.

Measured data does-riot exist for the welds; therefore the generic mean value of -56°F is used.

P*er the above defined methodology, the RTNOT values for surveillance weld metal and HAZ metal were determined to be equal to their NOTT values.

The unirradiated upper shelf energy was determined from Charpy V-notch tests using transverse specimen data, or by using longitudinal data multiplied by 65% to estimate transverse data.

The upper s.helf energy is the average of the transverse Charpy energy values for specimens exhibiting fully ductile behavior (i.e. 100% shear), at a given test temperature.

Typically, specimens are tested in sets of three at each test temperature.

The set having the highest average may be regarded as defining the upper shelf energy, as per ASTM E...;185-82..

4 Revision 1

I' I

NLR-N93149 The upper shelf energy values for beltline region plates B2402-1, B2402-2, and B2402-3 were determined by the average of three 100%

shear energy values using transverse Charpy v-notch data.

The upper shelf energy values for beltline region plates B2403-1, B2403-2, and B2403-3 and surveillance test plates were calculated by multiplying the average of the 100% shear longitudinal Charpy V-notch data by 65%.

The shelf upper energy values for the surveillance weld materials were determined by the average of the three 100% shear energy values.

Upper shelf energy values were not calculated for the beltline region welds because full Charpy V-notch curves were not generated for these materials.

The surveillance material Charpy and tensile 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 ASME Boiler and Pressure Vessel Code.

Combustion Engineering supplied Westinghouse Electric Corporation with sections of A533 Grade B, Class 1 plate used in the core region of the Salem Unit 1 reactor pressure vessel for use in the Reactor Vessel Radiation Surveillance Program.

The sections of material were removed from the 9-inch intermediate shell course of the pressure vessel.

Combustion Engineering, Inc, also supplied a weldment made from sections of the intermediate shell plate B2402-3 and adjoining lower shell plate B2403-1 using weld wire representative of that used in the original fabrication.

The plates were produced by Lukens Steel Co.

The heat treatment history of the pressure vessel.beltline region material and surveillance materials are given in Tables 2 through 16.

The Salem Unit 1 Reactor Pressure Vessel Surveillance Program also contains correlation monitor material.

Correlation monitor material was supplied by the Oak Ridge National Laboratory from plate material used in the AEC-sponsored Heavy Section Steel Technology (HSST) program.

This material was obtained from a 12-

• inch thick A533 Grade B, Class 1 plate (HSST Plate 02) which was provided to Subcommittee II of the ASTM Committee E-10 on Radioisotopes and Radiation Effects to serve as correlation monitor material in reactor vessel surveillance programs.

The plate was produced by Lukens Steel Co. and heat treated by Combustion Engineering, Inc.

The heat treatment history and

. quantitative chemical analysis of the correlation material are presented in Table 16.

3.

Addressees are requested to provide the.following information regarding commitments made to respond to GL ss-11:

a.

How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525°F were considered4

• In 5

Revision 1

I' NLR-N93149

Response

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

The PSE&G Operations Department performed a review of its policies and procedures to determine if the stated scenario, i.e., cold leg temperature below 525°F while at power, has occurred for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> total.

This review included Integrated Operating Procedure-3 "Hot standby to Minimum Load", which states that Tave must be verified greater than 541°F within 15 minutes of achieving criticality.

In addition, Technical Specification 3.1~1.5 requires that while in Mode 1 and 2, Tave must be greater than 541°F.

This LCO requires the temperature to be restored within 15 minutes or be in Hot Standby within an additional 15 minutes.

Based on department procedural requirements, it can be concluded that the outlined scenario has not occurred in the past and will not occur in the future at Salem*.

While historically there have been instances during plant transient, where RCS temperature may have gone below 525°F, the cumulative excursion time has been much less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Therefore, the effect of lower irradiation temperature on the reference temperature and Charpy upper shelf energy is negligible.

b.

How their surveillance results on the predicted amount of embrittlement were considered.

Response

As explained *in the PSE&G response to Generic Letter 88-11, the surveillance capsule analyses were conducted using the methods described in Regulatory Guide 1.99 Revision 2 to predict the effects of neutron radiation on the reactor vessel materials.

PSE&G has complied with its commitment to submit a License Change Request to include new heatup and cooldown curves.

Approval for the revised curves was received in January 1990 through Amendment 108 for the Salem Unit 1 Technical Specifications.

c.

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 Charpy upper shelf energy exceeds the value predicted using.the guidance in Paragraph c.1.2 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 6

Revision 1

NLR-N93149

Response

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

Table 17 presents measured and predicted 30 ft-lb temperature increases and upper shelf energy decreases for the three surveillance capsules T, Y and Z which have been removed from Salem 1.

The measured increase in reference temperature does not exceed the mean-plus-two standard deviation predicted by Regulatory Guide 1.99 Revision 2 for any of the surveillance capsule materials as indicated in Table 17.

The measured decrease in Charpy upper shelf energy does not exceed the value predicted using methodology specified in Regulatory Guide 1.99 Revision 2 for any of the surveillance capsule materials as indicated in Table 17.

7 Revision 1

I' NLR-N93149 TABLE 1 SALEM.UNIT 1 UNIRRADIATED AND CALCULATED UPPER SHELF ENERGY (USE) VALUES EOL USE, USE, ft-lb*

USE, ft-lbs (1) ft-lb* (2)

Material Descri~tion Unirradiated December 16, 1991 Augy.st 13 1 201~

Intermediate Shell

90. 7 <3l 73.7 61.9 Plate 82402-1 Intermediate Shell 98.o< 3l 86.2 66.6 Plate 82402-2 Intermediate Shell 104.3< 3l 91.0 72.8 Plate 82402-3 Intermediate Shell Long NA NA NA Weld 2-042A Intermediate Shell Long NA NA NA Weld 2-0428 Intermediate Shell Long NA NA NA Weld 2-042C Lower Shell Plate 92.5< 4l 72.2 67.5 82403-1 Lower Shell Plate 83.0< 4l 64.7 60.6 82403 Lower Shell Plate 05.0< 4i 66.3 62.1 82403-3 Lower Shell Long NA NA NA Weld 3-042A Lower Shell Long NA NA NA Weld 3-0428 Lower Shell Long NA NA NA Weld 3-042C Intermediate to Lower NA NA NA Shell Girth Weld 9~042 NA - Unirradiated upper shelf energy not available because tests were not performed.

In these cases, the December 16, 1991 and EOL USE values.were not determined.

( 1)

(2)

( 3)

( 4).

December 16, 1991 USE calculated at 1/4T location, based on fluences in PSE&G letter SCI-92-0357, 6/11/92, from J. Perrin to J. Chicots.

EOL USE calculated at 1/4T location, based on fluences fromPSE&G letter SCI-*92:-0319,_ 5/14/92, from J. Perrin to_.J. Chicots.

Unirradiated USE values-calculated from transverse test data.

Unirradiated USE values estimated from longitudinal data for beltline materials.

8 Revision 1 I

a*

I NLR-N93149 TABLE 2 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on January 11, 1967.

Component:

Intermediate Shell Plate, 82402-1 Heat No.:

C-1354-1 Mill Chemical Analvsia c

Mn p

s Si Ni Mo Cu

.25 1.43

.010

.013

.30

.52

.47

.24*

Per WCAP 10694 "Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 1 Reactor*vessel Radiation Surveillance Program,"

Table A-1, December 1984.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear Mils Lateral

-40 23.0 5

-40 27.0 5

-40 34.0 10

+10 40.0 20

+10 so.a 25

+10 39.0 20

+60 54.0 40

+60 61.0 50

+60 61.0 50

+110 83.0 85

+110 96.0 100

+110 92.0 85

+160 100.0 100

+160 99.0 100 Temp, OF Drop Weights NOT

-40 lF

-30 lF

-30 lNF

-20 2NF

-30°F 0

lNF Heat Treatment 1550-1G50°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

1225°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1150°F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled to 600°F.

• • Based on transverse test data (Table 2A) 9 Revision 1 Exp.

20 21 27 35 41 33 47 51 51 66 78 68 73 79 RTNDT USE**

12°F 90.7 ft-lbs

NLR-N93149 TABLE 2A SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from Material Testing Report, M.T.L. No.

4395, Westinghouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yariichko, 4/17/70 and Material Testing Report, M.T*L. No. 3582, Westinghouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yanichko, 9/3/70.

Component:

Intermediate Shell Plate, 82402-1 Heat No.:

C-1354-1 Transverse Charpy Impact Data Temp. OF

-100

-100

- 50 50 50 10 10

. 10 60 60 60 110 110 210 210 210 550 550 550 Enerqy, ft-lbs

% Shear 7.5 6

16 6

15 32.5 17.5 22 43 36 43 52 64 70 80 70 91.5 83 97.5 9

9 14 9

14 33 29 25 47 37 42 98 81 100 100 100 100 100 100 10 Revision 1 Mils Lateral Exp.

6 5

14 6

13 20 18 20 38 32 39 41 54 67 66 62 76 71 68

NLR-N93149 TABLE 3 SALEM UNIT.1 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8511, "PSE&G Co. Salem Unit No. 1 Reactor Vessel Radiation Surveillance Program," November 1975.

Component:

Surveillance Material Plate, 82402-1 Heat No.:

Chemical Analysis c

Mn p

s Si Ni Mo Cu Al Cr

.22 1.50

.013

.020

.27

.53

.46

.22

.028

.18 Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear

-100 9

9

-100 4.5 4.5

-100 6.5 9

-so 30 14

-50 14 9

-50 27 14'

+10 43 25

+10 40 25

+10 50 30

+60 59.5 43

+60 63 43

+60 62 48

+110 100 79

+110 80 69

+110 90 77

+160 110 100

+160 98 100

+160 115 100 Temp, OF Drop Weights ND.T Performed by CE 1550 -*1650°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1225°F +/- 25°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1150°F +/- 25°F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

-30oF (based on CE data)

Heat Treatment Water quenched.

Furnace cooled.

Furnace cooled.

11 Revision 1 Mils Lateral Exp.

11 0

9 27 12 22 34 34 37 48 52 49 78 63 71 83 75 84 RTNDT 45°F C-1354 Sn

.018 USE 70 ft-lbs

NLR-N93149 TABLE 4 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on January 11, 1967.

Component:

Intermediate Shell Plate, 82402-2 Heat No.:

C-13S4-2 Mill Chemical Analysis c

Mn p

s Si Ni Mo Cu

.22 1.42

.010

.014

.29

.so

.46

.24*

Per WCAP 10694 "Analysis of *Capsule Y from the Public Service Electric and Gas Company Salem Unit 1 Reactor Vessel Radiation Surveillance Program,"

Table A-1, December 1984.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs *

% Shear

-40 19 s

-40.

lS s

-40 17 s

+10 4S 2S

+10 42 2S

+10 so 2S

+60 S3 40

+60 73 so

+60 74 so

+110 85 60

+110 103 75

+110 96 60

+160 109 100

+160 112 100

+160 114 100 Temo. OF Droo Weiqhts NOT

-40 lF

-30 lF

-20 2NF

-30oF 0

lNF Heat Treatment 15S0-1650°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

122S°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Mils Lateral Exp.

16 16 18 36 34 39 S2 60 S9 67 71 67 83 86 80 RTNDT USE**

1S°F 98 ft-lbs

• 1150°F~ 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled to 600°F.

Based on transverse test data (Table 4A) 12 Revision 1 I

NLR-N93149 TABLE 4A SALEM UNJ:T 1

.MATERIALS CERTJ:Fl:CATJ:ON l:NFORMATJ:ON The following information was taken from Material Testing Report, M.T.L. No.

4396, Westinghouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yanichko, 4/17/70 and Material Testing Report, M.T.L. No. 3582, Westinghouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yanichko, 9/3/70.

component:

Intermediate Shell Plate, B2402-2 Temp. OF

-100

-100 50 50 50 10 10 10 60 60 60 110 110 210 210 210 550 550 550 Transverse Charpy Test Data Energy, ft-lbs

. % Shear 4

6 14 10 4

32 39 26 64 55 54 86 80 94 90 91 87 104. 5 -

102.5 9

9 14 14

.14 38 34 33 69 59 47 91 94 100 100 100 100 100 100 13 Revision 1 Heat No.:

C-1354'."'"2 Mils Lateral Exp

• 2 2

14

.11 10 29 38 24 52 47 47 71 60 79 78 73 65 71 73

NLR-N93149 TABLE 5 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8Sll, "PSE&G Co. Salem Unit No. 1 Reactor Vessel Radiation Surveillance Program, "November 197S.

Component:

Surveillance Material, Plate 82402-2 Heat No.:

C-13S4 Chemical Analysis c

Mn p

s Si Ni Mo Cu Al Cr Sn

.22 1.48

.016

.022

.32

.S4

.46

.23

.032

.18

.022 Longitudinal Charpy Impact and Fracture Tests Temp, "F Energy, ft-lbs

% Shear c

-100 12.S 9

-100 9.S s

-100 8

9

-so' 16.S 9

-so 31.S 14

-so 24.S 14

+10 41.S 34

+10 S2.S 34

+10 34.S 38

+60 64.S 43

+60' S4 43

+60 74 43

+110 108 79'

+110 101

'82

+110 107.S 84

+160 112 100

+160 120 100

+160 116 100 Temp, "F Drop Weights NDT

-30"F (based Performed bv CE on CE data)

• Heat *Treatment 1SS0-16SO"F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

122S"F +/- 2S"F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Furnace cooled.

llSO"F +/- 2S"F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled.

14 Revision 1 Mils Lateral Exp.

11 8

9 14 24 24 33 44 31 SS 4S S7 82 72 79 8S 81 80 RTNDT USE

-"SF 7S ft-lbs

NLR-N93149 TABLE 6 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on January 11, 1967.

Component:

Intermediate Shell Plate, B2402-3 Heat No.:

C-1397-2 Mill Chemical Analysis c

Mn p

s Si Ni Mo Cu

.20 1.22

.011

.025

.27

.so

.45

.22*

Per WCAP 10694 "Analysis of Capsule Y from the Public Service Electric and Gas Company Salem Unit 1 Reactor Vessel Radiation Surveillance Program,"

Table A-1, December 1984.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

• % Shear

-40 42 10

. -40 50 10

-40 60 15

+10 79 50

+10 84 50

+10 85 60

+60 100 65

+60 98 70

+60 107 70

+110 119 100

+110 121 100

+110 107 90

+160 125

. 100

+160 130 100

+160 127 100 Temp, OF Drop Weights NDT

-40 lF

-30 2NF

-20 lNF

-40°F 0

lNF Heat Treatment 1SS0-1650°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

1225°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1150°F., 40 h.ours.

Furnace cooled to 600°F.

Based on transerve test data (Table 6A) 15 Revision 1 RTNDT

-40°F Mils Lateral Exp.

37 41 52 66 65 65 59 72 77 90 83 86 81 93 90 USE**

104.3 ft-lbs

NLR-N93149 TABLE 6A SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from Material Testing Report, M.T.L. No.

4397, West*irighouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yanichko, 4/17/70 and Material Testing Report, M.T.L. No. 3582, Westinghouse Electric Corporation Atomic Power Divisions Materials Testing Laboratory, s. E. Yanichko, 9/3/70.

component:

Intermediate Shell Plate, B2402-3 Heat No.:

C-1397-2 Transverse Charpy Test Data Temp. OF Energy, ft-lbs

-100 8

-100 2

50 9

50.

15.5 50 20 10 56 10 61

. 10 37 60 72 60 77 60 55 110 106 110 99 210 105 210 104 210 104 550 79 550 111 550 107.5

% Shear 9

9 18 14 14 30 40 34 54 66 59 94 90 100 100 100 100 100 100 16 Revision 1 Mils Lateral Exp.

5 2

11 16 18 52 51 34 62 67 50 85 78 84 84 82 71 79 69

NLR-N93149 TABLE 7 SALEM UNJ:T 1 MATERIALS CERTJ:FJ:CATJ:ON :INFORMATION The following information was taken from WCAP-8Sll, "PSE&G Co. Salem Unit No. 1 Reactor Vessel Radiation Surveillance Program," November 197S.

Component:

surveillance Material, Plate B2402-3 Heat No.:

C-1397 Chemical Analysis c

Mn p

s Si Ni Mo Cu Al Cr Sn

.20 1.13

.012

.026

.27

.S2

.42

.22

.048

.12

.018 Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear

-100 s

-100 s.s

..,100 10.S

-so 32

-so 34

-so 19.S

+10 62

+10

72. s

+10

77. s

+60 104.S

+60 94 *. S

+60 87

+110 129

+110 130

+110 133

+160 128.'S

+160 137

+160 124.S Temp, OF Drop Weights NDT

-40°F Performed-by CE (Based on CE data)

• Heat Treatment 1SS0-1GS0°F! 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

water quenched.

122S°F +/- 2S°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Furnace cooled.

11S0°F +/- 2S°F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled.

17

• Revision 1 s s 9

14 20 18 4S 40 4S 66 61 S4 100 100 100 100 100 100 Mils Lateral Exp.

3 6

8 28 30 19 S2 60 64 78 78 71 86 92 9S 92 9S 86 RTNDT USE

-23°F 8S ft,.- lbs

NLR-N93149 TABLE 8 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from "Salem Units 1 and 2 Reactor Vessel Weld Data," CE Inc., Design Input File TOl.5-020, November 1985.

Component:

Welds 2-042A, 2-0428, and 2-042C Chemical Analysis c

Mn p

s Si Ni

.11 1.17

.016

.016

.20 1.00*

Estimated value.

Heat No.:

Flux:

Mo

.53 398196, and 348009 in tandem, in

• conjection w/Ni-200 wire

~Linde 1092, Lot No. 3692 Cu Cr

.18

.039 Charpy Impact and Fracture Tests Charpy tests riot performed for Heat No. 398196 and Heat No. 348009 in tandem.

Temp, OF*

Drop Weights No drop wt. test performed 1150°F for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

NOT Heat Treatment 18 Revision 1 RTNDT USE

-56°F (generic v~lue per 10 CFR 50.61)

NLR-N93149 TABLE 9 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion En9ineering, Inc. on April 21, 1967.

Component:

Lower Shell Plate, B2403-1 Heat No.:

C-1356-1 Mill Chemical Analysis c

Mn p

s Si Ni Mo Cu

.19 1.31

.011

.018

.25

.48

.47

.19*

Per PSE&G letter to the NRC of November 16, 1977, Docket No. 50-272, and Westinghouse Letter PSE-77-5 to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear

-40 10 0

-40 8

0

-40 10 0

+10 33 15

+10 29 15

+10 28 15

+40 71 35

+40 75 35

+40 66 30

+110 106 70

+110 102 70

+110 106 70

+160 147

. 100

+160 143 100

+160 138 100 Temp, OF Drop Weights NOT

-so lF

-40 lF

-30 2NF

'-40°F Heat Treatment 1550-1650°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

1220°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 1150°F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

~urnace cooled t~ 600°F~

19 Revision 1

.RNDT 4°F Mils Lateral Exp.

8 6

9 30 26 26 56 60 63 72 72 75 91 91 86 USE 92.S ft-lbs

NLR-N93149 TABLE 10 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report

• prepared by Combustion Engineering, Inc. on April 21, 1967.

Component:

Lower Shell Plate, B2403-2 Heat No.:

C-1356-2 Mill Chemical Analysis c

Mn p

s Si Ni.

Mo Cu

.20 1.34

.012

.018

.27

.49

.48

.19*

• . Per PSE&G letter to the NRC of November 16, 1977, Docket No. 50-272, and Westinghouse Letter PSE-77-5 to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear

-40 7

-40 9

-40 9

+10 24

+10 19

+10 40

+40 48

+40 41

+40 63

+110 110

+110 111

+110 95

+160 130

+160 130

+160 124 Temp, OF Drop Weights NDT

-70 lF

-GO 2NF

-so lN;F °F

'-30 lNF Heat Treatment 1550-16~0°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched 1220°F, 4.hours.

1150°F,. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled t6 600°F.

20 Revision 1 0

0 0

10 10 20 25 20 30 80 80 70 100 100 100 RTNDT

. 18°F Mils Lateral Exp.

6 8

9 22 16 33 40 35' 43 77 74 69 89 80

.86 USE 83 ft-lbs

NLR-N93149 TABLE 11 SALEM UNIT 1 MATERIALS CERTIFICATION :INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on April 21, 1967.

Component:

Lower Shell Plate, 82403-3 Heat No.:

C-1356-3 Mill Chemical Analysis c

Mn P.

s Si Ni Mo Cu

.21

1. 30

.010

.016

.25

.48*

.47

.19**

Per Test Certificate prepared by Lukens Steel Company on April 29, 1966.

• • Per PSE&G letter to the NRC of November 16, 1977, Docket No. 50-272, and Westinghouse Letter PSE-77-5 to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs

% Shear

-40 12

-40 6

-40 8

+10 45

+10 34

+10 36

+40 68

+40 73

+40 57

+110 105

+110 86

+110 109

+160 135

+160 124

+160 133 Temp, OF Drop.Weights NOT

-60 lF

-so lF, lNF

-40 lF

-40°F

-30 2NF Heat Treatment 1550-1650°F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Water quenched.

1220°F, *4 hours.

1150°F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

Furnace cooled to 600°F.

21 Revision 1 0

0 0

20 20 20 40 40 40 70 70 70 100 90 100 RTNDT 6°F Mils Lateral Exp.

8 4

7 36 27 29 53 57 47 74 70 77 89 76 73 USE 85 ft-lbs

NLR-N93149 TABLE 12 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from "Salem Units 1 and 2 Reactor Vessel Weld Data," CE Inc., Design Input File TOl.5-020, November 1985 and CE Welding Materials Qualification, June 8, 1967.

Component:

Welds 3-042A, 3-0428, and 3-042C Heat No.:

Chemical Analysis c

Mn p

s Si Ni Mo Cu

.016

.23 1.00*

.52

.19 Estimated value Charpy Impact and Fracture Tests Temp, °F 10 10 10 Energy, ft"".'lbs 84 71 90 A Full Charpy curve was not generated.

Temp, °F Drop Weights i

No drop wt. test performed 1150°F for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

NOT Heat Treatment 22 Revision*1 RTNDT.

-56°F (generic value per 10 CFR 50.61) 348009 w/Ni-200 wire Linde 1092, Lot No. 3708 Cr

.038 USE

NLR-N93149 TABLE 13 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from,;Salem Units 1 and 2 Reactor vessel Weld Data." CE Inc., Design input File TOl.5-020, November 1985 and CE Welding Materials Qualification, February 28, 1968.

Component:

Weld 9-042 Heat No.:

Flux:

Chemical Analysis c

Mn p

s Si Ni Mo

.26 1.33

.023

.014

.18

• 72

.44 Charpy Impact and Fracture Tests Temp, OF Energy, ft-lbs 10 85 10 77 10 81 A Full Charpy curve was not generated.

Temp, OF Drop Weights NDT RTNDT

-56°F 13253 Linde 1092, Lot No. 3791 Cu Cr

.25

.022 USE No drop wt. test performed (generic value 1150°F for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

Heat Treatment 23 Revision 1 per 10 CFR 50.61)

\\J

~LR-N93149 TABLE 14 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8511, "PSE&G Co. Salem Unit No. 1 Reactor Vessel Radiation Surveillance Program," Novmeber 1975.

Component:

Weld Metal Surveillance Material (Submerged arc weldment joining B2403-l and B2402-2)

Chemical Analysis c

Mn p

s Si Ni Mo

.08 1.14

.019

.016

.17 1.26

.53 Heat No.:

Flux:

Cu Al

.16

.01 Charpy Impact and Fracture Tests 39Bl96 Linde 1092 Lot No. 3617 Cr Sn

.04

.007 Temp, OF Energy, ft-lbs

% Shear Mils Lateral Exp.

-250

'3.

-250 4.5

-250 6.5

-200 19

-200 22.s

-200 8

-200 30

-200 31.5

-200 30

-150 38

-150 35.5

-150 43

-100 37

-100 40

-100 46

-so so.s

-so S2.S

-so 46

+10 80

+10 77.S

+10 70

+110 96.S

+110 116

+110 100 Temp, OF Drop Weights NOT Performed by CE 0°F Heat Treatment 11S0°F +/- 2S°F 14 hour1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />s~

Furnace cooled.

24 Revision 1 5*

2 5

2 9

7 9

17 9

20 5

8 18 23 14 29 14 27 20 32 20 30 27 33 30 32 36 36 30 38 40 4S 45 46 40 4S 77 67 77 67 71 S9 100 83 100 92 100 86

• RTNDT USE 00F 104.2 ft-lbs

0NLR-N9J149 TABLE 15 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8511, "PSE&G Co. Salem Unit No. 1 Reactor Vessel Radiation Surveillance Program," November 1975.

Component: Weld Heat Affected Zone Surveillance Material (machined from plate B2402-2 of a stress-relieved weldment joining plates B2403-1 and B2402-2)

Chemical Analysis Information not available (analysis not performed on HAZ)

Charpy Impact and Fracture Tests Temp, OF

.Energy, ft-lbs

-285 3

-285 7.5

-285 15

-250 6

-250 11

-250 11.5

-200 21

-200 18

-200 27.5

-150 52

-150 43

-150 44

-100 46.5

-100 51

-100 22

-50 45

-50 66

-50 48

+10 120.5

+10 61.5

+10 106.5

+no*

107.5

+110 106.5

+110 129 Temp,.* °F Drop Weights NOT Tests not performed QO]i*

Heat Treatment 1150°F +/- 25°F, 14 hour~.

Furnace ~coled.

25 Revision 1 Shear Mils 5

5 5

5 9

9 18 13 20 40 25 30 36 33 18 48 42 48 100 68 100 100 100 100 RTNDT QOF Lateral Exp.

1 4

14 2

7 8

11 11 18 36 24 31 30 37 16 36 44 34 78 55 67 77 85 76 USE.**

114.3 ft-lbs

• NLR-N93149 TABLE 16 SALEM UNIT 1 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8511, "PSE&G co. Salem Unit No. 1 Reactor Vessel Radiation surveillance Program," November 1975.

Component:

Correlation Monitor Material, A533 Grade B, Class 1 (HSST Plate 02)

Chemical Analysis Test c

• Mn p

s Si Ni Mo Cu Ladle 0.22 1.45 0.011 0.019 0.22 0.62 0.53 Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 Heat Treatment History - *correlation Monitor Material Material Temperature, correlation 1625 +/- 25 Monitor Plate -

1600 +/- 25 A533 Grade B, 1225 +/- 25 Class 1 1150 +/- 25 OF

Time, 26 Revision 1 4

4 4

40 hrs Coolant Air Cooled Water quenched Furnace cooled Furnace cooled at 600°F

~ **) w NLR-N93149 TABLE 17 SALEM UNJ:T 1 MEASURED VERSUS PREDICTED 30 FT-LB TEMPERATURE INCREASES AND UPPER SHELF ENERGY DECREASES ARTNDT,* °F (1)

Upper Shelt Energy (2)

Decrease, %

Material Capsule Fluence Measured Predicted Measured Predicted c1019 n/cm2\\

B24a2-l (long)

T a.24 laa 133 17.5 23.5*

B24a2-l (long) z 1.33 17a 2a9 16.5 35.a B24a2-2 (long)

T a.24 laa 13a 11.a 23.a B24a2-2 (long) z 1.33 165 2a3 12.a 34.5 B24a2-3 (long)

.* T a.24 75 125 a.a 22.a B24a2-3 (long) y a.891 11a 178 13.a 3a.a B24a2-3 (long)

• z 1.33 125 195 17.a 33.a Weld Metal y

a.891 165 26a 28.a 29.5 Correlation T

a.24 6a 97 6.5 16.5 Material Correlation y

a.891 125 133 16.5 22.a Material Correlation z

1.33 135 144 2a.5 24.5 Material (1)

Predicted ARTNDT includes 2cra as defined in Regulatory Guide 1.99, Rev. 2.

(2)

Predicted values based on Regulatory Guide 1.99, Rev~ 2 methodology.

27 Revision 1