ML18096A806

Rev 0 to Salem Unit 2 Response to Generic Ltr 92-01,Rev 1, 'Reactor Vessel Structural Integrity.'

ML18096A806
Person / Time
Site:	Salem
Issue date:	06/17/1992
From:	Public Service Enterprise Group
To:
Shared Package
ML17095A621	List: ML17095A620 ML18096A805 ML18096A806
References
GL-92-01, GL-92-1, NUDOCS 9207070035
Download: ML18096A806 (24)

Text

NLR-N92081 ATTACHMENT 2 SALEM UNIT 2 RESPONSE TO GENERIC LETTER 92-01, REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY JUNE 17, 1992 REVISION 0

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9207070035 920630 .

PDR ADOCK 05000272 p PDR

NLR-N92081 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 boldface 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 so:

Addressees who do not have a surveillance program meeting ASTM E 18S-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 so.

Addressees who plan to revise the surveillance program to meet Appendix II to 10 CFR Part so 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 so, 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-185-73 was the standard in place at the time the surveillance program was designed. The Salem Unit 2 surveillance program complies with ASTM E-185-73. 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 2 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 Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the 1

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• Response:

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 so.

Table 1 contains the unirradiated, December 16, 1991 and EOL (April 18, 2020) Charpy upper shelf energy for Salem Unit 2 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 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 forging and the heat number of wire and flux lot number used to fabricate the surveillance weld; (5) The chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillance material; and 2

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• Response:

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

The Salem Unit 2 reactor vessel was constructed to the 1965 Edition, through Winter 1966 Addenda to Section III of the ASME Code. Thus, the Salem Unit 2 reactor vessel was constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition. Tables 2 through 14 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 material test requirements and acceptance standards that were current at the time of the reactor pressure vessel construction.

(Note that the chemical composition of the welds was determined from the weld wire heat numbers of the actual welds, except for weld 9-442, in which the nickel content was estimated to be the upper limit of type MIL B-4 wire heats).

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 for each beltline region material by dropweight tests (ASTM E-208) performed by Combustion Engineering except for welds 3-442A, 3-442B, 3-443C and 9-442.

The unirradiated reference temperature (RTNDT) of the beltline region materials 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 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 10CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock."

The NDTT temperature, as determined by drop weight tests (ASTM E-208) is the RTNDT if, at 60°F above the NDTT, 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 lateral expansion are obtained on transverse Charpy specimens, minus 60°F.

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• 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.

If transverse Charpy V-notch specimens were 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-lbs and 35 mil lateral expansion temperatures for longitudinal specimens by 20°F.

If measured values of RTNDT 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 CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."

The Charpy V-notch data for each of the beltline region plates tested by Combustion Engineering were taken in the longitudinal direction. The RTNDT values for each of these materials were determined to be the higher of the (1) value obtained by increasing the temperature at which 50 ft-lbs and 35 mils lateral expansion were obtained for longitudinal specimens by 20°F (to estimate the temperature in the transverse direction for which 50 ft-lbs and 35 mils lateral expansion would have been obtained),

reduced by 60°F or (2) the NOTT. The RTNDT values for the intermediate shell longitudinal weld seams 2-442A, B and C and the surveillance weld heat affected zone material were determined to be equal to their NOTT values. Full Charpy curves were not tested for lower shell longitudinal welds and the intermediate shell to lower shell circumferential weld seams; therefore, the generic mean value of -56°F is assumed. The transverse Charpy test data was used to determine the.RTNDT for the surveillance test plate.

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

The upper shelf 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. The upper shelf energy values for the beltline region plates were calculated by multiplying the average of the 100% shear longitudinal Charpy V-notch data by 65%. The upper shelf energy value for the surveillance test plate was determined by taking the average of the three 100% shear energy values for the transverse data obtained in tests conducted by Westinghouse Electric Corporation.

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• The upper shelf energy values for the intermediate shell longitudinal weld and surveillance weld materials were determined by the average of the three 100% shear energy values. Upper shelf energy values were not calculated for the lower shell longitudinal welds and intermediate to lower shell circumferential 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 2 reactor pressure vessel for use in the Reactor Vessel Radiation Surveillance Program. The sections of material were removed from the 9 5/8-inch intermediate shell plate B4712-2 of the pressure vessel.

Combustion Engineering, Inc., also supplied a weldment made from sections of the intermediate shell plate B4712-2 and adjoining intermediate shell plate B4712-l using weld wire representative of that used in the original fabrication. The heat treatment histories of the pressure vessel beltline region material and surveillance materials are given in Tables 2 through 14.

3. Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:

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

In 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.

Response

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 3.1.1.4 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 ~dditional 15 minutes.

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• 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 complieq 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 86 for the Salem Unit 2 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 each beltline material as predicted for December 16, 1991, and for the end of its current license.

Response

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 15. The measured decrease in Charpy upper shelf energy exceeds the value predicted using methodology specified in Regulatory Guide 1.99 Revision 2 for the weld metal as indicated in Table 15~ Therefore the adjusted reference temperature and Charpy upper shelf energy of each beltline material as predicted by surveillance results for December 16, 1991.and end of current license are provided in Tables 16 and 17 .

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• TABLE 1 SALEM UNIT 2 UNIRRADIATED AND CALCULATED UPPER SHELF ENERGY (USE) VALUES USE, ft-lbs USE, ft-lbs< 1 > EOL USE ft-lbsC2)

Material Descri2tion Un irradiated December 16, 1991 A2ril 18, 2020 Intermediate Shell 90. o< 3 > 76.5 71.1 Plate B4712-1 Intermediate Shell 83. o< 3 > 69.7 64.7 Plate B4712-2 Intermediate Shell 75. 0(3) 64.5 60.8 Plate B4712-3 Intermediate Shell Long. lll.O 86.6 76.6 Weld 2-442A Intermediate Shell Long. 111.0 83.3 73.3 Weld 2-442B Intermediate Shell Long. 111.0 83.3 73.3 Weld 2-442C Lower Shell Plate 82.5< 3 > 70.1 66.0 B4713-1 Lower Shell Plate 88.0< 3 > 74.8 70.4 B4713-2 Lower Shell Plate 88. oC3) 74.8 70.4 B4713-3 Lower Shell Long. NA NA NA Weld 3-442A Lower Shell Long. NA NA NA Weld 3-442B Lower Shell Long. NA NA NA Weld 3-422C Intermediate to Lower NA NA NA Shell Girth Weld 9-442 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 calculated.

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

(2) EOL USE values calculated at 1/4T location, based on fluences from PSE&G letter SCI-92-0319, 5/14/92, J. Perrin to J. Chicots.

(3) Unirradiated USE values estimated from longitudinal data.

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• TABLE 2 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on November 12, 1968. (Corrected copy dated April 23, 1970).

Component: Intermediate Shell Plate 84712-1 Heat No.: C-4173-1 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.21 1.33 .012 .016 .22 .56 .54 .13*

• Per Salem 2 Table 6 of Westinghouse letter PSE-77-5 to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-80 8 0 4

-so 6 0 2

-40 20 10 16

-40 22 10 18

-40 21 10 17

+10 51 25 38

+10 52 25 39

+10 57 25 43

+40 69 35 51

+40 79 35 59

+40 75 35 56

+110 105 85 74

+110 111 85 82

+160 140 100 88

+160 136 100 90 Temp., OF Drop Weiqhts NDT RTNDT USE

+10 2NF QOF QOF 90 ft-lbs 0 lF 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 *

• 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 />. Furnace cooled to 600°F.

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• TABLE 3 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on October 3, 1968. (Corrected copy dated April 23, 1970).

Components: Intermediate Shell Plate B4712-2 Heat No.: C-4186-2 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.22 1.37 .011 .OlS .24 .60 .SS .14*

• Per Salem 2 Table 6 of Westinghouse letter PSE-77-S to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-40 7 0 8

-40 12 0 9

-40 10 0 8

+10 21 10 20

+10 37 lS 31

+10 30 lS 26

+40 62 30 46

+40 49 2S 39

+40 SS 30 42

+110 96 70 69

+110 98 80 70

+110 90 70 67

+160 124 100 86

+160 134 100 92

+160 12S 100 88 Tern

• I OF Dre hts NDT RTNDT USE o lNF

-10 2NF -20F 1°F 83 ft-lb

-20 lF Heat Treatment lSSO - 16S0°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 />.

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 to 600°F.

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• TABLE 4 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from WCAP-8824, "PSE&G Co. Salem Unit No.

2 Reactor Vessel Radiation Surveillance Program," January 1977.

Component: B4712-2 Surveillance Material Heat No.: C-4186-2 Chemical Analysis c Mn p s Si Ni Mo Cu

.23 1.34 .OlS .010 .30 .61 .SS .10 Transverse Charpy Impact and Fracture Tests Temp., OF Enerov, ft-lbs  % Shear Mils Lateral Exp.

-40 12 10 s

-40 8 lS 3

-40 12 10 7 0 34 2S 26 0 34 2S 27 0 24 18 18

+40 39 3S 28

+40 36 34 28

+40 43 34 34 RT S8 S2 46 RT S8 S2 47 RT 61 S2 49

+110 64 62 Sl

+110 74 68 61

+110 70 68 S7

+210 100 100 80

+210 87 100 71

+210 104 100 7S Temp., OF I Drop Weights NDT RTNDT USE Performed by CE -20°F 12°F 97 ft-lbs Heat Treatment lSSO - 16S0°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 />. Air cooled.

11S0°F +/- 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 S SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering. Inc., on October 4, 1968. (Corrected copy dated April 23, 1970).

Component: Intermediate Shell Plate B4712-3 Heat No.: C-4194-2 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.22 1.37 .010 .016 .26 .S7 .S2 .11*

• Per Salem 2 Table 6 of Westinghouse letter PSE-77-S to PSE&G of October 10, 1977.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Energy, ft-lbs  % Shear Mils Lateral Exp.

-40 12 0 10

-40 11 0 9

-40 9 0 7

+10 33 20 26

+10 34 20 24

+10 28 20 23

+40 49 30 49

+40 so 30 49

+40 38 2S 32

+110 100 8S 7S

+110 91 80 68

+110 79 7S 60

+160 116 100 82

+160 122 100 84

+160 110 100 80 Temp., OF Drop Weiqhts NDT RTNDT USE 0 lNF

-20 lNF

-40 2NF

-so lF -S0°F 22°F 7S ft-lbs

-60 lF Heat Treatment lSSO - 16S0°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 />.

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 to 600°F.

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• TABLE 6 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on March 15, 1971 and "Salem Units 1 and 2 Reactor Vessel Weld Data," CE Inc., Design Input File TOl.5-020, November 1985.

Component: Welds 2-442A, 2-442B, and 2-442C Heat No.: 13253 & 20291 (tandem)

Flux: Linde 1092, Lot No. 3833 Chemical Analysis c Mn p s Si Ni Mo Cu Cr

.12 1.29

• 019

• 012 . .19 .73 .48 .23 .028 Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-80 19 0 16

-80 26 10 21

-40 54 25 40

-40 39 20 31

+10 83 60 65

+10 94 70 72

+10 85 60 69

+40 83 70 63

+40 92 80 70

+40 97 80 76

+110 105 100 80

+110 108 100 84

+110 119 100 89

+160 113 100 87

+160 115 100 86 Temp. OF Drop Weights NDT RTNDT USE

-40 lF

-30 2NF

-20 lNF -40°F -40°F 111 ft-lbs 0 lNF Heat Treatment 1150°F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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• TABLE 7 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering. Inc. on August 19, 1969.

Component: Lower.Shell Plate B4713-1 Heat No.: C-4182-1 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.22 1.32 .010 .015 .22 .60 .54 .12*

• Per WCAP 10492, "Analysis of Capsule T from the Salem Unit 2 Reactor Vessel Radiation Surveillance Program," March 1984.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-80 9 0 6

-80 13 0 12

-40 25 10 21

-40 44 20 34

-40 16 5 12

+10 36 10 28

+10 65 30 46

+10 61 30 46

+40 60 30 47

+40 68 40 49

+40 64 35 48

+110 93 80 73

+110 108 85 75

+160 131 100 88

+160 123 100 88 Temp., OF Drop Weights NDT RTNDT USE

~

0 2NF

-10 lF -10°F -10°F 82.5 ft-lbs

-20 lF 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.

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 />. Furnace cooled to 600°F.

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• TABLE 8 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on October 28, 1969.

Component: Lower Shell Plate B4713-2 Heat No.: C-4182-2 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.22 1.28 .010 .015 .24 .57 .53 .12*

• Per WCAP 10492, "Analysis of Capsule T from the Salem Unit 2 Reactor Vessel Radiation Surveillance Program," March 1984.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Enerav, ft-lbs  % Shear Mils Lateral Exp.

-80 6 0 3

-80 13 0 7

-40 43 20 31

-40 13 5 12

-40 22 10. 16

+10 56 25 38

+10 42 20 31

+10 58 25 40

+40 79 35 58

+40 72 30 52

+40 89 40 63

+110 101 70 70

+110 112 80 78

+160 136 100 74

+160 135 100 76 Temp., OF Drop Weiqhts NDT RTNDT USE

-20 lF

-10 2NF 0 lNF -20°F -20 88 ft-lbs Heat Treatment 1550°F - 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 +/- 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 />. Furnace cooled to 600°F.

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NLR-N92081 TABLE 9 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on August 8, 1969.

Component: Lower Shell Plate, B4713-3 Heat No: B-8343-1 Mill Chemical Analysis c Mn p s Si Ni Mo Cu

.25 1.34 .012 .014 .28 .58 .53 .12*

• Per WCAP 10492, "Analysis of Capsule T from the Salem Unit 2 Reactor Vessel Radiation Surveillance Program," March 1984.

Longitudinal Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-40 19 0 10

-40 14 0 12

-40 12 0 10

+10 36 25 27

+10 52 30 37

+10 40 25 30

+40 63 35 46

+40 74 40 53

+40 50 30 38

+110 104 85 80

+110 108 90 82

+110 103 80 77

+160 135 100 85

+160 135 100 88

+160 136 100 91 Temp., OF Drop Weiqhts NDT RTDT USE

-20 lF

-10 lF 0 2NF -10°F .QOF 88 ft-lbs 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.

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 />. Furnace cooled to 600°F.

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• TABLE 10 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from "Salem Units 1 and 2 Reactor Vessel Weld Data," CE Inc., Design Input File TOl.S-020, November 198S and CE Welding Material Qualification, October 27, 1969.

Component: Weld 3-442A, 3-442B, and 3-442C Heat No.: 2193S and 12008 (tandem)

Flux: Linde 1092, Lot No. 3889 Chemical Analysis c Mn p s Si Ni Mo Cu

.11 1.38 .OlS .011 .lS .86 .SS .20 Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs 10 97 10 90 10 83 Full Charpy curve not performed.

Temp., OF I Drop Weights NDT RTNDT USE No drop wt. test performed. --- -S6°F (generic ---

value per 10CFR S0.61)

Heat Treatment 11S0°F for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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NLR-N92081

• TABLE 11 SALEM UNIT 2 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 Material Qualification Report, October 7, 1970.

Component: Weld 9-442 Heat No.: 90099 Flux: Linde 0091, Lot No. 3977 Chemical Analysis c Mn p s Si Ni Mo Cu

.14 1.20 .021 .013 .22 .20* .so .175

• Estimated value.

Charpy Impact and Fracture Tests Temp., OF Energy, ft-lbs 10 56 10 30 10 52 Full Charpy curve not performed.

Temp., OF I Drop Weiqhts NDT RTNDT USE No drop wt. test performed. --- -56°F ---

(generic value per 10CFR 50.61)

Heat Treatment 1150°F for 10.5 hrs.

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NLR-N92081 TABLE 12 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from "Salem*Units 1 and 2 Reactor Vessel Weld Data" CE Inc., Design Input File TOl.S-020, November 198S, and WCAP-8824, "PSE&G Co. Salem Unit No. 2 Reactor Vessel Radiation Surveillance Program,"

January 1979.

Component: Weld Surveillance Material Heat No.: 132S3 Weldment made from intermediate shell plates B4712-1 and Flux: Linde 1092, Lot No. 3833 B4712-2 and Linde 1092, Lot No. 3774 Chemical Analysis c Mn p s Si Ni Mo Cu Cr

.10 1.27 .017 .011 .29

• 71 .4S .23 .015 Charpy Impact and Fracture Tests Temp., OF Energy, ft-lbs  % Shear Mils Lateral Exp.

-100 4.S 2 1

-100 11 6 2.S

-100 6 lS 1

-so 11 18 7.S

-so 4 20 11

-so 3S.5 29 27 0 48.S S2 42 0 72 S2 36 0 63.S 62 so 40 71 59 S6 40 SO.S 5S 44 40 80.S 79 62 100 86 100 81 100 96.S 90 74.S 100 106.S 100 80 210 112 98 82 210 111. s 100 86 210 111. s 100 85.S Temp., OF I Drop Weights NDT RTNDT USE Performed by CE -40°F -20°F 111 ft-lbs Heat Treatment 11S0°F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 18 Revision o

• Ill ,,,

NLR-N92081 TABLE 13 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. March lS, 1971.

Component: Weld Heat Affected Zone (Seam No. 2-442)

Chemical Analysis Information not available (analyses were not performed)

Charpy Impact and Fracture Tests Temp., OF Enerqy, ft-lbs  % Shear Mils Lateral Exp.

-110 lS 0 12

-110 26 10 19

-80 41 20 31

-80 46 20 36

-40 4S 30 38

-40 SS 40 46

-40 47 30 39

+10 68 70 60

+10 87 80 76

+10 83 80 71

+40 lOS 9S 86

+40 100 90 83

+40 104 9S 80

+110 8S 100 84

+110 104 100 98 Temp., OF Drop Weiqhts NOT RTNDT USE 0 lF

+10 2NF 00F 00F 94.S ft-lbs Heat Treatment 19 Revision o

' "" ot_,

NLR-N92081 TABLE 14 SALEM UNIT 2 MATERIALS CERTIFICATION INFORMATION The following information was taken from the WCAP-8824, "PSE&G Co. Salem Unit No. 2 Reactor Vessel Radiation Surveillance Program January 1977."

Component: Weld Heat Affected Zone Surveillance Material (HAZ material obtained from B4712-2 of weldment made from intermediate shell plate B4712-1 and B4712-2)

Chemical Analysis Information not available (analyses were not performed on HAZ).

Charpy Impact and Fracture Tests Temp., OF Energy, ft-lbs  % Shear Mils Lateral Exp.

-125 51 33 24.5

-125 23 6 15

-125 28 12 10.5

-75 122 100 63

-75 36 17 17

-75 66.5 42 34

-25 79 66 48

-25 93.5 48 47

-25 76 51 37 25 131 96 69 25 111 88 72 25 74 60 45 100 101 100 66.5 100 114 100 70 100 179 100 78.5 210 118.5 100 79.5 210 115 100 72.5 210 106.5 100 75 -

Tern * , °F Dro hts NDT RTNDT USE Performed b CE 113 ft-lbs Heat Treatment 1150°F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

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- NLR..-N92081 TABLE 15 SALEM UNIT 2 MEASURED VERSUS PREDICTED 30 FT-LB TEMPERATURE INCREASES AND UPPER SHELF ENERGY DECREASES A RTNDT ( ° F) ( 1, 2 ) Upper Shelf Energy (1)

Decrease  %

Material Capsule Fluence Measured Predicted Measured Predicted 1019 n cm2)

B4 712-2 (long. ) T 0.276 50 99 6 14 B4712-2 (long.) u 0.57 70 118 8 17 B4 712-2 (long.) x 1.16 80 138 1 20 B4712-2 T 0.276 70 99 8 14 (transverse)

B4712-2 u 0.57 95 118 13 17 (transverse)

B4712-2 x 1.16 125 138 8 20 (transverse) ld Metal T 0.276 155 180 29(3) 27.5 Metal u 0.57 190 217 33(3) 32 x 1.16 195 255 22 38 (1) Predicted values based on Regulatory Guide 1.99, Revision 2 Methodology.

( 2) Predicted .6 RTNDT includes 2 6.6. as defined in Regulatory Guide 1. 99, Revision 2.

(3) Exceeds predicted value.

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t

' ... ct'.J ..

NLR-N92081

• TABLE 16 SALEM UNIT 2 ADJUSTED REFERENCE TEMPERATURES CART) AS PREDICTED BY SURVEILLENCE DATA ART (°F) ART (OF) screening Material Description December 16, 1991 April 18, 2020 Criterion 84712-1 96.3( 1 ) (96.3)< 2 > 131.9 (131.9) 270 84712-2* 85.4 (96.4) 123.8 (134.8) 270 84712-3 107.1 (95.l) 136.3 (124.3) 270 2-442A* 105.7 190.9 270 2-442B* 132.5 215.4 270 2-442C* 132.5 215.4 270 Weld Surveillance (152. 5) (235.4) 270 Material 84713-1 81. 7 (99.7) 115.3 (133.3) 270 84713-2 71.3 (99.3) 104.6 (132.6) 270 84713-3 91.4 (101.4) 124.9 (134.9) 270 3-442A 134.5 216.8 270 3-4428 108.5 190.0 270 3-442C 134.5 216.8 270 9-442 75.8 113.6 300

• ART based on surveillance capsule chemistry factor.

(1) Beltline plate, weld and HAZ material nos. are not in ().

(2) Surveillance material nos. are in ().

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I ...~

NLR-N92081 TABLE 17 SALEM UNIT 2 UPPER SHELF ENERGY CUSE) AS PREDICTED BY SURVEILLANCE DATA USE (ft-lb) EOL USE (ft-lb)

Material Description December 16, 1991 April 18, 2020 84712-1 76.5Cll ( 89. 3) (Z) 71.1 (83.0) 84712-2 (long.)* (114. 7) (111.0) 84712-2 (trans.)* 74.7 (87.3) 71.4 (83.4) 84712-3 64.5 (92.0) 60.8 (86.7) 2-442A* 86.6 75.5 2-4428* 83.3 72 .2 2-442C* 83.3 72.2 Weld surveillance (83.3) (72.2)

Material*

84713-1 70.1 (83.3) 66.0 (78.4) 84713-2 74.8 (87.6) 70.4 (82.4) 84713-3 74.8 (103.7) 70.4 (97.6) 3-442A, B and c NAC 3 l NA 9-442 NA NA Weld HAZ* 72.8 (87.0) 63.3 (75.7)

• USE based on surveillance data trending on Figure 2 of Reg. Guide 1.99 Revision 2. All other USE predictions based on Cu content, using Figure 2 of Reg. Guide 1.99 Revision 2.

(1) 8eltline plate, weld and HAZ material nos. not in ().

(2) Nos. in () are surveillance material.

(3) NA - Unirradiated upper shelf energy not available.

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