ML20101L857

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Forwards Response to Rev 1 to Generic Ltr 92-01 Re Reactor Vessel Structural Integrity,Consisting of Table Listing Effects of 550 F Irradiation on Notch Toughness Properties of Reactor Vessel Surveillance Capsule Matl
ML20101L857
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 06/30/1992
From: Skolds J
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NUDOCS 9207070385
Download: ML20101L857 (15)
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Text

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- ; So th Carolina Electric & Gas Company J hn ko ds Jenkinsville. Sc 29065 Nuclear Operations

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(803) 3454040 SCE&G neam w ...

June 30, 1992 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 _

Gentlemen:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 RESPONSE TO GENERIC LETTER 92-01, REVISION 1, REACTOR VESSEL STRUCTURAL INTEGRITY (LTR 920001)

South Carolina Electric & Gas Company (SCE&G) submits Enclosure 1 in response to the information required by the subject Generic Letter.

I declare that the statements i?t forth herein are true and correct to the best of my knowledge, information, and belief.

If you have any questions regarding this subject, please contact Mr. Manuel W. Gutierrez at (803) 345-4392.

Very truly yours, John L. Skolds hWG: led Enclosure c: 0. W. Dixon General Managers R. R. Mahan NRC Resident Inspector R. J. White J. B. Knotts Jr.

S. D. Ebneter RTS (LTR 920001)

G. F. Wunder File (815.14)

K. O. Cozens, NUMARC d

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NUCLEAR EXCELLENCE - A SUMMER TRADITION!

9207070385 920630 PDR ADOCK 05000395 ON l

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Enclosure to Document Control Desk Letter LTR 920001 Pag 6 1 of 14

SUBJECT:

Response to Generic Letter 92-01, Reactor Vessel Structural Integrity RE0 VEST 1-

"Certain addressees are requested to provide the following information regarding Appendix H to 10 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, 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 H 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 F to 10 CFR Part 50 under 10 CFR 50.60(b)."

RESPONSE 1:

The V. C. Summer (VCS) Unit i reactor vessel irradiation surveillance program, Reference 1, is in compliance with Appendix H to 10 CFR Part 50 by its use of ASTM E-185-73 to establish its surveillance program. The VCS program is not part of any other integrated surveillance program. The post-irradiation mechanical testing of Charpy V-notch and tensile specimens is being performed in accordance with ASTM E-185-82. Application of these standards has been applied to three irradiated reactor vessel specimens.

Results frem these analyses are documented in References 2, 3, and 4. .

REQUEST 2A:

"Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 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 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."

(

.__ _ _ - - - _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ . - - _ _ _ . _ - _ _ - _ _ - _ _ - _ _ _ _ _ __-_______-___-___-_a

Enclosure to Document Control Desk Letter LTR 920001 Pags 2 of 14 RESPONSE 2A:

Table 1 summarizes the Charpy upper shelf energies (USE) ar the unirradiated and three irradicted reactor vessel specimens' analyses as reported in References 1, 2, 3, and 4. A maximum 10 ft-lb, or 7.6%, decrease in USE was determined for plate A9154-1 (longitudinal orientation). This data is plotted on the Reg. Guide 1.99, Revision 2, curve--see Figure 1--using the guidance set forth in paragraph C.2.2 of Reg. Guide 1.99, Revision 2. Based on actual data, the projected decrease in USE at the projected end of life (32 EFPY) fluence of 3.87x1019 n/cm2, Reference 4, is expected to be less than 10%. Using the lowest initial USE value of 75 ft-lb for the intermediate shell plate A9154-1, the resulting end of life USE value is projected to be 67.5 ft-lb. This is above the 50 ft-lb limit and, therefore, is acceptable.

Using only chemistry (.10 Wt-% Cu for Base Metal) data and fluence data from Figure 1, the same projected decrease in USE is expected to be approximately 27%. Applying this to the same intermediate shell plate, A9154-1, the resulting end of life USE value is expected to be 54.8 ft-lb. Again this is above the 50 ft-lb limit and is, therefore, acceptable. No further actions are required as outlined in Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.

1

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l l-i I Table 1 JGT exo eo f EFFECT OF 550 F IRRADIATION ON NOTCH TOljGHNESS PROPERTIES OF THE M

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l V. C. SUMMER UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE - - MATERIAL l

l Capsule Fluence Average 30 Ft-Lb Average 50 Ft-Lb. Average Energy Absorption At Full Shear (Ft-Lb) 8' Referen Material Temp. ( F) Temp. ( F) .... ......................... g

( 1019 N/CM2 n l

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0 25 75 2 Int. Plate A9154-1 unirr. 55 105 3 $

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(Transverse) 1.47 65 73 4

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-70 130 0 -90 121 2 HAZ Metal unirr. -55 -30 3 U .639 -15 111 1.47 -45 117 4 V -25 2.46 -45 X

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Enclosure to Document Control Desk Letter LTR 920001 Page 4 of 14

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i Enclosura in Document Control Desk Letter '

LTR 9200'J1 Page 5 of 14 RE0 VEST 28:

"Certain addressees are requested to provide the following informa' ion regarding Appendix G to 10 CFR Part 50:

Addressees whose reactor vessels were constructed to an ASME Code earlier tnan 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 Ill.A of 10 CFR Part 50, Appendix G:

(1) the results from all Charpy and drop weight tests for all un'craiiated beltline materials, the unirradiated reference ten.perature for each beltline material, and the 09thod of determining the unirradiated reference tempers ' ce f rom the Charpy and drop weight test; (2) the heat treatment received by all beltline and surveillance materials; (3) the heat number for each beltline piate 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 -

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

RESPONSE 28:

The V. C. Summer Unit I reactor pressure vessel was constructed to the 1971 Edition of the ASME Code with no addenda; therefore, response to this request is applicable. For the Reactor Vessel Radiation Surveillance Program, Chicago Bridge and Iron Company supplied Westinghouse with Sections of SA533 Grade B Class 1 plate used in the core region of the V. C. Summer Unit No. 1 reactor pressure vessel, specifically from the 8-inch intermediate shell plate A9154-1. Also supplied was a weldment made from sections of plate A9154-1 and adjoining lower shell plate C9923-2, using RAC0 1NMM weld wire, heat number 4P4784, and Linde 124 flux, lot number 3930. This weld wire and flux combination is identical to that used in forming the longitudinal weld seams in the intermediate and lower shell courses and the girth weld joining these two shell courses. The following is a summary of the beltline materials used to fabricate the V. C. Summer Unit 1 reactor pressure vessel:

- - _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ -____________b

Enclosure to Document Control Desk Letter LTR 920001 Page 6 of 14 DESCRIPTION: Reactor Vessel Beltline Plate (Intermediate Shell)

HEAT NO.: A9154-1 SPECIFICATION NO.: SA533 Grade B Class 1 SUPPLIER: Lukens Steel Company HEAT TREATMENT: Austenitized at 1625*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Water Quenched lempered at 1280*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Air Cooled Stress Relieved at 1050"F for 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> Furnace Cooled / Air Cooled COMPOSITION: .10 %Cu, .51 %Ni, .009 %P, .015 %S DROP WEIGHT THDT: -20"F T RTNDT: 30*F (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO. 1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE A9154-1 (LONGITUDINAL ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

("F) (FT-LB)

+212 121, 144, 144

+70 80, 107, 112

+40 70, 88, 70

+10 54, 50, 48

-20 17, 10, 15

-100 5, 3, 4 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

("F) (FT-LB)

+212 82.5, 76.5: 83

+120 56, 64, 74

+90 51.5, 60.5, 58

+70 56, 36, 40

+40 36, 41, 38

-20 12.5, 20, 19 Additional Charpy V_ notch impact testing was performed on material from this intermediate shell plate, the weld metal, and heat affec: ' zone metal to establish full Charpy impact energy transition curves. Results of this testing are documented in Reference 1.

Enclosure to Document Control Desk Letter LTR 920001 Page 7 of 14 DESCRIPTION: Reactor Vessel Beltline Plate (Intermediate Shell)

HEAT NO.: -A9153 SPECIFICATION NO.: SA533 Grade B Class 1 SUPPLIER: _Lukens Steel Company HEAT TREATMENT: Austenitized at 1625*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Water Quenched Tempered at 1280*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Air Cooled.

Stress Relieved at 1050*F for 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> Furnace Cooled / Air Cooled COMPOSITION: .09 %Cu, .45 %Ni, .006 %P, .016 %S DROP WEIGHT TNDT: -20*F RTNDT: -20*F (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO. 1 REACTOR PRESSURE. VESSEL INTERMEDIATE SHELL PLATE A9153-2 (LONGITUDINALORIENTATION)

' TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+212 142, 146, 136

+70 108, 107, 105

+40 82, 83, 82

+10 68, 56, 74

_20 54, 58, 54

-100 9, 3, 8 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-; M 2h ibk ibb[ii .5

+120 93, 90, 94

+70 65, 71, 68

+40 59, 51, 50

+20 43, 42.5, 47

-20 14, 26, 18

Enclosure to Document Control Desk Letter LTR 920001 Page 8 of 14 DESCRIPTION: Reactor Vessel Beltline Plate (Lower Shell)

HEAT N0.: C9923-2 SPECIFICATION NO.: SA533 Grade B Class 1 SUPPLIER: Lukens Steel Company HEAT TREATMENT: Austenitized at 1600*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Water Quenched Tempered at 1260*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Air Cooled Stress Relieved at 1075*F for 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> Furnace Cooled / Air Cooled COMPOSITION: .08 %Cu, .41 %Ni, .005 %P, .015 %S DROP WEIGHT TNDT: -10*F RTNOT: 10*F (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT N0. 1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE C9923-2 (LONGITUDINALORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+212 164, 164, 155

+40 86, 76, 85

+10 56, 56, 55

-20 42, 50, 35

-50 8, 6, 12

-100 4, 5, 5 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+212 93, 94, 88

+120 82, 79, 84

+70 55, 50, 51

+50 40, 46, 53

+40 56, 42.5, 37.5 l 10 23, 31, 33.5 l

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Enclosure to Document Control Desk Letter <

LTR 920001 P6g'e 9 of 14 DESCRIPTION: Reactor Vessel Beltline Plate (Lower Shell)

HEAT NO.:- C9923_1 SPECIFICATION N0.: SA533 Grade B Class 1 SUPPLIER: Lukens Steel Company HEAT TREATMENT: Austenitized at 1600*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Water Quenched Tempered at 1260*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Air Cooled Stress Relieved at 1075*F for 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> Furnace Cooled / Air Cooled COMPOSITION: .08 %Cu, 41 %Ni, .005 %P, .014 %S DROP WEIGHT TNDT: -30*F RTHDT: 10*F (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO. 1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE C9923-1 (LONGITUDINALORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-L8)

~ 252 ik8,ik7, ISO

+40 92, 100, 89

+10 92, 70, 82

-50 52, 60, 44 l

-75 35, 17, 12

-100 7, 10, 9 l (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+212 104, 114, 100 j- +120 74.5, 81, 80

+70 68, 51, 51

+50 47, 44, 43.5

+30 37, 42.5, 35

-30 11, 9, 12 l

I

e Enclosure to Document Control Desk Letter LTR 920001 Pag' e 10 of 14 DESCRIPTION: Reactor Vessel Core Region Beltline Weld Metal WIRE HEAT NO.:' 4P4784 WIRE TYPE: RAC0 1NMM FLUX TYPE: Linde 124 FLUX LOT: 3930 FABRICATOR: Chicago Bridge and Iron Co.

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 /> WIRE COMPOSITION: .05 %Cu, .91 %Ni, .013 %P, .012 %S DROP WEIGHT TNDT: -50*F RTNDT: -44*F (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO. 1 REACTOR PRESSURE VESSEL CORE REGION BELTLINE WELD METAL (Reference 1)

TEST TEMPERATURE IMPACT ENERGY

("F) (FT-LB)

+210 91.5, 94, 93.5

+125 85.5, 89

+80 80, 84

+50 84, 71

+10 58, 64.5, 62

-25 30, 51 50 32, 34

-100 14.5, 15 4

Enclosure to Document Control Desk Letter LTR 920001

, Page 11 of 14 REQUEST 3A:

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

How the embrittlement ef#ects 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 3A:

This issue is not applicable to V. C. Summer Unit 1 since Technical Specifications do not allow the plant to be critical below 541*F. At no time has the plant been critical below 525*F.

REQUEST 38:

" Addressees are_ requested to provide the following information regarding commitments made to respond to GL 88-11:

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

RESPONSE 33:

In response to Generic Letter 88-11, Reference 5, measured results from two irradiated capsule analyses were used. The impact on using measured data is that the current P/T ' limits are bounded by the previous analysis and are still valid through 8 EFPY, Surveillance data was not used for the RTPTS determination. The RTPTS values for plates and welds at end of life (32 EFPY) were determined to be less than the 160*F PTS screening criteria. No actions were required as a result of V. C. Summer analyses that supported its Generic Letter 88-11 response.

RE0 VEST 3C:

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

If a measured increase in reference temperature exceeds the mean-plus-two standard deviations prediction by Regulation 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 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."

Enclosure to Document Control D'esk Letter l LTR 920001  :

Page 12 of 14 l

RESPONSE 3C:

No data from the V. C. Summer Unit 1 surveillance program has exceeded the mean-plus-two standard deviation bounds nor the upper shelf energy decrease as predicted by Reg. Guide 1.99, Revision 2. Table 2 summarizes the results from the V. C. Summer surveillance program.

i

urm-TABLE 2 $* $8 m&

COMPARISCN OF V. C. SlH1ER UNIT 1 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE StilFTS

~ U$$

AND UPPER SifELF ENEDeY DECREASES WITil REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS S

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8 8

30 ft-lb Transition Temo. Shift Upper Shelf Enerav Decrense g R.G. 1.99 Rev. 2 R.G. 1.99 Rev. 2 [

Fluence (Predicted) Measured (Predicted) Measured $

Material Capsule 10 I9 n/cm 2 (*F) (*F) (%) (%) S

?

Int. Plate A9154-1 0 0.639 57 40 17 1 E

(Longitudinal) V 1.47 72 60 21 8 g X 2.46 81 50 23 5  ;;

Int. Plate A9154-1 U 0.639 57 30 17 0 (Transve se) V 1.47 77 40 21 0 X 2.46 81 35 23 3 Weld Metal 0 0.639 59 30 18 4 V 1.47 75 45 22 7 X 2.46 102 35 25 7 IIAZ Metal U 0.639 --

30 --

7 V 1.47 --

45 --

15 X 2.46 --

45 --

10

Enclosure to Document Control Desk Letter LTR 920001

. Pag'e 14 of 14

REFERENCES:

1. Yanichko and Davidson, South Carloina Electric and Gas Company Viroil C.

Summer Nuclear Plant Unit No. 1. Reactor Vessel Radiation Surveillance Procram, WCAP-9234, January 1978.

2. R. S. Boggs, et. al., .alysis of Capsule U from the South-Carolina Electric and Gas Company Viroil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Procram, WCAP-10814 June 1985.
3. D. J. Colburn, et. al., Analysis of Capsule V from the South Carolina Electric and Gas Company Virail C. Summer Unit 1 Reactor Vessel Radiation Surveillance Procram, WCAP-11726, January 1988.
4. J. M. Chicots, et. al., Analysis of Capsule X from the South Carolina Electric and Gas Company Virail C. Summer Unit 1 Reactor Vessel Radiation Surveillance Procram, WCAP-12867, March 1991.
5. Letter from 0. S. Bradham (SCE&G) to USNRC Document Control Desk, "V. C.

Summer Nuclear Station Response to Generic Letter 88-11," January 12, 1989.

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