ML20101M504

Forwards Rev 1 to License NPF-3,requesting Info to Enable NRC to Assess Degree of Compliance W/Regulatory Requirements Re Reactor Vessel Integrity,In Response to Generic Ltr 92-01

ML20101M504
Person / Time
Site:	Davis Besse
Issue date:	07/01/1992
From:	Shelton D CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NUDOCS 9207080262
Download: ML20101M504 (32)

Text

, . -, _ . _

3

" GNTERROR 14t$ - -

EHNMEANGY -!

I Doneki C. SheNon 300 Madison Avenue l Vce President Nuclear Toledo OH 436520001

. Os 4 Besse (419)249 2300 Docket Number 50-346 License-Number NPF-3 Serial Number 2060-July 1, 1992 United States Nuclear Regulatory Commission Document' Control Desk.

Vashington, D. C. 20555

- Subj ec t : Response to NRC Generic Letter 92-01~, Revision 1, Reactor

_ Vessel Structural Integrity, for the Davis-Besse Nuclear Power Stetion Gentlemen:

This' letter provides Toledo Edison's response for the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), to the Nuclear Regulatory Commission (NRC) Generic Letter 92-01, Revision ~1, Reactor Vessel Structural lategrity, 10 CFR 50.54(f), dated March 6, 1992 (Toledo Edison Log Number 3705). Generic Letter 92-01, Revision 1, requests information to-enable the NRC to assess the degree of compliance with regulatory requirements regarding reactor vessel integrity.

i The information requested by Generic Letter 92-01, Revision 1, is provided'in Attachment 1 to this letter. The Attachment 1 responses identify the location of the requested detailed information applicable to the DBNPS. This detailed information is contained in a B&W Owners

~

Group report, BAV-2166, B&W Owners Group Response to Generic Letter 01, dated June 1992. BAV-2166 was prepared by the B&V Nuclear Services Company.for.the B&V Owners Group Reactor Vessel Vorking Group.

The B&V Ovners Group submitted BAV-2166 to the NRC by letter dated June L

-17, 1992-(0G-1036). For_ convenience,.the referenced portions of BAV-2166 applicable to the DBNPS are provided as Attachment 2.

000 00 9207080262 920703 PDR ADOCK 05000346 , Q.

P PDR v

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Operotmg Compan:es Clevelonc Etectnc muminating j  !

Toledo Edison i

Docket Number 50-346 License Number NPF-3

= Serial Number 2060 Page 2-If'you haveyany questions regarding the information provided by this letter, please contact.Mr. Robert V.-Schrauder, Manager - Nuclear-Licensing, at (419) 249-2366.

Very truly-yours,

/%% '&

PVS/dle attachments

-cc: K. 0. Cozens, NUMARC A. B. Davis, Regional Administrator, NRC Region III J. -B. Harkins, NRC/NRR DB-1 Senior Project Manager LV. Levit, NRC Region III, DB-1 Senior Resident Inspector Utility Radiological Safety Board i'

l I-l

- .~ . -

'Docke'. Number 50-346 Licer.se Number NPF-3 Serial Number 2060 Enclosure Phge 1-RESPONSE TO GENERIC LETTER 92-01, REVISION 1 FOR DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 This letter is subn,itted in conformance with Section 182a of the Atomic Energy Act of 1954 as amended, and 10 CFR 50.54(f). Enclosed is Toledo Edison's response to Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity.

By:

D. C.'Shelton, Vice President Nuclear Sworn and subscribed before me this 1st day of July, 1992.

dltf)LS. (l0 Notary PublYe', S' tate of Ohio EVELYNL dress gDPUDUC.SWE CcOHO 24*esmya m; l

l l

l

Docket Number 50-346-License Number NPF-3 Serial Number 2060 Attachment 1 Page 1 RESPONSE TO GENERIC LETTER 92-01, REVISION 1

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 (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 H to 10 CFR Part 50 are requested to indicate when the revised program vill 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 Patt 50 under 10 CFR 50.60(b).

Response

No response is required for the Davis-Besse Nuclear Power Station. -

Unit 1 (DBNPS). The DBNPS is listed in Enclosure 2 to Generic Letter 92-01 as a plant with an NRC epgroved integrated surveillance program.

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

a. 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 Nhc the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their corrent license for the limiting beltline veld 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.

, Response:

l The limiting Charpy upper shelf energy (UEE) is not predicted to be less than 50 foot-pounds prior to expiration of the operating license-(April 22, 2017) for either the limiting beltline veld material or beltline plate or forging.

. (

Reference:

BAV-2166, Davis-Besse, Table 2).

i l

l L

l - .- . _ - , ,__ -__ _ - _ __

' Docket Number 50-346 License Number NPF

• Serial Number 2060 A.ttachment 1 Page 2

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 50.63 and Paragraph III.A of 10 CFR Part 50, Appendix G:

(1) the results from all Charpy and drop veight tests for all unirradiated beltline materials , the unirradiated reference temperature for each beltline material, and the method of determ4. 'ng the unirradiated reference temperature from the Charpy and drop veight 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 vire and flux lot number used to fabricate each beltline veld; (4) the heat number for each surveillance plate or forging and the heat number of vire and flux lot number used to fabricate the surveillance veld; (5) the chemical composition, in particular the weight percent of copper, nickel, phosphorus, and sulfur for each beltline and surveillance material; and

(

l (6) the heat number of the wire used for Jetermining the veld metal chemical composition if different than Item (3) above.

Response

The DBNPS reactor vessel was constructed to the requirements

,-- for Class A vessels of the 1968 Edition of the ASME Code with Addenda through-Summer 1968. -(

Reference:

BAV-2166, Section 5.1, Supplementary Information, Construction Code).

(1) The results from all Charpy and drop veight tests for all unirradiated beltline materials, the unirradiated r

reference temperature for each beltline material, and the

. method of determining the unirradiated reference temperature from the Charpy and drop veight test are provided in BAV-2166, Davis-Besse. Table 3.

(2) .The heat treatment received by all beltline and surveillance materials is provided in BAV-2166, g Davis-Besse Table 4.

L .(3). The heat number for each beltline plate or forging and the l

heat numoer of vire and flux lot number used to fabricate each beltline veld are provided in BAV-2166, Davis-Besse Table 5.

Docket Number 50-346 License Number NPF-3 Serial Number 2060 A.ttachment 1 Page 3 (4) The heat number for each surveillance plate or forging and the heat nubber of vire and flux lot number used to fabricate the surveillance veld are provided in BAV-2166, Davis-Besse Table 6.

(5) The chemical composition, in particular the weight percent of copper, nickel, phosphorus, and sulfur for each beltline and curveillance material is provided in BAV-2166, Davis-Besse Table 7.

(6) The heat number of the wire used for determining the veld metal chemical composition is the same as those identified in the response to Item 2.b(3) above. (

Reference:

BAV-2166, Davis-Besse Table 7).

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

a. i'ov the embrittlement effects of operation at an .

tadiation temperature (cold leg or rteirculation suction

.c rature) belov 525'F vere considered. In particular licensees are requested to describe consideration given to determining the effect of lover irradiation temperature on the reference temperature and on the Charpy upper shelf energy.

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

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.l.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

a. The effect and consideration of irradiation temperature on embrittlement is discussed in BAV-2166, Section 4, Irradiation Temperature. Normal operation of the DBNPS precludes a minimum cold leg temperature, Tc, of less than 332 F, the minimum Tc at zero power (See BAV-2166, Figure -

4-2). Additionally, Technical Specification (TS) 3.1.1.4, Minimum Temperature for Criticality, requires the average reactor coolant system temperature, Tavg, to be greater or equal to 525 F when the reactor is critical. Two instances early in the operation of the DBNPS vere identified where

Docket Number 50-346 License Number-NPF-3 Serial Number 2060 A.ttachment 1 Page-4 Tavg was less than the TS 3.1.1.4 limit. In both of these cases, the TS 3.1.1.4 action statement was invoked. The action statement limits the maximum time in this condition

~

to'less than 30 minutes. This short duration is insignificant,

b. Surveillance results have not been used in determining the Charpy Upper Shelf Energy. Surveillance results were used in the determination of RTNDT following the guidance of Regulatory Guide 1.99, Revision 2, for the preparation of pressure-temperature limit curves for VF-182-1.and VF-233 veld material only. (

Reference:

BAV-2166, Davis-Besse Table 9). These curves are included in BAV-2125, Analysis of Capsule TEl-D, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1, Reactor Vessel Materials Surveillance Program, which was submitted to the NRC by Toledo Edison batter dated April 11, 1991 (Toledo Edison Serial Number 1906) and are applicable through 32 effective full power years (EFPY). The curves have not been incorporated in the Technical Specifications. The current Technical Specifications pressure-temperature limit curves which are applicable through 10 EFPY vere approved by the NRC in Amendment Number 116 to the DBNPS Operating License

-dated August 19, 1988 (Toledo Edison Log Number 2676). The NRC Safety Evaluation-associated with the issuance of Amendment Number 116 found that the current Technical Specifications pressure-temperature limits are based on values of RTNDI which are acceptable relative to the guidance provided by Regulatory Guide 1.99, Revision 2.

c. The measured increase-in RTNDT in no case exceeds the increase predicted by Regulatory Guide 1.99, Revision 2, plus two standard deviations. The measured drop in Charpy Upper Shelf Energy in no case exceeds the drop predicted by Regulatory Guide 1.99, Revision 2, Figure 2. (

Reference:

BAV-2166, Davis-Besse Table 10).

,w.. . , , , -,w , , , - , - , , - , , - , - - . - - -

Docket Number 50-346 l

• License Number NPF-3 Serial Number 2060 A.ttachment 2 Page 1 Referenced Portions of BAV-2166 Applicable to The Davis-Besse Nuclear Power Station

• 4 OWNERS GROUP kiansas n'ev & Light Canpany ANo.1 S+cramentn Marucymt unkry Om RancM M Ouke Power Cconpany Oconee 9.2.3 '

nwsa n-er Corparason GPU Nuctear Corporatum Cry r.1is.w s Tw.t l p ;h- ; g TJodo Edisan Campany r ,nn.,,,, y,u ,aurna,,,

DoesBoese w w, ,, ,

bJ W Nuclear Techrwogwe Working Together to Econornicalty Provide Rehable and Safe Electncal Power Suite 523 e 1700 Rockulle Pike

• Rockville.

June MD 20852 e (301) 230-2100 17, 1992 OG-1036 US Nuclesir Regulatory Comnission Washington, OC 20555 _

Attn: Document Control Desk

Subject:

NRC Generic Letter 92-01 To Holders of Operating Licenses or Conr.truction Permits For Nuclu r Power Plants

Attachment:

BAW-2166 B&W Owners . ,Jup . Response To Generic Letter 92-01 Gentlemen:

The document BAW-2166, is hereby submitted on behalf of the B&W Owners Group Reactor Vessel Working Group. This report provides the information requested in NRC Generic Letter 92-01 for the following plants:

Arkansas Nuclear One Unit 1 Crystal River 3 Davis Besse Ginna Oconee 1, 2, 3 Point Beach 1 and 2 Surry 1, 2 Three Mile Island 1 Turkey Point 3, 4 Zion 1, 2 Utility owners of these plants may rcference this lot.ter and the attachment BAW--2166, in their docket.ed responses.

Very truly yours,

+ov .JpmeT H. TA4foy James H. Taylor Manager Licensing Services JHT/DLH/mcl

BAW-2166 June 1992 B&W OWNfRS GROUP i RESPONSE T0. GENERIC LETTER 92 01 by ,

M. J. DeVan, L. B. Gross, and A. L. Lowe, Jr.

T BWflS Docum;nt No. 77-2166 00 (See Section 7 for document signatures.) 1 t

e Prepared for

. B&W Owners Group Reactor Yessel Workina Group Commonwealth Edison Company Duke Power Company Entergy Operations, Inc.

Florida Power Corporttion Florida Power & Light Company GPU Nuclear Corporation Rochester Gas and Electric Corporation - l' Toledo Edison Company Virginia Power Ccmpany Wisconsin Electric Power Company Prepared by B&W Nuclear-Service Company Engineering and Plant Services Division 3315 Old Forest Road P. O. Box 10935 Lynchburg, Virginia 24506-0935

. . . - . .._ . - . ~ . . . . - - _ - . - . . - , . , - . . - . - - - - , - . . . , - - . - . , . . - . - . - - -

1

4. IRRADIATION TEMPERATURE ,

Material sensitivity to irradistion embrittlement is directly affected by irradiation temperature. Over the temperature range that most light-water cooled reactors operate, the irradiation embrittlement is inversely related to irradiatien temperature. However, since current generation pressurir. : water cooled reactors operate over a the relatively narrow temperature range (i.e. 329-556F RV inlet temperature), the relative sensitivity of the beltline materials as a function of temperature is easily overshadoweo by other parameters such as variations in material properties and Charpy impact testing techniques. The development of Regulatory Guide 1.99, Revision 2, was based solely on surveil-lance data in the irradiation temperature range of 525 to 575F. Normally, the Regulatory Guide 1.99, Rev. 2 data is applied directly in the evaluation of a reactor vessel on the assumption that the reactor vessel temperature was always within this temperature range. However, as can be seen from a review of reactor coolant system temperature as a function of power, the inlet temperature can vary. This does not affect the monitoring of irradiation embrittlement of the reactor vessel because the surveillance capsules are located in the downcomer region of the reactor vessel and experience the same temperature history as the reactor vessel.

The reactor coolant system temperatures as a function of power for each plant included in this .eport are reviewed below. These data were provided by each plant owner and are as stated in their respective FSAR's.

4.1. B&W-Desianed 177-FA Plants Figure 4-1 shows the reactor vessel outlet temperaturs (Tno,) and the reactor

~

vessel inlet temperature"(Tc ) for the B&W 177-FA reactor vessels. This is 4-1 l

representative of all 177.FA plants except Davis-Besse. These operating limits are characterized by a constant system average temperature and an increase in the inlet temperature (Tcw) to 580F with a reduction in operating power. These temperature characteristics result from the fact that initial approach to power is controlled by the wder level in the steam generator followed by a change in operation to maintain the systa swrage temperature constant. The increase in inlet temperature may have the effect of minimizing irradiation embrittlement for these plants.

4.2. Davis-Besse Figure 4-2 shows the reactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tc.) for the Davis-Besse reactor vessel. The system behavior is similar to that of the other 177-FA plants with the exception that the change from level control to control of system average temperature is at approximately 28% power.

4.3. R. E. Ginna Figure 4-3 shows the reactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tc ) as a function of power for the R. E. Ginna reactor vessel. These operating limits are characterized by an increasing average temperature and a near constant reactor vessel inlet temperature for all power levels.

4.4. Point Beach Units 1 and 2 Figure 4-4 shows the eactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tc.) as a function of power for the Point Beach Units ~

1 and 2 reactor vessels. These operating limits are characterized by an increasing average temperature and a small decrease in reactor vessel inlet temperature as power increase to 100%.

l l The Point Beach Unit 1 operated at a reduced power from approximately December 1, 1979 to October 1, 1983, as shown in Figure 4-4. During this period, the reactor vessel was operaled at.a temperature of 511F at 80% to 522F at 0% power.

l.

4-2 l

Figure 4-1. Reactor Coolant System Temperatures as a Function of Power for B&W 177-FA Plants Except Davis-Besse

u. I I I 602F 2 600 -

Two, 3

E $80F T 580 Nominal - D' ' ' 8.' ,,,579F_

5 _ _ _ _ , ,, ..,

Q- Teola E

o 560 - -

H 556F 5 540 -

o 532F (Hot Standby)

O o 520 - -

E3 3 500 - Level -

m Control T4 ,,,,,, Control o

I I I I 480 O 25 50 75 100 Nuclear System Power, Percent figure 4-2. Reactor Coolant System Temperatures as a Function of Power for Davis-Besse O

u.  ; , ,

- 602F

$ 600 -

Tso m 580F Nominal "h 580 -

<------ U* ~ - - - 579 {

Q. / TCold E

o 560 - -

H 556F

~

@ 540 - -

~

o 532F (Hot Standby)

$ 520 - -

w

~

hm 500 -

Level -

Control T4 ,,,,, Control o  :

[ 480 O 25 50 75 100 Nuclear System Power, Percent 4-5

5. SUPPLEMENTARY INFORMATION 5.1. Construction Code The reactor vessels for the following plants were fabricated in accordance with Section !!! of the ASME Boiler and Pressure Vessel Code. The Edition and Addenda (where applicable) of the Code are noted.

Pl ant Section III Edition and Addenda

-Arkansas Nuclear One Unit 1 1965 Edition, Summer 1967 Addenda Crystal River Unit 3 1965 Edition, Summer 1967 Addenda Davis-Besse Unit 1 1968 Edition, Summer 1968 Addenda R. E. Ginna Unit 1 1965 Edition Oconee Unit 1 1965 Edition, Summer 1967 Addenda Oconee Unit 2 1965 Edition, Summer 1967 Addenda Oconee Unit 3 1965 Edition, Summer 1967 Addenda Point Beach Unit 1 1965 Edition Point Beach Unit 2 1965 Edition Surry Unit 1 Not available, final assembly by Rotterdam Surry Unit 2 Not available, final assembly by Rotterdam i Three Mile Island Unit 1 1965 Edition, Summer 1967 Addenda Turkey Point Unit 3 1965 Edition, Summer 1966 Addenda Turkey Point Unit 4 1965 Edition, Summer 1966 Addenda Zion Unit 1 ,

1965 Edition, Summer 1966 Addenda Zion Unit 2 1965 Edition, Summer 1966 Addenda 5-1

l l

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

l Davis-Besse Unit 1 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 ADB 203 Nozzle Belt 3.92E+17 2.35E+17 1.50E+18 9.01E+17 Forging AKJ 233 Upper Shell 2.80E+18 1.68Ee18 1.07E+19 6.4?E+18 Forging BCC 241 Lower Shell 2.80E+18 1.6BE+18 1.07E+19 6.43E+18 Forging WF-232 Nozzle Belt to 3.92E+17 --- 1.50E+18 ---

Upper Shell Cire. Weld (9% 10)

WF-233 Nozzle Belt to ---

2.35E+17 ---

9.01E+17 Upper Shell Circ Weld (91% 00)

WF-182-1 Upper Shell to 2.80E+18 1.68E+18 1.07E419 6.43E+18 Lower Shell Cire. Weld WF-232 Lower Shell to 1.57E+16 ---

6.00E+16 ---

Outchman Cire.

Wald (12% ID)

WF-233 Lower Shell to ---

9.42E+15 ---

3.60E+16 i- Dutchman Cire.

l Weld (88% 00)

L R. E. Ginna Unit 1 Fluence. 12/16/91 Fluence. 32 EFPY Material- Location !S _

T/4 IS T/4 L

123Pll3 val Nozzle Belt 2.05E+18 1.39E+18 3.69E+18 2.50E+18 Belt forging 125S255VA1 Interm. ShelT 1.86E+19 1.26E+19 3.35E+19 2.27E+19 Forging 5-5 l

TABLE 1. GENERIC LETIER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Davis Besse Unit I Question I: Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes / No a Question II: Is piant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, 4

Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zicn-2. Yes / No o IFANSk'(RIS"YES"TOEITHERQUESTIONIORQUESTIONII,PROCEEDTOTABLE2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED T0 QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

Response

Not applicable (see Question I and II above)

Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from HRC? l

Response

Not applicable (see Question I and II above)

NOTES: BAW-10100A: Surveillance Program Description ,

(ASTM E 185-73)

q .. ~. 1 l' ,

j . .i TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION'2, ITEM a c

Subject:

10CFR50, Appendix G, C,USE Requirements Plant: Davis-Besse Unit 1 i

i Column 1 Column'2 Column 3 Column 4 i.

Limiting Initial EFPY to reach If Column 2 is within license Action taken

! Material USE Cy USE<50 ft-lb period: C,USE at indicated time per IV.A.1

! ft-lb 4

Column 3A Column 3B- 't 2 12/16/91 EOL LIMITING'

! BELTLINE WELD I j l WF-233 70 (2) >32 NA NA - NA i LIMITING 1 BELTLINE PLATE

. OR FORGING 1  !

118 (3) >32- NA NA NA

E (?41

, e i

j NOTES: (1) Fluence values taken at %-thickness.

I

i. (2) BAW-1803

(3) BAW-1820 t 4

4 I i

j i a

i

]

4 TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Davis-Besse Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C.6 ,

1 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad Method of Notes  !

Materials Dropweight . Determng

• ' Col. 2a Col. 2b Col. 2c Col. 2d Test RTo , RT ,

Results T,

Cy ( C C y 10 F 30 ft-lb 50fl-lb 35 MLE ft-lb F F F r

FORGING ADB 203 71,70,67 +48 +65 Not avail. +50 450 NB-2331 (1,3,6) 118,113,102 AKJ 233 ND -15 +30 +15 +20 +20 NB-2331 (1,4,6)

BCC 241 ND -14 +27 +5 +50 +50 NB-2331 (1,4,6)

WELD WF-232 25,31,35 ND ND ND ND -5 Est. (2) (1,5)

WF-233 43,30,26 .NG ND ND ND -5 Est. (2) (1,5) -

WF-182-1 36,33,44 +5 +62 Not avail. -20 +2 NB-2331 (1,5)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

I t

{

I

TABLE 3 (CONTINUED)

NOTES 10 IABLE 3:

l (1) BAW-1820 (2) BAW-1803, Revisicn 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

j I (3) C,(+10F) values are for 40 hr stress-relief; other values for unknown stress-relief.

(4) Values are for 15's hr stress-relief.

(5) C,(+10F) values are for 48 hr stress-relief.

(6) Supplementary Mt. Vernon tests of excess surveillance program material.

S

t 4

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (2) ,

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties.Related to PTS and Fracture Toughness Requirements -- APPllCABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE

• EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Davis-Besse Unit 1 Column 1 Column 2 Col. 3 Mater:a1 Heat Treatment Kates BELTLINE MATERIALS ADB 203 1590110F-6h/WQ; 1240110F-14h/WQ; 1100-Il50F-15th/FC (cu=ul.) (1,2)

AKJ 233 1590110F-4h/WQ; 1240110F-6h/AC; 1100-Il50F-15h/FC (cumul.)

BCC 241 1590110F-4h/WQ; 1240110F-Sh/AC; 1100-Il50F-15h/FC (cumul.)

WF-232 Il00-1150F-14h/FC (cumul .)

WF-233 1100-1150F-14h/FC (cumul.) <

WF-182-1 1100-Il50F-15h/FC (cumul.)

WF-232 1100-1150F-14%h/FC (cumul.)

WF-233 Il00-1150F-14%h/FC (cumul.)

i SURVEILLANCE MATERIALS BCC 241 1590110F-4h/WQ; 1240110F-Sh/AC; 1100-Il50F-15%h/FC (1) .

AKJ 233 1590110F-4h/WQ; 1240110F-6h/AC; 1100-Il50F-15%h/FC ,

WF

____-182-1 Il00-Il50F-15gh/FC l

NOTES:

(1) BAW-1820 (2) Additional stress relief information per MT. Vernon process drawing.

(3) WQ - water quench [

AC - air cool t FC - furnace cool

__ _ ___ - ______-________m

6 TABLE 5. GENERIC LETTEC 92-0) RESPONSE: SECTION 2. IlEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Davis-Besse Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes Plateor! ,

• iumber Weld Heat Lot i Forging 1

N3 Forging ADB 203, 123Y317 NB to US Circ.(ID 9%): WF-232 8T3914 8790 (1)

US forging AKJ 233, 123X244 NB to US Circ.(OD 91%): WF-233 T29744 8790 LS Forging BCC 241, SP4086 US to LS Circ.: WF-182-1 821T44 8754 LS to Dutch C hc.(ID 12%): WF-232 8T3914 8790 LS to Dutch ' +7;.(00 88%): WF-233 T29744 8790 NOTES: (1) BAW-1820 (2) NB - Nozzle Belt US - Upper Shell LS , Lower Shell

i .

3

^

3 i '

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (4)  !

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture

! Toughness Requirements -- APPLICABLE- ONLY TO REACTOR VESSELS CONSTRL'CTED TO AN ASME CODE i EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA  ;

I Plant: -Davis-Besse Unit 1  !

Column'l Column 2 Column 3 Column 4 Column 5 [

Surveillance Surveillance Weld Wire Weld Flux- Notes  !

. Plate or Weld Heat Lot  !

3 Forgihg i Heat Number [

i BCC 241, SP4086 WF-182-1 821T44 8754 (1)

AKJ 233,.123X244 i

i i

r i

I

!' NOTES: (1) BAW-1820 i

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TABLE 7. GENERIC LETTER 92-01_ RESPONSE: SECTION 2 ITEM b, 1 (5) 2 Subject- ;10CFR50.61'and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture

! iToughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE

EARLIER THAN THE 1971 EDITION. SUtmER 1972 ADDENDA Plant: Davis Besse Unit 1 Column 1- Column 2 C. 3 Material Chemical Conosition, Weight Percent Notes 1

' Cr Ni

~

C Mn P S Si Mo Cu ,

BEL 1LINE MATERIALS ADB 203 0.23 0.70 0.007 0.009 0.29 0.39 0.68 0.63 0.04 (I) 4 AKJ 233 0.26 0.68 0.004 0.006 0.30 0.38 0.77 0.64 0.04 (1)  !

BCC 241 0.22 0.63 0.0'1 . 0.011 3.27 0.32 0.81 0.63 0.02 (1)

WF-232 0.06 1.30 0.016 0.011 0.47 0.11 0.64 0.37 0.18 (2)

WF-233 0.05 1.45 0.021 0.015 0.42 0.08 0.68 0.44 0.29 (2) l l WF-182-1 0.08 1.69 '0.014 0.013 0.45 0.14 0.63 0.40 0.24 (2) 4 i l

SURVEILLANCE l MATERIf3 BCC 241 0.22 0.63 0.011 0.011 0.27 0.32 0.81 0.63 0.02 (3) l 0.26 0.68 0.004 0.006 0.30 0.38 0.77 0.64 0.04 (3) i l AKJ 233 WF-182-1 0.09 1.69 0.014 0.013 0.41 0.15 1 0.63 0.40 0.21 (3) '

l 'REQUIREE. state heat number of weld wires used for determining above chemical composition if different Gm I

that in 1 (3). -- Not applicable --

NOTES:

(1) BAW-1820  ;

(2) BAW-2121P (3) BAW-1543, Revision 3 i I

i  :

y

^o' ?; . .tet, TABLE 8. MitdIC LETTER 92-01 RESPONSE: SECTION 3. ITEM a

,. #. c.

4 l

Subject:

Generic tetter 88-11 " p r ic Commitments Effect of Irradiation Temoerature Plant: Davis-Besse Unit I _

Cold Leg Temperature (T;,,a): 556 F (See Figure 4-2)

If T ,# is <525 F, state how this was considered in determination of embrittlement effects (C./J5E, RT,) in accordance with Regulatory Guide 1.99, Revisio.i 2:

r ,

I - '

Not applicable r 4 I I ,

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References:

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f

TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b f

Subject:

Generic Letter 88-11 Response Camitments: Utilization of Surveillance ,

Results Plant: Davis-Besse Unit 1 Were surveillance results useri in determining CoUSE? Yes a No /

Were surveillance results used in determining RTm,7 Yes / No o If any "yes" boxes were checked above, state hcw tN ::arveillance results were used:

l Determination pressere-temperature of RT, limit curves for WF-182-1 and WF-233 weld materials only.per Regulat I

l

Reference:

BAW-2125 l

4 TABLE 10. GENERIC LETTER 92-01 RESP 0 rise: SECTION 3, ITEM c 1

I

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Davis-Besse Unit 1 Question I. Does measured ART,,, exceed ARTwor + 2a predicted by Regulatory Guide 1.99, Revision 2?

Question II.' Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2? _

, Column 1 Column 2 Column 3 f Column 4 Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted Ouestion I Measured Predicted Question II i

Material ART,o, ART,o,+2a If "yes' CyUSE CyUSE If "yes"  ;

n/cm ~

(3) see Note (4) .

Drop Drop see Note'(4) i ADB 203. ---

ND ND --

ND -- --

1.29E+19 56 No 13(1) 17(1) No AKJ 233 2(1) 1.96E+18 24 9(2) 9(2) No i BCC 241 0(2) --

5.92E+18 34 9(2) 11(2) No 0(2) --

1.29E+19 44 No 4(2) 14(2) No 28(2) 9.62E+18 3(2) 40 No. 5(2) 12(2) No l ND ND ND ND --

WF-232 ---

WF-233 4.67E+18 211 No 18(3) 23 No 191(3) 1.08E+19 187(3) 257 No 24(3) 28 No 1.21E+19 263 No 19(3) 28 No i  ! 222(3)

WF-182-1 1.96E+18 152 No 6(3) 17(2) No [

127(3)  ;

5.92E+18 125(3) 200 No 13(3) 22(2) No '

1.29E+19 y 237 No 8(3) 26(2) No 175(3) 9.62E+18 l 223 No 16(2) 25(2) No 150(2) l l ,

1 t

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

i

[

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4 TABLE 10 (CONTINUED)

NOTES: (1) BAW-1882, Revision 1 (2) BAW-2125

' (3) BAW-1803, Revision 1 (4) No Statement required.

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7. REFERENCES

~

-'A. L. - Lowe, Jr. et al, " Analysis of Capsule OCl-F ' from Duke Power Company ,

Oconee Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1421. Revision 1, Babcock & Wilcox, Nuclear Power Generation Division, Lynchburg, Virginia, September 1975.

'A. L. Lowe, Jr. et. al, " Analysis of Capsule AN!-B from Arkansas Power & Light Company's ' Arkansas Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1698, Babcock & Wilcox, Nuclear Power Generation - Division, Lynchburg, Virginia, November 1981.

- A. S. Heller and A.- L. Lowe, Jr., " Correlations for Predicting the Effects of Neutron' Radiation on Linde 80 Submerged-Arc Welds," BAW-1803, Babcock & Wilcox, Utility Power Generation Division, Lynchburg, Virginia, January-1984.

A. L. Lowe,,Jr. and J. W. Pegram, " Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged Arc Welds, BAW-1803. Revision 1, b&W -

Nuclear Service Company, Lynchburg, Virginia, May 1991.

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'A. L. Lowe, Jr. et al, " Analysis of Capsule TEl-A, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1, Reactor Vessel Materials Surveillance

l. Program," BAW-1882. Revision 1, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, June 1989.

'This report .is available from B&W Nuclear Service Company, LynchburC, Virginia.

7-1 L

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L. Lowe, Jr. et al, " Analysis of Capsule CR34, Florida Power Corporation, Crystal River Unit 3, Reactor Vessel Materials Surveillance Program," BAW-1898, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, March 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsule CR3-D, Florida Power Corporation, Crystal River Unit 3, Reactor Vessel Materials Surveillance Program," BAW-1899, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, March 1986.

'A. L. Lowe, Jr. et al, "Analyr's of Capsule TMll-C, GPU Nuclear, Three Mile Island Nuclear Station-Unit 1, Reactor Vessel Materials Surveillance Program,"

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Information for Surry Units 1 and 2, North Anna Units 1 and 2," BAW-1908, Babcock

& Wilcox, Nuclear Power Division, Lynchburg, Virginia, February 1986.  !

'A. L. Lowe, Jr., " Reactor Pressure Vessel and Surveillance Program Materials Licensing Information for Surry Units 1 and 2," EAW-1909. Revision 1, Babcock &

Wilcox, Nuclear Power Division, Lynchburg, Virginia, August 1986.

'A. L. Lowe, Jr. et al, " Analysis of Qps910 CR3-LG1, Babcock & Wilcox Owners Group, Integrated Reactor Vessel Materials Surveillance Program," BAW-1910P, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, August 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsale DB1-LG1, Babcock & Wilcox Owners Group, Integrated Reactor Vessel Materials Surveillance Program," BAW-1920P, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, October 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsule CR3-F, Florida Power Corporation, Crystal River Unit 3, Reacto. Vessel Materials Surveillance Program," BAW-2049, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, September 1988.

'A. L. Lowe, Jr. et al, " Analysis of Capsule OCl-C, Duke Power Company, Oconee Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2050, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, October 1988.

72

'A. L. Lowe, Jr. et al, " Analysis of Capsule OCII-E, Duke Power Company, Oconee Nuclear Station Unit 2, Reactor Vessel Materials Surveillance Program," BAW-2051, Babcock & Wilcox, Nuclear Power Division, Lynchburg, firginia, October 1988.

'A. L. Lowe, Jr. et al, " Analysis of Capsule A.N1-C, Arkansas Pcwer & Light Company, Arkansas Nuclear One, Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2075. Revision 1, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, October 1989.

A. L. Lowe, Jr. et al, " Analysis of Capsule Y, Commonwealth Edison Company, Zion Nuclear Plant Unit 1, Reactor Vessel Materials Surveillance Program." BAW-20B2, B&W Nuclear Service Company, Lynchburg, Virginia, March 1990.

A. L. Lowe, Jr. " Properties of Weld Wire Heat Number 72105 (weld Metals WF-70 and WF-209-1)," BAW-2100, B&W Nuclear Service Company, Lynchburg, Virginia, To be l puM i shed.

K. K. Ycon and A. L. Lowe, Jr., " Low Upper-Shelf Toughness Fracture Analysis of Reactor Yessels of Turkey Point Units 3 and 4 for Load Level A and B Condition,"

BAW-2118P, B&W Nuclear Service Company, Lynchburg, Virginia, November 1991.

L. B. Gross, " Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," BAW-2121P, B&W Nuclear Service Company, Lynchburg, Virginia, April 1991.

A. L. Lowe, Jr. et al, " Analysis of Capsule TEl-D, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2125, B&W Nuclear Service Company, Lynchburg, Virginia, December 1990.

A. L. Lowe, Jr. et al, " Analysis of Capsule OCIII-D, Duke Power Company, Oconee Nuclear Station Unit-3, Reactor Vessel Materials Surveillance Program," BAW-2128, B&W Nuclear Service Company, Lynchburg, Virginia, May 1991.

A. L. Lowe, Jr. et.al, " Analysis of Capsule S, Wisconsin Electric Power Company, Point Beach Nuclear Plant Unit No. 2, Reactor Vessel Materials Surveillance Program," BAW-2140, B&W Nuclear Service Company, Lynchburg, Virginia, August 1991.

7-3 w ' -- we i- ~- -, m- , py

O K. K. Yoon and L. B. Gross, " Low Upper-Shelf Toughness fracture Analysis of Reactor Vessels of Zion Units 1 and 2 for Load Level A and B Conditions," BAW-2148P, B&W Nuclear Service Company, Lynchburg, Virginia, March 1992.

'H. S. Palme et al, " Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Appendix G," BAW-10046P, Babcock & Wilcox, Nuclear Powar Generation Division, Lynchburg, Virginia, March 1976.

W. A. VanDerSluys et al, "An Investigation of Mechanical Properties and Chemistry Within a Thick Mn-Mo-Ni Submerged Arc Weldment," EPRI NP-373. Electric Power Research Institute, Palo Alto, California, February 1977.

E. B. Norris, " Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4: Analysis of Capsule T," Final Recort SWRI Project No. 02-4221, Southwest Research Institute, San Antonio, Texas, June 1976.

E. B. Norris, " Reactor Vessel Material Surveillance Program for Capsule S -

Turkey Point Unit No. 3 (and) Capsule S- Turkey Point Unit No. 4," Final ReDort SWRI Project Nos, 02-5131 and 02-5380, Southwest Research Institute, San Antonio, Texas, May 1979.

P. K. Nair and E. B. Norris, " Reactor Vessel Matertal Surveillance Program for Turkey Point Unit No. 3: Analysis of Capsule V," Final Report SWRI Project No.

06-8575, Southwest Research Institute, San Antonio, Texas, August 1986.

S. E. Yanichko et al, " Analysis of Capsule T from the Florida Power and Light Company Turkey Point Unit No. 3 Reactor Vessel Radiation Surveillance Program,"

VCAP-8631, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1975.

S. E. Yanichko and S. L. Anderson, " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Surveillance Program," WCAP-9357, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, August 1978.

7-4 l

. .= -_- -. .

- . _ - - . . .= . - -

+ -- mm ,

S.' E. Yanichko et al, " Analysis of Capsule T from the Rochester Gas and Electric Corporation R. E. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP-10086, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, April 1982.

S. E. Yanichko et al, " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveil .

lance Program," WCAP-10736 Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1984.

S. E. Yanichko and V. A. Perona, " Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11415, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, February 1987.

S. E. Yanichko and V. A. Perone, " Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-11499, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, June 1987.

S. E. Yanichko et al, " Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-12396, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1989.

t l 7-5  ;

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l l 1 o