ML19317E313

Forwards Revised Pages to BAW-1437, Analysis of Capsule OCII-C Oconee Unit 2 Reactor Vessel Matls Surveillance Program, June 1977.Rept Submitted as Supporting Document in 770606 Request for Proposed Amend to Tech Specs

ML19317E313
Person / Time
Site:	Oconee
Issue date:	10/24/1977
From:	Parker W DUKE POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7912170539
Download: ML19317E313 (9)

Text

__

u.s. NUCLEAR nEGULATORY COMV' 16 N DOCKET

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NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL NRC FROM: Duke Pwr CO DATE OF OCCUMENT TO:

Charlctte, NC 10-24-77 W 0 Parker JR DATE RECEIVEo 10-28-77 CNOTORIZEO PRCP INPUT FORM NUMBER CF CCPIEs RECEIVED (LETTEn 26MIGINAL MNCLASSIFIE D OcePv f gigag ESCRIPTION ENCLOSURE 1p replacement pgs making corrections to report BAW-1437, dtd June 1977, ent'itled " analysis of Capsule OCII-C from Duke Pwr Co Oconee #2 Reactori Vessel Materials Surveillance Program" which was  !

submitted in support of their proposed tech specp

- change concerning pressurization, heacup & cool-down limitations...............

4p DISTRIBUTION FOR MATERIAL ON REACTOR VESSEL DATA PER R. INGPA'4 5-26-77 176= BAW-1437 PLANT NAME: . Oconee 1-3 l tib eact .

SAFFTY FOR ACTION /IT3 FORMATION l BRANCH CHIEF: (.3 ) l Sd//a/4//06# l 1 PSCL!EST :'A'!ACO l l

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! I Tic l I NSrc i i i - 773000262 1 ACRM 14 eve er*N e FNov 4 I Id012170 { f

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. RESULFTDOCXET FILECDPY DUKE POWER COMPANY Powra D':st.tsixo 422 SocTu Cucacu SrazzT, CrunwrTE, N. C. 2824a w w w o. Pa a a c a. s a. October 24, 1977

't%Ep=CN C: A A E4 70 4 V+Cr PotsrOt%f S Y t ana PeCO .C* ION 17 3- 4C 8 3 k[,

Director

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7 tp). CA Office of Nuclear Reactor Regulation */

U. S. Nuclear Regulatory Commission I d %, 8 C Washington, D.C. 20555 o g/S Jg RE: Oconee Nuclear Station \ *'% ,.U Docket Nos. 50-269, -270, -287  %

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Dear Sir:

Please find attached replacement pages making corrections to report BAW-1437, June, 1977, " Analysis of Capsule OCII-C from Duke Power Company Oconee Unit 2 Reactor Vessel Materials Surveillance Program".

This report was submitted as a supporting document in our June 6, 1977 request for a proposed amendment to the Oconee Technical Speci-fications revising pressurization, heatup and cooldown limitations for Oconee Unit 2. The replacement sheets correct an error in three figure titles.

Very truly yours, A

William O. Parker, Jr.

LJB:ge Attachment

+

773000262 l.

- +

THE BABC0CK & WILCOX COMPANY POWER GENERATION GROUP

" ~

_ To l Distribution From A. L. Lowe, Jr., Technical Staff nos eas.s Cust. File No.

Duke Power Company (Oconee Unit 2) Of E8f- BAW-1437 Su bj . Analysis of Capsule OCII-C From Duke Power Company Date Oconee Nuclear Station, Unit 2 - Report BAW-1437 October 5, 1977 lm.i....,......,..........,........i.....ir.

An error h 9 been discovered in three figure titles in report BAW-1437. Please replace contents pages v/vi and figure pages 8-5 through 8-8 in your copy of the report with the attached corrected pages. This correction does not involve text or tables.

ALL:be Distribution:

Duke Power Company (70) Merchent, JW Wimmer, LB c/o CD Russell, OFR Moore, KE Helmbrecht, HL/NED Barberton 88"' "E ' *# *#

• Behnke, HW/Mt. Vernon Ayres, PS/ Alliance ew n, Borsum, RB/Bethesda Palme, HS (2) Chulick, ET/LRC (2)

Dobel HF sse r ss, M RC Durant, WP/Mt. Vernon ,

Schuler, 'IM Poor, HH/ Alliance Sivashankaran, S/Mt. Rowe, JP/ Alliance ssi ZurLiPPe, CF/LRC (2)

Vern" Keyworth, WJ (3) Smith, RM evstek, DF Travis, CC/TRG

1. 7 Whitmarsh, CL (2) g g)

. /

)

, a *

'4 Tables (Cont'd)

Table Page B-3. Preirradiation Ter.cile Properties of Shell Plate Material, Heat AWG 164 . ........... . . . . . . . . . . .... B-4 B-4. Preirradiation TensiJe Properties of Shell Plate Material --

HAZ, Heat AWG 164 . . . . . . . . . . . . . . . . . . . .... B-5 B-5. Preirradiation Tensile Properties of Weld Metal -- Longitudinal, WF-209-1A .............. . . . . . . . . . .... B-6 C-1. Preirradiation Charpy Impact Data for Shell Course Material -

Longitudinal Orientation, Heat AAW 163 . . . . . . . . . .... C-2 C-2. Preirradiation Charpy Impact Data for Shell Course Material -

Transverse Orientation, Heat AAW 163 . . . . . . . . . .... C-3 C-3. Preirradiation Charpy Impact Data for Shell Course Material -

HAZ, Longitudinal Orientation, Heat AAW 163 . . . . . . .... C-4 C-4. Preirradiation Charpy Impact Data for Shell Course Material -

HAZ, Transverse Orientation, Heat AAW 163 . . . . . . . .... C-5 C-5. Preirradiation Charpy Impact Data for Shell Course Material -

Longitudinal Orientation, Heat AWG 164 . . . . . . . . .... C-6 C-6. Preirradiation Charpy Impact Data for Shell Course Material -

Transverse Orientati n, Heat AWG 164 . . . . . . . . . . .... C-7 C-7. Preirradiation Char ey Impact Data for Shell Course Material --

HAZ, Longitudinal Orientation, Heat AWG 164 . . . . . . .... C-8 C-8. Preirradiation Charpy Impact Data for Shell Course Material -

HAZ, Transverse Orientation, Heat AWG 164 . . . . . . . .... C-9 C-9. Preirradiation Charpy Impact Data for Weld Metal, WF-209-1A .. C-10 D-1. Detector Composition and Shielding . . . . . . . . . . . .... D-2 D-2. Oconee 2, Cycle 1 Neutron Dosimeters . . . . . . . . . . .... D-3 List of Figures Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule

~

Locations .... . . . . . . . . . . . . . . . . . . . .... 3-5 5-1. Impact Data From Irradiated Base Metal A, Longitudinal Orientation ......... . . . . . . . . . . . . . .... 5-6 5-2. Impact Data From Irradiated Base Metal A, Transverse Orientation ... . . . . . . . . . . . . . . . . . . . .... 5-7 5-3. Impact Data From Irradiated Base Metal A - HAZ, Longitudinal Orientation ............. . . . . . . . . . .... 5-8 5-4. Impact Data From Irradiated Weld Metal, Transverse Orientation . 5-9 5-5. Impact Data From Correlation Monitor Material, Transverse orientation .......................... 5-10 6-1. Fast Neutron Fluence of Surveillance Capsule Center Compared to Various Locations Through Reactor Vessel Wall for First 10 EFPY ........................... 6-8 7-1. Irrxdiated Vs Unirradiated Charpy Impact Properties of Base Metal, Longitudinal Orientation . . . . . . . . . . . . .... 7-5 7-2. Irradiated Vs Unirradiated Charpy Impact Properties of Base i Metal, Transverse Orientation . . . . . . . . . . . . . .... 7-6

..y_ Babcock a.Wilcox

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.A 1

'I Figures (Cont'd) l Figure Page _y I

6 7-3. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal, HAZ . . . . . . . . . . . . . .... 7-7 7-4. Irradiated Vs Unirradiated Charpy Impact Properties of Weld Metal, Transverse Orf ration . . . . . . . . . .... 7-8 7-5. Irradiated Vs Unirradiated Impact Properties of Correlation Monitor Material, Longitudinal Orientation . . . . . . .... 7-9 1 8-1. Predicted Fast Neutron Fluences at Various Locations -j Thrcugh Reactor Vessel Wall for First 10 EFPY . . . . . .... 8-5 8-2. Reactor Vessel Pressure-Temperature Limit Curves for _,

Normal _0peration - Heatup, Applicable for First 8 EFPY .... 8-6 8-3. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation -- Cooldown, Applicable for First 8 EFPY ... 8-7 8-4. Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests,

].i Applicable for First 8 EFPY . . . . . . . . . . . . . . .... 8-8 A-1. Location and Identification of Materials Used in -

Fabrication of Reactor Pressure Vessel . . . . . . . . .... A-5 C-1. Impact Data From Unirradiated Base Metal A, Longitudinal Orientation . . . . . . . .. . . . . . . .... C-ll y C-2. Impact Data From Unirradiated Base Metal A, t Transverse Orientation . . . . . . . . . . . . . . . . .... C-12  !

C-3. Impact Data From Unirradiated Base Metal A, HAZ,

Longitudinal Orientation . . . . . . .. . . . . . . . .... C-13 C-4. Impact Data From Unitradiated Base Metal A, HAZ,  ;

Transverse Orientation . . . . . . . . . . . . . . . . .... C-14

! C-5. Impact Data From Unirradiated Base Metal B, +

Longitudinal Orientation . . . . . . . . . . . . . . . .... C-15 0 -6. Impact Data From Unitradiated Base Metal B, Transverse Orientation . . . . . . . . . . . . . . . . .... C-16 C-7. Impact Data From Unirradiated Base Metal B, HAZ, Longitudinal Orientation . . . . . . . . . . . . . . . .... C-17 C-8. Impact Data From Unirradiated Base Metal B, HAZ, Transverse Orientation . . . . . . . . . . . . . . . . .... C-18 C-9. Impact Data From Unirradiated Weld Metal, Transverse Orientation . . . . . . . . . . . . . . . . .... C-19 i

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- vi - Babcock & Wilcox ,

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Figure 8-1. Predicted Fast Neutron Fluences at Various I.ocations Through Reactor Vessel Wall for First 10 EFPY 6.0 5.6 .-

5.2 + 18 nvt 5.2 -

"a u 4.8 -

c 2 4.4 -

> 3 g i ,

s 4.0 A

3 3.6 -

7a 3.2 -

e i e 2.9 + 18 nyt w ~ c/ @

S 2.8 -

he x 9 e 2.4 -

c 8

g 2.0 -

1.6 - . #

tj i't e , s

• 1.2 - L' 6.9 + 17 nvt

@cr 0.8 -

n S 0.4 - 3/4T tocation F Outside Surface 1 h:

0.0 0 1 2 3 4 5 6 7 8 9 10 o

O

' X EFPY a

1

Figure 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation -

lleatup, Applicable for First 8 EFPY 2400 Assumed RTET,F D G Beltline region 1/4T 132 Beltline region 3/4T 56 f 2000 -

Closure head region 60 Outlet nozzle 60 \

y 1800 -

Pressure, Temp, g Point psi _ F d

w 1600 -

A 450 60 g B 625 146 Applicable for a- C 625 273 Ileatup Rates

[y 1400 - D E

2250 625 302 273 up to 100F/h m g F 625 313 f, g 1200 - G 2250 342

' o u

e4 g 1000 -

g 800 -

Critical-u ity Limit

)

g B C

• 600 - F E

400

- -~ - --- -

g The acceptable pressure-temperature combinations are below 4 and to the right ut the limit c urve(s) . The limit curves Jo 0" not include the pressure dif ferential between the point of 200 mystem pre..ure e surement and the pres.ure on the reactor e3 -

o vemmel region controlling the limit curve, or any additional K margin of safety for possible instrument error.

P

- 40 80 120 160 200 240 280 320 360 o

(4 x Reactor Vessel Coolant Temperature, F i__  ! . _. L_- L_.- 1 ' ' -

a L_ J ' ) 1 J l - J ' i I

i i i ,

, I s

,1 a, e

Figure 8-3. Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation - ,

Cooldown, Applicable for First 8 EFPY 2400 .

Assumed RT NDT, F E

2200 -

Beltline region 1/4T 132 Beltline region 3/4T 56 2000 -

Closure head region 60 Outlet nozzle 60 j 1800 -

Pressure, Temp,

E Point psi F i )

$ 1600 -

A 250 70 3 B 625 119 Applicable for E C 625 205 Cooldown Rates U 1400 -

D 1120 213 up to 100F/h j

u E 2250 281 m e b o 1200 -

O D H

g 1000 -

g 800 -

u u

• B

• C

' # 600 -

The acceptable pressure-temperature combinations are below Q"

400 -

and to the right of the limit c urve(s) . The limit curvem do not include t he prensure dif f erential between the point of g bystem pressure measurement and t he pressure on t he- reactor

, O A vessel region cont rolling the limit curve, or any additional g

t 200 -

-s n a or . rety for possible inst rument error.

X" P

g 0 I i i I I i

= 40 80 120 160 200 240 280 320 o

a X Reactor Vessel Coolant Temperature, F

b Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curve fo r Inservice Leak and llydrostatic Tests, Applicable for First 8 EFPY 2600 Assumed RT NDT* f 2400 -

Behline region 1/4T 132 Bel tline region 3/4T 56 2200 -

Closure head region 60 Outlet nozzle 60

}

2000 -

Pressure, Temp,

, Point psi F E. 1800 -

A 330 70 ,- l

- B 625 131 '

E C 625 245 .

E e

1600 -

D 2500 272 [ ',

a e >

$ Applicable for Heatup /

co y 1400 -

and Cooldown Rates up j -

co eo $100F/h (<50F in any es ,

T 1/2-h period) .

g 8 1200 -

i e

E

,e 1000 -

o )

U 800 -

V

=

600 -

C G3 The acceptable pressure-temperature combinations are below

> - A and to the right of the limit curve (s). The limit curves do o not include the preneure dif ferential between the point of O system pressure measurement and the pressure on the reactor O vessel region cont rolling the limit curve, or any additional 200 .

margin of safety for possible instrument error.

P -

I_. O i i e i i E

9.4 60 100 140 180 220 260 300 e

x Reactor Vessel Coolant Temperature, F t

e

( ,_ k . .. $.. . L _. L. - l- - . . b - - - . - - + L-- *- -- ' -