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 TO:

NRC FROM: Duke Pwr CO DATE OF OCCUMENT Charlctte, NC 10-24-77 W 0 Parker JR DATE RECEIVEo 10-28-77 (LETTEn CNOTORIZEO PRCP INPUT FORM NUMBER CF CCPIEs RECEIVED 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

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3 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 V+Cr PotsrOt%f

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

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Director Office of Nuclear Reactor Regulation tp). CA

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U. S. Nuclear Regulatory Commission I d %,

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• '%,.U RE: Oconee Nuclear Station

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Docket Nos. 50-269, -270, -287 p'

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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

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Distribution From A. L. Lowe, Jr., Technical Staff nos eas.s Cust.

File No.

Of E8f-BAW-1437 Duke Power Company (Oconee Unit 2)

Date Su bj.

Analysis of Capsule OCII-C From Duke Power Company 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 ew n,

Ayres, PS/ Alliance 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 Vern" ZurLiPPe, CF/LRC (2)

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 y Impact Data for Shell Course Material --

e 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 Babcock a.Wilcox

..y_

~

.A

'I Figures (Cont'd)

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

,1 a

J a

- vi -

Babcock & Wilcox a

s i

y 4

e9.

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 5.2 + 18 nvt "au 4.8 c

2 4.4 3

i g

4.0 s

A 3

3.6 7a 3.2 e

i c/ @

2.9 + 18 nyt e

w

~

S 2.8 he x

9 e

2.4 c

8g 2.0 1.6 i't tj e

s 1.2 L'

0.8 6.9 + 17 nvt cr n

S 0.4 3/4T tocation F

Outside Surface h

0.0 0

1 2

3 4

5 6

7 8

9 10 1

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 f

Beltline region 3/4T 56 2000 Closure head region 60 Outlet nozzle 60

\\

y 1800

Pressure, Temp, g

Point psi _

F d

1600 A

450 60 w

g B

625 146 Applicable for a-C 625 273 Ileatup Rates

[

1400 D

2250 302 up to 100F/h E

625 273 y

g F

625 313 mf, g

1200 G

2250 342 o

u e4 g

1000 g

800 Critical-ity Limit

)

u 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 e3 200 mystem pre..ure e surement and the pres.ure on the reactor 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

(4x Reactor Vessel Coolant Temperature, F i

I

)

1 J

l

- J i__

L_-

L_.-

a L_ J 1

I

,1 a,

i i

i s

e Figure 8-3.

Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation -

Cooldown, Applicable for First 8 EFPY 2400 Assumed RT F

NDT, 2200 E

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

E 2250 281 u

m e

b 1200 o

O D

H g

1000 g

800 uu B

C 600 Q"

The acceptable pressure-temperature combinations are below 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 vessel region cont rolling the limit curve, or any additional O

t 200 A

g

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

a X"

Pg 0

I i

i I

I i

=

40 80 120 160 200 240 280 320 oaX 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 RTNDT*

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 1600 D

2500 272

[ ',

e a

e Applicable for Heatup

/

1400 and Cooldown Rates up j co y

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

200 vessel region cont rolling the limit curve, or any additional margin of safety for possible instrument error.

P I_.

O i

i e

i i

E 60 100 140 180 220 260 300 9.4 e

x Reactor Vessel Coolant Temperature, F t

e

(,_

k...

L L.

l- -..

b

- - +

L--

-