sg

e

ENCLOSURE 3

VOGTLE ELECTRIC GENERATING PLANTREQUEST TO REVISE TECHNICAL SPECIFICATIONS

REVISION TO REACTOR PRESSURE LIMITS

MARKED UP TECHNICAL SPECIFICATION PAGES

9410110196 941003PDR ADOCK 05000424PPDR

.. - . - . .. . _ . - _ - . -_

|'

i-,

| REPLACE u) TH rJ ew Pa*e 3/4 of 5|

3000,

URVE APPLICABLE FOR THE SERVICERIOD UP TO 13 EFPY

RTN . After 13 EFPYg ,

7 a.1 T < 110* FS b. 3/ T< 87'F LEAK TEST

LIMITgCRITICALITY$ LIMITyFOR 60*F/hrn. 2000 UNACCEP BLE

HEATUP[ OPERATI

> CRITICALITYm j- LIMIT$ jf n

FOR 100'F/hr'

$ HEATUPgo *

m$ 60*F/hro HEATUP

BASED ON INSERVICE 'g 1000 CURVE,f/ HYDROSTATIC TESTeTEMPERATURE (255'F)g g

FOR THE SERVICE>-PERIOD UP TO 13 EFPY$

h 100'F/hrkHEATUP ACCEPTABLE-

,

'

CURVE OPERATION|

I

0.00.0 100 200 300 400 500

LOWEST INDICATED RCS Tcold TEMPERA RE ('F)

MATERIAL BASIS

Copper Content: 0.10 Wt %

( c.e4 wt %)

I .40'FNDT ntttet: ;RT

(Actuel . 'F)After 13 EFPY e 1/4 110'F

RTNDTe 3/4 7 7'F

FIGURE 3.4-2a

UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 13 c PY

V0GTLE UNITS - 1 & 2 3/4 4-31 Amendment No. 36 ( nit 1)Amendment No. 16 (U it 2)

. _. ._; _ . _ . _

_ _ _

,

..

.

2500 g g ,

: LeakTest Limit '1 N I ,I

,

2250

2000;.

' 1750'

| Unacceptable| Operation

]'

3 1500 r

ta|

SE'

a 1250E Heatup Rate up to N Acceptablen. 60 F/hr Operation

1000js Heatup Rate up tos

.E 100 F/hr N750

f Criticality,Umit Based) / on Insenace Hydrostatic/ Test Temperature'

500 yp (246 F) for the ServicePeriod up to 16 EFPY- s

250

i ,

; OL

| 0 50 100 150 200 250 300 350 400 450 500

Indicated Temperature ('F)

MATERIAL BASIS

|Copper Content: Assumed - NAWT%

(Actual- 0.083 WT%)

NDT nitial: Assumed - NA *FIRT,

i (Actual- 20'F)I

RTNDT At 16 EFPY: @ 1/4T - 100.7*F@ 3/4T - 84.1*F

Figure 3.4-2a

| Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to 100'F/hr) Applicablefor the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors andMargin of 74 psig for Pressure Difference Between Pressure Instrumentation and ReactorVessel Beltline Region) .

i

f VOGTLE UNITS 1 & 2 3/4 4-31|

. . _ . _ _

|'| t

! REPcue u>irw Neno PM e- % 'l- 3| o~ |/'

UNIT 2 /* i

CURVE APPLICA8LE FOR THE /,

SERVICE PERIOD UP TO 16 EFPY /

'

|-

LEAK TEST_

l y LIMIT'

S 2000$ f f f

f, f CRIT bALITY LIMIT FOR3 060 F/hr HEATUP$,

w U EPTABLEE OPERATION

h.

HEATUP CURVE [ CRITICALITY LIMIT FORe 0100 F/hr HEATUP$I

8 ACCEPTABLE) OPERATIONu/m 1000 i

| $ 100 F/hr HEATUP CURVE |0

E;

..|, m

| 8'

< BASED ON INSERVICE |

;

| OHYDROSTATIC TEST |

'

E TEMPERATUR E (2880F)l 2 500FOR THE SERVICE-

PERIOD UP TO 16 EFPY

|

t

0.0 100 200 300 400 500

EST INDICATED RCS Tg TEMPER RE(OF),

MATEml A A$t$ I

| Cooper ce Asmened . 0.10 We %( Astwas . 0.05 Wt %I

INOT ""#g

r,,, ~ i . . .u.v.,a,,

j e 3/. T = S P ,I|

FIGURE 3.4-2b

UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 PY

\

V0GTLE UNITS - 1 & 2 3/4 4-31a

. . . . - . - . . _ - . .

.. -_ . _ _ . _ ._

.

2500 -

.

'

fILeak Test Limit

J fJ f

2250||

2000Unacceptable

17501

|

g; 1500l Heat Rate Acceptable

! up to 0 F/hr Operaton

| $ 1250 -

e,

' ec.uS 1000B ,f-3

Heatup Rate s

up to 100 F/hr |750

f 1

Criticality Umit Based 1

/ on Inservice Hydrostatic| / TestTem rature 1

; ,

5001 - (256 F) f r the Service !'

Period up to 16 EFPY j'

- s

\

250 |'

|

| I

! 0j 0 50 100 150 200 250- 300 350 400 450 500 l

Indicated Temperature ( F)

MATERIAL BASIS

Copper Content: Assumed - NA WT%

(Actual- 0.05 WT%)

NDT nitial: Assumed NA'FRT I,

! (Actual- 50*F)At 16 EFPY: @ 1/4T - 112'FRTNDT

@ 3/4T = 94*F

Figure 3,4-2b| Unit 2 Reactor Coolant System Heatup Limitations (Heatup rates up to 100 F/hr) Applicable

for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors andMargin of 74 psig for Pressure Difference Between Pressure Instrumentation and ReactorVessel Beltline Region).

,

VOGTLE UNITS 1 & 2 3/4 4-31a

!

. . _ . _ __ __ _ _ _ _ _ - - - . _ _ - - . _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - -

|! . .

!- Rehe wem uew pus J/+ 4 31-

UNIT 1 7 I

3000 g j f

CURVE APPLICABLE FOR THE SERVICEi

| PERIOD UP,TO 13 EFPY!

!

!

| 3 ' -

> . -

! O

! E .

o.

.E

|E 2000i! n.1 3i W! m UNACC TABLE

/| $ OPERA ON ACCEPTABLE i

OPERATIONj $) 5 ,

O.

i O '

; o; e: C

E 10002 ,

5-

COOLDOWNj e: o RATE

] $ ('F/hr) ;

4 < {u Op-

0 ff60 /

100

\0.0

0.0 100 200 300 - 400 500

LOWEST INDICATED RCS Tcold TEMPE URE ('F)

MATERIAL BASIS

copper content. . o.to wt %0.06 wt %)

NOT nfeet: 40*F^ - ' -IIT I(Actuel. * F)

After 13 EPPY e 1/4 T 110*FRTETe 3/4 T a "F

FIGURE 3.4-3a

UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 13 E I

YOGTLE UNITS - 1 & 2 3/4 4-32 Amendment No. 36 (Unit 1)Amendment No.16 (Unit 2)

.

..

.

2500

l2250

2000|

UnacceptableOperation

1750

;;;; 1500

bea 1250e Acceptable0- Operation

E 1000.h_E

CooldownRates 'F/hr

500

;8 9]Va0

250 RU"

00 50 100 150 200 250 300 350 400 450 500

Indicated Temperature ( F)

MATERIAL BASIS

Copper Content: Assumed - NA WT%

(Actual- 0.083 WT%)

NDT nitial: Assumed - NA 'FIRT

(Actual- 20*F)

RTNDTAt 16 EFPY: @ 1/4T - 100.7'F,t @ 3/4T - 84,1*F

Figure 3.4-3aUnit 1 Reactcr Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr)Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for InstrumentationErrors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentationand Reactor Vessel Beltline Region).

VoGTLE UNITS 1 & 2 3/4 4-32

|

-.

,

5

| kE@Lhce va t nt New PAae 3/+ y-32v1,

-

!

|UNIT 2

i i 1

CURVE APPUCABLE FOR THE'

; SERVICE PERK)O UP TO 16 EFPY.

-

!, t-

a

2000 - ('

wi E| |E' w

kM'

| t m0a >' z

$ UNACCE LE ;

OPERATI ;

g4

1 oE*

$ 1000 g

W i

E ACCE 'TANLE J-

COOLDOWN OPERATK)N '

i o I* w R ATE (*F/hr) '

1 e/ ;

] j 0< o M0 "20 -

g _- -%Z

to #; 100

\'

0.0; 0.0 6 200 300 #0 W;

LOWEST H409CATED RCS Tg TEMPERA RE (OF)

:

MATERI AL BACopper Content; Assurned 0.10 Wt %

4

( Actusi 0.05 Wt %)

RT Inttiel, med NADFNDT (A ual.50 F1

RT After 16 E FPY 1/4 T = 123*FNDT

'4,7 = 97'Fp

a

FIGURE 3.4-3b

!,

UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFa

,

i

a

VOSTLE UNITS - 1 & 2 3/4 4-323

.-. . . . . - . _ ,

:-

i-

.

2500,

|

2250 |li

2000 ||

f 1750^

Unacceptable |

Operation ;

g; 1500ToS ::

2a 1250$;- Acceptable

OperationoS 1000 ,

8e

|750

,

,

i Cooldown| Rates F/hr

l' 500 ^ .

| 2!. 0 j.gg / |

l

00 50 100 150 200 250' 300 350 400 450 500

Indicated Temperature ( F)

MATERIAL BASIS

Copper Content: Assumed - NA WT%

! (Actual- 0.05 WT%)

NDT nitial: Assumed NA'FIRT

(Actual- 50'F)

RTNDTAt 16 EFPY: @ 1/4T - 112'F@ 3/4T - 94'F

i

'

Figure 3.4-3bUnit 2 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100'F/hr)Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for InstrumentationErrors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentationand Reactor Vessel Beltline Region).

VOGTLE UNITS 1 & 2 3/4 4-32a

i

|.

,

: |J i,

- 3L:iLAC6 w im usw hA-6 e' J/+ 4- 3 7| |

UNIT 1M

N /| 1 -\ / |

! E -N /\ (242,726) / (350,726) |

y -\ i,

= - N I /'

O 700'

; - N I / :.

i ! - N / /i ! - N //

I - N / ;

e - \ //:2 000 s. ,-

| j /\/ |^

| ! - / Ai 1 F \,532)-(70g1 if

-

s'

_ /f \ft f f f f f f 1 1 1 f f f f f i f I t I i !j

50 100 150 200 250 300 360

TRTD-A IONEERED LOW MEASURED RCS TEMPER TURE (*F)

;

d

.'11

j

'

FIGURE 3.4-4a4

UNIT 1 MAXIMUM ALLOWABLE NOMINAL PORY SETPOINT;

i FOR THE COLD OVERPRESSURE PROTECTION SYSTEM

''0GTLE UNITS - 1 & 2 3/4 4-35'-

-. - . - ..

'

.

800

780

760

740-

y (212,726) (350 726)

S 720:E

! J 700' o

*| 680>,

C'

@ 660|

! og 640

f 3::S 620$@ 600'Ey 580

$560 '/ '

( 0,538) jm 540i

''

520

50050 100 150 200 250 300 350

TRTD - Auctioneered Low Measured RCS Temperature ( F)

!

Figure 3.4-4aUnit 1 Maximum Allowable Nominal PORV Setpoint forthe Cold Overpressure Protection System

VOGTLE UNITS 1 & 2 3/4 4-35

|

,

.

REPt/rce w en+ v e w 8kG c' 3/4 4- 3 fo.

UNIT 2800

\,

-\ /N /

-

'N (2a2,725) / (350,728)_

f{ '~

N /~ 700 -

s7u N /

-

! '\ I/-

~

/x

2 \ ['-

s '.N fe -

n .00

\-

/; </-

I / A-

i / \_

p

(70,505) [ \-

n. <g

\'

-

,.

/ \_

7

/ \'

50 100 150 200 250 \@B0 50

'

TRTD7'AUCTIONEERED LOW MEASURED RCS TEMPERA (*F)

FIGURE 3.4-4b

UNIT 2 MAXIMUM ALLOWABLE NOMINAL PORV SETPOINT FOR THECOLD OVERPRESSURE PROTECTION SYSTEM

V TLE UNITS - 1 & 2 3/4 4-35a

*5

800

780

760

740 (222,726) (350,726)'S 720

= |3, 700 ,

j: o* 680

|cco 660' /

I| f 640

[SR 620<] 600

/1:y 580

h560 /I1

w 540 //520 (70,516)

50050 100 150 200 250 300 350 i

TRTO - Auctioneered Low Measured RCS Temperature ( F)|

|

|| |

|

'

.

|Figure 3.4-4b

Unit 2 Maximum Allowable Nominal PORV Setpoint forthe Cold Overpressure Protection System

VOGTLE UNITS 1 & 2 3/4 4-35a

:!

_

. . . _ - - _ -. . ._

. -

!

}.*

!. .

!j REACTOR COOLANT SYSTEMi

| BASE 5

PRES 5URE/ TEMPERATURE LIMITS (Continued)2~

2. These limit lines shall be calculated periodically using methods providedbelow,

3. T.he secondary side of the steam generator must not be pressurized above200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100*F/h and200*F/h, respectively. The auxiliary spray shall not be used if thetemperature difference between the pressurizer and the auxiliary sprayfluid is greater than 625*F, and

5. System preservice hydrotests and inservice leak and hydrotests shall beperformed at pressures in accordance with the requirements of ASME Boiler'

L \ and Pressure Vessel Code, Section XI.M 4A

d q~Q The fracture toughness properties of the ferritic materials in the reactorZ vessel are determined in accordance with the NRC Standard Review Plan, ASTMM E185-82, and in accordance with additional reactor vessel reouirements. | Th;;;

;r:::rti:: " ::::rd:n:: with A;;:ndi C Of th: 1972 E r:r. . . . . . . . . . . . . . . . .

Add;nd; t: 5;;;i;r !!! ;f th; ASMC ";il;r ;nd "r;;;;re V;;;;l C;C; :ndi tnecalculation metnods described in WCAP-7924-A, " Basis for Heatup and CooldownLimit Curves," April 1975.

The heatup and co own limit curves.sho in Figures 3.4-2a an 3.4-3aare licable to Unit 1 up to 13 EFPY and a based on Westinghous |

/ develop generic curves whi were develo ed assum a 40*F initial RT$ and a copp content of 0.10 or the most limiting terial. These h ves

nh are applicabi o Unit i since its t limiting material ble B 3/4.4-la)2

has both a lower ' itial RTNDT (30*F) a lower copper cent (0.06 WT%).i e curves, however, re not applicable t nit 2, since its mos limitingmate 1 (Table B 3/4.4- has a higher initi RT (50 compared t40*F). arate heatup and idown limit curves yNdevelopedbasedo heactual mate al properties of t most limiting mater 1 for Unit 2 up to 1EFPY, The Un curves are shown Figures 3.4-2b and 4-3b.

Heatup and cooldown limit curves are calculated using the most limitingvalue of the nil-ductility reference temperature, RTNOT, at the end of theEffective Full Power Years (EFPY) of service life. The EFPY service life periodis chosen such that the limiting RT at the 1/4T location in the core region

NDTis greater than the RT f the limiting unirradiated material. The selection

NOTof such a limiting RT assures that all components in the Reactor Coolant

NDT

System will be operated conservatively in accordance with applicable Coderequirements.

V0GTLE UNITS - 1 & 2 B 3/4 4-8 Amendment No. 36 (Unit 1)Amenament No.16 (Unit 2)

. -.- - . . -. .. --. .- - - . - - . . .. . -

'.,

.

INSERT A

These properties are then evaluated in accordance with Appendix G of ASME Boiler

.

and Pressure Vessel Code, Section III, Division 1 - Appendices, " Rules for

| Construction ofNuclear Power Plant Components, Appendix G, Protection AgainstNonductile Failure," 1986 Edition and :

I!

INSERT B

The heatup and cooldown limit curves shown in Figures 3.4-2a and 3.4-3a for Unit Iand Figures 3.4-2b and 3.4-3b for Unit 2 are applicable for up to 16 EFPY and weredeveloped based on the actual material properties of the most limiting material. The

| most limiting material are shown in Table B 3/4.4-la for Unit I and Tables B 3/4.4-lb i

! for Unit 2.

I

ii

!I

it

!

L|-

1l'I

_ _ _ _ _ _ _ . _ _ _ _ = . . < - . , , ,- - - .. . 7

. - _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ . __ ____---- -_ _ - _ - _ _ _ - _ _

..

,

.

.- -

!'__ _ _ _ _ _ -- s

. _ ,,__ _ . .

.I AB_L 4.4-la

ONil 1 RI ACIOR _VI'$5t L_ IOUGHNL SS

ASML $0 I l-ll! AVI'RAGLc RI '

5 COMP AL CU Ni P N 35 MIL Nul HMWD* .

d C PJ CODE TYPE (%) @ @ 1*B Eg ( I ) (*r). (fI-tII) .

.

Closure Hea B8807-1 A533BCL.l .1 .67 .008 -50 15 15 88 |-

e- Closure Ilead Tor B8808-1 A5330Cl.1 . 14 . .010 -30 68 85Closure Head Flange 8801-1 A508Cl.2 - .70 11 20 (40 20 1 12

,

~

Vessel F lange B -1 A508CL.2 - .71 .01 0 <60 0 11Inlet Nozzle B8809- A508CL.2 - .86 .011 - <10 -20 10/Inlet Nozzle 88809-2 8CL.2 - .84 .014 -10 <50 -10 95

let Nozzle B8809-3 A50 - .82 .013 -10 < -10 11/Inl ozzle -B8809-4 A508CL.2 - .87 014 -20 <10 -20 105

Outlet le 88810-1 A508CL.2 - .82 .006 -10 <50 -i >124

Outlet Nozz 88810-2 A508CL.2 - 79 .006 -10 <50 -10 100i

Outlet Nozzle 88810-3 A508Cl.2 -. .006 -10 (50 -10* >

R Outlet Nozzle 8810-4 A508CL.2 - .80 . -10 <10 -10 /5 ;

Nozzle Shell B -1 A5338CL.1 .14 .62 .011 0 88 28 94*

i Nozzle Shell 88804- A533BCL.1 .10 .58 .006 -40 75 15 104

zzle Shell 88804-3 38CL.1 .14 .69 .013 -30 0 40 92*

In Shell B8805-1 A53 1 .08 .59 .004 0 6 0 90

Inter. 11 88805-2 A533BCL. .08 .59 .004 -10 80 40 100

Inter. Shel 88805-3 A533BCL.1 .60 .003 -20 90 36' 10/..

i Lower SheiI B8606-1 A5338CL.1 .05 59 .005 -50 80 20 .116

Lower Shell 8606-2 A533BCL.1 .05 . .009 -10 80 20 113\Lower Ehell B -3 A533BCL.1 .06 .64 . 7 -20 10 10 118 ;

Bottom Had Torus 88813- A5338CL.1 .13 .50 .009 -40 50 -10 88

ottom Head 00me B8812-1 3BCL.1 .10 .53 .009 -4 32 -28 122

In & Lower Shell Gl.43 SAW .03 .10 .00/ -80 -20 -80 *129

Verti Weld ''Seams an th

; Seam.

*0pper ShelI energy, norma i major working directions T. l

..i.

. s'

|' 15 D h ,

i 11"

k' )/. _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - - _ _ . - . .. ._- . _ _ ._- -__ __ __ _ _ _ _ _ _ __

_ . . _ . . . . .._-.._._._.._..~....__...._..._._____._.._.m.____. m.__..._.____._..__..

*.

.

!

i.

L

.

TABLE B 3/4.4-1a,

UNIT 1 REACTOR VESSEL TOUGHNESS gj

nm .

"

COMP CU NI INITIAL 16 EFPY RTNDT lil

COMPONENT CODE 1%1 L%1 B L ui El 1/4-t ( F) 3/4-t i El ]. . .

Closure Head Flange -- -- 0.70 20 --- --- !.

Vessel Flange -- -- 0.71' O --- ---

?

Intermediate Shell 88805-1 0.083 0.597 0 80.7 64.1 r

intermediate Shell* B8805-2 0.083 0.610 20 100.7 84.1 |

Intermediate Shell B8805-3 0.062 0.598 30 97.5 76.4 |

.

!' Lower Shell B8606-1 'O.053 0.593 20 77.6 57.6 i

Lower Shell B8606-2 0.057 0.600 20' 81.9 62.5 !Lower Shell B8606-3 0.067 0.623 10 80.8 60.6

Circ. Weld 101-171 0.039 0.102 -80 -21.7 -39.9 |

Long. Weld 101-124A 0.039 0.102 -80 -31.8 -48.5 ;

Long. Welt 101-124B 0.039 0.102 .-80 -30.0 -47.0'

Long. W 101-124C 0.039 0.102 -80 -30.0 -47.03'

Long. Weio 101-142A 0.039 0.102 -80 -30.0 -47.0 t

:

Long. Weld 101-1428 0.039 0.102 -80 -31.8 -48.5'' ,

Long. Weld 101-142C 0.039 0.102 -80 -30.0 -47.0

i.

!>.

* Limiting material|

|

I !!i |t

I

i-I

4

. - . _ _ . - _ . . _ . . _ . . _ _ _ _ _ _ _ _ - - - - - _ _ - - _ _ - - _ _ _ _ - _ . _ _ _ . - _ _ _ _ _ _ _ - _ _ _ _ _ - _ . _ _ _ _ - - - _ _ _ - - - - _ - - - _ - _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ - - - - _ _ _ - _ - _ _ _ - _ _ '

_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ . _ - _ _ _ . . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

-.

.

.

_ - . _ _

. xIABLE B 3/4.4-Ib N'\ xxo 'N uMIT 2 REACIOR VESSEL 100GHNE55 NS N ss '

g s

ASME AVI RAGEc- I NI?! COMP NAIERIAL C Ni P NDI N01 HMWI)"

d C3MPONENT CODE lYPE {%} {%) (*f) ( *1' ) ('r l- L8 )

[ Closure Head Gone R9-1 A5338 CL. 1 0.07 0.61 0.008 -40 -30 123

Closure Head Torus 0-1 A5338 CL. 1 0.07 0.64 0,010 -30 0 114,.

0.72 0.011. 10 to l ioClosure Head Flange R7- A508 CL. 2 -

Vessel Flange Rl-1 A508 CL. 2 - 0.81 0.011 \ 0 -60 115m

689806-1 08 CL. 2 0.07 0.84 0.010 -50s. -50 119

\ . Inlet NozzleInlet Nozzle 89806-2 A568' L. 2 0.06 0.83 0.009 -40 ''N ,-40 128~

thlet Nozzle RS-1 A508 CL. 0.09 0.87 0.008 -20 -20 141

Inlet''Nezzle RS-2 A508 CL. 2 0.08 0.85 0.009 -20 -20 . 134

Outlet Nott e R6-3 A508 CL. 2 0.69 0.011 -10 -10 122

Outlet Nozzle R6-4 A508 CL. 2 - 0.66 0.010 -10 -10 t40Outlet Nozzle 89807-3 A508 CL. 2 . 0.005 -30 -30 116-"

R Outlet Mozzle 9807-4 A508 CL. 2 - 0.64 0.010 10 10 ;132

* Nozzle Shell R3-1 A5338 CL. 1 0.20 0.67 15 0 20 19

? Nozzle Shell R3-2 \ A5338 CL. 1 0.20 0.67 0.0l3 0 40 19

y Nozzle Shell R3-3 '45338 CL. 1 0.15 0.62 0.010 IQ 60 84

CL. 1 0.06 0.64 0.009 -20 K 10 95A$33(8C-

' Intermediate She1I R4-1A533 1 0.05 0.62 0.009 -10 x10 104Intermediate Shell R4-2

Inteniihilate Shell R4-3 A5338 CL. x 0.05 0.59 0.009 0 30N 84

Lower Sheh 88825-1 A5338 CL. 1 D 45 0.59 0.006 -20 40 ''s 83

Lower Shell R8-1 A5338 CL. 1 0. g o.62 0.007 -20 40 'B/s'

Lower Shell** R8628-1 A5338 CL. 1 0.05 8 49 0.001 -20 50 85 's

0.012 -20 -20 89 H-Botton Head Torus 4tl2-1 A5338 CL. 1 0.17 0.64N '8.008 -30 -30 115 1 2)Bottom Head Dome Rll' A5338 CL. 1 0.10 0.62 n

Shell Vertical . ([g i oy ', /Intermediate & Lower Gl.60 SAW 0.07 0.13 0.007s -10 -10 141'

\ Ll(SeamsInte ' ate to Lower E3.23 SAW 0.06 0.12 0.001 -50 -30 90 (/

Shell Gt Weld 3

Seam

" Upper Shelf energy,Dr al to major working direction.** Limiting material.

. _ _ _ _ _ _ _ _ . . _ . _ _ _ _

_ _ . . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _

_ _ _ _ _ _ - _ - - _____ __ - _ _ _ _ _ - __ _-___ _-_ -- _ -___- - _ - _ _ - _ _ _ _ __ - - _ _-___-_ _ ___ _ _. _ __ - - _ -

.

-

.

TABLE B 3/4.4-1Ht

UNIT 2 REACTOR VESSEL TOUGilNESS

!COMP CU NI INITIAL 16 EFPY RTNDTCOMPONENT CODE @ @ RTndt 1/4-t (*F) 3/4-t(*F)

LD (

Closure IIcad Flange - - 0.72 10 - -

Vessel Flange -- - 0.87 -60 - -

Intermediate ShcIl R4-1 0.06 0.64 10 81 62

Intermediate Shell R4-2 0.05 0.62 10 72 54

Intermediate Shell R4-3 0.05 0.59 30 92 74

Lower Sell B8825-1 0.05 0.59 40 102 84

f-1i Lower Shell R8-1 0.06 0.62 40 111 92 ' Lt n

Lower Shell* B8628-1 0.05 0.59 50 112 94'

P,

-l'

Cire. Weld - 0.06 0.12 -30 55 31,

O'

Long. Weld - 0.07 0.13 -10 83 56;.

i

* Limiting Material,

4

4

,,;

.

l. . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _

, . . . _ _ _ _ _ _ _ _ _ _ . _ ____- ..._. ..

: .

j -.

!

!

!.

i REACTOR COOLANT SYSTEMi

!

| BASES

i.

! PRESSURE / TEMPERATURE LIMITS (Continued):

} The reactor vessel materials have been tested to determine their initialRTj NOT: the results of these tests are shown for Units 1 and 2 in Table

~

1 B 3/4.4-la and b, respectively. Reactor operation and resultant fast neutroni J (E greater than 1 MeV)-irradiation can cause an increase in the RT There-|

d ore,anadjustedreferencetemperature,baseduponthefluence,cohe.r content,y

= C an : n ::n r e content of the material in question, can be predicted using|Fi;;r: 5 2/' "-1 :nd th: ur;::t :: :: :" ^ % : r;;t:c by ::tn:q Regulatory;

Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation |Damage to Reactor Vessel Materials," Ler the Weetinenesee Cesser Trenc Curve;|

i 1 :n:r " "i;_r; O 2/4.4 2! The heatup and cooldown limit curves of Figures ;; 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) include predicted '

| f (o adjustments for this shift in RT at the end of il; (Jan l' :n: x (Unt ;:n |NOT| EFPY as well as adjustments for possible errors in the pressure and temperature I

sensinginstrumentsj p 60 p.3 and lov, ruput'gejy~

j:

| Values of MT determined in this manner may be used until the resultsNOT

j from the material surveillance program, evaluated according to ASTM E185~ are |,

: available. Capsules will be removed in accordance with the requirements of !; ASTM E185-82 and 10 CFR Part 50, Appendix H. The surveillance specimen with-i drawal schedule is shown in Table 16.3-3 of the VEGP FSAR. The lead factori represents the relationship between the fast neutron flux density at the loca-

tion of the capsule and the inner wall of the reactor vessel. Therefore, the@ results obtained from the surveillance specimens can be used to predict futurey,

$ radiation damage to the reactor vessel material by using the lead factor and' the withdrawal time of the capsule. The heatup and cooldown curves must bei

recalculated when the MT determined from the surveillance capsule exceeds iNOT

the calculated M T for the equivalent capsule radiation exposure.NOT

Allowable pressure-temperature relationships for various heatup andcooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix Gto 10 CFR Part 50, and these methods are discussed in detail in the following.paragraphs.

The general method for calculating heatup and cooldown limit curves isbased upon the principles of the linear elastic fracture mechanics (LEFM) tech-nology. In the calculation procedures a semielliptical surf ace defect with adepth of one-quarter of the wall thickness, T, and a length of 3/2T is assumedto exist at the inside of the vessel wall as well as at the outside of thevessel wall. The dimensions of this postulated crack, referred to in Appendix Gof ASME Section III as the reference flaw, amply exceed the current capaoilitiesof inservice inspection techniques. Therefore, the reactor operation limit cur-ves developed for this reference crack are conservative and provide sufficient

~

safety margins for protection against nonductile failure. To assure that theradiation emorittlement effects are accounted for in the calculation of the

V0GTLE UNITS - 1 & 2 B 3/4 4-10 Amendment No. 36 (Unit 1)Amenoment No.16 (Unit 2)

, .,

INSERT E

In addition, these curves include a pressure adjustment of 74 psig to account for thepressure difTerential between the wide range pressure transmitter and the belt lineregion.

'

I1

|

|I!1

l

j

i

i

|,

1

!

i

l

lI

|

|

|,

|

|!

I- , e -,

.. - _. -. - __

i

!- --

,

REACTOR COOLANT SYSTEM*

$^

\

j ' BASES |'

1

{ PRESSURE / TEMPERATURE LIMITS (Continued)!! be analyzed in order to assure that at any coolant temperature the lower valuei of the allowable pressure calculated for steady-state and finite heatup ratesj is obtained.

!

| The second portion of the heatup analysis concerns the calculation of; pressure-temperature limitations for the case in which a 1/4T. deep outside

surface flaw is assumed. Unlike the situation at the vessel inside surface,4

i the thermal gradients established at the outside surface during heatup produe.e| stresses which are tensile in nature and thus tend to reinforce any pressurei stresses present. These thermal stresses, of course, are dependent on bothj the rate of heatup and the time (or coolant temperature) along the heatupi ramp. Furthermore, since the thermal stresses at the outside are tensile and

increase with increasing heatup rate, a lower bound curve cannot be defined.;

i Rather, each heatup rate of interest must be analyzed on an individual basis.ij Following the generation of pressure-temperature curves for both the steady-j state and finite heatup rate situations, the final limit curves are prod ed asj follows. A composite curve is constructed based on a point-by-point c rison: of the steady-state and finite heatup rate data. At any given temperatusy, the| allowable pressure is taken to be the lesser of the three values taken fWe thei curves under consideration.

The use of the composite curve is necessary to set conservative heatuplimitations because it is possible for conditions to exist such that over the

'.course of the heatup ramp the controlling condition switches from the insideto the outside and the pressure limit must at all times be based on analysisof the most critical criterion.

Next, the composite curves for the heatup rate data and the cooldown ratedata are adjusted for possible errors in the pressure and temperature sensing

; instruments by the values indicated on the respective curves.|

| Finally, the new 10CFR50 Appendix G Rule which addresses the metal temperature( of the closure head flange and vessel flange regions is considered. This rule: states that the minimum metal temperature of the closure flange regions should be

~

th+1j at least .120*F higher than the limiting RT, for these regions when the pressure; exceeds reent of the preservice hydrostatic test pressure (621 psig for| For Unit I the minimum temperature of the closure flange.

) and v ions isJ40*F, since the limiting RT , is 20*F (see_

i Table B 3/4-4.la).104 "q,tle %;t 1 teetei, cerve ste r. er. Fisure -t.2e is| :: W::t;d by th; .n.100F",0 rule. ";_ .4;r, tra niitle %it I ceeld; , cerve <

i t h r_ 'n Ff;;r; 2 t. e is i;s,ectM to ite .... iGCFR3G ruiu. I For Unit 2, the

j minimum temperature of the closure flange and vessel flange regions is 130*F.j since the limiting RTwis 10'F (Table B 3/4-lbM The Wit 2 5::typ ;re;j s m " 9;re3.'-25 d th: :: ldr_; : r;: :hr_; in Fig;re 3.t-Ob er n;tj %_, _,. w w m. . . _ _. . _ i. n.e.n ,mr.

- ____ _, . .n suau.

1 /'

! HER E V0GTLE UNITS - 1 & 2 B 3/4 4-15.

i

._. . .__ _ . . - - - .-.-

_ .

.- ..

INSERT F

These values include margin of 10 F and 60 psig for instrumentation errors. Theheatup and cooldown curves as shown in Figures 3-4.2a and 3-4.3a for Unit I and theheatup and cooldown curves as shown in figures 3-4.2b and 3-4.3b for Unit 2 areimpacted by the new 10 CFR 50 rule.

|

||

|

|

||

- - - . . .. - - . ._- - -- - _-

1

| * " * REACTOR COOLANT SYSTEM

i

! BASES. -

ii PRESSURE / TEMPERATURE LIMITS (Continued)

| Although the pressurizer operates in temperature rcnges above those for whichthere is reason for concern of nonductile failure, operating limits are provided tos

| assure compatibility of operation with the fatigue analysis performed in accordancej with the ASME Code requirements.I

j COLD OVERPRESSURE PROTECTION SYSTEMS

$ The OPERABILITY of two PORVs, two RHR suction relief valves, a PORV and RHR SRV,or an RCS vent capable of relieving at least 670 gpa water flow at 470 psig ensures,

j that the RCS will be protected from pressure transients which could exceed the limits! of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or4 equal to 350*F. The PORVs have adequate relieving capability to protect the RCS from! overpressurization when the transient is limited to either: (1) the start of an idlei RCP with the secondary water temperature of the steam generator less than or equal toi 50*F above the RCS cold leg temperatures, or (2) the start of all three charging pumps

.

I and subsequent injection into a water-solid RCS. The RHR SRVs have adequate relieving |} capability to protect the RCS from overpressurization when the transient is limited to ;

; either: (1) the start of an idle RCP with the secondary to primary water temperaturei difference of the steam generator less than or equal to 25'F at an RCS tempera e ofi 350'F and varies linearly to 50*F at an RCS temperature of 200*F or less, or (2 thei start of all three charging pumps and subsequent injection into a water-solid A.

j combination of a PORV and an RHR SRV also provides overpressure protection for thei RCS.j The Maximus Allowed PORV Setpoint for the Cold Overpressure Protection System{ (COPS) is derived by analysis which models the performance of the COPS assumingj various mass input and heat input transients. Operation with a PORV Setpoint lessj than or equal to the maximum Setpoint ensures that the nominal M = im unn 1.nm; 16 EFPY = Udt 0 Appendix G reactor vessel NOT limits criteria will not be violated |j with consideration for a maximum possure overshoot beyond the PORV setpoint which can '

j occur as a result of time delays in signal processing and valve opening, instrumentuncertainties, and single failure. To ensure that mass and heat input transients more.

i severe than those assumed cannot occur, Technical Specifications require lockout of| all safety injection pumps while in MODES 4, 5, and 6 with the reactor vessel head in-j stalled and disallow start of an RCP if secondary temperature is more than 50*F abovei primary temperature. Additional temperature limitations are placed on the starting ofi a Reactor Coolant Pump in Specification 3.4.1.3. These limitations assure that the; RHR system remains withia its ASME design limits when the RHR relief valves are usedj to prevent RCS overpressurization.

j The Maximum- Allowed PORY Setpoint for the COPS will be updated based on theresults of examinettons of reactor vessel material irradiation surysOlance specimens'

i performed as required 'oy 10 CFR Part 50, Appendix H, and in accordance with thj-

i

schedule in Table 16.3-3 of the VEGP FSAR. !

4 3/4.4.10 STRUCTURAL INTEGRITY ;

u

i The inservico inspection and testing programs for ASME Code Class 1, 2, and 3j components Nure that the structural integrity and operational readiness of these !! components will be maintained at an acceptable level throughout the life of the plant. i

j These programs are in accordance with Section XI of the ASME Boiler and PressureVessel Code and applicable Addonda as required by 10 CFR 50.55a(g) except wherei

i specific written relief has been granted by the Commission pursuant to 10 CFR ;j 50.55a(g)(6)(i). '

ij V0GTLE UNITS - 1 & 2 B 3/4 4-16 Amendment No. 56 (Unit 1)i Amendment No. 35 (Unit 2)! !

?__ __

- . _ . .-

._ _ .

, . . .

\ '

i'

8 .

\ 1Ne .

,,

'

! ! 4'

\ pSURF 3.17 a 10''['\ N

\x2 '

p 1/4 T - 1.78 x 10 Cx

10''| | / b 's ag, '

'/ / 'N

!ss!'x'/ X 'x'

'

N 'N p 3/4 T 3.49 toesi

'! ' ''s N \ |2 *

> >, '. / fx 'N N\[ / 'N '\ N,s

N 'N'' x / 'x x' ''

'N 'x'

'x x 's,

s ,

8 'N \s5 10 N 15 20 30 36 s

- EFFEC VE FULL POWER (years) \'s

.

FIGURE B /4.4-1'N

FAST NEUTRON FL CE (E>1MeV) A.~ A FUNCTJON OF FULL POWER SERV LIFE

$-

V0 GILE JNITS - 1 & 2 B 3/4 4-11

- . . . _ . _ . _ ._. .. . . - _ . . _ .

.

.

'

<

O.-

\-

400 i i i y y ig i , , i g i y y sg i_

Upper Limit'

_

e.ses c mASE. e M4 m t D~

e as corren e' t. o 20s m t o.g . ,

\ 'N--~

x

__

Lower Limit

Nr / \]100 'x'

'N~" ~

yw _

x- egco se.e.,-mo<g , _

'e.wm c .se. . ws mto s

,,

; e.ws corren eq.e.ses ato40 x -

x _.

s -_

_, .

i l i i li i I I II11 1 !N I i 1 1 i a aa1

e s 10 '' N |'

10" 2 4 a 10 " 2 N42FAST NEUTRON FLUEPOCE (N/CM , E >l MeV) N '

r-IV \ 'x j

I -i N -

\m s

i flGURE B 3/4.

EffECT LUENCE AND COPPER CONTENT'ON IFT OF RTNDT

FOR REACT ESSELSEXPOSEDTORADIATIONA(5500F _

s

- _ - - - _ - - - - - _ _ - - - - - - - - - _ _ . _ - _ _ - - _ - _ - _ - _ - - - _ - - - - - - - - - _ - - -

>. .,

.

||

|

ENCLOSURE 4

VOGTLE ELECTRIC GENERATING PLANT, REQUEST TO REVISE TECHNICAL SPECIFICATIONS'

REVISION TO IEACTOR PRESSURE VESSEL LIMITS

EXEMPTION REOUEST

This enclosure requests, in accordance with the provisions of 10 CFR 50.12, an exemptionfrom certain requirements of 10 CFR 50.60, " Acceptance Criteria for Fracture PreventionMeasures for Light Water Nuclear Power Reactors for Normal Operation." As stated in10 CFR 50.60, " Proposed Alternatives to the Described Requirements in Appendices Gand H of this part or portions thereof may be used when an exemption is granted by theCommission under 50.12." This exemption is requested to allow the application of the

| American Society ofMechanical Engineers (ASME) Code Case N-514, " LowTemperature Overpressure Protection," in determining acceptable setpoints for VEGPUnits 1 and 2.

,Pressure / temperature (P/T) limits for low temperature overpressure protection are

| characterized by the system enabling temperature and the setpoint pressure for thepressure relieving device. Current regulatory guidelines require that the low temperatureoverpressure protection (LTOP) system must be enabled at temperatures less than orequal to R FNDT + 90 F, where RT iNDT s the adjusted reference temperature includingmargin, at the quarter thickness location. At temperature greater than RTNDT + 90 FLTOP protection need not be provided. The maximum LTOP system pressure is >

determined based on system-specific considerations, but is chosen so that the maximumpressure attained in the vessel will not exceed the P/ f limit curve defined by Appendix G

| to ASME Section III and Appendix G to 10 CFR 50.

The LTOP limits caused operational constraints by limiting the range available to theoperator to heat up and cool down the plant. The operating window through which thereactor coolant system is heated up and cooled down is determined by the difference

i

| between the maximum allowable pressure determined from ASME Section III, AppendixG, and the minimum allowable pressure determined by the differential pressure betweenreactor coolant system pressure and atmospheric pressure for the RCP seals. Aspreviously discussed, the Westinghouse methodology used to calculate LTOP setpointsdid not account for the differential pressure across the reactor core during RCP operation.

E4-1

|\

i

1

.

. _ _ _ _

,. .e

ENCLOSURE 4 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANTREQUEST TO REVISE TECHNICAL SPECIFICATIONSREVISION TO REACTOR PRESSURE VESSEL LIMITS !

EXEMPTION REOUEST

l

The existing LTOP setpoint curve is based on a plant-specific evaluation with PORVsetpoints selected within a range of allowable pressures at various temperatures. For the ;

existing curves, the 74 psi nonconservatism depletes the existing margin at approximately |

1200F for Unit I and 1450F for Unit 2.:1

ASME Code Case N-514 requires that the LTOP systems limit the maximun pressure in |

the vessel to 110 percent of the pressure determined to satisfy Appendix G limits.

|'

The ASME Working Group on Operating Plant Criteria developed code guidelines todefine LTOP limits that will avoid unnecessary operational restrictions, provide adequatemargins against failure, and reduce the potential for unnecessary activation of pressure-relieving devices used for LTOP The LTOP limits allow the pressure that may occur with

|activation of pressure-relieving devices at temperature less than 2000F to exceed the P/Tlimits, however, acceptable margins to vessel fracture are maintained during these events.The pressure vesselis protected both from LTOP events and the P/T limits in Technical

,

Specifications applicable for normal heatup and cooldown in accordance with Appendix G i

to 10 CFR 50 and Sections III and XI of the ASME code.

Some conservatism in Appendix G pressure and/or temperature curve calculations are:

1. Safety factor of 2 on the principal membrane (pressure) stresses.

2. The disregarding ofincreased mechanical properties of the vessel that accompanymaterial embrittlement (element yield strength and flow stress).

! 3. The limiting toughness is based upon a reference value (KIR), which is a lower boundof the dynamic crack initiation or arrest toughne s.

|

||

E4-2

|!

I

. _ . .

i

,o ,

9

ENCLOSURE 4 (CONTINUED)

VOGTLE ELECTRIC GENERATING PLANTREQUEST TO REVISE TECHNICAL SPECIFICATIONSREVISION TO REACTOR MESSURE VESSEL LIMITS

EXEMPTION REOUEST

4. A margin factor of 2a (sigma) applied in determining the adjusted referencetemperature (ART).

5. An assumed flaw in the wall of the reactor vessel with a depth equal to 1/4 of thethickness of the vessel wall and a length equal to 1 1/2 times the vessel wall thickness.

Bases for Exemotion

| The requested exemption to the regulations is authorized by law, will not present an undue,

risk to the public health and safety, and is consistent with the common defense andsecurity.

Georgia Power Company believes that the requested exemption meets the criteria in10 CFR 50.12 (a)(2)in that special circumstances are present. These includes:

10 CFR 50.12 (a)(2)(ii).

The application of the regulation in the particular circumstances would not serve theunderlying purpose of the rule or is not necessary to achieve the underlying purpose ofthe rule.

The basis for the LTOP setpoints is to preclude the potential for brittle failure ofreactor vessel material. ACME Code Case N-514 recognizes the conservatism of theAppendix G curves and allows establishing a setpoint that preserves the acceptablemargin of safety while maintaining operational rnargins for RCP operation at lowtemperatures and pressures. Setpoints established in accordance with Code Case N-514 will also minimize the unnecessary activation of protection system pressurerelieving devices. Therefore, establishing LTOP setpoints using Code Case N-514criteria satisfies the underlying purpose of the ASME code and regulations that nuclearpower plant systems and components are operated to ensure an acceptable level ofsafety and environment impact.

E4-3

:s. ,, ,

|

!|

|

$

| ENCLOSURE 4 (CONTINUED):

VOGTLE ELECTRIC GENERATING PLANTREQUEST TO REVISE TECHNICAL SPECIFICATIONSREVISION TO REACTOR PRESSURE VESSEL LIMITS

|

EXEMPTION REOUEST

Based on the above, the application of the regulation in the particular circumstances is notnecessary to achieve the underlying purpose of the rule.

10 CFR 50.12 (1)(2)(iii)*

Compliance would result in undue hardship or other costs that are significantly in

| excess of those contemplated when the regulation was adopted, or that arej significantly in excess of those incurred by others similarly situated.|

Administrative restrictions on RCP operations while at low RCS temperatures wouldresult in an unnecessary burden in that a long delay would be required to ensureminimum RCS temperature before starting the RCPs. The application of the code casewill minimize any delays in starting the plant. The proposed LTOP P/T limits providean acceptable margin against crack initiation and failure in reactor vessels. The P/T

j limits do not significantly change the likelihood of vessel failure associated with normal

| heatup and cooldown limits. The LTOP limits also reduce the potential forunnecessary activation of pressure-relieving devices. The new LTOP limits provideboth economic and safety benefit.

Therefore, compliance would result in undue hardship or other costs that aresignificantly in excess of those contemplated when the regulation was adopted.

Conclusion

ASME Code Case N-514 allows setting the LTOP setpoints such that the Appendix Gcurves are not exceeded by more than 10 percent. At this time, GPC only wants to usethis code case at temperature less than 2000F. The ASME code committee has concludedthe LTOP guidelines provide acceptable margin against crack initiation and failure inreactor vessel and will reduce the potential for unnecessary activation of protection systempressure-relieving desdces. Consequently, the proposed LTOP setpoints provide bothoperational and safety benefits with no adverse safety or environmental impact. GPCbelieves that use of Code Case N-514 provides an acceptable level of quality and safety.

E4-4

:!

*,

f. o,

|

:

|

ENCLOSURE 4 (CONTINUED)

j VOGTLE ELECTRIC GENERATING PLANT| REQUEST TO REVISE TECHNICAL SPECIFICATIONS

REVISION TO REACTOR PRESSURE VESSEL LIMITS

EXEMPTION REOUEST

|!

Compliance with the current approved Pff limits would result in economic hardship toGPC without a compensating increase in he level of quality or safety. Georgia PowerCompany requests that this exemption from certain requirement of 10 CFR 50.60 beprocessed by Maren 31,1995.

|

|

I

|

|

|

.

!

!

, .

E4-5

l

i

i

I

|

|