ML20062C620

Amends 147 & 143 to Licenses DPR-32 & DPR-37,respectively, Revising Heatup & Cooldown Curves to Be Effective for 15 EFPY of Operation

ML20062C620
Person / Time
Site:	Surry
Issue date:	10/24/1990
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062C621	List: ML20062C620 ML20062C623 ML20062C629
References
NUDOCS 9011010236
Download: ML20062C620 (21)

Text

_.

- - - - * - ~ ' ' ~ ~ ~ - ~ ~

~~^

~ ^ ' ~ ~ ~ ^ ~ ~

4

'[

j UNITED STATES yx

')

NUCLEAR REGULATORY COMMISSION j

1 wass NoroN,0. C. 20M5

,o VIRGINIA ELECTRIC AND POWER COMPANY DOCKET N0. 50-280 SURRY POWER STATION. UNIT NO. 1 AMENDHENT TO FACILITY OPERATING LICENSE 4

Amendment No. 147 License No. DPR-32 1.

The Nuclear Regulatory Comission (the Comission) has found that:

I A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated January 29, 1988, as supplemented-February 20, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii)'that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in a ance with 10 CFR Part 51 of the Comission's regulations anc' applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:

P

-=

. m

,.__-.w--

y

y. e

i J (B) Technical Specifications The Technical Specifications contained in Appendix A, 1

as revised through Amendment No.147. are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days, j

FOR THE NUCLEAR REGULATORY COMMISSION I

1 He ert N. Berkow, Director

.i Project Directorate 11-2 Division of Reactor Projects - I/II l

Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications l

l Date of Issuance: October 24, 1190 l

l l

i l

a

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

. _..~.--

i a nso

, #s uq),

UNITED STATES -

m NUCLEAR REGULATORY CDMMISSION a

WASHING TON, D. C. 20555

?

s, '.... j i

i VIRGINIA ELECTRIC AND POWER COMPANY' DOCKET NO. 50 281 s

SURRY POWER STATION UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.143 License No. DPR-37 1.

The Nuclear Regulatory Commission (the Commission)'has found that':

}

A.

The application for amendment by Virginia Electric.and Power Company.(the licensee) dated January 29, 1988, as' supplemented February 20, 1989, complies with the standards and requirements of the Atemic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of ~the Commission; C.

There is reasonable assurance (1)'that the activities. authorized by this amendment can be conducted without endangering;the health and safety of the o'ub ic, and (ii) that such activities'will be conducted in comp!. nce with the Co61 mission's regulations; D.

Theissuanceofthisamendmentwill'notbe:inimibaltothecommon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part' 1

51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

A cordingly, the license is amended by changesL to the; Technical

{

Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License.

No. DPR-37 is hereby amended to read as follows:

r 1--

- - - - - ~ - - -

e um.,

i i

l (B) Technical. Specifications I

l The Technical Specifications contained in Appendix A, l

as revised through Amendment No.143.. are-hereby incorporated in the license. The licensee shall

(

operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date'of its issuance and shall-be implemented within 30 days, l

FOR THE NUCLEAR REGULATORY COMMISSION

. W H rbert'N. Berkow, Director; Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 24, 1990 l

l l

I r

6

(

ATTACHMENT TO LICENSE-AMENDMENT AMENDMENT NO.

I47 FACILITY OPERATING LTCENSE NO. DPR-32 AMENDMENT NO.

143 FACILITY OPERATINC LICENSE NO.-DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:

Remove Pages insert Pages-TS 3.1-6 TS 3.1-6 TS 3.1-7 TS 3.1-7 TS 3.1-8 TS 3.1-8 TS 3.1 TS 3.1-9 TS 3.1-10 TS 3.1-10 TS 3.1-11 TS 3.1-11 TS-3.1-12 TS 3.1-12 i

TS 3.1-23 TS 3.1-23 TS 3.1-25 TS 3.1-25 Figure 3.1-1 Figure 3.1-1 Figure 3.1-2 Figure 3.1-2 Figure'3.1-3 Figure 3.1-3

' Figure 3.1-4 Table 3.1-1 Table 3.1-2 TS 4.2-4 TS~4.2-4 i.

r r

l i

'l

-l l

l

TS 3.1 6 B.

HEATUP AND COOLDOWN SDecification 1.

Unit i and Unit 2 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the l

pressurizer) shall be limited ~ in-accordance with TS Figures j

3.1-1 and 3.1-2.

Heatup:

Figure 3.1 1 may be used for heatup rates of up to 40'F/hr.

Cooldown:

Allowable combinations of pressure an'd temperature for specific cooldown rates are below and to the right of the.. limit lines-as shown in TS Figure 3.1 2.

This rate shall-not. exceed 100'F/hr.

Cooldown rates between-those -shown can be obtained-by interpolation between the curves on Figure 3.1-2.

l Core Operation:

During operation where the reactor core is in a critical condition (except for low level physics tests), vessel metal and fluid temperature shall be maintained above the reactor core criticality limits specified-in 10 CFR 50 Appendix G.

The-reactor shall not be made critical when the reactor coolant i

temperature is below 522'F as specified in T.S. 3.1.E.

\\

2.

.The secondary. side of the steam g'enerator must not be pressurized above 200 psig if the temperature of the vessel -is-below 70'F.

l l

l i

Amendment Nos. -l47 and'I43 i

e

+

~ TS 3.1-7 i

l 3.

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr. and 200*F/hr., respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320'F.

Basis i

The temperature and pressure changes during heatup and cooldown; are limited to be consistent with the, requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G.

1 1)

The reactor coolant temperature and pressure and system-heatup and cooldown. rates (with the exception-of the pressurizer) shall be limited in accordance with Figures 3.1-1 and 3.1-2.

E a)

Allowable combinations of pressure.and temperature.for specific.

l temperature. change rates are below -and to.the right of the limit lines shown. Limit lines for cooldown. rates 1between l

those presented may be obtained by interpolation.

I b)

Figures 3.1-1 and 3.1-2 define limits to assure' prevention of non-ductile failure only.

For normal operation, other inherent plant characteristics, e.g., pump heat'ad'dition and pressurizerf j

heater capacity, may limir the heatup and'cooldown rates that i

can be achieved over certain pressure-temperature-ranges.

2)

These limit lines shall be calculated: periodically using methods l

provided below.

i e

3)

The secondary side of the steam generator must not be. pressurized abeve 200 psig if theLtemperature of the steam generator is below 70'F.

)

L Amendment Nos.

147 and : 143 :

5 4

e

TS 3.1-8 4)

The pressurizer heatup and cooldown rates shall not exceed 100*F/hr.

and 200*T/hr. respectively.

The spray shall not be used if the e

temperature difference between the pressurizer and the spray fluid is greater than 320'r.

Although the pressurizer operates in temperature ranges above those i

for which there is reason _for concern of non-ductile failure.

operating limits are provided'to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

5)

System preservice hydrotests-and in-service-leak and hydrotesta shall be performed at pressures in accordance with the recuirewents-of ASME Boiler and Pressure Vessel Code,Section XI accord!.g t o the leak test limit line shown in Figure _3.1-1.

5 6)

The reactor shall not be made critical when the reactor coolant temperature is below 522'F in accordance with-Technical:

Specification 3.1.E.

The fracture toughness properties'of the ferritic materials in the reactor vessel are determined in accordance1vith the NRC Standard Review Plan, ASTM E185-82, and in accordance with additional reactor vessel requirements.

These properties-are then: evaluated _ in accordance with-Appendix G to Section III of the,-ASME Boiler.ar.d f

Pressdre Vessel Code.

I 4

1 Amendment Nos. 147 and 143 4

i

~

TS 3.1 9 Heatup and cooldown limit curves.are calculated using the most limiting value of the nil ductility reference temperature, j

RTNDT, at the end of (15) Effective Full Power Years of service life.

The (15) EFPY service life period is chosen such-that the limiting RTNDT at the 1/4T location in the core region is greater than-the RT of the limiting unirradiated material.

NDT The selection of such a -limiting RT assures that a l l --

NDT components in the Reactor Coolant System will be; operated i

conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine -

their initial RTNDT; the results of these tests 'are shown in Tables 3.1 1 (Unit 1) and 3.1-2 (Unit 2).

Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation-can cause an -increase -in the RT Therefore, an adjusted NDT.

reference temperature, based upon the copper and nickel content of the material and the fluence, can be predicted using. Figures 3.1-3 (Unit 1)' and 3.1-4 (Unit -2)l and the_ recommendations of ~

Regulatory Guide 1.99,-Revision 2 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

The heatup and cooldown limit curves of Figures 3.1-1"and 3.1 include predicted adjustments - for this shift 'in' RTNDT at. the end of (15) EFPY (as well as adjustments for-possible errors in the pressure and temperature sensing instruments).

i Capsules will be removed in accordance with the requirements of ASTM E185 82 and 10 CFR 50, Appendix H.

.The surveillance specimen withdrawal schedule is shown in the UFSAR.

The.heatup 1

and cooldown curves must be recalculated when.. the ARTNDT' determined from the surveillance capsule exceeds the calculated ART for the equivalent capsule radiation ' exposure, or when NDT the service period ' exceeds 15 EFPY prior to a scheduled refueling outage.

1 Allowable pressure-temperature relationships.for various heatup and cooldown. rates are calculated using. methods derived from '

4 Appendix G2 iniSection :111 of the ASME Boiler; and Pressure Vessel Code as required by Appendix G to.10'CFR rtrt 50.

, Amendment Nos.-147.'and:143' n.

TS 3.1-10 The general method for calculating heatup and cooldown limit curves l

is based upon the principles of the linear elastic fracture mechanies (1.EFM) technology.

In the calculation procedures la semi-elliptical surface defect with a deoth of one-quarter of the j

wall thickness, T and a length of 3/2T is assumed to exist:at the inside of the vessel _vall as well as at the outside of the vessel wall.

The dimensions of_ this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capcbilities of inservice inspection' techniques.

I Therefore,-the reactor operation limit curves developed for this reference crack are conservative.and provide sufficient safety margins for protection against non-ductile failure. To assure,that the radiation embrittlement effects are accounted for 'in the calculation of the limit curves, the most limiting value of the. nil ductility-reference temperture, RTg is used and this' includes the radiation-induced shift, ART

. corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for ! calculating the. allowable limit curves for l

various heatup and cooldown rates specifies thats the total stress intensity' factor, K, for the combined thermal and pressure stresses 7

at any time during heatup or.cooldown cannot be _ greater _than the l

reference stress intensity factor,' K

****1 ***P'#**"#' **

IR' that time. K is btained from the' reference fracture toughness IR curve, defined in Appendix G to the ASME Code. The K cu m s IR given ty the equationt 1

K

• 26.78 + 1.223 exp [0.0145(T-RTg +.160)}

(1)

IR where K is t e re erence stress intensity. factor as-a function 'of IR the metal. temperature T.

and the matal nil ductility reference temperature RT

Thus, the-governing equation

.for the g.

heatup-cooldown analysis is defined.in Appendix _G of the ASME Code l

as follows:

CKg+K7g IR sK where, K is the' stress intensity' factor : caused by membrance 73 (pressure) stress.

i Amendment Nos. 147 and 143

)

. - _ _ _ ~ _.

TS 3.1 11

[

Kit is the stress intensity factor caused by the thermal gradients l

K gg is provided by the code as a function of-temperature relative to the RT f the material.

NOT L

C = 2.0 for level A and B service limits, and C

1.5 for inservice hydrostatic and leak test

=

operations.

l At any time during the heatup or cooldown-transient,. Kgg is determined by the metal temperature at the. tip 'of the postulated flaw, the appropriate value for RT and the

NOT, reference fracture toughness > curve.

.The thermal stresses resulting.from temperature gradients through the vessel wall are calculated and then the-corresponding thermal stress intensity factor, K for the reference flaw is computed.

IT, From Equation ~(2)' the pressure stress -intensity factors are i

obtained and,- from

these, the allowable pressures are calculated, i

The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the -inside or outside wall controlling, for any i

heatup rate up to 40*F per hour. -The cooldown' limit curves of Figure 3.12 are composite curves which were prepared based upon the same. type analysis with the exception -that _the controlling location is always the -inside ' wall where. the.

cooldown-thermal gradients tend to produce ' tensile -stresses while producing compressive stresses at'the outside wall.

The heatup and ceoidown curves were prepared based upon the most limiting valut of. the predicted adjusted reference temperature at the end of.5 EFPY.

The edjusted reference temperature was calculated ustr.g results' from a-capsule removed after the:

eighth cycle.

The results Lare documented in WCAP 11415 and.

WCAP 114921 for Unit 1 and WCAP -11499 and WCAP 11505 for U 2.

q J

Amendment Nos.147 and Ll43'

t TS 3.1223 References (1) FSAR 4.2 l

(2)

FSAP 9.2 G.

Reactor Coolant System Overoressure Mitication Soecification 1.

The Reactor Coolant -system overpressure mitigating system-l shall be operable as described below, i

(

Whenever ' the reactor coolant average temperature is a.

greater than 350'F, a bubble-shall exist in the pressurizer with the 'necessary sprays and heaters-

operable, b.-

Whenever the reactor coolant average temperature is s 350*F and the reactor vessel head is bolted:

(1) A maximum of one charging pump operable (2) Two charging pumps shall be demonstrated I

inoperable at least once per 12 ' hours by.

verifying the motor circuit breakers have been -

removed from-their power supply or, the l

benchboard control switch is

.in, the

" PULL-TO-LOCK" position.

(3)- Two operable Power _- Operated' Relief _ Valves-(PORV's) with a lift setting of s 385 psig, or (4) A bubble in the' pressurizer with. a maximum pressurizer narrow range level of 337..; After a period of ' 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, two PORV's: must also be operable, or (5) The -Reactor Coolant system vented through : one opened PORV, or.an equival_ent size opening.

2.

The requirements ;of Specification 3.1.G.I.b' may' be modified as follows:

One PORV may be inoperable for a period not to exceed a.

7. days.

If the inoperable PORV' is snot restored to

.i operable status within 7 days, then depressurize the-RCS and open one PORV within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

. Amendment Nos. I47 and 143

..-.it

1 TS 3-1,25 capability to protect the Reactor Vessel from overpressurization when the.

transient is limited to either (1) the start of an idle Reactor Coolant Pump with the secondary water temperature of a steam generator' s 50'F above the RCS cold leg temperature or (2) the start of a charging pump.

and its injection into a water solid RCS.

The limitation for a maximum of one charging pump allowe'd operable and the surveillance required to verify that :two charging pumps aref i

inoperable below 350'F provide assurance that a mass addition pressure transient-can be relieved by the. operation of ' a single p0RV. or f

equivalent.

A maximum pressurizer narrow range level of 33% has-been selected.to[l' provide sufficient time, approximate 1y'10 minutes, for operator response in case of a malfunction resulting in maximum charging-flow from; one' charging pump (530 gpm). Operator action would be initiated by at least two alarms that would occur between the normal operating level and the maximum allowable level (33%). When both PORVs>are inoperable and'it is [

~

impossible to manually open at least one PORV. additional 1 administrative-

~

controls shall be implemented to prevent a pressure transient that would exceed the limits of Appendix G to 10 CFR'Part 50.

The requirements of this specification are only applicable when the Reactor Vessel-head is bolted. When th's Reactor Vessel head is unbolted, a RCS pressure of_< 100 psig will lift the head. -thereby. creating-a relieving capability equivalent to at least one PORV.

l s

.4 l

Amendment Nos.147 and 143

3,000 l

)'ggg.!.Rillit PRPflif Illil CIITRllE BIGIR: CIEWIBilt El

-i-tIPPit Willl:1.11 elI

}8ggg.i.llEl Mi!II: 1.N BI1

11111E Ildt: 1in tildt f!!!IIUP!:gilili,IN.8 og I 3/ll,194.1 det I 2,{QQ.i

_ 2,200

,leetTestlisil l

l m 2,000' g'

7

=1 UNACCEPTABl[ OPERATION a,800 t

E 1,600 l l

=-

1 I l

)

j 3,

U.l.I.' 9.d I/@

/

j l'7gg :

Validlen[IPT

[

l 1

'A,000 :

/

ACC[PTAll{OP[RAll0N 800i j

600l

/

s 400 !

200l i

0 0 20. 50 10 110 140 170 200.130 260 290 320 350 380 410 441 470 500 530 ColdlegTemperature;degf[

figure 3.1-1 RCS Heatup Limitations.

Amendment Nos.147 and 143

i 4

. tees,

i, Inside Surface l

1/4T f

ista 3/4T l

g M

5 i

9 W

W

'r g

5.i.

si W

O ANIMUTH i

l t.

4487 i

0 8

-4 6-8 10 - 18 14 - to 18 Os 5 - te M

as '30 32 l

SERWICE LFE ftFFtCTM FEL pgge 1830$1 l,

?

r i

l Figure 3.1-3

. Fast Neutron Fluence (E > 1 kV) as s' Function of Full Power i

ServiceLife(EFPY)forSurryUnit1 1

l 4

i l

Amendment Nos. 147 and 143

i i

1 i

1 I

t#[

IR

/4 T-teth i,

3/4 T "G

i S

6" i

g 5

b

-a 2

teth

((

O' Azimuth t

te#

8 khas as se as a Ngure 3.1-4 Fast Neutron Fluence (E > 1 WeV) as a Function of Full Power Service Life (EFPY) for Surry Unit 2 Amendment Nos.147 and 143 I

l

TABLE 3.1-1 UNIT I REACTOR VESSEL TOUCHNESS DATA (UNIRRADIATED)

(b)

UPPER SHFI,F HEAT OR, MATERIAL Cu Ni P

NDT NDT ENER(W MATERIAL CODE NO.

SPEC. NO.

(%)

(%)

(%)

(*F)

(*F)

(FT 1.8)

Closure head done C4315-2 A53313 Cl. 1

.14

.59

.011 0

0 75 Head flange FV-1894 A508 C1. 2-

.13

.64

.010 10

• 10 125 Vessel flange W-1870 A508 C1. 2

.10

.65

.009 10(*}

10 74-

.87

.007 60

• 60 64 Inlet nozzle 9-5078 A508 C1. 2 Inlet nozzle 9-4819 A508 C1. 2

.84

.008 60 *)

60 68 Inlet nozzle 9-4787 A508 C1. 2

.85

.007 60 *k 60 64 t'%tlet nozzle 9-4762 A508 C1. 2

.83

.007

'60 *}

I 60 85 Outlet ' nozzle 9-4788 A508 C1. 2

.84

.007 60(*

60 72 60 *)

I Outlet nozzle 9-4825

-A508 C1. 2

.85

.008 60 68 Upper shell 122VIO9 A508 C1. 2

.07

.74

.010 40 40 83 f

Intermediate-shell C4326-1' A533B C1. 1

..II

.55-

.008 10 10 I1S *I I

Intermediate shell C4326-2 A533B C1. 1

.11

.55

.008 0

0 93 Lower shell

'C4415-1 A533B CI. 1

.Il

.50

.014-20 20 103(*}

[

.R Lower shell C4415-2 A533B C1. 1

.11

.50

.014.

0 0

80 a-

-Bottom head ring 123T338.

A508 C1.,2

.69

.020 50 50 86

?

}'

h.

Bottom done

-C4315-3.

A533B C1.:I

.14

.59-

.011 0

0 85 a

Inter. & lower shell 8T1554 & Linde 80 flux

.18

.63

.014 0'*

O N/A I

vertical weld seam-

[

L1, L3,~& L4 a

w b

(L:

l TABLE 3.1-1 (Continued)

UNIT I REACTOR VESSEL TOUCHNESS DATA (UNIRRADIATED)

NHwD UPPER SIFT.F HEAT OR,

MATERIAL Cu Ni P

NDT NDT ENERGY MATERIAL CODE NO.

SPEC. NO.

(I)

(%)

(%)

(*F)

(*F)

(FT LH)

Lower shell vertical 299L44 & Linde 80 flux

.35

.67

.014 0*

O 70 "

weld sean. L2 O *)

I Inter. to' lower 72445 s Linde 80 flux

.21

.58

.016

-6 N/A

.shell girth seam t

t NOTES:.

.(a)

Estimated per N3C, standard review plan, NUREC-0800, Section MTEB 5-2 I,

~

(b) Normal to major working direction - estimated per NRC standard review plan. NUREG-0800. Section MTER 5-2 i

(c) Actual-values (d) Reactor Vessel'-Fabricator Certified Test Reports

-i a

a.

j..

a w

j z __

TABLE 3.1-7 UNIT 2. REACTOR VESSEL TOUCHNESS DATA (UNIRRADIATED)

,3(b)

UPPER SilFl.t' HEAT Olt MATERIAL Cu Ni P

NDT NDT ENERGY MATERIAL CODE NO.

SPEC. NO.

(I)

(I)

(%)

(*F)

(*F)

(FT !.8)

Closure head done C4361-2 A533B Cl. 1

.15

.52

.010

-20 7

81 Head flange ZV-3475 A508 C1. 2

.11

.60

.011

<10 *

<10 129 Vessel flange; ZV-3476 A508 Cl. 2

.10

.64

.013

-65 *

-65 129 Inlet-nozzle 9-4815 A508 C1. 2

.87

.008 60

• 60 66 Inlet nozzle 9-5104 A508 C1.-2

.84

.006 60(*}

60 73 Inlet nozzle 9-5205 A508 C1. 2

.86

.007 60

• 60 66 Outlet nozzle 9-4825 A508 C1. 2

.85

.009 60 *}

~

60 74 Outlet ~ nozzle 9-5086 A508 C1. 2-

.86

. 009 60

• 60 79-Outlet nozz1e 9-5086 A508 C1. 2

.87

.01I

~60

• 60 73 Upper shell 123v303 A508 C1. 2

.09

.73

.010 30 30 103 Intermedicte shell C4208-2 A533B C1. I

.15 55

.008

-30 94 f

94 *I Intermediate shell C4339-1 A533B C1. 1

.11

.54

.012

-10 11 k'

Lower shell C4331-2 A533B C1. 1

.12

.60

.009

-10 10 84 IE Lower shell C4339-2

.A533B C1. 1

.11

.54

.012

-20 10 93 R3 Bottom head ring 123T321 A508 C1. 2

.71'

.010 10 10 101 g

Bottom dome C4361-3 '

A533B C1. 1

.15.

.52

.010

-20

-15 80 Intermediate shell 72445 & Linde 80 flux-

.21

.59

.016

-6 70 vertical weld seam Lot 8579

}

g L3, & L4

~

o.

iI-Lower shell vertical welds

.s

~ C Sean L1 (ID 63%)'

8T1762 & Linde 80

. 2 9,

.55

.015

-6 70 Seam L2 (100%)

flux 8579 Sean L1 (37%)

871762 & Linde 80 flux 8632

.29

.55

.010

-6

'70 l

TABLE 3.1-2 (Continued)

UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

(10 UPPER SitF1F HEAT OR MATERIAL Cu Ni-P NDT NDY ENERGY MATERIAL CODE NO.

SPEC. NO.

(%)

(%)

(%)

(* F)

(*F)

(FT I.R)

Inter. to lower

-0227 Crau Lo Flux LW320

.19

.56 017 0*

O 90 shell girth eeam NOTES:

r (a) Estimated per NRC standard review plan, NUREG-0803, Section MTER 5-2.

(b) Normal to major working direction - estimated per NRC standard review plan, NUREC 'J800. Section MTER 5-2 (c) Actual value based on surveillance tests normal to the major working direction ag 3'

~

3w w-

.a.-

.~.

~ ~.

l TMWE 4.2-1 SBCFION A.

Isisty 2Jesum DRSEECTICIES Hegaired Regained -

l 1

Itan Examinatim hiriatim 11entative Inspection Ikt,_

Aram h

tharinga 10-Y-Ir*arval g

.i r

1.1 DELETED L

4 i

4

.i l

!i i

1.2

.Inw Head SIS Vimaal Dion-agplicable l'

Piping ImatM This pipe diall te vimaally in valve pit inspected at endt retteeling ainst w,

1 stata 1:

1 year -.

,---h to 1 year effective ft:11 panser operation.

gente 2:- The ruotants dtained fram these an==mismations sinall be aaned to tpdate FW,3.1-1 as regained. -

h

~i et.

- 2 O

i l

m Sim LI A

..a PJ

, Aw h

~

. _ ~. _

_