ML18152B359

Revised Tech Specs Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in UFSAR & Includes Ref to Applicable UFSAR Section

ML18152B359
Person / Time
Site:	Surry
Issue date:	08/23/1999
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18152B360	List: ML18152B359 ML18152B385
References
NUDOCS 9908270146
Download: ML18152B359 (11)

Text

e e ATTACHMENT 2 MARK-UP OF TECHNICAL SPECIFICATION BASIS VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 9908270146 990823 PDR ADOCK 05000280 P* PDR

e TS 3.1-9 12-28-95 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of 28.8 Effective Full Power Years (EFPY) and 29.4 EFPY for Units 1 and 2, respectively. The most limiting value of RTNDT (228.4°F) occurs at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material.,

This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to de~mH~ffli!lrlf-fnm~---.->...__./

RTNor; the results of these tests are presented in cepoFt 8AW 2222, "Response te ClesuFe Letters to NRG Generic Letter 92 01, Revision 1,*

dated June, 1994 and are reproduced in Tables 3.1 1 and 3.1 2. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation

- can cause an increase in the RTNDT* Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with 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 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively {as well as adjustments for location of the pressure sensing instrument).

Surveillance 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 and cooldown curves must be recalculated when the ~RTNDT determined from the surveillance capsule exceeds the calculated

~RTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EFPY for Units 1 and 2, respectively, prior to a scheduled refueling outage.

Amendment Nos. 207 ane 207

---~ - - - - - - ~ -

r. I TS 3.1-12 12=28 95 1<ef~~enc:e.s Cl) l.tFSAR, "Sech'of\ 4 l.l ,~es1~"1 ~csc:s

• cTl-f IS PAGE MAS BEEN INTENTIONALLY DELETED e..../

Amendment Nos. 207 aAEI 207

TABLE 3.1-1 UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)(d)

NMwo(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY MATERIAL CODE NO, SPEC. NO. L%l L%l Cfl w IFIW)

C4315-2 A53313 Cl. 1 .14 .59 0 0 75 FV-1894 A508 Cl. 2 .13 .64 1o(a) 1o(a) 10 e

Vessel flange FV-1870 A508 Cl. 2 .10 .65 Inlet nozzle Inlet nozzle A508 Cl. 2 A508 Cl. 2

.87

.84 50(8) 60 60 60 64 68 64 d

~

Inlet nozzle 9-4787 Cl. 2 (\)

n,...

.83 eo(a) 60 85 (i)

Outlet nozzle 9-4762

.84 50(a) 60 72 w

Outlet nozzle 9-4788

.85 50(a) 60 68 Outlet nozzle 9-4825

)::,,

3 fl) a.

3 fl) r+

Upper shell lntennedlate shell lntennediate shell 122V109 2 A533B Cl. 1 A533B Cl. 1

.11

.11

.74 40 10 40 10 0

83 115(c) 94

.11 .50 103(C) 2 0 Lower shell C4415-1 A533B Cl. 1 Ill C4415-2 A533B Cl. 1 .11 .50 0 83 .....

N

--i Vl N. I 0

123T338 A508 Cl. 2 .69 50 50 N 00 I

......w I

Bottom dome C4315-3 A533B Cl. 1 .14 .59 0 0 '° C.T1 N O'I

.18 .63 o(a) -5 Inter. & lower shell 8T1554 & Linde 80 flux vertical weld seam L1, L3, & L4

TABLE 3.1-1 (Contiooed)

UNIT 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRAOIATEO)(d)

NMWO(b)

UPPER SHELF HEAT OR MATERIAL Cu Ni ENERGY MATERIAL CCDE HO. SeEC, MC. CYal CYal (FTLB) 299L<< & Linde 80 llux .35 .68 -7 70/EMA(e) d Inter. to lower shell girth seam 72445 & Linde 80 llux .21 o(a) -5 n(a)1EMA(e) -

(Q

/II

.+-

fJ Upper shell lo Inter. .10 o(a) 0 EMA(e) t--1 shell girth seam NOTES:

rD

, (a) Estimated per NRC standard review plan, NURE a.

3 rD (b) Nonnal to major working direction - estimal rt-ff (c) Actual values Ill

,N

• a Ill a.

(e) The approved equiva G.

margins analysis In the Topical Reports BAW-2192PA and BAW-2178PA demonstrates co iance wlh lhe requiremen1s 1I

~

1--'

I.O I N CJ"IN 0 ........

TABLE 3.1-2 UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

NMwo(b)

UPPER SHE HEAT OR MATERIAL Cu Ni TNDT RTNDT ENERGY CODE NO. SPEC, NO, 00 00 ro w C4361-2 A533B Cl. 1 .15 .52 -20 7 Head flange ZV-3475 A508 Cl. 2 .11 .60 <1o(a) 129 Vessel flange A508 Cl. 2 .10 .64 -65 129 r-Inlet nozzle A508 Cl. 2 .87 60 66 A508 Cl. 2 .6()(8) 60 73 Inlet nozzle 9-5205 2 .86 so(a) 60 66 Inlet nozzle 9-4825 .85 so(a) 60 74 t""

Outlet nozzle L-1 9-5086 .86 so(a) 60 79 Outlet nozzle 9-5086 .87 so(a) 60 73 Outlet nozzle Upper shell 123V303 30 30 104 lntennediate shell A533B Cl. 1 .12 -10 -10 84

)>

3 lntennedlate shell A533B Cl. 1 .11 .54 -20 83 R)

I

a. A533B Cl. 1 .15 .55 94 3 Lower shell C4208-2 R)

I rl' A533B Cl. 1 .11 .54 -10 105(c)

Lower shell C4339-1

z ..... -t

o V,

123T321 A508 Cl. 2 .71 10 .. 10 N V'I I

NW 00

• A533B Cl. 1 .15 .52 -20 -15 I .....

C4361-3 I.O I U'1 N 00 72445 & Linde 80 fkJx .21 .59 -5 lntennedlate shell vertical weld seams Lot 8579 N

0

.._,, L3 (100%), L4 (0050%)

.20 .55 -5 EMA(d)

L4 (1050%) 8T1762 & Linde 80 lkJx 8597

• ~*

TABLE 3.1-2 (Continued)

UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATEO) wo(b)

UPPER SHELF HEAT OR MATERIAL Cu NI ENERGY CODE NO, SPEC, NO, 1%} 1%} (FTW)

Lower II veftical we Seaml2(1 ) 8T1762 & Linde 80 flux 8597 .20 -5 EMA(d)

Seam L 1 ( 100%) 8T1762 & Linde 80 flux 8597 .20 -5 EMA(d)

Seam l2 (0037%) 8T1762 & Linde 80 flux 8632 .20 -5 EMA(d)

Inter. to lower .56 o(a) 0 90(C)/EMA(d) shell girth seam Upper shell to Inter. .35 .10 o(a) 0 EMA(d) shell girth seam

(

NOTES:

):a, 3 (a) Estimated per NRC standard review plan 111 0..

3 111 (b) Normal to ma;or working direction - stlmated per NRC standard review plan, NUREG-rl' (c) Actual value based on surv.

• ance tests normal to the ma;or working direction

z 0

Ill nt margins analysis In the Topical Reports BAW-2192PA and BAW-2178PA demons les compliance with the requirements of

N Al'V'liArVIIX G.

0

...... -4 V,

. Ill N,

w 0..

I._,

NI CXIN I l,O

\0 U'I

e ATTACHMENT 3 REVISED TECHNICAL SPECIFICATION BASIS VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2

TS 3.1-9 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 28.8 Effective Full Power Years (EFPY) and 29 .4 EFPY for Units 1 and 2, respectively. The most limiting value of RT NDT (228.4°F) occurs at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to~lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RT NDT of the limiting unirradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than 1 MEV} irradiation can cause an increase in the RTNDT* Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculate~ in accordance with 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 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively (as well as adjus\ments for location of the pressure sensing instrument).

Surveillance 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 and cooldown curves must be recalculated when the

~RTNDT determined from the surveillance capsule exceeds the calculated ~RTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 28.8 EFPY or 29.4 EFPY for Units 1 and 2, respectively, prior to a scheduled refueling outage.

Amendment Nos.

e TS 3.1-12 References (1) UFSAR, Section 4.1, Design Bases Amendment Nos.

e e TS 3.1-26 Pages TS 3.1-26 through TS 3.1-29 have been deleted.

Amendment Nos.