ML20101L185

Proposed TS 3.6 & 4.6 Re Primary Sys boundary,4.6.A Re Thermal & Pressurization Limitations & Table 4.6.3 Re Reactor Vessel Matl Surveillance Withdrawal Schedule

ML20101L185
Person / Time
Site:	Pilgrim
Issue date:	07/01/1992
From:	BOSTON EDISON CO.
To:
Shared Package
ML20101L183	List: BECO-92-070, Forwards Response to Rev 1 to Generic Ltr 92-01 Re Reactor Vessel Structural Integrity,Consisting of NRC-approved Insp Program Governed by Procedures & Encl TS ML20101L185
References
NUDOCS 9207070004
Download: ML20101L185 (12)

Text

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LJMITINLCONDITION FOR OPERAILQL, SURVEltLANCE RE0VIREMENTS

' 3.6 PRIW' SYSTEM BOUNDAkY 4.6. PRIMARY SYST{M BOUNDART Aeolicability: Applicability:

Applies to the operating status of the Applies to the periodic examination and reactor coolant system. testing requirements for the reactor cooling system.

Qbiective: Obiettive:

To assure the integrity and safe To determine the condition of the operation of the reactor coolant system reactor coolant system and the operation of the safety devices related to it.

l Soecification: Smttifisation:

A. Tnermal and Pressurization A. Thermal and Pressurization Limitations Limitations

1. The average rate of reactor I. During heatups and cooldowns, with l coolant temperature change during the reactor vessel temperature less I normal heatup or cooldown shall than or equal to 450*F, the l not exceed 100*F/hr when averaged temperatures at the following i over a one-hour period except locations shall be n rmanently I when the vessel temperatures are logged at least every 15 minute

• above 450*F. The reactor vessel I until the difference between any flange to adjacent reactor vessel I two reaGings at individual I shell temperature differential locations taken over a 45 minute I hall not exceed 145'F. period is less than 5'F:

a. Reactor vessel shell Ujacent to reactor vessel l' n je l

b. Reactor vessel shell flange

c. Recirculation loops A and B

2. The reactor vessel shall not be 2. Reactor vessel shell temperatures, I pressurized for hydrostatic including reactor vessel bottom I and/or leakage tests, and head, and reactor coolant pressure [  %

critical core operation shall not shall be permanently logged at \

be conducted unless the reactor least every 15 minutes whenever the vessel temperatures are above l shell temperature is below 220*F those defined by the appropriate and the reactor vessel is not curv c on Figures 3.6.i, 3.6.2, vented.

and L 6.3. (Linear interpolation I between curves is perm';ted). At l Test specimens of the reactor stated pressure, the reactor l vessel base, weld and heat affected vessel bottoin head may be l zone metal subjected to the highest maintained at temperatures below I fluence of greater than 1 Hev those umperatures corresponding l neutrons shall be installed in the to the adjacent reactor vessel l reactor vessel adjacent to the shell as shown in Figures 3.6.1 l vessel wall at the core midplane and 3.6.2. I level. The specimens and sample program shall conform to the Revision 157 Amendment No. 82, 140 123 9207070004 DR 920701 ADOCK 05000293 PDR

l LIMITING CONDITION FOR OPfRATXON SURVEILLANCE RE00fREHENTS l

3.6.A Ihermal arid Pressurization 4.6.A Thermal and Pressurization Liuitations (Cont'd) g Limitations (Cont'd) l j In the event this requirement requirements of ASTH E 185-66. I is not met, achieve stable Selected neutron flux specimens reactor conditions with shall be removed at the frequency reactor vessel temperature required by Table 4.6.3 and tested above that defined by the to experimentally verify appropriate curve and obtain adjustments to Figures 3.6.1, ,

an engineering evaluation to 3.6.2, and 3.6.3 for predicted NDT l determine the appropriate temperature i' . .'ation shif ts. I course of action to take.

3. The reactor vessel head 3. When the reactd ,essel head bolting studs shall not be bolting studs are tensioned and the under tension unless the reactor is in a Cold Condition, the temperature of the vessel reactor vessel shell temperature head flange and the head is immediately below the head flange greater than 55'F. shall be permanently recorded.

4. The pump in an idle 4. Prior to and during startup of an recirculation loop shall not idle recirculation loop, the be started unicss the temperature of the reactor coolant temperatures of the coolant in the operating and idle loops within the idle and operating shall be permanently logged.

recirculation loops are within 50*F of each other.

5. The reactor recirculation 5. Prior to starting a recirculation pumps shall not be started pump, the reactor coolant unless the coolant temperatures in the dome and in the tempera b es between the dome bottom head drain shall be compared and the bottom head drain are and permanently logged, within 145*F.

6. Thermal-Hydraulic Stability .

Core thermal power shall not exceed 25% of rated thermal power without forced recirculation.

B. Coolant Chemistry B. Coolant Chemistry

1. The reactor coolant system 1. a. A reactor coolant sample shall radioactivity concentration be taken at least every 96 in water shall not exceed 20 hours and analyzed for microcuries of total iodine radioactivity content, per ml of water.

l b. Isotopic analysis of a reactor coolant sample shall be made at least once per month.

Revision 157 Amendment No 82, 140 124

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TABLE 4.6.3 REACTOR VESSEL HATERIAL SURVEILLANCE PROGRAM HITHDRAHAL SCHEDULE Effective Full l cal 9i's Power Years l Number (EFPY) l l

l 1 4.17 l l

2 15 l (approx.) l 1

3 32 l (End of Life) l 1 .

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Revision 157 Amendment No. 82, 140 124A

9 Attachment C to BECo Response to Generic Letter 92-01

Attachment C CHARPY V-NOTCH IMPACT RESULTS FOR PILGRIM BASE METAL Test Impact lateral Fracture Specimen Temperature, Energy, Expansion, Appearance,

. Identification F ft-lb mils Percent-Shear Y2/C -100 3.1 2.4 1 Yl/L -50 8.3 9.0 5 Y2/4 0 24.0 23.2 25 Y3/2 25 52.C 43.2 30 Y2/D 30 44.0 37.4 25 Yl/2 50 74.9 54.2 45-Y3/Y 50 71.3 61.2 40 Y2/1 50 56.4 50.0 45 Yl/0 80 83.8 67.0 60 .

Y2/2 115 110.0 85.2 85 Y2/5- 150 131 .9 75.0 100 Y3/M 250 130.3 92.0 100 Yl/7 300 130.2 84.2 100 Page 1 of 4

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P Attachment C CHARPY V-NOTCH IMPACT RESULTS FOR PILGRIM' WELD METAL Test Impact lateral Fracture Specimen Temperature, Energy, Expansion, Appearance, Identi fication F ft-lb mils Percent Shear -

Y4/B -200 11.0 9.4 0 r Y5/M -150 14.5 17.0 1

-Y6/U -125 6.8 6.8 5 Y6/5 -125 37.0 31.6 '10 Y6/Y' -100 41.0 54.0 20 Y4/5 -50 62.0 54.8 35 .

Y6/7 0 73.4 68.2 65 I i

YS/T 50 85.8 79.0- 75 t Y4/D 60 81.0 72.8 80 j Y5/U- 60 98.0 83.8 85

.Y4/7 60 89.0 77.8- 85 Y6/1 80 96.9 86.6 90.

. Y4/A 150 111.1 '93 100 y Y4/3 230 114.8 85 4 100 Y6/6 300 112.2 75.0 100 Page 2 of 4 l.

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d me - a /+ 424 e la$+ eG-4 -d 4- .#.# ,, LJA.uh4.&M,A 4' J ..%4-r J J lA4.inA..,, 3 Attachment-C ,

CHARPY V-NOTCH IMPACT RESULTS FOR PILGRIM HAZ METAL Test Impact l'ateral Fracture Specimen- Temperature, Energy, Expansion, Appearance, Identification F ft-lb mils Percent Shear YA/J -200 4.5 3.8 0 ,

Y7/P -165 21.5 16.0 5 Y7/A -135 13,0 10.8- 1 YB/1- -100 13.0 12.4 5 Y7/V- -

50 37.5 30.0 15- .

YA/3 - 40 68.6 53.4 35 YA/P -35 74.5 58.4 45 YA/T 0 68.0 60.4 75 Y7/2 35 97.5 84.0 85 YB/2 50 135.7 79.4 100 YB/6 81 107.9 77.2 90 Y7/7 150 99.9. 84.8 100 YB/E 300 112.0 87.8 100 Page 3 of 4 l

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t TENSILE TEST RESULTS FOR PILGRIM UNIRRADIATED BASELINE SPECIMENS 0.2 Percent "

Offset Ultimate Speci- Test- Yield Tensile Fracture Reduction Elongation, %

men Temp, Strength, Strength, Load, Strength. Stress, in Area, Uni- .i Material Number F. ksi ksi Ib ksi ksi percent form Total >

E Base YC-5 72 64.5 86.4 2800 55.7 185 69.90 12.4 25.5 h. I Dase YC-6 '550 57.7 62.8 2925 58.2 166 65.00 10.6 21.4 'E 3 Weld YD-T 72 68.8 78.4 2450 49.6 178 72.02 14.5' 28.6 i Weld YD-K 550 54.6 78.3 2725 54.7 148 63.04 12.0 23.0  :

IIAZ YE-D 72 61.7 79.6 2475 49.7 178 72.17 8.5 22.9  :

liAZ YE-E 550 57.2 78.1 2410 48.4 136 64.48 7.9 18.9 9

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Attachment D to BECo ,

Response to Generic Letter 92-01

Attachment D

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.- LIST OF PIECE, CODE AND HEAT NUMBERS Bottom Head Pc. No. Code No. Heat No.

336-02 G-3114 C-2888-3 336-03. G-3113-1, 2, 3 C-2913-3, A-2222-2, A-222 336-04 G-3111 C-2851-2 Lower Shell Assembly -

Pc. No. Code'No. Heat No.

337-01 G-3109-1, 2, 3 C-2957-1, C-2957-2, C-2973-1, 337-03 G-3108-1, 2, 3 C-2921-2, C-2945-1, C-2945-2 Upper Shell Assembly Pe. No. _ Code No. Heat No. ,

337-02 G-3109-4, 5, 6 C-2561-2, C-2973-2, C-3301-2 337-04 -

G-3107-1, 2, 3 A-2094-1, A-2094-2, C-2906-1 339-02 G-3101 2V-545 -

339-06 G-3142 Certification Vessel Support Skirt Pc. No. Code No. Heat No.

340-02 S-8529 VIEF-1225 340-04 G-3154 A-2933-5 340-06 G-3153 A-3118-2 Page 1 of 3

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RPVWELD DATA -

PILCR1M REACTOR VESSEL BELTLINE WELD DATA

SUMMARY

WIRE WELD GENERIC 1 CHERESTRY R.G. CHEARSTRY RCL CHEMISTRY RG.

) WELD FILLER WIRE FLUK FLUX REF. 1.99 REF. 1.99 REF. '1.99 i- DESCRIPTION ND. HEAT NO. TYPE TYPE LOT CU NI SolmCE & CU Ml SOURCE & CU N1 SOURCE CF 4

27204 B.4 LINDE 3774 0.12 1.06/ 4 161 0.32 0.82 234-

' PAOD) 1.07 0.24 1.00 2&3 239 LONGITUDINAL WELD 0.22 1.10 245 4

LOWER INTERMEDtATE 1-338 0.18 0.96 '209 SHELL A.B.C 12008 , B.4 1092 3774 0.13 0.99 4 172 '

O.13 0.99 .3 172 (MOD) 0.22 0.83 205 GIRTH WELD LNDE y LOWER INTERMEDIATE 1-344 21935 .B-4 1092 3869 0.13 0.71 4 151 0.22 0.71 2&4 184 0.21 0.68 2&3 177 .R es TO LOWER SHELL (MOD) 0.20 0.7f 4 177 0.22 0 83 '205 "

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$ LONGITUDINAL WELD 2-338 .27204 B-4 LNDE 3774 0.13 1.06/ 4 173 $

g LOWER SHELL A,B,C B.4 1092 3714 1.07 0.21 0.98 2&3 226 '[

TEST CAPSULE 1-366 13253 8-4 LNDE 3774 0.13 1.06/ 4 173 0.17 0.81 2 185 m

, & SEAM WELD 1092 1.07 5 u COMPOSITE WELD ALL y

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0.22 f.07 '>cg.

241 0.32 f.07 275 e

' WORST CASE" WELD ALL gz, .. p;- 0.35 f.07 .y 284 ;y

REFERENCES:

(1) SWRt REPORT (ATTACHED) (2) EPRI DATA NP-4797 (EXCERPTS & T. GR!ESBACH)

(3) G. E. DAl A (T. CAINE) i (4) C. E. DATA (RHfLLIS) l i

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Attachment D o

PILGRIM REACTOR VESSEL BELTLINE WELD DATA

SUMMARY

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