ML20011E053

Amends 108 & 86 to Licenses DPR-70 & DPR-75,respectively, Updating pressure-temp Limits for Unit 1 to 15 EFPY & to 10 EFPY for Unit 2,based on Results of Analysis of Surveillance Capsules.Limits Will Be Included in Next Updated FSAR Rev

ML20011E053
Person / Time
Site:	Salem
Issue date:	01/29/1990
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20011E054	List: ML20011E053 ML20011E055
References
NUDOCS 9002070191
Download: ML20011E053 (34)

Text

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e NUCLE AR REGULATORY COMMISSION a

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....*p PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY i

ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 ELEMGENERATINGSTATION.UNITNO.1 i

AMENDMENT TO FACILITY OPERATING LICENSE I

Amendment No.108 License No. DPR-70 1.

The Nuclear Regulatory Commission (the Comission or the NRC) has found that:

A.

The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated December P8,1988 and supplemented on July 31, 1969, October 18, 1989 and December 19, 1989 complies with the standards and requirementsoftheAtomicEnergyActof1954,asamended(theAct),

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 public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

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

The issuance of th'is amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Tecnnical Specifica-l tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2)ofFacilityOperatingLicenseNo.DPR-?Oishereby j:

amended to read as follows:

V 9002070191 900129 PDR ADOCK 05000272 P

PDC

7 I

2 v

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No.11 1, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

t 1

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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,[ L Walter R. Butler Director Project Directorate 1 2 Division of Reactor Projects I/ll

Attachment:

Changes to the Technical Specifications Date of Issuance: January 29, 1990 i

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c ATTACHMENT TO LICENSE AMENOMENT NO. 108 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET N0. 50-272 P,evise Appendix A as follows:

Remove Pages Insert Pages 3/4 4-26 3/4 4-26 f

a 3/4 4-27 3/4 4-27 i

3/4 4-27a B 3/4 4-6 B 3/4 4-6 B 3/4 4-7 B 3/4 4-7 i

f B 3/4 4-B B 3/4 4-B B 3/4 4-9 B 3/4 4-9 B 3/4 4-10 B 3/4 4-10 B 3/4 4-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 B 3/4 4-15 B 3/4 4-16 l

B 3/4 4-17 i

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caterial Prererty Initial RTNOT

'I*I RT After 15tFPY: 3/4T. 162'F NOT 1/47, 222.5'T 3888 1

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leeltatte ftuptsatung (ege.r) l' CONTAINS NO NkRGIN FOR POS$18LE INSTRUMENT ERRORS Sales Unit 1 Reactor Coolant Systes Heatup Limitations rigure 3.4-2 l'

Applicable for Heatup Rates up to 60'F/HR for the $ervice Period up to 15 EFPY sam - mT1 3/4 4-26 Amendment No.108 m

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inicarts truptmatung (sto.r) i h

r CONTAINS NO MARGIN FOR P055!BLE INSTRUMENT ERROR 5 ripre 3.4 3 Sales Unit 1 Reactor Coolant System Cooldown Liettations Applicable for Cooldown Rates up to 100'F/HR for the Service Period up to 15 EFPY sAI.m - Uw!T 1 3/4 4-27 Amendment No.108 I; -

. -. -... - - ~.

7 i

2 REACTOR C00LAh"T SYSTDI BASES 3/4.4.9 Pm!URE/TD(PERATURE LTMITS l

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

1)

The reactor coolant temperature and pressure and system bestup and cooldown rate (with the exception of the prescuriser) shall be limited in accordance with Figures 3.4-2 and 3.4 3 for the service period specified thereon.

a)

Allowable combinations of pressure and tadperature for specific temperature change rates are belew and to the right of the limit lines shown.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

j b)

Figures 3.4 2 and 3.4-3 define limits to assure prevention of nondoctile failure only. For normal operation, other ie.h m nt plant characteristics, e.g., pump heat addition and pressuriser heater capacity, may limit the heatup arid cooldown rates that can be s

achieved over certain pressure temperature ranges.

1 2)-

These limit lines shall be calculated periodically using methods provided below.

i 3)

The secondary side of the steam generator must not be pressuriged above 200 psig if the temperature of the steam generator is below 70 F.

l 4)

The,F/hr respectively.pressuriest heatup and cooldown rates shall not exceed.

200 The spray shall not be used if the temperature dif{F.erence between the pressuriser and the spray fluid is greater than 320 5)

System preservice hydrotests and in service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler j

and Pressure Vessel Code.Section XI.

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

These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Soiler and Pressure Vessel Code and the calculation methods described in WCAP-7924 A. " Basis for Heatup and Cooldown Limit Curves, April 1975".

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature. RT

, at the end of 15 affective full power years of service life. The 15 N Y service life period is chosen such that the limiting RT@ limiting unitradiated at the 1/4T location in the core region is greater than the RT of t The selection of such a91miting RT assures that all material.

components in the Reactor Coolant System will N7 operated conservatively in accordance with applicable Code requirements.

SALD1 UNIT 1 B 3/4 4 6 Amendment No.108

i

~

• p REACTOR COOLANT SYSTDI t

RAsts

,s The reactor vessel materials have been tested to determine their initial RT

the results of these tests are shown in Table B 3/4.4 1.

Reactor tion and resultant fast neutron (E greater than 1 MgV) irradiation can o

cause an increase in the RT based upon the fluence and N. An adjusted reference temperature (ART'#,

copper and nickel content of the material in

)

question, can be predicted.

The ART is based upon the largest value of RT cceputed by the methodolo Presented in Regulatory Guide 1.99, Revision b The ART for each material I

given by the following expressions s

ART = Initial RTET+

RTET

• M*'8 A" Initial RT is the reference temperature for the unitradiated material.

I the N adiation and is calculated as follows:isSemeanvalueoftheadjus RT RTET = Chemistry Factor x Fluence Factor The Chemistry Factor, CF (F), is a function of copper and nickel content.

It' is given in Table B3/4.4-2 for welds and in Table B3/4.4-3 for base metal (plates and forgings).

Linear interpolation is permitted.

The predicted neutron fluence as a function of Effective Full Power Years (EFPY) has been calculated and is shown in Figure B3/4.4-1.

The fluence factor can be calculated by using Figure B3/4.4-2. Also, the neutron fluence at any depth in the-vessel wall is determined as follows:

f = (f surface) x (e-0.24X) where "f surface" is from Figure B3/4.3-1, and X (in inches) is the depth into the vessel well.

Finally, the "Hargin" is the quantity in "F that is to be added to obtain conservative, upper bound values of adjusted reference temperature for the calculations required by Appendix G to 10 CFR Part 50.

2 Margin = 2

{ + cr3 If a maseured value of initial RT Ifgeneric$1ueofinitialRTfor the material in question is used, e may be t as sero.

I is used, e should be obtain fromthesamesegofdata. The standard N iations, fof, RT eAc,eed0.50timesthemeanvalueofare 28 7 for welds and 17 F for base metal, except th cf RT 3

3 ET ""#I"* *

• The heatup and cooldown Itait curves of Figures 3.4 2 and 3.4 3 include predicted adjustments for this shift in RT at the end of 15 ETPY.

ET SALEM - UNIT 1 B 3/4 t. 7 Amendment No.108

y REACTOR COOLANT SYSTEM BASES Values of O RT determined in this egnner may be used until the results from the materi@ surveillance program, evaluated according to ASN E185, are available.

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

The heatup and cooldown cerves must be recalculated when the ORT determined from the surveillance capsule encoeds the calculated -6 RT@T foNheequivalentcapsuleradiecionexposure.

Allowable pressure temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Soiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-7924 A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (1.EFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall 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 capabilities of inservice inspection techniques. Therefore, the reactor.

operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection egainst nonductile failure.

To assure that the radiation embrittlement effects are accounted fer in the i

calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RT is used and this includes the radiation induced cooldowncubsaregenerated.correspondh,totheendoftheperiodforwhichh shift, ORT The ASME approach for calculating the allowable limit cerves for various heatup and cooldown rates specifies that the total stress intensity factor, K

chc,ooldowncannotbegreaterthanthereferencestressintensityfactor,Kfo for the metal temperature at that time. K is obtained from the referenceyg, fracture toughness curve, defined in Appenkkx G to the ASME Code.

The K curve is given by the equation:

IR KIR " 26.78 + 1.223 exP [0.0145(T-RTE T + 160)) (1) where K is the reference stress intensity factor as a function of the metal temperabreTandthemetalnil-ductilityreferencetemperatureRT the governing equation for the heatup cooldown analysis is defined $. Thus, Appendix G of the ASME Code as follows:

L CKyg + KIT I IR 1

1 SALEM UNIT 1 B 3/4 4 8 Amendment No.108

i l

1

+

REACTOR COOLANT SYSTDI

&ASES l

where Kyg is the stress intensity factor caused by membrane (pressure) stress.

KIT is the stress intensity factor caused by the thermal gradients.

K is provided by the code as a function of temperature relative to the R

of the material.

l 1

C = 2.0 for level A and 8 service limits, and C = 1.5 for inservice hydrostatic and lec,k test operations.

At any time during the heatup or cooldown transient. K is determined by the metal temperature at the tip of the postulated flaw, tkI appropriate value for RT

, and the reference fracture toughness curve. The thermal stresses reNtingfromtemperatuegradientsthroughthevesselwallarecalculatedand (

then the corresponding (thermal) stress intensity factors, Kg, for the reference flaw are computed. FromEquation(2)thepressureItressintensity f actors are obtained and from those the allowable pressures are calculated.

C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cocidown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cocidown rates.

Allowable pressure-temperature relations are generated for both steady state and finite cooldown rate situations. From these relations composite limit i

curves are constructed for each cooldown rate of interest.

+

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the 1

vessel ID.

This condition, of course, is not true for the steady state situation.

It follows that at any given reactor coolant temperature, the a T developed during coeldown results in a higher value of K at the 1/4T locationforfinitecooldownratesthanforsteadystate$peration.

y Furthermore, if conditions exist such that the increase in K exceeds K the calculated allowable pressure during cooldown will be grd$ter than tN, steady-state value.

i-i The above procedures are needed because there is no direct control on temperature at the 1/47 location, therefore, allowable pressures may

(

unknowingly be violated if the rate of cooling is decreased at various l

intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire i

cooldown period.

SALEM - LHIT 1 B 3/4 4-9 Amendment No.108

REACTOR COOLANT SYSTEM BASES i

i Three_ separate calculations are required to determine the limit curves for i

finite-heatup rates. As is done in the cooldown analysis, allowable pressure-temperature reistionships are developed for steady state conditions as well as finite heatup rate conditions assuming the presence of a 1/47.

defect at the inside of the vessel vall.

produce compressive stress at the inside of the wall that alleviate theThe thermal g tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature therefore, the K for the 1/4T crack during heatup is lower than the K conditionsatthesamecoolanttedhrsture.forthe1/4TcrackkOringsteadystate q

During heatup, especially at the and of the transient, conditions may exist such that the effects of j

i compressive thermal stresses and different K s for steady-state and finite

)

TR heatup rates do not offset each cther and the pressure temperature curve based on steady-state conditions no longer represents a lower bound of all similar i

curves for finite heatup rates when the 1/4T flaw is considered. Therefore.

i both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates 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, the thermal gradients established at the outside surface during heatup produce 1

stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup j

ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analysed on an individual basis.

Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are i

produced as follows. A composite curve is constructed based on a point-by point comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

L L

1 l

l l

e SALDi - UNIT 1 B 3/4 4 10 Amendment No.108 l

l'

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Jt fC10R COOLANT SYSTEM I

J BASES Finally, the new 10CFR$0 rule which addresses the metal temperature of the closure head flange regions is considered. This 10CFR50 rule states that the i

metal temperatuge of the closure flange regions must exceed the material RT by at least 120 F for normal operation when the pressure encoeds 20 percent EI the preservice hydrostatic test pressure (621 psfg for Sales).

Table B3/4.41indicatesthatthelimitingRT"EIlowabletem i

of 28 F. occurs in the closure head flange,of Salen Unit 1. and the nintaum perature of this region is 148 F at pressures greater than 621 pois. Theso Itaits do not affect Figures 3.4-2 and 3.4-3.

l Although the pressuriser operates in temperature ranges above those for which there is reason for concern of non-ductile f ailure

• operating 1Laits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two POPSs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protectd from pressure transients which could exceed the limits of Appendix G to 10 gFR Part 50 when one or more of the RCS cold less are less than or equal to 312 F.

Either POPS has adequate relieving capability to protect the RCS from overpressurisation when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam gener6 tor less than or equal to 50'F above the RCS cold leg temperatures, or (2) the start of a safety injection pump and its injection into a water soli 6 RCS.

i f

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SALEM - UNIT 1 B 3/4 4-11 Amendment No.108 w

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Shell f:norgy.

Normal to 50 ft Ib Principal Principal 35-Mit Wt er k i n<;

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Plate tJo.

Material Cu Ni T

Temp-RT.

Direction Directior Component or We lti fao.

Tygve (t)

(1)

(*F) l'F)

(*F)

(f t -Ile)

(It-lb)

Cl It<1 Dome H2407-1 A553H, Cl.1 0.20 0.50

-30 99*

39 71.5*

110.

Cl 0,1 segment 192406-1 A513H, Cl.1 0.13 0.52

-20 89*

29 97*

125 Cl II.I segment H2406-2 A%33H, CA.I p.16 0.50

-30 H5*

25 79*

122 Cl st.I segment it2406-1 A5 3 385, Cl.1 0.10 0.53

-50 66*

6 R6*

I32 Cl II.I F l ange H2811 A%08, Cl.2 0.72 2R*

22*

28 329*

199 Ves*;o l Flange H2410 A50H, Cl.2 0.67 60*

0*

60 94*

145 Inlet th > z z l e 152408-1 A%08, Cl.2 0.68 50*

43*

50 94*

144 Inlet thiz z l e H240H-2 A508, cl.2 U.71 46*

26' 46 102*

157 Inlet t&.zzle H240R-3 A50R, Cl.2 0.66 47*

37*

47 105*

161

'I Inlet thiz z l e H240R-4 A50R, Cl.2 0.65 9*

17*

9 IOR.5*

167 that let tJe, z z l.*

I42409-1 ASOH, el.2 0.69 60*

95*

60 4R*

75 t ha t l e t.

thiz z l e 152409-2 A%0R, Cl.2 0.69 60*

95*

60 51*

7R M

that let th> zz l e it2409-3 A508, Cl.2 0.74 60*

10*

60 79*

121

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oint let thiz z l e H2409-4 A%08, cl.2 0.74 60*

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60 R2+

126 j_

lopper Shell H2401-1 A531H, Cl.1 0.22 0.48

- 10 R7*

27 74*

114 topsw r Stiel l it2401-2 A53 3tt, C1.1 0.19 U.4R 0

80*

20 79*

122 t ogipe r Shell H2401-3 A533H, Cl.I U.24 0.58

-10 114*

34 62*

96 I nt er Shell I42402-1 A533H, Cl.1 0.24 0.52

-30 los 45 73 97

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int er Shel l H2402-2 A531H, Cl.1 0.24 U.50

-30 55

-5 91 112 Inter Siro l l H2402-3 A5 3 385, Cl.I U.12 0.50

-40 57

-3 104 127 1.ow..r Shell 152403-1 A5 3 3H, Cl.1 0.19 0.4H

-40 70 10 99 143

1. owe r Shell 542403-2 A5 3 3tt, C1.1 U.19 0.49

-70 86 26 94 128

1. owe r Shell It240 5-3 A5 3185, Cl.1 U.19 U.48

-40 90

.30 102 I ll Hot 18<1 Segment _

H2404-1 A53 3tl, Cl.1 0.10 0.52 10 48*

19 78*

120 Ito t 11 1 Segmen t H2404-2 A533H, Cl.1 0.11 0.53

-50 60*.

o 86*

132 Ito t fiel Segment H2 4'J 4-3 A53 3H, Cl.1 U.12.

U.52 10 47*

10 82*

126 Ilo t 111 Dome 182405-I A5 3 3tt, Cl.1 0.15 0.50

. - 20 57*

-3 69*

106 Csrum. We l.1 Het 6-042 U.22 1.02

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Int. Shell 4

Cirum. We let 9-042 0.25 0.72 4

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Vertical Welst I A,ft,CI co

• Est imat eel tw*r NftC St anelatel Review Plan Sect ion 5.3.2.

I:st ima t eil. per Itequiatory Guile 1.99, Rev. 2 Es t i ma t eel' po r Pressur i re<1 Tiiermal Shock Rule, 10CFR50.61 t.

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TABLE D 3/4.4.2 M7 FMTS F98 MLM, 'F 31 hi. h-t l

E A @ LR E,3 @ & M m

=

=

=

0.01 0.00 31 N

N N

N 37 E

\\

0.08 30 M

41 41 41 41 41 j

0.N N

43 M

M M

M M

]'

0.08 m

M N

a a

a m

O.08 39 53 77 83 83

.83 83 0.07 33 M

85 M

88 M

M 0.00 M

M N

108 108 100 100 j

0.00 40 61 M

118 133 133 133

.l

-.n 0.10 44 M

97 138 133 188 1M 0.11 40 SS 101 1M-144 148 148

'l 0.12 63 73 103 135 163 161 - 161 1

l 0.18 M

78 108 IN 103 173 aft

~

i -

0.14' 41 79 100 143 188 ISS 100 0.15 M

N 118 tot 178 191 300 l

0.16 70 SS til 149 178 199 til 4

I O.17 78 93 '119 181 IM 307 331 0.18 79 95 133 ' 1H 187 314 330 l.

0.19 SS 100 136 157 191 330 338 0.30 SS 104 139 100 1H 333 345 j

0.31 99 100 133 1H 197 339 - 383 i

L 0.33 97 113 1E 167 300 333 387 j

O.33 101 117 140 let 303 336 363 l'

O.M 105 131 144 173 306 330 368

}

0.35 110 1M 148 178 300 943 373 0.36 113 130 151 100 313 344 776 0.37 119 1M 1M IN 316 349 300 l.

0.38 139 1M 100 187 318 341 384 O.39 138 143 1M 191 333 3H 387 O.30 131 tot 187 1H SS8 357 390 0.31 IN 151 173 1M 330 300 M3 0.33 140 1M 175 set 331 363 396 0.33 144 100 100 306 384 SH 399 L

O.M 149 1H 1H 300 330 369 303 0.35 153 188 187 313 341 373 305 0.36 1M 173 191 316 348 378 300 l

0.37 les 177 1H 330 Set 378 311 0.38 1H 183 300 333 340 381 314 0.39 171 1M 303 337 SH 365 317 0.40 178 1M 307 331 387 388 330 l

sAtrx t'N2; a 8 3/t. t.

13 Amendment No.108 i

1 l

t

TABLE 3 3/4.4-3 CBEMIST8Y PactOS PtB 8683 NBBAL, 'F

Copper, Ni-L1, bg b5 A gag g,g g,g g g g,g 9

90 30 30 30 30 SD 30 0.01 30 30 30 30 30 30 30 0.08 30 SD 80 30 30 2

30 0.03 30 30 30 30 30 30 SD 0.04 as N

N N

38 38 33 0.08 38 31 31 31 31 31 sg 0.06 38 37 37 37 37

- 37 37 0.07 31 43 44 44 44 44 g

0.00 34 48 81 81 81 Sg gg 0.08 37 83 88 88 p3 48 gg 0.10 41 88 88 88 87 87 87 0.11 48 63 73 74 77 77 77 0.13 49 67 79 83 86 86 88 0.13 83 71 88 91 98 98 98 0.14 87 78 91 100 108 106 100 0.18 61 80 99 ' 110 118 117 117 0.18 88 84 los n8 133 gg 13g 0.17 OS 88 110 137 133 138 135 0.18 73 93 118 134 141 144 144 0.19 78 87 130 143 180 184 184 0.30 83 103 138 149 189 1H 188 0.31 86 139 188 187 173 174 0.33 91 134 161 176 181 184 0.33 96 138 167 184 190 194 0.34 100 143 173 191 199 304 0.38 los 148 176 198 SOS 314 0.36 3*

181 100 908 316 331 0.37 A

188 184 311 336 330 0.N 138 180 187 316 333 330 0.39 134 143 184 191 331 341 348 0.30 139 148 187 194 338 349 387 0.31 134 181 173 198 338 388 388 0.33 139 188 178 300 331 980 374 0.2 led 100 100 28 334 384 383 0.34 148 164 184 308 388 388 390

0. N 183 168 187 313 341 373 398 0.N 188 173 191 316 348 378 308 0.37 163 177 198 330 348 378 308 0.38 188 183 20 333 380 381 313 0.39 171 188 303 327 384 388 317 0.40 178 189 307 331 387 388 330 SALF.M UNIT 1 B 3/4 4-14 Amendment No.108 l

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S A:.E.v. - UNIT 1 B 3/4 4-15 Amendment Nc 108

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C' 3/4.4.10- STRUCTURAL INTECRITY hM

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- The' inspection programs for ASME Code. Class li 2 and 3 components ensure:that.

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.the structural--integrity of these components will be maintained at an l

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-acceptable level'throughout the life of the plant.

To the' extent' applicable, i

the inspection program for these components is in compliance with Section XI i

of the ASME Boiler and Pressure' Vessel Code.

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION -

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y; WASHINGTON, D, C. 20$$6

.....,o PUBLIC SERVICE ELECTRIC &' GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-311 i

SALEM GENERATING STATION. UNIT NO. 2

[

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 86 License No. DPR-75 1.

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

A.

The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company,(Delmarva Power and Light Company and Atlantic City Electric Company the licensees) dated December 28, 1988 and supplemented on July 31, 1989, October 18, 1989'and December 19, 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; B.

The facility will operate in conformity with the application, the-

. provisions of the Act, and the rules and regulations of the Commission; H

C.

There is reasonable assurance: (1) that the activities authorized by

-this amendmert can be conducted without endangering the health and safety of the oublic, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; O.

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 accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been-satisfied.

2..

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby L

amended to read as follows:

-(2) ' Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices.A and B as revised through Amendment No. 86,areherebyincorporatedInthe license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective-as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE~ NUCLEAR REGULATORY COMMISSION

/ 't [

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L-Walter-R. Butler, Director Project Directorate-I-2 Division of Reactor Projects I/II e

Attachment:

Changes to the Technical Specifications Date of Issuance:. January 29, 1990

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' ATTACHMENT TO LICENSE AMEMDMENT NO. 86

-FACILITY OPERATING LICENSE;NO. DPR-75 DOCKET NO. 50-311' Revise Appendix A'as follows:

Remove Pages Insert Pages 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 B 3/4-4-7 B 3/4 4-7 B 3/4 4-8 B 3/4 4-8 B 3/4 4-9 B 3/4 4-9 i

B 3/4'4 10 B 3/4 4-10 B.3/A 4-11 B 3/4 4-11 B'3/4.4 B-3/4.4-12 B 3/4 4-13 B 3/4 4-13 B:3/4 4 B 3/4 4-14 8 3/4i4-15~

B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 B 3/4 4-17 8 3/4 4-17 B 3/4 4-18 t

7

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MATERIAL-PROPERTY BAS]S x.,

CONTROLLING MATERIAL: LONGITUDINAL WELD COPPER CONTENT:

NICKEL CONTENT:.

0.35 WT%

1.00 WTt INITIAL RTNDT:

-56'F RT AFTER 10 EFPY:

1/4 T. 178.6'F NDT 3/4 T. 116.1*F CURVES APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS 2500 i

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2250 L et Test Limit l

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2000 l

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Temp. (311*F) for 250 up iUll'tYP

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200 250 Jet 350 -

400 4!ie 300 INDIC&fE0 TEMPERAfWEt (Ste.F) 1

. Salem Unit 2 reactor coolant system heatup limitations applicable for the first 10 EFPY with maximum heatup rate of 60'F/hr FIGURE 3.4-2 SALEM UNIT 2 3/4 4-28 Amendment No. 86

c 1

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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: LONGITUDINAL WELD J

COPPER CONTENT:-

0.35 WTt NICKEL CONTENT:

1.00 WT1 1

INITIAL _RTNDT:

- 56'F RT AFTER 10 EFPY:

1/47, 178.6*F NDT 3/4T. 116.1*F CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS NO MARGIN FOR POSSIBLE INSTRUMEN 2100

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2210 l

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l for the first 10 EFPY-FIGURE 3.4-3 SALEM. UNIT 2 3/4 4-29 Amendment No. 86

!^

. REACTOR COOLANT SYSTD1 BASES s

A 3/4.4.10 PRESSURE / TEMPERATURE LIMITS 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.

I 1)

The reactor coolant temperature and pressure and system heatup and cooldown rate (with the exception of the pressuriser) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon.

a)

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. _ Limit lines for cooldown rates between those presented may be obtained by interpolation.

b)

Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile f ailure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

s 2)

These limit lines shall be calculated periodically using methods provided below.

3)

Thesecondarysideof.the'steamgeneratormustnotbepressurigedabove 200 psig if the temperature of the steam generator is below 70 F. '

4): 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 dif{erencebetweenthepressurizerandthesprayfluidisgreaterthan-320 F.

5)

System preservice hydrotests and in-service leak'and hydrotests shall be-periormed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor 7

vessel are determined in accordance with the NRC Standard Review Plan,' ASTM ElB5 82, and in accordance'with additional reactor vessel requirements. These-l properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975".

Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RT

, at.the end of 15 effective full power years of service life. The 15 N Y service life period is chosen such that the limiting RT at the 1/4T location in the.

coreregionisgreaterthantheRT@mitingRTof t b limiting unirradiated material. The selection of such a assures that all components in the Reactor Coolant System will bToperated conservatively

'i in accordance with applicable Code requirements.

SALEM - UNIT 2 B 3/4 4-7 Amendment No. 86 t

t REACTOR COOLANT SYSTEM k

i a

e BASES The reactor vessel materials have.been tested to determine their initial RT

the~results of these tests are shown in Table B 3/4.4 1.

Reactor tion and resultant fast neutron (E greater than 1 MEV) irradiation can o

cause an increase in the RT based upon the fluence and hI. An adjusted reference temperature, (ART),

l copper and nickel content of the material in question, can be predicted.

t I

The ART is based upon the largest value of RT computed by the methodology presented in Regulatory Guide 1.99. RevisionE The ART for each material is given by the following expression:

~

ART = Initial RTET + ARTg + Ma:nin Initial RT is the reference temperature for the unirradiated material.

ART is mean value f the adjustment in' reference temperature caused by the$Trradiation and is calculated as follows:

ARTET = Chemistry Factor x Fluence Factor The Chemistry Factor, CF (F), is a function of copper and nickel content.

It is given in Table B3/4.4-2 for welds and in Table B3/4.4-3 for base metal (plates and forgings).

Linear interpolation-is permitted.

The predicted neutron fluence as a function of Effective Full Power Years (EPPY) has been calculated and is shown in Figure B3/4.4-1.

a The fluence factor can be calculated by using Figure B3/4.4 2.

Also, the neutron fluence-at any depth in the vessel wall is determined as follows:

f =.(f surface) x (e-0.24X) where f surface" is from Figure B3/4.3-1, and X (in inches) is the depth into' the vessel wall.

4

-Finally, the " Margin" is the quantity in F that is to be added to obtain conservative, upper-bound values of adjusted reference temperature for the calculations required by Appendix G to 10 CFR Part 50.

Margin = 2 2, "6 2

y If a measured value of initial RT Ifgenerici11ueofinitialRTfor the material in question is used, o may be taken as zero.

is used, e shouldbe obtaingdfromthesamesegofdata. Thestandarddevbiions,forakT are 28 F for velds and 17 F for base metal, except that r need not e b e,dv, 3

0.50 times the mean value of ART 3

ET

""'I"***

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 incl.ude predicted ~ adjustments for this shift in RT at the end of 15 EFPY.

ET j

SALEM - WIT 2 B 3/4 4-8 Amendment No. 86 1

i 1*

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REACTOR COOLANT SYSTEM BASES Values of a RT determined in this manner may be used until the results fromthemateriksurveillanceprogram,evaluatedaccordingtoASTME185,are available.

Capsules will be removed in accordance with the requirements of AS'1H E185-82 and 10 CFR Part 50. Appendix H.

The heatup and cooldown curves must be recalculated when the 6 RT determined froci the surveillance capsule exceeds the calculated A RT I

h* '9"I"*It espsule radiation exposure.

NDT Allowable pressure temperature relationships for var hut heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR' 1

Part 50 and these snethods are discussed in detail in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves-is based L

upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one quarter oi the wall thickness. T, and a length of 3/2T is assumed to exist at the inside of the vessel wall 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 4

capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and t

provide sufficient safety margins for protection against nonductile failure.

To assure that the ridiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility l

reference temperature RT is used and this includes the radiation induced cooldowncuNsaregenerated.correspondk,totheendoftheperiodforwhichheatupan shift, A RT The ASME approach for calculating the allowsble limit curves for various

-heatup and cooldown rates specifies that the total stress intensity factor, y

K7 'for the combined thermal and pressure stresses at any time during heatup' i

or cooldown cannot be greater than the reference stress intensity factor, KIR' d

for the metal temperature at that time. K is obtained from the reference i

fracturetoughnesscurve,definedinAppenbxGtotheASMECode.

The K curve is given by the equation:

IR KIR = 6.78 + 1.223 exp [0.0145 M 160))

O)

NDT is the reference stress intensity factor as a function of the metal I

where K temperabreTandthemetalnil-ductilityreferencetemperatureRT the governing equation for the heatup-cooldown analysis is defined $. Thus, Appendix G of the ASME Code as follows:

CK73 + KIT IKIR SALEM - UNIT 2 B 3/4 4-9 Amendment No.86

r Y

[

RF. ACTOR COOLANT ' SYSTDi o

BASES 1

where K is the stress intensity factor caused by membrane (pressure) stress.

yg K

is the stress intensity factor caused by the thermal gradients.

IT K

is Provided by the code as a function of temperature relative to the RhR f the material.

T C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations, c

j At any time during the heatup or cooldown transient, K is determined by the metaltemperatureatthetipofthepostulatedflaw,tk$appropriatevaluefor RT

, and the reference fracture toughness curve. The thermal stresses rektingfromtemperatuegradientsthroughthevesselwallarecalculatedand then the corresponding (thermal) stress intensity factors. Ky, for the i

reference flaw are computed. From Equation (2) the pressure Ittess intensity factors are obtained and from these the allowable pressures are calculated.

i COOLDOW '

iFor the calculation of the allowable pressure versus coolant temperature i

during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel vall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile

{

stresses at the inside, which increase with increasing cooldown rates.-

i Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because a control of the cooldown procedure is bared on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the 1/4T vessel location is at a higher-temperature than the fluid adjacent to the vessel ID.

This condition, of course, is not true for the steady-state situation.

It follows that at any given reactor coolant temperature, the a T i

developed during-cooldown results in a higher value of K at the 1/4T locationforfinitecooldownratesthanfor=teady-state $peration.

y Furthermore, if conditions exist such that the increase in K exceeds K i

thecalculatedallowablepressureduringcooldownwillbegre!terthantkI, steady-state value.

3 The above procedures are needed because there is no direct control on temperature at.the 1/4T location, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cocidewn ramp. The use of the composite curve oliminates this problem and assures conservative operation of the system for the entire cooldown period.

SALEM - ll NIT 2 B 3/4 4-10 Amendment No. 86

^

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U REACTOR COOLANT SYSTEM

-l BASES 1

W HEATUP' Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable proscure temperature relationships are developed for steady state conditions as voll as finite heatup rate renditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup i

produce compressive str6ss at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature.Thereforethe K for the 1/4T crack during heatup is lower than the K for the 1/4T crack Ering steady state conditionsatthesamecoolanttedherature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K s r steady-state and finite IR heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates 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, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of coarse, cre dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather,'each heatup rate of interest must be analyzed on an individual basish r

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve.is constructed based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values-taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

SALEM - UNIT 2 B 3/4 4-11 Amendment No. 86

-b RIMCTOR COOLANT SYSTEM RAE ES '

Finally,-the new 10CFR50 rule which addresses the metal temperature of the-closure head flange regions is considered. This 10CFR50 rule states that the metal:temperatuge of the closure flange regions must exceed the material RT by at least 120 F for normal operation when the pressure exceeds 20 percent SI g

thepreservicehydrostatictestpressure(621paggforSalem).

Table B3/4.4 1 indicates that the limiting RTN f 28 F occurs in the closure head flangeofSalemUnit2,andtheminimumIIlowabletemperatureofthisregion is 148 F at-pressuren greater _than 621 psig. These limits do not affect Figures 3.4 2 and-3.4 3.

Although-the pressurizer operates in temperature ranges above those 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.

1 The OPERABILITY of two POPSs or an RCS vent opening of greater than 3.14 M

i square inches ensures that the RCS will be protectd from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of, the kCS cold legs are less than or equal to 312 F.

-l Either POPS has adequate relieving capability to protect the RCS from overpressurization when I

the transient is limited to either (1) the start of an idle RCP with the i

secondary water temperature of the steam generator less than or equal to'50 F above the ECS cold leg temperatures, or (2) the start of a safety injection jj pump and its injection into a water solid RCS.

j i

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SALEM - UNIT 2 g 3/4 4-12 Amendmen t No. 86

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Temp RT Direction Direction Component or Weld Nr).

Type

(%)

{t)

(*F)

(*F)

(*F)

(ft-lb)

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Closuto Ib! Dome P470H A51311CI.I 0.11-0.70

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-20 15*

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97*

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A5 3 3ttCl.1 0.10 0.55 0

51*

0*

84*

129 Clostare tid Peel H4707-3 A5 3 3ttCl.1 0.13 0.63 0

66*

6' 84*

129.5 Closure Ild Finq 144702-1 A5 0HCI. 2 0.68 28*

39*

28*

104*

160 Vessel Flanqo 1t5001 A50HCI.2 0.70 12*

4*

12*

107*

164 Inlet Nozzle 14470 1-1 A50 hcl.2 (1.69 60*

62*

60*

>72*

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Infot Nozzle 114701-2 A50 hcl.2 0.69 60*

25*

60*

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Intot Nozzlo It4 70 3-3 A 50 hcl.2 0.68 60*

32*

60*

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>109**

Inlet Nozzle B4703-4 A508Cl.2 0.81 60*

40*

60*

80*

121.5 Outlet Noz zl e H4704-1 A50 Hrf.2 0.84 60*

8' 6 fi

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126 c,

ou t l e t No z z l e f54704-2 A50 8Cl.2 0.77

-60*

20*

60*

75*

116 '

w2 Ou t l e t Nozzle R4704-3 A508CI.2 0.69 28*

8*

28*

82*

126 d

a that let Nozzle 144704-4 A50ftCl.2 0.71 60*

40*

60*

77*

179 tipser Shel l it4 71 1 - 1 A5 31stCt.1 0.11 0.55 0*

50*

0*

87*

134 w

tippe r Shell 144711-2 A5 3 3tict. I 0.14 0.56

-10 60*

0*

79*

122 Ifpper Shell H4711-3 A533 hcl.1 0.12 0.58

-10 88*

2 fl

• 69*

107 I nt er. Shell H4712-1 A5 3 38tCl.1 0.13 0.56 0

<60 0

105 138 Int er. Shell 154712-2 A533ncl.1 0.14 0.60

-20 72 12 97 127.5 Inter. Shell 114712-3 A5 3 38tcl.1 0.11 0.57

-50 70 10 107 116 Lower She l l 154713-1 A5 3 3ttCf.1 0.12 0.60

-10 68 8

98 127 I.ower Shell 154713-2 A5 3 3ItCl.1 0.12 0.57

-20 68 8

103 135.5 1.ower Shell 04711-3 A5 3 3tts't.1 0.12 0.58

-10 70 10 122 135.5 flo t t om lid Peel 194709-1 A513tlCf.1 0.12 0.60

-30 54*

-6*

90*

139 Ilot r om th! Peel h4709-2 A5 3 3 nCl.1 0.12 0.58

-20 42*

-18*

89*

137.5 Itottom tid Peel 144709-3 A513HCI,1 0.11 0.56

-20 71*

11*

93*

143 Bottom Ilead H4710 A5 3 3ttCl.1 0.12 0.60

-30 60*

0*

77*

118 y

Cnrum. We 1 1 Het 3-442 0.28 0.74

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Nozzle ShelI i

Int. Shell Cirum. Weld Hot 9-44/

0.175 0.20****

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Int. Shell A f

1. owe r Shel1 Int. Sheli 1-442 0.23 0.73

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Vertical Weld l A,f t,rl 1.o we r :;he l 1 3-44/

0.20 0.86

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Vertical Wold l A,It,C)

Es t i ma t e-d per NI<C Standard Review Plan Section.5.3.2.

100% Shear not reached

• • • Estimated per Pressurized Thermal. Shock Rule, 10CFR50.61

• • • • Cstimated value for t ype MIL D-4. wire heats.

---Q-

.;... i:

. ~6 =

L L

w.

..-.u_.

TABLE B 3/4.4.2 i

05B58787 FacRR 798 E.DS, 'F

Copper, stekel. 9t.-t -

h-E d M MRJ2M M M e

30 90 90 90 30 90 90 0.01

. 30 30 30 90 90 90 90

- 0.08 31 36 37 87 87 -

37 N

i 0.03-33 38 41 41 41 41 41 i

0.04 34 43 N

84 84=

84 84 l

0.08 38 40 67 68 08 68 68 O.06 99 83 77 83 83 83 83 0.07 33 88 88 98 96 98 98 0.04 36 88 90 106-108 108 108 0.00 40 61 94-118 122 -las 133 0.11 49

- 88 97 123 133 138 136 0.10 44-88 101 130 - 144 148 148 0.12 83 72-103 136 183 161 181 0.13' 88 76 106 1M 163 173 176 0.14 61 79 109 143 168 189 188 0.18 66 84 112 146 178 191 300 O 16 70 88 118 149 178 199 til O.17 78 92 119 181 IN 207 221-0.18 79 98 122-184 187 214 330 0.19 83 100 126 187 191 290 238 O.90 88 104 129 160 194333 248 i

r 0.21 92 108 - 133 164 197 229' 282 O.22 97-112 137 167 5)O 232 987

-)

0.23 101 117-140 169 303 334 - 343 0.24 108 121 144-173 906 339 - 968.-

0.28 110 196 148 176 900 943 272 0.26 113 130 181-180 312 -946 278 O.27 119-134 188 - 184 316 349 280 0.38 att 138 100 187 318' 381 -384 0.39

'133 142 164 191 292 384 287 1

0.80 131 148 167 194 228 387 290 0.31.

138 161 173 198 338 960 293 0.32 140 188 178 303 331 383 - 296 4

0.33 144 100 180 908 334 268 299 0.34 149 164 184 209 238 269 302 0.38 183 168 187 212-241 272 308 0.36 188 172 191 216 248 278 308 0.37 182 177 196 220 348 278 311 0.38 166 182 200 223 280 281 314 0.39 171 188 303 227 384 288 317 0.40 178 189 907 231 287 288 390 SALEM UNIT 2 B 3/t.

4-14 Amendment No. 86 I

v

i c.

31-

.c TAsLE B 3/4.4-3 l

1 M Y FA0fB F M E B W 341, 'F Oe9per Niekal. We,.E E, d $4E IeE. IAE Ia.E la.E d l

0 s0 s0 se s0 s0 30 s0 0.01-30 30 30 30

' 30.

30 30 1 0.00 =

30-30 30 30 2

30 30 0.03 30-30 30 30 30 30 30 0.04

~ 33 38 30 SS SS 39 -

M 0.05 38 31 31 31 31 31 31 0.00 38 37 37 37 N

-W 37 0.07 31 43 44 44 44 44 44 l.

0.00 N

48 51 51' 51 51 51 O.00 37 53 58 SS M

M 88 0.10 41.

88 M

85 Of

$7 87 0.11 45 83 73 74 77 77 77 0.13 49 67 79 83 80 M

88 O'.13 la 71 85 91 96 to 98 0.14-87 75 91 100 105 108 100 0.15 81

- to 99 110 118 117 117 l

0.16 85 N

104 118~ ISS ISS 135 4

0.17' 80 58 110 137 133 135 135 t-t 0.18 73 93 115 1H 141-144 144 0.19 78 97 130 143 150 1H IN 0.30 83 103 135 149 ~ ' 150 - 1H 185 O.21 SS 107 -139 155 187 173 174 0.33 91 113 134 161 176 181 184 0.33

- 95 117 138 '187 184 190 1H j

0.34 100 131-143-173 191 199. 304 0.38 104 -136 148-178 196 SOS-314 0.36 100 130 151 100 306 314-331 0.37 114 134 155 1N 311 - 385 330 0.30 119 1M 180 187 318 383 339 0.39 134 143 1H 191 231 341 348

'O.30 139 148 107 1H 335 349 : 357 0.31 134 151 173 100 338 355 386 0.33 139 155 175 SOS 381 900 374 0.33 144 100 100 SOS, 384 SS4 - 303 0.34 140 1H 1H-300 330 388 390 0.35 153 168 187 '313 M1 373 398 0.38 148 - 173 191 218 348 375 308 0.37 103 177 ite 330 348 378 300 0.38 1H 183 300 3B3 300 381 313 0.39 171 185 303 337 384~ 385 317 0,40 175 1FS 307 331 387 308 330

~

sALzM WIT a B 3/4 4-15 Amendment No. 86

c i,

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0' 11',

s

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10

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D

=

I I

1R g

19 5 10

'l s

d w

m THIS CURVE REPRESENTS THE FLUENCE AT THE INNER RADIUS OF THE LIMITING h.

=

LONGITUDINAL WELD SEAM LOCATED AT g

THE 30 AZIMUTH I

10.8 _

i 0

30 AZIMUTH i.

17 I

I I

I I

I l

l I

I l

l l

l l

l 10 0-2 4

6 8

10 12-14 16 18 20 22 24 26 28 30 32 i

L i

SEP.V!CE LIFE (EFFECT!VE FUL'. POWER YEARS)

Fast ' neutron fluence (E > 1 WeV) as a function of full power service life (DTY)

I, FIGURE B 3/4.4-1 SALEM 2 B 3/4 4-16 Amendment NO. 86 l~

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Fluence Factor for use in the expression for O RT

-[

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- Ficuas e 3/4.4-2.

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1-S '.

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REACTOR COOLANT SYSTDI

BASES 3/4.4. 11 STRUCTimE INTEGRITY

.I The inservice inspection and testing programs for ASME Code Class 1.- 2 and 3 components ensure that the structural integrity and operational readiness of -

these components will be maintained at an acceptable level through the life of the plant.

These programs are in accordance with Section XI of the 'ASME

.i Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(s) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(s)(6)(1).

SALEM UNIT 2 B 3/4 4-18 Amendment No.86 i

_