ML18093B367

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Proposed Tech Specs Modifying Plant Heatup & Cooldown Curves & Associated Bases Section
ML18093B367
Person / Time
Site: Salem  PSEG icon.png
Issue date: 12/28/1988
From:
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ML18093B366 List:
References
NUDOCS 8901040171
Download: ML18093B367 (54)
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  • Description of Change LCR 88-14 ATTACHMENT 1 Revise Salem Unit Nos. 1 and 2 Technical Specification Figures 3.4-2 and 3.4-3 and the associated Bases Section as provided in the attached pages.

The proposed changes reflect the results of the analysis performed on surveillance capsules removed from each.

reactor during the seventh and third refueling outages of Salem Units 1 and 2, respectively.

Reason for Change The heatup and cooldown curves are being updated to reflect the changes in reactor vessel material properties as identified from the examination of the surveillance capsules from each Salem Unit.

The bases section is revised to reflect the use of Regulatory Guide 1.99, Rev. 2 methodology in estimating the radiation embrittlement of reactor vessel materials.

Justification for Change 10CFR50 Appendix H requires that a sampling program exists to determine the effect of irradiation on reactor vessel materials.

Part of that program requires that if the results of the testing program so indicates, a change to the plant heatup and cooldown curves provided in the Technical Specifications be submitted.

The curves are being revised to ensure that the operation of the reactor will not exceed pressure and temperature limits imposed to prevent the potential for fracture of the reactor vessel or components.

The curves provided in this submittal reflect the results of the latest specimen analysis provided by WCAP 11955 and WCAP 11554, and forwarded to the NRC via NLR-N88169 dated November 7, 1988 and NLR-N87220 dated November 25, 1987.

The proposed changes are also consistent with the guidance.provided in Generic Letter 88-11 and Regulatory Guide 1.99, Revision 2.

Instrument uncertainties are being removed from the curves to clarify the actual requirements.

The use of heatup and cooldown curves without instrument uncertainties has been previously reviewed and approved by NRC for Diablo Canyon.

Instrument uncertainties in the limits, in addition to other conservatisms, are not required to prevent vessel damage and are not required by the regulations.

Significant Hazards-Consideration PSE&G has evaluated the hazards consideration involved with the proposed amendment, focusing on the three standards set forth in 10CFR50.92(c) as quoted below:

LCR 88-14 "The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a facility licensed under paragraph 50.2l(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if.

operation of the facility in accordance with the proposed amendment would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety."

The following evaluation is provided for the significant hazards consideration standards.

1.
2.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes do not increase the probability of an accident since they are being incorporated to ensure that existing safety limits are not exceeded due to changing conditions in the reactor.

The proposed changes are requested so that the results of materials testing is reflected in the operating limits pursuant to the requirements of 10CFR50, Appendix H and Appendix G.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes are being made to ensure that the Salem units do not operate such that the potential for a new kind of accident is created (e.g., fracture of the reactor vessel).

The proposed changes ensure that the operations remain within acceptable areas governed by previously analyzed areas.

The proposed changes do not make a physical change to the facility.

Therefore the proposed changes do not create a new or different type of accident from any previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

The results of the surveillance tests indicate that the reactor pressure vessel has adequate toughness for the continued safe operation provided the requested heatup and cooldown limitations are adhered to.

The proposed changes ensure that the existing margins of safety are met by modifying operating requirements to reflect the results of

LCR 88-14 specimen capsule analysis.

This analysis prescribes what the appropriate heatup and cooldown curves are and accounts for the effects of radiation on reactor vessel materials.

Changes to the-Bases Section of the Technical Specifications are being made to reflect Salem's compliance with later regulatory guidance regarding specimen analysis.

Therefore the proposed changes do not involve significant reduction in a margin of safety.

conclusion The Commission has provided guidance concerning the application of the standards for a No Significant Hazards determination by providing examples of actions not likely to involve a Significant Hazards Consideration in the Federal Register (48FR14870).

This proposed amendment corresponds in part to example II.2, a change constituting an additional limitation not currently in the Technical Specifications, in that the proposed change modifies existing requirements to reflect the results of a test program.

The proposed changes are considered to be more restrictive, since the new curves reflect new heatup and cooldown limits.

The remainder of the proposed change corresponds to example II.7, in that the Bases Section is being modified to reflect compliance to Revision 2 to Regulatory Guide 1.99.

These changes, as demonstrated above, do not constitute a Significant Hazards Consideration.

ATTACHMENT 2 TECHNICAL SPECIFICATION MARKUPS

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UPPER LIMIT OF REG. GUICE TRENO CURVES (FIGuRr SJ/4 4-2)

COPPER CONTENT

0.35 wn PHOSPHORUS CONTENT
0.012 wn AT INITIAL
0°F AT:+ AFTER 10 EFPY
1 /4T, 236°F
3/4T, 107°F CURY PPLICABLE FOR HEATUP RATES UP TO 60°F/HA FOR THE SERVICE PERIOD UP TO EFPY 3000

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INDICATED T!M'E~ATU~E (*F) 383.4 Figuf'e 3. 4-2 Salem UnH 1 *~eactof' Coolant System Heatup Lim;tations Applicable up to 10 EFPY SALEM - UNIT l 3/4 4-26 Amendment No. 75

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ES APPLICABLE FOR COOLOOWN RATES UP TO 1oo*F/HR FOR THE SERVICE PE UP TO 10 EFPY

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Fiqure 3.4-JA S1lem Unit 1 Reactor Coolant Syst1111 Cooldown Limitations Appli.-

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( Excludinq Inserument Error Marqi.ns l SAUM - WIT l 3/4 4-27 Amendment No. 75

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Salem Unit 1 Reaeto~ Coolant System CooldO'tiln Limitations Appli-cable up to l 0 crPY ( IncludinCJ Instrument Error Margins)

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SSURE/TEMPERATURE LIMITS All c nents fn the Reactor Coolant System are designed to withst

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the effects o yclic la.ds due to system temperature and pressure chi Th*se cyclic 1 s are intraduced by no,,..al load transients, l"ttlctor

fps, and startup and tdown operations. The various c1t99ories of loa cycles used for d1sign pu oses are pi-ovided fn Section 4.1.S of the FS During s~rtup and shutd the rat.ls of temperature and pressure cha s are limited so that them imum specified heatup and cooldown rat are consistent with the design assump ns Ind satisfy the stress limits fo yclic operation.

During he1tup, the the 1 gradients in the l"elCta essel wtll produce thennal* stresses which v1ry CClllpr*ssive at the i r wall to tensile at the outer wall. These the induced cC111pressiv str"9sses tend to 1levi1te the tensile stresses i cld by the inte 11 pressur1. Th1r1fore, a pressure-temperature curve bas n st11dy stl conditions (i.1., no thennal stresses),..presents 1 lOllM nd of s111111r curves for finite h1atup rates when the inner will of t s tr1ated as the governing location.

The heltup analysis also cavers the

""inatian of pressure-temperature limitations for the cas1 i th1* aut*r wall of th1 v*ssel becanes the controlling location. T gradients established during heatup produce tensile stresses 1t J of th1 vess*l. Th*se stresses are addi tfve to th* pres e induc9d te i le stresses which ire already Rresent.

The the,,..11 i c9d stresses at e outer will of the vessel are tensile and are de ijent on both the ra of heatup 1nd the time along tht he!tYP rl!!!Pi

_refore, 1 l~r bound rve similar to that described for the hea of the inner wtll cannot e defined.

Consequently, for the c1s in which the outer wall of vessel becC111es the stress controlling ation, each he1tup rate of inte t must be analyzed on an i nt:1vi 1 blS is.

The h*atup 11 curve, Figure 3.4-2, is 1 ce111posite cu was prepared by ennining t~e most conservative cas*, with ei inside or outsf wall contl"Olling, for any h.. tup rate up to 60 hour6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

The co own limit curves, Figure 3.4-3, 11"9 composite curv which were R pared based upon the same type analysis with the excep on that the c trolling location is always the inside wtll where the cooldown ennal gradients tend to produce tensile stresses while produ *ng cC111pre ve stresses at the outside will. The he1tup and cooldown curve were prepared based upon the most limiting value of the predicted dj ed reference temperature at the end of 10 EFPY.

B 3/4 4-S Amendment No.

tar vtss1l 111t1rfJ,ls hive betn tutld ta d1t1n1f 111 1nftf1l R

the results of thes1 tests 1r1 shown 1n T1bl1 I R11ctor o ion Ind r1sultlnt fast ntutron (E>l Miv) irT"adi caus1 1n incr st 1n the RTNDT.

Thlrlfart, an adjusted ref c1 t*perature, bl upon tht.. Tlu1nc1 and copper content of_

  • terf11 1n question, can predicted using Figures I 3/4.'91 a
  • 3/4.'92.

Thi h*tuP Ind coo 11*1t curvH (Figur11 3.'9Z I

.'93) fnclud*

preclfctld 1djust11en far th1s thfft in RT at the of 10 EFPY, 11

.-11 11 adjusm.nts possible errors 1n"i1 prts and t*perature stnsing instruMnts.

The ICtul 1 Shi ft f ft r111 wf 11 lie II tab 1 fshed P11"fodfe11ly during operat IYl1uating, 1n 1ccordlnc1 with AS1M E185-70, t"91ctor v ldf at1on surv11111nc1 spec1Mns installed n*r the the ractor v1111l tn the core...... Since the ntutron s thl frnd1at1on smplts and v11s1l fnsfd1 radius are 11sentt1 nt1ca1, the M1surld transition sh1ft for a s*pl1 can be applied canffdenc1 to ttl9 ldjactnt stct1on of th* rHCtor vessal. The hMtu cooldown curves must lie Ncalculatld Wtlft the ART dettnlfntd frcm 11111nc1 ClPSU11 is different frcm the c111f 1ted ARTNDT for 1 1quf ltnt capsule l"ld1at1on aposure.

The pnssure-t*pera been provided to usu 111nts of Appendix 6 on Figure 3.'9Z far hydrostatic ttst1ng hive tn. t*perature requtre-Th* IU!Hr o~*.ae~ vtsH1 irT'ad1at1on *su llanc1 speci*tns and thl frequ1nc111 rmavtng and t11t1ng thl11 SPIC 1111 are tn 1ccordlnc1 w1 th th* r~ 1 of. Append 1x H ta 1 O CFR Part Th*""'

tatfons fllposld on prtssurizw h*tuP and c ldCMt and spray w t*perature d1ffertnt1a1 1r1 provided ta usu thlt th*

prtssur..,. is operatld within the design crftar11 asSUl!ed r the fati 1Nl1s11 perforwed in accordlnc* with thl ASME Cade il"9ents.

SALEM - UNIT 1 B 3/4 4-7 Amendinent No. 75

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SERVICE LlrE (EFFECllVE rULl POWER YEARS)

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FLUENCE *.i-2 IE > 1111VI Figure Bl/4.4-2Effect of fluence ind Copper ind Phosphorus Contents on AllTNOT for Reactor VeHel Steels A Weld Metil (Based on Capsule Y results) e Shell Plate 82402-J (Based on Capsule Y resultq)

TABLE B l/4.4-l

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SALEH UNIT I RCACTOI vcssn TOUGl*ESS DATA (llUIUOIATEO) a 50 ft lb TIDT 15*111 I 11NDT Hlterl*I I-'

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de No.

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110 Cl Hd Segment t

A51ll, Cl. 1 O. ll 0.010

-20 eg*

115 Cl Hd Segment Cl54 A5ll8 1 Cl.I 0.16 0.012

-lO 85*

122*

Cl Hd Seglleftt. 85852 A51ll, Cl.I 0.10 0.009

-50

'6*

112 Cl Hd fl.age 121"09 A50I, Cl.2 0.010 28*

22*

I IJIJ Vessel fl.age 5PI 19J A508, Cl.2 0.009 MJ*

o*

145 4'1019 Inlet loule 12JP40l 12408-1 50 144 Inlet loule 125P544 12408-2 A

26*

46 151 Inlet lloule 12lP40l 12408-l

A508, n*

41 161 Inlet loule 125P544 12408-4 A508,* Cl.

9*

11*

9 I 61 UI Outlet Jloule 111550 12409-1 ASCII, CI. 2 60*

95*

60 15 Outlet lloule ZT2:.~o 12409-2 A508, Cl.z 60*

95*

60 18 w

Outlet lloule ZT2585 82409-l A508, Cl.2 60*

10*

60 121 Outlet lloule ZT2585 12409-4 A508 0 Cl.2 60*

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60 126 Upper Shell MM91 12401-1 A5l11, C 0.22

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27 114 I

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Upper SMll A0495 12401-2 A5l

.I 0.1' o.on 20 122 0

Upper She II A0512 12401-l A

I Cl.I 0.24 0.011 IJ4*

14 96 l*ter Shell CIJ54 12402-1 lll, Cl.I 0.24 0.010 45 91 l1tter Shell Cll54 9'402-A5Jll, Cl. I 0.24 0.010

-5 112 l1tter Shell Cll97 82 AS1l8, Cl.I o.n 0.011

-1 121 l.,...r Shell Cl156 OJ-I A5l18, Cl.I 0.19 0.011 0

141

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L.,...r Shell Cll56 82401-2 ASlll, Cl.I O.H 0.012 118 ID La.er Shell Cl 82401-J A5J18, Cl.I

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J lot Hd Sf911ftl 82404-1 ASJlll, Cl.I 0.10 O.OOt 120

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ID lot Hd s 112404-2 ASJlll, Cl. I 0.11 0.010 112

J lot Hd 12404-l A51JI, Cl.I
o. 12 0.-

126 z

llnl 112405-1 ASJJI, Cl.I 0.15 0.010 106 0

0.16 0.019 O* I.J8..

0

- Norwl to Major WDrttng Direction U1

- Major WDrll1t9 Direction

- htl*ted per

  • Sbndlrd llevtew Pl*n 8r1nch TechnlCll Po1ttlon. llJEB 5-2

- Actual tr1nsverse ct.tci obtatned frOll surveillance progn* (fr01111ini1iu11 d.ald points).

ACTOR COOLANT SYST~~

ihe ERABILrTY of two POPSs or an RCS vent ooening of great* than 3.14 sauar inches ensures that the RCS will be protected from o ssure transients

'ch could exceed the limits of Aooendix G to 10 CF.

Part 50 when one or the RCS cold legs are less thaPf or eQual to..

  • 2°F.

Either POPS ha deauate relieving caoability to protect th CS from overoressurizati when the transient is limited to tither 1) the start of an idle with the secondary water temperatur f the steam generator less than r eaual to 50°F above the RCS col eg temperatures, or (2) the start of afety injection pump and its i ection into a water solid RCS.

3/4.4.10 STRUCTURAL INT

  • The inspection program or ASME Code Cl l, 2 and 3 components ensure that the structural in grity of thes components will be maintained at an acceptable level through the lift tht plant. To the extent applicable, the inspection prog for t 1 components 1s in complfanct with Section XI of tht ASME Boilt essure Vessel Code
  • SALEM
  • UNIT l B 3/4 4.,;ll Amendment No. 75

KEp\\a.Q (td.

MAnA PROPERTY BASIS:

CONTRO l'fG MATERIAL.

WELO METAl. (UPPER BOUND OF REGUL.ATCFW GUICE TREND CURVE)

.RTNQT INI o*F RTNOT AFTE EFPY

11' T, 15'79F
3/4 T, 1s*F CURVES APPl.lC 5 FOF1 H!:.ATUP RATES UP TO 60' F/Hrl FOR THE SERVICE PEAIO PTO 7 E?PY ANO CONT~ S MAi!C:1NS OF 1o*F ANO 60 PSiG FOR POSSIBLE INSTRUMENT RCR:i

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M - UNIT 2 3/4 4-2G Amendment No. 47

T MATERIAL PROPERTY BASIS:

-Pe p\\QcE:

CONTROL.L.ING MATERIAL : WELO METAL (UPPER BOUNO OF REGULATORY GUI

. TRENO CURVE)

T MOT INITIAL o*F NOT AFTER 7 EFP,X.

114 T. 1679F

3/4 T, 1s*F cu APPLICABLE Foa COOLOOWN RATES UP TO 100* F/HR FOR THE s ICE PERIOD
  • UPT EFPY ANO CONTAINS MARGINS OF 10" F ANO 60 PSiG FOR POSSI

_INSTRUMENT ERRO 3000.0*

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Figure 3. 4-3 Salem Unit 2 Reactor Coolant System Cooldown Umitatlon Applicable Up to 7 EFPY Amendment No. 47 SAI.EM -

UNIT 2 3/4 4-29

~o+ed SecTiot.J~ Mod"i4'ied per o...t\\o...c.he.d

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3J REACTOR COOLANT SYSTEM BASES 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 ASHE Boiler and Pressure Vessel Code,Section III, Appendix G.

l)

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 3.4-2 and 3.4-3 for the first full-power service period.

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 failure 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.

2)

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

3)

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

4)

The pressurizer he;tup 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.

5)

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

SALEM - UNIT 2 8 3/4 4-7

REACTOR COOLANT SYSTEM BASES 82-Th* fractu,../'toughn*ss properties of the f*rritic ~t.rfals in the r*actor v*ss*l are.~etel"lllin*d in accordance wfth th* NRC Standard Review Plan, ASTH El85~-and in accordance with additional reactor vessel requirements.

These properties are then evaluated in accordance with Appendix G of the 1976 SUllllll9r Addenda to S*ction III of the ASME Bof l*r and Pressure Vessel Code and the calculation 111ethods dHcribed in WCAP-7924-A, "Basis for Heatup and Cooldown Li*it Curves, April 1975. 11

  • Heatup and cooldown"limit curves.are calculated using the most limiting valu* of the nil-ductf lfty reference temperatur*, r.J°NDT' at the end of.

IS )C' *ffectiv* full pow.r y.ars of service life. Th* JC" EFPY service lffe period is chos*n such that th* limiting RTNDT at the 1/4T location in the core region is great.r than th* RTNDT of th* lfmfting unirradiated material.

Th* s*l*ction of such a limiting *RTNDT assures that all COlllPOn*nts in the R*actor Coolant Syst.. will b* op*rat*d conservativ*lY in accordance with applicabl* Cod* r*quir91119nts.

Th* reactor vHsel mat.rials hav* bun ust.d to det*rmin* their initial RTNDTi th* results of th*s* tests are shown in Table B 3/4.4-1.

Reactor op*ration and resultant fast n*utron CE greaur than 1 MEV)*irradiation can cause an increas* in th* RTNDT" Tberefor*, ~n adjust.d reference telllp*rature, bas*d upon th* flu*nc* and copper content of the mater &l in quHtfon, can b* predicted lising F4gitPU B 3/4.4 l enel B 31'4.4 2.

The h*atup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustm.nts for this shift in RTNDT at the end of~ EFPY, ~

well a1 aajY1'tllieR~ fer pessible errors in the press~re and temperat~re SeRliRi iR1tr1ateRt~.

Values of aRTNDT det*rmined *in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available.

The first ea,ss~le will be r1110~ed at ~he end of the first eere eyele.

s~eeess1~e ~apsules will be removed in accordance with the requirements of ASTM El85-~ and 10 CFR Part 50, Appendix H.

The heatup

02.

SALEM -

UNIT 2 B 3/4 4-8

. :-:-_.._~-;--::::-

10 30 SERVICE LIFE, EFPY FIGURE B 314.4-1 l'AST NEUTRON FLUENCE (E>1 MEVI AS A FUNCTION OF FULL POWER SERVICE LI FE SAL.EM - UNIT 2 B 3/4 4-9

t.

200*~~;+;.~~.+::m~+-l-i-+-~+/--l~~~==-...... IF'.-~~

.. i SALEM - UNIT 2

~- -**

2 5

1019 2

5 Fl.U!NCE (N/CM2 >t MEVI is.,. II 314.4-2 Eff9ct of Fl1*a Ind ~

Comlnt en Shift of RT NDT 'or R-=r V....it Expotld ta 550° F Tllnl*WUM B 3/4 4*10

./

Vl TABLE 8 3/4.4-1

~

r

~

REACTOR VESSEL TOUGHNESS c :z

--i N

50 ft-lb Prtnctpal 1NDT 35 *tl Working p

I:;-;

Dtrectton C!!!!!onent Heat No.

m ru

{ft-lb}

Closure Head DOIMt C4377-3 A533Bil:Ll

-40 82.5*

127 DJ Closure Head Peel C4684-3 II 97*

149 zj w

II II II C4417-4 II O*

84*

129

[l)

II C4417-1 6*

84*

129.5

-0 I

r Closure Head Flange 123X270VA1 A508CIL2 28*

104*

160

}>

0 Vessel Flange 123W343VA1 12*

107*

164 rn Inlet Nozzle ZV-3265-1 II 60*

> 72*

) 111**

C7 II II ZV-3265-2 0.010 60*

> 94**

II II 0.010 60*

32*

> 109**

II II 0.011 60*

40*

60*

123.5 Outlet Nozzle 0.006 60*

8*

60*

II II u

0.006 60*

20*

60*

75*

II u

AV-2067 II 0.007 28*

8*

28*

82*

II AV-2099 II 0.007 60*

40*

60*

77*

I/)

TABLE B 3/4.4-1 (Continued)

)>

r

!i!

REACTOR VESSEL TOUGHNESS c :z

'.::1 Average Upper I\\)

Shelf Enerlll Hon11l to 50 ft-lb Principal TNDT 35 *11 RT NOT Working Cu p

Temp DfrecUon Component m m rn

{Of}

~ ft-1 Upper Shell C4194-l 0.11 0

50*

134 CD II II C4194-2

-10 60*

122 w

.:;rj II II C4171-1 II

-10 28*

69*

107

.p.

!Tl

.p.

Inter. Shell C4173-1 II 0

105 138

-0 I

I\\)

II II C4186-2 II 12 97 r

127.5 )>

II II C4194-2 II 10 107 116

('\\

IT)

Lower Shell C4182-1 A5338Cll 127 t7 II II C4182-2 0.010 68 135.5 II II 88343-1 0.12 0.012

-10 70 135.5 Bottom Head Peel II

0. 12 0.010

-30 54*

-6*.

139 II II II

0. 12 O.OJ.1

-20 42*

-18*

89*

II II 0.11 0.009

-20 71*

11*

93*

C4367-4 II

o. 12 0.009

-30 60*

O*

77*

c:

z

-I N

CO!Ponent Inter and Lower Shell Vert. Welds Inter to Lower Shell Girth Weld Heat Ho.

TABLE B 3/4.4-1 (Conttnuad)

REACTOR VESSEL TOUGHNESS Cu P

m m Principal Working Direct ton (ft-lb)

REACTOR COOLANT SYSTEM BASES and cooldown curves must be recalculated when the ART determined from th* surveillance capsule exceeds the calculated ARTND~+or the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various h**tup 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 Part SO and these methods ar* 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 (LEFM) technology.

In the calculation procedures a s... i-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 di1119nsions 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 develop~d for this reference crack are conservative and provide sufficient safety margins for protection against nonductil* failure.

To assure that the radiation embrittle111nt effects are accounted for in the calculation of the li*it curves, the 110st limiting value of the nil-ductility reference temperature, RT

, is used and this includes the radiation induced shift, ARTNRT' ~~responding to the end of the period for which heatup and cooldow curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that t.he t1:1ul stress intensity 1'actar, K, for the combined thermal and pressure stresses at any time during heatup o~ cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time.

K1; is obtained from the reference fracture toughness curve, defined in AppenaTx G to the ASME Code.

The KIR

. curve is given by the equation:

KIR = 26.78 + l.~3 exp [0.0145(T-RTNDT + 160)]

(1) where KyH is th* reference stress intensity factor as a function of the metal temperat re T and the metal nil-ductility reference temperature RTN

. Thus, the governing equation for the heatup-cooldown analysis is defined q~ Appendix G of the ASME Code as follows:

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

SALEM -

UNIT 2 B 3/4 4-14

REACTOR COOLANT SYSTEM BASES Kit is the stress intensity factor caused by the ther11al gradients.

KIR is provided by the code as a function of temperature relative to the RTNDT of the material.

C = 2.0 for level A.and 8 'ervice limits. and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient. K1R is determined by the metal temperature at the tip of the postulated flaw. tne appropriate value for RT

  • and the reference fracture toughness curve.

The thermal stresses re~BTting fro11 temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors. Krt* for the reference flaw are computed.

From Equation (2) the pressure stress intensity factors are o~tained and from these the allowable pressures are calculated.

COOLDOWN 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 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. Allowable pressure-temperature relations are generated for both steady-state and finite cooidown rate situations.

rrOll 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 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 l/4T vessel location is at a ~igher 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 4T developed during cooldown results in a higher value of K at the l/4T location for finite cooldown rates than for steady-state opl~ation. Furthermore, if conditions exist such that the increase in KIR exceeds Kit' the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on tempera-ture at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

SALEM - UNIT 2 B 3/4 4-15

REACTOR COOLANT SYSTEM BASES HEATUP Three separate calculations are required to determine the limit CUY"Ves for finite heatup rates.

As is done in the cooldown analysis, allowable* pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the pr9sence of a 1/4T defect at the inside of the vessel wall.

The thel'llal gradients during heatup.produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by intarnal pressure.

The 111tal temperature at the crack tip lags the coolant temperaturei therefore, the K for the l/4T crack during heatup f s lower than the Klg for the 1/4T crack du~¥ng steady-state conditions at the sime coolant temper ture. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K s for steady-state and finite heatup rates do not offset each other and th1Rpressure-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-te11perature 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 course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup r1111P.

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

Rather, each heatup rate of interest must bt analyzed 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 produced as follows.

A composite curve is constructed based on a point-by-point comparison of the steady-stat* and finite heatup rate data.

At any given temperature, the allowable pressure fs 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 limita-tions because it is possible for conditions to exist such that over the course of the heatup ramp the controlling ~ondition 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 4-16

ATTACHMENT 3 TECHNICAL SPECIFICATION CORRECTED PAGES

Material Property Initial RTNDT:

45°F RTNDT After lSEFPV:

3/4T, 162°F 2500 2250 2000 1750

~ 1500 11'1

~

~ 1250 11'1 11'1 lo.I f 1000 Q

bol..

u -

Q z 750 500 250 50 100 I

I I

I I

I I

I I

I I

I I

I I

I I

I 11 Leak Test I

Limit

'j j

I Hutup Ratu

. I Up To 5**F/Hr I/

J U naccaptab 1 a 11 Opartt1on 17 I/

I/

I/

~

Acceptable Operation 150 200 250 300 I

j IJ I/

I I

j I

I I

j II

.rt tfc:a 1 t)

Lfmft Bu-*

n Inse,-,fc
cr Hydrostatic Test Tel!IP.

(33i'9F) fo, ttle Sarv1c..

Parfod Up 15 EFPY I I I

I I

I I

I 350 400 INOICATED TEMPERATURE (DEG.r)

CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS 450 Figure 3.4-2

  • Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 60°F/HR for the Service Period up to 15 EFPY SALEM -

UNIT 1 3/4 4-26 I

I 500

Material Property Basis Initial RTNDT:

-56°F RTNDT After 15 EFPY:

l/4T = 222.5°F 2500 2250 2000 1750

~ 1500

~ 1250

~

w g:

1000 Q

w...

c

<.l -

Q z -

750 500 250 i-- Cool down Rates

  • F/Hr o-

~

E~

50 U111ccepuble Oper1t1on

//

1~

/

I' i....

~--

100 150 200 I

I I

I J

J

~

van I'/ '/J r_,,,,

'/ j Accept&Ole Oper1t1on 250 300 350 INDICATED TEMPERATURE (DEC.F')

CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS I

I I

Figure 3.4-3 Salem Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100°F/HR for the Service Period up to 15 EFPY SALEM -

UNIT l 3/4 4-27 500

REACTOR COOLANT SYSTEM BASES 3/4.4.9 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.

1)

The reactor coolant temperature and pressure and system heatup and cooldown rate (with the exception of the pressurizer) 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 telfiperature 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 nondoctile failure 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.

2)

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

3)
4)
5)

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

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.

System preservice hydrotests and in-service leak and hydrotests shall be performed 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 vessel are determined in accordance with the NRC Standard Review Plan, ASTM El85-82, 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 Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves, April 1975 11 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTllJDT' 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 RTNDT of the limiting unirradiated material.

The selection of such a-*Iimiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

SALEM - UNIT 1 B 3/4 4-6

~------------------------------"

REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are Shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT' An adjusted reference temperature, (ART),

based upon the fluence and tfie copper and nickel content of the material in question, can be predicted.

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

ART = Initial RTNDT *+

RTNDT + Margin Initial RTNDT is the reference temperature for the unirradiated material.

RTbIDT is-*ine mean value of the adjustment in reference temperature caused by the irradiation and is calculated as follows:

RTNDT = 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-l.

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:

-0.24X f = (f surface) x (e

)

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

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=2P.

If a meas~ value of initial RTNDT for the material in question is used, a-1 may be t~ as zero. If generic value of initial RTNDT is used. cr1, should be obtain3a* from the same se5 of data.

The standard aeviations. for ARTNDT' "fi,are 28 l for welds and 17 F for base metal. except that O""A need nof exceed 0.50 times ~he mean value of ARTNDT surface.

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

SALEM - UNIT 1 B 3/4 4-7

REACTOR COOLANT SYSTEM BASES Values of

~RTNDT determined in this manner may be used until the results from the materiaI surveillance program, evaluated according to ASTM El85, 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 curves must be recalculated when the.C:,RTNDT determined from the surveillance capsule exceeds the calculated

,6RTNDT for ehe equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various 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 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 (LEFM) 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 against nonductile.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 temperature, RTmJT' is used and this includes the radiation induced shift,.6RTNDT corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASHE approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K., for the combined thermal and pressure stresses at any time during heatup ot cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time.

KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.

The KIR curve is given by the equation:

KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)]

(1) where KIR is the reference stress intensity factor as a function of the metal temperaEure T and the metal nil-ductility reference temperature RTNDT"

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

(2)

SALEM - UNIT 1 B 3/4 4-8

REACTOR COOLANT SYSTEM BASES where KIM is the stress intensity factor caused by membrane (pressure) stress.

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

KIR is provided by the code as a function of temperature relative to the R'rNnT of the material.

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

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

, and the reference fracture toughness curve.

The thermal stresses resulting from temperatue gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT' for the reference flaw are computed.

From Equation (2) the pressure stress intensity factors are obtained and from these the allowable pressures are calculated.

COOLDOWN 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 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.

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 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 vessel ID.

This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the ~T developed during cooldown results in a higher value of KIR at the l/4T location for finite cooldown rates than for steady-state operation.

Furthermore, if conditions exist such that the increase in KIR exceeds KIT' the calculated allowable pressure during cooldown will be greater than tfie steady-state value.

The above procedures are needed because there is no direct control on temperature at th@ 1/4T location, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

SALEM - UNIT 1 B 3/4 4-9

REACTOR COOLANT SYSTEM BASES HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stress 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 therefore, the KIR for the 1/4T crack during heatup is lower than the KIR for the 1/4T cr~ck auring steady-state conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K1Rs for steady-state and finite 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 l/4T deep outside surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surf ace during heatup produce 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 ramp.

Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be 4efined.

Rather, each heatup rate of interest must be analyzed 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 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 limitatio"8 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 1 B 3/4 4-10

REACTOR COOLANT SYSTEM BASES 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 temperature of the closure flange regions must exceed the material RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Salem).

Table B3/4.4-1 indicates that the limiting RTNDI of 28°F occurs in the closure head flange of Salem Unit 1, and the minimum a lowable temperature of this region is 148°F at pressures 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.

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 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 312°F.

Either POPS has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the 0

secondary water temperature of the steam generator 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 solid RCS.

SALEM - UNIT 1 B 3/4 4-11

Cl z H

1-3 TABLE

  • 4-1 SALEM UNIT 1 REACTOR VESSEL TOUGHNESS DATA Plate No.

Material Cu Ni Component or Weld No.

Type

( % )

( % )

Cl Hd Dome B2407-l A533B, Cl.I 0.20 0.50 Cl Hd Segment B2406-l A533B, Cl.1 0.13 0.52 Cl Hd Segment B2406:-2 A533B, Cl.I 0.16 a.so Cl Hd Segment B2406-3 A533B, Cl.I 0.10 0.53 Cl Hd Flange B2811 AS08, Cl.2 0.12 Vessel Flange B2410 AS08, Cl.2 0.67 Inlet Nozzle B2408-l A508, Cl.2 0.68 Inlet Nozzle B2408-2 A508, Cl.2 0.71 Inlet Nozzle B2408-3 A508, Cl.2 0.66 Inlet Nozzle B2408-4 AS08, Cl.2 0.65 Outlet Nozzle B2409-l A508, Cl.2 0.69 Outlet Nozzle B2409-2 AS08, Cl.2 0.69 Outlet Nozzle B2409-3 A508, Cl.2 0.74 Outlet Nozzle B2409-4 A508, Cl.2 0.74 Upper Shell B2401-l A533B, Cl. l 0.22 0.48 Upper Shell B2401-2 A533B, Cl.I 0.19 0.48 Upper Shell B2401-3 A533B, Cl.I 0.24 0.51 Inter Shell B2402-l A533B, Cl.I 0.24 0.52 Inter Shell B2402-2 A533B, Cl.1 0.24 0.50 Inter Shell B2402-3 AS33B, Cl. l 0.12 0.50 Lower Shell B2403-l A533B, Cl.I 0.19 0.48 Lower Shell B2403-2 A533B, Cl. l 0.19 0.49 Lower Shell B2403-3 AS33B, Cl.I 0.19 0.48 Bot Hd Segment B2404-l A533B, Cl.l 0.10 0.52 Bot Hd Segment B2404-2 AS33B, Cl.I 0.11 0.53 Bot Hd Segment B2404-3 A533B, Cl.I 0.12 0.52 Bot Hd Dome B2405-l A533B, Cl.I 0.15 o.so Circum. Weld 8-042 0.22 1.02 Circum. Weld 9-042 0.25 0.72 Vertical Weld 2-042 0.18

1. 00**

Vertical Weld 3-042 0.19

1. 00**
  • Estimated per NRC Standard Review Plan Section 5.3.2.

-30

-20

-30

-so 28*

60*

50*

46*

47*

9*

60*

60*

60*

60*

-30 0

-10

-30

-30

-40

-40

-70

-40 10

-50 10.

-20

      • Estimated per Pressurized Thermal Shock Rule, 10CFR50.61 so ft lb JS-Mil Temp (OF) 99*

89*

85*

66*

22*

O*

43*

26*

37*

17*

95*

95*

10*

13*

87*

80*

ll4*

105 55 57 70 86 90 48*

60*

47*

57*

Average Upper Shelf Enerov Normal to Principal Principal Working Working RTNDT Direction Direction (OF)

(ft-lb)

(ft-lb) 39

71. 5*

llO 29 97*

125 2S 79*

122 6

86*

132 28 129*

199 60 94*

14S so 94*

144 46 102*

157 47 105*

161 9

108.5*

167 60 48*

75 60 51*

78 60 79*

121 60 82*

126 27 74*

ll4 20 79*

122 34 62*

96 45 73 97

-5 91 112

-3 104 127 i'o 99 143 26 94 128 30 102 131 10 78*

120 0

86*

132 10

~2*

1:.:.6

-3 69*

106

-56***

-56***

-56***

-56***

TABLE B 3/4.4.2

~~T JA.CTDI J'OI IELI>S. *r

Copper, lfickel 1 Wt-I lt.-1 o

0.20 o.40 o.50 o.ao 1.00 1.20 0

20 20 20 20 20 20 20 0.01

  • 20 20 20 20 20 20 20 0.02 21 28 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 64 M

54 0.05 28 49 87 51 58 88 ea o.oe 20 52 77 12 82

_82 82 0.07 32 55 as gs us OS 95 0.08 38 S8 go 106 108 108 108 0.09 40 81 M

us 122 122 122 0.10 55 97 122 133. 135 135 0.11 49 51 101 130 1"

148 148 0.12 52 72 103 135 153 151 181 0.13 SI 78 105 130 182 172 178 0.14 Bl 70 109 142 158 182 188 0.15 55 14 112 145 175 191 200 0.15 70 88 115 140 178 190 211 0.17 75 02 110 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 150 194 223 245 0.21 92 108 133 154 197 229 252 0.22 07 112 137 157 200 232 257 0.23 101 117 140 159 203 235 263 o.:u 105 121 1"

173 208 239 258 0.25 110 128 148 175 209 243 272 0.25 113 130 1S1 180 212 245 276 0.27 119 134 155 184 215 240 280 0.28 122 138 150 187 218 251 284 0.29 128 142 154 191 222 264 287 0.30 131 148 157 194 225 257 290 0.31 138 151 172 198

  • 228 250 293 0.32 140 155 175 202 231 253 295 0.33 1"

150 180 205 234 255 299 0.34 149 154 184 209 238 259 302 0.35-153 158 187 212 241 272 305 0.35 158 172 191 215 245 275 308 0.37 182 177 195 220 248 278 311 0.38 15G 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 SALEM UNIT.1 B 3/4 4-13

TABLE B 3/4.4-3 cmMISftY 1~C'IOI 101 BAD DUL, *r

Copper, llickel 1 1ft-I ft-I

-2.,. 2.:J2 ~

2.:,!2 0. IO 1.00 1. 20 0

20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02

~

20 20 20 20 20 20 0.03 20 20 20 20 20

~

20 O.CM 22 28 28 28 28 28 28 O.OI 25 31 31 31 31 31 31 o.08 21 37 37 37 37

-37 37 0.07 31 43 " " " " "

O.OI 34 41 51 61 11 11 61 0.09 37 53 S8 51 51 51 A

0.10 41 A

SS SS 87 87 87 0.11 45 82 72 74 77 77 77 0.12 49 87 79 13 18 18 18 0~13 53 71 u

91 98 H

H 0.14 67 76 91 100 105 108 108 0.15 81 80 99 110 116 117 117 0.18 SS 14 l<M 118 123 125 125 0.17 89 88 110 127 132 135 135 0.11 73 92 115 134 141 144 14' 0.19 71 97. 120 142 150 154 154 0.20 12 102 125 149 159 184 185 0.21 18 107 129 156 187 172 174 0.22 91 112 134 181 178 111 114 0.23 95 117 131 187 184 190 194 0.24 100 121 143 172 191 199 204 0.25 l<M 128 141 178 199 20I 214 0.28 109 130 151 180 205 218 221 0.27 114 134 156 18' 211 225 230 0.21 119 131 180 117 218 233 239 0.28 124 142 184 191 221 241 241 0.30 128 1'8 187 lM 225 249 257 0.31 134 111 172 191 221 255 288 0.32 139 151 176 -

231 280 274 0.31 14' 180 llO -

234 284 212 0.34 149 18' 114 208 231 281 280 0.31-153 181 117 212 241 272 281 0.38 151 173 191 218 245 276 303 0.37 102 177 191 220 241 271 30I 0.31 188 112 200 223 250 211 313 0.39 171 115 203 227 254 215 317 0.40 176 189 207 231 257 211 320 SALEM UNIT i B 3/4 4-14

'/

I lO~_D_

7 __ _

6----~

5 __ _

4 __ _

2---

11_1_1 II. -

7 __

6._

5 __

N' g*--

c

.... 3 __

c., =

~L c f.. =

z:

II.

7.
6.

5...

3.
2.

Fluence at Vessel IR 45° Az11111th (Max11111 Po1nt) 11 11.LL---L...... ---i~--..L....~'-'..,.+----.i.---~lS~--..L.---~2~0----L.-~2§=----.1.----::3~0----...... ---

l Service Life (Effective Full Power-Years)

Figure B 3/4.4-1 Fast neutron fluence CE>lMeV) as a function of of full power service life (EFPY)

SALEM -

UNIT 1 B 3/4 4-15

[/l

i:-

t""

t:tJ :s:

c:::: z H

t-3 I-'

1 0..

-~

0 0

t; *

u.

b:!

I w

1

p.

It I,_.

°'

z.

10" 10" 10"

,._...* n/cm1 CE > 1 MeVI Fluence factor for use in the expression for~RTNDT FIGURE B 3/4 4-2 117111 lo"

REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY 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 Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.SSa(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

SALEM UNIT 1 B 3/4.4-17

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:

COPPER CONTENT:

NICKEL CONTENT:

INITIAL RTNDT:

RTNDT AFTER 10 EFPY:

LONGITUDINAL WELD 0.35 WT%

1.00 WT%

-56°F 1/4 T, 178.6°F 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~ 'I I I I I I I

I I I I I I I I I I I I

'I I I I I I I I I I I r

ZZ50 Leak Test I

,j L 1mft I

zooo I

I I

1750 l

1500 Unacceptable Operation

a.

..... I 1250 w f 1000 Cll w

~

c

~

750 i

500 250 Heatup Rates

~

ug To 60 F/Hr I/

Crft1cal 1t,:* L fmit Acceptable Based on Inservf ce Operation Hydrostatic Test Temp. (J11*F) for the Service Period Up To 10 EFPY I I I I I I I I I I I I 1-.111111-i-i-IT 50 100 150 200 250 300 350 400 450 500 INDICATED TEMl'ERATUIC (DEG.r)

I 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

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:

COPPER CONTENT:

NICKEL CONTENT:

INITIAL RT NDT:

RTNDT AFTER 10 EFPY:

LONGITUDINAL WELD 0.35 WT%

1.00 WT%

-56°F l/4T, 178.6°F 3/4T, 116.1°F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100°F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS 2500 "lCI I

ZZ'50 zooo 1750 I

c;-

1500 IL -! 1250 w

a:

IL 0 w c

1000

~

7'50 0 z

'500 250

!/

unacceptable Operation rl Coolcown

.... ~

'/

Rates

~ -~

°F/Hr

,,.,~

Acceptable Operation L.....

0 20 -

v 40 -

  • ~

-*1 :~

50 10.{)

150 zoo 250 300 350 400 450 500 INDICATED TEa.ERATUltE (DEG.r) 1 Salem Unit 2 reactor coolant system cooldown limitations applicable for the first 10 EFPY FIGURE 3.4-3 SALEM -

UNIT 2 3/4 4-29

REACTOR COOLANT SYSTEM BASES 3/4.4.9 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.

1)

The reactor coolant temperature and pressure and system heatup and cooldown rate (with the exception of the pressurizer) 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 nondoctile failure 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.

2)

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

3)
4)
5)

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

  • 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.

System preservice hydrotests and in-service leak and hydrotests shall be performed 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 vessel are determined in accordance with the NRC Standard Review Plan, ASTM El85-82, 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 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 cooldo:wn limit curves are calculated using the most limiting value of the nil-ductility reference temperature, 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 RTNDT of the limiting unirradiated material.

The selection of such a-*11rniting RT T assures that all components in the Reactor Coolant System will ~ operated conservatively in accordance with applicable Code requirements.

SALEM - UNIT 2 B 3/4 4-6

REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT T"

An adjusted reference temperature, (ART),

based upon the fluence and ~e copper and nickel content of the material in question, can be predicted.

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

ART = Initial RTNDT + ARTNDT + Margin

. Initial RTNDT is the reference temperature for the unirradiated material.

ARTNDT is tfie mean value of the adjustment in reference temperature caused by the irradiation and is calculated as follows:

ARTNDT = 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:

-0.24X f = (f surface) x (e

)

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

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~I 2 + c:r-8 2

If a measured value of initial RTNDT for the material in question is used, err may be taken as zero.

If generic value of initial RTID)T is used, CTI should oe obtained from the same se5 of data.

The standard deviations, for ARTNDT'. a-8,

are 28°F for welds and 17 F for base metal, except that a-8 need not exceed 0.50 times the mean value of ARTNDT surface.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for tEis shift in RTNDT at the end of 15 EFPY.

SALEM - UNIT 2 B 3/4 4-7 I

I

REACTOR COOLANT SYSTEM BASES Values of 6 RTNDT determined in this.manner may be used until the results from the materiaI surveillance program, evaluated according to ASTM E185, are available.

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

The heatup and cooldown curves must be recalculated when the 6 RTNDT determined from the surveillance capsule exceeds the calculated 6 RTNDT for Ebe equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various 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 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 (LEFM) 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 against nonductile 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 temperature, RT~T' is used and this includes the radiation induced shift, 6RTNDT 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 various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined therma] and pressure stresses at any time during heatup ot cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time.

K1R is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.

The KIR curve is given by the equation:

KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)]

(1) where KIR is the reference stress intensity factor as a function of the metal temperaEure T and the metal nil-ductility reference temperature RTNDT"

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

(2)

SALEM - UNIT 2 B 3/4 4-8

REACTOR COOLANT SYSTEM BASES where KIM is the stress intensity fac~or 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 Rt~T of the material.

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

At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, tfie appropriate value for RTNDI, and the reference fracture toughness curve.

The thermal stresses resu ting from temperatue gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT' for the reference flaw are computed.

From Equation (2) the pressure stress intensity factors are obtained and from these the allowable pressures are calculated.

COOLDOWN 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 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.

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 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 vessel ID.

This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the ~ T developed during cooldown results in a higher value of KIR at the 1/4T location for finite cooldown rates than for steady-state operation.

Furthermore, if conditions exist such that the increase in KIR exceeds KIT' the calculated allowable pressure during cooldown will be greater than tfie steady-state value.

The above procedures are needed because there is no direct control on temperature at the l/4T location, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

SALEM - UNIT 2 B 3/4 4-9

REACTOR COOLANT SYSTEM BASES HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l/4T defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stress 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 therefore, the KIR for the l/4T crack during heatup is lower than the KIR for the l/4T crack auring steady-state conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K1Rs for steady-state and finite 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 l/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 l/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 course, are 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 basis.

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 4-10

REACTOR COOLANT SYSTEM BASES 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 temperature of the closure flange regions must exceed the material RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Salem).

Table B3/4.4-1 indicates that the limiting RTNDI of 28°F occurs in the closure head flange of Salem Unit 1, and the minimum a lowable temperature of this region is 148°F at pressures 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 fatigile 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 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 312°F.

Either POPS has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the 0

secondary water temperature of the steam generator 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 solid RCS.

SALEM - UNIT 2 B 3/4 4-11

c:: z H

t-3 I\\.)

4-1 SALEM UNIT 2 REACTOR VESSEL TOUGHNESS DATA Plate No.

Material Cu Ni Component or Weld No.

Type

( % )

( % )

Closure Hd Dome B4708 A533BCL1 0.11 0.70 Closure Hd Peel 85007-3 A5338CL1 0.12 0.57 Closure Hd Peel 84707rl A5338CL1 0.10 o.ss Closure Hd Peel 84707-3 A5338CL1 0.13 0.63 Closure Hd Flng B4702-l A508CL2 0.68 Vessel Flange B5001 A508CL2 0.70 Inlet Nozzle 84703-1 A508CL2 0.69 Inlet Nozzle 84703-2 A508CL2 0.69 Inlet Nozzle 84703-3 A508CL2 0.68 Inlet Nozzle 84703-4 A508CL2 0.81 Outlet Nozzle 84704-1 A508CL2 0.84 Outlet Nozzle 84704-2 A508CL2 0.77 Outlet Nozzle 84704-3 A508CL2 0.69 Outlet Nozzle 84704-4 A508CL2 0.71 Upper Shell 84711-1 A533BCL1 0.11 0.55 Upper Shell 84711-2 A5338CL1 0.14 0.56 Upper Shell 84711-3 A5338CL1 0.12 0.58 Inter. Shell 84 712-1 A5338CL1 0.13 0.56 Inter. Shell 84712-2 A533BCL1 0.14 0.60 Inter. Shell B4712-3 A533BCL1 0.11 0.57 Lower Shell B4713-l A5338CL1 0.12 0.60 Lower Shell 84713-2 A533BCL1 0.12 0.57 Lower Shell 84713-3 A533BCL1 0.12 0.58 Bottom Hd Peel B4709-l A533BCL1 0.12 0.60 Bottom Hd Peel B4709-2 A533BCL1 0.12 o.58 Bottom Hd Peel 84709-3 A533BCL1 0.11 0.56 Bottom Head B4710 A533BCL1 0.12 0.60 Circum. Weld 8-442 0.28 0.74 Circum. Weld 9-442 0.175 0.20****

Vertical Weld 2-442 0.23 0.73 Vertical Weld 3-442 0.20 0.86 Estimated per NRC Standard Review Plan Section 5.3.2.

100% Shear not reached T NOT (OF)

-40

-20 0

0 28*

12*

60*

60*

60*

60*

60*

60*

28*

60*

O*

-10

-10 0

-:.w

-so

-10

-20

-10

-30

-20

-20

-30

      • Estimated per Pressurized Thermal Shock Rule, 10CFR50.61
        • Estimated value for type MIL B-4 wire heats.

50 ft lb 35-Mil Temp

( o F) 45*

15*

51*

66*

39*

4*

62*

25*

32*

40*

8*

20*

8*

40*

50*

60*

88*

<60

n.

70 68 68 70 54*

42*

71*

60*

Average Upper Shelf Enerav Normal to Principal Principal Working Workinq RTNDT Direction Direction (OF)

(ft-lb)

(ft-lh)

-15*

82.5 127

-20*

97*

149 O*

84*

129 6*

84*

129.5 28*

104*

160 12*

107*

164 60*

>72*

>lll**

60*

>61*

> 94**

60*

>71*

>109**

60*

80*

123.5 60*

82*

126 60*

75*

116 28*

82*

126 60*

77*

119 O*

87*

134 O*

79*

122 28*

69*

107 0

105 138 Lr:

11 LO.'.>

,10 107 116 8

98 127 8

103 135.5 10 122 135.5

-6*

90*

139

  • -18*

89*

137.5 11*

93*

143 O*

77*

118

-56***

-56***

-56***

-56***

TABLE B 3/4.4.2

~T lACTOI 101 IBLDS 1 *r

Copper, Jfickel 1 Wt-I lt.-1 o

0.20 o."6 o.50 o.eo 1.00 1.20 0

20 20 20 20 20 20 20 0.01

  • 20 20 20 20 20 20 20 0.02 21 25 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.06 24 43 54 54 54 54 54 o.os 25 49 87 S8 58 SS 58 o.os 20 52 11 12 12 -12 12 0.01 32 55 85 95 95 95 95 0.08 35 58 90 105 108 108 108 0.09 40 81 N

115 122 122 122 0.10 44 es 97 122 133 135 135 0.11 49 SI 101 130 144 141 148 0.12 52 72 103 135 163 181 181 0.13 58 7S 1oe 139 182 172 178 0.14

&1 79 109 142 158 182 188 0.15 88 84 112 148 175 191 200 0.18 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 106 129 150 194 223 245 0.21 92 108 133 184 197 229 252 0.22 97 112 137 187 200 232 257 0.23 101 117 140 189 203 238 263 0.24 105 121 144 173 208 239 288 0.25 110 125 148 178 209 243 272 0.28 113 130 151 180 212 248 27.8 0.27 119 134 155 184 215 249 280 0.28 122 138 150 117 211 251 284 0.29 121 142 154 191 222 254 287 0.30 131 141 lfS7 194 225 257 290 0.31 138 151 172 191

  • 221 2'50 293 0.32 140 156 175 202 231 283
  • 208 0.33 144 150 180 20S 234 288 299 0.34 1.t9 184 184 209 238 289 302 0.35.

153 158 187 212 241 272 305 0.3&

158 172 191 218 245 275 308 0.37 lfS2 177 198 220 248 278 311 0.38 lfSIS 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 SALEM UNIT 2.

B 3/4 4-13

TABLE B 3/4.4-3 CBMISTIY PAC!O POI BASI MBTAL, *p

Copper, lfickd 1 'It-I ft-I o

0.20 o.40 o.eo o.so 1.00 1.20 0

20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 28 28 28 28 28 28 O.OI 25 31 31 31 31 31 31 o.08 21 37 37 37 37

-37 37 0.07 31 43 44 44 44 44 44 O.OI 34 41

&1

&l 51

&1 51 0.09 37

&3 58 SI SI SI SI 0.10 41 sa es es 17 87 97 0.11 4S 82 72 74 77 77 77 0.12 49 87 79 13 18 ae ae 0~13

&3 71 IS 91 98 98 98 0.14 57 75 91 100. 105 108 108 0.15 81 IO 99 110 115 117 117 0.18 es 84 104 118 123 125 125 0.17 9g 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 O.lD 78 97 120 142 150 154 154 0.20 82 102 125 149 159 184 1es 0.21 ae 107 129 155 187 172 174 0.22 91 112 134 181 178 181 114 0.23 96 117 138 187 184 190 1D4 0.24 100 121 143 172 lDl 19g 204 0.25 104 128 141 178 19D 208 214 0.28 109 130 151 180 205 218 221 0.27 114 134 155 114 211 225 230 0.21 119 138 ll50 187 218 233 239 0.29 124 142 184 191 221 241 248 0.30 129 148 187 lM 225 249 257 0.31 134 151 172 191 221 25&

288 0.32 139 15&

175 202 231 2l50 274 0.33 144 180 llO -

234 284 212 0.34 149 184 114 209 231 288 290 0.31 -

153 tel 187 212 2'1 272 298 0.38 151 173 191 218 24S 27&

303 0.37 182 177 198 220 241 271 308 0.38 188 182 200 223 250 211 313 0.39 171 115 203 227 254 215 317 0.40 175 189 207 231 257 211 320 SALEM UNIT 2 B3/4 4-14

~

'""'= 10 u

c:

19 IR THIS CURVE REPRESENTS THE FLUENCE AT THE INNER RADIUS OF THE LIMITING LONGITUDINAL WELD SEAM LOCATED AT THE 30 AZIMUTH 300 AZIMUTH 1017~...L-J...-L-.&....-1-..&..-L......L..-.L.....L.-..

....L-...L.......ii.-...i..------

o 2

4 6

8 10 12 14 16 18 20 22 24 26 28 30 32 SEP.VICE LIFE (EFFECTIVE FULL POWER YEARS)

Fast neutron f luence (E > 1 KeV) as a function of full power service life (EFPY)

SALEM 2

FIGURE B 3/4.4-1 B 3/4 4-15

" *.f t*

t, 00

i:-
2.

t:-<

ttl

.3:

c:: z H

t-3 J

0

  • ~

0 ii ti *

u.

I t.d I

w

+-

+-

I,_.

°'

Au.Ice. n/cm1 CE > 1 M*VI Fluence Factor for use in the expression for ll RTNDT FIGURE B 3/4.4-~

41'.,. !-""

REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY 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 Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the.

Conunission pursuant to 10 CFR Part 50.55a(g)(6)(i).

SALEM UNIT 2 B 3/4.4-17

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