Text

Proposed Tech Specs 3/4.6.K,updating Pressure-Temp Curves to 22 Effective Full Power Yrs & TS Bases
	ML20129D398
Person / Time
Site:	Dresden, Quad Cities
Issue date:	09/20/1996
From:	; COMMONWEALTH EDISON CO.
To:	;
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ML17187A595	List: ML20129D398;
References
	NUDOCS 9609300069
	Download: ML20129D398 (49)
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ATTACHMENT B j AFFECTED TECHNICAL SPECIFICATION PAGES i

LICENSE DPR-19/25 LICENSE DPR-29/30 VIII VIII

.'./4.6-19 3/4.6-19 3/4.6-20 3/4.6-20 4 3/4.6-21 3/4.6-21 B 3/4.6-6 B 3/4.6-5 B 3/4.6-7 B 3/4.6-6 B 3/4.6-8 B 3/4.6-7 ,

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I 9609300069 960920

~PDR ADOCK 05000237 P pop

0 TABLE OF CONTENTS TOC LIMITING CONDITIONS FOR O"4 RATION AND SURVEILLPfCE REQUIREMENTS SECTION PAGE J.[4.6

. PRIMARY SYSTEM BOUNDARY 3/4.6.A Recirculation Loops . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-1 I 3/4.6.B J et Pu m p s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-3 )l i

3/4.6.C Recirculation Pumps .................................. 3/4.6-5 ,

3/4.6.D Idle Recirculation Loop Startup . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-6 I i

3/4.6.E Sa fety Valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-7 l 3/4.6.F Relief Valves ....................................... 3/4.6-8 !

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3/4.6.G Leakage Detection Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-10 l

3/4.6.H Operational Leakage .................................. 3/4.6-11 3/4.6.1 Chemistry ......................................... 3/4.6-13 Table 3.6.1-1, Reactor Coolant System Chemistry Limits j 3/4.6.J Specific Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-16 :

k 3/4.6 K Pressure / Temperature Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-19

f. ?.6.M , *.'5'm; m .Mw V====l '1.;.; Tois.v i ow.. ... %.'.'::D

- - - - . _ ..._ _...._ / j

.6.L Reactor Steam Dome Pressure ........................... 3/4.6-22 ;

3/4.6.M Main Steam Line isolation Valves ......................... 3/4.6-23 i

3/4.6.N Structural Integrity ............................ . ..... 3/4.6-24 3/4.6.0 Shutdown Cooling - HOT SHUTDOWN . . . . . . . . . . . . . . . . . . . . . . 3/4.6-25 !

3/4.6.P Shutdo n Cooling - COLD SHQTDOWN . . % . . .m. .

_3/4__ 4 Visure. 3.co.vA, %m~% pc.kre. u~hRoc- Press <c. ras st as tg a ppy i

( ryc 3.c.cz,% -,- y u u_.uA-6 % % y- va..a v2ceer ;

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<>re. p e Q _

-. Art..el b '17 G F A/

DRESDEN - UNITS 2 & 3 Vill Amendment Nos. so's165 i

)

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS k Pressure / Temperature Limits K. Pressure / Temperature Limits he reactor coolant system temperature 1. During system heatup, coold n and '

\' a d pressure shall be limited in accordance inservice leak and hydros ic testing wi the limit lines shown on Figure 3.6.K-1 operations, the reactor oolant system (1) c rve A for hydrostatic or leak testing; temperature and pr sure shall be (2) cur B for heatup by non-nuclear determined to be ithin the required means, c oldown following a nuclear )

heatup and co own limits and to the !

/ shutdown d low power PHYSICS TESTS right of the li it lines of Figure 3.6.K-1 j and (3) curve for operations with a curves A, o B, as applicable, at least ;

critical core oth than low power PHYSICS once per minutes. l j TESTS, with: _

2. The r ctor coolant system

1. A maximum reactor oolant heatup of tem rature and pressure shall be

~

100'F in any one hour eriod, de rmined to be to the right of the i

c ticality limit line of Figure 3.6.K-1 /

j 2. A maximum reactor coolant c Idown urve C within 15 r2nutes prior to the f I of 100*F in any one hour perio , withdrawal of control rods to bring the reactor to criticality and at least once

~

3. A maximum reactor coolant per 30 minutes during system heatup.

temperature change of $20'F in any one hour period during inservice 3. he reactor vessel material surveillance h hydrostatic and leak testing operation spe' ens shall be removed and above the heatup and cooldown lim' examin , to determine changes in curves, and reactor pr ure vessel material properties in ordance with 10CFR

4. The reactor vessel flange and ead Part 50, Appendi .

flange temperature 2100*F hen reactor vessel head bolti studs are 4. The reactor vessel flan e and head under tension, flange temperature shall e verified to be 2100*F:

i APPLICABILITY: a. In OPERATIONAL MODE when k the reactor coolant temper ture is:

\ At all times.

1) s130*F, at least once p 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2) s110*F, at least once per 30 minutes.

I b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel

'b head botting studs.

DRESDEN - UNITS 2 & 3 3/4.6-19 Amendment Nos. 50 & 14

b %erY b K. Pressure / Temperature Limits K. Pressure / Temperature Limits The primary system coolant system 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal and pressure testing operations, at least temperature and pressure shall be limited as once per 30 minutes, specified below;

a. The rate of change of the primary

1. Pressure Testing: system coolant temperature shall be determined to be within the heatup and

a. The reactor vessel metal temperature cooldown rate limits, and and pressure shall be maintained within the Acceptable Regions as shown on b. The reactor vessel metal temperature Figures 3.6.K-1 through 3.6.K-3 with and pressure shall be determined to be the rate of change of the primary within the Acceptable Regions on system coolant temperature s 20*F per Figures 3.6.K-1 through 3.6.K-4.

hour,or

2. For reactor critical operation, determine

b. The rate of change of the primary within 15 minutes prior to the withdrawal of system coolant temperature shall be control rods and at least once per 30 s100*F per hour when reactor vessel minutes during primary system heatup or metal temperature and pressure is cooldown, maintained within the Acceptable Regions as shown on Figure 3.6.K-4. a. The rate of change of the primary system coolant temperature to be within

2. Non-Nuclear Heatup and Cooldown and the limits, and low power PHYSICS TESTS:

b. The reactor vessel metal temperature

a. The reactor vessel metal temperature and pressure to be within the and pressure shall be maintained within Acceptable Region on Figure 3.6.K-5.

the Acceptable Regions as shown on Figure 3.6.K-4, and 3. The reactor vessel material surveillance specimens shall be removed and

b. The rate of change of the primary examined, to determine changes in reactor system coolant temperature shall be pressure vessel material properties in s100*F per hour, accordance with 10CFR Part 50, Appendix H.

9

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1

3. Nuclear Heatup and Cooldown: 4. The reactor vessel flange and head flange temperature shall be verified to be

a. The reactor vessel metal 283*F:

temperature and pressure shall be maintained within the Acceptable a. In OPERATIONAL MODE 4 when Region as shown on Figure 3.6.K-5, the reactor coolant temperature is:

and

1) s113*F, at least once per

b. The rate of change of the primary 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

system coolant temperature shall be s100*F per hour. 2) 593*F, at least once per 30 minutes.

4. The reactor vessel flange and head flange temperature 283 "F when reactor b. Within 30 minutes prior to and at vessel head bolting studs are under least once per 30 minutes during tension. tensioning of the reactor vessel head bolting studs.

APPLICABILITY: I At all times.

l M

1 l PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l

l 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS ACTION:

With any of the above limits exceeded, P

p,;rM _

i

2. Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of -

the reactor coolant system and -

determine that the reactor coolant system remains acceptable for continued operation 'or r Lg36.12.

-- h**6)

)

3. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i i

1. Restore the reactor vessel metal temperature and/or pressure to within the limits within 30 minutes without

\ exceeding the applicable primary system coolant temperature rate of change limit, and a

DRESDEN - UNITS 2 & 3 3/4.6-20 Amendment Nos. 150 s 145

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l FIGURE 3.6.K-1 {

MINIMU EACTOR VESSEL METAL TEMPERAT RE VS. REACTOR VESSEL PRES RE 1600\ / -

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1400 55

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0 50 100 150 200 250 300 350

! Figures 3.b.6-1 s.oup s.L. < -5 *MU" "*R"k *M *"R[ VO DRESDEN - UNITS 2 & 3 3/4.6-21 Amendment Nos.

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses w '

ced vet from compressive at the inner wall to tensile at the outer wall. These thermal '

stresses tend to alleviate the tensile stresses induced by the int pressure.

compres Therefore, a pr re temperature curve based on steady state condit' , i.e., no thermal stresses, represents a r bound of all similar curves for fini atup rates when the inner wall of the vesselis treated as the rning location.

The heatup analysis also covers the dete of pressure-temperature limitations for the case omes the co ling location. The thermal gradients in which the outer wall of the vess established during heatup pro tensile stresses which are ady present. The thermal induced the vessel are tensile and are depende both the rate of heatup stresses at the outer at described for and the time alo - e heatup ramp; therefore, a lower bound curve similar e outer the heatu the inner wall cannot be defined. Subsequently, for the cases in w i wa the vessel becomes the stress controlling location, each heatup rate of interest mu nalyzed on an individual basis.

r, bis a q essure-temperature i h M 1 K-3 imit lines *shownuOhge 3.Sdfor operating conditions;gico dvhtet@ Testing hurve XO Non-Nuclear Heatup/Cooldown "urvc SJ and Core Critical

(' Operation ^ curve CR The curves have been established to b in conformance with Appendix G to 10 CFR Part 50 and Regulatory Guide 1.99 Revision 2, and ake into account the change in reference nil-ductility transition temperature (RTuoy) as a res it of neutron embrittlement. The I

adjusted reference temperature (ART) of the limiting vessel aterial is used to account for e fects. g y c,n'rradia

,. g. ie s -

,e

- Muss)Tessel regions are considered for the development of the pressure-temperaturegurvesF 1)

the core beltline region; 2) the non-beltline region (other than the closure flange regiom;@3) the p

clo e flange region. The beltline region is defined as that region of the reactor vessel that irectly surr s the effective height of the reactor core and is subject to an RTuor adjustment to closure flangFe~gion&receiveg account for radiation embrittlement. The insufficient fluence to necessitate an RTuor adjustmer non-beltlin@it. These regions'Wntain compone include; the reactor vessel nozzles, closure flanges, top and bcttom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core. Although the closure flange region %@ on-beltline region {Mreated separately for the development of the pressure- "

tempe ture ctirve o address 10CFR Par @ Accendix G requirements.

buy w - ed bortoq h ea Q CD _ _

In evaluating the adequacy of the steel which comprises the reactor vessel,jLijLnecesseg a

followingEE estabHM 1) the RTuor for all vessel a31ad;ommgmaTen fs~; 2) the relationship on-trux usr.ce, at aneroies greater than one Mev); and 3) the !

between RTuor and inte rat }

fluen.cmat-t ion of a postulated flaw.

.m d 9 ) the, bottem he rehn Amendment Nos. 150 s n,.

DRESDEN - UNITS 2 & 3 8 3/4.6-6

j PRIMARY SYSTEM BOUNDARY B 3/4.6 L -

BASES !

J I _-

Boltuo Temoerature ; 3 ,g ,

, m i Thhinitial RT y of the main closure flanges, the all and hea ' ' interi a connecting to these j flanges,@onnecting weldsf = -.7,,3the vertical electro g welds which

terminate immediately below the vessei flang: C.z: = h M Gr7 it . fore, the minimum l allowable boltup temperature is established --%mRT.y + 50"F) which includes a 60*F j conservatism required by the original ASME C e _ construction.

~ '

[ -..I,-lld g M i

j s indicated in curve A of Figure 3.6.K-1 for system hydrotesting, the minimum metal j- te rature of the reactor vessel shell is 100T for reactor pressures less than 312 ps . This

100*F ' mum boltup temperature is based on a RT., of 40*F for the electrosla eld immediately low the vessel flange and a 60T conservatism required by the ginal ASME

)

l Code of constru n. At reactor pressures greater than 312 psig, the m

um vessel metal temperature is estab as 130T. The 130V minimum tempo s based on a closure flange region RTuor of 40 nd a 90T conservatism required CFR Part 50 Appendix G '

j for pressure in excess of 20 the preservice hydrostatic pressure (1563 psig). At i I,r,cd g approximately 650 psig the effect pressurization er ore limiting than the boltup stresses of non-j at the closure flange region, hence a fa ar curves intersect the 130T vertical i line. Beltline as well as non-beltline curves een provided to allow separate monitoring of j the two regions. Beltline curves as a fun ssel exposure for 12,14 and 16 effective i full power years (EFPY) are presented allow the the appropriate curve up to 16 EFPY l j of operation.

A typical sequence invo in pressure testing is a heatup to the uired temperature and then pressurization a required pressure for the inspection. Durin a heatup, at 100*F/ hour or s, Curve B is the goveming curve. Since the vessel is ressurized during the heatup urves A and B are the same. When temperatures are stabilized t ithin

ins, at i 20

F/ r rates, at temperatures above those required by curve A, pressurization l w point Curve A is the goveming curve. During the inspection period with the ves at l he required pressure, temperature changes are limited to 20*F/ hour.

b v. ONNuclear Heatuo/Cooldown --

'(Ch.-e QFigure 3.6.K p'plies during heatups with non-nuclear heat (e.g., recirculation i pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).

! The curve provides the minimum reactor vessel metal temperatures based on the most limitina

( vessel stressM As indicated by the vertical 1001 line, the boltup stresses at th italigi sig. For reactor

( region ... m-+ limiting for reactor pressures below approxi

....anzation and thermal stresses become j l pressures greater than appma....:^ 'u 110 n-5.

more limiting than the es, which is rwi,.c ' hv the nonlinear portion of curve B.

} Jhe_ce-l'..... ortion of the curve is dependent on non-beltline and Demi... :fana. with the

!! beltline region temperature limits having been adjusted to account for vessel irradiation (up to a j - - -

- y i,e. o m da m he=Q /es,IJm ta4e. e4 too T/hwe is a rrt h lele .

Ameridment Nos. 150 s l DRESDEN - UNITS 2 & 3 8 3/4.6-7 i

i

INSERT B ;

I l

Flaures 3.6.K-1 throuah 3.6.K-3 Pressure Testina j As indicated in Figure 3.3.6.K-1 through 3.6.K-3 for pressure testing, the minimum metal temperature of the reactor vessel shell is 83 F for reactor pressures less than 312 psig.

This 83*F minimum boltup temperature is based on a RTuor of 23*F for the electroslag weld immediately below the vessel flange and a 60 F conservatism required by the original ASME Code of construction. The bottom head region limit is established as 68 F, based on moderator temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113*F. The 113*F minimum temperature is based on a closure flange region RTuor of 23 F and a 90 F conservatism required by 10CFR Part 50 Appendix G.

Beltline curves as a function of vessel exposure for 18,20 and 22 effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 22 EFPY of operation.

Figures 3.6.K-1 tnrough 3.6.K-3 are governing for applicable pressure testing with a maximum heatup/cooldown rate of 20*F/ hour.

I l

= PRIMARY SYSTEM BOUNDARY B 3/4.6 F,jum 3.L.K-5 BASES 3

//

T r

' vessel exposure of 16 EFPY). The non-beltline region is limiting between approximately 110 psig and 830 psig. Above approximately 803 psig, the beltline region becomes limiting.

[oc -

M dre: C4 Core Critical Operation e critical operation curve shown in Figure 3.6.K , generated in accordance

( with 10CFR Part 50_ Appendix G which requires core critical pressure-temperature limits to be t' l l 40*F above any(dE?c: ^ sr g? limits. Sinegree Tis more limiting,6'ir.Jis@urseEplus 3l ) .

h"N N_

F wte 3 6 K-4 es blished periodically during operation by The actual shift in RTur of the vessel material will and 10CFR Part 50, Appendix H, removing and evaluating,in accordance with ASTM E185 4 y irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in J 4 the core area. The irradiated specimens 5e" be used ";P cerM:n:Cin predicting reactor vessel it curves of Mgure 3.6.K-toshall be

( g , material [for.eWsr. to...nera;ura eh3 The operatingand li r. ecommend ions of Regulatory J

g adjustec, s required, on the basis of the specimen dat 3 ; Guide 1.99, evision 2. are used .s hp 3.Lk .5

*L , ,

i )j.i3/4.6.L Reactor Steam Dome Pressure e

s kgThe reactor steam dome pressure is an assumed initial condition of Design Basis Accidents a It ransients and is also an assumed value in the determination of compliance with reactor pressure l

4

)k i~h' vessel overpressure protection criteria. The reactor steam dome pressure of $1005 psig is an R initial condition of the vessel overpressure protection analysis. This analysis assumes an initial l

hmaximum reactor steam dome pressure and evaluates the response of the pre Qprimarily the safety valves, during the limiting pressurization transient. The determination o compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

1 1

3/4.6.M Main Steam Line Isolation Valves d

' Double isolation valves are provided on e.ach of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is require to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

4 Amendment Nos. 150 r,14 ORESDEN - UNITS 2 & 3 8 3/4.6-8

9 TABLE OF CONTENTS TOC-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ,

B SECTION EACif l 31.id ~ PRIMARY SYSTEM BOUNDARY !

3/4.6.A Recirculation Loops . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-1 !

3/4.6.B Jet Pumps .......................................... 3/4.6-3 3/4.6.C- Recirculation Pumps ................................... 3/4.6-5 I 3/4.6.D Idle Recirculation Loop Startup . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-0 3/4.6.E S a f ety Valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4.6-7 !

3/4.6.F R e lie f Valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6 8 3/4.6.G Leakage Detection Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-10 l 3/4.6.H Operational Leakage ................................... 3/4.6-11 ;

3/4.6.1 C he mi stry . . . . . . . . . . . . . . . . . . . . . . . . . . . . .............. 3/4.6-13 j Table 3.6.1-1, Reactor Coolant System Chernistry Limits 3/4.6.J Specific Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-16 ;

3/4.6.K g essure / Temperature Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-19 ;

' Figure 3.6.K-1, Minimum Reactor Vessel Metal Temperature vs. Rx. Vessel Pressurer l

_- - - _ - . . . _ . _ __j ;

3/4.6.L Reactor Steam Dome Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-22 +

3/4.6.M Main Steam Line Isolation Valves . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-23 l 3/4.6.N Structural Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-24 )

3/4.6.0 Residual Heat Removal - HOT SHUTDOWN . . . . . . . . . . . . . . . . . . . 3/4.6-25

-3/4.6.P Residug Heat-Removal - COLD SHUTDOWN . . . . . . . . . . . . . . . . . . 3/4.6-27 x _s --

Fyre %.to.Y-\ , Pru.socc. -%een%C-. l 'a %-%Mce T'o~g -

Vatid b IB E PPM

[ Ry e- hx -?_, P eu.<e-Te7.tu,4_u hs Coc Restw< c 'hbq- % t.cl 6 2.o E FPY Fi3 ur o .6.w - 4 R e e M9 ,,6,e L a 6c Reuvre Teh, _M I.A 4 22 E F9y Agw.a %.vA4, Precs.w- Te g,.k. <.L .t c C.v- 14:w-Muche.e- het.6p /Cootdee -

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.o re 3, to, sc-s, gra:u< e -Tey- r. 4 . n L.a s.

QUAD CITI NI G 1 & z x -Vith - ~~~-~AEdinent Nos. in s'in 1

- _ . l

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 1

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS i .i , ,,,

K. Pre sure/ Temperature Limits K. Pressure / Temperature L' its The eactor coolant system temperature 1. During system h atup, cooldown and and p ssure shall be limited in accordance inservice leak d hydrostatic testing with th limit lines shown on Figure 3.6.K-1 operations, th reactor coolant system (1) curve for hydrostatic or leak testing; temperature nd pressure shall be

, (2) curve 8 r heatup by non-nuclear determined o be within the required means, coold n following a nuclear heatup an cooldown limits and to the i shutdown and I power PHYSICS TESTS; right of t e limit lines of Figure 3.6.K-1 l and (3) curve C f operations with a critical core other th low power PHYSICS curves , or B, as arealicable, at least once r 30 minutes.

)

I TESTS, with: Y

2. Th reactor coolant system *

1. A maximum reactor co nt heatup of t mperature and pressure shall be 100 F in any one hour pe 'od, determined to be to the right of the criticality limit line of Figure 3.6.K-1 1

2. A maximum reactor coolant co own curve C within 15 minutes prior to the I of 100 F in any one hour period, withdrawal of control rods to bring the I reactor to criticality and at least once

3. A maximum reactor coolant per 30 minutes during system heatup, temperature change of s20*F in any I one hour period during inservice 3. he reactor vessel material surveillance hydrostatic and leak testing ope tions sp imens shall be removed and above the heatup and cooldo limit examt , to determine changes in curves, and reactor pr sure vessel material I properties in cordance with 10CFR )

4. The reactor vessel flang and head Part 50, Appen H. I flange temperature 21 F when reactor vessel head Iting studs are 4. The reactor vessel fla e and head under tension. flange temperature shal e verified to be 2100 F:

APPLICABILITY: a. In OPERATIONAL MODE hen j

~

the reactor coolant temperat te is:

At all times. t

1) 5130 F, at least once per '

j 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2) s110 F, at least once per

- 30 minutes.

\ b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting stud::. !

QUAD CITIES - UNITS 1 & 2 3/4.6-19 Amendment Nos. 171 & 167

i i

16%e.r-Y" k f

r K. Pressure / Temperature Limits K. Pressure / Temperature Limits The primary system coolant system 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal - and pressure testing operations, at least temperatur6 and pressure shall be limited as once per 30 minutes, j

specifica below

a. The rate of change of the primary j

1. Pressure Testing: system coolant temperature shall be l determined to be within the heatup and 4 a. The reactor vessel metal temperature cooldown rate' limits, and

) and pressure shall be maintained within i the Acceptable Regions as shown on b. The reactor vessel metal temperature Figures 3.6.K-1 through 3.6.K-3 with and pressure shall be determined to be the rate of change of the primary within the Ace $We Regions on Figures 3.6 K-1 through 3.6.K-4.

j system coolant temperature s 20'F per hour,or

2. For reactor critical operation, determine

b. The rate of change of the primary within 15 minutes prior to the withdrawal of i system coolant temperature shall be control rods and at least once per 30 5100*F per hour when reactor vessel minutes during primary system heatup or l cooldown, metal temperature and pressure is maintained within the Acceptable Regions as shown on Figure 3.6.K-4. a. The rate of change of the primary system coolant temperature to be within

2. Non-Nuclear Heatup and Cooldown and the limits,'and 4

t low power PHYSICS TESTS:

b. The reactor vessel metal temperature l

a. The reactor vessel metal temperature and pressure to be within the r

and pressure shall be maintained within Acceptable Region on Figure 3.6.K-5.

i the Acceptable Regions as shown on Figure 3.6.K-4, and 3. The reactor vessel material surveillance specimens shall be removed and i

b. The rate of change of the primary examined, to detennine changes in reactor system coolant temperature shall be pressure vessel material properties in

' accordance with 10CFR Part 50, Appendix s100*F per hour, H.

I

,e -, ~

~,n ---- .- , -

l A},SQl.'T A ( **W '

f

3. Nuclear Heatup and Cooldown: 4. The reactor vessel flange and head flange temperature shall be verified to be

a. The reactor vessel metal 283'F:

temperature and pressure shall be maintained within the Acceptable a. In OPERATIONAL MODE 4 when .

Region as shown on Figure 3.6.K-5, the reactor coolant temperature is: '

and l 1) s113*F, at least once per

b. The rate of change of the primary 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

system coolant temperature shall be s100'F per hour. 2) sg3*F, at least once per 30 minutes.

4. The reactor vessel flange and head flange temperature 283 'F when reactor b. Within 30 minutes prior to and at -

vessel head bolting studs are under least once per 30 minutes during tension. ~ tensioning of the reactor vessel head bolting studs.

APPLICABILITY:

At all times.

1

)

I 1

2 PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS

^

I ACTION:

With any of the above limits exceeded, +

l h j

j

}

estore the temperature a press smits within

}

3 es,and - - -

2. Perform an engineering evaluation to i determine the effects of the out-of-limit I

condition on the structural integrity of .

I the reactor coolant system and '

j' determine that the reactor coolant _

j system remains acceptable for g,s, continued operation or wi+Ala 73 i

3. Be in at least HOT SHUTDOWN within J 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN i

- within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

\

t 4

l

1. Restore the reactor vessel metal temperature and/or pressure to within the )

limits within 30 minutes without exceeding the applicable primary system coolant temperature rate of change limit, and t

OUAD CITIES - UNITS 1 & 2 3/4.6-20 Amendment Nos. 171 s 167

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

- - . ~. - - - - - hef

- - -QCG

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PRIMARY SYSTEM BOUNDARY 700 p s 3 . G . x -- 1 4 % 3, g , g _5 PT Umits 3/4.6.K j FIGURE 3.6.K-1 i M IMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL URE PRE i 1 00. / -

A _sysTru

- Hv.oncntst

_ uunr CURVE A

?' - NON-NUREMt HEATUP/

QRDOWN UWT.

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1400{ m m 1s m gmC la 14 0 f/

- i c-am I fI A i ! \ / N(/

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,l[ ' V AELTUNC:

MA 12 ar*r

(

8 l- 1000 i N '

//V / / :: = 4 s l \ u/ / /

9 E 800 :

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0 ' ininui,inuun u nu ni. unnui ununs, niniusi uninu O 50 100 150 200 250 300 350 .

WNIMUM REACTOR VC55EL METAL TEMPERATURE ('F)

~

QUAD CITIES - UNITS 1 & 2 3/4.6-21 Amendment Nos.

' I' ' "

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 3/4.6.J Specific Activity The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorobgical conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT l-131, but less than or equal to 4.0 microcuries per gram DOSE EQUlVALENT l-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

3/4.6.K Pressure / Temperature Limits All components in the reactor coolant system are designed to withstand the effects of cyc'ic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1.1 of the UFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the l stress limits for cyclic operation.

tup, the thermal gradients in the reactor vessel wall produce thermal stres ich Dunn vary from co ssive at the inner wall to tensile at the outer wall. These th induced compressive stresse to alleviate the tensile stresses induced b nternal pressure.

Therefore, a pressure temper curve based on steady stat itions, i.e., no thermal stresses, represents a lower bound similar curve mite heatup rates when the inner wall of the vesselis treated as the governing loc The heatup analysis also cover determination of pressu perature limitations for the case in which the outer wal e vessel becomes the controlling locati he thermal gradients established du eatup produce tensile stresses which are already presen . e thermal induced e outer wall of the vessel are tensile and are dependent on both the ra heatup stress the time along the heatup ramp; therefore, a lower bound curve similar to that describe Amendment Nos. 171 & 167 QUAD CITIES - UNITS 1 & 2 B 3/4.6-5

i J PRIMARY SYSTEM BOUNDARY B 3/4.6

! BASES

) .

l the hea u ot be defined. Subseque r

! wall of the vessel becomes the rate of interest must be anahczedam-ernr3HWgig asin.

@ m,,e 3.to l

t x -\ N 3.6.k d M ine pressure-temperature , limit lines {shownbu.e 0.0.K[1, for operating conditions;

crtet3 Testing 6"r>: /*, Non-Nuclear Heatup/CooldownT;u ;; 5*, and Core Critical fressors Operatiogcer>: C]. The curves have been established to b in conformance with Appendix G to (10 CFR Pa'rt ou and Regulatory Guide 1.99 Revision 2, an take into account the change in uit of neutron embrittlement. The I[ reference adjusted reference nil-ductility transition temperature temperature (ART) (RTuor) of the limiting vess Ias ar materialis used to account for -

(irradiation effects. _

3,g ad +k loo %

vessel regions are considered for the development of the pressure-temperature urves: 1)

M the core beltline region; 2) the non-beltline region (other than the closure flange region ;@3) the closure flange regio 3 The beltline region is defined as that region of the reactor vessel that

j. irectly surrounds the effective height of the reactor co and is subject to an RTm adjustment to 1 account for radiation embrittlement. The non-beltli losure flan egior& receive g insufficient fluence to necessitate an RTuor adjustm hese regions ntain components which f .

include; the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive j 14 penetrations, and shell plates that do not directly surround the reactor core. Although the closure

) flangegegionQpon-beltline regio treated s ar ely for the development of the oressure-f _

tempe(ture ggt ,ress 10CF a O ig equirements. ja bot 6-head k .3 ( following be established:In 1) the M i nor evaluaunw d.; r%nn@e

" . :!! ==al and : 'N...ov materials; 2)steel which compris the relationship

.i

[ f, between RTuor and inteoratad -1..... Gux (fiuence, at energi . v.;r +=n one Mev); and 3) the 2 fluan'= :: L iUEation~of a postulated flaw. - -

7l Boltuo Temperatur g 3 2.3*F im;hg The initial RTuor of the main closure flanges, the shell and hea materials connecting to these flanges,@ connecting weldsE'O*5 5:rz:r:Othe ver*a lectronlag welds which terminate immediately below the vessel flange @cre er "Tm e; 40"j. Therefore, the minimum allowable boltup temperature is established asGesaFJ(RTuor + 609) which includes a 60*F conservatism required by the original ASME Code of construction.

$* f" Ttnve A - Hydrotestina As indicated in curve Figure 3.6.K-1 for system hydrotesting, the minimum metal temperature of the reactor ves eli is 100*F for reactor pressures,M312 psig. This 100'F minimum boltup temperature is n a RTuo or the electroslag weld immediately below the vessel flange and a 6

atism required by the original ASME Code of construction. At react res greater than 3 the minimum vessel metal 6t temperature is esta as 130'F. The 130*F minimum temperatur sed on a closure flange r uor of 40*F and a 90'F conservatism required by 10CFR Part 5 dix G D

171 s 167 QUAD CITIES - UNITS 1 & 2 B 3/4.6-6 Amendment Nos.

INSERT B Fiaures 3.6.K-1 throuah 3 6 oressure Testina As indicated in Fig K-1 through 3.6.K-3 for pressure testing, the minimum metal temperature of the re. a vessel shell is 83 F for reactor pressures less than 312 psig.

This 83 F minimum boltup temperature is based on a RTuor of 23*F for the electrostag weld immediately below the vessel flange and a 60 F conservatism required by the original ASME Code of construction. The bottom head region limit is established as 68"F, based on moderator temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113 F. The 113'F minimum temperature is based on a closure flange ,

region RTuor of 23 F and a 90 F conservatism required by 10CFR Part 50 Appendix G. )

Beltline curves as a function of vessel exposure for 18,20 and 22 effective full power l years (EFPY) are presented to allow the use of the appropriate curve up to 22 EFPY of operation.

Figures 3.6.K-1 through 3.6.K-3 are governing for applicable pressure testing with a maximum heatup/cooldown rate of 20 F/ hour.

1 l

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES pressure in excess of 20% of the p eservice hydrostatic test pressure (1563 psig). A appr ately 650 psig the effects of p essurization are more limiting than the bolt tresses at the closu ange region, hence a fa nily of non-linear curves interssct th F vertical

. line. Beltline as w s non-beltline corves have been provided to allo parate monitoring of the two regions. Beltline es as a function of vessel exposu r 12,14 and 16 effective full power years (EFPY) are pre d to allow the use of appropriate curve up to 16 EFPY of operation.

A typical sequence involved in pressu sting is a tup to the required temperature and then pressurization to the requi pressure for the inspec. During the heatup, at 100'F/ hour or less, Cu is the governing curve. Since the v i is not pressurized during the heatup, Curv and B are the same. When temperatures are sta to within 20*F/ hour s, at temperatures above those required by curve A, pressuriza egins, at whic int Curve A is the governing curve. During the dspection period with the v at

, e required pressure, temperature changes are limited to 20'F/ hour.

n-Nuclear Heatuo/Cooldown Figure 3.6.K ies during heatups with non-nuclear heat (e.g., recirculation j .? pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram).

LN L l The curve provides the minimum reactor vessel metal temperatures based on the most limiting _

__ vessel stressJ As indicated by the vertical 100*F line, the boltup stresses at the closure flan i

0' '

F region are most limiting for reactor pressures below approximately 110 psig. For reactor i

, pressures greater than approximately 110 psig, pressurization and thermal stresses become l

' pp ,g i more limiting than the boltup stresses, which is reflected by the nonlinear portion of curve B.

j The non-linear portion of the curve is dependent on non-beltline and beltline regions, with the

!}gg / beltline region temperature limits having been adjusted to account for vessel irradiation (up to a l g1 vessel exposure of 16 EFPY). The non-beltline region is limiting between approximately 110

! j psig and 830 psig. Above approximately 803 psia, the beltline region becomes limitino.

Cr;: op r N i re critical operation curve shown in Figure 3.6.K- is generated in accordance i

with 10CFR Part 50 Appendix G which requires core critical pressure-temperature limits to be

! 40'F above any ure: .^. :: $ limits. Since is more lirn@ing,Crure:jis@pluss 40'F. 3.(o x- 4 , 34x g 1

j The actual shift in RTm of the vessel matorial will be est lished periodically during operation by j removing and evaluating, in accordance with ASTM E185 and 10CFR Part 50, Appendix H, i irradiated reactor vessel material specimens installed _near the inside wall of the reactor vessel in

the core, area. The irradiated specimensC;. 5 need - M P@in predicting reactor vessel i

materialtr~i+iaa

r:=- 2QThe operatindimit curves of Figur 3.6.K-1 hall be I

! adjusted, as required, onlhe basis of the specimen data and recommen ations of egulatory l

Guide 1.99, Revision 2. / 1 8 q re, os,ed ' reg h ;

3.t..K-E

, QUAD CITIES - UNITS 1 & 2 B 3/4.6-7 Amendment Nos. 171 & 167

" ' ~

7)g mnow heatflvnd.as rede.

of l e o

y-/ he d e is 8fh*O *

- _ - - ,, .- _ J

I ATTACHMENT B (Cont'd)

PROPOSED TECHNICAL SPECIFICATION PAGES h

h

=

4 ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT APPLICABILITY REVIEW The Commission has provided standards for determining whether a no significant hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an operating license involves a 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.

Comed proposes to amend Appendix A, Technical Specifications, Section 3/4.6.K of Facility Operating Licenses DPR-19, DPR-25, DPR-29 and DPR-30. The amendment request changes the pressure temperature (P-T) curves, Figure 3.6.K-1 and associated Bases.

Comed has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of Dresden Units 2 and 3 or Quad Cities Units 1 and 2 in accordance with the proposed amendment will not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because of the following:

The proposed changes merely adjust the reference temperature for the limiting beltline material to account for irradiation effects and provide the same level of protection as previously evaluated. The adjusted reference temperature calculations were performed utilizing the guidance contained in Regulatory Guide 1.99, Revision 2. The change is administrative in nature to reflect the extension of the operating limits to 22 EFPY. As such, these changes will not significantly increase the probability or consequences of a previously evaluated accident.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:

The proposed changes do not create the possibility of a new or different kind of accident previously evaluated for Dresden or Quad Cities Stations. No new modes of operation are introduced by the proposed changes. The revised operating limits are merely an updated of the old limits by taking into account the effects ofirradiation on the limiting reactor vessel material. Use of the revised P T curves will continue to provide the same level of protection as was previously reviewed and approved. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

The associated change to the P-T curves related to this proposed amendment does not affect any activities or equipment and are not assumed in any safety analysis to initiate any accident sequence for Dresden or Quad Cities Stations; therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

ATTACIIMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT APPLICABILITY REVIEW

3) Involve a significant reduction in the margin of safety because:

The proposed amendment reflect an update of the P-T curves to extend the operating limit to 22 EFPY. The revised curves are based on the latest NRC guidance along with actual data for the units. The new limits retain the margin of safety to the level expected for a new vessel, L adjusted for irradiation effects as required by 10CFR, Appendix G, thereby maintaining a conservative margin of safety.

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

Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change ,

requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant i

hazards considerations.

This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.

ENVIRONMENTAL ASSESSMENT Comed has evaluated the proposed amendment against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has i been determined that the proposed changes meet the criteria for a categorical exclusion as provided i under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes requested do not pose significant hazards consideration or do not involve a significant increase in the amounts, l and no significant changes in the types, of any effluents that may be released off-site. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure.

l ATTACHMENT B AFFECTED TECHNICAL SPECIFICATION PAGES LICENSE DPR-19/25 LICENSE DPR-29/30 Remove inKIl Remove ink 11 I VIII VIII VIII VIII l

3/4.6-19 3/4.6-19 3/4.6-19 3/4.6-19 3/4.6-20 3/4.6-20 3/4.6-20 3/4.6-20 3/4.6-21 3/4.6-21a 3/4.6-21 3/4.6-21a 1

3/4.6-21b --

3/4.6-21b 3/4.6-21c --

3/4.6-21c 3/4.6-21d --

3/4.6-21d l l

3/4.6-21e --

3/4.6-21e B 3/4.6-6 B 3/4.6-6 B 3/4.6-5 B 3/4.6-5 B 3/4.6-7 B 3/4.6-7 B 3/4.6-6 B 3/4.6-6 B 3/4.6-8 B 3/4.6-8 B 3/4.6-7 B 3/4.6-7 i B 3/4.6-9 --

TABLE OF CONTENTS TOC

. i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ;

3/4.6 PRIMARY SYSTEM BOUNDARY 3/4.6.A Recirculation Loops . . . . . . . . . . . . . . . . . .......... 3/4/6-1 '

3/4.6.B J et P u mp s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-3 3/4.6.C Recirculation Pumps. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-5 l l

3/4.6.D idle Recirculation Loop Startup. . . . . . . . . . . . . . . . . . . . . 3/4.6-6 3/4.0 E Safety Valves . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 3/4.6-7 3/4.6.F Relief Valves. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-8 I

3/4.6.G - Leakage Detection Systems. . . . . . . . . . . . . . . . . . . . . . . 3/4.6-10 i 3/4.6.H Operational Leakage. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-11 l

3/4.6.1 C hemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-13 I Table 3.6.1-1, Reactor Coolant System Chemistry Umits 3/4.6.J Specific Activity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-16 3/4.6.K Pressure / Temperature Limits . . . . . . . . . . . . . . . . . . . . . . 3/4.6-19 Figure 3.6.K-1, Pressure-Temperature Umits for Pressure Testing - Valid to 18 EFPY Figure 3.6.K-2, Pressure-Temperature Limits for Pressure Testing - Valid to 20 EFPY Figure 3.6.K-3, Pressure-Temperature Limits for Pressure Testing - Valid to 22 EFPY Figure 3.6.K-4, Pressure-Temperature Limits for Non-Nuclear Heatup/Cooldown - Valid to 22 EFPY Figure 3.6.K-5, Pressure-Temperature Limits for Critical Core I Operations - Valid to 22 EFPY 3/4.6.L Reactor Steam Dome Pressure . . . . . . . . . . . . . . . . . . . 3/4.6-22 3/4.6.M Main Steam Line isolation Valves. . . . . . . . . . . . . . . . . . 3/4.6-23 3/4.6.N Structural Integrity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-24 3/4.6.0 Shutdown Cooling - HOT SHUTDOWN . . . . . . . . ... 3/4.6-25 3/4.6.P Shutdown Cooling - COLD SHUTDOWN . . . . . . . . . . . . 3/4.6-27 DRESDEN - UNITS 2 & 3 Vill Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K. Pressure / Temperature Limits K. Pressure / Temperature Limits The primary system coolant system 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal and pressure testing operations, at least temperature and pressure shall be limited as once per 30 minutes, specified below;

a. The rate of change of the primary

1. Pressure Testing: system coolant temperature shall be determined to be within the heatup

a. The reactor vessel metal and cooldown rate limits, and temperature and pressure shall be maintained within the Acceptable b. The reactor vessel metal temperature Regions as shown on Figures and pressure shall be determined to 3.6.K-1 through 3.6.K-3 with the be within the Acceptable Regions on rate of change of the primary system Figures 3.6.K-1 through 3.6.K-4.

coolant temperature s 20*F pt r hour, or 2. For reactor critical operation, determine -

within 15 minutes prior to the withdrawal

b. The rate of change of the primary of control rods and at least once per 30 system coolant temperature shall be minutes during primary system heatup or

$100*F per hour when reactor vessel cooldown, metal temperature and pressure is maintained within the Acceptable a. The rate of change of the primary Regions as shown on Figure system coolant temperature to be 3.6. K-4. within the limits, and

2. Non-Nuclear Heatup and Cooldown and b. The reactor vessel metal temperature low power PHYSICS TESTS: and pressure to be within the Acceptable Region on Figure 3.6.K-5.

a. The reactor vessel metal temperature and pressure shall be maintained within the Acceptable 3. The rea:.Pr vessel material surveillance Regions as shown on Figure speciment hall be removed and 3.6.K-4, and examined, to determine changes in reactor pressure vessel material

b. The rate of change of the primary properties in accordance with 10CFR system coolant temperature shall be Part 50, Appendix H.

s100*F per hour.

DRESDEN - UNITS 2 & 3 3/4.6-19 Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS

3. Nuclear Heatup and Cooldown: 4. The reactor vessel flange and head flange temperature shall be verified to be

a. The reactor vessel metal 283*F:

temperature and pressure shall be maintained within the Acceptable a. In OPERATIONAL MODE 4 when Region as shown on Figure 3.6.K-5, the reactor coolant temperature is; and

1) s113*F, at least once per

b. The rate of change of the primary 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

system coolant temperature shall be s100*F per hour. 2) s93*F, at least once per 30 minutes.

4. The reactor vessel flange and head flange temperature 283 'F when reactor b. Within 30 minutes prior to and at vessel head botting studs are under least once per 30 minutes during tension. tensioning of the reactor vessel head botting studs.

APPLICABfLITY:

At all times.

ACTION:

With any of the above limits exceeded,

1. Restore the reactor vessel metal temperature and/or pressure to within the limits within 30 minutes without exceeding the applicable primary system coolant temperature rate of change limit, and

2. Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structuralintegrity of the reactor coolant system and determine that the reactor coolant system remains acceptable for continued operations within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or

3. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

DRESDEN - UNITS 2 & 3 3/4.6-20 Amendment Nos.

l

\

I

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-1 PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 18 EFP/

1400

I

(

1200 ,

I

_ l UNACCEPTABLE REGION l s i

S i o 1000 E

J E 800

Temperature Change Rate E / s;20*F/hr

> f h l ACCEPTABLE REGION' l U ;

5 BOTTOM a: 600 - -

HEAD 68'F

>- f 5 /

s , i E /

!4 i I

E e UPPER VESSEL AND :

8L l 312 PSIG l BELTLINE LIMITS l -

- BOTTOM HEAD e LIMITS ;

200 ! ,,,,,,

a REGION 83*F I

a 0

0 50 100 150 200 250 300 350 i

M!NIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

DRESDEN - UNITS 2 & 3 3/4.6-21 a Amendment Nos.

i PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 1

j i

FIGURE 3.6.K-2 i PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VAllD TO 20 EFPY 1400 .

I ;

e !

f 1200 g l UNACCEPTABLE REGION l I s .

1000

/ i I

t-

.J i

$ f T g M m C hnge M

800 s;20*F/hr f

5 f E

O l ACCEPTABLE REGION

l 0 <

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DRESDEN - UNITS 2 & 3 3/4.6-21 b Amendment Nos.

i l

1 PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 1 1 FIGURE 3.6.K-3 I 1

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DRESDEN - UNITS 2 & 3 3/4.6-21c Amendment Nos

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-4 PRESSURE - TEMPERATURE LIMITS FOR NON-NUCLEAR HEATUP/COOLDOWN - VALID TO 22 EFPY 1400

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0 O 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

DRESDEN - UNITS 2 & 3 3/4.6-21 d Amendment Nos.

! PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K l

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FIGURE 3.6.K-5 '

i PRESSURE - TEMPERATURE LIMITS FOR CRITICAL CORE :

OPERATIONS - VALID TO 22 EFPY !

l 1400 1 i

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DRESDEN - UNITS 2 & 3 3/4.6-21e Amendment Nos.

i I

, PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES shutdown, the rates of temperature and pressure changes are limited so that the maximum specified j heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for i cyclic operation. '

! t i The pressure-temperature limit lines are shown, for operating conditions; Pressure Testing, Figures

! 3.6.K-1 through 3.6.K-3, Non-Nuclear Heatup/Cooldown, Figure 3.6.K-4, and Core Critical Operation :

4 Figure 3.6.K-5. The curves have been established to be in conformance with Appendix G to 10 CFR l Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil- I

ductility transition temperature (RTuo7) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel material is used to account for irradiation effects.

1 i Four vessel regions are considered for the development of the pressure-temperature curves: 1) the i core beltline region; 2) the non-beltline region (other than the closure flange region and the bottom i head region); 3) the closure flange region and 4) the bottom head region. The beltline region is

defined as that region of the reactor vessel that directly surrounds the effective height of the reactor

! core and is subject to an RTuo7 adjustment to account for radiation embrittlement. The non-beltline, l closure flange, and bottom head regions receive insufficient fluence to necessitate an RTuo7

adjustment. These regions contain components which include; the reactor vessel nozzles, closure j flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly j surround the reactor core. Although the closure flange and bottom head regions are non-beltline j regions, they are treated separately for the development of the pressure-temperature curves to address i 10CFR Part 50 Appendix G requirements.

Boltuo Temoerature l

! The limiting initial RTuor of the main closure flanges, the shell and head materials connecting to

, these flanges, connecting welds and the vertical electrostag welds which terminate immediately i below the vessel flange is 23 F. Therefore, the minimum allowable boltup temperature is i established as 83 F (RTuor + 60 F) which includes a 60 F conservatism required by the original

) ASME Code of construction.

J Fiaures 3.6.K-1 throuah 3.6.K Pressure Testina As indicated in Figure 3.6.K-1 through 3.6.K-3 for pressure testing, the minimum metal j temperature of the reactor vessel shellis 83 F for reactor pressures less than 312 psig. This 83 F l minimum boltup temperature is based on a RT uor M MT h h eMdag wW hadaWy i below the vessel flange and a 60 F conservatism required by the original ASME Code of construction. The bottom head region limit is established as 68 F, based on moderator temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113 F. The 113'F minimum temperature is based on a closure flange region RT uor of 23*F and a 90 F conservatism required by 10CFR Part 50 Appendix G. Beltline curves as a function of vessel exposure for 18,20 and 22 effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 22 EFPY of operation. i DRESDEN - UNITS 2 & 3 B 3/4.6-6 Amendment Nos.

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES Figures 3.6.K-1 through 3.6.K-3 are goveming for applicable pressure testing with a maximum heatup/cooldown rate of 20 F/ hour.

Fioure 3.6.K Non-Nuclear Heatup/Cooldown Figure 3.6.K-4 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram). The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress. The maximum heatup/cooldown rate of 100 F/ hour is applicable.

Fioure 3.6.K Core Critical Operation The core critical operation curve shown in Figure 3.6.K-5, is generated in accordance with 10CFR Part 50 Appendix G which requires core critical pressure-temperature limits to be 40 F above any pressure testing or non-nuclear heatup/cooldown limits. Since Figure 3.6.K-4 is more limiting, Figure 3.6.K-5 is Figure 3.6.K-4 plus 40 F. The maximum heatup/cooldown rate of 100 F/ hour is applicable.

The actual shift in RTuor of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTI E185-82 and 10CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside w'all of the reactor vessel in the core area.

The irradiated specimens are used in predicting reactor vessel material embrittlement. The operating limit curves of Figures 3.6.K-1 through 3.6X-5 shall be adjusted, as required, on the basis of the specimen data and recommew ans of Regulatory Guide 1.99, Revision 2.

3/4.6.L Reactor Steam Dome Pressure The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria. The reactor steam dome pressure of s1005 psig is an initial condition of the vessel overpressure protection analysis. TMs analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

3/4.6.M Main Steam Line Isolation Valves Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not DRESDEN - UNITS 2 & 3 B 3/4.6-7 Amendment Nos.

I

- _ - . _ - _ . - = _ _ - -

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.6.N Etructural Intearity The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

The inservice inspection program for ASME Code Class 1,2 and 3 components will be performed 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 NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

3/4.6.0 Shutdown Coolina - HOT SHUTDOWN 3/4.6.P Shutdown Coolina - COLD SHUTDOWN Irradiated fuel in the reactor pressure vessel generates decay heat during normal and abnormal shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant.

This decay heat is required to be removed such that the reactor coolant temperature can be reduced in preparation for performing refueling, maintenance operations or for maintaining the reactor in cold shutdown conditions. Systems capable of removing decay heat are therefore required to perform ,

these functions.

A single shutdown cookog mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that attemate methods capable of decay heat j removal be demonstrated and that an attemate method of coolant mixing be in operation. I j

DRESDEN - UNITS 2 & 3 B 3/4.6-8 Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K. Pressure / Temperature Limits K. Pressure / Temperature Limits The primary system coolant system 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal and pressure testing operations, at least ;

temperature and pressure shall be limited as once per 30 minutes, j specified below: i

a. The rate of change of the primary

1. Pressure Testing: system coolant temperature shall be 1 determined to be within the heatup and !

a. The reactor vessel metal temperature cooldown rate limits, and I' and pressure shall be maintained within the Acceptable Regions as shown on b. The reactor vessel metal temperature Figures 3.6.K-1 through 3.6.K-3 with and pressure shall be determined to be the rate of change of the primary within the Acceptable Regions on system coolant temperature s 20'F per Figures 3.6.K-1 through 3.6.K-4.

hour. .

2. For reactor critical operation, determine l

b. The rate of change of the primary within 15 minutes prior to the withdrawal of l system coolant temperature shall be control rods and at least once per 30 l s100*F per hour when reactor vessel minutes during primary system heatup or l metal temperature and pressure is cu:,ldown, i maintained within the Acceptable Regions as shown on Figure 3.6.K-4. a. The rate of change of the primary system coolant temperature to be within

2. Non-Nuclear Heatup and Cooldown and the limits, and low power PHYSICS TESTS:

b. The reactor vessel metal temperature ;

a. The reactor vessel metal temperature and pressure to be within the i and pressure shall be maintained within Acceptable Region on Figure 3.6.K-5 i the Acceptable Regions as shown on !

l Figure 3.6.K-4, and 3. The reactor vessel material surveillance ,

specimens shall be removed and

!. b. The rate of change of the primary examined, to determine changes in reactor ,

j system coolant temperature shall be pressure vessel material properties in l l s100*F per hour. accordance with 10CFR Part 50, Appendix ;

! H.

1 e

i 1

i i

QUAD CITIES - UNITS 1 & 2 3/4.6-19 Amendment Nos.

1

- - .- - - , , e - ,m- - - - -

TABLE OF CONTENTS TOC Llh,! TING CONDITIONS FOR OPERATION AND SURVEILL \NCE REQUIREMENTS.

EQTION PAGE 3/4.6 PRIMARY SYSTEM BOUNDARY

3/4.6.A Recirculation Loops . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4/6-1 l 3/4.6.B J et P u mp s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-3 i

i 3/4.6.C Recirculation Pumps. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-5 l 3/4.6.D Idle Recirculation Loop Startup. . . . . . . . . . . . . . . . . . . . . 3/4.6-6 1

3/4.6.E S afety Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-7 j 3/4.6.F Relief Valves. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-8 3/4.6.G Leakage Detection Systems. . . . . . . . . . . . . . . . . . . . . . . 3/4.6-10 l 3/4.6.H Operational Leakage. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-11

! 3/4.6.1 Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-13 J

l Table 3.6.l-1, Reactor Coolant System Chemistry Limits 3/4.6.J Specific Activity. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-16 3/4.6.K Pressure / Temperature Limits '. . . . . . . . . . . . . . . . . . . . . . 3/4.6-19 Figure 3.6.K-1, Pressure-Temperature Limits for Pressure Testing - Valid to 18 EFPY Figure 3.6.K-2, Pressure-Temperature Limits for Pressure Testing - Valid to 20 EFPY Figure 3.6.K-3, Pressure-Temperature Limits for Pressure Testing - Valid to 22 EFPY Figure 3.6.K-4, Pressure-Temperature Limits for Non-Nuclear Heatup/Cooldown - Valid to 22 EFPY Figure 3.6.K-5, Pressure-Temperature Limits for Critical Core Operations - Valid to 22 EFPY 3/4.6.L Reactor Steam Dome Pressure . . . . . . . . . . . . . . . . . . . 3/4.6-22 3/4.6.M Main Steam Line isolation Valves. . . . . . . . . . . ...... 3/4.6-23 3/4.6.N Structural Integrity. . . . . . . . . . . ................. 3/4.6-24 3/4.6.0 Shutdown Cooling - HOT SHUTDOWN . . . . . . . . . . . . . 3/4.6-25 3/4.6.P Shutdown Cooling - COLD SHUTDOWN . . . . . . . . . . . . 3/4.6-27 QUAD CITIES - UNITS 1 & 2 Vill Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQ!JIREMENTS K. Pressure / Temperature Limits K. Pressure / Temperature Limits The primary system coolant system 1. During non-nuclear heatup or cooldown, temperature and reactor vessel metal and pressure testing operations, at least temperature and pressure shall be limited as once per 30 minutes, specified below;

a. The rate of change of the primary

1. Pressure Testing: system coolant temperature shall be determined to be within the heatup and

a. The reactor vessel metal temperature cooldown rate limits, and and pressure shall be maintained within the Acceptable Regions as shown on b. The reactor vessel metal temperature Figures 3.6.K-1 through 3.6.K-3 with and pressure shall be determined to be the rate of change of the primary within the Acceptable Regions on system coolant temperature s 20*F per Figures 3.6.K-1 through 3.6.K-4.

hour,or

2. For rt actor critical operation, determine

b. The rate of change of the primary within 15 minutes prior to the withdrawal of system coolant temperature shall be control rods and at least once per 30 s100*F per hour when reactor vessel minutes during primary system heatup or metal temperature and pressure is cooldown, maintained within the Acceptable Regions as shown on Figure 3.6.K-4. a. The rate of change of the primary system coolant temperature to be within

2. Non-Nuclear Heatup and Cooldown and the limits, and low power PHYSICS TESTS:

b. The reactor vessel metal temperature

a. The reactor vessel metal temperature and pressure to be within the and pressure shall be maintained within Acceptable Region on Figure 3.6.K-5.

the Acceptable Regions as shown on Figure 3.6.K-4, and 3. The reactor vessel material surveillance specimens shall be removed and

b. The rate of change of the primary examined, to determine changes in reactor system coolant temperature shall be pressure vessel material properties in s100*F per hour, accordance with 10CFR Part 50, Appendix H.

)

l QUAD CITIES - UNITS 1 & 2 3/4.6-19 Amendment Nos. !

i

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K F

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS

3. Nuclear Heatup and Cooldown: 4. The reactor vessel flange and head flange temperature shall be verified to be 283'F:

a. The reactor vessel metal temperature and pressure shall be maintained within a. In OPERATIONAL MODE 4 when the the Acceptable Region as shown on reactor coolant temperature is: ,

Figure 3.6.K-5, and j

1) s113*F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l

b. The rate of change of the primary j system coolant temperature shall be 2) s93*F, at least once per 30 minutes.

s100*F per hour.

b. Within 30 minutes prior to and at least

4. The reactor vessel flange and head flange once per 30 minutes during tensioning of temperature 283 'F when reactor vessel the reactor vessel head bolting studs.

head bolting studs are under tension.

I APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded,

1. Restore the reactor vessel metal temperature and/or pressure to within the limits within 30 minutes without exceeding the applicable primary system coolant temperature rate of change limit, and

2. Perform an engineering evaluation to determine the effects of the out-of-limit I condition on the structural integrity of the reactor coolant system and determine that the reactor coolant system remains acceptable for continued operathns within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or

3. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

QUAD CITIES - UNITS 1 & 2 3/4.6-20 Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 1

I FIGURE 3.6.K-1 i PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 18 EFPY i 1400 o

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QUAD CITIES - UNITS 1 & 2 3/4.6-21a Amendment Nos.

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

] PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K

! FIGURE 3.6.K-2 P_RESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 20 EFPY

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QUAD CITIES - UNITS 1 & 2 3/4.6-21b Amendment Nos.

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K i

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QUAD CITIES - UNITS 1 & 2 3/4.6-21C Amendment Nos.

l PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K !

FIGURE 3.6.K-4 PRESSURE - TEMPERATURE LIMITS FOR NON-NUCLEAR HEATUP/COOLDOWN - VAllD TO 22 EFPY I

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m __.-__ 1 '_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

1 PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-5 l,

j PRESSURE - TEMPERATURE LIMITS FOR CRITICAL i

CORE OPERATIONS - VALID TO 22 EFPY 1

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1 l

i i

I

, PRIMARY SYSTEM BOUNDARY B 3/4.6

)

BASES 3/4.6.J Specific Activity l The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. The values for the 4

limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of 4 typical site locations. These values are conservative in that specific site parameters, such as site i boundary location and meteorological conditions, were not considered in this evaluation.

i i The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT l-131, but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT l-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking j phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a l steam line rupture occur outside containment. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time i to take corrective action.

3/4.6.K Pressure / Temperature Limits

{

4 All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1.1 of the UFSAR. During startup and

shutdown, the rates of temperature and pressure changes are limited so that the maximum specified a

' heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The pressure-temperature limit lines are shown, for operating conditions; Pressure Testing, Figures

. 3.6.K-1 through 3.6.K-3 Non-Nuclear Heatup/Cooldown, Figure 3.6.K-4 and Core Critical Operation Figure 3.6.K-5. The curves have been established to be in conformance with Appendix G to 10 CFR

Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil-

ductility transition temperature (RTnor) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel material is used to account for irradiation effects.

Four vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non-beltline region (other than the closure flange region and the bottom head region); 3) the closure flange region and 4) the bottom head region. The beltline 4

J QUAD CITIES - UNITS 1 & 2 B 3/4.6-5 Amendment Nos.

i 1

PRIMARY SYSTEM BOUNDARY B 3/4.6 l

l BASES 1

i

region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core and is subject to an RTuor adjustment to account for radiation embrittlement. The non-i beltline, closure flange, and bottom head regions receive insufficient fluence to necessitate an RTuo1 4 adjustment. These regions contain components which include; the reactor vessel nozzles, closure i flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core. Although the closure flange and bottom head regions are non-beltline j regions, they are treated separately for the development of the pressure-temperature curves to address

10CFR Part 50 Appendix G requirements.

Boltuo Temperature

. The limiting initial RTuor of the main closure flanges, the shell and head materials connecting to these flanges, connecting welds and the vertical electroslag welds which terminate immediately j below the vessel flange is 23 F. Therefore, the minimum allowable boltup temperature is

{

established as 83*F (RTuor + 60 F) which includes a 60 F conservatism required by the original

,- ASME Code of construction.

Fiaures 3.6.K-1 throuah 3.6.K-3 Pressure Testina l

! As indicated in Figure 3.3.6.K-1 through 3.6.K-3 for pressure testing, the minimum metal i temperature of the reactor vessel shell is 83"F for reactor pressures less than 312 psig. This 83*F

minimum boltup temperature is based on a RTuor of 23"F for the electroslag weld immediately j below the vessel flange and a 60 F conservatism required by the original ASME Code of

construction. The bottom head region limit is established as 68 F, based on moderator -

i temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312

psig, the minimum vessel metal temperature is established as 113*F. The 113 F minimum l temperature is based on a closure flange region RTuor of 23'F and a 90 F conservatism required l by 10CFR Part 50 Appendix G. Beltline curves as a function of vessel exposure for 18,20 and 22

! effective full power years (EFPY) are presented to allow the use of the appropriate curve up to 22 i EFPY of operation.

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Figures 3.6.K-1 through 3.6.K-3 are goveming for applicable pressure testing with a maximum heatup/cooldown rate of 20 F/ hour.

1 i Fioure 3.6.K Non-Nuclear Heatuo/Cooldown i

! Figure 3.6.K-4 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during

cooldowns when the reactor is not critical (e.g., following a scram). The curve provides the minimum j reactor vessel metal temperatures based on the most limiting vessel stress. The maximum

heatup/cooldown rate of 100 F/ hour is applicable.

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l QUAD CITIES - UNITS 1 & 2 B 3/4.6-6 Amendment Nos, j

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PRIMARY SYSTEM BOUNDARY B 3/4.6 '

4' BASES l

Fiaure 3.6.K Core Critical Operation !

!' The core critical operation curve shown in Figure 3.6.K-5, is generated in accordance with 10CFR Part t i 50 Appendix G which requires core critical pressure-temperature limits to be 40'F above any Pressure j i testing or non-nuclear heatup/cooldown limits. Since Figure 3.6.K-4 is more limiting, Figure 3.6.K-5 is i

Figure 3.6.K-4 plus 40 F. The maximum heatup/cooldown rate of 100*F/ hour is applicable. !

3 The actual shift in RTwo7 of the vessel material will be established periodically during operation by i removing and evaluating, in accordance with ASTM E185-82 and 10CFR Part 50, Appendix H, irradiated j

[ reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

j The irradiated specimens are used in predicting reactor vessel material embrittlement. The operating limit l

curves of Figures 3.6.K-1 through 3.6.K-5 shall be adjusted, as required, on the basis of the specimen l data and recommendations of Regulatory Guide 1.99, Revision 2. '

I i 3/4.6.L Reactor Steam Dome P,rggrg

The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and transients j and is also an assumed value in the determination of compliance with reactor pressure vessel '

j overpreuure protection criteria. The reactor steam dome pressure of s1005 psig is an initial condition of ;

i the vessel overpressure protection analysis. This analysis assumes an initial maximum reactor steam I I dome pressure and evaluates the response of the pressure relief system, primarily the safety valves, i j during the limiting pressurization transient. The determination of compliance with the overpressure criteria !

j is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that l the assumptions of the overpressure protection analysis are conserved. .

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i 3/4.6.M Main Steam Line Isolation Valves r I

Double isolation valves are provided on each of the main steam lines to minimize the potential leakage

- paths from the containment in case of a line break. Only one valve in each line is required to maintain the ;

i integrity of the containment, however, single failure considerations require that two valves be OPERABLE. 4 i The surveillance requirements are based on the operating history of this type of valve. The maximum i closure time has been selected to contain fission products and to ensure the core is not uncovered

following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses -

l to prevent pressure surges. ;

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I 3/4.6.N StructuralIntearity l The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity .

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

i The inservice inspection program for ASME Code class 1,2 and 3 components will be performed in 3 accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as ,

I j QUAD CITIES - UNITF &2 B 3/4.6 7 Amendment Nos.

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PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

3/4.6.0 Residual Heat Removal- HOT SHUTDOWN 3/4.6.P Residual Heat Removal- COLD SHUTDOWN Irradiated fuel in the reactor pressure vessel generates decay heat during normal and abnormal shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant. This decay heat is required to be removed such that the reactor coolant temperature can be reduced in preparation for performing refueling, maintenance operations or for maintaining the reactor in cold shutdown conditions.

Systems capable of removing decay heat are therefore required to perform these functions.

A single shutdown cooling mode subsystem provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two subsystems be OPERABLE or that altemate methods capable of decay heat removal be )

demonstrated and that an attemate method of coolant mixing be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associtated piping and valves. The two subsystems have a common suction source and are allowed to j have a common heat exchanger and common discharge piping. Therefore, to meet the Limiting Condition i for Operation, both pumps in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems (the ability to take credit for a common heat exchanger and discharge piping only applies to the SDC mode of RHR).

l QUAD CITIES - UNITS 1 & 2 B 3/4.6-8 Amendment Nos.

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