ML20116C834

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Proposed Tech Specs,Reflecting Revised Heatup & Cooldown Curves & Revised Limits for PORV Setpoints
ML20116C834
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 10/29/1992
From:
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20116C828 List:
References
NUDOCS 9211050051
Download: ML20116C834 (31)
v • d • e


Text

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4 Attact; ment . to Enclosure 1 of TXX-92537 I

4 E0RV ANJLjiEATUP/C00LDOH8 TECHNICAL SPECIFICA110N CHANGE-(F IGURES__3.4-2. 3.4-L 3 4- 6) i I

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__ c FIGURE 3.4-2 REACTOR COOLANT SYS1EM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY COMANCHE DEAK - UNIT 1 3<4 4-21

, A rta<h.ma g j Jo L% Io w I c> -( T K K' Y2 [3 7 Sy %d MATERI AL PROPERTY ILMLS bgERT

~B CONTROLLING MATERIAL:

LOWER SHELL PLATE R1108-1 (UNIT 1) ,

INTERNEDIATf SNELL PLATE R3807-2 (UNIT 2)

INiilAL RTNDT: 0'F (UNIT 1), 10'F (UNIT 2)

ART AT 16 EFPY: 1/4T : 84'f IljNIT 1), 81'F (UNIT 2) 3/4T : 69'F (UNIT 1). 62'F (UNIT 2)

CURVES BOUhDING COMANCHE PEAK UNITS 1 AND 2.APPLICABLE FOR HEATUP RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP T0 lb EFPY. CONTAINS HARGINS OF 10*f AND 110 PSIG FOR POSSIBLE INSTRUMENTATION ERRORS.

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e *-n*.s-e t:w; ress crecer*1es of t~e 'erri++: .aters31s 4, t e ,en:::r

,essel tre :ete -'"e; 'n 3c:Or0Srce witn the '4RL itancar: Review ;'s , 45TM E;35-s; r: in at;crcance 4th 3:0itio"3 1 reacter vessel re0vire*erts. Trese

rt;ert'es are t en esal vat)d 'n 3cc *0se
e .ith Appenc1x G of t*e !?i6 E 1ti:n ,

to iectica !!! Of the ASP! E011er 3rd Eressure vessel Osde and the 031 : s 5 t i : n ,

+ein005 descr1 Led in aCAE-7324-4, '345 is for -eatu and Coolcc.n Limit Curses. '

Apt il 1975, i

r est.; ana ccclcown limit curves are calculatec using tne Tost limiting 3'ae Of tre nil-auctility referente tetterature, RTNDT, at he end ;f 16 ef'e:t' s e f ull p0wer years (ECL') of service life. The 16 [FPY service

'd<e cer'; is cnosen such that the limiting kT at the 1/4T locati;n i+

NOT the c;ie re:fon is greater inan the R}w,. of tne limiting unitradiate .

2 ateriel.

?*e seie:ticq of such a limiting RT FOT assures that all components - the

. Oeatt:r :: lant 5,, tem .419 te ;erated : nservatively in a:corcance *itn 3;plicat i e Coce requirements- ~**'~~

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' e reactor vessel materia's have been teste to etermine treir 'nitial Ri gu e . ; tne results of these tests are shown in Table : 3/4,4 1. React:r :; era-

'iO^ sr result 3nt f35t neutron (E greater than 1 Me; irradiation :ar cause an -rcrease in tre RT NDT. Therefore, an acjusted re? 'ence temperature, based s

seen t e #1cence, and the enemical content of tre mats dai in cuestien, has teea tredicted using Regulatory Guide 1.99, Revision 2, "Raciation E :r4 tile *ent i i of React;r Vessel "Sterials" The flu 6nct values for 15 EFDY is taAer ' rom t*e 26.5 :e;ree plot 'n Figure B 3/4.4-1. Tne heatue anc cooldown liuit car es of s

1gures 3.4-2 and 3.4-3 include precicted adjustments for this shift 'n RT,w. -n -

i i- ut t; la EFFT as -ell as adjustments for cossible errors in the pressure 3r:

l temperature sensing instruments.

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values of 2RT NDT determi ed in this manner may 'e used until-the results i fecm the material Surveillance program, evaluated according te AsiM E;5E, are availacle. Capsules will be removed in accorcance with the require *ents Of ASTM E155-82 and 10 CFR-50, Appendix H. The surveillance specimen with-crawal cchedule is shown in Table 4.4-2. The lead factor represents t"e rela- ,

tiensr + tetween the f ast neutron flux censity at the location of the capsule anc the inner-.all of tre reactor vessel. Therefore, the results 00tained from the surveillance specimens can te used to predict futu radiation camage to the reactor ' ssel material cy using the lead factor and he withdra.al .

time of tne cap e. The heatup and 0001cown curves must be recalculated onen the ART,40,i dete ned from the surveillance capsule exceecs the calcu l ate:

2RT ,, for the equivaler;t capsule radiation exposure.

Nui

MAN;-E :EAK - UNIT 1 E 3/4 c-7 t- i i -

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AEten0's 2, 't " 0 thei6 *et*C 5 ?re O't;,3se] **

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  • re q<reral ethod 'Or :a' ulati g heat-p anc oc! :.n limit ; r ses is based uncn the prirc';les o' the s irear Elastic Fracture u echanics (LE;M) tecn" ology In t e cal:siat'or croce0 ares a seM elli;t, cal surface de'e-t 1th a depth cf one quarter O' t*e -311 thic ress, * , and 5 length of 3/2T

- 's assu*ed to esist at the 'rs10e o' ine .essel all ai aell is at the -

outside o' tre vessel aal' ihe Oi e9sions O' t*is postulate Orack, referred to in Apterdi, G Of AiuE iection I!! as tre refererce flaa, 3L ij enceed t*e current catat,'4t,es o' ,nservice inspection techniaves.

' erefore, the reactor cterat*cn 'imit c u r >. 0 5 developed 'or this referecce cr3c. are ;;nservatt e anc trov ide su'ficient safet 1 *argins for prot 9: tic- .

sq31rst nor cuttile (311ure. ~ 3 assure inat t*e radiaticn emt,4ttle ent ef'ects are 30:Ounte11cn q ty _n l:alation of the limit curves, t e cst I'miting .3 1 ue cf the ' h.Ma.a.iJ,f f e f e r e n c e t e mp e r a t u r e ,

-+M"- T is usec - Ag7 and in i s i n c l de s t"e- r%dp9td a M m'Tp aucea snift, ;RT,g , correspording to

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t e ena of the perica for mch "estup and ccidown . .es are gener3tet.

Tne AIME app-encn for c3Ica'atirg ne allo-sble mit curses for sarious

' eat.T 3ra c;oldcan rates I;e:es t"at tne total 51 is intensit, ' actor, A,, 'r ' e cc"L'aec thermal ard pressure stresses at iny time during neatu:

Or ; cold:an :anect te greater than t*e ref erence stress intensi ty f 3ctor,  %. .

fc- t*e ~etal tet;erature at trat time. K)q 1s cttai ed f r:m the ref ere ce' '

'r 3;tre 1:uarnes s cur.e. ce#iaec in Apper dia G

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ahere: *, i s, is the ieference strdss .s j u s,k in' 4tensity 'ictor as a function [o' te metal temperature and the metal aM %+4M%)re f e re nc e te c e r a tuThus.

_. s y re the governing equation for the neatup-coalco n ar.alysis is definec in A;penois G of tne A5ME Coce as folloas:

Cayy + kit ' Kn, (2)

Where: g = the stress intensity ' actor caused by metorane (pressure) stress, y: tre St ess intensity factor caused by the thermal gradients, y D. = constant provice0 tv t*e Coce as a function of tem;erature relative to the f +uQ .of the p~aterial, _.

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't  !*. t' e Ou 4"; t*e *eatu; Or 0010;.n transient, K is Ceteraired tj I "- i t*e 'et al te*0etature at tre '.ip of the ;ostulated fla , the acerc0r' ate value for #1 [fandt*erefereace'racturetoughnesscurve. The thermal stresses "esulting dr09 te*;erits e gradients through ire vessel *all are calculatec and trer tre ;orres;;nding thermal stress intensity fact 0r, K,7 #0r 19e 1 refererce So is :D*cutec. Ircn Ecuativo (2) the pressure stress intensity fact;rs are cttairec at0, frc+ trese tr-e allowable Oressures are calculated.

- . .- ,. r. 4, Ter 're cal:ulatd:n c' t*e a'lowaDie pressure versus coolant tem erature durd"g c00'00 n, t'e Code reference flaw is assumed to exist at the ins 10e of

!*e essel -ati. D.r'r; c::ico.n, the controlling location of the flaa is al.ays et tre insice o' tre all because the thermal gracients produce tensile 1 tresses at t~e t s ce, nic" increase itn ircreasing cooldown rates. Ailowable '

' essure-temceratare relati
ns are generated for cott teady-state ana 'inite
c'00." rate s' t;at cr t. Tecm tnese relati0rs, com; ite limit curves are ,
nstructe: 'c- e3:- :: lc n rate of interest. ,

e use 0 t"e :otposite :ur<e in the cecico.n a' =ysis is necessary re:a se : ante:

0' tre c:clac n procedure is cased on easurement of rea; tor

c'a-t te recata<e ~"ereas t e limiting pressure is actually decencent on the ater'ai temterature at the tip of the assumed flawn During c:cleown, the ai cesse' tccat*:' 's at a higher temperature than tne fluid adjace .t to the

. esse  ::. 'nts ::acition, of course, is not true for '.ne steacy-state situa-t'a- :t ':1':-s init at any given reactor :colant temperature, the l' ce e ::e: ; r g c:cino.n results in a higner value of A,R at tne L 47 'l: cation A fhr fi nite c cla n rates than for steady-state operation. Furthermore, if conci;4:ns exist such that the increase in K n exceeds 4,,, :he calculated is 4. 3 alle.acle pressure caring c00ldown will be greater than the steady-state

.alue.

Tre ato.e ;roce0Lres are needed be:ause there is no cirect control on terce #ature at the 1/4T location; therefore. allowable cressures may unknowingly Oe .iciate; if tre rate of c00iing is uecreased at various intervals along a

coIdown camp. The use of tNe composite curve eliminates this pr: Diem and ass es conseriatise operation of the system for the ertire cc0ldo.n. period.

C: MAN:-! DEaK - UN:' i E 3/4 4-11 1

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i Atmachment 2 Enclosure 1 of to TXX-92537 Page 1 of 7 l t

DESCR1pTION AND ASSESSMENT I. Background During a technical review of Westinghouse calculations of-setpoints for the Cold Overpressure Mitigation System (COMS),

it was discovered that the differences in the pressures  ;

indicated by the wide range pressure transmitters (located on the Loop 1 and 4 hot loca) and the region of interest (typically, the core midplane elevation) had not been considered. Depending on the number of reactor coolant pumps in operation, this pressure difference could be as large as '

50 psi. In addition, dosimetry from the reactor ve t sel material irradiation surveillance capsule, removed during the "irst refueling outage of Unit 1, showed that the reactor vessel is being exposed to slightly less fluence than was ,

predicted in the original design. The changes proposed by this Licnnse Amendment Request have been prepared to account for this pressure difference and the results of the ,

surveillance capsule.

  • Even though the additional instrument uncetninty was not included in the COMS setpoint calculations and in the Appendix--

G heatup and cooldown limits, these limits have not been exceeded during operation of Unit 1. These heatup and cooldown limits are protected by the COMS or through the use of the RHR suction relief valves. The operation of the_RHR suction relief valves is unaffected by this iss"e. Because the actua) CoMS setpoints incorporated in the plant systems are lower than the revised ~ setpoints, the COMS was always

-wapable of providien adequate protection to ensure that the-Appendix G limits ber;rs not exceeded in the regions _ of- greatest l concern.

II. Description of' Technical-Specification Change Request--

i The af fected figuras are Technical Specification Figures 3.4--

l 2,- 3.4-3, and . 3. 4-4. In Figures 3.4-2 and 3.4-3, the i Appendix G heatup and cooldown curves _ are redrawn,:

incorporating the increased pressure indication _ uncertainty. ,

As a' result, the pressure limit was reduced by approximately

a. . - . . . . , _ . - . . - . _ . - - . - . . -

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1 i

Attachment 2 Enclosure 1 of to TXX-92537 Page 2 of 7 50 psi for each of the heatup rate, cricticality limit and cooldown cucves and for the leak test limit. In Figures 3.4-2, " Reactor Coolant System lleatup Limitations Applicable Up to 16 EFPY", and in Figure 3.4-3, " Reactor Coolant System Cooldown Limitations Applicable Up to 16 EFPY", the term

" RT.n " was replaced, where appropriate, with the more current ,

term " ART" (Adjusted Reference Temperature). This modification more closely follows the terminology of 10CFR50, Appendix G and Reference 1. In addition, the results_of the dosimetry obtained following removal of the first reactor vessel material irradiation surveillance capsule from the Unit i reactor vessel are incorporated. Specifically, the 1/4T ART decreased to 84F from 85F, and the 3/4T ART decreased from 70F to 69F. In addition, the criticality temperature limit, shown in Figure 3.4-2, increased from 229F to 230F. Added to the redraw.) figure 3.4-2 are the pressure-temperature curvcs for 20 F/ hour heatup rate. The revised Unit i heatup and ,

cooldown curves remain more restrictive than the Unit 2 curves.

In Figure 3.4-4, "PORV setpoints for Overpressure Mitigation-Applicable Up to 16 EFPY", the maximum PORV st tpoints for the COMS are presented. At temperatures less than-or equal-to '

237F, the maximum allowable PORV setpoint is reduced from 560 psig to 521 psig. The maximum allowable PORV setpoints for temperatures greater than 237F are unchanged.

The BASES.Section 3/4 4.8 was also revised to use the more current term " ART" where appreoriate.

l III. Analysis The Cold Overpressure Mitigation System _(COMS) is provided as  ;

~ ~

a backup to the reactor .norators in order to insure that the reactor vessel fracture _ toughness requirements as defined in 10CFR50, Appendix G are satisfied. The operatii ) of the COMS also. ensures that the forces generated by cycling of the l- pressurizer power operated relief valves (PORVs)-do nt c exceed .

the PORV piping and structural limits.

L

Attachment 2 1:ncl osu re 1 of to TXX-92537 page 3 of 7 The COMS is one of two available systems which are allowed, per Technical Specif ication 3.4.8. 3, f or use in the mitigation of low temperature overpressure (LTop) events. The Residual lleat Removal System suction relief valves may also be used for the mitigation of potential LTop events.

The COMS actuates one or both of the pressur!zer PORVu when the RCS pressure exceeds a temperature-dependent setpoint.

pressure indication is provided by wide range preocure _

transmitters located on the hot legs of Loops 1 and 4.

The Appendix G heatup and cooldown limits are presented as a maximum allowable pressure at a given temperature and are dependent on the heatup and cooldown rates of the reactor vessel. For an assumed temperature and heatup/cooldown rate, a maximum allowable pressure is calculated. Similar calculations are perf ormed on a component-by-component basis, and the NDT 1imits are compocito curves representing the

) maximum allowable pressure 1or any component in the system, llowever, the limiting location is typically near t' e core

\ midplane.

The Appendix G heatup and cooldown limits, with instrument uncertainties, are presented in Figures 3.4-2 and 3.4-3 of the plant Technical Specifications. These curves are used in the definition of the acceptable range of presE,ure and temperature in which operation is allowed. The wide range pressure instrumentation used to monitor compliance with the Appendix G heatup and cooldown limits is the same instrumentation used by the COMS. lle nce , the pressure dif ference identified in the previous discussion must also be applied to the heatup and cooldown limits presented in the Technical Specifications.

With respect to the changes depicted in Figures ^ 3.4-2 and 3.4-3, the reduction in the maximum temperature-dependent RCS pressure by 50 psia adds further assurance that the actual Appendix G heatup and cooldown limits will not be inadvertently exceeded due to instrument uncertainties.

I

. 1 Attachment 2 Enclosure 1 of to TXX-92537 l Page 4 of 7 l The pressure-temperature curves for 20 F/ hour heatup rates and for criticality were added to provide additional information to the operator. These curves were determined in the same manner as the other heatup curves and therefore have no additional impact on cafety.

In addition, dosimetry from the reactor vessel material irradiation surveillance capsule U, removed during the first refueling outage of Unit 1, showed that the reactor vessel is '

being exrved to slightly less fluence than was predicted in the original design. This lower fluence at 16 effective full power years (EFPY) caused the ART for the Unit i vessel material at the 1/4T and 3/4T locations to shift dow1 ward by 1F. ,

The IF increase in the criticality limit (from 229F to 230F) is caused by the relatively complex nature of the function used to calculate the criticality limit. This function is  ;

influenced by a variety of factors, which include the ART and the assumed instrument errors. The significant increase in j the pressure indication had a greater effect on. the l criticality limit than did the reduction in the ART; hence, the criticality limit was . increased. Relative to - plant - - -

operations and vessel life, the changes are insignificant.

Even though compliance with the heatup and cooldown curves is

required for all modes of operation, these curves do not I result in the imposition.of.real operational constraints if

! the RCS temperature is greater than 350F. Thus, the Cold l Overpressure Mitigation System would only be used, -if-L selected, in operational Modes 4, 5 and 6 when the reactor ,

vessel head was on. [

In order to address -- the impact of the pressure difference-Issue on the COMS PODV setpoints, Figure 3.4-4 was :tevised.

With respect to the w 4ximum PORV setpoints, the Appendix G limits are higher than- the PORV discharge piping stress limits. Due to elevation dif ferences, _ the RCS wide range

. pressure indication . used ' by the COMS _always indicates - a pressure which is higher than local pressure at the pressurizer PORV, thereby assuring that the PORVs will actuate t.

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Attachment 2 Enclosure 1 of 1 TXX-9253' Page 5 of 7 to wm al before the local pressure exceeds tbo applicable limit.

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p. , 4 At RCS temperatures below 237F, ther COMS setpojnts are h 3 elected in order to ensure compliance with the Appendix G the widt range pressure

,[ limits. As previously noted, sfu. itidications which are used by the COMS wil) indicate a pressure which is approximately 50 psi lower than the local prosau2a at the core midplane elevation. Following an investig :stion into the COMS analyses, a number of overly conservative assumptions were identified. Wher, 're realistic /

(but conservative) assumptions were used, it determ) that approximatelf 11 psi of the 50 psi pressi diffe7 -

could bi bsorbed in .u analysis. Thus, we ma>

allowabl- ORV sei ;;o i nt. wm reduced by 19 psi for ,J temperattu as below 37F.

Relative to the accident analyses presented in FSAR g Chapter 15, the prcposed revision to the maximum PORV setpoint has na effect. At the temperatures of concern for thesa events, the pressurizer FORVs are contrc11ed by a separate Pressurizer Pressure Control System which is independent of the COMS.

From the standpoint of normal plant operations, the proposed change has no impact. '1he PORV setpoints for the COMS which [

are actually used in the plant were lower than the maximum <

PORV setpoints presented in Figure 3 * ,. The plant PORV setpoints continue to be lower than ne revised maximuni PORV setpoints; thus, there is no impact on normal plant rg.e ra t i ons In summary, the proposed changer provide additional assur Ance that the actual reactor vessel heatup and cooldow. limits will not be exceeded.

IV. Significant llazardu Considerations Analys.is

1) Does the proposed change 1.avolve a significant increase in the probability or -

asequences of an accident .

previously evaluated?

l. -

l ..

Attachment 2 Enclosure 1 of to TXX-92537 Page 6 of 7 As noted in Section III, none of the accident analyses are affected by the proposed changes. 1herefore, the consequences of the analyzed accidents are unaffected by the proposed changes. In addition, the proposed changes add further assurance that the Appendix G heatup and cooldown curvec will not be exceeded; ther6y reducing the prcbability of occurrence of an event initiated by brittle fracture.

2) Does the proposed change create the possibility of a new or different kind of acciden- from any accident previously evaluated?

The proposed changes only affect the limits which define the acceptable regions for operation. No changes are proposed which could result ir, u 'lew or different kind of accident from any accident previously evaluated.

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

The margin of safety is based on the riif ference between the f ailure point of a particular 3 y item or component and the acceptance criteria established to ensure the fr.ilure point is not approached. For thj~ application, the acceptance criteria are defined by the heatup and cooldown limit curves and the margin of safety is mandated through the use of the metnods used to calculate the adjusted reference 1 nperature[1] and the heatup and cooldown curves [2]. Through the proper consideration of the capability of the pressure indication to represent the actual pressure of interest, tne propoced changes udd further assurance that the acceptance criteria will not be exceeded during normal plant operation.

Based on the above evaluations, TU Electric concludes that the activitic; associated with the proposed-changes satisfy the no significant hazards consideration standards of 10CFR50.92(c) and, accordingly, a no significant hazards corisideration finding is justified.

1

Attachment 2 Enclosure 1 of to TXX-92537 Page 7 of 7 V. Environmental rialuation TU Electric h u. evaluat. the proposed changes and has deteri:',ined that the changen Jo not involve (i) a e.ignificant hazards consideration, (ii) a significant change in the types or significant increase in the amount of uny effluents that may be released off-site, or (iii) a cignificant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed change meets the el3gibility -

criterion for categorical exclusion set forth in 10 CFR 51.22(c). Therefore, pursuant to 10 CFR 51.22(b), an environmental asseanment of the proposed change is not required.

VI. References

1. USNRC Regulatory Guic- 1.99, Re i sion 2.
2. ASME B&PV Code,Section III, Division 1, Subsection NB:

Clcss 1 Couponents, Appendix G.

_ _ _ _ _ _ . _ - _ _ - _ _ _ ____ ___ _ _-__-_____ __ - _ _a

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Enc?osure 2 to TXX-92537 1

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1 FEE 0 WATER ISOLATION VALVES L

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o TECHNICAL SPf.CIFICATION .

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SECTION PAGE 3/4.6.2 -DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System........................... ..... 3/4 6-11 Spray Additive System.................................... 3/4 6-12 3/4.6.3 CONTAINMENT ISOLATION VALVES... ..................... ... 3/4 6-13 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors.................... .............,..... 3/4 6-15 Electric Hydrogen Recombiners............................ 3/4 6-16.

, 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCmE Safety Valves.. .......... ....... ... .................. 3/4 7-1 TABLE 3.7-1 MAXIMUN ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING l

FOUR LOOP 0PERATION...................................... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER-L00P..................... 3/4 7-2 Auxiliary Feedwater System.. .......................... 3/4 7-3 Condensate Storage Tank.................................. 3/4 7-5 Specific Activity...................................... . 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..... ............................... 3/4 7-7 Main Steam Line Isolation Valves..................... ... 3/4 7-P.

Main Feedwatur Isolation Valves.................... ..... 3/4 7-9 Steam Generator Atmospheric Relief Valves................ 3/4 7r11 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-12 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4-7-13 3/4.7.4 STATION SERVICE WATER SYSTEM............................. 3/4 7-14 3/4.7.5 ULTIMATE HEAT SINX..../.................................. 3/4 7-15 3/4.7.6 ' FLOOD PRrTECTI0N......................................... 3/4 7-16 3/4.7.7 CONTROL ROOM HVAC SYSTEM................................. 3/4 7-17

3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS.. 3/4 7-20 3/4.7.9 SNUBBERS................................................. 3/4 7-22 3/4.7.10 AREA TEMPERATURE M0NITORING.............................. 3/4 7-23 TABLE 3.7-3 AREA TEMPERATURE MONITORING. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-24 3/4.7.11 UPS HVAC SYSTEM.......................................... 3/4 7-25 1 3/4.7.12 SAFETY CHILLED WATER SYSTEM..............................

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3/4 7-26 l !y% 7,13 t1Am FerownTcR Iso u4rron) VAttm Nt hsutE/77tsstergrutzg L1msf~ % y.3o COMANCHE PEAX - UNIT 1 viii Amendment No. 8

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A in&.a / h Em tonce 2o+ T vx- v 37

.d p 2 of7 INDEX

? BASES SECTIO PAGE 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS. . ... .. ... . . . .. B 3/4 5-1 3/4.5.4 REFUELING W/.TER STORAGE TANK. . . ... .. . .. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT. . ... . . .. . . ... .. . B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. . ....... .. .. B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES. .. . . ..... . . . . B 3/4 6-3 3/4.6.4 COMBUSTIELE GAS CONTROL... . .. . .. . .................. B 3/4 6-4 3/4.7 PLANT SYSTEM _S 3/4.7.1 TURBINE CYCLE. .. .... . . . . ... . ...... .. . B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.. .. . . B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYST9' .. ... .. .. ... B 3/4 7-4 3/4 7.4 STATION SERVICE WATER SYST97 . ... ... .. ........ B 3/4 7-4

/. S LLTIMATE HEAT SINK. . . .. ...... . . . . ~ . . B 3/4 7-4 3/4.7.6 FLOOD PROTECTION. . .

. ......... . .. . . ... .. B 3/4 7-4 3/1.7.7 CONTROL ROOM HVAC SYSTEM....,...... .........,............

l' .

B Q/4 7-5 3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS.. B 3/4 7-5 3/4.7.9 SNUBBERS. .. ... . ... .. .... . .. ......... B 3/4 7-5 3/4.7.10 AREA TEMPERATURE MONITORING., . ... . ... .. . .. . B 3/4 7-6 3/4,7.11 UPS HVAC SYSTEM. ..... ..... . . ....... .... . B 3/4 7-7 3/4.7.12 SAFETY CHILLED WATER SYSTEM..... .. .. ......... ... .. .. B 3/4 7-7 3/>l. 7. t 3 l1 Azw Fct:tuas. nowy .,; w 7 ,;w a y m n,n,,

3/4.8 ELECTRICAL POWER SYSTEMS umyr 6 h6/ 2--7 5/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION. . . . ........ .. ... . B 3/4 8-1 3/4.8.4 CTRICAL EQLIPMENT PROTECTIVE DEVICES.. ... . ........ B 3/4 0-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 80RON CONCENTRATION... .. ... ..... .... .... ... ... . B 3/4 9-1 3/4.9.2 INSTRUMENTATION.. . .. .... ................ . ........ B 3/4 9-1 3/4.9.3 DECAY TIME........ ... ..... ......... .. .... . B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONC.......... ... .. . ... B 3/4 9-1 3/4.9.5 COMMUNICATIONS....... .... . ........ ....... .. . ... .. B 3/4 9-1 3/4.9.6 REFUELING MACHINE. .. . .... ............. . . ..... 8 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAG: AREAS....... ..... ..... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REP 0/AL AND COOLANT CIRCULATION. . . . . . .. B 3/4 9-2 3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and IRRADIATED Ft;EL STORAGE . ... .. . .................. ... B 3/4 9-3 COR$NCHI PEAK - UNIT 1 xii Amendment No. 8

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

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fg'[3. 4: 7 d PLANT SYSTEMS 3/4.7.13 MAIN FEE 0 WATER ISOLATION VALVE PRESSURE / TEMPERATURE LIMIT LIFITING CONDITION FOR OPERATION J.7.13 The valve body and neck of each mahe feedwater isolation valve shall be greater than or equal to 900F, when feedwater line pressure is greater than 675 psig.

APPLJCABILITY: MODES 1, 2, 3 and during pressure testing of the steam generator or main feedwater line.

ACTION:

With one ter more main feedwater isolation valves outside of the abeve limits, restore main feedwater isolation valve pressure and/or temperature to within the limits w'ithin one hour, and perform in engineering evaluation to determine the effect of the overpressure on the structural integrit) of the main feedwater isolation valve (s) and determine that the main feedwater isolation valve (s) remains acceptable for continued operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2)RVEILLANCE REQUIREMENTS 4.7.13 Each main feedwater isolation valve shall be determined te be greater than or equal to 900F at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.*

  • Except in MODE-1 with the main feedwater isolation valve open.

COMANCHE PEAK -UNIT 3 1 AND 2 3 3/4 7-30 CRAFT TS i

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J /\)$fTC T b 3/4.7.13 MAIN FEE 0 WATER ' SOLATION VAL'!E PRESSURE / TEMPERATURE LIMIT The fracture toughness requirements are satisfied with a metal temperature of 900F for the main feedwater isolation valve body and neck, therefore, the*,e portions will b', maintained at or above this temperature prior to pressurintion of these vs.lves above 675 psig. ,

Minimum temperature limitations are imposed on the valve body and neck

' of main feedwater isolation valves HV-2134, HV-2135. HV-2136 and V-2137.

These valves do not need to be verified at or above 900F

'j, when in MODES 4, 5, or 6 (except during special pressure testing)

, since Tavg < 3500F which corresponds to a pressure at the valves of 140-150 psig or 'tess. The maximum pressuri:ation during cold conditions (valve temperature < 900F) should be limited to no more thr 20% of the valve hydrostatic test pressure (3375 psig x 20% = 675 psig). '

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3.4.9 Tne tra;;.<3: :stegrity of A5ME Csde (1955 1, 2. and 3 c0m;onents s

t srall ba "aintairec in act0rcance ai tn 50ecificaticn 4.4.i.

ADDtICAE!tITY: all MOCE3-ACTION:

a.

With tne structural inteccity of any ASME Code Class 1 ccmoonent(s) not conforming to tre acove requirements,-restore the structural integrity of the affectec component (s) to within its-limit or isolate

'he af fectec component (s) prior to increasing the Reactor Coolant L

Syst,m temperature.... ore than 50"F reove the minimum temperature equirec Oy NCT considerations.

b. With tre structural integrity of any ASME C0de Class 2 comc;nent(s) not conforming.tzo.tne above requirements, restore the structural integrity of t*e affected component (s) to within its limit or isolate
  • tre affecte component (s) prior to increasing the Reactor Cnciant System tem;erature above C00 F.
c. Witn tre structural integrity of any ASME C + Class 3 component (s) not conforming to the acove requirements, re-tore the structurai-integrity of tne affectec. component (c) to.w --in it: limit or isolate-the affected ccmoonent(s) from service.

SURVE!L;ANCE-RECUIREMENT5 --=

4.4,3

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in a~~ .icn to the requirements of Specification 4.0.5~. eacn reactor coolant oumo .swneel shall-ce inspected per.the recommencations of Rt,.ulatory Position C. 4.b of Regulatory Guice 1.14, Revision 1, August '1975.

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tiote: Separate Technical Specifications provide specific temperature / pressure limitations for specific components (e.g.,

3/4.4.8.1 for the Reactor Coolant System, 3/4.7.2 for, the Steam Generators, 3/4.7.13 for the Main Feedwater Isolation Valves, etc.)

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Attachment 2 to Enclosure 2 of TXX-92537 DESCRIPTION AND ASSESSMENTS 9

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Attachment 2 to Enclosure 2 of TXX-92537 Pege 1 of 3 MAIM FEEDk'ATER ISOLAT10N VALVE PRESSURE / TEMPERATURE LIMIT DESCRIPTION AND ASSESIMEHf

1. BACKGROUND 1he CPSES Unit 1 Technical Specifications do not currently include LCO, surveillance rcquiraments or BASrS for the Main Teedwater Isolation Valve Pressure / Temperature Limit. The pressure / temperature limit for these valves is presently administratively controlled via the Technical Requirements Manual (TRM 2,3). In order to be more consistent with other pressure / temperature limits are controlled ,

(e.g., the RCS in 3/4.4.8, the System Generator in 3/4.7.2, etc.),

this proposed technical specification cnange is provided for the 4 combines Unit Technical Specification to move the TRM requirements to the Technical Specifications 4 b

11. UESCRIPTION OF TECHNICAL SPECIFICATION CHANGE REQUEST This change adds Technical Specification (3/4.7.13) for the Feedwate-Isolation Valve Pressure / Temperature Lim.t to the Comanche Peak Stehm Electric Station (CPSFs) Unit 1 Technical Specifications (NUREG- <

1349, Reference 1). i ts change assures tnat appropriate valve temperatures (190 F) are maintained befcre and while these valves 'i are pressurized above 675 psig.

The change to the Technical Specification consists of the Feedwater Iso'ation Valve Limiting Condition for Operation (LCO) 3.7.13, .

including the mode applicability and the required actions based on 3.03, and system surveillance reqJiremants (4.7.13). This request ,

also pioposes that the basis for the new technical specification be -

added to the BASES.

The change to the Tectmical Speci ication (3/4.4.9) for the structural integrity is to add a fontnote which identifies specific

, temgerature/ pressure limitations for specific components of the CPSES j "Jnit 1 Technical Specification.

J 111. i,NALYSIS The requested specification provides explicit identification of the pressure / temperature limitation for feedwater Isolation Vaes in a manner consistent with the existing technical specifications for cther similar pressure / temperature limits.

The proposed LCO and surveillance requirements are consister,t with comparable safety related support systems and 3.03 of the Technical Specification for CPSES Unit 1. These controls and surveillance requirenents already exist in the TRM 3.3-and thctr relocation to-the combined unit technical specifications is primarily an administrative ,

change and has ro impact on safety. >

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Attachment 2 to Enclor.ure 2 of TXX42537 Page 2 of 3

!V. 57GNIFICANT HAZARDS CCdS10 ERAT 10N DETERMUiATION TU Electric 'nas ovaluated whether or not a significant ho2ards consideration is involved with the proposed thanges by focusing on the three standardt set forth in 10Clk50.92(c) as dist.ussed below:

Does the proposed change- .

1. Involve a significant increasc in the prcbability or ,

consequences of 6n accident previously evaluated?

4 The proposed LCC does not alter any of the assumptions used in

safety analy'es for CPSES Units ) because the capability of the Feedwater Isolation Valves to perform their intended function ,

under normal and accident conditions is maintained. The LCO -

, ensures suitable temperatures are maintained tu prevent flaws from propa. gating in the valm body.

I The proposed surveillanco requirements will enhaM e tre l

, reliability of the vdives hec 4use they will enforte 6 periodic survoillance during MODES of operation and condit inns where valve integrity cou'.d be challenged by pressure /tc10erature changes. This surveillance will assure all Feedwater ltolation valves will perform its safety function.

'. These surkeillance requirements are commensurate with the Containment isolation surveillance requirements, in addition essentially the same Controls ulready exist in the plants administrative controls (the TRM 3.3), making the change to add a technical specification more administrative in nature ---

than technical.

Therefore, the prcposed Technical Specification change has no effect on the probability or consequences cf any accident L previcusly evaluated for CPSES Unit 1. I E

P. Create the cessibility of a new or different kind nf accident g from any accident previously evaluated?

The proposea LCO, surveillance rerjuirements and BASES do not involve any hardware changes or any revisions in how the  :-

Feedwater Isolat ion Valves functions. Therefore, no new failure 1 modes are created ar.d no possibility for new or different kinds k of accidents is created. E 3.

W Involve a sigrificant reduttion in the margin of safety? g I

The margin of safety for the feedwater Isolation Valves relative y to Technical Specificaticns is preservetion of the integrity o' a The propcsed LCO with respect to the steam generaters o

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Attachment 2 to Enclosure 2 of TXX-92537 '

'Page 3 of 3 feedwter Isolation Valves Pressure / Temperature Lirdt satisfies '

the isolation criteria provided.

The Nuclear Regulatory Commission has provided a guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certein examples (51 FR 7751, Reference 7) of amendments that are considered not likely to involve significant hazards consideration. Example (1) relates to a change to achieve o snei'.'ency throughout the Technical Specifications. Example (ii).- .es to change that constitutes an additional limitation, restrict % or control not presently included in the technical spec nication',,

in this case, the prop m Limiting Condition for Operation desciibed' above is similar to r,L, ale (i) in that the feedwater Isolatlon Valve Pressure / Temperature Lim ts specification is consistent with other similar systems (e.g. Containment Isolation). The proposed

surveillance described above is similar to example (ii) in that adding the technical . specification for the Feedwater Isolation valves L Pressure / Temperature Limits constitutes an additional limitatio.,

(e.g. an additional surve*llance) not presently included in the Unit 1 Technical Specif(cations.

i' Bated on the above evaluations, TU Electric concludes that the activities associated with the above described changes present no l! significant hazards consideration under the standards set out in

( 10CFR50.92(c) and, accordingly, a finding by NRC of no significant j hazards consideration is justified.

I V. ENVIRONMENTAL EVALOATION l

l TU E!ectric has evaluated the proposed changes and ha_s determined

(- that tne changes do not involve (i) a significant hazards-l consideration, (ii) a significant change in the types or significant

! increase in the amounts of any effluents that may be released of fsite, or (iii) a significant_ increase in indiviCual or cumulative-occupational radiation exposure. Accordin0 1y,_the proposed changes

. meet the eligibility criter. ion for categorical _ exclusion set _forth in

'10CFR51,22(c). Therefore,pursuantto-100FR51.22(b),an

l. environmenta~l assessment ~of the proposed changes is not required, l-l VI. RCFERENCES B 1) NUREG - 1399, " Technical Specifications, Comanche Peak Steam Electric Station, Unit 1", Mted April, 1990.

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