ML20126B509
| ML20126B509 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/25/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126B506 | List: |
| References | |
| NUDOCS 8003130317 | |
| Download: ML20126B509 (27) | |
Text
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UNITED STATES g
g NUCLEAR REGULATORY COMMISSION y p s-.
WASHINGTON, D. C. 20665 n.g\\Wg/
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. OPR-52 j
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority i
(the licensee) dated October 4, 1979, as supplemented by submittals dated January 15, 1980 and January 29, 1980 complies j
with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) tnat the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
8 0 03130 J/7 l
1.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2} of Facility License No. DPR-52 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 54, are hereby incor-porated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
&? jn &-h-Thomas A/ Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 25, 1980
)
ATTACHMENT TO LICENSE AMENDMEllT NO. 54 FACILITY OPERATING LICENSE N0. OPR-52 DOCKET N0. 50-260 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages:
11/12 147/148 221/222 97/98 149/150 253/254 111/112 157/158 255/256 145/146 181/182 277/278 The underlined pages are those being changed; marginal lines on these pages indicate the area being revised. Overleaf pages are provided for convenience.
1 1
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SAFETY LIMIT LI.MITING SAFETY SYSTDi SETTING 1.1 Fuel Cladding Integrity 2.1 Fuel Cladding Intearity 1.
Core spray a.nd LPCI i 378 in, actuation--reactor above ves9el low water level zero J.
HPCI and RCIC i 470 in, actuation--reactor sLove vessel low water level zero K.
Main steam isola-1470 in.
tion valve closure-- above vessel reactor low water eero level 11 1
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Amendment No.
54 i
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FIGUP1 DELETF.D l
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k TA3LE 4.2.5 (Cunt.noed)
S 3
Instrument Check I
B Calibration e
Functional Test _
Function 5
none once/3 months (1) z Isotnament Channel O
i' Reactor Low Pressure (FS-48-93 & 94) none once/oPersting cycle (4)
Core Spray Auto Sequencin5 Timers (Itormal Power) none once/ operating cycle (4)
Core Spray Auto Sequencing Timers (Diesel Power) none once/ operating cycle 1
(4)
I LPCI Auto Sequencing Timers (Wormal Power) none once/ operating cycle 0
(4)
LPCI Auto Sequencing Timers (Diesel Power) none once/ operating cycle (4)
EBRSW A3, Bl. C3, D1 Timers (Bormal Power) none once/ operating cycle (4)
RHRSW A3, E1, c3, D1 Timers (Diesel Power) mone once/ operating cycle (4)
ADS Timer
4 TABLE 4.2.B (Continued)
Function Functional Test Calibration Instrument Check fastrument Channel (1) once/3 months none RER Fump Discharge Pressure Instrtarent Channel (1) once/3 months none Core Spray Pump Discharge Fressure Core Spray 5perger to RFT d/p (1) once/3 months once/ day Trip System Bus Power Monitor once/ operating cycle N/A none Instrument channel Coedensate Storage Tank inw Level (1) once/3 months mone Instrument n annel Seppression Chamber High Level (1) once/3 months none Instrument Oiannel Reactor Nigh Water Level (1) once/3 months once/ day Instrument Channel RCIC 7tarbine Steam Line Righ Flow (1) ouce/3 months mene Instrumment Channel BCIC Steam Line Space Righ Temperature (1) once/3 months moee e
l 3.2 BAMs instrumentation which initiates a in addition to reactor protectioninstrumentation has been provided which reactor scram, protective initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator er-This set of speci-rors before they result in serious consequences.
fications providen the limiting conditions of operation for the primary initiation of the core cooling systecs, con-system isolation function, The objectives of systems.
trol rod block and standby gas treatmentto assure the effectiveness of the protec-the Speciftentions are (1) tive instrumentse tnn when required by preserving its capability toof such nyqtems even d tolerate a ningt,' ti11 ore of any enmponent vuch sygtems are out of s e rv ic e for maintensnre, periods when port ionn of trip settings required to assure adequate per-and (ii) to prescribe the When necessary, one channel may be made inoperable for brief i
required functional tests and calibrations.
f o rmanc e.
)
intervals to conduct instrumentation that initiate or control core Some of the cettings on thecnoting have colerances explicitly stated where the high and containmentvalues are both critical and may have a substantial effect on points of other instrumentation, where only the high or and low safety. The set end of the setting has a direct bearing on safety, are chosen at a
inadvertent actua-low level away f rom the normal operating range to prevent tion of the saf ety system involved and exposure to abnor=al attuations.
is initiated by protective instru-Actuation of primary containment valvesTable 3.2.A which senses the conditions for which mentation shown 1:
Such instrumentation must be available whenever pri-lation is required.
mary containment integrity is required.
The inst rumentat lan which init iates primary system isolation is connected la i dual hos ar r. uni.cment.
The low water level inatrumentation set to trip at 171.7" (538" above the top of tho active fuel closen isolatiun ealves in vessel acro) abnvc and Drywell and Suppression Chamber exhausta and drains the RHR Syutem.
The low (Gr<uip ? and 3 isolatinn valves).
Reactor Water ("te utup Linew reactor water level instrumentation that is set to stip when reactor vnter above the top of the active fuel level ip 109.7" ( 470" above vessel zero)idolation Valves and Main St eam, RCIC, and !!PCI closes the Main Steam 1.ine Details of valve grouping and require <1 Drain Valven (Croup 1 and 7).
These trip settings are closing times are given in Specification 3.7, core uncovery in the case of a break in th adequate to prevent line assuming the maximum closing time.
instrumentation that is set to trip when reactor The low reactor water level above vessel zero) above the top of the active water level la 109.7" 470" also initiate the RCIC and HPC1, l fuel (Table 3.2.5)
Amendment No. 54
4
{.
1 I
i 3.2 BASES i
and trips the recirculation pumpe. The low I
reactor water level inntrumentation that la set to trip when reactor water level is 17.7" (378" above vensel zero) above the top of the active fuel (Table 3.2.8) initiates the LPCI, Core Spray Pumps, contributes to ADS in t t ia t ion and 9 t'a rt s the diesel generatora. These trip setting levels were chosen td be high ennny,h to prevent spurioug actuation but I
low enough to initiate CSCS operattun uo that post accident cooling can i
be accomplished and the guidelinen of 10 CFR 100 will not bc violated.
For large breaks up to the complete c,ircumferential break of a 26-inch r
j
. recirculation line and with the trip setting given above, CSCS initiation to initiated in time to ment the above criterta.
l The high dryveli presaure instrumentatton in a dive rne alsnal to tne i
water level inntrumentation and in addition to initiating CSCS, it causes j
inolat(on nf Groupa 2 and 8 innlatton valves.
For the breako dincuened above, this instrumentation will initiate CSCS operation at about the I
same time as the low water level instrumentation; thun the resulta given a
4 above are applicable here also.
Venturis are provided in the main steam lines as a means of measuring stenm flow and alno limiting the lose of mans inventory from the vennel I
during a nteam line t reak accident. Thn primary function of the instru-
)
mentation is to detect a break in the main steam line.
For the worst i
f case accident, main steam line break outside the dryvell, a trip setting of 140% of rated steam flow in conjunction with the flow limitern and
)
3 main steam line valvo closure, limita the maan inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F and release of radioactivity to the environs is well below 10 CFR 100
^
guidelines. Reference Sectton 14.6.5 FSAR.
l-i trentnre monitortna inst rumontat ion in provlihd in the mntn otram line j
inonel en.le t or i led n in t b.. n r asenn.
Tr(pn are piovided on thin inulru-l mensatton nna when currede.'. cann" einnure of 190lnflun volven.
1h" l
net 4 ing of
.HH1'F foi the main atenm linc tunnet detnetor l9 low ennogh to l
detect lenkn nf the erder ni li gpm; thus, it in capab1'e of covering the entire spectrum of breaks.
For Lars,e breaks, the high occam flow instru-4 eentation is a backup to the temperature instrumentation, liigh rad ia t ion monitors in the main steam line tunnel have been proviurd to detect p.ross fuel fallute as in the control rod drop accident. With the established netting of 3 timen normal background, and main occam line isolntion valve clnoure, fienton product releone is limited cn that
}
10 CFR 100 guidelinen are not anceeded f or this occident. Reference An sl):v, i/t to a n om\\nal set, poLnt, of 13 x
}
norma knf u('1 p. 2 FS AR. kr,round,
=
Secti l 6
.6 Droviced a so.
ower c1c Pressure instrumentation is provided to close the main occam isolation valves in Run Mode when the main steam line pressure drope below 8 0 4
Peig.
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112 Amendment No.
54 s
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(I!GTClO CO?lDIT!C:10 M9OTD*TWI SUSVE!LLAFE 9EC3 OT'1 3.5.R Hesidunt Heat Removal Svetem 4.5.5 Residual Heat Removal Svata-(RHRS) (LPCI and Containment (RKRS) (LPC1 and Containment Cooling)
Cooling) 1.
The RilRS shall be operable:
1.
a.
Simulated Once/
Autocatic Operating (1) prior to a reactor Actuation Cycle startup from a Cold Test Condition; or (2) when there is irra-b.
Pump Opera-Once/
disted fuel in the bility
- wnth reactor vessel and when the reactor vessel pres-c.
Motor Opers-Once/
sure is greater than ted vnive
- non th operability atmonpheric, except en epecified in specifica-tions 3.5.3.2, through d.
Pu=p 71ov Rata Once/3 taontho 3.5.B.7 and 3.9.3.3.
e.
Test Check Valve Once/
2.
With the resetor vessel pres-operating sure lesa than 105 pair., the Cycle rtHRS may he removed from ser.
vice (except that two RHR pumps-containment coolin;; mode and Each LKI pu:np :huti calty,.y associated heat exchangers must
?,000 g7m spinn u '.ndi:a: ad remain operable) for a period system prec:u-a *' '95 ps:. Two not to exceed 2I. hours while LM! pumps in the, sr+3 loop shell being drained of suppression deliver 15,000 g;m again:t 2n charter quality water and indicated :ys.em pre sure :r 200 psi.;.
filled with primary ecolant quality water provided that 2.
An air test ont,edryvel,ang n
s during cooldoen two locps with t rus heade s and no::len cha.1 one pump per loop or one loop vich be constucted onee/5 years, r,
two pumps, and associated diesel water test may be perforned o, Renerators, in the core spray syste:
the torus header in lieu of t'.::
are operable, air test.
3.
If one RilR pump (LPCI mode) 3.
When it is determined that one RHR in innperable, the reactor pump (LPCI mode) is inoperable at a may remain in operation for a time when operability is required, period not to exceed 7 days the remaining RIIR pumps (LPCI mode) provided the rem ining RHR and active components in both access paths of the PJiRS (LPC1' mode) and a
a in th RH the CSS and the diesel generators (LPCI mode) and the CSS and shall be demonstrated to be opera-the diesel generators ressin ble immediately and daily there.
after.
145 Amendment No.
54
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146 4
t
yli t'lG CiclulT ;otis FOR OPERAT ION
_ SU RV L t LL A?iO. It hilu t x).Qi jg,5 _
J.5.B - Rea ldual llent F.r mo ve i System 4.5.B Residual Hast R emova l Sy9 t em (RHRS) (LPC1 and Containment (RRRS) (LPCI and Containment Cooling)
Coolin3) 4, If any 2 RER peups (LPCI =cde),
4 No additional surveillance becoe.e inoperable, tne reactor l required chall be placed in the cold i
Shutdo'an condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
If one RHR puen (contain-
- 5. When it in de ts ru taed tha t one ment coolinr. wode) o < a n-RFR pump (containment cooling nociated hent ex: hanger is nodo) or t.ssociated heat inoperable, the reac:or exchanger is inoperabic at a may remain in opera tion f or time when operability 10 re-a period not ta exceed 30 quired, the recaining RHR days provided the rec.11ninA pumps (containment coolin?, mode).
RHR pumps (containment the aosociated heat exchangers cooling mode) and asso-and diesel generators, and all ciated heat exchangers snd activo components in the accesa diesel generstnes and all paths of the RXRS (containment access paths of the RHR5 cooling node) shall be demon-(containment cooling node) strated to ba operable impediately are operable.
and waekly thereafter untti the inopersble RHR pump (containment cooling mode) and associated heat exchanger is raturned to normal s e rvic e.
6.
If two RHR pumps (containment 6 When it 14 determined that evo cooling mode) or associatad RHA pumps (containment coolins heet exchangers ara inopers-mode) or associated hast cechangers bie, the reactor may remain are inoperable at a time when in operatton f.r a period o?ers'atlity is required, the not to exceed 7 days pro-re=aining RMR pumps (contair. ment vided the remaininz F.HR pumps cooling code), the associsted (contsinment cooling mode) heat exchangers, and diesel and associated heat exchansers 32nerators, and all active com-and all accesa pacha of the ponents in the access paths of RHR5 (containment cooling mode) ths RNR$ (concal mant coolinz W
l Amendment No.
54 t
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SURVEILLANCE REQUIREMENTS j
L; T1HC CONDITIONS FOR OPERATION 4.5.3 Residual Heat Recovel Svstem
)
3.5.5 Residual llent Renoval Svatem (RKRS)_ (LPCI snd Centainment (RHRS) (LPCI and Containment Cooling)
Cooling) 4 mode) ehall be demonstrated l
are operable.
to be operable immediately and daily thereafter until l
at least threo RHR pumpa (containment cooling mode) and associated heat exchangers d
j are returned to normal service.
i.
7.
If two accean paths of the 7.
When it to determined that one RHRS (containment cooling or more access paths of the s
mode) for each phase of the RHRS (containment cooling code) are inoperable when accens in mode (drywell sprays. sup-pression chamber sprays, required, all active componen:o j
and suppression pool cooling) in the accces patho of the RHR3 j
are not operable, the unit (containment cooling nede) shall l
may remain in operation f or a bo demonstrated to be operab1:
period not to exceed 7 days irmediately and all active com-provided at least one path ponento in the cecess paths or each phase of the mode which are not backed by a cecend remains operabic.
operable accase path for tha aame phase of the mode (dryvell sprays, suppression chamber sprays and suppression pool cooling) j shall be demonotrated to ba opera-i 4
ble daily the.reafter until tha second path is returned to nor-mal service.
8.
If specifications 3.5.3.1 8.
No additienal eurvaillsace through 3.5.0.7 are not met, required.
on orderly shutdown shall be initiated and the reactor shall be shutdown and placed in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
9.
When the r1 actor vessel pres-9.
When the reactor vessal pressure sure is atmospheric and irra-is atmospheric, the ?.H2 ; umps diated fuel is in the reactor and valves that arn recuirsd to i
vessel at least one RHR loop be operable shall bo demonstratad with two pumps or two loops to be operable monthly.
with one pump per loop shall be operable. The pumpa'sa so-cisted diesel generators must i
also be operable.
l 10.
If the conditions of specifica-10.
No additional surveillance required.
tion 3.5.A.$ are met. LPCI and containment cooling are not 148 required.
i l
Amendment No. 54 fI
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LIMIT LNG CO?fDlT10NS FOR OPERATION SURVEILLMC REC"I/1Ji:NTa 3.5.B Residual Heat Removal System I
4.5.3 Aasidual ile a t Rem.a si - *
.n (RHRS) (LPCI and Containment (UG -)
(LPCI and Containment Cooling)
Cooltng) idja-11.
L' hen there is irradiated fuel 11.
The RtG punt s en*
6 in the reactor ano the reactor cent units hirh oupplv vessel pressure is greater than !
cross-connect caplatiity atmospheric, 2 FRR pumps and snail be dcmat.att sted t e b-associated heat exchangers and l
operable conth'v -ra a t.13 valves on an adjacent unit must l cross-connect cuptbility be operable and capable of is required.
supplying cross-connect capabil-icy except as specified in spect -
fication 3.5.B.lZ below.
(No t e : Because cross-connect capability is not a short t e rm requirement, a component is not considered inoperable if cross-connect capability can be re-stored to service within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)
9 i
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/ 2.,
If three RilR pumps or l
12.
- C n L:
!" d e :d ' **
associated heat e xc hange r s that three UU
,u.m ;-
s l
usociated.w.it
'xenine.' -
located on the unit cross-connection in the adjacent located cn rite uni; cro:..-
units are inoperable for c onnec t ton L. che 1dieren-any reason (including vnive units are innretraoio at 149 l
Amendment No.
54 1
te LIMITING CON 0lT10NS FOR OPERATION SURVEILLANCE REQUIREteli'.
a time when operability inoperability, pipe break, is required, the re-etc), the reactor may remain maining RHR pump and in operation for a period not associated heat exchanger to exceed 30 days provided on the uni t cross-connec-the remaining RHR pump and ti n and the associated associated diesel generator diesel generator shall De are operable.
demonstra ted to be oper-able irrediately and every 15 days thereaf ter until the inoperable pump and associated heat exchanger
- 13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation a
for a period not to exceed 10
- 13. No additional surveillance required.
days unless such capability is res to red.
- 14. All recirculation puinp discharge valves shall te tested for operability
- 14. All recirculation pump during any period of discharge valves shall reactor cold shutdown be operable prior to exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if reactor startup (or operability tests have i
closed if permitted not been performed elsewhere in these during the preceding speci fications).
31 days.
Q 150 Amendment No. 54
1.iMITD;G CO:iDITTu% FOR "I'ERATItW S!!RVEIU.:lCh W UiNifCiU o L.;,
3.5.F Reactor Coro Iso la t_ ion Coolinu 4.5.F Reacto r Cor e Tso tat i...
m 1
2.
If the RCICS is. Ino pe rab le,
2.
When 16 13 d+.tu rm it.;J aa t ti.-
the reactor may renain in RCIC3 ir la p..u t.
1:
operatlon for a period not chall be de.or...t.
tec to v.
i to exceed 7 days if the operable f o aed !. cl.
ILPCIS is operabic during such time.
l 3.
If specifications 3.5 F.1 or 3.5.F.2 are not, met, an orderly t
- urziown shall be initiated and the reactor shall be depressuri.wed to loss than 122 psiy within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Automatic U rressuri.n tton G,
& turntif vjirg.
i..>.t; Evsta (.'i0 5 ;
System (ADS) 1.
Four or ti.e s i s. valves of 1.
During e.u.h r,wa u ue i
the Autmai ic Darre: suri-the tol.lowim.
L. A hn zatlon Systen shall be performed on the
.'A operablo:
- ..u ! -
acti:a t Lnt te-(1) ptfor to a.tartup from a Cold Cendition, pe r f o rr. e ri pr'or
)
after cach
.to.
i.
or,
- i i-ago.
Manual s s.
(2) whenever there is irra-of the reU M
.<a; e-diated fuel in tSo reactor ccm a mi, i n /.. '.'
vessel an<l the reactor vessel p re.v ute is greater than 10.; pi,ig, cr:ept as rpecitied i..t 3.i.G..! and 3.5.c.] belew.
2.
'.s h c o u it o..
.i...;
e.,
t h u. t..a o f 1
2.
If thtee..I rk. t s AtC 'alvea
. v m.h.
n tu are known to l'.
i.ne..pabic o f the h m :
a.
.o i
automatic olur.
tho
.o be un m ;
in reactar r.ay rc.:ain in opera-da.ily there, tion for.. periad num to Speci t ic.it. io i t.
'..ti.
cxceeri 7 u s,t y. p r o v del the itPCT sy.;t ":T U ot r.l i c.
(Note that the pra oire relief f uoetloa of i.hese valve.s is as<.ur.d by seetien 3.6.D of th...e speelfice.tfon; a ni t'o t ' Ls r,pec i fien r ien o:...
pr!i<c to the AD., t unei 1..)
li mero than three of t.e.d x li.n valves are kno'-'
to be inrap-abic of cutomat:c oacration, an Imraediat e o rihir l, 3,h u t hn.at shall be initi.m a.
ft5 the 157 reactor la a ho' chui dom con-dition in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown conditton in the following 18 houra.
Amendment No. 54
i 1,tMITINC CON 11tTIONS FOR DPERATION
$URVEtt.t.ANCE REOUIRdE?TTS S
$.C,Autoe,atic Denressurizatten 4.5.C Automatic _Deeressurizacien System ( At)S[
System (ADS) 3.
'If more than two ADS valves are
- 3. % cn it is determined
>>.at known to be incapabic of auto-more than two ADS valves are matic operation, the reactor incapable of automatic opera-may, remain in operation for a tion, the RFCIS shall be period not to exceed 7 days shown to be operable i=:e-penvided the HPCI is diately and daily thereafter
- operabic, as long as 3.5.C.3 applies.
4.
If specifications 3.5.C.2 2
i and 3.5.C.3 cannot be met, an orderly shutdown vill be initiated and the reactor vessel pressure shall be reduced to 105 psis or less withir. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
H.
Maintenance of Tilled Discharte H.
Maintenance of Tilled Discharec Pipe Pipe Whenever the core spra,y systems.
The following surveillance require-LPCI, HPCI, or RCIC.are required ments shall be adhered to to assure to be operabic, the discharge that the dischstge pipinr, of the piping f rom the pump discharge core' spray systems, LPCI, HPCI, and of these systems to the last ECIC are filled:
i block valve shall be filled.
l 4
4 158 4
m
t.f w ri t tp Mvn :r ton'. Fon CPt:RArtott SURvr:LiA W RESUf W EV-3.6.C Coolant Leakava 4.6.C Coolant Leekaec 3.
If the condi:Lon l'n 1 or 2 above cannot be met, an orderly shutdosn sna;; be initiated and the reactor anall be snut.
D.
Safe v and Relief valves down in the Cold Concition within 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />.
1.,
At least one safety valve and approxima:ety one-hal,;,;;
D.
Saf e v and telt=' va tves relief velves shall be bencn-checked or re;;sced vi:h a 1.,
When more than one relief bench-che ked va;ve eacn opera.
valve or one or more safety ting cycle. A;; 13 y,19,, (;
valves are known to be safety and 11 relief) vill have failed, an orderly shutdown been checked or replaced u;:n aball be iniciacad and the the cocolecion of every second reactor depressurized to Cycle.
less than 105 psig within 2.
Once during each operating 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
cycle, eacn relief valve shall be manually opened unti; the-a.
couples downs:reas of :ne va;ve indica:e steam ia floving from the valve.
3.
The integrity of the rellet '
safety valve bellows shall be continuously conitored.
4.
At least one relief valve sh4;;
be disassembled and inspec:ed asch operating cycle.
I.
Jet Pum:s g,
J.: pum,,
1.
Whenever the reactor is in the 1.
Whenever there is rceircula:ict startup or run modes, all jet flow vi:5 the reactor in the pumps shall be operable. If startup or run modes vi:h be:h it is determined tha: a jet recircula:1on y4=cs running, pump is inoperable, of if two jet purs operability she.'1 be or more jet pum7 flow instru-checked daily by verifying taa:
eent failures occur and can-the following condi:icas do ne:
not be corre :ed within 12 occur simul:aneously:
hosts, an orderly shutdevn shall bc ini:14:ed anc tne a.
The eso recir:uis: ion tr::s i
resetor shall be shutJew?. in have a flov imoalanca et the Cold Condi: ion within 24 13: er scre when :ne ;u::s hours.
are c:ere:ad a: :ne so a speed.
181 Amendment No.
35, 46
2 i
t.,t M i T I Nr., UPiril i I ttas rna of t AAT 10't S U H V L I L LA'iC P A f qu i l'..'.M r,it y l
- 1. ti. L.j e i,,,M; 4.6. E Je t Pumpe 3 6.T Jt t Pug 73 et-P,1 cestc h b.
The indicated value of core j
flow rate varies fro., the i
1 value derived from loop finv neerurements by core than 101.
c.
The di f f u t.c r t o Icn.e r p le mun differential presaure r e..d -
ing t.,n an individuel j e t pump varies f rem the ucan of all jet punp dif/cror.-
d tial presource by more c h a r.
j 10Z.
4 2.
Whenever there is recirculat st flow with the reactor in the 1,
The reactor she.l.1 not be Startup or Run Modt-and ene c-operated with one recirculatica circulation punp is opera.!nt
{
loop out, of servJ ce for more with the equeltzer 'relv.- c)e ud, than 24 hour:. With the reactor the dif fuoer to lover plenu-operatity;, if one recircu.1ation differential ptessure shall r loop is out of service, the ched ed dully and the dif/oc:n-4 plant shal.1 be placed in a hot tini pruaure of an indivii v al shutdovn conditior. within jet pu sp in e luer shall not 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> uruess the loop is very from the mean of all jet sooner returned to serrice.
P"*P ('if f erential pressures in that lonp by sure than 107..
l 2,
Tol.lovi,ng one pucp ope:rr. tion s F.
Jet Punp Flow litematch the diacharge valve of the low speed psep r.ay not be opened 1.
Recirculation pump speeds eht11 l
un.icss the speed of the faster be checked and logged at least pua.p is less than 50% of its once per day.
l rated cpaed.
3.
Steady state operation with both recirculation pumps out of ser-vice for up to 12 hrs is per-mitted. During such interval restart of the recirculation pumps is oermitted, provided the i
loop discharge temperature is within 750F of the saturation i
temperature of the reactor vessel water as determined by dome pressure. The total elapsed time in natural circula-i tion and o'ie pump oreration must C.
Structural I n t e r, r i t y be no greater than i'4 nrs.
C.
Structural Integrity 1.
Table 4.6.A together vich sup-1.
'lhe httuctural inter,rity of the primary nyst.em shall be piementa ry not es, specifies the 182 Amendment No. 54
~..
fk, '
b re N, ;,
3.6/4.6 n AS r.1:
the core flew rate measured by the If they do difier hy 10 p:, cent or more, the pump diffuser differential pressure system must be checked against jet flow to core flow core flow rate derived from the measured values of loop If the difference between menaured and derived core flow rate correlation.
or more (with the derived value higher) diffuser measurements is 10 percent will be taken to define the location within the vessel of failed jet punp no r. a le (or risor) and the unit shut down for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation the affected drive pump will "run out" to a pump is also reduced; hence, substantially higher flow rate (approximately 115 percent to 120 percent for a ntnr.le nozzle failure). If th= two loops are balanced in flow at the the resistance characteristics cannot have changed. Any same pump speed, tmbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakar,e path past the core thus reducing the core flow rate. The reverse flow throur.h the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
J failure could also generate the coincident failure of A nnr.zle-rimer sy9 tem a )ct pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
- 3. 6, F/ 4. 6. F Jet Pump Flow Mismatch k.
31 r
1
$n b
6 f*#
221 gi:-
k@-
Amendment No. 54 a; '
h
3.6/4 6 BASES:
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its' rated speed provides assurance when going from one to two pump operation that excessive vibration of the jet pump risers will not occur.
3.6.G/d.6.G Structural Integ-lty The requirements for the reactor coolant systems inservice
-inspection program have been identified by evaluating the need for a sampling examination of areas of high stres:, and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
Only proven nondestructive testing techniques will be used.
More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip. These welds were selected in respect to their cistance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems. Selection was based on judgement from actual plant obsrevation of hanger and support locations and review of drawings.
Inspecticn of all these welds during each 10-year inspection interval will result in there additional examinations above the requirements of Section XI of ASME Code.
An augmented inservice surveillance program is required to determine whether any stress corrosiun has occurred in any stainless steel piping, stainless components, and highly stressed alloy steel such as hanger springs, as a result of environmental conditions 1
associated with the March 22, 1975 fire.
222
TABLE 3.7.A (Continued)
Number of Power Maximum Actica on Operated Valves Operating No rma ',
Initiating Group Valve Identification Inbeard Oa-boat-d Tire frec.)
Position-Signal Standby liquid control system check valves CV 63-526 & 525 1
1 NA C
Process Feedwater check valves 2
2 NA 0
Process CV-3-558, 572, 554. & 568 Control rod hydraulic return check valves CV-85-576 & 573 1
1 NA 0
Process RLIRS - LPCI to reactor check valves CV-74-54 & 68 2
NA C
Process v
't e
e
NOTES FOR TABLE 3.7,A Key: 0 = Open C = Closed SC = Stays Closed GC = Coes Closed Note: Isolation groupings are as follows:
Croup 1: The valveo in Group 1 are actuated by any one of the followins conditions:
1.
Reactor Vessel Low Water Level (470")
2.
Main Steamline High Radiation 3.
Main Steamline High Flow 4.
Main Steamline Space High Temperature 5.
Main Stesaline Lcw Pressure Group 2: ThevalvesinGroup2areactuatedbyanyofthefolkowing conditions:
1.
Reactor Vessel Low Water Level (538")
2.
High Drywell Pressure Croup 3's The valves in Group 3 are actuated by any of the following conditions:
l.
Reactor Low Water Level (538")
2.
Reactor Water Cleanup System High Temperature 3.
Reactor Water Cleanup System High Drain Temperature Group 4: The valvco in Group 4 are actuated by any of the following conditions:
- 1. _ HPCI Steamline Space High Temperature 2.
RPCI Stesaline High Flow 3.
IIPCI Steamline Low Pressure Group 5: The valves in Group 5 are actuated by any of the followins condition:
1.
RCIC Steanline Space High Temperature 2.
RCIC Stessline High Flow 3.
RCIC Stesaline Low Pressure Group 6: The valves in Croup 6 are actuated by any of the following conditions:
1, Reactor Vessel Low Water Level (538")
2, liigh Drywell Pressure 3.
Reactor Building Ventilation High Radiation 254 l.
Amendment No. 54 l
l
\\
t i
1 Group 7: The valves in Group 7 are automatically actuated by only the following condition:
l1. Reactor vessel lov vater level (470")
Group 8: The valves in Group 8 are autecatically actuated by only the following condition:
2.
High Dryvell pressure 255 Amendment No. 54
1 TABLE 3.7.8 4
TESTABLE PEMETRATIOWS WITH DCtJ3LZ 0-21NC SEALS i
X-1A Equipment Batch l
x-13 X-4 EM Head Access Eatch 4
X-6 CRD Removal Hatch X-35A T.I.F. Driven X-353 s
X-35C 4
X-350 f
X-35F X-357 x-35c i
i l
X-4 7 Fower operation Test i
X-200A supp. Chamber Access Katch i
X-2003 a
a 4
X-213A Suppression Chamber Drain e
d DW Flange-Top Head le sn.., u, 1.m a.. w., n Natch #2 i
's a
w e,
g) n n
n m
gg i
o a
n en g3 g
=
=
=
=
gy pg 256
m DASES isolated by reactor vensel low water level Crcup 1 - process sinen are (400") in order et allow for removal of decay heat subsequent to a scram, yet isolats in time for proper operation of the core standby The valves in group 1 are alao closed when process coeling systemo, inotrumentation detects excenoive main steam line flow, high radiation, low prconure, or main steam space high temperature.
2 - isolation velves are closed by reactor vessel low Water level Group The group 2 isolation signal also " iso-(33S") or high dryvell pressure.
system.
lacco" the reactor 1,uilding and starts the standby gas treatment is not desirabic to actuate the group 2 1 solation signal by a tran-Itsient or spurious oignal.
lines are normally in use and it is therefore not desirable Crouq_2 - process to cause spurious isolation due to high dryvell pressure resulting f rom non-safety related causes. To protect the reactor from a possible pipe isolation is provided by high temperature in the becak in the eyotein, to the cleanup system.
cleanup nystem area or high flow through the inlet since the vessel could potentially be drained through the cleanup
- Also, system. a low icvel isolation is provided.
Grouj_,4 and 5 - process lines are designed to remain operable and mitigetc isolation of other the conocquencen of nn accident which resulto in the The cignals which initiate isolation of Group 4 and 5 proccan lines.
proccos linen are therefore indicative of a condition which would render them inoperable.
6 - lines are connected to the primary containment but not directly Croup to the reactor veoscl. These valves are isolated on reactor low water Icvel ($3S"), high dryvell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate pt imary containment isolation.
7 - procene lines are closed only on reactor low veter level (470").
Croup These close on the same signal that initiates HPCIS and RClCS to ensure the valves are not open when HPCIS or RCICS action is required.
that Cro,ur_8 - line (traveling in-core probe) is isolated on high dryvell pres-sure, This is to assure that this line does not provide a leskage path condition.
when. containment pressure indicates a possible accident The maximum closure time for the automatic isolation valves of the primary und reactor vessel isolation control system have been sulected containment in conalderation of the design intent to prevent core uncoverin3 following and the need to contain relemoed pipe breaks outside the primary containment fission products following pipe breaks inside the primary containment.
In satisfying this deoign intent an additional margin has been included in specif y ing maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times.
277 i
knendment No.
54 l-
I B AS ES In or<ter to amoure that the douce that may recult from a steam line break do not exceed the 10 CTR 100 guidelinen, it in neccseary that no fuel rod perforation resulting tros the accident occur prior to closure of the main nt enn line inolation valves. Analysee indica te that fuel rod cladding perforations vould be avoided for main steam valve closure
]
times, including instrument delay, as long as 10.5 seconds.
Thenc'valven are highly reliable, have lov service requirement and most
,I are normally cloaed. The initiating.nensors and associated trip logic are also checked to demonstrate the espability for automatic toolation.
The' test intervel of once per operating cyclc for automatic initiation j
reaults in a failure probability of 1.1 x 10'7 that a line vill not iso-late.
More f r ee.uent testing for valve operability results in a greater
{
assurance that the valve will be operable when needed.
The maf u at eam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.
I i
The primary containment la penetrated by several small diameter instru-ment lines connected to the reactor coolant system.
Each instrument linc contains a 0.25 inch restricting orifice inside the primary containment and an exceos flow check valve outside the prinary containment.
i 3.7.L/4.7.E Control Room Emergency Ventilntion i
The control room emergency ventilation system is desinned to filter the con-trol room atmosphere for intake air and/or for recirculation during control room isolation conditions. The control room emergency ventilation system
-is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.
i i
lijgb efficiency partienlace absolute (liF.PA) fil:ers are installed before the char-l coni adsorbers to prevent clogr.ing of the lodin., adsntbors. The charcoal ad-norbers are installed to reduce the potential ittake of radiciodine to the con-trol room.
The in-place test results should indicate a svstem leak tightness of less than 1 percent bypass leakage for the caarcoal adsorbers and a KEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory I
carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions. If the ef ficiencies of the HEPA filters and charcoal adsorbers are as specified, 4
the resulting doses will be less than the allovable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 4
j CTR Part 50.
Operation of the fans significantly different from the design flow will change the removal efficiency of the dEPA filters and charecal ad-sorbers.
If the system is found to be inoperable, there is not immediate threat to the control room and reactor operation or refueling operation may continue for a i
limited period of time while repairs are being made.
If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or refueling operations are terminated.
278 4
._,