ML20058A303
| ML20058A303 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/16/1978 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058A306 | List: |
| References | |
| NUDOCS 7812050009 | |
| Download: ML20058A303 (29) | |
Text
_
i i
f UNITED STATES e'.
'4 NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D. C. 20666
~
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 44 License-No. DPR-33 1.
The Nuclear Regulatory Commission (the Commission) has found j
that:
A.
The applications for amendments by Tennessee Valley Authority (the licensee) dated August 2,1978 and August 11, 1978, comply 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 conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and sa fety. of the public.
(ii) that such activities will be conducted in compliance with the Commission's regulations; I
D.
The issuance of this amendment will not be inimical i
to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 4
10 CFR Part 51 of the Commission's regulations and all t
applicable requirements have been satisfied.
i 7 812 0 5 0 0FC>T 4
' 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-33 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 44, are hereby incorporated 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 2c' Tho.nas A.
ppolito, Chief 1
Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
I Changes to the Technical Specifications Date of Issuance:
November 16, 1978 t
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t ATTACHMENT TO LICENSE AMENDMENT NO. 44 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:
Remove the following pages and replace with identically numberef pages:
33/34 35/36 51/52 73/74 75/76 113/114 131/132 193/194 197/198 240/241 292/293 304/305 Revise Appendix B as follows:
Remove the following page and replace with identically numbered page:
41/42 Marginal lines indicate revised area. Overleaf pages are provided for convenience.
I
,__i.
TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCPM ) INSTRUMDTATICN REQUIRDtEVT ttin. No.
of Operable I n s r..
Cl.a eme l s Modes in Which Function Must Be Operable Per Trtp Shut-S y :. t. m (1)
Trip Function Trip t.evel Setting Startup/itot down Refuel (7)
Standby __ Run 1
Mode Switch in Shutdown X
X x
x 1.A 1
Hanual Scraa X
X x
X 1.A IPM (16) 3 liigh Flux
."- IM/{ g ndicated y,
X (22)
X (3) 1.A 3
Inoperative X
X (5) 1A u
AP P.*1 (16) 2 High Flux See Spec.. 1.A.1 2
lit e.h rlux
< 15% rated power x
1.A 4 2
Inoperative (13) x(pt) ggg7)
(g3) 1.4 a 2
1:ounseale i 3 Indicated on Scale X(21)
X(17)
X l'A 8 (11)
(11)
X(12)
IA8 2
gig, it actor Pressure < 1055 pet t
. X(10)
X X
1A 2
H!r.h Drvvell
- 2 pstg Pressure (14)
X(8) x(g) x 1.A 2
Reactor Lov Water g $38
- above vesoci teto 1.evel (14)
X X
X lA 2
Itt g!i Jater Level in Screa
< $0 Ce11ons D11 charge Tani X
~
X(2) x g
1.A 0
TABLE 3.1.A (Continued)
Min. rio.
of Operable Modes in Which Function f"a$h rEls Must Be Operable Per Trip Startup/ Hot Systen (1)
Trip Functicn Trip Level Setting Refuel (7)
Standby
- Run, Action (l) 4 Mein Stea:' Line Isolation s 10% Valve Closure X(3)(6)
X(3)(6)
X(6)
- 1. A or 1.C Valve Closure 2
Turbine Cont. Valve Fast Upon trip of the fast X(4)
X(4)
X(4)
- 1. A or 1.0 Ciosure acting solenoid valves 4
'a rt. i ne Stop Valve Closure 1 10% Valve Closure X(4)
X(4)
X(4) 1.A or 1.0 2
Turbine Centrol Valve -
> 550 psig X(4)
X(4)
X(4) 1.A or 1.0 Loss of Control Oil y
3rcssure 2
Tsrbine First Stage i 154 psig X(18)
X(18)
X(18)
(19)
'ressure Parmissive 2
Tarbine Condenser Low
> 23 In. Hg. Vacuum X(3)
X(3)
X 1.A or 1.C
a c uum 2
"ein Stean Line High 1 3X tiormal Full Power X(9)
X(9)
X(9) 1.A or 1.C 72diation (14)
Background (20) 9 9
V
t
.3 s...n 4
1.
There shall be two operable or tripped trip syntems for each function.
i If the miniacu number of operable inatrument channula per trip syatem cannot be cet'for both trip systems, the appropriata actions listed below chall be taken.
A.
Initiate insertion of operabia rods and complete insertion of all operable roda within four hours.
].
B.
Reduce power level to IRM range and place mode switch in the j
Startup/ Hot Standby position uithin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Redoc.e turbine load and close main steam line isolation valves within S houro.
D.
Reduce power to less than 30% of rated.
2.
Scram discharge volume high bypasa may be used in ahutdown or refuel to bypass c.crn.m discharge volume scram with control rod block.for reactor protection ayaten reset.
3.
Bypasoed if reactor preseure < 1055 paia and mode avitch no: in run, 4.
Bypassed when turbine firar stage pressure is less than 154 psis.
i IRM'o are bypa6. sed when APRM'n are onscale and the reactor mode evitch iu in the run position.
6.
The decian permits closure of any two lines without a ser.tm being initiated.
7.
When the reactor is ouberitical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:
i A.
Mode suitch in shutdovn I
6.
Manual scram C.
liigh flux IRM D.
Scram discharge volume high level E.
Not required to be operabic when ertsary contaire.:ent integrity is not required.
9.
Not required if all r.ain steanlines are isolated.
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- 10. Not requirsd to b2 Cporiblo when tho rocctor prGscurQ vcsoci head is not bolted to the vessel.
- 11. The APRM downscale trip function is only active when the
.reactot mode switch is in run.
- 12. The ATRH downscale trip is automatically bypassed when the IRH instrumenta tien is cperable and not high.
- 13. Less than 1u operable LPR.P s vill cause a trip system trip.
14 Channel shared by Reacter Protection System and Pririarf Cent ainment and Reactor vessel Isolation Control System.
A channel failure may be a channel failure in e ach sy s t em.
- 16. Cha nnel sha red by Reactor Protection Systen and Reactor Manua.1 Control Syst em (Rod Block Portion).
A channel failure may be a channel f ailure in each system.
- 17. Not required while perfor: ming icv pcver physics tests at atmcapheric pressure during er af ter refueling at power levela not to exceed 5 m2(t).
- 18. Operability is required when normal first. stage pressure is belou 30% (< 154 Asig).
- 19. Action
- 1. A or 1.D shall be taken only if the pe_ issive f ails in such a manner to prevent the affeeted RPS logic freu perf e-ming its intended function.
Otherwise, no acticn is required.
- 20. An alarn setting of 1.5 times nor :a: back1raund at rated power shall be established to alert the cperater to ebror::al radiation levels in primary coolant.
- 21. The APRM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the Source Range Monitors are connected to give a non-coincidence, High Flux scram, at < 5 x 105 The SRM's shall cps.
~
be operable per Specification 3.10.B.1.
The removal of eight (8) shorting links is required to provide non-coincidence high-flux scram protection from the Source Range Monitors.
- 22. The three required IRM's per trip channel is not required in the Shutdown or Refuel Modes if at least four IRM's (one in each core quadrant) are connected to give a non-coincidence High Flux scram.
The removal of four (4) shorting links is required to provide non-coincidence high-flux scram protection from the IRMs.
36
1,1MITINC CON'DITICNS FOR OPER.ATION 511R7!!LI.AMCE RIQtJilDGMTS 3.2.3 Core and Containment coolint 4.2.3 Core and Contain.sent Coelint Systems - Initiation & Control Svetese - Insttation 6 Control are requirnd to be operable shall be considered operabis if thay are within the required surveil-lance testing frequency and there la ao raason to evepect that they are isoperable.
C.
Centrol tod Block Actuation C.
Contrel Red Stock Actuation I
The limiting conditions of Ia.trumentation shall be f unc t ion-operatico for the instrumen-ally tested. calibrated and checked tation that initiates control as indicated in Table 4.2.C.
rod block are given in Table 3.2.C.
Systen lexic shall be functionally tested se indicated in Table 4.2.C.
DELE"E
!;cv covered by note 7.c.
3, Off-Gas Post Treatrent 1 solation Off-Gas Post Treatment Isolathn D.
F me t ion Functions
- 1. Off Cas Post Treatment Mcniters 1, Off-Ca3 Post Tr ea t= en t Moniter19; Systen (a) Encept as optcified in (b)
Instrutentation shall ha func-belev. both off-gas tionally tested. calibrated and post treateent radiation checked as indicated in Table eeniters shall be eperable 4.2.D.
during reactor operation.
The isolaticn funct ton
$7e tem logic shall be function-trip settings for the ally tested se indicated in monitors shall be set.s t Tabla 4.2.D.
a value not to exce.d the equivalent of the stack release 11=it specified in, specificatien 3.8.3.1.
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LIMITINC CottDI?!0NS FCR OP MATTCM StntV17L1ANCE Rf7JIIDCDfTS 3.2.D Of f-Cas Post Treatment Isolation b.2.D Off-Cas Post Treatmcet Isolation Functions Function (b)" From and after the date that one of the evo of f-gas post treatment radiation penitors is made or found to be inoperable, continued reactor power operatien is permissible durLas the next seven days. previded that the inoperable monitor is tripped in the downscela positten. One radiation sweitor may be out of service for four hours for functional test and/
or calibration without the monitor being in a downeeslo trippsd condition.
)
(c) L'pon the loss of both of f-gas post treatment radia-tion monitors, initiate an orderly shutdcun and shut the main s te a: isolation valves or the of f-gas isolation valve within firyvellLa.beIection 7.
Dr-vell Lea's Detset tes E.
The 11titing conditiens of opera-Instrur.entirion ahz11 ha calibrated tien for the instruzantation that and checisd as Ladiccted Ln Tabla monitors drywell Isai detection 4.2.3.
are given in Tabis 3.2.Z.
F.
Survatiltnes In s t ruean t a t f on 7.
Burvalliance Int erw--nLEt'gi g
The limiting cenditiens for the Instrusentctica shall be calibr.stsd ins trustntation tha t pr ovid e s and chec' sed so indicated in Tabla surveillance intorr.ation readeuta 4.2.7.
)
are given in Table J..F.
C.
Centrol lees teolatten 0.
Centrol Meen f eetatte9 The lindting conditiena for Instrusantation shall be calibtstad ins t ruacnta tion tha t isolates and chacked as indicated is Table the control rene and initiatas 4.2.0.
the central roca asargency pressuritatics systec4 are given in Tabis 3.2.C.
53
8.
This function is bypassed when the mode avitch is placed la Run.
9.
This function is only active when the modo evitch is in Run.
This function is automatically bypassed when the IP.M instrumentation is operable and not high.
- 10. The inoperative trips are produced by the following functions:
a.
SAM and IRH (1) Local " opera t e-calibrat e" switch not in operate.
(2) Power supply voltage low.
(3) Circuit boards not in circuit.
b.
APLM 5
(1) Local " operate-calibrate" avitch not in opersta.
(2) Less than 14 L7RM inputs.
(3) Circuit boarda not in circuit, c.
inn (1) Local " operate-calibrate" rvitch not in operate.
(2) Circuit boards not in circuit.
(3) R5M fails to null.
(4) Lees than required number of LFt4 inputs for red selected.
11.
Detector traverse is adjusted to 114 1 2 inches, placing the detector lower position 24 inches below the lower core plate.
75
g TABLE 3.2.D OFT-CAS POST TREAT C;T ISOLATION INSTRIPT;TTATION hiin.No.
Operabic (1)
Function Trip I.evel Setting Action (2)
Remark.s 2
Of f-Cas Post Trestrent Note 3 A or B 1.
2 upscales, or 1 downscal-and I upscale, or 2 devn-Honitor scales will isolate off-gas line.
3 1.
One trip systes with auto 1
Of f-Cas Post Treat =ent Note 3 transfer to another source Isolation 5
NOT?.S:
1.
Whenever the minimu:n nu ber operable cannot be met, the indicated action shall be taken.
2.
Action A.
Refer to Section 3.2.D l.b B.
Refer to Section 3.2.D.l.c 3.
Trip setting to correspond to Specification 3.2.D.1.s O
m O
w-
.a
a 3.2 RA$rs The HPCI hir.h f;ow and temperature instrueentation art provided to detect a break in the HPCI steam piping. Tripp1ng of tbls lastrupentation re-sults in actuation of HPC1 1 solation valves. Tripping logic for the h1Rh flov is a 1 out of 2 logic, and all sensors are resulted to be operable.
Itigh temperature In the vicinity of the HPCI equipnent is sensed by 4 sets of 4 biretallic temperature switches. The 16 teaperature switches are arranRed in 2 trip systems with 8 temperature switches in each trip system.
The HPC1 trip settings of 90 ps! for hir,h flow er.d 20C*T for high tem-perature are such that core uncovery is prevented and fission product release la within 11 pits.
The RCIC h!Eh riow and temperature instrumentation are arra'nged the samt as that for the HPCI. The trip setting of 450" 18 0 for high flow and 7
200*P f or temperature are based on the same criterie as the HPCI.
liigh tenperature at the Reactor Cleanuo Syst,ee floor drain could indicate a break in the cleanup system.
When high te=perature occurs, the cleanup system is isolated.
The instrumentatinn whic'h initiates CSCS action is arranced in a dual bus systrn. As fer other vital instrumentation arranged in thls fashion, the 5pecification preserves the effectiveness of the systen even during periods when natntenance or testing is beinr. perforned. An exception to this is when locic functional testinr, is being performed.
The control red block f unctlons are provided to prevent excessive control rod withdrawal so that !!CPR does not decrea:* to 1.06. The trip lotic for this function is I cut of n:
e.g.. any trip on one of six APRM's, eight IRM's. or four SRM's will result in a rod block.
The ntntmus Instr ment channel require.nents assure sufficient instrumenta-tion to assure the cinP.le failure criteria is cet. Two REM channels are pre-vided and one of these may be bypassed from the console, for maintenance and/or testing, provided that this out of service cendition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day peried. This time period is only 3% of the operating time in a month end does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The APRH rod block function is flow biased and prevents a slanificant reduc-tion in MCPR, especially during eperation at reduced flow. The APAN pro-vides gross core protection.
i.e.. linics the gross core power increase f rom withdrawal of cent rol rods in the nor. mal wither.nal sequence.
The trips are set as that McPR is maintained greater than 1.06.
i The RBM rod block
- function provides local protection of the core; i.e.,
the prevention of critical power in a local region of the core, for a single rod withdrawal error f rom a limiting control rod pattern.
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3.)
sA%V%
If the IRM channels are in the worst condition of alloued byp49n, the sealing arrangement is such that for unbypassed IRM channels, a rod bintk signal is generated before the detected neutrons flux has increased by more than a factor of 10.
A dovnacale indication is an indication the instrun ne has failed or the instrument is nnt sensitive enough.
In either case tr.e inet:tment will not respond to changes in control rod cotion and thus, contrel rod totion is prevented.
The refueling interlocks also operate one logic channel, and are requirt.d for safety only when the code switch is in the refueling po.cition.
For ef f ect ive energency core ecoling f or ans11 pipe breaks, the HFCI ryr. ca must function since reactor pressure does not dec rease rapt.i enough to allow either core spray or LpCI to operate in eine. The aute.wtic pre:oare relief function is provided as a backup to the lipci in the (Nent.the ll?J!
does not operate. The arrangement of the tripping contacts is euch r.-
,')
t, provide this function when necessary and mininize spurious oper.stion. The trip settings given in the specification are adequate to ar.n:rc the obcvc criterin are net.
The specification preserves the effective.ac.c of the ayaten during periods of naintenance testing, or calibration, and alco ntnimizes the risk of inadvertent operation; i.e., only one inntrer. cat channel out of service.
Two poet treat =ent off-gas radiation monitors are provided and, then their trip point is reached, cause an isolation of the off-gas linn. Ir.olation is initiated when both instrumer ts reach their high trip point or one has an upscale trip and the other a downscale trip or both have a dottnscale trip.
Both instrunents are required for trip but the instruments are set so that any instrueents are act so that the instantaneous sts:k release rnte limit given in Specification 3.8 is not cxceeded.
Tour rad Lat ion nonit urn are proviilot for each." unit shich initicte Primary Containment isolation (Croup 6 isolation valves) Reactor Building isolation and operation of the Standby Cas Treatnent System. These inctrument charmels monitor the radiation in the Reactor zone ventilation exhaust due c cnd in the Refueling Zone.
Trip setting of 100 mr/hr for the nonttors in the Refueling Zcne are ba:cd upon initiating normal ventilation isolation and SOTS operstien so that none of the activity released during the refueling accident leaves the Reactor Buildinr, via the normal ventilation path but rather all ths activity is processed by the SCTS.
Flov integrators and sunp fill rate and punp out rate tieers f.rt used to deter =ine leaka g e in the dryvell. A systen whereby the tics interval to fill a known volune vill be utilized to provide a backup. An air sanplin; syotra is alco provided to detect lesk:r.e inc id: the prinary centair. tnt (See Table 3.2.E).
114
3/4.3 ff,9 aloce proitde the operator vish a visual indic:tioi of neu-tren Irsel.
The conweic,e ca u: re :ttv!cy scrider.ts art functionn of the ir.itial neutron f1v..
'the r.'qu iremer.c of at 1 cast 3 count i 1. a r accond aasures that.i.f trer.: lent,
should it n e c u r, ti.* < t n s a t ur.7 hove the liit t! al value of 10" o f r n t ed pwn r i.. ?d I i t!.: en$ly. af tratisients fruo cold c.nditions.
One osarabia !>It thinnel vould he adeluste to raunit or the approich to c r i t !:ali ty t.st as r.cso'.encou:s pat te rna of sc at t e r ed cont rol rod v.8 thdraval. A :sini.:c, of tvo opersble Srd's are provided as an eddeo conaervatin.
5.
The Red Block Monitor (RBMi ia de si;:ned t o autoestically preven: fuci de ar: in tne event of e roneous rod vit5draval from loestio-a og hith cover den:*1tv dutt7t hi,th power Icvel operacion.
Two REM channels.are providad and one of these may i
be bypassed fron the console for maintenance and/or testing.
Autematic rod withdrawal blocks from one of the channels vill block erroenous rod withdrawal soon enough to prevent fuel damare. The specified restrictions with one channel out of service censervatively assure that fuel damage vill net. occur due to rod withdrawal errors when this condition exists.
A liniting centrol red pattern is a pattern which results in the core being on a thtreal hydraulic linit, (ie, MCPE niven by figure 3.5.3 or LHCR of 18.5 for 7x7 er 13.'
for 8x8)
Durinn use of such patterns, it is judgec that testing of the R B!! system prior to with-drawal of such rods to assure i t r. operability vill assure that inproper withdrawal does not occur.
It is nernally the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated reds either when the patterns are initially established or as thov develop due to the occurrence of inoperable centrol rods in other than liniting patterns.
Other personnel qualified to per-form these functions may be designated by the plant superintendent to perform these functions.
Scram Insertion Times The contrel rod system is designated to bring the reacter suberitical at the r.i t e fast enough to prevent fuel daeane:
le, to prevent the ?fCTR f r o :: beceming less than 1.06.
The lietting po er transient is r,1ven in Reference 1.
Analysis of this trancient shows that tne n e p r. t i v e reactivity rates resulting from the scram with the averane response of all the drives as niven in the above specification provide the required pretection, and MCPR remaine greater than 1.06.
On an ear): TM. c e ri o degradation of erntrol rod scrac pe t f e re,a n c e occured dur!r.g plant startur and was determined e..
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a 1
131
L.
3.3/4.3 BAsti:
part iculate material (probably construction debris) p.'ui,ging an internal control rod drive filter. The design of the preatat control rod drive (Hodel 7RDB1443) is grossly inproved by the relocation of the filter to a lo stion out of the scrac drive path; i.e., it ton no longer interfere with acrca perfotsance, even if completely biccked.
The degraded performance of the original drive (CRD7 P.D D14 /. A) under dirty operating ronditions and the insensitivity of the rede=1gned drive (CRD 7 RDB144 5) has been demonstr.ited by a sertfe of engineering tents under cimulated reactor opc r. ting conditions. The successf ul perfornance of the ntw driv.. under actus1 operating conditions has also been demonttrated by cona:stently gooit in-service test renults for pinato unf..e. the new Irive and nay be inferred from ple,te using t!.o oldar redel driv with a modified tlarger screen size) intciv.a1 filtcr which
~5 in 1 ee prune to plugging. Data han been docursnted by r uevtil-lanca reports in various operating plasts. There include
)
Oyster Creek, Monticello, Dresden 2 an! Dreadtr. 3.
Appr. ::1ec c ely 3000 drive teste have been recorded to date.
Following identification of the " plugged filtet" problen, ver y frequent ocram testa were nececonry to ensure proper perferrarce.
However, the note frequent scram teste are nov ccnniderci tetrlly unnecessary and tnuise for the following reasona:
1.
Erratic scraes performance has been identific! as due to ar obstructed drive filter in type "A" driven. The drivar is.
BFNP are of the new "B" type design whose r. crees perictv.cnc e to unaffected by filter condition.
2.
The dirt load to primarily re1 caped during startup of the reactor when the reactor and its ayatens r.re first subjected to fleve and presn' ire and therr.a1 stressen. Special atten-T tion and ceae.uren e.re now being taken to osnure cleaner
)
systems. Resccord with drives identical or similt.r (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram serfortance. Thin preoperational and startup testing ic aufficient to detect anceelous drive perf orvance.
3.
he 72-hour outage limit which initiated the start of the iroquent serem testing is arbitrary, havin, no logical bas.o other than quantifying a " major outage" uhtch cight reasons-bly be caused by an event so severe se to poccibly of'tet drive performance. This requirement le unwise beca se it provides an incentive for shortcut actions to haste.. returnt i4 "en line" to avoid the additiont.1 testing due a 72-h:ur outa52-132
~
TAMI_16 W ggI; 1. p.gg g C
SHOCr SUPPRESSORS ( wimntne) w Snubbers Sn tber: in Htch Insteessible Snubbers F.n31ntion irca Daring Snuhbers Especiolly tha tng Normal Acces:1ble De 5,i,% r !?o.
- Svue, Gevntion C h u*. '. v m
- Dirricult to Pemove Orcration
_ Nor a l Orc ra Dlfi upper Dim 598 X
P16 It> var DR9 598 X
7ll PfUi' 555 X
920 upper Plf9
$!a9 X
P21 - cost Dia 5 72
- 1 P21 - vest pita ST2 X
P22 Dfiit ST)
X P2 '.
mrs 580 X
R25 Fim 579 X
i:26 DRD
$75 X
C fs ) inside Tsim
$55 X
r'al o2tald-RUR 555 X
P29 Dft2 hend spray 636 X
Fey Rim head Opray 636 X
TMLI 3.5 -H tJtGT 1 - peg 5 9t0CK Trrprvr r5 UN"MFDR)
G.
s Snubbers Snubbers in litch Insceessible Sr.utbers Radiation Area luricF.
Snu5bers Especially Du-Ina,!!o mni Acces:1ble Durt Snubber 16.
System Elevation Shutdo n
- Difficult to Penove
_ Operation For-21 ora att R1 Cora sprny Grf x
11 2 Core sprcy 6M X
116 - north Core r. pray Shis I
!!5 - south C yc sprny Sf 34 X
R3 Core spray 609 X
n9 Core spray 609 I
B13 - north Cose spray Shh X
R13 - south Core spray Shh X
D19 Stnnoby liquid 624 X
coitrol F21 Standby liquid 62h X
control F3 - north itPCI 542 X
R3 - south IIPCI SI 2 4
I FG IIPCI 563 X
X 99 firCI S47 X
B11 IINI 532 M7 IfrCI 532 X
147 IITCI 532 X
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e
l.
1ABLE 3.6.H UNIT 1 p,g,g g
540CK SUPPRF550RS,.(SNURBERS)
Snubbers Snubbers in High inaccessible Scubbers Radiation Area During Sr.ubbers [$pecially During Nomal Accessible During Snubber No.
System Elevation Shutdown *_
Difficult to Remove Operation
_Nomal Operation 55Z-4A PSC (ring header) 525 x
552-5A PSC (ring header) 525 x
55x-6A PSC (ring header) 525 3
55x-7A PSC (ring header) 525 i
55Z-8A PSC (ring header) 525 x
R2A' Fi.re Protection 601 x
R3A Fire Protection 601 1
R4 Fire Protection 601 x
~ R42 EECW 605 x
I 551-A Recirculation 556 x-551-8 Recirculation 556 x
552-A Recirculation 558 x
552-8 Recirculation 558 x
h 4
s g81.E 3.6.H UNIT I - page 9 SHOCK SUPPRESSORS (SNUBBERS)
Snubbers Snubbers in High Inaccessible Snubbers Radiation Area During Snubbers Especially During Normal Accessible During Snubber No.
System Elevation Shutdown
- D_lfficult to Renove Opera tion Nomal Operation 0
SS3-A(295 )
Recirculation 564 X
SS3-A(3350)
Recirculation 564 X
0 SS3-B(115 )
Recirculation 564 X
0
'553-B(154 )
. Recirculation 564 X
S$4-A Recirculation 570 X
554-B Recirculation 570 X
SS5-A(2620)
Recirculation 581 X
0 555-B(325 )
Recirculation 581 X
S$5-B(350)
Recirculation 581 X.
0 SSS-B(98 )
Recirculation 581 X
SS6-A Recirculation 568 X
S56-B Recircula tion 568 X
557 Recirculation.
564 X
558
' Recirculation-564 X
- Modificaticns to this Table due to changes in high radiation areas should be submitted to the NRC as part
~
~
of the next license 'aciendment.
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4 1,lMITINP. CONN 1T IO'i$ FOR O P C3/.T I O f f SURVF.II.L A.NCF REQU19. EMF.NTS 3.7.C Scenndary Containment 4.7.C Secondary containment 1.
Secondary containment inte-1.
Secondary containment surveil-grity shall be maintained in lance shall be performed as-the reactor zonc.st all times indicated below:
except as apccified in 3.7.C.2.
a.
A preoperational secondary containment capability test shall be conducted by iso-latinr, the recctor building and placing two standby gas treatment systen filter trains in operation. Sech test shall demenstrate the 4
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240 e
-r t
e 3.7.C Seco.dsry Containeenc 4.7.C Secon br,3 containnent capaoility to maintain 1/4 inch of water vacua, uncer calm wind ( < 5 mph) condi-tions with a sys te?. inlaakage rate of not more than 12,000 efs.
b.
Secondary containment capa-bility to maintain 1/4 inch :
vater vacuun under caln wing
( < 5 mph) conditions with a system inleakag.
rate of no t mo r e t ha n 12.000 c f n.
shall be demons t rated a t each refueling outage priar to refueling.
2.
If reactor z ne accondary con-2.
After a secondary containment tainment integrity cannot be violation is determined the maintained the follcwing con-standby gas treat =ent system ditions shall be met:
will be operated immediately a f t'e r the affected tones are
)
isolated from the remainder of a.
The reactor aball be F-ade the secondary containment to suberitical and Specifica-con f i rm its ability to esin-tion 3.3.A shall be e.et.
tain the remainder of the secondary contatament at 1/'-
b.
The reactor shall be ceoled inch of water negative pressere dovn below 212*T and the under calm vind cenditions, reactor coolant system vented.
c.
Fuel movement shall not be permitted in the reac-tor tone, d.
Pri=a y contair. ent integrity, r.a i nt a i ne d.
sl 3.
Secondary centsinnent integrity shall be maintilned in the re-l fueling zone, exce.,t as spect-
!!ad in 3.7.C.4 I
241 r
SURVElf.l.ANCE REQUIMCMENTS tJ. MITING CONDITIONS FOR OPERATION 3.9 AUKILI ARY ELECTRICAL SYSTDi 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability
.Appifcability Applies to the auxiliary elec-Applies to the periodic testin3 requirements of the s.uxiliary trical power system.
electrical systems.
Objective Objective f
To assure an adequate supply of Verify the operability of the electrical power for operation of auxiliary electrical cystem.
those systems required for safety.
Specification Specification A.
Auxiliary Electrical Equipment A.
Auxiliary Electrical Equipeent 1.
Diesel Generators A reactor shall not be started up (made critical) from the a.
Each diesel r,enerator cold condition unless four shall be manually started units I and 2 diesel generators and loaded once each raonth are operable, both 161-kV trans-to der.onstrate opercrional mission lines, two coenon sta-readinces. The tcct shall tion service transformers and continue for at lea t a one cooling tower transf omer one-hour period at 75% of are operable, and the require-rated load or greater.
ments of 3.9.A.4 through 3.9.A.7 are met.
During the monthly r,ene-rator test the diecci A reactor shall not be started generator starting. air up (cade critical) from the compressor shall be Hot Standby Condition unless checked for operation and
~
all of the following condi-its ability to recharge tions are satisfied:
air receivers. The opere-1.
At least one off-site 161-kV il trannf r pump shall transmissicn line and its be denonstrated, and the diesel starting tire to available and capable of reach rated.voltar.e and autocatically supplying speed shs M M lo w d.
auxiliary pcver to the shutdovn boards, b.
Once per oper.itin.; cy.
a test will he conducted 2.
Three units 1 and 2 diesel to demonstrate the ever-generators shall be operable.
gency diesel generators vill start and accept 3
An additional source of emergency load -vithin pover consisting of one of the tcllov1ng:
A second 161-kV trans-a.
mission line and its 292
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St!fVtitt UCE #fQUl8 M_'LTS LTMttthC CCCITIONS fos OPrutt0N 4.9.A Austlierr Flectrieel Equipment _
1.9.A Aulliarv tiectrical Equipme_nt_
the spectfled ttee sequence.
< ornon t r..rwI ornic r anil c.
Once a month the guantity L t".11 n.
t Se r t rans f.,rn.. r of diesel fuel oestlebte capable e t' e.upplying power shall be logged.
t i, t h.- si.ot d.iwn boa rds.
d.
tech diesel generator shall be given an ennual insM c-t.
A fi ut t n o;.e r b ' e units II'"
I"
'#*"#' " I U' 1 ar1 2 dtasel generator.
tos t ruc t ic,ns 1,e s ed on the I.. Puses and Boa-ds hallable tions.
e.
Once a essnth a sample of a.
Start buses LA and 18 e r e enereited diesel fael shall be checked for quellt7 The 9uell'1 d.
ire nt
I an'. 2 '*-hV ehell be within the accepte-stutdon tuards are ble limit
- epecified in energstei Table 1 of the latest ""igion e
to Am DOM
- logged, w
c.
The i.8 0.. shs' J,,vn board s associete: -ith tra unat 2.
D.C. Povet %.stt's - Unit Pattertes (250-vott) Diesel Generator are enerst:ed letteries (I M toit! and Shut down neerd Battertes (250 Vol:)
d.
Undervc!toer rele>s era r aM e r s '. a r t O'
I byges 1A and 18 and babV snutdevn teerds. A. B. C.
graetty and the valtale nf the ptiot call. and tespeta-
- n d I' ture of en adjacent cell and everall battesy voltage chall 3.
De 2 3 0-V ol t ur!t a r' d eSotdown be sesauteJ and legted.
boord t:s t t e r ie s anJ a battery charger for eacn battery and b.
Every t hr e e c.cnt hs t he me s-eseociated battery boards are surements ell Se "arte of
- opetable, yelgege of each cel*. to searest 0.1 eoIt. eret tlic 4.
Leste Systems gravtty of each cell. and te nerature of every fifth e.
Comnes accident stgral cell. Diese measure = ente logic speten le operable, shall be logged.
b.
440-7 load smedJtna logic A battery rated dis:harge c.
ereten t e ope r elle.
(capsett *) test shall be perfermet
'n' t"o V0l*84**
g g,, uA odout c u r r e.it 7
There s%stl be a auntowe of 101.300 gallone of diesel fuel
.,eature-omts shall be Icg(ed la the etsadt; diesel genere*
1s nct to exceed ter fvel tenke.
Ele tunt h s.
293 j
3.10.A Refueling Interlocks 4.10.A Refuelint, Interlocka refueling interlocks shall be operabis.
b.
A sufficient number of control rode shall be operable so that the core can be made sub-critical with the strongest operable cen-trol rod fully with-drawn and all other operable control rods fully inserted, or all dircetional control valves f or rer.aining control roda shall be dio.rned electrically and sufficient margin to criticality shall be demonstrated, i
c.
If naintenanec is to be l
perforned on two control rod drives they must be l
separated by nore than l
two control cello in any direction.
d.
An appropriate nunber of SRH's are available as defined in specifi-cation 3.10 A.
6.
Any number of control rods 3.
With the mode selection switch in may be withdrawn or removed the refuel or shutdown mode, no from the reactor core pro-control rod may be withdrawn until viding the following condi-two licensed operators have confirmed tiono are satiefied:
.that either all fuel has been removed from around that rod or that all s.
The reactor mode switch control rods in immediately adjacer.t is locked in the "re-cells have been fully inserted and fuel" position. The refueling interleck electrically disarmed, l
which prevents rare tF.an i
eno cor. trol rod from 3 04 6
e, t.lMIT tun Count Tioens rot OMRAf tois StTxvt11.LutCI arctl!REMIters 3.lO.A a.efuellan Interleeks 4.10.A Refuelinn Interieeks being withdrawn may be f
~
bypassed on a withdrawn y
control red after the fuel asseemlies in th cell containina (sen-trelled by) that ese-trol red have Wem re-ansved f rom ths ro as ter core. All a,ther re-fueltag interlocka shall be op4rable.
B.
Core $fenitoring 3.
Core Monitoring Frior to making any siteratione 1.
During core alterations, except
.}
to the cora the SIM's sh.all be as in 3.10.B.2, two STd!'s shall functionally tasted and checked be operable, in or adjacent to any I"***""'M""'
Th * " "
quadrant where fuel or control after, while required to be rods are being moved. For an 57.4 1 D' to be considered operable, the I'*'""***
followin;; shall be satisfied:
a.
The SFJf shall be inserted to the normal operating level.
(Use of special moveable, dunkire type detectors du. ring initial fuel loading and major core alterations in place of nornal detectors is per-missible as long as the detector s
is connected to the nor=al SRM circuit.)
)
b.
The SFJi shall have a minimum of 3 cps with all rods fully inserted in the core, if one ot' note fuel assemblies are in the core.
- 2. ' During a cceplete core remeval, I
the SFJi's shall have an initial, I
minim"un count rate of 3 cps prior to fuel removal, with sl1 rods t
fully inserted and rendered electrically inoperable. The count rate vill diminish during fuel removal. Individual control rods outside the periphery of the then existing niel matrix may be electrically armed and moved for maintenance after all fuel in the cell containing (c : r t roll e.i if hit :-.tr:1 i e.
1.n c bw an..-
t i.e reactor core.
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