ML18033B769
| ML18033B769 | |
| Person / Time | |
|---|---|
| Issue date: | 02/14/1991 |
| From: | Edgeworth P NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Lofy R PARAMETER, INC. |
| Shared Package | |
| ML18033B770 | List: |
| References | |
| CON-FIN-L-1343, CON-NRC-03-87-027, CON-NRC-3-87-27 NUDOCS 9102280104 | |
| Download: ML18033B769 (52) | |
Text
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Docke. Nos.
50-25 5 - 60 and 50-6 O.
~ UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 February 6, 1981 Mr. Hugh G. Parr is
'Manager of Power Tennessee Valley Authority 500 Chestnut"Street, Tower.II Chattanooga,.
Tennessee 37401 Dear Mr. Par.r is'.
The Commission has issued the enclosed Amendment Nos.
66, 62 and 38 to Facility License Nos.
DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Unit Nos. 1, 2 and 3.
These amendments.
which are in
'response to your applications dated June 13, 1980 (TVA BFNP TS 139) and
'October 16, 1980 (TVA BFNP TS 152),
change the Technical Speci.fications to (1) revise t'e definitions for "Limiting Conditions for Operation" ard "Operability" to be consistent with the definitions in the model technical specifications enclosed with our generic letter to you of April 10,
- 1980, and (2) revise the definition of "Cold Shutdown" to be ccnsistent with the "Standard BMR Technical Specifications".
Cc pies of the Safety Evaluation and Notice of Issuance are also enclosed.
Sincerely,
Enclosures:
l.
Amendment No.
66 to DPR-33 2.
Amendment No.
62 to DPR-52 3.
Amendment No.
38 to DPR-68
- 4. 'afety Evaluation'.
Notice rr~
c.>
Thomas~A'~ Ippolito, Chief Operating Reactors Branch 82 Division of Licensing cc w/enclosures:
See next page
4>
%a C.
J
gr.
Hugh G. Parris CC:
H. S.
Sanger, Jr., Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E 11B 33C Knoxvil.le, Tennessee 37902 Mr. Ron Rogers Tennessee Val ley Authority 400 Chesttiut Street, Tower II Chattanooga, Tennessee 37401 Nr. Charles R. Christopher Chairr.an, Limestone County Commission P, 0.
Box 188
=
Athens',. Alabama 3561'1
",Ira L. Myers, N.D.
State Health Officer S.ate Department of Public Health S.ate 0 fice Building Montgomery',
Alabama 36104 U. S. Environmental Protection Agency Region IV Office ATTN:
EIS COORDINATOR 345 Courtland Street Atlanta, Georgia 30308 Nr. Robert F. Sullivan U. S. Nuclear Regulatory Comnission P. 0.
Box 1863
- Decatur, Alabama 35602 Nr. John F.
Cox Tennessee Valley Authority
'W9-D 207C 400 Commerce Avenue Knoxville, Tennessee 37902 Mr. Herbert Abercrombie Tennessee Valley Authority P.
0, Box 2000
- Decatur, Alabama 35602 Pr.
H. ti, Culvert 249n',"'.B3 400 Cvmerqe Avenue Tenne=-.see Valley Author'ity Knoxville, Tennessee 37902 Athers Public Library South and Forrest Athens,'labama 35611
.Direc>or, Office of Urban 8 Federal Affairs 108 Pa~~way Towers 404 James Robertson Way Nashville, Tennessee 37219 Direc.or, Criteria and Standards Division Office of Radiation Programs (ANR-460)
U. S.
Environmental Protection Agency Washington, D. C.
20460
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C; 20555 I
TENNESSEE VALLEY AUTHORITY DOCKET NO.
50-260 BROWNS FERRY NUCLEAR PLANT, UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 62 License No.
DPR-52 1..
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Tennessee Valley Authority (the licensee) dated June 13, 1980 and October 1'6, 1980, comply with the" standards and requirements of the Atomic Energy Act of
- 1954, as amended (the Act), and the Commission's regulations set or gh in 10 CFR Chapter I; B.
Thy facility will operate in conformity with the applications,
'the provisions of the Act,'and the regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by thiq 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 Commission's regulations; I
D.
The issuance of this amendment will not be inimical to, the common
~
defense and security or to the heal.th and safety of the public; and E.
- he issuance of this amendment i's in accordance with 10 CFR Part 51 of the Commission's regul,ations.
and all applicable requi'rements have'een satisfi,ed.
2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and par~ graph 2..C(2) of.Facility License No.
DPR-52 is hereby amended to read as follows:
(2) Technical S ecificati'ons The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
62, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
0 N
f
2-3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A tachment:
Changes to the Technical Specifications'ate of Issuance':
February 6, 1981 Thomas
. Ippolito, Chief Operating Reactors Branch 82 Division of Licensing
0'
ATTACHMENT TO LICENSE AMENDMENT NO.
62 FACILITY OPERATING LICENSE NO.
DPR-52 DOCKET. NO. 50-260 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages:
1/2 3/4 33/34 61/62 63/64 77/78 175 T87/188 The underlined pages are the pages being changed; the marginal lines on these pages denote the area being changed.
The overleaf page is provided for convenience.
2.. Add the following new page:
2a
0 Y
I
'R
XBTRODUCTlON This document presents the technical specifications for the Brovns Ferry Nuclear Plant Unit 2 only.
.~endsheet:lo.+
3ilUg 2 g '~g$
1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the, specifications may be achieved.
A.
able maintenance of the cladding and primary systems are assured.
Exceeding such a limit requires unit shutdown and review by the Atomic Energy Commission before resumption of unit operation.
Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency sub)ect
,to regulatory review.
B.
Limitin Safet S stem Settin LSSS
- The limiting safety system
'setting are se'ttings on instrumentation which initiate the automatic protective action at a 'level such that the safety limits will not be exceeded.
The region between the safety limit and these settings represent margin with normal operation lying below these settings.
The mar'gin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.
C.
Limitin Conditions for Operation (LCO - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility.
When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.
l.
In the event a Limiting Condition for Operation and/or associated requirements cannot be satisifed because of circumstances in excess of those addressed in the specifi-
- cation, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the permissible discovery or until the reactor is placed in an operat'onal condition in wnich the spec'f'cation is not applicable.
Exceptions to these requirements shall be statec in the individual specifications.
This provides actions to be taken or circumstances not d'rectly provided for in the specifications and where occur. ence would violate the intent of the specification.
For example, if a specification calls for two systems (or subsystems) to be operable and provides for'xplicit requirements if one system (or subsystem) is inoperable, then if both systems (or subsystems) are inoperable the unit is to be in at least Hot Standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Snutdown within the followina 30 ho>>rs if the inoperable condition is not corrected.
c Amendment No.
62
l.0 DER IHITIONS (continued 2.
When a system, subsystem, train, component or device is determined to be inoperable solely because its onsite povex source is inoperable,:or solelybecause its offsite pover source is inoperable, it may be considered operable for the purpose of satisfying the zequirements of its applicable Limiting Condition For Operation, provided:
(1) its corresponding offsite or diesel povez source is operable; and (2) all of its redundant system(s),
subsystem(s),
train(s),
component(s) and device(s).are
- operable, or likevtse satisfy these requirements. 'Unless both conditions (1) and (2) are satisfied, the unit 'shall be placed in at least Hot Standby vithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and'n at'east Cold Shutdmw vithin the follovtng 30 hou=s.
This is not applicable if the unit is already in Cold Shutdovn or Refueling.
This provision describes vhat additional
~ conditions'must be sa-isfied to permit operation to continue consistent vith the specifications for pover sources, vhen a" offsite or onsite pover source is not opezable.
It specifically
. pzohibits operation vhen one division is inoperable because its offsite or diesel pover source is inoperable and a system, subsystem, train, component or device in another division is inoperable foz another reason.
This provision pemits the requirements. associated vith individual systems, subsystems,
- trains, components or devices to be consistent vith the requ'rements of the associated electrical pover source.
It allovs operation to be governed by 'the time limit of the zequizements associated vith the Limiting Condition For Operation for the offsite oz diesel paver source, not the individual requirements for each
- system, subsystem,
- train, component oz device that is dete~ed to be inoperable solely because of the inoperabQ.ity of i s offsite or diesel poorer source.
D.
DELETE)
Amendment No.
62.
II
1.Q DEFINITIONS (cont'd E.
erable -
erabilit - A system, subsystem, trainp component, or device shall be Operable, or have operability when it is capable of performing its specified function(s).
Implicit in
,this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or othez auxiliary equipment that are required'or the system, subsystem,
- train, component or device to perform its function(s) are also capable of performing their zelated support function(s).
p
~cretin
- Operating means that a system or component is performing its intended functions in its required manner.
G.
Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
H.
Reactor. Power eration - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the.
reactor critical and above 1Z rated power.
X.
Hot Standby Condition Hot standby condition means operation with coolant temperature greater than 212'F, system pzessure less than 1055 psig, the main steam isolation valves closed and the mode switch in the Startup/Hot Standby position.
J.
Cold Condition Reactor coolant temperature equal to or less than 212 F.
K.
L.
Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 212'F.
Cold Shutdown The reactor is in the shutdown mode and the reactor coolant temperature equal to or less than 212'F.
M.
Mode of Operation A reactoz mode switch selects the proper interlocks for the operational status of the unit.
The follow'ng are the modes and interlocks provided:
1.
Staztu /Hot Standb Mode - In this mode the reactor protection scram trips initiated by condenser low vacuum and main steam line isolation valve colsure, are bypassed when reactor pressure is less than 1055 psig, the reactor protection system is energized with IRM neutron monitoring svstem trip, the APRM 15K high flux trip, and control rod withdrawal interlocks in service.
This is o ten referred to as Just Startup Mode.
This 's intended to imply the startup/Hot Standby position of the mode switch.
Amendment No.
II
le 0
'DEF INITIOHS (Conc '
2 Rnn Mode - In this soda the rseotor systes pressers is at or above'880 psia snd the rest or protection systea is anardi=td vith APRM protection (excluding the 15X high flux trip) <<nd RBH interlocks ird service.
3.
Shutdovn Mode Placing the" mode avicch to the shutdovn posi-tion initiates a reactor scram and pover to the control rod drives is removed.
After a short time period (about 10 sec) 8 the scram, sign<<l is removed alloving a scram reset ancL restoring tha normal valve lineup fn the control rod drive hydraulic sys-tem', also, the main steim line isolation scram encL main con-.
denser lov vacucaa scram 'are bypassed if reactor vessel pressure is be}o< 1055 prig.
4.
Ref'uel Mode - With the mode svicch Xn the refuel position'interne
~ locks are established so chat one control rod only may be vithss dravn vhen the Source Range Monitor indicate at leastps snd the refueling crane is not over che reactor; also, the main steam line isolation scram and main condenser 1ov vacuum scram are bypassed if reactor vessel pressure is bsiov 1055 prig. If the refueling, crane is over the reactor, a11 rods must, be tully inserted,and none can be vichdravn.
N.
Rated Fover - Rated pover refers co operation at a r'eactor pover of 38293 MMc; chir ia also termed 100 percent pover and is the maximum pover level 88uchorired by the operating license.
Rated steam flov,
~
rated coolant flove racad neutron ftoxp dhnd racd..t) nuclear'system pressure refer to the, values of these parameters vhen tha reactor is at rated pover.
Design pover.
cho pover to vhich tha safety analysis
- applies, corresponds to 3440 NWt.l O.
prima.
Containment Inte rit Primary containment -integrit7 "means that the dryvell and pressure suppression chamber are intact and <<11 of the folloving conditions <<re satisfied:
All-non-automatic concainmenc isolation valves on lines connected to tha"reactor coolant system or. containment vhich are not.rsc)uirad to'be open during <<cctdent conditions are closed.
These v<<ives may be opened to perform necessary operation<<1 activities
~
2.
At 'ess t one door in each dcir lock is closed and sealed.
3, All=automatic containmeat isolation valves are oper<<bio or deacti-vated in the isolated position.
4.
All blind flangas and manvdLyr ~re cLoead.
'p.
Secondar Concainmroc Inceer)c
- Secondary contd8inmaat iacegrity means that the raactor bui'ding is intact and the folloviag condiee ti'oni are met:
TASLF. 3.1.A REACTOR PROTECT IOH SYSTEM( (SCRA.'I)
[asSTR'".EDiTATI".:R RKQUIRKHEHT Min. Ho.
of Operable Inst.
Clsanncis Per Trip Sin'R en (l)
Trio Punction Trt Level Settin Modes in Mhich Function Must he 0 arable Shut-Startup/Hot dnun
~Re.'uel 7
~Scandb Run
~Aeelen S
1 Mode Suitch in Shutdoun 1
Manual Scrass
'P..'I (l6) 3 High Flux 3
Inoperative
<<120/12$ Indicated nn scale 2
2 2
APP'I (16)
High Flur High Plur loup era t ive l:nRdnsca le See Spec. 2.l.h.l
<<152 rated pover (13)
~
3 Indicated on Scale 2
Hlg':R R actor Pressure
< 1055 paid X
X
"(22)
X (22)
X X
x x(21) x(D)
X 21 X(D)
X(10)
(5)
(5)
X (15-)
X X(1.2) l.h l.h 1.A or l,b 1.A or l.b 1.A or l.b 1 hor l..b l.h H!rh OrRSR:ell Pressure (14)
Reactor Lov Mater I.evel (14)
High Mater Level in Scree Discharge Tank
<<2.5 psig i 53S" above vessel toro
<< N Calloao X(8)
X(g)
I X(2) x I.h-l.h
II
+BE 3.1.A (Continu~d)
Hin. No.
of Oped (lb'.'nst Channels Pei Tr'ip Sg; t.c:
(Ql l
M Hodes in Mhich Function
<'us t Oe~Cnerable S ta rtoo/I,"ot Action 1)
Hain Stea:: Line Isolation Valve Closure
< 10>> Valve Closur X(3)(6) x(3)(e',
x(e) 1.A or 1.C Turbine Cont-. Valve Fast Closure Turbine Stop. Valve Closure Turbine Ccntrol Valve-Loss cf Control Oil Pressure
> 550 psig x(4)
~Jpon trip of the fast X(~i) acting solenoid valves 10". Valve Closure X(4) x(4) x(4)
V(4) x(4) x(4) x(4) 1.A or 1.0 1.A or 1.0 1.A or 1.0 Turbine First Stage Pressure Permissive
< 154 psig x(1B) x(18)
X(18)
(19)
Turbine Condenser Low Vacuum
> 23 In. Hg, Vacuum X(3) x(3)
X 1.A or 1.C Hain Steani Line High Radiation (14) 3X Normal Full Power X(9)
Background
(20) x(9) x(9) 1.A or 1.C
0 0'
~
t 6.
Channe1 shared by RPS and Pr.mary Con Control System.
h channel failure.may 7.
h t rain fn cnn indi rsd o trip system.
a inment 6 Reactor
~'eschel Tso'latinn be a chan'nel failure in each system.
Tvo out of three SCTS trains reqvi"..ed.
action A and F.
A failure o. more than one vill require 9.
There Es only one tr'p system vith auto transfer to too poMer sources.
4l
~
Tlute.E '3.2;0 fDS'Eked".Q10'ATfou TAAf fNTTTATRS OR COt~aOLS KfK CORI A3O3 COWAINKlfT COOLIHG SYSTNS absua Qs O~tabfa Ocr function Xuetnascnt Cbanncl-kc~ctor Lou 'Uatet Level Tri Level Set tin
> +70 above vessel cero.
Aattaa Renames
- l. below trip setting fnttfatcd EKf.
fnstaa-~t Channel-KaacCot ~ Mater Lcr~l
> O7> "above vesael tero.
- l. Multfplfet telays. inf.tiate RCfC.
Instant Channel-
> 370" above vessel seto.
1cactor 4w Voter Level (LLS-3-56M), % tl) 2(lt)
Iaetrmaeat Channel -
> 378" abore vestal aero.
reactor Low Meter Level (L?5-3-5SA-D, Qf tl)
- 1. left tt fp aegtfng fattfatcc
-
Hnlttplier relaya initiate LpCf ~
- 2. Hultfplfet.relay Croa CSS fnftfatca accident
~ fgnaL (l5),
1, g fov trip eettfnga fn c<<)un<<fen fifth dtyvell high pteasute, fov vatcr level pctatosfve, 125 scc.
dc4y tfact and CSS or RfE puap nmatng, fnftfatca ADS.
l(16)
Instr~at Chaancl-Rcactor Lov Mater Level.
>ctrdasfre (Lf$-)-fN l, la),
SM ii)
> 5C4" above veioei aero.
- l. Eclov ttip setting persfssfrc fot fnftiatfng signals on ADS.
lnst~nt Channel-
> yl2 5/16" abave maoel aero.
A Rcac:or 4w Mater Level (2/3 core beigbt)
(LITS-3 5'2 L &2 ~
SM Pl)
Taetrur cat Chaancl -
1 p
2 5 i
D~ff Ufah Prcesqte fPS-6(-56 r.-u)
- 1. Selmt trip aattfng prevents faadvet" tent operation of oontainnant aptay during accident condf,tion.
- 1. EeloM trip setting ptorcats faadrcr" tent operation of containment spray duting accident cavfftfons.
-ThbLK 3.2.1 (Continued}
'ffnfaue So.
Operable Pcr trt
~Sa (I)
Functfon Trt Level Setttn Actfon Renarks Instr~ant Channcl-Dryvc1 1 Iltth Prcssure (PS-64-58 A D, SM I2) tnstruoent Channel-Reactor Lou Mater Level (LS-3-56A, b, Ci D) 2.5 psig
. i +70 above vessel rcro
- l. Abov>> trip setting ln con}unccton vtth lou reactor pressure (ntttates CSS.
'.Iulctolter relays tnftiace llPCl.
2..'lultlplfer relay froca CSS initiates acctdcnt stRnal.(1$ )
- l. bc!v i trip setctnb trips rcctrcula-Cion pumps fostruacnt Channel Reactor Iltth Pressure (PS-a-LUC A, b, C, D) fnitrumcot Channel DryMclt Ilfbh Prcssure (PS-64-5SA-D, Sll Il)
< 1120 pstR
~ 2.5 psig
- 1. Above trtp sctctnb trtps rectrcuta-clan pumps
- l. Above trip setting in con}unctton Wh tou reactor pressure initiates LFCa 2(16) tnstramcnt Channel-Dryvcll Htph Prcasu".e (PS-64-5 lA-D) instrument Channel-Rcac'toe Lou Prcssure (PS-3-14 A f D, SM I2)
(PS-68-95, SM I2)
(PS-6D-96, SM I2) instrument Gunnel Reactor Lou Prcssure (PS-3-]CA L b, SQ Il)
(PS-6a-95, Su Il)
(PS-65-96, Ql Il) "
2.5 psig 450 pity + 1$
230 psfb + 1$
1, Above trip scctfnlf tn con)uncttoa vtth.
lou reactor uatcr level, dryvell hfjh,
- prcssure, 120 sec. delay cfocr sod (3S or IUD pmap runntng, fnttt ~ tcs ADS.
1 ~
clou tr t 3>>ttlnf p'rgt5<<ve for 'nits CSS and
. ~i adwtss'1:n val;es.
- 1. Re trculatton dischar je valve ac:uatlon.
I
~
TAAI.F. -), 2. B (Cont in< r<Q
~fnfavn Hn.
Operable Per Funct ion Instrument Channel Reactor I~M Pressure (PS-68-93
$ 94, S'M IL)
Tri l.eveL Sett.'n 100 paig
+
L5 Act<
on Remarks Bolo~ trip settfng in con)unction Mith contafiuuent isolation signal and both.
suction valves open Mill close HAIR (LPCI) admission valves.
Core Sp: sy h<jto Sequencing 6 < t <<8 secs.
Tfmers
',5) l.
1lfth diesel pover 2.
One per motor I.PCI Auto Sequencing Tfmers (5) 0< t <<1 sec.
RlIRS'M h, BL ~ C3, and OL 13
<< t < L$ sec.
Timers l.
Mfth diesel pover One per motor 1.
Mith diesel pover 2.
One per pump Core Spray and LPCI huto Sequencfng Timers (6) 0 < t < 1 sec.
6 <
r.
<< 8 sec.
]2
<< t
< 16'ec.
18
< t < Qr< sec.
1.
Mfth normal pouer 2.
One per CSS motor 3.
TMn per RllR motor RllRSU A <, Bl, C3, and OL Timers 21
< t <29 sec.
l.
Mfth normal poMer One per pump
i~
I it l
XABI.E 3.2.E INSTRQKNTATHNWHAT YONIT)RS LB@AGE INTO DRSKU S stan 2
Setpoints Action Remarks Equipment Drain Flow Integrator Sump Fill Rate Timer Sump Pump Out Rate Timer
>2O.L min.
<13.4 min.
l.
Used to determine identifiable reactor coolant leakage.
2.
Considered part of sump system.
floor Drairr Plow Integrator Surap Fill Rate Tim r Surrp Pump.Out
'ate Timer N/A
>80,4 rain.
<8.9 min.
l.
Used to determine unidentifiable reactor coolant leakage.
2.
Considered part of sump system,
.Dryvell Air Sampling Gas and Particulate 3 x Average
Background
(3) itOTES r (1)
Whenever a system ie required tc be operable, there shaLL be one operable system-either automatic or man~1, or the action required in Section 3.6.C.2 sh)LL bc taken.
{2) hn alternate system to dctcrminr the Leakage flow is a aanual system-whereby the time between sump prcap starts is monitored.
Thc time interval vtLL dctcrminc thc leakage flov because thc volume of thc sump vill bc known.
(3)
Upon receipt of alarm, immediate action will be taken to confirm the alarm and assess the possibility of increased leakage.
~I
i rr Nfnfsana t of Opctable Instrument Chsnnela Instrument f TABLE 3.2.F SURVEIKAlfCC INSTRUMClTAYIQN Inntrumcnt Type Indication lfoccs LI-3-46 A LI-3-46 B PI-3 "54 PI-3-61 Reactor %ster Level Reactor Pressure Indicator 107.5" to (1) (2) (3)
+107.$ "
Indicator 0-1200 pafg (1) (2) (3)
PR-64-50 Pf-64-67 TI&4-$2 IR-64-$ 2 TR-64-52 TI<4-$$
TIS-64-$ 5 LI&4-$4 A LI-64-66 NA Dryvell Pressure Drptel1. Zempeiature Supprcia ion Chanber Afr Tc)aporatura Suppreasfon Chamber Mater Ta"pernturo Suppression Chsuber Mater Lnlcl Control Rod Posftfon Neutron Honftorfng Recorder 0-80 paia Indicator 0-80 psfa Recorder
~ Indica tot 0-400'P Recorder 0-400'F Indicator, 0-400'P Indicator -25" to
+25" 6V Indfcatfn8
)
Lfghta
)
SRH, IR!l, LPRH
)
0 to 100I povet)
(1) (2) ())
(1) (2) ())
(1) (2) (3)
(1) (2) (3)
(1) (2) (3)
(z) (i) ()) (i:
~
pS-64-67 Drywall Pressure Li'-04-: A LI-U4-1>
CAD tanl: "A" level CAD tunk "0" level TR-64-52 and Dryvcll Temperature and PS-64-58 B and Prcssure and Tfmcr IS44-67 Alar(s at 35 psfg
)
)
Alarm ff temp.
)
> 281'F and
)
prcssure
~ 2.5 paid after 30 e)fnute
)
delay)
]n(liest()r 0 to 1N'S lnlicator 0 to l~
(I) (2) (3) (4:
~i 0
LIIGTIIVjCOI I}ITIOIIS FOB OPERATION SURYALIr'ANCE AEQUI'Kl~MIi;.
3.6.A C
Thermal and Pre suri=ation Limitat,i on" 4.6.A
'I'hc r.-.w1 and Pr os sur is.c tion Lir'u, t.io>>s During heatup by non'-nu'clear
- means, except when the vessel is vented or as indicated"in 3.6.A.4, cooldown following nuclear shutdown on low-level physics tests, the reactor vessel temperatures shall be
,at or above the temperatures of curve iI2 of figure 3.6.1.
4 ~
The reactor vessel shell temperatures during,inservice hydrostatic or leak testing
'shall be at or above the temperatures shown on curve
/31 of figure 3.6-1.
The applicability of this curve to these tests is extended to non-nuclear heatup and ambient loss cooldown associated with these tests only if the heatup and cooldown rates do not exceed 15'F per hour.
5.
The react:or vessel head bolting, studs may be partially tensioned (four sequences of the, seating pass) provided the studs and flange materials are above 70'F. Before loading the flanges any more, the vessel flange and head flange must be greater than 100'F, and must remain above 100'F while under full tension.
6.
The pump in an idle recircula-tion loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 50'F of each other.
7.
The reactor recirculation pumps shall not be started unless'the coolant temperatures between the dome and the bottom head drain are within 145'F.
Amendment No.
50 175 3 0 7 ~
Neu ron flux mres "'rt'-'e stalled in thc reacts; vcssc'd$
accnt to the r ac -.-
vu"sc')'1 at the core nLid,".Ic;.. 1cv<<i.
The i res shlL1l be c:;.o; c.
an" test,cd durir~ 'hc f':".; r:."u: inc, Outa}..C tO CXp<<ri:.".Cntu'y Vur~
the calculatccl values o'cu 0
~
flue>>cc at one-fourth of -}:c bcl li'nc shel) thic.'=.es t."a.
arc used to dc-c..'c
"~ 'lD sh'ft from F'~~i=c 3.6-2.
- ".ecd h.".cn the reactor ve""c'o
'ng studs ar<< tc.-.si.
.thc reactor is 'n a c"'=
-c.-.."crate "c i;=cd'a:e' hc head fl"r."c s"a c
mancntly recorded.
w
v%
vW Prior to cnd du"'ng sec=sup of cr. idle rccircu'ation
"=",
he tc=.pcraturc of thc rc:;"
~ ".
cco an'. i>> thc oocrat'- an. '"'
loops shall bc pcmnently logged.
Pr'or to s artin;a r.=':". a-tinn pump, thc rcactc:
c tc:".pc 'cttuca 'n thc dc.-..c and.
r.
thc bott.om }>cad dra>n "."" b!
corparcd and pc=~)ancn y lo-e,c'.
Tc"t sI!ccimcna rcI)resent'.np the reactor vc::..cl, tuse
-clr> and uc1d heat. affect.cd zone metal cha.
be installed in thc ree.cmr ves=c'djacent to thc vasss
.<<ci.'
~
the core m'dplane level.
Thc nmbcr an" type of spec'."..=n" w<Il hc in accordance rcport. Nii'0-10ll5.
il!<<specimens shall mc<<t t..e intent ofSTH 3.05-70.
Samples shai'e %th-e"ave at onc-fou~~h ~ad:hree-fou th-service 1'"e.
Amendment No.
62
0 0
I I
(8IN(: COND ITLONS FOR OPEPAT ION GVRVETLI;AN( ~. A19UIREI<ENTS PR vAPY SYSTEM 2OU'NDARY a
6
'P Bl; ~DRY sY sTK'I BoUNDA'R~
5 Zf the requirements of 3. 6. H. 1 and 3.6.H. 3 cannot.
be
- met, an orderly shutdown snail be initiated and the reactor shall be, in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> lf a snubber ls determined to be in-operable whiie the reactor is ';. tfie.
sfiutdown or refuel
Snubbers may be added
-'a'ety-related systems withou. orior license amendmen.
to Table ".6.H provided
,tha-a revision to Table
'3.6.H
',s included with a su".se"uen license amendment request.
Once each refueling cycle, a representative sample of 10 snubbers or approximately 10" of the snubbers, whichever is less, shall be functionally tested for operability including verification of proper piston movement, lock up and bleed.
For each unit and subsequent unit found inoperable, an additional 10" or ten snubbers shall be so tested until no more failures are found or all units have been tested.
Snubbers of rated capacity greater than 50,000 lb need not be functionally tested.
187
0
~ 1
Curve 8 1 Hin 1:-.. ",r ccml)cr-:u) ".
for prcssure ests such as required hy Section X[.
l20 l000 gc "c5-80 Carve'.
82 l".inimum t~mptraturi" for n.eci)an'ca1 h~.t
))p or cnoldcMn
) oil.owing n))c 1 ea c
~hu tdo)-w Curve
~3 llinimum terpcraturc.
for corp <n~.ratio.x (cr1tfrai it':)
I))cludu'ddi tiona:
margin ri quar~ d 1 OCo K )el lhppcnl) i,c.
Q
~
Par. ['.AC.
- "."0 i R
~ ]le pic c)1'."vi~s
>cit<crl 30'['o 1.l:c x ij'lit o" th)~ oriel) l) re t of curvrs'o 30 F, Tl)il sl)T' Qi.1 a 11 o'J these curv;!8 usi:8 thru ~.0 E."PY.
1/
r
,i
- PX)'VP.
I 2'.)0 l',!!l[~l!i~ Tr~lP'. PA")"iRE
( ")
188 Amendment No.
62
~I 1'
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I
+**y4 UNITEDSTATES NUCLEAR REGULATORYCOMMlSSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY TME OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 66 TO FACILITY OPERATING LICENSE NO.
DPR-33 AMENDMENT NO.,
62 TO FACILITY OPERATING LICENSE NO.
DPR-52 AMENDMENT NO.
38 TO FACILITY OPERATING LICENSE NO.
DPR-68 TENNESSEE VALLEY AUTHORITY BRO>lNS FERRY NUCLEAR PLANT UNITS NOS.
'1, 2
AND 3 DOCKET NOS.
50-259, 50-260 AND 50-296 1.0 Introduction By letters dated. June 13, 1980 (TVA BFNP TS 139) and October 16, 1980 (TVA BFNP TS 152), the Tennessee Valley Authority (the licensee or TVA) requested
.changes to the Technical Specifications (Appendix A) appended to Facility Operating Licenses Nos.
DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Unit Nos.
1, 2 and 3.
The letter of June 13, 1980 was in response to our gener ic letter of April 10, -1980 to "all power reactor.
licenses" requesting that they sub-mit proposed changes to their technical. specifications that incorporate
~
the requirements of the Model Technical Specifications (which were enclosed with our generic letter) to employ an explicit definition of the term OPERABLE for all components of safety related systems.
The letter of October 16, 1980 was a request to change the definition of "Cold Shutdown" to make it consistent with the "BMR Standard Technical Specification."
- Thus, the proposed amendments and r evised Technical Specifications would make the Browns Ferry Units 1, 2 and 3 Technical Specifications.consistent with the NRC Standard Technical Specifications
'n defining operability of safety related systems and cold shutdown.
- 2. 0 Evaluation Our. generic letter of April 10, 1980 discusses the basis for our request that all licensees review and, as necessary, revise the technical speci-fications for their facilities.
Basically, the thrust of the"model technical specifications is to insure that there are not only limiting conditions of operation (LCOs) that require all redundant components of safety related systems to be operable but that the LCOs address multiple outages of redundant components and the effects of outages of any support
0' y ~,>
rs
systems
- such as electrical power. or cooling water - that are relied upon to maintain the OPERABILITY of the particular system.
The changes to the definitions of "Limiting Conditions for Operation" and "Operability" proposed in TYA's submittal of June 13, 1980 are consistent with the definitions in the Model Technical Specifications enclosed with our generic letter of April 10, 1980.
The proposed changes are acceptable.
In the submittal of June 13, 1980, TVA also proposed several additional administrative type changes to the Technical Specifications for Units 1, 2 and 3 to correct errors or omissions.
Each of the proposed changes're discussed below.
On page 16 of the Unit 1 Technical Specifications, the value for the
.safety limit'inimum critical. power ratio (SLHCPR) shown in parenthesis
's.the old value of 1.06.
This limit only applied during the first fuel cycle.
Once Bx8 fuel was added to the core, the SLMCPR became 1.07;
~
a SLMCPR of 1.07 is the correct and accepted value for all three units.
The proposed change corrects an administrative error and, is acceptable.
By letter dated November 9,
- 1979, we issued Amendment Nc:. 53, 49 and 26 to Facility License Nos.
DPR-.33,. DPR-52 and DPR-68 in resp
.=.e to TVA's application of August 27, 1979.
The amendments
- char, Technical Specifications to increase the high drywell pressure
. level setpoint from 2.0 psig to 2.5 psig.
Besides the changes in T
- .:plication of August 27, 1979, there were additional pressure swit
-:or which the ranges shown in the Technical Specifications should hai been changed by the amendments but which TVA failed to include in their application.
The proposed changes on pages 33, 62, 63 and 78 of the Technical Specifications for Units 1
and 2 and the proposed changes on pages 32, 65 and 81 of the Technical Specifications for Unit 3 correct these omissions.
The changes are acceptable.
The proposed change on page 219 of the Unit 3 Technical, Specifications is to add two new snubbers to the list of snubbers to be inspected.
The two new snubbers are for the residual heat removal service water system.
The proposed additions are acceptable.
The proposed changes on page 188 of the Unit 1
and 2 Technical Specifi-cations and on page 201 of the Unit 3 Technical Specifications is to.
'substitute a revised Nil Ductility Temperature (NDT). operating curve.
These curves specify the minimum temperature that the reactor vessel must be. at for the range of primary coolant pressures.
The upper portion of the curves provides an additional 20'F shift from the original curves for protection because of uncertainty of radiation damage.
The lower portion of curves 2 and 3 reflect the limiting conditions for protection of the feedwater nozzles from degradation.
This lower portion includes the 40'F conservatism for,nuclear heatup.
These proposed curves are more conservative than those in the present technical specifications and have been administratively imposed.
The proposed changes are acceptable.
0 0'
I J'
The change'on p.
354 of the Unit 3 Fire Protection Inspection program is to change the time frame for the audit to make it consistent with the time frame specified for Units -'nd 2.
The proposed change is acceptable.
The audit by an outside consultant for the first 3 year period has been completed.
3.0
-Envi.ronmental Considerations Me have determined that these amendments-do not authorize a change in effluent types or total amounts.nor an increase in power level and will not.result in any significant environmental impact.
Having made this
,determination, we have further concluded that these amendments involve an action which is i:nsignificant from the standpoint of environmental impact; 'an'd pursuant to 10 CFR Section 51.5(d)(4} that an environmental impact statement, or negative declaration and environmental impact app'raisal need not be prepared in connection with the iss'uance of these amendments.
4.0 Conclusion Me have concluded that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2J there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, and (3) such activities will be conducted in compliance with the Gormission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
February 6, 1981
4l 0'
(,M'
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS.
50-259 50-260 AND 50-296 TENNESSEE VALLEY AUTHORITY NOTICE OF ISSUANCE OF 'AMENDMENTS TO FACILITY OPE TING ICEMSES The U. S. Nuclear Regulatory Commission (the Commission} has issued Amendment No. 66 to.Facility Operating License No.
DPR-33, Amendment No. 62 to:Facility Operating License No.
DPR-52 and Amendment No.
38 to Facility Operating License No.
DPR-68 issued to Tennessee Valley Authority (the licensee),
for operation of the Browns Ferry Nuclear Plant, Unit Nos. 1, 2 and 3, located in Limestone County, Alabama.
The amendments are effective as'of the date o$ issuance.
The amendments change the Technical Specifications to:
(1} revise the definitions.for "Limiting Conditions for Operation" and "Operability" to be consistent with the definitions in the model technical specifications enclosed with the Commission's generic letter of April 10, 1980 to "All Power Reactor Licensees" and (2)'evise the definition of "Cold Shutdown" to be consistent with the "Standard BWR Technical Specifications."
The applications for the amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Cqmmission s regulations.
The Commission has made appropriate findings as required by the Act and 'the Commi.ssion's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.
Prior public notice of these amendments was not required since the amendments do not involve a significant hazards consideration.
0 0 W>
'590-01 The Commission has determined that'he issuance of these amendments will not result in any significant environmental impact and that pursuant to 10 CFR Section 51.5(d}(4} an environmental impact appraisal need not be prepared in connection with issuance of these amendments.
,For further details with respect to this action, see (1) 'the applications for amendments dated June 13, 1980 and October 16,
- 1980, (2) Amendment No.
66 to License No.
DPR-33, Amendment No.
62 to License No.
DPR-52, and Amendment No.
38 to License No. 'DPR-68, and (3) the Commission's related Safety
'valuation.
All of these items are available for public inspection at the Commission's Public Document
- Room, 1717 H Street, NW., Washington, D. C.
and at the Athens Publ,ic Library, South and Forrest,
- Athens, Alabama 35611.
A copy of items (2) and (3} may be obtained upon request addressed to the
". S. Nuclear Regulatory Commission, Washington, D.
C.
20555, Attention:
girector, Division of Licensing.
~
Dated at Bethesda, Maryland, this 6th day of February 1981.
FOR THE NUCLEAR'EGULATORY COMMISSION c~
Thomas A. Ippolito, Chief Operating Reactors Branch. 82 D~.v>.sion of Li.censing
~
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gu cP e
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