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Amends 66,62 & 38 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Revising Tech Specs to Define Limiting Conditions for Operation & Operability to Conform W/Model Tech Specs & Cold Shutdown Consistent W/Bwr Tech Specs
	ML20148T877
Person / Time
Site:	Browns Ferry
Issue date:	02/06/1981
From:	Ippolito T; Office of Nuclear Reactor Regulation
To:	;
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ML20148T881	List: ML20148T877; ML20148T884; ML20148T891;
References
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.m arcoq jof UNITED STATES g

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g NUCLEAR REGULATORY COMMISSION Ti

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPEPiTING LICENSE Amendment No. 66 License No. DPR-33 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The applications for amendment by Tennessee Valley Authority I

(the licensee) dated June 13, 1980 and October 16, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 regulations of the Commission; C,

There is reasonable assurance (i) that 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 Commission's regulations; D.

The issuance of this amendment will not be inimical to the common l

defense and security or to the health and safety of the public; i

and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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A:cordingly, the license. is amended by changes 'to the Technical Saecifications as indicated in the. attachment to this license amendment and paragraph 2.C(2) of Facility Lic'ense No. DPR-33 is hereby amended to read as follows:

(2) Technical Soecifications

' The Technical Specifications contained in Appendices A and B, as revised through Amendment'No. 66, 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 COMMISSIO"

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M d.9. Q. A '

Thomas A. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing

"- ic-ert:

C a9;es to the Technical hs:ifications Cn e :' Issuance:

February C,1981 1

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I ATTACHMENT TO LICENSE AMENDMENT NO. 66 FACILITY OPERATING LICENSE NO. OPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

1.

Remove the following pages and replace with the identically numbered pages:

1/2 3/T T5/16 33/3T 6T/62 63/64 77/~78~

175 TET/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 o

INTRODUC!XCH This document presents the te:hr.ical specificatices for ?.h3 Brevns Fe.-ry Bucler.: Plant Unit 1 c=17 e

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1.0. DEFINITIONS The succeeding frequently'used terms are explicitly defined'so that a uniform interpretation of the specifications may be achieved.

A.

Safety Limit - The safety limits are limits below which the reason-able maintenance of the cladding and primary.syseens are assured.

J

. Exceeding such a limit requires unit shutdown and review by the i

Atomic Energy. Commission before resumption of unit operation.

l Operation beyond such a limi: may not in itself result in serious consequences but 'it indicates ' an operational. deficiency subject to regulatory review.

3.

Limiting Safety System Setting (LSSS) - The limiting safety system

' 1

' setting are settings on instrumentation which initiate the 4

automatic. protective action at a level such that the safety limits-will not be exceeded.

"he region _ between the saf ety lin1: and j

these settings represent margin with normal operation lying belov these settings. The : margin has been established so that with _

proper operation of che instrumentation the safety limi:s will never ba exceeded.

C.

. Limiting Conditions for Cteration (LCO)' - The limi:ing conditions for. operation specify the - ** = = acceptable levels of system performance necessary to assure safo startup and operstion of the facility. When these conditions are met, the plant can be opera:ed safely and abnormal situa: ions can be safely controlled.

1.

In the event a

'd

':ing Condi:1on for Operation sed /or associated require =ents cannot be satisifed because of circumstances in excess of those addressed in the specifi-

~ !

i cation, the uni shall be placed in at least Hot Standby

' within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shu:down within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months unless corrective seasures are comple:ed : hat per=1: operation under :he permissible discovery or until':he reactor is placed in an operational condition in which the specifica:ica is not applicable. Exceptions to these requiremen:s shall be stated in the individual specifica: ions.

his provides actions to be taken for circumstances not directly provided for in the specifica:iens and where occurrence would viola:e the intent of the specification. For example, if a specification calls for two systems (or subsystems) to be operable and provides for explicit requiremen:s if one system (or subsystem) is inoperable, then if both systems (or subsystems) are inoperable the uni: is to be in at least Hot Standby in l

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and.in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months if the inoperable condition is not corrected.

l Amend $entNo.66 l

1.0 DEFINITIONS (continued) 2.

When a systen, subsystem, train, component or device is dete.nined to be inoperable solely because its onsite power source is inoperable, or solelybecause its offsite power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicable Limiting Condition For Operation, provided:

(1) its co::esponding offsite or diesel power source is operable; and (2) all of its tedundant system (s), subsystem (s), train (s),

component (s) and device (s) are operable, or likewise satisfy i

these requirements. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Hot Standby withh 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least Cold Shutdown within the following I

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This is not applicable if the unit is already in Cold Shutdown or Refueling. This provision describes what additional conditions must be satisfied to permit operation to continue consistent with the specifications for power sources, when an offsite or onsite power source is not operable. It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperable and a system, subsystem, train, component or device in another division is inoperable for another reason. This provision permits the requirements associated with individual systems, subsysta=s, trains, components or devices to be consistent with the recuirements of the associated electrical power source. It allows operatien to be governed by the cine id d: of the requirements associated with the i d d ting Condition For Operation for the offsite or diesel power seur:e, not the individual requirements for each systen, subsyste=, train, ce=ponent or device that is determined to be inoperable solely because of the inoperability of its i-l-

offsite or diesel power source.

D.

DC.I D l

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e 2a k.endment No. 66

_. _ ~ _ _. _ _

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l 1.0 DEFINITIONS (cont'd) l E.

Operable - Operability - A system, subsystem, train, 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 other auxiliary equipment.that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

F.

Operatina - Operating means that a system or component is performing its intended functions in its required manner.

G.

Inamediate - Immediate means that the required action vill be l

initiated as soon as practicable considering the safe'operatien of the unit and the importance of the required action.

l H.

Reactor Power Ooeration - Reactor power operation is any operation-with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1% rated power.

l I.

Hot Standbv Condition - Het standby condition means o'peration with coolant temperature greater than 212*F, system pressure less than 1055 psig, the =ain stea= 1 solation valves closed and the mode.

switch in the Startup/Ect Standby position.

J.

Cold Condition - Reactor coolant temperature equal to or less than 212*F.

K.

Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 212*F.

L.

Cold Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature equal :o or less than 21 *F.

M.

Mode of Oce:ation - A reactor mode switch selects the proper interlocks for ne operational status of the unit. The folleving are the modes and interlocks provided:

1.

Startup /3ot Standby Mode - In this mode the reactor protection i

scram trips initiated by condenser low vacuum and main steam line isolation valve colsure, are bypassed when reactor pressure is lein than 1055 psig, the reactor protection l

system is energized with IRM neutron monitoring system ::1p, the APRM 13% high flux trip, ated control rod withdrawal interlocks in service. This ic of:en refer:ed to as just i

Startup Mode. Tc.is is intended to imply the startup/ Hot Standby position of the mode switch.

t 3

3 Amendment No. 66

1.0 DEFIN2780HS (Cons'd) -

2.

Run Mode - In thin mode the renc:or syste:; pressure is at or above 850 psig and the reactor protection system is energi:ed with APRM protec:iun (excluding the 15: high flux trip) and REM interlocks in s e rvic e.

3.

Shutdown Mode - Placing the mode svttch to the shutdosu posi-resete r scram and power to the control rod tion ini:iates a drives is removed. Af:er a short time period (about 10 sec),

the scram sigr.41 is removed allowin6 a scrss reset and restoring the normal valve lineup n the control red drive hydraulic sys-tem; also, the amin steem line isolation scram and main con-denser low vacuum scram are bypassed if rese:or vessel pressure is below 1055 psig.

4.

Refuel Mode Vith the mode svi:ch in the refuel position inter-locks are established so that one con:rol rod only may be with-drawn when :he Source P.aoge Monitor indicate at least 3 eps and the refueling crane in not over the resc:or; also, the main sceau line isolation scram and main enndenser low vacuum scram are bypassed if reactor vesaci pressure is belov 1055 peig. If the refueling crane is over :he reac:or all rods must be fully inserted and none can be withdrawn.

N.

Rated Fever - Rated power ref ers to opers: ion at a reactor power of 3,293 MW:; this is also t e rmed 100 pe r c en: power and is the maximun power level authori:ed by the operating license. Rated steam flow, rated coolant flov. rated neutron flax. end rn: del nuclear systen pressure ref er to the values of these parasatero when the reactor is at rated power. Design power, the safety a=alysis applies, corresponds to 3440 MW:.! power to vnich the C.

Primarv Contain en: Int e erity - Prim.ary c:n::incent integrity = mans that the dryvell and pressure suppressior enanber are intact and all of the following conditions are astisfiec; 1.

All non-auto =sti: contain=ent isola:::n v:1ves on lines tonnect:d to the reac:or :colant syst'os or con: sin =en: which are not required to'be open durin$ acciden: conditions are :losed. These valves may be opened :: perfer: necessary opera::: 41 activities.

2.

At least one door in each g irlock is c'.o s ed and sealed.

3.

All automA:ic CCUCainMcC: isolation Valves are operabl5 or desC:1-vated in the isolated position.

4 All biind flan 6es and =anways are c ic eed,

o F.

Secondary _ Con:ainme=: Integettv - Secondary con:ainment integriry means :ha: the reac:: building is ints : and :te follrving condi-tions are met:

4

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g.1 nasr.s t,rUEL CLAD 0!!:C 1RTr.CRITY S ATETY LIMIT j

I The fuel' cladding represents one of the physical barriers which separate radio-active materials fron environs. The integrity of this cladding barrier is related to its relative f reedom f rom perf orations or cracking. Although some corrosion or use-related cracking may occur during the lif e of the cladding, fissien product migration from this source is incrementally cumalative and,,

continuously sessurable.

Tuel cladding perf orations, however, can result f rom the rmal stresses which occur f rom reactor operation significantly sbeve design conditions and the protection system setpoints. While fission product migration fres

' cladding perf ormation is just as measurable as that f rom use-related cracking, the ther= ally-caused cladding pe-f orations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deteriora-tion. Theref ore, the fuel claddist saf ety limit is defined in terms of the reactar operating conditions which can rest ic in cladding perf oration.

The fuel cladding integrity Itait is set such that no calculated fuel damage would ectur as a result of an abnor=al operational transient. Because fuel damage is not direc t1'y observable, the fuel cladding Safety LLait ta defined with margin to the conditions which would produce onset transition boiling (K;7R of 1.0).

This establishes a safety Lir.it such that the nintaua critical power ratio (MCTR) is no less chan 1.07, MCFR >1.07 represents a conservative margia relative to the conditions required to maintain fuel claddi:g integrity, onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad snaperature and the possiblity of clad failure.

Since boiling transition is not a directly observable parameter, the sargin to oeiling tranuition is calculated f rom plant operating parameters such as core pese r, cors flow, feedwater temperature, and core power distribution. The margin f or each fuel assembly.is characterized by the critical power ratio (CFg) which is the ratio of the bundle power which would produce onset of transition heiltag divided by the actual bundle power. The mini =us value of this ratio for any buedle in the core is the minimum critical power ratio ()t:7R). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instru-eented variables, i.e.

normal plant operation presented on Figure 2.1.1 by the ne=fnal exaeeted flow canten1 lira.

The Sa fety ti= 1t (wc7R of 1.c7) han..dFicient conservatism to assure that in the event of an abnormal operational transient initif tet f rom a norr.41 operating condition (McFK > 11=1ts specifief. is specificatics 3.5.K)=cre than 99.9% of the fuel.

rods in the core are expected to avoid boiling transition. The marita between M ?.i of 1.0 (onset of transition boiling) and the saf ety limit 1.C7 __, is d erived a detailed statistical analysis considering all of the uncertainties in semi-from coring the core operating state including uncertainty in the boiling transition correla*. ion as described in Ref erence 1.

The uncertainties employed in deriving the saf ety limit are provided at the beginning of each fuel cycle.

15

cendnent No. 35, 47

. 1.1 BASES necause the boiling transition correlation is based on a large quantity of full scale data there is a very high confidente that operation af a fuel assembly at the condition of MCTR = 1.07. vould ut produce. toiling tran.

sition. Thus, although it is not required to establish the safety ljait additional margin. exists between the safety limit and the actual occurence of loss of cladding integrity.

However, if boiling transition vere to occur, clad perforation vould not be erpected. Cladding tersperatures woul'd increase to approximately 1100er vnich is belev the perfers. tion teroperature of the cladding material. This has been verified by tests in the General Electric Test 7teactor (CETR) where fuel similar in design to RTNP operstes' atove the critical heat flux for a significant period of time (30 mieutes) without clad perforation.

If reactor pressure should ever exceed ikCo psia during normal power operating (the limit of applicability of the boiling transitico corre-lation) it vould be assumed that the fusi cladding integrity Safety 1,imit has been violated.

In addition to the boiling transition limit (NCFR = 1.07) operation is constrained to a maximum LJICR cf 13.5 kv/f t for 7x7 fuel and 13.4 kv/f t for all 8x8 fuela. This linit is re. ached when the Core Haximum Tractica of Limiting Power Density squals 1.0 (CMT1.PD = 1.0).

For the case where Core Maxious T:setion of Linitinc Power Density exceeds the Trac: ion of Rated Ther:ssi Power, operation is permitted only at less than 100* of rated power and only with reduced APRM scrse settings as required by specificacia 2.1.A.1.

At pressures belov 800 psia, the core elevstion pressure drop (0 power, C fiev) is greater than k 56 psi.

At low powers and flevs this pressure differential is maintained in the bypast regico of' the ecre. Since the pressure drop in the bypass regien is essentially all elevatien head, the core pressure drop at lov pcvers and flov vill alvate be greater than b.56 psi. Analyses show that vith a flov of 26:C 0J lbs/hr bundle flov, bundle pressure drop is nearly independent of bundle power and has the bund.le flov vish a h.56 psi driving head a value of 3.5 psi. Thus,3 lbs/h. Tull scale ATLLS t.est data taxen vill be greater than 29x10 at pressures frors ik.T psia to 500 p.sia indicate that the fuel asse=bly critical power at this flev is apprcximately 3.35 HVt. Vith the design peaking factors this correspends to a core thermal pover of more than 5C%. Thus, a core thermal power limit of 255 for reactor pressures' below 800 psia is conservative.

For the fuel in the core during periods when the reacter is shut dovn, ccm-sideration must aise be given to voter level requirements due to the effe=t of deesy heat. If water level should drop below the top of the fuel during this time, the ability to remove decay. heat is reduced. This reduckjon in cooling capability cou.1d lead to elevated cladding temperatures and elad perforstion. As 1cag as the fuel remains covered vith vnter, sufficient cooling is available to prevent fuel clad perforation.

16 Amencment No. jlf, 66

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REACTOR FROTECTION Sv5Tei (SCRrt) 1:iSTR*f1ENTATIT4 REQUIRTHER

%D ttin. No.

I m

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Operable Inst.

Hades in Which f unction Cli.a nisci s

!tust Be Operable Fer Talp Shut-Startup/Itot s yn.t s a (1)

Trip Function Trip t.evel Setting down Refuel (7)

Standby M

Action (l) 1 Hode switch in shutdown y

g H

g 1,A 1

Hanual Scram x

K X

X 1.A w

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?-(22) X (22)

X

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Inoperative g

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High flus See Spec. 2.1.A.L 2

lite,h rius

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(LS) 1 A #f 1.3 2

taoperative (13)

X(21)

X(17)

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t;cunscate t 3 Indicated on Scale (11) 11)

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litpi seactor Pressure < 1055 petg X(kO)

X X

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7 A t ra Ln is considered a trip system.

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Os Amend,ent No.

f,66

d p

k=n n

3 et I ABLF. 3.2.5 (Continued) 2 P

Minte k.

k Operable ter isIa 3rs (11 Funct1on Trip Level 5ettinat Act1on Renarks

1. Above trip metting in conjunction with cn A

2 Instrument Channel -

1 2,5 pegg Dryvett litsh fressure low reacter pressuire inttletes C55.

(ri-64-18 A-0, SV M) tutstelter relays initiate ItrCt.

2. ?tuttipiter relay (tom CSS f nitiates accilent signal.(13) 2 tastrument Channel -

~. e/70'*above ves sel se co A

1. Beta, trip setting trips rectreuls-Reactor Lev unter Level tion pumps (13-3-36A, 8. C, D)

A

1. Above trip setting trips tectrcuta-2 Instrument Channel j i120 pstg Beafter Itt h Pressure tt an purops t

(ri-s-tut A, 3, C, 0) o, u

wi tla 2

In trument Channel -

1 2.5 psig A

1. Above trip setting in conjunction Drywell Illgh Pressure low reactor pressure taittates trCI.

(rs-44-36A-D, Su it)

A

1. At.ove trip setting in conjuncttom with 2(16)

Instr *=ent o ennel -

e j,5 p,g g

Dryvelt Illgh Frcasure low reactor water level, dryvelt high (rS-6t-HA-D) pressure,120 sec, delay tioer and CSS o r MtA puerp runn ing, 'i n t t i s t e s ADS.

I tantrurent Channel -

i+50 ps ts + 13 A

1.

Selow tr*r se'tlnc perntss tre tbr :pentr4 CSS and ' CI Ed stssi;r, valles.

J Reactor law Pressure (PS-3-74 A & t, Su d2)

(PS-68-95, su d2)

(FS-60-S6, su # 2) 2 lastr racot Chanact -

230 peig + 13 A

1. Re:trculatlon dischorde ValV4 ucassor low ricssure se:natlon.

(rs-1-JtA L h,

su st)

(rs-68-11, su st)

(r5-65 '>6, su it) i

-m m

~

.I IABLE 3.2.0 (Continued) j M

u ta.n :le,.

Operable Per 7t ip Sy s gt)

Function Trip Level Setting Action Remarks Instrument Channel -

100 pelg i 13 A

1.

Below trip setting in conjunction kith Reactor 1.ou Pressure conta ltsweut isolation sip,nel and both (P5-68-9) L 94, SW #1) suction valves open will close RilR'(LPCI).

admission valves.

i 2

Core Spray Auto Sequencing o< tj 8 secs.

B 1.

With diesel power Timers (5) 2.

One par motor 2

1PCI Auto Sequencing Oj t_<1 e'ec.

B 1.

With diecel power ilmers (1) 2.

One per motor on t.

i 1

RitR54 ^ ?. 81, C1, and D1 1] < t < 13 sec.

A 1.

Utth diesel power 4

j Timers 2.

Due per pump Z

Core Sprar and LPCI Auto Oj tJ l sec.

B 1.

With normal power l

Sequencing Timers (6) 6 < t < 8 sec.

2.

One per CSS motor 12 < t < !6 sec.

3.

Two per R11R motor 18 < t < 04 sec.

i i

i 1

RalR5u A1, 81, C3, and D1 2?j t_<29 sec.

A 1.

With normal power i

Timers 2.

One per pump l

.TARI.E 3.7.E INSTRUtIEITTATION TilAT Po?IIT rts 1.rAYACE INTO DRYME11 System (2)

Setpoints Action Remarks Equipment Drain (I) 1.

Used to determine identifiable reactor Flow Integrator N/A coolant leakage.

i Sump Fill R te 2.

Considered part of sump system.

Timer

>20.1 min.

i Stump Pump Out it a t e T ime r 113 4 "I"-

j Ploor Drain (I) 1.

Used to determine unidentifiable Flow integrator N/A teactor coolant leakage

$smop Fill kate 2.

Considered part of sury system.

}

Tir.er 380.4 min.

O Sunp Puop Out 4

Itate Timer 18.9 min.

Dryuell Air Sampling Cas and 3x Average (3)

Particulate Backgroun.1 i

1 2

i NOTES:

i l

(1) Whenever a system le required te be operable, there sha L1 be one operable system either automatic or canal, or the action required in Section 3.6.C.2 sh all be taken.

(2) An alternate system to detertaine the leak. ige flow is a sanual systers whereby the time between aump pump starte is monitored.

The. time interval u111 determine the. leakage flow because the voltsne of the sump will be known.

(3) Upon receipt of alarm, luunediate action will be taken to confirm the alarm and assess the possibility of.

increased leakage.

ev -

- ' ' ~

w v

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4 1

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TABLE 3.2.F 0

SURVEILLAllCE lilSlRlRIENTATION y

Itinians f of Operable Instrument Type indleation

~

i g;

Channels Testrument #

Instrument and Range Notes 2

Lt-3-46 A teactor Vater Level Indicator -107.5" te (1) (2) (3)

LI-3-46 B

+107.5" 2

73-3-54 Reactor Pressure Indicator 0-1200 pets (1) (2) (3)

]

FI-3-61 i

l 2

PR-64-50 Drywell Fressure Recorder 0-80 pela (1) (2) (3)

J PI-64-67 Indicator 0-80 pela i

l 2

T1-64-52 Dryve11 Teaperature Recorder, Indicator (1) (2) (3) 1R-64-52 0-400'F l.

1 TR-6 4-5 2 Suppression Chamber Air Recorder 0-400'T-(1) (2) (3)

Tcuperaturs y

a>

1 TI-44-55 Suppression Chasher Water Indicator. 0-400'F (1) (2) (3) tis-64-55 Tamperatura 2

LI-64-54 A Suppresaton Chanber Unter Indicator -25" to (1) (2) (3)

LI-64-6f, Level 425" 1

NA Control Rod Position

&Y Indicating

)

I.lghts

)

1 HA Heutron Ifoultoring SRH, IRM, LPPM

)

(1) (2) (3) (4.'

O to 100I power) 1 FS-44-67 Drywell Fressure Alarm at 33 psig )

)

i 1

TR-44-52 and Dryucil Temperatura and Alarm if temp.

)

FS-64-58 B and Precoure and Timer

> 281*F and

)

(1) (2) (3) (4' 15-64-61 pressure

2.5 paid after 30 minute

)

deley) 1 1.1 Uk-?A CAh tank "A" level In.11cator O to 10('i (1) 1 1.I-C4. l iA Call Lupk "C" level Inticator O to 100(

(1) i t

f

LI!C'" !P", C0f."JI"* IONS FCit OPE?AMON SUWEILIN!OE TE::UrtCC',

3. 6./s hrmal nnd Pressurization 4.6.A Th e r--e.l a nd. Pressurinctiem Li rrl ta t.i cn:

Litt: tat;en:

3 During heatup by non-nuclear 3

Te:t specimena reprece n.ing tr.e means, except when the vessel reactor venecl, buse.e;e anc we;j 3

is vented or as indicated in heat, uf fected zone metti :nall Le 3.6.A.4, cooldown following installed in the reactor ve:.el nuclear shutdown on low-level ad; scent to the ve :ci ::13 e, physics tests, the reactor the core radplane level. T3e vessel temperatures shall be m=ber ad type or :pecteer,:

at or above the temperatures vill be in accordance w;th

,E rep:rt E 0-1011$. The *- c-i -a ns of curve #2 of figure 3.6.1.

chc.3, meet the intent or 3TM -

14-70. Oste.ples che.11 be with.

4.

The reactor vessel shell d.rawn at one-fourth a..4 three-temperatures during inservice fourto: Se /i:e life.

hydrostatic or leak testing shall be at or above the N'

N8'tf ^ f M Vi#88 hf'l be in-U d i OC EC# VC-temperatures shown on curve "N " *#

U#-

1 of figure 3.6-1.

The

~

^# 1# '** '

applicability of this The

.re: :M be rce..:. t : e.n t curve to these tests is t c. c.., i...

.., c,.n....

.. ;ue_, ng m.

..e extended to non-nuclear expc.. c

.u._...y

.. c...s.,

ou..

.e.

r.,

heatup and ambient loss

+ h.. ~ * ~_, :,, a.

,<a

a...

o.

cooldown associated with riuen:c e.: ene.cour:n er :nc these tests cnly if the he'.tLine shcil th.cf.r.e::. :r.c:

heacup and cooldown rates are used to dc; err.ine :he N0r do not exceed 15'F per hift frc: Ficire

3. '~ - 2.

hour.

5 vnen the rcanor ve::c heae 5.

The reactor vessel head bolting t:1:in. : ul: are te.:1.ne; cn2 studs may be partially the rea:ter is in u ::.: ::..;:.

tensioned (four sequences of

tien, hc res:ter ves:u. :h d_

the seating pass) provided the tc ; crc rc 10: eta'e4 : '.:V studs and flange materials are the head f'.ange ch C :e pe.

"^~ C " ~' D # C O #d"d*

above 70*F. Before leading the flanges any more, the vessel

.,,,,a.,.....,.p y

o.

lange an nea :.ange must e c..,. c.. u... a...... _.,,..,

greater than 100'?, and must te pernure er ne 7e:.::,7 ec..,.

remain above 100*F while under ein; in the oper, ting an: i s.f. e full tension.

loop: * ' '

'e pe-:ane. ly 6.

The pump in an idle recircula-tion loop shall not be started 7.

p,. 4

..o o.. m ne a s....

..e

....m.

unless the temperatures of the tien p,.r p, the reac:cr e...:.n; coolant within the idle and tc=peratu c: in the Jc.e :..d.n operating recirculation loops the bance head drun :h 11 te are within 50*F of each other.

c:q cred and pe:c.anently logge:.

7.

The reactor recir:ulation pumps shall not be started unless the coolant temperatures between the deze and the bottom head drain are within 145'?.

i I

175 AmendmentNo,h,66 l

t

.:IMf7(NG CONDf780N5 TCR OPEPATICM SU DVEI LI.A NCT. M PO t !? RCHT.NTS 3.6 PR ! MA RY SYSTIN 200NoA.RY

- 4.6 P FSARY SY STI.'4 BOUNCARY u.

Once each refueling cycle, 4

f the requirements of 3.6.H.1 and a represe.ntative sa.ple.

3.6.H.3 cannot be of 10 snubbers or met, an orderly approximately 10% of shutdown sna11 be the snubbers, whichever initiated and the is less, shall be reactor shall be in a functionally tested for c I,d shutd m operability includine condition within 36 verification of proper hours.

pisten movement, lock um and bleed. For each unit 5.

If a snuS5er is and subsequent unit found determined to be in-inoperable, an additional operable while the 10% or ten snubbers shall reac ce is in :ne be so tested until no shutdown or refuel more failures are found mode, the snubcer or all units have been shall be made tested.

Snubbers of operable or replaced rated capacity greater prior to reactor than 50,000 lb neec not

startup, be functionally,.ested, 5.

Snubbers may be added to safety-related systems without prfor license amendment to Table 3.5.H provided that a revision to Table 3.6.H is included with a sucsecuent if cense amendmen recuest.

a 1S n

4

,-6

..r-r ev5e in

=-a' t--

4-

+ - -

e --

-a w

--t-we-c-

W T-w-

r

-T+r*+

-e

Figure 3.6-1 i

C.tv< 0t Min i:..u.n t empe r4. t u '.

for pressure tests such as required by Section XI.

1200 f

2 carve se f 'f 9-25-80 P.inimum ceeperatura for mechanical he.t up or couldcun following nuclear InOu" n hu td e*,:n.

(

.Cu.rve #3 M in itnur:. t e.pe ra t,. re.

l ft: core e eratica J

l O

! c r i - f ca 1:,.')

ge<%

Include:,, ad:!!: lor.,

v-j nargin-required.oe

5 l

10CFh50 A. pron.!!x 0, u

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P.1 r.

I','

A. 2. C.

c 2C

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l' s

t:i t e n 5i

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~

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'I 30*F.

This d ft.11' j

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2 u.;.:.! chru f. 0 F. TTY.

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M: ;P a p:H:vmPs

('T) 188 Amendment No. 66 I

l

ace:u

![o jeg UNITED STATES f'i NUCLEAR REGULATORY COMMISSION

,e p/.. y WASHINGTON. D. C. 20555 0

c

%,, '..... f TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS rERRY NUCt. EAR 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 16, 1980, comply with the standards and requirements of the Atomic Energy Act of 19 5 t., as amended (the Act), and the Commission's 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 regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized oy this amendment can be conducced without endangering the health and safety of the public, and (ii) that such activities will be

oncucted in compliance with the Commission's regulations; D.

he issuance of this amendment will not be inimical to the common

sfense 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Speci-fica-ions as indicated in the attachment to this license amendment and part;raph 2.C(2) of Facility License No. DPR-52 is hereby amended to rea d as follows:

(2) Technical Specifications 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.

, 3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

,e

'2 A

Thomas

. Ippolito, Chief Operat~ing Reactors Branch #2 Division of Licensing Attac hment:

Changes to the Technical Specifications' Date of Issuance:

February 6,1981 a

E 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 ET/62 63/6T 77)78--

175 T57/188 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

INTRODUC"' ION This document presents the tech.nical specificaticos fer the Brevns Fery Nucles.r Plant Unit 2 enly, s

1 l

l l

i

, NN

.0,.3 $%

1 3

A=end:er.1

'0-

s.

t

+

1.0 DEFINITIONS The succeeding frequently used terms ara explici:17 defined so that a uniform interpretation of the specifications may be achieved.

A.

Saf etv Limit - The safety limits are limits below which the reason-able maintenance of the cladding and primary systems are assured.

Exceeding such a limit requires unit shutdown and review by the.

l 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 subject to regulatory review.

B.

Limiting Safety Svstem Setting (LSSS) - The limiting safety system

- setting are settings on instrumentation which initiate the automatic protec:ive action at a level such that the safety limits will not be exceeded. The region between the saf ety limit a2d these settings represent margin with normal operation lying below these settings.. 'The margin has been established so that with proper overstion of the instrumentation the saf ety. limits will-never be exceeded.

The li=1:ing conditicus C.

T 4=4 ting Conditions for Overation (LCO) for opera:1on specify the inimum acceptable levels of system performance necessary to assure safe startup and operation of the'-

facility. When these conditions are sec, the plant can be opera:ed safely and abnormal situations can be safely centro 11ed.

1.

In the event a t4-d:ing Condition for Operation and/or associated require =ents cannot be satisifed because of circumstances in excess of those addressed in the specifi-cation, the un.d.: shall be placed in a: leas: Ect Standby 1

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months unless correc:ive measures are ecmple:ed that permit operation under the permissible discovery or until the reactor is placed in an opera:1onal. condition in which the specification is not appli:able. Exceptions to these requiremen:s shall be sta:ed in the individual specifications.

his provides actions to be takan for circumstances set direc:1y p;ovided for in :he specifications and where occurrence would viola:e :he intent of the a cification. For example if a specif t:ation calls for two systems (or subsystems) o be operable and provides for explicit requirements if cce system (or subsystem) is

. inoperable, then if both systems (or subsystems) are inoperable the uni: is to be in at least Hot Standby in e

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shu:down within the followina 30 bcuirs if the inoperable condition is net corrected.

b I

I j

.l.

V Amendment No. 62 l

i F

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i 1.0 DDTNIyIONS (centinued)_

2.

When a system, subsystem, train, component c device is determined to be inoperable solely because its ensite pows: sourca is ineperable, or solelybecause its offsite power source is inoperable, 1: may be considered operable' for the purpose of satisfying the requirements of its applicable Limiting Condition Tor Operation, provided:

(1)- 1:s cc :esponding offsite or diesel power source is operablei and (2) all of its redundan system (s), subsystem (s), train (s),

component (s) and device (s). are operable, or. likewise satisfy these requirements, Unless, both conditions _(1) and (2) are sa:1sfied, the unit shall be placed in at leas: Bot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in a: less: Cold Shutdown within the following 30 hoc:s. This is not applicable if the unit is already in Cold shutdown or Refueling. "his provision describes what additional condi: ions' must be satisfied to pe=1: operation to con:1sua consistant with the specifications for power seurces, when an offsite or onsi:e power source is ne: operable. I: specift: ally s

prohibits operatics when one division is inopcable because its offsite or diesel power source is inoperable and a sys:e=,

subs 7 stem, ::ain, cecponen: or device in ano-der division is inoperable f ar a=other reason. "his provision permits the

equirements associated vi h individual systems, subsysta=s,.

tradas, components or devices to be consisten with the requireme::s of the associa:ed electrical power source.

It allevs operation to be governed by the time 1' ': of :he requirements associated vi:h the t %d:ing Condizion 73: Operation for the offsite c diesel power source, no the individual requirements for each system, subsyste=, ::ain, ce=ponent er device that is dete==ined to be inoperable solely because of the inoperability of its offsite c diesel power source.

D.

DC.:..:.a o

e i

la Aren'ement No. 62

.?

1.0 DEFINITIONS (cont'd)

E.

Doerable '- Ooerabili:1 - A system, subsystem, train, component, or device shall be Operable or have operability when it is capable of perfor=ing 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 vatar, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

F.

Operating - Operating means that'a system or component is performing i

1:n intended functions in its required manner.

G.

T =adiate - Immediate means that the required action will be initiated a.s soon as practicable considering the safe operation of t

the unit and :he importance of the required action.

E.

Reactor power Ooeration - Reactor power operation is any opera:icu with the mode switch in the "Startup" or "Run" position with the reactor critical and abeve 1*. rated pcver.

I.

Hot Standbv Conditien - Eo: standby condi:1ca neans o'peration w1:h coolant te=perature grea:e than 212*F, system pressure less than 1055 psig, the sain s:ea: 1solatien valves closed and the mode switch in the 5:artup/ Hot Standby posi:1on.

J.

Cold Condi:1en - Reactor coolant te=perature equal to or less :han 212*F.

K.

Hot Shutdet.a - The reactor is in the shu:down mode and the reae:o coolan: tenpera ure grea:er than 212*?.

L.

Cold Shutdown

~he reactor is in the shu:devn node and the reacto:

coolant te=perature equal to or less than 212*F.

M.

Mede of Otera: ion - A reactor =ede svi::h selects the proper interlocks fo: :ne opera:1onal status of :he uni:. The following are the = odes and in:erlocks provided:

1.

Star:ue/Ho: Standbv Mode - In this mode the reactor protec:Los scram ::1ps initia:ed by condenser low vacuum and main steam line isolatien valve colsure, are bypassed when reactor pressure is less than 1055 psig, the reactor protection systen is energized vich IRM neutron monitoring system trip, the ApKM 13*. high flux trip, and con:rol red withdrawai interlocks is service. This is of ten referred :o as jus:

.Startup Mode. ~his is intended to inply the startup/ Hot Standby position of the mode switch.

3 Amendment No. 62

1.0 DEFINITIoH5_ (Cont'd) 2.

Run Mode - In thin mode the reactor system pressure is at or above 850 psig and the reactor protection system, is energi:ed with APRM protection (excluding the 151 high flux trip) and R3M interlocks in service.

Placing the mode switch to the shutdown posi-3.

Shutdevn Mode tion initiates a reactor scram a,nd power to the control tod drives is removed. After a simrt time period (about 10 sec),.

the scram signal is removed allowing a scram reset and restoring the normal valve lineup in the control rod drive hydraulic sys-tem; also, the main steen line isolation scram and main con-densar low vacuum scram are bypassed if reactor vessel pressure is below 1055 pais.

4.

Refuel Mode - With the mode switch in the refuel position' inter-

, locks are catablished so that one cuntrol rod eniv may be with-drawn when the Source Range Monitor indicate at least 3 cys and the refueling crane is not over the reactor; also, the nain steam line isolation scram and = sin enndenser low vacuum scras are bypassed if reactor vessel pressure is below 1055 psig. If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.

a r' actor power of H.

Rat e d F eve r - Ra t ed power re f e r s to operation at e

3,293 MVt; this is also termed 100 percent pvvce and is the saximun power level authorized by the operating license.

Rated steam flow, raced coolant flov, rated oeutron flax, and ratkd ndelear system pressure refer to the values of these parameters when the reactor is at rated power. Design power, the power to which the saf ety analysis applies, corresponds to 3440 MW:.l O.

Primarv Containnent Integritv - Primary containment -integrit7-= mans that the drywell and pressure suppression chenber sie' intact' add all of the following conditions are natisfied:

1.

All non-automatic contain=ent isolation valves on lines connected to the reactor coolant systes or containment which are not required to'be open during accident conditions are closed. These valves may be opened to perf or= necessary operational activities.

2.

At least one door in each dirlock is closed and sealed.

3.

All automatic containment isolation valves are operabis or deacci-(

vated in ene isolated position.

4 All blind flangee and =anways are closed, o

P.

Secendarv Containment Inteertry - Secondary containment isteE ri:7 means that the rasctor building is intact and the folleving condi-tions ars met:

l l

l

. m

N c

8 TABLE 3.1.A z

REACTOR PROTECTic.1 SYSTDt (SCRAM) 1 iSTR fMctTAT104 REQUIRDENT P

Q Htn. No.

of

~

i cn operable Inst.

Hades in Which Function Clunucts Hust Se Operable res Tsip Shut-Startup/ Hot s yi. t e m (1)

Ty p Function T,gtp I.evel Setting dow Rafuel(7)

Standby Run 3,,g,n{gg-_

1 Mode Switch in Shutdown y

X X

X l'A U

1 Hanuel Scraa X

X g

X I*A IFM (16) 3 11tsh Fium 1 $ Q Nejnd!ceted

.g g gg gj 3

Inuperative g

g (3) 1,4 AFP t (16) 2 High Fluu See Spec. 2.1.A.1 2

utch rtum i 15% rated power x

1.A or 1.8 X(21)

X(17).

(13)

I*A I*I 2

tauperattwe 01)

X(21) x(17)

X 1.A er 1.5 3

i;ounscale 3 1 Ind*cated on Scale (11)

(11)

X(12)

I*^ *r I B 2

litre Neactor Pressure j 1035 pe t s g({0)

X X

IA i

2 H!r.h arwell i

12.5 psig Pressure (14)

X(8)

X(4) x 1.A 2

Reactur Lov unter

> 538" above vessel seco Level (14)

X X

X lA

~

2 tilgh ustar Level in Scram Discharge Tank

-< 50 Cellone x

X(2) x-g 1.A

1 U.

- C.

C.

U.

V.

n w

C C

b b

b

'L L

L j

C O

O O

o C

v Cn er. -

{

s.o o=

l R

w

)

l i

e C

s=m.

o=.m m

m go m

n r.

.Ch, 2

aur sur eu CC

=

w we

=a x

x x

C

J

O g

o U L*

%2' em C-C. t 3.O P :

w e

6.c :

e

- w

=s m

go

=

m 8

6 m s.*

I

=4 sur aq*

aur M

C*

Lt 2 Ul @ B w

we we w

we we l

V C" '

J M

x x

O A

- =

.=

i 3O CD C

^

+s=

=J w

P%

e e3

.a=

C*.

O

~:~

3 w-

===

w e===

,="" ^

CC e

e==

y 1

Cr. l

.v

.v,

.M, Cs O

M w

-a x

x X

M x

x x

m

3 C'

e L

3 eJ G G

C.

Cd e>

3 C'

ei m-O F

o e8 L

6 M L

3 C.,

C

J 3

3 3

C a*

e G

e u

C O

E 'O O

C

- * = = =

U e'

2C w-bl

- C U

LN

%e

% C CC C

O C Gi U

C'n CD

=

9 "C CC

~

E c G

6 3 m

J m

- - e C

e e

L C

C.

C O c 6

C

d CD

=

2: b J

V C

C v

Cn c::

L C

C *=

C O

W m

Mw e:C H

C

J N

mv C. U m

Vl

3) 1 vl Al Vl Al V lCD 0

L C

9 C

~J s

~

C a

e.J eti C

6 V

t' 3

C 2

C O

C D

D C C

,J

.Cn c==

CM >

=

>C g-L C

.J U

ed @

G C

0 Ae e

o C

C==

3 C

O C C.

L b

J E o

- ac-O 6.

+

., l L sJ C

ed ed

J

.,,J em Ca.

3 C

w C C 6 0 w-i*, e C

/

U C

- Ca.

C C

~

L 4 C bJ VV U 6

V eC G -~

k D

U.*C y

sd V

?J G C

Ow 3 Q %

C sd 3

C E M *s i

M C L C

..C CJ e

..C e O

m j

y

.=

e

/5 e

.O

.O # ".'

O e O 2 C~

C.>

4 o

L L e 6

s. o 6 v 3

)

. e-c.

e1 J=

=

3 C, 0 3

6.

= c ec E>

>V

>=

lJ Ha t== >

E C::

W

-I e.c

-l c.

C e

.=

=

.O a u c.

.O n

=-

d=

=?

N v

N N

>w m.

C C C L

= .

.as f '.;.C O b.i C

C"', = J 2,. V.

14

,,,-v~,.

-w.,

.-..o%,-

..ww.. w, e-w e. w.,-..,,.._....._.e,.,rm-w,7,v.-r e--,

.v

.ew.,,-.--y m-w-*

r-we e

'n*--

6.

Channel sharad by F.?S'and Tri-a7 Contatsmant is Reactor Vesec1 Isolation

- (:en t rol S ys t em.

A cha. vel failure s.ay be a channel f ailure in each syst em.

7 A i ra les. l..

con sider mi a :rtp.nyates.

' (,

Two out of three SCTS trains required.

A' f ailure of more tha.n one vill require action A and F.

e 9.

~here is only one trip sys:e= with auto transf er to :ve power sources.

i e

P Jl 4

01 l

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

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

...,-.._.._,.-..._._.-~~...,_m...--,.-.

-.-.--,,--J

i i

4 N

a TL'1LE 3.2.0 M

TH AT lit t11 AILS OR CAIGROLS l'.lf. CORE AND CotrtAIM(Dft COOLING SYSinf 5 Its!! ttntDir Alldet i

=

?

ttel m a W 3.

1 D

Cperable for function Trip triel Setting Action Res e rk s 1

Trse Syp (1) cn A

1. Below trip setting inittsted EFCI.

14 2

Iustrument Channel -

> 970', above ve s s el sero.

Besctor im Unter Exvel 1 70*above veneet acro.

A

1. Hultiplier relays. initiate RCIC.

4 4

1 Instrur.at Chenac1 -

a. actor Lov unter Ic..t 4

1 2

Instauncat Channel -

> lis** etsove vessel seto.

A

1. Selow tt!p setting inttistes Cst.

llottlplier relays initiate 1.FCI.

deactor Low Unter 14 vel i

(Lis-1-35A-0. su #1) occident stanal (15).

2(It) le..:n aent omanel -

> 37a above vessel sero.

A

1. Below trip settfoge in conjunctica intth dryvell high pressure. Iow Keector low Ueter Level water level perateelve. 110 sec. dely (LIS-I-18 A-D. 54 # 2) tie:cr and CSS or EUA pansp runnlag.

1stristes ADS.

i l

1(16)

Instem. cat Cheunel -

> 544 strove veneel acro.

A

1. Below trip setting petitssive for Initiating signets on ADS.

teactor law Veter Levet Pcratestie (LIS-3-Ist t j

18), su it) i 1

Instrment Channel -

1 312 3/16 above vessel sero.

A

1. Beler trip settleg prevents inadver-Reactor Imv unter tavel (2/3 care height) tent operatlog of contatament spray (LIT 5-3-31 & 62. 3U #L) during accident con fittoa.

t A

1. Below trip settima prevente tasaver-2 Instrin.eur Channel -

le pe 2.5 pelg tent operation of conteinnent syrsy j

l Dr m it tilah fressore frS-64-38 E-u) during accideet'condittons, i

1 I

1

cu a

n M

a rt IABLE ).2.5 (Continued) 2

?

11ninum No.

"_ca operable ter Bla 3s (t) function Trip Level setting Actton Renarks os

1. Above trip setting in conjunction with N

A 2

Instrument Channel -

2.5 psig Dryvell lit h Pressure low reactor pressure inittstes C55.

~

t (rs-64-58 A-D, su J2)

Multtatter relays initiate lirct.

2. tutttpiter relay from CSS initiates acclient signal.(15)

A

1. Be!.rs trip setting trips rec tccula-2 Instrument Channel -

970**at.ove ve s sel ae ro Scactor tow water Level tion pumps (Ls-3-16A, S, C, D) e 2

Inst rument CIssuuel e1120 psig A

1. Above trip setting trips rec tacule-Beactor lii ti Frcssure (t an pureps t

o.

(Ps-s-JU4 A, B, C. 0) u

1. Above trip setting in conjunction ph l

2 Instrument Channel -

e 2.5 putg A

Dryvcll litati Pressure low reactor pressure initiates trCa.

4 (rs-64-18A-D, su #1) 2(16)

Instrument O$asuiel -

,c.2.5 palg A

1. Above trip settlag in conjunctica with-bryvell tilgh Frcasure low reactor water Icvel, drywell hidt (Ps-6&-MA-D) pressure,120 sec. delay tinct and CSS or RJtA puerp running, initiates AD5.

~:elow tr*r settlnc p4 missive t'r IIeninE

.2 Instrument Chanact -

LW psts i 13 A

1.

Reactor tow Fressure CSS a nd I.!-CI a.ireis s i r. val re s.

(PS-3-74 A & B, su #2)

(rs-6a-95, su #2)

(Ps-63-S6, su #2) 2 Instrument Channel -

230 psig i 13 A

1. Re:trculation discherge valve Re. actor low raessure ac:uation.

(Ps-1-14A &

t, su #1)

(rs 6a-95, su 21)

(Ps-65-96, su #1)

e I

I Ant.I 1.2.A (Cun t ini.cd )

4

,1tninen tin.

Operable Per f

Trip Sy d Q Functton

_ Trip _ Level Setting Actton Remarks 1

Instrument Ch4nnel -

100 pelg i 15 A

I-Below trip setting in conjianction with Reactor low Pressure cont a i nd.neu t isolation signal.1nd both.

l (FS-68-1) & 94. SW f1) suction valves open will close 311R (LPC1) admission valves.

2 Core Sp:iv Auto Sequencing 6 < t < G sec s.

B 1.

With diesel power i

Timers (3) 2.

One per motor 2

1.PCI Auto Sequencing 0<t<1 nec.

R 1.

Utth diesel power 1teers (3) 2.

One per antor en u

1 RitRSV.O, 81, C3, and 91 1 ) < t_< 15 s ec.

A 1.

With diesel power Timers 2.

One per pump 2

Core Spray and 1.PCI Auto O<t<1 sec.

8 1.

With normal power Sequencing Timers (6) 6,< t j 8 J e c.

2.

One per CSS motor 12 < t < 16 sec.

h DO P"'

E "0 EOF g g. g '2 g r, s e c,

1 RHR5u.' I. Bl. C1, and D1 21<t,f29 sec.

A 1.

Ut t h noruut1 power Timers 2.

One per pump

~

v-a

i 4

Y t

~

JAn!.E 3.2.E lilSTRU!!ElfrAT10tl TilAT l'ONiinS LEAKAGE INTO DRYWELL Sys t e:n (7)

Setpoints Action Remarks 7

(1) 1.

Used to deterutne identifiable reactor Equipment Drain Flow Integrator N/A

. coolant leakage.

2.

Considered part of sump system, Sump Fill Rate l

Timer

>20.1 min.

t Stnap Pmmp Out itate Timer 113.4 min.

\\

(1) 1.

Used to determine unidentifiable i

Floor Drain l

Flou Integrator.

N/A reactor coolant leakage.

Surup Fill Rate 2.

Considered part of sump system.

i Tiner

>B0.4 min.

O' Sunp Puop Out 1 9 min.

8 Itate Timer l

Dryuell Air Sampling Gas and 3 x Average (3)

Particulate

Background

}

I

^

1 Il0TES:

\\

j (1) Whenever a system la required to be operable, there shall be one operable systen either automatic. or mancal, i

or the action required in Section 3.6.C.2 sht11 be taken.

(2) An alternate system to determine the leakage flou is a aanual syste:m whereby the time between aump pusp starts is monitored. The time interval util detemine the leakage flow because the voltsee of the sump util be knoun.

(3) Upon receipt of alarm, inanediate action will be taken to confirm the alarm and ' assess the possibility-of

]

increased leakage.

j i

- - = -.

g TABLE 3.2.F SURVEILLANCE 11tSTRUllENTATI0li R

~

Bo Hintman i of z

operable Instrument Type Indicattoa

?

Chancele fostrument i Instruacnt and Range Notes 2

L1-3-46 A Reactor Vater Level Indicator -107.5" te (1) (2) (3)

LI-3-46 R

+107.5" 2

FI-3-54 Reactor Fressure Indicator 0-1200 pets (1) (2) (3)

FI-3-61 i

2 FR-64-50 Drywell Pressure Recorder 0-80 pela (1) (2) (3)

{

FI-64-67 Indicator 0-80 pela 2

T1-6 4-5 2 Dryve11 Tepperature Recorder. Indicator (1) (2) (3) 3 l

11-64-52 0-400*F 1

TR-6 4-5 2 Suppression Chamber Air Recorder G-400*F (1) (2) (3)

Tcoperature ym 2

TI-64-55 Suppression Chasber Water Indicator. 0-400*F (1) (2) (3)

TIS-64-55 To:7:ratura 2

LI-64-54 A Suppreeston Chanber Vater Ind2cator -25" to (1) (2) (3)

Lt-64-66 Level 425" 1

NA Control Kod Position 67 Indicating

)

t Lights

)

1 ItA Heutron Ifonitoring SRH IRM, LFKH

)

(1) (2) (3) (4 i

0 to 1001 power) 1 FS-64-67 Drywell Fresoura Alarm at 35 psig )

)

1 TR-64-52 and Dryvell Temperature and Alarm if temp.

)

FS-64-58 B and Frcosure and Tince

> 281'r and

)

(1) (2) (3) (4l 15-64-67 pressure

2.5 paid after 30 minute

)

Jelay) 1 la l!A.?A CAI) tank "A" level In,81catar O to 10Pi 1) 1 I.T-U4-l iA call tank "c" level inlicat or O to 100f 1)

L:!c-:r, C0f J-- 0 !S FOR OPE.N :C:7 SL7VEILw:*E FE:JJ Ect--

3.6.A hr rg and Pre::urizatien

!. 6.A 'Ther-rd aM Pressurization Limitat;on:

Lic.th.wnn 3

During heatup by non-nuclea-3 7c:t spec <:ena repre:enttn,,. ne means, except when the vessel ree.eter vencel, tu e e g.ana meld is vented or as indicated in heat affected sc.no mctc1 c.nall tse 3.6.A.4, cooldown following in:te.n ed in the resetor ye::e1 nuclear shutdown on low-level ad,'acent to the ve::c;,

.e ;1 v.

physics tests, the reactor the core =idplane level..

yne vessel temperatures shall be nunber and type of : pen;nen; at or above the temperatures vill he in accordance.c;ta I of curve #2 of figure 3.6.1.

CP rt I'~5 ~"1011%

h e :peci= ens che11 meet tne intent of STM I 103-70 Cumples shedi de with.

4.

The reactor vessel shell dr

't onc-fou:~ h and arte-temperatures during inservice four'~~.- a e.~..< e _'.<.*e.

hydrostatie or leak testing a,.

,.ie...._.,..,,x

.r,,,,s

.,.,,. ~ e.,,,

shall be at or above t,e n

..,.ed4...., e e s..... v..,,... _

temperatures shown on curve a.,, e.. o. n. e

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

1 of figure 3. 6-1.

The w.;l at the core nid;;:.ne icve.

. applicability of this

- ;c. g re., 3,6 ra1 1,e re e..g 3.,

curve to these tes:s is te:. ed ett-ing t.he fir;,;.

.due. tng extended to non-nuclear cu:cge :,o experimen.u ;y vur e,e heatup and ambien: loss the calcu'.sted v 1ue:; ol n :*... on cooldown associated wi:h flu ":e at one-fourth of the these t wts only if the b e'.

- wc11 thice.ne::. :nt.,

heatup a.! cocidown rates are used te detenine.ne :c;;

do not exceed 15'? per hif; frc: FiCdre 3d-2.

hour.

5 ner. the :cc::cr ver::1 heac tc;-ing ::ud: are ten:::ne; c..;

5.

The reactor vessel head b:lting

-he res:::r i: in e. c: : ::,r studs may be partially tien, the rea: tor v" tensioned (four sequences of

c. ;0rt ure i. :cdiately : 10.-

the seating pass) provided :he the nCad flanse :hcil :: Fr-studs and flange ma:erials are

---" ~.'

C e e *-s*e--

a above 70*F. Before 1:adin: the t.

h i:r to and during ::c c..

.c..

,,,,..c,..rtuo er rianges any more, t.ne vesse.

flange and head flange must be te..perature of the rea::.:r ecf..

greater than 100*F, and mus:

e.n: in the operating an: :C. e remain above 100*F while under iceps shr,

de per= ne.tly full tension.

legged.

6.

The pump in an 1.tle recircula-P,,...o s:cr,.n,,,

a,,..,,,,,

.s.

tion loop shall not de star:ed

..,......,. s. e c ac. c

,....s.,.

r.. e...

unless the temperatures c:. the te perg u,c3 in 3e ecce ug ;n coolant within the idle und the be:. cm head dra:n :hr.11 1e operating recirculation 1: ops cocpered and pe r.ar.cn ;y logge;,

are within 50*F of each c-her.

7.

.The reactor recirculation pumps shall not be started unless the coolant temperatures be: ween i

the dome and the bot:cm head drain are within l'5'F.

I Anendnen: No. 50 175 rt Amendment No. h, 62

,,; y :T f Nt; CrJt4D TTTONS P'On CPEp.ATICN.

CURVE LLANCI' nFOUTRC?iENTS 3.6 PP : MA PY S YSTI.M 20UN 0tJy u.6 P R-A RY sY sT!:4 BoUNCMY c.

If tne requirements 4

Once each refueling cycle, of 3.6.H.1 and a represe.ntative sampie 3.6.H.3 cannot be of 10 snubbers or met, an orderly approximately 10': of shutdown snall be the snubbers, whichever initiated and the is less, shall be reactor'shall be in a functionally tested for cold shutdcwn condition within 36 operability including hours.

verification of proper piston movement, lock up 5'

If a snuS5er is and bleed.

For each unft detarmined to be in-and subsequent unit found operable white the inoperable, an additional reactor fr in t5e

10. or ten snubbers shall shutwcwn or refuel be so tested until no mode, the snubber mere failures are found shall be mace or all mts have been operable or replaced tested. Snubbers of P"I#",, o reactor rated capacity greater s ta r. ',

than 50,000 lb need not be functionally tested, f.

Snubbers may be added

safe *y-rela *ed sys te:.s wi *h0U

prf0r license amencren: *:

Iable 3.6.H Or0 Viced

nat a revis'0E 00 Table 3.6.H is incluce0 witn a su:secuen: 'icense amencmen re uest.

157

Figure 3.6-1 Curve #1 Mini =n terpe r.tu. -

for pressurc tests stich as required by Section XI.

12OP-I 2

3 f

Carve 02 f

l l 9-23-80

..ini:nue tetspc m.ar) for n:echanical het.t itp or couldown following nuclear 1000 9hutdown.

l 2JflS ')3 Minirnua. ter.perstu ri for cor,e entration cy (critir alit y) 800---

I'

- + j In c. l ud e t. addi:1:w'

[

I margiti re'qu i r. d.. >

,1 G

10CFR$3 Appen.:i/ G.

Par.

I'.' A. 2. C.

~

j 2

603 -

.t:e t. e s.

~

$ i-l Thcnc ctirves ar+

.3 '

l shifted W T to th

'y r:y,ht of Llw origir :

C net of curves to y

ind ude a t.kr..,,_ of 30"F.

This sliYf: w :.1 '

f 4CO-l

,/

/

allev thesc curv.ni

'.c 2

ut.ed thru 4.0 ETIT.

/

o

/

l

/

200

/

I l

/

l

/

r i

+ -

, - P.0t.7 UP tem I.s ATUTC.,

C - ---

l l

O 10 0 20u 200 400 P : ':!".'t. Tu!P'.:U.n P E (c..

r>

133 Amendment No. 62

[pcro jo, UNITED STATES y g, g

NUCLEAR REGULATORY COMMISSION

,- E WASHINGTON, D. C. 20555

]

e l

%..%.'...f TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 38 License No. DPO S8 1.

The Nuc1 ear Regulatory Commission (the Commission) has found that:

A.

The applications for amendment by Tennessee Valley Authority (the licensee) dated June 13, 1980 and October 16, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; 3.

The 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) 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 Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and securicy or to tne health and safe y of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51

's ' the Comission's regulations and all applicable requirements

mve been satisfied.

2.

Accordingly, the licer,se is amended by changes to the Technical Speci-

" cations as indicated in' the attachment to this license amendment and paragraph 2.C(2) of Facility License No. DPR-68 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 38, are hereby incorporated in the license.

The licensee shall cperate the facility in accordance with f,

the Technical Specifications.

2-3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Q

,Q

/

Thomas A. Ippolito, Chief Operat',ng Reactors Branch #2 Division of Licensing Attac hment:

Changes to the Technical Specifications Date of Issuance: - February 6,1981 i

<a--,,

e--

y-g--,-,

y

ATTACHMENT TO LICENSE AMENDMENT NO. 38 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages:

2 3

32 65 81

,105 og 2'

Add the following new page:

2a The marginal lines on the above pages indicate the area being changed.

y+..

g,

.. ~

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

a

1.0 DEFINITIONS

..The succeeding frequently used terms are explicitly defined so that a uniform interpretatics of the specifications may be l

achieved.

A.

Safetv Limit Ihe safety limits are limits below which the reasonable tsaintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Aegulatory Commission before resunption of unit oper.stion.

Operation beyond such a limit may not in isself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

B.

Limiting Safety System Setting (LSSS)- Ths limiting safety system setting are settings on instrumentation which initiate the automatic protective action at a Icvel such that the safety limits will not be exceeded.

The region between the safety limit and these settings represent margin with normal operar:fsn lying below -hese settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

C.

Limitine conditions for Ooeration (LCol - 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 saf ely controlled.

1.

In the event a Limiting Condition and/or associated requireraents cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least Hot Standby within 6 hourt and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months unless corrective measures are completed that permit operation under the permisstble discovery or until the reactor is placed in an operacional condition in which tha specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications. This provides action to be taken for circumstances not directly provided for in the specifications and whose occurrence would violate the intent of the specification. For example, if a specifi-cation calls for two systems (or subsystems) to be operable and provides for explicit requirements if one system (or sub-systems) 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months if the operable condition is not corrected.

Amendment No. 38

'l.0 DETINITIONS (cont'd) 2-When a system, subsystem, train, component or device is determined to be inoperable solely because its onsite power source is inpperable, or solely because its offsite power source is inoperable, it may be considered opererable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding offsite or diesel power source is operable and (2) all of its redundant system (s), subsystem (s),

train (s), component (s) and device (s) are operable, or likewise satisfy these requirements. - Unless both conditions (1) sud (2) are satisfied, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This is not applicable if the uni: is already in Cold Shutdown or Refueling. This provision describes what additional cond1:1cus must be sa:isfied to permi operation to continue censistant with the specifica:1ons for power sources, when offsite or onsite power sources are not operable. It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperable and,a system, subsystem, train, component or device in anothet division is inoperable for another reason. This provision permits the requirements associa:ed with individual systems, subsystems, trm %, components or devices to be consis:ent with the requiremen:s of the associated electrical power source. It allows operation to be governed by the time limits of the requirements associated with the Limi:ing Condition for Opera:1cn for the of fsite or diesel power source, not the individual requirements for each system, subsystem, train, ce=ponent or device that is deternised to be inoperable solely because of the inoperability of its offsits or diesel power source.

D.

DEI.ZTED E.

Deerable - Ocerabilitv - A system, subsysten, : rain, component or device shall be operable or have operabili:7 when it is capable of performing its specified fune:1on(s). I=plicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical pcwer sources, cooling or seal water lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform 1:a functien(s) are also capable of perferning their related supper:

function (s).

F.

Deerating - Operating means that a system or componen: is performing 1:s intended functicus in its required ranner.

O.

I:znediate - Immediate means tha: the required action will be initia:ed as soon as practit.able considering the safe cperation of the unit and and importance of :he required ac:1on.

H.

Reacter ? ewer Ooeratien - Reac:or power operation is any operation wi:h :na mode sw1:ch in the "Startup" or "Run" posizion with the reactor critical and above 1:: rated power.

2a Amendment No.

M

,,.ns.

- -... +... - - -,

I.

Ho g_ St an d by Ccndition - Hot standby condition means operation with coolant temperature greater than 212*F, system. pressure less than 1055 reig, 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 2120F.

K.

Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 2120F.

L.

Cold shutdown - The reactor is in the shutdown mode and the-reactor coolant temperature equal to or less than 2120F.

3 M.

Mode of coeration -~A reactor mode switch selects the proper interlocks for the operational status of the unit.

The following are the modes and interlocks provided:

1.

Startue/ Hot Standbv Mode - In this mode the reactor protection scram trips initiated by condenser low vacuum and main steam line isolation valve closure, are bypassed when reactor pressure is less than 1055 psig, the reactor protection system is energized with IRM neutron monitoring system trip, the APRM 15% high flux trip, and control rod withdrawal interlocks in service.

This is often ref erred to as just Startup Mode.

This is intended to imply the Startup/ Hot Standby position of the mode switch.

2.

Run Mode

. In this mode the reactor system pressure is at or above 850 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and R3M interlocks in service.

3.

Shutdown Mode - Placi.74 the mode switch to the shutdown position initiates a reactor scram and power to the control rod drives is removed.

After a short time period (abo ut 10 sec), the ceram signal is removed allowing a scram reset and restoring the normal valve lineup in the centrol rod drive hydraulic system; also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is below 1055 psig, j

i i

l 3

Amend:en No.

3M

lY fD Tansc 3.t.n atACTom FatoitrT1*Me erstrit (t,O AH) IltsTRtettNTAT1010 St001BffttNT u

f u.

s" Min. Itc.

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operable twees in wblete Function i

Inst.

_._13 11 Ise opstable N

Shut-Startup/ Rot o.mpnet.

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jyq1g*J1, Idp rus.ction Jg[p_f.evel Se11[n3 pm3 ge t.je LH1 gt a_e..tbe gi N'

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3 X

X 1.A 1

planual Scran 3pH (36) 3 ulyta Flus E $20/125 Indicated E(22)

E (22)

E (S) 1.A on scale I

I (S) 1.A 1

3 Inoposative 4

AthM (141 E

3.A or 1.3 2

sigh rlum see spec. 2. 3.A.1 i

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+2 2

utqh clus

$ ISE rated power E (21) 3(17)

(13) 3.A or 1.3 u

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& 2 Indicated on scale 31 h pesctor treenwre i 1955 pelg E(to)

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seactor 8ow water tevel (34) t 5 34* stowe vessel sero X

X X

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algh water 14 e1 in plect.aage Tank n 30 Gallone R

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Main steam Line Isole-tion vatwo closure i 145 Valve Closure X(3)(48 2(3) (6)

X (8) 1.8L er 1.C 2

Turbine cent. volve epon trip et the test tant cloeuse acting molenoid valece E(4)

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2 Inst sument thansw! -

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1.

setow trip setting prevents i

ps ywell tingh tre =sure inadvestent operation of gr$-69-SS t-up containment mos ay duz ang accident conditions.

i

~ t.

tm e trip setti g in L

J Ip a t r ument Ch anne l -

s2.5 psig 2

Da ywell tilgh Ps es sus e conjunction with low reactsr (r5-64-58 A-D, SW $ 23 gsessure initiates CSS.

etulti pli es relsys initiate afPCs.

2.

uuttipites relay f rom Css.

I initiates accident signal.(IS) i 1

2 Instsument Chasuiet -

2 FMatuwe vessel aero A

1.

Below trip setting trips -

steactor tow Water I4 vel recirculation gumps (LS-3-56A, e, c, tit I

2 Instsument Clea nne l 19920 lisig A

1.

Ateve trit = sett ient tra ps Ecactor Itigh Pseasus e recirculat ion supps trS-3-204 A, B, C.

D) 2 Instrument Channet -

52.5 peig A

t.

Atme t rip,se tt a s>

in te ywet h atteh I's es mu s e con junct ie,,n tei t h low reactor trs-64-SSA-D. W $ t) pseasure initiates LFCI.

s i t at s.trument cl.annen -

s 2.5 peig a

s.

Atm e trip settt m to 14 ywell 4tsgh Pressus t con junet ton e.ith law reactor (PS-to-STA-D) water level, deyweth high psessuse, 120 sec, delay timer and C5s et - Ima pump r uswitseq, initiates AI4.

r i

s 4

i i

p,38 Amendment no.

4

-c

8

"A ne t i I.J.F a utidet 3.*.:6 e it..E*ltidENt' AT'l 1.4 Hissiesem 4 of Tyg.o I.hlle.at t-he end pse.se 443: e s Oper atale Inst s umes.t Ins t r uwesat 8 f n. t a *i. 3 it--

a hewiel e pe.actar 64.ter t.evel Indicatos -14,s.5* to

( t) ( 2) (3)

LI-1-46 A 2

101.5*

L1-3-46 a Indicatos 0-5200 peig (l) (2) (1)

FI-l-54 pee ct or Fasesinae 2

t3-3-64 secorder 0-40 paie til (2) (3)

Pk-64-50 twyvell Pressure Indicato: 0-80 pela 2

tt-64-61 pecorder, Initicator (t) (2) (3)

TI-64-52 t>rivel! Tesperature 8-400*r 2

79-64-52 pecorder 0-400*r (1) (2) (3)

Ta-64-52 suptseeston C3:eest.er Air 3

Te,sepe ratur e Indicatac, 0-480*F

( 1) (2) (3)

TI-64-55 suppression Ctiember stater 2

738-64-55 To;sperature InJ1cetor -25. to

( 1) ( 2) (1)

II-64-Se A suppreautoa Cheelser tieter

23*

2 8.5 66 E.evel Contaol 804 Position 6W Indicatin3

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)

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)

(1) (2) (3) (4) j W/A 9

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g 5

Tit-64-52 and Drywell Te perature and Aless if temp.

)

> 288*F and

)

( 3) (2) (3) (4)

Ps-64-54 3 and Paemeure and Timer am eneus e >2. 5 i

18-44-61 P81R after 30 minute delay 1

L1-04-2A CAD Tonk " A" IAWel

[edjCetor $ to IOO$

(1) 4 I

Lt-84-13A CAD Tank

h* favel Indicator 0 to lett (1) k y

)

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS w

3. 6 P7IMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY d.

Reactor vessel bottom head temperature e.

Reactor vessel shell adjacent to shell flange 2.

Reactor vessel metal 2.

Durs.nq all operations temperature at the witn a critical core, outside surface of other than for low the bottom head in level physics tests, the vicinity.of the the reactor vessel control rod drive shell and fluid housing and reactor temperatures shall be vessel shell adjacent at or above the to shell flange, temperature of curve shall be recorded at Number 3 of figure least every 15 3.6-1.

minutes during inservice hydrostatic or leak testing when the vessi pressure is

> 312 psig.

3.

Test specimens 3.

During heatup by non.

representing the nuclear means, except reactor vessel, base when the vessel is weld, and weld heat vented or as indicated aff e ted zone metal in 3.6.A.4, cocidown shall be installed in following nuclear the reactor vessel shutdown on low-level adjacent to the physics tests, the vessel wall at the reactor vessel core midplane level.

The number and type temperatures.shall be of specimens will be at or above the in accordance with GE temperatures of curve report NEDo-10115.

Number 2 of figure The specimens shall 3.6-1.

meet the intent of ASTM E 185-70.

Samples shall be withdrawn at one-fourth and three-fourths service life.

185 Amendment No. 38 m

... ~

LDfITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BCUNDARY 4.

The' reactor vessel shell temperatures 4

Neutron flux wires shall be installed in during inservice the reactor vessel hydrostatic or leak adjacent to the testing shall be at

. reactor vessel wall or above the at t.ne core midplane temperatures shown on l ev el.

The wires curve Number 1 of shall be removed and figure 3.6-1.

The tested during the applicability of this first refueling curve to these tests outage to is extended to non.

experimentally verify the calculated values nuclear heacup and f integrated neutron ambient loss cool-

.luence of one-fourth down associated of the belt line with these tests shell thickness that only if the heatup are used to determine and cooldown rates do the NOTT shift from not exceed 15*F per Figure 3.6-2.

hour.

5.

When the reacter 5.

The reactor vessel head vessel head beiting studs-are tencicned bolting studs may be and the reacter is in partially tensioned a Cold Ccndition, the (four sequences of the reacto-vessei-

-"el' C-seating pass) provided temperature the studs and flange 4mme.4 ately

,_w, w the a

materials are above head flance shall be 70*F. Before loading the permanently recorded.

flanges any more, the vessel flange and head 6.

Prior te and during flange must be greater startup of an idle than 100'F, and must recirculation loop, remain above 100*F while the temperature of the reacter c001 ant under full tension.

in the operatin: and i dle looEs s"'<-~'

6.

The pump in an idle permanently log' c.

recirculation loop shall not be started 7.

prior to startir.; a unless the E"*P' temperatures of the the reactor coolant coolant within the temperatures in the idle and opera' ting dore and in :"e recirculation loops so tom head crain are within 5007 of shall be ecmpared and a ch other.

permanently Icq;ed.

7 The reactor

. recirculation pumps shall not be started unless the coolant temperatures between the dome and bottom

. head drain are within 145or.

Amendment No. y[,38 186 l

Figure 3.6-1 Curva #1 Minir.um t ernpera tur.-

for pressure tes..

stich as required hf Section XI.

ky 0 I

2 3

v Carve #2

' f' 9-25 80 Minimu:n ter:perat:2re for mechanical he,.t up or couldown following nucleat 1000 ehutdost.

I Curve O '

Minimu::. ternperate.rt.

/

for core operation (criticality) b 8 00----

Includes additict.a3 g-

[

f I

rurnin required i.v I

10CFF.50 Appendix C, par. r 3. 2.c.

f.

n. rs

( Q.')- - -

Il Y

t:etes l

I G!

T{tc. G C C',1 PVQ S arff

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ahifted 'O F :o the J

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I 30"F.

This sliTf: wil'

-9

/

allev these etirva '.<-

ut,ed thru 4.0 ErPY.

5 ii a

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- -PN_T UP TEHDIR AT'JRE i

1' C-l l

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10 0 200 500 400 i

Mnint,?. T't21Y'.:iva.:FE

("F)

O 201 i

i Amendment No. 38 i

4 IABLE 3.6.H SillCE SUI'I'Iti!SSolts (SNiilitt1.Its) enut.be r s snutkie r s in utgn Inaccessible snobbers Radiatish Asea During Snubbers Especla!!y During Mormal Accessible Durinq Saiut,tser No.

system Elevation S hut.S own Ditficult to Demove Operation geormal Operation SSI-AtllS*3 R ecir cula tion 564 X

355-s(885*3 D ecir culat ion S6t

-M S53-8(154*3 m eci r cu ta t t an S64 X

Sst-A Rectscastatson SPD M

SSt-B Dec is cu lat ion Sie X

SSS-A (2 6 2 *)

pecisculation 36:

E 2

SSS-A(325*)

Recir cu lat ion Sei N

I SSS-B(IS*3 Recirculation SGI X

4 SSS-e(98*)

Recirculation Set 1

g S$6-A Sectrculation S6e x

SSk-8 Recirculatton See E

SSI Recis culat ton S64-X 4

SS6 Rec is cu t as lost

$64 x

x R-64 HilHSW Sil2 i

x it-f>2 ItllitSW Sil2 4

I i

i 4

4 J

%.is..tm./,38 4

- - 1

- --e n-3 e---

-m

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11 FTRE PROTErTTON SYSTEMS 4.11 FIRE PROT 1!X"I' ION SYSTEMS E.

Fire Protection System E.

Fire Protection Systems inspection Inspection 1.

An independent fire Any inspection or audit.

protection and loss will review and evaluate prevention inspection the ef fectiveness of fire and audit shall be prevention and protection performed annually by physical inspection of utilizing either plant f acilities, systems, qualified TVA and' equipment as related personnel or an to fire safety.

outside ' fire Evaluations will be made protection firm.

of, but not necessarily limited to, the following:

2.

An inspection and audit by an outside Administrative control qualified fire documentation, maintenance consultant will be of fire related records, performed at physical plant inspection, intervals no greater related historical than 3 years.

(The research and application, first inspection and and management interviews.

' audit will be during the period of June -

September 3977).

T.

If it becomes necessary to breach a fire stop, an attendant shall be posted on each aide of the open penetration until work is completed and the penetration is resealed.

G.

The minimum in-plant fire protection oroanization and duties shall be as depicted in Figure 6.3-1.

M Amendment No. 38 P

- ~.. -.

v r

-ww.,,y

=.

,,-+.r--

-s-----

=

y.,--

ML20148T877

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