ML18026A946

Proposed Tech Specs Reflecting Administrative Changes

ML18026A946
Person / Time
Site:	Browns Ferry
Issue date:	04/09/1984
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18026A945	List: ML18026A944 ML18026A946
References
TVA-BFNP-TS-197, NUDOCS 8404180193
Download: ML18026A946 (81)

Text

ENCLOSURE 1 PROPOSED REVISIONS TO TECHNICAL SPECIFICATIONS BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 TVA BFNP TS 197 8404i80i93 840409 PDR ADOCK 05000259 P PDR

PROPOSED CHANGES UNIT 1

G I>

1,0 Di FICTITIOUS (Cont'd)'0.

l.onic - A logic is an arrangement of relays, contaces, and other components that produces a decision output.

(a) 1nitiat:itti, - A logic that receive signals from citannels and produces decision outputs to the actuation logic.

(b) At;tu>eioit - A logic that r<<ccives signals (either (roti hei.t;i, tt.ion logic or channels) and produces decision outputs to accomplish a protective action.

W. p<<ttceional Tests - A functional test is the manual operation or initiation oi a system, subsystem, or components eo verify chat it functions within design colerances (e.g., the manual stare of a core spray pump co verify that it runs and that it pumps thc required volume of wat:er).

X. Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations arc being performed.

Y. Engineered Safeguard An engineered safeguard is a safety systt:m thc actions of which are essential to a safety action rcquircd in response to accid<<nts.

Z. Surveillance Interval - Each Surveillance Requirement shall be performed within ehe specified time interval with:

l. A maximum allowable extension not to exceed 25% of the surveillance interval, but:

2. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

fl

LXHXTI??G CO!?DXTXONS FOR OPERATXON SlJRVHXLLANCE REQUXRE?&??TS

3. 1 REACTOR PROTECTIO?$ SySTEM RFACTOR PROTECrXON SXSTE?~

A licabilit ~A lic~abilit Applies to the instrumentation Applies to the surveillance od and associated devices which the instrumentation and asso-initiate a reactor scram. ciated devices which initiate reactor scram.

~Ob ective Ob ective To assure the operability of the To speciiy the type and frequency reactor protection system. of survei.llance to be applied to the protection instrumentation.

S ecification 8 ecification A. When there is fuel in the vessel, A. Xnstrumentation systems shall the setpoints, minimum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.A and 4.1.3 respec-be operable for each position of tively.

the reactor mode switch shall be as given in Table 3.1.A.

B. Twn RPS power monitoring channels for each inservice'PS MG sets or alternate sour' shall h< nperabl e.

When it is isdetermined a channel that failed in the

1. With one RPS electric unsafe condition, the pnwer monitoring channel other RPS channels that.

for inservice RPS MG set monitor the same variably or alternate power supply shall be functionally inoperable, restore the tested immediately before inoperable channel to the trip system containing operabl.e status within the failure is tripped.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or rembve the The trip system containing associated RPS MG set or the unsafe failure may be alternate power supply untripped for short periods of time to allow from service. functional testing of the other trip system. The trip system may be in the untripped position for no more than eight hours per functional test period for this testing

n I

H

1'ABLE 3 2,A

')(I'nia."uz Zo. PRIMARy CO1 rAI1OKRT AND R ACTOR BUILOING ISOIATIOl4 INSTRlJMES ATIOR i~ I '4 L I

Ens trur.en<

Channels Operable (U 1 Fu io Level Settin aetio Remarks Instrument Channel-  ? 538" above vessel zero A or 1. Belov trip setting does the Reactor Lov Mater Level (6) (8 and t) foiloving:

a. Initiates Reactor Building Isolation b Inftfates Primary Containment Isolation ((:roups 2, 3, and 6) c Initiates SOTS Instrument Channel- 100 + 15 psig 1. Above trip setting isolates the Reactor Hfqh Pressure shutdovn cooling auction valves of the RHR system.

Instrument Channel- 2470" above vessel zero A 1. Belov trip setting initiates Main Reactor Lov Mater Level Steam Line Isolation (LIS 3-56A Dg SM 61)

Instrument Channel S 2.S psfs A or l. Above trip setting does the Sigh Dryvell Pressure (6) (B and t) folloving:

(PS 64 56A~D) Initiates Reactor Building Isolation b~ Initiates Primary Contafnmen Isolation

. c. Initiates SGTS Instrument Channel- 3 times normal rated 1 ~ Above trfp setting initiates.Rain Sfgh Radiation Main Steam full paver backgrourd (l3) Steam Line Isolatf.on Line Tunnel,(6)

Instrument Channel- 82S Psig (4) Belov trip setting initiates Mai'n Lov Pressure Nafn Stea~ Steam Line Isolatfon Line 2 (3) Instrument Channel- 6 'l40% of rated steam flaw B 1. Above trip setting initiates Kafn Bigh Plov Main Steam Line Steam Line Isolation

I

(

p

~

f l I

'E E

I

)

~ g Cq 1

I A

A II

TABLE 3 ~ 2 ~ A PRIMARC CONTQINNENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.

Instrumenr.

"hannels Operable Function Tri Level Settin Action 1 Ri'.war ks 2 Group 2 (Initiatir..) Logi c N/A A or 1. Refer to Table 3.7.A for list of (B and E) valves.

Group 2 (RHR Isolation-Actuation) Logic Group 8 (Tip-Actuacion) N/A Logic Group 2 (D~ell Sump N/A Drains-Actuation) Logic Group 2 (Reactor Building N/A P and G 1. Part of Group 6 Logic.

6 Refueling Floor, and Dry-vell Vent and Purze-Actudtion) Logic Group 3 (Initiatirg) Logic Yi/A 1. Refer to Table 3 7.A for list o=

valves.

Group 3 (Actuation) Logic N/A Group 6 Logic N/A F and G Refer to Table 3.7.A for list of valves.,

Group 8 (Initiating) Logic N/A 1. Refer to Table 3.7.A 'or list of valves.

2. Sane as Group 2 initiating logic.

Reactor Building Isolation N/A HorF Logic has perrcissive to refueling (refueling floor) Logic floor static pressure regulator.

Reactor Buildinv Isolation N/A BorG 1. Logic has peanissive to reactor (reactor xone) Logic or A xone static pressure rt.qulator.

I)

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require action A and F.

9. There is only one trip system with auto transfer to two power sources.

10. Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.

ll. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

12. A channel contains four sensors, all of which must be operable for the channel to be operable.

13. The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal back-ground at full power. The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.

61

If TABLE 3.2.b (Cont tnued)

Htntaua Ho.

Operable per T~rf 5 s (1) yunctton Trt Level Setttn Aetton 1(10) RHI Area Cooler yan Loete H/A 1 (10) Core Spray Area Cooler Fan Loate 'l/h 1 (ll) lnatruoent Channet- A 1. Scarce RHRSM pumps Al, B3, Cl, and D3 Core Spray Hotora A or C Start 1(11) Inatruaent Channcl- H/A l. Starta RHRSH pumpa Al B3 Cl ~ and D3 Core Spray Hotora b or D Start K(12) tnatruoent Channel A 1. Scarta RHRSQ punpa Al, B3, Cl, and D3 Core Spray Loop 1 Acetdcnc Stgnal (1$ )

1 (12) lnatrument Channel A 1. Starca RHXSV puopa Al, B3, Cl, and D3 Core Spray Loop 2 Accident Sternal (tS) l(13) RHRSM Intctate 4~tc H/A'14)

RPT logic N/A (17) l. Trips recirculation pumps on turbine control valve fast closure or stop valve closure. > 30$ pover,

~g J

TABLE 3 'eF SURVEILLANCE INSTRUNEHTATION Minimum 8 of Operable Instrument Type Indication Channels Instrument I Instrument and Range Notes 46 A Reactor Rater Level Indicator - 155" to (1) (2) t3)

LI 3-46 B 60H PI-3-54 Reactor Pressure Indicator 0-1500 psig t)) t2) 13)

PI-3-61 TI-64-52 Dryvell Temperature Recorder, Indicator (1) (2) 13)

R-64-'52 0-400oF TR-64-52 Suppression Chamber Air Recorder 0-4004F (1) 12) l3)

Temperature Control Rod Position 6V Indicating Lights )

N/A Neutron Monitoring SRM, IRM, LPRM ) (1) {2) 13) (4) 0 to 100f pover )

PS-64-67 Dryvell Pressure Alarm at 35 psiq )

)

TR-64-52 and Dr~ell Temperature and Alarm if temp.

28loF and

)

(2) (3) (4)

?S-64-58 B and Pressure and Timer ) (1)

IS-6%-67 pressure ) I 5 )

after 30 minuteygf8 )

delay )

LI-84-2A CAD Tank A Level Indicator 0 to 1004 LI-84"13A CAD Tank "B" Level Indicator 0 o 1001

3.2 BASES and trips the recirculation pumos.

The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top active fuel (Table 3.2.B ) initiates the LPCI, Core Spray Pumps, of'he contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of'0 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addit'on to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, tPis instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a brea'k in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000oF, and release of radioactivity to the environs is well below 10 CFR 100 guidelines . Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of isolation valves. The setting of 200 F f'r the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus,'t is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in'the control rod drop accident. Mith the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product, release is limited so that 10 CFR 100 guidelines are not exceeded f'r this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in Run Hade when the main steam line pressure drops below 825 psig o 112

,I Croup 7: The valvea in Croup 7 are autcnaotically actuated by only the tolloving condition:

l ~ Reactor veaoel los eater level (A)70")

Croup B: Th>> valvaa in Croup 8 are automatically actuated by only the Eolloving condition:

l. High Dryvell preaaure

2. Reactor vessel low water level (538")

255

0 Table 3.7.H (Continued)

X-1 07B Spare (testable)

X-108h Power X-108B CRD Rod Position Indic.

X-109 CRD Rod Position Indic.

X-110A Power X-110B CRD Rod Position Indic.

Suppression Chamber Vacuum Breaker 266

OASES

~crau l - proccaa Linea are iaolated by reactor vceael lou Mater level (490") in order to allou for removal of decay heat aubae<tuent to ~

~ cree, ycr. isolate in time for proper operation of the core ~ tandby cor Ling eyetcG<o. T'e valvee in group l are also closed uhen process fnetru<t2cntation detect ~ exccn<tive main ~ team line f lou, high radiation, lou prcnaure, or mein stcam apace high te<2<perature.

~Grou 2 - Isolation valves are closed by reacto) vessel low water level (538") or high drywell pressure. The group 2 isolation signal also "iso-1>>tes" the reactor building and starts the standby gas treatment system.

It is >>ot desirable to actuate the group 2 isolation signal by a transient or spurious sig>>al.

~Grow 3 Process lines ara normally in use, and it is, tberepore not ~

desirable to cause spurious isolation due"to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.

the consequences of an accident which results in the isolation of other proc<<ss lines. Tl>e sig>>als which initiate isolation of Groups 4 and 5 pr'ocess lines are therefore indicative of a condition which would render them inoperable.

Gro~ui 6 - I.ines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation.

~Grou 7 - Process lines are closed only on the respective turbine steam supply valve not fully closed. This assures that the valves are not open, when 11PCI or RCIC action is required.

(lribt

><<(tr>>v<<l i>>t i>>-<<or<<. prob<<),is is<3iate<l <>>> high dryweil ~ pr7ss<<rc'<br renct<yr low w>>ter level (538") . 'I'his is to <<>>sure that this J i>>e <ious not provi<l<d>> I<;aknge path when containment prcssure or reactor water level i>>dicates a possible accident condition. 'The maximum closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.'n satisfying this design intent, an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance prior to exceeding the design closure times. 277 D ~ A 'J Ad.', p ~ ~ perpr 'f Iver)'.:TR'3LS gyp ...... 6.8 l.in'um 8 ant =a.4C in C The minimum plant s" af fing c" mcn'to inc and conduct of operat'ons is as f olloMs.

l. p, licensed senior operator shall be presen a" the "'c all times when there is fuel in De reacto

2. A 1'ensed operatcr shall be in the con"=ol room Mhenever there is fuel in the reac or.

3 A license" senior operator shall -e i~ Bisect charge c' eactor re'ueling operation; i.e., able to devote ful ~ t'me to the re'ue'ing operatipn A health phys:cs echnician shall be present at the. .acil'ty at all t'~es the e is uel in the 'reactcr.

5. -.Mo 1'censed operators shall be in the ccntrol zoom dur'q any cold s" artups, while shutt'g down thetr'.

reactor, and during recovery rom unit

6. Eithe the plant supe in enoent or an ass'stan plar.-

super ntendent shall have acquired the experience ano -- 'n'nq ncrmelly realized for examine,-'on by the 1iRC ~ fcr a senior Beac"cr opera-or's 'cense, whether o" nc= ". e examinat'on 's taken. Zn acition, e =her the ope" ations superv'sor o" the. ass'stant ope=ations supervisor sna'1 have an SR@ -license.

7. h Sh'f Vechn'col Adv'.sor shall be present at the s':e a all es ~

6.9 Environmental Oualif ication A. Complete and auditable records shall be available and maintained at a central location which describe the environmental qualification method used for all 'safety-related electric cq6ipment in sufficient detail to document thc degree of compLiance with thc l)OR Gui<lelines or NUREG-0588. Such records shall be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified. 358 PROPOSED CHANGES UNIT 2

1. 0 DL'PILRIT IOtlS (Con t ')

10. l.ogic - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

(a) Tnitlati>>G - h logic that rcccivc signals from channels and produces decision outputs to thc actuation logic. (b) hctuatio>> A logic that receiver'ignals (citltcr from i>> i,t iat ion logic or cl>>>>tncls) and prod>>ccs dccisio>> out.puts to accompli>>It a prot.cct:ivc action. W. p>>>>etio>>el Tests - A functional test is the manual operation or initiation of a system, subsystem, or components to verify that it functions within design tolerances (e.g., thc manual start it it of a core spray pump to verify that runs and that pumps thc required volume of water). X. Shutdown Thc reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations arc being performed. gn~ince~red Safe uard An engineered safeguard is a " f rays system thc actions of which are essential to a safety action rcquircd in rcsponsu to accidents.

z. Surveillance Interval Each Surveillance Requirement shall be performed within the specified time interval. with:

l. A maxin>>tm allowable extension not to execu<i 25/ oE the surveillance interval, but:

2. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

0 TABLE 3 2 A PRINARY CONTAI%KRT AND'RECTOR BUILDING ISOLATION IHSTRUNESTATION Miniaug yo Lns trur..en'hannels Operable 1 Level Settin aetio Remarks Instrument Channel 538" above vessel sero A or Belov trip setting does the Reactor Lov Mater Level (6) {B and 8) foiloving:

a. Initiates Reactor Building Isolation bi Initiatea Primary Containment Isolation ((:roups 2, 3, any 6) c Initiates SOTS Instrument Channel- 100 + 15 psig Above trip setting isolates the Reactor High Pressure shutdovn cooling suction valves of the RHR system Instrument Channel- a 470" above vessel cero A 1 Belov trip setting initiates Bain Reactor Lov Mater Level Steam Line Isolation

{LIS 3-56A Dy SM S1) Instrument Channel- s 2.5 psis A or 1 Above trip setting does the Bigh Dryvell Pressure (6) (B and 8) folloving: (ps-6u 56A~D) a~ Znitiates Reactor Building Isolation b Initiate" Pri mar:r Containwn Isolation . c. Initiates SGTS Instruaent Channel- 3 times normal rated 8 1 Above trip setting initiates Nain Bigh Radiation Naia Steam full pover background (y3) Steam Line Isolation Line Tunnel f6) Instrument Channel-  ? S2S Psig tul 1. Belov trip setting initiates Nai'n Lov Pressure Nain Steam Steam Line Isolation Line 2 (3) Instrument Channel- 5 lu0% of rated steam fleas 8 Above trip setting initiates Nain Bigh Plov Nain Steam Line Steam Line Isolation TABLE 3;2.A PRIMAR'C CONTAINNENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Minimum No. Znstrument "hannels Operable Remarks er Tri S s(l) ly Funct ion Tri Level Settin Action 1 A or 1. Refer to Table 3.7.A for list of 2 Group 2 (Initiatin ) Logic N/A (B and E) valves. Group 2 (RHR Isolation- N/A Actuation) Logic Group 8 (Tip Actuation) N/A Logic Group 2 (Dryvell Sump Drains-Actuation) Logic FandG 1. Part of Group 6 Logic. Group 2 (Reactor Building 6 Refueling Floor, and Dry-vell Vent and Purqe-Actuation) Logic (Initiatirq) Logic: N/A Refer to Table 3.7.A for list of Group 3 valves. Group 3 (Actuation) Logic N/A Logic NIA F andG 1. Refer to Table 3.7.A for list Group 6 of valves.

1. Refer to Table 3.7.A or list of Group 8 (Initiatinq) Logic N/A, valves.

2. Same as Group 2 initiatinq loqic.

HorF 1. Logic haspermissive to refueling Reactor Building Isolation N/A floor static pressure regulator. (refueling floor) Logic HorG 1. Logic has permissive to reactor Reactor Building Isolation N/A or A xone static pressure r).gulator. (reactor xone) Loqic

6. Channel shared by RPS and Primary Containment & Reactor Vessel Isolation Control System. 'A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require action A and F.

9. There is only one trip system with auto transfer to two power sources.

10. Refer to Table 3.7.A and its notes for a listing of Isolation Valv<<

Groups and their initiating signals.

11. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

12. A channel contains four sensors, all of which must be operable for the channel to be operable.

13. The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal back-ground at full power. The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.

61 '?ARLY 3.2.5 (Continued) Kiabeua Ro. Operable Per ~Trl 5 I (I) Function Tri Level Sectin Action l(10) RHX Area Cooler fan Loeic H/A l(10) Core Spray Area Cooler Fan Logic '4/A A 1(ll) lnatruuent Channel- <</A A 1~ Statta RHRSM puapa Al, B3, Cl, and D3 Core Spray Hotota A or C Scarc l(ll) lnattunent Channel- H/A 1. Starta RHRSu punpa Al, B3, Cl, and D3 Core Spray Notora l ot -D Scarc 1(12) lnatruncnt Channel- W/A A 1. Starca RHRSM puapa Al, B3, Cl, and D3 Cote Sptay Loop 1 Accident Signal (1$ ) 1 (12) lnattuaenc Channel h 1, Starts RHRSM pe@pa Al, B3, Cl, and D3 Cote Spray Loop 2 Accident Signal (1$ ) l(13) RHRSM lalclate Logic H/A RPT logic N/A (17) l. Trips recirculation pumps on turbine control valve fast closure or stop ynlye closure. > 30$ pover, TAB'- 3.2.F SDRV EI~)cc. I <S. RO') )(inic:ua ) cf Og~~rable Instrument .ype indication Channels Instrument i Instru=ent and Rance -'I-3-e6 A Reac or 1(ater Lev! I Ind(cat - r - ) 55" to Li- 3-46 B ":- 3-5a Reactor Pressure Indxcator 0-1SOD psig Pi-3-61 PR-eq-50 PI-oe-67 Dry~ell Pressur Recorder Induce=o '- 0 8 0 psia =0 psia Tt ~ i~ 6Q TR-v4-52 .I-6e-55 52 Dryvell Temperature Suppress< cn Chas'L Temperature 'l r Recorder, In icator 0-o00oF Record. r 0-"00'oF Suppression Cha=.ber !'ater Ind jcatorg 0 400oF Trs-60-55 Tenperatu e LI"6e 5Q A Suppression Cha:2 r Rater Indicator -25" to LI 66 Level y 75o H/A Control Rod Position 6V Indicatinq ) Liqhts ) Neutron canxtcr i.".c; SR!(, IRN, LPR.'1 ) 0 to 100% pover ) PS-6t-67 Dryvel1, Pressure Alar~ at 35 psig ) TR-6a-52 and PS-6e-58 B and Dryvell Temperature Pressure an-" Timer and Alarm if temp. 281oF and ) ) ) IS-64-67 pressure ) 2 5 ~I< ) after 30 einute ) delay LI-84-2A CAD Tank A Level Indicator 0 to 100% LI 84 13A CAD Tank "B" Level 7 aj(a+~~ n o 1003 n>>d tr.i ps t:hc r rc irriil:>tion pumps. The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of t'.he active fuel (Table 3.2.8) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addit'on to initiating CSCS, causes isolation of Groups 2 and 8 isolation valves. For the breaks it discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in conJunction with the flow limiters and main steam line valve closure limits the mass inventory" loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000oF, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of isolation valves. The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in 'the control rod drop accident. With the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so 'that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also. Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig. 112 IH LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. 7 CONTAINMENT SYSTEMS 4~7 CONTAINMENT SYSTEMS

g. Local Lehk rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation valves, which are not part of a water-sealed system, et not less then 49.6 psig (except for the main steam isolation valves, see 4.7.A.2.i) and not less than 54.6 psig for water-sealed valves each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is closed, after being opened end at least once per operating cycle. Acceptable methods of testing are halide gas detection.

soap bubbles, pressure decay, hydrostatically pressurized fluid flow or equivalent. The personnel air lock shall be tasted at . 6-month intervals at an internal pressure of not lass than 49.6 psig. In addition, if the per sonnel air lock is opened during when containment 'eriods integr ity is not required, a test at the and of .such a period vill be conducted at not less than A9.6 psig. If the personnel air lock is opened during a period when containment integrity is required, a test atr2.5 psig shall be conducted 'ithin 3 days after being opened. If the air lock is opened more frequently than once every 3 days, the air lock shall be tested at least once every 3 days during the period of frequent openings. 231 I e 'l LIMITING CONDITIONS FOR OPERATION SVRVEILLANCE REQUIREMENT/

3. 7 CONTAINMENT SYSTEMS 4. 7 CONTAINMENT SYSTEMS The total leakage from all penetrations and isolation valves shall not exceed 60 percent of L per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage from containment isolation valves 'that terminate below suppression pool water level may be excluded from the total 1eakage provided a sufficient fluid inventory i's available to ensure the sealing function for at least 30 days at a pressure of 54.6 psig. Leak-age from containment isolation valves that are in closed-loop, seismic class I lines that will be water sealed during a DBA will be measured but will be excluded when computing the total leakage. Penctrations arid isolation valves are identified as follows: (1) Testable penetrations with double 0-ring seals - Table 3 ~ 7 ~ By (2) Testable penetrations with testable bellows Table 3.7.C, (3) Isolation valves with-out fluid seal " Table 3.7.D, (4) Testable electrical penetrations >> Table 3.7.N, and (5) Isolation valves sealed with fluid-Tables 3 7. E and 3 7 F 232 TABLE 3.7.A (Continued) Number of Power ~!aximum. lT,. "-. Action. Operated Valves ting .'.Normal

• Initiating On'pera

~Grou Valve Identif ication Inboard Outboard Tine (Sec.) 'oeicion ~Si nel

orus Hydrogen Sample Line Valves Analyzer A (FSV-76-55, 56) Note 1 SC

.orus Oxygen Sample Line Valves Analyzer A (FSV-76-53, 54) NA Note 1 SC ""..~ell Hydrogen Sample Line Valves Analyzer A (FSV-76-49, 50) NA Note 1 SC 3"p'ell Oxygen Sample Line Valves

=.alyzer A (FSV-76-51, 52) NA Note 1 SC

==...pie Return Valves Analyzer A FSR-76-57, 58) NA 0 GC

orus Hydrogen Sample Line Valves Analyzer B (FSV-76-65, 66) Note 1 SC

.orus Oxygen Sample Line ':alves-Analyzer B (FSV 64) NA Note 1 SC ~rn:ell Hydrogen Sample

ine Valves-Analyzer B

'":SV-76-59, 60) iNA Note 1 SC

ryT:ell Oxygen Sample Line

alves-Analyzer B (FSV 62) NA Note 1 SC

==...pie Return Valves- ".alyzer B (FSV-76-67, 68) 0 GC Note 1: ".alyzers are such that one is sampling crT~ell hydrogen and oxgen (valves from dr>well open-

alves from torus closed) vhile the other is sampling torus hydrogen and oxygen (valves from torus

.Fen - valves from drywall closed)

Croup 7: The valvca in Croup 7 are autcnaatically actuated by only the Coll ov I ng condition:

1. Reactor reaaal lov vatar level (47p>')

Croup S: The veivoa in Croup S ara automatically actuated by only the folloving condition: I. High Dryvell preaaure

2. Reactor vessel low water level (538")

255 0 Table 3.7.1 (Continued) X-107i Sp;:,: ( testable) PoM< I i tiur> Iau '.. X-100'-1088 CHL'orl Poa X-jc9 CRD Rod Position lni.'ic. X-11GA PoMcr X-1109 CRD Rod Position indlc. X-200A-SC S/RV Test Instrumentation (Temporary) X-219 Supnression Chamber Vacuum Breaker 266 ~Crn<< l - proccaft linea are iaolated by reactor vessel lov Mater level (4t30") ln order to allobf fOr remOVal of decay heat subsequent to a ~ cram, yet isolate in time for proper operation of thc cote ~ tandby cof ling ayatcso. The valves in group I are also closed vhen process inotrumcntation detect ~ excessive Dain steam line flov, high radiation, lou prcoaure, or main ateafa apace high temperature. ~Grow 2 isolation valves are closed by reactor vess 1 law water level (538") or high drywell pressure. The group 2 isolation signal also "iso-lates<< the reactor building and starts the standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal. ~Grnu 3 Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due "to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided. Grou s 4 and 5 - Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable. (lr~nu 6 Lines are connected to tha primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation. ~Grow 7 Process lines are closed only on the respective turbine steam supply valve not fully closed. This assures that the valves are not open when )1PCI or RCIC action is required. ('riiul3 8 - 1.i>>e (travel ing in-core prol>e) is isolated on high drywell pressure or reactor low water level (538"). This is to assure that this ll>>e does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition. The maximum closure time for the automatic- isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the pz'imary containment and the need to contain released fission products following pipe breaks inside tne primary containment. In satisfying this design intent, an additiona" margin has been included in specifying maximum closure times. This margin permits 'dentification of degraded valve performance prior to exceeding the design closure times. 277 ~ J yA'f T v c ) TR'9LS S=a. in The min:mum p'ant sta 'ing o" mcnitorinc and c~duct of ope"at'ons is as folloMs.

l. A licensed senior operator shall be presen a the : e at all times when there is fuel in ~De reactor.

2. P licensed. operate hall be in the con="ol room uheneve the e is fuel in the reac or.

A licensed senior opera-or shall be in direct charge o.' eactor re'ueling operation; i.e., able to devo" e fu 3 ~ time to the reueling operatipn 4 A health phys:cs technicia'n shall be present at the acility at all t'~es there is uel in the 'reactcr. ~

5. o licensed operators shall be in the con" rol room c'.urinq any cold startups, while shuttinq doMn the reactor, and dur'nq recove=y rom unit trip.

Kithe the plant superintenoent or an ass'stant plar.- super ntendent sha'1 have acquired the'experience ano ="ai ning normally required ~or examine ion by the bRC fcr a. Senior Beac"cr Opera o='s L'cense, whether or not the examina"'on is taken. Zn acquit on, ei=her the ope" at'ons supe=v'sor o" 4e. assistant opera iona superv'or sna'1 have an SR@ -license.

7. A Sh ~t 7 chnieal Advisor thal4 be present at the Wite a all ti es ~

6. 9 Environmental ualif ication A. Complete and auditable records shall be available and maintained at n central location which describe the environment il qualification method used for all safety-related electric equipment in sufficient detail to documenl the degree of compliance with the DOR Guidelines or NUREG-0588. Such records shall be updated and maintained

,current as equipment is replaced, further tested, or otherwise further qualified. 358 PROPOSED CHANGES UNIT 3 a protective trip function A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action ma;s require the tripping of a single trip system or. the conincident tripping of two trip systems.

7. Protect ve Acti - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8~ P otect ve nc o - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

9. simulated Automatic A utat on - simulated automatic acutation means applying a simulated signal to the sensor to actuate the circuit in question.

10. g~c - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

from channels and produces decision outputs to the actuation logic. (h} ~steat n - a lssjic that receives sicnals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances ( e.g. the manual start of a core spray pump to verify that volume of water). it~ runs and that it pumps the required X. ~c>hut~de - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed. Ys En n e ed Safe ard - An engineered safeguard is a safety system the actions of which are essential to a 'safety action required in response to accidents. Z. Surveillance Interval Each Surveillance Requirement shall be performed within the specified time interval with:

l. A maximum allowab'e extension not to exceed 25% of the surveillance inte "val, but:

2. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

TABLE .3 2+A PRIMARY CONTAINMENT AND REACTOR BUILDING ISOIATION IMSTRUMEMTATION .'(ini"...u=..io. Instr'-...c .st Charnels C",.era'ale per Tr.'- Svs(:)')>'unction i Level Settin action Remarks trip setting Reactor ~ Instrument channel-water Level {6) ? 538~ above vessel zero A or (B and E) Belav follovingz

a. Initiates does the Reactor Building Isolation

b. Initiates Primary Containment Isolation ((.roups 2, 3, and 6)

c. Instigates SGTS Instrument Channel- 100 + 15 psig Above trip setting isola es the Reactor High Pressure shutdovn caoling suction valves of the RHR system.

Instrument Channel-  ? 470" above vessel zero A 1. Belnr trip setting initiates Main Reactor Low Water Level Steam I.ine Isolation (LIS 3 56A Di SW 81) Instrument channel- 2.5 pair A or 1 Above trip setting does the High Dryvell Pressure (6) (B and E) folloving: {PS 6a 56A-D) a. Initiates Reactor Building Isolation

b. Initiates Primary Containment Isolation

c. Initiates SGTS Instrument Channel- 3 times normal rated B Above trip setting initiates Hain Hiqh Radiation Hain Steam full paver background (l3) Steam Line Isolatian Line Tunnel (6)

Instrument Channel-  ? 825 psig (a) 1. Belov trip setting initiates Hain Lov Pressure Main Steam Steam Line Isolation Line 2 (3) Instrument Channel 5 100S of rated steam f lac B Above trip setting initiates Hain High Flo~ Hain Steam Line Steam Line Isolation 2 (]2) Instrument Channel- 5 2000F 1 Above tri p setting initiates Main Steam Line Tunnel Hain Steam Line Isolation. High Temperatur~ TABLE 3 2,A PRINARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Hfnimum Vio. Inscruncnc Channels Operable (U Funct i Tri Level Settin Action 1 Remarks 2 Group 2 (Initiating) Logic N/A A or 1. Refer to Table 3.7.A for list of (B and E) valves. I'rip Group 2 (RHR Logic Group 2 Isolation-Actuation) Logic 8 (Tip-Actuation) (Dryvell Sump N/A N/A N/A Drains-Actuation) Logic Group 2 (Reactor Building N/A F and G 1. Part of Group 6 Logic. 6 Refueling Floor' and Dry-vell Vent and Purge-Actuation) legic Group 3 (Initiating) Logic N/A 1. Refer to Table 3 7.A for list of valves. 1 Group 3 (Actuation) Logic N/A o: Group 6 Logic N/A F and G Refer to Table 3.7.A for list of valves. Group 8 (Initiating) Logic N/A 1. Refer to Table 3.7.A for list of valves.

2. Same as Group 2 initiating logic.

Reactor Building Isolation N/A aorF 1. Logic has permissive to refueling (refueling floor) Logic floor static pressure regulator. Reactor Building Isolation N/A HorG 1. Logic has permissive to reactor (reactor xone) Logic or A xone static pressure regulator. t 4 J -'l~ g,4l 7 W ~ ~'k

3. There are four channels per steam Line of which two must be operable.

u. Only required in Run Node (interloc)ced vith Node Switch) .

5. Hot required in Run Node (bypassed:iy mode switch).

6. Channel shared by RPS and Primary Containment 0 Reactor-Vessel Isolation Control System. A channel failure may be a channel failure in each system.

7. A train is considered a trip system.

8. Two out of three SGTS trains required. A failure of more than one will require action A and F.

9. There is only one trip system with auto transfer to two power sources.

10. Refer to Table 3.7.h and its notes for ~ listing of Isolation Valve Groups

~ nd their iaitiatini signals. A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

12. A channel contains four sensors all ofo whi c h must be operable for the channel to be operable.

13. The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal back-ground at full power. The allowable setpoints for a1arm an) reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.

6" 1'able ). 2. a I.'ISTQP404TATIOV THAT IHltIATIS OI COIrrAOLS TIII CORI AIHI COHTAI I'nZHT C'c&LIHO .'TFTKn. II)n)nun Ro. Opec able per Sys +11 ~oh Renarts I (10) Instruct nt Channel - $ 100 ~ F Abnve tc)p set(rnp starts AHA Thernostat (AHA Area Cooler ar ta cno) tr I anss tan) 2 (10) Instrnnhnt Chhnni) I~ Swrts Core Spray area cooler Core Spray A or C St ~ ct (ah Mhin C'ore Spray ~ OtOr St ~ rtS 2 (10) Instrontnt Channel IJ Starts Colt Spray ares cooltc Core Spray 0 or O Can vhtn Core Spray notoc starts 1 (10) Inetronint Channel S Issiy 1 Above trip settlnp starts Core Thernostst (Core Spray Spray acts cooler (sna Arts Coolec'an) ~ 1 (10) RSI: Area Cooler tan Lopic 1 (10) Co! ~ Spcay Ares Cooler yan II/A Lop)c 1 (11) Instrnntnt Charms)- A lo Starts Rnssu p~ps A3 B] C3 and D] coct socay IIotore A oI C Searr InstcMntnt Charms)- Starts RHASQ pips A3s Bl C3, and D1 Core Spray IIotora B Or D Start 1 (12) Instr oJaent Channel 11/A I Starts RSlSM A3, Bl, C3, and Dl Core Spray Loop I Accfdent 0 (pna I ( 5) 'I 1 (12) Instcunent Channel Coca Spray I/sop 2 Accident A I~ Starts RIIRSM pimps A3, Bl, C3, and Dl Sipns I (I 5) RPT logic N/A (17) l. Trips recirculation pumps on turbine control valve fast closure or stop valve closure> 30$ pover. TABLE 3 2 P SDRV EILLANCE XNSTROHENTATION Hinimum 4 of Operable instrument Type Indication Channel. Xns tzument Instrument and Range Nates LX-3-i6 A Reactor Mate" Level Ind!.cator - 155" to 11) (2) (3) LX-3 46 8 + 60" PX-3-54 Reactor Pressure Indicator 0-1500 psig (1) {2) (3) PI-3-61 PR 50 Drywell Pressure Recorder 0-80 psia (1) (2) {3) PI-64-67 Indicator 0-80 psia TX-64-52 Dryvell Temperature Recorder, Indicator ~ (1) {2) (3) TR-54-52 0-4004P TR-64>>52 Suppression Chamber Air Recorder 0-vOOoP (1) (I) (3) Temperature TX-64-55 Suppression Chamber Mat er Indicator, 0-4004P { 1) (2) (3) TIS-64-55 Temperature Ll-64-54 A Suppression Chamber Mater Xndicator -25" to (1) (2) (3) LI-64 66 Level N/A Control Rod Position 6V Indicating ) Lights ) "1/A Neutron Honitoring SRH, XRHg LPR.'1 ) (1) (2) (3) (4) 0 to 100$ pose ) PS-64-67 Dr~ell Pressure .Alarm at 35 psig ) TR-64-52 and PS-64-58 8 and Dr>~e!,I Temperature and Pressure and Timer Alarm > 281OP ifandtemp. ) ) ) (1) (2) (3) (4) XS-64-67 pressure > 2 5>>X<)). after 30 ainute delay ) I-84-2A CAD Tank A Level Xnd'cator 0 to l004 'LX-84-l3A CAD Tank "8" Level Indicator 0 to XOO6 \ 3.2 BASES and HPCI and trips the recirculation oumos. The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCZ, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These tr ip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in con)unction with the flow limiters and main steam line valve closure limits the mass inventory. loss such that fuel is not uncovered, fuel cladding temper atures remain below 1000 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas. Trips are provided on this instrumentation and, when exceeded, cause closure of isolation valves. The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. Mith the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.2 FSAR. An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also. 109 only the following condition:

1. Reactor Vessel Low Water Level (470"..)

Group 8: The .valves in Group 8 are automatically actuated by only the following condition:

1. High Drywell Pressure

2. Reactor vessel low water level (538")

867 TABLE 3 7 H TESTABLE ELECTRICAL PENETRATIONS X 10 IB Spare {testable) X"108A . Power X-108B CRD Rod Position Indic. X-109 II II X-1 10A Power X-llOB CRD Rod Position Indic. X-219 SuDpres~in".a Chamber Vacuum Breaker 284 3.7.D/4.7.D Primar Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space'f the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential,.leakage paths from the containment in the event of a loss of coolant accident. ~crau 1 - process lines are isolated hy reactor vessel low water level (0-JO<) in order to aU.ow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiaticn, low pressure, or main steam space high temperature. P ~Grou 2 Isolation valves are closed by reactor vessel low water level (538") or high drywell pressure. The group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal. ~Grow 3 - Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting- from nonsafety-related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow thr ough the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided. the consequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of groups 4 and 5 process lines are therefore indicative of a condition which would render them inoperable. ~Grou 6 - Lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accord .nt and necessitate primary containment isolation. ~Grou 7 - Process li'nes are closed only on the respective turbine steam supply valve not fully closed. This ensures that the valves are not open when HPCIS or RCICS action is required. ~Grou 8 - Line (traveling in-core probe) is isolated on high drywall - pressure or reactor low water level (538"). This is to assure that this line'does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition. 294

6. 8 l in'um <~ ant e" a~+in The man mum p'ant sta "ang cr mcr. tossing znd conduct of operations is as follows.

l. Aat licensed senio" operato" shzll be present a the 'te all times when tnere is fuel in Ae reactor.

2. A 1'censed operatcr shall be in the con=-ol room whenever the e is fuel in the reac or.

3 A licensed senior operator shall be in direct. charge c' reac or re ueling ope=aticn; i.e., able to devo" e ~ to the refueling operatipn fu'ime A health physics ectmician shall be presert at the acility at all times the e is gael 'n the 'reactcr. 5..wo licensed operators shall be in he control room during any cold stzrtups, wh'le shutting cown the reactor, and during recove=y "om unit trip. E'the= the plant supe"intencent o" an assistant plat supe 'ntendent she'1 have'c'quired the experience z~o ""zining normally require" for examinzt'on by the bRC cr z. Sen'or Beactc>> Ope"a"c='s cense, whe her o" not the exzminzt'on 's zk .".. In acd't'on, e'ther the ope" at'rs supe=visa" o" 4e. assistant operations superv'sor snail have an SRQ.license.' ~ A Shift 'p chnical Advisor sha'1 bc present at the site a all eso

6. 9 Environmental uglification A. Complete and auditablc records shall be available and maintained at a central location which describe the environmental qualification method used for all safety-relntcd electric equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NURL'G-0588. Such records shall be updated and maintained current as equipment is'eplaced, further tested, or otherwise further qualified.

388 1~ ENCLOSURE 2 DESCRIPTION AND JUSTIFICATION TVA BFNP TS 197 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3

1. Units1 2 and3-Pa e7 Addition of definition: 'Surveillance Interval.'his definition would allow extension of a surveillance interval by 25 percent but not more than 3 .25 times the surveillance interval over three consecutive intervals. This definition is identical to specification 4.0.2 from the BWR Standard Technical Specifications.

Submittal of those teohnical spocifications was recommended in the NRC's Generic Letter No. 83-27 dated July 6, 1983. Safet Anal sis This change is being made to reflect the industry standard on the extension of surveillance intervals as specified in the Standard Technical Specifications and Generic Letter 83-27. The tolerances stated are necessary to provide operational flexibility because of scheduling and performance considerations. The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance aotivity is not significantly degraded beyond that obtained from the nominal specified interval. It, therefore, does not reduce the safety margin of the plant.

2. Unit 1 onl Pa e 31 Delete the letter 'C'n 4.1, 'Surveillance Requirements.'ection 4 .1 .B was added by Amendment No. 92 to license DPR-33 dated December 12, 1983 . Removal of the letter 'C'hould have been made at that time. However, as a result of an oversight it was not included in our submittal. Tho letter 'C'eeds to be removed for consistency of the specif'cations.

3 . Units 1 and 2 Pa es 55 61 and 112 Unit 3 Pa es 57 63 and 109 Add 'Groups 2, 3, and 6'o the reactor low water level trip setting at 538 inches. This is for clarification only. Add note (13) to the high radiation main steamline tunnel trip level setting. A similar ohango was made to and approved for table 3 .1 .A in TS 162. It was inadvertently omitted from table 3 .2.A. The bases were also ohanged to reflect this.

4. Units 1 and 2 Pa e 58 Unit9-Pa e60 Change group 2 (TIP-Actuation)logic to group 8 (TIP-Actuation) logic. The TIP system isolates with the gxoup 8 isolation logic. A similax'hange was made to and approved for table 4.2.A in TS 162.

It was inadvertently omitted from table 3 .2.A.

5. Units 1 and 2 Pa e 255 Unit 3 Pa e 267 He Add 'the reactor low water level (538 inches) trip setting to group

8. This signal isolates the TIP system. It was previously omitted from the technical specifications.

6. Units 1 and 2 Pa e 27 Unit 9 Pa e 294 The bases are being revised to more accurately describe the actuation logic associated with the group 7 isolation valves and to include the reactor low water level trip setting for the group 8 valves. This setpoint was previously omitted.

7 . Units 1 and 2 Pa e 70 Unit 3 Pa e 72 Revise RHR service water pumps which are started by the Core Spray motor starts. The pumps listed in the technical specifications are wrong. This change will not have any adverse impact on the safety of the plant.

8. Units 1 and 2 Pa e 78 Unit 3 - Pa e 81 The pxoposed amendment makes a correotion in table 9 .2 .F to the indicator xange of the reactor pressure instruments. The reactor pressure indicators in the control room have a span of 0-1500 lb/in2g. Tho range for this instrument was listed in the technical specifications as 0-1200 lb/in2g. This amendment proposes'o correct the range in the 'technical specifications to reflect the actual xange of 0-1500 lb/in2g.

This change is for accuraoy of the technical specifications only. The calibrated range of he indicators in service is 0-1500 psig. This change is to update the teohnical specifications to the existing equipment. This change will not affect the operation, safety mar'gina, accident analysis, or overall safety of the plant. 0 ~ I V 9, Unit 2 Pa es 231 and 232 Revise the testing requirements for personnel air lock doors. During unit 2 reload 4, strongbacks were attached to the personnel air lock doors. These strongbacks allow testing the personnel air lock doors at an internal pressure of greater than 49.6 psig in accordance with 10 CFR 50, Appendix J. The technical specifications associated with this modification were inadvertently omitted from the reload technical specifications. The 10 CFR 50, Appendix J testing procedure is more xestrictive and has been administratively imposed. The revision, therefore, does not have an adverse impact on plant safety. This change makes the unit 2 specification consistent with those for units 1 and 3.

10. Unit 2 - Pa e 251A Revise the location of the Hydrogen-Oxygen system isolation valves.

Tile Hays-Republic H2-02 monitoring system was installed during the unit 2 reload 4 outage. The system has two outboard primary containment isolation valves per line. The proposed revision is being made to reflect the correct configuration in table 3 .7 .A and is clerical in nature only.

11. Units 1 and 2 Pa e 266 Unit 3 - Pa e 284 Table 3 .7 .H is being revised to remove penetration X-230 and add penetration X-219. Penetration X-230 is not a testable penetration and, therefore, cannot be tested. Penetration X-219 is a testable penetration that was inadvertently omitted from the technical specifications. However, X-219 is included in the surveillance instruction and is being tested.

The revision to the technical specifications is the corrective action for a deficiency identified by an internal audit.

12. Units 1 and 2 Pa e 358 Unit 3 Pa e 388 Revise the requirements on environmental qualification by deleting paragraph 6.9.A which currently requires that all safety-related electrical equipment be qualified by no later than June 30, 1982.

(The new deadline specified in 10 CFR 50.49(g) replaced the previous deadline of June 30, 1982.) Also, renumber paragraph 6.9.B as 6.9.A and revise by removing the date of December'1, 1980, as the deadline for having complete and auditable records available. ENCLOSURE 3 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA BFNP TS 197 BROGANS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 Units 1 2 and 3 Pa e 7 Descri tion of Amendment Re uest The proposed amendment would revise the technical specifications to incorporate the definition of a surveillance interval. The application was submitted in response to NRC's Generic Letter 83-27 dated July 6, 1983, which clarifies NRC's position on surveillance intervals. NRC's letter reiterated the current practice concerning surveillance intervals, The proposed amendment would add the definition of a surveillance interval as it was described in the generic letter. Specifically, the definition would include a provision which permits any surveillance interval to be extended by 25 percent of the nominal interval provided that the total time interval does not exceed 3 .25 times the speoified surveillance interval over any three consecutive surveillance intervals. Basis for Pro osed No Si nificant Hazards Consideration Determination NRC has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) . The examples of actions involving no significant hazards consideration include a purely administrative change to the teohnical specifications. The proposed amendment fits the stated example due to the fact that it conforms to the staff's current practices and the standard teohnical specifications. On this basis, we propose to determine that the proposed amendment involves no significant hazards consideration.

2. Unit 1 Pa e 31 Descri tion of Amendment Re nest The proposed change is to remove the letter 'C'rom 4.1, Surveillance Requirements.'o existing requirements are proposed for removal. No new requirements are proposed.

Basis for Proosed No Si nificant Hazards Consideration Determination NRC has provided examples of amendments not likely to involve significant hazards consideration (48 FR 14870) . An example of an action involving no significant hazards consideration is '(i) a purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.'he change proposed to page 91 of the Browns Perry unit 1 specifications is encompassed by that example in that the paragraph now denoted as 'C's 'ompatible with paragraph A. As presently configured paragraph after the paragraph lettered 'C'proposed for deletion). The 'B'ppears proposed revision does not in any way change the requirements of the approved technical specifications. 3, Units 1 and 2 Pa es 55 58 61 112 255 2 Unit 3 Pa es 57 60 63 109 267 294 Descri tion of Amendment Re nest The following would make changes to the technical specifications in the following areas.

1. Table 9 .2 .A The reactor low water level trip setpoint of 598 inohes above vessel zero initiates primary containment isolation.

Fox clarification of exactly what is isolated, 'Groups 2, 3, and added to the note. 6'ere

2. Table 9 .2 .A, notes for table 3 .2 .A, and associated bases A note was added to the trip level setting of the high radiation main steam line tunnel which provides a range for the setpoint. The sign was also deleted. The notes for table 3 .2.A and the bases were revised to reflect the revision. Table 4.2.A currently reads as that proposed for 9 .2.A.

9 . Table .2 .A The TIP actuation logic group number was revised from '2'o 3 '8.'IP-actuation is from the group 8 isolation logic. This ohange will make table 3 .2 .A consistent with tables 4.2 .A and 3.7.A.

4. The 3.7/4.7 bases were revised to more accurately describe the actuation logic associated with the group 7 isolation valves and to include the reactor low water level trip setting for the group 8 valves.

5 . Notes fox table 3 .7 .A - The reactor low water level (538 inches) trip setting was added to group 8. This signal isolates the TIP system. It was previously omitted from the technical specificatinns. I Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided .guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) . One example of actions involving no significant hazards consideration include: (i) a purely administrative change to tcchnical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature;.'he changes proposed in the application for amendment are enoompassed by the example in that they are administrative in nature and are being made for clarification (change No. 1), to achieve consistency (change No. 2), and to oorrect errors (change Nos. 3, 4, and 5). Since the applioation for amendment involves a proposed change that is similar to an example for which no significant hazards consideration exists, we have made a determination that the application involves no

• significant hazards considerations.

4. Unit 1 Pa es 70 and 358 Unit 2 - Pa es 70 231 232 251A and 358 Unit 3 Pa es 72 and 388

'Descri tion of Amendment Re nest The amendment would revise the technical specifications of the operating license to (1) reflect the aotual RHRSW pumps that initiate on the core spray pump start signals for units 1, 2, and 3, (2) comply with 10 CFR 50.49 on environmental qualification for units 1, 2, and 3, (3) comply with 10 CFR 50, Appendix J on testing of the personnel air lock doors for unit 2, and (4) reflect the actual plant configuration . for the Hay-Republic H2-02 primary con ainment isolation valves for unit 2. Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided guidance concerning the application of the standards by providing examples of actions that are likely, and are not likely, to involve significant hazards consideration (48 FR 14870). Three examples of actions not likely to involve significant hazards considerations are: '(i) A purely administative change to technical specifications: for example, a change to achieve consistency throughout the teohnical specifications, correction of an error, or a change in nomenclature. (ii) A change that constitutes an additional 'limitation, restriction, or control not presently included in the technical specifications: for example, a more stringent surveillance requirement (vii) A change to make a license conform to changes in the regulations where the license change results in very minor changes to facility operations clearly in keeping with the regulations.'roposed change Nos. {I) and (4) are administrative revisions to correct errors. The revisions have no effect on either limiting condition of operations (LCOs) or surveillance requirements. Since this change is being made to correct errors, it is encompassed by example (I) of the guidanoe provided by the Commission. Proposed change No. (2) is encompassed by example (vii) in that the change is being made to comply with 10 CFR 50.49. Proposed change No. (3) is encompassed by example {ii) in that the strongbacks were installed on the personnel air lock doors to allow testing them at a higher pressure in accordance with 10 CFR 50, Appendix X. Since all of the changes to the technical specifications given are encompassed by an example in the guidance provided by the Commission of actions not likely to involve a significant hazards consideration, TVA has made a proposed determination that the application for amendment involves no significant hazards considerations. 5, Units I and 2 Pa es 78 Unit 3 - Pa e 81 Descri tion of Amendment Re nest The proposed amendment makes a correction in table 3 .2 .F to the indicator range of the reactor pressure instruments. The reactor pressure indicators in the control room have a span of 0-1500 lb/in2g. The range for this instrument was listed in the technical specifications as 0-1200 lb/in2g. This amendment proposes to correct the range in the teohnical specifications to reflect the actual range of 0-1500 lb/in2g. i~ + a Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided guidance concerning the appli'cation of, the standards by providing examples of actions that are not likely to involve significant hazards considerations (48 FR 14870) . One example of actions not likely to involve significant hazards considerations is a purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenolature. The proposed amendment is encompassed by this example since the revision is being made to correct an error in the instrument range for reactor pressure. Since the proposed amendment is encompassed by an example of actions not likely to involve a significant hazards consideration, WA proposes to determine )hat the proposed amendment does not involve a significant hazards cousideration. 6, Units 1 and 2 Pa e 266 Unit 3 Pa e 284 Descri Cion of Amendment Re nest Add penetration X-219 and remove penetration X-230 from table 3 .7.H. Basis for Pro osed No Si nificant Hazards Consideration Determination NRC has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) . The examples of actions involving no significant hazards consideration include '(;i) a purely administartive change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.'he proposed amendment fits this example because the proposed specification revisions reflect the actual testing requirements of the plant's configuration. Another example of the no significant hazards consideration is '(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications: for example, a more stringent surveillance requirement.'he proposed technical specifications fit this example since it reflects the addition of a ne<< testable pentration (X-219) to table 3 .7.H. On the basis we propose to determine that the proposed amendment involve no significant hazards consideration. A r e I