Text

Proposed Tech Specs Revising Action 21 in Table TS 3.5-2B,to Incorporate Statement Allowing 6 H to Trip an Inoperable Channel
	ML20126H211
Person / Time
Site:	Prairie Island
Issue date:	12/29/1992
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20126H167	List: ML20126H162; ML20126H200; ML20126H211;
References
	NUDOCS 9301050102
	Download: ML20126H211 (68)
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Attachment 2 Revision to Original Proposed Changes Marked Up on Existing Technical Specification Pages License Amendment Request Dated September 21, 1992 The existing and new pages affected by this License Amendment Request Revision are listed below:

EXTSTING PAGES NEW PAGES TS.1-5 TABLE TS.1-1 TS.1-7 TABLE TS.3.5-2A (Pages 1 through 3)

TS.3.10-1 TABLE TS.3.5 2B (Pages 1 through 8)

TS.3.10 2 TABLE TS.4.1 1A (Pages 1 through 3)

B.3.5 4 TABLE TS.4.1-1B (Pages 1 through 7)

B.3.10 1 TABLE TS.4.1-1C (Pages-1 through 4)

B.3.10 2 i

l l

~

9301050102 921229 PDR ADOCK 05000282 P .PDR l' -

TS.1 5 Mb94- 40/N/84 OPERABl.E - OPERABILIIX A system, subsys tern , train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Itaplic i t in this definition shall be the assumption that all necessary attendant instrumentation, controls, norinal 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 perfor: sing their related support function (s). )

When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperablo, or solely because its nortnal power source As inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting -!

Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s),

subsystem (s), train (s), cornponent(s) and device (s) are OPERABLE, or likewise satisfy the requirements of this paragraph.

The OPERABILITY of a system or conoponent shall be considered to be estab.

lished when: (1) it satisfies the Limiting Conditions for Operation in Specification 3.0, (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requireinents, and (3) its condition is consistent with the two paragraphs above.

Dyg(%T10NAlf MODE a HODE An OPERATI.ONAf# H0 DEL (1,#?,7HODE)fahall: correspondLt(anyfonghiclusive . .

combination?ofl cora rosetivity condition' power;1evelf and.. average) reactor.

coolent temperature;s[(ecifiedlin1Tablo TSsijlk l PilYSICS TESTS PilYSICS TESTS shall be those tests performed to measure the fundamental characteristics of the core and related instrumentation, PilYSICS TESTS are conducted such that the core. power is sufficiently_ reduced to allow for the perturbation due to the test and therefore avoid exceeding power-distribution limits in Specification 3.10,B, Low-power PilYSICS TESTS are run at reactor. powers less than 2% of rated power.

P&'iL OPERATION POWLOPERAT4C%ef-a-unl4-44-any--apent4+>g-comil-t4eighat, + e544a-when-4Ae reae44r e r that-utd4 --! = e r !4-leaWml-4 he - neu t : : fl un-power-unge-lus4ra-p.+nt*44en-tm14 ca t t ;;reater-than Asf RATED-THEFJ!AL PO"Eb.

TS.1 7 REV-86--10/3&oS BATED TilFRMAL POWER RATED TilERMAL POWER shall be the totel reactor core heat transfer rate to the reactor coolant of 1650 megawatts thertnal (MVt).

VFrunt A NG A-un14 -4 o-lu-tlia--RERIElJNG-ooi d144wi-+heu+

4. h. r4.-l e-fuel-4 * *-14 +*-+oas te r--v** *e l, 3,---W.-ves4,*1-heed-e1**u re- 4*o14 e-a r+.4***-t4.*n-f ull y4w ealenod--or-41 a head-4 e-4wev.+4.

3. Th *-n+ao4+ r-cool a n t-a ver a g*--t+mpe r a tu r e-4 *-4 *s4.-44 *a n-o r--equ a 14e 140*F, nnd-

^ The-4.orou-4xane4+nt-rat loit-of--t4.*-n+aator-coolant-*y*4wn-and-t4+e-n*fu*14ng uavity-4*-4,uff4o4*nt-o*-onoun+-.44 at-44io u or*-w,+nu4eblue-of-64 *-f*14 ewing co4*lltionn-l+-m*t +

a,---K A44r-o F b Bo resi.-ooiaw+nua14 e>@O40-pter P11'ORTAB1,E EVFE A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHIELD BUJJAULG_.U1TFGRITY SillELD BUILDING INTEGRITY shall exist when:

1. Each door in each access opening is closed except when the access opening -

g is being used for normal transit entry and exit, then et least one door 4 shall be closed, and

2. The shield building equipment opening is closed.

3. The Shield Building Ventilation System is OPERABLE.

TiiUTDOWN MARGlN SHUTDOWN MARGIN aball be the instantaneous arnount .of reactivity. by which;

1) the' reactor .ts saberitical

' or '

2) the reactor would be suberitical' fromi ttu present^ conditioti-~ assuming' all

~

rod- cluster. control assemblies are fully; inserted except _ f or the rod

' cluster' control' assembly of highest reactivity worth'which,in assum6d i t4 be fully withdrawn.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

TABLE TS.I.1 TABLE TS.I.1 OPERATIONAL MODES

~

REACTOR tRATED AVERAGE VESSEL llEAD*

REACTIVITY' TilERMAL' COOLANT C1hSURE BOLTS H@g - I.11LE @@lT10N ~ POWER._L TEMPERATURE FULIN TENSIONED 1 POWER OPERATION Critieni > 2n- HA YES

~

2 Il0T STAllDbY** . 3 Critic &l- 5 24 - ilA YES 3_ 110T SlWTDOWNM Suberitical NA_ a 350*r YES 4 INTERMEDIATE Subcritical' NA~ < 350'F LYES SilVTDOWN** h 200'F 5 COLD SilDTDOWN Subcritical= -NA < 200*F YES 6 REFUEL 1NO. NA* NA- lNA. :NO

- Boron concentration of tho- reactor _ coolmit~ system and the refueling cavity sufficient to ensure that the more_rentrictive of-the following conditions is not;

-n. Ke rg_$ 0.95, or _

b. Baron' concentration 1 2000_ ppm.

- -Prairlo Island specific'HODE title, not" consistent with Standard Technical Spocification MODE titles. MODE numbers are consistent?with Standard Technical Specification MODE numbers, 1

4:

4 1

TABLE TS.3.5-2A (Page 1 of 6)

REACTOR TRIP SYSTEM INSTFIMENTATION MINIMUM i TOTAL'NO. CHA*;NELS CHANNELS APPLICABLE FITNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

[- - 1. Manual Reactor Trip 2 1 2 1, 2 1 t 2 1 2 3(*), 4(*2, 5(*) '8

2. Power Range., Neutron Flux

a. High Setpoint 4- 2 3 1, 2 2

b. Low Setpoint 4 2 3 108, 2 2.

3. Power Range, Neutron Flux - 4- 2 3 1, 2 2.

High Positive. Rate

}' 4; Power Range, Neutron Flux,- 4 2 3 1,. 2 2 High Negative Rate

.5. -Intermediate Range, Neutron Flux 2- 1 2 10 ' , 2 3

6. Source Range,. Neutron Flux

a. Startup 2 1 2 2(*) 4

b. ' Shutdown 2 1 2 3(*), 4(*3, 5(*) 5

7. Overtemperature AT 4 2 3 1, 2 6 8 .. Overpower AT 4 2 3 1, 2 6 x-e E 2 $r -

e 3; g I (a) Lihen the' Reactor Trip Breakers are. closed and the Control' Rod Drive System is-capable of rod g withdrawal.: o.

Y (b) Below the F-10 (Lov..Setpoint Power Range. Neutron Flux Interlock) Setpoint.. O[>

(c) Below the.,P-6 (Intermediate. Range Neutron Flux Interlock) Setpoint.

t 4

, . - - , , _. pr. .,

e c. v., .

r-, , -.

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

ll 4 . t o y TABLE TS.3.5-2A (Page 2 of 6) ,

REACTOR TRIP SYSTEM INSTRUMENTATION ,

i-MINIMUM i TOTAL NO. CHANNELS CHANNELS APPLICABLE !

0 i FUNCTIONAL UNIT. OF CHANNELS TO TRIP OPERABLE MODES ACTION I 1 I

, - 9. Iow Pressurizer Pressure 4 2 3 1 6 l t

10. High Pressurizer Pressure 3 2 2 '1, 2 6 [

i

- - 11. Pressurizer High Water level 3 2 2 1 6 '

e j

, - 12. Reactor Coolant Flow Low 3/ loop.. 2/ loop 2/ loop 1 6

13. Turbine. Trip- !

i ,

I f

i a. Low AST Oil Pressure- 3 2 2 1 6 -

I~ .

l

b. Turbine Stop valve Closure 2 2 1 1 6 ;

4 14.-Lo-Io Steam Generator- 3/SG 2/SG in 2/SG in 1. 2 6 !

each SG

. Water. Level any SG !

15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1 11 !

'g 11 and 12 (Unit 2: 21 and 22) both bus !

]

buses l E9s

aeW E !

"U t

. av l' !

v. :

n b

3 g, .w- e , v: e --,r- e -u r**si' + == e + -q-e -**-- + * -- -- --------++~2 - - - - - - - '

. TABLE TS.3.5-2A (Page 3 of 6)

), REACTOR TRIP SYSTEM INSTRUMENTATION 4

1 MINIMUM

$ TOTAL NO. CHANNELS CHANNELS APPLICABLE RTNCTIONAL UNIT ' OF CHANNELS TO TRIP OPERABLE MODES ACTION I '16. Loss of Reactor Coolant Pump l

a. RCP Breaker Open 1/ pump 1 1/ pump 1 1

b. .Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one 1 11

, both bus

, buses i:

17. Safety Injection Input 2 1 2 1, 2 7 from ESF

18. Automatic Trip and Interlock Logic '2 1 2 1. 2 7-

2 1 2 3(*), 4(*). 5(*) 8

19. Reactor Trip' Breakers 2 l' 2' 1, 2 9 l 2 1 2 3(*), 4(*). 5(*) B

i. 20. Reactor Trip Bypass' Breakers' 2 1 1 (d) 10 L

]

f (a) When'the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod h withdrawal... gg

- (d) When the Reactor Trip Bypass Breakers are racked in and closed for. bypassing a Reactor Trip Breaker h and the Control Rod System is capable of rod withdrawal.. y ow 1' A3

+ >

y- y- p r. , + y , 4 - -

TABLE TS.3.5-2B (Page 1 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMrM TOTAL NO. CHA57;ELS CHANNELS APPLICABLE FUNCTIONAL UNIT ~ OF CHA*!NELS TO TRIP OPERABLE MODES ACTION 1.- SAFETY INJECTION

.a. Manual Initiation 2 1 2 1,2,3,4 23

'b. High Containment Pressure 3 2 2 1,2,3,4 24-

c. , Steam Generator Low Steam 3/ Loop 2 in any 2/ Loop 1,2,3N 24 Pressure /Ieop Loop

d. Pressurizer low Pressure 3 2 2 1, 2 , 3(*' 24

e. Automatic Actuation logic 2 1 2 1,2,3,4 20 and Actuation. Relays

2. CONTAI6 MENT SPRAY

a. Manual Initiation 2 2 2 1,2,3,4 23 bi Hi-H1' Containment Pressure 3 channels 1 sensor 1 sensor 1,2,3,4 21 with 2. per per sensors per channel channel channel in all 3 in all 3 channels channels NQY

c. Automatic--Actuation Logic and. 2 1 2 1,2,3,4 20 #$ $

Actuation Relays o-

- t ,W (a)' Trip function may be blocked in this MODE below a Reactor Coolant System Pressure of 2000 psig,

TABLE TS.3.5-2B (Page 2 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRLMENTATION MINIMEM TOTAL NO. CHANNELS CHANNELS APPLICABLE

- FLUCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

3. CONTAINMENT 'ISOIATION

a. . Safety Injection ' See 1 above for au safety injecten imtiating functie-is and reqmrements.

b .' Manual 2- 1 2 1,2,3.4 23

c. Automatic' Actuation Logic and- 2' 1 2 1,2,3,4 20 Actuation Relays 4 CONTAINMENT VENTIIATION ISOIATION

a. ' Safety Injection See 1 above for an Safety Injecuon initiating functkxis and requiremeras.

b. Manual 2 1 2 (b) 22

c. ' Manual Containment Spray ' See 2a aluve for Manual Containment spray ym.~e.

d.-' Manual Containment Isolation . See 3b above fer Manual Containmera Isolatim m,usrements.

e. High Radiation in Exhaust Air 2 1 2 (b) 22

f. Automatic Actuation I.ogic 2 1 2 (b) 22 and Actuation Relays en >

< a tz i

SC wg.

- .(b) Whenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in Sh operation. cmy 2

g-1 r,._ + -gw,, . , - a ww%. . ,.~ _ -4. . - - - - - a w o ., y, ,.g- in. . . . $w, # - -

t

TABLE TS . 3. 5- 2B (Page 3 o f 8 )

1 ENGINEERED SAFETY FEATl'RE ACTUATION SYSTEM INSTRL' MENTATION j MINIMlM TOTAL NO. CHANNELS CHANNELS APPLICABLE l FUNCTIONAL UNIT OF CHAVNELS TO TRIP OPERABLE MODES ACTION

!L

5. STEAM LINE IS01ATION l .a. _ Manual 1/Imop 1/ Loop 1/Ioop 1 , 2 , 3 ten 27

b. Hi-Hi Containment Pressure' 3 2 2 1 , 2 , 3(*) 24 4.

c Hi-Hi Steam Flow with Safety

Inj ection

l. .Hi-Hi Steam Flow 2/ Loop 1 in any 1/Ieop 1 , 2 , 3(*) 29 j IhoP l 2. Safety Injection see 1 atee for mII safety Irmten ind2 sting functwas and 1%aas.

d. Hi . Steam Flow and 2 of 4 Low

j. Tave with Safety Injection:

i. 1.. Hi Steam. Flow 2/ Loop 1 in any 1/Ioop 1, 2, 3(d3 29 Loop

{ .

f 2. Tave 4 2 3 1 , 2 ,. 3(d2 24 1

3. Safety Injection' see I atee for au safety injectwo maiarmg functmas and %=ements

[ n~n G75 (c) When either main steam. isolation valve-'is open.

~

TE y .

"M d-

. (d) When' reactor coolant- system average temperature is greater than 520*F and either main steam isolation c-

~ valve is open. 7" cm w NP CII E

,-.g a --g - e-m-g+ g 'lw? _i.ee p. , - + . - - m-*wya = u* *r'"-*-*' @f* "e +1' Mr"9 e P ~+ *s'am'2" T' --* e6 W.w- e et-*+-6'- at a

i TABLE TS.3.5-2B (Page 4 of 8)

- ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRlNE';TATION l MINIMLN TOTAL NO. CFJJC;ELS CHA'RELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. . STEAM LINE ISOLATION (continued)

e. .

Automatic Actuation Logic and 2 1 2 1 , 2 , 3t*> 25

. Actuation Relays 1

6. FEED'='ATER IS01ATION -

7. Hi-Hi' Steam Generator Level 3/SG 2/SG in 2/SG in 1, 2 24 any SG each SG

b. Safety Injection See 1 atee for an safety injecten initiating functens and .e.~.a

c. Reactor Trip with 2 of 4 Low Tave (Main Valves only):

1. Reactor' Trip 2 1 2 1, 2 28

2. Law.Tave 4 2 3 1, 2 24

d. Automatic; Actuation Logic 2 1 2 1, 2 28 and Actuation Relays ,

E2$

9 5; "d

(c) 'm' hen either main. steam isolation valve is open. Sh-OIm D*

c i f.

1 TABLE TS . 3. 5- 2 B (Page 5 of 8)

ENGINEERED SAFETY FEAR'RE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL l' NIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

!l 7. AUXILIARY FEEDWATER j'

)

a. Manual 2 1 2 1,2,3 26

b. Steam Generator Low-Low 3/SG 2/SG in 2/SG in- 1,2,3 24 j.. Water Level any SG each SG

c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29 11~and 12 (Unit 2: 21 and 22) both bus

[- (Start Turbine Driven Pump buses

only).

q d. Trip of. Main Feedwater Pumps l- 1. Turbine Driven 2 2 2 1, 2 26 j 2. Motor Driven 2 2 2 1, 2 26 i

e. Safety 'Inj ection' see I above for an Safety Injecuon inkia: mg functbas and .y a

.f. Automatic Actuation Logic 2 1 2 1,2,3 20 and Actuation Relays E2$

e ? 5; i

v. s I

o?

my i

m Ot 4

3 ',r-r-w y *e- -

w- w ..+..%

e - . - - m>.. . . . , _ . . . . , _ . .

4

TABLE TS.3.5-2B (Page 6 of 8) 1 ENGINEERED SAFET7 FEATURE ACTUATION SYSTEM INSTPWENTATION i

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE '

FUNCTIONAL UNIT OF CHANNELS - TO TRIP OPERABLE MODES ACTION I' [

.; 8. : LOSS OF PO'.'ER i .

, .a. Degrade'd' Voltage. ,4/ Bus 2/ Bus 2/ Bus 1, 2, 3. 4 29 I 4kV Safeguards Bus (2/ phase on 4

(1/ phase 2 phases) on 2 ,

i phases) i i

-b. Undervoltage' 4/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 29 !

j 4kV Safeguards Bus (2/ phase on (1/ phase ,

2 phases) on 2

phases). ,

I

, f i i

e - H '(

I re: m>

< ll D W or e gr2 u i

j-5 m 64 i tra i 0 -

co v.

se e w

4 W i

i 4

tr w ,- tv- w + -,y - a n w w r ~rr- ~s,, , -e w-.-.%_ _ - , - - - . 4- . - w

TABLE 3.5-2B (Page 7 of 8)

Action Statements ACTION 20: With the number of OPERABLE channels ACTION 23: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, restore tt. toperable Channels, restore the inoperable channel to OPERAB1J mi<cus within 6 channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at i- <

HOT SHUTDOWN hours or be in at least HOT' SHUTDOWN within the next 6 hor. and in COLD within the next 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 ; SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

'however, one channel may be bypassed for up to 8 hour:.; for surveillance

. testing per Specification 4.1, provided ACTION 24: With the number of OPERABLE channels-the other channel is OPERABLE. one less than the Total Number of Channels, operation in the applicable HODE may proceed provided the following ACTION 21: .Vith the number of OPERABLE channels conditions are satisfied:

less than the Total Number of Channels, operation may proceed provided the a. The inoperable channel is placed in inoperable channel (s) is placed in the the tripped condition within 6 tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and hours, and, the Minimum Channels OPERABLE requirement is met. The inoperable b. The Minimum Channels OPERABLE channel (s) may be bypassed for up to 4 requirement is met; however, the hours for surveillance testing per inoperable channel may be bypassed Specification;4.1. for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.1.

ACTION 22: With the number.of OPERABLE channels less than the Total Number of Channels, operation may continue provided the ggg containment purge supply and exhaust < m es

_ valves are maintained closed. SE

'n4 o-

~mw av h

w

i TABLE 3.5-2B (Page 8 of 8)

M ien Statements I ACTION 25: With the number of OPERABLE channels ACTION 28: With the number of OPERABLE channels one less thn the Total Number of one less than the Total Number of Channels, restore the inoperable Channels, restore the inoperable channel to OPERABLE status within 6 channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least. HOT SHUTDOWN hours or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation in within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, one HOT SHUTDOWN may proceed provided the channel may be bypassed for up to 8-main steam isolation valves are closed, hours for surveillance testing per if not, be in at least INTERMEDIATE Specification 4.1, provided the other-SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . channel is OPERABLE.

However, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance

' testing per Specification 4.1, provided ACTION 29: With the number of OPERABLE channels the-other channel is OPERABLE. less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following ACTION 26: With the number of OPERABLE channels conditions are satisfied:

one less than the Total Number of Channels, restore the inoperable a. The inoperable channel (s) is placed channel'to OPERABLE status within 48 in the tripped condition within.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be~in at least HOT SHUTDOWN hours, and, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in'at least INT ~luiEDIATE SHUTDOWN within the b. The Minimum Channels OPERABLE following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . requirement is met; however, the inoperable channel (s) may be b37assed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ACTION 27: With the number of OPERABLE channels surveillance testing of'other one less than the Total Number of channels per Specification 4'.1' x-e L Channels, restore the inoperable channel.to OPERABLE status within 48 2h o in ,

hours or be in at least HOT SHUTDOWN co s within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and close the o?

associated valve.

O$

E u > 1

~

'f =

TS.3'10 1 RE" 92 3/13/o0 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations, Obiective To assure 1) core suberiticality after reactor trip, 2) acceptable core power 'l distributions during POWER 40PERATION, and 3) limited potential reactivity insertions caused by hypothetical' 6ontrol rod ejection.

Specificati_oa A. Shutdown Marcin T4a chutdeen : rg!:. e4446-a14*wanc: fe r : ctuch cent 4c1 t ed nuerably ch 11 e*cced the--epp14 cable 'zelue chevn 1: Figur Ts.3.10-1 under 211

-dy- etate-operwng-em+4444ene -except r fer Physics TESTS , free acro.-to fu14-powa* r.-4nelud4 &fec tn of 2ni 1 peeer-41stributien. S e chutdown.

mar in-as-used-lier 7A-f4ned-ae-446*-amount by hich the recete: cere

-be ceb ritical at !!OT SHUTD0"" cend! tien if 211 centrel red-assemh14se-were--ct-4ppedr-assumin8-4,het,-446e-4d t hect uerth-sentr44-.+od mM4y ree!:ad-f*14y-w144* drawn, and--cosu=!ng ne=ehange: !r nenm+-os baron-concentrat4+n, 1h Reactor' CoolantrSyst'eFAVsfWWTemphratureT2004 TheTSliUTDOWN MARCINTsha1Ub6fgrehtWithaE66equalEtWtheTapplicable value ! shownfin : Figure ETS f 3 ;10 (li when 91 nil 10T ?STANDBYJ withik/rs91'. 0]I '~-

andtwhen inlHOTyliUTDOWrCandAINTERMEDIATEJSilyTDoky1 2{ Rsa e E6r" C6o Tan FSH t'sif Ave rus? TsisibW'it'ufiN 200 f F Th 65 SHUTDOWN f MARGIN ish511'Ebi4fektWBtihTnT6

~ ~ ~ ~ ~ ~ ~ ' ~ ' " ~ F bi['GilYE6TrinkK"~IwhilD

^ " ' ' ~ ~ " ~ " " ~ 16 COLDiSii,UTDOWNi 31;i Wi th7 ths S HUTDOWN :MARGI N i l e'ss? thain theta p p'lic abl ellisi1G13EdifindMn 3 .10,' A , lio n 3 ,10 ; AK. 23 aLove hwi thin 41 Mainute a Fi nit ia to tb o ra ti o ~'

restore SifUTDOWN; MARGIN;toiw.ithirgthofapplicablejltmlt$ '" " ~niti B. Power Distribution Limits

1. At all times, exce channel factors,-Fpt q during and Flow ta,power PliYSICS as defined TESTING, below and in the measured bases,hot shall meet the following limits:

RTP F"q x 1.03 x 1.05 s (Fo / P) x K(Z)

RTP F"6a x 1.04 s F6 u x (1+ PFDH(1-P)]

where the following definitions apply:

- Fo is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

RTP Fan is the Fan limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for F H N8Pecified in the CORE OPERATING LIMITS REPORT.

- K(Z) is a normalized function that limits F n (z) axially as specified in the CORE OPERATING LIMITS REPORT.

7 ir th; cere Night Icentie-P ic the free &ie- ef RATSD-THERMAL "0"r" nt hich the ceve in eperating.

In-the-F( l im i t- de t e rr i n t i c: ube- P :? . 50, cet ? - 0,50.

TS.3.10 2 REV 92 1/44/40 3.10.B.1.

Z.is:thelcore height? location; n Pfis : ths frac ti6n?of' RATED lTHERMAlflPOWERTat~whicif thW core)

~ operating 4 - In;the F 8g.11mit/ determination when Pjs0,50f setiP -10,50;

- F"o or FN y is defined as the measured Fn or Fui respectively, with the smallest margin or greatest excess of limit.

- 1.03 is the engineering hot channel factor, F g, E applied to the measured F"o to account for manufacturing tolerance.

- 1.05 is applied to the measured F"n to account for measurement uncertainty.

- 1.04 is applied to the measured FN g to account for measurement _

uncertainty.

2. Hot channel factors, F"o and F"an , shall be measured and the target flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 ef fective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10% or more of RATED THERMAL POWFR.

F"g (equil) shall meet the following limit for the middle axial 80%

of the core:

RTP N

Fg (equil) x V(Z) x 1.03 x 1.05 s (Fn / P) x K(Z) where V(Z) is specified in the CORE OPERATING LIMITS REPORT and _

other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3,10.B.1, reduce reactor power and the high neutron flux trip set-point by 1% for each percent that the measured HF g or by the factor specified in the CORE OPERATING LIMITS REPORT for each percent that the measured FN g exceeds the 3.10 B.1 limit. Then follow 3.10.B.3(c).

(b) If the measured FN g (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions:

1. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium configuration for which Specification 3.10 B.2 is satisfied, or

2. Reduce reactor power and the high neutron flux trip setpoint by 1% fer each percent that the measured F"g (equil) x 1.C 3 x 1.05 x V(Z) exceeds the limit.

l - _ _ _-

m - -- _. . _ .. ..

M TABLE TS.4.1-1A (Page 1 of 5)

REACTOR TRIP' SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE- . TEST TEST SURVEILIANCE IS REOUIRED L : Manual Reactor Trip. N.A. N.A. R u3) N . A .. 1, 2, 30) , 4(1), 5u) ;

1 :

2. Power Range, Neutron Flux ;

a) High Setpoint- S DD 7) Quer R 1, 2 ;

j., g(6, 7) -4 q(7,8) ,

b) Low Setpoint S R(7) S/U U7) R 1(3),2 i

3. Power Range, Neutron' Flux, N.A. R(7) Q R 1, 2 High Positive Rate

- 4. 4howerRange,NeutronFlux- 'N.A. R(7)- Q R 1, 2 High Negative Rate 'l

5. Intermediate Range,. ,S 'R(7) S/U") R -1(3), 2-Neutron Flux 6 .. Source Range, Neutron Flux

a. :Startup- S- R(7) S/U") R 1(2) m 4.

b. Shutdown S. .R(7) . Q U8) R 30) , 4u), 5(1) ~ E 2 $. -I 33: --

g ,4 7; ;Overtemperature AT S- R R 1, 2

~

Q-ww v.

-'8. . Overpower AT. ' S -. R- Q. R 1, ~ 2. >- ;

I

. . . . . - ., ,- ~

- ..c,.- 4 . , , . . - .

- i.E .{; [ i I;f{tj. fji[ i1 !t * , 2 ?

3 4N7P1D

> 1 4n

+t+ [~ewy 3 . -.

h 0" v%# . -

A M2M.S

'cM<.

D .

E R -

I L.

U Q

E R

H S i C I I .

H E V C N

R A O L F L I

S E E V 2 2 D R S O U , ,

T MS I 1 1 1 'l 1 1 1 N .

E M E E S R N T . . . . . . . .

I O S U P E A. A. A. A. A. A. A. A.

O S T N N N N N N N N E E R R E

C N

A 2 1

3 1

) L L 1 2 5 L A , .

I N s s f E O T t t o V I S U U R T E / /

2 U C T 'Q Q Q Q S S Q" Q S N .

e N R g

a O I

P T

( A E A T T N A

'l E R R R R R R R 1

M B R R U I 4 R L T A E S C L N B I A M T E T K A. A. A.

S C Y E S S S S N N S N S H C ,

P c I

R ,

T l R e O v T e e C e r l w e s A r u o r u u s r L u e B E

R s s

s e

e t w s

s l v' r o P e r a o e a t C r P W l r V a R P F P r m r p e V e . h' r g t l o" n K e z i n i t e 4 -

z i H a . . O S G r p e T

I i

r u

u s

're l o

o i

r S T e 'e nr ml ae g

a N s s z C T.. A i u ev t U s e i .

bs. t e l e r r r e. w. ro SL o L r P u o n o ul v A P s t i L TC or .r N h s c b Le e O w g e a .r t d I o i r e u . . oa n T L H P R .T a b L m U C

N '. .

. . . . . m U .

0 1 2 '3 4 5 F 9 1 1 1 1 1 - 1

,e p:'i; g. ' a 4i . i ,i j ; i; . , { . t1 !}; j4 4 ) ;j4 . :;

TABLE TS.4.1-1A (Page 3 of 5)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED FUNCTIONAL UNIT

16. Loss of Reactor Coolant Furp N.A. R S/U(') N.A. 1

a. RCP Breaker Open Underfrequency 4KV Bus N.A. R Q N.A. 1 b.

14 . A . N.A. R N.A. 1, 2

17. Safety Inj ection Input

18. Automatic Trip and Interlock N.A. N.A. MCS) R 1,2,3 0) , sti) , 5")

Logic N.A. N.A. Mt s.12) R 1, 2, 3"), 4"), 50) 19 Reactor Trip Breakers

20. Reactor Trip Bypass Breakers N.A. N.A. M"' ) Ri l" See Note (16)

NSY

<#en#

,b 5'

l . .. ..

l TABLE'TS.4.1-1B (Page 1 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL RESPONSE M' ODES FOR L'HICH CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED FUNCTIONAL UNIT:

1. SAFETY INJECTION.

a. Manual.' Initiation N.A. N.A. Rf1) N.A. 1, 2, 3, 4 R .Q N.A. 1, 2, 3, 4

b. High Containment Pressure S Steam Generator Low Steam- S R Q N.A. 1, 2, 3(2) c.

Pressure / Loop R Q N.A. 1 , 2 , 3<2)

d. . Pressurizer Low Pressure S Automatic Actuation logic N.A. N.A. M(33 N.A. 1,2,3,4 e.

and Actuation Relays

2. CONTAINMEN' aPRAY-N.A. N.A. R N.A. 1,2,3,4

a. Manual! Initiation

b. Hi-Hi Containment S R -Q N.A. 1, 2, 3, 4

' Pressure Automatic Actuation Iogic N.A. N.A. M(3) N.A. 1,2,3,4 c.

.and Actuation Relays

.EQ

- g

, ., y

=

i;;;

L ,

--_ , , g

. TABLE TS.4.1-1B (Page 2 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMESTS FUNCTIONAL RESPONSE MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED -

3. . CONTAINMENT ISOIATION

a. . Safety Injection See 1 above for all Safety injection Surveillance Requirements

b. Manual N.A. N.A. R N.A. 1,2,3,4

c. Automatic Actuation' Logic N.A. N.A. M(3) .N.A. 1, 2, 3, 4 and: Actuation Relays 4 CONTAINMENT VENTIIATION IS01ATION

=a. Safety Injection See 1 above for a!! Safety Injection Surveillance Requirements

b. Manual N.A. N.A. R N.A. See Note (7)

c. Manual Containment Spray See 2a above for all Manual Containment Spray Suneillance Requirements

d. Manual Containment ..' See 3b above for all Manual containment Isolation Surveillance Requirements Isolation e High Radiation-in ' Dt s) R M N.A. .See Note (7)

Exhaust-Air

. 'f . : Automatic Actuation-Logic N.A. N.A. M(3) N.A. See Note (7) and Actuation Relays xn4:

'bh

%W Ap ee G-

A TABLE TS.4.1-1B'(Page 3.of 7)'

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS -

FUNCTIONAL : RESPONSE MODES FOR TTrf1CH .

FUNCTIONAL UNIT- CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

5. STEAM LINE IS01ATION

a. Manual. N.A. N.A. R N.A. 1, 2, 3(*)

b. Hi.-Hi' Containment S R Q N.A. 1 , 2 , 3(')

Pressure

c. 'Hi-Hi Steam Flow with Safety Injection

1. Hi-Hi Steam Flow S- R Q N.A. 1 ,' 2 , 3(')

'2. . Safety Injeetion see I above for a!! safety Injection survei!!ance Requirements

d. H1' Steam- Flow and 2 of- 4 Low T,y, with.. Safety Injection

1. Hi Steam Flow.. S R Q N.A. 1 , 2 ,- 3!')

2. Tave! S R Q N.A. 1 , 2 ,- 3(5)

3. Safety Injection see I above for all safety injection surveillance Requirements

e. Automatic Actuation Logic N.A. N.A. M(3) - .N.A. 1 , 2 , 3!')

and Actuation Relays- *e - H :

EN 9 5; -

w ;;;

.O -

5

- P% p :

-U 7..

_ _ _ _ ._.____m _--_ wr w ' -'e--'y -

y w-

+v y

4 TABLE TS.4.1-1B (Page 4 of 7)

, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

6. FEEDWATER. ISOLATION

a. .Hi-Hi Steam Generator S R Q N.A. 1, 2 Level

b. Safety Injection see 1 above for all safety injection surveinance Requirements

c. Reactor. Trip with 2'of 4' Low T y. '(Main Valves

. Only)

1. . Reactor Trip N.A. N.A. R N.A. 1, 2'

' 2 .' Tave- .S -R Q N.A. 1, 2

d. Automatic Actuation Logic ~ N.A. N.A. M(3) N.A. 1, 2 and Actuation' Relays wn4; a k w S:5;-

.>n.

o*

, m .4 .

i !

t.. . , t n*s- r-r-' -

1 r ' r e i.4 e 2 s r

= . .

J i

l' TABLE TS.4.1-IB (Page 5 of 7)

ENGINEEPED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT, CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED :

7. AUXILIARY FEEDWATER

a. Manual N.A. N.A. R N.A. 1,2,3

b. Steam Generator Low-Low S R. Q N.A. 1,2,3

~ Water Level

c. Undervoltage on.4.16 kV N.A. R R N.A. 1, 2 Buses ll and 12-(Unit 2:

21'and 22)-(Start Turbine-Driven Pump'only)

'd. Trip of Main Feedwater

' Pumps l- 1. Turbine Driven N.A. N.A. R- N.A. 1, 2 i

2. . Motor-Driven N.A. .N.A. R N.A. - 1, 2

e. Safety Injec' tion .' See 1 above for a!! Safety injection Surveillance Requirements

.f. -Automatic Actuation Log'ic. 'N.A. N.A. M(33 N.A. 1,2,3

.and Actuation Relays

o m s k

%R .

O +

Pb ,A .

en i-1

s. , ,-.c. . , ,

TABLE'TS.4.1-1B (.Page 6 of 7) ,

ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS- _

FUNCTIONAL RESPONSE MODES FOR k'HICH RINCTIONAL UNIT- CHECK CALIBRATE TEST - T_EH SUR'.*EILIANCR IS' REOUIRED

. 8. IASS. OF POWER

a. ' Degraded Voltage N.A. R M N.A. -1, 2, 3, 4 ,

4kV Safeguards Bus

b. Undervoltage N.A. R M N.A. 1,2,3,4 4kV Safeguards Bus.: ;

1

'S b

T- y a

i i' p:r n .g.

. tus N > ' .

< m .. tn .

;l g;;

. B..

..tp .

h, 1

. > .,c,. e , ~-, , es w , r --

. . - +

a

]

] 1..

-}4i o .

TABLE'4.1-1B (Page 7 of 7)

TABLE NOTATIONS-FREQUENCY' NOTATION NOTATION, FREQUENCY ,

S'- Shift D .-Daily -

M .. Monthly- ,

Q Quarterly.'

R- Each Refueling Shutdown "

l N.A Not Applicable TABLE NOTATION (1)- One manual: switch shall be tested at each (7) ~Uhenever CONTAINMENT INTEGRITY is required refueling on'a STAGGERED TEST BASIS. -:and either of the containment purge systems ;

are in operation. '

.(2)

' Trip function may be blocked,in this MODE below a reactor coolant system. pressure of t 2000 psig. :

(3) Each train shall'.be tested at least every-two months'on a STAGGERED TEST BASIS. t

-(4) When either main steam isolation' valve is open.

(5). When reactor coolant system average"

-temperature-is greater than 520*F and either en m > ;

+

-main steam isolation. valve-'is open.

{

-(6) See Table'4.17-1.-

dJ i $ h, .,

.x ;

4

_ _ ,n . , . . - , , , .- _ , n.

TABLE TS.4.1-IC (Page 1 of 4)

MISCELLANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR VHICH

' FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED,

1. Control Eod' Insertion Monitor M' R S/Um N.A. 1, 2

'2. Analog Rod Position S- R -S/UW- N.A. 1, 2 , 3(2) 4(2), $(2)

3. Rod Position. Deviation M N.A. S/U(1) N.A. 1, 2

' Monitor

4. Rod Position Bank Sm N.A. N.A. N.A. 1, 2 , 3<2) , 4(2) 5(2)

Counters

5. ' Charging Flow S R- N.A. N.A. 1,. 2, 3, 4

6. -Residual Heat Removal S R N.A. N.A. 4N Sm , 6(s)

Pump Flow

7. Boric Acid Tank Level- D RN MN N.A. 1,2,3,4

-8. . Refueling; Water Storage' 'W R' M N.A. 1,2.3,4 Tank ~ Level

9. Volume' Control Tank- S, R- N.A. =N.A. 1,2,3,4'

'10. Annulus ~ Pressure N.A. R- R N.A. See Note (10)L (Vacuum: Breaker)' $Q $.

- < m to co t- -

11. Auto Load Sequencers 'N.Ai N;A. M- N.A. 1,-2,.3,.4

.[" d 12.' Boric-Acid'Make-up Flow

~

,N.A. R- ,

N.A. N.A. 1,'2, 3, 4 %[,6

' Channel 37 5

_ _ __ _ . . mm _ _ _ _ _ . . _ _ _ _ _ .

l' . h i

4 TABLE TS.4.1-1C (Page 2 of 4) l MISCELIANEOUS INSTRUMENTATION SURVEILLANCE REOUIREMENTS

~

FUNCTIONAL RESPONSE MODES FOR L'HICH ..

, ' FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

13. Containment Sump A, B and C N.A. R R N.A. 1, 2, 3,'4 Level .

14. Accumulator Level and S R R N.A. 1, 2,'3, 4 Pressure

-15. Turbine First S tag. t S R M .. N.A. 1

. Pressure.

16. Emergency Plan' Radiation M R M N.A. 1,2,3,4, 5, 6 '

Instruments (8)

, 17'. Seismic Monitors: 'R R N.A N.A. 1, 2, 3, 4, 5,'6

'18. Coolant Flow - RTD. S R M N.A. 1 , 2 , 3(5)

Bypass Flowmeter

-t

.19.~CRDM Cooling Shroud- S N.A. R N.A. 1, 2 , 3tz), 4tz), 5tz) .[

20.' Reactor. Cap Exhaust Air. S N.A. R' 'N.A. 1,2,3,4 Temperature.

21. Post-Accident Monitoring M R N.A. .N.A. 1, 2 - ;

Instruments ,e n e :

mc>

(Table TS.3.15-1)(7) #$omE. .I

-22. Post-Accident Monitoring 'D R M .N.A. 1, 2- ,d Radiation Instruments 4:.

'(Table.TS.3.15-2)' oL~

".?

,. c

.[

w I

~ .. c_ .

> - . . . . .~ .o . - . _ . . .. . . -

.i

. TABLE TS.4.1-1C (Page 3 of 4)

MISCELLANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS .

i RTNCTIONAL RESPONSE MODES FOR VHICH

- FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

- 23. Post-Accident' Monitoring. M R N.A. N.A. 1, 2 Reactor Vessel Level ;

Instrumentation (Table TS.3.15-3)-

24. Steam Exclusion Actuation W Y M N.A. 1, 2, 3 .

s 25.-Overpressure Mitigation N.A. R R N.A. 4W 5 ,

26. Auxiliary Feedwater- .N.A. R R N.A. l', 2, 3 Pump. Suction Pressure i

27.' Auxiliary Feedwater. .N.A. R R N.A. 1,2,3 .[

Pump Discharge Pressure i

28. NaOH. Caustic Stand Pipe 'W R M N.A. 1,2,3,4 Level

29. Hydrogen Monitors S Q M N.A. 1, 2 l

30. Containment Temperat.2re - M R N.A. N.A. 1, 2,.3, 4' Monitors ;

o - .4 i

31. Turbine Overspeed' N.A. R' M' N.A. 1 m ~J >

Protection Trip Channel #$amE -

W +3 .-

to o*

M ,R i S~. n*

y

' O J

y -

$ (- ,N _ p*; . * < m y a %-m a- e. 4 b e't.a @e-4e-<

1 1

TABLE 4.1-1C (Page 4 of 4)

TABLE NOTATIONS FREQUENCY NOTATION.

NOTATION- FREOUENCY S Shift D Daily W. Weekly M Monthly

.Q Quarterly S/U Prior to each startup 1Y- Yearly R' Each refueling shutdown N.A. Not applicable TABLE NOTATION -

~

Prior to-each-startup following shutdown in (7) Except for containment hydrogen monitors (1) which are separately specified in.this table.

excess of.tvo days if not done in previous 30 L ' days.

(8) When RHR'is'in operation.

(2) 'When the reactor trip system bteakers are closed and the control rod drive system is (9) When the reactor coolant' system average:

' capable'of rod withdrawal.' temperature is less than 310*F.

Following rod motion in. excess of:six inches (10) Whenever.. CONTAINMENT INTEGRITY'is required.

(3) when - the computer is. outE of service.

(4) ! Transfer logic to Refueling Water Storage . h ]' f '

l Tank. gg

' E! -

_ (5). [When~either main steam isolation valve.'is , F,.

cpen. .

4D:-T L (6) Includes those instruments' named in-the-' '

l emergency procedure.

" ' '' '"' ~ ' ' ' ' ' ' ' " '

a: ., ,,r, ,-

B.3.5-4 REV 91 10/N/M 3.5 JllSTRifMENTATION SYSTEd Bases contimied Limiting Instrument Setpoints (continued)

4. The steam line low pressure signal is lead / lag compensated and its set-point is set. well above the pressuce expected in the event of a large steam line break accident as shown in the safety analysis (Reference 3).

5. The high steam line flow limit is set at approximately 20% of nominal full-load flow at the no-load pressure and the high high steam line flow limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break -

accidents. The coincident low T,y setting limit for steam line isolation initiation is set below its hot shutdown value. The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

6. Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is discussed in the Bases of Section 2.3 of the Technical Specification.

7. liigh radiation signals providing input to the Containment Ventilation Isolation circuitry are set in accordance with the Radioactive Effluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the SITE BOUNDARY.

8. The degraded voltage protection setpoint is 294.8% and $96.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the minimum degraded voltage setpoint. Tha maximum degraded voltage setpoint is -

chosen to prevent unnecessary .tuation of the voltage restoring scheme at the minimum expected grid voltage. The first degraded voltage time delay of 8 1 0.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients (i.e.,

motor starting and fault clearing). It is also longer tnan the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded voltage condition to be corrected within a time frame which will not cause damage to permanently connected Class 1E loads.

B.3.10 1 REV 92 3/12/9^

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS j}ases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and

~

"RCCA(s)" are synonymous.

A. Shutdown Margin Tr4p-+haukewn--rec c t iv i ty I c p revi de d c ene4 c t e n t ul+h pl ca t en fe ty-m:n1ycce f nouwt>t4+nn . Onc pe re:no-*huulown--weg4*-4+-edequat+-+aoet+t for--t.he .e.t.e "h-ambyc ! n , "h!ch -r+qu! rec rere chutdevn-r+ast4-v44y due te th' ere egntive mdereter temperature-esof-f4eient et end-+0 life (when-borwi-eement-retier !c icu) . Figure TS.3.10-l 10 dr+wn aseer41*g4y, A sufficientISHUTDOWN'MARCIN'esssies?th"dtF(1)MhENeAEtf6f?Edn?l tie %Kds

$6bcritica1{ from?alltoperatinpconditions N(2)fthe9enetivity transients.~

a s socia ted: with po s tula ted"ac c identXc ondi tions s are9 contro11 able Evi th in~

accsp table "limi ts ; vand - ( 3)h the Iraaeto r ivill?be7 main tainedisu f ficien tly huh c r i t ic al? t o[p r e cl udelinayve r ten tfe r;i t i cal i ty;inll the !lsim t down ! c o ndi tibn]

SHUTDOW7MARO Nyrequirem6nts Nary (thr6dgh6ht7c6rFJ11fFWh{fdhEt16n"6f fuel;depletionsreactoricoolant? system?boronLeoncentration'andrreact6r coolant 9veragCt'emperatnreCfThe;niostfrsutrictivu0 condition [occura at%nd of lifel:and 1sfassociated 3 with;;a:postulatedisteamhinefbreak tascidestiand resulting1uncontrolledbreactorncool'antisystemicooldownt JInitheranalysis' o f ' thi s iac c ide ntb a f minimum 1SHUTDOWi ttARGIN ( ( shownlin WigurehTS . 3 t10s1(as a:functionioffequt11briumih6tlfullypow4rfboron1 concentration)iisfreq61 red t6 contro1Vthe;;reactiVityitransient G Accordingly'BthefSHUTDOWNiMARGIlf ~

~

re 'quirementsiafe bised uponfthis?timiting9 onditionfandlar;&

withi planti aafe ty Lanalysis fassumptions OWithf reictoricoolah{ tlsystem consist'ent

~

transtestauresultilh5

-fromsa; average? temperature 1csal thain200?F,ithesreactivitf(snaliand' ai1C Ak/~C postulated:steamtlinetbreak?cooldowniarefelui S}iyTDOWiMARGI{pfovid@adehnatej pjhtectipN ~~" ~""~ ' ^^ "

Inf POWERTOPERATION5EndWOT5STANDBEvi tlRKstriM1W SHUTDOWTHARGINBis ens uredi by?c omply ingni th?their6d ? ins e r t ion S11mi ta tions9 tni Sps' c ifi c a tioii 3 :10 ; D L lIn (HOT iSTAND BWwi thl k2M 1907ahdii NTH 0TSHUTDOW NINTERM EDI ATE SHUTDOWNjandl COLD SHUTDOWN MthecS11UTDOWNLMARCINire'qdirementasit( ~ ~ ~ ~

S poc i f ic ationf 3210i A$ a ro l appli sable ho l prov ide $df f ic ientinegatiW r

reactivity;tosmeetithe assumpti6ns(ofithehafetplanalysestdiscusssldMb'oVFH

' For lREFUELINGEthe (shutdown; rs ac tiivi_tyf reqpir,ementgarejspecifiedjlin3able]

.TS. l. i t. t WheifJiht POWER T OPERATIONNhd T HOTfSTANDBM S HUT DDWN 5 MARGIN ffiFas tWiiiins d hssuming thelfue1 Land #ino'derator' tamp ~eratnrestare?atithelnominh1Werofp"osisk s

~

~ ^' " ~ " " ~ " "' ~ '" ~

designi.tsmperaturejof(547(F4 " "

With74ny Wed fe l us te Ec o ht rol Tus s einb1pfhb tE saphh1FCo f E Using?fGT1y?i6nsr t sd ?

theireactivity worthMEthssr6dicidater?c6ntroEassemb1~p~ '

~~ ~~"~

mnstibelac66untsd-f6riinf thejdetermination]of$SHUTDOWNiMARCINg~~ ~ ^ ~

B. Power Distribution Control The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR in the core of greater than or equal to 1.30 for Exxon fuel and 1.1T for Westinghouse fuel during normal operation and in.short term transients, and (b) limiting the fission gas release, fuel-pellet temperature and cladding mechanical properties to within assumed design criteria. The ECCS analysis was performed in accordance with SECY 83-472. One calculation at the 95% probability level was performed as well as one calculation with

'B.3.10 2

.RF" 91- 10/27/89 3,10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued B. Power Distribution Control (continued) all the required features of 10 CFR Part 50, Appendix K. The 95%

probability level calculation used the peak-linear heat _ generation-. rate-specified in the CORE OPERATING LIMITS REPORT. The Appendix K calculation used the peak linear heat generation rate specified in the CORE OPERATING' LIMITS REPORT for the Fn limit specified in the CORE OPERATING LIMITS REPORT. Maintaining 1) peaking factors below the Fg limit specified in the CORE OPERATING LIMITS REPORT during all Condition I events and 2) the peak linear heat generation rate below the value specified in the CORE OPERATING LIMITS REPORT at the 95% probability level assures compliance with the ECCS analysis.

During operation, the plant staff compares the measured hot channel-factors, F"g and F"a , (described later) to the limits _ determined in_the transient and LOCA analyses. The terms on the right side of the equations in Section 3.10.B.1 represent the analytical limits. Those terms on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

FN g is the measured Nuclear _ Hot Channel Factor, defined as the maximum local heat flux on the surface of a fuel rod divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fue1' enrichment.

The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fn axially, The K(Z) value-is based on large and small break LOCA analyses.

V(2) is an axially dependent function applied to'the-equilibrium measured.

F"n to bound F"q's - that could be measured at non-equilibrium conditions.

This_ function is based on power distribution control analyses that-evaluated the effect of- burnable poisons, rod position,-axial 'e'ffects,'and xenon worth.

FE n, Entineerine Heat Flux Hot Channel Factor, is defined as the allowance

~

on heat flux required for manufacturing-tolerances. The engineering factor allows for local variations in enrichment, pellet density and '

diameter._ surface area of-the fuel rod and eccentricity of tho: gap between pellet and clad. Combined statistically the net effect is a factor of-

-1.03 to be applied _to fuel rod _ surface heat flux.

The'1.05 multiplier-accounts for-uncertainties; associated with. measurement of the power distribution with_the movable incore_ detectors _and the use of;

~

those measurements to establishfthe assembly local power distribution.

FN g (equil) is the measured limiting F"q obtained at' equilibrium _ conditions during target flux determination.

FN a, Nuclear Enthalny Rise Hot-Channel Factor, is dtfined as the-ratio of the integral of linear power along the rod with the _ highest integrated; power to the average rod power.

p- g,- w

Attachment 3 Revision to-Original Proposed Technical Specification Pages License Amendment Request Dated September- 21, 19921 The revised and new pages affected by this License Amendment Request Revision-

'are listed below:

REVISED _PAGES NEW PAGES TS.1-5 TABLE TS.1-1 TS.1-7 TABLE TS.3.5-2A (Pages 1 through_3)

TS.3.10 1 TABLE TS.3.5-2B_(Pages 1:through 8)

TS.3.10 2 TABLE TS.4.1 1A (Pages 1 through 3)

B.3.5-4 TABLE;TS.4.1-1B (Pages 1.through 7).

B 3.10-1 TABLE TS.4.1 1C (Pages 1 through 4)

B.3.10-2 i

t I

l l.

l:,-

, ~ , . .

1 TS ,1 5 ' i OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have

OPERABILITY whenuit 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 1 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). ,

Vhen a system, subsystem, train, component or device is determined-to be inoperable solely because its emergency power source is inoperable, or solely because its normal 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 corresponding normal or emergency 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 the requirements of this paragraph.

The OPERABILITY of a system or component shall be considered to be estab--

lished when: (1) it satisfies the Limiting conditions for Operation in Specification 3.0, (2) it has been tested periodically in-accordance with.

-Specification 4.0 and has met its performance requirements, and (3);its s condition:is consistent-with the two para 5raphs above.

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor- >

coolant-temperature specified in Table TS.1~.1.

FilYSICS TESTS PHYSICS TESTS shall be those tests performed. to measure the . fundamental; characteristics of the core and related instrumentation. . PHYSICS TESTS are

~

conducted such that the core- power is sufficiently . reduced to allow for- the perturbation due to the test and therefore avoid exceeding power distribution limits in: Specification 3.10.B.

Low power PHYSICS TESTS are run at reactor powers less than~2% of. rated power, i

1 h%+ - - .w', -' , , - , , , ,

' TS.1 7 RATED Ti!ERMAL POWER RATED THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant of 1650 megawatts thermal (MWt).

REPORTABLE EVENT A REPORTABLE EVENT shall be any of.those conditions specified in Section 50.73 of 10 CFR Part 50.

SHIELD BUILDING INTEGRITY SilIELD BUILDING INTEGRITY shall exist when;

1. Each door in each access opening is closed except when the access'opeutng is being used for normal transit entry and exit,.then at least one door shall be closed, and

2. The shield building equipment opening is closed. 1

3. The Shield Building Ventilation System is OPERABLE.

SHUTDOWN MARGIN-SilUTDOWN liARGIN shall be the instantaneous amount of' reactivity by which:

1)_the reactor is suberitical'

or

2) the reactor would be suberitical from its present condition assuming all.

rod cluster control assemblies are fully inserted except for the rod cluster-control assembly of highest reactivity worth 1which is assumed to be-fully withdrawn.

EJTE BOUNDARY The SITE BOUNDARY shall-be that' line beyond which the-land is neither owned nor;1 eased, nor otherwise controlled-by.the licensee.

. SOLIDIFICATION-

~

, SOLIDIFICATION shall be the conversion of-wer' wastes into a form that meets 1 shipping and burial ground requirements.

. SOURCE CHECK' i.

I A SOURCE CHECK shall be the qualitative 1 assessment ofichannel response'inen. _

- the channel sensor is> exposed to a source of-increased radioactivity.

l r(-* " c w iye--

. TABLE'TS,1-1 i

TABLE TS 1-1 OPERATIONAL MODES-REACTOR

% RATED AVERAGE VESSEL HEAD REACTIVITY THERMAL . COOLANT. CLOSURE- BOLTS MODE TITLE CONDITION POWER IEMPERATURE FULLY TENSIONED ~

1 POWER OPERATION Critical > 2% NA YES 2 110T STANDBY ** Critical 5 24 NA- YES i

3 Il0T SHUTDOWN ** Suberitical NA . 2 350*F YES 4 INTERMEDIATE Subcritical NA < 350*F YES SHUTDOWN ** z.200*F 5 COLD SliUTDOWN Subcritical NA < 200*F- YES 6 REFUELING NA* NA NA NO

Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive _of the following conditions.

is mot:

a. K,g 5 0.95, or

b. Boron concentration 2 2000 ppm.

- Prairie Island specific MODd title, not consistent with Standard Technica1L

- Specification MODE titles, MODE numbers are consistent with Standard.

- Technical Specification MODE numbers, f

l e

t , b + ,

w r --

-m

TABLE TS.3.5-2A (Page 1 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. . CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ,

1. Manual Reactor Trip ' 2 1. 2 1, 2 1 2 '1 2 3(*), 48*), 5(*) 8 i

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3- 1, 2 2

b. Low l Se tpoint 4 2 3 lW, 2 2 ,

3 Power Range, Neutron Flux,. 4 2 3 1, 2 2 High Positive Rate

4. Power. Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate

~i 5 Intermediate' Range,' Neutron Flux' 2 1 2 IN, 2 '3:

6. Source Range.. Neutron Flux

a. Startup 2 1 2 2(" .

4

b. Shutdown. 2 1 2 ' 3( ) , 4(*), 5(*) 5,

7. Overtemperature AT , 4 2 's 1, 2 6; i

- 8 '. . Overpower AT 4 2 3 1, 2 '6. :oa s.

S5D i 5; .

-(a) When the' Reactor Trip Breakers are-closed and the ' Control Rod Drive System is capable of rod -g'

. withdrawal. o-my-(b) B$ low;the P-10.(Low Setpoint Power. Range Neutron Flux. Interlock) Setpoint. O['>

(c) Below.the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. i

, a. . . .

. r t

TABLE TS. 3. 5-2A (i' age 2 of 6)

' REACTOR TRIP SYSTEM INSTRUMENTATION i

l MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL-UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1: 9. Low Pressurizer Pressure 4- 2 3 1 6

i. 10'. High Pressurizer Pressure' 3 2 2 1, 2 6 -

11. Pressurizer High Water Level 3 2 2 1 .6 l

12. Reactor Coolant Flow; Low- 3/ loop 2/ loop 2/ loop 1 6 '

-13. Turbine Trip

. a. Low AST 011 Pressure- 3 2- 2 1 6 n

b. Turbine'Stop ValveLClosure :2 2 1 1 6 14.:Lo-Lo Ste'am Generator 3/SG' 2/SG in 2/SG in 1, 2 6 :i Vater. Level 'any SG~ each SG-

'15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus.on 2 on one 1 11 .

bus. '

11' and 12 (Unit 2: 21_and 22) both buses i

'h$h "an n -

o,-g. 3 Shm !

1 4

i a

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

....u. - -

~

TABLE TS.3.5-2A (Page 3 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION i

j MINIMUM: l TOTAL NO. CHANNELS CHANNELS APPLICABLE *

. FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE' MODES ACTION

'N 16.' Loss of Reactor Coolant Pump

a. RCP Breaker Open 1/ pump 1 1/ pump 1 1: .i

b. Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one 1 11 .I both bus buses 5

17. Safety: Injection Input 2 1 2 1, 2 7 from ESF

[

18. Automatic Trip-and Interlock Logic 2 1 2 1, 2 ,,

7 ;

2 1 2 3t"), 4(*), 5(*) 8' '4

19. Reactor Trip Breakers' 2 1 2 1, 2 '9 r 2 1 2. 3(*), 4(*), 5(*) .8

20. Reactor Trip Bypass. Breakers 2- 1 1 (d) 10 1

-(a).When the Reactor Trip Breakers are closed and the Control Rod Drive System is c'apable of rod [

withdrawal. -co e i am.

(d),'When the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reactor Trip' Breaker"

~

d

, [

and the Control Rod System is capable'of. rod withdrawal.

- ;m y -

OY B?.

w; 4 4 e -, ~ , v ~.y .+

+

I TABLE TS.3.5-23 (Par 1 of 5) i E';GINEERED 3AFETY FEATURE ACITATIOw SYSTE(INSTRENENTATION :

MINIMUM '

TOTAL NO m CHA'M LS CHA';NELS APPLICABLE !

FU';CTIO';AL UNIT OF CHANNELS TO TRIP O'>ERABLE MODES ACTION f i

F

1. . SAFETY INJECTION ;

l

a. Manual' Initiation 2 1 2 1,2,3,4 23 i

b. High Containr ent Pressure 3 2 2 1,2,3,4 24 i t

c. Steam Generator Low Steam 3/ Loop 2 in any 2/ Loop 1 , 2 , 3(*) 24 I Pressure /Imop ,

Loop J. Pressurizer' Low Pressure. 3 2 2 1 , 2 , 3(*8 24 i

e. Automatic Actuation Logic 2 1 2 1,2,3,4 20 and Actuation Relays [

b r

2. CONTAINMENT SPRAY ' i r

a. Manual Initiation 2 2 2 1,2,3,4 23 l

b. Hi-Hi. Containment Pressure 3 channels 1 sensor 1 sensor 1,2,3,4 21 f with 2 per per i sensors per channel channel channel in all 3 in all 3 [

channels channels I EQ$ !

c. Automatic Actuation Logic and' 2 1 2 1, 2, 3, 4' 20 #$*"E !

Actuation Relavs - !

~g

o. .

i c= P* r'

. (a) Trip function may be blocked .in this ' MODE below a Reactor Coolant System Pressure of 2000 psig. ,

[

t E

4 i-4

[ TABLE TS.3.5-2B (Fage 2 of 8)

L

t. '

j. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRDfD.TATION 1

l MINIMDf 4

TOTAL NO. CRB7ELS CHG3ELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE- ACTION

~

MODES i 3. ' CONTAINMENT-ISOLATION l

< l

}

a. Safety Icjection see 1 above fer 3I1 Safety knecuon imtiaticg functims and .w. m.as.

I i b. Manual 2 1 2 1,2,3,4 23 a . .

i:

~

~ c. - Automatic Actuation logic and 2 1 2 1,2,3.4 20 Actuation Relays

, 4. CONTAINMDG VENTILATION ISOIATION

, a. Safety .Inj ection see 1 above for an safety L,icctica inMaing functims and mparements.

i f .b. Manual 2 1 2 (b) 22 e

c. Manual Containment Spray see 2a above for Manual Chma spray regaremems

d. Manual Containment Isolation see 3b above tw Manual cew isolation w_cas.

i.

e. High Radiation in Exhaust Air 2 li 2 (b) 22 1

f. . Autematic Actuation logic 2 1 2 (b) 22 x - -i i and. Actuat'.on Relays n -e >

< a e:r 4

ar am 0= MH to h (b) ~=atenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in Eh i

operation. cm u w

i..

l

, , , , , ,, y . s. , . . , - ..,x , .,m . , . - + - ._s . . . . , - 7 ..

r --.mmm. ..

=

TABLE TS.3.5-2B (Page 3 of.8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTPINENTATION MINIMI'M TOTAL NO. CHA'CiELS CHANNELS APPLICA3LE OF CRA'R;ELS TO TRIP OPEFAPLE MODES ACTION

' FUNCTIONAL' UNIT

5. STEAM LINE ISOIATION 1/ loop 1/ Loop 1/Ioop 1. 2 , 3(*) 27

a. Manual 3 2 2 1 , 2 , 3(*) 24

b. :Hi-Hi Containment Pressure

c. Hi-Hi Steam Flow with Safety Inj ection 2/ Loop 1 in any 1/ Loop 1. 2 , 3(*) 29

1. Hi-Hi Steam Flow IBOP

~. 2 . Safety injection see 1 above for an safety Ic;ation inkiating functims and requnemcras.

d. Hi Steam Flow and 2 of 4 Low Tave'with Safety Injection:

2/ Loop 1 in any 1/Ieop 1, 2. 3 88) 29

1. Hi. Steam Flow Ioop 4 2 3 1. 2, 3'd) 24

2. Tave-see 1 above br an safety tajection inkiarnig functions and .w a

3. - Safety Injection EQ$

<=m

-(c) When either main steam isolation valve is open.

SE we (d) When reactor coolant system average - temperature is greater than 520*F and either main steam isolation

~

o?

, valve is open.

",' Y w

I

TABLE TS.3.5-2B (Page 4 of 8)

ENGINEERED SAFETY FEATURE ACITJATION SYSTEM INSTRUMENTATION MINIMLN TOTAL NO. CHANNELS CHANNEIS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION FUNCTIONAL UNIT

5. STEAM'LINE' ISOLATION (continued) 1,2,3 M 25 e; . Automatic Actuation Logic and 2 1 2 Actuation Relays

6. FEEDWATER ISOIATION 24 1

a. Hi-Hi Steam. Generator Level 3/SG 2/SG in 2/SG in 1, 2 any SG each SG

b. Safety Injection see 1 above for an safety Irgectxm initiating functions mi m_ ~.:..

c. Reactor-Trip with 2 of 4 Low Tave'(Main Valves only):

2 1 2 1. 2 28

'1. Reactor Trip 4 - 2 3 1, 2 24 2.. Low Tave 2 1 2 1, 2 28

d. Automatic Actuation Logic and Actuation Relays 4 E7$

%G

2 (c) When either main steam isolation valve is open. Sh 3Y W

t l: <

i j .-

1 TABLE TS. 3.5-2B (Page 5 of 8)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i

]. MINIMCM

TOTAL NO. CHANNELS CHANNELS APPLICABLE d -

F1JNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION l'

e 4

-7. AUXILIARY FEEDb'ATER j' .a. Manual 2 1 2 1,2,3 26 f' b. Steam Generator Low-Lew 3/SG. 2/SG in 2/SG in 1, 2, 3 24 j L'ater Level any SG each SG'

c. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29

11 and 12 (Unit 2: 21 and 22)- both bus

- (Start Turbine Driven Pump buses only).

]-

l d. . Trip of Main Feedwater Pumps i-i 1. Turbine Driven 2 2 2 1, 2 26

2. Motor Driven ~2 2 2 1, 2 26

! e. Safety Injection See I above fbr aH Safety Injectam initiatmg fbnctums and iwoma

.- f. ' Automatic Actuation Logic 2 1 2 1,2,3 20 j .- and Actuation Relays

, :e n H w

%G i- u. r.n a

4

}; =b

{

of i

1 4

..y-- , - , s .v%., w _ _ - w . - 4 ,.- ,

.m....w., . , - -. +, , ..-.m. .-~.- . , _ _ _ _ , . _ . _ . _ -_m m _.- .

l1,ijiI ll)1 llI i!llI 1l lli 1

=rM 1-*".w. oM e

>Mf O nA D

w N 9 9 O 2 2 I

T C

A E 4 4 L

B , .

A 3 3 C

I S . .

L E 2 2 F D F O , .

N AM 1 1 O

I T ,

A T S E E' ML L J E E N

L T ;

A I R E

T N la E

S I R P N MC O 7 E I

)

8 M

f E S e ) >

)

o T L P s s s s S E I a e e e Y ;

R sh2 6 S '

2 T up s a sh2 s up a e N A B/ nh S/nh g

O H O / 1 op /1 op a I CT 2( 2(

P T

(

A B U 2 T

- C S n n 5 A L o) o)

E s s 3 E O ;

ee ee R ;

' N' . s s s s

- S U A aa aa T T L H shh shh A A C upp u ;p E E T B/ B/

L F O F /22 /22 B TO 4( 4(

i A

T l.

e r

A S

D E

R E

E N

I G

N E

s s u u eB B g

as s td d l r er oa ga

. R V ug au t g T G' d e l e I O ef of P d a va rS U' aS F r e

L O gV dV A ek nk N S D4 U4 O S I

D a T I b C

N T

F S

- I lll

TABLE 3.5-2B (Page 7 of 8)

Action Statenents ACTION 20: With the number of OPERABLE channels ACTION 23: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, restore the inoperable Channels, restore the inoperable channel to OPERABLE status within 6 channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least. HOT SHUTIXN3 hours or be in at least HOT SHLTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD within the next 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 ; ShTTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

however, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing per Specification 4.1, provided ACTION 24: With the number of OPERABLE channels the other channel is OPERABLE. one less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following conditions are satisfied:

ACTION 21: With the number of OPERABLE channels less-than the Total Number of Channels, operation may proceed provided the a. The inoperable channel is placed in inoperable channel (s) is placed in the the tripped condition within 6 tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and hours, and, the Minimum Channels OPERABLE requirement is met. The inoperable b. The Minimum Channels OPERABLE channel (s) may be bypassed for up to 4 requirement is met; however, the hours for surveillance testing per inoperable channel may be bypassed Specification 4.1. for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.1.

ACTION 22: With the number of OPERABLE channels less than the Total Number of Channels, w - -4 operation may continue provided the containment purge supply and exhaust E[he 2r valves are maintained closed. "

C

d m _La m

o?

r

- . = - = - _ __

t I

I i

TABLE 3.5-2B (Page 8 of 8) 4 i Action Statenents t

i ACTION 25: With the number of OPERABLE channels Channels, restore the inoperable I

one le.ss than the Total Number of channel to OPERABLE status within 6 j Channels, restore the inoperable hours or be in at least HOT SHUTD0k'N l 3 channel to OPERABLE status within 6 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, once i

. hours or be in at least HOT SHUTDOkT channel may be bypassed for up to 8

{ within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation in hours for surveillance testing per 1 l HOT.SHUTDOWi may proceed provided the Specification 4.1, provided the other l main steam isolation valves are closed, channel is OPERABLE. ;

if not, be in'at least INTEPREDIATE

[ SHUTDOkW within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

However, one channel may be bypassed ACTION 29: With the number of OPERABLE channels less than the Total Number of Channels, l for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance operation in the applicable MODE may l testing per Specification 4.1, provided proceed provided the following i the other channel is OPERABLE. conditions are satisfied: i i

a. The inoperable channel (s) is placed ACTION 26: With the number of OPERABLE channels in the tripped condition within 6 4

one less than the Total Number of hours, and, w

Channels, restore the inoperable channel to OPERABLE status-within 48 5. The Minimum Channels OPERABLE !

bours or be in at least HOT SHUTDOWi requirement is met; however, the i- within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least inoperable channel (s) may be i INTERMEDIATE SHUTDOk'N within ' the bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for I following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . surveillance testing of other channels per Specification 4.1

. ACTION 27: With the number of OPERABLE channels i 5-one less than the Total Number of :e - 4 1

Channels, restore the inoperable m i e:>

< [

l' channel to OPERABLE status within 68 SE hours or be in at least HOT SHUTDO'.;N m .4 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and close the e ."

associated valve. .

.- OY [

j ACTION 28: - With the number of OPERABLE channels '$

.one less than the Total Number of [

I i

, c !

i 7-f 4 g w. er y e- m . , = - e ---

e w ,r+-, :'w=r - n s w e-rw.+., . > - = - ,, -- .* -,,--ste-+, = . , -..w,. ._s ,sg

TS.3.10 1 ,

3.10 CONTROL ROD AND POWER DISTRIliUTION LIMITS 6ppifenbility Applies to the limits on core fission power distribution and to the limits on '

control rod operations.

Obiective l

To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during POWER OPERATION, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection, ;

Specification A. Shutdown Marr.in

1. Reactor Coolant System Avery e Temperature > 200'F j The SliUTDOWN MARGIN shall be greater than or equal to the applicable value shown in Figure TS.3.10 1 when in ll0T STAND 15Y with ke rr < 1,0, .

and when in ll0T SilVTDOWN and INTERMEDIATE S110TDOWN, ;

2. Reactor Coolant System Average Temperature s 200'F .,

The SHUTDOWN MARGIN shall be greater than or equal to itAk/k when in COLD SilUTDOWN.

3. With the SilVTDOWN MARGIN less than the applicable limit-specified in :

3.10.A.1 or 3.10.A.2 above, within 15 minutes initiate boration to !

restore SiluTDOWN MARGIN to within the applicable limit.

li . Power Distribution Limits

1. At all times, except during low power PilYSICS TESTING, measured hot channel factors, F No and Fan, N as defined below and in the bases, shall-meet the following limits:

RTP Fg 8

.x 1.03 x.1.05-s (Fn / P) x K(Z)

- RTP F"An x- 1.04 s F A u y (1+ PFDil(1 P)]

where the following definitions apply: ;

RTP Fn . is the Fn . limit at RATED TilERMAL POWER specified in the CORE !

OPERATING LIMITS REPORT,-

RTP

- F An is the - Fan limit at . RATED Ti!ERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

PFDit is the Power Factor Multiplier for F"Au specified.in the CORE OPERATING LIMITS REPORT. 1

!; -K(Z) is a normalized function -that limits Fn(z) axially 'as: specifiedJ in-

! .the CORE OPERATING LIMITS REPORT.

,, -- . ..,-_.~.....-._,._,m..._,2.-..,-._ -__..-__..__-.;_..,,- ,

=i

TS.3.10 2 3.10 B.1. - Z is the core height location.

- P is the fraction of RATED TilERMAL POWER at which the core is operating, in the F"n limit determination when P 50.50, set P - 0.50.

- F"q or 1*a is defined as the incasured Fn or Fan respectively, with the smallent inar61 n or greatest excess of liinit.

- 1.03 is the engineering hot channel factor, Fg, applied to the E

measured I"n to account for manufacturin6 telerance.

- 1.05 is applied to the measurad F"n to account for measurement uncertainty.

. 1.04 is applied to the measured F"tn to account for measurement uncert ainty.

2. Ilot channel factors, F"o and I"an , shall be measurod and the target flux ditference determined, at equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full power days in conjunction with the target flux difference determination, or (h) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10% or more of RATED TilERMAL POWER.

1*n (equil) shall ment the following limit int the middle axini 80%

of the core:

M i' F"o (equil) x V(Z) x 1.03 x 1.05 s (Fn / P) x K(Z) _

where V(Z) is specified in the CORE OPERATING LIMITS REPORT and other terms are defined in 3.10 B.1 above.

3. (a) If either measured hot channel factor exceeds its limit-specified in 3.10.B.1, reduce reactor power and the high neutron flux trip set point by It for each percent t.ha t the measured F N g or by the f actor specified in t he CORE OPERATING LIMITS REPORT for each percent that the measured F"3n exceeds the 3.10 B.1 limit. They. follow 3.10.B.3(c).

(b) If the measured F N (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take oi of the following actions:

Y

1. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place *') reactor in an equilibrium configuration for whicl_ pecification 3.10.B.2 is satisfied, or

2. Reduce reactor power and the high neutron flux trip setpoint by It for each percent that the measured F"n (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.

._ . m

' TABLE TS.4.1-1A (Page 1 of 5)

REACTOR TRIP SYSTEM INSTRLMENTATION SURVEILIANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR L'rIICH

' R7NCTIONAL'1TNIT - CHECK CALIBFATE TEST TEST SURVEILIANCE IS REOUIRED

1. Manual Reactor Trip. N.A. N.A. R(m N.A. 1, 2, 3m , 4m. 5m

.2. ' Power Range, Neutron Flux a) High Setpoint' -S D(5 'I Qmi R 1, 2 gte. 7) qtr. e)

. b)' Iow Setpoint S Rm S/U(173 R 1m. 2

.3. Power Range, Neutron Flux, N.A. Rm Q R 1, 2

'High Positive. Rate i :4 Power Range, Neutron Flux, N . A.' . R(7) Q R 1, 2

High Negative Rate.

5. Intermediate Range, S- .R m - 5/U N R 1m,2 Neutron. Flux

- 6. ;. Source Range . Neutron' Flux l

< la. 'Startup. S RW S/U W R. Im- :n , ,4 j- b. ' Shutdown S Rm. qm) R 3m,4m,5m Q7g 4 .

i-SE

-4 I 7. Overtemperature AT S- R Q R '1, 2' o-

+

-my i O7

' 8. Overpower AT .. S R .. Q. R, 1, 2 E l

1 1

, w#4 .. .w. , ,# . .. .. , # . , , - ,

TABLE 4.1-IA (Page 2 of 5) l l

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQi'IREMENTS FUNCTIONAL RESPONSE MODES FOR VHICH FTJNCTIONAL UNIT CHECK CALIBRATE TEST TEST . SURVEILLANCE IS REOUIRED

'9. Low Pressurizer Pressure S R .Q N.A. 1

'10. High Pressurizer Pressure S R Q N.A. 1, 2

11. Pressurizer High Vater Level S R Q N.A. 1

12. Reactor [ Coolant' Flow Iow S R Q N.A. 1

13. Turbine Trip a, Low AST Oil Pressure N.A. R S/Ut*, 10 N.A. 1

b. Turbine Stop Valve N.A. -R S/U('- 1U N.A. 1 [

Closure I 14 Lo-Lo Steam Generator 5 R Q N.A. 1. 2 Water Level t

15. Undervoltage 4KV RCP Bus N.A. R Q N.A. -1 1

i

e a a k l 2 5; !

mg .

o*

. f O[>-

m. . _ _ _ _ _ _ _ _

TABLE TS.4.1-1A (Page 3 of 5)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR WICH FUNCTIONAL UNIT- CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED

' 16. Loss of Reactor-Coolant Pump

a. RCP Breaker Open N.A. R S/U(') N.A. 1

b. :Underfrequency 4KV Bus N.A. R Q N.A. 1

17. Safety Injection Input N.A. N.A. R N.A. 1, 2 18.LAutomatic. Trip and Interlock N.A. N.A. M(8) R 1, 2, 3(1), 4(1), 5(1)

Logic

.19. Reactor. Trip Breakers N . A .- N.A. Mcs. 12) R 1 , 2 , 3(1)

, 4(1), 5(1)

20. Reactor Trip Bypass Breakers N.A. N.A. Mll') R(15) See Note (16)

< oe E, b Q >

_ .-%. . -e * ,m,. m y 1 -

c w ##,---, - w, r,,-%. . , . ,

-l t

~[

i- t

{ TABLE TS.4.1-IB (Page 1 of 7) ;

i ,

j ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REOUIREMENTS [

FT'NCTIONAL RESPONSE MODES FOR L'HICH f

_ FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS FEOUIRED [

1. SAFETY INJECTION [

. . . t

a. . Manual Initiation N.A. N.A. R"3 N.A. 1, 2, 3, 4 f i

b. .High Containment Pressure S R Q N.A. 1, 2, 3, 4 i 4 :

c. . Steam Cenerator Low Steam S R Q .N.A. 1 , 2 , 3(2) j l Pres.sure/Ioop i

d. Pressurizer Low Pressure. S R Q N.A. 1, 2 , 3t2)

[

{.

e. ' Automatic Actuation Logic 'N.A. N.A. MC33 N.A. 1, 2. 3, 4 r

'and Actuation Relays l

2. . CONTAINMENT SPRAY !

Manual Initiation N.A. N.A. R N.A. 1, 2, 3, 4 a.

t l

^

b. .Hi-Hi Containment S R Q N.A. 1, 2. 3, 4 Pressure i -.' c. Automatic Actuation Logic N.A. .N.A. M(3) N.A. 1, 2, 3, 4

and Actuation Relays i

r ,

I I nna

' kc e i 1: I

' rW-om ;

e e.* o.4 i m-O + i 9

PS .D ,

4 DY 6*

,- U f l

1 4 i 4

.i 4 -

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

l 3

TABLE TS.4.1-1B (Page 2 of 7) !

l t

. ENCINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUME'.'TATION SURVEILLANCE REOUIREMENTS !

4 I

. FUNCTIONAL RESPONSE MODES FOR L'HICH l .

FUNCTIONAL UtJIT CHECK' CALIBRATE TEST TEST .SURVEILIANCE IS REQUIRED 4

i CONTAINMENT ISOIATION 3.

a. Safety Injection: See i above for an Safety Injectxm Surveinance Requirements i

b. Manual' .N.A. N.A. R N.A. 1, 2, 3, 4 f

. c; Autcmatic Actuation Logic N.A. N.A. M(3) N.A. 1, 2, 3, 4 l

l. and Actuation Relays

4. CONTAINMENT. VENTILATION ISOLATION

., t L

a. Sa?ty.' Injection See I above for au Safety injecten Survedlance Reqtnremeres l' t

b. Manual N.A. N.A. R N.A. See Note (7) t l c. Manual'. Containment. Spray ' See 2a above for aD Manual Containmera Spray SurveiBance Requiremeras i

i- d. Manual. Contain:nent ' ' See 3b above for aD Manual Contamment Isolatkm Surve Bance R+ s .as Isolation t

e. High Radiation in DW R M N.A. See Note (7)

Exhaust Air j f. Automatic Actuation Logic N.A. N.A. M(3) N.A. See Note (7) and Actuation Relays .;

c - H tn N >

<. 2 t*?

~ e

. oI t %y N#

['

i b t:5 i

t

-~i...%- w -, -> -a

' , 4- 3--r2 m we e -e ,e..v.ii- m - ,... --e es4, ,.e.-,, e ---e , . -r m.- -,me m. , , r. .,m v., re

TABLE TS.4.1-1B (Page 3 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIR.W.ENTS FUNCTIONAL RESPONSE MODES FOR L'HICH FUNCTIONAL UNIT- CHECK CALIERATE TEST TEST SURVEILIANCE IS REQUIRED

5. STEAM LINE ISOIATION

a. Manual N . A.. N.A. R N.A. 1, 2, 3")

~

b. Hi-Hi Containment S R Q N.A. 1, 2, 3")

Pressure

c. Hi-H1 Steam Flow with Safety -Inj ection

. . l .. Hi-Hi Steam Flow S R Q N.A. 1, 2, 3")

2. Safety Injection see 1 above for au safety injectam Surveitance Ryo .~as

d. H1' Steam Flow and 2 of 4 Iow T ,,with Safety

'Inj ec tion '

1.

Hi Steam Flow S R Q N.A. 1, 2, 3")

2. .Tave S R Q N.A. 1, 2, 3(5)

3. Safety Injection See 1 above for an Safety Im surve Hance Rmu-.~a.

e. Automatic; Actuation Logic- N.A. N.A. M(3) N.A. 1, 2, 3")

and' Actuation Relays := ^ -4 22$

n t-

, 12 t1

[ i-k 0~

CF L

TABLE TS.4.1-1B (Page 4 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS _

FUNCTIONAL RESPONSE MODES FOR VHICH FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SUFVEILIANCE IS REQUIRED

6. FEEDVATER ISOIATION

- a. _Hi-Hi Steam Generator S R Q N.A. 1. 2 Ievel

b. ' Safety Injectit,n 5:e 1 above for aH Safety injecuen SurveaHance Regtmements

- c. , Reactor Trip.with 2 of 4 Lcer Tm (Main Valves Only)

1. Reactor Trip N.A. N.A. R N.A. 1, 2

2. Tave- S R Q N.A. 1, 2 i

d. Automatic? Actuation logic 'N.A. N.A. M(3) N.A. 1,. 2 4

and tctuation Relays 1 -

4 5

<: . =

j '. I i

enL1 O+

1 mb 37~

W g- y <%-. . 3 y e - = + , .M 9: -me. s see y, r<-%-t- e- --ew-vr---e -* *.*+e---

+e- *--- - - - - - - - - . - - - - - - - - - - - - - - - - - - -

TABLE TS.4.1-1B (Page 5 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION StiRVEILLANCE REQUIREMENTS FUNCTIONAL RESPONSE MODES FOR WHICH FUNCTIONAL UNIT- CHECK CALIBRATE TEST TEST SURVEILIRiCE IS REOUIRED

7. AUXILIARY FEEIL'ATER

a. . Manual N.A. N.A. R N.A. 3

b. Steam Generator Low-Low S R Q N.A. 1, 2, 3 Water Level

c. Undervoltage on 4.16 kV N.A. R R N.A. 1, 2 Buses 11 and 12 (Unit-2:

. 21 and 22) (Start Turbine.

Driven Pump only)l

d. Trip of Main Feedwater Pumps

'1. Turbine Driven N.A. N.A. R N.A. 1, 2

2. Motor Driven N.A. N.A. R N.A. 1, 2

e. Safety Injection . See 1 above for an Safety Injecten Save: Dance Requirerneras

f. Automatic Actuation Logic N.A. N.A. Mt s) N.A. 1, 2, 3 and Actuation Relays N

< 9., =$

d

?, b

i-...

TABLE TS.4.1-1B (Page 6 of 7)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REGUIREMENTS 4

.l FUNCTIONAL RESPONSE MODES FUR QTICH

- FU*iCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIASCE IS REQUIRED 1

8. IDSS OF PO*='ER i a.. Degraded Voltage N.A. R M N.A. 1,2,3,4 4kV Safeguards Bus

.. b. Undervoltage N.A. R M N.A. 1, 2, 3, 6 l '4kV Safeguards Bus i

b.

4 4

a a

4 M^N

? MT>

) .2 -

<3 =

l b i-- esva e

o.

V e Y

' M l

i . .

,,m ..r . . _ . . , . . - , , , - .m . . . ,

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

i TABLE 4.1-1B (Page 7 of 7)

TABLE NOTATIONS t t

i FREQUENCY NOTATION NOTATION' FREQUENCY i ,

j S Shift !

. 'D ~ Daily M- Monthly !

I Q Quarterly

[ R "ach Refueling Shutdown '

N.A. Not Applicable ;

t TABLE NOTATION-i

(1) One manual . switch shall be tested at each (8) Whenever CONTAINMENT INTEGRITY is required j refueling on a. STAGGERED TEST BASIS. and either of the containment purge systems j

- are in operation. j (2) Trip function may be blocked in this MODE i below a' reactor coolant system pressure of

.. 2000 psig. l

}i j (3) Each train shall be tested at least every two {

months.on.a-STAGGERED TEST BASIS.

I i . .

- (4) . 'When either main steam isolation valve is j open, r

!. (5) . When reactor coolant system average !

temperature'is greater than 520*F and either $Qy 'r 5.

main steam isolation valve is open. #$emE !

t 1

i, 1

I(6). .See Table'4.17-l'. "d o.-

mv v

t*

W !

4 !

i i

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

-i b

i TABLE TS.4.1-IC (Page 1 of 6)

MISCELI>SEOUS INSTRI' MENTATION SL7VEILIANCE REQUIREMENTS FINCTIONAL , RESPONSE MODES FOR VHICH FUNCTIONAL UNIT -CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED

1. Control Rod Insertion Monitor M R S/U(1) N.A. 1, 2

2. Analog Rod Position S R S/U(11 N.A. 1, 2 , 3(22 , 4(2', Scz)

'3. ' Rod. Position Deviation M N.A. S/U(1) N.A. 1, 2 Monitor'

'4 Rod Position Bank S(33 N.A. N.A. N.A. 1 , 2 , 3<2) , 4(2) , 5(2)

Counters

'5. Charging Flow- S R N.A. N.A. 1, 2, 3, 4

6. ' Residual Heat Removal S R N.A. N.A. 4") , 5ts) , 6'88 Pump Flow

7. - Boric Acid Tank level D' R"3 M"3 N.A. 1, 2, 3, 4 8 .. Refueling Water Storage 'W R M N.A. 1, 2, 3, 4 Tank Level

9. Volume Control Tank S R N.A. N.A. 1, 2, 3, 4

10. Annulus Pressure N.A. -R R N.A. See Note (10)

(Vacuum Breaker) EQY

< m en e= t-11.' Auto Load Sequencers **

N.A. N.A. M N . A .' l. 2, 3, 4

~d 12.-Boric Acid Make-up Flow N. A. ' R N.A. N. A.. 1,'2, 3, 4 ?h ~

Channel o~

"L o

ll1 ! jljil!lli{I! l)ll;1I)! jlll \j j!l ilI jjll!!l1- i l 1Il' NKn4a'eL t -r O Qgs w E ow N<

D E

R I

U O m E -

R S 6 6 ,

H S C I , ,

I 5 5 m H E 4 W C 4 4 4 4 ,- 4 R 'A O L , , , . W W ,

T L 3 3 3 3 3 3 3 -

I S E , , , , , , ,

E V 2 2 2 2 2 2 2 2 2 D R O U , , , , . , . , ,

MS 1 1 1 1 1 1 1 1 1 1 S -

T _

?' _

E E M

E S _

R N T . . . . . . . . . .

I O S A. A. A. A. A.

A A. A. A. A. _

P E U S T N N N N N N N N N N Q E E R -

R .

) E ._

4 C L N A f A o I N .

L O T . .

2 I S I

T E A. .

A.

E C T R R M M N M R R N M .

e g V N R .

a U U F

P S -

(

N ET O E _

I T

A A . .

T R ,

B A. A.

S_ I R R R R R R N N R R S c_ L T A R C E T L S B N ;

A I K .

T S C E A.

U H N S S M R S S S M D O C E

R I

L E C _

C S d

~

I _

n n g g .

M a o r n n ,

i i i i .

B d t A r rs n a -

o ot a e. i d t t t n .

A' g d D u s i m ie l a a T o u n ) nm) p e t R s Rr r a o 1 ou2 m v .S r e h h M - Mr-u e n o t S x 5 t5 .

S L . t aW t e E t 1 t s1 _

s l i wm g n nn T t r r' P s n ow n pe es3 eI3 I n o i yn t o l o i ar l. t d N e F Fl l - C ut i nS. i nS U

n m

t ae l r er e ce nm

' Mc t F o o ra ceT cm cot ci L i uu nu eu i ns C or Aue At e A al ms is gr m as t e - rl - al N t e us b s rt s l a M cp t tb tib O nv ce re es i op D am ssa sd a .

I CL oe cr ur mn e oy R ee onT oat T AP TP EI S. CB C RT PI( PR( -

C N . . . . . . . . . .

U 3 4 5 6 7 8 9 0 1 2 F 1 1 1 1 1 1 1 2 2 2 .

ii ii - 14 .* ;' ' 3 . ;l!j j[ 1 j [ ij i jj . iilil}: I ,' .

p 4

TABLE TS.4.1-IC (Page 3 of 4) 4 MISCELLANEOUS INSTRt' MENTATION St'RVEILLANCE REOUIREMENTS h

FU'!CTIO*;AL RESPO';SE MODES FOR WICH

- FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SLTIEILLANCE IS REQUIRED

!' 23. Post-Accident Monitoring ' M R N.A. N.A. 1, 2 Reactor Vessel Level Ins trumentation -

, (Table TS.3.15-3) i

; 24. . Steam Exclusion Actuation W Y M N.A. 1, 2, 3

. 25. Overpressure Mitigacion. N.A. R R N.A. 45, 5

} . 26. Auxiliary Feedwater N.A. R R N.A. 1, 2, 3 Pump Section Pressure i 27. Auxiliary Feedvater N.A. R R N.A. 1, 2, 3

: Pump Discharge Pressure I

.' 28. NaOH. Caustic Stand Pipe W R M N.A. ' 1, 2, 3, 4 Level' i'

29. Hydrogen Monitors- S Q M N.A. 1, 2 f 30. Containment. Temperature 'M R N.A. N.A. 1, 2, 3, 4 Monitors

e - H

- 31. Turbine Overspeed N.A. R M N.A. I m=>-

Protection Trip Channel #SemE

n a,

w 1

o 1

4-

- -.- .. .. . . r ,e4 m 4 . , , , . ....,--e

TABLE 4.1-1C (Page 4 of 4)

TABLE NOTATIONS FREQUENCY NOTATIO*i NOTATION' FREQUENCY S Shift D Daily V' L'eekly

.M Monthly

-Q Quarterly

'S/U. Prior to each startup Y Yearly R 'Each refueling shutdown N.A. Not applicable

TABLE NOTATION (1). Prior'to each startup following shutidown in (7) Except for containment hydrogen monitors excess of two days if not done in previous 30 which are separately specified in this table.

days.

(8) E en RHR is in operation.

(2) Wen the reactor trip system breakers are closed.and the control' rod' drive system is' (9) Wen the reactor coolant system average capable ' of rod withdrawal. temperature is less than 310*F.

(3) Following rod motion in excess of six inches (10) Wenever CCNTAI!DiENT INTEGRITY is required.

when the : computer is out of- service.

(4). Transfer. logic to Refueling L'ater Storage 1 NQ E!

2. Tank. <$ ~p on

, -(5):

E en either mal'n steam isolation valve is "U *

! open.. e

.aw .

(6). ' Includes those instruments named in the- g j emergency. procedure.

_ -._ - - m ._.. _ _. _ ~ _. _ . . _ _ _ _ _ _ _ . _ -

B,3.5 4 9

3.51[!STRUMENTATION SYSTIZ 4

Enses continued Limiting Instrument Setpoints (continued) ,

i

4. The steam line low pressure signal is lead / lag compensated and its set point is set well above the pressure expected in the event of a large stearn line break accident as shown in the safety analysis (Reference 3).

5. The high steam line flow limit is set at approximately 20% of nominal ;

full load flow at the no load pressure and the high high steain line

" low limit is set at approximately 120% of nominal full load flow at-the full load pressure in order to protect against largo steam break accidents. The coincident low T,y, setting limit for stearn line isolat. ion initiation is set below its hot shutdown value. The_ safety analysis shows that these settings provide protection in the event of '

a large steam break (Reference 3).

6. Steam generator low

low water icvel and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provice initiation signals for the

/sux111ary Feedwater System, Selection of these setpoints is-discussed in the Bases of Section 2.3 of the Technical-Specification. ,

7. liigh radiation signals providing input to the Contaitunent Ventilation .

Isolation circuitry are set in accordance with the Radioactive ,

Effluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR part 20 at the SITE BOUNDARY.

8. The degraded voltage protection setpoint is 2t94.8% and s96.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all !

safeguards loads will operate properly at or above the minimum degraded voltage setpoint. The maximum degraded voltage setpoint is chosen to prevent unnecessary actuation of the voltage restoring.

scheme at the minimum expected grid voltage. The first degraded voltage time delay of 8 1 0.5 seconds hac been shown by testing and analysis to be long enough to allow for normal transients (i.e.,

rnotor starting and fault clearing). It is also longer than the time required to start ao safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded' ,

voltage condition to be corrected within a time frame which will not cause damage to permanently connected Class 1E loads.

B.3.10 1 3.10 CONTR0h ROD AND POWER DISTRIBUTION LIM 1H hDEil Throughout the 3.10 Technical Specifications, the terins " rod (s)" and "RCCA(s)" are synonymous.

A. Shutdown Margin A sufiicient Sl!UTDOWN MARCIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients associated with postulated accilent conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SilVTbOWN MARGIN requirements vary throughout core life as a function _ of fuel depletion, reactor coolant system boron concentration and reactor cootant average temperature. The most restrictivo condition occurs at end of life and is associated with a postulated steam line break accident and resulting uncontrolled reactor coolant system cooldown. In the analysis of thin accident, a minimu:: S)!UTDOWN MARGIN (shown in Figure TS.3.10 1 as a function of equilibrium hot full 1ower boron concentration) is required to control the reactivity transient. Accordingly, the SilVTDOWN MARGIN ;

requirements are based upon this limiting condition and are consistent with plant safety analysis assumptions. With reactor coolant system average temperature less than 200'F, the reactivity transients resulting-from a postulated steam line break cooldown are minimal and a 1% Ak/k SilVTDOWN MARCIN provides adequate protection.

In POWER OPERATION and il0T STANDBY, with k.tr a 1, Sil0TDOWN MARGIN is ensured by coiuplying with the rod insertion limitations in Specification 3.10.D. In ll0T STANDBY with h err < 1.0 and in ll0T SilUTDOWN, INTERMEDIATE SilVTDOWN and COLD SilVTDOWN, t.he SilVTDOWN ttARGIN requirements in "

Specification 3.10.A are applicable t.o provide sufficient negativo reactivity to meet the assumptions of the safety analyses discussed above.

For REFUELING, the shutdown reactivity requirements are specified in Table-TS.1-1.

When in p0WER OPERATION and ll0T STANDBY, SilUTDOWN MARGIN is determined assuming the fuel and moderator temperatures are at the nominal zero power design temperature of $47'F.

With any rod cluster control assembly not capable of being fully inserted, th'e reactivity worth of the rod cluster control assembly must be accounted for in the determination of SilOTDOWN MARGIN.

B. Power Distribution Control The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR in the core of greater than or equal to 1.30 for Exxon fuel and 1.17 for Westinghouse fuel during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. The ECCS analysis was performed in accordance with SECY 83+472. One calculation at the 954 probability level was performed as'well as one calculation with-1

B,3.10-2 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITE Bases continued B. Power Distribution Control (continued) all the required features of 10 CFR Part 50, Appendix K. The 95%

probability level calculation used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT, The Appendix K calculation used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT for the En limit specified in the CORE OPERATING LIMITS REPORT. Maintaining 1) peaking factors below the To limit specified in the CORE OPERATINC LIMITS REPORT during ail Condition I events and 2) the peak linear heat generation rate below the value specified in the CORE OPERATING LIMITS REPORT at the 954 probability level assures compliance with the ECCS analysis.

During operation, the plant staf f compares the measured hot channel factors, FqN 393 pNg, (described later) to the limits determined in the transient and LOCA analyses. The terms on the right side of the equations in Section 3.10.81 1 represent the analytical limits. Those terms on the left side reprocent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

F8 g is the measured Nuclear Hot Channel Factor,. defined as the maximum local heat flux on the surface of a fuel rod divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment.

The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fu axially. The K(Z) value is based on large and small break LOCA analyses.

V(Z) is an axially dependent function applied to the equilibrium measured FN g to bound FN o's that' could be measured at non equilibrium conditions.

This function is based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth.

FE g, Encineerine Heat- Flux ' Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering-factor allows for local variations in enrichment,' pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect-is a factor of 1.03 to be applied to fuel rod surface heat flux.

The 1.05 multiplier accounts for uncertainties associated-with' measurement of the power distribution with the movable incore detectors -and the use of those measurements to establish the assembly local power distribution.

F"g .(equil) is the measured limiting FN .obtained at equilibrium conditions-during target flux determination.

F"a, Nuclear Enthalny Rise Hot Channel Factori is defined as the. ratio of the integral of. linear power along~the rod.with the highest integrated power to the average rod power.

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ML20126H211

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