Text

Proposed Tech Specs Including Corrections to Validated Unit 2 Values,Corrections Reflecting Previously Transmitted Changes & Revised Testing Frequency for Turbine Overspeed Protection
	ML20102A476
Person / Time
Site:	Comanche Peak
Issue date:	07/20/1992
From:	; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	;
Shared Package
ML20102A462	List: ML20102A459; ML20102A476;
References
	NUDOCS 9207270068
	Download: ML20102A476 (51)
	v • d • e

.

~ -

tl g

TABLE 2.2-1 (Continued)

[$

g TABLE NOTATIONS (Continued) m 5 NOTE 1: (Continued)

E*

o g

Y For Unit 2

,,4 a,

cro "G

(i) for q g between -52% and +5.5%, f (aq) = 0, where o and g am percent g

t g

du 5

RATED THERMAL POWER in the top and bottom halves of the core respectively, 88 M

e and qt+ab is total THERMAL POWER in percent of RATED THERMAL POWER, y

o&

w M

g (ii) for each percent that the magnitude of q g

ce t

b exceeds -52%, the M-16 Trip o4 uPJ Setpoint shall be automatically reduced by 2.15% of its value at RATED THERMAL 8S POWER, and 328 ma (iii) for each perce.it that the magnitude of g t

g exce s +5. R, W N-16 Trip Setpoint shall be automatically reduced by 2.17% of its value at RATED THERMAL POWER.

u b

NOTE 2:

The channel's maximump ri Setpo nt s 11 not exceed its computed Trip Setpoint by more than 1.8%

of span for Unit 1 or.

of span f.

Unit 2.

2.88 % *

4 m

JD 5so m

O E

m

Attach::nt to TXX*92318 Page 2 of 51 2.1 SAFETY LIMITS Kenew

,ASES 2.1.1 REACTOR CORE The restrictions of this Safet;' Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface tempera-ture is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relstion has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distribu-tions. The local DNB heat flux ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a 3 articular core location to the local heat flux, is indicative of the margin to

)NB.

The DNB design basis is that the minimum DNBR of the limiting rod during Con-dition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based 9 ire applicable experimental data set such that there is a 95 percent aro-e g

,) imum DNBR is at the DNBR limit.ility with 95 percent confidence level that DNB will not d

j7 n In meeting this design basis, uncertaint es in plant operating parameters, nuclear and thermal parameters t.nd fuel fabri-i cation parameters are considered such that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit.

In addition margin has been maintainedinthedesignbymeetingsafetyanalysisDNBRlimItsinperforming safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the safety analysis limit value, or the j

average enthalpy at the vessel exit is less th n the enthalpy of saturated

)

liquid, 1

[

l COMANCHE PEAK - UNITS 1 AND 2 B 2-1

i Attachaent to TXX 92318 Page 3 of 51 fR00F SAFETY llHITS BASES k

b

_ REACTOR CORE (continued) 1hese curves are based on a nuclear enthalpy rise hot channel factor, FN and a referen e@ial power shape. An allowance is included for an increase in F3g at reduced power based on the expression:

NF g,p P [1,0 + PF (1.0 - P)]

3g where:

P = the fraction of RATED THERMAL POWER (RTP),

FhP=theFhlimitatRTP specified in the CORE OPERATING LIMITS REPORT (COLR), and 3g=thepowerfactormultiplierforFh PF specified in the COLR.

These limiting heat flux conditions are higher than those calculated for

~

the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f

(41) function of the Overtemperature N-16 trip. Vhen the axial power t

imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature N 16 trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

P.1. 2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a waximum transient pressure of n 0% (2735) psig of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3107 psig) of design pressure, to demonstrate integrity prior to initial operation.

COMANCHE PEAK - UNITS 1 AND 2 B 2-2

Attachment to VXX 92318 Page 4 of 51 fRooF 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL

$YS $Y SNUT00W MARGIN - T, GREATER THAN 200'F LIMITING CONDITION FOR OPFRATION 3.1.1.1 The SHUTDOW MARGIN shall be greater than or equa 4304k/k.

APPLICABILITY: H0 DES 1, 2*, 3, and 4.

ACTION:

' b ' Nk f

b 2

With the SHLfiDOWN MARGIN less than 41r&304k/k mmediately initi and con-tinue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restorod.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to [4r634-Ak./J.:.

7-Within I hour after detection of an inoperable control rod (s) and a.

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTOOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b.

When in HODE 1 or H0DE 2 with K,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawai is within the limits of Specification 3.1.3.6; When in H0DE 2 with K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to c.

achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3,1.3.C; d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser*

tion limit of Specification 3.1.3.6;' and

See Special Test Exceptions Specification 3.10.1.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-1 l

Attach:ent to Txx 92318 Page 5 of 51 fROOF oEACTIVITY CONTROL SYSTEMS SHUT 00VN MARGIN - T,yLESS THAN OR EQUAL TO 200'F gMITINGCONDITIONFOROPERATION

~

[

3.1.1.2 1he SHUTDOWN MARGIN shall be greater than or equal t'.3 ok/k.

APPLICABILITY: H0DE 5.

ACTION:

With the SHUTOOWN MARGIN less tha

.3 ok/k, immediately initiate and continue l

MARGIN is restored.

boration at reater than or equal O a.gpm of a solution containing greater than or equa to 7,000 ppm toron or equivalent until the required SHUT 00WN SURVEILLANCE REQUIREMENTS 4.f... The SHUTOOWN MARGIN shall be determined to be greater than or equa l

4c 11..'

ak/k:

)p/

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at If the inoperable control rod is immovable or untrippable SHUT 00WN MARGIN shall be verified acceptable with an incre,ased the allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following fa; tors:

1)

Reactor Coolant System boron concentration, 2)

Control rod position, e

3)

Reactor Cociant System average tempferature, I

4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-3 l

Attachaont to TXX 92318 Page 6 of 51 fR00F REACT}VITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the boric acid storage tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System (RCS), and b.

Two flow paths frorr, the refueling water storage tank via centrifugal charging pum to the RCS.

A_PPLICABILITY: MODES 1, c, L and 4.*

ACTION:

With only one of the above required boron in ec on flow paths to the RCS OPERABLE,restoreatleasttwoboroninje o fl paths to the RCS to 3

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be into a SHUTDOWN MARGIN equivalent to at k/k at 200 F within the t leas T STANDBY and borated next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restare at least two flow pa ERABLE status within the next 7 days or de in COLD SHUTDOWN within he next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENYS 4.1.2.2 At least two of the d ove required flow paths shall be demonstrated OPERABLE:

i a.

At least once per 7 days by verifying that the temperature of the l

flowgathfromtheboricacidstoragetanksisgreaterthanorequal to 65 F when it is a required water source; b.: At least once per 31 days by verifying that each valve (manual, power-operated, or automatic; in the flow path that is not locked, sealed, or otherwise secured in position,-is in its correct position; and At least once per 18 months by verifying that the flow path required c.

by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

'A maximum of two charging pumps shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350'F except when Snecification 3.4.8.3 is not applicable. An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

f0MANCHE PEAK - UNITS 1 AND 2 3/4 1-8 1

Attach 2ent to TXX 92318 Page 7 of 51' fROOF REACTIVITY COMTPOL SYSTEMS BORATED WATER SOURCES - OPERATING

$h$Y LIMITING CONDITION FOR OPERATION 3.1. 2. 6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A boric acid storage tank with:

1)

A minimum indicated borated water level of 50%,

2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum indicated borated water level of 95%,

2)

A boron concentration between 2000 ppm and 2200 ppm, 3)

A minimum solution temperature of 40'F, and 4)

A maximum solution temperature of 120'F.

APPLICt.3ILITY: MODES 1, 2, 3, and 4.

ACTION:

With the boric acid storage tank inoperable and being used as one a.

of the above required borated water sources, restore the tank to OPERABL s within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the ne 6 hovt and borated to a SH'c700WN MARGIN equivalent to at leas

.3$ k/k at 200'F; restore the boric acid storage tank to OPERABL statu within the next 7 days o? be in COLD S"UTDOWN within tne next s hours, b5 With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY 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 .

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-13 i

n TABLE 3 23 h

2$

g REACTOR TRIP SYSTEM '

W-

,g E

e

+irilMUM EI A

TOTAL I;0.

CHANNELS CHANNELS APPLICABLE u

$E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

~$

b 1.

Nanual Reactor Trip 2

1 2

1, 2 1

Z 2

1 2

3, 4", 5 9

k 8

8 u,

~'

2.

Power Range Neutron Flux E

a.

High Setpoint 4

2 3

1, 2 2

a.

N C

b.

Low Setpoint 4

2 3

1,2 2

3.

F'ower Range, Neutron F1 x 4

2 3

1, 2 2

3 High Positive Rate 4.

Power Range, Netitron Flux, 4

2 3

1, 2 2

High Negative Rate wd 5.

Intermediate Range, Neutron Flux 2

1 2

1,2 3

C 6.

Source Range, Neutron Flux a.

Reactor. Trip and Indication b

1)

Startup 2

1 2

2 4

2)

Shutdown 2

1 2

3,4,5 5.1 h

b.

Boron Dilution Flux Doubling

2 1

2 3

,4, 5 5.

2 7.

Overtemperature N-16 4

2 3

1, 2 12 8.

Overpower N-16 4

2 3

1, 2 12 d

9.

' Pressurizer Pressure--Low 4

2 3

I 6' %

10.

Fressurizer Pressure--High 4

2 3

1, 2 6 $

Boron Dilution Flux Doubling requirements become effective for Unit I six months after crit Cycle 3 and for Unit 2.six months after initial criticality.

l i

I n

TABLE 4.3-1 m3 REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREMENTS 5

.o :r TRIP o5 m

ANALOG ACTUATING MODES FOR

'5 5

CHANNEL DEVICE WHICH E*

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

[

FUNCTIONAL UNIT CHECK CALIBRATION TEST TES_T LOGIC TEST IS REQUIRED U

z

?

3 1.

Manual Reactor Trip N.A.

N.A.

R(14)

N.A.

1, 2, 3, 4, 5 a

a 8

[

2.

Power Range, Neutron Flux g

a.

High Setpoint S

D(2,4),

Q N.A.

N.A.

1, 2 N(3. 4),

m Q(4, 6),

R(4,5) b.

Low Setpoint 5

R(4)

S/U(1)

N.A.

I,2 C

y 3.

Power Range, Neutron Flux, N.A.

R(4)

Q N.A.

N.A.

1, 2 High Positive Rate N

4.

Power Range, Neutron Flux, M.A.

R(4)

Q N.A.

N.A.

1, 2 High Negative Rate 5.

Intermediate Range, S

R(4, 5)

S/U(1)

N.A.

3,2 c

Neutron Flux 6.

Source Range, Neutron Flux 5 R(4, 13)

S/U(1),Q(9)

R(12)*

N.A.

2, 3, 4, 5 b

7.

Overtemperature N-16 S

Q N.A.

N.A.

1, 2 8.

Overpourr N-16 5

4),

Q N.A.

N.A.

1, 2 d

9.

Pressurizer Pressure--Low S R

Q(8)

N.A.

I 10.

Pressurizer Pressure--High 5 R

Q N. A.'

N.A.

1, 2

Boron Dilution Flux Doubling requirements become effective for Unit I six months af ter criticality fcr Cycle 3 and for Unit 2 six months after initial criticality.

I TABLE 4.3-1

,tinued) no yy h

REACTOR TRIP SYSTEN INSTRIMENi' 2 '9N SURVEILLANCE REQUIREMENTS oe g

{

m oa g

TRIP ANALOG ACTUATING MODES FOR o

g CHANNEL DEVICE milch

$g CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED g

c y

z 3

18.

Reactor Trip System Interlocks (Continued)

.{

[

b.

Low Power Reactor g

Trips Block, P-7 5

1) Power Range Neutron H.A.

R(4)

R N.A.

N.A.

1, 2 Flux P-10

2) Turbine First Stage M.A.

R R

N.A.

N. A.

I yg Pressure P-13 T

c.

Power Range Neutron N.A.

R(4)

R N.A.

N.A.

I t$

F1ux, P-8 d.

Power Range Neutron N. A.

R(a)

R N.A.

N.A.

1 Flux, P-9 Power Range Neutron N.A.

R(4)

R N.A.

N.A.

1, 2 Flux, P-10 19.

Reactor Trip Breaker N.A.

N. A.

N.A.

N(7, 11)

N.A.

3a a

20. Automatic Trip and Interlock N.A.

N.A.

M(7) 1

,3 Lagic a

21.

Reactor Trip Bypass Breaker N. A.

M.A.

N.A.

M(15), R(16)

N.A.

1 58 m

O E

I n

TABLE 3.3-2 (Centinued) yg E

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRtMENTATION m

=

MINIMUM

~3 h

FUNCTIONAL UNIT TOTAL NO.

CHANNELS CHANNELS APPLICABLE ob OF CHANNELS TO TRIP OPERABLE NODES ACTION

[

7.

Automatic Initiation of ECCS

$g Switchover to Containment g

Sump (Continued) 4 w

b.

RWST Level--Low-tow 4

2 3

1,2,3,4 26 5

g Coincident With: Safety u

Injection See Item 1. above for all Safety Injection initiating functions and requirements.

8.

Loss of Power (6.9 kV & 480 V Safeguards System Undervoltage)

R*

a.

6.9 kV Preferred Offsite 2/ bus bbus

'1/ bus Source Undervoltage 1

1,2,3,4 23 ca b.

6.9 kV A1 ternate Offsite 2/ bus 2/ bus 1/ bus f

Source Undervoltage 1, 2, 3, 4 23 6.9 kV Bus Undervoltage 2/ bus c.

9 2/ bus p 1/ bus 1, 2, 3, 4 23 d.

6.9 kV Degraded Voltage 2/ bus 2/ bus 1/ bus i 1, 2, 3, 4 23 480 V Degraded Voltage 2/ bus 2/ bus 1/ bus 1 2,3,4 23 e.

\\

1 f.

480 V Low Grid 2/ bus'.

2/ bus

)/ bus Undervoltage

,1, 2, 3 l 23 E

9.

Control Room Emergency m

10 Recirculation a.

Manual Initiation 2

1 2

All 24 m

D L

m

l TABLE 3.3-2 (Continued) c, h

IN e

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM m =r A

TOTAL NO.

CHANNELS CHANNELS APPLICABLE o$

][

y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION l

s<

b.

Safety Jnjection See Item 1. a~oove for all Safety Injection initiating functions ano l

, 5{

requirements a

e Y

h i

10.

Engineered Safety Features m

g Actuation System Interlocks g

I

>z a.

Pressurizer Pressure, 3

2 2

1, 2, 3 18 o

i I

P-11 b.

Reactor Trip, P-4 2

2 2

1,2,3 20 11.

Solid State Safeguards w}

Sequencer (5555) a.

Safety Injection 1/ train 1/ train 1/ train 1, 2. 3, 4 12 Sequence

~

b.

Blackout Sequence 1/ train 1/ train 1/ train 1, 2, 3, 4 25 l

~

N bo w4tol koe m E,,,,

kec:ceuf,;;,,..

$ s0

~

m O

E m

A n

TABLE 3.3-3 2R ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Q

n w5

/

?,

o$

SENSOR R

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z c

(S)

TRIP SETPOINT ALLOWABLE VALUE w

5 1.

Safety Infection (ECCS, Reactor Trip, U

Ut Feedwater Isolation, Control Room a

Emergency Recirculation, Emergency y

g Diesel Generator Operation, Contain-g

>5 ment Vent Isolation, Statfori Service Water, Phase A Isolation, Auxiliary m

Feedwater-Notor Driven Pump Turbine Trip, Component Cooling Water, Esser.tlal Ventilation Systems, and Containment Spray Pump).

w D

a.

Manual Initiation M.A.

N.A.

A b.

Automatic Actuation Logic M.A.

N.A.

ta and Actuation Relays c.

Containment Pressure--High 1 2.7 0.71

1. 7 1 3.2 psig 5 3.8 psig d.

Pressurizer Pressure--Low

1) Unit 1 15.0 10.91 2.0

>1 psig

> 1803.6 psig

2) Unit 2 15.0 11.3 2.0

[

psig i1803.6psig e.

Steam Line Pressure--Low

1) Unit 1 17.3 15.01

2. 0

> 605 psig*

> 593.5 psi

2) Unit 2 17.3 9.15 2.0

[605psig*

[578.4 psi 2.

Containment Spray g

a.

Manual Initiation N.A.

N.A.

M.A.

N.A.

b.

Automatic Actuation Logic N.A.

N. A.

N.A.

and Actuation Relays c.

Containment Pressure--High-3 2.7 0.71

1. 7 5 18.2 psig i18.8psigh E

d TAL1LE 4.3-2 (Ccatinued) n

,g h

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SE Mm SURVEILLANCE REQUIREMENTS 7{

a TRIP

., A g

ANALOG ACTUATING MODES x

CHANNEL DEVICE MASTER SLAVE FOR milch

'- o 8

CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE d

g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 7

3

3. Containment Isolation (Continued) w

c. Contlineent Vent Isolation 5

1) Manual Initiation See Item 3.a.1 and 2.a above. Containment vent isolation is manually 1,2,3,4 initiated when Phase "A" isolation function or containment spray m

function is manually 1:iitiated.

2) Automatic Actuation N.A.

N. A.

N.A.

M(1)

Q 1,2,3,4 Logic and Actuation mg Relays

3) Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.

w

4. Steam Lire Isolation

a. Manual-Initiation-N.A.

N.A.

R N.A.

N.A.

1, 2, 3

b. Automatic Actuation M.A.

N. A N.A N.A.

M(1)

M(I)

Q 1, 2, 3 Logic and Actuation Relays

c. Containment Pressure-S R

Q N.A.

N.A.

1, 2, 3 High-2

d. Steam Line 5

R Q

H.A.

P. A.

N.A.

2, 3 Pressure-Low

[g e, Steam Line Pressure-S R

N.A.

Negative Rate-High

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation N.A.

N.A.

M(1)

Q

,2 Logic and Actuation Relays

d n

h TABLE 4.3-2 (Continued) o

.[h ENGINEE'dED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATI Q

aa SURVEILLANCE REQUIREMENIS S

A R

TRIP 5

ANALOG ACTUATING

[

CHANNEL CHANNEL DEVICE MODES e-CHANNEL CHANNEL OPERATIONAL OPERATIONAL -ACTUATION RELAY RELAY SURVEILLANCE 7 MASTER SLAVE FOR WHICH Q

FUNCTIONAL UNIT N

CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQ 8 TIRED %

7. Autom& tic Initiation of

' ECCS Switchover to Contairment g

Sump (Continued) h O

N

b. RWST Level-Low-Low S

SR N

N.A.

M.A 1, 2, 3, 4 Coincident With f

f r

Safety Injection w

See Item 1. above for all Safety Injection Surveillance Requirements.

1

8. Loss of Power (6.9 kV &

J, 480 V Safeguards c'

System Undervoltage) 6.9 kV Prefeired Offsite a.

Source Undervoltage N.A.

R N.A.

(3,2)

N.A.

1, 2, 3, 4 b.

5.9 kV Alternate Offsite Source Un<iervoltage N.A.

R N. A.

(3, 2)

N.A.

1, 2, 3, 4 c.

6.9 kV Bus Under-voltage N.A-R N.A.

(3, 2)

N.A.

M.A.

N.A.

2, 3 d.

G.9 kV. Degraded Voltage N. A.

R N.A.

(3. 2)

N.A.

H.A.

N.A.

, 2, 3 e.

480 V Degraded Voltage N.A.

R N.A.

(3, 2)

N.A.

H.A.

N.A.

, 2, 3 4

f.

480 V Low Grid Undervoltage N.A.

R H.A.

(3, 2)

N. A.

N. f..

N.A.

. 2. - 3,g RI

Attach: nt to TXX-92318 Page 16 of 51 fRooF INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The explosive gas monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded.

APPLICABILITY:

n a 3.3-7.

ACTION:

( gl+3 1 pel 2.)

a.

th an explos eaa monitoring instrumentation channel Alarm /

Trl;rie ess conservative than required by the above specifi-cation, declare the channel inoperable and take the ACTION shown in lable 3.3-7.

b.

With less than the minimum number of explosive cas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-7.

Restore the inoperable-instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.

c.

The provisions of Specification 2.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.4 Each explosive gas monitoring instrumentation channel shown in Table 3.3-7 shall be demonstrated OPERABLE:

a. ' At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performance of a CHANNEL CHECK, b.

At least once per 31 days by performance of an ANALOG CHANNEL OPERATIONAL TEST, and c.

At least once per 92 days by performance of a CHANNCL CALIBRATION which shall include the use of standard with the manufacturer's recommendations. gas 1,amples in accordance COMANCHE PEAK - UNITS 1 AND 2 3/4 3-48

Attachment to TXX 92318 i

Page 17 of 51 ROOF INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION EWEW LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.

APPICABILITY: MODES 1, 2*, and 3*.

ACTION:

a.

With one stop valve or one control valve per high pressure turbine steam line inoperable and/cr with one stop valve or one control valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.

SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.

4.3.4.2 The above required o s

rotection system shall be tiemonstrated OPERABLE:

6 M5 a.

At least cuce p r 14 tys b cycling each of the following valves through 6t leas Ac e cycle from the running position using the manual test or Automatic Turbine Tester (ATT):

1)

Four high pressure turbine stop valves, 2)

Four high pressure turbine control valves, 3)

Four low pressure turbine stop valves, and 4)

Four low pressure turbine control valves.

b. At least once per 14 testing of the two mechanical overspeed devices using the omati c reine Tester or manual test c.

At least once pe 3h d$Ny frect observation of the movement of each of the above he ugh ons. complete cycle from the running position.

d.

At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats (if applicable), disks and stems and verifying no unaccept-able flaws.

If unacceptable flaws are found, all other valves of that type shall be inspected.

rNot applicable in MODES ? and 3 with all main steam line isolation valve: and associated bypass valves in the closed position.

COMANCHE PEAK - UNITS 1 AND 2 3/4 3-51

Attachaent to TXX-92318 Page 18 of 51 fROOF REACTOR COOLANT SYSTEM HOT SHUTDOWN gRS $Y$ $Y LIMITINGCONDITIONFOROPERATIONkodigue ACTION: (Continued) b.

With no loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), and/or RHR pump (s) if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE By verifying secondary side water level to be greater than or equal a.

to 10% (narrow range) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and b.

By performing the surveillances pursuant to Specification 4.0.6.

4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

COMANCHE PEAK - UNITS 1 AND 2 3/4 4-5

Attach ent to TXX-92310 Page 19 of 51 fRooF REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.5.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

Monitoring the Reactor Coolant System Leakage Detection System a.

required by Specification 3.4.5.1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; b.

Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; Performance of a Reactor Coolant System water inventory balance at c.

least within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving steady state operation" and at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter during steady state operation, except that no more than 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months shall elapse between any two successive inventory balances.

The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4; and d.

Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.5.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in 4

Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a.

At least once per 18 months, b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, except for valves 8701A, 8701B, 8702A, and 8702B.**

Prior to returning the valve to service following maintenance, c.

repair or replacement work on tho valve, and d.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following check valve actuation due to flow through the valve.

As c,utlined in the AE"E Ccde, SceMen XI, paragraph-IW-34C(b).

c.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

"T,yg being changed by less than 5'F/ hour.

C*This exception allowed since these valves have control room position indication, inadvertent opening interlocks and a system high pressure alarm.

COMANCHE PEAK - UNITS 1 AND 2 3/4 4-15

Attacha2nt to TXX 92318 Page 20 of 51 R00F EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

EWEW 2)

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion, At least once per 18 months, during shutdown, by:

e.

1)

Verifying that each automatic valve in the flow path actuates to its correct position on Safety Injection actuation test signals, and 2)

Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

4 a)

Centrifugal charging pumps, 4

b)

Safety injection pumps, and c)

RHR pumps.

v f.

By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1)

Centrifugal charging pump 1 2370 psid, 2)

Safety injection pump

> 1440 psid, and 3)

RHR purup

> 170 psid.

By verifying the correct position of each mechanical position stop g,

for the following ECCS throttle valves:

. 1)

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2)

At least once per 18 mont CCP/SI System Valve Number SI System Valve fiumber SI-8810A SI-8822A SI-8816A SI-8810B SI-8822B SI-8816B SI-8810C SI-8822C SI-8816C SI-8810D SI-8822D SI-8816D COMANCHE PEAK - UNITS 1 AND 2 3/4 5-5

Attachoont to TXb92318 Pago 21 of 51 R00F 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT EWEW CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

A_PPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I hour or be in at least HOT STANDBY 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 .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that all penetrations

not a.

capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 2.1.1 of the Technical Requirements Manual.

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and After each closing of each penetration subject to Type B testing, c.

except ti s containment air locks, if opened following a Type A or B

, test, by leak rate testing the seal with gas at a pressure not less than P, 48.3 psig, and verifying that when the measured leakage rate forth$sesealsisaddedtotheleakageratesdeterminedpursuantto Specification 4.6.1.2d. for all other Type B and C penetrations, the corabined leakage rate is less than 0.60 L,.

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed pesition. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. The blind flange on the fuel transfer canal need not be verified closed except after each drainage of the canal.

COMANCHE PEAK - UNITS 1 AND 2 3/4 6-1 1

m_._ _ _ _ _ - - - - -

Attachment to TXX-92318 Page 22 of 51 CONTAINMENT SYSTEMS 3/4.6,3 CONTAINMENT ISOLATION VALVES Kevo eoa LlHITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves shall be OPERABLE.#

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

"With one or more of the containment isolation valve (s) inoperabic, maintain at least one isolation valve OPERABLE in eacn affected penetration that is open and:

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , a.

or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least c.

one closed manual valve or blind flange, or d.

Be in at least HOT STANDBY 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 .

SURVEILLANCE REQUIREMENTS 4.6.3.1 The containment isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by perfonnance of a cycling test, and verification of isolation time.

The requirements of Specification 3.6.3 do not apply for those valves covered by Specifications 3. 7.1.1, 3. 7.1. 5, g3. 7.1. 6, ud J 7. l. 7.

CAUTION: The inoperable isolation valve (s) may be part of a system (s).

Isolating the affected penetration (s) may affect the use of the system (s).

Consider the technical specification requirements on the affected system (s) and act accordingly.

COMANCHE PEAK - UNITS 1 AND 2 3/4 6-13

Attachzent to TXX 92318 Page 23 of 51 ROOF PLANT SYSTEMS 3/4.7.4 STATION SERVICE WATER SYSTEM EWEW OPERATING LIMITING CONDITION FOR OPERATION 3.7.4.1 At least two independent station service water loops per unit and the cross connect between the Station Service Water Systems of each unit shall be OPERABLE.

,gj j 7,.]

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With only one station service water loop per unit OPERABLE, restore at least two loops per unit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or for the unit (s) with the inoperable station service water loop be in at least HOT STANDBY 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 .

With one or more of'the cross-connects or cr;s; c;nn;ct veh;(;) y/

b.

inoperable, within 72,,heure res, tor,e the cross.. con,nects to OPE.RABLE s

b,1,to,$)'chE$[kherwise'beIn'ai'IeastA0T55055Y5Ithin'th7next6-ive(

...........s

.m

~

hours and in COLD SHUTD0b within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.4.1.1 Each station service water loop shall be demonstrated OPERABLE:

At least once per 31 days by) verifying that each valve (manual, a.

power-operated, or automatic secc cing safety-related equipment

, that is not locked,

' its correct positio u d, or otherwise secured in position is in e

b.

At least once per 18 months during shutdown, by verifying that'each station service water pump starts automatically on a Safety Injection test signal.

4.7.4.1.2 At least once per 92 days the crofs-c nects shall be demonstrated OPERABLE by cycling the cross-connect valves dr' erifying that the valves are locked open.

1 COMANCHE' PEAK - UNITS 1 AND 2 2/4 7-14 d

Attachmont to TXX-92318 Page 24 of 51 fROOF PLANT SYSTEMS STATION SERVICE WATER SYSTEM EV# EW ONE UH1T SHUTDOWN pter LlHITlHG CONDITION FOR OPERATION 3.7.4.2 At least two independent s ation service water loops in the operating snit *, at least one station service ' pump in the shutdown unit ** and the cross-connects from the OPERABLE station service water pump (s) in the shutdown unit to the station service water loops of the operating unit shall be OPERABLE.

APPLICABILITY:

Unit 1 (Unit 2) in HODES 1, 2, 3 and 4 Unit 2 (Unit 1) in MODES 5, 6 and Defueled ACTION:

With one station service water loop in the operating unit

~

a.

inoperable, restore two loops in the operating unit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With one or more of he onnects-07 Orc:: ::nn::t v:1 between the OPERABLE station service water pump (s) in the shutdown unit and the station. service water loops in the operating unit inoperable, within 92 h:gre restore the cross-connects to OPERABLE st 1-v:atus.or :p:r th: Offected v:h:(:) :nd :. int:in th: :ff;;t:d

~

h ( O eper. Otherwise, place the operating unit in at least HOT STANDBY 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 .

5 If neither station service water pump in the shutdown unit is c.

o.

OPERABLE, restore at least one pump to OPERABLE status within V

h:urt or place the operating unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months andV OLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C 19 SURVEILL%NCE REQUIREMENTS Eachstatkonservicewaterloopintheoperatingunitshallbe 4.7.4.2.1 demonstrated OPERABLE per the requirements of Specification 4.7.4.1.1.

4.7.4.2.2 At least once per 92 days the cross-connect (s) between the OPERABLE station service water pump (s) in the shutdown unit and the station service water loops in the operating unit shall be demonstrated OPERABLE by c locked open.

A Unit in MODE 1, 2, 3 or 4 is designated as the " Operating unit".

- A unit is MODE 5, 6 or Defueled is designated as the " shutdown Unit".

COMANCHE PEAK - UNITS 1 AND 2 3/4 7-15

Attachoont to TXX 92318 Page 25 of 51 ROOF PLANT SYSTEMS

/M SURVEILLANCE REQUIREME 5 (Co"Wueb 4

w

~

b.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:

1)

Verifying that the filtration unit satisfies the in-place pene-tration and bypass leakage testing acceptance criteria of less than 0.05% by using the test procedure guidance in RegJatory Position C.5.a, C.S.c, and C.5.d of Regulatory Guide 1.52, Revi-sion 2, March 1978*, and the emergency filtration unit flow rate is 8000 cfm i 10%, and the emergency pressurization unit flow rate is 800 cfm i 10%;

2)

Verifying, within 31 days af ter removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2 March 1978*, meets the laboratory testing criteria of Regulatory Position C 6 a of Regulatory Guide 1.52, Revi-sion 2, March 1978*, for a methyl iodide penetration of less than 0.2%; and 3)

Verifying an emergency filtration unit flow rate of 8000 cfm i 10% and an emergency pressurization unit flow rate of 800 cfm i 10% during system operation when tested in accordance with ANSI N510-1980; Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, c.

within 31 days after removal that a laboratory analysis of a repre-sentativecarbonsampleobtaInedinaccor6ancewithRegulatoryPosi-tion C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, meets the laboratory testing criteria of Regulatory Position C.6.a of Regu-latory Guide 1.52 Revision 2, March 1978*, for a methyl iodide penetration of less than 0.2%;

ANSI N510-1980 and ANSI N509-1980 shall be used in place of ANSI H510-1975 and ANSI H509-1976, respectively.

COMANCHE PEAK - UNITS 1 AND 2 3/4 7-19

Attach 2ent to TXX 92318 Page 26 of 51 PLANT SYSTEMS CONTROL ROOH HVAC SYSTEM

$h$Y SHLITOOWN LIMITING CONDITION FOR OPERATION 3.7.7.2 Two independent control room HVAC trains shall be OPERABLE.

AFPLICABILITY:

H0 DES 5 and 6:

ACTION:

With one control room HVAC train inoperable, restore the inoperable a.

train to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE control room HVAC train in the emergency recirculation mode.

' With both control room HVAC trains inoperable, or with the OPERABLE

@b.

control room HVAC train required to be modebyACTIONa.,notcapableofbeink;the9mergencyrecirculation powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

,r SURVEILLANCE REQUIREMENTS 4.7.7.2 Each control room HVAC train shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.7.7.1.

e COMANCHE PEAK - UNITS 1 AND 2 3/4 7-21

Attach;ent to TXX-9231B Page 27 of 51 TABLE 3.7 AREA TEMPERATURE MONITORING EV# EwJ MAXIM AREA TEMPERATURE LIMIT ('F)

Normal Abnormal Conditions Conditions 1.

Electrical and Control Building Normal Areas 104 131 Control Room Main Level (E1. 830'-0")

80 104 Control Room Technical Support Area (El. B40'-6")

104 104 UPS/ Battery Rooms 104 113 Chiller Equipment Areas 122 131 2.

Fuel Building Normal Areas 104 131 Spent Fuel Pool Cooling Pump Rooms 122 131 3.

Safeguards Buildings Normal Areas 104 131 AFW, RHR, SI, Containment Spray Pump Rooms 222 131 RHR Valve and Valve Isolation Tank Rooms 122 131 RHR/CT Heat Excharger Rooms 122 121 Diesel Generator Area 122 131 Diesel Generator Equipment Rooms 130 131 Day Tank Room 122 131 4.

Auxiliary Building c C. P

) A rmal Areas No 104 131 AP ' R!;R, ;I -Cor,teire.f..t 0 prey-Pump Rnoms 122 131 CCW Heat Exchaager Area 122 131 CVC4 Valve and Valve Operating Rooms 122 131 Auxiliary Steam Drain Tank Equipment Room 122 131 Waste Gas Tank Valve Operating Room 122 131 5.

Service Water intake Structure 127 131 6.

Containment Buildings General Areas.

120 129 Reactor Cavity Exhaust 150 190 CRDM Shroud Er.haust 163 172 e

COMANCHE PEAK - UNITS 1 AND 2 3/4 7-26

Attachmont to TXX-92318 Page 28 of 51 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION [C d medh E[I E($j ACTION (Continued)

With one offsite circuit and one diesel generator of the above required c.

A.C. electrical power sources inoperable, demonstrate the OPERABILITY r the remaining A.C. offsite source by performing Surveillance e

Requirement 4.8.1.1 la. within I hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months t:

eaf ter, and, if the diesel generator became inoperable due to any cause other than preplanned preventative maintenance or testing, demon-t. ate the OPERABILITY of the remaining OPERABLE diesel generator b performing Surveillance Requirement 4.8.1.1.2a.4) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months *, y unless the OPERABLE diesel generator is already operating #Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN w A.C. power source (ithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore the'other offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of 3.8.1.1, ACTION statement

a. or b., as appropriate, with the time requirement of the ACTION statement based on A.C. power sourec. the time of initial loss of the remaining inoperable A successful test of diesel generator OPERABILITY per Surveillance Requirement 4.8.1.1.2a.4) performed under the ACTION s

diesel generator satisfies the diesel generator test "quirem ACTION statement a. or b.

d.

With one diesel generator inoperable, in addition to ACTION b. or c

bove, verify that:

1.

All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and 2.

When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STAWDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN s following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , With two of the above required offsite A.C. circuits inoperable, demon-e.

Surveillance Requirement 4.8.1.1.2a.4) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unles diesel generators are already operating #

in at least HOT STANDSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . inoperable With only one j

diesel generator is restored to OPERABILITY.This test is required to be c 0uring performance of surveillance activities as a requirement for ACTION statements, the air-roll test shall not be performed.

COMANCHE PEAK - UNITS 1 AND 2 3/4 8-2

Attachment to TXX-92318 Page 29 of 51 i

ELECTRICAL POWER SYSTEMS i

^ LIMITING CONDITION FOR OPERATION ((.mp;m,ec))

.k. g gg 3 g h J ACTION (Continued) offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , f.

With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveil-lance Requirement 4.8.1.1.la. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel gener-ators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY 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 . Restore at least two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class IE Do tribution System shall be:

Determined OPERABLE at least once per 7 days by verifying correct a.

breaker alignments, indicated power availability, and b.

Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) the 6.9 kV safeguards bus power supply from the preferred offsite source to the alternate offsite source.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a.

In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1)

Verifying the fuel level in the day fuel tank, 2)

Verifying the fuel level in the fuel storage tank, 3)

Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day fuel tank, 4)

Verifying the diesel rtarts from ambient condition and acceler-ates to at least 441 rpm in less than or equal to 10 seconds.*

All planned diesel engine starts for the purpose of this surveillance may be preceded by a prelube period in accordance with vendor recommendations.

COMANCHE PEAK - UNITS 1 AND 2 3/4 8-3

Attach ent to TXX 92318 Page 30 of 51 fR00F RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS

$h$Y LIQUID HOLDUP TANKS

LIMITING CONDITION FOR OPERATION

@ ou.11.1 3

The quantity of radioactive material c a ned in each unprotected tdoor tank shall be limited to less than o e tritium and dissolved or entrained noble ga s@ qual to 10 Curies, excludir.g APPLICABILITY: At all times.

ACTION:

a.

With the quantity of radioactive material in any unprotected outdoor tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b.

The provisions of Specifications 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1 The quantity of radicactive material contained in each of the unprotected outdoor tanks shall be determined to be within the above_ limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

u

Tanks included in this specification are those unprotected outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

COMANCHE PEAK - UNITS 1 AND 2 3/4 11-1

Attachoont to TXX-92318 Page 31 of 51 t

Kevo en BASES FOR SEClIDNS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS NOTE The BASES contained in succeeding pages sumarize 4

the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

d i

COPANCHE PEAK - UNITS 1 AND 2 B 3/4 0-0

Attachment to TXX-92318 Page 32 of 51 fROOF gPljCABILITY BASES EWEW Specification 3.0.2 establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the speci-fied time intervnl.

The pu. Jose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a Limiting Condition of Operation is restored within the time interval specified in the associated ACTION requirements.

Specification 3.0.3 establishes the shutdown ACTION requirements that must be implemented when a Limiting Condition for Operation is not met and the condi-tion is not specifically addressed by the associated ACTION irements.

The

@ purpose of this specification is to delineate the time lim Of r placing the unit in a safe shutdown MODE when plant operation cannot by (m intained within the limits for safe operation defined by the Limiting Conditi6 s for Operation and its ACTION requirements.

It is..vt intended to be used as an c,perational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to pre-pare for en orderly shutdown before initiating a change in plant operation.

This time permits the operator to coordinate the reduction in electrical genera-tion with the load dispatcher to ensure the stability and availability of the electrical grid.

The time limits specified to reach lower H0 DES of operation permit the. shutdown to proceed in a centrolled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE.

This reduces thermal stresses on components of the primary coolant system and i

the potential for a plant upset that could challenge safety systems under con-ditions for which this specification applies.

If remedial measures permitting limited continued operation of the facility l

under the provisions of the ACTION requirements are completed, the shutdown t

may be t.erminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation.

Therefore, the shutdown may be terminated if the ACTION require-ments have been met or the time limits of the ACTION requirements.have not expired, thus providing an allowance for the completion of the required actions.

The time limits of Specificat' ion 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in the COLD SHUTDOWN MODE when a shutdown is required during the POWER HODE of operation.

If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE of operation applies.

However, if a lower MODE of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other applicable MODE, is not reduced.

For example, if HOT STANDBY is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total time to l

reach HOT SHUTDOWN is not reduced from the allowable limit of 13' hours, l

l l

COMANCHE PEAK - UNITS 1 AND 2 8 3/4 0-2

Attachment to TXX-92318 Pago 33 of 51 fROOF APPLICABILITY EIkE N BASES co1vention of identifying valves; without the unit designator if the remainder o' the tag number is applicable to both units, with the unit designator if the tag is only applicable to one unit.

@When a specification is s aredidentifier "(Units 1 and 2) J er 3.0.Sc the ACTION section contains the Specifications 4.0.1 through 4.0.6 establish the general requirements applicable to Surveillance Requirements.

These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 60.36(c)(3):

" Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that the limiting conditions of operation will be met."

Specification 4.0.1 establit ss the requirement that surveillarces must be met during the OPERATIONAL MODES or other conditions for which the requirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement. The purpose of this specification is to

~

ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a MODE or other specified condition for which the associated Limiting Conditions for Operation are appli-cable.

Surveillance Requirements do not have to be performed when the facility is in an OPERATIONAL MODE for which the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements associated with a Special Test Exception are-only applicabit when the Special Test Exception is used as an allowable exception to the requirements of a specification.

Specification 4.0.2 establishes the limit for which the specified time interval for Surveillance Requirements may be extehded.

It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consider'ation of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities.

It also provides flexibility to accom-modate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18-month surveillance interval.

It is not intended that this provision be used repeatedly as a convenience to extend the surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of Specification 4.0.2 is based on engineering judgment and the recognition that the most prob-able result of any particular surveillance being performed is the verificaion of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillar.ce interval.

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 0-4

4 Attachnent to TXX 92318 Page 34 of 51 fR00F 3/4.1 REACTIVITY CONTROL SYSTEMS, i

EWEp BASES i

3/4.1.1 BORATION CONTPOL 3/4.1.1.1_ and 3/4.1.1.2 SHUTOOWNMARGJN A sufficient SHUTOOWN MARGIN ensures that:

(1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within ecceptable limits, and (3) the reactor will be saintained sufficiently suberitical to preclude inadvertent criticality in._the shutdrxLgondition. --

3

_tT s */, A W.6c (4 e, I (. 3 % K% kr M M j' SHUTDOWN MARGIN 4equire'E,elitt-vaey-thtoUg7imDre life as a function of fueldepletion.)Ctboronconcentration,andRCST,9g. The most restrictive a

ondition occuyt at E0L, with T et n 1 ad operating temperature, and is avg as ociated with a postulated steam line break accident and resulting uncon-j t 11ed pc4 t,A In the analysis of this accident, a minimum SHUTOOWN M

IN o J ulk s required to control the reactivity transient. Accord-

'fi ly, t e-SHUT RGIN reQwf and is consistent with FSAR4 e ent is based upon this limiting condition

)

nalysis assumptions. With T,yg less than 200'F, a SHUTDOWN MARGIN k/k provides adequate protection and is based on the results of t ilution accident analysis.

Since the actual overall core reactivity balance comparison required by 4.1.1.1.2 cannot be performed until after criticality is attained, this comparison is not required (and the provisions of Specification 4.0.4 are not applicable) for entry into any Operational Mode within the first 31 EFPD following initial fuel load or refueling.

3/4.1.1.3 MODERATORTEMPERAJUREC0 EFFICIENT The limitat eens on moverator temperature coefficient (MTC) are provided to ensure that the value or this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of pleri ;onditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accur4t co arison.

The most negative va ue equivalent to the most positive moderator density coef ficient (MC),.y obtaineo by incres.<ntally correcting the MDC used in the FSAR analyth re nominal operating cor,ditions. These corrections C0KANCHE PEAK - UNIT 1 AND 2

$ :,,4 1-1

Attachasnt to TXX-92318 Pago 35 of 51

? ROOF REACTIVITY CONTROL SYSTEMS BASES EW&W H0DERATOR TEMPERATURE COEFFICIENT (Continued) condition of al' rods inserted (most positive MDC) to an a condition and, a conversion for the rate of change of moderator density with temperaturo at RATF: THERML POWER conditions.

transformed inte the limiting End of Cycle Life This value of the MDC was then surveillance limit MTC value represents a conserv(ative value (with correctio EOL) MTC value.

The 300 ppu foi burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EOL HTC valu The Surveillance Requirements for measurement of the MTC at the be and near the end of the fuel cycle are adequate to confirm that the MTC remai within its limits since this coefficient change? slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY ith the Reactor Coolant System average temperature less limitatiop-equired to ensure:

This (1) the moderator temperature coefficient is withitt it$

alyzed temperature range, (2) the trip instrumentation is within its normals ating range, (3) the OPERABLE status with a steam bubble,presserizer is capable of being in an NDT temperature.

and (4) the reactor vessel is above its minimum RT 3/4.1.2 BORATION SYSTEMS availabledurinneachmodeoffacilityoperation.The Boron Injection Sys i

(1) borated water sourcesThe components required to perform this function include:

power supply from OPERABLE diesel generators.(3) separate flow (2) charging pumps p g (),3 3 %

7 3

With the RCS average temperature above 200 injection flow paths are required to ensure sin / b minimum of two boron fy va M gle functional capability in the event an assumed failure renders one of th boratien capability of either flow path is s flow paths inoperable.

The MARGIN from expected operating conditions of provide a SHUTDOWN i

cooldown to 200'F.

The maximum expected boration capability requirementafter xenon deca m _m

{

occurs at E0L from full power equilibrium xenon conditions and requires J15,700

[70,702)] gallons of 2000 ppm borated water from the r tank (RWST).

COMA"CHE PEAK - UNIT 1 AND 2 B 3/4 1-2 T

Attachoont to TXX 92318 Page 36 of 51 fROOF REACTIVITY CORTROL SYST[MS BASES E b_f BORAT10HSYSTEMS(Continued)

With the RCS temperature below 200'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron InjectionSystembecomesinoperable.

The limitation for a maximum of two charging pumps to be OPERABLE and the requirement to verify one charging pump to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The limitation for minimum solution temperature of the borated water soue s are sufficient to prevent boric acid crystallization with the highest allowabi boron concentr tie.

The bo on apa 1 y required below 200'F is sufficient to provide a SHbi-DOWN MARGI iof 1.3 A /k after u non decay and cooldown from 200'F to 140'F.

i This condition equ r either(1,100) the boric aBdatorde tanks or (7,113) gallons of 7000 ppm borated water from gallons of 2000 ppm borated water from the RWST.

As listed below, the required indicated levels for the boric acid storage tanks and the RWST include allowances for required / analytical volume, unusable volume, measurement uncertainties (which include instruatnt error and tank tolerances, as applicable), system configuration requirements, and other required volume.

Ta ter.

MODES Ind.

Unusable Required Measurement System Other Level Volume Volume Uncertainty Config. (gal)

(gal)

F RWST 5,6 24%

45,494 7,113 4% of span 57,857 N/A l

)

1,2,3,4 p5%

45,494 70,702 4% of span N/A 357,535" 9'

/

Boric 5,6 10%

3,221 1,100 6% of span N/A N/A Acid 5,6 20%

3,221 1,100 6% of span 3,679 N/A j Stora le (gravity feed)

Tank 1,2,3,4

,50%

3,221 15,700 6% of span N/A N/A TLe OPERABILITY of one Boron Injection System durinc REFUELING ensures that this system is available for reactivity control vhile in MODE 6.

" Additional volume required to meet Specification 3.5.4.

COMANCHE PEAK - UNIT I AND 2 8 3/4 1-3 l

l

+

Attach::nt to Txx 92318 Page 37 of I'l OFROOF POWER DISTRIBUTION LIMITS i

Keoew

, TACTOR (Continued) HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR r s

clad temperature will not exceed the 2200*F ECCS acce j

periodically as specified in Specifications 4.2.2 and 4.2.3.Each survelliance is sufficient to ensure that the limics are maintained provided.

This periodic

)

Control rods in a single group move together with no individual rod a.

insertion differing by more than i 12 steps, indicated, from the group demand position; i

b.

Control rod groupt are sequenced with overlapping groups as described in Specification 3.1.3.6; t

3.1.3.6 are maintained; andThe control rod insertion limits of Specif c.

d.

OIFFERENCE, is maintained within the limits.The axial powe ifgwillbemaintainedwithinitslimitsprovidedConditionsathrou

d. above are maintained.

The relaxation of FN. as a function of THERMAL POWER allows changes in 15e radial power shape for all permissible rod insertion limits.

g Fuel rod bowing reduces the value of the DNB catio.

offset 9.1% o,thi d"".*0n in the generic maroin. The Credit is available to and 10.1% for typical cells and 9. % for thimble, cells foreneric margin t r Uni Unit; completely offset any rod bow penalties.

the to p This margi,1 includes or Unit 1:

Design limit DNBR of 1.30 vs 1.28, a.

b.

Grid Spacing (K,) of 0.016 vs 0.059, Thermal Diffusion Coefficient of 0.038 vs 0.051, c.

d.

DNBR Multiplier of 0.86 vs 0.88, and e.

Pitch redurtion.

the safety analysis limit DNBR and the design Ifmit DNBR.The The applicable values of-rod bow penalties are referenced in the FSAR.

-COMANCHE PEAK - UNIT 1 AND 2 B 3/4 b 4

Attach:ent to TXX 92318 Paga 38 of 51 fROOF r

POWER DISTRIBUTION LIMITS BASES E YI E N 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parame-ters are maintained within the normal steady state envelope of operation as*

sumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR at or above the safety analysis limit value

+ throughout each analyzed transient.

The Unit 1 indicated T value of $92.7'F conservatively rounded to 592'F) and the Unit 1 indic1ted essurizer pres-f sure value of 2207 psig corresoond to analytical lin.its of 594.7'r and 2193 psig respectively, with allowance for measurement uncertainty. The Unit 2 indicated T,yg value of 592.8'F (conservatively rounded to 592*F) and the Unit d

2 indicated pressurizer pressure value of 2219 psig correspond to analytics1 limits of 595.16'F and 2205 psig respectively, with allowance for reasurement uncertainty. The indicated uncertainties assume that the reading from four channels will be averaged before comparing with the required limit.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation, and to detect any significant flow degradation of the Reactor Coolant System (RCS).

The additioral surveillance requirements associated with the RCS total flow rate are suff'.n ent to ensure that the measurement uncertainties are limited to 1.E as assumed in the Improved Thermal Design Procedure Report for CPSES.

Performance of a precision secondary calorimetric is required to precisely determine the RCS temperature. The transit time flow meter, which uses the N-16 system signals, is then used to aucurately measure the RCS flow.

Subse-quently, the RCS flow detectors (elbow tap differential pressure detectors) are normalizedtothisflowdeterminationandusedthroughoutthecycle.

COMANCHE PEAK - UNIT I AND 2 B 3/4 2-6 l

Attach ent to TXX 9231B Page 39 of 51 fROOF PJANTSYSTEMS i

[ASES EWEW

,3/4. 7.1.,2 AUXILIARY FEF0 WATER fySTEM i

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal ooerating conditions in the event of a total loss-of-offsite power.

Each electric motor driven auxiliary feedwater pump is capable of deliver-ingatotalfeedwaterflowof430ppmtotwosteamgenerstorsatapressureof 1221 psig to the entrance of the steam generators.

The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 860 gpm to four steam generators at a pressure of 1221 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temp-erature to less than 350'F when the Residual Heat Removal System may be placed into operation.

The Auxiliary Feedwater System is capable of delivering a total feedwater flow of 430 gpm at a pressure of 1221 psig to the entrance of at least two steam getrators while allowing for:

(1) any possible s design worst case break of the main feedwater line; (2) pillage through the the design worst case single failure; and (3) recirculation flow.

This capacity is sufficient to ensure that adequete feedwater flow is available to remove decay heat and reduce R6 actor Coolant System temperature to itss than 350'F at which point the Residual Hoat Removal System may be placed in operation.

The test flow for the steam-driven auxiliary feedwater pump at a pressure of greatsr than or equal to 1450 psid ensures this capability.

The auxiliary feedwater flow path is a passive flow path based on tM fact that valve actuation is not required in order to supply flow to the steam generators. The automatic valves tested in the flow path are the Feedwater Split Flow Bypass which are required to be shut upon initiation of the Auxiliary Feedwater System to meet the requirements of the accident analysis.

Both steam supplies for the turbine-driven auxiliary feedwater pump must be OPERABLE in order to meet the design bases for the complete range of acciden+

analyses. The allowed outage tire for one inoperable steam source is consistent with theelower probability of the worst case steam or feedwater line break accident.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months with steam discharge to the atmosphere concurrent with total loss-of-offsite power or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at HOT STANDBY followed by a cooldown to 350*F at a rate of 50'F/hr for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

The contained water volume limit includes an alPwance for water not usable because of tank discharge line loca-tion or other physical characteristics.

The required indicated level includes a 3.5 percent measurement uncertainty, an unusable volua.e of 12,100 gallons and a required usable volane of 24 y gall yo O, CPSES Station Service Water System, which ca 900 NUREG-0737, Item II.E.1.1 requires a backup source to the CST which is the in lieu of CST minimum water volume.

C0KANCHE PEAK - UNIT 1 AND 2 B 3/4 7-2

Attachu nt to TKK 92318 Page 40 of $1 fROOF CONTAINMENT SYSTEM $

[ BEN BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that:

(1) the containment structure is prevented from exceeding its design negative pressure diffeiential of 5 psid with respect to the outside atmosphere, and (2) the containment peak pressure does not exceed the desig u

-re of 50 during a LOCA.

/

@e Theindicatedcontainmentprgssurevalles(fh.3p g an i

ig anQS,pfig, resp cQvely, g and to analytical limits o 0

. wit owance for measurement un tO Th maximum peak pressure expected to be obtained from a LOCA event is Y@sa[B.i, which is less than design pressure and is consistent with the l

etrana yses. This value includes the limit of 1.5 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis for a LOCA or steam line a

break a:cident. The average temperature shall be by an adjusted averaging of at least 2 of the measurements made at the listed locations, by fixed or portable instruments with allowance for temperature measurement uncertainty.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that th$ structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48.3 psig in the event of a LOCA.

A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this tJpability.

L/4.6.1.7 CONTAINMENT VENTILATION SYSTEH The 48-inch and 12-inch containment and hydrogen purge supply and exhaust g

isolation valves are required to be locked closed during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining these valves locked closed during plant operation ensures that excessive quantities of radioactive meterials will not be released via the Containment Ventilation System.

To provide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator.

The use of the Containment Ventilation System during operations is restricted to the 18-inch pressu e relief discharge isolation valves (with an effective diameter of 3 inches) since, these venting valves are capable of closing during a LOCA or steam line break accident. Therefore, the Exclusion Area dose guideline of 10 CFR 100 would not be exceeded in the event of an accident during containment venting operation.

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 6-2

TABLE B 3/4.4-Ib n

,e g

UNIT 2 REACTOR VESSEL FRACTURE TOUGl! NESS PROPERTIES 30 2

9-m 1

T,

~g c

E5 h

50 FT-L8 AVG. SHELF AVG. SHELF x

T 35 MIL RT ENERGY ENERGY gg l

NDT MOT Code Co Ni P

TEMP.

MWD (b)

MMWD(c) g COMPONENT GRADE WD.

'F

'F FT-LB FT-LB C

l

\\

131 1

U Closure Hd. Dome A5338, C1.1 R3811-1 0.15

.65 0.014

-40 60 0

Closure Hd. Torus A5338, C1.1 R3810-1 0.15

.69 0.011

-30 30

-30 143 157

.71 0.013 40

<100 40 p Cicsure Hd. Flange A508, C1.2 R3802-1 Vessel Flange A508. C1.2 R3801-1

.70 0.009

-10

<50

-10 121 o

.84 0.009

-10

<50

-10 138 m

Inlet Monle A508, C1.2 R3803-1 136 Inlet Nozzle A508, C1.2 R3803-2 0.10

.91 0.008

-20

<40

-20 146

.91 0.010

-10

<50

-10 Inlet Nozzle A508. C1.2 R3803-3

.86 0.009

-20

<40

-20 136 Inlet Nozzle A508. C1.2 R3803-4

.64 0.006 0

<60 0

132

= Outlet Nozzle A508, C1.2 R3805-1 l

.66 0.005 0

<60 0

119 Outlet Nozzle A508, C1.2 R3805-2 I

w l

1 Outlet Nozzle A508, C1.2 R3805-3 117

.66 0.004 0

0

.67 0.005 C

260-- < 60 0 119

. Outlet Nozzle A508, C1.2 R3805-4 76 E Nozzle Site 11 A5338, C1.1 R3806-1' O.05

.61 0.010

-10

- 40 Nozzle Shell A5338, C1.1 R3806-2 0.06

.62 0.009

-30

'70 10 87 86 Nozzle Shell A5338, C1.1 R3806-3 0.06

.70 0.007

-30 100 40 Inter. Shell A5338, C1.1 R3807-1 0.06

.64 0.006

-20

<40

-20 133 108 L

Inter. Shell A5338, C1.1 R3807-2 0.06

.64 0.007

-20 70 10 122 101 Inter. Shell A5338, C1.1 R3807-3 0.05

.60 0.007

-20 40

-20 120 105 Lower Shell A5338, CLI R3316-1 0.05

.59 0.001

-30 30

-30 13E 107 Lower Shell A5338, C1.1 R3816-2 0.03

.65 0.002

-30 60 0

131 Lower Shell A5338, C1.1 R3816-3 0.04

.63 0.008

-40 20

-40 139 Botton Hd. Torus A5338, C1.1 R3813-1 0.12

.65 0.009

-60 0

-60 l

Botton Hd. Dome A5338, C1.1 R3814-1 0.12

.66 0.009

-70

-10

-70 Weld Metal (a) 0.05

.03 0.004

-60 10

-60 (Inter. to Lower Shell Girth Seam) g Weld Metal (b) 0.07

.05 0.005

-50

10

-50 te (Inter. & Lower Shell Long Seams) g (a) 84 Weld Wire Ht. 09833 & Linde 124 Flux tot 110. 1061 (b) 84 Weld Wire Ht. 89833 & Linde 0091 Flux Lot No.1054 (c) Normal to asjor working direction (d) Major working ' direction

Attachnent to TXX 92318 Page 42 of 51 hh ELECTRICAL POWER SYSTEMS BASES ggg;g A.C. SOURCES. D.C. SOURCES, and ONSITE POWER DISTRIBUTION (fo5t Nueh * * ~

The Fuel Storage System consists of the fuel oil storage tank and is equivalent to the ANSI N195-1976 definition for supply tank.

@dieselgeneratorsareinaccordancewiththerecommendationsofR The Surveillance Requirement for demonstrating the OPERABILITY of the i

Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971; 1.108, " Periodic Testing Diei,el Generator Uni Used ts Onsite Electric Power Systems at Nuclear Power Plants " Revision August 1977; and 1.137, " Fuel-011 S January 1978, Generic Letter 64*15,ystems for Standby Diesel Generators, and Generic Letter 83-26, " Clarification of Surveillance Kequireraents for Diesel Fuel Impurity Level Tests."

The Diesel Generator Test schedule, Table 4.8-1, is based on the recommenda-tions of Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1. August 1977, and NRC Technical Report A-3230. " Evaluation of Diesel Unavailability and Risk Effective Surveillance Test Intervals," May-1986, and Generic Letter 84-15

" Proposed Staff Position to Improve and Maintaili Diesel Generator Reliability."

The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries are based on the recomendations of Regulatory Guide 1.129 "Meintenance Testing and Replacement of Large Lead Storage Batteries fp

~

Nuclear Power Plants," Revision 1, February 1978. Regulatory Guide 1. 2

" Criteria for Safsty Related Electric Power Systems for Nuclear Power ts,"

Revision 2, February 1977, and IEEE STD 450-1980, S i Recommended Practice j

for Maintenance, Testing, and Replacement of Large Lead Storage Battet'ies for Generating Stations and Substations."

The operational requirement to energize the instrumer:t busses from their associated inverters connected to its associated 0.C. bus is satisfied only when the inverter'n output is from the regulated portion of the inverter and not from the unregulated bypass source via the internal static switch.

Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal witage on float charge, connection resistance values, and the performance of battery service and discharge tests discharg,tshe effectiveness v the charging system, the ability to handle high ensures e rates, and compares the battery capacity at that time with the rated capacity.

Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity.

The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity.

The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer s full charge specificgravitywithanaveragespecificgravityofalltheconnectedcells not more than 0.010 below the manufacturer s full charge specific gravity, ensures the OPERABILITY and capability of the batttry.

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 8-2

Attachm:nt to TXX 92318 Paga 43 of 51 fROOF ELECTRICAL POWER SYSTEMS

$YS SY BASES A.C. SOURCES. 0.C. SOURCES. and ONSITE POWER DISTRIBUTION (Continued)

Operation with t, battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period:

(1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average, specific gravity of all the cells, r,ot more than 0.020 below the manufacturer s recommended full charge specific grisvity, ensures that the decrease in rating will be less than the safety margin provided in sizing: (3) the allowable value for an individual cell's specifit; gravity, ensures that an individual celt's specific gravity will act be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's floh'. voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

i 3/4.8.4 ELEC'fRICAL EQUIPMENT PPOTECTIVE DEVICES

{

Contair. ment electrical penetre(tons and penetration conductors are pro-tected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILIT'! of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance. This is based on the recommendations of Regulatory Guide 1.63, Revision 2, July 1978, " Electric Penetration Assemblies in C9ntainment Structures for Light-Water-Cooled

?

Nuclesr Power Plants."

i The surveillance Rer,uirements applicable to lower voltage circuit breakers

' W w e provide assurance of breaker reliability by testing at least 10% of each manufacturer's brand of circuit breaker.

Each manufacturer's molded case and metal case circuit 'arsakers are grouped into representative samples which are then tested on a rr,tating basis.to ensure that all breakers are tested.

If a wide variety exists within any manufacturer's brand of circuit breakers, it is necetsary to divide that manufacturer's breakers into groups and treat each

. coup as a separate type of breaker for surveillance purposes.

All Class 1E motor-operated valves' totor starters are provided with thennal overload protection which is oermanently bypassed and provides an alarm funr.t(on only at Comanche Peak Steam Electric Station.

Therefore, there are na OPERABILITY or Surveillance Requirements for these devices, since they will not prevent safety-related valves from performing their function (refer to Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1 March 1977).

COMANCHE PEAK - UNITS 1 AND 2 8 3/4 8-3 l

1 Attachmont to TXX-92318 Page 44 of 51

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DFSIGN FEATURES 5

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Attachment to TXX-92318 j

P; 45 of 51 4

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l SECTION 6.0 1

ADMINISTRATIVE CONTROLS 1

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Attach:3nt to TXX 9231B Page 46 of 51 fROOF TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION EWEW TWO UNITS WITH A COMMON CONTROL R00H POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION BOTH UNITS IN BOTH UNITS IN ONE UNIT IN H0DE 1,2,3, or 4 MODE 1,2,3 MODE 5 or 6 AND or 4 or DEFUELED ONE UNIT IN FDDE 5 or 6 or DEFUELED SS 1

1 1

SRO 1

None**

1 R0 3*

2*-

3*

A0 3*

3*

STA 1***

None 1***

SS Shift Supervisor with a Senior Operator license SR0 Individual with a Senior Operator license RO Individual with an Operator license 3

AD Auxiliary Operator STA Shift Technical Advisor The shift crew composition may be one less than the minimum requirements of Table 6.'2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crea members provided immediate action is taken to restore the shif t crew composition to within the minimum requirements of Table 6.2-1.

This provision does not permit uny shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

I During any absence of the Shift Supervisor from the control room while the unit is in H0DE 1, 2, 3, or 4, an individual with a valid Senior Operator license shall be,dtwignated to assume the control room command function. During any absence'of the Shift Supervisor from the control roon while the unit is in MODE 5 or 6, an ),utividus) with a valid Senior Operator license or Operator license shall be d uignated to assume the control roca command function.

At least one of the required individuals must be assigned to the designated

@**AtleastonelicenaedSeniorOperatororlicensedSeniorOperato1{

position for each unit.

ted to Fuel Handling must be present during CORE ALTERATIONS on eith<

t, who has no other concurrent responsibilities..

- - The STA position shall be manned in H00ES 1, 2, 3 and 4 unless the Shift Supervisor or the individual with a Senior Operator license wets the' qualifications described in Option 1 of the Commission Policy Statement on Engineering Expertise (50 FR 43521, October 28,1985).

COMANCHE PEAK - UNITS 1 AND 2 6-3 i

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Attachoent to TxX 9231B Page 47 of 51 fR00F ADMINISTRATIVE CONTROLS TECHNICAL REVIEW AND CONTROLS (Continued) department manager as previousl Nuclear Operations, in writing.y designated by the Vice Prtsident, Individuals responsible dure reviews shall be members af the Nuclear Operations Mfor proce-anagement Staff previously designated by the Vice President, Nuclear Operations.

Changes to procedures which do not change the intent of approved pro-cedures may be approved for implementation by two members of the Nuclear Operations Management Staff, at least one of whom holds a Senior Operator License, provided such approval is prior to implemen-tation and is docueented.

Such changes shall be approved by the original approval authority within 14 days of is.plementation:

b.

Proposed tests and experiaants which affect plant nuclear safety shall be prepared reviewed, and approved.

mentshallberevIewedbyaqualifiedindividual/groupotherthahEach such test or experi-the individual / group which prepared the proposed test or experiment.

Proposed test and experiments snall be approved before implementation by the Plant Manager.

Individuals responsible for conducting such reviews shall be members of the Nuclear Operations Management Staff previously designated by the Vice President, Nuclear Operations; Proposed chang 6s or modifications to plant nuclear safety-related es c.

structures, systems and components shall be reviewed as designated by the Chief Engineer.

Each such modification shall be reviewed by a qualified individual / group meeting the expnrience requirements of ANSI N18.1-1571, Section 4.6 other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modifications.

Individuals / groups responsibis for conductin previously designated by the Chief Engineer.g such reviews shall be Proposed modifications to plant nuclear safety-related structures, systems and components sha11 be approved by the Plant Manager prior to inglementation; d.

Individuals responsible for reviews performed in accordance with the requirements of Specifications 6.5.3. Aa and 6.5.3.lb, shall be

. members of the Nuclear Operations Management staff previously designated by the Vice President, Nuclear Operat 9ns.

Each such

/

review shall include a determination of whether or not additional cross-disciplinary review is necessary.

If deemed necessary, such review shall be done in accordance with the appropriate nualification requirements; Each review shall include a detemination of whether or not an e.

unreviewed safety question is involved.

For items involving unreviewed safety C1tions, NRC approval shall be obtained prior to the Plant Ma geg preval for implementation; and f.

The Security Plan Emergency Plan, and implementing procedures, shall be reviewed at least once per 12 months.

Reconnended changes to the implementing ptocedures shall be approved by the Vice President, COMNCHE PEAK - UNITS 1 AND 2 6-12

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Attach 2ent to TXX 92318 Pago 48 of 51 (Roof ADMINISTR_ATIVE CONTROLS

_ PROCEDURES AND PROGRAMS (Continued)

EWEW e.

Radioactive Effluent Controls Program (Continued) bo implemented by ing procedures, and (3) shall include remedial actions to be tak never the program limits are exceeded. The program shall inc1 the following elements:

i 1)

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)

Limitations on the concentrations of radioactive material 1

released in liquid effluents to UNRESTRICTED AREAS conforming i

to 10 CFR 20, Appendix 8 Table II, Column 2, 3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, 4)

Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR 50, 5)

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM ct least every 31 days, 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment c

conforming to Appendix ! to 10 CFR 50, 7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY i

conforming to the doses associated with 10 CFR 20, Appendix B.

Table II, Column 1 8)

Limitations on the annual and quarterly air doses res"Iting fron noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR 50, 9)

Limitations on the annual and quarterly doses to a MEMBER OF i

THE PUBLIC from lodine-131, Iodine-133, tritiutu, and all l

radionuclides in particulate form with half-lives greater than l

8 days in gaseous effluents released from each unit to areas l

beyond the SITE BOUNDARY conforming to Appendix ! to 10 CFR 50, and COMANCHE PEAK - UNITS 1 AND 2 6-16

Attachmont to VXX 92310 Page 49 of 51 (ROOF ADMINISTRATIVE CONTROLS 6

@ MONTHLY OPERATING REPOL.S (Contir.uod) sommijsion,'v shall be submitted on a mohthly bas to the U J f Nucleap Regulatot Tocum CoistroFDesk, Wp51ngtuh

.2055}(withasfpytotheR ional

$1f81orMf thhtRgglQng]Ef_ e of thfiNLjno later than the 1 th of each month following th> calendar conth covered by the report.

CORE OPERATING LIMITS REPORT 4

6.9.1.6a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any reraining part

]

of a reload cycle for the following:

1).

%derator temperature coefficient BOL and E0L limits and 300 ppm sur-veillance limit for Specification 3/4.1.1.3, j

2).

Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 3). Control Rod Insertion Limits for Specification 3/4.1.3.6, 4). AXIAL FLUX DIFFERENCE Limits and target band for Specifica-tion 3/4.2.1.,

RTP 5). Heat Flux Hot Channel Factor, K(Z), W(Z), and F for Specifica-tion 3/4.2.2, 9

6). Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.

6.9.1.6b The analytical methods used to determine the core opermting limits are for Units 1 and 2, unless otherwise stated, and shall be those previously approved by the NRC in:-

1). WCAP-9272 P-A, "WESTINGH00fE RELOAD SAFETY EVALUATION METHODOLOGY,"

July 1985(yProprietary).

(Methodology for Specifications 3.1.1.3 -

i Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux

^ Difference, 3.2.2. - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

2).

WCAo-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES -

TOPICAL REPORT." September 1974 (y Proprietary).

(Methodology for Specifi-cation 3.2.1. - Axial Flux Diffiirence, [ Constant Axial Offset Control).)

3).

T. M. Anderson to K. Kniel (Cnist of Core Pe f nee Branch, NRC January 31, 1980--

Attachment:

Operation an Sqfet Analysis Aspects of an Improved Load Follow Package.

(Methodq) g 4 fo Specification 3.2.1 -

Axial Flux Difference (Constant Axial Offtatyr 1).)

4). NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Sec-l tion 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, i

Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.

l (Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Ln. trol).)

COMANCHE PEAK - UNITS 1-tND 2 6-20 l

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Attach 2ont to TXX 92318 Paga 50 of 51 i

ROOF ADMINISTRATIVE CONTROLS RECOR0 RETENTION (Continued)

Records of transient or operational cycles for these unit components e.

i identified in Table 5.7-1; f.

Records of reactor tests and experiments; Records of training and qualification for current members of the g.

unit staff; h.

Specifications; Records of inservice inspections performed pursuant to th i.

Records of qualit Assurance Manual y assurance activities required by the Quality j.

Records of reviews performed for changes made to procedurer or 4

j equipment or reviews of tests and experiments pursuant to 10 CFR 50 k.

Records of meetings of the 50RC and the ORC; 1.

required by the Technical Requirements Manual inc E

i maintenance records;which the service life commences and associate t

v^~

Records of secondary water sampling and water quality; and m.

Records of analyses required by the Radiological Envirornental n.

Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified tinies and QA records showing that these procedures were followed, L CALCULATION MANUAL and the PkOCESS CONTROL o.

5,11 RADIAT{0hPROTECTIONPROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained l

adhered to for all operations involving personnel radiation exposure, and 5.12 _HIGH RADIATION AREA 6.12.1 device" or " alarm signal" required by paragr(ap)h 10 CFR 2 radiation area, as defined in 10 CFR 20, in which the intensity of radiation

, each high is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radia+. ion penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be

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l COMANCHE PEAK - UNITS 1 AND 2

(-23

i Attach ent to TXX 92318 Pege 51 of 51 i

ADMINISTRAT!YE CONTROLS EbFW 6.13 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

Shall be documented and records of reviews performed shall be a.

retained ss required by Specification 6.10.30.

This documentation shall contain:

1)

Sufficient information to suppor t chan e together with the appropriate analyses or avaluati n5j tify ng the change (s) and 3

2)

A determination that the change will maintain the overall conftrmance of the solidified waste product to existing requirements of Federal, S' ate, or other applicable regulations, b.

Shall become effective after review and acceptance by the 50RC and the approval of the Vice President, Nuclear Operations.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCH)

Changes to the ODCH.

Shall be documented and records of reviews performed sr.all be i

s.

retained as re shall contain: quired by Specification 6.10.30.

This documentation 1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)

A determination that the change will maintain the level of radioactive affluent control required by 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and Apperdix I to 10 CFR 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b..r Shall become effective after review and ecce.nt ece by the SORC and the approval of the Vice President, Nucle-v Operations.

2 Shall be submitted to the Commission in the form of a complete, c.

legible copy of the entire ODCH as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the 00CH was made. Each change shall be identified by markings in the sargin of the affected pages, clearly indicating (e.g. area of the page that wat changed, and shall the indicate the date

, month / year) the change was implemented.

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COMANCHE PEAK - UNITS 1 AND 2 6-25 i

ML20102A476

Text

Attachment:

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