ML20083M864

Proposed Tech Specs,Removing Several cycle-specific Parameter Limits

ML20083M864
Person / Time
Site:	Comanche Peak
Issue date:	05/12/1995
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20083M855	List: ML20083M853 ML20083M864
References
NUDOCS 9505220100
Download: ML20083M864 (29)

Text

i c t

AFFECTED TECHNICAL SPECIFICATION PAGES (NUREG 1468)

(Pages iii. 2 1, 2 2, 2 3, 2 5, 2 6, 2 9, 2 10, 2 11 B 2 1, B 2 5, 3/4 1 1, 3/4 1 3, 3/4 1 8. 3/4 1 10, 3/4 1 13, 3/4 2 12 B 3/4 1 1, B 3/4 1 2, B 3/4 1 3, 6 20, 6 21, 6 21a) i i

1

- 9505220100 950512 PDR ADOCK 05000445 .

P. PDR s 1

. i Attachment 3 to TXX 95076 Page 1 of 23 '

M SAFETY LIMITS M LIMITIM SAFETY SYSTEM SETTIMS EfalllE .

EAE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 _O

-FIC"hE 2.1-16 Uni-T I "EACT0" CORC 3A ETrtfMIT. . . . . . . . . . . . . . . . .C

--fl80RE-t-i-itHMIY-f RCACTOR CORE-SAFETY-1.!MIT. . . . . . . . . . . . . . . . . 2-3

~ ?-24 ]

. . . O- s '

j '

2.2 LIMITIM SAFETY SYSTEM SETTIES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS...............2-/ z. T TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.. 2-/3 BASES -

SECTION gg 2.1 i SAFETY LIMITS -

2.1.1 REACTOR C0RE................................................. 8 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 8 2-2 2.2 LIMITIM SAFETY SYSTEM SETTIMS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... B 2-3 1

1 Com MCHE PEAK - UNITS 1 AND 2 111 l Unit 1 - Amendnent No.14 '

Attach:ent 3 to TXX 95076

.

• Pag) 2 of 23 2.0 SAFETY LIMITS Als LIMITING 1AFETY SYSTEM SETTimet 2.1 1AFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T ,) shall not exceed the lietts shown in

-71; = 2rk-4 R - +h e. Cee. Ope raw y Lo m . 4 R e e ce t. ( C ot *Q .

APPLICABILITY: MODES 1 and 2.

E.Ilm:

Whenever the point defined by the combination of the highest operating loop average temperati.re and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in H0T STAND 8Y within I hour, and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant Systes pressure shall not exceed 2735 psig.

)

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

l MODES 1 and 2:

l Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STAN08Y with the Reactor Coolant System pressure within its limit I within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within .

5 minutes, and comply with the requirements of Specification 6.7.1. I i

COMANCHE PEAK - UNITS 1 AND 2 2-1

ag o 2

~[ s( (~ N s

.70

. 1s '

N se0 %_^  % UWhABLE /

s m .. . === ,

650  % ,

640 s%  %

N 630 s, N_ \

3

% d 1985 f W.

e-  %

x n g'

,g 610 g 6-5" ACC8 m 8tf

/ ~D. '

580 N

7 ,

570 560 550 j

/ , , , ,, , ,

0 40 60 80 100 120 PERCENT OF RATED THERMAL POWER .

Y FIGURE 2.1-la C/ UNIT 1 REACTOR CORE SAFETY LIMITS COMANCHE PEAK - UNITS 1 AND 2 2-2 Unit, i - Amendment No. 44,21 Unit 2 - Amendment No. 7

Attachment 3-to TXX 95076

. Page 4 of 23 670 .

~

,. ................6. . ........

( '

i P=23ss Psc UNACCE blE OPE TON

)

650 -m - - - -- - - l .-

640 .

P=2235 PSIG ,

630 - - - - d -

l  :

i

P=i. i

.e  : .

• 620 -- -

. 610  ; ,

pgi ,p,, j 3

M .  :  :

g  :  : .  :  :

600 -

l l l l  :

590 -  :  ;

l  : . .

580  : .

i ACCEPTABLE i. i.

i OPERATON l i i

570 -

l l 560 .

52 20 40 60 80 100 12 PERCENT OF RATED THERMAL POWER k FIGURE 2.1 lb UNIT 2 REACTOR CORE SAFETY LIMITS I

COMANCHE PEAK - UNITS 1 AND 2 23 Unit 1 - Amendment No. 14 f

U

' mw W 5[ .

taste 2.2-1 Uw REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR d

+

FUNCTIONAL UNIT ALLONANCE ERROR 5 (TA) Z (S) TRIP SETPOINT 8

. 1. Manual Reactor Trip N.A.

Att0NABLE VALUE ,

y N.A. M.A. N.A. N.A

2. Power Range, Neutron Flux

a. Nigh Setpoint 7.5 4.56 1.25 sIO95 of RTP* s111.7% of RTP*

b. Low Setpoint 8.3 4.56 1.25 s25% of RTP*

3. 527.7 of RTP*

Power Range, Neutron Flux, 1.6 0.5 Nigh Positive Rate 0 55% of RTP* with 56.3% of RTP* with a time constant a time constant 22 seconds 22 seconds

4. Not Used

5. Intermediate Range, I i

17.0 8.41 0 Neutron Flux $25% of RTP* $31.5 of RTP*

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps 7.

$1.4 x 10% cps Overtemperature N-16 #* * #*

ra. ** *#

It I~ l 53 70

b. Unit g.,10+ See te 1 See te 2 ~ 4 10.0 6.7 1.0+ l%38+

7 -

See No 1 See Not 2 s

i

• RTP = RATED THERNAL POWER

!( M .0% span Q M-16 g r monitor,1 10% for T,,, RTDs and J2) 1. forhssurize ressure T ++ Aspan forw16 power nitor, 1. for T,,,, s and O. 76% for pre gurizer p essure s sors.

A s s pec..p,eci rs.

f% C ,e oge, COMANCHE PEAK - UNIT I AND 2 f, 3. L,m /3 Re ,,od, 2-5 Unit 1 - Amendment No. 2 N,2139 Unit 2 Amendment No. 7. 25

Ek

'e C ' -

I me g TABLE 2.2-1 (Continued) -

! REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS Uw E

TOTAL SENSOR A ALLOWANCE ERROR d N FUNCTIONAL UNIT fTA) Z fSi TRIP SETPOINT ALLOMABLE VALUE $

. =

l C

5

8. Overpower N-16 4.0 2.05 1.0+0.05* sil2% of RTP* s114.55 of RTP*- 3i l d 9. Pressurizer Pressure-Low

- a. Unit 1 4.4 0.71 2.0 21880 psig 21863.6 psig g b. Unit 2 4.4 1.12 2.0 21880 psig 21863.6 psig ro 10. Pressurizer Pressure-High

a. Unit 1 7.5 5.01 1.0 s2385 psig s2400.8 psig

b. Unit 2 7.5 1.12 2.0 s2385 psig s2401.4 psig

~

i

~

• 11. Pressurizer Water Level-High

a. Unit 1 8.0 2.18 2.0 $92% of instrument s93.95 of instrument span span

b. Unit 2 8.0 2.35 2.0 s92% of instrument 593.95 of instrument span span CC 11 12. Reactor Coolant Flow-Low

1. "_ a. Unit 1 2.5 1.18 0.6 2905 of loop 288.6% of loop

• Wstflow

• a s flowa *

b. Unit 2 2.5 1.25 0.87 290K of 1 . of 1 kk

==

de flow

• asu (ain1 mum flow *C surejd I kI

= s

"" (3) 1.05 span for N-16 power monitor and 0.05% for T,,,, Ms.

FF *** RTP - RATED THERNAL POWER d (

• g }-

Loop Wflow e; 5,:;T-as secc#,<4 m A e_ Ocem % L m u R*-'*

Co ,e.

g

] Q m::; =:ar- ==;.rd ::;r.; - n,=TQ

-6 >.

E. .C. .

4-O4

" TABLE 2.2-1 (Continued) 'A f

n

~- _. ' ' ~ -

TABLE NOTATIONS Uw g

5 NOTE 1: Overtemperature N-16 ~ y v 1+rs U g N -

K,-K2[1+rs i

T,-T,*] + K3 (P-P') - f, (aq) a n a g 5 i E

Q Where: N - Measured N-16 Power by ion chambers, \

~

T, - Cold leg temperature *F, R

T; - 560.5*F for Unit 1, 560.3*F for un 2 - Reference T, at RATED THERMAL POWER, m

K, - 1.150, K2 -

0.0134/*F for Unit 0.016856/*F for t2 m

1114 - The func generated by the lead-lag controller for I+r5z T, d ic compensation, e

r,, r2 - ime constants utilized in the lead-lag controller for

=

T,, r, ;t 10 s, and r, s 3 s, 0.000719/psig for Unit 1

[

0.000898/psig for Unit 2 /

l cc

=s w MM /

/

mw ee gg - h 33 a3 h.*'

\./0 ,

tt aa EI l

~ -

W

. - - = _ . _ ..

d EW

• E.

" *i l TABLE 2.2-1 (Continued)

~

}

n TABLE NOTATIONS.(Continued) 1 -

~

TE 1: -

j h

(Continued) \w x 7 P - Pressurizer pressure, psig, g E P' 2 2235 psig (Nominal RCS operating pressure), ,

5 8

[ S - Laplace transform operator, s",

$ and f,(aq) is a function of the indicated differepce between top and bottom halves of

, detectors of the power-range neutron ion chanbors; with gains to be selected based on measured instrument response during plant TUP tests such that:

For Unit 1 (1) for q -

RATED,THE between 5 and +4%, f (aq) - 0, where q, and g are percent.

in the top and, bottom halves of the core respectively, l'

and q, + g i

otal THENIAL POWER in percent of RATED THElWtu. POWER,

! o (11) for eac percent that the magnitude of q, - q exceeds -65%, the N-16 Trip

, Set at shall be automatically reduced by 1.,81% of its value at RATED T POWER, and cc (111 for each percent that the magnitude of q, - g exceeds +4%, the N-16 Trip 11 Setpoint shall be automatically reduced by 2.265 of its value at RATED

[ . .

THElutAL POWER.

gg De lA E. E .

Of .

N

.E.

.g O'.

e, n TABLE 2.2-1 (Continued) o2

$ TABLE NOTATIONS (Continued)

]R g f ~

g NOTE 1: (Continued)  %. g y, For Unit 2 3

3 (1) for q - A between -52% and +5.5%, f,(Aq) = 0, whe e

n c- RATED,THEIRAL POWER in the top and bottom hal

, and g are percent g of the core respectively, , y i

3 ui and g + g is total THEIMAL POWER in pe of RATED THEIMAL POWER, (11) for each percent that the magntJude of q, - g exceeds -525, the N-16 Trip g Setpoint shall be automatic Try reduced by 2.15% of its value at RATED THEIMAL }

o POWER, and N

(iii) for each percen at the magnitude of exceeds +5.5%, the M-16 Trip Setpoint sh be automatically reduced - q,7%

2.1 of its value at RATED THEIMAL ,

POWER. ,

m NOTE 2: The channel' i ximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.51% l'

~ of sp r Unit 1 or 2.88% of span for L%1t 2.

~

T)eik ,

)

==

8 8 an 2z D ~

~

JAttachment 3 to TXX 95076

,. Pag,e 10 of 23 i

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE ,

The restrictions of this Safety 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. DN8 is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNS. This relation has been developed to predict the DN8 heat flux and the location of DN8 for axially uniform and non-unifom heat flux distri-butions. The local heat flux ratio (DN8R), defined as the ratio of the heat flux that would cause DN8 at a particular core location to the local heat flux, is indicative of the margin to DNS.

The DN8 design basis is that the minimum DN8R of the limiting rod during Con-dition I and II events is greater than or equal to the DNSR init of the DNB correlation being used. The correlation DN8R limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence level that DNS will not occur when the minimum DN8R is at the DNBR limit. In meeting this design basis, uncertain-ties in plant operating parameters are considered such that the minimum DNBR for the limiting rod is greater than or equal to the DN8R limit. In addition, margin has been maintained in the design by meeting safety analysis DN8R limits in performing safety analyses.

ww Cae. 5,5 e.% w m u +s _

Th ecurves d r; ! Y show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DN8R is no less than the safety analysis limit value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

1 l

l l

COMANCHE PEAK - UNITS 1 AND 2 8 2-1 Unit 1 - Amendment No. 14

' AttachCent 3 to TXX 95076

'Pago 11 ol' 23 l

LIMITING SAFETY SYSTEM SETTINGS I

BASES l Intermediate and Source Rance. Neutron Flux  :

1 The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condi-4 tion. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. In addition, the Source Range Neutron Flux trip provides similar protection during shutdown operations with the reactor trip breakers closed and the rod control system capable of control rod with-drgwal. The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overta=nerature N-16 The Overtemperature N-16 trip provides core protection to prevent DNB for '

all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to piping transit delays from the core to the N-16 detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The setpoint is auto-matically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the cold leg temperature detec-tors, (2) pressurizer pressure, and (3) axial power distribution. With nomal axial power distribution, this Reactor trip limit is always below the eere-Safety Li::it :: :te., ir, f tpr; 2.1.. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detgs, the Reactor trip is automaticallgraducp.:j;:p t;)hef.;t: tier.#

Q Qe c care Sa Ve.tg Mh c.arve s.

Overnower N-16 -

The Overpower N-16 trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, l'aits the required range for Overtemperature trip, and provides a backup to the High Neutron Flux trip. The Overpower N-16 trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

Pressurizar Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pres-sure trip thus limiting the pressure range in which reactor operation is per-i mitted. The Low Setpoint trip protects against low pressure which could lead l l to DN8 by tripping the reactor in the event of a loss of reactor coolant  ;

pressure.

j COMANCHE PEAK - UNITS 1 AND 2 8 2-5

Nttach=nt 3 to TXX 95076

-

• Page 12 of 23 3/4.1 REACTIVITY CONTROL SYSTEMS

- 3/4.1.1 BORATION CONTROL f

SHUTDOWN "*?clu - T GREATER THAN 200'F LIMITING CONDITION FOR OPERATION J

3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% .k/k for't.

Unit-+-HT35 skfk fer Wi@ the. va /u.c. spe.c. f,e d i n +4 e C e re.

APPLICABILITY: MODES 1, 2*, 3, and 4. Y '~ "1 ACTION: s e fed ' n + A e. C ctA y With the SHUTDOWN MARGIN less than 1.6% akik Ter Uni 4-1 (1.3 ek/k for Unit 2) A _

imediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7,000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored. I SURVEILLANCE REQUIRENENTS 4.1.1.1.1 The SHUTDOWN PARGIN shall be determined to be greater than or equal to 1:6Fak/k fer UnM-1-Hv3Fak/k f r 5444M-va lu e. .s o c e . g. < d. sn AL c o uA. :

r k e-

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.

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

b. When in MODE 1 or MODE 2 with K greater than or equal to 1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying $at control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with K.,,less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical control red position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 55 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 i

l l

• See Special Test Exceptions Specification 3.10.1.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-1 Unit 1 - Amendment No. 14

'kt2achment3toTXX95076

• Page 13 of 23 REACTIVITY CONTROL SYSTEMS SHUTDOWN Mac4IN - T LESS THAN OR E00AL TO 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% .k/k.'E'-

+ k e. va I a e- Spec.,hul em %< Ce<e Gemky APPLICABILITY: MODE 5. c , m , +.s R e ,,o r + cc ou<),

ACTION: e, qjq e_ sp, e,g,< j , m jze cogg, With the SHUTDOWN MARGIN less than h 3% .k/hFiumediately initiate and continue boration at greater than or equal to 30 gpa of a solution containing greater than or equal to 7,000 ppa boron or equivalent until the required SHUTD0WN i MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be detereined to be greater than or equal i to 1r35-*k/kte a< g n /x 3,o e c, f,,s , n #4e c o c, g f

a. Within I hour after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.

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

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature, I

4) Fuel burnup based on gross thermal energy generation, I

5) Xenon concentration, and

6) Samarium concentration.

l l

COMANCHE FEAK - UNITS 1 AND 2 3/4 1-3 Unit 1 - Amendment No. 5

. Attach 0ent 3 to TXX 95076

,PagG ,14 of 23 l l

REACTIVITY CONTROL SYSTEMS FLOW PATHS - GPERATING LIMITING C0fmITION 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 from the refueling water storage tank via centrifugal charging pumps to the RCS.

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

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERA 8LE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y and borated ,

to a SHUTDOWN MARGIN equivalent to at least -i 3% .k/if at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow pathsgto OPERA 8LE status within -

the next 7 days or be in COLD SHUTDOWN withinithe next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the flow path from the boric acid storage tanks is greater than or equal 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, I or otherwise secured in position, is in its correct position; and

c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at east 30 gpa to the RCS. i

• A maximum of two charging pumps shall be OPERA 8LE whenever the temperature of I one or more of the RCS cold legs is less than or equal to 350*F except when  !

I Specification 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.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-8 Unit 1 - Amendment No. 5 l

l

- - - - - - - - _ _ - -_ _1

Attach ent 3 to TXX 95076 Page 15 of 23 REACTIVITY CONTROL SYSTEMS Qi&M ING Pt N S - OPERATING LIMITING coleITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE.

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

ACTIflN:

With only one charging pump OPERA 8LE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least H0T STAN08Y and borated to a SHUTDOWN MARGIN equivalent to at least 1.3". A/ Mat 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps \to OPERA 8LE status within the next 7daysorbeinCOLDSHUTDOWNwithinthenext)30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. _

SURVEILLANCE REQUIREMENTS l

4.1.2.4.1 The required centrifugal charging pump (s) shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5. -

4.1.2.4.2 The required positive displacement charging pump shall be ~

demonstrated OPERA 8LE by testing pursuant to Specification 4.1.2.2.c.

4.1.2.4.3 Whenever the temperature of one or more of the Reactor Coolant System (RCS shal) cold legs is less than or equal to 350*F, a maximum of two charging pumps l be OPERA 8LE, except when Specification 3.4.8.3 is not applicable.

When required, one charging pump shall be demonstrated inoperable' at least once per 31 days by verifying that the motor circuit breakers are secured in the open position.

i

• The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry i

i into MODES 3 and 4 for the charging pump declared inoperable pursuant to Specification 3.1.2.4 provided the charging pump is restored to OPERA 8LE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

3 "In MODE 4 the positive displacement pump may be used in lieu of cne of the i required centrifugal charging pumps.

8 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

! power removed from the valve operator (s) or by a manual isolation va(s) with lve(s) secured in the closed position.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-10 Unit 1 - Amendment No. 5

. Attachment 3 to TXX 95076

,. Pagg 16 of 23 REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - OPERATING 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 ppa, 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 2400 ppm and 2600 ppm,

3) A minimum. solution temperature of 40*F, and

4) A maximum solution temperature of 120*F.

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

ACTION:

a. With the boric acid storage tank inoperable and being used as one of the above required borated water sources, restore the tank to  !

OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at i least 1. % a.k/* at 200*F; restore the boric acid storage tank to OPERABLE status ithin the next 7 days or be in COLD SHUTDOWN within l the next 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />

+h e w is e s ee.ce fs ect m th e. Co M

b. With the RWST inoperable, restore the tank to OPERABLE status j

within I hour or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> .

and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l l

l l

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-13 Unit 1 - Amendment No. 5rM,26 Unit 2 - Amendment No. 5,12

.

• kttachrent 3 to TXX 95076

, Pqge 17 of 23 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DN8-related parameters shall be maintained within the stated limits:

a. Indicated Reactor Coolant System T ,s 592*F ,

l

b. Indicated Pressurizer Pressure 2 2219 psig*  !

++ \

c. Indicated Reactor Coolant System (RCS) Flow 03,400 gpm** for Unit 1 l 395,200 gpm** for Unit 2 l

APPLICABILITY: MODE 1. +he. a lu e spec' 4'e'l '" I ACTION:

Co<e. Om d*9 l#'U 4.e,ocr +- (Cot- O l

l I

With any of the above parameters exceeding its limit, restore the parameter to l within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of 1 RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the above parameters shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l 4.2.5.2 The RCS total flow rate shall be verified to be within its limits at least once per 31 days by plant computer indication or measurement of the RCS elbow tap differential pressure transmitters' output voltage.

4.2.5.3 The RCS loop flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The channels shall be normalized based on the RCS flow rate determination of Surveillance Requirement 4.2.5.4. ,

4.2.5.4 The RCS total flow rate shall be determined by precision heat balance measurement after each fuel loading and prior to operation above 85% of RATED l THERMAL POWER. The feedwater pressure and temperature, the main steam pres- l sure, and feedwater flow differential pressure instruments shall be calibrated i within 90 days of performing the calorimetric flow measurement. '

• Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

• • Includes a 1.8% flow measurement uncertainty.

COMANCHE PEAK - UNITS 1 AND 2 3/4 2-12 Unit 1 - Amendment No. +4 744,30 Unit 2 - Amendment No. 7,16

Attachment 3 to TXX 95076 j Page 18 of 23 ,

l 3/4.] Rf 4CTIVITY CONTROL SYSTDt3 1

BASES 3/4.1.1 RORATION CONTROL 4

3/4.1.1.1 and 3/4.1.1.2 SWTDOW MARefN 4

A sufficient SWTDOWN MARGIN ensures that: (1) the reactor can be made suberttical from all operating conditions (2) the reactivity transients asso-cisted with postulated accident conditions are controllable within acceptable limits, and (3) the reacter will be maintained sufficiently subcritical to

preclude inadvertent criticality in the shutdeun conditten.

SHUTDOWN MARGIN requirements vary throughout core life as a function of and RC5 T The most restrictive fuel depletion ccedition occurs RCS at E0L, boren concentration, with T at ne lead operaU n.

g temperature and is asso-

/V ciated with a postulated steam Une break accident and resultine un, controlled 05 cooldown. In the analysis of this accident, a miniaun SHUT 60W MARSIN ef-

.it stc/ir-for Wit ! '!T35 sk/t fer hit @is required to centrol the reactiv-jb 7" [# g tty trtnsients Accordingly, the SHUTDOWN MAR $1N requirement is based upon this ,

Y# limting concition and is consistent with FSAR safety analysis assumptions. l With T le SHUTDOW MAR $1N cf--1.0". .i ^i ,,Md:: ;d:gets '2-l "*7" ~y n bo e- ^M.

get::',h =ss than 200*F,da-is based on the resulin of the beren dikution acc .

g e e.pr o r g l

Since the actual overall core reactivity balance comparisen required by l 4.1.1.1.2 cannot be performed until after criticality is attained, this compari-scr. is n

t reqaired (and the provisions of Specification 4.0.4 are not appli- l cable) for entry into any Operational Mode within the first 31 EFPO following initial fuel load or refueling.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIEMI  !

The limitations on moderator temperature coefficient ensure that the value of this coefficient remains within the(MTC)lting 11a condition assumed in the F5AR accident and transient analyses.

The MTC values of this s plant conditions; accordingly,pecification verification ofare NTCapplicable to a specific values at conditions other set of than those explicitly stated will require estrapelation to these conditions in

order to permit an accurate comparison.

The mest negative NTC value equivalent to the most positive moderator density coefficient CMC) was obtained by incrementally correcting the MC used in the FSAA ana'yses to nominal operating condittens. These corrections l

l

COMANCHE PEAK - UNITS 1 A M 2 8 3/4 1-1 Unit 1 - Amendment No. 4, 18i

ettachment 3 to TXX 95076 Tage 19 of 23

,o . .

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting End of Cycle Life (E0L) MTC value. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppe equilibrium boron concentration and is obtained by making these corrections to the limiting EOL MTC value.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERAELE status with a steam bubble, and (4) the reactor vessel is above its minimum RT , temperature.

3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergngyM power supply from OPERABLE diesel generators. ( Ae. cegmed j With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoper The boration capability of either flow path is sufficient to provide %able. %HUTDOWN MARGIN from expected operating conditions of ' " " * '-- "-" ' (1.3% tA/lcf.

)-foc_ljn4t4) after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 15,700 gallons of 7000 ppm borated water from

the boric acid storage tanks or 70,702 gallons of 2400 ppm borated water from l the refueling water storage tank (RWST). j COMANCHE PEAK - UNITS 1 AND 2 8 3/4 1-2 Unit 1 - Amendment No. 5, M ,10,26 i Unit 2 - Amendment No.,6,12 <

4

. Attachment 3 to TXX 95076

,, Pagg 20 of 23 I

REACTIVITY CONTROL SYSTEMS BASES ,

BORATION SYSTEMS (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 Baron Injection System becomes inoperable.

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 sources are sufficient to prevent boric acid crystallization with the highest allowable boron concentration.

The boron capability required below 200*F is sufficient to provide e< d SHUTOOWN MARGIN ef '.3% e/Faf ter xenon decay and cooldown from 200*F to 140*F. This condition requires either 1,100 gallons of 7000 ppe borated water from the boric acid storage tanks or 7,113 gallons of 2400 ppe 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 instrument error and tank tolerances, as applicable), margin, and other required volume.

Ind. Unusable Required Measurement Tank N00ES Level Volume Volume Uncertainty Margin Other (gal) (gal) (gal)' (gal)

RWST 5,6 24% 98,900 7,113 4% of span 10,293 N/A 1,2,3,4 95% 45,494 70,702 4% of span N/A 357,535*

Boric 5,6 10% 3,221 1,100 65 of span N/A N/A Acid 5,6 205 3,221 1,100 65 of span 3,679 N/A Storage (gravity feed)

Tank 1,2,3,4 50% 3,221 15,700 6% of span N/A N/A The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in M00E 6.

I I

• Additional volume required to meet Specification 3.5.4.

COMANCHE PEAX - UNITS 1 AND 2 8 3/4 1-3 Unit 1 - Amendment No. 5d4,26 Unit 2 - Amendment No. 6,12

' Attachment 3 to TXX.95076 .

'Page 21 of 23 ADMINISTRATIVE CONTROLS

'n 1 9

i)MONTHLYOPERATINGREPORTS(Continuedi

' & $ j shall be .ubmitted on a monthly basis to the U.S. Nuclear Regulatory n E = Comunission, Document Control Desk, Washington, D.C. 20555, with a copy to the r# 4 E '\ Regional Administrator of the Regional Office of the NRC, no later than the

-5 2 3 n~ E 15th of each month following the calendar month covered by the report (Q{ "(. $ {0RE \ OPERATING LIMITS REPORT 0

k'3 6.9.1.6a Core operatin limits shall be established and documented in the n} $ k'COHEOPERATINGLIMITSRfPORT(COLR)beforeeachreloadcycleoranyremaining i

/

Ia i; , part of a reload cycle for the following:

j  ! a

\g@y, 3

$ 1). Moderator temperature coefficient 80L and E0L limits and 300 ppm sur-

. E veillance limit for Specification 3/4.1.1,3, f 'f , j; $ 2). Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,

'f.4

)g,2$ Lf* 4 d3). i }j Control Rod Insertion Limits for Specification 3/4.1.3.6, 4). AXIAL FLUX DIFFERENCE Limits and target band for Specification 0g,2*

a )= 3/4.2.1.,

$(3}

s. *O 5). Heat Flux Hot Channel Factor, K(Z), W(Z), F,*, and the F.*(Z) ggl$jjj) allowances for Specification 3/4.2.1,

j&j(4 x 6). Nuclear Enthalpy Rise Hot Char.nel Factor Limit and the Power Factor ea g ltiplier for Specification 3/4.2.3.

~#

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

\g9j l

/ , ~, ! $ f.

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

July 1985 (M Proprietary). (Methodology for Specifications 3.1.1.3 -

a) Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion

\

4(4' t

4 N, Limit 3.1.3.6 - Control Bank Inserti(.. Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot h annel Factor, 3.2.3 - Nuclear t

fIi :{I /( fnthalpy Rise Hot Ct:annel Factor [JQ 2). WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES -

i E O i 'N TOPICAL REPORT," September 1974 (W Proprietary). (Methodology for

/6 i Specification 3.2.1 - Axial Flux Difference F onstant Axial Offset

( 4 ,(,. t:y .

Control).)

s g

=

} 3). T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC)

(d:d w j ;9j $I. 7 \ January 31, 1980--

Attachment:

Operation and Safety Analysis s of an Improved Load Follow Package. (Methodology for Specification ,

3.2.1 - Axtil Flux Difference [ Constant Axial Offset Control).)

t ti@- 4). NUREG-0800. Standard Review Plan, U.S. Nuclear Regulatory Commission, 1 Section 4.3, Nuclear Design, July 1981. 8 ranch Technical Position g

qj]e]vOC$j t

(P8 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. (Methodology for Specification 3.2.1 - Axial Flux

' =5gg) Difference [ Constant Axial O'fset Control).)

- hv N w 4 =21 /

Q N ) COMANCHE PEAK - UNITS 1 ANO 2 6-20 Unit 1 - Amendment No GrH ,34 ,

Unit 2 - Amendment No. 20 ,

4

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

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

~. Attach ent 3 to TXX 95076

,;. Rage,22 of 23.

ADMINISTRATIVE Coi,'TROLS CORE OPERATING LIMITS REPORT (Continued) i2 / 5). WCAP-10216-P-A, Revision IA, " RELAXATION OF CONSTANT AXIAL 0FFSET I N CONTROL F. SURVEILLANCE TECHNICAL SPECIFICATION," February 1994 (W l- I Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot

[ , 's Channel Factor (W(z) surveillance requirements for F. Methodology).)

4 I 6). WCAP-10079-P-A, "NOTRUMP, A N00AL TRANSIENT SMALL BREAK AND GENERAL r4 d)l NETWORK CODE," August 1985, (W Proprietary).

7). WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING

},* THE NOTRUMP CODE", August 1985, W Proprietary).

N f') (*l (

,tm s

8). WCAP-11145-P-A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP CODE", October 1986, M Proprietary).

(kj5I 9). RXE-90-006-P, " Power Distribution Control Analysis and overtemperature t

O ,y ) N-16 and Overpower N-16 Trip Setpoint Methodology," February 1991.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 -

g 4;t Heat Flux Hot Channel Factor @ y

(\ .; 3k 10). RXE-88-102-P, "TUE-1 Departure from Nucleate Boiling Correlation",

January 1989.

f n v. .) i

\

~dV 11). RXE-88-102-P, Sup. 1. "TUE-1 DN8 Correlation - Supplement 1",

December 1990.

12). RXE-89-002, "VIPRE-01 Core Themal-Hyttraulic Analysis Methods for Comanche Peak' Steam Electric Station Licensing Applications", June 1989.

)

13). RXE-91-001, " Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", February 1991.

w -

\ " %. 5

\ k j) 14). RXE-91-002, " Reactivity Annealy Events Methodology", May 1991.

(Methodology for Specifice!,1* 3.1.1.3 - Moderator Temperature c j= j, Coefficient, 3.1.3.5 - Shu Myci Bank Insertion Limit. 3.1.3.6 -

, Control Bank Insertion Limm, 3.2.1 - Axial Flux Difference, 3.2.2 -

3 Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot

\ g .;3 ')

l ,j Channel Factor.)

5'/

, t's 15). RXE-90-007, "Large Break Loss of Coolant Accident Analysis (JJ=

Methodology", December 1990.

,E' 16). TXX-88306, " Steam Generator Tube Rupture Analysis", March 15, 1988.

m h )\

ii .* s 17). RXE-91-005, " Methodology for Reactor Core Response to Steamline Break

J Events," May, 1991.

f%. M A

((Methodologyfor Specifications 3/4.1.1.1, 3/4.1.1.2,

\ 3/4.l.2.2,3/4.1.2.4 and 3/4.1.2.6 Shutdown 6fargin.)

\ / V v WV U ._ /

COMANCHE PEAK - UNITS 1 ANO 2 6-21 Unit 1 - Amendment No. 1,5,:0,10,21,2",34 Unit 2 - Amendment No. 4r7rl4,20

Attachment 3 to TXX 95076 B ~Page 23 of 23 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

Reference 18) is for Unit 2 only: l 18). WCAP-9220-P-A, Rev.1, " WESTINGHOUSE ECCS EVALUATION MODEL- 1981 version", February 1982 (W Proprietary).

/ w

( 6.9.1.6c The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.6d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for sach reload cycle,

^

to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

, fV 3 -% ,

!9). RXE.94 001.A, " Safety Analysis of Postuintedinadvertent Boron Dilution

,\, Event in Modes 3,4, and 3," February I994. (Methodologyfor Specifications 3/4.1.1.1,3/4.1.1.2, 3/4.1.2.2,3/4.1.2.4 and 3/4.1.2.6 - Shutdown Margin.)

'~%,_/V s xy- -

s COMANCHE PEAK - UNITS 1 AND 2 6-21a Unit 1 - Amendment No. 21 Unit 2 - Amendment No. 7

.n  ; q 4 . rt +~ y * - A + hJ - A b n a.

f t

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^

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l

.. .t f l,, ,. 4 l + ' %.

i P

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)

I s

~

ENCLOSURE 1 TO TXX 95076  :

. GENERIC LETTER 88 16 REMOVAL OF CYCLE SPECIFIC PARAMETER  !

LIMITS FROM TECHNICAL SPECIFICATIONS l i

i y I i

i f

i I

, 0 ?

e b

P ' [o no  %'t, UNITED STATES NUCLEAR REGULATORY COMMISSION

{

)

j WASHINGTON, D. C. 20088 OCT 0 4 W TO ALL POWER REACTOR LICENSEES AND APPLICANTS

SUBJECT:

REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICATIONS (GENERIC LETTER 88-16)

License amendments are generally required each fuel cycle to update the values of cycle-specific parameter limits in Technical Specifications (TS). The processing of changes to TS that are developed using an NRC-approved method-ology is an unnecessary burden on licensee and NRC resources. A lead plant proposal for an alternative that eliminates the need for a license amendment t to update the cycle-specific parameter limits each fuel cycle was submitted for the Oconee plant with the endorsement of the Babcock and Wilcox Owners Group. On the basis of the NRC review and approval of that proposal, the en-closed guidance for the preparation of a license amendment request for this alternative was developed by the NRC staff.

Generally, the methodology for determining cycle-specific parameter limits is documented in an NRC-approved Topical Report or in a plant-specific submittal.

As a consequence, the NRC review of proposed changes to TS for these limits is primarily limited to confirmation that the updated limits are calculated using an NRC-approved methodology and consistent with all applicable limits of the safety analysis. These changes also allow the NRC staff to trend the values of these limits relative to past experience. This alternative allows continued trending of these limits without the necessity of prior NRC review and approval.

Licensees and applicants are encouraged to propose changes to TS that are consistent with the guidance provided in the enclosure. Conforming amendments will be expeditiously reviewed by the NRC Project Manager for the facility.

Proposed amendments that deviate from this guidance will require a longer, more detailed review. Please contact the Project Manager if you have questions on this matter. ,

Sincerely, 8810050058

• g )

Denn s M. Crutchfiel Acting Associate Di ector for Projects RECE D Office of Nuclear Reactor Regulation

Enclosure:

As stated OCT 211988 WILLIAM G. COUNSIL

. . . - _ - - e - . - - - - - _ - - . - - - - - -

m.

y<,e Enclosure Generic ' Letter 88- 16 GUIDANCE FOR TECHNICAL SPECIFICATION CHANGES FOR CYCLE-SPECIFIC PARAMETER LINITS .

4 INTRODUCTION A number of Technical Specifications (TS) address limits associated with reactor physics parameters that generally change with each reload core, requir-ing the processing of changes to TS to update these limits each fuel cycle. ,

If these limits are developed using an NRC-approved methodology, the license '

amendment process is an unnecessary burden on the licensee and the NRC. An alternative to including the values of these cycle-specific parameters in in-4 dividual specifications is provided and is responsive to industry and NRC ,

efforts on improvements in TS. j This enclosure provides guidance for the preparation of a license amendment request to modify.TS that have cycle-specific parameter limits. An acceptable r~ alternative to specifying the values of cycle-specific parameter limits in TS was developed on the basis of the review and approval of a leadip lant proposal for this change to the TS for the Oconee units. The implementation of this 4-alternative will result in a resource savings for the licensees and tne NRC by

- eliminating the majority of license amendment requests on changes in values of cycle-specific parameters in TS.

l

! DISCUSSION This alternative consists of three separate actions to modify the plant's TS:

(1) the addition of the definition of a named formal report that includes the values of cycle-specific parameter limits that have been established using an NRC-approved methodology and consistent with all applicable limits of the safe-ty analysis, (2) the addition of an administrative reporting requirement to sub-mit the formal report on cycle-specific parameter limits to the Commission for information, and (3) the modification of individual TS to note that cycle-specific parameters shall be maintained within the limits provided in the defined formal report.

In the evaluation of this alternative, the NRC staff concluded that it is essential to safety that the plant is operated within the bounds of cycle-specific parameter limits and that a requirement to maintain the plant within the appropriate bounds must be retained in the TS. However, the specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC-approved method-ology and consistent with all applicable limits of the plant safety analysis that are addressed in the Final Safety Analysis Report (FSAR). Additionally, it was concluded that a formal report should be submitted to NRC with the-values of these limits. This will allow continued trending of this information, even though prior NRC approval of the changes to these limits would not be  :

required.

The current method of controlling reactor physics parameters to assure conform-ance to 10 CFR 50.36 is to specify the specific value(s) determined to be with-in specified acceptance criteria (usually the limits of the safety analyses) using an approved calculation methodology. The alternative contained in this guidance controls the valu6s of cycle-specific parameters and assures conform-ance to 10 CFR 50.36, which calls for specifying the lowest functional 4

+r ,p- ,-m--+-- .,,-m,e3,-.-9,.-.e ,w.%,-,-+y -- ,.-.n ,, . - , . ,, _ . -- .,-m--e- . - - - - - - - - - - -

e.

C' s .. p Generic Letter 88 16 Enclosure performance levels acceptable for continued safe opention, by specifying the calculation methodology and acceptance criteria. Th h permits operation at any specific value determined by the licensee, using the specified methodology, to be within the acceptance criteria. The Core Operating Limits Report will docu-  !

ment the specific values of parameter limits resulting from licensee's calcula--

tions including any mid-cycle revisions to such parameter values.

The following items show the changes to the TS for this alternative. A defined formal report, " Core Operating Limits Report" (the name used as an example for the title for this report), shall be added to the Definitions section of the TS, as follows.

[ CORE] OPERATING LIMITS REPORT 1.XX The [ CORE] OPERATING LIMITS REPORT is the unit-specific document that provides [ core] operating limits for the current operating reload cycle. These cycle-specific [ core] operating limits shall be determined for each reload cycle in accordance with Specification 6.9.X. Plant operation within these operating limits is addressed in individual specifications.

A new administrative reporting requirement shall be added to existing reporting '.

requirements, as follows.

l

CORE: OPERATING LIMITS REPORT

6.9.X] [ Core] operating limits shall be established and documented in the

[ CORE] OPERATING LIMITS REPORT before each reload ejcle or any remaining part of a reload cycle. (If desired, the individual specifications that address [ core) operating limits may be referenced.-) The analytical methods used to determine the [ core] operating limits shall be those previously re-viewed and approved by NRC in [ identify the Topical Report (s) by number, title, and date, or identify the staff's safety evaluation report for a plant-specific methodology by NRC letter and date). The [ core] operating i limits shall be determined so that all applicable limits (e.g. , fuel therm-al-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The [ CORE] OPERATING LIMITS REPORT, in-cluding any mid-cycle revisions or supplements thereto, shall be provided i upon issuance, for each reload cycle, to the NRC Document Control Desk with l copies to the Regional Administrator and Resident Inspector.

. Individual specifications shall be revised to state that the values of cycle-specific parameters shall be maintained within the limits identified in the defined formal report. Typical modifications for individual specifications ,

are as follows. j The regulating rods shall be positioned within the acceptable operatina rance for reculating rod position provided in the [ CORE] OPERATING LIMITS REPC.RT. (Used where the operating limit covers a range of acceptable operation, typically defined by a curve.)

The [ cycle-specific parameter] shall be within the limit provide <1 in the

[ CORE] OPERATING LIMITS REPORT. (Used where the operating limit has a single point value.)

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Generic. Letter 88- 16 Enclosure  ;

SUMMARY

The alternative to including the values of cycle-specific parameter limits in' individual specifications includes (1) the addition of a new defined term for the formal report that provides the cycle specific parameter limits. (2) the addition of its associated reporting requirement to the Administrative Controls '

section of the TS, and (3) the modification of individual specifications to re-place of these these limits with a reference to the defined formal report for the values limits. With this alternative, reload license amendments for the sole purpose of updating cycle-specific parameter limits wi.11 be unnecessary.

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