ML20087K175

Proposed Tech Specs,Moving Shutdown Margin Limits from TS & Providing Guidelines for Removal of cycle-specific Parameter Limits from TS

ML20087K175
Person / Time
Site:	Comanche Peak
Issue date:	08/15/1995
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20087K174	List: ML20087K172 ML20087K175
References
NUDOCS 9508230068
Download: ML20087K175 (12)

Text

._ _

.; 1 3/4.1' REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T y GREATER THAN 20(fF LIMITING CONDITION FOR OPERATION l

3.1 1.1 Ty SHUTDOWN MARGIQall be greater than or equal tgM APPLICA  : MODES 1, 2*, 3, and 4. 1 Operatieg Mmits ACTION:

• V*lu* SPecNi'M (CocetRepoet (col.R). '

f the. coLR 4 3 l With the SHUTDOWN MARGIN less than E " ?!E ' r U .it 1 l:. : _h!' #~ m 4 + M, immediately initiate and continue boration at greater than or equal ;o : Ci gpm of a solution containing greater than or equal to 7,000 ppe boron or equivalent i until the required SHUTDOWN MARGIN is restored. ,

i SURVEILLANCE REQUIREMENTS

4. L 1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or eqdal d

to ' 3% ani m im vn n i ii . ; 1/% OT 0; i t 2'8 f the. valut ePt.c.ified M the COL,(U

a. wunin i nour after cetection 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required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

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

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 rod position is within the limits of Specification 3.1.3.6;

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 Unit 1 - Amendment No. 14 9508230068 950815 PDR ADOCK 05000445 P PDR

o ..

REACTIVITY CONTRDL SYST[MS

. SHUTDOWN MARCIN - T LESS THAN OR EDUAL TO 200*F LIMITING CONDITION FOR OPERATION '

3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to, 6 APPLICABILITY: MODE 5.

ACTION.

ha value spec ledl (4he VakeLivnt+s Operatteg spec.ified Report M *the Cort]1 (coLR u in 4ht COLR With the SHUTDOWN MARGIN less than6, insediately 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 SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.l.2 ThatSHUTDOWN MARGIN shall be determined to be greater than or equal tg,..- --phe. vcaue. specified in +ke COLR]

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,

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 Unit 1 - Amendment No. 5

p

'x I

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING f 1

' i LIMITING CONDITION FOR OPERATION l

i 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE: l

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 l i

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.* g gg1 ACTION: . t ni %e. COL.R J l With only one of the above required boron in action flow paths to the RCS OPERABLE, restore at least two boron inject n 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'iOT STAND 8Y and borated .

to a SHUTDOWN MARGIN equivalent to at leastn Z - - 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 paths to O'ERABLE status within i

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

SURVEILLANCE REQUIREMENTS l

4.1.2.2 At least two of the above required flow paths shall be demonstrated  !

OPERABLE J

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; I

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

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. j i

i

• A maximum of two charging pumps shall be OPERABLE whenever the temperature of f 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 m -w-o- ,e --,.,--w w ur -

er- v --w-e- -- - - - ' ' y =v' *e pr-e -

-7 w -=r-P+'-'&"

g REACI1VIII G.QlLT.BQL 1HIf.til CHARGINGPUMPS....-OPERATING LIMITING CONDITION FOR OPERATION .

3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE. 1 APPLICABILITY: MODES 1, 2, 3*, and 4* **.

ACTION:

4h mWe spec *edl  ;

ggcogg i With only one charging pump OPERABLE,1 restore at least two charging pumps to OPERA SHUTDOWN 8LE status MARGIN within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> equivalent to or at be(in at rleast least 1r m atHOT 200*FSTANDBY within the and nextborated to a )

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

{

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

1 4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specification 4.1.2.2.c. {

1 4.1.2.4.3 Whenever the temperature of one or more of the Reactor Coolant System-(RCS) cold legs is less than or equal to 350*F, a maximum of two chargin i shall be OPERABLE, except when Specification 3.4.8.3 is not-applicable. g pumps When required, one charging pump shall be demonstrated inoperable # at least I once per 31 days by verifying that the motor circuit breakers are secured in the open position.  :

• The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry 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 0PERABLE 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.

• • In MODE 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.

1. 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) i secured in the closed position.  !

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

i

-REACTIVITY CONTROL SYSTEMS

- 1 BORATED WATER SOURCES'- OPERATING LIMITING CONDITION FOR GPCRAT!ON .

3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: _i

~

a. A boric acid storage tank with: '

1) A minimum indicated borated water level of 50%, I

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:

i

1) A minimum indicated borated water level of 95%, ,

2) A boron concentration between 2400 ppm and 2600 ppm, f

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

4) A maximum solution temperature of 120*F. I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the boric acid storage tank inoperable and being used as one i 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 hour 1_and borated to a SHUTDOWN MARGIN equivalent to at l e a s t " - " "-'at 200'F; restore the boric acid storage tank to OPERABLE status"within the next 7 days or be in COLD SHUTDOWN within j the next 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />

e. Vebe specified M %e COLR] l

b. With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STAND 8Y 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 COMANCHE PEAK - UNITS 1 AND 2 3/4 1-13 Unit 1 - Amendment No. 4-44,26 Unit 2 - Amendment No 6,' 2

s

. -w 3 /4,) REACTIVITY CONTROL SYSTEMS BA!ES 3/4L1.1 BORATION CONTROL 144.1.1.1 and 3/4.1.1.2 SHUTDOWN MARefN 4

A sufficient SHUTDOWN 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 acceptable listts,.and (3) the reactor will be maintained sufficiently suberitical to i preclude inadvertent criticality in the shutdown condition.

p SHUTDOWN MARGIN requirements vary throughout core life as a function of -

r(*S st*CMd fuel depletion, RCS boron concentration, and RCS . The most restrictive ccedition occurs at E0L, with W h Co q{ ciated with a postulated steam Une break accident T at ne lead ope temperature, and is asso--

L resulting uncontroll -

+ R 5 coo komn. In the analysis of this ace' dent, a sinimum SHUTDOWN MARGIN l is required to control-the reac v-J ty transien3. Accorti ng y, e WIGIN requirement is based upon this '

Q'" t T M limting concition and i consistent with FSAR safety analysis ass tons.

i With T less than 200*F 5HUTD0lBI MARSIN

,s i chove - -

is based the results af ren di ution accident ana ysis,z goo.p 2 reguiced)

Since the actual overall core reactivity balance comparison required by 4.1.1.1.2 cannot be performed until after criticality is attained, this compari-scr. is as req.; ired (and the provisions of Specification 4.0.4 are not app 11- ,

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 COEFFICIENT i

The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the F5AR accident and transient analyses.

The MTC values of this s  !

plant conditions; accordingly,pecification verification ofare NICapplicable to a specific values at conditions set of other than those explicitly stated will require extrapelatten to these conditions in order to permit an accurate comparison. ,

The most negative NTC value equivalent to the most positive moderator.  !

density coefficient CleC) was obtained by incrementally correcting the ICC  :

used in the FSAR analyses to nominal-operating conditions. These corrections 1

. i

COMANCHE PEAK - UNITS 1 AND 2 8 3/4 1-1 Unit 1 - Amendment No. 4,14 4

_ ~ - . . . _ _ . _ . . . ~ _ _ __ .. _ . _ .

l

~ '-

REACTIVITY CONTROL SYSTEMS BASES I

MODERATOR TEMPERATURE COEFFICIENT (fontinued) involved subtracting the incremental change in the MOC associated with a core I condition of all rods inserted (most positive MOC) to an all rods withdrawn >

condition and, a conversion for the rate of change of moderator density with i temperature at RATED THERMAL POWER conditions.

This value of the MOC was then  :

transformed into the limiting End of Cycle Life (EOL) 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 ppm equilibrium boron '

concentration MTC value.

and is obtained by making these corrections to the limiting EOL l The Surveillance Requirements for measurement of the MTC at the beginning l 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 i principally to the reduction in RCS boron concentration associated with fuel  !

burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY I i

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 OPERABLE 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 pumos  ;

(3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency ,

power supply from OPERA 8LE diesel generators.

g

• With the RCS average temperature above 200'F, a minimum of two bo n i injection flow paths are required to ensure single functional capabil ty in the event an assumed failure renders one of the flow paths inoperab . The boration capability of either flow path is sufficient to provid HUTDOWN l

^

MARGIN from expected operating conditior:s -

Mr M ") after xenon decay and cooldown to 200 . e max' mum expec  !

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 the refueling water storage tank (RWST). l 1

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 1-2 Unit 1 - Amendment No. 5, M, M,26 Unit 2 - Amendment No. 6,12 i

i ~**

REACTIVITY CONTROL SYSTEMS- a BASES BORATION SYSTEMS-(Continued)-

't 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 Injection System becomes inoperable, .

The: limitation for a maximum.of two charging pumps to be OPERA 8LE 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 allowableare sufficient boron to prevent boric acid crystallization with the highest concentration, -

g The boron capability required below 200*F is sufficient to provide g "

SHUTDOWN MARGINE _ - r/ rafter xenon decay and cooldown from 200*F to '

140*F. This condiu on 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.- '[ l

' 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 MODES Level Volume Volume- Uncertainty Margin Other (gal) (gal) (gal) (gal)

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

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 Storage (gravity feed)

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

1

• Additional volume required te meet Specification 3.5.4.

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 1-3 Unit 1 - Amendment No. Gr M ,26 Unit 2 - Amendment No. 6,12 l

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

ADMINISTRATIVE CONTROLS

..,. i ..'

. MONTHLY OPERATING REPORTS (Continued)

. shall be submitted.on a monthly basis;to the U.S. Nuclear' Regulatory o

~ Commission, Document Control Desk, Washington, D.C. 20555, with a copy-to the

r- Regional Administrator of the Regional Office of the NRC, no later than the

.. d 15th of each month following the calendar month covered by the report.

j

},g.:.  !"

CORE OPERATING LIMITS REPORT 4- 5,

[a 6.9.1.6a Core operating limits shall be established and document'ed in'the if(v

• CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining i

part of a reload cycle for the following: .

[g fci {

s, 5 1). Moderator temperature coefficient 80L and E0L limits and 300 ppe sur- i veillance limit for Specification 3/4.1.1.3, i sf g' 8 . 2). Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, '

8 .7 ' 3). Control Rod Insertion Limits for Specification 3/4.1.3.6,

~~ #

h; 4). AXIAL FLUX DIFFERENCE Limits-and target band for Specification g; 3/4.2.1.,.

- ,. p 5). Heat Flux Hot Channel Factor, K(Z), W(Z)., F/*, and the F/(Z) '

allowances for Specification 3/4.2.2, 3l 6). Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor gd Multiplier for Specification 3/4.2.3.

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

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

July 1985 (W Proprietary). (Methodology for Specifications 3.1.1.3 - l 1

'E Moderator Temperature Coefficient, 3.1.3.5 - Shutdown' Bank Insertion Limit 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux a S 1 Difference, 3.2.2 - Heat Flux Not Channel Factor, 6 3.2.3 - Nuclear i Enthalpy Rise Hot Channel Factog.) 7

? r c' h.

n 2). WCAP-8385, " POWER DISTRIBUTION CONTROL Alm LGAD FOLLOWING PROCEDURES -

8 TOPICAL REPORT," September 1974 (M Proprietary).. (Methodology for r Specification 3.2.1 - Axial Flux Differencs [ Constant' Axial Offset

, )4 . cs Control).)

(' J .

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

.E January 31, 1980--

Attachment:

Operation and Safety Analysis Aspe as

~. r .

of an Improved Load Follow Package. (Methodology for Specification Em e c 3.2.1 - Axial Flux Difference [ Constant Axial Offset Control).)

a 8

_ 4). NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission,

!aEC L Section 4.3, Nuclear Design, July 1981. Branch Technical Position  ;

4 9 *h CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. (Methodology for Specification 3.2.1 - Axial Flux

e. E.

E~{t- Difference [ Constant Axial Offset Control).)

COMANCHE PEAK - UNITS 1 AND 2 6-20 Unit 1 - Amendment No. 6r+4. 34 Unit 2 - Amendment No. 20  !

3D l Amitid!$TRATI E CGIITROLE i

C0RE OPERATING LIMITS REPORT (Continued) j 5). NCAP-10214-P-A, Revision IA, " RELAXATION 0F CONSTANT AXIAL 0FFSET i CONTRDL F. SURVEILLANCE TECHNICAL SPECIFICATION." February 1994 (M j Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot. -

Channel Factor (W(z) surveillance requirements for.F. Metho1 ology).)

.j 6). WCAP-10079-P-A, "NOTRUMP, A N004L TRANSIENT SMALL BREAK AND GENERAL l NETWORK CODE," August 1985, (M Proprietary).

7). WCAP-10054-P-A, " WESTINGHOUSE L BREAK ECCS EVALUATION MODEL USING THENOTRUMPCODE", August 198{,(M reprietary).  ;

8). WCAP-ll145-P-A,"WESTINGHOUSESMALLBREAKLOCAEcrbKUATIONMODEL~ -I GENERIC STUDY WITH THE NOTRUMP CODE", October 198 toprietary). -l 9). RXE-90-006-P, " Power Distribution Control Analysis and Overtemperature ,

N-16 and Overpower N-16 Trip Setpoint Methodology," February;1991.  !

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - i Heat Flux Not Channel Factor.) l i

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

January 1989.

11). RXE-88-102-P, Sup. 1, "TUE-1 DN8 Correlation - Supplement 1", , j December 1990.

l t

12). RXE-89-002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", June j

' l 1989. '

a 13). RXE-91-001, " Transient' Analysis Methods for Comanche Peak Steam l

-Electric Station Licensing Applications", February 1991. j s

14). RXE-91-002, " Reactivity Anomaly Events Methodology" May 1991. 1 (Methodology for Specification 3.1.1.3 - Moderator Temperature l Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - i Control Bank Insertion Limits,. 3.2.1 - Axial Flux Difference, 3.2.2 -

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

Channel. Factor.) j 15). RXE-90-007 "Large Break Loss of Coolant Accident Analysis i Methodology", December 1990. j 16). TXX-88306, " Steam Generator Tube Rupture Analysis", March' 15, 1988, 17). RXE-91-uG5, " Methodology for Reactor Core Response to Steamline Break l Events," May, 1991. j 3

(tnagedd g hr SpeeMie. dens. V4.l l.I $ V4.l.l. h 'th.l.'4.2, l

("b/4.1.3,4 and 'a/4.1.2.(, - sbJtelww Mar $n.) . _

COMANCHE PEAK - UNITS 1 AND 2 6-21 Unit 1 - Amendment N' o. !,5,",!!,2

,20,34 I Unit 2 - Amendment No. t,7,10,20 l

I ADMINfETRATIVE rd*TROLE

.. l CORF DPERATING thifTS REPORT (Continued)

Reference 18) is for Unit 2 only:

28). 'WCAP-9220-P-A, Rev.1 ' WESTINGHOUSE ECC5 EVALURTIO Versten", February 1982 W Proprietary).

b .9.1.6c l

The core operating limits shall be determined so that all 11 cable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic l sits.

analysis limits) of the safety analysis are met.ECC5 limits, nuclea 6.9.1.6'd supplements thereto, shall be provided upon issuance, to the and NRC Document Resident Inspector. Control Desk with copies to the Regional Adelnistrator I

l 19). RXE-94-001-A, " Safety Analysis of Postulatedinadvertent Boron Dilution Event in Siodes 3, 4, and 5," February I994. (Alethodologyfor 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 31argin.) '

Q J 1

COMNCHE PEAK - INilTS 1 AfC 2 6-21a Unit 1 - Amendment No. 21 Unit 2 - Amendment No. 7 i

.5

' We

,' (q i

. t 8

R ENCLOSURE 1 TO TXX 95215 GENERIC LETTER 88 16 REMOVAL OF' CYCLE SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICATIONS 4

I 1

t i

a# **

[y,  %,,

' UNITE] STATES I

.!" o l NUCLEAR REGULATORY COMMISSION

! I waanewsrow, n. c.zosos  ;

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-a ,

ology is an unnecessary burden on licensee and NRC resources.pproved method-A lead plant proposal for an alternative that eliminates the need for a license amendment 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 4 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.

I 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 i

values of these limits relative to past experience. This alternative allows I continued and trending of these limits without-the necessity of prior NRC review approval.

Licensees and applicants are encouraged to propose chan i consistent with the guidance provided in the enclosure.ges to TS that Conforming are 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 i RECE D Acting Associate Di ector for Projects {

Office of Nuclear Reactor Regulation  ;

Enclosure:

As stated OCT 211988 i

WILLIAM G. COUNSIL i

ff C llCO Of b a='

_;< w Generic l Letter 88- 16 Enclosure  !

. . j GUIDANCE FOR TECHNICAL SPECIFICATIGN CHANGES FOR CYCLE-SPECIFIC PARAMETER LIMITS l

1 INTRODUCTION A number of Technical Specifications (TS) address limits associated with  !

reactor physics parameters that generally change with each reload core, requir- 'l

'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' a alternative to including the values of these cycle-specific parameters in in-  !

dividual specifications _is provided and is responsive -to industry and NRC -

efforts on improvements in TS.  !

This enclosure provides guidance for the preparation of a license amendment 1 request to modify TS that have cycle-specific parameter limits. An acceptable '

alternative to specifying the values of cycle-specific parameter limits in TS was developed on the basis of the review and approval of a lead ' plant proposal for this change to the TS for the Oconee units. The implementation of this .

i alternative will result in a resource. savings for the licensees and the NRC by  :

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

DISCUSSION ,

L 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 i values of cycle-specific parameter limits that have been established using an i NRC-approved methodology and consistent with all applicable' limits of the safe- l 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 l 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 1 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 i 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-l ance to 10 CFR 50.36 is to specify the specific value(s) determined to be with- i in specified acceptance criteria (usually the limits of the safety analyses)  ;

using an approved calculation methodology. The alternative contained in this guidance controls the values of cycle-specific parameters and assures conform-  ;

ance to 10 CFR 50.36, which calls for specifying the lowest functional '

l

_s ..

Generic Letter 88- 16 Enclosure performance levels acceptable for continued safe operation, by specifying the calculation methodology and acceptance criteria. This permits operation at any specific be withinvalue determined criteria.

the acceptance by the licensee, using the specified methodology, to 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 requirements, reporting requirement shall be added to existing reporting as follows.

CORE] OPERATING LIMITS REPORT

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

[ CORE] OPERATING LIMITS REPORT before each reload cycle 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 limits shall be determined so that all applicable limits (e.g., fuel therm-al-mechanical limits, core ther.nal-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 upon issuance, for each reload cycle, to the NRC Document Control Desk with 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.

The regulating rods shall be positioned within the acceptable operating rance for regulating rod position provided in the [ CORE] OPERATING LIMITS REPORT. (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 provided in the

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

m

t Generic Letter 88- 16 ,

Enclosure t

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 these limits with a reference to the defined formal report for the values of these limits. With this alternative, reload license amendments for the sole purpose of updating cycle-specific parameter limits will be unnecessary.  !

i i

1 1

w, -

1 Enclosure i i

LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Date of Letter No. Subject Issuance Issued To i l

88-15 ELECTRIC POWER SYSTEMS - 09/12/88 ALL POWER REACTOR  !

INADEQUATE CONTROL OVER LICENSEES AND i DESIGN PROCESSES APPLICANTS

.l 88-14 INSTRUMENT AIR SUPPLY 08/08/88 ALL' HOLDERS OF I SYSTEM PROBLEMS AFFECTING OPERATING LICENSES SAFETY-RELATED EQUIPMENT OR CONSTRUCTION -

PERMITS FOR NUCLEAR POWER REACTORS l 88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR '

EXAMINATIONS LICENSEES AND APPLICANTS FOR AN OPERATING' LICENSE. l 88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR.  !

REQUIREMENTS FROM TECHNICAL LICENSEES AND SPECIFICATIONS i APPLICANTS- .

88-11 NRC POSITION ON RADIATION 07/12/88 ALL LICENSEES OF EMBRITTLEMENT OF REACTOR .

OPERATING REACTORS VESSEL MATERIALS AND ITS AND HOLDERS OF ,

IMPACT ON PLANT OPERATIONS '

CONSTRUCTION PERMITS 88-10 PURCHASE OF GSA APPROVED 07/01/88 ALL POWER REACTOR  ;

SECURITY CONTAINERS LICENSEES AND  !

HOLDERS OF PART 95-  !

APPROVALS

{

88-09 PILOT TESTING OF FUNDAMENTALS 05/17/88 ALL LICENSEES OF ALL EXAMINATION BOILING WATER REACTORS AND APPLICANTS FOR A i BOILING WATER REACTOR ,

OPERATOR'S LICENSE i UNDER 10 CFR PART 55- I 88-08 MAIL SENT OR DELIVERED TO t

05/03/88 ALL LICENSEES FOR POWER THE OFFICE OF NUCLEAR REACTOR REGULATION AND NON-POWER REACTORS AND HOLDERS OF CONSTRUCTION PERMITS FOR NUCLEAR POWER i 1-REACTORS .!

88-07 MODIFIED ENFORCEMENT POLICY 04/07/88 ALL POWER REACTOR RELATING TO 10 CFR 50.49, LICENSEES AND

" ENVIRONMENTAL QUALIFICATION APPLICANTS OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY FOR l NUCLEAR POWER PLANTS" i