ML20044H003

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Proposed Tech Specs Supporting Use of in-house Reload Analysis Methodologies,Beginning W/Unit 1,Cycle 4
ML20044H003
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 05/28/1993
From:
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20044G997 List:
References
NUDOCS 9306070175
Download: ML20044H003 (13)
v • d • e


Text

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ATTACHMENT 3 to TXX-93204 AFFECTED TECHNICAL SPECIFICATION PAGES (NUREG - 1468)

(10 PAGES)

(2-2, 2-5, 2-6, 2-9, 2-10, 2-11, 3/4 2-12, 83/4 2-4' B3/4 2-6, and 6-21) 9306070175 930528 5 ADOCK 0500 gDR

, . I Attachment 3 to TXX-93024 i

,' Page'l of 12 go s E C A

[/ -~~-

OJ T N 7 670

/

/ 660  %

/ 650 385 PS 0 ER W 640  %

"* \  %

  • 2 xN ~~ NN e e' web NN

==

N TNN .

ct 590 NNN /

580 570 560 550 20 0 40 60 80 100 120 w

PERCENT OF RATED THERMAL POWER FIGURE 2.1-la t

UNIT 1 REACTOR CORE SAFETY 1.lHITS I

~

COMANCHE PEAK - UNITS 1 AND 2 2-2  !

Unit 1 - Amenhent No. 14 t l

  • Attachment 3 to TXX-93204 Page 2 of 12 rnseek A l

l 670 I l l

UNACCEPfABLE l _P= 2385 PSIG OPERATION j 650 i 640 Ix .N -

i P == 2235 PStG l  %

N, - '

i N l\

=a 1985 PSIG

@ 610 P= ,% l\

y) 600

\m <

o , ,

W ACCEPTABG OPERATION!

580 ,

! l l l i 570 i

560 ,

i l l 550 ,

0 20 40 60 80 100 120 PERCENT OF RATED THERMAL POWER

\

I

? #-

ee CD TABLE 2.2-1 " ' "

n B

2 REACTOR TRIP SYSTEM INSTRlHENTATION TRIP SETPOINTS TOTAL z"

SENSOR ALLOWANCE E

" ERROR w FUNCTIONAL UNIT (TA) Z (S) TRIP SETP0lNT ALLOWABLE VALUE E' E

. 1. Nanual Reactor Trip N.A. N.A. N.A. N.A. N.A C .

2. Power Range, Neutron Flux

- a. High Setpoint 7.5 4.56 1.25 $109% of RTP* slll.7% of RTP*

b b. Low Setpoint 8.3 4.56 1.25 s25% of RTP* $27.7 of RTP*

N

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

~ 4. Power Range, Neutron Flux, 1.6 0.5 0 $5% of RTP* with $6.3 of RTP* with S. High Negative Rate a time constant a time constant 22 seconds 22 seconds

5. Intermediate Range, 17.0 8.41 0 $25% of RTP* s31.5 of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 s105 cps s1.4 x 105 cps E 7. Overtemperature N-16

% a. Unit 1

b. Unit 2 l 10.0 5[8/ / / 3MS/

6.75

/1.f+d8[} See Note 1 See Note 2 s

1.0+1.38+ See Note 1 See Note 2 0.96t2)

{^f f 0,5 3 6.70 f.0 +f.10 t 0.76 ci a

W g,go s Pu y; R'rb.s and n. 76 % L p re.sswy P re sw<. s e.sso rs, f {g} t . O *'o s ea n k'N~'"P" ' ~,,'

E

"{

o

  • RTP - RATED THERMAL POWER g l(pf 1.2%/ span for defta-T/ RIDS) add 0.8% fgr press /rizer odssurdi (2) 1.0% span for N-16 power monitor,1.38% for Tc.:s RIDS and 0.96% for pressuriter pressure sensors k - - _ ___ -.______ -._____

M e-TABLE 2.2-1 (Continued) ',

n .

  • J.

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Sig 9

'" TOTAL SENSOR ALLOWANCE ERROR E fjjNCTIONAL UNIT (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE g

. 8. Overpower N-16 E e- y. 1 y/ n iAA / / / VM' / A1.0+0.05*

// /<X12 Var t1% / <A %3t dFM:+Y "

y y A. tM/ 4.0 2.05 il12%ofRTP* 2114.5%ofRTP'*

  • m .

- 9. Pressurizer Pressure-Low g a. Unit 1 4.4 0.71 2.0 21880 psig 21863.6 psig a b. Unit 2 4.4 1.12 2.0 21880 psig 21863.6 psig m

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 m 11. Pressurizer Water Level-High

& a. Unit 1 8.0 2.18 2.0 592% of instrument $93.9% of instrument span span

b. Unit 2 8.0 2.35 2.0 $92% of instrument s93.9% of instrument span span
12. Reactor Coolant Flow-Low g a. Unit 1 2.5 1.18 0.6 290% of loop 288.6% of loop
design flow ** design flow **

- b. Unit 2 2.5 1.25 0.87 290% of loop 288.8% of loop minimum measured minimum measured fl ow* *

  • flow ***

s

(

s (3) 1.0% span for N-16 power monitor and 0.05% for T RTP - RATED THERMAL POWER RTDs.

[

    • Loop design flow -lWQIlibCA 4-- 9 9,oso q pm o *** Loop minimum measured flow - ~90,500 gpa -

EN TABLE 2.2-1 (Continued) %C n

o TABLE NOTATIONS e og NOTE 1: Overtemperature N-16 %g

~

x mw 1 + r's -

m N -

K,-K2[I+rs z T,-T,*] + K3 (P-P') - f, (aq) g G t

. 7 g Where: N -

Measured N-16 Power by lon chambers, ~

8

~ 2 u

d T, - Cold leg temperature, *F, 560.5 g T,*

-WF for Unit 1, 560.3*F for Unit 2 - Reference T, at RATED THERMAL POWER, K, -

0.013_4 K2 -

F for Unit 1 m

. 56/*F for Unit 2 a 1.+ f l i -

The function generated by the lead-lag controller for I+T52 T, dynamic compensation, r,, 12 Time constants utilized in the lead-lag controller for T,, r,;t10 s, and 72 s 3 s, d 0.0o0719 K3 - gr#@psig for Unit 1 0.000898/psig for Unit 2 E

M

. -z

a

,=

TABLE 2.2-1 (Continued) Th omoe, g '

TABLE NOTATIONS (Continued) O '

G s'

NOTE 1: (Continued)

" NW P -

Pressurizer pressure, psig, g k w P' 2 2235 psig (Nominal RCS operating pressure),

c M a

5 5 -

Laplace transform operator, s,

$ M and f (aq) is a function of the indicated difference between top and bottom halves of i

2 g

detectors of the power-range neutron ton chambers; with gains to be selected based on c,

measured instrument response during plant STARIUP tests such that: ,

~

For Unit 1 l - 'S ?* *

  • 2 + +7*

4 o  ;

(1) for q - 4 between /}6f/pf)d/+/f)(, f (aq) - 0, where q, and q are percent '

RATED,THE151AL POWEL in the top and b,ottom halves of the core,respectively, y and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER,

-GS%

(11) for each percent that the magnitude of q - exceeds M he N-16 Trip i Setpointshallbeautomaticallyreducedby of its value at RATED '

THERMAL POWER, and I. si %

4

=

c (iii) for each percent that the magnitude of q -- ex eeds h the N-16 Trip 3 Setpointshallbeautomaticallyreducedby of its value at RATED THERMAL POWER.

- 2. 16 7 8

t 2

?

i-

N

%2 TABLE 2.2-1 (Continued)

~e n o R.

TABLE NOTATIONS (Continued) 'a

$ Gw G NOTE 1: (Continued) -

x O For Unit 2 d x

k (1) for q - between -52% and +5.5%, f,(aq) - 0, where q, and q are percent b RATED,THEbl POWER in the top and bottom halves of the core ,respectively, 2 e "

and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER,

_z .

d (11) for each percent that the magnitude of q - q, exceeds -52%, the N-16 Trip

- Setpoint shall be automatically reduced by 2.15% of its value at RATED THERMAL g POWER, and a

m (111) for each percent that the magnitude of q - q, exceeds +5.5%, the N-16 Trip Setpoint shall be automatically reduced by 2.17% of its value at RATED THERMAL POWER.

7 NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than of span for Unit 1 or 2.88% of span for Unit 2. 3,ss g t

t a

i &

l

Attachment 3 to TXX-93204 Page 8'of 12 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION .

3.2.5 The following DNB-related parameters shall be maintained within the stated limits:

a. Indicatad Reactor Coolant System T m 5 592*F 0
b. Indicated Pressurizer Pressure 2 2. 1 1 7 P.s8 g-
c. Indicated Reactor Coolant System (RCS) Flow 2GM(76@hm** for Unit 1 2 395,200 gpm** for Unit 2 APPLICABILITY: MODE 1. 404 + 0 0 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to 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 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 RE0VIREMENTS 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 />.

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 Surveillancs Requirement 4.2.5.4.  !

4.2.5.4 The ACS total flow rate shall be determined by precision heat balance measurement after each fuel loading and prior to operation above 75% of RATED THERMAL POWER.. The feedwater prescure and temperature, the main steam pres-sure, and feedwater flow differential pressure instruments shall be calibrated 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. 14 l

Sttachment 3 to TXX-93024

,.Page 9 of 12

~

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL fy g (Continued) and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position;
b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;
c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.  ;

F"3 will be maintained within its limits provided Conditions a. through N

d. above are maintained. The relaxation of F g as i a function of THERMAL POWER allows changes in the radial power shape for aT1 permissible rod insertion limits.  ;

l Fuel rod bowing reduces the value of the DNB ratio. Credit u available to t this reduction in the generic margin. The DNBR generic margin, totaling 17.1% or Unit I and 10.1% for typical cells and 9.5% for thimble cells for L  !

2 for DNBR completely offset any rod bow penalties.VThis argin nelude the 0110 ing or U t 1:

a. si limi DNBR of 1. vs .28, b Gr a ng ( ) of .046 s 0. 9,
c. Dif sion oeff ien of .038 s 0.0 ,

d DN Mul plie of 0 5v 0.8 , an

e. itch educ on.

uni + 4 mad The margin'for; Unit 2 is included by establishing a fixed difference between the safety analysis limit DNBR and the design limit DNBR equal to the percent margin of the safety analysis limit DNBR.

The applicable values of rod bow penalties are referenced in the FSAR.

COMANCHE PEAX - UNITS 1 AND 2 B 3/4 2-4 Unit 1 - Amendment No. 1, 14

Attachment 3 to TXX-93204 Page 10'of 12 POWER OISTRIBUTION LIMITS BASES 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 592.7'F (conservat vel rounded to 592*F) and the Unit 1 indicated p,ressurizer ressu CM value of psig correspond to analytical limits of 594.7'F and respective y, with allowance for measurement uncertainty. The Uni indicated T

m value of 592.8'F pressurizer pressure v(conservatively rounded to 592*F) and the Unit 2 indicated alue of 2219 psig correspond to analytical limits of 595.16'F and 2205 psig respectively, with allowance for measurement 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 th 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 additional surveillance requirements associated with the RCS total flow rate are sufficient to ensure that the measurement uncertainties are limited to 1.8% 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 accurately measure the RCS flow. Subsequently, the RCS flow detectors (elbow tap differential pressure detectors) are normalized to this flow determination and used throughout the cycle.

i 1

1 1

COMANCHE PEAX - UNITS 1 AND 2 B 3/4 2-6 Unit 1 - Amendment No. 1, 6 l

l

Attachment 3 to TXX-93024

,. Page 11 of 12 ADMINISTRATIVE CONTROLS l

l CORE OPERATING LIMITS REPORT (Continued) '

5). WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F, SURVEIL-LANCE TECHNICAL SPECIFICATION," June 1983 (H Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor NSER.T P; NI*) '""* """ "A" "**" * '

  • Methodology).)

Reference and re for Unit 1 only:

l+ IS

+) WCAP-8200, "WFLASH, A FORTRAN-IV COMPUTER PROGRAM FOR SIMULATION OF TRANSIENTS.IN A MULTI-LOOP PWR," Revision 2, June 1974 (W Proarie- p tary). I(M6thidglogy fArApecyficAtton 3.742/- Heg Fl#x Kot/Qiasmyll lf49tyrg)l '

is) 7 . WCAP-9220-P-A, " WESTINGHOUSE ECCS EVALUATION MODEL, February 1978 p Version." February 1978 (W Pro)rietary). l 44 hedoMg9 fArApfc}fyA

{ cht/ od M2/2 / H(at Vlug Hpt C)enptl fac r. )l

/6) 17) /8) /9)

Referencesf),M,17)and}E)areforUnit2only:

/6) WCAP-9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION MODEL- 1981 Version". February 1982 (W Proprietary)._ Re f9f IS6e/1Ittattorf 7.242/- W6at,Flyk hdtAhydr a oor/.)/

olpgyj/j g I7)U9. WCAP-10079-P-A, "NOTRUMP, A NODAL TRANSIENT SMALL BREAX AND GENERAL NETWORK CODE." Auaust 1985 (W Proprietary?. I hogo}cgy/fof Isoa/cifkat/on I J.I - Regt fl/x Hst CKanp61 F c rA / / / 9 I 8) WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAX ECCS EVALUATION MODEL USING THE NOTRUMP CODE". Auaust 1985. (W Proprietary).i ~t lAyf IIru' sn'acM4cztMn X.f.Y - W6at'FMr Aotsth/nn#1 IadtdrM,(

//A i9) WCAP-11145-P-A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL

.GFNERfC STUDY WITH THE NOTRUMP CODE". October 1986. (W Proprietary). 2 -

l>f4afh6d'oMof f6r S6edt#igittor(X2A - Afeaf/Fl# HoyChynnpr F,acpor/)/

,6.9.1.6c The core operating limits shall be determined so that all applicable limits (e.g., fuel thensal-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.

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

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

l t

l l l

l COMANCHE PEAK - UNITS 1 AND 2 6-21 Unit 1 - Amendment No. 1, 6, 14 f

.- Attachment 3 to TXX-93204

, Page 12 Of 12 knserb ky(). 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 - Heat Flux Hot Channel Factor.)

@,/). RXE-88-102-P, "TUE-1 Departure from Nucleate Boiling Correlation". January 1989.

8,8$ . RXE-88-102-P, Sup. 1, "TUE-1 DNB Correlation - Supplement 1",

Dacember 1990.

f g) . RXE-89-002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for -

Comanche Peak Steam Electric Station Licensing Applications",

June 1989.

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

ll /') . RXE 91 002, " Reactivity Anomaly Events Methodology", May 1991.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdewn 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. 3.2.3 - Nuclear Enthalpy ,

Rise Hot Channel Factor.)

(17). RXE-90 007, "Large Break Loss of Coolant Accident Analysis ,

Methodology", December 1990.

(J/f) . TXX-88306, " Steam Cenerator Tube Rupture Analysis ", March 15, 1988.

t

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