ML20058C862

From kanterella
Jump to navigation Jump to search
Amends 21 & 7 to Licenses NPF-87 & NPF-89,respectively, Consisting of Changes to Tech Specs Re Cycle 4 Operations, Specifically,Core Safety Limit Curves & N-16 Overtemp Reactor Trip Setpoints
ML20058C862
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 11/16/1993
From: Black S
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20058C864 List:
References
NUDOCS 9312020601
Download: ML20058C862 (21)
v • d • e


Text

(***'Goq k

UNITED STATES E D E

NUCLEAR REGULATORY COMMISSION I

(h

/

WASHINGTON, D.C. 20556K101 gv i

TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION. UNIT I DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 21 License No. NPF-87 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Texas Utilities Electric Company (TU Electric, the licensee) dated May 28, 1993, as supplemented by letter dated September 24, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

1 9312O20601 931116 PDR ADOCK 05000445 P

pyg

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-87 is hereby i

amended to read as follows:

2.

Technical Specifications and Environmental Protection Plan i

The Technical Specifications contained in Appendix A, as revised through Amendment No.

21, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The licensee shall. operate the facility i

in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance to be l

implemented within 30 days of issuance.

F0 THE NUCLEAR REGULATORY COMMISSION g}Q Q

i Suzanne C. Black, Dire r

Project Directorate IV-2 Division of Reactor Projects III/IV/V i

Office of Nuclear Reactor Regulation s

Attachment:

i Changes to the Technical Specifications i

Date of Issuance:

November 16, 1993

. y l

4

h

^

t UNITED STATES I

)H E

NUCLEAR REGULATORY COMMISSION i

j

[

WASHINGTON, D.C. 20555-0001

[

%,...../

V TEXAS UTILITIES ELECTRIC COMPANY COMANCHE fEAK STEAM ELECTRIC STATION. UNIT 2 DOCKET NO. 50-446 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 7 License No. NPF-89 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Texas Utilities Electric Company (TU Electric, the licensee) dated May 28, 1993, as supplemented by letter dated September 24, 1993, complies with the standards and requirescents of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense' and security or to the health and safety of the public; i

and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l r

j

i t I 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment-t and Paragraph 2.C.(2) of Facility Operating License No. NPF-89 is hereby i

amended to read as follows:

2.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

7, and the Environmental Protection Plan

+

contained in Appendix B, are hereby incorporated into this license.

TU Electric shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Suzanne C. Black, Direct Project Directorate IV-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation Attachment-Changes to the Technical Specifications Date of Issuance:

November 16, 1993

\\

l

t 4

I ATTACHMENT TO LICENSE AMENDMENT N05. 21 AND 7 FACILITY OPERATING LICENSE NOS. NPF-87 AND NPF-89 DOCKET NOS. 50-445 AND 50-446 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT l

2-2 2-2 2-5 2-5 2-6 2-6 2-9 2-9 2-10 2-10 2-11 2-11 3/4 2-12 3/4 2-12 B 3/4 2-4 B 3/4 2-4 B 3/4 2-6 8 3/4 2-6 6-21 6-21 6-21a 6-22 6-22 E

t l

p

4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERKAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,y) shall not exceed the limits shown in Figure 2.1-1.

APPLICABILITY: MODES I and 2.

ACTION:

I Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour, and comply with the require-ments of Specification 6.7.1.

l l

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

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

ACTION:

MODES I and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.7.1.

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.

l l

COMANCHE PEAK - UNITS 1 AND 2 2-1

?

670 i

i f

N UW CCEPTABLE N_ P =

2385 PSIG OPERATIO 1 650 N

N P = 223.5 PSIG N

640 N

\\

630 l

N i

l P W 1985 RSIG N

g 620

\\

\\

\\

e m5 estG N

\\

k 610 4N

\\

s N

v>

o 600

% D, ACCEPTABLE 580 l

I 570 560 l

550 l

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

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

TABLE 2.2-1 8

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Fi EE TOTAL SENSOR ALLOWANCE ERROR

,,j' FUNCTIONAL UNIT (TA) 1 (S)

TRIP SETPOINT ALLOWABLE VALUE l.

Manual Reactor Trip N.A.

N.A.

N.A.

N.A.

N.A E

t1 2.

Power Range, Neutron Flux

<n a.

High Setpoint 7.5 4.56 1.25 s109% of RTP*

$111.7% of RTP*

E b.

Low Setpoint 8.3 4.56 1.25 525% of RTP*

s27.7 of RTP*

3.

Power Range, Neutron Flux, 1.6 0.5 0

$5% of RTP* with s6.3% of RTP* with High Positive Rate a time constant a time constant 22 seconds 22 seconds 7'

4.

Power Range, Neutron Flux, 1.6 0.5 0

$5% of RTP* with

$6.3 of RTP* with 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*

c: c:

EL EL Neutron Flux rr 5

]3 ]"

6.

Source Range, Neutron Flux 17.0 10.01 0

$10 cps s1.4 x 10' cps

!TlT 7.

Overtemperature N-16

a. Unit I-10.53 6.70 1.0+1.10+

See Note 1 See Note 2

0. 76(U
b. Unit 2 10.0 6.75 1.0+1.38+

See Note 1 See Note 2 0.96 z) t E. E gm E!

  • RTP = RATED THERMAL POWER (1) 1.0% span for N-16 power monitor, 1.10% for T,,g RTDs and 0.76% for pressurizer pressure sensors.

(2) 1.0% span for N-16 power monitor, 1.38% for T,,tg RTDs and 0.96% for pressurizer pressure sensors.

l

. ~..

g TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 9

TOTAL SENSOR A

ALLOWANCE ERROR 3E FUNCTIONAL UNIT (TA)

I (S)

TRIP SETPOINT ALLOWABLE VALUE 8.

Overpower N-16 4.0 2.05 1.0+0.05(3) sll2% of RTP*

s114.5% of RTP*

d 9.

Pressurizer Pressure-Low a.. Unit 1 4.4 0.71 2.0 21880 psig 21863.6 psig

b. Unit 2 4.4 1.12 2.0 21880 psig 21863.6 psig g

10.

Pressurizer Pressure-High m

a. Unit 1 7.5 5.01 1.0 52385 psig

$2400.8 psig

b. Unit 2 7.5 1.12 2.0 s2385 psig

$2401.4 psig 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 CC 11 12.

Reactor Coolant' Flow-low

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 FF minimum measured minimum measured l

3@

fl ow* *

  • flow ***

&R a

(3) 1.0% span for N-16 power monitor and 0.05% for T,,tg RTDs.

&F RTP = RATED THERMAL POWER.

Loop design flow - 99,050 gpm l

]

      • Loop minimum measured flow - 98,500 gpm 6

4 w

e2---,,

r

.w, w

--c,

+y e

e

-.w, e,-v---

v

-.*w-

-w-

=w

-w&wv-r---+-r

-n e

4 v.,,e--m-

,4-*-.

v~--uw-

.-e.

Ej TABLE 2.2-1 (Continued) jE TABLE NOTATIONS 5

pi NOTE 1: Overtemperature N-16 g[1+7s T,-T, ] + K3 (P-P') - f (aq) m 1

9 N

K -K i

3 rs -

1+rsa t

5

{j Where:

N Measured N-16 Power by ion chambers, T,

Cold leg temperature, 'F, 3!

c' T*

560.5'F for Unit 1, 560.3*F for Unit 2 - Reference T, at RATED THERMAL POWER, N

K 1.150, i

K 0.0134/*F for Unit 1

=

g

-0.016856/*F for Unit 2 1+rs The function generated by the lead-lag controller for i

92 1+T52 T, dynamic compensation, e

Time constants utilized in the lead-lag controller for 7,

r 3

g T,, r, 210 s, and 72s3s, 0.000719/psig for Unit I K

3 0.000898/psig for Unit 2 c: c:

hk N-I I ea RR' aa no oo O

%d t

-r

..--..-.,e,

.e,w.**,.-.+-,.ib.~.....,.-.-

,r-

-a n.

.w.~

..v.-

e.r,..--

m.v

-,,u.,

,.ve.,-

er ww

.m-

--e

y TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) g NOTE 1:

(Continued)

']

P Pressurizer pressure, psig,

=

E P'

2 2235 psig (Nominal RCS operating pressure),

[

S Laplace transform operator, s",

=

EE deteck(ors)of the power-range neutron ion chambers; with gains to be selected based on and f aq is a function of the indicated difference between top and bottom halves of measured instrument response during plant STARTUP tests such that:

^

For Unit 1 (i) for q - a between -65% and +4%, f (aq) = 0, where q and q are percent l-3 RATED,THERhl POWER in the top and bottom halves of the cor,e respectively,

[

and q, + q3 is total THERMAL POWER in percent of RATED THERMAL POWER, o

(ii) for each percent that the magnitude of q

-q exceeds -65%, the N-16 Trip Setpoint shall be automatically reduced by 1.81% of its value at RATED THERMAL POWER, and c c:

(iii) for each percent that the magnitude of q. - o exceeds +4%, the N-16 Trip 11 Setpoint shall be automatically reduced by 2.36% of its value at RATED

[

THERMAL POWER.

aa it 3a n

O O.

~

4 O

v-v._,.

.,.m,.,-

y.

y

.c u

.y

,_,,,,-----w.,.

,..,y c.-..,.,

-,.,,,,_w...,

.r.,

. ~.

e TA 05 ION n d)

.my NOTE 1:

(Continued) l

. y For Unit 2 7:

(1) for q - q, between -52% and +5.5%, f (aq) = 0, where q are percent RATED, THERMAL POWER in the top and bo,ttom halves of the, and q3 i

core respectively, si; and q, + qb i s total THERMAL POWER in percent of RATED THERMAL POWER, (ii) for each percent that the magnitude of q - qe exceeds'-52%, the N-16 Trip a

Setpoint shall be automatically reduced by 2.15% of its value at RATED THERMAL G

POWER, and ro (iii) 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.

';o NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.51%

l of span for Unit 1 or 2.88% of span for Unit 2.

EE 23

~~

I 4 ea aa n

O.

O Sw'w'**w may'mW-awwwme&W 9vgw9-f*s"-'

49 WT v'v*mg-d

  • "'un W"W WW'T* vat'N*f'81 9*t9"*'T*
    • W*"We' ur W'MT'*
    • 'FY 9 ige-r g9-

,p&-e w +-

w g

,-w-,ie,w-y-ig,w+-r.

'-3 y-9'99,wwmwr' r

'%=

yy*

p'

-w1"-'="Tw-ai'Wt'--9,-f 9-v vve--"-

T'-$-w=*'-v*v7tw' r-ir'-ivw-v V-'

v

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

Calculating the ratio at least once per 7 days when the alarm is a.

OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO 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 either:

Using the four pairs of symmetric thimble locations or a.

b.

Using the Movable Incore Detection System to monitor the QUADRANT POWER TILT RATIO.

L 6

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

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS l

LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the i

stated limits:

a.

Indicated Reactor Coolant System T 5 592*F y

b.

Indicated Pressurizer Pressure 2 2219 psig*

l c.

Indicated Reactor Coolant System (RCS) Flow 2 403,400 gpm** for Unit 1 l

2 395,200 gpm** for Unit 2 APPLICABILITY: MODE 1.

I 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 THERMf'. 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 f

4.2.5.1 Each of the above parameters shall be verified to be within it:; 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.

l 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 i

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 75% of RATED l

THERMAL POWER. The feedwater pressure and temperature, the main steam pres-sure, and feedwater flow differential pressure instrumen2 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 I

RATED THERMAL POWEP,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. 44,21 Unit 2 - Amendment No. 7

100 I

I 90 g

I I

BO tr 1

i N

l TARGET FLUX 1

DIFFERENCE O 70 l

_J a

v tr 6 0 j

wI I

i o 50 W

l e

1 40 b

I 1

s 6 30 i

M l

W l

IL 20

[

1 i

I 10 l-O

-30%

-20%

-10%

O I0%

20%

30%

INDICATED AXIAL FLUX DIFFERENCE FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER i-COMANCHE PEAK - UNITS 1 AND 2

.B 3/4 2-3

~

4 i

POWER DISTRIBUTION LIMITS

[

i BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) i i

and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel j

clad temperature will not exceed the 2200"F ECCS acceptance criteria limit.

l t

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 i

demand position e

b.

Control rod groups are sequenced with overlapping groups as described

{

in Specification 3.1.3.6; c.

The control rod insertion lin.:ts of Specifications 3.1.3.5 and 3.1.3.6 l

1 are maintained; and i

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

I NFAH will be maintained within its limits provided Conditions a. through N

d. above are maintained. The relaxation of FaH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion i

limits.

Fuel rod bowing reduces the value of the DNB ratio. Credit is available to

[

offset this reduction in the generic margin. The DNBR generic margin, totaling 18.1% for Unit I and 10.1% for typical cells and 9.5% for thimble cells for l

Unit 2 for DNBR completely offset any rod bow penalties. The margin for Unit 1 l

l and 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

l safety analysis limit DNBR.

[

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

i t

i

[

i COMANCHE PEAK - UNITS 1 AND 2 B 3/4 2-4 Unit 1 - Amendment No. 4,44,21 l

j Unit 2 - Amendment No. 7 t

f i

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a l

3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor Fg(Z) is measured periodically and in-creased by a cycle and height dependent power factor appropriate to Constant Axial Offset Control (CAOC) operation, W(Z), to provide assurance that the limit on the heat flux hot channel factor, Fg(Z), is met. W(Z) accounts for the effects of normal operation transients within the AFD band and was determined from expected power control maneuvers over the range of burnup conditions in the core.

The W(Z) function is provided in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.6.

N When FAH is measured, an adjustment for measurement uncertainty must be included for a full-core flux map taken with the Incore Detector Flux Mapping System.

g Fn(Z) should t,9 measured with the reactor core at, or near, equilibrium conditTons. Therefo m, the effects of transient maneuvers, such as power increases, should be penitted to decay to the extent possible while assuring that flux maps are taken i: accordance with the specified surveillance schedules.

j 3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

[

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identification and correction of a dropped or mis-aligned control rod.

In the event such action does not correct the tilt.

the margin for uncertainty on Fg is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

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

i I

i POWER DISTRIBUTION LIMITS j

BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parame-i ters are maintained within the normal steady-state envelope of operation as-l 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 (conservatively rounded to 592*F) and the Unit 1 indicated,y, pressurizer pressure value of 2219 psig correspond to analytical limits of 594.7'F and 2205 psig l

respectively, with allowance for measurement uncertainty. The Unit 2 indicated T,y, value of 592.8*F (conservatively rounded to 592*F) and the Unit 2 indicated pressurizer pressure value 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.

l The 12-hour periodic surveillance of these parameters through instrument

[

F readout is sufficient to en'sure that the parameters are restored within their limits following load changes and other expected transient operation, and to detect any significant flow degradation of the Reactor Coolant System (RCS).

The additional surveillance requirements associated with the RCS total flow i

d 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.

i Performance of a precision secondary calorimetric is required to precisely determine ine RCS temperature. The transit time flow meter, which uses the N-16 i

system signals, is then used to accurately measure the RCS flow.

Subsequently, the RCS flow detectors (elbow tap differential pressure detectors) are t

normalized to this flow determination and used throughout the cycle.

I L

t t

i

\\

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 2-6 Unit 1 - Amendment No. 4,6,21 Unit 2 - Amendment No. 7 l

e

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) 5). WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F SURVEIL-o LANCE TECHNICAL SPECIFICATION," June 1983 (H Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor (W(z) surveillance requirements for F Methodology). )

a 6). WCAP-10079-P-A, "NOTRUMP, A N0DAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," August 1985, (W Proprietary).

l T

7). WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 1985, H Proprietary).

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

l 1

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 -

l Heat Flux Hot Channel Factor.)

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

January 1989.

11). RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",

December 1990.

l I

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

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

\\

Electric Station Licensing Applications", February 1991.

14). RXE-91-002, " Reactivity Anomaly Events Methodology", May 1991.

(Methodology for Specification 3.1.1.3 - Moderator Temperature i

Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 -

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

-l Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

15). RXE-90-007, "Large Break Loss of Coolant Accident Analysis Methodology", December 1990.

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

Reference 17) is for Unit 1 only:

l l

17). WCAP-9220-P-A, " WESTINGHOUSE ECCS EVALUATION MODEL, February 1978 Version," February 1978 (M Proprietary).

COMANCHE PEAK - UNITS 1 AND 2 6-21 Unit 1 - Amendment No. 4,6,44,48,21 Unit 2 - Amendment No. 4,7

ADMINISTRATIVE CONTROLS i

i CORE OPERATING LIMITS REPORT (Continued)

Reference 18) is for Unit 2 only:

18). WCAP-9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION MODEL-1981 Version", February 1982 (W Proprietary).

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 f

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.

i 1

?

I l

f i

b i

i.

l-l i

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

Unit 2 - Amendment No. 7

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, special reports shall be submitted to the Regional Admin-istrator of the Regional Office of the NRC within the time period specified for each report.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level; b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety; c.

All REPORTABLE EVENTS; i

d.

Records of surveillance activities, inspections, and calibrations required by the Technical Specifications, Technical Raquirements Manual, and Fire Protection Report, except as explicitly covered in Specification 6.10.3; e.

Records of changes made to the procedures required by Specification 6.8.1; f.

Records of radioactive shipments; g.

Records of sealed source and fission detector leak tests and results; and h.

Records of annual physical inventory of all sealed source material of record.

6.10.3 The following records shall be retained for the duration of the unit Operating License:

a.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; b.

Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; c.

Records of radiation exposure for all individuals entering radiation control areas; d.

Records of gaseous and liquid radioactive material released to the environs; COW.ANCHE PEAK - UNITS 1 AND 2 6-22 Unit 1 - Amendment No. I

-

Retrieved from "https://kanterella.com/w/index.php?title=ML20058C862&oldid=4475599"