ML20034E708

Southern Nuclear Operating Company Jm Farley Nuclear Plant Startup Test Rept Unit 1 Cycle 12

ML20034E708
Person / Time
Site:	Farley
Issue date:	02/22/1993
From:	Woodard J SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9303010227
Download: ML20034E708 (17)

Text

.... _. _ _-

Southern Nuclear Operating Company Post Offce Box 1295 Birmingham, Alabama 35201 l

Telephbne 205 868-5086 A

I g

h Southern Nudear Operating Company-J. D. Woodard

' Vee Presdent thesouthem electnc system 4

Farley Project February 22, 1993 Docket No. 50-348 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant Unit 1 Cycle 12 - Startuo Report j

l I

Gentlemen:

t Enclosed is the Startup Report for Unit 1 Cycle 12.

If you have any l

questions, please advise.

.f Respectfully submitted, l

%Qda J.

Woodard i

EFB:cht-Ulcycll2.efb 4

NEL-93-0076 Enclosure l

I cc: Mr S. D. Ebneter Mr. S. T. Hoffman Mr. G. F. Maxwell 1

L g1GC r "c.4 o

i 7

l.

f-.

U 9303010227 930222

/

I DR ADOCK 05000348 j)~

v.-

J PDR I

,w w

e v

gr*

r SOUTHERN NUCLEAR OPERATING COMPANY

~

JOSEPH M. FARLEY NUCLEAR PLANT STARTUP TEST REPORT UNIT 1 CYCLE 12 TABLE OF CONTENTS EA.GE 1.0 Introduction 1

2.0 Unit 1 Cycle 12 Core Refueling 1

3.0 Control Rod Drop Time Measurement.

8 4.0 Initial criticality 10 1

10 5.0 All-Rods-Out Isotheanal Temperature Coefficient and Boron Endpoint 6.0 control and Shutdown Bank Worth 11 Measurements i

7.0 Power Ascension Activities 12 8.0 Incore-Excore Detector Calibration 13 9.0 Reactor Coolant System Flow 15 Measurement APPROVED:

1 i

> a lt Technical Manager j

General Manager - Nuclear Plant

-l C:\\WP51\\ FILES \\SUFORM

s

+

1.0 INTRODUCTION

The Joseph M.

Farley Unit 1 Cycle 12 Startup Test Report addresses the teste performed as required by plant procedures following core refueling.

The report provides a brief synopsis of each test and gives a comparison of measured parameters with design predictions, Technical Specifications, or values in the FSAR safety analysis.

Unit 1 of the Joseph M. Farley Nuclear Plant is a three loop Westinghouse pressurized water reactor rated at 2652 MWth. The unit began commercial operations on December 1, 1977.

The Cycle 12 core loading consists of 15717 x 17 fuel assemblies, of which 105 are Westinghouse Low Parasitic (LOPAR) assemblies and the remaining 52 assemblies represent the first phase of transition to Westinghouse Vantage 5 fuel.

All thimble plug and burnable poison inserts have been removed from the Cycle 12 core and two new double encapsulated secondary source inserts were added to be activated for use in cycle 13. The new sources provide compatability with Vantage 5 fuel and the double encapsulation gives an additional margin of protection against source material leakage into the reactor coolant. The burnable absorber inserts have been supplanted by Westinghouse Integral Fuel Burnable Absorbers (IFBAs) incorporated in the Vantage 5 assemblies. The design burnup capability of the Cycle 12 core is 16000 MWD /MTU.

Previous Cvele Comoletion Dates and Averace Burnuos Date Start of EOL EOL Burnup EOL Burnup Total Cycle critical Cvele Date (MWD /MTU)

(EFPD)

EFPY l

08-09-77 08-18-77 03-08-79 15450 420.60 1.152 2

10-31-79 11-04-79 11-07-80 10177 276.70 1.910 3

03-25-81 04-03-81 09-10-81 5180 140.70 2.296 4

03-03-82 03-07-82 01-14-83 10622 288.10 3.085 5

03-28-83 03-30-83 02-10-84 11096 301.30 3.911 6

04-22-84 04-24-84 04-06-85 12238 333.58 4.825 7

05-26-85 05-27-85 10-03-86 17231 470.04 6.112 8

11-30-86 12-02-86 03-25-88 16190 443.26 7.326 9

05-20-88 05-21-88 09-23-89 17456 479.29 8.639 10 11-08-89 11-10-89 03-08-91 16910 464.17 9.911 11 05-18-91 05-21-91 09-25-92 17513 480.26 11.227 2.0 UNIT 1 CYCLE 12 CORE REFUELING REFERENCES 1.

Westinghouse Refueling Procedure FP-ALA-Ril.

2.

Westinghouse WCAP 13434 (The Nuclear Design and Core Management of the Joseph M. Farley Unit 1 Power Plant Cycle 12)

Unloading of the Cycle 11 core into the spent fuel pool commenced on 10-05-92 and was completed on 10-06-92 with no significant problems or delays. During the of fload, each fuel assembly was inspected with bino-culars for indications of damage or other problems.

No visual defects were noted.

Radiochemistry detection of an iodine peak following a reactor trip approximately a year before the end of cycle shutdown indicated the probability that the Cycle 11 core contained leaking fuel assembles.

Therefore, following core unload, each fuel assembly removed from the 1

e f

1 4

Cycle 11 core was subjected to _ ultrasonic leak testing (UT).

The UT program identified three leaking fuel assemblies (2B48, 2B30 and 2A47).

f In addition, high magnification TV examinations of the leaking assemblies disclosed that rod Q5 of assembly 2B30 had a hydride blister on the side and that the top end plug was displaced upward and to the side, leaving I

a visible opening. No visible defects could be found during the TV in-spection of assemblies 2B48 and 2A47.

Assemblies 2B30 and 2A47 were scheduled for discharge into the spent fuel pool for storage, but leaking assembly 2B48 was scheduled for reload into the cycle 12 core.

There-fore, 2B48 was replaced with assembly 2A14 in the cycle 12. core design.

[

f The cycle 12 Core reload commenced on.11-3-92 and was completed on 11-5-92.

The as-loaded, redesigned Cycle 12 core is shown in Figures 2.1

(

through 2.5

+

l I

9 i

I 9

i s

I 1

l 1

i 2-

. 1

i 4

FIGURE 2.1: UNIT 1 CYCLE 12 REFEREFCE LOADING PATTERN R

P N

M L

K J

H G

F E

D C

B A

2A58 2B32 2A68 1

2855 2004 2D31 2D05 2D34 2C11 2B54 2

R14 R44 R13 2B57 2C39 2D36 2005 2D14 2C15 2D42 2C61 2B56 3

l R05 ROI 2B58 2D01 2D46 2C27 2C60 2C28 2C46 2C21 2D48 2D02 2B46 I

4 R47 R03 SS05 R29 R35 2B52 2C47 2D49 2B41 2D16 2831 2D09 2B37 2D17 2B22 2D50 2CSI 2B50 5

RU M6 2001 2D43 2C19 2D22 2C10 2C34 2C50 2C45 2C31 2D23 2C18 2D30 2C12 6

R21 R42 R02 R45 R07 R36 R19 2A63 2D44 2C13 2C59 2B18 2C58 2D25 2C55 2D26 2C38 2B05 2C40 2C02 2D32 2A55 7

R12 R4i R09 R38 2B04 2D06 2D13 2C08 2D10 2C36 2C53 J24 2C48 2C56 2D11 2C22 2D24 2D07 2B23 8

R24 SSO9 R16 R34 SS10 R48 2A62 2D39 2C30 2C54 2827 2C63 2D27 2C43 2D28 2C37 2838 2C41 2C25 2D40 2A61 9

R39 R06 R22 R45 2C29 2D37 2C17 2D15 2C06 2C57 2C49 2C(2 2C14 2D20 2C20 2D3R 2007 10 RIS A*>7 R25 R30 R04 RIO R17 2B59 2CS2 2D51 2Bb; 2D21 2B09 2D12 2B35 2D18 2B39 2D45 2C64 2B53 11 R15 R28 2B51 2D03 2D52 2C16 2C42 2003 2C35 2009 2D47 2D04 2 861 12 R27 R08 SS06 R20 R32 21W9 2C44 2D33 2C32 2D19 2C26 2D35 2C33 2A14 13 R31 R33 4 North 2147 2C23 2D29 2D08 2D41 2 '.' 4 2B60 14 R23 R40 R11

- Fuel Anembly Serial Number 2A67 2B20 2A42 i

15

- Insert Serial Number The Original w/o U-235 enrichments were:

No. of Fuel Assemblies:

Region 9A (J) assemblies..... 3.597 %

Region 9A.....1 Region 11 A (2A) assemblies... 3.805 %

Region 11 A....1 Region 11B (2A) assemblies... 4.207 %

Region 11B... 8 Region 12A (2B) assemblies... 3.803 %

Region 12A...16 Region 12B (2B) assemblies... 4.193 %

Region 12B...15 Region 13A (2C) assemblies... 3.801 %

Region 13A... 32 Region 13B (2C) assemblies... 4.195 %

Region 13B... 32 l

Region 14A (2D) assemblies... 3.800%

Region 14A... 28 Region 14B (2D) assemblies... 4.200%

Recion 14B... 24 Total 157 C:WP51\\TILI5WP'\\ CORE;:.DIA 3

t l

+

i FIGURE 2.2: Control and Shutdown Pod Locations R

P N

W L

K J

H G

F E

D C

B A

I l

l 1

6 i

l i

A D

A 2

SA SA SP 3

i i

C B

SP B

C 4

SP SB SP SB

-5 i

A B

D C

D B

A 6

SA SB SB SP SA-

-7 1

909 D

SP C

SP C

SP D

-8 1

SA SP SB SB SA

-9 r

A B

D C

D B

A 10 i

SB SP SB SP 11 j

C B

SP B'

C 12 SP SA SA 13 A

D A

14 ABSDRBER MATERIAL: AG IN-CD 15 00 l

~ BANK NUMBER OF BANK NUMBER OF l

lDENTIFIER LOCATIONS IDENTIFIER LOCATIONS l

A 8

SA 8

B 8

SB 8

i l'

C 8

SP (SPARE) 13 l

D 8

l 4

l

i FIGURE 2.3: Bur. sable Absorber and Source Assembly I.ocations R

R P

N W

L K

J H

G F

E D

C B

A j

ll 4

1 7-321 321 32I 2

32I 80I 32I 3

7 64I 4SSA 641 4

64I 801 64I 80I 64I 5

l 32I 80I 801 32I 6

321 104I 104I 32I

-7 900 32I 801 4SSA 641 641 4SSA 80I 32I

-8 32I 104I 104I 321

-9 32I 801 80I 32I 10 641 801 64I 801 641 11 641 4SSA 64I 12 321 SOI 321 13 i

32I 321 32I 14 15 00 TYPE TOTAL fill..(NUWBER OF IFBA R0DS).................

2784 lSSA..(NUWBER OF SECONDARY SOURCE RODLETS)...

16 Note - Locations M-8 and D-8 contain, new dually-compatible Secondary Sources for first time irradiation in Cycle 12 5

1 FIGURE 2.4: Secondary Source Pod Configurations i

i i

i I

I i

i I

i I

i l

a l

4 I

i l

l 1

I I

I i-I i

i i

i 1

i SEXNWW SotAtCE ASSESLY 4

i t

I k

I i

i i

4 4

6 t

1 i

FIGURE 2.5: Burnable Absorber Configurations j

j el i iii l l l

4 1 I i

'o I i i i ! I i i i iiii i

i,eiel e

i !al i to: le e.

I !eiel i i ei i i i !eie i

4etel iol lei I I el el e e-let i !al le i I el ei I ei

ei l l :e 101 iel it! I I el

'el l I el el :e! 1e l l ei j ;ci el el lel :el :01 eI iel :ei 'e to

ei !ei e i

el 19! l l iel 191 I I !G1 lei l ISI I I l

i i 1 el I i i ei :ei el tel e! iel eI i !G! 'e! e ici ;e! :ej e i i

I t ;el !e. to ici 'el ;e i

i i !el i i e ici I

ai i

l l

.el I !el lei i iel lei l IM jel I lel lei l l9 foi i lei l

I i ;el le :01 ;el tel S i

! lGI i e ie!

I !e1 i

l 1 to e

et !Oi ici el e

'91 le

!G' tel l0 iej it i i

i

!el GI l iel el l lel le I

el l ! l l lel l I ei o

!O el e!

!G'

!G

!ei :e 19

ei e i j

5 l :01 te t

el l l lo: :el !el is l l el jel I :e el el !al l l 91 i

e!al *el lel l I 'el el e el 101

01

e. l l tel lo! ! el i

l :e:el ;e: I tel i tel :o e I

iote I

e i

I 'G;e: I et i l l l l l i i! i e

1 i I

e I i

I ii i

104 IFBA, ASSEWBLY 80 IFBA ASSEWBLY l

i i ! t i i i i i i ! 1 I

I ii it I (

i 10I l l l

I el I i l 101 i t i I e

t I iI tei i tel lel lI el 101 el i le e' l l :el 4

f I

I i I i 1

i I i i i :e= ;e! 191 lel l l l

1 I :01 el., I :ei i i '91 eI to

e 'I

!G I l :01 :O I

i III i l lei l i I II!l l ;el l I l l l l

l l l

l,e! I I el 19! 'e i ! e le~

t I

e' !

i i :ei i el ici I ;e l

I e

!e i

-el I 01 lei l I el 101 Q le Iel I iel i ll I e lel ii ei !

ei ei i i 4e i ! I el is i

i i

i ei i i el :si :ei i i e

!si i l i i

le, t il 191 l lilIiIi el i I iI

!III l l t i ei ei i,el ! I ei e

i I ie

el

01 It is e1 i i l i !e

'ei 'ei tel l I i il l

'I III i

el 101 1 I :o le e

i ;e le

_le le el i i '91 i i !si i I

e i !

i !(, I !

i ; i i i 1 I i i i i t i i i i i i i

I 64 !/BA ASSEWBLY 32 IFBA ASSEMBLY 7

J

,o 3.0 CONTROL ROD DROP TIME MEASUREMENT (FNP-1-STP-112)

PURPOSE The purpose of this procedure was to measure the drop time of all full length control rods under hot full-flow conditions in the reactor coolant system to insure compliance with Technical Specification Requirements.

SUMMARY

OF RESULTS l

[

For the hot full-flow condition (Tavg > 541 "F and all reactor coolant l

pumps operating) Technical Specification 3.1.3.4 requires that the drop time from the fully withdrawn position shall be < 2.7 seconds from the beginning of stationary gripper coil voltage decay until dashpot entry.

All full length rod drop times were measured to be less than 2.7 seconds.

The longest drop time recorded was 1.85 seconds for rod B-6.

The rod drop time results for both dashpot entry and dashpot bottom are presented in Figure 3.1.

Mean drop times are summarized belows i

TEST MEAN TIME TO MEAN TIME TO CONDITIONS DASHPOT ENTRY DASHPOT BOTTOM Hot full-flow 1.616 sec.

2.200 ser..

To confirm normal rod mechanism operation prior to conducting the rod drop test, the Verification of Rod Control System Operability (FNP-0-ETP-3643) was performed.

In this test, the stepping waveforms of the sta-tionary, lif t and movable gripper coils were examined for unomalies, rod speed was measured, the functioning of the Digital Rod Position Indicator (DRPI) and bank overlap unit were checked, and the bank overlap unit switch settings and functions were verified to be correct.

i i

8

F FIGURE 3.1: UNIT 1 CYCLE 12 DRIVE LINE " DROP TIME" TABULATION R

P N

M L

K J

H G

F E

D C

B A

1 1.633 1.583 1.617 2

i 2.183 2.150 2.283 1.583 1.650 3

2.183 2.183 1.633 1.617 1.620 1.617 l

2.217 2.182 2.196 2.200 1.600 1.633 5

2.183 2.250 1.617 1.643 1.617 1.567 1.567 1.570 1.lL40 6

2.167 2.200 2.200 2.150 2.150 2.167 2.517 1.617 1J50 1.583 1.650 j

7 2.200 2.183 2.167 2.200 l

1.617 1.550 1.567 1.700 1

8 4

2.200 2.083 2.117 2.317 1.617 1.533 1.550 1.617 9

I 2.217 2.150 2.133 2.200 1.617 1.582 1.617 1.600 1.600 1.612 1.683 i

10 2.183 2.183 2.200 2.133 2.183 2.233 2.250 l

1.600 1.617 l

11 l

2.167

~

2.150 f

1.567 1.548 1.617 1.ti33 12 l

2.167 2.133 2.185 2.233 l

l 1.600 1.667 t

13 2.167 2J83 4 North 1.733 1.633 1.667 14 2.367 2J1?

2.250 l

15 l

l TEMPERATURE -

547.667 PRESSURE -

2220 osic

%FIDiJ -

100

+- BREAKER " OPENING" TO DASHPOT ENTY - IN SECONDS

• BREAKER "0PENING" TO DASHPOT BOTTOM - IN SECONDS DATE:

11-28-92

'C:\\WP51\\ FILES \\WSM\\ CORE 1.DIA 9

b t

i I

e i

l

=

d i

4.0 INITIAL CRITICALITY (FNP-0-ETP-3601)

PURPOSE i

The purpose of this procedure was to achieve initial crit.ver.lity under carefully controlled conditions, establish the upper flux limit for the conduct of zero power physics tests, and operationally verify the calibration of the reactivity computer.

l

SUMMARY

OF RESULTS l

Initial reactor criticality for cycle 12 was achieved during dilution mixing at 1325 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.041625e-4 months <br /> on November 29, 1992.

The reactor was allowed to j

stabilize at the following conditions:

l RCS Pressure 2230.0 psig l

RCS Temperature 546.5 'F l

Intermediate Range Power 1.49 x 10' Amp RCS Boron Concentration-1916.5 ppm Bank D Position 184.5 steps Once criticality was achieved, the ptint of adding nuclear heat was determined in order to define the flux r. nge for physics testing, and the reactivity computer calibration was verified by making positive and negative reactivity changes and comparing the reactivity indicated by the reactivity computer with values determined using the Inhour Equation.

5.0 ALL-RODS-OUT ISOTHERMAL TEMPERATURE COEFFICIENT AND BORON ENDPOINT (FNP-0-ETP-3601)

PURPOSE The objectives of these measurements were to determine the hot, zero power isothermal and moderator temperature coefficients for the all-rods-out (ARO) configuration and to measure the ARO boron endpoint concentration.

SUMMARY

OF RESULTS The ARO, hot zero power teeperature coefficients and the ARO boron endpoint concentration are tabulated below:

ARD, HZP ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT l

Boron Measured ITC Design Acc. Calculated i

Rod Conc.

ITC Criterion MTC Confiouration vom ocm/*F pcm/*F pcm/*F All Rods Out 1949.0

+1.15

+0.92 1 2

+3.40*

10 l

- n

o

+

i where I

ITC = Isothermal Temperature coefficient, includes -1.96 pcm/*F Doppler coefficient MTC = Moderator Temperature Coefficient, corrected to the ARO f

condition

• MTC result was normalized to all rods out (ARO) and to the ARO critical.

f boron concentration _(1919 ppm).

t ARO, HZP BORON ENDPOINT CONCENTRATION

[

f Rod Confiouration Measured C.

foom)

Desion-oredicted C.

fromi All Rods Out 1952.0 1919 150 Since the measured MTC (+3.40 pcm/*F) was less positive than the Tech-nical Specification limit of +7.0 pcm/*F, no rod withdrawal limits were l

required.

The design review criterion.for the ARO boron concentration

[

was also satisfied.

}

6.0 CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS (FNP-0-ETP-3601)

PURPOSE The objective of the - bank worth.- measurements was - to determine the integral reactivity worth of each control and shutdown ' bank for comparison with the values predicted by design.

SUMMARY

OF RESULTS i

The rod worth measurements were performed using the bank interchange method'in whicht (1) the worth of the bank having the highest design j

worth (designated as the " Reference Bank") is carefully measured using the standard dilution method; then (2) the worths of tne remaining con-i trol and shutdown banks are derived from the change in the reference bank -

reactivity needed to offset full' insertion of the bank being measured.

For Cycle 12, control bank D was the reference bank. The measured bank worths satisfied the review criteria both for the banks measured indi-I vidually and for the total worth of all banks combined.

l

SUMMARY

OF CONTROL AND SHUTDOWN BANK WORTH MEASUREMENTS Control or Predicted Bank Shutdown Worth & Review

-Measured Bank Percent Bank Criteria focm)

Worth focm)

Difference A

296 1 100 314.7

, + 6.'3 2 l -

B 1122 1 168 1106.9

-1.35 l'

C 1017 1 152 973.2

-4.31 D (Ref.)*

1174 1 117

'1138.9

-2.99 SD - A 902 1 135 930.7

+3.18 SD - B 1052 1 158 972.8

-7.53 l

All Banks 5563 1 556.3 5437.2

-2.26 l

"The reference bank worth was measured by the dilution method.

4 11 i

-)

l

7.0 POWER ASCENSION ACTIVITIES Upon completion of HZP physics tests, the following activities were performed during power ascension, or at 100% power:

1.

Incore movable detector system alignment.

2.

Calorimetric thermal power measurement and adjustment of the power range NIS channel percent power indications.

3.

Measurement of NIS intermediate range channel currents in order to determine high flux trip and rod stop setpoints.

4.

Incore-excore AFD channel recalibration.

5.

Core hot channel factor surveillance.

6.

Reactor coolant system flow measurement.

7.

Rescaling OPAT and OTAT protection loops to the 100% loop ATs measured during the RCS flow test.

[

At approximately 10% - 12% power, the determination of the incore system core limit settings (FNP-1-ETP-3606) was performed. The purpose of this procedure is to align the system so that the movable detectors stop at the correct core heights during flux mapping.

In order to invoke Technical Specification 3.10.3 test exceptions for HZP physics tests, preliminary intermediate and power range trip setpoints of less than or equal to 25% power were used for initial reactor startup and physics testing.

In addition, since both NIS intermediate range detectors were replaced, the preliminary N35 and N36 channel trip setpoint and rod stop currents were set to 80% of the corresponding Cycle 11 setpoint currents.

Following the completion of physics tests, the NIS power range high range high flux trip setpoint was increased to 80% to allow power escalation above 25%.

The 80% setpoint (vice 109%) was administratively imposed to address the possibility that the power range channels initially could be indicating nonconservatively.

Due to the projected reduction in core neutron leakage between Cycle 11 and cycle 12, it was recommended that power ascension be limited to 22.75% indicated power prior to performing the first thermal power measurement in order to prevent inadvertently ex-ceeding 35% power. Therefore, a thermal power measurement was performed prior to increasing power above 22% and the NIS PR channels percent power indications were adjusted.

At 30% power, the thermal power measurement was repeated, the power range channels were recalibrated and currents were measured for determining the intermediate range high flux trip and rod stop setpoints.

Then, the reactor was ramped to 33%, the Incore-Excore test (described in Par. 8.0) was performed and the power range N41 - N44 delta flux channels were recalibrated. Following calibration'of the delta flux channels, a full-core flux map was performed at equilibrium xenon conditions at 33% power for core hot channel factor surveillance, and the power range NIS high l

flux trip setpoint was increased from 80% to 109%.

Subsequent power escalation to 99% proceeded smoothly with no high quadrant power tilt ratio (QPTR) indications.

I 12 l

j i

i l

i l

I l

i I

I At approximately 99% power, a Loop 3 Overpower AT rod stop alarm was received.

In addition, the percent AT channels had been indicating approximately 2% higher than NIS (calorimetric) power. Therafore, power escalation was stopped at 99%, the RCS flow test (described in Par. 9.0)

{

l was performed and the AT protection channels (OPAT and OTAT) were 1

rescaled to the AT values (normalized to 100%) measured during the flow test.

As summarized in Table 7.1, core hot channel factor surveillance was initially performed under non-equilibrium conditions using the incore-excore base case full core flux map taken at 33% power, and then under equilibrium conditions using full-core flux maps performed at 33% and 100% power.

TABLE 7.1

SUMMARY

OF POWER ASCENSION FULL CORE FLUX MAP DATA Parameter Fuel Tvoe Mao 287 Map 294 Eao 295 Avg. % power N/A 33%

33%

100%

Max FDH Lopar 1.4575 1.4560 1.4457 vantage 5 1.6273 1.6279 1.5570 Max power tilt' N/A 1.0090 1.0070 1.0062 Avg. core % A.O.

N/A

+7.959

+8.365

+5.023 Limiting FQ(Z)~

Lopar 1.9844 2.0312 1.8184 vantage 5 2.2557 2.2634 1.9347 FQ Limit Lopar 4.5364 4.5571 2.2779 Vantage 5 4.8453 4.8015 2.4056 Flux map N/A Non-equi-Equi-Equi-conditions librium librium librium i

" Calculated power tilts based or; assembly FDHN from all assemblies.

~ Based on percent to FQ limit.

j

\\

Fuel types refererced above are low parasitic (Lopar) fuel (105 assemblies),

and Vantage 5 fuel (52 assemblies).

8.0 INCORE-EXCORE DETECTOR CALIBRATION (FNP-1-STP-121)

PURPOSE The objective of this procedure was to determine the relationship between power range upper and lower excore detector currents and axial offset for the purpose of calibrating the control board and the plant computer arial flux difference (AFD) channels, and for calibrating the delta flux pen-alty input to the overtemperature delta-T protection system.

13

[,c l

i

SUMMARY

OF RESULTS At an indicated power of approximately 33%, a full core base-case flux map was performed at the AO (+7.959%) obtained immediately following power ascension. Five additional (quarter-core) flux maps were performed at various positive and negative axial offsets ranging from +24.6% to

-35.4% in order to develop equations relating detector current to incore axial offset.

(A sixth quarter-core map was performed, but the data was lost due to a plant computer malfunction.)

In addition, data was taken from intermediate range channels N35 and N36 to determine the ef fects of Bank D insertion on the detector currents. Prior to ascending above 33%

power, the power range NIS channels were adjusted to incorporate the re-vised calibration data.

During the refueling outage preceding the cycle 12 startup, the original analog power range channel detector current meters were replaced with permanently installed digital meters on all channels (N41 - N44).

The digital meters enhanced the accuracy and precision of detector current readings and reduced the error in the incore-excore test.

As a result, the excore quadrant power tilt ratio (QPTR) remained well within its t

l limits during the ascension to full power and, at 100% power, the maximum QPTR was only 1.0037.

In addition, an evaluation of the incore-excore calibration performed at 100% power using flux map data showed that the 33% power calibration was still within acceptable tolerance. Neverthe-l less, the detector equations were refined using full power flux map data and, once the core hot channel factors were verified to be satisf actory, the NIS delta-flux channels were adjusted to incorporate the revised calibration. The revised detector current vs AO equatione (ir which both the slopes and zero-offset currents were revised to account for core leakage changes between 33% and full power)~are tabulated below:

TABLE 8.1 DETECTOR CURRENT VERSUS AXIAL OFFSET EQUATIONS OBTAINED FROM INCORE-EXCORE CALIBRATION TEST CHANNEL N41:

I-Top 0.7718 *AO

+

145.97 uA

=

I-B: ;cm =

-0.8473 *AO

+

140.17 uA CHANNEL N42:

I-Top 0.7911 *AO

+

143.50 uA

=

I-Bottom =

-0.8752 *AO

+

136.69 uA CHANNEL N43:

0.7656 *AO

+

148.66 uA I-Top

=

I-Bottom

-0.9546 *AO

+

153.73 uA

=

CHANNEL N44:

I-Top 0.8105 *AO

+

142.63 uA

=

I-Bottom

-0.8786 *AO

+

139.44 uA

=

14

f l

9.0 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (FNP-1-STP-115.1)

PURPOSE The purpose of this procedure was to measure the flow rate in each reactor coolant loop in order to confirm that the total core flow met the minimum flow requirement given in the Technical Specifications. In addi-tion, the RCS loop 100% delta-T values measured during this test are used to evaluate and, if necessary, to rescale the OPAT and OTAT protection channels.

SUMHARY OF RESULTS The occurrance of a rod stop at 99% power and the disagreements noted between percent loop AT and NIS (calorimetric) percent power indicated the OPAT and OTAT protection loops required rescaling. Therefore, power escalation was stopped and the Unit 1 RCS flow measurement was performed at an average measured power of 99.7S%.

Since it was intended to use the data for rescaling the AT protection channels, 12 sets of flow test data l

(rather than the minimum of six) were taken to ensure accuracy.

In order to comply with the Unit 1 Technical Specifications, the total reactor coolant system flow rate measured at normal operating temperature l

and pressure must equal or exceed 267,880 gpm for three loop operation.

From the average of 12 sets of measurements, the measured RCS loop flows were:

Loop A = 94,476 gpm Loop B = 92,997 gpm Loop C = 94,431 gpm This gave a total measured core flow of 281,904 gpm, which satisfied the Technical Specification requirement.

The measured 100% loop ATs (normalized to 100.0% power) obtained during the RCS flow test were:

Loop A:

62.837

'F Loop B:

64.876 *F Loop C:

64.126 *F Scaling calculations were performed and the OPAT and OTAT protection channels were recalibrated to these revised 100% AT values.

)

i 15