abstract this report summarizes the results of the startup ... · 3.3 thermal power measurement 12...

INDIAN POINT UNIT NO. 2

STARTUP PHYSICS TESTrREP9RT

CYCLE, 5

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC. DOCKET NO. 50-247

AUGUST 1981

~. -~-x

810908007o7 810824 ( 7" PD ADc 05000247 P_____ PDR - ~--.~--'--~----7-

(' 'c (-,p

ABSTRACT

This report summarizes the results of the Startup Physics Tests

for Cycle 5 of the Indian Point Unit No. 2 reactor performed at

Hot Zero Power condition, and during reactor power level escala

tion. Results of these tests demonstrate adequate conservatism

of the Cycle 5 design for safe operation in accordance with the

Indian Point Unit No. 2 Technical Specifications.

Ai)

r,: ~

TABLE OF CONTENTS

PAGE8 NO.

ABSTRACT i TABLE OF CONTENTS ii LIST OF TABLES iii LIST OF FIGURES iv

1. INTRODUCTION

2. REACTOR CORE DESCRIPTION

2.1 REACTOR CORE CONTROL 4 2.2 REACTOR CORE INSTRUMENTATION 6 2.3 FUEL 6 2.4 CORE LOADING VERIFICATION 9

3. MEASUREMENT METHODS

3.1 REACTIVITY MEASUREMENT 10 3.2 CORE POWER DISTRIBUTION MEASUREMENT 12 3.3 THERMAL POWER MEASUREMENT 12

4. HOT ZERO POWER (HZP) TESTS

4.1 INITIAL CRITICALITY 13 4.2 END POINT BORON CONCENTRATIONS 16 4.3 RCC BANK DIFFERENTIAL AND INTEGRAL WORTHS 19 4.4 ISOTHERMAL TEMPERATURE COEFFICIENT 23 4.5 DIFFERENTIAL BORON WORTH 25

5 5. POWER TESTS

5.1- CORE POWER DISTRIBUTIONS 26 5.2 REACTOR COOLANT SYSTEM FLOW DETERMINATION 35

6. REACTOR INSTRUMENTATION CALIBRATION

38 6.1 EXCORE DETECTOR CALIBRATION 43 6.2 INCORE T/C AND WIDE RANGE RTD CALIBRATION 6.3 OVERPOWER AND OVERTEMPERATURE DELTA-T SETPOINTS 45

7. APPENDICES

A. CHRONOLOGICAL EVENTS 48 B. ACCEPTANCE CRITERIA 49

8. REFERENCES 50

(ii4

LIST OF TABLES

TABLE TITLE PAGE NO.

1.1 IP-2 Cycle 5 Startup Physics Test Program 2

3.1 Reactor Period and Reactivity Results 11

3.2 Delayed Neutron Data BOL Cycle 5 11

4.1 End Point Boron Concentrations 18

4.2 Control Rod Bank Integral Worth Summary 20

4.3 Isothermal Temperature Coefficients 24

5.1 Summary of Results of Incore Analysis, IP-2 Cycle 5 Startup Maps 27

5.2 Results of Reactor Coolant System Flow Determination at various Power Levels 37

6.1 Primary Loop Narr Range RTD Correction Factors as a Function of Tempe ture (OF) 44

6.2 Overpower And Overtemperature Delta-T 46 Setpoints

(iii)

LIST OF FIGURES

FIGURE

2.1

2.2

2.3

4.1

4.2

4.3

4.4

4.5

5.1

5.2

5.3

5.4

6.1

6.2

6.3

6.4

TITLE

Contol Rod Locations

In-Core Instrumentation, Thermocouples and Moveable Detectors

IP-2 Cycle 5 Core Loading Arrangement

Inverse Count Rate Ratio vs. Dilution Water

Inverse Count Rate Ratio vs. RCC Bank Position

Differential Rod Worth vs. Neutron Flux

Differential and Integral Bank D Worth

Differential and -Integral Bank C Worth

IP-2 Cycle 5 BOL and HZP Condition Power Map

IP-2 Cycle 5, 50% Power Map

IP-2 Cycle 5, 90% Power Map

IP-2 Cycle 5, 100% Power Map.

Excore Detector Current vs. Axial Offset Channel N-41.




PAGE NO.

5

7

8

14

15

17

21

22

28

30

32

34

39

40

41

42

(iv)

- .- . . Ii

I

Indian Point Unit No. 2 attained initial criticality

for Cycle 5 operation on May al, 1981. The startup

physics test Program described in Table 1.1 was con

* ducted. The objectives of this program were:

(a) To demonstrate that the core parameters dur

ing reactor operation 'would be within the

assumption of the accident analyses in FSAR

(Reference 1) and within Technical Specifi

cation (Reference 2) limits.

(b) To verify the nuclear design calculations.

(c) To provide the bases for the calibration of

reactor instrumentation.

Section 2 of this report gives a brief description of

the reactor core and Cycle 5 core loading. Section 3

deals with measurement methods used in startup tests.

In Section 4, results from Hot Zero Power (HZP) phy

sics tests are presented and in Section 5, physics

tests at different power levels up to 100% power are

described. Reactor instrumentation (excore detectors,

incore thermocouples (T/C) and resistance temperature de

tectors (RTDs)) calibration is treated in Section 6.

The test results reported here are compared against

Westinghouse nuclear design results (Reference 3).

-1-M

7.41 M

Table 1.1

Indian Point Unit No. 2 Cycle 5 StartupPhysics Test Program Outline

1. Pre-criticality Measurements

Incore Thermocouple and RTD calibrations.

2. Hot Zero Power (HZP), Xenon'Free Condition Tests

2.1 Initial Criticality

2.2 Isothermal Temperature Coefficient at control rod configurations given below.

(i) All Rods Out (ii) Control Bank D In

2.3 End Point Boron Concentrations for control rod configurations given below.

(i) All Rods Out (ii) Control Bank D In

2.4 Control Rod Worth (Integral and Differential) Measurements

(i) Control Bank D (ii) Control Bank C with Control Bank D In

3. Power Ascension Tests

3.1 Excore Detectors Calibration at 90% of reactor power.

3.2 Movable Incore Detector Flux Maps at power levels .<l%,-50%,--90% and- 100% of reactor power.

3.3 Reactor Coolant Flow Measurement.

3.4 Overpower and Over Temperature AT Setpoints.

2

-2-

The acceptance criteria for the Cycle 5 startup tests

are given in Appendix B. These criteria are used as

guidelines for interpreting startup test measurements.

- 3 -

2. REACTOR CORE DESCRIPTION

The Indian Point Unit 2 core consists of 193 fuel assem

blies of slightly enriched uranium dioxide. Each fuel

assembly contains 204 fuel rods with zirconium alloy

cladding, 20 rod'.cluster control (RCC) guide tubes for

inserting control rods, and a central instrument thimble.

Cycle 5 contains 24 fuel assemblies containing 12 burnable

poision (BP) rods each for a total of 288 fresh burnable

poison rods. Burnable poison rods are composed of a

borosilicate Pyrex glass. Burnable poison rods are in

serted in assemblies to provide a negative moderator

temperature coefficient during reactor operations, to

control excess reactivity in the beginning of life, and

to improve the power distribution. Installation of two

primary neutron sources was required due to a projected

long outage during Cycle 4/5 refueling. Two primary

neutron sources at locations G-2 and J-14 and two 'sec

ondary neutron sources at positions H-3 and H-13 are

added to the loading to maintain.a minimum count rate

during core loading.

2.1 Reactor Core Control

In addition to the chemical shim control by boric

acid, dissolved in the coolant water, control and

shutdown of the reactor is accomplished by 53

full-length Rod Cluster Control Assemblies (RCCA's).

The -latter consist:of four control and four shut

down banks. Figure 2.1 is an X-Y cross section of

the reactor core describing-RCC bank positions. -4-

2115 14 13 12 11 10 9 8 7

1800

2700

SYMBOL

0 0 0

,V 0 C)

NUMBER OF ROD CLUSTERS

9

• 6

8

Figure 2.1 Control Rod Locations

-5-

p

BANK SA • SB

sci SD $C A

D

6 5,

2. REACTOR CORE DESCRIPTION (Cont'd)

2.2 Reactor Core Instrumentatilon

The reactor core instrumentation consists of

excore neutron detectors, six movable incore

detectors (M/D) capable of scanning up to 50

fuel assemblies through4 the central instru

ments thimble and 65 incore thermocouples

(T/C) to monitor exit coolant temperatures.

Figure 2.2 shows the incore instrumentation.

2.3 Fuel

Cycle 1 through Cycle 4 were operated with a

high-parasitic (HIPAR) fueled core. Beginning

with Cycle 5, the first transitional cycle, the

core was refueled with a low-parasitic (LOPAR)

fuel region. The Cycle 5 core contains 125

HIPAR fuel assemblies and 68 LOPAR fuel assem

blies;, (See Figure 2.3). The active core height

of the HIPAR assemblies is approximately 142

inches while that of LOPAR assemblies is ap

proximately 144 inches. The RCC guide thimbles

and instrumentation thimbles in HIPAR fuel are

of type 304 stainless steel welded to the In

conel-718 spring clip grids (9 per assembly)

and to the top and bottom assembly nozzles.

In the LOPAR fuel, the RCC guide thimbles

and instrumentation thimbles are Zircaloy-4,

* -6-

L

j

J

r

D

C

THERMOCOUPLE

E~1

MOVABLE DETECTOR

-FLOW-mIXING DEVICE

FIXED DETECTOR,

Figure 2-2 In-Core .Instrumentation and Movable Detectors

ThermOcouPles

-- 7"

I.

Figure. 2.3

A 8 C D E

I607 G37 da3 4 32g

SI !.337 334 iZP114 I 655 DO? FI6 D64 .343 R0687 R51

.. 04 F25 F-38 FI1 I 304 118 RZ8 42

GI 6-5 034 F44 E49 64( E31 F07 652 662 E38 F29 P0 46 4 9 ( 0 3 312 16epl3 R37 68, / |2PIO. !06 13 RIT I2Ptio 210 108 R43 1,Pt 303

r-ir E32 F20 E23 651 119 F37 E40 F47 0/3 640 E37 F&3 E25 416 306 R50 76 41 MPl3 R03 29 R30 15 FA948itzpp.Z 86 70 R'3 316 613 672 E44 F02 E02 F04 E04 FIO E47 F13 E28 F43 E50 C,54 619 313 f6oO R09 120 R29 88 74 127 207 3( 112 17 R20 4Poi 31

421 E55 F09 1961 F30 E18 F48 P71 F49 03 F33 V&5 F15 E05 631 321 R16 107 R19 12. R40 99 R26 208 PR24 // R7I 85 R33 331 60 764 E29 F6 2 E1 F64 E58 F57 E45 F?8 E43 F41 E 22G57633 3?0 imtnO R15 89 01 08 I2, 105 67 100 R12 79 1R07 !Ltop 333 (Z11 E57 F24 E17 1656 D43 F17 E20 F681050 GI1 E35 F31 E27 60Z 311 R04 104 82 ,ploq F?48 51 R53/6 RV Ri!p 07 3 R44 302

632 66I 21 P1 0 ,9 E59jFG67E 46 663 E.?G F5210671 467~ 64 332 1pio2 R.2 48 125 12i~yo RIO 166 9712P107 94 .27 R52 j:11 oa 1Ai.

I 17 FOS F ,6 {349 65 R45 (7,| 05 D01 F6C

305. 1?08 28 I 9 622

G09 16 1.3091,P

G H 'J K L M " N P R z' ' 0 ' ' '0V

25 344 336131013301 I I

R4 ilPI5 R3 4 0 7 E:9 ~ 3GE'G18 4124 R4 7 1 IP1151 R 14 'PR 394plo43 B

F42 1I_5

IF34VE10F39 j03 F54 E39 F53 1F51 -03 184 /2 175 IR40I205 1209196 IR341 36

750 60116 15

,5I7 317

640 340 335

I I I

.3281 291C38 32913291338

I I I

I I

izPl/e 3Ol11315 . 614 I I

1 iI

1800

P54 P.41

E40

129Fob

LEGEND

G66 = Assembly Number

16PI0I w Insert M

S/S= Secondary Source

P L P/ : Primary Source

D - REGION 4

E - REGION 5 F-- REGION 6 G - REGION 7

Insert Number With Letter P" in it,is a Burnable .-Poison Assembly.

-8-

(HIPAR) (HIPAR) (HIPAR) (LOPAR )

FIGURE N

270.0- -

FZ7EI iFI2rE42 FOL fP45F'4(P26 C08

I1! 36 . [R 13 ZO6 R46 40 1R49 308(726 32(.

G27 327

02 6 E3 4 ,#35 24 F55. 063 F14 ID52 1623 4 I L I5 13 Z

I

_p 02 .1669 E53 j 711 1_ _ k-pto5IR02(; [_ v,-F1o31F321

13

-12

- II

-10

-9!

-8 -6I

-5

-3 9 7

-2

-I

900

I

2. REACTOR CORE DESCRIPTION (Cont'd)

with type 304 stainless steel sleeves posi

tioned at each axial location of an Inconel

718 spring clip grid. The sleeves are welded

to the nine grids, spaced approximately equi

distant along the length of the fuel assembly.

The Zircaloy-4 thimbles are fastened to the

top and bottom nozzles.

2.4/ Core Loading Verification

Cycle 5 core loading verification was carried

out by monitoring the movement of each assembly

during actual core loading. The location of

each assembly as it was loaded into the core

was verified using a detailed procedure pre

pared from the Cycle 5 loading pattern., The

final core loading verification was carried

out upon the completion of core loading, by

viewing the television screen. The iden

tification numbers of 193 fuel assemblies

and BP insert numbers of assemblies bearing

BP rods and neutron source assemblies were

verified against the Cycle 5 design loading

pattern.

-9- .--

.'r - -. .... . .- .

3. MEASUREMENT METHODS

The reactor was maintained at the just critical state

during the physics measurements and the reactor power

was held constant via control rod/boron exchanges and/

or control rod/coolant temperature .exchanges. Small

changes in core reactivity during the test were moni

tored by a reactivity computer. The axial power dis

tributions in the instrumented assemblies were obtained

using the movable incore detectors.

3.1 Reactivity Measurement

The absolute measurement of small changes in

reactivity was provided by the on-line solu

tion of the point-reactor kinetics equations

using an analog reactivity computer. The com

puter was checked out by comparing the reacti

vity obtained from the reactor period with that

given directly by the reactivity computer. This

comparison is shown in Table 3.1. A good agree

ment between reactivities obtained from two

sources demonstrated the reliability of delayed

neutron data, given in Table 3.2, which were used

as an input to the neutron kinetics equations

of the reactivity computer.

-10-

Table 3. 1

REACTOR PERIOD AND REACTIVITY RESULTS

Doubling Time Reactor Period (sec) (sec)

164.2 64.0

237.0 92.0

Reactivity Predicted

(pcm)

26.2 57.6

Reactivity Measured

(pcm)

26.0 57.5

Difference (M-P) %)

P

-0.76 -0.17

M = Measured Value

P = Predicted Value

Table 3.2

DELAYED NEUTRON DATA BOL CYCLE 5

0.00177

0.001263

0.001137

0.002336

0.000807

0.000276

A1i (sec - )

0.0126

0.0308

0.1174

0.3148

"1.2503

3.3329

L*= 15.09/Asec (Prompt Neutron Life Time)

= 0.970 (Delayed Neutron Importance Factor)

peff = 0.005816.

- i -

Group

1

2

3

4:

5

6

• } .7

3. MEASUREMENT METHODS (Cont'd)

3.2 Core Power Distribution Measurement

The Movable Detector (M/D) Flux Mapping System

was used to collect power distribution data dur

ing start up tests at various power levels. Data

from the M/D system provided input to the INCORE 2

code (Reference 4) to generate three dimensional

core power profiles. The INCORE 2 code combines

measured flux distributions with calculated de

sign flux distribution to yield measured hot

channel factors and F,, quadrant tilt, core

average axial offset and relative assembly

power distributions.

3.3 Thermal Power Measurement

Core thermal power was determined by performing

a heat balance across each of the steam genera

tors. This measurement required the determina

tion of steam generator pressure, feedwater

inlet temperature, feedwater and steam generator

blowdown flow and other parameters.

- 12 -

4. HOT ZERO POWER (HZP) TESTS

4.1 Initial Criticality

The Indian Point Unit No. 2 Cycle 5 attained ini

tial criticality on May 21,, 1981. The criticality,

at beginning of life (BOL), and HZP condition, was

obtained by the sequential withdrawal of RCC shut

down and control banks and by subsequently diluting

the borated reactor coolant. During the approach to

criticality, ICRR (Inverse Count Rate Ratio) plots

versus dilution water and RCC bank position (Figures

4.1, 4.2), were maintained. Measured critical boron

concentration, xenon free BOL, HZP and ARO (All Rods

Out) core condition, was equal to 1367 ppm compared

to the design prediction of 1360 ppm (Reference 3).

The difference of 7 ppm between measured and design

boron concentrations was well within the acceptance

limit (+50ppm of design) for this measurement.

Neutron flux range for HZP test was determined by

the following method. The upper flux limit was de

termined by observing the effect of nuclear heating

which-decreased reactivity and increased reactor

coolant temperature. The upper flux limit for test

ing was set below the'flux value at which nuclear

heating was observed..' The flux level for nuclear

heating was determined to be above l. 5x10- Amps,

as shown in Figure 4.3. The lower limit was governed

-13 .

It I' j.

~I'

f~) ~ t

litl

~tt Ih

lit

t ..

v-un. p. *v

,1

II I I! I * I.

'~1 1~I

*1' .1 jt I~

i

46 6010

FLUX LEVEL(FWhPS)

C

-J Ij

MIA

W.E 5' -- *AI 70CILE DlVImONs

- I H 0't

~.J.

L~J

.4.1 it .J

A~~

4. HOT ZERO POWER (HZP) TEST (Cont'd)

by the background noise level. In the presence

of background noise the reactvity computer

* underestimates changesiin core reactivity.

The lower limit was determined by measuring

the differential worth of control bank "D"

at various neutron flux levels. As shown in

Figure 4.3, above the lower limit of zero power

physics range (10- Amp) differential rod worth

is essentially independent of flux level. HZP

Physics tests were carried out at a neutron

flux level between 10-' and 10- Amps. on the

Keithley picoammeter,.over this-flux range the

effects of nuclear heating and background noise

are absent.

4.2 End-Point Boron Concentrations

In Table 4.1, measured end-point boron concen

trations, for different control rod configur

ations are presented. The corresponding

design values, from Reference 3, are also

listed. The maximum-deviation, as shown in

Table 4.1, is 8 ppm. This satisfied the accep

tance criteria of + 50 ppm.

4.3 RCC Bank Differential and Integral Worths

-Measurements of the differential and'integral

worth of individual RCC control banks (C and D)

0 . -17

wir7

Table 4.1

End-Point Boron Concentrations

Configuration

All Rods Out

D In

(1) Measured

(Ppnf)

1367

.1278

(2) Design (ppm)

(i) - (2) Deviation (ppm)

1360

1270

77777

4. HOT ZERO POWER (HZP) TESTS (Cont'd)

were carried out via boron/RCC exchange, while

the reactor was in the critical state. The re

* activitycomputer trace provided the change in

reactivity during insertion/withdrawal of an

RCC bank. The differential worth of bank is .4

defined as the amount of change in reactivity

per unit step of bank position around an average

bank position. The integral control bank worth

was obtained by summing the differential worths

for the bank positions during the insertion

or withdrawal of the RCC bank. In Table 4.2,

the integral worths of individual control banks

"C" and "D" are presented along with the design

values. In Figures 4.4 and 4.5 differential

and integral worths of control banks "C" and

"D" are shown.

Measured integral worths in all cases are

within + 10% of design value, (Reference 3),

the acceptance criterion for the RCC banks.

4.4. Isothermal Temperature Coefficient

Isothermal temperature coefficient measurements

were carried out for two control rod configura

tions (All Rods Out, and control bank-D In). Mea

surements involved heatup and cooldown of the

19-

TABLE 4.2

CONTROL ROD BANK INTEGRAL WORTH SUMMARY

CONFIGURATION

ARO

D IN

(1) (2)

MEASURED DESIGN

Worth (pcm) Worth (pcm)

840 819

896 909

(1) - (2) (1)

DIFFERENCE

+2.50

-1 .45

BANK

D

C

I t'j 0

44i

20 X f0 TO TE~E ICH MP 7 ', loit WEKEUFFE TO ESSER IC- wc It; l

'HI~- J-~ ,I

I, -17

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* 20 X 2.0 TO THE INCii - ?XV WE I<UUFFLI & ESSE.R CUO m- IN -

I

4. HOT ZERO POWER (HZP) TEST (Cont'd)

reactor coolant. In Table 4.3, measured as well

as design values of isothermal temperature coef

ficients for All Rods Out (ARO) and Bank D In

cases are presented. Measured values are obtained

from the reactivity versus temperature curves

provided by an X-Y plot recorder. The design

values shown in Table 4.3 are from Reference 3.

Measured isothermal temperature coefficients

were all negative and within the acceptance

criterion of + 3 pcm of design values.

Using the design value of Doppler coefficient at

HZP condition equal to -2.48pcm/°F, moderator

temperature coefficients calculated from measured

isothermal temperature coefficients are equal to

+0.lpcm/°F, and -0.12 pcm/°F at the rod configu

rations of ARO and Control Bank D In, respectively.

The Indian Point Unit 2 Technical Specification

(Ref. 2) requires operation of the reactor at a

negative moderator temperature coefficient, in

the operating power range. Therefore, the allowed

boron concentration limit below which reactor

operation is permitted, is established using the

negative moderator temperature coefficient limit.

The maximum allowed boron concentration of 1350 ppm

required partial insertion of control bank D for

plant startup operation.

-23-

Table 4.3

ISOTHERMAL TEMPERATURE COEFFICIENTS

Configuration

ARO

D In, C Out

(1) Measured (IPCM/0 F)

-2.38

-2.6

(2) Design

(PCM/OF)

-2.4

-3.4

(1) - (2) Difference (PCM/°F)

+0.02

+0.80

S.- - ~1~"

- 24 -

4.5, Differential Boron Worth

Based upon measured end-point boron concentra

tions and control rod worth data given in Tables

4.1 and 4.2, the measured differential boron

worth was calculated t o be equal to 8.78pcm/ppm

at 1322 ppm boron concentration compared to the

design value of 9.15 pcm/ppm (Ref. 3).

-25-

5. POWER TESTS

The power measurement tests consisted of: (a) relative

assembly power distributions at HZP <1%,-50%,90%,-l00%

of full power; .and (b) reactor coolant flow determination.

5.1 Core--Power Distributions

The purpose of the low 'power flux map testing was:

1) to verify that the Cycle 5 core loading

pattern was correct.

2) to verify that quadrant power tilts, hot

channel factors, and relative assembly

powers met the acceptance criteria.

3) to verify the design calcuations

Table 5.1 is a summary of the calculated INCORE parameters

for the four maps.

5.1.1" Hot Zero Power Map

The low level power flux map (HZP map)

was taken at a core power below 1%. The

HZP map was taken at a power level at

the onset of nuclear heating with D-bank

at 198 steps withdrawn.

Figure 5.1 gives;the measured versus de

sign assembly power distribution for the

HZP map. The measured hot channel peak

ing factors met the IP2 Technical Speci

fication limit.

-26 -" "-: "-

Table 5.1

SUMMARY OF RESULTS OF INCORE ANALYSIS, IP2 CYCLE 5 STARTUP MAPS

APPROXIMATE Power (%)

Axial Offset (%)

Quadrant Tilt (%)

Measured Peak FW*

Peak Fq**

HZP

22.510 9.050

3.73

1.562

2.473

1.80

1.479

1.964

90 100

3.258

1.27

1.440

1.846

1.220

0.88

1.451

1.852

*Includes measurement uncertainty factor of 1.04 **Includes measurement and engineering uncertainty factor

of 1.0815

4..697. .952. 1.104.

S16,2 -123- --' 12 ; 3.

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ii

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N . 12.5.

M ~ ,. I.0 1.007*

M . 11.0

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1.106. 1.261 -J. • 11.9. 11.4

1.144. 1.173

H 12.4. 11.9

1.095. 1.260 6 • 11.4.. 11.1

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1.090" -3.4,

.861. -7.2.

1 • 138. -8 , 9.

I . 077. -10.2.

1.134. - :1. (.) 0.

.816. -12 ,0.

1 ,024. -8 2.

. . 1.# 005

* , , . . +

1 .269. 11•9.

. ,

1 057, El 3.

I. 0, 1 1 •.095.• -5.6.

.987. -5.7.

1,154. -8 4

.963. -11 .2.

1.059. -12.0.

- .9#

1.108. -11,4.

. 0,

*.945. -8.8.

1 •052. -8.*2.

.911. -3,4.

1.221, 7.8.

• . 4

* , 1.147. 12.7.

. .$

1,177. 12,3.

o 116o 1.116. .7,

o 860. -7,3,

1,174. -7*4,

1 * 068, -11.0.

1.053# -12.6.

•793. -13.1.

1.050. --12.8.

1,054. --12.1.

1,129. -10*9.

. , ,61

B861o -7,.2.

1.082, -2.4.

1.092. 4.2.

1.130. 14.3.

1,231. 8.7.

.901* -4.5.

1 082. -5*6.

.951. -8.3.

1.114. -10. .

.962. -11,3. .o .•

1*065. -11.5.

.960. -1 1 *

1.102, -12.6.

.930. -1i.I.

1 .098. --5,4,

.965. -1.2.

1.146.

1.1 S..,

.700. .974. 1,069. 1.097. 1,013. .866. .614.

• 13.9. 12.8. 8.2. 7.8. 3.1. 2.2. 2#3.

MEASURED' AND PERCENT. DIFF. OF FIHN If:'2 CY R 1 1

.969, 12,3..-

.964. 4.7,

. , ,

1.102. -4.4. S. , .

.918. -4 .4.

1.020. --8,5.

.818.

1,121. .-11,*1,

1 •068•

-10o9.

1 . 108 --11I.3.

.812. -12.5. . . •

.962. -14.8.

•945. -2.4.

1.147. -• 8•

.884. -1.0.

.687. 11.9.

1.000#

.885. 3.7,

1.194.'

.2.

.847. -10.2.

1. 013. -10,2.

.955. -8.7.

1.157. -8,7.

.945. -8 * 8•

1,002o

-10,1.

.924. -2.1*

1.168 -2•.1

.849. •.

.927.

* 2,

.972. 7.1.

1.231. 6,5.

1 , , 1,.253. 1

2.3. 0 " . 4 1.221'.

2.3.

.991. 2,4.

1,083o -6.7.

, , ,

852.

-8. 3.

1 • 044. -7.2.

.8139. -7.3.

1 . 167. -2. 1

1.215. -. 8.

1.181. 2.9.

.934. 4.4.

.590. 9,2.

.719. 9.1.

.241. 8.1.

•867'. 2".3.

I *•: 18:L•

2..2.

,973. -. 4.

4 , .

1 0090. -1.6.

.91 8. -2.7.

1 .096. -4#8.

.835. -2.1.

1.162. '.•

#695. 5.6.

.590. 10.6.

.608. 13.9.

1*020. 14.0.

945.

,949.

6,2.

1 * 198

5.7.

7.1.

1 .166. 2.9.

,943.

2.5.

.962. 1.9.

.91 1

,570. 5.6.

gj*

. r S

• I11FF

re 5.1 RI ,5-21"-61 HZFEDIN=I198 STEPS

4 #

.612. 2.0.

.905. 6.8.

1•042. 6.0.

1 .093. 7,5.

1.056,.

6.9.

1.1.

.676. 10,2.

-3.2. -5.8. , . 4. 1.076. 1.038.

14,0. 12.7..

• • . . • . • • # l • • , @ • • @+

5. POWER TEST (Cont'd)

The quadrant power'tilt met the ac

c6ptance criterion :of 4% for HZP

flux map (see Appendix B).

The Cycle 5 HZP map demonstrated an

in-out power tilt, which was also

present in the HZP maps of earlier

cycles. The acceptance criteria for

the measured assembly power distri

bution were +15% for relative assembly

power (Pi) < 0.9 and +10% for (Pi) > 0.9.

Due to the presence of in-out tilt some

periphery and interior assemblies ex

ceeded the acceptance criteria for the

measured assembly power distribution.

See Figure 5.1.

5.1.2 50% Power Incore Flux Map

A full core map was taken at 50% power

on June 2, 1981 with control D-bank at

210 steps withdrawn. Figure 5.2 gives

measured versus calculated assembly power

distributions for the 50% power map.

Measured hot channel factors and quadrant'

power tilt were within Technical Specifi

cation limits and were, therefore acceptable.

-29-

y SURED AND PERCENT. DIFF. OF VR R * . . . 9 •

R .-...... .. 8...

P • 9.5.

S.0 .732# N . 9.9. 9.3.

.990. 1.235# m 7.5. 6,5.

.695. 1.011. .857. ....-L . 9.6. 5.1. -.5.

# .929. .979. 1.153. K * 4,7. 3.2. -.5.

1.041. * ........ J." 3.8.

-: . • 1.070.

H 4.2.

"-,. . 1.054. -.I 0.* 5.8.

• I • .901. F. 3,5.

* .659.

1.180. .-937. 3#4-. -. 9.

1.093. 1.076, 3,8. -2.7,

+9--.. ., . .

1.207. .970. 5,7. -.8.

•-... •.-. .. . .

.951. 1.154. 3.3. -.6.

1.003 .sR.1'- E . 6.4. 6.4.

II * .991, SlD l 9,1.

..... -* . .. ..... • •607 1. ...... .C . .... . 11.0.

3.8. ",.. + ..

1.196.

.716. 70.

" .. .625, * 13,1.

+

A

8.2,

1.210. 5.1.

1.•252.

2.8.

1. 177, -.6.

.948. -. 7,

1.107. -2.4,

.882, -4. 1

1.099, -4.2.

o913, -5.2.

1 *168.

-1o4, I -!o I 1.2.01. -1#4#

1.9225. 5.6,

1.032. 12.0.

FDHN IP2 CYCLE

re 5.2

2 57 OE STARTUP MAF,DIN=210 STEPS 6C

.689. .931. 1.049. 1.084. 1.077, .946. .600. 11.3. 6.9, 5.3. 5.6. 7.5. 6,7. 7.3.

11.3. 6.6, 5.0. 5.3. 4,1. 1.8. 3.3. 4.7. 7.3.

.952. 11.3,

1.163. -1.9.

.916, -1.8.

1*060, -3.2.

.90. -3.7.

,1 . .

-4.7,

.985, -4. 1.

1.051. -5.2.

.912* -2,4,

1. 156. -2.4.

.840. -2.4.

1065. 10.8.

. . •

1.189. 1.001. 1.099, .915. 1,121, .877. 1.208, .718. .618. 2.3. 2.3. -#7. -3.2, -3,2. 1.9, 4.2. 7.3. 13.0.

•944. 1.109, .881,. 1,092. .924, 1.174, 1,231. 1.233, 1.027. -1.8, --3,3. -4.2, -3.7. -3.2. -.8 1.1. 7.1, 13.1. " + : ' .. .. ....... .. .. " 8 , : : . . . . 1.07, .993. 1.186. .971. 1.046, 8B4. 1.198 .863. .949. .623. -1.9, -3*3. -4.2. -4.6.l -4*6. . -5.4,. ....1.1,... .9. .7. .6.

.874. 1.173i 1.105. 1.152, .857. 1.048, .973. 1.171. .946. .898 -4.1. -4.6, -5.5. -5.5, -5.9# -5.4, 1.1, .B. 2.8. 3.2.

1.160, 1.001. 1.112. 1.007, 1,166. .. 993. 1.133. .980, 1.169. 1.022, -4,8, -5.8. -5.9. -5.3. --5.1. -3,4. -1.2. .1. 2.4. 2. 6.

1,109. 1.112, 862, 1.122, 1,115. 1•202. 896, 1,103. 1.065.. 1.061. -5.2, -5.9. -6.1, -5.1, -4*6. -2.9. -2.5, -. 2. 3,1. 3.3.

1.171. 1,002. 1.104. 1.002. 1.154. .987. 1.100. -. 933. 1.152. 1.036. -4.7. -5.7, --6.6. -5,8. --5.4, -3,0. -3.0. -1.4. 1.0. 3+4.

.853. 1.147. 1.093. 1.143. .852. i.050. .917. 1.115. 958. .939. -6,3, -5.9. -6.5, -7.0. ..-6,5# -4.2, -3o9, -3*7. 1.0. 5.8.

1.042, .967, 1.162. .964. 1,012. .920,. 1165. B7 69+0 ,6.+ .8 7 .970,.•671. -4.9. -5,0. -6,2. -6,2, -8.7. -16,. -1.7. .- 1.6. .9. 5.9.

.909. 1.079. .877, 1.104. .941. 1.166. 1.211. 1.165, .925. -+4,8, -4,8, -4.6. -3,8. --2.2. -1.6. -. 6. .4. .4.

1.118, .925. "1,083, .960: 1.140. .844* 1*145* -*672. .555. --3.5. -2.2, -2.0, -1.8. -18 . -1.3, ..-. 6. #4. 4.

1.026. 1.181. 1.067, 1.132. .903, .931. .894. .549. 8.1, 3., 1*4. -.9, -1.9. -1,3, -1, ..

.702, .959. 1,040, 1.063. .998, •861. -613. -~ -..---...-- ~ ... .-. MEAS. 10,7. 8.2. 3.8, 3.6. .2. -. 0. -1.0,

DIFF4 * ~. 99 -*' . ~9 *~ - .-. 9-r- 9--------------------

, . .


In-out power tilt fbr the 50% map was

reduced compared to HZP map; however,

a few :assemblies exceeded the measured

power distribution acceptance criteria.


A 90% power map was taken on June 6, 1981

at approximately 150 MWD/MTU with bank-D

at 210 steps withdrawn. This map was a

base case map for the BOL Cycle 5 Incore/

Excore calibration. The measured hot chan

nel factors and quadrant power tilt were

within the Technical Specification limits.

The In-out power tilt was still present

in the 90% map. Figure 5.31 gives meas

ured versus design assembly power distr

butions for the 90% map.


A 100% power map was taken on July 2, 1981

with control bank-D at 211 steps withdrawn

and burnup of 899.7 MWD/MTU. Measured hot

channel peaking factors and quadrant power

tilt were within the Technical Specifica

tion-limits. In-out power tilt present in

the 90% power map persisted in the.,100% power

31.-

O. i . . . . . . . .

MEASURED AND PERCENT DIFF. OF FDHN IP2 CYC! R 1,1

.... .................... 5 ' 9"9

: -- -4 4 599*

... ... ......... . ... 6 .7.

*.607. .730. N * 7.2. 6.5.

. .969. 1.203. - M 5.4. 3.8.

* .691-, .900 .849. L • 9.4. 3.8* -2.3.

.. . 919. .969. 1.144. K • 5,8. 3.2. -. 9.

. 1.022. 1.167. .945. J . 5.7. 5.4. -.4.

* 1.048. 1.088. 1.087,

r H 6.0. 5.7s -1.4. ._-4... 4 -..-. 4,..-..... 4

. 1.009* 1,160. * 955.

G * 4,8. 4.6. -2.5.

- . .856, .913. 1.131. F * ,0. -. 1. -2.4.

-- -- 4 .- 4.- .... ... 4. 4 4+ '''' ' I

. .647, .976. .885. E , 4.3. 4.3. 2.4.

" . .. * . .984. 1.191. -D , 8.3. 3.5.

61A. 725, C * 9.4. 5..9

- o -. 4 4

. .626. B .. .. 10.5.

A

..681..'915. _9_..8. 6.9.

.963. 1.028. .976. 6.0. 9.8. 6.7.

1.176. .949, 1.204, 2.2. 9.8. 3,9

4 . .l . , 4 •

1.217. 1,169. .958. o2. -1.6. -1.6.

1.158. .925. 1.093. -2.4. -2,3. -1.6.

.942. 1.059. .891. -2.4. -3.7. --4.0.

1.108. .989, 1.171. -3,0. -4.2. -4#9.

. . 4 , . • • , •

.897. -4,0.

14107. -4.1.

.924. -5.0

.•j .0

1.171. -1.4.

1.198. -- 1.4.

1.209. 4.4.

1.014. 10.4.

1,188. -4.7.

.999. -4.0,

19055.

1,114, -.5.7,

1.173. -544.

* . , 4

.856. -7.7.

,• . .--.• ,. o

.923. 1.061. -2.5. -3.5. ' ' , ' 9 24 1.157. .932. -2.5. -3.5.

". o o • . •

.848. 1.114. -2,4. -3.5.

1,049. 1.028. 10.2. 9.4.

rure 5.3

I5FC3 ,6-06-81,150 MWD B.U.,DBANK 210 STEPS

* * -. - .1 • 9 # f

1.031. 1,061* 1.056. -7.1o, 7.4, 9.2.

1.185. 1.102 ...1.165,6.8#. 7.1. .- 5.2.

Power

.939. ,680. 8.0* 7.7.

,. -, -4 4: ,. - . 4 - . .>..- ... .. .

,962. .984, .956. .601. 2.4. ...3.3. 4.1, 6.1.

1,018, 1.097, .912, 1,111. .802. 1,200l .727. .620. 3.9. -.5# -3.9. -3#8, 1,5, 3.6. 6*1. + 10,5.

1.122. .893. 1,097. .929. 1.174. 1.223. 1.216. 1.005. -2.8. -4.5, -4.0. -3,+8. -1.1. .6. 5.6. 10.6.

1,011, 1,192. .902. 1.042* .893, 1.195. .058 . 929. .615. -2,8, -4,4. -4.9. -5.3. -5.7. .6. -. 8, -. 8. -. 8.

1.185. 1.113. 1.160. .067. 1.047. ,966. 1.150. .945. .088.

-4.5, -5.8 -5,8, -6,5. -5.7. -. 7, -.8. 3*4. 3.8.

1.017. 1.122, 1.021. 1.1'71. .996. 1.112. .973. 1.147. .998.

-6.1. -6,7. -5.0. -5.6. -4.3, -3.7. --.7. 3.3. 3.7,

1.122. .879. 1.134. + 1.119. 1,195. .894. 1.096. 1.002. 1.042.

-6.7. -7.1. -5.7. -5.3. -4.1. -4.3. - -. 6. 5.1. 5.4. 4.-. .. .. , ,4.4 . .. . 4. 9. ... ..- .. . . 4... . . . .

1.017, 1.122, 1.024, 1,161. .988, 1.097i .935. 1.131. 1.014.

-6.1. -6.7. -5.4, -5.7, -4.2. -4.b. -1. 4. 2.2. 4.8.

.159. 1.104. 1.161. ,864, 1.043. .927.,1.120. .957. -".930.-

-5,9. -6.6o -6.4. -6,9. -5.2. -4.1. -3.0. 1.9. 7.0.

.991, 1,178. .982. 1.016. 934. 1.172. .858. .966, .677.

-3.9. -5.5. -5,6, -8.5, -1,4. -1.3. -1.3. 1.5. •7.1.

1,098. .904. 1i17..956, 1,172. 1.210, 1.164. .923.

-3.9. -3.3. -32., -1,8. -1.4. -. 4. .,. ..

.932, 1.089, ,970. 1.159. .872. 1,103. .719. .595,

-1.7, -1.3. -. 2.: -.0. .9. 2,8, 5.0. 5,0.

1.171, 1,060. 1.123. .913. .946. .946. .614. 5.7.- , 3.0, 1,2. •-.2. - 1.0. 4.2, 9.4.

9 4 9 ,

.696. .951. 1,025# 1,044. .992, -.882. .641. .. .. ... . ...... .. MEAS

10.2. 9.4. 6,0. 5.7. 3.0. -3.1. 3,3,.. . DIFF .-

.8. . 4 .


map and a few assemblies exceeded the meas

ured power-distribution acceptance criteria.

However, it did not violate the Technical

Specification limit for FO and thus does

not have an impact on the safe operation

of the plant. Figure 5.4 gives measured

assembly power distributions versus de

sign values.

5.1.5 Summary of Incore Flux Maps

Table 5.1 gives a summary of four Incore

flux maps(HZP,-50 90% and"-lO0% power)

results. Measured hot channel factors

F~H and Fa for all four maps were within

IP2 Technical Specification limits.

Quadrant power tilt for the HZP map was

3.73% which was within the acceptance

limit of 4% quadrant tilt for the HZP

condition, and for the three'other maps

(-50%,-90% and100%) quadrant tilt was

lower than 2%. In-out power tilt was

predominant at HZP, and in general was

smaller at higher power. The relative

assembly power distributions of a few

assemblies at the periphery of the core

exceeded the acceptance criterion, given

- 33 -

.;j

J :

L •

K

-o - J

H.

CA G •

uj.

I. E.

.598. 5.5.

.606. .730. . 5.9. 5.3.

.971. 1.198. 5.6. 3.1.

.o .701. 1.000. .846. 12.3. 5.5. -2.9.

.o S919. .952. 1.139. 8.2. 4.9. -. 9.

1.025. 1.183. .947. 7.7. 7.3. -.4.

1.056. 1.105. 1.094. 8.1. 7.7. -1.3.

., 1.023. 1. 189. .968.

7.8. 7.6. -i.4.

S870. .919. 1.141. 3.8. 3.8. -1.1.

.649. .988. .890. 5.8. 5 .8. 2.6.

.980. 1.178. 7.7. 2.0.

.615. .724. 8.4. 4.4.

.625 [ % i:.9.1.

.970. 1.050. 6.5. 12.4.

1.163. .975. .6. 12.4.

1.215. 1.162. -. 4. -2.9.

1.158. .925. -3.1. -3.8.

.945. 1.064. -3.2. -5.3.

1.095. .979. -5.0. -6.1.

.89 .1171. -5.7. -6.5.

1.096. .990. -5.8. -5.7.

.921. 1.062. -6.3. -6.3.

1.152. .923. -3.7. -4.1.

1.174. 1.146. -3.7. -4.1

1.191. .836. 2.5. -4.1.

1.021. 1.071. 11.0. 13.0.

.967. 9.3.

1.228. 6.4.

.955. -2.9.

1.101. -2.9.

.889. -5.5.

1.150.

-6.8.

i .092 -7.4.

1 .155. -7.1.

.860. -8.6.

1•051. -6.4.

916. -6.2.

1.098. -4.5.

1.015. 11.8.

1.206. 9.1.

1.045. 6.4.

1.111.

-4.5.

1 .003. -4.5.

1. 170. -5.9.

.999. -7.3.

S0;3 -7.8.

996. -7.5.

1. 139. -7.7.

•979. -6..0.

1 .083. -5. 9.

934. -1.8.

1. 184. 7.4.

5.4 IC4,100% POWER., 07-02-81,D-BANK=211, BU=899.

1.072. 1.065. 9.7.-- 11. 9-.

1.122. 9.4.

1.110. .2.

887. -6.0.

1. 176. -6.1 .

1.096. -7.0.

1.098. -7.4.

834. -7. 9.

1.089. -8.2.

1.083. -8.1.

1.162. -7.2.

.896. -5.0.

1.098. ". 9.

1.070. 4.3.

1.183.

7.3.

.913. -3.9.

1.097. -4.8.

.97 6 -6.4.,

1.147. -7.1.

1.002. -7.0.

1.105. -6.8.

1 .000. -7.1.

1.144 -8.0.

.977. -7.0.

1.124. -3.4.

986. .4.

1.131. 2.4.

• .690. .917. 1.038. --o--- 12 .4_ 9--5. ---9-.4.

.946. ...1 .'4.

.951. 4 .7.

.1. 106. -3.7.

•938. -3.9.

i.050. -6.5.

.868. -7.8.

1.155. -7.1.

1.096. -7.0.

1.141. -7.5.

860. -8.6.

1.022. -9.9.

970. -1.4.

1.161. .7.

893. .9.

•681. .9 2.

.985. 4.0.

.883. 1.3.

1.179. -1.3.

•894. -7.1.

1.054. -7.1.

.990. -5.8.

1 .179. -5. 9.

980. -6.0.

1 .041. -7.3.

953. -1.0.

1. 185. -1 .0.

.882. 1.6.

• 952.

1 9.

.950. 3.3.

1.194. 2.8.

1.210.

-. 8.

1.187. -. 8.

1.029. 4.7.

1.108. -4.8.

.888. -5. 9.

1.096. -4.8.

930. -4. 7.

1.175. -1.7.

1.207. -1.1.

1.182. 2.3.

949. 4.3.

. 705. .950. 1.025. 1.049. .991. .878. .643.

13.0. 11.8. 7.7.. 7.4. 4.5. 4. 8. 4.8.

Fig IMEASURED AND PERCENT. DIFF. OF FDHN IP2 C

R 1 1

(D

Ln

D

B

B

.60 4. 5.5.

.731. 5.4.

1.210. 4.7.

.908. 4.7.

1.207. 4.7.

992.

1.099. -. 7.

939 . -1.3.

1.123. -2.3.

.857. -1.7.

1. 148. -1.2.

.722. 4.0.

.620. 9.3.

.625. 10.2.

1.003. 10.2.

.978. 4.7.

947. 7.0.

i.170. 5.9.

1.090. 6.3.

1.146. 4.0.

.947. 4.3.

972. 2.5.

.909. -1.2.

.596. 4.0.-

4.

•EAS

•DLIFF.;

PER MTU -

.642. 4.7.

.899. 7.3.

1.007. 6.2.

1.041. 6.6.

1 .020. 7.2.

942. 10.9.

.692. 10. 9.

.. ,

5.0 POWER TEST (Cont'd)

in Appendix B, but the hot channel factor,

was below the Technical Specification

limit. All safety related core parameters

associated with power distribution such as

hot channel factors, Fh and Fj, and quad

rant tilt met the acceptance criteria for

the measured values and are, therefore,

acceptable. In-out power tilt present in

the flux maps discussed above will have

no safety impact on the Cycle 5 operation

as long as safety related core parameters

are within Technical Specification limits.

5.2 Reactor Coolant System Flow Determination

The Technical Specification (Ref.2) requires that

the total RCS flow exceeds 340,800 gpm prior to 98%

of licensed power. Prior-to achieving 98% of power,

a number of calorimetric data sets, Delta-T readings[

and elbow tap differential pressure readings were

taken and the corresponding RCS flow rate determined!

to ensure that the total RCS flow exceeds the re

quired flow rate of 340,800 gpm.

Since the elbow tap transducer measurements of the

reactor coolant flow rate, are-sometimes inconsis

tent an independent method using the reactor thermal.

power-program, plus the RTD data for coolant leg

4 -35-

5.0 POWER TEST (Cont'd)

Delta-T was used. Usingthe known reactor thermal

power (Q), enthalpies h(T.) and h(T ) at hot and

cold leg.temperatures, flow rate (W) is given by

W= Q h (T ) - h(T')

The volumetric flow rate, then, is the mass flow rate

multiplied by the specific volume, at cold leg tem

perature.

On June 6, 1981, with the reactor power at 90% of

nominal, the calculated flow rate using this method

was 379,250 gpm.

This exceeds the acceptance flow rate (340,800 gpm)

by a margin greater than the analyzed measurement

uncertainty of 4.2%, thus Technical Specification

requirements are met. Measurements made at other

power levels, using the calorimeter as well as elbow

tap methods are given in Table 5.2. The results con

sistently show the RCS flow rate to be greater than

what is required.

- 36 -

Table 5.2

RESULTS OF REACTOR COOLANT SYSTEM FLOW DETERMINATION AT VARIOUS POWER LEVELS

Reactor Power (%)

50.6

70.0

90.0

Measured .Flow(%) of Design Calorimetric Elbow Tap

Method Method

106.2

107.7

105.7

104.5

104.3

104.4

Note: Measured flow exceeded 100% of design flow even after taking into/account measurement uncertainty of 4.2%

- 37 -

~0~* -- - -~----.: ~ '..,-'----,.-"--

,,- -- - I

6. REACTOR INSTRUMENTATION CALIBRATION

The calibrations of excore power range detectors, re

actor coolant loop resistance temperature detectors

(RTD's) and incore thermocouples are presented:in this

section.

6.1 Excore Detector Calibration

An excore detector calibration using movable

incore detectors was performed at90% of full

power. A range of axial offsets were obtained

by inducing an axial xenon oscillation by rod

insertion and subsequent withdrawal, following

xenon buildup in the upper portion of the core.

A full core map was taken at equilibrium con

ditions prior to the onset of the xenon oscil

lation. Quarter core (partial) maps were also

.,taken at various axial offsets. The full power

total excore detector output currents were ex

trapolated from the currents-obtained at reduced

power levels. For each map, top and bottom de

tector currents for each detector were normalized

to the extrapolated full power current. Plots of

detector current versus the. axial offset calcu

lated by the INCORE 2 code are given in Figures

6.1 through 6.4. A linear least square fit using

the axial offset and excore detector current data

was performed. The results of least square fit

* -38.-

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6. REACTOR INSTRUMENTATION CALIBRATION (Cont'd)

for top and bottom detector current vs-axial

offset were used to calibrate the-Delta Flux

(AI) Meters and Overpower Delta-T and Over

temperature Delta-T setpoints.

6.2 Incore T/C and Wide Ran@e RTD Calibration

During the Reactor Coolant System heatup elec

trical resistance of RTDs (including spares),

along with their lead wire resistance, core

exit thermocouple outputs and wide range RTD

amplifier outputs were monitored at approximately

50 0 F intervals from 320OF to 5250F. RTD resist

ances were converted into temperature using

factory-supplied calibration tables and were

plotted against time. The RTDs with minimum

correction factors were selected for use in

the protection and process circuits, and four

additional RTDs were selected for the saturation

meter. Table 6.1 shows the correction factors.

of each RTD from the average temperature for

data taken over 320OF to 525OF temperature

range. Incore thermocouples provide a continuous

on-line monitoring of assembly exit temperatures

of 65 assemblies which are about evenly.distri

buted throughout the core. Thermocouple and

wide range RTD correction .factors were obtained

by comparing their temperature readings with the

- 43 .

TABLE 6.1

PRIMARY LOOP NARROW RANGE RTD CORRECTION AS A FUNCTION OF TEMPERATURE (OF)

FACTORS

Temperatures

RTD

410A 411A** 412A* 410B 411B*

412B 420A*** 421A* 422A** 420B

421B 422B* 430A 431A* 432A**

430B 431B 432B* 44 0A** 441A*

442A 440B* 441B 442B

32,0 OF

-0.8 1.2 0.7 0.7

-0.4

-0.4

0.4 -0.5 -0.1

5.7 0.0 1.7

-0.1 -0. 1

-0.1 0.3 0.1

-0 0.4

0.1 0.2

-0.1, 0.1

4200 F

19.9 1.8 0.5 0.5

-0.3

-0.4

0.2 -0.8 -0.6

5.2 -0.5 4.2 0.2

-0.1

-0.1 0.3

0.2 -0.1 0.1

0.1 -0.2 -0.2 -0.2

470°F

-20.8 1.7 0.5 0.7 0.

-0.1

-0.3

-0.6 -0.2

5.5 -0.2 4.3 0.2 0.2

-0.4 0.1

-0.1 0.2 0.1

0.2 0.2 0.1

-0.1

Note:. *RTD **RTD ***RTD

initially used for protection'system used for RCS Saturation meter capped Correction factor = Best estimate temperature - measured temperature-

- 44 -

525°F

-0.9 1.0 0.6 0.6

-0.1

-0.5

-0.'5

-0.6 -0.3

5.6 -0.1 4.3

-0.1 -0.3

0.3

0.4 0.2 0.3

-0.

0.3 -0 -0.1 -0.4

777'

6. REACTOR INSTRUMENTATION CALIBRATION (Cont'd)

narrow range RTD readings taken during the

heatup of the primary coolant system.

6.3 Overpower and Overtemperature Delta - T Setpoints.

Reactor coolant Delta-T measurements, which are

used to determine the overpower and overtem

perature Delta - T setpoints were performed dur

ing power escalation. Thot and Tcold were measured

in volts and then converted into degrees Farenheit.

Delta-T, which is (Thot - Tcold), was plotted

against the power level-, up to 90% power for each

loop. The Prodac-250 computer Delta-T readouts for

each loop were also plotted on the same graph to give

additional data points.

The full power Delta-Ts used for the setpoints were

obtained from best estimated values extrapolated

from 90% power. Table 6.2 shows the Delta-T

setpoints for each loop.

-45-

TABLE 6.2

OVERPOWER AND OVERTEMPERATURE DELTA-T SETPOINTS

LOOP

21 22 23 24

AT (-F)

51.0 51.0 50.0 50.5

- 46 -

'S ~S

APPENDICES

- 47 -

d ; I -

APPENDIX A CHRONOLOGICAL EVENTS

DATE

10/17/80 12/30/80 2/21/81 5/14/81 5/14/81 5/18/81 5/19/81 5/19/81 5/21/81

5/21/81 5/21/81 5/21/81 5/21/81 5/21/81 5/21/81 5/21/81 5/21/81

5/21/81 5/22/81 5/22/81 5/22/81

5/22/81 6/02/81 6/06/81 6/07/81 7/02/81

TIME

8:26 9:30 8:30

11:00 5:45 7:40

7:45-11:02 11:02

13:07-14:35 14:35-15:10

19:30 20:00 22:30

23:50 1:45 3:45 4:30

5: 22-7: 20 10:22

14:32 6:20

14:57

TEST DESCRIPTION

Reactor Shutdown (End of Cycle 4)o Fuel Shuffle started Fuel Shuffle completed RC4 Flow Elbow tap data taken RTD and Thermocouple data @ 5250 F RTD and Thermocouple data @ 320°F RTD and Thermocouple data @ 420OF RTD and Thermocouple data @ 470OF Start RCC withdrawal to ARO condition Start to dilute boron to criticality Monitoring approach to criticality Reactor Critical Zero Power Testing Flux Range Reactivity Computer Check HZP Flux map completed ARO boron endpoint Isothermal temperature coefficient (ARO) Start control bank D measurement End control bank D in Boron endpoint bank D in Isothermal Temperature Coefficient, bank D in Control bank C worth measurements.1 50% Power Flux Map completed 90% Power Flux Map completed Incore-Excore data collection completed 100% power flux map completed

- 48 -

U-.

L ~|

Appendix B Acceptance Criteria

Parameter Acceptance Criteria

1. Boron Endpoint (a) ARO (b) Control Bank D In.

2. Integral Rodworth (a) Control Bank D (b) Control Bank C

3. Isothermal Temperature Coefficient

4. HZP (a) (b) (c)

Incore-Flux Map

Fq Quadrant tilt*

5. At Power Incore Flux Maps (a) FO (b) Fq (c) Quadrant tilt

Within +50 ppm of design value

Within +10% of design value

Within + 3 pcm/°F of design value

Within Tech. Spec. Within Tech. Spec. Less than 4%

Within Tech. Spec. Within Tech. Spec. Less than 2%.

* At HZP, higher quadrant tilt is anticipated.

criteria is based upon analysis.

limits limits

limits limits

Acceptance

Note: Acceptance criteria are guidelines which are used to

assist in evaluation of startup physics tests results.

- 49 -

8. REFERENCES

1. Docket No. 50-247 Final Facility Description and

Safety Analysis Report, Consolidated Edison Com

pany, of New York, Inc., Indian Point Nuclear Generating Unit No. 2

2. Docket No. 50-247 Technical Specifications as amended through Amendment No. 70 to Facility Operating License No. DPR-26 (Appendix A), Consolidated Company of New York, Inc., Indian Point Nuclear Generating Unit No. 2.

3. "Indian Point Cycle 5 Nuclear Design Report" by M, A. Kotun, L. L. Phelps, M. F. Muenks - WCAP-9881 May 1981.

4. The INCORE code by W. D. Legget, etal WCAP-7149 Rev., ..February 1972.

- 50-

abstract this report summarizes the results of the startup ... · 3.3 thermal power measurement 12...

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