ML20149M806

From kanterella
Jump to navigation Jump to search
Core Xiv Startup Physics Test Rept,Jul 1986
ML20149M806
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 07/31/1986
From: Stanford J
CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20149M795 List:
References
NUDOCS 8802290091
Download: ML20149M806 (30)
v • d • e


Text

[

t r

o e

I CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK, CONNECTICUT CORE XIV STARTUP PHYSICS TEST REPORT JULY 1986

/86 Prepared by:

h3 d

7 John t anford, Reactor Engineer

  • Approved by:

%bk d N on J6teph P. Drago,' Assistant Engineefddg Supervisor Reviewed by:

4.~ s-<

Lee-M -

7-G3-H Pierre F. L'Heureux, Engineering Supervisor 8802290091 860731 PDR ADOCK 05000213 P

DCD

xc.

A^ Mwt

,9,i '

- "q

.9 q

ff

_. 3

.1'-

t TABLE'0F' CONTENTS PAGE

.. INTRODUCTION...........'....................................

1 fl.

. 2. - ' CONTROL ROD DROP TIME MEASUREMENTS.........................

2

2.1_. COLD CONTROL ROD DR0P TESTS............................_2 2.2-HOT CONTROL ROD DROP TESTS.................'...........

3 l 3.--

.NEW CORE INITIAL APPROACH TO CRITICALITY...................

6 4.-

ALL RODS'0UT, JUST' CRITICAL'

. BORON CONCENTRATION..............................;.........

7 L5.

ISOTHERMAL TDfPERATURE COEFFICIENT MEASUREMENTS............

9 6.

CONTROL ROD BANK REACTIVITY WORTHS'AND EJECTED ROD W0RTN..................................................

11 6.1 INDIVIDUAL CONTROL ROD REACTIVITY TEST................

12 7.

BORON WOR 111................................................. 1 5 -

8.~

TWENTY-FIVE PERCENT. POWER FLUX MAP.........................

17 9..

EIGHTY PERCENT POWER FLUX MAPS.............................

19

10. FULL POWER FLUX MAP.......................<................

23

11. REACTOR CCOLANT SYSTEM FLOW TEST..........

25

12. REACTOR COOLANT SYSTD( RTD TESTING................,.........

28 9

e

  • 9 1

4 t

I i 4

'k, LIST OF~ TABLES TABLE PAGE 1.

HOT CONTROL RCD DROP TIME MEASUREMENTS.....................

5

'2.

DELAYED NEUTRON FRACTIONS..................................

8 3.

CONTROL ROD BANK W0RTHS....................................

13 4.

EJECTED ROD W0RTHS.........................................

14

-5.

INVERSE BORON W0RTH........................................

16 6.

SUMMARY

OF RESULTS FROM 25% POWER FLUX MAP CY-XIV-1

'l76...............................................

18 7.

SUMMARY

OF RESULTS FROM 80% POWER FLUX MAP CY-XIV-2-377...............................................

20 8.

SUMMARY

OF RESULTS FROM 80% POWEL FLUX MAP CY-XIV-3-378...............................................

21 9.

SUMMARY

OF RESULTS FROM 80% POWER FLUX MAP CY-XIV-4-379...............................................

22 10.

SUMMARY

OF RESULTS FROM 100% POWER FLUX MAP CY-XIV-5-380...............................................

24 11.

SUMMARY

OF RCS LOOP FLOW TEST..............................

27 12.

SUMMARY

OF RCS RTD TESTING.................................

30 l-t l

l

~

1l

~ ' 's yf[;

h i

m s

a c

+

s.

-/

c a

1.

INTRODUCTIO?p

/

/

/

p.

- Th'is report documents the Connecticut Yankee Core'XIV Startsp Physics Test program. -The testing sequence was completed as foll'owa:

Initial criticality.

May 6,-1986-5

'Zero power testing completed May 8, 1986 Turbine ~ phased to grid MaJ,10, 1986 802 perer flux mar- 00 pleted May 19, 1986 100% power, flux maps completed June 12, 1980 Core XIV is loaded with 153 stainless steel clad fuel assemblies and 4 zircaloy clad lead test assemblies located in core positions C8, N8, H13, and H3.

.t All acceptance criteria of the startup physics tests were met.

All

- requirements of the Technical Specifications were fulfilled. The reactor coolant flow measurement in the four loop configuration exceeded the minimum requirement. The tliree loop cont'iguration test did not meet the design analysis..The licensee has notified the NRC of the test results and han prohibited three loop operation pending review o1~ the design basis analysis. A cupplemental letter and revision of this start-up report will ise submitted following the resolution of the analysis.

l i.

L i

l l

L l

f' L

Page 1 of 30 I

hk.'

t s:

y.

.2N CONTROL ROD' DROP TIME MEASUREMENTS i

' 2.1 Cold Control Rod-Drop Tests-

' Objective The purpose of-the cold rod drop time measurements is to determine, prior to hot zero power conditions, if each of the 45' control rods will fall freely from its fully wichdrawn a

l position.

These rod drop time measurements are not a Technical Specification requirement. This test is done to identify any major rod drop problems before performing hot rod drop testing. The drop time is' measured for each of the 45 control rods. The drop time is the elapsed time for a control rod to travel from a fully withdrawn position to a fully inserted position.

t l

License Requirements There are no license requirements for cold rod drop testing.

Procedure l

l The rod position detector primary coil inputs for an entirc bank are connected to a Hewlett-Packard Multiprogrammer computer system. The bank is dropped using the MCB manual scram button. Pushing the scram l

button also triggers the computer system to begin collecting data.

The coil output voltage signals, which are a function of control ror velocity, are sampled by the computer system at one millisecond intervals for 3.5 secu;Js. These data are then analyzed by computer code which determines the time elapsed until the control rod strikes bottom.

Two graphs depicting the drop are also reproduced from the voltage signal for each rod.

Page 2 of 30 N9

!s.

u:

t Results The:results indicated thet_all 45 control rods were dropping satisfactorily..The cold rod drops were performed on May 4, 1986.

'The nominal'RCS pressure was 800 psig. The RCS' temperature was approximately 375'F.

2.2 Het Control Red Drop Tests Cbjective-The_ purpose of the hot control rod drop time measurements is to measure at operating temperature and pressure the time for each of the 45 control rods to travel from a fully withdrawn position to a fully inserted position.

. License Requirements Technical Spec'ification Section 4.2, Table 4.2-2, Operational Safety Items, requires the hot rod drop time for each control rod be determined each refueling.

Procedure The procedure described in Section 2.1 Cold Control Rod Drop Time Measurements is also applicable for the Hot dontrol Rod Drop Time Measuremencs.

Page 3 of 30

E~

.I

8

'n-

.Results All'43. control rods traveled from a fully withdrawn position to a fully. inserted position in less than 2.3 seconds with 4 RCPs operating, satisfying the Technical Specification requirement of 2.5 seconds. The hot rod drops'were' performed on May 5, 1986. The nominal RCS pressure was 1990 psig and the RCS temperature was 530 5'F.. The maximum drop time was 2.23 seconds (rod #18 of bank C) and the minimum drop time was 1.98 seconds (rod #3 of bank A).

The average drop time was 2.124 seconds; the standard deviation was 0.043 seconds. Data are presented in Table 1.

4 l

Page 4 of 30

C:;

~

V

.t kj,.

TABLE 1

, qb."

. CONNECTICUT YANKEE CORE XIV M

e, M

Rod'No.

Bank Core Location Drop Time (sec) 1, A

H8

~2.115 2-A K8 2.139

. 3 '-

A H6 1.984 A F8 2.055 5

A H10 2.160 6

D K6 2.154 7

D F6 2.173 8

D-F10 2.143

-:9 -

D K10 2.141 10.

B M8 2.115

'll B'

H4-2.045 12 B

D8 2.079 13 B'

H12 2.110 14~-

C M6 2.135 li C

K4 2.188 l16 C

F4 2.097 17 C

D6 2.135 18 C

D10 2.230 19 C --

F12 2.114 20 C

K12 2.104 21-C M10 2.146 22 A

N7 2.142 23 A

J3 2.124 241-A-

G3 2.137 25 A

C7 2.080

/

-26 A

C9 2.134 2 7._

A G13 2.082 28 A

J13 2.099 29 A

N9 2.180 30 B

M4 2.170 31 B

D4 2.124 32 B

D12 2.118 33 B

M12 2.095 34 D

NS 2.160 35 D

L3 2.101 36.

D E3 2.127 37 D

C5 2.133 38 D

C11 2.100 39 D

E13 2.101 40 D

L13 2.080 41' D

N11 2.137

.42 A

P8 2.200 43 A

H2 2.097

.i 44 A

B8 2.130

.45-A H14 2.117

_Page 5 of 30

C l

i 3.

NEW CORE INITIAL APPROACH TO CRITICALITY Objective The objective of the new core initial critical approach is to provide a safe and efficient means for achieving the initial criticality.

procedure At hot, zero power conditions, the reactor coolcnt system (RCS) boron concentration is first reduced from the refueling boron concentration to approximately 450-ppm above the predicted C/D/A @324 and B 0200 critical boron concentration.

1/M plots are maintained throughout the approach to criticality with a point plotted for each reactivity change. After the inititi dilution, control rod banks C, D and A are fully withdrawn and Bank B is withdrawn to 200 steps. The final approach to criticality is then made by additional RCS dilution and shimming of Bank B after the dilution has been terminated.

Results Core XIV initial boron concentration at hot conditions was approximately 2550 ppm. This concentration was reduced to 1997 ppm by adding demineralized water to the reactor coolant system. Control rod banks C, D, and A were then withdrawn to 324 steps and Bank B withdrawn to 200 steps.

Criticality was achieved at 0927 on May 6,1986 by adding approximately 9,540 gallons of demineralized water. The just critical conditions were 534*F, Bank B at 205 steps, and 1562 ppm boron. The corrected just critical boron concentration with Bank B at 200 steps and 535'F is 1558 ppm boron. The predicted boron concentration with Bank B at 200 steps is 1555 ppm. The difference between the corrected test data and the predicted critical boron concentration of 3 ppm is within the acceptance criteria of 100 ppm.,

Page 6 of 30

~ -

s-i e s

,4.

ALLRODSOUT,JUSTCRiTICALBORONCONCENTRATION Objective The objective is to measure the all rods out, just critical boron concentration at hot zero power conditions.

Procedure Bank B control rods are borated out to 324 steps. The remaining control rod banks are all fully withdrawn.

Results The all rods out just critical boron concentration was 1596 ppm at 535*F.-

The predicted hoe zero power boron concentration was 1609 ppm. The difference between the predicted and measured critical boron concentrations of 13 ppm was within the 100 ppm acceptance criteria.

Reactivity Computer Nuclear Instrumentation System channel 32A and 32B corresponding to the upper (uncompensated) and lower (compensated) ion chambers were used as inputs to the Westinghouse reactivity computer together with the hot zero power beta fractions for all zero power tests.

The reactivity computer was calibrated against several stable reactor periods varying from 50 seconds to 400 seconds.

The all rods out zero power beta fractions are listed in Table 3.

Page 7 of 30

p - = -

n f.

.1 s

TABLE 2

-CONNECTICUT YANKEE CYCLE XIV. DELAYED NEUTRON FRACTION AT 0:EFPD, ARO, HOT ZERO POWER GROUP BETA, eff LAMBDA, Sec ~I

.1 2~.0166E-4 1.28000E-2 2

~

1.2696E-3 3.15809E-2 3

1.1476E-3 1.21850E - 4' 2.4661E-3 3.20336E-1 5

9.0105E-4 1.37666E+0

- '6:

2.1882E-4 3.72999E+0 BETA (Total) =,6.49533E-3 Relative Importance (I) = 0.95527 BETA Effective = 6.2048E-3 Page 8 of 30

n a

5..

ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENTS Objective The objective.of these measurements is to (1) determine the isothermal temperature coefficient (ITC) for the new core at hot zero power conditions at two control rod configurations and (2) correct the ITC data at AR0/HZP to ARO/HZP/BOL, ARO/HFP/BOL and AR0/HFP/EOL moderator temperature coefficient (MTC) for verification of the Technical Specification limits.

License Requirements Technical Specifi~ cations, Section 3.16, Isothermal Coefficient of Reactivity, requires that the temperature coefficient be determined for a new core.

Procedure Hot zero power just critical conditions are established. The reactor-coolant temperature is then changed in approximately 5'F increments by controlling the atmospheric steam dump. The reactivity computer calculates the reactivity change due to the change in temperature and displays reactivity as a function of temperature on an X-Y plotter.

Temperature coefficients are obtained at the ARO and a rodded control rod bank configuration.

Results Four isothermal temperature coefficient measurements were obtained at

- the ARO configuration. The average isothermal temperature coefficient

. measured was -1.933 pcm/*F. The predicted value is -5.66 pcm/'F.

The difference between the measured and predicted temperature coefficients is within the acceptance criteria of 4 pcm/*F. This result was then converted to an equivalent hot, full power moderator temperature coefficient at BOL and EOL conditions.

Page 9 of 30

c.

i Results (Cont'd)

The hot full power BOL moderator temperature coefficient-(MTC) is-

-4.483 pcm/*F. The' predicted value is -8.21 pcm/'F.

The Technical

= Specification acceptance criteria that the BOL MTC shall be less positive than 5.0 pcm/*F was met.

The hot full power EOL MTC is -23.14 pcm/*F. The predicted value is -26.87 pcm/*F. The Technical specifica-tion acceptance criteria that the EOL MTC shall be less negative than

,-29.0 pcm/'F was met.

Five isothermal ~ temperature coefficent measurements were performed at hot zero power with Banks A, B, and D fully inserted and Bank C fully withdrawn. The average temperature coefficient measured was -12.6 pcm/'F.

The predicted value is -16.19 pcm/*F. The difference between the measured and predicted temperature coefficients is within the acceptance criteria of 4 pcm/*F.

-i 9

4 l

[

l l

l l

l Page 10 of 30

s 6.

CONTROL ROD BANK REACTIVITY WORTHS AND EJECTED ROD WORTH Objective The objective'of this test is to measure the differential and integral reactivity. worth of control rod Banks B, A and D, and the ejected rod worth of the two highest reactivity worth control rods in Bank B.

Procedure At hot zero power control rod banks are inserted into the core in small increments using the normal sequence for operating bank insertion.

Control banks B and A as well as shutdown bank D were measured. The

rate of insertion is controlled by the amount of reactor coolant dilution established by making up with demineralized water typically at 40 gpm. The reactivity of the core is continuously calculated and

-displayed on a strip chart by the reactimeter. The strip chart is then analyzed to datormine the reactivity worth of each control rod bank movement. The ejected control rod worths are determined either by pulling the single control rod or swapping the rod with a bank.

The ejected control rod worths are measured with the control rod bank's hot full power and hot zero ' power rod insertion limits.

Results The measured and predicted values for control rod banks and ejected rod worth are presented in Tables 3 and 4, respectively.

For Core XIV, the total rod worth measured was 3.62% less than the predicted worth. The results are within the acceptance criteria of 115% for individual control rod bank and 10% for total measured worth.

Page 11 of 30

T:

I

' The B&W Cycle XIV Design Report indicates that control rod (CR) in core location D4 (CR 31 in bank B) has the maximum ejected rod worths at BOC and EOC for both Hot Full Power (HFP) and Hot Zero Power (HZP)

- conditions. At the HFP control rod bank configuration, the measured worth of CR 31' was 24 pcm. The predicted worth was 26 pcm. This reactivity measurement met the acceptance criteria (1) the difference between predicted and measured is within 100 pcm and'(2) the measured worth is less than 170 pcm (Technical Specif.ications 3.10).

At the HZP control rod bank configuration, the measured worth of CR 31 was 230 pcm. The predicted worth was 301 pcm..

This difference meets the acceptance criteria (1) the difference between predicted and measured is'within 100 pcm and (2) the measured worth is less than 830 pcm (Technical Specification 3.10).

6.1 Individual Control Rod Reactivity Tests Objective The objective of this test is to verify that the control rod absorber section is connected to the drive shaft.

Procedure Af ter the All Rods Out Just Critical Boron Concentration procedure is completed, each control rod is inserted into the core until at least an 8 pcm change is observed on the reactivity computer. This is accomplished by disconnecting at the main control board panel all rods except the test rod in the test bank and driving the rod into the core.

The test rod is then returned to its initial position. The remaining control. rods are tested in this manner.

Results.

All control rods exhibited at least an 8 pcm reactivity change upon insertion into the core.

The acceptance criteria was met.

This is not a Technical Specification surveillance.

Page 12 of 30

~ ' ',,

.r

. TABLE 3 CONTROL ROD BANK WORTHS CONNECTICUT YANKEE CORE XIV I

CONTROL ROD BANK MEASURED WORTH, PCM PREDICTED WORTH, PCM-

% DEVIATION B

929 961 3.44 A

1929 2008 4.10 D

2114 2183 3.26 Total 4972 5152 3.62 5

i

' i j

% Deviation = Predicted-Measured x 100 i

Measured Page-13 of 30

u-

-4 I

Table 4 EJECTED ROD-WORTHS CONNECTICUT YANKEE CORE XIV.

1 Hot Full Power Rod Configuration-Rod No./ Bank / Core Location Reactivity Worth (PCM)

Measured Predicted 31/B/D 24' 26 t'

Hot Zero Power Rod Configuration

-Rod No./ Bank / Core Location Reactivity Worth (PCM)

Mee.sured Predicted 31/B/D-4 230 301 Page 14 of 30

s L7.0 BORON WORTH Objective

.The objective of this test is to measure the reactivity worth of the soluble poison in terms of pps/pem.

This is referred to as the inverse boron worth.

Procedure Reactor coolant.and pressurizer. boron samples are taken and analyzed at the.

equilibrium ARD and Banks B, A, and D inserted configurations.

The' critical boron concentrations are corrected for temperature and rod configuration.

The inverse boron worth is calculated by subtracting the

~

critical boron concentrations at the two control rod configurations and

. dividing by the sum of the bank worth inserted.

Results Table 5 presents the data of the predicted and corrected just critical boron concentration for the ARO and rodded configurations. The predicted and measured total reactivity worth of the banks B, A, and D is also shown.

T' Page 15 of 30

v,i

~

' Table 5' INVERSE BORON WORTH CORE XIV Measured Predicted

-ARO/HZP critical boron (ppm)'

1596 1609 Rodded HZP critical boron (ppm)

(Banks B, A, D inserted)'

888 864 Bank; worth (pcm)

(sua of B. A, D) 4972

~5152 Inverse boron worth (pps/pcm) 0.1424 0.1446 r

a L

i r

i 4

Page 16 of 30

4-

+..

9 8.

TWENTY-FIVE PERCENT POWER FLUX MAP-Obj ect'ive The objective of the nominal 25% power flux map is to determine if any gross neutron flux abnormalities exist.

License Requirements The flux map at 25% is not required per Technical Specifications.

Procedure One flux map is taken using the incore flux mapping system and evaluated using the INCORE computer code.

Results The results of the flux map ~ verified the operability of the flux mapping system and showed that no gross -core loading abnormality exists in the core. A summary of the results is shown in Table 6.

l k

i Page 17 of 30

u

- r Table 6 SURR1ARY OF RESULTS FR0rt INCORE RUN CV XIV-01-376 YJ POWER. 8 AT 320

' 8 tse/ttTU. 05/11/86 TAVG = 535.0 DEG F. BOROH Cot 4C. = 1462PPtt (3) 143 (II (2) ADJUSTED ALLOWABLE KWFT 153 (8)

EATCM AS$f.

FQH FXY2Q FZ34 FISIEFF F4Q3A0J KWFT KWF T HARGIN FIDEL HIN OtER 90 H08 1.326 1.058 1.254 1.042 1.539 8.54 14.30 67%

1.110 15.42387) 11A 208.

0.869 0.667 1.302 1.036 1.002 5.56 14.30 157%.

0.701 29.173873 84 J11 1.462 1.159 1.262 1.0 38 1.689 9.37 14.30 52%

1.217 16.52687)

ISA JRO 1.734 1.384 1.253 1.042 2.012 11.17 14.30 28%

1.453 13.682173 158 N08 1.454 1.142 1.273 1.036 1.677 9.31 11.00 18%

1.200 16.815471 16 N!!

1.810 1.397 1.296 1.036 2.087

!!.58 14.30 23%

1.466 13.337878 1.

FI5IEFF IS THE SPIKE FACTOR TO ttAXIt1IZE THE PRODUCT OF FtZI ASW FISI 2.

FtQlADJ 15 FQH ADJUSTED FOR THE EFFECTS DUE TO POWER SPIKE.ttEASURENENT UteCERTAlt4TY. A STATISTICAL FACTOR DOE TO DENSITY -

VARIATI0t45. THE EllGIt4EERItC FACTOR. STACK SHORTENIIG Ate THERMAL EXPAtlSION. THE POWER UtaCERTAlt4TT FACTOR. AtlD RADIAL TILT 3.

ADJUSTED FWFT IS FtQ3ADJa5.55 4.

SEE TECHNICAL SPECIFICATION 3.17 5.

FIDEL HIN IS FXYst.05 tielERE 1.05 ACCOUNTS FOR 5% UNCERTAINTYl.

6.'

PENALTT CORRECTION FACTOR It4CLUDED.1% PEHALTY IN DreR FOR EACH % QUALITY BELOW -15% TO -20%

'i 7.

OUT OF W-3 CORRELATIOtt RANGE O.

DtB CALCULATI0ftS ARE #30T QA CATEGORT 1 b

_,(h

&,. b c.

< 4l

=^"

l h

( n.

a.n.

dA sh/r&

k,.

t f

'Page 18 of 30' 1

g

q.

=..

{^

9.-

EIGHTY PERCENT POWER FLUX MAPS i

Objective The objective of the three 80% power flux maps is to confirm the predicted core power distribution and to establish th'e incore/excore axial offset correlation at 80% of rated power.

. License Requirements-Technical. Specifications, Section 3.17, Limiting. Linear Heat Generation Rate requires that the linear ' heat generation rate be determined from incore measurements.- Technical Specifications, Section 3.18, Power Distribution Monitoring and control requires that the power distribution measurement be performed before exceeding 80% of rated power.

e Procedure t

Three flux maps are performed at approximately 80% power. The measurements are adjusted to 100% of rated power and evaluated in comparison with predicted values.

A new incore/excore axial' of fset correlation is r

established based on these three maps.

The axial offset indication is calibrated and the y,wer distribution is evaluated prior to increasing power from 80 to 100%.

Results i

The results of the 80% power flux maps produced power distributions that compared well with predicted values and were within the Technical Specification 3.17 limit.

Based on the evaluation of the 80% power flux maps, power was increased to 100%.

A summary of the results is shown in Tables 7, 8, and 9.

i i

Page 19 of 30

s k'.

Tcble 7 i'.

~

StRitARY OF RESULTS FR0t1 It4 CORE RUH CY 'XIV-02-377 80% POWER. 8 AT 310.

43 t#40/NTU. 05/17/86 i

TAVG = 543.0 DEG F. BORON C0 tac. x 1263 ppt 1 (33 (4) i1)

-(2)

A0 JUSTED ALLOWABLE KW/FT-451 ISI BATCM ASST.

FQN FXYaQ FZaQ FISlEFF FtQlADJ KW/FT KW/FT MARGIN FIDEL HIN DNBR.

90 H08 1.242 1.054 1.179 1.036 1.432 7.95 14.30

'79%

1.106 8.06:173'.

11A 508 0.840 0.479 1.236 1.034 0.967 5.36 14.30 166%

0.713 12.940873 14 J11

1. 36 8 1.151 1.189 1.035 1.576 8.75 14.30'

~63%

1.209.

7.246t7)

ISA N09 1.643 1.363 1.205 1.034 1.892 10.50 14.30 36%

1.431 5.855178 e

158 M13 1.397 1.163 1.201 1.034 1.608 8.92 11.00 23%

1.221 7.114674 16 H11 1.673 1.366 1.224 1.034 1.926 10.69 14.30 33%

1'.435 5.76247) 1.

FtSIEFF IS THE SPIKE FACTL9t TO MAXIMIZE THE PRODUCT OF FtZ) AND FIS)

~

2.

FielA0J 15 FQN A0 JUSTED FOR THE EFFECTS DUE TO POWER SP!KE.ttEASUREt1ENT 124CERTATHTY. A STATISTICAL F ACTOR DUE TO DENSITY VARIATIONS. THE EIGIt!EERitG FACTOR. STACK SHORTEHitG AND THERMAL EXPAttSION. THE POWER UNCERTAINTf FACTOR. Aho RADIAL TILT a

i 3.

ADJUSTED KW/FT IS FlQtADJ=5.55 4

SEE TECHt41 CAL SPECIFICATI0t8 3.17 i

5.

Ft0EL Hlt4 IS FXY=1.05 EL44ERE 1.05 ACCOUNTS FOR 5% UNCERTAINTV 3.

]

6.

PENALTY CORRECT 10tt FACTOR ItaCLUDEO.1% PENALTT IN OHBR FOR EACH % QUALITY BELOW -15% TO.-20%

j 7.

DUT OF W-3 CORRELAT30tl RA! IGE j

O.

095 CALCULATI0t:5 ARE taci QA CATEGORY I a *

//'

PREPARED 3Y 1

l t Ac

(<,c/_ (4 DATE i

z REVIEWED BY

/)

~

W&

L- @

$ f Rudw

{

DATE

[

ro e 20 oe 30:

s

t Table 8 SuttARY OF RESULTS FROt1 INCORE fiul CY

  • XIV-03-378 8'lZ POWER 8 AT 290.

67 tEJD/HTU. 05/18/86 TAVG = 543.0 DEG F. BOROH CONC. * !!69 PPM 438

-(48 (18 428 ADJUSTED ALLOWADLE KW/FT ISS (Si BATCH ASST.

FQH FXf3Q FZ3Q FESIEFF F(QlA0J KW/FT KW/FT MARGIH FIDEL HH4 DIER

~

9D H08 1.264 1.066 1.186 1.035 1.457 8.09 14.30 76%

1.120

.7.999(73

!!A RCS 0.851 0.687 1.240 1.034 0.980 5.44 14.30 162%

0.721 12.774878 14 Jll 1.387 1.161 1.195 1.035 1.597 8.86 14.30 61%

1.219 7.18*(7)

~

15A tto?-

1.674 1.371 1.221

1. '32 1.924 10.68 14.30 33%

1.440 5.778478 155 H13 1.422 1.174 1.212 1.034 1.637 9.09

!!.00 21%

I.233 7.036(78 16 H11 1.709 1.388 1.231 1.032 1.9%

10.90 14.30 31%

1.458 5.650878 1.

FISIEFF IS THE SPIKE FACTOR To ttAXIMIZE THE PRODUCT OF FIZ3 Ate FESS 2.

FtQlADJ IS FQtt ADJUSTED FOR THE EFFECTS DUE TO POWER SPIKE.ttEASUREtttfiT UNCERTAINTY. A STATISTICAL FACTOR DUE TO DENSITY VARIATI0 TIS. THE EtlG1HEERIt G FACTOR. STACK SHORTEH193G ABC THERNAL EXPANSION. THE POWER UNCERTAIllTY FACinR. AND RADIAL TILT 3.

ADJUSTED KW/FT IS FtQlADJa5.55 4.

SEE TEClititC AL SPECIFICATION 3.17 5.

FEDEL H H4 IS FXYal.05 tutERE 1.05 ACC004TS FOR 5% UNCERTAINTYl.

6.

PEHALTY CORRECT 1018 F ACTOR ItaCLUDED. 1% PEllALTY IN OtER - FOR EACH % QUALITT BELOW -15% TO -20%

7.

OUT OF W-3 CORRELAT1084 RANGE C.

DND CALCULATlotas ARE NOT QA CATEGORY I y

PREPARED BY

.//-

x<

tr~

. (L, r e (ii

/

DATE 4

i 8

REVIEWED BY Tkn n<a

/?vS.. N has~

i

~

i3 Cft WY&

y

,qlg v

Page 21 of 30

,g D

w..

..L Tabla 9 Sutt1ARY OF RESULTS FR0tt INCORE RUN CY.XIV-04-379 80% POWER. 8 AT 270.

89 tSO/t1TU. 05/19/86 TAVG = 545.3 DEG F. BOROH C0 tac. = 1199pPrt (33 (4)

(1)

(2) ADJUSTED AELOWADLE kW/FT (5) tot 8ATCH A55f.

FQte FXYaQ FZ3Q F(SIEFF FtQlADJ KW/FT kl4/F T ttARGIN FIDEL HIH Ot40R 90 H08 1.276 1.074 1.188 1.033 1.468 8.15 14.30 75%

1.128 7.885478 11A ROS,

0.862 0.680 1.267 1.026 0.985 5.47 14.30 161%

0.714 12.658471 16 J11 1.398 1.164 1.201 1.033 1.607 8.92 14.30 60%

1.222 7.066t73 ISA N09 1.678 1.359 1.235 1.032 1.929 10.*J1 14.30 33%

1.427 5.65447) 158 H13 1.429 1.171 1.220 1.032 1.642 9.11 11.00 20%

1.230 6.88117) 16

'H11 1.716 1.385 1.239 1.029 1.%7 10.92 14.30 30%

1.454 5.540873 1.

FISl[FF IS THE SPIKE FACTOR TO ttAXIttIZE THE PRODUCT OF FtZI Ate FIS) 2.

FtQlADJ IS FQtt ADJUSTED FOR THE EFFECTS DUE 10 POWER SPIKE.itEASUREttENT UtaCERTAltITV. A STATISTICAL FACTOR DUE TO 0[14SITY VARI ATIOt35. THE Et3 git 4EERIt4G FACTOR. STACK SHORTErili4G Als THERt1AL EXPAllSION. THE PCHER Ut%ERTAINTT FACTOR. Als RADIAL TILT 3.

ADJUSTED kW/FT IS FtQlADJ=5.55 4

SEE TECHt4ICAL SPECIFICATI0tt 3.17 5.

FIDEL HIH IS FXYel.05 (14 TERE 1.05 ACCOUt4TS FOR 5% UNCERTAINTY).

6.

pet 4ALTY COWRECTI0ta FAC10R It4CLUDED. 1% P[NALTV IN DilDR FOR EACH % QUALITY EELOL4 -15% 10 -20%

7.

OUJ OF W-3 CORRELATI0t4 rat 3GE 0.

Ole CALCULATIO!!S ARE NOT QA CATEGORY I f

7 l

  • 7 '

PREPARED BY

/

A \\e

. n:. m/ \\

c M

OATE I

/

q REVIEWED BY l}"

j}

QL1kle (Dfu kh hw

~

05 J.;-//f/f/,

i a

Page 22 of 30

.-L-d

4

" 10. FULL' POWER FLUX MAP Objective The objective' of the full power flux map is to confirm the predicted core power distribution and to verify the.incore/excore axial offset correlation

~ determined at 80% power.

License Requirements Technical Specifications, Section 3.17, Limiting Linear Heat Generation.

Rate requires that ' the linear heat generation rate be determined from incore measurements.

Technical Specifications, Section 3.18, Power Distribution Monitoring and Control requires that the power distribution be measured and an incore-excore axial offset correlation /calfbration is performed.

Procedure A flux map is taken at full power.

Excore readings are also taken during the' flux maps.

Af ter the evaluation of the flux map the incore/excore axial' offset correlation is verified.

Results An incore flux map was performed at 100% power.

Data generated by this flux map were used in the evaluation of the incore-excore correlation. All incore results were within the Technical Specification 3.17 limit.

A summary of the results is shown in Table 10.

I-

?.

TL YI TT I

n 5

)

3 1

)

)

SL 6

7 6

6 6

6 NA 3

8 6

t t

1 t

1 DD EI u

8 9

4 4

1 8

4 w.

R.

1 6

3 2

9 6

A iB.

t.

3 9

0 7

9 OR a

T D.

5 9

4 4

4 3

D o

EN C(

UA g

N W

D I

SI RR H.

OO i

I 0

4 7

1 7

2 TT y

L 2

2 0

2 2

2 CC W

1 7

2 4

2 4

AA E.

D.

FF t

1 p

1 1

1 1

LY 9;3 F

2*,

AT D

D, u

CI I

9 4

/

h d

II 0

7 /

TH.

TA 2

/

FI.

% r ST IR 7

pM r

5 9

3

/G.

q i

A w

2 TE O

WR.

6 3,

2 AC T

t KA.

/

p TN 12 Y s SU 5

1 E AR 1

4L f

.E

.?

M0 W

4 OT t

YO W

h/,

AF

. W/

TP O

4 L

OW t J, t

1 lE E

LK.

t L

AH D

A TT R

Y 0

E.

T 0

l Cl I

3 8

) D 3ET.

4 6

3 9

0 4

NO L

Y 3

I A

B 9

3 6

2 9

4 8 T F.

L S

U f

5 6

S/

0 8

UW 7

5 8

0 8

0 TN Q

D o

/

JK.

1 1

NA R'

E 0

V 3

D I

EP 4

X Z

A I

2 i

A S

1 1

X

/

r I

EE P

c 2

t t

5 P

F R

l E

e 0

2

) J UL i

R e

l Y

5 20.

0 6

5 4

3 0

e SA C

P g

1 A.

3 6

5 5

0 8

At

)

A b

C 1

( l a

t U

1 4

9 5

8 6

8 A

ER Y

E E

T P

c N

T Q.

N. H N

R T

U 1

=

t 1

0 1

1 1

1

)

1 P

/

F Z

ET I

O D

f K

A F

C F

Ie T

E W

lt R

H

)

Pr R

O 0

1F F

SA E

R C

4 C

8 F 4

5 2

1 6

5 O

C E

I 9

E 3

2 3

3 2

2 RG N

3 P

I 1

H I

0 0

0 0

0 0

T E 1 U

O C

W1 O

S.

t R

I 1

1 1

1 1

1 U

Ot3 N

to O

F D

PE 5

I R

B O

T F

3 R

OR R

Y 0

P TO O

T 5

3 F

H F

L T

5 8

7 3

7 E

E5 A

L T

G Q.

6 2

7 9

0 1

H U

S 4

t 1

2 1

1 2

2 T

DK T

E U

A E

3.

C N

P S

D Z.

E 8

F.

1 1

1 1

1 1

E SA U

R 0

Z TT O

I I

CS C

1 F

R 6

t

[

C Y

1 O

E 5

I F.

A R

W 5

X FR D

O Y

O Q.

6 0

9 3

9 5

A EO 5

E G

t T

7 1

6 9

4 5

6 5

0 D

[

R P

a a.

U T

0 6

1 3

1 3

EC 1

A Y

o HA 5

1 L

A G

X 1

g t

5 3

C C

t 0

V F

1 0

1 1

1 1

T TF E

N E

U 0

A S

1 T

R RG 5

N R

I G

A T

FI J

I c

P t

Q O

ON

=

O

[

l A

a C

R D

T O

R T

l 3

7 3

5 3

8 A

DE A

A f

T 0

I H.

4 4

5 1

0 4

F EE l

C C

8 T l Q

I 5

A C

2 8

3 6

4 6

Q.

l E

SI t

F F

I E

0 F

1 0

1 1

1 1

K UC T

R T

I 4

I Jt C

l A

A t

P DE S

E e

0 L

S S

A I

P T

I E

It E

S X

T R

E H

T F

C R

0 l

f H

CT F

L E

O I

S S

1 9

8 1

T F

/

A 5

R C

T Y

O O

1 0

0 1

. u C

1 R

A 55.

H R

J t

N N

S S$

p I

O 3

L t

O D

H i

C U

a i

N A.

I I t t

W O

F JI E

C H

Y L

F DT T

E T

F A

E A A S

T L

L O

C I

l I U

E A

H S

eR J

E D

T S

t I

iA D

E I

E U

C A

A 8

T 0

1 4

S 5

6 F

FV A

S F

P D

D A

9 1

1 I

1 1

B 1

2 3

4 5

6 7

8

7

. a

11. REACTOR COOLANT SYSTEM FLOW TESTS Objective The purpose of the reactor coolant system flow test is to measure the total vessel flowrate (including core bypass flow) at rated power operating conditions. A precision heat balance was used and the results were corrected for all measurement uncertainties.

License Requirement Technical Specification Section 3.20, Reactor Coolant System Flow, Temperature and Pressure, Specification D requires that the reactor coolant system flowrate be determined by a he.at balance within 7 EFPD of achieving 100% rated thermal power af ter refueling.

This specification applies to the 4 loop configuration only.

There is no technical specification for 3 loop reactor coolant system flowrate.

The 3 loop RCS flow test was performed to verify the design basis assumption.

Procedure A precision heat balance wts established for each loop using the steam generators as the control volumes.

The following parameters were sampi,ed at a frequency of not less than once per minute for a period of one hour.

- reactor coolant system pressure

- hot leg temperatures

- cold leg temperatures

- feedwater temperatures

- feedwater flow rates

- feedwater pressure

- steam generator pressures Since steam generator blowdown flow is not directly measured, it was isolated during~the period of data acquisition.

The above data were used to calculate the following time averaged parameters for each of the four loopst steam generator heat transfer rate, primary coolant enthalpy change and cold leg specific volume.

The hot leg temperatures were corrected for stratification effects.by using empirically derived worst case stratification values.

The 4 loop test was conducted at 1818 MWeh, 562,0

'F average coolant temperature and 2030 psig average coolant pressure.

Page 25 of 30

s' s,

e The 3 loop configuration test was performed on June. 25, 1986 with loop 3 idled.

Since loop 3 demonstrated the largest loop flow, this 3 loop configuration assured that the minumum total RCS flow was achieved.

The 3 loop test was conducted at 1171.5 MWth, 535'F average coolant temperature and 2028 psig average coolant pressure.

An uncertainty analysis was performed for this measurement in accordance with NUREG/CR-3659, "A Mathematical Model for Assessing the Uncertainties of Instrumentation Measurements for Power and Flow of PWR Reactors." The analysis considered the effects of all sources of uncertainty in each instrumentation loop which was used.

The flow uncertainty value for the i

four loop configuration was established to be 2.963 % of the nominal flowrate 4.568 % for the three loop configuration.

Results l

l Four Loop Test The best estimate reactor coolant system flowrate was determined to be 267,067 gpm.

Corrected for measurement uncertainties, the flowrate was established to be 259,154 gpm.

This flowrate is 2154 gpm (0.84%) greater than the minimum value of 257,000 gpm, as required by Technical Specification 3.20.

Rated thermal power was reached at 6. 2 EFPD.

The reactor coolant system flow testing was completed at 10.1 EFPD, thus the flow test was conducted 3.9 EFPD af ter 100% rated thermal power operation was achieved. The flowrate test data is summarized in Table 11.

Three Loop Test The best estimate reactor coolant system flowrate was determined to be 212,214 gpm.

Corrected for measurement uncertainties, the flowrate was I

established to be 202,520 gpm.

This flow is 4980 gpm (2.4%) less than the minimum value of 207,500 gpm, as assumed in the safety analysis.

The NRC l

was informed via the 10CFR 50-72 immediate reporting requirements.

The licensee has prohibited three loop operation pending review of the saf ety analysis.

A supplemental letter and revision of this start up report will l

be submitted following the resolution of the analysis.

L nn m aua

c:

N-

.O

.A-i~

'8 r

Table 11-REACTOR COOLANT SYSTEM FLOW TEST RESULTS

'4 LOOP CONFIGURATION

-STRATIFICATION LOOP-LOOP -

-AT('F)-

CORRECTION (*F)

FLOWRATE (GPM)'

1 47.20'

+ 1.05 67225 2

47.20

+ 0.65 66587 3

45.95

+ 0.'4 5 67511

~

4 48.30

+ 1.05 65744 Total RCS Loop Flowrate 267,067 gpm 2 903% Uncertainty Penalty 7913 gpm Minimum guaranteed Flowrate 259,154 gpm

-3 LOOP CONFIGURATION STRATIFICATION LOOP LOOP AT(*F)

CORRECTION O FLOWRATE (GPM) r

'l 39.25

+ 0.10 70771 2

39.80

- 1.15 71710 3

IDLED 4

40.05

+ 0.10 697.33 Total RCS Loop Flowrate 212,214 gpm 4.568% Uncertainty Penalty 9694 gpm

' Minimum guaranteed Flowrate :

202,520 gpm Paae 27 o2 25)

2. -: s

. j;

(

sv

^

12.0 REACTOR COOLANT SYSTEM RTD TESTING Objective The purpose of this testing is to verify that the reactor coolant syscem temperattPs measured by the new RTDs are within the operating envelope which was used to establish the new variable low pressure trip constants for Cycle 14 operation.

i License Requirements As a condition of the Technical Specification Change Requesc for the Cycle la relo..', CYAPCo committed to documenting the RTD testing results in this rtport.

Background Information ww RTDs were installed in each reactor coolant system cold leg and hot leg. Two narrow range and one wide range RTD are located in each leg.

In addition, all cold leg RTDs were r elocated. from the suction to the discharge side of the reactor cociant pumps in order

  • o eliminate identified stratification.

Prior to the relocation of t'..e RTDs, it was predicted that measured rated po'er average temperature and steam generator delta temperature would both increase by five degrees for the same plant operating conditions.

Both the average coolant temperature and delta temperature are inputs to the reactor protection system.

L Variable los usure f '/i" protection is provided by comparing a calculated r

pressure v 100 and N # pressurizer pressure. When the se point is less the

-a cn e in two of three channels, a reactor trip is initiat-v i 10%).

The setpoint is calculated as a function o'

-~;-

c system average temperature and steam generator de1*a temperature.

Pags 28 of 30

~

M

. p

.s *

i.: 6.: e

~

4,'

^ha

~A Technical Specification Change Request 'was' submitted to the Staff to support the Cycle 14 reload.

This submittal included a request to allow CYAPCo to change the constants associated with the variable low pressure trip.. The. RfD testing program insured that ~ the measured temperatures at 2

power ' provided. conservative calcultted variable low ' pressure trip setpoints.

Specifically, the measured teactor vessel temperatere must be greater than or equal to 45 degrees at' rated power.-

Procedure 4

Holdpoints were established for the initial power ascension to rated power.

Data was gathered'at 70%, 80%, 90%, 95% and 100% of rated power. Data-was evaluated for acceptability prior to continuation of the power ascension.

l~

Soon af ter plant 'startup, additional testing was performed 'in three loop operation while. operating at 65% of-rcted power.

At each holdpoint, the fo'llowing parameters were measured and r(corded:

~

i

- Variable low pressure trip delta temperature input signal (voltage)

- RTD resistances

- Locp Tavg, delta temperature and cold leg temperature as indicated on the main control board.

Results All data gathered during both intermediate and. rated power operation m'et the sceeptance criteria.

The r.eu variable low pressure trip constants provide conservativs reactor protection for both four and three loop operation at all power levels.

The testing results are summarized on the following page in Table 12.

Nate that only the most limiting temperatures and. voltages are tabulated in Table 12.

s

.t.

Page 29 ef 30

f:;;.

y pw

4 y4.

. r.

y, Table 12

SUMMARY

OF RCS RTD TESTING Y

Minimum Required VLPT Input

_RTD Delta

' Power Level Delta Temp / Voltage Voltage ~

Temp.

66%

23.0 / 2.097 2.84 32.5

.79%

30.0 / 2.736 3.47 37.8 90%-

35.0 / 3.192 3.74 41.9 95%

38.0 / 3.470 3.96 44.3 100%

45.0 /.4.104 4.14 45.8 i

65.0%

34.5 / 3.150 3.53 38.6 (3 loop)

Page 30 of 30 L

_ _ _.

Retrieved from "https://kanterella.com/w/index.php?title=ML20149M806&oldid=4164279"