Text

Special Report No. 9, Palisades Cycle II Start Up Report
	ML18354A360
Person / Time
Site:	Palisades
Issue date:	08/27/1976
From:	; Consumers Power Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18354A360 (35)
	v • d • e

SPECIAL REPORT No. 9 PALISADES CYCLE II START UP REPORT - August 27, 1976 rec'd w/ltr dtd 8-30-76 CN# 8984 50-255 NOTICE -

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PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) FROM DOCUMENT FOR REPRODUCTION MUST BE REFERRED TO FILE PERSONNEL.

DEADLINE RETURN DATE RECORDS FACILITY BRANCH

SPECIAL REPORT NO 9 CONSUMERS POWER COMPANY PALIS.ADES CYCLE II ST.ART-UP REPORT

(.-- --:--..._\\.j

. August 27, 1976 Jfcc6?/r; //J/J P.-_?o-- ?'(;

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PURPOSE.AND SCOPE The purpose of this report is to fulfill the requirements of Paragraph 6.9.1.a of the Palisades Technical Specifications.

The report describes the results of test-ing performed at zero power and during power escalation at the start.of the second cycle of operation.

The following plant operating para.meters were measured during the course of the testing.

1. Critical boron concentrations.

2.

Control rod bank worths in nonoverlapping sequence.

3.

Regulating control rod bank worth in overlapping sequence.

4.

Dropped rod worths.

5.

Ejected rod worths.

6.

Boron differential worth.

7.

Moderator temperature coefficients.

8.

Power coefficient.

9.

Shutdown margin.

10.

Power distributions.

11.

Incore detector performance.

Test Instrumentation Reactivity measurements at zero power were made through use of a reactivity com-puter.

The scaler output from the two intermediate range power channels was summed and put into ~wide range linear channel.

This channel provided a 0-10 V linear input for the reactivity computer.

See the schematic diagram shown as Figure 1.

The reactivity computer uses precalculated 6-group delayed neutron fractions and decay constants to translate variations in input signal to reactivity in cents.

A multichannel recorder was used to simultaneously record flux level, control rod position, average moderator temperature, boron concentration (from the boronometer) and reactivity.

Upon reaching criticality, the reactivity meter response was checked against stable period measurements from the plant log rate recorders.

The resulting response as a function of reactivity is shown in Figure 2.

Most of the incremental reactivity measurements were in the range of -5 to +5¢ with the notable exception of the shut-down bank scram test which is mentioned later.

The random noise level in the reactivity meter trace amounted to as much as 2.5¢ peak to peak.

1

./

Measurements

1. Critical Boron Concentration

2.

With the reactor critical and all rods out except Group 4 at 100" withdrawn, the boron concentration was measured as 1205 ppmB.

The Group 4 rods were withdrawn all the way out with a corresponding reactivity change of +9¢ or 0.062% (Seff = 0.0069).

Using a reciprocal boron worth of 76.5 ppmB/%L\\p (as determined later in this report), the reactivity change corresponds to 5 ppmB.

Thus, the critical boron concentration is 1210 ppmB for all rods out (ARO), which* is within.!. 120 ppmB of the predicted value of 1222 ppmB as required by the acceptance criteria.

Dropped Rod Worths With the reactor critical and at an ARO configuration, dropped rod worth measurements were performed.

A constant dilution rate was established and the reactivity was permitted to increase slightly (~4¢). The reactivity was then made slightly negative by inserting rod 8 several inches.

The reactivity changes for each rod insertion were measured on the reactivity computer and summed for a total rod worth.

With rod 8 fully inserted and the reactor.

stable (dilution stopped), rod 6 was "traded" with rod 8 by withdrawing rod 8 and inserting rod 6.

This "tradingf' was done with all the control rods to be measured with the results listed in Table 1.

The dropped rod worths are all less than 0.2%p as required by acceptance criteria.

Table 1 - Dropped Rod Worths Control Rod 5

6 7

8 12 20 26 31 35 37 38 39 40 41 Measured

Dropped Rod Worth, %llp*

.lll

.110

.124

.llO

.047

.104

.095

.109

.076

.083

.099

.109

.094

.112 Exxon Calculated Dropped Rod Worth, %llp

.1125

.0693

.0974

.0904

.1095

.1035

.0803

.1044

- - 1004

.1004

Acceptance Criteria:

Measured Dropped Rod Worth <0.2%

2

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3.

Moderator Temperature and Pressure Coefficients c***)

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e With all rods out, the system was heated and then cooled at a constant rate with corresponding reactivity changes recorded*.

Reactor power was kept con-stant with Group 4 rods.

From this data, the moderator temperature coefficient (MTC) was calculated.

Later in the test, the same measurement was made with Groups 3 and 4 fully inserted and Group 2 at approximately 80 inches.

The results are listed in Table 2.

There was an observed bias between the mea-surements and calculations of about 0.27 x 10~4~p/°F. This was not unexpected as a similar bias was noted at the start of Cycle I in the first core calcu-lations.

The hot zero power measurement (HZP) was corrected to hot full power (HFP) using the method described in the Palisades Technical Specifications Section 3.12 (see Table 2).

The MTC is required by the acceptance criteria to be greater than -2.5 x 10-46p/°F and less than +0.5 x lo-4.

As seen in Table 2, this condition was met.

With all rods out and the reactor stabilized, the primary system pressure was decreased 74 psi and a reactivity of increase of 1¢ was measured, resulting in I

a negative pressure coefficient opposite in sign to the moderator temperature coefficient as required by the acceptance criteria.

Table 2 - Moderator Temperature Coefficient (MTC)

MTC MTC Exxon MTC(l,2)

Boron Temp Reactivity Measured Calculated Corrected Concentration-Change Change (HZP)

(HZP) 10-4~p/°F Measured EEmB OF lo-4 6p 10'-46p/°F 10-4~p/oF 1210 (ARO)

-* 9.1 1.9

.209

-.062

.23 1210 (ARQ)

-11.3

-2.24

.198

-.062

.22 1080 (Grp 2,3,4 in)

-12.4 1.9

-.153

-.427 1080 (Grp 2,3,4 in) 12.5

-1.73

-.138

-.427

1.

From Palisades Technical. Specifications

-4

-6 MTC (HFP) = MTC (HZP) + 0.1 x 10 x 0.16 x 10

- 0.11 x 10

- 0.15 x 10

(~ppmB) - 0.23 x 10-6 (6°F)

= MTC (HZP) +.021 x l0-4/°F.

Where 6ppmB is the difference in the. critical. boron concentration (ARO) between HFP and HZP (no xenon or samarium).

It is equal to the power defect at 100% FP expressed in ppmB.

Using the measured reciprocal. boron worth of 76.5 ppmB/%p and the calculated power defect of 0.88%6p, ~ppmB = 67.32 ppmB, where

~°F is the difference in Tav between HFP and HZP = 544°F - 532°F = 12°F.

2.

MTC acceptance criteria:

-2.5 x io-4~p/°F < MTC (measured,. corrected)

< +.5 X l0-46p/°F.

3

4.

Regulating Rod Worth in Overlapping Sequence With the reactor stable, the four regulating groups were then inserted into the core in the.sequential overlapping mode.

Insertions were performed by diluting in the groups in the same manner as was rod 8 in the dropped rod measurements.

The reactivity change at each rod insertion was measured and the results compiled to give the integral worth curve in Figure 3.

The maxi-mum reactivity insertion rate for the regulating groups in the sequential mode was determined, using Figure 3, to be 0.26 x lo-4 Lip/s for the rods above the full power insertion limit (Group 4 at 97") and 0.87 x lo-4 Lip/s for the rods above the zero power insertion limit (Group 2 at 75 11 ).

These numbers are based upon a 46 inch-per-min rod drive speed and are less than the maximum rates of 6.0 x lo-4 6p/s and 3.0. x lo-4 Lip/s, respectively, as set forth in the acceptance criteria.

5.

Ejected Rod Worth With Groups 3 and 4 fully inserted and Group 2 at approximately 80", ejected rod worths were measured by berating out rod 41 and then "trading" with rods 34 and 35.

The results listed in Table 3 show that the ejected worths were less than 0.963%6p as required by the acceptance criteria.

Table 3 - Ejected Rod Worth Control Rod 34 35 41

Measured Rod Worth, %Lip

.097

.129

.626

Acceptance Criteria: Measured Ejected Worth< 0.963%Lip

6.

Reciprocal Boron Differential Worth During the physics test, chemical samples wer.e taken before and after all dilutions and borations in order to determine reciprocal boron worths.

The results are listed in Table 4.

The reciprocal boron worth calculated for dilution of the regulatory group in the sequential. mode is considered to be the most accurate since this dilution produces the greatest change in boro~

concentration and reactivity.

Thus, the reciprocal boron worth is taken as 76.5 ppmB/%Lip.

Note that all of the values were less than 100 ppmB/%6p as required for acceptance.

The calculated value of reciprocal boron worth is 86 ppmB/%Lip.

4

*---*---. ____________ _:_~ __ __, ____. ____. _____.: __ ----.

Table 4:. Reciprocal Boron Worths*

Total Reciprocal Initial Final Reactivity Boron Worth Case EPrnB PErnB

%tip 1212mB/%6p Rod 8, Dilute In 1205 1198

.11 63.6 Rod 41, Borate *out 1075 1120

.626 71.9.

Group 1, Borate Out 890 1035 1.825 79.5 Group 2, Borate Out 1035 1095

.799 75.1 Group 3, Borate Out 1095 1160

.868 74.9 Group 4, Borate Out 1160 1205

.63 71.4 Reg Groups, Sequential, Dilute In 1205 900 3.985 76.5 Group A, Borate Out 1095 1217 1.32 92.4 Group B, Dilute In 900 780 1.463 82.0

Acceptance Criteria:

Measured Reciprocal Boron Worth < 100 pprnB/%6p

7.

Individual Control Rod Bank Worths Starting at all regulating rod groups inserted, Group B was diluted in and its integral worth was compiled as shown in Figure 4.

With the regulating groups and Group B fully inserted, the worth of Group A was measured by manually tripping and dropping the group into the core.

Following the Group A drop, ihe reactor was borated to the critical boron concentration for the regulating groups inserted', Group A withdrawn, and Group Bat 120".

Groups A.and B were then withdrawn and the "reactor was brought critical.

A constant boration rate was established and Group 1 was withdrawn from the core in the sequential nonoverlapping mode.

An integral group worth as function of rod insertion was compiled and shown in Figure 5.

A constant boration rate was again set up and Groups 2, 3 and 4 were with~ravm in the nonoverlapping mode in the same manner as Group

1.

The group worths are listed in Table 5 and integral worth curves shown in Figures 6, 7 and 8.

A summary of the group rod worth measurement results are presented in Table 5.

All of the values in the column labeled 11Measured Worth" were measured by the reactivity computer in small increments using the boron trading method except shutdown Group A.. The worth actually recorded by the reactivity computer during the rod bank scram test was -2.4%6p, a deviation from the expected value of -3.37%6p of 29%.

This difference is much larger than observed in any of the measurements using the incre-mental bo~on-trading measurement method, and was attributed to measurement Uncertainty rather than calculational error for reasons discussed below.

. As a check on the calculation for this rod group, its worth was measured out of sequence using the boron trading method.

This measurement deviated from a corresponding calculation by.o6%6p or 4.3% of the calculated value.

5

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Table 5 - Measured Group Rod Worths Measured Calculated Difference/

Control Rod Worth Worth Difference Calculated GrouI!

%tie

%tip

%tip Worth 2 %

4

.63

.70

-.07

-10.0 3

.87

,95

-.08

-8.4 2

.80

.72

+.08

+11.1 1

1.82 2.20

-.38 I

-17.3 Total Regs 4.12 Regs (Overlap) 3.98 Regs (Average) 4.05 4.57

-.52

-11.4 B

1.46 1.55

-.11

-7.1 Regs + B 5.51 6.12

-.63

-10.3 A

3.02*

3.37

-.35

-10.4 All Rods 8.53 9.49

-.98

-10.3 A (Out of Sequence) 1.32 1.38

--.06

-4.3

The Group A worth was estimated f!om the calculation and the total observed calculation/measurement difference for Groups 1, 2, 3, 4 and B.

3,37 x (1-.1ci3) = 3.02 The differences between predictions and measurements of the regulating rod banks were somewhat scattered with a maximU11l individual deviation of 0.38%tip (17.3%).

The overall difference of all the rods measured in sequence (regs plus B) using the incremental method coI!lIJared to the calculation was 0.63%tip or 10.3% of the cal-

culated value.

This bias of 10.3% was applied to the results of calculations to estimate the worth of shutdown Group A in sequence.

Both the maximum difference between the measurements and predictions of individual rod bank worth and overall rod worth were slightly outside the predetermined ac-ceptance criteria of 15% and 10%, respectively.

Accordingly, an evaluation of the accuracy of both measurements and calculations was undertaken to determine whether or not there was an impact on plant safety.

Sources of error in the measurement include:

6

(

a.

Errors in meter calibration (see Figure 2).

b.

Stated experimental errors in the values of delayed neutron fraction for uranium ( -v4%) *

c.

Meter noise and error introduced by manual reduction of strip chart data.

d.

Point-kinetics approximatioµ inherent in reactivity measurement technique.

This theory assumes that the power distribution in the core is constant during reactivity changes and the detector output is exactly proportional to the average flux in the core (detector sees the core as a point),

neither of which was the case.

The ability to check the measurements with boron endpoints is limited to the accuracy of the boron sampling and chemical analysis technique.

Based on the above, it is unreasonable to assume an experimental error less than + 10%.

Since the quoted accuracy of the calculations is + 10%, differences between the two of up to 20% for individual measurements can be expected.

Concerning the 17.3%

error in regulating Group 1, an independent c-alculation by Con~mmers Power of the worth of Group 1 indicates that the original predicted worth could be high by 0.16%~p. Inspection of the reactivity computer output associated with the mea-surements for Group 1 reveals abnormal behavior during the insertion and withdrawal of the last 20% of Group 1.

Such abnormal behavior, which is attributed to the proximity of the Group 1 rods and the. nuclear dete.ctors, indicates an unusually large uncertainty in the measured value.*

Most of the experimental uncertainties listed above are magnified in analyzing the rod drop measurement of shutdown Group A.

The calibration of the meter and lin-earity in the wide range channel over several decades is critical~ In addition, it was found that the time constants of the electronic circuits are significant compared to the speed of the reactivity insertion; both are on the order of 1 to 1.5 seconds?,/ Also, the power distribution change during the scram is large.

We understand that at other plants where this test is used, large correction factors

(>30%) are required to adjust the results to known values.* For these reasons, this test was considered unreliable and th~ results disregarded.

8.

Shutdown Margin Shutdown margin was reevaluated using the results of the qontrol rod worth measurements.

The total available rod worth is assumed to be 8.53%~p as quoted in Table 5.

Since the measured rod worths overall were 10% lower than calculations, the calculated value of stuck rod worth is conservative and was unchanged.

The calculated end-of-life total rod worth was lowered by the 10.3% measurement/calculational bias determined from the BOC testing.

A shutdown margin table based on the revised figures *for beginning-and end-of~life is shown as Table 6.

Reference 4 T

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Table 6 - Palisades Cycle II Shutdown Margin Doppler Defect Moderator Temperature Defect Moderator Void Defect PDIL Rod Insertion Worth Axial Flux Redistribution Total Reactivity Allowances Total Measured Full-Length Rod Worth Stuck Rod Worth Net Available Worth Shutdown Margin Required Margin HZP 0

0 0

1.68 0

1.68 8.53 3.22 5.31 3.63 3.40 BOC BOC (Measured)

EOC Rod Worth= EOC (Calculated) x BOC (Calculated)

- Based on BOC Integral. Rod Worth Curve 8

HFP 1.0

.2 0

.1 0

1.3 8.53 3.22 5,31 4.01 2.00

\\

HZP 0

0 0

1.68**

0 1.68 8.68*

3.52 5.16 3.48 3.40 EOC HFP 1.0

.8

.1

.5 2.5 8.68*

3.52 5.16 2.66 2.00

~,_

~* e Power Testing Following the zero power testing, the reactor was escalated to power.

Testing was conducted at 25%, 80%, and 100% of rated power at e~uilibr~um xenon.

1. Moderator Temperature Coefficient The MTC was measured at the three reference power levels by inserting and withdrawing Group 4 control rods to vary Tav about 5°, while maintaining a constant power level.

The worth of Group ~ control rods was then measured.

Using the temperature defect at constant power le.vel and the control rod worth inserted or withdrawn from the core, the value of the moderator temperature coefficient can be determined.

The measured MTC for each power level is given in Table 7.

Refer to *Figure 9 for a plot of MTC vs PCS boron concentration compared with calculated values.

Percent Power

2.

Power Coefficient 25 80 100 Table 7 PCS Boron (Ppm) 1,032 911.

860' Measured MTC

( t;p /OF)

..:4

+0.020 x 10

-0.16 x 10 -4

-0.25 x 10-4 The power coefficient was measured by varying the turbine load and matching the reactor power with the turbine by adjusting Group 4 control rods while maintaining Tavu constant.

The value of the power coefficient at each power level was then ~etermined from the rod worth inserted or withdrawn and the change in reactor power.

The values are given in Table 8 below.

These values are averages of more than one test at each power level, and the plot of these points in Figure 10 implies more precision than is inherent in the tests.

Uncertainties in rod worth and the magnitude of small power changes make the accuracy no better than !_20%.

Table 8 Percent Power Power Coeff t;p/% Pwr 25

-1.05 x 10-4 80

-.69 x io-4 97 4

-.56 x 10-9

/~. -*

3.

4.

(~--.)

'~'--'/

5.

~~*

By calculating the area under the curve in Figure 10, the power defect from zero to 100% power uas found to be -.88 x lo-2 Ap.

The average power coeffi-cient pf -.88 x 10-

~p/% agrees well with the Exxon prediction of -.88 x lo-4 ~p/% and falls within the acceptance criteria of 1.0 x lo-4 Ap/% + 30%

as given in Reference 2.

Worth of Group 4 Regu1ating Rods The worth of a small portion of Group 4 rods was measured by diluting the group in from some reference position while maintaining constant power level.

The worth of the group is then calculated by using the reciprocal boron worth.

For.the purposes of power escalation testing, the worth of Group 4 was mea-sured between 100 and 131 inches (full out).

The results of these measurements are shown in Figure 11 of this report~ The 80% power and 100% power rod worth data can be represented by the same curve.

The acceptance criteria for the small portion of Group 4 control rods t.hat is calibrated at power are any number that allows reasonable control of the plant.

The values obtained in this test are considered to be reasonable since the portion of Group 4 that was calibrated allows control of the plant during normal load changes.

Excore Channel Decalibration By maintaining constant reactor power (as determined by generator output) and varying Tavg' the excore channel decalibration was found to be.57%/°F.

The end-of-core.testing for Core 1 reported a value for excore decalibration of

.58%/°F.

Previous measurements from Cycle I tests used a decalibration of 0.5%/°F to 0.57%/°F.

Incore Detector Power Distribution Measurements During the refueling outage, 22 incore assemblies were removed and replaced maintaining 41 incore assemblies inserted out of a possible 45.

One hundred forty-three individual rhodium detectors are operable at present of the 164 detectors installed. At low power levels, significant differences were noted between the radial power distribution measured by the incore detectors and the predicted power distribution.

Inspection of the individual detector outputs from SY!IlIIletric assemblies revealed a*11% bias between detectors of different manufacture (Belfab and Reuter-Stokes).

No significant core tilt was measured

-by the symmetric sets of the same manufacture.

Detectors of both types had been used in Cycle I and no discrepancy was noted at that time.

All the Belfab detectors in Cycle II, however, are new and were delivered at the same time and are, therefore, suspect.

The bias is constant and does not seem to be a fun~tion of time or power level, so that a common design or manufacturing techni~ue causing a loss in sensitivity is considered to be the likely cause of the* low signals. Efforts to discover the particular s0urce of the problem have been unsuccessful.

10

6.

//--

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Figure 12 shows the normalized radial power distribution determined from the incore system.

In this figure, the assembly powers measured by the Belfab detectors have been increased by the observed 17% bias factor.

The result of this was to improve the agreement between measurements and calculations sub-stantially.

Since the limiting assemblies in the core (D assemblies not on the periphery) show very good agreement between calculations and measurements, no peaking factor penalty is reQuired because of the incore sensitivity problem.

Figures 13, 14 and 15 show the measured core average axial power distributio~s at various power levels.

The full power margin to thermal limits using the

incore power distribution (conservatively neglecting the Belfab incore correc-tion) including uncertainty factors was 3%.

Plant Performance.

Figures 16 through 21 show various primary and secondary system performance characteristics measured during the power escalation.

Included are:

a.

Reactor thermal power vs time.

b.

Gross electrical output vs _time.

c.

PCS boron vs time.

d.

Feed-water flows vs power level.

e.

Core ~T vs power level.

f.

Steam generator pressure vs power level.

g.

Governor valve position vs power level.

11

-------.. -, __, __ _,, _____ -----=--~-----*-*-----,...---.---*-**-- ----. -

1.

- 2.

3.

4.

REFERENCES Palisades Special Test Procedure T-84, Rev O, "Zero Power Test Program,"

and T-93, Rev 1, "Power Escalation Testing.for Palisades Cores," including test data.

Letter from D. A. Bixel of Consumers Power Company to R. A. Purple, "Palisades Plant - Answers to Reload 2/ECCS Questions," dated March 20, 1976.

Letter from Exxon Nuclear Company to Consumers Power Company, "Palisades Cycle 2 Physics Data," dated April_ 19, 1976_. - Contki.ins Exxon proprietary information~

Letter to N. C. Moseley from A. C. Thies, 110conee. Nuclear Station Unit 3 Start-Up Report," dated March 14, 1975, Docket No 50-287.

12

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-1.1

-1.2

-1.3

-1.4. 1/

A

-1.5

- 1.6 0

FIGURE 9 MODERATOR TEMPERATURE COEFFICIENT vs BORON CONCENTRATION 3

4 *

/

. iF

/

/~

I I

.v

/

I 1

IQ 2

5 6 A A -

/

I

/

, MEASUREMENTS II 1 HZE.:ARQ1

/

2 25% p

/

3 80% p I

4 100% p

/"

CALCULATIONS I'

5 HZP ARO

/

6 HFP BOL 7 HFP EOL 200 400 600 800 1000 1200 SOLUBLE BORON CONCENTRATION (ppmB) 0 A

1400

-~.

._d_.. --

~

0 FIGURE 10 POWER COEFFICIENT VS.

PERCENT POWER I.~,......----.-----.------.---~....-----,...---..,.....,------..-----,------.-----.

- I.3, i------+-----+------+------+-~--+------i------+---+------1---1

-1.2 ""'

~

-I.I i---~-+-~--~+~~~.-+-.~---'1--~-+-~~+-~--+~~-f-,-~~-1--~--1 r.01~----+----+----~r--...~.,--+------+------+------1:-----+-----1----~1

~

-,91------+------+------t------+---....,,,.,.+------i------+------+------t------1

~

~*

.si------+------+------1------+---~+------i-.~~.--------.--.-1-----1------1 d'e

a.

<J 1-

z UJ u

LL.

~

-. 7 t----+----t----t---+--t-----t---t--"'"dr--~

~-r------1

.61------+----+----t----+----+-------i------+-----+-----t-'------1

~

~ -.5 u

c::

UJ 3

,4t-----+---+----+----+-----~-----+----+----+------t------I 0 a..

-,31----+------+----+----+----+-----+----+------+-----+-----I

-,2i-----+-----+------+----+----~----+----+---+-----+-----I o~--__.. ___ __._ ___

_,__~_._ ___..__ __ __. ____

~-----'-----'--___..

0 10

. 20*

30 40 50 60 70 80 90 100 PERCENT OF RATED POWER

-~'" -- - ---*-----*-----* ****- :-- --*-** --------~----------

--.~---**,

L

-*. *--... - - --** \\_

FIGURE 11 GROUP 4 ROD WORTH FOR POWER TESTING 4J'..09 ----..-----~----.-----r-----.----.-,------,

/.

(

\\...__.../ e 25% PWR

/.

1032 ppmB I

I

.081----~-+------+------l--------l-~---+---~--+--------t I

I I

.071------+------+-----+---~-----ll--~-----1-------1>'-------t

.. f

.~;..

~~

I

.061-------+-----+-----+--,m.~c--c~---,-1-----J"'------t---+----+--------t

,01 0

130

. '~

125 120 115 110 ROD POSITION (INCHES) 105 1 00% _& 80% PWR 860 & 911 ppmB 100 95

-\\;

(-~-\\

~~/ -

/'

\\~/

e F

.843

.845*

+.002 CALC MEAS DIFF/

CALC FIGURE 12 - NORMALIZED ASSEMBLY POWERS - 100% POWER CALCULATION 100% POWER, 830 PPM, 700 MWD/MT, ARO MEASUREMENT lOOt POWER, 837 PPM, 750 MWD/MT, GRP 4-8% INSERTED E

E

1. 140 1.140 1. 1 7fr!:*

1.221

+.032

+.071 F

E

.897

1. 136

.900*

1.253'>':

+.003

+.103 F

-~09

.. 884*

. -.028 '

F E

D

.871

1. 141

.939 1.094

.782

.857*

1 ~230

.871

1. 135

.818

-.016

+.078

-.072

+.037

+.046 F

F D

.891

.946 1.409

.900

.:7'59

. 906*'"

.896>':

1. 400

.814

.748

+.017

-.053

-.006

-.096

-.014 E

E F

D D

1. 187

1. 228

.989

1. 183

.602 1. 203""

~. 238

.9oy' 1. 147

.620

+.013

+.008

-.087

-.031

+.030 E

F E

D.

1.239

.987 1.168

.909

1. 292

.896'"

1. 178

.902

+.043

-.092

+.009

-.008 D

D D

1. 312

.961

.537 1.293*

.949

. 533,*:

-.014

-.012

-.007 cr =.050

BELFAB DETECTOR SENSITlifltlES ADJUSTE TO NORMALIZE TO REUTER-STOKES DETECTO D BY 17%

RS

- NO DETECTOR, VALUE DERIVED FROM CALCULATIONS AND ADJACENT DETECTOR READINGS

r I 1.4 1.2

. l 1.0 1

j 1 0::

l ~-8

'0 a..

Cl w

N

_J

<(

r:

0::.6 0 :z

.4

.2 0

9()

~

/

"' ~

/

';~~

~

/

'\\ I\\

/

I\\*'

I

'.V

\\

I

..... ~

\\

~

I

/

\\'

' I

\\

j CORE AXIAL POWER DIST ~.5% RODS 120 l'NCHES

' I\\

v

\\

TOP OF CORE BOTTOM OF CORE

.02

.06

.!O

.14

.18

.22

.26

.30

.34.

,38

,42

~6.. 50 '.54 * !58

.G2

,66 *.* 70

.74

.78.* 82

86

90.. 94

.98 FRACTION OF CORE HEIGHT "Tl c;)

c:

o rn

~----------------

1.4 I

\\

1.2

,1.0

~.8

3:

0

'0.

Cl

w N

<(

' ::£ gs.6

z

.4

.2 0

_(

___ )

/ v I

I I

I/

) 7 TOP OF CORE v

v

~

/

t*'\\

\\*.**)',

\\'

l,....--- ~

CORE AXIAL POWER DIST 75%

--r-...__,,,

"' ~ I\\ [\\

[\\

\\

~ \\

~-

\\

l

~.

RODS: 115 INCHES

\\

i

\\

BOTTOM OF CORE

.02

.06

.10

.14

.1a

.22

.26

.Jo

.34

.Je

.42

.44

.50

,54

.. 58 *.s2

.66

.10,.74

_.78.. a2.. 86

.go

.94

.9a FRACTION O~ CORE HEIGHT c:

0 rn i

) :

1.4 1.2 1.0

.8 a::

w

.0 a..

Cl

'W N

_J

~.6 0 z

.4

.2 0

~---'

I J I

\\

J v I

I/

TOP OF CORE

/ v i/

i.----

/

~

/

v

\\

CORE AXIAL POWER DIST 95% RODS

'* -~

~

I\\

\\

109 INCHES

~

\\

BOTTOH OF CORE '..

.02

.06

.10

.14

,IB

.22

.26

.30.34.38

.42

,4q

.50.54

.58... 62

.66 i10

.14

.78

.82

.86

.90

.94

.98 FRACTION OF CORE HEIGHT

FIGURE 16 - GOVERNOR VALVE POSITION VS POWER VI

4 OPEN TO 2 %

0 60 a..

(

LLI

\\

l J

...J

_/ e

<:(

0 c.!J.

40 20 40 60 80 100 PERCENT POWER

-----*- _"_,.. ::::-*=*-=-*-=*=-* -=---~_:_:_:__ __

__;_____;....;__.:.,.__i;:_....;_...;;__...__ _________________

x

~

0 -'

LL.

0::

LIJ 1-:i*

FIGURE 17 - FEED-WATER FLOW VS REACTOR POWER X= FW FLOW #1 ON POL I

I 0: FW FLOW #2 ON POL 6 20001-------+-----+-----'------+-----+---------1-----1 LIJ LIJ LL.

I

/

'tl 10001-------+----"'--J'----+-----~----+----~------i REACTOR POWER (MWT) 0-----------------_.__ ____,.__ ___ __. ____ _,

0 400 800 1200 1600 2000 2400

1130 1110 1090 1070 10.50 E.

a..

(~"~

~j~ lo"30 es z

0 a::

0 a:l VI

~ 1010 990 970 950 930

\\

j 5/18 I

I I

I FIGURE 18 - 'PCS BORON RUNDOWN DURING POWER ESCALATION

\\

~

.. -'--. *.*.*-~

~**

\\

~

"' ~

TIME (DAYS)

I 5/19

s120 5/21 5/22 5/23 5/24*

L_ --

~----------.--**--------------~--*----_

.. ------~_

.. ------~~-~

... ~--,--

~---------*-

---r-....,--------~ -

L 50 45 40 35 30 1.l....

0 I-25

<l LIJ 0::::

0 u 20 15 ro 5

0 I

Fl GURE 19 -

}

v I

400 800 I

l

-=-* *-*-

-~ *--**. -* - -- -.

CORE LH vs REACTOR POWER j.- --

v I

I I REACTOR POWER (MWT)

I 1200 I ~00 I 2000 2400-

~ -- --------------. ---------------------

~ _,._ -~--- --- ------ --- - - --

r------

i 3

e::

2400 2000 1600

~ 1200 3

0 a..

800 400 0 0 I I I.

/

~

6 12 18

~

/

. z

/

.*~Reactor Power (MWT)

J

~

"- Generator

_/

Output MWE 24 30' 36 42 48 TIME HRS

/...

,w

/

54

.~

G')

c:

o rn N

0 C:.

z

-f

"'ti 0

~

rn

o Vl

-f 3:

rn I

I j

. i i*

f t-.

I

0 0 ao FIGURE 21

""'* STEAM GENERATOR PRESSURE VS REACTOR POWER I

I

. J I/

(,

(

ll 11 0

<O

/,

J ~

//~

/; v v

0 J

!1 I/

I I II I

I Sd 3~nss3~d ~Ol'tf~3N3~ W'tf3*1S l I

//

I I

0 0

C\\I C\\I 0

0 0

<\\I 0

0

~

0 0 <<o 0

0.q-0 0

l\\J*

0 0

0 ao 0

0

<D 0

0 O.q-0 f.D I-

3 0::

IJ.J

3 a

0...

0::

a I-u ct:

IJ.J 0::

..... ----- --*- -*- -- -------*----.. ----.......,.......-*-----------=------.- --------.....,..-----------*----- --:-----* -~-...,-----------~---~-----: ----.

ML18354A360

Text

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