ML19340B529

Cycle 3 Startup Rept.

ML19340B529
Person / Time
Site:	Crystal River
Issue date:	10/31/1980
From:	Richard Bright, Poole D FLORIDA POWER CORP.
To:
Shared Package
ML19340B528	List: ML19340B527 ML19340B529
References
TAC-46962, NUDOCS 8011110233
Download: ML19340B529 (68)

Text

. . . . - - . . . . - . . _ . . . _ .

4 h FLORIDA POWER CORPORATION

-CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 LICENSE N0. DPR-72 i

i 1

4 l

1 CYCLE 3 l

STARTUP REPORT i

! APPROVED BY: Nuclear Plan Manager.

i g .

Date /f)- 2 l- W 4

i Manager i Nuclear Support Services

4/

Date OM p'1,/ Mo

4 OCTOBER 1980 N$$$

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

On February 26, 1980, Crystal River Unit 3 experienced a power supply failure in the ICS (Integrated Control System) which even-tually resulted in ECCS initiation and spillage of approximately 43,000 gallons of primary cooling water in the containment.

Since the ' scheduled refueling outage was only a month away, it was decided to refuel early during this unplanned outage for con-tainment cleanup, at a core average exposure of 166.5 EFPD.

Fifty-two Batch 2 assemblies were discharged from the core; the remaining nine Batch 2 assemblies plus the Batch 3 and 4 assem-blies were shuffled into the inner two-thirds of the core; and fifty-two fresh Batch 5 assemblies were loaded on the core peri-phery. Further into the outage, a broken spring was discovered in one of the Batch 4 assemblies. This assembly and its three symetric counterparts were removed from the core. A small fuel shuffle was performed in which four more fresh Batch 5 assemblies were loaded into the core to replace the four Batch 4 assemblies removed. The final Cycle 3 core loading pattern is shown in Figure 1.1-1.

This report, prepared and submitted in accordance with Technical Specification 6.9.1 describes the precriticality, zero power, and power testing performed and the results obtained.

1.2

SUMMARY

Crystal River Unit 3 was brought to Hot Standby condition. Con-trol Rod Trip Testing and Source Range Testing were conducted.

The test methods used and results achieved are reported in Section 2.0.

Criticality was achieved on August 8,1980 at 2347 and Zero Power Physics Testing was performed. The results of these tests were reported in Section 3.0.

Zero Power Physics Testing was followed by Power Escalation Test-ing. The generator was synchronized on August 10, 1980 at 1530 and Power Testing was performed. The test methods used and re-sults achieved are reported in Section 4.0.

P

Core Loading Diagram Crystal River 3, Cycle 3 A 5 5 5 5 5

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. XX Cycle 2 Location B 5 5 5 5 5 5 3 3 3 Y Batch Number C D7 N3 L1 P8 L15 N13 D9 5 5 5 5 3 4 4 2 4 4 3 C7 03 M2 D5 R8 D11 M14 013 C13 5 5 D 5 5 3 4 4 3 4 3 4 4 3 G4 B12 F6 K5 K1 L8 K15 K11 F10 B4 C12 5 E 5 4 3 4 2 3 4 3 _

4 3 2 3 C12 Bil E9 ES D6 B10- D10 Ell E7 B5 C4 F 5 5 5 5 4 4 3 3 3 4 3 3 -3 4 4 C C6 A10 E4 A9 F4 B6 D8 F14 F12 A7 E12 A6 C10 5 5 3 4 3 4 3 4 3 4 3 4 3 4 3 5

C3 H14 HIS H10 F2 H4 H8 H12 Ll4 H6 H1 H2 K13 5

H 4 4 2 4 4 2 3 3 2 3 3 3 3 K 5 6 M0 M4 R9 M L2 N8 P10 L12 R7 M12 R6 UO 5

3 4 3 4 3 4 3 4 3 4 3 4 3 012 P11 M9 M5 N6 P6 N10 M11 M7 PS 04 L 5 5 5 5 4 4 3 3 3 4 3 3 3 4 4 M

K4 P12 L6 C5 Cl F8 C15 C11 L10 P4 K12 S 5 3 4 2 3 4 3 4 3 2 4 3 N 5 5 K3 C3 E2 N5 A8 N11 E14 Cl3 09 5 5 3 4 4 3 4 3 4 4 3 N7 D3 F1 B8 F15 D13 N9 0 5 5 5 5 3 4 4 2 4 4 3 5 5 L7 07 M 5 P 5. 5 5 3 3 3 F 5 5 5 5 5 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 I

Figure 1.1-1 .

i i

2.0 PRECRITICALITY TESTING 2.1 REACTOR'C00LANT FLOW AND FLOW C0ASTDOWN TEST The reactor coolant pump flow and flow coastdown test was not performed for this cycle in order to avoid aggravation of a minor problem 'with a reactor coolant pump seal. The test is not required at this time since the pump power monitors are not installed and the power upgrade has not been initiated.

2.2 CONTROL R00 DRIVE DROP TIME TEST 2.2.1 Purpose

, The purpose of the Control Rod Drive Drop Time test was to verify the functional trip capability of the Control Rod Drive System.

This was done by verifying the following acceptance criterion:

The total elapsed drop time from the initiation of the trip

signal until the control rod assembly is three-fourths in-serted, must be less than or equal to 1.66 seconds with Tavg

> 525*F and all reactor coolant pumps operating.

2.2.2 Test Method The Control Rod Drive Drop Time Test was performed using strip chart recorders to time the rod drops. Each control rod group was pulled to 100". withdrawn and then dropped into the core using the manual trip pushbutton. A zero time signal was furnished to the test recorders for each control rod assembly from a contact on the manual trip switch. A second isignal to indicate three-fourths insertion was furnished to the recorders by a reed switch located on the position indicator tube of each control rod drive. The test was conducted at the following conditions:

Flow Temperature Mode Four Pumps > 525 F Hot Standby 2.2.3 Evaluation of Test Results The results of the rod drop times are presented in Tables 2.2-1 and 2.2-2. The rod with the shortest drop tim' was CR4-9 with a rod drop time of 1.218 seconds. The rod with the longest drop time was CR1-8 with a rod drop time of 1.287 seconds. All of the control rods meet the acceptance criterion.

i 2.3 CHEMICAL AND RADI0 CHEMICAL TESTS Chemical 'and Radiochemical testing was not performed during this startup. These tests were conducted at initial startup and no i---+--++- -,si t- g- - m 19 ff T-

plant modifications have been made which would invalidate the results of those tests.

2.4 PRESSURIZER EFFECTIVENESS l No pressurizer effectiveness testing was done this startup since no modifications had been made to the pressurizer and results of the testing done at initial startup are still valid.

2.5 CALIBRATION AND NEUTRON RESPONSE OF SOURCE RANGE MONITORING

^

Precritical testing was completed following response checkout of the neutron source range detectors. The source range instrumen-tation was verified to provide a count rate of more than 2 counts /sec. A statistical test was performed to determine that the detector response data followed a normal distribution.

2.6 IN-SERVICE LOOSE PARTS AND VIBRATION MONITORING SYSTEM I No loose parts and vibration monitoring system testing was done during this startup since no modification had been made or cir-cumstances encountered which would invalidate the test results from the mid-Cycle 1 startup testing.

4 i

e

CONTROL R00 DRIVE DROP TIME l TEST RESULTS DROP TIME

CONTROL ROD CORE POSITION (Sec.)

4 CR1-1 C-11 1.269 CR1-2 E-13 1.271 i CR1-3 M-13 1.267 CR1-4 0-11 1.263

. CR1-5 0-5 1.263 CR1-6 M-3 1.262 i CR1-7 E-3 1.278 l CR1-8 C-5 1.287 CR2-1 E-9 1.279

? CR2-2 G-11 1.278 CR2-3 K-11 1.255 CR2-4 M-9 1.277 CR2-5 M-7 1.255

!' CR2-6 K-5 1.269 CR2-7 G-5 1.255

CR2-8 E-7 1.268 CR3-1 0-8 1.268 CR3-2 G-9 1.258

, CR3-3 E-11 1.260 CR3-4 H-12 1.250 CR3-5 K-9 1.285 CR3-6 M-11 1.263

CR3-7 N-8 1.267 CR3-8 K-7 1.263 i

CR3-9 M-5 1.268

, CR3 H-4 1.268 CR3-11 G-7 1.263 CR3-12 E-5 .1.280 CR4-1 H-8 1.244 i CR4-2 B-10 1.220 CR4-3 F-14 1.249 l' CR4-4 L-14 1.235 CR4-5. P-10 1.239 l

. CR4-6 P-6 1.239-CR4-7 L-2 1.234 CR4 F-2 1.235 CR4-9 B-6 1.218 Table 2.2-1

i

r CONTROL R00 DRIVE DROP TIME TEST RESULTS DROP TIME CONTROL ' 00 CORE POSITION (Sec.)

CRS-1 C-9 1.258 CRS-2 G-13 1.267 CRS-3 K-13 1.268 CR5-4 0-9 1.250 CR5-5 0-7 1.260 CRS-6 K-3 1.265 CRS-7 G-3 1.254 CRS-8 C-7 1.263 CR6-1 F-8 1.272 CR6-2 D-12 1.254 CR6-3 H-10 1.272 CR6-4 N-12 1.250 CR6-5 L-8 1.273 CR6-6 N-4 1.250 CR6-7 H-6 1.263 CR6-8 D-4 1.258 CR7-1 B-8 1.255 CR7-2 F-10 1.263 CR7-3 H-14 1.263 CR7-4 L-10 1.258 CR7-5 P-8 1.263 CR7-6 L-61 1.262 CR7-7 H-2 1.255 CR7-8 F-6 1.263 Table 2.2-1 (Cont'd) i

GROUP AVERAGE CONTROL R0D DRIVE DROP TIME AVERAGE DROP GROUP TIME (Sec.)

1 1.270 2 1.267 3 1.266 4 1.235 5 1.261 6 1.262 7 1.260 1-7 1.260

)

Table 2.2-2

3.0 ZER0 POWER PHYSICS IESTS The Zero Power Physics Tests were performed to verify the nuclear design parameters used in the safety analysis, the Technical Specification limits, and the operational parameters. Acceptance criteria for the tests are given in Table 3.0-1. A sumary of measured and predicted results obtained during zero power physics testing is given in Table 3.0-2.

3.1 INITIAL CRITICALITY Initial criticality was achieved on August 8,1980 at 2357 hours0.0273 days <br />0.655 hours <br />0.0039 weeks <br />8.968385e-4 months <br /> at reactor coolant conditions of 532*F, 2155 psig, and 1426 ppmB. Control rod groups 1 through 4 had previously been withdrawn as part of the heatup procedure with reactor coolant boron concentration at 1634 ppmB. The approach to critical began by withdrawing control rod group 8 to 37.5% withdrawn, groups 5 and 6 to 100% and group 7 to 85% withdrawn. The reactor coolant system was then deborated until criticality was achieved.

Throughout the approach to criticality, inverse multiplication curves versus time, reactor coolant system boron concentration, and quantity of demineralized water added were maintained for the source range detectors by two independent persons. At the end of each control rod group withdrawal and every 30 minutes during the deboration cycle, the count rate was taken from each source range detector by way of the scaler-counters. The ratio of the initial average count rate to the count rate at the end of each reactiv-ity addition was plotted and the criticality point determined by extrapolation.

3.2 NUCLEAR INSTRUMENTATION OVERLAP Technical Specifications state that prior to operation in the intermediate nuclear instrumentation range, at least a one decade overlap between the source range and intermediate range must be observed.5 This means that before the source range count rate equals 10 cps the intermediate range must be on the scale. If the one decade is not observed, the approach to the intermediate range cannot be continued until the situation has been corrected.

To. satisfy the above overlap requirements after initial critical-ity was reached, core power was slowly increased until the inter-mediate range channels came on scale. Detector signal response was thus recorded for both the intermediate and source range channel. This was repeated for another decade to assist in determining the number of decades of overlap.

f The results of the Nuclear Instrumentation overlap at 532 F and 2155 psig have been tabulated in Table 3.2-1. Examination of Table 3.2-1 shows that the average overlap between the source and l

l 1 1

L

intermediate range is 2.11 decades which is well above the mini-

'num of one -decade overlap required by Technical Specifications.

3.3 SENSIBLE. HEAT DETERMINATION Determination of the intermediate range current level at which the production of sensible nuclear heat occurs is important to the Zero Power Physics Test program in that it establishes the upper level limit. Thus, by restricting reactor power operation to a level reduced by a factor of three below the sensible heat level, the effects of temperature feedback are eliminated in the measurement of physics parameters.

The- test method was to increase the intermediate . range current level in one third decade increments until the detection of sen-sible heat. Since turbine bypass valves and pressurizer levels were maintained constant, heat production in the core was observed by change in the core outlet temperature and makeup tank level.

1 The poin at which there was a definite heatup rate was 6.0X10gamps. Therefore, during the test 6.0 X 10-8 amps on channel NI-3 as defined as the " sensible heat" point. From this',

the upper zero power physics test current limit was established at 1.8 X 10-8 amps.

After setting the upper zero power physics test current limit, temperature feedback measurements were performed to verify that the above limit was adequate. This was, done by adding +20 pcm at a neutron flux signal of 1.8 X 10-o and verifying that the reactivity remained constant. The results indicated that no temperature feedback effects were present at the upper zero power physics test current limit.

j An additional part of the measurement was to ensure that all power range channels were indicating between 0.8 to 2.0% full power at the point of sensible heat. This was done to ensure that the overpower trip protection was adequate. At the point of sensible heat, the indication on NI-5, NI-6, NI-7 and NI-8 was 1.5% full power. Thus, all power range channels were within their recommended tolerances. Since all channels were set near their maximum sensitivity, no calibration was performed.

3.4 REACTIVITY CALCULATIONS Reactimeter is the name given to the Babcock and Wilcox reactiv-

! ity computer which solves the one-dimensional, inverse kinetics equation with six delayed neutron groups for core net reactivity based upon periodic samples of neutron flux. In addition to reactivity and neutron flux, the reactimeter can also record 23 other analog and digital signals from the plant.

1 After initial . criticality and prior to the first physics measure-ments, an on-line functional check of the reactimeter was per-formed to verify its readiness for use the test program.

After steady state conditions with a constam neutron flux were established, a small amount of negative reactivity was irserted in the core by inserting control rod group 7. Stop watches were used to measure the doubling time of the neutron flux and the reactivity inserted was determined from period-reactivity curves. The measurement was repeated for several values of reactivlty inserted by rod group 7, from 0.026 to 0.100 ak/k.

The reactivities determined from doubling the measurents were then compared with the reactivities calculated by 1ne reacti-meter.

The results of the reactimeter verification measurements are sum-marized in Table 3.4-1. With the exception of Case 4, the reactivity calculated by the reactimeter was within the accep-tance criteria limit of 15% of the reactivity determined from doubling times.

3.4,1 "All Rods Out" Critical Boron Concentration The "all rods out" critical boron concentration was measured at one isothermal temperature test plateau of 532 F. The measure-ments actually were made with rod group 7 partially inserted, but the measured boron concentration was adjusted to the all rods out condition using the results of rod worth measurement to determine the reactivity worth, in terms of ppm boron, of the inserted con-trol rods.

The results are tabulated in Table 3.4-2. These results show that the measured critical boron concentration was within the acceptance criterion of 14301 100 ppmB. The results were also within the review criterion of 1430 1 50 ppmB.

3.5 CONTROL R00 GROUP WORTHS The layout of the core according to the standard alpha-numeric mesh showing the initial location of the control rod groups and the location of- the 52 incore dett.ctor strings is given in Figure 3.5-1. The number of rod , in each group and the reactiv-ity control function of each group is listed below.

Rod Group No. No. of Rods Control Function 1 8 Safety 2 8 Safety 3 12 Safety 4 9 Safety 5 8 Power Doppler 6 8 Power Doppler 7 8 Transient 8 8 Axial Power Shaping 69 1

w.-y ,, - .

l

, Predictions of control rod group worths for groups 1-4, 5, 6 and 7 at Hot Zero Power and Hot Full Power conditions were made in the Physics Test Manual. Measurements of control rod group worths for groups 5, 6 and 7 were made during Zero Power Physics

. Tests using the boron swap method. This method consisted of set-ting up a deboration rate and compensating for the change in reactivity by small step changes in rod group positions. The calculation of reactivity on a continuous basis is made by tha reactimeter. The output of reactivity in terms of percent milli p (PCM) is recorded on a strip chart and also on a magnetic tape by the reactimeter.-

The results of both the predicted rod group worths and the measured group worths are tabulated in Table 3.5-1 for a modera-tor temperature and pressure of 532 F and 2155 psig, respec-tively. Comparison between measured and predicted rod worth shows groups 5-7 agreed within 0.6 percent. Predicted and esti-mated rod worths at 579 F are also tabulated.

All acceptance criteria were met.

Integral measured rod worth curves for control rod groups 5, 6 d

c.d 7 at 532*F and 2155 psig are plotted in Figures 3.5-2 to 3.5-4.

3.6 SOLUBLE POIS0N WORTHS Soluble poison in the form of dissolved boric acid is added to the moderator to provide additional reactivity control beyond that available from the control rods. The primary function of l the soluble poison control system is to control the excess reactivity of the fuel throughout each core life cycle.

Measurement of the soluble poison differential worth has been completed at 532 F and 2155 psig. The measured value was deter-mined by summing the incremental reactivity values measured dur-ing the rod worth measurements over a known boron concentration range from 1426 to 1150 ppmB.

The result of the differential soluble poison worth measurement is tabulated in Table 3.6-1.

In summary, the measured differential boron worth at zero power was within 9.9 percent of the predicted worth which is within the acceptance criteria of + 15 percent.

3.7 EJECTED CONTROL R0D WORTH I Pseudo ejected control rod activity worth was measured at hot zero power conditions of 532*F, and 2155 psig for control rod 6-4. The purpose of this measurement was to verify the safety analysis calculations relating to the assumed accidentai ejection of the most reactive control rod which is fully inserted in , the core during normal power operation. The acceptance i

I y - w -q+ -

criteria for the ejected control rod test are that the average measured ejected rod worth be within 20% of tha predicted value,

- that the worst case measured ejected rod worth be less than 1.0% ak/k, and that each of the 4 symmetric ejected rod worths be within 0.05% ak/k of the average.

The ejected rod worth measurements were done with control rod groups 6 and 7 fully inserted and group 8 at 37.5% withdrawn.

Group 5 was 0% withdrawn for the boron swap measurement and 2S%

withdrawn for the rod swap measurements. The ejected worth of control rod 6-4 was measured by the boron swap method. Control rod 6-4 and the rods in the quarter core symmetric locatiors were also measured by the rod swap method. The rod swap test method was to iaitially position group 5 at 28% withdrawn with the

" ejected" control rod 0% withdrawn. The change in group 5 pesi-tion was then converted to reactivity by using a previously gen-erated group 5 rod worth curve to determine the measured ejected rod worth.

Since the measured rod position did not p"ecisely equal the mini-mum rod index, the measured ejected rod i orth was adjusted to the Technical Specifications position by using the equation below:

Where: p(ER) = F X p(R) = p(M) EC. (3.7.1) p(ER) = The measured ejected rod worth at i.ha maximum allowable inserted group worth.

F = Empirically derived constant of 0.45 which re-lated ejected rod worth and total inserted group worth, p(R) = The worth of roo:. at the minimum rod index minus the worth of rods Ot the final rod configuration, p(M) = The measured worth of the ejected rod which cor-responds to the worth changes in group 5.

The measured worst case ejected rod worth was within the accep-tance criteria of less than 1.0% ak/k and within 20% of the Physics _ Test Manual predicted value. One of the relative ejected rod worth values (rod 6-8) was 0.06% ak/k greater than the average, which was not within the acceptance criterion of t 0.05% ak/k of the average worth. As required in the Physics Testing procedures, each rod in all rod groups was exercised and a reactivity response verified to ensure that all rods were coupled. Coupling was confirmed and the Plant Review Committee reviewed the results prior to power escalation. The results of

, these measurements are given in Table 3.7.1.

3.8 TEMPERATURE C0EFFICIENTS OF REACTIVITY The temperature coefficient of reactivity is defined as the frac-l tional change in the excess reactivity of the core per unit

change in core temperature. The temperature coefficient is nor-mally divided into two components as shown in Equation (3.8-1).

aT = aM + oD EQ. (3.8-1)

Where: aT = Temperature Coefficient of Reactivity aM = Moderator Coefficient of Reactivity a0 = Doppler Coefficient of Reactivity The technique used to measure the 532*F and 2155 psig isotherr.al temperature coefficient at zero power was to first establish steady state condition > by maintainin3 reactor flux, reactor coolant pressure, turbine header pressure and core average temo-erature constant, with the reactor critical between 7 X 10-10 amps and the upper zero power physics test current limit set in the sensible heat determination experiment. Equilibrium boron concentration was established in the reactor coolant system, make-up tank and pressurizer to eliminate reactivity effects due to boron changes during the subsequent temperature swings. The reactimeter and the brush recorders were connected to monitor selected core parameters with the reactivity value calculated by the reactimeter and the core average temperature displayed on a two channel recorder.

Once steady state conditions were established, a negative heatup rate was started by opening the turbine bypass valves. As the reactivity changed about i 40 PCM, contro, rods were noved in step changes to keep the reactivity values on scale and also to keep core power in the range necessary to produce an adequate intermediate range signal. After the core average temperature decreased by about 5 F, coolant temperature and reactivity were stablized. This process was then reversed except that the core average temperature was increased by about 10 F. After stabili-zing coolant temperature and reactivity, the core average temper-ature was then returned to its original value. The measurenent of the temperature coefficient from the data obtained was then performed by dividing the change in reactivity by the correspon-ding changes in core temperature over a specific time period.

Isothermal temperature coefficient measurements were conducted at two different reactor coolant boron concentrations during the Zero Power Physics test program. The results of the measurements are summarized in Table 3.8-1 along with the predicted values which are included for comparison. In all cases, the measured results compared favorably with the predicted values. All measured temperature coefficients of reactivity were within the acceptance criteria of i 0.40 X 10-4 ak/k/ F of the predicted value.

The moderator coefficient cannot be directly measured in an operating reactor because a change in moderator temperature causes a similar change in the fuel temperature. However, since the moderator coefficient has safety implications, it is an important reactivity coefficient. As speci fied in Technical Specifications, the moderator temperature coefficient shall not be positive at power levels above 95 percent full power and shall be less than +0.9 X 10-4 Ak/k/*F at all other power levels. To obtain the moderator coefficient from the meaqured ter.ip?rature coefficient, a Doppler correction of -0.2 X 10-4 ak/k/*F nust be subtracted. Determination of the moderator coefficient has been

completed and shows that this coefficient is well within the limits stated above.

In conclusion, all temperature coefficients of reactivity that were measured at the 532*F and 2155 ppig plateau were within the acceptance criteria of + 0.40 X 10" ak/k/ F of the predicted value. In addition, calculation of the moderator coefficient indicates that it is well within the requirements of the Tech-nical Specification 3.1.1.3.

3.9 BIOLOGICAL SHIELD SURVEY A Biological Shield Survey was not done at Hot Zero Power since

. no plant modifications were made which would invalidate the Bio-logical Shield Surveys made during the Initial Startup Testing I Program.

l 3.10 EFFLUENT AND EFFLUENT MONITORING i

l No effluent or effluent monitoring testing was performed as these systems have been performing normally since initial startup and

no further testing was required.

i 3.11 CHEMICAL AND RADI0 CHEMICAL TESTS l

Chemical and radiochemical testing was not performed during this startup. These tests were conducted at initial startup and no plant modifications have been made which would invalidate the re-sults of those tests.

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= mags" 4- 4--4-4 --* b 4 _a ._ .aa_ . _ _ -.u_ a.A_ +s u +, _-*-_a.Me++- +* J--4 4 4- h_ pO4.-J,M -h+ - ..6-_--- 4--e 2- + +-u a-- W- - > + n ACCEPTANCE CRITERIA DEVIATION LIMITS BETWEEN MEASURED AND PREDICTED-VALUES ,

Allowable Deviation Between Core Physics Parameters Predicted and Measured Values A. Control Rod Worths Group Worths (5 - 7) i 15%

Total Worth (5 - 7) i 10%

i Ejected Rod Worth 1 20% .

l B. Temperature Coefficients + 0.4 x 10-4 Ak/k/*F t 57 C. Differential Boron Worth --+ 15%

. 55

• D. All Rods Out Boron Concentration + 100 ppmB ,

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i 4

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. _. . - _ - = _ _ _ _ _ _ . -. _. _ __

I COMPARISON CF KASURED AND PREDICTED RESULTS OBTAIED DURING ZERO POWER PHYSICS TESTS Physic Paraneter Units Measured Predicted Acceptance Requirements Cmparison

1. All Rods Out Critical Baron ppnB 1432 1430 Measured value nust be within OK(2) 100 ppnB of predicted value

2. Sensible Heat amps 6.00 X 10-8 None

3. NI Overlap decades 2.11 .Greata.r than 1.0 decade OK(>1.0)

4. Control Rod Group Worth  % ak/k Group 7 -0.78 -0.78 Percent deviation of group worth OK(0.0)

Group 6 -0.98 -0.98 should be less than 15% OK (1.0)

U Group 5 -1.36 -1.37 OK(0.7) 5 Percent deviation of total worth

[ TOTAL -3.12 -3.14 should be less than 10% OK(0.6)

S. EjectedRodWorth(6-4)  % Ak/k Measured 0.47 0.53 Percent deviation of ejected rod OK worth nust be within 20% of the predicted value.

Measured Worst Case 0.50 Value must be less than 1.0% Ak/k OK(<l.0%)

6. Temperature Coefficient 10-4 Ak/k F At'1428 ppnB +0.410 +0.210 UK Measuredgaluenustbewithin 0.4 X 10- ak/k*F of the pre-At 1145 ppnB -0.347 -0.425 dicted value. OK 1

COMPARISON OF WASURED AND PREDICTED RESILTS OBTAIED DURING ZERO F0WER PHYSICS TESTS Physic Paraneter Units Measured Predictg Acceptance Requirenents Cmparison

7. Moderator Coefficient 10-4 Ak/k F At 1428 ppnB 0.610 (;.406 Maximum positive noderator co- OK efficient nust be less than At 1145 ppnB -0.147 -0.227 0.9 X 10x-4 Akik/ F OK

8. Differential Boron Worth  % ok/k/ppnB At 1288 ppnB -0.0110 -0.00991 Percent deviation nust be less' OK(-9.9) y than 15%

X

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y NOTE: Percent Deviation is defined by the following fonnula:

% Deviation = Predicted Value - Measured Value X 100 S Measured Value s

9

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

SUPT %RY OF NUCLEAR INSTRUENTATION OVEMP KASUREMENTS DLRING ZERO POWEk PHYSICS TEST Socrce Range Indication Intermediate Range Indication Average SR Average IR .

Data Indication Indication Overlap Set NI-1 (CPS) NI-2 (CPS) NI-3 (AW S) NI-4 (AW S) (CPS) ( M S) (Decades) 1 01 1.0 X 104 1.0 X 104 2.0 X 10-11 2.0 X 10-11 1.0 X 104 2.0 X 10-11 2.30 02 2.0 X 104 2.0 X 104 3.0 X 10-11 3.0 X 10-11 2.0 X 104 3.0 X 10-11 2.18

03 1.0 'X 105 8.0 X 104 6.0 X 10-10 7.0 X 10-10 9.0 X 104 6.5 X 10-10 1.86 Average = 2.11

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NOTE (1): Overlap is obtained between Source and Intennediate

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l Range by using the average indications in the equation below:

Overlap = (6 - Log SR) + (Log IR + 11) h k

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- . .. m . . . m- .. _ _ . . _ _ . _ . _ . . . . . _ _ . _ . ~ . . _ . _ __ . _ _ - . , . ._ -- _ _ _ , . . ___.--. _ . . . . ..

1 j COMPARIS0N OF REACTIMETER AND DOUBLING TIME (DT) REACTIVITY E ASUREMENTS r

DT Converted Reactimeter Absolute Case DT Reactivity Reactivity Error No. (Sec) (% ak/k) (% Ak/k) (%) '

1 165.5 +0.0270 +0.026 3.8 .

2 114.0 -0.0585 -0.058 0.8 3 126.0 +0.034G +0.035 2.9

! 4 82.0 -3.0975 -0.089 9.6 -

5 191.0 -0.0310- -0.030 3.3 C

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NOTE: The Absolute Error Between DT Converted Reactivity and Reactimeter Reactivity is i Defined By the Equation Below:

E (%) = 100 (pR - pDT) / (pDT) 4 i

4

. - _ _ - - _ . - - - - _ - - _ - _ _ _ - - -- __.____--- - - ~ ~ -. e an ,-m- --

ALL RODS OUT CRITICAL BORON CONCENTRATION (PPM BORON)

Moderator Predicted Measured Temperature Results Results 532 F 1430 1432 (1)

1. NOTE (1): This number is based on the measured boron concentration with a E 4 prmB correction to account for the fact the rods were not quite at 100% withdrawn,

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__. . _ . . , _ . - . __ __ _ _ . . __ . _ _ _ . _ _-_.___ - . . _ . .. __ _ m .

COMPARIS0N OF PREDICTED AND MEASURED CONTROL R0D GROUP REACTIVITY WORTH AT B0C-3 A. Moderator Temperature at 532*F, APSR's at 37.5% WitLdrawn Predicted Measured Percent Rod Number Worth Worth Deviation Group Of Reds (% Ak/k) (% Ak/k) (%)

5 3 -1.37 -1.36 0.7 6 8 -0.99 -0.98 1.0 7 8 -[_;8 -0.78 0.0 Total -3.14 -3.12 0.6

.f B. Moderator Temperature at 579 F, APSR's at 37.5% Withdrawn

[ Predicted Estimated Rod Number Worth Worth Group Of Rods (% Ak/k) (% Ak/k) 5 8 -1.41 -1.40 6 8 -1.09 -1.08 7 8 -0.83 -0.83

-3.33 -3.31 NOTE: Estimated worth is determined by applying the percent deviations between predicted to measured results at 532 F to the predicted results at 579*F. . ,

b; 4

w ,- __ _

DIFFERENTIAL BORON REACTIVITY WORTH MEASUREMENTS OURING ZERO POWER PHYSICS TEST Rod Position, % wd Measured Average Delta Boron Differential Worth % Ak/k/ppmB Boron Conc Boron Conc Boron Conc Worth 1 2 3 4 5 6 7 8 (ppm) (ppm) (ppm) (A Ak/k) Measured Predicted 100 100 100 100 100 100 90 37.5 1426 1288 276 3.048 0.0111 0.00991 100 100 100 100 10 0 0 37.5 1150 x

e e

?

?

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

DETERMINATION OF THE RELATIVE -SYMETRIC EJECTED R00 WORTH AT ZERO POWER, BOL, ANO 532*F CONDITIONS

. Rod Position, % wd RCS: Baron Ejected P,ad Change In Ejected' Rod Worth, % A /k l Concentration Position Reactivity i l' .2 3 ~4 5 6 7 8 (ppm) -(% A/k) Measured Worst Case e

100 100 100 100 14.0 0 0 36.0 1176 100 0.440 0.470 0.54 100 100 100 100 28.3 0 0 36.0 0 100 100 100 100 6.2 0 0 36.0 100 0.489 i

100 100 100 100 28.3 0 0 36.0 0

[ 100 100 100 100 16.0 0 0 -36.0 100 0.407 l l

(

r' 100 100 100 100 28.5 0 0 36.0 0 100 10C 100 100 '16.1 ~0 0 36.0 100 '0.410 i

100 100 100 100 28.5 0 0 36.0 0  !

i

NOTL: 'he seasured ~e,iected rod worth has been normalized to the rod configuration oF Bank 5-7 0% withdrawn.

NOTE: The worst case ejected rod worth is obtained by application of an uncertainty

factor of approximately 1.10.

l

i

SUMMARY

OF MEASURED, CALCULATED, AND PREDICTED C0EFFICIENTS OF REACTIVITY AT ZERO POWER Rod Position, % wd - Average RCS Boron Reactivity Coefficients, 1.0E-04 % k/k/ F Temperature Concentration Temperature Coefficient Moderator Coefficient 1f2 3 4 5 6 7 8 (Deg. F) (ppmB) Measured P redict.ed Calculated Predicted 100 100 100 100 100 100 91 100 532 1428 +0.410 +0.210 +0.610 +0.406 100 100 100 100 100 10 0 38 532 1145 -0.347 -0.425 -0.147 -0.227 C

55 m

t'

?

I NOTL: "he calculated-noderator coefficient has been determ' ned from the neasured temperature oefficient by subtracting off the isothermal Doppler coefficient of -0.20 10-4 k/ F.

k

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4.0 , POWER TESTING 4.1 REACTIVITY C0EFFICIENTS AT POWER TEST l 4.1.1 Purpose The purpose of this test was to determine reactivity coefficients during power operation at 100 percent full power, and to verify that they were conservative with respect to the FSAR. The fol-lowing coefficients were either measured or calculated from the data obtained.

(a) Temperature coefficient of reactivity, defined as the frac-tional change in the reactivity of the core per unit change in fuel and moderator temperature, (b) Moderator coefficient of reactivity, defined as the frac-tional change in the reactivity of the core per unit change in nederator temperature.

l (c) Power Doppler coefficient of reactivity, defined as the j fractional change in the reactivity of the core per unit change in power.

4 Acceptance criteria are specified for the Reactivity Coefficients at Power Test and are listed below:

(1) The moderator coefficient of reactivity measured at 100%

full power shall be negative.

(2) The power Dopple coefficient of reactivity shall be nure negative than -0.55 X 10-4 Ak/k/%FP.

4.1.2 Test Method

< Reactivity coefficient measurements were made during the power escalation test program at 100% full power.

Differential rod worth measurements were performed during the reactivity coefficient measurement in order to generate rod worth data for the specific test conditions. For temperature coeffic-ients, average reactor coolant temperature was increased and decreased about 5*F and data recorded. For power Doppler coefficients, power was increased and decreased about 5 percent full power and data recorded. From the measured temperature and power Doppler coefficients, the moderator and Doppler coeffi-cients were calculated.

l

'l l

4.1.3 Temperature and Moderator-Coefficients The temperature coefficient of reactivity is defined as the fractional change in the reactivity of the core per unit change in fuel and moderator temperature. The temperature coefficient is normally divided into two components as shown in equation 4.1-1.

aT = aM + aD EQ. (4.1-1)

Where: aT = Temperature Coefficient of Reactivity aM = Moderator Coefficient of Reactivity a0 = Doppler Coefficient of Reactivity The moderator coefficient cannot be directly measured in an operating reactor because a change in the moderator temperature also causes a similar change in the fuel temperature. Therefore, the moderator coefficient must be calculated using equation 4.1-1 after the temperature and Doppler ccefficients have been deter-mined. Technical Specifications, Section 3.1.1.' requires that the moderator coefficient be less than +0.90 X 10-4 ak/k/*F at power levels below 95% full power and non-positive at or above 95% full power.

Temperature and moderat r coefficients were theoretically pre-dicted as shown in Figures 4.1-1 and 4.1-2 using the distributed moderator and fuel temperatures instead of the isothermal values which were used for zero power physics predictions. For these predictions, the normal mode of operation with critical boron and rod conditions was assumed which set the average core moderator temperature equal to 579 F.

The measurement method used at power is to change the reactor coolant temperature setpoint at the reactor control station with the integrated control. system in automatic effecting an approxi-mate 5 F change in the reactor coolant temperature. The reactiv-ity change caused by the temperature change of the core was measured by recording the change in the position of the control-ling control rod group and converting this change to reactivity using differential rod worth values measured during the test.

Prior to running the test, steady state equilibrium xenon condi-tions including a stable boron concentration and no significant control rod motion during the last 30 minutes prior to taking data were required as prerequisite system conditions.

The power Doppler coefficient relatt ^ ne change in core reactiv-ity to a corresponding change in bm . Theratical predictions of the power Doppler coeffic b.E w. . n:ade using the PDQ code' with thermal feedback. The s w a ocwer Doppler coefficients using this code are p' resented n Figurti 1.1-3.

i

The measurement method used was- to change the reactor power level 5 percent full power. This change in power level was initiated by manually ' decreasing the reactor power at the reactor master control station. After obtaining approximately ten minutes of steady - stata data at the reduced power level, reactor power was returned to the initial power.

The calculation of the power Doppler coefficient uses the mea-sured change in the controlling rod group position converted to an equivalent reactivity value and the measured change in reactor power determined by using the normalized core aT which is the primary side heat balance.

I 4.1.4 Differential Rod Worth at___ Power The method by which the differential rod worth was determined at power is the fast insertion / withdrawal method. In this measure-ment, the controlling rod group is inserted for approximately six seconds, followed immediately by a withdrawal for approximately six seconds. Since the total elapse ' dme is on the order of the primary loop recirculation time, the moderator temperature effects are eliminated and the reactivity versus time is essen-tially a combination of the effects due to the control rod mtion and the fuel power variation.

Determination of the differential rod worth was then found by using the measured reactivity and rod positions, compensating the data by a predicted fuel power correction factor.

The fuel power. corret. tion factor accounts for the time delay in-volved in fuel temperature change during the neasurement.

4.1.5 Evaluation of Test Results Th results of the measured temperature and calculated moderator coefficients at power are plotted in Figures 4.1-1 and 4.1-2 which also shows the predicted temperature and- moderator coeffi-cient results.

Examination of the calculated moderator coefficient-as plotted in Figure .4.1-2 indicates that the limit of a non-positive value is not to be exceed- d at 95 percent full - power. Le soluble poison concentration for equilibrium-xenon all-rods-out condition is about 992 ppmB. Thus during power operation at or above 95 per-cent full power the moderator coefficient will be negative even if all control rods are withdrawn from the core.

The results of the measured and predicted power Doppler coeffi-cient of reactivity are plotted in Figure 4.1-3. A list of the measured reactivity coefficients along with the conditions under which they were measured is given in Table 4.1-1.

. - v- t

The acceptance criterion for the measured power Doppler coeffic-ien 10 g ak/k/%FP.

is that the coefficient Figure 4.1-3must bethat shows morethenegative measuredthan -0.55 X coefficient is below this value and that the acceptance criterion is ade-quately met.

4.2 UNIT HEAT BALANCE Heat balance calculations were performed using both the Bailey 855 unit computer, and the IBM 5100 computer. The accuracy of these two methods was verified during initial startup by perform-ing hand calculations. No modifications were made during this shutdown which would require reverification of the heat balance calculation.

4.3 CORE POWER DISTRIBUTION TEST Core power distributions were taken as required during the power escalation test program as a part of the following individual tests.

Core Power Distribution Test Power Imbalance Detector Correlation Test Incore Detector Test Continuous monitoring of the core power density at 364 core loca-tions is accomplished by the incore monitoring system. This system is comprised of 52 detector strings each having 7 individ-ual neutron detectors equally spaced at seven axial elevations in the center of 52 fuel assemblies. This system is capable of pro-ducing detailed core power distributions for either eighth core or quarter core symetry conditions. A detailed description of the incore nonitoring system and the calibration test results are presented in Section 4.13. The output of the incore detectors is connected to the unit computer and is corrected for background, fuel depletion and the as-built dimensions to provide accurate outputs of relative neutron flux. The computer output of s ne corrected signals is used to develop core power distributions which provide power peaking information necessary to determine core p6cformance in terms of DNBR and LHR.

Implementation of the core power and core power imbalance safety limits, in terms of the reactor protection setpoints, is shown in Fi gure 4.3-1.

The out-of-core nuclear instrumentation provides the core power and core power imbaiance signals to the RPS, since the incore monitoring system does not imediately respond to prompt changes in core conditions.

l l

The results of 'the core power distributions taken during the test program as part ut the power escalation sequence are discussed by dividing this section into three subsections as follows:

-(1) Normal Operating Core Power Distributions (2) Worst N Minimum DNBR and Maximum LHR Calculations (3) Ou!drant Power Tilt and Axial Power Imbalance 4.3.1 Purpose The purposes of the steady state core power distribution measure-ments are as follows:

(a) To measure core power distribution and thermal-hydraulic data at the major test plateaus as required by the power escalation test program.

(b) To compare the measured and predicted core power distribu-tions at 40, 75, and 100% full power.

(c) To verify acceptable core thermal-hydraulic parameters at 40, 75, and 100% full power.

Four acceptance criteria are specified for the Core Power Ols-tribution Test and are listed below:

(1) The core power distribution and thermal-hydraulic parameters have been measured, evaluated, and deemed reasonable.

(2) The highest measured radial and total peaking factors are not more than 8.0 and 12 percent greater than predicted, respectively, :t 40% FP and 5.0 and 7.5 percent greater than predicted, respectively, at 75% FP and 100% FP.

(3) The measured worst case minimum DNBR is greater than 1.30 and the measured worst case maximum LHR is less than 18.00 kW/ft.

(4) The extrapolat?d worst case minimum DNBR is greater than 1.30 and the extrapolated worst case maximum LHR is less than 19.70 kW/ft., or the extrapolated imbalance falls out-side the power imbalance trip envelope as shown in Figure 4.3-1.

A review criterion was also added requiring the average root mean square of the calculated versus measured radial peak values of the 52 instrumented assemblies to be less than 0.072.

9

--- w , , -r~s

4.3.2 Test Method Computer printouts of the core power distribution and thermal hydraulict conditions were obtained after. establishing steady state conditions at the required power level and rod configura-tions as determined by the Controlling Procedure for Power Esca-lation, PT-120. Three-dimensional equilibrium senon was re-quired, the APSR's were maintained at a constant .,ition and the axial incore-imbalance was maintained within +/-24 FP of zero.

4.3.3 Evaluation of the Test Results 4.3.3.1 Normal Operating Core Power Distributions Normal operating, equilibrium xenon core power distributions were measured and predicted for each of the three major power escala-tion test plateaus.. The results of these measured core power distributions, given in Figures 4.3-2 to 4.3-7, at operating con-trol rod configurations indicate a maximum radial peaking factor of 1.34, and a maximum total peaking factor of 1.66 for the three cases studied. The total peaking factor observed during normal operation was well below the design maximum total peaking factor of 2.57.

The root mean square (RMS) value for the three radial peak power distributions were calculated via Tables 4.3-1 to 4.3-3. The RMS values at all three power levels were well below the review cri-terion os 0.072.

All acceptance and review criteria were met.

4.3.3.2 Worst Case Minimum DNBR and Maximum LHR Calculations To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal peaking conditions. The two primary core thermal limits which are indicative of fuel thermal perfor-mance are fuel melting - and departure from nucleate boiling.

These limits are independent; each must be evaluated to ensure core safety.for a given power peaking situation. Fuel melting is basically a function of the local power generated in the fuel which is a combination of radial and axial peaking. Ta maximum allowable linear heat rate limit is 19.70.

The upper boundary of the nucleate boiling region is termed

" departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperature and the possibility of cladding failure. The local DNB ratio, (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.

flow, temperature and pressure can be related to DNB through the use of the B&W-2 correlation. The B&W-2 correlation has been developed to predict DNB and the location of DNB for uniform and non-uniform axial heat flux distributions. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to 94.5% probability at a 99% confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operational conditions.

Worst Case Minimum DNBR Determination The worst case minimum DNBR values were calculated by the unit computer for each core power distribution taken as part of the power escalation test program. The results of various worst case minimum DNBR values calculated at each test plateau under normal rod configurations are_ plotted in Figure 4.3-8. These results indicate that all measured values were above the design worst case minimum DNBR versus power level and well above the minimum acceptable value of 1.30.

Three normal operating equilibrium xenon core power distribu-tions, as required by The Core Power Distribution Test, were obtained during the power escalation sequence. Each distribution was subjected to the following analysis:

(a) From each core power distribution, the worst case measured minimum DNBR was selected.

(b) The measured worst case minimum DNBR's were then extrapo-lated to the overpower trip setpoint and corrected for axial peak location and magnitude.

(c) Verification of acceptable core conditions at the present and next power level of escalation were then performed relative to 1.30.

(d) The measured worst case minimum DNBR's were then extrapo-lated to the LOCA and design overpower power level and cor-rected for axial peak location and magnitude.

Worst Case Maximum LHR Determination Worst case maximum LHR values were calculated using SP-104 " Hot Channel Factors Calculations" for each standard core power dis-tribution taken as part of the power escalation test program. In all cases, the above limit was met during the power escalation test program, Table 4.3-4.

i r- -

g

4.3.3.3 Quandrant Power Tilt and Av :1 Power Imbalance Quandrant Power Tilt Quadrant power tilt limits have been established in the Technical Specifications. These limit ;, when used in conjunction with the control rod position limits, assure that the design peak heat rate criterion is not exceeded during nomal power operation.

Quadrant power tilt is defined by the following equation and is expressed in percent.

.EQ. (4.3-2)

Quadrant Power = '[ Power in Any Core Quadrant I x 100 Tilt } Average Power of All QuadrantsJ During the startup testing program, maximum quadrant power tilt was determined using the corrected signals from the 16 symmetric incore monitoring assemblies. Figures 4.3-2 through 4.3-7 in-clude the maximum quadrant power tilt for each standard core power distribution taken as required by the Core Power Distribu-tion Test. Technical Specification tilt limits were not exceeded at any of the power plateaus.

Axial Power Imbalance Resul ts from the Standard Core Power Distributions taken at 40% FP during the performance of the Power Imbalance Detector Correlation Test, show that the imbalance trip envelope (Figure 4.3-1) of the reactor protective system is sufficient to protect the unit from exceeding the DNBR and the LHR limits under all core imbalance conditions when a gain factor of 4.50 is set into the delta flux amplifier. In addition, analyses indi-cate that the largest thermal margins (measured by DNBR and LHR) exis,t when a negative 5.0% to 10.0% incore axial offset is present.

The core imbalances measured in ccnjunction with the core power distribution of this section are included in Figures 4.3-2 through 4.3-7.

4.4 BIOLOGICAL SHIELD SURVEY A biological . shield survey was not conducted as part of this power testing program since no plant modifications were made which would invalidate the biological shield surveys made during the initial startup testing program.

4.5 PSEUD 0 R00 EJECTION TEST This test was not perfomed during the power escalation test pro-gram. This test was performed during zero power physics testing.

4.6 TURBINE / REACTOR TRIP TEST No turbine / reactor. trip testing was performed during this startup testing program since no modifications were made which would in-validate the original testing results.

4.7 INTEGRATED CONTROL SYSTEM TEST Since no modifications were made to the Integrated Control System during _ this outage no specific ICS testing was done during the startup testing program. Minor adjustments were made to the ICS during startup under normal maintenance and calibration proce-dures. The ICS functioned normally during the entire testing program.

4.8 UNIT LOSS OF ELECTRICAL LOAD No loss of electrical load testing was performed during this startup testing program since no modifications were made which would invalidate the original testing results.

4.9 UNIT LOAD TRANSIENT TEST No specific unit load transient testing was performed since no modifications were made - to invalidate the results of previous tests. No major problems were encountered during transient operations- during the testing program.

4.10 SHUTDOWN FROM OUTSIDE THE CONTROL ROOM No shutdown from outside the control room testing was performed since no modifications were made which would invalidate the re-sults during initial startups.

4.11 LOSS OF 0FFSITE POWER No loss of offsite power testing was performed since no modifica-tio.ns were made which would invalidate the results of the tests performed during initio. startup.

4.12 POWER IMBALANCE DFTECTOR CORRELATION TEST Imbalance of the neutron flux in a reactor results from tempera-ture distributions, fuel depletion, xenon oscillations, or con-trol rods positioned in the core. The amount of imbalance

• which is allowed in the core so that DNBR or LHR limits are not exceeded- is set in the reactor protection system and is a func-tion of the power level and the reactor coolant flow. Since this imbalance is determined using input signals from the out-of-core l

• Imbalance = % power in top of core - % power in bottom of core

\

detectors, it is essential that .they are calibrated to read the true imbalance as determined from the incore detectors.

4 This section of the report represents .the resul ts of incore imbalance to out-of-core imbalance induced on Crystal River Unit 3, Cycle.1 at 40 percent full power during the performance of Power Imbalance Detector Correlation Test.

4.12.1 Purpose The Power Imbalance Detector Correlation Test had the following

four objectives:

(a) To measure the relationship between core offset

• as indi-cated by the out-of-core power range nuclear instrumentation detectors and as indicated by the full incore monitoring system.

(b) To provide sufficient information to adjust the NI/RPS dif-ferential amplifiers should the measured correlation indi-cate a need.

(c) To verify that an acceptable relationship between the backup recorder syster and the full incore monitoring system indi-cation of core offset was observed.

(d) To verify acceptable core thermal-hydraulic parameters at each imbalance condition induced.

Three acceptance criteria are specified for the Power Imbalance Detector Correlation Test and are listed below:

(1) The measured correlation between each out-of-core detector offset to full incore monitoring system offset lies within the acceptance region shown in Figure 4.12-1. The correlation slope shall be greater than or equal to 1.00 and the correlation intercept should be less than + 3.5% offset.

(2) The measured relationship between the full incore monitoring system offset and the backup recorder offset lies within the acceptable region shown in Figure 4.12-2.

(3) The measured worst case minimum DNBR is greater than 1.30 and the measured worst case maximum LHR is less than 18.00 kW/ft.

• Offset = Imbalance Total Core Power

I l

4.12.2 Test Method During the Power Escalation Sequence at Crystal River 3, as part of the test program, imbalance measurements were made to deter-mhe the acceptability of the out-of-core detectors to detect j imoalance and to establish a basis for verifying that DNBR and LHR limits would not be exceeded while operating within the flux / delta flux / flow envelope set in the reactor protection system. These imbalance measurements were made at 40% FP with a gain factor of 4.50 applied to the multiplier on the delta flux amplifier output.

In performing the test, APSR's were positioned to obtain the desired full incore imbalance with reactivity compensations made by control rod groups 6 ana/or 7. At 40% FP, offset as indicated by the full incore system, out-of-core system, and backup record-er system was recorded as given in Table 4.12-1. From this data, plots of average out-of-core offset ' and backup recorder ' offset versus full incore offset were maintained as shown in Figures 4.12-1 through 4.12-2. Worst case minimum DNBR and maximum LHR data was monitored during the test. The results are plotted ver-sus full incore offset in Figure 4.12-3.

Based upon previous startup experience, the relationship between incore offset and out-of-core offset was determined to be a linear equation of the form below:

OC0 = M x ICO + B EQ.(4.13-1)

Where: OC0 = Out-of-Core Offset (Percent)

ICO = Incore Offset (Percent)

M = Slope of Relationship B = Intercept at Zero 100 The experimental slope and intercept could then be obtained using a linear least squares fit from the data obtained. If the mea-sured slope of the relationship of ICO to 0C0 was deter.ained to be unsatisfactory; i.e. <1.15, the gain of the out-of-core power range detector should be adjusted. The relationship of measured slope to gain factor is as follows:

GF = (M2/M1) x GFo EQ. (4.13-2)

Where: GF = Desired Gain Factor M2 = Desired Slope (i.e. >1.15)

M1 = Measured Slope GFo = Present Gain Factor (i.e. 4.50)

Verification of the adequacy of the power imbalance system trip setpoint was perfonned in conjunction with worst case analysis on each minimum DNBR and maximum LHR measured. The technique used  ;

1 I

l 1

l

was to extrapolate each measured point to the power / imbalance /

flow envelope boundary limits given in Figure 4.3-1. In this way, the adequacy of the imbalance system trip setpoints to pro-tect the unit from exceeding thermal-hydraulic limits could be veri fied.

4.12.3 Evaluation of Test Results The measurement of the offset correlation function between the full incore system and each out-of-core detector was determined during imbalance scans by APSR's and control rod group 6 and/or 7 at 40% full power, to be a linear relationship on all power range detectors. Figure 4.12-1 shows the average response in offset between the out-of-core power range detectors and the full incore system. Test data indicated that all measured offsets from the out-of-core power range detectors fell within the acceptable areas of the curve during the .perfonnance of the test. For each power range detector, a linear least squares fit was applied to the measured data points to obtain a value for the slope and intercept of the observed relationship. The results of these calculations are tabulated in Table 4.12-2. In all cases, the measured slopes were greater than the minimum allowable value of 1.15 which verified tha utilization of a 4.50 gain factor for the difference amplifiers.

The ability of the backup recorder to follow full incore offset was also verified as part of this test by collecting backup recorder data and performing the necessary calculations. The re-sults of this analysis are plotted in Figure' 4.12-2.

During all phases of testing, worst case minimum DNBR and maximum LHR were recorded against incore offset and a plot maintained as s'.own in Figure 4.12-3. The most limiting talue observed on the worst case minimum DNBR and maximum LHR was 9.35 and 5.22 kW/ft.,

respectively, which is well within the procedural acceptance criteria. As can be seen from the plots, both worst case minimum DNBR and maximum LHR are functions of incore offset.

4.13 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER TEST The purpose of the Nuclear Instrumentation Calibration at Power Test was to calibrate the power range nuclear instrumentation to within 2.0% full power of the thermal power as determined by_ a heat balance and to' within 3.5% of the axial offset as detennined by the incore monitoring system. Additional purposes during the power escalation program were as follows:

(a) To adjust the high power level trip setpoint when required by the power escalation test procedure.

1 l

(b) To confirm that the nuclear instrumentation calibration pro-cedure can adequately adjust the power range channels to a heat balance and an incore offset.

(c) To verify that at least one decade overlap exists between the intermeaiate and power range nuclear instrumentation.

Four acceptance criteria are specified for the Nuclear Instrumen-tation Calibration at Power Test as listed below:

(1) The power range nuclear instrumentation indicates the power level within +2.0% full power of the heat balance and wittiin 3.5% incore axial offset. - The incore axial offset criteria does not apply below 30% full pouer.

(2) The high power level trip bistable is set to trip at the de-sired value within a tolerance of +0.00 to -0.31% full power.

(3) The power range nuclear instrumentation can' be calibrated adequately using the nuclear instrumentation calibration procedure.

(4) The overlap between the intermediate and power range is in excess of one decade overlap.

The power range instrumentation channels were calibrated several times during the power escalation test program. The results of the calibrations met the criteria specified for the test.

4.14 EMERGENCY FEEDWATER FLOW TEST No emergency feedwater flow test was performed since no modifica-tions were made which would invalidate the results of the tests performed during Cycle 2 startup.

4.15 TURBINE / GENERATOR OPERATION No Turbine / Generator operational testing was performed during this testing program since no changes were made to the Turbine /

Generator which would invalidate the results of the tests con-ducted during initial startup.

4.16 DROPPED CONTROL R00 TEST The dropped rod test was not performed because sufficient thermal margin exists as indicated by B&W-2 correlation analyses.

b l

l 4.17- INCORE DETECTOR TEST Incore detector output was verified by comparing the corrected  !

detector response from similar core locations. All detector out-

> puts were normalized to the average detector output per assem-bly. Table 4.17-1 shows the results of this comparison at 40%

full power.

4.18 RCS HOT LEAKAGE TEST RCS hot leakage is monitored - on a regular basis during plant operation as required by Technical Specifications. No additional RCS leakage testing was performed at this time.

4.19 PIPE AND COMPONENT HANGER H0T INSPECTION AT POWER No pipe and component hanger hot inspection at power was done during this startup since no modifications had been made which would invalidate the results of the testing conducted at initial startup.

4.20 CHEMICAL AND RADI0 CHEMICAL TESTS Chemical and Radiochemical testing was not performed during this startup. These tests were conducted at initial startup and no plant modifications have been made which would invalidate the re-sults of those tests.

4.21 EFFLUENT AND F.FFLUENT MONITORING a

No effluent cr effluent monitoring testing was performed as these systems have been performing normally since initial startup and no further testing was required.

4 4 ---

i l

SUMMARY

.0F TEMPERATURE, MODERATOR, AND DOPPLER COEFFICIENTS OF REACTIVITY Maximum Linear Heat Rate (KW/ft.)

COEFFICIENT VALUE CONDITION Tem rature -0.519 99% FP, 960 PPM, 8 EFPD (10- ak/k/*F)

Mooc'ator -0.367 99% FP, 960 PPM, 8 EFPD (10

• ak/k/*F)'

Power Doppler- -0.923 98% FP, 960 PPM, 8 EFPD (10-4 Ak/k/%FP)

Table 4.1-1 I

=-- - . . -. . - . _ _ .-- -

DETERMINATION OF. RMS FOR CALCULATED VS. KASURED RPD'S @ 40% FP Detector Detector Measured Calculated 1/4 Calculated Detector Detector Measured Calculated 1/4 Calculated-

' String #. Location Peak Core Location Peak String # Location Peak Core Location Peak 1 H-08 1.083 H-08 1.016 27 D-10 1.230 N-10 1.197 2 H-09 1.197 H-09 1.114 28 C-10 1.238 0-10 1.257 3 G-09 1.214 K-09 1.258 29 t-09 1.204 o-09 '1.252 4- F-08 1.200 L-08 1.283 30 8-08 0.537 P-08 0.491 5 E-09 1.345 M-09 1.335 31 B-07 0.790 'P-09 0.809 6 F-07 1.161 L-09 1.131 32 C-06 1.240 0-10 1.257 7 E-67 1.312 M-09 1.335 33 D-05 1.268 N-11 1.265; 8 G-06 1.215 K-10' 1.132 34 3-04 1.280 M-12 1.292-9 G-05 1.301 K-11 1.338 35 F-03 1.265 L-13 11.264 10 H-05 1.219 H-11 1.166 36 G-02 0.859 K-14 0.813 11 K-05 1.289 K-11 1.338 37 H-01 0.501 H-15 ~0.457 U 12 L-06 0.815 L-10 0.714 38 L-02 1.102 L-14 1.103-E 13 M-07 1.283 M-09 1.335 39 L-03 1.266 L-13 1.264 14 N-08 1.337 N-08 1.3 72 . 40 M-03 1.063 M-13 1.021

? 15 N-09 1.120 N-09 1.176 41 N-04 1.0 52 N-12 1.052 -

Y

~

16 M-09 1.287 M-09 1.335 42 0-05 0.%5 0-11 1.014 17 M-10 1.090 M-10 1.054 43 0-06 1.225 0-10 1.257 18 L-11 1.104 L-11 1.057 44 P-06 1.060 P-10 1.099

19 K-1 1.283 K-11 1.338 45 R-07 0.518 R-09 0.543 20 K-12 1.188 K-12 1.180 46 R-10 0.518 R-10 0.507 21 H-13 0.993 H-13 0.963 47 0-10 1.219 0-10  ?.257 22 G-13 1.214 K-13 1.257 48 0-12 1.019 0-12 M 23 F-13 1.208 L-13 1.264 49 M-14 0.887 M-14 0.6, 24 F-12 1.163 2-12 1.204 50 L-13 1.242 L-13 1.264 25 G-11 1.289 K-11 1.338 51 D-14 0.561 N-14 0.596 26 E-11 1.084 M-11 1.016 52 C-13 0.670 0-13 0.706 RMS = E(Measured-Calculated)2 52 RMS = 0.0431 .

DETERMINATION OF RMS FOR CALCULATED VS. EASURED RPD'S 0 75% FP Detector Detector. Measured Calculated 1/4 -Calculated Detector Detector Measured Calculated 1/4 Calculated-String # Location Peak Core Location Peak String # Location Peak Core Location -Peak 1 H-08 1.0 96 H-08 1.012 27 D-10 1.191 N-10 1.185 2 H-09 1.210 H-09 1.104 28 C-10 1.230 0-10 1.250 3 G-09 1.219 K-09 1.242 29 u-09 ' 1.202 - o-09 1.244 4 F-08. .1.107 L-08 1.218 30 B-08 0.529 P-08 0.501 5 Es09 1.333 M-09 1.315 31 B-07 0.782 P-09 0.820 6 F-07 1.155 L-09 1.120 32 C-06 1.236 0-10 1.250 7 E-07 1.30 8 M-09 1.315 33 0-05 1.260 N-11 1.257 8 G-06 1.214 K -10 1.120 34 3-04 1. 270 M-12 1.283 9 G-05 1.303 K-11 1.317 35 Fs03 1.277 L-13 1.256 10- H-05 1.214 H-11 1.154 36 G-02 0.8 56 K-14 0.823.

11 K-05 1.296 K-11 1.317- 37 H-01 0.511 H-15 0.478-ed 12 L-06 0.812 L-10 0.713 38 L-02 1.049 L-14 1.109 55 13 Ma07 1.286 Ma09 .1.315 39 L-03 1.218 L 1.256 14 N-08 1.343 N-08 1.353 40 M-03 1.061 M-13 1.025 f' 15 N-09 1.105 N-09 1.165 41 N-04 1.052 N-12 1.055 I"

16 M-09 1.290 M-09 1.315 42 0-05 0.961 0-11 1.108 17 -M-10 1.088 M-10 1.047 43 0-06 1.219 0-10 1.250 18 L-11 1.107 L-11 1.0 50 44 P-06 1.076 P-10 1.105 la K-1 1.287 K-11 1.317 45 R-07 0.546 R-09 0.564

20 K-12 1.194 K-12 1.169 46 R-10 0.533 R-10 0.526 21 H-13 0.997 H-13 0.965 47 0-10 1.220 .0-10 1.250-22 G-13 1.210 K-13 1.249 48 0-12 1.028 0-12 1.088 23 F-13 1.20 7 L-13 1.256 49 M-14 0.900 M-14 0.888

- 24 F-12 1.145 2-12 1.193 50 L-13 1.238 L-13 1.256-25 G-11 1.292 K-11 1.317 51 D-14 0.569 N-14 0.613 26 E-11 1.085 M-11 1.014 52 C-13 0.681 0-13 0 .722 RMS = ' (Phasured-Calculated)2 I

52

)

RMS = 0.0468 I

e

DETERMINATION OF RMS FOR CALCULATED VS. KASURED RPD'S 0100% FP

, Detector Detector . Measured Calculated 1/4 Calculated Detector Detector Measured Calculated 1/4 Calculated String # Location Peak- Core Location Peak String # Location Peak Core Location Peak 1 H-08 1.096 H-08 0.999 27 D-10 1.190 N-10 1.175 2 H-09 1.216 H-09 1.087 28 C-10 1.229 0-10 1.246 3 G-09 1.220 . K-09 1.22o 29 C-09 1.202 0-09 1.237 4 F-08 1.191 L-08 1.199 30 B-08 0.529 P-08 0.508-5- E-09 1.323 M-09 1.293 31 B-07 0.782 P-09 0.830 6 F-07 1.150 L-09 1.102 32 C-06 1.231 0-10 1.246 7 E-07 1.300 M-09 1.293 33 D-05 1.256 N-11 1.251 8 G-06 1.208 K-10 1.103 34 3-04 1.274 M-12 1.277-9 G-05 1.304 K-11 1.296 35 F-03 1.275 L-13 1.252 10 - H-05 1.210 H-11 1.137 36 G-02 0.859 K-14 0.832 11 K 1.295 K-11 1.296 37 H-01 0.515 H-15 0.498 U 12 L-06 0.815 L-10 0.70 5 38 L-02 1.017 L-14 1.121 5 13 M-07 1.284 M-09 1.293 39 L-03 1.221 L-13 1.252 14 N-08 1.343 N-08 1.335 40 M-03 1.067 M-13 1.030 P 15 N-09 1.143 N-09 1.152 41 N-04 1.050 P-12 1.060 Y

16 M-09 1.290 M-09 1.293 42 0-05 0.959 l-11 1.024 17 M-10 1.0 94 M-10 1.036 43 0-06 1.217 0-10 1.246 18 L-11 1.111 L-11 1.038 44 P-06 1.0 72 P-10 1.117 19 K-1 1.291 K-11 1.296 45 R-07 0.549 R-09 0.585 20 K-12 1.195 K-12 1.156 46 R-10 0.538 R-10 0.54 7 21 H-13 1.000 H-13 0.963 47 0-10 1.227 0-10 1.246 22 G-13 1.226 K-13 1.242 48 0-12 1.022 0-12 1.102 23 F-13 1.209 L-13 1.252 49 M-14 0.898 M-14 0.905 24 F-12 1.126 2-12 1.181 50 L-13 1.240 L'-13 1.252 25 G-11 1.296 K-11 1.296 51 11 - 1 4 0.566 N-14 0.635 26 E-11 1.096 M-11 1.009 52 C-13 0.674 0-13 0 . 74 2 RMS = r (Measured-Calculated)2

\ 52 RMS = 0.0494

6 MAXIMUM LINEAR- HEAT' RATE PER INCORE DETECTOR LEVEL VERSUS POWER LEVEL Maximum Linear Heat Rate

~ (XW/ft.)

LOCA Limit IC Level 40% FP 75%'FP 100% FP. (KW/ft.)

I 1 2.94 '5.85 8.13 15.0 2 4.82 8.17- 10.41 15.8 3 4.56. 7.95. 10.19 16.8 4 '4.20 7.65 10.35 .18.0 5 - 4.31 8.32. -11.22 17.1 6' 4.38 8.83' 11.56 16.3 7 2.94 5.59 7.07 -15.4

}

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Table 4.3-4 i .

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m i e r - s, , -- , . , . . y-,,, - *. --'--

SUM %RY & TEST [ATA POWER IPEALANCE DETECT (R CORRELATION TEST 40% FP Power Rod Position, %wd Incore Offset, % Out-of-Core Offset, % Worst Case - Worst Case- ,

Date Level Maximum LIE Mininw: DfER Time (%FP) 1-5. 6 7 8 Full Backup NI-5 NI-6 NI-7 NI-8 (kW/ft) 8/12/80 40.354 100 85 11. - 27.0 13.11 11.60 14.75 14.90 14.95 14.23 4.95 9. , .'

1151 fi/12/80 40.117 100 85 11 31.0 6.29 6.45 6.32 6.41 6.50 6.38 4.63 9.89 1226

1. 8/12/80 40.445 100 85 11 34.5 -0.05 2.25 -0.73 -0.65 -0.48 -0.33 4.30 10.60

{ 1314

? 8/12/80 40.460 100 85 12 37.5 -6.28 -3.50 -8.46 -8.30 -8.18 -7.69 4.49 11.42 M 1344

,L .

8/12/80 40.667 100 85 11 41.0 -11.06 -18.21 -18.21 -18.01 -17.80 -17.33 4.% '10.33 1426 8/12/80 40.;94 100 85 11 42.5 -15.47 -24.32 -24.32 -23.92 -23.63 -23.15 5.22 9.77 1540

Summary of Least Squares Linear Regression Analysis for Each Power Range e Channel and Backup Recorders During the Power Imbalance Ccrrelation' Test Detector Gain Data Intercept Slope

. System. Factor Sets Calculated Allowable Calculated Allowable Power Imbalance Test at 40% Full Power NI-5 . 4. 5 .6 -1.044 +3.5 1.213 >1,15 NI-6 4.5 6 -0.885- T3.5. 1.204 ->1.15 NI-7 4.5 6 -0.754 T3.5 1.199 >1.15 -

4 NI-8 4.5 .6 -0.748 +3.5 1.164 >1.15 Backup ---

6 1.224 T5.0 0.849 1.00+0.20

. C 3-m.

G k

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

Comparison of Incore Monitoring Assenblies Flux Shapes at 40%, FP Using the Indicated Grouping Given in Incore Detector Test-Incore Fuel Detector Current to Avera9e Detector Current per Assenbly Group Detector Assenbly Nunber Nunber Location Level 1 . Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (BG) 01 05 E-09 0.798 1.155 1.200 1.027 1.083 1.099 0.637 0.144 07 E-07 0.785 1.148 1.1% 1.016 1.055 1.144 0.657 0.144 09- G-05 0.806 1.133 1.165 1.009 1.093 1.145 0.649 0.152 11 . K-05 0.809 1.137 1.189 0.998 1.089 1.134 0.644 0.148

13. M-07 0.815 1.144 1.171 1.016 1.081 1.127 0.647 0.130 16 M 0.810 1.142 1.194 1.006 1.0 72 1.137 0.638 0.126 19 K-11 0.820 1.158- 1.201 1.009 1.057 1.118 0.638 0.148 25 G-11 0.812 1.153 1.205- 1.018 1.063 1.122 0.628 0.150 .

1. 02 23 F-13 0.793 1.194 1.172 0.9 72 1.055 1.146 0.668 0.169 5 28 C-10 0.777 1.188 1.158 0.958 1.084 1.144 0.692- 0.112 32 C-06 0.775 1.197 1.197 0.9A) 1.0 52 1.130 0.678 0.158

? 35 F-03 0.762 1.156 .1.180 0.964 1.048 1.168 0.721 0.144 C 39 L-03 .0. 778 1.174 1.146 0.968 1.087. 1.1C 0.6% 0.172

d. 43 0-06' ' O . 7 74 1.193 1.141 0.962 1.091 1.150 0.688 0.152 47 0-10 0 . 7 74 1.174 1.163 0.958 1.082 1.167 0.684 0.126 50 L-13 0.782 -1.202 1.180 0 . 9 71 1.067 1.137 0.661 0.149 03 15 N-09 0.766 1.208 1.203 0.944 1.127 1.180 0 .5 73 0.692 17 M-10 0.866 1.164 1.166 0.934 1.045 1.138 0.687 0.157 18 L-11 0.857 1.201 1.187 0.953 0.985 1.126 0.693 0.159 20 K-12 0.777 1.177 1.202 0.939 1.037 1.131 0. 73 7 0.158 33 D-05 0.773 1.236 1.181 0.954 1.110 1.136 0.611 0.155 34 E-04 0.864 1.365 1.178 1.004 1.165 1.265 0.692 0.159 04 24 F-12 0 .8 71 1.336 1.313 0.691 0.903 1.164 0.722 0.141 27 D-10 0 .8 73 1.288 1.177 0.669 0.993 1.247 0.751 0.148 4

. L .

687 55 204330 62 53 9111 0

) 454 33 366784 37 55 68936 G 101 11 111112 11 1 1 1111 1 B

( 000 00 000000 00 00 00000 7 -

y 118 94 290896 11 91 25488 l l 868 47 034796 24 90 00704 b e 555 77 545556 67 67 77665 m v e

s e 000 00 000000 00 00 00000 L

s A -

P r 6 F e p

955 01 946563 10 28 54857 l 573 53 298065 72 35 37497

%t e v 001 12 190100 01 11 01100 0s t 4e n e 11I 11 101111 11 11 11111 t

T e r

L ar r o u st C 5 ec pe r 140 19 219150 24 93 15509 l 533 15 174483 37 56 89615 at o e 111 11 110190 10 00 0011 1 h e t v SD c e 111 11 111101 11 11 11111 e L xe t ur e l o D Fc 4 n e 048277

_ sI e a 9 l e

310 657 100 5%

0 09 529308 010109 54 14 00 46 16 09 10297 62203 0011 1

_ i n r v l i e v

e 111 10 11111.0 11 10 11111 tnn A L

ee

_ sv si t o

3 .

AG t 492 07 65 773768 73 88 71971 l 606 095369 95 56 83354

. gg n e 121 12 322300 11 11 11112 nn r e v

. i i rp e 111 11 111111 11 11 11111 r L ou u t o C i r nG r 2 o o 935 75 164482 24 1' 0 53963 Md t l 507 81 760186 81 68 38282 e c e 121 03 011200 11 11 10011 et e v ra t e 111 11 111111 11 11 11111 oc e L ci D nd I n .

I 1 f 480 97 800994 04 53 84446 oe l 264 30 217980 85 76 99231 h e 777 79 888811 77 77 77777 nt v

._ o sg L e 000 00 000011 00 00 00000 i n ri as pU yn l o m lbi 826 81 987668 53 39 72049 o ent 000 01 000000 01 10 00110

_ C uea HLP NE Fsc GFFGLB HH GG BGRMH so

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. ocb cen 184 034 46 12 346620 000013 01 12 29 22 16692 33440 nt u De N.

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Comparison of Incore Monitoring Assenblies Flux Shapes at 40% FP Using the Indicated Grouping Given in Incore Detector Test Incom Fuel Detector Current to Average Detector Curmnt per Assenbly Group Detector Assenbly Nunber Nunber Location Level 1 Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (BG)

. 11 37 H-01 0.801 1.0 73 1.124 1.114 1.104 1.143 0.643 0.185-45 R-07 0.781 1.156 1.137 1.125 1.113 1.156 0.532 0.162 48 0-12 0.750 1.164 1.344 .1.093 1.0 30 1.111 0.509 0.151 51 0-14 0.686 1.146 1.220 1.154 1.126 1.149 0.518 0.196 52 C-13 0.690 1.164 1.1% 1.144 1.136 1.142 0.528 0.168

- 12 40 M-03 0.757 1.166 1.154 0.984 1.113 1.182 0.643 0.178 77 41 N-04 0.793 1.231 1.237 1.085 1.143 1.133 0.3 79 0.123

if 42 0-05 0.795 1.215 1.146 1.027 1.086 1.139 0.591 0.158

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Figure 4.3-1

Comparison of Precicted and luasured Radial Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium Zenon, 40% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 40.0_ %FP Gps 1-5 100.0  % wd Boron Concentration ppmR Cp 6 87.3  % wd Core Burnup 2.0 EFPD Cp 7 12.3  % wd Axial Imbalance 0.07 %FP Cp 8 ~3373  % wd Max Quadrant Tilt 0.0 %

Measured Conditions Control Rod Group Positions Cure Power Level 39.85 %FP Gpa 1-5 100.0  % wd tioron Concentration 1090 ppmB Gp 6 87.0  %'wd Core Burnup 0.66 EFPD Cp 7 13.0  % wd Axial Imbalance T.47 %FP Cp 8 35.5  % wd Max Quadrant Tilt - f?U7 %

H K L M N O P R 8

1.033

-d:2 9 1.114 1.258 l.197 1.214

-6.9 3.6 1.234 1.132 0.714 1.200 1.188 0.815 10 2.8 -4.7 -12.4 1.166 1.338 1 .057 1.016 11 1.219 1.299 1 . 0 97 1.084

-4.3 3.0 -3.6 -6.3 1.377 1.180 1.204 1 .2 92 1.052 12 1.337 1.154 1 .! 97 1.274 1.052 3.0 2.3 0.6 1.4 0.0 0.963 'l.257 1 .2 64 1.021 1.084 0.706 13 0.993 1.209 1.238 1.014 1.019 0.670

-3.0 4.0 2.1 0.7 6.4 5.4 0.492 0.81 3 1 103

. 0.876 0.596 14 0.537 0.825 1.081 0.887 0.561

-8.4 -1.5 2.0 -1.2 6.2 15 0.457 0.543 0.506 0.501 0.518 0.518

-8.8 4.8 -2.3 X.XXX Predicted Results .

X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 1007.

Measured Figure 4.3

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

Comoarison of Predicted'and flaasured Total Peaking Core Power Distribution i- Results at Steady State, 3-D Equilibrium Xenon, 40% FP Conditions Predicted Conditions i

Control, Rod Group Positions ' Core' Power Level 40.0 %FP Cps 1-5 100.0 % wd- Boron concentration ppmn Gp 6 87.3 % wd Core Burnup 2.0 EFPD bg 7 12.3 % vd Axial Imbalance ~DID7 %FP Cp 8 ~~35I3 % wd Max Quadrant Tilt 0.0 %

Measured Conditions Control Rod Group Positions Core Power Level 39.85 %FP Cps 1-5 100.0 % wd Baron Concentration W ppmn Gp 6 TO % wd Core Burnup W EFPD Cp 7 13.0. % wd Axial Imbalance -lllCI%FP j Cp- 8 T % wd Max Quadrant Tilt M%

, H- K L 31 N O P R 1.238 8

1.280 I'

-3.3 1 . 34 8 1.520 9

1.432 1.432

-5.9 6.1 1.524 1.377 0.773 1.483 1.426 0.951 0

2.8 -3.4 -18.7 1.353 1.552 1.259 1.210 11 1.432 1.526 1.344 1.305

-5.5 1.7 -6.3 -7.3 1.571 1.419 1.611 1.521 1.255 12 1.508 1.395 1.660 1.572 1.305 i 4.2 1.7 -3.0 -3.2 -3.8 1.117 1.470 1.51 3 1.186 1.301 0.853 13 1.103 1.439 1.486 1.211 1.267 0.798 1.3 0.8 1.8 -2.1 2.7 6.9 0.532 0.933 1.288 1.036 0.722 14- 0.608 0.951 1.255 1.052 0.684 212.5 -1. 9 2.6 -1. 5 5.6 0.566 0.644 0.608 ,

15 0.583 0.634 0.621

-2 . 9 1.6 4.1 X.XXX Predicted Results X.XXX- Measured Results XX.X Deviation-(Predicted - Measured) x 100%-

Measured

Comparison of Predicted and iteasured Radial Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium Xenon, 75% FP Conditions Predicted Conditions l l

l Control Rod Group Positions Core Power Level 7L Q__ %FP Gps 1-5 100.0 % wd Baron Concentration ppen Op 6 8T3 % wd Core Burnup 3.0 EFPD Cp 7 12.3 % wd Axial Imbalance U.ZZ %FP Gp 8 32.0 % wa Max quadrant Tilt 0.0 %

Measured Conditions Control Rod Group Positions Core Power Level M %FP Cps 1-5 100.0 % wd Boron Concentration 1020 ppma Gp 6 88.0 % wd Core Burnup ___2.06 EFPD Cp 7 . 13.0 % wd Axial Imbalance 0.0 -%FP Gp 8 3.10 % wa Max Quadrant Tilt 1.25 %

H K L M N O P R 1.012 8 1.096

-7.7 1.104 1.242 9 1.210 1.219

-8.8 1.9 1.219 1.120 0.713 1.197 1.185 0.812 10 1.8 1.9 -12.1 1.154 1.31 7 1.050 1.014 11 1.214 1.299 1.098 1.085 4.9 1.4 -4.3 -6.5 1.3b6 1.169 1.193 1.283 1.055 12 1.343 1.149 1.168 1.265 1.052 1.1 1.7 2.1 1.4 0.3 0.965 1.249 1.256 1.025 1.091 0.722 13 0.997 1.210 1.231 1.011 1.028 0.681

-3.2 3.2 2.0 -1. 3 0.3 6.0 0.503 0.823 1.109 0.883 0.613 14 0.529 0.819 1.063 0.900 0.569 -

-4.9 0.5 4,3 -1. 3 7.7 15 0.478 0.564 0.526 0.511 0.546 0.533

-6.4 3.2 1.3 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

• Measured Figure 4.3 _4.

Comparison of Predicted and fleasured Total Peaking Core Power Distribution Results.at Steady St' ate, 3-D Equilibriun Xenon, 75% FP conditions PredicLed Conditionn Control Rod Group Positions Core Power Level '75.0 %Fr Gps 1-5 100.0 % wd Boron concentration ppmn Cp 6 lC  % wd Core Burnup W EFPD Gp 7 T '% wd Axial Imbalance 0.0 %FP Cp 8 M  % wd Max Quadrant Tilt V%

Measured Conditions _

Control Rod Group Positions Core Power Level 74.ll %FP cps 1-5 100.0 % wd Boron Concentration T020 ppmn Gp 6 ' 88.8 % wd Core Burnup 3 6 EFPD cp 7 13.0 % wd Axial Imbalance W %FP cp 8 31.0 % wa Max quadrant Tilt E%

II K L M N O P R 1.239 8 1.275

-2.8 1.342 1.507 9 1,410 1,418 5.6 6.2 1.497 1.346 0.763 1.431 1.384 0.927 10 4.6 -2.7 -17.7 1.334 1.521 1.238 1.198 11 1.391 1.486 1.285 1.282

-4.1 2.3 3.6 -6.5 1.53/ 1.387 1.b/E 1.434 1.275 ,

12 1.568 1.370 .l.558 1.500 1.268

-2.0 1.2 1.2 1.1 0.6 1.105 1.449 1.475 1.194 1.302 0.869 13 1.145 1.394 1.449 1.193 1.234 0.818

-3.5 3.9 1.8 0.1 5.5 6.2 0.542 0.951 1.305 1.053 0.745 14 0.593 0.947 1.241 1.043 0.688

-8.6 1.1 5.2 1.0 8.2 13 0.596 0.677 0.639 0.600 0.654 0.641

-0.7 3.5- -3.1 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100% ,

Measured Figure 4.3-5 f

Comparison of Predicted and !Teasured Radial Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium Xenon,100%-FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 100.0 %FP Gps 1-5 100.0 % wd Boron concentration 992 ' ppmn Cp 6 M  % wd Core Burnup 4.0 EFPD Cp 7 U  % vd Axial Imbalance D %FP Cp 8 W  % ud Max Quadrant Tilt 0.0, %

Measured Conditions Control Rod Group Positions Core Power Level 98.0 %FP l Cps 1-5 100.0 7. wd Baron Concentration 74F ppmn l Cp 6 80.0 % wd Core Burnup W EFPD Cp 7 12.0 % wd Axial Imbalance - W %FP Cp 8 T  % wd Max Quadrant Tilt 0.0~  %

11 K L M N O P R 0.999 8 1.096

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-3. 7 2.3 1.7 1.6 8.1 10.1 0.509 0.832 1.121 0.905- 0.635 14 0.529 0.821 1.045 0.898 0.566

-3. 7 1.3 7.2 0.0 ,12.2 0.498 0.584 0.547 15 0.515 0.549 0.538

-3.3. 6.3 1.6 X.XXX Predicted Results '

X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measu red -

Figure 4.3-6

i Comparison of Prddicted and !!easured Total Peaking Core Powar Distribution l Results at Steady St' ate, 3-D Equilibrium Xenon,100% FP Conditions Predicted Conditions Control Rod Group Positions '

Core Power Level 100.0. %FP Gps 1-5 100.0 % wd Boron Concentration 999 ppma Gp 6 87.3 % wd Core Burnup _4 a_ EFPD Cp 7 .12.3  % vd Axial Imbalance -1.17 %FP Cp 8 28.8  % wd Max Quadrant Tilt 0.0  %

Measured Conditions Control Rod Group Positions Core Power Level 98.0 %FP Cps 1-5100.0  % wd Boron Concentration 947 ppmB Cp 6 88.8  % vd Core Burnup 4.88 EFPD Cp 7 TZT6  % wd Axial Imbalance -0.94 -%FP Cp 8 ~E626  % vd Max Quadrant Tilt 0.0  %

H K L M N O P R I.228 8 1.268

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-6.7 4.9 1.461 1.326 0.791 '

l.429 1.377 0.916 10 3.6 -3.7 13.6 1.321 1.506 1.225 1.183 11 1.377 1.501 1.527 1.289

-4.1 3.3 -19.7 -8.2 1.527 1.376 1.541 1.506 1.293 12 1.579 1.377 1.563 1.522 1.304

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O.552 0.963 1.320 1.074 0.769 14 0.600 0.947 1.242 1.061 0.694

-0.8 1.6 6.2 1.2 10. 8 15 0.61 7 0.699 0.664 0.606 0.668 0.652 -

i 1.8 1.6 1.8 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

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