Text

Cycle 4 Startup Rept,Mar 1982.
	ML20041F540
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Site:	Crystal River
Issue date:	03/31/1982
From:	; FLORIDA POWER CORP.
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Crystal River Unit 3 Nuclear Generating Plant CYCLE 4 STARTUP REPORT MARCH 1982 Power

9. Florida C O R P O R ATIO N DOCKET NO. 50 - 302 LICENSE NO. DPR - 72 g

f$0Ao0ck*o!$0S

1.0 INTRODUCTION

AND

SUMMARY

1.1 INTRODUCTION

On September 28, 1981, Crystal River Unit 3 was shut down for re-fueling. The fuel reached a core average exposure of 322.9 EFPD or 9650.8 mwd /MTU. Nine assemblies from Batch 2, sixty assemb-l lies from Batch 3, and three assemblies from Batch 4 were dis-charged from the core. The remaining assemblies were then shuffled in the core and 63 assemblies from Batch 6 inserted. In addition, five assemblies from Batch 5 were inserted. These five assemblies were fresh assemblies which had not been inserted at the beginning of Cycle 3. Also, four assemblies from Batch 4 were reinserted. During the refueling outage between Cycle 2 and Cycle 3, an assembly from Batch 4 was discovered to have a broken spring, so it and its three symmetric assemblies were removed.

During Cycle 3, the assembly was repaired and made available for reinsertion. The final Cycle 4 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 SumARY

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

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

I Criticality was achieved on December 10,1981 at 1646 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 December 13, 1981 at 0230 and Power Testing was performed. The test methods used and results achieved are reported in Section 4.0.

Core L0ading Diagram Crystal River 3, Cycle M 7 8 9 10 11 12 13 14 15 i 2 3 4 5 6 X

5 5 5 5 5 A

R9 N2 R8 N14 R7 5 6A 5 6A 4C 6A 5 f 5 B M14 R10 C9 R6 M2 5 6B 5 6A 4C 6B 4G 6A 5 65 5 C 013 L2 E7 E9 L14 03 6A 4C 6A 4C 6A 4C 6A 4C 6B 5

$ 6B 4C D M12 L3 F8 L13 M4 P 5' P11 6A 4C 6A 4C 4A 4C 6A 4C 6A 5 6A 6A 5 E

B10 C7 D6 y(D DIO C13 B6 5 4C 6A 'C 4 6A 4C 6A 4C 6A 5 5 F 5 5 6A 4C D5 D8 D11 C6 L1 K1 K15 L15 C10 4C 6A 5 5 5 6A 4C 6A 4C 6A 5 5 5 4C 6A C4 C12 D13 F12 C11 B4 B12 C5 F4 4A 4C 5 4C 5 4C 4A 4C 65 4C 5 5 4C 6B 4C -Y HW- HIS C3 H6 J2 H4 D3 K9 N13 H12 g H10 K13 HI 6A 4C 6A 5 5 5 6A 4C 6A 4C 6A 5 5 6A 4C K N3 04 012 L12 K11 P4 P12 K5 L4 4C 6A 4C 6A 4C F 4C 6A 5 5 5 5 6A 4C 6A g N8 N11 06 F1 C1 CIS F15 010 N5 6A 4C 6A 4C 4A 4C 6A 4C 6A 5 6A 6A

5 09 P6 M P10 K3 N6 N14 N10 (Cv2) 6B 4C 6A 4C 6A 4C 6A 4C 6A 4C 68 5 5

N F3 L8 F13 E4 B5 B11 E12 5 6B 5 6A 4C 6B 4C 6A 5 65 5 0 F2 M7 M9 F14 C3 C13 l

5 6A 5 6A 4C 6A 5 5 5 P E14 A10 07 A6 E2 5 5 5 5 5 j A9 D2 A8 D14 A7 l

z Batch Previous core location Figure 1.1-1

i l . .

2.0 PRECRITICALITY TESTING 2.1 REACTOR COOLANT FLOW TEST 2.1.1 Purpose The purpose of Reactor Coolant Flow Test was to measure the reactor coolant flow characteristics with four reactor coolant pumps operating and compare the reactor coolant flow with design calculations to verify adequate core flow. This test was done by verifying the following acceptance criterion:

Steady state total reactor coolant flow with four pumps in operation at approximately 532*F and 2155 psig shall be l between 374,880 gpm and 411,840 gpm. The steady state loop l f flows shall be within 2% of each other. l 2.1.2 Test Method i The steady state temperature, pressure, and pressure drop across the two flow loops were taken every minute for ten minutes.

These values werr. than averaged and the reactor coolant flow for each flow loop was calculated.

2.1.3 Evaluation of Test Results The reactor coolant flow was calculated to be 400,127 gpm. After correcting this value for measurement uncertaintles,' the accep-tance criteria above was met. The difference in flow between the two loops was 0.03%.

2.2 CONTROL R0D 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 l 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 signal to indicate three-fourths insertion was furnished to the recreders 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 time was CRS-1 with a rod drop time of 1.26 seconds. The rod with the longest drop time was CR1-1 with a rod drop time of 1.35 seconds. All of the control rods meet the acceptance criterion.

2.3 CHEMICAL AND RAD 10 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 results of those tests.

2.4 PRESSURIZER EFFECTIVENESS 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 detennine that the detector response data followed a normal distribution.

l 2.6 IN-SERVICE LOOSE PARTS AND VIBRATION MONITORING SYSTEM {

No loose parts and vibration monitoring system testing was done I

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.

I i

J

l

)

l l

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

CR1-1 C-11 1.35 CR1-2 E-13 1.33 CR1-3 M-13 1.33 CR1-4 0-11 1.35 CR1-5 0-5 1.35 CR1-6 M-3 1.34 CR1-7 E-3 1.33 CR1-8 C-5 1.33 CR2-1 E-9 1.34 CR2-2 G-11 1.34 CR2-3 K-11 1.33 CR2-4 M-9 1.31 CR2-5 M-7 1.32 CR2-6 K-5 1.33 CR2-7 G-5 1.34 CR2-8 E-7 1.34 CR3-1 D-8 1.34 CR3-2 G-9 1.33 CR3-3 E-11 1.34 CR3-4 H-12 1.33 CR3-5 K-9 1.34 CR3-6 M-11 1.34 CR3-7 N-8 1.33 -

CR3-8 K-7 1.33 CR3-9 M-5 1.34 CR3-10 H-4 1.34 CR3-11 G-7 1.33 CR3-12 E-5 1.34 CR4-1 H-8 1.34 CR4-2 B-10 1.31 CR4-3 F-14 1.33 CR4-4 L-14 1.32 CR4-5 P-10 1.31 CR4-6 P-6 1.32 CR4-7 L-2 1.32 CR4-8 F-2 1.32 CR4-9 B-6 1.29 Table 2.2-1

CONTROL R00 DRIVE DROP TIME TEST RESULTS DROP TIME CONTROL ROD CORE POSITION (Sec.)

CRS-1 C-9 1.26 CR5-2 G-13 1.30 CRS-3 K-13 1.30 CRS-4 0-9 1.26 CRS-5 0-7 1.28 CRS-6 K-3 1.28 CRS-7 G-3 1.26 CRS-8 C-7 1.28 CR6-1 F-8 1.30 CR6-2 D-12 1.32 CR6-3 H-10 1.32 CR6-4 N-12 1.34 CR6-5 L-8 1.32 CR6-6 N-4 1.30 CR6-7 H-6 1.32 CR6-8 D-4 1.32 CR7-1 B-8 1.31 CR7-2 F-10 1.32 CR7-3 H-14 1.32 CR7-4 L-10 1.33 CR7-5 P-8 1.32 CR7-6 L-61 1.32 CR7-7 H-2 1.32 CR7-8 F-6 1.33 l

1 1

Table 2.2-1 (Cont'd)

_ - - - - - - _ - - - - - - - - _ . . - _ . - _ _ J

. = - - . . .

i I

GROUP AVERAGE

CONTROL R00 DRIVE DROP TIME

AVERAGE DROP GROUP TIE (Sec.)

i 1 1.339 2 1.331 3 1.336 3 4 1.318 i 5 1.278 6 1.318

. 7 1.321 l 1-7 1.321 i

i i

Table 2.2-2

3.0 ZER0 POWER PHYSICS TESTS 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 sunnary 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 December 10, 1981 at 1646 hours0.0191 days <br />0.457 hours <br />0.00272 weeks <br />6.26303e-4 months <br /> at reactor coolant conditions of 532 F, 2155 psig, and 1384 ppmB. Control rod groups 1 through 4 had previously been withdrawn as part of the heatup procedure with reactor coolant boron concentration at 2330 ppd . 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 90% withdrawn. The reactor coolant system was then deborated until criticality was achieved. j 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 I 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 rance and intermediate range must be ,

This means that before the source range count rate 1 observed.5 equals 10 cps the intermediate range must be on the scale. If ,

the one decade is not observed, the approach to the intermediate j range cannot be continued until the situation has been corrected. !

l I 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.

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

l

intermediate range is 2.36 decades which is well above the mini-mum of one decade overlap required by Technical Specifications.

3.3 SENSIBLE HEAT DETERMINATION Detennination 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 wu 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.

The point at which there was a definite heatup gate was 6.14 X 10-8 amps. Therefore, during the test 6.14 X 10-0 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 2.06 X 10-8 amps.

After setting the upper zero power physics test current limit, temperature feedback measurements wre performed to verify that the above limit was adequate. This was done by adding +20 pcm at a neutron flux signal of 2.06 X 10-8 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.

3.4 REACTIVITY CALCULATIONS deactimeter 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 samoles of neutron flux. In addition to reactivity and neutron flux, the reactimeter can also record 23 other analog and digital signals from the plant.

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 in the test program.

After steady state conditions with a constant neutron flux were established, a small amount of negative reactivity was inserted 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 reactivity inserted by rod group 7 or 8, from -68.9 to 76.7 pcm.

The reactivities determined from doubling the measurements were then compared with the reactivities calculated by the reactimeter.

The results of the reactimeter verification measurements are sum-marized in Table 3.4-1. The reactivity calculated by the reactimeter in each case was within the acceptance 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 baron 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 3.0-2. These results show that the measured critical boron concentration was within the acceptance criterion of 1359 1 100 ppmB. The results were also within the review criterion of 1359 1 50 ppmB.

3.5 CONTROL R0D 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 detector strings is given in Figure 3.5-1. The number of rods 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 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 the 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 3.0-2 for a modera-tor temperature and pressure of 532*F and 2155 psig, respective-ly. Comparison between measured and predicted rod worth shows groups 5-7 agreed within -0.9 percent.

All acceptance criteria were met.

Integral measured rod worth curves for control rod groups 5, 6 and 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 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 1375 to 1090 ppmB.

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

In summary, the measured differential boron worth at zero power was within -10.8% percent of the predicted worth which is within the acceptance criteria of + 15 percent.

3.7 EJECTED CONTROL R0D WORTH 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 accidental ejection of the most reactive control rod which is fully inserted in the core during normal power operation. The acceptance criteria for the ejected control rod test are that the average measured ejected rod worth be within 20% of the predicted value, that the worst case measured ejected rod worth be less than 1.0% Ak/k.

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. The ejected worth of control rod 6-4 was then measured by the rod swap method. Control rod 6-4 was then measured by the rod swap method. The rod swap test method was to initially position group 5 at 0% withdrawn with the " ejected" control rod 78.7% with-drawn. The change in group 5 position was then converted to reactivity by using a previously generated group 5 rod worth curve to determine the measured ejected rod worth.

Since the measured rod position did not precisely equal the mini-mum rod index, the measured ejected rod worth was adjusted to the Technical Specifications position by using the equation below:

Where: p(ER) = F X p(R) + p(M) EQ. (3.7.1) p(ER) = The measured ejected rod worth at the 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 rods at the minimum rod index minus the worth of rods at 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.

3.8 TEMPERATURE COEFFICIENTS OF REACTIVITY The temperature coefficient of reactivity is defined as the frac-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 + aD EQ.(3.8-1)

Where: aT = Temperature Coefficient of Reactivity t

aM = Moderator Coefficient of Reactivity a0 = Doppler Coefficient of Reactivity The technique used to measure the 532*F and 2155 psig isothermal temperature coefficient at zero power was to first establish steady state conditions by maintaining reactor flux, reactor coolant pressure, turbine header pressure and core average temp-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 'he reactivity value calculated by l the reactimeter and the core .serage temperature displayed on a two channel recorder.

l l

Once steady state conditions were established, a positive heatup rate was started by opening the turbine bypass valves. As the reactivity ch anged about i 5.0 PCM, control rods were moved 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 decreased by about 10'F. After stabili-zing coolant temperature and reactivity, the core average temper-ature was then returned to its original value. The measurement 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 specified in Technical Specifications, the modcrator temperature coefficient shall not be positive at power levgls above 95 percent full power and shall be less than +0.9 X 10-* ak/k/ F at all other power levels. To obtain the moderator coefficient from the measurgd temperature coefficient, a Doppler correction of -0.177 X 10-9 Ak/k/'F must 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 pgig plateau were within the acceptance criteria of 10.40 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 Program.

3.10 EFFLUENT AND EFFLUENT MONITORING No effluent or effluent monitoring testing was performed as these systems have been performing normally since initial startup and no further testing was required.

3.11 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.

ACCEPTANCE CRITERIA DElflATION LIMITS BETt3EEN MEASURED AND FREDICTED 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%

Ejected Rod Worth 1 20%

l l B. Temperature Coefficients 10.4 x 10-4 Ak/k/ F g C. Differential Boron Worth 1 15%

5 m All Reds Out Boron Concentration i 100 ppmB D.

?

i i

f3 9

COMPARISON OF EASURED Af0 PREDICTED RESULTS OBTAlfED DURItG ZER0 P0hD PHYSICS TESTS Physic Parameter Units Measured Predicted Acceptance Requiremente Conparison

1. All Rods Out Critical Boron ppnB 1386 1359 Measured value nust be within OK (27) -

100 pptB of predicted value

2. Sensible Heat anps 6.14 X 10-8 lbne

3. NI Overlap decades 2.36 Greater than 1.0 decade OK(>1.0)

4. Control Rod Group Worth % Ak/k Group 7 -0.985 -0.95 Percent deviation of group worth OK (3.0)

Group 6 -1.053 -1.02 should be less than 15% OK(-3.1) y Group 5 -1.129 -1.17 OK(-3.6)

Er Percent deviation of total worth

T TOTAL -3.167 -3.14 should be less than 10% OK(-0.9)

$ 5. Ejected Rod Worth (6-4) % Ak/k r'o Measured 0.503 0.510 Percent deviation of ejected rod OK worth nust be within 20% of the predicted value.

Measured Worst Case 0.746 Value nust be less than 1.0% Ak/k OK(<1.0%}

6. Temperature Coefficient 10-4 ak/k*F .

At 1379 pptB +0.156 +0.085 Measuredjalue cust be within OK 0.4 X 10 Ak/k F of the pre-At 1090 pptB -0.497 -0.598 dicted value. OK

COMPARISON T KAStRED AND PREDICTED RESULTS OBTAINED DURING ZERO F0WER PHYSICS TESTS Physic Paraneter Units Measured Predicted Acceptance Requirements Conparison

7. Moderator Coefficient 10-4 Ak/kT At 1379 pps 0.333 0.262 Maxinun positive noderator co- OK efficient must be less than -

At 1090 pp2 -0.320 -0.421 0.9 X 10x-4 Ak/k/T OK l

8. Diffemntial Boron Worth % ak/k/pps l At 1090 pp2 -0.01091 -0.00973 Percent deviation must be less OK(-10.8)

than 15%

E m

." 1

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N l

1 2 NOTE: Percent Deviation is defined by the following fonnula:

E

" _ % Deviation = Predicted Value - Measured Value X 100 ,

S Measured Value

SIMiARY OF NUCLEAR INSTRtNENTATION OVERLAP EASUREMENTS DlRING ZERO POWER PHYSICS TEST Source Range Indication Intermediate Range Indication Average SR Average IR Data Indication Indication Overlap Set NI-1 (CPS) NI-2 (CPS) NI-3 (AWS) NI-4 (AWS) (CPS) (AWS) (Decades) 01 9.0 X 103 8.0 X 103 1.0 X 10-11 2.0 X 10-11 8.5 X 103 1.5 X 10-11 2.28 02 1.1 X 105 1.0 X 10 5 1.5 X 10-10 2.0 X 10-10 1.1 X 105 1.8 X 10-10 2.21 l

03 4.1 X 105 4.2 X 105 1.0 X 10-9 1.5 X 10-9 4.2 X 105 1.3 X 10-9 2.49 2.0 x 105 2.0 x 105 4.0 x 10-10 8 x 10-10 2.0 x 105 6.0 x 10-10 2.48 04 Average = 2.36 E

m N

.L NOTE (1): Overlap is obtained between Source and Intermediate Range by using the average indications in the equation below:

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

COMPARIS0N OF REACTIMETER AND DOUBLING TIME (DT) REACTIVITY MEASUREMENTS DT Converted Reactimeter Absolute Case DT Reactivity Reactivity Error No. (Sec) (pcm) (pcm) (%)

1 255.0 -22.6 -23.0 -1.7 2 190.0 -24.5 -24.0 2.1 3 131.0 -49.7 -50.0 -0.6 4 65.5 60.3 +60.0 0.5 5 103.0 -68.9 -69.0 -0.1 C

5 6 47.7 76.7 77.0 0.4 a 1 w

T

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

E (%) = 100 (pR - pDT) / (pDT) l

ll' l Br .

p p d 3

/ e 7 k t 9

/ c 0 k i 0 A d e 0 r

P h

t r

T o S W E

T l 1 a 9 S i d 0 C t e 1 I n r 0 S e u Y r s 0 H e a

- P f e f M R i E D W

O P )

k O nh/ 8 R otk 4 E rrA 0 Z oo BWA s G (

N I

R c U n D o aC) 5 S t m 8 T l np 2 N eop E Dr(

M o E B R

U c S n 5 A eo E gC) 2 M a rnp m 3 2

H eop 1 T vr(

R Ao O B W

c Y d n T eo I rC) 5 0 V u snp m 7 9 I 3 0 T aop 1 1 C er(

A Mo E B R

5 5 N

O 8 7 7 R 3 3 O

B 0 0 L 7 0 A d 1 I w 0 T

N % 6 0 0 E 1 R ,

E n 0 0 F o 5 0 1 F i 1 I t D i 0 0 s 4 0 0 o 1 1 P

0 0 d 3 0 0 o 1 1 R

0 0 2 0 0 1 1 0 0 1 0 0 1 1 w a .h C

ill l lll

DETERMINATION OF THE RELATIVE SYMMETRIC EJECTED R0D WORTH AT ZER0 POWER, BOL, AND 532*F CONDITIONS Rod Position, % wd RCS Boron Ejected Rod Change In Ejected Rod Worth, % ak/k Concentration Position Reactivity 1 2 3 4 5 6 7 8 (ppm) % wd (% ak/k) Measured Worst Case 100 100 100 100 0 0 0 37.0 1081 0 1

100 100 100 100 0 0 0 36.0 1116 100 0.493 0.493

100 100 100 100 26.7 0 e 36.0 11!6 0 0.503 0.746 C

5 m

N

.L l

NOTE: The worst case ejected rod worth is obtained by application of an uncertainty l factor of approximately 1.10.

l k

SUMMARY

OF MEASURED, CALCULATED, AND PREDICTED COEFFICIENTS OF REACTIVITY AT ZERO P0 tier Red Position, % wd Average RCS Boron Reactivity Coefficients,1.0E-04 %Ak/k/ F Temperature Concentration Temperature Coefficient Moderator Coefficient 1 2 3 4 5 6 7 8 (Deg. F) (ppmB) Measured Predicted Calculated Predicted 100 100 100 100 100 100 91 375 532 1379 +0.156 +0.085 +0.333 +0.262 1

100 100 100 100 100 10 0 37.5 532 1090 -0.497 -0.598 -0.320 -0.421 a

u i co

,L l

l NOTE: The calculated moderator coefficient has been determined from the measured temperature cgefficient by subtracting off the isothermal Doppler coefficient of -:).177 10-%kff.

k

INCORE MONITOR AND CONTROL ROD MAP 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 O NI-1 O NI-3 A

O NI-5 CR4-9 CR7-1 CR4-2 O NI-7

5 5 CRI-8 CR5-8 CR5-1 CR1-1 m

CR6-8 CR8-8 CR3-1 CR -

lR6-2 .- D

] 27 .

Sj CR1-7 CR3-12 CR2-8 CR2-1 R3-3 lRI-2 E CR4-8 CR8-7 CR7-8 CR6-1 CR7-2 lR8-2 '

R4-3 p CRS-7 CR2-7 CR3-ll' CR3-2 :R2-2 lR5-2 g CR7-7 CR3-1C CR6-7 CR4-1 CR6-3 'R3-4 R7-3 H E 5 55 I A CR$-6 CR2-6 :R3-8 CR3-5 :R2-3 lRS-3 g CR4-7 CR8-6 :R7-6 CR6-5 CR7-4 'R8-3 '

R4-4 g CRI-6 CR3-9 'R2-5 CR2-4 lR 3-6 lR1-3 g CR6-6 :R8-5 CR3-7 CR8-4 lR6-4 y CRI-5 'R5-5 CRS-4 :RI-4 A

'R4d CR7-5 lR4-5 O NI-8 2 e Nr-6 e Nt-4 9 Ni-2

- Total Core Monitor

- Total Core and Z - Incore Monitor Number Syuusetry Monitor Z - Symmetry Monitor CRX-Y - Control Rod Number Y of Control Rod Group X Figure 3.5-1

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Figure 3.5-2

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Figure 3.5-3

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Figure 3.5-4

1 -

4.0 POWER TESTING 4.1 REACTIVITY COEFFICIENTS AT POWER TEST 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 moderator temperature.

(c) Power Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in power. q 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 Doppler coeffigient of reactivity shall be more negative than -0.55 X 10" Ak/k/%FP.

i 4.1.2 Test Method Reactivity coefficient measurements were made during the power escalation test program a,t 97.5% 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 as . 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.

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.

oT = aH + aD EQ.(4.1-1)

Where: aT = Temperature Coefficient of Reactivity aM = Moderator Coefficient of Reactivity aD = 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 coefficients have been deter-mined. Technical Specifications, Section 3.1.1.3, 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 moderator coefficients were theoretically pre-dicted as shown in Table 4.1-1 using the distributed moderator and fuel temperatures instead of the isothermal values which were u3ed for zero power physics pradictions. For these predictions, I 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.

i 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 relates the change in core reactiv-ity to a corresponding change in power. Theoretical predictions of the power Doppler coefficient were made using the PDQ code with thermal feedback. The predicted power Doppler coefficients l using this code are presented in Table 4.1-1.

1

" w T w - - - - - - - - - - -

e

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 state data at the reduced power level, reactor power was returned to the initial power.

The calculaticr. 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.

4.1.4 Differential Rod Worth at Power The method by which the differential rod worth was detennined 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 elapsed time 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 motion 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 correction factor accounts for the time delay in-volved in fuel temperature change during the measurement.

4.1.5 Evaluation of Test Results The results of the measured temperature ' :alculated moderator coefficients at power are shown in Tab' 4.1-1 which also shows the predicted temperature and moderator uefficient results.

The calculated moderator coefficient is negative and therefore meets the acceptance criteria of being non-positive.

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

The acceptance criterion for the measured power Doppler coeffic-ieng 10-is that the coefficient must be more negative than -0.55 X Table 4.1-1 shows that tb measured coefficient ak/k/%FP.

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 dur:ag 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 symmetry conditions. A detailed description of the incore monitoring 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 the corrected signals is used to develop core power distributions which provide power peaking information necessary to determine core performance 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 Figure 4.3-1.

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

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

(1) Normal Operating Core Power Distributions (2) Worst Case Minimum DNBR and Maximum I"R Calculations (3) Quadrant 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 Dis-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, at 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 extrapolated 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.

4.3.2 Test Method Computer printouts of the core power distribution and thermal hydraulics 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 equilibriua xenon was re-quired, the APSR's were maintained at a constant position and the axial incore imbalance was maintained within +/-2% 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.

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. The 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.

f l

1 l

1 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-unifann axial heat flux distributiens. 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 4.3.3.3 Quandrant Power Tilt and Axial Power Imbalance Quandrant Power Tilt Quadrant power tilt limits have been established in the Technical Specifications. These limits, when used in conjunction with the control rod position limits, assure that the design peak heat rate criterion is not exceeded during normal 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 -1 x 100 Tilt Average Power of All Quadrants During the startup testing program, maximum quadrant power tilt was determined using the corrected signals from the 16 symmetric incore monitoring assembli_ ligures 4.3-2 through 4.3-7 in-clude the maximum quadrar.t ,awer 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 Results from the Standard Core Power Distributions taken at 40% FP during the performance of the Power Imbalance Det& tor Correlation Test, show that the imbalance trip envelone (Figure 4.3-1) of the reactor protective system is sufficient (o 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) exist when a negative 5.0% to 10.0% incore axial offset is present.

The core imbalances measured in conjunction 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 l the initial startup testing program.

l 4.5 PSEUD 0 R0D EJECTION TEST This test was not performed 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 nonnal 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-tions were made which would invalidate the results of the tests performed during initial startup.

4.12 POWER IMBALANCE DETECTOR 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

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.

This section of the report represents the results 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 system 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.

l (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

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 verified.

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 oy 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 acc2ptable areas of the curve during the performance 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 the 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 shown in Figure 4.12-3. The most limiting value 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 determined 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.

(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 intermediate and power range nuclear instrumentation.

Four acceptance criteria are specified for the Nuclear Instrumen-tation Calibration at Power Test as 18sted below:

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

(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 poder 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 R0D TEST e

The dropped rod test was not performed because sufficient ther'31 margin exists as indicated by B&W-2 correlation analyses.

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

SUMMARY

OF TEMPERATURE, MODERATOR, AND D0PPLER COEFFICIENTS OF REACTIVITY COEFFICIENT E ASURED CALCULATED ACCEPTANCE REQ.

Temgrature -0.8638 -1.3334 --

J (10- ak/k/ F)

-0.7088 -1.1838 Less than zero Mode (10- ak/k/*F) {ator Power Doppler -1.0244 -1.0780-More negative than (10-4 Ak/k/%FP) -0.55 x 104 Ak/k/%FP Table 4.1-1 i

DETERMIMTION OF RMS FOR CALCU!ATED VS. KASURED RPD'S 0 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 1.136 H-08 1.110 27 D-10 1.075 L-12 1.010 2 H-09 1.254 H-09 1.220 28 C-10 1.259 L-13 1.275 3 G-09 1.294 K-09 1.252 29 C-09 1.106 K-13 1.095 4 F-08 1.217 H-10 1.149 30 B-08 0.543 H-14 0.557 5 E-09 1.155 K-11 1.194 31 B-07 0.857 K-14 0.960 6 F-07 1.271 K-10 1.252 32 C-06 1.230 L-13 1.275 7 E-07 1.158 K-11 1.194 33 0-05 1.283 H-12 1.245 8 G-06 1.290 K-10 1.252 34 E-04 1.275 M-12 1.245 9 G-05 1.157 K-11 1.194 35 F-03 1.224 L-13 1.275 10 H-05 1.351 H-11 1.318 36 G-02 0.920 K-14 0.960 11 K-05 1.174 K-11 1.1 94 37 H-01 0.440 H-15 0.438

-4 12 L-06 0.819 L-10 0.779 38 L-02 0.983 L-14 0.993 !

Er 13 M-07 1.137 K-11 1.1 94 39 L-03 1.224 L-13 1.275 T 14 N-08 1.158 H-12 1.147 40 M-03 1.115 M-13 1.118 4 15 N-09 1.263 K-12 1.277 41 N-04 1.083 N-12 1.036 i la 16 M-09 1.148 K-11 1.194 42 0-05 1.019 M-13 1.118 l J. 17 M-10 1.242 L-11 1.263 43 0-06 1.235 L-13 1.275 18 L-11 1.259 L-11 1.263 44 P-06 0.963 L-14 0.993 19 K-11 1.136 K-11 1.194 45 R-07 0.492 K-15 0.492 20 K-12 1.250 K-12 1.277 46 R-10 0.44 0 L-15 0.431 21 H-13 1.188 H-13 1.207 47 0-10 1.219 L-13 1.275 22 G-13 1.183 K-13 1.095 48 0-12 1.040 N-13 1.074 1 23 F-13 1.249 L-13 1.275 49 M-14 0.894 M-14 0.870 l 24 F-12 1.075 L-12 1.010 50 L-13 1.235 L-13 1.275 25 G-11 1.141 K-11 1.1 94 51 D-14 0.4 94 N-14 0.510 26 E-11 1.188 M-11 1.173 52 C-13 0.616 0-13 0.601 1

RMS = f;(Measured-Calculated)2 l v 52 l RMS = 0.0420

\

J DETERMINATION OF PRS FOR CALCULATED VS. MEASURED 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 1.114 H-08 1.103 27 D-10 1.034 L-12 1.010 2 H-09 1.229 H-09 1.209 28 C-10 1.214 L-13 1.267 3 G-09 1.259 K-09 1.239 29 C-09 1.074 K-13 1.0%

4 F-08 1.185 H-10 1.144 30 B-08 0.577 H-14 0.569 5 E-09 1.183 K-11 1.187 31 B-07 0.972 K-14 0.%9 6 F-07 1.240 K-10 1.241 32 C-06 1.250 L-13 1.267 7 E-07 1.189 K-11 1.187 33 D-05 1.226 M-12 1.237 8 G-06 1.256 K-10 1.241 34 E-04 1.224 M-12 1.237 9 G-05 1.189 K-11 1.187 35 F-03 1.247 L-13 1.267 10 H-05 1.306 H-11 1.305 36 G-02 1.037 K-14 0.%9 l

l 11 K-05 1.204 K-11 1.187 37 H-01 0.454 H-15 0.4M i --. 12 L-06 0.814 L-10 0.783 38 L-02 1.012 L-14 1.000 Er 13 M-07 1.172 K-11 1.187 39 L-03 1.248 L-13 1.267 l

E' 14 N-08 1.127 H-12 1.140 40 M-03 1.137 M-13 1.117 a 15 N-09 1.227 K-12 1.256 41 N-04 1.046 N-12 1.040 L 16 M-09 1.181 K-11 1.187 42 0-05 1.038 M-13 1.117

/o 17 M-10 1.206 L-11 1.251 43 0-06 1.259 L-13 1.267 l 18 L-11 1.227 L-11 1.251 44 P-06 0.994 L-14 1.000 l 19 K-11 1.173 K-11 1.187 45 R-07 0.526 K-15 0.507 l 20 K-12 1.214 K-12 1.265 46 R-10 0.460 L-15 0.445 l 21 H-13 1.155 H-13 1.200 47 0-10 1.241 L-13 1.267 1 22 G-13 1.062 K-13 1.094 48 0-12 1.069 N-13 1.076 l 0.878 l 23 F-13 1.271 L-13 1.267 49 M-14 0.918 M-14 24 F-12 1.035 L-12 1.010 50 L-13 1.258 L-13 1.267 i 25 G-11 1.176 K-11 1.187 51 D-14 0.516 N-14 0.521' 26 E-11 1.145 M-11 1.166 52 C-13 0.600 0-13 0.612 RMS=[1(Measured-Calculated)2 V 52 RMS = 0.0259

1 DETERMINATION OF RMS FOR CALCULATED VS. KASURED RPD'S 9 97.5% 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 1.112 H-08 1.106 27 D-10 1.038 L-12 1.007 2 H-09 1.225 H-09 1.207 28 C-10 1.222 L-13 1.261 3 G-09 1.252 K-09 1.234 29 C-09 1.079 K-13 1.092 4 F-08 1.177 H-10 1.142 30 B-08 0.578 H-14 0.575 5 E-09 1.178 K-11 1.180 31 B-07 0.948 K-14 0.973 6 F-07 1.232 K-10 1.233 32 C-06 1.256 L-13 1.261 7 E-07 1.185 K-11 1.180 33 0-05 1.226 M-12 1.231 8 G-06 1.243 K-10 1.233 34 E-04 1.227 M-12 1.231 9 G-05 1.184 K-11 1.180 35 F-03 1.250 L-13 1.261 10 H-05 1.302 H-11 1.293 36 G-02 1.014 K-14 0.973 l

11 K-05 1.197 K-11 1.180 37 H-01 0.475 H-15 0.467

-4 12 L-06 0.806 L-10 0.783 38 L-02 1.020 L-14 1.006 l Er 13 M-07 1.169 K-11 1.180 39 L-03 1.248 L-13 1.261 i 14 N-08 1.133 H-12 1.135 40 M-03 1.140 M-13 1.117 l

e. 15 N-09 1.231 K-12 1.256 41 N-04 1.049 N-12 1.042 l L2 16 M-09 1.178 K-11 1.180 42 0-05 1.040 M-13 1.117 do 17 M-10 1.204 L-11 1.241 43 0-06 1.261 L-13 1.261 18 L-11 1.225 L-11 1.241 44 P-06 1.003 L-14 1.006 19 K-11 1.168 K-11 1.180 45 R-07 0.526 K-15 0.520 1 20 K-12 1.220 K-12 1.256 46 R-10 0.470 L-15 0.4 58 21 H-13 1.153 H-13 1.191 47 0-10 1.245 L-13 1.261 22 G-13 1.068 K-13 1.0 92 48 0-12 1.071 N-13 1.075 23 F-13 1.272 L-13 1.261 49 M-14 0.929 M-14 0.889 24 F-12 1.040 L-12 1.00/ 50 L-13 1.260 L-13 1.261 25 G-11 1.170 K-11 1.180 51 D-14 0.516 N-14 0.531 26 E-11 1.143 M-11 1.161 52 C-13 0.608 0-13 0.620 RMS = (Measured-Calculated)2

52 RMS = 0.0219

t MAXIMUM LINEAR HEAT RATE PER INCORE DETECTOR LEVEL VERSUS POWER LEVEL Maximum Linear Heat Rate (KW/ft.)

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

8 3.04 4.77 5.74 14.8 7 4.31 9.25 10.30 15.3 6 4.32 9.33 10.46 16.3 5 4.74 7.92 10.40 17.1 4 4.26 7.97 10.64 16.2 3 4.21 8.36 10.87 14.9 2 3.72 7.68 10.33 13.4 1 3.21 7.42 9.33 12.7

=

\

Table 4.3-4

SUPHARY OF TEST DATA POWER INALANCE DETECTOR CORRELATION TEST 40% FP Power Rod Position, %wd Incore Offset, % Out-of-Core Offset, % Worst Case Worst Case Level Maxinun LHR Minimum Df0R

(%FP) 5-7 8 Full Backup NI-5 NI-6 NI-7 NI-8 (kW/ft) -

l 40.7 194.4 31.1 -1.03 -1.42 -1.27 -1.64 -1.18 l

41.2 191.9 36.7 -9.44 -11.53 -11.36 -11.78 -10.85 41.8 190.0 40.9 -15.15 -19.11 -18.80 -19.02 -17.88 41.1 191.2 22.9 8.43 10.17 10.47 10.56 10.71 g Rod Position, %wd er I 1-5 6 7 8 a

k 40.1 40.1 100 100 88.3 88.4 11.4 12.2 27.3 30.9 6.19 0.55 5.93 1.15 5.03 4.95 9.431 10.083 4,

40.1 100 88.5 11.7 35.0 -6.32 -3.90 5.55 10.188 40.1 100 88.3 11.0 36.9 -10.59 -7.27 5.70 9.730 40.1 100 85.0 8.6 42.9 -18.19 -13.89 5.83 8.670 40.1 100 76.1 1.1 50.9 -22.62 -21.94 6.28 7.437 l l

l l

e ,

Summary of Least Squares Linear P.egression Analysis for Each Power Range Channel and Backup Recorders During the Power Imbalance Correlation Test Detector Gain Data Intercept Slope System Factor Sets Calculated Allowable Calculated Allowable Power Imbalance Test at 40% Full Power NI-5 4.5 4 -0.163 +3.5 1.235 >1.15 NI-6 4.5 4 -0.070 T3.5 1.235 >1.15 NI-7 4.5 4 -0.099 T3.5 1.250 >1.15 NI-8 4.5 4 -0.385 T3.5 1.206 >1.15 Backup --- 6 -1.923 T5.0 0.740 1.00+0.20 1

1

. l E

N 1

e

Conparison of Incore Monitoring Assablies Flux Shapes at 40% FP Using the Indicated Grouping Given in Incore Detector Test Incore Fuel Detector Current to Average Detector Current per Assenbly Group Detector Assenbly Nunber Number Location Level 1 Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (BG) 01 05 E-09 0.733 1.001 1.200 1.257 1.239 1.095 0.475 10 07 E-07 0.756 1.025 1.181 1.206 1.205 1.089 0.538 42 i 09 G-05 0.745 1.021 1.096 1.112 1.198 1.142 0.685 14 11 K-05 0.850 1.051 1.162 1.275 1.193 0.999 0.469 25 13 M-07 0.732 1.072 1.093 1.138 1.246 1.133 0.586 47 16 M-09 0.742 1.073 1.104 1.102 1.255 1.166 0.558 45 19 K-11 0.803 1.065 1.155 1.177 1.204 1.015 0.580 28 25 G-11 0.750 1.131 1.116 1.045 1.228 1.146 0.585 48 02 23 F-13 0.782 1.024 1.091 1.077 1.199 1.153 0.674 23 E? 28 C-10 0.793 1.014 1.246 1.199 1.163 1.058 0.526 47 55 32 C-06 0.748 0.977 1.145 1.226 1.195 1.087 0.622 19 35 F-03 0.737 0.998 1.107 1.170 1.185 1.107 0.696 14 f' 39 L-03 0.751 1.054 1.076 1.070 1.230 1.166 0.653 22 03 43 0-06 0.762 1.082 1.106 1.148 1.229 1.100 0.574 33 d- 47 0-10 0.734 1.062 1.309 1.225 1.117 1.099 0.453 38 50 L-13 0.825 1.091 1.111 1.177 1.151 1.094 0.552 16 03 06 F-07 0.732 1.023 1.104 1.112 1.190 1.144 0.695 19 08 G-06 0.776 1.024 1.075 1.139 1.197 1.117 0.671 18 15 N-09 0.784 1.091- 1.139 1.221 1.187 1.102 0.477 14 17 M-10 0.736 1.065 1.110 1.153 1.243 1.207 0.487 14 10 L-11 0.808 1.053 1.163 1.156 1.203 1.021 0.596 35 20 K-12 0.750 1.084 1.131 1.101 1.235 1.130 0.569 44 33 D-05 0.783 1.026 1.208 1.201 1.200 1.060 0.522 44 34 E-04 0.714 0.981 1.142 1.200 1.203 1.091 0.669 13 04 24 F-12 0.783 1.038 1.103 1.094 1.185 1.133 0.664 19 27 D-10 0.772 0.968 1.160 1.196 1.203 1.104 0.596 15

+

c. _

Conparison of Incore Monitoring Assecblies Flux Shapes at 40% FP Using the Indicated Grouping Given in Incore Detector Test Incore Fuel Detector Current to Average Detector Current per Assecbly Group Detector Assenbly Nunter Nunber Location Level 1 Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (BG) 05 38 L-02 0.744 1.045 1.061 1.083 1.228 1.179 0.660 17 44 P-06 0.753 1.093 1.105 1.120 1.218 1.106 0.606 43 06 01 H-08 0.814 1.142 0.917 0.839 1.293 1.246 0.749 17 04 F-08 0.775 1.019 1.213 1.231 1.185 1.047 0.529 38 10 H-05 0.769 1.035 1.085 1.086 1.187 1.146 0.693 14 14 N-08 0.739 1.109 1.146 1.181 1.219 1.089 0.517 36 l U 26 E-11 0.779 1.015 1.197 1.207 1.190 1.072 0.540 37 41 N-04 0.726 1.077 1.085 1.136 1.257 1.129 0.590 43

[

P 07 02 H-09 0.717 1.039 1.090 1.098 1.231 1.152 0.673 16 y- 03 G-09 0.765 1.049 1.075 1.116 1.192 1.140 0.664 20 g 08 21 H-13 0.746 0.998 1.093 1.115 1.170 1.151 0.727 22 g 48 0-12 0.705 1.083 1.101 1.227 1.224 1.093 0.567 14 e

~

d 09 22 G-13 0.820 1.141 0.920 0.819 1.278 1.247 0.776 16 29 C-09 0.733 1.001 1.200 1.257 1.239 1.095 0.475 22 10 31 B-07 0.723 0.985 1.139 1.243 1.214 1.070 0.626 12 36 G-02 0.799 1.047 1.191 1.192 1.182 1.056 0.533 44 49 M-14 0.787 1.100 1.114 1.146 1.219 1.118 0.516 34 11 37 H-01 0.807 1.058 1.217 1.157 1.166 1.061 0.534 47 45 R-07 0.711 1.075 1.173 1.254 1.218 1.120 0.449 21 46 R-10 0.675 1.019 1.158 1.177 1.264 1.121 0.586 9 51 D-14 0.784 1.014 1.195 1.217 1.183 1.073 0.534 45 52 C-13 0.760 0.983 1.139 1.214 1.210 1.086 0.606 20

Cormarison of Incore Monitoring Asssblies Flux Shapes at 407, FP Using the Indicated Grouping Given in Incore Detector Test Incore Fuel Detector Current to Average Detector Current per Assertly Group Detector Assecbly Nunber Nunter Location Level 1 Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (BG) 12 40 M-03 0.721 1.061 1.184 1.148 1.242 1.072 0.573 37 42 0-05 0.761 0.947 1.121 1.214 1.265 1.096 0.596 16 -

13 12 L-06 0.744 1.057 1.100 1.120 1.248 1.144 0.586 46 30 B-08 0.727 0.974 1.152 1.216 1.243 1.070 0.618 24 N

5 (D

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(-7,107) - - (+2,N)7)

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(-32,90)t - - go >(+25pO)

PtW CPERATION

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(-7,79.92) (+2,79.92) 70

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50 -40 -30 -20 -10 O 10 20 30 40 50 60 Figure 4.3-1

Comparison of Predicted .and Measured Radial Peaking Core Power Distreibution Results at Steady State, 3-D Equilibrium Xenon, 40% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 40.0 %FP Gps 1-5 100 % wd Boron Concentration --

ppmB I

Gp 6 E 3 % wd Core Burnup 2.0 EFPD Gp 7 12.7 % wd Axial Imbalance 0.19 % FP Gp 8 28.8 % wd Max Quadrant Tilt --

Measured Conditions -

Control Rod Group Positions Core Power Level 39.9 %FP Gps 1-5 100 % wd Boron Concentration 1038 ppm 8 Gp 6 87 % wd Core Burnup 0.4 EFPD Gp 7 10 % wd Axial Imbalance 0.69 % FP Gp 8 31 % wd Max Quadrant Tilt 0.74 %

H K L M N O P R 1.14 8 1.11

-2.6 1.25 1.29 9 1.22 1.25

-2.4 -3.1 1.22 1.28 0.82 10 1.15 1.25 0.78

-5.7 -2.3 -4.9 1.35 1.15 1.25 1.19 %

11 1.32 1.19 1.26 1.17

-2.2 3.5 0.8 -1.7 1.16 1.26 1.08 1.28 1.08 12 1.15 1.28 1.01 1.25 1.04

-0.9 1.6 -6.5 -2.3 -3.7 1.19 1.10 1.23 1.07 1.04 0.62 13 1.21 1.10 1.28 1.12 1.07 0.60 1.7 0.0 4.1 4.7 -2.3 -3.2 0.54 0.89 0.97 0.89 0.49 14 0.56 0.96 0.99 0.87 0.51 3.7 7.9 2.1 -2.2 4.1 0.44 0.49 0.44 15 0.44 0.49 0.43 0.0 0.0 -2.3 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measured Figure 4.3-2

Compar.ison of Predicted and Measured Total Peaking Core Power Distribution Results.at Steady State, 3-D Equilibrium Xenon, 40% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 40.0 %FP Gps 1-5 100 % wd Boron Concentration -- ppmB Gp 6 371 % wd Core Burnup 2.0 EFPD Gp 7 12.3 % wd Axial Imbalance 0.19 % FP Gp 8 28.8 % wd Max Quadrant Tilt --

Measured Conditions Control Rod Group Positions Core Power Level 39.9 %FP Gps 1-5 100 % wd Boron Concentration 1038 ppmB Gp 6 - 37 % wd Core Burnup 0.4 EFPD Gp 7 10 % wd Axial Imbalance 0.69 % FP Gp 8 31 % wd Max Quadrant Tilt 0.74 %

H K L M N O P R 1.40 8 1.32

-5.7 L

1.55 1.58 9 1.46 1.49

-5.8 -5.7 1.51 1.54 0.91 10 1.42 1.49 0.87

-6.0 -3.2 -4.4 1.63 1.43 1.48 1.39 11 1.58 1.45 1.50 1.41

-3.1 1.4 1.4 1.4 1.35 1.49 1.35 1.55 1.32 12 1.35 1.53 1.32 1.54 1.30 0.0 2.7 -2.2 -0.6 -1.5 1.42 1.28 1.50 1.30 1.32 0.77 13 1.42 1.31 1.57 1.37 1.33 0.74 0.0 2.3 4.7 5.4 0.8 -3.9 0.61 1.06 1.39 1.09 0.61 14 0.61 1.14 1.21 1.07 0.63 0.0 7.5 1.7 -1.8 3.3 0.51 0.59 0.55 15 0.51 0.59 0.52 0.0 0.0 -5.5 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measured Figure 4.3-3

Comparison of Predicted ,and heasured Radial Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium Xenon, 75% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 75.0 %FP Gps 1-5 100 % wd Boron Concentration --

ppmB Gp 6 87.3 % wd Core Burnup 3.0 EFPD Gp 7 12.3 % wd Axial Imbalance -1.48 % FP Gp 8 19.1 % wd Max Quadrant Tilt --

Measured Conditions Control Rod Group Positions Core Power Level 75.56 %FP Gps 1-5 100 % wd Boron Concentration 916 ppmB Gp 6 H375 % wd Core Burnup 2.6 EFPD Gp 7 14.0 % wd Axial Imbalance -1.86 % FP Gp 8 2T!7 % wd Max Quadrant Tilt -0.40 %

H K L M N O P R 1.11 8 1.10

-0.9 1.23 1.26 9 1.21 1.24

-1.6 -1.6 1.19 1.25 0.81 10 1.14 1.24 0.78

-4.2 -0.8 -3.7 1.31 1.18 1.22 1.15 11 1.30 1.19 1.25 1.17

-0.8 0.8 2.4 1.7 1.13 1.22 1.03 1.23 1.05 12 1.14 1.26 1.01 1.24 1.04 0.9 3.3 -1.9 -0.8 -1.0 1.16 1.07 1.25 1.09 1.07 0.60 13 1.20 1.09 1.27 1.12 1.08 0.61 3.4 1.9 1.6 2.8 0.9 1.7 0.58 1.00 1.00 0.92 0.52 14 0.57 0.97 1.00 0.88 0.51

-1.7 -3.0 0.0 -4.3 -1.9 0.47 0.53 0.46 15 0.45 0.51 0.44

-4.2 -3.8 -4.3 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measured Figure 4.3-4

Comparison of Predicted and Measured Total Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium 'd. ion, 75% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 75.0 %FP Gps 1-5 100 % wd Boron Concentration -- ppmB Gp 6 87.3 % wd Core Burnup 3.0 EFPD Gp 7 12.3 % wd Axial Imbalance -1.48 % FP Gp 8 19.1 % wd Max Quadrant Tilt --

Measured Condi,tions Control Rod Group Positions Core Power Level 75.56 %FP Gps 1-5 100 % wd Boron Concentration 916 ppmB Gp 6 86.5 % wd Core Burnup 2.6 EFPD Gp 7 14.0 % wd Axial Imbalance -1.86 % FP Gp 8 21.7 % wd Max Quadrant Tilt -0.40 %

h K L M N 0 P R 1.34 8 1.36 1.5 1.46 1.51 9 1.50 1.54 2.7 2.0 1.45 1.49 1.06 10 1.48 1.56 0.92 2.1 4.7 -13.2 1.59 1.48 1.44 1.35 11 1.68 1.57 1.57 1.46 5.7 6.1 9.0 8.1 1.31 1.45 1.31 1.50 1.27 12 1.41 1.59 1.38 1.56 1.35 7.6 9.7 5.3 4.0 6.3 1.34 1.23 1.52 1.32 1.31 0.74 13 1.46 1.34 1.58 1.40 1.37 0.78 9.0 8.9 3.9 6.1 4.6 5.4 0.75 1.37 1.20 1.12 0.63 14 0.64 1.18 1.24 1.10 0.66

-14.7 -13.9 3.3 -1.8 4.8 0.54 0.62 0.56 15 0.54 0.62 0.55 0.0 0.0 -1.8 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measured Figure 4.3-5

Comparison of Predicted.and Measured Radial Peaking Core Power Distribu. tion Results at Steady State, 3-D Equilibrium Xenon, 97.5% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 100 %FP Gps 1-5 100 % wd Boron Concentration --

ppmB Gp 6 E7 1f % wd Core Burnup 4.0 EFPD Gp 7 12.3 % wd Axial Imbalance -5.22 % FP Gp 8 19.1 % wd Max Quadrant Tilt --

Measured Conditions Control Rod Group Positions Core Power Level 97.5 %FP Gps 1-5 100 % wd Boron Concentration --

ppmB Gp 6 85 % wd Core Burnup 4.02 EFPD Gp 7 11 % wd Axial Imbalance -5.18 % FP Gp 8 21 % wd Max Quadrant Tilt -0.32 %

H K L M N O P R 1.11 8 1.106 0.36 1.23 1.25 9 1.207 1.234 1.9 1.3 1.18 1.24 0.81 10 1.142 1.233 0.783 3.3 0.57 3.4 1.30 1.18 1.21 1.14 11 1.293 1.18 1.241 1.161

.54 0.0 -2.5 - 1. 8 g 1.13 1.23 1. 04 1.23 1.05 12 1.135 1.256 1.007 1.231 1.042

-0.44 -2.1 3.3 .08 0.77 1.16 1.97 1.25 1.09 1.07 0.61 13 1.191 1.092 1.261 1.117 1.-85 0.620

-2.6 -2.0 -0.87 -2.4 -0.47 -1.6 0.58 0.99 1.01 0.93 0.53 '

14 0.575 0.973 1.006 0.889 0.531 <

0.87 1.7 0.40 4.6 -2.1 0.48 0.53 0.47 15 0.467 0.520 0.458 2.8 1.9 2.6 X.XXX Predicted Results X.XXX Measured Results XX.X Deviation (Predicted - Measured) x 100%

Measured Figure 4.3-6 m_.__ _ _ _ _ _ _ _ _ _ _ _ - - _ - _

Comparison of Predicted and Measured Total Peaking Core Power Distribution Re'sults.at Steady State, 3-D Equilibrium Xenon, 97.5% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 100 %FP Gps 1-5 100 % wd Boron Concentration -- ppmB Gp 6 ~B771 % wd Core Burnup 4.0 EFPD Gp 7 12.3 % wd Axial Imbalance -5.22 % FP Gp 8 19.1 % wd Max Quadrant Tilt --

Measured Conditions Control Rod Group Positions Core Power Level 97.5 %FP Gps 1-5 100 % wd Boron Concentration -- ppm 8 Gp 6 85 % wd Core Burnup 4.02 EFPD Gp 7 11 % wd Axial Imbalance -5.18 % FP Gp 8 21 % wd Max Quadrant Tilt -0.32 %

H K L M N O P R 1.38 8 1.344 2.7 1.49 1.55 9 1.482 1.513 0.54 2.4 1.48 1.52 1.04 10 1.450 1.526 0.959 2.1 -0.39 8.4 1.61 1.51 1.43 1.34 11 1.632 1.533 1.522 1.422

-1.3 -1.5 -6.0 -5.8 1.34 1.45 1.31 1.48 1.27 12 1.377 1.546 1.342 1.516 1.328

-2.7 -6.2 -2.4 -2.4 -4.4 1.35 1.24 1.51 1.31 1.35 0.74 13 1.426 1.319 1.528 1.373 1.343 0.772

-5.3 -6.0 -1.8 -4.6 0.52 -4.1 0.75 1.21 1.20 1.15 0.64 14 0.686 1.165 1.223 1.099 0.658 9.3 3.9 -1.9 4.6 -2.7 0.56 0.64 0.57 15 0.546 0.622 0.554 2.6 2.9 2.9

- X.XXX Predicted Results 9

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

Measured Figure 4.3-7

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ML20041F540

Contents

Text

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

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

SUMMARY

Navigation menu

ML20041F540

Text

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

. _

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

SUMMARY

Navigation menu

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