3F0588-02, Cycle 7 Startup Rept

Cycle 7 Startup Rept

ML20153E914
Person / Time
Site:	Crystal River
Issue date:	04/30/1988
From:	Wilgus W FLORIDA POWER CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
3F0588-02, 3F588-2, NUDOCS 8805100167
Download: ML20153E914 (69)

Text

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DOCKET NO. 50 ~ 302 l

l LICENSE NO. OPR - 72 Crystal River Unit 3 CYCLE 7 STARTUP REPORT 4T-40 i APRll1988

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Table of Contents Page TABLE OF C0NTENTS........................................................... i LIST OF TABLES.............................................................. iii LIST OF FIGURES............................................................. iv

1.0 INTRODUCTION

........................................................... 1 2.0 PRECRITICAL TESTING.................................................... 4 2.1 Calibration and Neutron Response of Source Range Monitoring System............................................................ 4 2.2 Reactor Coolant Flow Test......................................... 5 2.3 Control Red Drop Time Tests....................................... 6 2.4 Chemical and Radiochemical Tests.................................. 7 2.5 Pressurizer Effectiveness Test.................................... 7 2.6 In-Se'rvice Loose Parts and Vibration Monitoring System Tests...... 7 3.0 ZERO POWER PHYSICS TESTING............................................. 12 3.1 Initial Criticality............................................... 12 3.2 Nuclear Instrumentation 0verlap................................... 12 3.3 Sensible Heat Determination....................................... 13 3.4 Reactimeter Check................................................. 14 3.5 All Rods Out Critical Boron....................................... 15 3.6 Moderator and Temperature Coefficients Measurements............... 15 3.7 Control Rod Group Worths.......................................... 18 3.8 Differential Boron Worth Determination............................ 19 3.9 Ej ected Control Rod Worth Measurement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.10 Biological Shield Survey.......................................... 20 3.11 Ef fluent and Ef fl uent Moni toring. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.12 Chemical and Radiochemical Tests.................................. 20 4.0 FULL POWER ESCALATION TESTING.......................................... 35 4.1 Turbine / Reactor Trip Test......................................... 35 4.2 Integral Control System Test...................................... 35 4.3 Unit Loss of Electrical Load...................................... 35 4.4 Unit Load Transient Test.......................................... 35 4.5 Reactivity Coefficients at Power Test............................. 35 4.6 Unit Heat Balance................................................. 38 4.7 Core Power Distribution Test...................................... 38 4.8 Biological Shield Survey.......................................... 42 4.9 P s e ud o Ro d Ej e cti o n Te s t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 4.10 Shutdown From Outside the Control Room............................ 43 4.11 Loss of Offsite Power............................................. 43 4.12 Power Imbalance Detector Correlation Test......................... 43 4.13 Nuclear Ir.strumentation Calibration at Power Test................. 46 4.14 Emergency Feedwater Flow Test..................................... 47 i

Table of Contents (Continued)-

Page 4.15 Turbine / Generator Operation....................................... 47 4.16 Dropped Control Rod Test.......................................... 47 4.17 Incore Detector Test.............................................. 47 4.18 Reactor. Coolant System Hot Leakage Test........................... 48 4.19 Pipe and Component Hanger Hot Inspection at Power................. 48 4.20 Chemi cal and Radiochemi cal Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 4.21 Effluent and Effluent Montitoring................................. 48 i ,

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List of Tables Page Table Summary of Precritical Testing Results............................... 8 2.0-1 2.3-1 Control Rod Drive Orop Time Tests Results............................ 9 2.3-2 Group Average Control Rod Drop Time.................................. 11 3.0-1 Acceptance Criteria Deviation Limits Between Measured and Predicted Va1ues............................................................... 21 3.0-2 Summary of Zero Power Physics Testing Results........................ 22 3.2-1 Nuclear Instrumentation Overl ap Test Resul ts. . . . . . . . . . . . . . . . . . . . . . . . . 24 3.4-1 Reactimeter and Doubling Time Reactivity Comparison.................. 25 3.6-1 Comparison of Measured and Predicted Reactivity Coefficients......... 26 3.6-2 Extrapolated Hot Full Power Moderator Coefficient of Reactivity...... 27 3.8-1 Differential Boron Reactivity Worth Test Results..................... 28 4.5-1 Summary of Temperature, Moderator, and Doppler Coefficients of Reactivity.......................................................... 49 4.7-1 Maximum Linear Heat Rate by Incore Detector Level.................... 50 4.12-1 Summary of Power Imbalance Detector Correlation Test, 75% FP. . . . . . . . . 51 4.12-2 Summary of Least Squares Linear' Regression Analysis of Power Range Channels and Backup Recorder During Power Imbalance Correlation Test, 75% FP............................................................... 52 4.12-3 Worst Case Minimum DNBR and Maximum Linear Heat Rate vs. Full Incore Offset, 75% FP....................................................... 53 4.17-1 Comparison of Incore Monitored Assemblies' Flux Shapes, 75% FP....... 54 l

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list of Figures j

Figure Page {

1.0-1 Core loading Diagram, Cycle 7........................................ 2 1.0-2 Fuel Assembly and Control Assembly Component Identifications......... 3 3.2-1 Nuclear Instrumentation Detector Locations........................... 29 3.2-2 Nuclear Instrumentation Detector Locations........................... 30 ,

3.7-1 Incore Monitor and Control Rcd Map................................... 31 3.7-2 Control Red Group 5 Integral Worth, BOC 7 at HZP. . . . . . . . . . . . . . . . . . . . . 32 3.7-3 Control Rod Group 6 Integral Worth, BOC 7 at HZP..................... 33 3.7-4 Control Rod Group 7 Integral Worth, BOC 7 at HZP..................... 34 4.7-1 Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance............................................................ 56 4.7-2 Comparison of Measured and Predicted Radial Power Peaking Factnrs with 3D Equilibrium Xenon at 75% Full Power............................... 57-4.7-3 Comparison of Measured and Predicted Maximum Total Power Peaking

• Factors with 30 Equilibrium Xenon at 75% Full Power................... 58 i

4.7-4 Comparison of Measured and Predicted Radial Power Peaking Factors with i 30 Equilibrium Xenon at 100% Full Power. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 4.7-5 Comparison of Measured and Predicted Maximum Total Power Peaking i Factors with 3D Equilibrium Xenon at 100% Full Power................. 60  ;

4.7-6 Hot Channel Minimum DNBR vs. Core Power Level........................ 61 4.12-1 Average Out-of-Core Offset vs. Full Incore Offset, 75% FP............ 62  :

4.12-2 Backup Incore Offset vs. Full Incore Offset, 75% FP.................. 63 4.12-3 Maximum LHR and Worst Case Minimum DNBR vs. Full Incore Offset, 75% FP 64 l

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1.0 INTRODUCTION

On September 18, 1987, Crystal River Unit 3 was shut down to refuel for Cycle 7. The Cycle 6 fuel had attained a core average exposure of 412.07 effective full power days (EFPD).

Cycle 7 has been designed for a cycle lifetime of 550 10 EFPD and l operation in a feed-and-bleed mode. The cycle utilizes burnable poison rod assemblies and gray axial power shaping rods (APSRs). The final Cycle 7 core loading pattern is shown in Figure 1.0-1. The fuel assembly and control component identifications for Cycle 7 are shown in Figure 1.0-2.

This report, prepared and submitted in accordance with Technical Speci- i fication 6.9.1, describes the precritical, zero power, and power testing i performed during the Cycle 7 start up and also summarizes the results of these tests.

Crystal River Unit 3 (CR3) was brought to Mode 3 (Hot Standby) for '

precritical testing. The results of these tests are reported in Section ,

2.0. Zero power physics tests began after criticality was achieved on January 8, 1988 at 2042 hours0.0236 days <br />0.567 hours <br />0.00338 weeks <br />7.76981e-4 months <br />. The zero power physics testing results are given in Section 3.0. Power escalation testing began after breaker - i closure on January 10, 1988 at 1034 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.93437e-4 months <br />. These test results are j summarized in Section 4.0. The all-rods-in (ARI) temperature l i coefficient, the hot zero ' power (HZP) ejected rod worth test, and one intermediate power distribution test plateau were not performed during  ;

Cycle 7 physics testing in accordance with a program for proving plant availability through the elimination of unnecessary tests i

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1 I"Reduced Physics Testing, Task Summary Report" document by Babcock I and Wilcox, B&W document number 86-1164722-00, dated March 1987 for l B&W Owners' Group Performance Committee. 1 l

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2 CORE LOADING DIAGRAM, CYCLE 7 rm ::asFtt 'an i 1

1 8 8 9 8 8 A .O 106 F K10 M12 g 8 9 8 9 8 9 I 9 8 LO3 F n03 F C08 F M '. 3 F L13 ,

50 9 8 9 8 9 8 9 8 9 5C Mll F K02 F K04 F K12 F K14 F a:8 C

CT4 CT4 8 9 8 9 7C 9 2C 9 7C 9 8 9 8 0 C10 F L11 F 806 F 707 F 810 F n06 F C06 Cv?

9 8 9 8 9 7C 9 7C 9 8 9 8 9 g

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CT2 CT2 9 8 9 8 9 7C 9 7C 9 8 9 8 9 M F 709 F Fil { 207 F 209 F 106 F 707 7 2C 8 9 8  !

8 9 8 9 7C 9 9 7C 9 N (10 F P06 7 809 F F10 F FCS F OC6 C10 F i Ct2

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Design Report  :: Crcle N t Far 8etase*siea FIGURE 1.0-1 2

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FUEL ASSEMBLY and CONTROL ASSEMSLY COMPONENT IDENTIFICATIONS Fun im!;!t m i 1

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L:11 4711 17 I 1712 4 11 till 1711 37Q JUA di 2YE 48! 2123 492 2:1 4A; 33:4 410 2YK siM JV! 371 C44  !!!7 A!  !!!O  ::J3 ISAT C046 I!aw C0Ci I!BF AO!!  !!AI C43

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Source: B&W Cy 7 Fuel eSign Report 4:te: %': :refues all Nt t.*e lat:s 2 'wel Asse :ty !: %-: e e s .

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FIGURE 1.0-2 3

l 2.0 PRECRITICAL TESTING During all precritical testing required by Section 13.4 Table 13-2' of I the CR3 Final Safety Analysis Report (FSAR), the actual reactor power 4 was maintained at a subcritical level. The results of the precritical tests are summarized in Table 2.0-1.

2.1 Calibration and Neutron Resconse of Source Range Monitoring System i Precritical testing began following response checkout of the neutron i source range detectors. -These detectors are two BF proportional counters which measure flux from the suberitical to s2 art up level, i This test verified that the source range ' instrumentation provided a count rate of more than 2 counts /sec. These tests were completed in

! accordance with Performance Testing Procedure PT-100.

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2.2 Reactor Coolant Flow Test The Reactor Coolant Flow test was performed to measure reactor coolant (RC) flow. The results were compared with design calculations to verify adequate core coolant flow. The test was executed in accordance with Surveillance Procedure SP-224.

The acceptance criteria of the test'are as follows:1 Steady state total reactor coolant system (RCS) flow (three or four pump operation) at 532t2 'F and 2155 100 psig shall be within the following limits: ,

6 4 RCS Pumps, Flow > 139.7x10 lb /hr 3RCSPumps, Flow [104.4x10 lb*/hr Additionally, with all 4 reactor coolant pumps in operation, steady state loop flows shall be within 2% of each other. To account for measurement uncertainty, the following must be t ue:

i Measured Flow x 1.025 f 411,840 gpm (Max. Flow)

Measured Flow x 0.975 > 374,880 gpm (Min Flow) 2.2.1 Method During this test, the RCS temp 3rature was maintained at 532 2*F with a pressure of 2155 100 psig.

The steady state temperature, pressure, and flow of the tw coolant flow loops were taken every minute for ten minutes. This data nas then used to calculate the average reactor coolant flow for each flow loop and for the entire system.

2.2.2 Results The total RC flow was calculated to be 153.765x10 6 lb /hr. The flow difference between loops was calculated to be 0.661%. Ithe calculation 4 for measurement uncertainty was 411,077 gpm for the maximum flow and  !

391,024 spm for the minimum flow. Therefore, the acceptance criteria I was met.

I SP-224, Section 4, Rev. 4 5

2.3 Control Red Droo Time Tests The Control Rod Drop Time tests demonstrated that the drop times are in accordance with Technical Specifications requirements. The acc criteria for these tests, per Surveillance Procedure SP-102, are:gptance The individual safety and regulating rod drop times from the fully withdrawn position shall be < 1.66 seconds from power interruption at the control rod drive breakers to three-fourths insertion, 25%

position, with T,yg 2 525 F and either 3 or 4 reactor coolant pumps operating.

In order to check for uncoupled control rods, the individual rod drop times were compared to their respective group average drop times. The individual red drop time should not exceed its group average drop time by more than 0.05 sec.

2.3.1 Method The Control Rod Drive (CRO) Drop Time tests were 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 chart recorders for each control rod assembly from a contact on the manual trip switch. A second signal to indicate three-fourths insertion was furnished by a reed switch located on the position indicator tube of each CRD.

2.3.2 Results The results of the Control Rod Orop Time tests are presented in Table 2.3-1. The average drop time by group is summarized in Tablo 2.3-2.

By examination of Table 2.3-1, it is found that the ?hortest drop time is 1.26 seconds for CR 5-5 and CR 7-6. The longest drop time was 1.33 seconds for CR 2-2. The weighted average drop time for groups 1 through 7 was 1.293 seconds.

Based on the results shown in Table 2.3-1, the acceptance criteria were met by all of the control rods, i

i I SP-102, Section 4, Rev.17 l

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a 2.4 Chemical and Radiochemical Tests Chemical and Radiochemical testing was not performed during this start-up. These tests were conducted at initial startup and no plant modifi-cations have been made which' would invalidate the results of those tests.

2.5 Pressurizer Effectiveness Test No Pressurizer Effectiveness testing was done this startup since no i modifications have been made to the pressurizer. Therefore, the results '

of the testing done at initial startup are still valid.

1 2.6 In-Service Loose parts and Vibration Monitorina system Tests l l

The Loose Parts Monitoring Subsystem was calibrated during the course of l the outage per Surveillance Procedure SP-152. The normal schedule of '

neutron noise readings, combined with routine core ph.ysics and Loose .

Parts Monitoring Systems, is sufficient to alert pl.6nt personnel of I degrading core internals, if it occurs, i l

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TABLE 2.0-1 SUPNARY OF PRECRITICAL TESTING RESULTS Precritical Testing Test (Reference) Units Results Acceptance Criteria Acceptability

1. Calibration and Neutron cps 2 2 cps Source range channels. OK Response of Source Range indicate 1 2 cps Monitoring Tests (PT-100) 6 2a. Reactor Coolant Flow lb"/hr 153.765x10 Flcw for four pumps- OK (SP-224) must be greater than 139.7x10 lb,/hr. _

, 0.661% Loop flow within OK' 2% of each other I

b. Reactor Coolant Flow gpm 411,077 Max flow must be UK Measurement Uncertainty < 411,840 391,024 Min flow must be Of

> 374,880

3. Control Rod Drive Time seconds Tests (SP-102)

Least (CR 5-5, 7-6) 1.26 All rod drop times must OK Greatest (CR 2-2) 1.33 be less than or equal. OK to 1.66 seconds ,

Group 1 Average 1.30 None Group 2 Average 1.31 Group 3 Average 1.29 Group 4 Average 1.31 Group 5 Average 1.28 Group 6 Average 1.29 Group 7 Average l'.28 Groups 1-7' Average 1.293 None

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TABLE '.3-1 CONTROL R00 ORIVE DROP TIME TESTS RESULTS Precritical Testing i

Orop Time Control Rod Core Position (seconds)-

CR 1-1 B-10 1.30 CR 1-2 F-14 1.32 CR 1-3 L-14 1.30 CR 1-4 P-10 1.29 l CR 1-5 P-6 1.30 l CR 1-6 L-2 1.30 '

l CR 1-7 F-2 1.31  !

CR 1-8 B-6 1.28 l l

l I CR 2-1 C-11 1.29 CR 2-2 E-13 1.33 CR 2-3 M-13 1.31 CR 2-4 0-11 1.31 CR 2-5 0-5 1.32 CR 2-6 M-3 1.32 CR 2-7 E-3 1.29 CR 2-8 C-5 1.32 CR 3-1 F-8 1.29 l CR 3-2 G-9 1.29 i CR 3-3 H-10 1.29 CR 3-4 K-9 1.29 CR 3-5 L-8 1.29 CR 3-6 K-7 1.31 CR 3-7 H-6 1.29 CR 3-8 G-7 1.30 CR 4-1 E-9 1.31 CR 4-2 G-11 1.29 CR 4-3 K-11 1.31 CR 4-4 M-9 1,32 CR 4-5 M-7 1.31 CR 4-6 K-5 1.32 CR 4-7 G-5 1.32 CR 4-8 E-7 1.31 l

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TABLE 2.3-1 CONTROL PrecriticalR00 ORIVE OROP TIME TESTS RESULTS Testing (Continued)

Drop Time

( secoiids)

Core Position Control Rod 1.29 C-9 1.28 CR 5-1 E-11 CR 5-2 1.28 G-13 1.30 CR 5-3 K-13 CR 5-4 1.26 M-11 1.28 CR 5-5 0-9 CR 5-6 1.28 0-7 1.29 CR 5-7 M-5 CR 5-8 1.30 K-3 CR 5-9 1.28 G-3 1.28 CR 5-10 E-5 1.28 CR 5-11 C-7 CR 5-12 1.29 B-8 CR 6-1 1.29 CR 6-2 F-10 1.29 H-14 1.30 CR 6-3 CR 6-4 L-10 1.28 P-8 CR 6-5 1.29 CR 6-6 L-6 1.29 H-2 CR 6-7 1.29 CR 6-8 F-6 1.28 CR 7-1 0-8 1.28 0-12 CR 7-2 1.28 H-12 CR 7-3 N-12 1.28 CR 7-4 1,27 N-8 CR 7-5 1.26 N-4 CR 7-6 1,29 H-4 CR 7-7 1.29 CR 7-8 0-4 i

Source: SP-102 10 1

TABLE 2.3-2 GROUP AVERAGE CONTROL ROD DROP TIME Precritical Testing Number of Average Orop Time Groue Rods (Seconds) 1 8 1.30 2 8 1.31 3 8 1.29 4 8 1.31 5 12 1.28 6 8 1.29 7 8 1.28 1-7 60 Average = 1.293 1

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3.0 ZERO POWER PHYSICS TESTING  ;

ine zero power physics tests (ZPPT) required by Section 13.4 Table 13-3 of  !

the CR3 FSAR were perforecd in accordance with Performance Testing Procedure PT-110. These tests were done to verify nuclear design parameters used in the-safety analysis, operational parameters, and limits set in the' Technical Specifications. ,

Acceptance criteria deviation limits for the ZPPT are given in Table 3.0-1. A summary of measured and predicted values obtained for the ZPPT is given in Table 3.0-2.

3.1 Initial Criticality ,

3.1.1 Method

.The reactor coolant conditions at criticality were 532'F, 2155100 psig,.

and 1965 ppmB. In order to verify that the source range instrumentation was operating properly, a Chi-Square Test was performed which confirmed ,

that a good fit to a Poisson distribution existed. The approach to  :

criticality began by withdrawing control rod group 8 to 25% withdrawn and  !

groups 5 and 6 to 100% withdrawn. Control rod groups 1 through 4 remained .

at 100% withdrawn. Rod group 7 was then withdrawn until criticality was l achieved (40% withdrawn).

Throughout the approach to criticality, curves of inverse multiplication i versus rod reactivity worth removed were maintained. The data was taken from two source range detectors by two people, independently. At the end of each control red group withdrawal, the count rate was taken from each source range detector using the scaler-counters. The ratio of the initial j average count rate to the average count rate at the end of each reactivity i addition was plotted and the._ criticality point determined by extrapola- l tion.

3.1..! Results Initial criticality for Cycle 7 was achieved on January 8,1988 at 2042-hours.

3.2 Nuclear Instrumentation Overlao The nuclear instrumentation (NI) detectors were used throughout testing to provide continuous reactor power information. The instrumentation consisted of eight measuring channels divided into three ranges; source (suberitical to startup), intermediate (startup to 150% full power), and power (0 to 125% full power). The location of these NIs is shown in Figures 3.2-1 and 3.2-2.

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Technical Specifications state that:

NI Source rgnge must overlap the intermediate range by a factor of 10 or more This means that before the source range count rate equals 10 5 cps, the intermediate range must be on scale. If the required one decade is not observed, the approach to the intermediate range cannot be continued until the situation has been corrected.

3.2.1 Method For this test, reactor coolant conditior.s were 532'F and 2155 psig. To verify the overlap requirements after initial criticality was reached, core power was slowly increased until the intermediate range channels came on scale. Detector signal response was thus recorded for both the intermediate and source range channels. This was then repeated for another decade.

3.2.2 Results The results of the Nuclear Instrumentation overlap test' are given in Table 3.2-1. This table shows that the average overlap between the source and intermediate range is 2.155 decades, which is above the minimum one decade specified as the acceptance criterion.

3.3 Sensible Heat Determination By determining the intermediate range current level at which the produc-tion of sensible nuclear heat occurs, the upper zero power physics test current limit is established. Thus, by restricting reactor power opera-tion to a level recuced by a conservatism factor of 3.3 below the sensible heat level, the effects of temperature feedback are eliminated in the measurement of physics parameters. The test for sensible heat was done according to performance Testing Procedure PT-116.

! 3.3.1 Method For this test, the intermtdiate range current level was increased in one-third decade increments until sensible heat was detected. The production of sensible heat in the core is indicated by an increase in the RCS T,yg, RCS loop Thot, and makeup tank level.

3.3.2 Results ,

The point at which there was a definite heatup rate was 1.1x10 amps measured on channel NI-4. This was defined as the sensible heat point.

8 From this, the upper current limit was established at 3.33x10 amps.

This was found by dividing the sensible heat point by a 3.3 conservatism factor.

I PT-110, Section 10.1, Rev. 12; Tech Specs Table 4.3-1, Note 5 13

3.4 Reactimeter Check The Reactimeter is the reactivity computer manufactured by Babcock and Wilcox 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 record 23 other analog and digital signals from the plant.

After initial criticality and prior to the first physics measurements, an on-line functional check of tda Reactimeter was performed to verify its readiness for use in the test program.

The Reactimeter check is subject to the following acceptance criterion:

The reactivity values computed from measured doubling times must agree withi 5% of the reactivity values measured on the Reactimeter.g 3.4.1 Method After steady state conditions with a constant neutron flux were established, approximately 25 pcm of negative reactivity was inserted into the core by inserting control rod group 7. Stop watches were used to measure tne doubling time of the neutron flux and the inserted reactivity was determined from period-reactivity curves. The measurements were repeated for several values of reactivity ranging from

-79.8 to +69.4 percent millitho.

3.4.2 Results The reactivities determined from doubling time measurements were then compared with tne reactivities calculated by the Reactimeter.

The results of the Reactimeter verification measurements are -summarized i in Table 3.4-1 The reactivity calculated by the Reactimeter was within the acceptance criterion limit of 15% of the reactivity determined from ,

doubling times in each case.

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3.5 All Reds Out Critical Boron Test  ;

L This test was used to provide information relating to core excess reac-tivity by determining the amount of soluble boron that must be added to the coolant water to maintain a critical level with all control rods  !

4 removed.

This test, performed in accordance with -PT-111, is subject to the following acceptance criterion:  !

The measured all rods out critical boron concentration... should  !

be withini 50 ppm 3 of the value given in the Physics Test f Manual t

3.5.1 Method The test measurements were made at a reactor coolant temperature of 532*F and a system pressure of 2155 psig. The portion of control rod group 7'which remained following deboration was withdrawn and the excess  ;

reactivity measured using the Reactimeter. From this value and the differential boron worth given in the Physics Test Manual, the all rods ,

out critical' boron concentration was determined. l 1

3.5.2 Results

• The excess reactivity was found to be 0.0225 Mk/k. From this and the j

. differential boron wcrth from the Physics Test Manual, it was determined -

that the all rods out critical boron concentration was 2033 ppmB. Since  :

the predicted value is 2032 ppmB, the acceptance criterion was met.  !

~

3.6 Moderator and Temoerature Coefficients Measurements ,

J l

The temperature coefficient of reactivity is defined as the fractional  :

change in the excess reactivity of the core per unit change in core l 3

temperature. The temperature coefficient is normally divided into two components as shown below.

c j T * "M + D

! where: a T = Temperature Coefficient of Reactivity a M = Moderator Coefficient of Reactivity a 0 = Doppler Coefficient of Reactivity I

i A

pT-111, Section 10.1, Rev. 7 15

For this test, performed in accordance with PT-114, the temperature and moderator coefficients at hot zero power were measured. Furthermore, an extrapolated hot full power moderator coefficient was calculated. Each of the above coefficients' was - determined. for an all-rods-out (ARO) control rod configuration, with rod bank 8 remaining at 25% withdrawn.

Previously, these calculations were repeated with regulating rod banks 5-7 ins This all-rods-in (ARI) calculation was eliminated-because:grted.

1. There is good agreement between predicted and measured 4 coefficients.

2. There is a very good correlation between the direction of predicted versus measured deviation for the ARO and ARI

. coefficients for a given. reload cycle.

The acceptance criterion for the hot zero power temperature coefficients is:

The calculated.got zero power temperature coefficient...shall be within Manual.

30.4x10 ak/k/*F of.the value given in the Physics Test The acceptance criteria for the hot zero power moderator coefficients of reactivity are:

The hot _4tero power moderator coeffici,yt shall e greater than

-3,0x10 ak/k/'F but less than 0.9x10 ak/k/'F For the extrapolated moderator coefficient, the criterion is as follows:

The $t full power moderator coefficient shall 'be less than Zero.

3.6.1 Method 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 temperature consynt. The reactor was maintained at a critical level between 5 x 10 amps and the upper zero power physics test current limit as defined by the sensible heat determination experiment. Equilibrium boron concentration was established in the reactor coolant system, makeup tank, and I"Reduced Physics Testing, Task Summary Report" document by Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for he B&W Owners' Group Performance Committee.

}pT-114,Section10.4,Rev.11 3

4 PT-114, Sections 10.1, 10.3, Rev. 11 PT-114, Section 10.2, Rev. 11 16

pressuri:er to eliminate reactivity effects from boren 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-pen recorder.

Once steady state conditions were established, a positive heatup rate was maintained by adjusting the turbine header pressure set point. As the reactivity changed, the controlling rod group was moved as necessary to produce an adeauate intermediate range signal. After the core average temperature increased by about 5*F, coolant temperature and reactivity were stabilized. This process was then reversed, and core average temperature was decreased by about 10'F. Finally, after stabilizing coolant temperature and reactivity, the core average temperature was returned to its original value. The measurement of the temperature coefficient from the data obtained was performed by dividing the change in reactivity by the corresponding changes in core temperature for a specific time period.

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 coeffi-cient has safety implications, it is an important reactivity coeffi-cient. To obtain the moderator coefficient from the,peasured tempera-ture coefficient, a Doppler correction of -0.151x10 Ak/k/*F must be subtracted.  ;

The extrapolated hot full power moderator coefficient is calculated based on the hot zero power moderator coefficient. Added to this are both a calculated control rod effect and a boron change effect, which include the Doppler and xenon effects.

3.6.2 Results The results of the hot zero power temperature and moderator coefficients measurements are summarized in Table 3.6-1 along with the predicted values which are included for comparison. In all cases, the measured results compared f avorably with the predicted values. All measured temperature coefficie.nas of reactivity were within the acceptance criterion of 10.40x10 ' ak/k/*F of the predicted value. In addition, calculation of the moderator coefficient indicates that it is well within the recuirements of the Technical Specification.

i Theextrapolatedhogfullpowermoderatorcoefficient,asgiveninTable l 3.5-2, is -1.41x10 3k /k/*F for all rods out. From this, it is seen that the acceptance criterion of a hot full power moderator coefficient of less than zero is met.

1 1

17 i

l 3.7 Control Rod Grouc Worths i The layout of the core showing the location of the control rod groucs I and the location of the 52 incore detector strings is given in Figure 1 3.7-1. The number of rods in each group and the reactivity control i function of each group is listed below.  ;

Red Groue No. No. of Rods Control Function 1 8 Safety 2 8 Safety '

3 8 Safety 4 8 Safety 5 12 Power Doppier 6 8 Power Ocppler 7 8 Transient '

8 _8 Axial Power Shaping Total = 68 The control rod worth tests were run in accordance with Performance Testing Procedure PT-112 to provide information about the reactivity worths of banks 5, 6, and 7. The results are also used to verify red i worth curves used during plant operation. The tests are subject to the following acceptance criteria:

The predicted reactivity worth of rod groups, 5, 6, and 7 in the Physic each be within 115% of the measured value.y Test Manual shall Additionally, the predicted total reactivity worth of the sum of red groups 5, 6, and 7 in th shall be within !10% of the measured total.g Physics Test Manual 3.7.1 Method Measurements of control rod group worths for groups 5, 6, and 7 were made during zero power physics tests using the boron swap methed. This method consisted of setting up a deboration rate and compensating for the change in reactivity by small step changes in red group positions.

The continucus calculation of reactivity was made by the Reactimeter.

The reactivity in percent millitho (PCM) from the Reactimeter and the l rod position were recorded on a strip chart and magnetic tape.

3.7.2 Results ,

The results of both the predicted rod group worths and the measured group worths are tabulated in Table 3.0-2 for a reactor coolant system temperature of 532*F and pressure of 2155 psig. Comparison betwean i measured and predicted rod worths shows groups 5-7 were well within the l specified 15% deviation, and the sum of groups 5-7 was within 3.029% i percent of the predicted value. Therefore, 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 3.7-2 through 3.7-4. ]

i fPT-112,Section10.1,Rev.7 PT-112, Section 10.2, Rev. 7  !

18 I i

3.8 Differential Boron Worth Determination 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 through-out the life of the cycle.

i The differential boron worth was measured in accordance with Performance Testing Procedure PT-112. From PT-112, the following acceptance criterion must be met:

ThepredicteddifferentialboronworthintgePhysicsTestManual shall be within !15% of the measured value 3.8.1 Method The test measurements of the boren differential worth was completed at reactor coolant conditiens of 532'F and 2155 psig. The measured value was determined by summing the incremental reactivity values measured during the rod worth measurements over a known boron concentration range from 2028 to 1635 ppmB.

3.8.2 Results The measufed differential boren worth was calculated as

-7.8117x1g 3k/k/ppmB, as compared to the predicted value of

-7.60x10 ak/k/ppmB. The deviation is therefore -2.71%, which is well within the acceptance criteria of 15%. The results of the differential soluble poison worth measurements are *.abulated in Table 3.8-1. '

. 3.9 Ejected Control Red Worth Measurement In previous cycles, this test was performed to verify the safety  !

analysis calculations relating to the assumed accidental ejection of the l most reactive control rod. From the existing B&W data base, there is )

good agreement between g the measured and predicted ejected rod worths, l It has been determined that the HZP ejected rod worth (ERW) test can be i eliminated if the predicted value of maximum HZP ejected red worth is less than 0.8 %3k/k. Since the predicted ejected rod worth from the Physics Test Manual is 0.434 %3k/k, the ejected control rod worth test has been eliminated from zero power physics testing for Cycle 7.

l 1

fpT-112,Section10.3,Rev.7 "Reduced Physics Testing, Task Summary Report" document by Babcock i and Wilcox, document number 86-1164722-00, dated March 1987 for the B&W Owners' Group Performance Ccmmittee.

l 19 l

. , - + , - - _ .y c , . - , , - , . -

l 3.10 Biological Shield Survey A Biological Shield Survey was not done at hot zero invalidate power since the Biological no plant Shield J

' modifications were made which would Surveys made during the Initial Startup Testing Program.

l

! 3.11 Effluent and Effluent Monitoring No Effluent or Effluent Monitoring testing was performedTherefore, as these systems have been performing normally since initial startup.

no further testing was required.

I 3.12 Chemical and Radiochemical Tests

' Chemical and Radiochemical testing was not performed during this start-up. These tests were conducted at initial startup and no plant modifi-cations have been made which would invalidate the results of those tests.

l l

e 20 t

l i

l TABLE 3.0-1 ACCEPTANCE CRITERI A DEVI ATION LIMITS BETWEEN MEASURED AND PREDICTED VALUES l

l Zero Power Physics Testing Allowable Deviation Between Core physics Parameters Measured & Predicted Values o All Rods Out Boron Concentration t 50 ppmB

~4 o Temperature Coefficient of Reactivity 0.4x10 ak/k/*F o Control Rod Worths Individual Group Worths (Groups 5, 6, & 7) -

15%

Total Group Worth (Groups 5-7)  ! 10%

o Differential Boron Worth  ! 15%

l 1

1 I

1 l

i e

21

Table 3.0-2 SUPMARY OF ZERO POWER PHYSICS TESTING RESULTS Zero Power Physics Testing Physics Parameter (Reference) Units Measured Predicted Acceptance Criteria Comparison

1. NI Overlap (PT-110) decades 2.155 Overlap must be OK (2.155) greater than 1.0 decade

2. Sensible Heat (PT-116) Amps

~

NI-3 1.1x10_f None NI-4 1.1x10

3. All Rods Out Critical ppmB 2033 2032 Measured value must OK (1)

Boron (PT-111) be within 50 ppmB of predicted value M

t l 4. Temperature Coefficient Ak/k/ F

! of Reactivity (PT-114) l

-4 -4 ARO, 2030 ppmB +0.225x10 +0.286x10 Measured value mugt OK be within 0.4x10 Ak/k/*F of the pre-l dicted,value l S. Moderator Coefficient Ak/k/*F -

! of Reactivity (PT-114)

• ~4 ~4

( ARD, 2030 ppmB 40.376x10 +0.453x10 Maximum positive OK '

! moderator coefficient ~

l must bg greater than OK

-3x10 k F and less j A/kf*Ak/k/*F.

than 0.9x10

^

i t _ _ _ _ , . - . . - . - - - - - _ _ . . - - - . - - - . - .- _ - . . - . . _ _ _ _ _ - - _ _ - _ _ _ _ - _ ____

'Zero Power Physics Testing (Continu:d)

Physics P:rameter (Raference) Units Measured Predicted Acceptance Criteria Comparison

6. Extrapo M:-d Moderator Ak/k/ F Coefficiut of Reactivity (PT-114)

ARO -1.41x10 -0 Extrapolated hot full OK power moderator coeffi-cient must be less than 0.0

7. Control Rod Group Worth  % Ak /k (PT-112)

Group 7 -0.8375 -0.881 Percent deviation of OK (5.13)

! Group 6 -0.8320 -0.927 group worth must be less OK (11.42)

Group 5 -1.4005 -1.355 than 15% OK (3.25)

Total -3.071 -3.163 Percent deviation of OK (3.029) total worth must be U . less than 10%

1

8. Differential Boron  % Ak/k/ppmB l Worth (PT-112) _3 1831.5 ppm 8 -7.812x10 -7.60x10 _3 Percent deviation must OK (-2.71) be less than 15%

l Percent Deviation is calculated as follows:

% Deviation = Predicted MeasuredValue - Measured Value Value x 100 l

TABLE 3.2-1 f40 CLEAR If1STRUMEfJTATI0rd OVERLAP TEST RESULTS Zero Power Physics Testing Average SR Average IR Case Source Range l'SR) Indication Intermediate Range (IR) Indication Indication Indication Overlap

• tJi-3 ( Amps) til-4 (Amps) (CPS) (Amps) (Decades) fiumber til-1 (CPS) fil-2 (" ',)

- sammmmmmmmmmmmum -

4 1 6x10 4

6x10 7x10'II 7x10'II 6x10 7x10'II 2.067

-10 -10 5 -10 5

7x10 7x10 7x10 7x10 7x10 2.000 2 7x10 5 -10 -10 -10 2.398 2x10 4x10 6x10 2x10 Sx10 3 2x10 .

Average Overlap = 2.155

% *0verlap between the Source (SR) and Intermediate Range (IR) average indications is obtained using the following equation:

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

Source: PT-110. Enclosure 4, Attachment 1, Console Values

TABLE 3.4-1 REACTIMETER AND DOUBLING TIME REACTIVITY COMPARISON Zero Power Physics Testing Doubling Average Reactivity Reactimeter Absolute Case Time (DT) from DTs Reactivity Error

• Number (Sec) (PCM) (PCM) (%)

1 223.3 -26.8 -26 3.1 2 202.0 +23.5 +24 2.1 3 118.7 -58.2 -56 3.9 4 97.7 +43.9 +44 0.2 5 95.7 -79.8 -76 5.0 6 54.7 +69.4 +69 0.6

• Absolute error (E%) between the Doubling Time reactivity (PDT) '"d Reactimeter reactivity (Pg ) is given by the following equation.

E(%)=100*l(P DT ~#

R) / I#R)!

Source: PT-110, Enclosure 5, Attachment 1 1

25 I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

TABLE 3.6-1 COMPARISON OF MEASURED AND PREDICTED REACTIVITY COEFFICIENTS Zero Power Physics Testing Control Rod Group Average RCS Boron Reactivity Coefficients ( Ak/k/*F)

Rod (Group Position, % Withdrawn) Temp Concentration Temperature Coefficient Moderator Coefficient Configuration 1 2 3 4 5 6 7 8 ( F) (ppm) Measured Predicted Measured Predicted

- amme amma e summ umas muun mens - -

ammmmmmmmmmmmm ammmmmmmmmmmm

~4 ~4 -4 -4 ARO 100 100 100 100 100 100 90 25 532 2030 +0.225x10 +0.286x10 +0.376x10 +0.453x10

- ' - All Rods Out p .1-114, Enclosure 1 k

m

TABLE 3.6-2 EXTRAPOLATED HOT FULL POWER MODERATOR COEFFICIENT OF REACTIVITY Zero Power Physics Testing Predicted Hot Full Power Moderator Red Configuration Coefficient of Reactivity (Ak/k/oF)

-5 All Rods Out -1.41x10 Source: PT-114, Enclosure 2 27

TABLE 3.8-1 DIFFEREf4TIAL BOR0ft REACTIVITY WORTH TEST RESULTS Zero Power Physics Testing Control Pod Group Measured Average Delta Boron Differential Boron Worth (Group Position, % Withdrawn Boron Conc. Boron Conc. Boron Conc. Worth (%\k/k/ppmB) 1 2 3 4 5 6 7 8 (ppm) (ppm) (ppm) (%\k/k) Measured Predicted ammu amme mums amm amm numa smin a- ammmmmmmmmme ammmmmmmmmmunsma smalammum - mummmmmmmmmmmmens 100 100 100 100 100 100 89 25.0 2028

-393 -3.070 -7.8117x10

-3 -7.60x10 -3 100 100 100 100 0 0 0 25.0 1635 1831.5 Source: PT-112 Enclosures 1 and 4 m

N1 1 PC P.-p B PrP A A

N1 5 UCiC f Nl.3 I

CIC l Hl . 7 UCIC

. 4 .

8, ,,.,*.

.*8 O

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NI.E NI.E UCIC UCIC NI-4 P. p D 2 Nl.2 C10 pg PpC CCSG B PC - PROPORTIONAL COUNTER SOURCE 'ANCE 1 DETECTOR Cic COMP!NSATED 10N CHAFSER - INTERNEDIATI RANCE DETECTOR.

UCIC UNCCMPEW5ATED IGN CNAMBER - PotER RANCE DETICTOR Sc=ce: FSAR, Fig. 7-18 Fig. 3.2-1 NUCLIAR IN S TRD'. INT A!!O N DITICTOR LOCATIONS 1

1 29

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12'- 5 g _

31' y 11'-7 g Fig. 3.2-2 NUCLEAR INSTRUMENTATION DETECTOR LOCATIONS 30

INCORE MONITOR N4D C0 lTROL ROD PAP 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 9 NI-1 ,

9 N1-3 A l 8 NI-5 CR1-a CRt 1 C;1 1

• NI-7

'CR2-F tRS-12 CR ( 1 ) CR2-1 b S D

CR7-8 Q

3 A

CR7-1 _

~A CR7-2 5

CR2-7 _fCRS-n CR4 a CRe ' Cy CR2-2 E

CR)-7 b

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? CR3-1 CR6-2 Y

" CR1-2 31 Mh b CR4 7 ICR3-S CPL ' CRt il G g ICR54C y ]e CR;g.21 y y CRS-7 'CR7 7 CR3-7 CF.3 3 CR7 3 5 h)CR53 CR5 9 CRe-6 CR3-6 ICR3-4 CR4-3' ICR5-4 K

g 7 Gr C R 1 - 61 CR6- 61 CR3 5 CR6-4 .

ICR1-3 CR2-6 CR5-8 CR4-E CR4 4' ICR5 5 CR2-3 h h _

d 'h CR CR Sl <

CR7-4 5'

CR' 'CR5-7 CRS-6 CRC-4 CR1-5 CRS-5 CR1-4 p

M e NI-8 e NI-6 9 N!-4 e N1-2 8

Total Core Monitor h Total Core & Symetry Monitor

{ Symetry Monitor

- Incore Monitor Number CRX-Y - Control Rod Y of Control Rod Group X Fi 7are 3.7-1 31

Control Rod Group 5 Inteoral Worth 80CT 7 at HZP L.H thilit4L b0Ltu

%4 (% Jt/c) 0.0 1.421 10.0 1.365 ta4  !!:i  !:!!!

14.0 1.311 19.3 1.219 16.5 1.248 17.5 1.253 18.5 1.232 19.5 1.2:6 20.5 1.113 21.0 1.165

1. M - 21.5 1.152 22.0 1.129 23.0 1.103 24.$ 1.C64

!$.5 1.C30 27.0 0.954 m 29.0 C.ll1 g 30.5 C.509

% 32.5 0.870 g 1.00 - 34.0 0.325 37.0 c.?s?

1 40.0 C.712 43.5 3 154 g 47.0 C.llt w 11.0 C.llt

$1.0 0.419 g 19.0 C.433 g 44.0 0.373 L 0.M - 78.0 78.0 C.300 c.224 0 81.0 C.164 3 84.0 C.Cil 91.0 e.est 100.0 C . C.$0 L

a j O.M -

C H

0.40 -

i 1

0.H -

" i i i i l e n a se se tw Rod Position (% Withdrawn)  :

Source: PT-112. Encl. 10 Figure 3.7-2 32

i I

Control Rod Group 6 Inteoral Worth

. BOC 7 at HZP l s.00 i thT(414 b04?w

%4 (% fdE)

C.0 C.832

0. e0 - 18.0 0.786 27.0 0.715 34.0 C.642

, 40.5 C.160 48.0 C.ill i

54,5 C.410 41.0 C.338 0.s0 - '

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0 M 40 00 00 600 l Rod Position (% Withdrawn) i S0urce: PT-ii2. Encl. 10 Figure 3.7-3 33

l r

Control Rod Group 7 Intecral Worth

. BOC 7 at HZP ,

a 1.00 .

Initten s:4TN

%4 (% f /t) 0.0 0.838

., 0.90 - to.0 0.7sl 1 24.5 0.734 4

33.0 0.674 1 39.0 0.621 48.0 0.536 I

$0.0 0.478 15.0 0.418 4

g , gg , 59,0 0.370

! 62,0 0.332 64.5 0.242 74,0 0.194 80.0 0.131 35.5 0.075  :

200,0 0.c00 l 0.70 -

i 2

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\

0.10 -  ;

8 i

H i

40 i

N i

N

\ $09

Rod Position (X Withdrawn 1  !

Source: PT-112. Encl. 10 Figure 3.7-4 i 34

4.0 ' FULL POWER ESCALATION TESTING Full power escalation tests were performed to verify the validity of the safety analysis assumptions and, therefore, the acceptability of the core design. Testing was performed at two major power plateaus; 75% and 95-100% full power (Fp). In previous cycles ' testing was performed- at three major power plateaus, but one intermediate = power distribution test plateau was eliminated as explained in the Introduction (Section 1.0) 4.1 Turbine / Reactor Trio Test No Turbine / Reactor Trip testing was performed during this startup since no modifications were made which would invalidate the original testing results.

4.2 Intearal Control System Test Since no modifications were made to the Integrated Control System (ICS) during this outage, no specific ICS testing was done during this start-up. Minor adjustments were made to the ICS under normal maintenance and calibration procedures. '

-i l

4.3 Unit loss of Electrical Load No Loss of Electrical Lead testing was performed since no modifications I were made which would invalidate the original testing results.

4.4 Unit load Transient Test  !

No specific Unit Load Transient testing was performed since no modifica-tions were made to invalidate the results of previous tests. No major ,

problems were encountered during transient operations throughout the l testing program.

i 4.5 Reactivity Coefficients At Power Test

. The purpose of this test was to determine reactivity coefficients at 100% full power, and to verify that they were conservative with respect to the .FSAR. "ihe following coefficients were either measured or 25

calculated from the data obtained:

a. Temperature coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in fuel and moderator temperature.

b. Moderator coefficient of reactivity, defined as the fractional 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.

d. Fuel temperature Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in fuel temperature.

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

1. The moderator coefficient of reactivity shall be less positive than

~

0.9x10 ' akt'k/ F at power levels below 95% full power, less positive than 0.0x10 ak/k/cF at power _1.evels at or above 95% full power and less negative than -3.0x10

• ak/'4' F at rated thermal power.y

2. ThefueltemperatureDop,plercoefjicientofreactivityshallbemore negative than -0.90x10 ak/k/ F 4.5.1 Method Reactivity coefficient measurements were made during the power escalation test program at 100.0% 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 coefficients, average reactor coolant temperature was increased and decreased about 5F and data recorded. For power Doppler coefficients, power was increased and decreased about 5% full power and data recorded. From the measured temperature and power Doppler coefficients, the moderator and fuel temperature Doppler coefficients were calculated.

4.5.1.1 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 measurement, the controlling rod group is inserted for six seconds, followed immediately by a withdrawal for 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 essentially a combination of the effects due to the control rod motion and the fuel power variation.

kPT-120,Saction10.2.7,Rev.11

'PT-120, Section 10.2.8, Rev. 11 36

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 corection factor accounts for the time delay involved in fuel temperature change during the measurement.

4.5.1.2 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 4.5-1. ,

T* M+CO * * }

Where: T = Temperature Coefficient of Reactivity (ak/k/ F) a =

a M = Doppler Moderator Coefficient Coefficient of of Reactivity Reactivity (ak/k/

(ak/k/ F) F) 0 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 4.5-1 after the temperature and Doppler coefficients have been determined.

Temperature, moderator, and Doppler coefficients were theoretically predicted as shown in Table 4.5-1 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 cr;tical boron and rod conditions was assumed which set the average core moderatar temperature 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 approximate 5 F change in the reactor coolant temperature. The reactivity change caused by 'the temperature change of the core was measured by recording the change in the position of the. controlling 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 conditions, including a stable boron concentration and no significant control rod motion during the last 30 minutes prior to taking dcta, were required as prerequisite system conditions.

The fuel temperature Doppler coefficient relates the change in core reactivity to a corresponding change in fual temperature. A theoretical prediction of the fuel temperature Doppler coefficient was made using the P0Q code with thermal feedback, and is presented in Table 4.5-1.

The measurement method used was to change the reactor power level 5%

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 calculation of the power Doppler coefficient uses the measured 37 w . - _ _ - _ , , . . , . .____ , . ,.

I 1

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 JT , which is the primary side heat balance.

This is then converted to a fuel temperature Doppler coefficient by multiplying by a theoretically derived factor, 3%FP/3*F.

4.5.2 Results ,

The results of the measured temperature and Ooppler coefficients, and -

calculated moderator coefficient at power are shown in Table 4.5-1 which )

also shows the predicted temperature, moderator, and Doppler coefficient results.

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

The results of the measured and predicted Doppler coefficient of reactivity are shown in Table 4.5-1. The acceptance criterion for the measured Doppler coeffgcient is that the coefficient must be more .

negative than -0.90x10 Jk/k/ F. Table 4.5-1 shows that the mesured coefficient is below this value and that the acceptance criterion is adequately met.

S 4.6 Unit Heat Balance Heat balance calculations were performed using the ModComp computer.

T'.e data was taken in accordance with Surveillance Procedure SP-312. No modifications were made during this shutdown which would require reveri-fication of the heat balance calculation.

4.7 Core Power ~ Distribution Test The Core Power Distribution test was performed at each power plateau (75% FP and 95-100% FP) to measure the core flux and power distri-butions. These measured powers were then compared to the predicted values. The test data was also used to evaluate core performance and the departure from nucleate boiling ratio (DNBR). Additionally, test results were used in the Power Imbalance Detector Correlation test and the Incore Detecter test.

The test, performed in accordance with PT-120, is subject to the follow-ing acceptance criteria:

The core power distribution and thermal-hydraulic parameters must be measured, evaluated, and deemed reasonable.*

I PT-120, Section 10.2.6, Rev. 11 38

l The highest measured radial and total peaking factors shall'not be greaterthan5%and7.5%,respectiyely,ofpredictedvaluesatthe 40-75% FP and 95-100% FP plateaus.

I The minimum DNBR is greater than 1.30 and the measured worsy case

! maximum linear heat rate (LHR) is less than 19.2 kW/ft when extrapolated to the next_ power plateau.

The extrapolated worst case minimum DNBR is greater than 1.30 and the extrapolated worst case maximum LHR is less than 19.2 kW/ft, or if the MLHR is - greater than 19.2 kW/ft, the extrapolated imbalance falls 4.7-1.gutside the power imbalance trip envelope as shown in Figure Continuous monitoring of the core power density at 364 core locations was accomplished using the incore monitoring system. This system is comprised of 52 detector strings each having 7 individual neutron detec-tors. These detectors are equally spaced at seven axial elevations in the center of 52 fuel assemblies. This system is capable of producing detailed core power distributions for either eighth core or quarter core symmetry conditions. The output of the incore detectors was connected to the unit computer and 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 was used to develop core power distributions which provide power peaking information neces-sary to determine 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 4.7-1.

The out-of-core nu: lear instrumentation provides the core power and core power imbalance signals to the reactor protection system, since the I incore monitoring system does not immediately respond to prompt changes in the core conditions. The out-of-core nuclear instrumentation (NI) is shown in Figure 3.7-1 as NI 1 through 8.

l 4.7.1 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 configurations. For this test, the APSRs were maintained at a constant position and the axial incere imbalance was maintained within 2% FP of the imbalance identified in the Physics Test Manual.

fpT-120,Section10.2.5,Rev.11 3

PT-120, Sections 10.2.3 and 10.2.2, Rev. 11 PT-120, Enclosure 2, Parts 5.0 and 6.2, Rev. 11 39

4.7.1.1 Normal Operating Core Power Distributions. Normal operating equilibrium xenon core power distributions were measured in this test. Data was taken at each power plateau; 72.9% FP, and 100% FP. Values were obtained over an eighth core for radial and maximum total power peaking factors. These results were then compared with the values predicted in the Physics Test Manual. Based on the maximum calculated and maximum measured peaks, percent deviations were determined. The percent devi-ation was calculated using the following equation:

% Deviation =

a cu a -Measure [

Measured x 100 EQ.(4.3-1)

The deviation was calculated for the radial and for the maximum total power peaking factors at each power plateau. These distributions are shown in Figures 4.7-2 through 4.7-5.

4.7.1.2 Worst Case Minimum DNBR. To maintain the integrity of the fuel cladding which prevents fission product release, it is necessary to prevent over-heating of the cladding under normal peaking conditions. The two pri-mary 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 19.2 kW/ft.

The upper boundary of the nucleate boiling region is called "departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures 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 pressJre Can be related to DNB throujh the use of the B&W-2 correlation. The B&W-2 correlation has been empirically developed by Babcock & Wilcox 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 99% confidence level that DNB will not occur. This is considered a conservative margin to DNB for all operational conditions.

The worst case minimum DNBR values were calculated by the unit computer for each core power distribution that was taken during the power escal-ation test program.

Two normal operating equilibrium xenon core power distributions, as required by the Core Power Distributioa Test, were obtained during the power escalation sequence. Each power distribution was subjected to the following analysis. From each core power distribution, the worst case measured mini, rum DNBR was selected. These DNBRs were then extrapolated I

40

l l

l l

to the overpower trip setpoint and corrected for axial peak location and I magnitude. Verification of acceptable core conditions at the present and next power level of escalation was then performed based on the I

acceptance criterion that the extrapolated DNBR must be greater than 1.30. The extrapolated DNBR values were supplied by the plant com-puter's PD0s based on actual conditions at the plateau. Next, the measured worst case minimum DNBRs were extrapolated to the Loss of l Coolant Accident and design overpower power level and corrected for l axial peak location and magnitude.

l 4.7.1.3 Worst Case Maximum Linear Heat Rate Determination. Worst case maximum l

linear heat rate (LHR) values were calculated using SP-104, "Hot Channel i Factors Calculations", for each standard core power distribution taken l

as part of the power escalation test program.

l l 4.7.1.4 Quadrant Power Tilt. Quadrant power tilt limits have been established I 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 tilts are subject to the following acceptance criteria:

The quadrant power tilts shall not e Technical Specifications Table 3.2-2.pceed the limits Furthermore, the specified in calculated incore tilts must be less than the error adjusted tilt limit.2 p wer Quadrant power tilt is defined by the following equation and expressed in percent:

P wer in Any Core Quadrant Quadrant Power : Average Power of All Quadrants -1 x 100 EQ.(4.3-2)

During the startup testing program, maximum quadrant power tilt was determined using the corrected signals from the 16 symmetric incere mon-itoring assemblies. Figures 4.7-2 through 4.7-5 include the maximum quadrant power tilt for each standard core power distribution taken as required by the Core Power Distribution Test.

4.7.2 Results 4.7.2.1 Normal Operating Core Power Distributions. The results of the core power distributions are shown in Figures 4.7-2 through 4.7-5. The per-cent deviation for each symmetric length core assembly is also given.

The acceptance criteria for this test were met, fPT-120,Section10.2.4,Rev.11 PT-120, Section 10.2.11, Rev. 11 41

4.7.2.2 Worst Case Minimum DNBR. The results of various worst case minimum DNBR values calculated at each test plateau under normal rod configurations are plotted in Figure 4.7-6. 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.

4.7.2.3 Worst Case Maximum Linear Heat Rate. In all cases, the acceptance criterion limit of 19.2 kW/ft was met during the power escalation test program. These results are summarized in Table 4.7-1.

4.7.2.4 Quadrant Power Tilt. Technical Specification tilt limits were not exceeded at any of the power plateaus.

4.7.2.5 Axial Power Imbalance. Results from standard core power distributions taken at 72.9% FP during the performance of the Power Imbalance Detector Correlation Test show that the imbalance trip envelope (Figure 4.7-1) of the reactor protective system is sufficient to protect the unit from exceeding the DNBR and the LHR limits under all core imbalance condi-tions. In addition, analyses indicate that the largest thermal margins (measured by DNBR and LHR) exist when a negative 5.0% to positive 10.0%

incore axial offset is present as shown in Figure 4.12-3.

The core imbalances measured in conjunction with the core power distri-bution of this section are included in Figures 4.7-2 through 4.7-5.

4.8 Biolooical Shield Survey A Biological Shield Survey was not conducted as part of this power test-ing program since no plant modifications were made which would invali-date the Biological Shield Surveys made during the initial startup testing program.

4.9 Pseudo Rod Ejection Test In previous cycles, the Pseudo Rod Ejection test was not performed during the power escalation test prograr. since this test had been performed during zero power physics testing. Since the maximum HZP ejected rod worth from the Physics Test Mpnual is less than 0.8 %_ik/k, it was unnecessary to perform these tests 1"Reduced Physics Testing, Task Summary Report" document by Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for B&W Owners' Group Performance Committee.

42

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 results of testing during the initial start up. -

l l

4.11 Loss of Offsite power No Loss of Offsite Power testing was performed since no modifications were made which would invalidate the results of the tests performed during initial startup.

4.12 power Imbalance Detector Correlation Test -

Power imbalance is defined as; Imbalance = % power in top of core - % power in bottom of core EQ. (4.5-1)

The imbalance of the neutron flux in a reactor results from temperature distributions, fuel depletion, xenon oscillations, or control rods positioned in the core. The amount of imbalance allowed in the core so that the DNBR or LHR limits are not exceeded is set in the reactor protection system. The imbalance is a function of the power level and the reactor coolant flow. Since this imbalance is determined using input signals from the out-of-core detectors, it is essential that they are calibrated to read the true imbalance as determined from the intere detectors.

The Power Imbalance Detector Correlation (PIDC) test was performed in accordance with Performance Testing Procedure PT-120. The PIDC was done to determine the relationship between core offset as indicated by the out-of-core power range NI detectors and core offset as indicated by the full incere monitoring system. Offset is the imbalance divided by the total core power. The PIDC also verified that an acceptable relation-ship between the backup and full incore monitoring systems offset was observed. Finally, the PIDC verified that the measured wcrst case mini-mum DNBR and maximum LHR were acceptable. These values were verified to assure that the plant operates within the assumptions made in the safety analysis calculations.

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

The measured correlation between each out-of-core detector offset to full incore monitoring system offset lies within the acceptance 43

region shown in Figure 4.12-1. The correlation slope shall be between 0.96 to 1.10.y The measured relationship between the full incore monitoring system offset and the backup incore recorder ffset lies within the acceptable region shown in Figure 4.12-2.2 The measured worst case minimum DNBR is greater than 1, measured worst :ase maximum LHR is less than 19.2 kW/ft.y0 and the 4.12.1 Method During the power escalation sequence at Crystal River 3, imbalance measurements were made to determine the acceptability of the out-of-core detectors' ability to measure imbalance and to establish a basis for r

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 72.9% FP.

In performing the test, the APSRs were positioned to obtain the desired full incere imbalance with reactivity compensations made by control rod groups 6 and/or 7. At 72.9% FP, the offset indicated by the full incore system, out-of-core system, and backup recorder system was recorded, and is shown in Table 4.12-1.

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:

OCO = (M x ICO) + B EQ. (4.5-2)

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

ICO = Incore Offset (Percent)

M = Slope of Relationship B = Intercept, when Incore Offset = 0 The experimental slope and intercept could then be obtained using a linear least squares fit from the data obtained. If the measured slope of the relationship of ICO to OC0 is outside the range of 0.96 to 1.10, then the gain of the out-of-core power range detectors would be adjusted. The relationship of measured slope to gain factor is as follows:

GF = (M2/M1) x GF o EQ. (4.5.3)

Where: GF = Desired Gain Factor M2 = Desired Slope M1 = Measured Slope GF, = Present Gain Factor fPT-120,Section10.2.9,Rev.11 ,

3PT-120, Section 10.2.10, Rev. 11 PT-120, Sections 10.2.3 and 10.2.2 44

Verification of the adequacy of the power imbalance system trip setpoint was performed in conjunction with the worst case analysis on each mini-mum DNBR and maximum LHR measured. Each measured point was extrapolated to the power / imbalance / flow envelope boundary limits given in Figure 4.7-1. In this way, the adequacy of the imbalance system trip setpoints to protect the unit from exceeding thermal-hydraulic limits could be verified.

I 4.12.2 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 APSRs and control rod group 6 and/or 7 at 72.9% 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 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 calcula-tions are tabulated in Table 4.12-2. In all cases, the measured slopes were in the range of 0.96 to 1.10, which verified the utilization of the gain factors used for the difference amplifiers. {

The abili ty 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 results cf this analysis are plotted in Figure 4.12-2, and show that the acceptance criterion was met.

During all phases of testing, worst case minimum ONBR and maximum LHR were recorded against incore offset. These results are given in Table 4.12-3 and Figure 4.12-3. The most limiting value observed for the maximum LHR and the worst case minimum DNBR was 10.63 kW/ft and 3.684, respectively, which is well within the procedural acceptance criteria.

45

i l 4.13 Nuclear Instrumentation Calibration At power Test L The Nuclear Instrumentation Calibration test was performed "to verify theabilitytoca1{bratepowerrangenuclearinstrumentationtomeasured core conditions". This test .was performed in accordance with l

procedures PT-120 and SP-113, and in conjunction with Surveillance Procedure SP-312.

The acceptance criteria for this test are:

The power range nuclear instrumentation (NI) is cal {brated to with-in !2% of the power calculated in the heat balance.

The higg level bistable trip is set to trip within the specified limits The absolute difference between the out-of-core detector axial power imbalance and the in-core detector axial power imbalance is less than 2.5%.4 Furthermore, the Technical Specifications require that the overlap between the interrrediate and power range nuclear instrumentation be in excess of one decade.

4.13.1 Me th o,d_

Reactor power was increased to the specified power level, while continuosly monitoring all the parameters that indicate power level change. A heat balance was then performed. Based on the results of the heat balance, the sensitivity of the linear a:nplifiers for each power '

range channel was adjusted, if ner.essary, and another heat balance was performed. This process continued until indicated power and heat balance power were within 2% of each other.

This test was performed at each of the major power plateaus; 75% FP and 95-100% FP. I 4.13.2 Results The results of the NI calibration met the acceptance criteria.

fFSAR, Table 13-4,#13 3

PT-120, Section 10.2.1, Rev. 11 4

SP-113, Section 9.3, Rev. 43 SP-113, Section 9.4, Rev. 43 46

4.14 Emercency Feedwater Flow Test No Emergency Feedwater Flow test was performed since no modifications were made which would invalidate the results of the tests performed during Cycle 2 startup.

4.15 Turbine / Generator Coeration No Turbine / Generator operational testing was performed since no changes were made to the turbine or generator which would invalidate the results of the test conducted during initial startup.

4.16 Droceed Control Rod Test The Dropped Rod test was not performed because sufficient thermal margin exists as indicated by B&W-2 correlation analyses.

4.17 Incore Detector Test The Incore Detector Test was performed to verify the adequacy of the system to orovide a description of core conditions. The test, performed in accordance with PT-120, is subject to the following acceptance criterion:

All detector outputs must be consistent and reasonable.1 4.17.1 Method The incore detector output was verified by comparing the corrected detector response from similar core locations. All detector outputs were normalized to the average detector output per assembly.

The values were taken from the Performance Data Output at 75% FP. From l this, the average level current for each detector, including background, I was found. Then, each current for each level was divided by the average current. This process was repeated for each detector string. The groupings in Table 4.17-1 were made based on symmetric or near-symmetric locations.

l 4.17.2 Results l The results of this comparison at 75% FP are shown in Table 4.17-1.

I PT-120, Section 10.2.6, Rev. 11 47 1

l

. 1 4.18 Reactor Coolant System Hot leakage Test The Reactor Coolant System (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 Pioe and Comoonent Hanger Hot Insoection at Power No Pipe and Component Hanger Ho't . 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 Radiochemical 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.

4.21 Effluent and Effluent Monitoring No Effluent and Effluent Monitoring testing was performed as these systems have been performing normally since initial startup. Therefore, no further testing was required.

48

TABLE 4.5-1 SUfEARY OF TEMPERATURE, MODERATOR, AND DOPPLER COEFFICIENTS OF REACTIVITY Full Power Escalation Testing ACCEPTANCE COEFFICIENT MEASURED CALCULATED REQUIREMENT

- mammmmmmemuss summmmmmmmmun ammmmmmmmmmmmmmmmmmma a c Temoerature -0.50048 -0.549 __

(10 3k/k/ F) 8 c Modegator -0.34947 -0.402 Less than zero, and (10 Ak/k/ F) less negative than

-3,0x10 ' 3k/k/ F Pael Teq. Doppler -1.564 b -1.51 d More negative than (10 ak/k/ F) -0.90x10

• 3k/k/ F a

Source: PT-120, Encl. 10 PT-120, Encl. 11 c

Physics Test Mcnual for CR3 C7, Table 12 Reload Report for CR3 C7, Table 5-1, BAW-1988 l

49

TABLE 4.7-1MAXIMUMLINEARHEATRATEBYINCOREDETECTORLEVEL Full Power Escalation Testing Maximum Linear Heat Rate In-Core (kW/ft)

Detector LOCA Limit Level 72.9% FP 100.0% FP (kW/ft)

-- mummmmmmmmmmmmus mummmmmmmmmmmmmu ammmmmmmmmmmummmmmmmmm 8 4.55 5.90 15.0 7 7.89 11.14 15.4 6 8.67 11.58 16.3 5 9.11 11.76 16.4 4 9.09 11.75 16.1 3 8.31 11.33 14.1 2 7.69 . 11.13 11.9 1 5.12 7.63 10.9 Source: SP-104, Encl. 1, Rev. 20 50

TABLE 4.12-1 SUW4ARY OF POWER IMBALANCE DETECTOR CORRELATION TEST 75% FP Full Power Escalation Testing Rod Position (% WD)* Incore Offset (%) Out-of-Core Offset (%)b Worst Case Worst Case Power Level" MLilR Minimum b i c d

(% FP) 1-6 7 8 Full Backup NI-5 NI-6 NI-7 NI-8 (kW/ft)d DNBR

- - - - - suussummmmmu ausmanns masammmy - - -- - - - -

72.86 100 87.04 52.26 +0.64 +1.06 +0.29 +0.39 +0.21 +0.21 9.76 4.161 72.91 100 87.47 48.90 +1.22 +1.90 +0.88 +0.96 40.80 +0.80 9.75 3.948 72.88 100 89.60 29.46 +6.63 +11.36 +6.08 +6.07 +5.73 +5.83 9.95 3.909 72.75 100 76.62 65.09 -5.27 -4.49 -6.50 -6.44 -6.34 -6.26 10.20 3.908 72.83 100 68.08 68.31 -13.09 -11.51 -14.47 -14.06 -14.15 -13.98 10.24 3.924 72.82 8

100 59.45 65.40 -21.63 -17.65 -22.33 -21.58 -21.73 -21.70 10.27 3.912 73.01 - - -

+12.91 +11.43 +11.36 +11.10 +11.04 +11.23 10.63 3.684 v.

Sources: ^ PT-120, Rev 11 PT-120, Encl. 6, Part 1, Rev. 11 l

l C

PT-120, Encl. 5, Rev. 11 d

PT-120, Er.cl. 4, Part 2, Rev. 11

^ Additional data point to meet the criterion on the sum of the squared differences between the average incore offset and the measured offset ( 2612 ).

l l

TABLE 4.12-2

SUMMARY

OF LEAST SQUARES LINEAR REGRESSION ANALYSIS OF POWER RANGE CHANNELS AND BACKUP RECORDER DURING POWER IMBALANCE CORRELATION TEST 75% FP Full Power Escalation Testing Intercept Slope (m)

Detector Data System Sets Calculated Allowable Calculated Allowable mmmmmme ammmmumumme summmmmmmmmmsca Out-of-Core" NI-5 7 -0.87 2.5 1.00 0.96fmfl.10 NI-6 7 -0.78 !2.5 0.97 0.965ms1.10 NI-7 7 -0.92 2.5 0.97 0.965mfl.10 NI-8 7 -0.83 2.5 0.97 0.96fmfl.10 b

Backup 7 1.31 0.92 a

Sources: PT-120, Encl. 6, Part 4, Rev. 11 PT-120, Encl. 7, Part 1, Rev. 11 C

52

TABLE 4.12-3 WORST CASE MINIMUM DNBR AND MAXIMUM LINEAR HEAT RATE VS. FULL INCORE OFFSET 75% FP Full Power Escalation Testing Worst Case Maximum LHR Worst Case Minimum DNGR Full Incore Offset (%) (kW/ft) Location Value Location immmumummmmmmus ammmmmmmmmmu asummmmmmmmmmme mummum mmmes ammmmmmmmmmmoss

-0.64 9.76 0-3 4.161 N-11

+1.22 9.75 D-3 3.948 D-3

+6.63 9.95 0-3 3.909 D-3

-5.27 10.20 D-3 3.908 D-3

-13.09 10.24 D-3 3.924 D-3

-21.63 10.27 D-3 3.912 0-4

+12.91 10.63 N-5 3.684 0-5 Source: PT-120, Encl. 4, Part 2, Rev. 11 53

l TABLE 4.17-1 COMPARISON OF INCORE MONITORED ASSEMBLIES' FLUX SHAPES 75% FP Full Power Escalation Testing Incore (Detector Level Current)/(Average Detector Current) Back-Group Detector Detector ground No. Location Level 1 Level 2 Level 3 Level 4 Level 5 Levei 6 Level 7 (amps) ammmmmmma m ammmmmmmu a- - umanummme m unisaamma =saiantums smsmemmes - saammensam 01 5 E-09 0.765 1.023 1.145 1.188 1.154 1.109 0.616 54 7 E-07 0.740 1.034 1.156 1.218 1.147 1.104 0.600 9 9 G-05 0.732 1.044 1.156 1.205 1.157 1.085 0.619 54 11 K-05 0.741 1.063 1.152 1.184 1.167 1.100 0.593 6 13 M-07 0.739 1.011 1.144 1.181 1.188 1.140 0.596 17 16 M-09 0.740 1.011 1.157 1.194 1.169 1.131 0.598 8 19 K-11 0.734 1.128 1.127 1.209 1.118 1.110 0.575 7 25 G-11 0.765 1.025 1.144 1.177 1.156 1.114 0.618 0 02 23 F-13 0.701 1.077 1.136 1.225 1.148 1.136 0.578 8 28 C-10 0.728 1.081 1.195 1.209 1.170 1.063 0.554 66 32 C-06 0.728 1.101 1.179. 1.202 1.167 1.056 0.568 65 ui

" 35 F-03 0.701 1.096 1.152 1.229 1.200 1.078 0.545 10 39 L-03 0.715 1.061 1.168 1.213 1.169 1.096 0.577 67 43 0-06 0.738 1.091 1.157 1.195 1.183 1.062 0.574 64 -

47 0-10 0.727 1.079 1.176 1.195 1.179 1.059 0.585 65 4 50 L-13 0.726 1.080 1.168 1.193 1.165 1.101 0.569 66 03 6 F-07 0.754 1.013 1.146 1.200 1.172 1.072 0.643 54 8 G-06 0.730 1.009 1.142 1.224 1.169 1.097 0.630 7 15 N-09 0.711 1.033 1.162 1.207 1.205 1.110 0.571 9 17 M-10 0.750 1.047 1.159 1.178 1.156 1.051 0.658 61 18 L-11 0.738 1.042 1.148 1.203 1.145 1.091 0.633 5 20 K-12 0.754 1.060 1.163 1.222 1.187 1.054 0.560 19 33 D-05 0.719 1.065 1.182 1.227 1.220 1.069 0.519 8 34 E-04 0.715 1.069 1.180 1.255 1.194 1.056 0.531 6 04 24 F-12 0.724 1.104 1.223 1.130 1.174 1.040 0.605 ~ 12 27 D-10 0.752 1.104 1.218 1.181 1.145 1.046 0.553 12 05 31 8-07 0.689 1.034 1.167 1.224 1.221 1.111 0.554 14 t

36 G-02 0.689 1.059 1.159 1.221 1.190 1.106 0.576 66

TABLE 4.17-1 COMPARISON OF INCORE MONITORED A5SEMBLIES' FLUX SHAPES l 75% fP Full Power Escalatinn Testing (Continued)

Incore (Detector Level Current)/(Average Detector Current) Back-Group Detector Detector ground No. Location Level 1 Level 2 Level 3 Level 4 Level 5 Level 6 Level 7 (amps) 06 38 L-02 0.694 1.0 4 1.168 1.217 1.204 1.076 0.566 0 44 P-06 0.681 1.034 1.160 1.220 1.208 1.082 0.614 11 07 22 G-13 0.733 1.037 1.151 1.183 1.204 1.118 0.575 69 29 C-09 0.693 1.018 1.206 1.224 1.194 1.089 0.576 13 08 40 M-03 0.719 1.043 1.176 1.195 1.188 1.105 0.574 0 42 0-05 0.718 1.044 1.179 1.198 1.187 1.098 0.575 8 09 1 H-08 0.623 1.053 1.142 1.253 1.165 1.115 0.648 15 2 11-09 0.749 1.049 1.122 1.197 1.142 1.096 0.646 51 3 G-09 0.730 1.004 1.162 1.211 1.151 1.108 0.634 10 us 10 4 F-08 0.711 1.026 1.167 1.196 1.175 1.079 0.645 8 10 H-05 0.708 1.012 1.158 1.252 1.225 1.042 0.605 18 12 L-06 0.738 1.032 1.140 1.197 1.141 1.079 0.674 45 14 N-08 0.742 1.050 1.199 1.257 1.221 1.101 0.429 10 11 37 H-01 0.606 1.016 1.161 1.204 1.218 1.211 0.583 0 45 R-07 0.653 1.039 1.142 1.147 1.227 1.204 0.588 3 51 0-14 0.652 1.023 1.195 1.219 1.235 1.135 0.541 4 52 C-13 0.627 1.068 1.222 1.240 1.269 1.093 0.479 8 12 26 E-11 0.821 1.054 1.175 195 1.166 1.041 0.549 9 30 B-08 0.685 1.033 1.170 1.206 1.211 1.138 0.556 13 41 N-04 0.756 1.099 1.213 1.223 1.221 1.114 0.377 66 13 21 11-13 0.734 1.027 1.168 1.206 1.198 1.108 0.561 10 46 R-10 0.600 1.012 1.148 1.231 1.241 1.161 0.607 0 48 0-12 0.646 1.054 1.177 1.232 1.244 1.123 0.526 3 49 M-14 0.665 1.058 1.177 1.197 1.239 1.117 0.547 3 Source: Group 1, 75% FP BKGD1, SPDNR1

Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance Full Power Escalation Testing

(-17,108). --

_ (17,108)

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COMPAR! SON OF MEASURED AND PREDlCTED RADlAL POWER PEAK!NG FACTORS WITH 30 EQU!LlBRlUM XENON AT 75% FULL POWER Full Power Escalation Testing Measured Predicted Control Rod Group Positions Groups 1-6 100.0% WO 100.0% WD Group 7 90.4% WD 93.3% WD Group 8 47.1% WD 41.0% WD Core Power Level 72.9% FP 75% FP Boron Concentration 1574 ppm -

Core Burnup 3.3 EFPD 3.0 EFPD Axial Imbalance 0.53% FP -1.46% FP l Maximum Quadrant Tilt 1.61% -

l H K L M N O P R 0.86 8 0.92 6.97 1.14 0.92 9 1.17 0.95 2.63 3.26 0.92 1.19 0.88 10 0.95 1.18 0.92 3.26 -0.80 4.54 1.26 0.95 1.24 1.12 11 1.20 0.98 1.21 1.17

-4.70 3,15 -2.40 4.46 0.90 1.31 0.93 1.34 1.14 12 0.96 1.26 0.97 1.28 1.11 6.66 -3.80 4.30 -4.40 -2.60 1.30 1.15 1.30 1.10 1.13 0.45 13 1.30 1.19 1.28 1.10 1.10 0.47 Source: PT-120, Encl. 2, 0.00 3.47 -1.50 0.00 -2.60 4.44 Parts 1 & 2 1.09 1.24 0.91 0.90 0.46 14 1.08 1.22 0.94 0.89 0.47

-0.90 -1.60 3.29 -1.10 2.17 x.xx Measured Results x.xxx Predicted Results 0.79 0.54 0.35 x.xx % Deviation 15 0.77 0.53 0.37

-2.50 -1.80 5.71

% Deviation = Predicted - Measured Measured x 100 Figure 4.7-2 57

COMPARISCN OF MEASURED AND PREDICTEC MAXIMUM TOTAL POWER PEAKING FACTORS WITH 3D EQUILIBRIUM XENON AT 75% FULL POWER Full Power Escalation Testing s

Measured Predicted Control Rod Group Positions Groups 1-6 100.0% WD 100.0% WD Group 7 90.4% WD 93.3% WD Group 8 47.1% WD 41.0% WD Core Power Level 72.9% FP 75% FP Boron Concer.tration 1574 ppm -

Core Burnup 3.3 EFPD 3.0 EFPD Axial Imbalance 0.53% FP -1.46% FP Maximum Quadrant Tilt 1.61% -

H K L M N O P R 1.05 8 1.09 3.81 1.36 1.09 9 1.39 1.12 2.21 2.75 1.08 1.44 1.04 10 1.13 1.40 1.09 4.63 -2.78 4.81 1.56 1.12 1.48 1.32 11 1.44 1.17 1.45 1.40

-7.69 +4.46 -2.03 6.06 1.11 1.58 1.11 1.65 1.38 12 1.15 1.52 1.18 1.55 1.36 3.60 -3.80 6.31 -6.06 -1.45 1.56 1.37 1.56 1.30 1.39 0.56 13 1.58 1.43 1.55 1.33 1.35 0.58 Source: PT-120, Encl. 2, 1.28 4.38 -0.64 2.31 -2.88 3.57 Parts 1 & 3 1,32 1.50 1.10 1.11 0.56 14 1.31 1.49 1.14 1.09 0.57

-0.76 -0.67 3.64 -1.80 1.79 x.xx Measured Results x.xx Predicted Results 0.96 0.65 0.43 x.xx % Deviation 15 0.94 0.65 0.45

-2.08 0.00 4.65

% Deviation = Predicted - Measured Measured x 100 Figure 4.7-3 58

COMPARISON OF MEASURED AND PREDICTED RADIAL POWER PEAKING FACTORS WITH 30 EQUILIBRIUM XENON AT 100% FULL POWER Full Power Escalation Testing Measured Predicted Control Rod Group Positions Groups 1-6 100.0% WD 100.0% WD Group 7 90.9% WD 90.1% WD Group 8 34.9% WD 27.9% WD Core Power Level 100.0% FP 100% FP Boron Concentration 1486 ppm 1541 ppm Core Burnup 6.6 EFPD 4.0 EFPD Axial Imbalance -1.69% FP -3.12% FP Maximum Quadrant Tilt 1.25% 0.00%

H K L M N O P R 0.86 8 0.93 7.53 1.15 0.92 9 1.17 0.96 1.74 4.35 0.92 1.18 0.89 10 0.96 1.18 0.93 4.35 0.00 4.49 1.24 0.94 1.23 1.13 11 1.19 0.98 1.21 1.17

-4.03 4.26 -1.63 3.54 0.90 1.29 0.94 1.34 l'12 12 0.95 1.25 0.98 1.27 1.10 5.55 -3.10 4.26 -5.22 -1.79 1.31 1.16 1.28 1.10 1.14 0.46 13 1.29 1.19 1.27 1.10 1.09 0.48 Source: PT-120, Encl. 2,

-1.53 2.59 -0.78 0.00 -4.39 4.35 Parts 1 & 2 1.10 1.23 0.92 0.92 0.48 14 1.08 1.22 0.95 0.90 0.48

-1.82 -0.81 3.26 -2.17 0.00 x.xx Measured Results x.xx Predicted Results 0.79 0.55 0.37 x.xx  % Deviation 15 0.77 0.54 0.38

-2.53 -1.82 2.70 .

% Deviation = Predicted - Measured Measured

-x 100 Figure 4.7-4 59 l

COMPARISON OF MEASURED AND PREDICTED MAXIMUM TOTAL POWER PEAKING FACTORS WITH 30 EQUILIBRIUM XENON AT 100% FULL POWER Full Power Escalation Testing Measured Predicted Control Rod Group Positions Groups 1-6 100.0% WO 100.0% WO Group 7 90.9% WD 90.1% WO Group 8 .

34.9% WO 27.9% WO Core Power Level 100.0% FP 100.0% FP Boron Concentration 1486 ppm 1541 ppm Core Burnup 6.6 EFPD 4.0 EFPO Axial Imbalance -1.69% FP -3.12% FP Maximum Quadrant Tilt 1.25% 0.00%

H K L M N O P R 1.03 8 1.10 6.80 1.33 1.07 9 1.38 1.12 3.76 4.67 1.06 1.39 1.02 10 1.13 1.40 1.09 6.60 0.72 6.86 1.50 1.07 1.41 1.29 11 1.43 1.16 1.42 1.39

-4.67 8.41 0.71 7.75 1.08 1.51 1.09 1.59 1.31 12 1.15 1.49 1.11 1.53 1.35 6.48 -1.32 1.83 -3.77 3.05 1.53 1.33 1.49 1.26 1.35 0.56 13 1.56 1.42 1.52 1.32 1.34 0.58 Source: PT-120, Encl. 2, 1.96 6.77 2.01 4.76 -0.74 3.57 Parts 1 & 3 1.28 1.45 1.07 1.08 0.56 14 1.30 1.49 1.14 1.09 0.58 1.56 2.76 6.54 0.93 3.57 x.xx Measured Results x.xx Predicted Results 0.94 0.64 0.44 x.xx  % Deviation 15 0.95 0.65 0.46 1.06 1.56 4.54 *

% Deviation = predicted - Measured x 100 Measured Figure 4.7-5 60

HOT CHANNEL MINIMUM CNBR VS. CORE POWER LEVEL Full Power Escalation Testing 25 .

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Florida

.P. o. . .w. .e. .r.

Walter S.Wilgus Vice President Nuclear Operations May 3, 1988 3F0588-02 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Cycle Seven Startup Report l

Dear Sir:

1 Florida Power Corporation (FPC) hereby submits the Crystal River Unit 3 Cycle Seven Startup Report. This report is submitted in accordance with Technical Specification 6.9.1.3 and Regulatory Guide 10.1, Item 170.  ;

If you have any quest' ions concerning this report, please contact this office.

Sincerely, W.S. W' rus, Vice President Nuclear Operations KRW/dhd Attachment xc: Dr. J. Nelson Grace Regional Administrator, Region II Mr. T.F. Stetka Senior Resident Inspector 4>9 s/

' \

(

f 3201 Thirtt ourth Street South P.O. Box 14042,St Petersburg, Florida 33733 813 866 5202

_ _ _ _ _ _ _ _ _ _ _ _ _ _