ML19329A928
| ML19329A928 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/08/1978 |
| From: | TOLEDO EDISON CO. |
| To: | |
| References | |
| NUDOCS 8001150832 | |
| Download: ML19329A928 (115) | |
Text
{{#Wiki_filter:50-346 DAVIS BESSE INITIAL STARTUP REPORT w/ltr.5-6-78. 781320039 s 0
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NOTICE - ! l l 1 (
'HE ATT ACHE D FILES ARE OF FICI AL RECORDS OF THE ! l vlVISION OF DOCUMEN T CON TROL THE Y H AVE BEEN i CHARGED TO YOU FOR A LIMITED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS F ACILIT Y BRANCH 016. PLEASE DO NOT SEND DOCUM E N TS j CH ARGED OUT THROUGH THE M AIL REMOV AL OF ANY P AGE (S) FROM DOCUMENT FOR REPRODUCTION MUST [
BE REFERRED TO FILE PERSONNE L I l' 1 OE ADuNE RETURN Dm l l
. . , * > 'e JQ,l--,- J ; 7 -4* l I
I RECORDS F ACILITY BR ANCH 1 O' ') l
l i t DAVIS-BESSE NUCLEAR POWER STATION UNIT ONE INITIAL STARIUP REPORT
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COVERING APRIL 23, 1977 THROUGH APRIL 5, 1978 SUPPLEMEIT 1 COVERING APRIL 5,1978 THROUGH JULT 5,1978 SUPPLDENT .2 COVERING JULY 5,1978 THROUGH OCTOBER 5,1978 SUPPLEMENT 3 COVERING OCI0BER 6,1978 THROUGH JANUARY 5,1979 ; SUPPLEMENT 4 COVERING JANUARY 6,1979 THROUGH APRIL 5,1979 ) SUPPLEMENT 5 COVERING APRIL 6,1979 THROUGH JULY 5,1979 ' SUPPLEMENT 6 COVERING JULY 6, 1979 THROUGH OCTOBER 5, 1979 )
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< TABLE OF CONTENTS Section Page,
1.0 INTRODUCTION
1-1 2.0 SLHMARY 2-1
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3.0 INITIAL FUEL LOADING 3-1 ! 4.0 POST FUEL LOAD PRECRITICAL HOT FUNCTIONAL TESTING 4-1 4.1 Reactor Coolant System Flow Measurement 4-1 4.2 Reactor Coolant System Flow Coastdown Measurement - 4-2 4.3 Reactor Coolant System Hot Leakage Test 4-2 4.4 Pressurizar Operational and Spray Flow Tests 4-3 4.5 Control Rod Drive System Operational Test 4-3 5.0 INITIAL CRITICALITY 5-1 5.1 Preliminary Approach to Criticality 5-1 5.2 Final Approach to Criticality 5-1 6.0 COREPERFORMANCEDURINGZEROPOWERPNYSICSTESTS 6-1 6.1 Nuclear Instrument Overlap _ 6-1 6.2 Sensible Heat Determination 6-1 6.3 Reactimeter Response Cher in 6-2
. 6.4 All Rods Out Boron Concentration 6-2 6.5 Temparcture Coefficient of Reactivity Measurements ,
6-3 6.6 Control Rod Reactivity Worth Measurements 6-3 6.7 Ejected Rod Worth Measurements 6-4
' 6. 8' Stuck Rod Worth and Shutdown Margin Measurements 6-4 6.9 Soluble Poison Worth Measurements 6-6 e
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Section Page, 7.0 CORE PERFORMANCE DURING POWER ESCALATION SEQUENCE TESTS 7-1 7.1 Nuclear Instrumentation Calibration at Power 7-1 b . 7.2 Reactivity Coefficients at Power 7-3 7.3 Rod Worth at Power 7-5 7.4 Core Pouer Distribution Tests 7-6 7.5 Pseudo Control Rod Ejection Test 7-6 7.6 Dropped Cont'rol Rod Test 7-7 7.7 Incore Detector Test 7-8 7.8 Power Imbalance Detector Correlation Test 7-9 8.0 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) PERFORMANCE 8-1 8.1 Unit Load Steady State. Test 8-1 8.2 NSSS Heat Balance Test 8-1 8.3 Integracad Control System Tuning at Power 8-2 9.0 UNIT PERFORMANCE DURING TRMSIENT AND ABNORMAL CONDITIONS 9-1 -
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9.1 Turbine / Reactor Trip Test 9-1 9.2 Unit Load Transient Test 9-2 9.3 Unit Power Shutdown Test 9-3 9.4' Unit Load Rejection Test 9-3 l l 9-4 9.5 Natural Circulation Test I 9.6 Loss of Offsite Power Test 9-4 l 9.7 Shutdown From Outside the Control Room 9-5 l
- l 10.0 SECONDARY PLANT PERFORMANCE AND STARTUP. EXPERIENCE 10-1 i 10.1 Turbine / Generator- 10-1 10.2 Condenser 10-2 10.3 Circulating Water System 10-3
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10.4 Peedwater Systems 10-4 ; l 11.0 UNIT MONITORING - CHEMISTRY AND HEALTH PHYSICS 11-1 11.1- Shield Survey 11-2
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Section h 11.2 Site / Station Survey 11-3 11.3 Reactor Coolant chemistry Test 11-3 11.4 Steam Generator Chemistry Test 11-4 11.5 Initial Radiochemistry Test 11-4 . 11.6 Process Area Radiation Monitoring Test 11-5 12.0 UNSCHEDULED UNIT TRIPS 12-1 13.0 CORE PERFORMANCE FOLLOWING BPRA AND ORA REMOVAL 13-1 13.1 Core Performance During Zero Power Testing 13-1 13.2 Core Performance During Power Escala. tion Testing 13-6 I i
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1.0 INTRODUCTION
i Davis-Basse Nuclear Power Station (DBNPS) Unit 1, located on the
.i southwestern shore of Lake Erie near Oak Harbor, Ohio, is a ,
, Babcock and'Wilcox pressurized water reactor rated at 2,772 MWt.
The turbine-generator is capable of a net electrical output of 906 l
- MWe. The Nuclear Steam Supply System (NSSS) employs once through steam generators.
t - The Facility Operating License (NPF-3) for DBNPS JJnit I was issued
~
. to the Toledo Edison Company on April 22, 1977. The first fuel assembly was loaded into the core on April 23, 1977, and fuel loading was completed on April 27, 1971, after.a total fuel load time of -
I 83 hours. Initial criticality was achieved on August 12, 1977, after i a Post Fael Load Precritical Hot Functional Test Program.
~2ero power physics testing commenced af ter achieving initial criti- !
p6 cality on August 12, 1977, and was completed on August 20, 1977. The zero power measurements of core performance were performed at a Reactor Coolant System temperature of 5300F, and a pressure of 2155 psi. ' l Power escalation constanced on August 24, 1977, and the turbine gen- ; erator was initially loaded on August 28, 1977. Further power level increases were successfully completed at each of the four major power level plateaus as defined by the Power Escalation Sequence Test Procedure. The four major power level plateaus and dates attained are as follows: - Power Level _Date 15% September 2, 1977 40% November 14, 1977 75% December 21, 1977 100% _,, _,._, April 4, 1978 ,, ,, _
^
Figures 1.0-1 through 1.0-12 show the chronological power history during the startup test program. Figures 1.1-1 through 1.1-8 show the chronolo-Sical core burnup during the startup test program. The initial transmittal on May 8,1978, of the Startup Report contained test data which summarized the startup test program and unit performance from initial fuel loading on April 23, 1977, through 100% full power opera-tion on April 5, 1978. Since the power escalation program was not com-
, placed by April 5,1978, it could not be included in the initial trans-mittal. -
Technical Specification 6.9.1.3 requires supplemental reports be submitted to the Startup Report on a quarterly basis until testing is completed and
.j- the unit resumes commercial power operation. Davis-Besse Unit 1 attained 4 commercial power operation on November 22, 1977. Davis-Besse Unit 1 was shucdown for. a maintenance outage and, therefore, no further testing was completed in the period from April 5,1978 through July 5,1978.
1-1 E 4 da
The second supplement updated the Startup Report to contain test results of testing completed between July 5,1978 through October 5,1978.
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The third supplement updated the Startup Report to contain the test results of testing completed between October 6,1978 through January 5,1979. The fourth supplement updated the Startup Report to contain test results of testing completed between January 6, 1979 through April 5, 1979.
- Since the unit was shutdown from March 31, 1979 to July 11, 1979, no further testing was completed. Therefore, only Chapter 1 was revised by Supplement 5 which covers from April 6,1979 to July 5,1979.
The sixth supplement updated the Startup Report to contain the test results of 6 testing complaced between July 6,1979 through October 5,1979. As this com-plates all testing, this is the final supplement. ThechangesmadetotheStartupReportbysupolementsbeindicatedhva vertical line in the lef t margin with a number to indicate by which supple-ment the revision was incorporated. C s t e l l i. s 1-2
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1 DBNPS - Unit #1 STARTUP REPORT l l 1978 EFPD HISTORY FIGURE 1.1-3 i i 1-19 I i I t
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7o JUI.Y I AUGUST l SEPTEMBER } OCI0BER DBNPS - Unit #1
. STARTUP REPOKr 1978 EFPD HISTORY FIGURE 1.1-4 1-20
O D D coE n D
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'
- I NOVEMBER
- DECEMBER JMUARY DBNPS - Unit #1
! STARTUP REPORT FIGURE 1.1-5 1 .
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JANUARY FEBRUARY -r7_.. . . JARG r l DBNPS - Unit #1
. STARTUP REPORT 1-22 4 4 ,,
FIGURE 1.1-6 o J*8* *' h _ _
D 0 oo ~ m' y 1979 EFPD HIST 6hY D A F7.GURE 1.1--7 i,
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- DBNPS - Unic #1 STARTUP REPORT 1 23 FIGURE 1.1-7 , -.. - . _ . . _ _ . . _ M' # '_ .Mh ig g .p ,
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210 - JULY I AUGUST l SEPTEMBER l OCTOBER t t DBNPS - Unit #1
. STARTUP REPORT FIGURE 1.1-8 1-24
. ~_, . _. . . _ _ .. . .._ . ... . _ _ _ _ _ _ _ _ _ _ _ , , , ,
'-w
2.0
SUMMARY
The unit has been operated at power levels up to and including 100% , full power since the completion of startup testing. The performance of the unit has generally been satisfactory. Testing and operation , of the NSSS and the turbine generator revealed some minor problems / conditions that were other than predicted, however, none of them adversely affected plant safety. The problems encountered were not unusual for the startup program of a unit this size. 4 A significant probles at a similar reactor did arise during power escalation that could affect Davis-Besse Unit 1. Two burnable poison rod assemblies (BPRA) were found outside of their fuel assen-blies at Florida Power Corporation Crystal River Unit 3 reactor. This initiated an investigation by the reactor vendor for both Crystal River Unit 3 and Davis-Besse Unit 1 Babcock and Wilcex. On April 5,1978, Toledo Edison personnel were notified a possible design deficiency could allow wear in the BPRA locking mechanism especially under high reactor coolant flow conditions. Although Babcock & Wilcox personnel felt the chance of such a failure due to wear during the first fuel cycle was extremely remote, they l } . recosamended, as a precautionary measure, the reactor coolant flow , be reduced. Reactor Coolant Pump 1-1 was shutdown on April 5, 1978. ' No BPRA locking mechanism failures have occurred at Davis-Besse, nor in five previous Babcock and Wilcox units using the same BPRA lock-ing mechanisms. All 68 BPRA and all 48 orifice rod assemblies were
~
i removed from the core by May 27, 1978 during the maintenance outage j as recommended by Babcock and Wilcox to insure 'no failures of the
~
lockius mechanism at Davis-Besse'. Modified orith e rod assemblies for the two neutron source holddowns were installet. l 2.1 INITIAL FUEL LOADING (SECTION 3.0) Initial fuel loading commenced on April 23, 1977 at 1357 hours. The : entire fuel loading sequence experienced only minor delays and was ' 6 accoglished .in approximately three and one half (31s) days. I J 2.2 POST FUEL LOAD PRECRITICAL HOT FUNCTIONAL TESTING (SECTION 4.0) Following initial fuel loading and prior to initial criticality, a Post' Fuel Load Precritical Hot Functional. Test Program was conducted l from July 2, 1977 to August 10, 1977. This testieg included a Reactor F Coolant System Flow Measutement, Reactor Coolant System Flow Coast-down, Pressurizer Operational and Spray Flow Test, and Control Rod Drive System Operational Test. All test results satisfied the Davis-
- Besse Unit 1 Technical Specifications and all test acceptance e criteria were met. The tests completed were, i (a) Reactor Coolant System Flow Measurement, TP 200.11 (b)' Reactor Coolant System Flow Coastdown Measurement, TP 200.11 1 (c) Pressurizer Operational and Spray Flow Tests, TP 600.13 I
- l. (d) Control Rod Drive System Operational Test, TP 600.17 l (e) Reactor Cociant System Hot Leakage Test, TP 600.10 (ST 5042.02) l- .
i 2 ___ _ __ _ _ _ _ _ . _ _ . d
l 2.3 INITIAL CRITICALITY, TP 710.01 (SECTION 5) Initial criticality was achieved at 1729 hours on August 12, 1977, at reactor conditions of 5300F and 2155 psig. Control Rod Groups 1 through 5 and 8 were withdrawn to the top lim.- (100%) and com- ; bined Groups 6/7 vere withdrawn to the 75% position. Criticality ' was then achieved by deborating from an initial reactor coolant boron concentrar. ion of 1843 ppm to a final concentration of 1520 ppm. I 2.4 CORE PERFORENCE DURING ZERO POWER PHYSICS TESTS, TP 710.01 (SECTIOM 6) l Following initial critical"ity, core performance measurements were conducted during the Zero Power Physics Test Program from August 12, 1977 to August 20, 1977. All test data and results satisfied Davis-Besse Unit 1 Technical Specifications and test acceptance criteria. The following parameters were verified: (a) Nuclear Instrumentation overlap (b) Sensible Heat Power Level (c) Reactimeter Response Checkout (d) All Rods Out Boron Concentration (e) Temperature Coefficient of Reactivity Measurements (f) Control Rod Reactivity Worth Measurementr-(g) Ejected Rod Worth Measurements (h) Stuck Rod Worth and Shutdova Margin Measurements (1) Soluble Poison Worth Measurements 2.5 CORE PERFORENCE DURING POWER ESCALATION SEQUENCE TESTS, TP 800.00 (SECTION 7.0) Core performance measurements were conducted during the Power Esca-lation Secuence T=st Program. Testing was conducted at the power level plateaus of 15%, 40%, and 75% of total thermal core power. All test data and results satisfied the Davis-Besse Unit 1 Technical Specifications and test acceptance criteria. The power escalation core performance data and measurements are contained in the following tests. 3 f (a) Nuclear Instrumentation Calibration at Power, TP 800.02 (b) Reactivity Coefficients at Power, TP 800.05 (c) Rod Reactivity Worth Test, TP 800.20 (d) Core Power Distribution Test, TP 800.11 (e) Pseudo Control Rod Ejection Test, TP 800.28 (f) Dropped Control Rod Test, TP 800.29 (g) Incore Detector Test, TP 800.24 (h) Power Imbalance Detector Correlation Test, TP 800.18 2-2
I l 2.6 NUCLEAR STEAM T'JPPLY SYSTEM (NSSS) PERFORMANCE (SECTION 8.0) A list of the tests performed during power operation related to the monitoring of the NSSS performing is presented below. In all, the performance of the NSSS was satisfactory, and as expected. (a) Unit Load Steady State Test, TP 800.12 , (b) NSSS Heat Balance Test, TP 800.22 ; (c) Integrated Control Syscam (ICS) Tuning at Power, TP 800.08 2.7 UNIT PERFORMANCE DURING TRANSIENI AND ABNORMAL CONDITIONS (SECTION 9) The purpose of the unit performance tests is to verify the unit can be maintained in a safe condition during and following load tran-sients and various abnormal conditions. In all, unit response to the following load transients and abnormal conditions was satisfac-6 tory.
~
('s) Unit Load Transient Test, TP 800.23 s (b) Unit Power Shutdown Test, TP 800.15 (c) Turbine / Reactor Trip Test, TP 800.14 (d) Loss of Offsite Power, TP 800.26 (e) Unit Load Rejection Test, TP 800.13 (f) Shutdown Frce Outside of the Contrei Room, TP 800.25 (g) RCS Natural Circulation Test, TP 600.04 , 2.8 SECONDARY PLANT PERFORMANCE (SECTION 10) This section provides a brief summary of the major difficulties encountered with the secondary systems during power escalation. 4 The secondary systems that are covered include: (a) Turbine-Generator (b) Condenser (c) Circulating Water System (d) Feedvater System ,
- 2.9 UNIT MONITORING (CHEMISTRY AND HEALTH PHYSICS) (SECTION 11)
This section presents a list of the unit monitoring and testing per-formed with regard to health physics and chemistry during various phases of the startup test program. Tests were conducted during initial fuel loading, reactor startup, power escalation and power operation. (a) Shield Survey, TP 800.01 (b) Site / Station Radiation Survey, TP 800.03 (c) Reactor Coolant Chemistry Test, TP 500.01 (d) Steam Generator Chemistry Test, TP 500.02 (e) Initial Radiochemistry Test, TP 500.03 6 (f) Process and Area Radiation Monitoring System , Pre-op Test, TP 360.01 2-3 e , 4
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2.10 UNIT TRIPS (SECTION 12) I l
,' This section is a listing of all unit trips and applicable information !
which occurred during the period from initial fuel loading through power escalation and operation. 2.11 CORE PERFORMANCE FOLLOWING BPRA'S AND ORA'S REMOVAL (SECTION 13) This section contains a description of physics' testing performed after the outage which removed the BPRA's and ORA's. Testing described includes: 3 - (a) " Post Refueling Physics Testing", ST 5010.03 (b) " Reactivity Coefficients at Power", TP 800.05 (c) " Core Power Distribution", TP 800.11 (d) " Rod Reactivity Worth Measurements", TP 800.20
, (e) "Incore Detector Test", TP 800.24 (f) " Pseudo Ejected Rod Test", TP 800.28 (g) "NI Calibration at Power", TP 800.02 (h) " Power Labalance Detector Correlation Test", TP 800.18 w
l [ l f \. 2-4
3.0 INITIAL FUEL LOADING 2 The Davis-Besse Unit 1 initial core contained 177 fuel assemblies, 53 control rod assemblics (CBA), 8 axial power shaping rod assemblies (APSRA), 68 burnabic poison rod assemblies (BPRA), 48 orifice red assemblies (ORA), and two installed neutron source assemblies. Further details and descriptions of the reactor core and components can be found in Chapter 4 of the FSAR. The initial fuel leading commenced when the first fuel assecbly (NJ004U) was removed from spent fuel pool location A-1 at 1357 on Ap'ril 23, 1977, and was completed when the final fuel asse=bly (NJ004C) was loaded into core position L-1 at 1808 on April 27, 1977. Actual fuel loading ti=e was approxi=ately 83 hours. The fuel loading was perforced in accordance with " Initial Fuel Load Procedure", PP 1302.04. The actual fuel loading sequence is illustrated 2 in Figures 3.0-1 through 3.0-8. The initial core configuration is shown in l Figure 3.0-9. The neutron count rate was monitored con;inuously and 1/M plots were maintained throughout fuel loading for both source range detectors (NI-i and NI-2) and both auxiliary neutron detectors (A and B). At least two independent 1/M calculations were completed for each asse=bly (one in the containment using an auxiliary neutron monitor and one in the Control Room using a source range detector) . Af ter each covecent of an a"- ;iary neutron monitor, the 1/M plots were either renormalized or a new baseline was obtained. The resulting 1/M plots were as expected and are su=narized in Figures 3.0-10 through 3.0-13. During fuel loading, several minor problems were encountered.' A description of the spccific problems and their resolution is given below:
- 1. .On April 23 at 1710, the west transfer techanism became stuck between containment and the spent fuel pool with fuel assembly NJ0048 in the carriage. The assembly and carriage were pulled to the spent fuel pool via the crane. After transferring the asse=bly from the west mechanism to the east, refueling recoc=enced using only the east transfer mechanism. The fuel handling area canal and the deep end of the refueling canal were drained to the top of the transfer tubes and the west transfer cechanism was repaired. The west transfer mech-anism was out-of-service for approximately 24 hours.
- 2. 'The clutch on the east transfer cechanism appeared to be slipping.
While the clutch was being repaired, only the west trans'er carriage was used. The east carriage was out-of-service for appro.<.mately 2 hours.
- 3. During fuel movecent, the hydraulic pump on the main bridge stopped twice and had to be restarted. Total delay was approxi=acely one hour.
- 4. Appeared to be having problems with the overload cutoff on the main bridge. On April 26 at 0110, fuel loading was suspended in order to test the bridge. The bridge was verified to be functioning properly.
3-1 .
It was determined that the upender was out of plumb. Both upenders were - replumbed and fuel loading recommenced.after a delay of approximately 3.5 hours. *
- 5. Near the end of the fuel loading, the east transfer mechanism hung-up again.- The west carriage was used exclusively for the remaining 7 hours of fuel loading.
- The Davis-Besse Unit 1 Technical Specifications state that during fuel movement, the boron concentration shall be such to ensure that the more restrictive of the following conditions are met:
- 1. Either a Keff of 0.95 or less which includes a 1% 4 K/K conservative allowance for uncertainties, or
- 2. A boron concentration of 2: 1800 ppm, which includes a 50. ppm conservative i- allowance for un~ certainties.
During initial fuel loading the second condition was the more restrictive. The measured reactor coolant boron oncentration ranged from 1829 to 1929 ppm during fuel movement. The initial boron measurement made within 12 hours of loading the first fuel assembly was 1888 ppm at 0720 on April 23. The overall average of the 8 boron measurements taken during fuel loading was 1883 ppm. .
- Following the completion of fuel loading, the incore detectors were inserted into the core. . Difficulties were experienced while inserting t.o of the incores Detector 46 could not be inserted into fuel assembly NJ0046, but could be e inserted into the dummy asse=bly. After a visual inspection of NJ0046, fuel element NJ0053 was reshuf fled from core location A-6 to core location R-10,
. while moving NJ0046 to core location A-6. Incore detector 46 was then insetted into NJ0053 with no difficulty.
f Detector 2, at core location H-9, was the other incore that experienced some problems while atte=pting to insert it into the core. After radiographing the.incore monitor tube, it was determined that one of the welds had excessive beads on the inside of the. cube. A wire was pushed through the tube to dislodge any loose obstructions. Detector 2 was then installed into assembly NJ002D with no additional difficulties. l l i L i L 3-2
'u _ _
7
.. o Figure 3.0-1 -
Davis-Besse Unit 1, Cycle I d Fuel Loading Sequence 11 of 8 N L K G F E D C B A R P O N M 11 1 NI-1 2 O 3 4
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l _ a I 7 B - 5 25b / 25a e 9
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Figure 3.0-2 Davis-Besse Unit 1, Cycle I L Fuel Loading Sequence'2 of 8 I N L K G F E D C B .. P. P O N M 11 1 b.
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~h-AuxiliaryNautronDetectorA Startup Report .
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. i Figure 3.0-3 Davis-Besse Unit 1, Cycle I Fuel Loadinsi "equence 3 of 8 j
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Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 4 of 8
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I"#" *~ h Attx111ary Neutron Detector B 3-6 N'
Figure 3.0-5 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 5 of 8
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7p/ws/% " o W E A W< 1 NI-2 15' Source Assemble . h Auxiliary Neutron Detector A . 'DBNPS - Unit 1 Startup Report , h Auxiliary Neutron Detcetor B Figure
Figure 3.0-6 Davis-Besse Unit 1, Cycle I Fuel 1.oading Sequence 6 of 8 G F E D. C B A R P O N M L K 11
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129 130 131 132l 133 134 135 136 137 , 138 139 140 1414 D3NPS - Unit 1 p Source Assemble Startup Report Fuel Loading Sequence h-Aur.iliary, cutron Detector A , Figure 3.0-6 . h- Auxiliary Neutron Detcetor B 3-8 .
3.0-7 Figure Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 7,of 8 N
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Figure 3.0--7 h Auxilia[y Neutron Detect.or B
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._ Fuel Assembly ID (preceded by "NJ00")
DBNPS - UNIT 1 Fuel Asse=bly Enrichment A = 1.98 wt. %U235 - STARTUP REPORT B = 2.63 wt. %U235 INITIALCORECONFIGURA C = 2.96 we. %U235 FIGURE 3.0-9 2lLBPEnrichment L = 1.09 wt. f3h0 M = 1.26 ut. 53hc 1 H = 1.43 wt. $3hc 3,11 j i
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4.0 POST FUEL LOAD PRTCRITICAL NOT FUNCIIONAL TESTING A Post Fuel Load Precritical Hot Functional Test Program was conducted following initial fuel loading. This section of the report presents the scope and results of that testing. Control rod drop times were obtained during the parformance of TP 0600.17, " Control Rod Drive System Operational Test." Measurements were taken at reactor coolant system conditions of approximately 265 F and 250 psig with 0 and 2 reactor coolant pumps in operation and at approximately 532 F and 2155 psig with 2 and 4 reactor coolant pumps running. Reactor coolant system flow and flow coastdown measuroments were con-ducted at reactor coolant system conditions of approximately 532*F l and 2155 psig to determine the core flow characteristics. RCS 1eakage measurements were performed to verify that the reactor coolant Pressurizer testing system leak rate was within acceptable limits. was also conducted at het conditions to adjust the spray and mini- , spray flow settings and to verify proper pressurizer heater and spray actuation setpoint,s. In all cases, the applicable test criteria and Davis-Besse Unit 1 Technical Specification requirements were met. 4.1 REACTOR COOLANT SYSTEM FLOW MEASUREMENT TP 200.11 . The Reactor Coolant System flow rates for various pump combinations were determined both before fuel loading and after fuel loading. Since the acceptance criteria applies only when the reactor core and l all 40 peripheral orifice rods are installed, the data acquired prior to fuel loading was not required to meet the acceptance values. Due to the relative insiSnificance of the pre-fuel load measurements, only the post-fuel load data is covered in this report. The RCS flow with all four reactor coolant pumps running simultane-ously was determined to verify that the total RCS flow was within the acceptable range. Likewise, the RCS flow rate with the three lowest flow pumps running simultaneously and the RCS flow rate with the lowest flow pump in each loop running simultaneously were de- . l termined to verify that the minimum flow requirement for three pump and two pump operation respectively were surpassed. After th,e BPRA 3 and ORA were removed, the RCS flowrate with all four RCPs in operation was re-determined. The acceptable criteria were again verified. The results of the flow measurements are compa red with their respective l acceptance criteria in Table 4.1-1. As shown, all measurements were j l well within their appropriate linits. TP 200.11 4.2 REACTOR COOLANT SYSTEM FLOW COASTDOWN MEASUREMENT RCS flow coastdown measurements were obtained prior to fuel loading and l again after the core was loaded. For the reason mentioned in Section l i 4.1, only the oost fuel load measurements are discussed in this report. I The flow coastdown test for a trip of one of four RCPs was repeated l= ' after the removal of the BPRAs and ORAs. i- l 4-1 l
With a RCS pressure of 2155 + 30 psig and a cold leg temperature of
-530 + 10*F, the following pump trips were initiated.
Casa Pumps Initially Ritnning Pumps Tripped 1 All four pumps Highest flow pump 2 Three lowest flow pumps Pump with highest flow in loop with 2 running pumps 3 All four pumps Highest flow pump in each loop 4 All four pumps All four pumps Prior to tripping a given pump combination, equilib'rium conditions were established and steady state flows were recorded with the computer line printer, brush recorders, and the reactir.eter. Following each trip, the resultant flow transient continued to be monitored on the recording devices mentioned above. The flow coastdown of each trip combination is compared with the appropriate acceptance criteria on 3l Figures 4.2-1 through 4.2-5. For each case, the coastdown flow exceeded.the limiting race. 4.3 REACTOR COOLANT SYSTEM HOT LEAKAGE TEST TP 600.10 The Reactor Coolant System (RCS) Hot Leakage Test and measurements were performed to accomplish the following: A) Determine the RCS leakage.
- B) Determine the accuracy of the RCS leakage measurement by imposing a simulated leak.
C) Verify that the RCS leakage is within the Davis-Besse Unit 1 Technical Specification limit. j D) Verify the adequacy of the RCS Water Inventory surveillance test. The RCS het leakage and s veillance test procedures were performed during the hot functional 2st program. RCS conditions were main-tained as steady as possible at about 532'T and 2155 psig throughout the test. l During the initial portion of the testing (prior to fuel load), the RCS leak rate was determined by performing TP 600.10 and ST $042.02, RCS Water Inventory Balance. Results indicated a total leak rate of less __ _ than 0.02 gpm. This calculated leak rate is well within the Davis-l Besse Unit 1 Technical Specification limit of 1.0 gpm unidentified
- leakage from the RCS.
A simulated leak rate of 1.0 gpm was then established through the leak test valves for the seal return isolation valve. Measurements were then taken in order to calculate the total leak rate. The calculated 4-2
[ f leak rate was then adjusted for the unidentified RCS leakage measured previously. The. resultant calculated value for the simulated leak rate was 0.95 gpa. This was in close agreement with che 0.91 spa (average over the test interval) simulated leak rate.. The percent deviation in the measured and actual simulated leak rate was 4% which meets the acceptance criteria of 4. 35%. . ., After fuel load was completed, the RCS leakage was measured at least every 72 hours while in steady state conditions using ST 5042.02. RCS leakage during the post fuel load precritical hot functional testing never exceeded the Technical Specification limits of 1.0 gpa unidentified leakage. 4.4 PRESSURIZER OPERATIONAL AND SPRAY FLOW TEST TP 600.13
. Pressurizer operational testing was conducted af ter fuel loading and prior to initial criticelity. This included the setting and testing of the pressurizer spray flow and mini-spray flow, and the testing of the presaurizar spray and heater actuation setpoints. The testing and verification of the pressurizer level setpoints, level control, and heater interlocks were performed during pre-fuel load hot function-al testing.
The technique used to set pressurizer spray flows was based upon bal-ancing the heat input to and the heat losses from the pressurizer. The pressurizer spray flow valve RC2 was adjusted near 190 gpm at re-duced RCS temperature and pressure so that the heaters could maintain a nearly constant pressure and temperature. The actual spray flow was then calculated using the measured data and equation in Table 4.4-1. The spray flow was calculated to be 190 gym which satisfied the test acceptance criteria of 184 to 209 gpm. The pressurizar mini-spray flow valve RC 49 was adjusted near 1.0 gpm in a similar manner for the spray flow valve, and the actual flow was also calculated using the measured data and equation in Table 4.4-1. The mini-spray flow was calculated to be 1.6 gpm which satisfied the test acceptance criteria of .75 to 3.0 gym. The pressurizer spray and heater actuation setpoints were tested by varying RCS press.ure using the pressurizer heaters and spray yalves. The results of these measurements along with respective acceptance criteria are nununarized in Table 4.4-2. AJ1 recorded setpoint data met the test acceptancs criteria. 4.5 CONTROL ROD DRIVE SYSTEM OPERATIONAL TEST TP 600.17 Testing of the control rod drive system was performed during the hot functional testing program prior to fuel loading. These tests were performed en assure proper operation of the control rod drive mechanisma under actual thermal operating conditions. All test results were acceptable. l l l~ 4-3 j l l i
-_ - - . _ , , - - ,. . . , . _ _ . _ ~ , _ , , . . , . _.
4
In addition, control rod drop times were measured and verification of. . . . . control rod full insertions were performed after fuel loading, prior to initial criticality. These control rod drop time measurements were taken to ensure conpliance with the requirements of the Davis-Besse Unit 1 Technical Specifications and the assumption stipulated in the FSAR accident analysis. . The control rod drop time measurements were performed for all rods in groups 1 through 7 at RCS conditions of approximately 265 F and 250 psig with 0 and 2 reactor coolant pumps running, and at approxi-mately 532 F and 2155 psig with 2 and 4 RCP's in operation. The actual procedure and measurements were performed in accrodacce with ST 5013.02, " Control Rod Assembly Insertion Time Test." Each rod was withdrawn to its fully withdrawn position and dropped into the core using the auxiliary power supply trip C and D switches. Test data was tabulated on brush recorders. Time signals were furnished to the recorders to show the initiation of each trip and closure of the 25% reference switch for the individual rods.
- All rod drop times from the fully withdrawn position were 41.30 seconds from power interruption at the control rod drive cabinets to 3/4 insertion. This met the Davis-Besse Unit 1 Technical Specification limit of f.1.58 seconds. The accident analysis requirement of drop times, from the fully withdrawn position being 41.4 seconds from power interruptior to 2/3 insertion was also satisfied. The actual rod drop times for the four pump condition are summarized in Figure 4.5-1.
e i f a 4-4
REACTOR COOLANT SYSTEM FLOW MEASURDENTS
- Reactor Coolant Pumps Miniatus Acceptable Maximum Acceptable Measured Flow Raee Flow Rate Flow Rate Pump Combination GPM CPM GPM No. _
*
- 102,400 1 1-1 99,30C 1 1-2
*
- 103,800 1 2-1
*
- 105,600 1 2-2 Lower flow pump 172,500
- 186,390 2
in each loop 1-2, 2-1 Three lowest 262,000
- 284,180 3
flow pumps 1-1, 1-2, 2-1 4 1-1, 1-2, 2-1, 2-2 352,000 410,100 378,890 Post BPRA and ORA Removal Retest 4 1-1, 1-2, 2-1, 2-2 387,200 429,400 403,818 3
- Indicates enat no acceptance criteria was established.
DBNPS - UNIT 1 STARTUP REPORT REACTOR COOLANT SYSTEM FLOW MEASURDIENTS 4-5 TABLE 4.1-1 1
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l l
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i l Flow Coastdown for Trip of 1 out of 4 RCP 360 - ;-
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- i 1 2 3 4 5 6 DBNPS - UNIT 1 Time After Pump Trip (Seconds) STARTUP REPORT FLOW COASTDOWN FIGURE 4.2-2 4-7 I
k
.R~
FLOW COASTDOWN FOR TRIP OF 2 OUT OF 4 RCP 38C
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, PRESSURIZER SPRAY AND HEATER ACTUATION SETPOINT DATA Measured Acceptance Value (psig) Criteria (psig)
Pressuri:er Spray OPENS 2205 220S 1 16 Flow Valve CLOSES 2165 2155 1 16 Hester Bank 1 ENERGIZES 2155 2155 1 16 (
- 2155 2.55 1 16 DE-ENERGIZES Heater Bank 2 ENERGIZES 2138 2135 1 16 DE-ENERGIZES 2155 2155 1 16 Heater Bank 3 ENERGIZES 2115 2120 1 16 DE-ENERGIZES 2140 2140 1 16 Heater Bank 4 ENERGIZES 2115 2105 1 16 DE-ENERGIZES 2125 212S 1 16 t
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9 PRESSURIZER SPRAY FLOW CALCULATIONS Spray Flow = Q (VRCS) ~ a h(K) where, Q = heater input (KW) . YRCS = specific volume of RCS cold leg (ft /lbm) 3 K = constant 2.3Sx10-3 ft - min - KW gal-BTU / Ah = enthalpy of pressuri:er - enthalpy of spray water (BTU /lbmT RCS RCS Pressurizer Pressurizer Calculated Temperature Temperature ' Heater Power Flow Pressure Pressuri:er 1224 KW 190 gpm Spray 1280 psig 532 F 577 F Pressuri:er 30 KW 1.6 gpm Mini-Spray 2150 psig 532 F 647 F DBNPS - UNIT 1 STARTUP REPORT PRESSURIZER SPRAY FLOW CALCULATION TABLE 4.4-1 4-11
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- P Report l
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. Figure 4.5-1 the Cabinets to 3/4 Insertion (.s ec . )
4-13
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5 0.9 o-- m O L z a 8 0.7 a w T.l c 0.5 2 4 6 8 10 Reactivity added by CRA withdrawal (% 6 k/k) DBNPS-Unit 1 Startup Report 1/M vs. CRA Withdrawal Figure 5.1-1 5-3
A new stable base counts for source range nuclear instrumentation NI-1 and NI-2 were determined prior to commencing deboration. At ten minute intervals during deboration, count races.were obtained; and these count rates were then plotted as inverse multiplication (1/M) values for both source range detectors versus deboration time and gallons of water added. In addition, at 30 minute intervals during deboration, the reactor coolant system was sampled for-boron concentration. These boron concentrations were then plotted versus inverse multiplication (1/M) values. These plots are provided in Figure 5.2-1. When criticality was achieved, the deboration process was stopped, maximum letdown flow to the makeup tank was established, and the boron concentrations of the RCS, makeup tank, and pressurizer were allowed to mix, while the reactor power was maintained at approximately 1.9 x 10-10 amps on the intermediate range by moving Group 6/7. The boron concentrations were essentially equalized with CRA Group 6/7 controlling at 56% withdrawn. Values of 1473, 1456 and 1493 ppa were recorded for the RCS, makeup tank and pressurizer respectively. The design RCS boron concentration for these conditions was 1453 ppm. During the zero power testing program, an all rods out, critical boron con-centration was measured. The results of this measurement are summarized in Section 6.4. t b { ! s i r 5-2 l
5.0 INITIAL CRITICALITY TP 710.01 The approach to initial criticality began at 2344 on August 11, 1977. Initial criticality was achieved at 1729 on August 12, 19 77, at reactor coolant conditions of 530 F and 2155 psig. Two out of core source range detector channels (NI-1.and NI-2) were used to monitor neutron flux. Inverse multiplication (1/M) plots, maintained during both the rod with-drawal and the subsequent reactor coolant system deboration, were used to predict criticality. Soron concentrations in the reactor coolant system, pressurizar, and makeup tank were determined by chemical analysis throughout the approach to criticality. Criticality was achieved in two major steps: a preliminary approach (step) and a final approach (step) . The preliminary approach consisted of a pre-determined control rod withdrawal sequence until rod groups 1 through 5 and 8-were at 100 percent withdrawn, and rod groups 6/7 were at 75 percent withdrawn. The final approach consisted of a reactor coolant system debora-tion from 1824 ppm boron to 1473 ppm boron. , 5.1 PRELIMINARY APPROACH TO CRITICALITY The initial boron concentration was determined to be 1824 ppm for the reactor
, coolant system,1828 ppm for the makeup tank, and 1833 ppm for the pressurizer.
The initial stable base counts for source range nuclear instrumentation NI-l and NI-2 were determined for the all rods in condition and then the following control rod group withdrawal sequence was performed: 1.) Orcup 1 was 100 percent withdrawn 2.) Group 2 was 100 percent withdrawn 3.) Group 3 was 100 percent withdrawn 4.). Group -4 was 100 percent wi-hdrawn 5.) Group 8 was 100 percent withdrawn 6.) Group 5 was 50 percent withdrawn
- 7. ) Group 5 was 75 percent withdrawn 8.) Group 6/7 was 25 percent withdrawn and Group 5 was 100% withdrawn 9.) Group 6/7 was 45 percent withdrawn-10.) Group 6/7 was 60 percent withdrawn 11.) Group 6/7 was 75 percent withdrawn After each of the above sequential control rod group withdrawals, stable counts were determined for each source range nuclear instrumentation NI-l and NI-2. Inverse multiplication (1/M) values were then determined and plotted versus rod group withdrawals (reactivity added) . These plots are provided in Figure 5.1-1.
5.2 FINAL APPROACH TO CRITICALITY Reactor coolant system deboration was commenced af ter the specified control rod withdrawal sequence was completed. A constant feed and bleed rate of approximately 45 gallons per minute of the reactor coolant system was performed. The boron concentration of the reactor coolant system, makeup tank, and pressurizer was re-verified, and the number of gallons of demin-eralized water needed to deborate to 1528 ppm boron was calculated. 5-1
. g L.
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-6.0 CORE PERFORMANCE DURING ZERO POWER PHYSICS TEST
~
Following initial criticality, a zero power testing program was conducted j to (1) confirm the nuclear design characteristics of the core, (2) validate the assumptions used in the safety analyses, and (3) validate the analytical models used for predicting plant responses. Measurements to determine the shutdown margin and the zero power (1) moderator coefficient, (2) control rod worth and (3) differential boron worth were made. Stuck rod and ejected rod worths were also measured during zero power testing. All testing yielded satisfactory results and ensured that initial operation of the reactor was within the limits of the Davis-Besse Unit 1 Technical Specifi-cations. The subsequent sections of this chapter summarize the results of the various tests performed at zero power in accordance with Zero Power l Physics Test, TP 710.01. 6.1 NUCLEAR INSTRUMENTATION OVERLAP In accordance with Technical Specification 4.3.1.1.1, during startup, an overlap of at least one decade between the source range nuclear instrumenta-tion and che intermediate range nuclear instrumentation must be verified. Overlap data for the nuclear instrumeatation was acquired immediately following initial criticality at reactor coolant system condition of 530*F and 2155 psig. The reactor pcwer was slowly increased from the just critical flux level. When the intermediate range detectors came on scale, their flux levels were recorded. The flux levels were recorded at several power levels until che source range high voltage cutoff level was obtained. An overlap of approximately 2 decades was observed between the intermediate 1 and source range detectors which satisfies the minimum allowable overl*ap of 1 1.0 decade. Overlap data for all of the nuclear instrumentation is shown in Figure 7.1-2. l 6.2 SENSIBLE HEAT DETERMINATION I Prior to zero power physics testing, the intermediate range level at which l nuclear heat (sensible heat) occurs was determined. Nuclear heat is defined i as the flux level at which detectable heat is being produced. After this level was determined, it was decreased by 20% and reactor power was restricted i to a level below this limit during zero power testing. ; I Nuclear heat was determined by increasing power in set increments, allowing !
. power to stabilize and then observing whether sensible heat was produced.
~
I The following parameters were used as an indication that nuclear heat was
-being produced:
- 1. Increase in RCS Tave. )
2 .' Increase in Thot.
- 3. Increase in the pressurizer level 4.'. Opening of the turbine bypass valves l Values for the parameters mentioned above are shown in Table 6.2-1 for each
! power increment. l l l' 6-1 1
)
L Nuclear heat determination was performed at 530F and 2155 psig and was found to occur at 1 x 10-7 caps on the intermediate range nuclear instru-mentation. All zero power measurements, except nuclear instrumentation overlap data, was obtained at a flux level below 2 x 10-8 amps on inter-mediate range NI-3 to ensure that temperature feedbacks were held to a miniumm. 6.3 REACTIMETER RESPONSE CHECKOUT Prior to using the reactimeter for reactivity measurements, the response of the. reactor to a change in reactivity was compared to the design responte. A second checkout of the reactimeter was performed just prior to the rod worth measurements. The purpose of these checkouts was to verify that the delayed neutron constants used by the reactimeter gave an accurate repre-sentation of the core. The checkouts were accomplished by initiating a reactivity excursion and measuring the doubling time of the flux. A plot of the reactivity inserted versus the doubling time was obtained and compared to the design value. These plots are shown in Figures 6.3-1 and 6.3-2. The design curves were obtained from an analytical solution of the inhour equation, using the same delayed neutron constants utilized by the reactimeter. If the theoretical delayed neutron constants were representative of the core, the reactivity obtained from the reactimeter for a given reactor period would be spproximately equivalent to the reactivity obtained from the analytical solution of the inhour equation for the same reactor period. Reactivity insertions of approximately +25, -25, +75, and -75 pcm were ' 1 obtained during each checkout of the reactimeter. Data obtained from these , measurements is summarized in Table 6.3-1. All measurements for the initial ]
, checkout were within the acceptante criteria of 2 5% of the design values. l The response check prior to the rod worth measurements yield a 6.25% deviation from the expected value for one of the doubling time measurements for the
-25 pcm reactivity insertion. However, when all of the. doubling times for that particular measurement were averaged, a deviation of less than 1 percent was realized, i
l 6.4 ALL RODS OUT BORON CONCENTRATION ! The all rods out, hot zero power (HZP), beginning of life (BOL) critical boron concentration was measured and compared to design. This comparison was used as one of the criteria for establishing the validity of the core i physics model. I
~ ~- - - ~
With Controfi7d Assembly (CRA) Group 8/7 controllTng E 88% wi5draim~ ~ ~ ~~ ~ and all othe;. ads- fully withdrawn, a boron endpoint measurement was per-formed to determine the all rods out (ARO) boron concentration. The meas-
~
ured boron concentration was 1518 ppm. The measured excess reactivity l
" worth of the inserted rods vis 54 pefcent~millif hica ~ l pcm = 10-5 ok/k). ~
~ ~ ~ ~
l Using the differential boron worth of 10.4 pcm/ ppm from Figure 2.3-2 of the ) , B&W Physics Manual, an ARO critical boron concentration of 1523 ppm was obtained. This is within the acceptance criteria of 1566 + 100 ppm. L
)
l 6-2 1 I l
, -r-+ .-- , - - - , - . , . w -,,w- 4, w -e -g r---- ' 3 - ,-- s--- w----
6.5 TDfPERATURE COEFFICIENT OF REACTIVITY The temperature coefficient of reactivity is defined as the fractional-change in the core reactivity per unit change in the average core temp-l erature. This quantity is the sum of the change in reactivity per unit change in fuel temperature (Doppler coefficient) plus the change in reactivity per unit change in the moderator temperature (moderator temperature coefficient). The temperature coefficient was measured directly by observing the reactivity change induced by a change in the average core temperature. From the BW Physics Manual, the Doppler coefficient at zero power is -2.0 x 10-3% a g/g, Using this value, the moderator temperature coefficient was indirectly obtained by subtracting the Doppler coefficient from the measured temperature coefficient.
~
Power ramps were initiated for each temperature coefficient measurement to obtain the approximate RCS temperattare change sequence of +50 (case (1) ),
-100 (case (2) ), +50 (case (3) ). A weighted average of the temperature coefficient was obtained where:
- -~~~ ~
- ~ ~T (weighted average) = f (t'otal)/
~ ~ - -
1 (total) and
,f(Total) = of case (1) - Af case (2) + Af case (3)
T (Total) = A T case (1) - A T case (2) + A T case (3)
~ ~-
Measurement results for the all r6ds out'~and other rod configurations are summarized'in Table 6.5-1. All moderator temperature coefficients measured satisfied the Davis-Besse Unit 1 Technical Specification limits of less positive than 0.9 x 10-4 0 k/k/0F (9.0 pcm/0F) and less negative than -3.0 x 10-4 d k/k/0F (30.0 pcm/0F) 6.6 CONTROL ROD REACTIVITY MEASUPEMENTS
~ ^~
During zero power physics testing at 5300F, measurements were made to determine ~ ~ the CRA group reactivity worths for the safety rods (CRA Groups 1-4), the regulating rods (CRA Groups'5, 6 and 7) and the Axial Power Shaping Rods (CRA Group 8). The reactivity worths of the control rod groups were calculated utilizing the React 1 meter. The " boron-swap" method was utilind to determine differential and integral worths for control rod groups 5, 6, 7 and 8. A rod drop measurement was used to determine the worth of the remaining rod groups (CRA's 1-4) instead of the " boron-swap" technique, i The " boron-swap" technique consists of establishing a slow boration er debora-tion rate while moving the control rods periodically to compensate for the reactivity change. Reactivity and control rod group position were continuously l . recorded versus time and analyzed'to determine the differential reactivity l' due to each rod movement. Integral coctrol rod worths.were obtained by adding the differential changes in reactivity over the range of control rod travel. The rod-drop method wss used to determine the total rod worth. This measure-ment and the results are discussed in Section 6.8. 6-3
Measured worths are tabulated and compared with predicted values in Table 6.6-1. Differential and integral rod worth curves obtained from these measurements are shown in Figure: 6.6-1 and 6.6-2. 6.7 EJECTED CONTROL ROD WORTH The ejected rod worth test was performed to measure the reactivity worth of the single predicted worst case ejected rod. Other ejected rods were also verified to be less reactive than the predicted worst ejected rod. CRA Rod 7-8 (and symmetric rods) was the predicted most reactive ejected rod. CRA Rods 6-11 and 5-10 (and their symmetric rods) were also identified as being high worth ejected rods. The ejected rod worth of CRA Rod 7-8 was determined from the Technical Specification 3.1.3.6 rod position insertion limit of 49% on CRA Group 5. Adjustments were made for measurement uncertainties and compared with the hot-zero power ejected worth limit of +1.0% 4 K/K assumed in the Safety Analysis. The measurement was initiated from the following rod configuration: CRA Groups 1-4 at 100% withdrawn CRA Group 5 at 49% withdrawn CRA Groups 6/7 at 0% withdrawn CRA Group 8 at 35% withdrawn CRA Rod 7-8 was borated out of the core and then inserted back into the core utilizing a rod swap with CRA Group 5. The worth of CRA Group 5 inserted during the swap was obtained from the rod worth measurements performed earlier in the testing program. This value was multiplied by 1.05 to account for measurement uncertainties. This resulted in an ejected rod worth of 0.78% A K/K. Following the ejected rod worth measurement of CRA Rod 7-8, CRA Rod 6-11 and 5-10 were swapped with CRA Group 5 first out and then back into the core. The results of this phase of the testing verified that CRA Rod 7-8 was more reactive than CRA Rod 6-11 or 5-10. 6.8 STUCK ROD WORTH AND SHUTDOWN MARGIN MEASUREMENT This test section accomplished three major objectives: (1) measured the total rod worth, (2) measured the worth of the predicted worst case stuck
~
rod, and (3) verified that at least 1.0% 4 K/K shutdown margin exists at the Technical Specification rod insertion limit of 49% withdrawn on CRA ; Group 5. Data for the stuck rod worth was obtained at the hot standby RCS condi- _ tions of 5300F and 2155 psig. The measurement consisted of three separate j sets of three rod drops. Each set of drops was performed from a different l
-initial rod configuration.
l 6-4
em-' aw e The first set of rod drops were initiated from a critical reactor with all rods withdrawn except CRA Groups 6/7 which were controlling at 80% withdrawn, and CRA Group 8 which was maintained at 35% withdrawn throughout the stuck rod testing. The second set of measurements were performed with the same rod configuration as was set one, except CRA Groups 6/7 were 24% withdrawn. Set three was conducted with CRA Groups 6/7 fully inserted and CRA Group 5 controlling at 54% withdrawn. Each set of rod drops consisted of withdrawing CRA Group 7 to the 100% withdrawn position while inserting the controlling group, or the next sequential rod group, to compensate for the change in reactivity. All control rods, except CRA Group 8, were then dropped into the core. This rod drop is signified as the symmetric rod drop. Reactivity inserted during the rod drop was measured with the reactimeter. Criticality was re-established at the pre-trip configuration and then all rods except CRA Rod 7-4 and CRA Group 8 were dropped. This rod drop is signified as the asymmetric rod drop. Reactivity inserted was again measured by the reactimeter. Finally, a third rod drop consisting of only CRA Rod 7-4 was performed. This drop was initiated for the sole purpose of obtaining the all rods in configuration. The data from the rod drop measurements was analyzed graphically to obtain: (1) the total rod worth, (2) the worth of the predicted most reactive rod and (3) the shutdown margin at the beginning of cycle I. For each case, the reactivity computed by the reactimeter for the specific rod drop was plotted against the reactivity inserted from the all rods out condition to obtain .the pre-trip rod configuration. Reactivity inserted is referenced
, as the equivalent boron worth because it represents the reactivity equivalent to the change in boron from the all rods out condition. This value was obtained from the reactivity worth of the inserted rods which was measured previously in the testing program. Extrapolating the symmetric rod drop points to the zero-intercept using a linear fit gives the measured ' total rod worth. A similar treatment of the asymmetric drop gives the N-1 rod worth. Taking the difference between these two measurements yield the worth of the predicted most reactive rod.
The third set of rod drops were obtained near the Tech Specs insertion limit ! of 49% withdrawn on CRA Group 5. A simple extrapolation of the rod worths l yield the measured equivalent boron worth for the insertion limit. Taking i the difference between this equivalent boron worth and the total rod worth l obtained previously gave the shutdown margin at the beginning of cycle I. This value was reduced by 50% to account for measurement uncertainties and compared to the 1.0% A K/K shutdown margin required by Tech Specs. Table 6.8.1 summarizes the rod drop measurements while Figure 6.8-1 repre-sents a graphical analysis of the data. All results satisfied the appropriate acceptance criteria. 1 J f 6-5
NOMINAL REACTOP. POWER (AMPS) - FROM NI-3 AND NI-4 ON CONSOLE
~
~
-9 PARAMETER NAME COMPUTER Ir 1 x 10-9 3 x 10 1 x 10 3 x 10- 1 x 10~7 NI-3 Intermediate Ranae Level (amps) R818 1.045 x 10-9 3.373 x 10~ 1.076 x 10-0 3.319 x 10-8. 1.130 x 10-1.140 x 10-9 3.690 x 10~ -8 ~7 NI-4 Intermediate Ranae Level (amps) R812 1.175 x 10-8 3.606 x 10 1.130 x 10 RCS Tave ( F) 'T709 530.4 530.3 530.3 530.4 530,7 Pre-curizer Level (inches) L769 69.07 70.49 71.83 73.18 75.45 RCS Loop 1, T-Hot, NR (OF) T719 530.5 530.5 530.5 530.6 530.9 "CS Loop 1, T-Hot, NR (OF)
. T720 526.8 526.7 526.7 526.7 526.9 RCS Loop 2, T-ilot, NR ( F) T728 530.4 530.4 530.4 530.4 530.7 RCS Loop 2, T-liot, NR (OF) T729 530.7 530.8 530.8 530.9 531.2 4
DBNPS - UNIT 1 7 N STARTUP REPORT SENSIBLE IIEAT DETERMINATION TABLE 6.2-1
6.9 SOLUBLE POISON WORTH MEASUREMENTS Data for the differential boron worths was obtained during the rod worth measurements. Coefficients at boron concentrations of approximately 1200 ppm, 1300 ppm, 1400 ppm and 1500 ppm were calculated and plotted with the predicted all rods out boron coefficients on Figure 6.9-1. Differences between the design and calculated values are due to measurement uncertain-ties and the fact that the calculated coefficients represent a rodded core. A summary of the calculations is given on Table 6.9-1. All measured boron coefficients are within acceptable ranges when the discrepancies between measured and design conditions have been taken into account. l I l , 6-6 l
REACTIMETER RESPONSE CHECK 0UT 1 INITIAL REACTIVITY CHECKOUT CONTROLLING CRA GROUP AVERAGE REACTIVITY DEVIA-DESIRED DO M ING ION REACTIVITY GROUP INITIAL FINAL DESIGN ACTUAL TIME NUMBER Pcm % wd % ud Seconds pcm pcm %
+25 6/7 55 58 144 +35 +34 +2.9
-25 6/7 58 54 283 -22 -23 -4.5
+75 6/7 54 60 58 +73 +72 +1.4
\
-75 6/7 60 53 118 -64 -62 +3.1 CHECKOUT PRIOR TO ROD WORTH MEASUREMENTS
+25 6/7 85 94 105 +46 +45 +2.2
-25 6/7 86 82 265 -24 -24 0.0
+75 6/7- 82 100 56 +73 +72 +1.4
-75 6/7 86 -
125 -60 -58 3.3 DBNPS Unit 1 l Startup Report l l Reactivity Checkout l- Table 6.3-1 ; t j 6-8
. . . l
. ._ ._. _ , - - ~ --
MODERATOR TEMPERATURE COEFFICIENT MODERATOR TEMPERATURE TEMPERATURE COEFFICIENT COEFFICIENT RCD CONFIGURATION (pcm/F) (pcm/F) CRA Groups 1-5 @ 100% wd CRA Group 6/7 @ 83% wd +3.4 +5.4 CRA Group 8 @ 100% wd CRA Groups 1-5 @ 100% vd CRA Group 6/7 @ 24% wd +0.2 +2.2 CRA Group 8 0 36% wd CRA Groups 1-4 @ 100% wd CRA Group 5 @ 51% wd CRA Group 6/7 @ 0% wd - -2.8 -0.8 CRA Group 8 @ 35% wd DBNPS Unit 1 Startup Report iemperature Coefficient Table 6.5-1 l S 6-9
TABLE 6.6-1 COMPARISON OF MEASURED AND PREDICTED CONTROL ROD GROUP REACTIVITY WORTHS AT 530 F Predicted Worth (1) From Physics Deviation Measured Worth Test Manual from CRA Position, % WD % AK/K % A K/K Predicted CRA 8 (maximum worth) - 0.385 -0.46 -16.3% CRA 6/7 (100 to 0) - 3.20 *
-3.34 - 4.4%
(CRA Group 8 at 35% vd) CRA 1-5 (100 to 0) - 6.79 -6.01 +13.0% (CRA Group 8 at-35% wd) i Total Rod Worth -10.38 -9.81 + 5.8% i l (1) Deviation Measurcd - Predicted from = Predicted Predicted l l i l
- j. l l
1 I l
'E l
DBNPS - UNIT 1 STARTUP REPORT l ROD WORTHS TABLE 6.6-1 6-10
. .o .
STUCK ROD AND SHUTDOWN MARGIN
SUMMARY
REACTIVITY FROM REACTIMETER EQUIVALENT SYMMETRIC ASYMMETRIC ROD BORON WORTH DROP DROP CONFIGURATION pcm (pcm) (pcm) CRA Groups 1-5 @ 100% wd - CRA Group 6 @ 74% wd 480 -6135 -6273 CRA Group 7 @ 100% wd CRA Group 8 @ 35% wd CRA Groups 1-5 @ 100% wd CRA Group 6 @ 12% vd CRA Group 7 @ 100% ud 1790 ' -5104 -4687 CRA Group 8 @ 35% wd I CRA Groups-1-2 @ 100% wd CRA Groups 3-4 @ 25% wd l CRA Groups 5-6 @ 0% vd 3730 j -4119 -3406 CRA Group 7 @ 100% vd CRA Group 8 @ 35% wd i Symmetric Asymmetric I l
- Linear Fit Y= 0.6119X - 6342 Y= 0.8149X - 6519 l
, Zero - Intercept 10,365 7,530 l Measured Design Acceptance ' (pem) (pem) % Deviation Criteria Total Rod Worth 10,365 9,810 +5.7 120% ) Stuck Rod Worth 2,835 2,980 -4.9 130% Shutdown Margin 1,900 NA NA 21,000 pcm
- NOTE I = EquivalenIt Boron Worth (pem)
Y = React 1 meter Reactivity (pcm) DBNPS - UNIT 1 STARTUP REPORT SHUTDOWN MARGIN TABLE 6.8-1 6-11
i I l DIFFERENTIAL BORON WORTHS ' l l
- j ROD POSITION BORON CONCENTRATION BORON COEFFICIENT Initial Final Change Measured Design INITIAL FINAL (ppm) (ppm) (pcm) pcm/ ppm pcm/ ppm CRA Groups 1-8 CRA Groups 1-5 100% withdrawn 100% withdrawn !
CRA Group 6/7 1523 1490 445 13.5 10.4 l 85% withdrawn l CRA Group 8 i 35% withdrawn CRA Groups 1-5 CRA Groups 1-4 ' 100% withdrawn 100% withdrawn CRA Group 6/7 CRA Group 5 85% withdrawn 99% withdrawn 1490 1359 1340 10.2 10.4 CRA Group 8 CRA Group 6/7 35% withdrawn 24% withdrawn CRA Group 8 ' 35% withdrawn CRA Groups 1-4 CRA Groups 1-4 100% withdrawn 100% withdrawn CRA Group 5 CRA Group 5 99% withdrawn 90% withdrawn CRA Group 6/7 CRA Group 6/7 1359 1234 1210 9.7 10.5 24% withdrawn 11% withdrawn CRA Group 8 CRA Group 8 35% vithdrawn 35% withdrawn CRA Groups 1-4 CRA Groups 1-4 100% withdrawn 100% withdrawn CRA Group 5 CRA Group 5 90% withdrawn 46% withdrawn CRA Group 6/7 CRA Group 6/7 1234 1168 760 11.5 10.5 11% withdrawn 0% withdrawn CRA Group 8 CRA Group 8 35% withdrawn 35% withdrawn l i DBNPS - UNIT 1 STARTUP REFORT i BORON WORTHS TABLE 6.9-1 6C _ - - ~ - - , - ,
i-REACTIVITY VS POSITIVE DOUBLING TIME 100 _ x x R J"'k g i I I
; N 1 , , , l
- . x_ ,.
50 _- .-
'A - -
n a - o v C. '~.'m m a 1 -u y I I I , N [3 , . ;- d e .
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l _ ll lll Illi lill l11lili,!!ll i ll 1 Il lli illl' lill lli I ll Ili IIl l ll llIl l 11 IIII llIl ll l 60 80 100 120 140 160 t t DOUBLING TIME (SECONDS) DBNPS - Unit 1 Startup Report l Reactimeter Checkout l Figure 6.3-1 6-13
REACTIVITY VS NEGATIVE DOUBLING TIME
-100 s
- \
1
-x.
,- V, , l
- 2.? _--- ._.
-50 -.
.'4 .
x1 7 'N
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it!. t.' , !!!, list sii' !!ai a!!! itii titt 9.'s !! i ii lisi iii' iiii li!* !! siti i'it o, .in u,, i., i oit u,6 on in ou iin no sin n i'i ni i.'sn .n un on iin o., nii ,.o on .!n
- p. In. ,,, nii on no ni lit nii no .o ini ini n i no nii ini ini on on an on un in iii un un on no nii lii ni siti iin ut. nii un ii i sin oh ilii siti nii iiii ilo int sin in llI IIli lill l;!! Sitt iill il I lit i il 16l llll ills lill ll l lill il l l661 llll ll jlll il(t ll41 ll t l HI n't tile ilh till iill !! I Uli i l li !HI Hi! 1111 ti i INI il I Ilil NH ll li i lill lli 11 1 I III lll! !!ll lili Ill! llil ill! ll! l ll ll'i llil 'll llll Illi l l I ll ll i l liIl ll' ll ll i
lill llH ltllIlli ill! I!ll !lli Ili ill lhi illl I il lil! lill I I i l II ll! lill lill !! II lll1 llIl ill1llll! lillilli ll11 lil' lill lill Illi Illi illi ll11 lill11 Illi liii lil Illi illi ilIl llll 11: 1 100 140 180 220 260 300 NEGATIVE DOUBLING TIME (SECONDS) DBNPS - Unit 1 Startup Report Reactimeter Checkout Figure 6.3-2 6-14
O CRA GROUP 8 WORTH 100 10
=
_ 1_. . 1-
' ~ ~'
c,
- a. 10
+_
M. 0 n - 8
, m- - -
y. w
/
f 0 Qi , j n
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/
4 d n 1 :-+. s E ' , S u o v Q. n e ,'1 1 L c. 2 = H
- ' % =
.: H ec m
'- C O -200 O
3
- ._s 4 , a
+ <
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, r.4 x
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-300 .
-4 5 1
. : -6'
' ' W
- h
- J-( e m.
-400 -
w I
-8 I*y f3 t
-10
.m' L- 2 N
20 40 60 80 100 CRA GROUP 8 (% WITHDRAWN) DBNPS - UNIT 1 STARTUP REPORT CRA GROUP 8 WORTH FIGURE 6.6-1 6-15
CRA GROUP 6/7 WORTH WITH CRA GROUP 8 AT 35% WITHDRAWN , l i 0.0 f 140
~~
7,. 120 l
-500
~
."",3W 'Eis :
_100
~
E'
-1000 .['
.-dE- 9
= 80 d
m ,. 0 :.# - 8 3 @~ 5 S
-1500 -
' l 5
=-
60 *
+ -
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.-~ 'l 2 I 5. -
8 -
-= g 8 .- m- -
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= %
.-. m a
~
b-E_ , -~ 20
-2500 B_.
-5 .
c
'^ E
E .-- _ 0
-3000 b 4b 60 80 100 I l 20l CRA GROUP 6/7 60 80 100 CRA GROUP 5 R0D POSITION (% WITHDRAWN) DBNPS - UNIT 1 l
STARTUP REPORT l ROD WORTH l FIGURE 6.6-2 6-16
STUCK ROD AND SHUTDOWN MARGIN GRAPHICAL ANALYSIS
-8000- ...
n Y 8 -6000 -*y ^ S ' t u-m
$ 't h -
5 .QCl '~
- x1 x _. -
_x, _ ~ . U d
-4000 N-'
$ x _,
'N.;_ e XL t
^
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.w +
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=
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'.X t= -
--t_
N: .lQ-
-'x 4- -
-: y k -l-3gd-E 2000 4000 6dOO 8000 10,000 EQUIVALENT BORON WORTH (pcm) i 1
l 1 DBNPS - UNIT 1 STARTUP REPORT SHUTDOWN MARGIN 6-U FIGURE 6.8-1
- s. ,
FIGURE 6.9-1 - DIFFERENTIAL BORON WORTH VERSUS BORON CONCD;TRATION
, , , , . . . , , , i
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i , i . . . . . . . ... . . . . . i . . . i,
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. . .<. . . . . . . . . .. . . . . . . . . . . ....... . . , ...e . . . .
i e.i i . . i i , .t ... . . . . , . .. . . . . i . i . ,
. . . . i . . .
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- .. . 4 . . . . . .. . e,. . . . . . . . . .. . . . . . . . . . . . . . . ,
4
. ,. . . . 4 . . . . . .. . , . , . . . . . . . . .. . . . . , , , , , . .
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-0.7 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
i 600- 800 1000 1200 1400 1600 BORON CONCENTRATION, ppm 3 DBNPS - UNIT 1 STARTUP REPORT BORON COEFFICIENT FIGURE 6.9-1
. . 6-18 m .-
, , . , , , . - , . - .,_r.-_ - _ - . , . , - . - - , - - - . . - . - - - , - . . ..----,.~,,--.-v
7.0 CORE PERFORMANCE DURING POWER ESCALATION SEQUENCE TESTS Af ter the operating characteristics of the reactor at low power had- been verified, a program of power escalation was undertaken. Testing has been conducted at three major power plateaus of 15%, 40% and 75% full power. Measurements to determine 1) the reactivity coefficients and 2) the relative power distribution in the core as a function of the power level were made. The response of the out-of-core power range detectors were also evaluated. Radial peaking factors were calculated throughout testing to ensure that the operating limits established in the Davis-Besse Nuclear Power Station Unit 1 Technical Specifications were not exceeded. An analysis of the significant parameters at each plateau was made prior to escalating . power above the plateau level. The following sections give more details relavent to the performance of the physics tests conducted during the escalation phase of testing. 7.1 NUCLEAR INSTRUMENTAT CALIBRATION AT POWER, TP 800.02 ' The nuclear inc;rumentation is designed to provide nuclear flux informa-tion over the entire operating range of the reactor. Three types of neutron detectors are employed to monitor the neutron flux over the required range. They are the source range (proportional counters), the intermediate range (compensated ion chambers) and the power range (uncompentated ion chambers) detectors. Their locations with respect to the reactor core are shown in Figure 7.1-1. The source range instrumentation consists of two redundant count rate channels. Each channel monitors the neutron flux over the range of 0.1 to 106 eps. These detectors provide readouts of log count rate and startup race for the operator's use. The high voltage of both detectors is automatically switched off when the flux level is greater than 10-9 amp in the intermed-inte range channels or a power level of 10% or more is indicated by any of the power range channels. The high voltage is automatically switched on when the flux level returns to within approximately one decade of the
" d = = useful operat_in_g_ range.
The intermediate range instrumentation consists of two redundant channels. Each channel has a separate adjustable high voltage power supply and an adjustable compensating voltage supply. The power range instrumentation consists of four redundant channels. The channels are calibrated to monitor the neutron flux over the range from 1% to 125% of rated power. Each channel is composed of two 72-inch sections with a single high voltage connection and two separate signal connections. The power signal is derived from the sum of the two sections. A signal generated from the difference in the currents from the .op and bottom sections of each detector is displayed on the control console to give the operator an indication of the axial power imbalance. i 7-1
Listed below are the bases for the acceptance criteria for TP 800.02,
" Nuclear Instrument Calibration at Power".
- 1. Section 7.8.1 of the Davis-Besse Nuclear Pover Station FSAR states that "a minimum of one decade overlap between ranges is provided".
The overlap between the source and intermediate range was established during the performance of TP 710.01, "Zero Power Physics Test".
- 2. Section 7.8.1.1 of the FSAR states that "each power range detector is calibrated to a heat balance by 2% or less".
- 3. Table 4.3-1 of the Davis-Besse Nuclear Power Station Unit 1 Technical Specification states that the out-of-core detectors are to, be calibrated to within the tolerances established in the following relationship:
R {APIo-API]<,3.5% eq. 7.1-1 where: RTP = U td Thermal Power . TP = Thermal' Power APIo = Out-of-core Axial Power Imbalance API7 = Incore measured Axial Power Imbalance During power escalation, the power range nuclear instrumentation was calibrated at various power levels to indicate within t2% of the reactor power as determined by a heat balance and to within the limits established in equation 7.1-1 of the incore axial offset as calculated from the incore monitoring system. TP 800.02, Nuclear Instrumentation Calibration at Power, controlled the power range nuclear instrumentation calibration during initial power escalation to 30% FP where the imbalance relationship in equation 7.1-1 was first established. The power range NI's were recalibrated per Surveillance Test Procedure ST 5030.11, RPS Power Range Calibration, at each new power plateau and as required by ST 5030.01, RPS Daily Heat Balance Check, to maintain a t2% accuracy with the heat balance and as required by ST 5030.10, RPS Monthly Imbalance Check, to maintain the relationship in equation 7.1-1. The reason that the excore detectors had to be recalibrated at various power levels is due to the fact that their signal is generated from leakage neutrons. Leakage from a core is a function not only of power level, but of rod position, xenon poisoning and boron concentration among other parameters. At each power plateau, several of these parameters, other than power level, changed and therefore affected the leakage neutron
~ flux seen by the detectors.
7-2
.The overlap between the-intermediate and power range nuclear instrumenta-tion was verified to be greater than one decade during the performance of TP 800.02. The overlap for all three nuclear instrumentations is shown in Figure 7.1-2.
TP 800.02 and TP 800.00, " Power Escalation Control Procedure", included the adjustment of the RPS Overpower Trip 31 stable Setpoints. The settings established at the indicated power levels are given below: Reactor Power Bistable Setpoint
% FP % FP 1 10 I 5 25 40 50 75 85 e 100 105.5 i
For each case, the setpoint was adjusted prior to escalating above the indicated power level. 7.2 REACTIVITY COEFFICIENTS AT POWER, TP 0800.05 A. DOPPLER COEFFICIENT AND POWER DOPPLER COEFFICIENT 4 The Doppler coefficient of reactivity is defined as the fractional change of core reactivity per unit change in the fuel temperature. This para-meter is essential when calculating the moderator temperature coefficient. The Doppler effect -introduces a negative reactivity contribution to the temperature coefficient due to the broadening of the U-238 capture resonance. , The Doppler coefficient cannot be measured directly in a commercial reactor since the fuel temperature is not monitored. An indirect method of calculating the Doppler coefficient (*% ) from the power coefficient (atp ) was used. g .m Af -* - a n
+ Af.
ATF I bTM a P. aTF AP eg. 7.2-1 or: ATm 6TP .
-r W Mp fu 6 JM gp D ~q eq. 7.2-2
.where'A T,and OT are y the temperature changes of the moderator and fuel respectively.and AP is the change in power.
7-3 4 .- , - -
From 15% to 100% full power, the average moderator temperature is held constant, so equation 7.2-2 becomes .
. 6IF eq. 7.2-3 CX p N D}* PD The quantity on the right, known as the power Doppler coefficient, was measured at the 40% and 75% power plateaus. The method used was to vary the reactor power by approximately 5% while maintaining all other parameters essentially constant. The following conditions were established prior to each measurement.
- 1. Pressurizer and Makeup Tank boron concentration were within 230 ppa of the RCS concentration.
- 2. Xenen reactivity was stable.
4
- 3. Tave was constant within + loF for 10 minutes prior to the measurements.
- 4. RCS pressure was constant within t25 psig for 10 minutes prior to the measurements.
Rod worth measurements were taken as per TP 800.20, Rod Worth at Power (see Section 7.3), before and after the power change. Using control rods to compensate for the power change then allowed a determination of the reactivity worth of the power change, and of the power Doppler (power) coefficients. Rearranging equation 7.2-3 yields aP CQ N M P eq. 7.2-4 The Doppler coefficients were calculated with this relationship using the measured power Doppler coef ficients and Figure 7.2-1, " Average Fuel Temperature vs. Reactor Power". No acceptance criteria were applied to the values of Doppler coefficient computed, but the power Doppler coefficient was limited to a maximum positive value of -3.7 x 10-5 A K/K/%FP. The values computed at the 40% and 75% power plateaus were -14.4 x 10-5 and
-8. 2 x 10-) a K/K/%FP respectively.
B. MODERATOR TEMPERATURE CO FTICIENT Section 3.1.1.3 of the Davis-Besse Nuclear Power Station Unit 1 Technical Specifications state: The moderator temperature coefficient (MTC) shall be:
- 1) Less positive than 0.9 x 10-4 AK/K/ F0 whenever THERMAL POWER is
< 95% of RATED THERMAL POWER.
- 2) Less positive than 0.0 x 10-' AK/K/0F whenever THERMAL POWER is E95% of RATED THERMAL POWER, and
- 3) Less negative than -3.0 x 10-4 aK/K/ F- at 0RATED THERMAL POWER.
7-4 k_ ._
The moderator temperature coefficient was calculated at the 40% and 75% power plateaus to verify that these requirements were not violated. Direct measurement of the moderator temperature coefficient is not possible in an operating reactor because a change in the moderator temp-erature is directly associated with a change in the fuel temperature. The moderator coefficient was calculated by measuring the temperature coefficient and then subtracting the calculated Doppler coefficient. The temperature coefficient was measured by changing Tave approximately 50F while holding all other paramecers essentially constant. The same conditions listed above for the power Doppler coefficient measurements were also established prior to each temperature coefficient measurement. As for the power Doppler measurement, rod worth measurements were taken before and after the, temperature changes and since only rods were used to compensate for reactivity changes, the temperature coefficient could be computed. The computed moderator temperature coefficients for the 40% and 75% power plateaus were -2 x 10-7 AK/K/0F and -1.1 x 10-5 gg/g/oF respectively. 7.3 ROD WORTH AT POWER, TP 800.20 TP 0800.20, " Rod Worth at Power", was performed to determine the average differential rod worth at power to provide data for TP 0800.05, Reactivity Coefficients at Power, and TP 0800.28, Pseudo Centrol Rod Ejection. This procedure employed the quick-insertion, quick-withdrawal method to obtain the rod worths. As the name implies, the desfm d rods were first inserted for approximately six seconds, immediately followed by an equivalent withdrawal. This measurement was performed rapidly to minimize feedback effects and since the 12 second duration time is less than the primary system loop time, the inlet temperature remained constant. To minimize adverse reactivity effects, the boron concentration in the Pressurizer and Makeup Tank was within 230 ppm of the RCS concentration and xenon equilibrium was realized prior to initiating this test. The following were also established before commencing the measurements.
- 1. Reactor power constant (10.5%) for 10 minutes prior to measurements.
i
- 2. Tave constant (21F) for 10 minutes prior to measurements. l
- 3. RCS pressure constant within 225 psi for 10 minutes prior to measurements.
. .During the measurements, reactivity, 7d position and core power were logged on tape by the transient monitor.
The change in reactivity is primarily ascribed to rod motion. The analysis corrected the rod worths for reactivity effects of fuel temperature changes. The resulting differential rod worths are summarized in Table 7.3-1. 7-5
7.4 CORE POWER DISTRIBUTION TEST, TP 0800.11 Power distribution data is provided by the incore monitoring system. During power escalation, this data was collected at specific steady-state conditions. The calculated eighth core power distributions were compared to the design values obtained from the B&W three-dimensional PDQ model for the Davis-Besse Unit 1, Cycle I core. These comparisons were made to benchmark the code's capability to predict core parameters at operating conditions. For each power level, a comparison of the measured and design values of radial peaks (total fuel assembly power / average fuel assembly power) and of total peaks (local power / average power) was made. Figures 7.4-1 through 7.4-6 summarize the results for the 15%, 40% and 75% power levels. Differences between measured and design values are partially attributed to differences in the core conditions that existed during the measurement and assumed for the design calculation. In lieu of this, reasonabic agreement was achieved between the measured and design values. This procedure verified that the operating limits for minimum DNBR and maximum LHR were not exceeded at the measurement conditions. Peak F and F qmeasurements were also shown to be within acceptab1'e limits. Worst case values for these parameters along with the measured quadrant tilts and core axial power imbalsace are given for the appropriate core conditions on Figures 7.4-1 through 7.4-6. 7.5 PSEUDO CONTROL ROD EJECTION TEST, TP 0800.28 A physical failure of a pressure barrier component in the control rod drive assembly could create a pressure differential that would eject the CRA from the core. A detailed analysis has been performed to demonstrate the inherent ability of the system to safely terminate this postulated reactivity excursion. A maximum CRA worth of 0.65% A K/K at rated power was used as the limiting value for this study. TP 0800.28, " Pseudo Control Rod Ejection Test", was performed at the 40% power plateau to verify that the worth of the most reactive control rod from its nominal full power position to its fully withdrawn position did not exceed this limit. Design calculations have determined the control rods in core positions E-5, E-11, M-5 and M-11 to be the most reactive CRAs at BOL under full power conditions. The ejected worth of control rod E-11 from its nominal full power position was measured during testing. Listed below is the aquence of the major events that occurred during the test.
- 1. Group 6 and 7 control rods were borated at their full power rod insertion limit.
- 2. TP 0800.20, " Rod Reactivity Worth Measurement", was performed to determine the differential worth of group 6 before the swap.
- 3. Control rod 7-3 (core position E-11) was swapped with group 6 rods until E-11 was to its 100% withdrawn position.
~ ' , 7-6 8
FIGURE 7.1-1 NUCLEAR INSTRUMENTATION DETECTOR LOCATIONS h , N NI-4 CIC % NI-8 NI-5 UCIC I UCIC / NI-l PC / 1 I e i O
/
o, o NI-2 PC O\ UCIC "t N NI-6 NI-3 \ l UCIC CIC tEcEND PC - Proportional Counter - Source Range Detector CIC - Compensated Ion Chamber - Intermediate Range Detector UCIC - Uncompensated Ion Chamber - Power Range Detector DBNPS - UNIT 1 STARTUP REPORT 7-10 NUCLEAR INSTRUMENTATION DETECTOR LOCATION FIGURE 7.1-1
, TABLE 7.3-1 i MEASURED DIFFERENTIAL ROD WORTHS AT POWER
% Differential Full Rod Group Position (: Withdrawn) Rod Worths Power 1-5' 6/7 8 (pem/% wd) 40 100 79 27 17.1 40 100 83 27 14.7 40 100 84 27 14.1 4C 100 85.5 27 14.6 75 100 81 17 15.9 75 100 82 17 13.7 75 100 83 17 13.8 t
DBNP3-NNIT1 STARTUP REPORT 4 ROD WORTHS
. 7-13 TABLE 7.3-1 bs J
1 1 l l 7.8 POWER IMBALANCE DETECTOR CORRELATION TEST, TP 0800.18 The relationship between the reactor power axial offset as indicated by the out-of-core power range nuclear instrumentation and the incore detectors was obtained during the performance of TP 0800.18. The capability to accurately duplicate the power imbalance calculated by the computer with the backup incore detector system was also demonstrated. During testing, the minimum DNBR and maximum LHR at various offsets were calculated to verify that their respective limits were not exceeded. This procedure was conducted during the 40% and 75% power plateaus. At each power level, the axial power shaping rods were positioned to obtain approximate offsets of -32%, -24%, -16%, -8%, 0%, and +7%. Conditions were allowed to stabilize at each power imbalance for a minimum of 15 minutes and then the following were obtained:
- 1. The backup incore detector readings.
- 2. Core power distribution data including incore offset values and com-puted maximum linear heat rates and minimum DNBR's.
- 3. Out-of-core offset values.
The offset measured by each out-of-core detector was plotted against the corresponding incore detector offset value. A least-squares fit line was plotted through these points and the computed slope was found to be between 1.017 and 1.042 at 40% FP and between 1.098 and 1.113 at 75% FP. This satisfied the acceptance criterion that the slope was greater than 1.000 in each case. A further requirement that each point be within 23.5% offset of the line was also satisfied. The values of out-of-core and incore offset are summarized in Table 7.8-1. The backup incore detector calculated offsets are also tabulated for each case of offset. 7-9 _ . o _
1 4
- 4. TP 0800.20 was performed to determine the differential worth of group 6 after the swap.
l
- 5. The ejected rod worth of E-11 was calculated using the average of the two differential rod wc:th measurements and the rod travel of group 6.
The measured ejected rod worth of E-11 was 0.0218 %K/K which is well below the 0.65% 6 K/K limit used in the safety analysis. 7.6 ' DROPPED CONTROL ROD TEST , TP 0800.29 Section 3.1.3.1 of the Davis-Besse Nuclear Power Station Technical Specifications states that "all control rods shall be OPERABLE and positioned within 16.5% of their group average height". Section 4.3.4.3 of the FSAR states:
"The reactor has a control function to protect against' a rod out of step with its group. The position of each rod is compared with the average of the group. If a fault is detected at power levels above 60% of rated power, a rod withdrawai inhibit is activated. If a rod is dropped, the ICS cannot ma'ntain core power to match demand by withdrawal of other rods, and the station is run back to 60% of rated power."
TP 0800.29, " Dropped Control Rod Test" was conducted to verify that these protection controls function as stated. Radial peaking factors uere calculated with the " dropped rod" at the 50% and 0% withdrawn position. Ndnimum DNBR and maximum LHR were extrapolated to 100% full power to verify that the limits established for fuel integrity during short term transients would not be exceeded for a dropped rod accident at full power. These limits are a minimum of 1.32 for DNBR and a maximum of 20.17 kw/ft. for LHR. This test was conducted in two phases. Each phase was performed at the 40% power plateau. During phase I, the seventh control rod in group 5 (core position N-8) was inserted to the 0% withdrawn position while the remainder of the group was held at the 100% withdrawn position. During the insertion, it was verified that the asymmetry alarm lamp.and the asymmetry fault lamps performed as expected. Core power distribution data was obtained with N-8 at the 100%, 50% and 0% withdrawn position. The minimum DNBR and maximum LHR, with the asymmetric rod fully inserted and 50% withdrawn, were extrapolated to the 100% full power condition. When extrapolated to 100%, the maximum linear heat race'value slightly exceeded the 20.17 kw/ft limit. This was not completely unexpected since this test method causes very unfavorable power distributions which in turn lead to unrealistic 1y high linear heat rate values. The test was re-run with a slightly improved set of initial conditions but the unrealistic 1y high linear L heat race values were still slightly above the limit. To resolve the
- problem, B&W reviewed the method used to calculate linear heat rates l and reduced some of the excess conservatisms in the calculations which 7-7
brought the maximum linear heat rate to within its limit of 20.17 kw/ft (19.27 kw/ft). The minimum DNBR limit was satisfied with all measurements; the lowest computed value was 1.55. Prior to performing phase II, the setpoints of ICS modules UL 11.6 and RC 16.5 were reduced to values below the 40% power level. These modules respectively determine to what power level a runback will go and the power level the plant must be above before a runback or a CRDC generated withdrawal inhibit can occur. To initiate phase II, an asymmetric fault uas simulated. It was verified that the reactor power was automatically runback to the setpoint of ICS
. module UL 11.6. The setpoint of ICS module UL 11.6 was then reset to 60% of full power. The setpoint of RC 16.5 was set approximately 2% above the power level and the unit load demand was increased past this setpoint and it was verified that automatic control rod withdrawal was inhibited at the setpoint of ICS module RC 16.5. The ICS module RC 16.5 was then reset to 60% power.
It was concluded that since these protection controls functioned properly when set at values below 40%, they would also perform satisfactorily when set at 60% power.
- 7. 7 INCORE DETECTOR TEST, TP 0800.24 The neutron flux within the core is monitored by 364 self-powered neutron detectors. These detectors are located in 52 fuel assemblies at seven axial positions. Each set of seven axial detectors are called a string.
The physical location of these strings in the core and their relationship to the control rods is shown in Figure 7.7-1. The output of the incore instrumentation is connected to the plant computer, which corrects the raw incore signal for burnup, differences in detector length, fuel enrichment and rod position. The computer also provides substitute values for inoperable detectors, and finally, provides a rcadout of the power distribution within the core. The incore instrumentation is also used to calibrate the out-of-core power range detectors in terms of power imbalance. TP 0800.24, "Incore Detector Test", was performed prior to calibrating the power range detectors at the 40% plateau to verify that the incore monitoring system was functioning properly. At the 40% power plateau, data for the incore self-powered neutron detectors was obtained from the computer. Ratios of computer corrected detector readings to average string readings were calculated and plotted against axial level. Symmetric detectors were compared for consistancy, while non-symmetric detectors were checked for reasonableness. After analyzing the rescits, the incore detector system was determined to be functioning satisfa ctorily. 7-8
. DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 1 FIGURE 7 2-1 AVERAGE FUEL TEMPERATURE vs. REACTOR POWER c
E 1200 j' A g 1000 ,e
.8 .-
5 . fa , g 800 _- H ',.- Q, : N .- g # h , g 600 ,. 5 ^
.- =
20 40 60 80 100
% FULL POWER i
i l ( l I ..\ DBNPS - UNIT 1
- STARTUP REPORT AVERAGE FUEL TEMPERATURE 7-12 vs. REACTOR POWER FIGURE 7.2-1 l
w
NUCLEAR INSTRUMENTATION FLUX RANGES
- 105 10' _
10
-10" 9
10 -
-103 10 3 8
10 -
- 125 -
- 102 - 10 " 100 0*
7 E# m ao
~
10 - c
-101 5 , _ 10 g5 So - b %* EU e 3 a .
3 - 100 6 gg _1 m g c c
,- 105_ n g .5 10-1 - 7 *3 m y u a .o e 10" - @ d5e 3 2 8 gg z
5 3 o *3 u e 10 - t 8 3 E -10 6 - to-9 3~ u cc u
- 10 2-
$ " -10 5 10 I
10 - 5 -10 4 m - 10-11 e - 0 2 100 - o
-103 m
- 10 6 *8 u Ei 10 S m ,8
- 10 7 -102 e =
3 0 10 2 - 6m 10-8 -101 10 3 -
'- 10 9 -100 DBNPS - Unit 1 l
l 1 Startup Report Nuclear Instrumentation i l Flux Ranges Figure 7.1-2 7-11 2
FIGURE 7.4 -1
, CORE POWER DISTRIBUTION Core Conditions -
. Measured ,
Design Measured Design GPS 1-4 at 100 % vd 100 % wd Power Level 14.6% 15 % 5 at 100 % vd 100 % wd Baron' cone 1309 ppe NA ppe
, 6 at 92.5% wd 87.1% wd Core Burnup 1 EFPD 0 EFPD 7 at 92.5% wd 87.1% vd Axial 8 at 35.5% vd 28.8% wd Imbalance -0. 89 %FP -0.16 %FP Radial Peaks 1.40 1.26 1.05 1.15 1.03 1.32 1.49 1.08 1.52 1.25 1.16 1.13 1.15 1.24 1.53 1.06 1.10 1.12 1.02 1.14 1.06 1.23 0.96 1.21 1.13 1.11 1.07 1.15 1.18 0.95 1.05 1.06 0.92 1.10 1.13 0.70 1.10 1.03 0.89 1.04 1.23 0.68 0.94 1.01 0.86 0.88 0.96 0.91 0.90 0.87 0.80 0.82 0.55 0.83 0.74 0.50 0.50 0.51 Quadrant Tilt
-0.084% -0.024%
l +0.10% +0 10%. t x..u -Measured X. XX -Design Minimum DNER = 24.30 Maximum LHR = 2.234 kw/ft 7-14
. , , - _ m -. . , - - .
FIGURE 7.4 -2
, CORE POWER DISTRIBUTION Core Conditions s Measured Design Measured Design GPS 1-4 at 100 % wd 100 % vd Power Level 14.6 % 15 %.
. 5 at 100 % v'd 100 % wd Baron" Cone 1309 ppm NA ppm
, 6 at 92.5% wd 87.1 % vd Core Burnup 1 EFPD 0 EFPD 7 at 92. 5 % wd 87.1 % wd Axial 35.5% wd Imbalance -0.89 %FP -0.16 %FP 8 at 28.8 % wd TOTAL PEAKS 1.78 1.56 1.28 1.40 1.28 1.62 1.93 1.40-2.18 1.75 1.61 1.53 1.56 1.67 2.17 1.51 1.34 1.34 1.25 1.37 1.34 1.56 1.21 1.67 1.53 1.52 1.48 1.57 1.63 1.34 1.28 1.31 1. ' 8 1.37 1.40 0.87 1.50 1.44 1.J8 1.44 1.71 0.96 1.1." .2" 1.03 1.09 1.3 1.28 1.23 1.19 0.97 1.03 0.62 1.15 1.00 0.68 0.53 0.69 Quadrant Tilt
* -0.084% -0.024%
~
+0.10% +0.10%
1 i X.XX -Measured X.XX -Design Minimum DNBR = 24.30 Maximum LHR = 2.234 kw/ft 7-15 ;
FIGURE 7.4-3 CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Design GPS 1-4 at 100 % vd 100 % vd Power Level 40.9 % 40 % 5 at Ivo % wd 100 % wd Boron' Cone 1170 DPm NA ppm 6 at vu % vd 87. It wd Core Burnup 7. 5 EFFD 4 EFPD 7 at 90 % wd 87.1% vd Axial 8 at 18 % wd 19.1% wd Imbalance +0.64 %FP +1.17 ZFP Radial Peaks 1.42 1.30 1.03 1.18 1.07 1.26 1.42 1.01 1.42 1.20 1.12 1.11 1.13 1.20 1.44 1.02 1.12 1.19 1.06 1.16 1.08 1.20 0.93 1.17 1.11 1.10 1.07 1.13 1-14 0.92 1.07 1.08 0.89 1.10 1.13 0.67 1.09 1.04 0.92 1.05 1.21 0.68 . 0.96 1.03 0.90 0.87 0.99 0.95 0.93 0.89 0.84 0.84 0.32 0.89 0.79 0.54 0.'50 Quadrant Tilt 0.56
-0.57% 0.091%
, -0.065% 0.031%
{ i
-Measured lA.A>
- 4. IX -nesign Minimum DNER = 7.79 Maximum LHR = 6.793 kw/ft 7-16
FIGURE 7.4 -4
, CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Design GP5 1-4 at 100 % vd 100 % wd Power Level 40.9 % 40 %
5 at 100 % wd 100 % wd Boron' Cone 1170 ppa NA ppa 6 at 90 % vd ' 87.1% wd Core Burnup 7. 5 EFPD 4 EFPD 7 at 90 % vd 87.1 % wd Axial 8 at 18 % wd 19.1% vd Imbalance _+0. 64 %FP +1.17 %FP Total Peaks 6 1.96 1.71 1.25 1.53 1.43 1.67 1.95 1.37 1.97 1.63 1.55 1.52 1.55 1.61 1.98 1.39 1.49 1.62 1.45 1.59 1.44 1.62 1.25 1.58 1.50 1.52 1.50 1.54 1.55 1.26 1.43 1.54 1.35 1.53 1.57 0.90 2 51 1.49 1.48 1.47 1.67 0.93 1.40 1.44 1.26 1.19 1.45 1.37 1.29 1.21 1.14 1.11 0.70 1.24 1.08 0.73 Quadrant Tilt
-0.057% 0.091%
-0.065% 0.031%
I j -Etasured g, g X.XX -Design Minimum DNBR = 7 79 m imum LHR = 6.793 kw/ft 7-17
FIGURE 7.4-5
, CORE POWER DISTRIBUTION Core Conditions
- Measure?, Design Measured Design GPS 1-4 at 100 % vd 100 % wd Power Level 73.8 % 75 %
- 5 at 100 % tid 100 _% wd Boron' Cone 1095 ppa NA ppa
, 6 at 93 % wd 87.1 % wd Core Burnup 26.3 EFPD 25 EFPD 7 at 93 % vd 87.1 % wd Axial 8 at 23 % wd 19.1 % wd Imbalance +1.09 %FP +2.3 %FP Radial Peaks 1.50 1.38 1.05 1.22 1.09 1.22 1.36 0.96 1.48 1.28 1.19 1.17 1.15 1.18 1.33 0.93 1.16 1.14 1.15 1.16 1.06 1.15 0.88 1.24 1.18 1.15 1.10 1.12 1.09 0.86 1.09 1. 15 0.88 1.12 1.10 0.65 1.15 1.09 0.94 1.03 1.13 0.64 0.99 1.05 0.89 0.83 1.02 0.97 0.92 0.85 0.87 0.84 0.52 0.90 0.79 0.53 0.50 Quadrant Tilt O.56 f
-0.50% +0.84% -
-0.28% -0.06%
i l l ; Xg -Measured X.XX -Design Minimum DNBR = 3.68 Maximum LHR = 12.134 kw/ft 7-18 l l i
~ . . _ . _ - . . , . ._ -._' . - . . _ . .
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FIGURE 7.4 -6
, CORE POWER DISTRIBUTION i
Core Conditions
- Measured ,
Design Measured Design gps 1-4 at 100 % wd 100 % wd Power Level 73.8 % 75 % ! 5 at 100 % wd 100 % wd Boron' Conc 1095 ppa NA ppa
, 6 at 93 % vd 87.1 % vd Core Burnup 26. 3 ET7D 25 ETPD 7 at 93 % wd 8 7.1 % vd Axial 8 a': 23 % vd 19.1 % wd Imbalance +1.09 gyp +2.3 %7P Total Peaks 2.03 1.77 1.33 1.56 1.39 1.53 1.80 1.25 2.07 1.76 1.67 1.60 1.59 1.60 1.85 1.28 1.50 1.48 1.48 1.54 1.38 1.49 1.14 1.71 1.61 1.60 1.55 1.53 1.49 1.17 1.39 1.53 1.21 1.47 1.47 0.85 1.61 1.57 1.53 1.46 1.57 0.88 1.37 1.40 1.17 1.08 1.51 1.41 1.29 1.16 1.14 1.08 0.68 1.26 1.08 0.72 0.68 Quadrant Tilt 0.77
- -0.50% +0 84% .
(. ,
-0.28% -0.06%
X.XX -Measured X.XX
-Design Minimum DNBR = 3.68 Maximum LHR = 12.134 kw/ft 7-19
_, . ----,m -..r-
U FIGURE 7.'7-1 INCORE MONITOR LOCATIONS 4 6-12 7- 6-2 31 h 30 3-4 2-8 1-1 4-1 32 @ 29h 28 C 52Q 5- L'. 8-8 5-1 'S-1 5-3 27h 51h 33 Q - 4-4 7-9 7-3 3-1 34Q 7C SC 26 Q 6-11 8-7 5-11 - 6-1 5-2 8-2 6-3 35Q 6Q 4Q 24Q 23 { 1-4 '2-7 2-1 1-2 36 Q 9h 8h 3h 25 C 22 Q 7-8 5-10 6-10 7-1 6-4 5-4 7-4 h 10 h 1h 2h i 21Q 2-6 2-5 2-3 1-2 11Q 19 { 20Q 6-9 8-6 5-8 6-7 5-5 Q-3 6-5 38 Q 39 { 12 Q 18 Q 50{ 3-3 7-7 40 13 16 17 h 49h 5-9 8-5 5-7 8-4 5-6 41h 14Q 15 Q 4-3 42h 43 41 48 6-8 7-6 6-6 - 44 Q LEGEND 46 45 Q Q - Sym etry Monitor Q - Total Core and Syn etry Monitor - DBNPS - UNIT 1 STARTUP REPORT h - Total Core Monitor ~ INCORE MONITOR LOCATIONS 1 X-Y FIGURE 7.7-1 IControI Ko'd NumDer T of-Control Rod Group X A~
-Incore Monitor Number 7 '
TABLE 7.8-1
SUMMARY
OF AXIAL OFFSET MEASUREMENTS , ICO OCO - % Power Backup Incore Of fset
% Power % Power NI-5 NI-6 NI-7 NI-8 % Power 41.2 -33.5 -31.7 -31.6 -31.8 -31.9 -25.6 41.4 -24.4 -24.7 -24.5 -24.7 -25.0 -19.5 41.6 -17.5 -16.5 -16.2 -16.5 -16.9 -13.6 41.2 - 8.6 - 7.1 - 6.6 - 7.1 - 7.4 - 6.6 40.7 - 3.1 + 0.7 - 0.8 - 1.4 - 1.5 - 1.1 41.9 + 1.2 + 2.3 + 2.8 + 2.3 + 2.1 Value not obtained 73.5 -29.8 -33.9 -33.0 -3~.0 -32.1 -22.1 73.3 -23.8 -27.1 - 2 6 . .'. -26.3 -25.3 -17.2 73.6 -17.3 -19.3 -if) .1 -18.4 -17.3 -12.0 73.7 -10.8 -10.9 - 9.7 - 9.9 - 8.8 - 7.0 73.7 0.8 0.5 1.8 1.3 2.5 1.5 ,
73.5 4.2 3.0 4.4 3.7 5.2 5.0 where: ICO = Top - Bot x 100% PTop + Pgg OCO = Channel Imbalance x 100% Channel Power DBNPS - UNIT 1 STARTUP REPORT AXIAL OFFSET MEASUREMENTS 7-21 TABLE'7.8-1
8.0 NUCLEAR STEAM SUPPLY SYSTDi (NSSS) PERFORMANCE The purpose of the tests described in this section is to monitor the performance of the Nuclear Steam Supply System (NSSS) to obtain baseline data and to verify the NSSS performs as cesigned. Four tests are used to complete this purpose. 6 Dele ted 8.1 UNIT LOAD STEADY STATE TEST, TP 0800.12 Primary and Secondary System steady state parameters were measured during power escalation to obtain baseline data. This information was gathered during Phase I of TP 800.12, " Unit Load Steady State Test", at approximately 0%,15%, 30%, 40%, 65%, 75%, 90% and 100% full power. Steady state condi-tions were established before any data was obtained. Several parameters were compared with design values to verify that the response for these parameters, as a function of power, was as expected. These comparisons are shown in Figures 8.1.1 through 8.1.7. Where appropriate, the recorded values were derived from an average of the measured readings. As shown on Figures 8.1-1 through 8.1-7, all parameters recorded responded within their acceptable bands. Phase II of TP 0800.12 was performed from 0% to 15% full power. Data was accumulated to check the relationship between Tave and reactor power. This information was used to adjust the OTSG low level setpoint to bring-Tave within 5,82 1 10F at 15% power. 8.2 NSSS HEAT BALANCE TEST, TP 0800.22 TP 0800.22, "NSSi Heat Balance Test", was performed during power escalacion with the intent of achieving the following objectives:
- 1. Verify the accuracy of the computer's heat balance calculation.
- 2. Provide baseline data for comparison with subsequent heat balance checks.
- 3. Determine the reactor coolant flow rate.
This test was conducted at power levels of 15%, 30%, 40%, ~ 65%, ~ 75%, 90%, i and 100% full power. Data for primary and secondary heat balances was i taken at each testing point. The balances were compared to the computer calculated heat balances. In all cases, the hand calculated and computer ; calculated values agreed to within the required 22% acceptance criteria. The results of these computations are summarized in Table 8.2-1. At 100% of full power, the hand calculated primary heat balances for each l loop were compared to their respective secondary heat balance. Since the deviation for both loops was greater than 1%, a new range for the primary
- 1
._... . flow meters for both loops, ha,s been calculated by setting th,e penary heac ,
balance eaual to the secondary heat balance. A retest was performed , to verify the deviation is less than 1%. ) [ 8-1 e
~ . . . . . _ . . - - . - - - - - - . . - - - - - - - - - - -- -- -- -
4
~.
8.3 INTEGRATED CONTROL SYSTEM TUNING AT POWER, TP 0800.08 This procedure was performed to verify that optimum plant performance and control is obtained by tuning of the integrated control system. Actual p ant trans ents fr a 00.23, Mt had Transient Test, were used to 6 evaluate NSS behavior. This transient data was carefully reviewed and g tuning adjustments were made to optimize plant performance. The major ICS related centrol functions tested are listed below:
- 1. Thermal efficiency between the primary and secondary system.
- 2. Electrical output versus feedwater fl'ow.
- 3. Feedwater temperature versus feedwater flow.
- 4. Steam generator startup level versus reactor power.
- 5. RCS inlet and outlet temperature versus reactor power.
- 6. Plant parameter signal levels which input to the ICS.
- 7. ICS capability to run the unit back to the desired load at the specified rate.
/ \ Selected functions are shown 'on Figures 8.3-1 through 8.3-6.
All plant parameters tested were within their respective acceptance criteria. e O 6 9 e 8-2
-- ' ~ ~-- - . _ . . _ . . _ . . . _ _ _ . _ _ _ . , , __
TABLE 8.2-1 1 HEAT BALANCE SUltfARY ' Nominal LOOP 1 LOOP 2 Power P1 (Comp) P1 (Hand) P2 (Comp) P2 (Hand) % DIFF P1 (Comp) P1 (Hand) P2 (Comp) P2 (Hand) % DIFF (%) MWt- MWt MWt MWt (P1-P2)(Hand) MWt MWt MWt MWt (P1-P2)(Hand) 15 193.56 184.63 0.32% 212.93 204.04 0.32% 30 427.44 435.98 414.54 443.40 0.27% 447.76 418.71 446.68 434.89 0.58% 40 564.80 572.82 599.39 588.80 0.58% 540.38 551.86 595.58 581.27 1.06%
-62.5 947.45 842.00 903.26 895.29 1.92% 817.24 824.77 906.05 887.905 2.28%
72.7 961.95 991.508 925.17 1027.33 1.29% 1042.70 951.058 1039.19 1020.04 2.49% 88.5 1151.82 1162.01 1269.53 1245.66 3.02% 1107.33 1127.35 1253.06 1227.96 3.63% l 100 1272.82 1287.09 1381.73 1380.78 3.38% 1226.92 1262.145 1374.40 1374.57 3.98% Whnra: P1 (Comp) = Primary computer heat balance P1 (Hand) = Primary hand heat balance P2 (Comp) = Secondary computer heat balance P2 (Hand) = Secondary hand heat balance DBNPS - UNIT 1 STARTUP REPORT HEAT BALANCE SUFMARY TABliE 8. 2-1
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> = - ==
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t .:=t = r.=sur =----
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L 20 40 60 80 10 0 NSSS POWER, PERCENT OF 2789 sth DBNPS - Unit #1 STARTUP REPORT FEEDWATER TEMP-ERATURE VS. POWER FIGURE 8.1.3 8-6 . g-. e i
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. r ._ , . _ . . . .tr: . .-- . t. . . . . . . . . . . _ . . . _ . _ _ _ _ . . _ _ . . . . _ . . . . . . _ _ .
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m .. -- -
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= -
LIMITS-==='==- - -- "- -~
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--== --- = n : : == =u . -- u =.=- .-_
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--- = :
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; tu -
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g
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O 20 ' 40 60 80 10 0 NSSS POWER IN PERCENT OF 2789 MW th 1 I DENPS - Unit #1 STARTUP REPORT STEAM GENERATOR OPERATING RANGE LEVEL VS. POWER FIGURE 8.1.4 8-7 l l l em
STEAM GENERATOR STARTUP RANGE LEVEL ( o PRESSURE) vs POWER
-::= = - :=:, ==.: w :; n : u ;.=-
O . _r n_-l m.=.p== _. . : r a- := - in = - :: =._:.. .u._a. : .- - =_:-- : := :== :.w : _ __:2 . _ _.. = = =: N . _
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Z 4 n. :. . t. . .:. :.:.:
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g : .=:c :=. ._.:.t: =c n =_:_:_.: .._- __u : .:
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v_ m.=..._ , Z --- = :=.: "r ..: . a==-,--- .:r u aua.._ -" ::- , , , _ . w -.c.:; :=. y .=.:.::- .. .=:t .- = :m. ::.. ,,=, LigI+*3 & e a u. ._2 =-_.. ._;. .. .=:'1.=. ._ . .:.=_:. :: ;, ,.-d: ".r4. -: :- ..q;= .=
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m .. _.:,:_ ...._. . ..,r . ._. . : a .. .: , . . L= . -- ,v..: .:. .=__.
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O 20 40 ~60 80 10 0 NSSS POWER IN PERCENT OF 2789 MW th DBNPS - Unit #1 STARTUP REPORT STEAM GENERATOR STARTUP RANGE LEVEL VS. POWER FIGURE 8.1.5 8-8 e
- LI .11TS Oil T-MI.T-MOT,AMD' T-COLD '.7tTH 4 GC PU".?S RU S!!NG
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