ML19329A928

Initial Startup Rept

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 I 4 1 1 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 l l OE ADuNE RETURN Dm '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 . e. 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 ) ) l i i l I .c i-TOLEDO mwmm h.s h d $ M U

TABLE OF CONTENTS Section

Page,

1.0 INTRODUCTION

1-1 2.0 SLHMARY 2-1 ~ 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 4-2 Measurement 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 6.1 Nuclear Instrument Overlap _ 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 e i

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 9.1 Turbine / Reactor Trip Test 9-1 9.2 Unit Load Transient Test 9-2 9.3 Unit Power Shutdown Test 9-3 l 9.4' Unit Load Rejection Test 9-3 l 9-4 9.5 Natural Circulation Test I l 9.6 Loss of Offsite Power Test 9-4 9.7 Shutdown From Outside the Control Room 9-5 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 ) 10.4 Peedwater Systems 10-4 11.0 UNIT MONITORING - CHEMISTRY AND HEALTH PHYSICS 11-1 ~ 11.1-Shield Survey 11-2 11-r-- ,-..p, p4. g._ w

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 e I l iii l )

~

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. l The turbine-generator is capable of a net electrical output of 906 MWe. The Nuclear Steam Supply System (NSSS) employs once through steam generators. The Facility Operating License (NPF-3) for DBNPS JJnit I was issued t 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 83 hours. Initial criticality was achieved on August 12, 1977, after a Post Fael Load Precritical Hot Functional Test Program. i ~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. 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. ~ 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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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 } 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 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 I l. (c) Pressurizer Operational and Spray Flow Tests, TP 600.13 (d) Control Rod Drive System Operational Test, TP 600.17 (e) Reactor Cociant System Hot Leakage Test, TP 600.10 (ST 5042.02) l-i 2 - d

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. 2.4 CORE PERFORENCE DURING ZERO POWER PHYSICS TESTS, TP 710.01 (SECTIOM 6) 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

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 (b) Unit Power Shutdown Test, TP 800.15 s (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. The secondary systems that are covered include: 4 (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 m-

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2.10 UNIT TRIPS (SECTION 12) 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: (a) " Post Refueling Physics Testing", ST 5010.03 3 (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 The Davis-Besse Unit 1 initial core contained 177 fuel assemblies, 53 control 2 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 27, 1977. Actual fuel loading ti=e was approxi=ately L-1 at 1808 on April 83 hours. The fuel loading was perforced in accordance with " Initial Fuel Load Procedure", PP 1302.04. The actual fuel loading sequence is illustrated l in Figures 3.0-1 through 3.0-8. The initial core configuration is shown in 2 Figure 3.0-9. rate was monitored con;inuously and 1/M plots were maintained The neutron count 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 allowance for un~ certainties. i-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 inserted into the dummy asse=bly. After a visual inspection of NJ0046, fuel e 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 i 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 R P O N M L K 11 G F E D C B A 1 NI-1 2 O 3 4 3 0 l a I 7 B 5 25b / 25a 9 e / / 24 23 20 10 B ' 22 16 17 19 11 11 13 14 18 21 12 5 7 9 12 15 13 3 4' 8 10 14 NI 1X 6 15* I I l \\ g Source Assemble Startup Report' '@-Auxiliary,NeutronDetectorA Fuel Loading Sequence + Figure 3.0-1 ~ h Auxiliary Neutron Detector B 3-3 r m

Figure 3.0-2 Davis-Besse Unit 1, Cycle I L Fuel Loading Sequence'2 of 8 I N P. P O N M L K 11 G F E D C B 1 / NI-1 2 b. -O 3 l 4 1 ,5 6 l l 43 45 36 38 40 42 31 33 35 41 46a 30 34 39 44 ~ 27 32 37 29 26 j /,u 28 pfff// l n o 1 NI-2 15" I p Source Assembie DBNPS - Unit 1 ~h-AuxiliaryNautronDetectorA Startup Report Fuel Loading Sequence h Auxiliary Neutron Detector B -Sure 3.0-2 2-s e'

i Figure 3.0-3 Davis-Besse Unit 1, Cycle I Fuel Loadinsi "equence 3 of 8 j N j s R P O N M L K 11 G F E D C B A } l 68 3 64 66 67 71 4 57 59 61 63 70 72 50 52 54 56 62 69 73 / 47 49 55 60 65 , [ [, '48 53 58 l/ l / 51 7 2/7/7###/W/ 7/ 7 # EA % % 7/ G 7####/W#/ l / / W%7 04 o 14 NI-2 15' Scurce ? -e::tble DBNPS - Unit 1, . Startup Report h-AuxiliaryLautronDetectorA Fuel Loading Sequence. l 'I"#*

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h Auxiliary Neutron Jctector S >3 u.e,--

~ Figure 3.0-4 { Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 4 of 8 N i k t R P O N M L K T1 G F E D C B. 'A F* '92a 91 90 89 88 87 86 85 84 / 83 82 81 80 (2k 79 78 77 I / YAV,Y4VMA% ,4 bYMAVMA% Vf/%V###A 7//%7##A% WA%7M/%D VA%%%7E/%G ~ Z%7##A% Vff/% o '4 NI-2 15' Source Assemble DBNPS - Unit 1 h-Aur.iliaryNeutronDetectorA .Startup Report Fuel Loading Sequence I"#"

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h Attx111ary Neutron Detector B 3-6 N'

Figure 3.0-5 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 5 of 8 ~ N i ~ R P O N M L K 11 G F E D C B A 123 122 121 120 119 118 117 116, / 115 114 113 112 / 111 110 109 103 107 / 106 105 104 103 102 / /, [ 101 100 99 98 [ 97 96 95 94 _93 / l %%W/X6%7A = 7p/ws/% W E AW< o 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 R P O N M L K 11 G F E D. C B A i ~ 0 / Y# #/3' /V/% 7X6 "'~b VYMX/#^ XM/%%i%%

1. (4%7#X//7#X/%%

7#'l/5%%%V/%%%VX4% ~ 'b'W##//Y XMX4%%% A'#//% V/ /% % % % %% %Y# #/% VX4 % % % / / / / / / ? l 124 125 / / 126 127 128 a5 / / / /, 129 130 131 132l 133 134 135 136 137 138 139 140 1414 p Source Assemble D3NPS - Unit 1 Startup Report h-Aur.iliary, cutron Detector A Fuel Loading Sequence Figure 3.0-6 h-Auxiliary Neutron Detcetor B 3-8

Figure 3.0-7 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 7,of 8 N R P O N 11 L K 11 G F E D C B A l i 'M/9% VEd % /' % M "'~ W/%?MWMM/7/W,,' b ~ '1 4 c-- '@ %%V/%%4W/VX/WO / / ~ [ ,/ / / 143 1 / / / 145 146 147 / . / 148 149 150 15 / 153 154 155 15l6 / j jl52 / / 142 157 158 159 160 161 l 1,;> '4%' # # # / V x Z % 1e2 1e> 1e4 1e> / / / / / /.166 167 168 s [ [ ', / / 169 170 171 s j / / l l 7 172 173a ~~ 1s' f/,7,7,/Q Q Source Assemble h Auxiliary N:utron Detector A up R por Fuel Loading Sequence - h Auxilia[y Neutron Detect.or B Figure 3.0--7 .e*m. ar m.,-

Figure 3.0-3 Davis-Besse (Jnit 1, Cycle I Fuel Loading Sequence 8 of 8 N

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L K II G F E 6 C B A I a [ [175a Y#/##M/#####/ WHMMM/####/ ~ WE(#XA/#######A A%7#HM/#M/ZHEM/ Uh%%%%'#/H#MMA%%% ~ %%7########AWHA% %%%%%V#######M// %%%%f/#>/##/MHA%%% [ %7#M#####/##/G WEEEEEN/EEEA c o 14 [ (( NI-2 174 I 176b - Source Assemble DBNPS - Unit i h-AuxiliaryEleutronDestector.} a cup R ort E " Auxiliary Neutron Detector B

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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 RCS and 2155 psig to determine the core flow characteristics. 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 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-termined to verify that the minimum flow requirement for three pump and two pump operation respectively were surpassed. After th,e BPRA and ORA were removed, the RCS flowrate with all four RCPs in operation 3 was re-determined. The acceptable criteria were again verified. The results of the flow measurements are compa red with their respective acceptance criteria in Table 4.1-1. As shown, all measurements were j 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 4.1, only the oost fuel load measurements are discussed in this report. i l= The flow coastdown test for a trip of one of four RCPs was repeated I 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

[ leak rate was then adjusted for the unidentified RCS leakage measured f 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-The mately 532 F and 2155 psig with 2 and 4 RCP's in operation. 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 Test core using the auxiliary power supply trip C and D switches. 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 The actual power interruptior to 2/3 insertion was also satisfied. 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 No. Pump Combination GPM CPM GPM 102,400 1 1-1 1 1-2 99,30C 1 2-1 103,800 1 2-2 105,600 186,390 2 Lower flow pump 172,500 in each loop 1-2, 2-1 3 Three lowest 262,000 284,180 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 3 4 1-1, 1-2, 2-1, 2-2 387,200 429,400 403,818

• 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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s l f 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 ( DE-ENERGIZES 2155 2.55 1 16 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 bBNPS-CaIT1 STARTt'P REPORT PRESSURI"ER SPRAY AND HEATER ACTUATION SETP01NT D.\\TA TABLE 4.4-2 4-12 i.

9 PRESSURIZER SPRAY FLOW CALCULATIONS Spray Flow = Q (VRCS) ~ a h(K)

where, Q = heater input (KW)

RCS = specific volume of RCS cold leg (ft /lbm) Y 3 2.3Sx10-3 ft - min - KW K = constant gal-BTU / Ah = enthalpy of pressuri:er - enthalpy of spray water (BTU /lbmT RCS RCS Pressurizer Pressurizer Calculated Pressure Temperature Temperature ' Heater Power Flow Pressuri:er Spray 1280 psig 532 F 577 F 1224 KW 190 gpm Pressuri:er Mini-Spray 2150 psig 532 F 647 F 30 KW 1.6 gpm DBNPS - UNIT 1 STARTUP REPORT PRESSURIZER SPRAY FLOW CALCULATION TABLE 4.4-1 4-11

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t T' N 19 3 3 3 3 3 33 '} N N -M NH in

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C C CC H N N O C C CC e M m M MM N N N n m <m e e e b k k k b b 4 b .o u o co o o o 1.1 I I I I I I I I Q "AL q, 0.9 .r. m = u L= so 0.7 0 A 2-0.5 23 v c i 3 2 4 6 8 10 a Reactivity Added by CRA Withdrawal (% A k/k) l.1 a G C. o .01 z { t---- =q g-y e, 5 0.9 o-- m O Lz a 8 0.7 a T.l w 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

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.

1/M vs. CB 1.0 .a

T g l

0.8 1 0.6 T E a -T-0.4 5 O g 0.2 U A E l O O j 1800 1700 1600 1500 RCS Boron Concentration (ppm) 1.0 ^ c. O ^ T -- =: 2 E 0.8 g E 0.6 7 = m s-0.4 '3 3m Q 0.2 'E C 1800 1700 1600 1500 RCS Boron Concentration (ppm) DBNPS-Unit 1 Startup Report l 1/M vs. Boron Concentration Figure 5.2-1 5-4

-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 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 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. 6.2 SENSIBLE HEAT DETERMINATION I Prior to zero power physics testing, the intermediate range level at which 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. Nuclear heat was determined by increasing power in set increments, allowing power to stabilize and then observing whether sensible heat was produced. 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 )

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 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. 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 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. ~ 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 ~ ~millif hica ~ l pcm = 10-5 ok/k). ~ " worth of the inserted rods vis 54 pefcent ~ ~ ~ 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 due to each rod movement. Integral coctrol rod worths.were obtained by l' 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 -initial rod configuration. 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- -8 ~7 NI-4 Intermediate Ranae Level (amps) R812 1.140 x 10-9 3.690 x 10~ 1.175 x 10-8 3.606 x 10 1.130 x 10 RCS T ( F) 'T709 530.4 530.3 530.3 530.4 530,7 ave 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 STARTUP REPORT N 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. I l 6-6 l

REACTIMETER RESPONSE CHECK 0UT 1 INITIAL REACTIVITY CHECKOUT DESIRED CONTROLLING CRA GROUP AVERAGE REACTIVITY DEVIA-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 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% -3.34 - 4.4% CRA 6/7 (100 to 0) - 3.20 (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 i j. 1 l 'E DBNPS - UNIT 1 STARTUP REPORT 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 -5104 -4687 1790 CRA Group 7 @ 100% ud 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

• Linear Fit Y= 0.6119X - 6342 Y= 0.8149X - 6519 Zero - Intercept 10,365 7,530 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 l DIFFERENTIAL BORON WORTHS 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 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 '~.'m C. v m a 1 -u y I I I N [3 d h .m 'n i.

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I lil l I lill I ll Illi lll1 ..,l_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 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 c. i i N g g. 4 u , x .m. i, ,n . i, w n i. 1 .u... 6; 9; I g;,j ig6 .;9 gilj ,g , ' Q.l6 e j.g3 I l'p,! Ii; 4 'g 44 .t O is .e .i-i vi i !st i .t. 6 . ?t lii st i, i! iiii 'IN '.'t~ i.! r iI. % . i a .i. 6 , 6 6.t in, on ..n on n. on .1 o ,o .,o .n. i. on it, pi !! i sie! ii lei lisi iii' iiii li!* !! siti s t i,. i'it list sii' !!ai lie a!!! itii titt 9.'s n 'i i.'s it!. t.' i ni n.n un on iin o., nii ,.o on .!n o, .in u,, i., oit u,6 on in ou iin no sin i

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 i l li !HI Hi! 1111 ti i INI il I Ilil NH ll li i lill lli 11 1 I HI n't tile ilh till iill !! I Uli 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- ~ ~' 10 c, a. .+_ M. 0 8 n , m- - - y. / w f 0 Qi , j n -a -100-4 d / n 1

-+.

s E S u o v e ,'1 L c. Q. n 1 2 = = H ec Hm O C 3 -200 O ._s 4 a + w e

/.

s s -2

4 x

M -.e. w m r.4 x N -300 ^ -4 5 1 -6

h W

J-( e m. -400 -8 w I f3 I*y 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 i 0.0 140 f ~~ 120 7,. -500 ~ 'Eis _100 ."",3W -1000 .[' E' ~ .-dE-9 = d m 80 0 8 3 @~ 5 S ' l 5 -1500 =- 60 '.a + 'l I 'z. .-~ 5. 2 8 g 5 -2000 .]'. -= 8 m-40 g g = m ~ a b-E_ 20 -~ -2500 -5 B_. '^ E E c 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-Y n 8 -6000 -*y ^ S t u -m 't h 5 .QCl x '~ U _x, x1 _ ~. N-d -4000 'N.;_ e XL ^ H x _, t l \\! N 5 v

x x

.w + U 'N ~' N' d -2000 '.X t= 'N., = --t_ 'N: .lQ-

-'x 4-y k

-l-3gd-E 2000 4000 6dOO 8000 10,000 EQUIVALENT BORON WORTH (pcm) i DBNPS - UNIT 1 STARTUP REPORT SHUTDOWN MARGIN FIGURE 6.8-1 6-U

s. FIGURE 6.9-1 - DIFFERENTIAL BORON WORTH VERSUS BORON CONCD;TRATION i , ie, ,, e e .., i.... ..,i,,i.. i i. 1 . i.. .. i.. I -1,k g... e e.. j ..,. 6,... 4 .., I t., 6 ,, e. t i i. j . i. . i.. i i., a..,, 1 I e 6 l i 4. 6 . 6. t l ,.. l 6 6, 6. 1.. t , 6 3... 6 6... , f. , i,, s .., i i. -1.3 1 ... e i, i. , i i. i... ... i ... i i m S g, . s. i,, i . i t...

1. .. e i,

e.. e.., i e.. g, 6... . i. l .. {. .. i. N . e i,. I i , f. A. ^ e . ei .. e e ,e e. it. i 4 i .. e t.. i a.. 8 l i.., l l Nj i 4.. e., * .. t. ,t., i t.... .. i, t,., .. l. .i . 1.. t... .. I <s -1. 2 .6,. ., 4. t.,, .e .( ,,. s . t l. 5. s . $ e i N i. .t ,, t e t.. ...t er 1 i. 6 , i.. ,,,a 6 ...i....... .. we.... 4.. ... 6 1 . 4,, .. i. p e- ..m g -1.1 ~ i i.e. .. e. w . i.. ...., t,. ,..,..... i.. .#,.-w i... i... t... .. <i ..i,... .. a.. CALCULATED VALUE . i,, m. l 3., a t i ....n.... .i.. a . i.. ... sc... i i... .s. 4 ~ -L.0 ...i i,, i i...... a m.i... m i v ...r us ..i. i, t. ,h. m .. i.i , i.. .......s.. i,. -0.9. e, e.,. ....,,, i ,w. . i.4 .. i. .. i.6,.., i. r . i. i i. . i. .. i, , 6 ...e i e.i i.. i i, .t i... .... i. i., r.... .ii.

eei, ei

+,,. 4. e,. 4 4 i,,. , i 4 6 . +. 6 .. i i se i e. , i i. . i. -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: {APIo-API]<,3.5% eq. 7.1-1 R where: RTP = U td Thermal Power TP = Thermal' Power APIo = Out-of-core Axial Power Imbalance API = Incore measured Axial Power Imbalance 7 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. Af a n Af. ATF + g.m I bT a P. aT AP M F eg. 7.2-1 or: AT 6TP m -r W Mp fu 6 JM gp D ~q eq. 7.2-2 . here'A T,and OT are the temperature changes of the moderator and fuel w y 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 0

AK/K/ F 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 RATED THERMAL POWER.

0 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. 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 measurements were also shown to be within acceptab1'e limits. q 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 O "t N NI-2 PC \\ UCIC 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 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

4 4. TP 0800.20 was performed to determine the differential worth of group 6 after the swap. 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 f a g 800 H Q N g h g 600 5 ^ = 20 40 60 80 100 % FULL POWER i i l ( l ..\\ DBNPS - UNIT 1 I STARTUP REPORT AVERAGE FUEL TEMPERATURE 7-12 vs. REACTOR POWER FIGURE 7.2-1 w

NUCLEAR INSTRUMENTATION FLUX RANGES - 105 10' 10 -10" 9 10 -103 10 3 8 10 - 125 - 10 " 100 0* - 102 E# ~ 7 10 - m aoc -101 5 _ 10 g5 So - EU b e 3 a. - 100 6 gg _1 g m 3c c 105_ g n 7

• 3

.5 10-1 y u a m 10" - d5 e .o e 3 2 8 gg o

• 3 5

3 z 10 - t u e 6 3~ 8 3 E -10 - to-9 cc uu 2-10 -10 10 5 I 10 - 5 -10 10-11 4 m e - 0 2 0 10 - o - 10 6 -103 m

• 8 u Ei

-1 S 10 ,8 - 10 7 -102 m 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 2.234 kw/ft Maximum LHR = 7-14 m

FIGURE 7.4 -2 CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Design s 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 8 at 35.5% wd 28.8 % wd Imbalance -0.89 %FP

-0.16 %FP 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 2.234 kw/ft Maximum LHR = 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 6.793 kw/ft Maximum LHR = 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

-0.50% +0.84% f -0.28% -0.06% i l l Xg -Measured X.XX -Design Minimum DNBR = 3.68 Maximum LHR 12.134 kw/ft = 7-18 ~

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% -Measured X.XX -Design Minimum DNBR = 3.68 X.XX Maximum LHR 12.134 kw/ft = 7-19 r

---

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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 33 Q 27h 51h 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 21Q i 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 17 h 49h 40 13 16 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 45 Q 46 Q - Sym etry Monitor Q - Total Core and Syn etry Monitor DBNPS - UNIT 1 STARTUP REPORT h - Total Core Monitor INCORE MONITOR LOCATIONS ~ FIGURE 7.7-1 1 X-Y 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 tuning adjustments were made to optimize plant performance. g 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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j 9.0 UNIT PERFORMANCE DURING TRANSIENT AND ABNORMAL CONDITIONS The purpose of the unit performance tests is to verify that the unit can be maintained in a safe condition during and following load transients and various abnormal conditions. A reactor trip test was completed at-the 40% of full power level. A turbine trip from 75% of full power was also completed. A unit load rejection test from 75% of full power was performed. A unit load rejection test was conducted at an initial power level of 100% of full rated power. The plant was shutdown from the Auxiliary Shutdown Panel from an initial power level of 15% of full rated power. A loss of offsite power test, including a loss of external load, was conducted from an initial power level of 15% of full power. 1 Load transient tests inciuled 10% full power transients at 40% and 75% of 6 full power. The 50% full power transients at 90% of full power were also completed. Proper operation of the integrated control system cross limits and rate limits were also verified. A natural circulation test was performed with the reactor at 3.8 - 3.9% of full power and all Reactor Coolant Pumps tripped. A unit power shutdown test was performed on October 18, 1977, from an initial power level of 15% of full rated power. 9.1 TURBINE AND REACTOR TRIP TEST TP'800.14 The reactor. trip from 40% of full power was successfully completed on December 15, 1977. The reactor trip was initiated manually, and the Reactor Trip Emergency Procedura, EP 1202.04, was implemented. A 75% of full power turbine trip was completed on April 2,1978. This pro-vidad more data to optimize the operation of the Integrated Control System. The 75% turbine trip was repeated on September 10, 1978, to test the blow-back of the main staar. safeties and changes made to the Integrated Control System during the BPRA removal maintenance outage. The main steam safeties operated properly during the test but the ICS displayed the need for further tuning. As,a result of the analv-is of the ICS ' performance af ter the trip,, ^ -~ adjustments ' vere made tio the I'CS. The load rejection tests at 75% and 100% of. full 7 aver verified proper opera, tion (see Section 9.4). b l-l 9-1 ~.

.y Data on a turbine trip from 100% of full power was obtained from the " Unit 4

Load P. ejection Test", TP 800.13. A reactimeter and brush recorders were used to record the applicable data. The results of the reactor and turbine trip tests are summarized in Figures 9.1.1 through 9.1.10. The collected data verified that the reactor coolant system remained within its safety limits. 9.2 UNIT LOAD TRANSIENT TESTS TP 800.23 1 The unit load transient tests demonstrated that the unit can be maneuvered l in a controllable manner at 5% FP per minute from 15% to 75: and from 75% i to 15% of full power. The unit load transient tests also verified proper l operation of the ICS cross limits and verified satisfactory low power level l response of the ICS to control the unit subsequent to the tripping of one reactor coolant pump or one main feedwater pump. Load transients of 10%~FP were conducted at 5% FP per minute from 40% to 30% to 40% of full power in the integrated control mode and turbine following mode, and at 3% FP per minute in the steam generator / reactor following mode. Af ter tripping Reactor Coolant Pump 1-1, a load transient of 20% FP was conducted at 5% FP per minute from 40% to 20% cf full power in the integrated control mode. Reactor Coolant Pump 1-1 was re-started and a load transient of 20 FP was conducted at 5% FP per minute from 20" to 40% of full power, also f n the integrated control mode. ICS cross limits were verified by: 1. Imposing a 5% FP per minute' load transient from 40% to 50% of full power with the feedwater control subsystem in the manual mode of control. The feedwater-toereactor cross limits limited the reactor demand to 45% of full power when the feedwater demand exceeded feedwater flow by 5%. 2. Imposing a 5% FP per minute load transient from 40% to 50% of full power with the reactor control subsystem in the manual mode of control. i The reactor-to-feedwater cross limits limited the feedwater demand to 45% of full power when reactor demand exceeded the neutron power by 5%. 3. Imposing a 5% FP per minute load transient ' rom 40% to 30% of full power with the reactor control subsystem in the manual mode of control. The reactor-to-feedwater cross limits limited the feedwater demand to 35% of full power when the neutron power exceeded the reactor demand by 5%. Load transients of 10% FP were conducted a' 5% [P per minute from 75% to 65% to 75% of full power in the integrat".a cuatrol mode and turbine following mode, and at 3: FP per minute in the steam generator /reacter following mode. With the reactor operating at 64% of full power, Main Feedwater Pump #1 was tripped, and the ICS initiated a plant runback to 59% of full power at 50% FP per minute, in the integrated control mode. TP 800.02 l 9-2 ~-

A list of unit load transient testing is. given in Table. 9.2.1. A reacti-meter and brush recorders were used to record the. applicable data. The collected data verified that the unit can be maneuvered at 5% FP per minute in the integrated control mode, af ter optimization, without a reactor or turbine trip, relief valve or turbine bypass valve actuation, or exceeding .any of the limits imposed by PP H01.01, "NSSS Plant Limits and Precaur'ons", thus satisfying the acceptance criteria. The high power positive rate limit was verifica 'oy imposing a 5% FP per minute load transient at 85% of full power in the inte-TP 800.23 grated control mode. 6 Load transients of 50% PP were conducted at 5% FP per minute - from 921 to 42% to-90% of fun power and at 3% FP per minute from 90% to 92% of fun power, in the integrated control mode. TP 800.23 Load transients of 30% FP were conducted at 5% FP per minute from 60% to 30% to 60% of FP in the integrated mode during three pumpTP 800.23 operation. 9.3 UNIT POWER SHUTDOWN TEST TP 800.15 The unit power shutdown test was performed to verify the adequacy of the Plant Shutdown and Cooldown Procedure, PP 1102.10, from 15% power to 0% power, and to obtain baseline data for subsequent shutdowns. j The shutdown wa.a performed from t.1 initial power level of 15% of full rated power. The cooldown was conducted to a final reactor coolant sys-tem temperature of $31*F and the reactimeter data was obtained by the Plant Computer's Ope:ator Special Summary Group. The results of the unit power shutdown test, summarized in Table 9.3.1, verified that a Turbine-Reactor Shutdown can be performed (Section 4 of Plant Shutdown and Cooldown Proce-dure, PP 1102.10) without exceeding the limits of the Nuclear Steam Supply Limits and Precautions, PP 1101.01, Section 1. Thus, the acceptance cri-teria was satisfied. 9.4 UNIT LOAD REJECTION TEST TP 800.13 The purpose of the Unit Load Rejection Test is to demonstrate the unit can be satisfactorily controlled when a loss of load occurs and to assure no ' Technical Specification safety limits are exceeded during or fo n owing the load rejection. On November 11, 1978, the Unic Load Rejection Test was performed with the - unit at 75% of fun power. At 0009 hours, the test was initiated by open-ing the main generator outlet breakers.. The Integrated Control Systen oper-ated properly to initiato a runback to approximately 15% of fun power, re-( ducing both reactor power and feedwater flow. Both the main steam safety valves and the turbine bypass valves maintained main steam line pressure within-limits, and the turbine speed returned to 1800 RPM. 1 9-3

At 0039 hours, the generator was synchronized with the grid by closing the main generator breakers. The test was successfully completed without vio-lation of any Technical Specification safety limits. A reactimeter and brush recorder were used to record the applicable data. The 100% of full power Unit Load Rejection Test was performed on January 14, 1979. At 0000 hours, the main generator outlet breakers were opened to initiate the test. The Integrated control System ran back reactor power to 15% of full power, reducing reactor power and feedwater flow as designed. The main steam safety valves and the turbine bypass valves maintained main steam line pressure within limits, and the turbine speed returned to 1800 4 RPM. The generator was re-synchronized and the main output breakers closed at 0052 hours, January 14, 1979. The tests' successful completion demonstrates that the unit can be satisfactorily controlled when a large loss of load occurs. All acceptance criteria were met and no Technical Specification safety limits were exceeded. Further details of the 'results of the 100% Unit Load Rejection Test are depicted in Figures 9.4.1 through 9.4.7. 9.5 NATURAL CIRCUIATION TEST TP 800.04 The purpose of the Natural Circulation Test is to verify that on a loss of all forced reactor coolant flow, natural circulation will provide adequate core cooling for all possible levels of decay heat generation. Since under natural circulation conditions cold leg temperature (and therefore density) is significantly different from the value at which the power range instrumentation was calibrated, Phase One of this test deter =ined a corree-tion factor for indicated power. On October 30, 1978, Phase One was com-pleted. At 1920 hours on November 3, 1978, the two operating Reactor Coolant Pumps were stopped with reactor power held at approximately 3.8 - 3.9% and a steam gen _;ator water level of 160 inches. Natural circulation flow was computed with the use of a heat balance to be over 5% of full flow, which is over three ti=es the minimun required flowrate. Steam generator level was redu:ed in five increments to the low level linit, each tLne computing the natural circulation flowrate. The lowest measured f1-wrate var 4.64% of full flow which is over 2.5 tLaes the required flow-r?te. The reactor 7as held at 3.8 - 3.9% pcuer for over seven hours on natural cir-culation with no operational limits violated. Further results o the Natural Circulation Test are depicted in Figures 9.5.1 through 9.5.3. S.6 LOSS OF EXTERNAL LOAD INCLUDING LOSS OF OFFSITE POWER TEST TP 800.26 The purpose of the Loss of Offsite Power Test is to demonstrate that the statior can be maintained in a safe condition following a loss of offsite power. 9-4 b

The loss of offsite power and loss of external load test is initiated by isolating the switchyard from the startup transformer supplying unit housepower. 1 The test uses a reactimeter and brush recorders to record the applicable data. At 1025 hours on January 15, 1979, with reactor power at approximately 15%, switchyard breakers HX01A and HX02B were opened isolating the unit's supply 4 of power. The loss of housepower resulted in the automatic starting of Emergency Diesel Generator (EDG) 1-2 which supplied essential bus Dl, as well as initiating Reactor Protection System (RPS) and Steam and Feedsater Rupture Control System (SFRCS) trips as designed. The SFRCS trip initiated auxiliary feedwater flow to the steam generators as designed. The automa-tic starting of EDG l-1 on a loss of power was also verified during.the conduct of the test. EDG l-1 was running supplying essential bus C1 at the time the loss of power was initiated. All acceptance criteria were met, and no Technical Specification safety limits were exceeded. 9.7 SHUTDOWN FROM OUTSIDE THE CONTROL ROOM TP 800.25 The purpose of the Shutdown From Outside of the Control Room Test is to demonstrate that the plant can be safety shutdown and maintained in a hot standby condition from th'e Auxiliary Shutdown Panel, C-3630, and to verify the adequacy of the Emergency Operation of the NSS Procedure, EP 1202.23. i The test is performed by tripping the reacto.- from outside the Control Room at 15% of full rated power, and establishing the plant in a safe, hot standby condition from the Auxiliary Shutdown Panel, C-3630. 4 The Unit Shutdown From Outside the Control Room Test was conducted on January 14, 1979. The test commenced with the tripping of the reactor remotely at 1015 hours. Operations personnel at the Auxiliary Shutdown Panel then brought the plant to a stable hot standby condition. The test was successfully completed; all acceptance criteria were met, and no Technical Specification safety limits were exceeded. t 9-5

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9 ~ 8 8 8 8 8 8 8 8 S 8 7 7 7 7 7 7 7 7 i 1 EE 7 / / / / / / / - / / TH /5 85 8580850080 85 85 90 t AI 451 1 1 11 21 31 21 1 1 0 1 21 2 i DT /0/9/9/2/2/J/3 /4/474 J1 J1 Jntr 11 1 0 1 0 11 11 I 1 1 1 1 1 I o ep s e sR e C B p I - u S st P i r NJ va M 0 O I 8 2 9 2 8 3 2 2 8 7 at 7 I ! 2 1 1 2 1 1 1 DS 8 NTE UAB T HI N I V t XE o ADE 5 r M V E t 6 5 1 5 0 6 8 6 3 0 T PO S i B 2 2 4 2 1 4 T A, F O

• lA l

5 5 F Nl 0 0 2 0 3 2 7 0 0 0 Y 0 OE I B 1 1 0 0 0 2 1 0 1 3 R 2 l T A 3 t l 5 UA ee HI ' t Tt V e S N XE I 1 D t T O I 1 S P .E E T CV 5 T E VO 0 5 2, 0 0 5 0 5 5 0 2 S AB l 1 1 1 0 3 2 0 3 9 T TA 1_ N E E L I e B S A N N T A I R HE T C 2 ZA 5 D R 7 A EE f Z 2 1 2 1 Z 1 1 D TV s 5 5 3 2 5 5 S 5 0 I AA 5 B T I NU 5 Z EI 2 5 C 2 2 Z 1 1 6 7 Z Z Z H 0 2 5 5 5 5 5 9 A 7 6 7 6 7 o o 6 7 5 liO t t CT o o o o o o o o t t t t t 2 t t t F 2 2 Z Z 5 5 PI 1 2 0 5 5 5 5 2 2 5 5 4 Z 6 7 6 7 6 7 6 7 6 6 e s s d d d e F o n n W ON M a a H O m m P 5 EI d e e F 2 DT e a g g gdl dl Dl Dl H 9 OA t n n n n na na aT a u u f k HR a ei ei i i a u au W P E r nw nw w w mn mn F nI n oc F SP gl o io Co Go e a e b a a CO e bl bl Sl 5l DH Dm hl htI eb RF3RF /l l t a t sn Z I t rl rl xo o xn xn on 10 n ou n uo uo I TF TF F I RI BI 1 I l P T' ~ N EI R SE NB A 6 7 8 9 0 1 2 3 4 5 R RI 1 1 1 1 2 2 2 2 2 2 TN 't e$

I 1 EE 5 8 8 8 8 8 8 8 8 8 8 8 8 8 TH I 7 7 7 7 7 7 1 7 7 7 7 7 7 t /0 075 /5 /5 /5 /0 /0 /0 /0 /0 /0 /0 i 5375052 AT /0 64 61 63 65 95 91 93 94 91 93 nt DT 30 /2 /2/3 /3 /0 /1 /2 /2 /9 /0 /2 /2 /3 /3 U2 8282 82 80 80 80 80 90 91 91 91 91 91 U ro ep s e sR e C W B p I N D - u S O I P I E 0 6 1 5 3 6 8 6 3 1 5 9 3 0 st MT B 2 1 3 3 1 1 4 i r 0 UA va 7 MI at 8 I V DS XE T AD N H I P O N E P T V 4 0 5 8 0 5 6 5 2 7 9 3 6 6 T O 2 5 3 1 3 E B S A F W O 0, 0 0 0 5 0 5 0 2 4 1 3 8 0 0 F I Y 0 N E R 2 O B 1 1 1 0 1 0 1 2 0 0 1 0 1 1 A 8 MI UT H 5 ll H A U T I I S N XV I AE T O HD T E C O 0 0 0 '1 0 0 1 0 1 1 0 0 0 " 4 3 2 5 2 0 2 5 7 2 0, 5 2 7 S P E . V 1 E T 0 2 S V B T A A n 9 N T E E I L S N B N I E A A HG T R A T %R 1 1 E 5 5 5 5 3 3 5 5 5 5 5 5 3 3 D EV A TA D AR I T I NU EZ G .N 0 0 0 0 0 0 0 0 5 5 5 5 5 5 A 3 4 3 4 3 4 3 4 5 7 5 7 5 7 II O CT o o o o o o o o o o o o o o t t t t t t t t t t t t t t PI F 0 0 0 0 0 0 0 0 5 5 5 5 5 5 4 3 4 3 4 3 4 3 7 5 7 5 7 5 o s . d d n n F a a m m e ON e e d d W O d d M EI e e g g g g Dl Ol e e g g g g DT t t n n n n a a t t n n n n i i Wu Wu a a ei ei 5 OA a a ei ei i i 2 HR r r nw nw w w F n F n r r nw nw w w a a g g i o io Go Go 9 E g g i o ioGo Go SP ee ee bl bl Sl Sl hH hH ee ee bl bl Sl Sl P CO t d td rl rl /l /l t t t d t d rl rl /l /l F I no no uo uo xo xo on on no no uo uo xo x o I M I H TF TF RF RF Bi Bi I M I M TF TF RF RF TN ER I E SR 6 7 8 9 0 1 2 3 4 5 6 7 8 9 NH 2 2 2 2 3 3 3 3 3 3 3 3 3 3 AU RN T Y* ~

TABLE 9.3.1 UNIT POWER SHUTDOWN TEST

SUMMARY

Initial Final , Parameter - Value Value Test Commenced / Completed 1009 hrs. 2359 hrs. Reactor Power / Source Range Level 15% FP 0% FP RCS Iave/RCS Tc WR 581 F 531 F RCS Pressure 2150 psig 2150 psig Pressurizer Level 195 in 188 in Makeup Tank Level 62.5 in 71 in SG #1 Operate Level 7% 47% SG #2 Operate Level 6% 46% SG #1 W Flow 0.55 x 106 lbm/hr - 0 lbm/hr SG'#2 W Flow 0.54 x 106 lbm/hr ~ 0 lbm/hr SG #1 Steam Outlet Pressure 882 psig 871 psig SG #2 Steam _ Outlet Pressure 888 psig 873 psig Feedwater Temperature-299 F 230 F RCS Boron Cc' centration 1339 ppmB 1304 ppmB RCS Contraction (Makeup) Volume 4000 gallons l l 9-9

7 l l 1050 y T-- _,t- ~ m .e __+ l l m t E 1000 b ~ ~ ~ ~ ^ t- --/ m / a: --f v. b y -S $ 950 ' o i r-x d l s i m f I e4 i 3 i l i e 900 l l l I 4 i i e i 1 850 m. 1 2 3 4 ue ELAPSED TDfE FROM TRIP (MINUTES) DBNPS - Unit #1 STARTUP REPORT SG #1 OUTLET PRESSURE VS. ELAPSED TIME FROM TRIP FROM 40% FP FIGURE 9.1.1 9-10 _s

1050 E l l n o

'q.

m l 6 1000 ~ ~ ' - h-fg E - J-M_ l .a s D 950 a H 1 i 2 i ~ I I t 6 i I H 6 i i i N 900 8Da l l __ c i ~ i I l 6 6 6 i i I + + i 850 t 1 2 3 4 s ELAPSED TIME FROM TRIP (MINUTES) DBNPS-Unit #1 STARTUP REPORT SG #2 OUTLET PRESSURE VS. ELAPSED TIME FROM TRIP FROM 40% FP l FIGURE 9.1.2 9-11

.. _j ..____.t._.__._. L_. {. _. _ [ i . l... _ _, f. _ _ + _ _ _.__ i 200 ...._.i. i i -r ..p -. ._ _ _ _ g. __7.__.

---

y..

175 ._ - ( - -t ___.p._. ,i i 0 i i s I f i i i E 150 = j u i i 25 i e i t l I i V"" ~ + 1 i M 1 kJ> I i DJ 6 6 e i 125 ~ a: l -~ i i W I N I M \\ x l b l \\ 2v3 f 8 un .i W M i 100 ca i i W l i g i i + ..( l j v3 d i __. s. 8 r 75 / i i - f, .___ H i i -i 1 i i t 50 - - * - - '~~ - i i + I I 25 m. I 1 2 3 4 f.* ELAPSED THE FROM TRIP (MDIUTES) s DENPS - Unit #1 STARTUP REPORT COMPENSATED PRES-SURT.ZER LEVEL VS. ELAPSED TIME i FROM TRIP 9-l*' FROM 40% FP FIGURE 9.1.3 1

75 h n .__. _ _ j _ g G i z l t i 50 s s, W t 9 U l f Qe

s H

i i M I - W-W -, w_ i 25 I m i i i l e [ r t I e --~ ~ ~ ~ p~ 4 i O l c. 1 2 3 4 ws ELAPSED TIME FROM TRIP (MINUTES) DBNPS - Uni #1 STARTUP REPORT SG #1 STARTUP LEVEL VS. ELAP5ED TIME FROM TRIP FROM 40% FP FIGURE 9.1.4 9-13 ~

75 ..____.i __.. .m __ __.._.4.____.. . _._. 9._.____ .. _- T. _._ _. _. _. __ _ f. _ __.. _____ y _._... g ._..4__.. ..p U ....__t'._______ g _. _ z 5 0 >t{.._ _ T., __..... .._.. _-.+ - -. O a

-a i

g i.. _ _. I ._._.._p._ a _. _ _._p _3,. _.___.g._______ .____}......_.. n ... _._ q

j...

Hg .._.p_.._. 3-Y-2 1 i g _.--t-- m 25 ._.p.._..._..j_.._.___. v_ _ _. _w _. 7.._.. _. _ _ _ -.p. __ __ __. e m 0 __= ' ,S 1 2 3 4 us ELAPSED TIME FROM TRIP (MINUTES) DENPS - Unit #1 STARTUP REPORT SG #2 STARTUP LEVEL VS. ELAPSED TIME FROM TRIP FROM 40% FP FIGURE 9.1.5 9-14 c

2200. p I.. _.1, "~~ _g 2100 91 d l i . /./ i i i i l 2000 i f I i i o I I / l g m 1900 E- \\ E a. \\g g i en 1 U F e I 1800 l 4 i t 3 l l t I t 1700 f f g i i i 4 l l i 1600. H __.4 S 1 2 3 4 uH EL4PSED TIME FROM TRIP (MINUTES) DBNPS - Unit #1 STARTUP REPORT RCS PRESSURE VS. ELAPSED TIME FROM TRIP FROM 40% FP FIGURE 9.1.6 9-15 l .a. +

2300 [.. _.. g _. f'__.. _.. _..._i__._._____; ._. _l _. J._ _ _ _._-.p.,__. p ~ p. _ _q. __.... . _. _t _{ + _ __.. 3 ._..a.-_-__ ..____L....__.__.. 4 ~ ~~ ' ' ~ ~"--b.. _ _. ...__4.__ 2200 .__..t...___.[.. ~~ ~~~ ~ ~ -T .__i +. _ ...._.[._... .......[... ._.._._a_._-_ ___4.___ ...-__.y p _ _ _ _ _ .. _ _... w..... a __ _. _E .__u. ._.___.__L.__ p . ____...i____._.__.4._ 2100 .._.4 -+ q _g l i e i I I e i n o g E 2000 ~ ~ I 1 I i as i i 4 as m vs y) f t b h.1 e i M i i b t v3 1900 i o m e + s 1 f I ( 8 1 i 1 -- i j 1 1800 i ~~ i 1 4 t 6 s i = 1700 4 -_ ' ; t t 1 ~ ~ - - = 1600 S 1 2 3 4 us ELAPSED TIME FROM TRIP (MINUTES) 1 DENPS - Unit #1 STARTUP REPORT l RCS PRESSURE VS. TIME FOR TURBINE TRIP FROM 75% FP FIGURE 9.1.7 9-16 4

125 h. .. i_.._ }._ _. 7-. 4_.... _. y. 9._ .L._.-. L- .Q . -.. b .M_ .._,.._ g ....._.y_ .__w..-.. -_.. r .._..'.4 ..__1..-.-. .... - + _.. _. 100 ~ _ _ g. - - = - j 4 3 i d... --- 1-my f X-_----- Am U 1 2; t 75 >I i l i I 6 8 g3 I I 1 i l .2 b i ~ j i m i .i_ i gg i i$ 50 l m i i y l f i i i t I e i i m i 25 i _ _.. j 4._ ___ m 1 2 3 4 uP ELAPSED TIME FROM TRIP (MIMTIES) DBNPS - Unit #1 STARTUP REPORT SG #2 STARTUP LEVEL VS. TIME FOR TURBINE TRIP FROM 75% FP FIGURE 9.1.8 9-17 L~

125 .._______.._.4.._.____..__._ .___3._.__ } ._.-.___r-_-- i t ._.1__... a...__..

w. __

._t_. _.4...__ _ j __ 100 j.__..... _ ..____t. . _.1 i i i i i m = 7 __ h r i y .p.. i = r i Z' i v 75 I M a f W i y s a e u y _.. x _ t _. 50 h ._ t. i m -i = l e m i 25 L_ _ 1 i 0 I r 1 2 3 4

• u H

ELAPSED TIME FROM TRIP (MINUTES) i DBNPS - Unit #1 STARTUP REPORT SG #1 STARTUP LEVEL VS. TIME FOR TURBINE TRIP FROM 752 FP FIGURE 9.1.9 9-18

250 ._.7_._... 7 _ _ _. .__.7.__. .......r._...._ . _ _ ~ .._..t.____. .L ___. _.. -I t 3 ....__H '~~~ ~~ ' ~ ~ ~ ~ ~ ~ ~ ~ + " ~ ~ ~ ~ ~ - ~ ' i ~' G 200 %. w l __ a _ ___. _l. _..._ E i ? i i e p.e I } I 6 gj t 3 I w i i w 15 0

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u h3 i ~ i ,i ,i _._. o4 D g p. t w-100 ' j y i g I - d l l I n. i r r i i A O 50 f I j 1 0 ~ ~~~I I S 1 2 3 4 us ELAPSED TIME FROM TRIP (MINUTES) DBNPS - Unit #1 STARTUP REPORT COMPENSATED PRES-SURIZER LEVEL VS. TIME FOR TURBINE TRIP FROM 75% FP i FIGURE 9.1.10 9-19

100% UNIT LOAD REJECTION (1-14-79) / 2325 ~T 1.: i.. t n

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- =gn.. 1 + u. ,i = (._ p ..___3.. ng __._.__a__ .m ,u__.-.p....._.. ._t_ + + r -t-H _~. ..AyygAcp_m i __ E t 4., n - .._,___w. y y_..- __.__t_.._ p -- .3 '. g ,__.a.~w....___ n. 3. u. F _.,_._z...... Y, s, _. 9. i a.._._..._. .s_ Au .... T. -- n... p.__... ). ._:1 _._t ..____.i_.__. 7 ..___.4.... _.. .5.__.& % r q. o _ _4 _ ,i_.- 1 ( ___6.. ~ -"'--p~^ . _. _ _. _ _ - - ~ _ _ 0~ - - * - - ' - - ~ '.. w u' - . _ _ _ _._ v._. _.. _.._. 3- ............._.. _ _....J _.. ...1._......__.j_.._.___ f..._..._.._ ...._._.. _,t._. __.. _.._...,.. _..__._+ a_..._..._ . s r__.. _._ px...,._..-.........s.._..,.._, a ...,.,..._...<.._.._..,s. ._._.__... ~. .m a . _. i i .3.. team. Generator _ Evel. _ I..EheU-n DBNPS - Unit #1 STARTUP REPORT Natural CircuAation Test Data 9-22 Figure 9.5.3

10.0 SECONDARY Pl.WT This section will provide a brief summary of the major difficulties encountere4 with the secondary systems during power escalation. The secondary systems that will be covered include: 10.1 Turbine-Generator 10.2 Condenser 10.3 Circulating Water System 10.4 Feedwater System 10.1 TURBINE-GENERATOR During startup, the turtiine and generator experienced relatively few major difficulties, but was plagued with numerous diversified prob-less. High vibration during the second turbine roll led to a re-alignment of the exciter in August. The first time the generator was loaded, grounds in the generator exciter bearing and Number 8 Generator Bearing were discovered. The bearings were subsequently disassembled and the grounding problem resolved. The turbine over-speed trip mechanism did not operate when first tested in September with the trip point approximately 12 RPM above the limit. The mech-anism was cleaned, adjusted and inspected. The Steam Generator 1-1 Turbine Bypass Valves were cycling open, then closed just prior to shutdown which caused damage to the turbine by-pass headers in tta High Pressure Condenser. The strap piping res-traints were replaced with rigid restraints to prevent excessive piping movement. In November, the Number 2 Turbine Control Valve position was found to be oscillating. This caused two unit shutdowns in which a function generator on the turbine EHC System and the servo valve for the Num-bar 2 Control Valve were replaced. The cause was found to be a defec-tive electrical connector which was the position feedback to the EHC. In January,1978, the turbine control rotor of the overspeed trip mechanism was replaced in an attempt to solve the problem with reset-ting the oil trip. In February, 1978, it was discovered a defective oil trip solenoid valve was the cause of the overspeed trip difficul-ties and the solenoid valve was replaced. The inboard bearing on the turbine main oil pump was replaced and orifice plugs installed in the - bearing oil supply lines of the front standard. ~ ~ During the January 1979 outage, an oil leak on turbine bearing number 8 was repaired by machining of the bearing cover sealing surfaces. An j inspection of the generator high voltage bushings by a vendor representa-i tive revealed the need for replacement of a bushing. This is to be dene during the first_ refueling outage. On February 22, 1979, electrical far.ts in the turbine electro-hydraulic control (EHC) system caused a runbt. (see Section 12. " Unscheduled Unit Trips"). The electric faults vert c avestigated by General Electric per-sonnel an'd several printed circuit cards were replaced. Further investi- ~ [ 6 gation byfGeneral Electric discovered defectivo capacitors in the circuit l-cards. l 10-1 j' OHMO M e4 4 ,g m

The unit experienced difficulties with the turbine throttle pressure limiter circuit supplied by General Electric. The purpose of this circuit is to provide backup header pressure control in the event of a failure of the normal mode of control. However, on August 22, 1979, a unit runback was initiated by a partial failure of the throttle pressure. limiter power supply which overrode the normal ICS signal and partially closed the turbine control valves. A complete failure of the power supply occurred on Septunber 27, 1979 which resulted in a unit trip. 6 Toledo Edison is presently in the process of replacing all four of the originally supplied pressure transmitters and the power supplies with transaitte s eM power supplies from General Electric but manu-l factured by a different vendor. The electro-hydraulic control (EHC) system has also caused unit trips on October 3,'1978 and September 18, 1979. On these occasions, the starting of a second EHC pump caused an instantaneous perturbation in pressure resulting in turbine trips on low EHC pressure due to sticking pump pressure controllers. Several design modifications are planned to reduce the sensitivity of the system. 10.2 CONDENSER The condenser has encountered problems with condensar tube leakage. In September,1977, approximately 28 tubes were plugged near a pene-tration where a high pressure steam header warmup drain from the main steam header had impinged on the tubes due to improper design of the baffle arrangement inside the condenser. A new baffle was installed which deflected the high pressure steam away from the condenser. On February 14, 1978, reactor power was increased to 90%. High con-ductivity was noted the next day which indicated a leak of a conden-ser tube. Power was reduced and one tube was plugged. On February 16, 1978, the unit returned to 90% power. On February 17, 1978, a tube leak was again reported. The unit was taken off line and five ^i leaking tubes along with eleven adjacent tubes were plugged. The unit returned to 90% power on February 19, 1978. On February 20, a small condenser tube leak was discovered and the power was reduced to.75%. The unit was shutdown on February 24 and an addy current inspection of various tubes indicated serious problems existed in numerous condenser tubes. Sixty-seven tubes were plugged. During the unit outage to remove the BPRA's, flow' diffusers were added to the condenser internals in an attempt to correct the con-denser tube failures. While the unit was at 100% of full power on September 25,1978, in-creasing condensate conductivity indicated condenser tube leakage. When the unit was shutdosu (due to a failed Reactor Coolant System Flowmeter), a condenser inspection revealed one tube had developed a leak. The defective tube was plugged. 10-2 ~. -..~___._~..~e.

1 l With the unit at 100% of full power o'n September 6, 1978, an increase in condensate conductivity was again discovered. When th6 unit was shutdown (due to defective Reactor Coolant Pump Seals), invs tigation revealed two leaking tubes which were then plugged. The condenser vendor, Ecolaire Condensers Incorporated, investigated the cause of these tube failures and installed some extra supports within the condenser while the unit was shutdc,wn for repairs of Reactor Coolant Pump seals during October,1978. In December 1978, abnormally low pressure was noted in the Low Pressure Feedvater Heaters.. The unit was shutdown, and the condensers were opened and inspected. Three expansa.on joints had failed in the conden-sers and. had caused some condenser internal damage. The High Pressure Condenser' extraction steam fairing door had been blown off and had dented eight tubes which were than plugged. Re-welding was required on some impingement baffles and a seam of one of the extraction steam fairing Instrument tubing in the vicinity of the expansion joints was replaced or repaired as required. The addition of extra tube supports was also completed during the outage. It is believed the failure of the expansion joints was due to botp under-designed expansion joints and excessive vibration. The original expan-sion joints were not designed to withstand' the superheated steam present in the extraction steam lines. Additipnal bracing to stop the excessive vibration was installed per Ecolaire recommendations and expansion joints designed to withstand the superheated steam were installed in an attempt to prevent a recurrence of the failure of the expansion joints. During a planned unit maintenance outage in January,1979, two 18" extraction steam line expansion joints were replaced as per Ecolaire's recommendation. One of these expansion joints was replaced in December 1978. Ecolaire's evaluation of the previous failures indicated a design problan with the joints, and as a result, the two new joints were installed. During the January 1979 outage, a total of 49 tubes in both the high and low pressure condensers were plugged. It is believed that most of the failures were caused by steam impingement; a sample of a broken tube is to be analyzed to determine the exact mode of failure. Only one additional tube was plugged in April,1979. Over three months 6 of almost entirely full power operation has not caused additional failures. 10.3 CIRCULATING WATER SYSTEM The Circulating Water System has experienced problems with failures of the liners of the 54 inch discharge valves. In August, 1977, the Number 2 Circulating Water Pump Discharge Valve was rebuilt and the rotation reversed to reduce the amount of turbulence. 10-1. ..a

During the plant cutage in September, fragments of a valve liner were fcund in the condensor water box. Subsequent investigations showed that the Number 3 Circulating Water Pump Discharge Valve was damaged. The valve liner was replaced and the rctation reversed on Number 3 Circulating Water Pump Discharge Valve. The amount of time these butterfly valves are throttled is now being limited to minimize the damage from the turbulence. t 6 All four of the discharge valves are to be replaced at the refueling outage. The Cooling Tower experienced some icing difficulties. In December, 1977, several internal fill support concrete beams wars broken and others damaged by ice buildup. Ice falling inside the veil also damaged some of the drif t eHminators and conduit was damaged from ice buildup. A revised operating procedure was provided by the vendor, Research-Cottrell to minimize the icing damage. Operating under the revised procedure has reduced the ice buildup and the damaged beams were rephced during the outage to remove the BPRA's. During clie.1978-79 winter, approximately 14 horizontal and diagonal fill columns and several deicing pipe end caps were damaged in varying amounts due to normal,ica. buildup. The fill columns do not perform any support function and were removed if heavily damaged. The deicing pipe end caps will be rep 1' aced-Although the d u ge to the cooling tower was not severs, the vendor, Research-Cottrell is investigating the possibility of modifying the deicing system to further eliminate icing difficulties. 10.4 FEEDWATER SYSTEMS Another area where major problems have been encountered is with the feedpumps. Main Feed Pump 1-2 was taken out of service in October, 6 1977 because of high vibration and flow race difficulties. The pump was disassembled and a piece of the pump impeller was found to be broken off. The entire impeller was replaced with the spare and the sharp radius corners between the imeplier vanes and sideplates were ground out to remove a potential high stress area on the impeller. The impeller of Main Feed Pump 1-1 was also inspected and ground. Both feedpumps were returned to service and further impeller diffi-culties have not occurred. In Jaisuan '1978 'it was determined the drain system of the Main Feed Pump Turbines was not operating. correctly and high turbine exhaust casing water levels were causing the turbines to trip. After an ex-tensive investigation of the drain difficulties during power operation, a modification which removed the lower source tap loop seal was com-plated in February. ~ 4

During January 1979, both Booster Feedwater Pumps were disassembled to replace the casing seal gaskets with a new type gasket to reduce casing leaks. At this time, ner seals were installed on Main Feedwater Pump 1-1 to reduce excessive seal leakage. The auxiliary feed pumps have had extensive difficulties in speed con-trol. In July and August 1977, repeated speed control relay failures rendered the auxiliary feed pimps inoperable. On August 10, 1977, a design modification was implemented which added a second set of identiaal speed relays in parallel to reduce the current carried by ear.n relay. This did not totally eliminate the speed control failures and in January 1978, the relays of -the speed circuit were replaced with relays of a larger current carrying capacity. Other design defieiencies were discovered in October when it was observed the Auxiliary Fr.ed Pump 1-2 Turbine Governor Valve wmid close under cer-tain vibrational conditions, rendering the Auxiliary Feed Pump 1-2 inopera-bla. A redesigned valve linkage was installed in which the force of a spring assured elimination of the vibrational closure. Feedwater chemistry control has encountered several problems during power escalation. The moisture separator reheater drain tanks concentrate silica and sodium. These tanks are located downstream of the condensate domineralizers and in 1977, it became necessary to return Number 5 Feedwater Heater Drain to the condenser to limit the silica and sodium concentration in the feedvater. This reduced the efficiency of the unit, therefore, a solenoid air control valve was added to allow Number 1 Mois-ture Separator Reheater Drain Tank to drain directly to the condenser. In November 1978, it was discovered that 19 tubes of Feedvater Heater 1-4-1 had failed. The failure was located within the drain cooler, approximately 8 feet from the tube sheet. The cause of the failure was attributed to high velocity flow which resulted in excess vibration of the tubes. The 19 failed tubes and 27 surrounding tubes were plugged. Investigation of the tube failures resulted in the vendor recommending the modification of the feedwater heater drain cooler inlet and the in-sta11ation of suci-vibration strips between the c,ubes. These modifications were made to Feedwater Heaters 1-4 and 2-4 during February and March 1979., Inspection of the tubes at that time revealed that water level contre 1 difficulty was not the source of the problem, but that the drain cooler inlet was improperly designed causing excessive . water velocities and tube damage. i 6 9 ~

• 10-5

.. c-a, -.~ . ~,.. - - - - -

O 11.0 U!!IT MO::ITCRI!!C - CHIMISTRY A';D HF_U.TH PHYSICS Unit monitoring tests were conducted to verify that the activity levels of the reactor coolant, feedwater and process fluids are within acceptable limits, that the dose rates resulting from direct radiation from sources contained within the station and on the site are within acceptable limits, that prcper chenistry control has'been achieved for the reactor coolant system t.nd stec:$ generators, and that the area and,cocess radiation coni-tors are capable of continuously detecting and recording asso-ciated radiation levels. Initial shield surveys were conducted on May 17, 1977. Shield surveys were conducted from August IS to 21, 1977, at 0% power, on September 3, 1977, at 15% of full power, from tiovember 17 to 18,1977, at 40% of full power and from Septe=ber 18 to 21,1978 2 at 100% of full power. Initial site and station radiation surveys were conducted on May 17, 1977. Site and station radiation surveys were conducted from August 19 to 21, 1977, at 0% power, on Septe=ber 3, 1977, at 15% of full oever, and from ::ovember 17 to 18,1977. at 40% 2 of full power, and from Sepec=ber 18 to 21, 1978 at L % of. full power. Proper water chemistry for the initial filling of the reactor TP 500.01 coolant system was verified on August 19, 1976, and was again verified on August 23, 1976, following the initial filling of the RCS. Proper water chemistry for the hydrotest of the reactor coolant system was verified on September 7, 1976, and was again 0 verffied on September 7, 1976, prior to exceeding 250 F

• t the reactor coolant system. During the power escalation prop am, reactor coolant chemistry was verified to be within specified limits,. on a daily basis; results included in this report are

.from s cples takea on August 24, 1977, at 0% power, on October 30, 1977, at 15% of full power, on Dece=ber 8,19,77, at 40% of full 2 power, on February 23, 1978, at 75% of full power, and on September 21, 1978 at 100% of full power. Proper water chemistry for the initial filling of the steam genera-TP 500.02 tors was verified on June 5,1976; following the filling of the steam generators, the steam generator layup water specifications were verified to be within li=its on July 26, 1976. Feedwater . cleanup was conducted, and proper feedvater chemistry was verified on January 9, 1977. During power ascension, prior to achieving 15% of full power, proper steam generator water chemistry was veri-fled on October 28, 1977, at a power level of 11% of full power. 'After achieving 15% of full power, proper feedwater chemistry was verified 'on a daily basis; this reporti includes the' results of samplas taken on Dece=ber 8, 1977, at 40% of full power, on Febru-ary 23,1978, at 75% of full power, and on Septetber 21, 1978, at } 2 100% of full power. J f t 11-1 1 f 6 f

~ TP 500.03 The following initial radiochemistry tests were performed to establish baseline activity levels for future operations: The reactor coolant, BWST water, 'eca= generators and 1. component cooling water were analyzed four ti=es during hot functional testing, from November 13, 1976 to Decem-ber 29, 1976. The spent fuel pool, fuel transfer canal and decay heat 2. system were analyzed just prior to the introduction of fuel daily during fuel movement and weekly thereafter. The results of the initial analyses conducted on Y. arch 4, 2 1977, and the results of the analyses obtained, follovira the completion of fuel loading, on April 27, 1977, were 2l normal background levels of radioactivity. The reactor coolant, BUST water, steam generators and ~ 3. component cooling water were analyzed in=ediately follow-the con-ing the closure of the reactor coolant syste=, at clusion of fuel loading, on April 27, 1977, and at speci-fied periods thereafter during power escalation. The typi-cal results of the reactor coolant analyses with the unit 2 at 100% of full power is shown on Table 11.5-1. TP 800.01 Process and area radiation monitor system tests were co=pleted on July 15, 1977. The tests included celibratien of the area radiation monitors and ion cha=ber area radiation monitors, and calibration of the gaseous process radiation =onitors and liquid process radiation monitors. Additionally, proper func-tioning of applicable alarms and interlocks was verifi,ed. 11.1 SHIELD SURVEY Shield surveys were conducted to designate locations for subse-quent surveys, to measure radiation levels at designated loca-tions adjacent to the shield building and secondary shielding and to obtain gamma and neutron background radiation levels for comparison with future measurements of radiatica due to activity buildup. All surveys have been completed up to and including the 100% power level. The shield surveys conducted indicate some radiation levels outside and inside contain=ent are in excess of the radiation zones state'd in the Final Safety Analysis Report (FSAR). This resulted in a 10 CFR 50.59 safety review and the limiting of access to se=e specific areas during power 2 operation. The existing dose rates do not result in any =easura-ble exposure to the public or any excessive. expcsure to plant per-sonnel as long as access to these areas while at power is limited. The areas exceeding the estinated radiation levels are inside the auxiliary and contain=ent building where access is controlled. The source of the radiation is neutron strea=ing enanating frca the reactor during power operation. The neutren streaming problem is still under analysis. 11-2 e

TP 800.03 11.2 SITE AND STATION RADIATION SURVEY The Site and Station Radiation Survey's are complete up to and 2 including the 100% of full power level, except for se=e data collection and review. Site, and station radiation surveys are conducted to: 1. Establish normal background radiation levels at the site boundary. 2. Verify that the dose rates resulting from direct radiation from sources contained within the station and on the site are at or below normal background levels. 3. Obta'in gamma and neutron radiation levels within the station at various power levels for comparison with future ceasure-ments of radiation due to activity buildup. 4. Establish station radiation zones for controlled entry by station personnel during various phases of reactor operation. S. heterminethealertandalarmaetpointsforthearearadia-tion monitors. 2h Baseline data has been established from the various survey measure-ments and alert and alarm values are determined for the area radia-tion monitors. Radiation levels in the following areas exceed the levels as listed in Chapter 12 of the FSAR (at 100% power): 1. 603' Elevation outside the equipment hatch 2 2. Main streamline access rooms

3. -Outside Containment Emergency Hatch but within the exterior entrance 4.

603' Elevation inside Containment (east side) All measurements except as noted above have been less than the 1evels established in Chapter 12 of the FSAR for the appropriate Radiation Zones. 11.3 REACTOR COOLANT SYSTEM CHEMISTRY TP 500.01 The Reactor Coolant System chemistry test procedure was implemented to ensure proper water chemistry was maintained during: 11-3

.e ~ 1. Initial RCS filling 2. RCS Makeup after initial fill a 3. Hydro test 4. Operation at temperatures above 2500F and Hot Functional Testing, and 5. Power ascension The water chemistry met the specifications listed in Tables 11.1 through 11.5 11.4 STEAM GENERATOP CHCIISTRY TEST TP 500.02 The steam gen trator chemistry test was i=ple=ented to establish minimus sa=pting frequency'and proper water chesistry for the condensata and feedvater syste=s and the steas generator during: 7.2 Feedwater Cleanup, 7.3 Hot Functional Testing, and 7.4 Power Ascension NOTE: Section 7.1 of TP 500.02.1 was deleted because steam generator che=istry, during initial fill was con-trolled and recorded under TP 200.09, STEAM GENERATOR SECONDARY SIDE HYDRO TEST PROCEDURE. The water chemistry met the specifications listed in Tables 11.1 through 11.4. 11.5 INITIAL RADI0 CHEMISTRY TEST TP 500.03 The initial radiochenistry test was rsn to: 1. Establish baseline activity levels during Hot Functional Testing.- 2. To monitor,the activity buildup in _various plant systems ~during initial fuel loading, reactor startup and power opera-tion so that rapid deter =ination of failed fuel and primary to secondary leakage is possible. .g 3. To monitor the radionuclide leakage from.the fuel pin to ~ .the reactor coolant, or from the reactor coolant to steam generator water, or from reactor coolant to component cool-ing water. 4.' To familiarize laboratory technicians with sampling tech- .niques, safety require =ents, radiochemistry procedures and the operation of all radiochemistry counting equipment.

The procedures.for collection and analysis of the samples were-

'verifieH, ~che background or MDA (Minimus Detectable Activities) kere, established, and radioactivity was monitored in the RCS. ~. 11-4 1_.

TP 360.01 11.6 PROCESS AREA RADIATION MONITORING SYSTEM TEST This test was performed to de=onstrate the ability of the Process and Area Radiation Monitoring System to continuously detect and record the radiation in the station effluents and protect station personnel from exposure to excessive radiation levels. The area monitors were calibrated per IC 2005.02 and IC 2005.01 (calibration procedures) and signed off on separated acceptance sheets for each monitor. The process instruments were calibrated and signed off on their respective acceptance sheets. NOTE: The alarms and interlocks and systgm performance require-ments of chapters 11 and,12 of the FSAR were proven by completion of the following TP's: TP 160.02, "CTMT and Penetration Room Purge" - RE 5052 A,B,C TP 170.05, "CTRM Heating, Ventilaring and Air Conditioning" - RE 2024 A,B,C and RE 2025 A,B.C. TP 170.01, " Aux Bldg Radiation Ventilation System" - RE 5403 A,B,C, RE 5405 A,B,C, F.E 8446, and RE 8447 TP 190.02, " Contact Data Logger Input Verification - Radiation Inputs" - TP 230.01, " Clean Liquid Radwaste System" - RE 1770 A,B TP 231.01, " Misc Liquid Radwaste System" - RE 1878 A,3 TP 232.01, " Gaseous Radwaste Pre-Op Test" - RE 1822 A,B TP 240.01, " Component Coolin's Water System" - RE 1412, RE 1413 4 d 11-5 s

TABLE 11.3-1 INITIAL FILL MATER QUALITY FOR RCS Analysis Specification Typical Sample Suspended Solids 0.1 ppm max. 0.04 ppm Chlorides as Cl 0.1 ppm max. 0.00 ppm Fluorides 0.1 ppm max. < 0.01 ppo Conductivity 1.0 umho/cm max

• 0.'92 umho/cm pH at 770F 6.5 - 7.5
• 6.2
• NOTE: Due to CO2 absorption specification pH may be lowered to 5.8 and conducievity increased to 2.5 ucho/cm.

TABLE 11.3-2 RCS WATER AFTER INITIAL FILL (AI!BIENT TEMPERATURE)*** Analysis Specification Typical Sample Suspended Solids 1.0 ppm max. 4 0.'02 ppm Chlorides as Cl-1.0 ppm max. 0.03 ppm Fluorides as F-1.0 ppm max.

4. 0.01 ppm Hydrazine 0.1 - 1.0 ppm 0.33 ppm-Total Dissolved Gas 100 sta ec/kg H O max

• 13 sta ec/kg H O 2

2 pH at.770F 6.0 - 8.0 ** 7.8 '

• NOTE: Required when a gas overpressure is maintained on the RCS.
  • NOTE: May be higher due to hydrazine.

May be lower due to CO2 absorption.

• • • NOTE: See Table 11.3-4 for higher temperature.

I e 11-6

TABLE 11.3-3 RCS WATER QUALITY FOR HYDROTEST. Analysis Specification Typical Chlorides as Cl-0.1 ppm max. 0.07 ppm Fluorides as F-0'.1 ppm max. 40.01 ppm Hydrazine 0.1 - 1.0 ppm 0.1 ppm Total Dissolved Gas 100 std cc/kg H O max

• 30 sta cc/kg H2O 2

pH at 770F 6.0 - 8.0 ** 10.2 Lithium 0.2 - 2.0 ppm 1.1 ppm

• NOTE: Required when a gas over pressure is maintained on the RCS
  • NOTE : May be higher due to hydrazine and its decomposition products (ammonia). May be much higher due to Lithium addition.

4 TABLE 11,3-4 RCS*** WATER QUALITY ABOVE 250"F & HOT FUNCTIONALS Analysis Specification Typical Chlorides as Cl-~ 0.' ppm max. 0.00 ppm Fluorides as F-0.1 ppm max. < 0.01 ppm Hydrazine 0.1 - 1.0 ppm

• 0.1 ppm Total Dissolved Gas 100 std cc/kg H O max.

30 std cc/kg H O 2 2 .pH at 770F

9. 5 - 10. S**

10.1 Lithium as Li7 0.2 - 2.0 ppm 1.0 ppm 0.1 ppm max. 0.01-ppm. Dissolved Oxygen as 02

• NOTE: Hydrazine spec. need not be maintained above 400 F if

.02 is well below 0.1 -pm.

• • NOTE: pH range is w/o boric acid, w/ boric acid (100 ppm) maylbe 7.5 to 8.5 t
    • NOTE:

Pressurizer is considered a part of the RCS. If Press *:rizer i:emp A 2500F, specs apply to Pressurizer regardless of RCS temp. o 11-7

TABLE 11.3-5 RCS WATER OUALITY AT OPERATING CONDITIONS DURING POWER ASCENSION Analysis Specification Typical Chlorides as C1~ 0.1 ppm max. 0.03 ppm Fluorides as F-0.1 ppm max. <0.01 ppm Hydrazine 0.1 - 1.0 pps

• O ppm Total Dissolved Gas 100 sta ec/kg H O max 31.6 sta ec/kg H O 2

2

• 15 -140 std cc/kg H O 31.2 sta ec/kg H2O 2

Dissolved Hydrogen as H2 pH at 770F 4.8 - 8.5 5.7 Lithium as $17 0.2 - 2.0 pps 0.45 ppm Dissolved Oxygen as 02 0.1 ppm max. 40.01 ppm Boric Acid 100 - 13,000 ppm 1093 ppm Boron or 6245 ppm H3B03 0 .1 p'pm each and

• NOTE: Not required at 4 200 F if Cl and F- (

not required if 02 4 0.1 ppm regardless of Cl-and F-concentration. TABLE 11.3-6 RCS WATER OUALI'N AT OPEP.\\ TING , CONDITIONS WITH C:IT AT 1007, OF FULL PCWER Analysis Specification Typical Chlorides as Cl-0.1 ppm max. 0,03 ppm 0.1 ppm max. <0.01 ppm Fluorides as F-Hydrazine 0.1 - 1.0 ppm

• O ppm 2

Total Dissolved Gas 100 std ec/kg H O max 45.0 sta ec/kg H O 2 2 Dissolved Hydrogen as H2 15 - 40 std cc/kg H O 26.8 std cc/kg H O 2 2 pH at 770F 4.8 - 8.5 6.2 _ Lithium as Li7 0.2 - 2.0 ppm 1.2 ppm . Dissolved Oxygen as 02 0.1 ppm max. 0.01 ppm Boric Acid 100 - 13,000 ppm 1086 ppm Boron or 6205 ppm H B03 3 0

• NOTE: Not required at < 200 F if C1~ and F~ (.1 ppm.each and not required if 02 < 0.1 ppm regardless of Cl-and F-concentration.

O i 11-8

DBNPS STARTUP REPORT ** ~ TABLE 11.4-1 FEEDWATER CYCLE CLEANUP TEST ACCEPTANCE CRITERIA TYPICAL Iron (Fe) 0.1 ppm max. 0.05 ppm - Cation Conductivity 1.0 pmho/cm. max. 0.5 pcho/cm. DBNPS STARTUP REPORT TABLE 11.4-2 ROT FUNCTIONAL FEEDWATER TESTING TEST ACCEPTANCE CRITERIA TYPICAL Cation Conductivity 0.5 umho/cm. max. 0.4 umho/cm. ~ Hydrazine 0.1 ppo min. 0.339 ppm Oxygen (02)

• • 7 ppb cax.

5 ppb Silica (sic 2) 20 ppb max. 2 ppm Iron (Fe) 0.1 ppm max. 0.01 ppb

• Copper (Cu) 2 ppb max.

O ppm

• Lead Not detectable Not detectable pil at 77F 9.3 - 9.5-9.4
• Copper and Lead analyses are weekly
  • May be increased to 100 ppb for a period not to exceed one week.

e s 11-9

DBNPS STARTUP REPOR'f TABLE 11.4-3 Operation at less than 15% Full Power Steam Generator i TEST ACCEPTANCE CRITERIA TYPICAL Chlorides 1.0 ppa max. 0.07 ppm Sodium 2.0 ppm max. 0.01 ppm Cation Conductivity 10.0 nmho/cm. max. 0.4 p=ho/cm. Silica (SiO ) 2.0 ppm max. 0.01 ppm 2 pH at 77F Set by feedwater pH 9.4 DENPS STARTUP REPORT TABLE 11.4-4 Operation greater than 15% Full Power ~ _Feedwater TEST ACCEPTANCE CRITERIA TYPICAL Cation Conductivity 0.5 pmho/cm. max. 0.2 pmho/cm. Silica (SiO2) 20 ppb max. 15 ppb pH at 77F 9.3 - 9.5 9.3 Hydrazine 20 - 100 ppb 31 ppb Iron 10 ppb 410 ppb 0xygen 7 ppb 5 ppb Copper 2 ppb 0 ppb Lead Not detectable Not detectable 11-10

' TABLE 11.4-5 i Operation at 100% of Full Power ,Feedvater 2 TEST ACCEPTANCE CRITERIA TYPICAL Cation Conductivity 0.5 nmho/cm. max. 0.2 pcho/co. Silica (SiO2) 20 ppb max. 5 ppb pH at 77F 9.3 - 9.5 9.4 Hydrazine 20 - 100 ppb 28 ppb Iron 10 ppb 10 ppb 0xygen 7 ppb 5 ppb Copper 2 ppb 0 ppb Lead Not detectable Not detectable p 6 e N ee,s A e 4-9 \\- O 11-11

TABLE 11.5-1 Typical RCS Radiochemistry At 100% of Full Power DISSOLVED uci/mi SUSPENDED uCi/c21 ANALYSIS 1.3E-1 Gross Beta (includes suspended) H-3 1.9E-1 Dose Equivalent I-131 . 1.1E-3 I-131 4.7E-4 I-133 1.7E-3 I-135 1.4E-3 CS-138 2.1E-3 KR-85M 1.0E-4 KR-87 2.1E-4 KR-88 9.0E-5 XE-133 1.0E-3 IE-135 6.8E-4 XE-135M 4.1E-4 2 AR-41 1.0E-3 F-18 6.3E-2 NA-24 5.8E-3 ~ 1.4E-4 CR-51 MN-54 2.1E-4 1.3E-5 2.1E-4 MN-56 2.8E-3 2.2E-5 FE-59 2.1E-4 3.1E Co-58 1.2E-5 CO-60 9.1E-5 NI-65 3.3E-6 ZR-95 1.3E-5 ZR-97 3.7E-6 NB-95 3.0E-5 NB-97 3.8E-6 TC-99M -W-187 4.2E-3 1.2E-4 4 e e 11-12

12.0 UNSCHEDULED UNIT TRIPS f 'f During the power escalation phase of the testing program, a number of unscheduled reactor / turbine trips occurred. Infor-4 mation from these trips has been used to improve plant perfor-mance by identifying tuning requiremente in the Integrated Control System. And by demonstrating system deficiencies for which corrective actions have been initiated. The following paragraphs briefly describe the unscneduled trips ^ 2 which occurred since initial criticality (Abgust 12, 1977). This summary is intended to present the ccnditions surrounding each event, and not to present a detailed evaluation of each trip. 9/2/77 During the initial escalation to 15% power, feedwater flow was erratic. The main feed water pump controller was placed in automatic prematurely. The Steam and Feedwater Rupture Control System (SFRCS) tripped on differential pressure betraen steam and feedwater, leading to a reactor trip on low RCS pressure. The ucessive blowdown of the main st am erfaty relief valves contributed to the. reactor trip on low pressure. All the relief valves were reset by use of a hydroset on September 16, 1977. l 9/24/77 With the turbine off line and the reactor at approxi-mately 8% power, a " half-trip" of the SFRCS caused the startup feedwater control valves to close. Reactor Coolant ",ystem (RCS) pressure increased and lifted the power relief valve on the pressurizar. After several cycles, this valve stuck open, blowing the rupture disc on the quench tank and causing a partial'depressurization of the RCS. The power relief block valve was closed, and the plant was shut down for repairs. 10/23/77 ~ An undetected half-trip of the SFRCS closed the startup feedwater control valve to steam generator 1-2. The steam generator water level decreased to 17 inches, giving a full trip of SFRCS and initiating auxiliary feedwater. The reactor tripped on low RCS pressure as a result of'the addition of 700F auxiliary feedwater to the steam generators, and due to lifting of the pres-j surizer power relief valve. 11/29/77 The unit was operating at 40% with the Reactor Protec-tion System (RPS) overpower trips set at 50%. A faulty p tch board was inserted into the startup test panel, producing a unit load demand signal equivalent to 50%. The plant responded to the increased demand, and the unit tripped on high flux when the reactor reached 50%. The automatic transfer of house loads from the auxiliary 12-1 x'

transformer to the startup transformers was defeated, resulting in a plant loss of AC power. Auxiliary feed-water initiated natural circulation flow through the reactor, and the diesel generators assumed the essen-tial loads until off-site power was restored. ( 12/16/77 In the unit startup following the reactor trip test (TP 800.14) from 40% power, the turbine-generator was on-line and the reactor was at approximately 11% when the startup feedwater control valves began to oscillate. These valve position swings resulted in overfeeding of steam generator 1-1. The reactor tripped on low RCS pressure. Additional tuning of the ICS was performed to minimize these valve oscillations during startups. 12/30/77 Following nine consecutive days of steady-state power operations at 72% power #1 main feed pump tripped on " indicated" high exhaust casing water level. An Inte-grated Control System (ICS) runback was initiated, but the pressurizer power relief valve lif ted resulting in a reactor trip on low RCS pressure. The response of the main feed pump speed controls was modified, using the data collected during this trip. 1/6/78 Two SFRCS trips occurred during startup operations. Both were caused by feedwater flow fluctuations which caused feedwater/ steam outlet pressure differential to exceed the limit. Following the second trip, Auxi-11ary Feedwater Pump 1-1 was declared inoperable because the speed control circuitry malfunctioned. A circuit modification was completed and tested to correct this problem. 1/21/78 To check out the main feedpump speed control changes made as a result of the 12/30/77 trip, a #1 feed pump trip test from 70% power was conducted. For approximately one minute the runback went smoothly. Then the running pump tripped on high exhaust easing level. The reactor and turbine were tripped manually, and the plant was controlled with auxiliary feedwater during the cooldown to 5320F. 1/31/78 An SFRCS trip at 67% power resulted in a high pressure RPS trip of the reactor. The SFRCS trip was caused by a spurious half trip in conjunction with an intentional half-trip of the system while performing the monthly surveillance test. The monthly surveillance test has been modified to reduce the likelihood of a recurrence of this problem. 12-2

2/24/78 A failed RCS flow' transmitter had placed RPS Channel 3 into a tripped condition. An erroneous RCS high tempera-ture signal to Channel 2 of the RPS tripped the unit off-line. Both problems were investigated and corrected prior to resuming power operations. - 79 3/1/78 The reactor was at 49% power. The level control valve to deaerator 1-2 failed closed. The main feed pump ran out of feedwater which initiated an SFRCS trip on feed-water / steam pressure differential. The loss of feed-water and closing of the main steam isolation valves increased RCS pressure which tripped the reactor on RPS high pressure. 3/29/78 An abrupt change in the setpoint of the Tave temperature controller by an operator placed the plant into a tran-sient condition. Return of the controller to its original setpoint produced a direction error in the Control Rod Drive (CRD) Control System, temporarily, transferring the CRD control station to " MANUAL", a condition in which the CRDs would not respond to ICS signal demands. The unstable plant condition coupled with the inability of the CRDs to respond to neutron error demands created a mismatch between reactor power and feedwater,,resulting in overfeeding the steam generators and tripping the ~ reactor on low RCS pressure. 4/2/78 A turbine trip test was performed at 75% power to evaluate piping modifications made on the extraction steam lines to the deaerator. The feedwater flow exceeded the feedwater demand during the runback, resulting in over-feeding the steam generators. This coupled with lifting of the pressurizer power relief valve caused a reactor trip on low RCS pressure. 4/5/78 While operating at 100% FP for the first ttse, B&W re-quested an immediate reduction in power and a change to 3 RC pump operation while a complete analysis of the LBPRA problem was conducted. The unit was reduced in power to 65% and RCP 1-1 was manually tripped. Feedwater 2 demands were noc properly ratioed and the feedwater valve d P error signal in the ICS affected the main feedwater e pump speed to such a degree that the feedwater system reached an uncontrollable oscillation, and the RPS tripped the reactor on low RCS pressure. Since that time, FCR 78-200 has been approved and i=plemented to de-tune the DP error signal during two MFP operations, and adjustments have been made to properly ratio feed-I water after an RCP trip. 12-3

l l 4/29/78 While the shutdown for the screen outage was in progress, the unit experienced a reactor trip from approximately 20% FP. This was the first shutdown attempted with only ] three reactor coolant pumps in operation. As #2 steam l generator approached " low-leve) lLnit", the operator used manual control of the main feed pump to maintain 45 psid i across the main feedwater control valves. This resulted j in overfeeding the steam generator, and although operator action was taken to stabilize the situation, a rapid cool-down took place, tripping the RPS on low RCS pressure, and initiating high pressure injection for approximate ;y 5 minutes. The Reactor Coolant System was. returned to 2155 psig/5300F and a normal controlled cooldown to Mode 5 was performed. 8/2/78 In preparation for 40% reactor physics testing, the six second rod insertion step for differential rod worth measurement was attempted. The rod movement resulted in a Reactor Coolant System upset. The positive temperature coefficient caused feedwater control of Tave to be un-stable. A divergent oscillation in feedwater lead to overfeeding of the steam generators, and resulted in an RPS low pressure trip. j 9/10/78 Conducted optional turbine trip test from 75% FP per TP 800.14. Excessive feedwater flow resulted in reactor 2 trip on low pressure. 9/28/78 While at 90% FP, the loop 2 RCS flow transmitter FT RC1A1 failed low. This low flow signal caused a trip of RPS Channel 1 and initiated an ICS runback at 20% per i minute. The runback stopped at 700 Tse and resulted in feedwater to the steam generator ratioed as if the erroneously indicated flow condition actually existed. The operator took manual control of loop 2 main feedwater control valve, attempting to maintain level in #2 steam generator. This action resulted in feedwater flow greater than that required for the existing reactor power level, and decreased RCS pressure to below the 1985 psig RPS trip setpoint. The plant was placed in Hot Standby (Mode 3) and the RCS flow transmitter was repaired, l L 10/3/78 While operating at 73% FP, the second EBC pump was started to investigate the recent reduction in EHC header pres-sure. A hydraulic perturbation was introduced, tripping i the turbine on low EHC pressure. The ICS initiated a . reactor power runback at 20%/u,% te. The increased steam l generator pressure and the F? "c; oss-1Laits" rapidly increased feedwater flow, ove; :oof,tng the RCS and caus-ing an RPS reactor trfe 2r Ic. RCC pressure 84 seconds after the turbine trip fb.palysis af this trip resulted in a recommended modif1 cation io the ICS cross-limits, reducing 1;he amount of feedwater 44d following any turbine trip. 12-4 .1-- e .._e

10/29/78 With reactor power at 4% of full power while lowering RCS temperature for the Natural Circulation Test, TP 800.04, the reactor operator was controlling Main Feed Pump Turbine (MFPT) 1-2 speed in manual. The MFPT 1-2 motor speed changer hung up at the high speed stop resulting in a high differential pressure across the control valve. As the operator kept trying to reduce the differential pressure, the speed changer was freed and ran MFPT l-2 back resulting in a low differential pressure and a Steam and Feedwater Rupture Control System (SFRCS) full trip at 10:43:06 hours. The resulting sequence of events resulted in a reactor trip from low reactor coolant pressure,(1985 psig) on Channels 1 and 3 of the Reactor Protection System (RPS). The cause of this trip is due to both the unusual plant conditions while performing the Natural Circulation Test and the M7PT speed changer difficulties. The speed changer was repaired. 4 11/13/78 The reactor was at 99% power. The main power supply fuse to Relay Cabinet RC-1718 blew. This resulted in the loss of the 125 VDC power to auxiliary relays of the RCP control circuits for Reactor Coolant Pumps 1-2 and 2-1. Reactor Coolant Pump 2-1 main breaker tripped eight seconds before Reactor Coolant Pump 1-2. This resulted in an RPS flux-delta flux-flow reactor trip. During the investigation, it was noted that the fuse had been improperly wired resulting in the loss of two Reactor Coolant Pumps for a single fault. This electrir.al deficiency has been corrected. 1/12/79 The reactor was at 100% of full power when an accidental short circuit caused the loss of the inverter supplying essential 120 VAC Bus Y2. The'resulting loss of Reactor Coolant System (RCS) flow, and neutron power indication to the Integrated Control System (ICS) initiated a reactor trip. The loss of power on the Y2 Bus also started a sequence of events which resulted in a loss of main feed-water flow and a full Steam and Feedwater Rupture Control 4 System (SFRCS) trip. The Auxiliary Feedwater System (AFWS) supplied feedwater to the steam generators. s All of the circuits supplied by the essential buses have been checked and all power distribution fuses are now quick acting types to minimize a short circuit from tripping the essential 120 VAC Buses. 12-5

2/13/79 The reactor was at 88% of full power when a loss of power to startup transformer 02 occurred due to Ohio Edison testing of Beaver Substation. This transformer was being fed from offsite and was supplying power to the 13.8 KV "B" Bus which in turn powers two of the Reactor Coolant Pumps (RCPs). The loss of power tripped the RCPs resulting in a reactor trip. The unit is temporarily operating with housepower loads supplied by the startup transformers in order to comply with FSAR commitments. This deficiency will be corrected at the first refueling outage. Ohio Edison has also been informed of the necessity of notifying Toledo Edison prior to the conduct of any relay testing of the sort which initiated this event. 2/22/79 With reactor power at 87% of full power, a malfunction in the turbine backup. speed control circuit of the Electro-Hydraulic Control (EHC) System caused inappropriate move-ments of the turbine control and combined intermediate valves. The resulting low main steam pressure initiated a full SFRCS trip as desig ad. The low feedwater flow caused Reactor Coolant System pressure to increase and the Control Room personnel manually tripped the reactor. The faulty turbine backup speed control circuit has been replaced. 9/18/79 The reactor was at approximately 99% of full power when an instantaneous perturbation in electro-hydraulic system (EHC) pressure caused a turbine trip and an Anticipatory Reactor Trip System (ARTS) trip of the reactor. The EHC pressure transient was due to a sticking pump pressure controller. General Electric has recommended several design changes to 6 reduce the sensitivity of the trip pressure switch. 9/26/79 The reactor was at approximately 100% of full power when a failed throttle pressure limiter power supply caused a clos-ing of the turbine control valves. The reactor tripped on high pressure followed by a turbine trip. Toledo Edison will replace all the identical pressure transmitters and their power supplies with an instrument supplied by Ceneral Electric but manufactured by a different vendor. 12-6 '=m>=e+em+=a. s,

=

ee+-. e,* m p4 m _h

13.0 CORE PERFORMANCE FOLLOWING BURNABLE POISON ROD ASSEMBLY (BPRA) AND ORIFICE ROD ASSEMBLY (ORA) REMOVAL With the removal of the BPRAs and ORAs from the core, two other core modifications had to be performed: (1) The two orifice rods which were holding down neutron sources were replaced by modified orifice rods (MORAs). (The MORAs have only four pins instead of the normal sixteen pins). A latch-ing mechanism was then placed over each MORA to hold the MORA to its fuel assembly. (2) Due to the removal of the BPRAs, eight fuel assemblies had to be shuffled to minimize power peaking in the core. The core map for modified core 1 cycle 1 is shown on Figure 13.0-1. 13.1 CORE PERFORMANCE DURING ZERO POWER TESTING Following the removal of the BPRAs and ORAs, a zero power test program ] was conducted to (1) confirm the nuclear design characteristics of the l core, (2) validate assumptions used in the safety analysis, aad (3) validate analytical models used for predicting plant responses. Measurements were made to determine the shutdown margin and the zero power (1) moderator coefficient, (2) "all-rods-out" RCS boron concen-tration, (3) control rod worths, and (4) differential boron worth. The ejected rod worth was also measured during zero power testing. All testing yielded satisfactory results which ensured initial opera-tion of the modified core was within the limits specified by the Davis-Besse Unit 1 Technical Specifications. The subsequent sub-sections of this section sunnarize the results of various tests per-formed at zero power in accordance with Post Refueling Physics Test-ing, ST 5010.03. 13.1.1 SENSIBLE HEAT CECK 2 An upper power limit of 2 x 10-8 amps was put on zero power testing to assure no nuclear heat (sensible heat) would be added to the RCS which would affect the testing. Prior to zero power testing, a sensible heat check was run to verify no sensible heat' produced below the upper power limit. The point of adding sensible heat was checked as described in Section 6.2. During this check, sensible heat was observed at approximately 1 x 10-7 amps (which is well above the power Jimit). At this power, the turbine bypass valves started to open and the makeup valve started to close. j 13.1.2 REACTIMETER RESPONSE CHECKOUT Prior to using the reactimeter for reactivity measurements, L -the response of the reactor to a change in reactivity was compared to the design response. The purpose of this check- ~ out was to verify the delayed neutron constants used by the .reactimeter gave an accurate representation of the core. 13-1 t._ ~

The checkout was accomplished as described in Section 6.3. A plot of the reactivity inserted versus the doubling time was obtained and compared to the design value. These plots are shown on Figure 13.1-1. Reactivity insertions of approximately +25, -25, +75, and -75 pcm were perforned for the checkout of the reactimeter. 4l-ments for the checkout were within the i 5% of design values Tests results are summarized in Table 13.1.2-1. All measure-as required by test accentance criteria. 13.1.3 ALL RODS OLTI BORON CONCENTRATION The all rod out, hot zero power (HZP), at 86 EFPD critical baron concentration was measured and compared to design. This comparison'was used as one of the criteria for establish-ing the validity of the core physics model. With Control Rod Assembly (CRA) Group 7 controlling at 78% withdrawn (wd) and all other rods fully withdrawn (except CRA Group 8 at 37.5% wd), a boron end point measurement was performed to determine the all rods out (ARO) boron concentration. The measured boron concentration was 1615 ppm. The measured excess reactivity worth of the inserted 1 pcm = 10-5 Group 7 rods was 122 pcm (percent millif,h of 10.06 pcm/ 4 k/k). Using the differential boron wort ppm from the B&W Physics Manual, an ARO critical boron con-centration of 1627 ppmB was obtained. This is within the acceptance criteria of 1661 + 100 ppm with CRA Group 8 at 37.5% wd. 13.1.4 TEMPERATURE COEFFICIEITI 0F REACTIVITY See Section 6.5 for a discussion of temperature coefficient and a description of the test method which is the same as used here. Measurement results for the three rod configurations at which temperature coefficients were measured is summarized in Table 13.1.4-1. All moderator temperature coefficients measured satisifed the Davis-Besse Unit 1 Technical Speci-fication limit of less positive than.9 x 10-4 4 k/k/0F (9 pcm/0F). Also, all temperature coefficients measured met the acceptance criteria of being within i 40 pcm/oF of the predicted values. 13.1.5 CO:TIROL ROD REACTIVITY MEASUREMENTS During zero power testing at 530 F, measurements were made to determine the CRA group reactivity worths for Groues 4. 5, 6 and 7. The reactivity worth of the control rod grouns were calculated utilizing the reactimeter. 13-2 _y ~~

The " boron swap" technique, which is described in Section 6.6, was used to determine differential and integral worths for the control rod groups. Measured worths are tabulated and compared with predicted values in Table 13.1.5-1. Integral rod worth curves for the regulating rods obtained from these measurements are shown in Figure 13.1.5-1. 13.1.6 EJECTED CONTROL ROD WORTH The ejected rod worth test was performed to measure the reactivity worth of the single predicted worse case ejected rod. With CRA Group 5 at the Technical Specification insertion limit of 55% ud, CRA 5 of Group 6 (6-5) was pre-dicted to have the highest worth. Adjustments were made for measurement uncertaiaties and compared with the acceptance criteria of.71 - 1.00% 4 k/k. (The upper limit is the numbers used in the safety analysis. The lower limit is 20% less than the predicted worth of.85% Ak/k). The measurement was initiated from the following rod config-uration: CRA Groups 1-4 at 100% wd 2 CRA Group 5 at 57% wd CRA Groups 6-7 at 0% ud CRA Group 8 at 37.5% vd CRA 6-5 was borated out of the core and then inser'ted 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. The measured ejected rod worth was then multiplied by 1.05 to account for measurement uncertain-ties. This resulted in an ejected rod worth of.595%4 k/k which corresponds to a certain amount of reactivity inserted. The measured integral rod worth curves, however, indicated a somewhat different shape from the predicted shape. This resulted in considerably less reactivity being inserted by the rods in the rod insertion limit than used in the predictionc. The effect of having less reacitity inserted is to reduce the measured ejected rod worth. For this reason, even through the worth we measured was more than 20% less than the predicted worth, it was considered acceptable. 13.1.7 SOLUBLE POISON WORTH MEASUREMENT Data for the differential boron worth measurement was taken at the end points of two rod worth measurements. CRA posi-tions were: 13-3 .e-

c 1st Measurement 2nd Measurement CRA Groups 1-6 @ 100% wd CRA Groups 1-3 @ 100% vd CRA Group 7 @ 78% wd CRA Group 4.@ 10% vd CRA Group 8 @ 37.5% vd CRA Group 5-7 @ 0% vd Boron Conc @ 1615 ppmB CRA Group 8 @ 37.5% ud Boron Conc @ 1156 ppmB From the rod worth measurements the total wo. th of CRAs from 10% wd on Group 4 to 78% wd on Group 7 was 5.10% 4 k/k. This reactivity change corresponds to a change in boron con-centration of 459 ppm (1615 - 1156).' Therefore, the differ-ential boron worth is: 5.10% 4 k/k 459 ppmB = 1.11 x 10-2% 4k/k/ppmB The predicted differential boron worth from the B&W Physics Test Manual is 1.006 x 10-2% 4 k/k/ppmB. The measured value is different from the predicted by -9.46%, so the acceptance criteria of + 10% was met. 13.1.8 SHUTDOWN MARGIN CALCULATION ~ Technical Specification 4.1.1.1.1.d requires: The shutdown margin should be determined to be A 1: 4k/k prior to initial operation above 5% Rated Thermal Power after each fuel loading by consideration of: (1) RCS boron concentration, (2) Control rod position, (3) RCS average temperature, (4) Fuel burnup based on gross thermal energy generation, (5) Xenon concentration, and (6) Samarium concentration, with the regulating rod groups at the maximum insertion limit of Technical Specification 3.1.3.6, (The insertion limit for the modified core is 55% wd on CRA Group 5). Shutdown margin is defined as the instantaneous amount of reactivity by which the reactor-is euberitical or would be subcritical from its present condition assuming no chssge in APSR position and all control rods in except for the maximum worth rod which is stuck out. l Considering the definition of shutdown margin and the required l surveillance testing, the following may be stated: i-l 13-4

The worth of Groups 1-4 at 100% wd and Group 5 at the inser-tion limit (55% wd) is 5.01% 4 k/k. (The worth of Group 5 was measured. The worth of Groups 1-4 was predicted by B&W and verified by measurements of Group 4).. If the stuck rod worth of 1.01% is subtracted from this, 4.0% d k/k remains. 2 .(The stuck rod is also calculated and verified by the rod measurements made). All other conditions affecting shutdown margin match those of the predicted or measured data. Therefore, the minimum shut-down margin is adequately satisfied. f I 13-5

4 13.2 CORE PERFORMANCE DURING POWER ISCALATION TESTING Following the completion.of the Post Refueling Physics Test, ST 5010.03, a program of power escalation was undertaken as per the Power Escalation Sequence Procedure, TP 0800.00. Core performance testing was conducted at three major power plateaus of 40%, 75%, and 100% full power. The following sections give more details relevent to the performance of the physics tests conducted during the escalation phase of testing. 13.2.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER, TP 0800.02 Core alternations described in Section 13.0 did not involve any alteration of the out-of-core nuclear in-strumentation. The Nuclear Instrumentation Calibration at Power Test, TP 0800.02, was conducted at various power levels as prescribed in the Power Escalation Sequence Procedure, TP 0800.00. The bases for the acceptance criteria are included in Section 7.1 13.2.2 REACTIVITY COEFFICIENIS AT POWER, TP 0800.05 The Doppler and moderator coefficients of reactivity were determined at the 90% power plateau by TP 0800.05, Reactivity Coefficients at Power. The method for deter-mining these coefficients is described in Section 7.2. The Doppler coefficient was calculated using the rela-tionship in equation 7.2-4 and Figure 7.2-1, " Average Fuel Temperature vs. Reactor Power". No acceptance critiera were applied to the value of the Doppler coef-ficient computed, but the power Doppler coefficient was limited to a maximum positive value of -3.7 x 10-5 3 k/k/% FP. The computed value of the power Doppler coefficient was determined to be -9.8 x 10-5 A k/k/% FP. The computed moderator temperature at the 90% power plateau was -2.05 x 10-5 6 k/k/0F. 13.2.3 ROD WORTH MEASURDfENTS, TP 0800.20 Differential rod worth measurements were performed at the 4li 40%, 75%, and 90% power plateaus using the six second

insert / withdraw method described in Section 7.3.

The change in reactivity was primarily due to rod motion, although the analysis did correct for reactivity effects [ of fuel temperature chang'es. The resulting differential rod worths are summarized in Table 13.2.3-1. s 13-6 s x m. ~

13.2.4 CORE POWER DISTRIBUTION TEST, TP 0800.11 Core power distribution data, described in Section 7.4, was collected at the 40%, 75%, and 100% power plateaus during steady state conditions. The results of the tests at'each power level are presented on Figures 13.2.4-1 through 13.2.4-6. 13.2.5 PSEUDO CONTROL ROD EJECTION TEST, TP 0800.28 The Pseudo Control Rod Ej ection Test', TP 0800.28, was per-formed 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 excead 0.65% A k/k. Design calculations determined the control rod in core position H-14 to be the mos.t reactive rod following BPRA and ORA removal. The ejected worth of control rod H-14 from its nominal full power position was measured during testing. Listed below is the sequence of major events that occurred during the test: 1. Group 5 control rods were positioned to their full out 2 limit of 100% withdrawn and Group 6 control rods were positioned about 90% withdrawn. 2. TP 0800.20, Rod Worth Measurement, was performed to determine the differential worth of Group 6 before the rod swap. 3. Control rod 7-3 (core location H-14) was swapped with Group 6 rods until H-14 was to its 100% withdrawn position. 4. TP 0800.20 was performed to determine the differential worth of Group 6 after the rod swap. 5. The ejected rod worth of H-14 was calculated using the average of the two differential rod worth measurements and the rod travel of Group 6. The measured ejected rod worth of H-14 was 0.18% d k/k which is below the 0.65% d k/k limit used in the safety analysis. 13.2.6 INCORE DEIECTOR TEST, TP 0800.24 The incore instrumentation system, as described in Section 7.7, was tested at the 40% and 75% power plateaus as dir-ected by the Power Escalation Sequence Procedure, TP 0800.00. l 13-7 4

Data for the incore self-powered neutron detectors was obtained from the computer. Ratios of computer corrected detector readings to average string readings were calcu-lated and plotted against axial level. Symmetric detectors were compared for consistency while non-symmetric detectors were checked for reasonableness. Af ter analyzing the results, the incore detector system was determined to be functioning satisfactorily. 13.2.7 POWER Lh1NCE DETECTOR CORRELATION TEST, TP 0800.18 The Power Imbalance Detector Correlation Test, TP 0800.18, was performed at the 40% power plateau in accordance with the Power Escalation Sequence Procedure, TP 0800.00. The objectives of the test were: 1. To determine the relationship between the indicated out-of-core offset distribution and to verify the re-lationship is within the assumptions of the safety analysis. 2. To deter =ine and set the proper gains for the power range nuclear instruments scaled difference amplifiers in order to obtain the desired relation in the previous objective. 3. To determine the relationship between the calculated offset from the Backup Incore Detector System (BIDS), and the calculated offset from the full incore detec-tor system. 4. Determine the core maximum linear her.t rate (MLER) and minimum departure from nucleate boiling ratio (DNBR) 2 at the various values of core offset as outlined in Sectica 7.8. The test, as outl~ned in Section 7.8, was performed and the values of MLER and.tinimum DNBR were obtained for the pre-scribed values of offset. The criteria in objectives 1 through 3 were not met due to thre fact that the out-of-core nuclear tastrunects scaled difference amplifiers were not J reset prior tc conducting the test. The Backup Incore Detector System critieria, mentioned in objective 3, also did not meet the acceptance criteria.. In lieu of rerunning the test in its entirety, the following modifications to the test were developed: 1. It was decided that the BIDS criteria not be run at 40% power and a test deficiency written to limit the use of the BIDS until their acceptance. The test will be per-formed again at 75% power following a 7 to 30 day operat-ing period at 100% power, at which time the BIDS will be tested to verify that the criteria in objective 3 will be met. 13-8 =

2. It was decided that a " mini" test be performed to check the criteria in objectives 1 and 2. An outline of the " mini" test is shown below: a. At the 40% power plateau, the axial power shaping rods were positioned to obtain approximate power in-balances of +5.6% FP and -9.6% FP. b. Conditions were allowed to stabilize at each power imbalance for a minimum of 15. minutes and then the following were obtained: - Core power distribution data including incore offset values. - Out-of-core offset values. The " mini" test was performed but the gain values calculated for the nuclear instruments' scaled difference amplifiars were found to be too low. Therefore, the gain values were recalculated and the " mini" test run for a cecond time. This time the slopes provicing the relationship Latween the out-of-core calculated offset and the full incore calculated offset were within the acceptance criteria and the " mini" test was signed off. The values of the computed slopes for each out-of-core detector were found m be between 1.28 and 1.32 at.40% F?. This satisifed the acceptance criterion that the slope be greater than 1.25 in each case. 'A further requirement that each point fall within the acceptable region of Enclosure 4 was also satisfied. The values of out-of-core and incore offset are summarized in Table 13.2.7-1. Af ter the seven to thirty day soak at 100% power, the test was performed at 75% power and the BIDS offset values were verified to be within the acceptance criteria. The values of the com-4 puted slopes for each out-of-core detector were found to be between 1.20 and 1.216 at 75% power. This satisfied the acceptance criterion that the slope be greater than 1.15 in each case. A further requirement that each point fall within the acceptabla region of Enclosure 4 was also satisfied. The values of out-of-core, BIDS, and incore offset are summarized in Table 13.2.7-1. The values for MLHR and DNBR for each offset were also calculated to verify that their respective Jimits were not exceeded. 13-9 i . ~..

I DAVIS-BESSE UNIT 1. MODIFIED CORE 1. CYCLE 1 June 28,1978 3lsl7 l8 l1 l t0 ' il 12 13 14 ' 15 2 3 4 i A A i i... ..m.,, ~ .m ..,,,,O 0" .o .c .m ., m m m t,,@ 00,,x.,, g),, ~ c 1, m .. m m .m .i V,, D .V,. w w m ., a m, .. m.... . ~ .m T.O. o O. x,

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• 8F"etry Monitor

/ g=MesitorTetel Core end Symmetry 3 - 12400 g 0235 sese 23 G)* / N C = 14000 g U235 ****** S C001 { lasere Deteeter String Ni.ber ,,W 6/1 Control Component ID and Type Seerees at F-4 & 5-12 C

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II ORA Betet teck dews LOOL Aenembly b ber sa 03AS-F-4 i teck deva LOO 2 control Component Croup / Red d g Prepared 37 Approved my [M%, hag / d Y # DBNES - Unit #1 STARTUP REPORT Core Map (Modified Core) 4 Figure 13.0-1 J 13-10

REACTIVITY vs. POSITIVE DOUBLING TIME l 100 V_ 'x M 50 y l 'm u 'x 2 S N 1 'x x D 'm a w m 20 Design Curve I f l 6 i, i,,, i i i.,, ,,, i 4 i,, i l '8 0 k1 a 6 l j$ ) l l g ,9 J g g t g g g g g g g 9 g g g g gg g g iet, I i I, t 6 i i i i i i ! i i s1 : t i i i i :, 4 sl l q i i 6j i 1 l 40 80 120 160 200 240 t l DOUBLING TIME (SECONDS) REACTIVITY vs. NEGATIVE DOUBLING TIME -100 Design Curve j ,A: ^-1" 1 . ~ m .6 n, w s' ) p 9 5 ../'-, N M ) U d -20 x , i,. , i i, i ii, . i,,

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,4 6, 1 -270 -230 -190 -150 -110 NEGATIVE DOUBLING TIME (SECONDS) DBNPS - Unit #1 .l STARTUP REPORT i Reactimeter Checkout l 1 0 - Measured Data Figure 13.1-1 13-11 w

REACTIMETER RESPONSE CHECKOUT REACTIVITY CHECKOUT A GE DESIRED CONTROLLING CRA GROUP REACTIVITY DEVIA-g REACTIVITY GROUP INITIAL l FINAL DESIGN ACTUAL TION TIME um Seconds pcm pcm X NUMBER % wd % wd 4 +25 7 77 81 205 +24.2 +23.5 +2.9 -25 7 77 72.5 -249 -24.7 -25.2 -2.0 +75 7 77 89 55.5 +72.0 +72.5 .7 i \\ -75 7 78 64.5 110 -67.2 -64.7 +3.7 l l l l i - DBNPS - Unit #1 l STARTUP REPORT l Reactimeter Checkout Table 13.1.2-1 13-12

MODERATOR TDiPERATURE COEFFICIENT MODERATOR TEMPERATURE TEMPERATURE ROD CONFIGURATION COEFFICIENT COEFFICIENT (pem/F) (pem/F) CRA Groups 1-5 0 100% wd CRA Group 6 @ 65.5% wd +5.54 +7.54 CRA Group 7 @ 0% wd CRA Group 8 @ 37.5% vd CRA Groups 1-3 0 100% wd CRA Group 4 0 30% wd -2.20 .20 CRA Groups 5-7 @ 0% wd CRA Group 8 0 37.5% vd 2 CRA Groups 1-4 @ 100% wd CRA Group 5 @ 54% wd +2.85 +4.85 CRA Groups 6-7 @ 0% ud CRA Group 8 @ 37.5% wd I i DBNPS - Unit #1 l STARTUP REPORT Temperature Coefficient t l Table 13.1.4-1 i 13-13 l-

~ s TABLE 13.1.5-1 i ~ ' / COMPARISON OF MEASURED AND PREDICTED CONTROL ROD GROUP REACTIVITY TORTHS AT 530 F Predicted Worth (1) From Physics Deviation CRA Position, % vd Measured Worth Test Manual From (CRA 8 @ 37.5% ud) % d k/k % O k/k Predicted CRA 4 -1.31 -1.19 +10.08% CRA 5 -1.53 -1.48 +3.38% (2) CRA 6 -1.48 -1.40 +6.43% (2) CRA 7 -1.01 -0.97 +4.12% (2) Total Rod Worth -5.34 -5.04 +5.95 (3) l (1) Deviation Measured - Predicted From-Predicted = Predicted (2) Acceptance Criteria =115% i l l (3) Acceptance Criteria + 10% l l I DBNPS - Unit #1 STARTUP REPORT-Table 13.1.5-1 13-14

1 1 i 1 /. i CRA GROUPS 5, 6, & 7 WORTH WITH CRA GROUP 8 AT 37.5% WD 0 ~ r~ - ' ~ ~ ~- f ,l ~ - - - ll I .p... _. .j: jj, [ p ".. . 4 _ .. _.i1 l l j* I 5 L t - 500-.-- -r ii. 's:. g J _ [

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TABLE 13.2.3-1 MEASURED DIFFERENTIAL ROD WORTHS AT POWER Differential Full Rod Group Position (% Withdrawn) Rod Worths Power 1-5 6 7 8 (pcm/% wd) '2 40 100 87.5 0 30 9.09 l l 40 100 69.7 0 30 9.55 75 100 87.4 0 30 9.17 1 75 100 81.75 0 30 11.4 ~ 90 100 91.2 0-30 10.23 90 100 90.3 0 30 10.22 90 100 94.0 0 30 7.66 90 100 92.3 0 30 8.98 90 100 87.1 0 30 11.32 1 13-16

FIGURE 13.2.4-1 CORE POWER DISTRIBUTION Core Conditions Measurdd Design Measured Desien Gps 1-4 at 100 % vd 100 % vd Power Level 39.5 % 40 5 at 100 % wd 100 % vd Baron Cone 1301 ppm NA ppm 6 at 92 .% wd 93.5 % Vd Core Burnup 89.6 EFPD 88 EFPD 7 at 0 % vd 0 . vd Axial 8 at 32 % wd 25.5 % wd Imbalance -0.02 gyp -0.00 yp l Radial Peaks 1.10 0.97 1.01 1.25 1.09 1.14 0.65 0.58 1.13 1.02 1.06 1.27 1.15 1.13 0.65 0.55 0.94 1.10 1.07 1.30 1.03 0.99 0.64 1.01 1.10 1.13 1.26 1.08 0.96 0.61 0.63 1.15 0.99 1.24 1.08 0.50 0.64 1.14 1.04 1.21 1.06 0.50 1.14 1.37 1.09 0.91 ~ 2 1.17 1.33 1.08 0.84 1.34 1.21 0.63 1.37 1.17 0.60 Quadrant Tilt O +0.67% -0.85% +0.53% -0.35% 9.18 X.XI - Measured Minimum DNBR = X.XX - Design Maximum LHR = 5,42 kv/ft 13-17

FIGURE 13.2.4-2 CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Design Gps 1-4 at 100 % ud 100 % wd Power Level 39.5 % 40 5 at 100 % vd 100 % ud Baron Cone 1301 pps NA ppm 6 at 92 % wd 93.5 % wd Core Burnue 89.6 EFFD 88 EFPD 7 at 0 % vd 0 % wd Axial 8 at 32 % ud 25.5 % wd Imbalance -0.02 %FP -0.00 %FP Total Peaks 1.37 1.15 1.18 1.44 1.30 1.32 0.75 0.69 1.37 1.23 1.25 1.49 1.37 1.33 0.77 0.65 1.11 1.30 1.27 1.54 1.23 1.18 0.74 1.20 1.29 1.33 1.52 1.28 1.14 0.73 0.75 1.39 1.31 1.48 1.29 0.58 0.76 1.38 1.39 1.47 1.26 0.60 2 1.37 1.63 1.27 1.04 1.40 1.61 1.29 1.00 1.69 1.44 0.74 1.64 1.39 0.71 89 Quadrant. Tilt 8 +0.67% -0.85% +0.53% -0.35% i 1 \\ 9.18 I.II - Measured Minimum DNBR = I.II - Design Maximum LHR = 5.42 kw/ft 13-18

FIGURE 13.2.4-3 CORE POIER DISTRIBUTION l 1 Core Conditions Measured Design Measured Design 100 % vd 100 % wd Power Level 75.1 ; 75.0 Gps 1-4 at 5 at 100 % ud 100 % vd Boron Conc 1195 NA ppm ppm 6 at

1. 4 % wd 9J.; % vd Core Burnuo 94 8 EFFD 94 EFPD 7 at u % vd

% ud Axial U 8 at 32 % ud 22.J % ud Imbalance -2. 64 %FP +0.06 %FP i Radial Peaks 1.14 1.01 1.04 1.26 1.09 1.14 0.65 0.59 1.14 1.03 1.07 1.26 1.14 1.12 0.66 0.57 0.97 1.12 1.08 1.29 1.04 0.99 0.65 1.02 1.10 1.12 1.24 1.08 0.97 0.63 0.65 1.15 0.99 1.23 1.08 0.51 0.65 1.13 1.04 1.20 1.06 0.53 2 1.14 1.35 1.08 0.90 1.16 1.31 1.08 0.85 1.31 1.19 0.63 1.34 1.16 0.62 Quadrant Tilt +0.44% -0.32% +0.23% -0.35% X.XX - Measured Minimum DNBR = 4.63 I.XX - Design Maximum LHR = 10.55 kw/ft 13-19

FIGURE 13.2.4-4 CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Design 100 % wd 100 % wd Power Level g 75.0 75.1 Ops 1-4 at 5 at 100

vd 100 % wd Boron cone 1195 ppm NA ppm 6 at 92.: vd 93.5 % wd Core Burnuo 94 8 EFPD 94 Eppp 7 at 0

% vd 0 % vd Axial 8 at 32 ". wd 22.3 % vd Imbalance -2.64%FP +0.06 %FP Total Peaks 1.38 1.17 1.19 1.48 1.36 1.37 0.76 0.68 1.37 1.23' 1.25 1.48 1.38 1.33 0.78 0.67 1.12 1.29 1.27 1.58 1.23 1.14 0.76 1.20 1.29 1.34 1.52 1.29 1.15 0.75 0.77 1.45 1.32 1.53 1.27 0.59 0.77 1.39 1.42 1.47 1.26 0.63, 2 1.38 1.69 1.29 1.07 1.41 1.60 1.30 1.01 1.73 1.48 0.74 1.61 1.38 0.74 '9 Quadrant Tilt 08 +0.44% -0.32% +0.23% -0.35% X.XX - Measured Minimum DNBR = 4.63 l X.XX - Design Maximum LBR = _10.55 kw/ft l 13-20 l i

FIGURE 13.2.4-5 CORE POWER DISTRIBUTION Core Conditions Measured Design Measured Desien Gps 1-4 at 100 % vd 100 % vd Power Level 00.8 inn 5 at 100 % vd 100, % vd Boron Conc 1100 ppm NA ppm 6 at 90.6 _% wd 93.5 % wd Core Burnup 109. 2EFFD 110 EFFD 7 at 0 % vd 0 % vd Axial 8 at 24 % vd 22.3 % wd Imbalance 0.0 %FP -2.99 %FP Radial Peaks 1.15 1.us 1.04 1.25 1.08 1.13 0.66 0.61 1.15 1.05 1.08 1.25 1.14 1.12 0.67 0.58 0.98 1.12 1.08 1.28 1.03 1.00 0.67 1.03 1.11 1.12 1.24 1.08 0.97 0.64 0.65 1.14 0.97 1.22 1.09 0.52 0.66 1.13 1.04 1.20 1.06 0.54 1.12 1.33 1.07 0.90 1.15 1.29 1.07 0.85 1.29 1.17 0.63 1.31 1.14 0.62 1 Quadrant. Tilt 1 l +0.60% -0.17% -0.58% +0.15% l 4 I.XI - Measured Minimum DNBR = 3.28 'l I.IC - Design Maximum LER = 13.13 kw/ft I 13-21 =%--

FIGURE 13.2.4-6 CORE POWER DISTRIBUTION Core Conditions Measured Desian Measured Design Gps 1-4 at 100 % vd 100 % vd Fower Level 99.8 g 100 ; 5 at 100 vd 100 % wd Baron Conc 1100 ppm NA ppm 6 at 90.6 _ % vd 93.5 % wd' Core Burnup 109.2EFFD 110 EFFD 7 at 0

• wd 0

% vd Axial 8 at 24 % ud 22.3 % vd Imbalance 0.0 %FF -2.99 %FP Total Pe.aks 1.44 1.23 1.21 1.48 1.35 1.30 0,78 0.72 1.39 1.25 1.27 1.50 1.36 1.33 0.78 0.69 1.18 1.33 1.28 1.53 1.23 1.18 0.77 1.22 1.31 1.33 1.48 1.27 1.16 0.76 0.77 1.33 1.29 1.44 1.29 0.60 0.76 1.34 1.36 1.43' 1.27 0.63 1.34 1,58 1.28 1.06 2 1.37 1.55 1.28 1.02 s-1.59 1.38 0.76 1.61 1.39 0.74 0.91 0.89 @adrant Tut +0.60% -0.17% -0.58% +0.15% X..U - M,easured Minimum DNBR =- 3.28 X.XX - Design Maximum LHR = 13.13 kw/fC 13-22

TABLE 13.2.7-1 f

SUMMARY

OF AXIAL OFPSET MEASUR. N.- S 1 ICO OCO - % Power \\ % Power % Power i NI-7 l Backup Incore Offset NI-5 l NI-6 NI-8 % Power 39.21 -1.04 -3.60 -1.84 -3.22 -2.29 Valve Not Obtained 39.20 +13.98 +15.93 +18.44 +16.27 +17.66 Value Not Obtained 39.40 +15.08 +16.59 +19.21 +17.00 +18.38 Value Not Obtained 39.25 +15.47 +16.47 +19.06 +16.94 +18.34 Value Noe Obtained 39.42 -18.45 -26.59 -25.05 -26.04 -24.79 Value Not Obtained 39.30 -19.23 -27.61 -26.06 -27.01 -25.78 value Not Obtained 39.42 -19.51 -27.49 -25.98 -26.90 -25.76 Value Not Obtained 75.14 -0.634 +0.276 -1.819 -1.214 -2.521 +1.2 75.33 +10.922 +11.903 +10.253 +10.316 +9.742 +11.6 74.52 +5.604 +6.812 +4.971 +5.304 +4.355 +6.4 74.60 -7.385 -8.068 -9.790 -9.330 -10.490 -3.6 74.40 -14.524 -17.951 -19.501 -19.085 -20.135 -9.24 74.54 -20.269 -25.730 -27.642 -26.846 -27.783 -14.9 l \\ l where: ICO = PTop PBot x 100% ETop + P3ag OCO = Channel Imbalance x 100% Channel Power l l c i 13-23 -}}