Text

Forwards Copies of Chapters 1,2,3,8,9,11,12 & 13 as Rev by Suppl 2 of Initial Startup Rept Re Testing Completed Between 780705 & 781005 IAW Reg Guide 1.16,Sec C.1.A. & Tech Spec 6.9.1 of App a to Facility License
	ML20062C909
Person / Time
Site:	Davis Besse
Issue date:	11/02/1978
From:	Jeffery Grant; TOLEDO EDISON CO.
To:	James Keppler; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
	1-37, NUDOCS 7811150124
	Download: ML20062C909 (112)
	v • d • e

{{#Wiki_filter:- - t. m e-- - 1 Totsoo Q:_, iiDISON j November 8, 1978 JAMES S. GAANT. I voce Paes feat Serial No. 1-3 Eneer,, hsa's 84t 9125 9-5232 dnVL ' Docket No 50-7y g 4 i License No. 07F-3 Mr. James G. Keppler Regional Director, Region III Office of Inspection and Enforcement i U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137

Dear Mr. Keppler:

Attache ( n.re. copies of chapters 1, 2, 3, 8, 9,10,11,12, and 13 as revised by Silpff8mut 2 o,f the Davis-Besse Unit 1 Initial Startup Report. This report, which covers the testing. completed during the period from July 5, j 1978 through October 5, 1978, is being submitted in accordence with Technical j Specification 6.9.1 of Appendix A to the Davis-3 esse Nuclear Potter Statien Operating License and the Regulatory Guide 1.16, Section C.l.a.. a very truly yours, 1 i kw a I cc: Dr. Ernst Volgenau, Director Office of Inspection and Enforcement i Enc 1: 25 ccpies Mr. William G. Mcdonald, Director Office of Management Information and Program Control Enc 1: 2 copies f Q ~ O t Y j THE TCLECO ECISON COMPANY E0tSON PLAZA 200 MAC 3ON AVENUE, TCLECO. CHIO 43652

Instructions for inserting Supplcnent 2 refision to Davis-Besse Unit 1 Initial Startup Report REMOVE INSERT Supplement 1 Title Page Supplement 2 Title Page Entire Table of Contents Entire Table of Contents Pages 1-1 thru l-S Pages 1-1 thru l-14 Pages 2-1 thru 2-4 Pages 2-1 thru 2-4 Pages 3-1 thru 3-15 Pages 3-1 thru 3-15 Pages 8-1 thru 8-15 Pages 8-1 thru 8-15 Pages 9-1 thru 9-16 Pages 9-1 thru 9-16 Pages 10-1 thru 10-4 Pages 10-1 thru 10-3 Pages 11-1 thru 11-10 Pages 11-1 thru 11-12 Pages 12-1 thru 12-3 Pages 12-1 thrt 12-4 Pages 13-1 thru 13-23 4 e e O 1 S

..-..__..._......._.--__y..._._ DAVIS-BESSE NUCLEAR P0;ER STATION UNIT ONE INITIAL STARTUP PIPORT COVERI:;G APRIL 23, 1977 THRCUGH A?RIL 5, 1978 SUPPLEMENT 1 COVERING APRIL 5,1973 TEROUGH JLLY 5,1978 SUPPLEME:iT 2 COVERING JULY 5,1978 THROUGH CCTOBER 5,1978 e '~' I , TCLEGO r:' '- 'e n t b

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,__.=. TABLE OF CONTENTS i l Page Section 1-1

1.0 INTRODUCTION

2-1 i 2.0 SUM:!ARY ' 3-1 3.0 INITIAL FUEL LOADING i 4.0 POST FUEL LOAD PRECRITICAL HOT FUNCTIONAL TESTING 4-l' i 4-1 4 4.1 Reactor Coolant System Flow Measurement \\ l

4. 2 -

" Reactor Coolant System Flow'Coastdown-~ 4-2 Measurement 4.3 Reactor Coolant System Hot Leakage Test 4-2 4.4 Pressurizer Opcrational and Spray Flow Tests _ ~4-3. 4.5 - Control Rod Drive System Operational Test 4-3 5.0 INITIAL CRITICALITY 5-1 i 5.1 Preliminary Approach to Criticality 5-1 5.2 Final Approach to Criticalfty 5-1 a 6.0 CORE PERFORMANCE DURING ZERO POWER PHYSICS TESTS. 6-1 ii 6.1 Nuclear Instrument Overlap 6-1 2 6.2 Sensible Heat Determination-6-1 l 6.3 Reactimeter Response Checkout 6-2 f 3 l 6.4 All Rods out Boron Concentration 6-2 6.5 Temperature Coef ficient of Reactivity Measurements 6-3 J 6.6 Control 'od Reactivity Worth Measurements 6-3 \\ d 6.7 Ejected Rod Worth Measurements 6-4 6.8 Stuck Rod Worth-and Shutdown Margin 6-4 Measurecents j 6.9 Soluble Poison Worth Measurements 6-6 a' i E e -,,---.-.,.--,,-,,,...,....-,,,,.n c- -c,.n.,.

l I f _Section Page l 7.0 CORE PERFOILMANCE DURING P0"ER ESCALATION SEQUENCE TESTS 7'l i 7.1 Nuclear Instrumentation Calibration at Power 7-l' 7.2 Reactivity _ Coefficients at Pcwcr 7-3 7.3 Rod Worth at Power' '7-5 7.4 Core Power Distribution Tests 7-6 7.5 Pseudo Control Rod Ejection Test 7-6 7.6 Dropped Control Rod Test '7-7 7.7 Incore Detector Test 7-8' 7.8 Power Imbalance Detector Correlation Test .7-9 8.0 NUCLEAR STEM 1 SUPPLY SYSTE'! (NSSS) PERF0Pl! RICE 8-1 ~ 8.1 Unit Load Steady State Test 8-1 8.2 USSS Heat Balance Test 8-1 8.3 Integrated Control System Tuning at Power 8-2 9.0 UNIT PERFOPJ4ANCE DURI';G TRANSIENT AND ABNOPJ4AL 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 10.3 SEC0"DARY PLA'iT PERFORMANCE A'.'D STARTUP EXPERIENCE 10 10.1 Turbine / Generator 10-2 10.7 Condenser 10-2 10.3 Circulating Water System 10-3 10.4 Feedwater Systems 10-3 11.0 UNIT MONITORING - CHEMISTRY AND HEALTH PHYSICS 11-1 11.1 Shield Survey 11-2 G ii

- T

m_._____

Section Page ~ 11.2 Site / Station Survey 11-3 11.3 Reactor Coolant Chemistry Test 11-3 11.4 's 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 U::1T TRIPS 12-1 l 13.0 CORE PERF0PXCCE FOLLC;TI :G BPRA Al;D ORA RE':0 VAL 13-1 13.1 Core Performance During Zero Pever Testing 13-1 l Core Performance During' Power Escalation Testing 13-6 13.2 l l l l l I l l i I i i I l iii e e

-..-- -. = l i { l.0 INTRODUCTION-Davis-Besse Nuclear Power Station (DBNPB) Unit 1, located on the southwestern shore-of Lake Erie near Oak Harbor, Ohio, is a l Babcock and Wilcox pressurized water reactor rated at 2,772 MWt. l The turbine-generator is capable of a not electrical output of.906 l MWe. The Nuclear Steam Supply System (NSSS) employs a once - through steam generator. The Facility Operating License (NPF-3) for DBNPS Unit 1 was issued to the Toledo Edison Company on April 22, 1977. The first fuel assembly was loaded into the core on April 23, 1977, and fuel loading was completed on April 27,197/, af ter a total fuel load time of 83 hours. Initial criticality was achieved on August 12, 1977, after a Post Fuel Load Precritical Hot Functional Test P,ro gram. Zero power physics testing com=enced after achieving initial criti-cality on August 12, 1977, and was completed on August 20, 1977. The zero power'meesurements of core ' performance were performed at a Reactor Coolant System temperature of 5300F,- and a pressure of 2155 psi. Power esesIlation coemenced on August 24, 1977, and-the turbine gen-erator was initially loaded on August 23, 1977. Further power level increases were successfully completed at ea h 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 1.0-8 show the chronological power history during the startup test program. Figures 1.1-1 thrcugh 1.1-4 show 6 the chronological core burnup during the startup test program. The initial transmittal on May 8, 1978, of the Startup Report contained test data which summari:ed 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 ces-pleted by April 5, 1978, it could not be included in the initial trans-t l 2 mittal. Technical Specification 6.9.1.3 requires supplemental reports be submitted' l to the Startup Report on a quarterly basis until testing is completed and the unit resumes commercial power operation. Davis-Besse Unit 1 was shut-down for a maintenance outage and, therefore, no further testing was com-pleted in the period from April 5, 1978 through July 5, 1978. i l l 1-1 l I = a e ,y,, .-v +w-, .--y , + - w-m-4 y, y---

The second supplement updated the Startup Report to contain test results of testing completed between July 5,1978 through October 9,1978. The changes made to the Startup Report by Supplement 2 are indicated by a 2 , vertical line and a "2" in the lef: hand margin. Supplemental reports will continue to be submitted on a quarterly basis. until the testing is completed. 6 e 9 1-2 O ^# ~ - humanes N M. - g,

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SUMMARY

The unit has been operated at power levels up to and including 1007. The performance full power since the completion of startup testing. Testing and operation of the unit has generally been satisfactory. of the NSSS and the turbine generator revealed sc=e =inor problecs/ conditions that were other than predicted, hcwever, none of then ~ adversely affected plant safety. The problems encountered were not this size. unusual for the startup pregram of a unit problen at a similar reactor did arise during power A significant Two burnable escalation that could af f ect Davis-3 esse Unit 1. poison rod asse=blies (SPRA) were found outside of their fuel assem-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-3 esse Unit 1, Babcock and Wilcox. On April 5, 1978, Toledo Edison p'ersonnel 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 extre ely remote, they recommended, as a precautionary =casure, the reactor coolant flow Reactor Coolant Pump 1-1 was shutdown on April 5,1978. be reduced. Davis-Besse, nor No BPRA locking techanism failures have occurred at in five previous Babcock and Wilcox units using the same BPRA lock-All 68 LPRA and all 48 orifice rod assemblies were ing mechanisms. 27, 1978 during the maintcnance outage removed frem the core by May as recenmended by Babcock and Uilecx to insure no f ailures of the 2 locking mechanien at Davis-Besse. Modified orifice rod assemblics for the two neutron source holddcwns were installed. 2.1 INITIAL FUEL LOADING (SECTION 3.0) Initi.1 fuel loading ccanenced on April 23, 1977 at 1357 hours. The entire fuel loading ser:ence experienced only ninor delays and was accomplished in approximately four days. 2.2 POST FUEL LOAD PRECRITICAL HCT FU: CTIONAL 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 testing included a Reactor Coolant Sys tem Flow.*. cat urement, Reactor Coolant System Flow Coast-down, Pressurizer Operational and Spray Flew Test, and Control Rod Drive System Operational Test. All test results satisfied the Davis-Besse Unit 1 Technical Specifications and all test acceptance criteria were met. The tests ccepleted were: (a) Reactor Coolant Systen Flow Measurement, TP 200.11 (b) Reactor Coolant Systen Flow Coas:dewn Measurecent, TP 200.11 (c) Prescuri cr Operational and Spray Flow Tests, TP 600.13 (d) Control Rod Drive System Operational Test, TP 600.17 (c) Reactor Cpolant Systen Hot Leakage Test. TP 600.10 (ST 5042.02) 2-1 ~

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 li=it (100".) and co=- bined, Groups 6/7 were withdrawn to the 75% position. Criticality was then achieved by deborating from an initial reactor coolant boren concentration of 1843 ppm to a final concentration of 1520 ppe. 2.4 CORE PERFORMANCE DURING ZERO POWER PHYSICS TESTS, TF 710.01 (SECTION 6) Following initial criticality, core perfor:ance taasurerents were conducted during the Zero Pcwcr Physics Test Progra= f ro= 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 Instrucentation Overlap (b) Sensible Heat Power Level (c) Reacticeter Response Checkout (d) All Rods Out Eoren Concentration (c) Te perature Coefficient of Reactivity Measure ents (f) Control Rod Reactivity Worth Measure:ents (g) Ejected Rod Worth Measure ents (h) Stuck Rod Worth and Shutdown Margin Measure ents (1) Soluble Poison Worth Measurecents 2.5 CORE PERFORMANCE DURING POWER ESCALATION SEQUE':CE TESTS TP S00.00 3 (SECTION 7.0) Core perfor ance casurencnts were conducted during the Power Esca-lation Sequence Test Frogree. Testing was conducted at the pcVer level plateaus of 15%, 400, and 757, of ' total thernal core power. All test data and results satisfied the Davis-Eesse Unit 1 Technical Specifications and test acceptance criteria. The pcwer escalation core performance data and nearure:ents are contained in the following tests. Test (h) below still requires sc=e retest data prior to con-pletion. (a) Nuclear Instrucentation Calibratica at Power, TP 800.02 (b) Reactivity Ccefficients at Powe r, !? E00.05 (c) Rod Reactivity Worth Test. TP 800.20 (d) Core Power Distribution Test, TP S00.11 (e) Pseudo Control Rod Ejection Test, TP S00.23 (f) Dropped Control Rod Test, TP 300.29 (g) Incore Detector Test, TP 800424 (h) Power I: balance Datector Correlation Test, TP 800.13 2-2

-..-.=

1 i l I 2.6 NUCLEAR STElli SUPPLY SYSTF21 (NSSS) PERFO?2!A';CE (SECTION 8.0) A list of the tests perforced during power operation related to the monitoring of the NSSS perferrance is presented belcw. - In all, the perfor:ance of the 5555 was satisfactory, and as expected. Tests (b) and (c) below still require sene additional ~ data prior to completion. l 2 Test '(d) below has yet to be performed. (a) Unit Load Steady State Test, TP 800.12 (b) NSSS Heat Salance Test, TP 800.22 (c) Integrated Control Systes (ICS) Tuning at Power, TP 800.08 (d) RCS Natural Circulation Test, TP 800.04 2.7 U'11T PERF0PECCE DURING' TR.'J*SIE';T AND A3503 MAL CONDITIONS (SECTION 9) The purpose of the unit perfor ance tests is to verify the unit can be maintained in a safe cendition during and following load tran-sients and various abnormal conditions. Tests (a), (c),-(d), (c), 2 and (f) listed below have not yet 'been cenpleted. (a) Unit Load Transient Test, TP 800.23 (b) Unit Power Shutdewn Test, TP 800.15 '(c) Turbine / Reactor Trip Test, TP 800.14 l (d) Loss of Offsite Pcwer TP 800.26 (c) Unit Load Rejectica Test, TP S00.13 (f) Shutdown fros outside of the Centrol Roos, IP 800.25 2.8 SECONDARY PLANT PERFORM /2;CE (SICTION 10) This section provides a brief s"--ary of the cajor difficulties encountered with the secondary systens during power escalation. The secondaty systens that are covered include: (a) Turbine-Generator (b) Condenser (c) Circulating L'ater Systes (d) Feedwater Systen 2.9 UNIT MONITORING (CED!ISIRY J2;D EE.'d.TH PHYSICS) (SECTION 11) This section presents a list of the unit ronitoring and testing per- . forced with regard to health physics and che=istry during various phases of the startup test progra=. Tests were conducted during initial fuel loading, reactor startup, power escalation and' power operation. 4 t (a) Shield Survey, TP 800.01 (b) Site / Station Radiation Survey, TP 800.03 (c) Reactor Coolant Che:istry Test, TP 500.01 (d) Steas Generator Che=is'try Test, TP 500.02 ] (e) Initial Radioche:istry Test, TP 5C0.03 (f) Effluent Monitoring Test, IP 360.01 2-3 \\ ~

2.10 C;IT TRIPS (SECTIO:I 12) This section is a listing of all unit trips and applicable infornation which occurred during the period from initial fuel loading through power escalation and operation. s 2.11 CORE PERFORMA.;CE FOLLOWING BPRA'S A';D ORA'S REM 0','AL (SECTION 13) This section contains a description of physics testing perfor=ed after the outage which removed the BPRA's and ORA's. Testing described includes: 2 (a) " Post Refueling Physics Testing", ST 5010.03 (b) " Reactivity Coefficients at Power", TP 800.05 (c) " Core Power Distribution", TP S00.11 (d) " Rod Reactivity Uorth Measurements", TP 800.20 (e) "Incore Detector Test", TP S00.24 (f) " Pseudo Ejected Rod Test", TP 300.23 (g) "NI Calibration at Power", TP 800.02 (h) " Power Imbalance Detector Correlation Test", TP 300.18 ( + 2-4 e en

.c 3.0 INITIAL FEL LOADING The Davis-Besse Unit 1 initial core contained 177 fuel assemblies. 53 control l 2 rod asne blies (CRA), 8 axial power shaping rod asse:blies (APSRA), 68 burnabic poison rod asse:blies (EPR.>.), 48 orifice rod asse:blies (ORA), and two installed neutron source asse:blies. Further details and descriptions of the reactor core and co:ponents can be found in Chapter 4 of the FSAR. The initial fuel loading cc=:enced when the first fuel assembly (NJ004U) was recoved from spent fuel pool location A-1 at 1357 on April 23, 1977, and was completed when the final fuel asse:bly (5J004C) was loaded into core position L-1 at 1808 on April 27, 1977. Actual fuel loading ti=c was approx 1:ately 83 hours. The fuel loading was perfor ed in accordance with " Initial Fuel Load Procedure", ?? 1502.04. The actual fual leading sequence is illustrated l in Figures 3.0-1 through 3.0-S. The initial core configuration is shown in 2 Figure 3.0-9. The neutron count rate was monitored centinuously and 1/M plots were maintained throughout fuel loading for both source range detectors (NI-l and NI-2) and both auxiliary neutron detectors (A and B). At least two independent 1/M calculations were co=pleted for each asse:bly (one in the containment using an auxiliary neutron monitor and one in the Control Room using a source range detector). After each cove =ent of an auxiliary neutron monitor, the 1/M plots were either renor:alized or a new baseline was obtained. The resulting 1/M plots were as expected and are sc :arized in Figures 3.0-10 through 3.0-13. During fuel loading, several ninor problems were encountered.' A description of the specific problems and their resolution is given below: 1. .On A'pril 23 at 1710, the west transfer sechanism became stuck between contain cat and the spent fuel povi with fuel asse:bly NJ0048 in the carriage. The assembly and carriage were pulled to the spent fuel pool via the cranc. Af ter transferring the assrably f rom the west ecchanism to the east, refueling recc cenced using only the cast transfer techanism. 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 techanis: was repaired. The west transfer cech-anis: was out-of-service for approxicately 24 hours. 2. 'The clutch on the east transfer =echanism appeared to be slipping. While the clutch was being repaired, only the west transfer carriage was used. The east estriage was out-of-service for approxi=ately 2 hours. 3. During fuel covement, the hydraulic pump on the main bridge stopped twice and had to be restarted. Total delay was approxi=ately one hour. 4. Appeared to be having proble=s with the overload cutoff on the = sin bridge. On April 26 at 0110, fuel leading was suspended in order to test the bridge. The bridge was verified to be functioning prcperly. 3-1

It was determined that the upender was out of plumb. Both upenders were replumbed and fuel loading reco==enced after a delay of approxi=ately 3.5 hours. 5. Near the end of the fuel loading, the east transfer =echanisa hung-up again. The west carriage was used exclusively for the re=aining 7 hours of fuel loading. The Davis-Eesse Unit 1 Technical Specifications state that during fuel covement, the boron concentration shall be r.uch to ensure that the core restrictive of the following conditions are cet: 1. Either a Keff of 0.95 or less which includes a 17. A K/K conservative allowance for uncercainties, or 2. A boron concentration of & 1800 ppm, which includes a 50.pps conservative allowance for uncertainties. During initial fuel loading the second condition was the more restrictive. The measured reactor coolant boron concentration ranged from 1529 to 1929 ppa during fuel movement. The initial boron measurement made within 12 hours of loading the first fuel assembly was 1888 ppa at 0720 on April 23. The overall average of the 8 boron measurements taken during fuel loading was 1883 pga. Following the completion of fuel loading, the incore detectors were inserted into the core. Dif ficulties were experienced while inserting two cf the incores Detector 46 could not be incerted into fuel asse=bly NJ0046, but eculd be inserted into the dummy asse bly. After a visual inspectica of NJ0046, fuel element NJ0053 was reshuf fled f ro= core location A-6 to core 1ccation ?.-10, while moving NJ0046 to core location A-6. Incore detector 46 was.then inserted into NJ0053 with no difficulty. 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 welcs had excessive beads on the inside of the tube. A wire was pushed through the tube to dislodge any loose obstructions. Detector 2 was then installed into asse bly NJ002D with no additional dif ficulties. d 3-2 PN8D 94fPN - - - _ _ m-*P are ees p ..w yeew

- -._----.~_ ___ Figure 3.0-1 Davis-Besse Unit 1, Cycle I I Fuel Loading Sequence 1 of 8 N R P O M 1, 1; 11 G F E D C B A l' NI-l 2 O 3 1 4 3 l l l -- 6 1 J 7 B d 5 25b / ,25a i j j. 9 e / ,/ 24 23 20 [ 4 10 i B l 22 16 17 19 11 11 13 14 IS 21 ~ 12 5 7 9 12 15 + 13 3 4 8 10 O 14 NI~ 2 IX 6 A 15' ! X Sourec.\\ssenble D3SPS - Unit 1 Startup Report @- Auxiliary, ::cutron t etecter J-Fuel Loading sequence h Auxill ry ::cu;;ren Uctector B Figure 3.0-1 3-3 e

W I Figure 3.0-2 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence'2 of 8 N 1 a j P. P O M M L K 11 G 'F E D C B A i i I 1 i l V NI-1 -2 O 1 l 3 l 1 4 5 g I 6 43 45 36 38

t. 0 42 31 33 35 41 l.6 a jf 30' 34 39 44

[, 32 37 27 29 26 8 / 28 ),, j i W WA l ^ o 1 NI-2 l .15' 3 I Source Asse= hic 3 DBNPS - Unit 1 ] @-AuxiliaryUcutronDetecto.A Startup Report Tucl Loading Sequence 1 h Aixiliary' Neutron Detector S sure 3.0-2 3-4 ~ ....w.r. a ,-c

I .\\ i Figure 3.0-3 Davis-Besse Unit 1, Cyc.'e I i Fuct Loading' Sequence 3 of 8 j. M R P O N M L K 11 G. 7 E D C B A 1 i NI-1 2 / .O 3 64 66 67 71 4 57 59 61 63 70 72 5 50 52 54 56 62 69 73 j/ 47 49 55 60 65 / / 48 53 58 l /// ,/ $1 f / b s / /_ V/%Q/V4%%V/GD WxMWS9xVx 22W/%%X/ M % V4 o 14 NI-2 13-Scurce Assembic DENPS - Unit 1 i h-Auxiliary ::eutron Detcetor A Startup-Report Fuel Loading Sequence, h Auxiliary ::cutron Detector.E 8"

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4 1 Figure 3.0-4 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 4,of 8 l N R P O M L K 11 G F E D C B. 'A l l .t F 92a 91 90 89 I ) 88 87 86 85 84 83 82 81 80 1l2h' 79 78 77 76 75 / l 74 /f o 7/ % % % y 'n 7A%'///4?w/7/A Vy,7/gy,g/y,/ 9 / '////////,d/g/7(D 7,g/;g7,y,7/g/,' W N A Y/ o 14 l NI-2 15' Source Assemble D3NPS - Unit 1 h-AuniliaryNeutronDetectorA .Startup Report 1 Fuel Loading Sequence E

h Auxiliary !!curren Detector B

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Figure 3.0-5 Davis-Besse Unit 1, Cycle I Fuel Loading Sequence 5 of 8 y 4 R P O N M L K II G F E D C B A 123 122 8 3 121 120 119 118 117 116 / 115 1114 113 112 111 110 109 108 107 106 105 104 103 102 7 101 100 99 93 97 96 95 94 93 / / %3V$%% Em o / 14 NI-2 g. i Source Assemble "DBNPS - Unit 1 h-Au::iliary Neutren Detector A Startup Report h.\\uxiliary Neutron Detcetor B j F g rc

f Figure 3.0-6 Davis-3 esse Unit 1, Cycle I Fuel Loadirg Sequence 6 of 8 N i R P 0

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Figure 3.0-7 Davis-3 esse L' nit 1, C.;cle I l l k Fuel Loac ir.; Sequence 7 of 8 N R P O N M L K 11 G F E D C B A I i 3 I !G)7/A///// $W/S////'VEN!2', "'~ / '//'WX/////##N//X/// ','~o e---A@I%VN//ENEXE/Y/N'//A 1 I %VN/7ENN//X///V///) ; ~ i l T'6M/HHMA////H/%% =, a ! {////##N//XHN//#/ius b \\ as '#/46VWX/XNN//Eka lu, i / l t,o ns i l 'E&/f'N///NN/%VA5 ua u= a

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n - - - ~ ~- F Figure 3.0-3 ~ l Davis-Besse Unit 1, Cycle I h Fuci 1.oaJit:g Sequence 8 ot' 3 N 4 I R P O N M L-K 4 G F E D C B A l 2 / // /175a, Y#/#XWHMM/%Vf/ 4%7 /X/#/MM#rW4 W.#Z% YAW ###LW/: 2M%%V/%%%V## YH#Z% / f f A f/'%%GW/#/%%%%W)%% / i %%tXVMU##EA%%%%%0 ? l W//'%WO% MY M'/M%% //%W7#MX6/72 MF/ / 79,gy,%7py/W/7px o [ p/ j HI-2 174 / 176b - Source Asse:able DBNPS - Unit 1 h-AuxiliaryMoutronDetector.) Startup Report Fuel 1.onding Sequence h Auxiliary Neutron Detector 3 ':'?

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5: s i 8.0 NUCLEAR STL*dt SUPPLY SYSTEM (USSS) PERFORMANCE The purpose of the tests described in this section is to conitor the performance of the Nuclear Steam Supply System (NSSS) to obtain baseline data and to verify the NSSS performs as designed. Four tests are used to complete this purpose. y +; l One of the NSSS perf ormance tests is entirely ccepleted - TP 80U.12. " Unit 1r. Load Stead ~y State Test". TP 800.22, "NSSS Heat Balance Test" requires a ~ _ ", 2 retcot after the range of the RCS flow =eters is completed. Also, TP 800.08, " Integrated Control System Tuning at Power", is complete except for the 100%

-j test data. The remaining NSSS performance test, TP 800.04, " Reactor Coolant y; f System Natural Circulation Test", has yet to be performed to cc=plete the jg testing of the MSSS performance.: o IK

? E;- 8.1 UNIT LOAD STEADY STATE TEST, TP 0800.12 je Primary and Secondary System steady state parameters were =casured during power escalation to obtain baseline data. This inforration was gathered l' '. a during Phase I of TP 800.12, " Unit Load Steady State Test", at approximately -~ 0%, J5%, 30%, 40%, 65%, 75%, 90% and 100% full power. Steady state condi- .o, tions were established bef ore any data was obtained. Sr.veral parame ters .{ J." ] were compared with design values to verifv that the response for thesc . c. parancters, 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 K, 4 values were derived f rom an average of the measured readings. As shown on A Figures 8.1-1 through 8.1-7, all parameter s recorded responded within their acceptable bands. Phase II of TP 0800.12 was performed f rom 0% to 15% full power. Data was accumulated to check the relatienchip between Tave and reactor power. ?E. This inforcation was used to adj ust the OTSG low level setpoint to bring vy..., Tave within SS2 i 1 F at 15% power. 3 J... 8.2 NSSS HEAT LALANCE TEST, TP 0800.22 i.jk TP OS00.22, "NSSS Heat Balance Test", was perfor:ed during power escalation ~., " with the intent of achieving the follcwing objectives: y 1. Verify the accuracy of the computer's heat balance calculation. r., 2. Provide baseline data for comparison with subsequent heat balance $Ji checks. i-$ 7 : [ :.. -. 3. Tetermine the reactor coolant flow rate. This test was conducted at powe r levels of 15%, 30%, 40%,~55%, ~75%, 90%, and 100% full power. Data for primary and secondary heat balances was 'c. taken at each testing point. The balances were co: pared to the computer J. calculated heat balanccs. In all cases, the hand calculated and computer calculated values agreed to within the required t2% acceptance criteria, y The results of these connutations are sunnarized in Table 3.2-1. At 100% of f ull power, the hand calculated primary heat balances for each .,$t. loop were ccapared to their respective secondary heat balance. Since the f deviation for' bcth loops was greater than 1%, a new range for the primary J?46* _ 2 flov meters for both loops has been calculated by setting the primary heat 3-balance equal to the secondary heat balance. A retest will be performed to verif y the deviation *is less than 1%. it ' 81

8.3 INTEGRATED CONTROL SYSTEM TUNI:;G AT PO'JER,.TP 0300.08 All but the 100% full power trancient testing portion of TP 0300.08, 2 "Integ ated Control System Tuning at Power", were performed threughout hot functional testing and during power escalation. This procedure was perf or=ed to verify that optimum plant perfor=ance and control is obtained by means of the integrated control system. The ocjor ICS related control functions tested are listed below: 1. Thermal efficiency between the primary and secondary system. 2. Electrical output versus feedwater flow. 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 paraceter 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 tesced were within their respective acceptznce criteria. B-2

/ z / i TABLE 8.2-1 IIEAT EAIJ.NCE SUFO!ARY ~ ~ ~., E LOOP 1 , LOOP 2 { Nominal P2-(Comp) P2 (lland)

7. DIFF P1 (Comp)

P1 (lland) P2 (Con,p) P2 (lland) %-D1FF j Power P1 (Comp) P1 Oland) __%) itJt tatt FrJt ? t'.J t (PI-P2)(lland) FRJt InJt FRJt !CJ:: (PI-P2)(IIand) I ( 184.63 0.32% 212.93 204.04' O.32% l- { l 1 J 3.% 4 15 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% I 62:5 947.43 842.00 903.26 895.29 1.92% 817.24 824.77 906.05 887.905 2.28% l 72.7 961.95 991.508 925.11 1027.'13 1.29% 1042.70 951.058 1039.19 1020.04 2.49% l l 88.5 1151.82 3162.01 1269.53 1245.66 3.02% 1107.33 1127.35 1253.06 1227.96 3.63% m i u' 100 1272.82 1287.09 1381.73 1380.78 3.38% 1226.92 1262.145 1374.40 1374.57 3.98% b L'here: P1 (Comp) = Primary computer Imat balance P1 (lland) = Primary hand beat balance P2 (Comp) - Secondary cony ote-heat balance P2 (IIand) - D condary hand beat balance D3NPS - UNIT 1 STARTUP REPORT llEAT BALANCE

SUMMARY

TABLE 8.2-1 1 i

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9.0 U'i1T PERF0PJfANCE DURING TRANSIENT AND A350?JiAL 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 unit load rejection test will be conducted at an initial power level of 100% of full rated power. A reactor trip test was completed at the 40% of full power level. A reactor trip test at 100% of full power level will be performed. A e ntbine trip f rom 75% of full power was also completed. Proper per-formance of the unit after a turbine trip frca 100% of full power-will be verified. The plant will be shutdown f rom the Auxiliary Shutdown Panel f rom an initial power 1cvel of 15% of full rated power. l l A loss of offsite rower test, including a loss of external lead, will I 2 be conducted from en initial power icvel of 15% of full power. Load transient tests included 10% FP transients at 40% and 75% of full 2 l power and 50% TP tronsients at 90% of full power. Proper operation of the integrated cont ol system cross limits and rate limits were also verified. A unit power shutdown test was performed en October 13, 1977, from an initial power le >c1 of 15% of full tated power. 9.1 TURBINE AND REACTOR TRIP TEST TP 800.14 The reactor trip from 40% of full power was successfully completed on Decc ber 15, 1977. The reactor trip was initiated manually, and the Reactor Trip Escrgency Procedure, EP 1202.04, was implemented. A rtactor trip test f rom 100% of full power will be done to verify the adequacy of the Nuclear Stean Supply System and provide suf ficient data to optimize the. performance of the control systems. /. 75% of full power turbine trip was completed on April 2,1978. This pun ided core data to optimi:e the operation of the Integrated Control Sys ;.em. TLe 75% turbine trip was repeated on September 10, 1978, to test the blow-back of the main steam safetics and changes cade to the Integrated Control System during the EPRA removal =aintenance outage. The main steam safeties 2 operated properly during the test but the ICS dicplayed the need for further tuning. As a result of the analysis of the ICS performance af ter the trip, adjustments were cade to the ICS. Either a load rejection or a turbine l trip test will be perferred to verify proper operation. The turbine trip was manually initiated from EhC' Panel #1 on C-5713, and tki Turbine Trip E=cr;ency Procedure EP 1202.03, was implemented. 9-1 O ~ -v MN'N* emeweeesesmane _ em .=

Data on a turbine trip frem 100% of full power will be obtained by a turbine trip if the data obtained from the '.' Loss cf Load Test", TP 800.13, is not sufficient to verify proper perfor:ance. A Reactimeter and Brush recorders were used to record the applicable The results of the reactor and turbine trip tests are su==arized data. in Figures 9.1.1 through 9.1.10. The collected data verified that the reactor coolant system remained within its safety liaits. TP 800.23 9.2 UNIT LOAD TRANSIENT TESTS The unit load transient tests demonstrated that the unit can be caneuvered in a controllable canner at 5% FP per minute from 150 to 73: and from 75% to 15% of full power. The unit load transient tests also verified proper operation of the ICS cross limits and verified satisf actory low power level to.the tripping of one response of the ICS to control the unit subsequent reactor coolant pump or ene main feedwater pump. Load transients of 10% FP vere conducted at 5% FP per ninute f rom 40% to 30% to 40% of full power in the integrated control mode and turbine following 3% FP per minute in the steam generator / reactor following mode. node, and at 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% of full power in the Reactor Coolant Pump 1-1 was r -started and a integratad control code. load transient of 20% FP was conducted at 5% FP per minute f rom 20% to 40% of full power, also in the integrated control code. ICS cross limits were verified by: ~ from 40% to 50% of full Imposing a 5% FP per minute load transient 1. power with the feedwater control subsystem in the canual code of control. The feedwater-to-reactor cross limits limited the reattor de:and to 45% of full power when the feedwater demand exceeded feedwater flou by 5%. Imposing a 5% FP per minute load transient from 40% to 50% of full 2. power with the reactor control subsystem in the manual =cde of control. The reactor-to-feedwater cross licits limited the feedwcter demand to 45% of full powe r when reactor demand exceeded the neutron power by 5%. Imposing a 5% FP per ninute load transient from 40% to 001 of full power 3. The with the reactor control subsystem in the =anual code of control. reactor-to-feedwater cross limits limited the feedwater de and to 35% ,of full power when the neutron power exceeded the reactor de:and by 5%. from 75% to Load transients of 10% FP vere conducted at 5% FP per minute 65% to 75% of full power in the integrated contrcl code and turbine following code, and at 3% FP per minute in the steam generator / reactor following mode. With the reactor operating at 64% of full pcwer, M.ain Feedwater Pump 01 was tripped, and the ICS ini iated a plant run'aack to 59% of full power TP 800.08 at 50% FP per minute, in the integrated control code. 9-2 O eN ew w ew *. m e--amme

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A list of unit load transient testing is given in Table 9.2.1. A reactimeter and brush recorders were used to record the applicable data. The collected data verified that the unit cdn be =aneuvered at 5% FP per minute in the integrated control mode, af ter optimizatien, without a reactor or turbine trip, relief valve or turbine bypass valve actuation, or exceeding any of the limits imposed by PP 1101.01 "NSSS Plant Linits and Precautions", thus satisfying the acceptance criteria. The following testing is yet to be completed: 1. The high power positive rate limit will be verified by imposing a 5% FP per minute load transient at 85% of full power in the integrated control ecde. TP 800.08 2 2. Load transients of 50% FP will be conducted at 5% TP per =inute from 92% to 42% to 90% of full power and at 3% FP per minute frca 90% to 92% of full pouer, in the integrated control mede and turbine following mode, and at 3% FP per minute in the steam generator /reac-tor following mode. TP 800.23 9.3 UNIT POWER Sl!UTDOWN TEST TP 600.15 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% pcwcr, and to obtain baseline data for subscquent shutdowns. The shuedown was performed from an initial power level of 15% of full rated power. The cooldown was conducted to a final reactor coolant _ system ~ temperature of 531 F and the reactimeter data was obtained by the Plant Cotputer's Operator Special Summary Group. The results of the unit power shutdown test, summarized in Table 9.3.1, verified :hst a Turbine-Reactor shutdown can be perforced (Section 4 of Plant Shutdown and Cooldown Procedurc. I? 1102.10) without exceeding the limits of the Nuclear Steam Supply Limits and Precautions, PP 1101.01, Section 1. Thus, the acceptance criteria was eatisfied. 9-3 1 e ~ ~ ' * ~ l

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TABLE 9.3.1 UNIT POWER SHUT 00'a"i TES T S Da>>.RY Initial Final Paraceter Value Value Test Coc=enced/Cocpleted 1009 hrs. 2359 hrs. Reactor Power / Source Range Level 15% FP 0% FP RCS Tave/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 L'evel 6% 46% 6 lb=/hr - O lb /hr SG #1 FU Flow 0.55 x 10 SG #2 FW Flow 0.54 x 106 lbu/hr -- O lb=/hr SG #1 Steam Outlet Pressure 882 psig 871 psig SG 02 Steam Oatlet Pressure 883 psig 873 psig Feedwater Temperature 299 F 230 F RCS Boren Concentration 1339 pp:3 1304 pp=3 RCS Contractica (Makeup) Volute 4000 gallens 9-6
1030 L_._..._, .._-_L l l i n i i 0 r i i r-m v -I w t n- -t= 1000 _ m _.__.t d ,__.__ a 4- - -}-..-..-.----. M ^ ___4 {.._._ A I _____7 U _ _ _ j.___. y-a i 950 t__ O g - i 3_ _.-.l, I g t s I _.4 M i t s_ = _. _ _ _.___._~1.-.____._..L.___- _-..__.7_-__ e 900 _... _ 5 i ._. _. 4.._ _ - }..... m _____4,___._. p_ g q -.,. --L._.-__.,._4 -.-t I _. d ___.._..... _...., _.. I __l..._. .. ~. ...._L.__. .....l _._.._.t _.-... I.. _... 850 ..___.4___ ..__p..___. A S 1 2 3 4 ue ELAPSED T11E FROM TRIP (MI:.'UTES) DENPS - Unit #1 STARTUP REPORT SG #1 OUTLET FRESSURE VS. ELAPSED TD2 FROM TRIP FROM 40*. FP FIGL7.E 9.1.1 9-7 . -.- e
a.,.-. _
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-- ~ v t,' 'l g e it 1050 _ 1y. [- y. 1 ^ .}_ ? o, -t }_ _ - _ -... m - t-- m 1000 1 i.. v _-} y p...{_. p _~ . 7._ l m N, _ y _ __ L.. _..Y.__. v . _ __3 g 950.;! .m ~_. sa a 4 y b ..__ _-,}.-_. _-4._. __.~.~.a. l }._._ _ 3 m m j._ 7 g L. _. 900. ___ _.- r___ p__.._ __g. j __._ __ e m
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_ _ _t._ } -p_.. m..._ 850 1 2 3 4 o. u H El APSED TD:E FEOM TRIP (MINUTES) 2 1 4I C DBNPS-Unic #1 STARTUP REPORT SG.#2 OUTLET PRESSURE VS. ELAPSED TDIE FROM. TRIP-FROM 40% FP 6 FIGURE 9.1. 2 9-8 4 A
j _v.____ t _. _i. l r__ _ 200 t, 7 1 i i 4 - i i 175 - I 1 t i a 8 a i A t 150 4 c i i 1 ~ I s 44 i i i v g i i m j l j i '-3 125 ~ r i w _{ s N p. g x D, i e u u va y }.__. _ ,j j .t p 100 1 i o i t i i i _ _ _ _ w h i -.i H-- l + v i j G3 + l__ i + ~ e. t- } s a I [.. 8 ? i L.. p __.....p. 4 L ..._..A t ~ ---_L._- 50 f ~ I i e, I i I a I 1 i t 25 _. o. t 1 2 -3 4 H ELAPSED TD5 FRCM TRIP (MI,'UTES) J D3NFS - Unit 01 STARIL*P RI? ORT CO l?I';SATID PRES-SURIZIA LIVIL VS. ELA?SID TD2 FROM TRIP i. 9-9 FPDM 40.". FP FIGURE 9.1.3 - ~. - ~. - -m-- -.r ~ ~ _.,,,, _j
75 -F,
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p 1 m t i i W ~:. o Z .~ i I r. v d 50 .. _ _ _4_. q 2 _.__ _ __} g ~ p.. n. 1 i - - - o i p i j g i -- t .. __. _i__ ~~._E__._._;__~_~~ '~ g d m, v,y. s.. %.., ,p o 25 i m . __._s a. - - -+. -.._-a___...., .._.___. ~._ _. . t--- -..- [. _ ___ _ ___ _.. F _. _ L__ ___ _,_i. _ a -._7_-_.._ -_ _r f -.. 1_ i -+-- e m __. _-.. m-f. _ p_.__ ___ _ a.. _ f.. --- a

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~~ - --~ 7 ~ f.. _ _.. -t. _ 0 ' ' ' ~ ~ " ~ ~ ~ ~ ...._ __ g._ - ---i------- T.I ~ o. u 1 2 3 4 u e ELAPSED TIME FROM TRIP (liII.'UIES) D2NPS - Unit #1 STARTUP REPORT SG Ill STARIUP LEVEL VS. ELAPSED TUIE FROM "?IP FROM 40% FF FIGURE 9.1.4 i 9-10 . 98* 8* *6 h remos _ og,ww.m.,
s 4 75 __.,4_.. _ J. ....j___. _I g .._.__7._..._ _ _ _ _ _. _..}...._. ..._.{._____ w
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.u. O ..t.. a 50 ;._ [-.. w:- .._.t....... u y.. co s Cd U 25 m n N-8::= 1 e ..}.._... m }.. _. p.._ ..i..... . _ 1.. l 0 4 1 2 3 4 bs ELAPSED TD5 FROM TRIP (MINUTES) 1 DBNPS - Unit #1 STARTUP REPORT SG #2 STARTUP LEVEL V.P.. ELAPSED TlHE FROM TRIP FROM 407. FP FIGURE 9.1.5 9-11 i _e apeg e.se. -- 3.' O + ' -
2200 f
i i I c i r -- 2100 -' i L 2000 ~' ~ ~~ f __. _ m e <n u 6 j __.}.___ ._d 1 ._.__[ p 1900 y _j p . t _.. ___y -_. _ _ _...._.j._ _. o. to +- F - -- g, .i _ _ g _. u r, - t -------- } t-------- 1800 _4_... 2 i__ j . _.. q. t t i l 4.__ 1700. L r. --.. p _-_ _ t - _. 1 _4 .L 1 L' ___.. L _ f 7 7, T****~~~ ^ * ' ' ' ~ ~ ~ - .7__ 1 2 3 4 us* ELAPSED T122 FROM TRI? (MI::UTES) D2:~PS - Unit #1 STARTUP Pl? ORT RCS ??2SSI.72 VS. ELAPSED TDZ FROM TRIP FROM 40*. FF FIGL*P2 9.1.6 9-12 _. =. m_-.._..--.. q
2300 ..._5_._. ..... L__. + } - - - --. f h.. -d: i t. 2200 1 i E ._? I_ y.._ } 2100 i i i i i i i n i t ? i e o f .F4 .i. I i 2000 m v 1. I d i' ..} nm m i L g j 2_. 1900 m u _..t }_. 7 _r x 4 ._._ _ L___ _ .;. a - } __...___._.).... =__ I L. a ---^--b-~~~-~~'_*t------I_-.... _ _y.__. - -- -~ ~ ~~ ' { f r -- ~~'-~~~ 1800 ~ ~ ~ ~ ~ ' ~ ~ - ~ ~ ~ ~ ~ ~ ' - ~ ~ p ___ __ g_ _ _ +. p p ..__ _. r

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_ 4. i .a: } _ ......[.._ 1700 b 1 .i. .... _ l._ -- t. L t 1 i t I . __4 _._ a p _ u 1 p _ _ _1 I j __+ 1600 S 1 2 3 4 -4us ELAPSED TIME FROM TRIP (MI';UTES) DBNPS - Ur.it fl STARTUP 72? ORT RCS PRESSURE VS. T Lv.:.-. FO3.. TL o.o., t. TRI? FFOM 75': ?? 9-13 FIGUP2 9.1.7
125 .L._.._ .._.a .k..._......._ l _ L... r_ 7 t. ~_ __ l __. _ _J 3_ i -t-p i I L 100 .'.. _. } r-__._.i__...
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l,_ A. _. ^ p i i G. ___.- -..._..._-r-_. r i g y 73 1 ,1 4 a . ~ _ _.. l u . ___ _L. { + u
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3 y .._ ___ }. g 50 m .._.....g..._.. .(. _.y g .__._7__. _ _ _ L. _ __._c.__._ _ _. _... - _ _ _.._._j.__.. _a - p _._ c4 i O .... i........ L,. _____. w .._[.._.. . t._ _ _ t__..___...._..t__.__. l 25 L .._._..__..:...____q___.___- -4.._._.. -.p_-._.._ l 4 _.a_.. -._.. _. _ _ _ _ _. _.. t.- -[_. u._ _... 3. 7,_..._...._..}. __ _ _._... _ _ p..,._ ._ - _ p ...q... 7 ... q... ~..... _ O 1 2 3 4 o. wus ELAPSED TIIIE FRO:I TRIP (::I:' UTIS) i r DBh*PS - L'ait #1 STARTL*? RIPORT S G L., S T<,.,..,v 3 LEVEL VS. TD'E FOR TG.3INE TRIP FROM 75 FP FIGL'RE 9.1.8 9-14
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..___-._f,_..._ _.._ ;.._ ._ N _-......__7__- } - a __.._t_.. j p ta -_.___}_.__...____.j._.._, i 1 L p__ ___ _ _t.__.__-. _ _ _. _ t l _..__L__.__ p ..__..m___._.. _..__.p__.._.. i g }_ l } _... ___L___ .._.._...___._.r g ._.__p.__. ._..y.._._._ ._.. m a _____7.__._.
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.. L __._.7,_.. _ . _ _ _ L_ _ __.._..___.H..._._.._ ._l.. .l__.. 1.__ 1 1_ .._}_._._._ .. J, ____- --- -------l.-------- - - - - - -..t --- -- - - l ---- -- j._._. _ 23 .......r- - ._..p._... _... g _ _.. ;. _t.__.._. ..__._,____}.__..}_____.. _.t_.._._._.__ ...___.9.._._,...___., 7..._.. .__ _ p_ ___.t_____._L______.. _t. _ _ _ _.. p_ { _. _ _ _.._.. _ 1 .__... _ 4.. ..... v.. ,.._....p_.. ] 7 ; _ _ }. __._ -.. F_. t.._. ___.__a..... q _ _ .__.._p_. ...t . [. .. _. _. +. _ _ _ p 0 o. u 1 2 3 4 ue ELAPSED TUE FROM TRIP (MI7UTES) D3h?S - P11t #1 STARTL? 1.R70RT SG #1 STARTL? L?_Lv VS. ~.' m ~ FOR TU?3'd'E TRIP FRCM 75% FP FIGL?.E 9.1.9 9 9-15 sp w ,--.h
250 r-i ..f ~ i i s _1 i _ _ _4 {. _l__- ._O -j _...._, l. 4....___. ) L{ r.t t ___ L. m v3 200 _. w t _... - _ _.. _ 4. - - 4. - L t,3 L_. +- i y _j j L_ 7 v j g a 150 _- { 3 _ [.- 1.- m n .s 1 }. _.. cr' I i +._._ A d ,_.__}__.-_ 1. _ -.. _ L_ _ _ _e e 9 p i_ _ ._ _ } .{ $o { _.___ t.o L... _ s 10 0 ! ~ ~ ~~~ ~-~ ~ ~ ' ' ' ' ' ' ~ " ~~ ~~~ { ~~ ~~~ ^~ A ...s.. -~ L.. ..._.} ...i p .. __._. p.. _ j .._ _7 ..a U2 2 }_ j._.._ 4_ 1 m .. _. __ _7,.._.. _y___.._.,_._....p_.= . _ q _____ d

+-------------t------

--.----t-------i-------- ---------------I__, --t--" u 50 .__T'...__. _. w [ __ _. ,._..[__..._ _._.4.._.. .._._. _ :._ _._ t__. a. _ __. . _ _._. _ i __. _. f. _... } __._. . _ _. _..._..__j__ - _ _ _ _.. _ _. _. g. g.. 7_
p..
..-. 4 ... _.4,._. l.. _.... },_. l g. ._._.p. p c-1 2 3 4 %4 t4 ' ELAPSED TU:E FROM TRIP (MIL *UTES) \\ l D3NPS - Unit #1 STAR'~J? FI? ORT CO'. > _=.'.'.t,4' ~. =. D v P v S.- SURIZIR LI?EL VS. TU:I FOR TU?3INE TRIP FROM 75% FP 9-16 FIGU?2 9.1.10 e G %***.e. g.e_, q ^
10.0 SECONDARY PLANT This section will provide a brief sunmary of the major dif ficulties The encountered with the secondary systems during power escalation. secondary systees that will be covered include: 10.1 Turbine-Generator 10.2 Condenser 10.3 Circulating Water Systen 10.4 Feedwater System 10.1 TURBINE-CENERATOR the turbine and generator experienced relatively few During startup, cajor difficultics, but was plagued with numerous diversified prob-High vibration during the second turbine roll Icd to a re-Icos. align =ent of the exciter in August. The first tire the generator was loaded, grounds in the generator exciter bearing and Number 8 Cenerator Bearing were discovered. The bearings were subsequently The turbine over-disassenbled and the grounding problem resolved. tested in September speed trip mechanism did not operate when first The mach-with the trip point approximately 12 RPM above the limit. anism was cleaned, adjusted and inspected. then The Steam Generator 1-1 Turbine Bypass Valves were cycling open, prior to shutdown which caused damage to the turbine by-closed just The strap piping res-pass headers in the High Pressure Condenser. traints were replaced with rigid restraints to prevent excessive piping cove =ent. .he Number 2 Turbine Control Valve position was found to In Nove:bcr This caused two unit shutdowns in which a function be oscillating. generator' ca the turbine EHC Systen and the servo valve for the Nun-The cause was found to be a defec-ber 2 Control Valve were replaced. tive electrical connector which was the position feedback to the EHC. In January,1978, the turbine control rotor of the overspeed trip cechanis: 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 dif ficul-2 ties and the solenoid valve was reolaced. The inboard bearing on the turbine cain oil pu=p was replaced and orifice plugs installed in the
bearing oil supply lines ci the f reat standard.
10.2 CO:? DENSER The condenser has encountered probices with condenser tube leakage. In September, 1977, approximately 28 tubes were plugged near a pene-header warcup drain from the tration where a high pressure stea: main steau header had impinged on the tubes due to i= proper design arrange:ent inside the condenser. A ncv baffle was of the baff ? installed which deflected the high pressure steam away from the condenser. I 10-1 ~ h *ene m i,r__
10.2 CONDENSER (Continued) 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 4 leaking tubes along with eleven adjacent tubes were plugge6. The unit returned to 90% power en February 19, 1978. On February 20, a small condenser tube leak was discovered and the power was reduced to 75%. The unit was shutdown en February 24 and an eddy current inspection of various tubes indicated serious problens existed in numerous condenser tubes. Sixty-seven tubes were plugged. During the unit outage to recove the EPRA's, ficw diffusers were to correct the con-added to the condenser internals in an attempt denser tube failures. While the unit was at 100% of full power on Septe=ber 25, 1978, in-creasing condensate cenductivity indicated condenser tube leakage. When the unit was shutdown (due to a f ailed Reactor Coolant System i Flowmeter), a condenser inspection revealed one tube had developed. 2 a leak. The defective tube was plugged. With the unit at 100% of full power on September 6, 1978, an increase in condensate conductivity was again discovered. When the unit was shutdown (due to def ective Reactor Coolant Pu=p Seals), investigation revealed two leaking tubes which were then plugged. i 1 f The condenser vendor, Ecolaire Condensers Incorporated, investigated l the cause of these tube failures and plans to install extra supports I within the condenser while the unit is shutdown for repaire of Reactor Coolant Fump Seals during October, 1978. 10.3 CIRCULATING UATER SYSIDI The Circulating Water System has experienced probIcos with failures of the liners of the 54 inch discharge valves. In August, 1977, the Kunber 2 Circulating Water Purp Discharge Valve was rebuilt and the rotation reversed to reduce the amount of turbulence. During the plant outage in Septecher, frag =ents of a valve liner were Subsequent investigations showed ,found in the condenser water box. that the Sunber 3 Circulating Water Pump Discharge Valve was da: aged. The valvo liner was replaced and the rotation reversed on Number 3 Circulating Water Funp Discharge Valve. The amount of ti=e these butterfly valves are throttled is new being litited to minimize the damage from the turbulence. The Cooling Tower experienced so=e icing dif ficulties. In December, 1977, several internal fill support concrete beans were broken and others damaged by ice buildup. Ice falling inside the veil also ds= aged sota of the drif t eliminators and conduit was damaged fron ice buildup. A revised operating procedure was provided by the vendor, Research-Cottrell to nini=1:e the icins damage. Operating under the revised procedure has reduced the ice buildup and the damaged bea=s 2* were replaced during the outage to remove the 3PRA's. g 10-2 une ws.n a v e,. awe -e se N-
10.4 FEEGUATER SYSICIS Another area where major prc5; ems have been encountered is with the f eed p u=ps. Main feed Pump 1-2 was taken out of service in Octo,ber because of high vibration and flow rate dif ficulties. The pump was disassembled and a piece of the pu=p i=peller was found to be broken off. The entire impeller was replaced with the spare and the sharp radiusicorners between the i=peller vanes and sideplates were ground The to recove a potential high stress area on the i=peller. out impeller of Main Feed Pu=p 1-3 was also inspected and ground. Both feedpu=ps were returned to service and further Ispeller difficulties have not occurred. In January,1978, it was determined the drain syste= of the Main Feed Pump Turbines was not operating correctly and high turbine exhaust After an ex-casing water levels were causing the turbines to trip. tensive investigation of the drain difficulties during power operation, a modification which recoved the lower source tap loop seal was com-pleted in February. The auxiliary f eed pu=ps have had extensive difficulties in speed control. In July and August, 1977, repeated speed control relay failures rendered the auxiliary feed pu=ps inoperable. On August 10, 1977, a design codification was implecented VSich added a second set of identical speed relays in parallel to reduce the current carried by each relay. This did not totally eliminate the speed control failures and in January,1978, the relaya of the speed circuit were replaced with relays of a larger current estrying capacity. Other design deficiencies were discovered in October when it was observed the Auxiliary Feed Pump 1-2 Turbine Governor Valve would close under certain vibrational conditions, rendcring the Auxiliary Feed Pu=p 1-2 inoperable. A redesigned valve linkage was installed 'in which the force of a spri'ng assured elimination of the vibrational closure. Feedwater chemistry control has encountered several problems during power escalation. The =oisture separator reb 2t.er drain tanks concen-trate silica and sodium. These tanks are les.ted do enstraam of tha condensate demineralizers and in 1977, it becate necessrci te return Number 5 Feedwater Heater Drain to the condenser to 11 cit the silica and sodium concentration in the feedwater. This reduced the effi-
ciency of the unit, therefore, a solenoid air control valve was added to allow Nu=ber 1 Moisture Separator P.eheater Drain Tank to drain directly to the condenser.
4 e 10-3 A.-. _A.-. .a-m a.x M T M r- = ~ v ~ 4
11,0 U'!IT MONITORING - C"EMISTRY A';D FIALTH PHYSICS Unit ccnitoring tests were conducted to verify that the activity levels of the reactor ccolant, feedwater and process fluids are with.in acceptabic limita, that the dose rates.resulting from direct radiation fres sources centained within the station and on the site are within acceptable limits, that proper chemistry control has been achieved for the reactor coolant system and steam generators, and that the area and procese radiation conl-tors are capabic of continuously detecting and recording asso-ciated radiation levels. Initial shield surveys were condue:cd on May 17, 1977, Shield surveys were conducted fre: August 13 to 21, 1977, at 0% power, on September 3, 1977, at 15% of full power, frco November 17 to 18, 1977,,at 40% of full power and froa September 18 to 21, 1978 2 at 100." af full power. Initial site and station radiatica surveys were conducted on May 17, 1977. Site and station' radiation surveys were conducted froa August 19 to 21,1977, at 0% power, on Septe:ber 3,1977, at 15% of full pouer, and f res Novc=ber 17 to 18,1977, at 40% 2 of full power, and frca Septe:ber 18 to 21, 1978 at 100% of full power. Proper water chemistry fer the initial filling of the reactor TP 500.01 coolant systen was verified en August 19, 1976, and was again verificd on August 2S, 1976, followit; the initial filling of the P.CS. Proper water chemistry for the hydratest of the reactor coolant system was verified on Septe:ber 7, 1076, and was again' 0 verified on Septe ber 7, 1976, nrior to exceeding 250 F in the reactor coolant systen. Duritg the power escalatica program, reactor coolant chemistrf was verified to be within specified limit :, on a daily batis; results included in this report are .f rem secpics taken on /.13ust 24, 1977, at 0% power, on October 30, 1977, at 15% of full powar, on Decc ber 8, 19,77, at 40% of full 2 power, on February 23, 1978, at 75% of full power, and on September 21, 19 78 at 100% cf full pcwer. Proper water chemistry for the initial filling of the steam genera-TP 500.02 tors was verified on June 5,197e; following the filling of the steam generators, the stes: generator layup water specifications were verified to be within limits en July 26, 1976. Fec.lwater , cleanup was conducted, and preper feedwater chenistry was verified i on January 9, 1977. During pcwer ascensica, prior to achieving 15% of full power, proper steam generator water che=istry was veri-ficd on October 28, 1977, at a power level ot' 11% of full power. Af ter achieving 15% of full power, proper feedwater, chemistry was verified on a daily basis; this report includes the results of sampics taken on Dece=ber 8, 1977, at 40% of full pcuer, on Febru-ary 23, 1978, at 75% of full power, and on September 21, 1978, at 2 100% of full power. l l 11-1 l I
r-r TP 500.03 The following initial radiochemistry tests were perforced to establish baseline activity levels for future operations: 1. The reactor coolant, BWST water, stes: generators and cenponent cooling water were analy:ed four ti=es during hot functienal testing, frc= Nove=ber 13, 1976 to Decc=- ber 29, 1976. The spent fuci pool, fuel transfer canal and decay heat 2. systc= were analyzed just prior to thc introduction of fuel daily _during fuel move =ent and weekly thereaf ter. 2 The results of the initial analyses conducted on March 4, 1977, and the results of the analyses obtained, following the conpletion of fuel leading, on April 27, 1977, were 2l normal background levels of radioactivity. 3. The reactor coolant, SWST water, steam generat' ors and co=ponent cooling water were analyzed h=2ediately follow-the con-ing the closure of the reactor coolant syste=, at clusion of fuel loading, on April 27, 1977, and at speci-ficd periods thereaf ter during power escalation. The typi-g cal results of the reactor coolant analyses with the unit ( at 100g of full power is shown on Table 11.5-1. Process and area radiation =enitor syste= tests were co=pleted TP 800.01 on July 15, 1977. The tests included calibration of the area radiation =enitors and ion cha=ber area radiation =enitors, and calibration of the gaseous precess radiatien =enitors and liquid process radiction conitors. Additionall/, prcper func-tioning of applicable alar:s and interlocks was verifi,ed. 11.1 SEIELD SUR"EY Shield surveys were cenducted to decignete locations for subse-to ceasure radiation levels at designated loca-quent surveys, tions adjacent to the shield building and secondary shielding and to obtain gan=a and neutron background ra'distien levels for conparison with future =easurtnents of radiation due to activity buildup. All surveys have been ec=pleted up to and including the 100% power level. The shield surveys conducted indicate se=e j radiation levels outside and inside containment are in excess of ' the radiation zones state'd in the Final Safety Analysis Repert (FSAR). This resulted in a 10 C7R 50.59 safety review and the lhaiting of access to sene specific areas during power 2 operation. The existing dose rates do not result in any measura. ble exposure to the public or any excessive exposure to plant per-sonnel as long as access to these areas while at pcVer is li= iced. The areas exceeding the esttnated radiation levels are inside the auxiliary and contain=ent building where access is controlled. The source of the radiation is neutron strea=ing e=anating frem the _l reactor during power operation. The neutron streaming problem is l still under analysis. l l 11-2 e _em l
......g TP S00.03 11.2 SITE A::D STATIC:: ?>.DIATIC:: St3VEY The Site and Station Radiar. ion Surveys are complete up to and 2 in.cluding the 100T of full power level, except for scoe data collection and review. Site, and station radiation surveys are conducted to: 1. Establich normal background radiation levels at the site boundary. 2. Verify that the dose rates resulting from direct radiation frcs sources contained within the station and on the site are at or belev normal background levels. 3. Obta'in gan=a and neutrer, radiation levels within the station at various power icvels for comparison with future measure-ments of radiation due to activity buildup. 4. Establish station radiation zones for controlled entry by station personnel during various phases of reactor operation. 5. Determine the alert and alarm setpoints for the area radia-tion conitors. 2l Baseline data has been established from the various survey measure-ments and alert and alara 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 1007. pcwct): 1. 603' Elcvatien outside the equiptent hatch 2 2. Main strec=line access rocas 3. Outside Containment E=crgency Hatch but within the exterior entrance 4. 603' Elevation inside Contain=ent (east side) All measurc ents except as noted above have been less than the 1cvels established in Chapter 12 of the FSAR for the appropriate Radiation Zones. 11.3 REACTOR C00tX.T SYSTEM CHDtISTRY TP 500.01 The Reactor Coolant Systen chemistry test procedure was i=plemented to ensure proper water chemistry was maintained during: 11-3
1. Initial RCS filling I 2. RCS Makeup after initial fill 3. Hydro test l 4. Operation at temperatures above 2500F and Hot Functional Testing, and 5. Power ascension l The water chcaistry net the specifications listed in Tables 11.1 thrcugh 11.5 11.4 STEAM GENERATOR CHEMISTRY TEST TP 500.02 The stea= generator chemistry test was imple=ented to establish minimum sa pling frequency and proper water chemistry for the condensate and feedwater systc=s and the steam generator during: 7.2 Feedwater Cleanup, 7.3 Hot Funct ional Testing, and 7.4 Fower 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 SECOND12Y SIDE HYDRO TEST PROCECU?2. The water chemistry act the specifications listed in Tables 11.1 thrcugh 11.4. 11.5 IN11 LAL RADI0 CHEMISTRY TEST TP 500.03 The initial radiochemistry test was run to: 1. Estnblish baseline activity levels during Hot Functional Testing. 2. To tenitor the activity buildup in various plcnt syste=s during initial fuel loading, reactor startup and power opera-tion so that rapid determination of failed fuel and primary to secondary leakage is possible. 3. To conitor the radionuclide leakage frc: 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 fa=iliarize 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 verifi+ M, the background or MOA (Minicu= Detectable Activities) were catablished, and radioactivity was conitored in the RCS. 11-4
11.6 PROCESS AREA RADIATION MONITCRING SYSTEM TEST TP 360.01 This test was perforced to demonstrate the ability of the Process and Area Radiation Monitoring Systen to continuously detect and record the radiation in the station effluents and protect station perconnel f roa exposure to excessive radiation levels. The area conitors were calibrated per IC 2005.02 and IC 2005.01 (calibration procedures) and signed off on separated acceptance sheets for each monitor. The process instrecents were calibrated and signed off on their respective acceptance sheets. I NOTE: The alaras and interlocks and systpn performance require-I ments of clapters 11 and.12 of the FSAR were proven by completion of the following TP's: l TP 160.02, "CTMT and Penetration Room Purge" - RE 5052 A,3,C l TP 170.05, "CTRM Heating, Ventilating and Air Conditioning" - RE 2024 A,3 C and RE 2025 A,3,C. l TP 170.01, " Aux Bldg Radiation Ventilation System" - RE 5403 A.B.C, RE 5405 A,3,C, RE 8446, and RE S447 TP 190.02, " Contact Data Logger Input Verificatien - Radiation Inputs" TP 230.01, " Clean Liquid Radwaste Systen" - RE 1770 A,3 TP 231.01, " Misc Liquid Radwaste Systes" - RE 1878 A,3 TP 232.01, " Caseous Radvaste Pre-Op Test" - RE 1822 A,3 TP 240.01, " Component Ccoling ilater System" - RE 1412, RE 1413 O e 9 11-5
TABLE 11.3-1 i INITLIAL FILL VATER OUALITY FOP. RCS Analysis Specification Typical Sa:ple Suspended Solids 0.1 ppm max. 0.04 ppo Chlorides as Cl 0.1 pp= max. 0.00 ppm Fluorides 0.1 ppm max. <0.01 pps Conductivity 1.0 umho/cm cax
0.92 umho/cm pit at 770F 6.5 - 7.5

6.2

NOTE: Due to CO2 absorption specification pH may be lowered to 5.8 and conducitvity increased to 2.5 unho/cs.
TABLE 11.3-2 RCS WATER AFTER INITIAL FILL (A'C4IENT TE!!?ERATL*RE)* ** Analysis Specification Typical Samole 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. (0.01 ppa !!ydcazinc 0.1 - 1.0 ppm 0.33 ppm. Tstal DJssolved Gas 100 sto cc/kg !! 0 max
13 std ec/kg E 0 2
2 pit at 770F 6.0 - 8.0 ** 7.8
NOTE:
Required when a gas overpressure is caintained en the RCS.
NOTE: May be higher due to hydrazine.
1tay be lower due to CO2 absorption.
NOTE: See Table 11.3-4 for higher teeparature.
e 11-6 o w.=w =
TABLE 11.3-3 RCS WATER OUALITY FOR HYDROTEST Analysis Specification Typical L Chlorides as Cl-0.1 ppo max. 0.07 ppe Fluo. rides as F-0'.1 ppa max. < 0.01 ppm Hydrazine 0.1 - 1.0 ppm 0.1 ppa Total Dissolved Gas 100 std cc/kg H O max
30 std cc/kg H2O 2
pH at 770F 6.0 - 8.0 ** 10.2 Lithium 0.2 - 2.0 ppm 1.1 ppm
NOTE: Ecquired when a gas over pressure is =aintained on the RCS
NOTE: Kay be higher due to hydrazine and its decemposition products (ammonia). May be much higher due to Lithium addition.
TABLE 11.3-4 RCS*** WATER OUALITY ABO'w'E 250 F $ HOT FUNCTIONALS ' Analysis Specification Typical Chlorides as Cl-0.1 ppm tax. 0.00 ppa Fluerides as F-0.1 ppm max. < 0. 01 ppm Hyd ra zine 0.1 - 1.0 ppn
0.1 ppa Total Dissolved Gas 100 std cc/kg H O max.
30 std cc/kg H O 2 2 pH at 770F
9. 5 - 10. 5"
10.1 Lithium as Li7 0.2 - 2.0 ppm 1.0 ppo Discolved Oxygen as 0 0.1 ppm max.
0.01 pps. 2 0
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) may be 7.5 to 8.5
NOTE: Pressurizer is considered a part of the RCS.
If Pressurizer temp A 250 F, specs apply to Pressurizer regardless of RCS teep. 11-7
l L TABLE 11.3-5 I P A WATER OUALITY AT OPERATING CO::DITICSS Dt' RING FO'.iER ASCESSIO:t Analysis' Specification Typical Chlorides as C1~ 0.1 ppm max. 0.03 ppm Fluorides as F-0.1 ppm max. <0.01 ppm liydrazine 0.1 - 1.0 ppm
O ppm Total Dissolved Gas 100 std cc/kg 110 max 31.6.td cc/kg !! 0 2
2 15 - 40 std cc/kg H O 31.2 std cc/kg H2O 2 Dissolved llydrogen as 112 4.8 - 8.5 5.7 pH at 770F, 7 0.2 - 2.0 ppm 0.45 ppm Lithium as Li Dissolved Oxygen as 02 0.1 ppm max. '.0.01 ppm Boric Acid 100 - 13,000 ppa 1093 ppm Boron of l 6245 ppm H3B03
NOTE:
Not' required at 4. 200 F if Cl and F- <.1 ppm each and not required if 02 < 0.1 ppm regardless of Cl-and F-concentration. TABLE 11.3-6 RCS WATER OUAI.ITI AT OPERATING COSDITIONS 1 ITil CIT AT 1007. OF ' FL'LL pot lER Analysi s Sp_eci fica tion Typical Chlorides as Cl~ 0.1 ppa max. 0,03 ppm Fluorides as F-0.1 ppm max. < 0. 01 p pm flydrazinc 0.1 - 1.0 ppm
O ppm 3
Total Dissolved Gas 100 std cc/kg H O max 45.0 std ec/kg H O 2 2 Dissolved Ilydrogen as 112 15 - 40 std cc/kg H O 26.8 std cc/kg H;0 2 pH at 770F 4.8 - 8.5 6.2 Lithium as.17 0.2 - 2.0 ppm 1.2 ppa Dissolved Oxygen as 02 0.1 ppm max. 0.01 ppm Boric Acid 100 - 13,000 ppa 1086 pp:a Baron or 6205 ppm H B03 3 0
NOTE: Not required at ( 200 F if Cl and F-(.1 ppm each and not required if 02 < 0.1 ppm regardless of Cl-and F-concentration.
11-8
DBNPS STARTUP REPORT
TABLE 11.4-1 FEED'JATER CYCLE CLEANUP TEST ACCEPTANCE CRITERIA TYPICAL Iren (Fe) 0.1 ppm cax.
0.05 ppa - Cation Conductivity 1.0 pnho/co. cax. 0.5 paho/cs. DBNPS STARTUP REPORT TABLE 11.4-2 IIOT FUNCTIONAL FEED'JATER TESTING TEST ACCEPTANCE CRITERIA TYPICAL Cation Conductivity 0.5 n=ho/cm. cax. 0.4 unho/ca. liydra zine 0.1 ppa =in. 0.339 ppo 0,xygen (02)
7 ppb =ax.
5 ppb Silica (sic 2) 20 ppb max. 2 ppo Iron (Fe) 0.1 ppm max. 0.01 ppb
Copper (Cu) 2 ppb cax.
O ppa
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.
11-9
l DENPS STARTUP K'iPORT l TABLE 1,1.4-3 Operation at less than 15% Full Power Steam Generator i TEST ACCEPTANCE CRITERIA TYPICAL Chlorides 1.0 ppm max. 0.07 ppa Sodium 2.0 ppm max. 0.01 ppa 0.4 paho/cm. Cation Conductivity 10.0 unho/cm. cax. Silica (510 ) 2.0 ppa max. 0.01 ppm 2 pH at 77F Set by feedwater pH 9.4 DBNPS STARTL*P REPORT TABLE 11./.-4_ Operation greater than 15% Full Power , Fee dwa te r TEST ACCEPTANCE CRITERIA TYPICAL Cation Conductivity 0.5 naho/cm. max. 0.2 paho/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 i 1 I G 11-10 L-
TA3LE 11.4-5 Operatica at 1007. of Full Power , Fee d.a t er 2. TEST ACCE?T/2;CE CRITERJJ TYPICAL Cation Conductivity 0.5 n ho/cs. max. 0.2 paho/ca. Silica (SiO2) 20 ppb =ax. 5 ppb pl! at 77F 9.3 - 9.5 9.4 liydra zine 20 - 100 ppb 28 ppb Iron 10 ppb 10 ppb 0xygen 7 ppb 5 ppb Copper 2 ppb 0 ppb Lead Not detectable Not detectabic he 4 e e 6 11-11
, ih 4 _Al-C
'wE,h '- h
TABLE 11.5-1 Typical RCS Radiochemistry At 100" of Full Pm.er SUSPENDED uCi/mi DISSOLVED uC1/n1 ANALYSIS Gross Beta (includes 1.3E-1 suspended) 1.9E-1 H-3 Dose Equivalent I-131 1.1E-3 4.7E-4 I-131 I-133 1.7E-3 1.4E-3 I-135 2.1E-3 CS-133 1.0E-4 KR-8SM 2.1E-4 KR-87 KR-88 9.0E-5 1.0E-3 XE-133 6.8E-4 XE-135 4.1E-4 XE-135M 1.0E-3 3 AR-41 6.3E-2 F-18 NA-24 ~ 5.SE-3 1.4E-4 CR-51 2.1E-4 1.3Ef5 MN-54 2.1E-4 MN-56 2.8E-3 2.2E-5 FE-59 3.1E Co-53 2.1E-4 1.2E-5 Co-60 9,1E-5 NI-65 3.3E-6 ZR-95 1.3E-5 ZR-97 3.7E-6 NE-95 3.0E-5 NE-97 3.8E-6 TC-99M 1.2E-4 4.2E-3 W-187 11-12 N'* _m e www _ rse w o* =
~, 12.0 UNSCHEDULED UNIT TRIPS Daring the power escalation phase of the testing program, a number of unscheduled reactor / turbine trips occurred. Infor-mA tion from these trips has been used to improve plant perfor-mance by identifying tuning requirements in the Integrated Control System, and by demonstrating system deficiencies for - which corrective actions have been initiated, j The following paragraphs briefly describe the unscheduled trips 2 which occurred since initial criticality (August 12, 1977). l This sumnary is intended to present the conditions' surrounding j 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 Fecdwater Rupture Control System (SFRCS) tripped on dif ferential pressure between steam and feedwater, Icading to a reactor trip on low RCS pressure. The excessive bicwdown of the main steam safety relief valves contributed to the reactor trip en' low pressure. All the relief valves were reset by use of a hydroset on September 16, 1977. j 9/24/77 With the turbine off line and the reactor at approxi-j mately 8% pcuar, a " half-trip" of the SFRCS caused'the-l startup feedwater contrcl valves to close. Reactor Coolant System (RCS) pressure increased and lifted the power relief valve on the pressurizer. After several cycles, this valve stuck open, blowing the rupture disc on the quench tank and causing a partiai depressurization l 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 secam generator 1-2. The i 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 0 a result of the addition of 70 F auxiliary feedwater to-the steam generators, and due to lifting of the pres-suri:cr power relief valve. 11/29/77 The unit was operating at 40% vith the Reacter Protec-tion System (RPS) overpower trips set at 50%. A faulty patch board was inserted into the startup test panel, producing a unit load de=and 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 e
transformer to the startup transfor=crs was defeated, i resulting in a plant loss of AC ppwer. Auxiliary feed-water initiated natural circulation flow through the reactor, and the diesel generators assu=cd the essen-l 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 approxi=ately 11% when the startup feedwater control valves began to oscillate. 1 These valve position swings resulted in overfceding of steam generator 1-1. The reactor tripped on low RCS Additional tuning of the ICS was 7erfor=ed to pressure. ninimize these valve oscillations during startups. l 12/30/77 Following nine consecutive days of steady-state power operations at 72% power, #1 main feed pu=p tripped on " indicated" high exhaust casing water level. An Inte-grated Control Systen (ICS) runback was initiated, but the pressurizer pcwer relief valve lif ted resulting in a reactor trip on low RCS pressure. The response of the main feed pu=p speed controls was codified, using the data collected during this trip. 1/6/78 Two SFRCS trips occurred during startup cperations. Both vere caused by feedwater flow fluctuations which caused feedwater/ steam outlet pressure differential to exceed the limit. Following the second trip, Auxi-liary Feedwater Pu=p 1-1 was declared inoperable because the speed centrol circuitry =alfunctioned. A circuit modification was co=pleted and tested to correct this problem. 1/21/78 To check out the main feedpunp speed control changes made as a result of the 12/30/77 trip, a #1 feed punp trip test ; rom 70% power was conducted. For approximately one:=inute the runback went s=oothly. Then the running pu=p tripped on high exhaust casing level. The reactor and turbine were tripped =anually, and the plant was controlled with auxiliary feedwater during the cooldown to 532 F. 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 conthly surveillance test. The conthly surveillance test has been codified to reduce the likelihood of a recurrence of this probica. 12-2 O
L I 2/24/73 A failed RCS ficw transmitter had placed RPS Channel 3 into a tripped condition. Ar. erroneous RCS high terpera-ture signal to Channel 2 of the P.PS tripped the unit off-line. Both proble s were investigated and corrected prior to resuming pcVer operations. 3/l/78 The reactor was at 49% pcwer. The levcl control valve to deserater 1-2 failed closed. The = sin feed pu=p ran out of feedwater which initiated an SFRCS trip on feed-water /stean pressure differential. The loss of feed-water and closing of the =ain steam isolatica valves increased RCS pressure which tripped the reactor en RPS high pressure. 3/29/78 An abrupt change in the cetpoint of the Tave te=perature controller by an operator placed the plant into a tran-sient condition. Return of the controller to its original setpoint produced a diiection error in the Control Rod Drive (CRD) Centrol Systc, tc porarily transferring tiaa CRD contrcl str. tion to "?'*C:UAL", a condition in which the CRDs would not respond to ICS signal dc: ands. The unstable plant condition coupled with the inability of the CRDs to respond co neutron error demands created a mismatch betweca reactor power and feeduater, resulting in overteeding the steam generators and tripping the e reactor on low RCS pressure. 4/2/78 A turbine trip test was perforced at 75% power to evaluate piping =cdifications cade on the extraction steam lines to the deserator. The feedwater flow exceeded i the feedwater dcrand during the runback, resulting in over-I feeding the stec= generaters. This coupled with lifting of the pressuricer pcVer relief valve caused a reactor trip on Icw RCS pressure. I 4/5/78 Uhile operating at 100" FP for the first time, B&U re-quested an 12:ediate reduction in power and a change to [ 3 RC pump operaticn while a complete analysis of the LBPRA problen was conducted. The unit was reduced in power to 65% and RCP l-1 was manually tripped. Feedwater 2 demands were not prcperly raticed and the feedwater valve { d P crror signal in the ICS affected the cain feedwater pump speed to such a degree that the feedvater system j reached an uncentrollable oscillaticn, and the RPS ( tripped the reacter on low RCS pressure. Since that l time, FCR 78-200 has been approved and i=plemented to de-tune the DP crror signal during two MFP operat.ons, and adjust =ents have been cade to properly ratio feed-water after an RCP trip. f . : ~4 12-3 j , M
4/29/78 While the shutdown for the. screen outage was in progress, the unit experienced a reactor trip f res approxi=ately 20% FP. This was the first shutdown atte=pted with only three reactor coolant punps in operation. As #2 s' team generator approached " low-levci li=it", the operator used =anual control of the = sin feed pump to maintain 45 psid across the =ain feedwater control valves. This resulted in overfeeding the steam generator, and although operator action was taken to stabili:e the situation, a rapid cool-down took place, tripping the RPS on low RCS pressure, and initiating high pressure injecticn for approx 1=ately 5 cinutes. The Reactor Coolant Systan was returned to 2155 psig/530CF and a normal controlled cooldown to Mode 5 was perfor=ed. 8/2/78 In preparation for 40% reactor physics testing, the six second rod insertica step for differential rod worth j measurement was attenpted. The rod movenent resulted in a Reactor Ceolant Systen upset. The positive temperature ccefficient 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. 9/10/73 Conducted optional turbine trip test from 75% FP per l TP 800.14. Excessive feedvater flow resulted in reactor 2 trip on low pressure. 9/28/78 While at 90% FP, the loop 2 RCS flow transmitter FT RCIA1 failed low. This low flow signal caused a trip of RPS Channel 1 and initiated an ICS runback at 20% per minute. The runback stopped at 7C0 MWe and resulted'in feedvater to the sten: generater ratioed as if the l crronecusly indicated fl:V cendition ac tually cxisted. l The operator took canual control of loop 2 nain f eedwater control valve, attempting to =aintain 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 belew the 1985 psig RPS trip setpoint. The plant was placed in Hot Standby (Mode 3) and the RCS flow trans=itter was repaired. 10/3/73 While operating at 73% FP, the second EEC pu=p was started + to investigate tha recent reduccica in EEC header pres-sure. A hydraulic perturbation was introduced, tripping the turbine en icw EEC pressure. The ICS initiated a reactor power runback at 20%/=inute. The increased steam generator pressure and the ICS " cross-limits" rapidly increased feedvater ficw, overcooling the RCS and caus-ing an RPS reactor trip on lev RCS pressure 84 seconds after the turbine trip. The analysis of this trip resulted in a rece== ended =odification to the ICS cross-limits, reducing the c=ount of feedwater added following any turbine trip. 12-4
CORE PERFORLU:CE FOLLOWING BURNABLE P'0ISON ROD ASSC3LY 13.0 AND ORIFICE ROD ASSEMBLY (ORA) REY.O*/AL 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 reds (MORAs). (The MORAs have A latch-only four pins instead of the normal sixteen pins). ing mechanism was then placed over each MORA to hold the MORA to its fuel assembly. Due to the removal of the 3PRAs, eight fuel assemblies had to (2) be,chuf fled to minimise powcr peaking in the core. The core cap for modified core 1 cycle 1 is shown on Figure 13.0-1. 13.1 CORE PERFORMANCE DURI';G ZERO POWER TESTING Following the removal of the BPRAs and ORAs, a zero power test program was conducted to (1) confirm the nucicar design characteristics of the corc; (2) validate assumptions used in the safety analysis, and (3) validate analytical models used for predicting plant responses. Measurements were made to determine the shutdcen cargin and the =cro (2) "all-reds-out" RCS boron concen-power (1) moderator coefficient, (3) control red worths, and (4) differential boron worth.
tratica, The ejected red worth was also reasured during =cro pcuer testing.
All testing yielded satisf actory ruhults which ensured initial opera-tion of the nodified core was within the limits specified by the Davis-Besse Unit 1 Technical Specificaticns. The subsecuent sub-sections of this section sucmarize the results of various tests per-f ormed at acro power in accordance with Post Refueling Physics Test-ing, ST 5010.03. 13.1.1 SENSIBLE HEAT CHECK 2 An upper power limit of 2 x 10-8 amps was put on scro pcwcr testing to assure no nuclear heat (sensibic heat) would be [ added to the RCS which would af fect the testing. Prior to I scro power testing, a sensible heat check was run to verify no sensibic heat produced below the upper power limit. The point of adding sensible heat was checked as described in Section 6.2. During this check, sensibie heat was observed at approximately 1 x 10-7 amps (which is well above the pcwer li=it). At this pcwcr, the turbine bypass valves started to open and the makeup valvo started to close. 13.1.2 REACTDIETER RESFONSE CHEC'<0U~' i Prior to using the reactimeter for reactivity =casurements, the response of the reactc to a change in reactivity was ecmpared to the design response. The purpcse of this check-out was'to verify the delayed neutron constants used by the reacti=eter gave an accurate representatica of the core. 13-1
The checkout was accomplished as described in Section 6.3. A plot of the reactivity inserted versus the doubling time These plots was obtained and compared to the design value. are shown on Figure 13.1-1. Reactivity insertions of approximately +25, -25, +75, and -75 pcm were perf ormed for the checkout of the reactimeter. All measure-Tests results are su=marized in Table 13.1-1. ments for the checkout were within the + 5% of design values (acceptance criteria) 13.1.3 ALL RODS OLTI BORON CONCENTRATION hot zero power (HZP), at 86 EFPD critical The all rod out, boron 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 Red Assembly (CRA) Group 7 controlling at 78% withdrawn (wd) and all other rods ful y withdrawn (except CRA Group 8 at 37.3% vd), a boron end point measurement (ARC) baron was perfor=ed to determine the all rods out The measured boren concentration was 1615 concentration. The neasured excess reactivity worth of the inserted Group 7 rods was 122 pen (percent milli /,1 pcm = 10-5 ppm. j Using the dif f erential boron worth of 10.06 pc=/ 4 k/k). ppm from the B5W Physics Kanual, an ARO critical boron con-2 centration of 1627 ppm 3 was obtained. This is within the l acceptance criteria of 1661 + 100 ppm with CRA Group 8 at ~ 37.5% vd. 13.1.4' TEMPERATURE COEFFICIENI 0F REACTIVITY See Section 6.5 for a discussion of temperaterc coef ficient and a description of the test method which is the same as used here. results for the three rod configurations at Measurc=ent which temperature coefficients were measured is se=marized in Table 13.1.4-1. All moderator temperature coefficients measured-satisif ed the Davis-3 esse Unit 1 Technical Speci-fication limit of less positive than.9 x 10-4 3 k/h/0F (9 pcm/cF). Also, all temperature coefficients measured met the acceptance criteria of being within + 40 pcm/oF of the predicted values. 13.1.5 CONTROL RCD REACTIVITY MEASURCENTS During zero power testing at 530 F, measurements were made 0 j to determine the CRA grouo reactivity worths for Groups 4. i The reactivity worth of the centrol rod groues 5, 6 and 7. were calculated utilizing the reacttneter. 9 13-2 --.r
The " boron swap" technique, which is described in Section 6.6, was used to determine dif f erential cnd integral worths for the control rod groups. Measured worths are tabulated and cc= pared with predicted values in Tchle 13.1.5-1. Integral red worth curves for N the regulating rods obtained frca these =casarc=ents are shown in Figure 13.1.5-1. 13.1.6 EJECTED CO; TROL ROD WORTH The ejected rod worth test was perforced to ceasure the reactivity werth of the single predicted worse case ejected rod. With CRA Group 5 at the Technical Specification insertion licit of 55% ud, CRA 5 of Group 6 (6-5) was pre-dicted to have the highest worth. Adjustments were made for measurement uncertainties and cc pared with the acceptance criteria of.71 - 1.00% o 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 measure =ent was initiated from the following rod config-uration: CRA Groups 1-4 at 100% vd 2 CRA Group 5 at 57% vd CRA Groups 6-7 at 0% ud CRA Group 8 at 37.5% ud CRA 6-5 was borated out of the core and then, inserted back into the core utilizing a red swap with CRA Group 5. The worth of CRA Group 5 inserted during the swap was obtained from the rod worth measurecents perfor:ed earlier in the testing program. The measured ejected rod worth was then multiplied by 1.05 to acccunt for measurement uncertain-ties. This resulted in an ejected rod worth of.595%4 k/k which corresponds to a certain a=ount of reactivity inserted. The ceasured integral rod worth curves, however, indicated a somewhat dif f ercnt shape from the predicted shape. This resulted in considerably less reactivity being ) inserted by the rods in the red insertica limit than used in the predictions. The effect of having less reacitity inserted is to reduce the neasured ejected rod worth. For this reason, even through the worth we =easured was more than 207. less than the predicted worth, it was considered acceptable. 13.1.7 SOLUBLE POISO:I UORTH MEASURECT Data for the differential boron worth =easure=ent was taken at the end points of two red vorth =easurements. CRA posi-tions were: ) 1 13-3 .o -e. ..a. 7 aw rr
9 1st Measurement 2nd Measurement CRA Groups 1-6 G 100% wd CRA Groups 1-3 G.00P. ud CRA Group 7 G 78% vd CRA Group 4 0 l'.a vd CRA Croup 8 0 37.5% ud CRA Group 5-7 0 0% ud Boron Conc 0 1615 ppm 3 CRA Group 8 0 37.5% wd Boron Cone 0 1156 ppnB From the rod worth measurc=ents the total worth of CRAs fren 10% vd on Group 4 to 73% ud on Group 7 was 5.10% A k/k. This reactivity change corresponds to a change in boron cen-centration of 459 ppa (1615 - 1156). Therefore, the differ-ential boron worth is: 5.10% 4 k/k 459 pp=B = 1.11 x 10-2% Ak/k/ppmB' The predicted dif ferential boren worth from the BS*' Physics Test Manual is 1.006 x 10-2% d k/t/ppmB. The =casured value is diff erent from the predicted by -9.46%, so the acceptance criteria of + 10% was met. 13.1;8 SHUTD0h':1 &\\RGI!! CALCUIATIO:i Technical Specification 4.1.1.1.1.d requires: The shutdown margin should be determined to be A l% 4 k/k prior to initial operation above 5% Rated Thermal Pcwer af ter cach fuel loading by consideration of: 2 (1) RCS boron concentration, (2) Control red positica, (3) RCS average temperature, (4) Fuel burnup based on gross ther=al energy generation, (5) Xenon concentration, and (6) Samarium concentration, with the regulating red groups at the maxi um insertion IL it of Technical Specificatica 3.1.3.6. (The insertion li=it for the modified core is 55% wd on CRA Group 5). Shutdown margin is defined as the instantanecas a=ount of reactivity by which the reactor is suberitical or would be subcritical frcs its present condition asse=ing no change in APSR position and all control rods in except for the maxt=un worth rod which is stuck cut. Considering the definition of shutdown margin and the required surveillance testing, the following =ay be stated: 13-4
The worth of Groups 1-4 at 100% vd and Group 5 at the inser-tien limit (55'; vd) is 5.01*. .d k/k. (The worth of-Group 5 l The worth of Grcups 1-4 was predicted by B&W vas ceasured. l and verified by ceasurements of Grcup 4). If the stuck rod l worth of 1.01'; is subtracted f re= this, 4.0% d h/k remains. (The stuck red is also calculated and verified by the rod 2 ceasurc=ents made). All other conditions affecting shatdeva cargin match those of the predicted or =easured data. Therefore, the minimus shut-down cargin is adequately satisfied. I r 4 e 13-5 ~ - - - - ~ - - - - - - - - - - -
COPI PERF0PX\\';CE DII3I ;0 POWER ESCAL\\TIO:i TESTI:G 13.2 Following the completica of the Post Refueling Physics Test, ST'5010.03, a pregran of power escalation was undertaken asCore per the Power Escalation Sequence Procedure, TP OS00.00. performance testing was conducted at three major power plateaus The following sections give more of 40%, 75%, and 100!. full power. to the perfornance of the physics tests conducted details relevent during the escalation phase of testing. 13.2.1 NUCLEAR 1:!STRUME;;TATIO:: CALIS3ATIO:I AT POWER, TP 0300.02 Core alternations described in Section 13.0 did not involve any alteration of the out-of-core nuclear in-strumentat ion. The Nuclear Instrunentation Calibration at Power Test, TP 0800.02, was conducted at varicus power levels as prescribed in the Pcwer Escalatica Sequence Procedure, TP 0800.00. The bases for the acceptance criteria are included in Section 7.1 13.2.2 REACTIVITY COEFFICIE::TS AT POWER, TP 0300.05 The Doppler and modcrator coefficients of reactivity were deter =ined at the 90% pcwer plateau by TP 0500.05, Reactivity Coefficients at Power. The method for deter-mining these coef ficients is described in Sectica 7.2. 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-ficicnt co=puted, but the power Doppler coefficieg.: vas limited to a maxinun positive value of -3.7 x 10-3
$ h/k/% FP.
The computed value of the power Doppler coef ficient was determined to be -9.8 x 10-5 6 k/k/% FP. The computed moderator tenperature at the 90P. power plateau was -2.05 x 10-5 6 k/k/0F. 13.2.3 ROD WORTil !EASUREME';TS, TP 0800.20 Differential rod worth measurements were perforced at the ^ 40%, 75%, and 100% pcwor platcaus using the six second insert / withdraw nethed described in Section 7.3. The change in reactivity was primarily due to rod totien, although the rnalysis did correct for reactivity effects of fuel temperature changes. The resulting differential rod worths are sumuariced in Table 13.2.3-1. 1 ~~ 13-6
13.2.4 CORE POWER DISTRIBUTION TEST, TP 0300,11 Core power distribution data, describcd 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. I 13.2.5 PSEUDO CONTROL ROD EJECTION TEST, TP 0S00.28-The Pseudo. Control Rod Ejection Test, TP 0800.28, was per- ~ formed at the 40% power plateau to verify that the worth. of the most reactive control red frem its nominal full power position to its fully withdrawn position did not exceed 0.C57;$ k/k. Design calculations d,etermined the control rod in core pasition H-14 to be the most reactive rod following EPRA and ORA re= oval. The ejected worth of control rod H-14 i ( from its nominal full power position was measured during testing. Listed belo:c is: the sequence of major events that occurred during the tesc: I 1. Group 5 control rods were positioned to their full out limit of 100% withdrawn and Group 6 control rods were 2 posicioned about 90% withdrawn. 2. TP 0300.20, Rod Ucrth Measure ent, was perforced to determine the diff erential worth of Group 6 before the rod swap. 3. Control rod 7-3 (core location H-14) was swapped' vith Group 6 rods'until H-14 was to its 100%. withdrawn position. 4. TP OS00.20 was perfor=ed to determine the differential worth cf Group 6 after the rod swap.1 5. The ejected rod worth of E-14 was calculated using the average of the two differential rod worth measure =ents and the rod travel of Group 6. The measured ejected red worth of H-14 was 0.18% [L k/k which i is below the 0.65%.i k/k 1 Lait used in the safety analysis. 13.2.6 INCORE DEIECTOR TEST, TP 0800.24 s The incore instru=entation system, as described in Section 7.7, was tested at the 40% and 75% power plateaus as dir-ected by the Pcwer Escalation Sequence Procedure, TP 0800.00. r F 9 5 e 13-7 y - ~, -, - -.,._
Data for the incere scif-pewered neutron detectors was obtained fren the ccuputer. ' Ratios of cc puter corrected detectcr readings to average string readings were calcu-lated and clotted against axial level. Syrretric detectors were compnred for consistency while non-syntetric detectors were checked for reasonableness. After analyzing the results, the incore detector systen was determined to be functioning satisfactorily. 13.2.7 PCliER D:B AI.A' CE DETECTOR CORRELATION TEST, TP Oc00.18 The Pcwer I balance Detectcr Correlatien Test, TP 0300.18, was perf e rred 'at the 40% pcuer plateau in accordance with the Power Escalatien sequence Precedure,, TP 0300.00. 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 assenptiens of the safety cnalysis. 2. To deternine and set the proper gains for the power range nuc' car instrutents scaled diff erence amplifiers in order us obtain the desired relation in the previous objective. 3. To determine the relationship between the calcularad offset f rca the Backup Incore Detector Systea (3 IDS), and the calculated offset f rca the full incore detec-tor system. 4. .t( <mine the core taximun linear heat rste (MLER) and ai:ua departure fron nucleate boiling ratio (DSSR) tne various values of core of f set as outlined in 2 l erion 7.8. The tect, as outlined in Section 7.8, was perf ormed and the values of '.UIR cnd minite DNBR were cbtained for the pre-scribed values of ef f set. The criteria in cbjectives 1 thrxqh 3 wer e not cet due to the fact that the cut-of-core nuclear instrunents scaled difference a:plifiers were not reset prior to condt;*.ing the t e s :.. The 3ackup Incore a Detector Systr: critier. centiened in objective 3, also did not cect the acceptance criteria. In lieu of rerunning the test in its entirety, the following odifications to the test were develcped: 1. -1: vas decided that the BIOS criteria not be run at 40: power and a test deficiency written to ifcit the use of the BIDS until their acceptance. The test will be per-formed again at 75% power follewing a 7 to 30 day operat-ing period at 103?. power, at which time the SIDS will be tested to verify that the criteria in objective 3 vill be met. ~
2. It was decided that a " ini" test be perforced to check the criteria in objectives 1 and 2. An cutline of the "nini" tcst is sho*a below: a. At the 40." pcwer plateau, the axial power shaping - rods were positiened to obtain approximate power kn-balances of +5.6" FP and -9.6% FP. i b. Ccaditions were alleved to stabilize at each power inbalance for a minimum of 15 minutes and then the following were obtained: - Core power distributica data including incore = offset values. 2 - Out-of-core offset values. The " mini" test was perferned but the gain values calculated l for the nuclear instru=ents' scaled difference anplifiers l ( were found to be eco lev. Therefore, the gain valaes were recalculated and the "nini" test run for a second eine. This ti=e the slopes providing the relationship between 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 cenputed slopes for each out-of-core de:ector were found c) be between 1.23 and 1.32 at 40". FP. This satisifed the acceptance criterion that the slope be greater than 1.25 in -each case. A further require =ent that each point he within A 3.5% offset of the line was also satisfied. The values oT out-of-core and in-core offset are su==ari:ed in Table 13.2.7-1. 4 O e O 13. 9 9 _ _ _ _ _ _ _ - _ _ _ _ _ _. _ _. _ _ _ _ = _ _ _ _ _ _ _ _
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REACTIVITY vs. POSITIVE DOUELI::G TE!E 100 3._..y. ; _, .....,7_.:_.. .._ _ p g g. +. i=_=_.-.. ;-__-- s :. gic#:- i'r i r]: _r-i t f( = ;i 'i-- _E-I-i_ii-~ r - -- :_- i.:) : _.: - __ : _. r-i : r i 2 --
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~_- __: =.:_ ::-- :::..:-. ---. :.= =. v .. _ _. -. _._:---=_-_-.:----_-:===:_. _- 2 _ _ = - - _. l y s U r--- t-- i r- - Design Curve d T e 20 f 6 i i i k b i l t I ,,i t jiii ie t i t 1 I } 40 80 120 160 200 240 DOUBLI:G TDZ (SECO DS) 2 REACTIVITY vs. !;EGATIVE DOUELI:?O TE!E -10 0 -.. _ - _= p
r.t. -. _.. = gr.-.
y _ ;.1= = -p 25 Lesign Curve _: r. g : _=- - : = =. . =;- := : :. 7 -_7 . : :. = : - : --- :: + = =_ ;_= :. 7= ET.= =_.'.r ri. _-
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__-._ -= L.___. _a ^ -._ _IL 6-F g 3 y=+4C===W M =. =11m==h=i.-4E-E========.h'ti=~m . <J-- - - - ._ -= p v---- y,=-..-__-- -__-.- - _ -__-._--__ s __=r=: = _. _ _ _ O a u -20 .i M 1 4 s i l l i i e i t'I l i i t ' l i I i e i l I i I -270 -230 -190 -150 -110 NEGATIVE DOUBLI::G TES (SECO::DS ) DBNPS - Unit #1 STARTUP REPORT Reacticeter Chechout Figure 13.1-1 O - Measured Data 1* 11 w _~
2 1-i i 1-2 i 1 REACTIMETER RESPONSE CHECKOUT i i i REACTIVITY CHECKOUT 2 A REACTIVITY DEVIA-CONTR0f:,ING CRA GROUP DOUBLING DESIRED DESIGN ACTUAL TION 3 REACTIVITY GROU? INITIAL' FINAL TDe 5 NUltBER % wd % wd Seconds PC3 PC j pen. +25 7 77 81 205 +24.2 +23.5 1.2 l 1 1 -25 7 77 72.5 -249 -24.7' -25.2 -2.0 4 2 l +75 7 77 89 55.5 +72.0 +72.5 .7 l -75 7 78 64.5 110' -67.2 -64.7 >+3.7 ii 1 i i i [ 1 4 T i i I i i .DBNPS - Unit #1 f STARTUP REPORT 1 Reactimeter Checkout Table 13.1.2-l' ~ 13, l s ,-,nn ,--c n ,~ ~ haMM - -
i I l i t MODERATOR TC4PERA'IUPI COEFFICIE;T MODERATOR TDfPERATUF2 TDTERATURE ROD CO:7ICURATION COEFFICID;I COEFFICIE;T (pem/F) (acm/F) CRA Groups 1-5 0 100% ud CRA Grcup 6 0 65.5". vd +5.54 +7.54 CRA Group 7 0 0" ud CRA Group 8 0 37.5* ud CRA Groups 1-30100* vd CRA Group 4 0 30*: ud -2.20 .20 CRA Groups 5-7 G- 0% ud CRA Group 8 0 37.5" ud 2 CRA Groups 1-4 0 100" ud CRA Group 5 0 54% ud +2.85 +4.85 CRA Groups 6-7 0 0" ud CRA Group 8 0 37.5" ud DBNPS - Unit.#1 STARTUP REPORT Temperature Coefficient ' Table 13.1.4-1 13-13 W s l~ N aA W - A A -> J6ea es 4 JamA _ M-a w 'V ~~L ' Mu. Aas n. x W ~ ~
l L... TABLE 13.1.5-1. l ( l COMPARISON OF MEASURED AND PREDICTED CONTROL ROD CROUP REACTIVITY fiORTHS AT 530 F Predicted Worth -(1) From Physics. -Deviation CRA Position, % ud Measured Worth Test Manual From (CRA 8 0 37.5". ud) % d k/k % 0 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) 2 Total Rod Worth -5.34 -5.04 +5.95 (3) (1) Deviation Measured - Predicted Predicted From = Predicted (2) Acceptance Criteria =115% (3) Acceptance Criteria + 10% DSNPS'- Unit #1 STARTUP REPORT . Table 13.1.5 13-14
t CRA GROUPS 5, 6, & 7 WORTH WITH CRA GROUP 8 AT 37.5% WD 0 l~ '-l l !I l ' '. l" ".l 0.[.'1 l]'[~},1[" ~ l7'"l~'-['{ ]-l, l l-] l r, l l ~ j ~i j -- 1 ~ i 3 'I a .[2! j _'y I i - 500- ~.l,. 7 L. j ,, j j {.;. -100G. - .f.l. l. [ p[ - l3_:4l Q i.l ! l.4 i.L.'{. 4 .J _8 ? c ( i. .j..z __ } q _L. . /p ..j_.- _ .j..{. j.j..j. l t. h_._._p .~..y d . +... - l . l.
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_L: -i _75 100 I 0 23 50 , L _ _,_b 0 2 50 7$ 100 Group 5 t i 1 I Group 7 0 25 50 75 100 i Group 6 l ROD POSITIO:t (*: WITHDRA't.;) DB:. S - Unit fl ? STARTU? REPORT Figure 13.1.5-1 13-15
TABLE 13.2.3-1 MEASURED DIFFEREITIAL RCD UORTHS AT PQWER Dif f erential Full Rod Groun Position (% Withdravn) Rod Worths Power 1-5 6 7 8 (pem/% vd) 3 40 100 87.5 0 30 9.09 l 40 100 69.7 0 30 9.55 75 100 87.4 0 30 9.17 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 i e e e 13-16'
FIGURE 13.2.4 1 l CORE PC"ER OISTRIBUTION Core Cenditiens Measurc'd Design Measured Design 'nq % ud ino % vd Power Level 3 5 'O ~ Cps 1-4 at 5 at 100 % Vd 100 % vd Borca Cene 13?! p;c '; A PPs 6 at 92
7. vd
~ i3.5 ': vd Core Burnu? 5 9 6 E7?D ES EFPD 7 at U ~ v' O ~ vd Axial 8 at 32 $vd 25.5 % Vd Imbalance -0.02 377 -0.00 grP Radial Peaks 1.10 0.97 1.01 1.25 1.09 1.14 0.65 0.58 0.55 l0.65 1.13 1.02 1.06 1.27 1.15 1.13 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 1.0S 0.50 0.50.l 0.64 1.14 1.04 1.21 1.06 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 0.75 Quadrant Tilt
0.,e 3 4 0.67%
-0.85% +0.53% -0.35% X.XX - MeasuEcd 9 18 Mininun D:GR = X.XX - Design Maxt=un LHR = 5.42 kv/ft 13-17 - - - -. ~, - - - --.m..
FIOURE 13.2.4-2 CORE -FOL*ER DIS""RIEUTIO Core conditions Measured Design Measured Desien Gps 1-4 at 100 % ud 100 % vd Power Level
39. 5 *:
40 5 at 100 % yd 100 % yd Boren Ccac 1301 ??: ';A PF3 6 at 92 % ud 93.5 % vd Core Eurnup 59.o_I??D 53 EFFD 7 at 0 % ud _ u T. vd Axial 8 at 32 % vd 23.a % vd Imbalance -0.02,yP -0.00 %rP Total Peaks 4 .44 1.30 1.32 0.75 0.69 1.37 1.15 1.18 1 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 1.2S 1.14 0.73 0.75 1.39 1.31 1.43 1.29 0.58 0.76 1.33 1.39 1.47 1.26 0.60 2 1.37 1.'63 1.27 1.04 1.40 1.61 1.29 'I .00 l 1.69 1.44 0.74-1.64 1.39 0.71 0*89 Quadrant Tilt
0. c, o.,
+0.67% -0.85% +0.53% -0.35% X. E - Ml suEed 9.18 Minbc M.'3R = 7.. n - Design Maxi =us M = 5.42 _ kv/ft 13-18
FICL*RE 13.2.4-3 CORE PCT. DISTRIE'; TION C_ ore Conditions Measured Design Measured Desien Cps 1-4 at g.,.3 100 % vd Power ' Level 75.1 - 75.0 g 100 SA 5 at 103 % ed 100 % vd Boron Conc 1195 pps PFm 6 at 14 % vd ">.3 % vd Core Burnuo s-3 EF?D 94 EFPD 7 at u % vd v % vd Axial 8 at 32 % ud 42.3 % vd Imbalance -2. 64 ".F? +0.06 %FF Radial Peaks 4 4 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 1.03 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 1.14 1.35 1.08 0.90 2 1.16 1.31 1.03 0.85 i 1.31 1.19 0.63 1.34 1.16 0.62 0.75 Quadrant Tilt 0.75 +0.44% -0.32% +0.23%- -0.35% X.XX - Measur'ed Mini =um DN3R = 4.63 Design Maxi =us LER = 10.35 kv/ft l X.XX 13-19 i. _m.._
FIGURE 13.2.4 -4 CORE POWER DISTRIBUTION Core Conditions Measured Design Measur dd"- Desien 100 % vd 100 % vd Power Level 75.1 g .75.0 ; Ops 1-4 at !!A ppm 5 at 100 % ud 100 % wd Boron Cone 1195 ppa 6 at 92 ,% vd 93.5 % vd Core Burnup 94 3 EFPD 94 EFPD 7 at 0 % ud 0 % vd Axial 8 at 32 % wd 22.3 % ud Imbalance -2.64%FP +0.06 %FP Total Peaks 1.33 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 . 0.63, 0.77 1.39 1.42 1.47 1.26 1.38 1.69 1.29 1.07 2 1.41 1.60 1.30 1.01 l 1 \\ 9 1.73 1.48 0.74 1.61 1.38 0.74 0.90 Quadrant Tilt 0.89 +0.44% -0.12% +0.23% -0.35% X.XX - Measuicd Mini =un D::ER = 4.63 X.XX - Design Maximus LER = 10.d5 kw/ft ,13-20 -~ - - ~ ~ - - - -. ._,n
e. .4-- I FIG'JRE 13.2.4 -5 CORE Foi.'ER DISTRIEUTION Core Conditions -Measured Desien Measured Desien 100 % vd 100 % ud Power Level Co.' 'no Cps 1-4 at pp= _FA pp 5 at ino % ud 100 % vd Eoren Cene s'ro 6 at 90.6 .% ud 93.5 % vd Core Burnup ilh IEFFD 110 ETFD 7 at 0 % wd 0 % wd Axial 8 at 24 % vd 22.3 % wd I= balance A.0 %FP -2.99 %FP Radial Peaks 1.15 1.03 1.04 1.25 1.08 1.13 0.66 0.61 1.15 1.05 1.03 1.25 1.14 1.12 0.67 0.53 0.93 1.12 1.08 1.2S 1.03 1.00 0.67 1.03 1.11 1.12 1.24 1.0S 0.97 0.54 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 2 1.15 1.29 1.07 0.S5 1.29 1.17 0.63 1.31 1.14 0.62 0.75 Quadrant Tilt 0.74 +0.60% -0.17% -0.5S% +0.15% pasnied Minfr a DN3R = 3.2S X. XY. Maxi =t=2 LER = 13.13-kw/ft X..U - Design 13-21 .w.h -m 6,.. + S
FIGU?2 13.2.6-6 CO?l PO*,'IR DISTRIBUTION Core Conditions Measured Desien Measured Design Cps 1-4 at 100 % vd 100 % vd Power Level 99.8 3 100 g 5 at 100 % ud 100 % wd Boron Cone 1100 _p;n NA ppm 6 at 90.A % wd 93.5 % wd Core Burnu> 109.2ETPD 110 EFPD 7 at 0 % vd 0 % wd Axial 8 at 24 % wd 22.3 % vd Imbalance 0.0_ 7.FP -2.99 %FP Total Peaks ~ 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 1.23 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 1.2! 1.44 1.29 0.60 0.76 1.34 1.36 1.43 1.27 0.63 1.34 1,58 1.23 1.06 2 1.37 1.55 '1.28 1.02 l 2 1.59 1.33 0.76 l 1.61 1.39 0.74 l l 1 0.91 @adrant Tilt 0.89 +0.60% -0.17% -0.58% +0.15% X'XX - M.casurcd Mini =t= DN3R = 3.23 X.XX - Design Maxi =ta U A e, _13. O - kv/ft ~ 13-22
TABLE 13.2.7-1 SinDIARY OF AXIAL OFFSET IIEASURD:E';TS l ICO OCO - % Power Backup Incore Of fset i Pcwcr % Pcwcr NI-5l NI-6 l NI-7 l :;I-S % Pcwer 39.21 -1.04 -3.60 -1.84 -3.22 -2.29 Value Not obtained l l 39.20 +13.98 +15.93 +18.44 ' +16.27 +17.66 Value Not Obtained l l 39.40 +15.03 +16.59 +19.21 +17.00 +13.33 Value Not Obtained 1 l+16.47 +19.06 +16.94 +13.34 Value Not obtained 39.25 +15.47 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 -25.75 Value Not Obtained i 39.42 -19.51 -27.49 -25.98 -26.90 -25.76 Value Not obtained l I l l l l where: ICO = P7c7 _ PEc: x 100'; PTop + PBot l r l OCO = chs nel I _ balance x 100% Channel Pcwer 1 t l 9 13-23 D:3 ' nap __ m. m h_ u _ _
m. ~- TABLE 13.2.7-1
SUMMARY
OF AXIAL OF? SET 1EASU?r:E;IS I .ICO OCO - % Power Backup Incore Offset I Pcver j % Power NI-5 l
I-6 [ NI-7 l NI-S
% Power i j 39.21 -1.04 -3.60 -1.84 -3.22 -2.29 Value Not Obtained 39.20 +13.98 +15.93 +18.44 +16.27 +17.66 Value Not obtained 39.40 +15.03 +16.59 +19.21 +17.00 +18.33 Value Not Obtained 39.25 +15.47 +16.47 +19.06 +16.94 +13.34 Value Not 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 whore: ICO = PTeo P2ct x 100% PTop + PEot G00 - Channel I-balance x 100% Channel Pcver em 13-23 END .m WO a w_.- ~~ m
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ML20062C909

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