ML20132E405
| ML20132E405 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/05/1983 |
| From: | Herring T, Phoenix W, Wright L SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13309B581 | List: |
| References | |
| NUDOCS 8510010079 | |
| Download: ML20132E405 (564) | |
Text
{{#Wiki_filter:. .. _. _ _ _ _ . - __ _ .. ___ 4... L gCE Southern California Edison Company SAN ONOFRE
- NUCLEAR GENERATING STATION UNIT 2 W AK:sP R@@K:
to the UNITED STATES JUCLEAR REGULATORY COMMISSIOb
%2220aTiE p"m aq'y FOR THE PERIOD ENDING AUGUST 8,1983 INCLUDING SUPPLEMENTS
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NUCLEAR GENERATNG STATX)N l LNT2 1
t i \, ../ INSTRUCTIONS TO RECIPIENTS OF SUPPLEMENT II 0F SONGS 2 INITIAL STARTUP REPORT Please replace the entire report with Supplement II. There were so few unchanged pages that it was decided to reissue the entire text. Please discard the old text and covers. The new covers slip into the jacket. Current plans are to use the SONGS. 2 Startup Report's 3-rin8 binder to include the SONGS 3 Initial Startup Report. The use of a single binder is expected to be a convenience to the reader. SONGS 3's startup report is expected to be easily accommodated by a single binder, q k i
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O d SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION UNIT TWO STARTUP REPORT TO THE UNITED STATES NUCLEAR REGULATORY COMMISSION LICENSE NUMBER NPF-10 DOCKET NUMBER 50-361 INCLUDING SUPPLEMENTS FOR THE PERIOD ENDING AUGUST 8, 1983
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k.) APPROVAL FOR ISSUE Supplement II of the San Onofre Nuclear Generating Station Unit Two Initial Startup Report is approved for issue by the Test Working Group whose members are listed below: For_Bechtel Power Corporation i % lo 5lS3
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For SCE Engineering _ g, , For SCE Station Engineering wo k ec/oJ/f3 0 For Combustion Engineering [eeldd fo//o/J.3
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For SCE Startup (TWG Chairman) M'd)/$t /6/5/r 3 y c/ Coauthors [,7. h4f,c /4/f/43
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20 $ c. Oldk isNN3 SONGS 2 Power Ascension Supervisor at . 10 C 8) Issue Date October 15, 1983 t'
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O ACKNOWLEDGEMENTS The Initial Startup Report for San Onofre Nuclear Generating Station Unit. Two was edited by A.F. Hays, A.T. Heinle, and W.C. Phoenix under the supervision of D.E. Frey and K.L. Johnson with contributions from the following startup personnel: G.S. Arnold ** C.E. Barbehenn** A.E. Bates ** R.W. Boynton** J.C. Chang ** S. Dolcemascolo** T.S. Earle D.E.-Frey** P.A. Gagne** A.F. Hays **
'~ "
A.T. Heinle** M.A. Herschthal** ** K.B. Holland ** Eg Os E. Holmes 5a J.E. Hughes J.E. Lyle W.A. Marcum* C.N. Pendleton T.E. Peterson** W.C. Phoenix ** J.B. Yee**
- Denotes contributor to Supplement- I.
** Denotes contributor to Supplements I and II.
J.. V
4
SUMMARY
OF GANGES SECOND SUPPLEMENT SECTION PAGE(S) CHANGE Acknowleg- v Update to second supplement. ments Preface -xvii Tense changes. Editorial, reworded three (3) sentences, added sentence. Contents xix-xviii Change page numbers. Deleted Section 6.1.7.4: 100% Turbine Trip. Mded Sections 6.12.4 and 6.16.7. Updated Section 6.1 description. Figures xxv-xxxiv Change page numbers. Md figures. Tables xxxv-xxxvi Gange page numbers. Md tables. 1.1- 1.2 Updated references to second supplement. Tense changes. 1.2 1.3 Editorial, added and deleted words.
- 1.2 1.4 Redrew figure.
'[ 1.3 1.5 Editorial, added words. Tense changes. Changed
" Guideline" to " Guide" twice.-
1.3 1.6 Editorial, changed wording three (3) times. , changed " post core" to "postcore" 1.5 1.8 Deleted " extremely". 1.5 1.13 Mded entry for Section 6.12.4. Incorporated 6.8.13 into 6.8.12
,, 1.5 1.14 Deleted entry for Section 6.17.4: 100% Turbine Trip. Mded entry for Section 6.16.7.. Deleted entry for Section 7.0: Conclusions 1.5 1.17 Deleted sentence.
1.5 1.18 Changed " disassembled" to- " adjusted". 1.5 1~. 20 Mded words to one (1) sentence. Tense changes. 1.5 1.21 Tense changes. Deleted one (1) sentence. 1.5 1.22 Mded test 2PA-298-01. Tense changes.
/~T V 1.5 1.23 Deleted test 2PA-344-04. Tense changes.
1.5 1.24 ' Dense changes , vil
SLNMRY OF CHANGES Q SECOND SUPPLEMENT SECTIOi PAGE (S) OWEE 1.5 1.25 Tense changes. Reworded one (1) sentence. 1.5 1.26 Tense changes. 1.5 1.27 h nse changes. Mded test 2ST-348-01. 2.0 2.2 Three (3) granmer changes. 2.0 2.4 Redrew Figures. 2.0 2.7 Deleted one (1) sentence. 2.0 2.15 Mded " Combustion Engineerirg". Uncapitalized
" Senior Reactor Operator".
2.0 2.lC Mded one (1) sentence. 3.0 3.2 Mded paragraph. 'Ihree (3) wording changes. 3.0 3.5 Rewarded paragraph. g 3.0 3.7 Reworded paragraph. 3.0 3.14 'Ihree (3) sentences changed. 'Iwo (2)
.w ntences added.
3.1 3.15 k nse change. 3.11 3.16 Tense change. Garged " Pre core" to "Precore". 3.11 3.17 Tense changes. Mded "hmperature" . 3.12 3.19 Tense changes. Cbrrected three (3) typ>- graphical errors. 3.12 3.20 Tense changes. Mded descriptive material. 3.12 3.21 Tense charge. 3.12 3.23 Tense changes, Mdal descriptive material. 3.12 3.24 Tense changes. Mded descriptive material. 3.13 3.26 Mded "the" two (2) times, Ganged "an" to "One". Shortened one (1) sentence. O viii
1 SLMMARY OF CHANGES SECOND SUPPLEMENP SECTIN PAGE(S) CHANGE 3.13 3.27 Garged RC to RCS. Tense change. Mded words to two (2) sentences. 3.14 3.29 Tense change. 3.1.5 3.32 Tense changes. changed " f" to " F". 3.1.6 3.41 Tense change. changed "1 1/2" to "1.5" and "4" to "4.0". 3.1.8 3.45 Tense changes. One (1) sentence reworded. 3.1.9 3.47 Tense change. 3.1.10 3.49 Tense changes. One (1) wording change. 3.1.10 3.53 Redrew figure. 3.1.11 3.54 Tense changes. 3.1.11 3.56 Tense changes. ( 3.1.12 3.57 Changed "J-Tubes" to " elbows" 3.2.2 3.63 Tense changes. 3.4 3.68 Tense change. 5.0 5.2 Tense changes. 5.2 5.7 One (1) sentence reworded. 5.2 5.10 Mded a note. 5.2.1 5.11 Tense change. 5.2.2 5.13 One (1) sentence reworded. 5.2.3 5.14 Mded explanatory paragraphs. 5.2.6 5.28 Mitorial Charge. 5.3 5.29 Mitorial Change. 5.3 5.30 Editorial Change. 5.3 5.31 Redrew figure. ix
SLE W OF CHANGES SEODND SUPPLEME?E SECTICN PAGE(S) CHA!EE 5.3 5.32 Editorial Charge. 5.3 5.36 Mitorial Charge. 5.3 5.37 Editorial Charge. 5.3 5.39 Redrew figure. 5.3 5.40 Mitorial Charge. 5.3 5.40 Mitorial Change. 5.3 5.43 Editorial Charge. 5.3 5.45 Fedrew figure. 5.3 5.46 D31torial Charge. 5.3 5.47 Editorial Charge. 5.3 5.48 Mitorial Charge. 5.3 5.49 Mxlified figure. 5.3 5.50 Editorial Charge. 5.3 5.51 Fedrew figure. 5.3 5.52 Mitorial Change. 5.3 5.53 Editorial Change. 5.3 5.54 Editorial Change. '.3 5.56 Redrew figure. 5.3 5.57 D3itorial Changer 0 0 X
O
SUMMARY
OF CHANGES STOND SUPPLEMENT SECTION PAGE(S) CHANGE 6.0 6.2 Update to 100%, tense changes. 6.1 6.3 Update to 100%, tense changes. Updated the figure to 100%.
~
6.1 6.16-6.19
- 6.2 6.20 Tense change.
i 6.2.1 6.21-6.33 Mded new writeup. 6.2.2 ~6.34-6.41 Rewrote the section. 6.3 6.42-6.45 Tense changes, rewrite Method ard report recalculated results. , I 6.4' 6.48-6.56 Md new writeup. t O e.5 e;57 Tenee che ,ees , d e1 e ed r u e ences m 1 * % Turbine Trip. 6.5.1 6.58 Tense changes, wordirg changes. J 6.5.1 6.59 . Tense change. 6.5.1 6.60 Mded 100% results, wording and tense changes. 6.5.1 6.61 Redrew figure. 6.5.1 6.66 Mded new figure. 6.5.2 ' 6.67 Tense and wording changes. 6.5.2 6.68- Mded a sentence, added retest results. Rewrote one paragraph. 6.5.3 6.70 Tense changes.
'6.5.3 6.71 Tense changes. Mded paragraph.
6.5.3' 6.72 Updated to 100%.
- 6. 5. '3 :6.77-6.78 Mded new figure.
6.5.4 6.79 Ten,e' changes. 01 argal contraction. xi
i
SUMMARY
OF CHANGES SECOND SUPPLEMEh7 SECTION PAGE(S) CHA?EE 6.5.4 6.80 'Iense changes, added 100% hst Method. 6.5.4 6.81 wording aM tense changes. 6.5.4 6.82-6.83 Updated to 100%. 6.5.4 6.86-6.87 Added new figures. 6.5.5 6.88-6.89 Updated to 1006, tense changes. 6.6 6.90-6.92 Tense changes, update to 1001. 6.6 6.95-6.96 Added new figures. 6.8 6.104 Tense change. Incorporated 6.8.13 into 6.8.12 6.8.1 6.105 Updated to 100%, tense changes. 6.8.1 6.106 Tense change. 6.8.1 6.107 Updated to 100%, added note. 6.8.2 6.108-6.109 Updated to 1001, tense changes, wording changes. 6.8.3 6.110 Updated to 100%, tense changes. 6.8.3 6.111-6.112 Uglated to 1001. 6.8.3 6.113-6.115 Typed the tables. 6.8.3 6.116 Mded new table. 6.8.3 6.120 Added new figure. 6.8.4 6.121 Tense chdnges. 6.8.4 6.122 Rewrote Results and Conclusions 6.8.5 6.123 h nse changes. 6.8.5 6.124 Updated to 100%, added new paragraph, wording changes. 6.8.5 6.125 Added new table. ; 6.8.6 6.126-6.127 Tense changes. h 6.8.6 6.128-6.129 Added 100% Results, reworded description, corrected tables. xii
l !O SUWARY OF CHANGES SECOND SUPPLEFIN" SECTION PAGE(S) CHANGE 6.8.7 6.130 Tense changes, reworded test method. 6.8.7 6.132 Tense changes, corrected typographical error. 6.8.8- 6.133 Tense changes. updated to 100%, editorial changes ' 6.8.8 6.134 R nse changes, add sentence. 6.6.8 6.136-6.137 Tense changes, updated to 100%, changed " error" to " correction factor". 6.8.8 6.137-6.138 Mded nore description to figure i 6.8.8 6.134 Mded new. data >
- 6'8.8
. 6'140
. Mded new figure.
6.8.9 6.145 Updated to 100%, editorial changes.
' 6.8.10 6.146 h nse changes, editorial changes.
6.8.10 6.148-6.149 Rewrote Results and Problems & Ccunents. 6.8.11 6.150 R nse changes. , 6.8.12 6.151 Mded results. r 5 7 J xiii r I
y , - - ~ -. ,. ..- i SLMMM0' OP CHANGES SECCt:D SUPPLEMENT SECTICN PAGE(S) CHANGE 6.9 6.153 Update to 100%, tense changes. I 6.9 6.154 Mded a paragraph, rewrote the sununary, corrected one (1) typographical error. I 6.9 6.155-6.156 Updated Table to incitzle 100% power values. 6.10 6.219 Tense changes. 6.11 6.222 Updated' to 100% power, tense changes, changed
" stratification" to " profile".
%nse changes, added a new paragraph.
B 6.11.1 6.224 6.11.1 6.225 Updated to 100%, added new paragraph. 6.12 6.226 Added references to new test. l 6.12.1 6.227 % nse changes. 6.12.2 6.243 w nse changes. 6.12.2 6.244 01anged " generator electrical" to " turbine", tense changes. l i 6.12.2 6.245 Tense changes. 3 6.12.3 '6.251-6.253 Tense changes, deleted sentence. 6.12.3 6'.254 (brrectal one (1) typographical error. 6.12.4 6.255 Mded new writeup. I i 6.13 6.256-6.257 Bewritten to incorporate 100% power results. 6.14 6.258 Tense changes, deleted paragraph, added sentence. 6.14 6.259 Updated to 100% power, correctal one (1) 4 typxjraphical error. 6.15 6.260 w nse changes, updated to 100% power. 6.15 6.262 Updated to 100% power. O xiv 9
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SlMMARY OF CHANGES b SECOND SUPPLEMENT (continued) SECTICE PME(S) CHANGE 6.16 6.263 Updated to 100% power, aded reference to new section. 6.16.1 6.264-6.267 Updated to 100% power, tense changes. 6.16.1 6.268 Updated to the present. 6.16.2 6.269 Updated to present, changed tenses. 6.16.2 6.270 Tense changes. 6.16.2 6.271-6.274 Redrew figures. 6.16.4 6.276-6.277 Tense changes, updated to 100%, corrected one (1) tyggraphical error. 6.16.5 6.279-6.280 Tense changes, updated to 100% power, editorial y changes. 6.16.6 6.282 Tense changes, added 100% results. 6.16.6 6.283 Mded 100% results. 6.16.7 6.284-6.285 Mded new writeup. 6.17 6.286 Deleted reference to 100% Turbine Trip, editorial changes. 6.17.1 6.287-6.315 Mded new writeup. 6.17.2 6.316-6.327 Mded new writeup. 6.17.3 6.328-6.340 Mded new writeup. 6.17.4 6.345 Redrew the figure. O XV
O PREFACE
%e Initial Startup 'Iest Report described the Initial Startup
-Test Program for San Onofre Nuclear Generating Station Unit
'I%o (SOK;S 2) . We progran was a series of tests developed by the Southern California Edison Ocenpany which satisfied require-ments of the Nuclear Regulatory Carsnission as detailed in the SONGS 2 Final Safety Analysis Report (FSAR) . We FSAR incorpor-ated requirements of Regulatory Guide 1.68, Revision 0, and the SONGS 2&3 Safety Evaluation Report, NUREG-0712. We progran was an orderly evolution of three categories of tests which n O
C started with initial fuel load and ended with cm mercial power d S operation (August 8, 1983). We tests were designed to confinn a certain design bases and demonstrate, where practical, that the plant is capable of withstandire anticipated transients and postulated malfunctions. Selected equipnent failures and control syste n malfune-tions which could reasonably be expected to occur in the plant's lifetime were also simulated. %e initial report addressed testing through the 50% power plateau. We first supplement addressed testing through most of the 80% power testing plateau. %is second , supplement addresses the end of the 80% plateau and all of the f remainder of the testing not previously reported. This supplement is k issued as the last installment of the Startup Report. i xvii l 1
.p TABLE OF CON'IDfrS O
l SECTICN DESCRIPTION PAGE f I 1.0 INm000CTICN AND SUMMAIN 1.1 2.0 INITIAL RJEL IDADING 2.1 3.0 POSitDRE IDF RJNCTIONAL TESr 3.1 3.1 Nuclear Steam Supply System 'Msts 3.15 3.1.1 Precritical Intercanparison of Plant 3.16 Protection Systen (PPS), Cbre Protec-tion Calculator (CPC) Systen, Main (bntrol Board, 'and Process 0:aputer , Input Paraneters 3.1.2 Reactor Coolant Systen Flow Measurenents 3.19 3.1.3 Reactor Coolant Systen Heat Ioss 3.26 I i 3.1.4 Reactor Coolant Systen 'Ihennal Expansion 3.29 ! ! 3.1.5 Cbntrol Elenent Drive Mechanism (CEm) 3.31~ l and Control Elenent Assembly (CEA) ( 3.1.6 Fixed Incore Instrunentation 3.41 j 3.1.7 ft>vable Incore Instrunentation 3.44 i i 3.1.8 Pressurizer Safety Valve Setting 3.45 [ 3.1.9 Pressurizer Performance 3.47 I i 3.1.10 Pressurizer Spray Valve ard Cbntrol 3.49 Adjust 2nent l l 3.1.11 Primary. and Secondary Water Oienistry 3.54 , t 3.1.12 Stean Generator Feedwater Rirg Integrity 3.57 l 3.1.13 Pressurizer Spray Effectiveness 3.60 3.2 Plant Tests 3.61 t 3.2.1 Auxiliary Feedwater Punps 48 Hour 3.62 ', Endurance Run O l g 3.2.2 Power Ascension Data Record 3.63 xix
TABLE OF CCNTENTS SECTION DESCRIPTION PAGE 3.3 Comunications 'Ibsting 3.64 3.3.1 UHF Radio 3.65 3.3.2 Telephone 3.66 3.3.3 Area Radiation Monitoring System 3.67 Audibility 3.4 Containment Isolation Valve Test 3.68 4.0 INITIAL CRITICALI'lY AND CEA EXERCISES 4.1 4.1 CEA Exercises 4.4 4.2 Dilution to Criticality 4.5 4.3 Withdrawal to Criticality 4.7 4.4 Neutron Detector Overlap 4.10 5.0 IIM IOWER PHYSICS TESTS 5.1 5.1 Zero Power Biological Shield Survey 5.6 5.2 Iow Power Physics 1bst Program 5.7 5.2.1 CEA and Part length CEA Symnetry 5.11 5.2.2 Isothermal Tenperature Coefficient 5.13 5.2.3 Shutdown and Regulating CEA Worths 5.14 5.2.4 Boron Worth Measurements 5.26 5.2.5 Critical Baron Concentration Measurements 5.27 5.2.6 Prcblems and Solutions 5.28 5.3 An Analysis of the SONGS 2 Iow Pcraer Natural 5.29 Circulation Denonstration 6.0 POWER ASCENSION TESTS 6.1 6.1 Sunmary of Power Range Testing 'Ihrough 6.3 100% Power 6.2 Core Power Distribution Tests 6.20 6.2.1 Dropped CEA 6.21 6.2.2 Pseudo Ejected CEA 6.34 XX
.. . . . - . - ~ _._ -. _ - . . .-
1. O == - SECIION DESCRIPTION . PAGE, 6.3 Variable T average (Moderator Temperature a d 6.42 Power Coefficient Detemination) 6.4 Unit Ioad Transient 6.48 6.5 Control Systems Checkout 6.57 6.5.1 SBCS Perfomance 6.58 I 6.5.2 SBCs capacity Checks 6.67 5 6.5.3 Feedwater Control System Perfomance 6.70 i 6.5.4 Reactor Regulating System Perfomance 6.79 6.5.5 Integrated Control System Perfomance 6.88 < r 6.6 BCS Chemistry and Radiochemistry 6.90 67 O. s"= tao =oet iae the co#tret a - 6 97 , j 6.8 Steacy-State Core Perfomance 6.104 ; 6.8.1 Nuclear Steam Supply System (NSSS) Calori- 6.105 , metric Detemination of Reactor Power ; ) 6.8.2 Incore Detector Signal Verification 6.108 j 6.8.3 Core Performance Record 6.110 ) 6.8.4 CDIES Power / Flow Verification 6.121 I 6.8.5 Nuclear and %ermal Power Calibration 6.123 l d 6.8.6 CDIES Secondary Pressure loss 6.126 Adjustment i 6.8.7 Linear Power Subchannel Calibration 6.130 ! 6.8.8 HCS Calorimetric Flow Measurements 6.133 6.8.9 RCS 4 T Power Detemination 6.145 O
- xxi ,
TABLE OF CONITNI'S SECTION DESCRIPTION PAGE 6.8.10 Mwable Incore Detector Check 6.146 6.8.11 NSSS fiand Calorimetric 6.150 ~ 6.8.12 Warranty Run 6.151 6.9 Biological Shield Effectiveness Survey 6.153 6.10 Xenon Oscillation Control Test Using 6.219 Part length CEAs 6.11 Interconparison of PPS, CPCs, and Process 6.222 Catputer Inputs and Outputs at Power 6.11.1 CPC/CDLSS Verification 6.224 6.12 Verification of CPC Power Distribution 6.226 Felated Constants 6.12.1 shape Annealing Matrix (SAM) Measurement 6.227 g 6.12.2 Tem;nrature Decalibration Verification 6.243 6.12.3 CPC Power Distribution Constants Verifi- 6.251 cation 6.12.4 CPC Bermal Power Decalibration 6.255 6.13 i<eactor Baseline Vibration Monitoring 6.256 6.14 Steady State Vibration 6.259 6.15 Dynamic Effects 6.260 6.16 Secondary Plant Information 6.263 6.16.1 Turbine Generator Benchmark Input / 6.264 Output Test 6.16.2 Circulating Water fleat Treatment 6.269 6.16.3 Turbine Overspeed Trip Test 6.275 6.16.4 Power Ascension %ennal Expansion 6.276 O xxii
i l . t 2 TABLE OF C NTENTS , t i f SECTIN DESCRIPTION PAGE ^ 6.16.5 HVAC Temperature Survey 6.279 l
! 6.16.6 Cable Temperature Monitoring Progran 6.282 1
6.16.7 Condensate, Feedwater and Heater Drain 6.284 i ! Pump Performance Test i I 6.17 Transients 6.286 i i Natural Circulation Verification 6.287 6.17.1 i (Frcan 80% Power) ( i I
; 6.17.2 Unit Ioad Rejection (100% Generator 6.316 4 Trip) >
4 i 6.17.3 Total Ioss of Offsite Power 6.328
- (Simulated Ioss of Onsite and i
- Cffsite Power) i ,
g 6.17.4 20% Reactor Trip 6.341 { 4 ! I j' ; i f I h i i i 1 ! i 1 1
- o I !
l ! l 4 xxiii 1
LISP OF FIGURES FIGURE NO. TITLE PAGE 1.2.1 Iscuetric View of Nuclear Stem Supply Syste 1.4 1.5.1 Overview of Planned SONGS 2 Startup frm Fuel 1.7 Ioad to Warranty Ibn 1.5.2 A Typical Power Plateau 1.8 2.0.1 Side View of Equipnent Used in Fuel Ioading 2.4 2.0.2 'Ibp view of RIuipnent Used for Fuel Ioadirg 2.4 2.0.3 Ioading Sequence (Fuel Ioading) 2.6 2.0.4 Initial Fuel Ioading - Ntaber of Ebel Bundles 2.9 toaded versus Time and Date 2.0.5 Inverse Multiplication Plot (Fuel Ioading) 2.18 2.0.6' Count Rate Data (Fuel Ioaling) 2.19 O 2.0.7 Core Inading Map - Fuel Bundle Location 2.20 2.0.8 Core Ioading Map - Control Elment Assably 2.21 Iocation 3.1.2.1 FSAR 4 Pmp (bastdown Curve (RCS Flow Measure- 3.25 ments) 3.1.3.1 Stem Generator Water Level Change During Stem- 3.28 down (RCS Heat Ioss Test) 3.1.4.1 'Ihemal Expansion Measurment - Lines in Contain- 3.30 ment 3.1.5.1 CInt Cold Drop Times 3.34 3.1.5.2 CEDM lbt Drop Times , 3.35 3.1.6.1 Sketch of Incore Detector Systm 3.43 O , XXV i l ._ l
LIST OF FIGURES ((bntinued) FIGURE NO. TITLE PAGE 3.1.9.1 Pressurizer (bntrols 3.48 3.1.10.1 Pressurizer Spray Valve and Bypass Valve 3.51 Incation 3.1.10.2 Pressurizer Spray Line Thermocouple locations 3.52 3.1.10.3 Depressurization Rates (Pressurizer Spray Valve 3.53 Control Adjust:nents) 3.1.12.1 Incation of Ebedring in Stean Generators 3.59 4.2.1 doron Dilution Evolution: Inverse (bunt Rate 4.6 Ratio and Boron Concentration versus Time 4.3.1 Inverse Count Rate Ratio versus CEA Group Position 4.8 During CEA Withdrawal to Criticality (Startup Gannels) 4.3.2 Inverse Cbunt Rate Ratio versus CEA Group Position 4.9 Durity CEA Withdrwal to Criticality (Safety Channels) 4.4.1 Startup and Safety (Iogarithnic) Gannel Overlap 4.11 5.0.1 Iow Power Physics Testirn (Activities and Time) 5.3 5.2.1.1 CEA Symmetry Pesults 5.12 5.2.3.1 Integral Bod Wrth versus Ibd Position - Banks 5.15 4, 5 and 6 (overlap) 5.2.3.2 Integral Pod Wrth versus Ibd Position - Banks 5.16 4, 5 and 6 (no overlap) 5.2.3.3 Integral Rod Worth versus Ibd Position - Banks 5.17 5 and 6 (no overlap) 5.2.3.4 Integral fed Wrth versus Ibd Position - Bank 5.18 4 (no overlap) 5.2.3.5 Integral Ibd Wrth versus Tod Position - Control 5.19 Banks (overlap) 5.2.3.6 Integral Pod Wrth versus Tod Position - Banks 5.20 1, 2 and 3 (r.o overlap) xxvi O
LIST OF FIGURES (@ntinued) O FIGURE NO. TITLE PAGE 5.2.3.7 Integral Ibd Wrth versus Ibd Position - Banks 5.21 4 and 3 (no overlap) 5.2.3.8 Integral Ibd Wrth versus Ibd Position - Banks 5.22 6 and 5 (no overlap) 5.2.3.9 Ibd Withdrawals from PDIL @ ibt Zero Power 5.23 5.2.3.10 Reactivity Insertion versus Part Length Ibd 5.24 Withdrawal 5.2.3.11 Integral Ibd Wrth versus Ibd Position - Banks 5.25 2 and 1 (no overlap) 5.3.1 Isometric View of SONW 2 Nuclear Stem Su[ ply System 5.31 5.3.2 rocation of Instrunents on the RCS 5.33 5.3.3 Trends of Significant Parmeters (Natural 5.34 Circulation) 5.3.4 Natural Circulation Flow Rates versus Reactor Power 5.38 5.3.5 Vessel aT and Flow versus Reactor Power 5.39 5.3.6 Ccenparison of Measured and Calculated foop Transit 5.40 Times 5.3.7 Isolated Stem Generator Dmonstration (Natural 5.42 Circulation) 5.3.8 Core Exit 'Ihemocouple Response Durig Isolated 5.44 Stem Generator Demonstration (Natural Circulation) 5.3.9 Ibric Acid Injection Point and Reactor Coolant 5.45 Syste 5.3.10 Reactivity versus Time for a Ten callon Injection 5.47 of Ibric Acid into One loop 5.3.11 Cbid IAg Fluid Entry Angles and Flow Directions 5.49 5.3.12 (bnponents mich Encourage MixiN in the RG 5.51 5.3.13 Sketch of (bmponents Associated with 5.56 Pressurizer Spray O xxvii
LIST OF FIGURES (Cbntinued) O FIGURE NO. TITLE PAGE 6.1.1 Power Range Testiry 6.4 6.2.1.1 Reactor PoWr vs Time (FIfEA Drop) 6.26 6.2.1.2 ICS Ebt Leg Temperature vs Time (FIfEA Drop) 6.27 6.2.1.3 Pressurizer Pressure vs Time (FILEA Drop) 6.28 6.2.1.4 Pressurizer Iovel vs Time (FILEA Drop) 6.29 6.2.1.5 Reactor Power vs Time (PIfEA Drop) 6.30 6.2.1.6 ICS lbt Leg Tm.perature vs Time (PIfEA Drop) 6.31 6.2.1.7 Pressurizer Level vs Time (PIfEA Drop) 6.32 6.2.1.8 Pressurizer Pressure vs Time (PICEA Drop) 6.33 6.2.2.1 Reactor Power vs Time (Pseudo Ejection) 6.38 6.2.2.2 ICS lbt Leg Tm.perature vs Time (Psenio Ejection) 6.39 6.2.2.3 Pressurizer Level vs Time (Psealo Ejection) 6.40 6.2.2.4 Pressurizer Pressure vs Time (Pseulo Ejection) 6.41 6.3.1 variable T average Test mthod 6.46 6.3.2 Strip 01 art Recordirgs of Temperature and Power 6.47 Durirg IK Measurments 6.4.1 Pressurizer Pressure and Level vs Time Durity 6.52 [ Unit Ioal Transient Test 6.4.2 Reactor Power and Generator Ioai vs Time During 6.53 Unit Ioai Transient Test 6.4.3 Stem Generator #2 Pressure vs Time Durirn Unit 6.54 Ioad Transient Test 6.4.4 lbt Inj 2 Tmperature vs Time Durirn Unit Ioal 6.55 Transient Test 6.4.5 Stem Generator #2 Level vs Time Duriry Unit Ioal 6.56 Transient Test O xxv111
.- .-. . -- - . .. -. - .. . . - -- .. . .~. . . . ~ . _ -
l I LIST OF FIGURES (Cbntinued) O
- FIGURE NO. TITLE PME l
6.5.1.1 Sketch of Stem Bypass (bntrol Syste 6.61
- 6.5.1.2 SBCS Master Controller Setpoint Progra 6.62 6.5.1.3 SBCS Response to Turbine Ioad Swings at 50% 6.63 l Power Prior to Design Change l 6.5.1.4 SBCS Response to Turbine Ioad Swirgs at 50% 6.64 Power Af ter Design Change 6.5.1.5 SBCS Response to Turbine Ioad Swings at 80% Power 6.65 6.5.1.6 SBCS Response to Turbine Ioal Swings at 100% Power 6.66 6.5.2 1 Stem Bypass Valve 2tN-8425 Capacity Curve 6.69 6.5.3.1 S/G Invel Response to Level Setpoint Perturbations 6.73 at 20% Power
- i 6.5.3.2 S/G Invel Response to Level Setpoint Perturbations 6.74 at 50% Power
]
6.5.3.3 S/G Invel Response to Level Setpoint Perturbations 6.75 i at 80% Power i l 6.5.3.4 S/G Ievel Response to Level Setpoint Perturbations 6.77 at 100% Power 6.5.4.1 Reactor Regulatirg Syste Test at 20% Power 6.84 t 6.5.4.2 Reactor Regulatiiv3 Syste Test at 50% Power 6.85 6.5.4.3 Reactor Regulatirg Syste Test at 80% Power 6.86 6.5.4.4 Reactor Regulating Syste Test at 100% Power 6.87 6.6.1 Gross Letdown Activity ard Power vs. Time 6.93 6.6.2 RCS Activity ard Power vs Time (May,1983) 6.94 6.6.3 RCS Activity ard Power vs Time (August,1983) 6.95 6.6.4 Iodine Ration vs Time 6.96 6.7.1 Parmeter Behwior - Shutdown Frm Outside 6.100 the Control Rom xxix I
LIST OF FIGURES (Cbntinued) O FIGURE NO. TITLE PME 6.7.2 Cooldown Fran Outside the (bntrol Roan 6.102 6.8.3.1 Radial Power Distribution at 20% Power 6.117 6.8.3.2 Radial Power Distribution at 50% Power 6.118 6.8.3.3 Radial Power Distribution at 80% Power 6.119 6.8.3.4 Radial Power Distribution at 100% Power 6.120 6.8.7.1 Excore Detector Signal Paths 6.131 6.8.8.1 RID Incation and Temperature Profile 6.137 (RCS Calorimetric Flow Measurement) 6.8.8.2 liot Leg il 'Iemperature Profile - Unrodded 6.140 6.8.8.3 lbt Leg 42 Temperature Profile - Unrodded 6.141 6.8.8.4 CEA Group Locations in SONGS 2 Core 6.142 6.8.8.5 lbt Leg il Temperature Profile - Rodded 6.143 6.8.8.6 lbt Leg #2 Tenperature Profile - Rodded 6.144 6.8.10.1 MICDS Pause Interval Determination 6.147 6.9.1 Bioshield Radiatico Survey Points 6.167 6.10.1 Group 6 Position, PIfEA Position and COISS ASI vs. 6.221 Time 6.12.1.1 Arrangenent of CPC Excore Neutron Detectors 6.236 6.12.1.2 COLSS ASI vs. Time (20% Test) 6.237 6.12.1.3 1/3 Core Fractional Powers vs. Time (20% Test) 6.238 6.12.1.4 Fractional Detector Respnse vs. Time (20% Test) 6.239 6.12.1.5 COLSS ASI vs. Time (50% Test) 6.240 6.12.1.6 Fractional Detector Response vs. Time (50% Test) 6.241 XXX 0 l
=- - _ -- .- - . . - _ . . . . -
M LIST OF FIGURES (Cbntinued) O
- FIGURE NO. TITLE PAGE 6.12.1.7 - 1/3 Core Fractional Power vs. Time (50% Test) 6.242 6.12.2.1 Raw knperature Shadow vs. T cold (CPC Gannel A) 6.247 6.12.2.2 Raw Tenperatre Shalow vs. T cold (CPC Channel B) 6.248 ,
6.12.2.3 Raw Tenperature Shadow vs. T cold (CPC Gannel C) 6.249 , 6.12.2.4 Raw Tenperature Shadow vs. T cold (CPC Gannel D) 6.250 ) 6.16.2.1 Circulating Water Systen 6.271 1 6.16.2.2 Intake Structure (Nonnal Flow) 6.272 i j 6.16.2.3 Intake Structure (Heat Treatreent of Discharge 6.273 j Cbnduit) l 6.16.2.4 Intake Structure (Heat Treatment of Intake 6.274 ) Cbnduit) i J 6.16.4.1 Pipe Hanger Locations for Main Stean and 6.278 j Feedwater Lines for hennal Expansion Testing 6.16.5.1 Typical Piping Penetration and 'IWnporary 6.281 2ennocouples (HVAC 'IWnperature Survey) i 6.17.1.1 Overview of Natural Circulation Test 6.302
) 6.17.1.2 Hot Leg and Cold Iag Tenperature - First Two 6.303 i Minutes (Nat. Circ.)
] 6.17.1.3 Hot Leg and Cold Iag 'Itmperature - First Fifteen 6.304 < a Minutes (Nat. Circ.) t 6.17.1.4 Pressurizer Level and Pressure During First 6.305 Fifteen Minutes (Nat. Circ.) 4 6.17.1.5 Reactor Vessel Differential Tenperature 6.306 i (Nat. Circ.) 6.17.1.6 Reactor Vessel Differential 'IWnperature and Cold 6.307 l IAg 'IWaperatTe vs Time (Nat. Circ.) l 6.17.1.7 Core Decay Heat vs Time (Nat. Circ.) 6.308 !O xxxi 1 l 1 j 1
-. ..,>,-._,.-~-.y-c-- ,,,,-,---..,,_.-,-_._m,-,,m,-y_.. , , - . - - , , , . . , . . . . , . , - _ . _ . - , . . . . . . . _ . _ - , . - - . _ , - .
1 LIST OF FIGURES ((bntinued) , FIGURE NO. TITLE PAGE 6.17.1.8 Water Usage During Natural Circulation Test 6.309 6.17.1.9 03rdensate Storage Tank Level vs Time (Nat. Circ.) 6.310 6.17.1.10 Difference frcm Average Temperature for RIDS 6.311 Around Hot Leg #1 vs Time (Nat. Circ.) 6.17.1.11 Difference frm Average Temperature for RIDS 6.312 Around Hot Leg #2 vs Time (Nat. Circ.) 6.17.1.12 Atmospheric Dunp Valves Position and Accunulator 6.313 Nitrogen Pressure vs Time (Nat. Circ.) 6.17.1.13 BCS Ebration - Doron Concentrations (Nat. Circ.) 6.314 6.17.1.14 Irtdown Flow and Pressurizer Pressure and Level 6.315 Durirg RCS Depressurization 6.17.2.1 100% Generator Trip - Hot Leg Temperature vs Time 6.320 6.17.2.2 100% Generator Trip - Pressurizer Level vs Time 6.321 6.17.2.3 100% Generator Trip - Pressurizer Pressure vs Time 6.322 6.17.2.4 100% Generator Trip - Stem Generator Pressure vs 6.323 Time 6.17.2.5 100% Generator Trip - Hot and Cold Leg Tenperature -6.324 vs Time 6.17.2.6 100% Generator Trip - Pressurizer Pressure and 6.325 Iavel vs Time 6.17.2.7 100% Generator Trip - Stean Generator 2(E088) 6.326 Paraneters vs Time 6.17.2.8 100% Generator Trip - Stean Generator 1(E089) 6.327 Paraneters vs Time 6.17.3.1 Pressurizer Pressure vs Time (Insa of Of fsite Power) 6.333 6.17.3.2 Pressurizer Level vs Time (Inss of Offsite Power) 6.334 6.17.3.3 Ioop fl Ibt ard Cold Irg Temperatures vs Time 6.335 (Inss of Of fsite Power) xxxii 9
LIST OF FIGURES (Continued) O FIGURE NO. TITLE PAGE 6.17.3.4 Ioop #2 Hot ard Cold Iag '1Dnperatures vs Time 6.336 (Loss of Offsite Power) 6.17.3.5 Stem Generator #2 Level vs Time (Ioss of 6.337 Offsite Power) 6.17.3.6 Ste m Generator #1 Level vs Time (Loss of 6.338 Offsite Power) 6.17.3.7 . Stem Generator 42 Level vs Time (Ioss of 6.339 Offsite Power) 6.17.3.8 Ste m Generator il Pressure vs Time (Loss of 6.340 Offsite Power) 6.17.4.1 Block Diagra of SONGS Unit 2 6.345 6.17.4.2 Pressurizer Parmeters frca the 20% MCB 6.346 Reactor Trip 6.17.4.3 SG E088 Parmeters from the 20% MCB Reactor Trip 6.347 6.17.4.4 SG E089 Parmeters from the 20% MCB aeactor Trip 6.348 O xxxiii
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l l l LIST OF TABLES v TABLE 10. TITLE PAGE 1.5.1 Test and FSAR Section Cross Reference 1.9 1.5.2 Synopsis of Testing 1.15 2.0.1 Bundle Loading Time Distribution (Fuel Icadirg) 2.13 3.0.1 Postcore Hot Functional Sequence of Events 3.4 3.0.2 Samary of Problens (Postcore Hot Functionals) 3.14 3.1.1.1 Devices in Wich Paraneters are Displayed 3.18 (Instrunent (brrelation) 3.1.2.1 RCS Flow - Measured Values 3.24 3.1.5.1 CEA Drop Times 3.36 3.1.5.2 Ten Drops of Fastest and Slowest CE0Ms 3.40 3.1.11.1 Primary Genistry Specifications 3.55 3.1.11.2 Secondary Genistry Specifications 3.56 5.2.1 Measured CEA ibrths at Hot Shutdown 5.9 5.2.2 Heatup and Pressurization Data 5.9 5.2.3 Measured Reactivity Worths at Hot Starriby 5.10 5.3.1 Natural Circulation Conditions 5.36 5.3.2 Gargirg velocities and Temperatures into 5.54 Cold tag 5.3.3 Natural Circulation and Chargirg Flow Velocities 5.54 5.3.4 Mixed Charging Flows ard Conditions 5.55 6.2.1.1 Data from Dropped CFA Test (FICEA) 6.24 6.2.1.2 Data from Dropped CEA Test (PIfFA) 6.25 6.2.2.1 Data fran Pseudo Ejected CEA Test 6.37 6.4.1 Sirgle Value Acceptance Criteria (Unit Ioad 6.51 Transient Test) O XXXV
l l LIST OF TABLES O' TABLE NO. TITLE PAGE 6.6.1 RCS Chmistry and Radiochmistry 'Ibst - RCS 6.92 Activity Levels 6.8.3.1 Ibwer Distribution Sur. mary 6.111 6.8.3.2 Axial Power Distribution at 20% Power 6.113 6.8.3.3 Axial Power Distribution at 50% Power 6.114 6.8.3.4 Axial Power Distribution at 80% Power 6.115 6.8.3.5 ~ Axial Power Distribution at 100% Power 6.116 6.8.5.1 Nuclear and 'Ihermal Power Calibration Results 6.125 6.9.1 Bioshield Radiation Survey Points 6.155 6.11.1 Parmeters Manitored for Process Variable 6.223 Interemparison 6.12.1 Shape Annealirg Matrix and Boundary Condition 6.235 Measurment 6.16.6.1 Cable Temperature mnitoring Results 6.283 6.17.1.1 Natural Circulation Verification Sequence of Events 6.288 O xxxvi
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5 f ) . ^ l t t ( I, , I I j SECTION 1.0: INIB000CTION AND SUK%IN 4 i { I I 4 4 I l
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1.0 INTRODUCTION
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SUMMARY
1.1 The Startup Report he following Startup Report for the San Onofre Nuclear ~", Generating Station Unit 'Iko (SCNGS 2) fulfills 'Ibchnical Speci- qq fication 6.9.1, testing frm issuance of low power operating license gg through cmpletion of testing. The Technical Specification requires a mm Startup Report be written which a3 dresses each test identified in the Final Safety Analysis Report (FSAR) . % e Startup Report is required to include a description of the measured values of the operating conditions or characteristics of the plant obtained during the test progran and a cm.parison with design predictions and specifications. Any corrective actions that were required to obtain satisfactory - operation must be described. Any additional specific details required ; in license conditions based on other crmmitments (such as the Safety g Evaluation Report) must also be included. lj t is report addresses the requirements by describing each of the l"J tests and problems encountered during the test. The report must be subnitted within (1) 90 days following cmpletion of the startup test S progran, (2) 90 days following resumption or commencement of ccmnercial 5 plant operation, or (3) 9 nonths following initial criticality, whichever is earliest. m is report originally satisfied the third time requirement. Since it did not cover all three evolutions (initial -n criticality, cmpletion of startup test progran, and cxmnencement of .. ccanercial power operation), the first sul.plemental report was 21 issued as required by 'Ibchnical Specifications three months later. @% mis second and final supplement addresses the balance of testing ** up to and including the start of ccmnercial operation.
.l.2 he Facility SotGS 2 is one of three pressurized water reactors located on -
federal land at the Pacific Ocean on the Carp Pendleton Military . Reservation in north San Diego County. It is just south of the city lE of San Clemente, approximately midway between Los Angeles and San ? Diego, California. It is owned primarily by Southern California
- Edison in conjunction with San Diego Gas and Electric, h e City of Anaheim, and he City of Riverside, California. It is operated by Southern California Edison. SCNGS 2 and its sister, SCtGS 3, are 3410 megawatt thermal reactors manufactured by Cmbustion Engineering, -
Incorporated. Each plant has a single turbine-generator rated'at a net 1100 electrical megawatts, manufactured by the General Electric Corporation, a British fim. The Ios Argeles Ibwer Division of ljg y Bechtel Power Corporation is the Architect-Ergineer and Constructor. 1.2 O l
1.2 %e Facility (Continued) i
%e Unit has a separate containment structure, safety equip-ment building, turbine building, diesel generator building and fuel handling building. Unit 2 shares an auxiliary. building and intake structure with Unit 3.
N l An isanetric view of the hSSS is shown in Figure 1.2.1. We - NSSS generates approximately 3410 Mwt and contains two primary l}g l
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coolant loops, each of which has two reactor molant pumps. Each coolant loop also-contains a stean generator, a 42 inch ID hot leg and - two 30 inch.ID cold legs. - An electrically heated pressurizer is connected to one of the loops. -Safety injection piping is connected to each of four cold legs and two hot legs. Reactor coolant punps are single stage cmbined centrifugal-axial types, driven by electric motors. We steau generators are of vertical U-tube design. f W e reactor core is fueled with uranium dioxide pellets enclosed I in Zircaloy tubes with welded end caps. % e tubes are fatri-cated into assenblies in which end fittings and grids prevent axial N and lateral motion of the tubes. %e control element assemblies - (CEA's) consist of NiCrFe alloy-clad boron carbide neutron absorber @ rods, which are guided by tubes located within the fuel assenbly. j
% e core m nsists of 217 fuel assemblies which were initially loaded with three different U-235 enrichments. h e NSSS full-thermal output ! is 3410 mt with a core thermal output of 3390 Mt. No provisions for stretch capability are included in the design of the NSSS.
4 A nore couplete description of the facility can be found in the Final Safety Analysis ReEnrt. ' i i r k. i 4 !
! 1.3 1
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Figure 1.2.1 l l ISOMETRIC VIEW 0F NUCLEAR STEAM SUPPLY SYSTEM (LOOKING WEST) 1 1.4
1.3 h e Test Program N p . (,) te Initial '1bst Prcgram was a series of tests developed by the Southern California Edison Ccxnpany to meet the requirements l}y of the Nuclear Regulatory Ccanission as detailed in Regulatory Guide 1.68, Revision 0, and the SONGS 2&3 Safety Evaluation Report (SER), NUREG-0712. Se Edison Ccanpany ccanitted to cmpleting the progran outlined in the SER and Chapter 14 of the FSAR, both of which detail objec- ' tives, methods, and acceptance criteria for each test. %e program was an orderly evolution which started with initial fuel load and e ended with commercial power operation. It was designed to confirm J; certain design bases and de:ronstrate, when practical, that the plant Eg is capable of withstandity anticipated transients and postulated 5g failures. It also simulated selected equipment failures and control system malfunctions which could reasonably be expected to occur during the plant's lifetime. ~ R e tests fall into three categories as defined by Regulatory Guide 1.68; precritical, low-power, and power range tests. Pre- lJE critical tests are those conducted after fuel has been loaded but 5 before tt.2 reactor sustains its first critical operation and are called Postcore Hot Functionals. Postcore Hot Functionals are - performed to insure that the reactor is in proper condition to begin j cperation. We reactor is brought frczn ambient to operating condi- g tions in a series of steps, or plateaus, and operability checks and j (_l final test measurements of equiptent which could be done only with V the core in place are performed. Examples of equignent are control m rods and incore neutron monitority instrumentation. Examples of . measurements include reactor molant system flow and flow coastdown following Reactor Coolant Pump trips. lI S Re second category, low-power physics testing, is conducted with the - reactor critical but producing no measurable heat. Ecw power testing ; consists of measurements which verify that radiation shielding and g core physics parameters are as expected, and that core reactivity g mefficients are as assumed in the Safety Analysis. S e first of two series of natural circulation tests, the Low Power Natural Circula-tion Derronstration, is performed after the reactor is brought to a power level where measurable heat is produced. Because powr is less than the low power operating license limit of 5% full power, the , tests are considered to be low power tests. , 2e third and final category, pwer ascension testing, occurs E after the receipt of the full power operatirg license and involves m testing at reactor power levels greater than 5% of full power. We full power license was received after the natural circulation demon-stration tests were performed. Power ascension l;N E a o
) 1.5 m
l l l l tests are performed to demonstrate the ability of the plant to withstand transients, to make final adjustments to equipnent, to show that natural circulation is capable of providing suffici- E ent core cooling and to provide operator training. At-power per- 3 formance data is measured to assure that core power distribution and other performance characteristics, including at-power reactivity coefficients, are as expected. Khereas the first two phases deal mostly with the reactor coolant system ard care, the ~ third involves integrated operation of the entire plant, including . the secondary system, control systems and operating and off-normal l~g procedures. j 1.4 'Ihe Initial Startup Report 'Ihe Initial Startup Report addresses testing which was performed to satisfy FSAR requirements and to test certain balance of plant cmponents. It does not address surveillance or other operating procedures except as they relate directly to the startup test progran. 'Ihe report is organized into the following:
- 1) General Overview of Testing
- 2) Initial Fuel Ioad
- 3) Postcore Hot Functionals
- 4) Initial Criticality
- 5) Iow Power Physics 'Ibsts
- 6) Power Ascension Tests Each of the tests is described in further detail. A cross reference between the tests, FSAR requirements, and report section follows in the Overview. 'Ihe tests are described by FSAR section so several tests will be incitded in a single category. A single section permits a more cohesive description of the tests and easier assessment of trends and plant behavior.
1.5 General Overview of Testing Figure 1.5.1 below, Overview of SONGS 2 Startup fran Fuel Inad n to Warranty Run, presents the concept of initial Startup which . -involves fuel load, followed by postcore hot functionals, l'd initial criticality, low power physics, and power range testing. *
?
Power range testing consists of four power plateaus, or periods of nearly constant reactor power, at 20, 50, 80, and 100% reactor power. A typical plateau is outlined in Figure 1.5.2.
'Ibsts performed at each plateau are listed in Table 1.5.1. A short description of each test is included in Table 1.5.2, starting with Fuel Ioad and concluding with pur ascension.
Table 1.5.1 is a cross reference anong Report Section, FSAR Section, and Test. Procedure Number, and shows how various tests satisfy portions of Chapter 14 of the FSAR. A detailed description is given for each evolution, Fuel Ioad, Postcore Hot Functionals, Initial Criticality and Iow Power Physics, and Power Range Testing. 1.6
~ . . . _ . . .. ..- -- _ _ _ . . . _ . _ _ . . . _ = . . - . _ _ . . _ _ . . - - - . - _
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) - O FIGUR E :1. 5.1 OVERVIIN OF PLANNED SONGS 2 STARTUP FROM FUEL IDAD TO WARRANTY RUN t
POSTCORE LOW POWER RANCE TESTING , 110T POWER < FUEL IAAD FUNCTIONALS PilYSICS 20%- 50% l 80% 100% WARRANTY j _ __ . . l . _. l "FL"..........." IIB"........"IC & LP"............... TEST DESIGNATOR ...."PA"...............................
~ COMMERCIAL r 4 OPERATION- l REACTOR COOLANT SYSTEM TEMPERATURE I
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- Turbine - Generator l Natural _ Trip [ Trip k- Shutdown Outside Circulation Verification 100%
Warranty Control Room - u ' Main Control 80% Plateau
) Plat-eau Run 3
/ Synchronize Main s h BoardTrips) 8 Generator 50% Plateau i a4 -
/. 20% Plateau g Total Loss Of Offsite Power i i 1
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FIGURE 1.5.2 A TYPICAL POWER Pl.ATEAll A typical power rande testing plateau is shown below. There are four major plateaus (20%, 50%, 80%, and 100% reactor power) which consist of three general parts. First, after a plateau's power level has " been initially attained, adjustments and data readings are made. Second, a period of steady-state power operation is maintained and tests are performed which do not require xenon stability. Finally, core - physics and other tests which require stable power levels and core power distributions are lg performed in the xenon stability portion. A trip will terminate the plateau; each trip has a specific a purpose.
^ -Initial Adjustments me Xenon Nonequilibrium -w- Xenon Equilibrium -> Trip Plateau And Readings Testing Testing Power
~
I'***I Initial adjustments Plant and core testing Physics testing and of control circuits which does not require other non-physics , and data taking. plant stability is testing requiring { performed. great stability is Power Initial performed. Level Power Increase From Previous 4 Power Time -+ Tests in the xenon nonequilibrium category may generally be performed in any order. Those requiring xenon equilibrium fall into two categories. The first requires all CEAs to be completely withdrawn (called, "unrodded") and are generally performed before the second, in which CEAs are inserted. O O O
TABLE 1.5.1 TEST AND FSAR SECTIN CROSS REFERENCE APPLICABLE TEST % POWER
- SECTION SECTICN AND TEST DESCRIPTION FSAR SECTION NUMBER LEVEL 2.0 INITIAL EUEL U%D 14.2.10.1 2FL-101-01 SC 3.0 POSICORE HOP EUNCTIONAL TESTS 2HB-213-01 SC 3.1 Nuclear Steara Supply Systm Tests 3.1.1 Precritical Intercomparison of Plant 14.2.12.78 2HB-458-01 SC Protection System (PPS), Core Protection 2HA-317-01 Calculator (CPC) System, Main Control Ebard, ard Process Canputer Input 4
Paraneters 3.1.2 Reactor Coolant System Flow 14.2.12.74 2HB-213-05 SC 3.1.3 Reactor Coolant Sytem' Heat Ioss 14.2.12.73 2HB-213-06 SC (Rotest) 3.1.4 Reactor Coolant System Thermal 14.2.12.72 2HB-102-01 SC Expansion (Retests) 3.1.5 Control Elenent Drive Mechanism (CEDM) 14.2.12.75 2HB-316-01 SC and Control Elenent Assembly (CFA) 3.1.6 Fixed Incore Instrumentation 14.2.12.67 2HB-310-01 SC 14.2.12.80 3.1.7 Itnable Incore Instrtunentation 14.2.12.80 2AC-310-02 SC 3.1.8 Pressurizer Safety Valve Test 14.2.12.58 2HB-313-02 SC 3.1.9 Pressurizer Performance 14.2.12.54 2HB-212-04 SC (Retests) 3.1.10 Pressurizer Spray Valve and Control 14.2.12.79 2HB-313-01 SC Adjustment
- All tests were performed suberiticsi (SC) .
A O 1.9
TABLE 1.5.1 TEST AND FSAR SECTION CROSS REFERENCE (continued) APPLICABLE TEST % POhiER** SECTION SECTION AND TEST DESCRIPTIO4 FSAR SECTION NUMBER LEVEL 3.1.11 Primary and Secondary Water 14.2.12.76 2HB-213-03 SC G enistry 14.2.12.77 3.1.12 Steau Generator Feedwater Ring 14.2.12.72t 2HB-201-01 SC Integrity (Retests) 3.1.13 Pressurizer Spray Effectiveness 14.2.12.79 2HB-313-01 SC 3.2 Plant Tests 3.2.1 Auxiliary Feedwater Pumps 48 Hour SER* 2ST-235-01 SC Endurance Run - 3.2.2 Power Ascension Data Record 14.2.12.101.3C 2ST-344-17 All 5 3.3 Comnunications Testing 3.3.1 UHF Radio 9.5.2.2.1.6 2AC-483-01 SC 9.5.2.2.3.1 14.2.12.4 3.3.2 Telephone 14.2.12.4.3 2AC-480-01 SC - 3.3.3 Area Radiation Itnitoring Systm None 2ST-340-01 SC k Audibility S 3.4 Containment Isolation Valve 'Ibst 6.2.4.2.5 2HB-101-04 SC 14.2.12.32 4.0 INITIAL CRITICALITY AND CEA EXEPCISES 14.2.10.2 2IC-301-01 SC 4.1 CEA Exercises 14.2.10.2 2IC-301-02 SC 4.2 Dilution to Criticality 14.2.10.2 2IC-301-01 SC 4.3 Withdrawal to Criticality 14.2.10.2 2IC-301-01 10-6g 4.4 Neutron Detector Overlap 14.2.10.2 2IC-301-01 10-6g
*SER is Safety Evaluation Report, NUREG-0712, Supplenent No.1, Section I.6.1
**All tests were performed subcritical except criticality, -
10-6% power. 1.10 O
l TAB E 1.5.1 l f^g 'IEST AND FSAR SECTION CROSS REFERENCE (continued) l U APPLICABM TEST % POWER SECTICE SECTION AND 'IEST DESCRIPTION FSAR SECTION NUMBER LEVEL 10-2%
~
5.0 ILW POWER PHYSICS 'IESTS Table 14.2.5 2LP-333-01 5.1 Zero' Power Biological Shield Survey 14.2.12.81 2LP-701-01 10-2% 5.2 Iow Power Physics Tests 2LP-333-01 I 5.2.1 CEA and Part Iength CEA 14.2.12.82 2LP-333-01 10-2g Symnetry Check 5.2.2 Isothermal Temperature 14.2.12.83 2LP-333-01 10-2g Coefficient Measurement
-2 5.2.3 Shutdown and Regulating CEA 14.2.12.84 2LP-333-01 10 %
Group Worth 5.2.4 Differential Boron Worth 14.2.12.85 2LP-333-01 10-2% Determination i Critical Boron Concentration 10-2g ('s ' 5.2.5 Measurement 14.2.12.86 2LP-333-01 5.2.6 Pseudo DcM and Ejected CEA 14.2.12.87 2LP-333-01 10-2% Worth 5.3 Natural Circulation SER* 2LP-333-02 1 - 3% 6.0 POWER ASCENSION 'IESTS Table 14.2.6 2PA-344-01 2ST-344-17 6.1- Stenary 6.2 Core Power Distribution Tests 6.2.1 Dropped CEA 14.2.12.100 2PA-457-02 50% _ 6.2.2 Pseudo Ejected CEA 14.2.12.99 2PA-457-01 50% 6.3. Variable T average (Moderator 14.2.12.89 2PA-344-08 20% 5 Temperature and Power 2PA-344-09 50% Coefficient Determination) 2PA-344-10 80% 2PA-344-ll 100% 1.11 e a
'IABM 1.5.1 TEST AND FSAR SECTION CFOSS REFERENCE (continued)
APPLICABLE 'IEST % POWER SECTICN SECTION AND TEST DESCRIPTICN FSAR SECTION NUMBER LEVEL 6.4 Unit Ioad Transient 14.2.12.90 2PA-344-02 50% 6.5 Control Systems Checkout 14.2.12.91 6.5.1 SBCS Performance 14.2.12.91 2PA-311-02 ALL* - 6.5.2 SBCS Capacity Checks 14.2.12.91 2PA-311-01 50% 6.5.3 Feedwat er Control System Performance 14.2.12.91 2PA-307-02 ALL 6.5.4 Reactor Regulating Systs Performance 14.2.12.91 2PA-349-01 ALL 6.5.5 Integrated Control System Perfomance 14.2.12.91 2PA-350-01 ALL 6.6 BCS Chemistry and Radiochemistry 14.2.12.92 SPC-002 ALL 6.7 Shutdown Outside the Control Pom 14.2.12.91 2PA-382-01 50% 14.2.12.95 6.8 Steady State Core Performance 14.2.12.101 6.8.1 Nuclear Steam Supply System (NSSS) 14.2.12.101 2ST-344-10 ALL Calorimetric Determination of Reactor Power s 6.8.2 Incore Detector Signal Verification 14.2.12.67 2PA-305-01 ALL 14.2.12.101 6.8.3 Core Performance Record 14.2.12.101 2PA-344-06 ALL 6.8.4 COLSS Power / Flow Verification 14.2.12.101 2PA-313-01 ALL 6.8.5 Nuclear and Thermal Power Calibration 14.2.12.101 2ST-344-09 ALL 6.8.6 Mjustment of COMS Secondary 14.2.12.101 2ST-344-16 ALL Pressure Ioss Term 6.8.7 Linear Power Subchannel Calibration 14.2.12.101 2ST-344-12 20, 50% 6.8.8 BCS Calorimetric Flow Measurements 14.2.12.74 2PA-344-12 80% 14.2.12.101
- ALL means at every major plateau, fa 1.12 S
t t TABLE 1.5.1
'IEST AND PSAR SECTION CIOSS RLFERENCE (continued)
APPLICABLE 'IEST % POhTR SECTION SECTION AND TEST DESCRIPTION FSAR SECTION NUMBER LEVEL , 6.8.9 RCS T Power 14.2.12.101 2ST-344-08 5,10,15,20i i l 6.~8.10. Movable incore Detector Verification- 14.2.12.101 2ST-344-14 ALL i 6.8.11 NSSS Hano Calorimetric 14.2.12.101 2ST-344-19 ALL 6.8.12 Warranty Run 2PA-204-01 1001 '} , 6.9 Biological Snield Survey 14.2.12.97 2LP-701-01' ALL 6.10 Xenon Oscillation Control Test Using 14.2.12.98 2PA-344-05 50% Part length CEAs
-6.11 Intercomparison of PPS, CPCs, and 14.2.12.102 2PA-458-01 'ALL Process Cm.puter Inputs and Outputs at Power 6.11.1 CPC/COLSS Verification 14.2.12.102 2PA-344-03 ALL 6.12 Verification of CPC Power Distribution - 14.2.12.103 Related Constants 6.12.1 Shape Annealing Matrix (SAM) 14.2.12.103 2PA-346-01 20, 50%
Measurement 6.12.2 Tenrerature Decalitration 14.2.12.103 2PA-347-01 501
' Verification 6.12.3 CPC Power Distribution Constants 14.2.12.103 -2PA-351-01 50%
Verification 6.12.4 CPC Thermal Power Decalibration None 2ST-348-01 100t 6.13 Reactor Baseline Vibration *>nitoring 14.2.12.104 2PA-105-01 ALL 4 l ~6.14 Steady State Vibration 14.2.12.72 2PA-102-02 ALL r
.14.2.12.104 6.15 Dynamic Effects 14.2.12.72 2PA-102-03 ALL 14.2.12.104 i
O 1.13
TABLE 1.5.1 IEST AND FSAR SECTION CROSS REFERENCE (continued) h APPLICABLE TEST % POWER SECTION AND TEST DESCRIPTION PSAR SECTION NUMBER LEVEL SECTION 6.16 Secondary Plant Information 14.2.12.26 2PA-611-01 ALL 6.16.1 Turbine Generator Benchmark Input / Output Test 6.16.2 Circulating Water Heat Treatment 14.2.12.16 2PA-299-01 80% 2ST-299-01 20, 50t 6.16.3 Turbine Overspeed Trip Test 14.2.12.26 2PA-602-01 15% 6.16.4 Power Ascension Thermal Expansion 14.2.12.72 2PA-102-01 ALL 6.16.5 HVAC Te:rperature Survey 9.4.1.1.1 2PA-110-01 ALL - 9.4.3.3.1 ;
=
6.16.6 Cable Temperature lenitoring Prcyram ibne 2ST-344-ll ALLl[ 6.16.7 Corriensate, Feedwater and Heater 14.2.12.19 2PA-298-01 601y h Drain Pump Performance E 3 6.17 Transients 6.17.1 Natural Circulation Verification 14.2.12.88 2PA-215-01 801 14.2.12.105 6.17.2 Unit Ioad Rejection 14.2.12.91 2PA-363-01 1001 (1001 Generator Trip) 14.2.12.94 6.17.3 Total Ioss of Offsite Power 14.2.12.96 2PA-381-01 201 (Si:r.ulated Ioss of Onsite SER Post 80i and Offsite Power) 6.17.4 20% Reactor Trip 14.2.12.91 2PA-401-01 20t , 5 LE 1.14 9
V G (V TAILE 1.5.2 SYNOPSIS OF TESTING FUEL IIRD AND POS'ICORE IK7F FUNCPIONALS TEST 4 TITLE DESCRIPTION 2FL-101-01 Initial Fuel Inad (bntrolling (bement for initial fuel loading. 2HB-213-01 Postcore Hot Functional Test Controlling docment for postcore tests. Test prcnided further assurance that the plant and its systems could satisfactorily support nuclear operation. 2HB-213-03 Primary & Secondary Water 2nitored primary aM secoMary water quality.
; Ch mistry 2HB-213-05 - Reactor Coolant Systm Flow Made preliminary measurement of steady-state reactor coolant Measurement system flow for Core Operating Limit Supervimry Systs (COLSS) and CPC's and prcvided data on flow coastdowns for
+
adjusting the Core Protection Calculators (CICs). (A flow u coastdown results when one or acre reactor coolant punps are de-energized.) (Final determination of steady state flow will be made during power range testing in 2PA-344-12.) I 2HB-310-01 Incore Instrmentation Measured core exit thermocouple temperatures during heatup, cooldown and at temperature to ensure proper calibration and response to temperature charges. 2HB-313-01 Pressurizer Spray Valve Adjusted control systems for proper spray & heater actuation. and Control Adjustments ReMjusted the mini-spray flow valves whose small flow keeps the spray line warm. (Mini-sprays needed Mjustment frca the precore hot functional setting because the core's hydraulic resistance caused spray flow to increase.) 2HB-102-01 'Ihermal Expansion Made measur ments on lines which moved due to thermal expansion.
TAIVE 1.5.2 (continued) SYtOPSIS OF TESTItG POS'IOORE IDF IUKTIO4AIS TEST # TITLE DESCRIPTION 2AC-310-02 Movable Incore Acceptance Verified proper operation of the novable incore
%st detector system using a portable control box.
211B-213-06 ICS Ileat Ioss (including Measured the heat leakage from the reactor coolant system. pressurizer) Provided an input to the secondary calorimetric determination of reactor power, anong other uses. Must be done during postcore hot functionals when there is no core decay heat. Containment Isolation valve Measured closure times under flow for containment isola- +[211B-101-04 Timing tion valve 2fN-9334. <n 2AC-483-01 GIF Radio Omnunications tests which verified that comunication devices 2AC-480-01 Telephone inside and outside the plant were operating properly. 2ST-235-01 Auxiliary Feedwater Pump Operated each of the three auxiliary feedwater pumps 48 Hour Run for at least 48 hours. 2ilA-317-01 PPS Instrument Correlation Tested inputs fran instrunent loops to the plant protective Retests system which were not tested during precore hot functionals. O O O
O O O TABLE 1.5.2 -(continued) SYNOPSIS OF TESTING POS'ICORE HOP FUNCTIONAIS TEST # TI'ITE DESCRIPTION 2HB-316-01 GDM Tests Mjusted the operation (voltage, currents, and controller
- settings) of the control element drive mechanisms control ;
system. Adjusted and exercised the control element drive E mechanisms control system for proper rod speed, ensured 5 that reed switches (which sense rod position) operated and gave proper readout in the cmputer and core protec-tion calculator, and denonstrated freedom of motion of '
." control rods by scram time measurements (at cold i G- (320 F) and hot (545 F) conditions) with RCS flow.
'Ihese tests were for single rods only (Multiple rods were tested just before initial criticality).
l 2HB-458-01 Precritical Cmparison of Ensured that values for various parameters were correctly PPS, CPC and Cmputer displayed by the plant ccanputer, the plant protective system (PPS), and the core protection calculator (CPC). I
- 2HB-212-04 Pressurizer Performance Verified pressurizer level and pressure control system
- actuation, readouts, alarms, and operation. E i E 2HB-201-01 Feedring Integrity Test Briefly exposed the feedring in the ' steam generators to a
, steam envirorrnent by lowering water level below the ring, and ,
^
then simulated energency feedwater flow. 'Ihe test was j performed at hot standby conditions and demonstrated the {. ability of the feedring to maintain its integrity.
}
4 i a
TAILE 1.5.2 (continued) SYNOPSIS OF 'IESTIfC IOS'ICORE IKTP FUNCTIO. JS AND INITIAL CRITICALITY TEST # TI'n2 DESCRII'fION 2HB-313-Ji Pressurizer Spray Effectiveness Verified sufficient normal pressurizer spray flow rate by measuring the rate of depressurization. 2HB-313-02 Pressurizer Safety Valve Used a hydroset device to sinrner set the pressurizer safety Setting valves under hot standby conditions. Although this test n was successfully perfonned during precore hot functionals, . m the valves were subsequently adjusted, requiring another 2 setting. S 2ST-340-01 Area Radiation Monitoring Verified under cperating conditions, that the audible alarms System Audibility for area radiation nonitoring systens could be heard in the areas which they were monitoring. INITIAL CRITICALITY 2IC-301-01 Initial Criticality Monitored count rate changes during CEA novement and durfngdilutiontocritical. Diluted to criticality at 325 P and 600 psia. Criticality was attained by having nost of the control rods empletely withdrawn frm the core, slowly reducing the soluble boron concentration of the RCS (i.e., dilution) by injecting demineralized water into the volume control tank, then withdrawing the renaining control rods.
'Iwo of the four reactor coolant Iunps were operating.
Iow temperature initial criticality was performed on Unit 2 since it is the first of the 3410 megawatt C-E plants. (Initial criticality for SotGS 3 and all other ogration for both units will be at normal hot standby conditions of 545 F and 2250 psia.) O O O
I O O l TABIE 1.5.2 (continued) SYNOPSIS OF TESTING INITIAL CRITICALITY AND IIM POWER PilYSICS TESTING TEST # TITLE DESCRIPTION 2IC-301-02 CEA Exercises Withdrew an3 inserted CEAs in groups for the first time and aljusted group movement as necessary. Ehsured the various conputer-generated interlocks such as out-of-sequence ard Eower-dependent insertion limits operated properly. Performed the first multired trips (scrams) . IIM POWER P!IYSICS TESTING 2LP-333-01 Iow Power Physics Testing Occurred after initial criticality and prior to power operation. Reactor was operated at low ( 0.1%) power. Provides physics information on the as-built reactor which verified I ysics h design paraneters and demonstrated conform-
." ance to aggus iate Technical Specifications, p ific 5 tests conducted at a reactor temperature of 325 P incitrial: (bntrol Elenent Assembly (CEA) Coupling Checks; Regulatirg CEA Wrths; Boron Wrth; Critical Boron Concentra-tion; Tenperature Coefficient Measurement; g and Rgactivity Computer 01eckout. 'Ihe heatup fran 325 F to 545 F incitdal measuranents of the presgure coefficient and tanperatureg coefficients. At 545 F, the tests described for the 325 F plateau were reperfonned arvi also incitried shutdown, ejected and dropped rod worths.
TABLE 1.5.2 (continued) SYtOPSIS OF 'IESTItC LOW PNER HlYSICS TESTItU AND 10hER IWEE 'IESTItE
'IEST # TITLE DESCRTPTIW 2LP-333-02 Natural Circulation Denonstration of natural circulation for operator Denonstration training. We reactor was critical at about 3% of full power, the reactor coolant punps were de-energized and '
natural circulation, driven by heat frcm the critical reactor and cooling through the steam generators, was E allowal to develop. l3 ~ - Natural Circulation at With natural circulation in progress, reactor pressure was !$ Reduced IKE Pressures reduced by de-energizing the pressurizer heaters. %e capability to establish and maintain natural circulation and zdequate margin to saturation without pressurizer heaters was denonstrated. Operation of the subccoled margin monitor was demonstrated. Operator trainirg was provided. Reactor power was approximately 1% to 2%. Natural Circulation Natural circulation with one steam generator isolated With Reduced lieat (the other renoving core heat) was shown to be sufficient Removal Capacity to cool the core. %e return to two loop natural circulation was denonstrated when the isolated generator was returned to service. Operator training was provided. POhER RANGE TESTItG 2IP-701-01 Biological Shield @tained radiation readirgs at various locations inside Effectiveness Survey the containment. Ongoing test through 100% power. n 2PA-102-01 Ibwer Ascension %ermal Visually inspected and neasured the positions of piping - Expansion as the plant was brought to greater power levels. g a 2PA-102-02 Steady State Vibration Determined piping and equiptent vibration cri safety cnd -safety related cunponents at various power levels and rating conditions.
o O O. 4 i NIE 1.5.2 (continued) SYNOISIS OF TI: STING POWER IWKE TESTING TEST S TITLE DESCRII'I' ION 2PA-102-03 Dynamic Effects Test Collected data during transients such as trips for , evaluation of pipe acceleration arxl component ~ movement. i 2PA-105-01 leactor Baseline Vibration At all operating plateaus, the reactor vessel core barrel Monitoring vibration data were taken using the loose Parts Monitoring System. 2PA-110-01 INAC Temperature Survey Verified proper cperation of the heating, venti 11ating, and air conditioning (ilVAC) systems. "~ L 2PA-204-01 100% Warranty Run Satisfied warranty and contractual obligations that the ! plant and nuclear steam supply system (NSSS) can operate at 1001 power for a substantial time (200 hours).
- E SPC-002 HCS Gesictry and Measured chemical concentrations and radioactivity in 5 j Radiochemistry Test primary coolant and followed trends which were indicative of the condition of the fuel. (Radiochemistry was monitored essentially continually.) - ;
I 2PA-215-01 Natural Circulation Simultaneously t. ripped all four reactor coolant pumps from 80% Verification reactor power. CPCs then tripped the reactor. Demonstrated attaining natural circulation with substantial stored and ljg decay heat. Maintained hot standby conditions and demon-strated boron mixing. Performed natural circulation cool-
- down to maximtzn shutdown cooling systen entry conditions
(<400 F,<376 psia) using safety related equipment, thereby simulating a cooldown under loss of offsite power corxlitions. Operated atmosF h eric dtmp valves manually. Satisfied NRC Branch Technical Position RSB 5-1 test requirements. , j t i t i .
TABLE 1.5.2 (continunl) SYTOISIS OF TFSrirO IOWER HAtMB TERPITC TITIE DESCRIPTION TEST # Condensate, Fealwater anl thnanstratel that the fen 1 water anr1 conlensate systen 2PA-298-01 lleater Drain Ptanp is capable of supplyinj feelwater to the stean Performance generators for nonnat operation. 2PA-299-01 Circulating Water ileat Danonstratai nethod of an1 procedure for controlling y Treatment Test nurine growth on the circulating water intake structure , by introducing heatal discharge water into the intake. ; 2PA-305-01 Incore Detector Signal Verifiel that the plant conputer receives proper signals Verification fran the incore detector amplifiers. Incore detectors determined the core power shape anl were used to calibrate
- a the COLSS anr1 CPCs.
ra 2PA-307-02 Feedwater Regulatin3 System Gatheral data anr1 observed the operation of the feed-Performance water regulating systen to optimize its operation. h
p p p V V V TABLE 1.5.2 (continued) SYNOPSIS OF TESTItU PERFDINPD DURItG IOWER IWEE TESTING TEST # TITIE DESCRIPI' ION 2PA-311-01 SBCS Ca[ucity Checks Measured atmospheric danp and SBCS valve flow capacity. 2PA-311-02 Stean Bypass Control Evaluated the performance of th'e stean bypass control System Perfonnance systen and made aljustments as necessary to optimize its performance. 2PA-313-01 CDISS Power and Flow Verified that algorithms in the COISS conputer progran cor-Verification rectly calculated reactor coolant systen flow and core tower. 2PA-344-01 Power Ascension Test Controlled power range testing. Progran m 2PA-344-02 Unit Inad Transient Test Changed reactor power in ranps and steps to demonstrate d
,- the ability of the plant to change power (maneuver) in a g m
g controlled fashion. 2PA-344-03 CIC/COISS Verification Verified that the calculations in CPCs and COLSS properly compute Departure fran Nucleate Boiling Ratio (DNBR) and 2 Incal Power Density (LPD). Evaluated the acceptability of & process and instranent noise levels on the calculations. 2PA-344-05 Xenon Oscillation Control Demonstrated that normal Xenon oscillations can be Tests - PIfEAS controlled using part length control element assenblies (PICEAs) . 2PA-344-06 Core Performance Record Denonstrated that steady-state core power distributions agreed with design predictions.
TABIE 1.5.2 (continued) SYNOPSIS OF TESTING PERF0fiMED [11 RING 10WER HANGE 'IESTI?G TEST # TITIE DESCRIPI' ION Variable T-Average Test obtained as-built values for moderator temperature and 2PA-344-08 (20% Power) power coefficients (core physics data) at the power 2PA-344-09 (50% Power) levels shown. 2PA-344-10 (80% Power) 2PA-344-11 (100% Power) 2PA-344-12 ICS Calorinetric Flow MMe final determination of reactor coolant system flow by Measurement neasuring core power frm secondary calorimetric and calculating RCS flow required to produce the measured reactor vessel differential temperatures. 2PA-346-01 Shape Annealinj Matrix and Provided data for adjusting the CICs so that the core ,_. Ibundary Condition Measurement power shape is properly calculated. m Ja $ 2PA-347-01 TemFerature Decalibration Provided data for assessing the decalibration of excore Verification neutron detector signals with changes in cold leg temp- j erature and adjustinj CPCs as necessary. 2PA-349-01 Ibactor Ibgulating System Observed and Mjusted the Reactor Regulating System. Performance 2PA-350-01 Integrated Control System @ served the plant's operation with all control systems Performance related to the Nuclear Steam Supply System (NSSS) in autmatic. 2PA-351-01 CPC Power Distribution Verified the radial peaking factors and CEA shadowing Constants Verification factors in the Core Protection Calculators at 50% rower and made Mjustments as necessary. O O O
TABIE 1.5.2 (continued) SYNOPSIS OF TESTING PERFORMED OURING IOWER RANT TESTING TEST # TITLE DESCRIPTION 2PA-381-01 Ioss of offsite Power Demonstrated that the plant tripped, renained controlled, (Simulated loss of Onsite and and was supplied by energency diesel power when all offsite . Offsite Power) power was lost. Selected onsite loads powered by energency diesel generators were also deenergized. 2PA-382-01 Shutdown fran Outside Demonstrated that the plant can be shut down, controlled (bntrol Ibom with 50% at hot standby conditions and cooled in a controlled man-Generator Inad ner fran outside the control roam.
- 2PA-383-01 100% Generator Trip /Ioad Demonstrated that the plant can sustain a full loM rejec-Rejection tion and can be shut down in a controlled manner. 'Ihis was
- a CESir (02nbustion Engineering Systens Excursion (bde)
L verification test.* cv 2PA-401-01 20% Main Control' Board Demonstrated proper operation of the plant subsequent to a ; (MCB) Reactor Trip reactor trip. First scheduls3 trip'of the power range test g , PDJran. 3 2PA-430-02 Main Generator Off-Line/ Mjusted the Main Generator's autanatic voltaje regulator Rated Speed Excitation Test with the generator at 1800 rpn but not synchronized. 2PA-430-03 Main Generator On-Line 01ecked response of voltage re3ulator at every power Excitation Test plateau. 2PA-457-01 Psuedo CEA Ejection Test Provided data for analysis of ejected CEA and comparison with physics predictions and CESDC verification program.
*1CPE: A generator trip always causes a turbine trip. Ibwever, a loss of , generator load does not produce a turbine trip unless load is above 55% power (or at lower powers if less than four SBCS valves are in service). A loss of load at full power (100% Generator Trip /Ioa3 Re-jection) is' identical to the 100% Turbine Trip. ; A CESEC verification test is one in which data was gathered to verify the Oxnbustion Engineering Systems Excursion Oode (CESEC) .
TABIE 1.5.2 (continued) SYNOPSIS OF TESTING PERFORMED OURItC IOWER PANGE 'IESTING TEST # TITLE DESCRIPTION 2PA-457-02 Dropped CEA Test Provided data for waluating radial power distributions and for CESEC verification af ter scrungning a single rod at power. 2PA-458-01 Process Variable Verified instrument readings at all power plateaus by Interconparison interemparison. 2PA-602-01 'Ibrbine Overspeed Trip Provided verification of the operability of a turbine Test protective device, the turbine overspeed trip. 2PA-611-01 Benchmark Turbine / Controlling document for the program by Q11ch data was " ,~ Generator Input / Output taken and the turbine was brought to full power. - g Test g 5 2ST-299-01 Special Biofouling Renoval Test Performed a heat treatnent of the circulating water intake and discharge conduits similar to 2PA-299-01, only at 50% power rather than 80%. 2SP-344-08 BCS A T Power Primary calorimetric determination of reactor power level. Used RG coolant flow rate and tenperatures in the tot and cold legs. 2SP-344-09 Nuclear and Thennal Adjusted CPC and PPS Fuer to match calorinetric determination. Power Calibration 2ST-344-10 NSSS Calorinetric Detennined core thermal power and adjusted ODIES power as required. 2ST-344-ll Cable Temperature Fkxiitoring Measured temperatures in wrtain safety-related cable Progran trays with the cables under load. O O O
7 (~ ,. TAIME 1.5.2 (continued)' SYNOPSIS OF 'IESTING PERIVINED [11 RING IOWER IWNGE 'IESTING l TEST 4 TITLE DESCRIPTION 2ST-344-12 Linear Power Subchannel Adjusted the anrplifiers which supply excore detector Calibrations signals to the core protection calculators. 2ST-344-14 Movable Incore Detector Verified the different modes of operation of the movable Checks incore detector drive system. 2ST-344-16 . Mjustment of 00ISS Secondary Determined pressure loss in main steam and feedwater lines as
.~ Pressure Ioss Term a function of steam flow and reactor power. Adjusted the y appropriate CDISS constants.
- 2ST-344 Power Ascension Data Record Recorded steady state and transient plant conditions for E analysis and trend analysis, a 2ST-344-19' NSSS Hand Calorimetric Plant data taken by hand was used to calculate core power.
2ST-348-01 CPC 'Ihermal Power Decalibration Verified that the thermal power indication cbtained frce the Core Protection Calculators (CPCs) was. not significantly affected by Control Element Assembly insertion.
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2.0 INITIAL FUEL IDADING (2FL-101-01) Sunmary % e inital fuel loading at San Onofre Nuclear Generating Station Unit Two was performed as specified in procedure 2FL-101-01 to satisfy FSAR requirement l "j 14.2.10.1. We procedure provided an organized and orderly method of loading i the core and safeguards to monitor and detect any unexpected reactivity addi- J tions. %e first of two hundred seventeen fuel bundles was loaded on Friday, February 19, 1982, two days after receipt of a low power operating license frcm the Nuclear Regulatory Cmmission. We last bundle was loaded approximately one week later. Although bundles could be loaded as rapidly as two per hour, an average tine of forty-seven minutes per bundle resulted frm equipnent failure, meal breaks, cleaning the refueling machine, and renoval of foreign objects frm the reactor vessel. A slight anount of inleakage frm the refueling water storage tanks to the reactor coolant system produced a very slight dilution; otherwise, there were no unexpected reactivity additions. We loading was " dry" " with low water levels in the refueling cavity and reactor vessel, which aided in ; detection of the inleakage and removal of the foreign objects. We core was & verified to be properly loaded and the procedure was cm.plete March 1,1982. 3 The remainder of this section will deal with the equipnent, overall novement of fuel, a history of the load and interruptions, administration and staffing, reactivity nonitoring, and loading verification. Dry Fueling O Initial fuel loading was dry; that is, the refueling cavity and reactor vessel were only partially filled with borated water so the reactor vessel flange was above water. We dry loa: ling was intended to allow observers to closely nonitor the fuel as it entered the core and seated on pins projecting frm the core support plate, and to minimize repair time in case of a malfunction of the refueling rachine. Because new fuel is only slightly radioactive, personnel could be continmusly stationed on the vessel flange except wnen the tw neutron sources were transported in fuel bundles to the reactor vessel. , Borated water was required in the refueling cavity to provide containment isolation and was used to lubricate the bearings of the fuel transfer tube's i equipnent. Water level was maintained above the fuel transfer tube and S provided a water seal which isolated the containment atnosphere as directed by h chnical Specifications. We reactor vessel was filled to just below the vessel rim with borated water which was circulated by the shutdown cooling system. We concentration of boron was higher than that required by Technical Specifications (>l720 ppn), which guaranteed the core would remain subcritical. 2.2 O
Equipnent and Fuel Movement l O We equipment used and the movement of fuel are described in Figures 2.0.1 and 2.0.2. W e side view shows the low water level in the refueling cavity and reactor vessel. Befueling cavity level was just above the fuel transfer tube and reactor vessel level was just above the hot leg. Observers were stationed as shown to observe bundles as they entered the vessel and to ensure they did not interfere with or danage those already loaded. Observers also insured proper seating of bundles on support pins by insuring the tops of the bundles were on the same elevation and noting air bubbles rising through the water as
~a ir in the lower end fittings of the bundles was displaced by the pins. Under-water lights aided the observers.
% e upenders are shown in the vertical position by broken lines and are required to place the tuel horizontally so it will pass through the fuel transfer tube.
Counters were connected to temporary detectors located inside the reactor ~ vessel and the permanent plant startup detectors were connected to counting equipnent in the control roan. Both sets monitored neutron count rate and from 4 this, the sub-critical neutron multiplication of the core. Speakers provided E audible indication of neutron count rate to observers on the flange as .well as m other workers in containment. New fuel was stored in air in the spent fuel building in both the spent and new fuel storage racks because the new fuel racks lacked capacity for an p entire new core. Fuel from the new fuel racks was transported to the spent fuel upender by the new fuel crane and~new fuel elevator. Fuel fran the spent j fuel racks was handled by the spent fuel machine. Once in the upender, the fuel g was lowered to horizontal, transported on a dolly through the fuel transfer a tube to the containment upender and returned to vertical. We refueling machine was positioned over the upender, grappled the fuel bundle, transported it to a precise location over the reactor vessel, and lowered it into the core. Fuel assemblies designated to have 5-finger CEA's were transferred and loaded with the CEA's inserted in them. Personnel exiting the refueling cavity were monitored for contamination by ; Health Physics personnel. Health Physics also monitored containment for g airborn contamination. Electronics for the temporary neutron detectors in the j reactor vessel were located at the physics counting station. In order to maintain clean conditions, shoe covers, paper coveralls, cotton gloves, and head covers were required around and in the refueling cavity. 4
\ 2.3
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0. L-J FIGURE 2.0.2. PLAN VIEW OF EQUIPMENT USED FOR FUR LOAD 2.4
History of the Ina31ng A history of core loading includes the order in which bundles were loMed, the number of bundles loaled as time progressed, and problens. Figure 2.0.3 shows the order in which the bundles were loa 3ed. %e first bundle, which contained one of the two neutron sources, was loaded in core location B-11 aM the rest of the core was loaded around it. Tenparary neutron detectors were placed at locations D-15 and D-7 before the first bundle was loaded and in conjunction with Startup Channel 12 provided the bulk of neutron nonitoring for the first 132 bundles. ne dumny fuel bundle containing the first neutron source was positioned near both temporary and startup detectors to provide a respnse check prior to loa $ing fuel. A background count rate with ro fuel or neutron sources in the core was also taken. %e neutron source was then loaied into the first bundle. The first bundle stood alone and unsupported until the secoM aM third bundles were loaded in locations A-12 and A-10. We additional bundles also placed fuel between the source and startup detector- and provided the detector with a strong supply of neutrons. We core was then loa 3ed around the source until it was nine bundles wide (fron coltuuns 5 to 17) and starting with number forty-eight (location G-5) was m extended east toward Startup Channel #1. A space was provided for tenporary
) detector B on the east side of the core by not loa 31rg.a bundle in location V-7; the location was bypassed and other bundles were loaded around it. After location X-ll was loaded with a bundle and murce the detector was relocated fran D-7 to V-7. The core was then nonitored by one permanent (startup) and one temprary detector on the west and east sides of the core.
Se discovery of a fragment of a rag and a piece of tape in locations S-5 aM W-15 caused the order of loading to be changed slightly. We procelure allowed the loading sequence (but not orientation or location of the bundles) to be changed. When the foreign objects were noted by the midnight crew the order was changed so loading could continue until further managenent assessment and correction could be made during the day. In order to not disturb the rag and tape, loadirg of positions ' S-5 and T-5 was deferred until after V-15 was loaded and the objects were renoved. We original sequence rest ned with location V-13. A long pole was assenbled fran a tool which had been intended to assenble the four-finger control elenent assenblies and was used to remove the tape and rag. One end was wrapped with approved tape (with the sticky side out), lowered into the vessel over the object, and pressed gently against each. We pole was then slowly and carefully withdrawn so the objects did not separate
. fran the tape.
O h 2.5
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FIGURE 2.0.3: LOADING SEQUENCE 2.6
The south side of the core was then loaded. A second and final rag fragment i was discovered and removed frm core location R-19. .W e bundles at locations T-20 and S-20 suffered slight (50 pounds or less) interference about halfway inserted, indicating that spacer grids were rubbing. Each bundle was slightly n withdrawn and successfully reinserted after the refueling machine's mast was . gently rocked. Se north side of the core was easily loaded. 2 S Finally, each of the temporary detectors was renoved and a bundle loaded in their place. Temporary detector B was removed frm location V-7 and a bundle loaded, then tentorary detector A was removed frcm D-15 and the final bundle loaded. The actual loading sequence is shown in Figure 2.0.3. Temporary detector A was located in D-15 until the last bundled was loaded; temporary detector B was removed frm D-7 to V-7 between steps 132 and 133, after the bundle with the second source was loaded in X-ll. R-1 and R-2, located in S-5 and R-19, respectively, were the approximate locations where rag fragments #1 and 2 were found. "T", shown in location W-15, was the location of the tape. "I" at T-20 and S-20 are the two core locations where sme minor grid interference to loading was encountered. "*" at G-5 was a bundle which leaned to the north approximately 3/8" when first seated but presented no interference to other bundles. "S" in locations X-ll and B-11 are neutron sources. 2.7
Progress of Ica3iry Progress of core loadiry incitdes the number of bundles in the reactor vessel with time, the nunber of bundles loaded by each shift, and interruptions. We number of bundles loaded is shown in Figure 2.0.4, which enconpasses the time fran the first at 1:22 pn on Friday, February 19, 1982, to the last at 2:14 pn the following Friday. Table 2.0.1 depicts the loa 3 by shift and day, and shows that roughly one third of the core was loaded by each shift with the midnight shift loadirg the nost. Interruptions Interruptions are also nated in Figure 2.0.4 anci Table 2.0.1. 'Ihe most regular was a meal break which occurred on each shift and consumed approximately eighty minutes. Maintenance, cleaning, and other work was often scheduled for the breaks. mise Bursts of noise plagued the permanently installed Startup Gannels, especially Gannel 42. 'Ihe Startup Gannels monitor the low nunber of neutrons present during fuel loadirg and when the reactor is rot operating. mise is electrical interference fran a nunber of saurces. It affects amplifiers and other canponents in the neutron nonitorirg circuits. Noise can overwhelm the true neutron signal of approximately one neutrcn count per second in brief bursts which appear as hundreds or thousards of counts. A steadily risirg asunt rate (counts per secon3) can indicate additions of reactivity which are Lm-portant to nonitor, especially during fuel load. Noise produces annoying but easily distinguishable increases in count rate which makes monitoring the core nore difficult. 'Ihe source and characteristics of raise were investi-gated during the first two days of core loadirg and its effect scmewhat reduced although roise persisted to scne extent throughout the loa 3ing. l Festoonity The refuelirg machine is supplied with power, control air and ccmnunications through cables which are supported by festooning as illustrated in Figure 2.0.1. Festoonirg consists of anali dollies which travel inside tracks made fran Unistrut. Bearirgs in the dolly wheels were insufficiently lubricated ard the dollies janmed in the track. Once jammed, the cables or their supports are highly susceptable to danage as the refueliry machine continues to nove. Danage occurred as the machine nxned against janmed dollies and bent sepports. Replacement dollies were also insufficiently lubricated. A generous application of nuclear-grade Ne/ersee7.e was applied to the wheels, bearings, and inside surfaces of the Unistrut on Tuesday mornirg, after which festconity ceased to be a concern. O 2.8
6*Z EUMBER OF FUEL BUNDLES LOADED o m N u > Z o o o o V I n f.._.i__..- . . . _ . . .. ~._. .._. .. . ... 7 _. 4 . _ ..... .2..-_ . _ _ . ... . ... ... u& -a e !
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Figure 2.0.4 Number of Bundles Loaded Versus Time (continued) e \ s, -
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- Friday, 2-26-82 2.12
~ - - _ _ _~ _ _
__ _ _ ~. -- 1 O O O l TABIE 2.0.1: BUNDLE LOADING TIME DISTRIBUTION Day's Bundles Loaded Bundles Per Shift ** Day Date Number Pe rcen t
- Midnight Day __ Swing Comments and Problems l
Friday 2-19-82 8 3.7% 0 1 5 Started on day shift. Cleaned refueling machine. Brief failure 'of communication. Containment side upender's " vertical" light failed. 12 15 Communication failure, noise on scalar-timer in the Sat' day 2-20-82 37 17.1% 12 control room. Sunday 2-21-82 32 14.7% 11 10 11 Communication failure, checked refueling machine's alignment, cable festooning on refueling machine jam.ned. l Monday 2-22-82 20 9.2% 10 9 1 Replaced damaged cable festooning parts. Replaced a l relay in the containment upender's " vertical" light circuit (no further light problems). l .N C -7 12 Discovered and removed portion of a rag and a piece Tuesday 2-23-82 33 15.2% 12 of tape from the core support plate. Wed. 2-24-82 26 12.0% 9 5 12 Moved temporary detectors per procedure, inspected refueling machine, found and removed a small piece of a rag from core support plate (no further rags). Thurs. 2-25-82 36 16.6% 10 12 14 Leaking valve from refueling water storage tank caused slight 1cvel increase in reactor vessel and approximately 25 ppm dilution. Friday 2-26-82 25 11.5% 14 13 0 Final bundle loaded on day shift. I Total 217 100.0% 78 69 70 Midnight shift loaded most bundles. l Percent of Total 36% 32% 32% Shifts loaded almost same percentage. Notes:
- Percent is percent of 217 bundles.
** Bundles per shift was based on shift rotation of Days: 7am to 3 pm, Swings: 3pm to 11pm, and Midnights: llpm to 7am, which does not coincide with calendar days starting at midnight. The discrepancy between the number and percent results from bundles being loaded by a shift which does not begin or end at midnight.
Cminunication te test procedure required continuous cmmunication anong personnel at the contairunent upender, refueling machine, physics station, vessel flange, in the control rom, and at the spent fuel upender, spent fuel machine, and spent fuel crane. We first of two systems, the permanent plant systen, is provided with jacks into which headsets can be plugged but was overloaded by the large ntznber of headsets. A tenporary system was connected through a spare contairinent penetration and was used almost exclusively. Several shorts and an anplifier failure interrupted loading. An electrician was stationed to monitor and maintain the tenporary systen and cmmunications worked well af ter Sunday. Containment Upender he containment uperxler is equipped with logic which prevents damage to the upender itself or the refueling machine if the upender is not vertical when the refueling machine is moved over it. A relay provides the logic as well as power to a light which signals the upender is vertical (the upender " vertical" light) . We relay failed and time was lost on mnday to identify and replace it. We vertical light had failed several times between the first bundle and the relay's failure, indicating a slow but sure deterioration of the relay with use. After repair the upender worked well. Valve teakage Valves which isolate the refueling water stcrage tank fran the low pressure safety injection system (wnich was also used for shutdown coolirg) leaked and caused dilute toric acid in the tanks to leak into the systen. Shutdown - cooling was beirg supplied to the reactor so the inleakage manifested itself as an increase in water level and decrease in boric acid concentration. W e l q. dilution from approximately 1935 to 1910 ppn was the only unexpected reactivity m? addition of the loading. Wrotshout the loadirg, boron concentration was well above the Technical Specification minimum of 1720 ppu. t 2.14
p Mministration and Staffing , t
'- Fuel Icad was directed frm the control rom where the procedure, fuel status boards, startup channel scalar-timers, and a reactor operator and startup engineer were located. (bntrol rom personnel were in constant ccmnunication with personnel in the containment artl spent fuel buildings. As fuel was noved frm its storage racks to cranes, machines, upenders, and the core, the control rom was informed of movement and location. %e precise location of the spent fuel and refueling machine was relayed to and verified by the control roan to insure fuel was being loaded in the correct position. (bntrol rocm status board was continuously ugiated to reflect the current location of all cmponents involved in the loading.
Staffing consisted of Westinghouse Electric Corporation personnel who operated the equipnent, Southern California Edison personnel who observed and aided the equipment operators, Cmbustion Engineering startup personnel who observed and E provided support to Southern California Edison personnel, a senior reactor 5 operator in the containment who observed and directed novement, a reactor operator and startup engineer in the control rom, and health physics and security personnel. Cleanliness control was accmplished by requiring personnel entering the vicinity of the reactor cavity to wear rubber tooties, cloth coveralls, gloves, head covers, and installing a sheet of plastic as shown cn Figure 2.0.3 frm the floor to the first bar of railing. We refueling machine was vacuumed to prevent dirt frcm falling into the vessel. Quality control personnel also /'~'s closely monitored cleanliness. ( ) Although there was essentially no radiation or contanination during the load, health physics personnel issued dosimeters, maintained a station with geiger counters and access control, and required personnel nonitoring. %e health physics station is shown on Figure 2.0.2. Access to the refueling cavity itself was controlled by guards stationed at the " access" control point shcun cn Figure 2.0.2 Entry into the cavity as well as the area inside the control point required special clothing to prevent entry of dirt and the removal of all loose objects such as pens, keys, etc. Access to the contairstent was limited by guards who insured that only authorized personnel entered. Entry to the spent fuel building was also controlled by guards and monitored by health physics personnel. (~ V) 2.15
Reactivity Monitoring Reactivity monitoring was accmplished with permanent startup detectors located outside the reactor vessel and temporary neutron detectors in the core. % e instru - ments detect a portion of the neutrons that enter the detector and produced a count . rate, the number of neutrons detected per unit time (normally expressed in E counts per second). Ganges in the reactivity of the core are reflected in E changes in munt rate. Increased reactivity produces higher count rates and can result frm the addition of fuel or decrease in boron concentration. Count rate charges due to fuel movement are expected and mst large changes are due to placing fuel close to a detector. Unexpected changes almost always result frm boron concentration changes. Even though the boron concentration was sampled twice per shift and although indications of a potential change such as increasing water level in the reactor vessel were monitored, count rates were also used to monitor reactivity changes. We minimum boron mncentration which is specified by the h chnical Specifications is sufficiently high that the core will remain suberitical even with all control rods fully withdrawn. We cmcentration in the reactor vessel and refueling canal was always above the minimum value. Changes in reactivity were specifically monitored by " Inverse Multiplication Plots", or, simply, IAi (one-over-m) plots, as shown in Figures 2.0.5 and 2.0.6. A base munt rate was obtained af ter the first bundle (with its neutron source) had been loaded and the count rates frm succeeding bundles were divided
- into the base count rate. We ratios which are produced are then plotted for each bundle and an extrapolation is made to the axis. If the line intersects the axis before the next bundle for more than one detector, fuel load was to have stopped. Although changes in the wre which resulted frm placing bundles near detectors caused one detector to extrapolate to the axis, there were no bundles which produced such extrapolations on two or m re channels.
Figure 2.0.5 shows a portion of the inverse multiplication plots for the detec-tors which read above background. Only the first twenty-three bundles are shown because the remaining bunales did little to cnange the count rate. Figure ,, 2.0.6 shows the background-corrected count rate for the temporary detectors qq and Startup Channel #2. No unexpectea behavior was observed in any of the gg 1/M data, mm Ioading Verification i Ioading verification insured that the fuel ard mntrol element assemblies were located and oriented in the reactor vessel as designed, and were properly aligned so the upper guide structure would not damage them when it was lowered. Orienta-tion and location verification was performed by a television camera suspended frm the refuelirs machine. A video tape was made of the inspection. Incation was " determined by reading the serial numbers stamped on one corner of the guide J tube's upper end fitting in tne case of fuel bundles, and on the webs (support- E ing members) in the case of control element assemblies. Each bundle's serial 5 number was in the northeast corner for proper orientation. 2.16
( Alignment was verified with the television camera mounted on the refueling machine. The refueling machine traversed two rows, F and S, and two coltunns, 6 and 16, as shown in Figure 2.0.7 and .the machine's coordinates were noted with the camera centered on the center control element assembly guide tube. Althotgh coltsun 6 and row S exceeded the quarter inch deviation allowed by the procedure, the measurenent was reviewed and found to be acceptable. Figures 2.07 and 2.08 show the location of fuel bundles and control element assenblies,- respectively, as loaded. Note that the four finger CEA's were loaded with the upper guide structure and are not shown on Figure 2.0.8. O i O 2.17 l l l 1
- 1. 0 - .
.-.-_....,._.y.
f, t .p r e .3. 0. ,.3: Inverse Jxultiplication
. A. $ tarts . ' .annel .J <,,. 7.
.+ E Plot.
- g. Temporars. - B . .... - * *
- r.
_ . . . . . . _ . . . . _ . . . . . . .........n.
. _.-... . . G.. Temporary A. . . . . -
- 0. 9 - .
g .., . .. .. . e ,...... . _.y v..
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. _ ... a ; _ ,
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0.8~ L
.- . . . _ - . ...t S a rtup- Chemne l 42 ~+- -
. . . _ , ._.._-,.2.,..-...
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__ .q.___.-._,.,.._.. . - , . . 7...-. The core loadinb' se9uence below
. . _ . . - .4.....
7
.__ . . . aL-._ . ~ _ . . .. +_. . - - . - ..
. . . . + . .
corresponds to the plot to the Icft.
- .n . . _ .
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.L .,..-.-.,.-e._
. . . .i. . 1 . . _. 4 -- -.4..--+.4-o
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to Ws 0.6 g . . . . . . _ . -- . _ . , .
. ,. . . . . _ . . . , et g3 it y' n
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.--. . . , _ _ - ,.q ..--p.. .
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._....-q._._.T~~ Note how the 1/M pior. falls (indicating
, i 0.3 [ __
. .. [__ ,. _._ 4p._._,.,.._._
count rate increases) when fuel bundles L ._ _-.- _;.-._.u...m.____.,.,-.
-s.. -1..-.. . . . .
are loaded between the source in B-11
.e..p
,g.
.___..2..__-.a.
.-~..".._..
(indicated by "s") and the various detectors.
- n. .-_.. . . . p. -
.. . . _ . - .q . . . . . , . . . _ . . ._; . The number of fuel bundles loaded 02 _._...M"o M.'/ ... ..- . . . . . and the loading sequence are identical.
. .. . . Temporary.B . . ,. .
p... . . __. .
. . . . . _ . - . - . . _ . , . + .
1 . . . . . _ . _ . - . . .
.u_..__.- '
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.A. -
;g. ... . .
_.: . L Noos 46; . 0.0d ! !. mi uu f. 1 5 10 15 20 25 Number of Fuel Bundles Loaded 2.18
Figure 2.0.6: COUNT RATE DATA C'l The following is the count rate data for the first twenty three bundles. Number of Temporary Detectors Bundles Startup #2 A B , Loaded CPS 1/M CPS 1/M CPS 1/M Comments 0 0.075 - 0.03 - 0.03 - Background count rate. 1 0.495 1.0 4.03 1.0 10.26 1.0 First bundle with source. 2 0.94 0.527 4.51 0.894 11.50 0.892 Large increases in permanent detector are due to fuel 3 1.25 0.396 4.69 0.859 11.56 0.888 being loaded between source and detector. 4 1.11 0.446 4.98 0.809 16.08 0.638 5 1.098 0.451 5.33 0.756 43.47 0.236 Large increase in Temporary B count rate is due to fuel 6 1.128 0.439 6.74 0.598 50.92 0.201 being loaded near it. 7 0.978 0.506 16.62 0.242 51.79 0.198 8 1.12 0.442 26.75 0.151 52.59 0.195 f) m 9 1.02 0.485 27.90 0.144 52.54 0.195 10 1.17 0.423 31.02 0.130 53.24 0.193 11 1.45 0.341 69.13 0.058 52.98 0.194 12 1.21 0.409 104.11 0.039 53.13 0.193 , 13 1.77 0.280 112.23 0.036 55.41 0.185 14 1.478 0.335 115.59 0.035 109.02 0.094 15 1.178 0.420 ~ 115.54 0.035 180.07 0.057 16 0.675 0.733 115.84 0.035 214.71 0.048 17 1.10 0.45 119.32 0.034 240.49 0.043 18 0.922 0.537 118.66 0.034 245.82 0.042 19 1.40 0.354 119.80 0.034 260.92 0.039 Count rate for the bundles after #23 remained fairly 20- 1.'268 0.390 285.60 0.014 278.19 0.037 constant. 21 0.692 0.715 287.43 0.014 278.44 0.037 O() 22 0.865 0.572 284.15 0.014 292.91 0.035 23 1.092 0.453 286.52 0.014 292.42 0.035 2.19
Figure 2.0.7: CORE LOADING MAP FUEL BUNDLE LOCATION COLD LEG COLD LEG 2B EAST 1A 8 10 12 14 I !l I 2 3 4 5 6 ~'T 9 -I!--- - 13DI Cy 7 18 l9N20 21
- ,- y s 3
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NORTH i ( 111 Ad &,Il Acd CeG Ac31 Es1 WL 6ct1 /sk SclL Lo s,r5 Ace Clcz
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\ \ ,c . . . ic .,
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.. n A
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Cofi c.at
+ , .
CM m.a + , . . Ccel COLD LEG COLD LEG 2A 1B 2.20
(3 ; Figure 2.0.8: CORE LOADING MAP CONTROL ELEMENT ASSEMBLY LOCATION COLD LEG COLD LEG 2B EAST 1A 8 10 12 I4 I
/
,2 3 4 56i 9 11 ! 13 ID l6 sl7 18 19N2O 21 pl--. s
\ ,/ / N N
,,,- .... ..e 4 .-,.
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, , 68 69
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() NOT LEG 44 26 il 18 27 'r1 s", '- - 2 ,, P R, ...- 56 6 10 7 II H 59 a>
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- L i -
50 22 (> t 4 20 98 i.. K ,. ,,., ,,,, ,,.., . . , ,,.,, ,,... . . . , ,,..., ,,..., ,,..., ,,,,, , , , ..., ...,,, { J ,i . - , - 65 36 15 2 11 23 60 >~ u ... .... ... ., .... ...,, .... .... ...,, .. .. _ y, , NORTH E D 9 5 3 A R
,,,, ...$ ,,... ,,... , , . , , . , . ,,.. ,,.,, ,,.... ...,.. .,..., ...,, ,,.. 1,... .,,.,
F 43 ii i3 24 25 40 E
\\ ,e,. . . ,
75 37 0 21 B 36 72
\ ... ..... .... .... ..., 4... .o . . . ..... 0 - ., m. ,..,. ...... ...,
D d il 3c li 41 61 C
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14
,c ic .
54
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53 33
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A NN ... . . . .. . . ..M . . m COLD LEG COLD LEG 2A 2.21 18
i l-i I i i, l j 1 l l I r i 1 ! l I i l SEcrICN 3.0: MIE POSIU)RE IDF RNCTICNAL TEST l { l i i j i i 1 l i 6 i f l i f l I g 3.1
3.0 POSTCORE HOP RJNCTIONAL TEST (2HB-213-01) Postcore hot functional testing provided the first opportunity to test and exercise integrate plant systms after the core had been loaded. Adjustments of equipnent and measurenents were made which could l;@ only be performed with the core in place. Exanples of equipnent requiring a Mjustments inclMed the minimum spray valves which bypass the main pressurizer spray valves aM electronics which control notion of the Control Elenent Drive Mechanisms (CE[Ws) . Examples of measure- .. ments included reactor coolant systen flow, CED4 drop times, and pres- I surizer performance. The Mdition of the core necessitated additional
- measurments. For instance, although coolant flow through the reactor coolant eycten o a was measured during precore hot functionals, the greater "
system hydraulic resistance with the core in place required the perfor-mance of plannoi Mditional measurenents. f S m The test sequence broo3ht the plant fran anbient pressure aM tenperature tototstandbyconditionsof545Pand3250 psia,andthenreturnedthe plant to hot shutdown conditions of 325 P and 600 psia in preparation for initial criticality. Objectives The objectives were: to demonstrate that the plant could be brought from O cold shutdoan coMition at the end of core loading to hot staMby and back n to hot shutdown in accordance with normal operatirg procedures; to schedule , the sequence of events for the postcore hot functional testirg; to verify 2 design flowrate through the boronaneter; to verify that S/G 104 pressure
- i and pressurizer low pressure trip set points functioned as designed; and to verify autanatic opening of Safety Injection Tank Isolation valves.
Four additional objectives of the Post Core Hot Functional Test Progran were: (1) to satisfy FSAR Pequirments as outlined in Chapter 14 of the FSAR, inclMirg Table 14.2-4; (2) to perform outstanding retests fran precore hot functional testing; (3) to perform for the first time certain Technical Specification Surveillance Procedures; and (4) to prepare for initial criticality by exercising ani denonstrating opra-bility of the plant, whien had essentially been idle for the year between pre- aM pst-core hot functionals. Postcore Hot Functionals required the three nonth period fran early , April to mid July,1982, approximately twice the originally scheduled , duration. Maintenance outages and the accanpanyirn reperformance of a Mode Entry surveillances accounted for the additional time. Rxie Entry
- surveillances are the testiry and checking of equipnent and instruments which are required each time certain operatiry conditions, such as increas-in) essure, ten rature, or reactor power, are atJ ained. The set of coM tions are ca led Operating Modes, or simply, Rdes.
3.2
V 3.0 POSTCORE LOP FUNCPIONAL TEST (2HB-213-01) (Continued) Three maintenance outages resulted in approximately thirty-eight days of delay. ' Den days were required for repairing leaking check valves at the boundary between the reactor coolant system and safety injection system, which had failed a Mode Entry surveillance due to excessive ba:kleakage. '14n days were also required to replace the four-stage mechanical seals on two of the reactor coolant pmps (P002 and P003) . Finally, eighteen days were required for replacanent of all four reactor coolant pmp seals and a realignnent of the pmps and motors. A simnary of problens is outlined in Table 3.0.2. Table 3. 0.1, below, is the sequence of events for Postcore Hot Functionals. Each of the tests is described in 4 further detail below. l 0 W n d 't 1 l I i i i
- 3.3
l I TABLE 3.0.1 POST CORE HOP EUNCTIONAL SEQUENCE OF EVENIS O 'Ihe following is a sequence of events for postcore hot functional testing.
'Ittchnical RCS RCS Specification
- Date Tenp. Pressure &x3e Comnents*
- 4/18/82 150 F 350 psia 5 Prepared to enter Mode 4 4/19/82 260 F 350 psia 4 Entered Mode 4 (>200 F) at 0430, make-up water denineralizer was out of service, resulting in water slurtage. 'Ihe Startup detectors experienced ruise problens. Canpleted precore retests on piping vibration.
4/20/82 260 F 350 psia 4 4/21/82 260 F 360 psia 4 Started Instrunent Correlation testing. Plant computer experi-enced restart problens, resolved late in the day. 4/22/82 260 F 360 psia 4 Statior: Ergineering performed Technical Specification leakrate testing of primary systen boundary check valves. 4/23/82 260 F 360 psia 4 01eck valve leakage testing continued. 4/2 V82 260 F 600 psia 4 Several loop isolation check valves failed Technical Speci-fication leakage limit of 1 gpn. 4/25/82 260 F 360 psia 4 Decision made to return to Mode 5 as a result of check valve failures. 4/26/82 190 F 360 psia 5 Preparai to repair loop check valves.
- Mode 5, RCS Tenp <200 F wxle 4, 200 F < RCS pp. <350 F Mode 3, RCS 'ItEP >350 F
- Problens are described in greater detail in Table 3.0.2 3.4 0
TABLE 3.0.1 POST CORE EDI EtJNCTIONAL SEQUENCE OF EVENTS A (Continued) Technical BCS RCS Specification Date '11!mp. Pressure Mode Cmments 4/27/82 190 F 360 psia' 5 Started 48 hour enduran run on motor-driven Auxiliary Feedwater pumps 2P504 and 2P141. Reperformed check valve leakrate
~
4/28/82' 260 F 2235 psia 4 test at elevated reactor pressure in attempt to seat the valves. Wree check valves still failed the test: S21204MU018 - 1.05 GPM S21204MUO75 - > 50 GPM S21204MUO21 - >100 GPM P004 Anti Reverse Ibtation Device (ARRD) low oil flow alarms; the ARRD p eps lose suction if not operated for long period of time. 4/29/82 260 F 2200 psia 4 Decision was made to drain /depressur-ize to repair the three leaking primary boundary check valves. Auxiliary feed-water pump 48 hour runs were cmplete on 2P504 and 2Pl41. m 0 4/30/82 120 F 100 psia 5 Wree RCP loop valves _were erroneously
- added to the leakage failure list due to E an error in the conversion of percent 5 Safety Injection Tank level to gallons.
After the error was corrected, the leak rate for these three valves was acceptable. ", 5/1/82 .i10 F abient 5 Drained to 39" below reactor vessel & flange for work on check valves. 5 5/2/82 110 F abient 5 Check valve rework continued. 5/3/82 110 F anbient 5 RCP Icop 2 check valves rework cmpleted. O 3.5
r TABLE 3.0.1 POST CORE HOP FUNCTIONAL SEQUENCE OF EVENTS (Continued) Technical PCS BCS Specification Date Tenp. Pressure Mode Comnents 5/4/82 110 F anbient 5 O'Nk valve rework continued. 5/5/82 110 F anbient 5 Icop 1 check valves rework completed. ICS leakage exceeded the Technical Specification allowable identified leakage of 10 GPM. 'Ihe main contributor to leakage was check valve bonnet leakage. Valve bonnet is designed to be seated with the aidition of pressure. Started RCS fill ani vent. 5/6/82 110 F 350 psia 5 Started RCP sweeps for RCS venting. 5/7/82 190 F 350 psia 5 Pressurizer bubble drawn. CEDM venting conpleted. 5/8/82 190 F 350 psia 5 Denineralizer restored to service. 5/9/82 210 F 350 psia 4 Entered !bde 4 at 1837; Auxiliary boiler tube leakage identified. 5/10/82 260 F 600 psia 4 Reperformed Station Engineering primary boundary check valve leakage test; response ttme testing on hat leg and cold leg RID's; Butt weld leaked on shut-down cooling line drain valve, systen drained for weld rework; Electromechanics (a vendor) installed m3difications on CEDM electric drawers. 5/11/82 260 F 2250 psia 4 Twa check valves still faile3 the leakage test; both just over 1.0 GPM, results acc3pted by Station Engineering; excessive noise on Startup Channel 2. 3.6 O
TABLE 3.0.1 POST CORE IUP FUNCPICNAL SEQUENCE OF EVENTS (Continued) {\ Technical FCS RCS Specification Date - Temp. Pressure Mode Caments 5/12/82 260 F 600 psia 4 Response time testing on RIDS. 5/13/82 340 F 600 psia' 4 Identified Mode 3 restraints (>350 F):
- Motor-Driven Auxiliary Feedwater Pung P504 Alignment
- Steam-Driven Auxiliary Feedwater Pump P140 Governor Valve
- Diesel Generator G003 Governor Controls
- Auxiliary Boiler Tube Repair Auxiliary Feedwater Pump 2P504 motor bearings seized during uncoupled run, estimated 2 days to replace with the motor fran the corresponding pump from Unit 3.
5/14/82 340 F 600 psia 4 p 5/15/82 340 F 600 psia 4 CDIES preoperational test was d aanpleted (plant ccrnputer); auxiliary boiler was returned to service. 5/16/82 320 F 600 psia 4 Cooled to 320 F to continue with cold CEDM drop time testing. 5/17/82 3200F 600 psia 4 Continued with CEDM drop time testing. 5/18/82 360 F 600 psia 3 Entered Mode 3 at 0530; instrument correlation testing; incore detector testing; thermal expansion retests. 5/19/82 470 F 1550 psia 3 Instrument correlation testing; a cooldown to Mode 4 was required , to repeat some response time testing , that had been done incorrectly due - to an error in the response time $ testing surveillance procedure. m 5/20/82 320 F 600 psia 4 Response time arrJ CEDM drop time , testing continued. 5/21/82 320 F 600 psia 4 Response time and.CEDM drop time testing cxantinued. 3.7
- - ~ , , , , - . ~ .
.v.. n.. n,,.-n, , , , - . - ,..n.- -. ~,, . -..-, .~ n., - .n - - -
TABLE 3.0.1 POSr CDRE IUr RJNCTIOtmL SEQUENCE OF EVENTS (Continued) Technical RCS RCS Specification Date Tenp. Pressure Made Ccxments 5/22/82 320 F 600 psia 4 Started RG Flow Test aP transmitter calibrations; started post accident g sample preoperational test; identified Made 3 Entry restraints:
- Technical Specification anendment to response time testing;
- Replaced iron bearings with babbitt in Auxiliary Feedwater Pump P140.
- CCW non-critical loop valve stroke time testing.
5/23/82 320 F 600 psia 4 5/24/82 340 F 1500 psia 4 Suspended cold CEm drop time testing for heatup; 65 of 91 rods tested to date. 5/25/82 360 F 1500 psia 3 Entered Mode 3 at 1300. 5/26/82 460 F 1500 psia 3 Reactor Coolant Pump 2P004 ARRD/ Oil lift pmps did not trip due to a faulty speed probe. 5/27/82 545 F 2250 psia 3 Started CEDM Ebt Ibd Drop testing; the upp2r seals failed on Reactor Coolant Pumps 2P002 and 2P003; Middle seals failed on reactor coolant Pump 2P002 at 2000 cminenced cooldown for seal replacement. 5/28/82 110 F anbient 5 Maintenance outa3e was started on the following:
- Replace seals on reactor coolant punps 2P002 and 2P003;
- Rewark leaking safety injection tank isolation valve;
- Rework CEm INAC ductwork;
- Upgrade Main Steam Isolation Valve Ilydraulic Skids;
- Repair misaligned speed probe wheel on RCP001;
- Safety injection tank (s) N2supply check valves;
- Condensate tank T121 level transmittor calibration;
- Auxiliary Feedwater Pump P141 babbitt bearing changeout.
3.8
TABLE 3.0.1 EOST 00RE HOT EUNCTIGIAL SEQUENCE OF EVENTS (Continued) Technical BCS BCS Specification Date Temp. Pressure Mode Ccanents 5/29/82 110 F abient 5 Maintenance Outiage 5/30/82 110 F abient 5 Maintenance Outage 5/31/82 110 F anbient 5 Maintenance Outage 6/01/82 110 F anbient 5 Maintenance Outage 6/02/82 110 F 350 psia 5 Started fill / vent 6/03/82 190 F 350 psia 5 Satisfied M e 4 restraints. 6/04/82 210 F 350 psia 4 Satisfied M e 3 restraints. 6/05/82 320 F 2250 psia 4 Tested primary boundary check valves for leakage. 6/06/82 500 F 2250 psia 3 &xie 3 entry at 1330. 6/07/82 545 F 2250 psia 3 Subcooled margin monitor test com-pleted; thermal expansion hot measure- l;g ments completed; RCP001 upper seal J, failed and middle seal pressure showed oscillation. 6/08/82 545 F 2250 psia 3 Movable incore path length measurenents canpleted; stean generator feedrirg test aborted due to auxiliary feedwater ptznp P140 governor valve problens. 6/09/82 545 F 2250 psia 3 Pressurizer relief setting test com-pleted; area radiation monitor audibility test cxampleted; performed special engineering test on auxiliary feedwater syaten (the cause of wat.er hanners observed at an earlier date in the auxiliary feedwater systen was determined to be from reverse flow through Kerotest valves) . O 3.9
TABLE 3.0.1 IOST CORE HOP EUNCTIOtaL SB2UENCE OF EVENfS (Continued) Technical RCS BCS Specification Date Temp. Pressure !bde Comments 6/10/82 545 P 2250 psia 3 Special instrtmentation setup on RCP001 to monitor pump startup; pump had been idle since its seal failure on 6/7/82. 6/11/82 545 F 2250 psia 3 Special Ergineering test investi-gating auxiliary feedwater system vibration transients resulting in failure of a weld on a vent valve. 6/12/62 545 F 2250 psia 3 Instrument correlation testing; stean generator feedrirg integrity test completed; RCS cold leg whip restraints impinged on pipire result-ing in mirur indentations; data was taken for CF/BPC evaluation. 6/13/82 545 F 2250 psia 3 CfDi ibt Drop Time Nasurenents 6/14/82 545 F 2250 psia 3 CEEM Hot Drops; RCP P002 speed indicators for CPC channels A & D were oscillating + 5 RPM; same indication as was observed on ICP P001 earlier, suspected speed ring misaligrment on shaft. 6/15/82 545 F 2250 psia 3 CECM Ibt Rod Drops Continued 6/16/82 545 F 2250 psia 3 Started RCS Flow Test 6/17/82 545 F 2250 psia 3 RCS Flow Test; RCP P001 declared inoperable at 1900 due to excessive pressure oscillations in the middle seal; coola3 down for seal replace-ment. 3.10
O TABLE 3.0.1 POSr 00RE HOT EUNCTIOPRL SB2UEtCE OF EVENIS (Continued) Technical RCS ICS Specification Date Temp. Pressure ' !tde Comnents 6/18/82 120 F ambient 5 Maintenance Outage included:
- Replace seals on all four RCP's and align notors ard pmps;
- Replace Pressurizer heaters E628 and E629;
- Clean RCS Hot ard Cold Iag RTD
'1hecnowells to inprove response time;
- RCP002 speed probe wheel alignnent;
- Inspect stean generator E088 feed-ring;.
- Repair RCS cold leg indentations fran whip restraint contact;
- Calibrate boric acid flow controller;
- Fix safety irdection tank nitrogen leaks;
- Rework RCS boundary check valves D)
- ich leak; .
- Performed feedring inspection, no damage or indication of excessive loading was indicated.
6/19/82 110 F- ambient 5 Maintenance Outa3e Continued 6/20/82 110 F ambient 5 Maintenance Outage Continued 6/21/82 110 F ambient 5 Maintenance Outage Continued 6/22/82 1100F anbient 5' Maintenance Outage Continued 6/23/82_ 110 F ambient 5 Maintenance Outage Continued 6/24/82 110 F ambient 5 Maintenance Outage Continued 6/25/82 110 F anbient 5 Maintenance Outage Continued 0 6/26/82 110 F anbient 5 Maintenance Outa3e Continued 6/27/82 110 F anbient '5 Maintenance Outage Continued 6/28/82 110 F ambient 5 Maintenance Outage Continued , O 3.11 I
TABLE 3.0.1 POST CORE HOP EUCPIONAL SEQUENCE OF EVENPS (Continued) Technical ICS RCS Specification Date Temp. Pressure Made Cam 1ents 1 i 6/29/82 110 F anbient 5 Maintenance Outage 6/30/82 110 F anbient 5 Maintenance Outaje 7/01/82 110 F anbient 5 Maintenance Outa3e 7/02s'82 110 F anbient 5 Maintenance Outage 7/03/82 110 F anbient 5 Maintenance Outage 7/04/82 110 F anbient 5 obintenance Outage 7/05/82 190 F 600 psia 5 Ccmnenced heatup and perfonned tbde 4 Entry surveillances. 7/06/82 260 F 600 psia 4 tbde 4 entry at 1200. 7/07/82 320 F 1500 psia 4 Continued CEDM cold drop time testing; shutdown cooling outage for flange rework. 7/08/82 320 F 1500 psia 4 Continued CEDM cold drop testing. 7/09/82 320 F 1500 psia 4 Continued CEDM cold drop testing. 7/10/82 320 F 1700 psia 4 Continued CEDM cold drop testing. 7/11/82 320 F 1700 psia 4 Continued CEDM cold drop testing. 7/12/82 320 F 1700 psia 4 Continued CEDM cold drop testing. 7/13/82 320 F 1700 psia 4 Shutdown coolirg maintenance can-pleted; Mode 3 entry surveillance completed. 7/14/82 400 F 1700 psia 3 lbde 3 Entry at 1700. 7/15/82 545 F 2250 psia 3 Continued CEDi hot rod drops. 3.12 O
O TABLE 3.0.1 POST CORE IDF EUNCTIONAL SEQUENCE OF EVENTS , (Continued) ICS RCS Tech Spec *
, Date Temp. Pressure Mode Consnents 7/16/82 545 F 2250 psia '3 CEDM hot drop testing and response time retests.
7/17/82~ 545 F 2250 psia 3 ICS heat loss tests; pressurizer performance tests; instrinnent mrre-lation. ICS Flow Test. 7/18/82 545 F 2250 psia 3 Continued Tests Started 7/17/82. 0 Iktests on pressurizer performance. 7/19/82 545 F 2250 psia 3 7/20/82 320 F 2250 psia 4 Gooled down to 320 F and 600 psia for preparation to Initial Criticality. O L O 3.13
TABLE 3.0.2
SUMMARY
OF PFOBLEMS
'1he following is a sumary of major prchlems and their solutions:
Proble Solution
- 1. Demineralizer supplying water to Installed temporary demineralizers, the plant was out of service and placed oemineralizer in service.
caused a water shortage.
- 2. Startup detectors were noisy. Instrument ground bus was separated frm plant ground bus, grounding schemes were reviewed. m Hardware maintenance was perfomed I;
- 3. Plant cortputer periodically failed.
and software changes were made. @ l
- 4. Check valves separating the primary Check valves disasserrbled, adjusted coolant system and safety injection and seats and disc polished and system leaked back slightly. carefully mated.
- 5. Auxiliary boiler tubes leaked, Repaired leaking tubes.
disabling steam supply to turbine sealing system and preventing vacuum frm being drawn. Steam generator blowdown was stopped which impacted the ability to control its chemistry.
- 6. Bearings on Auxiliary Feedwater Pump Motor frcm SOtES 3 was used to 2P504 seized during uncoupled operation replace the motor; bearings re-after repair of iron bearings. placed with different type (babbitt)
- 7. Reactor Coolant Pump Seals failed Replaced seals and aligned the "
due to vibration and temporary pump and pump motors to reduce loss of cceponent cooling water vibration. Alignment essentially '* flow. eliminated further failures. +i Improved operating procedures. fj
- 8. Main Steam Isolation Valves Upgraded hydraulic actuator failed to open. piping and controls.
- 9. Nitrogen supply to safety Repaired safety injection tank injection tank check valves. nitrogen supply check valves.
- 10. Pressurizer spray valve packing leaked. Replaced packing.
- 11. Speed sensing ring on one Reactor Realigned speed sensing ring.
Coolant Pump was misaligned.
- 12. RCS Cold Ix.g whip restraints impinged Reshinmed restraints, on RCS piping, resulting in minor Engineering evaluation was per-indentation. formed. m
- 13. CEDM upper grippers were sticking Upper and lower gripper coil causing scme CEAs to drop or slip. replacement was required. Worst lg sticking CEDMs were replaced after *m low power Fh ysics testing.
~
- 14. Primary Coolant Syr. tem Resistance .
coated RIDS Two with gold to reduce the gap between the RIDwere removed frm
'Itmperature devices (RIDS) responded 7 tco slowly. and well, reinser-ted, and retested satisfactorily. m 3.14
i- ! I i 4. o 4 I l
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the primary coolant system and related control systems, j i j' i 1 1 J. I . 4 V: i
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3.1.1 PRE CRITICAL COMPARISON OF CPC, PPS AND PLANT CCNPUTER (2HB-458-01) In accordance with the requirements of FSAR Section 14.2.12.78, this test verified that indications frm the plant cmputer, control N toard instruments, Core Protection Calculator (CPC) and Plant Protection System (PPS) for a given plant parameter all agreed j within specified tolerances. All open retests frm the Precore 1 Hot Functional Instrument Correlation procedure (2HA-317-01) were .g also retested. nsting was performed at the following temperature and pressure plateaus: Temperature Pressure RCS Pressure (psia) Tcold( F) 260 350 260 600 360 600 470 1550 530 1900 54 5 2250 te parameters and associated indication are listed on Table 3.1.1.1. Ebr each parameter the deviation frm the " Target Value" was calculated and compared to specified tolerances. We target value was calculated N as the average of the most reliable indication for each parameter, Normally,themostreliablewerethesafetygradeCPC'sfortheparametersljg which are in range; otherwise, the PPS was used. .A Test Exception j Report was generated whenever the calculated deviation frm the Target Value exceeded specified tolerances. 3.16
3.1.1 PRE CRITICAL COMPARISON OF CPC, PPS AND PLANT COMPUTER (2HB-458-01) (Continued)
~
A total of 77 Test Exception Reports were generated during this proce-dure. 'Ihese exceptions were divided into the following major categories. ldg m Category Approximate Percent of n Total 'Ibst Exceptions . E
- 1. Instruments that required only minor 48%
recalibration for deviations which just exceeded acceptance criteria. n
- 2. Failed indication that required 16% @
repair or recalibration. 5 3.- BCP speed probe or speed 13% indication problems.
- 4. Refueling water storage tank 13%
n O level indication was improperly
'd calibrated. ,
E o
- 5. Hot Leg Temperature Indication was 6%
added via a design. change which did not function properly (2TIO111BX).
- 6. Critical Functions Monitor System 4%
g
]
(CENS) indication. (System was undergoing . sof tware develognent. ) g m I d 3.17 t
TABIE 3.1.1.1 DEVICES IN WiICH PARMETERS ARE DISPLAYED Parameter Indication Instrument Cmputer CPC* PPS* BCS Cold Iag Temperature X X X RCS lbt Leg Tenperature X X X RCP Speed X X RCS Pressure X X X X Pressurizer Ievel X X Stea t Generator Level X X X Steam Generator Pressure X X X O Stean Generator Differential Pressure X X Containment Pressure X X X Refuelity3 Water Tank Level X X X
*CPC is Core Protection Calculator PPS is Plant Protection System 3.18
p b 3.1.2 REACIOR COOIANT SYSTEM FIOW MEASUREMENTS (2HB-213-05)
! Sumary
%e RCS Flow Measurement Wst determined both the primary coolant N
steady-state flow rate and the coastdown flow rates experienced upon the de-energization of all four reactor coolant pumps (RCPs) . M- ; ditionally, flow related algorithms ard mnstants used in both the Core & Protection Calculators (CPCs) and the Core Operating Limit Supervisory a System (CDLSS) were verified and/or determined. W e test began'on June 15, 1982 with the final data being collected on July 18, 1982, -" < following a plant outage for Reactor Coolant Pump seal replacement. j
%e four pump RCS flow rate and flow coast down were determined to be i acceptable. Performance of this test satisfied FSAR Section 14.2.12.74. j he Primary coolant steady-state flow rate, based on measured m pressure drops across the reactor vessel, provided a preliminary -
determination of RCS steady-state flow. We final determination E was made at power by procedure 2PA-344-12, RCS Calorimetric Flow j Measurements. (h U Objectives h e objectives of the RCS Flow Measurement W st were to: (1) Verify the as-built four pump steady state RCS flow rate; (2) Demonstrate that the J BCS flow coastdown was consistent or conservative with respect to safety & analysis; and (3) Verify the validity of the flow related algorithirs and 5 constants in the COIss and CPCs. Test Method he test method consisted of: ,
- 1) Determining RCS flow by measuring the differential pressures devel- d oped across the reactor vessel and reactor coolant pumps with speci- g fied reactor coolant pump ccabinations. m
- 2) Tripping the appropriate reactor coclant pump (s) for collection of coastdown data.
fm
. k__.) 3.19 l-N . .-- , _ . . - _ .. ,-
3.1.2 REACIOR COOLANT SYSTEM FII)W MEASUREMENIS (2HB-213-05) (Continued)
- 3) Verifying flow related algorithms by conparison with measured flows.
- 4) Tripping various RCPS to provide additional data not required by the FSAR.
- 5) Adjusting CPC and COLSS constants.
Eis test measured steady state and transient flows. Transient flow conditions existed just after one or more of the Reactor Coolant Pumps (RCP's) were tripped ani the system was readjusting to its new configura- ~" tion. Steady state was knen the system had stabilized. Stea3y state flow rates were determined fran differential pressures (dp's) developed across JJ the pumps. Rese curves were pp specific (developed by the vendor fran actual test loop data) ard were valid only when the punps were operatire am at rated speed. For this reason, transient flows were inferred fran the pung curves and RCP shaft speed usire the Affinity Laws. Due to problems encountered during pre-core reactor coolant flow measurements at SOKIS-2, the applicability of the vendor supplied pung casing dp curves was questioned. A method of determining flow rate lf p o based upon reactor vessel dp values was fonnulated. This was used to determine the four pump steady state flow, which was then used to adjust the vendor supplied pump casing dp curves. The adjusted curves were subsequently used to detecnine the transient flow rates. O 3.20
( 3.1.2 REAC'IOR @LANT SYSTEM FLOW MEASUREMENIS (2HB-213-05) (Continued) All of the data was corrected for calibration and density effects as per the following:
- 1) Se RCP dp values were corrected to conditions of 553 F, 2250 psia j since the BCP flow curves were normalized to those conditions, &
5
- 2) We SG and RV dp values were corrected to 545 F, 2250 psia, the ccnditions to which the RV dp flow curves were normalized, lj S
- 3) All values were corrected for the zero ptznp data (static zero of fset) .
Since one purpose of the test was to obtain baseline thernohydraulics $" data and since the flow rates were determined frca dp values, all qf available RCS dp instruments (as well as several tenporary instruments) gg O were monitored during the test. %is included eight steam generator dp instruments (four per loop), eight RCP dp instruments (two per
=m pump), and four reactor vessel (RV) dp instruments (one per loop, temporary).
4 O U 3.21 l
3.1.2 O REACIOR COOEANT SYSTEM ETDi MEASUREMENTS (2HB-213-05) (Continued)
, Steam generator dp instruments were connected to pressure taps on the "
bot legs ard discharge plenum of each generator. BCP instruments were ; connected to pressure taps at the suction and discharge of the pump, i and reactor vessel (RV) dp instruments were connected to a discharge 5 tap of the PCPs and cne of the hot leg taps. All sixteen of 3e differential pressure instrument loops were carefully calibrated prior to the test and were checked following the test to verify that no drift or decalibration had occurred. In addition to the dp data, RCP shaft speeds, RCS temperatures and pressures, RCP breaker status, and CPC calculated mass flow fractions were recorded. All of this information was digitized and stored on magnetic tape " using the startup cmputer (a miniemputer based data acquisition .; system). In addition, the sixteen dp signals were recorded on strip & charts to characterize noise. 3 O During the. steady state portions of this test, the plant cmputer (PMS) collected and printed predetermined COISS calculations and parameters for verification of the 00ISS flow algorithms. Data was also collected simultaneously by hand frm the CPCs for the same - purpose. During both transient (i.e. RCP trips) and steady state ,; runs, the startup cmputer was utilized in addition to the EMS. Hand g collected CPC data was not collected during the transient tests, due to g their short duration. 3.22
O V 3.1.2 REACIOR COOLANT SYSTEM FIN MEASUREMENIS (2HB-213-05) (Continued) RESULTS
~
t e full (four pep) reactor coolant systs flow rate was within the , acceptance criteria as shwn in Table 3.1.2.1 and satisfied one of the - objectives of the test. lE m We four pmp flow coastdown which resulted frm deenergizing all four reactor coolant peps also satisfied the acceptance criteria, as shwn - in Figure 3.1.2.1. The figure shws the FSAR 4-pmp coastdown and . that calculated frm the neerically smoothed data. lE? m PIOBIEMS AND RESOLUTIONS The data was found to be " noisy" due to fluctuations in the RCP dp signals used to calculate flow. The predminant frequency, as measured -N jj by a spectr a analyzer, was 19.75 Hz and was attributed to the RCP impeller vane passirg frequency. The flow coastdowns were fairly rapid gg transients and the noise initially interfered with efforts to calculate jg flow during the coastdowns. The resolution was to numerically smooth the RCP dp data and then calculate the coastdowns and CPC constants. f v l l 3.23 I.
TABLE 3 .1. '2.1 Steady State Full Flow The following RCS flows were measured by two rrethods; the reactor $ vessel dp and the RCS dp. Allfourreactorcoolangpumpswereoperating, qq 'Ihe values are corrected to design ccnditions (553 F Tcold)* SS mm Reactor Vessel dp Method RCP dp Method PSID PSID RCP Vessel D/P Vessel Flow RCP D/P RCP Flow 1A 58.386 449,000 gpn 100.106 113,000 gpn 1B 57.885 447,000 gpn 100.469 114,250 gpn 2A 58.463 450,000 gpn 101.97 112,000 gpn 2B 57.365 446,000 gpn 98.786 112,000 gpn O Total Flow (Vessel D/P Method) = (449,000 + 447,000 + 450,000 + 446,000)ggn 4
= 448,000 gpn Total Flow (RCP D/P Method) = (113,000 + 114,250 + 112,000 + ll2,000)gpn
= 451,250 gpn Design Flow = 4 RCP's
- 99,000 gpnVRCP = 396,000 gpn f=
5 Acceptance Criteria - Flow rate > 418,400 gpn and < 452,800 gpn 3.24
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3.1.3 RFJCIOR COOLANT SYS'IEM HEAT IDSS (2HB-213-06) We objective of the Reactor Coolant System Heat Loss Test was to determine the thermal heat loss frm the Reactor Coolant System under normal operating hot standby conditions. his test satisfied FSAR section 14.2.12.73. Heat Ioss frcm the Reactor Coolant System and pressurizer were measured separately. he tast was initially performed during precore hot functionals when portionc of the RCS were covered by blanket insulation " rather than the peur.anent reflective type. The precore test also failed its acceptance criteria; heat loss was higher than !* predicted. W e test was reperformed during postcore hot func- lR tionals when all permanent insulation was in plam. 5 Test Method We test method consisted of two portions, one for the RCS heat loss and the other for pressurizer heat loss. Both required estplishingstablehotstandbyoperatingconditionsof: 545 F and 2250 psia RCS temperature and pressure; all four reactor coolant pumps operating; a sirgle chargirs pump operating and being balanced by letdown flow; normal containment heating and ventilation; auxiliary feedwater supplying the steam generators; and steam flow being controlled by the atnespheric dump valves. We RCS heat loss portion involved a test technique called a "steamdown", in which steam generator water level was raised frcn its ncminal 69% narrow range level to approximately 85% narrow range level, as shown in Figure 3.1.3.1. Feedwater flow was then secured and steam generator pressure and RCS temperatures were maintained as constant as pssible by adjusting steam flow through the atmospheric dump valves. A gradual reduction in steam generator water level occurred as the water was boiled to steam. Steaming was required to maintain constant RCS tempera-tures since the 24 megawatt heat input frcm the reactor coolant pumps far exceeded all other heat losses and loads. Heat loads _ such as warming inccming charging flow were quantified by measur- . ing darging and letdown flow and tempratures. It was important 2 to perform the test under conditions where all heat sources and m sinks could te carefully quantified since the heat loss is a small number, approximatelv 0.1% of full power. Cuantifying heat sources required there be no decay heat so the test was performed prior to initial criticality. We steandown continued for approximately one hour during which time water level decreased through the essentially empty upper cylin- N drical portion of the steam generators. We rate of heat loss - due to steaming was easily quantified by knowing the volume of g saturated fluid converted to steam, the change in enthalpy at constant g steam pressure, and the time over which the steamdown occurred. All heat sink categories except cne, namely that due to insulation and other losses, were quantified and su:med. We a difference between heat losses and heat sources, was due to insulation and other normal operating static heat losses. 5 3.26
3.1.3 REACIOR COOLANT SYSTEM HEAT IDSS (2HB-213-06)
'Ibst Method (Cont.) ' ' - te pressurizer heat loss was obtained in a different manner.
Reactor Coolant System temperatures were maintained as constant as possible using the atnespieric dumps, auxiliary feedwater, and charging and letdown systems. Pressurizer pressure was main-tained within a.10 psi band about normal operating pressure of 2250
- psia by manually energizing and deenergizing one or more banks of-backup heaters as required. Only the backup heaters were used because their heat input could easily be quantified since they are N either "on" or "off" and do not provide a variable output like the proportional heaters. Once current and voltage to the backup heaters ;
l had been measured, the only source of heat into the pressurizer (the i heaters) could'be determined by knowing the total time the heaters y, were energized during the test. Were were two portions to the pressurizer heat loss test. We - first measured loss without spray flow. It was accomplished .
. by deenergizing the reactor coolant pumps which supply spray flow. E Heat loss from the test without spray quantified losses due to heat 5 leakage.through insulation and the uninsulated pressurizer support skirt, instrument fittings, and safety valves. Were was no measurable steam leakage through the safety valves; heat loss was due to insulation losses.
W e second portion was performed with normal steady state operating conditions and all four reactor coolant pumps operating. It involved O the additional heat input to warm inconing small flows of minimum spray flow. S e technique of manually cycling backup heaters was-used for both pressurizer heat loss tests. Results , We results of the BCS heat loss and pressurizer heat loss with spray tests lE
-fell well outside of the acceptance criteria, as shown below. However, m an engineering evaluation was performed and the results were accepted with no changes to the insulation. The results of the heat loss test n. '
. indicated that reflective insulation might be slightly less effective than .
the blanket insulation used during precore bot functionals. W e heat j losses obtained during pre-and post-core hot functionals were quite F similar, however. l As Found Heat Ioss Acceptance l Parameter Precore Hot Functionals Postcore Hot Functionals Criteria N RCS Heat Ioss 4.26 Mw 4.69 Mw 10.50 Mw } 5 4 Pressurizer Heat Loss l Without spray 113 Kw 131 Kw 1131.8 Kw .- O With Mini Spray 178 Kw 272 Kw ~<149.4 Kw I te tests were performed without problems. 3.27
Figure 3.1.3.1 STEAM GENERATOR WATER LEVEL CHANGE DURING STEAMDOWN P STEAM N0ZZLE- =
+
f A c
^"
STEAM DRYER DOME (TYPICAL) T MOISTURE SEPARATOR UPPER LEVEL TAP I3 (TYPICAL) [] APPR0XIMATE - (100%) { i; LEVEL CHANGE ~K=fr~ l
'}
FEEDRING , , FEEDWATER N0ZZLE , RISER '% (E46.8% C EL.71'-2 3/4" ' _ __00 0 P l I i 1999 .'
- BOTTOM OF FEEhATER LINE ,
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l l 1 s 3.1.4 'INERMAL EXPANSICN TEST (2HB-102-01) m his procedure, in conjunction with 2HA-102-01 (Precore Hot Functional Wermal Expansion) and 2PA-102-01 (Power Ascension %ermal Expansion), ; satisfied FSAR 14.2.12.72 requirements for monitoring selected primary g and secondary plant piping systems. y he Postcore Hot Functional Wermal Expansion procedure (2HB-102-01) was written specifically.to retest portions of the precore test where measurements were questionable. W e original test program did not include such testing in the Postcore Hot Functional program. I W e recorded pipe movements were transmitted to Bechtel Pipe Stress Analysis Group to detemine if pipe movement was acceptable. We test was performed satisfactorily with only one problem support, which was rewcrked acceptably. We test monitored the steam generator blowdown sample lines, a charging line to the RCS, steam generator sliding base, and see safety injection line snubbers. We testing was performed as retests from the Precore Hot Functional hermal Expansion 'Ibst.
, W e procedure initially measured each designated support or snubber in I
the cold shutdown condition, approximately 60 F. Measurements were taken using a mechanics ruler and a digital pyrmeter. Subseguent meagurements were made at reactor coolant'tsnperatures of 360 F, 4 545 F and again at ambient conditions. We pipes were monitored i for x, (north-south), y. (vertical) and z (east-west) orientation and novement. Se steam generator blowdown lines snubbers were monitored as shown in Figure 3.1.4.1. We lines moved as predicted with no binding. We reactor coolant pump supports and the steam generator sliding base grew away fr m the reactor vessel without restriction. h e only problem encountered involved a letdown line which bound at one point. We stnort was a " sliding T" type. A bracket in the shape of an inverted
~"T" was clamped to the pipe and was held in place by another set of brackets which allowed the pipe to nove, or slide, axially but not radially. %e clips holding the "T" were too tight and bound the entire support in one position. By removing the clips and rewelding them with a greater clearance between the clips and the "T", the problem was resolved and the pipe moved as predicted.
Results i
% e final measurement at cold shutdown conditions provided verification that the pipes returned to their original position.
O 3.29
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,, 3.1.5 CCNIlOL ELEMENT DRIVE MECHANISM (CET) AND 00talOL
( ) ELEMEtR ASSEMBLY (CEA) TESTS (2HB-316-01) G' he CEDM test verified proper operation of the Control Element Drive Mechanisms (CEDM) and their associated c3ntrols and position indications as required by FSAR Section 14.2.12.75. We test was conducted in stages frm April through July 1982. Each Control Element Assembly (CEA) was withdrawn and inserted twice (in both hot shutdown and hot standby conditions) to measure withdrawal speed,. drop time when the CEA was tripped, and the accuracy of alarm and position indicators. All CEAs performed satisfactorily. Sticking of the upper grippers was experienced in several CEDMS, but was overcme by cycling the grippers or, in four cases, by replacement of the CETs. Cbjective h e objectives of the test were toi
- 1. Denonstrate proper operation of the Control Element Drive Mechanism in both the CEA withdrawal and insertion nodes.
- 2. Verify the proper operation of the CEA position indication systes and alanns during withdrawal and insertion.
o 3. Measure CEA drop times to 90% insertion and 100% insertion frm fully withdrawn. Test Method Each CEA was tested individually at hot shutdown (320 F) with two Reactor Coolant Pumps (RCP) operating and at hot standby (545 F) with four RCPs operating. % e test sequence at each temperature was identical. Initially, the CEA was withdrawn to register the initiation of deviation alarms (indicating the deviation of.an individual CEA frm its group) and the de-energizations of CEA bottm (fully inserted) lights. %e withdrawal was continued, measuring the agreement of the position indicators at 22" increments, to the top of the core (150") . he CEA was then inserted, again measuring position indicator agreement at 22" intervals, to the bottm of the core. All position verifications, for both alarms and periodic rod heights, required that
-the Control Element Assembly Calculators, Plant Cmputer, and CEDM Control System step counter agree within 2.25 inches. During the withdrawal and insertion, strip chart recordings (Visicorder traces) _
were made of the current to the CEDM magnetic coils. Evaluation of the recordings determined either proper CEDM operation or tha need for adjustments to the control system. (q } V 3.31 L
3.1.5 CONTROL ELFRENT DRIVE MICHANISM (CEDM) AND CONrFOL ELEMENT ASSEMBLY (CEA) TESTS (2HB-316-01) Test Method (Cont.) Se CEA was then withdrawn a second time, pulling continuously frm the bottm to the top of the core. During this phase the withdrawal speed was measured. Finally, the CEA was tripped and the drop time frcm fully withdrawn to both 90% ard 100% insertion was measured. We fastest and slowest rod at each temperature plateau was redropped ten times each as required by Regulatory Guideline 1.60, Revision 0. Zero-flow drop testing, as suggested by the Guideline, was not performed because experience at previous CE Plants has shown the drop time with flow is more limiting than the drop time at na flow. S e slowing effect of greater RCS flow on drop times can be seen by evaluating the attached n disgributionplotsFigures3.1.5.1and3.1.5.2. he 90% drop time at . 320 F (X = 2.5 sgconds) with two RCPs operating was faster than the 90% l1 drop time at 545 F (;: = 2.63 secgnds) with four RCPs operating, even i though the lesser density of 545 F water would tend to make the drop time faster. Results All CEAs dropped to the 90% insertion point within the Technical Spec-ification requirement of 3.0 secorxis. W e 90% insertion time was used tecause the bottm 10% of the guide tube, into which the CEA falls, is - tapered to act as a hydraulic buffer and slow the descent of the CEA. E n us, the last 10% of travel was not representative of the CEA insertion 3 speed in most of the core. Additionally, at 90% the CEA has essentially inserted all of its reactivity. Se 90% drop times rarged frm 2.30 seconds to 2.84 seconds and are shown in Table 3.1.5.1. Table 3.1.5.2 shows additional drop times for the fastest and slowest CEAs. Se with-drawal speeds were acceptable cmpared to the nminal values of 30 inches / minute for full length CEAs and 20 inches / minute for part length CEAs. Se term "part length" refers to the length of the active poison section n (75" versas. approximately 150" for full length) . Se actual length of , the two types is virtually equal. Se part length CEA was adjusted to lq withdraw at a slower rate because it is about 25 pounds heavier than a g full length CEA due to the use of Inconel instead of B C in the bottm e 75" of each CEA finger. Se additional weight also a ts for the small peak in the attached drop time graphs at 2.30 - 2.40 seconds. All eight part length CEAs were in the lower end of the drop time spectrum. We CEDMs generally operated satisfactorily except as noted below in the problem section of the report. Position indicators agreed within 2.25" of each other, although sme required adjustment and retesting. 3.32
3.1.5 CONrROL ELEMENT DRIVE MECHANISM (CEDM) AND CCNTROL [-
\
ELEMENT ASSEMBLY (CEA) TESTS (2HB-316-01) (Continued) Probles 'and Resolutions te most significant problem encountered during the performance of this test was sticking upper grippers in t.Te CEDMs. We stickirg gripper caused the CEA to slip inward. Slippage occurred only when the CEAs were being noved. m slippage occurred from a rest position. In some instances, the proble was solved by cycling the grippers, as many ~as 20,000 times, until it operated freely. Ebur CEDMs were replaced in order to avoid potential pecbles during subsequent startups, shutdowns and surveillances. he replacements were made during a maintenance outage in Septaber, 1982. Operation of the CEDMs has greatly improved since that time. Several position indication problems were found during the initial round of testing, especially at the top of the core. Se problem was corrected by adjustment of the Paed Switch Position Transnitters (RSPIs) which provide CEA position signals. Additionally, four RSPIs malfunctioned and were replaced. Conclusion he test satisfied the requirements of the FSAR ard showed that the
] CEAs will drop within the required time limit ten tripped. Addi-tionally, the test produced data concerning the CEDM Control System that allowed the system to be adjusted for optimum performance.
^ 3.33
Figure 3.1.5.1: CEDM COLD DROP TIMES CEA DROP TIMES (90%) X = 2.56 sec 320*F 40 -
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CEDM TEST 2HB-316-01 Table 3.1.5.1 CEA DROP TIMES (second s) CEDM Visigorder9g% Visfcorder 10g% 4 320 F 545 F 320 F 545 F 1 2.73 2.63 3.03 2.94 2 2.52 2.61 2.82 2.89 3 2.51 2.61 2.80 2.90 4 2.57 2.60 2.85 2.88 5 2.52 2.62 2.79 2.88 6 2.57 2.63 2.85 2.91 7 2.64 2.67 2.92 2.92 8 2.57 2.66 2.87~ 2.93 9 2.58 2.65 2.87 2.92 10 2.54 2.62 2.83 2.88 11 2.58 2.65 2.85 2.92 E 12 2.53 2.63 2.83 2.90 3 13 2.54 2.65 2.84 2.90 14 2.55 2.59 2.85 2.89 15 2.46 2.55 -2.76 2.85 16 2.51 2.61 2.80 2.86 17 2.56 2.62 2.85 2.92 18 2.56 2.57 2.85 2.87 19 2.51 2.69 2.81 2.89 20 2.52 2.50 2.79 2.86 21 2.57 2.59 2.85 2.85 22 2.58 2.62 2.87 2.91 23 2.57 2.62 2.85 2.90 2.63 24 2.72 2.92 3.00 25 2.52 2.56 2.83 2.86 3.36
CEDM TEST 2HB-316-01 m Table 3.1.5.'l CEA DROP TIMES (seconds) CEDM Visigorder9g% visfcorder10g% 4 320 F 545 F 320 F .545 F 26 2.78 2.70 3.12 3.00 27 2.55 2.63 2.85 2.90 28 2.44 2.50 2.72 2.75 29 2.30 2.35 2.60 2.61 30 2.34 2.39 2.63 2.65 31 2.33 2.43 2.61 2.65 32 2.40 2.38 2.65 2.64 33 2.37 2.45 2.67 2.67 34 2.40 2.42 2.65 2.68 35 2.40 2.40 2.65 2.66
. (~) \- ' 36 2.62 2.65 2.92 2.95 .
E 37 2.57 2.67 2.87 2.95 & 38 2.45 2.62 2.78 2.87 39 2.55 2.62 2.58 2.89 40 2.60 2.62 2.9 2.90 41 2.53 2.57 2.83 2.87 42 2.55 2.65 2.85 2.90 43 2.55 2.66 2.85 2.91 44 2.51 2.65 2.83 2.90 45 2.50 2.62 2.79 2.87 46 2.60 2.65 2.88 2.90 47 2.55 2.70 2.87 2.95 48 2.58 2.72 2.87 2.97 ' O' 49 2.67 2.68 2.91 2.97 50 2.53 2.65 2.82 2.92 3.37
CEDM TEST 2HB-316-01 Table 3.1.5.1 CEA DROP TIMES (seconds) CEDM Visigorder9gt Visfcorder 10g% 320 F 545 F 4 320 F 545 F 51 2.56 2.63 2.85 2.89 52 2.60 2.70 2.87 2.95 53 2.61 2.69 2.92 2.98 54 2.53 2.62 2.82 2.90 55 2.59 2.65 2.90 2.95 56 2.66 2.73 2.94 3.00 57 2.59 2.68 2.90 2.96 58 2.74 2.73 2.10 3.03 59 2.84 2.73 3.18 3.03 60 2'.62 2.66 2.91 2.97 61 2.56 2.63 2.85 2.92 62 2.57 2.62 2.87 2.89 R 5 63 2.55 2.65 2.87 2.92 64 2.77 2.75 3.10 3.05 65 2.75 2.75 3.07 3.01 66 2.55 2.66 2.85 2.95 67 2.63 2.72 2.92 2.98 68 2.80 2.78 3.1 3.09 69 2.58 2.70 2.89 2.97 70 2.60 2.65 2.88 2.91 71 2.55 2.70 2.88 2.99 72 2.62 2.73 2.91 3.01 73 2.52 2.63 2.84 2.90 74 2.55 2.70 2.85 3.00 lh 75 2.81 2.80 3.15 3.15 3.38
CEDM TEST 2HB-316-01
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s Table 3.1.5.1 CEA DROP TIMES (seconds) CEDM Visigorder 9g% Visfcorder10g% 4 320 F 545 F' 320 F 545 P t 77 2.51 2.75 2.92 3.05 78 2.57 2.65 2.85 2.95 79 2.76 2.77 3.09 3.07 80 2.52 2.68 2.83 2.95 81 2.67 2.64 2.87 2.95 82 2.55 2.68 2.86 2.95 I 83 2.59 2.70 2.85 2.95 84 2.60 2.70 2.87 2.97 85 2.57 2.65 2.86 2.93 86 2.60 2.65 2.80 2.93 87 2.57 2.65 2.85- 2.95 & 5 88 2.50 2.59 2.80 2.85 89 2.40 2.42 .2.67 2.69 90 2.40 2.42 2.65 2.67 91 2.45 2.42 2.70 2.67 i O 3.39 e -
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CEDM TESTS TABLE 3.1.5.2 TEN DROPS OF FASTEST & SWIST CE&ls SECONDS DROP TIME 'IO 90% INSERTED 320 F 545 F RETEST SIIEEST FASTEST SD'EST FASTEST 4 CEA 59 CEA 29 CEA 75 CEA 29 l 1 2.80 2.33 2.75 2.37 2 2.80 2.35 2.75 2.35 3 2.85 2.37 2.75 2.33 4 2.85 2.33 2.73 2.33 5 2.82 2.33 2.77 2.33 6 2.82 2.33 2.73 2.33 7 2.82 2.33 2.72 2.32 8 2.82 2.35 2.73 2.35 9 2.82 2.35 2.72 2.32 10 2.82 2.33 2.73 2.32 l 3.40
3.1.6 FIXED INOORE INSTRUMENTATION (2HB-310-01)
'Ihe purpose of the Fixed Incore Instrumentation Test was to verify the proper functional performance of the fixed incore nuclear instru- !
mentation system. he fixed incore instrumentation has fifty-six strings of five self-powered incore neutron det.ectors. Sixteen of the strings also include a background detector. Each string has a single core exit thernocouple. Verification of proper operation involved cross-correlating core exit thermocouple temperature readings with high-accuracy Resistance 'Ibmperature Devices (RfDs) located on the hot and mld legs of the RCS and measuring the leakage resistance of the incore detectors at rominal hot standby conditions. We test satisfied portions of FSAR Section 14.2.12.80 and 14.2.12.67. %c remaining portion of the Section was satisfied under 2HB-310-02, Movable Incore Detector Acceptance 'Ibst. Objective . Since the self-powered Rhodium neutron detectors do not produce usable amounts of electrical current until approximately fifteen percent reactor power, pre-critical checking was limited to decking resistance of insulation and connectors and cross-correlating thermo- 1 couples which responded to temperature changes. W e general arrangement ~ of thermocouple and detector strings are shown in Figure 3.1.6.1. Each string consists of five neutron detectors (mounted every twenty percent of core height) and a core exit thermocouple nounted approximately h-) one foot above the top of the active fuel._ The incore string is lowered v into the center CEA guide tubes of selected unrodded bundles, (bundles with no CEAs). he string's long instrument lead is routed to a connector at the periFhery of the reactor vessel head, then to an amplifier and the m plant cmputer. We leads have high-resistance insulation. It is, however, possible for the insulation in the leads and the connectors - to be damaged during installation. m is was the reason for insulation a checks. j 2e core exit thernoccuples serve an important function in detecting inadequate core cooling. Cross-correlating their readings with high-accuracy RrDs demonstrated they were properly connected and are operating correctly.
'Ibst Methcd W e test method consisted of two parts. First, core exit thermocouple datg was taken at various temperature plateaus up to and including 545 F, the nominal hot zero power tenperature. RID temperatures were also remrded and used for cross-correlation. he emputer was connected to only forty-eight of the fifty six thermocouples because eight are dedicated to the Inadequate Core Cooling display on the reactor operator's panel. All thermocouples wgre monitored, however. %e u original acceptance critgria of +1.5 F was too restrictive
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and was widened to +4.0 F. E p) a 3.41
l l 3.1.6 FIXED IN00RE INSTRUMENTATION (2HB-310-01) Test Method (Cont.) O
'1he second part of the test, the fixed incore resistance check, performed to determine that insulation resistance was at least wag" 10 ohns.
Results
'Ihree core exit thermocouples, at the burdles in core locations R-09, L-04, and G-20, were found defective. Operation with these three failed thermocouples is a:ceptable. 'Ihe others performed satisfactorily. Besistance checks on all of the fixed incore detectors were successfully performed.
O 1 3.42 O
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? To Computer Connector Reactor Vessel Head i i
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C\ \ ) v 3.43
3.1.7 MOVABLE INCORE DETEC'IOR SYSTEM CHECKOUT (2AC-310-02) Re purpose of the Movable Incore Detector Checkout was to deconstrate O proper operation of the movable incore detector system at hot zero power conditions, including indexing and novement to all accessible locations, h e test satisified FSAR Section 14.2.12.80.1.C. Objectives he cbjectives of the checkout were satisfied by (1) measuring physical' cnd position indication (encoder) path lengths at ambient temperature and pressure, (2) verifying acceptable cable and incore detector asserbly resistance, (3) verifying leak detector operation and (4) verifying purge gas operation. Test Method A portable control box was used to actuate the drive motors and indexing equipnent from the reactor vessel head. Path lengths were measured phy-sically. Purge gas ficw, which is intended to maintain the drive medianism and indexer free of moisture, was verified. Operation of the leak detec-tion system was denonstrated and cable and detector assembly resistance were measured. We detector is a self-powered Rhodium neutron detector, quite similar to the fixed incore detectors, so the procedure for measuring its resistance was similar. Results %e novable detector's path lengths, lead and detector resistances were verified to be correct. He purge systen and leak detection system were verified operable.- Problems here were several problens with the physical path lengths and encoder path lengths dif fering by nore than the acceptance criteria. Software changes were implenented to correct these problems. We transfer machines were flooded through their drains to the Reactor Coolant Drain Tank. W e drain lines to the tank allowed water to enter the transfer machines when the RCDP was pressurized by inleakage frcm the reactor coolant systen. he drain line was renoved and routed to the floor. Detectors tended to stick in their tubes. A drive motor burned out-during checkout and was rewound and installed with no further problens. 3.44
3.1.8 PRESSURIZER SAFETY VAINE TEST (2HB-313-02)- We pressurizer safety valve test was perfonned to verify the setpoints and seat tightness of the pressurizer safety valves on June 9 and 10,1982. W e test satisfied Section 14.2.12.58 of the FSAR. Objectives he setpoint of each pressurizer safety valve was determined by m using a device called a hydroset, with the RCS at normal operating temperature and pressure. .ne hydroset unit provided additional - upward force cn the valve stem, until the valve just started lifting
~
I or "sinmering." Per section 14.2.12.58 of the FSAR, the required S setgoint is 2500 + 25 psia. Seat tightness following the setpoint determination was verified by monitoring the safety valve discharge line temperature and Quench Tank temperature, pressure and level. W is satisfied FSAR section 14.2.12.58.- Method Under the direction of the vendor representative (Dresser Industries)', D) a hydroset device was attached to each safety valve'(one at a time) . He hydroset device was essentially a hydraulic jack, which was attached to the valve spindle and utilized the valve bonnet to bear the force n developed by the jack. With the reactor coolant system at normal , pressure and tenperature, hydraulic pressure was supplied to the q hydroset by a hand pump and was monitored by a test gauge. Hydraulic g pressure was raised until the force developed by the jack plus the m force of the RCS pressure acting on the valve disc overcame the force of the safety valve spring. - Balancing the forces caused the valve to begin to open and produce audible flow. We hydraulic pressure required to open the valve was converted to equivalent system pressure using charts provided by the valve manufacturer and added to actual system pressure to determine the actual set point. Seat tightness.after valve operation was checked by monitoring safety valve discharge line temperatures and by monitoring RC quench tank parameters. Conditions prior to lifting were recorded and compared to post-test conditions. he Quench Tank was monitored for a 2.5 hour period after the final safety valve lift. Seat tightness was verified by a downward trend in the discharge line temperatures and stable Quench Tank parameters. We test was required even though both valves had been successfully set during Precore Hot Functionals, because they were disassembled for addition of stem position indication. Disassembly voided the previously obtained settings. ( 3.45 l l 1 i l
3.1.8 PRESSURIZER SAFETY VALVE TEST (23B-313-02) (Continued) ) Besults ne lift pressure of both 2PSV0200 and 2PSV0202 fell within the acceptance criteria of 2500 + 25 psia with ro adjustments necessary. Results were as follows: LIPT 2PSV-0200 2PSV-0201 1st 2499 psia 2519 psia 2nd 2510 psia 2513 psia %e seat tightness after lift was verified by nonitoring the quench tank pressure, level and temperature following the test. No changes - in these parameters were seen, indicating the safety valves reseated - tightly. g Ji Problems and Resolutions Sane mino equiFaent problems were experienced durirg this test. %ese prob'.e ns basically involved improper equipnent fit-up, requirirx3 sane modifications before cannencing the test. In Mdition a hand operated hydraulic pump, used to provide pressure to the hydroset unit, failed. his pump was replaced, and the test was recomenced. Cbnclusion ne setpoint and seat leak tightness were properly demonstrated. No Mjustments were required. %e FSAR conmibnent was satisfied. 3.46
3.1.9 PRESSURIZER PEREORMANCE (2HB-212-04) p yl
%e objective of the Pressurizer Performance test was to verify proper operation of the pressurizer pressure and level control -
systems. We test was designed to rigorously test the pressure , and level control systems including: proper annunciator operation; 2 prooer operation of the Auxiliary Spray System; satisfactory 2 operation of the Letdown Flow Control Valves (2LV-0110A & 2LV-01108); and proper operation of the pressurizer spray and heaters. The test satisfied Section 14.2.12.54 of the FSAR. Actual testing was performed on several dates which spanned frm June through Septerrber,1982. Although numerous problems were encountered, all the acceptarce criteria were met. Test Method
%e Pressurizer Performance Test was divided into three major sections: pressure control performance, level control performance, and auxiliary spray performance. Pressurizer level and pressure control loops are shown in Figure 3.1.9.1. %e pressure control loop performance section was performed by taking manual control of various portions of the control loop to change pressurizer pressure and actuate various equipnent.
Nine problems were documented on this portion of the test. All were concerned with calibration of various setpoints within the control loop.
% e level control performance section was divided into two sub-C sections. %e first manually changed the pressurizer level error (actual pressurizer level minus level setpoint) and verified the setpoints of the various control functions. W irteen problems were documented, consistirg of maladjusted setpoints of the various pieces of instrumentation in the level control loop. "
%e setpoints were adjusted. *
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B e second subsection concerned minimum and maximum letdown E flow rates. %e acceptance criteria was a minimm letdown flow l1 of 29+7.5 gpn and a maximtzn flow of 138+7.5 gpm. % e design of
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the letdown system and various problems with the letdown flow - control valves, 2LV-0110A and B, led to unstable letdown flow j rates, although the acceptance criteria were met. E 5 n e auxiliary spray portion was successfully performed by opening the auxiliary pressurizer spray valves to verify operability. Results
%e acceptance criteria were met after fine tuning of the pressure and level control loops. .We Auxiliary spray portion of the Pressurizer Performance Test was performed satisfactorily. ~
3.47
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f 3.1.10 PREHJRIZER SPRAY VALVE AND CCtCROL ADIUSIMENT (2HB-313-01) he objectives of the Pressurizer Spray Valve and Control Adjustment Test were to establish the proper settings of the continuous spray valves and to measure the rate at which pressurizer pressure could be reduced using normal pressurizer spray valves. %e acceptance criteria for both objectives were met, and the test satisfied FSAR Section 14.2.12.79. Test Method
%e first objective was to establish proper settings for the m continuous spray valves, also known as minimum spray valves. We .
valves are 3/4" needle valves that parallel the main spray valves. lE t ey permit a small spray flow which bypasses the main spray valves, & 2PV-0100A and 2PV-0100B. Se purpose of continuous spray is to maintain the spray line sufficiently warm so that thermal shock to the pressurizer spray nozzle is minimized when the main spray valves open. %e small flow also slowly mixes the pressurizer with the remainder of the reactor coolant system and encourages more uniform boron and chemical concentrations between the two. We bypass and main spray valves are shown in Figure 3.1.10.1. ~ Minimum spray flow was adjusted such that a small temperature loss d V was realized frca cold leg temperature to spray line temperature " measuring devices mounted approximately seven feet from the cold legs. We tenperature devices, 2TE-0103 and 2TE-0W4, were augmented by type J thermocouples strapped to the spray piping as shown in Figure 3.1.10.2. (It was discovered that the two spray line temperature measuring devices did not sense the coldest fluid in the line. 'n o additional thermocouples were installed and data from these therno-couples were used to satisfy the test requirements). %e original 1/4 turn opening on the g:ntinuous valves was increased to 7/8 turn in order to provide a 41 F decrease at the thermcouple located closest to the pressurizer.
%e minimum spray valve was not opened further (which would have decreased the temperature drop) because increased flow would have required greater pressurizer heater power to maintain pressure. With the valves opened 7/8 turn and insulation heat losses, one bank of backup heaters was required to maintain pressure. A balance between heater power requirements and tenperature decrease was obtained through iterative small adjustments of the continuous spray valves.Thefinaladjustmentresgitedinspraytemperatureatthe pressurizer of approximately 504 F (cold leg temperature was 545 F), with one bank of backup heaters (200 kw) and scue propor-tional heater capacity matching insulation heat lossen (described in Section 3.1.3) and continuous spray.
3.49
3.1.10 PRESSURIZER SPRAY VALVE AND 00tfrFOL ADIUSIMENT (2HB-313-01) (Continued) % e second objective was to determine depressurization rate using
~
each main pressurizer spray valve separately and then both together. 'Ihis objective satisfied FSAR section 14.g.12.79. Stable conditions of 2265 psia pressurizer pressure and 545 P cold leg tenperature were established and the valves were fully opened. The resulting pressure decreases are shown in Figure 3.1.10.3. The acceptance criteria of pressure dropping frm 2250 psia to 2100 psia in less than 93 seconds with both spray valves fully open was met. All pressurizer heaters were then energized to determine maximum repressurization rate frca 2100 to 2250 psia, which was 24.5 psi / minute. Were was no acceptance criteria for this part of the test. Results h e minimum spray valves were successfully adjusted as described above. % e depressurization rate was found acceptable. Problems Ihe original pressurizer spray valves suffered frm hydraulic instabilities, as a result of unbalanced trim design, which caused valve position to oscillate. W e oscillations lead to packing damage and forced cooldowns. We original 4" valve was changed to 2" and the trim was changed to a balanced design. We test was performed with the old and new designs and the results of depressurizations are plotted in Figure 3.1.10.3. While the new trim produces acceptable spray flow rate , maximum spray flow rate is significantly less than with the old trim. O 3.50
O - 8 A 2 2 P C R GC P R 2 = M . A4O f i . S N I C A 2 P 0 " 0 ' 1 M E T ? R S l P l O Y i Rl O ( S : t 5 5 E I R C t T P R lN iI it N- C SI S S A l g I V L O k R _ O ' _ 1 O I 0 1 O C A f 1
. C R 3
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l'igure 3.1.10.2 ISOMI'TRIC AND l.lNE DISTANCES SI' RAY I.INE TilFRMOCOtti'l.E 1.0t.AT10NS FOR SONGS 2 l'OSTCORE If0T FilNCTIONAI.S
+- 5':, ' -
#9 (On Top Of '
Pipe) n term couples are noted by dots ( e ) and
.# 13 (Itot t oni) except for #9 and #13 were removed after
#8
- 11' postcore hot functionals. #9 and #13 were used to adjust minimum spray valve opening.
inensi ns re approximate. Pressurizer o / (Typically ,, 2250 + 15 psia, ;
> 21' w approx. 65 3"F)
\,n- ~ it, e 1. f( m Access Walkway f6 10'
' #5 sr i Auxiliary
# ITSpray 1.inie Floor 45' Elevation u_.__ .;
./
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21'V 100lty TE103 2PV100A J' TEl04 y 1, 'j O (Main Spray Valve) } ;' l -',5
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#3 #7 MU043 #2 g{ MUO42
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To RCP p001
i Figure 3.1.10.3 DEPRESSURIZATION RATES Case Valve (s) Trim. Depressurization Rate Time
- l' 2PV-0100A ~ new 1.29 psia /sec 116.28 sec
-2 2PV-0100B new 1.32 psia /sec 113.79 sec 3 2PV-0100A,B new 1.82 psia /sec 82.82 sec 4 2PV-0100B old 2.11 psia /sec 71.17 sec-5 2PV-0100A old 2.08 psia /sec 72.06 sec 6 2PV-0100A,B old 2.28 psia /sec 65.78 sec 2260 ,
--- ~ " ~ - ~ ~ - " '~ " ~~
2240 - - n --- .
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2100 - 0 20 40 60 80 100 120 l Time (sec) 3.53
3.1.11 PRIMARY AND SECONDARY CHEMISTRY (2HB-213-03) Objective 'Ihis test was performed during the Post Core Hot Functional Test period " to establish proper control of primary and secondary water chemistry.
- By doing so, station sampling and analysis procedures were verified to .
l ,_1 be a3 equate. 'Ihis verification satisfied FSAR Section 14.2.12.76 and 14.2.12.77. In addition, the necessary sampling frequencies were 5 established to enable chemistry to be maintained within Technical Specification limits. Test Method This test was performed throughout the Post Core Ibt Functional Test period by monitoring the chemistry program. Chemistry analysis sheets were " reviewed daily to ascertain if the parameters were being maintained within - prescribed limits. If they were not, appropriate actions were taken to J correct the problem. o [ In order to verify that the analysis procedures were correct, a single sample was analyzed by both the station chemistry department and an independent lab. Table 3.1.11.1 contains the primary chemistry specifications and Table 3.1.11.2 contains the secondary chemistry specifications.
% st Results In general, primary and secondary chemistry wre maintained within the required limits throughout the Konitoring period. During a raintenance outage, however, scue parameters drifted slightly outside of these limits.
All parameters were verified back in specification per station chemistry procedures prior to reentry into the hot, zero power (Mode 3) testing plateau. 3.54
. . _ - . _ . . - ~ _ _ . . _ _ _ _ - - . _ . . . . _ _ . . _ _ _ _ - _ _ _ _ - . - . . _ _ - _ _ _ _ _ _ _ . . . _ _ . - . . . . . _ . - . _ _ . . . _ . _ . _ - _ . - .
T
- j. l
. I e' I i
e k I, TABLE 3.1.11.1 PRIl%RY CHEMISTRY SPECIFICATIONS j PARAETER SPECIFICATIOi NCTIES pH 4.5 to 10.2 4 ] Li+ 1.0 to 2.0 ppn 0 <10 ppb With RCS >l50 F 2 j Susperded Solids <.5 ppn Up to 2.0 ppn allowed during startup i j Cl~ <.15 ppn 1 ~ { F1 <.10 ppn NH 24 30 to 50 pp, With RG <l50 F 1
! E 3
<.5 ppa j Gross Activity <1Q0 g Ci/g E is the average energy of the j selected isotope i
i i i t i i } } t I 1 l 1 I d I 1 1 1 i } a l 3.55 4 1 u
TAnrP. 3.1.11.2 SECmDARY OIEMISTRY SPECIFICATIONS 3 STEAM GDiERA'IORS, MODE 5 PARAMETER SPECIFICATION pil 8.8 - 10.5 NH 75 - 200 ppm 24 Dissolved 0 2 <100 ppb Cl <1.0 ppm Na* <1.0 ppu N Overpressure >5 psig 2 Cation Corductivity <10 u: rho /an STEAM GENERA'IORS, MODE 1 PARAMETER SPECIFICATION pff 8.5 - 9.5 ~, NH 10 - 50 pEb 24 Dissolved 0 2 <10 ppb Nf <I # 3
~
C1 <100 ppb Na+ <100 ppn SO <100 p@ 4 Cation Conductivity <2 prrhof cn Specific Conductivity <10 umho/cm Silica <300 ppb Suspended Solids <100 ppb
~ *These are examples fran Special Chemistry Procedure SO2-SPC-004 Rev. O. They are ;
n to be used only during Startup. Limits are note restrictive durirq camercial operation. y 3.56 9
[
'% 3.1.12 STEAM GENERA'IOR FEEDWATER RING INTEGRITY TEST (2HB-201-01)
We objective of the Feedring Integrity Test was to demonstrate the ability of the steam generator feedwater ring (or sparger) to withstand the introduction of auxiliary feedwater with stem . generator water level below the feedring. S e test was conducted on June 12, 1982, after the feedrirs had been strengthened and replaced following collapse frca similar testing during precore hot functionals. An inspection of the feedring on June 21, 1982, showed no damage. M is test satisfied FSAR section 14.2.12.72t.
Background
%e feedring is a circular pipe tealve inches in dimeter which ,
distributes feedwater in the downcomer portion of the steam generators, as shown in Figure 3.1.12.1. All feedwater enters the steam generators through the feedring. . During the precore hot functional test, water , level was lowered and maintained below the feedwater ring for two hours. It is postulated that the feedwater ring drained such that when large flows of auxiliary feedwater were introduced into the , steam-filled pipe the resulting pressure reduction caused the feed- ' rity to collapse. Auxiliary feedwater was introduced in a manner Wich simulated the normal initiation of emergency auxiliary feed- , water, large flows (>800 gpn) in a fairly short time (<30 seconds) . O Although no unusual sounds or signs of damage were observed during the precore test, an inspection later revealed that the feedring had been crushed. We replacement feedring enployed several improvements which strengthened the feedring, opened greater flow area to steam, and sealed drainage paths for water. %e original feedring was 12" m Schedule 40; its replacement was 12" Schedule 160. Weorigingl76, ; 1" diameter elbows were replaced by 40, 31/2" extra strong 90 a elbows. Finally the thermal expansion slip joint at the junction S between the nozzle and feedwater ring was replaced by a mechanical seal which was intended to minimize leakage. Test Method
%e test method was to lower feedwater level below the feedwater ring for approximately twenty minutes, then refill with auxiliary feedwater as during the precore test.
O 3.57 l I i
3.1.12 STEAM GENERA'IOR FEEDWATER RING INIERITY TEST (2HB-201-01) (Continued) Results tb indications of waterhmmer, such as noise or vibration, were observed during the test. A subsequent visual inspection of the feedring itself when the plant was depressurized and cooled also showed no datage. O O 3.58
Figure 3.1.12.1 LOCATION OF FEEDRING IN SONGS 2 & 3 STEAM GENERATOR STEAM N0ZZLE = STEAM STEAM DRYER DOME (TYPICAL) T UPPER LEVEL TAP MOISTURE SEPARATOR (TYPICAL) q 3 (100%) { FEEDRING ^; l'N TOP OF ELBOWS FEEDWATER N0ZZLE % RISER 's ON FEEDRING
-- EL.71'-2 3/4" C
,k g gg } ,
46.8%
-) l j -* = 1__~ BOTTOM OF FEEDWATER LINE ,4 (, \ FEEDRING T
L 36.6%
,C EL.66'-11" - -
4 i TOP OF TUBES u 36.7% ~ O 00WNCOMER
- EVAPORATORj ;
j y l' JUNCTION 23% 2a WRAPPER-- _ ),, REGION
. ;q
! LOWER LEVEL
.[ ; j.
i TAP (0%) h ll . I { o it , i,L ,, .
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GENERATOR FEEDRING
/ WALL
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- SLEEVE PLAN VIEW OF FEEDRifiG
/ '
\ \ HOT COLD LEG 3.59
1 I l 9\ l 3.1.13 PRESSURIZER SPRAY EFFECTIVENESS See section 3.1.10, PRESSURIZER SPRAY VALVE AND C0tRT<OL paIUSm. O 3.60 l l
- _ .- _ _ _. _ _ _ - _ . . , _ . - . . . . _ _ _ _ _ _ , . ....~._. _ . _ . _ . - - _ - . . . . . ___-.m____. i i 3.2 PI. ANT TESTS te following describes two balance of plant tests. % e balance of - plant generally includes all but the Nuclear Steam Supply Systen i and ordinarily includes everything outside the containment. i t i [
? I l
-i i
I v k t p t i I I P 3.61 l l
3.2.1 AUXILIARY FEEDWATER PUMP NRIY-EIG1T HOUR ENDURANCE IUN (2ST-235-01) We objective of the forty eight hour erdurance run was to operate all three auxiliary feedwater punps ard denonstrate that the pumps remained within design limits with respect to bearing temperatures and vibration and that pump room anbient tenperature did not exceed environnental qualifications limits for safety-related equignent in the roon. We test satisfied commitments stated la NUREG-0712, Sup-plement 41, SONGS 2 & 3 SER. Test Method te test method consisted of operatiry the pumps for a minimum of forty eight hours. We plant had to be in tot standby corditions to supply stean to the steaMriven auxiliary feedwater punp. We main stean system was the only source of stean. Ibsults Alloftheacceptancecriteriawegemet. We criteria included (1) bearirg temperature less than 180 F, (2) motor vibration less than 1.5 mils peak-to-peak for the notor driven pumps, (3) turbine vibration lesg than 1.5 mils peak-to-peak, (4) punp room temperature & less than 104 F, and (5) punp vibration less than 1.0 mils W peak-to-peak. Problens ne babbitt metal bearirgs which were originally supplied with the notor driven pumps were replaced with iron bearings. We iron bearings siezed and were replaced with babbitt. Af ter bearity replacanent, + there was excessive vibration on the two notor driven pumps, but, after realigrinent, vibration was reduced to acceptable values. 3.62 O
l l l l 1 (G l 3.2.2 Power Ascension Data Record (2ST-344-17) m . Re Power Ascension Data Record provided the means to nonitor and record {
, primary and secondary plant performance throughout the test program, as y required by FSAR Section 14.2.12.101.3C. Records were maintained in'the form of plant ccuputer snapshots taken at regular intervals during steady I state operation, and trend data collection of critical parameters during power changes. A plant snapshot is a feature in the plant ccmputer which gathers the value of many plant parameters essentially' simultaneously and presents a reasonably complete picture of conditions in the plant.
%e major paraneters monitored include:
Reactor Power Generator Power Control Rod Positions Reactor Power Distribution Parameters RCS Temperatures and Pressure Secondary Pressures, Temperatures, and Ievels Reactor Coolant Punp Parameters 01arging and I4tdown Parameters ' Safety Injection Parameters Condenser Parameters Main Feedwater Parameters i Feedwater lieaters Status Generator Parameters Containment Atmosphere Parameters l All data was recorded on the Plant Canputer. Data collected during the y performance of Power Ascension testing was filed and maintained for future g reference. %e Power Ascension data record was performed through the " ctapletion of 1004 power testing. 3.63 l
-s - _ . - _ _ _ _ _ - ___ _ _ _ __ _.___..___.._____
3.3 COMMUNICATION TESTIE
'Ihe followirq acxumunications systens testing was performed to assure that the various otanunication ard warniry systens wre usable and atdible.
O I I 3.64
= __ - -
1 3.3.1 WF RADIO COM10NICATION SYSTEM (2AC-483-01) Cbjective Acceptance Test Procedure 2AC-483-01, Rev.1 satisfied the requirements of the Final Safety Analysis Report (FSAR) Sections 9.5.2.2.1.6, 9.5.2.2.3.1 and 14.2.12.4. Method te test demonstrated a$ equate, reliable cannonication of the WP portable two-way ra3io systen between designated locations throughout the plant, and fran fixed radio coneles in the Control Room and Central Alarm Station (CAS) to various designated locations in the plant. te' fixed radio consoles in the control room arti CAS were also verified to be accessible to the Unit 1 WF channels.
' Dest Results Clear and cancise connunication via the WF channel, was established between the off-site United State Marine Corps (USMC) Station and the fixed console in the Control Iban and CAS.
Testing the effects of WP radio frequency energy on low voltage solid state relays and controls resulted in an i engineering decision to prohibit the use of portable radios in designated areas of the plant where the effects would cause permanent damage to Plant Equipnent. te areas where radios are prohibited are: Remote Shutdown Panel 2L-42 Main Control Room Diesel Generator Buildings Canputer Rooms Main Generator Excitation Panel CEW Panels l l 3.65 O
3.3.2 INTERNAL TELEPHONE SYSTEM ACCEPTANCE TEST (2AC-480-01) l Cbjective ne test satisfisi the requirenents of FSAR sections 14.2.12.4.3. It demonstrated the ability of internal telephone systens to provide conmunication between various plant locations as follows: a) Proper operation of the plant comon battery telephone systen to provide comunication between priority operating locations. b) Proper operation of the internal PAX telephone systen to provide autanatic dial telephone facilities at all installed cperatirg locations throughout the plant. c) Proper operation of the intercan systen includirg verifica-tion of ability of master stations to originate calls with other master stations or speaker stations of the intercan system. Besults %e objectives listed abwe were met. 3.66
O 3.3.3 AREA RADIATION MJNTIORING SYSTEM AUDIBILITY (2ST-340-01) Objective - We objective was to deonstrate that the au3ibility of the ARMS alams was sufficient to alert personnel locally and in a3jacent access areas of a radiation hazard. We test satisfied no FSAR comitments. Method Each alam was initiated either remotely by depressing the TEST pushbutton on panel 2LO91 or by locally jmpering the alam. We perimeter at which the alam was easily audible above rurmal plant noise was then marked on floor plan sheets. his infomation was baseline data only and there were no specific acceptance criteria. O V n 3.67
3.4 CONTAINENT ISOLATION VALVES TEST (2HB-101-04) Se Safety Injection Tank Drain Hea$er to Refueling Water Tank Valve 2HV-9334 was tested for proper operation by this procedure. Containment Isolation Valves must be able to be opened and closed frm the control room, autmatically close upon a Containment Isolation Actuation Signal " (CIAS) or a Safety Injection Actuation Signal (SIAS), have a closing time <10 seconds, and fail "as is" upon a loss of power. Eis procedure J satis?ied FSAR Sections 6.2.4.2.5 and 14.2.12.32. % is procedure was 2 written as a retest to valve 2HV-9334 after it underwent modification due la to a design change. Method Each of the valve's required characteristics were tested in separate portions of the procedure. An engineer was stationed at the valve while a control rom operator cycled the valve open and shut. Eis ensured proper operation by the control rom was possible. - Se breaker for valve 2HV-9334 was opened while the valve was shut to observe ; whether the valve would remain shut. Powerwasrestored,thevalvewasopened,l3 2 and the breaker reopened. Again the valve failed "as is" and remained open. Operator surveillances require p riodic simulation of CIAS and SIAS signals to insure proper operation of all containment isolation valves. A copy of the nost recent empleted surveillance was attached to the procedure to - fulfill this portion of the testing requirements. Using a jumper, the surveillance simulated a CIAS signal and then verified that the proper valves l;g closed. Se procedure was then repeated using a SIAS signal. No problems y occurred during this portion of the test. Test Results The final part of the procedure required a closing time of <10 seconds under full flow and no flow conditions. A stopwatch was used to time trm when the control rom handswitch was pushed to when the limit switch on the valve actuator gave a closed indication. Under no flow conditions the l"J time was 9.67 seconds. S A procedural problem was encountered for the full flow condition stroke time l5 test. Se procedure required draining all four Safety Injection Tanks simultaneously to the Refueling Water Storage Tank. This would risk a violation of Technical Sgcifications if the SIT's were drained too far. l~
@e flow path was changed to the normal lineup for filling the SIT's. A J HPSI pump was used to provide the flow condition. W is change avoided E possibly draining the SIT's too far. Under full flow conditions the stroke 5 time was 10.0 seconds.
3.68 L
r 4 T i i ! Problems No mechanical or electrical problens were encountered durinj this procedure. l The flowpath was the only procedural problen encountered. All acceptance j criteria were met and the procedure was closed. i e L i i i i i l 4 i E I .. i 4 i i l i 1 k
- i t
l 3.69 l 3
it 1 s I t 1 I ( i a t I SECTICE 4.0: INITIAL' CRITICALITY 4 J l [ } I 1 1 i, (~. i
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i .I A 4 4 u I h, i I \ I. i 4.1 I h e- +--,r-----w- - - -ww--w.yw ,-w- w,+1-www--<e - e --e--ww m--weree-.,. .....e-
4.0 INITIAL CRITICALITY (2IC-301-01) AND CEA EXERCISES (2IC-301-02) Suninary Initial criticality satisfied the objectives of FSAR Section 14.2.10.2 by providing a safe organized method for attaining initial criticality and verifying that the startup nuclear instruments and logarithmic power circuits of the plant protective system exhibited at least one decade of overlap. Initial criticality occurred on July 26, 1982 and was the first time the SONGS 2 reactor attained a self-sustaining chain reaction. It was performed under hot shutdown conditions of 325 F ard tiOO psia in order to bring the plant to conditions required for the low temperature portion of low power physics testing, which followed immedi-ately. SONGS 2 is the first of the Cmbustion Engineering 3410 Megawatt - reactcrs; the low-temperature initial criticality and low-temperature , low power physics testing are not anticipated to be repeated on Sorgs 3 2 or any other 3410 megawatt plant. j CEA exercises were so closely intertwined with initial criticality that the two evolutions will be reported together. Test Methcrl Initial criticality was attained in five steps which consisted of: (1) demonstrating operability of the reactor trip paths, (2) fully with-drawing Control Element Assemblies (CEAs) while acnitoring the core's inverse multiplication ratio via startup neutron detectors, (3) denon-strating CEA Group overlap, (4) reducing the boric acid concentration of the Reactor Coolant System (RCS) by dilution, and (5) withdrawing CEAs to achieve criticality. The original intent of the procedure was to bring the reactor critical via dilution. However, nonents before criticality, a chemical sampling of the Volume Control Tank (VCr) erroneously indicated high oxygen and low hydrogen concentrations. 'Ihe VCT is used to adjust the che.ttical cmposition of the reactor coolant. High hydrogen and very low oxygen concentrations are required for proper chemical control with a critical reactor. As a precautionary measure, the CEAS were inserted to avert criticality, and additional samples were taken. After acceptable concentrations were verified, the reactor was made critical by withdrawing CEAs. 'Ihe test procedure provided steps to bring the reactor critical using CEAs, so no changes to the procedure were required. Special counting equignent was necessary to collect data during the approach to initial criticality due to the low count rates provided by installed plant excore neutron detectors. 4.2 O
c 4.0 INITIAL CRITICALITY (2IC-301-01) AND CEA EXERCISES (2IC-301-02) ( ) (Continued) v Test Method (Continued) Inverse Count Rate Ratios, also called Inverse Multiplication Ratios, are used to predict approximately when criticality will occur. 2 e concept is that count rates increase to large values as criticality is approached. If a neutron base count rate taken when the core is subcritical is divided by those taken as the reactor approaches criticality, the ratio will approach zero. Extrapolation of the line to zero will provide an estimate of the point (time, boron concentration, or CEA position) at which the reactor will becane critical. %e simple concept of inverse multiplication ratios was used extensively during fuel loa 31ng and initial criticality. During the initial criticality procedure, they verified the reactor would renain sub-- critical during CEA exercises and later provided estimates of critical baron concentrations and CEA positions. Overlap of neutron detectors means one detector channel registers a reliable signal before another channel reaches its upper limit. We two detector types are high sensitivity startup detectors and low-sensitivity, large range logarithmic detectors. Startup detectors are used to monitor the very low neutron fluxes when _tge reactor is far fran criticality (significantly sub-critical = 10 %) . As the ing at extranely A reactor low p3werapproaches levels (on criticality the order but is still oper_ag%) of kilowatts = 10 , the logarithnic
) detectors will start to resInnd. Startup detectors can be danaged if used in high neutron fluxes, which exist when the reactor is at p]wer. Rey age autanatically deenergized at a predetermined response level ( = 10 %) . Before they are deenergized, the logarithnic detectors must be reliably operating. One order of magnittrie, or decade, of overlap must exist. Eis procedure verified such an overlap. )
'O 4+}
4.1 CEA EXERCISES (2IC-301-02) t e first step, demonstrating operability of reactor trip paths, was O performed by raising all CEAs to six indles withdrawn and manually trip-ping the reactor. W e CEAS were observed to return to their deenergized (fully inserted) position. Step two, consisted of fully withdrawing CEAs and verifying that the core was still substantially sub-critical. It was performed by withdrawing single groups of CEAs in approximately equal reactivity steps of one per-cent and nonitoring the resulting increase in count rate and decrease in inverse count. rate ratio. Each group was withdrawn in turn, starting with Shutdown Group A and ending with Regulating Group 6, the final CEA Group. Se second step was the first time CEAs had been noved in Groups, which consist of four or nore CEAs. All previous testing had involved noving only single CEAs. Step Three, denonstration of group overlap, was performed with the CEAs in Manual Sequential. CEA group overlap occurs frm 0 to 50 and frm 100 to 150 inches withdrawn. W e most reactive axial segment of each group (frcm 50 to 100 inches) is not overlapped. Overlapped novement results in more uniform reactivity additions than is possible with individual group CEA notion. Power Dependent Insertion Limits (PDILs) were also verified. PDILs are limits on group insertions which depend on reactor power. W ey permit less ' insertion for greater power, such that, at full power, the CEAs must be alno fully withdrawn. PDILs and Group Overlap limits are generated from the plant computer which senses CEA position and reactor power. Steps One through tree occurred with the Reactor Coolant Systen at refueling baron concentration. Refueling boron concentration guarantees that the core will remain subcritical even with all CEAs fully withdrawn. Physics nonitoring was perfonned on the initial withdrawals so CEA interlock checks could proceed with little additional nonitoring. 4.4
4.2 DILUTION 'IO CRITICALITY O CEA Group 6 was then stationed approximately at core midplane, or 75" withdrawn. All other groups were fully withdrawn. Step Four, dilution of reactor coolant was started at a rate of 2 parts per million (ppn ) per minute. Wis brought the RCS boron concentration frcm refueling concentration (>1725 ppn) to approximately 1250 ppn. After a stabilization period of approximately four hours, a slower dilution rate of approximately 1 ppn.per minute was used until approximately 980 ppn. A five hour mixing period was followed by a dilution at approximately one ppn per minute to the estimated critical concentration of 830 ppn. Criticality would have been attained at approximately 870 ppn had CEAs not been inserted. SONGS 2 has no direct dilution path to the suction of the charging pumps, which would introduce demineralized water directly into the reactor coolant systen. All dilution must first enter the VCr. %e VCT was filled such that RCS boron concentration would reach the predicted criticality concentration of 830 ppn. W e resulting CEA notion to remain subcritical placed Groups 6 and 5 at their lower electrical limit and Group 4 at 60 inches withdrawn. Figure 4.2.1 shows inverse multiplication plots and boron concen-tration for the dilutions. Startup channel number one is shown although the other channel was similar. f) V We dilution frcm 1750 ppu to 830 ppn was periodically halted to enhance mixing between the RCS, pressurizer, and VCP. %e pressurizer receives small flows through the spray valves and requires hours to equilibrate with the main portion of the RCS. W e VCr and RCS mix at a faster rate but approximately one hour is still required for mixing. We three cmponents must be at chemical equilibrium; otherwise, RCS boron concentration can change significantly and require unde-sirable CEA notion, j 4.5
Fi:pire 4. 2.1: BORON DII.trl'10N EVOl.IITION: INVERSE COUNT RATE RATIO AND BORON CONCENTRATION VERSilS TIME (@ Co/Ci is Inverse Count Rate; 8 PI'M is Boron Concentration) l 1. 2 - ;, . r T --------~T
--- -~ -- -
2000 I t l _ - -
^ ~ ~ ' ~ ~~ ~'
1.0- b'g 1750 NLS-
] w +1 abLl '@ /,
2 ppt'fllNUTK liiff El CN GTI- _ 7
, ~
i
-]i- -
t o 0.8- 1 - --- - - ---- - --
-1500 t
m
} i t}
ppm C ) ING PI- R I l);) , f* o I o
-% g l 1 7 7 Boron
' ~ ~ ~
' '-'~ ~
m C. 1 N}.
.m
/
f -
~
l U 0.6 + - til 7-- ---1[c ro $ -1250 e wc a e ny 3
,w, m,m n - - -
-7 ; . _
o - 7..
.8 f~-g
<n i 9 y ;, l l- ly/g b tj 1 P. hi l.tH' l0B RUl: b-
.g s 7-3p - -
g h -
;Eh -
s
- ~ ~
7 $ 'G--._s
~
F
~
~
H 0.4 -- -- - 4--A -- SICIJEEU)l 1 1000
.__C,9' 1 ;I :...
E-~O @ El fl B 1 o EQ ;j atm10N- ] i % Mn f ; _ k ' S
'e -I 4 O cfq+oyg . -
hig) g3 y-r L.___ f 0.2 ---- - ---- - -.--- -- _____
-750 Q ( hs
.EMHFEER]TIO i,
p . u. ; O.0 . I - l-- r -- - r - - - t - - - 3--_- ;____-__ . . . _ . _ .
, rg 180g27 22m my 0200 0400 gd 0800 IM) 12W 1400 00 1
- 4.3 WTIEDRAWAL 'IO CRITICALITY i After boron concentration had stabilized, control rods were withdrawn in steps, and the reactor was brought critical in the fifth and final part of the evolution. Plots of inverse multiplication ratios for CEA withdrawal to criticality are shown in Figures 4.3.1 and 4.3.2. At the start of the withdrawal' plots, inverse cultiplication was normalized to unity.
i r t t O i
?
r d i i a f i l i J l 4.7 I l L
Figure 4.3.1 INVERSE COUNT RATE RATIO VERSUS CEA GROUP POSITION DURING CEA WITHDRAWAL TO CRITICALITY Startup Channels Only rr i r rr i . , , .iii , ,
' i r i 3
! ' ' i 1 ir i i i . , , ,
' t 8 i 3 i P i i i I 6i i i i i i i a e )
. i , . . , ,, ,
~
, , i 1.04 , ,
dx \e. .
. l w .
.. . . i
.C t- n + m &wnl c: i .
\\- [dahgiel i i 0.8 Y '
u '
. . i i ,
. .N. xs ' .
,>, i , . , , ,
i x\! , 3 v i i ,
\\ > ! . . i , i xM .
\Mi i . .
xA\' ' * , M\> . , , , i i o 0.6 , \A \ .' ' i , W ' ' .
!\'M. ! i e t I i i! i .
b i
!1 1\A . , e i . i =
- mx , , . ... , , i.
C . x x- , ! . , , 1 , . , ,, e _,0 _
.xx..,i . , , . , i
' ' '\\ ' '
H C '
< i \\
0.4 ! W
=
o :xu i U i : KAN x - mx r.n - \ x q XM a ' . X .\
> i I %\ \ t >
5 0.2 '
' x\xAW'i
! k l ~CFi E lief
! MClarLNi 0.Q-60"
. 4-70" 4-80" 4-90" 4-100" Y
4-no " CEA GROUP POSITION (INCHES WITHDRAWN) 5-0" 5-10" The count rate on the Startup detectors was sufficiently high that statistical scatter was minimized; the graph is fairly linear. Co was the background-corrected count rate at the beginning of the withdrawal and Ci was the back-ground corrected count rate at each position shown. Group positions are shown; 4-60 means, for instance, Group 4 was 60" withdrawn. 4.8
Figure 4.3.2 g(-- INVERSE COUNT RATE RATIO VERSUS CEA GROUP POSITION DURING CEA WITHDRAWAL TO CRITICALITY Safety Channel Detectors Only 1.2 , ,
., , . . 2 .. i i, ! .
i r i ;w i ! i _ i ! 6 i i i i. 4 r i! t ! i4 . W i. . . ! 4 *ie
- ii I i 8
! i i ! '
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./ g i , ,
. ., , i !*i i !
i
; fi -{ 6 3 } ; e ; e i i i 6
- f. g i i i i > i i 1 1.0g ' '
1, N 1 N. .\ . A ,I + NA 1- ^_cnfggy ch,nnal A* g: _ ,, !d ; , i itt
- \ pi i /.\-sar t:Ly %naHUEIL U;
\-1
- i i i . i i !
0.8 Vi
\l: .
! e e i
\11 . .
, , , i ! ! . . e i gg- , ! . i . i i i 6 i
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! . . , i i i t i , ! .
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,% gg , , i ! , 4 t . ii . . . .
Z l\. > i i i I + t i i i F
< 0.6 'i k im . ,
+ ' '
(:
. . 6 6 ia s s co i , . . w - , . .....+ - . . .
i < 1 .
, MT .! ! i .
i 2 H z {i ' g i ,
&t . . .
n
; ;\ P:
.Ai!
.\
i t i '
! t i t !
I , ; , I i 5 /E\ ; Y O 0,4 '_ Nw .x
. v x-
.g.
, \
m 1 ' "q' t ! x \ N i . w \ .
> ~\' d h_ ~Nx.. 4 Y\ . v$ n.. .
0.2. X
\ - x
\
b x.
^
\\ cr4rien14 ry R
! ~Deictared
. n p
0.0'4-60" 4-70" 4-80" 4-90" 4-100" 4-n 0" CEA GROUP POSITION (INCHES WITHDRAWN) 5-0" 5-10" Special counting equipment was used to moniter the extremely low count rates from the Safety Channels. The count rate was sufficiently low at the start of CEA~ withdrawal that statistical variations were important and points on the 8raph were scattered. As criticality was ap-O proached- count ( and statistical variations became less important. rate increased substantially 4.9
. _ . _ __ _ . . . . - - . . . _ _ _ - . . - ~
Q. l l l
- 4.4 NEUTBON DETECIOR 0/ERLAP During the dilution and subsequent CEA withdrawal the startup and logarithnic neutron detector channels were observed to exhibit at least one decale of overlap. Figure 4.4.1 illustrates the overlap.
O i i l i 4.10 l l
Figure 4.4.1 STARTUP AND SAFETY (LOGARITHMIC) CHANNEL OVERLAP { r 4 i . _ . . . . . . . . - . . . E' -EhE'- ~;b" * * . - - .
- s. "/ ^a
3 :'2 -- T'=~17 .' -13EE;:553: ri. *- -#~2 . .1*' #27=-....-"7" " "N - O= i ; - F c._ == gw = ==gg - . - _ . . .----}C_-~~~' - - - - - - -
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== q.===n i- .:n:1=b=y .if
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j I {g .~..-i. .' yr +.- ONE DECADE OF OVERLAP % l7-- /
-T- '
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/
f 7 ir ,
/ i g ==ti i u;- - ss - g= .- = =.:ii;i -s _ = == = = =- =2 " = = = - = = " - - - -
-j--@ = 4r 1X10 1_
=s -me = 5 =s; .- asE-__-__ rii: di=
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= -
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.-..-...._;.=.. 7. _ u._
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- -+ f . .= c ;;c.:21
====:,==;.. a.:a..
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f s= =v ..=.=.: -=
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m=!__. =I== := = ==
= = -E=~r == ==_:+=7 3 ':=r-- fi W.d : W i W--L=T_ -Tl i == EE- .=:fr- "}. c
- - n .=.=__=.11.: _._ ._ _- i.i._ii.c,-1__--- .-. . id_ -
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----_.p_____.r..----__.-t
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; 9. .=_"h. =u :_ t= _
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,-- _ ._ g _,
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- _ . . .g gtartup V; W i
i Channel I- I
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2 (CPS) w < -
.- /
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. - pep q g w_=ii==t u== =w== M-- . - . m. i- wp=-g p=m::n ig=.= m = .. -
=
IX10
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. - . .2
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=. y 7'~7.:.Mi.M=t5~. _ . , _. . _ . _..
. _""_Z _,
---- ~~-P"=E-f]
n= _ F :iii = L= . ._ ] 7
= .:: = -+
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2 4 - -- _: : 4._s. m g=. 2.:=_._,,.
--- , =-
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- - - - - - ,r,_, m e
a ...
; =;T_ . . 3,:._.._
.. .. +b :
M N N $-id iib}'!b
.:.. : . _- .. :. _ : . ..=_ : =_- _:_==.. _ _ _ _
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~
____ _-:_.iEII$$bl-bb bS~~dN_NM HN
- _._____..:= . = . . = . . ._ = _ . - . .a. .=._:
a d-M- -
- q. _;
. : . : : .;; .1= ,
=.=.:==-
- 1 =
Q ----'- - - - . . _ . .. - - -e- _.c==c =1-121: ;; -- - - -
-----: ~. ; _- ; == == - - . . n.= - - -
- 1. ;
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l== 1. :: =- -
- _ . . . =- == a=
= _ -_-=- _- -. .. -. _- -= = - .
.-~ .-- .
~~
....__.-t._.=
.- u . - . = .=. :.: ..+
= =
- - = :.:- t=f_g_
= u . __.-_f
=== + = =
+- = =- r --=
' . ^ _- _ . p' -== 3.2 :--
-p.- o
-- s-.
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- ,f O
= - - . .
IX10 ~ Z 3 4 5 6 i 8 9 1X10 '== 2 IX10 l 3 4 5 67 891 Safety Log Channel (Amps) 4.11 1 1 i I
1 i Results t e reactor was declared critical at 7:58 pn on July 26, 1982, with Regulating Groups 4, 5, and 6 at 110", 13", and 6" with-drawn, respectively, arx! all other CEAs cmpletely withdrawn frm the core. Ibron concentratign was 833 ppn, and reactor temperature and pressure were 326 F and 600 psia, respectively. An overlap of at least one decade was observed between the startup and safety channel logarithmic neutron detectors. Problems and Resolution Sme of the CEAs occasionally fell partially or fully into the core ~ while being moved. W e slipping and dropping was attributed to mechanical binding of the grippers. %e grippers are ratchet-like - devices in the Control Element Drive Mechanism (CEDM) which engace to R raise or lower the extension shaft. All ninety-one CEDMS are mounted S cn top of the reactor vessel head. W e extension shafts extend to the top of the core and are coupled to control elements (rods) which move in and out of the core. During normal novement, the upper and lower grippers are coordinated by electronics, such that one or the other grips the extension shaft except during a trip. Failure of - either gripper to engage as the other gripper disergaged caused the . extension shaft to partially (slip) or fully (drop) fall into the 2 core. Slipping and dropping CEA's prolonged CEA Exercises and zero
- S power physics testing.
Solutions to the problem involved first stationing technicians and engineers at the electronics control panels to observe electrical - control signals, which indicated if a gripper was moving when . cmmanded. If it was not noving, the technicians used pushbuttons to % override signals which would have permitted the other gripper to " @ disengage. Since the grippers nove in cycles, noving the CEAs 3/4 inch per cycle, the CEA could eventually be fully withdrawn even though one cycle was missed. We CEA withdrawal and insertion rates - were lowered frm 30" to 20" per minute. %e worst mechanisms were . replaced, a process which required cooldown and depressurization of E the PCS. 2 is occurred after low power physics testing. After @ replacenent, CEDM speeds were returned to 30"/ minute.
- 4.12
l t i j SECTION 5.0: W P0hTR PHYSICS TESTS - l J l f I ! t a f I 4 1 1 Il 5.1 } )
5.0 IIM POWER PHYSICS TESTS (2LP-333-01) Iow power physics. testing occurred frm July 27 to September 7,1982, and consisted of low power physics, natural circulation testing, and a biological shield survey. Se testing satisfied the reqaire-ments of FSAR Table 14.2.5. Iow power physics tests were perfonned under low temperature and high temperature conditions. %e low tenperature portion cmprised first-of-a-kind testing for SCtCS 2 and for the 3410 Megawatt generation of Cmbustion Engineering plants. Se high temperature portion expanded gpon the low temperature measgrements. During both the low (320 F, 600 psia) and high ~ (545 F, 2250 psia) temperature prtions the core was critical but was operating at pwer levels which were far below where sensible, or E measurable, heat was produced. Typical physics testing power levels were frm 0.01% to 0.1% of full power, or 340 to 3400 kilowatts of & core heat. During zero power testing, there was sufficient radiation in the containment frm the reactor, that it was pssible to perform a detailed radiation survey of neutron and ganma radiation around the primary coolant system and outside the biological shield which surrounds it. Zero power testing was the first time appreciable radiation fields existed. O After the low power measurenents had been perfonned, reactor power was raised to approximately 3% of full power (approximately 102 Megawatts thermal), and low power natural circulation denonstrations were performed. neir main purpose was to familiarize reactor operators with the operating characteristics of the plant under natural circulation. S e second of the successful tests is described in detail in Section 5.3, since it yielded a significant mount of useful infonnation thout the behavior of the plant. Figure 5.0.1 shows Operating Mode, Reactor %mperature, and Reactor Power during low power physics testing. Problems noted in the ccanent column are discussed in Section 5.2.6. 5.2
% FIGURE 5.0.1 LOW POWER PHYSICS TESTING un
.T ( ACTIVITIES AND TLME)
-- N $S
! ; e 4e. '
'v
/
7/23 4 320' 2IC-301-01 Initial Criticality comenced l CEA subgroup 11 breaker replaced. 7/24 i ,
- 2IC-301-02 CEA Exercises ll 7/26 l 2 .01: 2IC-301-01 completed. Reactor critical.
7/27- ! . l 2LP-333-01 Low Power Physics Testing comenced.
- l;. 2LP-701-01 Biological Shield Survey comenced.
7/28--- , ll'
+ i ,- End of low temperature portion of Low Power Physics 7/29 ,', l, Testing.
.x ..
.;i 545* Start of high temperature portion of Low Power 7/30-", , , , Physics Testing, 7/31 ', ,
. i.,
8/1 , ; i '
;, 9!' 3 Reactor manually tripped due to CEA problems.
Physics testing halted. 8/2 l! , ii [ ') ,,;,, l i
\s' 8/3 ,,,,'
.i i
', l i
i ' ? 8/4 ', ' 2 .01% 8/5 , Physics testing recommenced. t 8/6 i , :l 8/7 8/8 - 3 Shutdown for CEA speed changes. Physics testing halted. 8/9 , . 2 .01% Physics testing recomenced. 8/10 - 8/11 j. 3 Reactor manually shutdown. Physics testing halted.
,.1 Cooldown to outage started. j i
8/12 ,l /' 4
. li :
O 8/13 -- !/' l l ' , l i \'~' l I 5 8/14 ' 120' l l 5.3
hh FIGURE 5.0.1: LOW POWER PHYSICS TESTING
.g h h (ACTIVITIES AND TIME)
O O O r ec 8/14 -- 5 120 Major outage work:
' ' " " " **' *P' * #
8/15 - li; Cleaning CEA drives with Hydrazine Reactor Coolant Drain Tank relief valve repair l 8/16 - i i 4 8/17 - l'l, 'l i . i
,4 . ii ,
8/18 -
'l,'
. 4i 8/19 -( . ,' ,-
180* 8/20 e
'{,;l Started Mode 4 entry surveillances.
,i .
8/21 4 1 ,, 8/22
-j.ll. .' ii i 260*
8/23 -: s l
, . .... 340 E Started Mode 3 entry surveillances.
8/24 - .,2.l.. 1 ,
, Dl' 3 380*
8/25
,i,.
!Q!l ii l
545, 8/26 8/27: 8/28 . .' Started Mode 2 entry surveillances. 2 .01% Physics testing recotnenced. (Reactor critical) 8/30 8/31 - 9/1 9/2 9/3 2 .01% 2LP-333-01 Low Power Physics Testing complete. 9/4 9/5 - 2 .01% 2PA-34 -03 CPC M SS Ver Hication 5.4
L 1
% FIGURE 5.0.1: LOW POWER PHYSICS TESTING r , un o -
i e iII (ACTIVITIES AND TIME) 1- m a :s o oo E . H 3.
~
4
- 9/5 -
!i ' ' "
2 .01% ;
- . 3: 2LP-333-02 Natural. Circulation Demonstration commenced. !
! 9/6 i 3 Reactor trip on low flow. ,
j l 2 3% Reactor returned to power. Natural Circulation ! 9/7 i 3 I i l +1 - 2 3% i I 9/8 4'i' , ! 3 2LP-333-02 completed, i st i it E I k i : I i i I. e 4 j 1 4 1 2 e-i 1 i
!~
i. 5 1 I i ; 1 l t 2 i i , I 1
'5.5 i
. e I
, .,,~,r-,.,n,. ,. ., .,. , , , , , , . , , , . ,n,. ,..--,,.,-,n,,.n - , , , - . .,,,,%,,,,,,,,_,,,.,_,-,,,,,w_,.-.,.,,,,-,..,.,,,,,,m.m., ,
l 5.1 ZERO PCXER BIOLOGICAL SHIELD SURVEY (2LP-701-01) l W e zero power portion of the biological shield survey has been included in the main procedure which is discussed in Section 6.9 h e survey satisfied FSAR Section 14.2.12.81, Zero Power Biological Shield Survey Test. O i I l l 5.6
O ~5.2 IIM POWER PHYSICS TEST PROGRAM (2LP-333-01) Stamary te low power physica test progran at SONG 2 was perfooned between July. 27 and Septenber 3,1982, .to gaper as-built data on the Sq 2 core tmder hot shutdown (320 F, 600 psia) and hot starriby (545 F, 2250 psia) conditions. SONGS 2 is the first of the N Combustion Engineering 3410 Megawatt (thennal) cores and the hot - shutdown portion represents first-of-a-kind testing for SCNGS 2. g te tests consisted of measurenents which were designed to: g (1) show that no core loading or fabrication errors which result in measurable CEA asynenetries have occurred and to demonstrate that each CEA has been properly coupled to its drive mechanisn (FSAR Section 14.2.12.82). (2) Measure the isothermal temperature coefficient (I'IC) for various CEA configurations at zero power (FSAR Section 14.2.12.83).. (3) Determine' regulating (including part-length CEAs) arxl shutdown CEA group worths at zero power and verify that the CEA insertion limits are adequate to assure the required shutdown margin (at cold conditions only the full-length regulating CEAs were measured) (FSAR Section 14.2.12.84). (4) Measure the differential boron reactivity worth for various CEA configurations (FSAR Section 14.2.12.85). (5) Measure critical bcron concentrations (CBCs) for various CEA configurations (FSAR Section 14.2.12.86). (6) Measure the worth of the pseudo worst dropped CEA and PICEA (PSAR Section 14.2.12.87.lA). (7) Measure the worth of the psetzlo worst ejected CEA fran the zero power dependent insertion limit (ZPDIL) and from the full power dependent insertion limit (EPDIL) (FSAR Section 14.2.12.87.lB). All test objectives were adtieved and all acceptance criteria were met, as outlined.in Tables 5.2.1, 5.2.2, and 5.2.3. O 5.7 e
5.2 IfW POWER PHYSICS TEST PROGRAM (2LP-333-01) (Continued) In addition to measurements which satisfied acceptance criteria, O CEA worths were also measured in both the high and low temperature conditions. I<eactivity changes resulting frm increasing temperature and pressure frm hot shutdown to hot standby conditions were measured. 'Ihese results had no acceptance criteria (data was collected for information only), but are reported here for completeness. Each of the measurements is described in greater detail below in its respective section. A separate section will also discuss exceptions to the Technical Specifications which were necessary to perform the low temperature portion. Measuring Equipnent Physics measurenents were made in terms of " reactivity worth", with reactivity being that which changes reactor power. Reactivity can be psitive, which causes reactor power to increase; negative, which causes it to decrease; or zero, in which case the reactor is just critical with no increase or decrease in power level. tbst of low power physics involved very small changes in reactivity which caused reactor power to double or halve in four to five minutes or lorger. An analog reactivity emputer was used to determine reactivity. Two strings of two excore unccxngnsated ion chambers were connected to two picoammeters, and the voltage output frm the picoameters was supplied to the reactivity emputer. Core power, cmputed reactivity, ", temperature and, scmetimes, pressure were recorded. Data reduction was - performed by hand frcm the recorded data. a m Measurements , Physics data was gathered by a series of repetitive measurements as described below. m 5.8
TABLE 5.2.1 IIM TDIPERA1URE AND PRESSURE DATA
^
(320F, 600 psia) Measured Parameter Units Value Acceptance Criteria CBC (AR0) PPM 869. 8351100 O ITC (EARO) % ao/ F -0.143 x 10-2 -0.108x10-2+0.5x10-2 Group 6 Worth % ao 0.230 0.194 1 0.1 Group 5 Worth % ao 0.270 0.240 1 0.1 Group 4 Worth %ao 0.616 0.578 1 0.1 Differential Boron Worth PPM /%ap -64.84 -65.95 1 15 IK (Groups 6,5,4 0 LEL) % ap / F -0.346 x 10 0.318x10-2 0.5x10-2 Sequential Worth % ao 1.101 None (Groups 6,5,4) TABLE 5.2.2 HEAT-UP AND PRESSURIZATION DATA (INFORMATION ONLY) Measured Parameter Units Value Acceptance Criteria
-0.21x10 -2 U
IK (320-360 F) % do/ F None Pressure Coeff. %Ap/ psi -0.15x10 -4 he (600-1100 PSIA)
-0.25x10-2 O
IK (360-450 F) % ao / F None Pressure Coeff. %Ap/ psi -0.23x10~4 he (1100-2250 PSIA) IK (450-500 F) % do / F -0.33x10 -2 None IK (500-545 F) %ao/0F -0.33x10-2 None I CBC = Critical Boron Concentration ITC = Isothermal Temperature Coefficient ARO = All Rods Fully Withdrawn EARO = All Rods Essentially Fully Withdrawn 5.9
'IABLE 5.2.3 HOT STANDBY CONDITIONS DATA 0
(545 F, 2250 psia) Measured Acceptance Parameter Units Value Criteria CEA Symmetry Checks 4 0.86t (max) within 11.5 4 of group average Worst Drcpped PLCEA (P-30) % Ao 0.028 0.023 10.1 Worst Dropped SUBGP (P--1) % A0 0.108 0.0989 +0.1 Worst Dropped CEA (2-1)* % A0 0.085 0.0878 +0.1 CBC (ABO) PPM 833 805 + 100 Group 6 Worth % Ap 0.411 0.385 70.100 Group 5 Worth .% do 0.383 0.375 +0.100 Group 4 Worth % Ao 0.928 0.867 70.130 Group 3 Worth % Ap 1.029 1.031 70.155 Differential Baron Worth PPM /%Ao -74.15 -76.3 - 15 (CEA Groups 6, 5, 4, 3)
-2 0.5x10 -2 I'IC (Groups 6,5,4,3 @ LEL) %Ap[F -1.3x10-2 -0.93x10 Ejected CEA 5-45 (ZPDIL) % do 0.257 0.314 +0.1 Group 2 Worth % Ao 0.662 0.632 +0.1 Group 1 Worth % Ao 1.203 1.209 +0.181 Differential Boron Worth PPM /% do -69.74 -76.3 T 15 (CEA Groups 6 through 1)
-2 I'IC (Groups 6-1 @ LEL) %MF -1.83x10 -1.783x10-2+0.5x10-2 CEA Group P Worth % do 0.390 0.362 10.1 (Groups 1-6) n CEA Group B Worth % A0 3.143 2.970 +0.446 .
CEA Group A Worth *** % AD 3.501 3.160 70.474 *1 Total Full-Iergth CEA % 60 11.260 10.629 T 1.063 i Worth Sequential CEA Worth % Ao 4.601 None (Groups 1-6) I'IC (AE) %Ap/ F -0.380x10-2 -0.308x10-2+0.5x 0-2 MIC (Am) %Ao/ F -0.224x10-2 <0.13 x 10 -2 CEA 2-1 Worth % Ao 0.085 None CEA Group P Worth % Ap 0.211 0.194 +0.1 Ejected CEA Worth (FPDIL) % A0 0.0138 0.0134 +0.1 CEA 6-20 CBC (6, 5, 4, 3 @ LEL) Pai 629 CBC (6,5,4,3,2,1 @ LEL) PPM 499 452 i 100
- CEA 2-1 is the center CEA ZPDIL = Zero Power Pcwer Dependent N
** 'Ibch Spec Limit Insertion Limit - *** See Section 5.2.3 FPDIL = Full Power Dependent o Insertion Limit S LEL = Lower Electrical Limit RfC = Moderator Temperature SUBGP = CEA Bank Subgroup Coefficient CBC = Critical Baron Concentration ARO = All CEAs (Rods) Out of Core (fully withdrawn) 5.10 L
g
! )
V 5.2.1 CEA AND PARP LENG'IH CEA SYMMETRY CEA symmetry checks determine the relative reactivity worth of symetric CEAs. Relative reactivity worths determine how reactive one CEA is with respect to the average of the symmetric group. Symmetric CEAs are those whose rMial location and type (strh as part length or full-length) are symmetric with respect to the core center. Symmetric CEAs are fourx*. within Regulating, Shutdown, and Part Lergth groups. %e minimun nunber of CEAs in a symmetric group is four; the maximun is sixteen. A CEA is considered more worthy than its symmet-ric counterpart if when it is fully inserted in the core and its symmetric counterpart is withdrawn there is a net negative reactivity as cmpared to the reactivity of the opposite configuration. CEA symmetry checks are useful in deteunining that the core is properly loaded, that CEAs are coupled to their extension shafts, and if one portion of the core is more or less reactive than another (i.e., the radial flux distribution in the core is tiltd) . We test satisfied FSAR Section 14.2.12.82. CEA synnetry checks were performed by trMing one CEA with another in a symmetric location. %e trade was accanplished by first insert- " ing one of the symmetric CEAs and obtaining zero reactivity by
/] adjustire Regulating Group 6. W e CEA was then withdrawn in steps J
(,,/ Wile inserting a symetric counterpart CEA in roughly equivalent S reactivity steps. We reactor experienced snall reactivity changes 3 during the CEA trade. Once a CEA hd been trMed, the net reactivity was measured by the reactivity conputer, arxl the next synmetric CEA was trade 3. A va51 ation n CEA symmetry checks was performed at low tenperature (325 F) and involved movirg the CEAs into the core fran their fully withdrawn position only until negative reactivity resulting from CEA motion was observed. %e negative reactivity was final proof that CEAs were couple 3 to their extension shaf ts. We test was important because extension shafts can move' without being coupled to their respective CEAs. It is aivantageous to discover an uncouplai CEA early. %e test was called a (bupling Check. A coupling check was performed rather than synmetry check because the symmetry check could be nore accurately performed under hot conditions.
@e results of the Symmetry and Coupling check showed that all CEAS were couple 3. hhile the CEA worths appeared to be biased to one side 4 of the core, all acceptance criteria were met, as shown in Figure S 5.2.1.1. 5 cm i l
G/ 5.11
Figure 5.2.1.1 CEA SYMMETRY RESULTS (Relative worths in e reactivity.): 28 EAST 1A
'V 10 12 14 1
2 3 4 56 55 Dl6sl7 18 19%2O s 21 y um
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WEST A p sitive w rth means the COLD LEG COLO LEG CEA is worth more than the jg 2A average of the symmetric subgroup counterpart CEAs. ; 5.12 - J
f>
~5.2.2 ISOIHERMAL TEMPERA'IURE COEFFICIENT (IK)
We Isothennal 'IWnperature Coefficient (IK) test measured the change in reactivity resultire fran changes in reactor coolant tenperature for a variety of CEA insertions. W e test was perfonned by slowly heatirg or coolirg the RCS so the fuel and moderator were essentially in thennal equilibriun. %e resulting reactivity changes were calculated by the reactivity canputer and recorded as a function of RCS tenperature by an x-y plotter. One axis was RCS tenperature, the other reactivity. .We test satisfied FSAR Section 14.2.12.83. We IK is the sun of the Fuel Tenperature Coefficien.t which is always large and negative, and Pederator Temperature Coefficient. We Pederator Temperature Coefficient was obtained by subtracting the Fuel Temperature Coefficient ('upplied by Combustion Engineerirg) fran the Isothennal Temperature Coefficient. We bbderator Coeffi-cient is slightly positive for high boron concentrations and large " and negative for low concentrations. We resultirs IK was quite
- small and negative for high boron concentrations but becane more negative as CEAs were inserted and boron concentration was lowered. E 5
O O 5.13
5.2.3 SHUIDOWN AND REGUIATING CEA NORHE We reactivity worth of the shutdown and regulating CEAs was determined by trading the reactivity associated with CEA insertion with that associated with boron concentration. %e measurement was made first by diluting the RCS (which added positive reactivity) while balancing reactivity with CEA insertion to maintain reactor power approximately constant. We CEAs were then borated out of the core. We dilution portion of the test was initiated by adding demineralized water to the Volume Control Tank, waiting for the resulting dilution to enter the RCS and increase reactor power, then inserting CEAs to produce negative reactivity and maintain approximately just critical low power corditions. Dilution was strooth and gradual whereas CEA tretion was sudden. We resulting recorder trace of reactivity versus time appears as a zig-zag. We changes in reactivity associated with changes in CEA position can be added to produce total (or integral) worths or plotted individually for differential worth curves. h e measurement satisfied FSAR Section 14.2.12.84. Dilutions were used to measure individual group worths, whereas borations were used to measure CEAs in Sequential motion (the normal control method). Group worths are shown in Tables 5.2.1 and 5.2.3. Borations and dilutions were performed during both the hot standby and hot shutdown portions. Se resulting curves are shown in Figures 5.2.3.1 through 5.2.3.11. Dilutions and borations were used only to measure the worth of the regulating and part length CFAs and the second bank (Bank B) of the shutdown CEAs. It was not permitted to dilute all CEAs to their fully inserted position. We remaining worth of Group A was determined by CEA drop. A CEA drop involved tripping (or scrartming) those CEAs which were withdrawn and measuring the reactivity change. W e large core is so loosely coupled that drops worth more than approximately S3 (reactivity) are not calculated well by the analog reactivity ccmputer. We total worth of all CEAs was inferred frca dilution, boration, and drop measurements. We uncertainty associated with the measured worth of shutdown bank A n (Table 5.2.3) was larger than that associated with other rod worth . measurements due to the measurement technique which involved large ; rapid reactivity changes. B is measurement technique does not g prcduce as accurate a result as the smaller reactivity changes a rnrmally used to obtain reactivity measurements with the analog reactivity ccmputer. 5.14
.i O Figure 5.2.3.1 Sequential CEA Motion O O Integral Rod Worths vs. Position For Banks 4,5, and 6 BOL T-320*F P-600 psi
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[.
T A Lq> Cx> v Figure 5.2.3.3 Integral Rod Worths vs. Rod Position Banks 5 and 6 No Overlap 01, T-320*f P-600 psi.
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f~) p) L) GI Q Figure 5.2.3.5 Integral Rod Worths vs. Rod Position BOL T-547*F P-2250 psi ~
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Figure 5.2.3.6 Integral Rod Worths vs. Rod Position For Non Overlapping Banks 1,2, and 3 BOI. T=547 degrees F P=2250 psi
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5.2.4 BORON WORHi MEASUREMENTS Borcn worth measurements were performed by knowing the reactivity required to change frm one boron concentration to another. All necessary information was directly available frm the Shutdown and j Regulating CEA Group Worth Measurements since groups of CEAs were diluted into the core. Periodically there were Critical Boron Concentration (CBC) measurements which precisely defined boron concentrations. The amunt of reactivity needed to maintain the reactor just critical for the changing boron was known frm CEA l mtion. A simple division of boron concentration change by reac-tivity change produced boron worth. The Boron Worth Measurement satisfied FSAR Section 14.2.12.85. O 5.26 O
l O 5.2.5 CRITICAL BORCN CONCENTRATION MEASUREMENTS Critical boron concentrations (CBCs) were measurements during which the plant was operated as close to steady-state as possible so precise measurements of the variables which made the plant critical could be quantified. W e variables included CEA positions,~ iso-thermal reactor temperature, reactor pressure, and boron concentration. CBCs were required with all CEAs withdrawn (also known as ARO, or All Ibds Out), and with various groups inserted. In practice, CEA group position was not precisely fully withdrawn or fully inserted; rather, the CEAs were brought within .a few inches of the desired erdpoint position, and then were raised or lowered briefly and the equivalent reactivity difference added or' subtracted from the CBC boron concentration. CBCs required stable boron concentration and almost always preceeded ITC maasurements. W ey are extremely useful in benchmarking physics predictions. Measurenents of CBC satisfied FSAR Section 14.2.12.86. O 5.27
5.2.6 PROBLEMS AND SOLUTICNS 2ree significant plant problems were encountered which interrupted @ysics testing. First, slipping and dropped CEAs slowed testing considerably. W e problem had initially been encountered during Postcore Hot Functional Testing. During a maintenance outage a chemical treatment (hydrazene) was applied to the CEDM housings in an attspt to destroy any foreign material which could have interfered " with the upper gripper's movement. Adjustments to the electronics
- which energized and deenergized the CEDP. motor windings were made.
After the cmpletion of physics testing four of the CEDPs were S replaced. J! 2e second prcblem involved leaking packing in one of the pressurizer spray valves. . A seventeen-day maintenance outage was required to repair the packing and during this outage the chemical treatment of the CEDMS was performed. Wird, the plant cmputer occasionally failed. We plant cmputer performs many tasks, including permitting CEA notion near the top and bottcm of the core. hhen the cm.puter failed, CEA motion in the sequential mode of operation was stopped. We measurements were repeated after the cmputer was restarted, or a different mode of control was used. No problems were encountered with physics related acceptance criteria. 5.28
5.3 ANALYSIS OF THE SONGS 2 IDW POWER NATURAL CIRCULATION DEMONSTRATION (2LP-333-02) Sumary Iow power natural circulation demonstrations were mnducted at SCNC6 2. We demonstrations were a first for a Cmbustion Engineering two loop nuclear steam supply system. W e demonstrations were performed mainly to demonstrate various natural circulation mnditions to all licensed reactor operators prior to initially ex wding five percent reactor power per NUREG-0712. A considerable amoult of data was obtained with existing plant instrumentation, and a determination of natural circulation flow rate as a function of reactor power was made. We reactor coolant system, was shown to mix rapidly and tended to equalize boron and temperature mismatches. An analysis of equal cooling frm both steam generators, called symetric cooling, and the resulting nearly symetric flow was performed using three a steady-state reactor power levels. Natural circulation flow was shown to - increase rapidly with power to approximately four percent of full flow at g four percent power. A single ten gallon batch of highly concentrated boric g acid was introduced into one of the four cold legs. Its effect on the critical reactor was measured by a reactivity cmputer which received signals frm two of the excore nuclear detectors. The resulting changes in reactivity' allowed system transit and mixing times to be measured directly. A cmparison of transit time and flow rates showed that a portion of the reactor coolant system was not circulating. Analysis of boron behavior and loop and core exit temperatures indicated that mixing was occurring through-out the BCS with the exception of the upper reactor vessel above the hot legs. N Isolation of steam and feed flow to one stem generator resulted in asym- 4 metric cooling.. Natural circulation flow continued but slowed considerably R within an hour after isolation. We rapid recovery of the generator upon a resumption of feedwater and steam flow indicated natural circulation muld easily be restored. We isolated steam generator caused a large difference in cold leg temperatures across the reactor vessel. However, core exit and not leg temperatures indicated a uniform temperature distribution, verifying thorough mixing. Mixing is attributed to inherent features such as rotation of fluid in the dowcmer (due to the angle of entry of fluid frm the cold legs into the downcmer), unequal interloop flow rates, and unequal flow path lengths among steam generator tubes. Other components such as the stationary reactor coolant pump impellers and relatively high charging flow velocities may also contribute to mixing. O V 5.29
l l We respanse of the core to the concentrated boric acid showed that even a highly localized perturbation can be expected to be substantially mixed 2
.9'!
in less than five loop transit times (within a half hour) . & Analysis The following analysis consists of (1) determinations of natural circulation flow rates and reactor vessel temperatures as a function of reactor power; (2) cmparison of measured and calculated loop transit times; (3) boron mixing verification and determination; (4) isolated steam generator condi-tions; (5) a general discussion of mixing in the reactor coolant system; ard (6) depressurization and repressurization of the pressurizer. he normal method of determining reactor coolant system flow using reactor coolant pump or steam generator differential pressures could not be used because flow was so low that meaningful and reliable differential pressures , could not be obtained. We flows reported below depend entirely on temperature , differences between the hot and cold legs of the same loop (vessel differential q temperture). g w A brief description of the SONGS 2 nuclear steam supply system and its instru-mentation, and the low power natural circulation test program will preceed the analysis and is intended for familiarization.
% e Nuclear Steam Supply System The SONGS 2 nuclear steam supply system is shown in Figure 5.3.1. We Figure shows the two lcop configuration. 'No hot legs conduct water frm the core directly to two vertical U-tube steam generators, and four cold legs return aald water to the reactor vessel via four reactor coolant pumps. Temperature instruments mounted in the hot and cold legs provide a direct indication of fluid temperature in each. Nuclear detectors mounted external to the reactor vessel were calibrated to reactor power via a secondary calorimetric based on auxiliary feedwater flow.
]
his system is well designed for natural circulation flow since the heat i source (core) is at a lower elevation than the heat sink (steam generators) o and is connected by piping (hot legs) which have no bends capable of trappirg steam and isolating the heat sink frm heat source. We cold legs return cold fluid to the bottcm of the core via the downcmer. I pressurizer, connected to the top of hot Leg #1, maintains pressure in the reactor coolant system. 5.30
FIGURE 5.3.1. ISOMETRIC VIEW OF SONGS UNIT 2 NUCLEAR STEAM SUPPLY SYSTEM STEAM STEAM N0ZZLE N0ZZLE
% - )~ N STEAM . STEAM GENERATOR
)
' GENERATOR
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Location of Instruments in the Reactor Coolant System Most of the information presented below was obtained frm temperature instruments and nuclear detectors shown in Figure 5.3.2. Fluid temperatures were measured by high-accuracy resistance temperature devices (RIDS) which penetrate into the coolant approximately two inches and are a part of the normal Reactor Protective System instrumentation. W e hot leg temperature sensors are trounted 45 and 90 frm vertical as indicated in Figure 5.3.2. Temperature sensors on two of the four cold J legs were used. We cold leg RIDS are located in the upper quarter of the plie , m Two nuclear detectors mounted outside the reactor vessel approximately 180 apart were used to measure reactor power and reactivity. %e detectors received most of their neutron flux frcm the closest five or six bundles located on the periphery of the core. Since neutrons must pene- N trate the downcmer and will suffer increasing attenuation with decreasing ; cold leg temperature, a temperature correction of one half percent of g signal per degree Fahrenheit was made. m Pressurizer pressure and level, and steam generator pressures and levels were obtained frcm instruments connected directly to the vessels themselves. m Core exit thermocouples are located in the upper portion of selected fuel f bundles, as shown on Figure 5.3.1. g m Demonstration and Data Points he demonstration is sketched in Figure 5.3.3. W e reactor was brought to approximately 3% of full power and all four reactor coolant pimps were manually tripped as simultaneously as possible. It was noted that the pumps continued to turn considerably longer than the normal 200 seconds; one continued to turn as long as 480 seconds. Steam generator pressure was a manually controlled to maintain approximately 1000 psia (hot standby ; conditions), forcing hot leg and core average temperature to rise and a reactor power to fall due to the core's negative temperature coefficient. S A steady-state power of approximately 0.9% was measured twelve minutes after the pumps were tripped and forTned the first data point for flow determination. Reactor power was slowly raised by increasing steam flow. . Approximately 24 minutes after the pumps had trippA, the second data point at 2.2% reactor power was measured. 5.32 O
FIGURE 5.3.2: IOCATION OF INSTRUMENTS ON Tile RCS COLD LEG. TEMPERATURE RTDs CHARGING (T112CBA RCP 1B I
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O Isolated Steam
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5.34 -
I A final power increase to approximately 3.2% power was performed and data was obtained at approximately 35 minutes into the demonstration. A period of primary system depressurization and repressurization then ; followed. ' , A third portion of the test involved isolating one of the two stem gene-rators. Isolation was accmplished by stopping all stem and feedwater ! I flow fra conditions'of low power level and low (800 psia) stem generator pressure. After an hour the isolated generator was then returned to service.
! After a brief period of stability a single ten gallon injection of concen-l trated boric acid was made to the suction of the darging pups to determine i loop transit times and characterize boron mixing. m 4 J Determination of Natural Circulation Flow Rates Natural circulation flow rates were determined by measuring the temperature !
difference between the hot and cold legs of the reactor coolant system (as show in Figures 5.3.1 and 5.3.2), and reactor power frm nuclear , instrumentation. '1he temperature difference was converted to a dange in ; I enthalpy using 1967 ASME Stem Tables. Reactor power was calibrated prior ' to the start of the test by a secondary calorimetric using auxiliary O feedwater flow. Reactor power was divided by the change in enthalpy to obtain mass flow rates. Mass flow rate was converted to volumetric flow rate using appropriate conversion constants and specific volumes frm the stem tables. I
! All analysis assmed well-tnixed temperatures in both the hot and cold legs.
Data indicated that little (if any) stratification existed during the tests. l i i lO 5.35 L
Table 5.3.1 Natural Circulation Conditions Approximate Data Corrected % Icop #1 ( F) Icop #2 ( F) Average (oF) Point Reactor Power T hot cold hot T eold hot T eold aT 1 0.9 561.50 512.75 563.45 546.50 562.48 544.63 17.85 2 2.2 561.95 525.95 563.45 529.25 562.70 527.60 35.10 3 3.2 562.70 516.50 564.50 518.00 563.60 517.25 46.35
- The full power rating of the core is 3390 Megawatts thermal.
O O 5.36
~
Natural Circulation Flow Rate Figure 5.3.4 shows reactor coolant system flow rate in gallons per minute as a function of reactor power. Data was obtained at the three power levels during periods of stability. Data points are shown in circles; curve fitting was done manually. Reactor power was obtained frm nuclear detectors mounted outside the reactor vessel. Vessel differential tempera-tures were obtained from Resistance Temperature Devices (RIDS) mounted on the hot and cold legs of the two reactor coolant system loops, as mentioned above. Se error bars reflect first-order uncertainties based on engineering judgement and are intended to show that large uncertainties existed. Core , pawer was determined at ~3% of full power by a secondary calorimetric using , auxiliary feedwater, whose loop accuracy according to FSAR Table 7.5-1 is q 3% of 800 gpn full span. %e uncertainty bars in power reflect only the a uncertainty in auxiliary feedwater flow although other contributors existed. 5 Similarly, RCS flow was detgtmined by reactor vessel differential tempera-ture measurements whose 150 F spans are 1% accurate, with no other contributor. R e graph shows that an asymptotic single-phase flow will be reached at fairly low power.
%e practical limit for natural circulation is approximately 4% of full power and flow at which point an administrative full power vessel differen-tial temperature limit would be approached. Figure 5.3.5 shows vessel differential temperature and percent core flow as a function of reactor power.
Cm.parison of Measured and Calculated Loop Transit Times n Figure 5.3.6 cmpares measured loop transit time with that calculated using d the entire RCS volume, and with the BCS volume minus the reactor vessel i upper head (above the hot legs) . 2e graph shows that a portion of the reactor vessel did not circulate, although the isolated steam generator and boron mixing slug tests indicated that all parts of the coolant loops were well mixed. %e upper reactor vessel above the hot legs remained as the only likely stagnant location during this test. % e Natural Circulation Verification test, Section 6.17.1, indicated that the stagnant volume may be a short-term effect. O 5.37 l l L
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FIGURE 5.3.6: COMPARISON OF MEASURED AND CAIfULATED IDOP TRANSIT TIMES ne graph below empares data frm the slug of boron which produced a transit time of 4.8 minutes at a corrected power level of 1.8% full power to those calculated frm differential temperatures and reactor power with reactor coolant system (FCS) volume. We upper curve used on IG volume of 10300 cubic feet minus 43 cubic feet frm the surge line. We bottm curve also subtracted the regetor vessel upper head above the hot leg nozzles (approximately 8888 ft ) . l -
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O Re Isolated Ste m Generator Demonstration he isolated steam generator portion of the demonstration was corxiucted . after recovery frm the reduced primary pressure demonstration and pre- 2 ceeded the boron slug transient. Figure 5.3.7 is a plot of loop tem- E peratures, steam generator pressures, and temperature-corrected reactor power during the demonstration. At three minutes (plot time) the atnes-pheric dump valve on Steam Generator #2 was closed. Closing the valve cmpleted the isolation process in which all feedwater, blowdown, and sartpliry lines to Steam Generator #2 were isolated. Steam generator " pressure and cold leg temperature in the isolated loop increased. As warmer water returned to the core, reactor power decreased. Cold and hot J 1eg temperatures in the isolated loop converged at a rate which would have @ equalized temperatures and produced very low flow within an hour. S We plant responded normally to changes in the active steam generator. For instance, at approximately 181/2 minutes (plot time), the openity of the dump valve on the active (non-isolated) generator was increased, causing a n decrease in steam generator pressure which was reflected less than one , minute later by a decrease in cold leg temperature. At 32 minutes (plot q tine) the active generator's atmospheric dump valve was closed slightly, g causiry cold leg temperature to increase approximately one minute later, m Recovery was initiated at 381/2 minutes when the isolated generator N (n) V atmospheric dump valve was opened and feedwater was restored. A decrease ; in cold leg temperature was noted approximately two minutes after opening g the dump valve. j 2e initiation and recovery frcm isolated steam generator conditions was accmplished snoothly. Both hot leg temperatures increased at the same rate and with the same offset, indicating that there was flow in both hot legs and no significant stagnation. If stagnation had been significant the isolated generator's temperature would have increased as slower-moving, hotter water collected in the top part of the pipe in which the RID's are located. Separate nuclear detector signals were not recorded and it is not possible frm the recorded indications of reactor power to determine if one side of the cat's prcduced nore or less power due to non-isotherinal cold leg te70 erat./en. However, if fluid returning to the core frca the isolated n and ron-isolated steam generators did not mix, hot leg and core exit , temperatures should have diverged. Since no such divergence was observed, q cmplete mixing had to occur upstream of the core exit thermocouples, in g the downcomer. See Figure 5.3.8. Unequal flows between the loops n 4 o 5.41 1
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1 should have forced penetration in the downcmer and lower plenum of the i faster-moving fluid frm the active generator. As will be discussed later, i motion of fluid in the downcmer due to the configuration of the cold legs i appears to have substantially contributed to mixing. i " i A nonuniform core inlet temperature should have been reflected in the core
- exit thermocouples, even though small temperature feedback effects existed
! in the core. 'Ibe core exit thermocouple temperature distributions at the E beginning, middle, and end of the isolated steam generator cases were not 5 substantially different and showed no skewed temperature distribution, a very strong indication that core inlet temperatures were uniform.
'ite extent and likely location of interloop ccanunication, namely the downcmer and lower plenum, was demonstrated by the asymetric stean ,
. generator test.
; Beactivity Change Due to Boron Slug Injection Just prior to the end of the demonstration, a ten gallon slug of concen- n trated boric acid was introduced via the charging system into cne of the .
cold legs as illustrated in Figure 5.3.9. Information frm the slug gave E valuable insight into loop transit and mixing times, flow patterns within S
; the reactor vessel, arxl active (circulating) system voltane.
One of three charging pumps was operating because it was hoped that the low I flow would encourage the boric acid to enter the cold leg with as little dilution as possible; in other words, as a slug. i An analog reactivity cm puter, Wiich uses neutron signal fr a both nuclear detectors shown in Figure 5.3.9, was used to monitor the effect of the n boron slug on the core. Detector signals were averaged, not recorded - separately, so core power tilts which may have resulted frm radially g ! asymetric boron flow through the core or unequal core inlet temperatures 3 i were not recorded. i Figure 5.3.10 describes the reactivity transient which accmpanies the l boron slug. i l 4 5.43 I
536.499 23 F % / 514.7210.29 COLD LEG e lo 12 14 COLD LEG I 2 3 4 5Gi 5 E3 Dl6sl7 18 19N20 21
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, u -1.6 CORE EXIT TIIERMOCOUPLZ AVERACE COLD LEG 2A m PERA m 2 = 548.U t 1.17 9 COLD LEG 18 536.833 22 F 515.05+0.05 F FIGURE 5.3.8: TOP VIEW OF SONGS 2 CORE Sil0 WING FUEL BUNDLE LOCATIONS AND CORRESPONDING CORE EXIT TilERM0 COUPLE TEMPERATURE DIFFERENCES FROM AVERAGE (Individual core exit thermocouple temperatures minus the average of all core exit thermocouple temperatures are shown above for each instrumented fuel assembly) 5.44
O O O NUCLEAR REACTOR COOLANT PUMP (4) DETECTOR I n 9 (2) h/ Ar STEAM HOT T REACTOR ' H6T STEAM GENERATOR j LEG i VESSEL 2 LEG , GENERATOR
$; CHARGING #
N* o . CONTAINMENT REGENERATIVE
** BOUNDARY '
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BORIC ACID CHARGING PUMP TO LETDOWN SYSTEM & VOLUME C 4 a 1 r C FROM VOLUME CONTROL TANK CONTROL TANK 1 i FIGURE 5.3.9: BORIC ACID INJECTION POINT AND REACTOR COOLANT SYSTEM ! Suppl. 2
Mixing in the RG is %ought to Result Frm Many Mechanisms Two portions of the denonstration (the isolated steam generator and boron injection) indicated that the reactor coolant system tends to hmogenously mix within a relatively short time. At least four mechanisms are thought to exist which inherently mix fluids within the piping and, nest im;mrtant, within the reactor vessel. %ey include: (1) rotation of fluid in , the downcmer; (2) unequal flow between loops; (3) unequal tube lengths in , the steam generators; (4) and high charging velocities relative to that of q the cold leg fluid. Ibtation of fluid in the downcmer is thought to g result frm the configuration of the cold legs. Unequal flow between 1ceps e occurs any time steam generators steam at different rates. Unequal steaming results whenever the main steam isolation valves are closed. Unequal steam generator tube lengths cause portions of the coolant arriving at the hot leg side of the tube sheet to exit into the cold leg at various times. Finally, the cold leg fluid cbwnstream of the pu:r.ps should mix well with the incming charging flow whose velocity is many times that of the loop. Each mechanism will te discussed in detail below. O o 5.46
O k FIGURE 5.3.10: REACTIVITY VERS 11S TIME FOR A 10 CALLON INJECTION OF BORON INTO ONE LOOP The graph below is the raw reactivity signal from the reactivity computer with time after a ten gallon slug of concentrated boric acid (approximately 9.5% by weight) was injected into the suction of the single operating charging pump. The reactor was operating at approximately 1.8% of full power with both coolant loops in approximately equal natural circulation. Reactivity decreased rapidly as the slug initially entered the core starting at 2.4 minutes. Subsequent decreases are attributed to the boron returning from the coolant loops to the core and produce a loop transit time of 4.8 minutes. With each passage through the system, the slug's effeet was reduced, indicating.that it was mixing such that the system , was effectively mixed by the time the slug had completed five passes through the system. ,
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Downcmer Fluid Rotation Rotation of the fluid in the downcmer is the counter-clockwise notion of the fluid (as viewed frm above) which is driven by water entering frm the cold legs lA ard 2A. Such rotation was observed during open-vessel testing of the low pressure safety injection system. W e reactor vessel head and core support barrel had been removed so it was possible to see air bubbles in the water flowing frm the cold legs into the vessel at 3,000 to 4,000 gpn per cold leg. His flow rate was approximately equal to that measured in natural circulation testing. %e incming flow followed the directions shown in Figure 5.3.11. mis counterclockwise flow pattern suggests that a considerable arount of mixing will occur while cold fluid is descending through the downcmer. ~ J S 3 Unequal Flow Between Icops Almost any operational situation under natural circulation conditions will force an unequal flow rate between the two loops because heat transfer frm the steam generators will not be equal. Equal heat transfer means auxil-iary feedwater and steam flow rates must be equal, which in fact rarely happens when the main steam isolation valves are closed. As a result of unequal feed and steam rates, heat transfer will not be equal and the resulting natural circulation flow rates frm the generators will be unequal. Because unbalanced heat transfer will be the rule rather than the exception, flow will be inherently unbalanced. Unbalanced flows enhance mixing in the downcmer and lower plenum. 5.48 9
RCP IB Approximate Flow Directions RCP 2A
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RCS AS SEEN FROM ABOVE FIGURE 5.3.11: COLD LEG FLUID ENTRY ANGLES'AND FLOW DIRECTIONS P O 5.49
Unequal Steam Generator Tube Irngths Figure 5.3.12 illustrates the location of the steam generators and their tubes although only one loop is shown. W e reactor coolant pump impellers and different length tubes in the steam generators are sketched. he different lergths in steam generator tubes are important since any uniformly distributed concentration of boron arriving at the inlet tube sheet will be " discharged at the outlet tube sheet at various times. We shorter tubes will discharge borcn first and the longer coes later. We mixture will J then be spread over a much longer length of the cold legs than it occupied E in the hot legs. 5 Feactor Coolant Pump Impeller and Charging h e reactor coolant pump impeller may play an important part in insuring that the fluid in the cold leg is uniformly mixed. We impeller is a canbination centriftgal-axial type, and should force the fluid entering it to rotate. We rotating mixture should then mix with inccming charging fluid and warm it to temperatures which are quite close to the other cold legs. We Stagnation Line h e stagnation line shown in Figure 5.3.12 represents the boundary which is thought to exist between warm, less dense fluid trapped in the upper portions of the reactor vessel and calder denser fluid which is circulating. The location of the stagnation line should be just above the hot legs. As long as the fluid below the stagnation line remains colder than that in the head the stagnant fluid should remain trapped. However, just like in a i hctne hot water heater, if the fluid below beccmes botter, or the trapped fluid cooler, the head should mix. Starting a reactor coolant pump will N also destroy the stagnation layer and mix the system. ; a The Instulated stagnation line and related questions abcut storing het water j in the head formed a portion of the test plans for the 80% Natural Circula-tion Verification test's cooldown and depressurization, discussed later in Section 6.17.1. 5.50 O
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51
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UPPER GUIDE STRUCTURE FLUIDS) DOWNCOMER FUEL BUNDLE CORE
-- REACTOR VESSEL w DIFFUSER FIGURE 5.3.12: COMPONENTS k,. LOWER PLENUM WHICH ENCOURAGE MIXING IN 5.51 THE RCS
Other Observations We Hot Ie3 Is %oroughly Mixed The hot leg appears to be thoroughly mixed. Mixing prevents a situation in which hot water flows through the top of the pipe with the lower portion of the pipe filled with relatively stagnant colder water. Worough mixing has important ramifications. It provides proper temperature indication regard-less of RID locatica. With the fluid throughly mixed, correct calculations of flow velocities and rates in the reactor coolant system can be made unter natural circulation conditions. Flow Was Always in the Nonnal Direction troughout the denonstrations, hot leg temperatures were always greater than cold leg. B is clearly demonstrates that natural circulation flow remained in the nonnal direction, fran cold leg to hot leg in the reactor j vessel, to the steam generators, and returning to the vessel through the & cold legs. J No Convective Cells in the Core ne slug of boron entered and exited the core abruptly, as shown by the reactivity traces. E is shows no recirculating convective flow cells existed under the denonstration conditions. 5.52
1 f^' High Charging Flow Velocities Will Penetrate Slower-moving Cold Leg Fluid and Aid in Mixing Tables 5.3.2 and 5.3.3 indicate that charging fluid will always penetrate significantly into the cold leg under natural circulation conditions since its velocity at the point of entry (the charging nozzle) will be signifi-cantly greater than that of the bulk fluid. W e reactivity traces after the introduction of boric acid indicate that emplete cold leg mixing. occurred. The boron arrived in the core mixed with the cold leg fluid rather than settling in the lower plenum. Charging fluid does not greatly change the temperature or density of the mixture in the cold legs, as shown in Table 5.3.4. %e table is for a final mixed fluid, and shows the change in typerature and density of the mixture assuming an initial temperature of 545 F, a low power level of 0.89% (the power level at Data Point #1), and charging to a single cold leg. N l E Pressure Behavior During Reduced Pressure Demonstration 5 It is possible to control auxiliary spray flow using the main spray valves, rs as shown in Figure 5.3.13 because there are presently no check valves in
/,\j the spray lines. A design change is planned which will add check valves.
If the main spray valves are fully open, auxiliary spray currentlyflows backwards through the main spray valves into the cold legs. With one charging punp operating the main spray valves must be almost empletely closed to produce spray; with two charging pumps the valves were closed to approximately five percent open before a decrease in pressurizer pressure signalled the arrival of spray. flow. Figure 5.3.13 illustrates how a significant water column (approximately twenty five feet) must be overcome to force water into the pressurizer. Significant pressure reduction rates of 1 psi /second have been achieved with two charging pumps and fully closed valves. (m)
5.53 s . - -
TAmr 5.3.2 TEMPERA'IURES AND CHARGING VEIDCITIES INIO ' DIE COID LEG O U Number Of GPM At** F Fluid Fluid Velocity (Ft/Sec) Charging Charging Temperature At Charging Nozzles 3 Pumps Nozzle At Nozzle Single Nozzle Both Ccmnent o* 90 F 6.8 3.4 No Intdown. 1 (44 gpn) 44 GPM 1 (44 gpn) 52 1/2 GPM 445 F 8.1 4.0 'Ibst results of charging flow ard 2 (88 gpn) 102 GPM 405 F 15.7 7.9 temperature are frca 2HB-220-01, Attach. C, 3 (132 gpn) 150 GPM 380 F 23.2 11.6 "CVCS Integrated Test"
'Ihe single charging pump without letdown could cnly be a transient ,
situation and is not representative of normal operation. , C* E Corrected to temperature conditions at the nozzle. g-m TABLE 5.3.3 NA'IUPAL CIRCULATION AND CHARGING FIN VEIDCITIES A canparison of calculated flow velocities of 545 F fluid in each of the cold legs, assuming both reactor coolant system loops were operating normally and rejecting equal amounts of heat, is made below: Percent Feet / Minute Flow Velocity Reactor Assumine Equal Flow In I4 ops Power Hot Leg Cold Leg Cortments 0.89 1.32 1.15 Charging velocity is a maximum of 2 1/2 times loop velocity, 2.19 1.70 1.73 using cold chargirg (without letdown) . 3.23 1.88 1.92 Charging velocity is minin'um of 1.8 times loop velocity. 5.54 I
i 4 i r TABLE 5.3.4: MIXED TEMPERATURES IN SINGLE (DLD LE)G FOR VARIOUS CHARGING F WWS AND CONDITIONS : l GiARGING MIXED COID LEG FWID l CONDITION TEMPERATURE TEMPERATURE DENSITY QiANGE Single Charging Pump, No Letdown .90UF 537 F 1% Three 01arging Pumps, No Letdown 90 F 514 F 2.8% ; ) Single Charging Pump With Ietdown 445 P 543 F 0.2% f Three Charging Pumps With Ietdown 380 F 536 F 1.0% fore: 01arging without letdown is strictly a transient condition. O . i f .i . 1 I l 1 5.55
i FIGURE 5.3.13: SKETCH OF COMPONENTS ASSOCIATED WITH SPRAY SPRAY LINE 1 l l ) 4 1 PRESSURIZER e
". I 1 I. 1 t / AUXILIARY SPRAY LINE
{ (FROM REGENERATIVE HEAT EXCHANGER) If. . NORMAL ' WATER MINIMUM SPRAY VALVE LEVEL (OPEN APPROXIMATELY 7/8 TURN AT ALL TIMES)
'~
[' - MAIN SPRAY VALVE (TYP. OF 2)
' )
(CAPABLE OF BEING OPENED AND CLOSED
- [ FROM THE CONTROL ROOM)
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5.56
. Cbnclusion Natural circulation under core power levels which simulate realistic decay heat levels has been demonstrated to be stable and well controlled: (1) during normal steady-state hot standby conditions; (2) under reduced subcooling; and (3) with reduced heat removal capacity. We plant is easily controlled and attains and recovers frca various abnormal natural circulation conditions with no difficulty. N
! Natural circulation loop transit times typically are approximately five !
' minutes. Boron inequalities between loops equalized in less than five loop i transit times. Large temperature differences between loops failed to Jl appear at the top of the core or in the hot legs. %e reactor app ars to inherently mix under natural circulation.
l i r 1 5.57
.- - - . - _ . _ . . - ~ - _ _ .- .-. . _ .--.
i 4 SECTICN 6.0: POWER ASCENSION TESTS 1 e 'l 4 a l t I 6.1
' ~ ------__.-,,__ ,_,_,_ _ ' E* WW , . 9,9 wy
6.0 POWER ASCENSION TESTS - We objective of power ascension testirg was to danonstrate equipnent, l2S make adjustnents, and bring the plant to full power. W e test method
- n consisted of performirg the tests as the plant was brotsht to power in .
a series of orderly, closely-observed steps. Power ascension testirg - 2 was ampleted throtgh most of the 50% plateau when this report was - originally written. Testirg throtgh 80% was covered in the first I *g supplenent. %is second ard final supplenent covers the balance of ,S testing. O 6.2
6.1 SUMMAIU OF POWER RANGE TESTING 'Hil00Gil 100% POWER m
)
Power range testing was the evolution which brought the plant from .n low power physics testing through full power and the warranty run, .. which signified comercial operation. We original Startup Report 2% addressed testing only through the end of the 50% power plateau. The && first supplement covered testing through the 80% plateau. % is second and final supplement covers the balance of testing. Power range testing is discussed below in Figure 6.1.1. We figure shows operating node, temperature, and percent reactor power as a function of time; all values are approximate. O O 6.3
1
% FIGURE 6.1.1 POWER RANGE TESTING ww '
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= t- s
[ f .! ! T 9/9 . l. l' i 2 3% 9/10 ' ' '. , ',,l,l.,
, i ., i. .;
i . i < >i Turbine roll 9/11
,,,ii 1
l'; '!
.1% .
iii,i . ll 9/12 ,, 9/13 r . . 3%
! 4.5%
9/14 --- t 1
. i . i 1
9/15
! 3 Reactor trip on steam generator level.
9/17
. ll 2 4%
Reactor trip on steam generator level. 3
' 2 9/18 'h 5% 2PA-344-09 Nuclear and Thermal Power Calibration
.l 2ST-344-08 RCS Delta T Power Determination 9/19 (t.: ;' 1 10%
ex .
- e 2 i et Gi.
9/20 1 I
} ,
Initial turbine synchronization. 9/21 r 2 4%
\ 1 1- 15% Pressurizer spray valve failure.
9/22 / 3 Two HP turbine governor valves fail.
./ '.
9/23
/
/ 4 5
' 120 9/24 _t u. Resin beads found in the condenser.
~
Both pressurizer spray valve stems repaired. , 9/25 - Four CEA drives replaced. 2a 9/26 ,- 1 Steam Generators to be inspected for resin. 9/27 - O 6.4
Ed FIGURE 6.1.1 (continued) v
*T ao e a ::
( 2 6e N 9/27 - , .ir-5 120 9/28 - 9/29 - Decision made co flush resins from the steam
. generators. Resins had come from blowdown system. E i a New CEA drives installed, j 9/30 -
10/1 = 10/2 , i , Flush' of. staam generator E089 complete.
. i i i 10/3 l 4
10/4 ' Flush of steam generator E088 complete.
. Fill and vent'of reactor coolant system.
10/5 -- E a 10/6 =
,a ; .; 5 E7 , . (
- 10/7 -
180' 10/8 h ,' 4 245* 1 i
'10/9 ,
; Steam generator secondary manway gaskets leaking.
I: : , 5 180* 10/10 ; ;'
- if 150 Steam generator secondary manway gaskets require-
! replacement.
10/11 -
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. 6 4 i f I 10/12 M .'.'.
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10/13 l:;:';
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10/17 ; , l;*, 340'
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-545 6.5
w Figure 6.1.1 (continued) un
.T 4eI e 32 10/19 -
3 545* 10/20 ,-; ii 10/21 ,[ i 2 10/22 . { ; : 8 ! ! ! I 10/23 d' ' ' l I 3%
!!, 10%
10/24 , ,
, l 20% 2PA-458-01 Process Variable Intercomparison commenced.
10/25 .
.' 2ST-344-10 NSSS Calorimetric
, i- '
2LP-701-01 Biological Shield Effectiveness commenced. 10/26 2PA-305-01 Incore Detector Signal Verification. 10/27 , 2PA-349-01 Reactor Regulating System Performance.
,, i 10/28 -
,; i l.
. 5I .
10/29 -
,, 2PA-311-02 Steam Bypass Control System Performance.
~
l 10/30 % !Ei; 2ST-344-10 NSSS Calorimetric.
, . , l! 2PA-313-01 COLSS Power / Flow Verification.
10/31 - 2PA-344-06 Core Performance Record. 2ST-344-09 Nuclear and Thermal Power Calibration. 2ST-34 -12 Linear Power Subchannel Calibration. 11/2 -- l 2ST-344-14 Movable Incore Detector Test. 11/3 - 2PA-344-03 CPC/COLSS Verification. 2PA-344-12 RCS Calorimetric Flow Measurement. 11/4 2PA-346-01 Shape Annealing Matrix and Boundary Conditions Measurements. 11/5 - 11/6 - 11/7 - 2PA-344-08 Variable T average Test. 11/8 2PA-349-01 Reactor Regulating System Performance. 7179 _ 2PA-307-02 Feed Water Control System Performance. G 3 Reactor trip due to loss of feedwater control. 11/10 -- - 6.6
gp FIGURE 6.1.1 (continued) 8 it s
- 8 $ S 11/10.- t 3 545*
2 3*' 11/11 [ 1 5% - 3 Reactor trip due to slipped CEA. . 11/12 -
' g a 3
~20% 2PA-401-01 20% Main Control Board m S
11/13 y 9 Reactor Trip. i I 15% 11/14 W { 20% l ; 2PA-458-01 Process Variable Intercomparison complete.
.i . 2PA-311-02 Steam Bypass Control System retest. "
11/15 ' i< ll 2PA-350-01 Integrated Control System performed. *
.i
.; 30% 2 11/16 ;l jl ll1 ll 40" 11/17 Ml h . ll 50%
~
,7 ,'ll , 3 Reactor tripped on high steam generator water level. .
11/18 2 Z, i
'll
$ 180 Outage to repair condenser. tube leakage.
I i! 11/19 - l't n i f ' ? 11/20 H (' ! (! l 4 , 11/21 300* i ,.!
* ! l !
3 , 11/22
.,,,s, N ll.
,j!llI'... 2 545'
;w. c. , .
4.5% 11/24 i m$l l l .r . .
- 1. l ,
1 10% i>' l, , 37% Reduced power to repair.a relief valve on a 20% Moisture Separator Reheater drain tank. 11/25 2 l, , , . I 'M , *,l 42% Reduced power due to condenser tube leakage. 11/26 4 ' ', l 20% 1 i g 3 Reactor tripped on high steam generator
- gjg i,
,l. , ,
, , water level.
,..i 11/28 ,,'"l i i.i 11/29 '; ,
'l'
. 4 i t '-
2-11/30 h, ;:; ; , 1 18%
, 4 . . . 1%
12/1 1 .l 2ST-344-16 Adjustment of COLSS Secondary Pressure
./^\ 'N um Loss Term V<
' Sog 12/2 l 2PA-458-01 Process Variable Intercomparison 6.7 L
y FIGURE 6.1.1 (continued)
*T 4 is
- e A s 12/2 ., , i y
50% 1 i 2ST-344-10 NSSS Calorimetric 12/3 . l 2PA-305-01 Incore Detector Signal Verification li l 2ST-344-10 NSSS Calorimetric 12/4 4!;; l 2PA-344-06 Core Performance Record 2ST-344-09 Nuclear and Thermal Power Calibrat:.on - ii ii ! 2PA-313-01 COLSS Power and Flow Verification J 12/5 ;! , 2PA-344-12 RCS Calorimetric Flow l'j , @
,-l , ;,
2ST-344-12 Linear Power Subchannel Calibration j 12/6 . i
. .4 12/7 ,
2PA-344-03 CPC/COLSS Verification 12/8 !'l 12/9 2PA-347-01 Temperature Decalibration Verification 2PA-346-01 Shape Annealing Matrix and Boundary 12/10 ! . Condition Measurements 12/;. 1p; .';
- m . e !
l'$ ; C , 12/12 ' Reactor trip on faulty DNBR due to incorrect ; 3 c
'i -
Reed Switch Position Transmitter signal 12/13 i,,.!l RCP P004 upper two seals fail [
- i. /: , 4 f i . 4 12/14 -f 5 90 Outage to:
Replace all RCP seals 12/15
- Machine HP turbine governor valve internals Repair LP turbine valve shaft scoring 12/16 ,
Repair pressurizer spray valves Repair letdown valves 12/17 Move heater drain pump check valves - Replace auxiliary feedwater pump P504 motor. . 12/18 E 5 12/19 i 12/20 t 12/21 12/22 12/23 12/24 - ' 0 6.8
1 12v FIGURE 6.1.1 (continued) - o d. 5 mo aa oo r- x H a. ( 12/24-- ' ' 5 90* 12/25 - 12/26 12/27 -
'!i 4, kii 4 12/28 Di , , , ,
280
,e i 12/29 - -l , 310*
t, , ,
,, N.
3 12/30 j'((! l 545' 12/31 2 ',! 2 I , h I ' '
,1 .. .
1/1 %t. . , 1 5% 1 lll ; 2 Reduced power due to condenser tube leakage. 1/2 J I.. " i 4H e-
-c < . c 1/3 'E. l 40% hw @ ed h mm @ mm. d main turbine governor valves.
O W.l l l S i 1 50%
- 1/4 .
. H l 3 Reactor tripped while reducing power for i im .ii , .,
360a condenser tube leakage, i .t. 1/5 u; .i
. i .x .
545*
,:. . ... 2 1/6 s 1 ;
1 8% 3 , m . .. 20g 1/7 lUi i .s
~
. i 50% 2PA-344-10 NSSS Calorimetric -
1/8 -
. 2PA-311-02 SBCS Performance Test 1/9 ,.
ii 7ST-344-09 Nuclear and Thermal Power Calibration
- 2ST-307-02 Feed Water Control System Performance 1/10 -
, , i ...
1/11 l ! j 6 I .i 8 . - 1/1^' ; ll, 2PA-346-01 Shape Annealing Matrix and Boundary
'l Condition Measurement (retest) 1/13 -
, 2ST-344-09 Nuclear andl Thermal Power
' , ;' Calibration 1/14 - i i . , ,
.O + ,, ,-
(' 1/15- ' 6.9 i i
ER FIGURE 6.1.1 (continued) o
*Y.
a ca 3 $.E- A$
- t 1/15 l 1 50%
-l ,
i i i 1/16 il l ,' 2PA-344-09 Variable T average Test. ll l 1/17 'lll 4 2PA-349-01 Reactor Regulating System Performance . 3 Reactor tripped on high steam generator water . 1/18 ii 1 ;' 4 340 level while reducing power for condenser {n
1+i 4 3
tube leakage. m 2 vi
- , i Ni i i 1/19 ,,i ,!', 545*
. .' i'i 2 1/20 ,::l -
10% . 1 1/21 L l .' 2 20% Reduced power for condenser tube leakage. 3
', 3
' 360* Condenser opened for inspection.
1/22 , 1/23 2
,'i l,l ,
- i. .ii i4 1/24 XI.l! 545 I I f I f 1/25 i,
li>il 1 8% i * ' 2
, . 3%
1/26 . i ,ll YW E*l 1 20% 1/27 -- @ _ I' x 50% 1/28 . 2PA-307-02 Feedwater Control System Performance i 1 2PA-350-01 Integrated Control System Performance '/29 .:: , rw 2PA-351-01 CPC Power Distribution Constants Verification 1/31 E"' 2 2/1 ,,,, i .~ 1
^' '
2/2 20% 3 50% Reactor tripped by instability in high pressure ; 2/3 . 2 turbine governor valves. a C ' 1 50% 2PA-382-01 50% Generator Trip / Shutdown Outside
- 2/4 -
' 3 2 Control Room test successfully completed.
2/5 2/6' -
.L 6.10
;;;7 FIGURE 6.1.1 (continued) o
- 1:
no 11 e2 (~~s E AE t L./
) 2/6 . ,- i, . 2 .1%
2/7
' 1 2/8 -10%
N , l 20% 2/9 . i .
. 3 .
! i .
2/10 ' l ! 2ST-344-10 NSSS Calorimetric vm. - i 50% 2PA-311-02 Steam Bnass Control System Performance 2/11 ;;;; 2ST-344-10 NSSS Calorimetric "
; jjl i 3 Power reduced due to condenser tube leaka8e and l 2/12 , l l,ie l 4 reactor tripped on high steam 8enerator water level. @
* , 3
. .1,I ,l s m 5
2/13 ,! i . - 120o Condenser inspection reveals major damage to steam
*'i + bypass spargers.
1 f f 2/14 ' ll'; Major outage work: ii RCP P001 seal replacement 2/15 l!,ll l Steam bypass valve repair ll'll,' Condenser repaira 2/16 s'l,'ll Machining of RP turbine governor valve internals s; p. .a RCP stud inspection ( ? . i \m,/ 2/17 "l l,l Note: Drains on the turbine bypass lines ll l l;; downstream of the valves were enlarged I and the lines were insulated, after 2/18 -
,,'l,'
l which no further large condenser tube
; , leakage was experienced.
2/19 ,, , ,4
. .2 ,,.
i., ,, 2/204 -+t4--: , l ' l i , ,i,n _2/21 - l. 2/22 l: 4 e . f , p 2/23 l!!! ' l! Fill and vent of the RCS revealed that one steam -
. , , , generatop hot leg manway 8asket required g 2/24 l,'l*';' replacement. Replacement was initiated. j 2/25 !,l 2/26 ,
t t . 2/27 - l.. /"S 4 ( ) . . S/G gasket replacement complete.
~~
2/28'. 6.11
3;;p FIGURE 6.1.1 (continued)
* . if a a. o 15 5 $
s: H a. 2/28 t I I 5 120* l l I ! ! i 3/1 I i ! i 3/2 -jll l
' ' i 180*
=
3/3 ., 4 ii. lli! 4 i , , 4 230' 3/4 ,
!ll
, l! ;
i RCP P001 seal failed. 3/5 , 5
, p 3/6 -
; i .
3/7 . 3/8 l! 3/9 , ,
- e, !
L: 3/10 RCP P004 out of service due to failed fitting on
. f{ , '
__pp4 ; oil lift system. { } 3/11 e,,, w, RCS fill completed. 2 1; a I 3 3/12 = 150* Attempting to resolve mode restraints with the NRC. 3/13 - 3/14 - i 3/15 . l 3/16 = 3/17 = 3/18 Decision made that reactor trip breakers are a mode 2 restraint. 3/19 = 3/20 Water discovered in diesel generator G003 day tank Decision is made to recommence outage. Work 3/21 = includes: LP turbine intercept valves 3/22 - -' Feedwater bypass valves 6.12
v FIGURE 6.1.1 (continued) )
.u vwe i
. Le I
O CL 0 e ea
- m. e #2 3/22 q ,
5 150* 3/23 ,
'3/24 4 3/25 - ,
Main steam isolation valve 2HV4052 Marotta i oil control valve failed, repaired. 3/26 '.
.i r
.. . t 3/27- - l l
' ! 280* 3/29 4. , i
.i, i Safety Injection check valve inservice inspection
', , .' (ISI) performed.
3/30 .t . 4 l, l t 340* 3/31 -' '.
. l
. . '. :' Main Steam drain valve 2HV-8248 failed.
' +
4/1 4 '., 5 190* p -
. +,
4/2 ik w 4 -
P. 340* .
4/3 2 .' E. 2a
,, v.
. .e . 2 4/4 ,';,x .'il ,'l,
. 3
. .- 530, Pressurizer safety valve 2PSV-0200 found leaking.
4/5. l .
,,,i ,,
... 4 .
4/6 , .
, Hy&caet test of 2PSV-0200 performed.
!' 4/7 's
- f. 1 ! .
4 . .>f., . , 5 4/8 #l l.,l.'i 115' Outage recommenced. Work includes: Replacement of 2PSV-0200 4/9 =
', Replacement of 2 HP governor valve internals 4/10 =
s 4 . . 4/11 = l' .
. a , . . 1 .
p 4/12 -
,' , ,-i aj ' .4/13 d W W 4 i i 6.13 l
v w e4 FIGURE 6.1.1 (continued) O V
.u
@ Q. O e a2 O OO r s a.
4/13 .' -I 5 120' i 4/14 i 4/15 i [ i 1 4/16 lllll' 4/17 l',! ll
! i
. i . ; .i 4/18 l
, , , , , ,i l lll'l 4/19 Eight component cooling water containment ll'l'i l isolation valves required replacement of 4/20 'll',. their motor operators.
.4 ,.
4/21 .j'l,,;i
!? I i 5 e q e 1 3 4 4/22 4
~ i, l ll
,i 4/23 Z'. l' r 4 , i
!V 6 i !
4/24 ~l?>',,l!l 4 g
,, . i a a
4/25 '! 320*
"{
4/26 o. ,, Safety Injection check valve ISI completed. 4/27 7l l , l l li. 4 4/28 x 3 4/29 \ 545'
, . , NRC gives approval for mode 2 entry.
4/30 ", ,
=
5/1 l 5/2 2 5/3 zu s 5/4 y ' 1 SC i 155 5/5
-\
5 50: h 6.14
DG FIGURE 6.1.1 (continued) e, y o m. o e e2 o eo x e c. (m) 5/5 m 1 50% 20% Manual turbine trip due to HP governor valve G problems. T~ 5/6 " 50% 5/7 i ; 2PA-311-02 SBCS Performance retest. s. 5/8 .! l 2ST-344-09 Nuclear and Thermal Power Calibration
! performed.
5/9 , ! 3 0% Reactor trip due to electrical spike while j'!' ! switching PPS channel B power from inverter to 5/1C
, 1l- , [ l noninverter supply.
..!i +i ! 2 5/11 %! %; ; i 1 20%
, , _ . . m , 50%
1*; . wi .
=6 m.
5/12 . .. .i.
= 4 .
m . , 5/13 l' . 7 . i 2PA-344-05 PLCEA Xenon Control test performed.
!' l . !! '
5/14 i ,i i e> . 2PA-344-05 completed. 5/15 l' 1N, l 60%
-- . I 70%
80% 5/16 2LP-701-01 Biological Shield Survey commenced.
.,, 2ST-344-10 NSSS Calorimetric performed.
5/17
. l. 2ST-344-09 Nuclear and Thermal Power Cal. performed.
3 0% 2ST-311-02 SBCS Performance Test performed. - 5/18 . ', .': 2 Reactor trip on low S/G level due to inadvertent closure of feedwater containment isolation E 1iil ll:i 'l
, 1 20% valve 2HV-4048, 5 5/19 -' %i:ll
, W 50% Circulating water cooling reduced due to clogging 5/20 . .
by tuna crabs. 5/21
. . 1 i .
5/22 1:'. 2PA-457-02 Dropped CEA Test commenced.
!l
, i i 5/23 , . 'l',,
, i i 5/24 j f 2PA-457-02 completed.
5/25
-- i 5/26
, 2PA-457-01 Ejected CEA test commenced.
d 5/27 i 6.15
kG 0 v o &t FIGURE 6.1.1 (continued)
"0s c$
- O .
5/27 ...___;_, 1 50% 2PA-457-01 completed. 5/28 in 70% l 2PA-311-01 completed. 5/29 ; j 80% l 2PA-301-02 EVCS Performance Test performed. 5/30 H i l! i I . Circ. Water heat treatment performed. 5/31 ,'{', ! I 2ST-344-10 NSSS Calorimetric performed.
. . 2ST-344-09 Nuclear and Thermal Power Cal. performed.
6/1 ;i! il l 2PA-344-12 RCS Calorimetric Flow Measurement performed.
'!p 6/2
,i ;, i i
;_. 2PA-344-10 Variable Tavg Test commenced.
[ 6/3 l[jl iIh! NS ' 2PA-344-10 completed. 2PA-349-01 RRS Performance Test performed. 6/4 "l!$
'll'
, , l 2PA-350-01 Integrated Control System Performance 6/5 'l' o, Test performed.
,l
! 70%
Cleaning condenser waterboxes. 6/6 -
'l l, 6/7 i 2PA-311-01 SBCS Capacity Check retest performed. S-i ! n' l 55%
6/8 3 0% Reactor trip due to loss of feedwater pumps. l 6/9 =.
', {;; }
Reactor trip on high S/G level. i . i. 2 6/10 N l' ,,ll 1
- v. ^ 11 55%
6/11 , 6/12 - l
. ~ i 6/13 90%
='{ ,
lj'
,, 2PA-344-12 RCS Calorimetric Flow Measurement 6/14 - ' , l l r- retest performed.
- m 95%
2ST-344-09 Nuclear and Thermal Power Cal. performed. 6/15 ' 100% 2ST-344-10 NSSS Calorimetric performed. 3R3 4 SEiteFeh"Th%ilicMwer Cal. performed. 6/16 - 3 0% 2PA-344-12 RCS Cal. Flow Measurement performed.
/;, 4 345 Reactor trip due to high condenser backpressure.
I 6/17 - f- l RCP seals show indication of failure. An outage is
/' '
l , 5 required to repair them, 6/18 =! 145* 6.16
v tu n e
*T.
- o. o FIGURE 6.1.1 (continued) n4- ?8o
- c e n.
t -\ %) 6/18-i > ' 5 130' 6/19- , Draining of RCS restrained by work on CCW heat exchangers. 6/20- , i . 6/21 CCW heat exchanger work complete. 6/M -- , I i 6/23 l , s 6/24 i i J 6/25 ,
= ' '
6/26 Three broken and three loose bolts found on RCP 6/27 P004 rotating baffle. 8 i ! 6/28 N
! +
' 6/29 7 l . l i
, l
; ; Pressurizer spray valves repacked.
6/30 5 9 , ., ,; ,l l ', f , 7/1 -si'l, RCP seal replacement complete.
; 4 ..
,=, . .
7/2 ll l,!,
. t =- .
7/3 i 2: s 4 , i e je , , , 7/4 : l 4 4 ' ' RCS filling and venting commenced. 7/5
, ; i l: . :
7/6 ' ', , , 7/7 i. RCS filling and venting complete. 7/8 --
- p. 7/9 , ;
's"') 7/10 -
6.17
DG
'T
$ @$ FIGURE 6.1.1 (continued)
E $E 7/10 ;x ,
!y g 130' i :lj I I i \.
7/11 % >' , 3
+!i 545' 7/12 , ll! t t, 7/13 ", l,,
l'l ! 2 1 7/14 (; . , ! lN !! 50% 7/15- , 5, 60% i s-
,i , ,' 3 100% 2PA-204-01 Warranty Run recommenced.
7/16 , 2ST-344-10 NSSS Calorimetric performed.
' 2PA-458-01 Process Variable Intercomparison Test 7/17 i 'll performed.
7/18 ,',5 !) , i
;e ;. ,
'!! !' L 7/19 , ,;; :,ik 2PA-311-02 SBCS Performance Test performed.
,2 's -
7/20 2
';, :l i~,
2ST-344-10 NSSS Calorimetric performed. 2ST-344-09 Nuclear and Thermal Power Cal. performe 4 jil 2PA-344-06 Core Performance Record Test performed . 7/21' ',i
;C Reduced power to bump Circ Water pu mps. E
- a 2PA-307-02 Ek'CS Performance Test performed. S 7/22 Reactor Trip due to dropping a subgroup of part-length CEAs while controlling ASI.
7/23 .
,! l 'l' 2 7/24 x ll i- ? 3 100% 2PA -383-01 100% Generator Trip Test performed.
7/25 h '
'f3 Reactor Trip on high S/G 1evel.
7/26 'N 100% 2PA-349-01 RRS Performance Test performed.
. 2PA-350-01 Inte 7/27 , . Test performed. grated Control System Performance t "i 3 2PA-215-01 Natural Circulation Test commenced.
[L 4 330' 5 2PA-215-01 completed. N 7/29 1' 2 545* 7/30 7/31
' -hE 90%
20% Power reduction required due to seaweed clogging f circ water system.
~
** *
- E
- 8/1 6.18
n
'J kQ o a e
. T i8$
z e n. FIGURE 6.1.1 (continued) 8/1 y , , ,. !m 545% 2 l;;'; , l.i 3% Reaccor Trip on High Steam Generator Level 8/2 _ ,, ;,; , 2 '
. , , 1 8/3 .;w_A ':l s - - 50% Performing Unit Load Transient Test 2PA-344-02.
8/4 - , l }l, 65% 97 8/5 - ;
,.{l 8/6 -
i , .. 8/7 - - t i 8/8 ' Performing Variable T average Test 2PA-344-11. l' 8/9 = , , , , ', 8/10 , , i, 1u
. . , , o 8/11 8/12 _ _ ' -
8/13 , , 8/14 8'/15 - i 8/16 i 1 8/17 =
- ' i 8/18 -
Completed Warranty Run. Unit 2 Power Ascension Complete. 3 $ Q
~
~
5 s. O b. 6.19
A
\
I 6.2 CORE PCEER DISTRIBUTION TESTS
'Ihe two core power distribution tests, DROPPED CEA, 2PA-457-02,
- and PSEUDO EUECTED CEA, 2PA-457-01 were originally scheduled for performance at 50% pawer following the 80% plateau but were performed E at 50% power approximately midway through the 80% power plateau. 5 After starting the 80% testing, the condenser became partially -
blocked after an unusual migration of crabs forced a power reduction - to 50% power to allow cleaning of the condenser. It was decided to g do the 50%, post 80% testing at that time. By the time the 50% 3 testing was cmplete, the crabs were no longer a problem and the plant was able to return to the 80% testing plateau. O O 6.20
/O 6.2.1 DROPPED CEA (2PA-457-02) (/ Introduction & Sumary h e dropped CEA test consisted of dropping two Control Element Assemblies into the core and monitoring the subsequent nuclear and thermal-hydraulic respnses of the plant. First, a Full-Length Control Elment Assembly (FLCEA). was dropped; then, after reestab-lishing stable conditions, a Part Length Control Element Assmbly 1 (PICEA) was dropped. h is test was done only at 50% power, equil-brium Xenon, and with all CEAs fully withdrawn before each drop. Results were consistent with predictions and well within the accept-ance criteria. Measured core power distribations closely marched predicted. Pressurizer level, pressurizer pressure, RCS Hot Leg tmperatures, and steam generator pressures decreased to levels well within the Acceptance Criteria for bc,th rod drop events. Objectives This test had two independent objectives. The.first was to determine that the power distribution associated with the dropped CEAs were conservatively represented in the safety analysis. The'second objective was to verify acceptable plant performance by showing that specified parameters (i.e., pressurizer pressure and level, RCS Hot n
,) Leg temperatures, steam generator pressures) remained within accept- .
d able ranges following the CEA drop events. q S This procedure satisfied the comitments of Section 14.2.12.100 of a the FSAR. Test Method The first major step was stabilizing the reactor with the following initial conditions:
- ncminal 50% reactor power,
- bulk equilibrium Xenon, with axial Xenon stability,
- CEAs fully withdrawn,
- stable temperatures and pressure.
The CEA controllers (CEACs) were made inoperable, and Core Operating Limit Supervisory Systs (COLSS) adjustments were made to preclude CEAC/ CPC - generated trips or COLSS alarms during the test. O V 6.21
6.2.1 DROPPED CEA (2PA-457-02) (Cont.) Test Method (Cont.) Once all prerequisites were met, CEA 5-45 (see Figure 2.0.8 for location) was dropped by opening the appropriate CEA disconnect circuit breaker. Necessary data collection (CECOR snapshots, plant cmputer/startup cmputer/CFMS* trends, etc.) took place for one hour, af ter which the CEA was withdrawn. The conditions given above were re-established approximately 38 hours later and then CEA P-31 (a PICEA) was dropped. Af ter another hour of data collection, the PIfEA was withdrawn. Data reduction started with core power distribution cmparisons, which involved the following calculation for each of the 217 assemblies: (RPD fy ) (RPDiy ) = PR i (RPD. ) (RPD ) 1 i
. edicted ered
- where, RPD.y 1
= post-drop axial core averaged relative power density -
of assably i, RPD. 1
= pre-drop axial core averaged relative power density of assembly i, PR. = difference between the predicted arxl 1
measured post to pr H rop ratios of the relative power density of assably i. The associated acceptance criterion was that the absolute value of $3 PR. be less than 0.2. The measured RPD values were based on CECOR snapshots taken just prior to, and 15 minutes following the drop. - The predicted RPD values were based on calculations done by Combus-tion Engineering. The second phase of data reduction involved examination of the change in magnittxle of the various thermal-hydraulic parameters, which were trended by the startup cmputer, plant cmputer, and CFMS. The associated acceptance criteria are shown in Tables 6.2.1.1 and 6.2.1.2. A special test exception to the technical specifications was required to permit the dropped CEA to be misaligned from its group for nore than the normally permitted one hour in order to gather data for this test. Results For both CEA drop events, all values of PR (defined in previous section) were well within the acceptance criterion of +0.2 relative
- Critical Functions Monitoring Systm 6.22
O) (v 6.2.1 DROPPED CEA (2PA-457-02) (Cont.) Results (Cont.) power density units. Ebr the dropped full-leryth CEA (FILEA) (CEA 5-45) , the highest absolute value of PR was 0.033, which converts to a percent-age difference of 5.0%. %is value applies to an assembly digonally adjacent to the assembly containing the dropped CEA. %e assembly with the dropped CEA had a PR value of 0.022, while all other assem-blies had PR magnitudes of 0.025 or less. Ebr the dropped part-length CEA (PLCEA) (CEA P-31, see Figure 2.0.8 for location), the maximum value of PR was 0.03 which applies to the assembly containing the dropped PLCEA. All other assemblies had PR magnitudes of 0.011 or less. Pressurizer pressure, pressurizer level, reactor power, RCS Hot Iag temperatures, and stean generator pressures decreased as expected to levels well within the acceptance criteria. More specific informa-tion is provided in Table 6.2.1.1 (FILEA) and 6.2.1.2 (PIfEA) . te negative reactivity insertion frm the dropped full-length CEA caused a sudden power decrease, as shown on Figure 6.2.1.1. Se slight rise in power thereafter was the result of reactivity feedback O effects. (All figures related to this test have an 80-second x-axis, h where rod motions started at the 8 second point. Rad motion took seconds for each drop. All figures are based on startup ccuputer 2 data.) N We BCS Hot Leg temperatures decreased following the FILEA drop, as g indicated in Figure 6.2.1.2. tis behavior was due to a decrease in a reactor power coupled with a constant turbine load. Wis turbine / reactor mismatch also caused decreased stean generator pressures. Pressurizer level and pressure both decreased as shown on Figures 6.2.1.3 and 6.2.1.4 respectively. We lowered RCS temperatures led to a decrease in RCS volume, which caused flow out of the pressurizer and lower pressurizer level. t is decreased level increased steam volume, leading to lower pressurizer pressure. For the PLCEA drop, Figures 6.2.1.5 through 6.2.1.8 show that the changes in reactor power, RCS Hot Iag temperatures, and pressurizer
-level and pressure, were similar in direction but smaller in magni-tude than the changes associated with the FILEA drop. Relative to the FILEA, the part-length CEA is a weaker neutron absorber and inserts less negative reactivity into the core, thereby leading to a smaller perturbation of thermal-hydraulic parameters.
CEAs P-31 and 5-45 are closer to hot leg #2 and produce a greater effect I in hot leg #2 temperatures for reasons explained in the section on Pseudo Ejected CEA.(Section 6.2.2). n (/ Problems and Ccuments No significant problems were encountered during the test. 6.23
TABLE 6.2.1.1 DATA FIOM DROPPED CEA TEST FLCEA Initial Minimun Acceptance PARAMETER (Cmputer I.D.) Conditions Value Criteria Peactor Power (J007) 49.3% 46.6% - Reactor Power (J008) 50.5% 44.1% - Pressurizer Pressure 2250 PSIA 2195 PSIA ->2072.0 PSIA (P100X) Pressurizer Level 43.5 % 38.7 % >20.0% (L110X) m Stean Generator 1 Pressure 973 PSIA 934 PSIA ->814.0 PSIA J (P1013A) S Stean Generator 2 Pressure 978 PSIA 934 PSIA >814.0 PSIA S (P1023A) RCS Ibt Leg 1 Temperature 578.5 F 569.7 F >558.0 F (T111X) RCS Hot Leg 2 Temperature 579.0 F 574.9 F >558.0 F (T121X) NOI'E: Time of interest is fran initial rod notion to 60 seconds after rod is fully insertal. Initial and minimun values are . based on startup ca@ uter data. O 6.24
O -TABLE 6.2.1.2 DATA FEM DRCPPED CEA TEST PLCEA Initial Minimun Acceptance 4 PARAMETER (Ccaputer I.D.) - Conditions Value Criteria Reactor Power (J007) 48.1% 46.8% - Reactor Power (J008) 49.5% 47.8% - Pressurizer Pressure 2248 PSIA 2226 PSIA >2172.0 PSIA (P100X) " Pressurizer Level 43.4 % 41.6 % >29.0% (L110X) E 5 Stean Generator 1 Pressuce 948 PSIA - 925 PSIA ->865.0 PSIA (P1013A) (% Stean Generator 2 Pressure 958 PSIA 927 PSIA >865.0 PSIA (P1023A) RG Hot Leg l' Temperature 574.9 F 573.3 F >567.0 F (T111X) RCS Hot Leg 2 Temperature 575.0 F 571.9 F ->567.0 F (T121X) WrE: Time of interest is frca initial rod motion to 60 seconds after rod is fully inserted. Initial and minimun values are based on startup computer data. i 1 6.25
l 1 z toons g O I I Ii & I i h < O _ 1
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". $ s ob E 9 m & 8, 8,( -
e . H t g a < h O r3 < O L am 5 ) c 3 d I M - 9 _ m 9 P
- o
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f.2.2 PSEUDO EJECTED CEA (2PA-457-01) Introduction & Sumary O he pseudo ejected CEA test consisted of rapidly withdrawing a Control Elenent Assenbly (CEA) and nonitoring the subsequent nuclear and thermal-hydraulic response of the plant. Because the CEA was withdrawn, rather than ejected, the test is termed " Pseudo Ejected". This test was performed only at 50% power, equilibrium Xenon, and with the CEAS at the full power dependent insertion limit (FPDIL) . W e FPDIL is the deepest the rods can be inserted into the core at full power per Technical Specifications and represents the mximum reactivity available for an ejected rod. Results were consistent with predictions and well within the accept-ance criteria. Measured core power distributions closely matched predicted distributions. Pressurizer level, pressurizer pressure, RCS Hot Leg temperatures, and steam generator pressures increased, but to levels well within the Acceptance Criteria. Objectives This test had two. independent objectives. The first was to det' ermine that the power distribution associated with the pseudo CEA ejection fran the FPDIL CEA con #iguration was conservatively represented in the safety analysis. W e second objective was to verify acceptable n plant performance following the pseudo CEA ejection by showirg ~ that . specified parameters (i.e., pressurizer pressure and levsl, BCS Hot 1 Leg temperatures, steam generator pressures) renained within accept- @ able ranges. Wis procedure satisfied the ammitments of Section 14.2.12.99 of the FSAR. Test Method h e first major step was to stabilize the reactor with the following initial conditions: - naninal 50% reactor power, - bulk equilibrium Xenon, with axial Xenon stability, - CEAs at FPDIL (Group 6 at ~ 106.5 inches withdrawn) , - stable temperatures and pressure. De CEACs were made inoperable, and COLSS adjustments were made to preclude CEAC/CPC - generated trips or COISS alarms durirg the test.
- 6.34
6.2.2 PSEUD 0 EJECTED CEA (2PA-4E7-01) (Cont. ) Test Method (Cont.) Once all prerequisites were met, CEA-22 (a Group 6 CEA, shown in Figure 1.5 2.0.8) was minutes. continuously Appropriate withdrawn, data collectiona (CE00R process snapshots, which took approximately plant cmputer / start-up cmputer/CENS* trends, etc.) took place for one hour. Men CEA-22 insertion was empensated by CEA-21 (a symetric Group 6 CEA) withdrawal. In this way, the radial power distribution for a similiar ejected CEA was measured. After another hour of data collection, CEA-21 was inserted to the level of the remainder of Group 6. Data reduction started with core power distribution cmparisons, which involved the following calculation for each of the 217 assemblies:
'(RPD gy -
{RPD gy ) = PR f WD I ' ( 1 ed predicted where, RPD gy = post ejection axial core averaged relative power density of assenbly 1, , RPD g = pre-ejection. axial core averaged relative' power p o density.of assembly 1, PR g = difference between the measured and predicted post , and pre-ejection ratios of the relative' power. , density of assembly i. q S t e associated acceptance criterion was that PR. not differ from m the predicted ratio by greater than 20%. % e mbasured RPD values. ' were based on CIEOR snapshots taken just prior to, and 15 minutes following the ejection. The predicted RPD values were based on calculations done by Cmbustion Engineering. Se second phase of data reduction involved exanination of the change in magnittde of the various thermal-hydraulic parameters, which were trended by the startup cmputer, plant computer, and CFMS. The associated acceptance criteria are shown in Table 6.2.2.1. Results For CEA 6-22, the maximum value of PR (defined in previous section) was 0.033, which converts to a percentage difference of 2.8%. This value applies to the assacbly containing the CEA. Two neighboring assemblies had PR values near 0.01 while alnost all other assenblies
- Critical Functions Monitoring Systen 6,35 i
6.2.2 PSEUDO EJECTED CEA (2PA-457-01) (Cont.) Results (Cont.) had far smaller PR values. We ejection of CEA 6-21 (performed one hour after 6-22) showed similar, but slightly larger, PR values. m is small difference can be attributed to radial Xenon redistribu-tion caused by the previous withdrawal of CEA 6-22. Pressurizer pressure, pressurizer level, reactor power, RCS Hot Leg temperatures, and steam generator pressures increased as expected to levels well with the acceptance criteria. More specific infor: ration is provided in Table 6.2.2.1. We positive reactivity insertion frm the steady CEA withdrawal caused a smooth power increase, as shown on Figure 6.2.2.1. We smewhat constant power level thereafter is the result of reactivity feedback effects. (All figures related to this test have a 160-second X-axis, where red notion started at the 10-second point. We rod withdrawal required 90 seconds. All figures are based on startup cmputer data for the withdrawal of CEA 6-22.) We RCS Hot Leg temperatures increased following the pseudo ejection, as indicated in Figure 6.2.2.2. B is behavior was due to an increase in reactor power coupled with a constant turbine load. We turbine / " reactor mismatch also caused higher steam generator pressures.
- Pressure increased frm 920 psia to 960 psia but at no time approached the upper limit of 1002 psia. Pressurizer level and E pressure both increased as shown on Figures 6.2.2.3 and 6.2.2.4. We 5 higher RCS temperatures led to an increase in RCS volume, which caused flow into the pressurizer and an associated pressurizer level rise. tis increased level decreased stean volume, and led to a higher primary pressure.
CEA 6-21 is closer to hot leg #2, whose temperature increase was greater than that of hot leg #1 due to the absence of significant mixing in the upper plenum under forced flow. We dropped CEAs (see Section 6.2.1) are closer to hot let #2 and also produced a greater affeet on temp-eratures in that hot leg. Problems and Ccmrents No significant problems were encountered during the test. O 6.36
bV TABLE 6.2.2.1 DMA F10M PSEUDO EJECIED CEA TEST Initial Maximun Acceptance PARAMEIER (Ccaputer I.D.) Conditions Value Criteria i Reactor Power (J007) 49.0% 50.3% - , Reactor Power (J008) 50.5% 51.4% - Pressurizer Pressure 2242 PSIA 2271 PSIA -<2367 PSIA (P100X) Pressurizer Level 42.71% 44.71%- 149% (L110X)
< m Stean Generator 1 Pressure 920.7 PSIA 957 PSIA 11002 PSIA d (P1013A) g Stean Generator 2 Pressure 930.3 PSIA 959 PSIA m
-<1002 PSIA (P1023A)
RG Hot Leg 1 Temperature 574.1 F 575.7 F 1585 P (T111X) O RCS Hot Leg 2 Tenperature 574.6 F 577.0 F 1585 F (T121X) NOTE: Time of interest is fran initial rod notion to 60 seconds after rod is fully withdrawn. Initial and maximun values are based on startup cmputer data. j i O i 6.37 i
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l l 6.3 VARIABLE T average (2PA-344-08,-09,-10,-11) S e Variable T average tests were performed to determine reactor l irotnermal temperature reactivity coefficient and power reactivity emfficient at 20, 50, 80, and 100% reactor power. The tests are sequentially numbered by power plateau so that 2PA-344-08 was the 20% test, -09 the 50%, etc. The Variable T average test satisfied FSAR Ctanitment 14.2.12.89. Testing through 100% power has met all acceptance criteria. Method te variable T average test involved measurements of the charges in reactivity caused by changes in temperature and power. E is test had three parts which measured both of these effects separately and then the ratio of the tw. - Se temperature effect is known as the Isothermal Temperature Coefficient of reactivity (IK) and is a measure of the change in reactivity due to a change in average reactor coolant temperature. Eis measurement gives the test its name (i.e., variable N T avg). Note, however, that a change in reactor coolant temperature - at constant reactor power causes an equal charge in core average fuel g temperature, hence the term " isothermal." Relative to the change in g coolant temperature, this charge in fuel temperature has a smaller but gignificant effect on reactivity. We IK has units of delta rho /'F. S e IK was measured by changing T average using turbine loading and holding power constant using CEA Group 6 . Physics predicitions provided by the NSSS vendor were used to calculate the worth of the rod notion. %e IK was calculated frca the change in reactivity due to the change in rod position and the corresponding change in temperature. We change in reactivity caused by a change in reactor power (constant core average coolant temperature) is defined here as the Power Coefficient of reactivity (PC) . It has units of delta rho /% power. W e Power Coefficient was measured by holding temperature constant usirg tubine loading while varying power using Group 6 CEA motion. Se PC was then determined frcru the $hange in reactivity worth due to noving the rods and the change 16 power. We third part of the test was the ratio measurement. In this part of the test, the temperature was varied and power allowed to swing freely. Bis directly measured the change in power due to a charge in temperature which is the ratio of the IK to the PC. We ratio measurement has units of % F. At 20, 50, and 80% power the IK and PC were measured and the ratio of the two taken for information only. It was hoped that the 100% testing could be performed with no rods in the core. E is could have been done if the predicted and measured PC's were alIrost the same value. By measuring the ratio of the IK to the PC and having a very accurate preaiction of the PC, the ITC could be determined. While the measured PC's 6.42 O
6.3 VARIABLE T average (2PA-344-08,-09,-10,-11) (ov) (Continued) were within the acceptance criteria through 80%, an effort was made to improve the accuracy of the predictions based on actual test conditions. 21s included accounting for core burnup boron concen-
. tration and exact moderator temperature as well as recalculation of the rod _ worths based on the bottm 6" being made of AgInCd rather than B3C. Becaluclation of the results through 80% based on the new rod wotths gave closer agreement between measured and predicted values but was still not exact. Since the I'IC is an important value, it and the ratio measurements were made and frm these values a PC was calculated.
We three methods are illustrated below in Figure 6.3.1, with example data shown frm actual chart traces in Figure 6.3.2. We ITC is the sm of moderator temperature coefficient (MIC) and fuel temperature coefficient (PIC) . Moderator temperature coefficient is calculated by subtracting the fuel temperature coefficient frm the I'IC. . ne fuel temperature coefficient is precalculated by the reactor vendor. " A* E Test Results 5 p te test results for 20%, 50%, 80% and 100% power levels are shown below. There is acceptance criteria based on Rated termal Q Power (RrP) which is also listed. %e predicted Fuel Temperature coefficients (PICS) are also included. % e measured ratio is included so it can be cmpared to the measured I'IC/PC ratio to show the close agreement. 20% Power Iavel Parameter Value Acceptance Criteria TIC -0.628 x10 -4 A k/k/ F (-0.568_+0.5) x10
-4 A k/k/ F PC -1.065 x10 -4 A k/k/% (-1.1225+0.2) x10-4 6 k/k/%
MIC-20% -0.483 x10 -4 AMF None MIC-RTP -0.796 x10 -4 A k/k/ F -2.5x10-4<MIC<0.13x10-4 Ak/k/ F FIC -0.145 x10-4 A k/k/OF None Boron Concentration 660 ppu None Measured = U "U"' Ratio 0.624%/ F (s Measured = 0.589%/ F TIC /PC 1 6.43 1
y. ! 50% Power Level Parameter Value Acceptance Criteria
-4 -4 M/F O IK -0.824 x10 ok/k/ F (-0.82710,5) x10
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PC -1.104 x10 ok/k/% (-0.954310.2)x10
-4 MIC-50% -0.692 x10 ok/k/ F None MIC-RrP -1.027 x10-4 Ak/k/ F -2.5x10-4<MTC<+0.13x10-4 Ak/k/ F
-4 PIC -0.132 x10 ok/k/ F None Boron Concentration 559 ppu None Measured Ratio
= 0.744%/ F Measured "
= 0.746%/ F IK/PC N
S a 80% Power Level - Parameter Value Acceptance Criteria _
-4 IK -0.942 x10 -4 6k/k/ F (-1.04210. 5) X10 ak/k[F
-4 PC -0.946 x10-4 6 k/k/OF ( .87010.2)X10 Ak/k/%
-4 #
MIC-80% -0.819 x10 a k/k/ F None MrC-RTP -0.930 x10 -4 A k/k/ F -2.5 X10-4 <MK<+0.13X10-4 ok/k/ F
-0.123 x10 -4 ne U
PIC A k/k/ F Boron Concentration 512 ppn None Measurd = Ratio 0.960%/ F Measured "
= 0.995%[F IK/PC 6.44
1 l i i 100% Power Level Parameter Value Acceptance Criteria i
-1.037 x10-4
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. Calculated _4 _4 '
PC -0.947 x10 ak/k/t , ( .796_+0.2)x10 ak/k/t f _4 I MEC-RFP -0.919 x10 ak/k/ F -2.5x10_4<.VIC<0.13x10_4 ak/k/ F l N
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3 PIC - -0.118 x10 Ak/k/ F None ; a a ? i Boron Concentration.483 ppm None y j 4 , Mne Measured Ratio = -1.095%/ F i i Problems i 4 J No major problems were. encountered. All acceptance criteria were met. i 1 i s I 3 8 if 3 '1 6 h i i b e o
- 6.45 1
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6.4 UNIT IDAD TRANSIENT TEST (2PA-344-02) Purpose 2e purpose of this test was to dmonstrate that the unit can perfom as designed during had changes. In Mdition, data was collected to verify the ability of the NSSS simulator, Cmbustion Engineering Systems Excursion Code (CESEC), to mdel plant performance during a transient. %is test satisfied FSAR section 14.2.12.90. Method With the Reactor and Turbine at 50% power and all control systems in autmatic, a series of ramp and step changes of 10% power were made. First, a ramp decrease frm 50% to 40% was initiated at approximately 5% per minute. Bis was accmplished by decreasing turbine loading and allowing the Reactor Regulating System (RBS). to cmpensate for the Reactor /rurbine mismatch by driving rods into the wre. We RRS nonitors average moderator temperature (T average) and turbine load as measured by turbine first stage pressure. Using first stage turbine pregsure, the RRS calculates a T reference. If a mismatch of nore than 2 F is seen between T average and T reference, the RRS will demand withdrawal or insertign of the CEA's as appropriate until the tmperatures are within the 2 F band. If the mismatch is large N enough, the RRS will demand a high rate withdrawal or insertion as - appropriate. a After stabilization was achieved at 40% power, a step change was made to 30% turbine power. Af ter reactor power had stabilized at 30%, a step change to 40% turbine power was made. t en, after stable reactor power was achieved, a ramp increase in turbine pwer to 50% at approximately 1%.per minute was initiated. Wroughout the transients and stabilization periods, key plant parameters were monitored on the Plant Cmputer, Startup Cmputer, Critical Functions Monitoring System and Brush recorders. %e analysis of this data allowed verification of the proper perform-ance of the NSSS contrnl systems during the transients. Due to an incressing Xenon concentration at the initiation of the test and further Xenon concentration changes caused by power level changes during the test, the RRS withdrew the Group 6 CEA's to the upper electrical limit during the power increase frm 30% to 40% power. At that time, an approximate 2% mismatch between reactor power and turbine power existed. W is mismatch was then absorbed by the NSSS in the fom of a reauction in primary coolant temperature. O 6.48
. . . - - . - _ . _ . - - - . . - . .~.- . - . . . _ _
i F : ) ! i j 6.4 UNIT IDAD TRANSIENP TEST (2PA-344-02) (Cont.) ~i i Method (Cont.) i t
- An inability to achieve the desired ramp rates and precise step
.l changes was experienced. This was due to not having precise turbine ; controls. Ioad changes were made by pressing a load increase or , decrease button. The frequency and the length of time that the button i was pressed determined the amount and the rate of the load change. i Cold leg te 9 rature was reduced below the Technical Specification j limit of 544 F. 2 1s was because T reference does not track
- properly below 100%. (See Reactor Regulating Systen Test, Section.
- 6.5.4 for a further discussion of this problem). W us, when the RRS
systen is in automatic at low power levels, T average is depressed . below its programmed value.- , Results i Figures 6.4.1 through 6.4.5 show the response of various key NSSS [ parameters during the test. The beginning of each transient is < i marked with an arrow below the tbne axis. Listed below is a key to l
! the letters used:
! Designator Transient . 1 E j .A 50% to 40%'at 5% per minute & m l B 40% to 30% step change > I ~C 30% to.40% step diange I D 40% to 50% at 1% per minute ) In general, the figures illustrate that the plant responded as i
. designed. All single value. acceptance criteria from the CESEC predictions were met as shown in Table 6.4.1. A discussion of each' graph follows. ;
r i . Figure 6.4.1 shows pressurizer level and pressure as a function of-l time.. The pressurizer level.followed the power dependent setpoint with approximately a 2% overshoot. : Pressurizer pressure remained i relatively stealy during the transients. The maximum deviation was F . 20 psia from the setpoint. Figure 6.4.2 illustrates Generator Ioad and Reactor Power as a function of time. The graph clearly shows that ~the desired rang 6.49 t' "
6.4 (NIT IDAD TRANSIENT TEST (2PA-344-02) (Cont.) Results (Cont.) rates for transients A and D were not achieved due to turbine / generator control limitations discussed earlier. 'Ihe priod between transients C and D was of an extended duration due to the necessity of having to dilute CEA Group 6 back into the core. It had been necessary for the RRS to withdraw Group 6 to ccarpensate for increased Xenon concentrations due to the load reductions. Thus it was neces-sary to insert the rods back into the core prior to the ramp increase frm 40% to 50% guer. Figure 6.4.3 illustrates steam generator pressure as a function of time. Immediately af ter each transient was initiated, pressure peaked approximately 40 psia above the setpoint, but quickly returned ~ to normal. . 2 Figure 6.4.4 illustrates hot leg temperature as a function of time. & m In general, the temperature followed the T average program as demanded by the Beactor Regulation System (RRS) . Figure 6.4.5 displays steam generator level as a function of time. During the transients, level deviations of approximately 10% were observed. 'Ihe Feedwater Cbntrol Systm quickly restored the levels to the setpoint. Conclusions 'Ihe plant performed as designed during the transients. All accept-ance criteria were met. tb major problems were encountered. O 6.50
. _ . _ . . _ ._. _ ___. _ _ _ _ .=._. .._ _ _ _ _ _ _ . . . _ . _ _ _ _ . _ _ _ . _ _ _ _ _ __ _ _ _ _ . _ . _ _ .
e i ~ l L i I i i Table 6.4.1 i { Single Value Acceptance Criteria l 1. i-il i maad ; l Parameter Maximtsn or Minimm Acceptance Criteria
- ~
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Pressurizer Level 32% > 24% g a m . Pressurizer Pressure. 2250 psia i 2330 psia . Stean Generator Pressure 980 psia 11011 psia . T average 561 F > 556 F ! i i
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^ Figure 6.4.1 Pressurizer Pressure and Level vs Time j % During linit Load Transient Test o.
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Figure 6.4.2 ReactorPowerh../GeneratorLoadvsTime y During Unit Load Transient Test e u
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Figure 6.4.3 Steau Generator #2 Pressure vs Time During Unit Load Transient Test
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