Text

Startup Test Rept
	ML20210T484
Person / Time
Site:	Salem, 05000000
Issue date:	09/29/1977
From:	Czuchnicki E, Harrick J, Jackson J; Public Service Enterprise Group
To:	;
Shared Package
ML20210T454	List: ML20210T449; ML20210T457; ML20210T464; ML20210T470; ML20210T478; ML20210T484; ML20210T493; ML20210T500;
References
	FOIA-86-236 NUDOCS 8605300476
	Download: ML20210T484 (305)
	v • d • e

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O SALEM GEi4ERATING STATI0ti l UNIT 1

STARTUP TEST REPORT The Energy People

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8605300476 860521 PDR FOIA WILLIAM 86-236 PDR J7 ik. ; it, ; j.

,m PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION UNIT NO. 1 STARTUP REPORT BY REACTOR ENGINEERING STAFF E. P. CZUCHNICKI J. R. HARRICK J. G. JACKSON E. A. PEARCE E. V. ROSCIOLI APPROVED:

REACTOR ENGINEER: , . It h.A.dICHOLS !

CHIEF ENGINEER: d N <-t/c 5//f/

fJ.MIDURA/

STATION MANAGER: b h/ , , M H. I. HELLER

1 TABLE OF CONTENTS PAGE NO.

LIST OF FIGURES iv LIST OF TABLES xvii

1.0 INTRODUCTION

1 2.O INITIAL CORE LOADING 19 3.0 PRECRITICAL TESTING 53 3.1 SUP 70.2 - RTD BYPASS FLOW VERIFICATION 53 3.2 SUP 70.3 - REACTOR COOLANT FLOW COASTDOWN MEASUREMENTS 54 3.3 SUP 70.4 - PRESSURIZER SPRAY AND HEATER CAPABILITY 61 3.4 SUP 70.5 - REACTOR COOLANT SYSTEM FLOW MEASUREMENT 65 3.5 SUP 70.6 - ROD POSITION INDICATION SYSTEM 81 3.6 SUP 70.7 - ROD DRIVE MECHANISM TIMING 82 3.7 SUP 70.8 - CONTROL ROD DROP TIME MEASUREMENTS 85 3.8 SUP 70.9 - FULL LENGTH ROD CONTROL SYSTEM OPERATIONAL TEST FOR INITIAL CRITICALITY 95 3.9 SUP 70.10 - PART LENGTH ROD MECHANISM BRAKE TEST 96 3.10 SUP 70.11 - INCORE FLUX MAPPING SYSTEM 97 4.0 INITIAL CRITICALITY 100 1

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TABLE OF CONTENTS

.s 5.0 PHYSICS TESTING 113 ZERO POWER TESTS 5.1 SUP 80.3 - NUCLEAR DESIGN CHECK TESTS 113 5.2 SUP 80.4 - ROD AND BORON WORTH MEASUREMENTS DURING BORON DILUTION AND ADDITION 148 5.3 SUP 80.5 - RCCA PSEUDO EJECTION AT ZERO POWER 158 5.4 SUP 80.6 - MINIMUM SHUTDOWN VERIFICATION AND STUCK ROD WORTH MEASUREMENT 161 AT POWER TESTE 5.5 SUP 81.8 - POWER COEFFICIENT AND INTEGRAL POWER _.

DEFECT MEASUREMENT 164 5.6 SUP 81.9 - RCCA PSEUDO EJECTION AND RCCA ABOVE BANK POSITION MEASUREMENT 173 5.7 SUP 81.10 - STATIC RCCA DROP AND RCCA BELOW BANK POSITION MEASUREMENT 177 5.8 SUP 81.11 - INCORE AND EXCORE DETECTOR CALIBRATION 188 5.9 POWER AND BURNUP DISTRIBUTION MEASUREMENTS 201 6.0 INTEGRATED PLANT RESPONSE TESTS 213 l

6.1 SUP 81.5 - DYNAMIC AUTOMATIC STEAM DUMP CONTROL 213 l 6.2 SUP 81.3 - TURBINE OVERSPEED TRIP TEST 215 6.3 SUP 82.6 - LOSS OF OFF-SITE POWER 216 6.4 SUP 82.5 - SHUTDOWN FROM OUTSIDE THE CONTROL ROOM 217 6.5 SUP 82.1 - LOAD SWING TEST 220 6.6 SUP 82.4 - RODS DROP AND PLANT TRIP 229 6.7 SUP 82.2 - LARGE LOAD REDUCTION TEST 230 6.8 SUP 82.7 - STEAM GENERATOR MOISTURE CARRYOVER

, MEASUREMENT 237 l l

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TABLE OF CONTENTS 6.9 SUP 82.8 - NSSS ACCEPTANCE TEST 239 6.10 SUP 82.9 - GENERATOR TRIP FROM 100% POWER 241 6.11 RADIATION SHIELDING EVALUATION, EFFLUENT MONITORING, CHEMISTRY AND RADIOCHEMISTRY TESTS 247 7.0 CALIBRATION OF TEMPERATURE AND FLOW INSTRUMENTATION DURING POWER ESCALATION 253 7.1 SUP 80.7 - TURBINE CONTROL SYSTEM CHECKOUT AND STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM 253 7.2 SUP 81.7 - CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION AT POWER 255 7.3 SUP 81.12 - NUCLEAR AND TEMPERATURE INSTRUMENTATION CALIBRATION AND THERMAL POWER MEASUREMENT 258 8.0 XENON FOLLOW TEST 276 APPENDIX A OPERATING LICENSE AND AMENDMENT HISTORY 280 APPENDIX B 0% POWER FLUX MAPPING TECHNIQUE 284 ACKNOWLEDGEMENTS 288 I

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LIST OF FIGURES FIGURE TITLE PAGE NO.

1.1 STARTUP CHRONOLOGY I 7 1.2.1 STARTUP CHRONOLOGY II 8 1.2.2 STARTUP CHRONOLOGY II 9 1.3.1 DECEMBER 1976 POWER HISTORY 10 1.3.2 JANUARY 1977 POWER HISTORY 11 1.3.3 FEBRUARY 1977 POWER HISTORY 12 1.3.4 MARCH 1977 POWER HISTORY 13 i 1.3.5 APRIL 1977 POWER HISTORY 14 1.3.6 MAY 1977 POWER HISTORY 15 1.3.7 JUNE 1977 POWER HISTORY 16 1.3.8 JULY 1977 POWER HISTORY 17 e

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LIST OF FIGURES FIGURE TITLE PAGE NO.

2.1 KEFF VS. NUMBER OF ASSEMBLIES 25 2.2 1/M VS. NUMBER OF ASSEMBLIES 26 2.3 KEY SHEET FOR CORE LOADING SCHEMATICS 27 2.4 CORE LOADING SEQUENCE STEPS 1 TO 9A 28 1

2.5 CORE LOADING SEQUENCE STEPS 9B TO 9D 29 2.6 CORE LOADING SEQUENCE STEPS 10 TO 17B 30 2.7 CORE LOADING SEQUENCE STEPS 18 TO 71 31 2.8 CORE LOADING SEQUENCE STEPS 72 TO 98 32 2.9 CORE LOADING SEQUENCE STEPS 99 TO 105C 33 3

2.10 CORE LOADING SEQUENCE STEPS 106 TO ll7B 34 2.11 CORE LOADING SEQUENCE STEPS 118 TO 160 35 2.12 CORE LOADING SEQUENCE STEPS 161 TO 193 36

2.13 ICRR VS. NUMBER OF ASSEMBLIES LOADED TEMPORARY INCORE CHANNEL A 37 2.14 ICRR VS. NUMBER OF ASSEMBLIES LOADED ~

TEMPORARY INCORE CHANNEL B 38 2.15 ICRR VS. NUMBER OF ASSEMBLIES LOADED TEMPORARY INCORE CHANNEL C 39 6

t LIST OF FIGURES FIGURE TITLE PAGE NO.

2.16 ICRR VS. NUMBER OF ASSEMBLIES LOADED PLANT CHANNEL SR-31 40 2.17 ICRR VS. NO. OF ASSEMBLIES LOADED PLANT CHANNEL SR-32 41 2.18 BORON CONCENTRATION VS. TIME 42 2.19 CORE LAYOUT 43 2.20 INSTRUMENTATION LOCATIONS 44 6

LIST OF FIGURES i FIGURE TITLE PAGE NO.

I 3.2.1 CORE FLOW 2/4 COASTDOWN 57 3.2.2 CORE FLOW 4/4 COASTDOWN 58 3.2.3 CORE FLOW 2/3 COASTDOWN 59 3.2.4 CORE FLOW 3/3 COASTDOWN 60

l 3.3.1 PRESSURIZER SPRAY EFFECTIVENES'S 63 I

3.3.2 PRESSURIZER HEATER EFFECTIVENESS 64 3.4.1 LOOP 11 ERROR ANALYSIS 69 3.4.2 LOOP 12 ERROR ANALYSIS 70 3.4.3 LOOP 13 ERROR ANALYSIS 71 3.4.4 LOOP 14 ERROR ANALYSIS 72 3.6.1 NOMINAL CRDM STEPPING TEST TRACE 83 3.6.2 CRDM STEPPING TRACE CORE LOCATION H-8 84 3.7.1 ROD DROP TIMING TRACE - COLD NO FLOW 89 3.7.2 ROD DROP TIMING TRACE - COLD FULL FLOW 90 vii . . . . _ - . . - _ . - . . . - . . . - . . - . . - -

LIST OF FIGURES FIGURE TITLE PAGE NO.

4.1 INVERSE COUNT RATE RATIO VS. PART LENGTH BANK POSITION 103 4.2 INVERSE COUNT RATE RATIO VS. SHUTDOWN BANK POSITION 104 4.3 INVERSE COUNT RATE RATIO VS. CONTROL BANK POSITION 105 4.4 INVERSE COUNT RATE RATIO VS. RCS DILUTION TIME 106 4.5 INVERSE COUNT RATE RATIO VS. RCS DILUTION WATER 107 4.6 INVERSE COUNT RATE RATIO VS. RCS BORON CONCENTRATION 108 4.7 REACTIVITY COMPUTER ARRANGEMENT 109 4.8 NIS OVERLAP 110 e

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LIST OF FIGURES FIGURE TITLE PAGE NO.

5.1.1 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, ARO 122 5.1.2 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, C/B "D" AT 0 123 5.1.3 ISOTHERMAL TEMPERATURE COEFFICIENT-MEASUREMENT, C/B "D, C" AT 0 124 5.1.4 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, C/B "D, C, B" AT 0 125 5.1.5 ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, C/B "D, C, B, A" AT 0 126 5.1.6 ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT ,

VS. BANK POSITION 127 5.1.7 FT,1K MAP 104, HZP, ALL RODS OUT (FDHN, AVERAGE AXIAL KW/FT) 128 5.1.8 FLUX MAP 104, HZP, ALL RODS OUT (POWER TILTS, RELATIVE POWER) 129 5.1.9 FLUX MAP 102, HZP, C/B "D" AT 0, C/B "C" AT 204 130 5.1.10 FLUX MAP 103, HZP, BANKS "D" AND "C" IN (FDHN, AVERAGE AXIAL KW/FT) 131 5.1.11 FLUX MAP 103, F 4P, BANKS "D" AND "C" IN (POWER TILTS, RELATIVE POWER) ,

132 i 5.1.12 FLUX MAP 106, HZP, C/B "D" AT 5, C/B "C" AT 68, B-6 OUT (FDHN, AVERAGE AXIAL KW/FT) 133 5.1.13 FLUX MAP 106, HZP, C/B "D" AT 5, C/B "C" AT 68,

(. B-6 OUT (POWER TILTS , RELATIVE POWER) 134

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LIST OF FIGURES FIGURE TITLE PAGE NO.

5.1.14 FLUX MAP 107, 3% POWER, C/B "D" AT 120 (FDHN, AVERAGE AXIAL KW/FT) 135 5.1.15 FLUX MAP 107, 3% POWER, C/B "D" AT 120 (POWER TILTS, RELATIVE POWER) 136 5.1.16 INTERCHANGE BETWEEN 2.25% AND 3.30% ASSEMBLY 137 5.1.17 INTERCHANGE BETWEEN 2.25% AND'2.80% ASSEMBLY, BURNABLE POISON RODS BEING RETAINED BY THE 2.80%

ASSEMBLY 138 5.1.18 INTERCHANGE BETWEEN 2.25% AND 2.80% ASSEMBLY, BURNABLE POISON RODS BEING TRANSFERRED TO THE 2.25% ASSEMBLY 139 l

5.1.19 ENRICHMENT ERROR: A 2.80% ASSEMBLY LOADED INTO THE CORE CENTRAL POSITION 140 5.1.20 LOADING A 2.80% ASSEMBLY INTO A 2.25% ASSEMBLY POSITION NEAR CORE PERIPHERY 141 5.1.21 % DIFFERENCES BETWEEN MEASURED AND EXPECTED REACTION RATE INTEGRALS FOR THE ARO 0% POWER MAP 142 5.1.22 . FLUX MAP 108, D 9 108, D 9 127 (FDHN, AVERAGE AXIAL KW/FT) 143 5.1.23 FLUX MAP 108, D 9 127 (POWER TILTS, RELATIVE POWER) 144 5.2.1 DIFFERENTIAL AND INTEGRAL CONTROL BANK D WORTH 150 b 5.2.2 DIFFERENTIAL AND INTEGRAL CONTROL BANE C WORTH 151 X

LIST OF FIGURES FIGURE TITLE PAGE NO.

5.2.3 DIFFERENTIAL AND INTEGRAL CONTROL BANK B WORTH 152 1

5.2.4 DIFFERENTIAL AND INTEGRAL CONTROL BANK A WORTH 153 5.2.5 DIFFERENTIAL AND INTEGRAL SHUTDOWN BANK D WORTH 154 5.2.6 DIFFERENTIAL AND INTEGRAL SHUTDOWN BANK C WORTH 155 5.2.7 DIFFERENTIAL AND INTEGRAL CONTROL BANK OVERLAP WORTH 156 5.3.1 DIFFERENTIAL AND INTEGRAL RCCA B-6 WORTH 160 5.5.1 POWER COEFFICIENT VS. POWER 167 5.5.2 POWER DEFECT 168 5.5.3 DOPPLER ONLY DIFFERENTIAL POWER COEFFICIENT 169 5.5.4 DOPPLER DEFECT 170 5.5.5 AI VARIATION DURING POWER COEFFICIENT TEST 171 5.6.1

SUMMARY

OF NUCLEAR AND TEMPERATURE INSTRU-MENTATION VS. EJECTED ROD POSITION 176 5.7.1 MAP 114, 50% POWER, ARO (FDEN, AVERAGE AXIAL KW/FT) ,

180 5.7.2 MAP 114, 50% POWER, ARO (POWER TILTS, RELATIVE POWER) 181 b 5,7.3 MAP 115, 50%, C/B D @ 200, D-8 DROP (FDHN, AVERAGE AXIAL KW/FT) . 182 l

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LIST OF FIGURES FIGURE TITLE PAGE NO.

5.7.4 MAP 115, 50%, C/B D @ 200, D-8 DROP (POWER TILTS, RELATIVE POWER) 183 5.7.5 MAP 116, 50% POWER, K-6 DROP (FDHN, AVERAGE AXIAL KW/FT) 184 5.7.6 MAP 116, 50% POWER, K-6 DROP (POWER TILTS, RELATIVE POWER) 185 5.7.7 AVERAGE ASSEMBLYWISE POWER DISTRIBUTION, H-4 DROPPED 186 5.7.8 MAP 115, 50% POWER, D-8 DROP (MEASURED AND EXPECTED FDHN) 187 5.8.1 % POWER VS. TOTAL CHANNEL CURRENT 191 5.8.2 N-41 INCORE VS. EXCORE AXIAL OFFSET 192 5.8.3 N-42 INCORE VS. EXCORE AXIAL OFFSET 193 5.8.4 N-43 INCORE VS. EXCORE AXIAL OFFSET 194 5.8.5 N-44 INCORE VS. EXCORE AXIAL OFFSET 195 5.8.6 N-41 INCORE AXIAL OFFSET VS. EXCORE CURRENT 196 5.8.7 N-42 INCORE AXIAL OFFSET VS. EXCORE CURRENT 197 5.8.8 N-43 INCORE AXIAL OFFSET VS. EXCORE CURRENT 198 5.8.9 N-44 INCORE AXIAL OFFSET VS. EXCORE CURRENT 199 5.8.10 INCORE AXIAL OFFSET VS. EXCORE AXIAL OFFSET 200 e

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LIST OF FIGURES FIGURE TITLE PAGE NO.

5.9.1 MAP 121, 75% POWER, C/B D @ 226 203 5.9.2 MAP 121, 75% POWER, C/B D @ 226 204 5.9.3 MAP 133, 90% POWER, C/B D @ 195 205 5.9.4 MAP 133, 90% POWER, C/B D @ 195 206 5.9.5 MAP 138, 100% POWER, C/B D @ 220 207 5.9.6 MAP 138, 100% POWER, C/B D @ 220 208 209 5.9.7 MAP 141, 100% POWER, C/B D @ 207 5.9.8 MAP 141, 100% POWER, C/B D @ 207 210 5.9.9 COMPARISON OF DESIGN BURNUP WITH ACTUAL VALUES. 211 5.9.10 PERCENT DIFFERENCES BETWEEN DESIGN BURNUP AND 212 ACTUAL VALUES l

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l LIST OF FIGURES FIGURE TITLE PAGE NO.

6.5.1 LOAD SWING 30% POWER 226 6.5.2 LOAD SWING 75% POWER 227 6.5.3 LOAD SWING 100% POWER 228 6.7.1 LARGE LOAD REDUCTION 75% TO 25% POWER 233 6.7.2 LARGE LOAD REDUCTION 75% TO 25% POWER 234 6.7.3 LARGE LOAD REDUCTION 100% TO 50% POWER 235 6.7.4 LARGE LOAD REDUCTION 100% TO 50% POWER 236 6.10.1 GENERATOR TRIP FROM 100% POWER 245 6.10.2 GENERATOR TRIP FROM 100% POWER 246 6.11.1 REACT,'" COOLANT CHEMISTRY (02, CL) 250 6.11.2- REACTOR COOLANT CHEMISTRY (FL, Li,. pH) 251 6.11.3 REACTOR COOLANT RADIOCHEMISTRY (I-131, GROSS S-y) 252 i

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LIST OF FIGURES j FIGURE TITLE PAGE NO.

7.1 POWER RANGE DETECTOR CURRENT VERSUS REACTOR POWER 262 7.2 CALORIMETRIC DATA SHEET (EXAMPLE) 263 7.3 CALORIMETRIC CALCULATION (CALCULATOR EXAMPLE PRINTOUT) 264 7.4 LOOP AT VERSUS REACTOR POWER (EXAMPLE) 265 7.5 TURBINE FIRST STAGE PRESSURE VERSUS REACTOR POWER 266 7.6 STEAM GENERATOR PRESSURE VERSUS REACTOR POWER 267

. 7.7 TAVE VERSUS REACTOR POWER 268 7.8' FEEDWATER FLOW VERSUS REACTOR POWER 269 7.9 FEEDWATER VENTURI DIFF PRESSURE VERSUS FEED FLOW 270 1

7.10 STEAM RESTRICTOR DIFF PRESSURE VERSUS FEED FLOW 271 0

LIST OF FIGURES I

FIGURE TITLE PAGE NO.

8.1 XENON FOLLOW TEST 277 B-1 CONNECTIONS FOR 0% POWER FLUX MAP - NORMAL POWER SUPPLY 286 B-2 CONNECTIONS FOR 0% POWER FLUX MAP - BATTERY POWER SUPPLY 287 6

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LIST OF TABLES TABLE TITLE PAGE NO.

1.1 TESTING CHRONOLOGY 18 2.1 INITIAL CORE LOADING COUNT RATE DATA 47 3.4.1 REACTOR COOLANT LOOP FLOW CALCULATIONS 73 - 80 4.1 DELAYED NEUTRON DATA ,

111 4.2 RESULTS OF REACTIVITY COMPUTER CHECKOUT 112 5.1.1 MEASURED AND PREDICTED BORON ENDPOINT CONCEN-TRATIONS 122

] 5.1.2 ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENTS 123 5.1.3

SUMMARY

OF 0% POWER FLUX MAP RESULTS 124 7.1

SUMMARY

OF STATEPOINT DATA 272 7.2 NUCLEAR INSTRUMENTATION OVERLAP DATA 273 l

7.3 ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION

(

SUMMARY

OF RESULTS) 274 7.4 ~ CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION AT POWER (

SUMMARY

OF RESULTS) 275 8.1 -

SUMMARY

OF XENON FOLLOW DATA 278 - 279 l

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1 SECTION

1.0 INTRODUCTION

This report includes a description, analysis and comparison with pre-dicted results of tests and measurements conducted during the initial startup of Salem Generating Station, Unit No. 1. This report covers

the period from August, 1976, through July, 1977.

Salem Unit No. 1 is a four loop pressurized water reactor of 1135 MWe j

(gross) capacity. The nuclear steam supply system was supplied by Westinghouse Electric Corporation, the architect / engineer was Public Service Electric and Gas Company and the constructor was United ,

Engineers and Constructors, Inc.

I The facility operating license was issued on August 13, 1976, and the startup testing program began at 2226 on August 17, 1976, with core loading. Initial criticality was achieved at 1936 on December

) 11, 1976, with the zero power physics testing program starting i immediately thereafter with the search for the point of adding heat.

Following the completion of the zero power physics testing program, the power escalation program began with generator synchronization occurring at 1342 on December 25, 1976. The power escalation program consisted of a combination of physics and plant response tests that were conducted at various power plateaus between 10% and 100% power.

The startup testing program was completed on. July 30, 1977, with the GENERATOR TRIP TEST FROM 100% POWER.

The testing program ran for 348 days, of which 77 days were actually spent testing. TABLE 1.1, Testing Chronology, lists the major testing plateaus, the date the Station Operations Review Committee

! granted permission to perform the required tests at a given testing plateau and the actual time spent testing. FIGURE 1.1, Startup Chronology, outlines the power escalation program indicating those tests that require power reductions, for example, POWER COEFFICIENT MEASUREMENTS and LARGE LOAD REDUCTION TESTS. '

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Figures 1.2.1 and 1.2.2, Startup Chronology II, detail the testing at each power plateau with the respective planned and actual time spent at each significant level of testing. The Figures also provide an overview of the significant causes of delay encountered during testing. Figures 1.3.1 to 1.3.8, Thermal Power History, graphically l illustrates the daily thermal power history for each month of the ;

startup testing program. ,

f CORE LOADING ,

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Core loading began at 2226 on August 17, 1976, when the first fuel assembly 04C was placed in core position' H-15 and was completed at 0456 August 21, 1976, with the positioning of assembly 23C in core ,

location M-14. No major problems were encountered during core

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loading. However, several delays in the core loading sequence were ;

i caused by failures in the fuel transfer system equipment which cumu-latively added twenty hours to core loading. After the completion j of core loading, reactor vessel assembly began. During control rod j drive shaft latching, one control rod assembly, R-19 in core location ,

i H-02, exhibited a greater than allowable withdrawal load. Under the

h direction of the vendor, the red was exc'anged for one in storage.

The new rod was satisfactorily tested and control rod R-19 was returned to the manufacturer for refabrication. No further difficulties were encountered during vessel assembly. ,

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POST CORE LOADING f

I The two month period after core load and prior to initial criticality was primarily spent in correcting deficiencies in the service water system, the auxiliary building ventilation system and the radiation monitoring system. The modifications to the service water system ;

included replacement of all six service water pumps with Unit 2 pumps because of loose bowl couplings caused by galvanic corrosion. The modifications also included installation of a cathodic protection system, replacement of flexible hose with pipe on all five contain-

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ment fan coil units, construction of an enclosure around the service ,

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water traveling screens and installation of electric forced air f' heaters to prevent freezing of the screens. In addition, the nuclear headers were modified to satisfy pressure and flow sta-bility conditions of a STATION BLACKOUT TEST.

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4 Testing of the HEPA and charcoal filters in the auxiliary building ventilation system showed that system flow rates did not meet the requirements of the FSAR and TECHNICAL SPECIFICATIONS. Evaluation 4 of the test resulted in some physical changes and a change to the TECHNICAL SPECIFICATIONS that referenced ANSI N510-1975.

1 In the radiation monitoring system, investigation into continued i

i bearing failures of the containment air sample pump revealed that the t'

containment sample isolation valves were undersized and a different sample pump was required. All eight containment sample isolation 4

valves were replaced and a modified sample pump was installed. Of the 79 days between core load and initial criticality, 17 days were used for actual testing; 6 days were spent at cold conditions (<200*F) performing ROD MECHANISM TIMING TESTS and ROD DROP TESTS and 11 days were spent at normal no load operating temperature and pressure per-y forming ROD DROP TESTS, REACTOR COOLANT SYSTEM FLOW MEASUREMENT TEST,

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REACTOR COOLANT SYSTEM FLOW COASTDOWN TEST, PART LENGTH ROD MECHANISM j BRAKE TEST and the OPERATION OF THE ROD CONTROL SYSTEM.

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l INITIAL CRITICALITY Initial criticality was achieved at 1936 on December 11, 1976, with the zero power physics testing program starting immediately there-i after. The first test performed was the search for the point of add- ,

-8 ing heat. Subsequently, 3x10 amps (on the reactivity computer) was defined as the upper power limit for the zero power physics testing program. The zero power tests included rod warth measurements, boron worth measurements and boron endpoint measurements. Also performed I were a stuck rod worth measurement and measurements of the isothermal

, temperature coefficient. The tests were required to confirm that the values used in the Final Safety Analysis Report accident analysis and the nuclear design report for Unit 1 were accurate.

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10% POWER TESTING PLATEAU ,

l The period from December 19, 1976, to January 6, 1977, saw the first production of electrical power and completion of the tests required s s at 10% power. The usual startup difficulties were encountered since all systems were being operated manually. At this level, startup testing took two dazs during which several> tests were run including a DYNAMIC AUTOMATIC STEAM DUMP TEST, LOSS OF OFF-SITE POWER TEST and SHUTDOWN FROM CUTSIDE THE CONTROL ROOM. Two maintenance outages were required which occupied a total of eight days. One of the outages was used to correct improperly installed steam flow sensing lines and the other, to correct problems with the main steam isola-

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tion valves. ,

'30% POWER TESTING PLATEAU Thirty percent power saw the first tests of the automatic control systems (steam generator level, feed pump speed, rod contro.1) and t the first integrated plant test, LOAD SWING TEST. The aforemen-tiened test revealed a design problem with the unit's power operated atmospheric steam relief valves. Modifications were made to the controllers and the relief valves were successfully rete'ted s at fifty percent power. Significant delays were experienced at thirty percent power.for a variety of reasons. Among the reasons were oil line failures in two of the main steam isolation' valves, faulty steam flow transmitter sensing li,neu and continuous clogging of the con-densate and feed pump suctica " trainers. The problem with the main steam isolation valves ':4? 1e5 rmined to be a design problem and subsequently all main stcani me,lation valve oil lines were modified.

The steam. flow sensing lines again proved to be a design problem and an outage was required to reroute the lines.- The impact on the test-ing schedule of'the continuous clogging of(the. condensate and feed pump suction strainers was significant, delaying the testing schedule .

by at least twelve days. The continuous clogging of the feed pump suction strainers required a modification to the piping containing the strainers so that the piping did not have to be remo to clean

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them. The modification reduced the time to clean pump strainers from eight hours to two hours.

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i 50% POWER TESTING PLATEAU

-s At fifty percent power, there were three major tests, RCCA PSEUDO EJECTION, STATIC RCCA DROP, RODS DROP AND PLANT TRIP. The physics tests were required to confinn design assumptions used in the reactor safety analysis, while the integrated plant test verified that a reactor trip would occur if two rods were dropped into the core.

All tests were satisfactory and approval of the results of the fifty

, percent tests was obtained on March 7, 1977. Following the planned reactor trip, the unit underwent a scheduled maintenance outage.

The length of the outage was based on a critical path required to replace No. 12W main generator hydrogen cooler. The outage was extended to complete work on a hundred thirteen valves that required adjusting or repacking and to repair a crack in a bearing base plate of a containment fan coil unit. The unit was synchronized at 1005, March 19, 1977.

75% POWER TESTING PLATEAU There were five major tests scheduled at seventy-five percent power.

The plant response tests included STEAM GENERATOR MOISTURE CARRYOVER MEASUREMENT, a LOAD SWING TEST and a LARGE LOAD REDUCTION TEST from i seventy-five percent to twenty-five percent power. The physics tests were composed of an INCORE/EXCORE DETECTOR CALIBRATION and a POWER !

I COEFFICIENT MEASUREMENT. The moisture carryover of the steam gen-erators was measured using a radioactive tracer technique employing a sodium isotope (Na-24). All tests were satisfactory and approval of the test results was obtained on Apri.1 5, 1977.

90% POWER TESTING PLATEAU The ninety percent power testing plateau consisted of a flux map to

monitor core performance, collecting Statepoint data to confirm pre-dictions made at seventy-five percent power and a STEAM GENERATOR j MOISTURE CARRYOVER MEASUREMENT. Testing at this level took four days while problems associated with the repair and replacement of i

12 condensate pump motor (which failed because of the failure of a k/ motor cooler) took seven days to rectify and was thus a major delay in the testing program at this level.

100% POWEA YESTING PLATEAU The final tests of the startup program at 100% power were delayed because of various component failures. The first deley experienced was prior to achieving full power and was caused by faulty steam flow transmitters. As power was increased, the transmitters began inter

mittently producing high steam flow signals. The decision was made

therefore to hold at 90% until a design change could be implemented.

1 Full power was finally achieved on May 20, 1977. The next delay was i

caused by the failure of 11 main feed pump which sustained a cracked 1

and chipped impellor when the suction strainer failed. Repairs to the pump required six days. After returning to power, the NSSS ACCEPTANCE TEST was started. Sixty hours into the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months acceptance run, an electrical fault in the containment caused a reactor / turbine trip. Subsequent investigation revealed damage to a reactor coolant l

pump cable, pressurizer heater cable, a pressurizer heater transformer and the reactor coolant pump seal water return line. The damage resulted in a ten day outage for repairs. After returning to power, the remainder of the testing program was completed and the GENERATOR TRIP TEST was run on June 25, 1977. Due to the inadvertent operation

! of a 500 kV breaker fl.ashover relay, the turbine was tripped directly i by the opening of the generator output breakers. The turbine trip then initiated a reactor trip. The operation of the flashover relay caused electrical loss of one of the station power transformers which in turn resulted in loss of power to all the brush recorders which were connected to monitor various plant parameters. The result was that the test had to be rerun. Before the trip test could be rerun, the plant had to shutdown to repair valves MS168, a three way vent valve in the main steam system, and CV75, the pressurizer auxiliary ,

spray isolation valve. The repairs required four days. Finally, four days were required to repair 11 condensate pump. The startup test program finally came to an end on July 30, 1977, with the suc-cessful running of the GENERATOR TRIP TEST.

i This report is being written to comply with Regulatory Guide 10.1,

. Compilation of Reporting Requirements for Persons Subjected to AEC l Ni Regulations.

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TABLE 1.1 TESTING CHRONOLOGY Date Permission Granted Testing Plateau Testing Days 8-17-76 Core Load 5 Cold Pre-Critical 6 Hot Pre-Critical 11 12-08-76 Zero Power 11 12-19-76 10% Power 2 1-06-77 30% Power 7 2-13-77 50% Power 10 3-07-77 75% Power 9 4-05-77 90% Power 4 5-09-77 100% Power 12 TOTAL 77 O

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SECTION 2.0 INITIAL CORE LOADING FUEL DELIVERY During the period from October 14, 1975, to January 28, 1976, nuclear fuel was delivered to the Salem site in the form of 17 x 17 array fuel assemblies. The fuel assemblies and inserts were inspected.and placed in dry storage in the Unit 1 spent fuel pit. During the on-site inspection, it was discovered that some of the assemblies had reddish brown deposits near the upper and lower assembly nozzles. The cause of the deposits was attributed to improper cleaning following fuel assembly fabrication. A Westinghouse representative performed an on-site cleaning of the affected assembl'ies. This item was reviewed by the Westinghouse Nuclear Fuels Division and determined not to pose a problem.

Prior to removing each fuel assembly from its shipping container, a smear survey was conducted on the assembly in order to detect any

,,^

surface contamination of the fuel. The survey results showed all 2

contamination levels to be within Zone II limits (c <50 dpm/100 cm ,

2 B, y <1000 dpm/100 cm ),

During inspection of the fuel assembly inserts (part length rod cluster control assemblies, full length rod cluster control assem-blies, burnable poison rods, primary and secondary source assemblies),

it was discovered that five assemblies were shipped with incorrect inserts. These inserts were removed and placed in the proper assem-blies. The thimble plugging devices were being fabricated during the period of fuel shipment and therefore those assemblies that required a plugging device were shipped without an insert. The plug-ging devices were subsequently shipped and inserted in the respective assemblies. .

On January 28, 1976, the final truckload of fuel arrived on-site bringing the total delivery to 193 assemblies (88.792 MTU which included 2.47x10 6

grams of U235), )

-19' i

t i

- On August 10, 1976, two primary source rods were welded to their respective primary source assembly in preparation for initial core k loading. The source material was Californium 252 (Cf ' ) which had a spontaneous fission half life of 66 years. Each source rod con-I 52 which was expected to yield tained approximately .11 Curies of Cf 4.63x10 8 neutrons /see to aid in core reactivity monitoring during-initial core loading and subsequent criticality. The welding operation was performed without incident and with a minimum amount 1

j of radiation exposure to personnel involved in the operation.

On August 14, 1976, Facility Operating License No. DPR-70 was issued which permitted initial core loading and power operation up to 33.38 ;

i megawatts thermal for a period not to exceed 300 megawatts days of L integrated core burnup. An amendment to this license to permit i

! higher power operation would be contingent upon the outcome of

! Natural Resources Defense Council versus NRC, a court case dealing 1 i

with the environmental impact of the nuclear fuel cycle for a i

,~

generic light water reactor. l I Prior to initial core loading, a design calculation was performed l to ensure that under any core configuration the effective neutron

multiplication would be within the TECHNICAL SPECIFICATION limit.

The results of this calculation are presented in Figure 2.1 along j with the TECHNICAL SPECIFICATION limit. It is seen that the expected multiplication is well within the limit for each configuration.

Once the values of Reff for the various core configurations were l determined, a calculation was performed to determine the expected behavior of the inverse count rate ratio during core loading. The i following equation was employed to define the expected inverse count 1

rate ratio:

i .

I i

k=1-Keff 1 l I

1 i The values of g for the various configurations analyzed are pre-sented on Figure 2.2.

f; .

! ~

l L .. = - ---. -

i

\

i I

CORE LOADING Prior to the installation of the first assembly, reactor vessel con-ditions were established in conformance with the requirements of the INITIAL CORE LOADING PROCEDURE, SUP 60.0. Coolant circulation in -

the reactor vessel was established via the residual heat removal system by taking suction from the hot leg of primary loop 11 and discharging to the cold legs of all four primary loops. After veri-fying that no stratification existed within the reactor vessel, and that the boron concentration in the vessel was greater than the required 2000 ppm, initial core loading commenced. The RHR loop water was sampled and analyzed for boron concentration at regular intervals during the core loading. The average boron concentration was 2088 ppm, with a range of 2067 ppm to 2112 ppm (in the measure-ment data) .

Core loading operations began at 2226 on August 17, 1976, with the installation of fuel assembly 04C, containing a primary source cluster, into core position H-15. The detailed sequence of'the loading operation is shown schematically in Figures 2.4 through 2.12. l Figure 2.3 is a key sheet for interpretation of the symbolism em-ployed on the core loading schematics.

Operations were conducted during three 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months shifts per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period for the duration of the core loading operation. Total elapsed time was 78.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , with an average loading rate of 2.5 assemblies per hour. Major delays encountered during the loading operation were i

due to fuel transfer system failures (air motor, upender limit switches, selsyn gear, etc.). Factoring out delays due to repairs on the fuel transfer system, the core loading, tine is estimated at 58.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , corresponding to an average loading rate of 3.3 assem-blies per hour.

Shortly after the beginning of core loading, the LED failed on the scaler for temporary detector A. A replacement scaler / timer was obtained and installed, and counting on Channel A resumed at step 106 l of the core loading sequence. Temporary loss of this channel did I

~

\

not delay the loading process.

l l

i

- . . - .-. . . = - -. - . - - - -. - . _ --

t Noise spikes, traced to a power cable connector, were observed on the temporary channels. These spikes required occasional repeat i

)

{

count rate determinations, but were not responsible for core loading delays. The connector was eventually replaced. I Subsequent to the completion of core loading, final count rates of 3.8 cps and 3.4 cps were observed on plant channels 31 and 32, res-y pectively. The minimum count rates allowed by FSAR are 2 cps, f attributable to core neutrons. Core loading was completed at 0456 l on August 21, 1976. Following completion of core loading, a visual jl inspection was made of the assemblies in the core to verify the j proper core loading. The visual inspection was permanently recorded via video tape. Figure 2.19 depicts the as loaded core including insert locations and enrichments. Figure 2.20 illustrates the incore j and excore instrument locations.

i TEST EQUIPMENT / DATA

{ Three temporary neutron sensitive detectors (BF3 ion chambers) were l installed in the reactor vessel, just prior to the initiation of the f loading operation, in the locations specified by the core loading

! sequence. Count rate data for N31 and N32, the two plant-source ,

f, range channels, were obtained in the Control Room and the three j temporary detectors data were obtained at the reactivity station inside containment. Data was recorded at both places with communi-cation maintained between the two stations. Prior to introducing i the first assembly (containing one of the primary sources) into the vessel, the average background was taken on all five detectors.

Average background for all channels was a few hundredths of a count per second; several orders of magnitude less than the observed count I rate during the core loading operation. ;

Following the insertion of a fuel assembly into the vessel, the count rate was determined from the average of three 100 second counting intervals for the temporary detectors and three 100 second counting

. intervals for the plant source range channels. After subtracting

e. m

1

the appropriate background, the inverse count rate ratio (ICRR) was computed for each channel and a current plot of ICRR versus the number of assemblies loaded was maintained in the containment. The resultant plots ar presented on Figures 2.13 through 2.17 along with a tabulation of the count rate data in Table 2.1. At no time were loading operations interrupted as a result of eigher high count

! rates or unexpected changes in the ICRR.

l 1

Boron concentration measurements are shown as a function of time j during core loading in Figure 2.18. In an attempt to establish a i

possible time dependence of vessel boron concentration, a linear function was least squares fit to the bo,ron measurements. The resultant slope was essentially zero, suggesting that the boron concentration was unvarying over the duration of core loading.

DETECTOR RESPONSE Count rate changes, in response to fuel assembly additions, were as expected on all five source range detectors. Superimposed on this kas a gradual degradation of count rate (incrasing ICRR)

beginning at about core loading step 120 and terminating.at about step 170. This phenomenon was observed on all responding detectors.

Note that temporary channel A behavior (between steps 120 and 150) was heavily influenced by nearby fuel assembly additions, and there-fore, does not meet the definition of a responding detector. The behavior of the remaining channels is most likely attributable to neutron density changes in the vicinity of the detectors. A second possibility, although less likely, is a change in containment tem-perature and/or humidity that simultaneously, and in a similar way, effects the electronics of the several channel's.

Changes in neutron density are attributable to variations in temper-ature, boron concentration, and/or void formation. As previously observed, indiations are that there was no net change in vessel boron concentration over the core loading operation. However, because of data scatter, the time dependence of boron over the duration of the count rate degradation is not well defined. First observation would i

l lead to the conclusion of a decreasing boron concentration; a reac-

tivity addition rather than the observed loss. Unfortunately, t

l meaningful temperature data was not obtained during core loading.

I Void formation is mentioned as a possibility because of the large number of air bubbles continuously present in the vessel water during core loading. As assemblies are added to the vessel, flow is pre-ferentially channeled to vacant regions in the vessel, giving rise to the possibility of increased void formation in those regions.

VESSEL ASSEMBLY l

l After completion of fuel loading, the upper internals assembly,

! including control rod drive shafts, was placed in the reactor vessel.

I Latching of the drive shafts to the control rods then began. As i

part of the latching procedure, after a drive shaft is coupled to 4 .

its respective control rod, the entire assembly is withdrawn about j eight feet. This is done to ensure that each drive shaft is prop-

erly coupled and that a clear path exists for control rod motion.

l To do this, a spring scale is placed between the manipulato'r crane l' auxiliary hoist hook and the control rod drive latching tool. During each w'ithdrawal and insertion, the load indicated on the spring scale is monitored to ensure it remains between specified limits. It was

! during this procedural step that a higher than expected loading was

! indicated for control rod R,-19 in core location H-02. The loading l 1 was about sixty-five pounds greater than the maximum allowable load-j

ing.

i The vendor was contacted and it was determined that control rod R-19 was the most probable suspect component. Personnel were instructed to remove the upper portion of the guide tube, assembly and carefully i extract the control rod. A maximum overload restriction of one l hundred thirty pounds was imposed on the withdrawal. However, the RCCA was removed uneventfully and a spare control rod R-03 was I inserted. The new control rod was latched and the assembly with-

)'

drawn uneventfully. The upper portion of the guide tube assembly was j installed and vessel assembly continued. R-19, the control rod which q ...

was removed from the core, was inspected on site and scoring was l noted on several of the rodlets. It was returned to the manufacturer

! for refabrication.

Figure 2.1 C

l C I K EFF vs. NO. OF ASSEMBLIES

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4 Assembly loaded in permanent position Assembly loaded in temporary position N Assembly loaded into position during loading Step N 3

Location of temporary detector A (similarly for B and C)

)

. . 8 Assembly with primary source insert. (Cf 9 4 X 10 n/s)

M Assembly with secondary source insert

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Figure 2J Key Sheet for Core Loading Schematics

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c astee (c t f.ec.) scan Det. A (fe.parary) Det. 8 (Temporary) Det. C (fenporary) Plant Channel st-31 Plant Channel stE-31 l**,. ",Aege coge Poe.

Value Ref. Val. Value tel. Val. Value Ref. Val. Value Ref. Val. Value Ref. Val.

E-3I St-32 j * . B. .

J Als/s 55 W8 o455 . -

3sao 331.2 i 3.o 7 lo.Bo 4. lf, 3ers - - -

o A 4fo o.825 oso; ..

I sc. se esoa 548.n - 82.9 4 -

4.o o -

o.949 o.ss4 a.q44 i.

'7 FB 0 5 2*t. 348.B

13.5 6 RA4

oA49 o.79 6 oASA 58 E'8 05 5(m 3 49.7 -

12.71 4.o8 "

a.947 o.8 49 a .92.5 si L9 o554 3 5 o.1 ~ 1 A.53 "

4.o t. '

o.94(o o.797 o.93o so K9 o6so 3 60.7

13.42 '

3.9 I oA44 o.so4 o.9 c,5

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13.51 -

5.7 o -

o.945 o.199 i .o2.o i 61 14 9 06ES 348.8 -

13.1 7 3.93

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- 8 e 63 69 o ti4 . 349.1 -

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oA4a o.so6 o.441 '

64 F9 o 74 3 ! 348.4

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4.~3 6

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1 A.l3 - 4.02. -

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13 59 - 4.c o '

oA44 o.19 4 oA46 .

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s 363.5 a 14.14

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72 h sis 9 35o.1 -

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73 ii' 12.1 5 348.6

4J.52 3.afo " 8.c o s o . 3'lo I. ct3 $

74 8 12.3 5 .36o.2 =

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, _ o I.mo o.sil 1.o27 7 76 il 13o 4 '

l350A .". 90.4o " 4.o 3 " 0898 o 2.o4 oA80 .u

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Table 2.1(cont'd) i

, .Ceunt Rates (Counts /sec.) ICSR

! Det. A (Temporary) Det. 8 (temporary) Nt. C (Tesiposary) Plant Channel Ist-31 Plant Channel IIE-3J N' "" Asss'Cere

.a. res. tsee ,,,,, ,,,, ,,,, ,,; ,, ,,,, y,,, , , , , ,,,, ,,,, , , , , ,,,, ,,,, ,,,,, ,,,, ,,,,

a a c u-H u.32 Blen ist. 9 sk 2 i l 2. -

T 356n 398.4. iL.li 1.84 2.84 3AO V -

8.o01 0.13 6 i.o t t -

i<.

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I8.32 =

4.13 * .qqa o isq a.q 3 3

fas is 2:41 3s 1.1 -

i9A1 = 4. i r - a qq t., o. i4 t o q4.r l

18 1 $ 215 5 349.o e- 21.e4

2A4

i.ood a .12 4 i.oo l

j. . ses ,Ch 2,21o 35o.4 = 11.50 a 3A 2. a n.c oo 0.81t 1.oo6 i, - see i5 2217 J 50.9 -

21.s.G -

8.cs 1

aqqs o.i31 a.cA 3 Y ik 2244 349.1 -

21.94

4.01 *

1. o o 3 o.9 o A s-f, ( .

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7 1 233a 9 t.. (.81 96.69 3aco M=s o ' 47o.cl 470A 4.o9 4 o9 3.48 7.48- 1. c,c o e. coo s.ooo i. coo i. c o o I

.5/rohg, let. [*o oc o8 97.75

352.0 a 478 9 '*

4.13 a

4. o9

o.q sq er,9 4 oAq7 oqqo i.oit a s1 D7 o o 2.2. 97.49 =

350.1

4 13.1 -

4.21 a 4.o c, a oscl e o.qqq oaqs o s,i s.o41

. Ice 03 a o 2. I 95.99 -

3so.l 471.3

4.o 3 " h 3.35 -

1.co 7 o.c497 oAc4 s 1.oi4 s.o 4o

! ~ se9 09 ooS2 95.9o "

' r.l . l "

471.6

4.1 A -

3.43 "

r. co n oA66 a .ci9 8 oA G6 s. o 2.1 no cI ono r 9(.28 =

34IA

472.1 -

4.o 4 -

3.4 8 **

i . oc 4 1. c o s n.o i7 oto s.ooo ses c8 o222- 96.2.8 p 147A "

471/2 ~

4.oS

3.4 a -

t.co t 8.co S o .qq 7 f.oto I. coo

- ni c9 ona1 qcA l a 3484 =

471.5 a SAS a 3.43 a o.99 I s.co4 a.44A 1.o34 1.o17 j

4

" h3 B1 o152. 44.11 a 349A -

411.7 a 4.o fa a 3.49 a I.cos I.c o o oA98 1.n o 7 oA98

~

se4 8A d2c6 46.30

~

347.7 a 470.t. a 4.o9 a 3.Po a r.co3 i.cos 1. coo I.c o o dA+1 its 09 olis 96.3 1 a 351.1 -

4 69.l a 3.9 5 -

3.4(. -

1. oo 3 o A% i.co2, t.o34 s.oo r

. 18 . A7 o2.34 9(e.84 a 344.4

4 11.8 "

9A1 -

3.6 ) * .:.;o S 1. col o.99 8 s. o 2.8 0947.

"2 A9 0 24t's 96. Slo 251o * .8 6ff . l a 4.03 a f.4 6 a f.o o l o.:A '1 1 co2. i .ra i3 s.or t "l

l h A8 o2.63 2.38 2.33 - - - - " -

i . coo - - - -

8-o tre to 032.1 231 ~ 3so.ct __._

469.9 a 4.19 a 3.5 % "

Ano_s, oA9-) 1.o04- o97S o992. y O

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sei io s334 2.5[_, ~ _ , , 360.o -

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1. 531, c.394 1.001 o.991 .o63 7 i

fic N 034A 2.s o "

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4. 2.0 " 3.44 "

of-tS311. col I . c o 2. c.ce 12 s . o t 2, o l

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c % m s u ~4 0 o $ t w l w in ss y +N s %~ e y

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SECTION 3.0 PRECRITICAL TESTING This section deals with those tests that were conducted after initial core loading but prior to initial criticality under both cold shutdown and hot standby conditions.

3.1 SUP 70.2 - RTD BYPASS FLOW VERIFICATION The purposes of this startup procedure were to calculate the flow rate necessary to achieve the design reactor coolant transport time for each RTD, to mea,sure the flow rate to each RTD, to ensure that the coolant transport time met the design value and to. verify the low flow alarm setpoint.

Due to the increased sensitivity of the type of RTD's used at

' Salem, the required transport time was increased from 0.5 seconds to 1.0 seconds. The following tabulation shows that the minimum

- flow rates required to meet the 1.0 second transport time were exceeded.

LOOP MINIMUM REQUIRED FLOW ACTUAL' BYPASS FLOW llT C

25.5 GPM 196 GPM llT '

H 12T 5. GPM 185 GPM C

12T g 72.5 GPM 122.8 GPM 13T . GPM 201.7 GPM C

13T 9. 08.2 GPM H

14TC * '

14T .0 GPM 115.9 GPM H ,

All RTD low flow alarm setpoints were set and checked to trip within + 3 GPM of 90% of the total measured loop RTD bypass flow rate.

(__ _/ :

-53'-

i l

t

~% 3.2 SUP 70. 3 - REACTOR COOLANT FLOW COASTDOWN MEASUREMENTS l

i The purpose of this test was to measure the rate at which reactor ,

coolant flow changes, due to various combinations of reactor i coolant pump trips. In addition the test compared the coastdowns to that assumed in the Final Safety Analysis Report. Four dif-ferent coastdowns were run and analyzed:

l 4

l a) Four pumps operating, four pumps coasting down

)

b) Four pumps operating, two pumps coasting down c) Three pumps operating, three pumps coasting down d d) Three pumps operating, two pumps coasting down

! 2/4 Coastdown For this coastdown, besides verifying the validity of the

safety analysis, three other quantities were calculate,d

.a) Low flow time delay b) Undervoltage trip delay !

c) Underfrequency trip delay. ,

4 The low flow time delay is defined as the time from the cpening j of the second reactor ccolant pump breaker to the time of the first motion of the RPI signal plus the' time from the opening of 1 a reactor coolant pump breaker until the time all three low flow relays in that loop have tripped. The acceptance criteria called for the low flow time delay to be less than or equal to 1.90 seconds. The actual time was 1.72 seconds.

j The undervoltage trip delay was defined as the delay measured in the REACTOR PROTECTION TIME RESPONSE MEASUREMENT TEST plus j the time from the opening of the second reactor coolant pump breaker to the time of the first motion of the RPI. signal. The i ,

! acceptance criteria was that the undervoltage trip delay be less l J

(_-)- than or equal to 1.20 seconds. The actual time was 0.136 seconds.

1 I

i

- -. . - _ . _ _ _ . ~ . . , .. _ . . _ . . _ . _ _ . _ . . . . . . _ _ , _ . . ~ . _ ~ . . _ . , . . _ . _ - . - , _ . -.__._

The underfrequency trip delay was defined as the delay measured

s. in the REACTOR PROTECTION TIME RESPONSE MEASUREMENT TEST plus the time from the opening of the second reactor coolant pump breaker to the first motion of the RPI signal. The criteria was that the underfrequency trip delay be less than or equal to 0.60 seconds. The actual time was 0.239 seconds.

When the actual flow, corrected for flow sensor delay, was compared to the flow assumed in the Final Safety Analysis Report, the actual f)- 1 more slowly (Figure 3.2.1) . Flow sensor delay is defix.,t la the time at which the best straight line approximation to the inverse flow curve drawn in the 4/4 coast-down, between three and ten seconds, intersects the inverse flow value of 1.0.

4/4 Coastdown For this coastdown, two analyses were made. First, the actual flow (corrected for flow sensor delay) was compared to the flow assumed in the Final Safety Analysis Report. The actual flow always exceeded the assumed flow in the safety analysis (Figure 3.2.2). Second, the rate at which actual flow changed (between three and ten seconds) was compared to the rate at which flow changed in the safety analysis. The analysis showed that actual flow decayed at a slower rate than predicted.

2/3 Coastdown In this case, the actual flow (corrected for flow sensor delay) was compared to the flow assumed in the Final Safety Analysis Report. The actual flow always equaled or exceeded the flow predicted in the safety analysis (Figure 3.2.3).

3/3 Coastdown In this case, the actual flow (corrected for flow sensor delay)

(s was compared to the flow assumed in the Final Safety Analysis

-ss-i

i rReport. The actual flow always equaled or exceeded the flow predicted in the safety analysis (Figure 3.2.4).

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3.3 SUP 70.4 - PRESSURIZER SPRAY AND HEATER CAPABILITY There were three objectives to this test: first, to determine the effectiveness of the pressurizer heaters, second, to estab-lish the continuous spray bypass flow rate by adjustment of the pressurizer spray bypass valves 1PS2 and lPS4, and finally, to ;

determine the effectiveness of the pressurizer spray system.

Continuous Spray Bypass Flow Two requirements had to be met in setting the continuous spray bypass flow. The AT between the pressurizer and the spray line had to be less than 200'F and the spray line temperature had to be kept above the low teruperature alarm of 500*F. The spray by-pass valves 1PS2 and 1PS4 were adjusted in definite increments (1/4 turn, 1/2 turn) while the various temperatures were monitored.

No problems were encountered in setting 1PS4. However, due to a brok,en valve stem on 1PS2, final adjustment will have to await repair during the first major outage which requires primary

([.

system depressurization since lPS2 is unisolable.

Pressurizer Spray Effectiveness In order to compare the effectiveness of the pressurizer spray to the design curves, several plant parameters including pres-surizer pressure were instrumented using a brush recorder. In addition, all pressurizer heaters were put in manual and turned off. The pressurizer spray valves were then placed in manual i and fully opened until the pressurizer pressure decreased to approximately 2000 psig at which time all pressure control systems were returned to normal. The spray valves caused a pressure decrease of about 260 psi in two minutes which was well within the specified limits. The results of this test are

! shown in Figure 3.3.1. The spray valves have operated satis-t factorily during the entire power range testing program with i the exception of slight leakage through both spray valves.

I l

, 1

N Pressurizer Heater Effectiveness To verify the design of the pressurizer heaters, the procedure called for instrumenting various plant parameters including pressurizer pressure, placing the power operated relief valves in manual, energizing all the heaters ~and allowing' press'ure to increase about 50 psi before returning all systems to normL1.

The first time the test was run, the control heaters were left in automatic r.nd cycled off as the pressure was raised. This lowered the rate of pressure increase and the acceptance cri- ;

teria was not met. Analysis of the result indicated that in order to meet the acceptance criteria, all heaters would have to remain energized throughout,the imposed transient. During the second test, pressure control was put in manual and a maxi-mum demand signal was input. . System pressure increased 50 psi in about 3.3 minutes which followed the nominal ussponse and was well within the acceptable limits. The results of this test m are shown in Figure 3.2.2.

\

During the power escalation program, it was observed that the control heaters were not able to maintain plant pre'ssure without utilizing the small backVp group. Further testing traced the problem to leaking cpray valves, 1FS'l and-lPS3. These valves will be overhauled dufin'g the first major outage.

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-s 3.4 SUP 70.5 - REACTOR COOLANT SYSTEM FLOW MEASUREMENT The purpose of this test was to obtain data on reactor coolant pump input power and reactor coolant loop elbow tap differential pressure in order to calculate the actual reactor coolant system flow rate for comparison with the design flow rate as specified ,

in the TECHNICAL SPECIFICATIONS.

The reactor coolant flow measurement test was run over a period of two days. The procedure called for instrumenting a loop for RCP speed, elbow tap AP, reactor coplant loop temperature and reactor coolant pump power. With the loop instrumented, the ,

pumps were run in various combinations, from four pumps running to one pump running. Data was recorded at one minute intervals for ten minutes on each pump combination. The same procedure was used to collect data on the other three loops. No problems were encountered in data collection.

Once the data had been collected, the reactor coolant system loop ,

flow could be calculated. The procedure involved mathematically generating a pump performance curve for comparison with a pump.

curve supplied by the pump manufacturer. The manufacturer's curve was based on data obtained in a test loop using the plant impellors. From the raw data, the average of various parameters for all combinations of pumps was determined. Then for each loop with a specific pump configuration, the specific gravity was i calculated. The specific gravity was then used to correct the I measured pump power, normalizing it to the conditions under ,

)

! which the manufacturer's curve was determined. Next, the rela-tive coolant flow for each loop, under various pump configurations, was calculated using the following definition: -

6P - APo . S.G. !

g, (AP (4) - APo S.G. (4)) r j '

where: .

f - Relative coolant flo;<

AP - The diffarential prescure for a specific pump config-uration AP(4) - The differ &ntial pressure fer a specific 16cp when all pumps are operating APo - The differential pressure for a specific loop when no pumps a,te operating '

S.G. (4) - The specific gravity in a given loop when all ptops -

are operating ,

e S.G. - The specifig gravity for a given pun @ configuration.

The base case for the r61ative coolant ficw was the four pump operating condition. I'f the flow in this configuration was set equal to one, then as pmaps were secured, the flow in the operating ,

loops increased because of back flow through the idle '1. cops. With .

one pump operating, the ficw through that loop was in general ten ,

percent higher than the flow through the same . loop when all four pumps were operating. With two pur.p operating, the- iacrease in flow per locp was aheut eight percent. With three pumps operating, icop flow increased five percent over the four pump condition.

For each loop, a flow for the four pump.conficuration was assumed and using the relative caslant flow previously determined f the flow for the three, two and one pump configuration was calculated. I Using the assumed flow for the four, three, two and one pump

condition, the power required for each combination sss determined '

from the pump curves suppl:194 by the .nanufacturer. The difference in power between the power measured in the fccr pump configuration and the power determined from the mancfact'urer's curve was*calcu-lated.

i <

r This difference in power was added to the power used in the three, l

(-)

1 two, and one pump configuration as read on the manufactuer's curve. j l

i

This predicted power was then subtracted from the power actually :

) measured in the plant and the difference recorded as power error [

(the error could have been either negative or positive). The j errors for the three, two, and one pump configurations were i

added together and plotted versus the assumed flow. Three j i other flows were assumed for the four pump configuration. For {

each flow, the same error analysis was completed. When the four

points had been plotted, they were connected by a straight line and the point where the straight line intersected the flow axis ;

was the loop flow for the given loop with four pumps running. f r

! l Table 3.4.1, Reactor Coolant Loop Flow Calculations, and Figures l

3.4.1 to 3.4.4, Loop Error Analysis, show the type of analysis ,

that was performed. Design flow was 88,500 GPM/ LOOP, the cal-culated flow was: l t

1 I LOOP 11 99,250 GPM p LOOP 12 97,400 GPM [

! LOOP 13 100,000 GPM 4

LOOP 14 99,050 GPM

. 4 Historically, the results of the method just discussed, which

was used to calculate the reactor coolant system flow rate prior 1 i

to power operation, have been shown to be,6% to 8% high. A .

more accurate method, which can only be used during at power ;

,onditions, was used during the startup test program to confirm the results of the pre-critical flow measurements.

The at power procedure for the reactor coolant system flow measurement used an energy balance across the steam generators as its' basis. Over the course of an hour, data was collected for

eight calorimetrics and simultaneously eight sets of loop T H !

and T C temperatures were obtained from the protection racks. !

The calorimetrics for each steam generator were completed and l the change in enthalpy across each loop using the hot and cold I 1

-67'-

. . . I

leg temperatures previously recorded were calculated. Allow-m, ances were made in each calorimetric for losses due to reactor

~

coolant pump heat and steam generator blowdown. In order to calculate the reactor coolant loop flow, the following rela-tionship was used:

~

! ~

FLOW (GPM) =

(60) ( AH) -

where:

S/G - Steam generator output in 10' BTU /HR LOSSES - Reactor coolant pump heat.plus blowdown 7.48 - Gallons per cubic foot V - Specific volume at loop T }

C 60 - Minutes per hour AH - Enthalpy change across a loop (BTU /LBM) ,

, Calculations of reactor coolant pump flow were made throughout the power escalation program. The results of the calculations performed at 100% power are tabulated below:

LOOP 11 LOOP 12 LOOP 13 LOOP 14

1. 94,710 96,031 93,789 95,553 i

2. 93,622 96,382 92,328 94,760

3. 92,738 95,833 93,577 95,839

4. 93,068 95,445 92,933 95,284

5. 92,846 94,427 92,744 95,121 l

l '

6. 92,678 94,562 93,192 93,033 !

7. 93,856 95,358 92,760 94,960

8. 93,304 94,914

94,302

)

93,353 95,369 93,046 95,232 j

. O v

Failed temperature indicator, flow calculation not possible.

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ML20210T484

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