ML20040E792

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Startup Test Rept,Part 2.
ML20040E792
Person / Time
Site: Salem PSEG icon.png
Issue date: 01/29/1982
From: Schell W
Public Service Enterprise Group
To:
Shared Package
ML18086B287 List:
References
NUDOCS 8202050377
Download: ML20040E792 (198)


Text

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION UNIT NO. 2 STARTUP REPORT BY WILLIAM H. SCHELL I

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APPROVED:

REACTOR ENGINEER: M ,,

  1. ' 82-F. SCHNARR (DATE) l TECHNICAL MANAGER-SALEM L.

'((* !I1/h-K. MILLER (DATE)

I GENERAL MANAGER -

lI SALEM OPERATIONS: M<

H. M'lDUR'A w '4'd //d7[F (DATE)

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8202050377 820202

PDR ADOCK 05000311 l P PDR

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ACKNOWLEDGEMENT The success of the startup program depended upon many people.

The Reactor Engineer and his staff express their gratitude to the entire Salem Generating Station Staff. Special tenks to Fred Twogood, Ed Watj en , Al Ha ye s , Fred Baskerville and Lou Grubmeyer of Westinghouse for their contributions. Thanks to George Druf fner of the Energy Laboratory and to John Dal Pan

for his assistance in compiling this report.

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109D (1 ) : 28 I

i TABLE OF CONTENTS l PAGE NO.

LIST OF FIGURES iii - x LIST OF TABLES xi

1.0 INTRODUCTION

1 2.0 INTEGRATED TESTING 19

2. 1 SUP 81.5 -

DYNAMIC AUTOMATIC STEAM DUMP CONTROL 19

2. 2 SUP 81.3 -

TURBINE OVERSPEED TRIP TEST 23

2. 3 SUP 82.6 -

LOSS OF OFF-SITE POWER 26

2. 4 SUP 82.5 -

SHUTDOWN FROM OUTSIDE OF THE CONTROL ROOM 36

2. 5 SUP 82.1 -

LOAD SWING TESTS 51

2. 6 SUP 82.4 -

RODS DROP AND PLANT TRIP 57 <

2. 7 SUP 82.2 -

LARGE REDUCTION TEST 61

2. 8 SUP 82.7 -

STEAM GENERATOR MOISTURE CARRYOVER MEASUREMENT 69

2. 9 SUP 82.8 -

NSSS ACCEPTANCE TEST 74 2.10 SUP 82.9 - GENERATOR TRIP FROM 100% POWER 76 2.11 SUP 90.9 -

BORON MIXING AND COOLDOWN 80 2.12 RADIATION SHIELDING EVALUATION, EFFLUENT MONITORING, CHEMISTRY TESTS 85 2.12.1 SUP 81.13 - RADIATION MONITORING AND SHIELDING EVALUATION 85 2.12.2 SUP 81.14 - EFFLUENT MONITORING SYSTEMS 85 2.12.3 SUP 81.15 - CHEMISTRY AND RADIOCHEMISTRY TESTS 86 2.13 SUP 81.4 -

AUTOMATIC STEAM GENERATOR LEVEL CONTROL 89 2.14 SUP 81.6 -

AUTOMATIC REACTOR CONTROL 91 2.15' SUP 80.1 -

APPENDIX 6, FEEDWATER HAMMER TEST 94 I

109D (1 ) : 29 i

I TABLE OF CONTENTS PAGE NO.

3.0 CALIBRATION OF TEMPERATURS AND FLOW INSTRUMENTATION DURING POWER ESCALATION 98 I 3.1 SUP 81.12A - ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION 99

. 3.2 SUP 81.12B - STATEPOINT DATA 105 3.3 SUP 80.7 - TURBINE CONTROL SYSTEM CHECKOUT

. AND STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM 113 3.4 SUP 81.7 - CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION AT POWER 118 4.0 PHYSICS TESTING 146 4.1 POWER AND BURN DISTRIBUTION MEASUREMENTS 146 4.2 SUP 81.9 - RCCA PSEUDO EJECTION AND RCCA ABOVE BANK MEASUREMENT 153 4.3 SUP 81.10 - STATIC RCCA DROP AND RCCA BELOW BANK POSITION MEASUREMENTS 159 4.4 SUP 81.8 - POWER COEFFICIENT AND INTEGRAL POWER DEFECT MEASUREMENT 167 4.5 SUP 81.11 - fNCORE - EXCORE DETECTOR FLUX DIFFERENCE CALIBRATIONS 172 4.6 SUP 81.12C - INTERMEDIATE AND POWER RANGE CHANNEL HIGH VOLTAGE SETTING VERIFICATION 182

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109D:(1):30 1

I LIST OF FIGURES FIGURE TITLE PAGE NO.

1.1 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR MAY '81 5 1.2 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR JUNE '81 6 I 1.3 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR JULY '81 7 I 1.4 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR AUGUST '81 8 SALEM UNIT 2 CYCLE 1 I

1.5 HOURLY POWER FOR SEPTEMBER '81 9 1.6 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR OCTOBER '81 10 1.7 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR NOVEMBER '81 11 1.8 SALEM UNIT 2 CYCLE 1 HOURLY POWER FOR DECEMB'ER '81 12 1.9 CORE LOAD THRU HOT FUNCTIONAL TESTING - PLANNED / ACTUAL 13 1.10 LOW POWER THRU 10% PLATEAU TESTING - PLANNED / ACTUAL 14 1.11 30% TESTING - PLANNED / ACTUAL 15 1.12 50% TESTING - PLANNED / ACTUAL 16 1.13 75% THRU 90% TESTING -PLANNED / ACTUAL 17 1.14 100% TESTING - PLANNED / ACTUAL 18 I

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I LIST OF FIGURES FIGURE TITLE PAGE NO.

2.3.1 BLACKOUT SUP 82. 6 RCS LOOPS HOT / COLD LEG. TEM PERATURE 27 2.3.2 B LACKOUT SUP 82. 6 PRESSURIZER PRESSURE, PRESSURIZER LEVEL AND HOTTEST INCORE 28 2.3.3 LOSS OF OFF-SITE POWER NO. 24 STEAM GENERATOR 29 2.3.4 LOSS OF OFF-SITE POWER NO. 23 STEAM GENERATOR 30 2.3.5 LOSS OF OFF-SITE POWER NO. 22 STEAM GENERATOR 31 I 2.3.6 LOSS OF OFF-SITE POWER NO. 21 STEAM GENERATOR 32 2.3.7 LOSS OF OFF-SITE POWER

'I T,y - AUCTIONED / REFERENCE / LOOP 33 LOSS OF OFF-SITE POWER I

2.3.8 PRESSURIZER PRESSURE 34 2.3.9 LOSS OF OFF-SITE POWER PRESSURIZER LEVEL AND LEVEL SETPOINT 35 2.4.1 SHUTDOWN FROM OUTSIDE CONTROL ROOM RCS LOOPS HOT / COLD LEG TEMPERATURE 40 2.4.2 SHUTDOWN FROM OUTSIDE CONTROL ROOM RCS LOOP HOT / COLD LEG TEMPERATURE (CONTINUATION) 41 2.4.3 SHUTDOWN FROM OUTSIDE CONTROL ROOM PR2SSURIZER PRESSURE, PRESSURIZER LEVEL AND HCTTEST INCORE 42 2.4.4 SHUTDOWN FROM OUTSIDE CONTROL ROOM I PRESSURIZER PRESSURE, PRESSURIZER LEVEL AND HOTTEST INCORE (CONTINUATION) 43 I 2.4.5 SHUTDOWN FROM OUTSIDE CONTROL ROOM PRESSURIZER LEVEL AND LEVEL SETPOINT

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109D (1 ) : 32 ,

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I LIST OF FIGURES FIGURE TITLE PAGE NO.

2.4. 6 SHUTDOWN FROM OUTSIDE CONTROL ROOM PRESSURIZER PRESSURE 45 2.4. 7 SHUTDOWN FROM OUTSIDE CONTROL ROOM T avg - AUCTIONED / REFERENCE / LOOP 46 2.4. 8 SHUTDOWN FROM OUTSIDE CONTROL ROOM NO. 21 STEAM GENERATOR 47 2.4. 9 SHUTDOWN FROM OUTSIDE CONTROL ROOM NO. 22 STEAM GENERATOR 48 I 2.4.10 SHUTDOWN FROM OUTSIDE CONTROL ROOM NO. 23 STEAM GENERATOR 49 2.4.11 SHUTDOWN FROM OUTSIDE CONTROL ROOM NO. 24 5 STEAM GENERATOR 50 2.5. 1 LOAD SWING TEST 100% TO 90% FEED FLOW -

I STEAM FLOW - FEED PUMP DISC PRESS.

STEAM HDR PRESS. - STEAM GENERATOR LEVEL FEED PUMP SPEED - ROD CONTROL BANK D 53 2.5. 2 LOAD SWING TEST 100% TO 90% AUCT. T avg -

AUCT, AT - T Hot -T Cold AUCT. NUC FLUX - PRESSURIZER LEVEL 54

2. 5. 3 LOAD ShING TEST 90% TO 100% FEED FLOW -

STEAM FLOW - FEED PUMP DISC PRESS.

STEAM HDR PRESS - STEAM GENERATOR LEVEL FEED PUMP SPEED - ROD CONTROL BANK D 55

2. 5. 4 LOAD SWING TEST 90% TO 100%

AUCT. Tavg - AUCT AT - T Ho t -T Cold AUCT. NUC FLUX - PRESSURIZER LEVEL 56 2.6.1 SALEM NUCLEAR GENERATING STATION UNIT. 2 RCC INCORE DETECTOR AND THERMOCOUPLE LOCATIONS 60 2.7. 1 LOAD SWING TEST 100% TO 50%

AUCT. Tavg - AUCT AT - T Ho t -T Cold I, AUCT. NUC FLUX - PRESSURIZER LEVEL 67

- 2. 7. , 2 LOAD SWING TEST 100% to 50%

FEED LOW - STEAM FLOW - STEAM HDR PRESS.

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STEAM GENERATOR LEVEL - FEED PUMP SPEED ROD CONTROL BANK D 68 v

109D (1 ) : 33 l

I LIST OF FIGURES FIGURE TITLE PAGE NO.

1

, 2. 8.1 SERIES 51 STEAM GENERATOR 73

2.10.1 GENERATOR TRIP TEST FLUX - PRESSURIZER LEVEL - AUCT. Tavg AUCT AT - STEAM FLOW - FEED FLOW

-I STEAM GENERATOR LEVEL - 21 LOOP T Ho t 21 LOOP T Cold 79

.I 2.11.'1 BORON MIXING AND COOLING

'._ LOOP 21, 22, & 23 - Tavg & AT 83

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2.11.2 -

P250 COMPUTER TREND DURING NATURAL

- c'IRCULATION C00LDOWN SUP 90.9 84

!. 2.14.1 SUP 81.6 - AUTOMATIC REACTOR CONTROL

_ s RECOVERY FR0l4 LOWERED Tavg 92 j

LW

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2.14.2 SUP 81.6 - AUTOMATIC REACTOR CONTROL RECOVERY FROM RAISED Tavg 93 I-. 2.15.1 INSTRUMENTATION BRACKETS, FEEDWATER HANGER Fh H-21-17 96

!E 2.15 2' INSTRUMENTATION BRACKETS, FEEDWATER HANGER FWH-21-18 97 il i

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I LIST OF FIGURES FIGURE TITLE PAGE NO.

3.1.1 LOOP 21 - AT (FIGURE 1) 101 3.1.2 LOOP 22 - AT (FIGURE 2) 102 3.1.3 LOOP 23 - AT (FIGURE 3) 103 3.1.4 LOOP 23 - AT (FIGURE 4) 104 I 3.3.1 TURBINE FIRST STAGE PRESSURE VS.

REACTOR POWER 115 3.3.2 PROGRAMMED REFERENCE TEMPERATURE I VS. FIRST STAGE PRESSURE 116 3.3.3 STEAM GENERATOR PRESURE VS.

REACTOR POWER 117 3.4.1 FW FLOW VS. REACTOR POWER LOOP 21 125 3.4.2 FW FLOW VS. REACTOR POWER LOOP 22 126 3.4.3 FW FLOW VS. REACTOR POWER LOOP 23 127 3.4.4 FW FLOW VS. REACTOR POWER LOOP 24 128 3.4.5 FW FLOW VENTURI DIFF. PRESSURE VS. FEED FLOW - LOOP 21 129 I 3.4.6 FW FLOW VENTURI DIFF. PRESSURE VS. FEED FLOW - LOOP 22 130 3.4.7 FW FLOW VENTURI DIFF. PRESSURE VS. FEED FLOW - LOOP 23 131 3.4.8 FW FLOW VENTURI DIFF. PRESSURE VS. FEED FLOW - LOOP 24 132 3.4.9 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 21 (100% RX POWER) 133 3.4.10 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 21 134 I

109(D): 35 I

I LIST OF FIGURES FIGURE TITLE PAGE NO.

3.4.11 ' STEAM RESTRICTOR DIFF. PRESSURE VS. FEED PLOW - LOOP 22 135 3.4.12 STEAM RESTRICTOR DIFF. PRESSURE i

VS. FEED FLOW - LOOP 22 136 3.4.13 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 23 137 3.4.14 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 23 138 3.4.15 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 24 139 I 3.4.16 STEAM RESTRICTOR DIFF. PRESSURE VS. FEED FLOW - LOOP 24 140 3.4.17 NO. 21 STEAM GENERATOR STEAM / FEED I ~

FLOW AND LEVEL 8/29/81 141 3.4.18 NO. 21 STEAM GENERATOR STEAM / FEED FLOW AND LEVEL 9/11/81 142 bW ND VEL / /81 143 3.4.20 NO. 23 STEAM GENERATOR STEAM / FEED FLOW AND LEVEL 9/11/81 144 3.4.21 NO. 24 STEAM GENERATOR STEAM / FEED FLOW AND LEVEL 9/11/81 145 l

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I LIST OF FIGURES FIGURE TITLE PAGE NO.

4 .1.1 POWER TILTS VS. CORE POWER 148 4.1.2 POWER TILTS VS. CORE POWER AND AXIAL OFFSET 149 4.1.3 POWER DISTRIBUTION - 200 RTP 150 4.1.4 POWER DISTRIBUTION - 30% RTP 151 4.1. 5 POWER DISTRIBUTION - 100% RTP 152 4.2.1 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 RCC INCORE DETECTOR THIMBLE AND THERMOCOUPLE LOCATIONS 155 4.2.2 BASE CASE FLUX MAP FOR ID2 EJECTION 156 I 4.2.3 4.2.4 EJECTED ROD FLUX MAP INCORE POWER TILTS 157 158 4.3.1 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 RCC INCORE DETECTOR THIMBLE AND THERMOCOUPLE LOCATIONS 161 4.3.2 BASE CASE FLUX MAP FOR DROPPED ROD 162 4.3.3 DROPPED ROD FLUX MAP 163 4.3.4 INCORE POWER TILTS, BASE CASE, DROPPED ROD 164 4.3.5 BASE CASE THERMOCOUPLE MAP 165 4.3.6 DROPPED ROD THERMOCOUPLE MAP 166 4.4.1 SUP 81.8 - DOPPLER POWER COEFFICIENT MEASUREMENTS VS. FSAR CRITERIA 171 4.5.1 SALEM 2, REACTOR ENGINEERING MANUAL 100 % DWR 8/2 2/81, UNIT 2 CHANNELS N-41, N-42, N43, N -4 4 LINEARITY CHECK FOR IN/EX CALIBRATION 177 4.5.2 SALEM 2, REACTOR ENGINEERING MANUAL, 100% PWR 8 /2 2/81 UNIT 2, DETECTOR N-41 NORMALIZED

,I DETECTOR CURRENTS 178 I 1X 109D (1 ) : 37 a

I LIST OF FIGURES FIGURE TITLE PAGE NO.

4.5.3 SALEM 2, REACTOR ENGINEERING MANUAL 100% PWR 8/22/81 UNIT 2, DETECTOR N-42 NORMALIZED DETECTOR CURRENTS 179 4.5.4 SALEM 2, REACTOR ENGINEERING MANUAL 100% PWR 8/2 2/81 UNIT 2, DETECTOR N-43 NORMALIZED DETECTOR CURRENTS 180 I 4.5.5 SALEM 2, REACTOR ENGINEERING MANUAL 100% PWR 8/2 2/81 UNIT 2, DETECTOR N-44 NORMALIZED DETECTOR CURRENTS 181 4.6.1 N43 PLATEAU 10/26/81 99% RTP N43 PLATEAU 11/19/81 91% RTP 184 4.6.2 NIS FULL POWER CURRENT VS. BURNUP 185 I

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I LIST OF TABLES I

TABLE TITLE PAGE NO.

2.2.1 TURBINE OVERSPEED TRIP TEST HISTORY 25 2.6.1 ISCl ROD DROP TEST RESULTS 59 2.8.1 STEAM GENERATOR MOISTURE CARRYOVER MEASUREMENT, DATA REVIEW 72 2.12.3.1 RCS CHEMISTRY / STEAM GENERATOR CHEMISTRY 88 3.1.1 DIFFERENTIAL TEMPERATURE VS. POWER LEVEL 100 3.2.1 STATEPOINT DATA

SUMMARY

107 3.2.2 STATEPOINT DATA, DIGITAL VOLTMETER READINGS 108 3.2.3 STATEPOINT DATA, LOCAL READINGS 109 3.2.4 STATEPOINT DATA, NIS AND CONTROL RM. READINGS 110 3.2.5 SECTION 2.6, CALORIMETRIC CALCULATION DATA SHEET 111 3.2.6 STATEPOINT DATA -

SUMMARY

DATA SHEET 112 3.4.1

SUMMARY

OF DATA FOR CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION 123 3.4.2 STEAM FLOW DATA REVIEW 124 l- 4.4.1 POWER AND DOPPLER, COEFFICIENT REVIEW 170 4.5.1 100% POWER (8/22/8:), EXCORE DETECTOR FLUX DIFFERENCE CALIBRATION DATA SHEET 175 4.5.2 100% POWER ( 8/22/ 81) , EXCORE DETECTOR FLUX DIFFERENCE CALIBRATION WORKSHEET 176 I

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I SECTION

1.0 INTRODUCTION

This report is in addition to the Startup Report submitted on May 1, 1981 which describe the testing from Core Load to the completion of Zero Power Physics Testing including Natural Circu.lation testing. Included in this report is the Power Range Test Program from 10% to 100% power testing. The period covered is from August 29, 1980 through October 13, 1981 with additional comments up to January, 1982.

Salem Unit No. 2 is a four loop pressurized water reactor of 3411 mWt rated capacity. The Nuclear Steam Supply System (NSSS) was supplied by Westinghouse Electric Corporation, the Architect Engineer was Public Service Electric and Gas Company and the Constructor w.1s United Engineers and Constructors, Inc.

The facility's operating license was issued April 18, 1980.

Preparations for core load were completed by May 22, 1980 and core loading commenced on May 23, 1980. Core loading was completed by May 27, 1980. Initial criticality was achieved on' August 2, 1980 and the zero power physics test program was completed by August 12, 1980. Natural circulation tests were begun on August 23, 1980 and were partially completed August 29, 1980 before the Unit was shutdown to repair a leaking Control Rod Drive Mechanism vent. The natural circulation testing had been completed from a testing standpoint but were required to be reperformed for operator training of nine licensed operators as committed to in the license. The Unit entered Mode 5 (<200 F) to repair the leaky CRDM housing vent 109D (1) : 40

I and remained in Mode 5. On April 22, 1981 a heat-up was commenced to enter Mode 2 (547 F, <5% RTP) to complete Na t Circulation Testing in anticipation of receiving the Opers License allowing power asension testing.

1

! During the period of time from August 29, 1980 through Apr 1981 the unit remained in Mode 5. Fire Protection System fications were made and Post -TMI design changes incorport as Engineering Design Package and materials arrived on-sit major delaying factor was the completion and trial run of Emergency Plans of PSE&G and the states of New Jersey and The coordinated emergency drill was successfully conductec 8, 1981.

In late April, upon review of the work performed on the f:

system to meet regulatory requirements, it was determined additional modifications were required. The review, by t:

staff, consisted of examining the capability of the unit down from a remote location should a fire occur in the Co o r elsewhere that could af fect the safe operation of the the Control Room. The inspection team concluded that the up control system was adequate to safely shutdown the uni emergency. The team said , however , that some modificatio fire protection system must be made before the Facility F License could be issued.

I The modifications involved, in part, improved protection cables needed for the operation of the plant from a remot I location. In the event of a fire that forces the evacuat o f the Control Room, the plant must be able to be operate 109D (1 ) : 41

using alternate locations and controls. Other modifications included upgraded procedures for dealing with a fire emergency emergency and improvement of various fire barriers and automatic s pr in kl e r s .

Upon completion of those modifications the full power operating license was granted on May 19, 1981.

Figures 1.1 through 1.8 graphically display the power history o f Un it 2 f rom May 1981 through December 1981. Along with the graph is an explanation of the testing at the particular power level and the day the test was performed. Al so listed are the reasons for various trips and delays effecting the startup s ched ul e . Fig ures 1.10 through Fig ure 1.14 shc w the planned

.I vs. the actual number of days of testing at each power plateau.

The total planned days were 166 whereas the actual days were 186. Several of the larger delays were:

1) the replacing of an intermediate range channel (3 1/2 days)
2) steam flow sensing line modifications (8 days)
3) main turbine generator governor adj ustment (4 days)
4) condenser tube leakage requiring unit shutdown (5 days)
5) outage to modify steam generator separation equipment (15 days) l 6) unplanned reactor trips (10 days) i Following the steam generator outage in September 1981, to modify i

j the r.oisture separation equipment, the moisture carryover test (Section 2.8) was reperformed and verified the steam moisture content to be acceptable. The following day Salem Unit 2 was declared commercial, October 13, 1981, com pleting the startup test y 109D (1 ) : 42

I program. Several main feedwater pump trips had occurred between October 1981 and December 1981. A test program was formulated to determine the cause of the believed " low suction pressure" l

trips (see Section 2.7 for details). In December 1981 the high l

steam flow indications, observed prior to the steam generator l modifications., re-occurred. An additional moisture carryover test i

was conducted in January 1982 (see Sections 2.8 and 3.4 for details).

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I SECTION 2.0 INTEGRATED TESTING This section deals with testing performed to evaluate overall plant response to rapid load changes and changes in system parameters to evaluate integraded systems response. Included in this section are tests performed by the Chemistry and Ilealth Physics department which were part of the startup program.

2.1 SUP 81. 5 - DYNAMIC AUTOMATIC STEAM DUMP CONTROL

.I This test was performed at the 10% power testing plateau with

~

reactor power being varied between 1% and 10% depending on the test requi rem ents . The turbine generator was not on the line at this time. The objectives of the test were to verify the proper operation of both modes (turbine trip and load r ej ec tion) of the T avg steam dump control system and to obtain final settings for steam pressure control mode of the steam dump valves (12 valves) .

To test the turbine trip mode of the steam dump control system,

! with the steam dump control system in manual and reactor power at about 1%, T gyg was raised 3 F above the normal no load Tavg 1

o f 547 F. the steam dump system was then placed in automatic was controlled by the and since the turbine was tripped, Tavg turbine trip controller. The steam dump valves opened and Tavg l

( 109D (1 ) : 44 1

I was controlled within a degree of the no load Tavg V^1V**

Reactor power was then increased to about 6%, at 2%/ minute, by rod withdrawal. The steam dump valves opened as reactor power increased and T avg was controlled between 3 F and 5 F above the no load T,yg value. Reactor power was then decreased and the steam dump valves modulated closed, tracking reactor power.

In testing the response of the 1 css of load controller, the steam dump system control was initially placed in manual and reactor power increased to approximately 3%. The turbine was latched and the T ref input to the loss of load controller was disconnected. A test signal was injected in place of T ref signal, which was equivalent to a T f 4 F less than the ref no load of T f 547 . The steam dump controller was then ref switched to the T mode of control. T increased above avg avg the test signal by approximatelv 5 F due to controller dead band and then by another 2 F to provide a steam dump valve position equivalent to 3% reactor power. Power was then reduced to 1% and T returned to its no load value. The avg T mp rator was found faulty and replaced and tested.

I avg The response of the steam header pressure controller was

.I earily verified. The steam dump control was placed in the i 109D (1) : 4 5 i

lI l

I 1

I steam header pressure control mode with a controller set-point of 1005 psi . Reactor power was increased to approxi-

. mately 5% and the steam dump valves modulated open maintaining steam generator pressute at 1005 psi (T avg was increased from 547 F to 550 F) .

The only mechanical difficulty encountered during the tests were the " popping" open of the dump valves instead of modulating open as designed. The flow markings on the valve indicated the valve might have been installed incorrectly, but later investigations indicated the valves were installed c o r r ec tl y. The diaphram operated valve is designed to modulate from the fully closed position to the fully open position using supply air pressure of 9 psi thru 45 psi. Stroking of the valve required 25-35 psi to pop the valve off its seat at which point it would then modulate until closed.

. Disassembly of the valves for inspection of the intervals indicated no abnormal conditions. The valve vendor was contacted and arrived on site for observation of the valve operation during the 100% testing plateau. New internals were ordered and installed with no change in valve operation.

The internals are to be modified and retested until the operation of the valve is acceptable. The. modification consists of adj usting the trim of the internals o f the valve to relieve the off-balancing of the valve disc during the opening stroke which is causing the valve disc to bind on the seating l

109D (1 ) : 46

surface. Once the correct internals are designed for one valve the other 11 valves will be modified. This modification is also planned for Unit 1 valves.

The p; resent operational characteristics of the valves are acceptable for plant operation as determined during plant I trip tests and load swings; but the modulating rather than popping operation of the valves would provide smoother transient - for steam generator pressure and levels, and RCS temperatures and pressures.

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109D (1 ) : 47

i. g I
2. 2 SUP 81.3 - TUR31NE OVERSPEED TRIP TEST The test of the mechanical turbine overspeed trip device was performed during the 10% power testing program. The purpose of the test was to verify that the turbine overspeed pro-tection device would operate to trip the turbine in the event of an overspeed condition. Prior to the test, the turbine-generato'r was operating at approximately 10% power for eight hours in order to bring the machine to thermal equilibrium.

Af ter the turbine-generator was unloaded and prior to the overspeed~ test, the operability of the overspeed mechanism

~

was checked. At the pedestal end of the turbine, oil was intho'duced up to 48 psig to the overspeed trip mechanism to trip the; mechanism. The manual trip lever moved from the normal to the trip position indicating that the mechanism g was operating freely and had tripped.

,To : allow th'e turbine-generator to actually overspeed, the OVERSPEED PROTECTION ' CONTROLLER had to be removed from s'e rv ic e . This was. easily accomplished by use of a key

c. 8;

{ ..[ s -switch on- the contro1~ console. Using the E-H CONTROLLER, turbine 0speed'was I,ncreased at a rate of 50 rpm /m until the c ,

'/ ' UnittrIpped.

The m'aximum allowable overspeed is 1998 rpm. During the

three test r un s. , the Unit tripped at 1955, 2003 and 2000 rpm.- It wfs determined that a weight adjustment of

+ .

m i N'_'

i x

~

109D (1 ) : 48

I the governor was required. Af ter the weight adj ustment the oil-trip test was repeated with a trip oil pressure of 56 psig required. Oil pressure should have decreased following the adj ustment . An inspection of the governor adj ustment mechanism was made and retests of the oil-trip test were inconsistent (62 psig-90 psig) . The governor mechanism was disassembled and inspected with no abnormal conditions found.

Upon reassembly the oil-trip retest was consistent and the mechanical overspeed test performed with the trip speed still too high. A recheck of the oil-trip pressure came up with inconsistent oil-trip pressure. The governor mechanism was disassembled and new parts installed that were machined to increase the clearances to allow more freedom of movement.

Subsequent oil-trip retests and mechanical overspeed tests were successful. Sse Table 2. 2.1 for the sequence of events.

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I I

I I

109D (1 ) : 49 I

I TABLE 2.2.1 TURBINE OVERSPEED TRIP TEST HISTORY I

Date Event 6/5/81 Oil-trip 48 psig Mechanical trip 1955, 2003, 2000 rpm

6/6/81 Oil-trip after weight adjustment 58, 56, 56 psig
Oil-trip after governor inspection 62-90 psig ll im 6/7/81 Oil-trip (after governor disassembly and cleaning) 58, 58, 56 psig
6/8/81 Mechanical trip - 2000 rpm 6/9/81 Replacing governor internals / machining J
g 6/11/81 Oil-trip 24, 22, 21 psig Mechanical trip 1836, 1841, 1840 rpm

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!I ll 109D (1) : 50 lI

I 2.3 SUP 82. 6 - LOSS OF OFF-SITE POWER This startup procedure was completed during the 10% testing plateau. The purpose of the test was to demonstrate that the emergency power system was capable of maintaining the plant in a safe condition by carrying the required loads for at least thirty minutes following a plant trip caused by a I total lo ss o f o f f- si te po we r .

With the turbine generator on the line at minimum load (10%

reactor power) and with a normal electrical lineup, the black-out was initiated by opening the 13kV infeed breakers for 21 and 22 station power transformers followed closely by the operator opening the generator output breakers. All operations were carried out from the control console.

I All systems, equipment and indicators operated properly. The three diesel generators picked up their respective blackout loads and ran for the required thirty minutes. No problems were encountered during the test.

Fig ures 2. 3.1 thru 2. 3. 9 indicate plant trends fo r pressure, levels and temperatures of the RCS and steam generators during the transient using Plant Computer data and Control I Console recorder strip charts.

I I I 109D (1 ) : 51 I

1 i

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I 2.4 SUP 82.5 - SHUTDOWN FROM OUTSIDE OF THE CONTROL ROOM This was the last test performed at the 10% testing plateau.

The purpose of this procedure was to verify that the plant could be shutdown from outside the Control Room and be I maintained at hot standby for an hour utilizing the minimum shift crew. Limits on various parameters (pressurizer level, pressure, T,yg, steam generator levels) were included in the acceptance criteria. All control systems were kept in automatic and the Control Room was not evacuated for the test.

The procedure was modified to incorporate the requirements of Amendment 6 to the Operating License based on the results of a fire protection review. The additional requirements included:

1) local start of a diesel generator using alternative control power source
2) local operation of a 4kV breaker

.I 3) local start of the containment fan cooler unit.

I 4) local operation'of a motor operated and an air operated valve

5) local control of charging flow l This portion of the testing was performed while the plant was being maintained in HOT STANDBY at the remote control station.

l The plant was operating at 10% RTP with the Control Room manned by the regular shif t personnel (2 NCO's). The minimum lI 1090(1):52

!I

!I

I shif t crew was ccmprised of the following:

SRO - Senior Reactor Operator - 1

'NCO - Nuclear Control Operator - 2 EO/UO Equipment / Utility Operator - 3 STA - Shif t Technical Advisor - 1 Electrician - 1 The following stations were designated:

(1) Unit 1 Control Room SRO (1 ) , STA (1)

Electrician (1)

(2) Hot Shutdown Panel (213) NCO (1 )

(3) Reactor Trip Switchgear NCO (1 )

(4) Main Feedpump Local Control Panel EO (1 )

i I , (5) Auxiliary Feedwater Pump Panels EO (1)

! (6) Main Turbine Turning GEAR UO (1 )

The Shift Supervisor of Unit 2 simulated the evacuation of Unit 2 Control Room and established a Control Center in Unit 1 Control Room. From the center the shutdown and control of the plant in HOT STANDBY was maintained thru communications to the personnel at remote stations. The minimum shif t crew was assigned their positions at this point. The NCO tripped the plant at the Reactor Trip Switchgear and de-energized the Rod Drive MG sets to insure that an ATWS (anticipated transient without scram) event would not occur and that all rods would drop to the bottom of the core.

The EO and UO in the turbine building verified the 500 kV I

109D (1 ) : 53 l

l

breakers and field breakers were open and the group buses had transferred from the Auxiliary Power Transformer to the Station Power Transformer. They also verified that the main turbine generator and main feedwater pumps had tripped, and placed them on their turning gears when they coasted to a stop. The NCO and EO at the Hot Shutdown Panel (213) verified the following:

(1) Pressurizer level controlling automatically at 22% 1 5%.

(2) Pressurizer pressure controlling automatically 2235 psig 1 50 psig (3) Steam generator pressure 1000 psig i 25 psig and level (wide range) 58% - 69%

At this point (75 minutes from reactor trip) a one hour hold period was commenced to demonstrate the ability to maintain the plant in a HOT STANDBY condition f rom the HOT SHUTDOWN PANEL. The NCO and EO at the HOT SHUTDOWN PANEL manually opened and closed the auxiliary feedwater pump discharge valves to maintain steam generator levels as indicated at the HOT SHUTDOWN PANEL (the auxiliary feedwater pumps and associated valves are located next to this panel) within the wide range indication of 58% - 69%.

Following the one hour hold period, the second phase of this test started while the plant was being remotely maintained in HOT STANDBY. Using the appropriate sections from the Fire Hazards Analysis - Emergency Equipment Operation the following operations 109D (1) : 54

I were demonstrated as directed by the Unit 2 Shift Supervisor:

(1) Local start of #24 Fan Coil Unit (2) Local start of #24 Service Water Pump (3) Local operation of 22SW17 (header isolation)

(4) Local operation of 2CV55 (control charging)

(5) Local starting of a Diesel Generator These operations were performed by the Shift Electrician who performed the control circuit modifications and a NCO who operated the equipment. All phases of this test were demonstrated successfully. The following graphs (Figure 2.4.1-2.4.11) indicate plant trends for pressures, levels and temperatures of the RCS and steam generators during the transient and stabilization period using Plant Computer data

. and Control Console recorder strip charts.

I .

109D (1) : 55

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3 I 2. 5 SUP S2.1 - LOAD SWING TESTS

'The load swing tests were a series of integrated plant response I tests performed to verify that the plant was capable of auto-matically accepting a 10% step load change from 30%, 75% and 100% po we r . The load changes were initiated, using the turbine Electric Hydraulic control system, at a rate of 200% per minute.

Each test consisted of two parts: a 10% load decrease fo llo wed ,

a f.ter equilibrium conditions had been reached , by a 10% load increase.

The ' load swing was evaluated based on the following criteria :

(1) - The reactor and/or turbine did not trip (2) Sa fety inj ection was not initiated

'(3) Neither steam line relief valve or safety valve lifted

, (4) Neither pressurizer relief valve or safety valve

~'

lifted s

(5) No operator action required to restore plant conditions to steady state

'(6) Plant variables 'such as T,y , feedwater flow, steam flow, etc. should not incur sustained I1 (7) oscillations or large variations.

Nuclear po wer overshoot less than 3 % fo r the load decrease.

s The initial load-swing at 30% po we r wa s to 13%. The overshoot wa s d ue to . setting the turbine load reference too low.

Automatic steam generator level control cannot control s

at less than 15% RTP so operator action was required to control levels in the steam generators. The up po we r swing 109D (1 ) : 56

_ _ _ _ _ _ _ s____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _

I .

1 to 30% RTP had acceptable results. All other parameters during the 30% swings fell within the acceptable range.

Testing .at 7 5% and 100% had similar results. The results from the 100% RTP swings are depicted in Fig ures 2. 5.1 thru 2. 5. 4.

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2.6 SUP 82.4 - RODS DROP AND PLANT TRIP This test was performed at 50% power. The purpose of the test was to demonstrate the operation of the negative rate trip circuitry. Two rods were dropped from the rod group most difficult to detect by the excore detectors due to low worth and/or core location. In addition, it was the function of the test to review plant response and control systems behavior to a trip from an intermediate power level prior to the plant trip test from 100% power.

The two rods chosen were located in core positions B-4 and D-14.

The acceptance criteria specified that a reactor trip must occur as a result of two of the four negative rate bistables tripping. (see figure 2.6.1 for the core location of the rods vs. the excore detectors).

The rods selected were dropped simultaneously from the Rod Control System DC Hold Power Cabinet. The negative rate bistables for N42 and N13 tripped simultaneously (+ .01 seconds) followed by N41 and N44 negative rate bistables .5 seconds later. The N41 and N44 bistables tripped following the reactor trip from the N42 and 43 bistables at a point where the control rods were dropped approximately 50% into the core. Also measured during the reactor trip was the control rod drop time of IScl (core location E3) for comparison with the drop time at zero power.

This rod was also monitored during the 100% trig per SUP 82.9, 109D (1) : 58

i GENERATOR TRIP. See Table 2. 6.1 fo r a summary of the results.

1 The rod drop time was faster as measured during the reactor

g 5 trip f rom power operation compared . to zero power operation.

! Following the trip the plant was stabilized at the no load Tavg of 547 F successfully.

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109D (1 ) : 59 i

l__ - _ _ _ - - _ _ _ _ _ , _ _ . _ .- - - - - - -

TABLE 2.6.1 ISCl ROD DROP TEST RESULTS Initiations of Event Initiations of Event RCS Conditions To Da shpo t En tr y to Bo ttom o f Da shpo t 1

0

! 547 F, 4 RCP's 1.35 (Seconds) 1.88 0 % Po we r 563 F, 4 RCP's 1.26 1.82 50 % Po we r o

570 F, 4 RCP's 1.30 1.85 100 % Po we r I

I e O

I ll - Se -

109D (1 ) : 60

- __ _ 1

Figura 2.6.1 i SALEM NUCLEAR GENERATING STATION l( UNIT NO. 2 RCC INCORE DETECTOR THIMBLE AND THERMOCOUPLE LOCA TIONS S/G 22 Cold Leg S/G 21 Cold Lcq S/G 22 Hot Lcq S/G 21 Hot Leg I

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I 2.7 SUP 82.2 - LARGE REDUCTION TESTS The LARGE LOAD REDUCTION TESTS were conducted to verify the capability of the primary and secondary systems to automatically accept a 50% load reduction from 75% and 100 % of f ull po we r .

The tests were further used to evaluate the interaction between I control systems and to determine if system setpoints should be changed to improve transient r e s po r.s e .

75% POWER I

The first tests were conducted at the reactor power level of 75%.

The load changes were initated by using the Electric Hydraulic I system at a rate of change of 200% per minute.

The follcwing list shows the range of selected parameter move-ments during the power reduction:

1. Feod Water Temperature 401 F -

335 F

2. Steam Header Pressure 792 psi -

878 psi

3. Feed Flow 69?':vg) - 26% (avg)
4. Steam Flo w 69% ( vg) - 26% (avg)
5. Steam Generator Level 44% - 24% - 44%
6. Control Bank D 228 steps - 122 steps
7. Av erag e Loop AT 46.5 F -

22 F

8. Nuclea r Po we r Fl ux 71% -

32%

9. Pressurize r Level 39% -

29%

I 10. Pressurizer Pressure 2260 psi - 2175 psi - 2300 psi 2250 psi I 11. Auctioneered T,yg 563 F - 570 F - 554 F

12. Plant Load ( MW-G r o ss) 765 megawatts - 285 megawatts

~ ~

109D (1 ) : 61

I The following acceptance criteria items were met:

1. The reactor and turbine did not trip.
2. Sa fety inj ection was not initiated.
3. Pressurizer safety valves did not lift.
4. Steam generator safety valves did not lift.

I During the transient the steam dump valves operated (condenser d um ps) to restore T avg to its program value. The dump valves did not open until there was a maximum demand signal. They are designed to modulate open based on an increasing demand signal but instead they " pop" open when they receive a max imum demand signal . (See SUP 81. 5, Dynamic Automatic Steam Dump Control) . This type of operation causes the steam generator levels to drop to a lower level than they normally would due to the shrink effect in the steam generators caused by the rapid increase in steam generator pressure. Once the cause of the popping operation is corrected and the valves modulate with demand, the steam generator pressure will peak at a lesser pressure assisting in reducing the amount that the steam generator levels are lowered to before they are restored to normal. The present operation of the steam dump valves is acceptable but is being r ev ie wed .

100% POWER The LARGE LOAD REDUCTION f rom 100% power was performed prior to the plant trip test. The load change was initiated at a I I 109D (1 ) : 62 I

I E rate of 200% per minute using the Electric Hydraulic system.

The following acceptance criteria items were met:

1. The reactor and turbine did not trip.
2. Safety inj ection was not initiated.
3. Pressurizer safety valves did not lift.

The following list shows the range of selected parameter move-ment during the power reduction.

Feed Water Temperature 432 F -

380 F 1.

774 psi - 984 psi - 852 ps

2. Steam Header Pressure
3. Feed Flow 99 % (avg) - 48% (avg) 98 % (avg) - 46% (avg)
4. Steam Flow
5. Steam Generator Level 44% - 49% - 28%

228 steps - 141 steps

6. Control Bank D Av e rag e Ic op AT 62.5 F -

36.5 F 7.

I 8. Nuclear Power Flux 100% -

56%

9. Pressurizer Level 50% - 34% - 59% - 34%
10. Pressurizer Pressure 2235 psi - 2322 psi - 2108 p 2335 psi 571 F - 580 F - 559 F
11. Auctioneered T avg I 12. Plant Io ad ( MW-G r o ss) 1140 megawatts - 590 megawat During the transient, the steam dump valves operated at a lower demand signal providing a smoother transient for steam generator level and pressure than was observed during the 75% testing plateau.

I 109D (1 ) : 63

I Prior to the commencing the load reduction the condensate polishers were bypassed providing an additional 75 psi at the feedwater pump suction. This was done based on observation during the 10% load swing from 100-90-100%.

During the 10% load swing it was noted that the MSR Coil Drain Tanks dropped in level causing the Heater Drain Pumps to go in a recirculation mode reducing the feed' water pump suction pressure by approximately 60 psi . The cause of this transient is unknown at this time. To prevent reactor trip on loss of a feedwater pump on low suction pressure the polishers were bypassed to provide additional margin at the f eedwater pump suction. During the 50% load reduction the feedwater pump suction decreased by approximately 90 psi l again due to the loss of the Heater Drain Pump flow. This transient is being reviewed to determine its cause.

1 The overall transient was acceptable and is graphically l d isplayed in Fig ures 2. 7.1 and 2. 7. 2.

I Following several feedwater pump trips from 100% RTP, due to a loss of feedwater pump suction pressure, the feed and

[ condensate system was instrumented for continuous monitoring to determine the cause of the rapid reduction in suction pressure. Following a feedwater pump trip, the review of

- the data indicated the cause to be compounded by a rapid load reduction to restore suction pressure. Several load reductions (7) were made and reviewed to further determine the cause and effects.

E 109D (1 ) : 64 r

I During a load reduction rondition the H.P. turbine extraction pressure decreases accordingly leaving the saturated liquid of the heater drain system in an unstable condition (subject to flashing). The instability results in decay of heater drain tank level which calls for the heater drain pump discharge valves to close and therefore heater drain flow (1/3 total feedpump demand suppled to feedpump) is reduced.

The loss of flow supplied from heater drain system is pro-portionally more severe with greater load reduction resulting in a corresponding and instantaneous loss of feed pumps suction pressure.

Based on the load reduction tests, it has been determined that with the condensate polisher in service the system can correct for large reduction effects if the initial power level is below 85%. When operating at power levels in excess of 85% the following operation precautions are being observed to eleviate the possibility of a feedpump trip:

1) When operating above 85% power open bypass (2CN-47) around 23, 24 & 25 heaters, (gain of 30 psi on feed-water pump suction).
2) Heater Drain Pump recirculation (21HD17, 22HD17, 23HD17) valves are to be failed open to add stability to heater drain system.
3) The feedpump suction pressure alarm increased to 300 psig from 270 psig for early warning (feedpump trip at 215 psig).
4) If an alarm occurs and a rapid load reduction is necessary (provided their is no chem stry problem I present) the condensate polisher bypa s can be opened to recover feedpump suction pressure end the operator -

can then successfully complete a large load reduction (gain of 65 psi on feed pump suction) .

E - 6s -

g 109D (1) : 65

I

5) If an alarm occurs and a sudden load reduction is not necessary, the load should be reduced in 1-10% increments until suction pressure is restored.

The following design charges are being revie wed to provide stability to the feed pump suction during steady state operation and transients.

I 1) Increasing the capability of the existing condensate pumps for greater head at the same flow (impeller change) .

2) Diverting a percentage of condensate flow after the No . 2 feedwater heater to a spray sparger in the heater drain tanks.
3) The possibility of new condensate pumps, and an additional condensate pump or condensate polisher booster pumps .
4) Commonize the heater drain pump suctions so that one pump could be used as a standby pump. This change is I also expected to stabilize the heater drain pump flow at steady state and transient conditions.

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I 2.8 SUP 82.7 - STEAM GENERATOR MOISTURE CARRY OVER MEASUREMENT The steam generator moisture carryover measurement was orignally scheduled to be performed at three power levels; 75, 90 and 100%.

l l

Based on previous experience, Westinghouse ( the turbine gener-ator and NSSS supplier) recommended that we perform this test only at 100% RTP under various conditions. The variations included altering T , steam generator levels, and reactor power to determine their effects on moisture carryover. The determination was made to conduct the test as recommended.

The test procedure requires the use of a radioactive tracer, Sodium 24, in the form of an aqueous solution of sodium nitrate.

The traces had an activity of 1.2 curies at the time of its arrival on-site af ter being activated at the University of Missouri Test Reactor. The tracer was inj ected into the feed-water system using the chemical addition system. The phosphate feed pump (5 gpn) took suction from a 50 gallon drum to which the source was added (located on 100' elevation of the turbine building) and inj ected it into the main feedwater line of each steam generator j ust upstream of the main feedwater regulating valve. The inj ection lines were valved in and out to equalize the quantity of the source in each steam generator ( 3 minutes for each steam generator) .

I I 109D (1 ) : 67 Ii - -

I

I Three sets of samples were taken at each test condition. Each set consisted of a blowdown sample and a steam sample from each steam generator and a common feedwater sample for a total of 27 samples per test condition. The results were based on analysis of the blowdown and feedwater samples. The main steam samples were analyzed , but the results are used as a general indicator for comparison of each steam generator moisture removal performance.

During the performance of the first test, on 8 /2 8/81, two vials containing the source were received and used for inj ection into the feedwa ter system. This resulted in uneven distribution of the source from one steam generator to the next.

When the test was re-run on 9/10/81 and 10/12/81 four vials were used each containing equal strength. Each vial was added to the 50 gallon drum separately and fully inj ected into one steam generator, the drum flushed to the generator and then a second generator lined up. The process repeated for each steam generator. This method provided excellent distribution of the source with less than 5% deviation in activity levels between any tw'o steam generators.

I The perfo rmance of the first moisture carryover test was conducted o n Aug us t 28, 1981 a t 100% RTP and 44% steam generator levels, and at 100% RTP and 40% steam generator levels. With normal levels (44%) in the steam generators the average carryover was .33%.

~

j With levels reduced to 40% the average carryover was .20%. The tI 109D (1 ) : 68

'I

I steam samples indicated that #21 steam generator moisture carryover was significantly higher than the other three steam generators. It was also noted that #21 steam flow signals were 10-20% higher than the other steam generators (see SUP 81.7, calibration of stream flow and feedwater flow instrumentation). A retest on September 10, 1981 had similar results with #21 steam generator having the greatest carryover.

Even with reduced levels, the carryover on #21 steam generator was still greater than 1.0%. The determination was made to shutdown and inspect the internals of all four steam generators.

The inspection did not uncover any faults. The second stage separation equipment was modified to provide additional moisture drainage paths for each steam generator to reduce any carryover.

(See Figure 2.8.1)

A retest performed on October 12, 1981 indicated the average carryover was .13% with normal steam generator levels (44%).

The carryover measured for #21 steam generator had been reduced to .17%. The acceptance criteria was less than .25% moisture carryover and was easily met. It was also noted that No.21 Steam Generator steam flow signals were reduced in magnitude by 10-20%.

See Table 2.8.1 for a review of the carryover measurements.

Note:

An additional carryover test was performed on January 22, 1982 due to reoccurrence of high steam flows in December, 1981 for no apparent reason. The results from that test were similar I to the results in October, 1981. See Section 3.4 for explanation.

I 109D (1) : 69 I

M M M M M M M N m M M M M M M STEAM GENERA'IDR MOISTURE CARRYOVER MEASUREMENT DATA REVIEW Power SG SG T Feed with Level Levels Pressure ag Ioad Cahybr Temp carrhI5682 (%) Ph8Nure Notes Dates (%) (%) (PSIA) (UF) (MWC) (OF) (%) 21 22 23 24 (PSIA)

Before 8-23-81 99.7 44 802 572.2 1110 431 .342 - - - -

3.4 MOD's 99.2 40 808 572.6 1140 431 .200 - - - -

3.3 9-11-81 98.5 44 813 572.4 1134 432 .280 1.49 .69 .28 .02 2.3 Retest for 98.9 36 809 572.3 1140 432 .180 1.27 .44 .13 .01 2.3 Verification of 91.8 44 813 569.9 1053 426 < Min. Detectable Activity 2.2 Carryover 96.7 44 812 571.4 1110 430 .080 - - - -

2.4 After 10-12-81 100.1 36 791 570.6 1187 432 .125 .16 .14 .12 .01 1.7 SG 99.67 44 792 570.8 1180 432 .133 .15 .15 .11 .01 1.7 MOD's 94.33 44 792 567.6 1110 430 .05 - - - -

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I 2.9 SUP 82.8 - NSSS ACCEPTANCE TEST This was one of the last tests performed during the startup of Unit 2. The test had two purposes: first, to demonstrate the reliability of the NSSS by maintaining the plant at rated out-put of 3411 MW (+0%, -5%) for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> without a load red uc-tion or plant trip resulting from an NSSS malfunction and second, to measure the gross electrical output and the turbine heat rate to verify the capability of the Un i t .

I The 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> run was attempted several times during the 100%

testing plateau. Each time the unit was forced to red uc e power to clean condensate, heater drain or feedwater pump

.7 trainers. The maximum run at greater than 95% po we r wa s 69 hours7.986111e-4 days <br />0.0192 hours <br />1.140873e-4 weeks <br />2.62545e-5 months <br /> before a load reduction was required. Total accumulated time at greater than 95% power was 205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br />. Based on the fact that the load reductions were not NSSS related the 100

  • I hour run was accepted on an accumulated basis. Since this run, Unit 2 has accumulated better than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous operation at greater than 95% po we r .

I The second part of the NSSS acceptance test was the measurement of the gross electrical output and the turbine heat rate.

No rmal plant instrumentation was used as a backup to a primary data logger which was fed from test instrumentation installed for the test run. The data logger received a data scan every 2 minutes whereas the plant instrumentation was read every I 109D (1 ) : 72 I

30 minutes. Data was accumulated over a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period with reactor power and turbine power held steady at 100% RTP. The results were corrected to design conditions and are listed t Co r rected

  • Date Goss Electrical Output Heat Rate 8/27/81 1161.4 - (MW) 10,059 (BTU /KWH) 10/15/81 1172.4 9,963 10/21/81 1173.9 9,950 I *The test run on 8/27/81 had 5 out of 6 circulating water pump operating. To reduce the errors involved in correcting for backpressure, data was taken with 6 circulating water pump on 10/15 and 10/21/81. The calorimetric results using plant vs.

test instrumentation -indicated less than a .5% difference for all three test runs.

I I

I I .

I 109D (1 ) : 73 I --

I 2.10 SUP 82.9 - GENERATOR TRIP FROM 100% POWER This was the last test performed during the startup of Salem 2.

The test had three purposes : first, to verify the capacity of the primary and secondary plant systems to automatically accept a generator trip from 100% power and to bring the plant to a stable condition following the trip; second, to verify that the turbine overspeed mechanism operates to limit turbine speed in an actual overspeed condition; and finally, to deter-mine the overall response time of the reactor coolant hot leg resistance temperature detectors (RTD). The RTD response time was defined as the time interval between the point where the neutron flux had decreased by 50% of its initial value to the point where the hot leg temperature had decreased by 33-1/3%

of its initial loop AT value. In addition the rod drop time of a selected rod was monitored during the reactor trip.

l I The reactor was operating at 98% power when the trip was initiated at 0924 hrs. on September 2, 1981. The trip was initiated by the Control Room Operator simultaneously opening the main generator output breakers to the 500 KV bus (bus sections 1-9 and 9-10) to trip the generator. The main turbine tripped on overspeed at its trip setting of 183 5 R PM in less than 1 second causing a reactor trip. The main steam stop valves (21-2 4M S167 ) were closed immediately a f ter the reactor trip to contain as much heat as possible in the RCS to avoid delays in restoring Tavg required fo r SUP 90.9, BORON MIXING AND COOLDOWN TEST. This caused the 109D (1 ) : 74 I

I steam generator pressure to increase higher than normally e x pec ted . The atmospheric relief valves (21-24MS10) operated as designed to maintain steam generator pressure at approximately 1000 psig. On No . 23 Steam Generator the setpoint of the first safety valve was reached (2 3M S 15, 1070 psig) approximately 30 seconds following the trip.

The actual pressure reached was 1050 psig indicating the safety valve is set on the " light side". The valve was reset and did not lif t again.

I The remainder of the test went smoothly. Listed below are some of the parameters that were monitored and how they varied during the test:

I -

Pressurizer level ranged from 46% at the time of the trip, to a high of 46% and a low of 23%.

Pressurizer pressure ranged from 2240 psi prior to the test, to a high of 2240 psi and a low of 2010 psi .

When the generator output breakers were opened, turbine speed increased 90 rpm to 1890 rpm.

l Steam generator levels decreased from 44' to 0 % indicated level .

t T f 597 F at the start of the test to 547 F.

l H

T f 570 F at the start of the transient to 547 F.

avg Maximum steam dump demand was generated 5 seconds af ter the generator trip.

l 109D (1 ) : 75 l

I

J

, The steam adunp valves started opening six seconds af ter the g enerator trip.

The RTD response time was measured to be 5.5 seconds versus a maximum of 6. 3 seconds. The rod drop time was measured

', as 1.30 seconds from the fully withdrawn position at the time of decay of stationary gripper coil voltage to dashpot

_ entry versus the Technical Specification maximum of 2. 2 seconds. Th e rod measured was 1SCl.

I The transient response of selected parameters is graphically d isplayed in Fig ur es 2.10.1.

I .

O

!I I 109D (1 ) : 76

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I 2.11 SUP 90.9 - BORON MIXING AND COOLDOWN The objectives of this tests were:

(1) to borate the RCS and verify boron mixing while in the natural circulation mode.

(2) to demonstrate the capability to cooldown the RCS while in natural circulation.

This test was originally scheduled to be performed during the initial natural circulation test program af ter low power physics testing. During that time the reactor was maintained critical at approximately 3% RTP af ter the RCP's were secured.

To perform a more realistic test the test was delayed until the Generator Trip Test (Section 2.10) when the reactor core had built-in decay heat following 1310 MWD /MTU of core burnup and 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> at greater than 95% RTP accumulated (see fig ures 1.1 thru 1. 5 f o r reacto r power histo ry prio r to the plant l t r i p) .

The Ge ne r a to r Tr ip Te st wa s initiated on 9/2/81 at 0924 (see Section 2.10) and caused an immediate turbine trip and reactor trip. The plant was stabilized at a T f 547 F.

avg The pressurizer and RCS were sampled to obtain a baseline boron concentration (960 ppm). At 10 5 6 h r s . , the RCP's were tripped, t im e " O" on figure 2.11.1.

T avg and Delta T stabilized af ter 15 minutes. T avg increased to 558 F and delta T to approx. 18 F. Auxiliary spray was initiated to maintain pressurizer pressure at 2235 psig (charging isolated 109D (2 ) : 01

j and reinstated after 2 minutes). From this point on, auxiliary spray was used to control pressurizer pressure since normal spray is not effective without the driving head of the RCPs.

Starting at 1116 hrs. a boration of the RCS at 5 gpm was commenced to increase the RCS boron concentration by 100 ppm, from 960 ppm to 1060 ppm. By 1226 hrs.,367 gallons of boric acid had been added to the RCS, the boron concentration, as measured in loop 21 and 23 hot legs, was 1031 ppm. At 1410 hrs. the RCS samples indicated the corcentration had stabilized at 1090 ppm. Using the auxiliary spray flow, the pressurizer boron concentration was increased to equal the RCS boron concentration. As indicated in figure 2.11.1, loop 23 T ayg and Delta T indicated a sharp increase in temperatur-This was due to diverting all charging flow to the pressurizer to equalize the RCS and pressurizer boron concentration. The increase in flow to the pressurizer through the auxiliary spray line caused an outsurge from the pressurizer through the surge line. The hotter water was sensed by loop 23 hot leg RTD.

The second phase of this test was to verify the ability to cooldown the RCS while in natural circulation. Cooldown was commenced at 1600 hrs. using the steam generator atmospheric relief valves (MS-10fs) starting with an average T avg of 555 F The cooldown rate was limited to less than 25 F per hour with an average cooldown rate of 19 F per hour. The cooldown 109 D (2) : 02 I

I

was secured four (4) hours later at which time the average coolant temperatEce was 480 F. The only difficulty encountered during the cooldown was the controlling of the MS-10s to maintain the steam pressure difference between the steam generators to less than 100 psi to avoid a safety injection signal.

As can be seen in figure 2.11.2, between hours 2 and 3, the cooldown rate increased for a period of 15 minutes at

.I greater than 25 F per hour. Charging was increased to maintain pressurizer level greater than 20% indicated level.

The cooldown was temporarily secured until parameters stabilized and again initiated after 30 minutes of stabi-lization. During the first hour, pressurizer level and pressure increased due to manual control of pressurizer heaters and charging. Pressurizer pressure was allowed to slowly decrease during the cooldown to maintain the pressure /

temperature limitations.

I 109 D ( 2) : 0 3 I

I Figuro 2.11.1 I BORON MIXING & COOLDOWN HOURS 0 2 4 6 3

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