ML20197H341
| ML20197H341 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/24/1986 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 0560A, 5211-86-2070, 560A, NUDOCS 8605190142 | |
| Download: ML20197H341 (99) | |
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- STARTUP REPORT FOR RESTART '
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, TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
1.1 INTRODUCTION
1.2
SUMMARY
i 1.2.1 GENERAL 1.2.2 HOT FUNCTIONAL TESTING I 1.2.3 POWER ESCALATION TESTING 2.0 HOT FUNCTIONAL TESTING 3.0 LOW POWER PHYSICS TESTING 3.1 CONTROLLING PROCEDURE FOR LOW POWER PHYSICS TESTING 3.2 LOW POWER NATURAL CIRCULATION TESTING 4.0 POWER ESCALATION 4.1 CONTROLLING PROCEDURE FOR POWER ESCALATION 1
- 4.2 REACTOR TRIP ON LOSS OF FEEDWATER/ TURBINE TRIP i 4.3 THERMAL EXPANSION CHECKS FOR PIPING HANGERS AND SUPPORTS 1
- 4.4 UNIT LOAD STEADY STATE TEST
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4.S RCS OVERC00 LING TEST '
4.6
- ,. EMERGENCY FEEDWATER PUMP AUTO-START TEST 4.7 FEEDWATER SYSTEM OPERATION AND TUNING 4.8 INCORE THERMOCOUPLE TEST 4.9 INTEGRATED CONTROL SYSTEM TUNING 4.10 TUREINE GENERATOR OPERATIONAL TESTING 4.11 TURBINE BYPASS VALVE TESTING l i
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. 5.0 CORE PERFORMANCE TESTING 5.1 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER -
SUMMARY
5.2 CORE PERFORMANCE - MEASUREMENTS AT POWER -
SUMMARY
5.3 CORE PERFORMANCE - MEASUREMENTS AT ZER0 POWER 5.3.1 INITIAL CRITICALITY 5.3.2 NUCLEAR INSTRUMENTATION OVERLAP 5.3.3 REACTIMETER CHECK 0UT 5.3.4 ARO CRITICAL BORON CONCENTRATION .
5.3.5 TEMPERATURE COEFFICIENT MEASUREMENTS 5.3.6 CONTROL ROD GROUP WORTH MEASUREMENTS 5.3.7 DIFFERENTIAL BORON WORTH 5.3.8 SHUTDOWN MARGIN 5.3.9 EJECTED CONTROL R0D WORTH 5.4 CORE PERFORMANCE - MEASUREMENTS AT POWER 5.4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER 5.4.2 INCORE DETECTOR TESTING 5.4.3 POWER IMBALANCE DETECTOR CORRELATION TEST 5.4.4 CORE POWER DISTRIBUTION VERIFICATION 5.4.5 REACTIVITY COEFFICIENTS AT POWER t
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1.0 , INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
Following the shutdown order for TMI-l in August 1979, a detailed restart e test program was implemented to insure a safe and reliable return to i operation. The purpose of the program was to accomplish the following:
- (1) To provide a deliberate, methodical well planned verification of
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proper modification installation and performance in accordance with design.
! (2) Determination of plant transient response characteristics and i verification of acceptable integrated plant operation with modified systems / components.
(3) Verification of acceptable system readiness and plant operation with new modified plant operating, surveillance, emergency, abnormal and maintenance procedures.
I (4) Vertftcation of the adequacy of the OTSG Tube Repair Program by operational leak testing and on-line monitoring throughout the test program, i
(5) Performance of sufficient modified system / plant steady state and l transient operations to provide operator training and familiarization i with modified system / plant response throughout a range that they are j likely to experience during the design life of the plant.
i Hot Functional Testing (HFT) was conducted at various times during 1981,
- 1983, 1984, and 1985. Testing was conducted for operator training, OTSG j repair testing, and modification testing.
- Three Mile Island Nuclear Station Unit I was granted permission for restart
} on October 02, 1985. Initial criticality was achieved on October 03, 1985 1 at 1330.
l Zero Power Physics testing was conducted between October 03 and October 06,
.! followed by Power Escalation to approximately 3% power for Natural I
Circulation testing with subsequent recovery to normal operation at 5% power. ,
! Further increases in reactor power were made as testing was successfully
- completed at each of the power plateaus defined in the. Power Escalation
. sequence. Steam leaks in Main Turbine steam line High Pressure (HP) drain i lines between 25% and 40% plateaus caused a several day delay during testing l while repairs were completed.
-l Power Escalation continued until 40% power, where the reactor was tripped by tripping the Feedwater pump for the Loss of Feedwater Trip Test. Following
- the trip, the test program was reviewed and authorization was given to
- proceed to 48% power.
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I After, completing the 48% power testing a thirty (30) day training period was conducted. Following the training period, Power Escalation and testing continued until the Plant was raised to 75% power. Upon ccmpletion of the testing at.this plateau a second thirty (30) day training period was conducted. ;
While in the training phase at 75% power a Reactor Trip occurred due to a malfunction of the Turbine Generator over excitation relay. Repairs were completed and the Plant was returned to 75% power. At the completion of the training period the Plant was escalated to approximately 88% power. The Plant was limited to a maximum of 88% power due to high OTSG level restrictions caused by deposits on the OTSG secondary side. It was decided to perform all full power testing and data collection at that power level and then perform the Turbine Generator Trip Test to complete the power ascension test program.
During recovery from the 88% power trip and an unplanned low power Reac Mr Trip which occurred during return to power, the Steam Generators were found to no longer limit Reactor Power due to OTSG level. The Plant was escalated to approximately 1007. power where some testing and data collection was performed for this final plateau. At this point the Plant was turned over to Operations and the Power Escalation Testing program was considered completed on January 20, 1986.
1.2
SUMMARY
1.2.1 General Hot Functional testing at Three Mlle Island Unit I consisted of several phases conducted over a period of four years, followed by the restart of the Unit in October 1985.
The Unit has been operated at power levels up to and including 100% Power i during the conduct of Startup testing. In general, the performance of the Unit has been satisfactory, with an evaluation of the test program results concluding that the Unit can be safely operated at 100% Power.
A chronological sequence of the overall TMI-1 Restart Test Program beginning with August 1981 testing is given in Figure 1-1.2. A power history of the Power Escalation Test Program is given in Figures 2-1.2 through 5-1.2.
A summary of the test results addressed by the major sections of this report is given in the following paragraphs.
1.2.2 Hot Functional Testing Several series of HFTs were performed between August 1981 and October 1985, the following is a brief description of these HFT programs.
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(1) . August 1981 - Preliminary Heatup Hot Functional Testing (TP 600/1). A
. plant Heatup was conducted in steps, to 532*F, followed by a cooldown and return to cold shutdown. During this testing, various Plant surveillance procedures and miscellaneoJs modification testing occurred. Testing duration was approxi nately two weeks.
. (2) August 1983 - A series of plant heatups/cooldowns were conducted,_per TP 600/2 after repairs to Once Through Steam Generators were completed, to test effectiveness of repairs using Kr-85. During this testing, various plant surveillance procedures were performed, some data collection / testing was completed per enclosures of Controlling Procedure for Low Power Physics Testing (TP 700/l), and some modification testing was performed. Testing duration was
- approximately six weeks.
l (3) May 1984 - A plant heatup/cooldown was conducted per Controlling Procedure for Low Power Physics Testing (TP 700/1). During this
. testing various plant surveillance procedures, High Pressure Injection Functional Test, other miscellaneous tests, and Reactor Coolant Pump
'B' Shaft Replacement Tests were performed. In addition, data collection / testing was conducted per enclosures of TP 700/1. Testing j duration was approximately one week.
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! (4) April 1985 - A plant heatup/cooldown was performed per TP 700/1 for Krypton Leak Rate testing of Steam Generators. In addition to leak rate testing various plant surveillance tests were performed. The test duration was approximately three days.
(5) June 1985 - A plant heatup was performed in accordance with Plant procedures in anticipation of permission for Startup, following a favorable sourt ruling. The Plant was maintained in Hot Shutdown conditions for Operato' training while awaiting lifting of the court 4
issued rc:V of the 05s.1/85 NRC Restart Order.
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T MI-1 RESTART TEST PROG R A M IIOT RCS
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- AW i TO SUPPORT TP 600/2 TP 700/1 TP 700/1 TP 700/1 +B HEATUP (AUG-0CT 83) (MAY 84) (APR 85) (OCT 85) l l
l NATURAL POWER RETURN COMPLETION CIRCULATION ESCALATION TO 100% OF TESTING TESTING POUER y Bb TP 700/2 TP 800/1 TEST PROGRAM y (JAN 6, 86) (JAN 20, 86) l (OCT 85) (OCT 85 - h JAN 86)
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OVERSPEED TRIP TEST & DRAIN LINE REPAIR FURMANITE REPAIR OF 4 COMPONENTS
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1.2.3 Power Escalation Testing A summhry of the test results addressed by the major sections of this report is given in the following paragraphs.
(1) Reactor Trip on Loss of Feedwater/ Turbine Trip Test (TP 800/2) 4 The Loss of Feedwater Trip was conducted at approximately 407. power, and the Turbine Trip was conducted at approximately 87% power. On both trips the following results were observed i (a) The Reactor tripped prior to RCS pressure of 2300 psig on the i
1 anticipatory LOFW/ Reactor or Turbine / Reactor trip.
- (b) Containment isolation occurred.
(c) The main steam safety valves actuated and the turbine bypass valve setpoint changed to 1010 : 10 psig, but the bypass valves cculd not control at setpoint due to main steam safety valve actuation. Once the setpoint was slightly lowered to allow the main steam safety valves to seat, the turbine bypass valves controlled header pressure. .
(d) OTSG levels were controlled at 30 +2, -10 inches by EFW (40%) or MFW (87%).
On the Lost of Feedwater Trip, also observed that the EFW pumps 1
automatically started within allowed tolerances.
The acceptance criteria for both trips was met and the main steam safety valve performance, although still being evaluated, was considered appropriate for continued safe plant operation at all power levels.
(2) Thermal Expansion Checks for, Piping Hangers and Supports (TP 800/3)
This test monitored applicable system hangers / supports during both Plant Heatup (HFT) and PET at various temperature and Reactor power plateaus.
These hangers were adjusted, repaired, or reworked during this test period, with the results that all hangers / supports monitored were considered acceptable for continued plant operation.
(3) Unit Load Steady State Test (TP 800/5)
The Unit Load Steady State Test monitored the RCS and OTSG Steady State Operation at various power levels. Most parameters were within expected ranges at all power levels, and of those that were not, the only important carameters outside the expected ranges were those of 0TSG 1evels at cower levels of 75% and 87.57.. The OTSG high level limit restricted Reactor power to 87.5% prior to Reactor trip. Upon recovery i
from the Reactor trip at 87.5% power, the Steam Generator level was found to no longer limit power and power was escalated to 1007. power, all data taken at 1007. was within expected boundaries.
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(4) RCS Overcooling Test (TP 800/8)
Natural circulation was established with EFW flow maintained as continuous as possible, then OTSG level was increased until 0TSG pressure decreased to 750 psig. The Plant was then stabilized and returned to forced flow and normal Hot Shutdown conditions. Although the natural circulation portion of this test was terminated prior to reaching desired OTSG 1evel, the major objectives of the test were met and the test considered acceptable.
(5) EFW Pump Auto-Start Test (TP 800/9)
This test verified the two motor driven EFW pumps and the turbine driven EFW pump all started on loss of all RC pumps. All acceptance criteria,
- for this test, were met.
(6) Feedwater System Operation and Tuning (TP 836/1)
The feedwater heater and drain tank level controls functioned properly i
and were fine tuned during this test. Evaluation of the computer heat balances and comparison with GE heat balances indicat6d feedwater heat exchanger performance within expected tolerances. No major discrepancies were found and the performance monitored was considered acceptable.
I (7) Incore Thermocouple Test (TP 846/1)
The Incore Thermocouple Test verified the alarm for 100*F subcooled margin at 340*F and 525'F plateaus, it also verified the agreement of the symmetrical thermocouples to within i 1% of the average for that grouo, and the thermocouples agreed within 3 2% with the calculated values based on core power distribution at various power levels. The values for the Backup Incore Thermocouple Readout (BIRO) display and the Plant Comouter Operator's Group agreed within allowed t 16*F span.
The five highest incore thermocouples as selected by computer program did not correspond with the five highest readings from the computer grouas, except at 1007. power level. This difference was due to small differences in incore thermocouple readings and time differences between collection of data. These differences mean that small changes in flux or flow mixing could result in a different selection of five thermocouple! as highest for either means of data collection. The average value of the five highest incore thermocouples by each method was always within i 2*F '
of the other averages and computer calculated value. Also, the computed saturation margin was within 5'F of the computer calculated value at each power level.
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(8) Integrated Control System Tuning (TP 849/1)
The ICS subloop controllers were adjusted for optimum performance, the feed forward data was recorded and adjustments made as necessary for stable plant conditions, and the ICS was tested in various modes of control at several power levels and adjusted as necessary for optimum performance.
' The plant demonstrated stable ICS control during both steady state conditions and transient conditions. The transient conditions consisted of a 2% per minute power ramp change from 48% to 40% power and back to 48% power; a CRD runback from 75% power, during which the plant ran back to 55% full load; and two main FW pump trips, one at 75% power and the other at 87% power, during which the plant ran back to 60% full load design feedwater flow.
(9) Turbine Generator Ooerational Testing (TP 885/1)
Data was collected at thermal power levels of 5%, 15%, 25%, 40%, 60%,
75%, 87.5% and 100% rated power. All test data fell within allowable limits except at 100% power, where two thermocouples slightly exceeded j the acceptance criteria of 1 6*F of average value of all readings. The values were evaluated and determined not to be significant enough to
- cause concern, based on prior data and values of corresponding RTD's and l adjacent thermocouples. Two thermocouples and two RTO's were found to j be out of service, this was determined to have no significant effect on test results since there are 72 of each and the remaining instruments were performing correctly.
(10) Turbine Bypass Valve Testing (TP 885/2)
The Turbine Bypass valve response to setpoint and changes in setpoint was found to be satisfactory. The valves controlled header pressure at setpoint. During both Reactor Trips, the Turbine Bypass valve opening time could not be observed, since they never received a control signal high enough to open them to full open, due to Main Steam Safety valve actuation. The peak pressures were recorded and the control setpoint was verified to change by +125 psi following Reactor Trip.
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2.0 HOT FUNCTIONAL 'ESTING 2.1 PURPOSE Several Hot Functional Tests (HFT) were performed between August 1981 and October 1983, the following is a description of the purposes of each HFT Program:
(1) Preliminary Heatup Hot Functional Test Program (TP 600/1) - This program covered test and event performance beginning with steam bubble formation in the pressurizer, continued through Plant Heatup to 532*,
2155 psig and ended with Plant Cooldown. During this program Plant Operating and Surveillanca Procedures were verified and functional testing of plant modifications was completed. These functional tests included:
(1) High Pressure Injection System Functional Test - TP 655/1 (2) Olesel Generator Load test - TP 622/1 (3) PORV Flow Indication Functional Test - TP 664/1 (4) Tsat Functional Test - TP 645/1 (2) OTSG Repair Test program (TP 600/2) - This program was used to test the effectiveness of the OTSG Repair Program, to perform plant
- 4. survelliance procedures and to perform some functional testing of plant modifications which included TP 675/1 RCS High Point Vent (Pressurizer) and TP 664/1 PORV Flow Indication functional tests.
2.2 TEST METHOD The Preliminary Heatup HFT program established a controlling logic secuence for test and event performance for the entire HFT program, from Cold Shutdown through Plant Heatup and back to Cold Shutdown. This program established the various test plateaus and the tests to be performed at each plateau. All .
plant and functional testing was performed pe- the appropriate procedures at I each of the test plateaus.
The OTSG Repair Test program established a controlling logic for the sequence of events beginning with the Reactor Coolant System H 02 2 cleanup (TP 600/4) through several Plant heatup/cooldown phases and back to Cold Shutdown. Any plant and functional test procedures required were performed as designated by the test sequence according to the appropriate procedure.
2.3 TEST RESULTS Functional test results of modifications tested during HFT are included as part of those test procedures. A summary of significant test results follows:
(1) The T sat Functional Test was performed at various temoerature plateaus during plant heatuo. Data was recoraed and comparisons were maae between calculated ana indicated values. Test was comoleted satisfactorily.
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(2)
The Diesel Generator Load test was performed to ensure the engineered i safeguards system had the capacity and capability to accept the loads in the required time sequence. A review of the data indicated satisfactory results.
(3)
The PORV Flow Indication Functional Test was performed to verify proper operation of the valve, associated acoustic monitors, differential pressure indicators, and tall pipe temperatures. System performance during TP 600/1 was satisfactory, however, system instrument responses were questionable.
This test was repeated during TP 600/2 and valve position instrument responses were still a problem. Also computer data collection problens occurred. Test results were still unsatisfactory although PORV performance was acceptable. (This was later corrected as explainea in Section 3.1.3)
(4) The High Pressure Injection (HPI) System functional Test was performed to verify injection flow rates through HPI lines met estaullsned criteria after installation of the HPI cavitating venturi. This test experienced problems with test flow instrumentation at RCS pres:ure of 550 psig and could not be satisfactorily completed at this time.
The OTSG Repair test consisted of leakrate monitoring during 37 days at normal operating pressure and temperature and three separate cooldowns. This monitoring resulted in a measured primary to secondary leakage (in OTSG) during steady state periods at a relatively constant 1.0 GPH. The measured OTSG Primary to Secondary leakage increased, as expected, during the cooldowns where the peaks were 1.7 GPH and 2.6 GPH.
2.4 CONCLUSION
S Plant operation during HFT was monitored at various temperature and pressure plateaus during both Heatups and Cooldowns. All measured parameters and equipment operation fell within their respective maximum / minimum boundaries, or they were evaluated and a determination was made that the variation from the predicted range did not have an adverse effect on the safe operation or control of the unit. Unsatisfactory tests were scheduled to be repeated at a later date. Two tests scheduled to be repeated were High Pressure Injection '
System Functional Test and PORV Flow Indiction Functional Test, these were rescheduled for performance in TP 700/1, " Controlling Procedure for Low Power Physics Testing".
The OTSG Repair test resulted in establishing the administrative operating limit for OTSG Primary to Secondary leakrate as 7.0 GPH, based on a conservative baseline leakrate of 1.0 GPH, plus the allowable limit of 6.0 GPH above the baseline leakrate. This test also confirmed that the OTSG primary to secondary monitor leakrate monitoring method, using the off-gas radioactivity (RM-ASL), would provide the required sensitivity, trencing, and reliability with the higher off-gas activity from Xe-133 during power operation.
For further discussion see TDR 488.
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3.0 LOW POWER PHYSICS TESTING 3.1 CONTROLLING PROCEDURE FOR LOW POWER PHYSICS TESTING 3.1.1 PURPOSE l The purposes of the Controlling Procedure for Low Power Physics Testing (TP 700/1) were:
(1) To establish a controlling logic sequence for test and avent performance. The sequence included the initial steam bubble formation in the pressurizer through the Plant heatup and ended with achieving approximately 5% rated power. This sequence included completion of TP 700/2, Low Power Natural Circulation Test.
(2) To allow those systems from the Hot Functional Test (HFT) program which required retesting to be tested, and to provide pages for j recording test data on these retests.
l 3.1.2 TEST METHOD I i l The Controlling Procedure for Low Power Physics Testing (TP 700/1) established the various test plateaus and.the tests to be performed at each !
plateau. All plant and functional testing was performed per the appropriate j procedures at each test plateau.
This testing consisted of three phases, which were:
(1) A plant heatup/cooldown during which plant. surveillance procedures, High Pressure Injection (HPI) Functional Test (TP 655/1), PORV Flow Indication Functional Test (TP 664/1) and data collection / testing per enclosures of TP 700/1 were performed.
(2) A plant heatup/cooldown was performed for Krypton Leak Rate testing of the Once Through Steam Generators.
(3) A plant heatup to Hot Shutdown conditions for operator training and Low Power Physics Testing was completed. This procedure controlled-j the sequence of events from Hot Shutdown, through criticality to 3% !
power operations and finally through the Low Power Natural Circulation Test (TP 700/2). The completion of TP 700/2 signified.the end of the Low Power Physics testing.
4 4
, 14
3.1.3 RESULTS Functional test results of modifications tested during the Low Power Physics testing sequence are included as part of those test procedures. All results were within acceptable limits or evaluated and discrepancies resolved.
Results for specific procedures are as follows:
(1) During retest of TP 664/1 the PORV acoustic monitors, differential pressure indicators, and tail pipe temperatures were all verified to respond adequately to provide proper indication of PORV operation. The test was completed satisfactorily.
(2) The retest of TP 655/1 provided modified instrumentation for measuring injection flow rates. All test data accumulated was evaluated and analyzed. Based on the analysis it was concluded that the objectives of the HPI test were satisfied and the intent of all the required acceptance criteria was met.
(3) The Krypton Leak Rate testing of the Steam Generators resulted with an OTSG Primary to secondary leakrate, during steady state Hot Shutdown conditions, of approximately 0.5 gallon per hour (GPH).
(4) Per the controlling procedure (TP 700/1), at the completion of all testing to be performed at 525'F temperature plateau, and after permission was granted for restart, the Plant proceeded to initial criticality on October 03, 1985. After criticality was achieved and the Zero Power Physics testing was performed the Plant was escalated to 4
approximately 3% power. Following the completion of prerequisites, the Low Power Natural Circulation Test (TP 700/2) was performed and TP 700/1 testing was complete. The controlling document then became Controlling Procedure for Power Escalatica Testing (TP 800/1). See Section 3.2 for further discussion of TP 700/2, Low Power Natural ~ Circulation Test.
3.
1.4 CONCLUSION
S Plant operation during testing controlled by the Controlling Procedure for Low Power Physics Testing (TP 700/1) was monitored at various temperature / pressure plateaus during heatups and cooldowns, and during initial criticality and 3%
power natural circulating testing. All measured parameters and equipment operation fell within their respective maximum / minimum boundaries, or they were evaluated and a determination was made that the variation from the predicted range did not have an adverse effect on the safe operation m control of the unit.
l Tests scheduled to be repeated from Hot Functional Testing were the High Pressure Injection System Functional Test and the PORV Flow Indication Functional Test. These tests were both completed with satisfactory results.
The Krypton Leak Rate testing, wnich established an OTSG primary to secondary leakage as approximately 0.5 GPH, provides the basis for establishing the value for the OTSG primary to secondary " baseline" leakrate as 0.5 GPH. For further discussion see TOR 691.
1 I
l
- 15
3.2 - LOW POWER NATURAL CIRCULATION TEST 3.2.1 PURPOSE The purpose of the Low Power Natural Circulation Test (TP 700/2) was to verify the ability of the Integrated Control System (ICS) to maintain preset Once Through Steam Generator (OTSG) levels under loss of main feedwater and natural circulation conditions. This test also verified proper response of the Emergency Feedwater System as well as the establishment and maintenance of natural circulation under varying conditions.
Testing was conducted at approximately 37. rated thermal power, to simulate the decay heat load that would correspond to significant core burn up. Technical Specification 3.20.1 was in effect only for this test and provided for
, suspension of various RPS trip set points to allow Natural Circulation testing at 37. power. Technical Specification 3.20.1 also imposed new limitations for testing, which were incorocrated into TP 700/2 and were only in effect for this test. The data collected during the Natural Circulation testing was used in verifying adequate core cooling in the event of loss of main feedwater and/or reactor coolant flow, and providing operator experience and training for those types of transients.
Specific purposes were to:
(1) Demonstrate that both motor driven and the turbine driven Emergency Feedwater (EFW) pumps actuate and reach full flow conditions (discharge pressure 1010 psig) within specified times, upon loss of both Main Feedwater (MFW) pumps.
(2) Verify that the ICS will maintain preset OTSG water levels under various Plant conditions and casualties.
(3) Demonstrate that the operator can manually control OTSG water level in l the event of loss of ICS control.
(4) Allow the operators to observe and participate in a smooth transition from forced circulation to natural circulation and back to forced ;
circulation.
(5) Demon!trate that the operators can maintain a minimum of a 50*F subcooled margin in the Reactor Coolant System (RCS).
(6) Determine the effect on natural circulation flow of lowering 0TSG 1evel.
(7) Determine the effect of reduced cold leg temperature on indicated neutron flux in order to produce a correction factor which can be applied to indicated power while on natural circulation.
(8) Verify that the pressurizer heaters assigned to one emergency bus are !
sufficient to stabilize RCS pressure.
(9) Verify that the bottled air supply is capable of supplying air to valves EF-V30A/8, MS-V6, and MS-V4A/B for a minimum'of two (2) hours following the loss of normal and backup instrument air.
16
- -g - -- y. -, --
3.2.2 TEST METHOD This test was conducted with the reactor critical at approximately 3% power and at near equilibrium xenon, with the Plant operating normally and the Steam Generators being fed by Main Feedwater (MFW).
(1) The first part of the test was to determine the Reactor power correction factor. This factor was then used to calculate the revised hign flux trip setpoint, to insure a conservative trip setpoint in regard to Reactor Vessel downcomer temperature effects. The correction factor was '
found by performing a heat balance and collecting data starting at Tave -
541*F and reducing OTSG pressure to reduce T until Tcold-was approximately 526*F. cold in 5'F increments The heat balance and data was then used to determine a correction factor for each T qold level and this value was plotted; using this data a correction ractor was obtained for 526*F and the high flux trip setpoint was adjusted based on this value.
(2)
The second part of the test was to trip the operating MFW oump and verify that bcth motor driven pumps and the turbine driven pump actuated and achieved full flow conditions within specified times, and to verify that OTSG levels were controlled at 30 +2, -10 inches while the Steam Generators were fed by the Emergency Feedwater System. The motor driven EFW pumps were secured, then the normal air supply (both Instrument Air (IA) and Backup Air) to valves EF-V30A/8, MS-V4A/B and MS-V6 was isolated and these valves were cycled with load. This verified that the bottled air supply maintained a pressure 250 psig for at least two (2) hours.
The test then returned the air supplies to normal, started the motor driven EFW pumps and secured the turbine driven EFH pumps.
(3) The last part of the test established the conditions for natural circulation and then tripped all-four Reactor Coolant pumps to go into natural circulation. After natural circulation was established the operators placed the EF V30A/B controls in manual and fed the Steam Generators up to approximat'erly 50% on operating range, without overcooling the RCS. Then OTSG level control was placed in automatic and level was verified to control at 50 + 5% on operating range. Following 30 minutes of stable natural circulation flow, the pressurizer heaters were deenergized by placing the Control Room switches in the 'off' position, Group 7 and 9 breakers at 18 pressurizer heater MCC were opened and Group 8 breakers were verified closed. RCS pressure was allowed to decrease data. Group and the system was monitored every fifteen minutes by collecting 8 was then energized on the normal bus by placing the bank 4 control switch to "AUT0". After a fifteen minute wait data was collected ~
to verify RCS pressure stopped decreasing or began increasing. Grouc 7 and 9 breakers were then closed and the RCS pressure restored to 2155 t 50 psig. Recovery from natural circulation was then performed and the Plant was stabilized in preparation for reestablishing natural circulation. Natural circulation was established a second time by tripping the RC pumps, after steady state conditions were established the the OTSG Levels were decreased to 40% + 5% on operating range, levels were stabilized and data collected, then an attempt was made to lower '
OTSG level to 30% 1 5%. Asymmetrical cold leg temperatures develooed at '
approximately 33% on operating range and this portion of testing was terminated and forced flow was restored, EFW was secured, and Reactor Plant was restored to normal. !
i
(4) During both times that natural circulation was established and the plant was restored to forced circulation, the Operations and Training, Departments participated in allowing all the operators to observe and participate in the evolution.
3.2.3 RESULTS (1) The correction factor for Tcold of 526*F was determined to be 1.29 which resulted in an adjusted high flux trip setpoint of 5.4%.
(2) When the MFW pump was tripped the motor driven EFW pumps were verified to start and reach full flow conditions (discharge pressure 1010 psig) within three (3) seconds of receiving the actuation signal. The turbine driven EFH pump was verified to start and reach full flow conditions within 15 seconds of receiving the actuation signal. While the turbine driven EFH oump was in operation the OTSG levels were verified to control at 30 +2, -10 inches and the bottled air supoly was verified to maintain a pressure 250 psig for at least two (2) hours. The pressure dropped from 1860 psig to 1100 psig for the 'A' bank, and from 1880 psig to 1020 psig for the 'B' bank.
(3) Prior to starting natural circulation, the EFW control valves EF-V30A/B were placed in manual at the alternate feedwater station and it was verified that the operator could control OTSG water level manually. It was also verified that using manual control of pressurizer heaters, the operator can maintain RCS pressure.
Natural Circulation was initiated by securing all RC pumps, and after the Plant stabilized it was verified that Group 8 pressurizer heaters were sufficient to stabilize RCS pressure. For the two hours the Group 7, 8, 9 heaters were deenergized the depressurization rate was 71 psig per hour. The pressurization rate with Group 7 and 9 deenergized and Group 8 energized was :
36 psig per hour. Following this test Operations performed a recovery from ;
natural circulation and a return to natural circulation. After natural ;
circulation stabilized, the RCS was tested for the effect of lowering the OTSG level on natural circulation flow and found that the level can be lowered to 40% 2 5% in the operating range. During lowering of the level to 30% z 5%
asymmetric cold leg temperatures developed and testing was terminated and forced flow restored. Throughout the Natural Circulation testing the operators were able to maintain a minimum of 50*F subcooled margin in the RCS.
3.
2.4 CONCLUSION
S The primary purposes of this test were to train operators on Natural Circulation, to verify EFW flow and flow control, and to demonstrate the natural circulation capability of the Plant. Our conclusions are that this test has met those objectives. Specific conclusions for testing performed are as follows:
(1) Both the motor driven and turbine driven EFW pumps actuate and reach full flow conditions within required time limits.
18
(2) The ICS will maintain preset OTSG levels under various plant cend!ticas and casualties and the operators can manually control OTSG water level in the event.of loss of ICS control.
(3) The operators can make a smooth transition from forced circulatten to
- natural circulation and back to forced circulation, and the caeratcrs can maintain a minimum of a 50*F subccoled margin in the RCS.
! (4)
' The pressurl:er heaters assigned to one emergency bus are. sufficient to stabilize RCS pressure.
(5) The bottled air sucaly supplying valves EF-V30A/S, MS-V6. and MS 'ilA/3 is cacable of sucplying air to these valves for a minimum or, two (2) hcurs i- following the loss of normal and backup air. ,
(6) The test demonstrated that stable natural circulaticn occurs at CT3G levels less than 507. in the ocerating range. The test also sncus: :nat at acproximately 337. Incicatea level, asymmetric colo leg temcera:ures i
deveicced, with the cold leg temperature at the 3 RC? inlet crccoing # rem approximately 539'F to acproximately 518'F in acout ten (10) minutas. '
Natural circulaticn ficw, nowever, was maintainea in tne otner tnree Reactor coolant legs.
i i
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19.
l 4.0 POWER ESCALATION 4.1 CONTROLLING PROCEDURE FOR POWER ESCALATION 4.1.1 PURPOSE The purpose of the Controlling Procedure for Power Escalation was to establish a controlling logic sequence for test and event performance during Plant Power Escalation.
This procedure began with the completion of the Controlling Procedure for low Power Physics Testing (TP 700/1), defined the sequence of events to 100% power, and ended with the completion of Emergency Feedwater Pumo Auto-Start Test (TP 800/9). The intent of this procedure was to insure an orderly sequence of events in which quality testing and training could be achieved.
4.1.2 TEST METHOD This test procedure detailed the test seauence for Power Escalation Testing.
The procedure started at the completion of Low Power Physics Testing at approximately 5% power and continued through the major power test plateaus of 15%, 25%, 40%, 48% and 75% to 100% power. The procedure cetailed the tests to be performed during Power Escalation and in the major power test plateaus. It contained 30 day training periods at 48% and 75% power and authorization signoffs at various power test plateaus for proceeding to the next power plateau.
All plant and functional testing was performed per the appropriate procedures at the designated time or power level.
4.1.3 RESULTS Test results of the procedures run as a part of this test sequence are included as part of those individual procedures. Testing originally designated to be performed at 90%, 95%, and 100% power levels was performed at approximately 88% power.
This testing was performed at 88%r power versus at the designated levels operatingReactor because range). Power was limited by OTSG high level limits (approximately 92%
After performance of the Reactor Trip on Turbine Trip test at approximately 88% power, the Plant achieved 100% power and all required data for this power level was recorded. After completion of the test results i review, the Power Escalation Testing Program was considered completed satisfactorily on January 20, 1986.
4.
1.4 CONCLUSION
S Plant operation during the Power Escalation Test Program was monitored and evaluated at various power olateaus and during various upset conditions. All measurec parameters and eculpment operation fell within their respective maximum / minimum boundaries, or were evaluated and a determination was made that the variation frcm the credicted range did not have an adverse effect on the safe operation or control of the unit.
20
a Operator training was performed for two 30 day periods, one during operation at 48% power and the other during operation at 75% power.
The OTSG fouling was evaluated and it was decided to perform testing required for 90%, 95%, and 100% power levels at the maximum achievable power (approximately 88%).
This testing included the Reactor Trip on Turbine Trip Test. Following ccmpletion of testing after the Reactor Trip the Plant was restarted and found to be capable of achieving 100% power. Upon reaching 100%
power all required testing for this level was completed and the testing considered satisfactorily completed January 20, 1986.
21
4.2 REACTOR TRIP ON LOSS OF FEE 0 HATER / TURBINE TRIP TEST 4.2.1 PURPOSE The Reactor Trip on Loss of Feedwater/ Turbine Trip Test (TP 800/2) had two separate sections, one performed at approximately 40% power and the other performed at approximately 88% power. The purposes of this test were:
(1) At a power level of approximately 40%, both main feedwater pumps were tripped with the following to be verified:
(a) The Reactor trips on the anticipating LOFW/ Reactor Trip prior to the Reactor Coolant System (RCS) reaching 2300 psig.
(b) The turbine bypass valve setpoint is transferred to 1010 t 10 psig following the Reactor trip.
(c) The turbine trips coincident with the feedwater pump trip.
(d) Containment isolation on Reactor Trip and reset features function as designad.
(e) OTSG 1evels are controlled at 30 +2, -10 inches by E:nergency Feedwater (EFH).
(f) All thres EFW pumps start automatically.
(2) At a power level of approximately 88%, the turbine was tripped with the following to be verified:
(a) The Reactor trips on the anticipatory Turbine / Reactor trip prior to the RCS reading 2300 psig.
(b) The turbine bypass valve setpoint is transferred to 1010 t 10 psig following the Reactof trip.
1 (c) Containment isolation on Reactor trip and reset features function as designed.
(d) OTSG levels are controlled at 30 +2, -10 inches by Main Feedwater (MFH).
(3)
To verify proper operation of the acoustic monitor in the event any Mal.n Steam Safety Valves are actuated during both Reactor trips.
(4) To prepare plant conditions for the RCS Overcooling Control Test (TP 800/8) and the Emergency Feecwater Pumo Auto-Start Test (TP 800/9), :
which are done in conjunction with this test and also required the !
Reactor to be tripped. l 22 :
4.2.2 TEST METHOD At approximately 40% power the Main Feedwater (FW) pumos were tripoed from the Control Room on Console 'CL' and Abnormal Transient Procedure (ATP) 1210-1 was followed in parallel with the test procedure. Upon the FN oump trip, the Reactor and turbine were verified to trip and containment isolation verified to occur. The Emergency Feeawater (EFW) pumos were verified to start and control 0TSG level at 30 +2, -10 inches. The turbine bypass valve serpoint change and main steam safety valve actuation was also monitored. After following procedures ATP 1210-1 (Reactor / Turbine Trip) and 1210 4 (Lack of Primary to Seconcary Heat Transfer) to place the Plant in a normal, post trio, hot shutdown conaition. the Plant proceeded to testing per RCS Overcooling Control Test (TP 800/8) procecure.
At approximately 88% cower the main turbine was tricped from the Control Room on Console 'CL' and ATP 1210-1 was followed in parallel with the test procecure. The Reactor trip was verified to occur upon turoine trio, with containment isolation occurring, and MFW controlling 0TSG 1evel at 30 +2, -10 inches. The turbine bypass valve setootnt change and main steam safety valve actuation was also monitored. After following ATP 1210-1 to place.the Plant in a normal, post-trip, hot shutdown condition, the Plant proceeded to testing per Emergency Feedwater Pump Auto-Start Test (TP 800/9) procedure.
4.2.3 RESULTS At approximately 40% power, both main feedwater pumps were tripped with the following results:
(1) The Reactor tripped with a peak RCS pressure of 2155 psig.
(2) A containment isolation occurred on Reactor trip.
(3) The main turbine tripped coincident with-the feedwater pump trip.
(4) The motor driven EFW pumps reached full discharge pressure 1.8 sec (EF-P2A) and 2.0 sec. (EF-P28) after receiving actuation signals.
(5) The turbine driven EFW pump reached full discharge cressure 20.0 sec.
after receiving actuation signal.
(6) The EFW System controlled OTSG level at 30 +2, -10 inches.
(7) The main steam safety valves actuated to relieve pressure and acoustic.
monitors resconced.
(8) The turbine bypass valves did not control turbine header pressure at 1010 t 10 osig cue to main steam safety valve ' actuation. The controi setcoint aid cnange following the Reactor trip, anc wnen the setcoint was lowered to allow all the main steam safety valves .to seat. the turoine oypass valves aid control at that lowerec setpoint.
23
At approximately 88% power, the main turbine was tripped with the following results:
(1) The Reactor tripped on the anticipatory Turbine / Reactor trip with RCS peak pressure of 2205 psig.
(2) A containment isolation, on Reactor trip, occurred; valve MU-V3 did not close.
(3) The MFW System controlled OTSG level at 30 +2, -10 inches.
- (4) The main steam safety valves actuated to relieve pressure and the acoustic monitors responded.
(5) The turbine bypass valves did not control turbine heacer pressure at 1010 : 10 psig, due to main steam safety valve actuation. The control setcoint did change following the Reactor tria, ana wnen the set oint was lowereo to allow all the main steam safety valves to seat, the turoine bypass valves did control at that lowerea setpoint.
j 4.
2.4 CONCLUSION
S
-r The performance of all systems was determined to be satisfactory for continued plant oceration.
The Plant resconded to both the loss of feedwater and the Turbine Trip with a corresponcing Reactor trip at a considerably lower RCS pressure than the 2300 osig maximum. In both cases there was a containment isolation and OTSG levels were controllea at 30 +2, -10 inches. In the case of the loss of Feeawater, EFW pumps startea within required times (15 sec. for motor driven and 40 sec. for turbine driven) of actuation signal and controlled 0TCG level. On the turbine trip the main feedwater controlled level. Each time, the turbine bypass valve setpoints changea to 1010 1 10 psig, but due to the main steam safety valve actuation, could not control pressure at that point. After lowering turbine header pressure to reseat main steam safety valves the turbine bypass valves controlled pressure at the lowered setpoint.
Procedural changes were made to include guidance to reduce main steam safety valve relifts by using the turbine bypass valves during unplanned trios. This guidance will permit reducing the setting of the turbine bypass valves after initial MSSV blowdown. Main steam safety valve performance will be discussed in a later report.
The non closure of MU-V-3 on containment isolation was investigated, resulting in replacement of a timer and a retest which verified closure on a Reactor trip signal.
The reset satisfactory results. timer function (10 sec.) was also verified with See following figures 1-4.2 througn 7-4.2 for graphic illustration of plant resconse.
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24
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4 THERMAL EXPANSION CHECKS FOR PIPING HANGERS AND SUPPORTS 4.3.1 PURPOSE The purposes of the Thermal Expansion Checks for Hangers and Supports (TP 800/3) were:
(1) To perform thermal checks of:
(a) Those variable / constant supports (springs, countercoises and snubbers) on pipe systems which have been added or modifiea suoseguent to the original Plant Startup Hot Functional testing.
(b) Those variable / constant supports (springs, countercolses anc snuubers) on olpe systems wnich have not been modified subsecuent to the original Plant Startup Hot Functional testing, but may have been affected by those wnicn have been acdeo or mocifiea.
(c) All main steam heatea cost supports (18).
(d) All supports listed in TP 600/3 inspec ad during Hot Functional Testing.
(e) Rigid supports, inspected as necessary during walkdown.
(2) To verify that the applicable system (s) piping can expanc without I obstruction auring heatup to normal operating conditions.
(3)
To perform system balance, and establish integrity of pipe supports in the areas of operability and functional performance.
(4)
To inspect piping systems and their respective supports which are considered ' cold', i.e., no thermal or terminal movements.
4.3.2 TEST METHOD This test was Escalation performed during both Hot Functional Testing phase and Power phase.
The test consisted of system walkdowns, which included:
thermal expansion checks, monitoring for unobstructed expansion during system heatup, adjustment as necessary, system balancing, and hanger / support rework / repair curing testing. The walkdowns were performed during Hot Functional testing at various RCS temperature plateaus, and during Power Escalation Testing (PET) at various Reactor power plateaus.
4.3.3 RESULTS This test monitorea aaplicaole system hangers /suoports curing coth Plant Heatuo (HFT) and PET at various temoerature and Reactor cower olateaus. The hangers /succorts monitorea were aajustec, repaired, or reworkea as necessary during this test period; also system calancing was performed on applicaole systems. As _ a result of this test all hangers /succorts monitorec are l
- ensidered acceptacle for continuea Plant operation. I 1
l
. 32
o 4.
3.4 CONCLUSION
S This test was conducted over a considerable time period, due to various system requirements and the necessity to monitor the hangers / supports from ' Cold' conditions to normal system operating conditions. During this time period all hanger / support adjustment, repair, rework, and balancing was performed. All deviations between expected and actual data are either within the allowed tolerancesfor acceptable or continued the deviations have been evaluated and determined to be Plant operation.
r t
t I
I
~
l
, , 33 x J
4.4 UNIT LOAD STEADY STATE TEST i
4.4.1 PURPOSE The purpose of the Unit load Steady State Test (TP 800/5) was to measure the Reactor Coolant System and the Steam Generator Steady State parameters as a function of Reactor Power. Specific purposes are as follows:
(a) Measurement of primary and secondary operating parameters from 5 to 100 percent power.
(b) At each power plateau, plot and extrapolate selected parameters to the next power plateau to avoid exceeding any safety or operating limits, and to verify correct alignment of the Integrated Control System (ICS).
(c) To provide baseline data for the " Feed Forward Curves" used in the alignment of the Integrated Control System.
4.4.2 TEST METHOD This test measured the steady state performances of the Reactor Coolant System (RCS) and the Once Through Steam Generator (OTSG) at various power levels.
3 The power levels used during the Power Escalation Test Program were 5,15, 25, 40, 60, 75, 87.5 and 100 percent power. At each of the above power levels, steady state conditions were established and data listed in Table 1-4.4 was taken at (5) minute increments for (30) minutes.
From this data the average unit parameters were calculated and plotted, see Fig. 1-4.4 through 7-4.4 comparisons between the measured data and the cesign curves were made to verify the acceptable operation of the Unit. These comparisons System (ICS) were performed to verify the ability of the Integrated Control to maintain steady state conditions, and to eliminate or minimize any system oscillations. Adjustments were made, as necessary to improve the unit steady state response.
1 Prior to Power Escalation, the steady state parameters were extrapolated to the next plateau to estimate the acceptability of the measured variables at the next plateau. If the extrapolation indicated either a substantial i deviation from expected levels, or exceeding a limit, the condition was evaluated prior to escalation to the next power level.
4.4.3 TEST RESULTS The unit average parameters measured during steady state testing are shown on Table 1-4.4.
Comparison between the average measured data and expected design curves was performed by plotting the average measured data against the expected design curves. All measured unit average parameters fell within i their respective min./ max. boundaries except as discussed as follows: '
34
Y (a) r - ' The average measured steam temperature was lower than expected value at 5% power.
This is in a major transition area of the graph, and could be
- expected to vary slightly, especially with slight variations in system
. within operation at this low power level. All data taken above this level was expected bounds.
~
(b)
The average RCS Tave was lower than the expected 579'F at 15%-power.
This is in a major transition area of the graph and, could be expected to vary slightly especially since the OTSG levels were higher than the approximately 25" level excocted by B&W to result in a Tave of 579* at 15% power. All data taken above this level was within expected bounds.
(c) The average OTSG Startup Range level was greater than maximum expected levels at 75% and 87.5% power and 0TSG Operating Range level was greater than maximum expected level at 87.5% power. These' levels were evaluated and the OTSG high level ~ limit was reset to 92% of Operating Range with B&W concurrence. The higher OTSG levels are due to decosit builduc on the secondary side of the Steam Generators and the olant was limited to 87.5% power with the 92% limit in Operating Range. After the Reactor Trip at 87.5% power and a subsequent low power trip on recovery, tne Steam. Generators were found to be no longer level limited and the power was escalated to 100% power. Data taken at 100% power was within expected bounds.
4.
4.4 CONCLUSION
S Steady state operation of the Reactor Coolant System and the Steam Generators was monitored at,various pcwer levels during escalation to 100% power. All measured average unit parameters fell within their respective maximum /minimna bounds, or the paraceters were evaluated and a determination was made that the variation from preaicted range did not have an adverse effect on the safe operation or control of the Unit.
Based on data from this test the OTSG Performance, restriction. after repairs, was acceptable for continued coeration without For the OTSG levels and the secondary side fouling, the parameters were evaluated with and the OTSG high level limit was re' set to 92% cf Operating Range B&W; concurrence.
The Plant was initially limited to 87.5% power by the
' 92% Operating Range limit, however after the 87.5%' trip the fouling was apparently less severe and 100% power was reached, with OTSG levels' considerably ~1ess than the 92% high level limit and within expected range.
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4
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35 1
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Table 1-4.4 AVERAGE UNIT PARAMETERS AT VARIOUS TEST PLATEAUS FOR TMI-UNIT 1 DURING STEADY STATE CONDITIONS Average Unit Parameters Unit Parameter l Units l 5%FP l 15%FP l 15%FP* l 25%FP l 40%FP l 60%FP l 75%FP l 87.5%FP**l 100%FP**
Reactor Power (NI-5) l % l 4.77 l 15.5 l 14.0 l 26.4 1 41.2 1 61.6 1 74.5 1 86.2 l 98.7 OTSG 'A' Pressure l PSIG l 890.3 l 882.9 l 880.0 l 883.8 l 884.4 l 891.5 l 896.0 l 896.4 l 909.1 OTSG 'B' Pressure l PSIG l 894.8 l 887.1 l 885.1 l 889.4 l 889.8 l 895.3 l 899.8 l 896.4 l 909.6 Total Feedwater Flow l M lb/hr l .366 l 1.401 l 1.323 l 2.4 l 3.9 l 6.1 l 7.6 l 9.1 l 10.5 Feedwater Temperature l *F l 212.56 l 295.8 l 283.8 l 341.2 l 377.9 l 414.1 l 433.2 l 448.7 l 459.2 RC Outlet Temp. 'A' l *F l 546.9 l 570.3 l 572.8 l 585.8 l 588.9 l 592.7 l 595.7 l 597.1 l 600.1 RC Outlet Temp. 'B' l *F l 546.6 l 569.8 l 571.7 l 584.8 l 588.0 l 592.2 1 595.3 l 597.2 l 600.5 RC Inlet Temp. 'A' l *F l 544.5 l 562.9 l 565.8 l 573.4 l 569.7 1.564.9 l 562.0 l 558.7 l 556.5 RC Inlet Temp. 'B' l *F l 544.8 l 563.5 l 565.4 l 573.4 l 569.8 l 565.1 l 562.0 l 558.7 l 556.0 RC Average Temp. l *F l 545.7 l 566.6 l 568.9 l 579.35 l 579.1 l 578.1 l 578.8 l 577.9 l 578.5 OTSG S/U Level ' A' l in H2 O l 30.9 l 30.6 l 26.5 l 39.4 l 72.0 l 157.1 l 206.2 l 249.9 l 212.6 OTSG S/U Level 'B' l in H 2O .1 29.6 l 30.7 l 30.2 l 41.4 l 74.0 l 159.5 l 208.5 l 249.9 l 191.1 u, OTSG S/U OP Level 'A' l % l 4.15 l 5.77 l 6.02 l 8.71 l 14.42 l 47.42 l 69.36 l 89.96 l 79.52 m OTSG S/U OP Level 'B' 'l % l 3.68 l 5.21 l 4.92 l 8.17 l 14.27 l 46.55 l 71.60 l 90.79 l 72.03 OTSG Steam Temp. 'A' l *F l 535.3 l 569.6 l 571.1 l 585.5 l 588.4 l 589.4 l 589.0 l 584.9 l 579.9 OTSG. Steam Temp. 'B' l F l 534.5 l 568.6 l 569.7 l 584.0 l 587.1 l 590.1 l 590.5 l 588.5 l 589.5 Turbine Header Pressure 'A'l' PSIG l 893.4 l 886.3 l 877.4 l 884.6 l 884.4 l 886.1 l 887.4 l 881.8 l 888.3 Turbine Header Pressure 'B'l PSIG l 890.3 l 882.9 l 880.9 l 882.1 l 881.7 l 885.5 l 888.3 l 883.3 l 893.1 RC Pressure NR . l PSIG l 2150.4 l 2153.6 l 2150.7 l 2148.2 l 2149.6 l 2150.1 l 2151.6 l 2150.5 l 2150.2 Fressurizer Level l . inches l 140.6 l 221.0 l 201.0 l 224.7 l 225.3 l 220.0 l 220.4 l 219.0 l 221.3 Reactor Thermal- Power from l MWT l 113.7 l 376.0 l 363.89 l 671.6 l 1003.8 l 1558.55 l 1887.8 l 2206.4 l 2534.09 Current Heat Balance l l l l l l l. l l l
- 15% data taken af ter 40% Reactor Trip during recovery to 48% power, as requested by TAG.
- Power limited by OTSG High level limits (92.0% Op. Range) prior to Reactor Trip on Turbine Trip @ 87.5% power.
After trip OTSG 1evels not limiting factor, power escalated to 100% power.
FIGURE 1 - 4.4 STE AM GENERATOR OUTLET PRESSURE DEYlATION VS POWER 940
- tr_. : ,
p
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4 '
I O 0.5 1.0 1.5 2.0 2.5 3.0 NSS Power, nit x 10-3 i I 4.77% 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7%
% Full Power
. 37
. FIGURE 2 - 4.4 l l
, TOTAL FEEDIATER FLOW VS POWER l'6
+
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i i 0 0.5 1.0 1.5 2.0 2.5 3.0 NSS Power Mit x 10-3 I I 4.77% 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7%
I
% Full Power
. 38
I i 1
1 FIGURE 3 - 4.4 i
FEEDIATER TEMPERATURE VS POWER LEVEL 500
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ALLOWABLE DEvi ATION iEEEis ant == n4 EElEE55557:d' EEput**ntgir3EEEpIE EQ5E15E@l = .E- Et.3 Eriij
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I O 0.5 1.0 1.5 2.0 2.5 3.0 NSS Pour. Mit x 10-3 e 4.77" 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7%
% Full Power
. 39
RCS ALLOWABLE TEMPERATURE DEVIATION VS POWER FIGURE 4 - 4.4
, -AVERAGE TEMPERATURE GF 579'F AND 86,000 GPM 620 *
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re
= --r :J, .:i. ]:..;i.j:._.J_ .2=- c. ==
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== = L . j.. . u..l.:.
0 0.5 1.0 f.5 2.0 2.5 3.0 NSS Posu, MUt x 10 3 4.77% 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7%
% Full Power
. 40
FIGURE 5 - 4.4 OTSG.STARTUP RANGE LEVEL AP VS POWER m.i 240
!ISl!!!!ildil0!!iiillifii '!E!!!!iil!E!lI5i$il ~ili' !!i Mil!!!il!!!i !!$j5}l5!55E:551"E+N
'p"'*;uij"nng'::.. " .ajj;' n"
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= t. .. . .u:y2!l:'.'..'.l8":::'II' 2 20 ' " ' -
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- = t....
- n=t.p;- ---Agr ;_.. --d ~See notes
-=._3=..
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,e bottom.
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. . - .-- ..q 0 0.5 1.0 1.5 2.0 2.5 3.0 i
NSS Poser, H t x 10-3 4.77% 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7%
% Full Power Note 1: Reactor power limited to 87.5% FP by OTSG liigh Level limits, af ter Reactor trip at 87.5%, Reactor power was no longer limited and power was escalated to -100% FP. A 6 B OTSG levels were plotted separately at 100% Full Power.
Note 2: The maximum limic has been reset to 92% of Operating Range. See Section 4.4 for discussion. 41
FIGURE 6 - 4.4 OTSG OPER ATE RANGE LEVEL VS P0tLR 100 _. .___. _ , _ _ . _ _ _ __ _ _ _ ____._ i - : _
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NSS Power, Ht x 10-3 l l l 4.77% 15.5% 26.4% 41.2% 61.6% 74.5% 86.2% 98.7% 1 l
% Full Power Note 1: Reactor power limited to ~87.5% FP by OTSG High Level limits, after Reactor trip at 87.5%, Reactor power was no longer limited and power was escalated to ~100% FP. A & B OTSG levels were plotted separately at 100% Full Power.
Note 2: The maximua limit has been reset to 92% of Operating Range. See Section 4.4 for discussion.
. 42
FIGURE 7 - 4.4 STEAN TEMPERATURE VS POWER 610 -
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6 . 0 0.5 1. 0 1.5. 2.0 2.5 3.0 HSS Power, NWt I 10-3 4.77% 15.5% 26.4% 41.2% 61.6Z 74.5% 86.2% 98.7%
% Full Power 43
4.5 REACTOR COOLANT SYSTEM OVERC00 LING CONTROL TEST 4.5.1 PURPOSE The purposes of the Reactor Coolant System (RCS) Overcooling Control Test (TP 800/8) were: (1) To demonstrate that the Control Room operators can properly throttle emergency feedwater flow to prevent overcooling of the RCS following a loss of Reactor Coolant pumps with the Once Through Steam Generator level initially at 30" on the startup range. (2) To provide data and guidelines to revise Plant Procedure OP 1102-16 (RCS Natural Circulation). If revision is shown to be necessary. (3) To provide the basis for preparing guidelines for the transition from Natural Circulation to forced flow in the RCS. 4.5.2 TEST HETHOD Test TP 800/8 consisted of two parts, transition to Natural Circulation and transition from Natural Circulation to forced flow. This test was conducted after the Reactor trip at 40% power had occurred in accordance with TP 800/2, Reactor Trip on Loss of Feedwater/ Turbine Trip Test. These parts are i described as follows:
~
(1) Transition to Natural Circulation - This starts with steady state conditions established in TP 800/2 with Emergency Feedwater (EFW) feeding the Steam Generators while maintaining levels at 30" on the Startup Range. The Reactor Coolant Pumps were then tripped and verified to coastdown while EFW flow was manually controlled to maintain OTSG Pressure within 100 psig of desired pressure of 950 psig. The EFW flow was maintained as continuous as possible, while also maintaining bcth OTSG levels approximately the same. Once natural circulation was verified at 30" on Startup Range, attempts were made to raise OTSG level to 50% on the operating range. The test was terminated prior to reaching 50% on the operating range and the Plant was allowed to stabilize before proceeding to part 2. (2) Transition from Natural Circulation to Forced Flow - When the plant systems had stabilized, the first Reactor Coolant Pump was started and the Turbine Bypass Valves were used to manually control 0TSG pressure to prevent lifting the Main Steam Relief valves. After OTSG pressure peaked and was below 1010 psig the Turbine Bypass Valves were placed in automatic and then the remaining Reactor Coolant Pumps were started. A main feedwater pump was then started and MFW flow restored to the Steam Generators. Once positive MFW flow control was establisned the EFW { pumos were secured and the OTSG 1evels were steamed down to 30" on the Startup Range. 1
- L4
'1 ;
4.6.3 RESULTS l l Natural Circulation was established with EFW flow maintained as continuous as possibl e. The OTSG level was then increased, maintaining both OTSG levels approximately the same. The test was terminated with OTSG levels of approximately 3.6% (A) and approximately 2.9% (B) when pressure in the ' A' OTSG decreased to approximately 750 psig without a continuous level increase. The Plant was then stabilized and forced flow was established, using the procedure, to return the Plant to normal Hot Shutdown conditions. 4.
5.4 CONCLUSION
S Although the Natural Circulation portion of this test was terminated prior to reaching an OTSG 1evel of 50% in the operating range, the major objectives were considered to be met. These are described below: , (1) It was desired to verify that a transition to Natural Circulation would occur with low OTSG level and throttled EFW flow. A review of the data showed that a transition to natural circulation occurred successfully. Based on the test results, guidelines will be modified during the annual procedure review which will assist the operator in raising OTSG levels during conditions of extremely low decay heat. (2) It was also desired to verify that the operators could take manual control of EFW flow and throttle flow enough to prevent overcooling but still make a transition to natural circulation. This manual control was accomplished, the maximum cooldown rate was approximately 60*F/hr. (3) The procedural guidelines which were used to make the transition from natural circulation back to forced flow were shown to be effective in preventing overpressurization of the secondary side and monitoring the subsequent shrink. These procedural guidelines will now provide the basis for issuing guidelines to be used for this transition. It was understood that because of the very low decay heat and normal steam leaks we might not be able to reach the desired level without depressurizing the Steam Generators below 750 psig, and if this occurred the test would be terminated. The accomplishment of the major test objectives was not directly related to the OTSG level attained, and therefore, the test was successful. The operator training manual has been revised to incorporate the experience gained regarding RCP restarts from natural circulation conditions. Procedural guidance is being written and will be incorporated into the Abnormal Transient i Procedures by mid-May. , i B&W has also provided an assessment of the cold leg temperature anomaly l observed in this test and in TP 700/2 and supports the GPUN judgment that RCP seal flow caused the anomalous temperature response. The tests results are also being provided to the Operator Support Committee for their consideration of the issue on a generic basis. 45 l
Y 4.6 EMERGENCY FEEDWATER PUMP AUTO-START TEST l 4.6.1 PURPOSE The purpose of the Emergency Feedwater Pump Auto Start Test (TP 800/9) is to demonstrate the Emergency Feedwater (EFW) pump auto start on loss of all four Reactor Coolant pumps. 4.6.2 TEST METHOD i The test was conducted following the performance of the Reactor. Trip on Loss of Feedwater/ Turbine Trip Test (TP 800/2) at 88% power and the subsequent Plant recovery to a normal, post-trip, hot shutdown condition. With one main feedwater pump supplying both Steam Generators, the Steam Generators were fed up to approximately 60" on the Startup Range and the Plant was controlled in a stable non-trending status. The Main Feedwater (MFW) startup valves were then placed in manual and EFW valves EF-V-30A/B were placed in ' Hand' and verified fully closed. All four Reactor Coolant pumps were then trioped. The EFW pumps were verified to start on the trip of the RC pumos and then secured, as necessary, and the RC pumps were restarted and the Plant was returned to a normal, post trip, hot shutdown condition. 3 4.6.3 RESULTS 4 The two motor driven EFW pumps EF-P2A/B and the turbine driven EFW pump EF-P1 all started as required upon loss of all RC pumps. 4.
6.4 CONCLUSION
S i During Plant operation, if a loss of all RC pumps ocours, the EFW pumps (motor i driven and turbine driven) will all start, as designed. i i I l J .i
].
l
, 46
4.7 FEEDWATER SYSTEM OPERATION AND TUNING 4.7.1 PURPOSE The purposes of the Feedwater System Operation and Tuning Test (TP 836/1) were: (1) To verify that the Heater Drain System level controls function properly and adjust them as required. (2) To verify satisfactory performance of the feedwater heaters. (3) To verify that the Feedwater System operates without excessive oscillations during steady state and transient conaitions at power levels between 0 and 100% power. 4.7.2 TEST METHOD The feedwater heater and drain tank levels were monitored at all power levels and during all changes in power levels. Data was recoraed for 0%. 15%. 25%, 40%, 75% and 100% cower levels. The level controllers were fine tuned by adjusting the proportional band and reset and data was recorded whenever fine tuning was performed. 4.7.3 RESULTS The feedwater heater and drain tank level controls functioned properly and were fine tuned during the test program to operate within the desirea tolerance. Evaluation of computer heat balance and comparison with provided GE heat balance at 40%, 80%, 100% power levels indicated feedwater heater performance was within expected tolerances. System monitoring during both steady state and transient conditions verified that the Feedwater System operates without excessive oscillations. Discrepancies found during operation were: (1) Insufficient heater drain flow into the sixth stage heater drain tank prevented starting the second heater drain pump at 40% power; this second heater drain pump was started at approximately 65% power. This discrepancy resulted in a revision to the Plant Operating Procedures. (2) Due to chemistry concerns, no Moisture Separator Orain Tank pumps were fed forward, as indicated by Plant Operating Procedures, during power escalation. Levels were centrolled by the high level aumo valves. After the Reactor Trip at 88% power and upon return to 100% cower, the olant started feeding forward four pumos at a time. For enemistry concerns, the four operating pumps are continually cnanged, on a rotating basis, among the six pumos.
, 47
4.
7.4 CONCLUSION
S The feedwater heater and drain tank levels were monitored at all power levels, during steady state and transient plant operation, and found to perform satisfactorily and without excessive oscillations. The feedwater heaters performance at 40%, 807., and 100% was compared with expected values provided by GE and found to be satisfactory since values were all within expected tolerances. l l l I
. 48
4.8 INCORE THERMOCOUPLE TEST 4.8.1 PURPOSE 1 The purposes of the Incore Thermocouple Test (TP 846/1) were: (1) To verify proper operation of the incore thermocouples at nominal power plateaus of 15%, 40%, 75% and 100% rated power. 4 (2) To verify that the incore thermocouples are giving an accurate indication } of the temperature distribution in the core, and those thermocouples which are symmetric to one another give comparable readings. (3) To verify the following functions are available and functional on the j Plant Computer System: 1 , j (a) Determination of core exit temperature by averaging the five j highest valid incore thermocouple readings. (b) Displaying all fifty-two incore thermocouples. l
- (c) Identifying and deleting any malfunctioning incore thermocouple
] from any computer calculation. i (d) Computation of temperature saturation margin, based on the average of the five highest valid incore thermocouples. (e) Determination of the mean core exit temperature by weight averaging all of the valid incore thermocouple readings. (4) To verify that the sixteen selected incore thermocouples can be ' I accurately read from the Backup Incore Thermocouple Readout (BIRO) display. 4.8.2 TEST METHOD This test was performed during the Power Escalation testing with data - collection at 15%, 40%, 75%, 88%, and 100% power. The test consisted of a , 4 collection of various computer data and BIRO readings, performing calculations, and comparison of data for each power level listed to verify proper computer and thermocouple operation. 4.8.3 RESULTS Thermocouole data was collected at 15%, 40%, 75%, 88% and 100% power and calculations and comparisons were performed. It was verified that all recuired functions were available and functional on the Plant Computer System
! at all power levels. The symmetrical thermocouples were verified to agree ,
l within t 1% of the average for that particular group of thermoccupies. The thermocouoles were also found to agree : 2% with calculated values based on core power distribution. i i 1 49 i
- ~ _ , ---,-n,- - - - - , - , - - - - , . , . ,. --~n,- .,-.,,. .,,,,..v.----.-
- -- - ----, n~.--. - . , , , . , , -.
Except for 100% power level, the five highest incore thermocouples, as selected by computer program, did not correspond with the five highest readings from the computer groups. This difference was due to the small ' differences in incore thermocouple readings and the fact that there was a time difference between when the two computer data collections occurred. This time difference and small temperature difference means that at any given time, i small changes in flux or flow mixing could result in a different selection of five thermocouples being the highest by either means of data collection. The average value of the five highest incore thermocouples by each method was always within 3 2*F of the other averages, and also the computer calculated value. The computed saturation margin was within i 5*F of the computer calculated value at each power level. The values for the Backup Incore Thermocouple Readout (BIRO) display and the same values from the Plant Comouter Operator's Group were all in agreement within 3 16*F temperature scan allowed by the acceptance criteria. 4.
8.4 CONCLUSION
S The operation of the incore thermocouples was r.onitored
. at 15%, 40%, 75%. 88%
and 100% power. All data collected or calculated either fell within reautred tolerances or was evaluated and a determination was made that the variation from the expected value did not have an adverse effect on the safe operation or control of the unit. The operation of the incore thermocouples was
, determined to be satisfactory.
I i a I l I i ) ) l f
. 50
L . 4.9 INTEGRATED CONTROL SYSTEM TUNING 4.9.1 PURPOSE A The purposes of the Integrated Control System (ICS) Tuning at Power Test ? (TP 849/1) were: (1) To determine settings, find the natural period, and adjust the sub-loop controllers for optimum performance for the following: (a) Turbine Bypass Valve Control (b) Feedwater Pump Speed Control i (c) Steam Generator Level Control (d) Steam Generator Feed Flow Control (e) Turbine Control (f) Reactor Control (g) Tave Control (h) Delta Tc Control (2) To determine the actual relationship between basic Plant parameters and set the ICS Feed forward to these functions. I (3) To adjust the ICS in the various modes of control for optimum performance. 4 (4) To insure ICS controls the Plant response during plant transients and steady state conditions, including the following: (a) Transient testing at 48% power where the Plant was subject to load j changes at a rate of 2% per minute to 40% power and back to 48% 1 power while in fully integrated mode of control. 4 ) (b) ICS response to a CRD runback at 75% power. (c) ICS response to a single main feedwater pump trip at Reactor power l of 75% power and 87% power. t i I l ) 4.
. 51
l 4.9.2 TEST METHOD During Power Escalation Testing, at various power levels, each ICS subloop controller was adjusted. as necessary, to give a rapid yet stable response to changes in setpoint or controlled parameter. Also, at various power levels, the feed forward data was recorded and controls were adjusted, if necessary, for stable plant conditions. The operation of the ICS was tested at various power levels, in various control modes, and adjusted as necessary for optimum performance. At 48% power the Plant was subjected to load changes at a rate of 2% power per minute to 40% power and then ramped back to 48% power, wn11e in fully Integrated mode of control. At approximately 75% power, in Integrated mode of control, the Plant was subjected to a simulated CRD runback by installing a temporary jumper. After recove y the plant was suojected to a single main FH pump trip runuack by tripping one MFH pump. The plant was also subjected to a single main FH pump trip runback at approximately 87% power. 4.9.3 RESULTS The results of the adjustments performed, as necessary, at various power levels are the following: (i) The ICS subloop controllers were adjusted for optimum performance. (2) The ICS feed forward data was recorded and adjustments were made for stable plant conditions. (3) The ICS was tested in various modes of control at several power levels and adjusted for optimum performance. The Plant was subjected to a ramp change of 2% power per minute from 48% power to 40% power and back to 48% power. This ramp demor.strated the ICS control to be stable during plant transient conditions. At approximately 75% power the Plant was subjected to a simulated CRD runback, during which the Plant ran back to approximately 55% full load. The Plant was returned to approximately 75% power and then subjected to a main FW pump trip, which caused the Plant to runback to 60% full load design feedwater flow. At approximately 87% power the plant was again subjected to a main FH pump trip. This trip also caused the Plant to runback to 60% full load design feedwater flow. 4.
9.4 CONCLUSION
S The Plant ICS was monitored and adjusted at various power levels and during steady state and transient response. The conclusions of this test are that the Plant ICS is tuned and can control the Plant in various modes, during both transient and steady state conditions, in a stable manner. Another conclusion is that the ICS will runoack the Plaqt as required during a load limiting situation.
, 52
1 4.10 TURBINE GENERATOR OPERATIONAL TESTING 4.10.1 PURPOSE The purposes of the Turbine Generator Operational Test (TP 885/1) were: (1) To record turbine generator megawatt electrical output and reactor thermal output at power levels of 5%, 15%, 25%, 40%, 60%, 75%, 87.5% and 100% of Rated Thermal Power. (2) To record stator slot and coolant temperatures on the Turbine Generator at power levels of 5%, 15%, 25%, 40%, 60%, 75%, 87.5% and 100% of Rated Thermal Power. (3) Record vibration readings on the 12 Turbine Generator main journal bearings at power levels of 5%, 15%, 25%, 40%, 60%, 75%, 87.5% and 100% of Rated Thermal Power. 4.10.2 TEST METHOD The Turbine Generator was brought up to rated speed (1800 RPM), synchronized, and loaded in accordance with Plant Operating Procedure OP 1106-1, Turbine Generator Operation, and OP 1102-2, Plant Startup. Data was recorded at ) approximately 5% power with Turbine Generator secured, but in a warmed-up state, for baseline data at the request of Plant Engineering. With the Turbine Generator operating under load, data was recorded af ter steady state conditions were established at power levels of 15%, 25%, 40%, 60%, 75%, 87.5% and 100% of Rated Thermal Power. During testing, data was evaluated for each power level and approved prior to proceeding to the next power plateau, For each power level, generator electrical output (MWe) was plotted with % Reactor Power, see Figure 1-4.10. 4.10.3 TEST RESULTS All temperature and vibration data collected during the Power Escalation testing fell within the allowable limits except as discussed as follows: (a) Two thermocouples and two RTD's were found to be out of service. Review of results indicated that the loss of these instruments had no significant effect on test results since there are 72 thermocouDies and 72 RTD's and the remaining instruments were all performing correctly.
, 53
(b) Two thermocouples were found to slightly exceed the acceptance criteria of 3 6*F of average value of all readings at 100% power only. Discussions with Plant Engineering and General Electric concluded that the amounts (2.5*F and 0.5'F) were not significant enough to cause concern, based on prior data and temperatures of corresponding RTD's and adjacent thermocouples. 4.
10.4 CONCLUSION
S Steady State operation of the Turbine Generator was monitored at various power levels during escalation to 100% power. All measured parameters either fell within the allowable limits, or were evaluated and a determination was made that the variation did not adversely effect the safe operation of the Turbine Generator.
, 54
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4.11 TURBINE BYPASS VALVE TESTING 4.11.1 PURPOSE 3 The purposes of the Turbine Bypass Valve Test (TP 885/2) were: (1) Verify proper response of the Turbine Bypass valves to setpoint and changes in setpoint (small perturbations) at 10% to 15% power during testing performed by ICS Tuning Procedure (TP 849/1). (2) Determine bypass valve opening time and peak steamline pressure reached during the 40% power loss of FW trip and the 88% power Turbine trip, and verify correct setpoint change on Reactor trip. 4.11.2 TEST METHOD At approximately 10% power with the turbine stop valves closed, and turbine bypass valves controlling header pressure in manual, a valve step change was introduced to measure controller dead time. Then with the turbine bypass valves in auto a 10 psig step change in turbine header pressure setpoint was made, and the ability of the bypass valves to control header pressure at the new setpoint was determined. All Integrated Control System (ICS) adjustments were performed according to ICS Tuning at Power Procedure (TP P49/1). The bypass valve opening time, and peak pressure were observed during the 40% power loss of FH trip and the 88% power Turbine trip along with the turbine header pressure setpoint change. 4.11.3 RESULTS The Turbine Bypass valve response to setpoint and changes in setpoint was found to be satisfactory and the valves controlled header pressure at the setpoint. During both the 40% and the 88% power trips the Turbine 8ypass valve opening times could not be observed, since they never received a full open control signal due to Main Steam Safety Valve actuation. The peak Turbine header pressures were recorded, and
+125 psi following Reactor trip. the control setpoint was verified to change by 4.
11.4 CONCLUSION
S The Turbine Bypass valves respond satisfactorily to setpoint and changes in setpoint and do exhibit a change in control setpoint to approximately 1010 psig following Reactor trip. At the present time, the Main Steam Safety Valves lift and prevent the Turbine Bypass Valves from fully opening and from controlling header oressure at 1010 psig. WIth the Turoine Bypass Valve control setpoint reset to a lower value, the Main Steam Safety Valves reseated and the Turbine Bypass valves controlled the header pressure. The Turbine Bypass valves are operational and acceptable for use and the inability to measure their opening time does not affect this conclusion. 56
5.'0 CORE PERFORMANCE TESTING 5.1 CORE PERFORMANCE -_ MEASUREMENTS AT ZERO POWER -
SUMMARY
Core performance '.nasurements were conducted during the Zero Power Test Program which t aan on October 3,1985 and ended on October 5,1985. This section presents a summary of the zero power measurements. In all cases, the applicable test and Technical Specifications limits were met. A summary of zero power physics test results is included on Table 1-1.
- a. Initial Criticality Initial criticality was achieved at 1330 on October 3,1985.
Reactor conditions were 532*F and 2155 psig and control rod groups 1 through 6 were withdrawn to 100% while group 7 was positionea at 75% withdrawn. Control Rod group 8 was positioned at 37.5% withdrawn. Criticality was achieved by deborating the Reactor Coolant from 1805 ppm to 1167 ppm. Initial criticality was achieved in an orderly manner and good agreement was found between the predicted critical boron concentration of 1230 ppm at 75% on group 7 and the measured critical boron concentration of 1167 at 66.5% on group 7.
- b. Nuclear Instrumentation Overlap At least one decade overlap was measured between the source and intermediate range neutron detectors as required by Technical Specifications.
- c. Reactimeter Checkout An on-line functional check of the Mod-Comp reactimeter (using NI-3) was performed af ter initial criticality. Reactivity calculated by the reactimeter was within 5% of the core reactivity determined from doubling time measurements.
- d. All Rods Out Critical Boron Concentration i
The measured all rods out critical baron concentration of 1182 ppm 8 was in agreement with the calculated value of 1255 t 100 ppm 8. )
- e. Temoerature Coef ficient Measurements The measured temperature coefficients of reactivity at 532*F, zero power were within the acceptance criteria limits over the range of boron concentrations and rod positions that the measurements were made.
- f. Control Rod Group Worth Heasurements The measured results for control rod worths of groups 5, 6 ana 7 conducted at zero power (532*F) using the boron / rod swao method were in good agreement with predicted values. The maximum deviation between measured and predicted worths was 10.75% which was for group 5 worth. '
57
- g. Differential Boron Worth The measured results for the dif ferential boron worth at 532*F were 13.6% greater than the predicted values but within the bounds of the FSAR and B&W supplied limits'of t 15%.
- h. Shutdown Marcin Minimum shutdown margin verification measurements were performed at zero power (532*F) using the rod drop technique to determine the total worth of the safety groups. The shutdown margin with the most reactive rod stuck out of the core was calculated to be 3.30%
AK/K. This meets the acceptance criteria of 1% AK/K shutdown margin.
- i. E.iected Control Rod Worth The worth of the worst case " ejected rod" was measured using both the boron swap method and the rod swap method. The values obtained f rom the measurements were in good agreement with each other and with the predicted value. The average corrected ejected roc worth value of 0.691% AK/K was well within the Technical Specification limit of 1.0% aK/K.
58
TABLE 1-1 Summary of Zero Power Physics Test Results_ Cycle 5 Calculated Measured Pa rameter Value Value Critical Boron 1230 t 100 ppm 1171 ppm j Sensible Heat N/A 2.5 x 10-7 amos HI Overlap >l decade 1.5 decade All Rods Out Boron Concentration 1255 100 ppm 1 1182 ppm Temoerature Coefficient -2.45 pcm/*F -2.30 pcm/*F (1178 ppm) t 4 pcm/*F Moderator Coef ficient -0.49 pcm/*F -0.31 pcm/'F Integral Rod Worths (532*F) GPS-7 3.22 0.32% AK/K 3.21% aK/K Group 7 1.32 t 0.20% aK/K 1.400% aK/K Group 6 0.84 0.13% 4K/K 0.863% AK/K Group 5 1.06 0.16% aK/K 0.946% aK/K Temperature Coefficient -12.44 pcm/*F (901 ppm) -12.96 pcm/*F 4 pcm/*F Moderator Coefficient -10.46 pcm/*F -10.97 pcm/*F Ejected Rod Worth 0.65% AK/K 0.691% aK/K l I i i 59 i
- d. Core Power Distribution Verification Core power distribution measurements were conducted at 40%, 75%,
88% and 100% full power under steady state equilibrium xenon conditions for specified control rod configurations. The maximum measured and maximum predicted radial and total peaking factors are all in good agreement. The maximum difference between a measured and predicted value was 4.1% for total peaking at 100% FP. This met acceptance criteria of < 7.5%. The results of the core power distribution measurements are given in Table 4.4-1. All quadrant power tilts and axial core imbalances measured during the power distribution tests were within the Technical Specification and normal operational limits.
- e. Reactivity Coefficients at Power The temperature coef ficient measured at 88% FP was -8.70 pcm/*F.
i The measured power doppler coef ficient at 88% FP was
-8,585 pcm/% FP. All Technical Specification and Safety Analysis ! requirements were met.
l 4 I l l l l 60 l ! I
5.3 CORE PERFORMANCE - HEASUREMENTS AT ZERO POWER This section presents the detailed results and evaluations of zero power physics testing. The zero power testing program included initial criticality, nuclear instrumentation overlop, reactimeter checkout, all rods out critical boron concentration, temperature coefficient measurement, control rod worths, boron worth, shutdown margin verification and ejected control rod worth. 5.3.1 Initial Criticality Initial criticality for Cycle 5 was achieved at 1330 on October 3, 1985. Reactor conditions were 532*F and 2155 psig and
' control rod groups 1 through 4 were, previously withdrawn during the heatup to 532*F. The initial reactor coolant system (RCS) boron concentration was 1805 ppm.
The approach to criticality began by withdrawing control rod group 8 to 37.5% withdrawn, control rod groups 5 and 6 .to 100% withdrawn, and positioning group 7 at 75% withdrawn. Criticality was subsequently achieved by deborating the reactor coolant system to a boron concentration of 1167 ppm. The procedure used in the approach to critical is outlined below in three basic steps: Step 1 Control Rod Withdrawal Group 8 37.5% withdrawn a Group 5 100% withdrawn Group 6 100% withdrawn Group 7 75% withdrawn Step 2 Deborate using a feed and bleed flow rate of 70 gpm until the inverse count rate is between 0.2 and 0.3. At this t point, deborate at a feed and bleed flow rate of 40 gpm. i Step 3 Stop deboration and increase letdown flow to maximum (140 gpm) to enhance mixing between the makeup tank and the reactor coolant system. Achieve initial criticality and position control rod group 7 to control neutron flux as the rector coolant !gstem baron concentration reaches equilibrium. Throughout the approach to criticality, plots of inverse multiplication were maintained by two independent persons. Three plots of inverse count rate (ICR) versus control rod position were maintained during control rod withdrawal. Three plots of ICR versus RCS baron concentration and three plots of ICR versus gallons of demineralized water added were maintained during the dilution sequence. During each reactivity addition (boron dilution and control rod withdrawal), count rates were obtained from eacn i source range neutron detector channel and a temporary hookuo to one of the spare source range detectors." I 61
r During control rod withdrawal (step 1) ICR plots versus control rod group position were maintained from the outputs of three source range detectors. The withdrawal interval for each control rod group was limited to no more than half the remaining predicted distance to criticality as determined from the ICR plots. Deboration of the reactor coolant system was accomplished in two steps as indicated above. First, ceboration from 1805 ppm was commenced using a feed and bleed flow rate of 70 gpm (step 2). RCS boron samples were taken every 30 minutes and samples from the makeup tank and the pressurizer were taken hourly. Three ICR plots were maintained vs. galloni of demineralized water added, and three plots were maintained vs. RCS letdown concentration every 30 minutes. Deboration at a letdown rate of 70 gpm was continued until one of the ICR plots wass 0.25. At this time, demineralized water addition was reduced to 40 gpm. When one of the ICR plots indicated % 0.10, the letdown flow rate was increased to 120 gpm in a recirculation mode to expedite mixing in the RCS (step 3). The reactor went critical during the mixing process. The inverse count rate plots maintained during the approacn are presented in Figures 3.1-1 through 3.1-3. As can be seen from the plots, the response of the source range channels during reactivity additions was very good. Figure 3.1-1 is the plot of ICR versus control rod group withdrawal. Figure 3.1-2 is the ICR plots versus RCS boron concentration and Figure 3.1-3 is the ICR plots versus gallons of demineralized water added to the RCS. In summary, initial criticality was achieved in an orderly manner. There was good agreement between the measured and the predicted critical boron concentration. l l l l 62
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5.3.2 Nuclear Instrumentation Overlap
- a. Purpose Technical Specification 3.5.1.5 states that prior to operation in the intermediate nuclear instrumentation (NI) range, at least one decade of overlap between the source range NIs and the intermediate range must be observed.
- b. Test Method To satisfy the above overlap requirements, core power was increased until the intermediate range channels came on scale. Detector signal response was then recorded for both the source range and intermediate range channels. This was repeated until the source range high voltage cutoff value was reached.
- c. Test Results The results of the initial NI overlap data at 532*F and 2155 psig have shown a 1.5 decade overlap between the source and intermediate ranges.
- d. Conclusions The linearity, overlap and absolute output of the intermediate and source range detectors are within specifications and performing satisfactorily. There is at least a 1.5 decade overlap between the source and intermediate ranges, thus satisfying T.S. 3.5.1.5.
66
5.3.3 Reactimeter Checkout
- a. Purpose Reactivity calculations during the Cycle 5 test program were performed using the Mod-Comp Reactimeter. After initial criticality and prior to the first physics measurement, an online functional check of the reactimeter was performed to verify its accuracy for use in the test program.
- b. Test Method Af ter initial criticality and nuclear instrumentation overlap was established, intermediate range channel NI-3 was connected to the reactimeter and the reactivity calculations were started. After steady state conditions with a constant neutron flux were established, a small amount of positive reactivity was inserted in the core by withdrawing control rod group 7. Stop watches were used to measure the doubling time of the neutron flux and the reactivity was determined from the period-reactivity curves. The measurement was for t 25 pcm and 75 pcm. The reactivities determined from doubling time measurements were compared with the reactivity calculated by the reactimeter.
- c. Test Results The results of the reactimeter verification measurements are summarized in Table 3.3-1. The measured values were determined to be satisfactory and showed that the reactimeter was ready for startup testing.
- d. Conclusions An on-line functional check of the reactimeter was performed af ter initial criticality. The measured data shows that the core reactivity measured by the reactimeter was in good agreement with the values obtained from neutron flux doubling times.
67 _ _ _ _ - - - _ - _ _ _ - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - --- --- - - - - - - - - - - - - - - - - - -- ~ - - - - ~^ ~~ ~ ~
TABLE 3.3-1 ' i COMPARISON OF REACTIMETER AND DOUBLING TIME (OT) REACTIVI REACTIMETER AVERAGE AVG. REACTIVITY CASE PERCENT VALUE DT FROM DT NO. DIFFERENCE ' (PCM) __{SEC) (PCM) . (%) 1 +19 254.0 18.50 2.70 2 -21 268.3 -21.57 -2.64 3 +76 46.25 79.50 -4.40 4 -65 107.60 -66.50 -2.26 I a l l [ i 68
5.3.4 All Rods Out Critical Boron Concentration
- a. Purpose The all rods out critical boron concentration measurement was performed to obtain an accurate value for the excess reactivity loaded in the TMI Unit 1 core and to provide a basis for the verification of calculated reactivity worths.
This measurement was performed at system conditions of 532*F and 2155 psig,
- b. Test Method The Reactor Coolant System was ' borated to an almost all rods out condition with control rod groups 1-6 at 100% withdrawn and with group 7 maintaining criticality at 92% withdrawn.
Once steady state conditions were established, control rod group 7 was withdrawn to 100% and the resultant reactivity change was measured. The measured boron concentration with group 7 partially inserted was then adjusted to the all rods out configuration using the result of the rod worth measurement to determine the reactivity worth, in terms of ppm boron, of the inserted control rods.
- c. Test Results The measured boron concentration with group 7 positioned at 92% was 1178 ppm. An additional 4 ppm was added to this value that was derived from 41 pcm due to group 7 withdrawal to 100%
and a differential boron worth of 10.1 pcm/ ppm (i.e., measured result = 1178 ppm + 4 ppm = 1182 ppm).
- d. Conclusions The above results show that the measured boron concentrations of 1182 ppm is in agreement with predicted results of 1255 100 ppm. The results are also consistent with an anticipated reduction of approximately 70 ppm calculated by GPUN for the isotopic changes during the extended shutdown.
69
5.3.5 Temoerature Coefficient Measurements
- a. Purpose The moderator temperature coef ficient of reactivity can be positive, depending upon the soluble boron concentration in the reactor coolant. Because of this possibility, the Technical Specifications state that the moderator temperature coefficient shall not be positive while greater than 95% FP.
The moderator temperature coefficient cannot be measured directly, but it can be derived from the core temperature coef ficient and a known fuel temperature (isothermal Doppler) coefficient. The temperature coef ficient of reactivity was measured for two different boron concentrations to provide a comparison of the moderator temperature coef ficient with the design calculations prior to operation in the power range.
- b. Test Method Steady state conditions were established by maintaining reactor flux, reactor coolant pressure, turbine header pressure and core average temperature constant, with the reactor critical at approximately 10-9 amps on the intermediate range. Equilibrium boron concentration was established in the Reactor Coolant System, make-up tank and pressurizer to eliminate reactivity effects due to boron changes during the subsequent temperature swings. The reactimeter and recorders were connected to monitor selected core parameters with the reactivity value calculated by the reactimeter and the core average temperature displayed on a two channel strip chart recorder.
Once steady state conditicas were estaolished, a couldown rate was started by opening the turbine bypass valves. After the core average temperature decreased by aoout 5*F, core tempera,ture and flux were stabilized and the process was reversed by increasing the core average temperature by. about 10*F. Af ter core temperature and flux were stabilized, core temperature was returned to its initial value. This procedure j was completed with control rod group 7 at 91% wd. and then again at a rod index of 6% wd. During the test at 6% wd, the temperature swings were reversed since it was more ef ficient to decrease temperature by 10*F than it was to increase temperature by 10*F. Calculation of the temperature coef ficient f rom the measured data was then performed by dividing the change in core reactivity by the corresponoing change in core temperature over a specific time period. 1 i 70
- c. Test Results Isothermal temperature coef ficient measurements were conducted at two dif ferent reactor coolant boron concentrations during the zero power test program. The results of the measurements are summarized in Table 3.5-1. The calculated values are
, included for comparison. In all cases the measured results compare favorably with the calculated values.
- d. Conclusions The measured values of the temperature coefficient of reactivity at 532*F, zero reactor power are within the acceptance criteria of 4.0 pcm/*F of the predicted value.
Calculation of the moderator coefficient indicates that it is well within the limits of Technical Specifications 3.1.7.2. J 1 71
TABLE 3.5-1
SUMMARY
OF TEMPERATURE COEFFICIENT MEASUREMENTS AT HOT ZERO POWER CONDITIONS RCS MEASURED CALCULATED MEASURED CALCULATED BORON Irc ITC MTC (Dpm) MTC (Dem/*F) (Dcm/*F) (Dem/*F) (Dem/*F) 1178 -2.30 -2.45 4 -0.31 s5.00 901 -12.96 -12.44 4 -10.97 s,5.00 l l 1 72
-,--v.y-, ,-,g,-- - - --,--,wy, -, - , , , a- -
l 1 l i 5.3.6 Control Rod Group Worth Measurements
- a. Purpose The total amount of excess reactivity present at beginning-of-cycle (80C), hot (532*F), clean conditions is 13.34% AK/K. During reactor operations, nearly all of the excess reactivity is controlled by the soluble poison systems. Additional control is provided by moveable control rods. This section provides comparison between the calculated and measured results for the control rod group worths.
The location and function of each control rod group is shown in Figure 3.6-1. The grouping of the control rods shown in Figure 3.6-1 will be used throughout Cycle 5. Calculated and measured control rod group reactivity worths for the normal withdrawal sequence were determined at reactor conditions of zero power, 532*F and 2155 psi. The measured results were obtained using results of reactivity snd group position from the strip chart recorders.
- b. Test Method Control rod group reactivity worth measurements were performed at zero power, 532*F using the boron / rod swap method. Both the differential and integral reactivity worths of control rod groups 5, 6, and 7 were determined.
The baron swap method consisted of establishing a deboration rate in the reactor coolant system and compensating for the reactivity changes by inserting the control rod groups in incremented steps. The reactivity changes that occurred during the measurements were calculated by the reactimeter and differential rodrworths were obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each group were then summed to obtain the integral rod group worths.
- c. Test Results Control rod group reactivity worths were measured at zero power, 532*F conditions. The boron / rod swap method was used to determine differential and integral rod worths for control rod group 5 - 7 from 100% to 0% withdrawn.
The integral reactivity worths for control rod groups 5 through 7 are presented in Figures 3.6-2 througn 3.6-4.
, 73
These curves were obtained by integrating the measured differential worth curves. The integral worth of group 8 was not measured. The calculated worth of group 8 for 532*F, zero power is 0.43% AK/K. Table 3.6-1 provides a comparison between the predicted and measured results for-the rod worth measurements. The results show good agreement between the measured and predicted rod group worths. The maximum deviation between measured and predicted was 10.75%.
- d. Conclusions Differential and integral control rod group reactivity worths were measured using the boron / rod swap method. The measured results at zero power, 532*F indicate good agreement with the predicted group worths.
1 . 74
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N : : : 7: : 8 : 5 : : 8 : : 7 : : : i o : : : 1 : :6 : : 6 : : 1 : : : P : : : 3 : : 7 : : 3 : : : R : : : : : : , North Transfer Tubes 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 l X l = CONTROL ROD GROUP NUMBER ROD GROUP NUMBER CONTROL NUMBER OF RODS FUNCTION 1 8 Safety 2 8 Safety 3 9 Safety 4 4 Safety 5 12 Control 6 8 Control 7 12 Control 8 8 APSR
. 75
Figure 3.6-2 Integral Worth for CRG-5 Total Uorth = 0.946'; AK/K _. - ~ . _ . . . . 1.0 --
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0 -- , i , 0 10 20 30 40
, 50 60 70 80 90 100 CRG-5 POSITION ($ WD1 !
l l 76 1 1
Figure 3.6-3 Integral Worth for CRG 6 Total Worth = 0.863 %AK/K 1.0 - --
,r_.
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-e=, ,
. [
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. 77 l
Figure 3.6 4 Integral Werth for CRG-7 Total Worth = 1.40 %aK/K 1.3- __.g _ . . _ 1.2 _,s _ ._ w'
~~~
1.1
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/
m r 1.O f. a m 0.9 w I '~
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A x . . .. . 0 ' o r c 6 t ' i 0 10 2a 30 ku 50 60 70 8a 90 too CEG-7 PosITIO:t (.swo}
. 78
TABLE 3.6-1 COMPARISON OF PREDICTED VS MEASURED R00 WORTHS MEASURED PREDICTED PERCENT C P.G. WORTH WORTH OIFFERENCE NO. (4aK/K) (% AR/K) (%) 5 0.946 1.06 0.16 -10.75 6 0.863 0.84 0.13 2.74 7 1.400 1.32 t 0.20 6.06 5-7 3.209 3.22 0.32 -0.34
, 79
5.3.7 Differential Baron Worth
- a. Pproose Soluble poison in the form of dissolved boric acid is added to the moderator to provide additional reactivity control beyond that available from the control rods. The primary function of the soluble poison control system is to control the excess reactivity of the fuel throughout each core life cycle. The dif ferential reactivity worth of the boric acid was measured during thc zero power test.
- b. Test Method Measurements of the differential boron worth at 53?'F was performed in conjunction with the control rod worth measurements. The control rods worths were measured by the baron swap technique in which a deboration rate was established and the control rods were inserted to compensate for the changing core reactivity. The reactimeter was used to provide a continuous reactivity calculation throughout the i measurement. The differential boron worth was then determined by summing the incremental reactivity values measured during the rod worth measurements over a known boron concentration range. The average differential boron worth is the measured change in reactivity divided by the change in boron concentration.
- c. Test Results Measurements of the soluble boron differential worth were completed at the zero power condition of 532*F. The measured boron worth was 11.5 pcm/ ppm 8 at an average boron concentration 1035 ppm 8. The predicted value was 10.1 1.01 pcm/ppmB. '
- d. Conclusions The measured results for the soluble poison differential worth at 532*F was not within 10% of the predicted differential worth. The measured results were within 15% of the predicted ;
worth which is the acceptance criteria band specified by the NSSS supplier for this cycle. 80
5.3.8 Shutdown Marain
- a. Purpose Technical Specification 3.5.2.1 states that the available shutdown margin shall not be less than 1% aK/K with the most reactive control rod stuck out of the core. The purpose of the safety rod drop worth measurement at zero power, 532*F was to verify that sufficient shutdown margin exists with the most reactive control rod stuck out of the core.
- b. Test Method Critical equilibrium conditions were established with control rod groups 1-4 at 100% WD and groups 5-7 at 0% WD. The reactimeter started logging data at 0.2 second intervals. The reactor was then tripped manually and the negative reactivity inserted into the core by control rod groups 1-4 was measured.
It was assumed that the calculated worst case stuck rod had not dropped into the core and the available shutdown margin was calculated.
- c. Test Results i The most reactive control rod at zero power, 532*F was calculated to be in location M-13. Table 3.8-1 shows the results of the test. The boron concentration for the test was 898 ppmB. Correction factors were applied to the measured reactivity value from the reactimeter to correct for changes in drop.
the spatial flux distribution immediately after the rod The minimum available shutdown margin with the most reactive 3.30% control red stuck out of the core was measured to be AK/K.
- d. Conclusions Minimum shutdown margin verification was completed for the zero power condition at 532*F. The calculated worst case stuck rod worth of 2.24% aK/K in location M-13 was assumed for the stuck rod condition. The shutdown margin available under this condition was 3.30% AK/K (measured) which is well within the Technical Specification limit of 1.0% AK/K.
i
, 81
.~ . - - - . . - -.- -_ - _ . _
E TABLE 3.8-1 i SHUTOOWN MARGIN CALCULATION Initial Group Positions: CRG l-4 at 100%, CRG 5 at 3.5% CRG 6-7 at 0%, CRG 8 at 37.5% FOUR SEC REACTIVITY INTERVAL (ocm) 1 16 2 4641 Average = 4378 pcm 3 4627 4 CRG-5 Correction = -14 ocm 4579 5 ' 4401 Net = 4364 pcm 6 4476
- 7 4437 j 8 4384 j 9 4458 l 10 4552 i 11 4149 12 4459 13 4122 14 4364 1
Measured Value = Net x 1.41 = 6153 pcm Corrected Value = (Measured Value x 0.9)/(1000 pcm per % AK/K) = 5.54% 4K/K a Worst Case Stuck Roa = 2.24% 4K/K Shutdown Margin = 5.54 - 2.24 = 3.30% AK/K j r 4 1 (
, 82. .. . .- .. . ~ ... -- , , - ,. . . . . . . . . .- , . - , .
5.3.9 Ejected Control Rod Worth
- a. Purpose Technical Specification 3.5.2 states that the maximum worth of a single inserted control rod at zero power conditions of 532*F, 2155 psig shall not exceed 1.0% AK/K. A pseudo ejected control rod worth measurement was performed during the zero power test program to verify the safety analysis calculations relating to the hypothetical ejection of the most reactive control rod.
- b. Test Method Pseudo ejected control rod worths were measured for the rod in core location N-12 using two dif ferent techniques. The first technique was the boron-swap method during which the boron concentrationincreased.
continuously of the reactor coolant system was slowly and The pseudo ejected rod was withdrawn to compensate for the reactivity inserted by the boration and the reactivity change was measured by the reactimeter. The sum of the incremental reactivity changes gives the total worth of the ejected rod. In the second technique (rod swaD method), critical equilibrium conditions were established with the pseudo ejected rod withdrawn at 100%. The ejected red was then inserted into the core by swapping reactivity with group 5. The measured worth of the withdrawn group is taken as the worth of the ejected rod.
- c. Test Results Critical equilibrium conditions were established for the boron-swap measurement with an initial RCS boron concentration of 884 ppm and control rod group 5 at 6.5%, groups 6 and 7 at 0% and group 8 at 37.5% withdrawn. Control rod N-12 was withdrawn to 100% to compensate for borating the reactor ccolant to 957 ppm. The worth of rod N-12 from this measurement was 0.675% AK/K.
In the rod swap method, the reactor was just critical with rod group 5 at 12.5 withdrawn, group 8 at 36.1% and groups 6 and 7 at 0% withdrawn. position. Control rod N-12 was at the 100% withdrawn Under these conditions, rod N-12 was swapped into the core and its resultant reactivity worth was obtained f rom the the movement of group 5. Using the analysis presented in that section, the worth of the ejected rod by the rod swao method is 0.707% aK/K, which compares well with the boron swap result. 1 - 83
i
- d. Conclusions 1 Two different methods were used to measure the pseudo ejected rod worth at zero power, 532*F. The results from the boron swap and the rod drop techniques compare ' favorably. The average for the measured values, 0.691% aK/K, is in excellent agreement with the calculated value of 0.65% AK/K.
0 The Technical Specification requirement that the value not exceed 1.0% AK/K is satisfied. i. i l r i I i t i i i 1 l 1
, 84
._ -, _ _ _ .. _ _ --._ ._ _ . _ _ _ __ _ . . . _ _ . . . . _ . . , . . ~,. _
1 l I 5.4 CORE PERFORMANCE - MEASUREMENTS AT POWER This section presents the results of the physics measurements that were conducted with the reactor at power. Testing was conducted at four major power plateaus, 40%, 75%, 88% and 100% of 2535 megawatts core thermal power, as determined from primary and secondary heat balance measurements. Operation in the power range began on October 6,1985. Power escalations occurred as the required testing at each plateau was successfully completed with the exception of a 1 month hold at 48 and 75% power. Periodic measurements and calibrations were performed on the plant nuclear instrumentation during the escalation to full power. The four power range detector channels were calibrated based upon primary and secondary plant heat balance measurements. Testing of the incore nuclear instrumentation was performed to ensure that all detectors were functioning properly and that the detector inputs were processed correctly by the plant computer. Core axial imbalance determined from the incore instrumentation system was used to calibrate the out of core detector imbalance indication. The major physics measurements performed during power escalation consisted of determining the moderator and power Doppler coefficients of reactivity and obtaining detailed radial and axial core power distribution measurements for several core axial imbalances. Values of minimum DNBR and maximum linear heat rate were monitored throughout the test program to ensure that core thermal limits would not be exceeded. 85
5.4.1 Nuclear Instrumentation Calibration at Power
- a. Purpose The purpose of the Nuclear Instrumentation Calibration at Power was to calibrate the power range nuclear instrumentation indication to within 12% FP of the reactor thermal power as determined by a heat balance and to within 3.5% incore axial offset as determined by the incore monitoring system.
D. Test Method As required during power escalation, the top and bottom linear amplifier gains were adjusted to maintain power range nuclear instrumentation indication within t 2% of the power calculated by a heat balance. During top and bottom linear amplifier gain adjustment the ratio of their gains was maintained constant as long as the indicated axial offset nas within t 3.5% of incore offset; if not, their gains were adjusted to correct imbalance and heat balance mismatch at the same time. When directed by the controlling procedure for physics testing, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given below: Test Plateau Bistable Setpoint 4 FP % FP 40 50 75 85 100 105.5
- c. Test Results An analysis of test results indicated that changes in Reactor Coolant System boron and xenon buildup or burnout affected the power as observed by the nuclear instrumentation. This was as expected since the power range nuclear instrumentation measures reactor neutron leakage which is directly related to the above changes in system conditions. Each time that it was necessary to calibrate the power range nuclear instrumentation, the acceptance criteria of calibration to within 2.0% FP of the heat balance power was met without any difficulty. Also, each time it was necessary to calibrate the power range nuclear instrumentation, the 3.5% axial offset criteria as determined by the incore monitoring system was also met.
The high flux triD bistable was adjusted to 50, 85 and 105.5% FP prior to escalation of power to 40, 75 and 100% FP, respectively.
, 86
- d. Conclusions The power range channels were calibrated to within two percent of heat balance power several times during the startup program. These calibrations were required due to power level, boron, and/or control rod configuration changes during the p rogram. Acceptance criteria for nuclear instrumentation calibration at power were met in all instances.
i
- l d
. 87-
_ _ _ . _ _ _ _ _ . _ - - . . ~ _ - _ . _ - . _ _ . _ _ _ .
5.4.2 Incore Detector Testing
- a. Purpose Self-powered-neutron-detectors (incore detector system) monitor the core power density within the core and their outputs are monitored and processed by the plant computer to provide accurate readings of relative neutron flux.
Tests conducted on the incore detector system were performed to: (1) Verify that the output from each detector and its response to increasing reactor pcwer was as expected. (2) Verify that the background, length and depletion corrections applied by the plant computer are correct.
- b. Test Method The response of the incore detectors versus power level was determined and a comparison of the symmetrical detector outputs made at steady state reactor power of 40 and 75% FP.
Using the corrected SPND maps, calculations were performed to determine the detector current to average detector current values per assembly for each incore detector versus axial positions. Any detector levels which were determined to have f ailed were deleted from scan or substituted. ' At 40 and 100% FP, SP1301-5.3, Incore Neutron Detectors-Monthly Check, was performed to calibrate the back-up recorders to its incore depletion value,
- c. Conclusions Incore detector testing during power escalation demonstrated that all detectors except levels 2 and 3 of string 15 were functioning as expected, that symmetrical detector readings agreed within acceptable limits and that the computer applied correction factors are accurate. Symmetric string 20 was substituted for string 15. The backup incore recorders were calibrated at 40 and 75% FP and operational above 80% FP as required by the Technical Specifications.
, 88
, 5.4.3 Power Imbalance Detector Correlation Test
- a. Purpose The Power Imbalance Detector Correlation Test has four objectives:
- 1. To determine the relationship between the indicated out-of-core power distribution and the actual incore power distribution.
- 2. To demonstrate axial Xenon control using the Axial Power Shaping Rods (APSR's).
- 3. To verify the adequacy and accuracy of backup imbalance calculations as done in AP 1203-7, " Hand Calculation for Quadrant Power Tilt and Core ' Power Imbalance."
4. To determine the core maximum linear heat. rate and minimum DNBR at various power imbalances.
- b. Test Method This test was conducted at 40% FP to determine the relationship between the core axial imbalance as indicated by the incore detectors and the out-of-core detectors. Based upon this correlation, it could be verified that the minimum DNBR and maximum linear heat rate limits would not be exceeded by operating within the flux / delta flux / flow envelope set in the Reactor Protection System.
Specified plant data was recorded at the following target imbalances:
- 1. 7 to 8.5%
- 2. 3.5 to 4%
- 3. -14 to -16%
- 4. -10 to -14%
- 5. !
-5 to -10% !
- 6. -1 to +1%
As CRG-8 was moved to establish the above imbalances, the integrated control system automatically comoensated for j reactivity changes by repositioning CRG-7 to maintain a constant power level.
- c. Test Results The relationship between the ICD and OCD offsets was determined at 40% FP by performing an imbalance scan with the APSR's. The average slope measured on the four out-of-core detectors was 1.18%. The lowest sloDe was 1.12 for NI-7.
The scaled difference amplifier gain was 4.904.
, 89
A comparison of the incore detector (ICD) of fset versus the out-of-core (OCD) detector of fset obtained for each NI channel is shown in Table 4.3-1. Core power distribution measurements were taken in conjunction with the most positive and most negative imbalances at 40% FP and the values of minimum DNBR and worst case MLHR and compared to the acceptance criteria. 4 The worst case values of minimum DHBR and maximum linear heat rate determined at 40% FP are listed in Table 4.3-2. The worst case DNBR ratio was greater than the minimum limit of 1.3 and the maximum value of linear heat rate was less than the fuel melt limit of 20.15 kw/f t af ter extrapolation to 105.5 FP. These results show that Technical Specification limits have been met and that adequate protection is provided by the Reactor Protection system trip setpoints for the allowed axial imbalances during power operation. Backup imbalance calculations using AP 1203-7 agreed with computer calculated imbalances. Table 4.3-3 lists the computer calculated imbalances as well as imbalances obtained using the incore detector backup recorders.
- d. Conclusions Backup imbalance calculations performed in accordance with AP 1203-7 provide an acceptable alternate method to computer calculated values of imbalance. A revised difference amplifier K factor of 5.035 will provide a slope greater than 1.15 when OCD offset is plotted versus ICD offset.
Minimum DNBR and Maximum Linear Heat Rate parameters were well within Tcchnical' Specifications limitations. _ 90
y _ 4 TABLE 4.3-1 INCOREOFFSETVSbuT-OF-COREOFFSET INCORE 0FFSET OUT.-0F-CORE OFFSET (%)' (%) NI-5 NI-6 , _ NI-7 ~ NI-8 7.33 9.44 8.96 3.68 7.91 8.78 5.56 5.26 - 5.03
-47.25 -56.80 5.17
-58.06 -52.85 -54.01
-34.00 -40.43 -40.69
-14.62 -36.91 -37.96
-17.31 -17.14- -15.49 1.03 1.93 ;15.85 2.19 1.86 - 2.26 e
TABLE 4.3-2 j WORST CASE DNBR AND LHR IMBALANCE MINIMUM EXTRAP0l.ATED WORST CASE LHR (%) ONBR EXTRAP. WORST LHR MONBR (XW/FT) __ .(KW/FT) ___
-19.41 7.46 3.589 6.46
~
2.96 9.01 3.893 16.88 5.08 -13.23 TA8LE 4.3-2 FULL INCORE IMBALANCE VS BACKUP RECORDER IMBALANCE FULL INCORE BACKUP RECORDER IMBALANCE IMBALANCE (%) (%) y- '., 2.96 4'21
~
1.58 2.01
-19.41 '
-15.74
-13.73 -11.17
-6.15 -4.72 0.0 0.82 s 9
Y
. 9 tm
.]. d
-,: r 91 .
4
5.4.4 Core Power Distribution Verification
- a. Purpose To measure the core power distributions at 40, 75, 88, and 100 percent full power to verify that the core axial imbalance, quadrant power tilt, maximum linear heat rate and minimum DNBR do not exceed their specified limits. Also, to compare the measured and predicted power distributions.
- b. Test Method Core power distribution measurements were performed at each of the major power plateaus of the test program (40%, 75%, 88%
and 100% full power) under steady state conditions for specified control rod configurations. To provide the best comparison between measured and predicted results, three-dimensional equilibrium xenon conditions were established for all but the 40% FP measurements. Data collected for the measurements consisted of detailed power distribution information at 364 core locations from the incore detector system and the worst case core thermal conditions were calculated using this data. The measured data was compared with calculated results.
- c. Test Results i
A summary of the four cases studied in this report is given in Table 4.4-1 which gives the core power level, control rod pattern, cycle burnup, boron concentration, axial imbalance, maximum quadrant tilt, minimum DNBR, maximum LHR and power peaking data for each measurement. The highest Worst Case i MLHR was 12.81 at 100% FP which is well below the limit of 20.15 kw/ft. l The lowest minimum DNSR value was 3.88 at 100% FP which is well above the limit of 1.30. The quadrant power tilt and axial imbalance values measured were all within the allowable limits. Table 4.4-1 also gives a comparison between the maximum calculated and predicted radial and total peaks for an 1/8 core power distribution.
- d. Conclusions Core power distribution measurements were conducted at 40%,
75%, 88% and 100% f ull power. Comparison of measured and predicted results show good agreement. The maximum dif ference . between a measured and predicted value was <4.1% for total peaking at 100% FP. This met the acceptable criteria of
< 7.5%.
The measured values of DNBR and MLHR were all within the allowaole limits. All quadrant power tilts and axial core imoalances measurea during the power distribution test were within the Technical Specifications and normal operational , limits. 92
TABLE 4.4-1 CORE POWER DISTRIBUTION RESULTS POWER PLATEAU 40% 75% 88% 100% DATE 10-20-85 12-05-85 12-30-85 01-09-86 Actual Power (%FP) 40.71 75.8 87.9 CRG 1-6 100 99.62 (%WD) 100 100 100 CRG 7 85 (%WD) 86.7 86.4 85 CRG 8 (%WD) 25 26 26.6 21 Cycle Burnup (EFPD) 2.56 25.76 44.8 51 .7 5 Boron Conc. (PPM) 914 788 697 656 Imbalance (%) 0.18 -2.96 Maximum Tilt -4.53 -3.56 (%) 1.91 0.88 0.58 1.45 MDNBR 9.81 5.20 4.60 3.88 Worst Case MLHR (KW/FT) 5.224 9.073 12.00 12.81 Maximum Radial Peak Measured 1.291 1.288 Predicted 1.272 1.262 1.298 1.285 1.224 1.224 Difference (%) -0.54 0.23 3.92 Acceptance Criteria (%) 3.10 18% 15% $5% Maximum Total Peak SS% Measured 1.585 1.483 1.460 1.474 Predicted 1.532 1.436 1.416 Difference 1.416 (%) 3.46 3.27 3.32 4.10 Acceptance Criteria (%) 512% 57.5% $7.5% 17.5% I i 93
5.4.5 Reactivity Coefficients at Power
- a. Purpose The purpose of this test is to measure the temperature and power doppler. coefficients of reactivity at power. This information is then used to assure that Tech. Spec. 3.1.7.1, which states tha,t the moderator temperature coefficient shall not be positive at power levels above 95% of rated power, is satisfied.
- b. Test Method For measuring the temperature coef ficient of reactivity, the average RC temperature was increased and then decreased by about 5 Degree F. The reactivity associated with each temperature change was obtained from,the change in controlling rod group position, and the values for the coefficient were calculated.
For measuring the power doppler coefficient of reactivity, reactor power was decreased and then increased by about 5 percent FP. The reactivity change was obtained from the change in controlling rod group position, and the values for the coefficient were calculated. In conjunction with both reactivity coefficient measurements, differential controlling rod aroup worth measurements using the fast insert / withdrawal method were performed.
- c. Test Results At 88 FP, temperature and power doppler coefficient measurements were performed. The moderator temperature coefficient measured at 88% FP was -8.70 pcm/*F. This corresponds to MTC of -8.2 pcm/*F at ARO, HFP conditions and verifies that the moderator tempo: ature coef ficient is negative above 95% FP.
The measured power doppler coefficient at 88 FP was -8.585 pcm/%FP which corresponds to -8.305 pcm/%FP at HFP conditions. This meets the acceptance criteria of being more negative than -5.5 pcm/%FP but, is outside the predicted band of -9.1 to -17.1 pcm/%FP for the test conditions.
- d. Conclusions The measured moderator temperature coefficient (MTC) results indicate that the MTC will be negative above 95% F.P.
The measured power doppler coef ficient (PDC) results meet the l safety analysis requirement that the PDC be more negative than '
-5.5 pcm/%FP. However, this value is outside the band predicted for the test conditions. As a result of discussions with B&W and GPUN Nuclear Fuels Group, is was concluded that measured PDC results at S&W facilities are typically less neaative than those predicted by 8&W. B&W indicated that they place no major significant on this deviation as long as the saf ety analysis criteria is satisfied.
94 I
I 4 GPU Nuclear Corporation
' Nuclear :::ome:r8o s Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
April 24, 1986 5211-86-2070 Dr. Thomas E. Hurley Region I, Regional Administrator U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406
Dear Dr. Murley:
Three Mile Island Nuclear Station Unit 1 (TMI-1) Operating License No. DPR-50 Docket No. 50-289 Startup Report for Restart In accordance with Technical Specification 6.9.1. A, attached is the Startup Report for Restart. The report is considered complete and no supplementary reports are necessary. Sincerely, (
.lg H. D. H k'll Director, TMI-1 HDH/CWS/spb cc: J. F. Stolz Attachment 0560A
.1 GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporatl p'u 0\
$ N'{ 0 9 '
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