ML19276H210
| ML19276H210 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 11/26/1974 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| NUDOCS 7910150519 | |
| Download: ML19276H210 (250) | |
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- Docket No. 50-289 License No. DPR-50 THREE MILE ISLAND NUCLEAR STATION UNIT 1 INITIAL STARTUP REPORT METROPOLITAN EDISON COMPANY SUBSIDI ARY OF GENERAL PUBLIC UTILITIES CORPORATION PREPARED BY GENERAL PUBLIC UTILITIES SERVICE CORPORATION 414 002
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METROPOLITAN EDIS0N C0MPANY THREE MILE ISLAND NUCLEAR STATION UNIT I Docket No. 50-289 License No. DPR-50 INITIAL .STARTUP REPORT November 26, 1974 Prepared by General Public Utilities Service Corporation 1414 003
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. .
TABLE OF CONTEFTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
1.1-1
1.1 INTRODUCTION
1.2
SUMMARY
1.1-1 1.2.1 1.2-1 GENERAL 1.2.2 1.2-1 INITIAL FUEL LOADING 1.2-1 1.2.3 POST FUEL LOAD PRECRITICAL TEST PROGRAM 1.2-1 1.2.4 CORE PERFORMANCE - MEASUREENTS AT ZERO POWER 1.2-2 1.2.5 CORE PERFORMANCE - MEASUREENTS AT POWER 1.2-3 1.2.6 NUCLEAR STEAM SYSTEM PERFORMANCE 1.2-6 1.2.7 BALANCE OF PLANT TESTING 1.2-6 1.2.8 UNIT PERFORMANCE 1.2-8 2.0 INITIAL FUEL LOADING 2.0-1 3.0 POST FUEL LOAD PRECRITICAL TEST PROGRAM 3.0-1 3.1 REACTOR COOLANT PUMP FLOW TEST 3.1-1 3.1.1 PURPOSE 3.1-1 3.1.2 TEST ETHOD 3.1-1 3.1.3 TEST RESULTS 3.1-1 3.
1.4 CONCLUSION
S 3.1-1 3.2 REACTOR COOLANT PUMP FLOW COASTDOWN TEST 3.2-1 3.2.1 PURPOSE 3.2-1 3.2.2 TEST METHOD 3.2-1 3.2.3 TEST RESULTS 3.2-1 3.
2.4 CONCLUSION
S 3.2-1 3.3 CONTROL ROD DRIVE DROP TIME TEST 3.3-1 3.3.1 PURPOSE 3.3-1 3.3.2 TEST METHOD 3.3-1 3.3.3 TEST RESULTS 3.3-1 3.
3.4 CONCLUSION
S 3.3-2 3.4 PRESSURIZER TEST 3.4-1 3.4.1 PURPOSE 3.4-1 3.4.2 TEST ETHOD 3.4-1 3.4.3 TEST RESULTS 3.4-1 3.
4.4 CONCLUSION
S 3.4-1 3.5 REACTOR COOLANT SYSTEM LEAKAGE 3.5-1 3.5.1 PURPOSE 3.5-1 3.5.2 TEST ETHOD 3.5-1 3.5.3 TEST RESULTS 3.5-2 3.
5.4 CONCLUSION
S 3.5-2 4.0 CORE PERMRMANCE - MEASUREMENTS AT ZERO POWER 4.0-1 4.1 INITIAL CRITICALITY 4.1-1 4.2 NUCLEAR INSTRUMENTATION OVERLAP 4.2-1 1
1414 004
.
. .
Section Pm 4.2.1 PURPOSE 4.2-1 4.2.2 TEST METHOD 4.2-1 4.2.3 TEST RESULTS 4.2-1 4.
2.4 CONCLUSION
S 4.2-1 4.3 REACTIVITY CALCULATIONS 4.3-1 4.3.1 PURPOSE 4.3-1 4.3.2 TEST NETHOD 4.3-1 4.3.3 TEST RESULTS 4.3-1 4.
3.4 CONCLUSION
S 4.3-1 4.4 ALL RODS OUT CRITICAL BORON CONCENTRATION 4.4-1 4.4.1 PURPOSE 4.4-1 4.4.2 TEST METHOD 4.4-1 4.4.3 TEST RESULTS 4.4-1 4.
4.4 CONCLUSION
S 4.4-1 4.5 TEMPERATURE COEFFICIENT MEASUREMENTS 4.5-1 4.5.1 PURPOSE 4.5-1 4.5.2 TEST METHOD 4.5-1 4.5.3 TEST RESULTS 4.5-1 4.
5.4 CONCLUSION
S 4.5-2 4.6 SOLUBLE POISON WORTH 4.6-1 4.6.1 PURPOSE 4.6-1 4.6.2 TEST METHOD 4.6-1 4.6.3 TEST RESULTS 4.6-1 4.
6.4 CONCLUSION
S 4.6-1 4.7 CONTROL ROD GROUP WORTH MEASUREMENTS 4.7-1 4.7.1 PURPOSE 4.7-1 4.7.2 TEST METHOD 4.7-1 4.7.3 TEST RESULTS 4.7-1 4.
7.4 CONCLUSION
S 4.7-2 4.8 EJECTED CONTROL ROD WORTH 4.8-1 4.8.1 PURPOSE 4.8-1 4.8.2 TEST METHOD 4.8-1 4.8.3 TEST RESULTS 4.8-1 4.
8.4 CONCLUSION
S 4.8-1 4.9 SHUTDOWN MARGIN 4.9-1 4.9.1 PURPOSE 4.9-1 4.9.2 TEST METHOD 4.9-1 4.9.3 TEST RESULTS 4.9-1 4.
9.4 CONCLUSION
S 4.9-2 5.0 CORE PERFORMANCE - MEASUREMENTS AT POWER 5.0-1 5.1 BIOLOGICAL SHIELD SURVEY 5.1-1 5.1.1 PURPOSE 5.1-1 5.1.2 TEST METHOD 5.1-1 5.1.3 TEST RESULTS 5.1-1 5.
1.4 CONCLUSION
S 5.1-1 5.2 NUCLEAR INSTRL M ATION CALIBRATION AT POWER 5.2-1 5.2.1 PURPOSE 5.2-1 5.2.2 TEST METHOD 5.2-1 5.2.3 TEST RESULTS 5.2-2 5.
2.4 CONCLUSION
S 5.2-2 11 1414 005
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Section Page
, 5.3 INCOREsDETECTOR TESTING 5.3-1 5.3.1 PURPOSE
'
5.3-1 5.3.2 TEST ETH0D 5.3-1 5.3.2.1 Incore Detector System 5.3-1 5.3.2.2 Incore Detector Tests 5.3-2 5.3.3 TEST RESULTS 5.3-2 5.
3.4 CONCLUSION
S 5.3-3 5.4 POWER IMBALANCE DETECTOR Coli g ION TEST 5.4-1 5.4.1 5.4-1
-
PURPOSE 5.4.2 TEST METHOD 5.4-1 5.4.3 TEST RESULTS 5.4-3 5.
4.4 CONCLUSION
S 5.4-4 5.5 ROD WORTH AT POWER 5.5-1 5.5.1 PURPOSE 5.5-1 5.5.2 TEST ETHOD 5.5-1 5.5.3 TEST RESULTS 5.5-1 5.
5.4 CONCLUSION
S 5.5-1 5.6 REACTIVITY COEFFICIENTS AT POWER 5.6-1 5.6.1 PURPOSE 5.6-1 5.6.2 TEST ETHOD 5.6-1 5.6.3 TEST RESULTS 5.6-2 5.
6.4 CONCLUSION
S 5.6-2 5.7 DROPPED CONTROL ROD TEST 5.7-1 5.7.1 PURPOSE 5.7-1 5.7.2 TEST ETHOD 5.7-1 5.7.3 TEST RESULTS 5.7-2 5.
7.4 CONCLUSION
S 5.7-3 5.8 PSEUDO CONTROL ROD EJECTION TEST 5.8-1 5.8.1 PURPOSE 5.8-1 5.8.2 TEST METHOD 5.8-1 5.8.3 TEST RESULTS 5.8-1 5.
8.4 CONCLUSION
S 5.8-2 5.9 CORE POWER DISTRIBUTIONS 5.9-1 5.9.1 PURPOSE 5.9-1 5.9.2 TEST ETHOD 5.9-1 5.9.3 TEST RESULTS 5.9-2 5.9.3.1 Steady State Power Distributions 5.9-2 5.9.3.2 Minimum DNBR and Maximum LHR Calculations 5.9-2 5.9.3.3 Ouadrant Power Tilt and Axial Power Imbalance 5.9-4 5.
9.4 CONCLUSION
S 5.9-4 5.10 NUCLEAR STEAM SYSTEM HEAT BALANCE 5.10-1 5.10.1 PURPOSE 5.10-1 5.10.2 TEST ETHOD 5.10-1 5.10.3 TEST RESULTS 5.10-2 5.
10.4 CONCLUSION
S 5.10-2 5.11 REACTIVITY DEPLETION VERSUS BURNUP 5.11-1 5.11.1 PURPOSE 5.11-1 5.11.2 TEST ETHOD 5.11-1 5.11.3 TEST RESULTJ 5.11-1 5.
11.4 CONCLUSION
S 5.11-1 5.12 NEUTRON NOISE MEASUREMENTS 5.12-1 m
1414 006
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Section Page 6.0 NUCLEAR STEAM SYSTEM PERFORMANCE 6.0-1 6.1 REACTOR COOLANT SYSTEM PERFORMANCE 6.1-1 6.1.1 PURPOSI 6.1-1 6.1.2 TEST ETHOD 6.1-1 6.1.3 TEST RESULTS 6.1-1 6.1.3.1 Steady State Operation 6.1-1 6.1.3.2 Reactor Coolant System Transients t,.1-2 6.1.3.3 Reactor Coo] ant Pump Performance 6.1-4 6.1.3.4 Reactor Coolant System Leakage 6.1-4 6.
1.4 CONCLUSION
S 6.1-4 6.2 AUXILIARY SYSTEM PERFORTANCE 6.2-1 6.2.1 RADIOACTIVE WASTE MANJ'EMENT 6.2-1 6.2.2 PRIMARY AND SECONDARY YSTEM WATER "HEMISTRY 6.2-2 7.0 BALANCE OF PLANT TESTINu 7.0-1 7.1 TURBINE GENERATOR OPERATIONAL TESTING 7.1-1 7.1.1 PURPOSE 7.1-1 7.1.2 TEST ETHOD 7.1-1 7.1.3 TEST RESULTS 7.1-2 7.
1.4 CONCLUSION
S 7.1-3 7.2 TURBINE BYFASS SYSTEM TEST AND MAIN STEAM SAFETY VALVE OPERATION 7.2-1 7.2.1 PURPOSE 7.2-1 7.2.2 TEST ETHOD 7.2-1 7.2.3 TEST RESULTS 7.2-1 7.
2.4 CONCLUSION
S 7.2-2 7.3 FEEDWATER SYSTEM OPERATION AND TESTING 7.3-1 7.3.1 PURPOSE 7.3-1 7.3.2 TEST ETHOD 7.3-1 7.3.3 TEST RESULTS 7.3-1 7.
3.4 CONCLUSION
S 7.3-3 7.4 _E_MERGENCY FEEDWATER SYSTEM OPERATION AND TESTING 7.4-1 7.4.1 PURPOSE 7.4-1 7.4.2 TEST ETHOD 7.4-1 7.4.3 TEST RESULTS 7.4-1 7.
4.4 CONCLUSION
S 7.4-1 7.5 POWER ESCALATION CHECKPOINTS 7.5-1 7.5.1 PURPOSE 7.5-1 7.5.2 TEST ETHOD 7.5-1 7.5.3 TEST RESULTS 7.5-2 7.
5.4 CONCLUSION
S 7.5-3 8.0 UNIT PERFORMANCE 8.0-1 8.1 UNIT TRANSIENT RESPONSE 8.1-1 8.1.1 PURPOSE 8.1-1 8.1.2 TEST ETHOD 8.1-1 8.1.3 TEST RESULTS 8.1-2 iv \4
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Section M
8.1.3.1 Unit Load Transient Tests During Escalation 8.1-2 8.1.3.2 Reactor Coolant Pump Trio at 25% FP 8.1-3 8.1.3.3 Main Feedwater Pump Trip at 100: FP 8.1-3 8.1.3.4 Asymmetric Rod Runback 8.1-4 8.1.3.5 Turbine Trip From 76% FP 8.1-4 8.1.3.6 Generator-Reactor Trio From 100: FP 8.1-4 8.
1.4 CONCLUSION
S 8.1-5 8.2 L_OSS OF OFFSITE POWER 8.2-1 8.2.1 PURPOSE 8.2-1 8.2.2 TEST METHOD 8.2-1 8.2.3 TEST RESULTS 8.2-1 8.
2.4 CONCLUSION
S 8.2-2 8.3 SHUTDOWN FROM OUTSIDE THE CONTROL ROOM 8.3-1 8.3.1 PURPOSE 8.3-1 8.3.2 TEST METHOD 8.3-1 8.3.3 TEST RESULTS 8.3-2 8.
3.4 CONCLUSION
S 8.3-3 8.4 UNIT ACCEPTANCE TEST 8.4-1 8.4.1 PURPOSE 8.4-1 8.4.2 TEST METHOD 8.4-1 8.4.3 TEST RESULTS 8.4-2 8.
4.4 CONCLUSION
S 8.4-2 1414 108 V
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1.0 INTRODUCTION
AND
SUMMARY
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.
,
1.1 INTRODUCTION
Three Mile Island Nuclear Station Unit I was issued operating license DPR-50 on April 19, 1974 and the first fuel assembly was inserted into the core on April 20, 1974. Fuel loading was completed on April 25, 1974. Initial Criticality was achieved on June 5, 1974, upon completion of a Post Fuel Load Pre-Critical test program.
Zero Power Physics testing began on June 5, 1974 and was completed on June 10, 1974. The zero power measurements were performed at a Reactor Coolant System temperature of 532 F and a pressure of 2155 psi.
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Initial escalation of the reactor above zero power commenced on June 15, 1974 and the turbine generator was initially loaded on June 19, 1974. Further increases in power level were made as testing was successfully completed at each of the four major power plateaus defined in the jower escalatica sequence. The four major power levels and the dates they wr.re first attained are as follows:
Power Level Date 15: June 16, 1974 40 June 29, 1974 76% July 14, 1974 100% August 3, 1974 The power escalation test program was completed with successful performance of the unit acceptance test on August 26, 1974 and on September 2, 1974, TMI Unit I was declared commercial. Figure 1.0-1 gives the power history of Unit I from Initial Fuel Loading to the completion of startup testing.
This report is submitted in accordance with Technical Specification 6.7.3 and summarizes startup test program results and unit performance from fuel loading through 100% full power operation as of 1200 on August 27, 1974. The integrated burnup on the core at this date and time was 26.3 effective full power days (EFPD).
Figure 1.0-2 shows the integrated core burnup for core 1.
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1.2
SUMMARY
f 1 2.1 GENERAL Three Mile Island Unit I commenced connercial operation on September 2,1974 at a rated full power output of approximately 861 MWe (gross). The nuclear steam supply system was designed by the Babcock and Wilcox Company and was the third in a setles of systems to be put into service. The tandem compound turbine gener: tor was supplied by the General Electric Company.
The unit het been operated at power levels up to and including 100% FP since the completion o. startup testing. In general, the performance of the unit has been satisfactory. Testing and operation of the nuclear steam supply system and the turbine-generator revealed a few cor.ditions that were other than predicted and none which adversely affected plant safety. The problems enecentered were of a nature that would be expected during the startup of a unit of this size. An evaluation of the unit startup and Power Escalation Test Program results concluded that the unit can be safely operateu at full rated power.
A chronological log of the unit startup, beginning with post fuel load filling and venting of the Reactor Coolant System is given in Figure 1.2-1. A summary of
, the startup and power escalation test results addressed by the major sections of L.is repcet is given below.
1.2.2 INITIAL FUEL LOADING
- Initial Fuel Loading began on April 20, 1974 and was completed on April 25, 1974.
Loading of the core was accomplished under semi-dry conditions with the reactor vessel water level maintained six inches to three feet below the vessel flange.
One major delay occurred during the loading sequence when power was lost to the main fuel handling bridge. Overall, Initial Fuel loading was completed in less than five days and was conducted in a safe and orderly manner.
1.2.3 POST FU"., LOAD PRECRITICAL TEST PROGRAM Following Initial Fuel Loading and prior to Initial Criticality, a Post Fuel Load Pre-Critical Test Program was conducted from April 26, 1974 to June 4, 1974.
During this period, the Reactor Coolant Pump Flow and Flow Coastdown Tests, Control Rod Drive Drop Time Test, Reactor Coolant System Leakage and Surveillance Procedure Verification Test and Pressurizer Operational Test were conducted. In all cases, the applicable Technical Specification requirements and test acceptance criteria were met. A brief su= mary of each test follows:
(a) Reactor Coolant Pump Flow Test Reactor coolant (RC) flow measure =ents were conducted at 512 F, 2155 psi with the _
core installed and measured flow rates were within the range of maximum and minimum acceptable values for all RC, pump combinations tested. RC flow with 4 pumps
' operating is 146.0x10 lbm/hr or 111.2% of the design flow rate.
,
(b) Reactor roolant Pump Flow Coastdown Test The reactor coolant flow coastdown characteristics were measured at system condi-tions of 532 F, 2155 psi with the core installed and met flow decay criteria for all RC pump combinations tested. RC flow decays to 66.5x106 lbm/hr in 10 see when all 4 RC pumps are tripped.
1414 016 1.2-1
. . .
(c) Control Rod Drive Drop Time Test Control rod drop time measurements were conducted at 1500 F and 532 F under flow and no flow conditions with the core installed to ensure that the control rod assembly trip insertion times from 100% withdrawn to three-fourths insertion will not exceed 1.40 seconds under reactor coolant no flow conditions and 1.66 seconds with reactor coolant flow. All acceptance criteria limits were met.
(d) Pressurizer Operational Test The pressurizer sprar flow was set to 190.5 gpm and t."e spray bypass flow was set to 0.99 gpm. noch settings were well within the ac eptance criteria limits of 190.0 +19/-6 gpm and 1.0 +0.5/-0.25 gym, respectively.
(e) Reactor Coolant System Leakage Test and Surveillance Procedure Verification Reactor Coolant System hot leakage measurements were conducted during the hot functional and post fuel load pre-critical test programs. Measured results verify that the unidentified reactor coolant leakage does not exceed the Technical Specification limit of 1.0 spm and that the nor=n1 control instrumentation is sensitive enough to perform leak rate measurements. The value of normal evapora-
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tive losses, as used in Technical Specification 3.1.6.2, was established as 0.51 gpm.
1.2.4 CORE PERFORMANCE - MEASUREMENTS AI ZERO POWER Core performance measurements were conducted during the Zero Power Test Program l which best.n on June 5,1974 and ended on June 10, 1974. This section presents a summary of the zero power measurements. In all cases, the applicable test and Technical Specifications limits were met.
(a) Initial Criticalicy Initi;. .riticality was achieved on June 5,1974 at reactor conditions of 532 F and 2155 psig. Control rod groups 1 through 6 and 8 were withdrawn to 100% and group 7 was positioned at 75% withdrawn. Criticality was achieved by deborating the reactor coolant from 2086 ppm to 1545 ppm at an average deboration rate of 73 ppu per hour. Initial criticality was acnieved in an orderly manner and good agreement was found between the measured and predicted critical boron concentra-tions of 1609 ppm and 1625 ppm, respectively.
(b) Nuclear Instrumentation Overlap At least two decades overlap was measured between the source and intermediate ranges and the linearity and absolute output of the intermediate and source range detectors were within specifications.
(c) Reactivity Calculations An on-line functional check of the reactimeter( }was performed after initial criticality. Reactivity calculated by the reactimeter was within +2: of the core reactivity determined from doubling time measurements.
____
(1) A discussion of the reactimeter is given in section 4.3.
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(d) All Rods Out CriticL1 Boron Concentration The measured all rods out critical boron concentration of 1617 ppm was in excellent agreement with the predicted value of 1634 ppm.
(e) Temperature Coefficient Measurements The measured te=perature coefficients of reactivity at 532 F, zero power were well within the acceptance criteria limit of dO.4x10-iak/k/0F over the range of boron concentrations that the measurements were made.
(f) Soluble Poison Worth The measured results for the soluble poison differential worth at 5320F were within 1.25% of the predicted values.
(g) Control Rod Group Worth Measurements The measured results for the differential and integral control rod group reactivity worths conducted at zero power, 5320F using the boron / rod swap and rod drop techniques were in good agreement with predicted worths. Tne maximum deviation between measured and predicted worths was 8.33%.
(h) Ejected Control Rod Worth Two methods were used to measure the pseudo ejected control rod worth at zero power, 532 F. The results from the boron-swap and rod drop techniques were in good agree-ment. The best estimate for the measured result was 0.688%Ak/k from the boron swap method and the Technical Specification limit of 1.0%Ak/k was not exceeded.
(1) Shutdown Margin Elnimum shutdown margin verification and stuck control rod worth measurements were completed at the zero power, 5320F condition. The measured value of the most reactive control rod stuck out of the core with all other control rods inserted was 3.84:ak/k. The shutdown cargin available under this condition was at least 1.8% ak/k, which ensures that the Technical Specification limit is satisfied.
1.2.5 CORE PERFORMANCE - MEASUREMENTS AT POWER This section presents a summary of the physic; measurements that were conducted with the reactor at power. Testing was cond.~cted at the four major power plateaus of 15%, 40%, 76% and 100% of 2535 megawatts thermal core power, as determined from primary and secondary calorimetric measurements. Operation in the power range began on June 15, 1974.
Periodic measurements and calibrations were performed on the plant nuclear instru-mentation during the escalation to full power. The four power range detector channels were calibrated based upon primary and secondary plant heat balance mea-surements. Testing of the incore nuclear instrumentation was performed to ensure that all detectors were functioning properly and that the detector outputs were processed correctly by the plant computer. Core axial imbalance determined from the incere instrumentation system was used to calibrate the out of core detector imbalance indication. Radiation surveys of the biological shield and reactor and auxiliary buildings were conducted to obtain base line data on accessible work areas while the reactor is operating at power.
1.2-3 }4}k bkO
, . . -
The major physics measurements performed during power escalation consisted of
-
determining the moderator and power Doppler coefficients of reactivity, determining the worth and associated power distributions effected by simulated dropped and ejected control rods, and obtaining detailed radial and axial core power distribu-tion measurements for several core axial imbalances. Values of minimum DNBR and
=H== linear heat rate were monitored throughout the test program to ensure that core thermal limits would not be exceeded.
(a) Biological Shield Survey Radiation levels in all access!ble locations of the plant adjacent to the biological shield were measured. The max bum radiation levels found in all accessible areas were below 100 mrem /br. and the biological shield meets all design criteria.
(b) Nuclear Instrumentation Calibration at Power The power range channels were calibrated as required during the startup program to indicate within two percent of the total core power as determined by a primary or secondary plant heat balance. These calibrations were recuired due to power level, boron and/or control rod configuration changes during testing. The acceptance criteria were met in all instances.
,
(c) Incore Detector Testing Tests conducted on the incore detector system demonstrated that all detectors were functioning as expected, that symmetrical detector readings agreed within accentable limits and that the plant computer applied the correct background, length and deple-tion correction factors.
(d) Power Imbalance Detector Correlation The results of the Axial Power Shaping Rod (APSR) scans performed at 40% and 76% FP show that an acceptable incore versus out-of-core offset relationship is obtained by using a gain. factor.of.4.033 in.the. power range scaled difference amplifiers.
The measured values ef. minimum DNBIL and maximum linear heat rate for various axial core imbalances indicate that the React _or Protection Trip Setpoints provide ade-quate protection to.the core. Imbalance calculations using the backup recorder provided a reliable alternate to conputer calculated values.
(e) Rod Reactivity Worth Measurements Differential control rod reactivity worth measurements were performed in conjunction with the reactivity coefficients and pseudo ejected rod tests. The u.easured rod worths agreed with the design values well within the acceptance criteria limits of
+20%.
(f) Reactivity Coefficients at Power The measured results at 40%, 76% and 100% FP indicate that the moderator coefficient of reactivity will be negative during operation above 95% FP. The power doppler coefficient measurements indicate that the least negative value is -0.00710%Ak/k/% FP at full power conditions.
1414 019 1.2-4
,. . -
(g) Dropped Control Rod Test
~
The dropped control rod test performed at 40% end 76% FP met all required accep-tance criteric and the following conclusions were drawn as a result of the measurements:
(1) The 'cre power distributions and thermal conditions that developed from a control rod dropped into the core during operation at power showed adequate margins to minimum DNBR a' tum linear heat rate limits. The measured worth of the dropped rod was 0.094%Ak/k.
(2) Quadrant power tilt calculations using the backup recorder were accurate in comparison with the computer calculated values.
(3) The rod drive control system properly responded to an asymmetric rod condition.
(h) Pseudo Ejected Control Rod Test The measured worth of the pseudo ejected control rod i
- the rod swap technicue was 0.278%Ak/k, which is well below the Technical Specification limit of 0.65%Ak/k.
The measured values of maximum linear heat rate and minimum DNBR were 13.12 kw/ft and 4.85, respectively, with the ejected rod at 100% withdrawn. The maximum mea-sured radial power peak was 2.38 in the fuel assembly centaining the ejected rod. Sub-stantial margins were observed to core thermal limits in a. pseudo ejected rod accident.
(1) Core Power Distributions Core power distribution measurements were conducted at 15%, 40%, 76% and 100%
full power rod under steady configurations. state equilibrium Comparison xenondistributions of the measured conditions with for specified contg the PDQ-07 results shows good agreement. For the three cases studied at 40%, 76% and 100%
full power, the three largest measured and calculated radial peaks were chosen.
In each case, the measured values were within 8% of the calculated results.
The cesults of the minimum DNBR and m M ann LHR analyses are given in Table 5.9-6.
The margins to the minimum DNBR limit of 1.55 and the maximum LBR value of 17.1 kw/ft were 109% and 42%, Tespectively, after extrapolation to 112% TP. All quadrant power tilts and axial core imbalances measured during the power distribution tests were within the Technical Specification and normal operational limits.
(j) Nuclear Steam Supply System Heat Balance Good agreement was found between hand and computer calculated heat balances during power escalation. Preliminary calculations of totai reactor coolant flow based upon heat balance results indicate a flow rate of 108.6% of design at 100% FP.
(k) Reactivity Depletion Versus Burnup The measured critical boron concentration at 22.0 EFPD and 100% FP conditions was within 30 ppm of the predicted result and well within the acceptance criteria limit of 86 ppm.
(1) PDQ-07 is the analytical model used by Babcock and Wilcox for core design studies.
1.2-5 j 4 } 4 l120
,. . .
(1) Neutron Noise Measurements
.
Neutron noise data was recorded on the TMI Unit I Core during the startup test program to serve as baseline data for future periodic measurements. Initial analysis of the data indicates no major differences from the expected neutron noise signatures.
1.2.6 NUCLEAR STEAM SYSTEM PERFORMANCE A summary of the testing performed during power operation to monitor the performance of the nuclear steam system is presented below. The test results presented include reactor coolant system steady state and transient operation, reactor coolant pump performance, radioactive waste management and primary and secondary system water chemistry.
(a) Reactor Coolant System Performance Steady state operation of the reactor coolant system and the steam generators was monitored at various power levels during the escalation to 100% FP. The average values for reactor coolant inlet, outlet and average temperature; steam generator pressure, temperature and level; and feedwater flow and temperature followed the expected response with power. The response of the reactor coolant system to major unit transients has been satisfactory. One area that is under study is the low pressurizer level reached during a reactor trip. The reactor coolant pumps have performed well and produce flows in excess of their design values. Reactor coolant system leakage was maintained within the limits specified in the Technical Specifications.
(b) Auxiliary System Performance Radioactive vastes generated during power operation consist of liquid, gaseous and solid wastes. The vastes generated during the startup program were adequately processed, stored and/or disposed using plant and off-site facilities in accordance with the plant Technical Specifications. Primary and secondary system water chemistry have been maintained within the limits allowable for operation at power.
Radiochemistry analyisis of reactor coolant activity indicated that no fission product releases occurred during the startui test program.
1.2.7 BALANCE OF PLANT TESTING This section presents a summary of the results of balance of plant testing, adjust-ments and operation at power. Balance of plant systems consist mainly of the turbine generator, =ain steam, turbine bypass, atmospheric dump, condensate, feed-water, moisture separator, steam extraction and feedwater heating, henter drain, emergency feedwater, and cooling water systems. The cooling water systems consist of the circulating water, natural draft cooling tower, intermediate cooling water, nuclear service closed cooling water, nuclear service river water, secondary service closed cooling water, secondarv service river water and mechanical draft cooling tower systems.
A summary of the testing performed on the above systems is given below and includes the test results of the Turbine Generator Operational Test, the Turbine Bypass System Test and Main Stean Safety Valve Operation, Feedwater System Operation and Testing, Emergency Feed System Operation and Testing and Power Escalation Checkpoints.
1.2-6 1414 021
.. . .
(a) Turbine Generator Operational Testing
.
Turbine generator (TG) performance was verf satisfactory throughout the startup test program. Approximately 7 days of testing time were lost due to unscheduled turbine trips and turbine related problems. Water ingestion into the turbine through the 4B heater extraction line due to the isolation valve failure to close on high shell-side level caused by ruptured tubes was the only unanticipated startup problem which could have led to major damage and delays; however, subse-quent turbine operation indicates that the turbine suffered no damage. #3 bearing vibration is approximately 0.5 mils higher than acceptable for long term operation; balancing operations will be performed at the first convenient outage. TG output at 2535 Et is 861 We, when conservatively corrected to design vacuum conditions and compares well with a guaranteed vlaue of 837 NWe. Steam conditions are 10,621 000 #/hr at 5920F compared with design of 11,158,286 #/hr at 559 F. Due to the increased amount of superheat over design, the turbine operates at less than
" valves wide open" conditions. Gross heat rate is 9993 Btu / Kwhr compared with design of 10,002 Btu / Kwhr.
(b) Turbine Bypass System and Main Steam Safety Valve Operation Acceptable response of the turbine bypass valves in maintaining turbine header pressure setpoint and response to small changes in setpoint at reactor powers >15:
was attained af ter the difference between steam generator pressures was included
,
in the control system and a wiring reversal error was corrected. Peak to peak oscillations are +6 psi.
The turbine bypass valves, along with the main steam safety valves, function adequately to limit main steam pressure during turbine trips to 11100 psia. The longest valve opening tima was 2.1 seconds; peak steam pressure following the 100%
generator / turbine trip was 1082 psia.
Operation of the atmospheric dump valves was not required to limit main steam pressure to 31100 psia during the loss of offsite power test.
Final settings of the main steam safety valves appear adequate for cor;tinued plant operation; however, safety valve operation is one of several areas under study in an attempt to optimize plant response to major transients.
(c) Feedvater System Operation and Testing The condensate, feedwater, moisture separator, heater drain, feedwater heating, and steam extraction systems function acceptably to support steady state and transient operation at 100% power. Oscillations and transient response associated with these systems are acceptable; however, investigations are continuing in several areas in an effort to further optimize plant response. These areas are:
(1) Heater drain pump discharge valve control.
(2) Heater drain pump recirculation valve control.
(3) Ability of the feedwater pumps to supply feedwater to the steam generators when turbine header pressure increases rapidly.
(4) Ability .f the feedwater control valves to respond to changes in valve differ-ential pressure.
1.2-7
.. . .
.
(d) Emergency Feedwater System Operation a?d Testing With the amount of decay heat present durint; performance of the loss of offsite power test, the turbina driven emergency feedwater pump provided more than adequate flow to control RCS temperature and pressure. The EW valves had to be throttled to keep from exceeding RCS cooldown limits as steam generator levels began in-creasing to 95% on their operating range level indication. A setpoint of 50%
instead of 95% will be used to adequately remove decay heat without exceeding cooldown rate limitations.
(e) Power Escalation Checkpoints The secondary service closed cooling water adequately cooled its heat loads; SSCCW heat exchanger discharge temperatures were well below their design limit of 950F at 100% power.
The mechanical draft cooling tcuer effluent temperature and differential temperature between influent and effluent had to be controlled manually because the automatic controller was inoperative. Acceptao u operation could be obtained with continuous surveillance; however, until operators gained familiarity with tower operation, differential temperature limits were violated several times. Final testing of the cooling towers will be conducted at a later date.
The natural draft cooling towers sere performance tested under Summer conditions.
Capacity was determined as 104.1% of design. April and December performance tests will be conducted at a later date.
,
Powdex edlut.t chemistry analysis demonstrates acceptable capability to clean up the condensate system during 100% power operation.
1.2.8 UNIT PERFORMANCE A summary is presented in this section of the tests performed during and after escalation to 100% FP which ueasure the overall performance of the unit under normal operating, transient and emergency conditions. A summary of unit response to planned and unplanned major loed changes is presented in the section on Unit Transient Response. The Loss of Offsite Power and Shutdown From Outside the Control Room tests demonstrated the ability to safely control the unit under emergency conditions. The Unit Acceptance Test verified that the nuclear steam supply system can operate in accordance with the warranted design specifications. In all in-stances, eafe operation of the unit was demonstrated and the applicable Technical Specifications requirements weretmet.
(a) Jnit Transient Response Transient testing of the unit was conducted at specified ramp rates in the tutbine following, reactor /stea= genetator following and fully integrated modes of control at 40%, 76% and 100% of full power. The Integrated Control System (ICS) successfully maneuvered the plant.in all three modes of control.during the 40% and 76% FP tests.
In the 100%--50%-100% transient, the 10%/ min design ramp rate was accomplished within acceptable limits in the fully integrated ICS mode. The transient was completed at 8%/ min in the turbine following.= ode and at 6%/ min on the decrease and 4%/ min on the increase in the steam. generator / reactor following mode. The ability of the ICS to control the plant during a loss of reactor coolant pump, a loss of a main feedwater pu=p and a dropped control rod transient was satisfactory.
\k
.. . .
-
The reactor was successfully runback to 15: FP during the turbine trips from 76% FP. The reactor tripped on high RC pressure af ter a full load rejection at 100: FP.
(b) Loss of Offsite Power The station emergency blackout procedure was verified for use during a blackout and all required automatic action occurred as expected. RCS temperature and pressure remained well within their respective limits and no increase in fission product activity occurred as a result of this test.
(:) Shutdown From Oucside The Control Room The reactor plant can be maintained in a hot shutdown condition from locations outside of the main control room by a normal shift compliment. The alternate control center contains sufficient instrumentation and communications to permit satisfactory monitoring and direction of shutdown operations.
(d) Unit Acceptance Test The Three Mile Island Unit I nuclear steam supply system produces 2552.615 We gross energy output compared with the warranty value of 2449 s t. Main steam temperatureis 591.60F compared with the warranty of 5690F. These results indicate a substantial margin of NSS performance above warranty specifications.
1414 024 1.2-9
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=
_5 t 7/15 - CE 24 hr. restriction to Roll 8/12-Rz. at so:-treawient teattag ta C b TG/ Tube last repair in 4th stage p rosre as .
f== h'~ ; htr. in prog./MEC compt. S/U's 8/13-Rz. at 951/TF 800/34-TG trip-Az.
f g g (Traintag), trapped on bi RC press / plant cooldown/
c:-
-*
!
7/16 - Comm. 70 Roll 90530/Rs. to 762 F: s t. at hot s/D.
CE CIV Test Sat. 8/14-Plot in cooldown 260*F C345FSI/
- D 7/17 - IEC 1k. fCTf 3 f arced S/D - c e senced CIV screen rom. outage
,
a t N
_E.
comp 1. 30: TC ?;1p fl100/EBC Ik. repaired - returned to 76%.
7/18 - Ra. 9763 - Avaiting 3-D Equilb.
pint s/D/RC press 480 peig/ temp.145*F.
8/19-Sec. plat. vtr. chen. cleanup is progrese/ making prope for S/U.
2
- g sence. 8/20-Flant S/U delayed by saal inj. filter
% 4 7/19 - EBC fluid ik. 90320 - tripped leakage - Crouse wks. to repair /CRD TG - Ex. to 152 for EEC ik. & RCS venting in progress /coun. E/17
- ) repair /wks. MS hgrs./Rs. to 651 to 523'-
. $ 5 hr fuel soak /to 761. 8/22-iner. temp. to 34 /presa to 1500peig 9 _$ 7/20 - TC trip 01030 during TC over- 8/23-Res. temp. to 531)*F/Presa. to 2155pota
'>
- E opeed trip test - T; controla Rz. crit 40850-taer. pwr. to 13:.
C w g malfunction / recovered #1130 - 8/24-Rz. to 802 perr.
,, e Rz. to 80: - TG on line/ set M.S. 8/25-Rz to 902 pur.
. 2 Reliefs (RV-1 thru 6) . S/26-34. 4100% pur. 60345/c-ed U.A.b
=% 01000, com61. 01700/Rz. ehutdarn to g
- O 7/21 - Ex. 0761-/ Plant S/D 02300 for MEC AEC exaas aad S/U's. repair miac. leaka/ dept. te 1600peia, d
- I 7/22 - Ra 610"8 ampe (11) - Coll. date temp. to 400*F.
A } for E&W on sanon transient /ande 8/27 Rapaired leaka 4 comm. prepa. for
~7 *"""" insp. of Rx Sids. - cleaned up return to pwr.
~
""
=2{
g @M 4 eil under RC-PIC & ID/ working outage items.
8/281030-come. Rz. S/U/Rz. crit. 91047/
1629-T. trip on ov.-speed 6 ash, hood m thru?/23 - Ra enutdown 40300 for AEC auma/ temp. h1/reest turb. 4 syn./to 303 pwr.
4 -$ gQ y 7/24 Outage work items in progress. 2200 - FW esp. stra S. laak-reduced
% ; E 7/25 - Kinor delev on S/U due to 02 pur. to 202 pur.
~ M problem "A" cond. drn. viv.
M
}
- DcC was lef t open durios maint./tripp-4/29-0106-inct. pur, to 401 0920-Rz. to 602-500 Ni A > ed turbine due to 2 testing from 1200-iner. to 701 pvr.
E i M 1800 RFM/ excessive vib. on fil 8/30-7310 T. trip on til bear, vib /1312 az.
t: i [ bearing (7.5 mila) - Tuxt. off . rip on press /tese.
7 ~a line for repair /Rx to 70% pur.
02130 - comp 1. 5-br. soak period.
8/31-0135-comm. Rz. S/U 7 g 0720-Rz. 03:
.:::,- 7/26 - R2. to 762 00300 TG on line - 1840-Syn. m. turb. to 1800 RFM &
.E , -5 avait. 3D renon equilib. iner. Rx. pvr. to 402.
D _t 7/27 - Rx 0761 - TG on line/ comp 1. rod 9/1-0013-Rz. 070%
"71' worth & reactivity coeff. 0904-Rz. 0802
"::"" h; 7/28 - Reduction in FW Flow due to low 9/2-0001-TMI 1 commercial 000 Ex. fl002
-d- level in Borwell - reduced pvr. pur. 91100
- 26. to 402 "C" cond, boost. pump 1g -5 mech. bound up - holding pwr. at 40-502 for repair CD-P2C.
,,
""" 4 NOTES :
"C" 1.
- 7/29 - Ra. at 462 - Waiting CO-72C repair /incr. to 761-6th stage 1 - START POST FUEI,
-
- :
" SCEED. 78 DAYS l ED tk. viv. cont. vive, caused 14AD PRE-CRITICAL AL"!UAL 120 DAYS
(/) w n ICS oscillations - rumback TESTING TERU COMM-M O t ,, occurred - Ret. to 762. ERCIAL OPERATION.
U E 7/30 - Rz. at 76: - Attai.ed .enon 1 Equilib. 6 "0" imbalance; also 2 - INIT. CRIT. TO SCHED 49 DAYS yyi $
m NI adjust./atarted pwr. imbal. 1003 F0WER OPER. ACTUAL 59 DAYS 7/31 - Rz. at $5: due to savnetric tod y ? rumback (8-8) - blown fuse / ret. 3 - START 100 POWER 3CgED. 19 DAYS
- to 762 - computar failure /"Aa TESTING THRU COMM- ACTCA129 DAYS iN K condenser side tube leak - che ERCIAL CPERATION. (9 0F TRESE 29
-e istry probless a arced per.reduct. WERE ATTRIBUTED
$ to 50%. TO THE TURE.
r-
- 8/1 - Rz. at 502 - leaks in "A" Cond. PLANT SCREEN Repaired (lose 524 Ers.) OUTACE).
8/2 - Rx. to 8CI - Estab. sanon equilib, 8/3 - Rx. to 852 - comol. req 'd testing /
Rx. to 952 - compt. req'd testing /
-
l Ex. at 100% #2010/Rx. trip 92113
! due to flux /tabal./ flow on CE. A.
G. C 4 5103% pur. trip should have been:p106* by B&W revised flow s e r cal.
8/4 - Rz. c ri t . at 0700/Rz. to 80%
- 1600 - Avait. sence equilib./
Reduced to ::"-52 (2 hrs) to change oil in CO-P2C (wtr. in oil)/Rz.
to 80 - hold for menon equilib.
8/5-Ra. Pvt. to 100:
8/6-teduced pwr. toc 5752-oil leak at 3 Fw.
pump resulted in sa. fire / returned to 1001 pvr.
8/7-Decr. Rz. to 902-change powdez ual./re-turned to 1002-Es- runback to 552 due to loss of Cr.3 out limit light on sal. ret, to 80: pvr.
8/8-Lx. at 98.*!
8/9-Rz. at 100% pwr.-for T? 800/5 & 20.
8/10-Rz. at 7C2-at 43: pur.-reduced to 252 returned to 40%.
8/11-R2. et 98 -reduced to 60:-incr. to 81.51 F:GURE 1.2-1 (Cont'd) 1414 026
., , .
2.0 INITIAL FUEL LOADING Fuel loading was initiated with the insertion of fuel assembly 1C10 into core location 14-N on April 20, 1974 and was completed with the insertion of fuel assembly 1C40 into core location 15-F on April 25, 1974 Figure 2.0-1 presents a detailed map of the final core configuration, listing each fuel, burnable poison, control rod and orifice red assembly location. Table 2.0-1 provides a detailed sequence of events for inf.tial fuel loading of TMI Unit I.
Neutron countrate was monitored during the core assenbly on four seperate detector channels, with a minimum of two (2) channels operating whenever core geometry was changed. In addition to the permanently installed source range channels, NI 1 and 2, two (2) temporary incore BF3 proportional counters were used. Independent plots of inverse neutron countrate versus the number of fuel assemblies loaded were maintained from the ourputs of these detectors to ensure that the core r mained suberitical at all times.
Initial fuel loading at TMI Unit I was a semi-dry operation with the reactor vessel water level maintained six inches to three feet below the vessel flange.
The semi-dry loading improved visibility of the fuel assemblies during manipu--
lation and provided accessability to the vessel flange area when repositioning the incore detectors. Radiation levels were not overly restrictive due to the lower water levels in the fuel transfer and spent fuel pool canals. The maximum radiation level measured was 25mr/hr (S-Y) at the fuel handling bridge during transfer of the fuel assemblies with the sources.
Several minon problems and delays were encountered during fuel loading. A descriptica%f the problems and their resolution is given below:
(a) Hydrautic pressure on the west transfer carriage upender was lost several times due to a loose coupling between the hydraulic pump and the motor.
The coupling was retightened periodically to permit use of the upender.
(b) On the initial attempt to load fuel assembly IB01, interference occurred with an adjacent assembly. The fuel handling bridge 5:as re-indexed and 1B01 was inserted with no further problem.
(c) During insertion of fuel assembly 1316, the tube down light on the main fuel handling bridge failed to actuate at 2100 lbs and the low-load cut-out was obtained at 600 lbm. Seating of the assembly was verified visually and the 2100 lbm low-load interlock was. bypassed to ungrapple the assembly at the 600 lbm cut-out. The interlock was readjusted and no subsequent problems were encountered.
(d) While inserting fuel assembly 1C24, the 2100 lbm low load cut-out was obtained with the assembly three inches from the down position. The fuel handling bridge was reindexed and the fuel assembly was then inserted smoothly.
(e) Fuel assemblies 1B18 and 1B02 would not seat properly on the initial attempts to insert them. After the assemblies were re-grappled, both were loaded with-out any further problem.
2.0-1
.. . =
, (f) The major delay during the fuel load sequence occurred when power was lost to the main fuel handling bridge. The power loss resulted from shorting two leads in the main power cable which became frayed due to insufficient support as the cable moved with the bridge. The total delay was 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
In spite of the minor delays, initial fuel loading at TMI Unit I went smoothly and was conducted in a safe manner. Figure 2.0-2 depicts number of fuel assemblies loaded versus time.
1414 028 2.0-2
.. . .
FUEL ASSEMBLY LOADING SEOUENCE CRA - Control Rod Assembly ORA - Orifice Rod Asse=bly 3PRA - Burnable Poison Rod Assembly APSRA - Axial Power Shaping Rod Asse=bly DET( )- Auxiliary Neutron Detector STE ASSEMBLY NO. TYPE ID# FEATURE ID # CORE LOCA* ION ACTION 1 Support DET A 1h-H Insert 2 Support DET B 10-P Insert 3 Fuel 1C10 ORA 026 1h-N Insert (Neutron Source) k Fuel 1C22 ORA 005 2-D Insert (Neutron Source) 5 Fuel 1C37 13-0 Insert 6 Fuel 1C28 BPRA 328 13-N Insert 7 Fuel 1C26 BPRA 326 12-0 Insert 8 Fuel 1C53 1h-M Insert 9 Fuel 1Alk CRA C10 12-N Insert 10 Fuel 1A02 CRA C09 13-M Insert 11 hel 1355 BPRA B68 12-M Insert 12 Fuel 1327 BPRA 371 13-L Insert 13 hel 1357 3PRA 355 11-N Insert ik Fuel 1Ah9 APSRA A03 12-L Insert 15 Fuel 2A17 CRA C29 11-M Insert 16 Fuel 2323 BPRA 356 11-L Insert 17 hel 132h 3PRA 321 12-K Insert 18 Fuel 1326 SPRA 35h ;G-M Insert 19 Fuel 1A01 CRA Ch3 10-L Insert 20 Fuel 1A30 CRA Ch2 11-K Insert 21 Fuel 1306 3PRA 30h 10-K Insert 22 Fuel 13 59 3PRA 303 9-1 Insen 23 Fuel 1317 EPRA 3h6 11-H Insert T^3r2 2 -
1414 029
.. . -
STEP ASSEMBLY NO. TYPE ID # FEATURE ID # CORE LOCATION AC"' ION 2h Fuel 1A21 APSPA A0h 10-N Insert 25 Fuel 1A36 CPA Chh 9-M Insert 26 Fuel 13 33 BPRA 322 9-N Insert 27 Fuel 1308 BPRA 323 8-M Insert 28-1 Support DET B 10-F Remove 28-2 Support DET B 7-M Insert 29 Fuel MO4 CRA C11 11-0 Insert 30 Fuel 2323 3F3A 327 10-0 Insert 31 Fuel 1C21 OfA 018 12-P Insert 32 Fuel 1Ch2 11-P Insert 33 Fuel uk6 CRA C56 9-K I:isert 3h Fuel 1A29 CPA C57 8-L Insert 35 Fuel 2A15 CRA C55 10 -H Insert 36 Fuel 23 11 BPRA B07 8-K Insert 37 Fuel 1B01 BPPA B05 9-H Insert 38 Fuel 1335 CRA C61 8-H Insert 39 Fuel 1320 BPRA B07 7-L Insert 40 Fuel 1360 BPPA 318 10-G Insert kl Fuel 1A26 CRA C58 7-K Insert h2 Fuel 1A06 CRA C5h 9-G Insert h3 Fuel 135h 3PRA B08 T-H Insert hk Fuel 1312 3 PPA B09 8-G Insert h5 Fuel 1A10 CRA C60 7-G Insert h6 Fuel 1352 3 PPA 320 6-K Insert h7 Fuel 1319 3 PPA 316 9-F Insert h8 Fuel H56 CRA C59 6-E Insert h9 Fuel 1A27 CEA C53 6-F Insert TABLE 2.C-1 (eent'd)
)k)h
.. , a STEP ASSEMBLY NO. "'YPE ID # FEATURE ID # CORE LOCATIOl ACTION
,
50-1 Support DET A lk-H Re=ove 50-2 Support DET A 9-E Insert 51 Fuel 1361 BPRA 313 6-G Insert 52 Fuel 13h5 BPRA 311 7-F Insert 53 Fuel 2Akh CRA C51 6-F Insert 5h Fuel 13h9 BPPA 337 5-H Insert 55 Fuel 1304 BPRA 3 30 3-E Insert 56 Fuel 1A39 CRA C50 bG Insert 57 Fuel 1A51 CRA C52 7-E Insert 58 Fuel D28 3FRA 363 5-F Ins e.-t 59 Fuel 1B03 BPRA 361 6-E Insert 60 Fuel 1A5h CRA C35 5-E Insert 61 Fuel 1350 3FRA Sh1 h-G Insert 62 Fuel 1B40 BPRA 329 7-D Insert 63 Fuel 1All APSPA A07 h-F Insert 6h Fuel 1A22 APSPA A08 6-D Insert 65 Fuel 2337 3PRA 36h 4-E Insert 66 Fuel 1307 BPPA 362 5-D Insert 67 Fuel 1A35 CRA C22 L-D Insert 68 Fuel 1331 SPh 3h3 3-F Insert 69 Fuel 13h3 3FRA 333 6-C Insert 70 Fuel 1A13 CRA C21 3-E Insert 71 Fuel 1C33 EPRA Shh 3-D Insert 72 Fuel 1C36 2-E Intert 73 Fuel 1Ah2 CRA C23 5-C Insert 7h Fuel 1C31 3PRA Sho LC Insert Fuel
~
75 1Ch6 3-C Inse-:
76 Fuel 1 Chi CRA 035 h-3 Inse -
TA31E 2.0-1 (cent ' di 1414 03g :
=, ,
STEP ASSEMBLY NO. "YPE ID e FEATURE ID r CORE LOCATION ACTION 77 Fuel 1C23 5-3 Insert 78 Fuel M07 CRA Ch7 6-L Insert 79 hel 1A28 CRA Ch8 5-K Insert 80 Fuel 1309 SPRA 359 5-L Inser 81 Fuel 1A19 CRA Ch9 k-H Insert 82 Fuel 3 13 BPPA 336 h-K Insert 83 Fuel 1A09 APSRA A06 h-L Insert 8h Fuel 1A37 CRA C3L 3-G Insert 85 Fuel 2 53 BPRA 352 3-E Insert 86 Fuel 2Ah7 CRA C33 3-K Insert 87 Fuel 1B25 BPRA Sh2 3-L Insert 88 Fuel 1C01 CRA C20 2-F Insert 89 Fuel 2 16 BPRA 314 2-G Insert 90 Fuel 1C5h CBA C19 2-H Insert 91 Fuel Elk 3PRA 315 2-K Insert 92 Fuel 1C50 CRA C18 2-L Insert 93 Fuel 1Ch9 1-F Insert 9h Fuel 1017 1-0 Insert 95 Fuel IC29 1-E Insert 96 Fuel 1C16 1-K Insert 97-1 Support DET B 7-M Remove 97-2 Support DET B 1-L Insert 98 Fuel 1C56 CRA C12 10-P Insert 99 Fuel 1C2h 10-R Insert 100 Fuel 1Ahl CRA C30 9-0 Insert 101 Fuel Sh7 3PRA 3 02 9-? Insert 102 Fuel 1C58 9-R Insen TAB'E 2.0-1 ( cent ' d)
} k} k )3c
.. . .
STEP ASSEF3LY
.
NO. TYPE ID # FEATURE ID # CORE LOCA"ICN AC"TCN 103 Fuel 1A52 CRA Ch5 8-N Insert 104 hel 13k1 BPRA 32h 8-0 Insert 105 Fuel 2003 CRA C13 8-P Insert 106 Fuel ich5 8-R Inser 107 Fuel 1Ah5 CRA Ch6 7-M Insert 108 Fuel 1351 BPRA 332 7-N Insert 109 Fuel 1A38 CRA C31 7-0 Insert 110 Fuel 1318 3PRA 3 06 7-P Insert 111 Fuel 1C06 7-R Insert 112 Fuel 1302 BPRA 357 6-M Insert 113 Fuel 1A12 APSRA A05 6-N Insert 11h Fuel 1336 BPRA B38 6-0 Insert 115 Fuel 1C60 CRA C1h 6-P Insert 116 Fuel 1C20 6-R Inser 117 Fuel 1A55 CRA C32 5-M Insert 118 Puel 1322 BPRA 358 5-U Insert 119 Fuel 1A32 CRA C15 5-0 Inser 120 Fuel 1C25 5-P Insert 121 Fuel 1305 3PRA 360 h-M Insert 122 Fuel 1A18 CRA C16 h-N Insert 123 Fuel 1Ch7 BPRA 334 4-0 != sert 12h Fuel 1A25 CRA C17 3-M Insert 125 Fuel IC3h 3RRA 335 3-N Insert 126 Fuel 1C35 3-0 Insen 127 Fuel 1C52 2-M Insert 128 Fuel 1C51 2-N Insert 129 Fuel 1Ak3 CRA C39 10-F Insert 130 Fuel 1Ah8 CRA CLO 11-G Inse-t TA312 2 . C-1 ( :: t ' d }
)h\k
.. . .
STEF ASSEMBLY NO. n?E ID a FFATURE ID # CCRE LOCATION AC"' ION
-
131 Fuel 1310 BPRA 365 11-F Insert 132 Fuel 1A23 CRA Ch1 12-H Insert 133 Fuel 1321 BPRA ELT 12-G Insert 13h hel 1A50 APSRA A02 12-F Insert 135 Fuel 1A31 CRA C28 13-K Inse-t 136 Fuel 13hk BPRA Bh9 13-H Ins ert 137 Fuel 1A05 CRA C27 13-G Insert 138 Fuel 1320 3PRA 350 13-F Insert 139 Fuel '1C57 CRA CO8 14-L Insert 140 hel 133h EPRA B20 1L-K Insert 1h1 Fuel 1C13 CRA C07 1h-H Insert 142 Fuel 1315 BPRA 319 IL-G Insert ik3 Fuel 1C05 CRA C06 IL-F Insert ikk Fuel 1C15 15-L Insert ih5 Fuel 1C59 15-K Insert ik6 Fuel 1C39 15-H Insert ikT Fuel 1C08 15-G Insert Ih8-1 Support DET A 9-E Remove ik8-2 Support DET A 15-7 Insert 1h9 Fuel 1C 0k CRA C2h 6-3 Insert 150 Fuel 1018 6-A Ins ert 151 Fuel 1A20 CRA C36 7-C Inse-t 152 Fuel 1356 BPPA 312 7-3 Insert 153 Fuel 1C07 7-A Insert 15h Fuel 1A2h CRA C37 8-D Insert 155 Fuel 3h2 BPRA 331 8-C Insert 156 Fuel 1CO2 CRA Col 8-3 Insert 157 Fuel 1C30 8-A Insert TABLE 2.0-1 (cent'd) 1414 134
.. . .
STEP ASSEMBLY NO. TYPE ID # FEATURE ID # CORE LOCA* ION AC"' ION
-
158 Fuel 1A53 CRA C38 9-E Insert 159 Fuel 1B32 BPRA B39 9-D Insert 160 Fuel 1A40 CRA C25 9-C Insert 161 Fuel 1Bk6 BPPA B17 9-B Insert 162 Fuel 1C1h 9-A Insert 163 Fuel 13h8 BPPA 353 10-E Insert 16h Fuel 1A33 APSRA A01 10-D Insert 165 Fuel 1329 BPRA Bh5 10-C Insert 166 Fuel 1C12 CRA CO2 10-B Insert 167 Fuel 1Ch8 10-A Insert 168 Fuel 1A16 CPA C26 11-E Insert 169 Fuel 1358 BPRA B66 u-D Insert 170 Fuel 1A08 CRA C03 11-C Insert 171 Fuel 1C19 11-B Insert 172 Fuel 1B39 BPR*. B67 12-E Ins ert 173 Fuel 1A3h CRA CO4 12-D InseM 17h Fuel 1C27 BPRA Bh8 12-C Insert 175 Fuel 1A03 CRA C05 13-E Insert 176 Fuel 1C32 BPRA 351 13-D Insert 177 Fuel 1Chh 13-C Insert 178 Fuel 1C55 1h-E Insert 179 Fuel 1C38 lb-D Insert 180-1 Fue' 'CIO ORA 026 1k-N Re:Ove 180-2 Fuel 1C10 ORA 026 h-P Insert 181 Fuel 1C11 1h-N Insert
. 180-1 Fuel 1C22 ORA 00 5 2-D Re= eve 182-2 Fue' 'C22 ORA 005 12-B Insert TABLE 2.0-1 (cent'd) )k\
- . . . -
STEP ASSDIBLY NO. TYPE ID a FEATURE ID e CORE LOCATION ACTION r
183 Fuel ich3 2-D Insert 18h Support DET B 1-L Remove 185 Fuel 1C09 1-L Insert 186 Support DET A 15-F Re=ove 187 Fuel 1C40 15-F Inserc
~1414 036 TA3LE 2.C-1 s cont 'd)
.. e -
FINAL FUEL LOADING DISTRIBUTION FOR CORE 1, CYCLE 1
.
X A 1C18 1C07 1C30 1C14 1C48
_
FUEL TRANSFER \
1C41 1C23 1C04 1B56 1C02 1B46 1C12 1C19 1C22 CANAL 3
038 C24 B12 C01 B17 CO2 fg C 1C46 1C31 1A42 1B43 1A20 1B42 1A40 1B29 1A08 1C27 1C44 B40 C23 B33 C36 B31 C25 B45 C03 B48 D 1C43 1C33 1A35 1307 1A22 1B40 1A24 1B32 1A33 1B58 1A34 1C32 1C38 B44 C22 B62 A08 B29 C37 B39 A01 B66 C04 B51 1C36 1A13 1B37 1A54 1B03 1A51 1B04 1A53 1B48 1A16 1B39 1A03 1C55 E
C21 B64 C35 B61 C52 B30 C38 B53 C26 B67 C05 y .
1C49 1C01 1B31 1A11 IB28 1A44 1345 1A27 1B19 1A43 1B10 1A50 1B30 1C05 1C40 C20' B43 A07 B63 C51 Bil C53 B16 C39 B65 A02 B50 C06 1C17 1B16 1A37 1B50 1A39 1361 1A10 1B12 1A06 1B60 1A48 1B21 1A05 1315 1C08
-
B14 C34 B41 C30 B13 C60 B09 C54 B18 C40 B47 C27 B19 g p, IC29 1C54 1B53 1A19 1B49 IA56 1B54 1335 1B01 1A15 1B17 1A23 1B44 1C13 1C39 C19 B52 C49 B37 C59 B08 C61 B05 CS5 B46 C41 B49 C07 y
K - 1C16 1B14 1A47 1B13 1A28 1B52 1A26 1B11 1A46 1B06 1A30 1B24 1A31 1B34 1C59 BIS C33 B 36 C48 E10 C58 B01 C56 B04 C42 B21 C28 B20 g 1C09 1C50 1B25 1A09 1B09 1A07 1B20 1A29 1359 1A01 1B23 1A49 1B27 1C57 1C15 C18 B42 A06 B59 C47 B07 C57 B03 C43 356 A03 B71 C08 M 1CS2 1A25 1B05 1A55 1B02 IA45 1B08 1A36 1B26 1A17 1B55 1A02 1C53 C17 B60 C32 B57 C46 B23 C44 B54 C29 B68 C09 1C51 1C34 1A18 B22 1A12 1B51 1A52 1B38 1A21 1B57 1A14 1C28 1Cll N
B35 C16 358 A05 B32 C45 B22 A04 BSS C10 B28 0 1C35 1C47l 1A32 IB36 1A38 1B41 1A41 1B33 1A04 1C26 1C37 B34 l C15 B38 C31 B24 C30 327 C11 B26 1C10 1C25 1C60 1B18 1C03 1B47 1C56 1C42 1C21 g Cl4 B06 C13 B02 C2 2 018 R .
1C20 1C06 1C45 1C58 1C24 I i i l i 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 2
1A01 throsgh 1A56 - 2.06 wt. % fuel assemblies 1B01 through 1361 - 2.72 wt. % fuel assemblies 1C01 through IC60 - 3.05 wt. I fuel assemblies C01 through C61 - Control Rod Assemblies A01 through A08 - Axial Power Shaping Rod Assemblies 036, 005, 026, 018 - 0.ifice Rod Assemblies 301-324, 326-B68,371 - Burnable Poison Rod Assemblies
(*) - Assembly with Neu :ron Source H GURE 2.0-1
. . . ,
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.. . .
. 3.0 POST FUEL LOAD PRECRITICAL TEST PROGRAM A Post Fuel Load Precritical Test Program was conducted following initial fuel loading. This section of the report presents the scope and results of that testing.
The Control Rod Drive Drop Time Test was conducted at reactor coolant system conditions of 1500F, 450 psi and 5320F, 2155 psi with and without reactor coolant flow. The purpose of the test is to measure the total trip insertion time from trip '.nitiation to three-fourths insertion for each control rod assembly.
Reactor Coolant Pump Flow and Flow Coastdown measurenents were conducted at system conditions of 5320 F, 2155 psi to detemine core flow characteristics. Pressurizer testing was also conducted at hot conditions to determine the pressurizer spray valve and bypass flow settings. Reactor coolant system leakage measurements were performed to verify that RCS leak rate was within acceptable limits. In all cases, applicable test crite-ia and Technical Specification requirements were met.
1414 03o 3.0-1
.. .
- 3.1 REACTOR COOLANT PUMP FLOW TEST 3.1.1 PURPOSE The Reactor Coolant Pu=p Flow Test was performed with the core installed to determine the functional capabilities of the Reactor Coolant System and Reactor Coolant Pumps and to determine the reactor coolant flow characterisites for various pump operating co=binations.
3.1.2 TEST METHOD Reactor coolant loop flows were determined by means of loop flowrueter AP cells in called in the reactor coolant system. The output of the AP cells were con-verted to temperature compensated flow indication according to Equation 3.1-1.
- -
1/2 Flow = C f APh (Equation 3.1-1)
V, Where: C f =
Flow coefficient = 397100 dP =
Indicated flowmeter differential pressure V = Specific volume at reference conditions of 680 F,14.7 pst V, = Specific volume at system conditions For each reactor evolant pump combination, steady state temperature, pressure and flow was maintained and data was recorded by the plant computer, brush recorders and reactimeter. Measured flow rates were averag'i over a specified time and the results were compared with acceptance criteria.
3.1.3 TEST RESULTS Reactor coolant flow was measured at 5320F, 2155 psi for twelve (12) reactor coolant pu=p operating combinations. Table 3.1-1 lists the measured flow rates along with the minimum and maximr.m allowable flows. As can be seen from table 3.1-1, all measured flow rates were within the acceptance criteria.
3.
1.4 CONCLUSION
S 0
Reactor coolant flow measurements were conducted at 532 F, 2155 psi with the core installed and all measured flow rates were within the range of acceptable values.
1414 040 3.1-1
. . . .
t REACTOR COOLANT FLOW RATES AT 532 F, 2155 PSI WITH REACTOR CORE INSTALLED Pump Minimum Acceptable Maximum Acceptable Measured Case Combination Rate F1 p Rate F1 F1 Rate (Pumps Running) (Il0 lbm/hr) (X10plbm/hr) (X10plbm/hr, ,
1 A *
- 39.8 2 B *
- 40.48 3 C *
- 40.85 4 D *
- 40.72 5 A,B,C,D 138.5 154.5 146.0 6 A,B,D 103.2 154.5 110.13 7 A, D 67.8 154.5 73.75 8 B,C,D 103.2 154.5 109.7 9 A, C 67.8 154.5 74.25 10 A, B 62.4 154.5 80.35 11 C, D 62.4 154.5 81.17 12 B, D *
- 73.98
- - Indicates that no acceptance criteria was applied to the pump combination.
- 1414 041 TABLE 3.1-1
.. . .
3.2 REACTOR C00 TANT PUMP FLorJ COASTIORN TEST 3.2.1 PURPOSE The Reactor Coolant Pu=p Flow Coastdown Test was performed to determine reactor coolant flow characteristics for spec.ific reactor coolant pump trip combinations.
Test.ng was conducted at system conditions of 532 0 F, 2155 psi with the core f.nstalled.
3.2.2 TEST hETHOD Eight (8) different reactor coolant pump combinations were selected for the measurement of flow coastdown characteristics. The eight combinations and a description of each is su:marised in Table 3.2-1. For each pump combination, steady state conditions were established and data was recorded by the plant computer and test recorders. All or a portion of the coolant pumps were then tripped and data was recorded through the resulting flow transient. Reactor coolant flow indication was obtained from the loop flowmeter instrumentation as described in section 3.1. The hie.rarchy of single reactor coolant pump flows was de.termined during the Reactor Coolant Pump Flow Test, which was performed in conjunction with the coastdown test.
3.2.3 TEST RESULTS The results of this test at 5320F, 2155 psi with tne core installed are su=marized in Figures 3.2-1 through 3.2.-4. Measured reactor coolant flow versus time is plotted along with the acceptance. criteria limits for each reactor coolant pump combination. The flow. values plottei were obtained by dividing the indicated flow at a specific time, t, after the trip by the measured minimum initial core flow with the pump combination running. All mini =um core flow criteria were met. In addition to the minimum flow criteria, a further requirement was i= posed upon cases 4, 5 and.6, that.the reactor coolant flow decrease by a certain parcentage within a specified time after the pumps were tripped. This criteria was also met.in each Case.
3.
2.4 CONCLUSION
S The reactor coolant flow coastdown characteristics measured at system conditions of 532 F 2155 psi with the core installed met all applicable acceptance criteria.
)k\k 3.2-1
-
.
REACTOP. COOLANT PUMP FLOW COASTDOWN COMdINATIONS
.
.
Case Pump Initially Running Pumps Tripped 1 A,B,C,D A,B C,D 2 A, B, D - three lowest flow pumps A, B, D 3 A, D - lowest flow pumps in each loop A, D a
b 4 A,B,C,D C - highest flow pump N
5 A,B,C,D C, D - pumps in higher flow loop A
6 A,B,C,D B, C - higher flow pump each loop 7 B,C,D B - pump in loop with idle pump 8 B,C,D C - higher flow pump in loop with two pumps operating
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3.3 CONTROL ROD DRIVE DROP TIME TEST 3.3.1 PURPOS2 Technical Specifications 4.7 places limits on the control rod trip insertion times for reactor coolant system full flow and no flow conditions. The Control Rod Drive Drop Time Test measures the data to fulfill the Technical Specification limit and to establish data for future periodic testing.
3.3.2 TEST METHOD The Control Rod Drive Drop Time Test was performed using strip chart recorders to time the rod drops. Each control rod group was pulled to 100% withdrawn and then dropped into the core using the manual trip pushbu: ton. A zero time sigr.a1 was furnished to the test recorders for each control rod assembly from a contact on the manual trip switch. A second signal to indicate three-fourths insertion was furnished to the recorders by a reed switch located on t'ae position indicator tube of each control rod drive. The test was conducted at nominal reactor cool-ant system conditions of 150 0F, 450 psi and 5320F, 2155 psi under flow and no flow conditions. Control rod groups 1 through 7 were each withdrawn to 100% and tripped at each of the four (4) test conditions. After drop time measurements on all the groups were completed, the rods with the fastest and slowest trip insertion times were tripped ten additional times to demonstrate repeatability of the measurement. Measurements were performed on the group 8 control ro.ls to verify' that they do not drop into the core when power to the control rod drive trip breaker undervoltage coils is interrupted.
3.3.3 TEST RESULTS The measured results for the first test condition of 15007, 450 psi with no reactor coolant flow show that rod H-10 was the fastest at 1.104 sec. and rod M-7 was slowest at 1.16 sec. Ten additional drops on rods B-10 and M-7 produced drop times within 16 ms and 20 ms, respectively. The group 8 rods were withdrawn to 25%
and no rod movement was observed when the control rod drives were tripped, as required.
The measured results for the second test condition of 1500F, 450 psi with one reactor coolant pump operating show that rod F-10 was fastest at 1.128 seconds and rod M-9 was slowest at 1.176 seconds. Ten additional drops on rods F-10 and M-9 produced drop times within 16 ms and 24 ms, respectively.
The measured results for the third test condition of 532 F, 2155 psi with no reactor coolant pumps operating show that rod H-10 was fastest at 1.072 seconds and rod 0-5 was slowest at 1.135 seconds. Ten additional drops on rods H-10 and 0-5 produced drop times within 25 ms and 20 ms, respectively.
The measured results for the fourth test condition of 532 F, 0 2155 psi with 100%
reactor coolant flow conditions show that rod H-10 was the fastest at 1.225 seconds and rod M-5 was the slowest at 1.363 seconds. Ten additional drops on rods H-10 and M-5 produced drop times within 25 ms and 19 ms, respectively.
.
1414 048 3.3-1
.. . .
3.
3.4 CONCLUSION
S Control Rod Drop Time measurements conducted at 150 F and 532 F show that the control rod assembly trip insertion time from 100% withdrawn to three-fourths insertion will not exceed 1.40 seconds under reactor coolant no flow conditions and 1.66 seconds under reactor coolant flow conditions. The requirements of Technical Specification 4.7.1 were met in all cases.
)h\k
.
5 3.3-2
.. . .
3.4 PRESSURIZER TEST
.
3.4.1 PURPOSE Pressurizer Operational tes'.ing was conducted prior to initial criticality to set the pressurizer spray and-bypass flows at the prescribed setpoints.
3.4 2 TEST METHOD 1he technique used to set the presaurizer spray and bypass flows was based upon balancing the heat input to aLd the heat losses from the pressurizer. Initial steady state pressure and temperature conditions were established in the pressurizer without spray or bypass flow. The power input from the pressurizer heaters necessary to maintain steady state conditions was recorded. The additional heat input required to balance spray and bypass flow was then cal-culated using Equation 3.4-1.
F = K (aQ) (Equation 3.4-1) fp - hf RCS Where: F - is the spray or bypass flow K - is a constant = 9.03 AQ - is the difference between the heater input with flow and the h3ater input without flow hfp - is the enthalpy of saturated water at the pressurizer temperature hf RCS - 1s the enthalpy of saturated water at the RCS temperature Baat input to the pressurizer from the heaters was then increased by the amount calculated. The bypass and spray valve flows were increased to balance the additional heat input and maintain the pressurizer temperature and pressure at their initial values.
3.4.3 TEST RESULTS The measured results from setting the pressurizer spray and spray valve bypass flow are listed in Table 3.4-1. ::he bypass and spray flows were set at 0.99 gpm and 190.5 gpm, respectively. The measured pressurizer heat loss was in excess of 100KW at system conditions of 5320F and 2155 psi.
3.
4.4 CONCLUSION
S The pressurizer spray flow was set within the acceptance criteria limit of 190.0 +19/-6 gpm. The pressurizer spray bypass flew was set within the acceptance criteria limit af 1.0 +0.5/-0.25 gpm.
0
)&)h 3.4-1
.
.
'
. IEASURED RESULTS FDR DETERMINATION OF
- PRESSURIZER SPRAY AND BYPASS FLOW AT 532"F A. Pressurizer Spray Bypass Flow Test RCS RCS Pressurizer ileater Spray Bypass Conditions Pressure Temperature Temperatrre Power Flow initial 2157psig 532.0 F 644.1"F 106.9KW 0.0gpm a final 2159psig 531.0"F 644.2 F 124.5KW
$ 0.99gpm f;
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B. Pressurizer Spray Flow (with bypass flow)
Test RCS RCS Pressurizer lleater Spray Conditions Pressure Tempera ture Temperature Power Flow initial 1406psig 530.7 F 586.8 F 132.69KW 0.0gpm final 1404paig 533.6 F 587.1 F 1609.35KW 190.5gpm
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.. . .
.
3.5 REACTOR COOLANT SYSTEM LEAKAGE 3.5.1 PURPOSE The purposes of the Reactor Coolant System (RCS) hot leakage test were as follows:
- 2) Determine the accuracy of the method used to determine RCS leakage by imposing a "known" leak rate.
- 3) Examine systems containing reactor coolant to identify leakage.
- 4) Establish a value for" normal exaporative losses" as used by Technical Specifi-cation 3.1.6.2.
- 5) Verify the Surveillance Procedure for RCS leakage determination.
. 3.5.2 TEST METHOD The RCS hot leakage and surveillance procedure verification test was performed during the hot functional and post fuel load pre-critical test programs and its results served as a basis for conducting the Surveillance Procedura for RCS leak-age determination during the power escalation test program.
With the primary plant at 532 F and 2155 psig, pressurizer level, makeup tank level, reactor coolant drain tank (RCDT) level and RCS temperature were monitored as a function of time. Changes in RCS inventory were computed over a four hour period. These computations resulted in a measured 'eak rate of 0.821 gym during the four hours, when corrected for an RCP #3 seal purge flow addition of 0.07 gpm.
A known leak rate of .68 gpm was then established to determine the sensitivity of the above computations in yielding accurate values for leakage. Again, changes in RCS inventory were computed over a four hour period. The computations resulted in a leak rate of 0.659 gpm during the four hours, when corrected for RCP #3 seal purge flow addition and the known leak rate of .68 gpm.
Independent of the leakage monitoring operations above, a survey of all primary system boundary piping, valves, fittings, instrument connections, and flanges was
,
made in an attempt to measure and identify every drop of leakage that was noti evaporating to the containment, auxiliary building, or RCDT atmosphere. This survey resulted in a measured leakage of 0.231 gpm. The difference between the average value of computed leakage for the two four-hour runs minus the survey measured (identified) leakage is the established value of " normal evaporative losses" used in the Surveillance Procedure for leakage determination during normal plant operation.
RCS leakage was monitored every day during the power escalation program when the reactor was critical, as required by Technical Specifications. The Surveillance Procedure for leakage determination was used for this purpose.
3.5-1 kk
,. . .
3.5.3 TEST RESULTS The average value of computed leakage for the two four-hour runs was 0.740 gpm.
No value of leakage computed for any single hour out of the eight differed from the 0.740 gpm figure by more than +l_ gpm, thereby supporting the contention that
,
RCS level and temperature instrumentation is sensitive enough to detect a 1 apm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The total measured identified leakage was 0.231 apr. This results in a value of 0.51 gpm for tha " normal evaporative losses".
3.
5.4 CONCLUSION
S Reactor Coolant System hot leakage measurements were conducted prior to initial criticality. The measured results verify that the reactor coolant leakage does not exceed the Technicu Specification recuirements and that the normal control instrumentation is sensitive enough to perform leak rate measurements.
1AiA 053 3.5-2
.. . .
4.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POER Three Mile Island Unit One, Core 1 consists of 177 fuel assemblies,each con-taining 208 fuel rods,16 control rod guide tubes and one incore instrument guide tube. The arrangement of these assemblies is shown in Figure 4.0-1.
The inner 117 assemblies, which are arranged in a checkerboard pattern, are of two different enrichments - 2,06 and 2.72 we. % uranium - 235. An outer ring of 60 assemblies enriched to 3.05 vt. I uranium - 235 completes the core.
Lumped burnable poison is distributed throughout the core. A detailed loading map of Core 1, with each fuel assembly, control rod, orifice rod and lumped burnable poison assembly is given in Figure 2.0-1 of section 2.0.
The reactivity of the core is controlled by 61 full-length Ag-In-Cd control rods and soluble boron in the Reactor Coolant System (RCS). Eight (8) partial length control rods are provided for add $ tonal control of axial power distri-butions . The locations of the 69 control rods are also shown in Figure 4.0-1.
The important design data and calculated performance characteristics of Core 1 are tabulated in Table 4.0-1.
Core performance measurements were conducted during the Zero Power fest Program which began on June 5,1974 and ended on June 10, 1974. This section presents the results and an evile2 tion of the zero power tests, which included initial criticality, nuclear ins:rumentation overlap, verification of reactivity calcula .
tions, all rods out critical boron determination, temperature coefficient measure-ments, shutdown margin determination and soluble poison and control rod reactivity worth measurements. A comparison of meaaured and predicted results is given based on on-site analysis. In all cases, the applicable test and Technical Specification acceptance criteria were met.
)h\h 0 i 4.0-1
,. . .
Table 4.0-1. Core 1 Desien Data and Performance Characteristics
.
Reactor Design heat output, W t* 2535 Vessel coolant inlet temp, F 554 Vessel coolant outlet temp, F 603.8 Core coolant outlet temp, F 606.2 Core coolant operating pressure, psig 2185 Core coolant T,y ,F 579.3 Core and Fuel Assemblies Total number fuel assy in core 177 Nu=ber fuel rods per fuel assy 208 Number control rod guide tubes per assy 16 Number incore instr positions per fuel assy 1 Fuel rod outside diameter, in. 0.430 Cladding thickness (min) in. 0.026 Fuel rod pitch, in. 0.566 Fuel assembly pitch spacing, in. 8.587 Cladding material Zircaloy-4 (cold worked)
Fuel Material UO Form Di hed-end, cylindrical pettets Pellet diameter, in.(*}) 0.364 Active length, in. 141.2 Density (Unit 1, Core 1) , theor (*) 92.5 Heat Transfer and Fluid Flow at Rated Power (*
Total heat transfer surface 1n core, ft2 48,766 Average heat flux, Btu /h-ft 3 174,870 Maximum heat flux (at min DNBR), Btu /h-ft 7 469,873 Average power density in core, kW/i 82.31 Average ther=al output, kW/ft of fuel rod 5.69 Maximum thermal output, kW/ft of fuel rod 18.2 Maximum cicdding surface temp, F 650 Average fuel temp of hottest pin, F 3,237 Maximum fuel central temp at hot spot, F 4,953 Total reactor coolant flow , 10 1b 6 /h 131.32 2
Core flow area (eff for heat transfer), ft 49.19 Core coolant average velocity, fps 15.73 Coolant outlet temp at hot channel, F 647.1
--__
- Note: The core will be operated with a 100: FF value of 2535 Wt (Technical Specification limit) even though the rated power of this core is 2568 st.
TABLE 4.0-1 3434 055
.. . .
Power Distribution Maximum / average power ratio, radial x local 1.78 (Fah nuclear)
Maximum / average power ratio, axial 1.70 (Fg nuclear)
Overall power ratio (Fq nuclear) 3.03 Power generated in fuel and cladding, 97.3 Hot Channel Factors Power peaking factor (FQ) 1.011 Flow area reduction factor (FA)
Interior bundle cells 0.98 Perdpheral bundle tells 0.97 Local .w.t flux factor (F " 1.014 Hot spot maximum / average a he)t flux ratio 3.12 (Fq nuc and mech)
_DNB Data Design overpower, % rated power 112 DNB ratio at design overpower (W-3) 1.55 DNB ratio at design power (W-3) 2.0 Limiting DNB ratio at design overpower (W-3) 1.3 Fuel Assembly Volume Fractions Fuel 0.303 Moderator 0.580 Zircaloy 0.102 Stainless steel 0.003 Void 0.012 1.000 Total UO2 (BOL, First Core)
Metric tons 93.1
'
Core Dimensions, in.
Equivalent diameter 128.9 Active height (with/without densification) 141.1/144.0 Unit Cell H2O to U Atomic Rata (Fuel Assembly)
Cold 2.88 Hot 2.06 Full-Power Lifetine, days First cycle 460
)k)h TABLE 4.0-1 (Cont'd)
. . . .
Fuel Irraciation, mwd /mtU First cycle average 14,400 ,
Fuel Loading, wt : 235 g Core average first cycle 2.62 M trol Data Control rod material Ag-In-Cd Nu=ber full-length rods 61 Number APSES 8 Control rod cladding material SS-304
-___
(a) Following densification
\h\4 051 TABLE 4.0-1 (Cont'd)
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- 414 058 FIGURE 4.0-1
.. .*
4.1 INITIAL CRITICALITY Initial criticality was achieved on June 5,1974 at reactor conditions of 532 F and 2155 psig. Control rod groups 1 through 4 were previously withdrawn during the heatup to 5320F. The initial reactor coolant system (RCS) boron concentra-tion was 2086 ppm. The approach to critical began by withdrawing control rod groups 5, 6 arid 8 to 100% and positioning group 7 at 75% withdrawn. Criticality was subsequently achieved by deborating the reactor coolant system to a boron concentration of 1545 ppm. The procedure used in the approach to critical is outlined below in three basic steps.
Step 1 Control Rod Withdrawal Group 8 100% withdrawn Group 5 100% withdrawn Group 6 100% withdrawn Group 7 75% withdrawn Step 2 Deborate using a feed and bleed flow rate of 50 gpm until criticality is almost achieved, as indicated by any inverse count rate plot reading approximately 0.05.
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 reacto coolant system boron concentration reaches equilibrium.
Throughout the approach to criticality, plots of inverse multiplication were maintained by two independent persons. Two plots of inverse count rate (ICR) versus control rod position were maintained during control rod withdrawal. Two plots of ICR versus RCS boron concentration and two plots of ICR versus gallons of demineralized water added were maintained during the dilution sequence. At the end of each reactivity addition (boron dilution or control rod withdrraal),
count rates were obtained from each startup range neutron detector channel. The ratio of the initial average count rate to the count rate at the end of each reactivity addition is the value plotted.
During control rod withdrawal (Step 1) ICR plots versus control rod group position were maintained from the outputs of source range channels NI 1 and 2. The with-drawal interval for each control rod group was limited to no more than half t!.a 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. Firs t, deboration from 2086 ppm was commenced using a feed and bleed flow rate of 50 gpm (Step 2) . RC boren samples were taken every 30 minutes and samples from the makeup tank and the pressurizer were taken hourly. Two ICR plots were
=aintained vs. every 1000 gallons of demineralized water added, and two plots were maintained versus RC letdown concentration every 30 minutes. Deboration at a let-down rate of 50 gpm was continued until one of the ICR plots indicated 0.05. At this time, demineralized water additions were stopped and the letdown flow rate was increased to 140 gpm to expedite mixing in the RCS (Step 3) . RCS boron con-centration at this time was approximately 1650 ppm. After initial criticality was achieved, control rod group 7 was inserted to control neutron flux during the subsequent mixing.
n 4.1-1 g} k t)CG J/
.. . .
"Just critical" conditions were stabilized and maintained at an equilibrium boron concentration of 1545 ppm with group 7 at 26.5% withdrawn. The measured critical boron concentration with group 7 at 75% withdrawn was 1609 ppm which compares well with the predicted value of 1625 ppm. The inverse count rate plots maintained during the approach are presented in Figures 4.1-1 through 4.1-5. As can be seen from the plots, the response of the source range channels during reactivity additions was very good. Figure 4.1-1 is the plot of ICR versus control rod group withdrawal for data taken from NI channels 1 and 2.
Figures 4.1-2 and 4.1-3 are the ICR plots versus RCS boron concentration and Figures 4.1-4 and 4.1-5 arr. the ICR plots versus gallons of demineralized water added to the RCS, for source range channels NI-1 and NI-2, respectively.
In su= mary, initial criticality was achieved in an orderly manner. There was good agreement between the measured and the predicted critical boron concentration.
with the difference between the two directly attributable to the change in group 7 position from 75 to 26.5% withdrawn.
1414 060 4.1-2
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1414 061 FIGURE 4.1-1
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INVERSE COUNT RATEVERSUS BORON CONCENTRATION NI - 1 1.0 :
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.. . .
4.2 NUCLEAR INSTRUMENTATION OVERLAP 4.2.1 PURPOSE Technical Specification 3.5.1 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. Ihis means that before the source range count rate equals 105 cps the intermediate range NI must be on scale. In addition, the following number of NI channels must be in operation fer the test program to continue beyond initial criticality.
Channels Available Minimum Ooerating Source Range NI 2 2 Intermediate Range, NI 2* 2
- One channel was input to the react 1=eter but was operable.
4.2.2 TEST METHOD To satisfy the above overlap requirements after initial criticality was achieved, core power wr.s increased until the intermediate range channels came on scale.
Detector signal response was then recorded for both the source range and inter-mediate range channels. Th range channels approached 10ycps. was repeated for two.more decades until the source 4.2.3 TEST RESULTS s
The results of the initial NI overlap data at 532 F and 2155 psig are plotted on
.
I Figure 4.2-1. A minimum of two (2) decades overlap is observed between 'the source and intermediate ranges. The data is normalized to an estimated core power level of 10W(t) at an intermediate range signal of 1.0x10-7 ampere. This estimate was made by observing a 0.1907 temperature increase across the core at this signal level at 5320F. This core differential temperature has been shown to be equivalent to a 10W(t) heat addition to the reactor coolant system. Also plotted on Figure 4.2.-1 is the intermediate range detector response at power. The step change in 0
intermediate range signal between 5320F and 579 7 is due to a higher detector signal output for a given core power level caused by a reduction in reactor coolant hydrogen density and the reduction in the a=ount of boron per unit volume of reactor coolant.
4.
2.4 CONCLUSION
S Examination of Figure 4.2-1 shows that linearity, overlap and absolute output of the intermediate and source range detectors are within cpecifications and performing satisfactorily. There is at least two decades overlap between the source and in-termediate ranges.
- k
.. . .
NUCLEAR INSTRLHENTATION OVERLAP Detector Response, amp
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f FIGURE 4.2-1
)h\4 '
.. . .
4.3 REACTIVITY CALCULATIONS 4.3.1 PURPOSE Reactivity calculations during the THI Unit I test program were performed using the Reactimeter. After initial criticality and prior to the first physics measurement, an on-line functional check of the reactimeter was performed to verify its readiness for use in the test program.
.
4.3.2 TEST METHOD Reactimeter is the nene given to the Babcock and Wilcox reactivity-meter which solves the one-dimensional, inverse kinetics equation with six delayed neutron groups for core net reactivity based upon periodic samples of neutron flux.
In addition to reactivity and neutron flux, the Reactimeter can also record 23 other analog and digital signals frcm the plant. The computational and data recording capability of the Reactimeter were used cxtensively throughout the test program.
After initial criticality and nuclear instrumentation overlap were established, intermediate range channel NI-4 was input to the reactimeter and the reactivity calculations were started. Af ter 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 inserted was determined from period-reactivity curves. The measurement was repeated for several values of reactivity inserted by rod group 7, from 19 02% A k/k to fp.075%Ak/k. The reactiv-l ities determined from doubling time measurements were compared with the reactivity calculated by the reactimeter.
- 4.3.3 TEST RESULTS The results of the reactimeter verification measurements are summarized in Table 4.3-1. In each case, the reactivity calculated by the reactimeter was well within the acceptance criteria limit of 1;2% of the reactivity determined from doubling times.
4.
3.4 CONCLUSION
S An on-line functional check of the reactimeter was performed after initia 1 critical-ity. 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.
1414 068 4.3-1
. . . .
9
(
COMPARISON OF REACTIMETER AND DOUBLING TIME (DT)
REACTIVITT T.ASUREMENTS Measured ( Calculated Percent Case DT Reactivity Reactivity Difference No. (Sec) (%ll/k) (%Ak/k) (%)
1 305 +0.0177 +0.0179 -1.11 2 -271 -0.0233 -0.0230 +1.29
%.0440
~
3 110 +0.0442 -0.45
'
4 -150 -0.0470 -0.0478 -1.70
,
t 5 63 +0.0692 %.0695 -0.43 6 -127 -0.0580 -0.0582 -0.34
____
(1) Measured doubling times were determined from analyzing reactimeter traces of the neutron flux.
.
%
TABLE 4.3-1
.. . .
4.4 ALL RODS OUT CRITICAL BORON CONCENTRATION 4.4.1 PURPOSE The all rods out critical boron concentr. scion measurement is performed to obtain an accurate value for the excess reactivity loaded in the TMI Unit I core and to provide a basis for the verification of calculated reactisity worths. This measurement was performed at system conditions of 5320F and 2155 psig.
4.4.2 TEST NETHOD The reactor coolant system was borated such that control rod groups 1-6 and 8 were positioned at 100% withdrawn and group 7 was maintaining criticality at approxt=ately 80% withdrawn. Once staady state conditions were established, control rod group 7 was withdrawn to 100% and the resultant reactivity change was measured. The measured boron concentrt. tion 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 resetivity worth, in terms of ppm boron, of the inserted control rods.
4.4.3 TEST RESULTS The results of the measurement at 532 F are tabulated below.
ALL RODS OUT CRITICAL BORON CONCENTRATION ppm boron
'
Moderator Calculated Measured Temperature Result Result
'
532 F 1634 1617 The measured boron concentration with group 7 positioned at 75% was 1609 ppm. An additional 8 ppm was added to this value that is derived from 0.084% 4 k/k due to group 7 withdrawal to 100%, using a differential boron worth of 1.058% ak/k per 100 ppm boren.
4.
4.4 CONCLUSION
S The above results show that the measured boron concentrations are in excellent agrecment with predictions and are well within the acceptance criterion of +100 ppm.
,
a4i4 070 4.4-1
.. ..
4.5 TEMPERATURE COEFFICIENT MEASUREMENTS 4.5.1 PURPOSE The moderator temperature coefficient 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 at full power conditions. The moderator temperature coefficient cannot be measured directly, but it can be derived from the core temperature coefficient and a known fuel temperature (isothermal Doppler) coefficient at the zero power condition. For this reason, the temperature coefficient of reactivity was measured for several different boren concentr,tions at the zero power conditions of 5320F and 2155 osig to provide comoarison of the moderator te=perature coefficient with the design calculations prior to operation in the power range.
4.5.2 TEST METHOD The technique used to measure the isothermal temperature coefficient at zero power was to first establish steady state conditions by maintaining reactor flux, reactor coolan pressure, turbine header pressure and core average temperature constant, with the reactor critical at approximately 3x10-9 amps in the inter-mediate range. (The measurement began with the reactor critical at a slightly higher flux level if a negative feedback effect was expected from a temperature increase or at a icver flux level if a positive feedback effect was expected from a temperature increase.) Equilibrium boron concentration was established in the reactor coolant system, make-up tank and pressurizer to eliminate reactivity effects
, due to boren changes during the subsequent temperature swings. The reactimeter and the brush recorders were connected to monitor selected core parameters with the reactivity value calculated by the reactimeter and the core average te=perature displayed on an L&N two channel recorder.
Once steady state conditions were established, a positive heatup rate was started by closing the0 turbine bypass valves. After the core average temperature increased by about 10 F, core temperature and. flux were stablized and the process was re-versed by decreasing the core average temperature to the initial value by opening the turbine bypass valves. This procedure was completed two times at each boron concentration that the coefficient measurement was conducted to establish repeatability in the measured value. Calculation of the temperature coefficient-from the measured data was then perfermed by dividing the change in core reactivity by the corresponding change in core te=perature over a specific time period.
4.5.3 TEST RESULTS
.
Isothermal temperature coefficient measurements were conducted at four different reactor coolant boron concentrations during the zero power test program. The results of the measurements are sarized in Table 4.5-1 and in Figure 4.5-1. The calculated valuer are included for comparison. Good repeatability was demonstrated in all cases and the measured results compare favorably with the calculated values.
All measured te=perature coefficients.of resetivity were within the acceptance triteria of +0.4x10-'Ak/kOF of the predicted value. A calculation of the moderator
_
coefficient indicates that it is well within the requirements of Technical Specifi-cation 3.1.7.
4.5-1
){)k Gl\
.. . *
~
4.
5.4 CONCLUSION
S The measured values of the temperature coefficient of reactivity at 532 F, zero reactor power are within the acceptance criteria of 10.4x10-44k/k/0F of the predicted value. Calculation of the moderater coefficient indicates that it is well within the limits of Technical Specifications 3.1.7.
,
4.5-2
,
.
.
SUMMARY
OF 1EMPERATURE COEFFICIENT HEASUREMENTS
- AT TIIE ZERO POWER CONDITIONS OF 5320F AND 2155 PSIC
- RC Boron Control Rod Temperature Coefficient (x10~4Ak/k/ F) -
Concentration Position Heasurement Heasurement Average Calculated (jepm) (% withdrawn) (I) (II) (I) & (II) Results 1601 Cps 1-6 @100 +0.450 40.447 40.449 +0.488 Cp 7 0 78 Cp 8 @l00 1461 Cps 1-5 0100 40.306 40.302 40.304 40.200 cp 6 0 78 Gp 7@ 0 cp 8 @ 17 N (1) (2) (2)
,, 1269 Cps 1-3 @l00 -0.534 -0.520 -0.527 -0.710 L Cp 4095
& cps 5-7 0 0 Cp 8 @ 27 1245 Cps 1-3 @l00 -0.605 -0.603 -0.604 -0.860 Cp 4 0 50 Cps 5-3 @ 0 Cp 8 @ 27
, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
fIl (1) This la an average value based upon RC boron samples of 1261 ppm am, and 1276 ppm for measurements I and II, respectively.
cra (2) These results are from the heatup phase only.
N LN
.
.
.
.
TEffERATURE COEFFICIENT OF REACTIVITY VS "0R0t{ C0flCEllTRATI0tl 0 532 F, 2155 PSI, O EFPD 1
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4.6 SOLUBLE POISON WORTH 4.6.1 PURPOSE 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 feel throughout each core life cycle. The differential reactivity worth of the boric acid in terms of ppm boron was measured during the zero power test.
4.6.2 TEST METHOD Measurements of the differential boron worths at 532 F were performed in con-junction with the control rod worth measurements. The control rods worths were measured by the boron swap technique in which a boration/deboration rate was established and the control rods were withdrawn / inserted to compensate for the changing core reactivity. The reactimeter was used to provide a continuous reactivity calculation throughout the measurement. The differential boron worth was then determined by sumning 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.
4.6.3 TEST RESULTS Measurements of the soluble poison differential worth were completed at the zero power condition of 532 F. The measured results are plotted in Figure 4.6.-1 along with the calculated differential worths. The measured results are within 1.25% of the calculated worths and within the acceptance criteria limits of i.11%
dk/k/100 ppm. The results for only three out of five measurements are reported since the initial and final boron concentrations for two of the measurements were in question. The two results not reported, although within the acceptance criteria, are not considered representative.
4.
6.4 CONCLUSION
S The measured results for the soluble poison differential worth at 532 F were within 1.25% of the predicted values.
\h\h CTS 4.6-1
.
.
.
.
DIFFERENTIAL REACTIVITY WORT 110F SOLUBLE POISON VS BORON CONCENTRATION FOR H0DERATOR TEMPERATURE OF 532 F
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s Boron Concentration, ppn: ,
P O
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.. . .
4.7 CONTROL ROD GROUP k' ORTH EASUREENTS 4.7.1 PURPOSE The .otal amount of excess reactivity controlled at beginning-of-life (BOL),
hot (532 F), clean conditions is 17.8:Ak/k. During reactor operations, nearly all of the excess reactivity is controlled by the souble and lumped burnable poison systems. Additional control is provided by moveable control rods. This section provides comparison between the calculated and measured results for the control rod groun worths.
.
The layout of the core according to the standard alphebet-numeric mesh showing the initial location of the control rod groups is shown in Figure 4.7-1. The number of control rods and the reactivity control function of each group is given in Table 4.7-1. The grouping of the control rods shown in Figure 4.7-1 will be used until the core burnup is 250 EIPD. At that time, an interchange between groups 4 and 7 will be made. Calculated and measured BOL control rod group resetivity worths for the normal withdrawal sequence were determined at reactor conditions of 0
zero power, 532 F and 2155 psi. The calculated results were obtained using the PDQ code with either a two or three dimensional description of the core.
4.7.2 TEST ETHOD Control rod group reactivity worth measurements were performed at zero power, 532 F using the rod drop and boron / rod swap methods. The boron / rod swap method was used to measure the differential and integral reactivity worths of control rod groups 5 through 8 and parts of group 4. The total reactivity worth of rod groups 1 through 3 and part of group 4 was measured by the rod drop technique.
The boron swap meth.od consisted of establishing a deborstion rate in the reactor coolant system and compensating for the reactivity changes by inserting the control rod groups in incremental steps. In the rod swap technique (similar to the boron swap method), the reactivity changes caused by moving the rod group being measured are compensated for by moving another rod group. The reactivity changes that occarred during the measurecants were calculated by the reactimeter and differential rod worths were obtained from the k=own reactivity worth versus the change in rod group position. The differential rod worths of each group were then su=med to obtain the integral rod group worths.
In the rod drop method, critical equilibrium conditions were established with all the control rod groups to be measured withdrawn from the core. The control rod groups being measured were then trieped. The reactivity inser*ed in the core was calculated by the reastimeter. The total reactivity worth of rod groups 1 through 3 and part of group 4 was measured using the rod drop method.
4.7.3 TEST RESULTS Control rod group reactivity worths were measured at the zero power, 5320F condition.
The baron / rod swap method was used to determine differential and integral rod worths for control rod groups 5 through 8 from 100% to 0: wi~hdrawn, and for group 4 from 100 to SO: withdrawn. The differential and integral worth of group 8 frem 27 to 0: withdrawn was measured by the rod swap method using group 7. The rod drop method was used to obtain the total worth of groups 1 through 4 (group 4 from 50: to 0 withdrawn).
c.7-1 1414 0/7
.. . .
The results of the rod drop measurements on rod groups 1 through 4 are given in Table 4.7-2. Based on experience with previous startups, it was predicted that the rod drop measurements at TMI Unit I would yield values approximately 26% less than the correct value when considerably more than 1%Ak/k was being inserted. The deviation is caused by spatial flux changes in the core immedi-ately after the control rods drop and its sign and magnitude are a function of the total amount of reactivity inserted and the detector-control rod geometry.
The TMI results were consistent with these expectations, as seen in Table 4.7-2.
When a correction factor of 1.35 is applied to the measured value, the measured and calculated worths agree to within 2%.
The integral reactivity worths for control rod groups 4 through 8 are presented in Figures 4.7-2 through 4.7-6. These curves (1)were obtained by integrating the measured differential worth curves. A third order polynomial expansion was used to obtain a "best fit" differential worth curve from the measured rod worth data. The point of maximum reactivity insertion for the group 8 rods occurred at 27% withdrawn (0%Ak/k/: withdrawn differential worth) with a total worth of
-0.393%ak/k at this position. The integral worth of group 8 from 27% to full insertion was measured at +0.215%Ak/k. The group 8 rods were positioned at 27%
withdrawn during the reactivity worth measurements on group 1 through 7. Figure 4.7-7 is a plot of the total reactivity worth of groups 4 through 8 for the normal withdrawal sequence.
Table 4.7-3 provides a comparison between the predicted and measured results for the rod worth measurements. The calculated results were used as the best estimate for the worth of groups 1 through 4, based upon the rod drop results discussed above. The results show good agreement between the measured and predicted rod group worths. The maximum deviation between me.asured dand predicted was -8.33%.
Also presented in Table 4.7-3 are the expected control rod group worths at 579 F with the APSRs at 27% withdrawn. These values were obtained by applying the per-cent deviation between the meanured and predicted worths at 532 F to the predicted worths at 579 F.
4.
7.4 CONCLUSION
S Differential and integral control rod group reactivity worths were measured using the boron / rod swap and rod drop methods. The measured results at zero power, 532 F indicate good agreement with the predicted group worths.
--__
(1) Zero Power Physics rod worth data wa. processed by the Babcock and Wilcox computer in Lynchburg, Virginia to supplement on-site analysis. The results of that analysis are presented here as the best estimate of the measured control rod group reactivity worths.
4.7-2 }t'}k
.
.
.
REACTIVITY CONTROL ,
FUNCTION OF CONTROL ROD GROUPS .
Rod Group Number Control Number Of Rods Function 1 8 Safety 2 8 Safety 3 8 Safety i5 4 8 Safety n
L, 5 12 Power Doppler b
6 8 Power Doppler 7 9 Ttansient Xenon 8 _ Il Axial Power Shaping 69
-
M
-
5 O
N s f?
.
.
COMPARISON OF CALCULATED AND MEASURED CONTROL
- ROD GROUP REACTIVITY WORTilS FROM ROD DROP RESULTS
.
Moderator Temperature at 532 F APSRs at 27% Withdrawn Withdrawal Calculated Uncorrected Correcte Deviation From Rod Group Interval Worth Meas. Worth Heas. Worth Calculated Number (% Withdrawn) (%Ak/k) (%Ak/k) (IAk/k) (%)
g 1 0-100 0.89 -
2 0-100 3.01
.'* -5.99 4.33 5.85 -2 18 o
3 0-100 0.74 4 0-49 1.35 (1) Corrected measured worth is based upon an expected 26%
deviation between the measured and corrected results.
-
-
O
&
C.D
.
.
COMPARISON OF CALCULATED AND ,
HEASURED CONTROL ROD CROUP REACTIVITY WORTil ,
A. Moderator Temperature at 532 F, APSRs at 27% Withdrawn Predicted Measured Percent Rod Number Worth Worth Deviation I.O Croup Of Rods (%Ak/k) (%Ak/k) (%)
1 8 -0.89 -0.89 NA(3) 2 8 -3.01 -3.01 NA 3 8 -0.74 -0.74 NA 4 8 -1.86 -1.86 NA 5 12 -1.07 -1.03 -3.74 6 8 -1.22 -1.25 +2.46 7 9 -1.20 -1.10 -8.33 8 8 -0.38 -0.39 +3.42 h Total 5 -10.37 -10.27 r
"
.
[ B. Moderator Temperature at 579"F, APSRs at 27% Withdrawn Predicted Expec e Percent Rod Number Worth Worth 2 Deviation (l)
Croup Of Rods (%Ak/k) (%Ak/k) (%)
1 8 -1.40 -1.40 NAl3) 2 8 -3.59 -3.59 NA 3 8 -0.75 -0.75 NA 4 8 -1.47 -1.47 NA 5 12 -1.40 -1.35 -3.74 6 8 -1.48 -1.52 +2.46 7 9 -1.07 -0.98 -8.33
-"
8 8 -0.44 -0.46 +3.42
] Total 5 -11.60 -11.52 4
(1) Percent deviation is eniculated assuming predicted value is correct a (2) Expected worth is obtained by applying percent deviation between Co predicted and measured results at 532"F to the predicted results at
-
579"F (3) NA denotes that percent deviation is not applicable since the pre-dicted worths are used
., , .
CONTROL ROD GROUP LOCATI0h5 X
A B
(7) (4) (7)
C (5) (3) (3) (5)
D (4) (8) (6) (8) (4)
E (5) (6) (1) (1) (6) (5)
(7) (8) (2) (2) (2) (81 (7T G
(3) (1) (5) (5) (1) (3)
WH (4) (6) (2) (7) (2) (6) (4)
K (3) (1) (5) (5) (1) (3)
-
(7) (8) (2) (2) (2) (8) (7)
M (5) (6) (1) (1) (6) (5)
N (4) (8) (6) (8) (4) 0 (5) (3) (3) (5)
(7) (4) (7) l R
I i l i i 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 2
(X) : Control Rod Group Number 1414 082 Figure 4.7-1
. . . .
,
Control Rod Group k Integral Worth At Zero Power, 53207, O EFPD Total Worth = .k19% t.k/h 05 ' - -
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p 0
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. . . .
Control Red Group 6 Integral Worth At Zero Power, 5320F, O EFPD Total Worth = 1.25% ak/k 1.3 '
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b b,3 FIGURE 4.7-4
} k 'l k
. . . .
Centrol Rod Group 8 gtegre.1 Worth At Zero Power, 532 F, O EFPD 0.0 ' '
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Total Reactivity Worth Versus Rod Withdrawal
- At Zero Power, 532 F, O EFFD
- 0.0 g ,,,,,,4, ?.
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Withdrawal Sequence 1, 2, 3, is , 8 -'
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0 20 1:0 60 80 100 20 40 60 80 100 0 20 1:0 60 80 100 Group 8 Group 5 Group 7 m
4
-
4 Control Rod Position, % ud O
C N
.. . .
4.8 EJECTED CONTROL ROD WORTH 4.8.1 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.C:
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.
4.8.2 TEST METHOD Pseudo ejected control rod worths were measured at zero power using two different techniques. The first technique was the boron-swap method during which the boron concentration of the reactor coolant system was slowly and continuously increased.
The pseudo ejected rod was withdraws in quick steps 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, the rod drop method, critical equilibrium conditions were established with the pseudo ejected rod withdrawn to 100%. The ejected rod was then dropped into the core and the neutron flux and reactivity was logged by the reactimeter every 0.2 seconds. The measured instantaneous worth of the dropped rod is taken as the worth of the ejected rod.
4.8.3 TEST RESULTS Pseudo ejected control rod reactivity worth was measured at the zero power con-ditions of 532 F, 2155 psig. Rod worth calculations performed for several Doppler and Transient control rods indicated that core location F-2 (and those locations symmetrical to it) was the highest worth rod position. Control rod 8 in group 7 was selected for the ejected rod measurement. Figure 4.8-1 shows the location of rod 7-8 in the core.
Critical equilibrium conditions were established for the boron-swap measurement with an initial RCS boron concentration of 1269 ppm and control rod group 5 at 6%,
groups 6 and 7 at 0% and group 8 at 27.5% withdrawn. Control rod 7-8 was with-drawn to 100% to compensate for borating the reactor coolant to 1337 ppm. The worth of rod 7-8 from this measurement was 0.688%Ak/k.
In the rod drop method, the reactor was just critical with rod group 5 at 13% with-drawn, group 8 at 27.5% and groups 6 and 7 at 0% withdrawn. Control rod 7-8 was at the 100% withdrawn position. Under these conditions, rod 7-8 was dropped into the core and its resultant reactivity worth was obtained from the reactimeter.
The worth of the ejected rod by the rod drop method is 0.664%Ak/k, which compares well with the boron swap result. The calculated and measured results are compared in Table 4.8-1.
4.
8.4 CONCLUSION
S 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 best esti= ate for the measured value, 0.688:ak/k, is below the calculated worth, but this is more conservative with respect to an ejected rod accident. The Technical Specification requirement that the value not exceed 1.0%
ak/k is satisfied.
4.S-1
)h)k b
.
.
COMPARISON OF PREDICTED AND MEASURED PSEUDO .
EJECTED ROD WORTIIS AT 'IllE ZERO POWER, 5320F CONDITION .
I. Calculated . II. Measured Rod Positions Reactivity Worth Rod Positions Reactivity Worth
(% withdrawn) (% Ak/k) Method (% withdrawn) (% Ak/k)
Croups 1-4 0100% 1.0% Bo ron-Swap Groups 1-4 0100% 0.688%
Croups 5-7 0 0% Croup 50 6%
Group 8 037.5% Croups 6-7 @ 0%
g Group 8 @27.5%
t$
"
.
T' Rod Drop Groups 1-4 0100% 0.664%
"
Croup 5 @ 13%
Groups 6-7 @ 0%
Croup 8 027.5%
-
-
O CO D
.. . .
CONTROL ROD LOCATIONS FOR EJECTED ROD MEASUREMENT AT 532 F, 2155 PSI, O EFPD A
a O O C -
O O __
D -
e O 9 E
O O O O _
-
F -
e G O c _
O O H -
0 0 O K -
O O L -
O s @ O M
'
O O O O N - O O O O -- O O
'
P O O R
IIII I i
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 h APSR Location O CR croup, 5, 6, 7 Calculated Measured Rod Ejected Ejected Rod Worth Rods Inserted Ejected Rod Worth F-2 1.0%3k/k O ,6 Dopp 1er, 0.6ss:3kfkg Transient and 0.664 Ak/k APSR Groups (1) Boron Swap Result (2) Rod Drop Result
\g\h C93 FIGi'RE 4.8-1
.. ..
,
4.9 SHUTDOWN MARCIN 4.9.1 PURPOSE Technical Specification 3.5.2 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 stuck rod worth measurement at zero power, 5320F was to verify that the calculated stuck rod worths are conservative compared to the measured value.
4.4.2 TEST METHOD The minimum available shutdown margin and the worth of a simulated stuck control rod were measured by performing two rod drop measurements. In the first measure-ment, all control rods not in the core were dropped (except APSRs). The simu-lated stuck rod was then withdrawn to 100%, and the remainder of the safety groups were then withdrawn to establish critical equilibrium conditions at the same boron concentration as the first measurement. All control rods except the stuck rod and the APSRs were then tripped. The difference in the reactivity inserted in the two measurements was taken as the stuck control rod worth. The minimum shutdown margin available was obtained directly from the second rod drop.
4.9.3 TEST RESULTS The most reactive control rod at zero power, 532 F was calculated to be rod 7 in group 4 (core location H-2) and those control rods sy= metrical to it. Rod 4-7 was selected for the stuck rod measurement and all drops were made with the APSRs at 26% withdrawn. Figure 4.9-1 shows the core location of control rod 4-7. For the first drop, critical equilibrium conditions were established with group 4 at 49: withdrawn at a 1197 ppm baron concentration. The reactivity inserted in the core by dropping all the rods was 4.33%Ak/k. For the same boron concentration, control rod 4-7 was withdrawn to 100% and critical conditions were established with group 3 at 85% withdrawn. For the second drop, all control rods except rod 4-7 were tripped. The reactivity measured from inserting all control rods except the simulated stuck rod was 2.16:ak/k. The reactivity measured for the second drop gives the minimum shutdown margin available with the most reactive control rod stuck out. The difference in measured reactivity inserted in the two drops is the measured worth of the stuck rod.
Correction factors were applied to the measured reactivity value from the reacri-meter to corrtet for changes in the spatial flux distribution i= mediately after the rod drops. Table 4.9-1 lists the corrected and uncorrected measured worths and provides comparison with the predicted worths for the stuck rod. The corrected measured results for drops one and two are 5.85:ak/k and 2.01%ak/k, respectively.
This results in a stuck rod worth value of 3.84%Ak/k, which compares favorably with the predicted value of 3.91:ak/k. The minimum available shutdown margin with the most reactive control rod stuck out of the core was measured to be 2.01%Ak/k.
If a 10% uncertainty is assigned to the measurement, the minimum shutdown margin is at least 1.8%Ak/k and this ensures that the Technical Specification requirement
'is met.
4.9-1 }k}k 09
.. . .
4.
9.4 CONCLUSION
S Minimum shutdown mergin verification and stuck control rod worth measurements were completed for the zero power condition at 53207. The measured value of the most reactive control rod stuck out of the core with all other control rods inserted (except APSRs) was 3.84%Ak/k. The shutdown margin available under this condition was at least 1.8%Ak/k which guarantees that the Technical Specification limit of 1.0%Ak/k is satisfied.
1414 092 4.9-2
.
.
'
MEASURED AND CALCULATED VORTil 0F
'
STUCK CONTROL ROD AT ZERO POWER, 532 F Rod Uncorrected Corrected Predicted Drop Croups lieasured Worth Correction Measured Worth Worth No. Dropped (I) (%Ak/k) Factor (%Ak/k) (ZAk/k) 1 1, 2, 3 & 4 4.33 1.35 5.85 5.94 (4 at 49%)
2 1, 2 & 3 2.16 0.93 2.01 2.03(2)
(3 at 85%,
el Rod 4-7 remains fj nt 100%)
to Stuck Rod Worth 3.84 3.91
_ _ _ _ _
(1) APSRn were at 26% withdrawn for both measurements
-
(2) Based upon Stuck Rod Worth of 3.91% Ak/k M
-
5 O
W
(>4
-. .
.
CONTROL ROD LOCATION FOR STUCK ROD
.
MEASUREMENT AT 532 F, 2155 PSI, O EFPD A - -
B' O O O C O' 3 O O D O O O O O E o O O O O O F -
O 9 O O O 6 O c -
O O O O O O H -
O O O O O O K -
O O O O O O L -
O 4 O O O @ O l M
N H
-
O O
O O
O O
O O
O O
O o --
O O O O
'
P . O O O l l l l l l
\ \ l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 b APSR Locations O Au Rods in Core Corrected Corrected Measured Stuck Drop 1 Results Predicted Drop 2 Results Rod Worth Rod Worth Control Rod (All Rods) (All but 4-7) (%Ak/k) (%Ak/k) 7 5.85 2.01 3.84
_ 3.91 1414 094 FIGURE 4.9-1
.. ..
5.0 CORE PERFORMANCE - MEASUREMENTS AT POWER This section presents the results of the physics .neasurements that were conducted with the reactor at power. Testing was conducted at the four major power plateaus of 15%, 40%, 76% and 100% of 2535 megawatts themal core power, as determined fro =
primary and secondary calorimetr2c measurements. Operation in the power range began on June 15, 1974. Power escalations occurred as the required testing at each plateau was successfully completed.
Periodic measurements and calibrations were performed on the plant nuclear instru-mentation during the escalation to full power. The. Jour power range detector channels were calibrated based upon primary and secondary plan.: heat balance measurements.
Testing of the incore nucient instrumentation was performed to ensure that all detectors were functioning properly and that the detector outputs were processed correctly by the plant computer. Core axial imbalance determined from the incore inctrumentation system was used to calibrate the out of core detector imbalance indication. Radiation surveys of the biological shield and reactor and auxiliary buildings were conducted to obtain base line data on accessible work areas while the reactor is operating at power.
The major physics meacurements performed during power escalation consisted of determining the moderator and power Doppler coefficients of reactivity, determining the worth and associated power distriautions effected by simulated dropped and ejected control rods, and obtaining detailed radial and axial core power distri-bution 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.
A su= mary of the tests reported in this section, including the respective section number and power level at which they were performed, is given in Table 5.0-1.
The core power history and integrated burnup up to August 27, 1974 is presented in Figures 1.0-1 and 1.0-2, respectively.
)h\h 5.0-1
.
.
.
'
SUMMARY
OF TESTS REPORTED IN SECTION 5.0
.
Section Test Power Levels, % FP Number Title of Section - Test Procedure Number =5 15 25 35 40 50 65 76 85 95 100 5.1 Biological Shield Survey - TP 800/3 X X 5.2 NI Calibration at Power - TP 800/2 X X X X X X 5.3 Incore Detector Testing - TP 800/24 X X X X X 5.4 Power Imbalance Detector Correlation - TP 800/18 X X 5.5 Rod Reactivity Worth Heasurements - TP 800/20 X X X
$ 5.6 Reactivity Coefficients at Power - TP 800/5 I X X 5.7 Dropped Control Rod Test - 17 800/31 X X 5.8 Pseudo Ejected Control Rod Test - TP 800/33 X 5.9 Core Power Distributions - TP 800/11 X X X X 5.10 NSS lleat Balance - TP 800/22 X X X X X X X X X X X 5.11 Reactivity Depletion Versus Burnup - TP 800/16 X 5.12 Neutron Noise Heasurements X X X
-
__
C- )
NO p
.. . .
5.1 BIOLOGICAL SHIELIf SURVET 5.1.1 PURPOSE The purpose of the biological shield survey was to measure the radiation levels in all accessible locations of the plant adjacent to the biological shield and to obtain base line radiation levels for comparison with future measurements of radiation levels during plant operation.
5.1.2 TEST METHOD The biological shield survey was conduct ed at zero reactor power and at 40% and 100% of full power. The Reactor Building outside of the biological shield or areas designated as access areas were marked off in horizontal and vertical zones and readings were taken in discrete sections. All areas in the Auxiliary Building were also surveyed. The surveys were conducted using portable ionization and GM counters for gamma radiation and B7 e u ters r neutr ns. A rea Es were taken within one inch of the shield3 wall. Readings were taken af ter fifteen hours of steady state operation at the specified power was attained.
5.1.3 TEST RESULTS The results of the biological shield survey at each power level where the test was conducted are summarized in the table below.
4 Power Gamma / Neutron Gamma / Neutron Date Level Average (mrem /hr.) Maximum (mrem /hr.)
6/6/74 0% <0.03/0 0.15/0 7/1/74 A0% <2.34/1.0 14/4.8 8/8/74 100% <5.88/8.7 40/60 The above results apply to the inside of the Reactor Building only in those areas outside of the biological shield. The maximum radiation level measured was 60mr/hr neutron at elevation 365', which is above the shield area.
5.
1.4 CONCLUSION
S The maximum radiation levels found in all accessible areas were below 100 mrem /hr, and therefore, the biological shield meets all design criteria.
z41A 097 5.1-1
.. .*
5.2 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER 5.2.1 PURPOSE The purpose of Nuclear Instrumentation Calibration at Power was to calibrate the power range nuclear instrumentation indication to within 12% FP of the reactor thermal pcwcr as determined by a heat balance and to within 15 percent incore axial offset as determined by the incore monitoring system. Additional purposes during the power escalation program were as follows:
(a) To adjust the high power itvel trip setpoint when required by the power escalation procedure.
(b) To verify that at least one decade overlap exists between the intermediate and power range nuc3ae. instrumentation.
Two acceptance criteria are specified for nuclear instrumentation calibration at power as listed below.
(1) The power range nuclear instrumentatf.on indicates power level within 12% FP of the power level determined by heat balance and within 15 percent of the incore axiaJ offset as determined by the incore detectors.
(2) The high pouar level trip bistable is set ' a trip at the desired value,
+9 5% FF.
5.2.2 TEST METHOD As required during power escalation, the top and bottom linear amplific" gains were adjusted to maintain power range nuclear instrumentation channel power indication within 12% of the power calculated by a heat balance. During top and bottom licear amplifier gain adjustment, the ratio of their gains was maintained constant as long as the indicated axial imbalance was within 15% of incore i= balance; if not, their gains were adjusted to correct imbalance and heat balance mismatch at the same time.
Data was also taken to verify overlap between the intermediate and power range channels. The required overlap was a minimum of one decade between these two nuclear instrumentation ranges.
When directed by the power esenlation procedure and/or the unit startup precedure, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given below:
Test Plateau Bistable Setpoint IFF %FP ,
15 50 50 60 76 95 100 104.75
.
1414 093 5.2-1
.. . .
5.2.3 TEST RESULTS An analysis of test resulte indicated that changes in Reactor Coolant System boron and xenon buildup or uurnout affected the power as observed by the nuclear instrumentation. This was as expected since the power range nuclear instru=en-tation measures reactor neutron leakage which $s directily related to the above changes in system conditions. Channes in these system conditions resulted in a nuclear power range indication increase or decrease of approximately 5 to7%FP. Each time that it was necessary to calibrate the power range nuclear instrumentation, the acceptance criteria of calibration to within 12.0%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 +5% axial offset criteria as determined by the incore monitoring system was also met. Table 5.2-1 is a summary of the data taken during calibration at different power levels during power escalation testing. In all cases, the nuclear instrumentation was adiusted to within 2.0%FP of the heat balance and to within 15% incore axial offact.
The high flux trip bistable was adjusted to 50, 60, 95 and 104.75% FP prior to escalation of power to 15, 50, 76 and 100% FP, respectively. Acceptance criteria of adjusting the setpoint to the above values within 10.5 % FP was met each time without difficulty.
The overlap meacured during the startup program included the total span of the power range, exceeding the one-decade overlap requirement. Figure 5.2-1 shows the overlap of all three nuclear instrumentation channels.
5.
2.4 CONCLUSION
S The power range channels were calibrated to within two percent of heat balance powcr several times during the startup program. These calibrations were required due to power level, boron, and/or control rod configuration changes during the program.
Acceptance criteria for nuclear instrumentation calibration at power were net in all instances, qqQ'
\hhh U' 3.2-2
.
.
SUltfARY OF NUCLEAR INSTRUMENTATION CALIBRATION ',
AT POWER RESULTS PERET)RHED DURING POWER ESCALATION IIeat Balance Incore Power Before and Af ter Calib. , % FP Power Imbalance Imbalance Before and After Calib. . % FP
(%FP) (%FP) NI-5 NI-6 NI- 7 NI-8 NI-5 NI-6 HI-7 NI-8 12.96 16.87 16.87 17.0 16.5 12.96
-7.53 -4.00 -4.44 -4.39 12.2 13.0 13.0 14.0 -5.83 -3.28 -3.15 -3.70 30.40 28.0 27.6 27.5 27.5 -7.63 -2.75 -3.13 -3.38 30.40 29.1 29.0 28.5 29.3 -8.00 -3.13 -3.25 -3.88 s 40.05 1.42 39.1 39.6 38.8 40.0 5 -2.88 3.31 1.84 2.81 40.01 0.47 40.1 39.5 39.7 39.0 0.53 -0.91 N -1.05 -0.94
- v. 75.95 0.59 76.8 77.1 76.9 76.9 3.90 4.50
- 76.40 3.20 4.00 w 1.38 76.0 76.6 76.1 76.6 0.40 1.10 0 0.60 0.40 85.85 1,30 83.0 84.2 82.6 84,0 85.85
-3.70 -2.60 - 2. 0') 3 07 1.30 85.5 84.1 85.5 85.1 0.30 2.00 1.20 1.60 95.19 -1.30 93.6 91.7 93.7 92.9 -3.70 -1.90 -2.40 -3.30 95.30 -2.?1 96.0 94.4 95.5 94.9 0.00 0.20 -0.30 0.J0 99.6 -2.46 99.0 98.0 99.0 99.0 0.34 99.6 0.41 -0.16 0.59
-2.46 99.0 98.0 99.0 99.0 0.34 0.41 -0.16 0.59
-
--
_.
C.D
. . . .
NUCLEAR INSTRUMENTATION TLL7 RANGES
~ 104 9 -3 10 -
_g l
10 - -4 - 125 ,
- 10 uz
- 102 - 100 m: -
- 7 -5 Sy 10 ,
10
- 101 g - 10 $E
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E te
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-8
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- $ 103 -
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w 3 y 10 2 8 m
E 102 .
10' -10
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-
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100 ~
-6 10
_ g B8 10 - 10 5 7 om 10-2 _ ~ 10
_ 10~0 10 -3 _
10
,
20 ih\k \
Tigure 5.2-1
.. .a
.
5.3 INCORE DETECTOR M STING 5.3.1 PURPOSE A system of self-powered-neutron-detectors (SPNDs) is installed in the TMI Unit I reactor core. These detectors monitor core powerdensity within the core and their outputs are monitored and processed by the plant computer to provide accurate readings of relative neutron flux. Although the incore detector system serves no safety related function, it does provide detailed core power distribution data which will be used throughout core life for physics and fuel performance calcula-tions. In addition, infor=ation from these detectors formed an integral part of the startup test program.
Tests conducted on the incore detector system during power escalation were per-formed to:
(a) Verify that the output from each detector and its response to inercasing reactor power was as expected.
(b) Verify that the background, length and depletion corrections applied by the plant computer are correct.
(c) Calibrate the backup incore recorder.
5.3.2 TEST METHOD 5.3.2.1 Incore Detector System Power distribution within the core is measured at 364 locations (7 axial positions in 52 fuel assemblies) by the Incore Detector System's self-powered neutron detector. The 52 incore monitor assemblies are placed at preselected radial positions as shown in Figure 5.3-1. Seventeen detector assemblies are positioned to act as symmetry monitors and the remaining 35 assemblies, with 5 of the 17 symmetry monitors, monitor every other fuel assembly position assuming quatercore symmetry. Each asse=bly contains seven equally spaced flux detectors corresponding to seven axial core elevations to provide measurement of axial flux shape.
The self-powered incore detectors use rhodium wire detectors which undergo electron emmission when placed in a neutron environment. The capture of a neutron by rhodium - 103 produces the radioactive isotope rhodium - 104. The radioactive decay of rhodium - 104 emits a beta particle and creates a daughter product that requires one more orbital electron than the parent. This crbital electron is the source of the self-powered detector signal when the only free electron path to the detector is an electrical conductor in series with a current measuring instru-ment. Figure 5.3-2 shows the basic components of the self-powered neutron detector.
Since beta decay of rhodium - 104 is not the only source of electrons in the core, the current measured from the individual rhodium detectors must be corrected for a background current. The background current is measured by background detectors in each of the 52 incere monitor assemblies.
The outputs from the detectors are read and processed by the plant computer. The co=puter applies background, as built manufacturing and depletion correction factors to each detector reading. The incore detector infor=ation is then used in core physics calculatiens and/or provided for display to the reactor operators. The 5.3-1 14\4 \D7"
.. . <
readings from 36 incore detectors are also monitored by two 24 point recorders.
The recorders provide power distribution information to the operator at times the plant computer is not available.
5.3.2.2 Incore Detector Tests The response of the incore detectors versus power level was determined and a comparison of the symmetrical detector outputs made at reactor powers of 15, 25 35 and 40% FP. Once steady state conditions were achieved at each of the above power levels, a printout of corrected and uncorrected SPND maps for all F.tectors was obtained from the plant computer. Figures 5.3.-3 and 5.3-4 are sample print-outs of the uncorrected and corrected maps, respectively. The corrected and uncorrected rhodium detector readings and the background readings, in units of nanoamps, were then plotted versus reactor power level to verify that each detector was responding as expected. The values of all symmetrical detectors were compared to verify that they were the same within allowable deviations.
The plant co=puter applies background, as built manuf acturing and depletion correction factors to each rhodium detector, as mentioned above. Hand calculations were performed at 40% FP to verify the computer calculated corrections using uncorrected SPND outputs and SPNDI values from performance data output segment number 6. SPNDI is the accu =ulated nanoamp sum for each detector.
Data was collected to calibrate the backup incore recorder at 40% and 76% FP. The readings from two 24 point recorders located in the control room were recorded while obtaining a printout of the corrected detector readings from the plant computer.
The recorder indication was then adjusted to agree with the corrected detector outputs.
5.3.3 TEST RESULTS Incore detector testing during the startup program indicates that the detectors are responding as expected. Typical results of the tests are shown in Tables 5.3-1 and 5.3-2 and Figure 5.3-5.
Tables 5.3-1 and 5.3-2 show the comparison made between two sets of sy=metical detectors at 40% FP. Corrected SPND values for each detector at all seven axial levels were recorded and the highest and lowest detector readings at each level were deter =ined. The acceptance criteria for the test required that the difference between the average value and the highest and the lowest detector readings be within 5% of the average value for a given level. The 5% acceptance criteria was met in almost all cases; however, some detector readings were greater than 5% from the average value. The maximum difference between a detector reading and the average value was 11.6%.
The differences that were observed between the sy= metrical detector readings were attributed to errors caused by a loss of some of the detector signal. The a=ount of signal loss varies for each SPND, but is proportional to the value of the dropping resistors used to =easure the detector current. The results of measurements per-formed at 76% FP indicate that a reduction in the size of the dropping resistor will decrease the error associated with the measured signal. Plans are in progress to replace the present resistors.
5.3-2 1414 10 '>
.
.. 4 Both the corrected and uncorrected readings from each EPND and the background detector signals were plotted at 15%, 25%, 35 and 40: FP. An example of the response of an SPND with increasing reactor power is shown in Figure 5.3-5.
The outputs of the SPNDs increase linearly with power, as expected, and show some variations for different core locations and control rod position.
Hand calculations were perfomed at 40 FP to verify the length, background and depletion correction factors applied to each SPND by the plant computer. The hand calculations were performed for detector string 1, 7, 35 and 52 by obtain-ing the SPNDI (accumulated detector burnup) values from the plant ce=puter and then applying the appropriate length, background and depletion corrections. The correction factors derived from the hand calculations agreed to within 2% of the computer calculated values.
Calibration of the backup incore recorders was performed at 40% and 76: FP. The 36 detector signals were recorded from the backup recorders at the same time that corrected datector readings were obtained from the plant computer. A hand calculation of (P/Pc) was performed from the computer data and these values were factored into the adjustments made to the backup recorder readings. The advantage of using this technique is that in additon to having a readout of relative neutron flus: on the backup recorder, the informationdisplayed also incicates flux peaking in the core. A sample calculation for the backup incore halibration is given in Tables 5.3.-3 through 5.3-5.
5.
3.4 CONCLUSION
S Incore detector testing during power escalation demonstrated that all detectors were functioning as expected that sy= metrical detector readings agreed within acceptable limits and that the computer applied correction factors are accurate.
The backup incore recorder was calibrated and operational above 80: FP as required by the Technical Specifications.
)g\4 \D6 5.3-3
.. .i SYMMETRICAL DETECTOR COMPARISON 6/30/74 402 FP Dec. Core
-
Corrected Nanoamps No. Location 1 2 3 4 5 6 7 5 E=9 331 500 486 418 376 304 154 7 E-7 334 502 496 423 392 298 144 9 G-5 329 502 489 415 400 318 156 11 K-5 327 495 481 420 396 315 153 13 M-7 322 497 485 414 375 311 156 16 M-9 344 505 493 422 391 303 142 19 K-11 330 502 487 417 393 308 152 25 G-11 331 505 480 421 390 308 154 Average 331 501 487 419 389 308 151
._
Eighest 344 505 496 423 400 310 156 Lowest 322 495 480 414 375 298 142 13 4 9 4 11 10 4 (1) Difference 9 6 7 9 25 10 6 a
-
muumme mammam -um (1) Difference is taken between the average value and the highest and the lowest readings.
)k)h \
TABLE 5.3-1
.. .*
STHMETRICAL DETECTOR COMPARISON 6/30/74 402 FP Det. Core -
Corrected Nanoamps No. Location l1 2 3 4 I5 6 7 22 F-13 224 300 286 221 227 191 97 28 Colo 229 309 287 226 225 192 97 32 C-6 228 300 283 223 221 188 92 35 F-3 230 310 287 226 229 191 95 39 L-3 226 305 282 225 220 194 96 43 0-6 228 311 289 224 218 191 93 47 0-10 229 308 283 222 217 179 95 50 L-13 225 304 277 228 226 193 98 Average 227 306 284 224 ,223 190 95
,
Highest 230 310 289 228 229 194 98 Lowest 224 300 277 221 217 179 92 i
3 4 5 i 4 i 6 4 3 (1) Difference a embe a '
' e 2 .a (1) Difference ir caken between the average value and the highest asu the lowest readings.
)&\4 TAELE 5.3-2
.
I C. lcm).tten She.t 1 ..
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I Calculattoa Sheet :
CORE FoutR DISTRIBUTI'18 fro. 17tC0tt DETECTOR READIIsC5 ,
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,, Adjus ed incera Detector Readings (na)-telative Qser-- level nlI nel 2 Luel 3 Len t 4 Luet 5 et 6 Lad 7 Adjusted Weighting Relettve location number ** 3,,g,y g,,,,, ,,,,g n-a i '01.5 519. 555.3 520.5 473.6 396.7 209.1 3075.9 3075.9 c.s a 'e54.7 646.1 658.4 612.4 577.5 455.7 235.5 3640.3 4 1456,1.3 r-e 4 454.5 684.1 702.8 641.1 611.4 455.3 240.5 3789.7 ' BlO t-n 10 502.8 732.2 706.6 674.8 613.4 469.0 213.0 3811.7 4 15646.7 .
o- s i' 'e 38 . 9 677.8 634.1 593.5 588.0 328.2 150.1 3407.1 4 NG c-e at 't 64 . 9 688.1 659.5 611.3 572.4 409.6 207.9 3613.7 5 14454.8 s-S 20 464.9 704.1 711.4 661.3 620.h_ 468.0 204.8 3835.1 4 15340.4 A-s n 309.2 461.8 469.1 439.8 428_3 322.8 147.1 2578.3 4 10313.2 c-' 3 435.0 619.3 630.2 621.6 558.3 460.0 14Q 5 3565.3
- 14261.1 y '-'u 82 463.1 670.Q_ 645.0 580.2 571.6 444.3 210.8 3585.1 '
14340.4 8-18 26 515.1 782.8 610.3 3.65.5 666.4 337.0 153_9 2995.7 4 11982.9 D-:2 't 337.3 501.3 451.4 428.7 420.1 350.6 158.5 2648.0 4 10592.1 v' Co
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- .1-ic '6 156.7 224.6 230.9 218.3 204.8 159.8 80.4 1275.6 a 10204.0 m "si n '.27.0 611.3 518.1 463.9 600.1 327.9 194.6 3126,3_ e 25010.. F,,, , - 408.877 c-: 42 327.2 476.3 445.9 396.7 435.7 338.1 176.5 2576.8 s 20774.0
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P-ir u 165.1 244.5 243.4 229.8 224.6 111.3 77.3 1356.0 s WB C
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F[P CALIBRATICN SIEET for BACK'JF RECORDER A Computer Printout Reactor : THI-I Date7/17/76 Tina 0400 core Power Level 76 2 rF ,
Forria t 48 Initials of Analyst DAI.
.
Recorder Incore Detector Corrected Symsetric Power / Signal ' Adjustel P 3 ,g Calculated Initial Correction Present Correct pin
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incore Dec. FA location correction Det. Real. Recorder Recorder factor Recorder Readin3 r[ umber Reading ** in 1/8 core factor Fg P Reading,MR Reading 12, (MR /lfRg ) Read.N2 .'at OtRg /MRg )
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__,1 c-11-6 25 573 E- 9-6 0.7809* 447 1.093 83.1 70 1.187 .234MV .274MV 2 i:- s-6 1 330 it- s-6 1.024 338' .826 62.8 52 1.208 .162MV .202HV _
3 F- 3-6 35 352 c-lo-6 1.024 360 .880 66.9 65 1.062 .215MV .228 MV g
4 c-II-4 25 753 E- 9-4 0.7809* 5 5 8 . _ __ 1.438 109.3 102 1.072 .348MV .373MV h; 5 it- a-4 1 433 it- s-4 1.202 520. 1.272 96.6 72 1.342 .238MV .321MV M 6 z- 3-4 35 398 c-lo-4 1.024 408. .998 75.8 68 1.115 .239MV .255MV u 7 c-11-2 25 864 E- 9-6 a.7so9* 675. 1.651 125.5 125 1,004 .417MV 419MV y* 8 1t- 6-2 1 432 H- 8-2 1.202 519._ 1.269 96.5 71 1.359 .237MV .322MV 9 r- 3-2 35 525 c-lo-2 1.024 538. 1.316 100.0 82 1.219 .267MV .325MV 10
_ _
c- 6-: 32 .511 c-lo-2 1.024 523._ 1.279 97.2 86 1.130 .287MV 324MV al c-io-2 28 526 c-lo-2 1.02'+ 539._ 1.318 100.2 90 1.113 .29aMV .333MV 12 r-11-2 23 515 c-lo-2 1.024 527. 1.289 97.9 98 .999 .324MV .324MV 13 I.-13-2 50 525 c-lo-2 1.024 538. 1.316 100.0 94 1.064 .31&iV .336HV 14 0-10-2 47 534 c-lo-2 1.024 547. 1.338 101.6 98 1.037 . 3 31MV_ .341tv 15 o- 6-2 43 530 c-lo-2 1.024 543. _ 1.328 100.9 91 1.109 .307HV .34Giv 16 L- 3-2 39 496 c-lo-2 1.024 508. 1.242 94.4 87 1.085 .267MV . 29Qiv 17 K-Il-6 19 581 E- 9-6 0.7E03* 454. 1.110 84.4 70 1.205 .231MV .278MV la l H- 9-6 16 556 E- 9-6 0.7609* 434._ 1.061 80.7 72 1.120 .232HV .260HV
, *
- Corrected Incore Detector Reading (nanoamps)-
- Cioup 1 reds are 100% withdrawn vlenever ccrrected SPPO signals. Croup 48. caere power distribution is teken.
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INCORE DETECTOR LOCATIONS
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Y O s7==etr7 monit. ors g Total core monitors based on 3: core symmetry h Combination total core and sy= metry monitors r1 cure 3.3-1 1414 110
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Incore Self-Powered Neutron Detector
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UNCORRECTED SPND MAP
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L*lCORRECTED SPND MAP LEVEL 6 ALL LOCATIONS IN NAN 0 AMPS 06:01:01 06/2C/74
,, ,n 99 130 32 24 28 52
. no. itn . sns. e r.
33 27 51 107 145 41 3 r- 1 5 o r, 108
- 160*
- 165* 124 35 6 4 24 23
- 101- 13E 1E f- l'a + In**
36 9 8 3 25 22 106
- 172* 136 165
- 165* 131 37 la 1 ? 21 83 137 86 124 122 11 19 20
-
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3R 39 12 18 50 49
- 104* 166 127
- 103*
- , ,e ,, ,, o 4a 121
- 168*
- 163* 127 6F 41 14 15
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42 43 47 48 114
- 102*
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- 95* 79 54 45 46 7:. '1
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i FIGURE 5.3 - 3 }k\ '
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CORRECTED SPND MAP
. - . - - . . . - -
. - - . . . . . . .
CORRECTED SPND fMP LEVEL 5 ALL LOCATIONS lP' NANONiPS 06:01:01 06/29/74
+1 3a _
- _ _ _ _ _ . . . . . . . _ _ _
130 132 32 29 23 52
. ,;,. *ni . ses._ 3, 33 27 51 143 140 62
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3' 7 2c 149
- 228*
- 21F* 19F 3F 6 4 24 23
- 131- 1*7 234 1r* -
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36 9 8 3 25 22 147
- 231* 189 222
- 222* 187 y' in i 9 91 124 1RE 119 173 177 11 19 20
+ 229+ + 2?'+ !71 38 39 12 la 50 65
- 127* 226 170
- 130*
re it 17 rio ic 158
- 216*
- 226* IGP 83 41 14 15
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- 124*
- 124* 103
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)Q)h FIGURE 5.3 - 4
- . . .
J INCORE DETECTOR RESPONSE VS REACTOR POWER
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)h\h FIGURE 5.3-5
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5.4 POWER IMBALANCE DETECTOR CORRELATION TEST 5.4.1 PURPOSE The Power Imbalance Detector Correlation Test has three objectives:
(a) To determine the relationship between the induced power distribution as indicated by out-of-core instrumentation and the actual incore power dis-tribution.
(b) To verify the adequacy of the i= balance system trip setpoini:s.
(c) Verify the adequacy and a: curacy of backup imbalance calculations using OP 1203/7, " Power Imbalance and Quadrant Power Tilt Calculations Using the Backup Incore Detector System".
5.4.2 TEST METHOD This test was conducted at 40% and 76% FP to deter =ine the relationship between the core axial imbalance as indicated by the incore detectors and the out-of-core detectors. Based upor. 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.
The methed e= ployed to conduct the test at both power levels was the same. The sequence is outlined below:
(a)' Steady State conditions were est.ablished at the desired power level with core xenon concentrations at equilibrium.
(b) The Incore Monitoring System was verified as operational and the backup re-corders were checked as having been calibrated in accordance with TP 800/24, "Incore Detector Testing".
(c) The unit computer was verified as operational with applicable Nuclear Steam System (NSS) programs functioning properly.
(d) Prior to conducting this procedure, Nuclear Instrumentation Calibration at Power was used to calibrate the out-of-core detector imbalance to read within 0%,+1% of the incore imbalance.
(e) Baseline data consisting of the following was collected:
(1) Computer Group 33, Nuclear Instrumentation (2) Computer Group 55, Imbalance / Tilt / Insertion (3) Computer Group 34, 3-D Power Map (4) Computer Group 45, Uncorrected SPND Map for levels 1 thru 7 (5) Co=puter Group 32, Heat Balance (6) Core burnup and RCS boron concentration (f) Once the base line data was acquired, an imbalance was established using the group 8 control rods (APSRs). During group 8 movement, the integrated control system automatically compensated for the reactivity changes by repositioning group 6 to maintain constant power level.
5.4-1 k4
,. ..
(g) The imbalance previously established was observed for a minimum period of twenty minutes prior to obtaining the following data:
(1) Computer group 31, Fluid Conditions (2) Specified Operator Trend Group (3) Backup Incore Detector Recorder data (4) Computer group 20, Worst Case Thermal Conditions (5) Computer group 51. Normalized SPNP Map (6) Specified Operator Trend Group (second printout)
(7) Computer group 55, Imbalance / Tilt / Insertion (h) A new imbalancewas then established and the same data was recorded once again; this procedure was repeated until the msximum positive and negative imbalances had been established and the required data recorded.
The differences between the measurements at 40% and 76% FP were:
(a) The power imbalance limits observed were +20%/-25% at 40% FP and +14%/-23%
at 76% FP.
(b) The measurement at 40% FP used a gain factor of 3.20 set into the scaled difference amplifiers of the power ranste detector channels. The measurements at 76% FP used the gain factor determined during the 40% FP measurement to verify that the measured gain factor met the acceptance criteria.
As each imbalance condition was established, core power distribution and worst case thermal information was obtained from the plant computer to ensure safe conduct of the test. A } lot was maintained of incore offset versus out-of-core offset.
Based upon previous startup experience, it was determined that the relationship between the incore detector (ICD) and out-of-core (OCD) offsets was linear and of the form given in Equation 5.4-1.
OCO = M x K x ICO +B (Eauation 5.4-1)
Where: OCO = Out-of-Core Offset, % FP ICO = Incore Offser, % FP M = Slope of line B = Intercept at Zero ICO K = Gain Factor (difference amplifier)
The experimental slope could be determined from the plot of ICD versus OCD offset.
Once the experimental slope was known, the difference amplifier gain (K factor) required to meet the acceptance criteria was deter =ined from Equation 5.4-2.
K =
M2/My (Equation 5.4-2)
Where: K = Gain Tactor (difference amplifier)
M = Experimentally deter =ined slope 3
Mj=DesiredSlope
)h\
5.4-2
.. ..
5.4.3 TEST RESULTS The relationship between the ICD and OCD offsets was determined at 40% and 76%
FP by performing an imbalance scan with the APSRs. The measured results at 40% FP yielded an average slope of 0.734 for the ICD versus OCD offset reistion-ship. Initially, the minimum permissible slope was specified as 0.835, but was later increased to 0.920 to provide adequate margins to core imbalance limits.
Using the measured slope of 0.734 and an acceptance criteria (desired) slope of 0.920, the minimum K f actor was found to be 4.00. This value, however, would only give marginal assurance that the mintnun acceptance criteria would be met.
With an upper limit placed on the acceptable slope of 1.0 (K = 4.35) to avoid unnecessary restrictions on the OCD imbalance, it was decided to use a K factor of 4.033, which corresponds to a slope of 0.925.
Af ter adjustment of the difference amplifier K f actors to 4.033, the imbalance scan was performed at 76% FP to verify the results of the 40 FP measurement.
The average slope measured on the four out-of-core detectors was 0.991, corres-ponding to an " actual" difference amplifier gain of 4.072. This value compared well with the 40% FP results and the values established at 40% were accepted as providing the more conserva*.ive results.
A comparison of the incore detector (ICD) offset versus the out-of-core (OCD) detector offset obtained for each NI channel is shown in Table 5.4-1. The data taken at 40% FP was obtained with a difference a=plifier gain setting of 3.20 while the 76% FP data reflects the expertsentally determined K factors.
Core power distribution measurements were taken in conjunction with each of the imbalance measurements discussed above and the values of minimum DNBR and maximum linear heat rate were determined, extrapolated to the applicable overpower trip setpoint and compared to the acceptance criteria. The measured values of linear heat rate were multiplied by an uncertainty factor of 1.432 (discussed in section 5.9) to provide adequate conservatism in the comparison with the acceptance criteria. The worst case values of minimum DNBR and maximum linear heat rate determined at 40% and 76 FP are listed in Table 5.4-2, along with the extrapola-tions to the overpower trip setpoints for the next plateau in the power escalation sequence.
TABLE 5.4-2 Nominal Measured Measured LHR Extrapolations Power (%) DNBR x1.432(kw/ft) Power (%) DNBR LHR 40 8.48 7.90 95 3.6 18.8 76.4 4.56 13.56 105.5 3.6 18.73 The worst case DNB ratio was greater than the minimum limit of 1.55 and the maximum value of linear heat rate was less than the fuel melt limit of 19.6 kw/ft after extrapolation to 105.5% FP (the overpower trip setpoint) . These results show that Technical Specification li=its have been met and that adequate protection is pro-vided by the Reactor Prote: Hcn System trip setpoints for the allowed axial im-balances during power opera; :n.
Backup imbalance calculations using OP 1203/7 agreed with computer calculated tdaalances within the limits specified in the test. Table 5.4-3 lists the computer calculated imbalances as well as imbalances obtained using the incere detector backup recorders.
5.4-3 }4}47
.. .. TABLE 5.4-3 Nominal Computer Calculated Backup Recorder Power (%) Imbalance (%) Imbalance (%) 40 -17.40 -16.32 40 -13.66 -13.41 40 +0.56 +0.24 40 -13.6 -11.19 40 +3.67 +1.55 76 +11.86 +15.18 76 +8.64 +9.53 76 -19.11 -15.02 76 -0.66 40.17 5.
4.4 CONCLUSION
S Backup imbalance calculations performed in accordance with OP 1203/7 provide a reliable alternate method to computer calculated values of imbalance. Utilization of the Difference Amplifier Gain "K" factor as determined at 40% power, resulted in good agreement between Incore and Out-of-Core Detector Offset indica-tions. The two most important values verified as a result of this test were Minimum DNBR and Maximum Linear Heat Rate. Both of these parameters were well within Technical Specification limitations.
}h\4 5.4-4
.
.
~
MEASURED ICD AND OCD OFFSETS AT 40% AND 76% FP . Nominal TCD OCD Offset (%) Power (%) O f fset (%) NI-5 NI-6 NI-7 NI-8 40 -42.04 -33.01 -36.62 -33.92 -35.28 40 -32.63 -30.07 -33.59 -31.33 -32.08 40 -26.11 -28.77 -31.52 -28.37 -30.50 40 -15.32 -17.28 -21.34 -19.58 -21.05 40 -12.89 -18.20 -22.01 -19.48 -21.48 40 -0.67 -3.34 -7.13 -7.06 -7.25 g,j 40 1.70 -1.92 -5.47 -5.41 -5.46 Ls 40 8.85 6.70 3.60 2.60 2.63 N 76 -24.60 -26.57 -28.53 -24.40 -27.07 ui
- 76 -17.42 -20.69 -21.34 -18.17 -20.29 p 76 -8.15 -10.00 -10.58 -8.4 -9.71 H 76 -0.85 -2.26 -1.84 -1.06 -1.72 76 40.72 40.53 40.79 +1.45 40.53 76 +7.17 +5.55 +5.92 46.10 +5.42 76 +11.29 +6.87 46.84 +7.26 46.46 76 +15.20 +18.75 +17.62 +16.28 +18.04 where ICD OFFSET = POWUP-POWLW x100%
POWUP+POWLW OCD OFFSET = CHANNEL IMBALANCE x100% CHANNEL POWER
-
N s N
-
M Nb _
.. .. , 5.5 ROD WORTH AT POWER 5.5.1 PURPOSE The purpose of the Rod Worth at Power test vas to define a method for measuring differential control rod reactivity worths during equilibrium and transient power operation. This test was conducted in conjunction with Reactivity Co-efficient, Pseudo Rod Ejection and Dropped Rod measurements. 5.5.2 TEST METHOD The fast insert / withdrawal technique was used to measure differential rod worth under steady state and transient conditions. After placing the Diamond Control Station in hand, the selected control rods were withdrawn for six seconds and immediately inserted for six seconds. The Diamond Control Station was placed back in auto af ter waiting a minimi.m of 15 seconds following completion of rod motion. The reactimeter was used to calculate the change in core reactivity during the measurement. The differential worth was then obtained by dividing the known change in core reactivity by the change in rod position. The measured results were used to verify previously generated differential rod worth data. The following conditions were established in the reactor coolant system.(RCS) prior to recording test data: (a) RCS Average Temperature = 5790 F 11 F. (b) RCS Pressure = 2155 psig 125 psig. (c) Reactor Power constant within 11% for 20 minutes. (d) The RCS make-up tank was filled to ~90% and approximately two hours elapsed to allow system boron concentrations to equalize. The pressurizer was constantly sprayed for eight hours prior to the test in order te prevent any potential pressurizer out-surge from affecting test results. Pressurizer, make-up tank and RCS boron samples were obtained just prior to recording test data to verify equilibrium and to obtain a baseline boron concentration. 5.5.3 TEST RESULTS The results of differential rod worth measurements performed during power escala-tion testing are presented in Table 5.5-1. The core pcwer level, RCS boron con-centration and control rod positions that existed for each measurement are shown and the predicted results are included for comparison. Measured differential worths were well within the acceptance criteria limit of 120% of the predicted values. 5.
5.4 CONCLUSION
S Differential control rod reactivity worth measurements were performed as required during the power test progra=. Measured differential rod worths agreed with the design values well within the acceptance criteria limit of 120%. The maxi =um measured deviation was less than 4% from the design worth.
)k\k 5.5-1
.
.
~
MEASURED DIFFERENTIAL ROD WORTilS AT POWER
.
Power Baron Level Concentration Rod Group Position, Zwd Differential Rod Worth, 10-3%Ak/k/%wd (%FP) (ppmb) 1-5 6 7 8 Heasured Predicted 40 1192 100 72.2 0 33 9.44 9.77 40 1188, 100 65.5 0 13 11.39 NA II) 76 1120 100 85.5 15.0 20 15.03 15.3 s D y 100 1092 100 86.0 11.0 0 11.55 11.44
" .
u, 4 _______ (1) NA denotes that the predicted differential worth was not available. s N
-
N
-
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.. .. 5.6 REACTIVITY COEFFICIENTS AT POWER 5.6.1 PURPOSE Four coefficients of reactivity normally associated with lie nt water reactors are defined as follows: - (a) The Temperature coefficient of reactivity is defined as the fractional change in core net reactivity per unit change in fuel and moderator temperature. (b) The Moderator Temperature coefficient of reactivity is defined as the fraction-al change in core net reactivity per unit change in moderator temperature. (c) The power doppler coefficient cf reactivity is defined as the fractional change in core net reactivity per unit change in core power. (d) The doppler coefficient of reactivity is defined as the fractional change in core net reactivity per unit change in fuel temperature. The purpose of this test was to measure the temperature and the Power Doppler coefficients of reactivity at 40%, 76% and 100% of full power. The Moderator Temp-erature and the Doppler coefficients, which can be derived from the measured coefficients, were then calculated to verify that design limits were not exceeded. 5.6.2 TEST METHOD Reactivity coefficients were determined at 40%, 76% and 100% of full power by mea-suring the chance in core net reactivity caused by varying core average temperature and power level. The following prerequisite conditions were established at the start of each test to minimize reactivity effects not directly related to the coeffi-cient being measured: (a) Equilibrium xenon conditions were established. (b) Stable unit operating conditions were maintained - reactor power was held constant within +1% full power, reactor coolant temperature and pressure were maintained at 579 +2 F and 2155 +25 psig, respectively. (c) The soluble boron concentration differences between the Reactor Coolant, the Make-up Tank and the pressurizer were maintained in equilibrium. After the initial conditions listed above were established, the differential re-activity worth of the controlling rod group (s) were obtained using the Rod Worth at Power test to verify previously generated rod worth data. The temperature coefficient was measured by adjusting the Bailey Control Station to increase reactor coolant average temperature by SOF. The actual temperature change, rod movement, reactivity additions and any power changes were recorded on the plant computer, the reactimeter and the brush recorders. The individual reactivity effects were summed and then divided by the change in core average te=perature to obtain the te=perature coefficient. The measurement was repeated by decreasing core average temperature back to its initial value. 5.6-1
)h\4
.. .. The power doppler coefficient of reactivity was measured in a similar manner. After the prerequisite conditions were established, reactor power was decreased by 5% while the data was recorded. The reactivity effects of the power change were summed and then divided by the change in power to obtain the power doppler coefficient. The measurement was repeated by increasing core power by 5% to the initial level. The moderator temperature and doppler coefficients of reactivity were calculated from the measured tenperature and power doppler results. 5.6.3 TEST RESULTS The results of the reactivity coefficient measurements at power are sunnarized in Tables 5.6-1 and 5.6-2. Table 5.6-1 presents the measured temperature and power doppler coefficient results and provides the calculated results for comparison. The moderator temperature and doppler coefficients were calculated from the mea-sured results and are listed in Table 5.6-2. The measured coefficients are also plotted in Figures 5.6-1 and 5.6-2. The measured temperature coefficients are all within the acceptance criteria limit of 10.4x10-4ak/k/0F of the calculated values and are more negative. The moderator temperature coefficients calculated from the test results at 40%, 76% and 100% full power are all negative and trend more negative with increasing core power and burnup. Based upon the predicted results, the moderator coefficient will not be positive above 95% FP if the soluble poison concentration is less than 1155 ppm. Measured results from the Reactivity Depletion versus Burnup test (Section 5.11) show that the soluble poison concentration will not exceed 1155 ppm under equilib-rium xenon, beginning of life, normal control rod configuration and 100% full power conditions. These results show that the moderator temperature coefficient will not be positive during power operation at or above 95% FP. Comparison of the measured power doppler coefficient with the calculated values shows that the predicted reactivity deficit versus power is slightly less than pre-dicted. The estimated reactivity deficit from 0% to 100% full power based upon the measured results is 0.90%Ak/k as compared to the calculated value of 1.32:ak/k. The acceptance criteria limit that the power doppler coefficient be more negative than -0.55x10-4ak/k/: FP was met. 5.
6.4 CONCLUSION
S The measured results indiaste that the moderator temperature coefficient will be negative during power operation above 95% FP. The results of the power doppler coefficient measurements indicate that the least negative value of the coefficient is -0.00710%Ak/k/% FP at full power conditions.
,
. -
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.
.
MEASURED COEFFICIENTS OF REACTIVITY AT POWER
.
A. Temperature Coefficient of Reactivity
- Core Boron Avg. Differential Rod Group Position. 7. Ud Temperature Coefficient, Power Concentration Rod Worth X10-4 Ak/k/0F (7. PP) (ppmb) (% Ak/k/% Vd) 1-5 6 7 8 Measured Calculated 40 1192 0.00940 100 75 0 33 -0.135 -0.020 76 1120 0.01503 100 85 10 20 -0.251 -0.158 100 1090 0.01155 100 90 20 0 -0.329 -0.246 B. Power Doppler Coefficient of Reactivity Core Boron Avg. Differential Rod Group Position, 7.Wd Powe Power Concentration Rod Worth X10gDopplerCoefficient, Ak/k/% FP i (7. FP) (ppmb) (I Ak/k/% Vd) 1-5 6 7 8 Measured calculated
'D N 40 1192 0.00940 100 75 0 -0.890 33 -1.37 y 76 1120 0.01503 200 85 10 20 -0.849 -1.28
- 100 1090 0.01155 100 90 20 0 -0.734 -1.16
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SUMMARY
OF REACTIVITY COEFFICIENTS AT POWER , Core Boron Coefficient of Reactivity. X10 4 Ak/k Power Concentration (7. FP) (ppmb) Temperature Power Doppler Moderator Doppler 40 1192 -0.135 -0.890 -0.008 -0.127 76 1120 -0.251 -0.849 -0.144 -0.107 100 1090 -0.329 -0.734 -0.222 -0.107 a b M
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i . .. POWER DOPPLER COEFFICIENT OF PE. ACTIVITY VERSUS POWER LEVEL
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.. s . 5.7 DROPPED CONTROL ROD TEST 5.7.1 PURPOSE The purposes of the Dropped Control Rod Test conducted at 40% and 76% of full Power were as follows: (a) To verify the safety analysis relating to the accidental dropping of a control rod which is normally withdrawn from the core at full power. The measurement was performed at 40% FP. (b) To demonstrate the capability of the rod dri re control system to detect a control rod that deviates from its group average position by preset limits, to verify that reactor power is automatically reduced to a specified power level and that aut*matic control rod withdrawal is inhibited above a specified power level when a cont;rol rod asymmetric fault exists; this test was performed at 76% FP. (c) To verify the adequacy and accuracy of backup ouadrant tilt calculations using OP 1203/7, " Power Imbalance and Quadrant Power Tilt Calculations Using the Backup Incore Detector System", at 40% FP. (d) To insure reactor stability following an induced xenon occillation at 40% FP.
- 5. 7. 2 TEST METHOD The results of core thermal calculations show that control rod 7 in group 6 (core locatior. E-4), and those rods symmetric to it, would produce the most adverse thermal effects.if it.were dropped into the core during. operation at power. The pseudo dropped control rod test.was. conducted at 40% FP by measuring the worth of control rod 6-7 as it was inserted in the core and by measuring the resultant core power distributions with the dropped rod at 0% and 50% withdrawn.
The worth of control rod 6-7 was determined by performing a rod swap with control rod group 7. Differential reactivity worths were measured on group 7 using the fast insert / withdraw mechod at 20% insertion intervals as rod 6-7 was inserted into the core. The measured differential worth data for group 7 was then integrated to obtain the worth of rod 6-7. Core power distribution and worst case thermal con-ditions data was obtained from the plant computer when rod 6-7 reached 0% and 50% with-drawn. Data from the backup incore recorders was taken concurrent with the computer data and was used to verify the tiJt calculations of OP 1203/7. Control rod 6-7 was realigned with its group after completing the reactivity worth and power distribution measurements. Quadrant power tilt was monitored using the incore detector system for 24 hours following the test to verify that the induced xenon oscillation decayed away. Measurements at 76% FP were performed to demonstrate the response of the rod drive control system to an asymmetric rod. Control rod 6-7 was selected for individual control and inserted past the 7 in. and 9 in. deviation limits. Control rod maps were obtained from the plant computer when the asymmetric rod alarm and fault con-ditions occurred. The Diamond Control Station was then put back into automatic and the subsequent power runback and rod withdrawa3 inhibit were verified. 5.7-1 )k\
.. .. 5.7.3 TEST RESULTS The results of the pseudo dropped control rod test conducted at 40% and 76% full power are su=marized in Figures 5.7-1 through 5.7-4 and Tabics 5.7-1 and 5.7-2. Figure 5.7-1 shows the location of control rod 6--7 and its position relative to each core quadrant. During the insertion of rod 6-7 from its group average position to 0% withdrawn, control rod group 7 was withdrawn a total of 4 to compensate for the reactivity effects of the dropped rod. The initial and final positions of group 7 were 15% and 17% withdrawn, re:spectively, which was due to a periodic adjustment of the goup 6 and 7 positions to maintain proper overlap between the groups. The results of the differential rod worth measurements on group 7 yield a total dropped control rod worth of 0.094%Ak/k. Result: of the core power distribution measurements with control rod 6-7 in-serted in the core are summarized in Figure 5.7-2. These results show toe radial core power distribution with rod 6-7 at 85%, 50% and 0% withdrawn. A large de-pression of the flux in the quadrant containing the dropped rod was observed, as expected. The mavimum measured radial peaking factor was 1.708, in the core quadrant opposite to the dropped rod. The maximum quadrant power tilts measured were 14.87% and 14.36% in the XY and WX quadrants, respectively, with rod 6-7 fully inserted. Figure 5.7-3 is a plot of quadrant power tilt versus rod 6-7 position. The quadrant power tilt was subsequently monitored after aligning rod 6-7 with the group 6 average position. The quadrant tilt returned to less than 4% within 24 hours of realignment of the dropped rod which demonstrates the core stability to an induced radial xenon oscillation. Figure 5.7-4 shows the quadrant power tilt versus time after the test. Worst case thermal conditions were determined during the power distribution measure-ments taken for the dropped rod test and are summarized in Table 5.7-1. The values of maximum linear heat rate and minimum DNBR were measured with rod 6-7 at 85%, 50% and 0% withdrawn. The minimum value of DNBR was 9.50 and extrapolation of this value to 100% FP resulted in a DNBR of 3.8, which is well above the acceptance criteria limit of 1.55. The measured value of maximum linear heat rate was 5.02 kw/f t which, when multiplied by the 1.432 uncertainty factor, becomes 7.19 kw/f t. This value extrapolated to 100% full power gives a maximum linear heat rate of 18.44 kw/f t which is less than the fuel melt limit of 19.6 ka/f t. The results o, chis measurement show that even when the worst case uncertainties are assumed, sufficient margin exists to the core thermal limits with a control rod dropped in the core at 100% FP and a resultant. quadrant power tilt of 14.87%. Further, this evaluation doer not incorporate the reduction in total core power that would occur due to the reactivity inserted by the dropped rod. The results of the quadrant power tilt calculation performed with data from the backup incore recorders are su=marized in Table 5.7-2. The quadrant power tilts determined from the backup recorders were in good agreement with the tilts determined from the incores and within the acceptance criteria limit of 15 tilt. The asymmetric rod alarm and fault conditions occurred with rod 6-7 at 4.3% and 5.7% below its group average position. These results compare well with the acceptance criteria limits of 5 12% and 6.4 12%, respectively. Reactor power was automatically reduced from 76% to 56% FP at a runback rate of 32.3% FP/ min when ths asy= metric rod was detected by the rod drive system. 5.7-2
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.. ..
.
Automatic control rod withdrawal was inhibited above 60 FP following the runback, as required. 5.
7.4 CONCLUSION
S The dropped control rod test performed at 40% and 76 FP met all of the required acceptance criteria. The following conclusions were drawn as a result of the measurements. (a) The core power distributions and thermal conditions that developed from a control rod which was dropped into the core during operation at power showed adequate margins to minimum DNBR and maximum LHR limits. The measured worth of the dropped rod was 0.094%ak/k. (b) Quadrant power tilt calculations performed using backup incore recorder data were accurate in comparison to the computer calculated values. (c) The rod drive control system responded properly to an asymmetric rod condition.
.
\ q\ 4 \ 5D' 5.7-3
.
.
- CORE TilERHAL CONDITIONS HEASURED DURING illE ,
DROPPED CONTROL ROD TEST AT 40% MILL POWER Core CR 11-4 Hax. Extrap. to 100% FP Power Position Min. IllR LilR (%FP) (% wd) DNBR (kw/ft) DNBR (kw/ft) 39 86 11.14 5.64 4.34 14.46 39 49 10.21 6.04 4.01 15.49 39 0 9.50 7.19 3.8 18.44 n
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COMPARISON OF BACKUP RECORDER AND INCORE . DETECTOR TILT CALCULATIONS AT 40% FP CR 6'-4 Positi,n Data (% wd', Ou.drant Power Tilt (%) Source XY YZ ZW WX 86 Recorder 1.58 -1.58 1.06 -1.06 86 Incores 0.26 0.25 0.71 0.20 Dif ference (Recorder-Incores) 1.32 -1.83 0.35 -1.26 f t3 '" 49 Recorder 5.37 3.76 -1.08 -8.00 49 Incores 5.69 5.65 -6.01 -5.33 . { Difference (Recorder-Incores) -0.32 -1.89 4.93 -2.73 0 Recorder 14.87 12.30 -12.30 -14.36 0 Incorea 13.85 13.98 -14.07
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FOR MEASURDENT AT 40% FP A B C D E F - C - _ _ -
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)h\h FIGURE 5.7-1
'* ** HCASURED RADIAL CORE POUER DISTREUTICNS WIT!! DROPPED CONTROL ROD AT 85'4, 507. AND 0*4 WITHDRAWN FOR EQUILERIUM XENON, 40~4 FP CONDITIONS
. X 3.477 D.791 0.940 0.791 0.477 A 3.600 0.818 0.890 0.851 C.661 3.717 0.850 0.7710.845 l0.go? 0.479 0.7220.773 1.001 1.394 1.001 0.7733.722 o,479 _ B 43' O.734 0.790 1.031 1.440 1.066 0.8543.816 0.516 00.527 0.768 0.826 1.053 1.479 1.089 0.882 1.91 2 0.512 0.542 0.841 0.9070.982 1.069 1.2861.069 1.007 0,9n7 0.841 0.542 C P.488 0.873 0.8840.978 1.068 ino 1.057 0.982 0.976 0.572 1.2561.I51 1.111 1.010 1.034 0.612 0.5680.903 0.855 0.973 1.078 1.214I. 0.479 0.341 0.919 1.097 1.002 1.239 1.1981.239 1.0021.097 0.919 0.841 0.479 D 0.438 0.866 0.876 1.0460.948 1.219 1.218 1.268 1.039 1.167 1.008 0.984 0.510 0.484 0.865 0.807 0.9930.893 1.209 1.2551.316 1.091 1.249 1.071 1.072 0.551
- g. g.Pg {.fgg 1.064 1.281 1.203 1.3361.198 1.2811.064 1.106 .857 .722 E . .
o 970 1.201 1.183 1.2301.225 1.3111.118 1.174 .964 *833 0.673 0.697 0.902 0.857 L.123 1.172 1.1201.271 1.3801.200 1.280 .043 .902 0.477 0.781 1.005 1.034 1.269 1.209 1.338 1.2611.338 1.209 1.269 1.034 1.001 3.731 0.477 F - 0.753 0.879 0.894 1.139 1.138 1.303 1.2641.371 1.266 1.314 1.094 1.075 3.912 0.662 0 5 0.612 0.7120.707 0.990 1.064 1.274 l' 2 80 1.434 1.3671.401 1.175 1.165 3.969 0.* b4 . 0.769 0.791 bg b'*$J { g Q' 1.235 1.325 1.189 1.2251.189 1.3251.214 1.230 1.056 1.044 0.791 G - 0.765 1.071 1.219 1.140 1.2191.216 1.3631.273 1.296 1.133 1.087 0.878 0.707 0.676 0.6980.867 0.889 1.108 1.087 1.2191.264 1.4441.360 1.397 1.227 1.117 0.953 09 394 1.286 1.196 1.336 1.261 1.225 1.0821.225 1.2611.336 1.198 1.286 L.394 0 940 WH - 0 *.8b 1*324 1 1.1311.099 1.130 1.192 1.162 1.1101.263 1.347 L.343 1.347 L.382 L.538 0.934 Y 0.606 1.232 0.9330.983 0.88711.111 1.130 1.1411.315 1.4621.368 1.541 L.495 L.708 0.887 E791 1.044 1.056 1.230 1.224 1.325 1.189 1.2251.189 1.325 1.227 1.230 L.056 .044 0.791 K - 0.740 0.870 0.885 1.054 1.063 1.216 1.142 1*2261.225 1.374 1.286 1.304 L.136 .082 0.862 0.686 0.636 0.6830.857 0.884 1.107 1.091 3 m i.274 1.455 1.374.1.405 L.230 .111 0.935 3.47 7 0. 7 810.9 70 al.034 41.269 0. 209 (l.33811.261'1.3381.20981.269 0. 034 L.00010.7811 U #' " 6 - 3.54110.715 3.581 0.581 g 0 871tl.1251.133). 307 !1.2751.3871.284 - 0.9801.062 a .281 1.293 1.452 1.386E*f30 1.104).073 0.892 10.6181 i8 1.185 L.162 0.048 0.723l 1 1.1vb 1.064 1.281 1.207
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0.955 1.196 L.307 1.243 1.241 1.327 1.129 1.172 .942 0.781 0*639 0*6 0 $**N2 0.847 1.123 L.181 1.135 1.289 1.3981.211 1.278 )i .nonD.848 0.479 3.841 0.919 1.097 1.002 L.239 1.198 1.239 1.0021.097 1.919 )*841 0 479 N 0.410 3.841 3.858 1.036 D.947 L.188 1.228 1.277 1.0431.156 1.975 3.917 3.398 0.464 0.847 0.794 0.988 3.895 1.217 1.268 1.327 1.0961.239 1.038 1.005 3.437 3.542 0.541 0.907 3.983 1.069 1.286 1.069 0.980).907 3.841 .542 0 3.493 3.876 0.8893.984 L.225 1.257 1.099 1.030 L.156 3.884 .429 3.576 3.909 0.862 3.980 L.084 1.215 1.143 1.085 ).958 2.043 .469 3.479 0.722 J.173 L.001 1.394 1.011 0.773 0.722 3 '+ 7 9 P 3.475 0.764 3.807 L.033 1.423 1.023 0.777 1.156 3.333 3.566 o,794 3.840 L.053 1.460 1.045 0.3040.689 3.330 J " J.791 0.940 0.791 0.477 R l 3.807 0.817 0.847 1.099 0.590 3.741 0.839 0.721 0.748 0.537 l l l l l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 .15 2 - x.xx _ Rod 6-7 at 85*.' wd x.xx - Rod 6-7 at 50% vd x.xx _. Rod 6-7 at 0% wd
)h\h FIGURE 5.7-2
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QUADRANT POWER TILT DURING DROPPED CONTROL ROD TEST AT 40% 7P
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.. .- 5.8 PSEUDO CONTROL ROD EJECTION TEST 5.8.1 PURPOSE The purpose of the Pseudo Control Rod Ejection Test was to verify the safety analysis relating to the accidental ejection of a control rod normally inserted in the core during power operation. This was accomplished by measuring the rod worth and associated core power distribution as the pseudo ejected con-trol rod was withdrawn from the core. The ability of the rod drive control system to detect a control rod that deviates from its group average position was also determined during this test. 5.8.2 TEST METHOD The test was conducted at 40% FP with equilibrium xenon established in the core. Control rod groups 1 through 5 were positioned at 100% withdrawn, group 6 at 75%, group 7 at 0% and group 8 at 13% to maintain core axial imbalance at 0 +2%. All _ Integrated Control System Stations were in automatic,except the Diamond Control Station which was in the manual sequence mode of operation. Control rod 7-1 (core location H-8) was calculated to be the most reactive control rod in a rod ejection accident and this rod was selected for the measurement. Control rod 7-1 was withdrawn to 100% (in small increments) by performing a rod swap with control rod group 6. Differential reactivity worth measurements using the fast insert / withdrawal technique were performed on group 6 with control rod 7-1 positioned at 0% withdrawn and again at every 20% withdrawal interval. The total reactivity worth of rod 7-1 from 0% to 100% withdrawn was then obtained by integrating the differential worth data measured on group 6. Core power distribution measurements were taken with control rod 7-1 positioned at 100% withdrawn. The asymmetric rod alarm and fault verification was conducted by obtaining a control rod map from the plant computer when the asymmetric rod alarm and fault conditions occurred as rod 7-1 was withdrawn from the core. 5.8.3 TEST RESULTS The Pseudo Control Rod Eiection test was performed at 40% full power using control rod 7-1 (core location H-8) as the ejected rod. Figure 5.8-1 shows the location of control rod 7-1 in the core. A comparison of radial core power distributions with control rod 7-1 at 0% and 100% withdrawn is given in Figure 5.8-2. Control rod group 6 was inserted from 74.2% to 54.5% withdrawn to compensate for the reactivity effects of removing rod 7-1 from the core. An integration of the measured differential worth data on group 6 gives 0.229 Ak/k as the reactivity worth of the ejected rod. A reactor power level of 40.59% full power was established at the start of the measurement and was maintained constant using the excore instrumen-tation. However, the total effect of withdrawing control rod 7-1 from the core was partially masked by the other group 7 rods and actual core power increased to 46.08% of full power with the ejected rod fully withdrawn. Using the measured value for the Power Coefficient of -8.9x10-5Ak/k/% FP, a correctica factor of 0.0489%Ak/k was added to the integral worth result to give 0.278%Ak/k as the total ejected rod worth. This result is well below the acceptance criteria value of 0.49%Ak/k.
~1 5.8-1 \ i
)k)h
,. .- The effects of the pseudo ejected rod on core power distribution are sarized in Figure 5.8-2 and Tables 5.8-1 and 5.8-2. Core power distribution and thermal conditions were measured with control rod 7-1 at 0% and 100% withdrawn. The measured values of minimum DNBR and mrinum mR were 4.85 and 13.12 kw/ft, re-spectively, with rod 7-1 fully withdrawn. The measured value of LHR includes the uncertainty factor of 1.432 discussed in section 5.9.3.2. The maximum mea-sured radial power peak was 2.38 in core location H-8. The affect of the ejected rod on quadrant power tilt was small,since the rod was located in the center of the core. Table 5.8-2 presents the results of a hand calculation of quadrant tilt which show a marimum positive tilt of +0.73% in the YZ quadrant. Core axial power imbalance shif ted to the bottom of the core during the measurement dua to group 6 insertion from 74.2% to 54.5% withdrawn. The control rod asymmetric alarm and aspnmetric fault conditions were verified during the withdrawal of rod 7-1. The control rod asymmetric alarm should occur when rod 7-1 is 5% +1% above the group 7 average position and the asymmetric fault condition should occur when rod 7-1 is 6.5% +1% above the group average. Control rod group 7 was at 2.66% withdrawn when the asymmetric alarm occurred with 7-1 at 8.0% withdrawn. Control rod group 7 was at 2.88% withdrawn when the asymmetric fault occurred with rod 7-1 at 10.0% withdrawn. 5.
8.4 CONCLUSION
S The measured worth of the pseudo ejected control rod was 0.278%Ak/k, which is well below the 0.65%Ak/k limit set in Technical Specifications 3.5.2. The measured values of maximum LHR and minimum DNBR were 13.12 kw/ft and 4.85, respectively, with the ejected rod at 100% withdrawn. The marimum measured radial power peak was 2.38 in the fuel assembly containing the ejected control rod. The asymmetric control rod condition was detected by the rod drive control system within the acceptance criteria limits.
)h\k
.
5.8-2
.
.
.
.
CORE POWER DISTRIBUTION AND TilERMAL HYDRAULIC DATA DURING Tile PSHUDO ROD EJECTION TEST ROD INCORE QUADRANT TILT MAXIMUM M/JIMUM MINIMUM FUEL POSITION IMBALANCE RADIAL LIIR (1) DNBR ASSEMBLY (% Wd) (% Wd) WI XY WZ YZ PEAK (kg/ft) LOCATION 00 -1,90 -0.28 40.26 -0.21 +0.22 1.40 6.95 9.28 K-10 a D [,i 100 -8.37 -0.28 +0.01 -0.15 +0.42 2.38 13.12 4.84 11 - 0 8 1" Y e
. _ _
(1) The values of LilR include the 1.432 uncertainty factor. - N
-
N _- O
e . .
- RAND CALCU:ATI0ft of QUArJULMT POWER TILT from CORRECTED INCDRE DETTCTOR !/.ADINGS Cc.tcol Rod Group Positions Core Fever 40: FF xenon Conditicas Cps 1-4 E 2 vd Cp 6 M vd Baron cone. ppa o tierthWk/k Cp 5 _1D0.1 #d CP 7 M Z vd Core Burnup ETFD Equil. Conc.Y,gg, Cp 8 M 2 vd (yes or no)
Deto 7/5 /74. Time _122Q_, , Initials of Analyst _JAF QCADRANT W QUADRANT IT QUADRANT C QUADRANT Tz Core Lcc.l Level CoreLoc} Level Core 14c) Level Core Loc.l Level C-6 1 211 C-10 1 213 t-5 1 1AA K-11 1 97A 2 235 2 243 2 511 2 co' 3 242 3 248 3 mzo 3 mme 4 283 4 286 4 566 ' .-54.h 3 215 5 220 5 tot 5 toA 6 119 6 6 6 are iro 7 -rg 7 48 7 53 7 114 7 - 11 ^ _ E-7' 1 381 . E-s 1 17; L-3 1 e_ n e L-13 1 ?nA 2 91 4 2 Mi o 2 933 2 ?An
---
3 549 3 996 3 9 A r. 3 acn 4 550 4 mao 4 201 4 in 9_ 5 1AQ 5 ist 5 ?ni 5 ase 6 239 6 997 6 11A 6 stA 7 09 7 119 7 A7 7 44 F-3 1 706 F-13 1 ?nA M-7 1 7A/ M-9 1 109 2 240 2 296 2 991 2 527 3 259 3 796 3 <^^ 5 MA' 4 293 4 286 4 917 4 554 5 221 5 797 5 1Ao 5 109 6 140 g 147 6 260 6 257 7 55 7 Mo 7 117 7 101 C-5 1 372 C-11 1 171 0-6 1 910 0-lo 1 21 1 2 526 2 517 2 2.s,5 2 244 3 553 3 538 3 790 3 247 4 542 . 4 551 4 283 4 283 5 10 1 5 1A7 5 9no 5 207 6 272 g 261 6 137_ s 126 7 113 7 100 7. .,14 1 7 54 a MAL - 8274 x7 mn - 8265 e mn - 8252 Tz mn - 8344
+ + + T2 m quad M RACE = -
8_283.75 tnt n -
*^[-1 .
x100 .12: tnt u -
.-j]f x loo - .38%
TILT n - ."",*[-1 x100- .23% T:t? Tz - j]^[-1 : loo - A.73: TABl.E 5.8-2
.. .. FJECTED RCD UCRTH AliD LOCATIO:: AT LOS FP PLATEAU
^ _
3' C - D E - F _ G _ a
- ~
E _ _ _ - K - L _ M N O _ P_ . e
"
-
l l l 1I I I 1 2 3 4 5 6 7 8 9 10 11 12 13 1k 15 Technical Calculated Measured Specification CR Group Core CR Worth CR Worth Limit @l00% FP No. Locatien 5 Ak/k 5 ok/L % Ak/k 7 E-8 0.h9 0.278 0.65
.
1414 'A' rauR: 3.a-1
.. .. COMPARISON OF CORE POWER DISTRIBUTIONS WITH EJECTED ROD AT 0% AND 100% WITHDRAV.i, UNDER EQUILIBRIUM XENON, 40% FP CONDITIONS CORE CONDITIONS H-8 at 0% H-8 at 100% H-8 at 0% H-8 at 100% GPS 1-4 at 100 Nd 100 %Wd CORE POWER LEVEL 40 %FP 40 %FP 5 at 100 %Wd 100 _ %Wd BORON CONC. 1188 ppm 1188 ppm 6 at 75 %Wd 54.5 %Wd CORE BURNUP 3.75 EFPD 3.75 ~ ETPD 7 at 2 %Wd 2 %Wd AXIAL IMBALANCE -1.90 %FP -8.57 %FP 8 at 13 %Wd 13.0 %Wd MAX. QUADRANT TILT.+ .28 % + .42 : Date 7/5/7h , Tite 1220 , Initials of Analyst DAL _ 1.106 1.300 1.339 1.383 1.176 1.273 1.375 942 - 2.2h 1.982 1.627 1.458 1.093 1.166 1.253 .850
' 225 1.041 1.028 .7S0 271 1.404 1.265 .
.932 .704
- 1. 1 1.647 1.317 1.167 .959 68 1.282 .983 .971 .721 .456
- 1. 1.277 .915 .889 .663 .413 1.089 .853 .696
- 1. 5
.987 .767 .632
.9 920 .840 .476
. '6 .757 .429
.545 I
. 90 :
l
.
E X.XX Ejected Rod at 0% Nd X.XX Ejected Red at 100% Wd 1414 142 FIGUFI 5.8-2
.. .. 5.9 C,0RE POWER DISTRIBUTIONS 5.9.1 PURPOSE Detailed core power distribution measurements were perfomed under steady state conditions during the power escalation program to verify that the core axial imbalance, quadrant power tilt, maximum linear Leat rate and minimum DNBR do not exceed their specified limits. A su= mary of the steady-state core power distri-butions presented in this saction is given ic Table 5.9-1. The specific acceptance criteria applied to the measured core power distributions are listed below: (a) The combination of reactor thermal power and reactor power imbalance (power fraction in top half of the core minus power fraction in bottom half of the core) shall not exceed the safety limit as defined by the locus of points for the specified flow established in Technical Specification Figure 2.1-2 (Figure 5.9-1). (b) The quadrant power tilt, as defined below, does not exceed the limits specified in Technical Specification 3.5 2.4. Quadrant Power Tilt.= Power in any Core Ouadrant
-1 x100 Average Quadrant Power (c) The minimum value of DNBR shall be greatet than 1.55.
(d) The maximum linear heat rate shall not exceed the curve of Figure 5.9-2 (LOCA limit) for all combinations of reactor themal power and power imbalance that lie within the area bounded by the curve of Figure 5.9-3. For any combinations of reactor power and power imbalance that lie outside of the curve of Figure 5.9-3, the maximum linear heat rate shall not exceed 19.60 kw/ft (fuel melt limit). 5.9.2 TEST METHOD Core power distribution measurements were performed at the major power plateaus of the test program (15%, 40%, 76: and 100% full power) under steady state conditions for specified contrul rod configurations. To provide the best comparison between measured and predicted results, three-dimensional equilibrium xenon conditions were established for the measurements at 40, 76 and 100% full power. The first step to establish three-dimensional xenon equilibrium was to borate /deborate the reactor coolant system,as required,to maintain the contro111ne rod groups within plus or minus two percent of the desired position during and af ter escalation to the next power plateau. The axial power shaping rods were moved,as required,to establish the predicted offset on the core. Once the measurement power level was achieved, the axial power shaping rods were maintained in a constant position for six to eight hours prior to taking data. 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 themal conditions calculated by the plant computer. The calculated core power distributions for various core power levels, axial imbalances, control rod configura-tions, burnup, boron concentrations and xenon conditions were provided for comparison with test results from the three-d1=ensional PDQ-7 code with themal feedback. 5.9-1 A' 1;43
}k)k
.. .. 5.9.3 TEST RESULTS 5.9.3.1 Steadv State Power Distributions Steady-state equilibrium xenon core power distribution measurements were perfor=ed at the major test plateaus of the power escalation sequence for specific control rod patterns, boren concentrations, axial imbalances and core burnups. The measured results are tabulated in Tables 5.9-2 through 5.9-5. The tables present a complete 1/8 core power distribution map using the corrected in: ore detector outputs from 7 levels of 29 fuel assemblies which describe the entire core, assuming eighth core symmetry. A summary of the four cases studied in this report is given in Table 5.9-1 which gives the core power level, core burnap, control rod pattern, boron concentration, xenon conditions, axial imbalance, maximum quadrant tilt, minimum DNBR, maximum LHR and power peaking data for each measurement. Core power distribution calculations at steady-state eau 111brium xenon conditions were performed using the three-dimensional PDQ-07 code with thermal feedback. The four cases reported in this section are compared with the PDQ-07 results in Figure 5.9-4 through 5.9-7 to demonstrate the degree of agreement between the measured and calculated radial core power distributions. As can be seen from these figures, the comparison between measured and calculated distributions shows good agreement. Differences between measured and calculated results are attributed to differences in core conditions that were assu=ed for the calculations as compared to those rhat existed at the time of measurement. The results of measured core power distributions show maximum local peaking factors between1.80 and 1.87 for the four cases with the maximum value of 1.87 measured at 15% FP and the minimum value of 1.80 measured at 100: TP. In all cases, the maximum local peaking factor was below the 2.67 limit given in Technical Specification '.1. Calculations were performed to determine the maximum linear heat rate using the maximum local peaking factor measured in the four cases. The values of maximum linear heat rate were extrapolated to the overpower trip setpoint (105.5% FP) and the central fuel melt limit (112: FP). Examination of Table 5.9-6 shows that all extrapolated values of maximum linear heat rate were below the Technical Specification values of 17.1 kw/ft for the LOCA limit and 19.6 kw/ft for the fuel melt limit. Similarly, the extrapolated values of DNBR were well above the Technical Specifications minimum value of 1.30. 5.9.3.2 Minimum DNBR and Maxi =um IRR Calculations 5.9.3.2.1 Minimum DNBR Determination Minimum values of DNBR were calculated fer the core power distributions taken during the test program. The results of the DNER calculations, are plotted in Figure 5.9-8. The results show that the minimum DNER is greater than the acceptance criteria value of 1.55. The following analysis was used to deter =ine the DNBR whenever steady state core power distributions were obtained. (a) From each core power distribution, the fuel assembly which yielded the worst-case DNBR was selected. (b) Upon selection of the worst-case asse=bly, a radial peaking factor was calculated and seg=ent power levels were converted into axial peaking factors.
)h
.. .. (c) The radial peaking factors were adjusted by a factor of 1.05 (the local pin peaking multiplier) . (d) The radial and axial flux data from (b) and (c) were incorporated into the standard hot channel analysis and analyzed for operation at full power. All cases studied showed that the measured minimum DNBR values are greater than the minimum test acceptance criteria limit of 1.55. A minimum DNBR margin of 109 percent was observed after extrapolation to 112% FP. 5.9.3.2.2 W h Linear Heat Rate Determination Analysis for determining maximum linear heat rate was performed in conjunction with minimum DNBR analysis. After selection of the worst case asse=blies and de-termination of the radial and axial peaking factors, the maximum linear heat rate for each of the measured core power distributions was determined by Equation 5.9-1. NLHR = P R XPA XP I Q Rate X FNT (Equation 5.9-1) NA X NP X AL where: MLHR = Maximum Linear Heat Rate (kw/f t) PR = Radial Peaking Factor PA = Axial Peaking Factor Pt = Local Pin Peaking Multiplier (1.05) 3 QR ate = Rated Core thermal power (2535x10 KW) FNT = Fraction of power generated in fuel (0.973) NA = Number of fuel assemblies in core (177) NP = Number of fuel pins in each assembly (208) AL = Active length of fuel pin (12 ft) The results of the maximum linear heat rate calculations performed for the power distributions presented in this section are summarized in Table 5.9-6. The results of these worst case values were extrapolated to the overpower trip setpoint (105.5% FP) and fuel melt (112 FP) limits for each measurement. Substantial linear heat rate margins were observed with a minimum margin of 42% after extrapolation to 112% FP. In addition to the above analysis, an uncertainty factor was applied to the cal-culated values of linear heat rate during the startup test program to introduce additional conservatism in the comparison with acceptance criteria. The worst case values of linear heat rate determined during any core power distribution measurement were multiplied by 1.432 to account for uncertainties not incorporated in Equation 5.9-1. These uncertainties include: (a) nuclear uncertainty (b) power uncertainty (c) fuel densification effect ig} (d) power spike effect (e) quadraut power tilt effect }kkk i In each case where uncertainties were considered for incorporation into the 1.432 factor, worst case assumptions were used to provide adequate conservatism. For exa=ple, the quadrant power tilt f actor makes allowances for a quadrant tilt of 4% even though the measured values for tilt were much less than 4% during the tes: 5.9-3
.. .. program. The modified values of linear heat rate were then extrapolated and co= pared to the LOCA or fuel melt limit as specified in section 5.9.1 (d) . Table 5.9-6 includes the modified values of maximum linear heat rate determined during the four cases studied in this section. As can be seen from the table, the LOCA and fuel melt limits were not exceeded even virh the worst case assumptions used. 5.9.3.3 Ouadrant Power Tilt and Axial Power Imbalance Table 5.9-1 presents the maximum observed quadrant power tilts measured by the incore detector system during the core power distribution measurements. Quadrant power tilt limits are established by the Technical Specifications in coniunction with control rod position limits to assure that the design peak heat rate criterion is not exceeded during normal power operation. The quadrant power tilt measured during operation at power has been well within the limits established in the Technical Sp6cifications. The core axial power imbalances measured in conjunction with the core power dis-tribution of this section show that a substantial margin exists between the mea-sured reactor power / power imbalance combinations and the limits set forth in Technical Specifications Figure 2.1-2 (Figure 5.9-1). The core power imbalance measured by this incore detector system for each power distribution studied is summarized in Table 5.9-1. 5.
9.4 CONCLUSION
S Core power distribution measurements were conducted at 15%, 40%, 76% and 100% of full power during the power escalation sequence under steady state equilibrium xenon conditions for specified control rod configurations. Co=parison of the measured power distributions with the PDQ-07 results shows good agreement. For the three cases studied at 40%, 76% and 100% full power, the three largest measured and calculated radial peaks were chosen. In each case, the measured values were within 8% of the calculated results. The results of the minimum DNBR and maximu= LHR analyses are given in Table 5.9-6. The margins to the minimum DNBR limit of 1.55 and maximum LHR value of 17.1 kw/ft were 109% and 42%, respectively, after extrapolation to 112% FP. All quadrant power tilts and exial core imbalances measured during the power distribution tests were within the Technical Specifications and nor=al operational limits. 5.9-4
)h\k
.. .. Measured Core Power Distributions and Core Thermal Conditions for Various Control Rod Patterns and Core Power Levels of 15%, 40%, 76% and 100% FP. 15% 40% 76% 100% Date 6/20/74 7/01/74 7/21/74 8/09/74 Time 0713 1010 0844 0103 Power level, % FP 16.4 40.47 77 99 Xenon Equilibrium 2-D 3-D 3-D 3-D Rod Positions, % vd 1-5 100 100 100 100 6 76 75 77 93 7 0 0 5 19 8 34 33 27 14.5 Core Burnup, EFPD 0.45 1.8 9.87 21.2 Boron Concentration, ppmb 1340 1207 1090 1101 Axial Imbalance, % FP -2.0 -5.1 -12.43 -2.3 Maximum Quadrant Tilt, % 0.5 0.41 1.18 1.39 Minimum DNBR 25.82 10.28 5.42 4.01 Maximum IHR, kw/ft 1.86 4.49 8.33 10.06 Maximum Peaks Radial 1.44 1.40 1.39 1.37 Radial x Axial 1.87 1.80 1.81 1.83
)h\h TABLE 5.9-1
.. .' MEASURED CORE POWER DISTRIBUTION RESULTS Control Rod Group Positions Gps 1-4 100% vd Gp 6 76% vd Gp 5 100% vd Gp 7 0% vd Gp 8 34 % vd Core Power Level 16.4 % FP Boron Concentration 1340 ppm Core Burnup 0.45 ETPD Axial Imbalance -2.0 % TP Xenon Conditions Equilibrium Conc. No Yes or No Reactivity Worth NA % Ak/k Max Quadrant Tilt T% 1/8 Core incore Pmax/? core P/P Fuel FA Loc. Det. No. Local Assembiv H-08 1 1.15 1.11 G-08 2 1.73 1.32 F-Ub 4 1.74 1.32 E-08 10 1.83 1.41 D-08 14 1.61 1.10 C-08 21 1 72 1 .11 B-08 10 1 R7 1 44 A-08 37 1.22 0 04 G-09 3 1.67 1.27 F-10 12 1.66 1.29 E-11 26 1.43 1.06 D-12 41 1.15 0.80 C-13 52 0.66 0.52 F-09 6 1.83 1.42 E-09 5 1.62 1.26 D-09 15 1.60 1.25 C-09 29 1.38 1.04 B-Oir 31 1.31 1.07 A-09 45' 1.00 0.78 E-10 17 1.68 1.30 D-10 27 1.35 O.99 C-10 28 1.24 0.97 B-10 44 0.87 0.69 A-10 46 0.56 n.44 D-11 33 1.43 1.08 C-11 42 1.09 0.84 B-11 49 0.80 0.64 C-12 48 1.03 0.80 B-12 51 0 AA U.a* TABLE 5.9-2 kk
.
- * '
HEASURED CORE POWER DISTRIBUTION RESULTS Control Rod Group Positions Gps 1-4 100% vd Gp 6 75% vd Gp 5 __ 100% vd Gp 7 0% vd Gp 8 33% vd Core Power Level 40,47% FP Boron Concentration 1207 ppm Core Burnup 1.8EFPD Axial Imbalance -5.1% FP Ienon Conditions Equilibriun Conc. Yes Yes or No Reactivity Worth 1.91 % Ak/k Max Quadrant Tilt 0.41 %
.
1/8 Core incore F ax/Pcore P/P Fuel FA Loe. Det. No. Local Assemb1v H-08 1 1.39 1.12 G-08 2 1.66 1.29 F-08 4 1.74 1.17 E-08 10 1.78 1.38 D-08 14 1 90 1 17 _ C-08 21 1.66 1.20 B-08 30 1.80 1.38 A-08 37 1.20 0 07 G-09 3 1.61 1.26 F-10 12 1.64 1.26 E-11 26 1.47 1.04 D-12 41 1.16 0.92 C-13 52 0.69 0.55 F-09 6 1.80 1.40 E-09 5 1.59. 1.25 D-09 15 1.59 1.23 C-09 29 1.33 1.04 B-09 31 1.24 1.04 A-09 45 1.00 0.79 E-10 17 1.72 1.29 D-1G 27 1.42 1.00 C-10 28 1.25 0.97 B-10 44 0.88 0.72 A-10 46 0.55 0.45 _- D-11 33 1.42 1.09 C-11 42 1.11 0.85 B.11 49 0.86 0.69 , C-12 48 1.07 0.84 B--12' s1 0.e0 ' o 67 TABLE 5.9-3
}k}k
.. . . MEASURED CORE POWER DISTRIBUTION RESULTS Control Rod Group Positions Gps 1-4 100K vd Cp 6 77% Vd Gp 5 100% wd GP7 5% vd Gp 8 27% vd Core Power Level 77% FP Boron Concentration 1090 ppm Core Burnup 9.87EFPD Axial Imbalance -12.43 % FP Ienon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth H % Ak/k Max Quadrant Tilt U%
.
1/8 Core Incore Pmax/Fcore P/P Fuel FA Loc. Det. No. Local Asse~biv H-08 1 1.41 1.14 G-08 2 1.67 1.29 F-08 4 1.76 1.33 E-08 10 1.81 s 1.38 D-08 14 1.68 1.21
. C-08 21 1.70 1.26 B-08 30 1.74 1.35 A-08 37 1.15 0.91 G-09 3 1.58 1.26 F-10 12 1.66 1.26 E-11 26 1.52 1.05 D-12 41 1.22 0.42 C-13 52 0.69 0.54 F-09 6 1.79 1.39 E-09 5 1.65 1.28 D-09 15 1.70 1.24 C-09 29 1.41 1.05 B-09 31 1.23 0.99 A-09 45 0.95 0.75 E-10 17 1.81 1.30 D-10 27 1.50 0.99 C-10 2f 1.17 0 07 B-10 44 0.87 0.74 A-10 4s 0.57 0 6A D-11 33 1.48 1.10 C-11 L2 1.16 0.85 B-11 49 0.91 0.71 C-12 48 1.09 0 R4 b-14 51 0.40 0.ao TA3LE 5.9-4
}[
- - *'
MEASURED CORE POWER DISTRIBUTION RESLITS Control Rod Group Positions Gps 1-4 100% wd Gp 6 os: vd Gp 5 100% vd Gp 7 ,o: vd Gp 8 14.5% vd Core Power Level 99% FP Boron Concentration 1101 ppm Core Burnup 21.2I??D Azial Imbalance -2.3% TP Ienon Conditions Equilibrium Cone. Yes Yes or No Reactivity Worth 2.62
- Ak/k Max Quadrant Tilt 1.39 %
1/6 Core Incore Pnax/? core F/P Fuel FA Loc. Det. No. Local Assembiv H-08 1 1.50 1.20 G-08 2 1.54 1 20 F-08 4 1.63 1.28 E-08 10 1.83 1.37 D-08 14 1,64 1.23 C-08 21 1.60 1.25 B-08 30 1.65 1.33 A-08 37 1.09 0.89 G-09 3 1.58 1.23 F-10 - 12 1.61 1.23 E-11 26 1.45 1.08 D-12 41 1.26 0.92 C-13 52 0.66 0.52 F-09 6 1.75 1.35 E-09 5 1.60 1.26 D-09 15 1.65 1.24 C-09 29 1.33 1.05 B-09 31 1.15 1.00 A-09 45 0.88 0.75 E-10 17 1.83 1.30 D-10 27 ' 1.41 1.00 C-10 28 1.26 1.01 B-10 44 0.96 0.81 A-10 46 0.58 0 49 D-ll 33 1.48 1.12 C-11 42 1.18 0.86 B-11
'
49 0.88 0.72 C-12 46 1.07 0.83 n-12 51 0 41 ,
._6 47
~3ntv 3.9-5 k
.
.
T 4 IllllIt!Ulf ENBR AllD ltAXIlfUlf LINEAR llEAT RATE 11EASURED FOR S* E ADY STATE, EQUIL1BRIUlf XE!!0!I C0!iDITIO:IS POWER IllCORE 1!AXI!!Ulf AXIAL !!AXI!El ?!AXI! E l
'JATE LEVEL I!! BALANCE PEAK LOCATION TUEL LIIR(1) LilR(2) III!:I!!IDI TIllE (%FP) (%FP) (Pmax/Pcore) (SEC!E!!T) ASSEllDLY (ku/ft) (ku/ft) DI!BR 6/20/74 16.4 - 2.0 1.87 4 B-08 1.06 2.66 25.82 0713 105.5 11.97 17.11 4.01 112.0 12.70 13.17 3.78 g 7/01/74 40.47 - 5.1 1.80 3 B-08 4.49 6.43 10.28 %; 1010 105.5 11.70 16.76 3.94
[* 112.0 12.43 17.79 3.71 Y 7/21/74 77 -12.43 1.81 2 E-08 8.33 11.93 5.42 0844 105.5 11.41 16.35 3.96 112.0 12.12 17.35 3.73 8/09/74 99 - 2.3 1.83 4 E-08 10.06 14.41 4.01 0103 105.5 10.72 15.36 3.60 112.0 11.38 16.30 3.25 - 4 - 4
--
(1) Worst case values of LilR without uncertain:les (2) Worst case values of L11R multiplied by 1.412 to account for measurement uncertainties
. .. CORE PROTECTIO;c LAFCTY LIMIT 5 THPJE MILE ISLAND NUCLEAR STATION UNIT I Thermal Power Level, 5
-- 120 112.0 tr/ft Limit +I8'4 kw/ft Limit Slope = .96 16.9
-
- 100
,
86.7 h
- - 80 +19.3 2 slope = .1.46 DNDR Lim 8(
Slope = .63
~58,9 1 60
+25.6 ~h Slope = 78
-- 40
--
20 f I f I 1 l 60 40 -20 0 +20 +40 +60 Reactor Pouer I:balsnee, 5 CURVE REACT 03 C0^!.UJT Ft0"? (LB/HR) 1 131.3 x 106 2 98.1 x 106 3 04.4 x 106 -, g)k \rDO FIGURE 5.9-1
.
.
T 4 LOCA LIMITED MAX MUM ALLOWABLE LillEAR llEAT RATE 111REE MILE ISLAND NUCIIAR STATION UNIT I 20 _ _ 18 M R 5 9 Yn -
--
_=- _ [\ - P %
"
?
- 16
%
E!
%
5 2 14
- I
~ * -
l
-
p _- 12 I
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$ 0- 2 4 6 8 10 12 Axial Location from Bottom Of Core.Ft:
.. .. OPERATIONAL POWER IMBALANCE ENVELOPE TIIREE MILE ISLAND NUCLEAR STATION UNIT 1 POWER LEVEL, 5 RESTRICTE0 102 REGION
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.. .- COMPARISON OF MEASURED AND CALCULATED RADIAL CORE POWER DISTRIBUTIONS AT STEADY STATE DISTRIBUT1011S AND STEADY STATE, EQUILIBRIUM XENON, 15% FP CONDITIONS. CORE CONDITIONS CALCULATED ACTUAL CALCULATED ACTUAL GPS 1-4 at 100 %Wd 100 %Wd CORE POWER LEVEL 15 %FP 16.4 ZFP 5 at 100 %Wd 100 %Wd BORON CONC. 1441 opm 1340 ppm 6 at 75 %Wd 76 %Wd CORE BURNUP 0 EFPD 0.45 EFPD 7 at 0 %Wd 0 %Wd AXIAL IMBALANCE -5.04 %FP -2.0 *FP 8 at 37. 5 %Wd 34 %Wd MAX. QUADRANT TILT. O% 0.5 1.04 1.37 1.36 1.53 1.19 1.30 1.33 0.90
- 1. N 1.32 1.32 ' 1.41 1.19 1.31 1.44 0.94 j
0 1.57 1.29 1.31 1.00 0.93 0.73
- 1. 1,42 1.26 1.25 1.04 1.07 0.78 2 1.38 0.96 0.92 0.60 0.42
- 1. 1.30 0.99 0.97 0.69 10.44 1 7 1.09 0.77 0.63
- 1. 1.08 0.84 0.64 86 0.83 0.50
- 0. 0.80 0.44 59
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X.XX Calculated Results X.XX Measured Results, Group 36
\So, 1414 FIGURE 5.9-4
.. .. COMPARISON OF MEASURED AND CALCULATED RADIAL CORE PO*aTR DISTRIBUIIONS AT STEADY STATE DISTRIBUTIONS AND STEADY STATE, EQUILIBRIUM XENON, 40% FP CONDITIONS. CORE CONDITIONS CALCULATED ACTUAL CALCULATED ACTUAL GPS 1-4 at 100 Nd 100 Nd 5 at CORE POWER LEVEL 40 IFP _ an_av %FP 100 Nd 100 Wd BORON CONC. 1430 ppm 17n7 ppm 6 at 75 Nd 75 Nd CORE BURNUP 1. 6 ETPD 1_R EFFD 7 at OWd OWd AXIAL IMBALANCE -5. 7 %FP _s.11 %FP 8 at 30.8 Wd 33 Wd MAX. QUADRANT TILT. n% 0.41 % 0.98 1.29 1.26 1.44 1.12 1.28 1. n 0.96 _ 1.1h 1.29 1.32 1.38 1.17 1.29 1.38 0.03
.21 1.47 '1.21 1.27 1.00 0.97 0.78
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X.XX Calculated Results X.XX Measured Results, Group 36
}4}k !b,i FIGURE 5.9-5
.. .. COMPARISON OF M!.ASURED AND CALCUMTED RADIAL CORE POWER DISTRIBUTIONS AT STEADY STAII DISTRIBUTIONS AND STEADY STATE, EQUILIBRIUM XENON, 76: FP CONDITIONS. CORE CONDITIONS CALCULATED ACTUAL CALCULATED ACTUAL GPS 1-4 at 100Nd 100 Nd CORE PGJER LEVEL 75 %FP 77.1 %FP 5 at 100Nd 100 Nd BORON CONC. 1134 DPm 1000 ppm 6 at 75Nd 77.3 Wd CORE EURNUP 15. 2 EFPD 9.87 EFPD 7 at OWd 5. 3 Wd AXIAL IMBALANCE -13.0 %FP -12.43 %FP 8 at 3'2' 8Md 27.3 Nd MAX. QUADRANT TILT. O% 1.18 % 1.00 1.32 1.30 1.46 1.13 1.25 1.28 o on _g 1.:\ 1.29 1.33 1.38 1.21 1.26 1.35 0.91 1 '5 1.51 1.24 1.27 0.99 0.92 0.74
- 1. 1.39 1.28 1.24 1.05 0.99 0.75
.
- 1. 7 1.34 0.96 0.94 0.62 0.46
- 1. 1.30 0.99 0.97 0.74 0.46 05 1.12 0.82 0.69
- 1. 1.10 0.85 0.71 91 0.91 0.58
- 0. ' O.84 0.48 68
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X.XX Calculated Results q X.XX Measured Results, Group 36 u 4}k i FIGURE 5.9-6
.. .- COMPARISON OF .W.JSURED AND CALCL"dTED RADIAL CORE F0WER DISTPlBUTIONS AT STEADY STATE DISTRIBUTIONS AND STEADY STATE, EQUILIBRIUM XENON,100%FP CONDITIONS. CORE CONDITIONS CALCULATED ACTUAL CALCULATED ACTUAL GPS 1-4 at 100%Wd 100 %Wd CORE POER n LEVEL lon %FP oo %FP 5 at 100%Wd 100 %Wd BORON CONC. 96 A vpt: Si n, DF3 6 at 87.5%Wd 93 %Wd CORE BURNUP 25 EFPD 21.2 EFPD 7 at 12.5 %Wd 14.5 %Wd AXIAL IMBALANCE -2. 5 %FP -2.3 %FP 8 at 22.5 Wd 21.2 %Wd MAX. QUADRANT TILT. O% 1.39 % 1.02 1.28 1.j7 1.43 1.16 1.24 1.24 0 RR _g 1.:h 1.29 1.28 1.37 1.23 1.25 1.33 0.89
- 1. 2 1.45 1.23 1.27 1.00 0.92 0.74 1.2 1.35 1.26 1.24 1.05 1.00 0.75 25 1.33 0.98 0.95 0.66 0.48
- 1. 1.30 1.00 1.01 0.81 0.48 '
08 1.13 0.83 0.72
- 1. 1.12 0.86 0.72 92 0.90 0.58
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_ , h X.XX Calculated Results X.XX ::aasured Results, Group 36 FIGURE 5.9-7
F G $ ' HOT GANNEL MINIMUM DNBR VS REACTOR POER
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.. .. 5.10 NUCLEAR STEAM STSTEM HEAT BALANCE 5.10.1 PURPOSE The NSS Heat Balance test was performed as required durine power escalation with the following objectives: (a) To deter =ine the core thermal power between 0 and 15: power using the core differential temperature method. (b) To deter =ine the core thermal power based upon a primary side calorimetric measurement. (c) To determine core thermal power based upon a secondary side calorimetric measurement. (d) To provide a value for " Core Ther=al Power" to be used in the calibration of the power range nuclear instrumentation. (e) To demonstrate the accuracy of the conputer calculated heat balance. 5.10.2 TEST METHOD Steady state conditions were established, as indicated, below, at each of the specified povar plateaus prior to recording test data. (a) All four reactor coolant pu=ps operating. (b) The plant computer operating. (c) Reactor coolant pressure at 2155 125 psig. (d) RCS averare temperature at 57912 F (except for 0: to 15 FP) (e) Feedwater te=perature stable with less thas 12 F change in 15 min. (f) Feedwater average flow rate stable with less than 13 chanze in 15 min. When the required equilibrium conditions were verified, the following data was collected over a 30 minute interval. (a) Selected co=puter data points. (b) Computer group 32, Reactor Coolant Heat Balance. (c) Selected reactimeter-patch panel points of pri=a:7 and secondary parameters. (d) Computer group 41, Steam Generator Performance. Calculation of core thermal power by pri=ary and secondarv heat balance was then done by hand and cou: pared to the computer results. After the computer calculated values were verified at each major power plateau, the computer values for Core Ther=al Power were used for core power deter =ination. Since the pri=ary and sec-ondary heat balance calculations are inaccurate below 15: FP, core power was 5.10-1 \
e . .- date:=ined by core differential te=perature during the initial escalatien from 0 to 15 FP. Deter =ination of reactor coolant flow rate from heat balance data was performed by setting the pri=ary heat balance equal to the secondary best balance on each loop and solving for the loop primary flow rate. Total coolan: flow was then obtained from the sum of the loop flows. 5.10.3 TEST RESULTS Initial data analysis at low power levels indicated a possible source of error in the data taking method for the hand calculations. The computer used time averaged values for the calculations, wheras the hand calculations were perferned using single data sets. Normal plant oscillations introduced sizeable errors in the hand calculated values based upen single data sets when compared to the time averaged computer results. Average parameter values over a 15-20 minute period were then used for the hand calculations to obtain more representative data. Core thermal power was determined from core differential te=perature during the initial escalation to 15: FP. Figure 5.10-1 presents the normalized reactor coolant AT versus core power relationship used as a primary side heat balance. The results of primary and secondary heat balance measurements performed from 30* to 100: FP are presented and compared in Table 5.10-1. Good agreement was found between the hand and computer calculated heat balances. As can be seen from Table 5.10-1, the primary heat balances were consistently lower than the secondary balances. This difference is due mostly to the indicated reactor coolant flow being slightly less than the actual flow, as determined by setting primary and secondary heat balance's equal. Reactor coolant flows calculated by setting the primary side heat balance equal to the secondary side balance are listed in Table 5.10-2. Preliminary calculations show that the indicated flow is an average of 2: less than the calculated flow. The best estimate for total reactor coolant flow at 100: FP is 108.6% of design, based on a design flow rate of 131.32x10 61bm/hr. 5.10 4 CONCLUSIONS Primary and secondary heat balance calculations were performed during power escala-tion. Good agreement was found between the hand and com:puter calculated values for core thermal power. Primary side heat balances were consistently 2: lower than secondary side heat balances; therefore, actual or calculated RC flow is about 2: higher than indicated RC flow. The best es:imate for total reactor coolan: flow at 100: FP using normal plant temperature, pressure and flow indications is 108.6% of the design flow rate.
.
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IIEAT BALANCE CALC 8JfAT10N SUtetARY Nominal D1ff Diff Diff Diff Diff Power Pc Ph Ph-Pc Sc Sh Sc-Sh Pc-Sc Ph-Sh Be Bh Bh-Bc (% FP) (Wt) (Wt) (% FP) (Wt) (Wt) (% FP) (% FP) (% FP) (Wt) (Wt) (% FP) 30 774 781 +.28 814 794 .78 -1.6 .51 781 770 .43 35 934 894 -1.6 971 955 .63 -1.5 -2.4 944 908 -1.4 40 1054 1013 -1.6 1083 1066 .67 -1.1 -2.1 1063 1028 -1.4 50 1257 1202 -2.2 1285 1265 .79 -1.1 -2.5 1269 1226 -1.7 65 1677 1599 -3.1 1695 1691 .16 .71 -3.6 1688 1655 -1.3 h 75 V907 1880 -1.1 1941 1919 .87 -1.3 -1.5 1931 1908 .91 r.: m 85 2120 2100 .79 2139 2167 -1.1 .75 -2.6 2136 2155 +.75 L { 95 2374 2347 -1.1 2409 2388 .83 -1.4 -1.6 2407 2385 .87 100 2454 2426 -1.1 2504 2486 .71 -2.0 -2.4 2503 2484 .75 100 2438 2422 .63 2488 2466 .87 -2.0 -1.7 2487 2464 .91 Average Deviation: -1.3 .74 -1.3 -2.1 .89 ,,, Pc: Primary computer heat balance Sh: Secondary hand heat balance f[1 Ph: Primary hand heat balance Bc: Best estimate, computer
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REACTOR COOLANT FLOW CALCUIATION FR0ff PRIMARY - SECONDARY IIEAT BALANCE Nominal Difference Power Indicated Flow Calculated Flow Indicated-Actual Date Time (% FP) (x1061bm/hr) (x1061bm/hr) (Z) 7/14/74 0320 75 139.40 142.21 -1.98 8/03/74 0500 85 139.73 143.73 -2.78 8/03/74 1445 95 139.95 141.83 -1.33 g 8/06/74 0320 100 139.59 142.64 -2,14 r; , Y r ? u s N
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.. .. 5.11 REACTIVITY DIPLE* ION VEPSUS BURNOP 5.11.1 PURPOSE The purpose of the Reactivity Depistion v.. Burnup test was to determine the core excess reactivity based upon measured critical boron concentrations at various times in core life. Once the core excess reactivity is known, it can be used as the basis in a reactivity anomaly calculation. 5.11.2 TEST METHOD The depletion test was performed with the reactor at 100: TP with 2-D equilibrium xenon established. The normal operating control rod configuration was v. sed with group 6 at 87.5% and group 7 at 12.5% withdrawn. The pressurizer was sprayed previous to conducting the test to assure that RCS, pressurizar and make-up tank boron concentrations were at equilibrium. Three separate make-up tank, pressurizer and RCS boron samples were taken when all required steady state conditions were satisfied. The test was scheduled to be run at 20, 30, 40 and 50 ETPD during the startup test program. 5.11.3 TEST RESULTS The result of the measurements made at 22.0 EFPD are plotted in Figure 5.11-1.
- The average measured RCS boron concentration was 1091.67 ppm. This value was adjusted by 4.51 ppm to obtain 1087.16 ppm as the final result. The adjustment was made to account for minor deviations in control rod position and core power from the values used in the calculations. The measured critical boron concentratior was in good agreement with the predicted value of 1060 ppm.
The scheduled measurements at 30, 40 and 50 EFPD vere not performed as part of the test program but will be conducted as part of the nor=al plant surveillance testing. 5.
11.4 CONCLUSION
S The measured critical boron concentration at 22.0 EFPD and 100: TP conditions was within 30 ppm of the predicted result and well within the acceptance criteria value of 86 pp=. 5.11-1 $ (' b 4}k
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.. .. 5.12 NELTRON NOISE MEASt*.REMEhiS Neutron noise data was recorded on the TMI Unit I core during the startup test program to serve as baseline data for future periodic measurements. Signals from the four power range detectors and eight other primary and secondary plant parameters were recorded on a Honevve11 model 3600 m tape recorder af ter es-tab 11shing steady state three-dimensional equilibriu= xenon conditions at 40", 76: and 100% of full power. A brief summary of the measurement conditions is given below. Initial analysis of the data indicates no major differences fro:
*he expected neutron toise signatures.
Power ETP Boron Rod Positions, I ud Level Date Days (nom) Go 6 Go 7 Go 8 40: 7/1/74 1.9 1202 76 4 32 76: 7/19/74 8.5 1120 74 0 30 100: 8/8/74 20.7 1104 94 20 12 5.12-1 1
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.. .. 6.0 NUCLEAR STEAM SYSTEM PEPJORMANCE Several tests were performed during power escalation and operation at r'ull power to monitor the performance of the nuclear steam system. The test results pre-sented in this section provide a discussion of reactor coolant syste= perfor=ance under steady state and transient conditions, reactor coolant pump performance and a summary of radioactive vaste management and primary and secondary syste= vater chemistry. Steady state and transient operation of the reactor coolant system and steam generators was monitored at various power levels during the escalation to 100ll FP. The response of reactor coolant inlet, outlet and average temperature; steam generator pressure, temperature and level; and feedwater flow and temperature versus reactor power was iister=ined. Reactor coolant pump vibration and reactor coolant syste= leakage were maintained within the specified operational and Tech-nical Specification limits. Radioactive vastes generated during power operation were adequately processed, stored and/or disposed using plant and off-site facilities. Primary and secondary water chemistry have been maintained within limits allowable for operation at power. Radioche=istry analysis of reactor coolant activity indicate that no fission product releases occurred during the startup test program.
)k)h N 6.0-1
.. .. 6.1 REACTOR COOLANT SYSTEM PERIURMANCE 6.1.1 PURPOSE A number of tests were performed prior to and during the escalation of the unit to rated power to monitor the performance of the Reactor Coolant System _(RCS) and to venfy the capability of the individual components to support operation at power. Varicas RCS parameters were monite ed to identify and eliminate any oscillatory or unstable characteristics during thL escalation to power. This section presents the results of those tests and an evaluation of: (a) Steady State Operation of the reactor and steam generators at several reactor power levels. (b) The Reactor Coolant System response to major unit transients. (c) The performance of the reactor coolant numps. (d) Reactor Coolant Syste= leakage. 6.1.2 TEST METHOD During escalation of the unit to rated power, selected reactor coolant, steam gen-erator and reactor coolant pump parameters were recorded by the plant computer,- reactimeter and Brush recorders at the various power levels specified in the power escalation test sequence. The data collected for steady state testing was recorded over a thirty minute interval, af ter steady state steaming conditions were estab-lished at each plateau. Data was collected at a high recording frecuency during transient tests with several minutes of steady state data before and af ter the transients. The recorded values were plotted and/or tabulated, where applicable, < and then compared to predicted results. 6.1.3 TEST RESULTS 6.1.3.1 Steadv State Deeration Reactor Coolant System and steam generator parameters were monitored af ter establish-inn steady state conditions at 0%, 15%, 25%, 35%, 40", 50%, 65%, 76%, 85%, 95% and 100% of full power with four reactor coolant pu=ps operating. The average values for reactor coolant inlet and outlet te=perature; steam generator pressure, te=perature and level; and feedwater flow and te=perature were computed, plotted and compared to the predicted response of these parameters versus reactor power. Results are pre-sented in Figures 6.1-1 through 6.1-7. The plotted data at each plateau was extra-polated to the next power plateau to determine whether any operating limit would be exceeded during the escalation. Figure 6.1-1 shows the reactor coolant outlet, inlet and average te=peratures versus power. During the escalation to 15: FP, the reactor coolant average temperature did not follev the predicted response and was not maintained at 57907 at 13 FP. The steam generator level control setpoints were lowered from 30 in on the startuo range indication to 28 in. to obtain a 579 7 average RC tertperature at 15" FP. 1^
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6.1-1 )h\4
.. .. As can be seen from Figures 6.1-2 through 6.1-7, the response of the stea: 8enerator major parameters versus NSS power was as expected with the exception of the OTSG A and B outle: stes= pressures. Although the 0750 outlet pressures increased as expected with increasing power, the magnitude of the pressure reported by the plant computer was less than expected. This is attributed to a _2* ins:rument string error band, which is greater than the allowable parameter devia:1on specification of 11%. A more precise string calibration would yield values for steam pressure that lie within the allowable pressure band. This was verified by data recorded on the reactimeter, which did fall within the allowable deviations. The plot of OTSG A startup level on Figure 6.1-5 shows a step increase in level of about Si in. of water at 85: FP. There was no corresponding increase in operating level at this time. The change in startup level was attributed to a shift in the transmitter calibration. Table 6.1-1 lists the average values of the majo steam generator parameters recorded during escalation of the unit to 100: FP. Steam generator upper and lower downcomer temperatures were recorded at each power level listed above and are given in Table 6.1-2. These results show that the OTSG downcomers were always in an unflooded condition above 5 reactor power, since the upper downcomer te=pera:ures were always within 10 F of the lower downcomer temperatures. iunor oscillations were observed in the Reactor Coolant System and steam generator para =aters during the escalation to full power. Table 6.1-3 lists the period and amplitude of the oscillations observed in reactor coolant average temperature and turbine header pressure for various power levels. The amplitude of the oscillation in these parameters reached a - % of 10.81 F and 114.0 psi, respectively, at approximately 70% FP. The period of the oscillation is approxi-mately 4 seconds. At 100% power, the 4 second oscillation completely disappeared and only a low frequency oscillation in turbine header pressure was observed, with an amplitude of 13 psi. These oscillations stem from two sources. The 4 second oscillation was due to an oscillation in the steam generator boiling regions at partial flow-rates. The oscillation is not present under full load conditions. The cause of the low a=plitude oscillation at 100: FP is the slight drift of the average reactor coolant te=perature by about 10.3 F and the corres-ponding insertion or withdrawal of the control rods. The drift in reactor coolant average temperature is due to instrumentation drift within its deadband range. Overall, no maj or oscillatory problems were encountered during the startup and the reactor coolant syste= flows, temperatures and pressures matched their design values. 6.1.3.2 Reactor Coolant Svstem Transients The Reactor Coolant System was subjected to a nu=ber of varied transients during the startup program at TMI Unit I. Some of these transient operations were anti-cipated in the individual testing phases of the startup program but others resulted due to unpredicted plant startup problems. The nature of such transients and the corresponding behavior of the Reactor Coolant System parameters is important in regard to the nu=ber of transient cycles allowable for each component over the 40 year lifetime of the plant. Two differen: Reactor Coolant System transients have been selected for presentation in this section. Each represents a different transient operation for the Reactor Coolant Syste=. 6,1-2 k
.. ,e The first exa=ple of system transient behavior is the variation in Reac:or Coolan: Syste= parameters due to a turbine-generator load rejectice transient. The reactor was operating at 73 FP when the turbine-generator was inadvertently tripped during routine testing. In this type of transient, the turbine main stop valves shu: which decreased the heat transferred fro = the reactor coolant to the secondary side of each stean generator. A rapid buildup in the steam header pressure opened the main staa= safety relief valves to transfer some hea: across the steam generators. The redue: ion in heat transfer out of the reactor coolan increased the reactor coolant te=perature and the resultant expansion of the volume of the reactor coolant caused an increase in RCS pressure and pressurizer level. The reactor power was reduced to 15: FP by the in:egrated control syste=, thus reducing reactor coolant te=perature and pressure, which were subsequently controlled at their respective setpoints. Table 6.1-4 presents the envimum and mini =um values measured for reactor coolant pressure, temperature and pressurizer level during this transient. TABLE 6.1-4
."ASURED VARIATION LN R.C. SYSTEM PARAMETERS FOR AT TURBINE-GENERATOR TRIP Parameter Maximum Value Minimum Value RCS Pressure 2320 psig 1970 psig RCS Average Te=perature 590 F 570 F Pressurizer Level 296 in, water 200 in, water A second exa=ple of system transient behavior is a reactor trip from 100: FP.
Tripping the reactor will immediately trip the turbine-generator and cause the turbine stop valves to close and the bypass valves and main steam safety relief valves to open. As in th*: first exa=ple, heat is transferred from the reactor coolant through the steam generators and the reactor hot leg and cold leg tempera-tures decrease uniformly. The reduction in reactor coolant temperature and the corresponding shrinkage in coolant volume will cause a decrease in reactor coolant pressure and pressurizer level. The primary difference between this type of tran-sient and the first is the smaller increase in reactor coolan pressure as a result of tripping the reactor. Table 6.1-5 summarizes the maximu= variation in reactor coolant parameters as a result of this transient. IABLE 6.1-5 MEASURED VARIATION IN R.C. SYSTEM PARAME*ERS FOR A REACTOR TRIP FROM 100: FP Parameter _Fkxi=um Value Minimu= Value RCS Average Te=pera:ure 5790F 5450F RCS Pressure 2210 psig 1799 psig Pressurizer Level 251 in, water 60 in water Based upon the low pressurizer level reached during the uni: shutdown test conducted at 15* FP, the normal operating level of the pressurizer was raised from 220 in, to 240 in. The RCS te=perature decrease and the corresponding shrinkage in RCS volume during a reactor trip result in pressurizer levels below :he 80 in. low-low level setpoint. The pressuri:er heaters shut off below 80 in., and thus 11=1: operator control of RC pressure. Although the higher operating level has increased 6.1-3 1414 ! 7 >
a . . . the lov level reached during RCS transients, the reactor trip from 100: FP resulted in a pressurizer level of 60 in. Studies are now in progress to improve RCS response in this area. 6.1.3.3 Reactor Coolant Pumo Performance The performance of the Reactor Coolant Pumps of TMI Unit I has been satisf actory. These Westinghouse controlled seal leakage pu=ps have produced flows of 108.6% of the 88,000 gym design rate per pump. With the exception of minor oil leaks discovered in the pump oil coolers and upper bearing inspection plates during precritical testing and an 1= properly installed seal on the B RCP during hot functional testing, the overall mechanical performance of the pumps has been good. The measured shaf t vibrations have been relatively low at 5 to 12 thousandth of an inch (mils). The measured bearing te=peratures, component cooling and seal leakoff outlet temperatures have been within acceptable operating limits. 6.1.3.4 Reactor Ceolant Svstem Leakara Reactor Coolant System leakage was monitored during the startup test program as part of a periodic surveillance procedure. The method used to measure reactor coolant leakage was discussed in section 3.5. The results of measurements made while the reactor was critical are summarized below. The values listed are the marimum unidentified reactor coolant leakage measured during the specified time, period. The data shows that the m M m m unidentified leakage limit specified in the Technical Specifications of 1 gpm was not exceeded.
SUMMARY
OF RCS LEAKAGE DURING TEST PROGRAM Mart =mn Unidentified Period Leakare (gem) 6/8 - 6/14 +0.87 6/15 - 6/21 +0.61 6/22 - 6/28 +0.68 6/29 - 7/5 +0.11 7/6 - 7/12 +0.89 7/13 - 7/19 +0.43 7/20 - 7/26 +0.58 7/27 - 8/2 +0.74 6/3 - 8/9 +0.89 8/10 - 8/13 +0.43 8/22 - 8/31 +0.74 6.
1.4 CONCLUSION
S Steady state operation of the reactor coolant system and the steam generators was monitored at various power levels during the escalation to 100% FF. The average values for reactor coolant inlet, outlet and average te=perature; steam generator pressure, te=perature and level; and feedwater flow and temp-erature followed the expected response with power. The response of the reactor coolant syste= to major unit transients has been satisfactory. One area that is under study is the low pressurizer level reached during a reactor trip. The reactor coolant pu=ps have performed well and produce flows in excess of their design values. Reactor coolant system leakage was =aintained within the limits specified in the Technical Specifications. 6.1-4 1414 173
.
&
.
REACTOR C001,AFTP SYSTEM PARAtlETERS - AVERAGE VALUE VS. POWER - NSS OTSG OTSG TURB OTSG A OTSG B OTSG A OTSG B OTSG A OTSC B N0ff GEN A B llDR STM STlf FIM SU SU OP OP MM RX PWR IM PRESS PRESS PRESS TEt!P TEMP TEMP LEVEL LEVEL LEVEL LEVEL UF FI.0W TAVE PWR li tM(e) ps1R psig psig U F DF inches inches % % x10ibib/hr "F % 0 0 859 858 847 531 531 98.6 0.65
--- --- --- ---
531 0.97 15 102 872 881 881 581 582 292 26 35 7 3 1.5 578 17.0 25 164 886 882 882 584 585 285 38 36 8 7 2.3 579 26.4 35 240 884 888 879 586 587 287 51 48 13 11 3.3 579 35.9 h 40 278 869 880 881 588 588 376 55 53 15 13 4.1 579 39.4 m os 50 422 875 883 884 589 589 396 78 65 19
.
16 5.0 579 48.1 L f, 65 557 871 879 875 591 592 420 103 91 28 27 6.7 579 63.6 76 648 877 885 884 592 593 394 123 107 35 34 7.5 578 74.3 85 731 880 894 886 592 592 447 193 130 45 45 8.9 578 83.1 95 808 895 909 887 591 591 458 213 148 35 55 10.2 578 96.4 100 837 889 902 890 590 592 461 220 152 57 57 10.5 579 98.5 100 829 887 898 875 592 593 428 220 152 55 56 9.9 579 98.1 - N
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RCS ALLCE'ABI.E TEMPEPJ.TURE DETMON VS POWER 620
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IXMNC01ER TEMPERATURES F HSS OTSG OTSG OTSG OTSG Nominal A A Difference B B Difference Power (%) Lower Upper (Lowe r-Upper) Lower Upper (Lower-Upper) 0 527.6 527.5 0.1 524.2 --- 15 534.2 532.0 2.2 532.9 529.7 3.2 25 534.1 532.0 2.1 533.0 529.6 3.4 35 534.1 532.0 2.1 533.0 529.7 3.3 h 40 534.2 532.1 2.1 533.1 529.8 3.3 r os 50 535.2 532.5 2.7 533.7 530.3 3.4
's b 65
- 534.4 532.3 2.1 533.5 530.0 3.5 76 535.7 533.6 2.1 534.8 531.4 3.4 85 537.3 534.9 2.4 536.3 532.6 3.7 95 537.8 535.3 2.5 536.9 533.3 3.6 100 537.4 534.8 2.6 536.5 532.7 3.8
__ _.- CN
.. ,. PLANT STEAD ~ STATE OSCILLATIONS Oscillation in T AVE Oscillation in THS Power Level Period A=plitude Period Amplitude (I FP) (seconds) (07) (seconds) (esi) 44.0 4.0 10.41 4.0 15.5 50.0 4.0 10.44 4.0 17.9 58.6 4.2 _M . 38 4.1 18.6 62.5 4.3 10.56 4.3 114.0 71.3 4.4 10.81 4.4 110.5 76.0 4.6 _R . 50 4.6 18.0 97.5 none 10.32 none 13.0 1414 177 TABLE 6.1-3
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@ OTSG B 14\4 \l3 IIGURE 6.1-2
. . , . STFJL. GDERATOR OUTLET PRESSURE DEVIATION VS P0kTR 940 - . _ - _ _ . _ _ _ _
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.. .. 6.2 AUZILIARY SYSTEM pIFJORMANCE 6.2.1 RADICAC*IVE WASTE MANAGEMENT Radioactive wastes generated during power operation consist of liquid, gaseous, and solid wastes. Liquid radioactive wastes are basically generated from RCS letdown and RCS and auxiliary syste= leakage. Those wastes that are of reactor coolant grade chemistry are stored in the RC 31eed tanks (3 tanks at 88,000 gallons capacity, each) AC Elaed tank wastes can be filtered and/or de=ineralized, as required, to make minor adjustments in che=istry or to remove radioactive fission products and then evaporated at rates of up to 12h gpm in the RC Bleed evaporator. The distillate from the evaporation process is demineralized and stored in one of two 7,000 gallon capacity evaporator condensate tanks. This distillate is then either recycled back into the plant or dumped into the mechanical draft cooling tower effluent to the river at flow rates of up to 30 gym. The evaporator concentrate (high in boric acid concentration) is stored in either of the reclaimed boric acid storage tanks for reuse in the plant or in either of the two concentrated waste storage tanks for drumming. Those wates that cannot be practically cleaned up to reactor coolant grade che=1stry are stored in the 25,000 gallon capacity miscellaneous vaste storage tank. These wastes are then avaporated at rates of up to 12 gym in the miscellaneous waste evaporator. Evaporator aistillate is demineralized and stored in one of the evaporator condensate tanks. Agun, this distillate is either recycled back into the plant or du= ped into the mechanical draf t cooling tower effluent to the river at flow rates of up to 30 gpm. The evaporator concentrate is stored in the concentrated waste storage tanks for drumm hg. During the period fron initial criticality on June 5, 1974 through commencement of co=mercial operation on September 2, 1974, a total of 6.99x105 liters or 185,000 gallons of evaporator distillate was released from the evaporator condensate tanks to the river. Total radioactivity associated with these releases was 20.9 curies of tritium and 5.07x10-3 curies of other isotopes (mostly Co60 and Csl37). These values compare with the maximum allowable limit of 10 curies / quarter for iso-topes other than tritium and noble gases and a target limit of 1.25 curies / quarter for isotopes other than tritium and noble gases. I Gaseous radioactive releases are basically associated with one of three sources:
- 1) Purging of the reactor building
- 2) Nor=al_ventillation of the auxiliary and fuel handling buildings
- 3) Release'of a vaste gas decay tank During the period from initial criticality through co=mencement of commercial operation, the gaseous activity released from all three sources was .048x10-7uC1/
see of I131 and particulate with half lives >8 days and 3.3m /see 3 of gross gaseous activity. These values co= pare with the quarterly average li=1ts of .05uci/sec for Il31 and particulate with half lives >8 days and 1.9x10 43 m /see for gross gaseous activity; and the target a;terage limits of .024uci/see for Il31 and particulate with half lives >8 days and 4.8x103 :3/sec for gross gaseous activity.
.
;4\4 \V 6.2-1
.. .. Solid radioactive vaste consists of:
- 1) Solidified evaporator concentrates
- 2) Spent filter sulka floc and de=ineralizer resins
- 3) Miscellaneous solid wastes, such as spen: filter car: ridges During the power escalation test progra=, a total of 145 55-gallon drums of solidified evaporator bottoms were shipped off-site in two separate shipments.
Total radioactivity associated with these 145 drums was 2.286x10-3 curies. No other solid radicactive waste was disposed of during the power escalation test progra=. A =ajor portion of the radione:1ve liquid waste generated during the power escalation test program was the result of RCS leakage into the Reac:cr Coclant Drain Tank, (still considered a portion of the RCS for leakage calculations covered in sec: ion 3.5). Starting from initial criticality on June 5, 1974, this leakage was about 2.4 gpm. On June 19, the leakage had increased to 3.7 gym and caused RCDT te=perature to increase to =1900F. In order to keep RCDT te=pera-ture fro = exceeding 190 F, cool de=ineralized water was added :o the drain tank in large volu=es and the tank was pumped out to the miscellaneous waste storage rank or an RC bleed tank (depending upon which tank had enough reserve capacity to receive the contents of the RCDT). The high leakage was traced to the pressurizer spray valve ste= leakoff and was repaired on June 21. Leakage after the repair continued at the original value of 2.4 gps. On July 5, leakage rose to 5.5 gym and a significant number of RCDT "mixings and ducpings" were required to keep RCDT less than 1900F. The leak was located and isolated within 24 hours. Again, leakage returned to 2.4 gpm. As of Septe=ber 2, the leakage had risrn to =4 gp: and constitutad the major portion of liquid and solid (solidified evaporator botto=s) radioactive vaste generated. 6.2.2 PRIMARY AND SECONDARY SYSTEM WATER CHEMISTRY Water chemistry specifications were established for the primary and secondary systems to reduce the a=ount of corrosion that would occur over the lifetime of the plant. Periodic sa=ples of the pri=ary and secondary fluids were tahn during the startup test progra= to identify and el h b te any acverse conditions and to detect the presence of failed fuel assemblies. Pri=ary syste= vater chemistry was monitored through periodic sa=ples of the reactor coolant letdown, purifica: ion de=ineralizers, pressuriser, make-up tank, boric acid mix tank, core flood tanks, spent fuel pool, borated water storage tank and reclai=ed boric acid mix tanks. Secondary syste= vater chemis:ry was monitored through periodic sa=ples of the steam generator feedwater and condensate trains. Radioche=istry analysir of the reactor coolant was perferned during the startup test progra= to: (a) Monitor activity buildup in t' e reactor coolant during initial fuel loading, reactor startup and initial pever operation. (b) Establish base activity levels to deter =ine :he presence of failed fuel and pri=ary-secondary leakage. (c) Monitor for radienuclide leakage from fuel pins to coolant, fro coolant to steam generators or coolan: to closed cooling water sys:e=s during startup and initial operatien. 6.2-2 c
\ '
\h\h
.. .*
Overall, primary and secondary water chemistry specifications have been maintained during the startup test progra=. However, 'out-of-spec' conditions have occurred fron time to time due to operational or mache.ical problems. Water chemistry was returned to specification when the problems were corrected. No significant time delays were encountered during the test progra= due to che=1stry conditions. Radiochemistry analysis of the reactor coolant sys:e= yielded results which were reasonable and consistent with er.pected activities and identifiable isotopes. The data indicates that no failed fuel is present and iodine activities indicate less than the predicted activity for " tramp" uranium.
\\ Q[',
\h\h
_ 6.2-3
.. .. 7.0 BALANCE OF PLANT TESTIN0 This section presents the results of balance of plant testing, adjustments and operation at power. Balance of plan: syste=s consi=t mainly of the turbine generator, main steam, turbine bypass, atmospheric dump, condensate, feedwater, moisture separator, steam extraction and feedwater heating, heater drain, emer-gency feedwater, and cooling water syste=s. The cooling water systems consist of the circulating water, natural draf: cocling tower, intermediate cooling water, nuclear service closed cooling water, nuclear service river water, secondary service closed cooling water, secondary service river water and mechanical draft cooling tower systens. Turbine generator initial synchronicing, loading, monitoring, overspeed trip testing, and adjustments were performed in accordance with IP 800/9 - Turbine Generator Operational Testing. The turbine was tripped under load at 30% full power as part of TP 800/14 - Turbine / Reactor Trip Test; the generator was tripped under load at 100% full power as part of TP 800/34 - Generator Trip Test; several inadvertant turbine trips occurred during the power escalation test program. TP 800/9 will be discussed in this section whereas results of the major trips will be discussed in Section 8 of this report. The ability of the turbine bypass and atmospheric dump riystems to control steam pressure during steady state operation prior to loading the turbine, durins: tran-sients, and during tu:.bine trips was determined in TP 800/6 - Turbine Bypass System Test. A discussion of main steam safety valve operation is also included in this section. Monitoring of initial power operation of the condensate, feedwater, moisture separator, heater drain, steam extraction and feedwater heating systems, co=parison of operating flows, te=peratures and pressuresagainst design values at several power plateaus; and ind*'41 flushing of sections of the moisture separator and heater drain syste <ere performed in accordance with TP 800/7 - Feedwater System Operation and Testing. A discussion of the ability of the e=ergency feedwater system to maintain a supply of feedwater to the OTSG's during loss of the main feedwater pumps is presented in this section. Verification of the ability of the plant cooling water syste=s to adequately cool their service components plus one set of perfor=ance testing (summer conditions) of the natural and mechanical draft cooling towers were conducted per TP 800/30 - Power Escalation Test Checkpoints. Cooling tower perfor=ance testing under several other seasonal atmospheric conditions still need to be performed; Met Ed will perform these tests in their respec:ive seasons.
\
Q)h 7.0-1
.. ..
7.1 ITRBINE GENERATOR OPERA *IONAL TESTING 7.1.1 PURPOSE The purposes of turbine generator operational testing were to:
- 1) Monitor bearing vibration levels; stator, bearing, and valve chest tempera-tures; turbine shell, rotor and differential expansion; exhaus: hood spray operation; generator field current and voltage; generator hydrogen gas temp-erature and pressure; and turbine cycle performance during steady state and transient testing at the various power plateaus.
- 2) Verify the TG operating procedure for bringing the turbine up to rated speed
'
(1800 RPM), synchronizing it, and loading it.
- 3) Provide instructions for performing TG checkouts required prior to and during initial synchronization and loading; checkouts include voltage regulator ,
adjustment, overspeed trip testing, oil trip lockout testing, and protective relaying checks.
- 4) Monitor EG (electro-hydraulic control) operation during steady state and transient testing to verify acceptable control characteristics and stability.
- 5) Plot percent megawatts eleccrical VS percent megawatts thernal at 40%, 76 and 100% reactor power and compare with design values.
- 6) Perform the turbine generator acceptance test at 100% reactor power.
7.1.2 TEST METHOD The turbine generator was brought up to rated speed (1800 RPM), initially syn-chronized, loaded and tested in accordance with the steps in the test procedure (TP 800/9) and the operating procedure (OP 1106/1) in the following sequence, as directed by the Controlling Procedure for PET (TP 800/21):
- 1) With 15% full power stea= flow through the turbine bypass valves, varm the turbine steam chest, then accelerate it to 1800 RPM. Monitor bearing vibra-tion levels; stator, bearing and valve chest temperatures; turbine shell, rotor, and differen:ial cxpansion; and exhaust hood spray operation during this time.
- 2) Test oil trip and trip lockout systems.
- 3) Synchonize and load the turbine to 5 MWe, Re=ain at this load until bearing te=peratures equalize and exhaust hood te=peratures decrease below 125 7.
_.
- 4) Pick up generator lead until the bypass valves close and perfor= generator voltage regulator adjustments at 15 reactor power.
- 5) Unicad the generator and perform overspeed trip tests.
Synchroni:e and lead the turbine generator until the bypass valves close.
- 7) Print out computer groups 18 and 19 - turbine cycle perfor=ance and generator conditions and continue to monitor the parameters listed in 1) .
_ 7.1-1
)f)h \
.. ..
- 8) Perfor= step 7) at reactor power levels of 40, 76 and 100 .
- 9) Plot percent megawatts electric (corrected for differences in condenser vacuum between actual and design) VS percent megawatts ther=al at reactor power levels of 40, 76 and 100%.
. 10) Perfor= inservice protective relaying phase angle, current, and voltage checks at 40: reactor power.
- 11) Perform turbine generator acceptance test at 100* reactor power by calculating the TG gross heat rate.
- 12) Monitor the parameccrs of 8) during the design rate ramp unit load changes performed per the Unit Load Transient Test Procedure - TP 800/23.
7.1.3 TEST RESULTS During initial acceleration, synchronization and loading of the turbine generator at 15% reactor power, all TG parameters monitored were within limits of acceptance criteria. Several bearings exhibited high vibration (>3 mils), but not high enough to require a shutdown for balancing. One of the generator output breakers (GB1-02) could not be closed due to mechanical interference; it was repaired later on in the PET program and functioned satisfactorily. Also, a control problem developed with _ the turbine bypass valves whenever the turbine stop valves were open. This problem and its resolution are discussed in detail in Section 7.2. The oil trip lockout test was satisfactory; however, the mechanical trip finger could not be reset. The trip finger os replaced and the oil trip lockout test and reset were repeated with satisfactory results. The overspeed trip should be set at <1980 RPM (110% of 1800) and the backup overspeed _ set at (2016 RPM (112% of 1800) . Values deter =ined during the overspeed trip test were 1965 and 1990 RPM, respectively. - At 40 power, all TG parameters monitored were within limits of acceptance criteria. A very high frequency, low amplitude oscillation exhibited in turbine header pressure and in control valve servo currents was traced to turbine header pressure trans-mitter rack vibration; it was resolved by nreviding additional rack supports. Gen-erator output was 311 We vs a design prediction of 304.6 We. At 76 power, all TG parameters monitored were well within li=its of acceptance criteria except one - differential expansion approached its limit of 530 mils during a power runback from 76: to 55: power. Evaluation by General Electric resulted in a new limit of 560 mils. Several EHC system oil leaks at the main Control valves resulted in forced manual trips; however, replacement of the leaking fittings re-solved the problem. A reactor /curbine trip forn 76: power combined with tube leakage in the 4B high pressure heater and failure of the motor operated hi ?ter extraction steam isolation valve to close, resulted in water ingestion into the turbine; however, subsequent turbine operation appears nor=al and indicates mini =al, if any, damage was done. Investigation indicated that the tube leak (=1200 gpm) had existed for sone time and the failure was net caused by nor=al plant operation nor forces _ 7.1-2
}4}4 }[
.. ..
resulting fro = the trip transient. The heater float level control which should have closed the motor operated isolation valve was repaired and installation of a faster responding air operntor in place of the motor operator on a number of heaters is under consideration. Following heater tube repair, measured generator output was 668 We comapred with a design value of 661.5 We. A 30 HZ control valve oscillation which produced a 12 psi peak to peak turbine header pressure oscillation sas traced to the turbine speed sensor gear and noise pickup of associated cabling. The gear and cables were replaced at 100% power and the oscillation disappeared. At 100% power, all TG parameters were within limits of acceptance criteria except vibration on #3 bearing and generated megawatts. Bearing #3 had a vibration level of 3.5 mils, which is .5 mils greater than the acceptance criteria for long term operation. Short to intermediate ter= operation (up to a year or more) at this level is acceptable, however, with balancing to be performed at the first convenient opportunity. Generated megawatts at 100% power were 854 We at poorer vacuum conditions than design. When corrected to design vacuum by a very conservative calculation, gen-erated megawatts were 861 We. This is lower than the 870.4 on the reactor power vs generated megawatts design curve; however, GE only guarantees 837 We. Commercial plant operation has shown the vacuum correction calculations to be conservative because generation levels of 864 We have been obtainsd with vacuum poorer than design values . The gross heat rate at 100% reactor power was 9993 Beu/Whr compared with the acceptance criteria of y;10,002 Beu/Whr. Auxiliary plant load averaged 49 We. Currently, the clean steam generators produce steam with 33 degrees more superheat than minimum design; therefore, less steam is required to deliver a given amount of energy to the turbine and the turbine runs at something less than " valves vide open" conditions. Steam conditions currently are 10,621,000f/hr at 592 F0compared with turbine WO design of 11,158,286 #/hr at 5590F. 7.
1.4 CONCLUSION
S Turbine generator performance was very satisf actory throughout the startup test program. Approxi=ately 7 days of testing time were. lost due to unscheduled turbine trips and. turbine related proble=s. Water ingestion into the turbine through the 4B heater. extraction line due to the isolation valve failure to close on high shell side level caused by ruptured tubes was .the only unanticipated startup proble= which could have led to major damage and delays; however, subsequent turbine operation indicates that the turbine suffered no damage. #3 bearing vibration is approxi=ately 3.5 mils higher than acceptable for long term operation; balancing operations vill be performed at the.first convenient outage. TG output at 2535 W e is 861 W e, when conservatively corrected to design vacuum conditions, which ec= pares well with a guaranteed value of 837 se. Steam conditions are 10,621,000 #/hr at 592 F com-pared with design of 11,158,286 #/hr at 5590F. Due to the increased a=ount of superheat over design, the turbine operates at less than " valves vide open" condi-tions. Gross heat rate is 9993 5tu/Whr ec.= pared with design of ".0,002 Bcu/Whr. -
)h\
7.1-3
.. .. 7.2 TURBINE BYPASS SYSTEM *EST AND MAIN STEAM SAFE"T VA'VE OPERATION 7.2.1 PURPOSE The purposes of the turbine bypass system testing were to:
- 1) Deter =ine the capability of the bypass valves to control main steam pressure at turbine header pressure setpoint during steady state operation and following pressure perturbations at 10 to 15% full power.
- 2) Determine bypass valve opening times and peak main steam pressure following turbine trips from 30% and 100% full power.
- 3) Determine atmospheric dump valve opening times and peak main steam pressure foll'owing loss of offsite power testing at 15% full power.
In addition to the above tests delineated in TP 800/6, the main steam safety valves were monitored for lif t and reset pressures during the turbine, reactor, and gen-erator trip tests to verify initial setpoints and determine magnitude of setpoint shift as a result of repeated actuation. Results arc recorded in TP 800/14 and 800/34. 7.2.2 IIST METHOD At =10% power with the turbine stop valves closed, the ability of the turbine bypass valves to maintain constant curbine header pressure was determined and ICS adjust-ments made as required. Then . a 10 psi step change in header pressure setpoint was made and the ability of t?.e b fpass valves to control header pressure at the new setpoint deter =ined. All ICS adjustments were recorded in TP 800/8 - ICS Tuning at Power. Bypass valve opening times and ability of the bypass valves to control peak steam pressure following a 30% turbine trip and a 100: generator / turbine trip were to be determined by wiring up the valve limit switches to a Brush Recorder and monitoring main steam pressure. Atmospheric du=p valve opening times and ability of the bypass valves to control peak steam pressura following a 15: loss of offsite power were to be deter =ined by viring up the valve li=it switches to a Brush Recorder and mon-itoring main steam pressure. Main steam safety valve lift pressure's and shift in lift pressures during turbine, reactor and generator trips were determined by analyzing pressure recordings, monitoring thermocouples attached to each valve, and visual inspection. 7.2.3 TEST RESULTS Ability of the bypass valves to control turbine header pressure at setpoint with the turbine stop valves closed was satisfactory at 10% power. Peak to peak at:pli-tude of oscillations was an acceptable _H5 psi. Response to changes in header pressure setpoint was also satisfactory af ter it was discovered that the A header pressure controlled the B valves and B header pressure controlled the A valves and the wiring error was correctec. 7.2-1 4}( }9I
.. .. When the turbine stop valves were opened at 15 power in preparation for rolling and loading the turbine, it was observed that the two steam generators began steaming unevenly and the disparity increased with time. Analysis indicated that proxi=ity of A and B header pressure sensors to the steam chest resulted in a common pressure sensed by both when the stop valves were opened thus joining the A and B headers. Under these circumstances , small errors between actual and indicated pressures well within the range of calibration tolerances combined with the interaction of reverse flow stop check valves in the main stesn headers led to highly uneven steaming rates between the steam generators. In order to re-solve this problem, the control signal to each sat of bypass valves was modified to include the difference in pressures between the two steam generators. This modification . serves to converge the steaming rates. Since TP 800/6 did not cover pressure control by the bypass valves at 15: power with the stop valves open, a special operating procedure was written to cover it (SOP #46). Peak main steam pressure reached following the turbine trip from 30 power was 1025 psia; following the generator / turbine trip from 100% power, it was 1082 psia. This compares with a design peak of 1100 psia and a code allowable peak of 1170 psia. Valve opening time was not recorded during the transient due to limit switch misalignment caused by vibration; however, static times indicated a maximun open-ing time of 2.1 seconds compared with the acceptance criteria of 3 seconds. During the test program, it was discovered that design called for only four of the six bypass valves to be in operation at any given time; therefore, two valves were isolated and the 100% turbine / generator trip test was done with only four bypass valves in service. Peak main steam pressure reached following the loss of offsite power was 1008 psig in the A header and 1032 psig in the B; this was far less than the design peak of 1100 psia and the code allowable peak of 1170 psia. In fact, pressure reached such a low peak and decayed so rapidly that the atmospheric du=p valves did not need to open to control pressure; therefore, opening times were not measured. The main steam safety valves were originally set during factory testing. Their setpoints were checked and adjusted in place using a hydro set assist during hot functional testing. Subsequent hot functional testing indicated several valves were lifting at lower than setpoint so the valves were set a second time. During the 30% turbine trip test, one safety valve lifted =25 psi below setpoint and reseated at =25 psi below rescat pressure. The valves were reset for a third time; during resetting, it was deter =ined that an incorrect curve provided by rhe valve manu-f acturer for convers.ica.cf hydro set pressure to steam pressuri had been used on the previous field. sets. Traces of the 100% generator / turbine trip indicate that all safety valves lifted and resented at acceptable li=1ts; however, safety valve operation will be the subject of a continuing study on p1 Ant transient response opti=1:stion. 7.
2.4 CONCLUSION
S Acceptable response.ef the turbine bypass valves in. maintaining turbine header pressure setpoint and respense to.small. changes.in setpoint at reactor powers y_15 vss attained af ter the difference between stean generator pressures was included in the. control system and a wiring reversal error was corrected. Peak to peak oscillations are 16 psi.
\n7
'] w 7.2-2 kk
e . . . The turbine 'qpass valves, along with the t:ain steam safetv valves, function adequately to limit main steam pressure during turbine trips to <1100 psia. The longest valve opening time was 2.1 seconds; peak steam pressure following the 100% generator / turbine trip was 1082 psia. Operation of the atmospheric du=p valves was not even required to 11=1 main
-staam pressure to <1100 psia during the loss of offsite power test.
Final settings of the main stea= salety valves appear adequate for continued plant operation; however, safety valve operation is one of several areas under study in an atta=pt to opti= ire plant response to major transients.
\4 '
,. ,. 7.3 FEE %7ATER SYSTEM OPERATION AND TES*ING 7.3.1 PITRPOSE
.
The purposes of feedvater system testing were to:
- 1) Provide an organized approach to placing secondary cycle components into operation as a function of power level.
- 2) Verify that feedvater system and a.ssociated equipment perfor= without excessive oscillations during transients and steady state operations at 15, 40, 76 and 100% power.
- 3) Verify that feedwater system flow, temperature and cycle performance meet design specifications at 40, 80 and 100% steam flow.
7.3.2 TEST METHOD The secondary cycle systems and components (condensate, feedwater, moisture separator, heater drain, feedwater heating and steam extraction) were placed in operation as follows:
- 1) Prior to power operation, the level controllers on the heater drain tank, moisture separator drain tanks and feedwater heaters were set to provide anticipated .
stable control. Also, correct float cage centerline location with respect to tank or heater bottom was verified. Initial values of proportional band and reset were recordea for each controller.
- 2) As power level was increased (around the 40% power range), the moisture separator drain tanks, 6th stage heater drain tank, and feedwater heaters were flushed until their chemistry conditions were acceptable to permit no:=al valve lineup operation. .
- 3) At major power plateaus of 40, 76 and 100%, each controller was tuned for stability at steady state, if required..and then optimized by disturbing the controller flapper and monitoring the controller's ability to da= pen the resultant perturbation. If response was unacceptable, the proportional band and/or reset were tuned and the new values recorded along with the power level.
The final setpoints of each controllerwere recorded for each of the above power levels.
- 4) At 40, 80 and 100% full stea flow, feedwater cycle perfornance was compared with design. This included a comparison of feedwater flow, final feedwater temperature, ter=inal te=perature difference for each W heater and approach temperature for each W heater.
7.3.3 TEST RESITLTS The follcuing problems were uncovered and resolved to the point of supporting continued acceptable plant operation:
- 1) Three of the six moisture separator drain tank high level du=ps were blocked and three of the six MSDT pu-p discharge valves were either blocked or bound
-
up. This prevented moisture separator drain tank level controller tuning until af ter the turbine generator screen out. age shutdown and repair. Blocking of. these lines nay have been prevented by flushing the= to a drainage ditch for 7-14i4 194
- e. .-
initial cleanup prior to flushing to the condenser. Subsequent MSDT operation has proven satisfactory.
- 2) Feedwater flow oscillations were caused by oscillations of the heater drain pump discharge valves. Initially the two valves in the heater drain pump discharge header operated in series; in this mode, the first valve was greater than 70% open at 76: power and severe oscinations existed. Oscillations were significantly reduced by chatsging valve cams to somewhat linearize the equal percent valve plug, instaning sn%ers in the air supp.k lines to the diaphragm operators, and overlapping the control signals to permit paranel operation.
Currently, the first valve is 60% open and the second is 40% open at 100% fun power (both are <70%) . Even though oscinations are currently of low amplitude, a furthe reduction will be attempted by installing snubbers between the vdve diaphragm and plug.
- 3) Heater drain pump recirculation valve actuation also caused feedwater oscilla-tions. The recire valves were actuated at too low a flow rate by the 0-4000 gpm flow switches to narr.it accurate operation. An improvement in response was affected by changing the actuation setpoint from 400 to 800 gpm. Deirculation valve operation and its interaction with the integrated control. system is sein under investigation in an attempt to further optimize ICS performance.
- 4) Feedwater pu=p minimum speed for ICS control was lowered from 3300 RPM to 2800 RPM. This resulted in a lower flow at minimum speed, which caed the differential pressure .across the feedwater valves to drop to its control point of 35 psid at a lower power level. This also resulted in the nain feedwater flow block valve opening at a lower power level, which resulted in a smoother transition from startup feedwater valvo control to main feedwater valve control.
- 5) Severe transients on the control syste=, such as turbine trips, feedwater pump trips, and dropped control rods from high power levels revealed the fact that feedwate- flow to the steam generators during these transients decreased faster than the feedwater dmmmd signal from the ICS. Ihi, phenomenon is shown in Figure 8..-10 of the Section 8.1 discussion of plant transient tests and occurred because the high steam generator pressure caused by the transient pre-vented all of the demanded feedwater from entering the steam generators. This problem was resolved to the point of permitting continued plant operation by providing an error signal to the feedwater speed controller proportional to the difference between actual turbine header pressure and its setpoint. This is another area that is stin under investigation in an attempt to further optimize ICS performance.
- 6) Another phenomenon shown in Figure 8.1-8 of Section 8.1 is the increase in feed-water flow substantially above a decteasing feedwater de=and about 1-1/2 minutes after a feedwater pump trip from 100% power. This increase in flow also causes reactor power to increase for about 1-1/2 minutes; then both power and flow decrease to levels corresponding to feedwater demand. Prel1=inary investigation indicate that sluggish response of feedwater control valve position to valve differential pressure is responsible for this action and, even though response is acceptable to permit continued plant operation, =inor control system modifications will probably be made to further optimize ICS performance.
- 7) In section 7.1, the effects of ruptured tubes in the 43 heater on turbine gen-erator operation were discussed.
e 7.3-2 )k\
,. .. Review of information prior to the reactor trip at 76: power indicates that the ruptured tubes had existed for some time. For example, the "C" heater drain pump current was greater than full load current at 76: power with the normal nu=ber of drain pu=ps running (2) as are required for 100% power operation. Also, the enndensate booster pumps were running close to their full load current rating and cendensate flow was about 1200 gpu higher than expected. Heater performance data rec uired by TP 800/7 indicated less than acceptable heater performance at 40% and 3r 76: power; however, this was blamed on bad computer input data rather than feedwater heater leakage and the 40% and 76: heater perfor=ance data was not re-corded. Hindsight reveals that we should have paid more attention to these dis-crepancies . Heater performance data met acceptance criteria after the computer inputs were calibrated and condensate booster pump and heater drain pump currents were acceptable after repair of the 43 heater. 7.
3.4 CONCLUSION
S The condensate, feedwater, moisture separator, heater drain, feedwater heating, and steam extraction systems function acceptably to support steady state and transient operation at 100% power. Oscillations and transient response associated with these systems are acceptable; however, investigations are continuing in several areas in an effort to further optimize plant response. These areas are:
.
- 1) Heater drain pump discharge valve control.
- 2) Heater drain pump recirculation valve control.
- 3) Ability of the feedwater pumps to supply feedwater to the steam generators when turbine header pressure increases rapidly.
- 4) Ability of the feedwater control valves to respond to changes in valve differen-tial pressure.
Smoother initial operation of the moisture separator drain system may have been possible if we had flushed the high level dumps and pu=p discharge lines overboard prior to putting them into normal operation. This may have prevented several cases of valve blockage with debris. A review of heater drain pu=p and condensate booster pump current data and heater cycle perfor=ance data indicates that the 1200 gum tube leak in the 4B heater did not occur during the reactor trip at 76 and could possibly have been detected prior to the trip. Heater cycle perfor=ance meets design acceptance criteria at 100% power.
)h\k 7.3-3
,, . 7, 4 EMERCENCY FEEDWATER SYSTEM OPERAMON AND HSTING 7.4.1 PURPOSE The purpose of emerkency feedvater system testing during the power escalation progranvas to verify that the steam turbine driven emergency feedwater pump supplies adequate cooling water to the steam generators during the loss of off-site power test (TP 800/32) to remove decay heat generated in the core and limit the energy buildup in the reactor coolant system. During the hot functional test program, all automatic interlocks of the turbine driven emergency W pump were verified; also, head-flow curves of both 1/2 capacity motor driven and the full capacity turbine driven pu=ps were verified. 7.4.2 TEST METHOD During performance of the loss of offsite power test (TP 800/32), the stea= turbine driven emergency feedwater pu=p should start automatically. It should pump water into the steam generators at a rate great enough to increase their levels up to 95% on the operating range level instrumentation and maintain them at 95%, while removing all of the decay heat generated by the reactor. 7.4.3 TEST RESULTS The turbine driven emergency feedvater pu=p started feeding the steam generaters automatically, as expected. Emergency feedvater flow was so great and decay heat generation so lov that RCS pressure and temperature began decreasing as soon as the EfW pump started. Initial pressure was 2164 psig and te=perature was 582 F. In order to avoid exceeding a 1000F/hr cooldown rate, the emergency feedwater valve. were switched to r.anual as soon as both atea= generator levels were verified as increasing (the 95: level setpoint control had been verified during hot functional testing) and the feed rate slowed down by manually throttling the valves. Uneven feeding of the steam generators resulted from operator error, but final RCS pressure was 2004 psis and final te=peratures were 574 F out of the reactor, 522 F out of the A steam generator and 545 F out of the B steam generator. As a result of this test data, B&W has proposed a natural circulation level setpoint of 50* instead of 95: in order .o keep from exceeding cooldevn limits. Ihe 50 level is : ore than adequate for decay heat remeval by the EW syste=. 7.
4.4 CONCLUSION
S With the amount of decay heat present during performance of the loss of offsite power test, the turbine driven emergency feedwater pump provided more than adequate flow to control RCS temperature and pressura. In fact, the EW valves had to be throttled to keep from exceeding RCS cooldown limits as steam generator levels began increasing to 95% on their operating range level indication. A setpoint of 50* instead of 95: vill be used to adequately remove decay heat without exceeding cooldown rate limitations.
~1
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7.4-1
)h\4
.. .. 7.5 P0k'ER ESCA1.A" TON CHECKPOINTS
-
7.5.1 PURPOSE The purposes of perfor=ing power escalation checkpoints were to verify tha::
- 1) The secondary service closed cooling water system adequately cools its service components a: power levels of 15, 40, 76 and 1001.
- 2) The mechanical draf t cooling tower performs as designed at 15, 40, 76 and 100%.
- 3) The na: ural draf t cooling tower prforms as designed at 100% power under three different sets of climate and weather conditions.
- 4) The Powdex system performs as designed a: 100: power.
7.5.2 TEST METHOD At each maior power plateau (15, 40, 76 and 100 ) . the common secondary service closed cooling water inlet temperature to all SSCCL' heat exchangers and the outlet temperature of each SSCCh' heat exchaneer were recorded and valve adjustments were made,if required, to limit outlet temperatures to less than 95 F. Service com-ponent cooler flows were adjusted as recuired to maintain outlet temperatures within design limits. At each maior power plateau, the mechanical draf t cooling tower effluent tempera-ture and differential temperature between river water and effluent were recorded and compared with su=mer acceptance criteria of:
- 1) Differential temperature between effluent and inlet shall be no greater than
+7F or less than -3 0F during steady state operation for inlet temperatures less than 87 F.
- 2) If inle: temperature is 87 F or greater, then effluent temperature shall be maintained at or below inlet during steady state operation.
During RCS cooldown, the mechanical draf t cooling tower effluent temperature and differential temperature were recorded and the rate of change of differential te=perature was monitored and all were compared with the su=ner acceptance criteria of:
- 1) Differential te=perature between effluent and inlet shall be no greater than
+127 and this differential shall not decrease at a rate exceeding 2F0 /hr.
Perfor=ance testing of the MDCT was to be performed at 100% power operation and during cooldown by Marley Company (the supplier) in accordance with Gilbert Associates Specification 5572. The acceptance criteria are as follevs:
- 1) During cooldown with 3 fans running, the MDCT shgil cool 33,000 gpm from 108 F to 85 7 at an a=bient we: bulb te=perature of 78 F.
- 2) During cooldevn vi h 2 fans r-dag, the MDC shall cool 33,000 gpm from 1080F to 87.507 at an ambient vet bulb :e=perature of 78 F.
7.5-1
}k)k !9J
.. .-
- 3) During 100 power opera: ion, the MDCT shall cool 15,000 gp= from 110 F :o 8507 at an ambient we: bulb te=perature of 78 F.
Perfo =ance :esting of the na: ural draft cooling towers a: 100 power was per-fomed by Marley Co=pany a: one of the three conditions required. In order to de: ermine perfomance over the full range of climate and weather condi: ions no = ally experienced, perfo:=ance measurements are to be made under atmospheric conditions prevailing around April, August, and December - the August type perf omance evaluation was condue:ed soon af ter commencement of co=mercial operation in Septe=ber and pre 11=inary results are included herein. Acceptance criteria are in the fem of perfomance curves supplied by Marley. Powdex (condensate cleanup system) effluent chemical analysis was perfomed a: 100 power and compared with acceptance criteria. 7.5.3 TEST RESULTS The secondary service closed cooling veter system provides adeqtate flow to cool its ce=ponent heat loads at 100% power. With a SSCCW heat exchaager common inlet temperature of 87.6 F, outlet te=peratures at each of the four hast exchangers were 78 F, 74 F, 78"F and 790 F, respectively, compared with the acceptance criteria of 95'F. The differential temperature controller on the mechanical draft cooling tcwer was inoperable during the power escalation prea.am; therefore, effluent temperature had to be manually controlled. In the manual mode, both differential temperature and affluent temperature were within the limits of acceptance criteria at each power plateau and during cooldown when the readings were taken; however, changes in ambient conditions and failure, at times, of the control room operator to pro-vide continuous monitoring resulted in exceeding these limits several times. Performance testing of the MDCT at 100% power was not performed during the power escalation program due to tSe inoperability of the differential temperature controller. It will be performed at a later date. Perfomance testing of the natural draf t cooling towers resul:ed in an equivalent flow of 241,603 gp: per tower from the performance curves for August type condi-tiens compared with the design value of 232,000 gpe. This fs a capaci:y of 104.1% of design. April and Dece=ber type perfor=ance tests will be conducted by Marley at a later date. Powdex effluent che=ical analysis demonstrated satisfactory ability to control condensate che=istry. Analysis results are compared with acceptance criteria below: Analvsis Accentance Criteria
- 1) total dissolved solids 2 ppb 125 ppb
- 2) total suspended solids 0 110 ppb
- 3) dissolved SiO2 2rpb i 5 ppb
- 4) iron 0 1 5 ppb
- 5) copper 0 1 2 ppb 6' pH 9.41 9.3-9.5
- 7) conductivity .1 on highes: 1 1MME0/CM 3
vessel of 6 7.5-2 r;
.. .. 7.
5.4 CONCLUSION
S The secondary service closed cooling water adequately cooled its heat loads; SSCCW heat exchanger discharge te=peratures were well belov their design limit of 95*F at 100 power. The mechanical draft cooling tower effluent te=perature and differential temp-erature between inlet and effluent had to be controlled manually because the automatic con: roller was inoperative. Acceptable operation could be obtained with continuous surveillance; however, until operators gained fa=111arity with tower operation, differential temperature li=its were exceeded several ti=es. Final perfor=ance testing vill be conducted at a later date. The natural draf t cooling towers were performance tested under August type conditions. Capacity was determined as 104.1% of design. April and Dece=ber performance tests will be conducted at a later date. Powdex effluent chemistry analysis demonstrates acceptable capability to clean up the condensate syste= at 100% power operation. 1414 ^00 7.5-3
.. .. 8.0 UNIT PERFORMANCE The tests presen:ed in this section vere perforned during and after the escalation to 100: FP and measured the overall performance of the unit under normal operating, transient and energency conditions. A description of unit response to planned and unplanned major load changes is prescuted in the section on Uni: Transient Response. The Loss Of Offsite Power and Shutdown Fron outside the Control Roon Tests demonstrated the ability to safely control the unit under energency conditions. The Unit Accep:ance Test verified that the nuclear stean system can operate in accordance with the warranted design specifications. 14\4 ?D\ 8.0-1
.. ..
8.1 UNIT ""/JdSIENT RESPONSE 8.1.1 PURPOSE The capability of the Integrated Control Syste= (ICS) to maintain control of the turbine, steam generators and reactor under non-steady state conditions was demon-strated through various types of scheduled and unscheduled transients that occurred during the startup test program. Six different types of transient operation have been selected for presentation in this section to demonstrate ICS response to major load changes. The material presented includes a discussion of the following: (a) Transient testing perfor=ed at 40%, 76: and 100: FP that required load changes at rates equal to the design ramp rates while in the fully integrated, turbine following and reactor /stea= generator following modes of control. (b) ICS response during steady state and transient operations with three reactor coolast pumps operating at 25: FP.
~
(c) ICS response to tripping a main feedvater pump from 100: FP. (d) ICS response to a dropped control rod at 76 FP. (e) ICS response to a turbine trip fro = 76: FP. (f) ICS response to a load rejection at 100: FP. 8.1.2 TEST METHOD Tests conducted on the ICS at specified power plateaus during the escalation to full power were designed to optimize the perfor=ance of the sub-loop and feedforward controllers. The capability of the ICS to control the reactor, turbine, turbine bypass valves, feedwater pu=p speed, steam generator level and feed flow, and average reactor coolant temperature was monitored under steady state and transient conditions and sub-loop adjustments were made as required. Unit load changes were perfomed prior to conducting the =a,+ . transient tests to determine the actual relationship between reactor power, feedwater flow and generated megawatts and to set these functions into the ICS feedforward control. The Unit Load Transient Test conducted at 40%, 76: and 100% full power tested ICS response in the three modes of control and at rates up to and including the design ra=p rates. In each case, steady state conditions were first established and data recording comnenced using Brush recorders and the reactimeter. Unit load was then reduced to a specified power level and steady state conditions were again established while turbine header pressure, reactor coolant average te=perature and pressure, and other plant parameters were monitored. Unit Load was then increased to the original level. This procedure was repeated as required until the transient was - performed at the design ra=p rates and opti=um control in each mode was achieved. A similar method was used to perform the transient tests described in sections 8.1.1 (b) through 8.1.1 (f) . Steady stata. conditions were established with the ICS in the fully integrated mode of control and data recording co=menced. The respective component (feedvater pump, reactor coolant pu=p, generator, etc.) was then tripped to start the transient; data recording continued until steady state conditions were once again attained. In those cases where unplanned turbine and reactor trips occurred, data was obtained from the reactimeter and/or the Brush recorders, which were used as plant monitors during the startup program. 8.1-1 }k}k
.. .- 8.1.3 TEST RISULTS Initial testing of the Integrated Control System in preparation for initial plan: operation began prior to escalation in o the power range. The firs set of ad-justments to the syste= was to set all the =odules to :he calculated para =eters and to perform a basic viring check on the cabinets. Si=ulated plan: signals were then used to monitor the response of the control syste= to changes in the plant parameters. This provided a check of ICS response to abnormal plant con-dition withou: affecting the reactor or secondary system. The response of the ICS to all scheduled and unscheduled plant ::ansients and trips was analy:ed. The transient results discussed in this see:1on are su==arized in Table 8.1-1. Based upon the resul:s of the transien: tests, =inor adjustments were made to the ICS during the escalation to full power to provide the best possible response in all modes of operation. The following adjustments were made to the ICS as a result of the startup testing: (a) Adjustments were made to the sub-loop controllers to replace the theoretical relationship between generated megawatts, neu:ron power and feedwater to the measured one. (b) The range in which the ICS controls feedvater pump speed was changed to allow lower speeds and thus bstter control of feef valve differential pressure at lower flow rates. (c) An input from turbine header pressure was added to the feed pump control cir-cuit to increase pump speed rapidly when steam generator pressure rises due to a turbine or reactor trip. (d) The turbine bypass valve control was changed to provide balanced stea= genera-tor pressures. This was required because of the stop-check valve feature of the TMI Unit I main steam lines. (e) Several capacitors were added to the system as required to filter process and ac noise. 8.1.3.1 Uni: Lead Transient Tests Durine Escalation The results of the Unit Lead Transient Test performed a: 40%, 76: and 100: FP are presented in Tables 8.1-2 through 6.1-5 and Figures 8.1-1 through 8.1-6. Table 8.1-2 gives a su= mary of the tests performed at each power level. Tables 8.1-3 and 8.1-4 show the mardmum and minimum values reached for turbine header pressure, RC average te=perature and RC pressure during the 40% and 76: FP test, respectively. The results of the 100: FP transients, are presented in Table 8.1-5 and plot:ed in the figures. The testing at 40: FP maneuvered the uni: in a 55 -45%-55* transient in the fully integrated, turbine following and reactor / steam generator following modes of control at the design ramp rate of 5: FP/ min. The testing at 76: FP was conducted in the same three modes of con: o1 for a 76 76: ramp, again at 5: FP/ min. As can be seen fro: Tables 8.1-2 and 8.1-3, the ICS was able to maneuver the plan: :hrough a 10 load change without exceeding accep:able limi:s on primary and secondary plant parameters. The maximum rate that could be achieved in the reactor / steam generator following mode without exceeding of 50 psi turbine header pressure error was approxi-mately 5.0% FP/ min. 8.1-2 7 1414 :
.s03
.. . - After completion of the 10 load changes at 76: FP, s:cady state conditions were established again and one main feedvater pu=: eas taken off line. Reactor power was then i= creased with only one feedpu=p supplying both stea= generators. The results of this test shoved tha: the flow fro: a single feedwater pu=p would support opera:1on up to 78: FP. The results of the Uni: Load Transient Test performed at 100: FP are shown in Table 8.1-5 and Figures 8.1-1 through 8.1-6. The first part of the tes: maneuvered the plant in a 100%-90 -100 transien: a: 5: FP/ in. These ramps were successfully completed in all three modes of control. Ra=p rates up to 10: FP/ min were used in the second part of the testing a: 100: FP. The lead change ra:e li=1:er on the turbine EHC control was raised-from 6 /=in to 10:/ min as a result of the 100: FP testing. The 100 -50%-100 ra=p was completed without exceeding 140 psi turbine header pressure error and +2 F T AVE error in the fully integrated mode. Maneuvering in the turbine following " mode was acco=plished within acceptable limits at 8: FP/ min. The maximum rate obtained in :he reactor /stea= generator following mode without exceeding a 50 psi TEP error was 6: FP/=in decreasing power and 4: FP/=in on the increase. 8.1.3.2 Reactor Coolant Pume Trip at 25 FP Steady state conditons were established at 26: FP vith four reactor coolant pumps (RCP) operating. The ICS was in the fully integrated mode of control. The a RCP was tripped from this condition and the results are shown in Figure 8.1-7. The unbalanced reactor coolant loop flows resulted in a decrease in heat transferred to the A steam generator and an increase in core average temperature. The ICS corrected this condition by redistributing the total feedvater flow to increase the flow to the A loop. The marimum change in turbine header pressure and T AVE vere less than 120 psi and 15 F, respectively. These results show that the transition from four pu=ps to three pu=ps operating was acco=plished s=oothly by the steam generator ATc controller. 8.1.3.3 Main Feedwater Pu=0 Trie At 100: FP The effects of tripping one main feedwater pu=p while opera:ing a: 100: FP are shown in Figure 8.1-8. Once steady state conditions were established vich the ICS in the fully integrated mode of control, the A main feedwater pu=p was taken off line. Uni: load de=and was run back at 48:/cin and reactor power was reduced to 66: FP at 31.5% FP/=in. Both runback rates agreed well with the design ra:es of 50 / min and 30: FP/cin, respectively. Turbine header pressure error re=ained less than 20 psi. The final power level reached should have been approxi=ately 60: FP. This power level was no: reached until approximately five minutes after the trip, follow-ing a 2\ =inute period in which reactor power increased slightly before stabli:ing at 62: FP. As can be seen fro = Figure 8.1-8, feedwater flow decreased rapidly and then held nearly constan: un:11 the feedwater de=and signal reached the same value during the first two =dnutes of the transient. Feedwater flow then increased and decreased over the next 2h =inutes while the de=and for feedvater re=ained approxi=ately constant. The increase in feedwater flow was attributed in part to the input to the feedvater pu=p speed circui: which increases pu=p speed whenever a positive turbine header pressure error develops. The feedwater cross 11=1: signal, which requires reae:o power to =atch feedwa:er flow within 15%, caused reactor power to follow feedva:er flow. *he ICS corrected :he various error signals tha: develope. m *-
~~
1414 704
.. .a during the transient and feedwater flow matched the de=aad for feed a: 5 minutes after the trip. Reactor power was then decreased to the proper value. 8.1.3.4 Asv=cetric Rod Runback The asymmetric rod runback feature of the ICS was tested during the Droppel Con:rol Rod test at 76: FP (Section 5.7). Steady state conditions were established with the ICS in the fully integrated mode of control and doacrol rod 7 in group 6 inserted 9 inches below the group 6 average position. The results of the transient are shown in Figure 8.1-9. The ICS ran uni: load de=and back to 570
.W(e) a: 12 / min upon detection of the asy=cetric rod by the rod drive control circuit. Reactor power was reduced to 55: FP at approximately 30: FP/ min, which was within the acceptable li=it of 3b2.6% FP/ min. Feedwater cross limits developed
~
at about 2 minutes into the transien: vhich 11=ited the overall runback rate to ll: FP/ min. 8.1.3.5 Turbine Trip Fron 76: FP An inadvertant turbine trip occurred during a simulated turbine overspeed test with the reactor at 70; FP. The effects of this transient on selected system parameters are shown in Figure 8.1-10. Tripping the turbine caused the generator to trip and initiated a reactor power runback to 14 FP at 18.7% FP/ min. The rapid closing of the turbine stop valves opened the turbine bypass and main steam safety valves to relieve secondary side pressure. Main steam pressure reached a marimum value of 1062 psig and was subsequently controlled at 895 psig, as required. The maximum values for reactor coolant average te=perature and pressure were 5900F and 2320 psig, respectively. Pressurizer level reached a low value of 200 inches and a high of 296 inches. The ICS established level control in the steam generators, as required, in a power reduction to 15: FP. At the time of the turbine trip, the sudden increase in steam generator pressure crused a rapid decrease in feedwater flow, as is seen in Figure 8.1-10. The demand for feed flow re=ained high due to Btu limits on the generators. Feedwater flow recovered and remained spproximately constant from to 2 minutes into the transient, but was higher than the demand for feed flow. This caused a hold in power runback at 45: FP to correct the negative T AVE error which developed. Feed flow then followed feedvater de=and and power was : hen reduced to the proper value. 8.1.3.6 Generator-Reactor Trio From 100: TP A generator trip from 100: FP was conducted as part of the tes program. The tran-sient was initiated by opening the main generator breakers af ter establishing steady sta:e conditions with the ICS in fully au:omatic con:rol. The ICS started to run back uni: load demand a: 20 / min and the :urbine began to overspeed due to loss of load. The turbine governing valves started to close and turbine header pressure increased to 1040 psi. The turbine bypass and =ain stea= safety valves opened. Reactor coolant pressure increased to the high pressure li=it and the reactor tripped at approrincely 4 seconds af ter the generator breakers opened. The turbine tripped due to the reactor trip. Figure 8.1-11 shows the behavior of selectec primary and secondary system parameters for this transient. S.1-4 3414 .,0r;
.
e a . Af ter the reactor trip, turbine header pressure was controlled at 101.- psi, as required, and level control was established in the stea= generators. Reactor coolant average :e:perature increased to 5810F and then decreased af ter the reactor tripped. The behavior of feedwater flow af:er the trip is due to several error signals received by the ICS. The increase in turbine header pressure caused Btu 11=its to override the feedvater demand signal and the demand signal followed the Btu limits for several seconds after the ::1p. Feedwater pump speed decreased to minimu= speed due to the drop in feedwater de=and. After neutron power dropped to zero, the neutron error modifica: ion to the feedwater demand was e m ved and feedwater demand followed the Btu limits in a reduction of feedwater flow. 8.
1.4 CONCLUSION
S An analysis of scheduled and unscheduled transient results have led to several minor modifications to the ICS during the startup test program to optimize ICS performance. Transien: testing of the unit was conducted at the design ramp rates in the turbine following, reactor /stea= generator following and fully integrated mode of control at 40%, 76: and 100% full power. The ICS successfully maneuvered the plant in all three nodes of control at 40% and 76: FP and in the fully integrated mode at 100: FP. The 100%-50%-100% load sving was completed at 8 /cin in the turbine following mode and at 6:/ min on the decrease and 4 / min on the increase in the reactor / steam generator following mode. The ability of the ICS to control the plan: upon a loss of reactor coolant pump, a loss of main feedwater pu=p, a dropped control rod and a turbine trip at 76: IT was demonstrated. The reactor tripped on high RC pressure af ter a load rejection at 100% full power. Studies are in progress to determine the necessary corrective actions required to keep the reactor on line following a load rejection. 34)4 ?C6 8.1-5
.
.
.
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SUMMARY
OF TRANSIENTS REPORTED IN SECTION 8.1 Transient Data Presented in Number 8
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Description l-3 Unit I,oad Transient Test at 40% FP 8.1-3 X 4-6 Unit I.oad Transient Test at 76% FP 8.1-4 X 7-12 IInit I,oad Transient Test at 100% FP 8.1-5 8.1-1 to 8.1-6 13 Reactor Coolant Pump Trip at 25% FP X 8.1-7 g 14 Main Feedenter Pump Trip at 100% FP X 8.1-8 P' 15 Asymmetric Rod Runback at 76% FP X 8.1-9 Y " 16 Turbine Trip at 76% FP X 8.1-10 17 Generator-Reactor Trip at 100% FP X 8.1-11 s M
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SIDDIARY OF UNIT LOAD TRANSIE!Tr TESTS AT POWER Load Change Test Accept Transient ICS liode of Decrease Increase Ramp Rate Rate Number Operation (% FP) (% FP) (%/ min) ( min) _%/ A. Unit Load Transient Test at 40% Full Power 1 Fully Integrated 55-45 45-55 1 to 5 5 2 Turbine Following 55-45 45-55 1 to 5 5 3 Rx/SG Following 55-45 45-55 1 to 5 5 H D. Unit Load Transient Test at 76% Full Power D 4 Fully Integrated 76-66 66-76 1 to 5 5 N 5 Turbine Following 76-66 66-76 1 to 5 5 m 6 Rx/SG Following 76-66 66-76 1 to 5 5 'In C. Unit Load Transient Test at 100% Full Power 7 Fully Integrated 100-90 90-100 1 to 5 5 8 Turbine Following 100-90 90-100 1 to 5 5 9 Rx/SG Following 100-90 90-100 1 to 5 5 10 Fully Integrnted 100-50 50-100 1 to 10 10 11 Rx/SG Following 100-50 50-100 1 to 10 10 12 Tarbine Following 100-50 50-100 1 to 10 10
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4 e SIMIARY OF UNIT IAAD TRANSIENT TEST RESUI.TS AT 40% FP Power Transient ICS Mode of Change Rate, Z/ min AT AVE, "F ATc, F ATilP, pn1 Number Operation (% FP) Ave Max Hin Max Illn Max Hin Max 1 Fully Integrated 54-44 4.3 4.3 -1 G -0.2 10.2 0 +24 45-55 4.2 4.2 0 +0.5 -0.2 10.2 -30 0 2 Turbine Following 54-45 4.1 4.1 -1.5 40.5 -0.2 40.2 -6 46 45-55 4.2 5.4 -0.5 40.5 -0.2 40.2 -8 48
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SUIDIARY OF IJNIT LOAD TRANSIEllT TEST RESULTS AT 76% FP Power Transient ICS tiode of Change Ra'te, %/ min AT AVE, F ATIIP, psi thunber Operation (I FP) Ave Max Hin Max Hin Max 4 Fully Integrated 70-58 5.5 5.5 -0.5 0 0 +24 58-71 5.9 5.9 0 +1.0 -24 0 5 Turbine Following 68-58 3.8 3.8 -2.0 0 0 0 56-74 4,1 4.4 -0.5 +t.0 -6 0 [ ' l,; 6 Rx/SC Following 74-63 63-74 1.8 2.4 3.8 6.3
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OF UNIT LOAD TRANSIENT TEST RESULTS AT 100% FP Power Transient ICS Mode of Change Rate, %/ min AT AVE, F ARC Press, pel ATIIP, psi Number Operation (% FP) Ave Max Hin Max Hin Max Hin Max A. Ramp Rates at 5%/ min, 100%-90%-100% 7 Fully Integ;ated 100-88 6.0 6.7 -1.1 40.6 .28 +20 -2 417 88-100 6.5 6.5 -0.4 40.6 -4 +16 -16 +2 8 Turbine Following 98-85 3.9 7.4 -1.6 40.4 -20 448 -3 42 85-100 5.0 8.0 -1.0 +1.0 -4 .: ' 6 -1 +2 o D N 9 Rx/SG Following 100-87 5.6 12.0 -1.2 40.8 -8 448 0 440 P 87-98 5.7 15.0 -0.6 +1.2 -4 +16 -20 +12 Y v. B. Ramp Rates at 10%/ min, 100%-50%-100% 10 Fully Integrated 99-56 9.1 15.4 -2.0 40.6 -40 448 -12 416 55-91 10.8 10.8 -0.8 +1.0 0 +56 -36 416 11 Rx/SG Following 94-62 5.2 12 -1.6 40.1 -16 440 0 +76 62-91 4.9 14.4 -0.8 40.4 -8 +20 -72 0 _. - 4== __. 12 Turbine Fo11ceing 98-58 8.3 11.4 -1.6 40.6 -40 448 -12 +12 a=, 58-91 10.4 10.4 -3.8 41.4 -20 +96 -48 +24
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- 1) The automatic response of the reactor and auxiliary systems results in the plant being controlled in such a manner ac to prevent fuel damage and excessive pressure in the reactor coolant system.
- 2) The plant can be shutdown with power supplied from the station batteries and emergency diesel generators.
- 3) The emergency procedure to be followed during a Station Blackout Emergency (EP 1202/027 is a satisfactory document for verifying automatic actions and per-forming manual actions to maintain the plant in a safe condition and prevent equipment damage.
8.2.2 TEST METHOD The normal plant electrical lineup consists of the turbine generator output current passing tmugh voltage step up transformers and the_ through the generator breakers to the substation. The substation is of the " breaker and a half" design. Plant house load is supplied from the substation through breakers to two auxiliary transformers, either one of which is designed to supply 100% of house load. Prior to the test, all house load was placed on the B auxiliary transformer. The breaker between the substation and the A auxiliary' transformer was opened and placed in the " pull-to-lock" position. Reactor power was brought to 15". With this alignment, all incoming power to the plant could be cut off by opening the breaker between the substation and the B auxiliary trans former: if trouble developed, (for exa=ple. if the diesel generators failed to start up and assume load), the A auxiliary transformer could be imediately energized to supply required load by taking its feeder breaker out of " pull to lock". Brush recorders and the reactimeter were set up to monitor primary and secondarv plant parameter behavior and an equipment status log was kept to record the status (running or off) of equipment subject to automatic action following the loss of power. RCS pressure was monitored by a 0-3000 psi, 0.12 accuracy Heise gage and compared with the acceptance criteria of <2750 psis. The ::end of plant fission product activity was monitored before and af ter test performance to deter =ine if any fuel damage occurred during the test. Emergency Procedure 1202/02 (Station Blackout) was followed to verify all automatic actions and to perfo:n all required manual actions 'to maintain the plan in a safe shutdown condition and keep from da.taging equipment. 8.2.3 TEST REST 1TS All required automatic actions following the loss of power at 152 occurred as ex-pected and listed in EP1202/02. They are:
- 1) reactor trips
- 2) turbine generator trips
- 3) control room DC lighting cemes on
- 4) steam driven emergency feedvater pumo starts and raises stea= generator levels to 952 on operating range 8.2-1 -
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- 6) condensate, condensate booster and feedvater pu=ps trip
- 7) generator DC cil pu=p, RC pt=:p DC of.' pu=os and DC lube oil pu=ps start
- 8) A and B diesel generators start anc pick up block 1 (nen safeguard) leads EP1202/02 was satisfactorily verified for use during occurrence of a Statiet Bl .kout.
Both RCS te=perature and pressure decreased af ter the blackout; therefore, peak RCS temperature was the initial hot leg te=perature of 582 F and peak RCS pressure was the initial valut of 2164 psig. Minimum RCS pressure was 2004 psig just prior to normal per.'er restoration. Pressurizer level also decreased af ter the blackout from an initial value of 231 inches to a low point of 120 inches just prior to normal power restoration. Main steam peaked at 1008 psig in the A header and 1032 in the B header. As a result of this pressure differential, emergency feedvater preferentially went to the A steam generator; therefore the A steam generator lew1. increased rapidly while the B stea= generator level slowly decreased to 20 inches en the startup range level instrument and held there. In order to keep from ex-ceeding the cooldown rate limit of 100F /hr, both emergency W valves tre placed in manual control and closed when the A steam generator level mehed 28% on the operating range. At this point, the valve to the B steam generator should have been opened to bring. its level up even to the A, bta this was not rkna. Had the valves been left in auto, the valve feeding the A steam generator voud '.ve throttled devn at 95 and the valve feeding the B steam generator would have re-nained open until its level rose to 95* also. Following this test, B&W evaluated the total amount of natural circulation flow through the core to remove decay heat as a function of steam generator level and reduced the post blackout control level setpoint trom 95 to 50" on the operating range. Review of our test data iustifies this change, since RCS pressure and te=perature never increased above initial steedy state values af ter the blackout, and final control levels were 28: in one steam generator and 20 inches in the other. Radiochenical analyses before and af ter the blackout indicated no increase in fission product activity and therefore no failed fuel as a result of test perfor-cance. 8.
2.4 CONCLUSION
S The Station Blackout Emergency Procedure (IP1202/02) was verified for use during a blackout and all required automatic actions occurred as exnected. RCS pressure and te=cerature never increased above their initial steadv state values. Peak pressure of 2164 psig was far less than the 2750 p.ic upper limit. There was no increase in RCS fission product activity as a result of test perfornance. The emergency feedvater supoly valves were placed in manual and closed before steam generator levels increased to 95: to keep from exceeding the 100F /hr cooldown limit. 3&W has since revised the 95 setpoint to 50" to avoid exceeding the cooldown 11=it. Prior to EW valve closure, the steam generators were being fed unevenly due to the difference between A and B header pressure. Had the EF;; valves been lef t in auto, the supply valve to the A OTSG would have throttled down when A level reached 95: and the B OTSG vould have increased to 95 .
- 8. -2 1414 124
. = 8.3 SHUTDOWN FROM OUTSIDE THE CONTROL ROOM 8.3.1 PURPOSE The purposes of the shutdown from outside the control roo= test were to demon-strate the ability to remove decay heat from the reactor coolant syste= with control of all systems at locations remote from the control room and to verify porrions of Emergency Procedure 1202/37 (Cooldown from Outside the Control Room). 8.3.2 TEST METHOD Emergency Procedure 1202/37 (Cooldown fro = Outside the Control Room) provides instructions for shutting the plant down from power operation and cooling it down to 140 F in the event the control roo= must be evacuated. The procedure provides steps for the entire shutdown to be performed from outside the control room, but lists those actions which should be performed in the control room prior to evacuation as time per=its. Also, considerations regarding nuclear safety override considerations regarding potential equipment damage in shutting down and cooling down the plant. A remote control center has been established where sufficient instrumentation and communications channels exist for monitoring and directing a safe plant shutdown and cooldown. Equipment actuation is performed at various locations throughout the plant. In light of the above remarks, we established the following bases for perfor=ing the shutdown from outside the control room test:
- 1) Up to two minutes could be spent in the control room to perfom initial steps of the e=ergency procedure prior to evacuation.
- 2) The emergency procedure should be modified slightly to minimize the potential for equipment damage during the test.
- 3) Demonstration of the capability to bring the plant to hot shutdown conditions (Tave >525 F and at least 1% ak/k shutdown) is sufficient to satisfy the requirements of USAEC Regulatory Guide 1.68 (Pre-op and Initial Startup Test Programs for Water Cooled Reactors), so only those steps of the emergency procedure would be performed.
- 4) The test would be performed utilizing only the nu=ber of personnel scheduled on a nor=al shift to conduct operations; however, there would be additional personnel to act as observers and to take over operation from the control roo=
in the event assistance was required. Using these bases, the test was perfor=ed as follows:
- 1) Prior to evacuating the control roo=
a) the reactor was manually tripped b) both motor driven amergerev feedwater pu=ps were started c) both main feedwater pu=ps were tripped d) the letdown block valve was closed to ter=inate letdown flow e) the nuclear and turbine plant co==unications channels were cross tied f) the concentrated boric acid emergency injection valve was opened and both injection pu=ps were started. 1h\
~
o . ., ,
,,
_. _ . . . . . . e
- The emergency procedure was modified such that the reactor coolant pu=ps were not tripped (to keep from exceeding the 1007 /hr cooldown limit) and the operating makeup pu=p was not tripped (:o continue the supply of RC pump seal injection water to prevent damaging the seals) .
- 2) Following the above actions, the Shift Foreman announced the control room evacuation over the PA system. Then he and a control room operator pro-ceeded to the remote control center to monitor and direct plan: shutdown.
One of these men was available to turn pressurizer heaters on and off as required to maintain RCS pressure. Parametric indication available at the remote control center vas: a) pressurizer level b) makeup tank level c) RCS pressure d) RCS temperature e) both OTSG levels
- 3) Another operator proceeded to the che=ical addition control panel where he shutdown the concentrated beric acid injection pu=ps and then to the makeup valve manifold in the auxiliary buildingwhere he controlled pressurizer level and makeup tank level as directed by the Shif t Foreman.
- 4) Another operator proceeded to the intermediate building to the emergency feedvater valves to control steam generator levels between 5: and 95% on the operating range as directed by the Shift Foreman.
- 5) A second operator proceeded to the internediate building to the atu rpherie dump valves to control RCS te=perature as directed by the Shif t Foreman.
- 6) The Shif t Foreman made calculations to verify that the reactor was 12 Ak/k sub eritical.
The test was tereinated after stable conditions were maintained for twenty minutes. 8.3.3 TEST RESULTS The actions performed in the control roon prior to evacuation took one =inute and ten seconds to perform co=psred with the two minutes allotted. The remote control center instrumentation and co:=:unications proved adequate to monitor and direct plant operations to maintain the reactor plant at hot shutdown conditions. RCS pressure was maintained at =2155 psig without operation of the pressurizer heaters. Both pressurizer level and makeup tank level were adecuately raintained manually from the takeup valve manifold. Stea= generator levels were maintained between 5% and 95 by regulating the emergency feedvater valves in manual. down rates. Levels were kept near their initially low values to avoid excessivg cool-RCS te=perature was satisfactorily decrer. sed from 579 F to 527 F by controlling the atmospheric du=p valves in canual. The shutdown =argin was cal-culated to be 2.43: even after xenon reactivity decayed to zero. The only proble: encoun:ered during the test was the plugging of the concentrated boric acid pu=p suction strainer with boron crys:als and small pieces of the paper bags the boron is shirped in. A field change is being made to include redundant parallel hea: :: aced strainers to eli=inate :his preble=. S.3-2 3 p A '26 _ _ _ _
- - - -
. e The reactor plan: was satisfactorily maintained in a hot shutdown condition for greater than the required twenty minutes by the minimum shif: ce=plement of five men. 8.
3.4 CONCLUSION
S The reactor plant can be maintained in a safe hot shutdown condition from loca-tions outside the main control room by the ninimum shif: complement of five men. The alternate control center contains sufficies: instrumentation and communica-tions to permit satisfactory monitoring and direction of shutdown operations, Y h h h q }~/
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'
8.3-3
. e 8.4 UNIT ACCEPTANCE TEST r
8.4.1 PURPOSE The purpose of the unit acceptance test was to verify that the energy output from the nuclear steam supply syste= meets or exceeds the equivalent of 10,521,000 lb:t/hr steam flow at the steam generator outlet noceles at conditions of 925 psia and 569 7 when supplied with feedwater at conditions of 45507 and 1030 psia at the inlet to the steam generators, with a pri=ary coolant letdown flow of 55 gpu and makeup water supply temperature of 125 7. This energy output corresponds to a gross secondary side output of >2449 W t. 8.4.2 "T.ST ETHOD This test was conducted in accordance with the provisions set forth in the ASE Power Test Code 4.1-1964, Stea= Generating Units. It was performed as follows:
- 1) Prior to the test, an agreement was reached as to the specific instrunenta-tien to be used for recording test parameters.
- 2) The above instruments were calibrated within two weeks of test performance.
- 3) The test consisted of a preliminary four hour run during which data was re-corded every ten minutes. The data was then averaged over the four hour in .
terval.
- 4) The four hour data averages were then used to datermine the values for use in the following equation:
E t
=
WA (HAg - Egy) + WB (H BS - HBF) 3,412,142 Bru/st where We = gross NSS megawatts thermal output WA"l p A feedwater flow in ibm /hr HAS " lo p A main steam enthalpy in Bru/lbe HAF = 1 op A feedwater enthalpy in Btu /lbe WB"l p B feedwater flow in Ibm /hr HBS " lo P B main steam enthalpy in Btu /lbe HBF = 1 OP B feedwater enthalpy in Bru/lbe
- 5) Since the s calculated e above far exceeded 2449 W e, the preliminary run was accepted and therefore qualified as the first official run.
- 6) A second four hour run was made in the same mnner as 3) above and evaluated as in 4) above.
8 '-l 1414 223
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8.4.3 FST RESULTS Average gross NSS megawatts ther=al on:put during the firs: four hoar run was 2551.016 W. average for the second four hour run was 2554.213 W.. The average for'the entire unit acceptance test was 2552.615 E gas ce= pared wi:h the acceptance criteria of >2449 W g. Average loop A main stea: te=perature was 591.4 F and average loop 3 main steam temperature was 591.80F co= pared with the acceptance criteria of 15690F. Average main s:eam flow was 10,770,000 lbm/hr a: 591.6 F with feedwater at 461.10F. Gross electrical megawatts generated was 848.2 W , throughout the unit acceptance test. Condenser vacuu= vas lower than design values, which accounts for the lower than expected (870 W,) electrical output. The unit acceptance test was cr:cpleted on August 26, 1974, which completed the TMI Unit I power escalation tes: program. 8.4.4 CONCI.USIONS The Three Mile Island Unit I nuclear steam supply syste= produces 2552.615 W g gross energy output comparec. with the warranty value of >2449 Wg . Main steam te=perature is 591.6 F compared with the warranty of 1569 F. These results indicate a substantial margin of NSS performance above warranty specifications. 1414 729 \ 8.4-2}}