ML19261B359
| ML19261B359 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/30/1979 |
| From: | Baynard C, Beatty G FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19261B357 | List: |
| References | |
| NUDOCS 7902150391 | |
| Download: ML19261B359 (95) | |
Text
{{#Wiki_filter:_ _ _ _ . . ! o Crystal River Unit 3 Nuclear Generating Plant CYCLE 1 MID-CYCLE STARTUP REPORT DECEMBER 1978
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DOCKET NO. 50 - 302 LICENSE NO. DPR - 72 77o2/5o39/
FLORIDA PO'n'FR CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 LICENSE NO. DPR-72 CYCLE 1 - MID-CYCLE STARTUP REPORT APPROVED BY: Nuclear Plant Manager Date
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- Manager Nuclear Support Services Dat , , , um ) , /k7f y -)
DECEMBER 1, 1978
TABLE OF CONTENTS 1.0 Introduction and Summary 2.0 Precriticality Testing 2.1 Control Rod Trip Test 2.2 RC Flow and Flow Coastdown Test 2.3 Chemical and Radiochemical Tests 2.4 Pressurizer Effectiveness 2.5 In-Service Loose Parts and Vibration Monitoring System 2.6 Calibration and Neutron Response of Source Range Monitoring 3.0 Zero Power Tests 3.1 Initial Criticality 3.2 NI Overlap 3.3 Sensible Heat 3.4 Reativity Calculations 3.5 Control Rod Group Worth 3.6 Soluble Poison Worth 3.7 Ejected Rod Worth 3.8 Temperature Coefficients of Reactivity 3.9 Biological Shield Survey 3.10 Effluent and Effluent Monitoring 3.11 Chemical and Radiochemical Tests 4.0 Power Testing 4.1 Reactivity Coefficients at Power 4.2 Unit Heat Balance 4.3 Core Power Distribution Test 4.4 Biological Shield Survey 4.5 Pseudo Rod Ejection Test 4.6 Turbine / Reactor Trip Test 4.7 Integrated Control System Test 4.8 Unit Loss of Electrical Load 4.9 Unit Load Steady State Test 4.10 Unit Load Transient Test 4.11 Shutdown From Outside the Control Room 4.12 Loss of Offsite Power 4.13 Power Imbalance Detector Correlation Test 4.14 Nuclear Instrumentation Calibration at Power 4.15 Feedwater System Operation 4.16 Turbine / Generator Operation 4.17 Dropped Control Rod Test 4.18 Incore Detector Test 4.19 RCS Hot Leakage Test 4.20 Pipe and Component Ranger Hot Inspection at Power 4.21 Chemical and Radiochemical Tests 4.22 Effluent and Effluent Monitoring
1.0 INTRODUCTION
AND
SUMMARY
1.1 INTRODUCTION
On March 3, 1978, Crystal River Unit 3 was shut down to investigate a noise in the "B" Steam Generator. This investigation revealed a spider coupling assembly from a lumped burnable Poison Rnd Assembly plus 16 other pieces of BPRA on the upper tube sheet. The entire Reactor was then defueled and debris was removed from the Steam Generator Reactor Coolant Piping, Reactor and Fuel Assemblies. The remaining BPRA and most Orifice Rod Assemblies were removed from the core. Two modified Orifice Rod Assemblies were left in the core to secure the startup neutron sources. During Core reloading, a test weight was inadvertently dropped on a Fuel Assembly in the Spent Fuel Pool. This assembly was subsequently declared unfit for reinsertion. Four once-burned Fuel Assemblies from Oconee Unit 1, Cycle 1 were obtained from Duke Power Company to replace the damaged assembly and three symmetric assemblies. The revised core loading pat t e rn , shown in Figure 1.1-1, was verified with an underwater television camera at the completion of fuel loading. This report, prepared and submitted in accordance with Technical Specificaton 6.9.1, describes the subsequent precriticality, zero power and power testing performed and the results obtained. 1.2
SUMMARY
Crystal River Unit 3 was brought to Hot Standby condition. Control Rod Trip Testing, Reactor Coolant flow and Flow Coastdown testing, loose parts and vibration monitoring testing and source range testing was conducted. The test methods used and results achieved are reported in Section 2.0. Criticality was achieved on September 15, 1978, at 0637 and Zero Power Physics Testing was performed. Results for Zero Power Phyc;:s Tests are reported in Section 3.0. All acceptance criteria were met. Zero Power Physics testing was followed by Power Escalation Testing. The Generator was synchronized on September 18, 1978 at 1700. As Reactor power level was increased, the Incore Detector system began to indicate a significant flux tilt. During the resulting investigation, several possible causes were identified and considered. It was concluded that the most probable cause was an uncoupled control rod. One of the group 5 control rods was later confirmed to be uncoupled from its lead screw, and the plant was shut down for repair. Af ter return to criticality, Zero Power Tests for the worth of control rod groups 5 and 6, and the ejected worth were repeated. These tests were repeated to ensure that previous measurements were valid. The Generator was again synchronized on September 29, 1978, and Power Testing proceeded normally. The test methods used, and the results achieved are reported in Section 4.0. All results were satisfactory.
Figure 1.1-1 Core Loading Diagram for Remainder of Crystal River 3 Cycle 1 l A B C 1A36 3A53 3A32 1A05 D Oco-1 D-4 D12 3co-1 E F G II - . K L M N 1A04 3A05 3A04 1A01 Oco-1 N-4 N-12 L)c o -1 0 P R l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Note: Only shuffled assemblies are shown; all other assemblies are in their original evcle I positions. xxxx Fuel Assembly Ident. No. xxx Previous Location
2.0 PRECRITICALITY TESTING 2.1 REACTOR COOLANT FLOW AND FLOW COASTDOWN TEST Prior to reactor criticality, the reactor coolant pump flow and flow coastdown test was performed to ensure the functional capabilities of the reactor coolant cystem and the reactor coolant pumps during steady state flow and flow coast-down conditions. This test was conducted at hot conditions of 532*F and 2155 Psig with the core installed. The intent was to measure the normal flow rate with four reactar coolant pumps operating and the flew coastdown and time delay characteristics for the loss of four reactor coolant pumps from the normal operating conditions. 2.1.1 PURPOSE The purpose of the reactor coolant flow and flow coastdown test are listed below: (a) To measure the reactor coolant flow and flow coastdown characteristics at 532*F and 2155 Psig with four reactor coolant pumps operating and the core installed. (b) To compare reactor coolant flow with design requirements and to verify adequate core flow. (c) To verify that the reactor coolant system flow instrumentation response time is within acceptable limits. Four acceptance criteria are specified for the reactor coolant flow and flow coastdown test as listed below: (1) Steady state total reac*or coolant system flow with the core installed, and orifice rods and burnable poison rods removed, when corrected to 532*F and 2155 Psig shall be within the limits specified in Table 2.1.1. (2) With the reactor core installed, the snubbered RC flow signal must decrease to 97.6% of the minimum observed steady state flow rate within 2.0 sec. after the trip of all four RC pumps. (3) The reactor coolant system flow when corrected to 532*F and 2155 Psig, must be greater than or equal to the flow versus time relationship in Figure 2.1-1. (4) With all four reactor coolant pumps in operation, steady state loop flows shall be within 2 percent of each other. 2.1.2 TEST METHOD Reactor coolant steady state flows were determined by means of the temperature compensated loop flow transmitter voltages input to a Babcock and Wilcox supplied reactimeter and converted to loop flows in millions of pounds per hour (MPPH). With four reactor coolant pumps operating, ten sets of steady state coolant ficw data were taken on the reactimeter and the values averaged. These results were then corrected to reference conditions of 532*F and 2155 Psig using the equation
below: Fr = Fm (VM/Vr) Where: Fr = Flow Rate At Reference Conditions, MPPH Fm = Flow Rate At Measured Conditions, HPPH Vr = Specific Volume At Reference Condition, Ft3 /id, Vm = Specific Volume At Measured Conditions, Ft 3/lb. The measured flow rate corrected to the reference conditions was then compared to the respective acceptance criteria. Reactor coolant flow coastdown versus time was also determined from the temperature compensated loop flow transmitters. Steady state data was obtained with four reactor coolant pumps in operation. Subsequently, all of the pumps were tripped and data was recorded during the ensuing reactor coolant flow transient. The measured reactor coolant flow at various times during the coastdown transient was then plotted versus time and compared to the acceptance criteria. The snubbered flow signal ree-en:- was determined for the reactor coolant flow coastdown. Data was collect ao for the initial 2 second period of the coastdown. The final flow at the end of this period, expressed as a percentage of the minimum o) served steady state flow, was then compared to acceptar.re criteria. 2.1.3 EVALUATION OF TES T RESULTS TabIg 2.1-1 gives the mintien and maximmn allowable flow rate along with the measured flow rate for fou. reactor ccolant pump operation. It can be seen that the measured flow rat s is well wit hin the acceptance criteria, and that the flaw imbalance with fe ar reactor p tops was also well within the acceptance criteria of 2 percent. Figure 2.1-1 shows the minimum acceptab;e reactor coolant flow rate versus time for the four pump coastdown. Also included is the measured reactor coolant flow versus time for the coastdown of four reactor coolant pumps from an initial steady state o perating condition. It can be seen that during the four pump coastdown the f low remained above the acceptance curve. Snubbered flow signal res ponse during the four reactor coolant pump trip from an initial steady state operating condition is tabluated in Table 2.1-2. The data taken indicates that the resulting decrease was within the acceptance criteria. 2.
1.4 CONCLUSION
S The measure reactor coolant flow rate is approximately 113.2% of the design value. This flow provides adequate margin to both the maximum and minimum allowable flow rates. Snubbered flow signal time response was within the acceptance criteria. The measured reactor coolant flow rate versus time for a loss of our reactor coolant pumps was above the minimum allowable flow coastdown curve. Loop flow imbalance was well within the 2% mismatch criteria.
Comparison Of Measured Reactor Coolant Flow Rate To The Acceptance Criteria At Reference Conditions Of 532*F and 2155 Psig Minimum Maximum Maximum Measured RC Acceptable Acceptable Measured Acceptable Loop Flow Pump Flow Rate Flow Rate Flow Rate Loop Flow Difference Combination (1061bm/hr) (1061bm/hr) (1061bm/hr) Difference (%) (%) Four Pumps 145.7 155.8 153.3 2.0 0.65 Note: Design core flow rate is equal to 135.4 X 106 1bm/hr at 532*F and 2155 Psig. U a 7
Snubbered RC Flow Signal Response Following A Simultaneous Trip Of Four Reactor Coolant Pumps Corrected Corrected Total Time Flow Loop Flow Loop Corrected (Sec) A (mlbs/hr) B (mlbs/hr) Flow (mlbs/hr) 0.0 76.9 76.8 153.7 0.2 76.9 76.8 153.7 0.4 76.8 76.6 153.4 0.6 76.7 76.6 153.3 0.8 76.7 76.6 153.3 1.0 76.8 76.4 153.1 1.2 76.5 76.0 152.5 1.4 76.2 75.4 151.6 1.6 75.8 74.6 150.4 1.8 75.3 73.8 149.1 2.0 74.8 72.6 147.4 A. Minimum observed steady state total corrected flow = 151.67 mlbs/hr. B. Flow at 2.0 secs. as percentage of ninimum observed steady state flow = (147.4/152.67) X 100 = 96.!% Table 2.1- 2
Measured Reactor Coolant Flow Rate 400 Following The Trip Of Four Reactor Coolant Pumps From Initial Steady State Conditions of 532*F and 2155 390 Psig 380, 370 Measured Four Pump Trip Flow Coastdown m 360 7 x 350 m Four Pump
$ Trip Flow 340 Coastdown ]o Criteria w
u 330 5 1 0 320 w 3
% 310 E
3 300 3 290 4-T 0 1 2 3 4 5 ___ Iime After Trip (Sec.) Figure 2.1-1
2.2 CONTROL ROD DRIVE DROP TI!E TEST 2.2.1 PURPOSE The purpose of the Control Rod Drive Drop Time Test was to verify the functional trip capacity of the Control Rod Drive System. This was done by verifying the following acceptance criteria: (a) To serify, for each control rod assembly, that the total elapsed drop time from the initiation of the trip signal until the control rod assembly was three-fourths inserted, was less than or equal to 1.66 seconds with Tavg > 525*F and all reactog coolant pumps operating. 2.2.2 TEST MITHOD The Control Rod Drive Drop Time Test was p.rformed 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 pushbutton. A zero time signal was f urnished 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 the position indicator tube of each control rod drive. The test was conducted at the following conditions: Flow Temperature Mode Four Pumps >525*F Hot Standby 2.2.3 EVALUATION OF TEST RESULTS The results of the rod drop times are presented in Tables 2.2-1 and 2.2.2. As can be seen, the fastest dropped control rod was L-8 and the slowest dropped control rod was C-II. Analysis of the drop times showed that the slowest dropped control rod yielded a drop time of 1.31 seconds which is well within the accept _nce criteria of 1.660 seconds. Subsequent to the initial rod drop time tests, rod C-11 was discovered to be uncoupled from its lead screw. The plant was then shut down and rod C-Il was recoupled and drop tested. The new drop time was 1 21 seconds with the new drop time for this rod accounted for, the slowe.t dropped control rod becomes M-09 with a time of 1.26 seconds. 2.2.4 CONCLUS IONS The rod drop times were well within the acceptance criteria stated in section 3.1.3.4 of the Technical Specifications of 1.660 seconds at full flow and greater than 525 *F conditions. It was also discovered that an uncoupled control rod will result in a slower drop time for the af fected drive but the effect will be only on the order of +0.1 seconds.
2.3 CHEMICAL AND RADI0 CHEMICAL TESTS Chemical and Radiochemical testing was not performed during this startup. These tests were conducted at initial startup and no plant modifications have been made which would invalidate the results of those tests. 2.4 PRESSURIZER EFFECTIVENESS No pressurizer effectiveness testing was done during this startup since no modifications had been made to the pressurizer and results of the testing done at initial startup are still valid. 2.5 IN-SERVICE LOOSE PARTS AND VIBRATION MONITORING SYSTEM Upon starting of the first Reactor Coolant Pump and each pump thereaf ter during plant heatup, the Reactor Coolant system was monitored for indications of loose parts with the Loose Parts Monitoring System (LPMS). Baseline LPMS data of the Reactor Coolant System at varicus power levels was acquired during the zero power physics and power escalatien testing s eq ue nce . 2.5.1 P URPOSE The purpose cf the reactor coolant system startup noise monitoring test is: (a) To identify those points during plant startup at which data acquisition is required to establish baseline data for subsequent plant evaluation as performed using the Loose Parts Monitoring Sy s tem. (b) To identify those evolutions during plant startup requiring monitoring for loose parts in the reactor coolant system (RCS). Two acceptance criteria are specified for the reactor coolant system startup noise monitoring test as listed below: (1) No impacts greater than four volts peak-to peak with an "FS Range" switch setting of "10". (2) No impacts of any magnitude assessed to have a safety or damage potential. 2.5.2 TEST METHOD The Nuclear Steam Supply System was continuously monitored for indications of loose parts throughout the startup physics program. At specific hold points, additional monitoring, evaluation and recording was completed as directed by PT-102, "RC system Startup Noise Monitoring". The required hold points were as listed below: a) Af ter each RC pump start to " Hot Standby" conditions b) 20-30% Reactor Power - four RC pumps c) 40-50% Reactor Power - four RC pumps d) 70-80% Reactor Power - four RC pumps e) 90-100% Reactor Power - four RC pumps At each plateau specified, relevant plant parameters were recorded and all active LPMS channels were reviewed, both audibly by speaker and visually by displaying the incoming signal on an oscilloscope. In both ins tances , the
selected channels were reviewed for indications of signal quality and component impact, if any. The results of this review were logged and all active channel signals were tape recorded and retained as baseline data for later comparison. 2.5.3 EVALUATION OF TEST RESULTS No significant loose parts indications were detected which produced impacts of damage potential. Although indications of potentially loose parts were detected on channels IGT-1 and IGT-2, the levels if impact energy were below the alarm set point of 4 g's (pk to pk). All other channels exhibited normal background noise signals. 2.
5.4 CONCLUSION
S The Neclear Steam Supply System was free of any loose parts with significant damage potential during the Startup Physics Test Program. 2.6 CALIBRATION AND NEUTRON RESPONSE OF SOURCE RANGE MONITORING Precritical testing was completed following response checkout of the neutron source range detectors. The source range instrumentation was verified to provide a count rate of more than 2 counts /sec. A statistical test was performed to determine that the detector response data followed a normal distribution.
Control Rod Drive Drop Times For The Fastest And Slowest Control Rod Pumps Mean Drop Fastest Rod Slowest Rod Running Time (sec) Location Time (sec) Location Time (sec) 4 1.239 L-8 1.200 C-11 1.310 (1.237) (M-9) (1.260) Y m P n Note: The mean drop time is the average of all 60 control L rod drive drop times. The data in parenthesis above are the revised results based on recovery from the uncoupled rod condition discovered during the initial phase of power escalation testing.
Summary Of Control Rod Drive Drop Time Measurements Obtained During The Performance of Control Rod Drive Drop Time Test Core Rod Rod Rod Drive Drop Position Group Number Time In Seconds E9 1 1 1.250 Gil 1 2 1.250 K11 1 3 1.245 M9 1 4 1.260 M7 1 5 1.247 K5 1 6 1.250 G5 1 7 1.245 E7 1 8 1.245 H F8 2 1 1.245 $ F10 2 2 1.258 Y 111 0 2 3 1.251 N L10 2 4 1.244 N L8 2 5 1.200 b L6 2 6 1.251 11 6 2 7 1.245 F6 2 8 1.242 C9 3 1 1.250 G13 3 2 1.259 K13 3 3 1.259 09 3 4 1.238 07 3 5 1.259 K3 3 6 1.259 G3 3 7 1.245 C7 3 8 1.251 F14 4 1 1.231 B10 4 2 1.213 L14 4 3 1.238 P10 4 4 1.231 P6 4 5 1.238 L2 4 6 1.235 F2 4 7 1.238 B6 4 8 1.212
Summary Of Control Rod Drive Drop Time Measurements Obtained During The Performance Of Control Rod Drive Drop Time Test Core Rod Rod Rod Drive Drop Pos i t ion Group Number Time In Seconds C11 5 1 1.310 (1.210) G9 5 2 1.230 E13 5 3 1.220 M13 5 4 1.210 K9 5 5 1.240 011 5 6 1.225 05 5 7 1.210 K7 5 8 1.220 m M3 5 9 1.225 E E3 5 10 1.220 G7 5 11 1.225 P C5 5 12 1.210 Y D8 6 1 1.245 {f Ell 6 2 1.246 a 111 2 6 3 1.238
- M11 6 4 1.255 EU N8 6 5 1.244
, M5 6 6 1.245 11 5 6 7 1.257 E5 6 8 1.246 11 8 7 1 1.245 D12 7 2 1.231 B8 7 3 1.230 1114 7 4 1.243 N12 7 5 1.231 P8 7 6 1.231 N4 7 7 1.222 11 2
' 8 1.231 D4 / 9 1.242 Note: The second drop time for rod 5-1, given in parenthesis, is the result of recoupling the rod and repeating the drop test. Lod 5-1 was discovered to be uncoupled during the initial puase ut power escalation testing.
J.O ZERO POWER PHYSICS TESTS The Zero Power Physics Test was performed to verify the nuclear design paramete n used in the safety analysis, the Technical Specification limits, and the operational parameters. Acceptance criteria for the tests are given in Table 3.0-1. A summary of measured and predicted results obtained during zero power physics testing is given in Table 3.0-2. 3.1 INITIAL CRITICALITY Initial criticality was achieved on September 15, 1978 at 0637 hours at reactor coolant conditions of 532*F, 2155 Psig, and 1089 ppmB. Control rod groups I through 4 had previously been withdrawn as part of the heatup procedure with reactor coolant boron concentration at 1500 ppmB. The approach to critical began by withdrawing control rod groups 8, 5, and 6 to 100% and group 7 to 90% withdrawn. The reactor coolant system was then deborated until criticality was achieved. Throughout the approach to criticality, inverse multiplication curves versus time, reactor coolant system boron concentration, and quantity of demineralized water added were maintained for the source range detectors by two independent persons. At the end of eacn et ntrol rod group withdrawal and every 30 minutes during the deboration cycle, the count rate was taken from each source range detector by way of th( scaler-counters. The ratio of the initial average count rate to the count rate at the end of each reactivity addition was plotted. 3.2 NUCLEAR INSTRUMENTATION OVERLAP Technical Specifications state that prior to operation in the intermediate nuclear instrumentation range, at least a one decade overlap between the source range and intermediate range must be observed. This means that before the source range count rate equals 10 5cps the intermediate range must be on the scale. If the one decade '.s not observed, the approach to the intermediate range cannot be continued until the situation has been corrected. To satisfy the above overlap requirements after initial criticality was reached, core power was slowly increased until the intermediate range channele came on scale. Detector signal rasponse was thus recorded for both the Inte rmediate and source range channel. This was repeated for another deca d e to assist in determining the number of decades of overlap. The results 01 c.e N"elear Instrumentation overlap at 532*F and 2155 Psig have been tabulated in Table 3.2.1. Examination of Table 3.2.1 shows that the overlap between the source and intermediate range is constant at an average value of 2.10 decades which is well above the minimum of one decade overlap stated in Technical Specifications. 3.3 SENSIBLE HEAT DETERMINATION Dete rmination of the inte rmediate range current level at which the production of sensible nuclear heat occurs is important to the Zero Power Physics Test program in that it establishes the upper level limit. Thus by restricting reactor power operation to a level reduced by a factor of three below the sensible heat level, the effects of temperature feedback are eliminated in the measurement of physics parameters.
The test method was to increase the intermediate range current level in one third decade increments until the detection of sensible heat. Since turbine bypass valves and pressurizer levels were maintained constant, heat production in the core was observed by change in the core outlet temperature and makeup tank level. The point at which there was a definite heatup rate was 1.10 X 10-7 amps. Therefore, during the test 1.10 X ~7 amps on channel NI-3 was defined as the " sensible heat" po in t . From this the upper zero power physics test current limit was established at 3.00 X 10-8 amps. Af ter setting the upper zero power physics test current limit, temperature feedback measurements were performed to verify that the above limit was ad eq ua te . This was done by adding a +20 pcm over a current range of 1.5 X 10-8 to 3.0 X 10-8 amps and verifying that it remained constant. The results indicated that no temperature feedback ef fects were present over the current range indicated. An additional part of the measurement was to ensure that all power range channels were indicating between 0.8 to 2.0 percent full power at the point of sensible heat. This was done to ensure that the overpower trip protection was adequa te. At the point of sensible heat, the indication on NI-5, NI-6, NI-7 and NI-8 were 0.9, 0.8, 0.6 and 1.2 percent full power. Thus, all power range channels except NI-7 were within their recommended tolerances. Since all channels were set near their maximum sensitivity no calibration was perfo rmed. 3.4 REACTIVITY CALCULATIONS Reactimeter is the name given to the Babcock and Wilcox reactivity computer which solves the one-dimensional, inverse kinetics equation with six delayed neutron groups fc. 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 from the plant. Af ter initial criticality and prior to the first physics measurements, an on-line functional check of the reactimeter was performed to verify its readiness for use in the test program. Af ter steady state conditions with a constant neutron flux were established , a small amount of negative reactivity was inserted in the core by inserting 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 tod group 7, from t 0.025 to t 0.085% k/k. The reactivities determine from doubling time measurements were then compared with the reactivities calculated by the reactimeter. The results of the reactimeter verification measurements are summarized in Table 3.4.1 In each case the reactivity calculated by the reactimeter was within the acceptance criteria limit of + 5% of the reactivity determined from doubling times. 3.4.1 "ALL RODS OUT" CRITICAL BORON CONCENTRATION The "all rods out" critical boron concentration was measured at one isothermal temperature test plateau of 532 *F. The measurements actually were made with rod group 7 partially inserted, but the measured boron concentration was adjusted to the all roos out condition using the results of rod worth measurement to determic a the reactivity worth, in terms of ppo boron, of the inserted control rods.
The results are tabulated in Table 3.4.2. These results show that the measured boron concentration was within the acceptance criteria of 1096 +/- 100 ppmB. Subsequent tests revealed that control rod 5-1 was uncoupled, and was in the core when this measurement was performed. This rod has an estimated worth equivalent to 10 ppmB. 3.5 CONTROL ROD GROUP WORTHS The layout of the core according to the standard alpha-numeric mesh showing the initial location of the control rod groups and the location of the 52 iacore detector strings is given in Figure 3.5-1. The number of rods in each group and the reactivity control function of each group is listed below. Rod Group No. No. of Rods Control Function 1 8 Safety 2 8 Safety
~
3 8 Safety 4 8 Safety 5 12 Power Doppler 6 8 Power Doppler 7 9 Transient 8 8 Axial Power Shaping 69 Predictions of control rod group worths for groups 1-4, 5, 6, 7, and 8 at Hot Zero Power and Hot Full Power conditions were made in the Physics Test
. Manual. Measurements of control rod group worths _or groups 5, 6, 7, and 8 were esde during Zero Power Physics Tests using the boron swap method.
This nethod consisted of setting up a deboration rate and compensating for the change in reactivity by small step changes in rod group positions. The calculation of reactivity on a continuous basis is made by the Reactimeter. The output of reactivity in terms of percent milli p (PCM) is recorded on a strip chart and also on a magnetic tape by the Reactimeter. The results of both the predicted rod group worths and the measured group worths are tabulated in Table 3.5-1 for a moderator temperature and pressure of 532 *F 2155 Psig respectively. Comparison between measured and predicted rod worth shows groups 5-8 agreed within -1.3 percent. Predicted , and estimated rod worths at 579*F are also tabulated. The results of the measured rod worth shapes for control rod groups (5-7) at 532*F and 2155 Psig are plotted for group (8) at 35% wd in Figures 3.5-to 3.5-7. The rod worth for control rod group (8) was also measured and is shown in Figures 3.5-8 to 3.5-9. 3.6 SOLUBLE POISON WORTHS 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 f uel throughout each core life cycle. Measurement of the soluble poison dif ferential worth has been completed at 532*F and 2155 Psig. The measured value was determined by summing the
incremental reactivity values measured during the rod worth measurements over a known boron concentration range from 1106 to 772 ppmB. The result of the differential soluble poison worth measurement is tabula ted in Table 3.5-1. In summary, the measured differential boron worth at zero power was within 5.4 percent of the predicted worth which is within the acceptance criteria of + 10 percent. 3.7 EJECTED CONTk0L ROD WORTH Pseudo ejected control rod reactivity worth was measured at hot, zero power conditions of 532 *F, and 2155 Psig for control rod 7-8. The purpose of this measurement was to verify the safet; analysis calculations relating to the assumed accidental ejection of tba mc reactive control rod which is fully inserted in the core during normal er operation. The acceptance criterion for the ejected control rod teo. that the reactivity worth of the most reactive control does not exceed 1.0%a k/k at hot zero power conditions. Since the ejected control rod worth increases as more rods are inserted into the core, the measurement was performed near the minimum allowable rod index of 103 as specified by Technical Specifications. The ejected worth of control rod 7-8 was measured by the boron swap method. Control rod 7-8 and rods in 1/4 core symmetric locations were also measured by the rod swap method. The rod swap test method was to initially position Group 5 to 85 percent withdrawn with the ejected control rod 0% withdrawn. Then the ejected rod and control rod grou,< 5 and 6 were swapped until the ejected rod was 100% withdrawn. The chant in Groups 5 and 6 positions was then converted to reactivity by using a previously generated rod worth to curve to determine the measured ejected rod worth. Since the measured rod position did not precisely equal the minimum rod index, the measured ejected rod worth was adjusted to the Technical Specifications position by using the equation below: Where: p (ER) = F X p(R) + p(M) EQ. ( 3.7-4 ) p (ER) = The measured ejected rod worth at the maximum allowable inserted group worth. F = Empirically derived constant of 0.45 which relates ejected rod worth and total inserted group worth. p (R) = The worth of rods at the minimum rod index minus the worth of rods at the final rod configuration. p (M) = The measured worth of the ejected rod which corresponds to the worth changes in Groups 5 and 6. The relative ejected rod worth measurement was performed at two different times during the Physics Test program. Following completion of the initial ejected rod worth measurement, and an escalation to 40% full power, a 13% quadrant power tilt was observed which suggested that one of the control rods was uncoupled . This was confirmed through an individual rod movement measurement. The relative ejected rod worth measurements were perfo rmed a second time using the rod r'ap method. To ensure that the maximum measured
second time using the rod swap method. To ensure that the maximum measured ejected rod worth was conservative it was multiplied by an uncertainty factor of 1.10. The measured worst case ejected rod wortn is within the requirement of being less than 1.07. a k/k. The maximum ejected rod worth corrected to the rod insertion limit is within 2% of the predicted value. The results of the measurements are given in Table 3.7-1. 3.8 TEMPERATURE COEFFICIENTS OF REACTIVITY The temperature coef ficient of reactivity is defined as the fractional change in the excess reactivity of the core per unit change in core t em pe ra t ure. The temperature coef ficient is no rmally divided into two components as shown in Equa tion (3.8.1). aT= aM + aD EQ. (3.8-1) Where: aT= Temperature Coefficient of Reactivity aM= Moderator Coefficient of Reactivity aD= Doppler Coefficient of Reactivity The technique used to measure the 532*F and 2155 Psig isothermal temperature coef ficient at zero power was to first establish steady state conditions by maintaining reactor flux, reactor coolant pre ss ure , turbine header pressure and core average temperature constant, with the reactor critical at approximately 1 X 10-8 amps if . negative feedback ef fect wa s expected from a temperature decrease or at approximately 1 X 10-9 amps if a positive feedback ef fect was ext.cted. Equilibrium boron concentration was established in the reactor coolant system, make up tank and pressurizer to eliminate reactivity ef fects due to boron changes during the subsequent temperature swings. The reactimeter and the brush recorders were connected to monitor selected core pacameters with the reactivity value calculated by the reactimeter and the core average temperature displayed on a two channel recorder. Once steady state conditions were established, a negative heatup rate we; started by opening the turbine bypass valves. As the reactivity changed about + 40 PCM, control rods were moved in step changes to keep the reactivity values on scale and also to keep core power in the range necessary to produce an adequate intermediate range signal. Af ter the core average temperature decreased by about 5'F, coolant temperature and reactivity were stabilized. This process was then reversed except that the core average temperature was increased by about 10'F. After stabilizf coolant temperature and reactivity, the core average temperature was en returned to its original value. The measurement of the temperature coefficient from the data obtained was then performed by dividing the change in reactivity by the corresponding changes in core temperature over a specific time period. Isothermal temperature coef ficient measurements were conducted at two dif fe rent reactor coolant boron concentrations during the Zero Power Physics test program. The results of the measurements are summarized in Table 3.8-1 along with the predicted values which are included for c' aparison. In all cases the measured results compared f avorably virk the predicted values. All measured temperature ;;erricients of reactivity were within the acceptance criteria of + 0.40 X 10-4a k/k/*F of the predicted value.
The moderator coefficient cannot be directly measured in an operating reactor because a change in moderator temperature causes a similar change in the fuel temperature. However, since the moderator coefficient has safety implications, it is an important reactivity coefficient. As specified in Technical Specifications the moderator temperature coef ficient shall not be positive at power levels above 95 percent full power and shall be less than t0.9 X 10-'A k/k/*F at all oti.ar power levels. To obtain the moderator coefficient f rom the measured temperature coef ficient, a Doppler correction of -0.2 X 10-'tk/k/*F must be added. Determination of the moderator coefficient has been completed and shows that this coefficient is well within the limits stated above. In conclusion, all temperature coef ficients of reactivity that were measured at the 532 F and 2155 Psig plateau were within the acceptance criteria of + 0.40 X 10-*A k/k/ *F of the predicted value. In addition, calculation of the moderator coefficient indicates that it is well within the requirements of the Technical Specification 3.1.1.3.
3.9 BIOLOGICAL SHIELD SURVEY A Biological Shield Survey was not done at Hot Zero Powe since no plant modifications were made which would invalidate the Biological Shield Surveys made during the Initial Startup Testing Program. 3.10 EFFLUENT AND EFFLUENT MONITORING No ef fluent or ef fluent monitoring testing was performed as these systems have been performing normally since initial startup and no further testing was required. 3.11 CHEMICAL AND RADI0 CHEMICAL TESTS Chemical and radiochemical testing was not performed during this startup. These tests were conducted at initial startup and no plant modifications have been made which would invalidate the results of those tests.
Acceptance Criteria Deviation Limits Between tkasured and Predicted Values Allowable Deviation Between Core Physics Parameters Predicted and Measured Values A. Control Rod Worths Group Worths (5 - 8) + 15% Total Worth (5 - 8) + 10% g Ejected Rod Worth + _.20% E m u o B. Temperature Coef ficients -+ 0.4 X 10-4a k/k/*F s> L C. Dif ferential Boron Worth + 10% D. All Rods Out Boron Concentration + 100 ppmB
Comparison Of tkasured And Predicted Results Obtained During Zero Power Physics Test Physic Parameter Units Measure 6 Predicted Accept ace Requirements Comparison
- 1. All Rods Out Critical Boron ppmB 1106 1096 Measured Value must be within OK (10) 100 ppsb of predicted value
- 2. Sensi :.e Heat amps 3.00 x 10-8 None 3 NI O .rlap decades 2.10 Greater than 1.0 decade OK ( l.0)
- 4. Control Rod Group Worth
% a k/k Percent deviation of group worth Group 8 -0.42 -0.36 should be less than 15% OK (14%)
e Group 7 -1.15 -1.19 OK (-3.5%)
$- Group 6 -0.99 -0.93 OK (6.17)
E Group 5 -1.34 -1.37 OK (-2. '_I) u Percent deviation of total worth c"3 TOTAL -3.90 -3.85 shou'd be less than 10% (1.3%) T
- 5. Ejected Rod Worth (7-8) % a k/k +0.83 +0.85 Percent deviation of ejected rod (-2.4%)
worth must be within 20% of the predicted value. Also value must more negative than +1.0%Ak/k.
- 6. Temperature Coef ficient 10-44 k/k*F !!easured value must be within OK (0.04)
At 1103 ppm 3 -0.05 -0.09 0.4 X 10-4a k/k/*F of the pre- OK (0.21) At 895 ppmB -0.61 -0.09 dicted value.
Comparison of Measured And Predicted Results obtained During Zero Power Physics Test Physic Parameter Units Me_asured Predicted Accedptar.ce Requirements Compa ri son
- 7. Moderator Coef ficient 10-4 A k/k/*F Maximum positive moderator coefficient must be less than 0.9 x 10x-4 Ak/k/*F At 1106 ppm 3 +0.15 +0.11 OK At 895 ppmb -0.41 -0.20 OK
- 8. Dif ferential Boron Worth % A k/k/ppmB At 939 ppmB -0.0111 -0.0105 Pe rcent deviation must be less OK (5.4%)
E than 10% 5 -9 t4 E 8 D NOTE: Percent Deviation is defined by the following formula:
% Deviation = Measured Value - Predicted Value X 100 Measured Value
Summary of Nuclear Instrumentation overlap tie.isurements During Zero Power Physics Test Source Range Indication Intermediate Range Indication Average SR Average IR Data Indication Indication Overlap Set NI-l (CPS) NI-2 (CPS) NI-3 (A!!PS) NI-4 (AMPS) (CPS) ( A!!PS) (Decades) 01 9.0 X 103 9.0 X 103 1.0 X 10-11 1.0 X 10-II 9.0 X 103 1.0 X 10-11 2.05 02 1.0 X 104 1.0 X 10-4 1.0 X 10-11 2.0 X 10-11 1.0 X 104 1.5 X 10-11 2.18 03 1.0 X 105 1.0 X 105 9.0 X 10-11 1.5 X 10-10 1.0 X 105 1.2 X 10-10 2.08 g Average = 2.10 E 7 Note (l): Overlap is obtained between the Source and Intermediate Range by using the average indications in the equation below. Overlap = (6 - Log SR ) + ( Log IR + 11 )
Comparison Of Reactimeter and Doubling Time (DT) Reactivity !!easurements DT Converted Reactimeter Absolute Case DT Reactivity Reactivity Error No. (Sec) (% Ak/k),__ (% Ak/k) (%) 1 215.0 -0.0255 -0.0260 2.0 2 151.0 +0.0275 -0.0270 1.8 3 115.3 -0.0550 -0.0530 3.6 4 54.3 +0.0635 +0.0620 2.3 5 80.0 -0.0920 -0.0880 4.3 g 6 36.0 +0.0840 40.0830 1.2 E G t - NOTE: The Absolute Error Between DT Converted Reactivity and Reactimeter Reactivity is Defined By the Equation Below: E (%) = 100 ( PR - P DT) / ( PDT)
All Rods Out Critical Boron Concentration (PPM Boron) Modera tor Predicted Measured Temperature Results Results 532*F 1096 1106 (1)
- Note (1): This number is based on the measured boron concentration with a 3 ppmB correction to account for the fact the rods were not quite at 100% withdrawn.
Y
Comparison Of Predicted And Measured Control Rod Group Reactivity Worth At 268.8 EFPD A. tbderator Temperature at 5 32 *F, APSR's a t 35% Withd rawn Predicted !!eas ured Pe rcen t Rod Number Wo r t i. Worth Devia tion Group Of Rods (%Ak/k) (%Ak/k) (%) 5 12 -1.37 -1.34 -2.2 6 8 -0.93 -0.99 +6.1 7 9 -1.19 -1.15 -3.5 8 _8 -0.36 -0.42 +14.3 Total 37 -3.85 -3.90 + 1.3 (-9.48) B. Moderator Temperature at 579'F, AFSR's at 35% Withdrawn Predicted Estimated Rod Number Worth Worth Group Of Rods (%Ak/k) (%Ak/k) ~ 5 12 -1.60 -1.56 6 8 -1.04 -1.10 7 9 -1.23 -1.19 8 8 -0.40 -0.46 37 4.27 4.31 NOTE: Estimated worth is determined Ly applying the percent deviations between predicted to measured results at 532*F to the predicted results at 579"F.
1 2. 3 4 5 6 7 g 9 10 11 12 13 14 15
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NI-2 2 - Total Core Monitor
- Total Core and Symmetry Monitor
- Symmetry !bnitor Z - Incore Monitor Number CRX-Y Control Rod Number Y of Control Rod Croup X Figure 3.5-1
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Determination of The Relative Symmetric Ejected Rod Worth at Zero Power, BOL, And 532*F Conditions Rod Position, I wd Inserted RCS Boron Ejected Rod Change In Ejected Rod Work, %Ak/k Reactivity Concentration Position Reactivity 1 2 3 4 5 6 7 8 (%3k/k) (ppm) (%Ak/k) Measured Worst Case 100 100 100 100 100 84 0 33 1.69 965 0 0.68 0.77 100 100 100 100 100 18 0 35 2.37 100 100 100 100 100 *23 84 0 35 1.69 0 0.59 0.73 s 100 100 100 100 100 29 0 35 2.28 100 E 100 100 100 100 100 84 0 35 1.69 0 0.78 0.83 0.91 w s 100 100 100 100 95 0 0 35 2.47 100 100 100 100 100 100 84 0 35 1.69 0 0.63 0.75 100 100 100 100 100 25 0 35 2.32 100 I Note: The measured ejected tod worth has been normalized to the Technical Specifications minimum allowable rod index of 103. l Tjie worst I lcaseI ejected I I Note. rod worth is obtained by application of aa uncertainity factor of 1.10. l
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4.0 POWER TESTING 4.1 REACTIVITY COEFFICIENTS AT POWER TEST 4.1.1 PURPOSE The purpose of this test was to determine reactivity coefficients during power operation at 100 percent full power, and to verify that they were conservative with respect to the FSAR. The following coefficients were either measured or calculated from the data obtained. (a) Temperature coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in fuel and moderator t empera t ure . (b) Moderator coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in moderator temperature. (c) Power Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in power. Acceptance criteria are specified for the Reactivi., Coefficients at Power Test and are listed below: (1) The moderator coefficiegt of reactivity measured at 100% full power shall be less than 0.00 x 10-*a k/k/*F. (2) The power Doppler coefficient of reactivity shall be more negative than
-0.55 x 10-4A k/k/%FP.
4.1.2 TEST METHOD Reactivity coefficient measurements were made during the power escalation test program at 100% full power. Dif ferential rod worth measurements were performed during the reactivity coefficient measurement in order to generate rod worth data for the specific test conditions. For temperature coefficients, average reactor coolant temperature was increased or decreased about 5'F and data recorded. For power Doppler coefficients, power was decreased about 5 percant f ull power and data reco rd ed . From the measured temperature and power Doppler coefficients, the moderator and Doppler coefficients were calculated. 4.1.3 TEMPERATURE AND MODERATOR COEFFICIENTS The temperature coefficient of reactivity is defined as the fractional change in the reactivity of the core per unit change in fuel and moderator temperature. The temperature coef ficient is normally divided into two components as shown in equa tion 4.1-1. aT = aM + cD EQ. (4.1-1) Where: aT = Temperature Coef ficient of Reactivity aM = Moderator Coefficient of Reactivity aD = Doppler Coefficient of Reactivity
The moderator coefficient cannot be directly measured in an operating reactor because a change in the moderator temperature also causes a similar change in the fuel temperature. Therefore, the moderator coefficient must be calculated using equation 4.1-1 af ter the temperature and Doppler coef ficients have been d e te rmined . Technical Specifications, section 3.1.1.3, requires that the moderator coef ficient be less than +0.90 X 10-4 Ak/k/*F at power levels below 95% full power and non positive at or above 95% full power. Temperature and moderator coef ficients were theoretically predicted as shown in Figures 4.1-1 and 4.1-2 using the distributed moderator and fuel temperatures instead of the isothermal values which were used for zero power physics predictions. For these predictions, the normal mode of operation with critical boron and rod conditions was asnumed which set the average core moderator temperature equal to 579*F. The measurement method used at power is to change the reactor coolant temperature setpoint at the reactor control station with the integrated control system in automatic af fecting an approximate 5"F change in the reactor coolant temperature. The reactivity change caused by the temperature change of the core was measured by recording the change in the position of the controlling control rod group and converting this change to reactivity using differentiti rod worth valdes measured during the test. 'rior to running the test, ste'.uy state equilibrium xenon conditions including a stable boron concentration and no significant control rod motion during the last 30 minutes prior to taking data were required as prerequisite system conditions. The power Doppler coefficient relates the change in core reactivity to a corresponding change in power. Theoretical predictions of the power Doppler coefficient were made using the PDQ code wit thermal feedback. The predicted power Doppler coef ficients using this code are presented in Figure 4.1-3. The measurement method used was to change the reactor power level 5 percent full power. This change in power level was initiated by manually decreasing the reactor power at the reactor master control station. Af ter obtaining approximately ten minutes of steady state fata at the reduced power level, reactor power was returned to the initial power. The calculation of the power Doppler coefficient uses the measured change in the control 11rg rod group position converted to an equivalent reactivity value and the measured change in reactor power determined by using the normalized core T which is the primary side heat balance. 4.1.4 DIFFERENTIAL ROD WORTH AT POWER The method by which the dif ferential rod worth was determined at power is the fast insertion / withdrawal method. In this measurement, the controlling rod group is inserted fcr approximately six seconds, followed immediately by a withdrawal for approximately six seconds Since the total elapsed time is on the order of the primary loop recirculation time, the moderator temperature effects are eliminated and the reactivity versus time is essentially a combination of the ef fects due to the control rod motion and the f uel power variation. Determinaton of the dif ferential rod worth was then found by using the measured reactivity and rod positions, compensating the data by a predicted fuel pow 2r correction f actor.
The fuel power correction factor accounts for the time delay involved in fuel temperature change during the measurement. 4.1.5 EVALUATION OF TEST RESULTS The results of the measured temperature and calculated moderator coeff ? cients at power are plotted in Figures 4.1-1 and 4.1-3 which also shows the predicted temperature and moderator coefficient results. Examination of the calculated moderator coef ficient as plotted in Figure 4.1-2 indicates that the limit of a non positive value is not to be exceeded at 95 percent full power. The soluble poison concentration for equilibrium xenon all-rods out condition is about 775 ppmB. Thus during power operation at or above 95 percent full power the moderator coefficient will be negative even if all control rods are withdrawn from the core. The results of the measured and predicted power Dopper coefficient of reactivity are ploteed in Figure 4.1-3. pler coefficient is that the The acceptance coefficient mustcriterion be morefornegative the measured powerXDogAk/k/%FP. than -0.55 10- Figure 4.1-3 shows that all measured caef ficients are below this value and that the acceptance criterion is adequately met. 4.2 UNIT HEAT BALANCE Heat balance calculations were performed using both the Bailey 855 unit computer, and the IBM 5100 cocputer. The accuracy of these two methods was verified during initial startup by performing hand calculations. No modifications were made during this shutdown which would require reve rification of the heat balance calculstion. 4.3 CORE POWER DISTRIBUTION "2ST Core power distributions were taken as required during the power escalation test program as a part of the following individual tests. Core Power Distribution Test Power Imbalance Detector Correlation Test Incore Detector Test Continuous monitoring of the core power density at 364 core location: is accomplished by the incore monitoring system. This system is comprised of 52 detector strings each having 7 individual nautron detectors equally spaced at seven axial elevations in the center of 52 fuel assemblies. This system is capable of producing detailed core power distributions for either eighth core or quarter core symmetry conditions. A detailed description of the incore monitoring system and the calibration test results are presented in Section 4.18. The output of the incore detectors is connected to the unit computer and is corrected for background, fuel depletion and the as-built dimensions to provide accurate ouputs of relative neutron flur. The computer output of the corrected signals is used to develop core power distributions which provide power peaking information necessary to detei ne core performance in terms of DNBR and LHR.
Implementation of the core power and core power imbalance safety limits, in terms of the reactor pratection setpoints, is shown in Figure 4.3-1. The out-of-core nuclear instrumentation provides the core power and core power imbalant2 signals to the RPS, since the incore monitoring system does not immediately respond to prompt changes in core conditions. The results of the core power distributions taken during the test program as part of the power escalation sequence are discussed by dividing this section into three subsections as follows: (1) Normal Operating Core Power Distributions (2) Worst Case Minimum DNBR and Maximum LHR Calculations (3) Quadrant Power Tilt and Axial Power Imbd.ance 4.3.1 PURPOSE The purposes of the st idy state core power distribution measurementL are as follows: (a) To measure core power distribution and thermal-hydraulic data at the major test plateaus as required by the power escalation test program. (b) To compare the measured and predicted core power distributions at 40, 75, and 100% f ull power. (c) To verify acceptable core thermal-hydraulic parameters at 40, 75, and 100% full power. Four acceptance criteria are specified for the Core Power Distribution Test and are listed below: (1) The core power distribution and thermal-hydraulic parameters have been measured, evaluated, and deemed reasonable. (2) The highest measured radial and total peaking factors are not more than 8.0 and 12 percent greater than predicted, respectively, at 40% FP and 5.0 and 7.5 percent greater than predicted, respectively, at 75% FP and 100% FP. (3) The measured worst case minimum DNBR is greater than 1.30 and the measured worst case maximum LHR is less than 18.00 kW/ft. (4) The extrapolated worst case minimum DNBR is greater than 1.30 and the extrapolated worst case maximum LRR is less than 19.70 kW/f t., or the extrapolated imbalance f alls outside the power imbalance trip envelope as shawn in Figure 4.3-1. 4.3.2 TEST METHOD Computer printouts of the core power distribution and thermal hydraulics conditions were obtained af ter establiqhing steady state conditions at the required power level and rod configurations as determined by the Controlling Procedure for Power Escalation, PT-120. Three-dimensional equilibrium xenon was
required, the APSR's were maintained at a constant position and the axial incore imbalance was maintained within +/-2% FP of zero. 4.3.3 EVALUATION OF THE TEST RESULTS 4.3.3.1 NORMAL OPERATING CORE POWER DISTRIBUTIONS Normal operating , equilibrium xenon core power distributions were measured and predicted for each of the three major power escalation test plateaus, Table 4.3-1. The results of these me asured core power distributions, given in Figures 4.3-2 to 4.3-7, at operatinF control rod configurations indicate a maximen radial peaking factor of 1.2b, and a maximum total peaking factor of 1.68 for the three cases studied. The total peaking factor observed during normal operation was well below the design maximum total peaking factor of 2.67. In all instances, measured maximum radial and total peaking factors were within the acceptance criteria of not being more than 5.0 and 7.5 percent greater than those predicted in the Physics Test Manual.
- 4. 3. 3.2 WORST CASE MINIMUM DNBR AND MAXIMUM LHR CALCULATIONS To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal peaking conditions. The two primary core thermal limits which are indicative of f uel thermal performance are fuel melting and departure from nucleate boiling.
These limits are independent; each must be evaluated to ensut e core safety for a given power peaking situation. Fuel melting is basically a f unction of the local power generated in the fuel which is a combination of radial and axial pea king. The maximum linear heat race limit is 19.70. The upper boundary of the nucleate boiling region is termed " departure frcm nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperature and the possibility of cladding failure. The local DNB ratio, (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux at t ha t location, is indicative of the margin to DNB. Although DNB is not an observable parameter during reactor operation, the observable parnmeters of neutron power, reactor coolant flow, teuperature and pressure can be related to DNB through the use of the W-3 correlation. The W-3 correlation has been developed to predict DNB and the location of DNB for uniform and non-uniform axial heat flux distributions. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1. 30 corresponds to 94.5% probability at a 99% confidence level that DNB will not occur; this is considered a conserva tive margin to DNB for all operational conditions. WORST CASE MINIMUM DNBR DETERMINATION The worst case min tmum DNBR values were calculated by the unit computer for each core power distribution taken as part of the power escalation test program. The results of various worst case minimum DNBR values calculated at each test plateau under normal rod configurations are plotted in Figure 4.3-8. These
results indicate that all measured values were above the design worst case minimum DNBR versus power level and well above the minimum acceptable value of 1.30. Four normal operating equilibrium xenon core power distributions, as required by The Core Power Distribution Test, were obtained during the power escalation s eq uence . Each distribution was subjected to the following analysis: (a) From each core power c'istribution, the worst case measured minimum DNBR was selected. (b) The measured worst case minimum DNBR's were then extrapolated to the overpower trip setpoint and corrected for axial peak location and magnitude. (c) Verification of acceptable core conditions at the present and next power level of escalation were then performed relative to 1.30. (d) The measured worst case minimum DNus s were then extranolated to the LOCA and design overpower power level and corrected for axial peak location and magnitude. WORST CASE MAXIMUM LHR DETERMINATION Worst case maximum LHR values were calculated using SP-104 " Hot Channel Factors Calculations" for each standard core power distribution taken as part of the power escalation test program. In all cases, the above limit was met during the power escalation test program, Table 4.3-2. 4.3.3.3 QUADRANT POWER TILT AND AXIAL POWER IMBALANCE QUADRANT POWER TILT Quadrant power tilt limits have been established in the Technical Specifications. These limits, when used in conjunction with the control rod position limits, assure that the design peak heat rate criterion is not exceeded during normal power operation. Quadrant power tilt is defined by the following equation and is expressed in percent. Quad rant Power Tilt . Power in Any Core Quadrant -1 x 100 Average Power of All Quadrants EQ. (4.3-1) During the startup testing program, maximum quadrant power tilt was deteruined using the corrected signals from the 16 symmetric ine. ore monitoring assemblies. . Table 4.3-1 shows the maximum quadrant power tilt for each standard core power distribution taken as required by the Core Power Distribution Test. AXIAL ' 0WER r IMBALANCE Results from the Standard Core Power Distributions taken at 40 and 75% FP during the performance of the Power Imbalance Detector Correlation Test, show that the imbalance trip envelope (Figure 4.3-1) of the reactor protective system is suf fici ent to protect the unit from axceeding the DNBR and the LHR limits under
all core imbalance conditions when a gain factor of 3.90 is set into the delta flux amplifier. In addition, analyses indicate that the largest thermal margins (measured by DNBR and LHR) exist when a negative 5.0% to 10.0* incore axial of f set is present. The core imbalances measured in conjunction with the core power distribution of this section are shown in Table 4.3-1.
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Summary of Measured Core Power Distribution Results at 3-D Equilibrium Xenon Conditions for Core Power Levels of 40, 75, and 100 Percent Full Power Power Rod Position, %wd Core Boron Axial tbx imum Thermal Maximum level Burnup Conc. Imb. Tilt Da te Time (%FP) 1-5 6 7 8 (EFPD) (ppmB) (%FP) (%) DNBR LilR P(R) P(RXA) Predicted --- 40 100 90.3 15.8 25.5 271 NA -1.14 0 --- --- 1.248 1.632 9/30/78 0700 39.71 100 92 17 37 270 770 -2.06 1.08 9.91 5.04 1.280 1.676 Pred ic ted --- 75 100 90.3 15.8 22.3 273 NA 0.28 0 --- --- 1.238 1.518 10/2/78 1000 74.46 100 88 15 18 272 627 -0.39 1.54 4.71 8.63 1.279 1.466 f Predicted --- 100 100 87.1 12.6 12.6 273 NA 1.36 0 -- --- 1.244 1.441 10/4/78 1840 98.14 100 91 16 9 273 579 -2.01 1.27 3.4' , 11.26 1.279 1.452 Y 3
Maximum Linear Heat Rate per Incore Detector Level Versus Power Level Maximum Linear Heat Rate LOCA Limit (KW/ft.) (KW/ft) IC Level 40% FP 75% FP 100% FP 1 3.82 7.60 8.59 14.7 2 5.04 8.21 10.65 15.6 3 4.20 7.64 10.73 16.7 4 3.47 7.48 1 0.92 18.0 5 3.82 8.36 10.26 17.2 6 4.39 8.63 10.58 16.2 7 3.40 5.84 7.29 15.4 Table 4.3-2
Comparison of Predicted and Measured Radial Core Power Distribution Results at Steady State, 3-D Equilibrium Xenon, 40% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 40 %FP Gps 1-5 100 % wd Boron Concentration NA ppm 3 Gp 6 90.3 % wd Core Burnup 271 EFPD Gp 7 15.8 % wd Axial Imbalance -1.14 %FP Gp 8 25.5_ % wd Max Quadrant Tilt 0 % Measured Conditions Control Rod Group Positions Core Power Level 39.7 %FP Gps 1-5 100 % wd Boron Concentration 770 ppm 3 Gp 6 92 % wd Core Burnup 270 EFPD Gp 7 17 % wd Axial Imbalance -2.06 %FP Gp 8 37 % wd Max Quadrant Tilt 1.08 % H K L M N O P R 0.920 1.125 1.091 0.778 8 0.850 1.020 1.119 0.829
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Comparison of Predicted and Measured Total Peaking Core Power Distribution Results at Steady State, 3-D Equilibrium Xenon, 40% FP Conditions Predicted Conditions Control Rod Group Positions Core Power Level 40 %FP Gps 1-5 100 % wd Boron Concentration NA ppmB Gp 6 90.3 % wd Core Burnup 271 EFPD Cp 7 15.8 % wd Axial Imbalance -1.14 %FP Gp 8 25.5 % wd Max Quadrant Tilt 0 % Measured Conditions Control Rod Group Positions Core Power Level 39.7 %FP Gps}}