ML19316A458
| ML19316A458 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/16/1973 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7912110769 | |
| Download: ML19316A458 (400) | |
Text
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, 2 DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT 1 DOCKET NO. 50-269 LICENSE NO. DPR-38 I
STARTUP REPORT : I November Id, 1973 l f
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.1 TABLE OF CONTENTS Section Pm
1.0 INTRODUCTION
1.0-1 2.0
SUMMARY
2.0-1 3.0 INITIAL FUEL LOADING 3.0-1 4.0 TESTING PRIOR TO INITIAL CRITICALITY 4.0-1 4.1 REACTOR COOLANT PUMP FLOW TEST 4.1-1 4.2 REACTOR C00MNT FLOW COASTDOWN TEST 4.2-1 4.3 CONTROL ROD DRIVE DROP TIME TEST 4.3-1 5.0 CORE PERFORMANCE 5.0-1 5.1 INITIAL CRITICALITY 5.1-1 5.1.1 CRITICALITY 5.1-1 5.1.2 NUCLEAR INSTRUMENTATION OVERLAP 5.1-2 5.1.3 ALL-RODS-0UT CRITICAL BORON CONCENTRATIONS 5.1-2 5.2 CORE LIFETIME 5.2-1 5.
2.1 INTRODUCTION
5.2-1 5.2.2 INITIAL EXCESS REACTIVITY 5.2-1 5.2.3 REACTIVITY DEPLETION 5.2-1 5.2.4 REACTIVITY ANOMALY CALCULATIONS 5.2-1 5.3 REACTIVITY CONTROL 5.3-1 5.3.1 CONTROL ROD GROUP REACTIVITY WORTHS 5.3-1 5.3.2 SOLUBLE POISON REACTIVITY WORTHS 5.3-1 5.3.3 SHUTDOWN REACTIVITY MARGIN 5.3-2 5.3.4 EJECTED CONTROL R0D REACTIVITY WORTH 5.3-2 5.3.5 DROPPED CONTROL R0D REACTIVITY WORTH 5.3-2 5.3.6 STUCK CONTROL ROD REACTIVITY WORTH 5.3-3 5.3.7 XENON REACTIVITY WORTH 5.3-4 5.4 COEFFICIENTS OF REACTIVITY 5.4-1 5.4.1 TEMPERATURE COEFFICIENT 5.4-1 5.4.2 POWER DOPPLER COEFFICIENT 5.4-1 5.5 CORE POWER DISTRIBUTIONS 5.5-1 5.5.1 STEADY-STATE, EQUILIBRIUM XENON DISTRIBUTIONS 5.5-1 5.5.2 TRANSIENT DISTRIBUTIONS 5.5-2 5.5.3 MINIMUM DNBR AND MAXIMUM LINEAR HEAT RATE CALCULATIONS 5.5-4 5.5.4 QUADRANT POWER TILT 5.5-6 5.5.5 AXIAL POWER IMBALANCE 5.5-6 5.5.6 AZIMUTHAL OSCILLATIONS AT 40 AND 75% FP 5.5-7 5.5.7 AXIAL OSCILLATIONS AT 75% FP 5.5-8 5.5.8 AXIAL PROFILES FOR FUEL DENSIFICATION EFFECTS 5.5-9 5.6 NEUTRON NOISE ANALYSIS 5.6-1 5.7 NUCLEAR INSTRUMENTATION 5.7-1 5.7.1 OUT-OF-CORE NUCLEAR INSTRUMENTATION 5.7-1 5.7.2 INCORE NUCLEAR INSTRUMENTATION 5.7-1 i
TABLE OF CONTENTS (CONTINUED) Section g 6.0 REACTOR COOLANT SYSTEM PERFORMANCE 6.0-1 6.1 GENERAL SYSTEM PERFORMANCE 6.1-1 6.1.1 STEADY-STATE OPERATION 6.1-1 6.1.2 RESPONSE TO ABNORMAL CONDITIONS 6.1-1 6.1.3 OPERATION WITH THREE REACTOR COOLANT PUMPS 6.1-2 6.2 STEAM GENERATOR PERFORMANCE 6.2-1 6.2.1 THERMAL PERFORMANCE 6.2-1 6.2.2 HYDRAULIC PERFORMANCE 6.2-1 6.2.3 PERFORMANCE MARGIN TEST 6.2-2 6.3 REACTOR COOLANT PUMP PERFORMANCE 6.3-1 6.4 REACTOR COOLANT FLOW 6.4-1 7.0 REACTOR AUXILIARY SYSTEMS PERFORMANCE 7.0-1 7.1 RADIOACTIVE WASTE MANAGEMENT 7.1-1 7.2 WATER CHEMISTRY 7.2-1 8.0 STEAM AND POWER CONVERSION SYSTEMS PERFOREiNCE 8.0-1 8.1 TURBINE-GENERATOR PERFORMANCE 8.1-1 8.2 FEEDWATER AND CONDENSATE SYSTEMS 8.2-1 8.3 EMERGENCY POWER 8.3-1 l j 1 l 1 1
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1.0 INTRODUCTION
On February 6, 1973, the Atomic Energy Commission issued Facility Operating License DPR-38 to Duke Power Company for Oconee Nuclear Station, Unit 1. The first fueI~ assembly was inserted into the core on February 9, 1973, and initial fuel loading was completed on February 16, 1973. On March 6, 1973, during final, non-nuclear testing prior to initial criticality, an oil fire occurred in the vicinity of a reactor coolant pump. While the incident did not, and could not, have any significant effect on the nuclear safety of the unit, unit startup was delayed until recovery activities were completed. On April 11, 1973, the unit was returned to a status approximately equivalent to that which existed prior to the fire. Subsequently, on April 19, 1973, Oconee Nuclear Station, Unit 1, successfully achieved initial criticality. , Following the completion of Zero Power Physics testing, initial power level escalation was conducted on May 4, 1973, and further power level escalations occurred as required testing was satisfactorily completed. Major power levels, as defined by the power escalation testing sequence, were initially achieved as follows: Power Level (Percent of Full Power - %FP) Date 15 May 5, 1973 40 May 27, 1973 75 June 17, 1973 95 August 10, 1973 This report addresses unit startup and power escalation testing through 1200 hours, October 4, 1973. At that time, all or part of several power escalation tests remained to be completed as follows: (a) TP/1/A/0800/04, Natural Circulation Test - decay heat source portion. (b) TP/1/A/0800/10, Xenon Reactivity Worth and Rapid Depletion Test - ' I 95-30-95% FP transient portion. (c) TP/1/B/0800/13, Unit Loss of Electrical Load. (d) TP/1/A/0800/14, Turbine / Reactor Trip Test - 95% FP turbine trip portion. (e) TP/1/A/0800/29, Unit Load Transient Test - 75 and 95% FP portions. . Following completion of the above items, appropriate informstion, supplementary to this report, will be compiled and submitted concerning the results of the testing listed. I This report is prepared and submitted in accordance with Technical Specifi-cation 6.6.1.1A and addresses unit startup and power escalation testing, and the results thereof, with regard to Oconee Nuclear Station, Unit 1. 1.0-1
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2.0
SUMMARY
Oconee 1 is the first of a series of nuclear steam supply systems, designed by the Babcock and Wilcox Company and rated at approximately 886 HWe (net), to be placed into service. Therefore, due to Oconee 1 being a first-of-a-kind unit, the unit startup test program has been more extensive than a normal startup test program. As of 1200 hours, October 4, 1973, the unit has been operated at power levels up to, and including, 95% FP. In general, the performance of the unit has been acceptable. Testing and operation of the nuclear steam supply system has revealed few items which were other than predicted, and none which ad-versely affected unit safety. The deficiencies encountered have been of a nature that would be expected during the initial startup of a unit of this magnitude and, based on an evaluationof unit startup and power escalation testing, it is felt that the unit may be safely operated at full rated power. Oconee 1 startup and power escalation testing, as addressed by the various major sections of this report, is summarized below: (a) Initial Fuel Loading Initial fuel loading began on February 9, 1973, and was completed on February 16, 1973. The major difficulty experienced concerned the proper seating of assembly 1B06, due to what was concluded to be a slight misalignment of the fuel assembly upper,end fitting. In general, fuel loading proceeded in a safe and judicious manner. (b) Testing Prior to initial Criticality Following initial fuel loading of Oconee 1, certain testing was conducted ; prior to initial criticality. This testing included the Reactor Coolant Pump Flow Test, the Reactor Coolant Flow Coastdown Test and the Control Rod Drive Drop Time Test. In all cases, test results met the applicable ac-ceptance criteria. (c) Core Performance The Oconee 1 reactor core consists of 177 fuel assemblies, each containing 208 fuel rods, 16 control rod guide tubes and one incore instrument guide tube. Reactivity control is by means of 61 full-length silver-indium-cadmium control rod assemblies and soluble boron. Also, eight part-length control rod assemblies (axial power shaping rod assemblies - APSR's) provide ad- l ditional control for axial power shaping. The perfor=ance of the initial Oconee 1 core has been in agreement with predicted performance and acceptance criteria concerning core performance have been met. (d) Reactor Coolant System Performance The Reactor Coolant System consists of the reactor vessel, the pressurizer, two steam generators, four reactor coolant pumps and interconnecting piping. The system has been operated at various conditions ranging from cold, zero power to hot, 95% FP over a period of approximately 5 months. The per- . l l l 2.0-1 1
. e formance of the system has been satisfactory throughout this period. Total reactor coolant system leakage has been less than one gallon per minute and has remained essentially constant for a period of approximately four months.
The steam generators have proven to transfer heat somewhat more efficiently than was expedted. Reactor coolant flow has been determined to be 108.6 percent of design flow. This value is in excess of the minimum allowable and is also satisfactory in that it is lower than the maximum allowable. In general, Reactor Coolant System performance has exceeded nominal predicted values. (e) Reactor Auxiliary Systems Performance In general the performance of the radioactive waste management' systems has been adequate. However, larger than anticipated waste volumes have been en-countered. With only one significant exception, when testing indicated an out-of-limit chloride concentration, Reactor Coolant Systen chemistry specifi-cations have been maintained. This situation was corrected and measures taken to avoid a similar recurrence. (f) Steam and Power Conversion Systems Performance During the startup testing program, certain problems were encountered with
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the feedwater and condensate systems which caused several reactor trips. These were resolved and the operation of the steam and power conversion systems has been satisfactory. In addition, proper operationof the emergency feedwater system and proper unit response to a loss of offsite power were verified. t 2.0-2 i l
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3.0 INITIAL FUEL LOADING _ The first fuel assembly, 1C71, was inserted into the core on Februs.ry 9, 1973. Initial fuel loading was completed on February 16, 1973, with the installation df fuel assembly IC37. Figure 3.0-1 depicts the core con-figuration at the conclusion of initial fuel loading. Throughout fuel loading, two temporary incore detectors and the unit's two nuclear instrument iion source range channels, NI-l and NI-2, monitored ! l neutron count rate. Plots of inverse neutron count rate ratio were inde-pendently maintained from the outputs of these detectors. During the first several fuel loading steps, as each successive fuel assembly was inserted into the core, the flux incident upon the incore detectors increased signi- ] ficantly. These readings were subsequently determined to be due to the ) unsatisfactory positioning of the detectors. However, at the time this l situation occurred, in order to assure that an unsafe condition did not i develop and in order to comply with procedural requirements, fuel assemblics were inserted into the active core region in discreet increments, with an inverse neutron count rate ratio being plotted after each incremental in-sertion and prior to further insertion. I l At the time fuel loading began one of the two available fuel assembly up- i ender carriages was not operable. This situation was corrected on February ! 12, 1973, and no further difficulties vere encountered with this equipment. After loading the first fuel assembly, 1C71, a leaking hydraulic hose reel required replacement on the main fuel handling bridge. The second assembly ~ to be inserted into the core,1C67, was found to be bowed and the assembly was returned to the vendor's fabrication facilities for investigation and repair prior to loading into the core. Difficulty was also experienced in properly seating assembly 1B06 due to what was concluded to be a slight mis-alignment of the fuel assembly upper end fitting. The a%ove listed incidents occurred relatively early in the fuel loading sequence and were satisfactorily resolved, although they did have some adverse effect on the fuel loading schedule. The fuel loading procedure initially required that after inserting an as-sembly into the core, and prior to unlatching from that assembly, that data be taken and the inverse neutron count rate ratio plots be completed. After approximately twenty fuel assemblies had been loaded, the procedure was revised to allow the fuel handling bridge operator to unlatch from an assembly once a stable neutron count rate had been verified. In this manner, the procedural matters of data taking and plotting could proceed in parallel with prepara: ion of the fuel handling bridge for the subsequent assembly loading. Verification of the inverse neutron count rate ratio plots continued to be required prior to the movement of each successive assembly into the core region. This procedural change is estimated to have decreased the time required for fuel loading by at least 48 hours, without adversely affecting the safety of the operation. Initial fuel loading at Oconee Nuclear Station, Unit 1, was completed in sevcn days, however, the majority of the fuel assemblies were inserted into the core during the final two-day interval. The several equipment and procedural dif- i 3.0-1 l I
ficulties which initially caused delays in the fuel loading sequence were satisfactorily resolved and fuel loading proceeded in a generally orderly manner and was conducted in accordance with the applicable writtea approved
- procedures.
l l f I I W i i 4 ? I s 1 I l 3.0-2 i
INITIAL FUEL LOADING FINAL CORE CONFIGURATION IC11 1C24 1C26 1C49 1C05 1C18 1C06 1C07 1C08 1C25 1C27 1C47 1C48 1C20 C21 006 C31 056 C42 1C14 1C63 1C17 IB20 IB28 1B01 1B29 1B42 1C68 1C69 1C22 C23 038 C29 001 C32 080 CS2 1C23 1C15 1C38 1B36 1B45 1B22 LA28 1526 1B31 1B32 1CO2 1C70 1C50 053 C39 060 A04 039 Cl3 050 A05 058 C03 076 1C12 1C19 1B17 1816 1B12 LA30 1A18 1A35 1B39 1B33 1B34 1C73 1C39 C22 036 C20 032 C18 031 C27 077 C33 059 C57 1C57 1C56 LB44 1809 1853 1A29 1A25 LA26 1A27 1A32 1856 1B35 1B18 1C35 1C37 C49 093 A03 099 C17 029 C16 030 C24 096 A06 037 C38 1C58 1C55 1B51 1825 1A31 1A37 1A36 1A38 1A05 1A14 1A33 1824 1849 1C28 1C36 087 CSS 049 C19 017 C28 026 C43 023 C25 048 C46 061 1C60 1C59 1B07 LA24 LA15 1A21 1A39 1A16 1All LA12 1A40 1A34 1B15 1C44 1C30 C50 010 C15 025 C14 020 C12 021 C10 027 C26 035 C41 1C61 1C62 1B48 1B43 1A22 1A03 1A04 1A08 1A01 LA02 1A13 1B52 1B23 1C31 1C32 092 C51 081 C60 016 C07 019 C05 022 C06 047 C54 063 1C64 1C03 1350 1B11 1819 1A20 1A07 1A06 1A09 1A23 1B46 IB06 1B08 1C33 1C34 CO2 015 A07 083 C44 018 C08 028 C09 034 A02 012 C37 1C65 1C66 1B13 1B14 1B40 LA10 LA17 LA19 1B10 1B04 1B37 1C74 1C10 C53 033 C48 078 C45 024 C11 014 C04 051 C40 1C16 1C80 1C40 IB38 1B54 1B41 1A41 1B47 1802 1B05 1C01 1C75 1C71 04' C30 057 A08 079 C61 086 A03 009 C01 102 1C21 1C78 1C76 1B55 1B30 1B03 1B27 1B21 1C77 1C79 1C67 C58 100 C36 007 C34 013 C59 1C13 1C43 1C29 1C46 1C45 1C51 1C52 1C09 1C72 C35 073 C56 082 C47 1C04 1C41 1C42 1C54 1C53 Fuel Transfer Canal Or Legend Items Description LA01 through 1A41 2.00 w/o Fuel Assemblies 1801 through 1B56 2.10 w/o Fuel Assemblies 1C01 through IC80 2.15 w/o Fuel Assemblies C01 through C61 Control Rod Assemblies 001 through 108* Orifice Rod Assemblies A01 through A08 Axial Power Shaping Rod Assemblies l *1nstalled orifice rod assemblies not numbered consecutively ( due to the re= oval of 44 assemblies to increase core bypass flow. l
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i 4.0 TESTING PRIOR TO INITIAL CRITICALITY Following initial fuel loading of Oconee 1, certain testing was conducted prior to initial criticality. This testing included the Reactor Coolant Pump Flow Test, the Reactor Coolant Flow Coastdown Test and the Control Rod Drive Drop Time Test. In all cases, test results met the applicable ac-ceptance criteria. 1 4 4.0-1
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4.1 REACTOR COOLANT PUMP FLOW TEST Prior to initial criticality, the Reactor Coolant Pump Flow Test was per-formed with the reactor core installed. The purpose of the test was to determine the" functional capabilities of the Reactor Coolant System and the reactor coolant pumps, and to determine reactor coolant flow for various, specified reactor coolant pump operating combinations. Reactor coolant loop flows were determined by means of both the unit computer and loop flowmeter AP cells. For each pump operating combination, five sets of steady-state data were read from the computer, and the indicated flows, temperatures and pressures were averaged and these average values used to calculate properly compensated flow valuee. From the loop flowmeter AP cell indications, flows were calculated as follows: Flow = Cf AP
. Vs.
Where: Cf = Flow Meter Coefficient = 397,100 AP = Indicated AP (psi) Vc " SP ecific volume at reference coaditions (68 F, 14.7 psig) Vs = Specific volume at system conditions Table 4.1-1 gives the minimum and maximum allowable flow rates for four different pump combinations, along with the measured flow rates for each listed condition. It can be seen that the measured flowrates are within the acceptance criteria. In those instances when both reactor coolant pumps in a loop are not operating, and one or two pumps in the other loop are operating, coolant flow through the idle loop is reversed and the net flow through the reactor core is equal to the difference in the indicated loop flows. Table 4.1-2 presents forward and reverse reactor coolant loop flows, determined from the loop flowmeter 6P cells, during single-loop operation. 4.1-1 0 _ a -_
e o REACTOR COOLANT FLOW FOR VARIOUS PUMP COMBLNATIONS Minimum Maximum e Acceptable Acceptable Measured Flow Rate At Flow Rate At Flow Rate At Pump 2155 psig, 532 F Combination 2155gsig,532F 2155 gaig 5320 F (10 lbm/hr) (100 lbm/hr) (10 lbm/hr) Four Pumps 138.4 150.0 145.2 Three Pumps 103.2 150.0 110.3 Two Pumps, One Loop 63.4 150.0 67.1 One Pump Each Loop 67.8 150.0 73.7 1 1 i l l l Table 4.1-1
FORWARD AND REVERSE REACTOR COOLANT FLOWS DURING SINGLE LOOP OPERATION r Pump Totg1 Flow combination LoogAFlow LoogBFlow (10 lbm/hr) (10 lbm/hr) (100 lbm/hr) 1A1 40.63 -7.20 33.43 1A2 39.93 -7.07 32.86 1A1 + lA2 80.52 -12.31 68.21 1B1 -6.70 39.54 32.84 IB2 -7.55 39.82 32.27 1B1 + 1B2 -11.95 79.00 67.05 i Table 4.1-2
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4.2 REACTOR COOLANT FLOW COASTDOWN TEST The purpose of the Reactor Coolant Flow Coastdown Test was to determine reactor coolant flow versus time for various, specified reactor coolant pump trip combinations; and to compare these test results with minimum acceptable flow coastdown criteria. Various combinaticna of reactor coolant pumps were operated and steady-state data acquired. Subsequent?.y, all or a portion of the operating pumps were tripped and data were recorded during the insuing reactor coolant flow transient. Steady-state data were again taken following the flow transient. Reactor coolant flow, at various times during the coastdown transients, was determined from loop flowmeter AP cell data according to the equation given in Section 4.1. Figure 4.2-1 shows the minimum acceptable reactor core coolant flows versus time for both single-pump and multi-pump coastdowns. Measured reactor core coolant flows versus time are presented in Figures 4.2-2 through 4.2-4 for typical coastdowns from four pump initial operating conditions. For all test conditions, reactor core coolant flows versus time exceeded the ap-plicable acceptance criteria. 4.2-1
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4.3 CONTROL ROD DRIVE DROP TIME TEST The purpose of the Control Rod Drive Drop Time Test was to verify the inte-grated, functional trip capability of the Control Rod Drive System and to determine for each control rod assembly, the total elapsed drop time from the initiation of a trip signal until the control rod assembly was three-fourths inserted. This test was conducted at various combinations of reactor coolant flow, pressure and temperature as follows: Test Condition Flow Pressure Temperature 1 No Flow 3,350 psig 5,400 F 3,1700 psig 2 One Pump Each Loop 2,350 psig 5,4000F 5,1700 psig 3 Four Pumps 2155 1 30 psig 532 1 10 F At each condition, Control Rod Groups 1 through 7 were driven, sequentially, to the fully withdrawn position. A manual trip of all control rod drives was then initiated and, coincidentally, a time signal provided to the data logging equipment. As each control rod assembly reached the three-fourths insertion position, a second time signal was provided to the data logging devices from a switch located on each control rod drive's position indi-cator tube. The total elapsed time from the initiation of a trip signal until three-fourths insertion was then determined for each control rod drive from the data acquired. The test was repeated a second time for all control rods at each test condition. An analysis of the drop times at Test Condition 1 indicated that the fastest drop time was 1.127 seconds and the slowest was 1.177 seconds. For Test Condition 2, 1.186 seconds was the fastest drop time and 1.236 seconds was the slouest. Under Test Condition 3 the fastest drop time was 1.267 seconds and the slowest was 1.308 seconds. The rod with the slowest drop time (1.308 seconds) and the fastest drop time (1.127 seconds) were dropped an additional 25 times and produced drop times consistent within 45 milliseconds and 27 milliseconds respectively. 1 As expected, the drop times under flow conditions were longer than under no flow conditions. Rod drop times were well below the acceptance criteria, as i l identified in Technical Specification 4.7, of 1.66 seconds for full flow conditions and 1.40 seconds at ne flow conditions. l i 4.3-1
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. o 5.0 CORE PERFORMAl;CE The Oconee 1 reactor core consists of 177 fuel assemblies, each containing 208 fuel rods, 16 control rod guide tubes and one incore instrument guide tube. Reactiv'ity control is by means of 61 full-length silver-indiesm-cadmium control rod assemblics and soluble boron. Also, eight part-length control rod assemblies (axial power shaping rod assemblies - APSR's) provide additional control for axial power shaping. The performance of the initial Oconee 1 core has generally been as predicted and acceptance criteria concerning core per-formance have been met. l l i l 5.0-1
5.1 INITIAL CRITICALITY 5.1.1 CRITICALITY Initial criticality was achieved on April 19, 1973, at 1215 hours. Reactor Coolant System temperature and pressure at the time of initial criticality were 250 F and 800 psig, respectively. Control Rod Groups 1 through 5 and Control Rod Group 8 were fully withdrawn, Group 6 was 75 percent withdrawn and Group 7 was fully inserted at initial criticality. Criticality was achieved by deboration. The intended procedure for the approach to initial criticality was as follows: (a) Control rod group withdrawal; Groups 1-4 100% Withdrawn Groug 5 100% Withdrawn Group 6 100% Withdrawn Group 7 25% Withdrawn Group 8 100% Withdrawn (b) Deboration from 1800 ppm to 1250 + 100 ppm. (c) Withdrawal of Control P.od Group 7 to approximately 42 percent withdrawn. Deboration, item (b) above, was conducted in two steps. Initially deboration from 1800 ppm to 1500 ppm was accomplished using a 70 gpm letdown and makeup rate for approximately 2h hours. A second deboration was accomplished at a lower makeup and letdown rate of 45 gpm for a period of approximately three hours. The exact amount of deboration during this second step was intended to be limited to that which would produce a final inverse neutron count rate ratio of about 0.003 or an effective multiplication factor, keff, of approxi-mately 0.997. However, a systematic error in the reported boron concentrations resulted in criticality being achieved during the second deboration step. This error was due to the fact that a standard sodium hydroxide titration solution had been prepared and used for several months and during- that period of time the normality of the standard had changed. The following corrective actions have been taken to prevent a similar recurrence: (a) Fresh batches of sodium hydroxide solution are prepared frequently for use in titrating boron samples. The quantities prepared are kept small so that new solutions must be prepared before the normality of the solution can change significantly. (b) A lab boron standard is checked at least once each day to assure that analyses drift does not occur with time. (c) Each sample is run at least twice on the boron titrator to assure pre-cision. If the two results differ a third analysis is performed. O 5.1-1
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(d) The boronometer provides an independent check of lab results. An inves-tigation is initiated to determine which is in error, if they disagree. This incident was considered an unusual event and a written report was sub-mitted to the' Atomic Energy Commission on July 17, 1973, which describes the incident further. Throughout the approach to criticality, two independent plots of inverse neutron count rate ratio were maintained and criticality was achieved in a safe manner. 5.1.2 NUCLEAR INSTRUMENTATION OVERLAP Technical Specification 3.5.1.5 requires that prior to operation in the nuclear instrumentation internediate range that at least one decade of overlap between the source range and intermediate range must be observed. Initial nuclear instrumentation overlap data at 250 F and 800 psig are plotted on Figure 5.1-1, which also includes intermediate range detector data at power. Examination of Figure 5.1-1 shows that the linearity, over-lap and absolute output of both the source range and intermediate range detectors are acceptable and that the detectors are performing 0 satisfactorily. The step change in the intermediate range signal between 250 F and 579 F is due to coolant density and boron concentration effects. 5.1.3 ALL-RODS-0UT CRITICAL BORON CONCENTRATIONS Critical baron concentrations were measured at the three moderator tem-perature test plateaus of 250, 400 and 532 F with Control Rod Group 7 partially inserted. These measured boron concentrations were adjusted to the all-rods-out condition using the results of rod worth measurements. All-rods-out critical boron concentration results are presentcd in Table 5.1-1. 5.1-2
9 s . I a ALL-RODS-0UT CRITICAL BORON CONCENTRATIONS Moderator Predicted Measured Temperature Concentration Concentration ( F) (ppe Boron) (ppe Boron) 250 1334 1329 400 1378 1403 532 1477 1476 i 1 I i Table 5.1-1
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- 5.2 CORE LIFETIME 5.
2.1 INTRODUCTION
The operationrof Oconee 1 through the 95% FP plateau has resulted, as of 1200 hours, October 4, 1973, in an initial core burnup of 56.6 effective full power days (EFPD). Plots of the core power level history for this period are given in Figures 5.2-1 through 5.2-6 and plots of integrated core burnup during this startup period are presented in Figures 5.2-7 through 5.2-12. 5.2.2 INITIAL EXCESS REACTIVITY The initial excess reactivity of the core, at beginning-of-life zero power conditions, was measured in terms of soluble poison concentration at moderator temperature conditions of 250, 400 and 532 F. These measurements are dis-cussed in Section 5.1.3 and the results are presented in Table 5.1-1. The boron concentration data were converted to reactivity, using measured boron reactivity worths, to determine excess reactivity versus moderator temperature. For the clean, zero power, all-rods-out, 579 F condition excess reactivity was determined to be 17.52% Ak/k - see Figure 5.2-13. Using measured power doppler reactivity deficits from the 15, 40, 75 and 95% FP test plateaus, a full power deficit of 1.09% Ak/k was predicted. This results in an excess reactivity for the clean, full power,all-rods-out, 579 F condition of 16.43% Ak/k. 5.2.3 REACTIVITY DEPLETION Reactivity depletion in terms of core excess reactivity and boron concen-tration has been measured through 56.6 EFPD. A tabulation of 100% FP, equilibrium xenon, all-rods-out critical boron concentrations is given in Table 5.2-1. The results of this data, together with measured excess re-activity data, are extrapolated to the end of the initial cycle in Figures 5.2-14 and 5.2-15. From these figures it can be seen that an initial core lifetime of approximately 385 EFPD is predicted. The power escalation test program included a reactivity depletion test which was completed during 95% FP testing. The results of this depletion test agree with data presented in Table 5.2-1. 5.2.4 REACTIVITY ANOMALY CALCULATIONS Technical Specification 4.10 requires that the measured reactor coolant boron concentration be periodically compared with the predicted concentration, following normalization of the computed boron concentration as a function of burnup. If the difference between the observed and predicted concentrations reaches the equivalent of one percent in reactivity, an evaluation must be made and the incident reported to the Atomic Energy Commission. In order to fulfill this requirement a plot of boron concentration versus core burnup is maintained - see Figure 5.2-16. As evidenced by this figure, to date no appreciable reactivity anomaly has been observed. 5.2-1
MEASURED CRITICAL BORON CONCENTHMIONS ADJUSTED TO 100% FP, EQUILIERIUM XENON, 579 F, ALL-RODS-0UT CONDITIONS s se Core Core Measured Differential Power Power Xenon Xenon Worth of Cantrol Rad Critical Boron Concentration Boron Boron Doppler Doppler Reactivity Baron Inserted Boron Adjusted to 100% FP, Nurober Burnup Power Conc. Worth (Za React.Adj. Boron Adj. Adj. Adjustment (butral Rx!s Adjustment Equilibrium, Xenon, All-Rods (EFPD) (% FP) (ppa) k/k/100prun) (% t.k /k) (ope) (%"k/k) (ppm) (tek/k) (ppa) Out Conditions. (ppa) 1404 W 0 0 100 L 0.8 15 1362 1.17 -0.93 -80 -1.46 -125 0.11 + 9 1166 2 1.4 15.6 1350 1.17 -0.92 -79 -1.44 -123 0.11 + 9 1157 3 1.6 15.2 1335 1.17 -0.93 -80 -1.46 -125 0.59 + 50 1180 4 5.5 42.5 1145 1.18 -0. ( 3 -53 -0.61 - 52 1.31 +111 1151 5 5.6 44.0 1130 1.18 -0.61 -52 -0.59 - 50 1.24 +105 1133 6 7.0 40.0 1097 1.18 -0.65 -55 -0.66 - 56 1.74 +147 1133 p 1152 W 7 15.9 77.0 1172 1.18 -0.25 -21 -0.21 - 18 0.22 + 19 h 8 20.0 76.0 1105 1.18 -0.26 -22 -0.16 - 14 0.60 + 51 1120 y 9 27.2 75.5 988 1.19 -0.28 -24 -0.16 - 13 1.72 +145 1096 [ 10 31.4 75.8 1000 1.19 -0.25 -21 -0.16 - 13 1.71 +144 1110 11 36.6 95.8 1075 1.19 -0.05 -4 -0.03 - 3 0.51 + 51 1119 12 40.8 72.0 1008 1.19 -0.27 -23 -0.20 - 17 1.53 +128 1096 13 48.5 94.5 935 1.19 -0.05 -4 -0.03 - 3 1.59 +134 1062 14 51.0 73.0 955 1.19 -22
-0.26 -0.20 - 17 1.64 +138 1054 15 16 17 18 NOTE (1) Estimated BOL critical boron concentration at 100% FP, all-rods-out, no xenon conditions, NOTE (2) This result is in error by approximately 25 ppa due to a probable error in the recor4ed boron concentration.
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a .
. CORE EXCESS _ REACTIVITY VERSUS LIFETIME
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I 3 0 50 100 15 0 ' 200 250 300 350 40:: Lifetime, EFPD 1 l i l Figure 5.2-15
i < BORON CONCENTRATION VARIANCE VERSUS LIFETIME FOR REACTIVITY ANOMALY CALCUIATIONS
+ '! ..
psee ' E: " ~ l- . l--. -.. .-
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. _ :. _l
/000 f, ,
1...-__.__i g
- o. 900 ..
l ; ;..._ a . _ _ . . . . _ L.. . . i . . , . 2 . c 3 900 '. Normalized Calculated Results j s for 579F, 100%FP, All-Rods-Out, Equilibrium Xenon Conditions y.m . y / ..
-^
3 400 // - e ' 500 '., _,. - - - .
)
e too . . _ . . 100 '. O L"O So too f50 200 2so 500 .sgo m l Lifetime, EFPD Figure 5.2-16
o e 5.3 REACTIVITY CONTROL 5.3.1 CONTROL ROD GROUF REACTIVITY WORTHS The Oconee 1 control rod configuration is shown in Figure 5.3-1. Differential reactivity worths, as a function of withdrawal position, were measured for each group of control rods except Group 1 and part of Group 2, which were not determined due to reactivity shutdown requirements. The method used to determine these differential worths was to establish a boration or deboration rate and compensate for the resulting reactivity change by small step changes in rod group position. Also, rod drop measurements were made in an attempt to measure the reactivity worth of Group 1 and that portion of Group 2 which was not inserted into the core. However, this technique for measuring control rod worth is inherently inaccurate unless the results are corrected for geometry effects, detector sensitivity, etc. Consequently, it was decided to determine the Group 1 and 2 worths by assuming the same percentage deviation between the measured and calculated worths as that determined for Group 3. The results of the measured differential rod group worths at 532 F are plotted in Figurea 5.3-2 and 5.3-3 for part-length rod group positions of 35 percent withdrawn and 100 percent withdrawn. Normalized integrated rod worth curves are shown in Figures 5.3-4 and 5.3-5 for zero power conditions at 5320F. These c trves were calculated by inte-grating the differential curves given in Figures 5.3-2 and 5.3-3. Normalized integral worth curves were d oeloped for full power, 579 F conditions by assuming the same percentage deviation between measured and calculated worths at these conditions as detas..ained for 532 F, zero power conditions. Normalized
- 5790F, full power integral sorth curves are shown in Figures 5.3-6 and 5.3-7 for part-length rod group positions of 35 percent withdrawn and 100 percent withdrawn Calculated rod group worths were determined using the PDQ-7 code, with either a two or three-dimensional description of the core. A comparison of calcu-lated and measured beginning-of-life control rod group reactivity worths, for the normal withdrawal sequence, at the three moderator temperature test plateaus of 250, 400 and 532 F is given in Table 5.3-1.
5.3.2 SOLUBLE POISON REACTIVITY WORTHS The reactivity worth of the boric acid, in terms of ppm boron, was calculated . by using the one-dimensional LIFET code and the multi-dimensional PDQ-7 code. These calculations were performed at four moderator temperature and pressure conditions as follows: Moderator Moderator Temperature Pressure (OF) _(psig) 250 800 400 1500 l 532 2155 l 579 2155 l 5.3-1 1
- 1 I
D Measured differential soluble poison worths are compared with calculated worths in Figure 5.3-8 and integral worths are similarly compared in Figure 5.3-9. Thesecomparisonsindicatecloseagreement,withamaximumdeviation in the integral worths of 0.8% ak/k at 579 F, 1400 ppm. 5.3.3 SHUTDOWN REACTIVITY MARGIN Technical Specification 3.5.2.1 requires that the available shutdown margin shall not be less than 1% ak/k with the highest worth control rod fully with-drawn. Shutdown margins calculated at 5.6 and 36.6 EFPD were 6.57 and 7.33% ak/k respectively. Both values are well in excess of the minimum required. 5.3.4 EJECTED CONTROL ROD REACTIVITY WORTH Ljected control rod reactivity worths were measured at zero power conditions of 532 F, 2155 psig for four different control rod assemblies. Also, at the 40% FP test plateau, the ccatrol rod assembly in core position L-6, see Figure 5.3-1, was borated from the fully inserted position to the fully with-drawn position and the effect on core reactivity and core power distribution was measured. Ejected rod worths at zero power were measured using the rod drop method. This technique for measuring control rad worth is inherently inaccurate un-less the results are corrected for geometry effects, detector sensitivity, etc. However, the measured results differed by no more than ten percent from the calculated values, with the exception of one case in which the measured result was 25 percent less than that calculated. The calculated and measured ejected rod worths are listed in Table 5.3-2. At the 40% FP test plateau, a differential reactivity worth curve was generated as the subject control rod assembly was moved from the fully inserted to the fully withdrawn position concurrent with boration of the reactor coolant. Ejected rod worth was then determined by integrating this differential worth curve and, also, by converting the observed change in boron concentration into equivalent reactivity. The rod worth measurements yielded an ejected rod wo2th of 0.21% ak/k and the change in boron concentration resulted in a worth 0.26% ak/k. The effect of the control rod assembly withdrawal on core power distribution was determined using the incore detectors. The results are plotted in Figure 5.3-10 and indicate a maximum positive quadrant power tilt of 37 percent for quadrant WZ and a maximum negative quadrant power tilt of 26 percent for quadrant XY when the rod was at its fully withdrawn position. 5.3.5 DROrF S CONTROL R0D REACTIVITY WORTH Dropped control rod worto was measured for the control rod assembly located in core position H-12, sea Figure 5.3-1. The selection of this assembly was based upon the results of ?DQ-7 calculations which indicated that the control rod assembly located in position H-12 would produce the most adverse thermal effects if dropped into the core during power operation. 5.3-2 o e
The purpose of the rod drop test, beyond determining the reactivity worth of - the control rod assembly, was as follows: (a) To demonstrate that the position indicator alarm and asymmetric rod alarm function. (b) To demonstrate that reactor power is automatically reduced to a prede-
- termined power when a control rod assembly is asymmetric (nine or more inches from the group average position).
(c) To demonstrate that rod withdrawal is inhibited beyond a predetermined power. (d) To measure the core power distribution with the asymmetric rod at 50 percent withdrawn and at the fully inserted position. A differential reactivity worth curve was generated as the subject control rod assembly was moved from the fully withdrawn to the fully inserted position concurrent with deboration of the reactor coolant. Dropped rod worth was then determined by integrating this differential worth curve and, also, by converting the observed change in boron concentration into equivalent re-activity. The rod worth measurements yielded a dropped rod worth of 0.20% Ak/k and the change in boron concentration resulted in a worth of 0.14% Ak/k. At 40% FP automatic runback control acaion was successfully demonstrated by inserting the control rod assembly until it became asymmetric with respect to the remainder of Group 2, i.e. , approximately nine inches halow the group average position. This action resulted in power being aut..aatically reduced to the test setpoint of 30% FP. Rod withdrawal inhibit was then successfully verified by attempting to go beyond 35% FP, which was the test setpoint for rod withdrawal inhibit. The position indicator alarm and the asymmetric rod alarm both responded according to specifications. The effects of the " dropped" control rod on core power distribution were determined by calculating the quadrant power tilt for each core quadrant at the 50 percent and 100 percent inserted positions. The resulting quadrant power tilts are plotted in Figure 5.3-11 and indicate a maximum positive tilt equal to 15 percent in quadrant WX and a maximum negative tilt of 16 percent l in quadrant XY. 5.3.6 STUCK CONTROL ROD REACTIVITY WORTH Stuck rod worth was determined by measuring the worth of a control rod group using the rod drop method, then repeating this measurement less the " stuck" rod. The results of both calculated and measured stuck rod worths are given in Table 5.3-3. The difference between the measured and calculated worths is attributed to use of the rod drop measurement method. Based on these calcu-lations and measurements, a conservative maximum calculated stuck rod worth of 2.2% Ak/k is used in determining available shutdown margin. 5.3-3
1
. +
l 5.3.7 XENON REACTIVITY WORDI Although no specific test was performed during the startup test program for the purpose of obtaining steady-state xenon worth, data at just critical, zero power, xenon free conditions (following a minimum two-day shutdown) have been compared with critical data af ter xenon buildup to equilibrium to obtain estimates of equilibrium xenon worth. The measured critical boron concen-trations given in Table 5.2-1 were utilized to obtain a value of 2.77% ak/k at 100% FP for equilibrium xenon. l l 5.3-4
! l I
I COMPARISON OF CALCUIATED AND MEASURED CONTROL RCD GROUP REACTIVITY WORTHS 3 A. Moderator Temperature 250F. APSR's 100% wd a i Rod Group Number of Calculated Measured Number Rods Worth. %ik/k Worth. %ik/k 1 8 2.20 2 2.16 12 1.56 1.53 3 12 2.30 2.26 4 5 0.51 ] 0.52 5 8 0.62 ! 0.64 6 8 0.75 ! 0.74 7 8 0.88 ! 0.87 61 8.82 8.72 j B. Moderator Temperature 400F, APSR's 100% vd 4, Rod Group Number of l Calculated Measured
-Number Rods Worth, %ik/k Wo rth. %8k/k 1 8 l 2 2.67 2.49 12 1.87 3 12 1.74 2.70 2.52 4 5 0.59 0.58 5 8 i
6 0.70 0. 76 8 0.89 7 8 0.83 1.04 0.96 61 10.46 9.88 C. Moderator Temperature 532F. APSR's 100% wd Rod Group Number of Calculated Number Meas ured Rods Worth %ik/k Worth, mik/k 1 8 2.95 2.60 2 12 2.10 1.85 3 12 2.99 2.63 4 5 0.70 0.64 5 8 0.76 0.68 6 8 1.02 0.97 7 8 _1.17 1.11 i 61 11.69 10.48 D. Moderator Temperature 532F. APSR's 35% vd Rod Group Number of Calculated { Mumber Measured Rods Worth, E k/k Worth. Zik/k 1 8 3.24 2.92 2 12 2.24 3 2.02 12 3.12 3 4 2.81 5 0.62 5 0.59 8 0.68 0.69 6 8 1.21 7 1.10 8 1.19 8 1.11 8 0.57 I 0.52
- 60 12.87 11.76 1
Table 5.3 . . - ,_ ._.- -. .-. .- - _ - . . _ _ , _ , _ . , . . . _ . ~ . _
EJECTED CONTROL ROD ASSEMBLY REACTIVITY WORTHS AT ZERO POWER, 532 F Calculated Measured
. Core Ejected Rod Worth Ejected Rod Worth Position (% Ak/k) (% Ak/k)
L-10 0.306 0.27 N-12 0.265 0.20 K-9 0.149 0.15 H-2 0.324 0.33 i l i i Table 5.3-2 i -
, a 2
^ STUCK CONTROL ROD ASSEMBLY REACTIVITY WORTHS AT ZERO POWER, 532 F Calculated Measured Core Stuck Rod Worth Stuck Rod Worth Position (% Ak/k) (% Ak/k) H-12 1.0 0.12 M-11 2.2 1.0 K-13 1.6 0.77 H-14 1.55 0.33 N-12 2.01 0.4 M-13 1.85 0.47 J i I 1
, Table 5.3-3
CONTROL ROD ASSEMBLY CORE LOCATIONS A
- B (2) (6) (2)
C (5) (3) (3) (5) (7) (8) (2) (8) (7) E (5) (4) (1) (1) (4) (5)
#~
(2) (8) (7) (3) (7) (8) (2) 0~ (3) (1) (6) (6) (1) (3) H-(6) (2) (3) (4) (3) (2) (6) K-(3) (1) (6) (6) (1) (3) E-(2) (8) (7) (3) (7) (8) (2) (5) (4) (1) (1) (4) (5) (7) (8) (2) (8) (7) 0 (5) (3) (3) (5) P (2) (6) (2) R 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 (X)e-Control Rod Group Nu:nber P Figure 5.3-1
b l l 1 l 1 1 DIFFERF.NTIAL CONTROL ROD GROUP REACTIVITY WORDI FOR BEGINNING-OF-LIPS, 532*F, ZERO POWER CONDITIOW __.r. . _4 ._ .. . _w. _ .
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ua m a.. -
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ube=thn
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~
i m:nt: m. =t= m 7.._u m - s.. .+ 'i- 7_ --::a=== : ~.- " Ej i l M/MV % 'tG20M MTATaseau l 1
- Figure 5.3-9
MEASURED, QUADRANT POWER TILT DURING EJECTED CONTROL R0D TEST s
l :i.+ -Ei~ =1
- g. .. ; , _._ . . . . -;;. _ , -
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.A 6" o 20 do . do 33 /gt3 Position of CRA in Core Position L-6, % Wd Figure 5.3-10
e a MEASURED QUADRANT POWER TILT DURING DROPPED ROD TEST
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i MO 80 60 40 20 \ .O, Position of CRA in Core Position, H-12, % Wd 1 1 l l l Figure 5.3-11
e = 5.4 COEFFICIENTS OF REACTIVITY 5.4.1 TEMPERATURE COEFFICIENT The temperature coefficient of reactivity is defined as the fractional change in the reactivity of the core per unit change in core temperature. Temper-ature coefficients were calculated and measured for various soluble poison concentrations and core temperatures. Calculated and measured temperature coefficients are plotted in Figures 5.4-1, 5.4-2 and 5.4-3 for moderator temperatures of 250, 400 and 532 F. These curves also contain the calculated moderator coefficient of reactivity which is de-fined as the fractional change in the reactivity of the core per unit change in moderator temperature. The temperature coefficient was calculated for various power levels and measurements of the temperature coefficient were made at each of the major test plateaus during the power escalation test program. The measurement method used at power was to effect an approximate 5 F change in the reactor coolant temperature and observe the resulting reactivity change by recording the change in the position of the inserted control rod group and converting this change to reactivity. Calculated and measured temperature coefficients at power are given in Table 5.4-1. 5.4.2 POWER DOPPLER COEFFICIENT The power doppler coefficient relates the change in core reactivity to a corresponding change in fuel temperature. Calculations of the power doppler coefficient were made using a three-dimensional PDQ code with thermal feed-back. Measurements of the power doppler coefficient were conducted at each of the major test plateaus during power escalation. The measurement technique used was similar to that described for the temperature coefficient at power, except that the power level was increased by about 5% FP and the core average reactor coolant temperature was maintained at 579 F. Ihe determination of the power doppler coefficient then used the measured change in the controlling control rod group position, converted,to an equivalent reactivity value, and the measured change in reactcr, power, determined by using the recorded outputs of the power range nuclear instru-mentation channels. The acceptance criterion for the measured power doppler coefficient was that the coefficient must be more negative than -0.55 x 10-4 ok/k/%FP. All measured coefficients were between -0.96 x 10-4 Ak/k/%FP and -1.17 x 10 -4 ak/k/%FP.
'n s
, 5.4-1
e ? TEMPERATUl'E COEFFICIENT OF REACTIVITY AT POWER
)
Power Boron TemperatgreCoefficient Level Concentration (10- ak/k/ F) (% FP) (ppm) Calculated Measured 16 1335 +0.59 +0.61 41 1247 +0.30 +0.38 76 974 -0.40 -0.18 95 930 -0.54 -0.39 i l I l I 4 l r Table 5.4-1
~ - , _ , _ _ , . _ _ ,
O A TEMPERATURE AND MODERATOR COEFFICIENTS OF REACTIVITY VERSUS BORON CONCENTRATION AT 2500F e 800 PSIG 1i. : 1. i[ :Y E ,(__,_j_,,_ h ~i5lSd$.EM E;5hE5 5hN [
~ ' ]EEEEll{iE.g::iy -
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. Temperature Coefficient EiEs@s== E2j5MiIsMEEiiEHg=E . , ,
o ;_[ Moderator Coefficient El-\55 :~# p lifdfEI5 151EM55M
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Boron Concentration, ppm Figure 5.4-1
TEMPERATURE AND liODERATOR COEFFICIENTS # OF REACTIVITY VERSUS BORON CONCENTRATION AT 400 F 1500 PSIG
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? Temperature Coefficient
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3. See 956 AdSO /AGO /ASO /580_L_- .l . .: _../_40.. 0.- . Boron Concentration, ppa Figure 5.4-2 ( t
TEMPERATURE AND MODERATOR COEFFICIENTS OF REACTIVITY ~ VERSUS BORON CONCENTRATION AT 532 F, 2155 PSIG
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