ML19322C171
| ML19322C171 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/09/1974 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 8001090579 | |
| Download: ML19322C171 (222) | |
Text
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', DUKE POWER COMPANY
. OCONEE NUCLEAR STATION UNIT 2 DOCKET No. 50-270 LICENSE No. DPR-47 STARTUP REPORT l
4 AucusT 9, 1974 N. <
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D TABLE OF CONTENTS 4
t e Section Pye, e
1.0 INTRODUCTION
AND
SUMMARY
1.1 -1
- .1 INTRODUCTION 1.1-1 1.2
SUMMARY
1.2-1 1.2.1 GENERAL. 1.2-1 1.2.2 INITIAL FUEL LOADING 1.2-1 1.2.3 TESTING PRIOR TO POWER ESCALATION 1.2-1 1.2.4 POWER ESCALATION TESTS -1.2-2 2.0 INITIAL FUEL LOADING 2.0-1 1.0 TESTING PRIOR TO POWER ESCALATION 3.0-1 3.I REACTOR COOLANT PUMP FLOW TEST 3.1-1 3.L.1 PURPOSE 3.1-1 3.1.2 TEST METHOD 3.1-1 3.1. 3 EVALUATION OF TEST RESULTS 3.1-1 3.
1.4 CONCLUSION
S 3.1-1 3.2 REACTOR COOLANT FLOW COASTDOWN TEST 3.2-1 3.2.1 PURPOSE 3.2-1
, 3.2.2 TEST METHOD 3.2-1
\ 3.2.3 EVALl* ION OF TEST RESULTS 3.2-1 3.2.4 CONC ,SIONS 3.2-1 3.3 CONTROL ROD DRIVE DROP TIME TEST 3.3-1 3.3.1 PURPOSE 3.3-1 3.3.2 TEST METHOD 3.3-1 3.3.3 EVALUATION OF TEST RESULTS 3.3-1 3.
3.4 CONCLUSION
S 3.3-2 3.4 ZERO POWER PHYSICS TEST 3.4-1 3.4.1 PURPOSE 3.4-1 3.4.2 TEST METHOD 3.4-1 3.4.3 EVALUATION OF TEST RESULTS 3.4-2
- 3. 4. 3.1 Initial _ Criticality 3.4-2
- 3. 4. 3. 2 Nuclear instrumentation Overlap ! 3.4-3 3.4.3.3 "All Rods Out" Critical Boron Concentration '
3.4-3
- 3. 4. 3.4 - Control Rod Group Worths 3.4-4 3.4. 3. 5 Soluble Poison Worths 3.4-5 3.'4.3.6 Ejected Control Rod Worths 3.4-5
- 3. 4. 3. 7 Stuck Control Rod Worth 3.4-6 13.4 . 3.8 Temperature Coefficiint of Reactivity 3.4-6 9 3.
4.4 CONCLUSION
S .
3.4-6
-3 4.0 POWER ESCALATION TESTS 4.0-1 4.I NUCLEAR INSTRUMENTATION CALIBRATION AT POWER 4.1-1
. 4.1.1 PURPOSE 4.1-1 C 4. I '. 2 TEST MEDIOD 4.1-1 4.1.3 EVALUATION OF TEST RESULTS _4.1-2:
4.-l.4 CONCLUSIONS 4.1-2 t
Section Page 4
l 4.2 BIOLOGICAL SHIELD SURVEY 4.2-1 4.2.1 PURPOSE 4.2-1
, 4.2.2 TEST METHOD 4.2-1
- 4.2.3 EVALUATION OF TEST RESULTS 4.2-1 4.
2.4 CONCLUSION
S 4.2-1 4.3 REACTIVITY COEFFICIENTS AT POWER 4.3-1 4.3.1 PURPOSE 4.3-1 4.3.2 TEST METHOD 4.3-1 4.3.3 EVALUATION OF TEST RESULTS 4.3-1 4.3.4 -CONCLUSIONS 4.3-2 -
, 4.4 CORE POWER DISTRIBUTION 4.4-1 4.4.1 PURPOSE. 4.4-1
- 4.4.2 TEST METHOD 4.4-1 4.4.3 EVALUATION OF TEST RESULTS 4.4-1 4.4.3.1 General 4.4-1 4.4.3.2 Steady-State, Equilibrium Xenon Distributions 4.4-2 4.4.3.3 Minimum DNBR and Maximum LHR Calculations' 4.4-2 4.4.3.4 Quadrant Power Tilt and Axial Power Imbalance 4.4-4 4.
4.4 CONCLUSION
S 4.4-5 4.5 ROD WORTH AT POWER 4.5-1 4.5.1 PURPOSE , 4.5-1 4 4.5.2 TEST METHOD 4.5-1 4.5.3 EVALUATION OF TEST RESULTS 4.5-1 4.5.3.1 Acceptance Criteria 4 . 5-1 . -
4.5.3.2 Integral Rod Worth at Power 4.5-2 4.5.3.3 Differential Rod Worth at Power 4.5-2 j 4.
5.4 CONCLUSION
S 4.5-2 4.6 POWER IMB.. LANCE DETECTOR CORRELATION -4.6-1 4.6.1 PURPOSE 4.6-1 4.6.2 TEST METHOD -4.6-1 4.6.3 EVALUATION OF TEST RESULTS 4.6-2 4.
6.4 CONCLUSION
S 4.6-2 4.7 NSSS HEAT BALANCE 4.7-1 4.7.1 PURPOSE 4.7-1 4.7.2 TEST METHOD 4.7-1 4.7.3 EVALUATION OF TEST RESULTS 4.7-1 4.7.3.1 Primary and Secondary Heat Balance Calculation ( 4.7-1 j 4.7.3.2 Reactor Coolant Flow Determination 4.7-2 4.
7.4 CONCLUSION
S 4.7-2 4.8 UNIT LOAD STEADY-STATE TEST 4.8-1
- 4.8.1 PURPOSE 4.8-1
( 4.8'2
. TEST METHOD 4.8-1
- 4.8.-3 . EVALUATION OF. TEST RESULTS 4.8-2 i( 4.8. 4 - CONCLUSIONS '4.8-2 4.9 UNIT LOAD TRANSIENT TEST 4.9-1.
l 4.9.1 PURPOS E 4.9-1 m' 4.9.2 TEST METHOD 4.9-1
, , 4.9.3 EVALUATION OF TEST RESULTS 4 .9-2 4.9.3.1 Integrated Control System Transient Test at 40 %FP 4.9-2 l
(" 4.9.3.2
'4.9.~4 Integrated Control System Transient Test at 75 %FP CONCLUSIONS 4.9-2 4.9-2 11
Section Pace c
, 4.10 PSEUDO CONTROL RLD EJECTION TEST 4.10-1 4.10.1 PURPOSE 4.10-1 4.10.2 TEST METHOD 4.10-1
- 4.10.3 EVALUATION OF TEST RESULTS 4.10-1
. 4.
10.4 CONCLUSION
S 4.10-2 4.11 DROPPED CONTROL ROD TEST 4.11-1
't.ll.1 PURPOSE 4.11-1 4.11.2 TEST METHOD 4.11-1 4.11.3 EVALUATION OF TEST RESULTS 4.11-2 4.
11.4 CONCLUSION
S 4.11-2 APPENDIS A RE-ESCALATION TO POWER i
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/ 1.0 , INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
. On October 8, 1973, the Atomic Energy Commission issued Facility Operating License DPR-47 to Duke Power Company for Oconee Nuclear Station, Unit 2.
Dua first fuel assembly was inserted lato the core on October 9,1973, and initial fuel loading was completed on October 13, 1973. On November 11, 1973,.0conee Nuclear Station, Unit 2, successfully achieved initial criti-cality.
Zero Power Physics testing, which commenced on November 11, 1973, was suc-cessfully completed on November 29, 1973. This program was conducted primarily at reactor coolant temperatures of 300 and 5320F.
Initial power level escalation was conducted on December 1,1973, 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 tallows:
Power Level (Percent of Full Power - %FP) Date 15 December 1, 1973 40 December 13, 1973 i 75 December 25, 1973 ,
100 June 19, 1974 On January 4, 1974, during power escalation testing at 75 %FP, a malfunction in the switchyard caused a turbine / reactor trip. Following this trip, a foreign object was detected in the bottom of the reactor vessel. After evaluations by Duke and -he Babcock & Wilcox Company, the nuclear steam supply system vendor, ans with Atomic Energy Commission concurrence, power operation was resun,d. While at 15 %FP, a seal failure in Reactor Coolant Pump 2B2 occurred. Since repairs required the unit to be shut down for an extended period of time, it was decided to remove the fuel from the reactor and inspect the r.eactor vessel interior and the vessel internals.
Due to,this incident, further unit operation was delayed until May 23, 1974.
Details of these events were reported to the AEC cn January 10,14, {and 17, February 1, and April 12, 1974. A summary of retesting performed between fuel relcading and the resumption of the power escalation testing program is given in Appendix A of this report.
This report addresses unit startup and power escalation testing through 0500
, hours on June 24, 1974; at that time, all or part of several power escalation
, tests remained to be completed as follows:
(a) Turbine / Reactor Trip Test o (b) Unit Loss of Electrical Load
, (c) Unit Loss of Control Room (d) Unit Load Transient Test - 100 %FP portion 1.1-1
i Following completion of the above items, appropriate information, supplemen-
- tary to this report, will be compiled and submitted concerning the results of the testing listed.
This report is prepared and submitted in accordance with Technical Specifi-cation 6.6.1.1 and addresses unit startup and power escalation testing and the results thereof with regard to Oc. nee Nuclear Station', Unit 2.
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1.1-2
1.2
SUMMARY
e l.2.1 G ENERAL -
Oconee Unit 2 is the second of a series of nuclear steam supply systems, designed by the Babcock and Wilcox Company and rated at approximately 871 MWe (net), to be placed into service.
As of 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, on June 24, 1974, the unit has been ope .:ted at. power levels up to, and including, 100 %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 adversely 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 evaluation of unit startup and power es-calation testing, it is felt that the unit may be safely operated at full rated power.
Oconee Unit 2 startup and power escalation testing, as addressed by the various major sections of this report, is summarized below.
1.2.2 INITIAL FUEL LOADING 4 Fuel loading was initiated on October 9,1973, and was completed on i October 13, 1973. During fuel loading, two fuel assemblies did not seat on
[ the initial attempt and were returned to the spent fuel pool for inspection.
\ After inspection had revealed that the assemblies were in good condition i
and when adjacent fuel assemblies had been loaded, these two assemblies were i inserted with no further problems. Initial fuel loading at Oconee Unit 2 i
was completed in four days; and in general, the fuel loading proceeded in an orderly manner.
4 1.2.3 TESTING PRIOR TO POWER ESCALATION Following initial fuel loading of Oconee Unit 2 and prior to power escalation, the Reactor Coolant Pump Flow Test, the Reactor Coolant Flow Coastdown Test, the Control Rod Drive Drop Time Test, and the Zero Power Physics Test were conducted. In all cases, applicable Technical Specification requir.ements were i
met. A brief summary of each of these tedts follows:
{
(a) Reactor Coolant Pump Flow Test .
The measured reactor coolant flowrates_were within the range of acceptable values and provided adequate margin to both the maximum and minimum allowable flowrates, e
(b) Reactor Coolant Flow Coastdown Test
. For all test conditions, reactor core coolant flow versus time exceeded the minimum applicable criteria.
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. (c) Control Rod Drive Drop Time Test e
The rod drop times were well below the acceptance criteria stated in Secticn 4.7 of the Technical Specifications, of 1.66 seconds at full flow ana 1.40 seconds at no flow conditions. 3 (d) Zero Power Physics Test
~
Zero Power Physics Testing commenced on November 11, 1973,with initial criti-cality occurring at I'+10 hours. The Zero Power Physics Test was completed on November 29,1973,with good agreement between measured and predicted results.
1.2.4 POWER ESCALATION TESTS Following the completion of Zero Power Physics Testing, initial power es-calation commenced on December 1, 1973,with the first electrical power produced at 1435 on December 5,1973. The power escalation test program was conducted at four major test plateaus of 15, 40, 75, and 100 percent full power with minor testing performed at intermediate power levels as required by the con-trolling prccedure for power escalation. The average daily power level history.during the power escalation test program is shown in Figure 1.0-1.
Power level escalations occurred as required when testing was satisfactorily -
completed. A brief summary of each of these tests follows:
(a) Nuclear Instrumentation Calibration at Power The power range channels were calibrated to within two percent of total power several times during the startup program. Total power was determined by a heat balance, and imbalance was determined using the incore instrumentation as input to the calibration. These calibrations were required due to power level, boron, and/or control rod configuration changes during the program.
The calibration procedure has, therefore, been thoroughly tested and has proven extremely satisfactory. Acceptance criteria for nuclear instrumen-tation calibration at pcwer were met in all instances.
(b) Biological Shield Survey The maximum radiation level found during power escalation testing ups 11 mrem /hr gamma.which was well within the acceptance criteria of 100 inRem/ hour in all accessible areas.
(c) Reactivity Coefficients at Power The measured results indicate that the moderator coefficient will be negative
, during operation at or above 95 %FP.
Analyzed data for the power doppler coef ficient versus power level indicate that the least negative coefficient is -0.78 x 10-4 Ak/k/%FP. The total
- power _ doppler deficit from 0 to 100 %FP from this measured data is estimated a
to be -0.92 %Ak/k.
v 1.2-2
, (d) Core Power Distribution o
Comparison of the five equilibrium xenon core power distributions taken at 40, 75, and 100 %FP at various control rod positions with PDQ-7 calculations showed that the measured values for the maximum radial peaking factor and
- for the maximum radial times axial peaking factor were within 5 percent and 7.5 percent, respectively of the calculated values.
Minimum DNBR margins sere calculated for each core power distribution with respect to the limiting DNBR of 1.55. All cases analyzed resulted in a sub-stantial DNBR margin. The margins to the limiting linear heat rate were calculated for both the central fuel melt limit and the LOCA ltait. Maximum linear heat rate analyses indicated substantial margins to the limiting l i noa r heat ra t es .
Results of the quadrant tilt and the axial power imbalance calculations indi-cated that all tilts determined during normal power operation were well within the Technical Specification limit of four percent and that the Reactor Pro-tective System would provide suf ficient protection against exceeding DN3R and LHR limits when the delta flux amplifier had a gain factor of 3.3.
(e) Rod Worth at Power Integral rod worths at full power developed from Zero Power Physics results for part-length rods both 100 and 35 %wd predicted rod worth adequately.
Comparison of predicted worths using these curves against measured results using the fast insertion / withdrawal technique gave marginal results due to the fact that the measured results were not done with the part-length rods at either 100 or 35 %wd.
Differential rod worth measurements at power using the fast insertion /
withdrawal method gave minimum and maximum differential rod worths well within the limits of the acceptance criteria.
(f) Power Imbalance Detector Correlation Results of power imbalance detector correlation testing at 40 and 75 %FP are listed below: i (1) The slope of the equation relating tne incore and out-of-core imbalance was independent of the technique used to produce the imbalance.
(2) The imbalance trip envelope as set in the Reactor Protective System will protect the reactor core from exceeding LHR and DNBR limits when a gain factor of 3.30 is set in the circuit.
(3) Extrapolating the minimum DNBR and maxim. LHR to the overpower trip set-point (105.5 %FP) yields values of 2.67 and 13.4 kw/ft respectively.
e 1.2-3
. (g) Nuclear Steam Supply Syster Heat Balance All primary and secondary heat balance calculations met their respective ac-ceptance criteria, which are that all calculated primary heat balances are within 5 percent of each other and that all calculated secondary heat balances
- are within 5 percent of each other. Preliminary calculations of reactor coolant flow rate using normal unit instrumentation for flows, temperature, and pressures indicated a flow rate of 106.4 percent of design during initial escalation and 109.5 percent of design during re-escalation at 75 %FP.
(h) Unit Load Steady-State Test The average of the measured unit parameters during the test period fell within their respective minimum and maximum limits, except for steam generator outlet steam pressure. Analysis of unit parameter stability indicates that all variables are relatively stable, even though feedwater flow did not meet the stated acceptance criteria. However, neither item has any adverse effect on the safe operation of the unit.
(i) Unit Load Transient Test From analysis of all test data taken to date, the following conclusions may be <
made:
(1) All transients were performed without exceeding the limits of the unit Technical Specifications.
(2) All transients were completed without causing the Reactor Protective System to actuate.
(3) During the transients, none of the acceptance criteria parameters varied outside the acceptance band.
I (j) Pseudo Control Rod Ejection Test The measured worth of the most reactive control rod was found to be 0.34
%Ak/k, which is less than the maximum value of 0.50 %Ak/k specified in Section 3.5.2 of the Technical Specifications. j i
Analyzed core power distribution and thermal hydraulic data indicated a large perturbation to the steady-state core power distribution, as was predicted, with a maximum linear heat rate of 8.48 kw/f t, a minimum DNBR of 5.22 and a maximum power peak of 3.25 measured with the rod in core position H-08 at 100 %wd.
- (k) Dropped Control Rod Test Upon analysis of the' Dropped Control Rod Test data, the following conclusions were made:
(1) Analyzed core power distribution and thermal hydraulic data indicated g_j suf ficient margin to minimum DNBR and maximum linear heat rate limiting criteria. The perturbation to the steady-state power distribution was as expected with a maximum tilt of 12.64 percent.
l.2-4
. +
. (2) The measured worth of the control rod which produces the most adverse
. thermal effects in the core, if it is inadvertently dropped, was found to be 0.13 %Ak/k.
(3) The Integrated Control System accurately detected the asymmetric control rod and initiated the appropriate alarms.
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A
INITIAL FUEL LOADING
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2.0
. Fuel loading was initiated with the insertion of fuel assembly 2C52 into the core on October 9,1973, and was completed with the loading of 2C53 on October 13, 1973. Figure 2.0-1 depicts the final configuration of the core at the conclusion of initial fuel loading.
Neutron count rate was monitored throughout the fuel loading sequence utilizing the unit's two nuclear instrumentation source range channels, NI-l and NI-2, and two temporary incore detectors. Independent calculations and plots of the inverse neutron count rate ratio were obtained from the output of these detectors. During the first several fuel loading steps, the inverse neutron count rate ratio from an incore detector gave an overly conservative estimate of the number of assemblies required for criticality and successive fuel assemblies were lowered in discrete increments into the active core region. Nine assemblies were loaded in this manner. Due to this experienc.
it was felt that the fuel loading procedure could be revised to take into account the data from all neutron detectors. The revision stated that, prior to loading the second neutron souce, if the assembly to be loaded was in a direct line .between the source and the detector indicating the most con-servative predication of the number of assemblies to criticality, and if the two next most conservative detectors indicated that the next assembly could be loaded safely, then the entire assembly could be loaded. This revision significantly decreased the time required to complete fuel loading without adversely affecting the safety of the operation.
(" During fuel loading two fuel assemblies did not seat on the initial attempt and were returned to the spent fuel pool for inspection. After inspection had revealed that the assemblies were in good condition, and as soon as all adjacent fuel assemblies had been loaded, these two assemblies were inserted with no further problems.
Initial fuel loading at Oconee Unit 2 was completed in fcur days; and, except for the items noted :above and minor equipment probim * , the fuel loading pro-ceeded in an orderly manner.
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2.0-1
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FINAL FUEL LOADIN'r DISTRIBUTION FOR OCONEZ 2, CORE 1 6
SPENT FUEL %
CANAL '
A 2C54 2C47 2C15 2C30 2C35 11e B
C37 2C09 2005 2342 2C50 2B36 2C07 2C12 2C40 009 C51 Bl4 C54 B17 C57 035 C 2C23 2C22 2A32 2B43 2A27 '2B59 2A01 2B18 2A39 2C32 2C49 B30 C23 339 CS2 B41 C56 B26 C59 B28 D 2C43 2C04 2A10 2335'2A51 2B53 2AS4 2360 2A49 2B06 2A11 2C51 2C42 B24 C21 B56 A04 B46. C53 B47 A08 B58 C60 B29 E 2C13 2A22 2B19 2A34 2320 2A29 2B14 2A19 2B26 2A14 2B03 2A56 2C34 C22 B61 C20 B60 C19 B51 CSS 364 C58 B66 C61 y_ 2001 2C24 2B24 2A05 2B21 2A25 2B16 2A46 2B04 2A06 2B47 2A08 2B30 2C58 2C53 C29 B27 A03 357 C17 312 C16 B19 C43 367 A07 B42 C50 G 2C33 2Bf6 2A23 2B44 2A31 2308 2A28 2B12 2A03 2340 '2A20 2B54 2A38 2B39 2C17 B20 C27 B23 C18 B10 C14 B06 C13 B01 C44 B33 C47 B08 H-2C21 2C26 2B13 2A47 2B51 2A55 '2B11 2B57 2B10 2A35 2B37 2A18 2B50 2C48l2C57 116 C30 B48 C26 B70 C15 C11 B03 BOS C10 349 C45 B21 C49 l 11c K-2C18 2B38 2A40 2309 2A17 27,15 2A45 2B49 2A04 2B25 2A36 2B45 2A52 2352 2C27 Bil C18 B31 c25 . uns c?9 nn7 ene n11 ent Bai c46 nna L_ 2C44 2C10 ,2332 2A26 2307 2A50 2333 2A09 2B05 2A44 2B27 2A16 2B22 2C59 2C39
. C31 'B40 A05 B68 C24 B18 C09 B16 C04 355 A01 B37 C48 M 2C19 2A21 2B46 2A41 2341 2A07 2B48 2A43 2B34 2A15 2B02 2A37 2C12 C42 B54 C39 B62 C36 B69 C06 B59 C03 B63 CO2 N 1C25 2C56 2A48 2B31 2A53 2B27 2A30 2B58 2A24 2B01 2A12 2C28 2C29 .
B25 C41 B65 A06 B.'i? C34 B45 A02 B53 C01 335 0 2C60 2C55 2A13 2B29 2A02 2B55 2A42 2B17 2A33 2C14 2C46 B38 C40 B36 C37 B44 C33 B22 C07 B34 P 2C52 2C38 2C31 2B28 2C41 2B61 2C02 2CC3 2C16 021 C38 B15 C35 B09 C32 004 P '?C45 2C20 2C08 2C36 2C06 l 117
- l l l l l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15
- 2A01 thrcugh 2A56 = 2.06 vt % Fuel Assemblies 2B01 through 2B61 = 2.72 vt 5 Fuel Acac 'les -
2001 through 2C60 = 3.05 v: % Fuel Assemblies Col thrcugh C61 = Centrol 20d Assemblies A01 thrcugh A08 = Axial Pever 2h vin- T ri Asse=blies B01 through B70 = Burnable ?cisen Asse:blies (Excluding B50 and B52)
, 00k,009,021,035 = orifice Rod Asse=blies 4, 115.116,117,118 = special Orifice Rod Asse=blies (Irradiation Samples) v Figure-2.0-1
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, 3.0 TESTING PRIOR TO P0k'ER ESCALATION Following initial fuel loading of Oconee Unit 2 and prior to power escalation, the Reactor Coolant Pump Flow Test, the Reactor Coolant Flow Coastdown Test, the Control Rod Drive Drop Time Test, and the Zero Power Physics Test were i , conducted. This section of the report presents the results of these tests.
In all cases, applicable Technical Specification requirements were met.
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.o 3.0-1
a e 3.1 REACrdR COOLANT PUMP FLOW TEST ,
'. 3.1.1 PURPOSE Prior to power escalation, the Reactor Coolant Pump Flow test was performed
, 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.
3.1.2 TEST NETHOD 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 values. From the , loop flowmeter AP cell indications, flows were calculated as follows:
_ _ y Vc Flow = C g AP g-
- a Where: C = Flow Meter Coefficient = 397,100 f
Vc = Specific volume at reference conditions (68 F, 14.7 psia)
Vs = Specific volume at r,scem conditions 3.1.3 EVALUATION OF TEST RESULTS Table 3.1-1 gives the minimum and maximum allowable flow rates for four diff-erent pump combinations, along with the measured flowrates for each listed con-dicion. It can be seen that the measured flowrates are within the acceptance criteria.
3.1. 4 CONCLUSIOT The measured reactor coolant flowrates were within the range of acceptable values, indicating that power operation could safely commence.
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3.1-l~
REACTOR COOLANT FLOW FOR VARIOUS FUMP COMBI!iATIOIS a
Minimu::t Ma:dmum Acceptable Acceptable Measured Flow Rate At- Flow Rate At . Flow Rate at NP 2155 psig, 532 F 2155 psig, 532 F 2155psig, 532 F Cembination (106 lbm/hr) 6 (10 lbm/hr) 6 (10 lbm/hr)
Four Pumps 138.4 151.h 146.6 Three Pumps, 103.2 151.h 111.7 Two Pumps, One Loop 63.h 151.h 76.4 One Pu=p Each Loop 67.8 151,h 76.4 j.
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. 3.2 REACTOR COOLANT FLOW COASTDOWN TEST 3.2.1 PURPOSE
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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. .
3.2.2 TEST METHOD Various combinations of reactor coolant pumps were operated and steady-state data acquired. Subsequently, all or a portion of the operating pumps were tripped and data were recorded during the ensuing reactor coolant flow tran-sient. Steady-state data were again taken following the flow transient.
Reactor coolant flow, at various times during the coastdown transients, was determined from loop flow meter data according to the equation given in Section 3.1.
3.2.3 EVALUATION OF TEST RESULTS Figure 3.2-1 shows the minimum acceptable reactor coolant flows versus time for both single-pump and multi-pump coastdowns. Measured reactor core coolant flows versus time are presented in Figures 3.2-2 through 3.2-4 for typical coastdowns from four-pump initial operating conditions.
3.
2.4 CONCLUSION
S For all test conditions, reactor coolant flow versus time exceeded the minimum applicable acceptance criteria.
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, 3.3 CONTROL ROD DRIVE DROP TIME TEST 3.3.1 PURPOSE
- 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.
3.3.2 TEST METHOD This test was conducted at various combinations of reactor coolant flow, pressure and temperature as follows: Test Condition Flow Pressure Tempera ture 1 No Flow 3,350 psig < 400 F 1 1700 psig 2 One Pump Each Loop 3,350 psig < 400 F 1 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 was 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 indicator 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. The total drop time for each rod by core location in milliseconds yas re- , corded along with the date, time, number of reactor coolant pumps pperating, reactor coolant flow, recctor coolant temperature, and reactor coolant pressure. 3.3.3 EVALUATION OF TEST RESULTS . An analysis of the drop times indicated that at Test Condition 1
. rod B-8 was fastest at 1.114 seconds and rod 0-9 was the slowest at 1.175.
seconds For the second trip B-8 was fastest at 1.122 seconds and D-12 was slowest at 1.171 seconds. . For Test Condition 2, rods B-8 and C-ll were fastest at 1.183 seconds,
- and 0-9 was slowest at 1.243 seconds. For.the second trip B-8 was fastest
( at 1.179 seconds, and F-10 was slowest at 1.238 seconds. 3.3-1
For Test Condition 3, rod C-11 was fastest at 1.263 seconds, and F-10 was slowest at 1.349 seconds. For the second trip C-11 was fastest at 1.251 seconds, and M-9 was slowest at 1.342 seconds. Rods M-9 and F-10 (the slowest) and C-11 (the fastest) were dropped an additional 10 times and produced drop
. times within 33 ms and 16 ms, respectively.
3.
3.4 CONCLUSION
S The rod drop times were well below the acceptance criteria stated in Section 4.7 of the Technical Specifications, of 1.66 seconds at full flow and 1.40 seconds at no flow conditions. Also, as would be expected, all drop times under flow conditions were longer than under no flow conditions. e I l I 9 9 3.3-2
m, _ - - - - r . 3.4 ZERO POWER PHYSICS TEST 3.4.1 PURPOSE e The purpose of the Zero Power Physics Test was to verify the nuclear design
, parameters used in the safety analysis, the Technical Specification limits, and the operational parameters. This test was conducted after initial fuel loading and before escalation into the power range. Testing was performed at two major temperature plateaus, 300 F and 5320F. The test included the following measurements:
(a) Initial criticality (b) Nuclear instrumentation overlap between the source and intermediate range (c) "All Rods Out" critical boron concentration (d) D'ifferential and integral rod worth (e) Differential and integral boron worth (f) Ejected control rod worth (g) Stuck rod worth (h) Temperature coefficients as a function of boron concentration 3.4.2 TEST METHOD Initial criticality was achieved by control rod withdrawal and boron dilution of the Reactor Coolant System after the system had been heated up to 3000F using one reactor coolant pump in each loop. During the initial approach to criticality a plot of inverse neutron count rate ratio cersus boron con-centration and time was maintained by using channels NI-l and NI-2 of the units' nuclear instrumentation. After achieving criticality, nuclear power was increased and the source and intermediate range nuclear instrumentation
- overlap was verified to be in excess of one decade. During this same increase in power, the nuclear heatup point was determined and the upper power limit for Zero Power Physics testing was established.
Physics testing at 300 F, 800 psig was then begun. The test program at this testing plateau included the following measurements: (a) The "All Rods Out" boron concentration (b) The temperature coefficient of reactivity at three boron concentrations (c) The differential and integral rod worth of Control Rod Groups }, 6, and part of Group 5 by the rod / boron swap technique (d) The integral rod worth of Control Rod Groups 1, 2, 3, and 4 and part of Control Rod Group 5 by the rod drop technique (e) Differential boron worth for large changes in boron concentration
. Upon completion of the above measurements, the Reactor Coolant System was , heated up and pressurized to 532 F and 2155 psig, respectively. At this .
temperature and pressure, physics testing was conducted including the ! following measurements: (
. (a) The "All Rods Out" boron concentration (b) The temperature coef ficient of reactivity at four boron concentrations
(- (c) The differential and integral. rod worth of Control Rod Groups 8, 7, 6, 5, and part of Group 4 by the rod / boron swap technique 3.4-1 L .
(d) The integral rod worth of Control Rod Groups 1, 2, and 1 and part of Control Rod Group 4 by the rod drop technique (e) Differential boron worth for large changes in boron concentration (f) Stuck rod worth by the rod drop technique
* (g) Ejected rod worth by the rod drop technique 3.4.3 EVALUATION OF TEST RESULTS
- 3. 4. 3.1 Initial Criticality Initial criticality was achieved on November 11, 1973, at 1410 hours at reactor coolare conditions of 300 F and 800 psig. Control Rod Groups 1 through 6 and Group 8 had previously been fully withdrawn to 100 percent and Control Rod Group 7 positioned at 75 percent withdrawn, with the Reactor Coolant System boron concentration at 1879 ppmB.
After positioning of the control rods, deboration commenced and continued until initial criticality was achieved at 1542 ppmB. The procedure for the approach to initial criticality was as follows: (a) Control rod group withdrawal: Groups 1-4 100% withdrawn Group 5 100% withdrawn Group 6 100% withdrawn
! Group 8 100% withdrawn Group 7 75% withdrawn i
(b) Deboration from 1879 ppmB to 1665 1 50 ppmB using a feed and bleed rate of approximately 70 gallons per minute. (c) Deboration from 1665 1 50 ppmB to the critical boron concentration using a feed and bleed rate of approximately 20 gallons per minute. Control Rod Group 7 inserted to maintain criticality if required. Throughout the approach to criticali*y, two plots of the inverse neutron count rate ratio were maintained independently. At the end of each reactivity addition, three count rates were taken from each startup range neucron detector. The ratio of the initial average count rate to the average count rate at the end of each reactivity addition was then plotted. As indicated above, deboration was carried out in essentially two steps. First, deboration from an initial boron concentration greater than 1800 ppmB
. to 1665 1 50 ppmB'was accomplished'using a 70 gpm-letdown and makeup rate.
This step brings the Reactor Coolant System boron concentration to within 100 ppmB of the predicted critical boron concentration (1565 ppmB). Boron concentration changes during this period were determined by taking reactor coolant samples at least every 30 minutes. The second deboration step was
. accomplished at a lower makeup and letdown rate of 20 gpm. Again, Reactor Coolant System boron samples were taken at least every 30 minutes, and inverse count rate ratio plots were maintained versus boron concentration
.and versus time. Deberation was terminated upon initial ' criticality and 3.4-2
9 Control Rod Group 7 was inserted as necessary to maintain the reactor
~
critical. Plots of inverse neutron count rate ratio versus time and versus boron concentration during the approach to initial criticality are presented in Figures 3.4-1 through 3.4-4.
. Initial criticality occurred at a Reactor Coolant System boron concentration of 1542 ppmB and was achieved in a safe and orderly manner. Analyzed results indicate goed agreement between predicted and measured criticality endpoints.
3.4. 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 ranga and intermediate range must be observed. This means that before the source range count rate equals 105 cps the intermediate range must be on scale. If the one decade is not observed, the approach to 10-9 amps on the intermediate range cannot be continued until the situation has been corrected. In addition, the following number of channelt must be in operation for the test program to continue beyond initial critittlity. Channels Available Minimum Operating
, Source Range NI 2 2 Intermediate Range NI 2 2 To satisfy the above overlap requirements, after initial criticality was ,
reached, core power was slowly increased until the intermediate range channels came on scalc. Detector signal response was thus recorded for both the inter-mediate and source range channels. This was repeated for two more decades until the -ource range signals approached 106 cps. t l The results of .he initial nuclear instrumentation overlsp data at 300 F, 1500 psig are plotted on Figure 3.4-5 and are normalized to an estimated core power level of 10 MWt at an intermediate range signal of 7 x 10-8 amps. This estimate was made by observing a heatup rate of 36 F/hr at this signal level at 300 F. This heatup rate has been shown to be equivalent to a 10 MWt heat addition to the Reactor Coolant System. Also plotted on Figure 3.4-5 is tha intermediate range detector response at power. }
- i
, Examination of Figure 3.4-5 shows that the linearity, overlap, and absolute l output of both the intermediate and source range detectors are within specifi-cations and performing satisfactorily. The step change in the intermediate range signal between 300 F and 579 F is due to increased neutron leakage at the higher temperature level.
. 3.4.3.3 "All Rods Out" Critical Boron Concentration The "All Rods Out" critical' boron concentrations were measured at the two moderator temperature test plateaus of 300 and 532 F. The measurements were . made with Control Rod Group 7 partially inserted. The measured boron con-centrations were adjusted to the "All Rods Out" condition using the results
, [
~~
of rod worth measurements to determine the reactivity worth, in terms of l boron concentration, of the inserted control rods. 3.4-3
l l 1 The results are tabulated in Table 3.4-1. These results show that the measured boron concentrations compare quite favorably with the predicted l " results and are within the acceptance criterion of + 100 ppmB.
. 3.4.3.4 Control Rod Group Worths I
The Oconee Unit 2 initial control rod group configuration is shown in Figure 3.4-6. t Calculated and measured beginning of life (BOL) control rod group reactivity worths, for the normal withdrawal sequence were determined at the Zero Power Physics test plateaus of 300 and 532 F. The calculations were made using the PDQ-7 computer code with either a two or three-dimensional description of the j core. The rod / boron swap method was used to determine integral and dif- ! ferential worth for Control Rod Goups 7, 6, and part of 5 at 300 F and Control hod Groups 8, 7, 6, 5, and part of 4 at 532 F. This method consisted of
, establishing a boration or deboration rate and compensating for the change in reactivity by small step changes in rod group positions.
The rod drop method was used to determine the worth of control rod groups not meanured by boron swap. For each measurement, the reactor was adjusted to criti-cality with all of the control rod groups to be measured out of the core and ! at a power level near the Zero Power Physics test upper power limit. The control rod groups being measured were then tripped. 1 Based on experience with Oconee Unit 1, it was predicted that rod drop measure-ments on Oconee Unit 2 would yield values approximately 74 percent of the correct value when considerably more than 1 %Ak/k was being inserted. The Oconee Unit 2 results were consistent with these expectations. Table 3.4-2 4 compares the calculated, expected, and measured results for the rod drops at 300 and 5320F. The expected values were computed by applying a correction i factor of 0.74 to the calculated results. The results show that the measured values were within three percent of the expected value. The approximate magnitude of the correction was confirmed through an overlap in rod cali-bration by rod drop and boron swap at 532 F. In this particular case, the worth of Control Rod Groups 4 and 5, from 59.8 percent on Group 5 to 53.8 percent on Group 4 was determined using both the rod / boron swap technique and the rod drop results. , Table 3.4-2 gives the result of the rod drop with Control Rod Group'4'at 13.8
%wd as 4.70 %Ak/k. The second drop was made with Control Rod Group 5 at 59.8
%wd and gave a value of 5.57 %4k/k. The portion of the rod worth between these two positions was, therfore, measured by both rod / boron swap and rod drop (difference of the two drops). The worth obtained by rod / boron swap was
. 1.21 ik/k and by rod drop 0.87 %ak/k. In this case, the rod drop measured 72 percent of the value by rod / boron swap which agrees well with the expected 74 percent.
l < The results of both the calculated rod group worths and the measured group
, worths are tabulated in Table 3.4-3 for moderator temperatures Lof 300, 532,
- j. and 579 F. Based on -the results of the rod drop measurements _ considered l
/ above, it was decided to use the calculated worths as the best estimate of l
the worth of Groups' I through 4, as listed in . Table 3.4-3. In the case of 1 3.4-4 1
. . . . -~ . < s rod group worths with Control Rod Group 8 at 100 %wd at 532 F and 579 F, ~
the estimated worths listed assume the same percentage deviation between the calculated and measured worths as determined for the case with Control Rod Group 8 centered at 532 F. The results indicate good agree-
- ment between measured and calculated control rod group worths.
The results of the measured differential control rod group worths at 300 and 532 F are plotted in Figure 3.4-7 for Control Rod Group 8 positions of 35 %wd (centered in core) and 100 %wd. i The shape of the integral worth curves are shown in Figure 3.4-8 for 532 F, zero power conditions. These curves were calculated by integrating the dif-ferential curves given in Figure 3.4-7 and considering the change in integral worth shape that would occur between 300 and 532 F as negligible in the case of the Control Rod Group 8 at 100 %wd curve. Normalized integral worth curves were developed for full power, 5730F conditions by assuming the same per-centage deviation between measured and calculated results exists at these conditions as that determined for 532 F, zero power conditions. Plots of the normalized 579 F, full power integral worth curves are shown in Figure 3.4-9 for Control Rod Group 8 positions of 35 %wd and 100 %wd. 3.4. 3. 5 Soluble Poison Worths Measurements of the soluble poison differential worths were made at the two zero power test plateaus of 300 and 532 F. The measured values were deter-mined by summing the incremental reactivity values measured during the rod worth measurements over a known boron concentration range. Measured differential soluble poison worths are compared with predicted results in Table 3.4-4 and in Figure 3.4-10. The integral reactivity worth curves shown in Figure 3.4-11 were then established by integration of the measured differential boron worth curves and by reactivity balance calculations. The integral worth curve at an average coolant temperature of 579 F, obtained during power operation, is also included. In summary, all measured values at zero power were within 1.4 percent of the calculated worths. l
' }
- 3. 4. 3. 6 Ejected Control Rod Worths j Pseudo ejected control rod reactivity worths were measured at zero power
, conditions of 532 F, 2155 psig for two dif ferent control rods. The purpose of these measurements was to verify the safety analysis calculations relating to the assumed accidental ejection of the most reactive control rod during
, power operation. The acceptance criterion for these measurements is that the . reactivity worth of the most reactive control rod does not exceed 1.0 %Ak/k at 532 F, 2155 psig, zero power conditions.
The ejected rod worths at zero power were measured using the rod drop method.
. As stated above, this technique for measuring control rod worth is inherently inaccurate by at least i 25 percent for most reactors. The calculated and
( measured ejected rod worths are tabulated in Figure 3.4-12 which also 3.4-5 l
, gives the core location of each control rod measured. The maximum ejected control rod warth was determined to be 0.72.%Lk/k which insures that Technical Specification 3.5.2 will always be met at zero power.
3.4.3.7 Stuck Control Rod Worth The purpose of the stuck rod worth measurements at zero power were to verify that the calculated stuck rod worths are conservative compared to the measured results. The method used in measuring the simulated stuck rod worths involves taking the difference between a rod drop during which Groups 1 and 4 are essentially fully withdrawn and another rod drop during which the stuck rod remains fully withdrawn. The results of both calculated and measured stuck rod worth are given in Figure 3.4-13. The difference between the measured and calculated worth is attributed to the use of the rod drop measuremcnt method. This_does not, however, affect unit safety, since the maximum calculated stuck rod worth of 4.27 %Lk/k is used in determining available shutdown margin. 3.4. 3. 8 Temperature Coefficient of Reactivity The temperature coefficient of reactivity is defined as the fractional change in the reactivity of the core per unit change in core temperature. Tem-perature coefficients were calculated and measured for various soluble poison concentrations and core temperatures. The measurements were made by initially increasing and then decreasing the
- reactor coolant temperature by approximately 10 F. The change in the position of the inserted control rod group was recorded and converted to a change in reactivity.
The results of both the calculated and measured temperature coefficients are plotted in Figures 3.4-14 and 3.4-15 for core temperatures of 300 and 532 F. These curves also contain the calculated moderator coefficient of reactivity which is defined as the fractional change in the reactivity of the core per unit change in moderator temperature. All measured temperature coefficients of reactivity at the 300 F j and 532 F glateaus were within the acceptance criteria of + 0.4 xi 10-4 ak/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 3.1.7. J 3.
4.4 CONCLUSION
S
- Zero Power Physics Testing commenced on November 11, 1973, with initial criticality occurring at 1410 hours. The Zero Power. Physics Test was completed on November 29, 1973, with good agreement between measured and predicted results. A summary of each of the measurements performed during the Zero Power Physics Test is given below.
s.- 3.4 ,
a y .- -
-, . g . -p-I . < '(a) Initial Criticality Initial Criticality was achieved at 1410 hours on Sunday, November 11, 1973. The approach to criticality was performed in a safe and orderly i manner. Results indicated good agreement between predicted and measured o criticality boron endpoints.
(b) Nuclear Instrumentation Overlap . Nuclear instrumentation overlap was verified to be in excess of two decades between the source and intermediate range. The minimum ac-ceptable overlap is one decade.
- (c) "All Rods Out" Boron Concentration The measured "All Rods Out" boron concentrations were 1555 ppmB at 300 F and 1630 ppm 3 at 532 F as compared to the predicted values of 1571 and
] 1634 ppmB respectively. A maximum deviation of -1.0 percent from the f predicted values was observed. (d) Control Rod Group Worths Control rod-group integral and differential reactivity worths were calcu- . lated by rod / boron swap and rod drop measurements. The measurement of l rod worth by rod / boron swap at 300 and 532 F indicated good agreement i between the measured and calculated group worths. Good correlation was obtained for rod drop measurements when ccmpared with expected measured worths. (e) Soluble Poison Worths Measured differential boron reactivity worths of 1.25 %ak/k/100 ppm at 1450 ppm and 300 F, and 1.08 %ak/k/100 ppm at 1434 ppm and 532 F were determined. Comparison of these values to the predicted values showed that the measured values were within 1.4 percent of calculated worths. (f) Ejected Control Rod Worths , The maximum ejected control rod worth was determined to be 0.72 %Ak/k, which insures that Technical Specification 3.5.2 w111'always be met at zero power. (g) Stuck Control Rod Worth The measured stuck rod worth of 1.63 %Ak/k was less than the calculated value due to the rod drop measurement method utilized. .This does not however, affect unit safety, since the naximum calculated stuck. rod worth of 4.27_%Ak/k is used in determining available shutdown margin. (h)' Temperature Coefficient.of Reactivity o o V Measured temperature coefficients of reactiv3ty at 300 F and 532 F were 4 Ak/k/0F of. the predicted within the acceptance criteria cf i 0.4 x 10
'3.4 . value. In addition, calcula- of the moderator coefficient indicates that it is well within the r utrements of Technical Specification 3.1.7.
0 Y s... O f l O e 4 w . 3.4-6
sh,- m b t 6 l ALL RODS OUT CRITICAL BORON CONCENTRATIONS t ppm boron , , Moderator Predicted l i Temperature Rese'ts 4 Measured Results ! l 4 o l l 300 F 1571 1555 :
- i
! 532 F 1634 1630 4 i P t i i IJ 1 i 4 i 2
.) <1 i I i
i' a l 2 , r a t
.,' l Table 3.4-1 ..
na f yt V y *- y w - r,. ,m- w -e e - wyrpwT w = c 6 --*4 - - - - 'q w
,-~.. ~ %
~(
Cc=parisen of Calculated and Measured Control Rod Group Reactivity Worths
'A. Moderator Temperature at 300F, APSR's at 100% wd { ROD DROP) acd Group Position Calculated Expected Measured
!!urber Deviativ
- Deviation Interval Worth, 5 Ak/k Measured Worth, % Ak/k Fron Calculated Fro:s Expected G Vd Worth, % Ak/k Vorth Meas. *** orth 1 0 + 100 0.89.-
2 0 + 100 3.01 - 6.19 h.58 4.70 -2h% 35 3 _0 + 100 0.74 e-3
& h 0 + 53.8 1.55 _s 5
Y c-L - B. Moderator Temperature at 532F, APSR's . at 35% vd (ROD DROP) Rod Group Position Calculated Expected Measured Deviation Deviation
'%:=b er Interval Worth, 5 Ak/k Measured Worth, % Ak/k From Calculated From Expected y vd Worth, % Ak/k Worth Meas. '4 orth 1 0 + 100 0 70- ,
2 0 + 100 2.13
-5.81 4.30 4.28 - -26% -0.5%
3 0 + 100 0.70 *
.h 0 + 100 1 73
- 5 0+ 57 0 55-L
CCMPARISON OF CALCULATED kid EASUPID COUTROL RCD GROUP REACTIVITY WORTH
, A. Moderater Te=perature at 300F, APSR's at 100% vd Predicted Measured Percent Group No. Rods Worth, %AY/K Worth, %AK/K Deviation o
1 8 -0.70 -0.70 NA 2 8 -2.13 -2.13 NA 3 8 -0. 70 -0.70 NA 4 8 -1.73 -1.73 NA 5 12 -0.69 -0.72 +4.3 6 8 -0.96 -0.86 -10.4 7 9 -1.04 -1.19 +14.4 Total 61 -7.95 .8.03
- 3. Moderator Te:perature at 532F, APSR's at 35% vd Predicted Measured Percent Group No. Rods Worth, % eX/K Worth, ZaK/K Deviation 1 8 -1.07 -1.07 NA 2 8 -2.90 -2.90 NA 3 8 -0.79 -0.79 NA 4 8 -1.71 -1.71 NA 5 12 -1.08 -1.14 +5.6 6 8 -1.23 -1.21 -1.6 7 9 -1.21 -1.15 -5.0 8 8 -0.38 -0.37 -2.6 Total 69 -10.37 -10.34 C. Moderator Temperature at 532F, APSR's at 100% vd ,
Predicted Estimated Percent Group No. Rods Worth, %dX/K Worth, %AK/K Deviation 1 8 -1.07 -1.07 NA 2 8 -2.90 -2.90 NA 3 8 -0.79 -0.79 NA
- 4 8 -1.71 -1.71 NA 5 12- .-l.15 -1.21 +5.6-
~
6 8 -1.30 -1.28 -1.6 7 9 -1.10 -1.05 -5.0 , Total 61 -10.02 -10.01 i / ( Table 3.h l
COMPARISON OF CALCULATED AND 25:ASUPID CCHTROL' ROD GROUP PIACTIVITY WORTH
. D. Moderator Temperature at 579F, APSR's at 35% vd 1
. Predicted Estimated Percent Group No. Rods Worth, %aK/K Worth, %AK/K Deviation 1- 8 -1.40 -1.40 NA 2 8 -3.59 -3.59 NA 3 8 -0.75 -0.75 NA 4 8 -1.47 -1.47 NA 5 12 -1.40 -1.48 +5.6 6 8 -1.48 -1.46 -1.6 7 9 -1.07 -1.02 -5.0 8 8 -0.44 -0.43 -2.6 Total 69 -11.60 .
-11.60 s
E. Moderator Temperature at 579F, APSR's at 100% vd Predicted Estimated Percent Group No. Rods Worth, %aK/K Worth, %AX/K Deviation 1 8 -1.82 -1.82 NA 2 8 -3.19 -3.19 NA 3 8 -0.86 -0.86 NA 4 8 -1.28 -1.28 NA 5 12 -1.42 -1.50 +5.6 6 8 -1.58 -1.55 -1.6 7 9 -0.99 -0.94 -5.0 Total 61 -11.14 -11.14 I i Note (l): NA indicates that deviations are not applicable to tod groups 1 h, since the predicted worths are used. Note (2): Estimated worth is determined by applying the percent deviations determined for rod groups 5-8 at 532F, with
- group 8 at 35% vd, to groups 5-8 at other conditions.
Note (3): Percent deviation ,is determined. assuming that the pre-dicted values are correct. s_- l Table 3.h-3 (continued)
' DIFFERENTIAL BORON REACTIVITY WORTH Ar 30(PF AND 532 F0MODERKIOR TEMPERATURES ,
Critical Conditions Heas ured Calculated Temp Boron Conc. Avg. Boron A Boron Ap Di f fe rential Diffe rential F Rod Position ppm Concentration Concentration % Ak/k Boron Worth Baron Worth ppm ppm I 26/k/100 ppm % Ak/k/100 ppm _y 300 All Out 1539 RG 8 Ob%wd 1362 532 All Out 1630 5 32 ~CRG 4 44% wd 1434 39 6 4.27 1.08 1.065 CRG 8 35% wd 1234 1 4 9
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!- Boron Concentration (ppmB) Figure 3.h-3 I
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l 1 l NUCIIAR INSTRUMENTATION OVERLAP l Detector Response, amp
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CONTROL ROD GROUP LOCATIONS A a (7) (4) (7) C (5) (3) (3) (5) D (4) (8) (6) (8) (4) E - (5) (6) (1) (1) (6) (5) r- (7) (8) (2) (2) (2) (8) (7)
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COMPOSITE DIFFERENTIAL SHAPES OF CONTROL ROD GROUPS 5-7 VERSUS WITHDRAWAL POSITION, FOR BEGINNING OF LIFE, ZERO POWER CONDITIONS
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REACTIVITY WORIll 0F CONTROL ROD GROUPS 5-7 5320F. , BOL, ZERO POWER MMMI5 =E!E b~= Zi5EEA =5I5 E=is -b Ai==i=EIE
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4.0 POWER ESCALATION TESTS Following the completion of Zero Power Physics testing, initial power escalation commenced on December 1, 1973 with the first electrical power produced at 1435 on December 5, 1973. The power escalation test program was conducted at four major test plateaus of 15, 40, 75, and 100 percent full power with minor
, testing performed at intermediate power levels as required by the controlling procedure for power escalation. Power level escalations occurred as required testing was satisfactorily completed. .
Power Level (Percent of Full Power - %FP) Date 15 December 1, 1973 40 December 13, 1973 75 December 25, 1973 100 June 19, 1974 On January 4,1974, during power escalation testing at 75 %FP, a malfunction in the switchyard caused a turbine / reactor trip. Following this trip, a foreign object was detected in the bottom of the reactor vessel. After , evaluations by Duke and the Babcock & Wilcox Company, and with Atomic Energy Commission concurrence, power operation was resumed. While at 15 %FP,
- a seal failure in Reactor Coolant Pump 2B2 occurred. Since repairs required
, the unit to be shut down for an extended period of time, it was decided to remove the fuel from the reactor and inspect the reactor vessel interior and the vessel internals. Due to this incident, further unit operation was delayed until May 23, 1974. The tests reported in this section cover power escalation testing as of 0500 hours, June 24, 1974. A summary of tests reported, along with the appropriate section number of this report and the power ' level at which the tests were performed, is given in Table 4.0-1. 4 i i 4 I~ l . i i ! s-l 4.0-1
a . , , . - . e s Summary of Tests Reported in Section 4.0 t Test Power Levels (% FP) Report Section Title of Section s5 15 30 40 65 75 90 100 ' 4.1 Nuclear Instrumentation Calibration at Power X X X X X a 4.2 Biological Shield Survey X X X X m e o 4.3 Reactivity Coefficients at Power X X X j - 4.4 Core Power Distribution X X X X 4 4.5 Rod Worth at Power X X X
! 4.6 Power 1mbalance Detector Correlation Test X X X i
4.7 Nuclear Steam Supply System lleat Balance X X X X X X X X l 4.8 Unit Load Steady-State Test X X X X X X X 4.9 Unit Load Transient Test X X 4.10 Pseudo Rod Ejection Test X 4.11 Dropped Control Rod Test X X
, 4.1 NUCLEAR INSTRUMENTATION CALIBRATION AT POWER 4.1.1 PURPOSE The purpose of Nuclear Instrumentation Calibration at Power was to calibrate
, the power range nuclear instrumentation indication to within 12 %FP of the reactor thermal power as determined by a heat balance and to within 15 percent.
incore axial offset as determined by the incore monitoring system. Addition-al purposes during the power escalation program were as follows: (a) To adjust the high power level trip setpoint when required by the power escalation procedure. (b) To verify that at least one decade overlap existed between the inter-
. mediate and power range nuclear instrumentation.
Two acceptance criteria are specified for nuclear instrumentation calibration at power as listed below. (1) The power range nuclear instrumentation indicates the power level within 12 %FP of the power level indicated by heat balance and within 15 percent incore axial offset as determined by the incore detectors. (2) The high power level trip bistable is set to trip at the desired value within +0.5 %FP. 4.1.2 TEST METHOD As required during power escalation, the top and bottom linear amplifier gains were adjusted in order that the power range nuclear instrumentation channels would indicate the power calculated by heat balance. The gains were adjusted, keeping their ratios the same if the imbalance were within toler- . ance of 15% axial offset as determined by the incore monitoring system. If the imbalance needed to be adjusted, the gains were adjusted to correct the imbalance and heat balance mismatch at the same time. During each adjustment, data were also taken to verify' overlap between the intermediate and power range channels. The required overlap was ajminimum of one decade bccween these two nuclear instrumentation ranges. ; When directed by the power escalation procedure and/or the unit startup pro-cedure, the high flux trip bistable setpoint was adjusted. The major settings during power escalation are given below:
. Test Plateau Bistable Setpoint
%FP %FP
'15 - - 35 40 50
, 75 85
, 100 ,105.5 L ~
4.1-1
y. l 4.1.3 EVALUATION OF TEST RESULTS An analysis of test results indicated that changes in Reactor Coolant System boron, changes in control rod configuration, and xenon buildup or burnout
. affected the power as observed by the nuclear instrumentation. This was as expected since the power range nuclear instrumentation measures reactor neutron leakage which is directly related to the above changes in system conditions. Changes in these system conditions resulted in a nuclear power range increase or decrease of approximately 3 to 5 %FP. Each time that it was necessary to calibrate the power range nuclear instrumentation, the acceptance criteria of calibration to within i 2.0 %FP of the heat balance power was met without any difficulty. Also, each time it was necessary to calibrate the power range nuclear instrumentation, the + 5% axial offset criteria as determined by the incore monitoring system was also met. Table 4.1-1 is a summary of the data taken during calibration at dif ferent power levels during power escalation testing. In all cases, the nuclear instru-mentation was adjusted to within i 2.0 %FP of the heat balance and to within i 5% incore axial offset.
The high flux level trip bistable was adjusted to 35, 50, 85, and 105.5 %FP prior to escalation of power to 15, 60, 75, and 100 %FP, respectively. Acceptance criteria of adjusting the setpoint to the above values within 1 0.5 %FP was met each time without difficulty. The maximum trip error observed was 0.1 %FP when setting the high flux trip at 50 %FP. The overlap measured during the startup program included the total span of the power range, exceeding the one-decade overlap requirement. Figure 4.1-1 shows the overlap of all three nuclear instrumentation channels. 4.
1.4 CONCLUSION
S The power range channels were calibrated to within two percent of total power , several times during the startup program. These calibrations were required due to power level, boron, and/or control rod conf 1guration changes during the program. The calibration procedure has therefore been thoroughly tested and has proven extremely satisfactory. Acceptance criteria for nuclear instrumentption calibration at power were met in all instances, i 4 I a e r 4.1-2
( s
' Summary of Nuclear Instrumentation Calibrations at Power Performed As Required By The Power Escalation Program and Section 4.1 of Technical Specifications Core AT Incore 'laximum Power Before and Af ter Calib. % FP Imb. Before and Af ter Calib. . % PP Power Imbalance Quad Tilt
(% FP) (% FP) (%) NI-5 NI-6 NI-7 NI-8 NI-5 NI-6 NI-7 NI-8 14.8 19.1 17.8 17.5 18.1 -3.7 -1.4 -2.2 -1.9 14.8 14.7 14.9 14.6 14.4 -2.9 -1.8 -2.3 -1.9 48.1 -8.9 -1.40 57.5 55.9 55.6 55.7 -10.1 -10.1 -10.1 -9.7 45.7 -9.2 46.6 46.3 46.6 46.7 -8.9 -9.0 -9.0 -8.9 Y o
.65.7 -16.9 -1.13 60.0 60.2 60.3 61.2 -19.9 -20.8-
-19.9 -20.7 s- 65.6 -16.1 -1.10 66.2 65.4 65.4 65.4 -15.6 -16.3 -15.8 -16.3 74.8 -19.4 -1.29 69.9 69.8 70.3 70.3 -17.3, -17.4 -17.5 -17.5 75.1 -19.7 -1.80 74.9 74.8 74.7 74.8 -18.7 -18.6 -18.6 -18.2 74.8 -7.8 -1.59 78.0 77.7 77.8 75.3 -7.3 -8.0 -7.4 +0.8 75.5 -5.2 -1.63 75.8 75.8 75.8 75.7 -3.9 -4.7 -4.1 -3.5 78.7 -1.5 -1.44 82.0 81.5 81.9 81.5 -2.0 -1.6 -1.7 -1.5 76.9 -4.0 -0.86 76.2 76.2 76.1 76.2 -2.7 -2.5 -2.6 -2.2 99.3 0.9 -0.89 95.5 95.6 95.5 95.6 -0.2 -0.1 0.7 1.2 100.8 1.0 . -0.88 99.9 99.7 100.0 99.7 -0.3 0.3 0.5 0.6 100.0 4.1 -0I.lii 95.6 95.3 95.5 95.7 -1.5 -1.1 -1.8 -2.0 99.9 3.58 -0.73 99.7 99.8 99.5 99.3 -n.8 -0.3 -0.9 -1.7
j . 4 i Detector Neutron Flux, nv i o o o o o o o o o i- o o _ o p w s m m N (D @ - : 2 e .l ~ - ' i I t I g g , g g g g g g 3 ' l ' 3 I I I I I i 3 3 i - - 1 o o o o o o o- o-- w w
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9 4.2 BIOLOGICAL SHIELD SURVEY 4.2.1 PURPOSE , The purpose of the Biological Shield Survey was to measure radiation levels in all accessible locations of the unit adjacent to the Blological Shield, and to obtain base-line radiation level data for comparison with future mea-surements of radiation levels during operation. 4.2.2 TEST METHOD Background surveys were performed on all Reactor Building levels at locations of interest prior to receiving fuel on site. Surveys were conducted during power escalation at the following power levels: 0, 15. 40, 75, and 100 %FP.
~
The Reactor Building, outside of the Biological Shield, or areas designated as access areas were marked off in discrete sections and readings taken. All areas in the Auxiliary Building and the Restricted Area yard that are adjacent to the outside of the Reactor Building were also surveyed. 4.2.3 EVALUATION OF TEST RESULTS The test results for each different phase are shown in the table below: Gamma / Neutron Gamma / Neutron Phase Date Average (mrem /hr) Maximum (mrem /hr) Background 08/15/73 0.06/0 .08/0 0% 11/12/73 0.1/0 0.1/0 IST 11/02/73 <2.0/<1.0 <2.0/<1.0 40% 12/13/73 0.41/<1.0 11.0/<1.0 75% 01/04/74 <1.0/<1.0 4.0/<1.0 100% 06/20/74 1.7/<1.0 2.4/2.5 The above results are for the inside of Reactor Building only. Thq maximum radiation level fcund was 11 mrem / hour gamma on contact with pipingf Normal levels were less than 0.5 mrem / hour gamma through the shield. 4.
2.4 CONCLUSION
S Since the maximum radiation level found during the power escalation, 11 mrem / hour gamma, is well within the acceptance criteria of 100 mrem / hour, the Biological Shield meets applicable design criteria. e
~- .
4.2.1
4.3 REACTIVITY COEFFICIENTS AT POWER . 4.3.1 PURPOSE , The purpose of this test was to measure reactivity coefficients during power operation at 40, 75, and 100 %FP. 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 temperature. (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. (d) Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in fuel temperature. 4.3.2 TEST METHOD Measurements of the temperature, moderator, doppler, and power doppler co-efficients were made at each of the major test plateaus during the power escala_ tion test program. The major test plateaus were at core power levelc of 40, 75, and 100 %FP. Rod worth measurements were executed prior to reactivity coefficient measurements in order to verify previously generated r6d worth data. For temperature coefficients, average reactor coolant temperature was increased or decreased about 5 F and data recorded. For power doppler coefficients, power was increased or decreased about 5 percent at 40 %FP and about 10 percent at 75 and 100 %FP, and data recorded. From the measured temperature and power doppler coefficients, the moderator and doppler coefficients were calculated. 4.3.3 EVALUATION OF TEST RESULTS . 1 The results of the measured temperature and moderator coefficients at power are plotted in Figures 4.3-1 and 4.3-2 which also show the calculated temperature and moderator coefficient results. A tabulation of the measured results and the calculated results for each measurement is also given in Table 4.3-1. Examination of the measured moderator coefficients plotted in Figure 4.3-2 indicates that the limit of a non-positive value will not be exceeded at 95 %FP_ unless the soluble poison concentration is in the range of 1290 ppm. , Measured results of the soluble poison concentration versus core lifetime for equilibrium xenon, all rods out, beginning-of-life conditions indicate a maximum beton concentration of 1215 ppm under these conditions. These
.i
results coni.rm that during power operation at or above 95 %FP the moderator coe'fficient will be negative. 4.3-1 a- -.a--.-.- ..-:- . - - - - - - - - - - - - - - - - .
The results of the measured and calculated power doppler coefficient of reactivity are plotted in Figure 4.3-3. A tabulation of the measured and calculated doppler and power doppler coefficient is also presented in Table 4.3-1. Examination of the measured power doppler coefficient indicates a slightly lower measured value than predicted which results in a reactivity deficit versus power less than originally calculated. The total measured reactivity deficit between 0 and 100 percent full power is estimated at -0.92 %ak/k as compared to a predicted value of -1.32 %ak/k. This difference is a net gain of + 0.40 ok/k/%FP in core excess reactivity available for Cycle 1 lifetime. coefficient is that Theacceptancecriterionforthemeasuredpowerdopple{ak/k/%FP. the coefficient must be more negative than -0.55 x 10- Figure 4.3-3 shows that all measured coefficients are well below this value and that the acceptance criterion is adequately met. In addition, all calculated doppler coefficiencs were well below the acceptance criterion. 4.
3.4 CONCLUSION
S The measured results indicate that the moderator coefficient will be negative during operation at or above 95 %FP. Analyzed data for the power dcppler coefficientvegsuspowerlevelindicatethat the least negative coefficient is -0.78 x 10- Ak/k/%FP. The total power doppler deficit from 0 to 100 %FP from this measured data is estimated to be -0.92 %ak/k. f. 6 4 e f v. 4.3-2
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k r A e p d o b d o D
- e M 3 0 r 7 7 h 7 5 h 4 1 u 0 0 2 2 1 1 1 1 s -
a - - - - - - - - y e t M i y i v t t d i c e v a t 0 6 i e a 2 3 2 h 8 3 3 8 t R l 0 0 2 2 r 2 1 c u e e . . a f c r 0 0 0 0 l 1 1 1 1 e o l u - - - - p - - - - R a t p t C a o f n r D o e e i p r s c d m e t i e e w n f r T o e f u 2 1 8 o P 5 0 8 0 i e s 2 2 3
- h. 9 9 8 8 c o a i C e - - - - - - - -
f M f e o ) d C E o V I t d A e ( % t / 0 7 7 0 6 7 0 0 a h k h 6 9 1 4 6 9 1 l d/ 1 0 0 1 1 0 0 1 u /k 0 0 0 0 0 0 0 0 c #A l d% a C d n ) t t a n5 n n d ocm rnn e 9 3 6 1 8 0 5 9 e 3 9 6 1 8 0 5 9 i i e r oon B Gl c 1 1 2 1 0 1 0 1 c 1 1 2 1 0 1 0 1 i i u f f s f f a e e e o o M C C 5 0 5 7 5 0 5 7 f d 8 r 3 0 3 2 r 3 0 3 2 _ o V o 1 e 1 t l _ y % a p _ r r p a , e 0 5 0 5 o 0 5 0 5 m n 7 d 0 7 0 1 D 0 7 0 1 . m o o u i M d S t n i d A s 6 n 5 0 5 5 5 0 5 5 o A 7 0 7 9 r 7 0 7 9 P 1 e 1 e l d r p o 5 u p
, R - t 0 0 0 0 o 0 0 0 0 1 a 0 0 0 0 D 0 0 0 0 r 1 1 1 1 1 1 1 1
_ e r l p e
= e m w v e o e T P e eL) g P
~~
arf . 0 5 5 0 . 7 0 1 5 r re A 4 7 7 0 B 3 7 7 9 - (, e w% 1 - - v o( - AP - a, ". Yr -
l , k Temperature Coefficient of Reactivity Versus Boron Concentration for Critical Boron and' Rod Conditions at 579 F, 2155 psig, and 0 EFPD. fl l !! h[ ' (( [! {! !
!lf' I i l ! O I!llifI!IhIf}Illh IIl}} r!!!((j !
! lll! lIl 0* ! !
, fi: l[!j !! !!O IN 00 ![!I
'F.l'ITI 'f F R. i! n!! f( :; :l'D.i!!-. . m qI. l q!:p; :- yl'l!!! h W 'n!l I im.it t m~ 'fpJ q' r. . !!
lP
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gt' 7 [f, -> t 1 i-pI pll1. Q !:' l } ! !! k l 1 1 i t : If I : i ! ! ! . t h I 't i ! l! .1: ;, ! t I . l l i ' il l' 'i lj il '*li ": I j! j li
- il I l l : i : l t :
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l l 'l l
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l l j l !!I! ill l.!i !!!! gii ig
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ilii ;! j N, -. L l' .f,! . .* ! ! !!
- 1
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i ;
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- i. ' i ! i i
.: u . . .i l .
l l w g!!!
.fl itiI ! . M .!I!
p l I {1 ! I I .I (1 i
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~'i i ll! ! i .
i [}j ll l H!j j l l j j i ji ((l [~ I} I [ ijji j{l,! Calculated Temperature ai ; i ; - m p; p" ft;*fF M . * * "!l[ -t~I # tri- r Q T
'li I> : it"il !'!
d l t. l I
; l; '
i tl(( i A- % " : -
- ~D I ' '
15 :41.INI ! m r Jn .. ..i. + 40 %FP ~ If.[.I! ..il!..!.! !. {} ki 3 7..
!p
!J;,p;.M h-.DM p, ;;j!- i.
di1 - MM. . . - t! i. .l.d.!iO 'n; Oh- -!!:
!!i
++ +
!ie':i M .p.
1 o 4 ..g m ! 100 %FP x :, jih [} { jili .- t I j! c - . - 3 i ,11! ilg n
- s. 5 5L". .i .'..ljily qi {:li!!! :1! .t+U r -
llr! ! !!'] IM s'N ~, i
, .{
{l l
-i ! ll ! 7 W!! I!M !
ll0 l i i [j i !:
*['t .!' 'I [j i l!!l M i h-
*!j l+i pd "TiiIIM}
.et # r fp it?'IIi' 7' 4; -*
d :'ii i 1 !~ } l:Jj i '!*i
.c-r+4 :
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.. M t W,:21i iat it it'l
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--*+
j g-i
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i I ll 'l i!11d !'
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rt ~ *, t nr i i!! u 8I . l l k-."iS+ 1.M.jj!.Mi" 31 j.Lg 5 - a t i. ! I I , l l
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.<n e'
<3di$MfifiM.fF,I ar- iisgi
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}
dlhf d }_ _I. -..; I"I : .M. !!! l9 ira .rM.. e4 !L~b .: .
. *:l"!!! l 4 .j!!~
8 ' j L 9eMI!!kfif ij . M[ V i di lii iii i' i ! ! , II! il I i j]!1.' j lf .
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- W_rF j .i- .i f.J.i . p
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iI ii ' i t i ii*- - 1 i -
"I e I M - l fl i 1
.l l I
. I~ i l *I; ; il il -} [*-
L-* "H .' j j ; .j; j Measured Temperature -
, -- !j +1 Coefficient Results -
7 I' i i I i i e __ 40 %FP _ l-. ' [ [ f O- 75 %FP
! [:. l 6 -_ 100 %FP i -
I .;l.
..y m Ql li i I
"$h; Du y0
%: l
'ld i li lW IKO I i :0 I I lu I I Boror. Concentration, ppm
l ,,. ( i
! Moderator Coefficient of Reactivity Versus Boron Concentration for Critical Boron and Rod Conditions at 579 *F, 2155 psig, and 0 EFPD.
~
ji.
- l. l t 4 9.1 }{ .y i!'. .Tl l! !' ..!i !!!! j g g !g!.y l [.
l l i .l L{I! I'-lj
!!..i!!!ll!. i !
.![.!!OII!h !N 0!!!f;-
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; q t ;
l . i M{..,q ii q ..j. ;;; .yt n a
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j j , [~- j
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- l i il!
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+ '
l' y
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!L
';; . Calculated Modkrdtor.
p d h f} 0 [ ' h [ jdidh$hN " jihy!! n
% U i- t- Coefficient Results l l !!!! g ! ll
}l {1F p' i [*
i ! j t i g-[(( '
;; h !!%l
,;;0 $-$ !!0(!!
M g V g 1 15 $FP i
..jL j p--- h e-
}.t g l *3/ :. $ f}pi
" T jl o 40 %FP g [
li 100 %FP \ c.s s i }
}
1 4 r- " 1 Y ~"t1-
<* 4 l
"i l }<fl.
.-- 1'{
.*g st lill i~i t-W' <W'E pi .
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- El M ' i t
i
- Measured Moderator
.a i i J:1 .
Coefficient Results : I O -- 40 %FP - e -- 75 %FP A -- 100 %FP j 1
- j -
1
. r } .
i i . c
'l K i
- > K lL i
.lMI !UC 1 '""'
- MIK -M N IfIEl .
{ { .. ll' !.. l. } I Boron Concentration, ppm
. Power Doppler Coefficient of Reactivity Versua Pcwer Level _~.. . .b..(..., .{ -*-. h. .1I_
=,._ g _... . . , , . . _ . . _ . _ . . . _ . . . __..
.___ . __.7.
^~E ~~Z z Acceptance Cr1teria of .55 x 10- AK/K/%FP =-
==tE==1- _, w. -
=E!== = * := =1:-- ---d=-a._, . ._;
1 - _ 2.- , z=_ a_ ,i_ . .
+=
e- _h=.M-
. = . _ - - _._ ..._
_y._. _ _ _ . .__. t.
@* _E*= n _.L._. ; --
m = _ . . _ = . .. m._
. _ . , . _ .. . , _ . , . __.- , .m-s ---.g_-- - , _ --.+-
_ . .-t_-.
- e _
==t= =t=- - - + - ----t-. =P- 7.:s .f _ _j . ._ =-t-~,M. ,,_L._. . .___
ggjgg EE$5==_r---==t=- =iEETu:g}.j. _;;[_;-y52.-45"=- ~~ _a _ . 4 _ , _. - - - _ _ - _ .
-y .v n=_ =.. =E; ;_=!== =t. m=E
- p --
-==L_
._1=.
a 2-_ . . . = m ._ . , _ _ __1_ ._. . . . _ . H E :- -+_ -...u_.. ._2 . . .i; ____ , .
=t= .._; . _. p.. -
=1= =
M
=_.z._ -._ _ __- . _ . ,
_. s . .._. _&_ __ _.. . . . _ . _ _ a.. _. _.
<-4 _EEigg IEEZ ZiEgl=j~== + - _t.==Z-~ =-ZE- ~~'-~
= E~=
._=.y---- _ -
= t== a - - - - _
~z=--- . .- ;= -,- - - . .
--.z_ ,--
o , . . .
,==1_ _m._. _ ._ z_ .=1== _ . . . _.._. . . . . _ a_ _.a_.. , . _.
g 7_._
,_... ..__,_f
, _ __ _ - . ..__ __ L.
y g_
% __ _ :_.. =+- -+-
.t ._ t._p
^
.5 ~ii-ri----
- -i= --1._ _- *-:-l'~ \
u e _ _ . . _ .___ _ . ..!_. . . . _ . _._+__
. . . . . . __ __ Calculated Results".. -
__,__ z__ _ . . . _+_ g - . . =t- _ . . _ _ . __1_. ._1. _ _.r-.--- 1._ O --~- ~~ ""~ a 3= ...g=_
., .E;~ - . *. . . . -": Z
-~
ZMeasured Power Doppler-
~~~~
o - - + = . _. . _ _. . . _ .. t- - o, _L_. - - E == ~~
=
Coefficient Results 0 -- h0 %FP
' ==:t=
~ _
i -- 75 %FP m f* __ __. A - 100 %FP - m_ s_., ._. . . .
~
- s. _ _ .- __,_. _.
~.__.. __e. =-1 = _,
e~ i t-- --
,: ._u = =_...
Pcwer Level, % Full Power 4 4 I. Figure 4.3-3
4.4 CORE POWER DISTRIBUTION
. 4.4.1 PURPOSE
, Steady-state core power distribution measurements were intended co verify
~
that the core axial power imbalance, quadrant power tilt, minimum DNBR and maximum linear heat rate do not exceed their specified limits. Acceptance criterla specified for the Core Power Distribution Test included: (a) The combination of reactor power and reactor power imbalance does not exceed the safety limit as defined by the locus of points established in Figure 2.3-2B of the Technical Specifications (Figure 4.4-1) . (b) DNBR is greater than 1.55. (c) The quadrant power tilt, as defined below, does not exceed four percent, except during physics testing. Quadrant Power Tilt = Power in any Core Quadrant
-1 x 100 Average Power of All Quadrants (d) The maximum linear heat rate is less than 17.3 kw/f t, based on Figure 3.5.2-4 of the Technical Specifications (Figure 4.4-2) .
4.4.2 TEST METHOD When required by the test program, core power distributions were determined after establishing steady-state conditions at the required power level and rod _ configurations as determined by the controlling procedure for pewer escalation. In order.to compare measured results to predicted results, some cases required two-dimensional or three' dimensional equilibrium xenon. When three-dimensional equilibrium xenon was required, the part-length control rods were maintained at a constant position and the axial incore imbalance was maintained to within plus or minus three percent of an equilibrium im-balance value for an eight-hour period prior to taking data. In other cases, data were taken as per the conditions required fori the individual tests. { 4.4.3 EVALUATION OF THE TEST RESULTS 4.4.3.1 Ceneral During the startup program, c' ore power distributions were determined as required by the different test procedure a the power escalation sequence
~
and covering various control rod patterns, core power levels, core axial imbalances, core quadrant tilts,'and xenon conditions. Table 4.4-1 summarizes
. core power distributions taken during the performance of the. Core Power Distribution Test.
ps e 4.4-1
4.4.3.3.1 Minimum DNBR Deteruination Minimum DNBR values were calculated for core power distributions taken as part of the power escalation test program. The results of various DNBR , values, including the worst-case values, calculated at each test plateau under normal rod configurations are plotted in Figure 4.4-14. These results indicate that all measured values and calculated values result in DNBR's greater than 1.55. Six steady-state, equilibrium xenon core power distributions, as required by the Core Power Distribution Test, were obtained during the power escalation sequence. Each distribution was subjected to the following analysis for determining DNBR. (a) From each core power distribution, the fuel assembly which yielded the worst-case DNBR was selected. (b) Upon selection of the worst-case assembly, a radial peaking factor was calculated and segment power levels were converted into axial peaking factors. (c) The radial peaking factors were adjusted by a factor of 1.05 (the local pin peaking multiplier). (d) The radial and axial flux data from (b) and (c) were incorporated into the standard hot channel analysis and analyzed for operation at full power. The results of these calculations are presented in Table 4.4-8. All cases studied resulted in substantial DNBR margins. A minimum DNBR margin of 71.0 percent was observed after extrapolation to 102 %FP. 4.4.3.3.2 Maximum Linear Heat Rate Determination Maximum linear heat rate values were calculated for core power distributions taken as part of the power escalation test program. The results of these worst-case values at each test plateau under normal rod configurat%ons were then extrapolated to 102 %FP. The results indicated that none werp greater than 17.3 kw/f t as required in the power escalation sequence procedure. Analysis for determining maximum linear heat rate was performed in conjunction with minimum DNBR analysis. After selection of the worst-case assemblies and the determination of the radial and axial peaking factors, the maximum linear , heat rate for each case was determined for operation ac full power. The
, formula utilized to calculate the maximum linear heat rate is as follows:
MLHR = P R* A* L* # ** NA x NP x AL f 4.4-3
4.4.3.2 Steadv-State, Equilibrium Xenon Distributions Steady-state, equilibrium xenon core power distributions were measured and calculated for various control rod patterns at each of the major power es-
. calation test plateaus. Measured results of six core power distributions covering various control rod patterns and core power levels are tabulated in Table 4.4-2 through 4.4-7. These tables give a complete 1/8 core power distribution map using the corrected signal outputs from 203 incore detectors located in 29 different fuel assemblies which describe the entire core assuming eighth core symmetry. A summary of each measured core power distribution presented is given in Table 4.4-1, which tabulates the core power level, control rod positions, core burnup, boron concentration, axial imbalance, maximum quadrant tilt, maximum LHR, minimum DNBR and power peaking data for each measurement. The measurements covered the following control rod patterns and core power levels:
Power Level Control Rod Position (% ud) Equilibrium Number (f FP) 1-5 6_ 7 8, Xenon 1 15 100 75 00 35 No/0-D 2 40 100 75 00 100 Yes/2-D 3 40 100 75 00 35 Yes/3-D 4 75 100 100 75 100 Yes/2-D 5 75 100 75 00 35 Yes/3-D 6 100 100 90 15 14 Yes/3-D The results of these measured core power distributions indicate a mcximum local peaking factor between 1.81 and 2.00 for all cases examined with the local peaking being maximum at 15 %FP with a value of 2.00 and the minimum at 100 %FP with a value of 1.8. Both values are well below the 2.67 maximum local peaking factor limit given in Technical Specification 2.1. Using the maximum local peaking factor for the six cases reported above, calculations were performed to determine the maximum linear heat rate. As can be seen in Table 4.4-8, all extrapolations to the LOCA limit (102 %FP) were below the Technical Specification limit of 17.3 kw/ft and all extrapolations to the central fuel melt limit (112 %FP) were below the Technical Specification limit'of 19.8 kw/ft. Similarly the minimum DNBR extrapolations to the LOCA limit and the central fuel melt limit were well above the TechnicaljSpecifi-cation minimum value of 1.30. 'i Core power distribution calculations at steady-state conditions were deter-mined using the three-dimensional PD0-7 code with thermal feedback. The six cases reported in this section have been compared (measured versus calculated) in Figures 4.4-3 through 4.4-13 to demonstrate the degree of
, agreement between the calculated and measured power distributions. It can be seen from these figures that the comparison between measured'and calculated j radial and radial times axial power distributions shows favorable agreement.
i
. 4.4.3.3 Minimum DNBR and Maximum LHR Calculations e
I 4.4-2
- i. _ _ -__ - - m
Where: MLRH = Maximum linear heat rate (kW/ft) PR = Radial peaking factor
- PA = Axial peaking factor PL = Local pin peaking multiplier (1.05) >
- Q Rate = Rated core thermal power (2568 x 103 kW) FNT = Fraction of power generated in the fuel (0.973) NA = Number of fuel assemblies in the core (177) NP = Number of fuel pins in each fuel assembly (208) AL = Active length of each fuel pin (12 f t.) The results of these calculations are presented in Table 4.4-8. All cases studied resulted in substantial linear heat rate margin. A minimum linear heat rate margin of 33.1 percent for central fuel melt was observed af ter extrapolation to 112 percent full power. 4.4.3.4 Quadrant Power Tilt and Axial Power Imbalance
- 4. 4. 3. 4.1 Quadrant Power Tilt Quadrant power tilt limits have been established in the Technical Specifi- .
cations. 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. Table 4.4-1 shows the maximum quadrant power tilts for core power distributions taken as required by the Core Power Distribution Test at various power levels and incore axial offsets. 4.4.3.4.2 Axial Power Imbalance Results from the core power distributions taken at 40 and 75 %FP during the performance of the Power labalance Detector Correlation Test, Section 4.6, show that he imbalance trip envelope (Figure 4.4-1) of the Reactor Protective Sys tem is suf ficient to protect the unit from exceeding the DNBR and LHR limits u. der all core imbalance conditions when a gain factor of 3.3 is set into the delta flux amplifier. Thermal analyses indicate that the optimum operating conditions for the unit are obtained when a negative six percent incore axial. offset is present. j i The calculation of axial core imbalance is performed using signals from the incore detectors and the exc re power range detectors. The excore axial imbalance is . calibrated to the incore -axial imbalance whenever a five ' percent or greater discrepancy occurs. 4 4-S- v
' 4 .~4-4
4.
4.4 CONCLUSION
S \
. Comparison of the five equilibrium xenon core power distributions taken at 40, 75 and 100 %FP at various control rod positions, with PDQ-7 calculations , showed that the measured values for the maximum radial peaking factor and ;
for the maximum radial times axial peaking factor were within 5 percent and 7.5 percent, re=pectively of the calculated values. The results of the minimum DNBR and maximum linear heat rate analyses are presented in Table 4.4-8. Minimum DNBR margins were calculated for each core power distribution with respect to the limiting DNBR of 1.55. The , margins to the limiting linear heat rate were calculated for each case for both the central fuel melt limit and the LOCA limit. On the basis of this study, the following determinations were made: (a) All cases analyzed resulted in a substantial DNBR margin. A minimum DNBR margin of 71.0 percent was determined when calculated at the LOCA limit (102 %FP). (b) Maximum linear heat. rate analyses indicated substantial margins to the limiting linear heat rates. Minimum margins of 33.1 percent for central fuel melt limit and 30.3 percent for the LOCA limit were determined for the cases analyzec. The results of the quadrant tilt and the axial power imbalance calculations for a variety of different core power distributions taken during the power escalation test program yield the following conclusions: (a) All tilts determined during normal power operation were well within the Technical Specification limit of four percent. (b) The Reactor Protective System will provide sufficient protection against exceeding DNBR and LHR limits when the delta flux amplifier has a gain factor of 3.3. f s s e l 4.4-5 i M ,
. x.~c O O e e cJ - m c. - 1
'U
- t-a nu O. O. . cc. CS. Ch. C.t G. e.
g x< ei a - - - - - - -
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= m M m = M a
u N . . . . . =. . =. a.
~ ~ ~ ~ ~ ~ ~ ~ ~
n m es a a eJ eJ a g Lt% 4 4 ed
. O. C.N N. . \C.
g 5 e a o - e e Ch O e
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s.t t--2= 0
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% M M CJ O-va-x - e c"- = c O e w = o O O a
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ci c1 H H ed rH
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M O M O M M C M H T 74 "0 % > m - - U 3 L .* O 4 %'t M =0 A CO 3 O*
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cw O
- AO Ln CJ O N Q O AC C C 5 t c$ 8 e & s - O & & O r o , ,
ue ~ ~ A o t 3> . . . , ta a A - m '8 9b% 0 o d .O M e I O O O O O O O O O O O O O C O O O O c - ~ ~ ~ ~ - - m m - 3 L aO r.o O C-O 7+ g g , , , g g
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= - 3tr J - d.
m
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, COM 2 M :/\ t- in O CJ V - 'n A t.% C M N 04 - M O s
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. Lts e-4 M
CJ V j p3 3 3 I i e t i I t O # eJ M m O O o a e O O e O e_ O Table k.k-1
l MEASURED CORE POWER DISTRIBUTION RESULTS AT 15 % FULL POWER Control Rod Group Positions Cps 1-5 100 % wd GP 7 02 % wd , Gp 6 77 % wd GP 8 3h % wd Core Power Level 15.h % FP Boron Concentration 12o?._ PPM Core Burnup Q EFPD Axial Imbalance -2.4 % FP Xenon Conditions Equilibrium Conc. No Yes or No Reactivity Worth NA % Ak/k Max Quadrant Tilt +1.gh % 1/8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te cto r Factor Assembly i nen a nn Numb e r Peon Irca H-0 8 1 1 1.h? 1.10
< G-08 2 4 1.77 1.30 F-08 4 4 1.70 1.25 E-08 10 4 1.91 1.38 D-0 8 14 4 1.65 1.20 C-08 21 4 1.86 1.37 B-0 8 30 4 1.77 1.35 A-0 8 37 4 1.20 0.86 G-09 3 4 1.80 1.34 F-10 12 4 1.72 1.33 E-11 26 4 1.52 1.11 D-12 41 4 1.23 0.90 C-13 52 4 0.73 0.54 F-09 6 8 1.92 1.44 E-09 5 8 1.77 1.31 D-09 15 8 1.73 1.26 C-09 29 8 1.h6 1.08 B-09 31 8 1.33 1.00 A-09 45 8 1.09 0.80 '
E-10 17 8 1.82 1.30 D-10 27 8- 1.18 0.96 C-10 28 8 1.26 0.93 B-10 44 8 0.94 0.72 A-10 46 8 0.60 0.45 D-11 33 8 1.37 1.02 C-ll 42 8 1.14 0.82 . B-ll 49 8- 0.78 0.62 C-12 48 8 1.07 0.78 B-lZ 31 0 0.69 0.53 f v Table 4.4-2
- _ - . . . ~ . _ . _ _ . . . . . _
e MEASURED CORE POWER DISTRIBUTION RESULTS AT 40 % FULL POWER e Control Rod Group Positions a Gps 1-5 100 ud cp 7 00 % wd
, Gp 6 72 % vd GP 8 W~% wd Core Power Level h0.0 % FP Boron Concentration 1240 PPM Core Burnup 2.6 EFPD Axial Ir. balance -9. 3 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth-192 % ak/k Max Quadrant Tilt -1 10 %
1/ 8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te c to r Factor l?corg Assembly tncar4cn Nedbe r ocs_ H-0 8 1 1 1.35 0 95 G-08 2 4 1.07 1.17 F-08 4 4 1.77 1.22 E-0 8 10 4 2.00 1.34 D-08 14 4 1.69 1.09 C-08 21 4 1.83 1.25 B-0 8 30 4 1.dk 1.30 A-08 37 4 1.26 0.86 G-09 3 4 1.76 1.21 F-10 12 4 1.73 1.22 E-11 26 4 1.72 1.11 D-12 41 4 1.43 0 96 C-13 52 4 0.8h 0.57 F-09 6 0 1.88 1 30 E-09 5 8 1.8h 1.25 D-09 15 8 1 90 1.26 C-09 29 8 1.56 1.07 B-09 31 8 1.h0 0.98 A-09 45 8 1.11 0.76 E-10 17 8 1.95 1.31 D-10 27 8 1.6h 1.11 C-10 28 8 1.48 1.02 B-10 44 8 1.02 0.72 A-10 46 8 0.65 0.46 D-11 33 8 1.6L 1.10 C-11 42 8 1.28 0.87
, B-11 49 8 1.03 0.70 , C-12 48 8 1.2h 0.85 B-12 31 6 0.80 0.58 s
Table 4.4-3
e MEASURED CORE ?OWER DISTRIBUTION RESULTS AT LO % FULL POWER e Control Rod Group Positions Cps 1-5 100 % vd GP 7 Ok % wd
. Gp 6 76 % wd GP 8 35 % wd Core Power Level bl,$ % FP Boron Concentration 1208 PPM Core Burnup 12.6 EFPD Axial Imbalance o_.6_% FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.01 % ak/k Max Quadrant Tilt -1.25 %
1/ 8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te cto r Factor Assembly tncn H nn Number PeoE Lo c H-08 1 1 1.hh 1.08 G-08 2 4 1.77 1.27 F-08 4 4 1.74 1.26 E-0 8 10 4 1.97 1.37 D-0 8 14 4 1 72 1.15 C-08 21 4 1.85 1.33 B-0 8 30 4 1.79 1.32 A-08 37 4 1.16 0.83 G-09 3 4 1.83 1.33 F-10 12 4 1.77 1.31 E-11 26 4 1.69 1.08 D-12 41 4 1 36 0 93 C-13 52 4 0.77 0.55 F-09 6 8 1 92 1.41 E-09 5 8 1.82 1.31 D-09 15 8 1.84 1.26 C-09 29 8 1.55 1.07 B -09 31 8 1.41 0.99 . A-09 45 8 1.07 0.78 I E-10 17 8 1.97 1.30 D-10 27 8 1.61 0 98 l C-10 28 8 1 37 0 95 l B-10 44 8 0 9o 0.73 A-10 46 8 0.64 0.h6 D-11 33 8 1.58 1.05 C-ll 42 8 1.23 0.83
. B-ll 49 8 0.93 0.66 . C-12 48 8 1.15 0.81 j B-12 31 6 0.74 0.55 l I
I I Table 4.4-4 l
MEASURED CORE POWER DISTRIBUTION PISULTS AT 75 % FULL POWER e Control Rod Group Positions
- 76 g yd Cps 1-5 100 % wd GP 7 4
. Gp 6 100 % wd GP 8 100 % ud Core Power Level 75 0 % FP Boron Concentration 1277 PPM Core Burnup %15 EFPD Axial It:6alance Tf.9 % FP Xenon Conditions.
Equilibrium Conc. Yes Yes or No Reactivity Worth -2.56 % ak/k Max Quadrant Tilt +1. 37 % 1/ 8 Co re Incore Weigh ting Pmax/ p/i> Fuel Fuel Assy. De te c to r Factor gory Assembly i n r 2 H nn Nurbe r ca H-0 8 1 1 1.03 1.33 G-08 2 4 1.05 1.19 F-08 4. 4 1.39 1.06 E-0 8 10 4 1.00 1.19 D-08 14 4 1.41 1.08 C-08 21 4 1 70 1.30 B-08 30 4 1.62 1.34 A-08 37 4 1.18 0.86 G-09 3 4 1 59 1.16 F-10 12 4 1.41 1.10 E-11 26 4 1.37 1.05 D-12 41 4 1.34 0.92 C-13 52 4 0.70 0.52 F-09 6 0 1.50 1.20 E-09 5 8 1.52 1.14 D-09 15 8 1.57 1.18 C-09 29 8 1.44 1.09 B-09 31 8 1.48 1.11 A-09 45 8 1.20 0.88 E-10 17 8 1.58 1.20 D-10 27 8 1 39 1.02 C-10 28 8 1.46 1.09
~
B-10 44 8 1.54 1.03 A-10 46 8 0.88 0.63
^
D-11 33 8 1.43 1.05 C-ll 42 8 .1.10 0.c6
.- B-11 49 8 1.02 0.70
, C-12 48 8 1.09 0.80 B-12 al 6 0.76 0.56 i
( Table 4.4-5
a
, MEASURED CORE POWER DISTRIBl' TION RESULTS AT 75 % FULL POWER Control Rod Group Positions
~
Cps 1-5 100 % wd GP 7 00 % wd
- 35 % wd Gp 6 75 % vd GP 8 Core Power Level 75.? % FP 4
Boron Concentration 1081 PPM - Core Burnup 26.0 EFPD Axial Imbalance -17. 0 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth-2.59 % Ak/k Max Quadrant Tilt -1. 5 % 1/8 Core Incore Weigh ting Pmax/ P/P Fuel l Fuel Assy. De te ctor Factor Peore Assembly i nnnH nn Numbe r B eal H-0 8 1 1.47 1.11 G-08 2 4 1.73 1.30 F-08 4 4 1.74 1.26 E-0 8 10 4 1.98 1.39 D-08 14 4 1 7h 1.19 C-08 21 4 1.66 1.36 B-0 8 30 4 1.77 1.29 A-08 37 4 1.12 0.80 G-09 3 4 1.79 1.33 F-10 12 4 1.79 1.3h E-ll 26 4 1.69 1.11 D-12 41 4 1.33 0.93 C-13 52 4 0.75 0.55 F-09 6 0 1.94 1.hh E-09 5 8 1.01 1.32 D-09~ 15 g 1.84 1.26 Cs09 29 8 1.52 1.07 B-09 31 8 1.37 0 98 ,
' A-09
' ' ~
45 8 1.03 0.76 E-10 17 8 1.99 1.32
~~ D 27 8 1.61 0.99 C-10 28 8 1 37 0 95 B-10 44 8 0.97 0.72
^
A-10 46 8 0.62 0.h5 _ D-ll 33 8 1.65 1.10 C-11 42 8 1.2h 0.85 B-ll 49 8 0.87 0.56
. C-12 48 8 1.15 0.c0
(\, .s B-12 51 6 0.72 0.52 Table 4.4-6
, MEASURED CORE POWER DISTRIBUTION RESULTS AT 100 % FULL POWER e
Control Rod Group Positions ' ~ Gps 1-5 100 % wd GP 7 15 % vd
= Gp 6 90 % vd GP 8 13.5% wd Core Power Level 97.6 % FP Boron Concentration 1077 PPM Core Burnup 31.5EFPD Axial Imbalance +2. 5 3 % FP Xenon Conditions Equilibrium Conc. Yes Yes or No Reactivity Worth -2.77 % ak/k Max Quadrant Tilt +1. 74 1/ 8 Core Incore Weigh ting Pmax/ P/P Fuel Fuel Assy. De te c to r Factor Asserbly inn,cyn, NuMbe r Peon Mca .
H-0 8 1 1 1.37 1.12 G-0 8 2 4 1.62 1.29 F-08 4 4 1.53 1.23 . E-0 8 10 4 1.81 1.38 , D-0 8 14 4 1.59 1.26 . C-0 8 21 4 1.69 1 37 B-0 8 30 4 1.03 1.29 A-08 37 4 0.98 0.78 G-09 3 4 1.60 1.29 F-10 12 4 1.65 1 31 E-11 26 4 1.54 1.18 D-12 41 4 1.18 0.93 C-13 52 4 0.69 0.54 j F-09 6 0 1.75 1.h1 E-09 5 8 1.64 1.29 D-09 15 8 1.73 1.22 C-09 29 8 3.20 1.07 B-09 31 8 1.27 n.o8 A-09 45 8 0.00 0.75 E-10 17 8 1.76 1.33 D-10 27 8 1.43 0 97 C-10 28 8 1.28 0.97 B-10 44 8 0.68 0.75 A-10 46 8 0.56 0.46 D-11 33 8 1.46 1.12-C-11 42 8 1.08 0.84
- B-11 49 8 0.72 0.58
. C-12 48 8 1.02 0.80
/ B-12 31 8 o,71 o,51
( Table 4.4-7
m u m _ _ 851 4 78 36 9 697 h01 765 iH 063 728 995 140 nB T. 2 8 6. h. 0 iN 422 932 822 533 h32 333 MD 2 d e ) t m t au f 667 572 742 116 064 185 l mR/ 767 768 891 6 78 902 590 ll w I S ui.k( cxI l a 112 h12 h13 802 823 002 1 1 11 11 11 11 11 1 S aM Y C LS AN yn NO l o AI T bmti. 7 5 5 8 0 5 EI ea 0 0 0 0 1 0 TD sc - - - - - - AN so F I 1 1 t 1 RO AL 1 1 1 f 1 C T AN ) E0 nt Hh on E ia RX l t m A akag EM i aco 3 3 2 h 2 4 NU xeon I I ApL( LR B e MI r UL MI m oe u I tJ P XQ m _/ ik 2 9 0 0 7 9 3 9 9 9 8 1 AE M xa ax 2
, ae m MP 1 1 1 1 1 DE P NT AA e T c RS e nI B rap 0 0 0 0 3 NY ol F 4 3 6 9 0 5 DD ca .
A nbf 2 9 9 6 7 2 ME I m( - - - - 1 + C7P f 1 I - I - NH - I ) I - MM P - rlF 400 500 600 000 300 600 ee wv4 522 1 22 122 522 522 722 oe( 101 401 401 701 701 901 PL 1 1 1 1 1 1 11 11 11 4 3 4 4 4 4 7 ( 7 7 7 7
- 0 - 5 2 6 - 1 - 5 e e 42 60 71 1 3 50 31 t m 26 1 0 20 0h 13 22 , ai - 1 - 1 2 - 1 - 1 - 0 DT 5 2 1
5 6 6 6
,I 9 $a *'.
' r _
e
. Protective Systen ::aximum Allowable Setpoints
-- 120 I
Four Pump - 100 h One Pump in Each Loop Set Points
@ Two Pumps in One Loop 80 l Three Pump T Set Points
. . 60 Two Pump Set Points
.. 40 2
t i
- .. 20 ;
i i f f i i 60 40 20 0 20 , 40 60 Power imbalance 5 i O 1 S Figure 4.h-1
LOCA Limited Maximum Allowable Linear lleat Rate 20 18 x jf , n = a y , 16 Y x 3 to 2 2m o C 14 4 12 0 2--- 4 6 8 10' 12 Axial Location from Bottom Of Care.ft.
COMPARISON OF MEASURED AND CALCULATED RADLiL CORE POWER DISTRIBLTION RESULTS AT STEADY STATE, EQUILIBRIUM XENON,1% FP CONDITIONS Me asurement Conditions Control Rod Group Positions Core Power Level 15.h gyp Gps 1-4 100 % wd Boron Concentration 1397 ppm Gp S 100 % wd Core Burnup 11.6 E FPD Cp 6 77 % wd Axial Imbalance -2.h %FP Cp 7 02 % wd Max Quadranc Tilt +1. o % Cp 8 3h % wd 1.0h 1.31 1 30 1.h5 1.16 1.25, 1.28 0,.92 g
- 1. [bs 1 30 1.25 1.38 1.20 1.37 1.35 0.86 1 5 1.h8 1.2h 1.26 1.01 0 95 0.76 8 1. 1.hh 1.31 1.26 1.08 1.00 0.80 1,27 1 32 0 98 0.9h 0.6h 0.h7 1.'h3s 1.30 0 96 0 93 0 72 0.h5
'l 07 1.08 0.80 0.68 1.1 1.02 0.82 0.62
'd:89 0.86 0.55
- 0. 9'bs 0.78 0.53
\
O'.63
' O.5hs s
i l E 6 I
~
X.XX Calculated Results F.ev (1)
, X.XX Measured ~ Restilts PD0 Segment (2) l f
v Figure h.h-3 L -
COMPARISON OF MEASURED AND CALCULATED RADI AL CORE POWER DISTRIBLTION RESULTS AT STEADY STAT 1, EQUILIBRIUM XENON, LO % FP CONDITIONS Measurement Conditions Control Rod Group Positions Core Power Level _j0. 0 %FP Cps 1-4 100 % ud Boron Concentration 1240 ppm Cp 5 100 % wd Core Burnup 2.0 EFPD Cp 6 72 % wd Axial Imbalance -0.? %FP Cp 7 02 % wd Max Quadrant Tilt -1,1 % Cp 8 100 % wd 0.95 1.20 1.21 1 38 1.13 1.22 1.23 0,.88 g 0, 1.16 1.22 1 3h 1.09 1.25 1.30 0.86 16 1.39 1.22 1.28 1.01 0 92 0 73 s 1. 0 1 30 1.25 1.26 1.07 0.98 0 76 1 23 1.36 1.08 0 99 0.65 0.k5
- 1. 1.31 1.11 1.02 0.72 0.k6 1 10 1.16 0.85 0 70 1.1 1.10 0.87 0.71
' N 01 3 0.89 0.57 09 0.85 0 58 h66 i
0.k s f i a b
~
X.XX Calculated Results , Pev (1) X.XX Measured Results, PLO segment (2) Figure k.h h
t Comparison of Measured and Calculated Axial Core Power Distritiution Results at Steady State, 2-D Xenon L ullibrium,10% l 4 FP Conditiens Calculated: Measured: O J l L [-
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i ! COMPARISON OF MEASURED AND CALCULATED RADIAL CORE POWER DISTRIBUTION I
- RESULTS AT STEADY STATE, EQUILIBRIUM XENON, k0 % FP CONDITIONS Measurement Conditions Control Rod Group Positions Core Power Level kl.6 %FP Cps 1-4 100 % wd Boron Concentration 1208 ppm Cp 5 100 % wd Core Burnup 12. 6_ E FP D Cp 6 76 % wd Axial Imbalance o.6 %FP Gp 7 nL % wd Max Quadrant Tilt -1.2s %
Cp 8 os % wd 0.98 1.29 1.27 1.h3l1.12 1.28 1.35 0.96 ___ g
- 1. 5k[ 1.27 1.26 1.37 1.15 1 33 1 32 0.8'_
1 1.h7 1.21 1.27 1.00 v.iT 0.78 8 1. 1.k1 1.31 1.26 1.07 0 99 0 78 1 ab 1.32 0 95 0 9h 0.6h 0.h8
- 1. 1.30 0 98 0 95 0 73 0.h6 i
i g02 1.11 0.82 0.71 14 1.05 0.83 0.66
' Ds 0 0.91 0.58
- 0. 0.81 0 55 N
06.67 8 0.5Bs s i i e b X.XX Calculated Results , P.ev (1) X.XX Measured Results , PD0 Segment (2) Figure h.h-6
m ( Comparison of Measured and Calculated Axial Core Power Distribution Results at St.eady State, 3-D Xenon Equilibrium, h0% FP Conditions Calculated: Measured: O
]
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COMPARISON OF MEASURED AND CALLULAfED RADIAL CORE POWER DISTRIBLTION - RESULTS AT STEADY STATE, EQUILIBRIUM XENON, 75 % FT CONDITIONS Me asurement Conditions Control Rod Group Positions Core Power Level 75 %FP Cps 1-4 100 % vd Boron Concentration 1277 ppm Cp 5 100 % Vd Core Burnup M5 EFPD Cp 6 100 % wd Axial Imbalance -o.9 %FP Cp 7 -6 % wd Max Quadrant Tilt +1.37 % Cp 8 100 % wd 1.28 1.19 1.03 1.15 1.0h 1.20 . 1.31 0 96 G 1.k s 1.19 1.06 1.19 1.08 1.30 1.3h 0.86 1 05 1.15 1.02 1.15 1.04 1.11 0 90 1 1.1 1.20 1.1h 1.18 1.09 1.11 0.88 1 02 1.13 0 99 1.11 1.12 0.Th 1 g 1.20 1.02 1.09 1.03 0.63 s i b 98 1.05 0.89 0.92
. 1. C 1.05 0.86 0.70 i hSh 0.85 0.60 0.80 0.56 0.'4 x
0.61 i 0.$ \ I i i Note: Calculated results are for [ all rods out. X.XX Calculated Results , ."av (1) X.XX Measured Results , F00 3e ; ent (2) Firure L.L-S
. 4 i.
Comparison of Measured and Calculated Axial Core Power Distribution Results at Steady State, 2-D Xenon Equilibrium, 75% FP Conditions Calculated: Measured: O r ,r-l.y_ _,_ _ __ _... _. _ _ _} _ 2 3_ 2.1 y
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COMPARISON OF MEASURED AND CALCULATED RADIAL CORE POWER DISTRIBLTION
. RESULTS AT STEADY STATE, EQUILIBRIUM XENON, 75 % FP CONDITIONS Meas urement Conditions Control Rod Group Positions Core Power Level . 75.3 %FP Gps 1-4 100 % wd Boron Concentrationlodl ppm Gp 5 100 % wd Core Burnup 26.0 EEPD Cp 6 76 % wd Axial Imbalance -17 %FP Gp 7 00 % wd Max Quadrant Tilt -1.5 %
Cp 8 as % wd 1.01 1.27 1.26 1.ho 1.12 1.24 1.28 0 95 E 1.15( 1 30 1.26 1.39 1.19 1.36 1.29 0.80 1 1.k3 1.21 1.24 1.02 0 97 0 79 8
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COMPARISON OF MEASURED MD CALCULA1ED RADIAL CORE POWER DISTRIBLTION
. RESULTS AT STEADY STARE, EQUILIBRIUM XENON,100 % FP CONDITIONS ~
Me as urenent Conditions Control Rod Group Positions Core Power Level 97.6 %FP Cps 1-4 100 % vd Boron Concent ration 1077 ppm Gp 5 100 % wd Core Burnup 31.5 EFPD Cp 6 90 % vd Axial Imbalance +2. 5 3 %FP Gp 7 15 % wd Max Quadrant Tilt +1. 7h % Cp 8 13.5 % wd 1.02 1.28 1.27 1.h3 1 16 1.2h, 1.2h q.88 g 1.A s 1.29 1.23 1.37 1.26 1 37 1.29 0.78 1 22 1.h5 1.23 1.27 1.00 0 92 0.74 5 1. 1.h1 1.29 1.22 1.07 0.98 0 75 1 25 1 33 0 98 0 95 0.66 0.h8 1.. 1 33 0 97 0.97 0 75 0.h6 i - 1 08 1.13 0.83 0.72 1.1 1.12 0.8h 0 58 6:92 0.90 0.58 0.93s 0.80 0 51 s 047 s 0 5h g i
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. X.XX !!easured Result.<, P30 Seg:ent (2)
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4.5 ROD WORTH AT POWER
. 4.5.1 PURPOSE . The purpose of this test was to measure differential rod worth at power.
From these data, integral rod worth over the region of measurement is obtained by integration of the differential rod worth curve. 4.5.2 TEST METHOD The method by which the differential rod worth was determined at power is the insertion / withdrawal method. In this technique, the controlling group is inserted for six seconds and then withdrawn for six seconds. The determination of differential rod worth is then found by use of the equation given below: dp/dh = (2p2 - pl - p3) / (2H2 - H1 - H3) Where: H1 = CRA(s) position (%) prior to motion H2 = CRA(s) position (%) after the insertion but prior to withdrawal H3 = CRA(s) position (%) after the withdrawal is terminated p1 = Reactivity prior to CRA motion (ak/k) p2 = Reactivity af ter the insertion but prior to the withdrawal (ak/k) p3 = Reactivity af ter the withdrawal is terminated (ak/k) Differential rod worth measurements were also performed at 75 %FP during rod configuration changes. These tests were performed to check differential rod worth minimum and maximum limits, and to obtain data for ccmparison to previously generated integral rod worth data. The measurements were performed over the following two changes in controlling group position at different times during 75 %ir ee: ting. Initial Control Final Control Rod Group Positions Rod Group Positions Case 1-5 6 7 8 1-5 6 7 8 Number % Withdrawn % Withdrawn 1 100 100 65 100 100 100 41 10 , 2 100 100 41 10 100 94 16 00 { 4.5.3 EVALUATION OF TEST RESULTS 4.5.3.1 Acceptance criteria Two acceptance criteria were specified for rod worth at power measurements as listed below: (a) The maximum dif{erential reactivity worth of any control rod group is less than 3.03 x 10' ak/k/%wd at power conditions. (b) The r. :nimum differential reactivity worth of any control rod group is ('- gsm*ter than 8.35 x 10-6 ak/k/%wd at power conditions. However, this limit does not apply for Control Rod Group 5 less than 25 %wd or Control
# Rod Group 7 greater than.75 %wd.
4.5-1
4.5.3.2 Integral Rod Worth at Power
.Since the integral rod worth is necessary for adequate reactivity control of the unit, integral rod worth curves were developed using the measured results
- as determined during zero power physics testing. From these results integral rod worth curves were developed for two dif ferent part-length rod positions of 100 and 35 %wd for zero power, 532 F conditions as shown in Figures 4.5-1 and 4.5-2. In similar fashion, integral worth curves were developed for full power, 575 F conditions as shown in Figures 4.5-3 and 4.5-4. These full power -
curves were generated by assuming that the same percentage deviation between measured and calculated total rod worth measurements determined at zero power, 532 F conditions would exist at full power, 579 F conditions. During power escalation testing at the 75 %FP plateau, differentia.1 rod worth measurements were performed on Control Rod Groups 6 and 7 as rod position was being adjusted between two test conditions. By integrating the differential curve, the integral rod worth was determined and comparison made to the pre-dicted results as given below: Measured Projected Case Integral Worth Integral Worth Number (% Ak/k) (% Ak/k) 1 0.21 0.29 2 0.25 0.30 Since the projected integral worth was bassi on fixed part-length rod position of either 100 or 35 %wd and because part-length rod position varied signifi-cantly, the above comparison, as was expected, gave marginal results. It is important to note that the projected integral worth curves have been well established by use of reactivity balances at power. In similar fashion the zero power curves were also verified. 4.5.3.3 Differential Rod Worth at Power
~
The determination of differential rod worth at power was accomplished by employing the equation given in Section 4.5-2. j i Table 4.5-1 is a summary of dif ferential rod worth measurements performed at power under steady-state conditions. Comparison of these values to the ac-ceptance criteria stated in Section 4.5.3.1 shows that these measured results are well within their respective limits.
. 4.
5.4 CONCLUSION
S Integrcl rod worths at full powar developed from Zero' Power Physics results for part-length rods at both 100 and 35 %wd, predicted rod worth adequately. Comparison of predicted worth using these curves against measured results
. using the insertion / withdrawal technique gave marginal results due to the f fact that the measured results were not done with the part-length rods at
(_ either 100 or 35 %wd. 4.5 ,- - -- - - ----
Differential rod worth measurements at power using the insertion / withdrawal method gave minimum and maximum dif ferential rod worths well within the limits of the acceptance criteria. I i e e e t 4.5-3 t.--- ,-,- __ - . . _ _
1 . . , . . . . , I i 1 t Heasure Differential Rod Worths at Steady State Conditions During the Performance of Reactivity Coefficient at Power t larvel Conc. Rod Group Position,%wd Differential Rod Worth, %Ak/k/% wd (% FP) (comb) 1-5 6 7 8 Pull (1) Pull (2) Pull (3) Average Predicted 40 1170 100 84 09 26 .0145 0142 .0152 .0146 N/A 40 1192 100 84 09 26 .0141 0138 .0140 N/A
; 75 124e 100 100 73 100 .00e7 . 0065 .00e9 .00e7 .00e7 Y
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