ML20116D234
| ML20116D234 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/15/1985 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20116D231 | List: |
| References | |
| NUDOCS 8504290334 | |
| Download: ML20116D234 (89) | |
Text
.
, START-UP TEST PROCEDURE l-2 CHEMICAL AND RADIOCHEMICAL MEASUREMENTS
- 1. PURPOqlE l
A. The principal objectives of the Chemical and Radiochemical Tests are 1 to secure information on the chemistry and radiochemistry of the reactor coolant, and to determine that the sampling equipment, -
procedures, and analytical techniques are adequate to supply the data i required to demonstrate that the chemistry of all parts of the entire reactor system meet specifications and process requirements.
4 B. Specific objectives of the test program include indirect observations of fuel clad integrity, evaluations of domineralizer operations by direct and indirect methods, measurements of filter performance, confirmation of condenser integrity, measurement of off-gas system parameters, and confirmation of certain process instrumentation calibratices. Data for these purposes is secured from a variety of sources: plant operating records, regular routine coolant analysis, radiochemical measurements of specific nuclides, and special chemical tests on fluids.
- 2. CRITERIA Level 1 A. Water quality must be known at all times and must remain within the guidelines of the water quality specifications.
t B. The activities of gaseous and liquid effluents must be known and must conform to license limitations.
- c. Chemical factors defined in the Technical Specifications and Fuel t warranty nust be maintained within the limits specified.
D. Radiation Monitoring Instrumentation must be responsive to radionuclide sources and/or effluents containing radionuclides.
Level 2 .
A. None i
8504290334 850415 PDR ADOCK 05000374 PDR p
DOCUMENT 0862r
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- 3. RESULTS Test condition open vessel Chemical process systems were tested. Sample instrumentation and laboratory equipment were calibrated during preoperational testing. Prior to fuel load, reactor water, condensate polisher inlet and outlet, and r feedwater samples were collected and analyzed. After fuel load, a reactor water sample was collected and analyzed. The results demonstrated acceptable water quality. All applicable acceptance criteria were satisfied during the performance of the test at this test condition.
Test conditions Heat-Up 1, 3, and 6 Analysis of radiolytic gas in steam and chemical / radiochemical tests of reactor water, condensate domineralizer inlet and effluent, feedwater, off-gas pre-treatment and post-treatment, and plant vent were conducted.
Measurements of stored water (clean demineralized water storage tank, cycled condensate storage tank, condenser hotwell, and heater drain tank) quality and condenser /feedwater systems filterable iron concentrations were taken.
All applicable acceptance criteria were satisfied during the performance of the test at these test conditions.
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i DOCUMENT 0862r L
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STARTUP TEST PROCEDURE 2-2 RADIATION MEASUREMENTS
- 1. PURPOSE The purpose of this test is as follows:
A. To determine the background radiation levels in the plant environs prior to operation for use as base data on activity build-up.
B. To monitor radiation at selected power levels to identify potential ,
deficiencies and assure the protection of personnel during plant operation.
C. To provide sufficient data (exposure rate and dose equivalent rates) to allow comparison of the actual dose rates with the design dose equivalent rates outside selected plant shield structures and room <
entrances for potentially radioactive equipment.
- 2. CRITERIA Level 1
- A. The radiation doses of plant origin and the occupance times of personnel in radiation zones shall be controlled consistent with the guidelines of the standards for protection against radiation as outlined in 10CFR20, NRC General Design Criteria.
Level 2 -
A. None
- 3. RESULTS Test Condition Open Vessel, Heatuo, 1, 2, 3, 5 end 6 This test was performed at test conditions open vessel, heatup, 1, 2, 3, -
5, and 6. The results of the radiation surveys showed that the radiation levels were within the above criteria.
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l DOCUMENT 0802r 3
STARTUP TEST PROCEDURE 3-2 FUEL LOADING
- 1. PURPOSE A. The purpose of this test is to load fuel safely and efficiently to 3 - the full core size.
- 2. CRITERIA Level 1.
A. The partially loaded core must be subcritical by at least 0.38%
AK/K with the analytically strongest rod fully withdrawn.
Level 2. ,
A. The core must be loaded in the analyzed core loading pattern as indicated on Table 1.
- 3. RESULTS Test Condition Open Vessel This test was performed during Test Condition Open Vessel. The partial core shutdown margin demonstrations were completed as follows:
16 bundles loaded : 12-31-83 _
64 bundles loaded : 1-4-84 144 bundles loaded : 1-4-84 At each of the 'bove core configurations the shutdown margin was demonstrated (i.e., the core remained subcritical with the analytically highest worth rod withdrawn). After the core had been loaded, the loading pattern was verified to be correct by video taped i
visual inspection. All applicable acceptance criteria were I satisfied.
l DOCUMENT 0760r/0010r 4
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STARTUP TEST PROCEDURE 4-2 FULL CORE SHUTDOWN MARGIN
- 1. PURPOSE A. The purpose of this test is to demonstrate that the reactor will be subcritical throughout the first operating cycle wit'. . 'ny single control rod fully withdrawn.
- 2. CRITERIA Level 1.
A. The shutdown margin of the fully loaded core with the analytically strongest rod withdrawn must be at least 0.38% delta K/K plus an additional margin for exposure of 0.261h delta K/K.
Level 2.
A. Criticality should occur within i 1.01k delta K/K of the predicted critical.
- 3. Results Test Condition Open Vessel The shutdown margin was demonstrated to be greater than 0.381k delta K/K plus an additional margin for exposure (0,261h AK/K). The actual value for shutdown margin was determined to be 2.791k AK/K.
l The actual initial critical was demonstrated to be within 0.0861h AK/K of the predicted critical. The Shutdown Margin Calculation and critical Eigenvalue Comparison Calculation have been included in Table 1 and Table 2.
All applicable criteria were met during the performance of this test.
l DOCUMENT 0761r/0010r 9
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Table 1 Shutdown Margin Calculation Kanal = KGl + KG2 + KG3 + KMT KG1 = 0.048 E (worth of all group 1 rods withdrawn).
K KG2 = 0.018 E (worth of all group 2 rods withdrawn).
K KG3 = 0.0023 E (worth of 18 group 3 rods @ 04).
K
+ 0.0052 E (worth of 9 group 3 rods @ 08).
K
+ 0.0002 E (worth of 1 group 3 rod @ 06).
K Total KG3 = 0.0077 E K
KMT = (-5.5 x 10-5) ( p )
TM = 150*F KMT = -0.00451 E (worth of reactivity to conspensate for neg.
K temp. coeff.). _
Kanal = 0.048 + 0.018 + 0.0077 - 0.00451
= 0.06919 E K
SDR = Kanal - KP - KSR KP = 0.0005254 E (reactivity correction for T = 140 seconds) l l
KSR = 0.0408 E (worth of strongest rod)
! K SDR = 0.06919 E -0.nnn%?%4 E - 0.0408 $ ^
K K K SDR = 0.02786 E = 2.8% E j K K l
l DOCUMENT 0761r/0010r to
Tablo 2 CRITICAL EIGIDIVALUE COMPARISON CALCULATIONS DK = K12 + K3 + KMT - KP - 1 * (100)
K12 = 0.9982 E (Keff with groups 1 & 2 withdrawn).
K K3 = 0.0023 E (worth of all group 3 rods @ 04).
K
+ 0.0052 E (worth of 9 group 3 rods @ 08).
K
+ 0.0002 E (worth of 1 group 3 rod @ 06).
K Total K3 = 0.00~17 E K
KMT = (-5.5 x 10-5) (TM-68)
TM = 150*F KMT = -0.00451 E (worth of reactivity to compensate for K neg. temp. coeff.).
KP = 0.0005254 E (reactivity correction for T = 140 seconds)
K DK=l0.9982+0.0077+(-0.00451)-0.0005254-1l*(100)
DK = 0.08646%
DOCUMENT 0761r/0010r ,
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. STARTUP TEST PROCEDURE 5-2 CONTROL ROD DRIVE SYSTEM
- 1. PURPOSE A. The purpose of this test is as follows:
- 1. To demonstrate that the control Ead Drive (CRD) System operates properly over the full range of primary coolant temperatures and pressures from ambient to operating.
- 2. To determine the initial operating characteristics of the entire CRD System.
- 3. To demonstrate the optimum settings for the CRD flow control loop by analysis of the transients induced in the CRD Hydraulic System by means of CRD flow setpoint changes.
- 4. To verify that the flow control valve (FCV) closes to a minimum position within 10 to 30 seconds in response to the maximum error signal (scram).
- 5. To demonstrate that the FCV maintains a constant flow within +3 gym as the reactor pressure changes from a shutdown condition to the normal operating pressure.
- 2. CRITERIA Level 1 A. Each CRD aust have a normal withdrawal speed less than or equal to 3.6 inches per second indicated by a full 12 foot stroke in greater than or equal to 40 seconds.
B. The mean scram time of all operable CRDs with functioning accumulators must not exceed the following times: (Scram time is
! measured from the time the pilot scram 5,alve solenoids are
( de-energized).
Position Inserted From Fully Scram Time Withdrawn (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 l
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DOCUMENT 0762r/0010r IL l
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C. Tha mean scram time of tha thre2 fectS2t CRDs in a two by two Errty
. must not exceed the following times: (Scram time is measured from the time the pilot scram valve solenoids are deenergized).
I Position Inserted From Fully Scram Time withdrawn (Seconds) 45 0.45 39 0.92 25 2.05 05 3.70 D. The scram insertion time of each control rod from full out to position 5, based on de-energization of the scram pilot valve selenoids as time zero, shall not exceed 7.0 seconds.
Level 2.
A. Each CRD aust have a normal insertion or withdrawal speed of' 3.01 0.6 inches per second indicated by a fall 12-foot stroke in 40 to 60 seconds.
B. With respect to the CRD Friction testa, if the differential pressure variation exceeds 15 psid for a continuous drive in, a settling test must be performed, in which case, the differential settling pressure should not be less than 30 psid nor should it vary more than 10 psid over a full stroke. Lower differential pressures are indicative of excessive friction.
Level 3 _
A. Upon receipt of a simulated or actual scram signal (maximum error),
the FCV must close to its minimum position within 10 to 30 seconds.
B. The CRD system flow should not change by more than i 3.0 m as reactor pressure varies from zero to rated pressure.
C. The decay ratio of any oscillatory controlled variable must be less than or en,ual to 0.25 for any flow setpoint changes or less than or l equal to 0.50 for system disturbances caused by the CRD's being stroked.
- 3. RESULTS Test condition open vessel The following CRD system testing was performed at test condition open vessel; l'
i DOCUMENT 0762r/0010r i '3 i
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- Ins rt and withdraw timing fcr all drivas
- Position indication verification for all drives
- Coupling checks for all drives.
- Friction testing for all drives.
- Single rod scrans with normal accumulator pressure for all drives.
- Single rod scrans (Eight drives - four sequence A and four sequence B with low accumulator pressure.)
All CRD testing met the applicable acceptance criteria during this test condition.
The only deficiency occurred during the position indication verification for the drives, position 48 for CRD 06-15 did not appear on the 4-rod display. A work request was written and the problem was corrected. This had no impact on the startup test.
Test Condition Heatup The following CRD system testing was performed in test condition heatup (rated reactor pressure unless noted otherwise);
- Flow control valve controller tuning at reactor pressures of 0 psig and rated.
- System flow response as reactor pressure is increased from 0 psig to rated.
- Normal insert and withdraw timing for eight drives (sequence A and B).
- Friction testing for eight drives (sequence A and B).
- Single rod scrams (eight drives) at intermediate reactor i pressures (600 and 800 psig).
- Single rod scrans with normal accumulator pressure for all drives.
- Single rod scrans (eight drives) with zero accumulator pressure.
}
During the continuous friction tests, CRD 34-35 exhibited a differential pressure variation which exceeded 15 psid (Level 2 criteria). The settling (single-notch) friction tests on CRD 34-35 could not be performed due to the following equipment problems; l - Failure of the compression fittings on the high pressure l hoses used for friction testing.
- HCU withdraw line vent valve for CRD 34-35 was discovered r
to be galled and unable to be closed. .
Since CRD 34-35 was movable and trippable (90% scram time of 2.51 seconds), the station and General Electric concluded that there was no problem leaving test condition heatup with the settling (single-notch) friction test outstanding.
During system flow controller tuning, the closure times of the flow control valves to a minimum position during a simulated scram signal exceeded the level 3 acceptance criteria (10-30 seconds). The station and General Electric found the closure times acceptable since the CRD pump does not runout during a scram, which is the basis for the level 3 closure time criteria. Also, tuning the flow control valves for a quicker response at rated pressure would have caused unstable valve performance at
! lower operating pressures. All other applicable criteria were satisfied l during this test condition.
h DOCUMENT 0762r/0010r
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See T-blo 1 fcr the technical specification reisted ccrau times obtained
. during this test condition.
Test condition 3 CRD scram timing was performed in conjunction with the Turbine Stop Valve Trip (STP-27-2).
Scram times for 26 drives were obtained during the scram, and their 90%
times were within the level 1 acceptance criteria (s 7.0 seconds). See Table 1.
All CRD testing at test condition 3 was satisfactorily completed.
Test Condition 6 CRD scram timing was performed in conjunction with the MSIV Full Isolation (STP-25-2) and the Generator Load Rejection (STP-27-2).
Scram times for 27 drives were obtained in conjunction with both scrams, and their 90% times were within the level 1 acceptance criteria (s 7.0 seconds). See Table 1.
Continuous and settling (single-notch) friction tests were performed on CRD 34-35. Upon the station's and General Electric's analysis of the friction traces obtained, the drive's response was found acceptable and no further investigation was required.
All CRD testing at test condition 6 was satisfactorily completed.
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DOCUMENT 0762r/0010r
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- TABLE 1 SCRAM TIME DATA (HEATUP)
AVE SCRAM TIMES SLOWEST AVE SCRAM TIME POSITION OF ALL CRDs (sec) IN A TWO-BY-TWO ARRAY (sec).
45 0.316 0.368 39 0.622 0.704 25 1.361 1.461 05 2.430 2.584 Slowest 90% scram time; CRD 22-03, 2.92 seconds.
Slowest 90% scram time; CRD 34-2*1, 2.68 seconds.
Slowest 90% scram time; CRD 30-51, 2.83 seconds.
DOCUMENT 0762r/0010r 16
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STARTUP TEST PROCEDURE 6-2 l SRM PERFORMANCE AND CONTROL ROD SEQUENCE i
4 1. PURPOS8 4
- A. The purpose of this test is to demonstrate that the operational i sources, source range monitor (SRM) instrumentation, and rod withdrawal sequences' provide adequate information to achieve
- criticality and' increase power in a safe and efficient manner. The
. effect of typical rod movements on reactor power will be determined.
I b 2. CRITERIA .
i
. Level 1 f A. There must be a neutron signal count-to-noise ratio of at least 2:1
I B. There must be a minimum count rate of 0.7 counts /second on the required operable SRM's or Fuel Loading Chambers.
C. The IRM's must be on scale before the SRM's exceed the rod block setpoint.
Level 2 A. None
- 3. RESULTS Test condition Open vessel, Heatup, Test Condition 1 s This test was successfully completed during test condition open vessel, test condition heatup and test condition 1.
4 During fuel loading and prior to the initial critical, all operable Fuel unding Chambers and Source Range Monitors, were demonstrated to have .
neutron signal-to-noise greater than 2:1 and a count rate greater than 0.7 CPS (Data included on Tt.ble 1).
O control rods were successfully withdrawn using both the A (test condition
- 1) and B (Test conditions open-vessel and heatup) control rod sequence to achieve criticality and initiate a power _ increase. Proper SRM-IRM overlap was demonstrated during the performance of this test.
All applicable criteria were satisfied during the performance of this test.
DOCUMENT 0764r/0010r "l
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TABLE 1 SOURCE RANGE MONITOR MINIMUM COUNT RATE AND SIGNAL-TO-NOISE RATIO
- 1. Fuel Loading Chamber prior to Fuel Lodding (O.V.)
FLC Count Rate (CPS) S/N Date/ Time Inserted withdrawn A 5.0 0 . ~1 6:1 12-30-83/1455 B 5.0 1.0 4:1 12-30-83/1455 c 10.0 3.0 2.3:1 12-30-83/1455 D 15.0 3.0 4:1 12-30-83/1455
- 2. Source Range Monitor during Fuel Loading (0.V.)
SRM Count Rate (CPS) S/N - Date/ Time Inserted withdrawn A 2.0 0.5 3:1 1-9-84/0930 B 1.9 0.2 8.5:1 1-9-84/0930 C 3.9 0.1 38:1 1-9-84/0930 D 6.5 0.18 35:1 1-9-84/0930 n
DOCUMENT 0764r/0010r G
TABLE 1 (Cont.)
SOURCE RANGE MONITOR MINIMUM COUNT RATE AND SIGNAL-TO-NOISE RATIO
- 3. Source Range Monitor prior to startup (O.V.)
SRM Count Rate (CPS) S/N Date/ Time Inserted withdrawn A 1.0 0.1 9:1 3-10-84/0300 B 1.8 0.1 17:1 3-10-84/0300 C 12.0 0.3 39:1 3-10-84/0300
, D 4.5 0.2 21:1 3-10-84/0300 l
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, DOCUMENT 0764r/0010r 9
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2 STARTUP TEST PROCEDURE 9-2 WATER LEVEL MEASUREMENT
- 1. PURPOSE The purpose of this test is a follows:
A. To check the calibration of the various narrow range and wide range indicators.
B. To measure the reference leg temperature and recalibrate the narrow and wide range instruments if the measured temperature is different than the value assumed during the initial calibration.
C. To collect plant data which can be used to investigate the effects of core flow velocity, carry-under, the subcooling on indicated wide range level.
d
- 2. CRITERIA Level 1 A. None Level 2 A. The narrow range level indicator readings on the instruments used for feedwater level control (2C34-R606A-C) should agree within i 1.5 inches of their average reading. ,
B. The narrow range level indicator readings on the instruments not used for feedwater level control (2B21-N024A-D, 2B21-NO38A-B, 2B21-M100A-B, and 2821-N101A-B) should agree within i 3.0 inches of l
j their average reading.
C. The wide range level system indicators (2B21-N026A-D, 2B21-N031A-D, 2B21-NO36A-D, 2B21-NO37A-D, 2B21-R604, 2C61-R010, 2B21-R884A-B should agree within i 6 inches of the average reading.
- 3. RESULTS Test condition Heatup All narrow and wide range level instruments satisfied their respective Level 2 criteria.
No new end point calibration calculations were required.
During the performance of STP 9-2, for Test condition Heatup through TC#6, the four fuel zone instruments (2B21-N044A, 2B21-N044B, 2B21-R610, and 2821-R615) read upscale. This indication is expected since the range of these instrtaments is -311 to -111", and the vessel level during the test was set at +36".
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DOCUMENT 0789r/0010r W
o t
, Tost condition 1 All narrow and wide range level instruments satisfied their respective Level 2 criteria.
Test Condition 2 All narrow and wide range level instruments satisfied their respective Level 2 criteria.
Test condition 3 All narrow and wide range level instruments satisfied their respective Level 2 criteria.
Test Condition 4 All narrow and wide range level instruments satisfied their respective Level 2 criteria during TC #4.
Test condition 5 All narrow and wide range level instruments satisfied their respective Level 2 criteria during TC #5.
Test Condition 6 All narrow and wide range Level instruments satisfied their respective Level 2 criteria.
i e
DOCUMENT 0~189r/0010r 11
STARTUP TEST PROCEDUE 10-2 IRM PERFORMANCE
- 1. PURPOSS To adjust the Intermediate Range Monitor (IRM) System to obtain an optimum overlap with the Source Range Monitor (SRM) System and the Average Power Range Monitor (APRM) System.
- 2. CRITERIA Level 1 A. Each IRM channel must be adjusted so that it is on scale on the lowest range before any SRM exceeds the rod block setpoint.
B. Each IRM channel must be adjusted so that it is not upscale on the highest range before all APRM's clear the APRM downscale trip setting.
C. The IRM's must produce a scram at 96% of full scale.
I.evel 2 A. None
- 3. RESULTS Test conditions Open vessel, HeatuD, Test Condition 1 and Test Condition 2 This test was performed at test conditions open vessel, heatup and 1.
Adequate SRM/IRM overlap was demonstrated immediately following the initial critical. At test condition heatup, each IRM demonstrated proper overlap with the SRM'S. All IRM's were adjusted to provide adequate IRM/APRM overlap. In addition, each IRM was adjusted to obtain proper range 6/7 correlation. At test condition'1, each IRM demonstrated adequate overlap with the APRM's. After ensuring adequate IRM/APRM ,
overlap, the SRM/IRM overlap was re-verified during the next startup which 1 occurred in test condition 2. At each test condition, all applicable test criteria were satisfied.
DOCUMENT 0763r/0010r u-1
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'- 11. . . __ _ . . _ _ _ . , . .
9 STARTUP TEST PROCEDURE 11-2 LPRM CALIBRATION
- 1. PURPCSE A. To verify proper response of the u> cal Power Range Monitoring (LPRM)
System to local changes in the reactor power level.
B. To calibrate the LPRM System.
~
- 2. SgJINELA_
Level 1 A. None.
Level 2 A. Each LPRM reading will be within i 10% of its calculated value (as determined by a process computer or offline calculation, from Traversing Incore Probe Power Distribution Data).
Level 3 A. At least three LPRM detectors in each LRPM string must respond to a local change in neutron flux to assure the proper connection of the LPRM detectors to their cables. If less than three detectors in a string ar.; operable, all operable detectors in that string must respond properly.
- 3. RESULTS Test Condition Heatup Because of the limited local and average core power levels encountered during test condition heatup, verification of proper connection of the LPRM detectors and readout equipment was completed for 13 of the 1*12 LPRM detectors. Per the procedure, the remainder of this test section was completed during test condition 1, when a sufficient number of LPRM detectors were indicating 5/125 or greater. No deficiencies were generated as a result of this testing.
Test condition 1 The remaining LPRM response to neutron flux testing was completed. All acceptance criteria were satisfied, as at least three detectors in an LPRM string indicated a response to the control rod pulls.
8 DOCUMENT 0765r/0010r M
At cpprcximataly 21% cora tharmal power thD LPRM'c wera calibrated with the TIP system. After the calibration, it was determined that 18 LPRM's were outside of the + 10% acceptance criterion. Based on startup testing on Unit 1, a recalibration was not performed since repeating the calibration did not improve the results during the Unit 1 test. This is due to the low pcneer level at which the testing was done. After consulations with General Electric, it was determined that this criterion failure should not stop the plant from proceeding to higher power test conditions where a more accurate calibration would be possible.
Although Startup Test Procedure 11-2 was not performed again until Test condition 3, at Test condition 2 an LPRM calibration was performed per normal plant procedures in conjunction with Process Computer Testing per Startup Test Procedure 13-2, and all applicable acceptance criteria were satisfied.
Test Condition 3 As stated above, the LPRM detectors were teuccessfully calibrated to read proportional to the neutron flux at their locations. This was accomplished by using the TIP system.
The test was satisfactorily completed at test condition 3 and all applicable acceptance criteria were satisfied.
Test Condition 6 The LPRM detectors were calibrated by the TIP system to read proportional to the neutron flux at their locations. The test was satisfactorily completed at test condition 6 and all applicable acceptance criteria were satisfied. _
l DOCUMENT O'165r/0010r
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STARTUP TEST PROCEDURE 12-2 APRM CALIBRATION
't
- 1. PURPOSE A. The purpose of this test is to calibrate the Average Power Range )
Monitor (APRM) system.
- 2. CRITERIA Level 1 A. The APRM channels must be calibrated to read greater than or equal to the actual core thermal power. However, recalibration of the APRM system will not be necessary from safety considerations if at least two APRM channels per RPS trip circuit have readings greater than or equal to actual core thermal power.
B. The APRM scram and rod block setpoints shall be set no higher than the limits specified in the Technical Specifications and the fuc1 warranty document.
C. In the STARTUP mode, all APRM channels must produce a scram at less than or equal to 15% of rated core thermal power.
Level 2 A. If the above Level I criteria are satisfied, then the APRM channels will be considered to be reading accurately if they agree with the heat balance to within i 7% of rated core thermal power.
- 3. RESULTS Test Condition Heatup During test condition heatup, the reactor was placed on a relatively constant heatup rate for approximately 60 minutes. Due to the limited" range of the APRM amplifier gains, the APRM gains were set as high as possible. The as left settings provided conservatism, as shown by the as left APRM gain adjustment factors which ranged from 0.57 to 0.76. This did not affect the plant operation and the gains were more accurately adjusted during subsequent test conditions. The upscale alarm and the neutron trip setting were demonstrated to be less than or equal to 12% and 15% thermal power respectively.
DOCUMENT 0766r/0010r 15
A Test condition 1 i Testing at test condition 1 was successfully completed, and all applicable criteria were satisfied. The APRMs were adjusted to the results of a
- manual heat balance, and alarm and trip settings were demonstrated to be less than or equal to their respective limits.
Test condition 2 i
Testing at test condition 2 was successfully completed, with the APkMs being adjusted to the results of a P-1 heat balance. As the calculation of MFLPD/FRP indicated a value greater than 1.0, the APRM's were adjusted to read greater than 100%
- MFLPD. This conservative adjustment of each APRM alleviated the need to adjust ths flow biased alarm and scram setpoints. The alarm and trip actings were subsequently verified to be less than their respective limits.
Test' Condition 3 Testing at test condition 3 was successfully completed, and all applicable criteria were satisfied. The APRMs were adjusted to the results of an OD-3 heat balance, and alarm and trip settings were demonstrated to be less than or equal to their respective limits.
Test Condition 5 Testing at test condition 5 was successfully completed, and all applicable criteria were satisfied. The APRMs were adjusted to the results of an OD-3 heat balance, and alarm and trip settings were demonstrated to be less than or equal to their respective limits.
Test condition 6 Testing at test condition 6 was successfully completed, and all applicable criteria were satisfied. The APRMs were adjusted to the.results of an 00-3 heat balance, and alarm and trip settings were demonstrated to be less than or equal to their respective limits.
WARRANTY RUN Testing during the fuel warranty run was successfully completed, and all applicable criteria were satisfied. The APRMs were adjusted to the results of an OD-3 heat balance, and alarm and trip settings were demonstrated to be less than or equal to their respective limits.
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DOCUMENT 0766r/0010r l N
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STARTUP TEST PROCEDURE 13-2 PROCESS COMPUTER' i l
- 1. PURPOSE A. The purpose of this test is to verify the performance of the process computer under plant operating conditions.
- 2. CRITERIA Level 1 A. None Level 2 6
. A. Programs OD-1, P1 and OD-6 will be considered operational when:
7 1. The MCPR calculated by BUCLE and the process compu:er either:
- a. Are in the same fuel assembly and do not differ in value by more than 2%, or
- b. For the case in which the MCPR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly, the MCPR calculated by the two methods shall agree within 2%.
- 2. The maximum LHGR calculated by BUCLE and_the process computer either:
- a. Are in the same fuel assembly and do not differ in value by more than 2%, or
- b. For the case in which the maximum LHGR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly, the MRPLHGR calculated by the tw2 methods the maximum LHGR's calculated by the two methods shall agree within 2%.
- 3. The MAPLHGR calculated by BUCLE and the process computer either:
- a. Are in the same fuel assembly and do not differ in value by more than 2%, or
- b. For the case in which the MAPLHGR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly, the MAPLHGR's calculated by the two methods shall agree within 2%.
- 4. The LPRM calibration factor calculated by BUCLE and the process computer agree to within two percent (2%).
- 5. The remaining programs will be considered operational upon successful completion of the static and dynamic testing.
1 DOCUMENT 0~195r/0010r u
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- 3. RESULTS Test condition Open - vessel Tecting at test condition Open vessel consisted of the static system test case. During this test the TIP System, process computer interface and program OD-1 (Whole Core LPRM Calibration) were verified to be consistent with the system design. The General Electric plant simulation software was used to simulate the appropriate plant conditions. This test was successfully completed on 2-14-84. All applicable test criteria were satisfied during this test condition.
Test Condition Heatup Testing at test condition heatup consisted of a TIP system hot and cold alignment. This section was successfully completed on 4-1-84.
All applicable test criteria were satisfied during this test condition.
Test condition 1 Program testing at test condition 1 consisted of TIP spacer dip verification, LPRM trip set polnt program (OD-18) testing, and the initial OD-1 (whole core LPRM calibration) testing. One computer software problem was encountered during the full core TIP set (OD-1). The computer was incorrectly storing the LPRM calibration constants (C). A software patch was developed and implemented by the General Electric and cr==nnwealth Edison Software Engineers. This testing was reperformed on 5-13-84 and the LPRM calibration constants were verified to be correct. 'his section was successfully completed and all applicable test criteria were satisfied during this test condition.
Test Condition 2 The Dynamic System Test Case, static OD-ll (pre-conditioning) program testing, nuclear thermal limits verification, and miscellaneous program testing at test condition 2 was successfully completed on 6-8-84. Five minor deficiencies (due to low power and exposure) were
( encountered during this test condition. The above problems had no adverse affects on the results of this test and were resolved in later test conditions. The thermal limits (maximum linear heat generation rate, minimum critical power ratio, and maximum average planar linear heat generation rate) calculated by the process computer were demonstrated to be in the same bundle locations and have values within 2% of the BUCLE results. A summary of this l
i comparison is provided below:
DOCUMENT 0795r/0010r 6
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- 1. MLHGR Comparison ' l METHOD MFLPD LOCATION A. P1 0.246 19-20-14 BUCLE. 0.245 19-20-14 B. P1 0.323 19-20-5 BUCLE 0.322 19-20-5
- 2. MAPLHGR Comparison METHOD MAPLH3R_ LOCATION A. P1 2.85 19-20-14 BUCLE 2.85 19-20-14 B. P1 3.65 27-26-12 BUCLE 3.64 27-26-12
- 3. MCPR Comparison METHOI), MCPR LOCATION A. P1 3.8282 19-28 BUCLE 3.827 19-28 B. P1 3.065 19-28 BUCLE 3.067 19-28 The testing also demonstrated that the LPRM calibration factors calculated by the process computer were within 2% of the BUCLE values. All applicable test criteria were satisfied during this test condition.
Test condition 3 Program testing at test condition 3 consisted of a verification of the asymmetric periodic nuclear log (P-1), and a nuclear thermal limits verification. A summary of this verification is provided below:
- l. MLHGR Casparison METHOD MFbFD LOCATION l' P1 0.616 23-10-12
- l. BUCLE 0.616 23-10-12
- 2. MAPLHGR Comparison DOCUMENT 0795r/0010r l.
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I METHOD MAPLHGR LOCATION P1 7.42 23-10-12 BUCLE 7.41 23-10-12
- 3. MCPR Comparison METHOD MCPR LOCATION P1 1.9377 25-28 BUCLE 1.940 25-28 The testing also demonstrated that the LPRM calibration factors calculated by the process computer were within 2% of the BUCLE values. This test condition was successfully completed on 7-3-84.
All applicable criteria were satisfied during this test condition.
Test Condition 6 Program testing at test condition 6 consisted of the Dynamic OD-ll (Pre-conditioning) program testing, nuclear thermal limits verification, and miscellaneous program testing. A sununary of the thermal limits verification is provided below:
- 1. MLHGR Comparison METHOD MFLPD LOCATION P1 0.888 13-16-5 BUCLE 0.887 13-16-5
P1 10.57 19-10-4 BUCLE 10.55 19-10-4
- 3. MCPR Comparison METHOD MCPR LOCATION P1 1.506 13-16 BUCLE 1.507 13-16 It was also demonstrated that the LPRM calibration factors calculated by the process computer were all within 2% of the BUCLE values. This test condition was successfully completed on 8-10-84. All applicable test criteria were satisfied during this test condition.
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DOCUMENT 0795r/0010r 50 l
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STARTUP TEST PROCEDUR3 14-2 REACTOR CORE ISOLATION COOLING SYSTEM
- 1. PURPOSE A. To verify the proper operation of the Reactor Core Isolation Cooling (RCIC) System over its expected operating pressure range.
- 2. CRITERIA Level 1 A. The average pump discharge flow must be equal to or greater than the 100% rated value after 30 seconds have elapsed from initiation on auto starts at rated reactor pressure.
B. The RCIC turbine shall not trip on overspeed or isolate during auto or manual starts.
Level 2 A. The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere.
B. The differential pressure switch for the RCIC steam supply line high flow isolation trip shall be adjusted to actuate at 290% of the maximum required steady state flow, with the reactor assumed to be near the pressure for Main Relief Valve actuation.
C. In order to provide an overspeed and isolation trip avoidance margin, the transient start first and subsequent speed peaks shall not exceed the rated RCIC Turbine Speed.
D. The speed and flow control loops shall be adjusted so that the decay ratio of any RCIC system related variable is not greater than 0.25.
E. During rated pressure cold quick starts, the margin-to-trip of the turbine steam exhaust pressure shall be at least 10 psi.
F. The average pump discharge flow must be equal to or greater than 100%
of the rated value after 30 seconds have elapsed from automatic initiation at 150 psig.
- 3. RESULTS DOCUMENT 0~16'ir/0010r 31
Test Condition Heatup RCIC Manual (controlled) and hot quick starts to the condensate Storage Tank (CST) were made to demonstrate proper controller operation and system performance at 150 psig and rated Reactor Vessel Pressure. During the 150 psig testing, rated RCIC flow of 600 gym could not be attained (535 gym maximum) through the full flow test line due to the high back pressure in the test piping. Upon further testing, it was determined that the discharge pressure of the RCIC Pump is more than sufficient to achieve rated flow with the reactor vessel at 150 psig. As a result, operability of the RCIC system is shown by achieving at least 450 gym through the test line with at least 400 psig discharge pressure at greater than 2100 rpm turbine speed at a reactor pressure of 150 t 15 psig. A RCIC cold quick start to the reactor vessel from the control room was performed, at 150 psig, but the flow controller operated erratically, so the system was shutdown and controller adjustments were made.
RCIC testing at test condition Heatup has been satisfactcrily completed, with the following deficiencies remaining open at the completion of test condition Heatup:
- 1. Minor steam leakage to the atmosphere was observed coming from the governor valve and the turbine gland. A work request was written to determine and repair the cause of this leakage.
- 2. The peak turbine exhaust pressure did not meet the 10 psig margin to trip acceptance criteria. This was scheduled to be monitored during future turbine starts.
Test condition 1
~
RCIC Manual (controlled) and a hot quick start to the condensate Storage Tank (CST) were made to demonstrate proper controller operation, after making controller adjustments at the end of test condition Heatup. RCIC cold quick start injections to the reactor vessel and the CST at rated reactor pressure and 150 psig were initiated from the control room. When the cold start at rated pressure was performed, the turbine tripped on high exhaust pressure. A test gauge was installed to monitor exhaust pressure at the suppression pool and another cold quick start was performed. The maximum exhaust pressure observed during this start was 5 l
psig. Based on this run and other starts performed during test condition l
1, the exhaust pressure was determined acceptable. Cold quick start vessel injection at rated reactor pressure was initiated from the remote shutdown panel, followed by flow step changes while injecting high (540 gym) and low (330 gym) RCIC flows into the vessel. A cold quick start vessel injection at 150 psig reactor pressure was also performed. These tests demonstrated proper controller settings and system operation.
The steam leakage from the governor and governor end turbine was so slight that it presents no insmediate hazard to personnel or equipment, and was determined to be acceptable by the Station and General Electric.
DOCUMENT 0767r/0010r 3 L.
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RCIC tsating wa3 complsted in T22t Condition 1 and tpplic.nble critsris were satisfied. Although all testing on the RCIC system was completed during Test condition 1, the steam flow isolation instrustentation setpoints which are calculated based on the change in bypass valve steam flow were recalculated during Test Condition 3 as a part of STP 26-2 when data was obtained on Bypass Valve Capacities (See STP 26 '!).
i DOCUMENT 0767r/0010r 33
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r STARTUP TEST PROCEDURE 16-2 SELECTED PROCESS TEMPERATURES
- 1. PURPOSE l A. To assure that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operations.
B. To identify any reactor operating modes that could cause temperature stratification.
C. To determine the proper setting of the low flow control valve limiter for the recirculation pumps to avoid coolant temperature stratification in the reactor pressure vessel bottom head region.
D. To familiarize plant personnel with the temperature differential limitations of the reactor system.
- 2. CRITERIA Level 1 A. The reactor recirculation pumps shall not be started nor flow increased unless the coolant temperatures between the steam done and bottom head drain are within 145'F (81*C).
B. The recirculation pump in an idle loop must not be started unless the loop suction temperature is within 50*F (28'C) of the active loop suction temperature if one pump is idle or the steam done temperature if two pumps are idle. -
Level 2 A. During two pump operation at rated core flow, the bottom head coolant temperature, as measured by the bottom drain line thermocouple, should be within 30*F (17'C) of the recirculation loop temperatures.
- 3. RESULTS Test condition Heatup At Test Condition Heatup, all of the applicable acceptance criteria were l
satisfied. The proper setting for the low flow control valve limiter for the recirculation pumps was determined to be the minimum valve position
- (5%). With the Recirculation FCV's at minimum valve position, Reactor l Water Cleanup Flow was decreased from 360 GPM to 100 GPM causing the bottom head drain temperature to decrease from 520'F to 503*F. The control rod drive flow was increased frca 46 GPM to 87 GPM which resulted in an additional decrease to the bottom head drain temperature from 503*F l' to 500*F.
I' DOCUMENT 0768r/0010r Y , ,
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Test conditions 1 and 2 At test condition 1 and 2, all of the applicable criteria were satisfied.
The temperature difference between the steam done and the bottom head drain was 25'F and 20*F (for Test Condition 1 and 2 respectively).
Therefore, at these test conditions minimal temperature stratification exists in the bottom head region.
Test condition 3 Analysis of the test data showed no evidence of thermal stratification in the event of double recirculation pump trip. All applicable acceptance criteria were satisfied. The following maximum delta temperatures were observed during testing.
One Pump Trip: Steam Dome - Botton Head 21*F Idle Loop - Active Loop 5'F
! Two Pump Operation: Botton Head - Recirc Loop 6*F Test condition 4 Analysis of the test data taken while both recirculation pumps were tripped for test condition 4 showed no evidence of thermal stratification. The following maFDEun delta temperatures were observed during natural circulation testing.
Steam Dome - Botton Head 36*F Idle Loop - Steam Dome 28'F Testing at test conditions 3 and 4 was successfully completed with all applicable criteria being satisfied.
Test Condition 6 Analysis of the test data showed no evidence of thermal stratification in the event of single recirculation pump trips. The following maximum delta temperatures were observed during testing.
One Pump Trip: Steam Dome - Botton Head 26.9'F Idle Loop - Active Loop 8'F Two Pump Operation: Botton Head - Recirc Loop 0.5'F Testing at Test condition 6 was successfully completed with all applicable criteria being satisfied.
1 1
1 DOCUMENT 0768r/0010r
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STARTUP TEST PROCEDURE 17-2 SYSTEM EXPANSICW
- 1. PURPOSE f
, A. Verify that the reactor drywell piping system is free and unrestrained with regard to thermal expansion.
B. Verify that suspension components are functioning in their specified i manner.
t
- 2. CRITERIA i
Level 1 A. There shall be no evidence of blocking of the displacements of any
- system component caused by thermal expansion of the system.
B. Electrical cables shall not be fully stretched.
C. Mangers shall not be bottomed out or have the springs fully stretched.
D. Snubbers shall be in the operating range about the midpoint of the total travel range at operating temperature.
E. The measured steady state displacement of the recirculation and main steam systems shall not exceed the allowable values.
Level 1 A. At a steady-state condition, the displacements of instrumented points with displacement measuring devices shall not vary from the calculated values. If measured displacements do not meet these criteria, the piping design engineer must be contacted to analyze the data with regard to design stresses.
B. During the heatup cycle, the trace of instrumented points on the main steam and recirculation systems shall fall within a range of 150 percent of the calculated value from the initial cold position in the direction of the calculated value and 50 percent of the calculated value from the initial position in the opposite direction of the calculated value.
C. Mangers will be in their operating range between the hot and cold settings.
DOCUMErr 0809r/0011r 36
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- 3. RESULTS j Test condition open Vessel i Cold readings for hangers and snubbers were obtained and recorded 4 from construction line walks performed during PT-RR-201. The data
- was reviewed by the design engineers and adjustments were made where
- required.
1
. Inspections of hangers and snubbers at 250*F were obtained and recorded from construction line walks performed during PT-RR-201 for the subsystems required. The results were found to be acceptable.
j Hot readings for hangers and snubbers at rated temperature were obtained and recorded from construction line walks performed during PT-RR-201 for the subsystems required. The results were found to be
- acceptable.
i
- Data was obtained and recorded from the Reactor Pressure Vessel Stablizer Shims' adjustment performed following the first major thermal cycle during PT-RR-201.
All hard-wired instrumentation on the Main Steam, Feedwater, and Recirc systems was verified installed properly. The testing was successfully completed and all applicable criteria were satisfied during this test condition.
Test condition Heatup f
Inspections of the remaining hangers and snubbers were performed at 250*F. This inspection turned up no abnormalities.
~
All applicable I criteria were satisfied.
i Hot readings for the remaining hangers and snubbers at rated i temperature were obtained and recorded. The data was reviewed by the j design engineers and adjustments were made where required.
t Displacements of instrumented points were recorded every 50*F for the Recirculation, Main Steam, and Feedwater systems. Displacements for seventy (70) points exceeded the Level 2 acceptance criteria and seven (7) points fell outside of the Level I criteria. Analyses by General Electric and Sargent and Lundy showed these exceptions to be
, minor with actual displacements yielding acceptable stress levels and l- recommended that testing continue. Testing was successfully completed and all other applicable criteria were satisfied during this test condition.
^
Test condition 1 Not readings for hangers and snubbers which had been adjusted tvere i obtained at rated temperature and recorded. General Electric and Sargent and Lundy have determined the drywell piping suspension to be i
acceptable. All applicable criteria were satisfied.
I
! DOCUMENT 0809r/0011r 37 I
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Dicplacements cf instrumented points were rsecrd2d at stsady state power. Displacements for eighteen (18) points exceeded the Level 2 acceptance criteria and three (3) points fell outside of the Level 1
^
criteria. Analyses by General Electric and Sargent and Lundy have shown these exceptions to be acceptable with actual displacements
+
yielding acceptable stress levels and recommended testing continue.
Following the third major thernal cycle, hanger and snubber readings for the Recirculation and Main Steam systems were recorded. General Electric found the suspension to be acceptable. All applicable
- criteria were satisfied.
During the course of testing, it was determined by General Electric that three (3) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test because the measurements from the functioning instrumentation indicated values !
which were well within acceptance criteria limits. The test was successfully completed and all other applicable criteria were satisfied at this test condition.
Test condition 3
- Displacements of instrumented points were recorded at steady state j power Displacements of seventeen (l*1) points exceeded the Level 2 thermal expansion criteria. Analyses by General Electric have shown these excepticas with respect to the Main Steam and Recirc Systems to be acceptable with actual displacements yielding acceptable stress
- levels. All Feedwater piping was found to expand as predicted and no level criteria were violated.
During the course of testing., it was determin_ed by General Electric that six (6) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test results because the measurements from the functioning instrumentation indicated values which were well within acceptance criteria limits. This test was successfully completed and all other applicable criteria were satisfied at this test condition.
Test condition 6 Displacements of instrumented poit.ts were recorded at steady state power. Displacements for sixteen (16) points exceeded the Level 2 acceptance criteria but analyses by General Electric on the Main
! Steam and Recirc System piping have shown these exceptions to be I
acceptable with actual displacements yielding acceptable stress levels. All feedwater piping was found to expand as predicted and no level criteria were violated.
l l During the course of testing, it was determined by General Electric
! that four (4) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test because the measurements from the functioning instrumentation indicated values which were well within acceptance criteria limits.
This test was successfully completed and all other applicable i criteria were satisfied at this test condition.
DOCUMENT 0809r/0011r JB l . - _ _ ,. - -
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STARTUP TEST PROCEDURE 18-2 CORE POWER DISTRIBUTION
- 1. PURPOSE A. To determine the core power dist.ribution in three dimensions.
B. To determine the reproductibility of the Traversing In-core Probe (TIP) system readings.
- 2. _ACCEPTANCit CRITERIA Level 1 A. None ,
Level 2 A. The total TIP uncertainty (including random noise and geometric uncertainties) obtained by averaging the uncertainties for all data sheets must be less than 6.0%.
NOTE A minimum of two and a maximum of six data sets may be used to meet the above criterion.
- 3. RESULTS Test Condition 3 Tip data was taken at this test condition. As this is the first data set of STP-18-2 and a minimum of two data sets are required for criteria evaluation, the evaluation will be performed following the second data set acquisiticn scheduled for test condition 6. No anomalies were noted.
Test Condition 6 Tip data taken at this test condition was evaluated along with that taken at test ccndition 3. The total TIP uncertainty for both data sets was found to be 1.89%. The applicable acceptance criteria were satisfied during this test condition.
DOCUMENT 0773r/0010r 34
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STARTUP TEST PROCEDUR3 19-2 CORE PERFORMANCE
- 1. PURPOSE A. The purpose of this test is to evaluate the following core performance parameters at test conditions 1 through 6 and the warranty run:
- 1. Maximum Linear Heat Generation Rate (MLHGR).
- 2. Minimum Critical Power Ratio (MCPR).
- 3. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).
- 4. Core Thermal Power (CTP).
- 2. CRITERIA Level 1 A. The Maximum Linear Heat Generation Rate (MLHGR) of any rod during steady state conditions shall not exceed 13.4 kw/f t.
B. The steady state Minimum Critical Power Ratio (MCPR) shall not exceed the limits as specified in the plant technical specifications.
C. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) shall not exceed the limits as specified in the plant technical specifications.
D. Steady State reactor power shall be limited to 3323 Mwt and values on or below the analyzed flow control line. -
Level 2 A. None
- 3. RESULTS Test Condition 1 i
l All core performance parameters were demonstrated to be within the range as required by the Technical Specifications during the performance at test condition 1. All applicable acceptance criteria were satisfied during the test condition. The calculated values are shown in the following table:
i TEST CONDITION 1 RESULTS OBSERVED LEVEL 1 CORE PARAMETER LOCATION VALVE CRITERIA LIMIT CTP ~122.6 Mwt < 1595 Mwt MLHGR 13-16-5 4.0~1 KW/ft 513.4Kw/ft MCPR 11-44 4.022 2 1.655 MAPLHGR 11-44-4 3.49 Kw/ft s 11.8 Kw/ft DOCUMENT O'l99r/0010r ao
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Test condition 2 j All core performance parameters were demonstrated to be within the range i as required by the Technical Specification auring the performance at test condition 2. All applicable acceptance criteria were satisfied during the
- test condition. The calculated valves are shown in the following table
I. TEST CONDITION 2 RESULTS 1
OBSERVED LEVEL 1
) CORE PARNEETER LOCATION VALVE CRITERIA LIMIT h
CTp 1000.1 Mwt $ 1662 Mwt MLHGR 11-28-12 4.21 Kw/ft < 13.4 Kw/ft
- l MCPR 41-18 3.233 E1.645 i MnPusaR i Fuel Type:
i 8CR 183 11-26-12 3.63 Kw/ft i 11.89 Kw/ft j 8CR 233 41-20-12 3.67 Kw/ft s 11.99 Kw/ft
,f 8CR 711 49-08-12 1.45 Kw/ft $ 11.53 Kw/ft Test condition 3 All core performance parameters were demonstrated to be within the range j as required by the technical specifications during the performance at test l
condition 3. All applicable acceptance criteria were satisfied during j this test condition. The calculated values are shown in the following
- table
- .
4 r
I TEST CONDITION 3 RESULTS 4
i OBSERVED LEVEL 1 CORE PARAMETER LOCATION VALVE CRITERIA LIMIT C't"? 2327.6 met s 3090 Mwt 4 L iGR 39-54-11 8.21 Kw/ft s 13.4 Kw/ft MCPR 25-34 1.9% 1 1.265 l MRPLHGR Fuel Type:
8CR 183 39-54-11 7.37 Kw/ft i 11.94 Kw/ft r 8CR 233 25-10-11 7.01 Kw/ft i 12.05 Kw/ft.
8CR 711 49-08-12 2.*18 Kw/ft $ 11.49 Kw/ft.
Test Condition 4 All core performance parameters were demonstrated to be within the range l as required by the technical specifications during the performance at test condition 4. All applicable acceptance criteria will satisfied during this condition. The calculated values are shown in the following table:
l DOCUMENT 0799r/0010r un
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TEST C M DITION 4 RESULTS OBSERVED LEVEL 1 CORE PARAMETER 14 CATION VALVE CRITERIA LIMIT CTP 1327.3 Mwt i 1595 Mwt MLHGR 37-36-4 5.36 Kw/ft i 13.4 Kw/ft MCPR 35-23 2.077 Kw/ft 1 1.693 MAPLHGR 4
Fuel Type:
) 8CR 183 37-36-4 4.61 Kw/ft i 11.98 Kw/ft j 8CR 233 25-26-4 4.68 Kw/ft s 12.06 Kw/ft 8CR 711 53-50-12 1.59 Kw/ft i 11.48 Kw/ft Test Condition-5 i All core performance parameters were demonstrated to be within the range m
- as required by the technical specifications during the performan
- e at test i condition 5. All applicable acceptance criteria were satisfied during the test condition. The calculated values are shown in the following table:
IEET CONDITION 5 RESULTS ossERv8D LEVEL 1
] CORE PPRAMETER LOCATION VALVE CRITERIA LIMIT CTP 2217.9 Mwt 1 2226.4 Mwt MLHGR 37-36-4 8.28 Kw/ft i 13.4 Kw/ft MCPR 1.7768 2 1.4942 Puel Type:
8CR-183 09-38-4 7.37 Kw/ft i 11.97 Kw/ft 8CR-233 37-32-12 7.24 Kw/ft $ 12.05 Kw/ft 8CR-711 11-54-11 2.84 Kw/ft i 11.48 Kw/ft Test Condition 6 .
I All core performance parameters were demonstrated to be within the range as required by the technical specifications during the performance at test condition 6. All applicable acceptance criteria were satisfied during this test condition. The calculated values are shown in the following table:
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DOCUMENT 0799r/0010r 4L
TEST CONDITION 6 RESULTS 'l OBSERVED LEVEL 1 CORE PARAMETER 14 CATION VALVE CRITERIA LIMIT CTP 2217.9 Mwt S 3289.8 Mwt MLHGR 13-16-5 11.87 Kw/ft s 13.4 Kw/ft MCPR 47-46 1.521 1 1.245 ;
MAPLHGR j Fuel Type: I 8CR 183 21-10-4 9.85 Kw/ft $ 12.02 Kw/ft 8CR 233 13-44-5 9.39 Kw/ft $ 12.24 Kw/ft 8CR 711 49-54-12 3.90 Kw/ft i 11.43 Kw/ft Warranty Run All core performance parameters were demonstrated to be within the range as required by the technical specifications during the performance during the warranty run. All applicable acceptance criteria will satisfied during this test condition. The calculated values are shown in the following table:
l Warranty Run !
l OBSERVED LEVEL 1 CORE PARAMETER LOCATION VALVE CRITERIA LIMIT CTP 3249.6 Mwt s 3265 Mwt MLHGR 21-10-4 11.02 Kw/ft $ 13.4 Kw/ft MCPR 47-16 1.522 1 1.245 l MAPLHGR ,
Fuel Type: (
8CR 183 21-10-4 9.85 Kw/ft s 12.02 Kw/ft 8CR 233 13-44-5 9.39 Kw/ft .< 12.24 Kw/ft 8CR *lll 49-54-12 3.90 Kw/ft 3 11.43 Kw/ft i
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DOCUMENT 0799r/0010r d3
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l STARTUP ThST PROCEDURE 20-2 l
STEAM PRODUCTION
- 1. PURPOSE A. To demonstrate that the Nuclear Steam Supply System is providing steam sufficient to satisfy all appropriate warranties as defined in the contract.
- 2. CRITERIA Level 1 A. The NSSS Parameters, as determined by using normal operating procedures, shall be within the appropriate License Restrictions.
Level 2 A. None Level 3 A. The NSSS shall produce an amount of steam of not less than 99.M quality at a pressure of 985 PSIA at the second isolation valve in accordance with the following equation while the reactor is operating at warranted core thermal power of 3293 MWT:
WPR ( Lb/Hr) = 11223.4 x 106 + 32,000 1191.5 - hFW _
Consistent with the 1967 ASME Steam Tables
- 3. R1SULTS Note: STP 20-2 was performed during the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty run following the completion of Test condition 6.
Warranty Run STP 20-2 was conducted following the completion of all TC 6 Startup Testing, with the exception of the No Cleanup Test of STP l-2, which was conducted concurrently.
The Unit operated satisfactorily within all thermal limits.
The measured steam flow was found to be .016% Less than rated, when analyzed in the manner specified by STP 20-1 (Unit #1)
DOCUMENT 0788r/0010r ua
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A second method found the mer ured stsam flow to be .298% below warranted.
i though both measured values are less than warranted, the results are I acceptable,' based on the fact that the accuracies of the calculations and rounding off of values are on the order of 1% and no contract action is required for deviations < 1% between measured and rated
- steam flows.
I
- The pressure at the 2nd MSIV was found to be 987 PSIA vs the Level 3 criteria of 985 PSIA, resulting in a pressure drop from the Dome to
, the 2nd MSIV of f APr s (STP-20-2) = 1013 - 987 = 26 PSIA I As compared to the rated value of APr s (Rated) = 1020 - 985 = 35 PSID i
The steam quality calculations found this pressure drop to be acceptable, with the moisture increase from the Dome to the 2nd MSIV being within design.
' Warranted conditions were maintained throughout the tesc with all applicable level criteria being satisfied.
The measured parameters were as follows LHGR = 11.015 i MRPLHGR = 9.85 MCPR = 1.523 _
Steam Quality = 99.843%
2nd MSIV Pressure = 987.36 PSIA The NSSS satisfactorily demonstrated the ability to supply steam near rated conditions without exceeding license limits.
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DOCUMENT 0788r/0010r U t*
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. STARTUP TEST PROCEDURE 21-2 CORE POWER-VOID MODE RESPONSE
- 1. PUltP065 A. The purposes of this test are:
- 1. To measure the stability of the core power-void dynamic response in test conditions 4 and 5.
- 2. To demonstrate that the behavior of the stability of the core power-void dynamic response is within specified limits in test conditions 4 and 5.
- 2. CRITERIA Level 1 A. The decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response.
Level 2 A. The decay ratio must be less than or equal to 0.25 for each total core process variable that exhibits oscillatory response.
- 3. RESULTS Test condition 4 The stability of the response to a rapid reactivity insertion (control rod motion and simulated pressure regulator failure) was demonstrated to be within the acceptance crit cion of less than 0.25 decay ratio. This test was successf211y completed at this test condition.
l Test condition 5 The stability of the response to a rapid reactivity (control rod motion and simulated pressure regulator failure) was demonstrated to be within the acceptance criterion (decay ratios less than 0.25). This test was successfully completed at this test condition.
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l DOCUMENT 0774r/0010r db I
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STARTUP TEST PROCEDURE 22-2 PRESSURE REGULATOR-
- l. pVRPOSE A. Tb demonstrate the optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulators.
B. To demonstrate the takeover capability of the backup pressure regulator via simulated failure of the controlling pressure regulator and to set the regulating setpoint difference between the two regulators to an appropriate value.
C. To demonstrate smooth pressure control transition between the turbine control valves and bypass valves when the reactor steam generation exceeds the steam flow used by the turbine.
- 2. CRITERIA Level 1 A. The transient response of any EHC System - related variable to any test input must not diverge.
Level 2 A. System - Related variable may contain oscillatory modes of response.
In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.
B. The response time from pressure input until the pressure peak of the pressure regulator sensed pressure must be'less than or equal to 10 seconds, with the recirculation flow control system in the position en-marut mode only.
C. Pressure control system deadband, delay, etc., shall be small enough that steady state limit cycles (if any) shall produce steam flow variations no larger than 1 0.5 percent of rated steam flow.
D. The normal difference between regulator setpoints must be small enough that the peak neutron flux and peak vessel pressure remain
.below the scram settings by 7.5 percent and 10 psi respectively, for
. the regulator failure test perfor:aed at test condition 6.
Level 3 A. Dynamics of both pressure regulators shall be essentially identical.
B. The variation in incremental regulation (ratio of the maximum to the airsimum value of the Quantity, " Incremental change in pressure L control signal / incremental change in steam flow," for each flow range) shall meet the following: ,
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DOCUMENT 0775r/0010r ~
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} % CF STERN FLOW WITH VALVES WIDE OPEN VARIATION 0 To 85% $ 4:1 i 85% TO 97% < 2:1 I 85% TO 99% 55:1 i
- 3. RESULTS l
^
Test condition 1 i
i Primary and backup pressure regulator testing verifying the basic stability of the steam loop and demonstrating optimum control system
- response was satisfactorily completed. Negative and positive step changes of 10 psi were induced into the pressure regulators with the turbine generator off line and load reference set for transients to be handled by the turbine bypass valves. Backup pressure regulator takeover capability
- was verified by failing the primary pressure regulator and observing proper system response. Test results and overall system performance were determined acceptable in test condition 1.
Test condition 2 Pressure Regulator Testing was performed with the turbine generator on line and load reference set for transients to be handled by 1) the turbine I control valves, 2) the turbine control valves and turbine bypass valves, l and 3) the turbine bypass valves. For each of the above operating conditions a negative and positive step change of 10 psi was induced into the pressure regulators and the backup takeover capability was demonstrated for both pressure regulators. Test results and overall system performance were determined acceptable in test condition 2.
i/- Test Condition 3 I
pressure regulator system testing was performed with load reference set
- for transients to be handled by the turbine control valves. Pressure setpoint step changes and simulated regulator failure tests were performed with the recirculation system in the position command and flux command j modes.
i l All applicable test results were acceptable with the exception of the margin to scram criteria of ;t 7.5% below the neutron flux scram setpoint for presssure regulator backup takeovec capability testing with the ,
recirculator system in flux command mode. Due to uncertainties in extrapolating data to the 100% rod line from data taken while at the 70%
rod line, and based on overall stable system performance during failure of the primary pressure regulator it was determined'to proceed to the next test condition and reevaluate the margin to scram when the recirculation system is in flux ccousand mode at later test conditions as unit two I approaches the 100% rod line. !
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DOCUMENT 0775r/0010r ug i i
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Test condition 5 l , Pressure Regulator Testing was performed with load reference set for 2
transients to be handled by 1) the turbine control valves, 2) The turbine control valves and turbine bypass valves, and 3) The turbine bypass valves. For.each of the above operating conditions, a negative and positive stepchange of 10 psi was induced into the pressure regulators l with the recirculation system in the position cosmand and flux cossaand modes. The backup takeover capability was demonstrated for both pressure regulators.with load reference set for transients to be handled by the turbine control valves and the recirculation sytem in the position cosmand and flux - rut modes. Attention was given to the margin to scram criteria when failing the primary pressure regulator with the recirculation system in the flux command mode. Extrapolating data to the 100% rod line from data taken while at the 96% rod line indicated that more than sufficient margin to scram exists, therefore all test results and overall system performance were determined acceptable in test condition 5.
Test condition 4 Pressure Regulator Testing was performed with load reference set for transients to be handled by 1)The turbine control valves, 2) The turbine control valves and turbine bypass valves, and 3) The turbine bypass valves. For each of the above operating conditions a negative and positive step change of 10 psi was induced into the pressure regulators and the backup takeover capability was demonstrated for both pressure regulators. Test results and overall system performance were determined acceptable in test condition 4.
i Test condition 6 ,
Power ascension data was obtained from when the' turbine generator was put on line and terminated at 95% power to determine the variation in incremental regulation (ratio of the maximum to the minimum value of the quantity, " incremental change in pressure control signal / incremental chenge in steam flow"). This specifies the linearity of the turbine control system from the point of receipt of the pressure regulation demand to turbine admission valve steam flow. Pressure regulator testing was performed with load reference set for transients to be handled by 1) the turbine control valves, 2) the turbine control valves and turbine bypass valves, and 3) the turbine bypass valves. For each of the above operating
> conditions a negative and positive step change of 10 psi was induced into the pressure regulators and the back up takeover capability was demonstrated for both pressure regulators. Test results and overall system performance were determined acceptable in test condition 6.
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DOCUMENT 0775r/0010r M
- 1. PURPOSE A. To demonstrate satisfactory reactor water level and feedwater flow rate control. Measurements of feedwater system stability and performance are analyzed for this determination.
I' 2. ACCEPTANCE CRITERIA Level 1 A. In the automatic mode, the response of any level system controlled
- variable to any test input change or disturbance must not diverge.
Level 2 T. . Level control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.
B. The average rate of response of the feedwater turbines to large (greater than 20%) step disturbance shall be between 10 percent to 25 percent of pump rated flow /second. This average respcase rate will be assessed by determining the time required to pass linearly through the 10% and 90% response points of the flow transient.
C. The dynamic flow response of each feedwater actuator (turbine or valve) to small (less than 10%) step disturbances in the manual mode shall be:
- 1. Dead time $ 1.0 sec.
- 2. Maximum time to 10%* $ 1.1 sec.
- 3. Maximum time from 10% to 90%* $ 1.9 sec.
- 4. Settling time to within i 5%*
of the final value s 14.0 sec.
- 5. Peak overshoot * $ 15%
j' * % of input step disturbance.
i Level 3 l
l- 1. Transmitters, square root converters, susmers, recorders, etc. of the
- total feedwater flow and steam flow circuits shall be calibrated and
<! adjusted properly so that the total steam flow recorder indication l: matches the total feedwater flow recorder indication within i 4% of rated feedwater flow.
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DOCUMENT 0776r/0010r i.
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- 3. EEE&It Igg,t condition 1 At test condition 1, the Motor Driven Reactor Feedwater Pump was subjected to level step response testing with the feedwater control system in the single element mode of control. This testing was completed suc(assfully and all applicable criteria were satisfied.
Iggt Condition 2 At Test Condition 2, each feedwater pump actuator was dynamically tested under loaded conditions. Optimum controller settings were chosen for low to mid-power operation. With the control system configured in this marmer, the Level 2 dynamic response criteria (decay ratio, rise time, overshoot, etc.) and average rate of response were not satisfied at all times. Even though these criteria were not satisfied, satisfactory feedwater control system performance and transient performance was indicated in the single-element and three-element modes of control.
Test condition 3 l The following tests were successfully performed in this test condition; open flow loop steps for MDRFP, TDRFP "A" and TDRFP "B", manual steps into closed flow loops for IGRFP, TDRFP "A" and TDRFP "B", level setpoint steps on startup controller for TDRFP "A" and TDRFP "B", level setpoint steps on Master controller for MDRFP, TDRFP "A", TDRFP "B", MDRFP and TDRFP "A",
1 ICRFP and TDRFP "B", and 2-TDRFP. Controller settings were chosen to j obtain stable level response at higher (normal operating) flows and to match the response of the reactor feed pump turbines, which is necessary for optimum level response when both turbine driven reactor feed pumps are in 3-element control. With the control system tuned in this manner, the ;
level 2 dynamic response criteria (decay ratio, overshoot, etc.) and average rate of response were not satisfied at all times. Even though i
these criteria were not satisfied, the feedwater control system is able to achieve steady control of reactor water level and provides stable response (recovery) from a transient. ;
Test condition 4 Level setpoint steps with both turbine driven reactor feed pumps in 3-element were successfully completed. There were no criteria violations during the test.
Test condition 5 Level setpoint steps with both turbine driven reactor feed pumps in 3-element were successfully completed. There were no criteria violations during the test.
Test condition 6 At test condition 6, level setpoint steps with both turbine driven reactor feed pumps in 3-element were successfully completed. All applicable criteria were satisfied.
DOCUMENT OT16r/0010r fi l _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ __ _ _ _ _
STARTUP TEST PROCEDURE 23-B-2
. FEEDWATER SYSTEM, LOSS OF FEEDWATER HEATER
- 1. PURPOSE A. To demonstrate adequate response to feedwater heater loss.
- 2. CRITERIA Level 1 A. The maximum feedwater temperature decrease due to a single failure case must te less than or equal to 100'F. The resultant MCPR must be greater than the fuel thermal safety limit.
B. The increase in simulated heat flux cannot exceed the predicted level 2 value by more than 2%. The predicted value will be based on the actual test values of feedwater temperature change and power level.
Level 2 A. The increase in simulated heat flux cannot exceed the predicted value referenced to the actual feedwater temperature change and power level.
- 3. RESULTS Note: STP 23-B-2 was performed during Test condition 6 (only)
Test Condition 6 Testing was satisfactorily completed during Test Condition 6. The feedwater temperature change and minimum MCPR observed when the extraction steam to both high pressure heaters was isolated were AT = 58.975'F and MCPR min =1.549. Both values were well within their respective Level I criteria.
In addition, the 1st & 2nd stage drain tank isolation valves (from the A &
B reheaters) to both H. P. Heaters were closed simultaneously with the Extraction Steam Isolation.
The reasoning for this action was that a postulated accident at th'e H. P.
Heater Local Control Panel (2PL36J), which would cause the extraction steam valves to close, would also cause the aforementioned drain valves to close. This action was in accordance with the intent of the test procedure, since it represented a further isolation of the H. P. Heaters.
The observed delta heat flux of 7.8% was well within the predicted value of 8.7%.
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DOCUMENT 0787r/0010r l !?
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- STARTUP TEST PROCEDURE 23C-2 FEEDWATER SYSTEM, FEEDWATER PUMP TRIP i 1. PURPOSE A. To demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump. l B. To demonstrate the ability of the standby motor driven feedwater pump to maintain water level if the turbine driven system is totally lost.
- 2. CRITERIA Level 1 A. None Level 2 A. A scram must not occur from low water level following a trip of one of the operating feedwater pumps. There should be greater than 3 inch water level margin to scram for a feedwater pump trip initiated at 100% power conditions.
- 3. RESULTS Test Condition 2 Testing was successfully completed at Test condition 2. The ability of the standby motor driven feedwater pump to maintain water level if the turbine driven system is totally lost was successfully demonstrated. A scram did not occur at test condition 2. The level margin to scram criteria was not applicable at test condition 2.
\
Test condition 6 Testing was successfully completed at test condition 6 and all applicable criteria were satisfied.
Test results determined a 12.9" margin to a low level scram at 100% core thermal power.
Analysis of transient recorder data determined that the level controller behaved as predicted in the control Systems Design Report (CSDR) in
- response to the reactor water level drop induced by the feedwater pump
! trip.
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DOCUMENT 0777r/0010r S3
STARTUP TEST PROCEDURE 23D-2 FREDWATER SYSTEM, MAXIMUM FREDWATER RUNOUT CAPABILITY.
. 1. PURPQS5 A. The purpose of this test is to determine the maximum feedwater runout capability.
'2. CRITERIA
Level 2 A. None.
Level 3
, A. None.
- 3. RESULTS Test condition 1 Testing of the 2A & 2B Turbine Driven Reactor Feed Pumps (TDRFP) High Speed limiters (HSS) was not performed in Test Condition 1 because only 1 TDRFP was available with the condition that 2 TDRFP would not be operated in parallel until the High Speed Limiters had been verified.
Test condition 2 The TDRFP2A HS3 was found to be 4690 RPM and the TDRFP2B HSS was found to be 4380 RPM. The endpoints of the turbine function generators were determined. The results of this testing satisfy all applicable test criteria.
Test condition 6 The maximum F/W Turbine Speed in manual control was satisfactorily performed (for both the A & B pumps) during Test Condition 6 with the plant operating between 93.4 and 94.M CMWT.
The TDRFP2A HSS was found to be 4791 RPM and the TDRFP2B HSS was found to be 4812 RPM. Both Speeds were determined via a certified stroboscope.
The combined maximum flow of each TDRFP is 19.03 x 106 Lb/HR, which is within the $ 19.2 x 106 LB/HR Level I criteria.
Upon an initial HSS analysis conducted between TC 2 and TC 6, a HSS value of approximately 5000 RPM was derived and accordingly the TDRFP2A & 2B settings were increased. Therefore, the HSS values measured during TC 6 were greater than those measured in TC 2.
DOCUMENT 0790r/0010r f4 p
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The limiting case was 135% flow Et 1075 PSIA Rx preEIure vargus 146% flow at 1020 PSIA, resulting in a maximum (conservative) HSS Setting of approximately 5500 RPM.
All applicable test criteria were satisfied.
I DOCUMEWT 0790r/0010r SC
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? STARTUP TEST PROCEDURE 24-2 1
( TURBINE VALVE SURVEILLANCE I
I
- 1. PURPOSE
- A. The purpose of this test are as follows:
3
- 1. To demonstrate acceptable procedures and maximum power levels for surveillance testing of the main turbine control, stop, and bypass valves without producing a reactor scram.
- 2. To establish baseline data for evaluation test condition acceptability with respe::t to PCICIEt during future startup !
1 tests.
- 2. CRITERIA Level 1 A. The decay ratio of any oscillatory response must be less than 1.0.
Level 2 A. The peak neutron flux must be at least 7.5 percent below the scram I trip setting. The peak heat flux must remain at least 5 percent below its scram trip setting. The peak vessel pre:ssure must remain at least 10 psi below the hitih pressure scram _sett:Ing.
, B. The peak steam flow in each I.ine must remain 10 percent below the high steam flow isolation trhp setting.
1 C. The decay ratio of any oscillatory response must be less than 0.25, when operating above the min:laus core flow for the recirculation master manual mode.
- 3. RESULTS L Test condition 3 l-Performance of turbine control valve, turbine stop valve, and turbine bypass valve testing was conducted at 70% core Thermal Poseer with load reference set high so that control valves maintain control of reactor pressure. Margin to scram criteria for neutron flux, heat flux and vessel
!! pressure were determined acceptable fet control valve, stop valve and bypass valve testing. Scee oscillatocy behavior was observed in the monitored plant variables however overall system performance was determined acceptable in Test condition 3.
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I DOCUMENT 0778r/0010r 86
Test condition 6 Performance of turbine control valve, turbine itop valve and turbine bypass valve testing was conducted at ~10%, 80%, 90% core ther.nal power with load reference set high so that control valves maintain control of
- reactor pressure. Itvaluation of margin to scram critoria for neutron flux, heat flux and vessel pressure and the nodal power changes during control valve, stop valve and bypass valve testing at all power levels dictated that control valve surveillances be conducted at power levels no
- higher than 90% of rated with special directives contained in the 7
operating surveillance procedures for pre-conditioning ramps to comply with PCIOMR limits. Stop valve and bypass valve survnillances may be
' conducted at 100% power as determined from Test condition 6 results. Some oscillatory behavior was observed in the monitored plant variables however
! overall system performance was determine acceptable at all power levels in l test condition 6.
4 DOCUMENT 0*Ml3r/0010r
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STARTUP TEST PROCEDURE 25-2 MAIN STEAM ISOLATION VALVES
- 1. PURPOSE A. To functionally check the s,ain steam line isolation valves (MSIV's) for proper operation at selected power level.
B. To determine isolation valve closure times.
p' C. To determine the maximum power at which full closure of a single valve can be performed without a scram.
'; D. To determine the reactor transient behavior resulting from the l
simultaneous full closure of all MSIV's.
p l 2. CRITERIA Level 1 A. MSIV closure time, excluding electrical delay shall be no faster than 3.0 seconds and including electrical delay shall be no slower than 5.0 seconds (each valve, not averaged).
B. The positive change in vessel done pressure occurring within 30 seconds after the full MSIV closure from greater than 95% of rated power must not exceed the Level 2 criteria, 6.2.D, by more than 25 psi. The positive change in simulated heat flux shall not exceed the Level 2 criteria, 2.D, by more than 2% of rated value.
C. Feedwater control system settings must preven't flooding of the steam lines.
Level 2 A. During full closure of individual valves:
- 1. Peak vessel pressure must be 10 psi below scram.
- 2. Peak neutron flux must be 7.5% below scram.
- 3. Steam flow in individual lines must be 10% below the isolation trip setting.
- 4. Peak heat flux must be % less than its trip point. .
B. Initial action of RCIC and HPCS shall be automatic if the level 2 setpoint is reached, and system performance shall be within specification.
C. The relief valves must reclose properly (without leakage) following the pressure transient.
DOCUMENT 0779r/0010r
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D. For t'.te full MSIV closure from grGoter than 95% power, predicted analytical results based on beginning of cycle design basis analysis, assuming no equipment failures and applying appropriate parametric corrections, will be used as the basis to which the actual transient is compared. The following table specifies the upper limits of these criteria during the first 30 seconds following initiation of the indicated conditions:
INITIAL CONDITIONS CRITERIA Dome Increase In Increase In i Power Pressure Heat Flux Dome Pressure
! (%) (psia) (%) (psi) 0
- 100 1020
- To be determined based upon actual plant conditions at the time the test is performed.
Test Condition Heatuo and 1 The MSIV's were " slow closed" to functionally demonstrate proper valve operations. Each MSIV was " fast closed" to determine valve closure times. The above closure time data has been included in Table 1.
This test was successfully completed for test conditions heatup and I with all applicable criteria met.
Test condition 2 -
The MSIV's were " slow closed" to functionally demonstrate proper valve operation. The fastest MSIV (from Test Conditions Heatup and 1) was " fast closed" a second time to verify that the above single valve closure would not cause a reactor scram. The above closure time data has been included in Table 1.
This test was successfully completed for test condition 2 with~ ail applicable criteria met.
Test Condition 5 l
The MSIV's were " slow closed" to functionally demonstrate proper valve operations.
This test was successfully completed for test condition 5 with all applicable criteria met, c
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l DOCUMENT 0779r/0010r 54
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Test condition 6
- The MSIV's were " slow closed" to functionally demonstrate proper valve operation. Individual valve closure, of the fastest MSIV, was performed twice during the ascension to test condition 6, to determine the maximum i power level at which a single valve can be closed. The maximum power, determined by extrapolation of data, was determined to be 90%. A simultaneous full closure of all MSIV's was initiated at 96.4% of rated power, and the reactor transient was observed. Individual valve closure times were determined and these times have been included on Table 1.
This test was successfully completed for Test condition 6 with all acceptance criteria being satisfied.
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i DOCUMENT 0779r/0010r 60
TABLE 1
)
MSIV CLOSURE TIME PERFORMANCE I i
MAIN STEAM ACTUAL CLOSURE CLOSURE TIME PERCENT TEST :
ISOLATION VALVES TIME
- LIMIT + POWER CONDITION 2821-F022A 4.17/4.33 3.0/5.0 6.0% Heatup f
2B21-F022B 4.07/4.29 3.0/5.0 6.0% Heatup 2B21-F022C 3.95/4.09 3.0/5.0 6.0% Heatup 2B21-F022D 3.71/4.0 3.0/5.0 6.0% Heatup 2B21-F028A 4.24/4.37 3.0/5.0 6.0% Heatup 2B21-F028B 4.11/4.39 3.0/5.0 6.0% Heatup u
2B21-F028C 4.43/4.61 3.0/5.0 6.C% Heatup 2B21-F028D 4.18/4.36 3.0/5.0 6.0% Heatup 2B21-F022A 4.18/4.31 3.0/5.0 19.5% 1 2B21-F022B 4.1/4.35 3.0/5.0 19.5% 1 2B21-F022C 3.95/3.99 3.0/5.0 19.5% 1 2B21-F022D 3.61/4.09 3.0/5.0 19.5% 1 2B21-F028A 3.89/4.17 3.0/5.0 19_.5% 1 !
2B31-F028B 4.0/4.4 3.0/5.0 19.5% 1 2B21-F028C 4.54/4.67 3.0/5.0 19.5% 1 2B21-F028D 3.94/4.27 3.0/5.0 19.5% 1 2B21-F022D 3.72/3.92 3.0/5.0 46% 2
- MSIV closure tiw;, excluding electrical delay / including e:.ectrical delay
+MSIV closure time,' minimum / maximum values l
t l
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DOCUMENT 0779r/0010r 61 -
TABLE 1 1
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MSIV CLOSURE TIME PERFORMANCE HAIN STEAM ACTUAL CLOSURE CLOSURE TIME PERCENT TEST i ISOLATION VALVES TIME
- LIMIT + POWER CONDIT1 M j 2B21-F022D 3.57/3.85 3.0/5.0 70.3% 6 2B21-F022D 3.65/3.85 3.0/5.0 80.6% 6
! 2B21-F022A 4.28/4.4 3.0/5.0 96.4% 6 l 2B21-F022B 4.07/4.29 3.0/5.0 96.4% 6 2B21-F022C 3.65/3.69 3.0/5.0 %.4% 6 2B21-F022D 3.65/3.99 3.0/5.0 %.4% 6 1
l 2B21-F028A 3.96/4.28 3.0/5.0 %.4% 6 1 2B21-F028B 4.07/4.19 3.0/5.0 %.4% 6 2B21-F028C 4.42/4.75 3.0/5.0 96.4% 6
! 2B21-F028D 3.93/4.12 3.0/5.0 96.4% 6 I
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i STARTUP TEST PROCEDURS 26-2 i
j RELIEF VALVES j
l 1. PURPOSE
- A. To verify the proper operation of the primary system relief valves.
f B. To verify that the discharge piping is not blocked.
C. To verify that each relief valve reseats following operation.
D. To obtain a transient recorder signature of each relief valve operation for subsequent comparisons.
E. To confirm proper overall functioning of the low-Low Set Pressure Relief Logic.
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- 2. CRITERIA Level 1 A. There should be positive ~ indication of steam discharge during the manual actuation of each valve.
B. The Low-Iow Set Pressure Relief logic shall function to preclude
, subsequent simultanious SRV actuations following the initial SRV actuation due to the original pressurization transient.
Level 2 -
A. No observable leakage shall exist following reclosure.
. B. The pressure regulator must satisfactorily control the reactor transient and close the control and/or bypass valves by an amount .
equivalent to the relief valve steam flow.
C. The transient recorder signatures for each valve must be analyzed for a relative system response comparison. The delay time (between trip and motion) shall be less than or equal to 0.1 seconds, and the
! response time (main disk stroke time) shall be less than or equal to 0.15 seconds.
D. The selected MSRV with the highest nominal safety spring setting must indicate full open when manually actuated with its accumulator air
, supply isolated and vented.
E. When the Low-Low Pressure Relief logic functions, the open/close actions of the SRV's shall occur within i 13 psi and i 20 psi of ,
- their design setpoints respectively.
F. Steam flow through each relief valve as measured by the initial and final bypass valve positions shall not be less than 10% of valve position below the average of all valve position responses.
DOCUMENT 0780r/0010r fo3
- 3. RESULTS Test condition Heatup Each safety relief valve was opened and closed demonstiating proper '
- functioning. Positive steam discharge for each valve actuation was indicated by bypass valve position indication and SRV tailpipe temperature increase responses. valve reseating was demonstrated by proper tailpipe temperature decrease following valve closure. The safety relief valve with the highest spring setting was manually opened and closed with its accumulator air supply header isolated and vented.
This test was successfully completed for test condition heatup with all applicable criteria being satisfied.
Test Condition 3 As a part of this testing, a bypass valve calibration was performed. This consisted of a set of data taken to determine the relationship between bypass valve position and bypass valve flow. This data was used'~to verify the setpoints of the RCIC steam supply line high flow isolation trip.
This data showed that the existing setpoints are conservative and therefore would not be changed. Each SRV was cIcled manually, from the control room, in order to determine the delay and stroke times of each SRV. The Level 1 and 2 criteria for the Low-Low Set Pressure Re1.ief Logic was not applicable at Test Condition 3 because the pressure spike resulting from the Turbine Trip (STP 27-2) did not reach the trip setpoint for the SRV's.
This test was successfully completed for test condition 3 with all applicable criteria being satisfied. _
Test Condition 6 Relief valve testing was performed in conjunction with the MSIV Closure and Generator Load Reject Tests at greater than 9% of rated power.
During both tests, the Low-Low Pressure Relief Logic performed according to design.
This test was successfully completed for test condition 6 with all criteria being satisfied.
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- STARTUP TEST PROCEDURE 27-2 GENERATOR LOAD REJECTION
- 1. PURPOSE s A. The purpose of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and the generator.
- 2. CRITERIA Level 1 A. Por Turbine and Generator trips there should be a delay of less than 0.1 seconds following the beginning of control or stop valve closure before the beginning of bypass valve opening. The bypass valves should be opened to a point corresponding to approximately 80 percent of their capacity within an additional 0.2 seconds, or 0.3 seconds total, from the beginning of control or stop valve closure motion.
B. Feedwater system settings must prevent flooding of the steam lines folicwing these transients.
C. The two recirculation pump drive flow coastdown transient during the first three seconds must be equal to or faster than that specified in this procedure.
D. The positive change in vessel done pressure occurring within 30 seconds after either generator or turbine trip must not exceed the
. Level 2 criteria by more than 25 psi. -
E. The positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2% of rated value.
F. Turbine speed does not reach the point where a mechanical overspeed
.i .
turbine trip would occur.
h Level 2 A. There shall be no MSIV closure in the first three minutes of the transient and operator c.ction shall not be required in that period to avoid the MSIV trip.
0:
- B. The positiver change in vessel done pressure and in simulated heat flux which occurs within the first 30 seconds after the initiation of i either generator or turbine trip must not exceed the predicted g values.
U ll C. Electrical load transfers occur as designed.
U P D. The reactor shall not scram for initial thermal power at less than or t,? equal to 25% of rated.
l DOCUMENT 0781r/0010r i:
E. If the Leval 1 criterion (2.c of this rsport) fcr tha two
. recirculation pump drive flow coast down transient is passed, the i data shall be analyzed within 3 weeks for compatibility with the safety analysis.
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- 3. RESULTS Test condition 2 6
A generator load rejection within bypass valve capacity was initiated by opening the generator circuit breakers.
This test was successfully completed for test condition 2 with all
- criteria being satisfied.
! Test Condition 3 A trip of the main turbine, from an intermediate power level, was initiated by depressing the Main Turbine Master Trip pushbutton. ,When the data was analyzed for the Recirculation Pump coastdown, it was determined that the pumps coastdown exceeded the Level 2 criteria. This data was submitted to G.E., San Jose for analysis. The data was reviewed by the Plant Transient Performance Engineering and ECCS Engineering departments and they determined the coastdown was acceptable.
This test was successfully completed for test condition 3 with all the criteria being satisfied.
4 Test condition 6
! A generator load reject from 95.5% of rated power was initiated by opening the generator circuit breakers. As in test condition 3 the Recirculation Pump Coastdown exceeded the Level 2 criteria. G.E. San Jose again reviewed the data and determined the Recirculation Pump Coastdown to be acceptable.
This test was successfully completed for test condition 6 with all the criteria being satisfied.
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DOCUMENT 0781r/0010r b6 .
TABLE 1 Parameter Worst Case Measured Temperature at the tube side outlet of the non-regenerative heat exchangers Blowdown Mode 93.0*F Normal Mode 101.0*F Pump available NPSH Normal Mode 2197.2 ft.
Blowdown Mode 2216.2 ft.
Hot Standby 2116.0 ft.
Both pumps 9 2~10 gpa 27.56 ft.
Cooling water supplied to the non-regenerative heat exchangers flow and temperatue Normal Mode T = 149'F W = 442 gym Blowdown Mode T = 180*F W = 445 gym Botton head flow indicator /RWCU _
flow indicator deviation 5 gpm Pump vibration all values less than
.20 ips I
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DOCUMENT 080lr/0010r VI
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STARTUP TEST PROCEDURE 29-2 I
RECIRCULATION FLOW CONTROL SYSTEM i
f 1. PURPOSE The purposes of this test are:
1 A. To demonstrate the core flow system's control capability over the entire flow control range, including valve position, core flow,
- neutron flux, and load following modes of operation.
i B. To determine that all electrical compensators and controllers are set for desired system performance and stability.
- 2. CRITERIA Level 1 A. Position Loop Criteria.
- 1. The position loop response to test inputs shall not diverge.
B. Flow Imop Criteria.
- 1. The flow loop response to test inputs shall not diverge.
C. Flux Loop Criteria.
- 1. The flux loop response to test inputs shall not diverge.
D. Load Following Loop Criteria.
- 1. The load following loop response to test inputs shall not diverge.
E. Scram Avoidance and General Criteria.
- 1. None.
F. Flow Control Valve Duty Test Criteria.
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[ 1. None.
4 G. Flux Estimator Criteria l-l 1. None.
Level 2 1'
A. Position Loop Criteria.
4 I
DOCUMENT 0812r/00llr 66
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q Gains and limitcra Ehall be cct to obtain the following rsrponsa:
i
- 1. Maximum steady state rate of change of valve position shall be
, between 9 & 11% per second for a 100% position demand input.
a (Initial valve velocity may exceed this limit for a short time.)
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- 2. Gains shall be set to give as fast a response as possible for
. small position demand input within the overshoot criterion and 1 I
without additional valve duty cycle. (See FCV duty criterion
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(F.1) for valve duty cycle requirement.)
- 3. The decay ratio of any oscillatory controlled variable must be s f'
0.25, when operating above ..he minimum core flow for Recirculation Master Manual mode. Below this minimum core flow, the decay ratio must be s 0.50 with the recommendation that each control system be adjusted to meet s 0.25 unless there is an identifiable performance loss involved at higher power levels.
B. Flow Loop Criteria.
- 1. The decay ratio of any oscillatory controlled variable must s 0.25, when operating above the minimum core flow for Recirculation Master Manual mode. Below this minimum core flow, the decay ratio must be s 0.50, with the recommendation that each control system by adjusted to meet $ 0.25 unless there is an identifiable performance loss involved at higher power levels.
- 2. The flow loops provide equal flows in the two loops during steady state operation. Flow loop gains should be set to correct 90% of a flow imbalance in 20 1 5 sec.
C. Flux Loop Criteria.
- 1. The decay ratio of any oscillatory controlled variable must be s O.25, when operating above the minimum core flow for Recirculation Master Manual mode (loops A and B receive command ,
j from a common point). Below this minimum core flow, the decay ratio must be s 0.50, with the recommendation that each control
! system by adjusted to meet $ 0.25 unless there is.an identifiable performance loss involved at higher power levels.
- 2. For small flux command step changes of between 1%-5%, at near rated power, the following apply:
j a. Deadband, percent rated flux demand: $ 0.5.
- b. Delay time for flux demand steps, sec.: $ 0.8. #
}
- c. Response time for flux demand steps, sec.: $ 2.5.
1 d. Maximum allowable flux overshoot, for step demand of s 20%
of rated is, in percent: 2.
- e. Flux settling time, sec.: 5 15.
i DOCUMENT 0812r/0011r M
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- 3. Switching between estimated and actual flux should not sxceed 5 i e times /5 minutes at steady state.
- 4. During flux step transient there should be no switching to actual flux or if switching does occur, it should switch back to estimated flux within 20 seconds of the start of the transient.
- 5. The deadband of the flux controller for a flux demand step shall be s .5% of rated flux demand.
D. Imad Following Loop Criteria.
- 1. The decay ratio of any oscillatory controlled variable must be $
l 0.25, when operating above the minimum core flow for Recirculation Master Manual mode. Below this minimum core flow, the decay ratio must be $ 0.50, with the recommendation that each control system be adjusted to meet 1 0.25 unless there is an identifiable performance involved at higher power levels.
- 2. The response to a step input of less than 10% in load demand shall be such that the load demand error is within 10% of the magnitude of the step within 10 seconds.
- 3. When a load demand step of greater than 10% is applied (N%), the load demand error must be within 10% of the magnitude of the step within N seconds.
- 4. For large Auto Load following Recirculation system maneuvers p along the 100 percent rod line, 90 percent of the commanded step power change (P) must be completed within (t) seconds:
- a. For 10 percent change, 9 percent within 10 seconds,
- b. For 20 percent change, 18 percent within 20 seconds.
- c. For 35 percent change, 31.5 percent within 35 seconds.
- 5. The automatic load following range along the 100 percent (Flow control) rod line shall be at least 35 percent power (ie.,
between 65 percent and 100 percent power).
E. Scram Avoidance and General Criteria. For anyone of the above loops' a test maneuvers, the trip avoidance margins must be at least the following:
- a. For APRM 1 7.5%.
- b. For simulated heat flux 1 5.0%.
- c. The system response in any mode response shall produce steady steam flow limit cycle variations no larger than 0.5% of rated steam flow.
F. Flow Control Valve Duty Test Criteria.
- 1. Se flow control valve duty cycle in any operating mode shall not exceed 0.2% -Hz. Flow control valve duty cycle is defined as:
DOCUMENT 0812r/0011r
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, Total valve travel (%) (% - Hz)
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2x time span in sec.
G. Flux Estimator Criteria t
- 1. Switching between estimated and actual flux should not exceed 5 times per 5 minutes at steady state.
- 2. During flux step transients there should be no switching to actual flux, or if switching does occur, it should switch back g to estimated flux within 20 seconds of the start of the transient.
i Level 3 A. Position Loop Criteria
- 1. Position loop deadband shall be 0.25% of full valve stroke (hot only).
- 2. Overshoot after a small position demand input (0.5 to 5%) step shall be $ 10% of magnitude of input.
- 3. Performance for 0.5 to 5% steps:
Delay: $ 0.15 seconds Response: $ 0.45 seconds
- 4. Time required from peak of first overshoot for output of 0.2 to 5% steps to settle within a range about the final value is s one second. The range will be + 5% of the step change or + 0.05%
full stroke, whichever is graater.
- 5. Any limit cycle in the sensed feedback position of the closed position loop must be less than 0.1% (peak to peak) of the full range as observed on the position feedback transmitter.
B. Flow Loop Criteria.
- 1. Deadband from function generator for valve position demand input to flow shall be $ 0.5%. .
- 2. The delay time for flow demand step 1 to 5% shall ba 5 0.4 sec.
- 3. The response time for flow demand step 1 to 5% shall be $ 1.1 i sec.
- 4. The maximum allowable flow overshoot for step demand of 1 to 5%
of rated shall be 6% of rated.
- 5. The flow demand step settling time shall be $ 6 sec.
i DOCUMENT 0812r/0011r 9
- 6. Linerrization cf th2 open flow control loop, using function generators must be adjusted so that the graphical slope changes do not exceed a factor of 2 to 1 over the entire valve position range.
Test Condition Open Vessel At the above referenced test condition, with the recirculation pumps idle, the flow control valve position controller was demonstrated to have a stable response. Proper valve speeds were also demonstrated.
One of the four hydraulic subloops did not satisfy the Level 2 criteria of full stroke speed during the initial test. This problem was subsequently resolved. The position controller response satisfied all applicable acceptance criteria concerning control syster operation. This testing was successfully completed and all other applicable acceptance criteria were satisfied during this test condition. j l
Test Condition 1 l 1
At the above referenced test condition, with the recirculation pumps !
I operating at low speed, the flow control valve position controller was demonstrated to have a stable response and meet the criteria concerning system deadband. A Level 3 exception was encountered during testing as the response time of the A flow control valve was slightly slower than the criteria value. This exception was evaluated by r'r= mnwealth Edison and General Electric control system engineers and this response was determined to be acceptable. This testing was successfully completed and all other acceptance criteria were satisfied during this test condition.
Test Condition 3 At the above referenced test condition, data was taken to calibrate the control system function generators to acheive a linear relation ship between flow demand and flow feedback. Stable control system performance was demonstrated by the performance of position demand steps, flow demand steps, flux demand steps and load demand steps.
With the control system tuned for a stable response, all criteria relating to control system response were not satisfied at all times.
Even though these criteria were not met, the recirculation flow control system is able to acheive a stable control system response
! and provides for an acceptable transi.ent response. Each of the exceptions was reviewed by the cosmoonwealth Edision and General ,
Electric Control System Engineers and each was found to be l acceptable. This testing was successfully completed and all other ;
acceptance criteria were satisfied during this test condition.
DOCUMENT 0812r/00llr
Test Condition 6 At the above referenced test condition, stable control system response was demonstrated by the performance of position demand steps, flow demand steps, flux demand steps and load demand steps an the 100% flow control line at approximately 9% core flow. With the flow control system tuned for a stable response, all criteria ,
relating to control system response were not satisfied at all times, i Even though these criteria were not satisifed, the recirculation flow l control system is able to acheive a stable control systest response and provides for an acceptable transient response. Each of the exceptions was reviewed by r w alth Edison and General Electric control System engineers and each was found to be acceptable. This testing was successfully completed and all other acceptance criteria were satisfied during this test condition.
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DOCUMENT OS12r/00llr
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STARTUP TEST PROCEDURE 30-2 I RECIRCULATION SYSTEM l
! 1. PURPOSE I A. To obtain recirculation system performance data under different operational conditions, such as pump trip, flow coastdown, pump restart, and flow induced vibration.
. B. To verify that no recirculation system cavitation will occur in the operating region of the power-flow map.
- c. To verify that during the trip of one recirculation pump, the feedwater control system can satisfactorily control water level without a resulting turbine trip or scram.
- 2. CRITERIA Level 1 A. The two pump drive flow coastdown transient during the first 3 seconds must be equal to or faster than that specified in the analysis.
Level 2 i A. The water level, APRM and transients of simulated heat flux, pressure, drive and core flow for the one pump trip shall not exceed the predicted values.
B. The reactor water level margin to avoid a high level trip shall be greater than or equal to 3.0 inches during the one pump trip.
C. The simulated heat flur (TPM) margin to avoid a scram shall be greater than or equal to 5.0 percent during the one pump trip.
D. The recirculation system cavitation runback feature shall be adjusted such that a flow runback (transfer of recirc. pump power supplies from 60 Hz to 15 Hz) occurs prior to any observable cavitation in the L Recirculation System.
L
. E. During recirculation pump restart (s) the scram trip avoidance margins must be at least the following:
- 1. For APRM, greater than or equal to ~1.5%.
l
- 2. For simulated heat flux, greater than or equal to 5.0%.
F. If the level I criteria for the two pump trip coastdown transient is met the data shall be analyzed within two weeks to ensure compatibility with the safety analysis.
G. The two pump drive flow coastdown transient during the first 3 seconds must oe equal to or slower than that specified by analysis, s
DOCUMENT 0814r/0011r
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Test condition 2 ,
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y Steady state data was recorded during this test conditiosy. Minor 3 s discrepancies between the recirculation drivg flows and the flow <
control valve positions were noted. Theseproblemswereresolyed
, during Test condition 3.
! , 7 A non-cavitation verification test was satisfactorily completed.
During this test, control rods were inserted to lower core power m ,^
l maintaining a cormtynt core flow until the recirculation pumps , ,
transferred to low speed. The pumps transferred to low speed at approximately 38% power. Cavitation was not observed during this. S test. All applicable test criteria,were satisfied during this testy
+
condition. , ,
s Test condition 3 c1 ;
, i Steady state data was recot.1M during this test condi.tici.. No f' '
exceptions were noted. i
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A single pump trip anditestart was performed. Tha reactor wetetN "
~
.- level, APRM, drive flow',~ cora flow, reactor pressurs,.and simulated j
/
heatsfluxresponsecomparedjavorablywit'stheexpectedresponse.
Athpumptripwasperformed. TheLev911$critert regafdingpump
- coastdown was satisfied but tte Le s l'2 criteria (2.0)was slightly exceeded. Analysis of the .coastdoun response _was performed by the .
y General Electric Transient Response and ECCS Enginsecing Group and ,
conc (udedthatthecoastdounperformancewasaccepteble. s
.. r .-
A reencu ation flow runback was performed by simulating a loss of
', .one feedwater pumpJn ' conjunction with a low reactor water 1cvel signal. An exception was encounterd because the A-flow control val @ locked at 20V position rather than reaching tt9 10% final position. The problem has been determined to be cochanical friction and a work request was initiated to inspect the salve interr.als ,
during an outage. With this exception outstanding, fiatisf W ory operation of this runback was demonstrated by the paOJonastne "
of
! STP-2'JC'2;'Feedwater System - Feedwater PumF Trip.
's Non cavitation verification tests were perlo'emed at this test condition. No exceptions were'.notsd..
> The testing was successfully canpleted during this test condition and all other acceptance criteria were satisfied.
\
_Tsat condition 4 2 . ,
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s I % Steady state data was recorded for this test. condition. No t exceptions were rote 4 ; This test 2ng was successfully c.nrtleted '
and
- all applicable acceptart;;e criteria were satisfied.-
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, Test condition 5 i
.; Steady state data was recorded for this test condition. No exceptions were noted. This testing was successfully completed and all applicable acceptance criteria were satisfied.
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! Test condition 6 i Steady state data was recorded during this test condition. No exceptions were noted.
A single pump trip and restart was performed. The reactor water level, APRM, drive flow, core flow, reactor pressure and simulated
, heat flux compared favorably with the expected response. This test was successfully completed during 1.51s test condition, and all-
- applicable acceptance criteria were satisfied.
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l STARTUP TEST PROCEDURE 31-2 LOSS OF OFFSITE POWER
! 1. PURPOSE t
l A. To determine electrical equipment transient performance during a loss of auxiliary power.
- B. To determine reactor system transient performance during a loss of aux.iliary power.
j 2. ggIIggE Level 1 A. Reactor Protection System actions shall prevent violation of neutron flux and simulated fuel surface heat flux thermal power limitations.
B. All safety systems, such as the Reactor Protection System, the diesel-generators, and HPCS must function properly without manual assistance. HPCS and/or RCIC system action, if necessary, shall keep the reactor water level above the initiation level of the Low Pressure Core Spray, LPCI, and Automatic Depressurization systems.
Level 2 A. Normal cooling systems should be able to maintain adequate suppression pool water temperature, maintain adequate drywell cooling, and prevent actuation of the auto-depressurization system.
- 3. RESULTS Test condition 2 This test, performed during Test condition 2, demonstrated adequate i station electrical eluipment and reactor system transient performance during a sustained leas of offsite power. The heat flux valve and APRM neutron flux level were well within limits. Division 1, 2, and 3 diesel generators functioned properly without manual assistance. Drywell and suppression pool temperatures were well within limits. The plant remained
, isolated from the power grid for 30 minutes. During the test, 2 PCIS valves failed to show full closed. This was later corrected by a work request. All other acceptance criteria were satisfied.
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DOCUMENT 0782r/0010r
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- STARTUP TEST PROCEDURH 33-2 DRWELL PIING VIBRATION
- 1. PURPol5B A. The purpose of this test is to verify that the main steam, reactor recirculation, and feedwater piping have acceptable vibration.
- 2. C3, ITER._M i
Level 1 4 1 i A. The measured amplitude for vibration of the recirculation system during recirculation pump
- trips and subsequent coast down shall not 4 exceed the allowable values.
B. The measured amplitude for vibration of the main steam lines during
- relief valve operation shall not exceed allowable values.
! ' C. The measured amplitude for steady state vibration of the recirculation and main steam systems shall not exceed allowable values.
b D. The measured amplitude for vibration of the main steam lines due to turbine stop valve trip and relief valve operation shall not exceed ,
allowable values.
Level 2 A. The measured amplitude of vibration of the main steam system following relief valve operation and turbine stop valve trip should not exceed the expected values.
B. The measured amplitude of vibration of the main steam and recirculation systems during steady state operation should not exceed the expected values.
C. The measured vibrational stresses induced in the feedwater system following trip of one and both turbine driven feed pumps and dur!.ng steady state operations should not exceed the expe::ted stresses.
- 3. RESULTS Test Condition Heatup At Test Condition Heatup, all accessible RCIC piping greater than two inches in diameter, outside containment, used in the CST to CST and CST to RPV flow modes was inspected for perceivable vibration under steady-state conditions. RCIC instrumentation lines outside containment were inspected. All acceptance criteria were met.
DOCUMENT 0815r/0011r
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Test condition 2 At Test Condition 2, steady-state vibration measurements were made and found to be acceptable. Transient vibration measurements were also made at this test condition. Vibration in the main steam lines due to the generator trip (STP-27-2) was within the acceptance
Vibration induced in the main steam lines due to turbine trip following loss of offsite power (STP-31-2) satisfied the acceptance j criteria.
During the course of testing it was determined t: sat four (4) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test because the measurements from the
- functioning instrumentation indicated values which did not even approach the acceptance criteria limits. This testing was successfully completed at this test condition.
Test condition 3 At Test condition 3, steady-state vfbration measurements were made and found to be acceptable. Transient vibration measurements were also made at this test condition. Vibration in the main steam lines due to relief valve capacity checks (STP-26-2) was found to be acceptable. Vibration induced in the recirculation lines due to recirc pump trip and start (STP-30-2) was within criteria limits.
Vibration induced in the main steam lines due to main turbine trip (STP-27-2) was found to be acceptable.
. During the course of testing it was determined by General Electric that eight (8) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test because the measurements from the functioning instrumentation indicated values which did no! even approach the acceptance criteria limits. This testing was successfully completed at this test condition.
Test condition 6 At Test Condition 6, steady state vibration measurements were rAde and found to be acceptable. Transient vibration measurements were also made at this test condition. Main steam line vibration due to a generator trip.(STP-27-2) was found to be acceptable. Main steam line vibration induced by a full MSIV isolation (STP-25-2) was within criteria limits. Recirculation system vibration induced by RHR shutdown cooling operation (STP-71-2) and by recirculation pump trip and restart (STP-30-2) was within criteria limits. Vibration induced in the feedwater lines due to the feedwater pump trip (STP-23C-2) was found to be acceptable.
During the course of testing it was determined that twelve (12) of the installed sensors were inoperative. The loss of this instrumentation had no impact on the test because the measurements from the functioning instrumentation indicated values which were well l
within acceptance criteria limits.
DOCUMENT 0815r/0011r
s j .' he inspection cf RCIC piping in the Supprs".,aion Pool to RPV flou
- path which was originally scheduled for Test Condition Heatup was performed during Test condition 6 since this flow path was not utilized in STP-14, RCIC System. This inspection was performed under LST-84-ll3 and could not be completed in earlier' test conditions due 1 to the water quality of tne suppression pool. All acceptance
! criteria were inet.
This testing was successfully completed at this test condition.
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1 STARTUP TEST PROCEDUR2 35-2 l
RECIRCULATION SYSTEM FLOW CALIBRATION
- 1. PURPOSE
- A. The purpose of this test is to perform a complete calibration of the installed recirculation system ficw instrumentation.
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- 2. ACCEPTANCE CRITERIA
) Level 1
- 1. None.
Level 2
- 1. Jet pump instrumentation shall be adjusted such that the jet pump total flow recorder will provide a correct core flow indication at rated conditions.
- 2. The APRM/RBM flow bias instrumentation shall be adjusted to function properly at rated conditions.
Test condition 3 The recirculation system flow calibration was successfully performed at Test condition 3. The core flow instrumentation was adjusted to provide accurate core flow indication based on jet pump flows. No adjustments were required on the loop drive flow instrumentation since these flows were already conservatively _ adjusted per normal plant surveillance procedures. During the performance of the testing a higher than nominal flow distribution factor was found for jet pump
- 1. Additional data was taken and an evaluation of the results suggested that testing at higher power levels should proceed and this exception should be addressed again at Test Condition 6. All other applicable acceptance criteria were satisfied.
Test condition 5 The recirculation system flow calibration was successfully performed at Test Condition 6. The core flow indication was shown to provide accurate core flow indication based on jet pump flows.
During the performance of Test Condition 3 testing,an anomaly was found in the flow distribution factors for jet pump #1. Additional data was evaluated and General Electric - San Jose determined that the flow distribution factors were within acceptable levels. All
- other criteria were satisfied during this test condition.
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DOCUMENT 0816r/00llr il
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l STARTUP TEST PROCEDURR 70-2 REACTOR WATER CLEANUP
- 1. PURPOS5
! A. To demonstrate specific aspects of the mechanical operabiliy of the i
Reactor Water Cleanup System with the reactor at rated pressure and temperature.
l 2. CRITERIA Level 2 4
t l A. The temperature at the tube side outlet of the non-regeneratiave heat exchangers shall not exceed 130*F in the blowdown mode and shall not exceed 120*F in the normal mode.
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i B. The pump available NPSH shall be 13 feet or greater for all modes of Reactor Water Cleanup System operation.
C. The cooling water supplied to the non-regenerative heat exchangers shall not exceed the flow and outlet temperature limits of 150*F in <
! the normal mode, 180*F in the blowdown mode, and 762 gpa in either 4
, mode.
D. Recalibrate bottom head flow indicator (R610) against RUCU flow indicator (R609) if the deviation is greater than 10 gps.
E. Pump vibration measured on the casing between_the motor and pump )
shall be less than or equal to 2 mils. 1
- 3. RESULTS Test Condition Heatup l
Reactor Water Cleanup System (RWCU) performance was evaluated for the l normal, blowdown, and hot standby modes of operation. In addition, a comparison of the bottom head flow indicator and the RWCU flow indictor i was conducted. All test criteria were satisfied. See Table 1 for test data.
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DOCUMENT 0801r/0010r i
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STARTUP TEST PROCEDURE 71-2 4
- RESIDUAL HEAT REMOVAL SYSTEM
- 1. ]LURP085: l 4 l
{ A. To demonstrate the ability of the Residual Heat Removal (RHR) System l to remove residual and decay heat from the Nuclear System so that refueling and nuclear servicing may be performed. This will be demonstrated from both the control room and the remote shutdown panel.
] 2. CuTanzA Level 1 A. None Level 2 A. The RH2 System shall be capable of operating in the suppression pool cooling and shutdown cooling modes (with either heat exchanger operating) at the flow rates and heat transfer indicated on the process diagrams.
" Note: During the shutdown cooling mode of operation, it may be necessary to open the bypass Around the RHR Hx in order to prevent exceeding the 100*F in 1 Hr Rx cooldown rate. If such an action is required, only a controllable cooldown rate can be demonstrated. Hx performance cannot be calculated.
B. In the steam condensing mode, for small disturbances, the shellside pressure and level decay ratio must be less than 0.25 throughout each controller's expected operating range. The RHR System shall be capable of operating in the steam condensing mode.
C. The RHR System performance in the shutdown cooling mode shall provide a controlled cooldown during shutdown. The RHR System Heat Exchangers shall meet the heat transfer of the process. diagram for the suppression pool cooling mode.
- 3. RESULTS
, with NRC Region III present on 16 April 1984 and reconfirmed by NRR on 20
' August 1984. The change is scheduled to be documented in the next updata to the LaSalle UFSAR.
Note: STP 71-2 was performed during Test conditions 1, 3, & 6 (only) i DOCUMENT 0791r/0010r
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i Test condition 1 )
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- The Steam Condensing Mode Inlet Pressure and Level Controller Adjustment 1 Test was performed during TC 1 with the following cosaments.
i The RHR A/B Hx response to auto / manual pressure / level steps in the l
- increasing / decreasing direction were verified to be acceptable.
> System response was of reasonable speed, allowing the system to arrive at a stable condition within a reesonable period of time.
I Response to level changes was slow enough to minimize pressure transients.
With the pressure controller in auto, slight non-divergent pressure oscillations were noted after system disturbance stablized with a frequency of approximately .025 Hz and an amplitude of approximately 2.5 psig, which is + 1.25% of the net pressure.
Due to their low frequency and amall amplitude, the oscillations had essentially no effect upon the :.tHR Hx's ability to operate in the steam condensing mode.
During level transients, momentary pressure spikes greater than the 200 psig limit were achieved, with the maximum pressure value being 350 psig.
,4 Since the pressure values > 200 psig were only momentary, they had no effect upon the Hx Design Safety Limits and were well below the 500 psig settings of the shell side pressure relief valves.
The B loop pressure regulator valves movement was not sa:cth (i.e. the valves travel was in small jumps) indicating a binding at the valve stem .
Upon resolution, no reperformance of STP ~11-2 was required. A simple a valve cycling and verification of a smooth stroke was satisfactcry.
The A Hx maximum capability was found to be 94.22 x 10 6 BTU /Hr, 63.6%
of rated; and the B Hx maximum capability was found to be 69.37 x 106 BTU /Hr, 46.8% of rated.
All pressure and level decay ratios were satisfactorily below the 0.25 level 2 criteria. No other level criteria were applicable in Test condition 1.
Test condition 3 The RHR Suppression Pool Cooling Mode Startup Test was performed (in conjunction with STP 26-2 (Relief Valve testing) during TC 3 with the following comments.
All recorded flow rates were within reasonable agreement with the process diagram values.
1-DOCUMENT 0791r/0010r .
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The recorded temperatures wera 1ses than the procasa diagram valuss, which was due to:
a) The lake temperature being below 100*F and b) The suppression pool Tech spec temperature limits of < 105'F (under test conditions) and < 100*F (under normal conditions)
The final calculated values for the transfer coeffLcients (U*A), to be used in T.C. 6, were as follows:
l Loop A: U*A = 2.98 x 106 BTU /HR*F Loop B: U*A = 2.33 x 106 BTU /IIR*F Since the suppression pool cooling process diagras U*A value is 2.3 x 106 BTU /HR*F, the aforementioned measured (actual)
. U*A values satisfactorily meet the relation.
(U*A) actual 1 1 (U*A) design In conclusion, the performance of STP 71-2 in TC 3 satisfactorily verified the operability of the RHR system in t.he suppression pool cooling mode.
All applicable level criteria were satisfied.
Test condition 6 The RHR shutdown cooling mode startup test was perforned during TC 6 with the following consents.
All recorde<1 flow rates were within reasonable agreement with the process diagram values.
A vessel cooldown via the 'A' Loop was performed with the Hx bypass valve closed, whereby Q shell and Q tube values were obtained which were in reasonable agreement with each other.
However, since the Hx shell side outlet and inlet temperatures were approximately 70*F below th's process diagram values, the measured Log Mean Temperature Difference (LMTD) < the Process Diagram Valve LMTD.
Since the intent of the test procedure was to verify the system's ability to re mve heat and provide a controlled vessel cooldown, the less than expected LMfD had no effect upon the performance of STP 71-2 sit.ce the measured Qs and Qt values (with the Hx bypass valve closed) were acceptable.
For all other cases, including from the remote stlutdown panel, the Hx bypass valve was open and reliable heat transfer parameters could not be calculated, since the shell side outlet temperature element and flow sensing element were dowr. stream of the system's tie in with the bypass loop.
However, by throttling the Ex discharge valve (s), various heat removal rate 1 and vessel cooldown rates could be obtained.
All cpplicable level criteria were satisfied.
DOCUMENT 0791r/0010r KT
STARTUP TEST PROCEDURE 74-2 OFF-GAS SYSTEM a
- 1. PURPOSE The purpose of this test is as follows:
A. To verify the proper operation of the Off-Gas System over its expected operating parameters.
B. To determine the performance of the activated carbon adsorber.
- 2. CRITERIA Level 1 A. The release of radioactive gaseous and particulate effluents must not exceed the limits specified in the Technical Specifications.
B. There shall be no loss of flow of dilution steam to the non-condensing stage when the steam jet air ejectors are pumping.
Level 2 A. The system flow, pressure temperature, and relative humidity shall comply with the design specifications.
B. The catalytic recombiner, the hydrogen analyzer, the activated carbon beds, and the filters shall be operating properly during operation, i.e., there shall be no gross malfunction of these components.
Level 3 None
- 3. RESULTS Test Condition Heat Up, < 250 psig The release of radioactive gaseous and particulate effluents was within f the applicable limits of the Technical Specifications. ,
i The initial rate of dilution steam flow was slightly less than the Level 1
! criteria however at the time the steam pressure to the second stage air ejector was less than rated. The dilution steam flow was found within the Level I criteria in all other test conditions and was therefore evaluated as satisfactory.
There were no gross malfunctions of any components and the parameters were within the applicable test criteria with the following deviations that were corrected by system adjustments or repair.
DOCUMENT 0792r/0010r f6
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., 1. Gas Roh rtsr outict Dew point High.
Gas heater outlet Temperature Low (as system flow rate decreased 2.
this criteria was satisfied.)
- 3. Hydrogen Ant.lyzers Inoperable.
The adsorber vault temperature was lower than design specification due to inoperable ventilation heaters and less than optimum sensor location.
Test Condition Heat Up at rated pressure There were no gross malfunctions of any components and the parameters were within the applicable test criteria with the following deviations that were corrected by system adjustments or repair.
- 1. Hydrogen Analyzers Incperable.
- 2. Steam Jet Air Ejector /Preheater Inlet Pressure Indicator Inoperable.
- 3. Gas reheater outlet dew point indicator inoperable.
The following parameters were not within the design specifications but were evaluated by the station and General Electric as acceptable' based on the overall system performance and release rates being satisfactory.
- 1. Adsorber Vault Temperature (see T/C H/U < 250 psig)
- 2. Adsorber Vessel Temperatures (slightly less than or greater than design but greater than gas dew point)
- 3. Gas Flow rate high (flow rate was attributed to condenser in leakage caused by low load and not all boundaries satisfactorily sealed such as Turbine Driven Reactor Feed pumps and Feedwater Heaters and Extraction steam Lines. The flow rate was found acceptable during subsequent test conditions).
Test Condition 1 The radionuclide residence times for Xe 135 and Xe 133 in the A charccal train were inconsistant and during a unit shutdown the charcoal beds were inspected, whereupon water was discovered in the first bed of both the A &
B train.
Noting an increase in the 'A' profilter AP, a D.O.P. test was performed, whereupon water damage was discovered in the 'A' prefilter. The B profilter was not damaged.
In addition, the middle 2 beds of each train were tested for moisture and found to have between 16.*1 and 21.1% water by weight (an expected value is approximately 4 to 5%).
The source of this water was suspected to be from startup problems occurring on 4/19/84 through 4/20/84 in restoring condenser vacuum at a Rx
- pressure of 250 psig.
Accordingly, the O.G. System was placed on continuous purge, a work request was generated to repair the 'A' profilter, the 'B' profilter was placed in service, and operating personnel were notified to check all off gas system drains for moisture and to ensure that all loop seals were filled with their fill sources isolated.
DOCUMENT 0~192r/0010r M
l Upon complction cf the aferementioned, tha unit was cgain started up and !
new radionuclide samples were taken, with improved results.
Since the system's dew point was low, the adsorbers moisture content :
decreased with prolonged operability, as noted in the improved TC 3 and TC
, 6 residence times. .
I Based on the improved results of the 2nd TC 1 sample and the satisfactory 00CM values (u Tech specs); the TC 1 radionuclide results were considered acceptable.
The aforementioned deficiency was signed off per the satisfactory TC 3 results.
The 'A' profilter was replaced and verified operable per the satisfactory AP reading obtained in TC 3.
The following parameters were not within the design specifications but were evaluated by the station and General Electric as acceptable based on the overall system performance and release rates being satisfactory
- 1. Adsorber Vault Temperatures (see T/C H/U < 250 psig)
- 2. Adsorber vessel Temperatures (see T/C H/U) i Test conditions 3 and 6 The following parameters were not within the design specifications.
- 1. Adsorber vault Temperature Low (See T/C H/U)
- 2. Adsorber Vessel Temperatures (slightly low but average within specification and above dew point) _
- 3. Off Gas Condenser Outlet Temperature High Upon review of previous STP 74-2 data, the high condenser outlet temperature appears to be unique only to the B Loop.
The temperature indicators were verified to be in calibration via their satisfactory comparison when the system was shutdown and stabilized.
Via the 2N62-N008B thermister output, the 'B' loop off gas condenser outlet temperature waz later found to be 185.6*F (at 89.76% Reactor Power); 25.6*F above the design specification of 160'F.
Additional data associated with the Unit 2 B off gas condenser were taken during TC 6 whereupon H2 concentrations, moisture separator outlet temperatures and adsorber vessel temperatures were found to be below their uaximum design limits. Therefore the station and General Electric evaluated the off gas condenser performance acceptable.
- 4. Reheater outlet Dew Point High Dewpoint readings were inconclusive during TC 3.
l Determination of a relative humidity via the moisture separator and gas reheater outlet temperatures resulted in a TC 3 value of 26.8%, well within the 22 to 38% expected value range. A similar calculation during TC 6 resulted in a satisfactory value of 27.3%.
DOCUMENT 0792r/0010r i
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i A TC 6 R31ctiv3 Humidity wa3 found to be 5~1.6% with a d swpoint of 59.2*F.
i j All of the aforementioned TC 3 and TC 6 RH/ dewpoint ve. Lues were below the j adsorber temperatures, thereby preventing any water cordensation in the l charcoal.
Furthermore, review of the TC 3 & TC 6 data does not iridicate any
' difficulties with high humidity, such as high drain tarik levels or filter damage. Residence times, ODOt calculations, and post treatment readings
' were also satisfactory, along with acceptable charcoal bed AP values.
As similarly stated in previous test conditions, all of the measured TC 6 Level 2 values in question were determined acceptable by the Station and 5
General Electric based on:
- 1. satisfactory redundant measurements / indications being obtained or
- 2. The parameter in question improving with continueil system operation at higher power levels or
- 3. The measurement of all parameters associated with the criteria in question being within their design limits or
- 4. The acceptable overall performance of the off gas. system in TC 6 with satisfactory radionuclide residence times and 00CM values u T. S.
Level 1 The release of radioactive gaseous and particulate effiluents was demonstrated to be within the limits specified in the Technical specifications during Test condition 6.
There was no loss of dilution steam to the noncondens!.ng stage when the SJAR's are pumping during Test Condition 6. ,
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DOCUMENT 0*192r/0010r N ,
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