ML20069E711
| ML20069E711 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/11/1983 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20069E709 | List: |
| References | |
| NUDOCS 8303220166 | |
| Download: ML20069E711 (74) | |
Text
s 4
STARTUP TEST PROCEDURE 1 CHEMICAL AND RADl0 CHEMICAL MEASUREMENTS 1.
PURPOSE A.
The principal objectives of the Chemical and Radiochemical Tests are 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 required to demonstrate that the chemistry of all parts of the entire reactor system meet specifications and process requirements.
B.
Specific objectives of the test program include Indirect observations of fuel clad integrity, evaluations of demineralizer operations by direct and indirect methods, measurement of filter performance, confirmation of condenser integrity, measurement and calibration of the off gas system, and calibration of certain process instrumentation.
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 I
Level 1.
A.
Water quality must be known at all time 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 l
and Fuel Warranty must be maintained within the limits specified.
D.
Radiation Monitoring Instrumentation must be responsive to radionuclide sources and/or effluents containing radionuclides.
3.
RESULTS Open Vessel Chemical process systems were tested; sample instrumentation and laboratory equipment were calibrated during preop testing.
Prior to fuel load and after fuel load (before heatup), chemistry of the reactor water, the condensate polisher system, and the feedwater system were tested. The results demonstrated acceptable water quality.
8303220166 830311 PDR ADOCK 05000 P
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4 Test Conditions Heatup and 1 Analysis of radiolytic gas in steam and chemical / radiochemical tests of reactor water, condensate demineralizer inlet and ef fluent, feedwater, of fgas pre-treatment and plant vent were conducted.
Measurements of stored water (clean demineralized water storage tank, cycled condensate storage tank, suppression pool, condenser hotwell, and heater drain tank) quality and condensate /feedwater systems filterable iron concentrations were taken.
The results were within limits except for those parameters which are expected to be out of GE specifications during initial startup.
These deficient conditions are being addressed, and discussions with GE are taking place concerning the validity of their criteria at this time.
No tech spec limits were exceeded at any time, i
i 8
s STARTUP TEST PROCEDURE 2 RADIATION HEASUREMENTS 1.
PURPOSE The purposes of this test are:
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 occupancy 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.
Results This test was performed at test conditions open vessel, heatup, I and 2.
The results of the radiation surveys showed that all radiation levels were well within all test criteria.
i k
STARTUP TEST PROCEDURE 3 i
FUEL LOADING 1.
PURPOSE A.
The purpose of this test is to load fuel safely and efficiently to the full core size.
2.
CRITERIA A.
Level 1.
1.
The partially loaded core must be subcritical by at least 0.38% 4 K/K with the analytically strongest rod fully withdrawn.
D.
Level 2.
1.
The core must be loaded in the analyzed core loading pattern as indicated on Attachment A.
l 3
RESULTS This test was performed during Test Condition Open Vessel.
The partial core shutdown margin demonstrations were completed as follows:
i 16 bundles loaded : 4-18-82 64 bundles loaded : 4-20-82 144 bundles loaded : 4-21-82 At each of the above core configurations the shutdown margin was
(
demonstrated (i.e., the core remained subcritical with the analytically l
highest worth rod withdrawn). After the core had been loaded,the loading pattern (Attachment A) was verified to be correct by video taped visual inspection.
All applicable criteria were met.
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STARTUP TEST PROCEDURE 4 FULL CORE SHUfDOWN MARGIN 1.
PURPOSE A.
The purpose of this test is to demonstrate that the reactor will be subcritical throughout the first operating cycle with any single control rod fully withdrawn.
2.
CRITERIA A.
Level 1.
1.
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 0.26% delta K/K.
B.
Level 2.
1.
Criticality should occur within + 1.0% delta K/K of the predicted critical.
3 Results This test was performed during Test Condition Open Vessel.
The shutdown margin was demonstrated to be greater than 0.38% delta K/K plus an additional margin for exposure (0.26% A K/K). The actual value for shutdown margin was determined to be 2.89% A K/K.
The actual initial critical was demonstrated to ba within 0.193% o K/K of the predicted critical. The Shutdown Margin Calculation and Critical Eigenvalue Comparison Calculation have been included in Attachment A.
All applicable criteria were met during the performance of this test.
i
h ATTACHMENT A j
Shutdown Margin Calculation Kanal = KGl + KG2 + KG3 + KMT AK KGl = 0.048 (worth of all group i rods withdrawn).
g AK i
KG2 = 0.018
-(worth of all group 2 rods withdrawn).
g KG3 = 0.0023 K (worth of all group 3 rods @ 04).
+ 0.0049 o K (worth of 8 group 3 rods @ 08).
g
+ 0.00015 A K (worth of I group 3 rod 9 06).
K Total KG3 = 0.00735 g
KMT = (-5.5 x 10-5) (TM-68)
TM
= 125 F KMT = -0.003135 AK (worth of reactivity to compensate K
for neg. temp. coeff.).
Kanal = 0.048 + 0.018 + 0.00735 - 0.003135 AK
= 0.070215 g
_SDM = Kanal -KP -KSR KP = 0.000488 (reactivity-correction for T = 153 seconds)
KSR = 0.0408 o K (worth of strongest rod) g AK SDM=0.070215(-0.000488y-0.0408 g
oK SDM = 0.0289
= 2.89% oK g
g a
J CRITICAL EIGENVALUE COMPAR160N CALCULATIONS J
DK =
K12 + K3 + KMT - KP - 1
- (100)-
4 K12 = 0 9982 (Keff w!th groups 1 & 2 withdrawn).
K K3 = 0.0023 A K (worth of all group 3 rods @ 04).
g
+ 0.0049 o K (worth of 8 group 3 rods @ 08).
g
+ 0.00015 O K (worth of I group 3 rod @ 06).
g TotalK3-0.00735h KMT = (-5.5 x 10-5) (TM-68)
TM = 125 F V.MT = -0.003135 (worth of reactivity to compensate for g
neg. temp. coef f.).
KP = 0.000438 AK (reactivity correction for T = 153 seconds) g DK = f 0.9982 + 0.00735 + (-0.003135) - 0.000488 - I f * (100)
DK = 0.193%
4
STARTUP TEST PROCEDURE 5 CONTROL ROD DRIVE SYSTEM 1.
PURPOSE A.
The purpose of this test is as follows:
1.
To demonstrate that the Control Rod 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 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 gpm as the reactor pressure changes from a shutdown condition to the normal operating pressure.
2.
CRITERIA Level 1.
A.
Each CRD must 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, i
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 valve solenoids are de-energized).
l Position Inserted From Fully Scram Time l
Withdrawn (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49
C.
The mean scram time of the three fastest CRDs in a two by two array must not exceed the following times:
(Scram time is measured from the time the pilot scram valve solenoids are de-energized).
Position Inserted From Fully Scram Time Withdrawn (Seconds) 45 0.45 i
39 0.92 1
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 solenoids as time zero, shall not exceed 7.0 seconds.
Level 2.
A.
Each CRD must have a normal insertion or withdrawal speed of 3.0 + 0.6 inches per second indicated by a full 12-foot stroke Tn 40 to 60 seconds.
B.
With respect to the CRD Friction tests, if the differential pressure variation exceeds 15 psid for a continuous drive in, 1
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 j
to 30 seconds.
B.
The CRD system flow should not change by more than + 3.0 gpm as reactor pressure varies from zero to rated pressure.
C.
The decay ratio of any oscillatory controlled variable must be less than or equal to 0.25 for any flow set point changes or less than or equal to 0.50 for system disturbances caused by the CRD's being strokes.
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3.
RESULTS Test Condition Open Vessel The following CRD system testing was performed in test condition open vessel;
- vessel insert and withdraw timing for all drives.
- indication verification for all drives.
- coupling checks for all drives.
- friction testing for all drives.
- single rod scrams with normal accumulator pressure for all drives.
- single rod scrams (eight drives - four sequence A and four sequence B) with low accumulator pressure.
Twelve (12) drives had insert and/or withdraw times which did not meet the level 1 (withdraw time 2 40 seconds) or level 2 (insert and withdraw 40 to 60 seconds) acceptance criteria. The appropriate directional flow control valves were adjusted to bring the times within the test criteria. Refer to Table 1 for information on the drives and their times.
All drives were verified to be properly coupled.
CRD 10-11 was the only drive which did not meet the continuous friction ($ 15 psid variation) and settling friction (230 psid total, 6 10 psid variation) acceptance criteria (level 2).
Following an appropriate engineering evaluation, G.E. found the drive acceptable.
All of the CRD scram times were within the level 1 acceptance criteria (individual 90% scram times, mean scram times of all the drives, two-by-two array scram times).
See Table 2.
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 pressurcs of 0 psig and rated.
- system flow response as reactor pressure is increased from 0 psig to rated.
- normal insert and withdraw timing for four drives (sequence A).
- friction testing for four drives (sequence A).
- single rod scrams (eight drives) at intermediate reactor pressures (600 and 800 psig).
- single rod scrams with normal accumulator pressure for all drives.
- single rod scrams (eight drives) with zero accumulator pres.sure.
h System Flow controller tuning, performed at reactor pressures of 0 psig and rated, met the level 3 acceptance criteria (decay ratio s.25 for flow step changes and 6.50 for CRD stroking). With the final controller settings, 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). GE 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.
The CRD system flow did not change by more than 3 gpm as reactor pressure was increased from zero to rated (level 3 acceptance criteria).
Drive timing (normal insert and withdraw), friction testing, and scram testing met the applicable test acceptance criteria.
Sea Table 3 Test Condition 1 During the Remote Shutdown Panel scram (STP-28), the procedure called for the acquisition of the following data:
- closing time of the "on-line" flow control valve.
The closure time of the "on-line" flow cor.crol valve was not obtained because the CRD controller input signal (system flow),
which was being monitored during the scram, could not be used to determine the minimum valve position due to the fact that system flow pegs upscale during a scram. The loss of this data (closure time) had no impact on the test since the CRD pump does not runout during a scram.
G.E. concurred with this evaluation.
Scram times for 27 drives were obtained during the scram, and their 90% times were within the level 1 acceptance criteria ( 6 7.0 seconds).
See Table 4.
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\\ tNITIA1. TIME'(SECONDS)
ADJUST TIME (SECCNDS)
~ INSERT WITHDRAW INSERT WITHDRAW s
64.5 46.0 52.7 18-27
,39.9 '.,
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. 22-07 39.4 38-07 37.5 61.0 c,
49.0 50.4
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49.2 N/A 46-19 60.9 SAT.
48.0 N/A
- s s
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45.6 N/A 34-07 37.4-SAT.
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A.
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5 Fas tes t Wi tlidrhw sTime: CRD 22-15, 40.1 seconds.
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J TABLE 2 OPEN' VESSEL SCRAM TIMES j
l Ave. Scram Times Slowest Ave. Scram Time Position of All CRDs in a Two-By-Two Array 45 0.26 sec.
0.27 sec.
39 0.44 sec.
0.46 sec.
J 25 0.90 sec.
0 95 sec.
05 1.63 sec.
1.71 sec.
Slowest 90% scram time; CRD 50-51, 1.97 seconds.
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t TABLE 3 HEATUP SCRAM TIMES Ave. Scram Times Slowest Ave. Scram Time Posi tion of All CRDs in a Two-By-Two Array 45 0.32 0.34 39 0.62 0.66 25 1.34 1.47 05 2.39 2.62 Slowest 90% scram time; CRD 14-43, 3.0 seconds.
HEATUP DRIVE TIMING FASTEST INSERT TIME; CRD 14-43, 48.0 seconds SLOWEST INSERT TIME; CRD 30-19, 52.6 seconds FASTEST WITHDRAW TIME; CRD 30-43, 4'3.1 seconds SLOWEST WITHDRAW TIME; CRD 34-31, 51.0 seconds I
I i
F TABLE 4 TEST CONDITION 1 SCRAM TIMES Slowest 90% scram time; CRD 10-11, 2.85 seconds.
9
STARTUP TEST PROCEDURE 6 SRM PERFORMANCE AND CONTROL R0D SEQUENCE 1.
PURPOSE A.
The purpose of this test is to demonstrate that the oper-ational 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.
2.
CRITERIA A.
There must be a neutron signal count-to-noise ratio of at least 2:1 on the required operable SRM's or Fuel Loading Chambers.
B.
The.e must be a minimum count rate of 3 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.
3.
RESULTS This test was performed at Test Conditions Open Vessel, Heatup and 1.
During fuel loading and prior to the initial ciritical, all oper-able Fuel Loading Chambers and Source Range Monitors, were demon-strated to have a neutron signal-to-noise greater than 2:1 and a count rate greater than 3 CPS (data included on Table 1).
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 power increase. Proper SRM-IRM overlap was demonstrated during the performance of this test.
All applicable criteria were met during the performance of this test.
0 TABLE 1 SOURCE RANGE MONITOR MINIMUM COUNT RATE AND SIGNAL-TO-NOISE RATIO 1.
Fuel Loading Chamber prior to Fuel Loading (0.V.)
S/N Date/ Time Inserted Withdrawn A
9.5 0.5 18: 1 4-18-82/1121 B
6.0 0.3 19: 1 4-18-82/1121 C
10.0 0.5 19: 1 4-18-82/1121 D
7.5 0.5 14: 1 4-18-82/1121 2.
Source Range Monitor during Fuel Loading (0.V.)
~
S/N Date/ Time inserted Withdrawn A
6.5 0.1 64: 1 4-28-82/0700 B
6.5
< 0.1 64: 1 4-26-82/2138 C
8.0 0.1 79: 1 4-24-82/0740 D
3.0 0.1 29: 1 4-23-82/0910 l
l
~-
TABLE I (Cont.)
SOURCE RANGE MONITOR MINIMUM COUNT RATE AND SIGNAL-TO-NOISE RATIO 3
Source Range Monitor prior to startup (0.V.)
S/N Date/ Time incerted Wi thd rawn A
7.0 0.1 69: 1 6-21-82/0226 B
7.5 0.3 24: 1 6-21-82/0226 C
6.0 0.1 59:1 6-21-82/0226
_}
C 6.0 0.1 59: 1 6-21-82/0226
STARTUP TEST PROCEDURE 9 WATER LEVEL MEASUREMENT 1.
PURPOSE A.
The purpose of this test is as follows:
1.
To check the calibration of the various narrow range and wide range indicators.
2.
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.
3 To collect plant data which can be used to investigate the effects of core flow velocity, carry under, and subcooling cn indicated wide range level.
2.
CR:TERIA A.
Test criteria prfor to revision 6 dated 12-2-82 1.
Lerel 2 A.
The narrow range indicators that provide feadwater flow and water level control functions (1C34-R606A-C, IB21-N024A-D, and 1821-NO38A-B) should have readings that agree within j 1.5 inches of the average reading.
B.
The wide range level system indicators (IB21-N026A-D, 1821-NO31A-D, I B21 -N036A-D, 1821 -NO37A-D, 1821 -
C604, IC61-R010, IB21-R623A-B should agree within +
6 inches of the average reading.
8.
Test criteria after revision 6 dated 12-2-82 1.
Level 2 j
A.
The narrow range level indicator readings on the instruments used for feedwater level control (1C34-R606A-C) should agree within j-1.5 inches of their L
average reading.
B.
The narrow range level indicator readings on the instruments not used for feedwater level control (IB21-N024A-D, IB21-N038A-B, 1821-N100A-8, and 1821-N101A-B) should agree within + 3.0 inches of their average reading.
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C.
The wide range level system indicators (IB21-N026A-D, IB21-N031A-D, 1B21-N036A-D, 1B21-N037A-D, 1B21-R604, IC61-R010, 1821-R623A-B should agree within +
6 inches of the average reading.
~
These criteria changes have been discussed with and approved by the NRC.
They will be incorporated in a future FSAR ammendment.
3.
RESULTS Test Condition Heatup The cold water calibration on the reference leg temperature was successfully verified to be within tolerance (50 F) of the assumed reference leg temperature of 135 F.
The measured reference leg temperature for test condition heatup was 143.5 F and 142.5 F.
During the performance of this test, two of the narrow range in-struments 1821-N024C and 1C34-R606B were found to be outside the level 2 criteria (+ 1.5 inches of their average reading). The actual data for the above narrow range instiuments are included in Table 1.
Instruments IB21-H024C and IC34-R6068 wure recallbrated j
and this test repeated at a later date.
At the recommenoition or G.E. feur additional narrow range level instruments (1821-N100A/B nd 1821-N101A/8) were included in the procedure and data analysis.
During the retest of this procedure the narrow range instruments IB21-N024C and IC34-R606B verified to be within tne + 1.5 irch criteria. However, it was determined that four cthe7 narrow in-struments were indicating outside of the + 1.5 inch criteria.
The actual data for the above narrow range instruments (1821-NO38A, IC34-R606C, 1B21-N100A, and IB21-N101A) are incitded in Table 2.
After consulting with G.E.
It was determined that the criteria had not been met but, it would be safe to proceed to test condition 1.
The test has been transmitted to G.E. San Jose for evaluation of the + 1.5 inch criteria.
Test Condition 1 During the performance of this test at test condition 1, three narrow range instruments (1821-N038A, 1821-N101A, and IC34-R606B) failed the + 1.5 inch criteria (actual data included on Table 3).
The above instruments were not recalibrated and this test repeated (per the procedure) due to the following:
1.
It was shown during test condition heatup, that a couple of the instruments were indicating within their tolerance but still failed the test criteria.
2.
The data was transmitted to G.E. San Jose for further evaluation to determine the cause for the recurring level 2 failure.
After consulting G.E., it was determined that the above criteria violation would not limit proceeding to test condition 2.
Test Condition 2 During the performance of this test at test condition 2, all the level indicators were reading within the "new" (revision 6) criteria.
The test results from test conditions heatup and 1, were also re-evaluated, and found to be within the "new" (revision 6) criteria.
This test was successfully completed for test condition heatup, 1, and 2 with all applicable criteria met.
l 1
TABLE 1 NARROW RANGE LEVEL INSTRUMENTS (TEST CONDITION HEATUP)
INSTRUMENT VALUE AVERAGE ERROR NUMBER VALUE 1821-N024C 31.8 33.62
-3.38 1C34-R606B 37.0 33.62 1.82 TABLE 2 NARROW RANJE LEVEL INSTRUMENTS (TEST CO?iDir!ON WEATUP)
INSTRUMENT VALUE AVERAGE EREOR NUMBER VALUE 1821-N038A 32.1 34.98
- 2.88 IC34-P606C 36.5 34.98 1.52 l
1821-N100A 37.5 34.98 2.52 IB21-N101A 33.2 34.98
- 1.78 TABLE 3 NARROW RANGE LEVEL INSTRUMENTS (TEST CONDITION 1)
INSTRUMENT VALUE AVERAGE ERROR NUMBER VALUE IB21-NO38A 32.5 34.94
- 2.44 1C34-R606B 36.5 34.94 1.56 1821-N101A 33.4 34.94
- 1.54 4
i
STARTUP TEST PROCEDURE 10 IRM PERFORMANCE 1.
PURPOSE 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 cny 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% cf full scale.
3.
RESULTS This test was performed at test conditions open vessel, heatup and 1.
Adequate SRM/lRM 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 IP.M was adjusted to obtain proper range 6/7 correlation. At test condition 1, each IRM demonstrated adequate overlap with the APRM's.
At each test condition, all applicable test criteria were satisfied.
/
l L
STARTUP TEST PROCEDURE 11 LPRM CAllBRATION 1.
PURPOSE A.
To verify proper response of the Local Power Range Monitoring (LPRM) System to local changes in the reactor power level.
B.
To calibrate the LPRM system.
2.
CRITERIA Level 2 A.
Each LPRM reading will be within + 10% of its calculated value (as determined by a process computer or offline calculation, from Traversing incore Probe Power Distribution Data).
Level ?
A.
At least three LPEM detectors in each LFRM string must respond to a local change in neutron flux to assure the proper connection of tne LPRM detectors to their cables. I f less thar, three detectors in a string are 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 oetectors and readout equipment was completed for 10 of the 172 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 i
l The remaining LPRM response to neutron flux testing was completed.
1 All acceptance criteria were met, as at least three detectors in an LPRM string indicated a response to the control rod pulls.
Three attempts were made to satisfy the calibration criteria of +
10%. After each recalibration, many of the LPRM's were still l
outside the criteria.
In addition, 8 LPRM's could not be calibrated because they were still reading downscale at the low power level of this test condition.
Following the final calibration at this test condition 58 LPRM's were outside the + 10% criteria.
l l
p An analysis of this problem determined that an accurate calibration could not be performed at such a low power level ~ (15%).
After consulations with G.E.,
i t was determined that this criteria failure should not stop the plant from proceeding to higher power a
test conditions where a more accurate calibration would be possible.
Although Startup Test Procedure 11 will not be 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.
During this testing at 20% power, only 2 low reading A level LPRM'S in non-critical locations failed to meet the + 10% criteria (14% maximum deviation) and 3 additional LPRM's could 'not be calibrated due to low readings.
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STARTUP TEST PROCEDURE 12 APRM CAllBRATION 1.
PURPOSE A.
The purpose of this test is to calibrate the Average Power Range Monitor (APRM) system.
2.
CRITERIA A.
Level 1 1.
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.
2.
The APRH scram and rod block setpoints shall be set no higher than the limits specified in the Technical Specifications and the fuel warranty docunent.
3.
In the STARTUP mode, all APRM channels must produce a scram at less than or equal to 15% cf rated core thermal power.
B.
Level 2 1.
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 + 7% of rated core thermal power.
3 RESULTS During test condition Heatup the reactor was put on relatively constant heatup rate for approximately 75 minutes.
Due to the limited range of the APRM amplifier gains, the APRM gains were set as low as possible. The as left settings provided between 30% and 70% conservatism. Although the +7% level 2 criteria could not be met, 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 4
less than or equal to 12% and 15% thermai power respectively.
Testing at Test Conditions 1 and 2 was successfully completed, and all applicable criteria were met.
The APRM's were adjusted to the results of the heat balance (manual heat balance during test condition 1 and OD-3 heat balance during test condition 2).
The Upscale Alarm, Thermal Trip, and the Neutron Trip settings were demonstrated to be less than or equal to their respective limits.
This test was successfully completed for Test Conditions Heatup, 1, and 2 with all applicable criteria met (except as explained above).
Further testing of STP-12, APRM Calibration will be performed at subsequent test conditions.
i
STARTUP TEST PROCEDURE 13 PROCESS COMPUTER 1.
PURPOSE A.
The purpose of this test is to verify the performance of the process computer under' plant operating conditions.
2.
CRITERIA A.
Level 2 1.
Programs OD-1, D1 and OD-6 will be considered operational when:
A.
The MCPR calculated by BUCLE and the process con.puter either:
1.
Are in the same fuel assembly and do not differ in value by more than 2%, or 2.
For the case in which the MCPR calculated by the process computer is in a different assembly than thet calculated by BUCLE, for each assembly, the MCPR calculated by the two methods shall agree within 2%.
B.
The maximum LHGR calculated by BUCLE and the process computer either:
1.
Are in the same fuel assembly and do not differ in value by more than 2%, or 2.
For the case in which the maximum LHGR calculated by the process computer is in different assembly than that calculated by BUCLE, for each assembly, i
l the maximum LHGR's calculated by the two methods I
shall agree within 2%.
C.
The MAPLHGR calculated by BUCLE and the process computer either:
I 1.
Are in the same fuel assembly and do not differ in value by more than 2%, or i
2.
For the case in which the MAPLHGR calculated by the process computer is in a different assembly i
l than that calculated by BUCLE, for each assembly, the MAPLHGR's calcuiated by the two methods l
shall agree within 2%.
i i
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4 D.
-The LPRM calibration factor calculated by BUCLE and the process computer agree to within two percent (2%).
E.
The remaining programs will be considered operational upon successful completion of the static and dynamic testing.
3.
RESULTS The Static System Test Case using plant simulation software was success-fully completed and all applicable criteria were met at Test Condition
~
Open -Vesse l.
Testing at Test Conditions Heatup and consisting of TIP system hot alignment and Process Computer /TIP system Interface testing was success-fully completed and all applicable criteria sere met.
The Dynamic System Test Case and all other testing at Test Condition 2 was successfully completed. The thermal limits (Maximum Linear Heat Generation Rate, Minimum Critical Power Ratio, MaxiNJin Average Pianar Heat Generation Rate) calculated by the process computer were demonstrated to have the same bundle laccticos and nave valuer within 2% of the BUCLE resuits (table 1 has thermal limits for 21% and 46% power).
It was also demonstrated that the LPAM calibratier factorr calculated by the process computer were within 1% of the BUCLE values.
(All actuel values v:cre 0%
1.e. exact agreement.)
This test was successfully completed for Test Conditions Open-Vessel, Heatup,1, and 2 with all applicable criteria met.
+ -. -
TABLE 1 Thermal Limit Data A.
Maximum Linear Heat Generation Rate (MLHGR)
POWER LOCATION LOCATION MLHGR MLHGR
(%)
FROM BUCLE FROM P-1 FROM BUCLE FROM P-1 ERROR 21%
27-20-5 27-20-5 2.64 2.64 0%
46%
37-14-4 37-14-4 6.02 6.03 0.22%
B.
Mimimum Critical Power Ratio (MCPR)
POWER LOCATION LOCATION MCPR MCPR
(%)
FROM BUCLE FROM P-1 FROM SUCLE FP.0M P-1 ERROR 21%
33-42 33-42 4.378 4.375 0.073%
fo,g4g 46%
53-38 51-38 2.513 2.512 C.
Maximum Average Planar Heat Generation Rate (MAPLHGR)
POWER LOCATION LOCATION MAPLHGR MAPLHGR
(%)
FROM BUCLE FROM P-1 FROM BUCLE FROM P-1 ERROR 21%
09-18-4 09-18-4 2.29 2.29 0%
46%
25-12-4 25-12-4 5.17 5.17 0%
STARTUP TEST PROCEDURE 14 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 any reactor pressure between 150 psig and rated.
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 differertial 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 tne 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.
- 3.. RESULTS Test Condition Heatup RCIC Manual (controlled) and Hot Quick Starts to the Condensate Storage Tank (CST) and the Reactor Pressure Vessel (RPV) were made to demonstrate proper controller operation and system performance.
Flow step changes while injecting high and low RCIC flow to the CST and RPV were performed during system tuning. A manual RCIC Hot Quick Start to the CST from the Remote Shutdown Panel (RSP) was performed follcwed by flow step changes demonstrating proper
RCIC RSP controller operation. RCIC testing at test condition Heatup has been satisfactorily completed, with only the following deficiencies remaining open at the completion of test condition Heatup.
1.
600 gpm flow was not achieved at 150 psig reactor pressure due to high back pressure in the test piping. The capability to achieve 600 gpm flow at 150 psig reactor pressure will be demonstrated during vessel injection testing prior to exceeding 25% reactor power.
2.
RCIC tripped on overspeed during a cold quick start. The RCIC cold quick start capability will be demonstrated by further testing prior to exceeding 25% reactor power.
Test Condition 1 Turbine overspeed trips were encountered during quick start testing.
These problems were eliminated by correcting a misalignment problem between the valve and valve stem of the turbine control valve.
RCIC cold quick start injections to the reactor vessel and the cundansate storage tank at rated reactor pressure and 150 psig vere initiated from the control room. Manual (controlled) and cold quick start vessel injections at rated reactor pressure vare initiated from the remote shutdown panel, followed by flow step changes while injecting high (540 gpm) and low (330 gpm) RCIC flows into the vessel. A cold quick start vessel injection at 160 psig reactor pressure w4s also performed.
These tests demonstrated proper controller settings and system operation. RCIC testing ct cest ecnditien 1 has been satisfactcrily coupleted, with only the following deficiency remaining open at the completion of test condition 1.
1.
The time for RCIC to reach 600 gym during the 150 psig vessel injection cold quick start was 34.8 seconds instead of the required 30 seconds. This deficiency must be resolved prior to exceeding 80% power.
Test Condition 2 The RCIC 150 psig vessel injection cold quick start was successfully retested. A RCIC cold quick start injection to the condensate storage tank was performed at 150 psig reactor pressure.
Due to excessive flow restrictions in the test line, it was only possible to achieve 568 gpm flow. A detailed analysis revealed that these restrictions prevented rated flow in the RCIC system even if the system is fully capable of providing 600 gpm to the vessel. As a result, operability of the RCIC system is shown by achieving at least 450 gpm through the test line with at least 450 psig discharge pressure at greater than 2100 rpm turbine speed.
The technical specification requirements have been changed so that RCIC operability is based on vessel injection of 600 gpm which was demonstrated previously at 150 psig during test condition 1.
RCIC testing for test condition 2 has been satisfactorily completed.
Although all testing on the RCIC system has been completed, the steam flow isolation instrument setpointo which are calculated l
based on the change in bypass value steam flow will be recalculated at Test Condition 3 when more accurate valve flow data becomes I
available using previously obtained RCIC start data.
l
STARTUP TEST PROCEDURE 16 SELECTED PROCESS TEMPERATURES 1.
PURPOSE A.
The purposes of this test are as follows:
1.
To assure that the measured bottom head drain ten'.perature corresponds to bottom head coolant temperature during normal operations.
2.
To identify any reactor operating modes that could cause temperature strati fication.
3 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.
4.
To familiarize plant personnel with the temperature d i f ferential limitations of the reactor system.
2.
CRITERIA A.
Level 1.
1.
The reactor recirculation pumps shall not be started nor flow increased unless the coolant temperatures between the steam dome and bottom head drain are within 145 F (81 C).
2.
The recirculation pump in an idle loop must not be started unigss the loop suction temperature is within 50 F (28 C) of the active loop suction temperature if one pump is idle or the steam dome temperature if two pumps are idle.
B.
Level 2.
1.
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 At Test Condition Heatup, all of the applicable criteria were met.
The proper setting for the low flow control valve limiter for the recirculation pumps was detemined to be the minimum valve position (3%). With the Recirculation FCV's at minimum valve position, Reactor Water Cleanup Flow was decreased from 140 GPM to 100 GPM cauging the bottom head drain temperature to increase from 510 F to 520 F.
The control rod drive flow was increased from 50 GPM to 69 GPM which resulted in no additional change to the bottom head drain temperature.
3 At-test condition 1 and 2, all of the applicable' criteria were met.
The temperature difference between the steam dome and the bottom head drain was 24 F and 25 F (for Test Condition 1 and-2 respectively).
Therefore, at these test conditions minimal temerature stratification exists in the bottom head region.
-This test was successfully completed for test condition heatup, 1, 4
and 2 with all applicable criteria met.
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1 STARTUP TEST PROCEDURE 17 SYSTEM EXPANSION i
1.
PURPOSE j
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 the specified manner.
2.
CRITERIA A.
Level 1 1.
There shall be no evidence of blocking of the displacements of any system component caused by thermal expansion of the system.
2.
Electrical cables shall not be fully stretched.
3 Hangers snall not ce bottomed out oc have the spring fully stretched.
4.
Snubbers shall be in the operating range about the midpoint of the total travel range at operating temper-ature.
5.
The measured steady state displacement of the recircula-tion and main steam systems shall not exceed the allowable values.
B.
Level 2 l
1.
At a steady-state condition, the displacements of in-l strumented points with displacement measuring devices shall not vary from the calculated values.
If measured displacements do not meet these criteria, the piping l
design engineer must be contacted to analyze the data with regard to design stresses.
2.
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 l
calculated value and 50 percent of the calculated value from the initial position in the opposite direction of L
the calculated value.
3 Hangers will be in their operating range between the hot and cold settings.
I m-.
3.
RESULTS Open Vessel Cold readings for hangers and snubbers were obtained and recorded. Also all Lanyard potentiometers on the Main Steam, Feedwater, and Recirc system were verified installed properly.
Heatup A drywell inspection was made at 362 F.
This inspection turned up no abnormalities. All applicable criteria were met.
i Three drywell inspections were made at rated temperature to record hanger and snubber hot readings.
General Electric and Sargent and Lundy have found drywell piping suspension to be acceptable.
All applicable criteria were met.
Following the third major thermal 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 met.
Displacements of instrumented points were recorded every 50 F for the Recirculation, Main Steam, and Feedwater systems.
Displacements for certain points exceeded Level 2 criteria but analyses by Gene.*al Electric and Sargent and Lundy have shown these except!onc to te minor with actual displacements yielding acceptable stress levels.
i Following the first major thermal cycle, Reactor Pressure Vessel Stabilizer Shims were adjusted.
Test Condition 1 Displacements of instrumented points were recorded at steady state power.
Displacements for certain points exceeded Level 2 criteria but analyses by General Electric and Sargent and Lundy have shown these exceptions to be acceptable with actual displacements yielding acceptable stress levels.
STARTUP TEST PROCEDURE 19 CORE PERFORMANCE 1.
PURPOSE A.
The purpose of this test is to evaluate the following core performance parameters at Test Conditions 1 through 6:
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 A.
Level 1 1.
The Maximum Linear Heat Generation Rate (MLHGR) of any rod during steady state conditions shall not exceed 13.4 Kw/ft.
2.
The steady state Minimum Critical Power Ratio (MCPR) shall not exceed the limits specified in t.he plant technical specifications.
3.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) shall not exceed the limits shown specified in the plant technical specifications.
4.
Steady State reactor power shall be limited to 3323 MWt and values on or below the analyzed flow control line.
3.
RESULTS The core performance parameters were demonstrated to be within the range required by the Technical Specifications for test condition 1 and 2 (table 1 and 2 respectively).
This test was successfully completed for test conditions 1 and 2 with all applicable criteria met.
TABLE I CORE PERFORMANCE DATA (TEST CONDITIOh 1)
CORE OBSERVED LEVEL 1
' PARAMETER LOCATION VALUE CRITERIA LIMIT CTP 578 Mwt
<1600 Mwt MLHGR 27-34-11 2.84 KW/ft 613.40 KW/ft MCPR 27-36 4.651 21.921 MAPLHGR Fuel Type; 8CR 183 27-34-11 2.48 KW/ft 612.0 KU/ft 8CR 233 27-36-11 2.46 KW/ft 511.88 KW/ft
[
wt 36.0 Mlb/hr N/A TABLE 2 CORE PERFORMANCE DATA (TEST CONDITION 2)
CORE OBSERVED LEVEL 1 PARAMETER LOCATION VALUE CRITERIA LIMIT CTP 1494 Mwt
< 1940 Mwt MLHGR 23-48-4 5.81 KW/ft 5 13.40 KW/ft MCPR 9-38 2.571 2 1.567 MAPLHGR Fuel Type:
8CR 183 19-42-4 4.99 KW/ft 5 12.00 KW/ft 8CR 233 25-50-4 4.99 KW/ft 6 11.90 KW/ft 8CR 711 11-54-4 1.83 KW/ft 6 11.52 KW/ft Wt 46.8 Mlb/hr N/A i
l
5 1
l STARTUP TEST PROCEDURE 22 PRESSURE REGULATOR 4
4 1.
PURPOSE A.
To determine the optimum settings for the pressure control loop by analysis of the transients induced in the reactor l
pressure control system by means of the pressure regulators.
~
B.
To demonstrate the takeover capability of the backup pressure l.
regulator via simulated failure of the controlling pressure regulator and to set the regulator 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 l.
~
A.
Level 1.
i 1.
The transient response of any EHC system-related variable j,
to any test input must not diverge, e
I B.
Level 2.
1.
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 O.25 2.
The response time from pressure setpoint input until the pressure peak of the' pressure peak of,the pressure regulator Inlet pressure must be less than or equal to 10 seconds, with the Recirculation Flow Control System in the Position Command Mode only.
3, 3.
Pressure control system deadband, delay, e,tc., shall be i
small enough that steady state limit cycles (if any) i shall produce steam flow variations no larger than + 0.5 percent of rated' steam flow.
~
t 4.
The normal difference between regulator setpoints' must be
~
j small enough that the peak neutron flux and peak vessel o
pressure remain below the scram settings by 7.5 percent and 10 psi respectively, for the Regulator Failure Test performed at Test Condition 6.
C.
Level 3 l.
Dynamics of both pressure, regulators shall'be' essentially i
identical.
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2.
The variation in incremental regulation (ratio of the maximum to the minimum value of the quantity
" incremental change in pressure control signal / incremental change in steam flow" for each flow range) shall meet the following:
% of Design Valve Steam Flow (with Valves Wide Open)
Variation 0 to 90%
- $. 4
- 1 90% to 97%
- $1 2
- 1 90% to 99%
- <. 5
- 1 3.
Control or Bypass Valve motion must respond to pressure inputs with deadband (insensitivity) no greater than +
0.1 psi.
3.
RESULTS Test Condition 1 Primary and backup pressure regulator testing verifying proper control system response was satisfactorily completed.
These sections were repeated for two different pressure regulator lead / lag settings, showing that the original settings were most optimal.
Backup regulator takeover capability was tested and found acceptable.
Test Condition 2 Pressure regulator system tuning was performed with the turbine-generator on line and the turbine-generator load selector set so that pressure tranFients were controlled by 1) the turbine control valves, 2) the turbine control valves and the turbine bypass valves, or 3) the turbine bypass valves. For each'of the above operating conditions, pressure setpoint step change testing and simulated regulator failure testing was performed.
Test results were acceptable.
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STARTUP TEST PROCEDURE 23A FEE 0 WATER CONTROL SYSTEM 1.
PURPOSE A.
The purpose of this test is as follows:
1.
To demonstrate satisfactory reactor water level and feedwater flow rate ' cont rol. Heasurements of feedwater system stability and performance are analyzed for this determination.
2.
ACCEPTANCE CRITERIA A.
Level 1 1.
In the automatic mode, the response of any level system controlled variable to any test input change or disturbance must not diverge.
B.
Level 2 1.
Level contro!.spstem-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.
2.
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 response rate will be assessed by determining the time required to pass linearly through the 10% and 90% response points of the flow transient.
3.
The dynamic flow response of each feedwater actuator (turbine or valve)~to small (less than 10%) step disturbances in the manual mooe shall.be:
a.
Dead time 6 1.0 sec.
5.
Maximum time to 10%*
6 1.1 sec.
c.
Maximum time from 10% to 901*
6 1.9 sec.
Settling time to within +5%*:of o.
the final value 614.0 sec.
e.
Peak overshoot
- 6 15%
- % of input step disturbance.
C.
Level 3 1.
Transmi tters, square ' root conve rters, summers, recorde rs, etc.
i of the total feedwater flow and steam flow circuits shall be calibrated ana adjusted properly so that the total steam flow recorder indication matches the total feedwater flow recorder indication within +4% of rated feedwater flow, i
t
i 3.
RESULTS A.
Test Condition 1
\\
At Test Londition 1, the Motor Driven Reactor Feedwater Pump and Turbine Driven Reactor Feedwater Pump IB were subjected to level step response testing with the feedwater control system in the single element mode of control. The results of this testing indicated satisfactory performance of the feedwater control system.
5 Turbine Driven Reactor Feedwater Pump 1A was originally scheduled for 4
testing at this Test Condition. This testing was delayed to Test Condition 2 as a result of a speed stability problem with the turbine..The deletion of this testing at Test Condition 1 was discussed and approved via telecon with fir. Fred Lieberbach (NRR) on 10/08/82.
B.
Test Condition 2 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 manner, the Level 2 dynamic response criteria (decay ratio, rise time, overshoot, etc.) was not satisfied at all times.
Even though this criteria was not satisfied, satisfactory feedwater control system i
performance was indicated in the single-element and three-element modes of control.
Control system performance will be reevaluated at Test Condition 3 4
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. _. ~ _ _ _
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I STARTUP TEST PROCEDURE 23C FEEDWATER SYSTEM, FEEDWATER PUMP TRIP r
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a g
1.
PURPOSE A.
To demonstrate the capability of the automatic core flow 3
runback feature to prevent low water level scram following the q
trip of one feedwater pump.
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 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 The ability of the standby motor driven feedwater pump to maintain 3
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 i
2.
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STARTUP TEST PROCEDURE 23D FEEDWATER SYSTEM, MAXIMUM FEEDWATER RUNOUT CAPABILITY 1.
_P_URPOS E The purpose of this test is to determine the maximum feedwater runout cability.
2.
CRITERIA Level 1 A.
The feedwater flow runout capability must not exceed the assumed value in the FSAR.
3 RESULTS At Test Condition 1, setup of the turbine driven reactor feedwater pumps was performed. The feedwater turbine high and low speed stop setpoints were verified and function generator data was obtained on both turbines during uncoupled runs.
Results of this testing was satisfactory.
Feedwster flow runout capability is not determined until Test Condition 6, so the level I criterion is NA.
STARTUP TEST PROCEDURE 25 MAIN STEAM ISOLATION VALVES 1.
PURPOSE A.
The purpose of this test is as follows:
1.
To enctionally check the main steam line isolation va!ves (MSIV's) for proper operation at selected power level.
2.
To determine isolation valve closure times.
3 To determine the maximum power at.which full closures of a single valve can be performed without a scram.
4.
To determine the reactor transient behavior resulting from the simultaneous full closure of all MSIV's.
2.
CRITERIA A.
Level 1 1.
MSIV closure time, excluding slectrical delay shall be no faster than 3.0 seconds and licluding electrical delay shall be no slower than 5.0 seconds (each valve, not averaged).
2.
The positive change in vessel dome 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, 6.2.D, by more than 2% of rated value.
3 Feedwater control system settings must prevent flooding i
of the steam lines.
B.
Level 2.
1.
During full closure of individual valves:
2 a.
Peak vessel pressure must be 10 psi (0.7 Kg/cm )
1.
below scram, b.
Peak neutron flux must be 7.5% below scram.
(
[
c.
Steam flow in individual lines must be 10% below the j
isolation trip setting.
l d.
Peak heat flux must be 5% less than its trip point.
l c.
2.
Initial action of RCIC and HPCS shall be automatic if the level 2 setpoint is reached, and system performance shall be within specification.
3.
The relief valves must reclose properly (without leakage) following the pressure transient.
4.
For the full MSIV closure from greater 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 Power Pressure Heat Flux Dome Pressure l
(%)
(psla)
(%)
(psi) 100 1020 0
To be determined based upon actual plant conditions at the time the test is performed.
3 RESULTS A.
Test Condition Heatup 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.
B.
Test Condition 2 The MSlV's were " slow closed" to functionally demonstrate proper valve operation. The fastest MSIV (from test condition heatup and 1) was " fast closed" a second time to verify that the above single valve closure will not cause a reactor scram.
The above closure time data has been included in table 1.
This test was successfully completed for test conditions heatup, 1, and 2 with all applicable criteria met.
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o TABLE 1 4
MSlV CLOSURE TIME PERFORMANCE MAIN STEAM ACTUAL CLOSURE CLOSURE TIME PERCENT TEST ISOLATION VALVES TIME
- LIMIT
- POWER CONDITION 1821-F022A 3.83/4.06 3.0/5.0 3%
Heatup 1821-F0228 4.13/4.31 3.0/5.0 3%
Heatup 1821-F022C 3.76/4.03 3 0/5.0 3%
Heatup IB21-F022D 3.64/3.82 3.0/5.0 3%
Heatup 1821-F028A 3.82/4.15 3.0/5.0 3%
Heatup 1821-F028B 3.71/3 99 3.0/5.0 3%
Heatup 1821-F028C 3.46/3 70 3.0/5.0 3%
Heatup IB21-F028D 3.41/3.80 3.0/5.0 3%
Heatup 1821-F022A 3.62/3.97 3.0/5.0 24%
1 IB21-F0228 4.00/4.19 3 0/5.0 24%
1 1821-F022C 3.98/4.15 3.0/5.0 24%
1 IB21-F022D 3.67/3.74 3.0/5.0 24%
1 IB21-F028A 3.74/4.18 3.0/5.0 24%
1 IB21-F028B 3.79/4.15 3.0/5.0 24%
1 1821-F028C 3.53/3.69 3.0/5.0 24%
i IB21-F028D 3.68/3.81 3.0/5.0 24%
1 1821-F028C 3.35/3.63 3.0/5.0 45%
2
- MSIV closure time, excluding electrical delay / including electrical delay
STARTUP TEST PROCEDURE 26 RELIEF VALVES 1.
PURPOSE A.
The purpose of this test is as follows:
1.
To verify the proper operation of the primary system relief valves.
2.
To determine each relief valve's capacity.
3 To verify that the discharge piping is not blocked.
4.
To verify that each relief valve reseats following operation.
5.
To obtain a transient recorder signature of each relief valve operation for subsequent comparisons.
6.
To confirm proper overall functioning of the Low-Low Set Pressure Relief Logic.
7.
To verify proper safety / relief valve discharge line backpressure.
2.
CRITERIA Level 1.
A.
There should be positive indication of steam discharge during the manual actuation of each valve.
B.
The sum total of the percentage corrected flow rates must be greater than 111.5% of the Nuclear Boiler warranted steam flow at 103% of the spring setpoint pressure of 1165 psig.
C.
, n: Low-Low Set Pressure Relief logic shall function to preclude subsequent simultaneous dRV 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.
No individual relief valve may have a flow rate (corrected to the setpoint pressure) that, considering measurement uncertainties, is less than 90%, or greater than 122.5%, of its expected flow rate of 862,400 lbs/hr at 103% of the spring setpoint pressure of 1146 psig.
E.
No more than 25% of the installed relief valves may have an individual corrected flow rate that is between 90% - 100% of their expected flow rates.
F.
The total flow capacity of the safety relief valves used in the Autogatic Depressurization System must be equal to or greater than 4.8 x 10 lbs/hr. at 1125 psig when the valve having the highest measured capacity is assumed to be out of service.
G.
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.
H.
Discharge line backpressure shall be comparable with information presented on the Nuclear Boiler Process Diagram.
l.
When the Low-Low Pressure Relief logic functions, the open/close actions of the SRV's shall occur within +13 psi and +20 psi of their design points respectively.
3.
RESULTS This test was perforned at Test Condition Heatup.
Each safety relief valve was opened and closed demonstrating proper functioning.
Positive steam discharge for each valve actuation was indicated by bypass valve position decrease and SRV tailpipe temperature increase responses.
Valve rescating 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. All applicable test criteria were satisfied at test condition heatup.
STARTUP TEST PROCEDURE 27 GENERATOR LOAD REJECTION 1.
PURPOSE 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.
For 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 following 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 dome 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 turbine trip would occur.
Level 2.
A.
There shall be no itSIV closure in the first three minutes of the transient and operator action shall not be required in that period to avoid the MSIV trip.
B.
The positive change in vessel dome pressure and in simulated heat flux which occurs within the first 30 seconds efter the initiation of either generator or turbine trip must not exceed the predicted values.
C.
Electrical load transfers occur as designed.
D.
The reactor shall not scram for ini tial thermal power at less than or equal to 25% of rated.
E.
If the Level I criterion (6.1.c of this procedure) for the two recirculation pump drive flow coast down transient is passed, the data shall be analyzed within 3 weeks for compatibility with the safety analysis.
1 1
3 RESULTS
)
At test conditton 2, generater load rejection within bypass capa.-lty was l
initiated by opening the generator circuit breakers. The feedwater system i
functioned properly and there was no reactor scram. There was no MSIV i
closure trip and automatic-bus fast transfer was successful. As the main generator OCB's were opened the generator lockout relay caused a turbine trip as expected. All level criteria were met.
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STARTUP TEST PROCEDURE 28 SHUTDOWN FROM OUTSIDE THE CONTROL ROOM 1.
PURPOSE A.
The purpose of this test is to demonstrate that the reactor can be brought from a normal initial power level to the point where cooldown is initiated and under control with Reactor Vessel pressure and water level controlled from outside the Main Control Room.
2.
CRITERIA Level 2 A.
During a simulated main control room evacuation, the reactor must be brought to the point where cooldown can be initiated, and the reactor vessel pressure and water level must be controlled using equipment and controls outside the main control room.
3.
RESULTS On 9-10-82, the LaSalle Unit 1 Reactor was shutdown satisfactorily from outside the control room following a simulated main control room evaluation. The reactor was scrammed, the MSIV's closed, the reactor vessel level and pressure were controlled from outside the Main Control Room by a team of operating personnel representing the minimum shift manning required by Technical Specifications.
Ability to control suppression pool temperature from outside the Main Control Room was also demonstrated.
The test criteria was satisfied.
STARTUP TEST PROCEDURE 29 REClRCULATION FLOW CONTROL SYSTEM 1.
PURPOSE The purposes of this test are:
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.
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 Loop 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.
1.
None.
Level 2.
A.
Position Loop Criteria.
Gains and limiters shall be set to obtain the following response:
1.
Maximum steady state rate of change of valve position shall be between 9 s 11% per second for a 100% position demand input.
(Initial valve velocity may exceed this limit for a short time.)
2.
Gains shall be set to give as fast a response as possible for small i
positien demand input within the overshoot criterion and without additional valve duty cycle.
(See FCV duty criterion (6.2.F.1) for valve duty cycle requirement.)
4
3 Tha decry ratio of any oscillatory controlled variable must be 10.25, when operating above the minimum core flow for Recirculation Master Manual mode.
Below this minimum core flow, the decay ratio must be 10.50 with the recommendatio-that each control system be adjusted to meet 40.25 unless there is an identifiable performance loss inv Hved at higher power levels.
B.
Flow Loop Criteria.
1.
The decay ratio of any oscillatory controlled variable must be 4 0.25, when operating above the minimum core flow for ifecirculation Master Manual mode.
Below this
~
minimum core flow, the decay ratio must be 10.50, with the recommendation that each control system be adjusted to meet 40.25 unless there is an identifiable performance loss inv Eved 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 + 5 sec.
C.
Flux Loop Criteria.
1.
The decay ratio of any oscillatory controlled variable must be 10.25, when operating above the minimum core flow for Recirculation Master Manual mode (loops A and B receive command from a common point).
Below this minimum core flow, the decay ratio must be 10.50, with the recommendation that each control system be adjusted to meet 10.25 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:
a.
Deadband, percent rated flux demand: 10.5 b.
Delay time for flux demand steps, sec.: 10.8.
c.
Response time for flux demand steps, sec.:
4.2.5 d.
Maximum allowable flux overshoot, for step demand of 120% of rated is, in percent:
2.
e.
Flux settling time, sec.: 1 15 3
Switching between estimated and actual flux should not exceed 5 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 i.5% of rated flux demand.
i
D.
Lo:d Following Loop Criteria.
1.
The decay ratio of any oscillatory controlled variable must be 10.25, when operating above the minimum core flow for Recirculation Master Manual mode.
Below this minimum core flow, the decay ratio must be 10.50, with the reconmendation that each control system be adjusted to meet 10.25 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 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 (i.e., between 65 percent - 100 percent power).
E.
Scram Avoidance and General Criteria.
For anyone of the above loops' test maneuvers, the trip avoidance margins must be at least the following:
1.
2.
For simulated heat flux _> 5.0%.
3 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.
The flow control valve duty cycle in any operating mode shall not exceed 0.2% -Hz.
Flow control valve duty cycle is defined as:
Total valve travel (%)
(% - Hz) 2x time span in sec.
Level 3 A.
Position Loop Cri teria.
1.
Position loop deadband shall be 0.25% of full valve stroke (hot only).
l 2.
Overshoot after a small position demand input (0.5 to 5%)
step shall be 410% of magnitude of input.
3 Performance for 0.5 to 5% steps:
Delay:
4_0.15 seconds 1
Response
4,0.45 seconds 4.
Performance for steps less than 0.5%, delay / response combination must be within the acceptable region of i
Figure 12.10.
5 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 g one second. The range will be + 5% of the step change or + 0.05% full stroke, whichever is greater.
6.
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.
Incremental gain from flow demand input to sensed drive flow shall not vary by more than 2 to 1 over the entire flow range.
2.
Deadband from function generator for valve position demand input to flow shall be 4_0.5%.
3 The variation in open flow loop response times shall not i
exceed a factor of 3 to 1.
This applies to both the symmetry of positive and negative flow changes and to responses at the low and high ends of the Master Manual mode range.
4.
The delay time for flow demand step I to 5% shall be
/ 0.4 sec.
5 The response time for flow demand step I to 5% shall be 4_1.1 sec.
6.
The maximum allowable flow overshoot for step demand of 1 to 5% of rated shall be 6% of rated.
7 The flow demand step settling time shall be 4_6 sec.
8.
Linearization of the 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.
3 RESULTS Test Condition Open Vessel.
With the recirculation pumps idle, hydraulic actuator position controllers were adjusted to meet their specifications.
This was done by the performance of Flow Control Valve stroking tests from full open to full shut and by the performance of small step position demands from various initial positions.
Data analysis indicates that overall control system operation is satisfactory although certain level 2 and 3 test criteria were not satisfied.
Test Condition 1 With the recirculation pumps in operation on the Low Frequency Motor Generators the recirculation loop hydraulic actuator position controllers (servo controllers) were adjusted to meet their specifications.
This was done by the performance of a series of small position demand inputs to the flow control valve with valve initial positions of from 10% to 90% open.
Evaluation of the test results indicates that overall control system operation is satisfactory.
Although some level 2 and 3 criteria were violated, GE has found the results to be acceptable.
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STARTUP TEST PROCEDURE 30 RECIRCULATION SYSTEM 4
l 1.
PURPOSE 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 l
In the operating region of the power-flow map, i
{
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 and/or scram.
D.
To record and verify acceptable performance of the recirculation two pump circuit trip system.
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 on l
Figure 14.2-7 of the FSAR.
Level 2 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.
I B.
The reactcr water level margin to avoid a high level trip shall be. greater than or equal to 3.0 inches during the one i
pump trip.
i i
C.
The simulated heat flux (TPti) margin to avoid a scram shall bc l
greater than or equal to 5.0 percent during the one pump trip.
l l-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 Recirculation System.
l E.
- During recirculation pump restart (s) the scram trip avoidance l
margins must be at least the following:
t l
1.
For APRit, greater than or equal to 7.5%.
l l
2.
For simulated heat flux,' greater than or equal to 5.0%.
i F.
If the level I criteria for the two pump trip coastdown transient L
is met, the data shall be analyzed within two weeks to ensure compatibility with the safety analysis.
1 l
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3 RESULTS At test condition 2, the recirculation pump transfer from 60 Hz to i
15 Hz occurred with feedwater flow at 32.2% of rated prior to any observable cavitation.
Recirculation system performance data was taken. All applicable test criteria were satisfied.
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STARTUP TEST PROCEDURE 31 LOSS OF 0FFSITE POWER 1.
PURPOSE A.
To determine electrical equipment transient performance during a loss of auxiliary power.
B.
To determine reactor system transient performance during a loss of auxiliary power.
2.
CRITERIA 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, i f 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 On 12/2/82 this test was performed, demonstrating adequate station electrical equipment and reactor system transient performance during a sustained loss 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.
All acceptance criteria were satisfied.
STARTUP TEST PROCEDURE 33 DRYWELL PIPING VIBRATION 1.
PURPOSE A.
The purpose of this test is to verify that the main steam, reactor recirculation, and feedwater piping have acceptable vibration.
2.
CRITERIA Level 1.
A.
The measured amplitude for vibration of the recirculation system during recirculation pump trips and subsequent coast down shall not 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.
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 during steady state operations should not exceed the expected stresses.
3.
RESULTS 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 were inspected for perceivable vibration under steady-state conditions. RCIC instrumentation lines outside containment were inspected. All acceptance criteria were met.
The inspection of RCIC piping in the CST to RPV flow path which was originally scheduled for Test Condition 1 was performed during Test Condition Heatup since all necessary conditions and prerequisites were met at that time.
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) was within the acceptance criteria. Vibration induced in the feedwater lines due to the feedwater pump trip (STP-23C) was found to be acceptable. Vibration induced in the main steam lines due to turbine trip following loss of offsite power (STP-31) satisfied the acceptance criteria.
STARTUP TEST PROCEDURE 34 REACTOR INTERNALS VIBRATION 1.
PURPOSE A.
The purpose of this test is to obtain vibration measurements on the Jet pumps to confirm the mechanical integrity of the system with respect to flow induced vibration and to verify the accuracy of the analytical vibration model.
This test is in conformance with Regulatory Guide 1.20 requirements for Non-Prototype. Category ll Plants (similar to prototype but some component differences).
2.
CRITERIA Level 1.
A.
The peak stress intensity may exceed 10,000 psi (single amplitude) when the component is deformed in a manner corresponding to one of its normal or natural modes but the fatigue usage factor must not exceed 1.0.
Level 2.
A.
The peak stress intensity shall not exceed 10,000 psi (single amplitude) when the component is deformed in a manner corresponding to one of its normal or natural modes.
This is the low stress limit which is suitable for sustained vibration in the reactor environment for the design life of the reactor components.
Level 3 A.
The measured flow shall not differ from the calculated flow by more than:
1.
15% - for flow less than 50%.
2.
10% - for flow less than 90% and greater than or equal to 50%.
3 5% - for flow greater than or equal to 90%.
i 3
RESULTS OPEN VESSEL TEST CONDITION During open vessel testing, the vibration measurements obtained on the jet pumps did not violate the level 1 or level 2 acceptance criteria. The peak stress intensity observed was 6100 psi.
During the course of testing it was determined by G.E. that two (2) of the installed accelerometers and four (4) of the installed strain gauges had failed.
A deficiency was generated to indicate the failure of this instrumentation. The loss of this instrumentation had no impact on the test because; I
l
-the remaining accelerometers and strain gauges indicated results with good correlation to the measurements taken during preoperational testing, thus, vibration at those failed locations could be predicted.
-the measurements from the functioning instrumentation Indicated values which did not even approach the acceptance criteria limits.
The remaining deficiencies generated dealt with testing not. rela ted to vibration of the vessel internals, but rather to tests which G.E. required LaSalle to repeat as a part of STP-34.
The deficiencies dealt with the following:
-loop flows at minimum flow control valve position were not between 20%-25% of rated.
-loop flows were not within 3% of one another for the same flow control valve positions.
-loop flow hysteresis was not less than 2%.
Upon an engineering analysis and evaluation, G.E. found the values received during STP-34 to be acceptable.
I
STARTUP TEST. PROCEDURE.70 REACTOR WATER CLEANUP 1.
PURPOSE A.
To demonstrate specific aspects of the mechanical operability of the Reactor Water Cleanup System with the reactor at rated pressure and temperature.
3.
CRITERIA Level 2 A.
The temperature at the tube side outlet of the non-regenerative heat exchangers shall not exceed 130 F in the blowdown mode and shall not exceed 120 F in the normal mode.
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 be within the flow and outlet temperature limits of 150 F in the normal mode, 180 F in the blowdown mode, and 762 gpm in either mode.
D.
Recalibrate bottom head flow indicator (R610) against RWCU flow indicator (R609) if the deviation is greater than 10 gpm.
E.
Pump vibration measured on the casing between the motor and pump shall be less than or equal to 2 mils.
3 RESU LTS This test was performed at Test Condition Heatup.
Reactor Water Cleanup System (RWCU) performance was evaluated for the normal, blowdown, and hot standby modes of operation.
In addition, a comparison of the bottom head flow indicator and the RWCU flow indicator was conducted. All test criteria were satisfied.
See Table I for test data.
RESULTS (Continued)
TABLE 1 Parameter Worst Case Measured Temperature at the tube side outlet of the non-regenerative heat exchangers Blowdown Mode 107 5 F Normal Mode 115.3 F Pump available NPSH Normal Mode 2206.3 ft Blowdown Mode 2178.3 f t.
Hot Standby 2112.4 ft.
Both pumps @ 270 gpm 27.56 ft.
Cooling water supplied to the non-regenerative heat exchangers flow and temperature Normal Mode T -= 149 F W = 510 gpm Blowdown Mode T = 180 F W: 510 gpm Bottom head flow indicator /RWCU flow indicator deviation 9 gpm Pump vibration all values less than 2 mils l
i i
i
STARTUP TEST PROCEDURE 71 RESIDUAL HEAT REMOVAL SYSTEM 1.
PURPOSE A.
To demonstrate the ability of the Residual Heat Removal (RHR)
System 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.
B.
To demonstrate the ability of the RHR system, in conjunction with the Reactor Core Isolation Cooling (RCIC) System, to condense steam.
2.
CRITERIA Level 2 A.
The RHR System shall be capable of operating in the steam condensing (with both I and 2 heat exchangers), suppression pool cooling, and shutdown cooling modes (with either heat exchanger operating) at the flow rates and temperature differentials indicated on the process diagrams.
B.
In the steam condensing mode, for small disturbances, each variable must have a decay ratio less than 0.25 throughout each controller's expected operating range.
C.
The time to place the RHR heat exchangers in the steam condensing mode with the RCIC using the heat exchanger condensate flow for suction shall be one-half hour or less.
D.
The RHR System performance in the shutdown cooling mode shall not be less than that indicated on the process diagram.
3.
RESULTS Test Condition 1 Time constant, overshoot, and decay ratio (level 2 criteria violation) parameters for the RHR Loop "A" and "B" inlet pressure controllers exceeded the recommended values during some of the pressure step changes. However, satisfactory controller responses were observed and no controller setting adjustments were made. Time constants for the RHR Loop "A" and "B" level controllers exceeded the recommended values for some of the level step changes.
However, since the system crocess response is slower than the controller response, no controller setting adjustments were made.
Due to low and unstable RCIC flow RCIC suction pressure controller tuning could not be pe r forned. However, proper design function operation of the RCIC suction pressure controller was verified.
i e
y s.%
p.,
y
The time to establish the RHR heat exchangers in the steam condensing mode with RCIC using the heat exchanger condensate flow for suction was in excess of the level 2 criteria limit of 30 minutes. This is not a cause of significant concern as it is not a safety requirement.
RHR steam condensing mode system performance was not in agreement with the process diagram. Also, it was not possible to establish a single steam condensing loop with RCIC using the HX condensate flow for suction. These are all level 2 criteria violations. However, after consultation with GE it was determined that since the steam condensing mode of RHR is not a safety system, and the HX's were shown to operate in a stable manner while in the steam condensing mode, the level 2 criteria violations should not restrict proceeding with further testing at higher power levels.
4 4
STARTUP TEST PROCEDURE 74.
OFF-GAS SYSTEM l.
PURPOSE A.
The purpose of this test is as follows:
s 1.
To verify the proper operation of the Off-Gas System over its expected operating parameters.
2.
To determine the performance of the activated carbon absorber.
2.
' CRITERIA A.
Level 1 1.
The release of radioactive gaseous and particulate effluents must not exceed the limits specified in the site Technical Specifications.
2.
There shall be no loss of flow of dilution steam to the non-condensing stage when the steam jet air ejectors are pumping.
B.
Level 2 1.
The system flow, pressure temperature, and relative i
humidity shall comply.with the design specifications.
2.
The catalytic recombiner, the hydrogen analyzer, the jactivated carbon beds,'and the filters shall be operating
' properly during operation, i.e.,
there shall be no gross malfunction of these components.
l 3.
RESULTS Test Condition:
lieatup and I
/The release of radioactive gaseous and particulate effluents was demonstrated to be within the limits specified in the Technical Specification.
For both Startup.and llormal Mode, the dilution steam flow failed to satisfy the level I criteria.
After consultation with Genera! Electric it was determined that it 'would be safe to proceed with testing as long as dilution steam flow remained greater than 550dlb/hr-and reactor power'was less than 25%of rated power.
The above problem with low dilution steam flow was resolved during test condition 1, by re-calibrating the second stage SJAE' flow
(
indicator.
Test Condition:
Heatup" f'
The following level 2 criteria were not satisfied during test u -
condition heatup:
X s
i f
M'
..w.
r-
A.
Gas Reheater Outlet Temperature (Startup Mode).
B.
Loop B (Standby) Catalytic Recombiner Temperature (Normal Mode).
C.
Adsorber Vessel Temperature IN62-N023 (Normal Mode).
D.
Off gas Flow to After Filter (Normal Mode).
The actual values and criteria are included in Table 1.
After 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of steady state operation the gas reheater outlet temperature was indicating within the level 2 criteria (74 F).
The low Standby Recombiner Temperatures were identified as a system performance problem during the preoperational testing.
I t was determined that the standby Recombiner would require a longer heatup period prior to admitting off gases to the recombiner. This additional heatup time will not effect the overall performance of the Recombiners.
The high off gas flow was evaluated as acceptable with the understanding that as condenser performance improved and in leakage drops, so will off gas flow.
The low adsorber vessel temperature IN62-N023 was evaluated as acceptable result.
The following level 2 criteria were not satisfied ouring Test Condtion 1:
A.
Moisture Separator Outlet Temperature (Normal Mode).
B.
Adsorber Vessel Temperature IN62-N023 (Normal Mode).
C.
Off-Gas Flow to After Filter (Normal Mode).
The actual values and criteria are included in Table 2.
The Moisture Seperator Oulet Te:rperature is roughly twenty degrees less than the Adsorber Vessel Temperatures indicating the relative humidity at the charcoal beds should be reduced by the cooling / heating process. This would indicate the Moisture Separators are operating properly.
As anticipated above (Test Condition Heatup), both the Adsorber Vessel i
Temperature (IN62-N023), and Off-Gas Flow to After Filter values moved closer to their expected range with the increase in power.
This test was successfully completed for Test Condition Heatup, and 1.
All applicable criteria were met except for the afore mentioned exceptions and their justifications.
Further testing of STP-74, Off gas System will be performed at subsequent test conditions.
9
i TABLE 1 DESIGN SPECIFICATION
-DEVIATION (TEST CONDITION: HEATUP)
PARAMETER ACTUAL CRITERIA VALUE VALUE Gas Reheater Outlet 68 F 70 - 80 F Temperature (Startup)
Loop B (Standby) Catalytic Recombiner Temperature (Normal)
Top 310 F 325 - 375 F Middle 310 F 325 - 375 F Bottom 310 F 325 - 375 F Adsorber Vessel Temperature 66 F 72 - 82 F IN62-N023 (Normal)
Off-Gas Flow to After 35 SCFM 6 - 30 SCFM Filter (Normal)
TABLE 2 DESIGN SPECIFICATION DEVIATION (TEST CONDITION:
1)
PARAMETER ACTUAL CRITERIA VALUE VALUE Moisture Separator 51 F 33 - 500F Outlet Temperature Adsorber Vessel Temperature 71 F 72 - 82 F IN62-N023 Off-Gas Flow to After 32 SCFM 6 - 30 SCFM Filter
. - -