PY-CEI-NRR-1093, Post-Refueling Startup Test Summary Rept,Cycle 2
| ML19324B753 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/03/1989 |
| From: | CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| Shared Package | |
| ML19324B749 | List: |
| References | |
| PY-CEI-NRR-1093, NUDOCS 8911080143 | |
| Download: ML19324B753 (37) | |
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THE CLEVELAND ELECTRIC ILLUMINATING COMPANY b
PERRY NUCLEAR POWER PLANT UNIT 1 POST-REFUELING STARTUP TEST
SUMMARY
REPORT CYCLE 2 i
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1.1 INTRODUCTION
l This report presents a summary of the results from the post-refueling startup tests which were conducted in preparation for Cycle 2 at Unit 1 of the Perry Nuclear Power Plant. This report is submitted pursuant to Technical Specification 6.9.1.1.
1.2 PLANT DESCRIPTION The Perry Nuclear Power Plant is operated by the Cleveland Electric Illuminating Company (CEI) and is located near Lake Erie in Lake County, Ohio. Unit I has a Boiling Vater Reactor (BVR) nuclear steam supply system designed and supplied by the General Electric Company (GE) and designated BVR/6, with a Mark III containment. The balance-of-plant was designed by Gilbert Associates, Inc., Reading, Pennsylvania, as architect-engineer.
The rated core thermal power is 3579 HVt with a gross electrical output of 1250 MVe. The turbine is an 1800 rpm tandem compour.d, six flow, reheat unit consisting of one double flow high pressure stage in tandem with three double flov low pressure stages. The generator is a direct coupled 60 Hz, 22 KV, three-phase unit with a hydrogen cooled stator and rotor.
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SUMMARY
OF ACTIVITIES DURING REFUELING OUTAGE 1 The first refueling outage at Unit 1 of the Perry Nuclear Power Plant began on February 22, 1989 (main generator off-line) and was completed on August 5, 1989 (main generator on-line).
l The outage duration was 164 days.
Key features of this refueling outage were the performance of a complete core-offload for Technical Specification flexibility, the l
replacement of 272 (out of 748) fuel bundles, the replacement of the main generator rotor due to concerns of possible " copper dust" buildup, the l
installation of over 200 design changes, the completion of a large number of corrective and preventive msintenance activities, and the performance of required Technical Specification surveillance tests including a Containment Integrated Leak Rate Test.
Start Date 22 Feb., 1989 Stop Date 5 Aug., 1989 Duration 164 days Vork Orders 3,705 Design Changes 215 Surveillances 507 Repetitive Tasks 1,749 -
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PY-CEI/NRR-1093 L 2.1 DIFFERENCES IN FUEL DESIGNS g
Unit 1 at the Perry Nuclear Pover Plant used the following General on
,'A Electric fuel designs for Cycle 1:
P8 SIB 071-NOG-120M-150-T 92 bundles, GE6B-P8 SIB 176-4GZ-120M-150-T 232 bundles, GE6B-P8 SIB 219-4GZ-120M-150-T 424 bundles and the following General Electric fuel designs for Cycle 2:
GE6B-P8 SIB 176-4GZ-120M-150-T 52 bundles, GE6B-P8 SIB 219-4GZ-120M-150-T 424 bundles, GE8B-P8 SIB 301-5GZ-120M-150-T 136 bundles, GE8B-PBSIB301-7GZ-120M-150-T 136 bundles.
Complete descriptions of these General Electric fuel designs are given in GESTAR II, General Electric Standard Application for Reload Fuel. The rajor differences in the GE8 fuel relative to the GE6 fuel include:
1)
An increase in the number of gadolinium bearing rode in the bundle.
(5 or 7 vice 0 or 4) 2)
A decrease in the maximum veight percent of gadolinium in the rods.
(3.3 or 3.6 vice 5.0) 3)
The absence of fuel in the top six inches of gadolinium rods.
4)
An increase in maximum average percent enrichment in the bundle.
(3.01 vice 2.19) 5)
An increase in the peak Linear Heat Generation Rate (LHGR) limit.
(14.4 kW/ft vice 13.4 kW/ft) 6)
The use of lattice types vice fuel types in the determination of I
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits.
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Lover pressure drop across the upper tie plate which enhances reflooding characteristics during ECCS performance.
3.1 POST-REFUELING STARTUP TEST PROGRAM During refueling operations and the subsequent return to power, I
activities were controlled under normal administrative programs rather than a separate, formally defined post-refueling startup test program.
These administrative programs cover areas of normal operation such as:
Design Changes / Post-Modification Testing Post-Maintenance Testing Technical Specifications Surveillance Inservice Inspection Special Nuclear Material Control Periodic and Special Tests Computer Software Modification.
Radiation Control The acceptance criteria for these tests were derived from the requirements of these administrative programs. The reactor conditions during which the tests were conducted were guided by requirements in the I
appropriate administrative program...
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s ATTArdMENT 1 PY-rEI/NRR-1093 L 3.2 POST-REFUELING STARTUP TEST REPORTS As required by Technical Specification 6.9.1.2, this report addresses each of the startup tests identified in the Final (Updated) Safety Analysis Report Subsection 14.2.12.2.
Each test was evaluated by the Responsible System Engineer who determined whether the test was impacted by any refueling activity. Those tests which were determined not to have been impacted by any refueling activity are listed in Table 3.2-1.
For those tests which vere impacted by refueling activities, this report lists:
A Test Objective--from the appropriate subsection of USAR 14.2.12.2, and A Discussion--which includes the following, as appropriate 1.
A description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications:
2.
A description of any corrective actions that were required to obtain satisfactory operation;
'3.
Any additional specific details required in license conditions based on other commitments.
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14.2.12.2.2 Test Number 2 - Radiation Monitoring i
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Test Objective The purposes of this test are to determine the background radiation levels in the plant environs prior to operation for baseline data on activity buildup and to monitor radiation at selected power levels to assure the protection of personnel during plant operation.
Discussion In response to a post-licensing commitment identified in the NRC SER for Chapter 12 of the FSAR, Perry evaluated the potential radiation streaming from a spent fuel bundle in the Inclined Fuel Transfer System-(IFTS).
The test TXI-0069, " Inclined Fuel Transfer System (F42) Shielding Survey" vas performed twice to determine the adequacy'of the shielding at accessible locations adjacent to the IFTS. The first time was with two r
peripheral' spent fuel bundles and the second time was with two interior spent fuel bundles.
No shielding problems were observed during the test. The transfer tube mechanism was stopped at seven points during the transfer of spent fuel bundles from containment to the Fuel Handling Building. The points with the highest readings corresponded to the points where the potential for l
streaming was greatest. The highest readings were 1000 mr/hr on contact ard 125 mr/hr at 18 inches. In all cases, the measured values satisfied tha requirements for control of personnel radiation dose and occupancy times consistent with the guidelines of the standards for protection against radiation as outlined in 10CFR20, " Standards for Protection Against Radiation."
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14.2.12.2.3 Test Number 3 - Fuel Loading Test Objective i
The purpose of this test is to load fuel safely and efficiently to the full core size.
Discussion Fuel unloading and loading was conducted in Operational Condition 5 under 101-9, " Refueling," with all control rods fully inserted. Fuel movement followed a predetermined plan in accordance with a Fuel Movement Checklist per FTI-D09, "Use of the Fuel Movement Checklist." The core vas fully off-loaded in order to provide Technical Spacification flexibility.
The core offload started with the peripheral bundles in single bundle support pieces. Then, bundles in the control cells suspected to be in the vicinity of a leaking fuel bundle were removed for l
sipping.
Finally, bundles in East-West rows of control cells were removed starting with the southern rows and progressing northward.
After the core was off-loaded, the neutron sources were removed from their core locations.
Four of the neutron sources were placed in half blade-guides and later used as portable sources to provide minimum Source Range Monitor (SRM) count rates during the verification of instrument operability.
All fuel channel lifetime criteria vere met since no irradiated fuel bundles were rechanneled and only new fuel channels were installed on new fuel bundles. No control rod blades were replaced since no control rod blades were predicted to exceed the 34% depletion limit by the end of cycle 2.
The fuel on-load technique used for Cycle 2 varied in many ways from that used for Cycle 1.
In part, this was to ensure that all fuel loads were monitored by operable SRMs. After achievement of the minimum SRM count rate via the portable neutron sources, the first eight fuel bundles were loaded (two fuel assemblies per SRM) in cells around each of the four SRMs.
Fuel loading continued with the establishment of an continuous East-West " bridge" of fully loaded control cells between two of the SRMs.
Further fuel loading in those two quadrants was accomplished by loading control cells immediately adjacent to the previously loaded cells.
After all control cells in the first two quadrants were loaded, control cells in-the next quadrant were loaded after the establishment of a bridge of fully loaded control cells to the SRM in that quadrant.
Finally, the peripheral bundles in single fuel support pieces vere loaded with a camera verification of proper seating. -
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PY-CEI/NRR-1093 L After all fuel bundles were loaded, the location, seating, and orientation of every fuel assembly were verified to conform to the designed core' configuration per FTI-D01, " Core Verification."' No discrepancy was found. Since no intermediate core configuration exceeded the Cycle 1 and Cycle 2 fuel loading designs and since both designs were demonstrated to have the required shutdown margin throughout their respective cycle, the core remained suberitical with the required shutdown margin during all intermediate core configurations.
The suberiticality, shutdown margin, and SRM tests performed during fuel load were considered physics' tests in USAR 14.2.12.2.3.
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.i ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.4 Test Number 4 - Full Core Shutdown Margin
' Test Objective The purpose of this tast is first to demonstrate that the reactor is suberitical throughout the fuel cycle with any single control rod fully
~vithdrawn and second to determine quantitatively the shutdown margin of the as-loaded core.
t Discussion The shutdown margin and reactivity anomaly for Cycle 2 were demonstrated to be acceptable.
SVI-B13-T0005, "Insequence Suberitical Shutdown Margin Demonstration," was performed 3 days after core verification was complete. The moderator temperature was 91'F, and core flow was about 15 Hlb /hr. Control rods were withdrawn to a reactivity configuration which demonstrated shutdown margin was adequate at beginning of cycle. This facilitated installation of Reactor Prctection System shorting links for control rod testing and startup.
l Full core shutdown margin and reactivity anomaly were demonstrated to be 1
within their Technical Specification requirements during the first startup of Cycle 2.
SVI-B13-T0001, "Insequence critical Shutdown Margin Calculation," was performed at 0% power, 130*F moderator temperature, and i
24 Mlb/hr total core flow. The minimum shutdown margin for Cycle 2 was measured to be l.52% delta-K/K, compared to the Technical Specification l
minimum of 0.38% delta-K/K. The measured reactivity anomaly at the beginning of cycle was 0.245% delta-K/K, compared to the Technical Specification maximum of 1.0% delta-K/K.
i This was considered a physics test in USAR 14.2.12.2.4.
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PY-CEI/NRR-1093 L 14.2.12.2.5 Test Number 5 - Control Rod Drive System Test Objective The purposes of the control rod drive system test are to demonstrate that the Control Rod Drive (CRD) system operates properly over the full range of reactor coolant temperatures and pressures from ambient to operating, and to determine the initial operating characteristics of the entire CRD system.
Discussion Control rod insert /vithdrav timing was performed in accordance with SOI-Cll (CRDH), " Control Rod Drive Hydraulic System," on all control rods. This evolution began during the no-fuel period, stopped during core load, and resumed after the shutdown margin demonstration.
All rods were verified to have insert-and withdraw times of 40 to 60 seconds.
All tests were performed at atmospheric pressure and various temperatures.
Eleven control rods were friction tested per PTI-Cll P0003, " Control Rod t'
Friction Testing," after core verification and shutdown margin demonstration vere completed.
Eight of these rods had their associated control rod drive mechanisms replaced during the outage.
The other three had their associated fuel suppott piece removed and replaced during the outage. None of the drives that vere friction tested had differential pressure deviations exceeding 15 psid. The absence of control blade interference for the remaining 166 control rods was confirmed by an evaluation of the results of the control rod insert /vithdrav timing and the control rod scram timing tuts.
Control rod scram timing was performed for all control rods in accordance with SVI-Cll-T1006, " Control Ro:1 Maximum Scram Insertion Time." Control rod maximum scram insertion times were determined per Technical Specifi-cation 4.1.3.2.b for those rods whose associated control rod drive mechanisms were replaced, and those rods whose hydraulic control units had maintenance during the outage. This testing was performed at rated temperature and pressure prior to entering Operational Condition 1.
Scram timing was performed on the remaining rods in the core per Technical Spec.ification 4.1.3.2.a prior to exceeding 40% of rated thermal power. Buffer times (the slowing down time at position 00 during a scram) were not measured. During all previous tests, Perry has never l
4 experienced a control rod drive with unacceptable buffer time.
All rods were " fast" per Technical Specification 3.1.3.2. on their first attempt with the following three exceptions whose behavior was attributed to slov ope dng of the scram salves:
Rod 34-47 'slov' on first attempt, fast on second attempt Rod 34-51 'slov' on first attempt, fast on second attempt l
Rod 26-35 'slov' on first attempt, fast on second attempt l
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A fourth rod, rod 10-19, was slov on first attempt and slov on second attempt.
This rod was identified for further investigation.
The Technical Specilication action statements aciociated with these rods were satisfied. These four rods were rescheduliid for scram timing during the next performance of SVI-C11-T1006.
f Ganged rod timing was not performed because all centrol rods satisfied the individual insert /vithdrav timing requirements.
t No other hydraulic testing was performed on the C11 system as there were no changes to the system that vould have affected these parameters.
The timing tests were considered physic tests in USAR 14.2.12.2.5.
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14.2.12.2.6 Test Number 6 - SRM Performance and Control Rod Sequence Test Objective The purpose of this test is to demonstrate that the neutron 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.
Discussion The neutron sources verc removed during the refuel outage.
But, the irradisted fuel, which was reinstalled for cycle 2, generated a sufficient number of neutrons to ensure an adequate count rate for SRH operability during control rod withdrawal for the approach to criticality in acccrdance with IOI-1, " Cold Startup." A non-saturation check of the SRM's was not performed since no SRH vas replaced.
This test was considered a physics test in USAR 14.2.12.2.6.
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ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.7 Test Number 8 - Rod Sequence Exchange e
Test Objective The purpose of this test is to perform a representative sequence exchange of control rod patterns at a significant power level.
Discussion Because Cycle 2 utilized a Control Cell Core fuel loading pattern, the reactor operates in the same sequence for the entire cycle.
Therefore, no control rod sequence exchange was performed.
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Discussion IRM C has been inoperable since the refueling outage. This IRN was identified as a potential Technical Specification impact on future operation and was identified for corrective maintenance when the reactor is shutdown. The remaining seven IRMs were confirmed to have a half decade overlap with both the SRMs and the APRMs during performance of 10I-1, " Cold Startup."
This test was considered a physics test in USAR 14.2.12.2.8.
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14.2.12.2.9 Test Number 11 - LPRM Calibration Test Objective The purpose of this test is to calibrate the Local Power Range Monitoring (LPRM) system and to verify the LPRM flux response.
Discussion LPRM flux response was performed with the reactor power between 8% and 35% of rated power in accordar.ce with PTI-C51-P0001, " Verification of Proper LPRM Connection." Proper LPRM connection was verified by moving an adjacent control rod past each LPRM and observing that the LPRM reading changed accordingly. All of the 164 LPRMs responded satisfactorily.
Near 35% power, SVI-C51-T5351, "LPRM Calibration," vas performed to calibrate all LPRMs using the resulting calculation of the process computer programs, OD-1 and Pl.
A total of 8 LPRMs (32-33C,40-17D, 48-170,48-41D, 56-25B,56-25C, 56-33B, and 56-41B) vere found to be outside the 10%-band in the SVI's acceptance criteria. However, the LPRMs outside the desired band were not considered inoperable because the process computer and Average Power Range Monitor (APRM) calibration correct for any variations. The 21 LPRMs that were replaced during the refuel outage had the initial calibration current set to 700 micro amps.
These were subsequently calibrated in SVI-C51-T5351 with the other LPRMs.
This test was considered a physics test in USAR 14.2.12.2.9.
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6 ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.10 Test Number 12 - APRM Calibration Test Objective The purpose of this test is to calibrate the Average Power Range Monitor (APRM) system.
Discussion During the startup following refueling, the APRMs were calibrated twelve times in accordance with SVI-C51-T0024, "APRM Gain and Channel Calibration" as the reactor was brought to full power. The APRMs were calibrated to read within 2% of actual core thermal power on the following dates and power levels:
8-5-89 19.8, 23.4, 25.26 8-6-89 24.4, 25.2 8-7-89 33.7, 51./
8-8-89 62.9 8-9-89 73.3, 74.7, 94.8 l
8-11-89 100 Technical Specification and fuel warranty limits on APRM scram and rod block vere not exceeded. The startup APRM scram functions were checked in the SVI-C51-T0030, "APRM Channel Calibration," series.
This test was considered a physics test in USAR 14.2.12.2.10.
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1 ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.11 Test Number 13 - NSSS Process Computer Test Objective The purpose of this test is to verify the performance of the NSSS Process 1
Computer and online NSS computer programs under plant operating conditions.
Discussion The Traversing In-core Probe (TIP) tubing undervessel was removed / installed during the refuel outage per a Vork Order. The TIP alignment'was checked, after installation of the tubing, per ICI-C-C51-5, f
"TIP System Drive Control Unit (DCU) Draver Calibration."
The NSS software for Cycle 2 was updated and tested via Computer Program / Modification Request 89-25 under PAP-0506, " Computer Access and Software Control." The P1 calculations for thermal limits were ecmpated against those from the BUCLE calculations in order to establish the NSS software's acceptability for Technical Specification application.
In addition, the P1 calculations for thermal limits were compared against those from an independently updated copy of the same NSS software in order to evaluate this method for establishment of the process computer's acceptability for Technical Specification application. The results from this independently updated copy of NSS softvare are listed under P1-Backup. The' Local Power Range Monitor (LPRM) calibration factors l
calculated by P1 vere not compared against those calculated by BUCLE; i
however, the LPRM calibration factors calculated by OD1 were verified by manuel calculations.
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l In preparation for the above comparisons, the input data and intermediate i
calculations were evaluated. At the end of Cycle 1, records were generated for the NSS database for fuel constants, exposure and j
isotopics, LPRM exposures, and control rod exposures. While the plant 1
was shutdown for reactor refueling, the NSS database was updated to i
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reflect the new core design and a statistically significant number of l
entries in selected arrays were checked for reasonableness.
During the i
reactor startup from 0% to 25% of rated thermal power, the process i
computer values for control rod positions, LPRM readings, and thermal i
hydraulic parameters vere compared against readings from appropriate plant sensors. While slightly less than 25% of rated thermal power, the l
process computer calculation of reactor thermal power was compared i
rgainst a manual calculation per FTI-B05, " Core Heat Balance," and l
verified to agree to within 1% of rated thermal power. Actual agreement i
L was within 0.1% of rated thermal power.
After validation of the core L
thermal power calculation, OD15 was initiated to start the P4 core and generator energy accumulation which was checked via manual calculations.
Reactor power was raised to approximately 35% of rated thermal power and OD1 was performed to calibrate the LPRMs for the process computer. The process computer calculations for TIP normalize. tion factor, LPRM i
calibration constants, and LPRM substitute values were verified against I
appropriate manual calculations.
Prior to physically adjusting the 1
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e ATTACHMENT 1 PY-CEI/NRR-1093 L amplifier gains on ' the LPRMs, the power, time, flow, and run flags of t'a4 P1 program were checked for reasonableness and, the thermal limits from P1 vere compared against those from BUCLE. The results of that comparison vere as follows:
SOURCE P1 BUCLE P1-Backup TEST CONDITIONS:
Date 07-AUG-89 07-AUG-89 07-AUG-89 Core Thermal Power (MVt) 1284.6 1284.59 1277.4 Core Flov (Mlb/hr) 36.25 36.25 36.25 Core Inlet Subcooling (Btu /lb) 31.13 31.12 31.08 Reactor Dome Pressure (psia) 986.
985.57 986.
NSS SOFTWARE RESULTS:
MCPR Location: 33-20 2.331 2.306 2.346 Difference (%)
1.084 0.000 1.735 LHGR Location: 35-36-5 (kV/ft) 5.48 5.47 5.44-Difference (%)
0.183 0.000
-0.551 MAPLHGR Location: 35-36-5 (kV/ft) 4.73 4.73 4.70 Difference (%)
0.000 0.000
-0.638 The OD6 calculation of thermal limits was compared against the corresponding P1 calculations with no differences being identified. At this point the OD1, P1 and OD6 NSS programs were considered operational.
A second confirmatory check of these programs was performed at reactor conditions closer to rated core power and rated core flov.
In this second check the power, time, flow, and run flags of the P1 program were again checked for reasonableness; and, the thermal limits from P1 vere again compared against those from BUCLE. The results of this comparison vere as follows:
SOURCE P1 BUCLE P1-Backup TEST CONDITIONS:
Date 09-AUG-89 09-AUG-89 09-AUG-89 Core Thermal Power (MVt) 3383.7 3383.69 3383.7 Core Flov (M1b/hr) 89.22 89.22 89.22 Core Inlet Subcooling (Btu /lb) 23.87 23.87 23.87 Reactor Dome Pressure (psia) 1041.
1041.06 1041.
NSS SOFTWARE RESULTS:
HCPR Location: 19-34 1.300 1.297 1.301 Difference (%)
0.231 0.000 0.308 LHGR Location: 39-10-4 (kV/ft) 13.16 13.19 13.16 Difference (%)
-0.227 0.000
-0.227 MAPLHGR Location: 39-10-4 (kV/ft) 11.60 11.62 11.60 Difference (%)
-0.172 0.000
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'The ger.erator and core energy accumulation features of the P1 program c
were checked. After P1 vas considered operational, the daily generator i
and core energy accumulation and the LPRM exposure accumulation features of the P2 program were checked.
After P2 was considered operational, the i
total geaerator and core energy accumulation and the control rod exposure accumulation features of tne P3 program were checked..Pinally, the LPRM drift detection and reset feature of the P5 program was checked. At this point, all programs were considered operational.
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PY-CEI/NRR-1093 L 14.2.12.2.12 Test Number 14 - RCIC System The purpose of this test is to verify the proper operation of the Reactor i
. Core Isolation Cooling (RCIC) system over its expected operat ng pressure range and to demonstrate reliability at power conditions and during RCIC startup.
Discussion
.No changes were made to the RCIC system which could affect system operation.
l However, in a RCIC related matter, the Reactor Pressure Vessel (RPV) level nozzle shields temporarily installed during the initial Startup Test Program were replaced with the permanent design per DCP 87-385.
SXI-41, "RCIC Level Instrument Modification," which injected RCIC to the reactor vessel'at low power, verified no level anomalies occurred during the retest performed following implementation of the DCP. The test i
criteria vere equivalent to the criteria that had been used to verify the adequacy of the temporary nozzle shields installed during the initial
.Startup Test Program.
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ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.15 Test Number 18 - Core Power Distribution l
Test Objective The purpose of this test is to determine the reproducibility of the Traversing In-core Probe (TIP) system readings.
Discussion l
FTI-A16, " Total TIP Uncertainty," was performed in an octant symmetric control rod pattern at 95% of rated thermal power. The total TIP uncertainty was measured to be 1.7221%. A comparison to the 6.6% value assumed in the safety analysis shoved the TIP reading reproducibility to i
be acceptable.. The 6.6% resulted from the removal of two uncertainty components assumed in the safety analysis [ Local Power Range Monitor (LPRM) reading uncertainty and calculation uncertainty) from the 8.7%
limit on TIP reading uncertainty required by Technical Specifications.
This test was considered a physics test in USAR 14.2.12.2.15.
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14.2.12.2.16 Test Number 19 - Core Performance r
Test Objective The purposes of this test are to evaluate the core thermal power and core t'
flow and to evaluate the following core performance parameters:
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Maximum Linear Heat Generation Rate (MLHGR).
2.
Minimum Critical Pover Ratio (MCPR).
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Maximum Average Planar Linear Heat Generation Rate (MAPLHGR).
Discussion The fuel thermal limits were monitored on a shiftly basis under the Technical Specification Rounds Instruction. The core thermal power was continuously monitored and controlled under 10I-3, "Pover Changes."
During post refueling testing, there vere no fuel thermal limit violations and no steady-state operation above the more limiting of rated i
thermal power or the bounding licensed load line. The following lists the average thermal power (CMVTA) and the most limiting values for each of the thermal limits for every day in August that reactor power was above 25% of rated thermal power.
CMVTA MCPR MLHGR (kV/ft)
MAPLHGR (kV/ft)
Date (MVt)
Value Limit Ratio Value Limit Ratio Value Limit Ratio 07-AUG-89 1087 2.313 1.890 0.817 6.72 14.40 0.467 5.81 9.47 0.614 08-AUG-89 2089 1.829 1.431 0.782 8.85 14.40 0.615 7.41 10.33 0.717 09-AUG-89 2724 1.297 1.214 0.936 13.06 14.40 0.907 11.51 12.28 0.937 l
10-AUG-89 3564 1.257 1.201 0.955 13.99 14.40 0.972 12.09 12.40 0.975 e
11-AUG-89 3577 1.240 1.187 0.957 13.51 14.40 0.938 11.86 12.61 0.941 12-AUG-89 3396 1.457 1.373 0.943 13.50 14.40 0.937 11.85 12.61 0.939 13-AUG-89 3052 1.518 1.404 0.925 12.66 14.40 0.879 11.12 12.17 0.913 14-AUG-89 3505 1.254 1.220 0.973 13.15 14.40 0.976 11.54 12.61 0.979 15-AUG-89 3575 1.284 1.181 0.947 13.37 14.40 0.929 11.75 12.62 0.931 l-16-AUG-80 3575 1.279 1.180 0.922 13.39 14.40 0.930 11.77 12.63 0.932 17-AUG-89 3472 1.327 1.248 0.941 13.05 14.40 0.906 11.20 12.40 0.903 18-AUG-89 3577 1.274 1.180 0.926 13.03 14.40 0.905 11.18 12.40 0.902 19-AUG-89 3576 1.273 1.181 0.928 13.11 14.40 0.910 11.26 12.40 0.908 l
20-AUG-89 3326 1.264 1.198 0.947 13.05 14.40 0.906 11.21 12.40 0.904 21-AUG-89 3576 1.265 1.181 0.934 13.10 14.40 0.910 11.26 12.40 0.908 22-AUG-89 3577 1.268 1.180 0.931 13.07 14.40 0.908 11.24 12.40 0.907 I
23-AUG-89 3578 1.267 1.180 0.931 13.10 14.40 0.910 11.27 12.39 0.909 24-AUG-89 3379 1.266 1.180 0.932 13.08 14.40 0.909 11.26 12.40 0.908 25-AUG-89 3578 1.264 1.180 0.934 13.08 14.40 0.908 11.26 12.40 0.908 26-AUG-89 3578 1.260 1.181 0.937 13.14 14.40 0.913 11.32 12.40 0.913 27-AUG-89 3515 1.326 1.264 0.953 13.05 14.40 0.906 11.24 12.40 0.907 28-AUG-89 3579 1.256 1.181 0.940 13.06 14.40 0.907 11.26 12.40 0.908 l
29-AUG-89 3578 1.258 1.185 0.942 13.07 14.40 0.908 11.27 12.40 0.909
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30-AUG-89 3578 1.256 1.188 0.946 13.08 14.40 0.909 11.28 12.40 0.910 l
31-AUG-89 3578 1.249 1.189 0.952 13.13 14.40 0.912 11.32 12.39 0.914 This test was considered a physics test in USAR 14.2.12.2.16.
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14.2.12.2.17 Test Number 20 - Steam Production Startup Test Test Objective To demonstrate that the reactor steam production rate is sufficient to satisfy all appropriate warranties as defined in the contract.
Discussion This was a warranty test which was not performed for Cycle 2.
- However, the process input for the control Rod Drive (CRD) temperature was changed p
from an assumed value of 80'F to a temperature sensor on the suction line of the CRD pump. As a result, reactor thermal power was recalculated to be approximately 0.3 MVt lover. This small decrease in calculated reactor thermal power caused a corresponding slight increase in the full power adjusted total steam flow, which better satisfied the test objective than the original value.
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ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.18 Test Number 21 - Core Power-Void Mode Response Test Objective The purpose of this test is to measure the stability of the core power-void dynamic response and to demonstrate that its behavior is within specified limits.
Discussion s
This test was not performed during the return to power following the refueling outage because appropriate plant conditions could not be obtained withcut impacting plant operations.
Later in the cycle, the test was attempted when reactor pressure vas raised from Cycle l's operating value of 1020 psia to Cycle 2's operating value of 1040 psia; but, this evolution was performed too slowly to generate meaningful results. The performance of'this test does not warrant an impact of plant operation since, according to General Electric, the decreased pressure drop across the upper tie plate of the new fuel is expected to result in increased reactor stability. The performance of this test vill be reconsidered when appropriate plant conditions can be obtained without impacting plant operations.
This test was considered a physics test in USAR 14.2.12.2.18.
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d; ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.20.1 Test Number 23A - Feedvater System Test Objective The purpose of this test is to verify that the feedvater system is adjusted to provide acceptable reactor vater level control.
Dis'cussion DCP 88-305 was implemented during the first refuel outage to slow the (C34) feedvater system's response time. Before the change, the Perry Plant was capable of producing very rapid power changes (1%/second).
That capability was determined to be unnecessary and even overly sensitive to certain plant transients such as B33 flow control runback.
Experience with large power plant installations had shown that a lower power change rate (1%/ minute) was required.
E The response of the Turbine Driven Reactor Feedpumps "A" and "B" are rate.
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limited by DCP 88-305 to 5% per second. The Motor Feed Pump (MFP) has been limited to 15% per second because it has a much smaller capacity than the Turbine Driven Reactor Feedpumps.
No-level sensor, flov sensor or function generator was altered during the refuel' outage.
N27-023A.2 FV Low RX Level Controller Test The lov feedvater flow reactor level controls were not tested during the i
startup after the refuel outage. The lov flow control transmitter 1N27N0126 was respanned to 0-2000 gpm by a Vork Order. This satisfied a i
commitment. This VO is still open pending final tuneup of the control loop. It was decided not to conduct the lov flow test until new feedpump recirculation valves are installed.
N27-023A.3 FW Startup RX Level Controller Steps On August 5, 1989, the Turbine Driven Reactor Feedpump "A" was tuned (per l
PTI-C34-P0001, "Feedvater Control System Tuneup Procedure") on the "B" pump by inserting step inputs into the N27 system via step generators located in the C34 circuitry. The reactor power was approximately 75%.
The "B" pump was in master automatic control and "A" pump vss on its flow controller in manual.
The level response to step changes was satisfactory up to approximately a 7% change downward where flow to the vessel ceased and the minimum flow valve vent vide open.
The step changes that were input into the system were conditioned by the rate limiter IC34K669D such that a true step input could not be passed to the i
feedpump actuator. Review of the data shoved all acceptance criteria were met.
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PY-CEI/NRR-1093 L N27-023A.5 Master RX Level Controller On August 5, 1989, the Turbine Driven Reactor Feedpump "A" was tuned (per PTI-C34-P0001) on the Master Automatic Controller, IC34R600, with the reactor at 24% power.
Step inputs of $1, 3, and 6 inches were inserted in the system via~ level setpoint changes by the reactor operator.
The step inputs produced a stable, convergent, non oscillatory level response with very little overshoot.
No further adjustments were necessary.
Prior to transfer to the Master Automatic Control, adjustments were
.necessary on 1C34R601A01's reset such that'the reset frequency was reduced by 75%, namely 20 repeats per minute to 5 repeats per minute.
The reduction in reset was made to control the rate limiting that might otherwise occur due to DCP 88-305.
On August 10, 1989, both Turbine Driven Reactor Feedpumps "A' and "B" were placed in Master Automatic with the reactor at 100% pownr.
Step l
changes of 21, 3, and 6 inches were made to the reactor setpoint by the reactor operator. The response at 100% power was very stable end l'
non-oscillatory. The response, in fact, was better at 100% power than it was at 24% power.
l On August 18, 1987 with the reactor at 100% power, a minor adjustment was made to the gain on 1C34R601B01 so as to achieve the best available load sharing characteristics. The gain was reduced from about 1.4 to 1.1 and then step changes of 21, 3, and 6 inches were inserted into the reactor i
level setpoint per PTI-C34-P0001. The response was stable, convergent and nonoscillatory.
i Thus, the feedvater system was adjusted to provide acceptable reactor water level control.
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e' ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.20.3 Test Number 23C - Feedvater Pump Trip 1
Test Objective The purpose of this test is to demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedvater pump.
Discussica During the refuel outage the response rate of the automatic core flov runback feature was changed. The Recirculation Flow Control Valves
. runback rates vere sloved down from 11% position per second to 5.5%
position per second per DCP 88-305. This essentially reduced the runback rate.in half, t
Although not a retest requirement of DCP 88-305, performance of the Feedvater Pump Trip test is anticipated during the next planned shutdown.
During the original performance of this test large margins were exhibited.
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ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.20.4 Test Number 23D - Maximum Teedvater Runout capability Test Objective 1
The purpose of the test is to determine the maximum feedvater runout capability.
Discussion Doring the refueling outage, Reactor reed Pump b impeller was changed out and the high speed stops for both Reactor Teed Pump Turbines were raised back to their original design setpoints. The high speed stops had been lowered following the initial Startup Test Program to limit the runout capability of both Reactor Teed Pumps to vithin the originally analyzed USAR limits.
The reload licensing submittal for cycle 2 included results of analysis to support a change of the USAR runout capability limit to 143%.
SXI-0039, "Feedvater Runout capability - Retest Due to Impeller Replacement RTP-t' vas performed during the Cycle 2 power ascension testing to verify that the capability did not er:ceed the new limit.
The runout capability was measured to be 139.4%. -_
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ATTACHMENT 1 PY-CEI/NkR-1093 L 14.2.12.2.26.1 Test Number 29A - Recirculation Flov Control - Valve i
Position Control Test Objective I
i The purpose of this test is to demonstrate the recirculation flow control system's capability while in the loop flov manual mode.
Discussion Final flow control valve position testing was performed at 50% of rated reactor power with the Recirculation Pumps in FAST speed.
Testing was performed per PTI-833-P0001, Section 5.1, "FCV Position and Velocity l
controller Tuneup (Loop Flow Manual)."
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During the refueling outage, the Recirculation Flow Control System vas retuned to provide r. slover response rate than described in USAR Chapter 14 per DCP 88-305. The overall goal was to provide a slov, stable response from the Recirculation Flow Control System. Once the Flow Control Valve speeds had been changed, surveillance test SVI-833-71158, " Reactor Recirculation Control Valve Functional Test," vas satisfactorily completed, i
The transient response of any recirculation systcm related variable to any test input did not diverge.
Final Response Characteristics for a 5% valve position step changes
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Parameter DCP 88-305 Acceptance Criteria 1.
5% Step Open Maximum Rate of Change (Speed)
FCV A 1.5%/see 1.5% 2.5%/sec FCV B 1.4%/sec Delay Time FCV A 1.08 see f.65 sec(I)
TCV B
.78 see i
Rise Time FCV A 2.7 wee
$ 4 sec
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FCV B 2.79 see I
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5% Step close
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Maxieue Rate Of Change (Speed)
FCV A 5.2%/sec 5.5% t.5%/see FCV B 5.0%/sec i
Delay III FCV A
.37 see f.22 sec FCV B
.27 sec l
Rise Time TCV A
.71 sec
$.8 see FCV B
.80 sec Settle Time, himit Cycle, Decay Ratio, Duty cycle, Overshoot and scram avoidance were all negligible and thus not calculated.
NOTE (1): Due to the difference in valve speeds in the opening and closing directions it was not possible to tune to achieve both the delay and rise criteria and still achieve the valve speed goals. The slightly longer delay times (compared to the DCP 88-305 Acceptance criteria) were not noticable to the reactor operator and thus vere deemed acceptable.
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PY-CEI/NRR-1093 L 14.2.12.2.26.2 Test Number 29B - Recirculation Flov Control-Flov Loop Test Objective The purposes of this test areq 1.
To demonstrate the core flow system's control capability over the entire flow control range, including both flov loop and neutron flux loop modes of operation, and 2.
To determine that all electrical compensators and controllers are set for desired system performance and steH 11ty.
Discussion 1.
The transient response of any recirculation rystem related variable i
to any test input did not diverge.
2.
A partial system linearity test was performed on the 75% rod line with Recirculation Pumps in FAST speed.
This test was performed per PTI-B33-P0001, Section 5.6, " System Linearity Test." The Flow Control Valves were opued from 12% to 80% resulting in a core flow change from 53 to 109 M1b/hr. Overall the system had a linear response to changes in flow demand. The incremental gain was less than 1.25 to 1 (DCP 88-305).
Although the Control System provides a linent response, the Flow Demand to Valve Position Function Generators over-estimate valve position (thus actual drive flov) by about 3%.
From data collected from this test the function generators vill be recalibrated as plant conditions permit.
3.
A one point Flow Controller tuneup vas performed at 60% core flov vith the flow control valves at 25% optn. This test was performed per PTI-B33-P0001 Section 5.2, " Flow Centro 11ers Tuneup (Flux Manual)."
DCP 88-305 sloved down the expected response of the Flow Contro11ere.
Final Response Characteristic for a 5% Flow Step Change:
DCP 88-305 Parameter Acceptance Criteria 1.
5% Step Up Delay Time FCV A 1.87 sec
$.9 see FCV B 1.75 sec l
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FCV A 5.0 sec
$ 4.55 sec l
FCV B 4.7 sec l
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5% Step Down Delay Time FCV A
.9 sec
$.5 sec(I)
FCV B
.95 sec i
Rise Time FCV A 1.47 sec
$ 1.35 sec(I) j FCV B 1.75 see
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i Overshoot, Settle Time and Decay Ratio were all negligible and thus vere not calculated.
Scram avoidance and Flow Control Valve Duty were also very small thus vere not calculated.
Balance Times vere not calculated since loop flovs remained within 5% of each other during step changes.
NOTE (1): Due to differences in opening and closing speeds along vith a strong desire to have a slow tesponse rate from the 4
Flow controllers (reduce valve vear/ cycling resulting from Bi-Stable Flow) further tuning was not pursued even though l
the delay and response time criteria vere not met.
These minor deviations have a negligible effect on Flov Control operations.
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Flux Controller (Master Manual) tuning was performed at 2 different i
flov states 60% Core Flow /75% Pover and 80% Core Flov/90% Pover.
l Flux step changes vere made with the Flux Estimator Bypassed and with the Flux Estimator in Normal. Testing was conducted per PTI-B33-P0001, Section 5.3, " Flux Controller Tuneup (Master Manual)."
i DCP 88-305 also slowed down the expected response of the Flux Controller.
Final Response Characteristic for a 5% Flux Step Change:
Flux Estimator-Bypassed DCP 88-305 Parameter Acceptance Criteria 1.
5% Step Down Delay Time 60% Flov 1 see f 5 sec 80% Flov 1 see t
Rise Time 60% Flov 2 sec
$ 25 cee 80% Flov 1.5 see,
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60% Flov 1%
$ 2%
j B0% Flow 0%
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5% step Up I
Delay Time i
60% Flov 1.5 see f 5 sec i
80% Flov 2 sec l
Rise Time 60% Plov 6.5 see f 25 sec 80% Flov 4 sec Overshoot 60% Flov 0%
$ 2%
80% Flov 0%
Settle Time, Decay Ratio and Scram Avoidance vere all negligible and thus not calculated.
B.
Flux Estimator-Normal Farameter DCF 88-305 Acceptance Criteria 1.
5% Step Down t
Delay Time 60% Flow 1 see f 5 sec 80% Flov 1 see i
Rise Time 60% Flov 4.5 see f 25 sec 80% Flov 5 see j
Overshoot 60% Flov 2.2%
$ 2%(2) t 80% Flov 1.5%
F 2.
5% Step Down Delay i
60% Flov 1.5 sec
$ 5 see 80% Flov 1 see r
Rise Time t
I2) 60% Flov 6 see f 2.5 sec 80% Flov 9.55 see t
Overshoot 60% Flov 2.2%
$ 2%
80% Flov 0%
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l Settle Time, Decay Ratio and Scram Avoidance vere negligible and i
thus not calculated.
NOTE (2): With the Flux Estimator ir. Normal, the Overshoot was
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slightly greater than the acceptance criteria.
Flux j
Controller tuning was then stopped pending discussion with i
General Electric. General Electric has since recommended a tuning adjustment f or the Flux Estimator. When plant conditions allow, further Flux Controller tuning vill be performed.
Continued operation in Master Manual is administratively prohibited until further tuning and satisfactory response characteristics are obtained.
During testing, the Flux Estimator was observed not to be switching and as a result Flux Estimator svitching rates vere not measured.
5.
Master Controller tuning was not performed since Master Auto operations are not currently used at Perry.
t 6.
Automatic Flow Demand Limiter tuning vas performed with the reactor at 97% of rated thermal p)ver. This testing was performed per PTI-B33-P0001, Section 5.7. " Automatic Flow Demand Limiter Tuning and Testing." The acceptance criteria vere that the AFDL limit core flow by limiting flov Control Valve position with a smooth, stable response. For 2% changes in setpoint and flow demand, the AFDL exhibited a smooth stable response. Although there vas no acceptance criterion on delay time, the delay time in response to i
the step change was viewed as being excessive.
Investigations are being performed to determine the exact cause of the delayed response.
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ATTACHMENT 1 PY-CEI/NRR-1093 L 14.2.12.2.28 Test Number 31 - Loss of Turbine-Generator & Offsite i
Power Test Objective The purpose of this test is to determine the reactor transient performance during the loss of the main generator and all offsite power, and to demonstrate acceptable performance of the plant electrical supply system. Loss of offsite pover is maintained for sufficient time to demonstrate that necessary equipment, controls, and indications are available following loss of offsite power to remove decay heat from the core using only emergency pover supplies and distribution systems.
Discussion The Loss of Turbine-Generator & Offsite Power Test was not required to be performed during this refueling outage. The following surveillance tests vere performed to test 2ndividual Division LOOP and LOOP /1.0CA system responses:
SVI-R43-T1327, " Division 1 Standby Diesel Generator 18 Honth Functional Test" SVI-R43-T1328, " Division 2 Standby Diesel Generator IP Month Functional Test" SVI-R43-T5366, "LPCS/LPCI A Initiation and Loss of Off-site Power Time Response Test" SVI-R43-T5367, "LPCI B&C Initiations and Loss of Off-site Power Time Response Test" SVI-E22-T1329 " Division 3 HPCS Diesel Generator 18 Honth Functional Test" SVI-E22-T5397, "HPCS Initiation and Loss of Off-site power Time Response Test" These tests are performed every 18 months.
Each test was successfully performed in accordance with USAR or Technical Specification l
requirements.
SVI-R43-T1330, " Simultaneous Start of All Diesel l
Generators," vas not performed because this is a ten year surveillance.
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PY-CEI/NRR-1093 L 14.2.12.2.31 Test Number 35 - Recirculation System Flow Calibration i
Test Objective i
The purpose of this test is to perform complete calibration of the installed recirculation system flow instrumentation.
Discussion The jet pump flow instrumentation was checked at both low core flow and near rated core flow. As part of the NSS softvare verification via Computer Program / Modification Request 89-25 under PAP-0506, " Computer Access and Software Control," the measured core flow vas verified (at approximately 35% of rated core flow) against a core flov/ drive flow correlation established during Cycle 1.
The comparison shoved good agreement and no adjustment of the instrumentation was performed. At an indication of approximately 102% of rated core flow, the instrumentation vas calibrated because the measured core flow did not show good agreement with the previously established correlation. The gain adjustment factors
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for the summers were determined from the ERIS computer program.
le Following the adjustment of the gains on both jet pump loop summers, the gain adjustment factors vere recalculated by the ERIS computer program and were determined to be within 1%.
Date 08-AUG-89 09-AUG-89 Total Core Flow Recorder B33-R613 107 100 Computer Point B33NA001 107.5 100.5 Summer B33-K613 Measured 106.36 101.00 calculated 104.15 101.53 Gain Adjustment Factor 0.979 1.005 Jet Pump Loop Flow At Recorder B33-R612A
$4 50 Summer B33-K611A Measured 53.28 49.69 l
Calculated
$2.13 50.02 Gain Adjustment Factor 0.978 1.007 Jet Pump Loop Flov B Recorder B33-R612B 54 50 Summer B33-K611B Measured 53.32 51.48 t
Calculated 52.02 51.51 Gain Adjustment Factor 0.976 1.001 Except for the Gain Adjustment Factor, the values above are in M1b/hr.
l The electronics for the APRM flov-bias instrumentation vere calibrated in SVI-C51-T0030, "APRM Channel Calibration for IC51-K605," series and the flov transmitters vere calibrated in the SVI-C51-T0029 "APRM Flov Reference Transmitter IB33-N014 and IB33-N024 Calibration," series.
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1 14.2.12.2.37 Test Number 100 - Integrated HVAC j
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Test Objective y
i To demonstrate the ability of ventilation systems to maintain specified Unit 1 and common area temperatures and relative humidity within i
specified limits during plant operation.
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Discussion i
The Dryvell Area experienced temperature problems during Cycle 1.
The l
I maximum average dryvell temperature anticipated during design was 135'F.
This was also the original Technical Specification maximum allovable i
limit. The normal average dryvell temperature during Cycle 1 operation i
vas 133'F.
During the outage, it was discovered that local hot spots in the upper portions of the dryvell had damaged some cables [ Local Power Range Monitor (1.PRM) signal cables to Average Pover Range Monitor (APRM)
A and E, power cables to Intermediate Range Monitor (IRM) drive motors A, B, C F, G, and H, power cables to Source Range Monitor (SRM) drive motors A B, and C, limit switches for the under vessel platform] such that replacement was varranted.
Due to insufficient insulation in the upper dryvell region, heat radiated from the reactor vessel and damaged the cables discussed above. In order to correct these temperature problems additional insulation was installed and some minor steam leaks vere repaired. Also, TXI-0088, " Upper Dryvell Temperature Monitoring" vas written to monitor specific equipment to ensure Environmental Qualification limits are not violated and to provide data to Perry's Nuclear Engineering Department to initiate corrective actione/ modifications as required per licensing commitments. The nornal average dryvell temperature during cycle 2 operation was measured to be 123'F.
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ATTACHMENT 1 PY-CE1/NRR-1093 L Table 3.2-1 -- USAR 14.2.12.2 tests not impacted by refueling activities 14.2.12.2.1 Test Number 1 - Chemical and Radiochemical 14.2.12.2.13 Test Number 16A - Selected Process Temperatures 14.2.12.2.13.1 Test Number 16B - Vater Level Reference Leg Temperature 14.2.12.2.14 Test Number 17 - System Expansion 14.2.12.2.19 Test Number 22 - Pressure Regulator 14.2.12.2.20.2 Test Number 23B - Loss of Feedvater Heating 14.2.12.2.21 Test Number 24 - Turbine Valve Surveillance 14.2.12.2.22.1 Test Number 2bA - Main Steam Isolation Valses Function Tests 14.2.12.2.22.2 Test Number 25B - Full Reactor Isolation 14.2.12.2.22.3 Test Number 250 - Main Steamline Flov Venturi calibration 14.2.12.2.23 Test Number 26 - Relief Valves 14.2.12.2.24 Test Number 27 - Turbine Trip and Generator Load Rejection i -
14.2.12.2.25 Test Number 28 - Shutdown From outside the Control Room 14.2.12.2.27.1 Test Number 30A - One Pump Trip 14.2.12.2.27.2 Test Number 30B - RPT Trip of Two Pumps 14.2.12.2.27.3 Test Number 300 - System Performance 14.2.12.2.27.4 Test Number 30D - Test Deleted 14.2.12.2.27.5 Test Number 30E - Recirculation System Cavitatien 14.2.12.2.29 Test Number 33 - Dryvell Piping Vibration 14.2.12.2.30 Test Number 34A - Vibration Hessurement 14.2.12.2.32 Test Number 36 - Isolated Reactor Stability 14.2.12.2.33 Test Number 70 - Reactor Vater Cleanup System 14.2.12.2.34 Test Number 71 - Residual Heat Removal System 14.2.12.2.35 Test Number 74 - Offgas System 14.2.12.2.36 Test Number 99 - ERIS 14.2.12.2.38 Test Number 113 - Service Vater System 14.2.12.2.39 Test Number 114 - Emergency Closed Cooling System 14.2.12.2.40 Test Number 115 - Nuclear Closed Cooling System 14.2.12.2.41 Test Number 116 - Turbine Building Closed Cooling System 14.2.12.2.42 Test Number 117 - Emergency Service Vater 14.2.12.2.43 Test Number 118 - Circulating Vater Systet Suppression Pool Cleanup System 14.2.12.2.44 Test Number 119 14.2.12.2.45 Test Number 120 - Teedvater System 14.2.12.2.46 Test Number 121 - Extraction Steam System 14.2.12.2.47 Test Number 122 - BOP Piping Expansion and Vibration 14.2.12.2.48 Test Number 123 - Concrete Temperature Survey 14.2.12.2.49 Test Number 124 - Main and Reheat Steam System 14.2.12.2.50 Test Number 125 - Condensate System 14.2.12.2.51 Test Number 126 - Main, Reheat Extraction and Misc. Drains 14.2.12.2.52 Test Number 127 - LP/HP Heater Drains and Vents 14.2.12.2.53 Test Number 128 - Condensate Demineralizer System 14.2.12.2.54 Test Number 129 - Steam Seal System 14.2.12.2.55 Test Number 130 - Condenser Air Removal System 14.2.12.2.56 Test Number 131 - Offgas Vault Refrigeration System 14.2.12.2.57 Test Number 132 - Turbine Plant Sampling 14.2.12.2.58 Test Number 133 - Locse Parts Monitoring System 14.2.12.2.39 Test Number 134 - Equipment Area Cooling