ML20198N201
| ML20198N201 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/12/1998 |
| From: | Bordley P, Moffitt S, Orogvany C CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20198N182 | List: |
| References | |
| NUDOCS 9801200291 | |
| Download: ML20198N201 (20) | |
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PY CEI/NRR 2247L
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Page I of 20 i
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i TIIE CLEVELAND ELECTRIC ILLUMINA, TING COMPANY
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PERitY NUCLEAR POWER I,LAN's UNIT 1 POST-REFUELING STAIG11P TEST S11MMARY REPORT CYCLE 7 e
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Reviewed By:
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Attachment i PY CEl/NRR-2247L
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1.1 INTRODUCTION
Thldeport presents a summary of the results from the post refueling startup tests which were conducted in preparation for Cycle 7 at Unit I of the Perry Nuclear Power Plant. This report is submitted pursuant to Operational Requirements Manual Administrative Requirement 7.6.1.3.
r 1.2 PLANT DESCRIPTION The Perry Nuclear Power Plant is operated by the Cleveland Electric illuminating Company (CEI) and is located near Lake Eric in Lake County, Ohio. Unit I has a iloiling Water Reactor (llWR) nuclear steam supply system as designed and supplied by the General Electric Company (GE) and designated IQ.R6, with a Mark 111 containment. The balance of plant was designed by Gilbert Associates, ine., Reading, Pennsylvania, as architect engineer.
The rated core thermal power is 3579 MWt with a gross electrical output of 1250 MWe The
- turbine is an 180 rpm tan em compoun, s xi ilow, reheat unit consisting of one double flow 0
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high pressure stage in tandem with three low pressure stages. The generator is a direct coupled 6011z,22 KV three-phase unit with a water cooled stator and hydrogen cooled rotor.
1.3
SUMMARY
OF ACTIVillES DURING REFUELING OUTAGE 6
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The sixth refueling outage at Unit I of the Perry Nuclear Power Plant began September 12,1997 (main generator off line) and was completed October 23,1997 (main generator on-line). The outage duration was 41 days,22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />,51 minutes.
Key activities of this refueling outage were: the performance of a core shuffle, replacement of 252 fuel bundles (out of 748), the completion of a large number of corrective and preventive maintenance activities, and the performance of required Technical Specification surveillance tests. A major design change was the installation and satisfactory testing of a low approach velocity Emergency Core Cooling System suction strainer which satisfied the issues of NRC llulletin 96 03.
Start Date September 12,1997 Stop Date October 23,1997 Duration 41 days Work Orders 1639 Design Changes 37 Surveillances 663 Repetitive Tasks 623
Attachment i PY-CEl/NRR-2247L Page 3 of 20 2.1 PliSCRIPTION OF DlFFERENCE IN FUEL DESIGNS Unit k at the Perry Nuclear Power Plant used the following General Electric fuel designs fer cycle 6:
GE8B-P8SQB320-9GZ-120M-150-T 16
- bundles, GE8'3-P8SQB322-7GZ-120M-150-T 24
- bundles, CE10-P8SXB306-10GZ2-120M-150-T 136
- bundles, GE10-P8SXB306-!1GZ3-120M 150-T 336
- bundles, Gell-P9SUB338-10GZ-120T-146-T 152
- bundles, gel 1-P9SUB338 120Z-120T-146-T 84 bundles.
and the following General Electric fuel designs for cycle 7:
g' GE10-P8SXB306-10GZ2-120M-150-T 24
- bundles, GE10-PSSXB306-1 IGZ3 120M-150-T 236
- bundles, gel 1-P9SUB33810GZ-120T-146-T 152
- bundles, 1
gel 1-P9SUB338-12GZ-120T-146-T 84 bundles.
GE12-P10SSB369-14GZ-120T-150-T 92 bundles.
GE12-P10SSB369-12GZ-120T-150-T 160
- bundles, Complete discriptions of these General Electric fuel designs are given in GESTAR 11 General Electric Standard Applicatic,n for Reload Fuel. The major difference in the GE12 fuel relative to the GElI fuel is an increase in the number of fuel rods per btindle (92 for GE12 vice 74 lbr gel 1, arranged in a 10x10 array rather than a 9x9 anay).
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l Attachment t PY-CEl/NRR 2247L Page 4 of 20 3.1 POST-REFUELING STARTUP TEST PROGRAM
' During refueling operations and the subsegient return to power, activities were controlled under normal administrative programs rather thu. sepante, formally defined, post-refueling stanup test program. These administrative programs cover areas of normal operation such as:
Design Changes / Post-modification Testing Post-maintenance Testing Technical Specification Surveillances inservice Inspection U
Special Nuclear Material Control Periodie and Special Tests
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Computer Soltware Modification Radiation Control l
The acceptance criteria for these tests were de.ved from the requirements of the administrative programs. The reactor conditicas for conducting the tests were guided by requirements in the appropriate administrative program.
E 3.2 POST-REFUELING STARTUP TEST REPORTS As required by Operational Requirement 7.6.1.2, this report add esses each of the startup tests identified in USAR Subsection 14.2.12.2. Each test was evaluated by the Responsible System E
Engineer who determined whether the test was impacted by any refueling activity. Those tests F:
determined not to be impacted by any refueling activity are listed in Table 3.2-1.
b-For those tests which were impacted by refueling activities, this report lists:
gs 1.
A description of the measured values of the operating conditions or characteristics obtained during the test program end a comparison of these values with design predictions and specifications; 2.
A description of any corrective actions that were required to obtain sttisfactory operation; 3.
Ar.y additional specific details required in license conditions based on other commitments.
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nttachment I PY-CEl/NRR 2247L PJge 5 of 20
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3.3 I'OST-REFUELING STARTUP TEST REPORTS -- SUMMARIES 14.2.1 (2.3: ' Test Number 3 - Fuel Loading Test Objective :
p The purpose of this test is to load fuel safely and efficiently.
Discussion F.iel unlor. ding and loading was conducted in MODE 5 under 101-9," Refueling " Fuel -
r.iovement followed a predetermined plan in accordance with a Fuel Movement Checklist in accordance with FTI D39,"Use of the truel Movement Checklist." A core shuffle (vice an
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ofiload/ reload) was performed to minimize fuel movement.
i Fuel movement was pe formed in three phases. First, fuel was ofDoaded from selected control
- cells in order to preps e for performance ofin Vessel Visuai inspection (IVVI) of six source range monitor / intermediate range monitor dry tubes, and to reduce the reactor shutdown margin.
Following the IVVI, the core shuffle was then started. The core shuffle consisted ofintegrated moves which removed discharged fuel from the core, inserted new fuel, and rearranged reload fuhl within the reactor vessel. All fuel moves involving the reacter were analyzed for shutdown
- margin, within a minimum analytical margin of 1.00% Ak/k. This enstired that the acceptance criteria of at 1:ast 0.38% Ak/k was met for all core configurations.
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Attachment l' PY CEl/NRR 22471T Page 6'of 20_
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' 14.2.i2.2l1 !- Test Number 4 Full Core Shutdown Margin -
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Tht Ohlective -
?The purpose of:his test is first, to demonstrate the reactor is suberitical throughout the fuel cycle f
with any single control rod fully withdrawn and second, to determme quantitatively the '
j shutdown margin of the as-loaded core.
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- - DISCUSSIOM t
- Full core shutdown' margin and reactivity anomaly were deme,nstrated to be within their.
. Technical Specification requirements during Cycle 7 startup (reactor startup Number 79).
SVI-Ill3 T0001. " Shutdown Margin Calculation," was performed at 0% power and 160 *F
- moderator temperature. The minimum shutdown margin for Cycle 7 was measured to be g
- 0.92% Ak/k, which exceeded the Technical Specification minimum of 0.38% Ak/k. The
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measured reactivity anomaly _ calculated in accordance with SVI-Ill3-T0004," Reactivity -
- Anomaly Calculation During MODE 1" was less than the Technical Specification maximum of i1.0% Ak/k.
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PY-CEl/NRR 2247L
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Page 7 of 20
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14.2.12.2.5 Test Number 5 - Control Rod Drive System
' Test obiective The purposes of the control rod drive system tests 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 deteimine the initial op-rating characteristics of the entire CRD system.
Discussion Five control rods were friction tested per PTI-Cl1-P0003," Control Rod Friction Tcsting." The absence of control rod blade interference for the remaining 172 control rods was confirmed by an evaluation of the results of the control rod scram timing tests.
Control rod scram timing was performed for al! control rods in acco: lance with SVI-Cil-T1006,
" Control Rod Maximum Scram insertion Tim:." Control rod maximum scram insertion times were determined in accordance with Technical Specification SR 3.1.4.3 for one control roc' whose hydrculic control unit had maintenance performed during the outage which may have impacted scram times. This testing was completed at a reae.ar pressure of 0 psig. Scram timing was perfonned on all rods in the core in accoidance with Technical Specification SR 3.1.4.1 during the initial reactor staitup prior to exceeding 40% of rated thermal power. Buffer times (the slowing down time at position 00 during a scram) were not measured because Perry has not experienced a control rod drive with an unacceptable buffer time in previous te-ts.
All control rod scram insertion times were acceptable, i.e., not considered slow as described in Note i for Technical Specification Table 3.1.4-L No other hydraulic testing was performed on the control rod drive system as there s to changes to the system that would have affected these parameters.
The timing tests were considered physics tests in USAR 14.2.12.2.5.
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Attachment I PY-CEl/NRiv2247L Page 8 of 20
- 14.2.12.2.6 Test Number 6 - SRM Performance and Control Rod Sequence Test 3bjective 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 four SRMs indicated greater than 0.7 cps with a signal to noise ratio greater than two to one for approach to criticality in accordance with 101-1," Cold Startup."
This test was considered a physics test in USAR 14.2.12.2.6.
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i PY-CEl/NRR 2247L Page of 20 14.2.12.2.7-Test Number 8 - Rod Sequence Exchange Test bbjective The purpose of this test is to perform a representative sequence exchange of control rod patterns at a significant power level.
- Discussion llecause rod sequence exchanges are performed only approximately every 1400 MWD /ST. no control rod sequence exchange was performed. Nothis:g in the design of the GE-12 fuel is -
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projected to impact the ability to perform a sequence exchange
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Attachment i l'Y-CEl/NRR 2247L Page 10 of X
-14.2.12.2.8 Test Number 10 - Intermediate Range Monitor Performance Test Objective The purpose of this test is to adjust the Intermediate Range Monitor (IRM) system to obtain an optimum overlap with the Source Range Monitor (SRM) and Average Power Range Monitor
- (APRM) systems.
Discussion Two IRM detectors (IRMs "E and "II") were replaced in RF06. During the reactor startup at the beginning of Cycle 7, IRM "E" failed to meet the half decade SRM/lRM overlap in accordance with 101-1," Cold Startup." IRM "E" was subsequently adjusted per work order 97-185 to optimize the IRM/APRM overlap. The remaining IRMs were confirmed to have a half decade overlap with both the SRMs and APRMs during the performance oflOI-1," Cold :
Startup."
This test was considered a physics test in USAR 14.2.12.2.8.
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Attactment 1 PY-CEl/NRR 2247L Page iI of 20 14.2.12.2.9 Test Number 11 - LPRM Calibration Tes't Objectivs The purpose of this test is to calibrate the Local Power Range Monitoring (LPRM) system and to verify the LPRM flux response.
Discussion LPRM llux response was verified with the reactor power between 8% and 60% of rated power in accordance with PTI-C51-P0001," Verification of Proper LPRM Connection." The testing procedure involves moving an adjacent control rod past each LPRM and observing the appropriate change (sigrofe ant in magnitude and correct in direction) in the LPRM reading.
Two I PRMs did not respond; they were bypassed and work requests were initiated. Five other LPRMs were bypassed prior to the initial reactor startup and one was bypassed during the startup. The remaining 156 LPRMs responded satisfactorily.
Near 60% reactor power, Process Computer Program ODI,"Whole-Core LPRM Calibration and Base Distribution," was performed. Based on this program, the LPRM readings are internally adjusted by the Process Computer to be used in the core power distribution and thermal limits calculation. At this time, the gain adjustment factors (GAFs) were not manually adjusted. Although the GAF array ranged between 0.8 and 1.2 (outside the desired 10% band).
the LPRMs outside the desired range were not conridered inoperable becc.se the Process Ccmputer and Average Power Rany Manitor (APRM) calibration correct for any variation. In addition, the GAFs were not adjusted because of the non-steady state power conditions and non-equilibrium xenon conditions.
SVI-C51-T5351,"LPRM Calibration," was performed to manually adjust the LPRM GAFs on November 20.1997. This delay was r.cceptable since the last performance of SVI-C51 T5351 in Cycic 6 was still ctaent.
This test was considered a physics test in USAR 14.2.12.2.9.
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Attachment i I'Y-CEl/NRR 2247L
' l' age 12 of 20 14.1.12.2.10 Test Number 12 - APRM Calibration Test Obiective The purpose ot'this test is to calibrate the Average Power Range Monitor (APRM) system.
. Discussion During startup following refuelint, the APRMs were calibrated several 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 at the identified power levels.
% Rated Dalc
.10 2 10/24/97 29-10/25/97 30 10/27/97 29,-52 10/28/97 85,95 10/29/97 100-Technical Specification and fuel warranty limits on APRM serr.m and rod block were not -
exceeded. The startup APRM scram functions were checked in the SVI-C51-T0030 series, "APRM Channel Calibration."
This test was considered a physics test in USAR 14.2.12.2.10.
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Attachment I PY-CEl/NRR 237L Page.13 of 20 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 Computer and on-line NSSS computer programs under plant operating conditions.
Discussion No undervessel Traversing in-Core Probe (TIP) tubing was removed during the refuel outage.
The ' LIP alignment was checked per ICl-C-C51-5, "TIP System Mechanical Drive System Calibration."
Ten LPRM detectors were replaced and the initial gains were set in acenrdance with work order 96-3367.
The NSSS software fbr Cycle 7 was updated and tested via Computer Program / Modification Request C91-97-28 in accordance with PAP-0506," Computer Access and Sof1 ware Control."
The Core Monitor calculations for thermal limits based on non-adaptive predictions (based exclusively on the vendor's code) were compared to those from an adaptive case (based on actual on-line data) to establish the NSSS sof1 ware's acceptability for Technical Specification application. The GE program Backup Core Thermal Limits Evaluation (BUCLE) is no longer used as the backup thermal limit calculation method for Perry, therefore. no BUCLE results were compared during the startup tests for Cycle 7. The LPRM calibration factors calculated by ODI were verified by manual calculations.
In preparation for the above comparisons, the input data and intermediate calculations wete evaluated. At the end of Cycle 6, records were generated for the NSSS database for fuel constants, exposure and isotopics, LPRM exposures, and control rod exposures. While the plant was shutdown ihr reactor refueling, the NSSS databass was updated to reflect the new core design and a statistically significant number of entries in selected arrays were checked for reasonableness. During the reactor startup f:om 0% to 25% of rated thermal power, the process computer values for < ore thermal power, control rod positions, LPRM readings, and thermai hydraulic parameters were compared against readings from appropriate plant sensors After validation of the core t!ennal power calculation, the Start / Restart program was initiated to start the Core Energy Increment program. Reactor power was increased and the Whole-Core L PRM Calibration and Base Distribution program (ODI) was performed to calibrate the LPIWs for the a
process computer. The process computer calculations for TIP normalization factor, LPRM calibration constants, and LPRM substitute values were verified against appropriate manual calculations. Prior to actually physically adjusting the amplifier gains on the LPRMs, the power, time, flow, and run flags of the Core Monitor program were checked for reasonableness.
The control rod exposure. generator and core energy accumulation features of the Core Monitor program were checked. At this point, all programs were considered operationa!.
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Attachment i PY-CEl/NRR-2247L Page 14 of 20 14.2.12.2.15 Test Number 18 - Core Power Distribution.
Test Objective The purpose of this test is to determine the xproducibility of the Traversing In-Core Probe (TIP) system readings.
- Discussion The licensing basis TIP uncertainty for the MCPR Safety Limit is 8.7% (standard deviation).
' This value is based on a model uncertamty of 4.6%, an LPRM uncertainty of 3 4%, and a TIP measurement uncertainty of 6.6%. He TIP measurement uncertainty is based on 6.4% ge netry uncertainty and_l.2% random noise u.' certainty. Referenev NEDE-31152P," General Electric Fuel Ilundle Designs," and letter PY-GLN/CEl-2961 dated 9/6/89 from N. R.13arker (GE) to K. R. Pech (CE.I).
FTI-A16," Tota! TIP Uncertainty," was perfomied in an octant symmetric control rod patterc at approximately 100% of rated thermal power to determine the TIP measurement uncertainty. li.e TIP measurement uncertainty was determined to be 1.2%. A comparison to the 6.6% value for TIP measurement uncertainty assumed in the licensing basis showed the TIP reading reproducibility to be acceptable.
This test was considered a physics test in USAR 14.2.12.2.15.
7 Mtachment i PY.CEl/NRR 2247L Page 15 of 20 14.2.12.2.16 Test Number 19 - Core Performance
. Test' Objective Tlic purposes of this test are to evaluate the core thennal power and core flow and to evaluate the following core perfonnance parameters:
1.
Maximum Linear IIcat Generation Rate (MLHGR).
2.
Minimum Critical Power Ratio (MCPR).
3.
Maximum Average Planar Linear lleat Generation Rate (MAPLllGR).
N Discussien The fuel thennal limits were monitored on a once per shift basis under the Tecimical Specification Rounds Instruction. The core thermal power was continuously monitored and controlled under IOI-3. " Power Changes." During post refueling testing, there were no fuel thermal limit violations and no ;teady-state operation above the more limiting of rated thermai power or the bounding licensed load line. The following lists the average thermal power (CMWTA) and the most limiting values for each of the thennal limits for 15 days following reactor startup that reactor power was above 25% of rated thermal power during the day.
CMWTA Date (MW!)
MFLCPR MFLPD MAPRAT 10/25/97 1064 0.718 0.379 0.580 10/26/97 607 0.711 0.372 0.569 10/27/97 1432 0.779 0.700 0.861 10/28/97 2856 0.869 0.9C0 0.942 10/29/97 3318 0.921 0.968 0.924 10/30/97 3543 OS38 0.929 0.881 10/31/97 3576 0.911 0.930 0.880 11/01/97 3576 0.910 0.930 0.878
!l/02/97 3576 0.913 0.930 0.877 11/03/97 3576 0.912 0.931 0.877 1I/04/97 3577 0.895 0.901 0.852 11/05/97 3577 0.892 0.R99 0.850 11/06/97 3578 0.887 0.897 0.850 11/07/97 3578 0.886 0.894 0.848 11/08/97 3577 0.885 0.895 0.848 11/09/97-3577 0.833 0.894 0.848 This test was considered a physics test in USAR 14.2.12.2.15.
Attachment I PY-CEl/NiiR 2247L Page 16 of 20 14.2.12.2.18 Test Number 21 - Core Power-Void Mode Rdsponse Test'Obiective 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.
Diccussion This test was not performed during the return to power since the low pressure drop spacers used in the new GE12 fuel increases reactor stability. This fuel design maintains stability equivalent to the GE8 fuel design.
This test was considered a physics test in USAR 14.2.12.2.18.
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Attachmcnt i PY-CEl/NRR 2247L l'
Page 17 of 20
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14,2.12.2.31 Test Number 35 - Recirculation System Flow Calibration
. Test'Obiective 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 stead a te conditions with core flow a
indicative of an operating core flow for the cycle. As part c. the NSSS software verification via Computer Program Mooiticat:an Request C91-97-28 under PAP-0506,"Co.nputer Access and Software Control," the measured core flow was verified against an established core flow / drive flow correlation. The comparison indicated good agreement and no adjustment of the instrumentation was performed.
At approximately 97% of rated core flow, FTI-A13. " Core Flow Calibration," was perfonned and the instrumentation was calibrated because the measured core flow did not show good agreement with the previously established correlation as required by FTI-A13," Core Flow Calibration."
The gain adjustment factors for the summers were determined to be within 1%.
Date:
11/19/97 Jet Pump Loop Flow A:
Recorder B33-R612A 49 Summer B33-K611 A Measured 48.47 Calibrated 48.97 Gain Adjustment Factor 1.010 Jet Pump Loop Flow B:
Recorder B33-R612B 49 Summer ll33-K61111 Measured 48.76 Calibrated 48.40 Gain Adjustment Factor 0.993 Total Core Flow:
- Recorder Il33-R613 95 Summer
[133-K612 Measured 96.83 Calibrated 97.37 Gain Adjustment Factor 1.006
. Except for the Gain Adjustment Factor, the values above are in Mlb/hr.
Atta:hment 1 PY-CEl/NRR 2247L Page 18 of 20 14.2.12.2.31 Test Number 35 - Recirculation System Flow Calibration (continued) e The electronics for the APRM flow-bias instrumemation were calibrated in SVI-C51-T0030, "APRM Channel Calibration for IC51-K605," series and the flow transmitters were calibrated in the SVI-C51-T0029,"APRM Flow Reference Transmitter IB33-N014 and IB33-N024 Calibration," series. The flow biased APRM system was verified to be within the established core flow / drive flow correlation in SVI-B33-Tl160," Jet Pump Operability."
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PY-CEl/NRR-2247L Page 19 of 20
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, Table 3.2-L The following tests from USAR 14.2.12.2 were evaluated by the Responsible System -
Engineers and determined not to have been impacted by any refueling activities.
14.2.12.2.1 Test Number 1
Chemical and Radiochemical 14.2.12.2.2 Test Number 2 - Radiation Monitoring 14.2.12.2.12-Test Number 14 RCIC System 14.2.12.2.13 Test Number 16A - Selected Process Temperatures 14.2.12.2.13.1 Test Number 16B - Water Level Reference Leg Temperature System Expansion 14.2.12.2.14 Test Number 17 14.2.12.2.17 Test Number 20 - Steam Productir a Startup Test Pressure Regulator l_4.2.12.2.19 Test Number 22 14.2,12.2.20.1 Test Number 23A - Feedwater Control System 14.2.12.2.20.2 Test Number 23B - Loss ofFeedwater Heating 14.2.12.2.20.3 Test Number 23C - Feedwater Pump Trip 14.2.12.2.20.4 Test Number 23D - Maximum Feedwater Runout Capability 14.2.12.2.21 Test Number 24 Turbine Valve Surveillance 14.2.12.2.22.1 Test Number 25A - Main Steam Isolatica Valves Function Tests
.14.2 12.2.22.2 Test Number 25B - Full ReactorIsolation 14.2.12.2.22.3 Test Number 25C - Main Steamline Flow Venturi Calibration 14.2.12.2.23 Test Number 26 ReliefValves 14.2.12.2.24 Test Number 27 - Turbine Trip and Generator Load Rejection Shutdown From Outside the Control Room 14.2.12.2.25 Test Number 28 14.2.12 2.26.1 Test Number 29A - Recirculation Flow Control-Valve Positica Control 14.2.12.2.26.2 Test Number 29B - Recirculacion Flow Control-Flow Loop 14.2.12.2.27.1 Test Number 30A - One Pump Trip 14.2.12.2.27.2 Test Number 30B - RPTTrip ofTwo Pumps 14.2.12.2.27.3 Tee Number 30C - System Performance 14.2.12.2.27.4 Test Number 30D - Test Deleted 14.2.12.2,27.5 Test Number 30E - Recirculation System Cavitation 14.2.12.2.28 Test Number 31
- Loss of Turbine-Generator and Offsite Power 14.2.12.2.29 Test Number 33
- Drywell Piping Vibration 14.2.12.2.30 Test Number 34 - Vibration Measurement 14.2.12.2.32 Test Number 36 - Test Deleted Reactor Water Cleanup System 14.2.12.2.33 Test Number 70 14.2.12.2.34 Test Number 71 - Residual Heat Removal System 14.2.12.2.35 Test Ntimber 74 - OITgas System 14.2.12.2.36 Test Number 99 - Emergency Response Information System 14.2.12.2.37 Test Number - 100 - Integraicd HVAC 14.2.12.2.38 Test Number - 113 - Service Water System 14.2.12.2.39 Test Number 114 - Emergency Closed Cooling System 14.2.12.2.40 Test Number i15 - Nuclear Closed Cooling System 14.2.12.2.41-Test Number 116 - Turbine Bui! ding Closed Cooling System
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- PY-CEl/NRR 2247L; 4-Page 20 of 20 ;
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- Table 3.2-1 (continuedi i
14.2.12.2.42-Test Number I17_- Emergency Service Water 14.2.12.2.43-Test Number.118 ~- Circulating kater system
- 14.2.12.2.44 Test Number-119 Suppression Poo! Cleanup System
- 14.2.12.2.45
. Test Number. = 120 - Feedwater System -
14.2.12.2.46- - Test -Number 121~- Extraction Steam System.
14.2.12.7.47 Test Number 122 - -' _DOP Piping Expansion and Vibration
-14.2.12.2.48:
Test Number.1.13... 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
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14.2.12.2.51 Test Number 126 - Main, Reheat Extraction and Miscellaneous -
j Drains 14.2.12.2.52.-
Test Number 127 - LP/HP lleater 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 OITgas Vault Refrigeration System-
- 14.2.12.2.57 Test Number 132 - Turbine Plant Sampling -
133 ' Loose Parts Monitoring System
- 14.2.12.2.58 Test Number-
. 14.2.12.2.59
- Test Number 134 - Equipment Area Coo!ing F
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