Text

Startup Rept,Supplemental - 1
	ML20214S388
Person / Time
Site:	Perry
Issue date:	06/03/1987
From:	Cantlin J, Edelman M; CLEVELAND ELECTRIC ILLUMINATING CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	PY-CEI-NRR-0661, PY-CEI-NRR-661, NUDOCS 8706090251
	Download: ML20214S388 (150)
	v • d • e

. - _ . _ - _ . . - . - .- . . . _ _ . _-

. > \

r ,

4 The Cleveland Electric illuminating Company r ,

STARTUP REPORT Supplemental - 1 4

i

.a

,,T-~ tT: ,. , , -

W ' ' z.

-4 s . . , . ..

' ' ~

,T.=

, ,yP{T?' H ...s

. .:f:,((* ,

i

.: i

..,3

a an

(

PERRY NUCLEAR POWER PLANT L

UNIT 1 J

l \ [

l B706090251 870603 PDR ADOCK 05000440 P pop

The Cleveland Electric Illuminating .

Compang Perry Nuclear Power Plant Unit 1 STARTEP MPORT Supplemental - 1 0/b, 4k 0.afhdI ( =)-f'7 Startpp Test Element Supervisor Data khW 4tu 6 -:- 37 Startup Test Program Director Data c he 6-247 Perry PlagTechnical Department Manager Date LLAA 11 Y7 Perry PJ as .t;erati a D partment Manager ' Da'te

$/dl, . /w 6-5-P7 Vice Presid t,' Nuclear Operations Division Date ht 6l1lD Senior V se President Nuclear Group Date sej'

FOl6&RD this Supplemental Startup Report (Supplemental-1) for the Perry Nuclear Power Plant, thit 1, covers the period from

, March 1, 1987 to May 15, 1987, and is subnitted as required by Regulatory Guide 1.16 and Perry Technical Specifications section 6.9.

i

,c.

l TABLE OF COtCENTS Title Pace SECTICN 1 - DESCRIPTICH 1.1 Introduction 2 1

SECTICN 2 -

SUMMARY

2.1 Summary of Testing Since March 1, 1987 4 2.2 Startup Test Program Chronology 5 SECTICU 3 - STARWP TEST INSTRUCTIONS No Change From Initial Report SECTION 4 - STARWP TEST RESULTS 4.1 P35-001 Chemical and Radiochemical 11 4.2 D21-002 Radiation Measurements 13 4.3 C11-005 Control Rod Drive System 15 4.4 c51-10 IRM Performance 18 4.5 C51-011 LPRM Calibration 19 4.6 C51-012 APRM Calibration 21 4.7 C91-013 Process Computer 24 4.8 E51-014 Reactor Core Isolation Cooling 49 4.9 C91-019 Core Performance 55 4.10 C85-022 Pressure Regulator System 57 4.11 N27-023A Feedwater Control System 63 4.12 B21-026 Safety Relief Valves 73 4.13 B21-027 Generator Load Reject Within 77 Bypass Valve Capacity 4.14 C61-028 Shutdown from Outside the 79 Control Room 4.15 B33-029A Recirculation Flow Control-Valve 81 Position Loop 4.16 B33-030E Recirculation System Cavitation 84 4.17 R43-031 Loss of Turbine-Generator and 85 '

Offsite Power 4.18 B21-033/P99-122 Drywell Piping Vibration / BOP 91 Piping Expansion and Vibration 4.19 E12-071 Residual Heat Removal System 102 4.20 C95-099 Emergency Response and 116 Information System 4.21 M99-100 Integrated HVAC 118 4.22 P42-114 Emergency Closed Cooling System 119 4.23 P45-117 Emergency Service Water System 122 ii

TABLE OF CONTENTS Title Page SECTION 4 - STARTJP TEST RESULTS (OONr.)

4.24 G42-119 Suppression Pool Cleanup System 124 4.25 N27-120 Feedwater System 125 4.26 T23-123 Concrete Temperature Survey 129 at Low Power 4.27 R63-133 Loose Parts Monitoring System 130 (LPMS) Baseline Data 4.28 M99-134 Equipment Area Cooling. 131' SECTION 5 - ATTACHMENTS Attachment 1 Startup Test Performance Log 136 for all Testing Through Test Condition 2 iii

SECTION 1 DESCRIPTIO1 1

1.1 INIRODUCTICH ,

This report consists of a summary of the Startup Test Program performed at Unit 1 of the Perry Nuclear Power Plant. This supplemental report covers the period March 1, 1987 to May 15, 1987. During this period, Test Condition 1 was completed, a brief post-TC-1 maintenance outage was performed, 'IC-2 testing completed and TC-3 has been entered.

Since the Startup Test Program is still in progress, Supplemental Startup Reports will be submitted at least every three months until the program is completed.

Previous Startup Report submittals are listed below:

Report Period Covered Initial Startup Report June 6, 1986 to March 1, 1987 2

SECTION 2 SU.?ARY 3

2.1

SUMMARY

OF TESTING 'IO DATE Since the submittal of the initial Startup Report which covered up to March 1,1987, the following significant activities have been successfully completed within the scope of the Startup Test Program:

1) All Test Condition 1 (K-1) testing was completed with the exception of 10 tests which were rescheduled to Test condition 2 (K-2).

2) A Post TC-1 outage, to perform minor repairs and modifications to various systems and components, was completed.

3) K-2 was completed.

4) TC-3 was started.

4

2.2 STAR'IUP TESTING PROGRAM CHRONOLOGY 3-01-87 Continued testing in Test Condition 1 3-02-87 SCRAM 87-02-Cause: Loss of Feedwater due to a personnel error while trying to troubleshoot the hot surge tank high level indication switch N21-N337.

3-04-87 After recovering from the Scram, criticality was achieved and testing was resumed. Testing in this period included pres-sure regulator tests, feedwater low flow valve testing with the turbine feed pumps and computer room data as well as two process computer (DSTC) tests. Plant shutdown commenced on March 6 at 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months due to high steam tunnel temperatures caused by steam leakage.

3-07-87 Corrective actions were taken during this shutdown period to repair feedwater minimum flow and check valves, control rod drive pump oil leaks, and minor maintenance work.

3-12-87 The plant achieved criticality on March 12 at 0920 hours0.0106 days 0.256 hours 0.00152 weeks 3.5006e-4 months .

Startup testing included three (ERIS - Emergency Response Information System) diesel event marker tests, Safety Relief Valve (SRV) and related piping tests, feedwater master level controller steps on the turbine feed pumps, chemistry data, suppression pool cooling loops A & B, suppression pool cleanup performance, (LPRM) nuclear instrumentation calibrations steam condensing capacity testing, flow control valve (FCV) position loop and deadband, and various process computer (DSTC) tests.

Planned Scram #1 (87-03) from the remote shutdown panel occurred on March 24 at 1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months .

5

l 2.2 STARRIP TESTING PROGRAM CHROIOLOGY (CONT )

3-25-87 The plant was placed in a planned outage (Post TC-1 outage) to make modifications to the reactor core isolation cooling (RCIC) isolation valve, reactor vessel instrumentation (B21),

residual heat removal (E12) maintenance work as well as Division I, II, and III diesel and other minor maintenance work. During this period TC-1 testing was reviewed and approved by PORC to complete TC-1 on 4-10-87.

4-10-87 Test Condition 2 started when criticality was achieved at 0455 hours0.00527 days 0.126 hours 7.523148e-4 weeks 1.731275e-4 months on April 11 and special test SXI-0012, for the RCIC system, was performed at reactor pressures of 250#,

400#, 700# and 950# to evaluate the B21 instrumentation changes made during the outage.

4-13-87 A manual scram (87-04) was initiated at 1817 hours0.021 days 0.505 hours 0.003 weeks 6.913685e-4 months due to a loss of vacuum caused by a leak that developed on a high pressure condenser drain manifold. Work completed in this outage included; hooded inserts on the remaining B21 level instrumentation channels, E12 pipe whip restraints, MOVATS testing on HPCS, and maintenance work.

4-26-87 Startup commenced when criticality was reached at 03:49.

Special test SXI-013 was performed at reactor pressures of 250# and 950# to verify and test the B21 level instrumenta-tion changes performed during the outage. Other startup tests included RCIC discharge to the vessel, APRM/IRM overlap, BOP piping and NSSS thermal expansion / vibration testing. Feedwater Controls System Tuneup on the startup level controller was performed prior to reaching ~50% core thermal power on May 1, 1987, at 14:38 hours.

6

2.2 STARTUP TESTING PROGRAM CHRON0I.OGY (CONT.)

5-1-87 A manual scram (87-05) was initiated at 1737 hours0.0201 days 0.483 hours 0.00287 weeks 6.609285e-4 months when a loss of feedwater occurred when the feedwater valve N21-F0230 closed (air line break) causing a hot surge tank low level signal to trip the feedwater booster pumps.

5-5-87 After repairing the air line on the 1N21-F230 valve criti-cality was achieved at 1141 hours0.0132 days 0.317 hours 0.00189 weeks 4.341505e-4 months on May 5, 1987. Startup testing was the critical path until the (planned scram #2, 87-06) loss of offsite power test was successfully run at 1919 hours0.0222 days 0.533 hours 0.00317 weeks 7.301795e-4 months on May 10, 1987. This test completed Test condition 2 testing.

7 l

l SECTIOT 3 STARnTP TEST INSTRUCTIO15 (No Change From Initial Report) 8

SECTICN 4 STAR WP TEST RESULTS 9

STARWP TEST RESULTS m is section contains summaries of the test results for those tests which were completed during the period covered by this report. nese summaries are not intended to modify any items in the official test results packages. Attachment 1 contains a listing of all test result packages completed to date.

We test results summaries consist of a brief description of the purpose or scope of the test followed by a listing of the applicable acceptance criteria and a description of the actual results. When appropriate, more detailed information is provided in a " discussion" section.

10 i

STARTUP TEST RESULTS (CCtff.)

4.1 STI-P35-001 Chemical and Radiochemical (TSN-129, 197) Chemistry Data and Power Ascension This test verifies that chemical and radiochemical parameters of the reactor coolant and selected support systems meet acceptable limits and determines, using approved plant procedures, the. adequacy of sampling equipment and analytical procedures / techniques for sampling. Additional objectives for this test are to evaluate fuel performance, evaluate demin-eralizer integrity, measurement and calibration of the off-gas system and certain process instrumentation, and demonstrate proper steam separator-dryer operation. During Test conditions 1 and 2, samples were obtained from Reactor Water Cleanup, Feedwater, CRD Water, and Off-Gas.

Acceptance Criteria Level 1

1. Chemical factors defined in the Technical Specifications and fuel warranty must be maintained within the limits specified.

2. The Activity of gaseou ..nd liquid effluents must conform to license limitations.

3. Water quality must be know at all times and must remain within the guidelines of the Water Quality Specifications.

11

STAPIUP TEST RESULTS (CO!C.)

STI-P35-001 (cont.)

Results _

Results *

1. FWCU Influent TC1/TC2 Criteria

a. Conductivity ( mho/cm 0 25'C) .406/.690 $ 1.0

b. Chloride (ppm) .0024/.0026 5 0.2

c. pH (0 25'C) 8.0/8.1 5.6 to 8.6

d. Boron (ppm) < 1.0/<1.0 $ 5.0

e. Silica (ppm) .245/.320 $ 5.0

2. Feedwater

a. Conductivity (vmho/cm 0 25'C) .071/.063 $ 0.1

b. pH (0 25'C) 7.0/7.0 5.7 to 7.5

c. Metallic Impurities (ppm) .001/.006 5 0.015

d. Cu Impurity (ppm) .0002/.0003 $ .002

e. Oxygen (ppm) .066/.060 0.02 to 0.2

3. CRD Water

a. Conductivity ( mho/cm 0 25'C) 0.068/.072 $ 0.1

b. Oxygen (ppm) 0.05/.029 5 0.05

4. Condensate

a. Conductivity ( mho/cm 0 25'C) .077/.073 t;/A

b. Chloride (ppm) < .001/<.001 IJ/A

c. fuspended Iron oxide (ppm) 50/.075 !!/A 12

STARWP TEST RESULTS (CCNT.)

STI-P35-001 (Cont.)

Results *

5. off-Gas K1/Ic2 Criteria

a. Pre-Treatment (pci/sec) None detected /19.2 < 3.58E5

b. off-Gas Vent (pci/sec) < limits /< limits **None

TC1 at ~30% power, K2 at -50% power

- 0ff-Gas vent activity is converted to dose to the body in accordance with Technical Specifications.

A1.' data collected satisfied the criteria of this test.

4.2 STI-D21-002 Radiation Measurements (TSN 115, 137) he purpose of this test during Test condition 1 is to monitor plant radiation levels after reaching 20% reactor power and to perform a survey prior to, during, and after Reactor Water Cleanup (RWCU) resin transfer.

Acceptance Criteria Level 1 h e 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 10CTR20,

" Standards for Protection Against Radiation."

13

STARIUP TEST RESULTS (CCNr.)

STI-D21-002 (Cont.)

8.1 - General Survey Perfornance This test obtained neutron and gamma readings throughout the plant at 20%

power. In addition to general area surveys, there were a total of 203 radiation bare points monitored during this test. The entire survey took _. .-

40 Health Physics manhours to complete. All areas surveyed were less than ,

or equal to 0.2 mrem /hr gamma and less than or equal to 0.5 mrem /hr neutron with the exceptions of the Reactor Water Cleanup Rooms A & B (15.0 and 18.0 mrenVhr) and the Reactor Sample Station area (0.6 mrenyhr).

The results of the radiation surveys were compared with the maximum radiation fields described in FSAR figures 12.3-1 through 12.3-8. In all cases, the measured values were well below the FSAR maximum values, as expected for a 20% power level. All Acceptance Criteria were satisfied and no Test Exceptions were issued for this test.

8.2 - RNCU Resin Transfer Survey Health Physics personnel performed Radiological Surveys prior to, during, and after a RNCU Resin Transfer in accessible areas adjacent to FWCU resin transfer piping.

'the highest contact reading of 6 R/hr was recorded before the transfer began. This reading was located at the RNCU Backwash Receiving Tank where the resin was stored. When the transfer was complete the reading dropped to 200 mR/hr.

Level 1 Acceptance Criteria was satisfied during the performance of this test. [

14

STARIUP TEST RESULTS (CONT.)

4.3 STI-Cll-005 Control Rod Drive System (TSN-171)

The purpose of the CRD System test is to determine that the CRD System operates properly over the full range of primary coolant temperatures and pressures from ambient to operating. In conjunction with planned scrams, the scram times of selected CRD's and the in-service CRD flow control valve's response time will be measured.

Acceptance Criteria Level 1

1. For vessel pressures between 950 psig and 1050 psig the maximum scram times of individually fully withdrawn CRDs shall comply with the following table: (Note: Performance rated with charging headers at 1750 psig)

The scram insertion time of each control rod from the fully withdrawn position, based on opening of contacts of main scram contactor (deenergization of scram pilot valve solenoids) as time zero, shall not exceed the following criterion:

Maximum Insertion Times (sec)

From opening of contact of main scram contactor (de-energization Reactor Pressure of scram pilot valve solenoids) psig to Notch Position

43 29 13 950 0.31 0.81 1.44 1050 0.32 0.86 1.57

For intermediate vessel dome pressure, the scram time criteria are determined by linear interpolation at each notch position.

15

STAR'IUP TEST RESULTS (COhT. )

STI-Cll-005 (cont.)

2. If the maximum scram insertion time of one or more control rods exceeds criterion 1 above, then the following criteria are applicable:

a. W e individual scram times of a drive exceeding the times of criterion 1 above shall not exceed the following table:

Maximum Insertion Times (sec)

From opening of contact of main scram contactor (de-energization Reactor Pressure of scram pilot valve solenoids) psig to Notch Position

43 29 13 950 0.38 1.09 2.09 1050 0.39 1.14 2.22

b. We total nu=ber of drives failing criterion 1 but meeting criterion 2.a shall not exceed 7.

c. The average scram times of the remaining (i.e., those that meet criterion 1) individual control rod drives shall be less than the following table:

Maximum Insertion Times (sec)

From opening of contact of main scram contactor (de-energization Reactor Pressure of scram pilot valve solenoids) psig to Notch Position

43 29 13 950 0.30 0.78 1.40 1050 0.31 0.84 1.53 16

STAR 7UP TEST RESULTS (COtTr. )

STI-Cll-005 (Cont.)

d. A drive failing criterion 1, but meeting the criteria under 2 shall not occupy an adjacent location in any direction, including the diagonal, with another slow or inoperative drive. Note that a drive that fails criterion 2 is considered to be inoperative.

Level 2 Upon receipt of a simulated or actual scram signal (maximum error), the FCV must clore to its minimum position within 10 to 30 seconds.

Results During the scram for Shutdown from outside the Control Room test, ERIS data was collected. Based upon previous startup testing, the four slowest rods which were at position 48 were selected for analysis. All four rods scrammed well within the time limits of the Acceptance Criteria and no further analysis on these rods was required by the test. In addition, the CRD cooling water flow control valve closure time was determined to be 12.2 seconds, satisfying its limits for a closure time between 10 and 30 seconds.

ERIS Scram Timing was planned to be used to measure the CRD scram times for 108 fully withdrawn rods but appeared to receive the initiation signal

-120 msec late. This resulted in the control rods being ~120 msec faster than expected. Correcting for this delay, all 108 rods appeared to be measured correctly and were satisfactory. A Dranetz Time Events Analyzer was used in parallel to monitor the four sl est rods and functioned correctly. The Dranetz was used to validate the ERIS Scram Timing feature during single rod scrams, this was the first use of ERIS for multiple rod 17

STARIUP TEST RESULTS (COtC. )

STI-C11-005 (cont. )

timing. A TER was written to document that the times measured by ERIS vere not correct and to use the times obtained from the Dranetz Time Event Analyzer for the four slowest rods.

4.4 STI-C51-010 IRM Performance (TSN-179)

The purpose of this test is to verify, and if necessary, adjust the IM gains to obtain the required overlap with the APM system. This test was performed in TC-2 at the first opportunity following the APM calibration performed in TC-1.

Acceptance Criteria Level 1 Each IRM channel must be on scale before the SMs exceed their rod block setpoint. Each APM must be on scale before the IRM's exceed their rod block setpoint.

Level 2 Each IRM channel must be adjusted so that a half decade overlap with the SRMs and one decade overlap with the APRMs are assured.

18

STARTUP TEST RESULTS (COTr.)

STI-C51-010 (Cont.)

Results he IM-APM overlap test was performed during a reactor heatup. All IMs behaved as expected with the exception of IM-D, which could not be adjusted below 100/125 on range 8 with the APRtis indicating approximately 4%. TER 179-1 was issued since STI step 8.3.7.2 and the Level 2 analysis could not be performed. Upon further investigation, it was discovered that the preamp of IM-D had been replaced on 4/24/87, and would require further adjustments. A work order (W.O. 87-3726) was issued to perform gain adjustments on IM-D. Resolution of TER 179-1 showed that IM-D could be adjusted to produce acceptable overlap. This was verified on a subsequent reactor startup.

4.5 STI-C51-011 LPM Calibration (TSN-107, 108) his test verified the flux response of the LPMs (retest per a Test Exception Report) and calibrated the LPMs.

Acceptance Criteria Level 2 Each LPM reading will be within 10% of its calculated value.

19

STARIUP TEST RESULTS (CONT.)

STI-C51-Oll (Cont.)

Results The LPRM Verification Test was originally performed in Test Condition lleatup. At that time 65 of the 164 LPRMs failed to respond to control rod withdrawals. Four (4) of the 65 were subsequently retested satisfactorily during TC IIeatup. The remaining 61, documented by TER-066-1, were to be retested in TC-1 at higher flux levels. Their lack of response was not unexpected at the low flux levels of TC Heatup. All 164 LPMs were re-tested at approximately 20' power. 160 of the 164 LPM's responded to flux changes. The four (4: LPMs which failed to respond to flux changes are "A" level LPRMs and were socumer:ted in Test Exception Report 107-1.

These four (4) LPM's were declared INOP and will be investigated during future Startup Testing. These four LPRMs are: 08-41A,32-49A, 32-25A, and 32-57A.

A LPRM calibration was performed using Surveillance Instruction SVI-C51-T5351. Since the Process Computer program OD-1 was not declared opera-tional, the General Electric Mark III computer, using the BUCLE program, was utilized. This encompassed gathering LPM data, Digitized Traversing Incore Probe data and other reactor data required to run BUCLE. The LPRM gain adjust:nent factors (GAF's) calculated by BUCLE were then used to ad-just each of the LPM amplifier gains to yield the calibrated LPM readings. Due to low flux levels in the lower core,12 LPMs were found not to be reading within 10% of their calculated values. Test Exception Report 100-1 documented this and the partial teruns of SVI-C51-T5351 for the 12 LPRMs. Seven LFMs remained outside the Level 2 Acceptance Criteria. Since these LPRMs are all "A" level LPMs, it is expected that at higher power levels these out-of-limit LPM Cars can be more easily adjusted. This will be documented in a future Startup Test.

20

STARWP TEST RESULTS (COtfr. )

4.6 STI-C51-012 APM Calibration (TSN-155)

This test was performed to adjust the APRM gains using the Process Computer program, OD-3, to obtain a heat balance. Previous APM calibrations were performed at very low powers and used a Constant Heatup Rate Heat balance.

Acceptance Criteria Level 1

1. The APM channels must be calibrated to read equal to or greater than the actual core thermal power.

2. Technical specification and fuel warranty limits on APM scram and Rod Block shall not be exceeded.

Level 2 If the above criteria are satisfied, then the APM channels will be con-sidered to be reading accurately if they agree with the heat balance or the value required by Technical Specifications (based on the ratio of CMFLPD to TRTP), to within !2% of rated power.

Results

1. K -1 Performance The APM calibration was performed using Surveillance Instruction SVI-C51-T0024. Core thermal power was determined, using the Process computer 0D-3 program, to be 28.02% of rated. The APM readings were compared to actual core power. Only one APM (channel E) required amplifier gain adjustment to read consistent with Acceptance criteria. All others were acceptable.

21

STARWP EST RESULTS (COtTr.)

STI-C51-012 (Cont.)

A review of the most recent performances of SVI-C51-T0027 (A-H) demonstrated that APM scram and Rod Block Setpoint were consistent with Technical Specifications.

APM Readings (First Perforr.ance)

APM As-Found Desired

As-Le f t A 29.0 28.82 29.0 B 29.0 28.82 29.0 C 29.0 28.82 29.0 D 29.0 28.82 29.0 E 28.5 28.82 29.0 r 29.0 28.82 29.0 G 29.0 28.82 29.0 H 29.0 28,82 29.0

Desired = APM readings to be greater than or equal to actual Core Thermal Power Due to significant APM gain changes since the last IM-APM overlap verification test, a review of the results was perforred. It was determined that significant APM gain reductions had occurred since the last overlap test and that a reperformance was required. Since the IM-APM overlap test was already scheduled, no Test Exception Report was required.

22

STARIUP TEST RESULTS (CCNr.)

STI-c51-012 (Cont.)

2. TC-2 Performance Core thermal power was determined to be 47.66% of rated, using the Process Computer CD-3 program, and the APM gains were adjusted to be consistent with the heat balance. Due to a power transient, caused by a Feedwater Master level gain controller spurious failure, during the performance of svI-C51-T0024, the STI Acceptance criteria could not be verified. Consequently, per the STI, svI-c51-T0024 was re-peated after reactor power had stabilized. The STI was successfully completed with the "as-left" APM readings verified to be within the acceptance criteria of +2, -0% of rated thermal power.

APM Readings (second Performance)

APM As-Found Desired

As-1 eft A 48.5 47.66 48.5 8 48.0 47.66 48.0 C 48.0 47.66 48.0 D 47.0 47.66 48.0 E 47.0 47.66 48.0 r 47.0 47.66 48.0 0 49.0 47.66 49.0 H 47.0 47.66 48.0 i

Desired = APM readings to be greater than or equal to actual Core Thermal Power 23

r ,

V STARWP TEST RESULTS (CONT. )

4.7 STI-C91-013 Process Computer (TSN-126, 131, 132, 133, 134, 135, 136, 142, 143, 144, 145)

LPRM Calibration and TIP Plotter Adjustment (8.3) h e purpose of this test is to verify that when process computer program 00-1 (LPM Calibration BASE Distribution) is run:

1) the arrays containing LPRM calibration data are updated;

2) the traversing in-core probe (TIP) x-y plotter is properly aligned with respect to core top and bottom;

3) the CD-1 traverse edits are aligned consistently with the x-y plotter traces.

Acceptance Criteria W ere are no acceptance criteria applicable to this test.

Results After the TIP drive control unit (DCU) amplifier gains and x-y plotter gains were adjusted, the beginning and end points of the various analog traces were verified to be in agreement with those generated during the initial system setup.

24

l l

sTAnwr Test nestmTs (cour.)

i STI-C91-013 (Cont.)

i Machine-digitized data were also verified to track closely the analog traces generated by the plotter in that the flux peaks lined up well, he data were not coincident, however, due to the yet-to-be-optimized plotter y-axis gain.

)

After an 0D-1 was successfully run to completion, the arrays BASLP and I

BLPP in the Monicore VAX were demonstrated to have been overlaid with fresh data, indicating that their contents were updated when the programs PCA003/MCA003 (Transfer 0D-1 Data to MC) ran.

During pre-test hardware interface checkout, it was discovered that the difference between core top and core bottom as expressed by the relation NCCT-NCCB had to be equal to 145 inches instead of the previous 144 inches. A pulse broadening circuit serves to delay slightly the trans-l mission of the data from the TIP DCU to the process computer. Se logic circuitry which controls the end of the data transmission process is not subject to this delay, so the link to the process computer is cut off before the last pulse is received by the computer, and 0D-1 consequently aborts due to counting out low.

Setting the bottom core limit 145 inches below core top causes 145 data pulses to be sent to the computer, although for the reason already cited,

-only 144 are registered by the software. %erefore, this adjustment has no effect on any other hardware or software setting. It should be noted that the upper core limit was not readjusted since it is fixed by TIP tube length and detector turnaround constraints.

1 25

- - - _ - _ - , . -- . - , . - , . . , , , = -

STAR 7UP TEST RESULTS (CCNr.)

STI-C91-013 (Cont.)

once the new values for core bottom were set into the DCU hardware, it was necessary to readjust the individual machine x-axis gains and channel zeros such that the odd (145th) inch would fall at the bottom of the stroke, and the TIP traces would always begin at core top (NCCT). It was found that channel zero settings were the same as the "as-left" values of the hot TIP alignment phase of this test (Section 8.2).

On a representative sampling of analog TIP traverses, the corresponding computer-digitized, machine-normalized, and full power adjusted transverse data from 00-1 were plotted. In all cases the agreement, as measured in the coincident axial deviations of the flux peaks, was as close as could be expected when comparing discrete (digitized) vs. continuous (analog) data.

he BASLP and BLPP arrays (containing OD-1 data) in the Monicore VAX were verified to have been updated by the running of OD-1, demonstrating that the programs PCA003 and MCA003 ran before 00-1 terminated.

No hardware or software problems or any procedural difficulties were noted in the performance of this test, and all test objectives were satisfactorily met.

DSTC - Data Initialization by OD-15 (8.6)

W e purpose of this test was to verify the proper operation of process computer program 0D-15, Computer outage Recovery Monitor (CORM). In this evolution, the computer software was initialized for the first time.

Cumulative thermal and electrical generation energies were manually calculated and entered into the computer in response to the OD-15 prompts.

26 m - , -

., -- . ~ . - - .

STARWP TEST RESULTS (CCNT. )

STI-C91-013 (Cont.)

Acceptance Criteria Level 1 None Level 2

1. Programs OD-1, P1 and OD-6 will be considered operational when:

a. 'Ihe MCPR 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 MCPR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly, the MCPR and CPR calculated by the two methods shall agree within 2% (8.10).

b. The maximum MGR calculated by EUCLE 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 MGR calculated by the precess computer is in a different assembly than that calculated by BUCLE, for each assembly, the maximum MGR and MGR calculated by the two methods shall agree within 2% (8.10).

27

STARWP TEST RESULTS (CCNT.)

STI-C91-013 (Cont.)

c. he MAPLHGR 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 MAPLHGR calculated by the process computer is in a different assembly than that calculated by BUCLE, for each assembly, the MAPLHGR and APLHGR calculated by the two methods shall agree within 2% (8.10.

d. W e LPRM calibration factors calculated by BUCLE and the process computer agree to within 2% (8.10).

2. W e remaining programs will be considered operational upon successful completion of the static and dynamic testing.

Results h e process computer software was initialized for the first time during this test. CD-15 was verified to be functioning correctly, reflecting properly the accumulated core thermal and electric energies generated by the plant to the time of the test. No problems were noted in the performance of this section of the Dynamic System Test Case (DSTC), and all test objectives were met satisfactorily.

At the start of this test, the hardware link between the Honeywell 4400 (process computer) and the VAX (minicomputer) was verified to be working properly via the program DSPINK.

28 i.

STAR'IUP TEST RESULTS (CONT. )

STI-C91-013 (Cont.)

The process computer program CD-15 (CORM) was run and the following variables were input:

TIME -

present time and date MOP -

generator synchronization time GNEOP - average generator power from 'INOP to TIME CTPOP -

average core power from 'INOP to TIME The generator synchronization time was arbitrarily set to 00:00:01 on 18 February 1987 both because the time coincided with the start of the then-current on-line run and the average thermal energy generation rate for the cycle was approximately the same for the 11+ day average as the core thermal power output at the time of the 0D-15 run. Per information supplied by Reactor Engineering, the assumed average core thermal power and electrical output of 1120 mt (=CIPOP) and 115 Ne (=GNEOP),

respectively, from the time of generator synchronization (20P) to the 0D-15 time /date would have yielded the values of thermal and electrical energy of 13167 NDt and 1357 N De, respectively.

Transfer of new OD-1 data to the VAX minicomputer was required per the 0D-15 prompt before program P1 (Periodic Core Evaluation) could be demanded.

'Ihe computer determinations of ECOR (core thermal energy accumulated during the computer outage, i.e., 20P to TIME) and EGEN (electrical energy accumulated from 'INOP to TIME) were checked against manual calculations. The only differences noted between the two calculations were attributable to the fact that it was assumed that OD-15 would be run on 1 March 1987 at 1800 whereas it actually happened at 1805. Core thermal power and electrical output as recorded by the computer at the time of the 0Q-15 were verified to be within approximately 1% of a manual heat balance calculation and control room front panel indication.

29

STARTUP TEST RESULTS (CCNr.)

STI-C91-013 (Cont.)

h e time and date of the most recent P1 exposure update was verified to be identical to Tf10P, the time of computer software initialization.

Finally, P1 Option 1 was demanded and failed (as required) due to the OD1 data transfer flag being set.

DS1C - Plant Sensor Checks (8.7)

The purpose of this test was to verify correct operation of process computer programs CD-3 (Core t ermal Power and APRM Calibration), 00-7 (Present Control Rod Positions), and OD-8 (Present LPRM Readings). The OD-3 core thermal power determination was verified to be within 1% of a manual heat balance calculated at the same plant power level. Other plant variables edited by OD-3 were verified to be reflecting plant operating conditions based on a review of control room indicators. OD-7 control rod positions were determined to be in agreement with those shown on the RC&IS display. Finally, the OD-8 LPRM readings were verified against those of the meters on the Control Room back panels.

Acceptance Criteria Same as Section 8.6 Results he results of the heat balance computations, manual vs. computer-generated, were as follows:

Core power, manual heat balance 751.0 mt Core power, 00-3 755.8 mt Percent difference 0.638%

30

STAR WP TEST RESULTS (CONT.)

STI-C91-013 (Cont.)

In addition, the comparison of the 00-7 control rod positions with RC&IS indications revealed no discrepancies,' and the 00-8 edited LPRM readings with those of the Control Room showed only differences attributable to noise and momentary fluctuations.

No problems with either the process computer hardware or software were noted in this test and at its conclusion, the subject programs were declared to be operational.

At the beginning of this test, the computer was reinitialized,'and using the DSPINK program, it was verified that LPRMs and other NSSS data points were being scanned every 5 seconds. We one-minute averaging routine was also verified to be functioning correctly for the feedwater flow signals.

Next, the only failed sensors appearing on the OD-3 edit and in the CT array (scanned plant data at time of last P1) were determined to be the recirculation pump A&B powers. his was expected since the computer ,

points only treat the power supplies to the pumps when they are operating ,

at high speed, and the pumps were on the LFMG throughout 'IC-1.

he OD-3 core thermal power was verified to be within 0.638% of the reactor power level determined by a manual heat balance. ne manually calculated APRM gain adjustment factors (AGAF) and the APRM calibration i constants (CAP) were also shown to be approximately equal to the computer-generated values, any differences being attributed to normal APRM signal fluctuations.

J t

31

STAR 7UP TEST RESULTS (CCNr. )

STI-C91-013 (Cont.)

In addition to the verification of the OD-7 and 0D-8 edits with indepen-dent plant instrumentation, the respective edits requested from the Honeywell 4400 (process computer) and the VAX (minicomputer) were shown to be identical in the case of 00-7 and substantially equivalent (due to the aforementioned statistical variations in LPRM signals) in the case of OD-8. No problems with either the computer hardware or software were noted in the course of this test.

DSTC - Verification of Programs Enabled by OD-15 (8.8)

The primary purpose of this test was to check that the process computer program 00-15 (CORM) properly initializes programs P4 (Ten-Minute Core Energy Increment) and OD-18 (LPRM Alarm Trip Recalculation). The total feedwater flow rate value determined by the process computer square root routine was verified to be correct by means of a manual calculation.

Initial checkout of OD-18 could not be carried out due to a problem later identified as a failure of the process computer scan function to recognize

-core flow as a variable whose values are to be checked against previously defined limits. A software patch was implemented which, upon a partial retest of the applicable test section, permitted the successtui demonstra-tion of an 00-18 operational check. At the conclusion of this test, program P4 was declared operational.

Acceptance Criteria Same as Section 8.6 32 b

STARWP TEST RESULTS (CCNT.)

STI-C91-013 (Cont.)

Results The program P4 was verified to be running every ten minutes, as required, j and to be updating core thermal and electric energies correctly (as l checked against a manual calculation).

The process computer square root routine was then verified to be function-ing properly by performing a manual calculation to determine the total feedwater flow rate.

Finally, an operational check of program OD-18 was made wherein the values of variables bracketing core flow were arbitrarily changed to zero and later verified to have been restored by the triggering of co-18.

A checkout of the program P4 was accomplished by recording initial values of ECOR (core thermal energy accumulation), EGEN (accumulated electrical energy generation), CTP2 (current core thermal power), GMN2 (current plant electrical output), and TIM 2 (time of last P4 updating). The program was verified to have operated three times at ten-minute intervals by observing the "Show process" utility on a CRT display. Afterwards the same variable values as noted earlier were recorded and respective differences computed.

These differences were then checked against manual calculations using the

' computed average of the thermal and electric powers from the first and last P4 runs. Although this P4 check was made by adding relatively small numbers (30 minutes energy accumulation) to large ones (approximately two weeks' energy accumulation), agreement was very close, and P4 was therefore declared operational.

33

STARTUP TEST RESULTS (CONT.)

STI-C91-013 (cont.)

ne process computer square root routine was next checked by performing a manual calculation for the two branches of feedwater flow. The test

. procedure started by obtaining millivolt readings (proportional to the differential pressure across the feedwater flow element taps) and used the same fit coefficients as the computer. The two feedwater branch flows were then sunned and compared with the total feedwater flow edited by OD-3, the values being 4.0591 Mlbn/hr for the manual calculation and 4.06 MlbnVhr for OD-3.

An operational check of program 0D-18 was attempted next. The procedure called for reading WID (flow bracket low limit), JW (index of EKF array -

contains flow fractions from 0.25 to 1.00), and WHI (flow bracket high limit) prior to setting them to 0,1, and 0, respectively. Instead of having the values 26, 2, and 57.2 respectively, they already indicated the values to which they were to be changed. 0D-15 had caused them to be set to the values indicated upon initialization, but OD-18 had not run to bracket the actual core flow rate. TER-133-1 was initiated to track this problem.

Se source of the CD-18 problem was traced to a software routine that did not recognize core flow as a variable whose values are to be checked against defined limits. A software patch was implemented and partial re-tests of the test section were performed per the ITR closecut resolution.

We values of WID, JW, and WHI were changed as prescribed and then 00-18 initiated to restore the values that had existed previously. A manual calculational check of these values was performed and all were satisfactory.

No hardware problems nor any other problems other than the one already mentioned were noted in this test.

! 34

STARWP TEST RESULTS (CONT.)

STI- C91-013 (Cont. )

DS M - LPRM Calibration (8.9) he purpose of this test was to verify that process computer program CD-1 (LPRM Calibration and BASE Distribution) functions properly prior to its use in the validation of program P1 (Periodic Core Evaluation). It was performed in conjunction with a routine LPRM calibration. Process computer hardware and software performed satisfactorily during this test, and no problems were noted.

Acceptance Criteria

' here are no criteria applicable to this test.

Results All TIP traverses were run for all machines in a complete OD-1 run. The program was terminated normally at the conclusion of the TIP set. We BASE and BASLP arrays and the machine normalization constants, Am, were verified correct by' performance of manual calculations. We BLIPP and BLPP were verified to have been set to zero by CD-1, the RC set equal to

BASLP, the TSV array equal to the CD-1 run date, the CRDSV array was

. verified to reflect accurately the control blade coverage of the LPRM positions, the CX array was correctly set to 1.0 at all positions, and finally, the C, NPAS arrays, and the CD-8 edits were verified to contain the same data as the respective CD-1 edits. Digitized TIP traverse data from CD-1 edits were plotted on the corresponding analog plots, and axial alignment of the two respective traces was shown to be good.

I i

35 i._ - - - - - .- _ - - . _ _ - - _

. ~ _ _ - -_ ._ _ _ _ _

STAR'mP TEST RESULTS (CCtfr. )

STI-C91-013 (cont.)

Discussion h e CD-1 program was run in conjunction with the generation of a set of TIP traces for a routine LPRM calibration per a plant surveillance procedure. The resulting analog traces were for' hand-digitization since the OD-1 program was not administratively operable at the time of this test. Several CD-8's were run during the course of TIP data acquisition to show that power was approximately constant. Shortly before the last TIP trace was run, a DC electrical buss was inadvertently de-energized, a recirculation pump tripped, and reactor power level dropped several percent. This was determined to not adversely impart the continuation of test and the performance of the final TIP trace.

Manual calculations were performed to check that the 0D-1 software logic correctly calculated the arrays TPP (TIP readings at LPRM elevations),

BASE (base axial power shape), BASLP (BASE values at LPRM axial eleva-tions) and A (machine normalization factors) were verified to be correct i by comparison with manual calculations.

h e BLIPP and BLPP arrays (correction to effective LPRM reading distribu-

tion due to control pattern changes at fuel segment centers and LPRM elevations, respectively) were zeroed out by the running of 0D-1. he RC array (LPRM substitute values) was shown to be equal to the BASLP array (point-by-point). W e C (LPRM calibration constants) array was verified to be equal to the respective values calculated by OD-1 and shown on the edits. We NPBAS and VBAS (control rod positions at time of BASE deter-mination and void fraction at LPRM elevations, respectively) were shown to have been edited properly by CD-1.

I u

36

STARWP TEST RESULTS (CCNT.)

STI-C91-013 (Cont.)

h e values of RP (LPRM readings) on the 00-1 edits were demonstrated to be roughly equal to the CD-8 edited values of the time nearest to that at which the TIP trace was made.

S e TSV (times of LPRM substitute value determination) was verified to be uniformly equal to the date of the CD-1 run. W e CRDSV (control rod den-sity or " coverage" at the LPRM elevations at time of LPRM substitute value determination) contained the same valves as appeared on the 00-1 edits.

We C array (LPRM exposure correction factors) were shown to be uniformly equal to 1.0, as expected.

Finally, the flag IGAF (for calculation of GAfD, exposure-dependent GAF multiplier) was verified to have been set to one by the running of OD-1, indicating that P1 should calculate new LPRM gain adjustment factors (GAF's).

DSTC - Power Distribution and termal Limits (8.10)

The purpose of this was the validation of process computer program P-1 (Periodic Core Evaluation). hermal limits (and their core locations) and LPRM calibration constants determined by P-1 were in agreement with the values calculated by the off-line computer program BUCLE. Se P-1 soft-ware performed according to its specifications, and all acceptance criteria were satisfied.

At the conclusion of this test, process computer programs P-1, P-6, OD-1, 00-6, OD-9, OD-10, and OD-16 were declared operational.

Acceptance Criteria Same as Section 8.6 37 1

\

STARIUP TEST RESULTS (CONT.)

STI-C91-013 (Cont.)

Results The following tables present the principal results of this test. In each case, the four most limiting bundles / core locations are shown because although the thermal limits are the same (due to core symmetry), the computer codes P-1 and BUCLE order these bundles / locations differently.

MCPR Process Computer BUCLE Fuel Ass'y MCPR Fuel Ass'y MCPR Percentage Coordinates Coordinates Difference 27-18 2.899 27-44 2.899 0 33-18 2.899 33-18 2.899 0 27-44 2.899 33-44 2.899 0 l

33-44 2.899 27-18 2.899 0 LHGR Fuel LHGR Fuel LHGR Percentage Coordinates Coordinates Difference 1

27-26-15 4.40 33-26-15 4.40 0 33-36-15 4.40 33-36-15 4.40 0 27-36-15 4.40 27-26-15 4.40 0 33-26-15 4.40 27-36-15 4.40 0 38

STARTUP TEST R55M.TS (CCNT. )

STI-C91-013 (Cont.)

MAPMGR Fuel MAPLHGR Fuel MAPLHGR Percentage Coordinates Coordinates Difference 27-36-15 3.83 33-26-15 3.83 0 33-26-15 3.83 33-36-15 3.83 0 33-36-15 3.83 27-26-15 3.83 0 27-26-15 3.83 27-36-15 3.83 0 In addition, the LPRM calibration factors, C, as calculated by P-1 and BUCLE were equal at each respective LPRM location.

The first phase of this test was to demand P-1 (Periodic Core Evaluation).

Since P-1 zeros out the hourly accumulations of core thermal and plant electrical energies, a recheck of program P-4 was performed this time -

comparing numbers of approximately the same magnitude, i.e., the 10-minute increments are only about 25-33% of the accumulated energy at the time of the test. No problems were noted with the P-4 program.

Next, verifications cf power and flow distributions, thermal limits, LPRM gain adjustment factors, and LPRM alarm trip settings were performed wherein programs OD-6 ('Ihermal Data in a specified Fuel Bundle), 00-9 (Axial Interpolation Data), and OD-16 (Target Exposure and Power Data) for the purposes of comparison with the quantities edited by P-1 and the performance of selected manual calculations, i.e., RC (LPRM substitute values), and ER (LPRM effective readings).

39

STARWP TEST RESULTS (CONT. )

STI-C91-013 (Cont.)

The program 0D-10 (List / Alter NSS Data Files) was used to edit a number of arrays to check against the 00-6, and CD-9 edits and for cesparison with manually calculated data, e.g., DRLP (differences at LPRM elevations) and ATSP (presently applied alarm trip setpoints).

The off-line version of P-1 (BUCLE) resident on the General Electric Mark III' computer was used to generate an' independent P-1 edit. The high-speed (4800-baud) link from the on-site VAX (process computer) to the Mark III was employed to transfer the required data files to permit BUCLE to run.

A check of P-1 exposure accumulation and void fraction determination was made (via manual calculations) on the variables: CICEX (total LPRM exposure), DICEX (daily LPRM exposure), CREX (control rod exposures), EXF (fuel exposure), EXVF (exposure-weighted void fraction), and VF (fuel void fraction). No problems were noted in P-1 exposure updating.

Finally, P-1 was demanded again and steps involving the OD-6, OD-9, and 00-16 comparison, BUCLE run, and exposure updating were repeated to note the effects of incrementing exposure.

- No minor software problems were noted in the performance of this test section: 1) the ATS array (LPRM alarm trip setpoints) being written to incorrect core locations, and 2) a coding error in a P-1 subroutine which was discovered by comparison with the BUCLE edit. Both of these problems were resolved as soon as they were encountered by vendor personnel assisting in the test effort.

All applicable test criteria were met and process computer programs P-1, l P-6, OD-1, OD-6, OD-9, 00-10, and OD-16 were declared operational upon the

( conclusion of this test.

40

1 .

.' STARTUP TEST RESULTS (CCNr.)

STI-C91-013 (Cont.)

DSIX - Daily and Monthly Logs (8.11)

S e purpose of this test was to validate process computer programs P-2 (Daily Core Performance Summary) and P-3 (Monthly Core Performance Susmary). Data obtained from P-2 runs were verified to be correct by comparison with hand calculations and reconciling with P-1 edits. P-3 edited data were determined to be correct from hand calculations and reconciliation with P-2 edits.

No software problems were noted and all acceptance criteria were satisfied. At the conclusion of this test, programs P-2, P-3, OD-13, and 00-20 were declared operational.

Acceptance Criteria None Results Data edited by program P-2 were verified correct by comparison with P-1 data and manual calculations. .Similarly, P-3 data was verified against the data edited by P-2. Good agreement was obtained in every case.

m is test began by manually demanding P-2 and verifying that core thermal and electrical energies accumulated for the month were correct. In addition, the indicated core thermal limits (and their core locations) as edited by P-2 were shown to be consistent with those of the P-1 runs made up to that point. Finally, the cumulative hour spent at or below the specified fractions of rated power were determined to be correct as well.

41

STARWP TEST RESULTS (CCNT.)

STI-C91-013 (Cont.)

Since a P-2 edit was obtained automatically (and inadvertently) during the course of P-1 testing, the P-2 requested in this test was not the first one generated by the Perry process computer. herefore, it was necessary to use both of the edits to verify the correct operations of the Perry P-2 i software. Minor accr=vvhtions to the procedure were made via a startup test change notice to reflect the earlier P-2.

Next, a check of the following data was made in order to verify the proper i

functioning of the software by comparison with data edited during P-1 testing.

EBUN -

bundle exposure (verified against average fuel segment exposure for the first bundle around LPRM string 1)

EXVF -

exposure-weighted void fraction DICEX - daily LPRM chamber expcsures CICEX -

cumulative LFRM chamber exposures CX -

LPRM sensitivity correction factors GAF - LPRM gain adjustments factors (These could not be checked directly since 00-15 was run before the start of the test to reinitialize the computer. However, since P-2 sets GAF = GAFO

CX, this calculation was verified to have been i done correctly.)

ne last phase of the initial P-2 checkout was to verify the P-2 log data against the P-2 edits obtained for this test for the variables listed in 4

the previous paragraph.

Process computer program OD-13 (LPRM Sensitivity) was verified to be editing LPRM exposures, LPRM sensitivity, and gain adjustment factors.

42

STARTUP TEST RESULTS (CONT.)

l l

STI-C91-013 (Cont.)

At this point, the test was paused in order to permit a significant-accu mlation of energies and exposure before the running of another daily log. After several P-l's were generated, the exposure updating capability of the software was verified by performing the identical steps as were done in the first part of the test, including another OD-13 check.

l After P-2 was allowed to run automatically at 2330 hours0.027 days 0.647 hours 0.00385 weeks 8.86565e-4 months on 21 March 1987, the test was put on hold overnight to allow another accumulation of data ,

for a daily log.

After the P-2 edit was generated control rod exposures were edited via CD-10, and P-3 was demanded. .

i The control blade exposures were verified to have been updated when P-3 i ran, as well as the total cumlative days of power operation at the ,

specified power fractions. The monthly energy accumulation registers were also verified to be reset to zero by P-3, as expected.

e Finally, a security log file copy was performed to preserve the MCPIVALS.IAT datafile used in this test for later examination, if required.

No hardware or software problems were noted in the course of this test, and at its conclusion, programs P-2, P-3, 0D-13, and CD-20 (Refueling Update Monitor) were declared operational and made available for use by the plant staff.

i:

43

+.-~4- .., ..--

STARTUP TEST RESULTS (CCtif.)

STI-C91-013 (Cont.)

DSTC - OD-18 Verification (8.12)

The purpose of this test was to verify proper operation of process computer program OD-18 (LPRM Alarm Trip Recalculation). LPRM alarm trip setpoints were manually changed to values lower than the then-current LPRM readings. 00-18 confirmed the existence of LPRM alarms, then recalculated the trip setpoints and nullified the alarms.

No hardware or software problems were noted in this test and CD-18 was declared operational.

Acceptance Criteria Same as Section 8.6 Results One ATSP (Alarm Trip Setpoint value - Presently Applied), was adjusted to make it lower than its then-current LPRM reading. OD-18 initiated after 15 seconds, indicated the presence of LPEM alarms followed by a message indicating that the LPRM alarms had been nullified. Next, the ATS values and associated ATSP were reduced to values less than the associated LPRM readings. CD-18 initiated to confirm the presence of LPRM alarms and

. printed a message advising running P-1. P-1 was then run to clear the advisory message.

The LPRM charber at core-coordinates 16-09 and elevation 'C' was selected for the evolutions of this test.

44

STAR 1UP TEST RESULTS (CCNr.)

STI-C91-013 (Cont.)

The ATSP value for this LPRM was changed from the existing 34.489 to 10, which made it less than the then-current LPRM reading of 22. CD-18 confirmed the presence of LPRM alarms, i.e., one or more LPRM's whose readings were greater than their respective ATSP values. From the current value of core flow, OD-18 them calculated a new ATSP value by linear interpolation of the ATS values (LPRM alarm trip setpoints as a function of core flow), determined that an alatm condition was no longer in effect, and printed "LPRM AIARM NULLIFIED" on the I/O typer.

Next, the ATS points between which 0D-18 interpolates to arrive at the ATSP were changed from 38.69 and 46.85 to 1.0, followed by changing the ATSP value from 34.46 to 1.0.

As before, this action caused OD-18 to register the presence of LPRM alarms. Since the ATS points themselves were less than the reading of LPRM 16-09C, OD-18 was not able to clear the LPRM alarms by interpolating between the ATS values to arrive at an ATSP greater than the LPRM reading, and correspondingly printed an advisory message on the I/O typer to demand P-1 to recalculate updated LPRM alarm trip settings (ATS).

P-1 was then demanded, and new ATS values were transmitted to the process computer and updated ATSP values computed.

No problems relating to either process computer hardware or software were noted, and accordingly program OD-18 was declared operational.

45

STAR'IUP TEST RESTETS (CONT. ) .

STI-C91-013 (Cont.)

DS*It - OD-2 verification (8.13)

'1he purpose of this test was to validate the process computer program CD-2 (Specified LPRM Substitute Value and BASE Distribution). An OD-2 run was made for an LPRM string which had a failed detector and also appeared on the " BASE Criticality Code" list on the most recent P-1 edit. (Note: A

" BASE Criticality Code" is notification that OD-2 should be run for the LPRM string in question in order to maintain good accuracy in the axial power determination in the vicinity of the LPRM(s) on the list.) OD-2 was verified for the failed detector and cleared the " BASE Criticality Code" for the selected LPRM string.

No process computer hardware or software problems were encountered in this test and 00-2 was declared operational.

Acceptance Criteria Same as Section 8.6 Results Program CD-2 successfully supplies a substitute LPRM value for the failed LPRM sensor at 32-57 'A' and cleared the BASE Crit Code for the same LPRM string.

Inspection of the most recent P-1 edit at the time this test began resulted in the choice of the LPRM string at core location 32-57 since it appeared on the " BASE Crit. Code" list and one of its sensors was failed (A-elevation) since both functions of the OD-2 program could be checked out at once. .

i l

i 46

. _= - _ . - .

STARWP TEST RESULTS (CORI.)

STI-C91-013 (Cont.)

A series of 00-10 edits of various data arrays was requested prior to running 00-2 to use for baseline comparison purposes. Among these arrays were CRDSV, TSV, CX, C, A(M), BASLP, RP, RC, BLPP, DRLP, LPF, ER, BASE, BLIPP, DR, and LL. An CD-9 edit was requested for LPRM string 32-57 and verified the earlier P-1 edit.

CD-2 was demanded for the LPRM string followed by a P-1 and an O>9. The recently generated 0D-2 data were currently shown on the 00-9 edit, e.g.,

BLIPP and BLPP zerced out, DR reduced in magnitude, etc.

We same 00-10 edits requested before running 00-2 were redemanded and compared with the earlier edits. The OD-2 generated TIP data was properly reflected. The LBC array was also edited and showed that the LPRM string was eliminated from the list of BASE Crit codes, (confirming the the post-OD-2 P-1 run).

No hardware or software problems were noted in this test and program CD-2 was declared operational.

BOP Calculations (8.29)

The purpose of this test was to verify that the Perry balance of plant performance calculations are being correctly performed by the process computer software. A series of special computer logs were established to acquire the same balance of plant process data used by the computer software routines. Wese data were then used in manual calculations using the BOP specification algorithms for comparison with the computer-generated results.

47

STARWP TEST RESULTS (CONT.)

STI-C91-013 (Cont.)

TER-145-1 was initiated after various computer calculations did not agree with the manual calculations. mis TER also covered various plant sensors which did agree with normal values expected at this power level. Since the results of this test were unsatisfactory, this test must be repeated when all problem areas have been fixed, before declaring the BOP performance calculations operable.

Acceptance Criteria Same as Section 8.6 Results BOP Results Comparison Computer Manual Quality Symbol Result Result Mainsteam Flow NMS D M1b/hr 4.0308 M1b/hr Unit Auxiliary Power MHAUX 0.0 m 0 m Station Net Generation NN 258.5 m 263.4 m Q Added to Cycle by pumps QSUPP N/A MBW/lb 14.7 MBW/lb Q Added to Working Fluid QADD N/A MBW/lb 3661.4 MBW/lb Gross Turbine Heat Rate GTHR D B W/kwh 13,416 B W/kwh Q rejected to Main Condenser QREJ N/A MBW/lb 17,537.6 MB W /lb Net Plant Heat Rate NPHR D B W/kwh 14,040 BW/kwh Net Plant Efficiency NEPFF D % 24.3 %

Gross Plant Efficiency GPEFF D % 25.3 %

NCyrE: D = the computer calculation is not functioning. See TLR-145-1 for additional information.

N/A = Computer Point Not Available At This Time.

48 l

STARIUP TEST RESULTS (CorTr.)

STI-C91-013 (Cont.)

six special computer logs were established before this test began in order to gather the 178 data point values in an efficiently and orderly fashion.

S e test itself consisted of demanding the special logs to obtain the input data for manual calculations to check the software and getting the next subsequent BOP periodic log (normally generated every hour).

S e manual calculations are detailed in Tables 1, 2, and 3 (see pertinent data Section). Numerous deficiencies were noted in the BOP performance calculations done by the process computer. These included instrumentation wrongly connected or out of calibration, process computer points not performing as expected, etc. Rese problems will be addressed by plant work orders and this test will be reperformed at a later date.

4.8 STI-E51-014 Reactor Core Isolation Cooling (TSN-185, 196, 199)

Se purpose of this test is to verify the proper operation from the control room and remote shutdown panel of the RCIC system over its expected operating pressure and flow ranges, and to demonstrate reliability in automatic starting from cold standby.

Acceptance Criteria Level 1

1. The average pump discharge flow must be equal to or greater than the 100% rated value after 30 seconds have elapsed from automatic initiation (or manual push button start) at any reactor pressure between 150 psig and rated.

49

STARTUP TEST RESULTS (CONI. )

STI-E51-014 (Cont.)

2. Se RCIC turbine shall not trip or isolate during auto or manual start tests.

Level 2

1. In order to provide an overspeed and isolation trip avoidance margin, the transient start first and subsequent speed peaks shall not exceed 5% above the rated RCIC turbine speed.

2. ne speed and flew control loops shall be adjusted so that the decay ratio of any RCIC system related variable is not greater than 0.25.

3. W e delta P switches for the RCIC steam supply line high flow isolation trip shall be calibrated to actuate at the value specified in the plant Technical Specifications (about 300%).

4. Motor operated valves open satisfactorily (within specified time) with start differential pressure conditions simulated to the extent practicable.

Results TC-2 testing of RCIC consisted of

1. Section 8.3 - RCIC Discharge to Vessel (Rated)

2. Section 8.4 - RCIC Cold Quick Start to Vessel (Rated)

3. Section 8.5 - Operation of RCIC from Remote Shutdown Panel 50

l STAR'IUP TEST RESULTS (CONT.)

STI-E51-014 (Cont.) 1 The Discharge to vessel test involved the demonstration of flow controller stability and a RCIC quick start at rated pressure (920 psig).

Demonstration of flow controller stability was performed by inserting 10%

step changes into the controller at 700 gym and 350 gym. Also, 54 speed changes at high and low turbine speeds were done as part of the tuneup of the system. We system response was satisfactory.

Quick start capability was demonstrated using the manual initiation -

pushbutton on 1H13-P601. S e ability of the system to achieve and maintain 700 gym flow within 30 seconds without RCIC turbine overspeed was satisfactorily demonstrated.

During the test, RCIC steam line differential pressure data was collected to verify the adequacy of the high steam flow setpoint.

RCIC operation from the Remote Shutdown Panel was verified by operating it in the CST to CST mode. 10% step changes were made from 700 gp and 350 gp with the test return valve throttled to provide a discharge pressure of approximately 100 psi above reactor pressure.

All Level 1 and Level 2 acceptance criteria were satisfied.

It should also be noted that all of the TC-2 RCIC Startup Tests were performed subsequent to the installation of " hooded sleeves" on the reactor water level reference legs and that no level anomalies were observed at any point during this testing.

51 i

- . . . , _ _ - , ,_,m., - .,__,__,______.,..,-m._,,..._ , , - - - . . , _ _ _ - _ , - - , - - _ . . . - - , , . _ , - - . _ , , . -

STI-E51-Ol4 RCIC STEP C.MANGrES

..,........g........

i i .

I I ; : . :

i i . : :

i, : i i i .i . i. :

3: : 3: : 3:

g .

. . . . . . .. ...... .. . .. m .

.g .Y[ . - - ........i.....i.....i............i.

i : : : :

RCirc flow .

i i

i i Co&'rROLL.QR l i i i ! i C UTPVY

i i i ! I

. . . . . . ~ .

........y............,

. I; .

..-. l:....

. g

. .I . g i. .

! Sf %ii ! :

i

i i

s i

i i t

: 4lld : BPM : i e

. j :

l

: . : : [. .

j ). :

i

: . : : : ?

. . . . . . . . . . . . . . . .............:......-. 3. 7. 4.. 0. ...F. . .M... .... . _ . . .

j i

i

.i: i

?

.eM ;

[ :

. : : .: . : . .t j : :

.........:......-.....:........j...

........__:.j .. . :.

j g

. j I ! : :* ; !

: a 3

8 i: .

.i -

. . . . . . . . .. ....a.........,.... .......g ...g.........g- . . .g..... . . . . . . . . .

3 . . . g .

8 CEC TURBINE l ! l i !

SPEE05: i

! , i : ;

. .. .. . 1r...

. +

v _-2 = ; ; ; i ; ;

i.i3890 i! ! ! i

! i j

..4..............

4.....i R P M,... .. .... . ,r 3 70

, . G PM

... .l...i,....

\RCIC PDMP :

FL'OW :

i i

l g

l l

! g : .

. a. . : :

i '*

........m . .. .. .... .: I. .. . . .. : .!. 5......%.. :

....S...T...E.

...R....I...N. :

.C....A. 4..E....A.....S.....E...l...

4. .-

m.. : : .

4 g.

gr !. : ig !.

I. !. Q O

.. l [ j I i : i ..

=

~ !

: 1 i o o :

, i. . i. .

52 i

1

STl- E51- 014 RC.IC. STEP CM AMCTES kl Yo ! ! ! !

L i. ..g i ............,g,.........

. . . . . . . . . . . . . . . . . . . . . . . . . . ...........i............g.. ..... ..... s... ..

j i ....R. C.ONT t. .l.C. OL .i. . .F.R. . .O.. .W.f .Q.... .... ... ! 5 OU.TPUT i: !.

! ! . i.

e . . ~ . . - . . . . . . . . . . .

.~ **a.a a a.. . naa........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.......1.. -

.2 .

.. . . . . . . . . . . . . . . . . . ....... . .......... . :: 4. .1.1. C. P.f, .M. .... i... .... ....... ...........!......

.R..c.i.c. ... T.... X DU: R.Bl. 4. E....... :

. :. i S.PE . . .

i : .g...o :

. ; . ~ . . .

- = ,

. RCICi. PUMP Fu.oW.

t . .

............g........

. . . . . . . . . . . . . . . . . .. ..g...................

. g . . . . . . . . . . . . . . . . .g

... ..... ...... . . . . . . . . . . . . . . . g g . . . . . . . .

t . . .

* : ..a.....

Y .

__ ~ . . . _ . . . . . .. . . . . . . . . . .

o~ . .

0.. i i !. !. !. 4

l3 i 8 ..................i . 5.. ..'..o.....S. ..T.;...E.....P.....D...E.C.. ! R... ..E.. 8 A..S....E.. -.3"~

i o

g. ;. .. t . .........ga.ao.... .... . .a .

e .

. e

. .e *. . *

. e .

. *

O g . .

. g .

'.e 6 . 5 8 i

l t

i l

53 l.

i

, - - , , _ .,, - , , , - , - . , - - - . . . - - - - .,_p y e.-,-- ma .,,g ,,,,r-

1 l

l l

STI-E5l-Old. RCIC. QUlcK. STAR.T j ! ! 8

t

: i :

a : si gl2 Psm i e i . I

~

j i i N A R R O W R N Cr i = w eswr ; i

.. i _._

OM E! P E...SS.._U.. RE.. _ ..

I i.............;.....

i -

i ! I i ! ! : i i :-

i ' ! !

i i 4wh i :4300 W di arm C TURE>lNE

! i 3

.gCl$ PEED' i C

i i: : i .

v 1i v : :

_.! ,! ,iRc.it idJ VLV,POS i !

i i' i

s.

siRCic. IND' VLV ,

1 i i ! :

a: : I i m .

i-1:.

i f(eSo RPM i j !. W:i.e:ss:n.5 I W "lSi :

8

-! j L360 i l otN:52 I i ** !Regc ' ulTIATION l ! ! !

i _

! r s A,-us si ! s: -

._ lunitAvioN ! ! ! ! i !

SY"" !* Rc.ic! PUMP l !

O

) -

DISCp PRESS p ,

j i i ! : i- KOW; ; e i i : i: : i

! ! 3! ! ! !

! 2M PS3tr ! '

I !

SI #75 PSIG- t +

i;:

e- 5 y. .. . i. ....,,

j g.

Reic vt ow j k;[ big!g. --- , , ; $OTP

I ! i i.

e

. . i.

i: : i a

i ! -

i I , e

- : I . .

! : I

! ! I

! ?

! I i ! ! ,

i i j i i i .

I i ! RC.1C Q01c.X START -

54

t STARIUP TEST RESULTS (CONT.)

j. STI-C91-019 (Cont.)

l Results Section 8.1, CTP Determination Using the value of core flow (34.5%-TC1, 64%-TC2) obtained from the OD-3 edit, the bounding licensing power was obtained by finding the intersec-tion of the core flow value and the MEOD load line. This resulted in a core thermal power limit of 2362 MWt or 66% of rated for TC1 and 3293 MNt or 92% of rated for TC2, thus, the Level 1 acceptance criteria was easily satisfied.

Section 8.2, Core Thermal Limits Determination The following table shows the values of the localized parameters, their location in the core and the associated limiting value for the reactor conditions at the time of this test.

l~

r Core thermal / hydraulic l Core Location l Limiting Tech. Spec.

parameter l x-y-z

l value i I TC1 MLHGR 4.41kw/ftl 33-36-15 l f 13.4 kw/ft (b)

TC2 5.54kw/ftl 13-46-12 l l 1 K1 MCPR 2.907 l 27-18 l 1 2.011 (c)

K2 2.510 ] 33-44 l 2 1.539 I I El MAPLHGR 3.84kw/ftl 27-36-15 l $ 7.45 kw/ft (d)

TC2 4.87kw/ftl 39-54-15 l f 8.61 kw/ft i

x-y bundle coordinates z bundle node (not applicable to MCPR) 56 I

i

4' STARTUP TEST RESULTS (CCNr.)

STI-C91-019 (Cont.)

NCYTE: S e right column represents the Level 1 Acceptance Criteria.

S e letter to the right in parentheses' indicates which acceptance criteria it is associated with. We values on the left column are the most limiting values calculated by the process computer. The results indicate that all Level 1 Acceptance Criteria were satisfied.

4.10 STI-C85-022 Pressure Regulator System (TSN-138, 139 202, 203) f h e purposes of this test are:

a. To demonstrate the adequacy of the-control settings for the pressure control loop, l

b. To demonstrate the backup capability of the pressure regulators via simulated failure of the controlling pressure regulator,

c. To demonstrate smooth pressure control transition between the turbine i control valves and bypass valves when the reactor steam generation exceeds the steam flow used by the turbine, and i' d. To demcnstrate that other affected parameters are within acceptable l

limits during pressure regulator induced transient maneuvers.

Acceptance Criteria Level 1 i-4 The transient response of any pressure control system related variable to f

any test input must not diverge.

l l

57 l

l .

l l

STARTUP TEST RESULTS (CONT.)

t STI-C85-022 (cont.)

Level 2

-1. Pressure control system related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled arde of response must be less than or equal to 0.25.

2. The pressure response' time from initiation of pressure setpoint change to the turbine inlet pressure peak shall be less than or equal to 10 seconds.

3. Pressure control system deadband, delay, etc., shall be small enough that steady state limit cycles (if any) shall produce steam flow variations no larger than' O.5 percent of rated steam flow.

4.- For all pressure regulator transients the peak neutron flux and/or peak vessel pressure shall remain below the scram settings by 7.5 percent and 10 psi respectively (maintain a plot of power versus the peak variable values along the bounding licensed rod line).

5. 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:

i

% of Steam Flow obtained j

With Valves Wide open Variation l' (Less than or equal to) i

! O to 85% 4:1 85% to 97% 2:1 85% to 99% 5:1 t

l P

58 i - ~ - - -.. . . _ . - . , . - . . . . , _ _ _ - . . _ _ _ - _ _ , , - ___ _. , _ _ _ _ _ _ _,

STARETP TEST RESULTS (CONT.)

STI-C85-022 (Cont.)

Level 3

1. Additional dynamics of the control system outside of the regulator compensation-filters, shall be equivalent to a time constant no greater than 0.10 seconds. This also includes any dead time which may exist.

2. Control or bypass valve motion must respond to pressure inputs with deadband (insensitivity) no greater than t0.1 psi.

Results

'No tests were performed on the Steam Bypass & Pressure Regulator System during TC-1. These were the Pressure Regulator test with Bypass Valves and the Pressure Regulator Deadband test. The latter test verifies that the system, controlling on either the Turbine control valves (TCVs) or the Bypass Valves (BPV's) will respond to a pressure change of i0.1 psi. It was found that normal limit oscillations were greater than 0.1 psi, and no response was observed on either the TCVs or the BPVs. Preoperational test data was reviewed to verify the other Level 3 for control system time con-stants. 'Ihese were found to exceed the 100 msec limit for both the TCVs and BPVs. Both of these deficiencies have been evaluated by General Electric and found to be acceptable.

The Pressure Regulator Test with Bypass Valves verified the ability of the system to control pressure when controlling pressure with the BPVs.

This was done by inputting steps changes of 5 psi and 10 psi into each pressure regulator channel and observing system response. The test showed that the turbine bypass valves can satisfactorily regulate reactor pres-sure by opening or closing as necessary to maintain pressure at or near the setpoint.

59

STARIUP TEST RESULTS (CONT. )

STI-C85-022 (cont.)

The test also demonstrated that a failure of the controlling pressure I regulator channel will result in an automatic transfer of control to the i other channel. Failure of each channel was simulated by inserting'a

- 20 psi step change into the controller. Control transferred automatically to the other channel and plant parameters stabilized without operator intervention.

Based on these two tests, the performance of the Pressure Regulator' system

~

is acceptable. Since all Level 1 and 2 criteria evaluated were satisfied, the Level 3 violations found in the Deadband test does not restrict System operation. The ERIS plot included for the response of Pressure Regulator channel 'B' to a 10 psi step increase shows the systems response. ,

The Pressure Regulator Tests were reperformed during Test Condition 2 as follows.

1. Section 8.1 - Pressure Regulator Test with Control Valves, which consists of performing transients with load set at the high end of the band to allow only the control valves to respond to changes in

^

the pressure regulator demand.

j. 2. Section 8.3 - Pressure Regulator Test with Bypass Valves, which consists of performing transients with the load set down so that the bypass valves respond to changes in pressure regulator demand. This was a repeat of the TC-1 test.

- Section 8.1 and 8.3 also include a simulated failure of the operating pressure regulator channel to the standby channel. This was accomplished by inputting a 20 psi step change into the operating regulator and i observing transfer to the other.

9 60

STAR'IUP TEST RESULTS (CONT.)

STI-C85-022 (Cont.)

Section 8.1, Pressure Regulator Test with Control Valves was performed on 5-12-1987. All Level 1 and 2 Acceptance Criteria were verified satisfac-tory during the performance of this test.

Section 8.3, Pressure Regulator Test with Bypass Valves was performed on 5-11-1987. All Level 1 and 2 Acceptance Criteria were verified satisfac-tory during the performance of this test.

l l

t i

F i

f l

61 t

t STi-C85- 02. 2.

PRESSURE REGULATOR E5 10 PSI STEP INCREASE.

t g

-= _.

/ ~K a .

d I APRM A FLUX i

E MAtM TURStuE. STM Ft.OvJ DEMAMD A ef 9

l l-v

_'2 I I 932'*- I

._ zi r. .--

> X.5% . ,

2e

r. Tor ^'_ eve v'v Pos D PRESS. CONT. $ SENSED PRESS ~s -

g -933PGGr

. woc amu niDOME. PRESSURE qw o%k__ f -

d22Ps4 /, =i s

. v g

,,,l ARvr crura PRExuRc t

_ _i f

k w .. -l-.

..I.t_ _

s_ _ ,

-_..,n-_-

MARROW StAMQE RK DOME PRESSust E.

1 I I

~

- . Tow _dwwetaw

,! _h.,. ' . '

Tarau usuw w to "74 o i 1

/

'.g

=

L .9 f ^ y * ';

2

~-

u .

)) mary / v SV V-v _-- i w 1,g o

I.. 1%*

N L Y N E 8 5 3.so o o m 0 62 l

I STARIUP TEST RESULTS (CONT.)

4.11 STI-N27-023A Feedwater Control System (TSN 118, 119, 120, 147,'148, 182, 188, 189, 190, 191, 192, 193, 204, 205, 206)

The purpose of this test is to verify that the various components of the Feedwater Control System have been adjusted to provide acceptable control of reactor level in all operating conditions.

Acceptance Criteria j

i Level 1 The transient response of any level control system-related variable to any test input must not diverge.

Level 2

1. Level control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.

2. The open loop dynamic flow response of each feedwater actuator (turbine or valve) to small (less than 10%) step disturbances shall be less than or equal to: (8.4)

a. Maximum time to 10% of a step disturbance 1.1 see

b. Maximum time from 10% of a step disturbance 1.9 sec

c. Peak overshoot (% of step disturbance) 15%

d. Maximum Settling time, 100% 5% 14 sec 63

_ ._. , _ . . _ . . _ _ . _ _ _ . _ _ _ _ . - . - . _ _ . _ _ ___. _ _. . ....,_ _ _, _ _ . ~ _

4 STAR'IUP TEST RESULTS (CONT. )

STI-N27-023A (Cont.)

4 -

3. The average rate of response of the feedwater actuator to large (greater than 20% of pump flow) step disturbances shall be between 10 percent and 25 percent rated feedwater flow /second. Wis average response rate will be assessed by determining the time required to pass linearly through the 10% and 90% response points. (8.4)

Level 3

1. ' ne dynamic response of each individual level or flow sensor shall be as fast as possible. Band width must be at least 2.0 radians /second (faster than 0.5 second equivalent time constant), except for the

{!

steam flow sensors which must have band width of at least 1.0 radian /

l second equivalent time constant).

l- 2. Vessel level,.feedwater flow, and steam flow sensors must be installed with sufficiently short lines and proper damping adjustment so that no resonances exist.

3. All auxiliary controls which have direct impact on reactor level and feedwater control (e.g., feedpump minimum recirculation flow valve control) should be functional, responsive, and stable. For example, testing should demonstrate that the minimum flow valve controls are fast enough to avoid pump trips and yet slower than the feedwater startup level controller to avoid possible reactor flux scrams due to l a cold water slug.

i I

l 64 t

- - - _ . - - _ . - . , _ _ . - _ - . . . _ , _ . . _ . . - ~ . _ . . _ . . . . . . , _ . , - - . _ _ - - . . . . _ , . _ - . . _ . - - .

STAR'IUP TEST RESULTS (CONT.)

STI-N27-023A (Cont.)

Results A. N Low Flow Valve Test (8.1)

Section 8.1 was performed using each of the'feedwater pumps. The

-testing involved establishing a feedpump in operation with some recirculation flow and decreasing / increasing 6 inch Startup Reactor Level Controller tapeset changes (controller in AUIO).- For each step change the system's response was monitored and recorded. The data obtained during testing was analyzed to verify system stability and correct controller response. All acceptance criteria was satisfied.

B. N Startup Controller Steps (8.2)

Section 8.3 of STI-N27-023A, reedwater Control System was performed using each of the feedwater pumps. The testing involved initiating six inch increasing / decreasing steps in reactor water level (controller in AUIO). W roughout each maneuver, the system's responses were monitored and recorded. Se data obtained during testing was analyzed to verify system stability and correct controller response.

The only problem identified during this. testing was that for each pump, the feedwater recirculation controller output did not control the feedwater recire valve in Tcl. Wis violated Level 3 criteria number 3. A Test Exception Report was written for each pump. This Level 3 acceptance criteria was verified satisfactorily in TCl during the performance of Section 8.1 of STI-N27-023A, TSN 118, 119, and 120. Attached are ERIS plots of the 6 inch increase and decrease Startup Rx Level Controller tapeset step changes for the TDFPB which are typical.

65

STARTUP TEST RESULTS (CCNr.)

STI-N27-023A (Cont.)

C. FW Pump Open Loop Steps (8.4)

Section 8.4 was performed individually on TDFPA, TDFPB and the MFP approximately 8% and 22% (of pump full flow demand) open loop decreasing and increasing steps allowing the system to stabilize between each step.

n e average rate of response to the 22% steps for all 3 pumps was outside the Level 2 Acceptance Criteria. Also the peak overshoot time response for TDFPB and the 0-10% on 10-90% rise time response for the MFP were outside the Level 2 Acceptance Criteria. All other Acceptance Criteria was met satisfactorily. The Level 2 violations for the MFP were due largely to using Feedwater Flow versus Control Value Position to determine rise time (due to ERIS points not being available for MFP Control Valve Position). TER's were written against all Level 2 violations. Wese tests will be reperformed as scheduled tests in TC3 after further tuneups to optimize the systerrs are completed.

D. FW Master Rx Level Controller Steps (8.5)

Section 8.5 was performed in TCl using Turbine Driven Feed Pump A and Turbine Driven Feed Pump B, each tested individually with the Master Rx Level Controller in Atno. In TC2, TDFPA and TDFPB, MFP and TDTPA, and MTP and TDFPB combinations were tested with the Master Reactor Level controller in Atno. We testing involved initiating six inch increasing / decreasing closed loop steps using the Master Rx Level Controller tapeset which was in AInc. During the step changes the system's responses were monitored and recorded. The data obtained 66 i

STARWP TEST RESULTS (CONT.)

STI-N27-023A (Cont.)

during testing was analyzed to verify system stability and correct controller response.

All acceptance criteria was satisfied.

E. FW Control Linearity (8.6)

Section 8.6 was performed in TC2 for the Motor Feed Pump by adjusting flow to the vessel using the MFP with its controller in MANUAL.

Fourteen steady state data points were collected at MFP flows ranging from 2600-7600 gpm. The maximum slope was determined to be 1.47 times the minimum slope. See attached graph. All Level 3 Acceptance Criteria was met satisfactorily. During the test, the MFP Manual /

Amo Station Flow Demand output ranged between 46% and 65% with a 40%-100% demand output expected. An FCR was written to address this, and subsequent evaluation has determined 100% output demand to be equal to approximately 9000 gpm. Section 8.6 of STI-N22-023A for the MFP is scheduled to be performed again in TC6.

67

STi-N 2.7-o23A TDFP 6 G" IN C.REASE.

1

. A. .l MT SURCA. TAMK Li. VFL .

j [

. l e a MTNEL L LE.V EL., __e !

w RAug (I.VM.A

_/ I TOFP B FLOW Tt VESSEL g r- ^% -

l

~~

.. _ e- .y 4 ED P S NVAJTAT quTPOT ,

Sb; LtvtLE 2, Rod

~

. m. M _1 rw C.

roe tva.

oLLEkouTPUT ' '

... W.r.. - f i

%>W -

q FLOW {.OW DEMAND i --

i 7"C I I !

4 2 8 8

% h I

68

STI-N27 023A TDFP 'B ' 6 " oecns a ss m ._ .._. ... . '... . _ . .. ._

war _ . . _ _

h i ,8

' sun 6E TA V

Lvt

. = -

e

.. . MOTWELL L VL ,

'. m

~ +

~us" ~

nno** RAnne tvtA,l'

%s ^%!

I 119

. .- .. fbFP f f c- _

J - __

~'Gf 6

_.7,,,,:we &~h trmou urp'R-'~ e' i RX LV/. ERROR

~.a y, '

, . _ - r- - f%

t

_. 3.__, .. ..

-- - r -

~+S,f ~ l l ~ 53g .

' ~

FW START 0P LVL s4g% y'

_ . . __ conreotte& ..our.eu.r- .

-// *[ l m 20 o m. .

o O m m N N 69

' B... h-

m. ;

~ a% ..Ig. .

J p,y,# i1 t c h j

, s

, i.

i

. y , - u.io

$._.: _.4+sl a . ., . , -

i -

.-5 I. a t r y

q d

. d .) " "'**

O_ d !

2 I i<'3 S. 3< ? '

r.2 u Y a

Y< s >

. u...

- 9 A ~i eq. a a m

] k.__1_ _l @ 0

__l.

g g_

a e6

< 6- hq h{

u 4 [b 1 d !

.h

w 'a f 25h i [._ L 'S ! _ ,._

J S E!5 9 -t i!: 4 ;

/

gp '

*" ' 3 jI .

_.,_ . i. '

.L .- or v:za

< 1

[ , l j I si f.

,,[.

, *mo 7

q 1 o g h< h Ih6. i. . .

n k 1 .oe:Utoo 1 -_bh $ >k b^

, g-q-

M -- lT ~~*.

' ' a ~--" W u

' . _ . l.

..?

h-g gg

$ h i

es :aeiso t[_,

..._ i.__. -

s

-_._.A. F

, .._. I i ( aJ e{3 t

9

=

2oa o 4

i ! -

. , , , , , , , CG W 20

. . t . . . . . .

n i. . ..AL . -

,y, g 3.,. i . . .n i iiim;i rii-,. rim iim. E " . ..; u...

8 434tMet RAstra Ltytt tesiteLLta 087797 8 8.sse 33. ass 3 434tMf 4 alAtfee StVtt taase Istett

3.sse 3. sos 4 C34tMeg flipegnP e coeftettte ggtry? 3 g,33g . 3g.33g 9 C34tAsti Fatetust a n/A STAftes eWTPTT S 3.303 33.333 8 434tAtti 79fP S Ptew 70 998884 sits /un 3.3ss 3.sas F C34tAFT4 85 AAA80W RAst$ EtVik A BBCutt ? 8.999 l$4.800 0 ClitMF1 Af tM A FLUS S POWta 2,80s Of.sce 9 83FtM18 TDf P G tusalat SPtte APM tag,gge 3g33.333 88 ESPEMll 197P 0 fust 5057 VALVE P983 3 $.333 gg.333 70 1

l

-g J-) _! -ga w e,Y._l q)y_AY v

-l Q-

\

n =

t

=

.%=.,k,,

s w e==m f._ ; di

v. -

_v.'

. l} l l

. ~ .

3 5 J lI i) ,-

s 6

. jt-q< 8- - =-

p - * * 's n F i n,.__2

,,' e I ;,

t>-] , w._ , ,

4 k-f ,

'y

g. ,

, ]'._ $<

wnos i _ -__ ._i, e 7

i

,, l .(,s ,, . ,, ,

, - w.s ai

, l .-l i

li. I %, ,

, i waas

> l g'

f . _.

' ,, q ._

, __.) .

1 . .. A 4'.

0 .

l'. wwn

,k i 2 hi, 1

j i, l,- I -

l a

}(d d y i.l _. , y I) __ ., 4wu j Y 4 ,i

, g- -

9 L I w ..,

t, Es 4- w m.

e t

.[ J ... . , .[ ., o4 18:05 LP S._. e 1 . . . .

. e ~e e e i . . e e

= n se sum m ,g= ;.y amie =,en

~iiimitFilRiiin. E" " ~' e. = . ;;

eu.ememuruamtosmuta mmt e e.a. n .=

a neune uana sms e== e==. e.= e.s=

e n.um enw.= e ===t= ==, e e. . n..=

e u.uan unw.= e m nsn= wms e e. m se. m su uns ween n wout mena e. m e. m e u.eme u n=e= une seu a s ==. e. = s e..

e nium a na er . a .= w.m e musse wo e mera eme m u.m eme.m a unme me e me ear un me e e. m u.m 71

- _~ _ _ .

i STI-N27-023A NFP FLOW LINEARITY i

66 --

64- -

/

s 62 - -

60 - -

i n 5g . .

/

j N/A Stetton

Output 56 - - '

I 2 Flow Demand 54- -

i 52 - -

j 50 - -

48 - -

46 : : : : : : : : : : :

, 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 7500 0000 1

MFP Flow to Vessel - GPM i

STARIUP TEST RESULTS (CCNr. )

4.12 STI-B21-026 Safety Relief Valve - Flow Test (TSN 150)

The purposes of this test are: a) to verify that the safety relief valves function properly (open and close manually), b) to verify that the safety relief valves reseat properly after operation, c) to verify that there are no blockages in the relief valve discharge piping, and d) to demon-strate overall plant stability during single relief valve operation.

Acceptance Criteria Level 1 There should be positive indication of steam discharge during the manual actuation of each valve.

Level 2

1. During the 250 psig functional test the steam flow through each relief valve, shall not differ by more than 10% from the average relief valve steam flow as measured by bypass valve position.

2. During the rated pressure test the steam flow through each relief valve, as measured by PHe, shall not be less than 0.5% of rated twe less than the average of all the valve responses.

3. Pressure control system - related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.

4. 'Ihe temperature measured by thermocouples on the discharge side of the valves shall return to within 10*r of the temperature recorded before the valve was opened.

l 4

73

STAR'IUP TEST RESULTS (CONT. )

STI-B21-026 (Cont.)

Results In 'IC-1, all SRV's were individually manually lifted with the generator on-line. For all nineteen Safety / Relief Valves (SRVs), there was positive indication of steam flow. This was determined by observing a decrease in the generators output, and indication of tailpipe pressure switch actua-tion during each SRV lift. The decrease in generator output for all SRV's was within limits.

The pressure control system operated properly. This was evident by the fact that no oscillatory response was observed during any of the SRV actuations.

One test exception documented that all SRV tailpipe temperatures did not return to within 10'F of pre-lift temperatures. This was not considered significant since 16 SRVs indicated weeping prior to the test. As a result of the test, five SRys showed no change in tailpipe temperature, two showed a small increase (-5'r), ten showed small decreases (5-20*r),

and two showed larger decreases in tailpipe temperatures.

All Safety / Relief Valves operated as designed and no blockages exist in their flow paths. The pressure control system is capable of properly responded to SRV actuations. 'Ihe attached ERIS plot for the actuation of B21-F051G is representative of the plant response.

74

STAR'IUP TEST RESULTS (CONT. )

STI-B21-026 (Cont.)

1 Safety Relief Valve Operation 4

PMe Change in Relief Valve No. Generator Load.

~

1B21-r051D 60.1 1821-r047B 54.0

! 1821-r047F 55.4 '

1B21-r041G 55.0 IB21-r051B 55.3 1B21-r041D 60.7 IB21-r051A 60.9 IB21-r047C 62.0 i

1821-r041K 58.9 1B21-r047G 60.6 1821-r041C 58.8 IB21-r047D 58.6 1821-r051G 58.6 1821-r041B 57.3 1B21-r041E 58.5 1B21-r041r 60.3 1B21-r041A 56.6 1B21-r051C 58.7 1821-r047H 58.2 4

1

! 75

(~ J ! J d l g_'* f .

STi-521-odG,

-){.. a, q 2

SRV ACTUATION z24s oo y,_ t . y .

p q .. 1 l>

i .= >

1

, . h y .,. . .. . . , , -1 -

g) s u..... .! ._

-(

~ . .

'- } i t p i ?

r- t r '

( L~

ac h ..,; t.-

. ;kk. 22:44:30

, e, ,

e i , "~ij u

1 a **

< , rJ. '

cir q '*;iil h

13 j . .

i I ! ! 1. J

~ ~~ ~

(*

., g h[ [l }e l

o YF~ ~ ~ ~

~~~~

'l. .

' ~~

-I [~ 2d

~

j '; b l g . '.; , ,

22M 00 t5 .; .

.... L

..-.f .- . . < ,

_ j - -

t 2 _ ,l _ _ , _ . ,,, =g. , , 4 _

.h .0 . . --~~

r - L, I. , l

- 22CJO i t at i 1

i .

2 . .

. . . . . 9 .. u.

= ,,i. . ,y,= pg .......n

. .e .,Att t Obhtbh NT Ph$k5ee S&h erhe 0. ewe h h6 e n.tuin veta as enu et= mine e.in ..

8 (3484884 80 ausw SAa64 4898L & tatset 8.300 lps,sse 4 434 taste MAsw SAsat la tem Histent tele 8. sos pts. ass 8 (Measet Mile Geef A Mette P9848 Hee 8. ape set.ses 4 99iSAdet Itsta tes4iM 7914 49 Pee 1fles 3 8.ast 33.ame f gettaset stale fgstlN Geef V499 el Pete S 8.8e9 16, des 0 040t&#33 stage essetAtm eeMS Ih4 90 4 88.000 tot,See 76

STARIUP TEST RESULTS (COtfr.)

Generator Load 4.13 STI-B21-027 Reject Within Bypass Valve Capacity (TSN-204)

This test is performed at a power level within Bypass valve capacity to verify that the reactor will not scram and feedwater settings are sufficient to avoid both a level 8 scram and flooding of the main steam lines. This test was scheduled in It2 but not actually performed.

Instead an unplanned turbine trip which occurred due to unrelated testing was evaluated and determined to be sufficient to meet all acceptance criteria for this test. The following analyses were verified.

Acceptance Criteria Level 1

1. Verify feedwater system settings prevented flooding of the Steam Lines,

a. Determine maximum reactor level as recorded by upset range level instrumentation.

b. Verify maximum level is below main steam line level.

Level 2

1. Verify for the Generator Trip within the bypass valve capacity, the reactor did not scram.

2. Verify feedwater level control avoided loss of feedwater due to a high level (level 8) trip during the event.

77 l

t

I STARIUP TEST RESULTS (CCNT.)

STI-B21-027 (Cont.)

a. Determine maximum level measured on narrow range level instruments during the transient.

b. Verify maximum level is below level 8 feedwater pump trip setpoints.

All acceptance criteria were met. Maximum reactor level reached by the narrow range level instrumentation as measured by the ERIS computer was 198 inches. Normal level is 197 inches. Reactor level by upset range instrumentation peaked at 197 inches. The reactor did not scram.

Qualitatively, this was a very minimal transient on the reactor. Pressure rise was minimal ~10 psi, level fluctuated only about 6 inches, power peaked at approximately 12% greater than initial briefly, and all other parameters showed little change.

'Ihe differences between the unplanned transient and the test were that:

1. The unplanned transient was a turbine stop valve closure versus a turbine control valve fast closure.

3

2. The initial reactor power at the start of the transient was 25% of rated versus 30% of rated if the test were to have been actually performed.

, 3. Due to other plant conditions, Main Steam Line 'C' was isolated.

The unplanned turbine trip was evaluated by the NSS vendor, General Electric, and determined to be a sufficient substitute for the generator load reject test. The reason for the unplanned turbine trip was investigated and corrected as a separate issue. (Reference Condition Report #87-240) i 78

STARWP TEST RESULTS (COffr. )

4.14 STI-C61-028 shutdown from Outside Control Room (TSN-169, 170)

The purpose of this test was to demonstrate that in the event of an evacuation of the Main Control Room, the Perry Power Plant has the instructions, equipment, and controls to safety shutdown and cooldown the reactor froc outside the Main Control Room.

Acceptance Criteria Level 2 During a simulated control room evacuation, the reactor must be brought to the point where cooldown is initiated and under control, and the reactor vessel pressure and water level are controlled using equipment and controls outside the control room.

Results

1. 8.1 - Shutdown from outside the Control Room The test was performed by evacuating the Main Control Room with the reactor at 24% power. Only the minimum shift complement as defined by plant Technical Specifications, were used in this test. Safety monitors were stationed throughout the Main Control Room and at selected plant locations. Actions to protect equipment (such as placing the main turbine on the turning gear) were allowed but were documented and evaluated to ensure they did not adversly affect the test. A plant operator was dispatched to W-1-A & B and W-1-A & B were then opened, momentarily, causing a manual scram and completing ONI-C61, Evacuation of the Control Room.

79

i STARIUP TEST RESULTS (CONT.)

j STI-C61-028 (Cont.)

10I-11, Shutdown from Outside the Control Room, was then entered and control of appropriate plant equipment was transferred to the Division 1 Remote Shutdown Panel (RSP). After transfer, the following equipment was started: ESW "A", ECC "A", RCIC in the CST to CST mode and RHR "A" in suppression pool cooling mode. Einally Reactor Pressure was decreased by ~150 psig with the opening of SRV 1821-r051C. The test went as planned, maintaining the plant in hot shutdown for more than >30 minutes from the RSP.

Prior to the start of testing, one instruction related deficiency was identified. This was in determining full insertion of all centrol rods, before reclosing the EV-1-A & B breakers. The problem was a control rod with an inoperable position indication channel, causing a scram event tag on the RSP ERIS Terminal to indicate control rods stuck out after the scram. Since it was identified prior to the evacuation, OD-7 was demanded from the process computer which, after an approximate 6 min. delay, was run using the Monicore terminal and all control rods were verified to be full-in.

2. 8.2 - Cooldown from outside the Control Room

'Ihis test commenced prior to completing STI-C61-028, 8.1, with necessary plant equipment already operating from the RSP and the reactor at - 605 psig. Reactor pressure was reduced to - 95 psig, using each of the three SRV's and RCIC, while maintaining RPV level using RCIC, with the excessive water in the suppression pool trans-ferred to radwaste. RHR "A" was flushed, warmed and placed into shutdown cooling from the RSP.

Reactor temperature was lowered by ~68'r, at which point the test was concluded. The RSP ERIS terminal was extensively used, especially to monitor RCIC discharge pressure.

1 80

i-STARIUP TEST RESULTS (CONT. )

(

t STI-C61-028 (Cont.)

One test exception report (TER-170-1) was generated when Control Room personnel reset the isolation signal to valve lE12-F049. This allowed the valve to remain open following manual opening. A change to the operating instruction will correct this potential problem.

4.15 STI-B33-029A Recirculation Flow Control - Valve Position Loop (TSN-175)

The Position Loop and Deadband Test in TC-1 was performed to verify stability in the inner most control loop of the recirculation system.

Acceptance Criteria Level 1 The transient response of any Recirculation system-related variables to any test input must not diverge.

Level 2

1. Recirculation system related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response muct be less than or equal to 0.25,

2. Maximum rate of change of valve position shall be 10 1%/sec.

3. Overshoot after a small position demand input (1 to 5%) step shall be

<10% of magnitude of input.

I 81

STARTUP TEST RESULTS (COtfr. )

STI-B33-029A (Cont.)

Level 3

1. Position loop deadband shall be <0.2% of full valve stroke.

Small steps were input to the A and B recirc position loops to change flow control valve position. No divergence was observed thus satisfying the Level I criterion. The maximum velocity of the valve actuators was greater than 11% per second which was out of the Level 2 Criterion band.

A Test Exception Report (175-1) was written to address this problem. The maximum instantaneous velocity obtained was 17% per second, however, the average valve positions velocity over 10 to 90% of the position change for the 5% step changes were all on the order of 10% per second. The resolu-tion to the TER was to perform a B33 tuneup of the FCVs as scheduled in TC-3 on Fast Speed Pumps.

Both Flow Control Valves were tested to have a deadband less than 0.2% by observing response to a 0.1% step input.

Included is a plot of a +5.0% step change to 'B' position controller, i

l l

l

! 82

1 i I am-b33-ot.% Fcv b no% STEP 3 I I i _

j' 8-wsneo= 0 t enz.

I RECRC FCV B PotlT5M l l l _

/

J l w m

_1_il

/ _!__

-- ,!z -

I L, .= = - - . - - ' -

__ i I

L- m' m.tw #-

STl-833-o29A FCV 6 \O% STEPS 83

STARIUP TEST RESULTS (C0hT.)

i 4.16 STI-B33-030E Recirculation System Cavitation (TSN-207)

The purpose of this test is to verify that no recirculation system cavitation will occur in the operable region of the power-flow map.

During 60Hz recirculation pump operation, the FCVs' NPSH is provided primarily from the subcooling effect of the feedwater. The required NPSH is proportional to the differential pressure across the FCVs, which is a function of FCV position. At low reactor power with the FCVs near their minimum position, cavitation could occur. A cavitation interlock is provided to automatically transfer the reactor recirculation pumps to the LFMGs, avoiding conditions conductive to cavitation. This interlock actuates when feedwater flow is less than -22.3% of rated.

Acceptance Criteria Level 1 None Level 2 ,

Recirculation cavitation interlocks shall have settings adequate to prevent operation in areas of potential cavitation.

Results With reactor power between 35% and 45% and the recirculation FCVs at their minimum positions, reactor power was reduced, via control-rod insertion, until the interlock was reached and the recirculation pumps transferred to-l the LENGs. .No indications of cavitation were observed during the test.

The "A" pump tripped to off instead of downshifting to slow, TER 207-1 was issued to document this problem.

84

STARIUP TEST RESULTS (COtTI.)

STI-B33-030E (Cont.)

he transfer occurred at 3.41 mlbnyhr (22%) of feedwater flow and at approximately 24.7% core thermal power.

W e recirculation system's low feedwater cavitation interlock prevents cavitation from occurring on the Flow Control Valves.

4.17 STI-R43-031 Loss of Turbine-Generator and Offsite Power (TSN-208)

The purpose of this test is to determine the reactor transient performance during the Loss of the Main Turbine Generator and All Offsite Power, and to demonstrate acceptable performance of the plant electrical supply system. This test was performed by tripping the main turbine at approxi-mately 25% rated core thermal power and simultaneously tripping both startup supply breakers to busses L10 and L20.

Acceptance Criteria Level 1 Reactor protection system actions shall prevent violation of fuel thermal limits.

All safety systems, such as the Reactor Protection System, the diesel-generators, and HPCS must function properly without manual assistance, and HPCS and/or RCIC system action, if necessary, shall keep the reactor water level above the initiation level of Low Pressure Core Spray, Low Pressure Coolant Injection, Automatic Depressurization System, and MSIV closure.

Diesel generators shall start automatically.

85

l STAR 7UP TEST RESULTS (CONT.)

STI-R43-031 (Cont.)

The turbine steam bypass valves shall remain operable until the MSIVs are closed or until the low condenser vacuum signal closes the bypass valves.

If any safety / relief valves open, no more than one valve shall reopen after the first blowdown.

An MSIV isolation trip event due to RPS MG set coastdown shall not occur for at least 2 seconds after the initiation of the transient.

Level 2 Proper instrument display to the reactor operator shall be demonstrated, including power monitors, pressure, water level, control rod position, suppression pool temperature, and reactor cooling system status. Displays shall not be dependent on specially installed instrumentation. Temporary interruption of instrument display is acceptable, provided the operator has sufficient information available to properly assess the plant state for long term operation.

If the low-low set pressure relief logic functions, the open/close actions of the SRVs shall occur within 15 psi and 20 psi of their design setpoints, respectively.

NCTTE: Although the acceptance criteria call for reclosing of the SRV's within 20 psi of their setpoints, plant technical specifica-tions are more restrictive at 15 psi.

If safety / relief valves open, the temperature measured by thermocouples on the discharge side of the safety / relief valves must return to within 10'F of the temperature recorded before the valve was opened. If pressure sensors are available, they shall return to their initial state upon valve closure.

86

STARIUP TEST RESULTS (C0tTr.)

STI-R43-031 (Cont.)

Results

1. 8.1 - Loss of Turbine Generator and Offsite Power This test was performed by tripping the main turbine at approximately 25% rated core thermal power and simultaneously tripping both Startup Supply Breakers L10 and L20.

As expected, an MSIV isolation signal was received due to a loss of leak detection power and the reactor scramed as the MSIV's were going closed. All three diesel generators auto started on under voltage. The main generator tripped on reverse power and the loss of the RPS caused a BOP isolation. SRV's F051D and F051C (on Lo-Lo set) opened and closed after -24 seconds. .Later, SRV F051D opened again for -28 seconds. RCIC was manually initiated to control reactor level. After 30 minutes elapsed, L10 and L20 were reenergized and plant restoration began.

TER 208-1 was initiated when 2 SRV tail pipe temperatures (F051C and F051D) did not return within 10*F of their original temperature after reclosure. (Note 4 other unopened SRV's showed high post event temperatures due to the pressure transient). This was a Level 2 acceptance criteria violation and has previously been evaluated.

All systems functioned satisfactorily with minor exceptions:

1. P47-B001A chiller tripped on low condensing flow. Work was initiated to troubleshoot and repair this problem.

87 1

STAR 7UP TEST RESULTS (CCNr.)

STI-R43-031 (Cont.)

2. Scram discharge volume drain valves 1C11-F010 and 1C11-F181 showed intermediate position after the scram. This problem was determined to be with the limit switches - they were fixed and retested satisfactorily on 5-12-87.

3. Div. 2 diesel generator bus indicated load oscillations. This was determined to be a governor problem and governor adjustments were completed and the diesel generator retested satisfactorily on 5-12-87. None of these problems encountered had adverse impact on satisfying performance of this test.

a T

88

5 'a d To l :

i. .. . i.

~

I I.

l

... MAtu Tumaiut . ... i .. .. !

i . .

ToTAI %YPAbis 8

! vAi.vt PosiTiou i:

I. -

i i. : :.

I I ! !. .. !.

j ! 1 8

l

muuGeo= t.o SEcoue :

l 1 i -

. , i 3 : l : : 4 e j )jo% ' !;-

.o. - !. --

g .

oY. I i :

I l !.

[ .! I I I ! ! TI j OUTeo*Ab M5iv PoWThod a f j j ?

7; state.oei7 i 7:

. s......-....

l :

! ! anammomsw I 8 I.

! : 2 mur.im am u o.s . :

6' 6' I 6: '

6*

i -

i i i .

8 i ! t' i i Si

}

oertemme utev memaa is :- i , aute.oaa 4 . ,

E ! 5! 5 - :

.g...

3 : : :

i : i i !

! ; : : 31 g 1 ..~. ~.iv mes u i m2ite. oT8 9

1. ! 3: ! / Si

.i .!

2.-

! iutoama www posmou l i g mast.c.07.5 i ! :

E : 2i 21 ! ,

I i-i i i !

i + :

i woo.=a m.v = =

I ! ; -

,i azice ovop , i i i :

i j .

: i

i. . i I :

i :

i i:

i : i.

STi-R43-o31 MSM POSITioMS 1

89

1 -

i i 1

p ,

l q+

i t

i.-.,,.. . STi-R43 031

. g . , i. iJ SRV a I e

f ACTUATION 3

t

) : ci, '

_ _ __.) .

I k 'O kq s ./ -

-i i s L-- $3nv _.s A

K, Z---4 "* ,

1 f

-+ ._ ,

D ' R h =

4 .L_. 4..... .4. , , , . _ . , , _ , _ , , , , , , , , , , ,

I .__

I

N. .! 11 U, i s

[

. j _,. ; _.

_4 i .

g

\ d a s- -.

a 2 j * .2 ,

O -- ~

-l - E J!

yJ f l-)

p p

3 _ .

l - . ,

k j .

g .. ._ ;_.

i i

. 2 1 3 I t.

I mm 2s '

.;_ l -. 1!q _

_x, .

I y

l i

_ .. 4._ .i..

. : , . , j , . -

7 8

- un + - . - - .

WOP '

t;--. .

__ f, i m s, _ _,,

I I. h h 'd 1 >. 3 . . _ , / ', . _ - .

e

F- l.

=

i 3, i s :

l l l l

' .!*g ',, I i 90 l

l

\

STARIUP TEST RESULTS (CONT.)

4.18 STI-B21-033 NSSS Vibration STI-P99-122 BOP 'Ihermal and Vibration Steady-State

1. Vibration Power Plateau TSN 124 - 8.1 NSSS Steady State Vibration Power - Plateau STI-B21-033 NSSS steady state remote vibration readings were recorded using lanyard potentiometers through the ERIS computer at 20%-30% of rated steam flow. This test verified that the NSSS piping scope of. recirculation, main steam and RCIC steam piping vibration was within acceptance limits during steady state operation with main steam flow at the specified range. Piping response was recorded and analyzed using the ERIS computer. Strain gage signals were also nenitored and determined as acceptable although not required by the test specification. Recorded vibration levels were well within the Level 2 Acceptance Criteria values.

2. TSN 125 - 8.3 BOP Expansion and Vibration - Power Plateau STI-P99-122 Designated portions of BOP Piping were remotely monitored for steady state vibration using lanyard potentiometers and acceler-ometers through the ERIS computer. The Test Specification called for the test to be run at steady state conditions with the plant at a reactor power level between 20% and 30% of rated power. Sensor signals were processed through the ERIS computer and PCM Tape Recorder. Statistical analysis, time history plots and time series analysis reports were generated for data analysis. Peak vibration levels were determined for comparison to criteria. The vibration levels measured during this test were within the Level 1 and 2 displacement and acceleration limits. j l

91

.r -

STARETP TEST RESULTS (CCNr.)

STI-B21-033 and STI-P99-122 (Cont.)

l

Transient Manual Discharge

3. Vibration of Safety Relief Valve at Rated Pressure TSN 153 - 8.3 NSSS Transient Vibration - SRV Manual Discharge Test STI-B21-033 Remote monitoring of all NSSS vibration and strain sensors on the main steam, recirculation and RCIC Piping during the per-formance of the SRV Manual Test (STI-B21-026, TSN 150). All remote sensor signals were recorded during the discharge of each SRV to evaluate vibration and strain levels of the associated piping. Signals were analyzed through the ERIS computer using time history plots and the statistical analysis reports to determine the peak vibration displacement or strain at the time of valve discharge. Results of the analysis showed that all sensors recorded vibration values which were within the limits of the Level 1 and 2 Acceptance Criteria.

Transient Vibration Manual Discharge

4. and Thermal Expansion of Safety Relief Valve at Rated Pressure TSN 154 - 8.2 BOP Expansion and Vibration - SRV Manual Discharge STI-P99-122 Remote monitoring of designated portions of BOP Piping for transient vibration and thermal expansion during the discharge of selected safety relief valves was performed. At the time of the SRV discharge, the ERIS computer was used to record vibra-tion and thermal expansion signals from lanyard potentiometers

. mounted on the SRV discharge tailpipes, SRV vent lines, and SRV local instrumentation lines. A pulse code modulation (PCM) tape recorder was used to record signals from accelerometers mounted I

92 L

. - - . - - - , _ - . _ _ _ . , _ , _ . . . _ , , . . , - _n. . - . , , - - . , , , - . _ , , , ,

STAR'IUP TEST RESULTS (CONT. )

STI-B21-033 and STI-P99-122 (Cont.)

on the same associated piping. Recorded signals were analyzed using statistical analysis and time history plots, and values of vibration and expansion were coupared to criteria. Lanyard vibration and expansion were measured directly as displacement in inches, while accelerometers provided acceleration in "g's".

~ Analysis methods using the ERIS time series analysis provided fnr the conversion of acceleration to displacement when

. quired.

Level 1 violations were identified on a total of nine sensors.

Two lanyard pots recorded displacement and seven accelerometers recorded acceleration which exceeded criteria. These violations were on the 10 inch discharge piping (5), the 2 inch SRV bonnet vents (3), and the 3/4 inch discharge pressure instrument lines (1). Three Level 2 violations were also identified, one for acceleration and two for thermal expansion which was less than predicted. The Nuclear Engineering Department is presently _

analyzing the recorded data to determine the required course of action. Presently, it appears that the resolution may involve the installation of lanyard potentiometers at selected loca-tions, and a retest to validate the acceleration / displacement values recorded during this test. This work is scheduled to be complete prior to entry into 'ICS.

Steady-State and Transient

5. Vibration and 'Ihermal Expansion Steam Condensing TSN 156 - 8.4 BOP Expansion and Vibration - Steam Condensing STI-P99-122 Performance of this test required visual and remote monitoring of BOP Piping for thermal expansion and steady state / transient 93

STARWP TEST RESULTS (CONT.)

STI-B21-033 and STI-P99-122 (Cont.)

vibration during initiation and operation of the steam condensing mode of the residual heat removal system. Testing consisted of both visual inspection by teams of Qualified Piping Inspectors (QPI) and Qualified Piping Engineers, (QPE) and also remote monitoring using lanyard potentiometers. Piping scope inspected during this test consisted of portions of the RHR (E12) and RCIC (E51) piping and supports.

Visual and remote transient vibration testing was performed during the initiation of steam to the RHR heat exchangers. This was performed on two separate occasions, once for each loop.

Because of the possibility for waterhammer due to condensate in the system, this portion of the test presented the greatest concern prior to testing. In fact, the initiation was virtually undetectable due to the slow opening of the E12-F051 and F052 valves. This warning is provided in the system operating instruction, and proved to be successful in preventing water-hammer and possible resultant system damage. Following initia-tion, steady state vibration inspection was performed with the heat exchanger between 200 and 220 psig in the suppression pool lineup. Visual inspection was satisfactory, and although not required, remote monitoring was also performed and results were satisfactory.

Visual thermal inspection was performed to measure the expansion of the system at operating conditions. 'Ihis was attempted during the suppression pool lineup on Loop A. Review of the data indicated that the system was not in an appropriate lineup for test performance, as identified by several Level 2 94

STARIUP TEST RESULTS (CCNr.)

STI-B21-033 and STI-P99-122 (Cont.)

violations. It was decided to reperform this step for both loops when the system was in the RCIC lineup, with steam condensing at maximum capacity. At these conditions, two separate walkdowns were again performed on Loop A and B. Visual inspections were made for thermal expansion and steady state vibration. Vibration results were satisfactory. However, thermal expansion results identified Level 2 violations on both loops. 'Iwo test exception reports were issued accordingly.

During dual loop operation, remote testing was performed for vibration and thermal expansion. Vibration results were satis-factory and thermal results again identified several Level 2 violations. Wese violations were previously identified during earlier thermal tests, and a test exception was issued to document the violations in this test, overall, vibration results for this test were satisfactory and demonstrated the ability to initiate steam condensing while ,

avoiding waterhammer. Thermal results indicate deviation from predicted pipe response. These exceptions are with NED for resolution. The exceptions do not presently restrict plant operation and are scheduled for resolution prior to entry into TC5.

Se visual portion of this testing utilized sixteen Qualified Piping Inspectors (QPI's) for the transient tests and six QPI's for the steady state tests. Wey were supervised, and the data reviewed, by three Qualified Piping Engineers (QPE's) at each test step.

95

STARTUP TEST RESULTS (CCNr.)

1 STI-B21-033 and STI-P99-122 (Cont.)

Transient-

6. Vibration Remote Shutdown from Outside Control Room TSN 174 - 8.3 NSSS Transient Vibration - Remote Shutdown STI-B21-033 Performance of this test required monitoring the NSSS scope of main steam and recirculation piping for transient vibration /

strain during the remote shutdown from outside the Control Room.

Lanyard potentiometer and strain gage signals were analyzed through the ERIS computer to determine peak values for comparison to criteria.

'Ihe shutdown transient is actually a series of events rather I than a single event, such as an SRV lift. Data was collected

for fifteen minutes after the scram and reviewed to determine which events were significant with respect to pipe vibration.

This review identified the scram and turbine trip as causing the highest vibration levels. Accordingly, the highest vibration level was used for comparison to acceptance criteria.

A single exception was identified for violation of Level 2 strain limits on the recirculation loop piping. The strain was actually induced by thermal expansion of the piping rather than the transient vibration of interest. The condition was evalu-ated by General Electric Engineering and determined to be acceptable. With this exception noted, results were satisfac-tory and well within the Level 2 allowable limits.

r i 96 l

l

STAR'IUP TEST RESULTS (C0tfr. )

STI-B21-033 and STI-P99-122 (Cont.)

Transient Vibration Remote Shutdown

7. and 'Ihermal Expansion from Outside the Centrol Room TSN 176 - 8.2 BOP Expansion and Vibration - Remote Shutdown STI-P99-122 Performance of this test required remote monitoring of lanyard and accelerometer sensor signals for transient vibration and thermal expansion during performance of the remote shutdown from outside the Control Room. Vibration Signals included accelero-meters and lanyard potentiometers on drywell/ steam tunnel piping and instrumentation, and RCIC piping. Thermal expansion was monitored on the CRD insert / withdraw piping and discharge volume.

J piping.

i Peak levels for vibration and expansion were determined in identical fashion as the NSSS piping, by reviewing sensor response for a period of time after the scram.

A single exception for excessive vibration was identified on a

). . vessel level instrument line. This exception is with NED for engineering evaluation and resolution. With the exception noted, all test results were satisfactory and within Level 2 Criteria limits. Of particular interest was the thermal response of the CRD piping following the scram due to the short duration of the transient. The piping was found to expand in a manner consistent with the predicted analytical expansion.

- 97

-- ._ _ . - _ _ . _ . . . . . _ _ _ _ . . _ _ _ . . ~ _ _ . - _ _ _ _ _ _ - . _ . _ - . ~ _ . . - , . . .

STARWP TEST RESULTS (CONT.)

STI-B21-033 and STI-P99-122 (Cont.)

8. Thermal Expansion Reactor Heatup Retest TSN 178 - 8.1 BOP Expansion - Reactor Heatup (Retest)

STI-P99-122 Performance of this test was required due to a Level 1 violation for thermal expansion identified on a remote sensor located in the steam tunnel. During a previous plant heatup, the sensor indicated improper expansion of the Main Steam Drain Line and a Test Exception (TER-097-01) was issued. Based on analysis of the exception a design change was issued and several supports on the subject line were modified.

Results of the retest indicated that the modifications achieved the desired effect. Expansion of the drain line was satisfac-tory meeting Level 1 and 2 criteria limits. The TER Action Item was closed accordingly.

9. Transient Vibration RCIC Trip / Restart TSN 184 - 8.2 BOP Transient Vibration - RCIC Trip / Restart STI-P99-122 The purpose of this test was to demonstrate, by means of visual inspection and remote ERIS signals from lanyard potentiometers, that transient induced piping vibration levels were within acceptable limits for designated portions of the E51 system during the RCIC Trip and Restart. Visual inspection was per-formed by Qualified Piping Inspectors (QPI) under the super--

vision of Qualified Piping Engineers (QPE). Inspectors are Nuclear Engineering Department (NED) Engineers and designers who have been trained on inspection requirements and qualified in the Startup Test Program as QPI's and QPE's. A total of ten (10) QPI's and four (4) QPE's were involved in the performance of this test.

98

STARWP TEST RESULTS (CNr. )

STI-B21-033 and STI-P99-122 (Cont.)

Testing methodology for the visual inspection required the QPE to determine the maximum viewing _ distance for individual seg-ments of the piping system, based on allowable bending stress.

Inspectors are positioned at this distance prior to the tran-sient initiation. During the transient event, if motion in the piping system is not detectable to the inspector, the stresses are acceptable and the criteria has been met. Test methods for the remote monitoring required recording the vibration signal from the lanyard potentiometer. This recorded signal is then analyzed using time history plot and statistical analysis func-tions of the ERIS computer. The maximum transient displacement is determined from this analysis and compared to criteria for acceptability.

Vibration levels were minimal and well below the Level 1 and 2 criteria limits. Visual observation showed no sign of vibration or waterhanmer.

Steady State

10. Vibration / Strain Reactor at 50% Rated Steam Flow TSN 186 - 8.1 NSSS Steady State Vibration / Strain - Steam Flow at 50%

STI-B21-033 Performance of this test required the evaluation of vibration /

strain levels on NSSS piping with the reactor at 50% of rated steam flow. Data was collected using remote monitors on the GE scope of the Main Steam and Recirculation piping. Lanyard potentiometer and strain gage signals were processed through the ERIS computer and analyzed to determine vibration / strain magni-tude for comparison to acceptance criteria. i l

99

STARTUP TEST RESULTS (CONT.)

STI-B21-033 and STI-P99-122 (Cont.)

Data archival was executed at the required steam flow, however, subsequent analysis revealed that the required data was not recorded due to an electrical problem in the ERIS. It was determined that this cendition had been caused by concurrent testing of the building lighting circuits. Coincident with this discovery, the reactor was scrammed due to a feedwater problem, and the plant was placed in Operating Condition 3. The test was suspended with a plan to reperform the test step when the plant returned to 50% steam flow.

Prior to the restart, a problem with the Main Steam Leakage Control (E32) system prompted the decision to complete TC2 with the main steam line "C" isolated. 'Ihis condition rendered the proposed reperformance of this test invalid. It was realized at this time that the required data for this test had also been recorded during the performance of TSN 187, BOP Piping Test -

Reactor at 50% power. Review of the data from TSN 187 showed that steam flow was only slightly outside of the required band (43.8% vice 45%-55%). It was decided to complete analysis using this data, and issue TER 186-1 to document the inadequate steam flow and seek GE approval of test performance at this condition.

All vibration / strain results were satisfactory. The levels were below the Level 1 and 2 acceptance criteria by a wide margin.

GE review indicated that the test results at 43.8% steam flow were acceptable for satisfying the 50% steam flow test.

i 100

STAR W P TEST RESULTS (CONr.)

STI-B21-033 and STI-P99-122 (Cont.)

2 Steady State vibration

11. and Thermal Expansion Power Plateau TSN 187 - 8.3 BOP Vibration and

'Ihermal Expansion - Reactor at 50% Power STI-P99-122 The purpose of this test was to verify that vibration and thermal expansion of designated portions of BOP piping are within acceptable limits during steady state operation at a reactor power level of 50% of rated. Similar testing has been performed at the 25% level and are planned to be performed at

! the 75% and 100% power levels.

Piping systems were remotely monitored with lanyard potentio-meters and accelerometers. Sensor signals were processed through the ERIS computer and Pulse Code Modulation (PCM) tape recorder. Statistical Analysis Reports, Time History Plots, Time Series Analysis Reports and ERIS EXPAND reports were generat:d for data analysis. Peak vibration levels and position due to thermal expansion were determined for comparison to acceptable criteria.

Results from the analysis showed that all vibration data recorded at 50% power were within the Level 1 and 2 criteria limits and were satisfactory. Level 1 analysis on thermal expansion data identified three remote sensors on the main steam drain lines which recorded expansion outside of the acceptable limits, as recorded on TER-197-1. A 50% power restriction was imposed on plant operation while the Nuclear Engineering Department (NED) evaluated the violations. One sensor was found l

l o

l' 101 l

STARIUP TEST RESULTS (C0tTr.)

STI-B21-033 and STI-P99-122 (cont.)

damaged in the field, with retert specified at 100% power. The remaining two sensors require further analysis and disposition, however, NED did determine continued operation was acceptable and the power restriction was lifted. A PORC Action Item has been assigned to complete the required analysis and retesting.

Level 2 analysis identified six additional thermal criteria violations, as recorded on TER 187-2. These violations remain with NED for final resolution. ,

4.19 STI-E12-071 Residual Heat Removal System (TSN-151, 152, 163)

1. Suppression Pool Cooling (8.2) he purpose of this test was to demonstrate the ability of the RHR system to remove heat from the Suppression Pool. The Suppression Pool Cooling tests for BHR Loops A and B were performed individually in parallel with STI-B21-026, Relief Valves. 21s ensured that an adequate temperature differential existed between the suppression Pool and Emergency Service Water. RHR Loop A and B were run individu-ally in Suppression Pool Cooling for 20 minutes intervals while data was collected. The data was used to calculate the heat transfer across the RHR Loop A and B Heat Exchangers. The results were then compared to a pre-calculated value representing the required heat exchanger performance during accident conditions.

Acceptance Criteria Level 1 None 102 I i rig i -

STARWP TEST RESULTS (CONT. )

l STI-E12-071 (Cont.)

i unl 2

he RHR System shall be capable of operating in the Suppression Pool Cooling Mode. Each RHR loop shall be independently tested in this mode.*

It shall be demonstrated with either suction from the Reactor vessel or Suppression Pool that each heat exchanger can remove 166.4 x 10 6B W/HR when the inlet and outlet conditions in mode B-1 of FSAR Figure 5.4-14 exist.

RHR Loop A and B Heat Exchangers were demonstrated to be capable of I removing 166.4 x 106 f B W/HR, thus satisfying the Level 2 Analysis criteria. 21s was accomplished through the calculation of RHR Loop A and B Heat Exchanger heat transfer rates at lower Suppression Pool tem-peratures than design conditions and then extrapolating the data. he UA t (overall heat transfer coefficient) value calculated based on the test results was compared to the UA value calculated for design conditions.

i. h e UA values calculated from the test were all greater than 2.83 x 10 6

.B W/HR*F, thus satisfying the Level 2 Analysis Criteria. Also attached are the data tables used to perform the Level 2 Analysis.

Heat Rate = heat removed from RHR System in MBW/hr

= ESW from Mx enthalw - ESW to Hx enthalW (BW/lb) ,

3 Specific volume (ft /lb) 1:

3 ft x ESW flow (galanin) x 60 bx 7.48 gal hr 106 BW i.

k 103 l

l

STARWP TEST RESULTS (COtfr.)

STI-E12-071 (Cont.)

< Substituting instrument numbers and simplifying: <

Heat Rate = 5 enhlw - P45EA001 (P45EA005) enhlw x R602 x Specific volume 8.02 x 10-6 where the specific volume is at the ESW pressure and the average of the heat exchanger ESW side inlet and outlet temperatures.

0 Og = 166.4 x 10 BW/HR

" (T -T ) - (T -T )

RHR IN ESW OUT RHR CXJr ESW IN T -T l .

LN l RHR IN ESWOUTl

l. T RHR M ~T ESW IN where: (In Mode B1) TRHR IN = 5r TRHR OU'r = 138.1*r TESW IN = 0'r.

T ESW M

= 125.6*r utTD = (185 - 125.6) - (138.1 - 80) = 58.74*r 185 - 125.6

138.1 - 80 6 6 UA = OHX = 166.4 X 10 BW/hr = 2.83 x 10 BW/hr*r urID 58.74'r l 104 4

i

STARWP TEST RESULTS (CCtfr. )

STI-E12-071 (Cont.)

Suppression Pool Cooling Data Data for RHR Loop A l l l Time l l Parameter l Units l 2011 l 2016 l 2021 l 2026 l 2031 l

l A N to W l 7 l 37 ! 37 I 37 ! 37 I 37 !

l P45-R091A(B) l l l l l l l l ESw fm ax l I I I I I I l B A-E12-N005A l 'F l 64 l 68 l 67 l 67 l 66 l l B-E12-N005B l l l l l l l

! C ESW flw l gg, l 7300 ! 7300 ! 7300 l7300 ! 7300 !

l E12-R602A(B) l l l l l l l

! o ESW Press l .F ! 90 I 90 ! 90 l 90 I 90 l l P45-R102A(B) l l l l l l l l ESw to ux I I I I I I I

l E E12-N004A(B) l 'T l 98 l 100 l 100 l 100 l 99 l l 1E12-R601(pt.1,2)l l l l l l l l Raa fm ax 1 I I I I I I l F E12-N027A(B) l 'T l 69 l 69 l 69 l 69 l 68 l l 1E12-R601(pt.5,6)l l l l l l l

l RHR F1w l gp, l6800 ! 6800 ! 6800 ! 6800 I 6800 !

l E12-R603A(B) l l l l l l l l c Enthalpy In ! ! 5 .33 ! 5 .33 ! 5 .33 I 5 .33 !

l l BW N l 5.33 (at A 'F,D-psig) l l l l l l Enthalpy out ! 36.34 ! 35.35 ! 35.35 I 34.35 I

! H lBW/lb+! 32.35 l (at B *F,D-psig) l l l l l l l I I ^v9 T*:P I 'r I 50.5 I 52.5 l52.0 l52.0 l 51.5 l l 1/2(A+B) l l l l l l I 105

STAR W P TEST RESULTS (CONT.)

STI-E12-071 (Cont.)

Suppression Pool Cooling Data (Cont.)

Data for RHR Loop A C

J MC ft/lb+!.01602!.01602!.01602 !.01602!.01602 !

l (at I *F,D-psig) l l l l l l l l Heat !

l MBTU/hr ! 98.75 113.33! 109.71I 109.71 !106.00 I l Rate l l l l l l l l Delta T l 27 ! ! ! ! !

'F 31 30 30 29 l B-A l l l l l l l l IMID l 'F ! 32.99 ! 32 ! 32.50 I 32.50 ! 31.99 !

l l l l l l i I

! D'"1 ! **/ ! 2.99 ! 3.54 ! 3.38 ! 3.38 ! 3.32 !

l Capacity l hr*F l l l l l l

May obtain data from ERIS point (Attachment 2) instead of meter.

+ Thermodynamic data from 1967 ASME Steam Tables Heat Rate = x C x 8.02 x 10-6 J

g ,(E-B) - (T-A) (If use of this formula results in zero IRID, use g (E-B) IMID = AT = (E-B) + (F-A))

(F-A) 2 )

Thermal Capacity = Heat Rate l

106 l .

L_

STARWP TEST RESULTS (CONT. )

STI-E12-071 (Cont.)

Suppression Pool Cooling Data (Cont.)

Data for RHR Loop B l l l Time l l Parameter l Units l 2323 l 2328 l 2333 l 2338 l 2343 l

l A

'F ! 38 I 38 ! 38 ! 38 ! 38 !

l P45-R091A(B) l l l l l l l l m fm Hx i I I I I I I l B A-E12-N005A l 'F l 63 l 67 l 67 l 66 l 65 l l B-E12-N005B l l l l l l l

! c " fl* I y I 7300 I 7300 ! 7300 l7300 I 7300 I l E12-R602A(B) l l l l l l l

! D ress l psig ! 95 I 95 I 95 ! 95 l 95 I l P45-R102A(B) l l l l l l l l ESW to Hx l l l l l l l

l E E12-N004A(B) l 'F l 92 l 95 l 95 l 96 l 91 l l lE12-R601(pt.1,2)l l l l l l l l RsR fm Hx l I I I I I I l F E12-N027A(B) l 'F l 67 l 69 l 69 l 65 l 65 l l 1E12-R601(pt.5,6)l l l l l l l

l RHR F1 w l gp, l 6900 l6900 ! 6900 l6900 l6900 !

l E12-R603A(B) l l l l l l l

! G Enthalpy In l B W/lb+ ! 6 .34 ! 6 .34 ! 6 .34 ! 6 .34 ! 6 .34 !

l (at A 'F,D-psig) l l l l l l l

! H alg M lBW/lb+! 31.37 ! 35.36 ! 35.36 ! 34.36 I 33.36 I l (at B 'F,D-psig) l l l l l l l l r Av9 T**P l 'r I 50.5 I 52.5 l52.5 ! 52.0 l 51.5 I l 1/2(A+B) l l l l l l l 107

STARRTP TEST RESULTS (CCNT.)

STI-E12-071 (Cont.)

Suppression Pool Cooling Data (Cont.)

Data for RHR Loop B

! J Sp cific Vol 3 lft/lb+!.01602 !.01602 !.01602!.01602 !.01602!

l (at I 'F,D-psig) l l l l l l l l Heat ! 106.06! 106.06! 102.40 ! 98.75 !

l MBW/hr ! 91.47 l Rate l l l l l 1 l l Delta T l 7 l 25 ! 29 ! 29 ! 28 ! 27 !

B-A l l l l l l l l

! LMTD ! 'F ! 29.0 ! 29.47 ! 29.47 ! 28.47 ! 26.5 I l l l l l l 1 l l Thermal l MBW/ l 3.15 ! 3.60 I 3.60 ! 3.60 ! 3.73 !

l Capacity l hr'F l l l l l l

May obtain data f rom ERIS point (Attachment 2) instead of meter.

+ Thermodynamic data from 1967 ASME Steam Tables Heat Rate = E x C x 8.02 x 10-6 J

, g ,(E-B) - (F-A) (If use of this formula results in zero LMTD, use g (E-B) LMTD = AT = (E-B) + (F-A))

(F-A) 2 )

Miermal Capacity = Heat Rate IKrD 108

STARW P TEST RESULTS (CONT.)

STI-E12-071 (Cont.)

2. Steam condensing capacity and Stability (8.1) h e purpose of this test was to demonstrate the steam condensing capacity of RHR System Loop A and B Heat Exchangers. In addition, Heat Exchanger Loop A and B level and pressure step changes were performed to demonstrate stable system response.

Acceptance Criteria Level 1 1

The transient response of any system-related variable to any test input must not diverge.

Level 2

1. W e RHR System shall be capable of operating in the Steam Condensing Mode at 150.4 MBW/HR for single heat exchanger operation. The simultaneous operation of both RHR loops and single loop operation 1 - shall be tested in this mode.

2. System-related variables may contain oscillatory modes of response.

In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.

The Level 1 Acceptance Criteria was met satisfactorily. All system related variables exhibited stable response characteristics.

i 109 L.

1 g, :, v .

kh'" '

i

.e STARWP TEST RESULTS (CCNT.)

\ v STI-E12-071 (Cont.)

The RHR Loop Aland B Heat Exchangers steam condensing capacity was demonstrated by ' operating RHR A and B loop individually, collecting data

' ', and verifying the design heat removal' rate was achieved. For both RHR l

Heat Exchanger loops a' heat removal rate of 150.1 MBW/HR was achieved.

Resed on additional EN flow and heat exchanger level available the Level. 2 Acceptance Criteria was satisfied. Sar,isfactory dual Loop steam

. condensing operation was also demonstrated which satisfies the Level 2 Acceptance criteria. Attached are the data sheets for both single and dual loop Steam Condensing performance during the STI.

Steam Condensing Data Data for RHR Loop A Single Loop Operation Time l l l 'l l Parameter l Units l 1545 l 1550 l 1555 l 1600 l 1605 l 1610 l

! ! I 40 l 39 !

l 0 ! ! 39 ! 40 40 40 A 'F l l l P45-R091A(B) l l l l l l

I I I I I 1 m fm Hx I I I l B E12-N005A l *F l 79 l 80 l 80 l 80 l 80 l 80 l l l l l l E12-N005B l l l l

l l l l l l M fm Hx l l l l C-E12-N003A l *F l 40.l 40 l 40 l 40 l 40 l 40 l l l l D-E12-N003B l 'l l l l l M Flow ! 7500 I 7500 ! 7500 ! 7500 !

l l l7500 ! 7500 l l l l l E12-R602A(B) l l l l

! # #'l ! % ! 85 I 85 I 85 ! 85 ! 85 I 85 !

l l l l l l l E12-R604A(B)-1l l

! ! ! 215 I l E Press l g l 215 ! 215 ! 215 215 215 E12-R606A(B)-1l p,g l l l l l l l l

110

STAR *IUP TEST RESULTS (CONr.)

STI-E12-071 (Cont.)

Steam Condensing Data (Cont.)

Data for RHR Loop A Single Loop Operation l l l Time l l Parameter l Units l 1545 l 1550 l 1555 l 1600 l 1605 l 1610 l l Ex to RCIC l l l l l l l l l Pressure lpsig l 20 l 20 l 20 l 20 l 20 l 20 l l E12-R605-1 l l l l l l l l l RCICSuctionl l l l l l l l l Pressure lpsig l 26 l 26 l 26 l 26 l 26 l 26 l E51-R604 l l l l l l l l l

l RCIC Flw l gpn ! 700 ! 700 ! 700 ! 700 ! 700 ! 700 !

E12-R606 l l l l l l l l l

l Steam Temp l 7 l 388 ! 388 ! 388 ! 388 I 388 ! 388 !

E12-N004A(B) l l l l l l l l l

l Condensate l 'F l 67 l 67 l 67 l 67 l 67 l 67 l l Tem I I I I I I I I l l l l l l l l E1QN002A(B) l l Delta T l .F ! 40 ! 40 ! 40 ! 40 ! 40 ! 40 !

B-A I l l l l l l l l

Data for RHR Loop B Single Loop Operation l l l Time l l Parameter l Units l 0245 l 0250 l 0255 l 0300 l 0305 l l ESW to Hx l 7 l 40 ! 40 I 40 ! 40 ! 40 I A

P45-R091A(B) l l l l l l l l

! B" * ! 'T ! 80 ! 80 ! 80 ! 80 ! 80 !

l E12-N005A/B l l l l l l l 111 i

r- -

l l

STAR'IUP TEST RESULTS (CONT. )

STI-E12-071 (Cont.)

Steam Condensing Data (Cont.)

Data for RHR Loop B Single Loop Operation l l l Time l l Parameter l Units I 0245 l 0250 l 0255 l 0300 1 0305 l l ESW fm Hx l l l l l l l l C-E12-N003A l 7 l 37 ! 37 I 37 ! 37 ! 37 !

l D-E12-N003B l l l l l l l l ESN Flw l ggg l 7500 I 7500 '! 7500 ! 7500 ! 7500 l l E12-R602A(B) l l l l l l l l m Level l g l 78 ! 78 ! 78 ! 78 ! 78 !

l E12-R604A(B)-1] I l l -

l l l l E Press l ! 200 l 200 ! 200 ! 200 ! 200 I

-E12-R606A(B)-1l psig l l l l l l l

l Hx to RCIC l l l l l l l Pressure l psig l 17 l 17 l 17 l 17 l- 17 l l

l E12-R605-1 l l l l l l l l RCICSuctionl l l l l l l Pressure lpsig l 24 l 24 l 24 l 24 l 24 l l

l E51-R604 l l l l l l l l RCIC Flow l p l 420 ! 420 ! 420 ! 420 ! 420 I E12-R606 l l l l l l l l

l Steam Temp l .F ! 390 ! 390 ! 330 ! 390 ! 390 l l E12-N004A(B) l l l l l l l l Condensate l l l l l l l l Temp l 'F l 67 l 67 l 67 l 67~ l 67 l E12-N002A(B) l l l l l l l l

l Delta T l 7 l 40 ! 40 ! 40 ! 40 ! 40 I B-A l l l l l l l l

112

. . . _ . . __ _ .- .. _ . _ . _ ~_ .

STAR'IUP TEST RESULTS (CONT.)

, . STI-E12-071 (Cont.)

Steam Condensing Data (Cont.)

Data for.RHR Loop A Dual Loop Operation j- l l l Time l l Parameter -l Units l 1355 l 1400 l 1405 l 1410 l 1415 l l l 39 l 39 l ! ! I

, IA ESw to Hx .F 40 40 40 l P45-R091A(B) l l l l -l l l-ESw fm Hx l l 66 l 66 l 67- l 67 l 67 l l 3 .,

l E12-N005A/B l l l l l l l 1 ESw fm Hx l l l l l l l l C-E12-N003A~ l *F l 39 l 39 l 39 l 39 l 39 l l o-E12-N003B l l l l l l l

l ESw rlw l ,,, l7500 l 7500 l 7500 ! 7500 l7500 l E12-R602A(B) l l l l l l l 4

l l HX Level l g l 90 ! 90 ! 90 ! 90 I 90 !

i-l' E12-R604A(B)-1l l l l l l l

l E Press l l ! 215 ! 215 I 215 I 215 !

g 215 E12-R606A(B)-1lp,g l l l l l l l

l Ex to RCIC l .l l l l l l Pressure lpsig l 20 l 20 l 20 l 20 l 20 l

'l E12-R605-1 l l l l l l l l

l RCICSuctionl l l l l- l l Pressure lpsig l 26 l 26 l 26 l 26 l 26 l

, l E51-R604 l l l l l l l l

l RCIC Flw l gp , l 700 ! 700 ! 700 ! 700 ! 700 I E12-R606 l l l l l l l l

l Steam Temp l ! ! ! ! 390 ! 390 I

.F 390 390 390 E12-N004A(B) l l l l l l l l

l l

113 v w w -,w .<w.m-- . - - ,.n - v-.,v . - -- --y , y, - --,r+-,. v-.. --.% -,- w,----- w- evw--. - -,.w ..s,m.---%----e,-. .-~

.f STARR7P TEST RESULTS (CONT.)

~

STI-E12-071 (Cont.)

Steam Condensing Data (Cont.)-

Data for RHR Icop 1 Dual Loop. Operation l l l Time l' ,

.l Parameter l Units l 1355 l 1400 l 1405 l 1410 l 1415 l l Condensate l l l l l l l l Temp l *F l 60 l 60 l 60 l 60 l 60 l l E12-N002A(B) l l l l l l l l Delta T- l 7 l 27 ! 27 ' ! 27 ! 27 ! 27 !

l B-A l l l l l l l Data for RHR Icop _B_ Dual Loop Operation-l l l Time l l Parameter l Units l 1355 l 1400 l 1405 l- 1410 l 1415 l lA m to k l 7 l 40 ! 40 ! 40 I 40 ! 40 I l P45-R091A(B) l l l l l l l l B *F ! 67 ! 67 ! 67 l 67 ! 67 l l E12-N005A/B l l l l l l l I m f= nx I I I I I I I

-l C-E12-N003A -l *F l 36 l 37 l 37 l 37 l 37 l l D-E12-N003B l l l l l l l l m F1w l gg, l7500 ! 7500 ! 7500 ! 7500 ! 7500 !

l E12-R602A(B) l l l l l l l

! " ""1 ! % ! 84 ! 84 ! 84 ! 84 ! 84 !

l E12-R604A(B)-1l l l l l l l l n Fress l l I 2a5 I 205 ! 205 ! 205 l 205 l

E12-R606A(B)-1lp,19 l l l l l l 114

STARIUP TEST RESULTS (CONT.)

STI-E12-071 (cont.)

Steam Condensing Data (Cont.)'

Data for RHR Loop -B Dual Loop Operation-l l l Time l l Parameter l Units l 1355 l 1400 l 1405 l 1410 l 1415 l l Hx to RCIC l l l l l l l

~l Pressure lpsig l 20 l 20 l 20 l 20 l 20 l l E12-R605-1 l l l l l l l l RCICSuctionl l l l l l l l Pressure lpsig l 26 l 26 l 26 l 26.l 26 l l E51-R604' l l l l l l l l RCIC F1w .

l gp, l 700 ! 700 ! 700 I 700 I 700 !

l E12-R606 l l l l l l l l Steam Temp l ! ! ! ! ! I

'T 390 390 390 390 390 l E12-N004A(B) l l l l l l l l . Condensate l l l l l' l l l Temp l *F l 60 l 60 l 60 l 60 l 60 l l E12-N002A(B) l l l l l l l l Delta T l .F ! 27 ! 27 ! 27 ! 27 ! 27 !

l B-A l l l- l 1 l l The Level 2 Analysis Criterion 2 was satisfied by performing 20'psig pressure and 5% level step changes into the RHR Heat Exchanger Loop A and B controllers and verifying stable system response. All system.

responses were satisfactory except for one case of minor oscillatory type behavior experienced during the 20 psig step decrease for the RHR B Loop Heater Exchanger. After further analysis this was determined to be satisfactory since the problem was minor and did not occur during later testing.

115

STARWP TEST RESULTS (CCtfr. )

4.20 STI-C95-099 ERIS (Emergency Response and Information System)

(TSN-141, 146, 157, 172, 173)- t

'Ihe ERIS testing conducted in this portion of the Startup Test Program i consisted of Whole Core Scram Timing Verification and Event Target verification on the SPDS screens. The whole core scram timing analysis feature of ERIS is run on the Transient Recording Analysis computer (TRA) and the SPDS screens are run on the Real Time Analysis and Display computer (RTAD).

-1. Transient Recording Analysis (TRA)' Scram Timing During Planned Scrams 1

Acceptance Criteria Level 2 l a. The control rod scram timing function shall properly indicate selected control rod status, i

b. The control rod scram timing function shall indicate scram times of selected rods to the appropriate notch positions to within i 0.01 second of an independent measurement.

Results This test was performed to verify the whole core scram timing feature of ERIS. This was the first time this program of ERIS had been run.

The individual rod scram times were verified in ' Inst Condition Open Vessel and Heatup. One hundred and eight (108) rods were initially at position 48 and were scrammed in parallel with the Shutdown From outside the Control Room test. All 108 rods appeared on the ERIS edits as being scrammed.

116

STARWP TEST RESULTS (CCNr.)

STI-C95-099 (Cont.)

TER 173-1 was written to address an observed problem that the indicated scram times on the edits (ERIS) show the control rods having a time to technical specification notch positions approxi-mately 100 as less than expected. 'Ihis is being investigated further by General Electric.

2. Real Time Analysis for Display (READ) SPDS Screen Verification Acceptance Criteria Level 2 Selected BASIC ERIS event targets (safety relief valve, MSIV, SCRAM) shall agree with actual plant status.

Event Target Verification Results Sections of this test were performed to verify that selected ERIS event markers correctly reflect actual plant conditions. Three (3)

Diesel Generator Event Tag Tests were performed to test the screen change involved as the Division 1, 2, and 3 Diesel Generators receive their start signal and proceed to rated voltage and synchronize to their respective buses. The Division 2 and 3 Diesels failed to indicate a yellow " Diesel Connanded" event tag which was attributed to the loads and processing rates on the ERIS computer at that time.

The SCRAM event log was verified on various SPDS screens in conjunction with planned scram #1, Shutdown From Outside the control Room. During the scram, the scram event tag was verified to turn yellow " SCRAM CMD" and then it turned red " SCRAM RODS OUT" rather than the expected green " SCRAM RODS IN" event tag. This was caused by bad data coming from one of the position indicating probes (PIP) of Channel A of RCIS.

117

STARWP TEST RESULTS (CONT.)

Integrated HVAC 4.21 STI-M99-100 Computer Room Data and Integrated HVAC Data (TSN 130, 140)

'Ihe purpose of these tests was to demonstrate the ability of the area ventilation systems including the computer room HVAC system to maintain Unit 1 and common zone temperatures and relative humidities within design limits for normal operation with ivactor and turbine power greater than 15% power.

Acceptance Criteria Level 1 None Level 2

1. Area ventilation systems maintain Unit 1 and common zone temperatures and humidity within design limits with reactor and turbine power greater than or equal to 15% power and greater than or equal to 95%

power as specified by the testing cor.o.tions.

2. The Computer Room HVAC System (M27) can maintain each Computer Room at a temperature of 75' l'F DB and a relative humidity between 40%

and 50%.

Results The test results, when compared to the Level 2 Acceptance Criteria, indicated 49 out of 148 specified temperatures and humidities were outside their allowable value bands. Thirty were temperatures below allowable l values, 13 were humidities below allowable values, 3 were temperatures j above allowable values and 3 were humidities above allowable values.

l 118

STARIUP TEST RESULTS (CONT.)

STI-M99-100 (Cont.)

Further engineering evaluation of the 49 temperatures and humidities outside their Level 2 Acceptance Criteria value bands determined that all values were acceptable, one temperature which was above its allowable value (DN-4, Drywell Dome) was found to be caused by a defective tempera-ture element which was replaced and retested satisfactorily via work order. One humidity which was above its allowable value (DW-2, outside RPV Shield Wall-At Core Midplane) was found to be caused by a dirty dew-point cell which was cleaned and retested satisfactorily via work order.

The Process Computer Room humidity Level 2 Acceptance Criteria band is planned to be changed from 40-50% RH to 40-60% RH via a M27 Test Specification change. The Process Corputer Room humidity, which was above its allowable value, will fall within the allowable value bands prescribed in the Test Specification change. The 49 temperature and humidity values outside their Level 2 Acceptance Criteria bands were evaluated and determined to not degrade the performance of the HVAC systems and those systems effected.

The Integrated HVAC Data and Computer Room Data tests were performed to demonstrate the ability of the Unit 1 and common area ventilation systems to maintain temperatures and relative humidities within design limits at greater than 15% power. Per Engineering evaluation, all HVAC systems performed adequately. This test will be reperformed at higher power later in the test program.

4.22 STI-P42-Il4 Emergency Closed Cooling System (TSN-162, 166)

'Ihe purpose of this test is to demonstrate that the Emergency Closed Cooling Air Handling Units (OM28-B001A and OM28-PT3) are capable of maintaining the required temperatures present in the areas containing the 119 1

. . I

STAR'IUP TEST RESULTS (CONT. )

STI-P42-114 (Cont.)

Emergency Closed Cooling Pumps and Heat Exchangers under the design limit heat load and to demonstrate that the Emergency Closed Cooling Water System is capable of maintaining the following equipnent within design limits:

1. The Control Ccatplex Chiller (OP47-B001A, OP47-B001B and OP47-B001C).

2. The RHR Room Coolers (1M39-B001A, 1M39-B001B, 2nd 1M39-B002).

3. The RCIC Pump Room Cooler (1M39-B004).

4. The LPCS Pump Room Cooler (1M39-B006).

Acceptance Criteria Level 1 None Level 2

1. The Control Complex Chillers (OP47-B001A, OP47-B001B, and OP47-B001C) are capable of maintaining a chilled water outlet temperature of 45'r or less.

2. The RHR Pump Room Coolers (1M39-B001A, 1M39-B001B, AND 1M39-B002) are capable of maintaining the temperature of the rooms containing the RHR Pumps and Heat Exchangers within the required temperature limits:

Minimum 71'r Maximum 138'r 120

---____________-----_____J

STARWP TEST RESULTS (CONr.)

STI-P42-114 (Cont.)

3. %e RCIC Pump Room Cooler (1M39-B004) is capable of maintaining the temperature of the room containing the RCIC Pump within the required temperature limits:-

Mininnam 98'r Maximum 137'r

4. h e LPCS Pump Room Cooler (1M39-B006) is capable of maintaining the temperature of the room containing the LPCS Pump within the required temperature limits:

Minimum 71*r Maximum 143'r

5. The Emergency Closed Cooling Areas Cooling Air Handling Units (OF28-B001A and OM28-B001B) are capable of maintaining the temperature in the areas containing the ECC Pumps and Heat Exchangers within the required temperature limits:

Minimum 81*r Maximum 89'r Results ECC System - Loop A (8.1) h is section of the test was performed on March 25, 1987, during Test Condition 1. here was one TER written on this test since the control Complex Chiller A outlet temperature was three degrees higher than Accept-ance Criteria. Due to winter conditions, the ECC pump area ambient temperature was several degrees lower than expected. Resolution of this TER is to retest the affected parts of the system during TC3 to ebtain more data.

121

STAR'IUP TEST RESULTS (CCNT. )

STI-P42-114 (Cont.)

ECC System - Loop B (8.2) h is section of the test performs the same function as 8.1 using the B Loop. 2 1s Loop when performed showed very much the same results as 8.1. One exception was that the Coninon Control Complex Chiller "C" was to have been tested. his Chiller was out of service and will be tested when available and/or included in the Unit 2 Startup. The Chiller is not needed for Unit 1. The second exception was that Chiller B outlet temperature was three degrces higher than allowed by Level 2 Acceptance Criteria and the ECC Pump Ambient temperature was several degrees lower than allowable. Section 8.2 will be reperformed in TC-3 to obtain data for Engineering to evaluate.

4.23 STI-P45-117 Emeroency Service Water System (TSN-160, 165)

The purpose of this test is to demonstrate that the Emergency Service Water (ESW) System Loops "A" and "B" are capable of supplying sufficient coolant to the RHR Heat Exchangers and the Emergency Closed Cooling Heat Exchangers while removing design heat loads. Flow was supplied to the Diesel Generator Jacket Water Heat Exchangers, but no heat load existed since the Diesel Generators were not in operation. Loop "C" of ESW was tested during preoperational testing and was not included as part of this STI. (Loop C is independent of steam loads.)

W e test began by placing the "A" or "B" Loop of ESW, Emergency Closed Cooling (ECC), and RER - Steam Condensing mode in operation. A flow of approximately 7500 gpm was established to the RHR Heat Exchangers and approximately 50 gpn to the ECC Heat Exchanger. ESW flow to the ECC Heat Exchanger was throttled to bring ECC temperature up to 90-95'r per requirements for STI-M99-134 (Equipment Area Cooling). With ESW at 38'r, ECC temperature was approximately 56*r before throttling.

122

STARWP TEST RESULTS (CONT. )

STI-P45-ll7 (Cont.)

Acceptance Criteria Level 1 None Level 2

1. The ESW System provides a sufficient amount of coolant to the RHR Heat Exchangers providing a heat removal rate of greater than or equal to 150.4 x 10 6BW/hr while in the Steam condensing Mode.

2. The ESW System provides a sufficient amount of coolant to the ECC Heat Exchanger maintaining ECC Heat Exchanger outlet temperature at less than or equal to 95'F.

Results 6

h e calculated heat transfer rate during the test was 150.1 x 10 BW/HR, 6

therefore, the 150.4 x 10 BTU /HR acceptance criteria was not met.

However, ESW capacity available was more than enough to transfer the additional heat required to exceed the acceptance criteria value based on subsequent Engineering evaluation. ESN outlet temperature from the RHR Heat Exchangers was only = 76'F, more heat could have been removed by simply lowering RHR Heat Exchanger level so more tubes (ESW side) could be exposed to reactor steam.

W e test demonstrated that the ESW system is capable of supplying sufficient coolant to the RHR Heat Exchangers and ECC Heat Exchangers while removing design heat loads.

123

STARTUP TEST RESULTS (CONT.)

.4.24 STI-G42-119 'SPCU Performance Test (TSN-149) h e Suppression Pool Cleanup (SPCU) System demonstrated its capability to maintain suppression Pool water chemistry within specified limits.

Acceptance Criteria Level 1 None Level 2

1. he suppression pool cleanup demineralizer G42-D002 can keep water chemistry within the following limits:

Influent Effluent

a. Conductivity at 25'c f 10 pmho/cm f 3.0 paho/cm

b. Chlorides f 0.5 ppm f 0.1 ppm

c. pH at 25'C 5.3 to 8.6 6.5 to 7.5

d. Suspended solids 1 5.0 ppm f 1.0 ppm

2. Radiation measured (at platform Elev. 599' 9") is less than or equal to 2.5 mR/hr.

Results The test was performed on March 17, 1987, following SRV Flow Testing by placing the SPCU system in service for 24 hrs. and drawing samples from both the demineralizer influent and effluent. % e test failed the Level 2 Acceptance Criteria for effluent pH (5.9). %e demineralizer effluent pH was determined to be acceptable as the value was within the limits currently in use by the chemistry department's administrative program.

124 L. . . . . . . .

STARRTP TEST RESULTS (CONT. )

STI-G42-119 (Cont.)

The radiation monitoring section of the test was performed during the initial performance of the test on 3/17/87. All points were less then 0.2 mR,/hr.

4.25 STI-N27-120 Feedwater System Data - Low Power (TSN-109)

The purpose of this test was to gather data on the feedwater system at low power levels and compare this data to expected operating values.

Acceptance Criteria Level 1 None Level 2 Compare actual readings against acceptance criteria listed on attached sheets.

Results This test was run at approximately 18% reactor power with the following feedwater pumps in operation:

Reactor Feed Booster Pumps "B" and "D" Turbine Driven Reactor Feed Pump "A" Feedwater Seal Injection Pump "B" 125

l STARTUP TEST RESULTS (COtTf. )

STI-N27-120 (Cont.)

Vibration, temperature, flow, and pressure data was collected on the operating equipment and compared to the acceptance criteria. Reactor Feed Pump Turbine accessory oil pressure and seal water pump discharge pressure were outside of the acceptance criteria band. Evaluation by the appropri-ate vendors determined that the equipnent is operating properly. All other acceptance criteria were satisfied.

Additional feedwater system data will be gathered at higher power levels.

Acceptance Criteria Data Sheet: Feedwater System Data - Low Power l Step l l Acceptance Parameter l Reading !

l8.1.5.()l l Criteria l l

! 11.a ! RFPT A Vib at HP Brg l !

2 mils max O mils !

l 1 l l l

! 11.b ! !

RFFT B Vib at HP Brg 2 mils max ! O mils !

l I I I I

! 12.a ! RFPT A Vib at LP Brg I !

2 mils max 0.5 mils !

l l l l l

! 12.b ! RTPT B Vib at LP Brg ! !

2 mils max 0.4 mils !

l 1 I l l

! 13.a ! RFP A Vib at Inbd Brg !

2.0 mils max 1.1 mils !

l l l l l

! 13.b ! RTP B Vib at Inbd Brg ! 2 .0 !

mils max 0.2 mils !

l l l l l

! 14.a ! !

l l RTP A Vib at Outbd2 .0 Brgmils max l! l 1.6 mils I l

! 14.b ! RFP B Vib at Outbd Brg ! ! !

2 .0 mils max 0.1 mils l I I I I 126

STARIUP TEST RESULTS (CONT.)

STI-N27-120 (Cont.)

Acceptance Criteria Data Sheet: Feedwater System Data - Low Power (Cont. )

l Step. l Parameter l Acceptance l Reading !

l8.1.5.()! l Criteria l I

! 15.a I RFPT A Hyd Hdr Press ! 195to225psig! 225 psig I l 1 I l l

! 15.b ! RFPT B Hyd Hdr Press ! 195to225psig! 225 psig !

l l l l l

! 16.a lRFPTAOilFilter l15psidmax ! 0.75 psid !

l l Diff Press l l l l16.b lRFFTBOilFilter l15psidmax ! 1.0 psid !

l l Diff Press l l l

! 17.a ! RFPT A Brg Oil Press ! 10 to 17 psig ! 13.5 psig !

l l l l l I 17.b ! RTPT B Brg Oil Press ! 10 to 17 psig ! 14.5 psig l l l l l l

! 18.a I RFP A Brg Oil Press l 8 to 15 psig ! 13.0 psig !

l l l l l

! 18.b !RFP B Brg Oil Fress 8 to 15 psig ! 12.0 psig !

l l l l l lgg,, lRTPAAccessoryOil l48to62psig ! 65 psig !

l l Press l l l l19.b lRFPTBAccessory011 l48to62psig ! 66 psig !

l l Press l l l l21.a l M A E BRG DWN l180'Fmax ! 118.2*F !-

l l OIL TEMP l l l

! 21.b l M B E BRG DEN l180'rmax ! 105.7'F l l OIL TEMP l l l 127

STARTUP TEST RESULTS (CONT.)

STI-N27-120 (Cont.)

Acceptance Criteria Data Sheet: Feedwater System Data - Low Power (Cont. )

l Step l Parameter l Acceptance l !

Reading l8.1.5.()l l Criteria. l l l , lRFPTATHRUSTBRG l180'Fmax l 117'T l l DRAIN OIL TEMP l lg lRFPTB11{RUSTBRG l180'Fmax ! 106.8'F l l DRAIN OIL TEMP l

! 23.a l M A R BRG DRAIN l180'Fmax l 114.8'F l l lOILTEMP l l l I 23.b l R M B D BRG DRAIN l180'Fmax ! 109.0'F l l OIL TEMP l l l l lRFPTAINBDBRGDRAIN l180'Fmax ! 112.5'F l l OIL TEMP l l24.b lRTPTBINBDBRGDRAIN l 180'F max ! 107.0*F l l OIL TEMP !

l25.a l RFPT A OLTIBD BRG DRAIN l 180'F max ! 120.5'F l l OIL TEMP l l l l25.b l RM B OUTBD BRG DRAIN l 180'F max ! 108.5'F l l OIL TEMP l l lRFPTASealLeakoff l 170'F max ! 105'F l l Temp (R008A) l l26.a lRFPTASealLeakoff l170*Fmax ! 110*F l l Temp (R009A) l l lRTPTBSealLeakoff l 170'F max ! 170'F l l Temp (R008B) l l23,3 l RrrT B Seal Leakoff l170.rmax I 155'r l l Temp (R009B) l l lRFPTAExhaustHood l250'Fmax 82*F l l Temp 128

STARIUP TEST RESULTS (CONT.)

STI-N27-120 (Cont.)

Acceptance Criteria Data Sheet: Feedwater System Data - Low Power (Cont. )

l Step l Parameter lA eptance l Reading i l8.1.5.()l l Criteria- l l RFPT B Exhaust Hood l250*Fmax ! 135'F l M.b l l Tew I

! 28.a jSealWaterAPump l360to l 510 psig !

l l Disch Press l440psig l l l28.b lSealWaterBPump l360to l 510 psig !

l l Disch Press l 440 psig l I 4.26 STI-T23-123 Concrete Temperature Survey at Low Power (TSN-110)

The purpose of this test was to demonstrate the ability of natural heat transfer to cool the concrete surrounding selected pipe penetrations.

Acceptance Criteria lj Level 1 Exterior Drywell and Shield Building concrete temperatures at selected penetrations shall not exceed a temperature of 200'F minus loop accuracy.

Results Due to the high radiation fields present in the Steam Tunnel, remote monitoring was performed by attaching thermocouples to the concrete and f

! routing the thermocouple wire to low radiation areas.

l l

129

t STARIUP TEST RESULTS (CONT.)

STI-T23-123 (Cont.)

Concrete temperature data was collected from near the main steam, feedwater, and reactor water cleanup penetrations. The maximum allowable concrete temperature was determined to be 195'r based on a 5'r maximum thermocouple loop-digital thermometer error. The following table lists the maximum temperatures recorded:

Maximum Recorded Temperature (*r)

Location Drywell Shield Bldg Main Steam 113.7 156.8 reedwater 106.2 101.3 Reactor Water Cleanup 107.6 137.3 The maximum recorded temperature was 38.2'r below the Level 1 analysis limit, thus satisfying all criteria.

4.27 STI-R63-133 Loose Parts Monitoring System (LPMS) Baseline Data (TSN-183)

'1he Loose Parts Monitoring System (LPMS) was successfully used to collect steady state baseline data during Test condition 'No (TC 2).

Acceptance Criteria Level 2 Initial Baseline data was obtained for steady state conditions established for Test Condition Twc (TC 2).

130

STARIUP TEST RESULTS (CONT.)

STI-R63-133 (Cont.)

For 'It 2, the reactor power was at 35% of rated with both recirculation 1 pumps on fast speed with minimum flow control valve position. The generator was synchronized to the grid.

All twelve vibration and loose parts channels for.the reactor were monitored for this test. We baseline data was collected on three 90 minute cassette tapes lasting approximately 11 minutes each. All baseline '

data was collected satisfactorily with the noise plots showing the expected characteristic waveform distribution.

4.28 STI-M99-134 Equipment Area Cooling (TSN-158, 159, 161, 164, 167, 168)

The purpose of this test was to verify that the RCIC, LPCS, HPCS and RHR "A", "B", and "C" room coolers are capable of removing the postulated post-accident der,ign heat loads. The six rooms were tested in two groups of three each. One group was comprised of Division 1 equipment: RHR "A",

LPCS and RCIC. The other group was comprised of Division 2 and Division 3 equipment: RHR "B", RHR "C", and HPCS. Data collection occurred when the respective Division 1 or 2 was placed in Steam condensing mode. The ECCS pumps were placed in test with return to the suppression pool, for the rooms being tested. W e Emergency Closed Cooling System "A" or "B" was in operation supplying cooling water between 90-95'r to the room coolers.

131

STARWP TEST RESULTS (CONT. )

STI-M99-134 (Cpnt.)

And for at least one hour, the Auxiliary Building Ventilation System was secured. Data collection consisted of room, hall, cooling water, suppression pool, and air temperatures. Cooling water flow rates to the room coolers were also recorded.

Acceptance Criteria Level 1 None Level 2

1. We RHR A Pump Room Cooler maintains the room average temperature between 140'r and 145'r with a cooler load between 362,250 BW/hr and 402,500 BW/hr.

2. % e RHR B Pump room Cooler maintains the room average temperature between 138'r and 143' with a cooler load between 342,900 BW/hr and 381,000 BW/hr.

3. The RCIC Pump Room Cooler maintains the room average temperature between 133'r and 138'r with a cooler load between 69,750 B W/hr and 77,500 DW/hr.

4. The RHR C Pump Room Cooler maintains the room average temperature between 120'r and 125'r with a cooler load between 211,455 BW/hr and 234,950 B W/hr.

132 h - - - -- - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _

STAR MP TEST RESULTS (C0ffr.)

STI-M99-134 (Cont.)

5. The HPCS Pump Room Cooler maintains the room average temperature between 125'r and 130*F with a cooler load between 392,805 BW/hr and 436,450 BW/hr.

6. 'Ihe LPCS Pump Room Cooler maintains the room average temperature between 125'r and 130*r with a cooler load between 264,443 BW/hr and 293,825 BW/hr.

Results

1. (TSN-158) mm "A" Pump Room Cooler heat load was not between 362,250 BTUAm and 402,500 BWAm (actual 30,032 BTU /HR), and the average RHR "A" pump room temperature did not fall between 140*r and 145'r (actual 81.7'r).

2. (TSN-159) RCIC Pump Room Cooler heat load was not between 69,750 BW/HR and 77,500 BWAm (actual 14,742 BWAm), and the average RCIC pump room temperature did not fall between 133'r and 138'r (actual 107.4'r).

3. (TSN-161) LPCS Pump Room Cooler heat load was not between 264,443 BWAm and 293,825 BWAm (actual 27,940 BWAm), and the average LPCS pump room temperature did not fall between 125'r and 130'r (actual 89.0*r).

4. (TSN-164) RHR "C" Pump Room Cooler heat load was not between 211,455 BWAm and 234,950 BWAm (actual 209,885 BWAE), and the average mm "C" pump room temperature did not fall between 120'r and 125'r (actual 94.0'r).

I i

l l

133

STARWP TEST RESULTS (CCNT.)

STI-M99-134 (Cont.)

5. (TSN-167) RER "B" Pump Room Cooler heat load was not between 342,900 BWAfR and 381,000 BW/HR (actual 208,493 BW/HR), and the average BW/HR (actual 208,493 BW/HR), and the average RHR "B" pump room temperature did not fall between 138'F and 143*F (actual 79.2'F).

6. (TSN-168) HPCS Pump Room Cooler heat load was between 392,805 BWAIR and 436,450 BW/HR (actual 401,960 BTU /HR). This room passed it is required heat load, because Emergency Service Water cools this room and it was at 38'F. The average HPCS pump room temperature did not fall between its acceptance criteria band of 125'F and 130'F (actual 80.7'F).

Engineering was forwarded a copy of the data and answered the test exceptions in the following way:

An EDDR or NR is not required. System design criteria has been satisfied or exceeded as described in Sections 3, 8, & 14 of the FSAR.

1. Low Temperatures The temperature ranges shown per addendum "E" to CAI TS-M39 were for maximum room temperatures. These values were shown in the acceptance criteria in error. The allowable temperature ranges for the rooms should have been 60*F minimum to the values shown in the acceptance criteria (maximum values). With these temperature ranges room temperature were acceptable. In addition, temperatures as required by FSAR Table 3.11-3 are well within the acceptable limits.

134

STARWP TEST RESULTS (C0tn.)

STI-M99-134 (Cont.)

2. Low Heat Load

'Ihe test data taken to prove the M39 coolers are adequate to remove the maximum anticipated heat load (found via calculation), showed the actual heat loads measured are far below the anticipated heat loads and that the cooling system in place is acceptable.

The lower than expected temperature measured in the pump room during this test indicates the heat load calculations performed were conservative.

This would also suggest the calculations performed to generate the values in the FSAR environmental tables would also have some degree of conserva-tiveness. Therefore, temperatures shall be taken in the pump room during normal conditions when the ambient outdoor temperature is equal to or greater than 85'F. This data shall be submitted to Nuclear Engineering Department for further review.

135

Attachment 1 Startup Test Performance Log Through the End of Test Condition 2 Test Serial No. STI No. - Section STI Title (Section)

Pre-Test Condition OV 001 D21-002-8.1 General Survey Performance 4 002 D21-002-8.3 Transfer Tube Survey ,

003 P35-001-8.1 Chemistry Data Pre-ruel Load Test Condition Open Vessel 1

004 OV-200-8.1 Open vessel Testing 005 C11-005-8.1 CR-Insert / Withdraw Timing s 006 C11-055-8.2 CR-Friction Testing ,

007 J11-003-8.1 ruel Loading 008 C11-005-8.3 Single Rod Scram Testing 009 C95-099-8.4.1 Scram Timing verification 010 C51-006-8.1 SRM Functional Testing 011 J11-033-8.3 Installation of Cr-252 Neutron Sources 012 J11-003-8.2 Verification of Core Loading 1 013 G42-119-8.1 SPCU Performance Test 014 C95-099-8.2.12 Drywell Temperature - Cold 015 B33-029A-8.1 riow Control valve Stroking 016 G42-119-8.1 SPCU Performance Test (Retest) 017 C95-009-8.4.1 Scram Timing Verification (Retest) 018 C11-005-8.3 Single Rod Scram Timing (Retest) 019 J11-004-8.1 rull Core Shutdown Margin 020 C51-010-8.1 SRM-IRM Overlap 021 D21-022-8.1 General Survey Performance - Post ruel Load 136

?

Attachment 1 (cont.)

Startup Test Performance Log Through the End of Test Condition 2 (Cont.)

Test Serial No. STI No. - Section STI Title (Section)

Test Condition Open Vessel (Cont.)

022 C11-005-8.4 Ganged Rod Timing 023 P35-001-8.2 Chemistry Data Pre-Heatup 024 C91-013-8.1 TIP Alignment After TIP System -

Installation 025 C91-013-8.5 DSTC-Pre-Startup Operation Test Condition Heatup 026 HU-201-8.1 Heatup Plateau Testing

. j g 027 B21-033-8.1 NSSS Steady State Vibration / Strain Measurements (1) 028 B21-033-8.3 NSSS Transient Vibration / Strain Measurements (2) 029 N33-129-8.4 Steam Seal Flowpath verification

'i 3

030 F41-034-8.1 Pre-Nuclear Heatup Flow Test 031,. P44-116-8.2 TBCC System-Mech. Vac. Pump Data

'032 B21-0168-8.2 Shutdown Range Data 033 B21-017-8.1 NSSS Piping Expansion, Thermal cycle No 1 034 B21-033-8.2 NSSS Visual Vibration 035 P99-122-8.1 BOP Piping Test - Reactor Heatup 036 .

. B21-033-8.3 NSSS Transient Vibration / Strain - SRV i

runctionals 037 P99-122-8.2 BOP Piping Test - Transients 038 C11-005-8.6 CED Hydraulics Test 137

Attachment 1 (Cont.)

Startup Test Performance Log

'Ihrough the End of Test Condition 2 (Cont. )

Test Serial No. STI No. - Section STI Title (Section)

Test Condition Heatup (Cont.)

039 C51-010-8.1 SRM and IBM Overlap 040 C51-006-8.2 SRM Non-Saturation 041 C51-010-8.2 IRM Range Correlation Adjustment 042 C95-099-8.2.1 Reactor Pressure - Low 043 E51-014-8.1 RCIC Discharge to the CST 044 C95-099-8.2.3 RPV/ Press /2D Plot / Limit Tag (Low Press) 045 B21-026-8.1 SRV Functional Test 046 C95-099-8.3.4 SRV Test - ERIS 047 N33-129-8.1 Steam Seal Ivaporator 048 C51-012-8.1 Constant Heatup APRM Calibration 049 E51-014-8.1 RCIC Discharge to CST 050 C11-005-8.5 Scram Timing of Selected Rods -

600 psig 051 N62-130-8.1 SJAE Low Steam Flow 052 N64-074-8.1 off Gas System Test - Startup 053 C11-005-8.5 Scram Timing of Selected Rods -

800 psig 054 P99-122-8.2 BOP Piping Test-Transient RCIC Hot Quick Start 055 P99-122-8.1 BOP Piping Test-Reactor HU-RCIC 2 Hr. Run 056 B21-033-8.1 NSSS Steady State Vibration / Strain-RCIC 2 Hr. Run 057 D21-002-8.1 General Survey Performance 058 P35-001-8.3 Chemistry Data Heatup Tests 059 C11-005-8.2 Friction Testing 138

Attachment 1 (Cont.)

Startup Test Performance Log

'Ihrough the End of Test Condition 2 (Cont.)

Test Serial No. STI No. - Section _

STI Title (Section)

Test Condition Heatup (Cont.)

060 G33-070-8.1 FWCU Normal Mode 061 B21-016B-8.1 Reference Leg Temperature Meas.

062 T23-123-8.1 concrete Temperature Survey at Low Power 063 G33-070-8.2 Rx Vessel Bottom Head Drain Flow Ind.,

Cal. Check

064 065 C51-010-8.3 IRM-APRM Overlap 066 C51-011-8.1 LPRM Verification 067 E51-014-8.1 RCIC Disch. to CST (Rated Press) 068 E51-014-8.2 RCIC 2 Hr. Run 069 C11-005-8.3 Single Rod Scram Timing 0 > 950 psig 070 B21-025A-8.1 MSIV Function Test 071 C95-099-8.4.1 Scram Timing Verification - Single Rod 072 C51-012-8.1 Constant Heatup APRM Calibration

073 074 C51-011-8.1 LPRM Verification (Retest) 075 B21-016A-8.1 FCV Minimum Flow Adjustment

076

077

078 079 N64-074-8.1 off-Gas System - Normal Ops

- 080 081 R63-133-8.1 LPMS, Data 082 C11-005-8.3 Single Rod Scram Timing (Retest) 139

Attachment 1 (Cont.)

Startup Test Perforrance Log Throuch the End of Test Condition 2 (Cont.)

Test Serial No. STI !?o. - Section STI Title (Section)

Test Condition Heatup (Cont.)

083 B21-017-8.2 NSSS Piping Expansion - Therral Cycle #2 084 P99-122-8.1 BOP Piping Test - Rx Heatup (Retest) 085 C51-010-8.1 SRIVIRM Overlap (Retest) 086 C95-099-8.2.1 Rx Press - Low (Retest) 087 C95-099-8.2.3 RPV Press /2D Plot / Limit Tag - Low Press (Retest) 088 C91-013-8.2 TIP Alignment @ Rx Startup (Retest) 089 C11-005-8.2 Friction Testing (Retest of 10-15) 090 C95-099-8.2.2 Reactor Press - Pated 091 C95-099-8.2.3 RPV Press /2D Plot / Limit Tag - Rated Press.

092 G33-070-8.3 RNCU Blowdown Mode 093 C95-099-8.2.7 Reactor Power - Low 094 C95-099-8.2.16 Containment Temperature - Norral Test Condition One 095 LP-202-8.1 Test Condition 1 Testing 096 B21-017-8.3 NSSS Piping Expansion Thermal Cycle #3 097 P99-122-8.1 BOP Piping Test - Rx Heatup (Retest) 098 C51-012-8.1 Constant HU APRM Calibration (Retest) 099 C95-099-8.2.6 Reactor Level - Limit Tags 100 C95-099-8.1.3 Reactor Vessel Temperature Constants at TC-1 101 C95-099-8.2.11 Drywell Pressure 140

Attachment 1 (Cont.)

Startup Test Performance Log Through the End of Test Condition 2 (Cont.)

Test Serial No. STI No. - Section STI Title (Section)

Test Condition One (Cont.)

102 C95-099-8.2.15 Containment Pressure 103 C95-099-8.2.13 Drywell Temperature - Normal 104 C95-099-8.2.17 Suppression Pool Temperature - Normal 105 C95-099-8.2.18 Suppression Pool Level 106 N64-074-8.1 Off-Gas System Test 107 C51-011-8.1 LPM 4 Verification (TER Retest) 108 C51-011-8.3 LPRM Calibration w/o Process Computer 109 N27-120-8.1 N Data - Low Power 110 T23-123-8.2 Concrete Temperature @ High Power 111 N27-023A-8.3 (RFP A) N Startup Rx Level Controller Steps

112 113 R63-133-8.1 LPMS Baseline Data 114 N27-023A-8.3 N Startup Rx Level Controller (MFP) 115 D21-002-8.1 General Survey Performance 116 N27-023A-8.3 N Startup Rx Level Controller (RFP-B)

117 118 N27-023A-8.1 (MFP) N Low Flow Valve Test (MFP) 119 N27-023A-8.1 (RTP-A) N Low Flow Valve Test (RFP-A) 120 N27-023A-8.1 (RFP-B) N Low Flow Valve Test (RFP-B) 121 P44-116-8.3 TBCC System - Motor Driven Feed Pump 122 G33-070-8.3 RNCU Blowdown Mode 123 P44-116-8.3 TBCC System-MDFP Data (Retest) 124 B21-033-8.1 NSSS Steady State Vibration / Strain (1) 125 P99-122-8.3 BOP Piping Test-Power Plateaus (2) 126 C91-013-8.3 LPMi Cal. & TIP Plotter Adjustment 127 C91-019-8.1 Core Thermal Power Determination 141

l Attachment 1 (Cont.)

Startup Test Performance Log Through the End of Test condition 2 (Cent.)

Test Serial No. STI No. - Section STI Title (Section)

Test Condition One (Cont.)

128 Col-019-8.2 Core Thermal Limits Determination 129 P35-001-8.4 Chemistry Data & Power Ascension 130 M99-100-8.2 Integrated HVAC Data 131 C91-013-8.6 DSK-Data Init. by OD-15 132 C91-013-8.7 DSTC-Plant Sensor Checks 133 C91-013-8.8 DSTC-Verif. of Prog. En. by OD-15 134 C91-013-8.9 DSTC-LPRM Calibration 135 C91-013-8.10 DSTC-Power Dist. & Th. Limits 136 C91-013-8.11 DSK-Daily & Monthly Logs 137 D21-002-8.2 RNCU Resin Transfer Survey 138 C85-022-8.3 Press. Regulator Test w/ Bypass Valves 139 C85-022-8.5 Press. Regulator Deadband 140 M99-100-8.1 Computer Room Data 141 C95-099-8.3.1 Event Target Ver.-Div. 1 Diesel Gen.

142 C91-013-8.8 DSTC-Verif. of Prog. En. by OD-15 143 C91-013-8.12 DSTC-OD-18 Verification 144 C91-013-8.13 DSTC-OD-2 Verification 145 C91-013-8.29 BOP Calculations 145 C95-099-8.3.2 Div. 2 Diesel Event Markers 147 N27-023A-8,5 EW Master Controller RFP A 148 N27-023A-8.5 EW Master Controller RFP B 149 G42-119-8.1 SPCU Performance Test 150 B21-026-8.2 SRV Flow Test 151 E12-071-8.2 Supp. Pool Cooling Loop "A" 152 E12-071-8.2 Supp. Pool Cooling Loop "B" 153 B21-033-8.3 NSSS Pip Vib.-SRV Manual Test 142

Attachment 1 (Cont.)

Startup Test Performance Log Through the End of Test Condition 2 (Cont.)

Test Serial No. STI No. - Section STI Title (Section)

Test Condition One (Cont.)

154 P99-122-8.2 BOP Pipe-Transient SRV Manual Test 155 C51-012-8.2 APRM Calibration at High Power 156 P99-122-8.4 BOP Piping Test Steam Condensing 157 C95-099-8.3.3 DIV. 3 Diesel Generator (Event Marker) 158 M99-134-8.1 RHR A Pump Room Cooler Data 159 M99-134-8.3 RCIC Pump Room Cooler Data 160 P45-117-8.1 ESW System - Loop A 161 M99-134-8.6 LPCS Pump Room Cooler Data 162 P42-114-8.1 ECC System - I4cp A 163 E12-071-8.1 Steam condensing Capacity & Stability 164 M99-134-8.4 RHR C Pump Room Cooler Data 165 P45-ll7-8.2 ESW System - Loop B 166 P42-114-8.2 ECC System - Loop B 167 M99-134-8.2 RHR B Pump Room Cooler Oata 168 M99-134-8.5 HPCS Pump Room Cooler Data 169 C61-028-8.1 Shutdown From Outside Control Room 170 C61-028-8.2 Cooldown From Outside Control Room 171 C11-005-8.7 Scram Timing During Planned Scram 172 C95-009-8.3.5 Scram (B-1) Test 173 C95-099-8.4.2 Scram Timing During Planned Scram 174 B21-033-8.3 NSSS Pipe Vibration-Remote S/D 175 B33-029A-8.2 Position Loop & Deadband Test @ TC-1 176 P99-122-8.2 BOP Piping Test-Transient Remote S/D 143

Attachment 1 (Cont.)

Startup Test Performance Log

'Ihrough the End of Test Condition 2 (Cont. )

Test Serial No. STI No. - Section STI Title (Section)

Test Condition Two 177 MP-203-8.1 Test condition 2 Testing 178 P99-122-8.1 BCP Exp. Cycle #4 (Retest) 179 C51-010-8.3 IRM-APRM Overlap

180

181 182 N27-C23A-8.6 m Control Linearity I'IP) 183 R63-133-8.1 LFPS Baseline Data 184 P99-122-8.2 BOP Piping Test - Transients - RCIC Trip / Restart 185 E51-014-8.3 RCIC Discharge to Vessel (Rated) 186 B21-033-8.1 NSSS SS Vib/ Strain - Rx 0 50% Steam Flow 187 P99-122-8.3 BOP Pipe Vib. - Power Plateau -

Rx @ 50% Power 188 N27-023A-8.4 m Pump Open Loop Steps (RFPA) 189 N27-023A-8.4 m Pump Open Loop Steps (RFPB) 190 N27-023A-8.4 m Pump Open Loop Steps (MFP) 191 N27-023A-8.3 m S/U Rx Lvl Controller Steps (RFPA) 192 N27-023A-8.3 m S/U Rx Lvl Controller Steps (RFPB) 193 N27-023A-8.3 m S/U Rx Lvl Controller Steps (MFP) 194 C91-019-8.1 Core Thermal Power Determination 195 C91-019-8.2 Core Thermal Limits Determination 196 C51-012-8.2 APRM Calibration High Power 197 P35-001-8.4 Chemistry Data Power Ascension 144

Attachment 1 (Cont.)

Startup Test Performance Log -

'Ihrough the End of Test Condition 2 (Cont. )

Test Serial No. STI No. - Section STI Title (Section)

Test Condition Two (Cont.)

198 E51-014-8.4 RCIC Cold Quick Start to Vsl - 1st '

199 E51-014-8.5 Operation of RCIC from Remote Shutdown

~

Panel

200

201 202 C85-022-8.1 Pressure Reg Test with Control Valves 203 C85-022-8.3 Pressure Reg Test with Bypass Valves 204 N27-023A-8,5 FW Master Rx Level Controller Steps -

RFPA & B 205 N27-023A-8,5 FW Master Rx Level Controller Steps -

MFP & RFPA 206 N27-023A-8,5 FW Master Rx Level Controller Steps -

MFP & RFPB 207 B33-030E-8.1 FCV Cavitation Interlock Test 208 R43-031-8.1 Loss of Turbine - Generater & Offsite Power 209 B21-027-8.1 Generator Load Reject within Bypass Valve Capacity

Void

- Cancelled M.M.-3 #83 4

145 -

'a' y

T, ! s! t.r (a' ) i:t. il< f' a,i_ ,n, Iit, .,[1 ?.?

-.m i, C o i f' ti T . f : u i

J !! ! it t,',! q i :! ;j (1

,t% o s. t 1 .. - i, u P )T'

> le D ,&, M,, V

,,, . 2 i

P O Box 5000 - CLEVELAND, OHIO 44101 - TELEPHONE (216) 622-9800 - ILLUMINATING BLDG. - 55 PUBLIC SQUARE =

Serving The Best Location in the Nation c RJRRAY R. EDELMAN

$R VCE ME59ENT NUC LE A R June 5, 1987 PY-CEI/NRR-0661 L Document Control Desk U.S. Nuclear Regulatory Commission Washington, D. C. 20555 -

Perry Nuclear Power Plant Docket No. 50-440 '

Supplemental Startup Report ,

(Supplement #1)

Dear Sir:

Attached is the first Supplemental Startup Report for Perry Unit I covering the time period from March 1, 1987 through May 15, 1987, inclusive. This report is submitted in accordance with Regulatory Guide 1.16 and Perry Technical Specifications, Section 6.9.

Please feel free to contact me if you have any questions.

Very truly your ,

Murray Edelman Senior Vice President Nuclear Group M MRE:njc Attachment cc: T. Colburn K. Connaughton USNRC Region 111 Director, Office of Resource Management b

? 1 i\

- - - - ------

PY-CEI-NRR-0661, Startup Rept,Supplemental - 1

Contents

Text

SUMMARY

SUMMARY

Dear Sir:

Navigation menu

PY-CEI-NRR-0661, Startup Rept,Supplemental - 1

Text

SUMMARY

SUMMARY

Dear Sir:

Navigation menu

Search