ML20206E757
| ML20206E757 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/10/1988 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | |
| References | |
| NUDOCS 8811180186 | |
| Download: ML20206E757 (14) | |
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BEAVER VALLEY l POWER STATION l i
UNIT 2 l l l i
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STARTUP REPORT !
SUPPLEMENT 3 l l
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DUQUESNE LIGHT COMPANY j OHIO EDISON COMPANY l THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY ,e hhh p
hth h b 4 a 12 PE4
. TABLE OF CONTENTS
- .P_agg i TABLE OF CONTENTS i 11 LIST OF ABBREVIATION! 11
1.0 INTRODUCTION
1 2.0 DEFERRED POST-COMMERCIAL OPERATIONS TEST PROCEDURES 1 2.1 Verification of Plant Performance Following Plant 2 Load Rejection / Trip From Power (IST-2.04.06) 2.2 Verification of Reactor Plant Setpoints 3 (PO-2.01A.03) 2.3 Fuel Pool Cooling System Test (PO-2.20.01) 4 2.4 Spent Fuel Pool and Refueling Cavity Leak Test 4 (PO-2.20.02) 2.5 Fuel Handling Equipment (PO-2.66.01) 4 2.6 Cranes and Lifting Equipment (PO-2.66.02) 5 2.7 Verification of Performance Calculations 5 (S0V-2.05A.03) 2.8 Solid Waste Disposal System Test (SOV-2.18.01) 6 2.9 Automatic Steam Genert.or =*cter Level Control Test 6 (S0V-2.240.01) 2.10 Steam Generator Blowdown System Test (SOV-2.25.01) 7 3.0 POST-COMMERCIAL OPERATIONS CHRON0 LOG 9 l
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ANSI /ANS American National Standards Institute /
American Nuclear Society BVPS Beaver Valley Power Station DF Decontamination Factor EFPH Effective Full Power Hours F Degrees Fahrenheit FSAR Final Safety Analysis Report I&C Instrumentation and Control OST Operations Surveillance Test PCP Process Control Program RCS Reactor Coolant System RTD Resistance Temperature Detector Th Reactor Coolant Hot Les Temperature AT Change in Temperature 11-
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- L 1.0 IhTRODUCTION
- . The third supplement to the Beaver Valley Power Station Unit 2 (BVPS-2) ,
i Startup Report covers the period of testing deferred after Commercial *
- Operations between May 15, 1988 and August 13, 1988. This supplement +
- report is prepared in accordance with Regulatory Guide 1.16 "Reporting j, of Operating Inforraation - Appendix A. Technical Specifications", and l addresses the results of deferred testing identified in the BVPS-2 ,
Final Safety Analysis Report during this period. Those tests which j were deferred, but not completed by August 13, 1988, will also be ,
identified in this report, but a comprehensive discussion on these ;
tests will be provided wnen they are completed. Another supplement ;
report will cover those tests completed subsequent to August 13, 1988. !
l The Unit ran at approximately 100% power for 71 days since t.ho last [
reactor trip on April 4, 1988. On June 15, 1988, reactor power was l reduced dus to condenser back pressure and high hotwell temperature, j At 1701 hours0.0197 days <br />0.473 hours <br />0.00281 weeks <br />6.472305e-4 months <br /> the main turbine was shutdown. Soon after, condanser ;
vacua started to increase and hotwell temperature decreased. On June {
16, 1988 the Unit returned to 100% power, but reduced power to approximately 55%, 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> later for maintenance on the "A" sain (
feedwater pump and "C" water box. The Unit was returned to j approximately *00% power on June 20, 1988. Unit 2 continued to operate ['
at 100% power until 1441 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.483005e-4 months <br /> on July 27, 1988, when a reactor trip occurred during the performance of OST 2.1.1, "Control Rod Partial i Movement Test". I&C was in the process of troubleshooting to determine !
why the "A" shutdown bank rods failed to move, when they pulled the 1AC !
regulation firing card causing all rods controlled frn this cabinet to I fall. This caused a negative rate trip. Repaire to the rod control i system were made and the Unit returned to 100% pwer on July 29, 1988.
Unit 2 continued full power operation through the remainder of this report's period. Section 3.0 of this supplement report contains a :
Post-Commercial Operations Chronolog for additional information. !
2.0 DEFERRED POST COMMERCIAL OPERATIONS TEST PROCEDURES The test procedures found in this section could not be completed prior t to Commercial Operations, and were deferred as concurred with by the !
Onsite Safety Committee cad formally identified to the NRC. All {
remaining test procedures which could not be completed in the time j frame of this supplement report have a scheduled completion date of the ;
first refueling. This section will discuss test results for those test j procedures completed after Commercial Operations between May 15, 1988 i and August 13, 1988. Another supplement report will be issued covering
- the period of August 13, 1988 to November 11, 1988. !
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221 Verification of Plant Performance Following Plant Load Rejection / Trip From Power (IST 2.04.06) 2, . was designed with the Beaver Valley Power Station, Unit capability of sustaining a 100% full load rejection without a turbine trip or a reactor trip. This test performed a full load l rejection from approximately 100% power prior to Commercial !
Operations, but did not satisfy all acceptance criteria. A detailed discussion about the full load rejection can be found in Sect!on 7.2.5 of the original Startup Report. ,
The other portion of this test was to perform a turbine trip from i 100% power. Data obtained from an inadvertent turbine trip from !
approximately 100% power prior to Commercial Operations was !
ctaluated and found to be acceptable to satisfy the performance of j the turbine trip portion of this test. A discussion about the turbine trip can be found in Section 7.2.5 of the original Startup Report.
As discussed in Section 7.2.5 of the original Startup Report, data collected during the inadvertent turbine trip and during the unsuccessful full load rejection turbine trips, could not be used for the Reactor Coolant Hot Les Resistance Temperature Detector (RTD) time response evaluation. Retest 04-002 was generated to obtain the required RTD response time during the next inadvertent reactor trip from 100% power. Since the plant was declared Commercial on the day this retest was initiated, the retest and IST-2.04.06 were both deferred until prior to the first refueling.
I&C was directed to connect a memory type strip chart recorder to the Th RTD outputs at the primary process racks to gather the required data for Vestinghouse during the next reactor trip from 100% power. The recorder would automatically start upon a reactor trip, and Westinghouse would use this information to determine plant RTD time response. An inadvertent reactor trip occurred on November 17, 1987, subsequent to Commercial Operations, but prior to installing tho strip chart recorder. Consequently, the RD response time data was not obtained during this trip.
The Startup Report, Supplement 1 Section 2.3 discusses the results of data obtained during an inadvertent reactor trip from approximately 96% power on January 27, 1988. This data was evaluated by Westinghouse and determined not usable fer the purpose of RTD response time evaluation. On April 4, 1988, an inadvertent reactor trip from approximately 100% power occurred due to a low RCS flow in the "A" loop. Since the analysis .
assumptions for the January 27, 1988 trip were the same for this trip, the data was determined not usable for the purpose of RTD response time evaluation. The analysis assumptions for development of RD response time criteria is that the full reac'or coolant pump flows are maintained in all loops for the duration of the required RTD response time (7 10 seconds). The f ow increase or decrease in any one loop can thue affect the RTD bypass flow and can invalidate the acceptance cel',eria.
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On July 27, 1988 a reactor trip occurred that resulted in the collection of plant data that could be used in the evaluation of overall RTD response times. The data showed that overall RTD response times exceeded the startup test accept ance criteria by approximately 0.25 seconds. Westinghouse performed an initial review of the plant trip data and concluded that, based on experience, the response times may be acceptable since the methodology used to develop the test acceptance criteria typically includes several tenths of a second margin. On several plants, Westinghousa has been able to recalculate the acceptance criteria based on the actual plant configuration at the time of the test, thereby removing margin from the startup test acceptance criteria.
One of the most significant sources of margin is the assumed fuel heat transfer which is a function of fuel burnop. In general, the mora burnup, the greater the fuel heat transfer and the lower the response time acceptance criteria. The analysis performed to determine the startup test acceptance criteria assumes that the fuel burnup at the time of the test is 3000 EFPH. Since fuel heat transfer increases from initial criticality to 3000 EFPH, this is I a conservative assumption for the startup test and provides a j window for which the acceptance criteria is applicable.
Westinghouse subsequently requested additional data from the subject reactor trip, including fuel burnup, and performed a more in depth evaluation of the Beaver Valley Unit 2 data and the plant configuration at the time of trip. This evaluation shosed that the source of margin (i.e., fuel heat transfer) that is usually available may not he available for Beaver Valley since the reactor trip occurred at 6.083 EFPH. Consequently, additional evaluation and calculation of the overall RTD response time acceptance criteria is being performed by Westinghouse. Results of this i evaluation will be discussed in another supplement report. Until then, the Overpcwer and Overtemperature AT reactor trip setpoints ramain conservatively set at 3% below technical specification values in accordance with BVPS Unit 2 Westinghouse Precautions Limitations, and Setpoints Manual.
2.2 Verification of Reactor Plant Setpoints (PO-2.01A.03)
The initial reactor plant setpoints were verifiad by using the latest plant documentat!.on and/or actual instrumentation module settings ensuring the equipment was aligned for the initial startup setpoint values. All setpoints fell within the required Technical Specification limitations. Final setpoint verification
- was started with the plant at 100% power, but could not be completed prior to Commercial Operations because several systems did not have the final setpoint values in place. The remainder of this test was deferred with a required completion date of the first refueling.
At the time of this supplement report, gathering of final setpoint values was still in progress. The results of final setpoint ,
verification will be discussed in another supplement report after obtaining all final metpoint values.
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2.3 Fuel Pool Cooling System Test (PO-2.20.01)
This test was satf rractorily completed prior to core load with the exception of Retest 20-003 which will measure Fuel Pool Cooling Pump 21A vibration the next time the Spent Fuel Pool is filled.
Enginevring resolution for high pump vibration during original i
- esting modified the motor support stand by welding on stiffening plates. The pump motor was retested yielding acceptable vibration levels. The pump and motor were then recoupled, but the Spent Fuel Pool had already been drained. The Spent Fuel Pool will not be refilled until prior to the first refueling so the test was deferred with a required completion date of prior to the first refueling.
At the time of this supplement report, the Spent Fuel Pool had not yet been refilled. The results of Retest 20-003 will be discussed ,
in another supplement report subsequent to completion of the !
retest.
2.4 Spent Fuel Pool and Refueling Cavity Leak Test (PO-2.20.02) ;
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, This test was satisfactorily completed prior to core load with the !
exception of Retests20-001 and 20-002. Retest 20 001 was 1 generated when the gates betwr m the Fuel Pool and Cask Area and '
between the Fuel Pool anu Fuel Transfer Canal exhibited c.onsiderable leakage at the bottom seals. Repairs were made to both gates but the retest cruld not be performed until the next
- time the Spent Fuel Pool was filled. Retest 20-002 was generated when the Spent Fuel Pool instrumentation test could not be performed due to unavailability of the Spent Fuel Pool level transmitters. This retest requirad the Spent Fuel Pool to be filled to its high level. The Spent Fuel Pool will not be L refilled until prior to the first refueling so both retests were i deferred with a required completion date of frior to the first l refuS11ng. l At the time of this supplement report, the Spent Fuel Pool had not yet been refilled. The results of Retests 20 001 and 20 002 will L be discussed in another supplement report subsequent to completion -
of the retests.
2.5 Fuel Handling Equipment Test (PO 2.66.01)
This test verified control logic and demonstrated operability of the Fuel Handling Equipment and provided a functional -
demonstration of a simulated fuel transfer between the Fuel )
Building and Reactor Containment. Fuel Handling Equipment testing I was satisfactorily completed to support core load although l portions of the test were not dono due to unavailability of I equipeent.- Those portions of the test not available for testing prior to core load were completed prior to Commercial Operations except for testing of the RCC Change Fixture (Retest 06 005). A discussion on that portion of the test coeplated prior to Commercial Operations can be found in Section 8.4 of the original !
Startup Report. Testing of the RCC Change Fixture (Retest 06 005) i
was deferred with a required completion date of prior to the first refueling.
At the time of this supplement report, testing of the RCC Change Fixture had not yet been started. The results of Retest 06 005 will be discussed in another supplement report subsequent to completion of the retest.
2.6 Cranes and Lifting Equipment Test (PO 2.66.02)
Control logic testing for the Reactor .ontainment Polar Crane and operability of its auxiliary hoist under load was successfull';
completed prior to core load. Polar Crane main hoist operability under load was checked during its load test and .*tisfactority ;
completed during reactor vessel head installation. K.Lest 66-007 was generated to test an indicating light on the telescoping work platform which was inoperative when in the maximum down position.
Outstanding rework on the Spent Fuel Cask Crane was not completed prior to core load because it was thought to be safer to do it without fuel in the Fuel Storage area. Retest 66-007 and the remainder of the test which tests the Spent Fuel Cask Crane, were deferred with a required completion date of prior to the first refueling.
At the time of this supplement report, electrical maintenence repair of the indicating light on the Polar Crane telescoping work platform, and testing of the Spent Fuel Cask Crane had not yet been performed. The results of Retest 06 007 and Spent Fuel Cask Crane testing will be discussed in another supplement report subsequent to completion of testing.
2.7 Verification of Performance Calculations (ETi-2.05A.03)
This test was performed at 30, 50, 75 and 100% power plateaus during Startup, and verified that the computer generated performance calculations, Nuclear Steam Supply System (NSSS), and Bslance of Plant (B0P) programs were accurate at these power levels. Deficiencies at lower power levels were cleared at subsequent power levels or Maintenance Work Requests were issued against deficiencies to perform troubleshooting with any required rotests to be performed by existing plant procedures. i This procedure only satisfied a portion of the BVPS-2 Final Safety Analysis Report (FSAR) test objective that was required to be performed. This limited scope was justified by the fact that a .
comprehensive set of initial operating preceduras fer verifying i the accuracy of the plant performance calculations was being developed and performed by the Computer Integration Group. These tests perform thorough verifications of all the programe and !
calculations related to the plant performance applications of the ;
process co=puter system. The tests are designed in accordance )
with accepted software validation and verification ssandards, '
using tout data sets to prove the accuracy of the program output.
The accuracy of the field sensors was verified by the Phase 1 ;
The accura.y of the raw value conversion into testing program. ,
engineering units was verified by SOV-2.05A.03 as ment'oned above.
- The total of final test results after completion, will fully
- satisfy all.FSAR test objectives. The Computer Integration Group ;
is continuing the testing per the Initial Operating Procedures.
Deferral to beyond Commercial Operations was approved with the ;
estimated completion date for tais testing prior to initial criticality after the first refueling.
At the time of this supplement report, testing by the Computer ,
Integration Group had not yet been completed. All test results e will be discussed in another supplement report following i i
completion of all testing. j 2.8 Solid Waste Disposal System Test (S0V-2.18.01) I
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This test was partially performed prior to core load and deferred [
until needed to provide solid waste disposal capabilities at BVPC- !
- 2. As of Commercial Operations, only a portion of the test was !
i satisfactorily completed. A discussion or that portion completed !
L prior to Commercial Operations can be found in Section 8.8 of the [
original Startup Report. Section 2.13 of Supplement 1 to the ,
Startup Report discusses testing completed oa the Solid Waste i Disposal System af ter Commercial Operations and prior to February (
l 15, 1988. Section 2.8 of Supplement 2 to tho Startup Report j discusses testing completed and in progre,s between February 15, 1988 and May 15, 1988.
l Verification of the Solid Waste Disposal Drumming Station wac i successfully completed on July 20, 1988 by developing the proper ;
mixtures for the various waste streams and combinations of waste i i streams that result in a solidified waste product that meets the j i requirements of 10 CFR 61 "Licensing Requirements for Land ;
Disposal of Radioactive Waste" and ANSI /ANS-55.1-1979, "Solid [
l Radioactive Waste Processing System for Light Water Cooled Reactor j 1 Plants". In this test, waste streams of simulated evaporator i f bottoms, bead resin, powdered resin (sludge) were drumed as well l 1 as mixes of bead resin and simulated evaporator bottoms, and [
j powdered resin and simulated ovaporator bottoms. Starting with 1 I
the vendor (Stock) recommended formuistions, mixture adjustments were made until arrising at a solidification prodect that would l 9
meet all of the test acceptance criteria and code requirements.
l This was accomplished through experimentation with the required
! ingredients by varyttig th$ amounts of each to be plar ed into the
- t j drum. After successful completion, the results of the testing !
j were forwarded to Operations for inclusion into the Process [
j Control Program (PCP) which will be used to control future waste i solidification, j 2.9 Automatic Steem Generator Water Level Control Tent (SOV 2.24C.01) l 1 l
! This test was performed at various modes and power levels during ;
I the Startup Testing Program. A detailed discussion on that t l portion of the test completed prior to Comercial Operations can be fo ud in Section 7.3.3 of the original Stattup Report. ,
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. I During Startup at the 30, 50, 75, and 100% power plateaus, each j
.i steam and feedwater flow transmitter output signal was compared to the associated calorimetric flow creputed from test gauges in i 1 accordance with IST-2.02. ,16 "Thermal Power Calorimetric Test". *
- At each power level, the fee'. water and steam flow rates which were i j calculated did not agree with the calorimetric flow rate criteria. j
- A copy of the test data was forwarded to I&C to make the necessary adjustments to both the feedwater and steam flow transmitters. ,
j hese adjustments and the remainder of the test were not cumpleted 1
) prior to Commercial Operations, and were deferred with a required j completion date of the first refueling. The last section in this :
test will obtain additional flow data at 100% power to determine ;
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the need for further adjustments, and to demonstrate repeatability .
, of the feedwater and steam flow transmitters at 30, 50, 75, and ;
l 100% power during a subsequent plant startup. l At the time of this report, I&C had not made the necessary adjustments to both the feedwater and steam flow transmitters. [
nis :.till precludes the completion of the remainder of this test. ;
Note that I&C had determined th*t there was no problem with a ;
staam flow / feed flow mismatch and recommended that plant operation j continue until final adjustments and checks can be performed. .
Test results will be discussed in another supplement report upon l completion of the remainder of testing. ;
2.10 Steam Generator Blowdown System Test (SOV-2.25.01)
This test was completed prior to Commercial Operations except for clearing of minor deficiencies which occurred prior to Commercial Operations, and verification of operating parameters for the Cleanup Ion Exchangers and the "B' Steam Generator Blowdown Evaporator. A discussion on that portion of the test completed prior to Commercial Operations can be found in Section 8.11 of the original Startup Report. Section 2.10 of Supplement 2 to the Startup heport discusses testing coepisted on the A & B Cleanup Ion Exchangers, A & B Cleanup Strainers and Cleanup Filter.
During this report period, both Evaporators A & B were tested.
Review of test data on Evaporator A from prior to Commerica)
Operations indicated that it operated satisfsetorily. However, ratestint was performed to verify Evaporator A operation at a higher bottoms boron concentration with radioactive liquid waste ,
present in the feedwater source tanks which was not present during the pre-commercial operations testing.
Evaporator A performed satisfactorily except t.at the calculated 6
Decontamination Factor (DF) was low out of specification (780 vs.
> 1000 required), ne DF was determined by dividing the Co 58 activity levels in the evaporator bottoms sample by the Co 58 activity levels in the evaporator distillate discharge line sample. An analysis was pirfntmed using other available data (i.e. , boron concentrations and non Co 58 activity levels obtained from the same samples) and the Chemistry Group was consulted for their opinion. The DF was considered acceptable for the follwing reasons:
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- 1) Operation of the evaporators during this test were at distillate production rates much higher than normal.
- 2) No 'ofoaming agents were used which would have been an acceptable means c' reducing carryover.
l 3) An alternate DF calculated from boron concentrations determined from the s ame s amples yielded a significantly higher end acceptable value of 5953.
Evaporator B also performed satisfactorily. Initial operation d utilized demineralized water as a feed source in place of the present normal liquid waste feed source. This deviation from the intended mode of operation was requested by the Operations Group to help maintain Evaporator B cleanliness conditions for backup boron recovery operations and ALARA considerations. In order to operate in this mode, a temporary operating procedure was developed. Since borated, radioactive liquid waste was not present during this mode of operation, (except due to residual liquid waste in the feed lines and bottoms transfer lines from cperation of Evaporator A), the test results did not include operation at or near 12% boron concentration in Evaporator B bottoms and determination of its DF. However, operation in this mode did verify satisfactory operation of Evaporator B within design temperature, pressure and flow limits.
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3.0 POST-COMM[RCI AL OP[ RATIONS C4NONOLOG ,
PLANT CGNDITIONS EVENT DATE -
YYYYYoooo$ooob
<> m 4=t.a N ee seeeee se ee ee ** ** me l PLANT AT APNtOX8PATELY 100% POWER. 5/15/88 R[ ACTOR POWER REDUCED CUE TO CCADENSER BACK 6f 15/88 PRESSURE AND HIGH HOTWELL TEMPERATURE. MAIN TURBINE SHUIDOWN AT 1!O1 HOURS.
PLANT BACK AT APPROX 1f1Ai[LY 100% POWER. 6/16/88
( RIDUCE D POWER TO APPROKIMAi[LY $$% ROWER FOR 6/17/88 MAINYENANCE ON THE "A" MAIN FEEDWATER PUMP AND "C" WATER BOX.
) PLANT BACK AT APPROXIMAT[LY 100% POWFR 6/20/88
() COMPLEilON OF OP[ RATION OF THE "B" STEAM 7/13/88 CINtRATOR BLOWDOWN [VAPORATOR WiiH BORON SOLUTION TO COMPLETE THE STEAM GENERATOf-BLOWDOWN SYSTEM T[ST (Vil.0, SOV-2.25.01).
() COMPLETION OF TESTING THE DRUMMING STATION 7/20/88 TO COMPLETE THE SOLID WASTE DISPOSAL SYSTEM i[ST (Vil.G, SOV-2.18.01).
l? I' R E ACIOst TRIPPED DUE TO A NECATIVE RATE TRIP. 7/27/88 THE 1AC REGULATIO86 FIRING CARD WAS PULLED G 1441 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.483005e-4 months <br /> DURING TROUBLESHOOilNG TO DETERMINE WHY THE "A" SHUTDOWN BANK RODS FAILED 10 MOVE CAUSING Att ROOS CONTROtt[D fROM THIS C21 NET TO TALL.
IST-2.04.06 (RETEST 04-002) DATA WAS DETERMINLD TO BE USABLE f0R IHE PURPOSE Of RID R[SPONSE TIME
[ VALUATION AND FORWARDED TO WESTINGHOUSE.
REACTOR CRITICAL. 7/28/88
) PLANT BACK AT APPROXIMAi[LY 100% POWER. 7/29/88 j g PLANT CONT 4NLING 100% POWER OP[ RATION. 8/13/88 I
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$$ La Duqu@esne L.idit 1.--,m mm wa~
SNgengport PA 15077&J04 November 10, 1988 U. S. Nuclear Regulatory Commission Attn Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPP-73 BVPS-2 Startup Report, Supplomont 3 Gentlement In accordance with Regulatory Guide 1.16, "Reporting of Operating Information - Appendix A, Technical Specifications", the attached Summary Report of plant startup testing is being submitted. This report is due 90 days following submittal of the BVPS-2 Startup Report, Supplement 2, for Beaver Valley Power Station, Unit No. 2.
This report addresses the results of all startup testing identiiled in the Beaver Valley Power Station Unit No. 2 Final Safety Analysis Report during the period from May 15, 1988 to August 13, 1988.
Completion of the startup test program has not been fully accomplished due to some deferred testing. A supplemental report will be it, sued within 90 days to update the progress of the startup test program.
Very truly yours, f
k . ,&W
. D. Sieber Vice President Nuclear Group Attachment cc: Mr. J. Beall, Sr, Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr. P. Tam, Project Manager i
COMMONWEALTH OF PENNSYLVANIA)
) SS:
COUNTY OF BEAVER )
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on this -
//7 day of / u p ,e /[A / , 1988, before me, e,s /) 5 ~4 rz, a Notary Public in and for said u
v /
Commorwealth and ' County, personally appeared J. D. Sietsr, who being P duly sworn, deposed, and said that (1) he is Vice Presider.t of Duquesne Light, (2) no is duly authorized to execute and file the [
foregoing submittal en behalf of said Company, and (3) the statements set forth in the submittal are true and correct to the best of his knowledge, information and belief.
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