ML17229B013
| ML17229B013 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/12/1999 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| L-99-32, NUDOCS 9902260043 | |
| Download: ML17229B013 (44) | |
Text
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CATEGORY 1 REGULATORY INFORMATION DISTRIBUTIOOSYSTEM (RIDE)
ACCESSION NBR;9902260043 DOC.DATE: 99/02/12 NOTARIZED: NO FACIL:50-..335 St. Lucie Plant, Unit 1, Florida Power
&, Light Co.
, 50>589.St.
Lucie Plant, Unit 2, Florida Power
& Light Co..
AUTH.NA?4E..
AUTHOR AFFILIATION STALL,J.A.
Florida Power
&. Light Co.
RECIP:NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)
SUBJECT:
Forwards 1999 plant reference simulator certification rept for St Lucie,Units 1
& 2,per. 10CFR55.45(b)(5)(ii) 10CFR55.45(b) (5) (vi).Rept contains description
& schedules for correcting test failures
& testing completed.
DISTRIBUTION CODE':
ENCL 1
- SIZE, TITLE: Simulator Facility Certification - GL-90-08 DOCKET 05000335 05000389 A
T E
NOTES:
RECIPIENT ID CODE/NAME GLEAVES,W COPIES LTTR ENCL 1
1 RECIPIENT ID CODE/NAME COPIES LTTR ENCL INTERNA E CENTER 0 1 EXTERNAL: NRC PDR 1
1 1
1 D
E N
NOTE TO ALL MRIDSM RECIPIENTS:
PLEASE HELP US TO REDUCE WASTETH TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTR1BUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)
ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED:
LTTR 3
ENCL 3
Florida Powers Light Company,6351 S. Ocean Orive, Jensen Beach, FL34957 February 12, 1999 L-99-32 10 CFR 55.45 10 CFR 50.4 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 RE:
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 1999 Plant-Referenced Simulator Certification Re rt Pursuant to 10 CFR 55.45(b)(5)(ii) and 10 CFR 55.45 (b)(5)(vi), attached is the 1999 Plant-Referenced Simulator Certtftcatt'on Report for St. Lucie Units 1 and 2.
This report is required every four years.
The original certification was submitted on February 21, 1991, by FPL letter L-91-48.
FPL letter L95-04 submitted the 1995 Plant Referenced Simulator Certijication Report on February 15, 1995.
The rcport is required to contain a description and schedule for correcting test
- failures, a description of the testing completed, and a description and schedule of testing, if different, to be performed during the next four year interval.
The required information is included in the enclosed report. Section 1 is a list of certification tests performed by year during the past four years.
Section 2 is a list of open deficiencies identified during the interval tests and includes the scheduled completion dates.
Section 3 identifies the single plant change/modification that has not yet been incorporated into the simulator.
Section 4 is a list of additions, deletions, and revisions to the certification test program.
Section 5 is the test schedule for the next interval.
Section 6 includes abstracts of new test procedures to be included in the next interval.
Please contact us ifthere are any questions about this submittal.
Very truly yours, J. A. Stall Vice President St. Lucie Plant JAS/GRM 2 t> 0 r~'r 4 7 cc:
Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant 9902260043 9902i2 PDR ADQCK 05000335 P
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L
D St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 1 ST. LVCIEUNITS 1 AND2 1999 PLANTREFlHU<RICED SIMULATOR CERTIFICATIONREPORT
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 2 SECTION 1 CERTIFICATIONTEST LIST FOR PERIOD 1995-1998
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L99-32 Enclosure Page 3 Cert. Test 0'RN-001 Reactor Tri Certification Tests Performed in 1995 Certification Test Title TRN-002 Loss OfAllFeedwater TRN-003 Main Steam Valve Closure TRN-004 Loss Of AllRc s
TRN-005 Loss Of One RCP TRN-006 Turbine Tri From (15% Power TRN-007 Maximum Rate Power Ram TRN-008 Lar e Break LOCA With LOOP TRN-009 Mslb Inside Containment TRN-010 Failed 0 en PZR Safet Valve With No HPSI MAL-002 Ra id Gross Failure OfMulti le Steam Generator Tubes MAL-005 Small Break Loca MAI 006 Failed 0 en PORV With Loss Of Off-Site Power MAL-008 Loss OfInstrument AirCom ressors MAL-010 Loss Of Off-Site Power With Both Diesel Generators Failin MAL-017 Loss Of Condenser Level Control MAL-025 Loss Of Both Main Feedwater Pum s
MAL-046 Wide Ran e Nuclear Instrumentation Failed Hi h MAL-050 T-Cold Failed Hi h n ut To RPS MAL-054 Rcs Flow Steam Generator Differential Pressure Instrument Failure MAL-068 Antici ated Transient Without Scram Atws NPE-001 Reactor Plant Heatu Cold To Hot Standb NPE-002 Nuclear Startu From Hot Standb NPE-003 Turbine Startu And Generator S nchronization NPE-005 Plant Shutdown From Rated Power To Cold Shutdown SST-001 Stead State Test At100% Power SST-002 Stead State Test At 30, 50 And 70% Power
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I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L99-32 Enclosure Page 4 Cert. Test ¹ TRN-001 TRN-002 TRN-003 TRN-004 TRN-005 TRN-006 TRN-007 TRN-008 TRN-009 TRN-010 MAL003 MAL-009 MAL 012 MAL-013 MAL-026 MAI 030 MAL-032 MAL-035 MAL-042 MAL-045 MAL-047 MAL-051 MAL-055 MAL-059 MAL-062 MAL-064 MAL-066 MAL-067 MAL-068 SST-001 SST-002 Certification Tests Performed in 1996 Certification Test Title Reactor Trip Loss Of AllFeedwater Main Steam Valve Closure Loss OfAllReps Loss Of One RCP Turbine Trip From (15% Power Maximum Rate Power Ramp Large Break LOCAWith LOOP Mslb Inside Containment Failed Open PZR Safety Valve With No HPSI LOCA Outside Containment In The Letdown System Loss Of Off-Site Power Loss OfMAInstrument Bus Loss Of Non-Safety VitalAC Loss OfAllFeedwater Pumps One Dropped Rod Test Freeze Control Rod Drive System Trip Generator From 100% Power Main Steam Line Break Outside Containment Large Feedwater Line Break Inside Containment Linear Power Range Failed High T-Cold (Rrs) Failed High Feedwater Flow (Input To 3-Element Controller)
Rwt Level Transmitter (Safety Channel) Failure Alarm Window Incorrectly Actuates ESFAS Failure With Small Break LOCA MSIS Fails To Actuate AFAS Fails To Actuate Anticipated Transient Without Scram (Atws)
Steady State Test At 100% Power Steady State Test At 30, 50, And 70% Power
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 5 Cert; Test ¹
'ertification Tests Performed in 1996 Certification Test Title SUR-002 Isothermal Temperature Coefficient Determination SUR-004 Aro Critical Boron Concentration Determination SUR-005 Plant Heat Balance SUR-008 Surveillance Requirements For Shutdown Margin, Modes 2, 3, 4, And 5 Subcritical SUR-018 Boron Flow Test
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A St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 6 Cert. Test 0 TRN-001 Reactor Trip Certification Tests Performed in 1997 Certification Test Title TRN-002 TRN-003 TRN-004 TRN-005 TRN-006 TRN-007 TRN-008 TRN-009 TRN-010 MAL-001 MAL-004 MAL-007 MAL-011 MAL-021 MAL-022 MAL-024 MAL-028 MAL-029 MAL-031 MAL-034 MAL-038 MAL039 MAL-040 MAL-041 MAI 043 MAL-048 MAL-052 1VGG 056 MAL-058 MAI 060 Loss OfAllFeedwater Main Steam Valve Closure Loss OfAllReps Loss Of One RCP Turbine Trip From (15% Power Maximum Rate Power Ramp Large Break LOCAWith LOOP Mslb Inside Containment Failed Open PZR Safety Valve With No HPSI Complete Rupture Of One Steam Generator U-Tube Large Break LOCA With LOOP Loss Of Instrument Air: AirHeader Rupture Loss OfA Safety Related AC Bus Loss Of Shutdown Cooling From Suction Valve Closure Loss Of One CCW Header RCS Leak Into CCW From RCP Seal Cooler One Stuck Rod One Uncoupled Rod During Startup One Slipped Rod Test Turbine Trip From (15% Power Pressurizer Pressure And Level Control Failures Reactor Coolant Volume Control Failures Reactor Trip Initiated By A Low Steam Generator Level Double Ended MSLBIn Containment Failed Open Main Steam Safety Valve Steam Generator Level Instrument Failure Rcs Hot Leg Temperature Monitor (Control Channel) Failure Steam Flow (Input To 3-Element Controller) Failure Containment Radiation Monitor Failure (Esfas)
Annunciator Panel Failures
I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 7 t
Cert. Test ¹ Certification Tests Performed. in 1997 Certification Test Title MAI 065 RAS Fails To Actuate NPE-004 Reactor Trip And Recovery To Rated Power SST-001 Steady State Test At 100% Power SST-002 Steady State Test At 30, 50, And 70% Power SUR-003 Rod Worth Measurement SUR-009 Reactor Coolant System Inventory Balance SUR-010 Wide Range Nuclear Instrumentation Channels Functional Test SUR-012 SUR-013 Reactor Protection System Periodic Logic Matrix Test Cea Periodic Test SUR-014 Turbine Valve Testing SUR-015 Hydrogen Recombiner Test
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 8 Cert. Test ¹ TRN-001 Reactor Trip CertiTication Tests Performed in 1998 Certification Test Title TRN-002 TRN-003 TRN-004 TRN-005 TRN-006 TRN-007 TRN-008 TRN-009 TRN-010 MAL-014 MAI 015 MAL-016 MAL-018 MAL-019 MAI 020 MAL-023 MAL-025 MAL-027 MAL-033 MAI 036 MAI 037 MAL-044 hGQ 049 MAL-053 MAL-057 MAL-061 MAT 069 MAL070 SST-001 SST-002 Loss OfAllFeedwater Main Steam Valve Closure Loss OfAllReps Loss Of One RCP Turbine Trip From (15% Power Maximum Rate Power Ramp Large Break LOCA With LOOP Mslb Inside Containment Failed Open PZR Safety Valve With No HPSI Loss Of 2B/2BB DC Bus Loss OfAllReps; Natural Circulation Cooldown Slow Condenser Vacuum Leak Loss Of One ICW Header Rupture Of One ICW Header Loss Of Shutdown Cooling Rupture Of "B" CCW Header Loss Of Both Main Feedwater Pumps Failed Power Supply To One RPS Channel Excessive Reactor Coolant Activity Inadvertent Dilution At Power Steam Bypass Control System Valve Pails Open Small Feedwater Line Break Outside Containment Containment Pressure Transmitter Failure Rcs Hot Leg Rtd Failure Steam Generator Pressure Transmitter Failed Low Alarm Window Fails To Actuate Hot Shutdown Control Panel Cooldown Loss Of One Heater Drain Pump From 100% Power Steady State Test At 100% Power Steady State Test At 30, 50, And 70% Power
I fl St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L99-32 Enclosure Page 9 Cert.; Test"¹..
. Certification Tests Performed in 1998
',.-:--~---':-':,".: Certification-Test;-.:Title SUR-006 SUR-011 Moderator Coefficient Determination At Power Diesel Generator Monthly Test Note:
The following certification test numbers were deleted as noted in Table 3-2, St. Lucie Certification Test Matrix, of the Initial Certification Report dated February 15, 1991 SUR-001 MAL-063 SUR-007 SUR-016 SUR-017 SUR-019
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 10 SECTION 2 OPEN DISCREPANCY REPORTS ON CERTIFICATIONTESTS
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 11 During the previous certification interval, a total of 21 deficiency reports have been written during certification tests.
Of these 21 deficiency reports, none remain open at the conclusion of the interval.
I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 12 SECTION 3 PLANTCHANGESMODIFICATIONS(PC/M)
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 13 Of the plant change/modifications (PC/M) reviewed and approved for incorporation in the Simulator, only two remain outstanding.
- 1. PC/M 95178-2 has replaced the outdated Tracor Westronics recorders on the Unit 2 HVAC panel with a newer model recorder.
Replacement recorders are on order for the Simulator HVACpanel.
The expected completion date for this PC/M is July 2000.
2.
The reactor turbine gage boards (RTGB) and other cabinets in the plant have been repainted a different color. The schedule for painting the Simulator has not been determined.
I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 14 SECTION 4 ADDITIONS,DELETIONS, ANDREVISIONS TO THE CERTIFICATIONTEST PROGRAM
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 15
- Pa'ordure No.
MAL-069 MAL070 MAL-071
'.-':-:Pr ocedure Bescriptioii',,:...,
Hot Shutdown Control Panel Cooldown The original initialcondition of this certification test has the Simulator in hot standby with control already established at the Hot Shutdown Panel.
In order to increase the dynamic scope of this test, the initial condition of the test has been changed to steady state 100% power.
This change allows the certification test to include the Simulator's response to all of the applicable initial steps of 2-ONP-100.2, Control Room Inaccessibility.
These steps include the tripping of the reactor and turbine, the securing of RCPs and other system components, and finall the transfer of control to the Hot Shutdown Panel.
Loss of One Heater Drain Pump The last Simulator Four-Year Report committed to developing and performing a new Malfunction Certification Test.
The new test involved the loss of one Heater Drain Pump (HDP) from 100% power.
The report stated that the HDP would be lost due to a malfunction with its electrical breaker.
Subsequently, it was decided to use a failure of the 4B Feed Water Heater Alternate Drain to trigger the malfunction and the ultimate loss of the 2B HDP. This scenario was chosen in an attempt to replicate the events of IHE-92-067 closer.
This failure examines a larger portion of the Simulator's "balance ofplant" response.
The final condition of the certification test has the Simulator stable at approximately 93% power with the 2B HDP off. To restart the 2B HDP and return the Simulator to 100% power, as stated in the report is not necessary.
Certification Test NPE-003, Turbine Startup and Generator Synchronization, examines a similar Simulator response when the second HDP is started durin the ower increase'o 100%.
Plant Transient to Solid Plant Operation Due to an increased interest in solid plant operations, a certification test for this condition willbe developed and performed during the ensuing four year cycle.
The initial conditions willbe steady state 100% power. A series of plant malfunctions willthen be utilized to take the plant solid. The final condition of the test willbe solid plant operations with the RCS temperature and pressure stable and within acce table limits.
St. Lucie Units 1 snd 2 ~
Docket Nos. 50-335 and 50-389 L99-32 Enclosure Page 16 SECTION 5 NIt22T FOUR YEARTESTING SCMU)ULE
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St. tstcie Units 1 snd 2 ~
Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 17 Cert. Test 0 Certification Test Schedule for 1999 Certification Test Title TRN-001 TRN-002 TRN-003 TRN-004 TRN-005 TRN-006 TRN-007 TRN-008 TRN-009 TRN-010 MAL-002 MAL-005 MAL-006 MAL-008 MAL010 MAL017 MAL-022 MAL-025 MAI 035 MAL-046 MAL-048 MAL-050 MAL-054 MAI 058 MAL-066 MAL-068 MAL-071 NPE-005 SST-001 SST-002 Reactor Trip Loss OfAllFeedwater Main Steam Valve Closure Loss OfAllReps Loss Of One RCP Turbine Trip From (15% Power Maximum Rate Power Ramp Large Break LOCA With LOOP Mslb Inside Containment Failed Open PZR Safety Valve With No HPSI Rapid Gross Failure OfMultiple Steam Generator Tubes Small Break Loca Failed Open PORV With Loss Of Off-Site Power Loss OfInstrument AirCompressors Loss Of Off-Site Power With Both Diesel Generators Failing Loss Of Condenser Level Control Loss Of One CCW Header Loss OfBoth Main Feedwater Pumps Trip Generator From 100% Power Wide Range Nuclear Instrumentation Failed High Steam Generator Level Instrument Failure T-Cold Failed High (Input To RPS)
Rcs Flow (Steam Generator Differential Pressure) Instrument Failure Containment Radiation Monitor Failure (Esfas)
MSIS Fails To Actuate Anticipated Transient Without Scram (Atws)
Plant Transient To Solid Plant Operation Plant Shutdown From Rated Power To Cold Shutdown Steady State Test At 100% Power Steady State Test At 30, 50, And 70% Power
St. Lncie Units i snd 2 ~
Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 18 Cert. Test 0'RN-001 Reactor Trip Certification Test Schedule for2000 Certification Test Title TRN-002 Loss OfAllFeedwater TRN-003 Main Steam Valve Closure TRN-004 Loss OfAllReps TRN-005 Loss Of One RCP TRN-006 Turbine Trip From (15% Power TRN-007 Maximum Rate Power Ramp TRN-008 Large Break LOCAWith LOOP TRN-009 Mslb Inside Containment TRN-010 Failed Open PZR Safety Valve With No HPSI IVGQ 003 LOCA Outside Containment In The Letdown System MAL-009 Loss Of Off-Site Power MAL-012 Loss OfMAInstrument Bus MAL-013 Loss Of Non-Safety Vital AC MAL-026 Loss Of AllFeedwater Pumps MAL-030 One Dropped Rod Test MAI 032 Freeze Control Rod Drive System MAL-042 Main Steam Line Break Outside Containment MAL-045 Large Feedwater Line Break Inside Containment MAL-047 Linear Power Range Failed High MAL-051 T-Cold (Rrs) Failed High MAL-055 Feedwater Flow (Input To 3-Element Controller)
MAL-059 Rwt Level Transmitter (Safety Channel) Failure MAL-062 Alarm Window Incorrectly Actuates
~064 ESFAS Failure With Small Break LOCA MAL-067 AFAS Fails To Actuate NPE-001 Reactor Plant Heatup Cold To Hot Standby NPE-002 Nuclear Startup From Hot Standby NPE-003 Turbine Startup And Generator Synchronization SST-001 Steady State Test At 100% Power SST-002 Steady State Test At 30, 50, And 70% Power
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St. Lucie Units i snd 2 ~
Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 19 Cert. Test ¹ SUR-002 Certification. Test Schedule for 2000 Certification Test Title Isothermal Temperature Coefficient Determination SUR-004 Aro Critical Boron Concentration Determination SUR-005 Plant Heat Balance SUR-008 Surveillance Requirements For Shutdown Margin, Modes 2, 3, 4, And 5 Subcritical SUR-018 Boron Flow Test
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 20 Cert. Test ¹ TRN-001 Reactor Trip Certification Test Schedule for2001 Certification Test Title TRN-002 TRN-003 TRN-004 TRN-005 TRN-006 TRN-007 TRN-008 TRN-009 TRN-010 MAI 001 MAL-004 MAL-007 MAL-011 MAL021 MAL-024 MAL-028 MAL-029 iVGG 031 MAL-034 MAL-038 MAL-039 MAL-040 MAL-041 MAI 043 MAL-052 MAI 056 MAL-060 MAI 065 NPE-004 SST-001 Loss OfAllFeedwater Main Steam Valve Closure Loss OfAllReps Loss Of One RCP Turbine Trip From (15% Power Maximum Rate Power Ramp Large Break LOCAWith LOOP Mslb Inside Containment Failed Open PZR Safety Valve With No HPSI Complete Rupture Of One Steam Generator U-Tube Large Break LOCAWith LOOP Loss Of Instrument Air: AirHeader Rupture Loss Of A Safety Related AC Bus Loss Of Shutdown Cooling From Suction Valve Closure RCS Leak Into CCW From RCP Seal Cooler One Stuck Rod One Uncoupled Rod During Startup One Slipped Rod Test Turbine Trip From (15% Power Pressurizer Pressure And Level Control Failures Reactor Coolant Volume Control Failures Reactor Trip Initiated By A Low Steam Generator Level Double Ended MSLB In Containment Failed Open Main Steam Safety Valve Rcs Hot Leg Temperature Monitor (Control Channel) Failure Steam Flow input To 3-Element Controller) Failure Annunciator Panel Failures RAS Fails To Actuate Reactor Trip And Recovery To Rated Power Steady State Test At 100% Power
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 21 Cert. Test 0 Certification Test Schedule for 2001 Certification Test Title SST-002 8UR-003 SUR-009 SUR-010 SUR-012 SUR-013 SUR-014 SUR-015 Steady State Test At 30, 50, And 70% Power Rod Worth Measurement Reactor Coolant System Inventory Balance Wide Range Nuclear Instrumentation Channels Functional Test Reactor Protection System Periodic Logic Matrix Test Cea Periodic Test Turbine Valve Testing Hydrogen Recombiner Test
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 22 Cert. Test I/
TRN-001 Reactor Trip Certification Test Schedule for 2002 Certification Test Title TRN-002 Loss OfAllFeedwater TRN-003 Main Steam Valve Closure TRN-004 Loss OfAllReps TRN-005 Loss Of One RCP TRN-006 Turbine Trip From (15% Power TRN-007 Maximum Rate Power Ramp TRN-008 Large Break LOCAWith LOOP TRN-009 Mslb Inside Containment TRN-010 Failed Open PZR Safety Valve With No HPSI MAL-014 Loss Of 2B/2BB DC Bus MAL-015 Loss OfAllReps; Natural Circulation Cooldown MAL-016 Slow Condenser Vacuum Leak MAL-018 Loss Of One ICW Header MAL-019 Rupture Of One ICW Header MAL-020 Loss Of Shutdown Cooling MAL-023 Rupture Of "B" CCW Header MAL-027 Failed Power Supply To One RPS Channel MAL-033 Excessive Reactor Coolant Activity MAL-036 Inadvertent Dilution At Power MAI 037 Steam Bypass Control System Valve Fails Open MAI 044 Small Feedwater Line Break Outside Containment MAL-049 Containment Pressure Transmitter Failure EGG 053 Rcs Hot Leg Rtd Failure MAI 057 Steam Generator Pressure Transmitter Failed Low MAL-061 Alarm Window Fails To Actuate
~069 Hot Shutdown Control Panel Cooldown MAL-070 Loss Of One Heater Drain Pump From 100% Power SST-001 Steady State Test At 100% Power SST-002 Steady State Test At 30, 50, And 70% Power SUR-006 Moderator Coefficient Determination At Power
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 23 Cert. Test ¹ Certification Test Schedule for 200? "'"'""-"""';"-.;. -',
Certification Test Title SUR-011 Diesel Generator Monthly Test Note:
The following certification test numbers were deleted as noted in Table 3-2, St. Lucie Certification Test Matrix, of the Initial Certification Report dated February 15, 1991 SUR-001 MAL-063 SUR-007 SUR-016 SUR-017 SUR-019
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 24 SECTION 6 ABSTRACTS OF NEW TEST PROCEDURF8
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 25 MAL-070 LOSS OF ONE HEATER DRhJN PUMP
5.0 DESCRIPTION
5.1 Approach This scenario is modeled after a Unit 2 event, IHE-92-067, where the 2B Heater Drain Pump tripped while at 100% power.
During the event, the 4B Feed Water Heater Alternate Drain, LCV-11-18B, failed to re-close when the water level in the 4B Peed Water Heater returned to normal.
The water drained from the 4B Peed Water Heater to the low-level setpoint, tripping the 2B HDP. This certification test willbe initiated by failing high the controller for LCV-11-18B. The controller willsignal LCV-11-18B to open fully. The level in the 4B Feed Water Heater will drain down to the low-level setpoint and thereby, trip the 2B HDP. A downpower willthen be performed in order to stabilize the Simulator at approximately 93% power.
5.2 Initial Conditions/Final Conditions Initial Conditions:
100% Power, Steady State, Middle Of Life.
Final Conditions:
Stable at approximately 93% power with the 2B HDP off.
5.3 Options The simulator is capable of a loss of a HDP from several different causes.
For the purpose of this test, the HDP willtrip due to a low level in the 4B Peed Water Heater.
This scenario was chosen in an attempt to replicate the events ofIHE-92-067.
5.4 Limitations and Assumptions This test involves only a loss of the 2B HDP, and does not include any other possible malfunctions.
Due to different idiosyncrasies of individual operators, there willbe differences in how the downpower is performed and the how the Simulator is stabilized.
Therefore, the various graphs recorded during subsequent certification tests may not agree exactly.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L99-32 Enclosure Page 26
5.0 DESCRIPTION
- (continued) 5.5 General Certification Test Instructions The Certification Test (CT) provides the specific instructions for operator actions which would be performed in the plant procedures.
For each step required to be performed, a space is included in this procedure for the initials of the individual who is performing the test.
6.0 OBJECTIVES
The objectives of this test are as follows:
6.1 Verify the proper response of the simulator to a loss of one HDP pump.
(ANSI 3.5, Section 3.1.2. (18))
6.2 Ensure that the operator was required to take the same action on the simulator to mitigate the consequences of a loss of one HDP pump as would have been required on the reference plant using the plant procedures.
(ANSI 3.5, Section 3) 6.3 Verify that the operators/instructors did not observe a difference between the response of the simulator control room instrumentation and the reference plant.
(ANSI 3.5, Section 3.1) 6.4 Ensure that the transient showed plant operations of the reference plant, which occurred continuously and in real time. (ANSI 3.5, Section 3.1.1 and 3.1.2) 6.5 Verifythat the critical parameters and the other parameters, which were important to the successful completion of this evolution, were displayed on the appropriate instrumentation and provided proper alarm or protective system action, or both.
(ANSI 3.5, Section 3.1.1) 6.6 Verify that the loss of one HDP pump interaction with the other simulated systems provides total system integrated response.
(ANSI 3.5, Section 3.3.1) 6.7 Verify that the simulator does not fail to cause an alarm or automatic action that would occur in the reference plant and conversely, does not cause an alarm or automatic action that would not occur in the reference plant for this evolution.
(ANSI 3.5, Section 4.2.1(c))
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 27
6.0 OBJECTIVES
(continued) 6.8 Verify that the operator was able to control the transient to a steady state condition ifthe simulator operating limits were not exceeded.
(ANSI 3.5, Section 3)
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 28 1VGG 071 Plant Transient to Solid Plant Operation
5.0 DESCRIPTION
5.1 Approach This scenario uses a series of malfunctions to place the RCS into a solid condition.
The first malfunction is a leak in the auxiliary spray system into the Pressurizer (PZR). This causes the PZR pressure to drop below the Thermal Margin Low-Pressure Trip (TMLP) setpoint resulting in an automatic reactor trip. Following the reactor trip, a second malfunction causes a rupture in the main steam line which blows down one of the Steam Generators.
The RCS cooldown and resulting RCS shrink causes the RCS pressure to drop rapidly. The pressure drop is sufficient to initiate SIAS and allow HPSI injection.
Once the Steam Generator has blown down, the RCS cooldown is halted and a RCS heat up starts.
A third malfunction temporarily disables both of the Atmospheric Dump Valves on the unaffected Steam Generator.
This limits the ability of the secondary system to depressurize and effectively remove decay heat from the RCS.
The RCS heat up causes a rapid swell in the RCS liquid.
The RCS pressure increases above the PORV setpoint causing them to lift, venting the steam bubble in the PZR to the containment.
Once the steam bubble is completely vented, the RCS becomes a solid system.
Once the RCS is solid, one of the ADVs becomes operable again, allowing the removal of decay heat from the RCS.
The test is complete when the RCS pressure and temperature is stabilized within acceptable limits. This willbe performed by controlling the charging flow and adjusting steam flow through the available ADV.
5.2 Initial Conditions/Final Conditions Initial Conditions:
100% Power, Steady State, Middle Of Life.
Final Conditions:
RCS in solid condition with the RCS pressure and temperature stabilized and within acceptable limits.
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-99-32 Enclosure Page 29
5.0 DESCRIPTION
- (continued) 5.3 Options The simulator is capable of being placed in solid plant operations from several different causes.
For the purpose of this test, RCS is to be placed in solid plant operations by a rapid cooldown followed by a rapid heat up.
5.4 Limitations and Assumptions This test involves only the placing the plant in solid conditions, and does not include any other possible malfunctions.
5.5 General Certification Test Instructions The Certification Test (CT) provides the specific instructions for operator actions, which would be performed in the plant procedures.
For each step required to be performed, a space is included in this procedure for the initials of the individual who is performing the test.
6.0 OBJECTIVES
The objectives of this test are as follows:
6.1 Verify the proper response of the simulator to solid RCS Plant operations. (ANSI 3.5, Section 3.1.2. (18))
6.2 Ensure that the operator was required to take the same action on the simulator to mitigate the consequences of solid RCS Plant operations as would have been required on the reference plant using the plant procedures.
(ANSI 3.5, Section 3) 6.3 Verify that the operators/instructors did not observe a difference between the response of the simulator control room instrumentation and the reference plant.
(ANSI 3.5, Section 3.1) 6.4 Ensure that the transient showed plant operations of the reference plant which occurred continuously and in real time. (ANSI 3.5, Section 3.1.1 and 3.1.2)
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 99-32 Enclosure Page 30
6.0 OBJECTIVES
(continued) 6.5 Verify that the critical parameters and the other parameters, which were important to the successful completion of this evolution, were displayed on the appropriate instrumentation and provided proper alarm or protective system action, or both.
(ANSI 3.5, Section 3.1. 1) 6.6 Verifythat the solid RCS Plant operations interaction with the other simulated systems provides total system integrated response.
(ANSI 3.5, Section 3.3.1) 6.7 Verify that the simulator does not fail to cause an alarm or automatic action that would occur in the reference plant and conversely, does not cause an alarm or automatic action that would not occur in the reference plant for this evolution.
(ANSI 3.5, Section 4.2.1(c))
6.8 Verify that the operator was able to control the transient to a steady state condition ifthe simulator operating limits were not exceeded.
(ANSI 3.5, Section 3)