ML20236W996
| ML20236W996 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/09/1998 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20236W985 | List: |
| References | |
| NUDOCS 9808070042 | |
| Download: ML20236W996 (300) | |
Text
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INDEX NRC FORM 474 . TMI CERTIFICATION REPORTMEMO ATTACHMENTS - EXCEPTIONS TO ANSI /ANS 3.5-1985 ATTACHMENTS- CERTIFICATION TESTABSTRACTS A TTACHMENT C - SIMULATOR CERTIFICATION PLAN TESTING SCHEDULED A TTACHMENT D - TMI SIMULA TOR RE-HOST SYNOPSIS ATTACHMENT E - PLANT OPERA TING EXPERIENCE EVALUATIONS O S 0
NRC FORM 474 U.S. NUCLEAR REGULATORV COMMISSION QPPQOtfED BY OM8; NO. 3160 0138 EAPIREQ: C8/31/9[ (eas) Estimated burden per response to compir with th,a mandatory mtormeia cohocw
,.qu.si iro n-s n,s mto,mM,on m used io comfy a s~aw foe.in, ro,wa,d 4 c h'MULATION FACILITY CERTIFICATION comen.nts reoarding bu, den eso,nai. io me ineo,mM,on .nd Records Mans-eni
" (" '. u 8 NucM- R n"001, and to the Paperwo* Reductum Prolect Mory Connissa. w-n.,gion. DC 20sw and 7,udget. Washmgton, DC 20503 NRC may not conduct or sponsar, and a peson Ofree(31504138).
et Managemeit a not requved to respond to, a conection of information unless it dispisys a currently vahd OMB controi number INS TRUCTIONS; This form m to be f%d for niial comfcation, recertdcateon (d requred) and for any change to e simulaleon faculty performance testmg plan made after ntiai submittal or sucn . pian. Provide the following .ntormatm and check the appropnate box to mdcate reason for submittal F ACILITY DOCKET NUMtiLR Three Mile Island 50 289 LICLNSLL DATL GPUN, Inc. This is to certify that 5/28/98 1 The above named facdsty hcensee a using a simulate factiety consisting soiety of a plantveforenced simulator that meets the requwements of 10 CFR $5 45 2 Docurn.niMa is a adabie for NRC revow m ccordance with 10 CrR ss 4s(b> 3 This armulation facdity meets the guidence contained m ANSI /ANS3 6-1985or ANSlJANS3 6-1993. as endorsed by NRC Regulatory Guide i 149 e ine,e are .ny ExCEPneas io in. c.rticcai,on of inia .iem. CHECK uERE : g .nd do.cnb. fusiy on .ddiieonai pages as noc.ssary s ee At t a c hme n t A hAML got other rountrAcahon) AND LOCATION OF SIMULAllON F ACILITY Three Mile Island Referenced Simulator GPU Nuclear Inc., Route 441 South, P. O. Box 480 s Middletown PA 17057-0480 s X SIMULATION FACILITY PERFORMANCE TEST ABSTRACTS ATTACHED (For performance tests conducted m the penod endng wrth the Jafe of thes costrhcahon ) DL6CRIPTiON OF PERFORMANCE TESTING COMPLLTLD (Attach a00rtoonal pages as necessary ana ooennty the stem oescnptuon beng conhnvec) l See Attachment B for Three Mile Island Plant Referenced Simulator certification testing abstracts. O SIMULATION FACILITY PERFORMANCE TESTING SCHEDULE ATTACHED (For the conduct of appicasmately 25'of peermance tests per year fbr the fbur-year penod X commencmg wrth the dare of thas cemhcatson) DLbCRr?T*')N OF PERFORMANCE TLbTING TO BL CONDUCTED (Attach a0@toonalpages as necessary and roentr9y the stem Gescnptron semg conhnved) e Attachment C for certification testing plan for the next four year period. PERFORMANCE TESTING PLAN CHANGE (For any modMcarron to a perfbemance teshng plan submrtred on a prevrous comRcahon) DE SCR! piton OF PERFORMANCE TLSTING PLAN CHANGE fAnach moWtoonal pages as necessary an0 noenth the stem Gescnptu beng contmund) DECERTIFICATION (Descnbe correcfrve acfnons taken, attach results atcompoetedpersbemance reshng se accordance enth 10 CFR 55 4S(bH5)(v) (Attach adthonalpages ss necessary and scontrfy the stem descnptuve borng contrnued) Any faise statement or om.ase m this document, mcluding attachments. Inay be subject to civil ano enmine: sancreons I certify uncer penalty of perjury that the information m inis _ document and attachments e true and correct
$lGNATURL - AUTHORIZLD RLPRE NI AYlVL TITLE - DATE M NQicePresidentandDirector,TMI in -sufe w th 10 CF R 65 5. Communcatsud. tras form snan be submitted to tne NRC es k9hf fonows BY DELIVERY IN PER$0N ONE WHITE FLINT NORTH BY Matt ADDRESSED TO DIRECTOR, OFFICE OF NUCLEAR REACTOR REGULATION TO THE NRC OFFICE AT 11588 ROCKVILLE PIKE U S. NUCLEAR REGULATORY COMMIS$10N ROCKVILLE, MD WASHINGTON, DC 2086541001 NRC F ORM 474 (TE; FRINTED ON RECYCLEDTXUR
\ G NUCLEAR U.
Subject:
1998 SIMULATOR CERTIFICATION Date: 05/15/98 FOUR YEAR REPORT TO THE NRC From: Robert Parnell Location: TMI Nuclear Station 6211-98-042 To: FILE
References:
a) 6221-PLN-2820.01, Three Mile Island Unit 1 Plant Referenced Simulator Certification Plan b) 10 CFR 55.45.b.5, Certification of Simulation Facilities c) ANSI /ANS 3.5-1985, Nuclear Power Plant Simulators for Use in Operator R Training G This report is being generated in accordance with the requirements of References a) and b). The certification tests required for years one through four and the annual tests as described in Attachment C of the four year certification report submitted to the NRC in 1994 were completed. The abstracts describing the tests and results are provided in Attachment B. The Simulator's operability is documented per Reference a) based on the following evaluations:
- Simulator Physical Fidelity
- Simulator Functional Fidelity
- Simulator instructor Console Functions
- Simulator Design Control The statistics for the fcurs years covered by this report are as follows:
- 111 deficiency reports were written and 70 closed
- 99 modification packages were opened and 72 were closed
- 7 physical fidelity exceptions were opened and 44 were closed
- 48 functional fidelny exceptions were opened and 39 were closed 9 There were also five ANSI /ANS 3.5-1985 (Reference c)) exceptions taken for modifications. These exceptions are described in Attachment A.
I i
[ In accordance with Reference a) and the Simulator Training procedures relating to simulator fidelity and design control the following documents have been reviewed for changes:
- 10 CFR 55 Operator Licenses
- Reg Guide 1.149 Nuclear Power Plant Facilities For Use In Operator License Examinations
- ANSI /ANS 3.5-1985 Nuclear Power Plant Simulators For Use in Operator Training
- NUREG 1258 Evaluation Procedure For Simulation Facilities Certified Under s
10 CFR 55 . There have been no changes to the documents affecting simulator certification requirements and the TMI Plant Reference Simulator continues to meet the requirements of these documents except as described in this report and in the June 1994 Simulator Certification Report submitted to the NRC. ' During the period covered by this report the TMI Plant Reference Simulator computer system was upgraded to a PC based system as described in Attachment D. Sufficient certification testing was performed to verify there was no significant change in plant response to the previously benchmarked tests. The schedule of tests for the four year period covering 1998-99 though 2001-02 is included in Attachment C in accordance with Reference b). TMI Training conducted three Plant Operating Experience evaluations which are provided in V Attachment E.
?#R. L. Parnell ill cc:
James Langenbach Randy Hess Larry Noll Michael J. Ross CARlRS: O Yes Om O
ATTACHMENT A j EXCEPTIONS TO ANSI /ANS 3.5-1985 The following is a list of the modifications that were not installed on the simulator within one year as required in Simulator Modification Control Procedure,6221-ADM-2820.07 Section 4.1.3. CYCLE 10 CORE UPDATE MODIFICATION This modification was identified as a simulator upgrade and accepied as a task February 23,1994. Data specific to the simulator core model was needed to perform the model upgrade. The GPUN Fuels group utilizing B&W cycle specific information produced this data. The engineers generated data tables and constant changes to be sy plied to the simulator engineers who in turn calculated simulator specific model values. When the calculations were initially performed by Fuels the simulator assumptions were incorrect, which made the data supplied to the simulator engineers invalid. After determining the origin of the error corrections were made and new data tables generated and supplied to the simulator engineers. The data was incorporated as required into the core model and the modification tested satisfactorily. The time required identifying and correcting the error, and calculating new data was lengthy. These delays resulted in exceeding the required installation date. The model modification was accepted for use for operator training and examination on February 23,1996.
Reference:
DCCE001, Modification 3 #94-003. As a result of the problems encountered a procedure was developed to allow GPUN to perform this task independent of B&W. GPUN lhis and GPUN simulator engineers developed a process and proceduralized it to allow totalin-house modification of the simulator core model. As a result subsequent core model upgrades will be accomplished more expeditiously with less reliance on outside sources. This has also developed a deeper understanding of the simulator model and its capabilities. RC-P-l A/D #3 SEAL PURGE DELETION The modification to delete the #3 Seal Purge flow from the Reactor Coolant Pump model l was incorporated into the Reactor Coolant Pump Seal Package Upgrade. The RCP Seal Package Upgrade was not scheduled for completion until after the one yen date for the
#3 Seal Purge removal. #3 Seal Purge flow is not readily apparent to the operators in the i control room and is a simplification in the Reactor Coolant Pump Seal Package model l requiring only an open valve on a control room panel. Delaying removal was determined to have no impact on training or license examinations. Delaying the removal of the seal purge flow impacted no malfunction or dynamic response. The purge supply valve position was changed in the controlled initial conditions indicating to the operator that the s seal purge flow was terminated. The fact the software had not been updated was totally I
i transparent to all simulator users. No design control exception was initiated for this f modification.
Reference:
Modification #94-004. U NEW PPC POINTS FOR DC DISTRIBUTION MONITORING This modification to the Replica Plant Process Computer was not installed within the required 12 months due to an oversight. The modification was installed on the pkmt computer December 28.1995 and installed l on the sunulated plant computer hiarch 12,1997. RCP VIHRATION MONITOR PANEL REPLACEMENT The control room panel that displays Reactor Coolant Pump and hiotor vibration was replaced with a new digital and analog display. The digital technology involved required a new interface component be manufactured to provide simulator I/O connection. The component was designed and construction began by vendor. That vendor determined his design would not work and abandoned the project. A new vendor was selected and the component successfully manufactured but the delay caused installation and testing to go beyond the required one-year date. The modification was initiated hiarch 20,1996 and completed on
. January 19.1998. Reference hiodification #95-041.
NEW PDMS COMPUTER POINTS New computer points for monitoring conditions in Three hiile Island Unit 2 were added to the phmt pnx:ess computer. These points were not added to the simulator plant process computer within the required one-
] year due to an oversight. These points are not simulated so installation had no impact on training or examination of licensed operators. Installation is performed only to keep the software on the Replica Plant d- Process Computer the same as the software on the Plant Process Computer.
Reference:
hiodification #96-006. s i J
g ATTACIIMENT H I TIIREE MILE ISLAND PLANT-REFERENCED SIMUL ATOR k TESTING COMPLETED YEAR #1 TESTS (1994-1995) Benchmark Tests TTS07 RCS Safety Valve Failure TTS19 less of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Steam Leak Inside Reactor Building TTS56 Manual Reactor Trip TTS57 Simultaneous Closure of All Main Steam isolation Valves TTS58 Imss of One Reactor Coolant Pump TTS59 Maximum Rate Power Ramp TTS60 Loss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Simulator Stability , SSP 02 Simulator Accuracy Real Time Test RTT01 Re:d Time Test O Transient Tests 1TS01 OTSG Tube Leak g TTS02 OTSG Tube Rupture ) TTS03 RCS Leak Inside Contaimnent ' 1TSM _ RCS Leak Outside Containment TTS05 Large Break LOCA TTS06 Small Break LOCA TTS08 RCS PORV Failure TTSO9 Loss ofInstrument Air TTS10 Station Blackout TTSil DC Distribution Failure TTS12 Emergency Diesel Generator TTS13 6.9 KV Bus Fault TTS14 4.16 KV Bus Fault Nornud Operations Tests l NOT01 Plant Heatup j . NOT02 Plant Startup l NOT03 Reactor Trip and Recovery NOT04 3 RC Pump Operation NOT05 Zero Power Physics Testing NOT06 Core Flood System Valve Operability Test NOT07 Emergency Power System
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} THREE MILE IS W SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT i ]/
j C Test Identification: RCS Safety Valve Failure Certification Test Transient Test TTS07 ANSI / ANS - 3. 5 Reference (s): B.2.2.(10) 3.1.2(1b) 3.1.2(1d) Test Date: 05/24/95 Malfunction (s) Tested:* ES01A ESAS Failure to Actuate at HPI Setpoint Channel A ES01B ESAS Failure to Actuate at' HIP Setpoint Channel B RC27A Pressurizer Safety Valve Fails Open 25% (100% = ; Full Open)
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization; Protected Initial Condition IC-15 1 100% Reactor Power f Equilibrium Xenon Beginning of Cycle { j l Point of Test Termination: This test was terminated with the reactor tripped, J
,/'] Reactor Coolant System saturated, pressurizer filled and I
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system inventory decreasing. Simulator Run Time: 10 minutes } Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test TTS07 dated 02/27/90.
,Right Direction Analysis.
Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of l a simulated malfunction. i This transient resulted in the loss of Reactor Coolant System inventory through a partially open Code Safety Valve on the pressurizer. The pressure in the Reactor 7 Coolant System decreased to saturation and the pressurizer filled solid. The Engineered Safeguards Actuation System was prevented from operation.
/mi \ _,/ The event was initiated by failing a Pressurizer Code Safety Valve partially open. As Reactor Coolant System pressure decreased the Reactor Protection System actuated tripping the reactor and initiating Reactor Trip Containment Isolation. Pressure continued to decrease and Engineered Safeguards Actuation System High Pressure
Injsetion actustion satpoint was reached. Tha system did not actuated dus to the malfunctions inserted. Reactor Coolant System pressure continued to decrease; neactor Vessel and Reactor Coolant System voiding caused the pressurizer to fill. Reactor Coolant Drain Tank pressure increased causing rupture disk rupture with subsequent activity release to containment. Due to model changes the RCDT variable could not be plotted. RCDT pressure was observed on the control room instrument and instructor station' monitored parameter plot. RCDT pressure response was correct before and af ter rupture disk operation. During the conduct of this test simulator dynanic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Management. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from e:tpected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to ! those expected from the actual reference plant and do not violate the physical j laws of nature. 1 The simulator satisfied this requirement with no identified deviations, f
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CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE Completed By ,,, ._ .
- William Orle Date 06/29/95
- O Approved By ,/ g Robert L. Parnell Date 06/30/95 h
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'I J Y' TRER MILE ISLAN SIMULATOR MANAGEMENT '
CERTIFICATION TEST ABSTRACT l Test Identification: Imse of Forced Flow Certification Test J Transient Test TTS19
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' ANSI /ANS - 3.5 Rafaranca(s): .
3.1.2(4) Benchmark Test B2.2(4) Test Date: . 01/09/95 Malfunction (s) Testad:* RC 35 A/B/C/D - RC Pump Trip A/B/C/D
- Refer to Malfunction Cause and Effects Documents.
Test Initialization: Protected Initial Condition IC-15 1004 Reactor Power Equilibrium Xenon Beginning of Cycle l i Point of Tant Termination: React'or tripped with Raactor Coolant System heat transfer by stable natural circulation.
^
l Simulator Run Time: 10 minutes j 's4=nlator Evaluation Tf==: 6 hours. Baselina Evaluation Data: Previous Certification Test TTS19 dated 02/28/90. g - Right Direction Analysis. RETRAN Data. Current Controlled Copies of: Reference Plant Alars Response Procedures !- Reference Plant Emergency Procedures l
- Overall-Tant Rasulta! SATISFACTOg1 Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of that simulator to perform correctly following activation of j < a simulated malfunct!~- l This transient n,sulted in the loss of forced flow in the Reactor Coolant System. l~ Decay heat remreal was accomplished by the use of Emergency Feedwater flow to the OTSGs
'and steamir; so the Main Condenser with natural circulation avident.
TN event was initiated by tripping all four Reactor Coolant Pumps on overload. The
,- reactor. tripped and . initiated Reactor Trip Containment Isolation. The Emergency Esedwater System actuated and filled the OTSCs to 50% level on the Operating Range.
The plant transitioned to the natural circulation mode. The current revision of ATOC Procedure 1210-10 was utilized to verify evidence of natural circulation. The RCS pressure rose' due to the Makeup System regaining pressurizer level, compressing the steam bubble. -Because this test was performed from a Beginning of Cycle Initial l Condition, decay heat levels were relatively low, causing the average differential I temperature between hot and cold leg temperatures to be less than the procedurally ) stated 30-50 degrees' Fahrenheit. '
'CS19 i i The transioit was terminated following the verificaties of natural circulation f1'ow and overall plant response utilizing ATOC Procedur.e 1210-10.
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations fra expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6211-ADN-2820.02. ihe results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CElTERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws el nature.
. The simulator satisfied this requirement with no identified deviations.
CRITERIA #2: - The simulator shall not fail to cause an . alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. ( Corrective Actions /Coqmmants: NONE l I Completed By - ff William A. Fraser Date 05/02/95 Approved By Robert L. Egenall . Date 3/9f l 4 e v .
4
.js THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT
' ~'N, Test Identification: Loss of All Feedwater Certification Test
! Transient Test TTS27 ANSI /ANS - 3.5 Reference (s): B.2.2.(2) 3.1.2(10)
Test Date: 05/24/95 Malfunction (s) Tested:* FW15A/B Main Feedwater Pump Trip 1A/1B FW17 Emergency FW Pump Trip (EF-P-1) FW18A/B Emergency FW Pump Trip (EF-P-2A/B)
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated with the reactor tripped, Reactor Coolant System heatup in progress with Prim'ary System pressure controlled by the pressurizer Power rN Operated Relief Valve. Pressurizer level was increasing ( ) and RCS subcooling margin was decreasing.
'J Simulator Run Time: 10 minutes Simulator Evaluation Time: 4 hours Haseline Evaluation Data: Previous Certification Test TTS27 dated 02/27/90.
Right Direction Analysis. Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures Qyerall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of all secondary side heat removal with Reactor Core heat removal via the PORV. During the duration of this test, heat transfer from tf s, the RCS was insufficient for removal of decay heat as expected. I ( ( _ ,/ The event was initiated by failing both Main Feedwater Pumps and preventing the start of all Emergency Feedwater Pumps. The Reactor Protection System tripped the reactor and initiated Reactor Trip Containment Isolation. The Heat Sink Protection System actuated to initiate Emergency Feedwater but the malfunctions prevented the pumps from i L _ _ . . _ _ . . _ _ _ _ _ _ _ ..______________J
,TTS27 ,
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starting. As etcam ganarator invsntory was lost Rasetor Coolent System temperature increased causing system pressure to increase resulting in opening of the pressurizer spray valve then the PORV. At the end of the test the PORV was cycling to limit n Reactor Coolant System pressure and the Reactor Coolant system was approaching [ saturated conditions. N.l) The transient was terminated following OTSG dry-out, with RCS haatup in progress and the PORV controlling RCS pressure. During the conduct of this test simulator dynamic response, annunciator operation, and I automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TM1 Simulator Training, i The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. { Identified deviat i.ons from expected performance have been evaluated for impact and ' corrective acti u in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: ( The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: [m ' The simulator shall not fail to cause an alarm or automatic action if the
\ reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
The simulator satisfied this requirement with no identified deviations-. Corrective Actions / Comments: None Completed By h, 6742M Norman J. Monson Date 06/16/95 Approved By _ _ Robert L. Parnell Date 06/20/95 . I 1 1 O t 1 v
- - _ _ _ _ = THREE MILE ISLAND SIMULATOR MANAGEMTNT CERTIFICATION TEST ABSTRACT [)
\ /
Test Identification: Turbine Trip Certification Test Transient Test TTS35 ANSI /ANS - 3.5 Referencets): 3.1.2(15) Benchmark Test B2.2(6) Test Date: 06/16/95 Malfunction (s) Tested: TC01 Turbine Trip
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-13 50% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: The test was terminated after plant stability was reached at the lower reactor power level following the Main Turbine trip. Simulator Run Time: 10 minutes
,/N Simulator Evaluation Time: 2 hours
\' /
Baseline Evaluation Data: Previous Certification Test TTS35 dated 03/01/90 Right Direction Analysis.
Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this ce' ratification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a turbine tri, ',rith reactor power below the high power trip setpoint. The event was initiated bv activation of the turbine trip malfunction af ter power was
- . reduced to 41% in order to prevent a reactor trip.' The Integrated Control System responded by inisiating a power reduction, since the control system transferred to the Tracking mode with megawatts generated equal to zero when the turbine was tripped.
Reactor power change due to increased temperature was less significant than the initial fg test due to incorporation of a new core model with diffcrent temperature coefficient l 1 values. Reactor Coolant System pressure and temperature decreased significantly as i_/ s reactor power decreased to less than 22%. SASS actuated a mismatch indication for i T-cold transmitters. Reactor Coolant System pressure initially increased resulting i 1 l
I TTS35 .in operation of the Pressurizer spray valve. . g' The transient was terminated when primary and secondary pressures and temperatures were returning to stable conditions. Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test, simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. T1.e TMI representatives have held an NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact .and corrective action in accordance with Functional Fidelit Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA 2: ( t The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely,
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the simulator shall not cause an alarm or automatic action if the reference plant would not cause'an alarm or automatic action. The simulator satisfied this requirement with one identified deviation. Corrective Actions / Comments:
Description:
Incorrect Alarm Actuation Number of Identified Deviations: One Corrective Actions: One discrepancy report. RCS Pressure exceeded G-3-8 setpoint but it didn't alarm. Completed By W W Norman J. Monson Date 06/16/95 Approved By / Q U/ Robert L. Parnell Date 06/16/95
.. Manager, TMI Simulator r
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THREE MILE ISLAND SIMULATOR MANAGEMENT' CERTIFICATION TEST ABSTRACT y
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Test Identification: Main Steam Leak Inside the RB Certification Test Transient Test l TTS42 j
, ANSI /ANS'- 3.5 Reference (s): Benchmark Test B2.2(9) 3.1.2(20)
Test Date: 06/23/95 Malfunction (s) Tested:* MS02A Main Steam Leak in the RB 100% (100% - 6,000,000 LBM/HR)
-
- Refer to Malfunction Cause and Effects Documents .
for options available. Test Initialization: Protected Initial Condition IC-16 100%-Reactor Power Equilibrium Xenon Middle of Cycle'
' Point of Test Termination: . This test was terminated following the reactor
. trip, HPI, and Emergency Feedwater actuations due to the effects. of a main steam line rupture inside containment.
Simulator Run Timca 10 minutes Simulator Evaluation Time: 4 hours ELaseline Evaluation Data: Previous Certification Test TTS42 dated 03/02/90. Right Direction Analysis. RETRAN Data. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Dyerall Test Results: SATISFACTORY _I Results
Description:
In accordance with ANSI /ANS-3.5-1985. this certification test ww conducted to
- l. dersonstrate the ability of the' simulator to perform correctly following activation of i
a simulated malfunction.
'The event was initiated by causing a taaximum size steam rupture inside containment on the OTSG 1A. ' The. resultant steam release caused a ' reactor and turbine trip on
. Overpower with subsequent Reactor Trip Containment Isolation actuation. The containment pressure and temperatures increased causing Engineered safeguards
- actuation, and a f alse Reactor Building fire alarm. The rapid blowdown of OTSG 1A caused _a significant cooldown of the RCS. A-SASS actuation occurred due to the A significant mismatch between the A and B Main steac header pressures. Feedwater flow
.(-
- was isolated to the OTSG 1A on HSPS Low Pressure. The Emergency Feedwater System actuated on high Reactor Building pressure and fed the OTSG 1A. High pressure injection flow recovered the RCS inventory lest by shrinkage and compressed the pressurizer steam bubble, elevating RCS pressure. The spray valve automatically actuated to dampen the pressure increase.
-__x_____-_-__---_ _ _ _ _ _ _ _ _
TTS42 ,The transient was terminated following Engineered Safegnards and Emergency Fesuwater y ' System actuations with elevated Containment pressures and temperatures. s
") This test differed f rom the initial benchmark test in 1990 due to the impact of lowered OTSG levels on the simulator. The 1990 test produced high OTSG level Feedwater isolation which limited the post-trip cooldown. The 1995 test resulted in lower RCS pressures and temperatures, and lower Pressurizer level as a result of continued feed flow. This then caused a longer period of time for conditions to return to more normal post-trip conditions after OTSG 1A isolation.
Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reantor Operator licenses for TMI Unit 1. Identif ud deviations from expected performance have been evaluated for impact and corrective actior. in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws'of nature. The simulator satisfied this requirement with no identified deviations. O (\ j) CRITERIA #2: . The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or autor,atic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: 4 NONE Completed By 8h ' Willinma b Frase_r._ Date 06/27/95 Approved By - Robert L. Parnell Date 06/29/95
-p
( x 1
THREE MILE ISIAND j
..'? ,,
"; SIMUIATOR MANAGEIGNT CERTIFICATION TEST ABSTRACT 1
, Test Idenet fication- Manual Reactor Trip Certifica.t. ion Tent 3 Transient Test
( TTS56 ANSI /ANS w 3. 5 Reference (s): Benchmark Test B.2.2(1) l
.Tast'Datai 01/09/95 h ifunction R Tanted:* None - In tiation of Manual Reactor Trip ;1 I
Iggg Initia11rationi Protected Initial Condition IC-17 1004 Reactor Power .i Equilibrium Xenon ' End of Cycle Point of Tast m inatic,n: This tast was. terminated with the reactor tripped and proceeding to stable Hot Shutdown conditions. , s h astor nua Ti==: '10 minutes h iator Evaluation T&ggi 2 hours j
.I ammelina Evaluation Data': Previour Certification Test TTS56 dated 03/07/90 Right Direction Analysis.
Current Controlled copies of: y Reference Plant Alarm Response Procedures l; Reference Plant Emergency Procedures l Overa'11' Tant Ranulta! SAIll/AM i Ranults Damerintion: In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to I demonstrate the ability of the simulator to perform correctly following activation of 'l
- a manual reactor trip. " '
The event was initiated by dep;essing the Manual Reactor Trip pushbutton on the Main Console and reducing the pressurizer level control setpoint to 100 inches. All control rod drive breakers opened and Reactor Trip Containment Isolation actuated. The Main
- . Turbine tripped on interlock. The Integrated, Control System reduced Feedwater flowe and stabilized OTSG 1evels at Low Level Limits. The Turbine Bypass Valves operated to maintain OTSQ pressure and remove decay heat.
[ The transient was terminated with the reactor tripped and the plant approaching stable l Hot Shutdown conditions. n . !: During the condu-t of this test simulator dynamic response, annunciator operation, and ; automaticLsafety system actuations were evaluated. Test results have been evaluated j against ANSI /ANS-3.5 criteria by GPUN personnel representing TNI Simulator Training. The TMI representatives have held NRC Senior Reactor. Operators licenses for TMI Unit 1 Identified deviations from expected performance have been evaluated for impact and O .'co.- rrective action in accordance with Functional Fidelity Procedure 6211-ADM 2820.02.
, TTS56 -2 The rccitico of checs svaluatiens cro ce follows: .
The results of this test are satisfactory, based upon the following: CRITERIA #1: The ' observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
-The simulator satisfied this a quirement with no identified deviations, CRITERIA #2:
The simulator shall not fail to cause an alarm 'or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant
.would not cause an alarm or automatic accion.
The simulator satisfied this requirement with no identified deviations. Corrective Actions /ce- nts: NONE Completed By #
. - I', -
W iliam A. FrA33I ,, Date 4 h"-- Approved My a C- , Robert L. Parnell .Date AI!4a"~ ES 9 9 ('
THREE MILE ISLAND - d7g ' SIMUIATOR MANAGEMENT
' CERTIFICATION TEST ABSTRACT Tant Identification:
Simultaneous Closure of All MSIVs Certification Test Transient Test TTS57 ANSI /ANS - 3. 5 Rafaranca(s): Benchmark Test 8.2.2(3) Tant Data: 01/04/95 4 Malfimetion(s) Tasted!* MS08A/B/C/D Main Steam Isolation Valve Closure MS-V-1A/15/1C/1D Tant initialization: Protected Initial Condition IC-16
, 1004 Reactor Power Equilibrium Xenon Middle of Cycle Point of Tant Termination: This test was terminated with the reactor tripped and Emergency Feedwater removing decay heat via the Atmospheric Dump Valves.-
S t =>1 ator aim Tf == : 20 minutes
- St-il ator Evaluation' Tf ==:
4 hours , 4 'ammelina Evaluation Data! Previous Certification Test'TTS57 dated l; 03/13/90. Right Direction Analysis. Current Controlled copies of: Reference Plant Alarm Response Procedures Reference Plant P & ID Prints l Overall Tant Ranults: SATISFACTORY tamulta Damerintion! In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. l i l This transient resulted in a reactor trip on high pressure and Emergency ) Feedwater System actuatiop due to a loss of Main Feedwater. The event was' initiated by simultaneous closure of all four Main Steam Isolation Valves. Ihis caused OTSC pressures to increase and Main Steam Meader Pres'sure to decrease. The Main Turbine transferred to manual control
. automatically upon receipt of a sustained 40 PSIG Header Pressure Error Signal. .As OTSG pressures increased the Turbine Bypass, Atmospheric Dump and Main Steam Safety. Valves operated to limit the pressure excursion. Reactor Coolant System pressure and temperatures increased due to th' transient. The reactor tripped on high pressure and actuated Reactor Trip containment Isolation. When all Main Steam Isolation Valves had closed the Main Feedwater Pumps slowed, reducing Feedwater flow to the OTSGs. The Emergency Feedvater System actuated to feed the OTSGs at low byel Limits. Gland Sealing steam was lost when the OTSG'15 MSIV's closed, causing a loss of vacuum to the Main Turbine and Main Feedwater Pumps. Decay heat was then removed by automatic action of the Atmospheric Dump Valves, since the Turbine Bypass Valves closed
TTS57 M d ue to tha low Candenscr Vecuum csndition (ccuand by th3' loso of Clend Saal Stsas). The transient was terminated with the reactor tripped and decay heat being removed by the Emergency Feedwater System via the Atmospheric Dump Valves. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by CPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operators licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in I direction to those expected from the actual reference plant and do not j violate the physical laws of nature. l I All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. . CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or autcmatic action, and conversely, the simulator shall not cause an alarm or. automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. Corrective Actions /C--- -- nts : NONE Completed By _ M 4, , William A. Fr a Dato 05/21/95 Approved By / Robert L. Parnell Dats I t
THREE MILE ISLAND SIMULATOR MANAGEMENT
-g c) CERTIFICATION TEST ABSTRACT !
g'~ ~N, Test Identification: Loss of One RC Pump Certification Test i,
) Transient Test v TTS58 i
ANSI /ANS - 3.5 Reference (s): 3.1.2(4) Benchmark Test B2.2(5) { l { l Test Date: 05/24/95 Malfunction (s) Tested: RC35A RC Pump Trip Test Initialization: Protected Initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: The test was te rminated following reactor trip with Primary and Secondary Plant pressures and temperatures approaching stability. I ! E_/s Simulator Run Time: 10 minutes Simulator Evaluation Time: 3 hours Baseline Evaluation Data: Previous Certification Test TTS58 dated 03/15/90 Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures 0terall Test Results: SATISFACTORY Results
Description:
i l In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip. Reactor Trip initiated a Main Turbine trip p,_ , and Reactor Ttip Containment Isolation. l \
) Th' event was initiated by failing Reactor Coolant Pump 1A. The loss af the Reactor Coolant Pump a~t full power resulted in the Reactor Protection System w-_--__--___._-___ _ _
. TTS58 l
,a i
initiating Rasetor Trip dus to ths reduction in Rsactor Coolant System flow. The Reactor Trip resulted in a Main Turbine trip. Reactor Trip Containment Isolation closed specified Containment Isolation Valves. Reactor Coolant System pressure and temperature , decreased to post trip values and were maintained. SASS actuated a i mismatch indication for T-cold and T-hot transmitters. The transient was terminated when Primary and Secondary pressure and temperature were approaching post trip stable conditions. l During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuacions were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator certification for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, I 6221-ADM-2820.02. The results of these evaluations are as follows: ) The resultis of this test are satisfactory, based upon the following: l CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to I those expected from the actual reference plant and do not violate the physical
. laws of nature. {
The simulator satisfied this requirement with no identified deviations. I CRITERIA 2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. 4 The simulator satisfied this requirement with one identified deviation. i Corrective Actions / Comments:
Description:
Incorrect System Response Numb *er of Identified Deviations: One Corrective Actions: Modification - Feedwater Upgrade R92-043 1 l Completed By _ Daryl Wilt Date 06/22/95 l Approved By Robert L. Parnell Date 06/22/95 Supervisor, Simulator Training V l
1 T THREE MILE ISIAND y. SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT
)
Test Identification: Maximum Rate Power Ramp Certification Test Transient Test TTS59 ANSI /ANS - 3.5 Rafarance(sii Benchmark Test 5.2.2(7) 5.4.1(2) . Tant Date: .01/09/95 Malfunction (s) Tasted
- None - Manual Plant Maneuvering f
Test Initialization: Protected Initial-Condition IC-17 1004 Reactor Power Equilibrium Xenon ! End of Cycle Point of_ Test Termination: This test was terminated with the plant stable at 1006 power. Simil ator nun Ti = : 15 minutes St=ilator Evaluation Ti=: 1 hour- . t a==alina Evaluation Data Previous Certification Test TTS59 dated 3/12/90 Right Direction Analysis. ' Current Controlled Copies of: Reference Plant Operating Procedures
,0verall Test Results: SATISFACTORY L- Rasults Damerintion!
In accordance with ANSI /ANS-3.5-1985 this certification- test was conducted to. demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. l The event was initiated from steady-state full power with the Integrated Control System L in the integrated mode. The Unit load Demand controller was set to approximately 75' percent load. The plant reduced power at the maximum rate of ten percent per minute and. stabilized - at 75 percent power. The plant was allowed to stabilize for ! Approximately 3.5 minutes. At the end of the stabilization' period, power was ramped . back up at the maximum rate of ten percent .per minute. When the plant reached 90 L
. percent . load the Integrated Control System automatically reduced the rate of power increase to 5 percent per minute. During the period of time at reduced power the Xenon l,
concentration began to increase. The plant was. allowed to stabilize at full power. The transient was terminated with'the plant at full power. l Dynamic response was. compared to the dynamic response of the initial certification test. During the . conduct of this test ' simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have .
'been evaluated . against. ANSI /ANS 3.5 criteria by GPUN personnel representing TMI ; Simulator Training. The TMI representatives have held NRC Senior Reactor Operator \
licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity
- Procedure 6211-ADM-2820.02. The results of these evaluations are as follows:
TTS59 Th3 results of this tsse cro cctiefcetary, based uprn th2 folicwing: CRTTERTA #1: The observable changes in simulator parameters shall correspond ino dire thoseofexpected laws nature. from the actual reference plant and do not violatee th physical All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRTTERTA #2: The' simulator shall not fall to cause an altra or automatic action if the the simulator shall not cause an alarm or automatic ce plant a would not cause an alarm or automatic action. All CRITERIA 2 evaluators cotv,arred that during this test, the simulator satisfied this requirement with no ids.stified deviations. Corrective Actions /Cc-- nta: NONE Completed By ' William A. Fraser __ Date 05/21/95 Approved By h_ Robert L. Parnell ._ Date ff
+
l s u t l l 3 ( l t i ( ,
THREE MILE ISLAND -
?- ,b, SIMULATOR MANAGEMENT L
I CERTIFICATION TEST ABSTRACT Test Identification: Loss of Offsite Power with Large Break LOCA Certification Test Transient Test ( TTS60 ANSI /ANS - 3.5 Referencefsf1 3.1.2(1b) Benchmark Test B.2.2(8) l Test'Date: 02/09/95 4 Malfunction (s)-Testedt* ED01 Station Blackout TH04A RCS LOCA at Hot Leg Nozzle 100% severity Test' Initialization!. Protected Initial Condition IC-16 1004. Reactor Power Equilibrium Xenon ^ Middle of Cycle Point of Test Termination: This test wa's terminated with the reactor tripped, Core Flood tanks emptied, and reactor core cooling being provided by High Pressure and Low Pressure Injection. Reactor Luilding pressure and temperature were decreasing from peak conditions. Simulator Run T4==* 10 minutes ' Simulator Evaluation'Timet 10 hours
-Baseline Evaluation Datar Previous Certification Test TTS60 dated 03/08/90.
Right Direction Analysis. Simulator Malfunction Cause and Effects Document.
}- current controlled Copies of:
l / Reference Plant Alarm Response Procedures. I < Reference Plant Electrical One-Line Diagrams
- . Reference Plant Emergency Procedures l Overall Test Results
- SATISFACTORY Results Description!
.In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation
.of a simulated malfunction.
o This transient resulted in the loss ot ,ffsite power coincident with a'large break L loss of coolant accident. The reactor was-tripped due to a loss of power to the. l Reactor Coolant Pumps and the, Control Rod Drive System. Safety syrtem power was
- l. restored by the Emergency Diesel Generators. A large break in the Reactor coolant
! . System caused rapid depressurization of the coolant and a rapid increase"in Reactor ; _ Building. pressure. The Engineered Safeguards System actuatodi initiating safety '
; injection and Reactor Building Isolation and cooling.
The' event was initiated by fault in'the 230 KV Substation coincident with a large
~
size rupture of the Reactor Coolant-System Not Leg-piping. The Reactor Prot 3ction
' System initiated a Reactor trip and Main Turbine trip. Reactor Trip Containment Isolation closed specified Containment Isolation valves. The Heat Sink Protection System detected a lo(s of Reactor Coolant Pumps and initiated Emergency Feedwater to the OTSGa r co,ntrolling level at 50% on the operating range. The Emergency Diesel
- Generators. started and the output breakers closed to re-energize their 4.16 KV Buses. SASS-actuated a mismatch indication for RCS Hot Leg Temperature j Transmitters. !
l i
'\ I l
=
I TTS60 L, Tha loss of inventory and pressure from the Reactor Coolant System resulted in discharge of the Core Flood Tanks and actuation of the Engineered Safeguards Actuation System. High Pressure Injection, Low Pressure Injection, Reactor Building
#
- Spray,.and Reactor Building Isolation and. Cooling were initiated due to the pressure loss from the Reactor Coolant System and the high pressure detected in the Reactor
) Building, j
The transient was terminated when Reactor Building pressure and temperature were decreasing and reactor core cooling was verified. - Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were evaluated. Test results have been-evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. .The TMI representatives have held NRC Senior Reactor Operators licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity {' Procedure 6211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA 51: The observabi'e changes in simulator parameters shall correspond in direction to th'ose expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA 92: The simulator shall not fail to cause an' alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversuly, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with three identified deviation. Corrective Actions / Comments:
Description:
Incorrect Annunciator Alarm Operation (AA-2-1 SBO DIESEL TROUBLE alarm failed to actuate) Na=her of Identified Deviati2R11 One j Corrective Actions: One Discrepancy Report correction
Description:
Incorrect Computer Alarm Operation
-(T-hot alarm and High Power Trip SOE points)
N"=ker of Identified Deviations: Two f corrective Actions: Two Discrepancy Report correction IL I l I t l
,< x .
f l Completed By .1 Charles E. Husted Date_06/30/95 q/J' . ,/ Approved By (3Le Robert L. Parnell Date 06/30/95 i
'DIREE MIIE ISIAND SDEJIA'KR MANAGEMENT CERTIFICATICH TEST AB9'IRACF Test Identificatim: Simlator Stability Certification Test O Steady State Performance Test SSP 01 ANSI /ANS - 3.5 naference(s): 5.4.1(2) 4.1 Benchmark B2.1 Test Date: 05/19/95 Test Initiali*atim: Protected Initial candition IC-17 100% Reactor Power F
?ilihium Xerum End of cycle Point of Test Tarninatim: 100% Reactor Pouar.
579 Degrees Tave. i - ICS in Full Automatic sf =ila+rn- Run Tim = I hour l
--1 ins Evaluation Data: Plus/minus 2%.
Overall Test Results: SATISFACIGN i Results
Description:
In amnrdance with ANSI /ANS-3.5-1985. this certification test was carvhv+='i to demonstrate the ability of the sinulator to operata at stable full power, in automatic ocz1 trol for a 66 period. During carukact of this test similator dynamic operation was evaluated against ANSI /ANS 3.5 critaria. 'Ihis evaluation was em1 ducted by GRM perscrmal representirry 'IMI Sinulator Training. 'Ihe evaluators have held an NRC Senior Reactor Operator licanoe . for 'IMI Unit 1. 'Ihm results of these evaluations are as follows:
'Ihm results of this test are satisfactory, based upon the following:
GTIERIA #1:
'Ihe ' simulator ocoputed values for steady str%, full power operation with the referenced plant ocntrol system configuratica shall be stable and not vary more than plue/minus 2% of the initial values over a 60-minute period.
'Ihe similator this requirunent with no deviations. -
j ocspleted By /E Charles E. Hustad Data 06/20/95 I
]
Approved By / Robert L. Parnall. Data 06/20/95 l ( ) l
THREE MIIE ISIAND 4, SIMJIAttR MANAGEMENT CERTIFICATION TEST ABS 1RACT Test Identification: Siza11ator Accuracy certificatica Test Steady State Performance Test SSP 02 ANSI /ANS - 3.5 Reference (s): 4.1(2) 4.1(3) 4.1(4) 5.4.1(2) Band 1 mark B2.1 Test Date: 05/26/95 Test Initializatirm: Protected Initial Candition IC-17 100% Reactor Power , Fmi14heigg Xgngn ! End of Cycle l Point of Test Termination: After maamitionant of similator ocmputed values and principal mass / energy balance determinations were made at 1004, 80%, and 60% rated power. Simslator Run Maal. 3 hours Baseline Evaluation Data: Previous Certification Test SSP 02 dated 03/31/94. , Reference Plant operating Data Reference Plant Operating Prdives l Overall Test Results: SATISTACKRY Results
Description:
. I I
In amrwdance with ANSI /ANS-3.5-1985 this certificatics) test is conducted to demonstrate the ability to perform within the accuracy requirement of plus/minus 2% for l critical parameters and plus/minus 10% for noncritical parameters pertinent to plant j operation. Principal mass and energy balances shall be satisfied. f During conduct of this test similator dynamic operation was evaluated against ANSI /ANS 3.5 criteria. This evaluation was ocnducted by GPW personnel i.s ---_Idrg 1MI i similata Training. She evaluators have held an NRC Senior Reactor Operator license for 1MI Unit 1. The results of these evaluations ars as follows: The results of this test are satisfactory, based upon the following: GITERIA #1: The sinalator ,ocuputed valtas of critical parameters shall agree within plus/minus' 2% of the reference plant paammstars and shall not detract from training. She paramstats displayed cn control panels may have the instrument error added to the ocmputed values. The siallator supported this requirement with no deviations. GITERIA #2: 1hs calculated values of noncritical plant parameters pertinent to plant operation, that are included on the similator ocmtrol panels, shall agree i
1 SSP 02 _2
-,.s~.
within plus/ininus 10% cf the reference plant N and shall not h t a trainim. su parameters displayed on control penals may have the instnment error added to the empted values. ( i h sMatar supported this requirement with no deviations. GITERIA #3: ! ipal mass and ener w balances shall be satisfied at three different p f
- s tor - esa - with n. su_.
Corrective Actions /Ctaments: NCHE l Ca pleted By Charles E. Hbated Date 06/TW
- , Tv# - _ , , _ _ , ,
L I-l I l I i W e D 0 8 4
THREE MILE ISIAND o' SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test' Identification;, Real Time Test RTT01
' ANSI /ANS - 3.5 Reference (s): Appendix A3(1)
Test Dates . , 06/27/95 & 06/28/95
. Test Initialization: Protected Initial Condition IC-32 1004 Reactor Power Equilibrium Xenon End of Cycle Point of Test' Termination This test was terminated following collection of all required data.
' Simulator Run 'Ilme 2 hours Sinn11ator Evaluation Tima: 10.5 hours Rameline Evaluation Data! . Valve stroke times per controlled copies of applicable Reference-Plant surveillance procedures. Known values for event timers.
Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to
- demonstrate the ability of the simulator co&puters to run in real time.
,During the conduct.of this test simulator dynamic performance was~ evaluated against ANSI /ANS 3.5-1985 criteria. Thame evaluatim.c were. conducted by CPUN
. personnel representing TMI Simulator Training who hold or have held NRC Senior
- Reactor Operator licenses for TMI Unit 1.
This test verified the capability of.the simulator computers to run in real time by measuring known values for selected valve stroke times and event- ' timers. The' correct valve stroke times were drawn from controlled copies of Reference Plant surveillance procedures. LKnown values for event timers. wore
.. drawn from Reference Plant system design data. A calibrated stopwatch was utilized to time each item, g' Additionally, the computer system processing capabilities were measured to serve as an annual check on each processor's performance. The evaluation of this data serves as a computer system management tool. .
f. t The results of this test are satisfactory with all valve stroke times and event timers corresponding to the reference data from the Reference Plant with
. minor differences (less than plus one second) due to the use of a manual stopwatch for timing. Computer processing results (spare times) are being analyzed to evaluate the effectiveness of the configuration changes made this past year.
l' a-______ -. __
1 Corrective Actions / Comments: e Thero cra na correctiva ceticna rsquired cs a result of thin tact. Measurement of processor performance was performed on a test disk pack incorporating the Digital Turbine Control system and the Cycle Ten Core [ \ Upgrades. This was done to ensure that adequate processor performance would be (,_,) maintained after these significant model upgrades. Due to the inclusion of the Cycle Ten core, a test Initial Condition (32) was utilized which was drawn , from IC-17. l l Completed By # William A. FraseL Date: 06/29/95 Approved By Robert L.Parnell Date: 06/29/95 Lead Instructor, Simulator Training s-O i-j
'f,.- THREE M LE ISLAND T= .
SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT' Test identification: OTSG Tube Leak Certification Test Transient Test ( . TTS01
. ANSl/ANS - 3.5 Reference (s): 3.1.2.(1.a)
Test Date: 01/09/95 Malfunction (s) Tested. TH17B OTSG Tube Rupture Low 0.1% (100% = 50 Tubes)
- Refer to Malfunction Cause and Effects Documents for options-available.
. Test initializabon: Protected initial Condition IC-36
, 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the plant stable at 100% power, RM-G-27 steam line monitor in alert, RM-A-5 and RM-A-15 off-gas monitors increasing. Makeup flow to the Reactor Coolant System increased in response to the tube leak.
L
+ Simulator Run Time. 20 minutes Simulator Evaluation Time: 1 hour
- Baseline Evaluation Data. Previous Certification Test TTS01 dated 04/29/90.
Right Direction Analysis Simulator Malfunction Cause and Effects Document - TMI-l Tube Leak incident March 6,1990
- OverallTest Results. SATISFACTORY Results Description.
In accordance with ANSl/ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the sirnulator to perform correctly following activation of a simulated malfunction. , , This transient resulted in activity released via the plant off-gas stack due to a slight tube leak in OTSG 18. During lc this test the plant remained at 100% power with increased makeup flow replacing the coolant released through the tube leak. t
)
l: l i V
,, T,TS01 . l The event was initiated by causing a small tube leak in the lower portion of OTSG 18. Condense, off-gas monitors showed activity increasing. Reactor Coolant System makeup flow increased and maintained pressurizer level at the normal value. The plant remained at 100% power for the duration of this test. RM-G-27 went into
,8% Alert. )
)
(
\ )
The Radiatior' Wait 4ng System response during this test differed from the 1990 test. The RCS activity in the initial conditioc sas reduced from 10 microcuries to 1.6 microcuries. This caused RM-G-27, RM-A-5 and RM-A-15 response to be less severe with the same size tube leak. Also, improvements in radiation system modeling j produced more realistic response, j The transient was terminated following verification of proper makeup flow, pressurizer level and Radiation Monitoring System Response. 1 During the conduct of this test simulatcr dynamic response, annunciator operation, and automatic safety system j actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives hold or have held an NRC Senior Reactor { Operator licenses for TMI Unit 1. Identified deviations frcm expected performance have been evaluated for impact and corrective act;on in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results of these evaluations are as follows: a The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direct}on to those expected from the actual reference plant and do not violate the physical laws of nature, i The simulator satisfied this requirement with no identified deviations. I CRITERIA #2: ' l h The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. l The simulator satisfied this requirement with no identified deviations. , Corrective Actions / Comments: NONE Completed By [ I
. WY _ W. A. Fraser Date [ 7!4[
/ /
Approved By Lead instructor, Simulator Training R. L. Pamell Date MJb[
' /
l G' i
THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: OTSG Tube Rupture Certification Test
.A Transient Test
'( TTS02 .
ANSI /ANS - 3.5 Reference (s): 3.1.2.(1.a) Test D81E 03/11/95 Malfundion(s) Tested: TH17A OTSG Tube Rupture Low 3 2% (100% = 50 Tubes) J l
*Referto Malfunction Cause and Effects Documents for options i available.
Test initialization; Protected 8nitial Condition IC-17 100% Reactor Power Equi librium Xenon End of Cycle Point of Test Termination: This test was terminated fc!!ow:ng verification of initial responses, Just prior to reactor trip and Engineered Scfoguards actuation due to the excessive OTSG tube leakage. Condenser off-gas activity monitors were in alarm. Simulator Run Time: 8.5 minutes f-w Simulator Evaluation Time: 1 hour (
% Baseline Evaluation Data: Previous Certification Test TTS02 dated 04/29/90.
Right Direction Analysis Simulator Mal;dnction Cause and Effects Document Overall Test Results: SATISFACTORY Results
Description:
)
in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulatorto perform corredly following activation of a simulated malfunction. This transient resulted in activity released via the plant off-gas stack due to a tube rupture in ths OTSG 1 A during this test. i I i f~~N r i
+TTS02 2-The event was initiated by causing a tube rupture in the lower portion of OTSG 1 A. Condenser off-gas monitors showed activity increasing and went into high alarm. The RM-A-5 MAP 5 sampler started. Reactor Coolant Makeup flow increased to its maximum value. Pressurizer level and Reactor Coolant System pressure decreased. Feedwater flow to OTSG 1 A decreased slightly due to flashing of the coolant leaking into the OTSG causing a slight steam pressure increase.
The Radiation Monitoring System response during this test differed from the 1990 test. The RCS activity in the initial condition was reduced from 10 microcuries to 1.0 microcuries. This caused RM-G-27, RM-A-5 and RM A-15 response to be less severe with the same size tube leak. Also, improvements in Radiation system modeling produced more realistic response. The test was terminated following verification of initial responses, just prior to reactor trip and Engineered { Safeguards actuation due to low RCS pressure. l During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held an NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for - ( impact and corrective action in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results of these evaluations are as follows: l 1 The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: . The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: l
'NONE Completed By ' Ob tt/Mh[
s _ W. A. Fraser Date / r
/ /
Approved By R. L. Pamell Date [ f!I L'ead Instructor, Simulator Training / / l l l r t . 1
n e THREE MILE ISLAND SIMULATOR MANAGEMENT .
. CERTIFICATION TEST ABSTRACT .
( Test identification. RCS Leak inside Containment Certification Test Transient Test t TTS03 6NSI/ANS - 3.5 Reference (s): - 3.1.2.(1.b) [ 3.1.2,(1.c) Test Date: 02/09/95 Malfunction (s) Tested. TH02A RCS Leak at top of Hot Leg 50% severity ramped over 240 seconds
- Refer to Malfunction Cause and Effects Documents for options available.
l Test initialization. Protected initial Condition IC-16 . L
'100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated with the reactor operating at rated p power and Reactor Building pressure slightly below the reactor -
trip setpoint. Reactor Building isolation and Cooling actuation I had occurred. I Simulator Run Time: 12 minutes l
- p. Simulator Evaluation Time:
t ( 2 hours I Basehne Evaluation Data. Previous Certification Test TTS03 dated 01/26/90. Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: , Reference Plant Alarm Response Procedures !- Reference Plant Electrical One-Line Diagrams OverallTest Results. SATISFACTORY Results Descnotion. l In accordance with ANSI /ANS-3.51985 this certification test was conducted to demonstrate the ability of the . ( ~simulator to perform correctly following activation of a simulated malfunction. o
- This transient resulted in the leakage of Reactor Coolant System high energy water into the Containment -
E
. Building. Building pressure and temperature had increased to the point of actuation of Engineered Safeguards Reactor Building isolation and Cooling. The reactor had not been automatically tripped at the point of test
. termination. Airbome radioactivity levelin the containment was increasing. The event was initiated by producing a small leak in the Reactor Coolant System Hot Leg that was not large enough to result in Reactor Protection
' System actuation of a reactor trip. The high energy leakage caused containment temperature and pressure to increase to the aduation setpoint of Reactor Building Isolation and Cooling which started the High Pressure injection System.
A.* TTS03 The transient was terminated following Reactor Building Isolation and Cooling actuation on high containment pressure. Reactor Coolant System inventory was being maintained and containment temperature and pressure p were being reduced. t I
\v/ During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system ~
actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held an NRC Senior Reactor Operatorlicenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical 19ws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shat! not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations.
- Corrective Actions / Comments: / s i
k RJ
/ NONE I Completed By / 1M Charles Husted Date 4 2J 9[
Approved By ~ Lea'd instructor, Simulator Training R. L. Pamell Date 4!Z3 9I l t (~ N_)i {
7
+
THREE MILE ISLAND p CERTIFICATION TEST ABSTRACT SIMULATOR MANAGEMENT ,
' Test Identification: RCS Leak Outside Containment Certification Test Transient Test TTS04
)
' ANSI /ANS 3.5 Reference (s): 3.1.2(1b)
Test Date: 03/11/95' Malfunction (s) Tested:* MU13 Letdown Line Relief Valve Fails (MU-V-105) 100% (100% = Full Open)
- Refer to Malfunction Cause and Effects Documents. for ' options
. available.
Test Initialization: Protected Initial Condition IC-15 100% Reactor Power
- Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated with Miscellaneous Waste Storage Tank level- increasing and Makeup Tank 4 decreasing with constant Makeup flow to the RCS. ,V l
Simulator Run Time: 25 minutes Simulator Evaluation Time: 30 minutes Baseline Evaluation Data: Previous Certification Test TTS04 dated 03/21/90. Right. Direction Analysis. i Simulator Malfunction Cause and Effects' Document
'Overall~ Test Results: SATISFACTORY
, .Results
Description:
i In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability .of the simulator to perform correctly following activation of a simulated malfunction. This test resulted in reactor coolant leakage from the !.etdown System outside the Containment Building to the Miscellaneous Waste Storage Tank. The event was. initiated by failing open MU-V-105 - the Letdown Line Relief Valve. This caused reactor coolant to leak to the Miscellaneous Waste Storage Tank. Level in the MWST increased. The level in the Makeup Tank decreased as a result of the leakage. Pressurizer level was maintained at setpoint by the Automatic
./ ^ \ Level Control System.
(,,/'
r. Vi .. TTS04 p V This test was terminated following verification of proper water transfer, Makeup Tank level response, and automatic control of. pressurizer level. During the conduct of this test simulator dynamic response annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator
' licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action 'in accordance with Functional
- Fidelity Procedure, 6211-ADM-2820.02. The results of these evaluations are as follows:
- The results of this test are satisfactory, based upon the following:.
CRITERIA #1: The observable changes in simulator parameters shall corres)ond in direction to those expected from the actual reference plant anc do not violate the physical laws of nature. . ALL CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement wit no identified deviations. CRITERIA #2: p\
\
V The simulator shall not fail to cause an alarm or automatic action if the
, reference conve.rsely,the plant simulator'shall would havenot caused causethe alarm or an alarm or automatic automaticaction action, if and
- the reference plant would not cause an alarm or automatic action. '
ALL CRITERIA 2 evaluators concurred that during this test. the simulator satisfied this requirement with no identified deviations. . Corrective Actions / Comments: , The results of this test were satisfactory with no deviations from expected response. There is not corrective action as a result of this test.
-) -Completed B - A, # - . William A. Fraser Date 5/21/95 -Approved By '
Leid Instructor, Simulator Training 2 e AAvr tLo Date h/W n >-k v
]
___., ___,_.____.u,_ _ _ - - . - - - - - - - - - - - - - - - - - - -
I
~ k-THREE MILE ISLAND
, SIMULATOR MANAGEMENT .
CERTIFICATION TEST ABSTRACT Test identification: Large Break LOCA Certification Test Transient Test TTS05 ANSI /ANS - 3.5 Reference (sh 3.1.2.(1.b) 3.1.2.(1.c) Test Date. 03/11/95-Malfundion(s) Tested. TH06 RCS LOCA at A RCP Discharge
' 50% (100% = Full Area)
Ramp-300 seconds
- Refer to Malfunction Cause and Effects Documents for options
, available. l Test Initiahzation. Protected initial Condition IC-17 100% Reactor Power l Equilibrium Xenon l , End of Cycle - Point of Test Termination: This test was terminated following reactor trip, High Pressure and Low Pressure injection actuated, Core Flood Tanks discharged, and Reactor Building Spray actuated. Simulator Run Time: 12 mint [tes Simulator Evaluation Time: 8 hours , Baseline Evaluation Data. Previous Certification Test TTS05 dated 02/01/90. j Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of:
. Reference Plant Alarm Response Procedures Reference Plant Operating Procedures Overall Test Results. SATISFACTORY Results Description.
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. l-This transient resuhed in a significant loss of RCS coolant. The reactor tripped on low RCS pressure and the Engineered Safeguards Actuation Systems actuated at the appropriate setpoints. The reactor core was being cooled by a combination of high and low pressure injection flow. The Core Flood Tanks discharged to reflood the core. Reactor Building pressure increased to 30 PSIG. Reactor Building Spray flow reduced both pressure and temperature in the Reactor Building.
f ->~ 4 l TTS05 2-1 Reactor Trip Containment Isolation, Engineered Safeguards Actuation Systems (including Containment isolation), and the Emergency Feedwater System actuated on high containment pressure.
, Reactor coolant pumps operated throughout this transient which resulted in the pumping of saturated fluid through s / the RCS as w' ell as out the break. The results of this were evidenced by coolant pump cavitation and heat transfer from the Steam Generators. The OTSGs depressurized, causing isolation of Main Feedwater. With the loss of OTSG pressure, main steam was lost resulting in decreasing speed and vacuum on Main Feedpumps as well as the Main Turbine.
The transient was terminated with the RCS at containment pressure and core cooling provided by high and low pressure injection. Engineered Safeguards component actuations were verified using the current revision of reference plant operating procedure OP 1105 Safeguards Actuation System. During the conduct of this test simulator dynamic respatse, annunciator operation. and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN persont.el representing TMI Simulato; Training. The TMi representatives hold or have held an NRC Senior Reactor Operator licenses for TMl Unit 1. Identified deviations from expected performance have been evaluater! for impact and corrective action in accordance.with Functional Fidelity Procedure,6211-ADM-2820.02. The results ;
- of these evaluations are as follows
! ) i The results of this test are satisfactory, based upon the following: i CRITERIA #1: l The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. j The simulator satisfied'this requirement with no identified deviations, j CRITERIA #2. l The simulator shall not fail to cause an alarm or automatic action if the reference plant would have ] caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation.
)
Corrective Actions / Comments:
Description:
Incorrect Replica Plant Process Computer Alarms Number of identified Deviations: One Corrective Actions: Corrected immediately Completed By /m William A Fraser Date d-23-9y~ I Approved By ( p _ R. L. Pamell Date 4-7 3 M Lead Iristructor, SinTuYaT5r Training b) b 1 l
_j ' h THREE MILE ISLAND l .6" SIMULATOR MANAGEMENT j. CERTIFICATION TEST ABSTRACT Test identification. Small Break LOCA Certification Test
- Transient Test TTS06 l(
ANSI /ANS - 3.5 Reference (s): 3.1.2.(1.b) l 3.1.2.(1.c) i i Test Date. 03/11/95
.]
Malfunction (s) Tested: TH04A RCS LOCA: Hot Leg Nozzle I 1% (100% = Full Area) )
" Refer to Malfunction Cause and Effeds Documents for options -
available. { 1 l Test Indiahzation. Protected initial Condition IC-16
]
100% Reactor Power [ Equilibrium Xenon
)
l Beginning of Cyde , l-Point of Test Termination: This test was terminated with the reactor tripped, hi0h pressure injection flow initiated, increasire RCS void fraction and steam generators maintaining low level limits. !: Simulator Run Time: 12 minutes 1 - i 6 hours ! r) Simulator Evaluation Time. [fV Baseline Evaluation Data: Previous Certification Test TTS06 dated 02/01/90. Right Direction Analysis l Simulator Malfunction Cause and Effects Document ] (' Current Controlled Copies of: Reference Plant Alarm Response Procedures Overalf Test Results: SATISFACTORY I Results Description. In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the
' simulator to perform correctly following activation of a simulated malfunction.
This transient resulted in the loss of Reactor Coolant to the Reactor Building. High ' pressure injection flow - provided the heat removal mechanism post-trip. Cooldown proceeded along the saturation curve.
j . TTjgo6 ~ The event was initiated by insertion of 1% of full area leak at Loop 1 A hot leg nozzle. The Reactor Protection System tripped the reactor and initiated Reactor Trip Containment isolation. The Engineered Safeguards Actuation System activated to provide high pressure injection flow for core cooling and recovery of lost inventory. I The Heat Sink Protection System actuated to initiate Emergency Feedwater on high containment pressure but did , not feed the tieam generators as Reactor Coolant Pumps and Main Feedwater remained in operation. Cooldown i of the RCS continued along the saturated curve with steam pressure tracking cold leg temperature. Reactor Coolant Pumps showed evidence of cavitation and the pumping of saturated fluid. Containment pressure and i temperature increased in response to the LOCA. The transient was terminated with the RCS in a saturated condition, full h gh pressure injection flow, elevated containment pressure and temperature and steam generators maintaining low level limits. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives. hold or have held an NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference pla'nt and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2. The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE Completed By 1 /F M William A Fraser Date Approved By R. L. Pamell Date 4 I Lead Instructor, Simulator Training ' i
.. THREE MILE ISLAND
,- SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: RCS PORV Failure Certification Test O Transient Test TTS08 ANSI /ANS - 3.5 Reference (s): 3.1.2.(1.d)
Test Date: 3/11/95 Malfunction (s) Tested: TH08 PORV Leakage 10% (100% = 100,000 LBM/HR)
*Referto Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor tripped RCS pressure being controlled by the Pressurizer Code Safety Valves. High pressure injection flow was at the maximum value and nad filled the RCS to a water solid condition. Reactor Building pressure was increasing due to the bursting of the Drain Tank Rupture Disc. O Simulator Run Time: 53 minutes Simulator Evaluation Time: 6 hours gaggne Evaluation Data: Previous Certification Test TTS08 dated 01/31/90. Right Deection Analysis Simulator Malfunction Cause and Effects Document . Current Controlled Copies of: Reference Plant Alarm Response Procedures ,
. l
- Overall Test Results: SATISFACTORY .
I Results
Description:
i in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the l simulatorto perform correctly following activation of a simulated malfunction. l l
I l TTSG % The event was initiated by failing the Pilot-Operated Relief Valve (PORV) on the pressurizer to ten percent open. RCS pressure decreased which initiated a reactor trip on low RCS pressure. This initiated Reactor Trip Containment isolation. As pressure reduced further an Engineered Safeguards actuation occurred due to low IRCS pressure. Injection of high pressure water filled the system to a water-solid condition which caused pressure to increase to the point of actuatin0 the Pressurizer Code Safety Valves repeatedly. RCS subcooling margin did r not decrease below 28 degrees Fahrenheit and forced flow was maintained. The Reactor Cociant Drain Tant Rupture Disc ruptured, releasing coolant and activity to the Containment Building and causing containment pressure to increase. - The response shown by this test differs from the same test run in January of 1990. Improvements in pressurizer modelir$ caused a longer time period to ocwr prior to Reactor trip. Reactor Coolant Drain Tank modelng i improvements produced a more linear response in pressure and ievel indications. The test was terminated with water being discharged from the Pressurizer Safety and Relief Valves to the Containment Building. During the conduct of this test simulator dynansic response, annunciator operation, and automatic safety system j actuatior's were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN perscnnel repr9senting TMI Simulator Training. The TMI representatives hold or have held an NRC Senior Resctor l Operatorlicense for TMl Unit 1. Identified deviMions from expected performance have been evaluated forimpact i and corrective action in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results of these-evaluations are as follows: t The results of this test are satisfactory, based upon the following: CRITERIA #1 I The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature . i g The simulator satisfied this requirement with no identified deviations. CRITERIA #2 i l The simulator shall not fail to cause an alarm or automatic action if the reference plant would have I I caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. )
- The simulator satisfied this requirement with no identified deviations.
Corrective Actions / Comments NONE Completed By t
/1/M W. A. Fraser Date i
2 93 l Approved By . R. L. Pamell Date l Lead instructor, Simulator Training / ' l l
g'; , .* - THREE MILE ISLAND.
~
SIMULATOR MANAGEMENT i CERTIFICATION TEST ABSTRACT Loss of Instrument Air Certification Test
~ fD. . Test identification. Transient Test TTS09 -
ANSI /ANS- 3.5 Reference (s): 3,1.2.(2) Test Date. 03/19/95 Malfundion(s) Tested: lA01A instrument Air' Compressor Trip lA-P-1A - IA01B instrument Air Compressor Trip IA-P-1B 1A01C Instrument Air Compressor Trip IA-P-4 IA03A Service Air Compressor Trip SA-P-1B IA03B Service AirCompressorTrip SA-P-1B IA05 Header Loss of Instrument Air IA-P-1 A/B lA07 Primary Side Loss of instrument Air 100% (100% -Total Loss) lA08 Secondary Side Loss of Instrument Air 100% (100% - Total Loss)
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initishzation: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon i . Middle of Cycle . bf Point of Test Termination: , Reactor tripped due to loss of Reactor Coolant Pumps. The Reactor Coolant Pumps tripped due to: Loss of Seal
- Injection and intermediate Closed Cooling Water, instrument air pressure at 0 PSIG.
Simulator Run Time: 20 minutes-l Simulator Evaluation Time- 6 hours 1 ,l Baseline Evaluation Data: Previous Certification Test TTS09 dated 03/06/92. Right Direction Analysis . Simulator Malfunction Cause and Effects Document
- Current Controlled Copies of
Reference Plant Emergency Procedures Reference Plant P & ID Prints ] Reference Plant Alarm Response Procedures OverallTest Results: SATISFACTORY Results Description. In accordance with ANSI /ANS-3.51985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. l I A() l l (
,,n TTS09 l
]
This transient resulted in the complete rupture of a main two-inch and two three-inch Instrument Air header pipe with the added failures of all main air compressors. This caused the subsequent loss of adequate air pressure to , maintain control of plant systems. During this test, components realigned themselves to their fail-air state. I Instrument and Service Air Pressure decreased to the setpoints for starting of standby compressors. The Main Turbine tripped on loss of vacuum. All Reactor Coolant Pumps tripped due to loss of seal injection flow. Emergency feedwater actuated as designed and the reactor tripped on turbine trip. Reactor Trip Containment isolation aduated due to the reactor tilp. The current revision of plant emergency procedure EP 1202-36 (Loss of Instrument Air) was utilized to verify plant and component responses. The transient was tenninated afterthe reactor tripped and verification of component response per EP 1202-36. Final instrument A.ir pressure was O PSIG with backup instrument air compressors running. Dynamic response was compared to the response of the previous certification test. The response M this loss of instrument Air differed from the test conducted in 1992. The main Turbine tripped on low condenser vacuum, causing the reactor to trip prior to the loss of Reactor Coolant Pumps. Due to the loss of condenser vacuum, the Turbine Bypass valves were maintained closed on interlock, causing the Atmospheric Dump valves to reject all steam. Once-Through Steam Generator response differed due to modeling improvement, resulting in lowered OTSG levels. An improved pressurizer model resulted in slightly different system pressure response. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system ! actuations were recorded. Test results have been evaluated against ANSI /ANS 3,5 criteria by GPUN personnel j representing TMl SimulatorTraining. The TMI representatives hold or have held an NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for L impact and corrective action in accordance with Functional Fidelity Procedure,6211-ADM-2820.02. The results j_ of these evaluations are as follows: l The results nf this test are satisfactory, based upon the following: l CRITERIA #_1,: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. ' CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have I caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or l automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation. Qprrective Actions /Cowdents: Description. Incorrect Replica Plant Process Computer Alarms Number of Identified Deviations: One Corrective Actions: One Discrepancy Report Correction Alarm L2277 Completed By y/W William A Fraser Date ff Approved By R. L. Pamell Date 70 I Lead Instructor, Simulator Training C . l
1
)
g .; , 4 THREE MILE ISLAND
' SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT' Test Identification: Station Blackout Certification Test Transient Test P( TTS10 ANSI /ANS - 3.5 Reference (s): 3.1.2.(3)
Test Date. 03/30/95 Malfunction (s) Tested. ED01 Station Blackout Test initJahzation. Protected initial Condition IC-15 i 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated with the reactor tripped, decay heat being removed by natural circulation and the Emergency Diesels supplying power to essential equipment. The Emergency Feedwater System was supplying Feedwater to both OTSGs and steam was being dumped to the atmosphere. Simulator Run Time: 17 minutes Simulator Evaluation Time. 8 hours Baseline Evaluation Dataf Previous Certification Test TTS10 dated 02/05/90. Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY Results Descriptigg in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulatorto perform correctly following activation of a simulated malfunction. The transient resulted in the loss of off-site power causing a loss of power to the 6.9 KV,4.16 KV, and lower voltage buses. The Reactor Protection System tripped the reactor due to a loss of all Reactor Coolant Pumps snd loss of power to the Control Rod Drive Mechanisms. The Heat Sink Protection System actuated to initiate Emergency Feedwater to the OTSGs and the Atmospheric Dump Valves opened to remove decay heat. Natural , circulation cooling was established. Engineering Safeguards actuated and High Pressure injection was i repressurizing the Reactor Coolant system. 1
)
e O 4
-.___m._ _ _ _ _ _ - - - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _
glTG16 The event was initiated by a fault in the'230 KV substation resulting in the opening of all substation breakers. This resulted in a loss of power to the station Auxiliary Transformers. The loss of the Auxilialy Transformers j caused the Reactor Coolant Pumps to be lost. The Reactor Protection System tripped the reactor due to loss of g ali Reactor Coolant Pumps. The Heat Sink Protection System sensed a bss of all Reactor Coolant Pumps and initiated a start-up of the Emergency Feedwater System. The SASS actuated a mismatch indication for Turt>ine Header Pressure, T-hot and T-cold. I Electrical power was lost to all Secondary and Primary Systems. This resulted in a loss of Main Feedwater and ! Turbine support equipment. The power loss to the Primary Support Systems resulted in a 1.oss of seal injection to the Reactor Coolant Pumps and normal Makeup to the Reactor Coolant System. A low voltage condition was sensed on the Engineered Safeguards 4.16 KV buses and resulted la the automatic start-up and loading of the Emergency Diesel Generators. They supplied power to the Vital AC and DC Electrical Systems. Reactor Coolant System pressure decrease resulted in Engineering Safeguards Actuation repressurizing the system.
]
- The transient was terminated following the establishment of natural circulation heat removal with 50% Operating Range levelin the OTSGs.
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system
- actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held an NRC Senior Reactor Operator licenses for TMl Unit 1. Identified deviations from expected performance have been evaluated for
- impact and corrective action in accordance with Functional Fidelity Procedure, 6211-ADM-2820.02. The results of these evaluations are as follows
The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the j - actual reference plant and do not violate the physicallaws of nature. L,/ The simulator satisfied this requirement with no identified deviations. CRITERIA #2 i The simulator shall not fail to cause an alarm or automatic action if the reference plant would have
- caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
- The simulator satisfied this regtjrement with one identified deviation.
Corrective Actions /Cc6wents: l Descnotion. Incorrect Annunciator Alarm Operation - PLF-1-9 Number of identified Deviations: One Corrective Actions: Modification Completed By / Charles E. Husted Date 4 29 RI Approved By R. L. Pamell Date 2-9 I [ead Instructor, Simulator Training ' '
p J.
' THREE MILE ISLAND fN SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: DC Distribution Failure Certificr+. ion Test Transient Test TTS11
, ANSI /ANS 3.5 Reference (s): 3.1.2(3)-
i l Test'Date: 03/19/95 Malfunction (s) Tested:* ED08ADClDistributionSystemFailure(ASide)
- -
- Refer to Malfunction Cause and L Effects Documents for options available.
Test Initialization: Test Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor tripped and the plant at stable Hot Shutdown conditions. O Simulator Run Time: 10 minutes kf ' Simulator Evaluation Time: 4 hours- - Baseline Evaluation' Data: Previous Certification Test TTS11 dated 01/15/91. Right Direction Analysis. Simulator Malfunction 'Cause and Effects
. Document.
Current Controlled Copies of: Reference Plant Emergency Procedures - Reference Plant Alarm Response Procedures Overall. Test Results: SATISFACTORY Results
Description:
cIn accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly ~ following activation of a simulated malfunction. This transient resulted in a plant. trip due to a loss of DC power to the Turbine
' Control System. The plant stabilized at. Hot Shutdown conditions. Reference plant emergency. procedure EP 1202-9A (Loss of "A" DC Distribution System) was utilized to verify. proper simulator response. The reactor tripped from full power due to a turbine trip. Reactor Trip Containment ' Isolation occurred.
Control power to a.significant number of plant electrical buses was lost. f ,$s - p
i h TTS11 )
;d The transient was terminated when overall plant stability was achieved and .
pressurizer level and RCS pressure were recovering. Dynamic response was compared to the dynamic response of the initial certification test dated 02/09/90. During the conduct of this test simulator dynamic response annunciator operation and automatic safety system actuations were re' corded. Test results have been evaluated against ANSI /ANS 3.5 criteria .by GPUN personnel representing TMI
-Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure. 6211-ADM-2820.02. The results of these evaluations are as follows:
The results of this test are satisfactory, based upon the following: CRITERIA #1: The' observable changes in simulator parameters shall corres)ond in direction to those expected from the actual reference plant anc do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations.
/'"\
b CRITERIA #2; The simulator shall not fail to cause an alarm or automatic action if the reference I conversely,the plant would have simulator caused shall not the cause analarm alarmororautomatic automatic action, action ifand j the reference plant would not cause an alarm or automatic action. i The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: i NONE. Completed By [ / / M - 8 # -- - William A. Fraser Date 6/13/95 Approved By ._ _M T L. /% N6ct Date NNW Lead Instructor. Simulator Training , 1 i O l l
je. O '
.THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT f Test identification: Emergency Diesel Generator Certification Test O Transient Test
.Q TTS12 ANSI /ANS - 3.5 Reference (s): 3.1.2.(3) l l Igg [Ratt; 3/30/95 Malfunction (s) Tested. ED01 Station Blackout t l ' EG07A Emergency DG Trip- A 1 t'
EG07B Emergency DG Trip- B
~
Test initialization: Protected initial Condition IC-16 i 100% Reactor Power Equilibrium Xenon Middle of Cycle g Point of Test Termination: This test was terminated with the reactor tripped and the Reactor Coolant System on natural circulation heat removal using Er..ergency Feedwater. Electrical power was being supplied to . instrumentation and controls by the station battery. Simulator Run Time: 33 minutes Simulator Evaluation Time: 24 hours Baseline Evaluation Data: Previous Certification Test TTS12 dated 3/8/91. Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Pr'ocedures Reference Plant Electrical One-Line Diagrams Reference plant Emergency Procedures
-'Overall Test Results: SATISFACTORY Results Descnotion:
' In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the
, simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of offsite power with failure of the Emergency Diesel Generators to start. This
~
resulted in a loss of all 6.9 KV buses and 4.16 KV buses and lower voltage buses.' The Instrument and Control Systems were then powered from the station batteries. The reactor was ieipped, nii Reactor Coolant Pumps were lost, and Emergency Feedwater was supplied to the OTSGs to establish natural circulation cooling of the Reactor Cooling System. i V \ n,
TTS12 The event was initiated by introducing a fault on the 230 KV Substation buses resulting in the tripping of the Substation breakers and the Main generator breakers. With all breakers open there was no power to the Auxiliary Transformers. The loss of voltage to inplant buses resulted in the Reactor Protection System tripping the Reactor
; due to loss of all Reactor Coolant Pumps and initiating Reactor Trip Containment isolation. T he Emergency
( Diesels attempted to start on undervoltage but were prevented by malfunction of the Oil System. The Heat Sink Protection System sensed a loss of all Reactor Cooiant Pumps and initiated Emergency Feedwater with a 50% level control setpoint. Natural circulation cooling of the reactor was established. i As pressurizer level and Reactor Coolant System pressure decreased due to system shrinkage, without normal ! Makeup, the Engineered Safeguards System actuated. Because there was not voltage on the ES buses, the safety equipment did not start. The transient was tecninated when OTSG level was stable at 50% Operating Range and natural circulation cooling was established in the Reactor Coolant System. The Emergency Feedwater System was in operation maintaining the OTSG level and the Atmospheric Dump Valves were controlling OTSG pressure thus RCS temperature. l l During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives hold or have held an NRC Senior Reactor , Operator license for TMI Unit 1. Identified deviations from expected perfomiance have been evaluated for impact ) and corrective action in accordance with Functional Fidelity Procedure,6511.ADM 2820.02. The results of these , evaluations are as follows: f The results of this test are satisfactory, based upon the following: CRITERIA #1: I The observable changes in simulator parameters shall correspond in direction to those expected from the [, actual reference plant and do not violate the physicallaws of nature. (
\
The simulator satisfied this requirement with no identified deviations. CRITERIA #2: , ! The simuistor shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or , l automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with 6 identified deviations. l l l i I l l l
,a f
-TTS12 Corrective Actions / Comments:
Description incorrect Annunciator Alarm Operation V Number of identified Deviations: Corrective Actions: Four Four Discrepancy Report corrections j 1 Description
- Incorrect Replica Plant Process Computer Altams Number of Identified Deviations _:_ Two Corrective Actions: Two Discrepancy Report corrections
. Completed By _
_ R. L. Pamell Date S 2//7 Approved By _ f M& nager, Simulator Management R. L. Pamell Date $ht [ 4 l 1 O
THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT l 6.9 KV Bus FaultCertification Test (' n) Test Identificatiore: Translant Test
's_,/ TTS13 ,
j ANSI /ANS - 3.5 Reference (sr 3.1.2.(3) Test Date 0$/24/95 Malfunction (s) Tested: ED04B 6.9 KV Bus 1B Fault
- Refer to Malfunction Cause and Effects Documents for options available.
Test initiWration: Protected initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle 1 Point of Test Termination: The test was terminated following reactor trip with primary and l secondary pient pressures and temperatures approaching . normal post inp values. Simulator Run Time: 10 minutes I Simulator Evaluation Time: 6 hours I
/ Dgspline Evaluation Data; Previous Certification Test TTS13 dated 02/08/91.
(j Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Referenca Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY s Results Descrioticn: In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of two Reactor Coolant Pumps and subsequent reactor trip. The reactor trip initiated a Main Turbine trip and Reactor Trip Containment isolction. O I l L:
..r2. TTS13 The event was initiated by causing a fault on 6.9 KV Bus 18. The Reactor Coolant Pumps powered from the bus were lost and the Reactor Protection System initiated a reactor trip. The Main Turbine was tripped and Reactor Trip Containment isolation closed specified Containment isolation Valves. The Main Feedwater System reduced flow and established normal post trip level in the OTSGs. Reactor Coolant System pressure and temperature decreased to post trip values and were maintained. SASS actuated due to Turtaine Header Pressure transmitter differences.
The transient was terminated when primary and secondary pressures and temperatures were approaching post trip stable conditions. Results were compared to the dynamic response of the initial certification test dated February 8,1990. During the conduct of this test simulator dynamic response, annureciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TM1 representatives hold or have held an NRC Senior Reactor Operator licenses for TMi Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, C211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the fohowing: CRITERIA #1' The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
, The simulator satisfied this requirement with no identified deviations.
CRITERIA #2 Og The simulator shall not fall to cause an alarm or automatic action if the reference plant would have 4 1 caured the alarm or automatic action, and conversely, the simulator shall not cause an alarm or V automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions /Coininents: N'ONE
/
Completed By M , Daryl L. Wilt Date [ [f Approved By Lead instructor, Simulator Training R. L. Pamell Date /f6 l b (
'f
j 2 THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT 4.16 KV Bus Fault Certification Test Transient Test TTS14 ANSI /ANS 3.5 Reference (s): 3.1.2.(3) Jest Date: 06/05/95 Malfunction (s) Tested *: EDOSA 4.16 KV Bus 1 A 15ault
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition 10-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor operating at 70% power, one Main Feedwater Pump operating, and the plant stable at the lower power level. Simulator Run Time: 10 minutes Simulator Evaluation Time: 6 hours Previous Certification Test TTS14 dated 03/07/91. Right Direction Analysis Simulator Malfunction Cause and Effects Document
. Current Controlled Copies of:
Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of a Condensate Pump, which caused the trip of a Condensate Booster Pump and Main Feedwater Pump with a s.bsequent plant runback. Other support equipment was lost but cf no consequence to overall plant operation. . The event was initiated by introducing a fault onto 4.16 KV Bus 1 A. This de-energized secondary plant equipment powered from this bus. The loss of Condensate Pump 1 A without automatic start of the Standby Pump caused the trip of one Condensate Booster Pump and Main Feedwater Pump 1B. The Integrated Control System sensed the loss of the Main Feedwater Pump and initiated a High Load Limit runback. Power was reduced in the integrated mode until the High load Limit was cleared. Stability was achieved at the lower power level. The transient was terminated when stability was reached at the lower power level. During the conduct of this test simulator dynamic responso, annunciator operation, and automatic safety system f ^; ions were recorded. GPUN personnel representing TMI Simulator Training have evaluated test results against ( VlANS 3.5 criteria. The TMl representatives have held NRC Senior Reactor Operator license for TMl Unit 1
O. G TTS14
, Page 2 U-
[ dified deviations Tunctional Fidelityfrom expectedADM-2820.02. Procedure,6511 performance Thehave resultsbeen evaluated of these evaluationsfor are impact and corrective action in as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. E.BlIft918E The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement whh 2 identified deviations. , Corrective Actions / Comments:
Description:
Incorrect Annunicator Alarm Operation Number of identified Deviations. Two Corrective Actions: Two Discrepancy Report correction. J b Completed by:
/ M_ Charles E. Husted /
/
f[ Ofate L
' Approved by:
~
Supervisor Stinu'Titor Training Robert Parnell f!ze!9I Dafe m
TH EI MIIE ISUWB s 4. SINJumat Muguateur
, ?t - CEREIFIGTIG6 Her m >
l
- i L Test Idartification: Plart Heat 2p certifi'=*im Test .
l Nmum1 Operatie s Test NOICL q
- l.
- ABEI/AIS - 3.5 Referenon(s): 4.2.1 l
3.1.1(1) l
. . _ :.. 01/06/95
- l. Referanon Plart ProcedLaru: CP 1102-1 Plart Heatap To 525'F Revision 135 Test Initiallratim:. Protectet Initial canditim IC-03
- Cold Saatdcun Reacter coolart Systen Filled steen aabble j Pre Heatup, End of Cycle {
1 L I Poirt of Test Terminatim: , RCS toeparature 525'F. RC5 pressure 2155 PSIG l Safeties Rods cut ; l l simulater nun Time: 12 hours i Baseline Evaluation Data: Prwricus certification Test NDIO1 dated 01/17/90.
, Tactile Feel by Experienced Operators.
Ri@st Direction Analysis. i Ctarrant controlled copies of:
, Deferenos Plart Operating Procedures overall Test nesults: SMFISFAC1tRY Results
Description:
In accordanon with AISI/A15-3.5-1985 this certifidim test is condiacted to domestrate the ability to cparata the simulator in accordarica with similar reference piart %. Ling puw.22res, using ely cii ..Lcz artim rwmuni to the referanos plant. During the ooruttact of this test simulator dynamic performance, armanciator operatim,' and autmatic actims were evaluated against ANSI /Als 3.5-1985 oriteria. Tactile feel of simulator controls used diaring the test was also evaluated. These evaluations were conmacted by anu permannel representing sur Simulater Training. All the evaluatore have held.an lac sonice Reactor operator lionnme or certifidim for.5MI Unit 1. . Procedure steps not perfemund and identified deviations frtza expected performance have been evaluated for iW arid wwu.ctive acticut in accordanos with FLanctimal Fidelity Procediate 6211-A[M-2820.02. She'resalts of these evaluations are as follous: The results of this test are satisfactory, based sqxst the following: OtITERIA #1:
- 1ha simulater has met the @T.ie critaria of tis referenos plant p.- tme(s) f( used diaring the candtact of this test.
- 1ha simulator sigotted $his requirement with no deviatims.
Imol- # QtITERIA #2 1ha cheervable damsigns in simulatm parametere corresped in direction to those espected fram the *=1 referenom plant ard do not violate the physical laws of nature. The siunalator supported this regirement with no deviations. OtITERIA #3: The similator shall not fail to muss an alana or autcaustic actuation if the referenom plant inculd have caused an alam or automatic action, and conversely, the sinanlator shall not cause an alam or automatic actica if the reference plant idauld not cause an alarm or automatic actimi. The simulator supported this requirerient'with no deviaticais. QtIIERIA #4: Tactile feel of Plant-Rafersnood Simulator control device (s) cxmpares to those of the referenom plant. She simulator supported this requirement with no deviations. Corrective Actions /qaments: Durirg the conduct of this test, referenos plant ph steps not perfenmed 1dere identified as test Wims and evaluated for iM ard ce Jve action requirements in acocedanon with Functional Fidelity Procedure, 6211-ADE-2820.02. 1ha results of.these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test operator to violate the r e in ceder to proceed with the evolutism.
- 2. Steps not parfarund did not result in aboarvable differences in the control roan.
- 3. Steps not perfarund did not prevent the uman=Aal compl#4m of the p --ins in accordanon with plant limits and proomutions, tedinical specifications or p.- ins acomptanos criteria.
l , 4.. In accordanoa with 6211-ADI-2820.02, these test Wims have been l dars and ciceed as appropriate. I Ompleted By Onarles E. Itastad Data 04/21/95 approved By N L. Fam=11 Data 04/21/95 Imed Simulator Instructor { i l l l
.: .cs J THREE MILE ISLAND SIMULATOR MANAGEMENT l ) CERTIFICATION TEST ABSTRACT L
)
Test Identification: Plant Startup Normal Operations Test NOT02 ANSI /ANS 3. 5 Reference (s) - 4.2.1 3 .1.1 ( 2 ) 3 .1.1 ( 3 ) 3 .1.1 ( 6 ) Test Date: 05/19/95 Reference Plant Procedure OP 1101-2; Plant Startup Revision 119 Test Initialization: Protected Initial Condition IC-06 Hot Shutdown Safety Rods Out Xenon Free Point of Test Termination: 100% Reactor Power 579 Degrees Tave ICS in Full Automatic Control Rods at Normal Full Power
/ ; Positions
( /
'V Simulator Run Time: 6 Hours Baseline Evaluation Data: Tactile Feel by Experienced Operators.
Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures Reference Plant Alarm Response 1 Procedures Overall Test Results: SATISFACTORY
~
Results
Description:
l In accordance with ANSI /ANS-3.5-1985 this certification test is ! conducted to demonstrate the ability to operate the simulator in l accordance with similar reference plant operating procedures, using only operator actiod normal to'the referenced plant. During the . conduct of this test simulator dynamic performance, annunciator operation and automatic actions were evaluated against i. ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls
.used during the test was also evaluated. These evaluations were conducted by select GPUN personnel representing TMI Operator (q
)
\m/
l
- NOT02 Training. All the evaluators have held an NRC Senior Reactor t Operator license / certification for TMI Unit 1. Procedure steps not {
performed and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with i Functional Fidelity Procedure 6221-ADM-2820.02. The result's of l these evaluations are as follows: i The results of this test are satisfactory, based upon the following: l CRITERIA #1: l The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator supported this requirement with one deviation. l CRITERIA #2: The observable changes in simulator parameters correspond in
~
direction to those expected from the actual' reference plant and do not violate the physical laws of nature. The simulator supported this requirement with no deviations. CRITERIA #3: The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not , cause an alarm or automatic action if the reference plant L would not cause an alarm or automatic action. The simulator supported this requirement with no deviations. CRITERIA #4: Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. { l The simulator supported this requirement with no deviations. Corrective Actions / Comments: l
Description:
Incorrect System Response Number of Identified Deviations: 'One Corrective Actions: Discrepancy Report Written to correct 4th stage heater level control oscillation. l During the conduct.of this test, reference plant procedure steps not performed were identified and evaluated for impact and O. ' I-
i-i
. . ~s NOT02 ! p
,~ ,
l .,
, corrective action requirements in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: ,
l
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limite and precautions, technical specifications or procedure acceptance criteria.
7-
)
m_- Completed By _
! , Charles E. Husted Date 6/16/95 m.
( Approved By L L- Robert J. Parnell Date 6/16/95
,e s Lead Instructor, Simulator Training t
)
4 _/
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i THREE MILE ISLAND ' SIMULATOR MANAGEMENT' [ CERTIFICATION TEST ABSTRACT ' Test Identification: Reactor Trip and ' Recovery Certification Test i Normal Operations Test NOT03 ANSI /ANS 3.5 Reference (s)- 4.2.1 3 .1.1 (4 ) 3 .1.1 ( 6 )
. Test Date: 05/17/95 Reference Plant Procedure: ATP 1210-1, Reactor Trip Revision 29 OP 1102-2, Plant Startup Revision 119 Test Initialization: Protected Initial Condition IC-16 100% Reactor Power .
Equilibrium Xenoh Middle of Cycle Epint of Test Termination 1 100% Reactor Power p 5?9 Degrees Tave ICS in Full -Automatic Control Rods at Normal Full Power ! . Positions t l Simulator Run Time: 8 Hours l ! Baseline Evaluation Data: ' Tactile Feel by Experienced ! Operators. -) l Right Direction Analysis. i Current Controlled Copies.of: L Reference Plant Operating l Procedures Reference Plant Alarm Response
.,,p Procedures SATISFACTORY overall T _%lgJtesulta l
Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the referenced plant.
]
During the conduct of this test simulator dynamic performance, annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls ( 5 - 1 _ ]
a , s. m. -y - . s. . . _ s. q-I NOT03 i l used during the test was also evaluated. These evaluations were conducted by select GPUN personnel representing TMI Simulator Training. All tihe evaluators have held an NRC Senior Reactor Operator license / certification for TMI- Unit 1. Procedure steps not performed and identified deviations from expected performance have { j been evaluated for impact and corrective action in accordance with l Functional Fidelity Procedure 6221-ADM-2820.02. The results of l these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA ~#1: 1 The simulator has met the acceptance criteria of the reference { plant procedure (s) used'during the conduct of this test. The simulator supported this requirement with one deviation.
. CRITERIA #2:
, The observable changes in~ simulator parameters correspond in l direction to those expected from the. actual reference plant l and do not violate the physical laws of nature.
t p\ V The simulator supported this requirement with no deviations. CRITERIA #3: The simulator shall not, fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the raference plant would not cause an alarm or automatic action. The simulator supported this requirement with no deviations. CRITERIA #4: Tacti '; feel of Plant-Referenced Simulator control device (s) comp ~ m to those of the reference plant. The simulator supported this requirement with no deviations. l During the conduct of this test, reference plant procedure steps l not performed w6re identified and evaluated for impact and I corrective action requirements in accordance with Functional i Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did.not result in the need for the
. V
,s . . .
$ s 4 NOT03 test operator 'to violate the precedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful' corrpletion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
Corrective Actioris/ Comments: '
Description:
Incorrect System Response Number of Identified Deviations: One Corrective Actions: Discrepancy Report written to correct 4th stage heater level control oscillation. O . k
....?
I 1 Completed By Charles E. Husted - Date 6/16/95 Approved By Robert J.- _ rnell Date 6/16/9f,i Lead Instructok Simulator Training l w/ d
( THREE MILE ISLAND SIMUIATOR NANAGEMENT l CERTIFICATION TEST ABSTRACT l Test Identification Three RC Pump Operations Certification Test Normal Operations Test l NOT04 { l ANSI /ANS 3.5 Reference (s)r 4.2.1 3.1.1(6) 3.1.1(7) Test Date: , 04/27/95 i Reference Plant Procedure: OP 1102-2, Plant Startup Revision 119 OP 1102-10, Plant Shutdown j Revision 76 J Teat Initialization Protected Initial Condition IC-09 Hot Zero Power . -
, 1E-8 Amps Xenon Free End of Cycle, ,
]
t
)
O Point of Test termination Hot Shutdown - 532 Degrees Tave Safety Rods Out Simulator Run Timar 6 Hours Baseline Evaluation Data: Previous Certification Test NOT04 dated 01/19/90. Tactile Feel by Experienced operators.
, Right Direction Analysis. '
Current Controlled Copies oft' Reference Plant Operating Ficcedures
,y, Reference Plant Alarm Responsa gij, Procedures. .
Overall Test Rasults: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in
, accordance with similar reference plant operating procedures, using only operator action normal to the referenced plant.
During the conduct ot this test simulator dynamic performance, t . annunciator operation and automatic actions were evaluated against
. l
~g e .
NOT04 ANSI /ANS 3.5 1985 criteria. Tactile feel of simulator controls used during the test was,also evaluated. These evaluations were conducted by GPUN' personnel representing TMI Simul'ator Training. All the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6211-ADN-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the j following . I CRJTERIA #1: The ' simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test.
~
The simulator supported this requirement with no deviations. CRITERIA 92: The observable changes in simulator parameters correspond in - direction to those expected from the actual' reference plant and do not violate the physical laws of nature. l The, simulator supported this requirement'with no deviations. CRITERIA #3:
#fhe simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
The simulat w supported this' requirement with one deviation. CRITEpt$r94: Tact i' eel of Plant-Referenced Simulator control device (s) l compares to those of the reference plant.
.The simulator supported this requirement with no deviations.
Corrective Actions / Comments: Description Invalid Main Turbine Lube Oil temperature alarm RPPC point A0328. N=her of Identified Deviations: One i s l
,--v
't . .
NOT04 corrective Actions: one Discrepancy Report correction During the conduct of this test, reference plant procedure steps not performed were identified as N/A or exceptions and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure, 6211-ADet-2820.02. The results of these evaluations are as follows: ,
- 1. Steps not performed did not result in the need for the ,
test operator to violate the procedure in order to - proceed with the evolution.
- 2. Steps 'not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant i limits and precautions, technical specifications or i procedure acceptance criteria.
- 4. In accordance with 6211-ADIT-2820.02, ..these test exceptions have been documented and closed.
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- i. . . '
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1.
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. / ~
Complated By , _ Charles E. Hustad_ Date 04/27/95 gow dx Y y//s/4r I (. . .
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.. 95 TEE MIIE ISIJ80 GRrIFImrIm Talrr Aamuce l
Test Identificatient Zero Power Riysics Testing certification Test
. Normal Operations Test NCTIO5 ANSI /ANS - 3.5 Refenmon(s): 4.2.1 3.1.1(9) f 5.4.1(2) l Test Date: 05/01/95 Referanon Plant Procedure; RP 1550-02, Zero Pcuer Rysics Testing Revision 13 Test Initia14ratim: Protected Initial Candition I0-04 Hot Shutdcun Safeties out Xenon Free Begiming of cycle Point of Test Termiriatim: Reactor Critical at IE-4 Amps.
- 532 Degrees Tave.
, MS Pressure 2155 HEIG.
4 Simlator Run Tiine 8 hours ammalina Evaluatim Data: Previous Certification Test 10105 dated 03/23/91. Tactile Feel by Ekporianoud Operators. Right Direction Analysis. QIrrent controlled Copies of: Reference Plant Operating Prmartnus
. Reference Plant Alarm Response Prmartnw Overall Test Results: SATISFACIGN Basilts_nascrMirm )
In accordance with AISI/Al5-3.5-1985 this certification test is conchacted to demonstrate the ability to operate the simulator in accordance with sinflar reference plant operating y ;- twas, using only operator action normal to the reference plant. During the MM of this test minnlator dynmaic performance, amunciator cperation, and ='_*==*4hims were evaluated against AtBI/AIS 3.5-1985 critaria. Tactile feel l of simlehMuls used during the test was also evaluated. Shaos evaluaticms were conducted by@ tat perscrmal representing.1MI Simlater Training. All the evaluators have held an ISC Sonice Reactor GW.h licones for 1MI thit 1. PEMwe steps not perferned and identified'dsviations from WM performance have been evaluated for l 4W and hJve actica in accordance with Punctional Fidelity Pzmarkns 6211-AIM-i 2820.02. His resultat of these evaluaticms are as follows: The results of this test are satisfactory, based upon the following: CRIM!FJA #1: The simlator parameters has met the acceptance criteria of the referunoe plant I p 2 inu(s) used during the conduct of this test. The simlater supported this requirement with no deviaticms.
^- ,
, NDPOS ,
CRr1ERIA #2:
'Dyn aboarvable changes in simlator parameters m. .--d in direction to those empacted fram the e=1 reference plant ard do not violate the physical laws of
\ nature.
'Ihm simulator wtad this requirement with no deviations.
CRn'ERIA #3: .
'the simulater shall not fail to ames an' alarm ce automatic actuation if the referanos plant uculd have caused an alarm er automatic accias, and conversely, ,
the mimilater shall not cause an alarm er autcmartic acticri if the reference plant would not; cause an alarm or automatic action.
'Ihe si=11 mter supported this requirement with no' deviations.
OtITERIA #4: Tactile feel of Plant-Referenced Shmilater ocritrol device (s) campares to those of the reference plant. . ,
'Ihe simlater supported this requirunent with no deviations.
Carrective Actions /Ommaants: . During the cordet of this test, reference plant we' steps not perfcceed were identified and evaluated for 4M and =sWve action requirumments in accordance with functional Fidelity PrMwe, 6211-AD6-2820.02. 'Ihm results of these evaluations l are as follows: 1
- 1. Staps not performed did not result in the need for the test operator to violate the rh in order to e M with the evolution.
- 2. Steps not performed did not; result in cheervable differences in the control rocat.
l l
- 3. Steps not p b-- t did not prevent the sucomesful cxmpleticm of the ph in accordance with plant limits and pm* ions, technical specifications ce r.. twa acceptance critaria.
^
l D. > completed *er'! O t ; ciarles E. Hustad Date 06/20/95 ARaxpvud By N L. Patmm11 m 06/20/95 t 1' A%
;-- >. -e + y.cs w .a ; ,. .,
y ?, 4 THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Core Flood System Valve Operability Surveillance Normal Operations Test NOT06
- ANSI /ANS ~3. 5 Reference (s) : 4.2.1 <
3.1.1(10) Test Date: 06/06/95 Reference Plant Procedure: SP 1303-11.21 Core Flooding System ' valve Operability Test
, Revision 12 Test Initialization: Protected Initial Condition IC-03 Cold Shutdown Reactor Coolant System Filled <
' Steam Bubble-
- l Pre Heatup, End of Cycle :
Heatup and Raise Pressure to 530-PSIG ' Simulator Run Time: 30 minutes l Baseline Evaluation Data: Tactile Feel by Experienced Operators. . Right Direction Analysis. l ' Current Controlled Copies of: Reference Plant Operating i Procedures Reference Plant Alarm Response l Procedures Overall Test' Results: SATISFACTORY Results
Description:
In' accord < with ANSI /ANS-3.5-1985 this certification test is
= conducted W demonstrate the' ability to operate the simulator in accordance1dth similar reference plant operating procedures, using
.onlyloperator action normal to the referenced plant. .
During the - conduct .of this test simulator dynamic performance, annunciator operapion and automatic actions were evaluated against { ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls ! used during.the test was also evaluated. These evaluations were i conducted by select GPUN personnel representing TMI Simulator Training. All the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure stepe not perforr.;ed and
,- identified deviations from expected performance have been evaluated l
NOT06 , for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The, simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator supported this requirement with no deviations. CRITERIA _12: The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the. physical laws of nature. The simulator supported this requirement with no deviations. CRITERIA #3 i The' simulator shall not fail to cause an alarm or automatic ,i actuation if the reference plant would have caused an alarm or - automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant ; would not cause an alarm or automatic action. The simulator supported this requirement with no deviations. CRITERIA #4r i Tactile feel of PlantaReferenced Simulator control device (s) compares to those of the reference plant. 1 The simulator supported this requirement with no deviations. l f. CorrectivePActions/ Comments: . During the conduct of this test, reference plant procedure steps not performed were identified and evaluated for impact and corrective action. requirements in accordance with Functional b Fidelity Procedure, 6221-ADM-2820.02. Tho results of these evaluations are as follows:
- 1. Steps'not performed did not result in.the need for the test' operator to violate the procedure in order to O
-Qf proceed with the evolution.
O
s .- + ., . r* , , ,. 4 ,j NOT06 ' 1
- 2. Steps not performed did not result in observable differences in the control room. .
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria. )
t J i i I 1
.: 7 ,
k' i Q'. n , Completed By Charles E. Husted Date 6/20/95 hpproved By _
-- Robert L. Parnell Date 6/20/95 Manager, Simulator Management IO .
,4 , . .
THREE MILE ISLAND SIMULATOR MANAGEMENT } I CERTIFICATION TEST ABSTRACT g Test Identification: Emergency Power System Surveillance j Certification Test Normal Operations Test NOT07 ANSI /ANS 3. 5 Reference (s) - 4.2.1
'3 .1.1 ( 5 )
3.1.1(10) Test Date: 02/08/9.5 - Reference Plant Procedure: SP 1303-4.16 Emergency Power System Revision 73
~
Test Initialization: Protected Initial Condition IC-09 Hot Zero Power 1E-8 Amps - Zenon Free End of Cycle Point of Test Term 4 nation: Completion of Test of Emergency Diesel Generator-1A. ' Hot'Zero Power, 1E-8 Amps. s Simulator Run Time: 1 hour 30 minutes Baseline Evaluation Data: Previous Certification Test Dated 03/06/91. Tactile _ Feel by. Experienced
- Operators.
Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures ' Reference Plant Alarm Responst w Procedures OverallTab[Results: SATISFACTORY Results Descrioti'on: , In1 accordance with ANSI /ANS-3 5-1985 this certification test is conducted to demonstrate the-ability to operate the simulator in accordance with similar reference plant operating procedures, using _
, only operator. action normal to the referenced plant.
During the conduct of this test simulator dynamic performance,
~ annunciator operation and automatic actions were evaluated against
-ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls i ,
I 4 used during the test was also evaluated. These evaluations were
, conducted by select GPUN personnel representing TMI Simulator ~\ Training. All the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-28'20.02. The results of these evaluations are as follows:
The results of this test are satisfactory, based upon the l following: l CRITERIA #1: The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator supported this requirement with no deviations. CRITERIA #2: The observable changes.in simulator parameters correspond in direction to those expected from the actual reference plant and do,not violate the physical laws of nature. The simulator supported this requirement with no deviations. CRITERIA #3: The simulator shall not fail to cause an alarm or automatic. actuation if the reference plant would have caused an alarm or automatic action, and conversely., the simulator shall not cause an alarm or automatic action ~ if the~ reference plant would not cause an alarm or automatic action. The simulator supported this requirement with no deviations. CRITERIA #4: Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator supported this regt:irement with no deviations. Corrective Actions / Comments: During the conduct of this test, reference plant procedure steps not performed were identified and evaluated for impact and corrective action requirements in accordance with Functional Fidelity. Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: I c l'. Steps not performed'did not result in the need 'for the test operator to violate the procedure in order to 6 ______._______ _______________________m_ _ _ . . _ . _ _ _ _ . _ . _
aa , . . . NOT07 \ proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the precedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
'S
! l i Completed By s .ta ,, Charles E. Husted Date 4/21/95
- a Manager, Sin.ulator Management t
U_- -
l
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l
,e3 ATTACIIMENT B j (w' ). TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR l TESTING COMPLETED l YEAR #2 TESTS (1995-1996) 1
( Benenmark Tests TTS07 RCS Safety Valve Failure TTS19 Loss of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Steam Leak Inside Reactor Building TTS56 Manual Reactor Trip TTS57 Simultaneous Closure of All Main Steam Isolation Valves TTS58 Loss of One Reactor Coolant Pump ' TD59 Maximum Rate Power Ramp
. TTS60 Loss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Simulator Stability SSP 02 Simulator Accuracy Peal Time Test RTT01 Real Time Test
/ ,m) Transient Tests g%dj TTS15 480 V Bus Fault '
TTS16 480 V MCC Fault TTS17 ICS Auto Power Failure TTS18 Inverter Failure TTS20 Loss of Condenser Vacuum TFS21 Condenser Level Control Failure TTS22 Loss of Service Water TTS23 Loss of Shutdown Cooling TTS24 Loss of Component Cooling TTS25 Loss of Normal Feedwater TTS26 Normal Feedwater System Failure TTS28 Loss of Protective System Channel TTS29 Stuck Control Rod Nonnal Ooerations Tests NOT08 RB 30 PSIG Analog Clumnels
. NOT09 Main Steam isolation Valves Surveilhtnce NOT10 Main Steam Isolation Valves Monthly Surveilkmce NOTI 1 Shift and Daily Checks Surveillance l NOT12 Weekly Surveilhtnce Checks
! NOT13 RCS Leakrate O (x l
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# THREE MILE ISLAND y2 SIMULATOR MANAGEMENT CERTIFICATION TEST. ABSTRACT .
s Test identification: RCS Safety Valve Failure Certification Test Transient Test TTS07
-t ANSI /ANS - 3.5 Reference (s): B.2.2.(10) 3.1.2(1b) 3.1.2(1d)
. Test Date: 01/31/96 '
Malfunction (s) Tected:* ES01A ESAS Failure to Actuate at HPI Setpoint Channel A ES01B ESAS Failure to Actuate at HPI Setpoint. Channel B-RC27A Pressurizer ' Safety Valve Fails Open 25% (100% = Full Open)
'* Refer to Malfunction Cause and Effacts Documents for options available.
Test-Initialization Protected Initial Condition IC-15' 100% Reactor Power
- Equilibrium Xenon Beginning of Cycle i~
Point of Test Termination: This test was' terminated with the; reactor tripped,
. Reactor Coolant System saturated, pressurizer filled and -
system inventory. decreasing. Simulator Run T4==: 10 minutes
' Simulator Evaluation Ti==: 2' hours-2 Baseline Evaluation Data: Previous Certification Test TTS07 dated 02/27/90.
Right Direction Analysis.
*imulator Malfunction Cause and Effects Document Jurrent Controlled Copies'of:
Reference Plant Alarm Response Procedures Referencs Plant' Electrical One-Line Diagrams
. 'Overall Test Results: SATISFACTORY l-
.Results'
Description:
lC In1 accordance . with ANSI /ANS-3.5-1985 this certification test was conducted to' L ' demonstrate the ability of the simulator to perform correctly following activation of-a' simulated malfunction.
= This transient resulted in the loss of Reactor Coolant. System inventory ' through a partially ' open Code . Safety Valve on the pressurizer. The pressure in the Reactor Coolant . : System decreased to saturation and the pressurizer fillsd -solid. The
~2ngineered Safeguards Actuation-System was prevented from operation. ,
The' event was initiated by failing a Pressurizer Code Safety Valve partially open. As 4 Reactor Coolant; System pressure .. decreased the Reactor Protection System actuated tripping the reactor.. and initiating Reactor Trip Containment Isolation. Pressure 1 continued to decrease and Engineered Safeguards Actuation System High Pressure
TTS07 ' Injsetion actuttion satpoint wts ratchsd. Tha cyattm did not actuate du2 to tha malfunctions inserted.
,_, Reactor Coolant System pressure continued to decrease, Reactor Vessel and Reactor
/ j Coolant System voiding caused the pressurizer to fill solid. Reactor Coolant Drain
' j Tank pressure increased causing rupture disk rupture with subsequent activity release to containment. Due to model changes the RCDT variable could not be plotted. RCDT pressure was observed on the control room instrument and instructor station monitored parameter plot. RCDT pressure response wac correct before and af ter rupture disk operation.
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations.
/~'N CRITERIA #2:
i ) The simulator shall not fail to cause an alann or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Garrective Actions / Comments: NONE i Completed By ,_/ Charles Husted Date 05/31/96 Approved By ,
- Robert L. Parnell Dated!/
I
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~ _ _ _ _ _ _ _ _ _ _ ____
THREE MILE ISI).ND
,, SIMULATOR MANAGEMENT p: CERTIFICATION TEST ABSTRACT i
. Test Identification:
Loss of Forced Flow Certification Test
/ s Transient Test
) TTS19 ANSI /ANS - 3.5 Reference (s): 3.1.2(4)
Benchmark Test B2.2(4) Test Date: 02/22/96 l Malfunction (s) Tested:* RC 35 A/B/C/D - RC' Pump Trip A/B/C/D !
- Refer to Malfunction Cause and Effects l' Documents.
Test Initialization: Protected Initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: ' Reactor tripped with Reactor Coolant System heat transfer by st'able natural circulation'. Simulator Run T4==: 10 minutes S.igniator Evaluation Tim =: 6 hours- - ( l (v Baseline Evaluation Data: Previous Certification Test TTS19 dated 02/28/90, o 01/19/95 ' Right Direction Analysis. j RETRAN Data. Current Controlled Copies of: Reference Plant Alarm Response Procedures Referv.nce Plant Emergency Procedures Overall Test Results: SATISFACTORY.1 Results
Description:
In' accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of , a simulated malfunction. This transient resulted in the loss of forced flow in the Reactor Coolant System. Dacay heat removal was accomplished by the use of Emergency Feedwater flow to tha. OTSGs ,j and steaming to the Main Condenser with natural circulation evident. i The event was initiated by tripping all four Reactor Coolant Pumps on overload. The reactor tripped and initiated Reactor Trip Containment Isolation. The Emergency , l Feedwater System actuated and filled the OTSGs to 50% level on the Operating Range. !
- l. The plant transitioned to the natural circulation mode. The current revision of ATOG l Procedure 1210-10 was utilized to verify evidence of natural circulation. The RCS
.n pressure rose due to the Makeup System regaining pressurizer level, compressing the -l i steam bubble. Because this test was performed from a Beginning of Cycle Initial V Condition', decay heat levels were relatively low, causing the_ average differential temperature between hot and cold leg temperatures to be less than the procedurally stated 30-50 degrees Fahrenheit.
f
TTS19 e, The trcusisnt wzo terminated following the verification of natural circulation flow and overall plant response utilizing ATOG Procedure 1210-10. p During the conduct of this test simulator dynamic response, annunciator operation, and
't automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training.
The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact f and corrective action in accordance with Functional Fidelity Frocedure 6221-ADM-2820.02. The results of these evaluations are as follows: j i The results of this test are satisfactory, based upon the following: CRITERIA #1: { The' observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical f laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: ) j The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. g The simulator satisfied this requirement with no identified deviations.
\ Corrective Actions /Cc.-- nts:
NONE
~
Completed By
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/ William A. Fraser Date 03/07/96 Approved By Robert L. Parnell Date 7!hd
.4*
C, t
i f i THREE Mii E ISL91D . g.d* SIMULATOR MANAGEMENT
' CERTIFICATION TEST AB3 TRACT
\ f Test Identification: Loss of All Feedsater Certification Test ! - l\ Transient Test
'TTS27 i I
1 ANSI /ANS - 3.5 Reference (s): B.2.2.(2) 3.1.2(10) I Test Dg.31 01/31/96 i Malfunction (s)' Tested:* . W15A/B Main Feedwater. Pump Trip 1A/1B W17 Emergency W Pump Trip (EF-P-1) W18A/B Emergency W Pump Trip (EF-P-2A/B)
- Refer to Malfunction Cause and Effects Documents for options available.
' Test Initialization: Protected Initial Condition IC-16 100% Reactor Power Equilibrium Xenon-
, Middle of Cycle '
Point of Test Termination: This test was terminated with the reactor tripped, Reactor Coolant System heatup in progress with Primary . System pressure' controlled by the pressurizer Power Operated Relief Valve. Pressurizer level was increasing 4 v ) and RCS-subcooling margin was decreasing. Simulator Run Tima: 10 minutes Simulator Evaluation Time: 4 hours Baseline Evaluation Data: Previous Certification Test TTS27 dated 02/27/90. Right Direction Analysis. Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures I
, Reference Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures , !
1 Overall Test Results: SATISFACTORY Results
Description:
In:accordance with ANSI /ANS-3.5-1985. this certification test was conducted to. demonstrate- the ability.of the simulator to perform correctly following activation of a simulated malfunction. This transient. resulted .in the loss 'of all secondary side heat removal with Reactor Core' heat removal via the PORV. During the duration of this test, heat transfer from
, the RCS was insufficient for removal of decay heat as expected.
) The event was initiated by failing both Main Feedwater Pumps and preventing the start of all Emergency Feedw'tera Pumps. The Reactor Protection System tripped the reactor ,
and initiated Reactor Trip' Containment Isolation. The Heat Sink Protection System actuated to initiate Emergency Feedwater but the malfunctions prevented the pumps from 4 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ . . .m_____
TTS27 -2 W *sterting. As etsca ginarotor invsntory was lost R2cctor Coolant System tsmparature increased causing system pressure to increase resulting in opening of the pressurizer spray valve then the PORV.. At the end'of the test the PORV was cycling to limit Reactor Coolant System pressure and the Reactor Coolant system was approaching [] - saturated conditions. The transient was terminated following CTSG dry-out, with RCS heatup in progress and the PORV controlling RCS pressure. During the conduct of this test sinulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in'accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this' requirement with no identified deviations. CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the
-(^ reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
The simulator satisfied this requirement with two identified deviations. Corrective Actions / Consents:
Description:
Replica Plant Process. Computer Alarms S2046, S2051 and L2618 incorrectly alarmed. Number of Identified Deviations: Two { Corrective Actions: One discrepancy report and PDR file update. i Completed By l
/
e Charles Husted Date 06/19/96 l Approved By / '
, Robert L. Parnell Date 06/19/96
(
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I: .6 4 *
- THREE MILE ISLAND 3' SIMULATOR MANAGEMENT
+ CERTIFICATION TEST ABSTRACT P
> Test Identification: Turbine Trip Certification Test Transient Test TTS35' L ' ANSI /ANS - 3.5 Reference (s): 3.!. 2(15)
BenchmarkJTest B2.2(6)
- Test Date: 02/01/96 l; . . . ,
' Malfunction (s) Tested: TC01 Turbine Trip t
i
*. Refer to Malfunction Cause and Effects Documents for h options available.
L Test Initialization: Protected Initiil Condition IC-13 ll 50% Reactor Power- ' pa = Equilibrium Xenon. J End of Cycle - Point of Test Termination: The test was . terminated after plant. stability was L reached'at the lower reactor power level following the Main. Turbine trip.
- S4==ilator Run T4==: 10 minutes V
jj q S4==ilator Evaluation T4==: 2 hours ' l' ) . Baseline Evaluation Data:
/7 Previous Certification Test TTS35 dated 03/01'90.
/ ,
[ Right.Direetion Analysis. D Simulator Malfunction Cause and Effects Document. Current Controlled. Copies of: l . Reference PlantL Alarm Response Proceduras. Reference Plant Electrical One-Line. Diagrams Overall Test Rasults: SATISFACTORY Results
Description:
In accordance . with ANSI /ANS-3.5-1985 this certification test was conducted to-p- demonstrate the ability of the simulator to perform correctly following activation of - a' simulated malfunction.
.This transient'resulted.in a turbine trip with reactor power below the high power trip
,,' setpoint.
L The event was .' initiated by activation of the turbine trip malfunction after power was I reduced tol41% in order to prevent a. reactor trip. The Integrated Control System. E responded by initiating a power reduction, 'since the control system transferred to the I Tracking mode with megawatts generated equal to zero when~the turbine was tripped. Reactor power change due to increased temperature was less significant than the initial test due to : incorporation of a new core modal with different temperature coefficient
-values.: Reactor = Coolant System pressure and temperature decreased significantly as a[
t reactor;' power decreased to less. than 22%. SASS actuated a mismatch indication for
. T-cold transmitters. Reactor Coolant System' pressure initially increased resulting i
I L_ a:_ _L- _. _ - - _ _ _ . - _ - - - - _ _ _ _ .
TTS35 - 2 . in operation of the Pressurizer spray valve. Emergency Feedwater Actuation occurred f'~'N due to low level in OTSG-B. Main Feedwater underfeed occurred on OTSG-B only due to ! ) it's high steam demand supplying the Gland Steam System demands. sa The transient was terminated when primary and secondary pressures and temperatures were returning to stable conditions. Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test, simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA r1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfP 5 this requirement.with one identified deviation. (] CRITERIA 2: (N/ / The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation. Cprrective Actions / Comments:
Description:
Incorrect Computer Alarm Actuation Number of Identified Deviations: One ferrective Actions: One discrepancy report. RCP void fraction alarms.
Description:
Incorrect System Response Number of Identified Deviations: One Corrective Actions: None - unable to reproduce on this date. Completed By Charles Husted Date 96/20/96 l l Approved By , Robert L. Parnell Date 06/20/96 I
- p. Supervisor, Simulator Training
( ) \_/ s l l l j
THREE MILE ISLAND o SIMULATOR MANAGEMENT 6' l CERTIFICATION TEST ABSTRACT i s
, Test Identification:
Main Steam Leak Inside the RB Certification Test [m} . Transient Test C/ TTS42 l ANSI /ANS - 3.5 Reference (s): Benchmark Test B2.2(9) l 3.1.2(20) l Test Date: 03/01/96 Malfunction (s) Tested:* MS02A Main Steam Leak in the RB
,100% (100% - 6,000,000 LBM/HR) !
- Refer to Malfunction Cause and Ef fects Documents for options available.
Test Initialization: Protected Initial Condit! t IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated following the reactor trip, HPI, and. Emergency Feedwater actuations due to the effects of a main steam line rupture inside containment.
.p / j Simulator Run Time: 10 minutes \ )
Simulator Evaluation Time: 4' hours Baseline Evaluation Data: Previous Certification Tests TTS42 dated 03/02/90, 06/23/95. Right Direction Analysis. RETRAN Data. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated by causing a maximum size steam rupture inside containment on the OTSG 1A. The resultant steam release caused a reactor and turbine trip on Overpower with subsequent Reactor Trip Containment Isolation actuation. The containment pressure and temperatures increased causing Engineered Safeguards actuation, and a f alse Reactor Building fire alarm. The rapid blowdown of OTSG 1A caused a significant cooldown of the RCS. A SASS actuation occurred due to the rm significant mismatch between the A and B Main steam header pressures. Feedwater flow (v) was isolated to both OTSGs on HSPS Low Pressure. The Emergency Feedwater System actuated on high Reactor Building pressure and fed the OTSGs. High pressure injection flow recovered the RCS inventory lost by shrinkage and compressed the pressurizer steam bubble, elevating RCS pressure. The spray valve automatically actuated to dampen the pressure increase, j l
TTS42 'j Th transient was terminated following Engineered Safeguards and Emsrgency Feedwater System actuations with elevated Containment pressures and temperatures. e This test differed from the initial benchmark test in 1990 due to the impact of lowered [ m\ OTSG 1evels on the simulator. The 1990 test produced high OTSG 1evel Feedwater , () isolation which limited the post-trip cooldown. The 1996 test resulted in lower RCS pressures and temperatures, and lower Pressurizer level as a result of continued feed flow. This then caused a longer period of time for conditions to return to more normal post-trip conditions after OTSG 1A isolation. Dynamic response was . compared to the dynamic response of the initial certification t e s t -, During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TML Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have l been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: i
. The results of this test are satisfactory, based upon the following:
QRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations.
/rm CRITERIA #2:
The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: . NONE Completed By _ #,, d '~ William A. Fraser Date 05/22/96 Approved By 4 _ W Robert L. Parnell Date r ( - (
THREE MILE ISLAND v
- A, s. SIMULATOR MANAGEMENT 2 - CERTIFICATION TEST ABSTRACT Test-Identification: Manual Reactor Trip Certification Test
' ' Transient Test TTS56 ANSI /ANS - 3.5 Reference (s): Benchmark Test B.2.2(1)
Test Date: 03/01/96 Malfunction (s) Tested:* None - Initiation of Manual Reactor Trip Test Initialization: Protected. Initial' Condition IC-17 100% Reactor Power Equilibrium Xenon . End of Cycle.
-Point of Test Termination: This test was terminated with the reactor tripped and proceeding to stable Hot Shutdown conditions.
l Simulator Run T4= : 10 minutes J Si=nlator Evaluation T4=: 2 hours . Baseline Evaluation Data: ' Previous. Certification Tests dated '03/07/90, 01/09/95 Right Direction Analysis. Current Controlled Copies of: ' (% ' Reference Plant Alarm Response. Procedures Reference Plant Emergency Procedures-Overall Test Results: SATISFACTORY I: -Results
Description:
y LIn . accordance with- ANSI /ANS-3.5-1985 this certification test was conducted to
~
demonstrate the ability.of .the simulator to perform correctly following activation of [ a manual reactor trip. The event was initiated by depressing the Manual Reactor Trip pushbutton on the Main-Console and reducing the pressurizer level control setpoint to 100 inches. . All control' rod drive breakers opened and Reactor Trip Containment Isolation actuated. The Main t
-Turbine tripped on interlock. The' Integrated Control System reduced Feedwater flows l.
and stabilized OTSG 1evels at Low Level Limits. The Turbine Bypass Valves operated to maintain OTSG pressure and remove. decay heat. The transient was terminated with the reactor tripped and the plant approaching stable .. Hot Shutdown conditions. 1
' During the conduct of this test simulator dynamic response, annunciator operation, and automatic-safety system actuations were evaluated. --Test results have been evaluated
~
against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Siinulator Training. The-TMIErepresentatives have held NRC Senior Reactor Operators licenses for TMI Unit ( 1 '. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. o
TTS56 ,,
'3 ,
The racults of thass evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator paranieters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this-requirement with no identified deviations. CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied,this requirement with no identified deviations. Corrective Actions / Comments: NONE Completed By _ i William A. Fraser Date 04/10/96 fm Approved By Robert L. Parnell Date 6Z 1 [
,s.
THREE MILE ISLAND SIMULATOR MANAGEMENT - CERTIFICATION TEST ABSTRACT N_ Test identification. Simultaneous Closure of All MSIVs Certification Test x Transient Test TTS57 ANSl/ANS - 3.5 Reference (sk Benchmark Test B.2J.(3)
, Test Date: 03/01/96 Malfunctions) Tested. MS08A/B/C/D Main liteam isolation Valve closure MS-V 1 A/1B/iC/1D
*Referto Malfunction Cause and Effects Documents for options available.
Test initialization. Protected initial Condition IC-16 100% Reactor Power -
. Equilibrium Xenon Middle of Cycle
- Point of Test Termination: This test was terminated v.ith the reactor tripped and Emergency Feedwater removing decay heat via the Atmospheric Dump Valves. -
Simulator Run Time: 20 minutes Simulator Evaluation Time. 4 hours Baseline Evaluahon Data: Previous Certification Tests dated 03/13/90,01/04/95 Right Direction Analysis.
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Current Controlled Copies of: Reference Plant Alann Response Procedures Reference Plant P & ID Prints Overall Test Results: SATISFACTORY Results Descnotion: in accordance with ANSI /ANS-3.5-1985 this certification test'was conducted to demonstrate the ability'of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip on high pressure and Emergency Feedwater System actuation due to a l
, loss of Main Feedwater..
l- 1 I J f~~ ( I. . , i i
._______._.____.m__ _____.-----.- -- - -
.c ~', TTS57 The event was initiated by simultaneous closure of all four Main Steam isolation Valves. This caused OTSG pressures to increase and Main Steam Header Pressure to decrease. The Main Turbine transferred to manual
,3 control automatically upon receipt of a sustained 40 PSIG Header Pressure Error Signal. As OTSG pressures increased the Turbine Bypass, Atmospheric Dump and Main Steam Safety Valves operated to limit the pressure excursion. Reactor Coolant System pressure and temperatures increased due to the transient. The reactor tripped on high pressure and actuated Reactor Trip Containment isolation. When all Main Steam isolation Valves had closed the Main Feedwater Pumps slowed, reducing Feedwater flow to the OTSGs. .The Emergency Feedwater System actuated to feed the OTSGs at Low Level Umits. Gland Sealing steam was lost when the OTSG 1B MSIV's closed, causing a loss of vacuum to the Main Turbine and Main Feedwater Pumps. Decay heat was then removed by automatic action of the Atmospheric Dump Valves, since the Turbine Bypass Valves closed due to the low Condenser Vacuum condition (caused by the loss of Gland Seal Steam). '!
The transient was terminated with the reactor tripped and decay hea't being removed by the Emergency Feedwater System via the Atmospheric Dump Valves. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 cdteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold o'r have held NRC Senior Reactor Operator licenses for TMi Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,622-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1- > The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
) The simulator satisfied this requirement with no identified deviations.
J . CRITERIA #2
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The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments NONE Completed By [ /MAI, Ah n -- - _ W. A. Fraser Date Y-/d- 9h Approved By - Lead instructor, Simulator Training R. L. Pamell Date
/5N4 ;
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THREE MILE ISLAND SIMULATOR MANAGEMENT j CERTIFICATION TEST ABSTRACT f-~3 Test Identificat12n: Loss of One RC Pump Certification Test
/ j Transient Test tj TTS58 ANSI /ANS - 3.5 Referencq111: 3.1.2(4)
Benchmark Test B2.2(5) Test Date: 02/01/96 Malfunction (s) Tested: RC35A RC Pump Trip _ Test Initialization: Protected Initial Condition IC-15 100% Reactor Power Equilibrium Xenon l Beginning of Cycle I Point of Test Termination The test was te rminated following reactor ' trip with Primary and Secondary Plant pressures and temperatures approaching stability. j t% , ( ) ( ,/ Simulator Run Time: 10 minutes
)
Simulator Evaluation Time: 3 hours l Baseline Evaluation Data: Previous Certification Test TTS58 dated 03/15/90 Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY Results
Description:
1 In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. l This transient resulted in a reactor trip. Reactor Trip initiated a Main Turbine trip and Reactor Trip Containment Isolation. ( ( ) The event was initiated by failing Reactor Coolant Pump 1A. The loss of the N Reactor Coolant Pump at full power resulted in the Reactor Protection System
TTS58 - initiating Rasctor Trip dua to tha reduction in Reactor Coolant System flow. The Reactor Trip resulted in a Main Turbine t rip ., Reactor Trip Containment ' Isolation closed specified Containment Isolation Valves. Reactor Coolant System pressure and l l temperature decreased to post trip values and were maintained. SASS actuated a mismatch indication for T-cold and T-hot transmitters. b ' The transient was terminated when Primary and Secondary pressure and temperature were approaching post trip stable conditions. During the conduct of this test simulator dynamic responsc, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator License for TMI Unit
- 1. Identified deviations from ' expected performance have been evaluated for impact and corrective ' action in accordance with Functional Fidelity Procedure , 6221-ADM-2820.02.
The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: . The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA 2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, I the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. I The simulator satisfied,this requirement with one identified deviation. Corrective Actions / Coa = ants:
Description:
Incorrect Computer Alarm Response Number of Identified Deviations: One L Corrective Actions: Discrepancy Report - R94-1013 Completed By Charles Husted Date 05/30/96 i Approved By Robert L. Parnell Date h Supervisor, Simulator Training
THREE MILE ISLAND SIMULATOR MANAGEMENT b'" CERTIFICATION TEST ABSTRACT Test Identification: Maximum Rate Power Ramp Certification Test
/] Transient Test TTS59 ANSI /ANS - 3.5 Reference (s): Benchmark Test B.2.2(7) 5.4.1(2)
Test Date: 01/25/96 Malfunction (s) Tested:* None - Manual Plant Maneuvering Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was. terminated with the plant stable at 100% power. Simulator Run Time: 15 minutes Simulator Evaluation Time: I hour Baseline Evaluation Data: Previous Certification Test TTS59 dated 3/12/90 Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. I Th'e event was initiated from steady-state full power with the Integrated Control System in the integrated mode. The Unit Load Demand cor. troller was set to approximately 75 percent. load. The plant reduced power at the maximum rate of ten percent per minute
- and stabilized at 75 percent power. The plant was allowed to stabilize for approximately four minutes. At the end of the stabilization period, power was ramped back up at the maximum sate of ten percent per minute. When the plant reached 90 percent load the Integrated Control System automatically reduced the rate of power increase to 5 percent per minute. During the period of time at reduced power the Xenon concentration began to increase. The plant was allowed to stabilize at full power.
The transient was terminated with the plant at full power. Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test simulator dynamic response, annunciator
,n operation, and automatic safety system actuations were reccMed. Test results have i been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI
( Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6211-ADM-2820.02. The results of these evaluations are,as follows:
TTS59 '" T The racults cf this tact cro actisfcctory, becad upsn th2 following: CRITERIA #1: O The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. ) CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA. 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. Corrective Actions /Co - nts: NONE Completed By [AM, e - William A. Fraser Date 02/02/96 Approved By 9/ Robert L. Parnell Date 02/05/96 G e sp
4 THREE MILE ISLAND' SIMULATOR MANAGEMENT i . .
- CERTIFICATION TEST ABSTRACT-
[ 4 . (4 [. Test Identification: . Loss of Offsite Power with Large Break LOCA L Certification Test Transient-Test. TTS60 l . ANSI /ANS - 3.5 Referencets): 3.1.2(Ib) i Benchmark Test B.2.2(8) I Test Date' 12/11/95
'Malfunctionful Tested
- ED01 Station Blackout TH04A RCS LOCA'at Hot Leg Nozzle 100%-severity L
- Test > Initialization':
Protected-Initial Condition'IC-16 i ! 100% Reactor Power Equilibrium Xenon p , Middle of Cycle l; ' ' ' Point'of Test Termination: 'This' test was terminated with.the reactor. tripped,
, Core Flood tanks emptied, and reactor core cooling being provided by High Pressure-and Low Pressure Injection. . Reactor. Building pressure and temperature were decreasing from peak conditions.
l , ' Simulator Run Time: 10 minutes simulator Evaluation Time: = 9 hours [_ Baseline Evaluation Data Previous Certification' Test TTS60 dated'03/08/90. Right Direction Analysis. L l Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: L.\ Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line' Diagrams; ,
. Reference' Plant Emergency Procedures Overall Test Results: SATISFACTORY Eggglts
Description:
-In accordance.with: ANSI /ANS-3.5-1985 this certification test was conducted to
~ demonstrate the ability of.the simulator to perform correctly following activation p ;of:a simulated malfunction.
This transient-resulted in the loss of offsite power coincident with a large break loss of coolant accident. 'The reactor was tripped due to a loss of power to the Reactor' Coolant Pumps and the Control Rod Drive System. Safety system power was
< restored by.the Emergency Diesel Generators. A large break in the Reactor Coolant
- j. . System caused-rapid depressurization of the coolant and a rapid increase in Reactor Building pressure. .The Engineered Safeguards System. actuated, initiating-safety
! injection _and Reactor Building. Isolation and cooling.
'The' event was initiated by fault in the 230 KV: Substation coincidentLwith a-large size _ rupture of the. Reactor Coolant System Hot Leg piping.~ The Reactor Protection System initiated a Reactor trip and Main Turbine-trip. ' Reactor Trip Containment Isolation closed specified Containment Isolation. valves. The Heat Sink Protection
- System detected a loss of Reactor Coolant Pumps and initiated Emergency Feedwater to
.the OTSGs, controlling level at,50% on the operating range. The Emergency Diesel Generators started and the output breakers closed to re-energize their 4.16 KV Buses. SASS-actuated a mismatch indication for RCS Hot Leg Temperature
-Transmitters.
)
9 ;
TTS60 .T'ha loss of inventory and precsure from tha Racetor Coolant Systrm resulted in s' ' discharge of the Core Flood Tanks and actuation of the Engineered Safeguards Actuation System. High Pressure Injection, Low Pressure Injection, Reactor Building Spray, and Reactor Building Isolation and Cooling were initiated due to the pressure loss from the Reactor Coolant System and the high pressure detected in the Reactor {A Building. The transient was terminated when Reactor Building pressure and temperature were decreasing and reactor core cooling was verified. Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test simuistor dynamic response, annunciator operation, and automatic safety system actuations were evaluated. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operators licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference' plant and do not violate the physical laws of nature.
,The simulator satisfied this requirement with no identified deviations.
CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the g reference plant would have caused the alarm or automatic action, and g , conversely, the simulator shall not cause an alarm or automatic action if the
\g reference plant would not cause an alarm or automatic action. d The simulator satisfied this requirement with no identified deviations.
Corrective Actions / Comments: None Completed By / Charles, E. Husted Date 06/28/96 g Q
,d Ah % 6. O ge 3r, O aks) '
Approved By - Q . Robert L. Parnell Date 06/28/96 c) i 1 L i Y w___-__-____.___---- _-
c
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THREE MIIE ISERO SIMEATOR MANAGEMENT CERTIFICATION TEST ABSTRACT I Test Identification: Sim11ator Stability Certification Test Steady State Performance Test SSP 01
. ANSI /ANS - 3.5 Reference (s): 5.4.1(2) 4.1 Benchmark B2.1
. Test Date: 01/12/96 Test Init4al 4 vat 4nry Protected Initial Condition IC-17 100% Reactor Power Pqi414hritER Xenon End of Cycle Point of Test Taminariari: 100% Reactor Power.
579 Degrees Tave. ICS in Full Autanatic Mmilator Run Tima: I hour ' ( Baseline Evaluatinri Data: Plus/minus 2%. Overall '1%st Ra= nits: SATISPACitRY t [ Results Descrinkinrl: i In erdance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the similator to operate at stable full power, in autcznatic ' control for a 60aninute period. i l During conduct of this test similator dynamic operation was evaluated against ANSI /ANS i 3.5 criteria. This evaluation was conducted by GPUN personnel representing 'IMI Sinulator Training. The evaluators have held an NRC Senior Reactor Operator license for 'IMI Unit 1. The results of these evaluations are as follows: The results of this test are satisfactory, haamri upon the following: CRTIERIA #1: The simulator ocmputed values for steady state, full power operation with the referenced plant control system configuration shall be stable and not vary more than plus/minus 2% of the initial values over a 60aninute period. The simila e supported this requirment with no deviations. *
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Ccmpleted By /, Charles E. Husted Date 01/30/96 Approved ay m D Robert'L. Parnell 'Date 01/30/96 r s l 1 i I
THREE MILE ISIAtt jj . . SIKKATOR MANAGEMENP
..< CERTIFIChrza usT AnsTRscr g Test Identification: Sittulator Accuracy Certification Test Steady State Performance Test SSP 02 ANSI /ANS - 3.5 Reference (s): 4.1(2) '
4.1(3) - 4.1(4) 5.4.1(2) Benchmark B2.1 Test Date: 01/12/96 Test Init 4 al 4 ***4 em : FivumTGid Initial Condition IC-17 1004 Reactor Power
.. rm41N-ium Xenon End of Cycle Point of Test Tamina&4rvt: After umasurement of sinulator ocuputed values and principal mass / energy halanr= determinations were made at 100%, 80%, and 60% rated power.
Sjggdator Run T4==: 3 hours n===14ne Evaluation Data: Previous Certification Test SSP 02 dated 05/26/95. e , Reference Plant Operating Data (- [
\
Reference Plant Operating Prnemrinres Overall Test Results: SATISFACTORY l Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to perform within the accuracy requirement of plus/minus 2% for j critical pam _ /.ig. and plus/minus 10% for noncritical parameters pertinent to plant operation. Principal laass and energy balances shall be satisfied. Duri:v; conduct of this test sinulator dynamic operation was evaluated against ANSI /ANS 3,5 criteria. This evaluation was conducted by GPW personnel representing 1MI Sinulator Training. The evaluators have held an NRC Senior Reactor Operator license for 1MI Unit 1. The results of these evaluations are as follows: The results of this test are satisfactory, haamri upon the following: [ CRI1ERIA #1: l The sinnlator- omputed - values of . critical parameters shall agree within plus/minus 2% of the reference plant parameters and shall not detract from training. The parameters displayed on control panels may have the instrunent error added to the computed values. The sinulator supported this requirement with no deviations. , CRI1ERIA #2: The <almlated values of noncritical plant parameters pertinent to plant L__ _ _ _ _ _ '___
SSP 02 4
.?' within plus/minus 10% of the reference plant parameters and shall not detract fran training. 'Ihe parameters displayed on control panels may have the instrunant error added to the computed vedues.
'Ihe sintiator supported this requiianent with no deviations.
QE227tIA #3: Principal mass'and energy balances shall be satisfied at three different power levels.. The similator sgyad.ed this requirement with no deviations. Corrective Actions /Ccanants:' Nutt [ ) Ca pleted By '(13'
- IV/' / char,ma E. mmtad Date 02/20/96 JE wy F# - s...a >, o.eew>rge
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THREE MILE ISLAND SIMULATOR MANAGEMENT . CERTIFICATION TEST ABSTRACT Test Identification: Real Time Test ) RTT01 '
) ANSI /ANS - 3.5 Referenceis): Appendix A3(1)
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Test Dates: 06/13/96 I Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Terr.ination: This test was terminated following collection of all required data. Simulator Run Time: 2 hours simulator Evaluation Time: 10.5 hours Baseline Evaluation Data: Valve stroke times per controlled copies of applicable Reference Plant surveillance procedures. Known values for event timers. Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator computers to run in real time. s During the conduct of this test simulator dynamic performance was evaluated against ANSI /ANS 3.5-1985 criteria. These evaluations were conducted by GPUN [ - personnel representing TMI Simulator Training who hold or have held NRC Senior
\
v ) Reactor Operator licenses for TMI Unit 1. This test verified the capability of the simulator ccaputers to run in real time by measuring known values for selected valve stroke times and event timers. The correct valve stroke times were drawn from controlled copies of Reference Plant surveillance procedures. Surveillance procedure valve stroke times were used as reference points to validate simulator computer performance. There is no intention to exactly match reference plant procedurally required values. Known values for event timers were drawn from Reference Plant system design data. A calibrated stopwatch was utilized to time each item. Additionally, the computer system processing capabilities were measured to serve as an annual check on each processor's performance. The evaluation of this data serves as a computer system naanagement tool. + The results of this test are satisfactory with all valve stroke times and event timers corresponding to the reference data from the Reference Plant with minor differences (less than one second) due to the use of a manual stopwatch for timing. Computer processing results (spare times) were analyzed to evaluate the effectiveness of the configuration changes made this past year. Corrective Actions / Comments: There are no corrective actions required as a result of this test. l r l'/ Completed By #/ William A. Fraser Date: 06/19/96 Approved By Robert L.Parnell III Date: 06/19/96 (/ 'Supervi.sor, ' Simulator Training __2 _ _ _ --
.- THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT I 480 Volt Bus Fault Certification Test i Transient Test TTS15 ANSI /ANS 3.5 Referencels): 3.1.2(3) l Test Date: 02/06/96 Malfunctions) Tested *- ED06N 480 Volt Bus 1S Fault
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC-16 100% Reactor Power , Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated after collected data for ten minutes. Plant stability was i rnaintained throughout. j 1 Simulator Run Time: 10 minutes ' Simulator Evaluation Time: 1 hour Baseline Evaluation Data: Previous Certification Test dated 02/02/90. Right Direction Analysis. Current Controlled Copies of: [sT Reference Plant Operating Procedures y/ Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results
Description:
I in cccordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to j perf rm correctly following activation of a simulated malfunction. 1
- Th3 cvent was initiated by tripping the feeder breaker to 480 Volt Bus 1S, Various etauipment lost power as well as the
. downstream electrical buses. The current revision of reference plant operating procedure OP 1107-4 (Electrical i Distribution Panel Listing) was utilized to verify component responses. Overall plant stability was maintained. I' The transient was terminated after ten minutes of data collection and verification of component responses per OP 1107-4. i.
\
/'~'N "
E5-1 GPU Nuclear,Inc. ' t leddletown, PA
%# Copyr$ht 1997 ;
I I L_ -_
,,During th3 conduc* of this tist simulator dynamic response, annunciat:r operation and automatic safety syst m ' actuations were recorded.; Test results have been evaluated against ANSl/ANS 3.5 criteria by GPUN personnel representing TMt Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses f~ . [ TMI,.ance Unit with
- 1. Identified deviations Functional Fidelity from expected performance Procedure,6221-ADM-2820.02. have The results been of these evaluated evaluations forimpact are as follows: and correc U
The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference niant and do not violate the physicallaws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERi&.S . 1 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the ' alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this taquirement with two identified deviations.
)
Corrgctive Actions / Comments:
Description:
Incorrect Power Supply Number of Identified Deviations: Two Corrective Actions: Two Discrepancy Report Corrections J Completeo by: # William A. Fraser d Dite Approved by: ,. Supervisor, Simulator Training Robert Pamell a/a 5/94 l Dat6 i-1 ES-2 GPU Nuciew. h:.
' leddletown PA coprneht 1997 L
y,. el THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Idagification: 480V MCC Fault Certification Test Transient Test f TTS16 ANSI /ANS 1.5 Reference (cli 3.1.2(3) Test Date: 02/06/96 Malfunction (s)' Tested:* ED0726 1AES MCC (480 1P) Fault
- Refer to Malfunction Cause and Effects Documents t for options available. I Test Initialization: Protected Initial Condition IC-15 1 j
100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated after verifying that overall plant stability was maintained. The 1A-ES 480 V MCC was de-energized along with its associated loads. Shpulator Run Timut.;, 10 minutes i i S4=+1= tor Evaluttion T4==: 2 hours Baselin. Evaluation Data: Previous Certification Test dated 02/21/90. Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures j Reference Plant Alarm Response Procedures Overall Test Rasults: SATISFACTORY J. I Results Deseriotion: ' In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator ' to perform correctly following activation'of a simulated malfunction ^. This transient resulted in the loss of the 1A ES 480V Motor Control Center due All associated electrical loads were de-
~
to its feeder breaker tripping. energized with identified exceptions. Overall plant stability was maintained. l The current revision of reference plant operating procedure OP 1107-4 (Electrical Distribution Panel Listing) was used to verify component responses. The plant remained at 100% power throughout this transient test.
TTS16 V The transient was terminated af ter verifying plant stability and component responses. 3 During the conduct of tM9 test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. evaluated' against ANSI /ANS Test results have been 3.5 criteria by GPUN personnel representing TMI Simulator Training. Operator licensesThe for TMI TMIrepresentatives Unit hold or have held NRC Senior Reactor ,
- 1. {
Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADH-2820.02. evaluations are as follows: The results of these ; The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied.this requirement with no identified deviations. CRITERIA #2: 1 k The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the . alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if l the teference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with four identified deviations. j Corrective Actions /C. nts; ' Descrintiont Improper Component Operation on Power Loss i Nn=her of Identified Deviations: Four Corrective Actions: One Discrepancy Report correction 1; Completed By *
- William A. Fraser Date 02/23/96 Approved By Robert L. Parnell Date N a I b Supervisor, Simulator-Training i -
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1 1 O e i
< 1 e
f THREE MILE ISLAND
,r] SIMULATOR MANAGEMENT l
i ; CERTIFICATION TEST ABSTRACT b/ , Test Identification: ICS Auto Power Failure Certification Test Transient Test TTS17 ANSI /ANS 3.5 Reference (s): 3.1.2(3) Test Date: 03/01/96 Malfunction (s) Tested:* ED13 ICS Auto Power Fa11ure
- Refer to Malfunction Cause and Ef fects Documents for options available.
Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with all Integrated Control System stations in manual, maintaining plant stability. Reactor Coolant System pressure
,7m. and pressurizer level were continuing to rise. ? i
( ,/ Simulator Run Time: 10 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Tests dated 02/10/90, 04/10/92. Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Emergency Procedures Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following j activation of a simulated malfunction. ) 1 This transient resulted in a loss of ICS Auto Power which normally supports the plant's automatic mode and provides power for various instrumentation. The ICS j automatically switched all control stations to HAND and maintained the last automatic output demands. This allowed the plant to remain stable. The MU-V-17 q I ',__. (RCS Makeup Control Valve) switched to hand while partially open causing l ( pressurizer level and thereby RCS pressure to rise. The cutrent revision of i s / 1
C* l TTS17 l
\
V) reference plant emergency procedure EP 1202-42 (Total or Partial Loss of ICS/NNI Auto Power) was used to verify component responses and actuations. A significant number of console indications lost power. The transient was terminated af ter a period of ten minutes to verify plant stability and verification of components lost per emergency procedure EP 1202-42. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have'been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6211-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond ~ in direction to those expected from the actual reference plant and do not violate the physical laws of nature. I The simulator satisfied this requirement with no identified deviations.
/
,%.s CRITERIA #2:
The simulator shall not fail to cause an alarm or automatic action if the' reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation. Corrective Actions / Comments:
Description:
Incorrect Instrument Response to Loss of Power. A0426 Number of Identified Deviations: One Corrective Actions: Discrepancy Report Correction Completed By [ [ C. William A. Fraser Date 05/31/96 Approved By - Robert L. Parnell III Date $bb4~ 4 Supervisor, Simulator Training I b - V _ __ _ o
y r THREE MILE ISLAND i ' SIMULATOR MANAGEMENT
/
CERTIFICATION TEST ABSTRACT Test Identification: Inverter Failure Certification Test Transient Test TTS18 ANSI /ANS 3.5 Referenceis): 3.1.2(3) Igst Date: 02/07/96 Malfunction (s) Tested:* ED09B Inverter 1B Failure
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated following verification of plant instrument and c;ntrol system responses. (
,,_ Simulator Run Time: 10 minutes l
\'a/ Simulator Evaluation Time: I hour '
laseline Evaluation Data: Previous Certification Tests dated 02/12/90,
, 04/13/92.
Right Direction Analysis. Current Controlled Copies of: ' Reference Plant Operating Procedures
, Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY
?esults
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. Initiation of the malfunction resulted in the loss of Vital Instrumentation Bus 1B. With identified exceptions, all plant instrumentation and control functions powered from this bus were affected. The event was initiated by failing Inverter IB, which caused the 120 Volt AC Vital Bus 1B to be~de-energized. The loss of this bus caused a loss of power to various plant instruments and control functions. i One channel each of ESAS and RPS actuated due to loss of power. Current l f'~' revisions of reference plant operating procedures (OP 1107-2 Emergency Electrical _( x
t t TTS18 /% b System, and OP 1107-4 Electrical Distribution Panel Listing) were used to verify simulator instrumentation and dynamic responses. The test was terminated after ten minutes, after instrument and control system responses to the loss of power bad occurred. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The ~MI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected ( performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these ( evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1:
' The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
The simulator satisfied this requirement with no identified deviations. k CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the re ference , plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with three identified deviations. Corrective Actions / Comments:
Description:
Incorrect Replica Plant Process Computer Alarm Operation Number of Identified Deviations: Two Corrective Actions: One Discrepancy Report correction
Description:
Incorrect instrument Response Number of Identified Deviations: one Corrective Actions: One Discrepancy Report correction Completeo By /M#M william A Fraser Date 3/5/96 Approved By ' ~ x Robert L. Parnell Date'7/7/N ' Supervisor, Simulator Training ( \ 1 b
)
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.I 1 (4
l [ THREE MILE ISLAND gg SIMULATOR MANAGEMENT
! ) CERTIFICATION TEST ABSTRACT '% )
Test Identification: Loss of Condenser Vacuum Certification Test Transient Test TTS20 ANSI /ANS 3,5 Refgrence(s): 3.1.2(5) Test Date: 03/01/96 Malfunction (s) Tested:* FW35 Main Condenser Vacuum Leak 100% (100% = 3500 CFM) Ramp - 240 seconds
- Refer to Malfunction Cause and Ef fects Documents for options available.
Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the turbine and
-~s reactor tripped, at stable Hot Shutdown with RCS
[ heat removal via the atmospheric dump valves and
\s_,/} emergency feedwater, ,
i Simulator Run Time: 20 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Tests dated 02/12/90, 05/13/92. . Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies.of: Reference Plant Alarm Responses Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to i demonstrate. the ability of the simulator to perform correctly following L activation of a simulated malfunction. l ) This transient resulted in the loss of Main Condenser vacuum and subsequent turbine and reactor trip, with Emergency Feedwater actuation and heat removal via the Atmospheric Dump Valves.
,A i uns l
l l l L_ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ . _
.. i
+
TTS20 2-
\ h V The event was initiated by simulating a vacuum leak in the Main Condenser and increasing the severity to 100% over a 240 second period. Main Condenser vacuum began to decrease, causing a reduction in turbine efficiency. Generated megawatts started to d-..aase. The Integrated Control System sensed the reduction in load and incraa .d reactor power and Feedwater flow in an attempt to recover the lost megawatts. The turbine tripped on low vacuum which produc-d a reactor trip and Reactor Trip Containment Isolation. Due to the increased ' ak pressure in the Main Condenser, the Auxiliary Condensers for the Main Feed Pumps filled with .
condensate, covoring the tube bundles. This caused the Main Feed Pumps to trip and the Emergency Feedwater System to actuate and establish feed to the OTSGs. Coincident with the turbine trip, control for steam pressure transferred to the Atmospheric Dump Valves and the Turbine ~ Bypass Valves were latched closed (by interlock) to prevent pressurizing the Main Condenser. At the end of the test the plant was stable at Hot Shutdown with Emergency Feedwatet maintaining OTSG levels and the Atmospheric Durp Valves maintaining OTSG pressure. The test was terminated following verification of Emergency Feedwater control of OTSG level and pressure control by the Atmospheric Dump Valves. During the conduct of this test simulator. dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI [m} Simulator Training. The TMI representatives hold or have held NRC Senior Reactor (f Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure , 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. \ O i qj
)' -
.( 'l..,* -
TTS20 3 Corrective Actions /Co - nts: . NONE' i; Completed By . vv - /h William A. Fraser .Date 03/14/96,
' ~
-Approved By R L NMLc e- Date' 7/h 7/
Supervisor, Simulator Training
+
1 t 4-
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4 9 A 9
1 . THREE MILE ISLAND
,, SIMULATOR MANAGEMENT /h CERTIFICATION TEST ABSTRACT N ']
Iest Identification: Condenser Level Control Failure Certification Test Transient Test TTS21 ANSI /ANS 3.5 Reference (s): 3.1.2(5) Test Date: 03/01/96 Malfunction (s) Tested:* FW23 Condenser Reject Valve Fails Open (CO-V-6)
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated af ter verification of plant stability and 'Hotwell Level Control
,_, response.
( ) i Simulator Run Time: 20 minutes x_ - Simulator Evaluation Timg1 30 minutes Baseline Evaluation Data: Previous Certification Test dated 03/26/90, 04/13/92. Right Direction Analysis. ' Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant P & ID Prints Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. ! This transient resulted in a decreased level in the Main Condenser Hotwell. Normal Makeup and Emergency Makeup flow stabilized Hotwell level at a final lower value. . The event was initiated by failing open Condensate Reject Valve CO-V-6 which I
/'^N directed Condensate flow back to the Condensate Storage Tanks. Main Condenser i
LJ \ \
. .s -
TTS21 (U ) Hotwell level decreased. The normal Makeup Valve Co-V-7 opened to attempt to recover level but its capacity is less than the Reject Valve capacity. Level continued to decrease until the Emergency Makeup Valve Co-V-8 opened to maintain the Hotwell level at a lower value. Slight differences between previous tests and this test exist. A higher main condenser level is maintained. Due to a , higher overall Main Feedwater flow rate and subsequent lower Condensate pump discharge pressure, CO-V-6 and CO-V-7 exhibit a slightly lower flowrate. A less stable Feedwater system made it difficult to quantify the impact of the CO-V-6 opening. The transient was terminated following verification of plant stability and Hotwell level control response. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been l evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI l Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: [] CRITERIA #1:
\
v The obse rvable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE Comp eted By // F William A. Fraser Date 03/18/96 Approved By ( + Robert L. Parnell Date N/ f/N n Supervisor, Simulator Training I ! h
)
Q.y* l l THREE MILE ISLAND SIMULATOR MANAGEMENT
/ CERTIFICATION TEST ABSTRACT U Test Identification: Loss of Service Water Certification Test Transient Test
., TTS22 ANSI /ANS 3.5 Reference (s): 3.1.2(6)
Test Date: 03/01/96 Malfunction (s) Tested:* RWOI A/B/C Secondary river Water Pump Trip A/B/C
- Refer to Malfunction Cause and Effects Documents for options available.
-Test Initialization: , Test Initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated after a main turbine-generator high temperature Stator Cooling runback occurred, i Simulator Run Time: 53 minutes -
)
a Simulator Evaluation Time: I hour Baseline Evaluation Data: Previous Certification Tests dated 02/22/90, 11/05/91. Right Direction Analysis. Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACIQB1 Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a high temperature Stator Cooling runback due to the loss of the Secondary River Water pumps. The event was initiated by tripping SR-P-1A/B/C to cause a loss of ecoling to the Secondary Closed Cooling Heat Exchangers. The loss of flow caused the Secondary [ Closed Cooling System temperature to increase. The temperatures of affected i components began to rise, tracking the increasing SCCW temperature. The plant remained stable at 100% power until a high Main Generator Stator Water Cooling ("~N outlet temperature interlock actuated a protective Stator Cooling runback which
*/s (
l ! TTS22 . O
- (] automatically transferred the main turbine to manual control and reduced generator output.
I The test was terminated af ter the initiation of the Stator Cooling runback. During the condugt of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI 1 Simulator Training. The TMI representatives hold or have held NRC Senior Reactor . j Operator licenses for TMI Unit 1. Identified deviations from expected- ' performance have been evaluated for impact and corrective action.in accordance with Functional . Fidelity Procedure , 6221-ADM-2820,.02. The results of- these evaluations are as follows: i The results of this test are satisfactory, based upon the following: { CRITERIA #1: ! The observable changes in simulator parameters shall correspond in i direction to those ' expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. l O CRITERIA #2: k l The simulator shall not fail to cause an alarm or automa. tic action if the j reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic' action. The simulator satisfied this requiremen't with no identified deviasions. Corrective Actions /Co - nts: NONE Completed By - M illiam A. Fraser Date 03/18/96
- Approved By Robert L.'Parnell Date 3 Supervisor, Simulator Training I
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O ' V 1 _ ____ _____ -__A
THREE MILE ISLAND jk)y SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT
/N
' Test Identification: Loss of Shutdown Cooling Certification Test j
Transient Test TTS23 V ANSI /ANS - 3.5 Referencets): 3.1.2.(7) Test Date: 01/26/96 Malfunction (s) Tested: DH01A Decay Heat Pump Trip 1A ;
- Refer to Malfunction Cause and Effects Document for options available
' Test Initialization: Protected Initial Ccndition IC-03 Cold Shutdown, RCS Filled' Steam Bubble Pre-heatup, End of' Cycle Point of Test Termination: This test was terminated with the reactor in cold shutdown, no active means of decay heat removal in operation and.a slow heatup in progress.
Simulator Run Time: 12 minutes simulator Evaluation Times I hours O Y [ l Baseline Evaluatio'n Deta: Previous Certification test dated 02/06/90
\
\~' Right Direction Analysis. .
i Simulator Malfunction Cause and Effects Document. Current Controlled Copies'of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY Enaults
Description:
1
- n accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction.
This transient resulted in a loss of all active means of reactor core decay heat removal. Heat transfer from the Reactor Coolant System was-via loss to ambient and temperature was slowly increasing. The event was initiated by failing the operating Decay Heat Removal Pump. The forced l L flow of water through the core was stopped and natural circulation flow was being established. Without forced flow through the decay heat removal heat exchanger reactor i !. coolant hot leg temperatures increased resulting in the establishment of natural ) l circulation flow in the Reactor Coolant System. There were no automatic actuations. The transient was. tenminated when a heatup trend was established. i i During the conduct of this test simulator dynamic response, annunciator operation, and l automatic safety system actuations were recorded. Test results have been evaluated : [ against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. I The TMI representatives have held an NRC Senior Reactor Operator license for TMI Unit
\s_, 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6211-ADM-2820.02.
The results of these evaluations are as follows:
TTS23 ' 'i Tha results of this tsst are satieftetory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those laws ofexpected nature. from the actual reference plant and do not violate the physical The simulator satisfied this requirement with no identified deviations. CRITERIA 2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions /Ca== ants: none A-' Completed By ,_ Charles E. Husted Date 05/14/96 Approved By a Robert L. Parnel'1 Date Y. Eupervisor, Simulator Training ' N I l l-
- t. = /*
THREE MILE ISLAND
,_s SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: , Loss of Component Cooling Certification Test Transient Test TTS24 ANSI /ANS 3.5 Reference (s): 3.1.2(8)
Test Date: 02/15/96 Mal functJ9n( s ) Tested:* CC03A/B/C Nuclear Se rvices Closed Cooling Pump Trip A/B/C
- Refer to Malfunctiori Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated after the reactor tripped on loss of all Reactor Coolant Pumps and Emergency Feedwater had initiated. ( Simulator Run Time: 9 minutes L S_imulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Tests dated 02/19/90, 05/28/92. Right Direction Analysis. Current Controlled Copies of: Reference Plant Emergency Procedures Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results Descrintiorn In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a loss.of Nuclear Service Closed Cooling water flow. Reactor Coolant Pump temperatures increased rapidly. Motor conditions actuated i RC pump breaker trips which caused a reactor trip and Emergency Feedwater actuation. The event was initiated by tripping NS-P-1A/B/C which stopped cooling flow to major reactor support equipment, notably the Reactor Coolant Pump motors and oil L/ L___-_-__-__-_--_. _ _
et
.^
TTS24 coolers and the operating Makeup Pump. Temperatures on all Reactor Coolant Pump bearings and stators began to rise rapidly due to the loss of cooling water flow. All RC Pump vibrations and eccentricities went into alarm condition. The Reactor Coolant Pumps tripped on overload, initiating a reactor trip. Reactor Trip Containment Isolation actutated and Emergency Feedwater actuated to begin raising OTSG 1evels to 50% to establish natural circulation cooling. The test was terminated af ter the reactor trip and Emergency Feedwater actuation had occurred. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit s
- 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional ' Fidelity Procedure, 6211-ADM-2820.02. The results of these evaluations are as follows:
The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in ( p) direction to those expected from the actual reference plant and do not (_/ violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. I CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with five identified deviations. Corrective Actions / Comments:
Description:
Replica Plant Process Computer Alarms L2910, L2208, l L2209,L2211, C4018 did not actuate as expected. Number of Identified Deviations: One l Corrective actions: None - Memo 6211-96-0018 Completed By /
- n Hliam A. Fraser Date 05/20/96 Approved By I -[ . M A D G !~ Date D I Supervisor, Simulator Training j
p , i THREE MILE ISLAND SIMULATOR MANAGEMENT l , . j CERTIFICATION TEST ABSTRACT l; t. j '
- Test Identification:
Loss of Normal Feedwater Certification Test L Transient Test .
.TTS25 l' 3 ( nar/u s - 3.5 Referencets): 3.1.2.(9)
Test Date: 02/01/96 l- , Malfunction (s) Tested: FW15A'B
/ Main Feedwater Pump Trip 1A/1B Test Initialization: . Protected Initial Condition IC-15 100%-Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was~ terminated with the reactor tripped, Reactor Coolant System pressure and temperature stable and OTSG level controlled by Emergency Feedwater System at setpoint.
simil ator Run Ti==: . 10 minutes l
. Si=dator Evaluation T4=: 3 hours l' k Baseline Evaluation Data: Previous Certification test dated 02/15/90 i
.Right Direction Analysis.
i Simulator M41 function Cause and Effects Document, Current Controlled Copies of: Reference Plant-Alarm Response Procedures Reference Plant Emergency Procedures ' l overall Test Results: SATISFACTORY l- Results
Description:
-In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of E a simulated malfunction.
i This transient resulted in a reactor trip due to a loss of both Main Feedwater Pumps. Heat Sink Protection System actuation started Emergency Feedwater System operation and OTSG conditions stabilized.
' The ~ event was initiated by f ailing both Mah Feedwater Pumps. The Reactor Protection System detected the' loss of both Main Feedwater Pumps and initiated Reactor Trip, Main Turbine tripe and Reactor Trip Isolation of the Containment. The Beat Sink Protection
. System. actuated to initiate Emergency Feedwater start-up of all three pumps. SASS actuated a mismatch indication for Turbine Header pressure.
~'evelt L in the OTSGs decreased to the normal post trip level and was maintained at setpoint by the Emergency Feedwater Systemi Normal full power steady state OTSG level was reduced since. conducting the initial test resulting in Emergency Feedwater flow starting sooner and maintaining-a higher flow for a longer period of time. Reactor
=
i Coolant System pressure returned to. normal and pressurizer level was increasing to
.setpoint. , ,
The transient was terminated when OTSG 1evel was being maintained at setpoint using the
' Emergency Feedwater System.
s 4
y- .
. .- - TTS25 ' (
"L During th2safety automatic conduct of this system toct siEulator actuations weredynamic recorded. rssponse, annunciator operation, and '
Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator license for TMI Unit
- 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6211-ADM-2820.02.
The.results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA 81: . l The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. ' The simulator satisfied this requirement with no identified deviations. CRITERIA 2: l The ' simulator shall not fail to cause an alarm or automatic action if the - reference plant would have caused the alarm or automatic action, and conversely, I the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with two identified deviations. 3 l Corrective Actions /C- nts j
Description:
Inco.irect Replica Plant Process Computer Alarms
~{N Number of Identifiad Deviations: Two Corractive Actions: One Discrepancy Report correction and removed non-simulated point from monitor.
Completed By Charles E. Husted Date 05/13/96 Approved By - Robert L. Parnell Date f 3 bb' Supervisor, Simulator Training 1 I - 1 i t
l .- l l THREE MILE ISLAND SIMULATOR MANAGEMENT
' }i CERTIFICATION TEST ABSTRACT Test Identification:
Feedwater Flow Transmitter Failure Certification Test Transient Test TTS26 ANSI /ANS 3.5 Reference (s): 3.1.2(9) Test Date: 02/23/96 Malfunction (s) Tested:* FWO2B Loop A Main Feedwater Flow Transmitter Failure 0% (100% = 6,000,000 LBM/HR) , l i
- Refer to Malfunction Cause and Effects Documents for options available. l j
Test Initialization: Protected Initial Condition IC-17 100% Reactor Power j Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated after restoration of l
/3' plant stability.
k Simulator Run Time: 10.0 minutes Simulator Evaluation Time: 0.5 hours Baseline Evaluation Data: Previous Certification Tests dated 03/22/90, ! 03/17/92 Right Direction Analysis. Simulator Malfunction Cause end Effects Document. Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to l demonstrate the ability of the simulator to perform correctly following j activation of a simulated malfunction. ' This transient resulted in the loss of the selected Feedwater flow transmitter for Feedwater Loop A. This caused a significant loop and total Feedwater flow error signal to be generated instantaneously to the Integrated Control System. The system became Cross Limited and reduced reactor power slightly. The total flow error signal boosted Main Feedwater flow momentarily. The SASS System l I I sensed the failure of the transmitter and selected the alaternate transmitter for )
,e ICS input. The Cross Limit then cleared and the plant stabilized at 100% reactor v
)
v' TTS26 (Q power. Two computer alarms (L2697, L2701) did not acutate apparently due to the combination of the SASS swtich actuation (approx.1/2 second) and the one second scan rate of the RPPC points. The test was terminated following restoration of plant stability. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior R,eactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: - The results of this test are satisfactory, based upon the following: CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. O
\ I CRITERIA #2:
The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and i conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: j l The results of this test were satisfactory with no deviations from expected response. There is no corrective action as a result of this , test. Completed By William A. Fraser Date 03/07/96 Approved By
/2x4 h Robert L. Parnell Date N' '7/ NI Supervisor, Simulator Training
,Q
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(= , , J THREE MILE ISLAND g SIMULATOR MANAGEMENT D) CERTIFICATION TEST ABSTRACT E Test Identification: Loss of Protective System Channel Certification Test Transient Test TTS28
-ANSI /ANS 3.5 Reference (s): 3.1.2(11)
Test Date: 03/01/96 Malfunction (s) Tested:* NI26C Loss of RPS Channel - C
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition 1C-15 100% Reactor Power Equilibrium. Xenon Beginning of Cycle Point of Test Termination: This test was terminated with Reactor Protection Channel C de-energized and tripped. Channel C of [m Reactor Trip Isolation was actuated. N.)} Simulator Run Time: 10 minutes Simulator Evaluation Time: 0.75 hours Baseline Evaluation Data: Previous Certification Tests dated 02/08/90, 03/17/92. Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Resyl3;p.1 SATISFACTORY Ee sulta ' Description :
.In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to l
demonstrate the,' ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of a reactor' protection channel. The event was initiated by failing power to the C Channel of the Reactor Protection System. As designed, the channel failed in the tripped state placing the system in a one out of three trip logic. Channel C of Reactor Trip Isolation actuated but no components actuated as a two of three logic must be satisfied. _ _ . _ _ ___.__________.____._______________________J
c 4 TTS28 ,O The transient was terminated after verification of plant stability and a check of the affected RPS cabinet for proper loss of power indication. During the conduct of this test simulato: dynamic response, annunciator operation and automatic safety system actuations were recorded. Test resulta have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor i Operator licenses for TMI Unit 1. Identified deviations from expected i perfomance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: l The results of this test are satisfactory, biased upon the following: l CRITERIA #1: The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2:
\j The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
The simulator satisfied this requirement with no identified deviations. Corrective Aqtions/ Comments: NONE Completed By '
#' I William A. Fraser Date __03/18/96 Approved By - / . NMM Date ., / /
Supervisor, Simulator Training i. k V
~
r% .
- THREE MILE ISLAND
/'~i SIMULATOR MANAGEMENT '
L 5 CERTIFICATION TEST ABSTRACT Test Identification: Stuck Control Rod Certification Test Transient Test TTS29 ANSI /ANS 3.'5 Reference (s): 3.1.2(12) Test Date: 03/01/96 Malfunction (s) Tested:* RD0225 Stuck Rod #25
- Refer to Malfunction cause and Effects Documents for options available.
Test Initialization: Test Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor at s 87.5% power, one control rod stuck at 92.5% and rod control in automatic. s : Simulator Run Time: 14 minutes Simulator Evaluation Timel I hour , i Baseline Evaluation Data: Right Direction Analysis. Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Emergency Procedures Reference Plant Alarm Response Proceduras Reference Plant Operating Proa.edures Previous Certification Tests 02/09/90, 05/22/92. dated Qyerall-Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conduc demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This the' core. transient resulted in a decreased power level and an induced flux tilt in
/N 1u) .
o - _ - _ - _ - - -
a
- TTS29 ,,
(.I The event was initiated by sticking rod number 25 located in Rod Group Seven, Rod Number Six near NI-6 power range detector. Rod control was placed in manual and rods were inserted from 93% to 84%. This caused asymmetric rod alarms and indications. Rod control was returned to automatic. An Out-Inhibit on the Rod Control System prevented any rod withdrawal while in automatic. Power was then reduced to 717 megawatts. Due to .ut-motion by Control Rods being prevented, plant stability was maintained with Cross Limits actuation. This test. differed from previous tests in that the Unit Load Demand potentiometer setpoint was reduced from 10 percent to O percent (2 megawatts / minute) in the Initial Condition. This allowed the system response to be more stable. The test was terminated following restoration of plant stability. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2826.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following:
\
f'~'s CRITEEIA #1:
) J The observable changes in simulator parameters shall correspond in I direction to those expected from the actual reference plant and do not l violate the physical laws of nature.
The simulator satisfied this requirement with no identified deviations. CRITERIA #2: The simulator shall not fail to cause an alarm or automatic action if the reference plant would have m sed the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE ! / ! Completed By
- &- 4# William A. Fraser Date 04/08/96 Approved By /k . /hudi Date N T4 em Supervisor, Simulator Training
/ %
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pr k THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: RB 30 PSIG Analog Channels Surveillance Certification Test Normal Operations Test NOT08 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(5) 3.1.1(10) Test Date: 02/08/96 Reference Plant Pr >cedure: SP 1303-4.14, RB 30 PSIG Analog Channels Surveillance Revision 24 Test Initialization: . Protected Initial Condition IC-09 Hot Eero Power IE-8 Amps Xenon Free End of Cycle Point of Test Termination: This test was terminated following completion of surveillance. Simulator Run Timmi 1 hour > Baseline Evaluation Data: ' Tactile Feel by Experienced Operator. Current Controlled Copy of: Reference Plant Surveillance Procedure Reference Plant Alarm Response
. Procedure Overall Test Results:- SATISFACTORY Results
Description:
..b In accordance with ANSI /ANS-3.5-1985 this certification test is '
l conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the referenced plant. During the condupt_' of this test simulator dynamic performance, annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls
. used during the test was also evaluated. These evaluations were conducted by GPUN personnel representing TMI Simulator Training.
All the evaluators have held an NRC Senior Reactor Operator license for TMI- Unit 1. -Procedure steps not performed and identified deviations from expected performance have been evaluated for impact
, U r
NOT08 l l and corrective action in- accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1:
' The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test.
The simulator supported this requirement with no deviations. CRITERIA #2: 1 The observable changes in simulator parameters correspond in j direction to those expected from the actual reference plant and do not violate the physical laws'of nature. The simulator supported this requirement with no deviations. CRITERIA 53: The simulator shall not fail to cause an alarm or automatic actuation-if the reference plant would have caused an alarm or automatic . action, and conversely, the simulator shall not . cause an alarm or automatic action if the reference plant i would not cause in alarm or automatic action. The simulator supported this requirement with no deviations.
~
CRITERIA 44: Tactile feel of Plant-Referenced Simulator control device (s) compares to-those of the reference plant. The simulator supported this requirement with no deviations.
,7 Corrective Actions /C-- nts: 1 During the conduct of this test, reference plant procedure steps impact and not performed were identified and evaluated for corrective action requirements in accordance with Functional 6221-ADM-2820.02. The results of these Fidelity Procedure, evaluations are as follows:
'. 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to
% proceed with the evolution.
. 2. Steps not performed did not result in observable r
~
1 NOT08 g differences in the control room.
- 3. Steps not performed did not prevent the successful completion limits of the procedure in accordance with plant and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6211-ADM-2820.02, these test exceptions have been documented and closed.
l l
.gy, l
l Completed By Charles Hunted Date 02/14/96 Approved By Robert L. Parnell Date 3f /J9#II, a Supervisor, Simulator Training 1 (
1 ? r.
- THREE MILE ISLAND SIMULATOR MANAGEMENT
% CERTIFICATION TEST ABSTRACT l l
Test Identification: Main Stean Isolation Valves Surveillance Certification Test
- Normal Operations Test NOT09 ANSI /ANS'3.5 Reference (s)
- 4.2.1 3.1.1(5) 3.1.1(10)
Test Date: 06/10/96 I Reference Plant Procedure: SP 1303-11.22, Main Steam Isolation f valves Surveillance Revision 13 ) Test Initialization: Protected Initial Condition IC-08 Hot Zero Power Xenon. Free Startup . Middle of Cycle - Point of Test Termination: This test was terminated following completion of surveillance. (} Simulator Run Time: 20 minutes ! \- Baseline Evaluation Data: Tactile Feel by Experienced Operator. I Current Controlled Copy of: Reference Plant Surveillance Procedure Reference Plant Alarm Response 3 Procedure I Overall Test Results: SATISFACTORY l Results
Description:
l In accordance with ANSI /ANS-3.5-1985 this certification test is L conducted to demonstrate the ability to operate the simulator in l 'accordance with similar reference plant operating procedures, using L only operator action ncrmal to the referenced plant. During the conduct of this test simulator dynamic performance, annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls i used during the test was also evaluated. These evaluations were conducted by GPUN personnel representing TMI Simulator Training. All the evaluators have held an NRC Senior Reactor Operator license for . TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact t and corrective action in accordance with Functional Fidelity
.N[-Y l
NOT09 f3 ( ' Procedure 6221-ADM-2820.02. The results of these evaluations are
\d } as follows:
The results of this test are' satisf actory, based upon the 1 following: j j CRITERIA #1: The simulator has met the acceptance criteria of the reference j plant procedure (s) used during the' conduct of this test. The simulator supported this requirement with one deviation. l CRITERIA #2: 1 T h e o b s e r v a b l e c h a n g e s i n s i m u l a t o r' p a r a m e t e r s c o r r e s p o n d i n ; diraction to those expected from the actual reference plant { and do not violate the physical laws of nature. The simulator supported this requirement with no deviations. CRITERIA #3: i The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or
; /- s
! $ {
automatic . action, and conversely, the simulator shall. not b cause an alarm or automatic action -if the reference plant would not cause an alarm or automatic action. { The simulator supported t.his requirement with no deviations. J i CRITERIA #4: ] Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator supported this requirement with no deviations. Corrective Actions / Comments: 1 During the conduct of this test, reference plant procedure steps ) not performed were identified and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. .The results of these evaluations are as follows: t i
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps. not performed did not result in observable
/. differences in the control room.
i. i NOT09 h' 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant - limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2.820.02, these test ;
exceptions have been documented and closed.
, )
Corrective Actions / Comments:
Description:
Incorrect System Response Number of Identified Deviations: One Corrective Actions: Setpoint change request to change stroke time to 90 seconds. l 1 0 1
)
1
-l 1
I 1 1 Completed By Charles Husted Date 06/10/96 Approved By Robert L. Parnell D' ate b /# I Supervisor, Simulator Training \.
l l { THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Main Steam Isolation Valves--Partial Stroke Surveillance certification Test Normai Operations Test NOT10 ANSI /ANS 3.5 Reference (s)- 4.2.1 3.1.1(10) Test Date: 01/17/96 Reference Plant Procedure: SP 1303-4.17, Main Steam Isolation Valves--Partial Stroke Surveillance Revision 14 Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon i End of Cycle Point of Test Termination: s This test was terminated following (- completion of surveillance. I f Simulator Run T4== 15 minutes Baseline Evaluation Datar Tactile Feel by' Experienced Operator. - Current Controlled Copy of: i Reference Plant Surveillance Procedure Overall Test Results: SATISFACTORY Results Descrin_ tion l In accordance with ANSI /ANS-3.5-1985 this certification test is ' L conducted'to demonstrate the ability to operate the simulator in l accordance.with similar reference plant operating procedures, using only operator action normal to the referenced plant. During. the conduct of this test simulator dynamic performance,
~
annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls used during the fest was also evaluated. These evaluations were conducted by GPUN personnel representing TMI. Simulator Training.
, All the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified f deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are i __ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _
__j
w . , NOT10 . as follows: The results of this test are satisfa'ctory, based upon the following: CRITERIA #1: The simulator has. met the accepte.nce criteria of the reference l plant procedure (s) used during the conduct of this test. . The simulator supported this requirement with no deviations.
- CRITERIA _j2,,1.,
- The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
The simulator supported this requirement with no deviations. CRITERIA #3: The' simulator shall not fail to cause an alarm or automatic
- ' actuation if the reference plant would have caused an alarm or automatic action, and conversely, the. simulator shall not ca/se. an alarm or automatic - action if the reference plant would not cause an alarm or automatic action.
I The simulator. supported this requirement with no deviations ~. CRITERIA #4: , Tactile feel'of Plant-Referenced Simulator control device (s)
. compares.to those of the reference plant. -
The simulator supported this requirement with no deviations. l Correctiva- Actions /C. nts: . ! t During therconduct of this test, reference plant procedure steps I L not performed were identified and evaluated for impact and l l corrective action requirements in accordance 'with Functional . I Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test' operator to violate the procedure in order to proceed with the evolution. ,
- 2. Steps not performed did not re'sult in observable differences in the control room. .
3
. \
NOT10 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits ' and praeautions, technical specifications or procedure acceptance criteria. . 4. In accordance with 6211-ADM-2820.02, these test ! exceptions have been documented,and closed. l l I i i l l e r j t Completed By I- . charles Husted Date , b, I Approved By Robert L. Parnell Date N
' Supervisor , Simulator Training a
4
- v. >
THREE MILE ISLAND SIMULATOR MANAGEMENT, CERTIFICATION TEST ABSTRACT Tast Identification: Shift and. Daily Checks Surveillance Certification Test Normal Operations Test NOT11
~
ANSI /ANS 3.5 Reference (s): 4.2.1
, 3.1.1(10)
Test Date: 04/04/96 Reference Plant Procedure: SP 1301-1, Shift' and Daily Checks Revision 118 Test Initialization: Protected Initial 1 Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This. test was terminated following collection'of specified data. Simulator Run' Time: 4.5 hours Baseline Evaluation Data: Tactile Feel by Experienced
. Operators.
Right Direction Analysis. l Current controlled Copies of:
-Reference Plant. Surveillance Procedures Overall Test Results- SATISFACTORY Results
Description:
In' .accordance with ANSI /ANS-3. 5-1985 this certification test is L ' conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using l only operator action. normal to the referenced plant. During' the conduct. of this' test simulator dynamic performance, L annunciator _ operation and' automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria.. Tactile feel'of simulator controls used during.the' test was also evaluated. These evaluations were conducted by ~ select < GPUN personnel' representing TMI Simulator i Training. All the evaluators- have held an NRC Senior Reactor Operator license for TMI Ur.it 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these
. evaluations are-as.follows:
i O\ t d'.
1 I NOT11 l t The results of this test are satisfactory, based upon the folleving: < l l
-CRITERIA #1:
The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. 1 CRITERIA 1 evaluators concurred that during this test, the simulator supported this requirement with no deviations. CRITERIA #2: i The observable changes in simulator parameters correspond in I direction to those expected from the actual reference plant and do not violate the physical laws of nature. CRITERIA 2 evaluators concurred that during this test, the simulator supported this requirement with no deviations. 4 CRITERIA #3: The simulator shall not fail to cause an alarm or automatic g actuation if the reference plant would have caused an alarm or f automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would.not cause an alarm ~or automatic action. CRITERIA 3 evaluators concurred that during this test, the simulator supported this requirement with no deviations. CRITERIA #4: Tactile feel of Plant-Referenced Sinulator control device (s) I compares to those of the reference plant. CRITERIA 4 evaluators concurred that during this test, the , simulator supported this requirement with no deviations. j i Corrective Actions / Comments: l During the conduct of this test, reference plant procedure steps not performed were identified and evaluated for impact and l-corrective action. requirements in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: l
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
O Lb .
NOT11 2. Steps not performed did not result iis observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of . the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6211-ADM-2820.02, these test exceptions have been documented and closed.
l I l l l l l Completed By [8 ^^
- n. A. Fraser Date 04/04/96 Approved By Robert L. Parnell Date 9 Supervisor, Simulator Training '
n
. .. , Q THREE MILE ISLAND O ' Test Identification:
SIMULATOR' MANAGEMENT CERTIFICATION TEST ABSTRACT Weekly Surveillance. Checks Certification Test Normal Operations Test' NOT12 ANSI /ANS 3.5 Reference (si- '4.2.1-1: 3.1.1(10) Test Date: 03/01/96 Reference Plant Procedure: SP 1301-4.1~ Weekly . Surveillance [- Checks Revision 61 Test Initialize. tion: Protected Initial Condition IC-15 100% Reactor-Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated following collection of specified data. l Simulator-Run Tima: 2 hours Baseline Evaluation Data: Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant ' Surveillance Procedures Overall' Test Results: SATISFACTORY
'Basults
Description:
In . accordance' with . ANSI /ANS-3.5-1985 this certification test is ' conducted to demonstrate the ability to operate the simulator in
.accordance with similar reference plant operating procedures, using only. operator action normal to the referenced plant.
- During. the . conduct of this test simulator dynamic performance, annunciator' operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls used during.the test was also evaluated. These evaluations were conducted by GPUN personnel' representing TMI Simulator Training.
All' the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and = corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are aslfollows: 4
t, N ". NOT12 p The.results of this test are satisfactory, based upon ' the l following: 9RITERIA #1: , The simulator has met the acceptance criteria of the reference l plant procedure (s) used during the conduct of this test. The simulator supported this requirement with no deviations. CRITERIA #2: j The observable changes.in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator supported this requirement with no deviations. CRITERIA #3: The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or ) automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator supported-this requirement with no deviations. CRITERIA #4: Tactile feel.of Plant-Referenced Simulator control device (s)
~
compares to.those of the reference plant. The simulator supported this requirement with no deviations. Co rrective Actions /Commants: ) i During the conduct of this test, reference plant procedure steps l not performed were identified and evaluated for impact and I corrective action requirements in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these 1 evaluations are as follows:
)
- 1. -Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room. l
T .'. NOT12 N
- 3. Steps not performed did not prevent the successful completion of the procedure 'in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
O l 5
\
\
f Completed By ! T2 '^ W. A. Fraser Date 04/11/96 Approved By Robert L. Parnell Date A TA74
' Supervisor, Simulator Training 'l '
t . I
,T; THREE MI,LE. ISLAND ppj SI.MULATOR MANAGEMENT ,
CERTIFICATION TEST ABSTRACT Test Identification: RCS Leak Rate Surveillance Certification Test Normal Operations Test ' NOT13 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(10) Test Date: 04/04/96 Reference Plant Procedure: SP 1303-1.1 Reactor Coolant System Leak Rate Revision 28 Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End'of Cycle i Point of Test Termination: This test was terminated following collection of specified . data and calculations completed.
- p. Simulator Run Timer j 2 hours 10 minutes-Baseline Evaluation Data: Tactile Feel by Experienced Operators.
Right Direction Analysis. Current Controlled Copies of:
-Reference Plant Surveillance Procedures j Overall Test Results: SATISFACTORY Results
Description:
In ' accordance with ANSI /ANS-3.5-1985 this certification test is t conducted'to demonstrate the ability to operate.the simulator in L accordance with similar reference plant operating procedures, using l only operator action normal to the referenced plant. L .During the conduct of this test simulator dynamic performance, annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985- criteria. Tactile-feel of simulator controls used.during the test was also evaluated. These' evaluations were conducted by' select GPUN personnel representing TMI Simulator Training. All the evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in, accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these
4
,?
NOT13 O, pI; ( uJ evaluations are as follows: f ' The results of this test are satisfactory, based upon the following: g CRITERIA #11 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. CRITERIA 1 evaluators concurred during this test ,-- the simulator supported this requirem,that ent with no deviations. i J CRITERIA _fif, - The observable changes in simulator parameters correspond in ! direction to those expected from the actual reference plant i and do not violate the physical laws of nature.
.{
{ CRITERIA 2 evaluators concurred that during this test, the simulator supported this requirement with no deviations. i CRITERIA #3: , i [m()} The simulator shall not fail to cause an alarm or automatic l actuation if the reference plant would have caused an alarm or !
' automatic action, and conversely, the simulator shall not i cause an alarm or automatic action if the reference plant would not cause an al' arm or automatic. action. '
CRITERIA 3 evaluators concurred that during this test, the ' simulator supported this requirement with no deviations. CRITERIA #4: Tactile feel of Plant-Referenced Simulator control device (s) compares to those.of the reference plant. CRITERIA 4 evaluators concurred that during this test, the simulator supported this requirement with no deviations. Corrective Actions / Comments: NONE-Completed By 4 ' dr 7. A. Fraser Date 04/04/96 l Approved By MH,hm Robert 6.On/e.Jr. oWos/N cacMg)Date -04/04/^6 L. P'arnell e
, X Supe'rv{[Mr/ Simulator Training ,
u\ h_ l L_____-______--- .
o l p' . ATTACilMENT li I t I Tif REE MILE ISLAND PLANT-REFERENCED SIMULATOR { d TESTING COMPLETED YEAR #3 TESTS (1996-1997) Benchmark Tests TTS07 RCS Safety Valve Failure TTS19 Loss of Forced Flow TTS27 1.oss of All Feedwater TFS35 TurbineTrip TTS42 Main Steam Leak Inside Reactor Building TTS56 M:mual Reactor Trip TTS57 Simultaneous Closure of All Main Steam Isolation Valves TTS58 Loss of One Reactor Coolant Pump TTS59 Maximum Rate Power Ramp TTS60 Loss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Simulator Stability SSP 02 Simulator Accuracy Real Time Test p RTF01 Real Time Test
\v Transient Tests TFS30 Continuous Rod Insertion TFS31 Dropped Rod TFS32 Uncoupled Rod TTS33 Inability to Drive Rods TFS34 Failed Fuel
' TFS36 Generator Trip TTS37 Inadvertent OTSG Isolation TFS38 Inadvertent OTSG Overfeed TTS39 Pressurizer Level Control Failure TTS40 Pressurizer Heater Failure TTS41 Reactor Trip TFS43 - Main Steam Leak Outside Reactor Building Normal Operations Tests NOT14 Control Rod Movement NOT15 R.B Cooling and Isolation System Logic Channel And Component Test NOT16 Loading Sequence and Component Test and
- l. -
HPI Logic Channel Test I' NOT17 High Pressure injection NOT18 RB Emergency Cooling System NOT19 ES System Emergency Sequence and Power Transfer Test
)
- Q/
g "' THREE MILE ISLAND Numbw
' TRAINING S EDUCATION NUCLEAR , SIMULATOR MANAGEMENT PRS TTS07 TITLE Re M I
RCS SAFETY VALVE FAILURE CERTIFICATION TEST b THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Testidentification: RCS Safety Valve Failure Certification Test Transient Test TTS07 ANSI /ANS 3.5 Reference (s): B.2.2.(10) 3.1.2(1 b) 3.1.2(1d) Test Date: 11/20/96 Malfunction (s) Tested *: ES01 A ESAS Failure to Actuate at HPl Setpoint Channel A ES01B ESAS Failure to Actuate at HIP Setpoint Channel B RC27A Pressurizer Safety Valve Fails Open 25% (100% = Full Open)
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 10-15 100% Reactor Power (p\ Equi!ibrium Xenon (/ Beginning of Cycle ; I Point of Test Termination: This test was terminated with the reactor tripped, Reactor Coolant System saturated, !. pressurizer filled and system inventory decreasing. Simulator RJn Time: 10 minutes Simulator Evaluation Time: 2 hours , Baseline Evaluation Data: Previous Certification Test TTS07' dated 01/31/96. l Right Direction Analysis. Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY l l 1 l ES-1 [ s GPU Nuclear, Inc.
\ Mddletown. PA copynght 1997 s
- THREE MILE ISLAND Numbw
(-g TRAINING G EDUCATION PRS-TTS07 NUCLEAs SIMULATOR MANAGEMENT _ TITLE Revision No. 3 RCS SAFETY VALVE FAILURE CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of Reactor Coolant System inventory through a partially open Code Safety Valve on the pressurizer. The pressure in the Reactor Coolant System decreased to saturation and the pressurizer filled solid. The Engineered Safeguards Actuation System was prevented from operation. The event was initiated by failing a Pressurizer Code Safety Valve partially open. As Reactor Coolant System pressure decreased the Reactor Protection System actuated tripping the reactor and l'nitiating Reactor Trip Containment isolation. Pressure continued to decrease and Engineered Safeguards Actuation System High Pressure injection actuation setpoint was reached. The system did not actuated due to the malfunctions inserted. R: actor Coolant System pressure continued to decrease; Reactor Vessel and Reactor Coolant System voiding caused the pressurizer to fill. Reactor Coolant Drain Tank pressure increased causing rupture disk rupture with subsequent cctivity release to containment. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified revelations from expected performance have been evaluated for impact and corrective action paccordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: ( / u . The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference' plant and do not viola,te the physical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an a! arm or automatic action if the reference plant would have caused the ' alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that d' uring this test, the simulator satisfied this requirement with no
' identified deviations.
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\ GPu Nuclear ine. J Middletown PA copynght 1997 l
(_ - _ _ -_______ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ THREE MILE ISLAND Number g TRAINING & EDUCATION NUCLEAn SIMULATOR l AANAGEMENT PRS-TTS07 l TITLE Revision No. t 3 f RCS SAFETY VALVE FAILURE CERTIFICATION TEST Corrective Actiona/ Comments: NONE Completed by; Charles E. Husted 5/2/97 Date Approved by: / M Supervisor, Simulator Training Robert Pamell (f9fk/ Date i i o U i i l 1 i f
/ ES-3 GPU Nuclear,Inc Mddletown, PA Copynght 1997
,,.- THREE MILE ISLAND Numbw TRAINING a EDUCATION suctsAs SIMULATOR MANAGEMENT PRS-TTS19 M.E Re h p LOSS OF FORCED FLOW CERTIFICATION TEST 4 j THREE MILE ISLAND SIMULATOR MANAGEMENT )
CERTIFICATION TEST ABSTRACT IE3Ltdentift ution: Loss of Forced Flow Certification Test Transient Test TTS19 ANSI /ANS 3.5 Reference (s): 3.1.2(4) Benchmark Test B2.2(4) Test Date: 11/26/96 Malfunction (s) Tested *: RC 35 A/B/C/D - RC Pump Trip A/B/C/D
- Refer to Malfunction Cause and Effects Documents for options available.
Test init!alization: Protected initial Condition 1C-15 100% Reactor Power Eq'uilibrium Xenon Beginning of Cycle Reactor tripped with Reactor Coolant System heat transfer by stable natural circulation. ulator Run Timq; 10 minutes i Simulator Evaluation Time: 6 hours Basellne Evaluation Data: Previous Certification Tests TTS19 dated 02/28/90,02/22/96. Right Direction Analysis. RETRAN Data. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY l l
\ ES-1
- \ GPU Nuclear,Inc.
i Mddletown, PA Copynght 1997
/ THREE MILE ISLAND Number TRAINING & EDUCATION afpCLEAs SIMULATOR MANAGEMENT PRS-TTS19 TITLE Revleion No.
LOSS OF FORCED FLOW CERTIFICATION TEST 4 V Results Description; 1
' in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simul'ator to -
perform correctly following activation of a simulated malfunction.-
'i This transient resulted in the loss of forced flow in the Reactor Coolant System. Decay heat removal was accomplished s
= by the use of Emergency Feedwater flow to the OTSGs and steaming to the Main Condenser with natural circulation i
Svident. The event was initiated by tripping all four Reactor Coolant Pumps on overload. The reactor. tripped and initiated Reactor Trip Containment isolation. The Emergency Feedwater System actuated and filled the OTSGs to 50% level on the Operating Range. The plant transitioned to the natural circulation mode. The current revision of ATOG Procedure - 1210-10 was utilized to verify evidence of natural circulation. The RCS pressure rose due to the Makeup System
- regaining pressurizer level, compressing the steam bubble. Because this test was performed from a Beginning of Cycle
- initial Condition, decay heat levels were relatively low, causing the average differential ternperature between hot and cold leg temperatures to be less than the procedurally stated 30-50 degrees Fahrenheit.
The transient was terminated following the verification of natural circulation flow and overall plant response utilizing. ATOG Procedure 1210-10. t 1 During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel Nresenting TMl Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses jTMI Unit 1. Identified revelations from expected performance have been evaluated for impact and corrective action Jcccordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following- , CRITERIA #1
- . The observable changes in simulator parameters shall correspond in direction to those expected from the actual i= reference plant and do not violate the physical laws of nature.
p l All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
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All CRITERIA 2 evaluators corx:urred that during this test, the simulator satisfied this requirement with no identified deviations. l - ES-2 GPU Nuclear, Inc. IAddletown, PA Copynght 1997 1
- THREE MILE ISLAND Numbw TRAINING Q EDUCATION NUCLEAn SIMULATOR MANAGEMENT -
PRS TTS19 TITLE RevWon No. LOSS OF FORCED FLOW CERTIFICATION TEST 4 Corrective Actions / Comments: NONE Completed by: / ] 8' ' eete- = - William A. Fraser 00/16/97 (/V Date Approved by: / ~ Supervisor, Simulator Training Robert Parnell Yfh3 97
, Date 8
f v j i l 1 s i s ES-3
'~ GPU Nuclear,Inc.
Mddletown, PA Copynght 1997 ? t
THREE MILE ISLAND Number TRAINING & EDUCATION NUClKAN SIMULATOR MANAGEMENT PRS-TTS27 _ TITLE Revision Ns . ' 2 LOSS OF ALL FEEDWATER CERTIFICATION TEST THREE MILE ISLAND ) SIMULATOR MANAGEMENT I CERTIFICATION TEST ABSTRACT Test identification: Loss of All Feedwater Certification Test Transient Test 1TS27 , l 1 ANSI /ANS 3.5 Reference (s): B.2.2.(2) ) 3.1.2(10) T. pat Date: 11/22/96 Malfunction (s) Tested *- I FW15A/B Main Feedwater Pump Trip 1 A/1B ! FW17 Emergency FW Pump Trip (EF-P 1) FW18A/B Emergency FW Pump Trip (EF P 2A/B) Test Initlaimation: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle This test was terminated with the reactor tripped, Reactor Coolant System heatup in progress with Primary System pressure controlled by the pressurizer Power Operated Relief Valve. Pressurizer level was increasing and RCS subcooling margin was decreasing. Simulator Run Time: 10 minutes Simulator Evaluation Time: 4 hours Baseline Evaluation Data: Previous Certification Test TTS27 dated 01/31/96. Right Direction Analysis. Simulator Malfunction Cause and Effects Diagram Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Reference Electrical One-Line Diagrams Overall Test Results: SATISFACTORY s (V) ES-1 "usfoO*A Copyrght 1997 t
THREE MILE ISLAND Number TRAltdlNG & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS27 TITLE _ Revison No. 2 LOSS OF ALL FEEDWATER CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly foliowing activation of a simulated malfunction. J This transient resulted in the Icss of all secondary side heat removal with Reactor Core heat removal via the PORV. During the duration of this test, heat transfer from the RCS was insufficient for removal of decay heat as expected. Tha event was initiated by failing both Main Feedwater Pumps and preventing the start of all Emergency Feedwater I Pumps. The Reactor Protection System tripped the reactor and initiated Reactor Trip Containment Isolation The Heat l Sink Protection System actuated to initiate Emergency Feedwater but the malfunctions prevented the pumps from starting. As steam generator inventory was lost Reactor Coolant System ternperature increar.,ed causing system pressure to I,ncrease resulting in opening of the pressurizer spray valve then the PORV. At the end of the test the PORV was cycling to limit Reactor Coolant System pressure and the Reactor Coolant system was approaching saturated conditions. Th: transient was terminated rollowing OTSG dry-out, with RCS heatup in pmgress and the PORV controlling RCS pressure. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1, Identified deveiations from expected performance have been evaluated for impact and corrective action in ordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIM1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. L - The simulator satisfied this requirement with no identified deviations. l 1 fK ES-2 Mut tw A Copyrght 1997 i i
l THREE MILE ISLAND Number ' TRAIN'NG & EDUCATION muctr4m SIMUU'. TOR MANAGEMENT PRS-TTS27 Revmon No,
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f'_ TITLE - 2 (v) LOSS OF ALL FEEDWATER CERTIFICATION TEST l l Corrective Actions / Comments: l NONE j Completed by:
. Charles E. Husted j!//97 Date Approved by:
Operations Training Manager Robert Pamell !'l b .7
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THREE MILE ISLAND Number g TRAINING & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS35 TITLE Revision No. [ \ TURBINE TRIP CERTIFICATION TEST
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THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Turbine Trip Certification Test Transient Test TTS35 ANSI /ANS 3,5 Reference (s): 3.1.2(15) Benchmark Test B2.2(6) Test Date: 11/26/96 Malfunction (s) Tested *: TC01 Turbine Trip g
' Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 10-13 50% Reactor Power Equilibrium Xenon End of Cycle This test was terminated after plant stability was reached at the lower reactor power level following the Main Turbine trip. Simulator Run Time: 10 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test TTS35 dated 2/1/96. l Right Direction Analysis. ' Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diag, rams l OverallTest Results: SATISFACTORY O E5-1 GPu Nuclear,Inc. (<\' Mddletown, PA Copynght 1997 i- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
THREE MILE ISLAND Number TRAINING & EDUCATION mucum SIMULATOR MANAGEMENT PRS-TTS35 TITLE Revision No. rw 3 ( ) TURBINE TRIP CERTIFICATION TEST U Beaults Description in accordance wlth ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a turbine trip with reactor power below the high power trip setpoint. The event was initiated by activation of the turbine trip malfunction after power was reduced to 41% in order to prevent a r: actor trip. The Integrated Control System responded by initiating a power reduction, since the control system transferred to the Tracking mode with megawatts generated equal to zero when the turbine was tripped. Reactor power change due to increased temperature was less significant than the initial test due to incorporation of a new l core model with different temperature coefficient values. Reactor Coolant System pressure and temperatu're decreased significantly as reactor power decreased to less than 22% SASS actuated a mismatch indication for T-cold transm'ars. R: actor Coolant System pressure initially increased resulting in operation of the Pressurizer spray valve. E.. ,ency Feedwater Actuation occurred due to low level in OTSG-A&B. Main Feedwater underfeed occurred on both OTSGs. The underfeed is a result of initially reduced levels and higher than normal OTSG pressures resulting in level dropping below initiation setpoin'. Underfeed is expected to occur in the plant during low power transients. The transient was terminated when primary and secondary pressures and temperatures were returning to stable conditions. ynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test mulator dynamic response, annunciator operation and automatic safety systern actuations were recorded. Test results ve been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: 'j The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. , CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. O) ( (/ ES-2 GPU Nuclear,Inc. Mddletown, PA Copynght 1997 r e t__ . _ __ _ _ _ _
n THREE MILE ISLAND Numbw TRAINING G EDUCATION muetsAn SIMULATOR MANAGEMENT PRS 'I i 335 TITLE Re e M. [%/ TURBINE TRIP CERTIFICATION TEST Corrective Actions / Comments: NONE i Completed by: v
Charles E. Husted I' Dats
!77 Approved by: E!!P-Supervisor, Simulitor Training Robert Pamell 5'/7[47
' Date r
I I h l -( O ES-3 GP'J Nuclear, Inc. c
- b. . . i Mddletown, PA Copynght 1997
M' THREE MILE ISLAND Nunt w TRAINING & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS42 TITLE Ravision No. 4 MAIN STEAM LEAK INSIDE THE RB CERTIFICATION THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Tegt Identification: Main Steam Leak inside the RB Certification Test Transient Test TTS42 ANSI /ANS 3.5 Reference (s): Benchmark Test B2.2(9) 3.1.2(20) Test Date: 11/22/96 Malfunction (s) Tested *: MS02A Main Steam Leak in the RB 100% (100% - 6,000,000 LBM/HR)
- Refer to Malfunction Cause and Effects Documents for options available.
Testinitialization: Protected initial Condition 1C-16 I 100% Reactor Power Equilibrium Xenon Middle of Cycle - This test was terminated following the reactor trip, HPl, and Emergency Feedwater actuations due to the effects of a main steam line rupture insido containment. Simulator Run Timm 10 minutes Simulator Evaluation Time: 4 hours Beseline Evaluation Data: Previous Certification Tests TTS42 dated 03/02/90, 6/23/95, 3/1/96. I Right Direction Analysis l RETRAN Data I Simulator Malfunction Cause and Effects Document I Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Roulla; SATISFACTORY i E5-1 t, )
- GPU Nuclear, Inc.
Mddletown. PA Copynght 1997 1
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THREE MILE ISLAND Numbw
*g TRAINING G EDUCATION sucurAN SIMULATOR MANAGEMENT PRS TTS42 TITLE Revleion No.
4 9 MAIN STEAM LEAK INSIDE THE RB CERTIFICATION Results Description; in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction.
- The event was initiated by causing a maximum size steam rupture inside containment on the OTSG 1 A. The resultant steam release caused a reactor and turbine trip on Overpower with subsequent Reactor Trip Containment isolation actuation. The containment pressure and temperatures increased cauting Engineered Safeguards actuation, and a false Reactor Building fire alarm. The rapid blowdown of OTSG 1 A caused a significant cooldown of the RCS. A SASS actuation occurred due to the significant mismatch between the A and 8 Main steam header pressures. Feedwater flow was !solated to both OTSGs on HSPS Low Pressure. The Emergency Feedwater System actuated on high Reactor Building pressure and fed the OTSGs. High pressure injection flow recovered the RCS inventory lost by shrinkage and compressed the pressurizer steam bubble, elevating RCS pressure. The spray valve automatically actuated to dampen the pressure increase.
The transient was terminated following Engineered Safeguards and Emergency Feedwater system actuations with elevated Containment pressures and temperatures.
)
Dynamic response was compared to the dynarnic response of the initial certification test. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by select GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: = _QBJTERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. PRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the
. reference plant would not cause an alarm or automatic action.
The simulator satisfied this requirement with no identified deviations. I l l l O
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ES-2 .
\s) GPu Nuclear,Inc.
I ' \ Middletown, PA Copynght 1997
t-
, THREE MILE ISLAND Numter TRAINING Q EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS42 E Revision Nc,.
( 4 x r MAIN STEAM LEAK INSIDE THE RB CERTIFICATION
%d Corrective Actions / Comments:
NONE Completed by:
-v v-
/ r 9 W. A. Fraser SM 9gg, i(pproved by: / -
Operations Tialning Manager Robert Pamell NF//97 Dats l
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(Q )) E5-3 GPU Nuclear,Inc. Mddletown, PA Copynght 1997
THREE MILE"lSLAND Numbw TRAINING & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS56 mTITLE Revtsion No. ( MANUAL REACTOR TRIP CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Manual Reactor Trip Certification Test Transient Test TTS56 ANSI /ANS 3.5 Reference (s): Benchmark Test B.2.2(1) Test Date: 11/25/96 Malfunction (s) Tested *: None - Initiation of Manual Reactor Trip Test initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor tripped and proceeding to stable Hot Shutdown conditions, ulator Run Time: 10 minutes Simulator Evaluation Time: 2 houns , Baseline Evalcation Data: Previous Certification Tests dated 03/07/90,01/09/95,03/01/96 Right Direction Analysis Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SAT 4SFACTORY i ES-1 l \ GPu Nuclear,Inc. Middletown. PA Copynght 1997
THREE MILE ISLAND Numtw a(
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TRAINING C3 EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS56 { Revtsion No. 4 MANUAL REACTOR TRIP CERTIFICATION. TEST w Results
Description:
In accordance with ANSI /ANS 3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated by depressing the Manual Reactor Trip pushbutton on the Main Console and reducing the pr:ssurizer level control setpoint to 100 inches. All control rod drive breakers opened and Reactor Trip Containment Isolation actuated. The Main Turbine tripped on intericek. The integrated Control System reduced Feedwater flows and i stabilized OTSG levels at Low Level Umits. The Turbine Bypass Valves operated to maintain OTSG pressure and l remove decay heat. The transient was terminated with the reactor tripped and the plant approaching stable Hot Shutdown conditions. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were evaluated. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives ha've held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expceted performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: 1 The results or this test are satisfactory, based uoon the following: l CRITERIA #1 { m : J l ) The observable changes in simulator parameters shall correspond in ' direction to those expected from the actual V reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no ider.tified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. l l I 9 I ES-2 i ( GPU Nuclear, Inc. i Mddletown. PA Copynght 1997
;F THREE MILE ISLAND . Number TRAINING & EDUCATION NUCLEAN SIMllLATOR MANAGEMENT PRS-TTS56 TITLE Rom No.
[ MANUAL REACTOR TRIP CERTIFICATION TEST V Corrective Actions / Comments; NONE Completed by: William A. Fraser #~F/ Date Approved by: Operations Training Wanager Robert Parriell I!>7 !4 7 Dafe (3 V l (
- / E5-3 GPU Nuclear,Inc.
' Mddletown, PA -
Copynght 1907 L______________----- - _ - -
THREE MILE I! LAND Numts G'W NUCLEAn TRAININ3 6 EDUCATION SIMULATOR MANAGEMENT PRS-TTS57 TITLE Revtsson No.
/7 3
(%d ) SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Simultaneous Closure of All MSIVs Certification Test Transient Test TTS57 ANSI /ANS 3.5 Reference (s): Benchmark Test B.2.2(3) Test Date: 11/25/96 Malfunction (s) Tested *: MS08A/B/C/D Main Steam isolation Valve Closure MS-V-1 A/1B/1C/10 Test initialization: Protected Initial Condition IC 16 100% Reactor Power Equilibrium Xenon Middle of Cycle Point of Test Termination: This test was terminated with the reactor tripped and Emergency Feedwater removing decay heat via the Atmospheric Dump Valves, ulator Run Time: 20 minutes Sjmulator Evaluation Time: 4 hours Basellne Evaluation Data: Previous Certification Test dated 01/04/95 Right Direction Analysis l Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant P & ID Procedures e Overall Test Results: SATISFACTORY l l O ES-1 V " O Z ': Copynght 1997 l
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l l THREE MILE ISLAND Nurnber TRAINING & EDUCATION NUCMAn SIMULATOR MANAGEMENr PRS-TTS57 TITLE Revision No. SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST Results
Description:
l In accordance with ANSI /ANS-3.51985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulbd in a reactor trip on high pressure and Emergency Feedwater System actuation due to a loss of Main Feedwater. The event was initiated by simultaneous closure of all four Main Steam Isolation Valves. This caused OTSG pressures t3 increase and Main Steam Header Pressure to decrease. The Main Turbine transferred to mane 3alcontrol automatically ' upon receipt of a sustained 40 PSIG Header Pressure Error Signal. As OTSG pressures increased the Turbine Bypass, Atmospheric Dump and Main Steam Safety Valves operated to limit the pressure excursion. Feactor Coolant System i pressure and temperatures increased due to the transient. The reactor tripped on high pressure and actuated Reactor Trip Containment Isolation. When all Main Steam isolation Valves hhd closed the Main Feedwater Pumps slowed, reducing Feedwater flow to the OTSGs. The Emergency Feedwater System actuated to feed the OTSGs at Low Level Limits. Gland Sealing steam was lost when the OTSG 1B MSIV's closec8 causing a loss of vacuum to the Main Tur~oine cnd Main Feedwater Pumps. Decay heat was then removed by automats action of the Atmospheric Dump Valves, since the Turbine Bypass Valves closed due to the low Condenser Vacuum condition (causeo by the loss of Gland Seal Steam). The simulator response differed from previcras tests due to the decrease in MS-V-1 stroke time from 120 seconds to 90 seconds. The transient was terminated with the reactor tripped and decay heat being removed by the Emergency Feedwater System via the Atmospheric Dump Valves. f3 f ( iring the conduct of this test simulator dynamic response, annunciator operation, and automatic r:afety system kluations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Trainin0. The TMl representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evakations are as follows: The resuks or this test are satisfactory, based upon the follow!ng: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identDied deviations. 931TIBlM2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. [ ES-2 t I i\s/ GPU Nuclear. Inc, Middletown. PA Copynght 1997 4
THREE MILE ISLAND Annber TRAINING 0' EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS57 TITLE
- Revision No.
3 SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST I Corrective Actions / Comments: t i NONE Completed by:
/ '
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William A. Fraser SI #^ b) Date Approved by: Operations Training Manager Robert Parnell NN' Date
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l [ ( l l l l [ E5-3 GPU Nuclear, Inc. I Mddletown. PA Copynght 1997 _______________________________l
, THREE MILE ISLAND Number
- ,W TRAINING & EDUCATION
' NUCLEAN SIMULATOR MANAGEMENT PRS-TTS58 TITLE Revtsion No.
,bh LOSS OF ONE RC PUMP CERTIFICATION TEST U
TH.REE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Loss of One RC Pump Certification Test Transient Test TTS58 ANSI /ANS 3.5 Reference (s): 3.1.2(4) Benchmark Test B.2.2(5) Test Date: 12/04/96 Malfunction (s) Tested *: RC35A RC Pump Trip Test initialization: Protected initial Condition 1C-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated following reactor trip with Primary and Secondary Plant pressures and temperatures approaching stability. 10 minutes Simulator Evaluation Time: 3 hours
- Baseline Evaluation Data
- Previous Certification Test TTS53 dated 02/01/96 l Right Direction AnaF/ sis Simulator 14alfunction Cause and Effects Document.
Current Controlled Copies of: Raference Plant Alarm Response Procedures Re:srence Plant Emergency Procedures Overall Test Results: SATISFACTORY I
?
O \ ES-1 GPU Nuclear,Inc. Cop ngh i99
r- J f- ! i q ! THREE MILE ISLAND Number
~ j TRAINING S EDUCATIOfJ ;
NUCLEAg SIMULATOR MANAGEMENT PRS-TTS58 i TITLE Revision No. ! 3 I [D LOSS OF ONE RC PUMP CERTIFICATION TEST esults
Description:
l- In accordance with ANSl/ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to ! perform correctly following activation of a simulated malfunction. (
' This transient resulted in a reactor trip. Reactor Trip initiated a Main Turbine trip and Reactor Trip Containment isolation.
The event was initiated by failing Reactor Coolant Pump 1 A. The loss of the Reactor Coolant Pump at full power resulted in the Reactor Protection System initiating Reactor Trip due to the reduction in Reactor Coolant System flow.- The Reactor Trip resulted in a Main Turbine trip. Reactor Trip Containment isolation closed specified Containment isolation Valves. Reactor Coolant System pressure and temperature decreased to post trip values and were maintained. SASS actuated a mismatch indication f or T-cold and T hot transmitters. l l . The transient was terminated when Primary and Secondary pressure and temperature were approaching post trip stable conditions.
~ During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by select GPUN personnel l representing TMI Simulator Management. The TMI representatives have held NRC Senior Reactor Operator license for
, TMI Unit 1. Identified deviations from expected performance have been evaluated for Npact and corective action in '
- acordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evalautions are as follows: i l
g The results of this test are satisfactory, based upon the following:
? \ CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
The simulator natisfied this requirement with no identified deviations. CRITERIAJ2 h , The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the ! alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. I ! The simulator satisfied this requirement with no identified deviations. l l l l o ES-2 [ . GPU Nuclear,Inc. t
\ Mddletown, PA i V
l Copynght 1997 l I
1 1 THREE MILE ISLAND ' Number TRAINING & EDUCATION NUCLEAR SIMULATOR MANAGEMENT FRS-TTS58 Revtson No. LOSS OF ONE RC PUMP CERTIFICATION TEST Corrective Actions / Comments: NONE s Wmpleted by: 1 Charles E. Husted / 77 Cate 7 l Approved by: Robert Pamell I /I Operations Training Manager Date O - i V l l 1 ES 3 ( - GPU Nudear, Inc. L Middletown. PA Copyngnt 1997
~
.: l'.
[^ THREE MILE ISLAND Number
\ TRAINING G EDUCATION NUCLEAir SIMULATOR MANAGEMENT PRS-TTS59 TITLE Revtsion No.
~
MAXIMUM RATE POWER RAMP CERTIFICATION TEST v THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Maximum rate Power Ramp Certification Test Transient Test TTS59 ANSI /ANS 3.5 Reference (s): Benchmark Test B.2.2(7) 5.4.1(2) , Test Date: 11/25/96 Malfunction (s) Tested *: None - Manual Plant Maneuvering Test initialization: Protected initial Condition 10-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination; This test was terminated with the plant stable at 100% power, 15 minutes Simulator Evaluation Time: 1 hour Baseline Evaluation Data: Previous Certification Test TTS59 dated 01/25/96 Right Direction Analysis . Current Controlled Copies of: Reference Plant Operating Procedures
- i. OverallTest Results: SATISFACTORY i
i O ES-1
"==%
Copyrgni 1997 O
C THREE MILE ISLAND Number TRAINING O EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS59 TITLE Rev6sion No.
" 3 MAXIMUM RATE POWER RAMP CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated from steady-state full power with the Integrated Control System l'n the integrated mode. The Unit Load Demand controller was set to approximately 75 percent load. The plant reduced power at the maximum rate of ten percent per minute and stabilized at 75 percent power. The plant was allowed to stabilize for approximately four minutes. At the end of the stabilization period, power was ramped back up at the maximum rate of ten percent per minute. When the plant reached 90 percent load the integrated Control System automatically reduced the rate of power increase to 5 percent per minute. During the period of time at reduced power the Xenon concentration began to increase. The plant w:s allowed to stabilize at full power. The transient was terminated with the plant at full power. Dynamic response was compared to the d/namic response of the initial certification test. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221 ADM-2820.02. The results of these evaluations are as follows: p The results or this test are satisfactory, based upon the following: I "k./ CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.' The sirautator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the l alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. [ \ ES-2 (' ) GPU Nuclear. inc.
'y/ Middletown. PA Copyngnt 1997
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THREE MILE ISLAND Numbw TRAIN 8N9 6 EDUCATION NUCLEAn SIMU* 'rtMANAGEMENT 2 PRS-TTS59 TITLE RevlWon No, , MAXIMUM RATE POWER RAMP CERTIFICATION TEST Corrective Actions / Comments: NONE Completed by: V
/A-* '
/
"f'- William A. Fraser F 2,P-Pp Date Approved by:
~
Robert Pamell $/&#l[h 7 Supervisor, Simulator Training Date i t /" \v O ( ES-3 (j) GPU Nuclear, Inc. Mddletown, PA Copynght 1997
~ '
THREE MILE ISLAND Number TRAINING G EDUCATION NUCLEAa SIMULATOR MANAGEMENT PRS-TTS60 TITLE Revtsion No. (v) LOSS OF OFFSITE POWER WITH LARGE BREAK LOCA CERTI THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT . Test identification: Loss of Offsite Power with Large Break LOCA Certification Test Transient Test TTS60 ANSI /ANS 3.5 Reference (s): 3.1.2(1b) Benchmark Test B.2.2(8) Test Date: 11P.5/96 - Malfunction (s) Tested *: ED01 Station Blackout TH04A RCS LOCA at Hot Leg Nozzle 100% severity Test initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle I This test was terminated with the reactor tripped, Core Flood tanks emptied and reactor core cooling being provided by High Pressure and Low Pressure Injection. Reactor Building pressure and temperature were decreasing from peak conditions. Simulator Run Time: 10 minutes Simulator Evaluation Time: 9 hours Baseline Evaluation Data: Previous Certification Test TTS60 dated 12/11/95. Rignt Direction Analysis. Simulato,r Malfunction Cause and Effects Diagram Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY
/ E5-1 \d '*
u$io$.N Copynght 1997 l I l L_-_____________-__.
.(g THREE MILE ISLAND TRAINING G EDUCATION Number NUCLEAn SIMULATOR MANAGEMENT PRS-TTS60 TITLE Revision No.
LOSS OF OFFSITE POWER WITH LARGE BREAK LOCA CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the loss of offsite power coincident with a large break loss of coolant accident. The reactor was I tripped due to a loss of power to the Reactor Coolant Pumps and the Control Rod Drive System. Safety system power was restored by the Emergency Diesel Generators. A large break in the Reactor Coolant System caused rapid depressurization of the coolant and a rapid increase in Reactor Building pressure. The Engineered Sefeguards System actuated, initiating safety injection and Reactor Building Isolation and Cooling. l The event was initiated by fault in the 230 KV Substation coincident with a large size rupture of the Reactor Coolant !
~
System Hot Leg piping. The Reactor Protection System initiated a Reactor trip and Main Turbine trip. Reactor Trip Containment Isolation closed specified Containment isolation valves. The Heat Sink Protection System detected a loss of Reactor Coolant Pumps and initiated Emergency Feedwater to the OTSGs, controlling level at 50% on the operating range. The Emergency Diesel Generators started and the output breakers closed to re-energize their 4.16 KV Buses. SASS actu&ted a mismatch indication for Turbine Header Pressure Transmitters and RCS Pressure Transmitters.
]
The loss of inventory and pressure from the Reactor Coolant System resulted in discharge of the Core Flood Tanks and actuation of the Engineered Safeguards Actuation System. High Pressure injection, Low Pressure injection, Reactor j Building Spray, and Reactor Building Isolation and Cooling were initiated due to the pressure loss from i e Reactor Coolant System and the high pressure detected in the Reactor Building. I S,he transientTninated waswhen te:Reactor Building pressure and temperature were decreasing and reactor core cooling res verified. Dynamic response was compared to the dynamic response of the previous certification tests. During the conduct of this t;st simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test r;sults have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. , The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMl Unit 1, Identified deviations { from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity 1 Procedure,6211-ADM-2820.02. The results of these evaluations are as follows: I The results or this test are satisfactory, baced upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. , 1 All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with one identified deviation: Containment Temperature.
/
i E5-2 I \ GPU Nuclear, Inc. Middletown, PA Copynght 1997
P THREE MILE ISLAND Number TRAINING & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS60 TITLE Revision No.
~
) LOSS OF OFFOlTE POWER WITH LARGE BREAK LOCA CERTIFICATION TEST
\J CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments:
Description:
Incorrect System Response Number of Identified Deviations: One Corrective Actio.nn: One Discrepancy Remrt correction to resolve the plotting' response problem with CHTT655D. O mpleted by: ' Charles E. Husted [ /97
' Date
' Approved by. Robert Parnell .$7f / 9 7 Supe'rvisor, Simulator Training Dhte i
I I O ( ES-3 GPu Nuclear,Inc. N Mddletown, PA Copynght 1997 t
THREE MILE ISLAND Number 3f' TRAINING & EDUCATION WUCLKAN SIMULATOR MANAGEMENy PRS-SSP 01 TITLE Rom No. R 4 (v) SIMULATOR STABILITY CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Simulctor Stability Certification Test Steady State Performance Test , SSP 01 ANSI /ANS 3.5 Reference (s): 5.4.1(2) 4.1 Benchmark B2.1 1 Test Date: 11/27/96 Test initialization: Protected initial Condition 10-17 100% Reactor Power Equilibrium Xenon '
'End of Cycle Point of Test Termination: 100% Reactor Power 579 Degrees Tave
, ICS in Full Automatic
\
' t Simulator Run Time: 1 Hour I Baseline Evaluation Data: Plus/minus 2%
OverallTest Results: SATISFACTORY Results
Description:
1 in accordance with ANSI /ANS-3.5-1984 this certification test was conducted to demonstrate the ability of the simulator to operate at stable full powel, in automatic control for a 60-minute period. {
)
During conduct of this test simulator dynamic operation was evaluated against ANSl/ANS 3.5 criteria. This evaluation was conducted by GPUN personnei representing TMI Simulator Training. The evaluators have held an NRC Senior Reactor Operator license for TMI Unit 1. The results of this evaluations are as follows: I d GPU Nuclear, Inc. , Middletown, PA ) Copynght 1997
, THREE MILE ISLAND Number
' ~ gg TRAINING & EDUCATION NUCLKAN SIMULATOR MANAGEMENT PRS-SSP 01 TITLE Revision No.
pI g SIMULATOR STABILITY CERTIFICATION TEST o V The results of this test are satisfactory, based uoon the following:
\
i CRITERIA #1: The simulator computed values for steady state, full power operation with the referenced plant control system configuration shall be stable and not vary more than plus/minus 2% of the initial values over a 60-minute period. The simulator supported this requirement with no deviations. l I Completed by: Charles E. Husted /S!T7 i D' ate Approved by: Operations Training Manager Robert L. Pamell /6[47 Dhte l l [ l E3-2 GPu Nuclear. Inc. Mkidletown, PA Copynght 1997
I THREE MILE ISLAND Number TRAINING & EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-SSP 02 TITLE Revision No. p . SIMULATOR ACCURACY CERTIFICATION TEST 5
\b') THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification; Simulator Accuracy Certification Test Steady State Performance Test SSP 02 ANSI /ANS 3.5 Reference (s): 4.1(2) 4.2(3) 4.1(4) 5.4.1(2)
Benchmark Test B2.1 Test Date: 12/09/96 Test initialization: Protected initial Condition 1C-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test termination: After measurement of simulator computed values and principal mass / energy balance determinations were made at 100%,80%, and 60% rated power. 9 simulator Run Time: 3 hours Baseline Evaluation Data; Previous Certification Tests SSP 02 dated 01/12/98. . Current Controlled Copies of: Reference Plant Operating Data Reference Plant Operating Procedures Overall Test Results: SATISFACTORY j l I e t
)
, s j E5-1 L b/ GPU Nudaar,Inc. I Middletown. PA Copynght 1997
' . THREE MILE ISLAND Number
- TRAINING & EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-SSP 02 TITLE Revision No.
SIMULATOR ACCURACY CERTIFICATION TEST 5 - Fleaults
Description:
l In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to perform within the accuracy requirement of plus/minus 2% for critical parameters and plus/minus 10% for noncritical parameters pertinent to plant operation. Principal mass and energy balances shall be satisfied. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified devoiations from expected performance have been evaluated for impact and corrective action in I accordance with Functional Fidelity Procedure,6221-ADM 2820.02. The results of these evaluations are as follows: I The results or this test are satisfactory, based upon the following: CRITERIA #1 l The simulator computed values of critical parameters shall agree within plus/minus 2% of the reference plant parameters and shall not detract from training. The parameters displayed on control panels may have the instrument error added to the computed values. The simulator satisfied this' requirement with no identified deviations. i pg CRITERIA #2
. The calculated values of noncritical plant parameters pertinent to plant operation, that are included on the simulator control panels, shall agree within plus/minus 10% of the reference plant parameters and shall not detract from training. The parameters displayed on control panels may have the instrument error added to the computed values. 1 The simulator satisfied this requirement with no identified deviations.
Corrective Actions / Comments: NONE 1 i
] 3 06/04/97 Completed by:
v-- h L< < ' Charles E. Husted pg,g Approved by: - Operations Training Manager Robert Pamell 'd/4[f7 ' D' ate f (\ l ES-2 GPU Nuclear, Inc. lWodletown. PA Copynght 1997
- l. ._ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
' j THREE MILE ISLAND e
l SIMULATOR MANAGEMENT
-%< N. .. - !
CERTIFICATION TEST' ABSTRACT Test ^ Identification: -Real Time Test RTT01 f
' ANSI /ANS - 315 Reference (s): Appendix A3 (1) 1 N_. -
Test-Dates: 06/17/97 & 06/23/97 Test-Initialization: . Protected Initial Condition IC-17 .
'100%-Reactor Power i Equilibrium Xenon -{
End of. cycle i Point of Test Termination: This test /was terminated following collection'of all required data. Simulator'Run Time , 1 hours
, Simulator Evaluation Times 2 hours
- Baseline Evaluation Data Valve stroke times per controlled copies'of applicable Reference Plant surveillance procedures.' Enown values for event timers.
Overall Test Results: SATISFAC"3RY-s Results
Description:
In accordance-with ANSI /ANS-3.5-1985 this certification test'was conducted'to
. demonstrate the ability of the simulator computers to run in real time.
During the conduct of this test simulator dynamic performance was evaluated
- (N
? IL against ANSI /ANS 3.5-1985 criteria. These evaluations'were conducted by GPUN personnel representing TMI Simulator. Training who hold or have held NRC Senior
\s_ / Reactor Operator licenses for'TMI Unit 1.
-This test verified the capability of the simulatorLeomputers to run in real time by measuring known values for selected valve stroke times and event ,
1 timers. The correct valve stroke' times were drawn from controlled copies of' Reference Plant surveillance. procedures. . Surveillance procedure valve-stroke
times.were used as reference points to validate simulator computer-
' performance. There is-no intention.to exactly match reference plant procedurally required values. Known values for' event: timers were drawn from j
= Reference Plant system design data. A calibrated. stopwatch was utilized to '
time =each' item. Additionally the computer system processing capabilities were measured'to
. serve as.an annual check on each processor's performance. The evaluation of. , .this data servea'as a computer system management: tool, h, DuringIthis certification' period the simulator computer 'ystem s was. upgraded.
i This required ainew baseline for evaluation of processor performance. The !
- > testing method was revised to
- focus on the periods of time where maximum (processor. operations-occurred and to incorporate new performance monitoring W software.
l I LThe results'of this test.are satisfactory with &ll valve stroke times and' I , eventt timers-performing as expected. Minor differences (less than 0.5-second) exist'due to the use of~a manual stopwatch for timing, computer processing results. were baselined to allow future performance monitoring' under repeatable conditions. i n L's_/1 L 1 L l 1 1 Y +- .'
- _ . _ _
6 gj ,lr . l.
' Corrective Actions / Comments:=
Description:
MU-V-14A stroke Time Different i !' Nirmher of Identified Deviations: One ~ Corrective Actions: One modification ( i .- ' [ Completed By' N- ~ William A. Fraser Date: 06/25/97
, Approved By. Robert L.Parnell-III Date: 06/25/97' Operations Training Manager i
n I-I l~ I f
- 1. .
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THREE MILE ISLAND Number gh TRAINING & EDUCATION pUCLEAN SIMULATOR MANAGEMENT PRS-TTS30 TITLE E Revision No. 7:s 1 i CONTINUOUS ROD INSERTION CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Continuous Rod Insertion Certification Test Transient Test TTS30 ANSI /ANS 3.5 Reference (s): 3.1.2(12) Test Date: 11/25/96 Malfunction (s) Tested *; RC05C - Continuous Rod Insertion Manual and Auto Group 7
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 1016 100% Reactor Power Equilibrium Xenon Middle of Cycle This test was terminated with reactor power being reduced by the effects Xenon build-up. Main generator output was reducing, Simulator Run Time: 15 minutes S!mulator Evaluation Time: 2 hours and 30 minutes Baseline Evaluation Data: Previous Certification Test TTS30 dated 5/28/93. Right Direction Analysis. Simulator Malfunction Cause and Effects Diagram Overall Test Results: SATISFACTORY
,O ES-1 N dd eto .PN Copynght 1997
THREE MILE ISLAND Numtwr e TRAINING G EDUCATION wecurAR SIMULATOR MANAGEMENT ~ PRS TTS30 TITLE Flevision No. l'"\ 1 v CONTINUOUS ROD INSERTION CERTIFICATION TEST Results
Description:
in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the plant operating at a reduced power level with Control Red Group Seven inserted fully. The eve nt was initiated by inserting a malfunction to cause continuous insertion of all Group Seven rods. The Integrated Control System sensed the mismatch between reactor power and ICS reactor cfamand and initiated a Cross-Limit. This caused the ICS to transfer to the Tracidng mode and reduce secondary plant output to remain within five percent of reactor power. The feedwater flow reduction due to the Cross-Limit allowed T-ave to recover as the power reduction continued. This caused RCS pressure to increase and reach the spray valve setpoint. The plant continued to reduce power. The Cross-Limit operation maintained system stability. The build-up of Xenon due to the power reduction continued to cause reactor power to reduce after Rod Group Seven was fully Iraerted. T-ave remained stable and generator output continued to follow the power reduction. The certification test conducted this year differed significantly from the 1993 test due to the installation of an up-graded reactor core model. The present core has different reactivity characteristics which include a higher Group Seven rod worth. This resulted in an additional ten percent reduction in power at the termination of the transient. The transient was terminated following Rod Group Seven insertion, with Xenon build-up slowly reducing reactor power. Main Feedwater flow and generator output were decreasing as the turbine control valves slowly closed to control steam ssure. ring'the conduct of this test simulator dynamic response, annunciator operation and automatic 3afety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel
, representing TMl Simulator Training. The TMl representatives have heid NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows:
i I The results or this test are satisfactory, based upon the following- i CRITERIA #1 1 , The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. . J All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no
' identified deviatioris.
CRITERIA #2 ' The simulator shall not fall to cause art alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the l reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurmd that during this test, the simulator satisfied this requirement wkh no identified deviations. (
,a) ES-2 GPu Nuclear, Inc.
Middletown. PA Conyngrit 1997
- i l
. THREE MILE ISLAND Number .
! TRAINING G EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-TTS30 ME Revision No, CONTINUOUS ROD INSERTION CERTIFICATION TdST i Corrective Actions / Comments: NONE ' w Completed by: //' M*#~ W. A. Fraser I @) 7" Date 'i Approved by: Robert Pamell 5"/#'7/4 7 ) Supeivisor, Simulator Training Date ; l J t. l i L v ("') l t l . l l l l l l I e' ES-3 if GPU Nuclear, Inc. i(%. [ Meddietown. PA l Copynght 1997 1
t y.. v. THREE MILE ISLAND Number
' TRAINING & EDUCATION AfDCLEAN SIMULATOR MANAGEMENT PRS-TTS31 TITLE Rewson No. /^g 2
( i DROPPED ROD CERTIFICATION TEST
'wd THREE MILE ISLAND l SIMULATOR MANAGEMENT l
CERTIFICATION TEET ABSTRACT ? Test identification: Dropped Rod Certification Test Transient Test TTS31 ANSI /ANS 3.5 Reference (s): 3.1.2.(12) 5.4.1(2) l Test Date- 11/25/96 Malfunction (s) Tested *: RD0160 Dropped Rod Test Initialization: Protected initial Condition 1C-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle This test was terminated when stability was achieved at the lower power level. 7.0 minutes Simulator Evaluation Time: 2 hours ' Baseline Evaluation Data: Previous Certification Test TTS31 dated 01/25/96. Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference PlantEmergency Procedures OverallTest Results: SATISFACTORY l C ES-1 GPU Nuclear, Inc. ( Mddielown. PA copynght 1997
4 THREE MILE ISLAND Number
., TRAINING & EDUCATION wuctr4m SIMULATOR MANAGEMENT PRS-TTS31 TITLE Revision No.
2 DROPPED ROD CERTIFICATION TEST
'Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to peiform correctly following activation of a simulated malfunction. ' This transient resulted in a plant runback due to a dropped control rod. Cunditions stabilized afts. the high load limit was cleared. The event was initiated by dropping a control rod in core location L-6 and allowing the plant to runback until the high load limit condition was cleared. When the rod was dropped, an out-inhibit condition was indicated on the Diamond control panel and a high load limit runback was initiated. Reactor Coolant System pressure increased high enough during the runback to open the Pressurizer Spray Valve. The transient was terminated when Reactor Coolant System pressure and temperature were stable and the high load limit was cleared. Dynamic response was compared to the dynamic response of the most recent certification test. During the c'onduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator license for l TMI Unit 1. Mentified deviations from expected performance have been evaluated for impact and corective action in l acordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evalautions are as follows: j n I I ()\ The results or this test are satisfactory, based upon the following: j i CRITERIA #1 ) i The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. j The simulator satisfied this requirement with no identified deviations. ! CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant owuld have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. l
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- ES.g GPU Nuc;'** mc.
l v' %00ielown, PA Ccpyngm ,gg i e
i
THREE MILE ISLAND NJmber TRAINING & EDUCATION muets4n. SIMULATOR MANAGEMENT PRS-TTS31 TITLE Revison No.
I DROPPED ROD CERTIFICATION TEST
%s l Corrective Actions / Comments: .
NONE Completed by:
. Charles E. Husted 618[T7 Date Approved by: N Operations Training Manager Robert Pamel! "7h7 Date
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[h) g sj ES-3 GPU Nudear,Inc. Middletown, PA Copynght 1997 l
.e THREE MILE ISLAND rannber
^h NUCLEAn TRAINING S EDUCATION SIMULATOR MANAGEMENT PRS-TTS32 ,
TITLE Revision No. . UNCOUPLED ROD CERTIFICATION TEST _
. THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identifica% Uncoupled Rod Certification Test Transient Test TTS32 ANSI /ANS 3,5 Reference (s): 3.1.2(12)
Test Date: 11/25/96 Malfunction (s) Tested *: RD0165 Rod Drop- Group 8 (
- Refer to Malfunction Cause and Effects Documents for options available. t Test initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle I -
This test was terminated with one Group 8 rod uncoupled and reactor power at 100% ; Plant stability was maintained. ) Simulator Run Time: . 6 minutes Simulator Evaluation Time: 30 minutes Baseline Evaluation Data: Right Direction Analysis. Simulator Malfunction Cause and Effects Document OverallTest.Results: SATISFACTORY Results Descr!otion: , in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated by uncoupling and dropping a Group Eight Axial Power Shaping Rod. This is a part-length poison rod. This caused a flux shift and power increase as seen by the power range nuclear instruments. Average core j power increased slightly and then was reduced to 100% by the ICS. l The transient was terminated with the plant stable at 100% power and at a slightly less negative imbalance than at the start of the test due to the positive reactivity added by dropping the part-length rod to the bottom of the core. l h g
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E5-1 GPu Nuclear, Inc. Mddletown, PA Copynght 1997
l
- h. THREE MILE ' ISLAND Numbw I
CW NUCLEAs TRAINING S EDUCATION SIMULATOR MANAGEMENT PRS-TTS32 RE Revtsion No. UNCOUPLED ROD CERTIFICATION TEST This test differed from the test previously conducted in 1993 due to the installation of lower poison content power-shaping rods and a longer-lived core. The combination of flatter power peaking, doppler feedback and lower worth rods all contributed to a reduction in malfunction impact. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel - representing TMl Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses f:r TMI Unit 1. Identified deviations fcom expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
]
i All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with one ) identified deviation. , CRITERIA #2 I i [')i t ( ,/ The simulator shall not fall to cause an alarm or automatic action if the reference plant would tave caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the ] reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. Corrective Actiona/ Comments: Description- Group Eight Rod Bottom Light incorrectly Lighted Msvlations: One Corrective Actions Discrepancy Report R95-1019 Completed by: h- William A. Fraser 6 92 Date Approved by: Operations Training Manager Robert Pamell [!A[4 7
'Date I
CN I E5-2 GPu Nuclear, Inc. ( Middletown, PA copynght 1997 l f
;, 1['ggp3(g THREE MILE CLAND TRAINING G EDUCATION Numbw NUCLKAn SIMULATOR MANAGEMENT PRS-TTS33 TITLE ,
Revision No. 2 INABILITY TO DRIVE RODS CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Inability to Drive Rods Certification Test Transient Test TTS33 ANSI /ANS 3,5 Reference (s): 3.1.2(12) ! Test Date: 11/25/96 - Malfunction (s) Tested *: RD10G in-Motion Command Block (Auto) - Group 7 RD11G In-Motion Command Block (Manual)- Group 7 j l
- Refer to Malfunction Cause and Effects Documents for options available. {
Test initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle This test was terminated with the Reactor at 100% power and Tave at 581*F following verification of plant stability. Simulator Run Time: 18 minutes Simulator Evaluation Time: 1 hour Baseline Evaluation Data: Previous Certification Test dated 02/22/93 ) Right Direction Analysis. I Simulator Malfunction Cause and Effects Document l Current Controlled Copies of: Reference Plant Alarm Response Procedures OverallTest Results: SATISFACTORY ES-1 GPU Nuclear,Inc. f ( f MN3dletown, PA (/ Copynght 1997
( THREE MILE ISLAND Numbw TRAINING & EDUCATION NUCLEAM SIMULATOR MANAGEMENT PRS-TTS33 TITLE Revlsion No.
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{ 2 l INABILITY TO DRIVE RODS CERTIFICATION TEST l Results Description; in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated by blocking all in-motion commands to Control Rod Group Seven. Rod Control was placed in manual and an attempt was made to insert rods. This was unsuccessful, as expected. Rcd Control was retumed to the . cutomatic mode, and the Unit Load Demand signal was reduced to 600 megawatts electric to attempt to reduce n%nt i load. Reactor power remained constant while the Secondary Plant responded to the power reduction. When Neutron l Error reached five percent a Cross Linilt occurred which caused the Integrated Control System to switch to the Tracking mode, canceling the power reduction. Cross Limits due to reactor power being five percent greater than the demanded signal repeatedly came in over the next few minutes. Final plant stability was achieved with a e, lightly higher Tavo, Feedwater flow slightly lower, and reactor power slightiv lower. The transient was terminated following verification of plant stability in the off-normal configuration. This certificct'on test differed from the previous test due to a difference in the Unit Load Demand setting. The previous test had ten percent selected in the initial Condition. The present 10-16 has 0.25 percent selected. This dampened the transient response. During the conduct of this test simulater dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ArvSI/ANS 3.5 criteria by GPUN personnel presenting TMI Simu!ator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses r TMi Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in cordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results.07 these evaluations are as follows: The results or this test are satisfactory, Msed upon the following: CRITERIA #1 ? The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. GRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no j identified deviations. f g
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ES-2 GPU Nuclear. Inc.
\ Mddletown. PA Copynght 1997
, 1 * # THREE MILE ISLAND Numtw TRAINING G EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS33 TITLE' --
W No. INABILITY TO DRIVE RODS CERTIFICATION TEST Corrective Actions /C~~ mig; NONE
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Completed by: , William A. Fraser $-t-if Date 1 Approved by: Operations Training Manager Robert Parnell [r!k/47 Date ' l l l l O l l \ l i j E5-3 i GPU fluclear, Inc
\
Mtdetown, PA r/s; w ht1997 J
d.
/* THREE MILE ISLAND tArabw l ( TRAINING O EDOCATION i NUCLEAn SIMULATOR MANAGEMENT PRS-TTS34 l TITLE gge m, g :
l FAILED FUEL CERTIFICATION TEST i
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l l THREE MILE ISLAND ' i SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Failed Fuel Certification Test Transient Test TTS34 ANSI /ANi;LEReferord 3.1.2(14) . Test Date: 11/25/96 Malfunction (s) Tested *: TH01 Failed Fuel 100% (100% - Categoy 10 Release)
' Refer to Malfunction Cause and Effects Docr.ments for options availadle.
l 1 Test initialization: Protected initial Condition 10-15 j 100% Reactor Power Equilibrium Xenon End of Cycle This test was terminated with the maximum amount of activity released to the RCS. Containment radiation levels and conde nser off-gas activities were elevated. Simulator Run Time; 20 minutes Simulator Evaluation Time: 1 hour i Baseline _EraluatloftData; Previous Certification Test TTS34 dated 2/22/93. J Right Direction Analysis. Simulator Malfunction Carse and Effects Document , Current Controlled Copiera of: i Reference Plant Alarm Response Procedures Reference Plan'. Alarm Response Procedures QyarAILInsMasults; SATISFACTORY l Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perfonn correctly following activation of a simulated malfunction. This transient resulted in the maximum amount of activity released from the fuel to the RCS. The plant remained at 100% power throughout. (N
- E5-1
( f GPu Nuclear, Inc. (j Mdsetown, PA Copynght 1997 (
Q THREE MILE ISLAND l Number TRAINING & EDUCATION MUCLEAR SIMULATOR MANAGEMENT PRS-TTS34 Tm.E Revision No. 2 1 FAILED FUEL CERTIFICATION TEST j
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The event was initiated by failing all fuel assemblies to induce the maximum activity release to the RCS. Reactor Coolant System activity increased, which was detected by the Letdown System radiation monitors. Letdown was automatically isolated by the high Letdown activity interlock. Condenser off-gas activity increased to alarm conditions due to baseline leakage in the OTSGs. The condenser off-gas MAPS sampler actuated on high alarm from the off-gas monitor. Radiation monitors went into high alarm on all Containment Radiation Monitors. The transient was terminated following verification c,7 proper indications and intedock actuations. The Radiatice Monitoring System response differed from the 2/22/93 test due to discrepancy corrections dealing with monitor response to RCS activity changes. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses f:rTMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not vivate the physicallaws of nature, b All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or autornatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. D. ES-2 [ j . GPU Nuclear. Inc. k" / Mddletown. PA Copynght 1997 (
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J THREE MILE ISLAND Number TRAINING & EDUCATION NUCLEAs SIMULATOR MANAGEMENT PRS-TTS34 TITLE Revision No. FAILED FUEL CERTIFICATION TEST Corrective Actions / Comments: NONE 1 1 Completed by: // vvv - William A. Fraser I- 17-f p 9gg, l Approved by: & Supervisor.
- Robert Pamell 5 17[7 7 Tiulator Training Date l
ES-3 i GPU Nuclear, Inc. l ( Mdcnetown, PA Copynght 1937 L____________----
)
THREE MILE ISLAND Number
, gg TRAINING & EDUCATION NUCLEAR SIMULATOR MANAGEMENT PRS-TTS36
]
_E Revision No. l GENERATOR TRIP CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Generator Trip Certification Test Transient Test TTS36 l i ANSt/ANS 3.5 Reference (s): 3.1.2(16) Test Date: 11/26/96
]
1 Malfunction (s) Tested *- EG028 Generator Trip - 86GS Exciter Ground j
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle i nt of fest Termination: This test was terminated with the reactor tripped at stable hot shutdown conditions. Simulator Run Time: 15 minutes Simulator Evaluation Time: 1 hour and 30 minutes Baseline Evaluation Data: Previcus Certification Test TTS36 dated 5/11/93. Right Direction Analysis. Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY l Results
Description:
In cccordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. ] This transient resulted in a turbine / reactor trip which proceeded to Hot Shutdown conditions. The event was initiated by causing an exciter field ground which caused a Generator and Turbine trip. The reactor tripped due to the turbine and r cctuated Reactor Trip Containment Isolation. The plant then proceeded to Hot Shutdown conditions with steam pressure l being controlled by turbine bypass valves and Main Feedwater controlling OTSG levels at setpoint.
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- O E5-1 GPu Nuclear,Inc.
Middletown. PA i Copynght 1997 l
/J"'N THREE MILE ISLAND Number (GPU NUCLEAa TRAINING & EDUCATION SIMULATOR MANAGEMENT PRS-TTS36
> E Revision No. 1 GENERATOR TRIP CERTIFICATION TEST The transient was terminated with the plant approaching normal post-trip Hot Shutdown conditions. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMt Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in ccc;rdance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this tcst are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the (o)
%/
reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE C:mpleted by: W ' AF William A. Fraser [ d-4/ Datu Approved by:
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Robert Parnell d 2 Operations Training Manager Date l V E5-2 GPu Nuclear, Inc. Middletown. PA CGpynght 1997 i
THREE MILE I! LAND Number
'y TRAININ2 & EDUCATION
, NUCLEAR SIMULATOR MANAGEMENT PRS-TTS37 TITLE Rewston No.
1 [%} INADVERTENT OTSG ISOLATION CERTIFICATION TEST l V THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Inadvertent OTSG lsolation Certification Test Transient Test TTS37 ANSI /ANS 3.5 Reference (s): 3.1.2(17) Test Date: 11/26/96 Malfunction (s) Tested *: IC34B Inadvertent OTSG B H!gh Level isolation Actuation
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition IC-15 1 k 100% Reactor Power Equilibrium Xenon Beginning of Cycle oint of Test Termination: This test was terminated with the reactor tripped and Emergency Feedwater f maintaining OTSG levels. Simulator Run Time: 20 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test TTS37 dated 5/11/93. I Right Direction Analysis. { Simulator Malfunction Cause and Effects Document j Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant P & ID Prints Overall Test Results: SATISFACTORY Results Derriotion: In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the isolation of Feedwater to the OTSG 18 and a reactor trip on high pressure. Emergency Feedwater maintained OTSG levels and removed decay heat. I p ES-1 GPU Nuclear, Inc. ( ~ Middletown, PA j Copynght 1997 1
THREE MILE ISLAND Numtw
.. TRAININ3 & EDUCATION arUCLEAsr SIMULATOR MANAGEMENT PRS-TTS37 TITLE Revis6on No.
INADVERTENT OTSG ISOLATION CERTIFICATION TEST y
- T; .e event was initiated by causing the OTSG 1B to isolate Feedwater on a spurious false high level. This caused the Feedwater block valves FW-V-5B and FW V-92B to begin closing. Loop B Feedwater flow began decreasing rapidly.
The Feedwater flow rapidly increased to the OTSG 1 A, increasing its level s'gnificantly. Integrated Control System sensed the Feedwater flow error signal and initiated Cross Limits in an attempt to control the transient. Reactor power stirted to decrease and Loop A Feedwater flow received a signal to increase. Reactor power was continuing to decrease due to the Cross Limits from the loss of Feedwater flow. Generator levels decreased and the reactor tripped on high prIssure, initiating Reactor Trip Containment isolation and a Main Turbine trip. OTSG levels continued to boil down until Emergency Feedwater actuated on low OTSG levels. The OTSG levels were subsequently recovered and maintained at setpoint. This certification test differed from the previous tests in that the operating make-up pump, MU-P-1B, tripped near the end of the test.. RCS pressure and pressurizer level dipped lower than in the previous test due to a lower Tav (approximately 2.5'F) existing prior to stabilizing at approximately the same values. The decreased pressure allowed a higher makeup flow rate from the makeup tank for a longer period of time. This emptied the tank faster, cavitating MU-P-1B and causing the pump to trip.
. Tha transient was terminated following OTSG parameter stabilization and verification of overall plant stability.
1 During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system Ectuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in tecordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: ( CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the pnysical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the t reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluafbrs concurred that during this test, the simulator satisfied this requirement with no ! identified deviations. i I. E5-2
. GPu Nuclear, Inc.
M4dietown, PA Copynght 1997
THREE MILE ISLAND l Number
*g1. * . wuctsAn TRAININ3 a EDUCATION SIMULATOR MANAGEMENT PRS-TTS37 TITLE Revision No. /N 1
( ) INADVERTENT OTSG ISOLATION CERTIFICA ciON TEST D Corrective Actions / Common',t; NONE Completed by: AM ~- William A. Fraser M-M-77
/ Date Approved by: Robert Pamell [,[/5'/f 7 Operations Training Manager Date d .
G 6 4 E5-3 M t n. N Ccpyngnt1997
THREE MILE 1: LAND ' Number TRAINING 8 EDUCATION l
*i WUCLEAn SIMULATOR MANAGEMENT PRS-TTS38 N '
Revison No. INADVERTENT OTSG OVERFEED CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Inadvertent OTSG Overfeed Certification Test Transient Test TTS38 , ANSI /ANS 3.5 Reference (s): 3'.1.2(17) Test Date: 11/26/96 Malfunction (s) Tested *: FW12A Main Feedwater Valve Failure - FW-V-17A 100% (100% = Full Open)
- Refer to Malfunction Cause and Iffects Documents for options available.
Test initialization: Protected initial Condition IC-17 . l 100% Reactor Power Equilibrium Xenon End of Cycle in f Tes T rmin lon* This test was terminated with the reactor tripped. Plant parameters were approaching I stable post-trip conditions.- i Simulator Run Time: 20 minutes i Simulator Evaluation Time: 1 hour
)
Baseline Evaluation Data: Previous Certification Test TTS38 dated 4/13/93. I Right Direction Analysis. Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant P & ID Prints Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.51985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of'a simulated malfunction, i This transient resulted in unbalanced OTSG heat transfer conditions during power operations. Decreased RCS { . t mperatures produced positive reactivity effects which caused reactor power to increase.
}C ES-1 GPU Nuclear. Inc.
Middletown. PA Copyright 1997
THREE MILE ISLAND Number
.TRAININ2 a EDUCATION
~~ _aruCLEAN SIMULATOR MANAGEMENT PRS-TTS38 TITLE ,
ReWeion No; INADVERTENT OTSG OVERFEED CERTIFICATION TEST U The event was initiated by failing the Loop A Main Feedwater regulating valve (FW-V-17A) to the 100% open position. This pction produced a rapid increase in Loop A Feedwater flow and a reduction in Loop B Feedwater flow (supplied from the same piping header). The Integrated Control System (ICS) automatically increased both Main Feedwater Pump speeds to restore Main Feedwater Valve differential pressure to 35 PSIG, resulting in additional total Feedwater flow to the OTSGs, and reduction in RCS temperatures. The integrated Control System re-ratioed Loop A and Loop B Feedwater Demands in response to the difference in col.d leg temperatures produced by the transient flow conditions. The ICS Feedwater subsystem also responded to the increase in total Feedwater flow by reducing both Loop A and Loop B feedwater Demand signals. These actions t:rminated the decrease in RCS temperature and the resulting increase in reactor power. Since FW-V-17A did not respond to this ICS action, the Loop A Feedwater flow reduction was affected by automatic throttling of the startup Feedwater Valve FW V 16A. Approximately two minutes after FW V-17A failed open, the Main Turbine transferred to Hand control on header pressure crror. The turbine control valves had opened fully yet were unable to control header pressure. OTSG 1 A pressure e,xceeded 950 psig immediately prior to the reactor trip on overpower. After the reactor tripped, fw-v-Sa automatically closed, isolating flow through FW-V-17A. At the teanination of the test both primary and secondary systems were approaching standard post-trip conditions. The results of this transient test differed significantly from previous tests due to enhanced flow characteristics on FW-V-17A and FW V-17B; the installation of the Digital Turbine Control System and previously unidentified OTSG energy transfer characteristics,
,. ,-Curing the conduct of this test simulator dynamic response, annunciator operation and automatic safety system
( -yuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by GPVN personnel Vpresenting TMl Simulator Training. The TMl representatives bold or have held NRC Senior Reactor Operator licenses for TMI Unit 1, Identified revelations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CBITERIM 1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with one identified deviation. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the ! - alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant wo,uld not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with one identified deviation. E5-2 t $, N I . M Copynght 1997 8
THREE MILE ISLAND Number
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TRAININ2 a EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS38 TITLE RevWon No. INADVERTENT OTSG OVERFEED CERTIFICATION TEST
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Corrective Actions / Comments: Desctfotion: OTSG Pressure Response Number ofIdentitled Deviations: One Qorrective Actiops: RETACT UPGRADE
Description:
Main Turbine Switched to Manual Control Number ofidentitled Deviations: *One 6 Phase il DTCS Installation C:mpleted by: I ~~ William A. Fraser -/If7page v' - 1 6!8b7 Aporoved by: - Robert Pamell Supervisor, Simulator Training Date ( 4 i ES-3 l
\ uNt$,'IA Copynght 1997
THREE MILE ISLAND Numter h NUCLEAR TRAINING & EDUCATION SIMULATOR MANAGEMENT PRS-TTS39 TITLE RevieJon No. 9 PRESSURIZER LEVEL CONTROL FAILURE CERTIFICATION TEST 1 (V THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Pressurizer Level Control Failure Certification Test Transient Test TTS39 ANSI /ANS 3.5 Reference (s): 3.1.2.(18) Igat Date: 11/26/96 MP'f9Detion(s) Tested *: MUO5 Pressurizer Level Control Valve Fails MU-V-17 100% (100% - Full Open)
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle This test was terminated with pressurizer level increasing due to increased Makeup flow rate. The Pressurizer Spray Valve RC-V-1 operated to limit the increase in RCS pressure. Simulator Run Time: 9.5 minutes Simulator Evaluailon Time: 30 minutes Baseline Evaluation Data: Previous Certification Test dated 2/24/93 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures OverallTest Results: SATISFACTORY l i ES-1 GPU Nuclear, Inc. k Mddletown PA i Copyngnt1997
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, THREE MILE ISLAND Numbw TRAINING G EDUCATION sucts'AA SIMULATOR MANAGEMENT PRS-TTS39 TITLE Revision No, 1
PRESSURIZER LEVEL CONTROL FAILURE CERTIFICATION TEST
' Results
Description:
In accordance with ANGI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in an increasing pressurizer level with the spray valve controlling RCS pressure. The event was initiated by failing to open the normal Makeup Valve MU-V-17. This failure resulted in an increase in Makeup flow to the RCS. Pressurizer levelincreased, raising RCS pressure. The Pressurizer Spray Valve RC-V-1 opened to reduce pressure. Pressurizer temperature decreased due to the combined effects of spray flow and the
-insurge from the hot leg.
The transient was terminated with pressurizer level increasing, the spray valve open and RCS pressure retuming to the normal value. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMl representatives hold or have held NRC Senior Reactor Operator license for TMl Unit 1. Identified devutions from expected performance have been evaluated for impact and corrective action in a.ccordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: [m) s
\ CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. ! { CRITERIA #2 ' The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. 4 i l ! 1 E5-2 GPu Nuclear, Inc. (#/ Middletomi, PA Copynght 1997
> , THREE MILE ISLAND Numtw
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TRAINING & EDUCATION NUCLEAN . SIMULATOR MANAGEMENT PRS-TTS39 TITLE Revision No. 1 PRESSURIZER LEVEL CONTROL FAILURE CERTIFICATION TEST l l Corrective Actions / Comments: NONE Completed by: #C - W. A. Fraser [a J~-9'J Approved by: Operations Train:ng Manager Robert Pamell [ ' 5 97 bate i L a f U l [N } ES-3 GPU Nuclear,Inc. (j Mddletown. PA Copyngnt 1997
, THREE MILE I! LAND Number wuctaAs Q TRAININ2 & EDUCATION SIMULATOR MANAGEMENT PRS-TTS40 TELE Revision No.
1 PRESSURIZER HEATER FAILURE CERTIFICATION TEST THOEE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Pressurizer Heater Failure Certification Test Transient Test ' TTS40 ANSI /ANS 3.5 Reference (s): 3.1.2.(18) Test Date: 11/26/96 Malfunction (s) Tested': RC32A/B/C Pressur! or Heater Fails Off SCR Bank 1/2/3
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 10-15 100% Reactor Power . - Equilibrium Xenon Beginning of Cycle fT T rm 1 This test was terminated with pressurizer temperature and RCS pressure slowly decreasing due to the loss of SCR controlled pressurizer heaters. Simulator Run Time: 15 minutes j Simulator Evaluation Time: 20 minutes Baseline Evaluation Data: Previous Certification Test dated 5/17/93 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY l I I I b E5-1 I (s GPu Nuclear, Mc. Mddletown, PA Copynght 1997
, THREE MILE ISLAND Numter TRAININ3 & EDUCATION
' mucurAN SIMULATOR MANAGEMENT PRS-TTS40 TJ1LE Revtsion No.
1 PRESSURIZER HEATER FAILURE CERTIFICATION TEST Results
Description:
In accordance with' ANSI /A14S-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. Thb transient resulted in a slowly decreasing pressurizer temperature a'nd RCS pressure due to ambient losses and spray valve bypass flow. ' The cvent was initiated by falling the normal SCR-controlled Pressurizer heaters off. Pressurizer temperature and
. pressure decreased as a result of ambient losses and spray valve bypass flow.
The transient was termina'ad following verification of the resulting temperature and pressure reduction. During the conduct oC this test simulator dynamic response, annunciator operation, and automatic safety system cctuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by select GPUN personnel a representing TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator license f:r TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in cccordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows:
. l The results or this test are t, satisfactory, based upon the following:
CRITERIA #1 (p) V The observable changes in simulator parameters shall correspond in diro:: tion reference plant and do not violate the physicallaws of nature. All CRITERIA 1 evaluators concurred that during this test, the simulator. satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or s.utomatic sction if the reference plant would not cause an alarm or automatic action. All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. m l ES-2 ( GPu Nuclear,Inc. g Mkksetowrt PA Copynght 1997 t _ _ _ _ _ _ _ _ _ _ _ _ _
r .
,, THREE MILE ISLAND Number g -
TRAINING & EDUCATION PRS-TTS40 NUCLEAR SIMULATOR MANAGEMENT f l TAILE Revision No. 1 i PRESSURIZER HEATER FAILURE CERTIFICATION TEST ) i Corrective Actions / Comments-l NONE Completed by: / Y W. A. Fraser '9/ Dafe Approved by: - /> Operadons Training Einager Robert Pamell f7 Date l l l
)
1 1 l L I ES-3
%/ GPU Nuclear, Inc.
! Middletown, PA ( D IM l
A f THREE MILE ISLAND Number lGPU ' NUCLEAa TRAINING & EDUCATION SIMULATOR MANAGEMF.NT PRS-TTS41 TjTLE Revisen No.
/ )/ REACTOR TRIP CERTIFICATION TEST 2
THREE MILE ISLAND SIMULATOR MANAGEMENT
\
CERTIFICATION TEST ABSTRACT Test identification: Reactor Trip Certification Test Transient Test TTS41
)
ANSl!ANS 31Heference(s): 3.1.2.(19) Test Date: 11/2&96 Malfunction (s)lested*: RD29 Reactor Trip Test initialization: Protected Initial Condition 10-17 ) 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated fo; lowing reactor trip with ed nary and secondary plant pressures and temperatures approaching normal po:t trip values. elator Run Time: 10 minutes Simulator Evaluation Time: 5 hours l Baseline Evaluation Data: Previous Certification Test dated 05/11/93 Right Direction AnaFysis Simulator Malfunction Cause and Effects Docanent Current Controlled Copies of: Reference Plant A! arm Re:ponse Procedures Reference Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures I
. Q191311 Test Resuhs: SATISFACTORY I
i l 1 ( n51 EE.N Copynght 1997 I L
/ THREE MILE ISLAND Number TRAINING & EDUCATION AfuCLEAN SIMULATOR MANAGEMENT PRS-TTS41 f ' TITLE Revison No. / T 2
( U
/ REACTOR TRIP, CERTIFICATION TEST Results
Description:
I J in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to l perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip. The teactor trip initiated'a Main Turbine trip and Reactor Trip Containment isolation.
. The event was initiated by activation of the reactor trip malfunction. The reactor tripped resulting in a Main Turbine trip.
R: actor Trip Containment isolation closed specified Containment isolation Valves. The integrated Control System
- i. responded by initiating a runback due to Feedwater Cross Limits causir,g the Main Feedwater System to reduce flow and establish normal post trip level in the OTSGs. Reactor Coolant System pressure and temperature decreased to post trip viluis and were maintained.
The transient was terminated when primary and secondary pressure and temperature were approaching post trip stable conditions. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by GPUN personnet repriset. ting TMI Simulator Training. The TMI representatives have held an NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corective action in acordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evalautions are as follows:
-( O The resuJts or this test are satisfactory, based upon the following:
CRITERIA g1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant owuld have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would rut cause an alarm or automatic action. The simulator satisfied th*s requirement with no identified deviations. I l
.(A) 8 ES-2 (v/ GPU Nudear. Inc.
Mddletown. PA Copynght 1997
l l
/ 1HREE MILE ISLAND Number TRAINING & EDUCATION !
MUCLEAn SIMULATOR MANAGEMENT PRS-TTS41 i TITLE Revison No.
. 2 REACTOR TRIP CERTIF10ATION TEST
_ { 1 Corrective Actions / Comments: NONE Completed by: M///97 Charles E. Husted
- I, i gggg i ,
Approved by: Operations Training Manager Robert Pameli T[/Ih7 Da'te 1 I p O 4 0 [3; 1 E5-3 eeus. . y/ 14ddletown, PA Copynght 1997
l
.* THREE MILE CLAND Number
]
TRAININ3 & EDUCATION AruCLEAg SIMULATOR MANAGEMENT PRS-TTS43 TITLE Revtston No. 1 MAIN STEAM LEAK OUTSIDE THE RB CERTIFICATION TEST 1 l THREE MILE ISLAND SIMULATOR MANAGEMENT f CERTIFICATION TEST ABSTRACT { ' Test identification: Main Steam Leak Outside the RB Certification Test Transient Test TTS43 ANSI /ANS 3.5 Reference (s): 3.1.2(20) IftgtOste: 11/26/96 Malfunction (s) Tested *: MS03B Main Steam Leak Outside RB 20% (100% - 6,000,000 LBM/HR) 4
- Refer to Malfunction Caute and Effects Documents for options available.
Test initializa_ tion: Protected initial Condition 10-15
~
100% Reactor Power Equilibrium Xenon Beginning of Cycle fT Termination- This test was terminated with reactor power at 99%, Tave (stable at 570 degrees Fahrenheit) and generated megawatts low due to the steam leak. Simulator _Run Time: 15 minutes Simulator Evaluation Time: 1 hour i Basellne Evaluation Data: Previous Certification Test TTS43 dated 06/30/93 - Right Direction Analysis , Simulator Malfunction Cause and Effects Document j Current Controlled Copies of: 1 Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Reference Plant P & ID Prints Overall Test Results: SATISFACTORY l l E5-1 G) *M*o .'?i Copyright 1997 l
THREE MILE ISLAND Nurrber t g TRAINING G EDUCATION NUCHAN SIMULATOR MANAGEMENT PRS-TTS43 _7T!!.LE Revision No. 3 1 V[ MAIN STEAM LEAK OUTSIDE THE RB CERTIFICATION TEST Results Deswist;cn: in accordance with ANS!/ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. { { The cvent was initiated by causing a significant steam leak in the intermediate Building upstream of the Main Steam , Isolation Valves for the OTSG 18. The Main Steam Safety Valve audio monitor indicated significant noise in the ' d'ac;ed area. Main Steam header pressure dropped. Generated megawatts decreased as the Main Turbine Control V;lves throttled to recover header pressure. The Integrated Control System sensed the decrease in megawatts and sent signals to the Fedwater and Reactor subsystems to increase their demands. Control rods reached their outlimits but Feedwater flow continued to increase. ICS BTU Limits and Cross Limits then actuated to reduce Feedwater flow to the OTSGs. ROS pressure and temperatures then increased and the spray valve actuated to dampen the pressure increase. The transient was terminated with a reactor power - megawatt mismatch, and the plant trending toward a stable condition with the steam leak in progress. This trst differed from the previous certification test. The system response was impacted by the rapid response of the installed Digital Turbine Control System to error signals. This held up generated megawatts which in tum held up steam flow causing a lower T-ave, initiall higher power spike, and RCS pressure response. Conditions at the termination of the test, however, were similar. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system ' actu;tions were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by select G UN personnel p)senting ( ll Unit 1. TMI Simulator Identified Training. deviations from The TMI have expected performance representatives hold and been evaluated for impact or have correctcheld action in NRC Se L.ardance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in sfmulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature.
- The simulator satisfied this requirement with no identified deviations.
CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action it the reference plant owuld have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations.
/ ES-2 I
( GPu Nuclear, Inc. Middletown. PA Copynght 1997
THREE MILE ISLAND Number { t TRAININ3 G EDUCATION d MUCLEAn SIMULATOR MANAGEMENT PRS-TTS43 TITLE Ree No. MAIN STEAM LEAK OUTSIDE THE RB CERTIFICATION TEST
.V Corrective Actions / Comments:
'NONE Completed by: W. A. Fraser b #!f7
. Approved by' Robert Pamell Operations Training Manager Date I
i 4 w l i T E5-3
). ,
oPU Nucioar, km. Mddletown, PA Copynght 1997
I NOT14 7
) THREE MILE ISLAND S.IMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Control Rod Movement Surveillance Certification Test Normal Operations Test NOT14 ANSI /ANS 3.5 Reference (s)- 4.2.1 3.1.1(5) l 3.1.1(10) 5.4.1(2) l Test Date.1 11/20/96 Reference Plant Procedure: SP 1303-3.1, Control Rod Movement Surveillance Revision 21 Test Initialization: Protected Initial-Condition IC-17 100% React'or Power 579 F Tave Equilibrium Xenon End of Cycle l ) Point of Test Termination: This test was terminated following
'/ completion of surveillance.
Simulator Run Time: 30 minutes Baseline Evaluation Data: Tactile Feel by Experienced Operator. Current Controlled Copy of: Reference Plant Surveillance Procedure Reference Plant Alarm Response { Procedure i I Qverall Test Results: SATISFACTORY l t Results
Description:
J In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using
. only operator action normal to the referenced plant.
During the conduct of this test simulator dynamic performance, i annunciator operation and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. N __,/ w
s *
,k NOT14 ..
I 'l Tactile feel of simulator controls used during the test was also d evaluated. These evaluations were conducted by GPUN personnel representing TMI Simulator Training. All the evaluators have held an NRC Senior Reactor Operator license . ) for TMI Unit 1. Procedure steps not performed and identified i deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: l CRITERIA #1: The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator supported this requirement with no deviations. CRITERIA #2: The observable changes in simulator parameters correspond in
,m direction to those expected from the actual reference plant and do not violate the physical laws of nature.
[(s-) The simulator supported this requirement with no deviations. CRITERIA #3: The simulator shall not fail to cause an alarm or automatic .I actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator supported this requirement with no deviations. CRITERIA #4: Tactile feel of Plant-Referenced Simulator control device (s) compare (s) to those of the reference plant. The simulator supported this requirement with no deviations. j' %, Q_,l l
~
jg. l _ NOT14- > ym I c
). Corrective Actions /Commentsi During the conduct of this test, reference plant procedure steps not performed, were identified and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure, _ 6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result '-in the need _ for the test operator to violate the procedure in order .to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of . the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
. 4. In accordance with 6211-ADM-2820.02, these test exceptions have been documented and closed.
i-
/:
-Completed By _
- William A. Fraser Date 12/04/96 Approved By Robert L. Pa'rnell Date b N l Supervisor, Simulator Training kw ): .
k___,________ ____..__-_m. - - - - - -
l THREE MILE ISLAND Numbst TRAINING G EDUCATION wuctrAs SIMULATOR MANAGEMENT PPS-NOT15 TITLE p Revison No.
1 ( RB COOLING AND ISOLATION SYSTEM LOGIC CHANNEL AND COM*)ONENT U SURVEILLANCE CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: RB Cooling and Isolation System Logic Channel and Component Surveillance Certification Test Normal Operations Test NOT15 ANSI / ANS 3.5 Reference (sh 4.2.1 3.1.1(5) 3.1.1(10) ItaLData; 11/14/96 Reference Plant Procorture: SP 1303-3.1, Reactor Building Cooling and Isolation System Logic Channel and Component Test Revision 45 Test initialization: Prc.tected initial Condition 10-07 Hot Zero Power 1E-8 Amps Xenon Free (/ Beginning of Cyc4 Point of Test Termination: Completed e n ia at Automatic Actuation Channel A and Manual Actuation Channel A. ' Simulator Run Time: 5 Hours l Baseline Evaluation Data: Previous Certification Test dated 05/05/93 Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures Reference Plant Alarm Response Procedures l Overall Test Results: SATISFACTORY j Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was condacted to demonstrate the ability to operate the l simulator in accordance with similar reference plant operating procedures, using only operator actior. normal to the i reference plant. / e
\ E4-1 da o E.PA Copyngnt 1997
. THREE MILE ISLAND Number TRAINING & EDUCATION NUCLEAsr SIMULATOR MANAGEMENT PRS-NOT15
~ TITLE Revison No.
[, m 1 V) RB COOLINGSURVEILLANCE AND ISOLATION CERTIFICATION SYSTEM TEST LOGIC CHANNEL AND CO During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMl representatives have held NRC Senior Reactor Operator license for TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance enteria of the reference plant procedure (s) used during the conduct of this test. The simulator satisfied this requirement with no identified deviations. CRITERIA #3 . The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. O CRITERIA #3 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. l i h g / E4-2 GPU Nuclear. Inc. f (/ Mddletown. PA ( Copynght 1997 i l t _ - - - - - - - - - - - - - - - . - - - - - -
- i
, ,, THREE MILE ISLAND Number 1 RAINING G EDUCATION muets4a SIMULATOR MANAGEMENT PRS-NOT15 TITLE Revtson No.
Q ; RB COOLING AND ISOLATION SYSTEM LOGIC CHANNEL AND COMPONENT 1 !(_ / SURVEILLANCE CERTIFICATION TEST Corrective Act,l.ons(Commen,is: NONE During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and evalauted for impact and corrective action requirements in accordance with Functional Fidelity Procedure, 6221-ADM-h 2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and .
precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
l Completed by: Charles E. Husted /// 77
'I Ddte '
Approved by: n
~ Operations Training Manager Robert Pamell I//6/97 Date
{ i E4-3 (- GPU Nuclear,Inc.
# Mddletown, PA Copynght 1997
~
THREE MILE ISLAND Numbw gg TRAINING S EDUCATION NUCLEAm SIMULATOR MANAGEMENT PRS-NOT16 TITLE Revision No. G 1 LOADING SEQUENCE AND COMPONENT AND HPI LOGIC CHANNEL SURVEILLANCE CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Loading Sequence and Component and HPl Logic Channel Surveillance Certification Test Normal Operations Test NOT16 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(5) 3.1.1(10) Test Date: 11/15/96 Reference Plant Procedure: SP 1303-5.2, Loading Sequence and Component and High Pressure injection Logic Channel Test Revision 55 Test initialization: Protected initial Condition 1C-09 Hot Zero Power 1E-8 Amps O Xenon Free ( End of Cycle Point of Test Termination: Completion of Emergency Core Cooling and Loading Sequence - Automatic Actuation A and Manual Actuation A Test Sections. 1 Hot Zero Power. I 1E-8 Amps Simulator Run Time: 4 hours 30 minutes j Baseline Evaluation Data: Previous Certification Test dated 05/17/93 Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures Reference Plant Alarm Response Procedures OverallTest RePutta; SATISFACTORY Results
Description:
( in accottlance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability to operate the j simulator in accordance with similar reference phnt operating procedures, usMg only operator action normal to the stference plant. O E41
'c) *%::: %
Copynght 1997 C-- ---- ' '
THREE MILE ISLAND Numtw TRAINING G EDUCATION mucurAn CIMULATOR MANAGEMENT PRS-NOT16 TrrLE RevWon No. LOADING SEQUENCE AND COMPONENT AND HPl LOGIC CHANNEL
) SURVEILLANCE CERTIFICATION TEST During the conduct of this test simulator dynamic response, annunciator operation, and automat!c safety system actuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMl representatives have held NRC Senior Reactor Operator license for TMI .
Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Funct;onal Fidelity Procedure 6221 ADM 2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The smutator satisfied this requirement with no identified deviations. CRITERIA #2 The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violata the physicallaws of nature. ' The simulator satisfied this requirement with no identified deviations, ( CBIIERIA13 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would rat cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. CRITERIA #4 Tactile feel of Plant-Refero/nced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. l E4-2 y GPU Nuclear, Inc. Mddletown. PA Copynght 1997 ____________.__________--_-________m
THREE MILE ISLAND Numts TRAINING & EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-NOT16 TITLE Revision No. O 1 LOADING SEQUENCE AND COMPONENT AND HPI LOGIC CHANNEL SURVEILLANCE CERTIFICATION TEST Corrective Actions / Comments: NONE During the conduct of this test, reference plant procedure steps not performed were identified and evelauted for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not perforrned did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steos not performed'did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM 2820.02, these test ext:eptions have been documented and closod.
fm k ompleted by: Charles E. Husted [4!97
/ Dhte Approved by:
OperationsTraining Manager Robert Pamell /4 h 7 4 Dats
)
i i i l i I
,/ E4-3
, 1 ) e_,_. m
's_/ Mu$dletown. PA Copynght 1997 l
L _-_ _ _
3 '+ f4 ( (g, y THREE MILE ELAND TRAINING G EDUCATION Nurnber sucrJrAn SIMULATOR MANAGEMENT PRS-NOT17 TITLE Rem M l N.J [^\ HIGH PRESSURE INJECTION CERTIFICATION TEST 1 THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: High Pressure injection Certification Test Normel Operations Test NOT17 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(10) ' Test Date: 05/30/97 Reference Plant Procedure: SP 1301 11.8 High Pressure injection Revision 31 Test inillalization: Protected initial Condition 1C-03 Cold Shutdown Reactor Coolant Steam System Filled Steam Bubble Pre Heatup, End of Cycle Completion of Manual Actuation A and B Testing. Reactor Coolant System Filled. . Steam Bubble. Simulator Run Time: 6 hours Baseline Evaluation Data: Previous Certification Test dated 5/17/93. Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Surveillance Procedures Reference Plant Alarm Response Procedures OverallTest Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the ref:rence plant. During the conduct of this test simu!ator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSl/ANS 3.5 criteria by GPUN personnel l representing TMl Simulator Training. The TMl representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221 ADM-2820.02. The results of these evaluations are as follows: p r E41 GPu Nuclear. Inc. i
\ Mddletown. PA Copynght 1997 L
THREE MILE ISLAND Number
' TRAINING & EDUCATION WUCLEAN SIMULATOR MANAGEMENT PRS-NOT17 TITLE Rev6eson No.
Og 1 i HIGH PRESSURE INJECTION CERTIFICATION TEST The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used dur'ng the conduct of this test. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 - The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfi6d this requirement with no identified deviations. GBITERIA #3 The simulator shall not fall to cause an alarm or automatic actuation if the reference plant would have caused an alarm of automatic ECllon, and conversely, the simulator shall not cause an atrarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. CRITERIA #4 Tactile feel of Plant Referenced Simulator control device (s) compares to those of the reference plant. The simulator sati'sfied this requirement with no identified deviations.
)
4 1 l E4-2 i GPu Nuclear, Inc.
.b Mddletown, PA Copynght 1997
. ,, THREE MILE ISLAND Numbw TRAINING & EDUCATION wuCLEAN SIMULATOR MANAGEMENT PRS-NOT17 TITLE Revtaky No.
1 HIGH PRESSURE INJECTION CERTIFICATION TEST Corrective Actions / Comments: NONE
)
During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and cvaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution. ,
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance tvith plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
I mpleted by: ' Charles E. tuted I/7/f1 )
' D6te Approved by:-
Operations Traini5g Manager Robert Pamell h!3 97 i l Date I l l 1 l I { (
/~T
- E4-3 4
GPu Nuclear,Inc. Mddletown, PA k- copynght 1997 4 J
F ] THREE MILE ISLAND Number
# TRAINING & EDUCATION MUCLEAN SIMULATOR MANAGEMENT PRS-NOT18 TlTLE R6nson No.
1 l RB EMERGENCY COOLING SYSTEM SilRVEILLANCE CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CEFiTIFICATION TEST ABSTRACT Test Identification: RB Emergency Cooling System Surveillance Certification Test Mormal Operations Test f40T18 ANSl/ANS 3.5 Reference (s): 4.2.1 3.1.1(10) Test Date: 06/20/97 Reference Plant Procedure: SP 100311.9 Reactor Building Emogency Cooling System Revision 49 Test inillalization: Protected initial Condition IC-03 Cold Shutdown Reactor Coolant Steam System Filled Steam Bubble Pre Heatup, End of Cycle Point of Test Termination: Cold Shutdown ( N, Reactor Coolant Steam System Filled - i Steam Bubble Completed Testing of Actuation A Components Simulator Run Time: 6 hours Baseline Evaluation Data! Previous Certification Test dated 5/13/93 Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Surveillance Procedures Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the r:fererwe plant. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system { actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel i representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows:
/7 E4-1 I
) GPU Nuclear, Inc.
k'd / Mddletown, PA Copynght 1997
THREE MILE ISLAND Number 3 gg TRAINING S EDUCATION atuCLEAN SIMULATOR MANAGEMENT PRS-NOT18 TITLE Revtsion No. RB EMERGENCY COOLING SYSTEM SURVEILLANCE CERTIFICATION TEST The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator satisfied this requirement with three identified deviations. CRITERI A #2 The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #3 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. V The simulator satisfied this requirement with no identified deviations. CRITERIA #4 Tactile feel of Plant Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: Description; Unacceptable Valve Stroke Time Number of Identified Deviations: One Corrective Actions: Modification Request
Description:
Unacceptable System Flow Rate Number of identified De,yhtions: One Corrective Actiont Discrepancy Report
Description:
Unacceptable Discharge Pressure /Detta Pressure Number of identified Deviations: One ! Corrective Actions: Discrepancy Report I
/N I E4-2 GPU Nuclear, Inc.
( \ \d Middletown. PA Copynght 1997
, THREE MILE ISLAND Number TRAINING G EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-NOT18 TITLE Revision No.
1 RB EMERGE.4CY COOLING SYSTEM SURVEILLANCE CERTIFICATION TEST During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and j cvaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM- l 2820.02. The results of these evaluations are as follows: 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with .I the evolution. l
- 2. Steps not performed did not result in observable differences in the control room.
)
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
1
- 4. In accordance with 6221-ADM 28?O.02, these test exceptions have been documented and closed.
06/23/97 Completed by: o 2. Charles E. Husted Date Approved by;' Robert Parnell [! 7 p ' Operations Training Manager ' Ddte
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l
/'x{ E4 3 j GPU Nuclear. Inc.
r l
\\ Mddletown. PA l
l Copyngni 1997 I i ! I
THREE MILE ISLAND Nurnber 4-TRAINING & EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-NOT19 TITLE Revuxm No. 1 G ES SYSTEM EMERGENCY CERTIFICATION TEST SEQUENCE AND POWER TRA THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification ES System Emergency Sequence and Power Transfer Surveillance Certification Test Normal Operations Test NOT19 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(5) 3.1.1(10) Test Date: 06/20/97 Reference Plant Procedure: SP 1303-11.10 Engineered Safeguards System Emergency and Power Transfer Test Revision 31 Tee initialization: Protected initial Condition 10-05 Hot Shutdown Safeties Out Xenon Free Middle of Cycle Termination: Hot Shutdown Safeties Out Completed Testing Actuation B Components Simulator Run Time: ' 4 hours Baseline Evaluation Data: Previous Certification Test dated 05/24/93. Tactile Feel by Experienced Operators. Right Direction Anaiysis. Current Controlled Copies of: Reference Plant Surveillance Procedures ' Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the r ferenceplant. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel r: presenting TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator license for TMl Q E4-1 dd town. P Copyrght 1997
l THl1EE MILE ISLAND Number
, g TRAINING G EDUCATION mucurAn SIMULATOR MANAGEMENT PRS-NOT19 TITLE Revtston No.
1
\
9 ES SYSTEM EMERGENCY CERTIFICATION TEST SEQUENCE AND POWER TRANSl J l Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in ! accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: ' The results or this test are satisfactory, based upon the following: 4 CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this j test. j The simulator satisfied this requirement with no identified deviations. I
]
I CRITERIA #2 The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature.
)
The simulator satisfied this requirement with no identified deviations. - CRITERIA #3
)
h\j The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. ! The simulator satis' fied this requirement with no identified deviations. CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. l The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments. None l During the conduct of this test, reference plant procedure steps not periormed were identified as test exceptions and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: I
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with ;
the evolution. l I l l O , j E4 2 GPU Nucks. Inc. \\j . Middletown. PA Copynght 1997
1 1 THREE MILE ISLAND Number TRAINING & EDUCATION
,' NUCLKAN SIMULATOR MANAGEMENT PRS-NOT19 TlTLE Revmon No. '
1 S ES SYSTEM EMERGENCY CERTIFICATION TEST SEQUENCE AND POWER TRANS
- 2. Steps not performed did not result in observable differences in the control room or the difference observed had no
' impact on performance of the procedure.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2620.02, these test exceptions have been documented and closed.
I Completed by: [ L/ 1 06/24/97 Charles E. Husted Date 1 Approved by: Operations Training Manager Robert Pamell 7 Date (%,, l
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l i i i l A E4-3 f ;; GPU Nacisar. inc i x ' Modletown PA
'w Copyrigni 1997 E
l ATI'ACIIMENT 11
&~' TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR TESTING COMPLETED YEAR #4 TESTS (1997-1998)
Benchmark Tests TTS07 RCS Safety Valve Failure TTSl9 laiss of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Ste:un Leak inside Reactor Building TTS56 Manual Reactor Trip TTS57 Simultaneous Chisure of All Main Steam isolation Valves TTS58 laws of One Reactor Coolant Pump TTS59 Maximum Rate Power R;unp TTS60 laiss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Simulator Stability SSP 02 Simulator Accuracy Real Time Test
,,Dg RTT01 Real Time Test ! l \d Transient Tests TTS44 Main Feedwater Line Break inside Reactor Building TTS45 Main Feedwater Line Break Outside Reactor Building TTS46 NI-5 Failure TTS47 NI-6 Failure TTS48 Pressurizer Level Control TTS49 Feedwater Flow Transmitter Failure TTS50 RC Cold Leg Temperature Transmitter Failure TTS51 OTSG Pressure Transmitter Failure TTS52 Emergency Diesel Generator Failure TTS53 Emergency Feedwater Failure TTS54 ESAS Actuation Failure TTS55 ATWS Normal Ornations Tests NOT20 Turbine Overspeed Testir.g NOT21 Low Pressure injection NOT21 Fnmt Shutdown NOT23 Plant Cooldown NOT24 Shutdown Margin and Reactivity Balance l NOT25 Heat Balance Calculation n
L) i ? L - i
' THREE MILE ISLAND Numte TRAINING & EDUCATION muetsAs SIMULATOR MANAGEMENT PRS TTS07 TITLE Revison No.
RCS SAFETY VALVE FA; LURE CERTIFICATION TEST
.v THREE MILE ISLAND -
SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: RCS Safety Valve Failure Certification Test Transient Test TTS07 ANSI /ANS 3,5 Referenceifs): B.2.2.(10) 3.1.2(1b) - 3.1.2(1d) Test Date: 12/22/97 Malfunction (s) Tested *: ES01 A ESAS Failure to Actuate at HPl Setpoint Channel A ES01B ESAS Failure to Actuate at HIP Setpoint Channel B RC27A Pressurizer Safety Valve Fails Open 25% (100% = Full Open)
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 1015 100% Reactor Power
- l O /
Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated with the reactor tripped, Reactor Coolant System saturated, pressurizer filled and system inventory decreasing. Simulator Run Time: 10 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test TTS07 dated 11/20/96. Right Direction Analyds. - Simulator Malfunction Cause 'cd Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY I I h) y
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ES-1 GPU Nuclear,Inc. Mddletown PA Copynght 1997
THREE MILE ISLAND Number TRAINING & EDUCATION nuctmr SIMULATOR MANAGEMENT PRS-TTS07 TITLE Revision No. RCS SAFETY VALVE FAILURE CERTIFICATION TEST Results
Description:
l t in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malftt% ion. This transient resulted in the loss of Reactor Coolant System inventory through a partially open Code Safety Valve on the pressurizer. The pressure in the Reactor Coolant System decreased to saturation and the pressurizer filled solid. The Engineered Safeguards Actuation System was prevented from operation. ' The event was initiated by failing a Pressurizer Code Safety Valve partially open. As Reactor Coolant System pressure decreased the Reactor Protection System actuated tripping the reactor and initiating Reactor Trip Containment isolation. Pressure continued to decrease and Engineered Safeguards Actuation System High Pressure injection actuation setpoint was reached. The system did not actuated due to the malfunctions inserted. Reactor Coolant System pressure continued to decrease; Reactor Vessel and Reactor Coolant System voiding caused the pressurizer to fill. Reactor Coolant Drain Tank pressure increased causing rupture disk rupture with subsequent activity release to containment. During the conduct of this test simulator dynamic response, annunciator operation and autoniatic safety system actuations were recorded. GPUN personnel representing TMl Simulator Training have evaluated test results against ANSl/ANS 3.5 criteria. The TMl representatives have held NRC Senior Reactor Operator licenses for TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective aeion iri accordance r~$h Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows:
/ i '\j The results or this test are satisf actory, based upon the following:
CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. i 1 1 I tO)j E5-2 GPU t wcisar. Inc. Mddl atown. PA Copoght 1997 1
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THREE MILE ISLAND Number TRAINING & EDUCATION nuctamr , SIMULATOR MANAGEMENT PRS-TTS07
~IiiLE Ranson No. !( RCS SAFETY VALVE FAIL'URE CERTIFICATION TEST Corrective _ Actions / Comments; NONE Con,pleted bv: +
Charles E. Husted 32
/ D6te Approved by: Robert Parnell .A F / T' /
Operations Training Manager ' Date l
^
[ ES-3 dd ato n', A Copfight 1997
l 1
,, THREE MILE ISLAND Number
- TRAINING O EDUCATION
( suctsAs SIMULATOR MANAGEMENT PRS-TTS19 TITLE Reviskx) No. LOSS OF FORCED FLOW CERTIFICATION TEST 4 l -- THREE MILE ISLAND l SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT l Test identification: Loss of Forced Flow Certification Test ! Transient Test TTS19 1 l ANSI /ANS 3.5 Reference (s): 3.1.2(4) , Benchmark Test B2.2(4) Test Date: 01/07/98 Malfunction (s) Tested *: RC 35 A/B/C/D - RC Pump Trip A/B/C/D
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 10-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle
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' Reactor tripped with Reactor Coolant System heat transfe? by stable natural circulation.
ulator Run Time: 10 minutes Simulator Evaluation Time: 4 hours Baseline Evaluation Data: Previous Certification Tests TTS19 dated 11/26/96. Right Direction Analysis. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY i i I l t i E5-1
\ GPU Nuclear, Inc.
Mddletown, PA Copynght 1997
1 I 4 THREE MILE ISLAND Numbw TRAINING a EDUCATION Muctras SIMULATOR MANAGEMENT PRS-TTS19
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TITLE
- Redston No.
LOSS OF FORCED FLOW CERTIFICATION TEST 4 Results Description l l in accordance with ANSI /ANS-3.5-1905 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activatiin of a simulated malfunction. ) j
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Th'3 transient resulted in the loss of forced flow in the Reactor Coolant System. Decay heat rernoval was accomplished by the use of Emergency Feedwater flow to the OTSGs and steaming to the Main Condenser with natural circulation evident. ) The event was initiated by tripping all four Reactor Coolant Pumps on overload. The reactor tripped and initiated Reactor Trip Containment isolation. The Emergency Feedwater System actuated and filled the OTSGs to 50% level on the Operating Range. The plant transitioned to the natural circulation mode. The current revision of ATOG Procedure 1210-10 was utilized to verify evidence of natural circulation. Because this test was performed from a Beginning of Cyclo initial Condition, decay heat levels were relatively low, causing the average differential temperature between hot and cold leg temperatures to be less than the procedurally stated 30-50 degrees Fahrenheit. The transient was terminated following the verification of natural circulation flow and overall plant response utilizing i ATOG Procedure 1210-10. l During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel re resenting TMI Simulator Training. The TMI representatives hold or have held NRC Senior Reactor Operator licenses Tpl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in ordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. i All CRITERIA 1 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the i alarm or automatic action, and corrversely, the simulator shall not cause an alarm or automatic action if the l
- reference plant would not cause an alarm or automatic action. ;
i All CRITERIA 2 evaluators concurred that during this test, the simulator satisfied this requirement with no identified deviations. p Q) E5-2 GPU Meadnc. Mosetown PA Copynght 1997 [
4 THREE MILE ISLAND Numbw TRAINING O EDUCATION wuctsAs SIMULATOR MANAGEMENT PRS-TTS19 TITLE
- Revislan No.
LOSS OF FORCED FLOW CERTIFICATION TEST 4 Corrective Actions / Comments: NONE Completed by: ~ ed ' William A. Fraser 2 - 5% 9p Date Approved by: Robert Pamell I/4# Operations Training Manager date b l 4
.f 's ES-3 GPU Nuclear, Inc.
Mddletown, PA Copynght 1997 i l \ l
THREE MILE ISLAND Number g TRAINING & EDUCATION muetsAs SIMULATOR MANAGEMENT PRS-TTS27 TITLE R4Mson No. LOSS OF ALL FEEDWATER CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Loss of All Feedwater Certification Test Transient Test TTS27 ANSI /ANS 3.5 Reference 3h B.2.2.(2) 3.1.2(10) Test Date: 12/22/97 Malfunction (s) Tested *: FW15A/B Main Feedwater Pump Trip 1 A/1B FW17 Emergency FW Pump Trip (EF-P-1) FW18A/B Emergency FW Pump Trip (EF-P-2A/B) Test initialization: Protected initial Condition 1C-16 100% Reactor Power Equilibrium Xenon Middle of Cycle in fT T rmin I - This test was terminated with the reactor tripped, Reactor Coolant System heatup in I progress with Primary System pressure controlled by the prepurizer Power Operated I Relief Valve. Pressurizer level was increasing End RCS subcooling margin was I decreasing. Simulator Run Time: 10 minutes
$1mulator Evaluation Time: 4 hours Baseline Evaluation Data: Previous Certification Test TTS27 dated 11/22/96. s Right Direction Analysis. j Simulator Malfunction Cause and Effects Diagram Current Controlled Copies of:
Reference Plant Alarm Response Procedures j Reference Plant Emergency Procedures i Reference Electrical One-Line Diagrams Overall Test Results: SATISFACTORY n) ( (./ E5-1 i T57,E Copyngm 1997 r l , t :
l n 1 THREE MILE ISLAND Number TRAINING & EDUCATION nuctsas SIMULATOR MANAGEMENT PRS-TTS27 TITLE Rension No. 3 LOSS OF ALL FEEDWATER CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to , perform correctly following activat!on of a simulated malfunction. This transient resulted in the loss of all secondary side heat removal with Reactor Core heat removal via the PORV. l During the duration of this test heat transfer from the RCS was insufficient for removal of decay heat as expected. I l The event was initiated by falling both Main Feedwater Pumps and preventing the start of all Emergency Feedwater I Pumpt The Reactor Protection System tripped the reactor and initiated Reactor Trip Containment isolation. The Heat j Sink Frotection System actuated to initiate Emergency Feedwater but the malfunctions prevented the pumps from Carting. As steam generator irn 9ntory was lost Reactor Coolant System temperature increased causing system pressure { 12 increase resulting in opening of Be pressurizer spray valve then the PORV. At the end of the test the PORV was I cycling to limit Reactor Coolant System pressure and the' Reactor Coolant system was approaching saturated conditions. l The transient was terminated following OTSG dry-out, v.ith RCS heatap in progress and the PORV controlling RCS pressure. During the conduct of this test t;imulator dynamic response, annunciator operation and automatic safety system actuations were recorded. GPUN personnel representing TMI Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMl Unit 1. Identified dovlations from expected performance have been evaluated for impact and corrective action in accordance ith Functional Fidelity Proceduro,6221-ADM-2820.02. The results of these evaluations are as follows: (
' The results or this test are satisf actory, based upon the following:
CRITERIA #1 The observable changes in rdmulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the reference plant wcld have caused the alarm or automatic action, and coriversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no dentified new deviations. One previously identified deviation i remains uncorrected -TTS27E0021. I
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( ) E5-2 (/ GPU Nuclear. Inc Miodierown PA Copyright 1997
l THREE MILE ISLAND Number I TRAINING & EDUCATION suctaAs SIMULATOR MANAGEMENT PRS-TTS27 TITLE Revision No. 3 LOSS OF ALL FEEDWATER CERTIFICATION TEST Corrective Actions / Comments: NONE Completed by: Charles E. Husted
~
/ bate Approved by:
Operations Training Manager Robert Parnell O >> /9/ Date I i c . k
%ws e
t l (s/ E5-3 M dd et n PA Copyn0ht 1997 L.
( THREE MILE !SLAND Number gpgg TRAINING & EDUCATION nuct. san SIMULATOR MANAGEMENT PRS-TTS35 TITLE Revision No. TURBINE TRIP CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: Turbine Trip Certification Test Transient Test TTS35
, ANSI /ANS 3.5 Reference (s): 3.1.2(15)
Benchmark Test B2.2(6) Test Date: 12/23/97 Malfunction (s) To'sted*: TC01 Turbine Trip
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition 1C-13 50% Reactor Power Equilibrium Xenon End of Cycle rmin - This test was terminated after plant stability was reached at the lower reactor power level following the Main Turbine trip. Simulator Run Time: 10 minutes ' Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test TTS35 dated 11/26/96. I Right Direction Analysis. l Simulator Malfunction Cause and Effects Document Current Controlled Copies of: { J Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY l I l i l l [ ES-1 ( ' GPU Nuclear, Inc. Mddistown. PA Cooright 1997 l l l .
4 THREE MILE ISLAND Numbw TRAINING & EDUCATION NUCLEAs SIMULATOR MANAGEMENT PRS-TTS35 , Revisson No. 4 WTLE
- TURBINE TRIP CERTIFICATION TEST Results
Description:
In accordance with ANSl/ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perforrr, correctly following activation of a simulated malfunction. This transient resulted in a turbine trip with reactor power below the high power trip setpoint. The event was initiated by activation of the turbine trip malfunction after power was reduced to 41% in order to prevent a
. r: actor trip. The integrated Control System responded by initiating a power reduction. The control system transferred to the Tracking mode with megawatts generated equal to zero when the turbine was tripped.
Reactor power change due to increased temperature was less significant than the initial test due to incorporation of a new core model with different temperature coefficient values. Reactor Coolant System pressure and temperature decreased significantly as reactor power decreased to less than 22% SASS actuated a mismatch indication for T-cold transmitters. Reactor Coolant System pressure initially increased resulting in operation of the Pressurizer spray valve. Emergency Feedwater Actuation occurred due to low level in OTSG-A&B. Main Feedwater underfeed occurred on both OTSGs. The underfeed is a result of initially reduced levels and higher than normal OTSG pressures resulting in level dropping below initiation setpoint. Underfeed is expected to occur in the plant during low power transients. The transient was terminated when primary and secondary pressures and temperatures.were returning to stable conditions.
,m
( '7namic response was compared to the dynamic response of the initial certification test. During the conduct of this test ( ,dulator dynamic response, annunciator operation and automatic safety system actuations were recorded. GPUN
' personnel representing TMI Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations fro:n expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221.-ADM-2820.02. The results of these evaluations are as follows:
The results or this test are satisf actory, based upon de following: CRITERIA #1 ' The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. ! l CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the i alarm or automatic action, and conversely, the simulator shall not cause an a' arm or automatic action if the reference plant would not cause an alarm or automatic action. l The simulator satisfied this requirement with no new identified deviations. One exception, TTS35E014, remains uncorrected. l [\ l ( N ,I ES-2 GPU Nuclear. Inc. Mddletown, PA CopyngN 1997 l L_______ - - _ _ - -
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THREE MILE ISLAND Number TRAINING & EDUCATION suctu4n SIMULATOR MANAGEMENT PRS-TTS35 _ TITL.E. Revisen No. 4 TURBINE TRIP CERTIFICATION TEST Corrective Ac60ns/ Comments: i NONE Completed by: _ e Charles E. Husted _ Date' Approved by: M - operations Training Manager Robert Parnell __ 2 /I![
' Date h
( .. L i ES-3 U "d%:!M Copyright 1997
f THREE MILE ISLAND Number du)pg) TRAINING G EDUCATION suctaan SIMULATOR MANAGEMENT PRS-TTS42 TIT E Nsion No. 4 MAIN STEAM LEAK INSIDE THE RB CERTIFICATION THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identificatioru Main Steam Leak inside the RB Certification Test ! Transient Test TTS42 l ANSI /ANS 3.5 Reference (s): Benchmark Test B2.2(9) 3.1.2(20) Test Date: 01/08/98 Malfunction (s) Tested *: MS02A Main Steam Leak in the RB 100% (100% - 6,000,000 LBM/HR) ,
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Prctected initial Condition 10-16 f 100% Reactor Power Equilibrium Xenon Middle of Cycle eint of Test Termination: This test was terminated following the reactor trip, HPI, and Emergency Feedwater actuations due to the effects of a main steam line rupture inside containment. Simulator Run Time: 10 minutes Simulator Evaluation Time: 4 hours Baseline Evaluation Data: Previous Certification Tent TTS42 dated 11/22/96. Right Direction Analysis Simulator Malfunction Cause and Effecta Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY l l I e } i [ ES-1 GPU Nuclear, Inc. l h4ddletown, PA Copynght 1997 I , 1
f r I
., THREE MILE ISLAND Number TRAINING & EDUCATION t
muctaAs SIMULATOR MANAGEMENT PRS-TTS42 TITLE Revision No. 4 l MAIN STEAM LEAK INSIDE THE RB CERTIFICATION Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to ! perform correctly following activation of a simulated malfunction. l The event was initiated by causing a maximum size steam rupture inside containment on the OTSG 1 A. The resultant i steam release caused a reactor and turbine trip on Overpower witti subsequent Reactor Trip Containment isolation actuation. The containment pressure and temperatures increased causing Engineered Safeguards actuation, and a false i Reactor Building fire alarm. The rapid blowdown of OTSG 1 A caused a significant cooldown of the RCS. A SASS l
. actuation occurred due to the significant mismatch between the A and B Main steam header pressures. Feedwater flow was isolated to both OTSGs on HSPS Low Pressure. The Emergency Feedwater System actuated on high Reactor Building pressure and fed the OTSGs. High pressure injection flow recovered the RCS inventory lost by shrinkage and compressed the pressurizer steam bubble, elevating RCS pressure.
The transient was terminated following Engineered Safeguards and Emergency Feedwater system actuations with clevated Containment pressures and temperatures. 1 l Dynamic response was compared to the dynamic response of the initial certification test. During the conduct of this test I Cmulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results ) have been evaluated against ANSI /ANS 3.5 criteria by select GPUN personnel representing TMI Simulator Training. The l TMl representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from ' cxpected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity i Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: , I b The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. j l l 1 i I L I l l l ES-2 (
, GPU Nuclear. inc.
Miosetown, PA Copyngn 1997
]
(oM
, THREE MILE ISLAND Numtw TRAINING G EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS42
_TrrLE Revtalon No. MAIN STEAM LEAK INSIDE THE RB CERTIFICATION Corrective Actiona/ Comments: NONE Completed by: I/ M , MM# vvv- - - W. A. Fraser /- 2 9 -PT 9,,, 1 Approved by: . '[ddhN Robert Pameli 2/y/9,C 1 Operations Tr gnager / tate Ov 4 ES-3 ('
*22':2Yi Copynght 1997
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1 THREE MILE ISLAND Number
- g pgg TRAINING & EDUCATION muetsAs SIMULATOR MANAGEMENT PRS-TTS56 TITLE .
Revision No. A 4 MANUAL REACTOR TRIP CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Manual Reactor Trip Cet;ification Test Transient Test TTS56 ANSI /ANS 3.5 Reference (s): Benchmark Test 8.2.2(1) Test Date: 01/08/98
)!latfunction(s) Tested *: None - Initiation of Manual Reactor Trip Test initialization: Protected initial Condition 10-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the reactor tripped and proceeding to stable Hot Shutdown conditions.
10 minutes Simulator Evaluation T!me: 2 hours Baseline Evaluation Data: Previous Certification Test dated 11/25/96 Right Direction Analysis Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY i GPU Nuclear, Inc. l l
' Mddletown, PA Copyngnt 1997 l
T THREE MILE ISLAND Mmucumr TRAINING & EDUCATION SIMULATOR MANAGEf *ENT Number PRS-TTS56 TITLE Rewsson No. 4 MANUAL REACTOR TRIP CERTIFICATION TEST ! Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to i perform correctly following activation of a simulated malfunction. ' The event was initiated by depressing the Manual Reactor Trip pushbutton on the Main Console and reducing the pressurizer level control setpoint to 100 inches. All control rod drive breakers opened and Reactor Trip Containment Isolation actuated. The Main Turbine tripped on interlock. The Integrated Control System reduced Feedwater flows and Cabilized OTSG levels at Low Level Limits. The Turbine Bypass Valves operated to maintain OTSG pressure and remove decay heat. The transient was terminated with the reactor tripped and the plant approaching stable Hot Shutdown conditions. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were evaluated. Test results have been evaluated against AMSl/ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordanco with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisf actory, based upon the following: CRITERIA #1 m f) V The observable changes in simulator parameters shall correspond in direction to those expected from the actual reforence plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. l l I l l l l
/ E5-2
((j ) e,o m_. m Mddletown. PA Copynght 1997
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THREE MILE ISLAND Number gh TRAINING G EDUCATION suctsAA SIMULATOR MANAGEMENT PRS-TTS56 TITLE _ Reusion No. 4 MANUAL REACTOR TRIP CERTIFICATION TEST Corrective Actions / Comments:
.NONE
)
Completed by: #W-Mc #- William A. Fraser 2.-3-9A Date Approved by: ~ (fte.fS4) Robert Pamell M3//7"' Operations'Trgg%Iar,ager Date
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I l I I O E5-3 sto n. Copynght 1997
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# THREE MILE ISLAND Number G TRAINING S EDUCATION suCMAn SIMULATOR MANAGEMENT PRS-TTS57 TITLE Revtskm No. i
/N 3 l l I SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST l
\d i THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Simultaneous Closure of All MSIVs Certification Test Transient Test TTS57 '
ANSI /ANS 3,5 Reference (s): Benchmark Test B.2.2(3) Test Date: 01/08/98 Malfunction (s) Tested *: MS08A/B/C/D Main Steam' Isolation Valve Closure MS-V-1 A/1B/1C/1D I Test Initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middleof Cycle Point of Test Termination: This test was terminated with the reactor tripped and Emergency Feedwater removing decay heat via the Atmospheric Dump Valves. ulator Run Time: 20 minutes Simulator Evaluation Time: 4 hours Baseline Evaluation Data; Previous Certification Test dated 11/25/96 Right Direction Analysis Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant P & ID Procedures Overall Test Results: SATISFACTORY l
*O E5-1 d "%%%
Copynght 1997
THREE MILE ISLAND Number TRAINING G EDUCATION NUCLEAn SIMULATOR MANAGEMENT PRS-TTS57 TITLE ReWsion No. SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.51985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip on high pressure and Emergency Feedwater System actuation due to a loss of Main Feedwater. j The event was initiated by simultaneous closure of all four Main Steam Isolation Valves. This caused OTSG pressures f t) increase and Main Steam Header Pressure to decrease. The Main Turbine transferred to manual cor rol automatically i upon receipt of a sustained 40 PSIG Header Pressure Error Signal. As 01 SG pressures increased the Turbine Bypass, Atmospheric Dump and Main Steam Safety Valves operated to limit the pressure excursion. Reactor. Coolant System f j pressure and temperatures increased due to the transient. The reactor tripped on high pressure and actuated Reactor ^ Trip Containment isolation. When all Main Steam isolation Valves had closed the Main Feedwater Pumps slowed, reducing Feedwater flow to the OTSGs. The Emergency Feedwater System actuated to feed the OTSGs at Low Level Umits. Gland Sealing steam was lost when the OTSG 18 MSIV's closed, causlag a loss of vacuum to the Main Turbine and Main Feedwater Pumps. Decay heat was then removed by automctic action of the Atmospheric Dump Valves, since i the Turbine Bypass Valvee closed due to the low Condenser Vacuum condition (caused by the loss of Gland Seal J Steam). The transient was terminated with the reactor tripped and decay heat being removed by the Emergency Feedwater System via the Atmospheric Dump Valves. j Firing the conduct of t'his test simulator dynamic response, annunciator operation, and automatic safety system ( luations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel
%@ resenting TMl Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMl Un!! 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
The results or this test are satisfactory, based upon the following: CRITERIA #1 l 1 1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual j reference plant and do not violate the physical laws of nature, l l The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the alarm _ or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation.
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O) ( ES-2 GPU Nuclear. Inc. Mddletown, PA Copynght 1997
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THREE MILE ISLAND Number -
' TRAINING G EDUCATION NUCLEAs SIMULATOR MANAGEMENT PRS TTS57 TI M ReWsion No, n,
g SIMULTANEOUS CLOSURE OF ALL MSIVs CERTIFICATION TEST 3
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Corrective Actions / Comments:
Description:
Incorrect Computer Alarm Operation Number of Identified Deviations One Corrective Actions: One Discrepancy Report Correction Completed by: / ' h- ~ William A. Fraser' 2-S-9P Date Approved by: Robert Pamell ! Operations Training Manager Date 0 1 i j l I I i j l i [ ES-3 , . GPU Nuclear, Inc. (/ Mddletown, f A Copynght 1997
THREE MILE i3 LAND Number
,, g TRAINING & EDUCATION suctrasr SIMULATOR MANAGEMENT PRS-TTS58 TITLE Revtson No.
[ LOSS OF ONE RC PUMP CERTIFICATION TEST V 3 THREE MILE ISLAND l SIMULATOR MANAGEMENT l CERTIFICATION TEST ABSTRACT Test identification: Loss of One RC Pump Certification Test Transient Test TTS58 l ANSl/ANS 3.5 Reference (s): 3.1.2(4) Benchmark Test B.2.2(5) Test Date: 12/22/97 Malfunction (s) Tested *: RC35A RC Pump Trip Test initialization: Protected initial Condition 1C-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated following reactor trip with Primary and Secondary Plant pressures and temperatures approaching stability. mulator Run Time: 10 minutes Simulator Evaluation Time: 3 hours Baseline Evaluation Data: Previous Certification Test TTS58 dated 12/04/96 Right Direction Analysis Simulator Malfunction Cause and Effects Document. Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Emergency Procedures Overall Test Results: SATISFACTORY l O ==Copyngnt 1997 l L___-----
I
, THREE MILE ISLAND Number j
,'. g TRAINING & Er UCATION '
NUCLEAn SIMULATOR MANAGEMENT PRS-TTS58 . TITLE Reasion No. I LOSS OF ONE RC PUMP CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3,5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. ) ( This transient resulted in a reactor trip. Reactor Trip initiated a Main Turbine trip and Reactor Trip Containment Isolation. The event was initiated by failing Reactor Coolant Pump 1 A. The loss of the Roactor Coolant Pump at full power 1 resulted in the Reactor Protection System initiating Reactor Trip due to the reduction in Reactor Coolant System flow. The Reactor Trip resulted in a Main Turbine trip. Reactor Trip Containment isolation closed specified Containment Isolation Valves. Reactor Coolant System pressure and temperature decreased to post trip values and were maintained. SASS actuated a mismatch indication for T-cold and T-hot transmitters. The transient was terminated when Primary and Secondary pressure and temperature were approaching post trip stable co'nditions. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. GPUN personnel representing TMI Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI representatives have held NRC Senior Reactor Operator license for TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidality Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: o The results of this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reterence plantsand do not violate the physical laws of nature. The simulator k.sstied this JJquirement with no identified deviations. CRITERIA #2 The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. 4 3 1 I
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ES-2 )
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- GPU Nuclear. Ira.
I (j Mddletown. F Copright 1997 l l
THREE MILE ISLAND Number
g TRAINING & EDUCATION wuctrAs SIMULATOR MANAGEMENT PRS-TTS58 TIT E Rawsion No.
LOSS OF ONE RC PUMP CERTIFICATION TEST I Corrective Actions /Commentti 1 j NONE l l Completed by: v-w Charles E. Husted 3 /8 fS
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Date Approved by: ~ h Operations Trainh:g Manager Robert Pamell
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1 f GPU Nuclear,Inc.
\j Mddletown, PA Copynght 1997
, . . THREE MILE ISLAND Number
'GmuctaAN W TRAINING G EDUCATION !
, SIMULATOR MANAGEMENT PRS-TTS59 l TITLE
- ntMalon No. ;
3 MAXIMUM RATE POWER RAMP CERTIFICATION TEST i - THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Maximum rate Power Ramp Certification Test Transient Test l TTS59 l ANSI /ANS 3.5 Reference (s): Benchmark Test B.2.2(7) 5.4.1(2) Test Date: , 01/08/98 I l Malfunction (s) Tested *: None - Manual Plant Maneuvering
)
1 Test Initialization: Protected initial Condition 10-17 I l 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated with the plant stable at 100% power. 15 minutes Simulator Evaluation Time: 1 hour Baseline Evaluation Data: Previous Certification Test TTS59 dated 11/25/96 1 Right Direction Analysis Current Controlled Copies of: Reference Plant Operating Procedures Overall Test Results: SATISFACTORY l E5-1 l,o ==% , Copynght 1997 j
e i THREE MILE ISLAND Number TRAINING & EDUCATION suctsan SIMULATOR MANAGEMENT PRS-TTS59 Tn'W
- Revtsion No.
3 MAXIMUM RATE POWER RAMP CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated from steady-state full power with the integrated Control System in the integrated mode. The Unit Load Demand controller was set to approximately 75 percent load. The plant reduced power at the maximum rate of ten percent pe: minute and stabilized at 75 percent power. The plant was allowed to stabilize for approximately four minutes. 3 At the end of the stabilization period, power was ramped back up at the maximum rate of ten percent per minute. When i the plant reached 90 percent load the Integrated Control System automatically reduced the rate of power increase to 5 ! percent per minu'e. During the peried of time at reduced power the Xenon concentration began to increase. The plant was allowed to stabilize at full power. .
)
{ l The transient was terminated with the plant at full power. Dynamic response was compared to the dynamic response of the iriitial certification test. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMt representatives have held NRC Senior Reactor Operator licenses or certifications for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these eveluhtions are as follows: The results or this test are satisfactory, based upon the following: k- CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fall to cause an alarm or automatic action if the refererce plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. ( V E5-2 Cj GPU Nuclear, Inc. Mddletown, PA Copynght 1997
THREE MILE ISLAND Number TRAINING S EDUCATION spetsAn SIMULATOR MANAGEMENT PRS-TTS59 TITLE Rev.skm No. [~],- MAXIMUM RATE POWER RAMP CERTIFICATION TEST 3 , V Corrective Actions / Comments: j NONE Completed by: - M William A. Fraser D ~ S~-97 VV " Date Approved by: Robert Pamell / 7 /9# ' Operations Training Manager Date i l o I (
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I. E5 3 V GPU Nucieur. Inc. Mddletown, PA Copyngnr 1997
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t- THREE MILE ISLAND Number W muetsAs TRAlMNG & EDUCATION SIMULATOR MANAGEMENT PRS-TTS60 TITLE ' Rwison No. 5 LOSS OF OFFSITE POWER WITH LARGE BREAK LOCA CERTIFICATION 7EST ,_ THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TFST ABSTRt.CT Tgg Identification: Loss of Offsite Power with Large Break LOCA Certification Test Transient Tect TTS60 I ANSI /ANS 3.5 Reference (s): 3.1.2d D) Benchmark Test B.2.2(8) Test Date: 12/22/97 Malfunction (si Tested *: ED01 Station Blackout TH04A RCS LOCA a'. Hot Leg Nozzle 100% severity Testinitialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middleof Cycle in iT T rmin i n- This test was ter.ninated with the reactor trippod, Core Flood tanks emptied and reactor core cool!ng being provided by High Pressure an(i Low Pressure Injection. Reactor Building pressure and temperature were decreasing from peak conditions. Simulator Run Ting 10 minutes Simulatsf Evaluation T!me: 8 hours paseline Evaluation Data: Previous Certificatbn Test TTS60 dated 11/25/96 Right Direction Analysis. Simulator Malfunction Causs and Effects Diagram Current Controlled Copies of: Reference Plant Alarm Response Procedures Heference F; ant Electrical One-Line Diagrams Reference Plant Emergency Procedures Overall Test Resdits: SATISFACTORY l t ["N) ES-1 GPU Nuclear. inc. I ( j/ Mddletown. PA Copynght 1997 l
i 1 THREE MILE ISLAND Number i 1 TRAINING & EDUCATION mucer S!fgtVLATOR MANAGEMENT PRS-TTS60 Trd.E Revison No. fm 5 t l
. {OMpSS OF OFFSITF POWER WITH LARGE BREAK LOCA CERTIFICATION TEST
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Results
Description:
In accordance with ANSl/ANS-3.5-1905 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation o' a simulat El malfunction. i This transient resulted in the loss of offsite power coincident with a large break loss of coolant accident. The reactor was ' tripped due to a loss of power to the Reactor Coolant Pumps and the Control Rod Drive System. Safety system power was restored by the Emergency Diesel Generators. A large break in the Reactor Coolant System caused rapid depressurization of the system and a rapid increase in Reactor Building pressure. The Engineered Safeguards System actuated, initiating safety injection and Reactor Building isolation and Cooling. The event was initiated by a fault in the 230 KV Substation coincident with a large size rupture of the Reactor Coolant System Hot Leg piping. The Reactor Protection System initiated a Reactor trip and Main Turbine trip. Reactor Trip Containment isolation closed specified Containment isolation valves. The Heat Sink Protection System detected a loss of Reactor Coolant Pumps and initiated Emergency Feedwater to the O_TSGs, controlling level at 50% on the operating
' ange. The Emergency Diesel Generators started and the output breakers closed to re-energize their 4.16 KV Buses. 4 SASS actuated a mismatch indication for Turbine Header Pressure Transmitters and RCS Pressure Transmitters. !
I The steam supply to Emergency Feedwater Pump EF-P-1 turbine is the Main Steam system. As Main Steam system pressure decreased EF-P-1 steam supply pressure decreased, as indicated on console instrumentation. When EF-P-1 pressure reached 48 PSIA it increased to about 110 PSIA then continued a decreasing trend. This response is incorrect. I e loss of inventory and oressure from the Reactor Coolant System resulted in discharge of the Core Flood Tanks and uation of the Engineered Safeguards Actuation System. High Pressure injection, Low Pressure injection, Reactor ilding Spray, and Reactor Buildir#g isolation and Cooling were initiated due to the pressure loss from the Reactor Coolant System and the high pressure detected in the Reactor Building. l The transient was terminated when Reactor Building pressure and temperature were decreasing and reactor core cooling tcs verified. . Dynamic response was coKcared to the dynamic response of the previous certification tests. During the conduct of this test simulator dynamic respoinse, annunciator operation and automatic safety system actuations were recorded. GPUN personnel representing TMI Simulator Training have. evaluated test results against ANSI /ANS 3.5 criteria. The TMl representatives. hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity i Procedure,6211 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in sirnulator parameters shall correspond in direction to those expected from the actual l , reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with two identified deviations: Containment Temperature and EFP Steam Supply pressure response. Containment temperature was identified during the last test. j ES-2 i opu wew. ine ( Mddletown. PA Copright 1997 ( - i ( t
THREE MILE ISLAND Number
'( sucurAn TRAINING & EDUCATION SIMULATOR MANAGEMENT PRS-TTS60 TITLE Raision No.
5 LOSS OF OFFSITE POWER WITH LARGE BREAK LOCA CERTIFICATION TEST CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have ca'used the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulatof sati$fied this requirement with no identified deviations. Corrective Actions! Comments:
Description:
. Incorrect System Response Number of identified Deviations: Two Corrective Actions: One Discrepancy Report correction to resolve the plottinfresponse problem -]
with CH'T655D has not been corrected. A new discrepancy report to ' correct EFP-1 steam supply pressure response. l
)
( omp'l'e ted by:
/ Charles E. Husted 3 /8 f8 D' ate
( Approved by: ] # 4- - Robert Parnell 3 l ! h,f . Operations Training Manager Date I l
/ ES-3 (N GPU Nacisar, Inc. 1 Mddietown. PA Copyr.gnt 1997 ;
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- THREE MILE ISLAND TRAINING & EDUCATION Number nuct. san SIMULATOR MANAGEMENT PRS-SSP 01 TITLE Rwtson No.
6 SIMULATOR STABILITY CERTIFICATION TEST l THREE MILE ISLAND l SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT l l 1 Test identification: Simulator Stability Certification Test Steady State Performance Test
)
l SSP 01 ANSI /ANS 3.5 Reference (s): 5.4.i(2) { 4.1 J Benchmark B2.1 l Test Date: 03/20/98 - Test Initialization: Protected initial Condition 10-17 - 100% Reactor Power Equilibrium Xenon ( End of Cycle i Point of Test Termination: 100% Reactor Power g 579 Degrees Tave [ ICS in Full Automatic I Sim. ulator Run Time: 1 Hour Baseline Evaluation Data: Plus/minus 2% Overall Test Results: SATISFACTORY l l Results
Description:
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In accordance with ANS!!ANS-3.5-1984 this certification test was conducted to demonstrate the ability of the simulator to operate at stable full power, in automatic control for a 60-minute period. During conduct of this test simulator dynamic operation was evaluated against ANS!/ANS 3.5 criteria. GPUN personnel representing TMl Simulator Training conducted this evaluation. The evaluators have held a NRC Senior Reactor Operator license for TMI Unit 1. The results of this evaluation are as follows: I { t i E3-1 GPU Nuclear. Inc P.eddletown, PA Capright 1997
f d- > THREE MILE ISLAND Number TRAINING & EDUCATION mucurAs SIMULATOR MANAGEMENT PRS-SSP 01 TITLE Revison No.
/*% 6
( ) SIMULATOR STABILITY CERTIFICATION TEST i v j 4
. The results of this test are satisfactory, based upon the following:
s CRITERIA #1: The simulator computed values for steady state, full power operation with the referenced plant control system configuration shall be stable and not vary more than plus/minus 2% of the initial values over a 60-minute period. The simulator supported this requirement with no deviations. Completed by:
~
d, Charles E. Husted 7 24 78
/ Date Approved by:
Operations Training Manager Robert L. Parnell >/ /k[ Date [\ U (s) ( E3-2 GPU Nue: ear. Inc. Mddistown, PA Copyrignt 1997
i 1 l g> THREE MILE ISLAND Number * { TRAINING & EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-SSP 02 ! TITLE Reson No. l (N SIMULATOR ACCURACY CERTIFICATION TEST 6 \ \ i {
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l THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Simulator Accuracy Certification Test Steady State Performance Test SSP 02 ANSl/ANS 3.5 Reference (s): 4.1(2) l 4.2(3) j 4.1(4) 1 5.4.1(2) i Benchmark Test B2.1 Test Date: 12/23/97 Test initialization Protected initial Condition 10-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: After measurement of simulator-computed values and principal mass / energy balance determinations were made at 100%,80%, and 60% rated power. 9 simulator Run Time: 3 hours i Baseline Evaluation Data: Previous Certification Tests SSP 02 dated 12/09/96. Current Controlled Copies of: Reference Plant Operating Data Reference Plant Operating Procedures Overall Test Results: SATISFACTORY l
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J i ES-1 GPU Nuclear,Inc. MMelown. PA Copynght 1907 i
e THREE MILE ISLAND Number j G muctaM TRAINING & EDUCATION - i SIMULATOR MANAGEMENT PRS-SSP 02 ) TITLE Rension No. SIMULATOR ACCURACY CERTIFICATION TEST O Results
Description:
6 In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to perform within the I accuracy requirement of plus/minus 2% for critical parameters and plus/minus 10% for noncritical parameters pertinent to plant operation. Principal mass and energy balances shall be satisfied. During the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. GPUN personnel representing TMl Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM 2820.02. The results of these evaluations are as follows: The results or this test are satisf actory, based upon the following: ' CRITERIA #1
)
i The simulator-computed values of critical parameters shall agree within plus/minus 2% of the reference plant parameters and shall not detract from training. The parameters displayed on control panels may have the ] I instrument error added to the computed values. j The simulator satisfied this requirement with no identified deviations. 4
] CRITERIA #2
{ The calculated values of noncritical plant parameters pertinent to plant operation, that are included on the simulator control panels, shall agree within plus/minus 10% of the reference plant parameters and shall not detract from training. The parameters displayed on control panels may have the instrument error added to the computed values. The simulator satisfied this requirement with no identified deviations.
- l Corrective Actions / Comments:
NONE l l l Completed by: ' 4 '
/ Charles E. Husted j 24 98 I
/ bate Approved by: a Operations Training Manager Robert Parnell h7 bate [N
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l ES-2 GPU Nctear, Inc. i Mddletown, PA i f' Copright 1997 I
1 I g THREE MILE ISLAND. Number
$h mucurAs TRAINING & EDUCATl!N PRS-RTT01 SIMULATOR MANAGEMENT TITLE - Renskm No.
REAL TIME TEST CERTIFICATION TEST 3 THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT
' Test identification: Real Time Certification Test RTT01 ANSI /ANS 3.5 Reference (s): Appendix A3(1)
Test Date: 03/29/98 Test Initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle Point of Test Termination: This test was terminated following collection of all required data. Simulator Run Time: 30 minutes j Simulator Evaluation Time: 1.5 hours In vI in - Valve stroke times per controlled copies of applicable Reference Plant surveillance procedures. Known values for event timers. Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator computers to run in real time. During the conduct of this test simulator dynamic performance was evaluated against ANSI /ANS 3.5-1985 criteria. These evaluations were conducted b; GPUN personnel representing TMI Simulator Training who hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1. This test verified the capability of the simulator computers to run in real time by measuring known values for selected valve stroke times and event timers. The correct valve stroke times were drawn from controlled copies of Reference Plant surveillance procedures. Surveillance procedure valve stroke times were used as reference points to validate simulator computer performance. Known values for event timers were drawn from Reference Plant system design data. l A calibrated stopwatch was utilized to time each item. Additionally, the computer system processing capabilities were measured to serve as an annual check on each processor's performance. The evaluation of this data serves as a computer system management. GPU Nuclear. Inc. Md@etown, PA Copright 1997
, .. THREE MILE ISLAND Number TRAINING & EDUCATION muctsan SIMULATOR MANAGEMENT PRS-RTT01 TITLE Rwision No.
REAL TIME TEST CERTIFICATION TEST 3
- I The results of this test are satisfactory with all valve stroke times and event timers corresponding to the reference data from the Reference Plant with minor differences due to the use of a manual stopwatch for timing. Computer processing results were also satisfactory.
Corrective Actions / Comments: There are no corrective actions required as a result of this test. Completed by: [ #/ ' C
/
William A. Fraser M-P/ Date Approve'd by: C Robert Pamell /[Y[ Operations Training Manager Da't6 / V[ > I ( l' I (' )
,U E3-2 GPU Nuctsar. Inc.
Mddletown, PA Copright 1997
, THREE MILE ISLAND Number M>gg TRAINING & EDUCATION ,
NUCLEAs SIMULATOR MANAGEMENT PRS TTS44 TITLE Reviskm No. j 2 1 MAIN FEEDWATER LINE BREAKJNSIDE THE RB CERT!FICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification- Main Feedwater Line Break Inside Certification Test Transient Test TTS44 ANSI /ANS 3.5 Reference (s): 3.1.2(20) Test Date: 12/23/97 Malfunction (s) Tested *: FWO9B Feedwater Line Break Just inside RB 10% Severity
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle This test was terminated following Engineered Gafeguards actuation due to #4 Reactor Building pressure. The reactor was operating tvith power being reduced due to boron injection into the RCS. Simulator Run Time: 7.0 minutes Rimulator Evaluation Time: 5 hours Baseline Evaluation Data: Previous Certification Test TTS44 dated 01/19/94 Right Direction Analysis RETRAN Data Simulator Malfunction Cause and Effects Dcm' ient Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY l I i GPU Nuclear, Inc. MKidistown. PA Copyright 1997
, THREE MILE ISLAND Number TRAININ3 S EDUCATION
'muetsAs SIMULATOR MANAGEMENT PRS-TTS44 TjTLE Revtsson No.
2 MAIN FEEDWATER LINE BREAK INSIDE THE RB CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to pe! form correctly following activation of a simulated malfunction. This transient resulted in an increase Reactor Building pressure actuating the Engineered Safeguards System. The integrated Control System responded to the reduction in Main Feedwater to OTSG 18 and boron injection. The cvent was initiated by rupturing Loop B Main Feedwater line just inside the Reactor Building, equivalent to 10% o'f the pipe size. The reduction in Feedwater to the OTSG 1B caused a difference in T-cold and the Integrated Control Syst;m attempted to correct by increasing Loop incre: sed pressure above the setpoint ered for Engine,B Safeguards Main Feedwater causing flow. actuation. The The resulted actuation leakagein into the injection Reactor of borated water into the Reactor Coolant System reducing power, and increasing pressurizer level and Reactor Coolant System pressure. The transient was terminated shortly after the Engineered Safeguards Actuation System actuation occurred. The Re:ctor Coolant System was responding to the rapid reduction in power and Reactor Building pressure was slowly increasing, due to the leak. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety systern cctu:tions were reccided. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Operations Training. The TMl representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in , 7"~7ance with Functional Fidelity Procedure,6511-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. O I \ V '
. ES-2 GPU Nuclear,Inc.
Middletown, PA Copyright 1997
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, THREE MILE ISLAND Number TRAININ3 & EDUCATION NUCLEAN SIMUL.ATOR MANAGEMENT PRS-TTS44 TITLE- Revision No.
f%. 2
) MAIN FEEDWATER LINE BREAK INSIDE THE RB CERTIFICATION TEST Corrective Actions / Comments:
NONE Completed by: Charles E. Husted 3//8/98 Date Approved by: - Operations Training Manager Robert Parnell 3// ' hV Date - O 4
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t O ES-3 GPU Nuclear,Inc. Mddletown, PA Copyright 1997
THREE MILE ISLAND Number g% TRAINING & EDUCATION aructEAs SIMULATOR MANAGEMENT PRS-TTS45 l TITLE RevisKin No.
] ,
2 f MAIN FEEDWATER LINE BREAK OUTSIDE THE RB CERTIFICATION TEST J l THREE MILE ISLAND i SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT-Test identification: Main Feedwater Line Break Outside The RB Transient Test l TTS45 l l- ANSI /ANS 3.5 Reference (s): 3.1.2(20) Test Date: 01/09/98 l Malfunction (s) Tested *: FW10A Feedwater Line Break Outside RB 100% severity
,
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC-16 100% Reactor Power g Equilibrium Xenon Middleof Cycle Point of Test Termination: This test was terminated with the reactor and turbine tripped. Reactor Coolant System pressure and pressurizer level were hereasing. OTSG level and pressure were l approaching normal post trip values.
- Simulator Run Time: 10 minutes Simulator Evaluation Time: 3 hours Baseline Evaluation Data: Initial Certification Test dated 01/27/94 Right Direction Analysis 1
Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Overall Test Results: SATISFACTORY l [' GPU Nuclear, Inc. l Mddletown, PA Copynght 1997 4
Y THREE MILE ISLAND Number
- TRAINING & EDUCATION i wuct.sAs SIMULATOR MANAGEMENT PRS-TTS45
{ yLE Revision No.
/ 2 l lq) MAIN FEEDWATER LINE BREAK OUTSIDE THE RB CERTIFICATION TEST
\
j Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip due to high Reactor Coolant System pressure. The reactor trip initiated a Main Turbine trip and Reactor Trip Containment isolation. The Heat Sink Protection System actuated Emergency Feedwater
' 12 control the OTSG levels.
l The event was initiated by rupturing the Main Feedwater line to OTSG 1 A at a location outside the Reactor Building. The reduction in flow to OTSG 1 A resulted in an increase in Reactor Coolant System temperature resulting in pressure ! increasing to the high-pressure trip setpoint. The Reactor Protection System initiated a reactor / turbine trip and Reactor l Trip Isolation of the containment. The increase in Reactor Coolant System pressure resulted in ' opening of the !' pressurizer spray valve and the PORV. SASS actuated a mismatch on out of core nuclear instruments. The Heat Sink Protection Syster* actuated when OTSG level reached the low level setpoint for each OTSG, starting the Emergency Feedwater Pump; and controlling OTSG level at the low level setpoint. The 1:st was terminated with the reactor and turbine tripped. Reactor Coolant System pressure and pressurizer level were increasing towards normal post trip values. OTSG level and pressure were also approaching normal post trip values. ' l p l ( ) the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system L: ions were recorded. GPUN personnel representing TMl Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI rersresentatives have held NRC Senior Reactor Operefor license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: c'r'g5ggy The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 l The simulator shall not f all to cause an alarm or automatic action if the reference plant would have caused the alarm or automa:lc action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not caua an alarm or automatic action. The simulator satisfied this requirement with no iden3ad deviations. l
)
ES-2 I [\ p i GPU Nociser. Inc. Mddletown. PA
, Coppight 1997 L
THREE MILE ISLAND Number
- TRAINING & EDUCATION wuctsAs SIMULATOR MANAGEMENT PRS-TTS45 TITLE Ramslon No.
2 MAIN FEEDWATER LINE BREAK OUTSIDE THE RB CERTIFICATION TEST t Corrective Actions / Comments: NONE
\
Completed by: Charles E. Husted
/
/[Date f8 l Approved by: e Operations Training Manager Robert Parnell
_)'//9Dhte98 O I s I i l ! l ES-3 GPU Nuclear,Inc. [ Mddletown, PA Coppight 1997
,3 THREE MILE ISLAND Number G NUCLKAa
'fg TRAININ2 G EDUCATION SIMULATOR MANAGEMENT PRS-TTS46 TITLE
- Revision No.
NI-5 FAILURE CERTIFICATION TEST 2
~
THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: NI-5 Failure Certification Test Transient Test TTS46 ANSI /ANS 3.6 Reference (s): 3.1.2(21) , Test Date: 01/08/98 Malfunction (s) Tested *- N116A Power Range Upper Detector Linear Amp Signal Falls High - NI-5
' Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated with the Channel 1 A of the Reactor Protection System tripped and after verification of failure responses. Aator Run Time: 6 minutes Simulator Evaluation Time: 30 minutes Baseline Evaluation Data: Previous Certification Test dated 01/22/94 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY ES-1 (, j GPU Nuclear,Inc, IAddiotown, PA Copyngtd 1997
,4 THREE MILE ISLAND Number j
CJeff ~ TRAINING & EDUCATION wuctsAn SIMULATOR MANAGEMENT PRS-TTS46 i TITLE
^ Rension No.
NI-5 FAILURE CERTIFICATION TEST 2 RFsults
Description:
In accordance with ANSl/ANS-3.5-1985 this certification test was conducted to denionstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in the trip Channel 1 A of the Reactor Protection System. Overall plant stability was maintained. The event was initiated by failing the power range upper detector linear amplifier signal high for Ni 5. This caused two modules in Reactor Protection System Cabinet 1 A to trip (Overpower and Power / Flow /lmbalance). A SASS actuation occurred which caused the ICS controlling signal to switch to Ni-6. The initial f alse overpower signal initiated a Cross Limit which initiated insertion the Group Seven Control Rods. The Cross Limit immediately cleared. Plant stability was ! maintained at 100% power. l The transient was terminated following an inspection of the RPS Cabinet 1 A and verification of plant stability. During the conduct of this test simulator dynamic rr sponse, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Simulator Training. The TMi representatives have held NRC Senior Reactor Operator licenses or certifications for TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and ; corrective action in accordance with Functional Fidelity Procedure,6511-ADM-2820.02. The results of these evaluations aro as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 []) The obs3rvable changes
\
V r in simulator paramete s sh ll a correspond in direction to those 'xpected from the actual reference plant and do not vlotate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 l l The simulator shall not fall to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the l' reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE Completed by: ddaPrM e William A. Fraser .3 +9# l / Date Approved by: # Operations Training Manager Robert Parnell / DateNW
>O 5 l ES-2 GPU Nuclear,inc,
(/ M$dletown. PA Copfight 1997 ) i
y THREE MILE ISLAND Number TRAININ3 & EDUCATION nuctsAs SIMULATOR MANAGEMENT PRS-TTS47 TITLE Revision No. 2 NI-8 FAILURE CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: NI-6 Failure Certification Test Transient Test TTS47 ANSl/ANS 3.5 Referencels): 3.1.2(21) Test Date: 01/12/98 ICalfunction(s) Tested *: N!20B Total Power Range Summer Amplifier Signal Faiis High - NI-6
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected Initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle "t of Test Termination: This test was terminated with the Channel 1B of the Reactor Protection System tripped
. due to a false high power signal and after verification of Nl failure responses.
Simulator Run Time: - 10 minutes Simulator Evaluation Time: 30 minutes Baseline Evaluation Data: Previous Certification Test dated 07/12/93 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall TeJt Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perfirm correctly following activation of a simulated malfunction. This tr nsient resulted in the trip Channel 1B of the Reactor Protection System. Overall plant stability was maintained. The cvent was initiated by failing the power summer amplifier signal high for NI-6. This caused several modules in the R: actor Protection System Cabinet 1B to trip. These include Overpower, Power / Flow / Imbalance and Power / Pumps. A SASS actuation occurred as a result of the malfunction. However, since NI-6 was the non-selected instrument, no central system instrument transfer occurred. TM transient was terminated following evaluation of the RPS Channel 1B and verification of maintained plant stability.
\
l [V
p .w ' THREE MILE ISLAND Number
\ djDU TRAININ3 S EDUCATION NUCLEAR SIMULATOR MANAGEMENT PRS-TTS47 TIT,LE Revison No. / \ 2 g f NI-8 FAILURE CERTIFICATION TEST v
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel r: presenting TMI Simulator Training. The TMl representatives hold or have held NRC Senior Reactor Operator licenses for TMl Unit 1, Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6511 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator narameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. ' The simulator satisfied this requirement with no identified deviations. l GRITERIA #2 l The simulator shall not fail to cause an alarm or automatic action if the reference plant owuld have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the j reference plant would not cause an alarm or automatic action. 1 i The simulator satisfied this requirement with no identified deviations. - j
- V Corrective Actions / Comments
NONE Completed by: [I///#M P William A. Fraser b*W Date Approved by: Robert Pamell /7 N Operations Training Manager Date i l i
THREE MILE ISLAND Number p (GP(f
~
TRAININS G EDUCATION NUCLEAN SIMULATOR MANAGEMENT PRS-TTS48 i TITLE Revision No.
'p) .
PRESSURIZER LEVEL TRANSMITTER FAILURE CERTIFICATION TEST 2 J THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification.DI Pressurizer Level Transmitter Failure Certification Test Transient Test TTS48 ANSI /ANS 3 5 Reference (s): 3.1.2(22) Test Date: 01/12/98 Malfunction (s) Tested *- RC04A RC PZR Level Transmitter Failure 4 RC-1-LT1 ! 0% (100% = 400 Inches) Ramp 240 Seconds
- Refer to Malfunction Cause and Effects Documents for options available.
Iest initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon , Middie of Cycle P5 int of Test Termination: This test was terminated with real pressurizer level and RCS pressure slowly decreasing while indicated level was increased. Simulator Run Time: 10 minutes Simulator Evaluation Time: 30 minutes Baseline Evaluation Data: Previous Certification Test dated 01/22/94 - Right Direction Analysis i Simulator Malfunction Cause and Effec 3 Document Current Controlled Copies of: Reference Plant Alarm Resp ,nse Procedures Reference Plant Emergency Procedures Ovgrall Test Results: SATISFACTORY Results
Description:
l In accordance with ANSl/ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. Th3 cvent was initiated by failing the selected pressurizer le' vel differential pressure transmitter to zero inches DP over a 240 second ramp. This caused a false high level s.gnal for the pressurizer level control circuit t.nd the console recorder. Since the operator-selected level transmitter also controls the normal Makeup Valve MU-V-17, makeup flow dropped to zIro. Letdown flow of approximately 45 gallons per minute was maintained. This resulted in a reduction of primary inventory. The non-selected pressurizerlevel transinitters indicated the actuallevel decrease. Although [( )Jrizer heater demand increased in response to the pressure drop caused by the inven L_:ued to slowly decrease due to the pressurizer out-surge. l ,
.. THREE M:LE ISLAND Number !
g)E(J TRAINING u EDUCATION l wuct m SIMULATOR MANAGEMENT PRS-TTS48 TE LE Reason No.
' l 2 j PRESSURIZER LEVEL TRANSMITTER FAILURE CERTIFICATION TEST i
The transient was terminated with actual pressurizer level and RCS pressure decreasing. ' During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system
. actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel j
) repr:senting TMI Simulator Training. The TMl representatives hold or have held NRC Senior Reactor Operator licenses for TMl Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in tccordance with Functional Fidelity Procedure,6511-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisf actory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the 5.Q f
/reference plant would not cause an alarm or automatic action.
U The simulator satisfied this requirement with no identified deviations, cllve Actions / Comments: i NONE i C:mpleted by: _[ ///m_/ A William A. Fraser )d7-F Date Approved by: 7MM Cperations Training Manager
/ Robert Pamell
/ Y Dale l
i j (m) .
dd
\GN THREE MILE ISLAND TRAINING & EDUCATION Number l
l suctr4a SIMULATOR MANAGEMENT PRS-TTS49 l TITLE Revision No. 4 NORMAL FEEDWATER SYSTEM FAILURE CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Normal Feedwater System Certification Test Transient Test TTS49 ANSl/ANS 3.5 Reference (s): 3.1.2(22) Test Date: 01/19/98 Malfunction (s) Tested *: FWO2A - Loop A Main Feedwater Flow Transmitter Failure SP-8A-DPT1 0% (0-100% = 0-6,000,000 LBM/HR) FWO2B - Loop A Main Feedwater Flow Transmitter Failure SP-8A-DPT2 0% (0-100% 6,000,000 LBM/HR)
- Refer to Malfunction Cause and Effects Documents for options available.
Test initialization: Protected initial Condition IC 15 100% Reactor Power (q V i Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated with OTSG .1 A and 1B on Low lev 91 Limits, the reactor and main turbine tripped. Simulator Run Time: 30 minutes Simulator Evaluation Time: 3 hours Baseline Evaluation Data: Previous Certification Test dated 11/09/93 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Procedures Reference Plant P & ID Prints Overall Test Results: SATISFACTORY
/~~'T I k'N ES-1 GPU Nuclear,Inc, Idddletown, PA Copynght 1997 N
' I d THREE MILE ISLAND Number Fpgp TRAINING & EDUCATION
" nucuas SIMULATOR MANAGEMENT PRS-TTS49 i
l TITLE Rewston No. 4 NORMAL FEEDWATER SYSTEM FAILURE CERTIFICATION TEST Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. The event was initiated by failing both Loop A Main Feedwater Flow Transmitters to zero. The Integrated Control System (ICS) sensed the significant Feedwater flow error and initiated Cross-Limits to reduce plant power to match the indicated flow reduction. The resultant Loop A Feedwater Flow error caused the Loop A Feedwater Regulating valve to open wide. Tha increase in Loop A Feedwater flow induced a differential in cold leg temperatures, and the ICS Delta Tc circuit incr:ased Loop B Feedwater flow. The increased feedwater flow to both OTSG's caused a reduction in RCS pressure. The cross limits reduced Loop B-feedwater flow in an attempt to control total feedwater flow. RCS pressure continued to decrease until the reactor tripped on low pressure. L ow level limits then overrode all other feed flow control signals, closing FW-V-16A and allowing FW-V 163 to control level at setpoint. OTSG "A" level boiled down until it reached the low level control setpoint. FW-V-16A then opened to control OTSG level. RCS pressure and pressurizer level recovered until near the end of the transient when MU P-1B experienced a cavitation-induced failure due to a loss of makeup tank level. This test differed significantly from the previous two tests. In this case, the reactor tripped instead of remaining at power. Investigation revealed that two calibration constants were changed in the simulator in response to a plant change. ICKBTUG4 and ICKBTUBIAS were changed. This effectively prevented BTU limits from limiting the feedwater flow increases, and caused the overcooling condition to trip the reactor, Rerunning the test with the original constant values yWided results similar to the 1994 test. The test results are being investigated by Training and the system l&C engineer in ('~] to determine if this is the correct response. A Discrepancy Report will be generated if it is determined that the
'! cse is not in accordance with plant response.
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training The TMl representatives have held NRC Senior Reactor Operator licerises for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6511 ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation. '
'l \bj E5-2 GPU Nuclear,Inc.
Mddletown. PA Copyright 1997
d THREE MILE ISLAND Number TRAINING & EDUCATION ' suctrAs SIMULATOR MANX 3EMENT PRS-TTS49 T]LE Revision No. 4 NORMAL FEEDWATER SYSTEM FAILURE CERTIFICATION TEST Corrective Actions / Comments:
Description:
Incorrect Replica Plant Process Computer Alarms L2699, L2702, L2704 Number of Identified Deviations: One Corrective Actions: One Discrepancy Report Correction Completed by: / // /8 + # v -- - - sh William A. Fraser 3-/7-9eF ng,, Approved by: a Robert Parnell S!) / Operations Training Manager Date O l
)
i l 1 ! h GPU Nuclear,Inc. Mddletown, PA copyright 1997
i I' THREE MILE ISLAND Number Agg TRAINING & EDUCATION nuctEAN SIMULATOR MANAGEMENT PRS TTS50
~
TITLE Revtsson No. (G
\__/ C COLD LEG TEMPERATURE TRANSMITTER FAILURE CERTIFICATION TEST 4
l THREE MILE.lSLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: RC Cold Leg Temperature Transmitter Certification Test Transient Test TTS50 Atgj!ANS 3.5 Reference (s): 3.1.2(22)
)
Iggi,pgra 01/09/98 Malfunction (s) Tested *: RC08B RC Cold Leg Narrow Range Temperature Transmitter Wilure RC-5A-TE3- 0% RC08A RC Cold Leg Narrow Range Temperature Transmitter e
- jure RC-5A-TE1 - 0% '
Ramp 300 seconds
- Refer to Malfunction Cause and Effects Documents for options available.
Test Initialization: Protected initial Condition IC 17 100% Reactor Power Equilibrium Xenon End of Cycle e of Test Tennination: This test was terminated with the reactor operating at 101% power with stable Feedwater flow and OTSG levels. Reactor Coolant System pressure and pressurizer level were stable. Simulator Run Time: ' 12 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Initial Certification Test TTS50 dated 02/01/94 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY l l
\ \j E5-1 GPU Nuclear,Inc.
Mddletown. PA Copptght 1997 I t
q l THREE MILE ISLAND Number Mpgg TRAINING & EDUCATION I aructsan SIMULATOR MANAGEMENT PRS-TTS50 j 1 TITLE
- Redsion No. j 4
; COLD LEG TEMPERATURE TRANSMITTER FAILURE CERTIFICATION TEST )
Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. This transient resulted in a shift of Reactor Coolant System heat removal such that more heat was being removed'via OTSG 1B. Operating range level was higher in OTSG 1B and Feedwater flow to OTSG 1B was higher. Pressure was slightly higher in OTSG 1B. The cvent was initiated by f alling the unselected T-cold instrument (RC-5A-TE3) to 520 degrees causing SASS to automatically select ti.a fully functional instrument (RC-SA-TE1). A malfunction was then entered failing the selected instrument to 520 degrees over a 300-second ramp period. The Integrated Control System re-ratioed Feedwater flow att:mpting to correct the mismatch in cold leg temperature between the Loop A and Loop B of the Reactor Coolant System. As a result the level increased in OTSG 1B and decreased in OTSG 1 A. There was a small, short-term oscillation in Reactor Coolant System pressure and pressurizer level. Reactor power increased to 101%. . The transient was terminated when Feedwater flow and OTSG level had stabilized. During the conduct of ! is test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorucd. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMI Operations Training. The TMI representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure, 6221-ADM-2820.02. The results of these evaluations are as follows: ( i
. v) The results of this test are satisfactory, based upon the following:
CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified devi9 ions. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic c.ction, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. I Corrective Actions / Comments: ' NONE
)
>O k)v E5-2 GPU Nuclear, Inc.
Mddletown, PA
, copright 1997
_ __. ____b
THREE MILE ISLAND Number TRAINING & EDUCATION nuctrAs SIMULATOR MANAGEMENT PRS-TTS50 TfLE Revision No. 1 COLD LEG TEMPERATURE TRANSMITTER FAILURE CERTIFICATION TEST
/)
Completed by: !
,7 Charles E. Husted 7 /7 !98 i t$ age Approved by:
Operations Training Manager
~
Robert Parnell 3[JI/#// Date c) I I a r \ ES-3 A to n. N Copynght 1997 1.
THREE MILE ISLAND Number
" (G(
nuctsAR TRAINING & EDUCATION PRS-TTS51 SIMULATOR MANAGEMENT TITM Renson No. t OTSG PRESSURE TRANSMITTER FAILURE CERTIFICATION TEST v THREE MILE ISLAND l SIMULATOR MANAGEMENT l CERTIFICATION TEST ABSTRACT Test Identification: OTSG Pressure Transmitter Certification Test Transient Test TTS51 ANSl/ANS 3.5 Reference (s): 3.1.2(22) Test Date: 01/19/98 Malfunction (s) Tested *: MS15A SGA Pressure Transmitter Failure SP6A-PT1 100% (100% - 1200 PSIG) MS15B SGA Pressure Transmitter Failure SP6A-PT2 100% (100% = 1200 PSIG)
- Refer to Malfunction Cause and Effects Documents for options available.
4 Test initialization: Protected initial Condition IC-16 100% Reactor Power Equilibrium Xenon Middle of Cycle ,)
' Point of Test Termination: This test was terminated with Main Feedwater maintaining OTSG levels, the Main Generator at low load and reactor power at 28%. OTSG 1 A Turbine Bypass and
, Atmospheric Dump valves were full open in response to the Control System input failure. Simulator Run Time: 10 minutes Simulator Evaluation Time: 7 hours Baseline Evaluation Data: Previous Certification Test dated 03/11/94 Right Direction Analysis Current Controlled Copies of: ] Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams ( Reference Plant Emergency Procedures l Reference Plant Operating Procedures OverallTest Results: SATISFACTORY Results
Description:
i t in accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simulated malfunction. ,_ i ES-t N l GPU Nuclear. Inc. Mddletown. PA Copynght 1997 l
- \
o
e (g' THREE MILE ISt AND Number
\t TRAINING G EDUCATION j mucuran SIMULATOR MANAGEMENT PRS-TTSS1 TITLE Revision No.
2 I OTSG PRESSURE TRANSMITTER FAILURE CERTIFICATION TEST !
' The event was initiated by failing both OTSG 1 A steam pressure transmitters to 1200 PSIG The first transmitter was failed instantaneously to 1200 PSIG to cause a SASS actuatior and transfer to the second transmitter. The second transmRter was failed high on a 120-second ramp to 1200 PSIG. This false high prescure indication caused OTSG 1 A Turbine Bypass Valves and Atmospheric Dump Valve to open fully'in an attempt to retum OTSG pressure to within normal limits.
R actor powerinitially increased due to a reduction in Tave, which was caused by the increase in steam flow from the open OTSG 1 A Turbine Bypass Valves and Atmospheric Dump Valve. The ICS also increased reactor and feedwater d: mand signals due to the resulting decrease in unit load (megawatts generated). This resulted in Control Rod Group 7 being withdrawn to the almost full out position. Main Steam header pressure decreased duo to the steam flow through the open Atmospheric Dump and Turbine Bypass Valves. The ICS system also opened the Main Turbine control valves in an attempt to maintain target unit load. The failed (high) OTSG 1 A pressure signal developed an invalid BTU Limit within the ICS on Feedwater Loop A, reducing Loop A Feedwater Demand. As a result of the rapid reduction in actual Loop A Feedwater Flow, FW Cross Limits developed, and the ICS was automatically transferred to the Tracking Mode. From this point, ICS automatically reduced reactor power demand, following actual unit load (megawatts generated). The Main Turbine control valves were automatically throttled to maintain steam header pressure at 885 F SIG, until the DTCS switched to manual control on h;ader pressure error. Th3 automatic Loop A Feedwater Flow reduction (by BTU limits) was terminated when OTSG 1 A reached ICS Low Level Limits. This action, combined with the continuing reduction in reactor ' power, resulted in a decrease in OTSG 1 A ressure, an increase in Loop A Feedwater Flow and s decrease in Loop B Feedwater Flow. Loop A and B Feedwater ows then began to oscillate attemately, responding to OTSG pressure cycles. OTSG 1B Turbine Bypass Valves sponded correctly to the pressure oscillations (the setpoint bias shifted to 10 PSI at low load). OTSG 1 A Main Steam isolation Vahres (stop check) closed due to low DP with low steam flow to the Main Turbine.
"h3 transient was terminated with reactor power at 28% and the Main Generator at low load. OTSG 1 A Turbine Bypass Valves and Atmospheric Dump Valve were full open. OTSG 1B Turbine Bypass Valves and Atmospheric Dump Valve were responding in automatic to the steam pressure oscillations.
This test differed from the 1994 test in that the Main Turbine did not trip. The Digital Turbine Control System responded more smoothly to the transient and remained on-line in manual control. This made the system response smoother than the pievious test. During the conduct of this tent simulator dynamic response, annunciator operation, and automatic safety system cctuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by select GPUN personnel representing TMI Simulator Training. The TMI representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated forimpact and corrective action in eccordance with Functional Fidelity Procedure,6511 ADM-2820.02. The results of these evaluations are as follows: l l l i i ES-2 ( (j CPU Nuclear. Inc. [ M,ddletown. PA Copynght 19S7 ! ______________J
THREE MILE ISLAND Number TRAINING & EDUCATION nucuAn SIMULATOR MANAGEMENT PRS-TTS51 TITLE Revision No. 2 OTSG PRESSURE TRANSMITTER FAILURE CERTIFICATION TEST The resuits or this test are satisf actory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the ' alarm or automatic action, and conversely, the simulator shall not causa an alarm or autotratic action if the reference plant would not cause an alarm or automatic action.
~
The simulator satisfied this requirement
N GPU Nuclear,Inc. fWhddletown, PA Copynght 1997 ,
y
@ THREE MILE ISLAND Number TRAININ2 & EDUCATION .
wuctaAA SIMULATOR MANAGEMENT PRS-TTS52 l TITLE Revisxm No. l ! EMERGENCY D!ESEL GENERATOR FAILURE CERTIFICATION TEST 3 afeguards equipment. P ant control was maintained at full power, in automatic mode The current revision to the Reference Plant Operating Procedure OP 1107-4, Electrical Distribution Panel Listing was used to verify affected component responses. Data plots showed a peculiar sine wave oscillation in RCS pressure, temperature and pressurizer level. The test was re-run in an attempt to identify the source of the oscillation. The oscillation was not present and the RCS was responding as expected. Discussion with Process Computers personnel failed to identify a cause for either the oscillation or the differences in the responses between the two tests. The transient was terminated following verification of plant stability and loss of power per OP 1107-4. I During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by select GPUN personnel representing TMI Simulator Training. The TMl representatives have held NRC Senior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corective action in l acordance with Functional Fidelity Procedure,6511-ADM-2820.02. The results of these evalautions are as follows: l The results or this test are satisfactory, ba ~' upon the following: CRITERIA #1 l I The observable changes in simulator parameters shall correspond in direction to those expected from the actual j reference plant and do not violate the physical laws of nature. I l The simulator satisfied this requirement with no identified ceviations. l O \
! CRITERIA #2
'w.
The simulator shall not fail to cause an alarm or automatic action if the reference plant owuld have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. Corrective Actions / Comments: NONE 1 1 Completed by: Nr OC William A. Fraser W l" Date Approved by: df - Robert Pamell 7k Operations Training Manager Date ( , i / ES-2 1
\ GPU Nuclear. Inc. l Middletown. PA !
Copynght 1997 l 1
., THREE MILE ISLAND Number TRAINING & EDUCATION
! nuctsAn SIMULATOR MANAGEMENT PRS-TTS53 TITLE Revision No. 2 EMERGENCY FEEDWATER CERTIFICATION TEST ( THREE MILE ISLAND l SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Emergency Feedwater Failure Certification Test Transient Test TTS53 ANSI /ANS 3.5 Reference (s): 3.1.2(23) Test Date: 01/19/98 { Mfunction(s) Tested *: FW19A Emergency Feedwater Control Valve Failure EF-V-30A 0% (100% - Full Open) FW19D Emergency Feedwater Control Valve Failure EF-V-30D 0% (100% - Full Open) ED01 - Unit Blackout
- Refer to Malfunction Cause and Effects Documents for options available.
-Test initialization: Protected Initial Condition 10-17 3
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100% Reactor Power Equilibrium Xenon Endof Cycle Point of Test Termination: This test was terminated with a loss of Off-site Power, Emergency Diesel Generators operating and removal of decay heat via the Emergency Feedwater System feeding OTSG 18 only. Simulator Run Time: 40 minutes Simulator Evaluation Time: 2 hours Baseline Evaluation Data: Previous Certification Test dated 03/14/94 Current Controlled Copies of: Reference Plant . Alarm Response Procedures Reference Plant Electrical One-Line Diagrams Reference Plant Emergency Procedures Reference Plant Operating Procedures Overall Test Results: SATISFACTORY i
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\\/ ) EG-1 GPU Nuclear Inc.
Mddletown PA Copynght 1997 L,______. _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _
, THREE MILE CLAND Number l
TRAINING & EDUCATICN ' muetsAN SIMULATOR MANAGEMENT PRS-TTS53 TITLE
]
Revision No. l ' p) 2 l (V Results
Description:
EMERGENCY FEEDWATER CERTIFICATION TEST l In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to perform correctly following activation of a simLlated malfunction. l This transient resulted in a loss of Off-site Power with removal of decay heat via OTSG 18 on natural circulation. i The event was initiated by failing OTSG 1A Emergency Feedwater Regulation Valves closed and initiating a loss of Off- { sit 3 Power. The reactor and Main Turbine Tripped. All non-vital powered equipment lost power. Reactor Trip { Containment Isclation actuated on loss of all Reactor Coolant Pumps. The Emergency Feedwater Regulating Valves for OTSG 1 A ud not open. OT6G 1B was being fed from the Emergency Feedwater System and removing decay heat via q the Atmospheric Dump Valve. With nc Feedwater flow and modeled Steam Leakage present, OTSG 1 A boiled dry and began to depressurize. OTSG 1B level increased to the level required for natural circulation. Natural circulation was confimled through OTSG 18 only. The test operator utilized the current' revision of ATOG Procedure 1210-10, Abnormal Trsosients Rules, Guides, and Tables to evaluate this response. The transient was termiriated following the confirmation of natural circulation conditions through OTSG 18.
. The 1994 test evidenced RCS loop A back flow during natural circulation of RCS loop B. This phenomenon was not present in this test.
During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel presenting TMl Simulator Training. The TMl representatives have held NRC Senior Reactor Operator license for TMl it 1. Identified deviations from expected performance have been evaluated forimpact and corrective action in rdance with Functional Fidelity Procedure,6511-ADM-2820.02. The results of these evaluations are as fo! lows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physicallaws of nature, The simulator satisfied this requirement with one identified deviation. Corrective Actions / Comments:
Description:
Lack of Backflow in 'A' Loop Number of Identified Deviations: One porrective Actions: One Discrepancy Report Correction (g3 es2 ) Meddlatown. PA Copynght 1997
THREE MILE ISLAND Number M NUCLEAR TRt?;W"J G EDUCATION SIMULATORS MANAGEMENT PRS-TTSS3 TITLE Revision No. 2 EMERGENCY FEEDWATER CERTIFICATION TEST CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant ~.ould have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or aute matic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with one identified deviation.
' Corrective Actions / Comments:
Description:
Incorrect RPPC Alarms Operation - L2781, L2792, L2803, L2814 Number of Identified Deviations: One Corrective Actions: One Dir repancy Report Correction Ccmpleted by: <P'M '#7 - William A. Fraser 3 ~M- M Date proved by: Robert Pamell Operations Training Manager Date 9
\ ) ES-3 sf GPU Nuclear,Inc.
I Middletown, PA Copynght 1997 J
) 'h THREE MILE ISLAND Number (GN sucurAN TRAINING & EDUCATION SIMULATOR MANAGEMENT PRS-TTS54 TITLE Revision No.
ESAS ACTUATION FAILURE CERTIFICATION TEST l THREE MILE ISLAND SIMULATOR MANAGEMENT ! CERTIFICATION TEST AB. STRACT j i Test identification: ESAS Actuation Failure Certification Test i Transient Test ! TTS54 . ANSI /ANS 3.5 Referencefsi: 3.1.2(23) Tut Date: 01/09/98 Malfunction (s) Tested *: ES01B ESAS Failure to Actuate at HPI Setpoint Channel B TH0S RCS LOCA at C RCP Suction at .15% Test Initialization: Protected initial Condition 10-16 100% Reactor Power Equilibrium Xenon Middleof Cycle Point of Test Termination: This test was terminated following reactor trip and ESAS actuation with Reactor 9 Simulator Run Time: Coolant System pressure decreasing. OTSG pressure and temperature were approaching normal post trip values. 3.5 minutes Simulator Evaluation Time: 3 hours Baseline Evaluation Data: Certification Test TTS54 dated 02/01/94 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY I ES-1 C ~ l oa$aY$i Copynght 1997 u__ _ _ - _ _ _
1 THREE MILE ISLAND Number TRAINING G EDUCATION nucurAs SIMULATOR MANAGEMENT PRS-TTS54
'rlTLE Revtskri No.
ESAS ACTUATION FAILURE CERTIFICATION TEST Results Descrimligm
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In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to ' perform correctly following activation of a simulated malfunction. This transient resulted in a reactor trip followed by Engineered Safeguards actuation on low Aeactor Coolant System pressure. The reactor trip initiated a Main Turbine trip and Reactor Trip Containment isolation. The event was initiated by causing a small rupture of the Reactor Coolant System after failing the B-ESAS channel High Pressure injection actuation circuits. The Reactor Protection System tripped the reactor on low Reactor Coolant System pressure. The Main Turbine tripped and Reactor Trip Containment isolation closed specified containment isolation v;lves. Reactor Coolant System pressure decreased to the actuation setpoint for High Pressure injection. The A-ESAS channel actuated -- but B-ESAS channel did not actuate. The transiert was terminated following verification of the B-ESAS channel f ailure to actuate. with Reactor Coolant System inventory loss being supplied by the A-High Pressure injection System. Reactor Building pressure and temperature were increasing due to the LOCA in progress. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. GPUN personnel representing TMI Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMl representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance ith Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisf actory, based upon the following: CRITERIA #1 The observable changes in simulator parameters shall. correspond in direction to those expected from the actual 4 reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 1 The simulator shall not f ail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. I i
. Corrective Actions / Comments:
f- NONE o E5-2
,f g l
CPU Nuclear. Inc. Mddlstown, PA ( f Copynght 1997 l
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THREE MILE ISLAND Number l
- TRAINING & EDUCATION muetsAs SIMULATOR MANAGEMENT PRS-TTS54 l l
. TITLE Remsson No.
' ESAO ACTUATION FAILURE CERTIFICATION TEST
*.} l l
l ! Completed by: Charles E. Husted Nf f8 Date l Approved by: d.A. Im Robert Parnell 3 A' 9f Operations Training Manager Dats 1
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l l l l l l l P GPU Nuclear,Inc. Mddletown, PA
.( Coppight 1997
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k THREE MILE ISLAND Number TRAINING & EDUCATION MUCLEAN SIMULATOR MANAGEMENT PRS-TTS55 TITLE Revision No. A)
! ATWS CERTIFICATION TEST 4
LJ l THREE MILE ISLAND i ' SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT l l Test identification: ATWS Certification Test l Transient Test l TTS55 l ANSl/ANS 3.5 Reference (s): 3.1.2(24) Test Date: 01/09/98 Malfunction (s) Tested *: RD28 Reactor Auto Trip Block RD32 Diverse Scram System Automatic Actuation Failure TC01 Turbine Trip Test Initialization: Protected Initial Condition 10-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle This test was terminated with the reactor ope,ating at 24% power, the turbine tripped and all power being removed through the Turbine Bypass Valves to the condenser. Reactor Coolant System pressure and pressurizer level were stable under automatic control. Simulator Run Time: 12 minutes Simulator Evaluation Time: 6 hours Baseline Evaluation Data: Previous Certification Test TTSS5 dated 01/25/96 Right Direction Analysis Simulator Malfunction Cause and Effects Document Current Controlled Copies of: Reference Plant Alarm Response Procedures Reference Plant Electrical One-Line Diagrams OverallTest Results: SATISFACTORY L E5-1 [ GPU Nuclear, Inc. Mddletown, PA copyr:ght 1997
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- THREE MILE ISLAND Number TRAINING & EDUCATION muctr4m SIMULATOR MANAGEMENT PRS-TTS55 TITLE Revision No.
4 ATWS CERTIFICATION TEST I Results
Description:
f l In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the ability of the simulator to l perform correctly Iollowing activation of a simulated malfunction. { This transient resulted in the loss of the Main Turbine as a heat removal path for the reactor and the Reactor Protection l System failing to respond as required. The Integrated Control System responded to the loss of the Main Turbine i generator output reducing reactor power to within the heat removal capacity of the Turbine Bypass Valves. The event was init#.ed by tripping the Main Turbine and preventing the automatic trip of the reactor by failing the Reactor Protection System and the Diverse Scram System. The integrated control System responded to the loss of generator' output and reduced reactor power by continuous Control Rod insation. The increase in Reactor Coolant system temperature resulted in a reduction in reactor power due to temperature coefficient feedback. Tripping of the Reactor Protection System resulted in actuation of Reactor Trip Isolation of the containment. Reactor Coolant System pressure increased causing the Pressurizer Spray Valve to open to prevent an overpressure condition. The spray valve closed when normal pressure was reached. Oscillations in OTSG level resulted in actuation of Heat Sink Protection System low level start-up of the Emergency Feedwater Pumps. The OTSG Pressure oscillation was initiated by Main Steam Relief valve operation while steam flow exceeded the capacity of the bypass valves. When power decreased to within the capacity of the bypass valves the OTSG Pressure oscillation severity began to decrease. OTSG level and Feedwater flow began to stabilize at setpoint under automatic control. A'he transient was terminated when OTSG level and Feedwater flow approaching stability. The reactor was operating
/ Th a high load limit imposed, oscillating reactor power with reduced T-average.
U During the conduct of this test simulater dynamic response, annunciator operation, and automatic safety system actuations were recorded. GPUN personnel representing TMl Simulator Training have evaluated test results against ANSI /ANS 3.5 criteria. The TMI representatives have held NRC Senior Reactor Operator license for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure,6511-ADM-2820 02. The results of these evaluations are as follows: The results or this test are satisf actory, based upon the following: ; CRITERIA #1 The observable changes in simulator parameters shall correspond in direction to those expected f rom the actual I reference plant and do not violate the physical laws of nature.
.The simulator satisfied this requirement with no identified deviations.
CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviations. l ES-2 7'N GPU Nucinar, Inc. l
) Mddletown, PA
{ j V Copynght 1997 l i
.A THREE MILE ISLAND Number TRAINING & EDUCATION muctrAs SIMULATOR MANAGEMENT PRS-TTS55 TIT E Revision No.
r% 4
-[
V(__ ATWS CERTIFICATION TEST Corrective Actions / Comments: NONE
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Completed by: , Charles E. Husted 3kk8' Date Approved by: Operations Training Manager Robert Parnell 3!1>h8' Date
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1 l l l l E5-3 O GPU Nxiear,Inc. Mdaletown. PA Coppght 1997 l i m___.___.__.._____.__.____
O THREE MILE ISLAND Number TRAINING G EDUCATION suctsas SIMULATOR MANAGEMENT PRS-NOT20 TITLE Revtsson No. [~NI 2 TURBINE OVERSPEED TESTING SURVEILLANCE CERTIFICATION TEST V . THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Turbine Overtpeed Testing Surveillance Certification Test Nonnal Operations Test NOT20 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(10) Test Date: 01/19/98 Reference Plant Procedure: SP 1303-11.19 Turbine Overspeed Testing Revision 17 Test initialization: Protected initial Condition IC-08 , Hot Zero Power { 1E-8 Amps I Xenon Free Startup Middle of Cycle - T - This test was terminated when overspeed testing was complete. imulator Run Time: 30 minutes Baseline Evaluation Data: Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Surveillance Procedures Overaillest.Results; SATISFACTORY o 1 GPU Nuclear,Inc. Middletomi, PA Copynght 1997
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THREE MILE ISLAND Nunber TRAINING & EDUCATION wucurAs SIMULATOR MANAGEMENT PRS-NOT20 TITLE
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Revision No. 2 TURBINE OVERSPEED TESTING SURVEILLANCE CERTIFICATION TEST j Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the reference plant. During the conduct of this test simulator dynamic response, annunciator operation, and automatic safety system actuations were recorded. Test results have been evaluated against ANSI /ANS 3.5 criteria by GPUN personnel representing TMl Simulator Training. The TMI representatives have held NRC SSnior Reactor Operator licenses for TMI Unit 1. Identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator satisfied this requirement with no identified deviatior; . l CRITERIA #2 The observable changes in simulator parameters correspond in direction to those expected from the actual (^)
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reference plant and do not violate the physicallaws of nature. I l The simulator satisfied this requirement with no identified deviations. CRITERIA #3 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. l CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. V E4-2 GPU Nuclear,Inc. hmddletown. PA Copynght 1997
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THREE MILE ISLAND Number j (~ TRAINING & EDUCATlHN ! aructaAs SIMULATOR MANAGEMENT PRS-NOT20 TITLE Remsm No. 2 TURBM OVERSPEED TESTING SURVEILLANCE CERTIFICATION TEST 4 7 Corrective Actions / Comments: 1 During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
)
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
Completed by:
/d##M '
4 - _- - William A. Fraser f- is -9T Date 9 Approved by: N Robert Parnell 3-/! Operations Training Manager Date l l 1 l"% (y E4-3 - GPU Nuclear,Inc. Modletown. PA Coppight 1997 , I L _ _ . _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
THREE MILE ISLAND Number TRAININ3 S EDUCATION NUCLEAs SIMULATOR MANAGEMENT PRS-NOT21 TITLE Revtsion No. 3 LOW PRESSURE INJECTION SURVEILLANCE CERTIFICATION TEST THR" . MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Low Pressure injection Surveillance Certification Test Normal Operations Test NOT21 ANSI /ANS 3.5 Referencels): 4.2.1 . 3.1.1(10) , rest Date: 03/02/98 Reference Plant Procedure: SP 1303-11.54, LPI Test Revision 14 Test initialization: Protected initial Condition 1C-01 Refueling Shutdown Fot Legs Drained Beginning of Cycle . Point of Test Termination: This test was terminated following conduct of testing per SP 1303-11.54. 30 minutes Baseline Evaluation Data: Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: , Reference Plant Surveillance Procedures Reference Plant Alarm Response Procedures. Overall Test Results: SATISFACTORY O E41 b "do"oO.'A Copyright 1997
[ h THREE MILE ISLAND Number
' TRAININ3 G EDUCATION
, suctsaa SIMULATOR MANAGEMENT PRS-NOT21 TITLE
^ Revision No.
l ' 3 LOW PRESSURE INJECTION SURVEILLANCE CERTIFICATION TEST Results
Description:
i in accordance with ANS!!ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the reference plant. i During the conduct of this test simulator dynamic performance, annunciator operation, and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls used during the test was also evaluated. These evaluations were conducted by GPUN personnel representing TMI Simulator Training. All the evaluators have held NRC Senior Reactor Operator licenses for TMI Unit 1. Procedure steps not performed and identifal deviations from expected performance have been evaluated for !mpact and conective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: I The results or this test are satisfactory, based upon the following: J
' 1 CRITERIA #1
)
1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this " test. . The simulator satisfied this requirement with no identified deviations. CRITERIA #2 (h The observable changes in simulator parameters correspond in direction to those expected from the actual b) reference plant and do not violate the physicallaws of nature. The simulator satisfied this requirement with no identifiod deviations.
^
CRITERIA #3 The simulator shall not fall to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied ' tis r6quirement with no identified deviations. l t O E4-2 { GPu Nuclear, Inc.
'v Middletown PA Copynght 1997
b THREE MILE CLAND Numbw TRAINING G EDUCATION MUCf.KAm SIMULATOR MANAGEMENT PRS-NOT21 TITLE Revision No. 3 LOW PRESSURE INJECTION SURVEILLANCE CcRTinCATION TEST Corrective Actions / Comments: During the conduct of this test, reference plant pmcodure steps not performed were identified as test exceptions and evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result 41 the need for the test operator to violate the procedure in order to proceed with the evolution.
P. ~ fPaps not performed did not result in observaNs differaf.cas in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. lit accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
C:mpleted by: [///h4
#N William A. Fraser 30 O Date 4pproved by: 4., & Robert Pamell d/ I Operations Ttsining TVlanager Date l
l ['ij E4-3 GPU Nuclear,Inc.
\
/ leddletown, PA Copynght 1997
THREE MILE ISLAND Number s TRAINING & EDUCATION nuctsAs SIMULATOR MANAGEMENT PRS-NOT22 d TITLE Rension No. (~N) i 2 PLANT SHUTDOWN CERTIFICATION TEST V THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Plant ShJtdown Certification Test Normal Operations Test NOT22 ANSI /ANS 3.5 Reference (sh 4.2.1 3.1.1(6) t 3.1.1(8) Test Date: 09/25/97 Reference Plant Procedure: OP 1102-10, Plant Shutdown Revision 82 i Test initialization: Protected initial Condition 1C-17 100% Reactor Power Equilibrium Xenon l End of Cycle j rmin - Hot Shutdown. 532 Degrees Tave. Safety Rods Withdrawn. Simulator Run Time: 2 Hours Baseline Evaluation Data: Tactile Feel by Experienced Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures Reference Plant Alarm Response Procedures Overall Test Results: SATISFACTORY I
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, t GPU Nucf ear. Inc. Mddletown. PA Copregne 1997 A
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THREE MILE ISLAND Number
.' {
TRAINING & EDUCATION t NUCLEAN SIMULATOR MANAGEMENT PRS-NOT22 '* ) i TITLE
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Reasson No. 2 PLANT SHUTDOWN CERTIFICATION TEST Results Description- I I In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in accordance with reference plant operating procedures, usirg only operator action normal to the reference plant. During the conduct of this test simulator dynamic performance, annunciator operation, and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls used during the test was also evaluated. GPUN personnel representing TMI Simulator Training conducted these evaluations. The evaluators have held a NRC Senior Reactor Operator license for TMl Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in accWance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluatons are as follows: The results of this test are satisf actory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 f f The observable changes in simulator parameters correspond in direction to those expected from the actual C reference plant and do not violate the physical laws of nature.
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The simulator satisfied this requirement with no identified deviations. CRITERIA #3 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. i CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. 0
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E4-2 GPU Nuclear,Inc. j ( Mddletown. PA Copynght 1997
r THREE MILE ISLAND Number l l
#, TRAININGB EDUCATION nuct. san PRS-NOT22 SIMULATOR MANAGEMENT s TITLE
- Ramsson No.
2 PLANT SHUTDOWN CERTIFICATION TEST l Corrective Actions / Comments: During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and ! evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows: l
- 1. t S'eps not performed did not result in the need for the test operator to vlotate the procedure in order to proceed wit the evo!ution.
- 2. Steps not performed did not result in observable differences in the control room.
t
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and I precautions, technical specifications or procedure acceptance criteria. l
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
Charles Husted Completed by: _ fo/t _ M ,, JM f8 i D' ate roved by:
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Operations Training franager Robert Pamell Al N ; Da'te { I l l s e l I 1 I
/ E4-3 ! GPU Nuclear,Inc.
k.) . Mddletown. PA Copynght 1997
THREE MILE ISLAND Number l 33g TRAINING & EDUCATION n uctreur SIMULATOR MANAGEMENT PRS-NOT23 TITLE Revtskm No. 2 PLANT COOLDOWN CERTIFICATION TEST THREE MILE ISLAND l SIMULATOR MANAGEMENT l CERTIFICATION TEST ABSTRACT Test identification: Plant Cooklown l Norrnal Operations Test NOTM ! ANSI /ANS 3.5 Reference (s): 4.2.1
- 3.1.1(8) l
' Test Date: 09/29/97 Reference Plant Procedure: OP 1102-11, Plant Cooldown Revision 11.5 Test Initialization: Protected initial Condition 10-06 H1 Shutdown Sa.ety Rods Out Xenon Free End of Cycle P n fT T rm n i n- 140 Degrees Tave.
RCS Pressure 40 PSIG. Steam Bubble in Pressurizer. Decay Heat Removalin Operation. Simulator Run Time: 8 hours Baseline Evaluation Data: Tactile Feel by ExperierrAd Operators. Right Direction Analysis. Current Controlled Copies of: Reference Plant Operatirg Procedures Reference Plant Alarm Response Procedures. Overall Test Results: SATISFACTORY l I i i p) E4-1 GPU Nuclear, Inc. N / Mddletown. PA Copyr.ght 1997 l
THREE MILE ISLAND Number TRAINING & EDUCATION
', muetsm SIMULATOR MANAGEMENT PRS-NOT23 TITLE Revison No.
2 PLANT COOLDOWN CERTIFICATION TEST i l Results
Description:
l In accordance with ANSI /ANS-3.5-1985 this certification test is conducted to demonstrate the ability to operate the simulator in accordance with reference plant operating procedures, using only operator action normal to the reference plant. ! l During the conduct of this test simulator dynamic performance, annunciator operation, and automatic actions were evaluated against ANSI /ANS 3.5-1985 criteria. Tactile feel of simulator controls used during the test was also evaluated. GPUN personnel representing TMI Simulator Training conducted tnese evaluations. All the evaluators have held a NRC Senior Reactor Operator license for TMI Unit 1. Procedure steps not performed and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2820.02.
- The results of these evaluations are as follows:
The results of this test are satisfactory, based upon the following:
.Q. BIT.E.Ble 11 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test.
The simulator supported this requirement with no deviations. CRITERIA #2 o\ The observable changes in simulator parameters correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. (d The simulator satisfied this requirement with no identified deviations. CRITERIA #3 The simulator sha!! not fail to cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and converstely, 'he simulator shall not cause an alaim or automatic action if the reference plant would not cause an alarm or automatic action. The simulator satisfied this requirement with no identified deviation. CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. O E4-2 I GPU Nuclear, Inc. l q Mddletown, PA Copynght 1997
THREE MILE ISI.AND Number l TRAINING G EDUCATION ! 1, muctaAs SIMULATOR MANAGEMENT PRS-NOT23 { TITLE Revtskri No. 2 l PLANT COOLDOWN CERTIFICATION TEST {
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{ Corrective Actions / Comments: NONE During the conduct of this test, reference plarit procedure steps not performed were identified as test exceptions and evaluated.for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221 ADM-2820.02. The results of these evaluations are es follows:
- 1. Steps not performed did not result,in the need for the test operator to violate the procedure in order to proceed with the evolution. .
I
- 2. Steps not performed did not result in observabl6 differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria. ,
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have beep documented and closed.
Dmpleted by: m tel- ff Charles E. Husted '3 M 98
/ date Approved by: F Robert Pamell I //
O% rations Training Manager Date i
/ \ E4-3
( GPU Nuclear,Inc. l
\. M:Idletown, PA l Copright 1997 j
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, THREE MILE ISLAND Nurrber 1
CU muets4n TRAININ3 G EDUCATION SIMULATOR MANAGEMENT PRS-NOT24 TITLE Reviskr1 No. A 2 l V) SHUTDOWN MARGIN & REACTIVITY BALANCE CERTIFICATION TEST
~THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Shutdown Margin and Reactivity Balance Certification Test '
Normal Operations Test NOT24 ANSI /ANS 3.5 Referencois): 4.2.1 3.1.1(9) Test Date: 01/26/98 Reference Plant Procedure: OP 1103-15A, Shutdown Margin and Reactivity Balance Revision 25 Test initialization: Protected initial Condition IC-17 100% Reactor Power Equilibrium Xenon End of Cycle This test was completed when the data was collected and calculations completed per OP 1103-15A. , Simulator Run Time: 30 minutes Baseline Evaluation Data: Current Controlled Copies of: Reference Plant Operating Procedures Overall Test Results: SATISFACTORY Results
Description:
In accordance with ANSI /ANS-3.5-1985 this certification test was conducted to demonstrate the simulator in accordance with similar reference plant operating procedures, using only operator action normal to the reference plant. During the conduct of this test simulator dynamic response, annunciator operation, and automatic actions were evaluated against ANSI /ANS 3.51985 criteria. Tactile feel of simulator controls used during the test was alsn evaluated. These ev luations were conducted by GPUN personnel representing TMI Simulator Training. All the evaluators have held NRC Senior Reactor Operator licenr.es for TMI Unit 1. Procedure steps not perfonned and identified deviations from expected performance have been evaluated for impact and corrective action in accordance with Function Fidelity Procedure 6221 ADM-2820.02. The results of these evaluations are as follows: l I E4-1 - dd et .N Copynght 1997 l I
THREE MILE CLAND Number
/
TRAINING S EDUCATION
, muetsAN SIMULATCR MANAGEMENT PRS-NOT24 TITLE Revtsson No.
SHUTDOWN MARGIN & REACTIVITY BALANCE CERTIFICATION TEST The results or this test are satisfactory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the reference plant procedure (s) used during the conduct of this test. CRITERIA 1 evaluators concurred that during this test, the simulator supported this requirement with no deviations. CRITERIA #2 l The observable changes in simulator parameters correspond in direction to those expected from the actual ! reference plant and do not vioiste the physical laws of nature. CRITERIA 2 evaluatois concurred that during this test, the simulator supported this requirement with no deviations. . j CRITERIA #3 The simulator shall not fail to cause an alarm or automatic actuation if the reference plant would have caused an ['] alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. (U) CRITERIA 3 evaluators concurred that during this test, the simulator supported this requirement with no
. deviations.
CRITERIA #4 Tactile feel of Plant-Referenced Simulator control device (s) compares to those of the reference plant. CRITERIA 4 evaluators concurred that during this test, the simulator supported this requirement with no deviations. 1 . E4 2
/, ,_ g GPU Nuclear,Inc.
\ l Middletown. PA
/ Copynght 1997
THREE MILE CLAND Number TRAINING & EDUCATION
,- muetsas SIMULATOR MANAGEMENT PRS-NOT24 TITLE
- Revtsx5iUo.
2 SHUTDOWN MARGIN & REACTIVITY BALANCE CERTIFICATION TEST Corrective Actions / Comments: During the conduct of this test, reference plant procedure steps not performed were evaluated for impact and corrective action requirements in accordance with Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
- 1. Steps not performed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Steps not performed did not result in observable differences in the control room.
- 3. Steps not performed did not prevent the successful completion of the procedure in accordance with plant limits and precautions, technical specifications or procedure acceptance criteria.
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented "J closed.
(7:mpleted by: " [88MM '
&# William A. Fraser 3-/W9P l ) Date pproved by: 8 Operations Training Manager Robert Pamell 4//D Date l
l E4-3 f\ GPU Nuckar,Inc.
/
) .
Middietown. PA Copynght 1997
]
- ' THREE MILE ISLAND Number
# TRAINING & EDUCATION
, muctawr SIMULATOR MANAGEMENT PRS-NOT25 TITLE Revison No.
HEAT BALANCE CALCULATION CERTIFICATION TEST THREE MILE ISLAND SIMULATOR MANAGEMENT ] CERTIFICATION TEST ABSTRACT Test identification: Heat Balance Calculation Certification Test Normal Operations Test NOT25 ANSI /ANS 3.5 Reference (s): 4.2.1 3.1.1(9) Test Date: 03/20S 8 Reference Plant Procedure: OP 1103-16. Heat Balance Calculations Revision 28 Test Initialization: Protected initialCondition 10-17 100% Reactor Power Equilibrium Xenon Endof Cycle Point of Test Termination; Calculation of Heat Balance Completed. - 100% Reactor Power. 9 579 Degrees Tave. ICS in Full Automatic. Steam Bubble in Pressurizer. Simulator Run Time: 20 minutes Baseline Evaluation Data: Previous Certification Test NOT25 dated 3/21S4. Tactile Feel by Experienced Operators. . Right Direction Analysis. Current Controlled Copies of: Reference Plant Operating Procedures OverallTest Results: SATISFACTORY 9 E41 So"dNo74.IA Copright 1997
i i THREE MILE ISLAND Number
- TRAINING & EDUCATION
; nucuuur SIMULATOR MANAGEMENT PRS-NO T25 TITLE Revision No.
' 2 i HEAT 9ALANCE CALCULATION CERTIFICATION TEST l
' Results
Description:
in accordance with ANSl/ANS-3.5-1985 this certification test is conducted to demonstrate the aoility to operate the simulator in accordancC with similar reference plant operating procedures, using only opC E4t. 5 n normal to the reference plarrt. During the conduct of this test simulator dynamic performance, annunciator operation, and automatic actions w9re evaluated ' against ANSI /ANS 3.5-1985 criteria. Tactile fevl of simulator controls used during the test was also evaluated. GPUN personnel representing TMl Sinidator Training conducted these evaluations. All the evaluators have held a NRC Senior Reactor Operator licerce for TMI Unit 1. Procedure steps not performed and iderstified deviations from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure 6221-ADM-2620.02. The results of these eduations are as follows: The resuits of this test are satisf actory, based upon the following: CRITERIA #1 The simulator has met the acceptance criteria of the refsrence plant procedure (s) used during the conduct of this test. The simulator supported this requirement with no deviations. CRITERIA 62
-m The observable changes in simulator parameters correspond in direction to those expected from the actual
[ ) reference plant and do not violate the physical laws of nature.
~~
The simulator satisfied this requirement with no identified deviations. CRITERIA #3 The simulator shall not fail to' cause an alarm or automatic actuation if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an aiarm or automatic action. The simulator satisfied this requirement with no identified deviation.
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CRITERIA #4
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Tactile feel of Plant Referenced Simulator control device (s) compares to those of the reference plant. The simulator satisfied this requirement with no identified deviations. 1 f] g E4-2 GPU Nuclear. Inc. ( l
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) Mddletown PA Copright 1997 l
THREE MILE ISLAND Number
- TFAINING & EDUCATION g suct m SIMULATOR MANAGEMENT PRS-NOT25 TITLE Revision No.
_ HEAT BALANCE CALCULATION CERTIFICATION TEST Corrective Actions / Comments: During the conduct of this test, reference plant procedure steps not performed were identified as test exceptions and evaluated for impact and corrective action requirementsin accordance with Functional Fidelity Procedure 6221-ADM-2820.02. The results of these evaluations are as follows: 1. Steps not peiformed did not result in the need for the test operator to violate the procedure in order to proceed with the evolution.
- 2. Gieps not performed did not result in observable ditforences in the control room.
- 3. Steps not [ performed did not prevent the sumessful comple'. ion of the procedure in accordance with pl precautions, technical specifications or pacedure acceptance criteria.
- 4. In accordance with 6221-ADM-2820.02, these test exceptions have been documented and closed.
Completed by: 4 Charles E. Husted ' 3 78 j_m / Date roved by: Operations Training Manager Robert Pamell M/Y Date [
}; GPU Nuclear. Inc.
Mddistown. PA l (/ Copynght 1997
rm ATTACilMENT C ( TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR
\ SIMULATOR CERTIFICATION PLAN TESTING SCIIEDULED Anaual Simulator Test Requirements 1(X)% of ANSI /ANS 3.5-1985 Benchmark Tests (10) 25% of Transient Test Series (TFS) 25% of Normal Operations Tests (NOT) 100% Steady State Performance Tests (SSP) 100% Meter AccuracyTests Real Time Test (RTT)
ANNUAL, TESTS llewchmark Tests TTS07 RCS Safety Valve Faihire TTS19 less of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Steam Leak Inside Reactor Building TTS56 Manual ReactorTrip TTS57 Simultaneous Closure of All Main Steam Isolation Valves TTS58 Loss of One Reactor Coolant Pump TTS59 Maximum Rate Po'wer Ramp . TTS60 less of Offsite Power with Design Basis LOCA Steadv State Tests SSP 01 Siraulator Stability SSP 02 Simulator Accuracy Real Time Test RTT01 RealTimeTest s i t _.
,m_ A'ITACIIMENT C i
s J TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR V SIMULATOR CERTIFICATION PLAN TESTING SCHEDULED YEAR #1 TESTS (1998-1999) Henchmark Tests TTS07 RCS Safety Valve Fa'ilure TTS19 Loss of Forced Flow TTS27 1.oss of All Feedwater TTS35 TurbineTrip TTS42 Main Steam Leak inside Reactor Iluilding TTS56 ManualReactorTrip TTS57 Simultaneous Closure of All Main Steam Isolation Valves TTS58 less of One Reactor Coolant Pump TTS59 Maximum Rate Power Ramp TFS60 1.oss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Siuulator Stability SSP 02 Simulator Accuracy Real Time Trgq
'] RTT01 Reid Time Test Transient Tests TTS01 OTSG Tube Leak 1TSO2 OTSG Tube Rupture TTS03 RCS Leak Inside Containment TTSO4 RCS Leak Outside Contairunent 1TS05 Large Break LOCA TTS06 Small Break LOCA 7TS08 RCS PORV Failure TTSO9 LossofInstrument Air TTS10 ' Station Blackout TTSI1 DC Distribution Failure TTS12 Emergency Diesel Generator TTS13 6.9 KV Bus Fault TTS14 4.16 KV Bus Fault Nornud Operations Tests i
NOT01 Plantlicatup < NOT02 Plant Startup )
.- NOT03 Reactor Trip and Recovery )
NOTO1 3 RC Pump Operation ; i NOT05 Zero Power Physics Testing ! NOT06 Core Fhni System Valve Operability Test q NOT07 Emergency Power System x f I
p A'ITACHMENT C . t TIIREE MILE ISLAND FLANT-REFERENCED SIMULATOR
' V) SIMULATOR CERTIFICATION PLAN TESTING SCHEDULED YEAR #2 TESTS (1999-2(KMI)
Benchmark Tests TTS07 RCS Safety Valve Failure TTS19 Loss of Forecd Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Stetun Leak Inside Reactor Building 1TS56 Manual Reactor Trip
'ITS57 ' Simultaneous Closure of All Main Steam isolation Valves TTS58 Loss of Or.e Reactor Coolant Pump TTS59 Maximum Rate Power Ramp TTS60 Loss of Offsite Power with Design Basis LOCA Steady State Tests SSPoi . Simulator Stability SSP 02 Simulator Accuracy Real Time Test
/O RTT01 RealTimeTest
)
. Transiera Tests TTS15 480 V Bus Fault TTS16 480 V MCC Fault 1TS?7 ICS Auto Power Failure TTS18 Inverter Failure 1TS20 Loss of Condenser Vacuum TTS21 Condenser is: vel Control Failure 1TS22 Loss of Service Water TTS23 Loss of Shutdown Cooling TTS24 Loss of Component Cooling ,
1TS25 Loss of Normal Feedwater TTS26 Normal Feedwater System Failure l TTS28 Loss of Protective System Channel 1TS29 Stuck Control Rod Normal Operatiim Tests . NOT08 RB 30 PSIG Analog Cluumels NOT09 Main Steam Isolation Valves Surveillance NOT10 Main Stenen Isolation Valves Monthly Surveillance NOTI 1 Shift and Daily Checks Surveillance NOT12 Weekly Surveilhmce Checks NOT13 RCS Leakrate f i /
'V I j
73 ATTACIIMENT C C'; ' TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR SIMULATOR CERTIFICATION PLAN TESTING SCIIEDULED YEAR #3 TESTS (2(HMI-2(H)]) Benchmark Tests TTS07 RCS Safety Valve Failure TTS19 Loss of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip TTS42 Main Steam Leak Inside Reactor Building TTS56 Manual Reactor Trip TTS57 Simultaneous Closure of All Main Steam Isolation Valves TTS58 Loss of One Reactor Coormt Pump TTS59 Maximum Rate Power Ramp TTS60 Loss of Offsite Power with Design Basis LOCA Steady State Tests SSP 01 Simulator Stability . SSP 02 Simulator Accuracy Real Time Test Cg RTT01 Real Time Test
\ J V Transient Tests TTS30 Continuous R(xlinsertion TTS31 Dropped Rod TTS32 Uncoupled R<x!
TTS33 Inability to Drive R(xis TTS34 Failed Fuel TTS36 GeneratorTrip TTS37 Inadvertent OTSG Isolation TTS38 Imidvertent OTSG Overfcui TTS39 Pressurizer Level Control Failure TTS40 Pressurizer Heater Failure TTS41 Reactor Trip TTS43 Main Steam Leak Outside Reactor Building Normal Operations Tests NOT14 Control R(xl Movement NOTIS RB Cooling and Isolation System Logic Channel And Component Test NOT16 Loading Sequence and Component Test and ! ' HPI Logic Channel Test NOT17 High Pressure injection NOT18 RB Emergency Cooling System NOT19 ES System Emergency Sequence and Power Transfer Test
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m ATTACIIMENT C f\ k TIIREE MILE ISLAND PLANT-REFERENCED SIMULATOR SIMULATOR CERTIFICATION PLAN TESTING SCIIEDULED YEAR #4 TESTS (2(Hil-2(N)2) Benchmark Tests TTS07 RCS Safety Valve Failure TTS19 Loss of Forced Flow TTS27 Loss of All Feedwater TTS35 TurbineTrip iTS42 Main Ste:un Leak Inside Reactor Building TTS56 ManualReactorTrip TTS$7 Simultaneous Closure of All Main Steam Isolation Valves TTS58 Loss of One Reactor Coolant Pump TTS59 Maximum Rate Power Ramp TTS60 Loss of Offsite Power with Design Basis LOCA Steadv State Tests SSP 01 Simulattr Stability SSP 02 Simulator Accuracy Real Time Test
/ RTT01 Real Time Test Transient Tests TTS44 Main Feedwater Line Break inside Reactor Building 7TS45 Main Feedwater Line Break Outside Reactor Building TTS46 NI-5 Failure TTS47 NI-6 Failure TTS48 Pressurizer Level Control TTS49 Feedwater Flow Transmitter Failure TTS50 RC Cold Leg Temperature Transmitter Failure TTS51 OTSG Pressure Transmitter Failure TTSS2 Emergency Diesel Generator Failure TTS53 Emergency Feedwater Failure TTS54 ESAS Actuation Failure TTS55 ATWS Normal Operati<m Tests NOT20 Turbine Overspeed Testing . ,
NOT21 Low Pressure Injection l NOT22 -Plant Shutdown . NOT23 Plant Cooldown l NOT24 Shutdown Margin and Reactivity Balance l NOT25 Heat Balance Calculation j l I 1 (. I i
IP eI ATTACHMENT D
' ' The TMI Replica Simulator was re-hosted in January of 1997 Re-hosting was necessary due to the age, reduction of vendor support, and increasing inadequacy of the original.
simulation computer systems to support future modeling enhancements. The 1984 - vintage ENCORE 9780 and ENCORE 9750 mini computers were replaced with a single , . twin 200M9' "C running OPENSIM Simulation Support Software. A single 200MHz , PC pre , ides server support to the new simulation PC and three base level 200 MHz PC's use f as Instructor Stations. Each of the new systems utilize the Microsoft NT 3.51 op.: rating system and, as an integrated system, has provided significant increases in
, pro ,essor speed, storage capacity, reliability, as well as savings in operating costs and maintenance costs.
The original simulation models have been updated to take advantage of the new, more efficient compilers, however the basic model logic and dynamics has not changed. Microsoft's SQL Data Base Manager handles the database, resident on the Server. System drawings and various productivity enhancing software utilities for use by L Instructors' and Developer's are now resident, and readily available, on the Instructor L Station consoles, n- The I/O front end (Master Controller) was replaced. The original 15 year old wire-(- wrapped Master Controller was unreliable and has not been supported by the vendor for some time. The new front end consists of a High Speed Data (HSD) board emulator to , replace the aging and expensive ENCORE HSD, and a highly reliable digital Master - Controller. In total, the new platform is providing a very reliable, efficient, and user friendly training - device that also allows for significant enhancement capabilities and cost savings through lower maintenance and operating costs. ll l
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. . l THREE MILE ISLAND Number Cpgg TRAINING & EDUCATION NUCLEAR SIMULATOR MANAGEMENT OES-014 Revision No.
SE TURBINE VALVE FULL STROKE TESTING O
\
J THREE MILE ISLAND SIMULATOR MANAGEMENT CERTIFICATION TEST ABSTRACT Test identification: Turbine Valve Full Stroke Testing Certification Test OES014 ANSI /ANS 3.5 Reference (s): N/A NUREG 1258 3.1.1.4 Test Date: 12/29/97 Malfunction (s) Tested *: None Test initialization: Protected initial Condition IC-15 100% Reactor Power Equilibrium Xenon Beginning of Cycle Point of Test Termination: This test was terminated followilg completion of the valve testing. Simulator Run Time: 60 minutes ulator Evaluation Time: 2 hours seline Evaluation Data: Right Direction Analysis. Data Trends from the Plant Process Computer E-mail from Paul Dojka Current Controlled Copies of: Operating Procedure OP 1106-1 Appendix C.1.2.3 Overall Test Results: SATISFACTORY l l
'N y,
GPU Nuclear, Inc. Mddletown. PA Copynght 1997 L__-_---_-
, .s THREE MILE ISLAND Numtxw
(~~)3(f TRAINING & EDUCATION uuctm SIMULATOR MANAGEMENT OES-014
'E Remion No 0
TURBINE VALVE FULL STROKE TESTING
\
Results
Description:
1 In accordance with NUREG 1258 this certification test wa.s conducted to demonstrate the ability of the simulator to reproduce ref erence plant transient responses during a documented operational event This transient was a result of Main Tuttine full stroke testing of Stop valve (SV) and intermediate Stop and Control valves (CIV) conducted on the Main Turoine on 12/18/97. The plant response to valve testing was as expected by Plant Engineenng. There was minimal oscillation of plant parameters during the conducting of the test.' Th3 transient was initiated by reducing power to approximately 90%. When the plant was stable the Turbine Valve test was conducted utilizing Appendix C.1.2.3 of the current revision of OP 1106-1. Testing was performed on SV 1.2.3.4 but , not CIV-12.3,4,5,6 due to outstanding deficiencies in the Turbine Control System modification in progress at test time. When testing was completed the simulator was returned to full power. During the test ICS remained in full automatic control. During testing on the simulator OTSG pressure and Main Feedwater flow changes were in the same direction and similiar magnitude when compared to the plant. Turbine valve response to the simulated Digital Turbine Control System was the same as the plant. The test was terminated following completion of stop valve testing. Dunng the conduct of this test simulator dynamic response, annunciator operation and automatic safety system actuations were recorded. GPUN personnel representing TMI Simulator Training have evaluated test results against
.f>ySI/ANS 3.5 criteria. The TMi representatives hold or have held NRC Senior Reactor Operator licenses for TMI Unit 1.
j' 5tified deviations f rom expected performance have been evaluated for impact and corrective action in accordance
) Functional Fidelity Procedure,6221-ADM-2820.02. The results of these evaluations are as follows:
The results of this test are satisfactory, based upon the following: CRITERI A #1 , i The observable changes in simulator parameters shall correspond in direction to those expected from the actual reference plant and do not violate the physical laws of nature. The simulator satisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused the alarm or automatic action, and conversely, the simulator shall not cause an alarrn or automatic action if the reference plant would not cause an alarm or automatic action. The simulator-satisfied this requirement with no identified deviations.
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(_,/ E5-2 GPU Nuclear, inc. i-
'Aadietown. PA l6 CCDyngnt 1997 l u ;
)
p THREE MILE ISLAND Number i i GN muets4n TRAINING & EDUCATION SIMULATOR MANAGEMENT OES-014 Revision No. O TURBINE VALVE FULL STROKE TESTING Corrective Actions / Comments: NONE Completed by: Charles E. Husted 3 Al f8 Ddte Approved by:
- Operations Training Manager Robert Parnell' M!!
Date [h L.) 9 l
.g Q
c., = ;
o n .,
- ((
[..- Chuck Husted l 05/26/98 02:5 t PM
.. i
;To:
- Robert Pamell
' cc: _. ~
.-c j .
Subject:
~ Certification Test OEs-15 TMI-1 Loss of Off-site Power
.I have conducted the above OES and am in.the process of evaluatinti the ' collected data.' There are a number of issues that will require significant research to determine
- tl.e validity of the simulator response. - As s' result of the amount of effort required I will-not be able to ' complete the evaluation and documentation of the resalts prior to the
- submittal date.1I am making this a high priority task with completion expected prior to
. June 30,-1998.
. A thorough review is being conducted of annunciator and computer alarms received at' the plant and on the simulator. This evaluation is taking a'significant amount of- l
.research. - Discrepancy reports will be written for response that is not the same as'the-rplant. l l
The procedure for conduct of the test is attached for inclusion in ite report.-
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THREE MILE ISLAND Number gpgg TRAINING & EDUCATION NUCLEAa SIMULATOR MANAGEMENT PRS-OES15 Revision No SE Applicability / Scope TMI-1 LOSS OF OFFSITE POWER 0 Responsible Oftice Ettective Date TMl TRAINING 05/21/98 TMl PLANT-REFERENCED SIMULATOR This document is within OA plan scope X l YES NO Safety Reviews Required l YES X NO List of Effective Pages j EaQR Revisiori E.agg Revision Eage Revision Ea92 Revision 1.0 0 l 2.0 0 { 3.0 0. l 4.0 ' 0 5.0 0 - 1 6.0 0 l l E1-1 0
.' \ E21 0 I E2-2 0 E3-1 0 E4-1 !
ES 1 0
, // , Sigpature Date l l a st. S rnulator Training C 2/ 7S Concurred by: '
I
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p rat o s T ining Manager b [ /!W \ f%
~.
x 10 GPU Nucisar. Inc. Modletown, F A Copyright 1997 l L _.. _ _ _ _ _ . _ . _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . _ _ . - - _ _ _ _ _ _
(ngpgg NUCLEA2
.THREE MILE ISLAND TRAINING & EDUCATION SIMULATOR MANAGEMENT Number PRS-OES15 TITLE Revision No.
0
. TMI-1 LOSS OF OFFSITE POWER
. PREPARED BY; EFFECTIVE'- .
. REVIEWED BYf REVISION :DATE- DESCRIPTION OF CHANGE ~ APPROVED BY:'
0 05/21/98 Original revision. N, i l l l' l I n N 2.0 GPU Nuclear. Inc. Mddletown PA Copright 1997
"A .
THREE MILE ISLAND Number (fpg TRAINING & EDUCATION cuct34s SIMULATOR MANAGEMENT PRS-OES15 TITLE Revision No. 0 {s~ i TMI-1 LOSS OF OFFSITE POWER v TABLE OF CONTF,NTS
- f. age 1 COVER PAGE 1.0 DOCUMENT HISTORY 2.0 TABLE OF CONTENTS 3.0 1.0 PURPOSE 4.0 2.0 APPLICABILITY / SCOPE 4.0
' 3.0 DEFINITIONS 4.0 4.0 PROCEDURE 4.0 4.1 Overview - 4.0 4.2 Simulator Initialization 4.0 .
4.3 Sequence of Events 5.0
' 4.4 ~ Data Acquisition . 5.0 4.5 Point of Termination 5.0 4.6 Certification Test Close-out 5.0 4.7 Performance Evaluation 5.0 ., RESPONSIBILITIES 5.0 \.- REFERENCES 6.0 7.0 EXHIBITS ' 7.0 .
7.1 Exhibit 1 - Simulator Performance Evaluation Form E1-1 ! 7.2 Exhibit 2 - Data Collection E2-1 7.3 Exhibit 3 - Annunciator Evaluation Form E3-1 3 74 Exhibit 4 - Plant Process Computer (PPC) Alarm Evaluation Form E4-1 7.5 Exhibit 5 - Certification Test Abstract ES-1 a 6 3.0 GPU Nuclear. Inc. 4 Wtdistown. PA l Copynght 1997
THREE MILE ISLAND Number gpgg TRAINING & EDUCATION NUCLCAN SIMULATOR MANAGEMENT PRS-OES15 TITLE Rension No. O TMI-1 LOSS OF OFFSITE POWER 1.0 PURPOSE The purpose of this test is to provide guidelines for recreation of a documented operational event and comparison of simulator responses to the associated plant performance data for the purpose of evaluating simulator functional fidelity. 2.0 APPLICABILITY / SCOPE This test applies to the recreation and evaluation of a documented plant operational event conducted at the TMI-1 Plant-Referenced Simulator. The scope of this test includes:
- Overview
= Simulator initialization
= Sequence of Events
- Data Acquisition
- Point of Termination e' Operational Test Close-out f ;;
- Performance Evaluation 3.0 DEFINITIONS l None, i
4.0 PROCEDURE 4,1 Overview
]
4.1.1 The intent of this test is to demonstrate the ability of the simulator to reproduce reference plant transient responses during recreation of a documented operational event. i 1 4.1.2 The test operator SHALL operate the simulator to reproduce as accurately as possible operating conditions immediately prior to the event, the initiator or ca.ase for the event, and documented operator control panel actions during the event. 4.1.3 - The Operations Training Manager SHALL designate individuals to perform and evaluate sivulator operational event tests. 4.2 Simulator Initialization i IC 15 - 100% Power, Equilibrium Xenon, BOC. f Ln( i t/ 4.0 GPU Nuc:sar, :nc. h4adletown PA Copyngnt 1997 _ _ .____________________.--_____a
THREE MILE ISLAND Number , (~ pgf TRAINING & EDUCATION {
" NUCLPAN SIMULATOR MANAGEMENT PRS-OES15 TITLE l Revision No 0
TMI-1 LOSS _ OF OFFSITE POWER 4.3 Seauence of Events 4.3.1 initialize as described in section 4.2 above. 4.3.2 initialize data collection in accordance with Exhibit 2, Data Collection. and insure the PPC alarm printer is on line. 4.3.3 Place the simulator in RUN and insert malfunction ED01 Electrical Blackout. Start MU-P-1 A and MU-P-1C within the first couple of minutes. 4.3.4 Continue the test until the point of termination condition desenbed in section 4.5 is reached, and then FREEZE the simulator. 4.3.5 With the simulator in FREEZE record all panel annunciator alarms ACTIVATED. Utilizing Exhibit 3, Annunciator Evaluation Form, document the cause for each annunciator or assign an El#. 4.3.6 Obtain the PPC alarm printout and utilize Exhibit 4, Plant Process Computer Alarm Evaluation Form, to document the cause for each alarm or assign an El#. 4.3.7 Obtain data tistings and/or plots as per Exhibit 2, Data Collection. 4.4 Data Acquisition [ Obtain the following simulator data:
}
^
= PPC alarm printout.
. Data collection printouts and plots.
- List of all annunciator alarms activated.
4.5 Point of Termination This operational event test SHALL be terminated when OTSG pressure has stabilized at norma! post trip pressures. 4.6 Certification Test Close-out Following completion of this test, the Test Operator SHALL forward all test documentation to the Operations Training Manager for subsequent evaluation and companson of simulator response to plant data. 4.7 Performance Evaluation Simulator response SHALL be evaluated using Exhibit 1, Simulator Performance Evaluation Form by individuals designated by the Operations Training Manager. 5.0 RESPONSIBILITIES - i i
/ ~N 5.1 The Operations Training Manager is responsible for: !
(O' ~l . 5.1.1 Designation of individuals to perforrn and evaluate simulator operational event tests. 5.0 PU Nucisar. inc.
*Addletown PA Copynght 1997,
THREE MILE ISLAND Number gW TRAINING & EDUCATION I, ea .EAs SIMULATOR MANAGEMENT PRS-OES15 TITLE Revision No. 0 TMI-1 LOSS OF OFFSITE POWER . 5.1.2 Review and approval of the following: 5.1.2.1 This Operational Event Test. 5.1.2.2 Operational Event Test Abstracts. 5.2 The Test Operator is responsible for: 5.2.1 Performance of the operational event test in accordance with the guidelines established. 5.2.2 Acquisition of the data listed in section 4.4 of this test. 5.2.3 Evaluation of all Annunciator and PPC alarms actuated during the test. 5.2.4 Evaluation of simulator dynamic responses during the test.
6.0 REFERENCES
6.1 ' Simulator Functional Fidelity Procedure,6221-ADM-2820.02. 6.2 ANSI /ANS 3.5 - 1985, Nuclear Power Plant Simulators for Use in Operator Training. fs 6.3 NUREG 1258, Evaluation Procedure for Simulation Facilities Certified Under 10 CFR 55.
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5
., ,/ 6.4 Transient Assessinent Report TAR-TMI-024.
7.0 EXHIBITS 7.1 Exhibit 1 - Simulator Performance Evaluation Form 7.2 Exhibit 2 - Data Collection 7.3 Exhibit 3 - Annunciator Evaluation Form 7.4 Exhibit 4 -Plant Process Computer (PPC) Alarm Evaluation Form 7.5 Exhibit 5 - Operational Event Test Abstract . j
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N/ l 6.0 . l l GPU Nuclew nc. ! Mddletown, PA Copynght 1997
THREE MILE ISLAND Number C)lDgg TF,ilNING & EDUCATION cruct:4s SIE LATOR MANAGEMENT PRS-OES15 TITLE - Revision No.
. O i TMI-1 LOSS OF OFFSITE POWER )
EXHIBIT 1 SIMULATOR PERFORMANCE EVALUATION FORM i
.1. Execpt as noted on the test abstract, the observable change in the parameters correspond in direction to those obtained from actual referenced plant data or from a best estimate for the simulated transient and do not violate the physical laws of nature.
- 2. Except as noted on the test abstract, the simulator did not f all to cause an alarm or automatic action if the reference plant would have caused an alarm or automatic action. Conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.
Evaluated By Concur * (Signature) Name (Y/N) Date p
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*Pl:ase attach documentation if "NO".
I i Appro'ved By Date Operations Training Manager
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'( l E1-1 l GPU Nuclear. Inc ;
Mddienmn. PA Copright 1997 L
p THREE MILE ISLAND Nu..ibw
\ CPUcuctrea TRAINING & EDUCATION SIMULATOR MANAGEMENT PRS-OES15 TlTLE Rwiston No.
0 TMI-1 LOSS OF OFFSITE POWER EXHlBIT 2 DATA COLLECTION Frequency: 5 Second Parameter Description Point ID
- 1. FEEDWATER FLOW -LOOP A FWA0560
- 2. FEEDWATER FLOW -LOOP B FWA0562
- 3. FEEDWATER TEMPERATURE-LOOP A FWA0090
- 4. EFW FLOW -LOOP A FWA0907
- 5. EFW FLOW -LOOP B FWA0908
- 6. MAIN STEAM TEMPERATURE A MSA0553
- 7. MAIN STEAM TEMPERATURE B MSA0554
- 8. OTSG PRESSURE-OTSG A MSA0556
- 9. OTSG PRESSURE-OTSG B MSA0557 e 10. REACTOR POWER - NI-5 f x NILPRS(1) 11.OTSG OPERATING LEVEL-OTSG A RCA0002 12.OTSG OPERATING LEVEL - OTSG B RCA0005 13.RCS LOOP A WIDE RANGE PRESSURE RCA0505
- 14. COLD LEG TEMPERATURE-LOOP A RCA0510
- 15. COLD LEG TEMPERATURE - LOOP B RCA0511
- 16. COLD LEG TEMPERATURE - LOOP C RCA0513
- 17. COLD LEG TEMPERATURE - LOOP D RCA0514
- 18. HOT LEG TEMPERATURE - LOOP A RCA5064
- 19. HOT LEG TEMPERATRUE - LOOP B RCA5065 20.RCS FLOW - LOOP A RCA5067 21.RCS FLOW - LOOP B F CA5068
- 22. SATURATION MARGIN-LOOP A RCTMAFSAT(1)
- 23. SATURATION MARGIN - LOOP B RCTMARSAT(2) t _
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v ! E2-1 aoU.$a$.*eI Copynght 1997
THREE MILE ISLAND Number
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'~~ TRAINING & EDUCATION cuct An SIMULATOR MANAGEMENT PRS-OES15 TITLE i Revision No. '
0 TM11 LOSS OF OFFSITE POWER EXHIBIT 3 ANNUNCIATOR EVALUATION (Typical) Page xx of xx I Annunciator Description Cause El# I J
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Completed By Date: 1 I l l j' \
/ 'l m.)
E3-1 GPU Nuclear, Inc. Mad'atowrt FA Copptgnt 1997 I
. THREE MILE ISLAND Numoer
'~ p(f TRAINING & EDUCATION suctrA2 SIMULATOR MANAGEMENT PRS-OES15 3
- TITLE Revmion No
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TMI-1 LOSS OF OFFSITE POWER EXHIBIT 4 PLANT PROCESS COMPUTER (PPC) ALARM EVALUATION FORM (Typical) Page xx of xx PPC Alarm Description Cause El# i Completed By Date: I
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E4-1 GPU Nacisar, snc. Mddle'isown. PA Copyngnt 1997 l
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THRES MILE ISLAND Number i pgp ) TRAINING & EDUCATION ' rucuA2 SIMULATOR MANAGEMENT PRS-OES15 TITLE Revision No f (, TM11 LOSS OF OFFSITE POWER { LJ \ EXHIBIT 5 j j THREE MILE ISLAND l SIMULATOR MANAGEMENT I CERTIFICATION TEST ABSTRACT (Typical) { Test identification _: ANSI /ANS 3.5 Reference (s): Test Date: Malfunction (s) Tested *- Test initialization: Point of Test Termination: Simulator Run Time: Simulator Evaluation Time: l $aseline Evaluation Data: 1 Overall Test Results: ' Results
Description:
I Corrective Actions / Comments:
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Comoleted by: j Date Approved by: Operations Training Manager Date l /~N k- ) ! E5-1, GPU Nuclear,Inc. I hAddletown. PA Copyright 1997 .i i
I THREE MILE ISLAND p (G)3(J TRAINING & EDUCATION Number
' NUCMA8 SIMULATOR MANAGEMENT OES-016 LE Revision No.
O I TURBINE VALVE TEST AT 50% POWER i I THREE MILE ISLAND SIMULATGR MANAGEMENT CERTIFICATION TEST ABSTRACT Test Identification: ' Turbine Valve Test at 50% Power Certification Test OES016 ANSI /ANS 3.5 Reference (s): N/A NUREG 1258 3.1.1.4 Test Date: 05/21/98 Malfunction (s) Tested ** None Test Initiallration: Protected initial Condition IC-16 100% Reactor Pcwer Equilibrium Xenon Middle of Cycle Point of Test Terrnination: This test was terminated following completion of the valve testing. Simulator Run Tirne: 24 minutes ulator Evaluation Time: 2.5 hours seline Evaluation Data; Right Direction Analysis. Data Trends from the Plant Process Compute E-mail from Paul Dojka Operating Procedure OP 1106-1 Revision 91 Appendix C. Overall Test Re sults: SATISFAC_TM m f GPU Nuclear, Inc. j Mddletown, P A ' Copyngnt 1997
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THREE MILE ISLAND Number j GPU JVUCLEAa TRAINING & EDUCATION
,, SIMULATOR MANAGEMENT OES-016 RHvistor1 No SE TURBINE VALVE TEST AT 50% POWER 0 i
)
l Resujts
Description:
In accordance with'NUREG 1258 this certification test was conducted to demonstrate the ability of the simulator to reproduce reference plant transient responses dunng a documented operational event.
~ This transient was a result of Main Turbine full stroke testing of Stop valve (SV) and Control valves (CV) conducted on i the Main Turbine on 04/25/98. Due to a faulty relay in the Integrated Control System. the turbine control system suddenly increased load. Operator action was initiated to mitigate the transient. The plant was stabilized and valve j tssting was completed. )
Th] transient was initiated by reducing power to Approximately 50%. The "A" side circulating water pumps were secured ; to mimic plant conditions. When the plant was stable the Turbine Valve test was conducted utilizing Appendix C,1.2.3 of R; vision 91of OP 1106-1. Testing was completed on ail turbine valves following the performance sequence of the ret:rence plant. Th3 simulator response differed in two aspects from the reference plant because: (1) the reference plant experienced a div:rgent oscillation which necessitated putting the Main Feed pumps in ICS Hand control and (2) another transient occurred which was caused by a f aulty relay. Operator actions were in response to the sudden oscillations and subsequent sudden increase in turbine load and its impact on plant stability. The simulator remained stable. There was no reason to perform extraordinary measures although they were performed to maintain conformity With the scenario. ring the conduct of this test simulator dynamic response, annunciator operation and automatic safety system ations were recorded. GPUN personnel representing TMl Simulator Training havc ; valuated test results against 1/ANS 3.5-1985 criteria. The TMl representatives hold or have held NRC Senior Reactor Operator licenses for TMI it 1. Identified devictions from expected performance have been evaluated for impact and corrective action in accordance with Functional Fidelity Procedure. 6221-ADM-2820.02. The results of these evaluations are as follows: The results of this test are satisfactory, based upon the folicwing: CRITERI A #1 The observable changes in simulator parameters shall correspond in direction to those expected from the actual re9tence piar.t and do not violate the physical laws of nature. The simulator Mtisfied this requirement with no identified deviations. CRITERIA #2 The simulator shall not fail to cause an alarm or automatic action if the reference plant wouid have caused the alarm or automatic action. and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action. I The simulator satisfied this requirement with no identified deviations.
"d' E5-2 )
GPU NL.: lear. inc.
' Mddmtown PA i Copynght 1997
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L o_ - _ - - - - -_ J
THREE MILE ISLAND Number
' ~G TRAINING & EDUCATION j - ' nuct m SIMULATOR MANAGEMENT OES-016 Revision No 0
TURBINE VALVE TEST AT 50% POWER Corrective Actions / Comments:
.NONE:
. Completed by: C :! M~ William A. Fraser S' LP i P
/ - Date Approved by: 2 Robert Parnen lli N Operations Training Manager Date -
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i l p l. 1 ES-3 GPU Nur. lear,Inc. Mdaiatown, PA l, ( 1 Coppgnt 1997 1 8 i
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