ML20083C102
| ML20083C102 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/18/1991 |
| From: | Boldt G FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20083C101 | List: |
| References | |
| NUDOCS 9109260030 | |
| Download: ML20083C102 (679) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ - _ _ _ _ _ _ FLORIDA POWER CORPORATION O Cav51At river uNii 3 INITIAL CCRTIFICATION SUBMITTAL PACKAGE TABLE Of CONTENTS Y9LUtil.1 Introduction General Information Exceptions Table 0.1 - ANS/ ANSI 3.5(1985)/ Certification Submittal Cross Reference 1.0 Simulator Information 1.1 General Information 1.2 Control Room Information 1.2.1 Control Room Physical Arrangement 1.2.2 Panels / Equipment 1.2.3 Systems 1.2.4 Simulator Control Room Environment 1.3 Instructor Interface 1.3.1 initial Conditions 1.3.2 Malfunctions O 1.3.3 1.3.4 Remote functions Instructor Station features 1.4 Operating Procedures 1.5 Changes Since Last Report figure 1.1. Simulator Control Room Layout Table 1.1 - Simulator Initial Conditions 2.0 Simulator Design Data 3.0 Simulator Tests 3.1 Overview 3.2 Test Procedure Development 3.3 Test Procedure format 3.4 Scope of Tests 3.5 Critical Parameters 3.6 Conduct of Performance Testing and Test Evaluations 3.7 Perft,rmance Test Schedule Table 3.1 Performance Test / Plant Transient Cross Reference Table 3.2 - Simulator Performance Tests Table 3.3 - Performance Test /ANS-3.5 Requirements Cross Reference Table 3,4 - Performance Test /fSAR Accident Category Cross Reference Q Table 3.5 Surveillance Tests 9109260030 910919
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FLORIDA POWER CORPORATION ' cavsta' aivta unit 3 O INITI AL CERTIFICATION SUBMITTAL PACKAGE T ABLE OF CONTCHTS 4.0 Simulator Discrepancy Resolution and Upgrade Program 4.1 Overview 4.2 Simulator Modifications 4.3 Work Packages 4.4 lesting 4.5 Documentation 4.6 1 rouble Reports Appendix A - Performance Test leam Member i Qualifications Appendix B - Simulator Control Panel liardware Discrepanc es Simulator Systems Appendix C the Appendix 0 - Plant Modifications Requiring implementation on Simulator Appendix E - Numbered Malfunctions Appendix f - Simulator Handlers Appendix G - Performance lest Data Base 1 1 rg Appendix 11 - Training Department Procedure TDP-401, Mm RQLument Lyaluation R f Plan 1/ Appendix ! - Training Department Procedure TDP-402, _2yiew inJ11111YJREB11B911RAll il J1 Lapse Appendix J - Training Department Procedure TDP-403, Sisj g EltDJittd1 Open Simulator Trouble Reports Appendix K V.RWiL11 Simulator Performance Test Abstracts 1.0 Computer Real Time Test Steady State and Normal Operations Tests 2.0 3.0 Transient Tests 4.0 Malfunction Tests
FLORIDA POWER CORPORATION cavsta' nivon unit 3 O INITIAL CERTIFICATION SUBMITTAL PACKAGE TABLE OF CONTENTS Y0WILL111 Simulator Performance Test Package PIM9C/01SG Overfill O O
FLORIDA POWER CORPORATION O custat aivonu"i' 3 sinut^'on CERTlflCA110N SUBMITTAL INTRODUCTION GENERAL INFORMATIQB 1he information contained in this report is p.ovided to demonstrate compliance of the Crystal River Unit 3 Simulator with the Certification requirements contained in ANSI /ANS-3.5(1985) and Regulatory Guide 1.149, Revision 1 April 1987. , The report is organized into three volumes. Volume I contains the Certification Submittal for the Crystal River Unit 3 Simulator. This information is organized and presented in accordance with the format specified in ANSl/ANS-3.5(1985) Appendix A. " Guide for Documenting Simulator Performance" - Volume 11 contains the abstracts for the performance test s conducted to demonstrate compliance with the referenced standards. Volume lit contains one complete test package which is typical of each of the completed tests in terms of level of detail, acceptance criteria, data collection and simulator response. The Simulator Certification was performed in accordance with the guidance contained in Training Department Procedures TDP-405, Simulator Per_formance O, Test Ina and TDp-406, NRC Simu.lder_.CertififJLt1Rnfrocess. Preparation of the performance tests, conduct of the tests, documentation of the results and preparation of the certification submittal were performed over a nine month period from December,1990 to August,1991. These activities were performed entirely by permanent Florida Power Corporation personnel within the Nuclear Licensed Operator Training Group; the reviews and evaluattors of the test results were conducted by a panel of experts selected from both the Nuclear Licensed Operator Training Group and the Nuclear Operations Drpartment. The qualifications for each person involved in the conduct and/or review of the performance tests are contained in Appendix A, Table 0.1 contains a cross reference of the ANS/ ANSI 3.5('985) simulator certification requirements and the section(s) of this submittal which address each of those requirements. i
__ _ _ ~ . _ . _ _ _ _ . . _ . _ _ _ _ _ . _ _ . _ - - - _ . _ _ FLORIDA POWlR CORPORA 110N O CRYST AL RIVER UN113 SIMULA10R C[RilflCA110N SUBM111AL , EXCEPTIONS The Crystal RivJr Unit 3 Simulator meets all of the applicable requirements 1 of ANS/ ANSI-3.S(1985) with the exception of those deviations identified in , the applicable sections of the sebmittal. Each of these deviations has been-evaluated by FPC; for those deviations censidered to be unavoidable due to such factors as the simulator control room physical layout or overriding training requirements / considerations, or which were determined to have no training impact, no further action will be taken. All other identified deviations are being tracked and corrected in accordance with section 4.0 of this submittal. None of the deviations are considered to impact simulator fidelity or training / testing effectiveness to a sufficient degree to warrant taking an exception to any requirement of the standard. In addition to these deviations, the following non applicable testing requirements are taken as exceptions to the standard.
.o Section 3.1.2, " Plant Malfunct ions", item (12) Control rod f ailure including stuck rods, unconW ed rods, drifting rods, rod drops and ,
misal'7ned rods. The design of the CR 3. control rod drive system precludes drif ting _ O- rods. ror tse CR-3 ioox eerformence iest tist, this 4 tem has been replaced by degraded rod motion. o Section 3.1.2, " Plant Malfunctions", item (25) Reactor pressure control system failure including turbine bypass failure (BWR). This item is only required for Boiling Water Reactors o_ Section 4.1,-" Steady State Operation", item (2) Principal mass and energy balances shall be satisfied: (b) Reactor Coolant System ' temperature to Steam Generator Pressure and (e) Mass balance of steam generator. < These items are not applicable to Babcock and Wilcox Plants with Once-Through Steam Generators (OlSGs). o Section 4.1, " Steady State Operation", item (3)(f) Recirculation flow. This item is only applicable to Boiling Water Reactors, o __ Appendix B, Section B2, "PWR Simulator Operability Test" Several subsections of the section require recording pressurizer pressure as a critical parameter. Since Babcock and Wilcox Plants do not monitor this parameter, narrow range RCS pressure has been l: l Q substituted. l
l FLORIDA POWER CORPORAllbN O J CRYS 1AL RIVER UN113 SlhUIA10R CERTlflCA110N SUBMIT 1 AL TABLE 02 1 MS/A8il _h.hDM 5)lCIRLifl0 A110fL5DM111AL GR015_BLfElmCI MS/A85J-L5 SECTIM/.REQMMMMI AP111CelllI_5MHilIAL51Gil0HLSJ 3.1.1 Normal Plant Evolutions Volume 1 Sections 3.1, 3.4 Volume 11, Section 2.0 3.1.2 - P1&nt Malfunctions Volume 1, Sections 1.3.2. 3.1, 3.4 Volume 11, Section 3.0, 4.0 3.2.1 - Degree or Fanel Simulation Volume 1, Sections 1.2.1, 1.2.2 3.2.2 - Controls on Panels Volume I, Section 1.2.2 3.2.3 - Control Room Environment Volume 1, Section 1.2.4 3.3.1 - Systems Controlled from the Volume 1, Section 1.2.3 ( Control Room 3.3.2 - Systems Operation or functions Volume 1, Section 1.3.3 controlled Outside of the Control Room 3.4.1 - Initial Conditions Volume 1 Section 1.3.1 3.4.2 - Malfunctions Volume 1, Section 1.3.2, 1.3.4.3 3.4.3 - Other Control features Volume 1, Section 1.3.4 3.4.4 - Instructor Interface Volume 1, Section 1.3.3 4.1 Steady State Operation Volume 1, Sections 3.1, 3.4 Volume 11, Section 2.0 4.2 Transient Operation Volume 1, Sections 3.1, 3.4 Volume 11, Sections 3.0, 4.0 4.3 Simulator Operating Limits Volume 1, Section 1.3.4.17 4.4 - Monitoring Capability Volume 1. Sections 1.3.4.11, 1.3.4.12 O S.1 - Simulator Design Data Volume I, Section 2.0
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FLORIDA POWER CORPORA 110N l O cavs'^' aivra UNIT 3 SIMULATOR CERTlflCATION SUBM111AL IABLLh1 AMS/AN119.i.5(1HERCIRIlFlfATION SilEMIIAL CA01S_REfIRfELE(continued) i ANS/ARSIALSE011M/IEMIRMIM APlLICABLE_MBRIIIALSIC110M(M 52 - Simulator Upgrade Design Data Volume 1 Sections 2.0, 4.0 5.3 - Simulator Modifications Volume 1, Section 4.0
-5.4.1 - Simulai r Performance Testing Volume 1, Sections 3.0, 4.0 Volume 11. Sections 2.0, 3.0, >
4.0 5.4.2 Simulator Operability Testing Volume 1 Section 3.0 Volume 11, Sections 1.0, 2.0, 3.0, 4.0 Al.1 - General Information Volume 1, Section 1.1 O Ai.2 - Centroi Room informatioa voiume i. Section i.2 A1.3 - Instructor Interface Volume 1. Section 1.3 A1.4- - Operating Procedures for Volume I, Section 1.4 Reference Plant A1.5 - Identification of Changes Volume 1, Section 1.5 Since Last Report A2 - Simulator Design Data Volume 1, Sectior. 2.0 A3.1 - Computer Real Time Test Volume 1, Sections 3.1, 3.4 Volume 11 Section 1.0 A3.2 - Steady State and Normal Volume 1 Section 3.1, 3.4 Operations _ Tests Volume 11, Section 2.0 A3.3 - 1ransient Tests Voluma 1, Sections 3.1, 3.4 Volume 11, Section 3.0 A3.4 - Malfunction Tests Volume I, Sections 3.1, 3.4 Volume 11, Section 4.0 pJ - A4 - Simulator Discrepancy Resolution Volume 1 Section 4.0 l and Upgrading Program I
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FLORIDA POWER CORPORAliDN FRYSIAL R1?ER U11113 S!MUL A10i1 dc CERTIFICATION SUBHillAL IABLLLI MS/AN S I - 3 . 5 {120)fCEAULLCAI1M,_5MSM1UAL CDJiLKEEEEEE(continued) MS/MU:L1_1E0110EiHEMIREffEI MELLCARL1@ELUALEG119MI51 B2.1 - Steady State Performance Volume 1, Sections 3.1, 3,4 Volume 11, Section 2.0 P,2.2 - Transient Performance Volume 1, Sections 3.1 3.4 7 Volume 11, Section 3.0 O O
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t 1 TLORIDA POWER CORPORA 110N , 4 O cavstat aivta u"i'a sinu'^'oa C[R11 FICA 110N SUBK 1TAL 1.0 $ltiUtAID3_lEQRMATIDE 1.1 GENERAL INFORMATION Owner: florida Power Corporation (fDC), Crystal River Unit 3 Plant (CR 3) Operator: The maintenance and operation of the CR-3 Simulator is the responsibility of the ; Licensed Operator Training Group of the FPC . Nuclear Operations Training E,cpartment. i Manufacturer: CAE Electronics Ltd., of Montreal, Canada, t Reference Plant / Type / Rating: CR 3 is a B&W 177 Pressurized Water Reactor licensed for 2544 MWt. The plant began commercial operation in 1977. The NSSS was ' supplied by Babcock and Wilcox, the turbine was supplied by Westinghouse, and the A/E was Gilbert Associates. O Date avanabie for Training: Simulator training began May 28, 1990. FPC control of the simulator configuration began April 1, 1991. Type of Repert: Initial Certification Report. I t-
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FLORIDA POWER CORPORA 110N O CRvSut RmR uNin SiMuu,0R CERilfICATION SuBM111AL 1.2 CONTROL ROOM INFORMATION LL1 Cont r01. . RaptEttylini Arrarigment The physical arrangement of the simulator control room is depicted in figure 1.1. All cabinets and equipment in the reference plant control room which are normally accessed by the operators t.re replicated in the simulator in terms of size, physical appearance and location with the following deviations:
- 1. Due to the location of the Instructor's Booth, the locations of both the fire Service Panel and to : Reactor Coolant Pump Vibration Monitoring Cabinet in the simulator are different than in the control room. Both of these cabinets have limited training applications.
- 2. Due to the nrientation of the instructor's Booth, access to the simulator control room is not in the same location as the entrance to the plant control room. >
- 3. Cabinets in the plant control room which are not accessed by the operators, and therefore have no training Oi applications, are not included in the scope of simulation. These cabinets include the Annut.ciator Cabinets, the ICS Cabinets, the NNI Cabinets and the Plant Process Computer Cabinets.
- 4. The support column in the plant control room is replicated in the simulator in terms of physical location and length and width dimensions; the column in the simulator, however, does not extend to the ceiling in order to allow personnel in the Instructor's Booth an unobstructed view of the main control boards. >
- 5. The simulator control room contains three microphones and two cameras not included h the actual control room.
These cameras were installed to videotape NRC Requalification Examinations and simulator training sessions for use in operating crew self-critiques. The training benefits gained from the use of these cameras is considered to outweigh the fact that they represent a deviation fren the reference plant physical arrangement.
- 6. The simulator control room contains a storage cabinet next to the Reactor Coolant Pu.np Vibration Monitoring Cabinet which is not contained in the reference plant control room. The cabinet is used to fill in the area under the visitors' gallery not taken up by the Vibration O
f LORIDA POWER CORP 0RA110N O CRvstat RivcR uni 13 SiMutAiOR CERTiflCA110N SUBMITTAL 1.2.1 Contiql l apm Phy1ical Arranaement(Continued) Monitoring Cabinet and fire Service Panel and can be removed to allow access to the back of this equipment for maintenance. ,
- 7. There are cabinets and lockers behind the Main Control Boards in the plant control room for storage of chart paper / expendable supplies and use by operating shif t personnel. These furnishings are not included in the simulator control room.
- 8. There is a file cabinet in the simulator for storage of 208-series drawings for use by the operators. Storage of these drawings in the plant is outside the control room in-a support office.
- 9. The simulator control room includes noth a chalk board and a Mobile Instructor Station which are not in the plant control room. The training uses of these items is considered to outweigh this minor deviation.
1.2.2 110 tis /Eauipment The simulator control panels included in the scope nf simulation were originally fabricated to replicate the reference plant control room as of the design data freeze date of June 40, 1987. During construction of the simulator, it was decided to upgrade the panels to include major panel modifications performed during the plant's sixth refueling outt.ge which occurred after the design data freeze date. Remaining Reft.el VI and post-Refuel VI modifications which affect the panels and have not yet been implemented on the simulator are listed in Appendix D. These modifications will be , incorporated within the required time frames identified in the Standard. Within the boundaries of the current scope of simulation, the controls on the panels duplicate the size, shape color and configuration of the functionally simulated hardware of the reference plant; information displayed to the operators is in the same form and units as the reference plant; and components and displays that function during normal, abnormal and emergency evolutions are included in the scope of simulation. Since installation of the simulator, a detailed comparison of tte simulator and reference plant panels has been conducted O using the following methodology:
FLORIDA POWER CORPORATION O CRYSTAL RIVER UHli 3 SIMULATOR CERTiflCATION SUBMITTAL 1.2.2 M ngh Kgy.ipm3Di(Continued) 1-. A complete set of photographs of the main control boards, i Reactor Protection System cabinets, ES Actuation Cabinets and ES relay cabinets was taken in December of 1990.
- 2. A detailed comparison of these photographs with the ;~
equivalent simulator panels / cabinets was performed, and all differences documented.
- 3. Each noted difference was tracked to its source, i.e.,
plant modification not yet implemented on the simulator. ' vendor construction deviation, etc. For differences which were tracked to plant modifications, corrections will occur as the modifications are implemented on the i simulator. for differences based on other sources, the , difference was evaluated for training impact. For minor ' cosmetic differences such as those discussed in section 3.2 of the standard, no further action was taken; for all other differences, Trouble Reports were written to initiate corrective action.
.O A compiete iist of panei differences requiring corrective action at the time of this submittal is contained in Appendix ;
B.
?n addition to Appendix B, the following items are included in thd submittal as deviations to the requirements of the 1 standard:
- 1. With the exception of the lockout relays, the electrical protective relays on the Turbine Generator Relay (TGR) panel are simulated through a combination of front face
. visual simulation and functional drop targets / reset ,
levers.
- 2. ' Simulation of the Loose Parts Monitoring Panel is limited L to the Reactor Coolant Pump Vibration Monitoring t i equipment. Since the remainder of the equipment / controls-l cssociated this cabinet are not used by the operators, this limitation hat no training impact.
- 3. The control room contains a closed circui. telcvision system for monitoring leakage on the main stean-relief ,
L valves and atmospheric dump valves. At this time, leakage or opening of these valves is-indicated in the simulator control room by backlighting designated annunciator windows. A simulator modification is currently in
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FLORIDA POWER CORPORA 110!4 "O tavsi^t aivcaunii 3 sinut^1on CER11 FICA 110f4 SUBM11TAE 1.2.2 fatlellfiggippant(Continued) progress which will replicate the control room display of these valves through a PC-driven graphics package and monitor.
- 4. The Vital Bus Distribution panels located in the main control room are not included in the scope of simulation due to limited training applications; however, the Vital Bus Circuit Monitoring Panels are fully simulated. in addition, each of the individual breakers in the Vital Bus Distribution Fanels, and the loads associated with these breakers are accessible as remote functions in the Instructor facility.
- 5. DC Distribution panels DPDP-7A and .78, located in the main control room are not included in the scope of simulation due to limited training applications; however, the breakers for each of the individual loads on these panels which are addressed in plant procedures are accessible ~ as remote functions in the Instructor facility.
- 6. He ES . indicator Power Transfer Panels located in the main control room are not included in the scope of ,
simulation due to limited training applications; however, the transfer functions associated with these panels are ! accessible as remote functions in the instructor facility.-
- 7. The dedicated Plant Process Computer (PPC) printer in the simulator _is not the same model as -that in the plant control room; however, the differences between the two are minor and have no training impact.
- 8. The RECALL system for the STA workstation located in _the control room is not included in the simulator. Current plans call for incorporation of these features in the '
simulator after the planned upgrade to the PPC and RECALL systems in the plant.
- 9. The simulator does not, at this time, include the Emergency Dose Assessment System (EDAS). Since the -
o)erators do not currently interface with this system
-tais difference is considered to have no training impact.
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l 1 I ft0RIDA p0WfR CORPORA 110N ; O cavstat aivtRUN113 SIMULATOR CERllflCA110N SUBM111AL 1.2.3 S.ytt.cm1 The CR 3 simulator dynamically models all the reference plant systems required to train operations personnel over the full range of normal evolutions, operator conducted surveillance tests, abnormal and emergency conditions. A complete listing of the simulator systems is included as Appendix C to this submittal, i The scope of simulation was originally set by the simulator i procurement specificatioh and was further refined during the construction of the simulator through Design Review Meetings, pre Acceptance lesting and f actory Acceptance lesting. The scope of each system is described in its associated model report. The-simulator sof tware was originally designed and written to reflect the reference plant configuration as of the design data freeze date. During the construction of the simulator, it was decided to incorporato the significant plant modifications incorporated during the CR-3 sixth refueling outage. Plant n modifications since design data freeze have been tracked and V evaluated in accordance with the CR 3 Simulator Configuration Control process described in Section 4 of this submittal. Florida Power Corporation assumed configuration control of the simulator on April 1,1991. Since that time the simulator maintenance and engineering staff has been actively implementing in service plant modifications on the simulator. A list of'in-service plant modifications which have not been incorporated on the simulator is included as Appendix D. These , modifications, which are being tracked by the Configuration - Management System, will be incorporated within the required time frames specified in the Standard. , 1.2.4 11mylator ControlR091!LECLYitonment in addition to those items discussed in sections 1.2.1, 1.2.2 and 1.2.3 above, the following features have been incorporated in- the simulator to replicate, to as close a degree as possible, the physical environment of the reference plant L control room: l I O-
FLORIDA POWER CORPORA 110N cavsi^t aivcauwit a sinut^1oa
-O C[RTlflCA110N SUBMITTAL 1.2.4 Simulator CQattolloom EnvironLnent(Continued)
- 1. Control Room -Lighting -
the lighting in simulator I replicates both the normal and emergency lighting in the , reference plant control room in terms of intensity, ! physical arrangement, and sower dependency. A loss of i power to normal lighting wl'1 result in actuation of the emergency lighting system.
- 2. Sound Systems - the simulator employs a digital sound system which replicates the following sounds heard by the ,
operators in the reference plant control room: ! ventilation system background noises, the site i evacuation, auxiliary building evacuation, reactor building evacuation and fire alarms, control room audible annunciator alarms, the lifting of main steam safety valves, and steam line breaks outside the containment.
- 3. furnishings - replications of the following reference plant control room furniture are included in the _j simulator: shift supervisors desk, assistant shift supervisors desk, nuclear- operators dest, _and communications table.1he type and arrangement of ceiling tiles, type and color of carpeting, and paint colors in !
the simulator' control room match those in the reference plant control room. -
- 4. Communications - The simulator control room contains all the communications systems used by the operators during both normal plant operation and emergency _ conditions. '
These systems include: ,
- a. Hand held radios used for normal communications between the control room and in plant operations personnel,
- b. The Commercial- Telephone System used for communications between the control room and personnel in the 1&C, Electrical and Mechanical ;
Health Physics, Chemistry, Energy Control-Shops Center, CR 1&2, CR 4&5, etc.
- c. The Gaitronics intra-plant Communication System (PAX, Page and Party Lines), i
- d. The Emergency intra plant Communication System.
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ILO?lDA l'0W[R CORPOPATION O cavStat nivratirii13 Sii4ui Aloa CER11F1 cal 10ti SUBlillTAl. 1.2.4 11wl1tqr _G9ntto_LRgemJnykonmeni (c ont i nue d )
- e. Dedicated phones ur.ed for euergency communications with the control room and direct communications with the t1RC Emergency llotification System, flational Warning Sy st em ('1 AWAS) , State Hot Ringdewn, and Babcock & Wilcox.
All calls initiated from the control room ring in the Instructor's Booth at the phone console or the dedicated State Hot Ringdown o'* f1RC phones. Calls initiated in the instructor's Booth ring at the appropriate station / phone in the simulator control room. When a number is dialed, a message appears on the phone console in the Booth, informing the instructor of the specific location being called. Background noises on the Party Line phone system can be inserted at the instructor's discretion to emulate the sounds of the turbine, emergency diesel generators, condensate pumps, inverters and control complex ventilation equipment. O O l
ft0RIDA POWER CORP 0RA110f4 O CRYST AL RIVE R Utill 3 SIMUL A10R CERTIfICAT10f4 SUBM111AL 1.3 INSTRUCTOR INTERFACE 1.3.1 InLtigLgrdiligni The simulator is capable of storing fifty-cne (51) initial conditions. At the time of this submittal ICs 1 10 are under configuration control and are maintained as permanent storepoints for use in training. These points include a variety of pl ant operating conditions, fission product poison concentrations and times in core life. ICs 11 -20 are reserved for additional permanent training storepoints are they are identified and developed. ICs 31 - 50 are temporary points created for specific short-term training applications, sof tware or training material development or for use in troubleshooting performance problems. IC 51 is the designated " Snapshot" point. The permanent storepoints currently being used for training are listed in Table 1.1 The permanent training storepoints were all originally created from a full power -teady state, beginning-of-life condition, from that point, the plant was maneuverert to the required conditions using plant operating procedures. Middle- and end-O of-core life conditions were created using utilities to modify appropriate core variables to produce the desired operating conditions. 1.3.2 tial functLor.1 The CR-3 simulator is capable of performing all the required categories of malfunctions specified in Section 3.1.2 and Appendix B, Section B2.2 of the Standard with the exception of the non-applicable items discussed in the introduction, in addition, the simulator is capable of supporting the types of accidents specified in the FSAR which are deemed appropriate for training. For those malfunctions where operator action is a function of degree of severity, the simulator has an adjustable range to reflect the malfunction conditions. The introduction of malfunctions on the simulator does not alert the operator in any manner other than that which would occur in the reference plant. A detailed discussion of the ways in which malfunctions can be inserted cn the simulator is contained in Section 1.3.4.3 of this submittal. The malfunction capabilities of the simulator can be broken down into three (3) categories: O l _ ._
FLORIDA POWER CORPORATION O cav5'^' aivc a u"i' 3 Si"ut AiOR CER!iflCATION SUBMillAl 1.3.2 lialfunctioni(Continued)
- 1. Numbered malfunctions - the simulator contains a total of 197 numbered malfunctions. The selection of these specific malfunctions, which defined the scope of the malfunction and transient test portions of the simulator Acceptance Test program, was performed by FPC personnel based on the requirements discussed in Section 3.1 of this submittal. A representative subset of these malfunctions forms the basis of the 100% Certification Test list. The complete list of numbered malfunctions is included as Appendix C of this submittal.
- 2. Panel Overrides - this feature of the simulator allows the simulator instructor to fall any switch, meter, recorder. potentiometer, light, or alarm located on the simulator control boards or cabinets to any desired condition or position.
- 3. Component f ailures - rather than develop unique code for every modeled component in the scope of simulation, the CR-3 simulator utilizes an in line " handler approach to
( model generic types of components. The CR 3 simulator uses 98 handlers to model components such as pumps, motors, transmitters, relays, pressure switches, valves, tanks, heat exchangers, etc. The majority of these handlers include one or more f ailures for the component being addressed, e.g, each type of breaker handler will include fail open, fail close, fail as-is capabilities; each type of valve handler will include fail open, fail closed, fail as is and seat leakage capabilities, etc. A complete list of the CR 3 handlers and the failures associated with each is included as Appendix f. In addition to the capabilities discussed above, many of the simulator remote functions, described in Section 1.3.3 below, can be used to initiate malfunctions, e.g., closure of a manual suction isolation valve for a pump can be used to cause purrp cavitation and an eventual trip of the pun,p. O
e FLORIDA p0WER CORP 0 RAT 10ll CRYSTAL RIVER UNIT 3 SIMULATOR CER11 FICA 110N SUliMITTAL 1.3.3 Remote functions The CR 3 simulator utilizes remote functions to allow the instructor to simulate those actions required to be performed by plant personnel outside the control room during normal evolutions, surveillance testing, and abnormal and emergency conditions. The specific remote functions required by the CR-3 simulator were initially defined when the scope of simulation was set. Additional capabilities were added as a result of the design review procen, pre Acceptance and f actory Acceptance Testing. Each remote function required during the conduct of the full range of simulator performance testing was verified as , part of the performance test procedure. See Section 1.4 for further explanation of the verification process. A number of the available remote functions are implemented through handlers, discussed in section 1.3.2 above. Appendix f includes the remote functions associated with each handler, A list of the remote functions, generic malfunctions and ^ component failures (handler malfunctions) for each system is ( t contained in each system's associated model report. Due to the extremely large number of remote functions and malfunctions available on the CR-3 simulator (>l8,000 discrete items), a full listing of these features is not included with this submittal. Since a significant numbei of these malf unctions have variable ranges, a complete listing of variable < malfunctions and their ranges is also not included. A full list of all remote functions and malfunctions can be generated if required. For each variable malfunction included in the scope of performance testing, a range is specified in the associated performance test abstract. 1.3.a intrElor Station Entym 1.3.4.1 Instructor Facility Layout The Instructor facility (1/f) provides the communication link between the instructor and the simulator. By using the training features of the 1/f tne instructor can insert and remove failures,. monitor simulated plant parameters, monitor operator performance and play back training sessions. The 1/f consists of two fixed workstations, one mobile workstation and a portable workstation. The fixed stations consist of two high resolution color graphic CRTs and their
-associated touch sensitive screens, special function keypads O and alphanumeric keyboards. The mobile workstation consists of a high resolution color graphic CRT and its associated touch
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ILORIDA POWER CORPORAT10!1 h CRYST AL RIVERUNIT 3 SIMULATOR CER11FICAT10!4 SUBM111AL
-1,3.4.2 Organization of CRT Pages sensitive screen, special function keypad and alphanumeric keyboard. A color printer can be accessed f rom either of these workstations. The portabic workstation coresists of the TELXON
-P10 71 hand held controller.
The CRis display pages of information which f all into one of the following categories:
- 1. System Schematics - System schematics are graphic representations of plant systems containing monitored parameters and device statuses. From the schematics the instructor can access failures, bar charts and graphic trends for any active device displayed on that schematic.
- 2. Matrix Pages - These pages allow quick access to other pages or schematics = The overall system matrix allows the instructor to access any system matrix. The system matrices allow the instructor to call up any of the system related schematics.
O 3. Centroi ca9es Coatroi na9es are those na9es which aliow access to 1/F features such as initial Condition Select, Backtrack, Playback, Parameter Control, etc.
- 4. Stylized Pages -
Stylized pages are graphic representations of the simulated panel and cabinets. From these pages the hstructor can implement panel overrides or monitor the status of any panel device. 1.3.4.3 Failure Insertion and Removal failures and remote functions can be inserted in six ways:
- 1. Immediato insertion
- 2. Keyboard ramp This feature allcws the ramping of any analog . failure over time using the UP and D0hN function keys on the keypad.
- 3. Software Ramp This ramp feature allows for ramping of up to 30 failure variables. This is accomplished by selecting the desired cialfunction and entering a final value followed by the Q- ramp time.
l FLORIDA POWER CORPORATI0ff I O cavs'^'aivonu'iitasinu'ATon CERTIFICAT10ft SUCMITTAL 1 l i 1.3.4.3 Failure Insertion and Removal (Continued)
- 4. Time Delay Time Delay allows for selected failures to enter the system after a certain time period has elapsed. This is-accomplished by selecting the desired malfunction and entering a destination value followed by the delay time, i
- 5. Arming The instructor can save up to 30 failures on a system stack so that they can be entered at a later time.
Activation of the failures in the stack is accomplished by pressing the ACT STAK key on the keypad.
- 6. Conditional The ins *f ator can insert a failure based on a predefined condition being met. The instructor types in the condition followed by the destination value that the failure will take on when the condition is met.
Failures can be removed from the system by either entering the , non failure state of the malfunction or by deleting the failure on the Active Malfunction Summary page. 1.3.4.4 Panel Overrides Panel Overrides are instructor controllable failures of the , panel hardware. Using this feature, the control panels can be made to indicate instructor selected values / conditions, in place of those calculated by the modelling sof tware. Up to 30 overrides can bn inserted into the system et any one time. 1.3.4.5- Alarm Overrides-Alarm Overrides are instructor controllable overrides of the clarm windows and the alarm messages appearing on the annunciator CRTs. Using this feature, the instructor can override an alarm /or alarm point to one of three states: . permanent on, permanent off, or cyclic. 1.3.4.6 IC Select and Switch Check The-1/F allows the instructor to call up or store up to 50 - Initial Conditions and a Snapshot point. The 50 ICs are password protected to prevent inadvertent overwriting; the Q snapshot point is used to quickly store an existing simulatoi m
l FLORlDA POWER CORPORA 110N l O cavst^t aivra u"i' 3 si"ut^'on CER11 FICA 110N SUBM111AL i 1.3.4.6 1C Select and Switch Check (Continued) condition for use in training or sof tware troubleshooting, and is not password protected. When an 1C is selected and loaded, the panels go inte switch check mode to identify any control board components that are out of position for the selected 10, 1.3.4.7 Backtrack The Backtrack feature automatically stores the status of the ! simulator at an instructor selectable interval. The simulator can be reset to any of 60 available backtrack points. Using this feature portions of a lesson or exercise can be repeated by resetting to the desired point. 1.3.4.8 Record RecordP ,, is a feature that generates a report of all actions performed by the trainee and instructor during an exercise. The format of the report is a time-stamped message indicating the O switches turned, the controllers moved and the failures / instructor controlled functions inserted. 1.3.4.9 Playback Using the Playback feature, a period of the lesson in progress can be_ replayed, allowing the students to observe all the actions performed during the exercise as they are automatically performed. The sin,ulated plant response will be the same as if the inputs were being made from the panels. 1.3.4.10 Speedup And Slowdown This feature allows portions of the simulation to be run in other than real time. The entire simulator can be run slower than real time; the following portions of the simulation can be run faster than real time:
- 1. Xenon and Samarium Buildup and Decay 2, Decay Heat 3.- RCS Heatup/Cooldown
- 4. Dilution /Boration O- 5. Turbine Metal lemperatures
FLORIDA p0WER CORP 0RA110fi O cars'^' aivi a uriii 3 Sinut A30R CERilflCATI0ld SUBMITTAL 1.3.4.10 SpeedupAndSlowdown(Continued) , d
- 6. Pressurizer Heaters
- 7. Condenser Vacuum 1.3.4.11 Graphic Recorder The Graphic Recorder is a plotting package that provides the instructor with the ability to store data and review it in graphic form at the end (or during) a lesson. The instructor can create up to twenty menus of four parameters each and have these parameters recorded during a lessor At any point during the lesson, a menu can be plotted .hout disturbing the recording of its data.
1.3.4.12 Pen Plotter The pen plotter feature allows the astructor to record up to eight parameters on a Could eight pen plotter. 1.3.4.13 Lesson Plan A lesson plan is a sequence of failures activated by the instructor in any order and at any time during a training exercise. Lesson plans enable an instructor to casily execute the same sequence of events many times in an exactly repeatable manner. Each step of a lesson plan can be manually triggered by the instructor or automatically initiated using a " delta time" for that particular step. 1he timer counts down until it reaches zero. At that point. the step is executed. Steps can be executed in any order desired, except those with a delta time. 1.3.4.14 Honitored Parameters This feature allows the instructor to monitor up to 20 specified parameters at a time during a training session. 1.3.4.15 Numbered Halfunctions This feature allows direct insertion of any one of the designated numbered malfunctions without first calling its associated system page. O
i l i FLORIDA POWER CORPORA 110l4 h CRYSTAL RIVER Uf4113 SIMUL ATOR CER11flCAT10f4 FVBMITTAL 1.3.4.16 Parameter Control The parameter controller allows for setting up of a series of malfunctions for use in a simulator scenario. Using this i feature, malfunctions can be entered using combinations of the malfunction insertion modes discussed in section 1.3.4.3 above , e.g., the time delayed conditional insertion of a malfunction l which ramps to a final value, in addition, the malfunction set i can be saved to a file and recalled at any time. 1.3.4.17 Out-of-Scope Conditions The simulator continuously monitors a specified set of > parameters against predetermined limits which reflect the simulation boundaries for the software. If any of these parameters reaches its limit, the simulator reverts to IREEZE, an OUT LlHIT SIM light illuminates oh the 1/f and an audible alarm annunciates in the Instructor's Booth, lhe instructor can then call up a CRT page which will list the out of-scope parameter and its value at the time the simulation was frozen. The instructor has the option at this time to continue the simulation if desired. The following parameters and their associated limits are used to determine an out-of scope condition:
- 1. RCS Pressure 2800 psig
- 2. fuel Temperature 4500 F
- 3. Cladding Temperature 2300 F
- 4. Containment Pressure 58 psig
- 5. Containment Temperature 500 F
- 6. flormalized Reactor Power 125't. (Rated Thermal Power)
- 7. Turbine Speed 2250 rpm l
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l l l i FLORIDA POWER CORPORATION I-) CRYST AL R1VER VH1T 3 SIMULATOR CERTIFICATION SUBMITTAL 1.4 OPERATING PROCEDURES All simulator performance testing and training is conducted using current plant procedures. A controlled set of all surveillance, normal, alarm response, verification, abnormal and emergency proceJures is maintained in the simulator control room. During the conduct of the performance tests, each step of each plant procedure used was evaluated for applicability on the simulator. For all referenced procedures, all steps were found to fall into one of the following categories:
- 1. Steps which could be performed in the control room exactly as written - for thet.e steps, no further action was required.
- 2. Administrative steps - steps calling for notification of personnel, chemistry samples, etc. were marked N/A for the performance tests. During training sessions, these steps are performed with appropriate feedback being provided by the simulator instructor.
- 3. Steps which could be performed exactly as written using remote functions - for each of these steps, a Remote function Verification was performed the first time the remote function was encountered. The verificat'.on ensured that the action could, in fact be performd, and that the results of tae action were as expected.
- 4. Steps which could not be performed, or could not be performed exactly as written - for each of these steps, a simulator Scope Limitation Evaluation was performed the first time the limitation was encountered. Each Scopa Limitation received one of the following dispositions:
- a. If the inability to perform the step was totally transparent to the operators in terms of control room indications and/or simulator response, no further action was taken,
- b. If the desired results of the step could be accomplished through some other mechanism, such as the operation of a valve in series with the requested valve, and the component substitution was transparent to the operators in terms of control room indication and simulator response, no further action was taken.
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FLORIDA POWER CORPORATION O CRvS1At RivEa UNIT 3 SIMULATOR CERTlflCATION SUBMITTAL 1.4 OPERATINGPROCEDURES(Continued)
- c. If the desired results could not be accomplished, and this fact was visible to the operators in terms of control room indication and/or simulator response, a Trouble Report was initiated to incorporate the required component in the scope of simulation.
Appendix K includes all open ' rouble Reports generated as a result of this process. The simulator is currently one refueling cycle behind the reference plant. Due to design differences between the actual and simulated core loads, ECP calculations using current plant procedures / curves will not yield the correct results Until a simulator core update is performed, this problem is being addressed by adjusting RCS boron concentration (transparent to the operators) to achieve the ECP calculated using current plant procedures. O v-
FLORIDA POWER CORPORAT10f4 cavst^' aivca urii> S SinulA,0R O CERTiflCA110N SUBM11TAL 1.5 CHANGES LAST REPORT Since this is the initial CR-3 Simulator Certification Submittal, there are no changes to identify since the last report. M O O
ELERIDA POWER CORPORAllpN CfLYSIAL RIVER UNIT 3 SIMULATOR g LLGWD PANEL NUP,8fR DESCRIPTION Al Engineered Safeguards Panel A2 Primary Systems Auxiliary Panel A3 Integrated Control System Panel A4 Turbine Generator Panel AS Station Service Panel (Electrical , Distribution) A6 thru A8 Heating, Ventilation and Air Conditioning Panels A10 thru A12 Station Service Relaying, Turbine Generator Relaying and Turbine / Generator Supervisory instrumentation Panels A21 thru A28 Reactor Protection System Cabinets A29 thru A30 Engineered Safeguards Actuation Relay Cabinets A31 thru A33 Engineered Safeguards Actuation Channel Cabinets A34 Radiation Monitoring Panel A35 PPC Alarm Printer A36 A37 PPC Line Printer Reactor Coolant Pump Vibration Monitoring h Panel A38 Fire Service Panel A40 thru A43 Vital Bus Circuit Monitoring Panels l 9
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O O O FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICATION SUBMITTAL TABLE 1.1 SIMULATOR INITIAL CONDITIONS CORE RCS RCS POWER IC# AGE TEMP PRESS LEVEL XENON BORON DESCRIPTION 1 MOL 579 2155# 100% EQUll 957 Full power, steady state 2 MOL 537 2155# 0% 0% 1350 Mode 3, Safety Rods out, ready for reactor startup 3 MOL 155 25# 0% 0% 1410 Mode 5 4 MOL 579 2155# 75% 2.5% 958 Power decrease from 100%, Xenon increasing 5 BOL 579 2155# 100% EQUIL 1486 Full power, steady state 6 EOL 579 2155# 100% EQUIL 70 Full power, steady state 7 MOL 580 2155# 19% .026% 1347 Ready to sync the main generator with the grid 8 EOL 579 2155# 0% 0% 455 Ready for reactor startup 9 BOL 555 2155# 0% 2.6% 1465 Mode 3, ready for reactor increasing startup following a trip f 10 MOL 579- 2155# 45% 2.6% 957 One MFW Pump in operation, increasing MFW Block Valves closed
FLORIDA POWER CORPORATION s CRYSTAL RIVER UNIT 3 SIMULATOR
' CERTlflCATION SUBMIT 1AL 2.0 DJSIGN DATA SA51 The design of the Crystal River Unit 3 simulator is based on plant data initially submitted to the vendor in 1986. This initial submittal was modified by periodic updates transmitted until the design data freeze date of June 30, 1987, and selected plant modifications from the plant's sixth Refueling outage which were added to the simulator scope after data freeze. The submitted data included all applicable plant drawings, technical / vendor manuals, plant procedures, plant modifications (in-service at the time of submittal or scheduled for implementation), plant process computer listings, design calculations, plant physics test results plant operational and transient data and analytical data including the CR-3 final Safety Analysis Report and B&W ATOG data / analyses. Additional information requested by the vendor during the construction and testing of the simulator was supplied through data requests.
Each simulator :;ystem model report contains a listing of the specific design items used for that module; the entire design data base listing is maintained in the simulator configuration management system, O All plant modifications are tracked, reviewed and evaluated for impact and required incorporation on the simulator. As modifications are implemented, all applicable simulator docu;.entation, including the design data base, is updated accordingly. This process is discussed in further detail in section 4 of this submittal. Due tu the size of the design data base (>5000 records), a listing has not been included with this submittal, If required, the configuration management system can generate a report listing all design data base items. O
FLORIDA POWER CORPORATION O U CRYSTAL R1VER UNIT 3 SIMULA10R CERTIFICATION SUBMITTAL 3.0 SIMULATOR TESTS 3.1 OVERVIEW Simulator performe.we testing was performed in accordance with the requirements ,ontained in ANSI /ANS-3.B(1985). Test results were evaluated by a panel of experts against existing reference plant data, analytical data (B&W ATOG selected transients) and/or expert judgement. Due to the extensive capabilities of the CR-3 simulator, discussed in section 1.3.2 above, it was determined that the scope of performance testing could not address the entire spectrum of possible simulator malfunctions. It was decided that a subset of malfunctions would be tested which would satisfy the following requirements:
- 1. Demonstrate the simulator's ability to comply with each of the transient tests identified in Appendix B2, "PWR Simulator Operability Tests" of the Standard.
- 2. Demonstrate the simulator's ability to comply with O Sectioa 3.i.2 of tne Steodard 4n terms of both the totei number and types of malfunctions specified, including FSAR accidents which result in observable control room indications and which are determined to be appropriate for training.
- 3. Demonstrate the simulator's ability to accurately reproduce plant data on actual plant transients. Table 3.1 contains a list of those performance tests based on actual plant transients.
- 4. Demonstrate the simulator's ability to comply with each of the normal plant evolutions identified in Section 3.1.1 of the Standard.
l 5. Demonstrate the simulator's ability to comply with l Appendix A, Section A3, " Simulator Tests" of the l Standard.
- 6. Demonstrate the simulator's ability to support a l sufficient number of malfunctions and transients to allow l
for effective, varied training and support the requirements of NUREG 1021 regarding the operaticaal i portions of both initial license and requalification
- examinations.
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FLORIDA POWER CORPORATION (l V CRYST AL RIVER UNIT 3 SIMULATOR CERTlflCATION SUBMITTAL 3.1 OVERVIEW (Continued) Based on the above criteria, a total of 102 tests were conducted to demonstrate overall simulator capabilities and compliance with ANSI /ANS-3.5(1985). Table 3.2 lists each of the tests conducted; Table 3.3 provides a cross reference of the conducted tests to the applicable requirements contained in the Standard; Table 3.4 provides a cross reference of performance tests to the Safety Analysis Section of the FSAR. Test Abstracts /results are contained in Volume 11 of the submittal. Each test abstract contains all of the information recommended in Attachment A of " Guidelines for Performance Testing Requirements Under 10CFR55 and Regulatory Guide 1.149" by Neal K. Hunemuller, USNRC. One complete test package is included as Volume 111 of the submittal. The test simulates a steam generator overfill and subsequent reactor / turbine trip, and is based on an actua' plant transient that occurred on August 26, 1983 as documented in V0ER 83-11. In addition to the standard information contained in a completed performance test package, graphic data (. 3_) comparison of major plant and simulator parameters is included for ease of evaluation of the transient. G V
FLORIDA POWER CORPORATION r CRYSTAL RIVERONIT 3 SIMULATOR -] CERTIFICATION SUBMITTAL 3.2 TEST PROCEDURE DEVELOPMENT The format and technical content of the simulator performance tests is based on the simulator acceptance test (ATP) procedures developed and conducted as part of CR-3 Simulator procurement. The scope, format and general content of these procedures were specified by the FPC simulator project team members in order to verify simulator fidelity, scope of simulation, compliance with ANSI /ANS-3.5(1985) and Regulatwy Guide 1.149, Revision 1, and the terms of the simulator procurement specification. The acceptance test procedures were developed as a joint effort between FPC and the simulator vendor, CAE Ltd. The development methodology for the test procedures was as follows:
- 1. For each identified simulator malfunction, a cause and effects document was developed. Each of these documents was reviewed for technical accuracy by a member of the simulator project team.
- 2. For each simulator acceptance tes', an acceptance test O Procedure was developed. ror each trensient and malfunction test, the technical content of the procedure was based on the associated cause and effect document and where applicable, plant reference data and/or analytical or design data. Normal evolutions tests were developed based on plant reference data and the actual plant procedures used to conduct each of the required evolutions. A detailed review of each procedure for technical accuracy was performed by a member of the simulator project team.
- 3. Each of the simulator acceptance tests was performed as' part of simulator pre-acceptance testing. The tests' were conducted by members of the project team and/or members of the plant operating staff. Errors in the test procedure identified during the performance of the test were corrected at this time.
- 4. Each of the simulator acceptance tests was performed as part of simulator acceptance testing. The tests were conducted by members of the project team and/or members of the plant operating shifts. Any additional errors in the test procedure identified during the performance of the test were corrected at this time, n
FLORIDA POWER CORPORATION O. CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICATION SUBMITTAL J.2 TEST PROCEDURE DEVELOPMENT (Continued) The simulator performance tests consist primarily of a subset of the simulator acceptance test procedures. For these selected tests, the technical content is based on the corresponding acceptance tests; some reformatting was needed to ensure that all required information was presented in an easily identifiable manner and that appropriate portions of the tests could be utilized as the test abstracts. Some new tests were developed as part of the simulator performance testing effort; the majority of these tests were developed to include additional reference plant transients in the tested scope. Each of f he new tests was developed by a member of the Nuclear Licensed Operator Training Group and was prerun to ensure completeness and accuracy prior to conduct of the formal performance test. O l'- i L l
FLORIDA POWER CORPORA 110N h CRYSTAL RIVER UNIT 3 SIMULATOR CERTiflCA110N SUBMITTAL 3,3 TEST PROCEDURE FORMAT Each of the simulator performance test procedures contains the following information:
- a. Test Number / Title Contains the test title a unique alpha numeric designator for the test,
- b. Date Performed Indicates the date on which the test was satisfactorily completed.
- c. les_t Duration Indicates the time required to run the test. This includes set-up time if required, simulator run time and data collection time.
- d. ANSI /ANS-3.5(1985) Indicates the section of the Cross Reference standard the test is intended to q satisfy.
V
- e. Test Description Presents a brief description of the test. For malfunction and transient tests, a description of the problem is included as well as a brief summary of the simulator response. If the test uses a variable malfunction, the specific value used during the test is indicated,
- f. JRLtial Conditions Describes the condition of the simulator at the start of the test. The majority of the tests-were initiated from a full power steady state middle-of-life condition with all plant engipment in service and all ct.isu ci systems in~ full autonatic. If the test required a different start point or if equipment was required be out-of-service or in an of f-normal condition, this is indicated.
A V
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FLORIDA POWER CORPORATION (_
- CRYSTAL RIVER UNIT 3 SIMULATOR CERTlflCATION SUPilTl AL 3.3 TEST PROCEDURE FORMAT (Continued) 9 final Conditioni Describes the plant conditions at the point when the test was terminated, in general, these represent stable end points for the test in question,
- h. Plant Procedures Lists the alpha-numeric Used Durina the Test designators, titles and revision levels of all procedure which were used during the test or which might be referenced by the operators during the course of the tested transient / malfunction.
- i. RaselintData Lists the specific plant reference and/or analytical or
-design data used to evaluate the fidelity of simulator response during the test. If no information is listed, the yO evaluation was based on expert judgement,
- j. Test Results/Cqments Contains the evaluation of the transient. In the event that test results differ significantly from expected simulator response or from baseline data, either justification for the differences is included, or the difference is identified as a problem requiring correction /
resolution,
- k. Lest Data lists the parameters recorded during the test. For each parameter the software label, description, recording device, range tnd recording frequency is specified.
- 1. {Lb.iectives Lists the objectives of the test.
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FLORIDA POWER CORPORATION -O CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICATION SUBMITTAL 3.3 TEST PROCEDURE FORMAT (Continued)
- m. Test Acceplancs This section lists the general Criteria acceptance criteria for the test.
These criteria reflect the generic requirement s :;peci fied in Section 4 of the Standard.
- n. Procedure This section provides the specific actiors required to perform the test, specific acceptance criteria for the test and the termination criteria.
Each step in the procedure which is performed satisfactorily is initialed by the person conducting the test. In the event the actions or results identified in a step are not satisfactory, the number of the Trouble Report generated as a result of the problem is O indicated in place of initials.
- o. Test Status This section lists any Trouble Reports generated due to.
unsatisfactory test results, it also contains the signature blocks for the personnel who reviewed and approved the final test results,
- p. Appendix This section contains any supplemental information required for data collection, conduct or evaluation of the test.
Sections a - k and o form the test abstracts contained in Volume 11 of the submittal. In addition to the completed test procedure, each performance test package contains the following information as applicable:
- a. A copy of any baseline data used in the development / i evaluation of the test. '
O b. The alarm typer printout for the transient. -l l l I l
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR .O' CERTlflCATION SUBMITTAL 3.3 TEST PROCEDURE FORMAT (Continued)
- c. A copy of the simulator test data collected during the test, d.. A copy of any Trouble Reports generated during the test.
- e. A simulator Scope limitation Evaluation for any scope limitation encountered during the test if it was not evaluated during another test,
- f. A Pemote function Verification for each remete function performed during the test if it was not verified during another test.
- g. For normal plant evolution tests, the completed plant procedures used to conduct the test.
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FLORIDA POWER CORPORATION Q. CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICA110N SUBMITTAL 3.4 SCOPE OF TESTS As previously discussed, the 100% Performance Test List was developed using the general criteria listed in Section 3.1 above. The specific methodology used to finalize the Test List was as follows:
- a. Eech of the tests identified in Appendix B, Section B.2.2, " Transient Performance" of the Standard was included. These Transient Tests are designated PTTl through PTT10.
- b. The 'ast iaentified in Appendix B, Section B.2.1, " Steady State Performance" of the Standard was included. This Steady State Test is designated PlSI.
- c. The test identified in Appendix A, Section A.3.1,
" Computer Real Time Test" of the Standard was included.
This Computer Real- Time Test is designated PTCl.
- d. Each of the tests identified in Section 3.1.1, " Normal Plant Evolutions" of the Standard was included. These O Normai Tests are des 49 neted eTNi tnrou9h TNiG end ere broken down as follows:
PTN1 includes 3.1.l(1) Plant startup - cold to hot standby; 3.1.l(2) Nuclear startup from hot standby to rated power; 3.1.l(3) Turbine startup and generator synchronization; and 3.1.l(5) Operation at Hot Standby. PTN2 corresponds to 3.1.l(4) Reactor trip followed by recovery to rated power. PIN 3 corresponds to 3.1.l(6) Load changes. PTN4 corresponds to 3.1.l(7) Startup, shutdown and power operations at less than full reactor coolant flow. l PTN5 corresponds to 3.1.l(8) Plant shutdown from rated power to hot standby ana cooldown to cold shutdown conditions. PIN 6 corresponds to 3.1.l(9) Core performance testing. i This test was conducted as in December of 1989 as part of simulator acceptance testing. Since no modifications have been made to the core models since that time, this test was not repeated as part of the simulator performance test program. The test will be rerun when the next core Q cycle is implemented on the simulator. i
FLORIDA POWER CORPORA 110N O cavST^' aivea uni 13 sinut^1oa CERTIFICATION SUBMITTAL 3.4 SCOPEOFTESTS(Continued) PTN7 corresponds to 3.1.l(10) Operator conducted surveillance testing on - safety related equipment or systems. Additional surveillance testing was also conducted as part of normal plant evolutions such as plant heatup, cooldown and power escalation. A list of all the surveillance tests conducted during performance testing is included as Table 3.5. Surveillance Test SP-130, Engineered Safeguards Monthly functional Test, was performed as part of simulator acceptance testing. Since recent modifications to the Engineered Safeguards System have not yet been incorporated on the simulator, the current revision of this procedure cannot be satisfactorily run at this. time. The current revision of this test will be performed as part of the validation process for these Es modifications after they are implemented on the simulator.
- e. Each type of generic malfunction identified in Section 3.1.2, " Plant Malfunctions" of the Standard was included, I with the exception of item (25) Reactor pressure control system failure including turbine bypass failure (BWR). In order to demonstrate full compliance with this section of the Standard, >75 separate Malfunction Tests were conducted. These tests are designated PTM1A1 through PTM248. At least one Malfunction Test per malfunction type was performed. In most cases it was determined that more than one test would be performed to demonstrate either simulator capabilities or to replicate an actual plant transient. -
O r I - -
FLORIDA POWER CORPORA 110N cavstat aivcauNIT 3 SIMULA10R _.O . CERTiflCATION SUBMITTAL 3.5 CRITICAL PARAMETERS for the majority of conducted tests, simulator performance was documented by plotting appropriate plant process parameters using the graphic recorder feature of the 1/F, Specific parameters, their ranges and sampling frequencies for each test were determined by the members of the test team based on the following considerations:
- 1. For each test in Appendix B of the Standard, the applicable specified parameters were recorded at the required sampling frequency. Additional parameters were included for each test as deemed appropriate.
- 2. For other tests, parameters required to document integrated plant respanse were recorded at a frequency appropriate for the duration of the test.
- 3. For tests with minimum integrated plant response, the list of monitored parameters was limited to those directly affected by the test.
O 4. ror tesis wnere pleot resnonse was primeri,y digite, in nature, e.g., loss of individual electrical busses, power supplies, etc., with no integrated plant impact, parameters were not recorded. Documentation of proper response for these tests was accomplished through the use of the test procedure and checklists. For each performance test, a list of recorded parameters is included 'as part of the test procedure. Asterisked (*) parameters for each test are those considered to be " critical" in accordance with the definition contained in the standard. p ( !O
FLORIDA POWER CORPORATION O CRYSTAL R1VER UNIT 3 SIMULATOR CERTiflCATION SUBMITTAL 3.6 CONDUCT OF PERFORMANCE TESTING AND TEST EVALUATIONS Each of the perfornance tests was conducted by one or more members of the Nuclear Licensed Operator Training Group in accordance with Training Department Procedure 1DP-405, Simulator P 1 rforence Testino, Every person involved in the conduct of the tests holds or has held either an NRC SRO License or SR0 Certification on Crystal River Unit 3. For tests requiring a significant amount of operator intervention, a +. least two people were involved in the performance of the test. Data collection was accomplished through the Instructor Station Graphic Recorder, the Simulator Alarm Typer CTS monitoring files and where required, the Plant Process Computer. Discrepancies found during the conduct of the test were noted on the test procedure and followed up by the generation of a Simulator Trouble Report (TR) to initiate resolution of the problem. A list of the TRs generated during performance testing that have yet to be closed via simulator changes are included in Appendix K. The results of each test were initially reviewed at the O conclusion of the test by the person (s) who performed the test. This review included a verification that each of the spccific test criteria contained in the proceduie were met as well as a review of the parameter plots, alarm typer output and any plant process computer printouts. . A final review of each test was conducted by a panel consisting of a member of the Licensed Operator Training Group, a currently licensed SR0 from the Nuclear Operations Group and a member of the Simulator Maintenance and Engineering Group. This review consisted of an evaluation of the test results, including all collected data, against the test acceptance criteria and baseline data when available in order to ensure consistency of evaluation over the full range of performance tests, an evaluation checklist was used. This checklist is contained in TDP-405, Simulator Performance Testi_ng. Any additional discrepancies discovered as a result of this review resulted in the generation of Trouble Reports. Final approval of each test was performed by the Nuclear Licensed Operator Training Manager or the Nuclear Simulator Training Supervisor. O
FLORIDA POWER CORPORATION L.A CRYSTAL RIVER UNIT 3 SIMULATOR d CERTiflCATION SUBMITTAL 3.7 PERFORMANCE TEST SCHEDULE The scope of the performance testing to be conducted over the next four year cycle currently consists of the 100% Performance Test List used to conduct Initial Certification Testing. Changes, deletions or additions to the Performance lest List will be accomplished in accordance with Training Department Procedures TDP-405, Simulator PerfnDnance Teji_ingi and TDP-406, ERC Simytlator Certification Pr.qqgn, which require that the NRC be notified of all planned scope changes to the Performance Test List. The schedule for conducting performance tests over the next four-year recertification cycle is contained in Training Department Procedure TDP-405 and the CR-3 Simulator Configuration t'anagement System as part of the Performance Test Data Base. A copy of the data base is included 3: Appendix G of this submittal. The frequency field of the data base indicaces the performance periodicity of each test based on a four year cycle (A for annual performance, I for first year performance, 2 for second year performance, etc.) The schedule ensures that all tests specified in Appendix B, Section B2 of the standard will be performed annually, that approximately 25% of all the other performance tests will be tested annually, and that all performance tests will be performed at least once over the next four-year recertification cycle. Compliance with this schedule is the responsibility of the Nuclear Simulator Tra' sing Supervisor and will be tracked by the CR Simu'a .or Configuration Management System through the generatioi, of periodic reports indicating required performance test completion dates. O
L 1 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR O CERTIFICATION SUBMITTAL IAELE_L1 EERE9RMANCE_lEH / PLANT TRANSIIHT CROSS _ REFERENCE If311 UILE P L ANT TRANS I ENT/J'.!P_P_0RT I NG__ DAI8 PTTI REACTOR TRIP U0ER 87-02, Reactor 1 rip During Vital Bus Transfer RECALL Data File 0070287 Reector Trip from 89% Power During Vital Bus Transfer. 07/02/87 0 09:53:00 PTT3 SIMULTANE0US CLOSURE OF ALL U0ER 82 15 Reactor Trip MAIN STEAM ISOLATION VALVES Initiated by inadvertant Main Steam Line Closure 06/20/82 0 00:33:06 PTT4 LOSS OF 0FFSITE POWER FROM U0ER 81-10, Reactor Trip and FULL POWER Loss of Offsite Power 06/16/81 PTT5 REACTOR COOLANT PUMP TRIP U0ER 82-11, Successful Runback i on loss of One Reactor Coolant Pump 03/27/82 PTM3A LOSS OF 0FFSITE POWER (LOW U0ER 89-0-02, Reactor Trip and DECAY HEAT) Emergency Feedwater Actuation Due to a Degradation of Offsite Power RECALL Data Fila 0061689. Loss of Offsite Power from 12% Reactor Power 06/16/89 PTM4A REACTOR COOLANT PUMP SHAFT UOER 86-01, failure of RCP-3Al SHEAR RECALL Data File D123185, RCP Shaft Shear from 91% Power 01/01/86 PTM4B LOSS OF 2 REACTOR COOLANT PUMPS U0ER 85-11, Reactor Trip RECALL Data File D120385, Reactor Trip / Loss of B 6900 VAC Bus from 93% Power 12/03/85 P1M6D PARTlAL LOSS OF CIRCULATING Draft U0ER 91- , Reactor Trip WUfR and Emergency Feed Water Actuation Due to a Trip of a Main Circulating Water Pump Q 04/20/91
l l 1 I FLORIDA POWER CORPORATION O CRv51^t river uNil 3 SiMUtA10R CERTIFICATION SUBMITTAL IADLE 3.1 EERQ.R1ANCE TEST / PLANT TRANSIENT C3055 REFERENCE (Continued) IESI! IIILE PLANT TRBLSIENT/SUPPORTIE_.DAIA PTM9A MAIN FEEDWATER PUMP TRIP V0ER 87-01, A MFWP Trip and AND RUNBACK Successful Runback RECALL Data File 0030787, MFP Trip / Runback from 97% Power 03/07/87 PTM9B LOSS OF BOTH MAIN FEEDWATER U0ER 88-02, Reactor 1 rip Due to PUMPS Excessive Steam Generator Depressurization Following Output Breaker Closure RECALL Data File D102888, Reactor Trip /MS Isolation During Startup 10/28/88 n PTM9C OTSG OVERFILL U0ER 83-11, Reactor Trip on V High RCS Pressure following a Main feedwater Pump Trip RECALL Data File 0082683, OTSG Overfill / Reactor Trip on High RCS Pressure 08/26/83 PTM12 DROPPED CONTROL ROD RECALL Data File 0113089, Dropped Control Rod 3-2 from 98% Power 11/30/89 [
FLORIDA POWER CORPORATION
'O CRYSTAL RIVERUNIT 3 SIMULATOR CERTiflCATION SUBMITTAL LABJL32 100% PERFORMANCE TLST LIST TE$T # IIILE PINI PLANT STARTUP - MODE 5 TO FULL POWER PTN2 REACTOR TRIP AND RECOVERY TO RATED POWER PTN3 LOAD CHANGES AT POWER PTN4- THREE REACTOR COOLANT PUMP STARTUP, SHUTDOWN AND POWER OPERATIONS PTN5 PLANT SHUTDOWN AND C00LDOWN PTN6- CORE PERFORMANCE TESTING PTN7 SURVEILLANCE TESTING 3 PTS 1 STEADY STATE PERFORMANCE TEST vp PTCl COMPUTER REAL TIME TEST PTTI REACTOR TRIP 9 PTT2 TOTAL LOSS OF FEEDWATER
~
PTT3 SIMULTANE0US CLOSURE Of ALL HSIVS PTT4 LOSS OF 0FFSITE POWER FROM K'LL POWER PTT5 REACTOR COOLANT PUMP TRIP PTT6 MAIN TURBINE TRIP PTT7 MAXIMUM RATE POWER RAMP PTT8 LARGE BREAK LOCA (DBA) PTT9 MAIN STEAM LINE BREAK INSIDL CONTAINMENT PTT10 PORV FAILURE WITH HPI INOPERABLE PTM1A1 STEAM GENERATOR TUBE LEAK PTM1A2 STEAM GENERATOR TUBE RUPTURE Q PTMlB LETDOWN LEAK DOWNSTREAM 0F MOV-49
v FLORIDA POWER CORPORA 110N CRYSTAt river uni 13 SiMutA10R
' O- CERTIFICATION SUBMITTAL IABLE ? '
101% PERFORMANCE TES: LLSI(Continued) II.51_f 11111 PTHICI REPRESSURIZlNG SMALL BREAK LOCA PTMIC2 SMALL BREAK LOCA PTMIC3 SMALL BREAK LOCA WITN HPI INnprRABLE PTMID PRESSURIZER SAFETY VALVE Fait.bnt. PTM2 LOSS OF INSTRUMENT AIR PTM3A LOSS OF 0FFSITE POWER (LOW DECAY HEAT) PTM3B1 LOSS OF EMERGENCY OFF-SITE SOURCE PTM3B2 LOSS OF EMERGENCY OFF-Si1E SOURCE / FAILURE OF ONE EDG PTM3Ci LOSS OF NNI-Y 24 VDC PTM3C2 LOSS OF HNI-Y 120 VAC PTM3D1 LOSS OF ICS DC POWER PTM3D2 LOSS OF ICS-AC POWER PTM3E LOSS OF VITAL BUS PTM3F LOSS OF 4B0VAC ES BUS A t-O PTM3G LOSS OF DC DISTRIBUTION PTM4A REACTOR COOLANT PUMP SHAFT SHEAR PTM4B LO3S OF TWO REACTOR COOLANT PUMPS PTM4C REACTOR COOLANT PUMP SEAL FAILURE PTM4D REACTOR COOLANT PUMP LOCKED ROTOR PTM5A LOSS OF CONDENSER VACUUM DUE TO AIR IN-LEAKAGE PTM5B LOSS OF CONDENSER LEVEL CONTROL PTM6A LOSS-0F NUCLEAR SERVICES RAW WATER PTM6B LOSS OF CIRCULATING WATER PTMGC LOSS OF DECAY HEAT RAW WATER PTM60 PARTIAL LOSS OF CIRCULATING WATER PTM7 LOSS OF DECAY HEAT REMOVAL PIMBA LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING PTM88 LOSS OF SECONDARY SERVICES CLOSED CYCLE COOLING PTM8C LOSS OF-DECAY HEAT CLOSED CYCLE COOLING PTM9A MAIN FEE 0 WATER PUMP TRIP AND RUNBACK PTM9B LOSS OF BOTH MAIN FEEDWATER PUMPS PTM90 OTSG OVERFILL PTM9D LOSS OF'BOTH CONDENSATE PUMPS i O r
FLORIDA POWER CORPORA 110N CRYSTAL RIVERUNIT 3 SlHULATOR O CERTIFICATION SUBMITTAL IRLLL2 JpQ% PERFQRMANCE TEST LIST (Continued) IE1Li IllLE PTM10 TOTAL LOSS OF FEEDWATER PTMllA FAILURE OF RPS +15V CABINET POWLR SUPPLY PTMllB FAILURE OF ES CABINET POWER SUPPLY PTM12A MECHANICALLY STUCK R0D PTM12B UNC0VPLED R0D PTM12C DEGRADED R00 MOTION
.PTM120 DROPPED CONTROL R00 PTM13 INABILITY TO DRIVE CONTROL RODS PTH14A HIGH RCS ACTIVITY PTM14B WASTE GAS DECAY TANK LEAK PTM16 MAIN GENERATOR TRIP PTM17A CONTROL R00 OUT INHIBIT FAILURE PTM17B CONTROL R0D GROUP COMMAND ENABLE FAILURE PTM17C FAILURE OF THE MAIN TURBINE TO TRIP PIM17D ATMOSPHERIC DUMP VALVE FAILURE PTM17E TURBINE BYPASS VALVE FAILURE PTM17F CONTINUOUS R0D WITHDRAWAL PTM17G DILUTION ACCIDENT PTM18A MAKEUP PUMP TRIP PTM18B PRESSURIZER LEVEL CONTROL VALVE FAILURE PTM18C- LETDOWN ISOLATION VALVE FAILURE PTM180 PRESSURIZER HEATER CONTROLLER FAILURE PTM18E PRESSURIZER SPRAY VALVE FAILURE PTM18F RCS NARROW RANGE PRESSURE TRANSMITTER FAILURE PTM20A MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT PTM20B MAIN FEED LINE BREAK INSIDE CONTAINMENT PTM20C MAIN FEED LINE BREAK OUTSIDE CONTAINMENT PTM21A POWER RANGE CHANNEL FAILURE PTM21B INTERMEDIATE RANGE CHANNEL FAILURE PTM210 SOURCE RANGE CHANNEL FAILURE LO 1
l
FLORIDA POWER CORPORATION O CRYSTAL RIVERUNIT 3 SIMULATOR CERTIFICATION SUBMITTAL TABLE 3.2 100*. PERFQ.RMAECE TEST LIST (Continued) lESI_f IIILE PTM22A RCS Thot _ TRANSMITTER FAILURE PTM22B RCS Tcold TRANSMITTER FAILURE PTM22C MAIN FEEDWATER FLOW TRANSMITTER FAILURE PTM220 OTSG OPERATING RANGE LEVEL TRANSMITTER FAILURE PTM22E ICS RUNBACK FAILURE PTM22F ICS HEADER PRESSURE BIAS FAILURE PTM22G MAIN FEEDWATER TEMPERATURE TRANSMITTER FAILURE PTM22H FAILURE OF ICS TO TRANS~~.R TO TRACKING MODE PTM221 REACTOR DEMAND SIGNAL FAILURE PTM22J SELECTED TAVE FAILURE PTM23A MS ISOLATION ACTUATION FAILURE PTM23B MFW ISOLATION ACTUATION FAILURE PTM23C ES CHANNEL FAILURE TO ACTUATE HPl PTM23D ES CHANNEL FAILURE TO ACTUATE RB ISOLATION AND COOLING PTM23E it.. AUTO ACTUATION FAILURE PTM24A ATWS (N0 OPERATOR INTERVENTION) ! PTM24B ATWS (OPERATOR INTERVENTION) l O
FLORIDA POWER CORPORA 110N
'O CRvS1At RivERuNi13 SinutA10R CERTIFICATION SUBMITTAt IML E 3.3 PERFORMANCE TEST /ANS-ALBJAU_lRMENT CROSS INDEX TEST # ANSI /ANS-3.511985)_REQM E M MI PTN1 Section 3.1.1(1)(2)(3)(5), Appendix A Section A3.2 PTN2 Section . 3.1.l(4), Section 3.1.2(19), Appendix B Section B2.2(1),
Appendix A, Sections A3.2, A3.3 PTN3 Section 3.1.l(6), Appendix A Section A3.2 PTN4 Section 3.1.l(7), Appendix A Section A3.2 PTN5 Section 3.1.l(8), Appendix A Section A3.2 PTN6 Section 3.1.l(9), Appendix A Section A3.2 PTN7 Section 3.1.l(10), Appendix A Section A3.2 PTSI Appendix. B Section B2.1, Appendix A Section A3.2 PTCl Appendix A Section A3.1 PTTI Section 3.1.2(19), Appendix B Section B2.2(1), Appendix /. Sections A3.3, A3.4 PTT2 Section 3.1.2(9)(10), Appendix B Section B2.2(2), Appendix A Sections A3.3, A3.4 PTT3 Appendix B Section B2.2(3) PTT4 Section 3.1.2(3)(4), Appendix B Section B2.2(4), Appendix A Sections A3.3, A3.4 PTT5 Section 3.1.2(4), Appendix B Section B2.2(5), Appendix A Sections A3.3, A3.4 PTT6 Section 3.1.2(15), Appendix B Section B2.2(S), Appendix A , Section A3.4 PTT7 Section 3.1.l(6), Appendix B Section B2.2(7) PTT8 Section 3.1.2(1)((b)(c)(3), Appendix B Section B2.2(8), Appendix A Sections A3.3, A3.4 PTT9 Section 3.1.2(20), Appendix B Section B2.2(9), Appendix A Sections A3.3, A3.4 PTT10 Section 3.1.2(1)(c)(d)(23), Appendix B Section B2.2(10), Appendix A Sections A3.3, A3.4 PTMIAl Section 3.1.2(1)(a), Appendix A Section A3.4 PTM1A2 Section 3.1.2(1)(a), Appendix A Section A3.4 PTMIB Section 3.1.2(1)(b)(c), Appendix A Section A3.4 PTHICl Section 3.1.2(1)(c), Appendix A, Sections A3.3. A3.4 PTMIC2 Section 3.1.2(1)(c), Appendix A, Sections A3.3, A3.4 PTMIC3 Section 3.1.2(1)(c), Appendix A, Sections A3.3, A3.4 PTMID Section 3.1.2(1)(d), Appendix A Sections A3.3, A3.4 PTM2 Section 3.1.2(2), Appendix A Section A3.4 \
l ft0RIDA POWER CORPORATI0ff O CRvS1 At RivcRurii13 SiMutA10R CERT 1 FICA 110ti SUBMITTAt leRE.12 f1B[QRBARCI TEST /ANS-3.5 REQMM[NT CPOSLINQU(Continued) IL$IJ AMi]/ANS-3.5f1985) REQUI R$181 PIM3A Section 3.1.2 3)(4), Appendix A Sections A3.3, A3.4 P1M3B1 Section 3.1.2 3), Appendix A Section A3.4 PTM3B2 Section 3.1.2 3), Appendix A Section As.4 P1M3Cl Section 3.1.2(3)(22), Appendix A Section A3.4 PIM3C2 Section 3.1.2(3)(22), Appendix A Section A3.4 PTM3D1 Section 3.1.2(3)(22), Appendix A Section A3.4 PTM3D2 Section 3.1.2(3)(22), Appendix A Section A3.4 P)M3E Section 3.1.2(3)(11)(21)(22), Appendix A Section A3.4 PIM3F Section 3.1.2(3), Appendix A Section A3.4 PTM3G Section 3.1.2(3)(23), Appendix A Section A3.4 , PTM4A Section 3.1.2(4), Appendix A Sections A3.3, A3 i P1M4B Section 3.1.2(3)(4), Aprendix A Sections A3.3, A3.4 PTM4C Section 3.1.2(4), Appendix A Section A3.4 PIMSA Section 3.1.2(5), Appendix A Section A3.4 PTMSB Section 3.1.2(5), Appendix A Section A3.4 PTM6A Section 3.1.2(6), Appendix A Section A3.4 O PIM6B PIM6C Section 3.1.2(6)(8), *.ppendix A Section A3.4 Section 3.1.2(6)(7)(B), Appendix A Section A3.4 P1M60 Section 3.1.2(6)(8), Appendix A Sections A3.3, A3.4 PTM7 Section 3.1.2(7), Appendix A Section A3.4 PTM8A Section 3.1.2(8), Appendix A Section A3.4 PTM8B Section 3.1.2(8), Appendix A Section A3.4 PTM8C Section 3.1.2(7)(8), Appendix A Section A3.4 PTM9A Section 3.1.2(9), Appendix A Sections A3.3, A3.4 PTM9B Section 3.1.2(9), Appendix A Sections A3.3, A3.4 PIM9C Section3.1.2(9),AppendixASectionsA3.3,A3.4 PTM9D Section 3.).2(9), Appendix A Section A3.4 PIM10 Section 3.1.2(10), Appendix A Sections A3.3, A3.4 PIMllA Section 3.1.2(11)(17)(21), Appendix A Section A3.4 PTMllB Section 3.1.2(11)(17), Appendix A Section A3.4 P1M12A Section 3.1.2 12), Appendix A Section A3.4 PTM128 Section 3.1.2 12), Appendix A Section A3.4 PTM12C Section 3.1.2 12), Appendix A Section A3.4 PTM12D Section 3.1.2(12), Appendix A Sections A3.3, A3.4 PIM13 Section 3.1.2(13)(17), Appendix A Section A3.4 PTM14A Section 3.1.2(14), Appendix A Section A3.4 PTM14B Sectinn 3.1.2(14), Appendix A Sectinn A3.4 PTM16 Section 3.1.2(16), Appendix A Section AJ.4 PTM17A Section 3.1.2(17), Appendix A Section A3.4 PIM178 Section 3.1.2(17), Appendix A Section A3.4 P1H17C Section 3.1.2(17), Appendix A Section A3.4 O eini7o PTM17E Section 3.1.2(>7)(20),^Pnendix^Sectioa^3.4 Section 3.1.2(17), Appendix A Section A3.4
_ _____ .. . _ _ _ _ _ _ _ _ _ _ _ _ . . . . _ . . _ ...__.m. _. _. fl0RIDA POWER CORPORA 110N l O CRvSiAtRivcauni13SiMutAiOR CIR11 FICA 110N SUliM111Al IABIL1J PlRE0AMEL1HI/ANS hLREQUJRBGLCROSLINDEX(continued) IESL1 M l/ ANSA & 11BM _REQVJREM MI PIM17f Section 3.1.2 7), Appendix A Section A3.4 PIM170 Section 3.1.2 7), Appendix A Section A3.4 , PIM18A Section 3.1.2 8) Appendix A Section A3.4 ' PIM18B Section 3.1.2 18 , Appendix A Section A3.4 P1M18C Section 3.1.2(18 , Appendix A Section A3.4 PIM18D Section 3.1.2 18 , Appendix A Section A3.4 P1H18E Section 3.1.2 18), Appendix A Section A3.4 P1H18f S:ction3.1.218), Appendix-ASectionA3.4 P1M20A Section 3.1.2(20), Appendix A Sections A3.3, A3.4 PIM20B Section 3.1.2(20), Appendix A Section A3.4 PIM20C Section 3.1.2(20), Appendix A Section A3.4 P1M21A Section 3.1.2(21 , Appendix A Section A3.4 P1H21B Section 3.1.2 , Appendix A Section A3.4 P1M210 Section 3.1.2 , Appendix A Section A3.4 PIM22A Section 3.1.2 , Appendix A Section A3.4 n P1M220 Section 3.1.2 22 , Appendix A Section A3.4 U P1H220 Section 3.1.2 9)(22), Appendix A Section A3.4 P1M220 Section 3.1.2 9)(22), Appendix A Section A3.4 P1M22E Section 3.1.2(22), Appendix A Section A3.4 PIM22F Section 3.1.2(17)(22), Appendix A Section A3.4 PIM22G Section 3.1.2(9)(22), Appendix A Section A3.4 P1M22ii Section 3.1.2(22), Appendix A Section A3.4 PIM221 Section 3.1.2(17)(?2), Appendix A Section A3.4 P1H22J Section 3.1.2(22), Appendix A Section A3.4 PTH23A Section 3.1.2(23), Appendix A Section A3.4 PIM23B Section 3.1.?(23), Appendix A Section A3.4 P1M23C Section 3.1.2(23), Appendix A Section A3.4 PTH23D Section 3.1.2(23), Appendix A Sect'on A3.4 P1M23E Section 3.1.2(23), Appendix A Section A3.4 PIM24A Section 3.1.2(17)(24), Appendix A Section A3.4 PIM24A Section 3.1.2(17)(24), Appendix A Section A3.4 O
FLORIDA POWER CORPORA 110N O CRYSTAL RIVER UNIT 3 SIMULATOR CER11FICAT10N SUBMIT 1AL IMLEA4 RRf0RMEIlI11/11 AIL 5AEEILANAL11113R05SlEEBLEI UNCOMPENSATED OPERATING REACTIV11Y CHANGLS The ability of the control rods and the makeup and purification system (boration and dilution) to compensate for reactivity changes due to fuel depletion and xenon burnup/ buildup is verified in the following Performance Tests: PINI - Power Escalation PIN 2 - Reactor Trip and Recovery to Rated Power PIN 3 - Load Changes at Power STARTUP ACLlD[td Proper simulator response to a continucus rod withdrawal accident during a reactor startup is verified in the following Performance Test: O i mi7r - Continuous Rod withdrawai RODllTilDRAWAL AT RA1ED POEILQRBAI.10N ACCIDEfd Since CR 3 operates at rated power with the control rods fully (>90%) withdrawn, it was determined that this accident was not suitable for simulator training. MODERAT0_R DILUTION ACLlD.WI Proper simulator response to an inadvertent dilution with the reactor at power is verified in the following Performance Test: PIM17G - Dilution Accident COLD WATER ACCIDEtG The classic cold watcr accident is the start of an idle Reactor Coolant System Loop which has been isolated by primary system isolation valves. Since the CR-3 design does not include primary isolation valves, the classic cold water accident cannot occur. O
FLORIDA POWER CORPORA 110N Q CRYSTAL RIVERUNIT 3 SIMULA10R CER11IICAT10N SUBM11TAl j IAi!LLLA HRE0AWG LIESI/11MLS ALEILAML151LCROISJ tEGMC C (Cont i nue d ) LQ15 Of C00LAN1 FLOW ACC10ENI Proper simulator response to partial and full losses of Reactor Coolant System forced flow due to Reactor Coolant Pump Trip (s) is verified in the following Performance Tests: l PTT4 - Loss of Offsite Power from full Power P1TS Reactor Coolar t Pump Trip PTT4B - Loss of Two Reactor Coolant Pumps PIM4C - Reactor Coolant Pump Seal failure PTM3A - Loss of Offsite Power (Low Decay Heat) Proper simulator response to a partial loss of Reactor Coolant System forced flow due to an RC Pump shaft shear is verified in the following Performance Test: PIM4A Reactor Coolant Pump Shaft Shear O eroPer simulator response to a Parti 11oss of Reactor Cooiant System forced flow due to an RC Pump locked rotor is vartfied in the following Performance Test: PTM4D Reactor Coolant Pump Locked Rotor Sly & ll STUCK 0VT OR DR0FPED CONTROL R0D ACCIDJ1(1 Proper simulator response to control rod malfunctions is verified in the following Performance Tests: PTH12A - Mechanically Stuck Rod P1M12B - Uncoupled Rod P1M12C Degraded Rod Motion PTM120 - Dropped Control Rod l
FLORIDA POWER CORPORATION O CRYSTAL R1VERUN113 SlHULATOR CERilflCATION SUBMITTAL IMLL LA P.LRIQRNGLlL1ILESAR SAFIlLMAL11l3_CRQ11_MELRIEE(Continued) LOSS Of ELECTRIC POWER ACCID W1 The CR 3 Electrical Distribution System design precludes a total loss of AC power without the occurrence of at least four separate failures within the system 1.e., a simultaneous failure of both EDGs, the station startup truisformer and the emergency offsite source. This is not regarded as a (redible scenario, requiring multiple failures and placing the sirtttr electrical distribution system in an off normal lineup. Howevv, preer sir;'ator response to losses of of f-site /on-site power so'rces is ve<ii:ed in the following Performance Tests: PT18 - Large 5 $ LOCA (DBA) PIMM - Loss of Ottsite Power (Low Decay H< cat) PT14 Loss of 400 VAC and 4160 VAC Unit Busses from full Power P1M381 - Loss or Emergency Off-site Source PTM302 - Loss of Emergency Off-site Source /f ailure of One EDG STEAM LINE FAILURE ACCIDENT O Proper simulator response to steam lines breaks of various sizes and location both inside and outside the containment is verified in the following Perfortrance Tests: PTT9 - Main Steam Line Break Inside Containment P1T20A - Main Steam Line Break Outside Containment P1T170 - Atmospheric Dump Valve failure STEAM GEWERATOR TUBE RUPTURE ACCIDifil Proper simulator response to steam generator tube ruptures of various sizes is verified in the following Performance Tests: PIM1A1 - Steam Generator Tube Leak PIM1A2 - Steam Generator Tube Rupture FUEL HANDLING ACCIDENTS These accidents are considered outside the scope of simulation. BRD_Ld[fT10N ACCID M1 This accident was determined not to be suitable for training. O
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1 FLORIDA POWER CORPORA 110N O CRYS 1 AL RIVER UNIT 3 SIMUL A10R CERilflCATION SUBMITTAL IMLE 3.4 EEEEQfLMKE TEST /fSARJAFETY AliA11111_CRMLBEFEREEE(Continued) LOSS Of C00LANT ACCIDEN11 Proper simulator response to loss of coolant accidents of various break sizes and locations is verified in the following Performance Tests: PTIO - Large Break LOCA(DBA) PTT10 - PORV f ailure with HPl Inoperable PTMICI - Repressurizing Small Brea( LOCA PIMIC2 - Small Break LOCA PIMIC3 Small Break LOCA with HPl Inoperable PIMID - Pressurizer Safety Valve failure MAKEUP SYSTEM LEJDOWN LINE FAILURE ACCIDENT Proper simulator response to this accident is verified in the following Performance Test: O eTTiB - tetdown tine teak Downstreem of Muv-49 MAXIMUM HYP0THETICAL ACCIDEfH This accident is covered under loss of Coolant Accidents WASTE GAS DECAY 1ANK_BUPTVRE ACCIDENT Proper response to a Waste Gas Decay Tank leak is verified in the following Performance Test: P1M148 - Waste Gas Decay Tank Leak LOSS Of FEEDWATER AND FEEDWATER LINE BREAK ACClHEHLS Proper response to partial and complete losses of main feedwater, a total loss of feedwater and feedwater line breaks of various sizes and locations is verified in the following Performance Tests: PTT2 - Total Loss of feedwater PTM9A - feedwater Pump Trip and Runback PTM98 - Loss of Both Main feedwater Pumps PTM90 - Loss of Both Condensate Pumps PTH10 - Total Loss of feedwater PIM20B Main feed Line Break inside Containment
- Main feed Line Break Outside Containment Q PIM20C
FLORIDA POWER CORPORATION O CRYSTAL RIVER UNIT 3 SIMULATOR CERTiflCATION SUBMITTAL IABI L h5
$ RYfRLANCE TL$li 110EE0ME_! 111LI II1Li SP-130 Engineered Safeguards Monthly functional Test PIN 7 SP-300 Operating Daily Surveillance Log PTS 1 SP-301 Shutdown Daily Surveillance Log PIN 7 SP-312 Daily Heat Balance Power Comparison PIN 2/PTSI SP-317 RC System Water Inventory Balance PIN 7 SP-321 Power Distribution Breaker Alignment and Power Availability Verification PTN7 SP-323 Evacuation and fire Alarm Demonstration PIN 7 SP-125 Turbine Generator Checks PIN 7 SP-332 Monthly Steam Line and feedwater Line Isolation functional Test PTN7 SP-333 Control Rod Exercises PIN 7 SP 343 Main Steam isolation Valves Part-Stroke Exercising PTN7 SP 349A EfP-1 Operability Surveillance PTN1 O. SP-3498 EfP-2 and Valve Operability Surveillance PIN 1 SP 354A Monthly functional Test of the Emergency Diesel Generator EGDG-1A PIN 7 SP-356 Operations ES Monthly Manual Actuation Channel functional Test for RBl&C PIN 7 SP 358A Operations ES Monthly Automatic Logic functional Test #1 PIN 7 SP-358B Operations ES Monthly Automatic Logic functional Test #2 PIN 7 SP-358C Operations ES Monthly Automatic Logic functional Test #3 P1N7 SP-390 Startup Surveillance Log P1N1/PTN2 SP-422 RC System Heatup and Cooldown Surveillance PTNS O
ft0R10A POWER CORPORA 110f1 Q CRYST AL RIVER VillT 3 SIMUL A10R CER'ilflCAT10f1 SUBMITTAL , 4.0 51MLATOR DISGMPM0Y AND UE0MDE PRQAM!i 4.1 OVERVIEW Configuration control of the CR-3 Simulator utilizes a combination of procedural guidance, formal written instructions and a PC-based relational data base to ensure compliance with the requirements of Section 5, " Simulator Design Control" of the Standard.
~
The configuiation control process, which is based on INP0 Guideline 10 504, ' Simulator Configuration Management System", is conducted in accordance with the following Training Department Procedures: 10P-401, Simulator Smtrce Dotqment Evalya11gn 10P-402, Review _pf Plant / indust ry Ongrating. tvenh 1DP 403, Simulator Chance S.1p_nAgrds n These procedures, which are included as Appendices H, I and J O to this submittal, define the major portions of the simulator configuration control process including: review and evaluation of plant modifications / changes for possible incorporation on the simulator; review of plant and industry operating events for impact on the simulator and potential modifications to the 100% Performance Test List; implemcntation, testing and documentation of plant modifications and elective enhancements . on the simulator; and correction and documentation of performance problems and/or hardware problems within the existing scope of sim'ilation. in addition to these procedures, detailed guidance on impicmenting specific portions of the configuration control process are defined in the following Simulator Operating Instructions: S01-08, S.LMEX Confiagrhn_AnJLLqr.qgtgnad.Jan_agengni 501-11, Simul.ptor 1rquble Ratgrt s 501-12, Conf _Lquration itemunJLDasj.gdaltRaic S01-13, Common Data BaidpAate lpagina Protem S01-14,1/F Schematic Modifications ,
-_----__-._________-_____m.. . _ - . . _ _ _
FLORIDA POWER CORPORA 110N cavs'^' aivta unit a sinu'^'on O CERTIFICA110N SUBM111AL 4.1 OVERVIEW (Continued) finally, the CR 3 configuration control process utilizes a Configuration Management System PC based relational data base to track the mejor steps of the process and maintain / update all required data bases. 4.2 SIMULATOR H0DIFICATIONS Changes to the simulator originate from one of three sources: 4.2.1 11Lnt Modific3figm All plant modification packages are transmitted to the Simulator Group as part of the ' plant modification review and approval process controlled by fPC Nuclear Engineering Department procedures. Each modification is reviewed by the Nuclear Simulator 1 raining Supervisor and the Simulator Maintenance and Engincering Supervisor for potential impact on the simulator, and the results of their evaluations are reviewed by the Nuclear Licensed Operator Training Manager. Each modification package, along with its evaluation results, O is entered in the Configuration Management System data base. On a monthly basis, the plant modification data base is downloaded to the Simulator Configuration Management System and compared to the CMS modification data base. This process serves two purposes: first, it serves as a check to ensure that all modification packages have been transmitted to the Simulator Group for review and evaluation; second, as modifications are placed in service in the plant, CMS automatically assigns required simulator completion dates to the modifications based on their in-service dates. All plant modifications which were in service as of the simulator design data freeze are included in the scope of simulation. All modifications implemented in the plant after design freeze have been identified and evaluated, and those determined to have impact have been assigned required completion dates. Appendix D contains a list of all plant modifications requiring implementation on the simulator as of the date of this submittal. O
FLORIDA POW [R CORPORA 110f4 O CRYSl AL RIVERUNil 3 SIMUL A10R CER11 FICA 110tl SUBMlllAL t 4.2.2 flantandntrLDP.eraling.Irentti Plant / industry operating events, in the form of LERs, 50ERs, 00ERs, etc. are reviewed for potential training impact in accoroance with Training Department Procedure 1[iP-105, trian ILlan Prepar1110L If the incorporation of tht event as either a training exercise or a performance test requires modifications to the simulator software, the initiating document is entered into CMS for tracking and eventual implementation. 4.2.3 Eartpmementi Desired elective modifications or additions to the simulator design data base are entered and tracted in CMS for eventual implementation. 4.3 WORK PACKAGES Changes to the simulator are tracked and documented through simulator work packages. For each change requiring p implementation, a work package is assigned to the responsible Simulator Computer Controls Specialist (hardware) and/or Simulator Systems Engineer (software). Tracking of progress on each change is accomplished through periodic status updates of each work package in CMS. 4.4 TESTING Prior to closure of a simulator work package (s), the implemented change must be tested and approved. Wherever possible, testing is based on existing performance tests, simulator exercise guides, or plant procedures. The fluclear Simulator Training Supervisor is responsible for ensuring that all performance tests affected by a simulator change are updated to accurately reflect the change. 4.5 00CUMENTAT10fi After satisfactory completion of all required testing, and as a prerequisite to closure of a work package (s) all documentation affected by a change is updated to reflect the modification, O
FLORIDA POW [R CORPORAT10f1 cavsi^t aivo n u'<it a sinui ^'on
~O CLRilflCA110N SUBMillAL 4.6 TROUBLE REPORTS in addition to changes which affect the simulator design data base, performance anomalies / hardware problems within the existing scope of simulation are identified through simulator 1 rouble Reports (1Rs). Once identified, correction of the problem, testing of the resolution, and documentation of the correction are conducted in accordance with the guidelines discussed above.
Apsendix K lists all open 1 rouble Reports at the time of this su>mittal. O O
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR O' CERTlflCATION SUBMITTAL AEEIMD.lLA 11EORM ANC LIDl_)IMlDiBIR_QUAllf1CAIl0H1 i This Appendix consists of abstracts from the resumes of the members of the Crystal River Unit 3 Training and Operations Departments involved in the conduct and evaluation of the simulator performance tests. Rhn A. Lind (Nuclear Lictnsed Oppntor Traininst Manuttl Responsible for the design, development and impicmentation of Licensed 0)erator and Senior Reactor Initial and Requalification Training Programs, tie simulator portions of the Shift Technical Advisor and Engineer Training Programs, all hardware and sof tware maintenance and modification activities on the- Crystal River Unit 3 Simulator ano all simulator certification activities. Served as the Project Manager for the simulator procurement effort. Previously served as the FPC Senior Simulator Systems Engineer responsible for the review of simulator design documents, participation in simulator design review meetings, initial development of the simulator configuration management system and development of the simulator Acceptance Test Program. O ^> a coasultaat, served in various Posit'ons inciudin9: Manager of the Beaver Valley Simulator Training Center where he was responsible for all aspects of simulator operation, including the design, development and implement ation of training programs, supervision of the instructional and simulator hardware /sof twara maintenance staffs and the development of the simulatnr configurat f on management system; NUS Project Engineer with responsibility for the training and operational aspects of simulator procurement projects, including the V.C. Summer, Beaver Valley and Kewaunee simulators. Provided , operations support to the NUS Consulting Division on various projects including the Industry Degraded Core Rulemaking (IDCOR) Program, and conducted customer training programs on the Sequoyah, Surry and Cumberland City simulators; Supervisory Engineer at the Babcock and Wilcox Training Center, where he was responsible for all aspects of customer classroom and simulator training and simulator hardware and software maintenance and upgrades. Provided operations support for the B&W Transient Assessment Program and development of the Small Break LOCA Guidelines and the B&W ATOG Program; O
, fl0R10A POWER CORP 0 iia 110N
-O CRvStat RivcR unii a Sinut^1oR CIR11 FICA 110N SUBM111AL ,
John A. LinLuing.lpar_LittanLQptrnor ItaluiD9JiBMgtIJ(Continutd) Member of the B&W Haster Services leam for Crystal River Unit 3 initial fuel load, zero power physics testing and power escalation testing, where he served as the B&W site representative to ensure proper operation and testing of B&W supplied equipment and overall plant operation in compliance with lechnical specifications, supplied on-shift operational and technical advice to IPC operations personnel, and coordinated testing activities; Test Engineer for the Midland Project, where he was responsible for the cogeneration unit test schedule and preoperational and ! operational test procedure development; Engineering Watch Supervisor and Reactor Controls Division Lead Petty - Officer assigned to the U.S. flavy f ast Attack Submarinc VSS Jack (SSN 605).
-He has twenty one (21) years of Nuclear Power experience, including fif teen (15) years commercial experience on B&W and Westinghouse PWRs. He holds a Bachelor's Degree in Mathematics from Boston College, and has held an NRC SRO License on Crystal River Unit 3 and an SRO Certification on Beaver Valley Unit 1.
datkson L.jpr_tttger_{11uc1 ear Sj!nglator Trajning_Sypgr_yis.pij Responsible for the supervision and coordination of all training activities on the CR 3 Control Room Simulator, development and maintenance of NRC simulator examination banks, evaluation of all plant modifications for
-impact on the simulator, design / development / implementation of the simulator portions of all_ CR-3 training programs, implementation and maintenance of the Simulator Certification Test Program. Served as a member of the Simulator Project Team, responsible for all training /operatio aspects of ,
the simulator project including determination of the scope of .lation and malfunction capabilities, and the design and conduct of tr. Simulator Acceptance Test Program. Past experience includes the following: Florida Power Corporation Nuclear Operations Instructor responsible for the preparation and conduct of Initial and Requalification Operations Training programs for licensed and Non licensed personnel. Also participated in scenario development and served as Control Rom implementor for three annual emergency exercises. Currently holds an SR0 certification on CR 3. As a florida Power arporation Nuclear Plant Operator, held all Operations positions from Assistant Nuclear Auxiliary Operator (ANAO) O through Chief Nuclear Operator (CNO). [arned a reactor operator license and senior operators license.
FLORIDA POWER CORP 0RA110N O CRYSTAL RIVER UNil 3 SIMULA10R CER11FICAT10N SUBM111AL hEALQu.L_.$PIIDECLlHVGINr_51!WinRLIL61010L6VRCLY11Qr) (C ont i nued ) As a florida Power Corporation Nuclear Plant Mechanic, participated in various maintenance activities prior to initial startup of CR-3, including reactor vessel and internals assembly and the replacement of three reactor coolant pump impellers. Served-in the U.S. Coast Guard where he achieved a rate of Engineman Second Class (ES). i He has seventeen (17) years of commercial nuclear power experience. He is currently enrolled as a senior with the University of Alabama. B9LL. LiMOD_lilEd319LMLIMtDage_anLDigintering,$ypmilor) Responsible for the supervision and coordination of all simulator-related engineering, maintenance, and contractor personnel to assure adupacy of CR-3 simulator activities; i.e., reference plant modification reviews, ; development and testing of simulator modifications, discrepancy disposition, verification of simulator fidelity, contract management, budgets, parts procurement, and maintenance of technical / administrative procedures. Past experience includes the following: florida Power Corporation Senior Simulator Engineer, responsible for the maintenance, modification, and documentation of the nuclear simulator software program. Provided technical guidance and direction to junior simulator engineers. Developed the configuration control process and the PC resident dBase IV application software / data bases for compliance with ANS-3.5. florida Power Corporation Simulator Engineer. Resident engineer at the simulator vendor's (CAE) facility while participating in a 21/2 year 0J1 program. Lead engineer for real time modeling and hardware integration of the main steam system, MSR/fWH drains, reheat control system, turbine support systems, and other BOP systems. Activities included code (f 0R1RAN) development, background testing, tuning,1/0 check out, foreground integration, discrepancy resolution, and documentation. Experience with both GOVl0 (32/97) and VAX Operating Systems, florida Power Corporation Performance Engineer. Lead engineer in the development and implementation of BOP performance monitoring / enhancement activities, Provided guidance to Operations and Maintenance to resolve system anomalies which degrade station efficiency or plant safety based on performance analysis and root cause evaluation. Was also responsible for plant thermal / hydraulic heat balances, maintenance of PEPSE code models, performance data base management, and PPC softwara changes to enhance or correct O performance monitoring calculations.
floalDA POWER CORPORATION O ca's'^' aiVER UNIT 3 SIMULATOR CERTiflCATION SUBMITTAL JLoy L. Linton (Simulator Maintenance and EncineeringJypervisgr)(Continued) Supported the simulator project team in discussions with the simulator vendor and outside consultants regarding plant design data, , plant performance data, configuration management, plant drawing systems, and documentation reviews. Babcock and Wilcox Company Systems Engineer. Lead engineer on the RCP Seal System upgrade. Modeled the thermal / hydraulics of the MU&P system which resulted in operational and design modifications to reduce RCP seal thermal stresses. Many miscellaneous projects required nodalization and dynamic modeling of large scale thermal energy systems and associated controls. Systems engineer in support of B&W's Energy Management System Program. Performed energy studies to assess the energy efficiency of industrial powerhouses, identified potential improvemerit s , and performed design and cost-benefit analyses to optimize energy consumption. He has 10 years commercial nuclear experience. Holds a B.S. degree in Mechanical Engineering from Virginia Tech and is enrolled in the Masters
- program, Engineering Administration, at George Washington University.
Charles D. Arbuthnot (Nuclear Operations Instructorl Responsible for the conduct of sir.ul ator training for the following programs: replacement operator training, senior operator upgrade training, licensed operator requalification training, shift operations technical advisor initial and requalification training, and engineer training. Additional responsibilities include: developme__nt and evaluation of performance based training programs for the simulator Served as a member of the Simulator Project Team responsible for participation in design review meetings, review of Acceptance Test procedures and conduct of the Simulator Acceptance Test Program. Currently holdt an SRO Certification on CR 3. Past experience includes the following: Florida Power Corpora
- n n Emergency Team Instructor, responsible for the conduct of trainir for Emergency Team Preparedness in the areas of medical, emergency repair, radiation emergency, fire emergency, environment survey, and accident assessment.
Florida Power Corporation Chief Nuclear Operator. In this position his duties included . assisting other licensed operators as required, providing direction and assistance to non-licensed operators in the performance of assigned tasks, providing assistance in any area of the plant as needed, and making periodic in-plant rounds to obterve operating equipment. Q
FLORIDA POWER CORPORA 110N O CRYSTAL RIVER UNIT 3 SIMULATOR CER11FICAT10N SUBM111AL Charles D. Arbuthnot (Hgglut.Aprations InstHC10rl(Continued) Florida Power Corporation Nuclear Operator and Assistant Nuclear Operator. Duties in these jobs included 'aperating all equipment controlled from the main control board area 1.e. manipulating controls, monitoring equipment and system parameters, documenting evolutions and significant events related to those systems and/or components controlled from the main control board, initiated or performed operator actions required by all procedures applicable to the main control board area. He was also responsible for initiating power reductions, reactor trips, etc. if conditions required such actions. As an Assistant Nuclear Operator he was required to assist in the operation of the plant, make rounds to observe and record operating data, and execute routine operations. Held an NRC reactor operator license on CR-3. Auxiliary Equipment Operator at Toledo Edison. Responsibilities in this position included operation of plant auxiliary equipment, and assisting the Nuclear Operator and the Assistant Nuclear Operator. Served as an Electrical Operator / Instructor in the U.S. Navy Nuclear power Program. O ne has 20 years or nucieer Power experieace iacludin9 fifteea reers commercial experience on B&W plants. fL David Demontfort (Nuclear Operations Instrug.19t). Responsible for the conduct of simulator training for the following programs: replacement operator training, senior operator upgrade training, licensed operator requalification training, shift operations technical advisor initial and requalification training, and engineer training. Additional responsibilities include: development and evaluation of performance-based training programs for the simulator. Served as a member of the Simulator Project Team responsible for participation in design review meetings, review of Acceptance Test procedures and conduct of the Simulator Acceptance Test Program. Currently holds an NRC SR0 License on CR-3. l l Past experience includes the following: Florida Power Corporation Assistant Nuclear Shift Supervisor where he was responsible for directing the Reactor Operators to ensure proper
- performance of their. duties and coordinating their activities with other Operations and plant personnel. In addition, he coordinated the operation of plant equipment and systems, surveillance testing, l and tagging operations. He supervised the Auxiliary Operators in the performance of their duties and performed the duties of the Shift Supervisor in his absence. Also acted as Fire Team Leader. Assisted in revision / maintenance / development of plant Emergency Operating O- Procedures.
FLORIDA POWER CORPORA 110N O cavsT^t. nivca u"i' 3 sinut^ ton CERIIfICA110N SUBM11TAl. R. David Demontfort (Nuclear Operations InstrE12t)(Continued) Florida Power Corporation Nuclear Operator and Assistant Nuclear Operator. . Duties in these jobs included operating all equipment controlled from the main control board area i.e. manipulating controls, monitoring equipment and system parameters, documenting evolutions and significant events related to those systems and/or components controlled from the main control board. Initiated or performed operator actions required by all procedures applicable to the main control board area. He was also responsible for initiating ; power reductions, reactor trips, etc. if conditions required such I actions. As an Assistant Nuclear Operator he was required to assist in the operation of the plant, make rounds to observe and record operating data, and execute routine operations, i l' Florida Power Corporation Nuclear Auxiliary Operator and Assistant Nuclear Auxiliary Operator. Responsibilities in these positions . required that he operate the plant auxiliary equipment, and assist I the Nuclear Operator and the Assistant Nuclear Operator. As an Assistant Nuclear Auxiliary Operator he functioned as an assistant to the Nuclear Auxiliary Operator, performing work functions dictated by plant conditions, made rounds to observe and record operating data, and executed routine operations. (q.) ' He has 10.5 years of commercial nuclear power experience. He received an Associates Degree from Central florida Community College and is working towards his BS from Thomas Edison College. tLalcom P. Holmes (Nuclear Oprations Jmtructor) Responsible for the conduct of simulator training for the following programs: replacement operator training, senior operator upgrado training, licensed operator requalification training, shift operations technical advisor initial and requalification training, and engineer training. Additional responsibilities include: development and evaluation of performance-based training programs for the simulator. Served as a member of the Simulator Project Team responsible for participation in design review meetings, review of Acceptance Test procedures and conduct of the Simulator Acceptance Test Program. Currently holds an SRO Certification on CR 3. Past experience includes the following: Florida Power Corporation Nuclear Operations Instructor responsible
.for the preparation and conduct of Initial and Requalification Operations Training programs for Licensed and Non licensed personnel.
He also participated in all phases of the CR-3 initial INPO accreditation effort. O
It0RIDA POWER CORPORA 110N O cavstat nivtn uni, 3 SinutA40n CER11 FICA 110N SuBnllTAL tiplqom P. Holmes (Nuclear OptriticatJnuruciprJ(continued) Babcock & Wilcox Senior Engineer. Participated in developing the initial draft of the Bellefonte Nuclear Plant (1VA) operating procedures. This work was performed at the Bellefonte plant site as part of-the Master Services contract with Babcock & Wilcox. florida Power Corporation Assistant Shift Supervisor. Ilis < responsibilities in this position included the direction of reactor operators to ensure proper performance of their duties, coordination of the activities of the reactor operators with other Operations and plant personnel, coordination of the operation of plant equipment and systems, surveillance testing, and tagging operations, lie also supervised the auxiliary operators in the performante of their duties. In addition, he performed the duties of the Shif t Supervisor in his absence and performed other duties as delegated by the Shift , Supervisor. Also acted as fire 1e m Leader. Held an NRC SRO License on CR 3. florida Power Corporation Nuclear Operator, ile participated in all phases of the CR 3 startup including llot functional Testing, Low . Power Physics Testing, and initial power escalation. O Served in the u. S. Navy where he x n4eved a re u ag of n x ninis u a e first Class (E6). , He has'18 years of commercial nuclear power experience on B&W plants. (b2np_bf E. Smith (Nuclear Operations inttructor) Responsible for the conduct of simulator training for the following programs: replacement operator training, senior operator upgrade training, licensed operator requalification training, shift operations technical advisor initial and requalification training, and engineer training. Additional responsibilities include: development and evaluation of performance-based training programs for the simulator. Served as a member ; of the Simulator Project TL..1 responsible for participation in design review meetings, review of Acceptance Test procedures and conduct of the Simulator Acceptance Test Program. Currently holds an SR0 Certification on CR-3. Past experience includes the following: Florida Power Corporation Nuclear Operations Instructor responsible for_ the preparation and conduct of Initial and Requalification Operations Training programs for Licensed and Non-licensed personnel. O.
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TLORIDA POWER CORPORATION Q CRYSTAL RIVER UNIT 3 SIMULA10R C[RilflCA110N SUBMITlAL i T D.QD114 fu_hillL@lCICE_0R9I1119ni_lu11rRted(Continued) florida Power Corporation Nuclear Operator. Duties in this jcb included operating all equipment controlled from the main control board area i.e. manipulating controls, monitoring equipment and system parameters, documenting evolutions and significant events related to those systems and/or components controlled from the main control board. Initiated or performed operator actions required by all procedures applicable to the main control board area. He was diso responsible for Mitiating power reductions, reactor trips, etc. if conditions required such actions. Held an NRC R0 license on CR 3. Louisiana Power & Light Company Chief Nuclear Operator. In this position his duties included assisting other licensed operators as required, providing direction and assistance to non licensed operators in the performance of assigned tasks, providing assistance l in any area of -the plant as needed, and making periodic in-plant rounds -to observe operating equipment. Held an NRC R0 license on Waterford. Served in the " 5. Navy auclear power program. He has 20 years of nuclev y ret operience, including fourteen years of p/ (. commercial nuclear experkt.a un lif,W and CE PWRs. He has earned 55 credits towards a degrt.e in Geology. JAmJts._LEBtiARMenior Simulator Enginced Senior Sin"il ator Systems Engineer responsible for the maintenance, todificrt+9n, and documentation of-the simulation software program for the Crystal Riv r Unit-3 Power Plant Simulator System, including supervision and coordination of simulator software engineering activities to ensJre conpliMce with scope of design, budgetary constraints, and all regulatory comnitments; performance of sof tware development / enhancement projects in the areas of reactor neutronics and thermal hydraulics, plant electrical distribution system, primary support systems, and process computer simulation; computer system administration; and performance of plant design modification revi N for simulator software impact. Past experience includes the following: florida Power Corporation $1mulator Systems Engineer, in this position, his responsibilities included the review, maintenance, and modification of simulation sof tware for the Crystal River Unit Power Plant Simulator System. Duties included the review of interim sof tware documentation for design scope and technical accuracy; participation in the Simulator factory Acceptance Testing including Q development-and performance of the Reactor Core Physics Acceptance
ft0RIDA POW [R CORPORA 110!4 O CRYSTAL RIVER UNil 3 SIMULATOR CER11FICAT10N SUBMITIAL
@ mas L. O!BditL(ignior SimulalonInsitiged(Continued) lest Procedure; development of Simulator Operating Instructions; and utility software development which includes a Dispatcher ConfiWion Display Utility, an-Initial Conditions Editor, and i@ e L a ton of the Reactor Core Coupling and Perturbation (dPJ% m hility.
Cy A V M if And Light Company, Senior Specialist. Simulator Sj4 W L- Ly f osition r3 quired the maintenance and modification of com m M tuare for the Shearon liarris Nuclear Power Plant Simulator system O.:luding the supervision of all sof tware activities i.e. implementation and maintenance of configuration control procedures and verification of mathematical model changes; system i administration including operating system maintenance and ; modification activities; upgrade of simulator logic sof tware and ! mathematical models to incorporate plant design chantes including l control rod drive system logic and reactor neutronics modifications . due to plant control bank redesign; installation of Radiation Monitoring System control panels including a complete restructure of the _ simulator input / output system and development of interface software; and general upgrades as requested to enhance the simulator i capabilities such as nodal restructure of the two-phase reactor . O coolaat system for improved naturai circuiet4en response, restructure of the reactor core model for improved fuel cycle response, and the installation of an automated reactor core cycle initialization and i physics testing module for core refueling. i Carolina Power And Light Company Simulator Specialist. Duties in i this position included operational support and testing of the Shearon Harris Nuclear-Power Plant-Simulator, i.e. the review, approval, and performance of Simulator factory Acceptance Tests during procurement; ! development and implementation of software for a computerized simulator database _ tracking system for plant- design changes and simulator service requests; performance and tracking of plant design change analyses for simulator applicability; and software modification for simulator upgrado activities including mathematical models and instructor system programs. Carolina Power And Light Company, Specialist, Operator Training, in , this position he served as a PWR Simulator Instructor interfacing > primarily with the Shearon Harris Cold License Certification Program. Responsibilities were simulator training program development > i including lesson plan development for- the Cold -License - Program; l simulator instruction during cold license training; and classroom instruction duties in Heat Transfer, fluid flow, Math, Chemistry Radiation Protection, Nuclear and Reactor Theory. O , b-l
l lLORIDA POW [R CORPORA 110N O CRYS 1AL RIVER UNil 3 SIMULA10R CER11 FICA 110N SUBM111AL ! i LaMi L. 0'Brinn_digipr_lhyhtgr_Lnghur)(Continued) 1 Carolira Power And Light Compat,y , Junior Specialist, Operator Training where he was the BWR Simulator Instructor interfacing primarily with the initial training development- during simulator procurement. This involved initial training aids development for the l simulator including lesson plan developmenti preliminary functional testing of the Brunswick Steam Electric Plant Simulatort and classroom instruction and course development in Heat Transfer, fluid flow. Math, Radiation Protection. Chemistry, DWR Systems,and BWR 1ransient Analysis. North Carolina State University, licensed Reactor Operator, Nuclear Reactor Pi ogram. This position required reactor operations and ! control board instruction during power plant operator training in the Reactor Operator 1 raining Program and reactor operations during experimentation and sample irradiation activities. He received his B.S.in Nuclear Engineering from North Carolina State University in 1981. He has 10 yrws of commercial nuclear power experience. ECILY_la zRose -(ilmJhtstJngingtt) O ResPonsibie for maintenence of modeiin9 sof tware for various systems on the CR-3 Simulator. These systems include the Reactor Coolant System (RCS), the i Once Through Steam Generator (01SG), the Reactor Butiding (Containment), the Waste Gas. System, the Instructor facility (1/f), the Safety Parameter Display System (SPOS) and the Main Steam Relief valve Monitor. Past experience includes the following: CAL Electronics Ltd., Montreal, Quebec, Technical Staff. He was one of a three man team which developed CAE's advanced, real-time, two-phase thermo-hydraulic code. This code has now been applied to the Reactor Coolant System (RCS) and the Once lhrough Steam Generator (0150) on the CR 3, Davis Besse and Rancho Seco. Pressurized Water Reactors (PWR) Nuclear Power Plant Training Simulators. Responsible as system leader for implementing this code in modeling the RCS for one of the above mentioned plants (Rancho Seca), as well as for
- integrating the RCS with the 01SG, the main steam and the reactor core models. Responsible for developing CAE's novel, 2-0 steam i tables using boundary-fitted coordinate techniques. He has also done control and logic modules for reactor shut-down system and moderator cover-gas system on CAE's Gentilly 2 (CANDU) simulator.
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i FLORIDA POWER CORPORA 110N Q CRYSTAL RIVER UNil 3 SIMULATOR CER11 FICA 110N SUBM111AL , Perr_v L u Rqi d lim & tgr_ A gi n tt)(Continued) McGill University, Montreal, Quebec. Developed software components for McGill's engineering plotting package for IBM PC's and compatibles. Specifically, developed assembly language (Microsof t) routines to manipulate the PC video hardware and directly address the . bit mapped memory of the graphic screen, as well as FORTRAN routines : to manage fonts and draw text on the screen. McGill University, Montreal, Quebec, Summer Research Assistant. ! Constructed a 30 finite-difference computer code for convection diffusion problems. Generated numerical solution to problems in heat transfer and bioengineering, He has 5 years experience in real-time modeling of advanced thermal / hydraulic codes for training simulator applications. He received a Bachelor of Engineering - Mechanical Degree with a Minor in Computer Science from McGill University, Montreal, Quebec, graduating as " University Scholar". He is presently pursuing his Master of Engineering Mechanical Degree from McGill University, Montreal, Quebec, with an expected completion date of 1992. His Thesis topic is "An Experimental and Numerical Study of Natural Convection in Ventilated Enclosures". O ne was awarded numerous s m iarsnins ina uding NSERC Postgraduate Scholarship, TCAC Postgraduate Scholarship, John Bonsall Porter Scholarship, J. W. McConnell Scholarship, Iron Ore Company of Canada Scholarship, and NSERC Undergraduate Summer Research Award. He received 1he British Association Medal for Great Distinction in Mechanical Engineering for graduating second in a class of 111. He has written various computer programs including: 1-0, 2 0 and 3-D finite differ 7nce solutions to conduction, convection and fluid 4 flow; boundary fitted coordinate grid generation; plotting package; function plotting; contour plotting; 3 D wirc+ mesh graphics with hidden line removal; pop-up window screen management package. Also he is co author of a paper entitled " Experimental and Numerical Investigation of Heat Transfer in Polymethylmethacrylate Bone coment During Exothermic Polymerization", which was presented at the ASME Winter Annual Meeting, November, 1983. Richard J. Gtymbir_,L$h&tpr_ Enginer.) Responsible for maintenance of modeling software for various systems on the Crystal River Unit 3 simulator. These systems include the integrated Control System (ICS), Reactor Protection System (RPS) Emergency feedwater Initiation & Control (EFIC), Engineered Safeguards (ES), Electrohydraulic Control System (EHC), Control Rod Drive (CRD), Non Nuclear instrumentation ' (NNI), and all common services model (includes Service Water. Raw Water, Demineralized Water, etc.). In addition to these responsibilities he is
FLORIDA P0KFh CORP 0RA110N Cavsi^t aitca uait 3 sinu'^2oa O CER11rlCA110N SUBM111AL Richard 6_Stg@ltJ1LnMlB19r_En91Dtet)(Continued) also responsible for foreground maintenance and Common Data Base Updates. He is currently a member of the B&W Owner's Group Advanced Control System Task force. Past Experience includes the following: CAE Electronics, Ltd., Montreal, Quebec, Canada, Integration Team Member. Member of the Toledo Edison Simulator Integration Team, responsible for the control systems integration (ICS, RPS, CRD, SFRCS, ARTS, & NNI). CAE Electronics, Ltd., Montreal, Quebec, Canada, Acting Group Leader. Responsible for the code development and model designs for the Controls Group at CAE for the Toledo Edison (Davis-Besse), florida Power (CR-3), and SMUD (Rancho Seco) simulators. Responsibilities included serving as CAE's Engineering representative at Design Review meetings, assisting junior engineers in the development of their software models from conception to integration, and allocation and estimation of manhours for specific tasks and projects associated with the Controls Group, i O Cu Eixtronics, tid., entreai, Quebu, Canada, nodeler. Deveioped and implemented CAE's models of the ICS & NNI for B&W simulator l applications, in addition, developed, coded, and integrated the Moderator system for the Gentilly 11 CANOV simulator. He has 6 years experience in real-time modeling of plant control and l thermal / hydraulic systems for training simulator applications.in He holds a B.S. degree in Mechanical Engineering from Concordia University in Montreal Canada. Ronald D. TytigEperations Department) System Engineer responsible for engineered safeguards system including maintenance and revision of ES testing procedures and refueling integrated testing. Duties also include providing technical support to Maintenance, Operations and Training departments for ES actuation-related issues. i Past experience includes the following: Florida Power Corporation Assistant Nuclear Shif t Supervisor where he was responsible .for directing Reactor Operators to ensure proper performance of their duties and coordin; ting their activities with other Operations and plant personnel. In addition, he coordinated the operation of plant equipment and systems, surveillance testing, and tagging operations. He supervised the Auxiliary Operators in the performance of their duties and performed the duties of the Shif t i
'Q Supervisor in his absence. Also acted as fire Team Leader.
fl0RlDA p0WIR CORP 0RA110N O cavsi^t RivfR uNil 3 Sinuts10R CfR11IICATION SuBMllTAL BonalLDJyrlL10 Peta 1193LDe p3rLmeni) ( c o n t i n u e d ) florida power Corporation Chief Nuclear Operator, in this position his duties included assisting other licensed operators as required, providing direction and assistance to non licensed operators in the performance of assigned tasks, providing assistance in any area of the plant as needed, and making periodic in plant rounds to observe operating equipment. L liorida power Corporation Nuclear Operator and Assistant Nuclear Operator. Duties in these jobs included operating all equipment controlled from the main control board area 1.e. manipulating controls, monitoring equipment and system parameters, documenting evolutions and significant events related to those systems and/or components controlleo from the main control board. Initiated or performed operator actions required by all procedures applicable to the main control board area. He was also responsible for initiating power raductions, reactor trips, etc. if conditions required such actions. As an Assistant Nuclear Operator he was required to assist in the operation of the plant , make rounds to observe and record operating data, and execute routine operations. Florida Power Corporation Nuclear Auxiliary Operator and Assistant O Nuclear Auxiliary Operator. Responsibilities in these positions required that he operate the plant auxiliary equipment, and assist the Nuclear Operator and the Assistant Nuclear Operator. As an Assistant Nuclear Auxiliary Operator he functioned as an assistant to the Nuclear Auxiliary Operator, performing work f unctions dictated by plant conditions, made rounds to observe and record operating data, and executed routine operations. He has eleven (11) years of commercial nuclear power experience. He currently holds an NRC SRO license on CR 3, Garrett P. Hebb (tluqlear Shitt_$up_grylior) Nuclear Shift Supervisor, responsible for all activities of the operating shift. This responsibility, normally carried out from the Control Center, is met through direct command of the operating shif t and ongoing review of operations, maintenance, and support functions. Past experience includes the following: florida Power Corporation Assistant Nuclear Shif t Supervisor where he was responsible for directina the Reactor Operators to ensure proper performance of their duties and coordinating their activities with other Operations and plant personnel. In addition, he coordinated the operation of plant equipment and systems, surveillance testing, and tagging operations. He supervised the Auxiliary Operators in the Q
FLORIDA POWER CORP 0RAll0N I O CRYSTAL RIVER UNil 3 SIMULATOR CERilflCATION SUBH111AL Dittelt P. HeML(kglp3r ShifLivneniipfj(Continued) performance of their duties and performed the duties of the Shift Supervisor in his absence. Also acted as fire leam leader. Florida Power Corporation Nuclear Operator and Assistant Nuclear Operator. Duties in these jobs included operating all equipment controlled from the main control board area 1.e. manipulating controls, monitoring equipment and system parameters, documenting evolutions and significant events related to those systems and/or components controlled from the main control board. Initiated or performed operator actions required by all procedures applicable to ; the main control board area. He was also responsible for initiating
~
power reductions, reactor trips, etc. If conditions required such actions. As an Assistant Nuclear Operator he was required to assist in the operation of the plant, make rounds to observe and record operating _ data, and execute-routine operations. Vermont Yankee Nuclear power Corporation, Vernon, Vermont. Ills duties included operation of all equipmont exterior to the main control room, assisting in all testing of plant systems during - construction and power escalation, assisting in fuel testing and replacement, and writing procedures for operation of the reactor O auxiliary and secondary equipment. Served in the U.S. Navy Nuclear Power program as a Reactor Operator and Engineering Watch Supervisor. He has 26 years of nuclear experience. including 20 years of commercial experience on B&W and GE plants, lie holds an NRC SRO License on CR 3. lie received his BS Degree in Applied Science and Technology from Thomas A. Edison State College. O n
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4 O O O Page No. 2 OPEN PMYSICAL FIDELITf DISCREPA4CIES 09/09/91 RESOLUT!ou IMPACT cmp _DATE S!M,iD ' CATAGORf DESC' DR_WUM PLANT _1D
%081002-11 / /
20 ESF- (RIP) A141M20 SCALE SCALE OtFFERENCE SCALE DIFFEREu2 , FUNCTION LABEL SMOULD BE M 081002 11 / / 21 ESF- (RIP) A1A1M21 SCALE,LASEL ELAME.
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.o o ;o Page No. 4 OPEN *HYSICAL f!DELITY DISCREPANCIES 09/09/91
'$1M ID CATACORY DESC RESOLF ION 1MFACT CCP*P- DATE DR_NUM PLAMT_10 58 ESF- (RIP) A1A1M59" SCALE,LAEEL SCALE & FUNCT. LABEL'$HOULD BE BLAhK, TAG 8408:002-11 89102301A 32050320 / /
LASEL..tMAR82050320$ 59 ESF- (RIP) A1A1M60 SCALE, LABEL SCALE & FUNCT. LABELS SMOULD BE BLANK, TAG 84081002-11 89102301A / / LABEL..(82050320) 60 ESF- (RIP) A1A1M61 . SCALE SCALE DIFFERENCE 84081002-11 / / 61 ES-A-KW1(1-5,6) A1A2A1 (1-5,6) UBEL ANNUNCIATOR T!LE INCORRECT. 88052401 / 4
^
62 ES A-KW1(2-2) A1A2A1(2-2) LABEL ANNUNCIATOE TILE INCORRECT. 88052401 / / 63 ES-A-KV1(2-3,4) A1A2A1(2-3,4)' . LABEL AlsNUNC!ATOR TILE INCORRECT. 88052401 / / 64 ES-A-KW1(2-5,6) A1 A2A1(2- 5,6) ' LABEL ANNUNCIATOR TILE IM.ORRECT 88052401 / / (5 ES-A-KW2(3-6) A1A2A2(3 6) LABEL ANNUNCI A"A TILE INCORRECT. 88080101 / / 66 ES-A-KW2(7-6) A1A2A2(7-6) label ANNUNCIATOR TILE INCORRECT. 88080101 / / 69 ESF- A1A2DS'*9... LOC, ABS,LA8, 6 BACKLITED ES WINDOWS ADDED,SOME MOVED, LABEL 88052401 / / I ADCED,WIMIC... 70 ESA- A1A2DSSI MARKING ES BACKLITED WINDOW CMANGED TO BLANK. 88352401 / / 71 ESA- A1 AMS122,DS123 ABS, LAB,MIM MISSING BACKLITED ES WINDOW,... tABEL & M!MIC 88052401 / / TO BE ADDED. t 72 ESF- A1 A2DS142,DS177 ABS, MARKING 1 ES BACKLITED WINDOW ADDED & MARr1NCS 88052401 / / f t
. CHANGED ON DS177.
i 73 ESF-A A1A2M8 SCALES SCALE INCREMENTS DIFFERENT (BOT 8: SCALES). SNAG-06236 TR-00031 / / ; 74 ESF-A A1A2M9 & M10 MARK 1NG TECH SPEC LIMITS HI-LITED 'A. WiM, NL*BERS 82050321 ./ / I LARGER ON M9. 75 ESF-A A1 A2DS575,DS576 MARKING ES BACKLITED WINDOW ENGRAVINGS / MARKINGS 87011931-02 / / 1 CHANGED. 76 ESF-A A1A2DS377,DS436 MARKING ES BACKLITED WINDOW EWGRAVED INCORRECTLY. 87011901-02 /./. 77 ESF-A A1A2DS597 ABS, MARKING ES BACKLITED WINDOW ENG8AVED INCORRECTLY. ONE 88052401 / / TO BE .CCCD. 78 ESF-A A1A2DS$34,DS536 EXTRA (ALSO S23)/SV & IND. LITES REMOVED: 84081018 / ./ COVERPLATE ADDED.
O- .O O Page No. 5- OPEN OMYSl(.AL FIDELITY DISCREPANCIES i 09/09/91 DR_WUM PLANT ,1D $1M,,ID CA1AG0ftY CsESC RESOLil7104 IMPACT ' COMP,DATE 82 ESF-A A1A2M25 & M26 DEVICE BOTH 'RY' thD. MAVE BEEN REPLACED BY 82050320 /, / IETERNATIONAL INST ... I 84 .ES-B-JH1(1-5,6) A1A2A4(1-5,6) MAR (!ss A=wusCI AT0ft DISCREPAsCY. 88052401 /, / 85 ES-B-JM1(2-2) A1A2A4(2-2) MADEING AWNUNCIATOR TILE DISCREPAECY. 88052401 /I / 86 E S-B-J N1(2-3,4) A1A2A&t2-3,4) Ma#L!wG AtWUvCIATOR T!LE DISCRCPAwCY. 88052401' //
'87 ES-B-JM1(2-5) A1A2A4(2-5) MARK!kG ANGUNCIATOR TILE DISCREPA4CY. 88052LJ1 //
88 ES-B-JH2(3-6) A1A2A5(3-6) MARKING A44UwC1ATOR TILE DISCREPAuCY. 88080101 // 89 ES-B-JN2(4-6) .A1A2A5(&-6) MARCING ANNUMCIATOR TILE DISCREPAtCY (PARENTPESIS TR-0001fs // MISSING.) 90 ES-B-J'12(5-6) A1A2A5(5-6) MARK 1=G AwwueCI A!DR TILE DISCREPAhCV. (et!SSIMG TR-C00*8 / '/ PARENTNESIS). 91 E S-B -J H2(7-6) A1A2A5(7-6) MARrtwG AWNUNCIATOR TILE DISCREPANCY. 88080101 // 92 ES*-B A1A2DS114,... LOC, ABS, LAB, 6 ES BKLITED WINDOWS ADDED,50K MOVED, LABEL & 88052401 // MIMIC ADDED. 93 ESF-8 ATA20S127,DS12S ABS. LAB,MIM. MISS1hG 1 ES BKL11ED WINDOW, DS127 SHOULD EE 88052401 // BLAT <,LBLS& MIN 94 ESF-8 A1A2DS157 ASSENT ES BitLITED W1EDOW MISS14G, >=wE .. MEW 88052401 // EWGRAV!hG. 97 RM C29-R!R,. . A6A2A6 !* -
'8SE N T MISSINO 2 STRIP CFART RECOPDERS. (am-G29 RIR 84C81018 //
& Wi-015-tIR) 98 SS-141-CIR A/ A2A6 uABEL LABEL POSITitus Ih^0RRECT AND 8.4GR AVING 84081018 //
D1FFERENT. 99 BS-93-PIR 'T? A6A2A9 LASEL LASEL D1FFEREECF (BI-93-PIR OR BS-065-PR ???) 84081018-01 // 107 PSA-Y-6-4 A2A2A2 ARCI NG WINDOW EwGRAV NG Of 0CPE>AeCY. S8050401 87100301-01 // 103 PSA-x-6-8 A2A2A1 MARE!WG WINDOW ENGRAVlWG DISCREPA4CY. 84081001-20' // 123 FW-39 A1A2S280 EXTRA SWITCel,LA*PS, ESC. PLATE & LASEL REPLACED BY 86060005 // COVE R. 134 LRV-71-SV A1A25270 MAR (twG ESCLTTCHEON PLATE DISCREPAeCY, ESC //
t O O O Page No. 6 idEN PHYSICAL FIDELITY DISCCEPANC!ES 09/09/91 CATAGORY DESC FESOLUTION IMPACT COMP _DATE DR_NUM PLANT _10 SIM_ID 135 LRV-73-SV A1A25271 LASEL PLANT 'V3DP-6 BKR5', SIMutATOR *VBDP-4 BtR33' 87011901 '/ / 139 A1A2545 LABEL PLANT 'HPI SEAL IN RESET', $1M. *BSP-18 AUTO ' 88052401 / / START PERMIT..' 158 A1A23240 ENGRAVING ESCUTCHEON PLATE DISCREPANCY. ESC /. / 167 FWV-40 A1A25259 EXTRA DEVICE SW.,2 LAMPS, ESC.PL & LABEL REMOVE & REPLACED 86060905 / / BY COVER PLATE. 178 ANF-1A #1A25231 ENGRAVING ENGRAVINE ON COVER PLATE DIFFERENCES. ESC / / 180 A1A2S253 LASEL PLANT 'VBOP-5 BER 14', $1MULATOR 'VBDP-3 BKR 87011901 /-/ 33' 184 A1A2S44 LABEL CHANGE TO "HPI SE AL IN REST", WAS "BSP-1 A 88052401 / / AUTO START PERM.." 190 A1A2M29/M30 DEVICE NPI FLOW INDICATORS ARE DIFTERENT. NEW ONES 82050320 / / ARE DUAL INO. 191 A1A2DS549,557,. EXTRA (ALSO $28) THESE ITEMS HAVE BEEN REPLACED BT 84081018 / / COVERPLATES. 193 A1A2DS462 MARKING PLANT " LEV-71:RS: PURGE:..." /SIM. 87011901-02 / /
*AHV-44:RS:PUEGE:..."
194 9012A A1A2DS403 MARKING PLANT "LRV-73:RB: PURGE :..." /SIM. 87011901-G2 / /
"AHV-40:RB:PURCE:..."
196 ES STATUS LITES A1A2DS599 MADE!NG PLANT "HPI: SEAL:1N* /S!M. 88052401 / /
" AUTO: START:PEPMIT:BSP-18" 197 ES STATUS LITFS A1A2DS579/580 NARKING PL ANT "ANV-1 A RB PURGE..." /SIM. "tRV-71 RB 87011901-02 / / ;
PURGE..." 199 CF-1-Pf2 ?-Lil, A1A2M19 SCALE SCALING NUMBER SITE IS SIGNIFICANTLY LARGER 82050321-01 / / IN THE PLANT. 201 BS-91-PI A1A2M18 SCALE SCALING INCDeSISTENT: PLANT "O,25,50..." SNAG 06236 / /
/SIM. "O,5,10...."
202 ES P t A1A2M18 SCALE SCALING ikCONSISTENT: PLANT "...,50,60,70" SNAG-06236 / /'
/SIM. "..,50,70",
205 MSV STAT /TST LT AZA2DS1-DS6 DEVICE TFE 6 LENSES ARE BLUE AND SHOULD BE WHITE. 84081018 / / , 219 RCV-11 A3A2536 ENGR AVING -ESCUTCHEON PLATE DISCREPA=CY. ESC / / 221 SP-18-L11 'AIA2M19 SCALE PLANT IS C-4,1-4,2-4... /SIM. IS 0,2,3,... . 59061901 / / ALSO UNITS...
O O OL Page No. 7 OPEN PHYSICAL ~1DELITY DISCREPANCIES 09/09/91 DR_NUM PLANT _ID SIM_ID CATAGORY DESC RESOLUTION IMPACT COMP _CATE 222 SP-3B-TI A3A2M20 SCALE PLANT HAS tMITS AT TOP, SIM. DISPLAYS UNITS 89061901 / / AT THE 30TTG4. 228 SP-1A-Lf1 A3A2M18 SCALING tmITS ABOVE METER MISSING, ALSO SCALING IS 89061901 / / - WRONG. 232 RC-1-t!2 A3A2M25 SCALING PLANT IS 0,100,200,320 & THE SIMULATOR IS SCA / / . 0,10,20,32 x10. I 235 ICS-CY2-1-3 A3A2A2 MARKlWG WINDOW ENGRAVING DISCREPANCY. 84062702 / / t I 1 236 ICS CY2-3-3 A3AZA2 MARKING WINDOW ENGRAVING DISCREPANCY. 84060702 / / 237 ICS-CY2-1-2 A3A2A2 MARKING WINDOW ENGRAVING DISCREPANCY. 84060702 / / 2 238 1CS-CY2-1-1 A3A2A2 MARK 1NG WINDOW ENGRAVING DISCREPANCY. 1K060702 / / , 261 CD-100 & 101-L1 A2A2M37 SCALING THE SCALING ON THESE 2 METERS ARE INCORRECT. 88040501-01 / / 289 PSA-2-2-6 A2A2A3 MARKING PLANT " SUMP LEVEL:HIGH" /SIM. " - BLANK " 8?O31601 / / , 290 PSA-2-7-1 A2A2A3 MARKING PL ANT "EF T ANK: LEVEL: LOW-LOW" /SIM. "EMER TR-0CO22 84081001-20 / / , FW: TANK: LEVEL L.." I 291 PSA-2-6-2 A2A2A3 MARK!bG PLANT ".. ..." /SIM. "... !$0L ..." 84081001-20 / / 292 PSA-2-2-8 A2A2A3 MARKING PL ANT "SEC SAMPLE...: ALARM" /SIM. "SEC 34081001-20 / / .! SAMPLE...: TROUBLE" , 299 RC-201-Lt.L11, A2A?M8 BELOW ABSENT DEV. METER MISE 3hG AND ASSOCIATED LABELS 87072803 87072302 / / i 300 MSRV MONITOR A2A1 DEVICE THE SONY MONiiOR FOR THE MSRV CAMERAS IS 82G50316 / / MISSING. 301 NI-014/015-41-1 A3A2M36/M41 SCALE SCALING ON THIS DEVICE IS D1FFERENT, BOTH SCA / / INDICATORS. i 308 SP-18-MS2 A3A2568 ENGRAVING PLANT "CHAhWEL SELECT:SP-18-MS2* /SIM. ESC 84081018 / /
*RCSG-1B START LEVEL.
312 RCP-1D (ABCVE) A3A2DS260 DEVICE 'PZR HTR LEVEL AUN RELAY PWR FAIL LAMP ' 88110501 / /
' REMOVED TROM PLANT.
322 SP-8A A3A2S46 EhGRAVING PLANT "...:SP-8A" /SIM. "...: SP-BA-MS" ESC / / i 323 SP-83 A3A2S51 ENGRAVING PLANT "... esp-88" /SIM. "...: SP-88-MS" ESC / / 6 w
+s ,
O O O Page No. 8 OPEN PHTSICAL FIDELITY DISCREPAkCIES 09/09/91 DR,WUM PLANT _ID S!M ,lD CATAGORT .DESC RESOLUTION IMPACT COMP,DATE i
?
327 FWV-47 A3A2S3 ENGRAVING PLANT: "FWP-1 A RECIRC:FWV-47: /SIM:. "FWSP A ESC / /- RECI RC: F WV-47: . . 331 ICS-CY4-4-5 A3A2A4 MARKING PLANT "FWP: SEAL POT LEVEL:HIGH" /SIM. 34081001-21 84081001-20 / /
" . . . L E VE L : LOW" 332 !CS-CY4-3-1 A3A2A4 MARKING PLANT: "FWBP A: TRIP" /SIM: "FWPB A: TRIP" TR-00022 / /
333 ICS-CY3-6-6 A3A2A3 MARr!NG PLANT "...:NOT FULL CLSD" /SIM: "...:NOT Futt $4081001-20 / / CLOSED" 334 RC-003A-PIR2 A3A2A29 SCALE SIM.NAS "PSIG X 1000" FOR UNITS, PLANT. ONLT 8408101C-12 / / HAS "K" ABOVE 2.5 343 RC-003A-PIR1 A3A2A20. SCALE PLANT "K", /SIM "X 1000" - S!M. UNITS *PSIG", 84051010-12 / / PLANT DOES NOT ; 352 RB PURGE SUPPLY A7R1 DEVICE PLANT POT SCALE 0-10 (BLACK), SIM. IS 0-0 AND 84081010-20 / / SILVER. . 356 P/A0705 A7577 LABEL LABEL ABSENT AND PANEL MARKING IS INCORRECT. LAB ESC MIM / / 357 AHF-21A A7535 LABEL PLANT LABEL FOR AHF-21 A DIFFERS FROM ESC / / SIMULATOR LABEL. 362 VIBRA SW RESET A7570 LABEL SWITCH LABEL DISCREPANCY. ESC / / 363 Aku-2 A7DS272 LABEL LABEL DISCPEPANCY. ESC / / 365 ANF-219 A7550 LABEL PLANT LABEL FDP AHF-21B DIFFERS FROM TME ESC / / , SIMULATOR LABEL. 367 3ESBR/ RESET A7582 LABEL CONTROL COMPLEX HVAC ISOLATE /kESET LABEL ESC LA8 / / ABSENT FROM SIM. 368 RC-203-J1 AS DEVICE INDICATOR RC-203-J1 IS ABSENT FROM THE 82050317 / / ; SIMULATDR. 369 RC-204-J1 A8 DEVICE INDICATOR RC-204-J1 IS ABSENT FROM THE 82050317 / / SI M'JLA T OR . 3 70 SSF ASA357/8/11/21 EXTRA TWITCHES & LAMPS FOR 1691 & 1692 BKR TRIP 89081103 /-/ COIL, LABELS,COVTR, 371 SSF ASA2DS10/20 ABSENT GREEN & RED TRIP /CLOSE LAMPS FOR SKRS 1691 & 89081103 / / 1692 SEPARATE.. 373 $$F ASA2 NIMIC START UP TRANS. MIMIC MOVED LEFT IN PLANT..., 89081103 / /
" BUS A".. MIMIC 374 SSF ASA2 ABSENT BKR 4900 &4902 TRIP /CLOSE SW. LAMPS & SYNCH. 89081103 / /
SW. ASSENT SIM. ; w, .4 .
O O O' Page No. 9 OPEN PHYSICAL FIDELITT DISCREPANCIES 09/09/91 S!M_tD JCATAGORY DESC RESOLUTION IMPACT ' COMP _DATE DR_NUM PLANT _10 .
-375 SSF A5A2M29/30 *AARKING LAB ' RED LINE MARE FOR DIESEL CEN..., MAX's 8801T201-10 / /
LABELS..., LABELS... 376 SSF ASA2 MIMIC OFFSITE POWER SUPPLY MIMIC & LABEL MIS $1NG ON 89081103 / /. PLANT... 377 SSF / iA2 MIMIC MIMIC OF BUS CONNECTING TO STARTUP 89031103 / / TRANSFORMER MOVED TO LEFT 380 SSF A5A2 ABSENT MINUTE COUNTERS *EDG-3A&B ELAPSED TIME ABOVE 88051501 88051501-02 / / 3250 KW", ... 384 SSF ASA2DS70 LABEL LABEL ON PLANT READS " DIESEL GEN A/B 84081018 / -- / CROSS-TIE ENABLE", ... 385 SSF ASA2 ABSENT AMBER LAMP AND
- BLOCK CLOSICS ACTUATED 3209" 84081018 / /
'(&3210) NOT ...
387 SSF A5A2 ABSENT BKR 3321 TRIP /CLOSE SWITCH, GREEN! RED LAMP 88110501 / / AC BUS MIMIC. .. 390 NVR A6A1A1 ABSENT, SCALE BOTTCM SCALE ( A&B) 04 RECORDER RC-154-PR/TP SNAG-05547 / / MISSING ON SIM. 395 HVR A6A1A5 ' LABEL RECORDER RC-19A-PR1 MAS TYPE WRITTEk PAPER 87091303 / / LABEL ... 396 HVR A6A1A6 LABEL RECORDER RC-198-PR1 NAS TYPE WRITTEN PAPER 87091303 / / LABEL ... 37' HVR A6A1A7 LASEL RECORDER RC-19A-PR2 NAS TYPEWRITTEN PAPER 87091303 / / LABEL ... 398 Hva A6A1A8 LABEL EEL & DER RC-198-PR2 MAS TYPEWRITTEN PAPER 87091303 / / LABEL ... 433 1CS-21-MCS A3A2A53 DEVICE JE TEMP. SET Po1NT POT HAS AN ALLM!Ntpa COVER JLS / / OVER IT. 435 FWM-1A A3A2592 ENGRAVING PLANT "FWM 1A:lJRB A:..." /SIM. "FWM-1A: ESC S4081018 / /
-BLANK- : ...*
439 FWV-37 A3A2S!29 DEVICE, LABEL THE VALVE POSITION INDCATIONS & LASEL ARE 86050905 / / MISSING. 440 FW 40 A3A25128 (NEAR) DEVICE, LABEL THE VALVE POSITION I C ICATIONS AND LABEL AEE 86050905 / / MISSING. 442 FW PUMP TURE A A3A2DS179 DEVICE.LAaEL LAMP DS179 REMOVED, A PLUG CO.'ERS If NOW. 89020901 /^/ ALSO LABEL REMOVED 444 FWV-38 A3A2Sf31 DEVICE, LABEL' VALVE POSITION INDICATIONS AND LABEL ARE B6050905 ' /. /. MISSING. 445 FW-39 A3A2S130 (NEAR) DE VICE,L ABEL VALVE POSITION INDICATIONS & LABEL ARE 86050905 / / MISSING. l j
O O- O Page No. 10 OPEN PHYSICAL FIDEL 1TT DISCREPANCIES 09/09/91 DR,NUM PLANT _10 SIM_ID CATAGORY DESC RESOLUTION IMPACT COMP,CATE 447 FWM-1B A3A25113 ENGRAVING PLANT "FWM-18:TURB 8:DISEN..." /SIM. ESC 84081018 / /
"...FWM-18: -SLANK-:...
451 MS LINE MONITOR A4 DEVICE THE MAIN STEAM LINE MONITOR IS NOT IN THE 82050315 / / SIMULATOR. 452 TGF-Ax2-1 1 A4AZA2 MARKING PLANT " -BLANK " /SIM. 84081001 / /
" CONDENSER: VACUUM: LOW" 454 SP-004A-T!R A4A2A6 SCALE PLANT SCALE HAS
- F" DISPLAYED BETWEEN 550 & SCA / /
650, & GRAD. .. 457 18-240-P1 A4A2M14 SCALE PL ANT 0,500,1000,1500, . . . /SIM. SCA / / 0,750,1500,2250,3000 462 TGF-Ax3-4-10 A4A2A3 MARKING PLANT "TURS: AT ZERO SPEED" /SIM. " - BLANK - 86062103 /'/ e 463 TGF-Ax3-4-9 A4A2A3 *kx(ING PLANT "TURB GEN: SPEED CHANNEL: TR.." 86062103 / /
/SIM."TURB:AT ZERO ..."
464 .TGF-Ax3-5-6 A4A2A3 MARKING PLANT "TURB GEN: CONDITION MON:HIGH" /SIM. " - 86062103 / / BLANC " 465 TGF-Ax3-5-5 A4A2A3 MARK!NG PLANT *TURB GEN: RADIO FREQ: SIGNAL HIGH" 86062103' /./
/SIM." EXCITER: FIR...
466 TGF-Ax3-6-8 A4A2A3 MARKING. PLANT
- EXCITER:PAISE:lNHIBIT" /SIM."TURB 86062103 / /
GEN: SPEED CHAN..." 467 TGF-Ax3-6-7 A4A2A3 MARKING PLANT
- EXCITER: LIMITED:PRETRIP" 86062103 / /
/SIM.* EXCITER: RAISE:lWHIBIT" 468 TGF-Ax3-6-6 A4A2A3 MARKING PLANT
- EXCITER: GROUND" 86062103 //
/S!M." EXCITER: LIMITED:PRETRIP" 469 TGF-Ax3-6*5 A4A2A3 MARKING PLANT "EXCITEE:F! RING CKT: POWER LOSS" 86062103 / /
/SIM." EXCITER: GROUND" 4 70 TGF-Ax3-5-3 A4A2A3 MARKING PLANT"TURB. AMSAC:CMANNEL:1N TEST" /St4."TURB S4060702 / /
LIM!TED: SY... 471 TGF-Ax3-5-2 A4A2A3 MARKING PLANT "TURB.AMSAC:t0W Flux:dvPASS" /SIM."TURB 84080702 / / THROTTLE: ... 4 72 TGF-Ax3 3-4 A442A3 MARKING PLANT"TURB. THROTTLE: PRESS.:HIGH/LDWa /SIM." - 84080702 // BLANK 4 73 TGF-Ax3-5-1 A4AZA3 MARKING PLANT"TURB.AMSAC:CMANNEL: TRIP" /SIM.* THROTTLE 84030702 / / PRESS:TRAN ... 4 75 TB-3S6-VAR!R A4AZA5 SCALE DIFFERENT GRADUATIONS. SCA / / 476 TB-355-JIR A4A2A4 SCALE FLANT 0,200,400,600,... /SIM. 0,20,40,80, ... SCA /-/ l i l l l
n . . , i O O' O: 11 OPEN PNYSICAL FIDELITY DISCREPANCIES Page No. 09/09/91 kF % UTION IMPACT COMP ,DATE CATAGORY DESC DR_NUM PLANT _ID,. SIM_ID
/ /
496 Exv-65 A4A2S93 MARKING PLANT *EKDT-12 BYPASS" /SIM.*EXDT-11 SYPASS" ESC ESC
/ /
517 TDV-10 A4A2S118 MARKING PLANT "STM CHEST:X-TIE DRAIN" /SIM."C.V. CHEST X-TIE" 69081103 / / 537' SSF-A1-2-1 ASA2A1 MARKING PLANT "0FFSITE PWR: SOURCE XFMR:FAUL1d
/SIM."6.9EV:BUSA: ... /,/
53S SSF-A1-2-2 ASA2A1 MARKING PLANT " BREAKER:4900/4902: TRIP" /SIM."4KV: UNIT 89081103 BUS A: DEAD" 80081103 / /.
$39 SSF-A1-3-1 ASA2A1 MARKING PLANT "6.9KV: BUS A:DE AD" /SIM."6.9KV:8US 8: DEAD" 89081103 / /
540 SSF-A1-3-2 ASA2A1 MARKING PLANT "4 KV: UNIT BUS A: DEAD" /SIM."4KV: UNIT BUS 8: DEAD" 89081103 / / 541 SSF-Al-4-1 ASA?A1 MARKING PLANT "6.9KV: BUS 8:DE AD" /SIM."INVER?ER A: FAILURE"
/ /
542 SSF-A1-4-2 ASA2A1 MARKING PLANT "4KV: UNIT BUS 8: DEAD" /SIM."INVERTEP A: 89081103 TROUBLE"
/ /;
5-1 85-1 IN THE PLANT ARE THE SAVE AS 4-1 & 89081103 543 SSF-A1-5-1 ASA2A1 MARKING 4-1 04 SIM.
/ /-
544 SSF-A1-6-1 A5AZA1 MARKING 6-1 & 6-2 IN THE PLANT ARE THE SAME AS 5-1 & 89081103 5-1 04 51M. 87081103 -/'/- 545 SSF-A1-7-1 ASA2A1 MARKING 7-1 87-2 IN THE PLANT ARE THE SAME AS 6-1 & 6-1 04 S1M. 89051103 / / 546 SSF-A1-8-1 45A2A1 MARKING 8-1 &S-2 ON THE PLANT ARE THE SAME AS 7-1 & 7-2 ON $1M. 89081103 / / 547 SSF-Al-8-3 ASA2A1- MARK 1NG 8-3 &S-4 IN THE PLANT ARE THE SAFE AS 8-1 & 8-2 IN THE SIM.
/ /
548 SSF-A1-7-3 ASA2A1 MARKING 7-3 & T-4 IN THE PLANT ARE THE SAME AS 6-3 & 89081103 6-4 DN SIM. 89081103 / / ASA2A1 MARKikG PLANT "OFFSITE PWR:SOURCF RFMR:NAJOR At*PM" 549 SSF-A1-2-3
/SIM."480V: ..." / /
PLANT "0FFS!TE PWR: SOURCE XFwR: MINOR ALARM" 8'7181103 550 SSF-A1-2-4 ASA2A1 MARKING
/SIM."480v: ..." / /
PLANT: - 6LAN( - /SIM." 89081103 ASA2A1 MARKING
$51 SSF-A1-3-3
/ /;
FLANT "480V:TURs Aux EUS A: DEAD * /SIM." 89031103 552 SSF-A1-3-4 A542A1 . MARKING
~
89051103 / /- 553 $$F-Al-4-4' ASA2A1 MARKING PLANT "480 V:TUR8 AUW BUS 8: DEAD" /SIM.
~ BLANK-y
0 0 , O Page No. 12 OPEN PHYSICAL FIDELITT DISCREPANCIES 09/09/91 DR_NUM PLANT _ID .SIM_ID CATAGORT DESC RESOLUTION IMPACT CCPP_DATE 554 $$F-Al-8-6 ASA2A1' MARKING 8-5 & 8-6 IN PLANT MATCH 7-5 & 7-6 OM 89081103 / /- SIMULATOR. 555 S$F-Al-7-6 ASA2A1 MLRKING 7-5 & 7-6 IN PLANY MATCH 6-5 & 6-6 ON TNE 89081103 / /. SIMULATOR. 556 SSF-A1-8-7 ASA2A1 MARKING 8-7 & 8-8 IN PLANT MATCH 7-7 & 7-8 ON THE 89081103 / / SIMULATOR. 557 SSF-A1-7-7 ASA2A1 MARr!NG 7-7 & 7-8 IN PLANT MATCHES 6-7 & 6-8 IN THE 89081103 / /
$1MULATOR. 5 558 SSF-A1-2-5 ASA2A1 ' MARKING PLANT
- BREAKER:4900/4902: TROUBLE" /SIM.*480 89081103 / /
V:!NTAKE ... " 559 SSF-A1-2-6 ASA2A1 MARKING PLANT -8 LANK- /SIM."480 V PLANT: AUX BUS: DEAD" 89081103 / / 560 SSF-A1-2-7 ASA2A1 MARK!NG PLANT -BLANK- /SIM.*480 V: NON-ESS 89081103-08 / / LOADS: TRIPPED" 561 SSF A1-2-8 ASA2A1 MARKING PLANT -BLANK- /SIM."480 V:MCC BREAKER:0 PEN" 89081103 / / 562 SSF-A1-3-5 A5A2A1 MARKING PLANT "480 V:Rx AUX GUS A: DEAD" /SIM."480 89081103 / / V:18iTAKE BUS . .
- 563 SSF-A1-3-6 ASA2A1 MARKING PLANT *480 V:!NTAWE BUS A. DEAD" /SIM."4RO 89081103 / /
V:HIG ,WX .. . " 564 SSF-A1-3-7 ASA2A1 MARK!NG PLANT "480 V PLANT: AUX BUS: DEAD" /$1M.*480 80081103 / / V: XFMR TEMP..." 565 SSF-A1-3-8 ASA2A1 ' MARKING PLANT -BLANK- /SIM."P2R:MCC BREAKER:0 PEN" 89081103 /-/ 566 SSF-A1-3-9 ASA2A1 ' MARKING PLANT "480 V:MCC BREAKER:0 PEN" /SIM. 89081003 / /-
-BLANK-567 SSF-Al-4-5 A5A2A1 MARKING PLANT "480 V:RX AUX BUS B: DEAD" /SIM. 89081103 / /
-BLANK-568 SSF-Al-4-6 ASA2A1 MARKING PLANT "480 V:!NTAKE BUS B: DEAD" /SIM. 89081103 / / 3
-BLANK-I 569 SSF- Al 7 ASA2A1 MARKING PLANT "480 V PLANT:HTG BUS: DEAD" /SIM. 89081103 / /
-BLANK-570 SSF-A1-4-8 ASA2A1 MARKING PLANT "480 V:xF9R TEMP:NIGH" /SIM. -BLANK- 89081103 / /
571 S$F-A1-4-9 ASA2A1 MARKING PLANT "P2R:MCC BREAKER:0 PEN" /SIM. -BLANK- 89081103 / / 5 72 SSF-A2-3-2 A5A2A2 MARKING PLANT " DIESEL GEN A:OVER 3250 Kw:FOR 24/29 88051501 88051501-02 871019014 / / MIN"
O O- O Page No. 13 OPEN PHYSICAL FIDELITY DISCREPANCIES 09/09/91 DR ,NUM PLANT _ID SIM_10 CATAGORY DESC RESOLUTION IMPACT' C0"P_CATE 5 73 SSF-A2-3-7 ASA2A2 MARKING PLANT -BLANK- /SIM."480 V:ES BUS A: DEAD" 86051501 87101901 A / / 574 SSF-A2-3-5 A5A2A2 MARKING 3-5 & 3-6 IN PLANT MATCH 2-5 & 2-6 IN THE 83051501 87101901 A ./ / SIMULATOR. 575 $$F-A2-7 8 ASA2A2 MARKING PLANT "480 V:ES BUS B:UV LOCKOUT ACT" /SIM. 89081103 / /
-BLANK-576 SSF-A2-8-2 ASA2A2 MARKING PLANT " DIESEL GEN B:OVER 3250 KW:FOR 24/29 88051501 88051501-02 / ' /
MIN" /SIM -BLANK-577 SSF-A2-8-1 ASA2A2 MARKING PLANT " DIESEL GEN B:0VER:3250 KV" /SIM. 88051501 88051501-02 / /
-BLANK-578 SSF-A2 3*4 ASA2A2 MARING PLANT "4KV ES A:UNDERVOLTAGE: TRIP BLOCKED" 88051501 / /
/SIM."4KV ES ..."
5 9 SSF-A?-3-3 ASA2A2 MARKING PLANT -BLANK- /SIM." DIESEL GEN 88051501 87101901 A / / A: BREAKER: CLOSED
- 580 SSF-A2-3-1 ASA2A2 MARKING PLANT " DIESEL GEN A:0VER:3250KW" /SIM."4KW:ES 88C51501 88051501-02 87101901A / /
BUS A: DEAD" 581 SSF-A2-2-4 ASA2A2 MARKING PLANT "4KV ES A: CROSS-TIE: BLOCKED" /S!M."4KV 88051501 87101901A / / ES A: UNDER..." 582 SSF-A2-2-3 ASA2A2 MAnKING PLANT " DIESEL GEN A: BREAKER: CLOSED" /SIM. 88051501 87101901A / /
~ BLANK-583 SSF-A2-2-2 ASA2A2 MARKING PLANT
- DIESEL GEN A:BREAr.ER: TRIP" /SIM. 83051501 87101901A / /
-BLANK-584 SSF-A2-2-1 ASA2A2 MARKING PLANT "4KY:ES BUS A: DEAD" /SIM. -BLANK- 88051501 87101901A / /
$85 SSF-A2-2-8 ASA2A2 WUKING PLANT "480 V:ES BUS A:UV LOCKOUT ACT" /SIM. 89081103 / /
-BLANK-586 SSF-A2-2-7 A*;A21.2 MARKING PLAhT "480 V:ES SUS A: DEAD" /SIM. -BtANK- 68051501 87101901 A / /
590 A10A1A155/A156 MARKING THE OTF/04 ESCUTCHEON PLATE FOR SWITCHES ESC / / MISSING 591 252/3209 A10 DEVICE 2 SYNC. CHECK RELAYS ARE MISSING FROM 88041101 / / SIMULATOR. 592 TB-448-A1 A10 DEVICE Th!S DEv!CE (TB-448-A1) 15 MISSING FROM 86062103 / / SIMULATOR. 593 A10 DEVICE THE AUXILLIARY RELAY IS MIS $1NG. 88111101 88111101-01 / /
$94 SW #1873 A10A6S5 DEVICE CEVICES MISSING. 89081103 / /
i O O .O
-Page No. 14 OPEN PHYSICAL FIDELITY DISCREPANCIES 09/09/91 DR Nt!M PLANT _ID SIM_ID CATAGORY DESC RESOLUTION IMPACT COMP ,DATE
] 595 A10A642 LABEL PLANT "230 KV PLT.LINE 48/V:RLY 86 SU/PL4/1" :88111101 / /
/SIM. *230 ..."
l 596 868U/PL4-1 A10A6A1 (NEAR) CEVICE LOCKOUT RELAY MISSING. 88111101' / / 621 86 X 27/EiA A12A52 DEV!CE- 86X2//ESA L/0 RELAY IS MIS $1NG, LOCATE BELOW 88110501-07 / / A23A52. 622 RM-c29-RIR A12A64 DEVICE TMIS DEVICE ho LONGER EXISTS AT THE PLANT. 84081018 i/ / 623 A12A58 LABEL PLANT "OFFSITE POWER: SOURCE XFMR" /SIM." UNIT 89081103-03 / / < . #1 & #2:ST ..." j 624 A11439/40/41 DEVICE .THESE 3 DEVICES NO LONGER EXIST IN THE PLANT. 83070491 / / 683 PHONE ST HOT. PH0hE ST HOT. O!hER STATE HeD PH. NEEDS TO BE RELOCATED, ..., TR-00026 / / , REPROGRAM *ED. 700 A29A1A1DS17 LABEL A29A1A1DS17 LABELS 00 #0 AGREE. 84C82401 / / 701 A29A1A1 IND.,LASELS' PLANT HAS 2 ADDITIONAL LAMDS & LABELS ON THE 88052401 / / LEFT SIDE. 702 A29A2A10$17 LABEL PLANT LABEL FOR A29A2A1DS17 DOES NOT MATCH 88052401 / / S!MULATOR. l 703 A29A2A1 LAMPS / LABELS PLANT LEFT HAS 2 MORE LAMPS & LABELS THAN THE 85052401 / / SIMULATOR. l 704 A29A3A1DS17 LABEL A29A341DS17 LABEL DIFFE8ENT BETWEEN PLA%T & 88052401 / / . S1M. 706 A29A3A1 LMP,L ABEL PLANT PANEL LEFT MAS 2 MORE LAMPS / LABELS THAT 88052401- / / { SIM. i 707 A29A4A1 LAMP, LABELS PLANT LEF1 SIDE MAS 3 MORE LAMPS & LABELS 88052401 /'/ THAN SIM. ... , 709 A29A4A2S1 SVITCM, LABEL PLANT HAS SW. & ASSOC. LABEL ... WEAR 88080101 / / A29A4A2S1 ... ? 710 A29 '1A1DS10 LABEL A29AB1A1DS10 LABELS DIFFER-PLANY "BSP-18", 87070601 / /
/SIM."ASB: START" 711 A29AB1A1DS22 LABEL PLANT " SPARE" /SIM."CCP-1B" 88052401 / /
712 A29AB1A1 LAMPS / LABEL PLANT MPI-3 SECTION MAS 2 LAMPS & 1 LABEL, 88052401 / / SIM. DOESN'T 713 A29AB1A1 LAMPS, LABEL PLANT LPI-1 SECTION HAS 2 MORE LAMPS & 1 88052401 / / ; LABEL ThAN SIM.
O O- O Page No. 15 OPEN PHYSICAL f1DELITY DISCREPANCIES 09/09/91 SIM_ID CATAGORY DESC RESOLUTION IMPACT COMP _ ATE-DR_NUM PLANT _ID 714 A29AB1A1DS67/74 LABEL LABELS FOR DS67 & DS74 ARE NOT THE SAME FOR 88052401-05 / / PLANT & SIM. 715 A29AB1A1DS1/15 LASEL ...DS1 & DS15 LABELS IN HDI-1 SECT. IS 88032401 / / DIFFERENT PLANT /SIM. 716 A29AB1A1DS$2/58 LASEL ...DS5210S58 LABEL ON HPI-2 SECT. DIFFER !.d - 88052401 / / PLANT & SIM. 717 A29AB2A1DS9/23 LABEL DS9 & 0523 LABEL 04 RSC-4 SECT. DIFFER IN '87011901 / / PLANT AND SIM. 718 A29AB2A1DS4/18 LABEL ...DS4 & DS18 LABEL 04 RSC-3 SECT. DIFFER IN 87011901 / / 'i PLANT & SIM. 719' A30A4A1DS30,34 LABEL .. 0530 THRU DS34 LABEL DIFFER IN PLANT & 88052401 / / SIL 720 A30AB1A1DS8/16 LABEL ...DSB&DS16 ON HPi-3 SECT. PLArT "BSP-1A", 87070601 / / ' !
/SIM."ASA '; TART" 721 A30AB1 AIDS 36/45 LABEL ...D536 & DS45 LABEL ON MP3-3 SECT. PLAN 83052401 / /
" SPARE", /SIM."...
722 A30AB1 AIDS 37/46 LABEL ...DS37 & DS46 LABEL 04 HPI-3 SECT. PLANT 88052401 / /
"EFP-1: START", ... ?
723 A30AB1DS37/46 BEZEL .. 0537 & D546 BEZELS PLANT BLACK (NEON) & 88052401 / / SILVER... IN $1M. 724 A30AS1A1 ' LAMPS, LABEL PLANT HAS 2 MORE LAMPS & T MORE LABEL 04 88952401 // f NPI-3 SECT., ... 725 A3 CAB 141 LAMPS, LABEL SECT. LP3-1,LPI-2 & LPI-3 ON A30AB1A1 ARE 88052401 / / RADICALLT DIFFER'T 726 A30AB1A1DS1/18, LABEL (ALSO 2/19,3/20) 3 LABEL ON HP!-1 SECTION 83052401 / / DIFFER. + 728 A30AB1A1DS M/63 LABEL A30AB1A1DS56&DS63 LABEL ON HPI-2 SECTION 88052401 / / PLALT
- SPARE", ...
- 730 A30AB2A1DS10/27 LABEL A30AB2A1DS10 & D527 LABEL ON RBC-4 SYCT. 87011901 / /
4 PLANT "LRV-72", ... 732 A30AB2A1054/21 BEZEL, LABEL" .. 04 RBC-3 SECTION PLANT *LRV-70"', 87011001 / / ' i SIM."AHV-1C*, ... 733 A30A3A10S17 LABEL ttHT ON BLK, PLANT "62-1,2,3,4,5,6:RC3", 88052401 / / SIM."62-1,2,3:RC3" 734 A30A3A1 1 AMPS, LABELS LEFT SIDE PANEL MAS 3 MORE LAMPS / LABELS IN 88052401 / / , I PLANT THAN SIM. I 735 A30A2A1DS17' LABEL PLANT "62-1,2,3,4,5,6:RC-2" 8805240' / /. ; 4 /51M."62-1,2,3:RC-2", ... j ;
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR (])- CERTIFICATION SUBMITTAL APlENQJLG SIMULATOR SYSTEES SYSTEM DESIGNATOR llILE AA GAS WASTE PROCESS AV GAS WASTE VALVES & MOTORS Al GAS WASTE CONTROL & LOGIC AB LIQUID WASTE PROCESS AW LIQUID WASTE VALVES & MOTORS A2 LIQUID WASTE CONTROL & LOGIC B0 DECAY HEAT REMOVAL PROCESS FU DECAY HEAT REMOVAL VALVES AND MOTORS f,1 DECAY HEAT REMOVAL CONTROL & LOGIC BM MAKEUP & PURIFICATION PROCESS BV MAKEUP & PURIFICATION VALVES AND MOTORS B2 MAKEUP & PURiflCATION CONTROL & LOGIC B20 MAKEUP & PURIFICATION BATCH CONTROLLER BF CORE FLOODING PROCESS BW CORE FLOODING VALVES AND MOTORS B3 CORE FLOODING CONTROL & LOGIC p) (_ BC SPENT FUEL COOLING PROCESS BX SPENT FUEL COOLING VALVES AND MOTORS B4 SPENT FUEL COOLING CONTROL & LOGIC BG CHEMICAL ADDITION PROCESS BY CHFMICAL ADDITION VALVES AND MOTORS B5 CHEMICAL ADDITION CONTROL & LOGIC BE LIQUID SAMPLING PROCESS BS REACTOR BUILDING SPRAY PROCESS / VALVES & MOT 0RS B6 REACTOR BUILDING SPRAY CONTROL & LOGIC CA CONTAINMENT PROCESS
-CV CONTAINMENT VALVES & MOTORS C1 CONTAINMENT CONTROL & LOGIC CI VENTILATION PROCESS CU VENiiLATION VALVES & MOTORS C7 VENTILATION CONTROL & LOGIC CH CHILLED WATER PROCESS CJ INDUSTRIAL COOLING PROCESS CL METEOROLOGICAL SYSTEMS CM RADIATION MONITORING DD HEATER DRAINS & VENTS PROCESS DQ HEATER DRAINS & VENTS VALVES & MOTORS D2 HEATER DRAINS AND VENTS' CONTROL & LOGIC DF EXTRACTION STEAM PROCESS / VALVES AND MOTORS D3 EXTRACTION STEAM CONTROL & LOGIC DC CONDENSER PROCESS (3
V
FLORIDA POWER CORPORATION f'] CRYSTAL RIVER UNIT 3 SIMULATOR CERTiflCATION SUDMITTAL APPENDIX _Q SIMULATORSYSTEMS(Continued) SYSTEM DESIGNATOR IllLE DB CONDENSER AIR REMOVAL PROCESS DW CONDENSER AIR REMOVAL VALVES & MOTORS D4 CONDENSER AIR REMOVAL CONTROL & LOGIC DM MSR HEATER DRAINS & VENTS PROCESS DZ MSR HEATER DRAINS & VENTS VALVES & MOTORS D7 MSR HEATER DRAINS & VENTS CONTROL & LOGIC El 6.9 KV CONTROL LOGIC & PROTECTION E2 4,16 KV CONTROL LOGIC & PROTECTION E3 480 V CONTROL LOGIC & PROTECTION E6 120 VAC CONTROL LOGIC & PROTECTION E9 STATION CONFIGURATION EL STATION LOADS ET DISTRIBUTION TRANSFORMER COOLING FA- CONDENSATE & DEMINERAllZER PROCESS FV CONDENSATE & DEMINERAllZER VALVES & MOTORS F1 CONDENSATE & DEMINERAllZER CONTROL & LOGIC (q_/ FF FEEDWATER & CYCLE START UP PROCESS FW FEEDWATER & CYCLE START UP VALVES & MOTORS F2 FEEDWATER & CYCLE START UP CONTROL & LOGIC F3 EFIC FK EMERGENCY FEEDWATER PROCESS FY EMERGENCY FEEDWATER VALVES AND MOTORS F4 EMERGENCY FEEDWATER CONTROL AND LOGIC FB SEAL & SPRAY WATER PROCESS FX SEAL & SPRAY WATER VALVES & MOTORS F6 SEAL & SPRAY WATER CONTROL & LOGIC FT FEEDWATER & CONDENSATE THERMODYNAMICS GO SYNCHRONIZING GA- SYNCHRONIZING UNIT GS SYNCHR0 SCOPE DRIVER GG MAIN GENERATOR STATOR GF EXCITER. G1 MAIN utNERATOR CONTROL & LOGIC GD EMERGENCY DIESEL GENERATOR PROCESS GS EMERGENCY DIESEL CONTROL & LOGIC GP EMERGENCY DIEEEL SUPPORT HC RCS HEAT TRANSFER HH1 RCS HYDRAULICS PART 1 HH2 RCS HYDRAULICS PART 2 HQ1 RCS HEAT TRANSPORT PART 1 HQ2 RCS HEAT TRANSPORT PART 2 HS1 RCS THERMODYNAMIC PROPERTIES PART 1 O~' HS2 RCS THERMODYNAMIC PROPERTIES PART 2
FLORIDA POWER CORPORATION () CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICATION SUBMITTAL APPENDIX C SIMULATOR SYSTEMS (Continued) 4 SYSTEM DESIGNATOR lilLE HK REACTOR COOLANT DRAIN TANK HN RCS PUMP SEAL INJECTION HU RCS PUMP MOTORS HV RCS VALVES HI RCS CONTROL & LOGIC JA ANNUNCIATOR SYSTEMS JP PLANT PROCESS COMPUTER JS SAFETY PARAMETER DISPLAY SYSTEM KA SW & DH RAW WATER PROCESS KV SW & DH RAW WATER VALVES AND MOTORS K1 SW & DH RAW WATER CONTROL & LOGIC KD SECONDARY SERVICES COOLING PROCESS KX SECONDARY SERVICES COOLING VALVES AND MOTORS K3 SECONDARY SERVICES COOLING CONTROL & LOGIC KK CIRCULATING WATER PROCESS KY CIRCULATING WATER VALVES AND MOTORS
.() K4 KF CIRCULATING WATER CONTROL & LOGIC INSTRUMENT & STATION AIR PROCESS KZ INSTRUMENT & STATION AIR VALVES AND MOTORS K5 INSTRUMENT & STATION AIR CONTROL & LOGIC KE DECAY HEAT CLOSED CYCLE COOLING PROCESS KW DECAY HEAT CLOSED CYCLE COOLING VALVES AND MOTORS K6 DECAY HEAT CLOSED CYCLE COOLING CONTROL & !.0GIC KJ DOMESTIC WATER PROCESS KU DOMESTIC WATER VALVES & MOTORS K7 DOMESTIC WATER CONTROL & LOGIC KH DEMINERAllZED WATER PROCESS KT DEMINERALIZED WATER VALVES AND MOTORS K8 DEMINERAllZED WATER CONTROL AND logic KB NUCLEAR SERVICES COOLING PROCESS KS NUCLEAR SERVICES COOLING VALVES A!;D MOTORS K9 NUCLEAR SERVICES COOLING CONTROL & LOGlC KC CARBON DIOXIDE GAS KG NITROGEN GAS KL HYDROGEN GAS KN FIRE SERVICE KP INTAKE STRUCTURE L1 ICS TRANSMITTERS L2 -ICS DIGITAL CONTROL L3 ICS ANALOG CONTROL L4 REACTOR PROTECTION SYSTEM L5 CONTROL R0D DRIVE LOGIC
' (~J' k- LI CONTROL R0D DRIVE POSITION INDICATION
-~ FLORIDA POWER CORPORATION ( _) CRYSTAL RIVER UN11 3 SIMULATOR ~ CERTIFICA110N SUBMITTAL 6.PPENulX C SIMULATOR SYSTEMS (Continued) SYSTEM DESIGNATOR IJILE L6 ESAS L7 EHC TURBINE CONTROL LA ANNUNCIATOR WINDOWS LB ANNUNCIATOR INPUTS LS REMOTE SHUTDOWN PANEL N1 NNI SPENT FUEL, BUILDING SPRAY, & CONTAINMENT INSTRUMENTATION N2 NNI MAKEUP & PURIFICATION INSTRUMENTATION N3 NNI CORE FL000 & DECAY HEAT INSTRUMENTATION NP NN1 POWER SUPPLIES P1 TRANSFORMER CONTROL LOGIC & PROTECTION P8 SWITCHYARD CONTROL & LOGIC PG ELECTRICAL GRID COMPUTATION PI ISOLATED PHASE BUS COOLING PT TRANSFORMER COOLING t- Q1 125/250 VDC CONTROL & LOGIC (g/ QL DC DISTRIBUTION R3 EXCORE DETECTORS R4 INCORE DETECTORS RD DETECTOR SIGNALS RA REACTOR KINETICS RL COUPLING COLfFICIENTS RR REACTIVITY & THERMAL POWER SR- MAIN STEAM THERMODYNAMICS SA MAIN STEAM PROCESS SV MAIN STEAM VALVES & MOTORS S1 MAIN STEAM CONTROL & LOGIC S2 REHEAT CONTROL & LOGIC ST AUX SfEAM THERMODYNAMICS SF AUX STEAM PROCESS SX AUX STEAM VALVES & MOTORS SS AUX STEAM CONTROL & LOGIC SG OTSG UA LUBE Olt PROCESS UW LUBE OIL VALVES & MOTORS U3 LUBE Oil CONTROL & LOGIC UB TURBINE EH FLUID PROCESS UX TURBlNE EH FLUID VALVES & MOTORS U4 TURBINE EH FLUID CONTROL & LOGIC UV TURBINE HYDRAULIC VALVES UC GLAND STEAM PROCESS (]) UY US GLAND STEAM VALVES & MOTORS GLAND STEAM CONTROL & LOGIC
FLORIDA POWER CORPORATION O CRYS 1At river uni 1 3 SinutA10R CERTIFICATION SUBM111AL 8EEMQlLG SIMULATOR SYSTER$(Continued) l SYSTEM DESIGNATOR IIILE US GENERATOR OIL SEAL PROCESS / VALVES & MOTORS V8 GENERATOR SEAL Oil CONTROL & LOGIC VJ HYDR 0 GEN COOLING PROCESS / VALVES & MOTORS U9 HYOROGEN COOLING CONTROL & LOGIC UF TURBINE SUPERVISORY INSTRUMENTATION UK TURNING GEAR O O
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p p n U V d j SDs IN-SERVICE, w/ IMPACT, & NOT ON SIMULATOR daga No. 2 09/0*/91 S0tRCE DOWpENTS IW-SERVICE fluAL EVAL SIMJLATOR DATE IMPACT DATE DATE CCM TITLE DOC DATE REE IVED STAT SOURCE DOC f REY TTPE 01/19/90 S 01/04/89 / / 900081
/ / 12/14/88 OP 8704 401 8 0 MAR SA CONkECTION TO AH COMPRESSOR RECE1 VERS 900147
/ / 03/04/90 OP 06/16/90 S 09/10/90 / /
87C,050 s 1 FCN ADDITION OF REVISED VENDOR DRAWINGS 900006
/ / 04/25/88 OP 06/16/90 S 06/02/88 / /
8704060" 7 0 MAR EFIC-VECTOR MODUtE LOGIC CHANGE 900031
/ / 06/24/88 (P 12/01/88 $ 01/29/90 / /
8703c501 8 0 MAR DECOMMIS$10N RCP SP SWITCHES 900003
/ / 06/28/88 OP 03/28/89 MS 07/01/88 / /
87022501 8 0 MAR RCITS NUISANCE ALARM
/ / 900062
/ / 11/21/88 OP 06/07/89 5 08/23/90 87011902 8 0 MAR RB DEPRESSURIZATION THROTTLE VALVE 900125
/ / 03/20/89 OP 06/07/89 S Or./24/89 / /
87011901 5 TCM A00!T10N OF RELAY 03/02/89 / / 900136
/ / 02/07/89 OP 06/07/89 N 87011901 2 FCN REVISE STATUS LAMP ENGR AVING
/ / 900076
/ / 12/05/88 OP 06/07/89 MS 08/23/90 87011901 8 0 MAR RB DEPRESSUR12ATION
/ / 900051
/ / 09/20/89 OP 10/20/88 S 01/29/90 8 0 MAR CAD BREAKER UV SENSOR SETp01NT CHANGE 87011301 01/25/90 / / 900036
/ / 06/24/88 OP 01/t8/89 5 86121003 8 0 MAR ES ACTUATION SIGNAL SEAL IN FOR BSV3 & BSV4
/ / 910038
/ / 06/25/88 CP 12/12/88 S 02/20/91 86100901 7 FCN ADD CONTACT BLOCK
/ / 910037
/ / 06/24/88 OP 12/12/88 S 02/20/91 86100901 4 FCN RELOCATION OF SIGNAL MONITORS 910036
/ / 02/20/91 Op 12/12/88 S 02/20/91 / /
h6100901 8 0 MAR LOW TEMP OVERPRESSURIZATION PROTECTION ALARM ADDIT
/ / 900138
/ / 07/06/90 OP 10/24/90 # 08/22/90 86062103 11 FCN MAIN GENERATCR UPGRADE MONITORING
/ / 900126
/ / 07/06/90 Op 10/24/90 = 08/22/90 66062103 10 FCN MAIN GENERATOR UPGRADE MONITORING
/ / 900066
/ / 11/04/89 OP 10/24/90 FS 12/08/89 86062103 7 0 MAR MAIN CENERATOR L7 GRADE MONITORING INSTPUMENTATION 900104
/ / C5/29/90 0F 10/24/90 S 09/10/90 / /
86062102 3 FCN ADD FikAL TEST PROCEDURE 08/24/90 / / 900096
/ / 09/10/89 OP 10/24/90 S 06062102 1 FCN ADD PRELIMINARY TEST PROCEDURE
/ / 900082
/ / 12/23/88 op 10/24/90 S 08/24/90 86062102 7 0 MAR MAIN GENERATOR STATOR REFLACEuf4T
/ / 900009
/ / 02/18/89 OP 09/14/89 HS 03/13/89 86050905 8 0 MAR VPI FOR MFw STARTUP AO LOW LOAD PEG VALVES 900130
/ / 01/18/90 OP 10/03/89 S 03/12/90 / /
86050812 2 FCN REPLACEMENT OF ICS/NW1 F0WER SUPPLY
/ / 000105
/ / 01/19/90 OP 05/09/90 S 03/12/90 86050807 1 FCN NN! ALARM STATUS Mw 900034
/ / 02/22/89 OP 05/09/90 5 09/10/90 / /
86050807 6 0 MAR NN! ALARw STATUS MODIFICATION 9031?6
/ / 11/19/89 OP 09/14/89 5 09/10/90 / /
86050805 1 FCN BACKUP CONTROLS FOR PRESS LEVEL & CCWOEN FLOW 900077
/ / 07/08/89 OP 09/14/89 5 08/22/89 / /
86050805 8 0 MAR BACKUP CONTROLS FOR PRCSS. LEVEL & COCENSATE FLOW
/ / 900123
/ / 03/21/88 OP 03/22/88 5 06 d2/78 86020301 11 FCN NEW ALERT AND DANGER SETPOINTS 910244
/ / 05/15/91 OP 01/26/38 5 07/15/71 / /
8 0 MAR RV Pump U USN WATER 85090501 09/10/90 / / 90010-2
/ / 01/12/00 OP 06/26/90 M 84081010 22 FCW EAME PLATE CORRECTION 06/26/90 M 08/24/87 / / 900119
/ / 07/29/89 OP 84081010 20 FCN MAIN CONTROL BOARD MODIFICATION
/ / 900132
/ / 06/28/88 09 06/26/90 M 07/01/88 8*081010 14 FCN ADDITIONAL CORRECT!OkS 900113
/ / 04/18/88 OP 06/26/90 M 06/02/88 / /
84081010 12 FCN MCB NAMPLATE CORRECTIONS
/ / 910031
/ / 02/19/91 Op 10/05/88 s 02/19/91 84081003 4 FCN R.G. 1.97 UPGR ADE 9*002S
/ / 02/19/91 OP 02/27/89 M 02/19/91 / /
84081002 11 FCN CHANGE R.I.P. METER SEALES i / 910034
/ / 02/19/91 op 02/27/89 5 02/19/91 84081002 10 FCN REMOVE TEST SWITCH FOR VALVE E l e .'M l
t ; V kJ %/ Page No. 3 SDs IN-SERVICE, w/ IMPACT, & NOT ON SIMULATOR 09/09/91 SOUR & Dot 1MNTS IN-SERVICE FluAt EVAL SIMJLATOR SutRCE DOC # REV 1TPE TITLE DOC DATE RECEIVED STAI DATE IMACT DATE DATE CCN 84060702 9 FCN ATWAS / / 05/08/91 OP 03/26/90 S 05/14/91 / / 9*0194 84060702 3 0 MAR ANTICIPATED TRANSIENT WITHOUT SCRAM - OUTAGE / / 06/22/89 OP 03/26/90 S 08/02/89 / / 900015 84060701 6 FCN CHANGE 0UT TO START-UP FEEDWATER FLOW / / 03/24/90 OP 06/27/90 S 09/10/90 / / 900137 83070401 8 0 MAR GENERATOR BUS VOLTAGE TELEMETRY & METERING UPGRADE / / 01/27/90 OP 02/12/90 HS 04/12/90 / / 900048 83041901 2 FCN ADDITION OF LESS THAN 22% FLUX RELAT / / 02/25/91 OP 01/12/89 5 02/25/91 / / 910052 82091902 7 FCW ADDITION OF NAMEPLATES / / 06/24/88 OP 01/05/89 d 01/25/90 / / 900118 l 82050321 1 FCN CORE FLOOD T ANK XMITTER REPLACEMENT / / 02/10/90 OP 01/31/90 MS G4/12/90 / / 900103 82050320 4 0 MAR REG 1.97 HIGH PRESSURE UPGRAN / / 03/09/90 OP 03/23/90 MS 11/28/90 / / 900149 82050317 8 0 MAR R.G. 1.97 PRESSUR12ER HEATER STATUS !NDICA110M / / 04/13/89 OP 08/15/90 MS 05/23/89 / / 900004 82050316 8 0 MAR MAIN STEAM RELIEF VALVE POSITION INDICATION / / 01/16/89 OP 02/08/89 HS 09/10/90 / / 900039 82050307 9 FCN REG GUIDE 1.97 MOD COMPUTER / RECALL INSTALLATION / / 03/22/88 OP 01/06/88 s 06/02/88 / / 900127 82050307 4 FCN RESCALE POINT x-333 / / 02/12/91 OP 01/06/88 S 02/12/91 / / 910020 82050306 23 FCN REW1RE CIRCS RCR 327 & 331 TO PROVIDE 150LAT104 / / 06/28/88 OP 12/27/89 5 09/10/90 / / 900144 82050301 21 FCN CORRECT COMPUTER POINT x-333 / / 02/'2/91 OP 01/11/89 S 02/12/91 / / 910019 82050301 15 FCN REVISE DATA SHEETS / / 02/12/91 OP 01/?t/89 5 02/12/91 / / 910018 80121403 8 0 MAR MSR TUSE BLWDLE INSTRUMENTATION INSTALL. / / 09/15/88 AC 05/08/89 5 09/10/90 / / 900045 80031801 4 FCN REVISE REFERENCE LEG DIMENSION / / 02/12/91 OP 11/21/89 S 02/12/91 / / 910017 THC TOTAL NtMBER OF RECORDS FOR THIS REPORf IS: 87 END OF REPORT
FLORIDA POWER CORPORATION f~T- CRYSTAL RIVERUNIT 3 51MULATOR '~ CERTIFICATION SUBMITTAL
&E.P_EED_Ill NUMBERED MALFUNGTIONS MALFUNCJ10N NUMBB I_IILE BS-001 RX BUILDING SPRAY PUMP TRIP CD 001 HOTWELL CROSSOVER PIPE BREAK CD 002 HOTWELL LEVEL CONTROLLER FAILURE CD-5 LC CD-003 CONDENSATE PUMP TRIP CD-004 #1 CONSENSATE HEATER TUBE LEAK CD 005 #2 CONDENSATE HEATER TUBE LEAK CD-006 #3 CONDENSATE HEATER TUBE LEAK CD-007 CONDENSER AIR IN-LEAKAGE CF-001- CORE FLOOD TANK GAS SPACE LEAK CF-002 CORE FLOOD TANK WATER SPACE LEAK CF-003 RCS LEAKAGE INTO CORE FLOOD TANK CR-001 STUCK CONTROL R0D (ANY OR ALL)
CR-002 DROPPED CONTROL R0D (ANY OR ALL) CR-003 CONTROL R00 MECHANICAL BINDING (ANY OR ALL) CR-004 UNC0VPLED CONTROL R0D (ANY OR ALL) 73 CR-005 DEFEAT CONTROL R00 GROUP COME.AND ENABLE (_/ CR-006 --DELETED-- CR-007 --DELETED-- CR-008 BLOCK CONTROL R00 "00T INHIBIT" CR-009 BLOCK CONTROL R00 " INSERT / WITHDRAW" COMMANDS CR-010 --DELETED - CW-001 CIRCULATING WATER PUMP MOTOR TRIP CW-002 CONDENSER TUBE LEAK DC-001 DECAY HEAT CLOSED CYCLE COOLING PUMP MOTOR TRIP DC-002 DECAY HEAT CLOSED CYCLE COOLING SYSTEM LEAK DC-003 DECAY HEAT CLOSED CYCLE COOLING HEAT EXCHANGER TUBE LEAK DH-001 DECAY HEAT REMOVAL PUMP MOTOR TRIP DH-002 DECAY HEAT REMOVAL SYSTEM LEAK DH-003 DECAY HEAT REMOVAL COOLER TUBE LEAK DH-004 DECAY HEAT SUCTION ISOLATION VALVES FAIL CLOSED ED-001 LOSS OF STARTUP TRANSFORMER ED-002 LOSS OF 6.9 KV BUS ED-003 LOSS OF UNIT 4.16 KV BUS ED-004- LOSS OF ES 4.16 KV BUS ED-005 LOSS OF ES 480 V BUS ED-006 LOSS OF REACTOR 480 V BUS ED-007 LOSS OF TURBINE 480V BUS ED-008 LOSS OF INTAKE 480V BUS ED-009 LOSS OF HEATING 480V BUS ED-010 LOSS OF PLANT AUX 480V BUS ED-011 LOSS OF VITAL BUS /~' \') ED-012 ED-013 LOSS OF RX 480V MCC LOSS OF TURBINE 480V MCC
FLORIDA POWER CORPORATION CRYST AL RIVERUNIT 3 SIMULATOR I~)'
'~
CERTIFICATION SUBMITTAL AP.fEHQ1LE NUMBERED MALEMHCTIONS(Continued) MALFUNCTION NUfLBEB IjlLE ED-014 LOSS OF WATER TREATMENT 480V MCC ED-015 LOSS OF VENTILATION 480V MCC ED-016 LOSS OF INTAKE 480V MCC ED-017 LOSS OF PRESSURIZER 480V MCC ED-018 LOSS OF ES 480V MCC ED-019 LOSS OF DC DISTRIBUTION ED-020 BLOCK AUTO BUS TRANSFER FROM AUXILIARY TO STARTUP TRANSFORMER EF-001 EMERGENCY FEEDWATER PUMP TRIP EF-002 LOSS OF EFIC POWER SUPPLY EF-003 EMERGENCY FEEDWATER PUMP FAILS TO START EF-004 EFIC BAILEY H/A STATION FAILURE EF-005 EFIC AUTO ACTUATION FAILURE EF-006 MAIN FEEDWATER ISOLATION ACTUATION FAILURE EF 007 MAIN STEAM ISOLATION ACTUATION FAILURE EG-001 EMERGENCY DIESEL GENERATOR Fall TO START p), (_ EG-002 EMERGENCY DIESEL GENERATOR TRIP EG-003 ERRATIC EMERGENCY DIESEL GENERATOR V0LTAGE CONTROL EG-004 BLOCK EMERGENCY DIESEL GENERATOR FIELD FLASH EH-001 MAIN TURBINE TRIP EH-002 BLOCK MAIN TURBINE TRIP EH 003 BLOCK ICS TO INPUT ' REFERENCE" CONTROL EH-004 BLOCK TRANSFER TO LOAD CONTROL EH 005 Fall EHC MEGAWATT TRANSDUCER LOW EH-006 Fall EHC MAIN SPEED CHANNEL LOW EH-007 TURBINE GOVERNOR VALVE FAILURE EH-008 ' TURBINE THROTTLE VALVE FAILURE EH-009 EHC FLUID PUMP MOTOR TRIP EH-010 EHC UNL0ADER FAILURE EH-Oll EHC SYSTEM FILTERS CLOG ES-001 SPURIOUS TRIP OF HPI CHANNEL ES-002 SPURIOUS TRIP 0F LPI CHANNEL ES-003 SPURIOUS TRIP 0F REACTOR BUILDING ISOLATION & COOLING E5-004 FAILURE OF ES CA'ilNET DOWER SUPPLY ES-005 FAILURE OF +15 V 1C ES CABINET POWER SUPPLY ES-006 FAILURE OF -15 V C ES CABINET POWER SUPPLY ES-007 ES CHANNEL FAILURE TO ACTUATE HPI ES-008 ES CHANNEL FAILURE TO ACTUATE LPI ES-009 ES CHANNEL FAILURE TO ACTUATE RB ISOLATION & COOLING ES-010 ES CHANNEL FAILURE TO ACTUATE RB SPRAY ES-Oll DIVERSE CONTAINMENT ISOLATION FAILURE
~r N FW-001 MAIN FEEDWATER LINE BREAK INSIDE RB '- FW-002 MAIN FEEDWATER LINE BREAK OUTSIDE RB l
l
FLORIDA POWER CORPORATION CRYSTAL RIVERUNIT 3 SIMULATOR (]) CERflFICATION SUBMITTAL AP.EENDJ2__E NUMBERED MALFURCTIONS(Continued) MALFUNCTION NUMBER 111LE FW-003 MAIN FEEDWATER LINE BRiAK OUTSIDE RB FW-004 MAIN FEEDWATER B0OSTER PUMP TRIP FW-005 MAIN FEEDWATER PUMP TRIP FW 006 MAIN FEEDWATER BLOCK VALVE FAILS AS IS FW-007 MAIN FEEDWATER LOW LOAD BLOCK VALVE FAILS AS IS FW-008 MAIN FEEDWATER STARTUP FLOW TRANSMITTER FAILURE FW-009 MAIN'FEEDWATER FLOW TRANSMITTER FAILURE FW-010 MAIN FEEDWATER TEMPERATURE TRANSMITTER FAILURE FW-Oll MAIN FEEDWATER VALVE TRAIN DP TRANSMITTER FAILURE FW-012 #5 FEEDWATER HEATER TUBE LEAK FW-013 #6 FEEDWATER HEATER TUBE LEAK FW-014 MAIN FEEDWATER PUMP AUTO TRIP FAILURE FW-015 STARTUP FEEDWATER BLOCK VALVE Fall AS IS FW-016 MAIN FEEDWATER PUMP-SUCTION VALVE Fall AS IS FW-017 OTSG STARTUP LEVEL TRANSMITTER FAILURE FW-018 OTSG OPERATING LEVEL TRANSMITTER FAILURE i FW-019 OTSG HIGH RANGE LEVEL TRANSMITTER FAILURE FW-020 OTSG LOW RANGE LEVEL TRANSMITTER FAILURE GS-001 GLAND STEAM DRAIN TANK LEVEL CONTROL VALVE FAIL OPEN HD-001 HEATER DRAIN LEVEL CONTROLLER FAILURE IA-001 UNIS0LABLE LEAK IN INSTRUMENT AIR HEADER 1C-001 MAIN BLOCK VALVE FAILURE (FWV-29/FWV-30) 10-002 BLOCK FEEDWATER CROSS 0VER VALVE 0% LIMIT SWITCH INPUT 1C-003 BLOCK ICS RUNBACK CAPABILITY 10-004 ICS HEADER PRESSURE BIAS FAILURE (0, 50, 125) 10-005 BLOCK TRANSFER OF Tave CONTROL (10 FEEDWATER/TO REACTOR) 10-006 BLOCK ICS TRACKING FUNCTION 10-007 REACTOR DEMAND SIGNAL FAILURE TO ZERO VOLTS 10-008 TOTAL FEEDWATER DEMAND FAILS TO ZERO VOLTS 10-009 LOOP FEEDWATER DEMAND FAILS TO ZERO VOLTS 1C-010 STEAM GENERATOR / REACTOR DEMAND FAILS TO ZERO VOLTS 10-011 STARTUP CONTROL VALVE STATION FAILS TO ZERO VOLTS 10-012 LOW LOAD CONTROL VALVE STATION FAILS T0 ZERO V0LTS 10-013 FEEDWATER PUMP CONTROL STATION FAILS TO ZERO VOLTS 1C-014 BLOCK FEEDWATER TRANSFER TO FLOW CONTROL l IC-015 TURBINE-BYPASS VALVE CONTROL STATION FAILS TO ZERO V0LTS 1C-016 FEEDWATER CONTROL VALVE ERRATIC CONTROL
- 1C-017 ICS POWER FAILURE
! IC-018 Fall ICS Tc INPUT TO ZERO VOLTS 10-019 FAIL ICS N1 INPUT TO ZERO V0LTS 1C-020 Fall ICS Tave INPUT TO ZERL VOLTS 10-021 Fall ICS MEGAWATT INPUT TO ZERO VOLTS l (]) 1C-022 Fall ICS LOOP FEEDWATER FLOW INPUT TO ZERO VOLTS I
FLORIDA POWER CORPORATION CRYSTAL R!VERUNIT 3 SIMULATOR O- CERTIFICATION SUBMITTAL APPEND [1_f NUM BLRID_tlAlBBCT I ON S (Con t i nu e d ) MALFUNCTION Hl)M ER llILE 10-023 Fall ICS OTSG PRESSURE INPUT 10 ZERO VOLTS 10-021 Fall ICS OTSG STARTUP LEVEL INPUT TO ZERO VOLTS MS-001 STEAM LINE BREAK IN RB MS-002 STEAM LINE BREAK OUTSIDE RB MS-003 MAIN STEAM SAFETY VALVE LEAKS / FAILS 10 RESEAT MS-004 MAIN STEAM ISOLATION VALVE Fall CLOSED MS 005 ATMOSPHERIC REllEF VAtVE FAILURE MS-006 TURBINE BYPASS VALVE FAILURE MS-007 TURBINE HEADER PRESSURE TRANSMITTER FAILURE MS-008 STEAM GENERATOR PRESSURE TRANSM11TER FAILURE MS-009 EFIC OTSG PRESSURE TRANSMITTER FAILURE MU-001 MAKE-UP PUMP MOTOR TRIP MU-002 LETDOWN ISOLATION VALVE FAIL CLOSED (MUV-49) MU-003 PRESSURIZER LEVEL CONTROL VALVE FAILURE (MUV-31) MU-004 SEAL INJECTION CONTROL VALVE FAILURE (MUV-16) MU-005 LETDOWN COOLER TUBE LEAK _) MU-006 LETDOWN LINE RUPTURE INSIDE RB MU-007 LETDOWN LINE RUPTURE IN AUX BUILDING DOWNSTREAM MVV-49 MU-008 NORMAL MU LINE BREAK IN RB MU-009 MAKE-UP HEADER BREAK AT MVP DISCHARGE NN-001 LOSS OF NNI +24 VDC AND -24 VDC NN-002 REMOTE SHUTDOWN POWER Fall LOW NN-003 NNI 118 VOLT FIELD SUPPLY FAILURE FAIL LOW NN-004 SELECTED Tave FAILURE RC-001 HOT LEG BREAK RC-002 COLD LEG BREAK RCP SUCTION RC-003 COLD LEG BREAK RCP DISCHARGE RC-004 OTSG TUBE LEAK RC-005 PORV FAILURE RCV-10 RC-006 RCS SAFETY VALVE FAILURE RC-007 RE/'. TOR COOLANT PUMP TRIP RC-008 REALTOR COOLANT PUMP SHAFT BREAK RC-009 REACTOR COOLANT PUMP LOCKED ROTOR RC-010 RCP FIRST STAGE SEAL FAILURE RC-Oll RCP SECOND STAGE SEAL FAILURE RC-012 RCP THIRD STAGE SFAL FAILURE RC-013 PRESSURIZER L)W LEVEL INTERLOCK (Fall TO ACTUATE) RC 014 PRESSURIZER HEATER CONTROLLER FAILURE RC-015 PRESSURIZER LEVEL TRANSMITTER FAILURE RC-016 RCS NARROW RANGE PRESSURE TRANSMITTER FAILURE RC-017 RCS WIDE RANGE PRESSURE TRANSMITTER FAILURE RC-018 RCS NARROW RANGE Th FAILURE (]) RC-019 RCS NARROW RANGE Tc FAILURE
FLORIDA POWER CORPORATION CRYST AL RIVERUNIT 3 SIMULA10R (]) CERTlflCATION SUBMllTAL APE M il.1 El@)MD HALFUNCT10NS(Continued) ILA.lf0NCT10NJURG IIILE RC-020 PRESSURIZER TEMPERATURE TRANSMITTER FAILURE RC-021 PRESSURIZER SPRAY VALVE FAILURE RC-022 RCS Tave FAILURE RP-001 SOURCE RANGE NI DISCRIMINATOR FAILURE RP-002 50VnCE RANGE N1 DElECTOR POWER SUPPLY FAILURE Fall LOW RP-003 INTERMEDIATE RANGE N1 COMPENSATION POWER SUPPLY FAILURE RP-004 INTERMEDIATE RANGE N1 DETECTOR POWER SUPPLY F61 LURE Fall LOW RP-005 POWER RANGE N1 POWER SUPPLY FAILURE Fall LOW RP-006 POWER RANGE NI SUMMING AMP FAILURE RP-007 POWER RANGE NI DIFFERENTIAL AMP FAILURE RP-008 LOSE JF +1SV RPS CABINET POWER SUPPLY RP-009 LOSS OF -lSV RPS CABINET POWER SUPPLY RP-010 LOSS OF 120V RPS CABINET POWER SUPPLY RP-Oll REACTOR TRIP MODULE FAILS TO TRIP RP-012 REACTOR TRIP MODULE TRIP O RW-001 SC-001 RAW WATER PUMP MOTOR TRIP SECONDARY SERVICES CLOSED CYCLE COOLING SYSTEM LEAK SC-002 SECONDARY SERVICES CLOSED CYCLE COOLING PUMP MOTOR TRIP SC-003 GENERATOR HYDROGEN COOLER CONTROL VALVE FAILURE SC-004 HYDROGEN COOLER TUBE LEAK SC-005 SECONDARY SERVICES CLOSED CYCLE COOLING HEAT EXCHANGER TUBE LEAK SW-001 NUCLEAR SERVICES CLOSED CYCLE COOLING SYSTEM LEAK IN AUX BUILDING SW-002 NUCLEAR SERVICES CLOSED CVCLE COOL!NG SYSlEM LEAK IN RB SW-003 NUCLEAR SERVICES C10 SED CYCLE COOLING PUMP MOTOR TRIP SW-004 LOSS OF HUCLEAR SERVICES CLOSED CYCLE COOLING TO RCr SW-005 LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING TO LETDOWN COOLERS SW-006 NUCLEAR SERVICE: E12*ED CYCLE COOLING HEAT EXCHANGER TUBE LEAK SW-007 RC DRAIN TAN 4 COOLER LEAK SW-008 RCP INTERNAL COOLER TUBE LEAK O ,
FLORIDA POWER CORPORA 110N I O' CRYSTAL RIVER UNIT 3 SIMULATOR CERTIFICATION SUBMITTAL APPENDlX F SIMULATOR RANQl.ERS prNTROL SYSTEM HANALERS
- 1. 2 INPUT AUCTIONEER Remote functions: Polarity switch SI status
- 2. 3 INPUT AUCTIONEER Remote functions: Polarity switch SI status, polarity switch S2 status
- 3. ANALOG MEMORY UNIT
- 4. FUNCTION GENERATOR Remote Functions: Breakpoint coordinates f- 5. PULSER
( Remote functions: Deadband voltage, maximuia triangular wave voltt;3
- 6. RATE '.lMll' SIGNAL FOLLOWER Remote functions: Positise slope setpoint, negative slope setpoint 7, SUMMER + BIAS Remote functions: Bias setpoint, output gain, gair. on input #3, gain on input #4, gain on input #5, switch Sl status
- 8. SIGNAL CONVERTER
- 9. SIGNAL GENERATOR Remote functions: Calibrated output signal
- 10. SUMMER + INTEGRAL Remote functions: Gain on input #3, gain on input #4, bias setpoint, integrating rate, switch S1 status,
(] integrating rate x10, select configuration pin 5 - pin 9
FLORIDA POWER CORPORA 110N O cars'^' aivca u"I' 3 sinu'^ ton CERilflCA110N SUBM11TAL , APPENDIX f 1))iULATOR ll&f!011R$(Conttnued) p.Q!i1RQLSY$ TEM HA'f0LERS
- 11. SIGNAL LlHITER Remote functions: High limit, low limit
- 12. $1GNAL MONITOR Remote functions: Positive trip setpoint, negative trip setpoint, deenergized or energized operation
- 13. SUMMER + PROPORTIONAL + INTEGRAL Remote functions: Gain on input #3, gain on input #4, bias ,
setpoint, pror9rtional gain, switch S1 status, integrating rate x10, proportional gain x5
- 14. SQUARE ROOT EXTRACTOR Remote functions: Select polarity
- 15. RESISTANCE TEMPERATURE DETECTOR failures: fail to position, drift, noise, lag, loss of power, f ail high, fail lu, fail as-is 16, GENERAL TRANSMITTER failures: Fall to position, drift, noise, lag, loss of power, fail high, fail low, fail as-is
- 17. PNEUMATIC 1RANSDUCER failures: f ail to position, drift, lag, f ail low, fail high, fail as-is, loss of air
- 18. THERH000VPLE r
failures: fail to position, drift, noise, lag, offset, 1, as of power, fail high, fail low, f ail as-is
- 19. GENERAL TRANSMITTER failures:
Q fail to positien, drift, noise, lag, offset, loss of power, fail high, fail low, fail as-is '
-- +a w 'e'- % vy 1-x w- p tuwwa- wg yr -
y-- -- y - w-
.- . _- _ =. _-_ _ __ - - _- . . . . ._-_ - .- _ _ - . . .
FLORIDA POWLR CORPORA 110N O cavst^t aivca u">' 3 sinu'^'on CERTiflCA110N SUBM111AL l AEEMDllf l 111MALOR H6801ER$(Continued) GMIRQL $YSTEH llMDLERS
- 20. TlHE DELAY Remote functions: Time delay select, multiplication factor
- 21. TRI-STABLE RELAY l
Remote functions: Positive trip setpoint, negative trip setpoint, l doenergized or energized operation
- 22. $PLIT RANGE HULTIPLIER Remote functions: Gain high limit, gain low limit
- 23. RATE LlHITED SIGNAL FOLLOWER Remote functions: Limiter on/of f ELECTRLCAL 5YSTE3_tl6!iQlf.[Li
- 24. AIR CIRCUIT BREAKER failures: fall closed, fail open, fail as-is, fail to reset
- 25. OIL circuli BREAKER failures: fail closed, fail open, fail as-is, fail to reset
- 26. 4,16 KV DHP BREAKER failures: fail closed, fail open, fail as-is, fail to retet Remote functions: Select breaker position (test, operate, rack-out)
- 27. 4B0 V AK25 BREAKER failures: Fail closed, fail open, fall as-is kemote functions: Local pushbutton close, local mechanical trip
- 28. 4B0 V AK50 BREAKER failures: f ail closed, fail open, f ail as-is, f ail to reset Remote functions: Local pushbutton close, local raechanical trip
FLORIDA POWER CORPORA 110N O carsi^' aivca unit 3 sinut^1oa CER11flCA110N SUBMIT 1AL APfJNDIX F i
$ltn!LALOR HANDLER $(Continued) t UICIRlCAL SYSl[M HANQL[ES
- 29. 480 V ITE BREAKER Failures: Fail closed, fail open, fail as is, fail to reset Remote functions: Select breaker position (test, operate, rack-out) 30: 6.9/4.16 KV BREAKER Failures: Fail closed, fail open, fail as-is, fall to reset Remote Functions: Select breaker position (test, operate, rack out)
- 31. TDPU TIME DELAY RELAY
- 32. TODO TIME DELAY RELAY
- 33. BISTABLE RELAY O 34 BISTABLE MODULE Failures: Fail ts is, fails to trip, fails to reset Remote functions: Bistable trip
- 35. OVERCURRENT RELAY 51 TYPE C0H-5 Failures: Fails to actuate, delayed actuation
- 36. OVERCURRENT RELAY 51 TYPE CO-2 Failures: Fails to actuate, delayed actuation
- 37. OVERCURRENT RELAY 51 TYPE CO-7 Failures: Fails to actuate, delayed actuation
- 38. OVERCURRENT RELAY 51 TYPE CO-9 Failures: Fails to actuate, delayed actuation
- 39. TIME DELAY /INSTANTANE0US OVERCURRENT RELAY TYPE LAC 66K Failures: Fails to actuate, delayed actuation l
O
FLORIDA POWLR CORPORA 110N O cavst^' aivca unii 3 sinutaion CER11 FICA 110N SUBMIT 1AL AITINQJLE SJ M ATOR HANDLER $(Continued) R ECIRl0A k i1&IEM HANDLERS
- 40. DEf! NITE TIME UNDERVOLTAGE RELAY 27 ITE ,
failures: f ails to actuate, delayed actuation
- 41. OVERCURRENT RELAY 51 TYPE C0-5 failures: f ails to actuate, delayed actuation
- 42. OVERCURRENT RELAY 51 TYPE CO 11 Failures: fails to actuate, delayed actuation
- 43. THERMAL OVIRLOAD RELAY 49 (>50 HP) failures: Spurious actuation, fails to actuate, delayed actuation
- 44. THERMAL OVERLOAD RELAY 49 (<50 HP) failures: Overload failure
- 45. SYNCHR0 VERif!ER RELAY failures: f ails to actuate, delayed actuation
- 46. SYNCHR0 CHECK RELAY Failures: fails to actuate, delayed actuation
- 47. INSTANTANE0US 0/C RELAY 50 failures: fails to actuate, delayed actuation
- 48. TIME UNDERVOLTAGE RELAY GE failures: fails to actuate, delayed actuation 49, 86 LOCKOUT RELAY ELECTRICALLY RESET failures: Fails to actuate, delayed actuation O
FLORlDA POWER CORPORATI0ff O cavst^t a uta unii 3 siMutaion CERliflCATION SUBMillAL AEffhpjX F SIMULATORHANDL[R$(Continued) ILLC1RLCAALLY31ItU1ARD11R1
- 50. 86 LOCK 0UT RELAY MANUALLY RESET Failures: Fails to actuate, delayed actuation
- 51. OVERCURRENT RELAY TYPE IAC53A(B)
Failures: Fails to actuate, delayed actuation
- 52. TlHE UNDER/0VERVOLTAGE RELAY failures: fails to actuate, delayed actuation
- 53. TIME UNDER/0VERVOLTAGE RELAY GE Failures: Tails to actuate, delayed actuation O 54. TlHE UNDER/0VERV0 GAGE REMY TYPE MV53A(D)
Failures; fails to actuate, delayed actuation
- 55. THERMAL OVERLOAD RLLAY failures: Overload failure
- 56. RECTIFIER
- 57. VOLTAGE REGULATOR
- 58. BATTERY
- 59. ANALOG INDICATOR failures: Power loss, signal loss, fail high, fail low, fail as-is, drift
- 60. SYNCHROSCOPE DRIVER
- 61. FUSE STATUS Remote functions: fuse removal / reset
FLORIDA POWER CORPORA 110N Q CRYSTAL RIVER UNIT 3 SIMULATOR CER11 FICA 110N SUBMITTAL b1ECH0.1L1 11MQLA10LBMIRLER$(Continued) MECHAMLCAL11511LlLAMLERS
- 62. CENTRIFUGAL PUMP failures: Bearing friction
- 63. P051TIVE DISPLACEMENT PUMP failures: Bearing friction
- 64. LINEAR TANK failures: Leak
- 65. HON-LINEAR TANK failures: Leek O 66. s1EAn 1 RAP failures: fail closed, fail open, fail as is, fail to specified position
- 67. FAN
- 68. ION EXCHANGER failures: Channeling
- 69. FIXED POWER HEATER failures: Short circuit, open circuit
- 70. VARIABLE POWER HEATER failures: Short circuit, open circuit
- 71. HEAT EXCHANGER failures: fouling Remote functions: Select ambient temperature O
FLORIDA POWER CORPORATION O cavst^' aivca u"it 3 sinut^ ton CERTiflCAT10N SUBMllTAL APPENQ1LE ilfit(ATOR HANQLEM(Continued) E0]iANICAL SYSTEM HANDLERS
- 72. COOLER failures: fouling Remote functions: Select ambient temperature
- 73. PNEUMATIC FIELD CONTROLLER failures: fail high, fail low, fail as-is, loss of input signal Remote functions: Proportional gain, reset time
- 74. BAILEY TYPE 701 CONTROLLER failures: fail high, fail low, fail as-is, loss of input signal Remote functions: Proportional gain, reset time
- 75. BATCH CONTROLLER failures: Fail high, fail low, fail as-is, loss of input signal Remote functions: Valve holding, ramp shutdown slope
- 76. PNEUHATIC INDICATOR failures: Internal power loss, signal loss, fail high, fail low, fail as is, drift
- 77. LARGE INDUCTION HOTOR failures: Sheared shaft, shaft seizure
- 78. SMALL INDUCTION H010R failures: Sheared shaft, shaft seizure
- 79. SIMPLIFIED PUMP / MOTOR failures: Sheared shaft, bearing failure
- 80. SHUNT FIELD DC HOTOR O Failures: Sheared shaft
FLORIDA POWER CORPORA 110N O cavs'^t nivER unit 3 SinulATOR CERTiflCA110N SUBM11TAL 8PJLNDR_E
$180LAl0R HANQlER$(Continued)
MECRAljCAL SYSTEM HANDLER $
- 81. ELECTRICAL SWITCH Failures: Power loss, signal loss
- 82. MECHANICAL SWITCH WITH FAILURES Fa'1ures: Fail open, fall closed, fail as is
- 83. HECHANICAL SWITCH WITHOUT FAILURES
- 84. DUAL SETPOINT MECilANICAL SWITCH
- 85. TRIPLE SETPOINT HECHANICAL SWITCH
- 86. AIR TO-0 PEN CONTROL VALVE O raiiures: raii open, faii ciosed, rei, as-is, faii to specified position
- 87. AIR TO-CLOSE CONTROL VALVE Failures: Fail open, fail closed, fail as-is, fall to specified position
- 88. MOTOR-OPERATED VALVE Failures: fail open, fail closed, fail as is, fail to specified position
- 89. SAFETY /REllEF VALVE Failures: Fail open, fail closed, fail as-is, fail to specified position
- 90. CliECK VALVE Failures: Leak by
- 91. AIR-ACTUATED CHECK VALVE failures: Leak-by, fail open, fail closed, fall as-is, fail O to sPecified Positio"
FLORIDA POWER CORPORA 110f4 O cavst^t RivcR Uriii 3 SinutATOR CER11FICAT10f4 SUBM111AL Afffl4Q1LE
$111MLA10R HANDLEAS(Continued)
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- 92. HOTOR OPERATED CHECK VALVE failures: Leak by, fail open, fail closed, fail as is, fail to specified position
- 93. AIR AND HOTOR OPERATED CHECK VALVE Failures. Leak-by, f ail open, f ail closed, Tail as is, f all to specifica position
- 94. AIR-ASSISTED CHECK VALVE Failures: Leak-by, fail open, fail closed, fail as-is, fail to specified position
- 95. SPECIFIC VALVE PORT AREA Failures: Seat leakage
- 96. GENERIC VALVE PORT AREA Failures: Seat Leakage
- 97. DOWNSTREAM PRESSURE REGULATOR Failures: Fail open, fall closed, fail as-is, fail to specified position
- 98. UPSTREAM PRESSURE REGULATOR failures: Fail open, fail closed, fail as-is, fall to specified position O
l
Pope No, i PA R f DRMANCE 1E 515 09/D9/91 Test REV SIAT 111LE f RE Q AhS_RLF , P'C1 1 CU COMPUTER REAL 11ME 1E$1 1 A3.1 PIM10 1 CU TOTAL Lots OF IEEDWATER 4 3.1.2(10), B2.2(2) PIM11A i CU LOSS OF *1$v CABikEl POWEk SUPPLY 1 3.1.2(11) P1M118 1 CU F AILURE OF El CAlthEl POWER SUPPLY 2 3.1.2(11) P1M12A 1 CU MECHAh!CALLY $1UCK R00 3 3.1.2(12) PTM128 i CU UNCDUPLED ROD 4 3.1.2(12) P1M12C 1 CU DECRADED kpD M0110N 1 3.1.2(12) P1M120 i CU DRDPPID CONTROL ROD 2 3.1.2(12) P1M13 1 CU lhABIL11Y 10 DRIVE CONTROL RODS 3 3.1.2L13) PTM14A 1 CU MIGH RCS ACilV11Y 4 3.1.2(14) PIM14B 1 CU WA51E CAS DECAV 1 AkK LE AK 1 3.1.?(14) PIM16 i CU MAIN GEhtRA1DR 1 RIP 1 3.1.2(16) P1M17A 1 CU CDNfROL ROD 0U1 IkHlBit FAILURE 2 3.1.2(17) P1M178 i CU CDNTROL ROD CROUP CDMMAND ENABLE f AILURE 3 3.1.2(17) - P1M17C 1 CU FAILURE 08 1HE MAIN TURB!ht 101 kip 4 3.1.2(17) PIM17D 1 CU ATMOSPHERIC DUMP VALVE FAILURE 1 3.1.2(17) PTMi?E i CU TURBlhE BYtAts VALVE FAILURE 2 3.1.2(17) P1M17f 1 CU CDh11NUOU$ kOD W11HDRAWL 1 3.1.2(17) PIM17G 1 CU DILUllDN AtCIDENT 1 3.1.2(17) PTM1BA 1 CU MAKEUP PUMP 1 RIP 3 3.1.2(18) PTMiB8 1 CV PRESSURl!Ek LEVEL CON 1ROL VALVE FAILURE 4 3.1.2(18) PTM18C 1 CV LE1DOWN l%0LellDN VAL'VE FAILURE 1 3.1.2(18) P1M1bD i CU PRESSUR12ER HEA)ER COW 1 ROLLER F AILURE 2 3.1.2(18) PTM1BE 1 CU FRESCURIItt SPRAY VALVE FAILURE 3 3.1.2(18) P1M18F 1 CV RCS ht FRE55URE TRANSM111ER FAILUkE 4 3.1.2(18) PIM1A1 1 CU $1E AM LEhERAIDR TUBE LE AK 1 3.1.?(1)(a) O PIM1A2 P1M1B 1 CU 1 CU STEAM GENERA 1DR TUBE RUP1URE LE1DOWN LEAK DOWNSTREAM OF MUV 49 2 2 3.1.2(1)(e) 3.1.2(b)(c) PIM1C1 i CU REPRESSUR12thG SMALL BREAK LOCA 3 3.1. 2(1)(b)(c ) PIM1C2 1 CU SMAtt BetAK LDCA 3 3.1.2(1)(b)(c) PIM1C3 1 CU SMALL BRE AK LDCA W11H HPl lhDFERABLE 3 3.1.2(1)(b)(c) P1M1D 1 CU FRE55URl!ER SAf ETY VALVE F AILURE 4 3.1.2(1)(d) PIM2 i CU Loss of INSTRUMEN1 AIR 1 3.1.2(2) PTM?0A 1 CU MAIN STEAM Llkt BREAK DU1 SIDE CONTAINMEh1 1 3.1.?(20) _ PTM208 1 CU MAIN FEED LlhE BREAK INSIDE CONTAINMEk1 2 3.1.2(20) PIM20C 1 CU MAIN FEED LINE BREAK OU15tDE CONTAINMEN1 3 3.1.2(20) P1M21A 1 CU POWER RANGE CHANNEL FAILURE A 3.1.2(21) PIM218 1 CU IN1EPMCDIA1E RANCE CHAbNEL FAILUWE 1 3.1.2(21) PTM21C 1 CU SOURCE RAhGE CHANNEL FAILURE 2 3.1.2(21) PIM224 1 CU RC! 1,M01 1R ANSM111ER F AILURE 3 3.1.2(22) PTM228 1 CU RCS 1 , COLD TRANSM111ER FAILURE 4 3.1.2(22) PTM22C 1 CU MAIN FW FLDW TRANSM111ER FAILURE 1 3.1.2(22) PIM22D 1 CU 01SG OP RANGE LEVEL TRANSM111ER FAILURE 2 3.1.2(22) PTM22E 1 CU ICS RUNBACK FA! LURE 3 361.2(22) PTM22F 1 CU ISC HE ADER PRESSURE BI AS f AILURE 4 3.1.2(22) PTM22G 1 CU MAIN FE TEMFERATURE TRANSM111ER FAILURE 1 3.1.2(22) PIM22M i CU f AILURE OF 1C5101RANSf ER 101 RACK 2 3.1.2(22) P1M221 1 CU REACIOR DEnAND SIGNAL FAILURE 3 3.1.2(22) PTM22J 1 CU - $ELEC1ED 1 AVE F AILURf 4 3.1.2(22) PIM23A 1 CU MS ISOLA 110N ACTUA110N fAILUNE i 3.1.2(23) PTM238 1 . CU MFW ISOL AllDN ACTUA110N F AILURE 2 3.1.2(23) PIM23C 1 CU ES CHAhNEL FAILURE 10 ACTUATE HP] 3 3.1.2(23) PIM23D 1.CU E5 CHAANEL FAILURE 10 ACTUA1E RBl&C 4 3.1.?(23) PIM23E 1 CU EFIC AUTO ACTUAllDN FAILURE 1 3.1.2(23) O PIM244 P1M248 i CU 1 CU A1WS (ho OPER410R INTEkvEN110N) A1WS (OPERATOR INTERVENTION) 2 3 3.1.2(24) 3.1.2(24) l l
I Page No. 2 Pikl0&MAhCf lisil 09/09/91 1151 RIV $1A1 Illit f kl0 AkS_kti PTMIA 1 CU LOS$ Of 0Ff $11t PWR (LOW DICAV ht A1) 2 3.1.2(3) PIM361 1 CU LOS$ Of futRGINCY Off$ lit 50($Ct 3 3.1.?(3) PfM382 i CU Lost OF (MtRGINCY Of f $ lit $0VRCf /IDG ! AlLURt 3 3.1.?(3) PTM3C1 1 CU Loss of kNI Y + & - 24 VDC 4 3.1.i(3)(22) PIN 3C2 1 CU LOS$ Of kN1 f 120 VAC 1 3.1.2(3) PIM3D1 1 CU LO$$ of IC$ DC POWik 2 3.1.2(3) P1M3D2 i CU LO$$ Of ICS AC 00WtR 3 3.1.2(3) PIM3t i CU LOS$ Of Vil AL Bus 4 3.1.2(3) PIM3F 1 CU Lost OF 480 vaC t$ BUS A 1 3.1.2(3) PIM3G 1 CU L0ss 07 DC DistRibu110N 2 3.1.2(3) . PTM4A 1 CU RtAC10R COOLANT PUMP $ HAFT $ktAt 3 3.1.2(4) I PIM4B 1 CU L0ss OF 2 RCPs 4 3.i.2(4) l PIM4C 1 CU RtACTOR COOLANI PUMP $tAL fAILURI i 3.1.?(4) i PfM4D 1 CU Rt AC104 COOL ANT PUMP LOCKID A010R 1 3.i ?(4) PTMSA 1 CU Lost 0F CONDth$tR VACUUM 2 3.1.2 d ) PTM5B 1 CU LO$$ OF CONDth$lt Ltytt CONTROL 3 3.1.2(), P1M6A 1 CU Loss of huCLE AR $lRVICil RAW WAf tR 4 3.1.2(6) PIM6B 1 CU LO$5 Of CIRCULAtlNG WATER 1 3.1.2(6) PfM6C 1 CU LOS* OF DICAY HE AT RAW WAlta 2 3.1.2(6) PfM60 1 CU PAPilAL LO$$ OF CIRCULATING WAltR 3 3.1.2(6) PIM7 i CU LO$$ OF DECAY NE AT RLMovAL 4 3.1.2(7) P1MBA i CU LOS$ OF N$CCC i 3.1.2(8) PIMBB 1 CU Los$ Of $$CCC 2 3.1.2(B) PIM8C 1 CU Loss of DHCCC 3 3.1.2(B) PfM9A 1 CU MAIN titDWAtta PUMP 1 RIP AND kub6ACK 4 3.1.2(9) Pir7B i CU LO$$ Of BOTH MAIN ILIDWAftR PUMPS 1 3.1.2(9) 1 CU 3.1.2(9) O PIM9C PTM?D PIN) 1 CU 1 CU 01$G OVERFILL Los$ OF BOTH CONDth$ Aft PUMPS PLANT $1ARTUP . M00t 5 to FULL POWIR 2 3 1 3.1.2(9) 3.1.1(1)(2)(3)(5) P1h2 1 CU REACTOR 1 RIP AND RfC0VERY 10 IUtt PostR 2 3.1.1(4)(5), 3.1.?(19) Pih3 1 CU LOAD CHANCES Al POWlR 3 3.1.1(6) Pfk4 1 CU 3 RCP $1ARfUP/$HUIDOWN/ POWER OrtRAflObs 4 3.1.1( 7) P1k5 1 CU PLANT $HU1DOWN AND COOLDOWN 3 3.1.1($)(8) PIN 6 i CV CORE PERIORMANCE it$11NG 2 3.1.1(9) PTN7 1 CU $URVf]LLANCE It$11NG 1 3.1.1(10) P151 1 CU $1EADY $1 Aft PtRFORMANCE ft$1 A 4.1 P111 1 CU REAC10R TRIP A 3.1.?(19), B2.?(1) P1110 1 CU PORV FAILURE W11H HP1 OPERABLE A 3.1.2(c)(d), B2.?(10) P112 i CU TOTAL Loss OF FtEDWAftR A 3.1.2(10), B2.2(10) P113 1 CU $1MULTAht0V$ Closure OF ALL M51VS A B2.?t3) P1f4 i CU LO$$ OF OF F$ lit POWER PROM FULL POWER A B7. 2( /i) P115 1 CU REACTOR COOLANT PUMP TRi> A 3.1,2(4), S2.2($) Pii6 i CU MAIN TURBikt IRIP A 3.1.2(16), B2.2(6) Pii7 1 CU MAtlMUM Raft POWtk RAMP A 3.1.1(6), B2.2(7) PTT8 1 CU LARCC BREAK Lt4A (DBA) A 3.1.2(1)(b)(c), B2.2(8) P119 i CU MAIN STEAM t!Nt BREAK INSIDI CONTAINMLN1 A 3.1 ?(2), B2.2(9) O
NUCLEAR OPERATIONS TRAINING Da $ 10)$5 0 9l@5-wa
,W.W, DEPAir MENT PROCEDURE Page: 1 of 19 SIMU Jt0R SOURCE DOCUMENT EVALUATION CONTR0!!ED COPY E IGNATURE APPROVAL COVERSHEET APPROVAL:
tiam DjLts (x) NTS: N 1 /1 /#11 [x] NfTH: __ _J /'//41-(x) NLOTH: __
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[x] DNOT m k,- _.2 [ // 1j /9/ q- i1 1 , () VPHP: . () OTHER: _ O _ REVIEW / CONCURRENCE:
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() NLOTS: (( NNLOTS: () _ NTTS: _ .__ () SNST: [x] NTAS: d. 3 /t/f / [x] NTCSt "ft . M /f/ [x] NSTS: h .n 't t 91 () DNP0: () OTHER: hS
TDP 401 Rev. 4 NUCLEAR OPERATIONS TRAINING 0 9...........- DEPARTMENT PROCEDURE Page: 2 of 19 nm SIMULATOR SOURCE DOCUMENT EVALUATION TABLE OF CONTENTS Section No. Paae No. 1.0 PURPOSE . . . . . . . . . . . . . . . . . . . . . . 3 2.0 SCOPE . . . . . . . . . . . . . . . . . . . . . . . 3
3.0 REFERENCES
.................... 3 4.0 DEFINITIONS . . . . . . . . . . . . . . . . . . . . 4 5.0 RESPONSIBILITIES. . . . . . . . . . . . . . . . . . $
i 6.0 ACTIONS . . . . . . . . . . . . . . . . . . . . . . 6 . 6.1 Source Document Classifications . . . . . . . . . . 6 6.2 Initial Evaluation ................ 8 6.3 Final Evaluation of Source Documents ....... 10 O 6.4 Source Document Resolutions 6.5 Source Document Closure 12 12 6.6 Generation of Simulator Change Notices (SCNs) . . . 12 7.0 ATTACHMENTS . . . . . . . . . . . . . . . . . . . . 13 8.0 QUAllTY RECORDS . . . . . . . . . . . . . . . . . . 14 9.0 INTERPRETATION CONTACT. . . . . . . . . . . . . . . 14 10.0 REVISION HISTORY ,................ 14 11.0 REVIEW HISTORY ,................. 14 ATTACHMENT 1 ........................ 15 ATTACHMENT.2 ........................ 16 ATTACHMENT 3 ........................ 18 ATTACHMENT 4 ........................ 19 0
9IM 10P 401 Rev 4 NUCLEAR OPERATIONS TRAINING A I- DEPARTMENT PROCEDURE Page 3 of 19 V lif LL SINULATOR SOURCE DOCUMENT EVALUATION 1.0 WRPOSE This procedure provides instra tions and defines responsibilities for the evaluation of Source Documents. The Source Document review process will identify, evaluate, and tra;k pctential changes to the simulator which originate from reference plant changes and/or industry operating experience. Simulator def t:tencies that exist or develop due to Source Document evaluations are resolved via TOP 403, Simulator Change Standards. This procedure also provides instructions and defines responsibilities for initiating Simulator Change Notices (SCNs). An SCN describes a proposed change to the simulator. The SCN is considered a Source Document andiis ' evaluated by the same criteria as other Source Documents. . 2.0 LGAf% - The scope of work addressed in this procedure begins with the receipt of a Source Document and continues until a Final Evaluation of the Source " Document has been completed and logged in the Configuration Management System (CMS). See Attachment 1, Procedure Flow Chart. All changes to the simulatnr configuration and/or design data base which are required to resolve deficiencies originating from a Source Document evaluation must be made in accordance with TOP 403, Simulator Change Standards. 3.0 HEFERENCE$ 10CFR55.45; ' Operating Tests" Regulatory Guide 1.149; Nuclear Power Plant Simulation Facilities for Use in Operator Licersing Examinations ANSI /ANS-3.5(1985); American National Standard for Nuclear Power Plant Simulators for Use in Operator Training NUREG 1258; Evaluation Procedures for Simulation Facilities Certified Under 10CFR55 INPO TQ-504; Simulator Configuration Management System TT TOP 403; Simulator Change Standards U Simulator Operating Instruction 501-10; Configuration Management System Simulator Operating instruction 501-11; Simulator Trouble Reports
.u m v.,
TDP-41 Rn 4 M NUCLEAR OPERATIONS TRAINING I DEPARTMENT PROCEDURE Page 4 of 19 _) , wa
$1MULATOR SOURCE DOCUMENT EVALUATION 4.0 QIFINITIONS 4.1 REFERENCE PLANT Crystal River Unit 3 4.2 DESIGN DATA BASE - those documents that describe the existing performance and physical appearance standards chosen as the simulator design boundaries; it encompasses a subset of all reference plant documents (and Industry Event Reports).
4.3 SOURCE DOCUMENT - a document which 1) identifies or describes a change to the reference plant design or operation, 2) identifies a possible ' deficiency in the simulator design data base, or 3) an elective enhancement of the simulator capabilities which has no corresponding - change in the reference plant.
^
/ 4.4 SlHULATOR CHANGE NOTICE (SCN) - a type of Source Document which originates V) from within the simulator orgar,1zation. SCNs are reserved for changes whi-h describe an elected enhancement to the simulator. 4.5 TROUBLE REPORT - a document which identifies a discrepancy between the simulator performance / appearance and the reference plant as defined by the existing simulator design data base. Resolution of_A TR cannot result in a chance to the desian data bast. See 501-11 for information regarding irs. 4.6 SIMULATOR CONFIGURATION - those items that define the simulator's scope (software source code, documentation text & drawings, engineering drawings, Handler code / text, malfunctions, and I/F schematics). 4.7 CHANGE CONTA0L NUMBER (CCN) - a unique numeric designator assigned to track impact on the simulator configuration. 4.8 SlHULATOR CHANGE - the software and/or hardware modifications which are necessary to implement an " initiating document" (Source Document or Trouble Report) on the simulator. 4.9 SIMULATOR CHANGE RECORD (SCR) - a document which tracks a required change in the existing simulator configuration and/or design data base.
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TOP 401 Rev 4 iM NUCLEAR OPERATIONS TRAINING {
.Dd DEPARTMENT PROCEDURE Page 5 of 19 wa SIMULATOR SOURCE DOCUMENT EVALUATION IJW85E-4.10 WORK PACKAGE the paperwork hardcopy which describes and documents the simulator change in conformance with TOP 403, Simulator Change Standards.
A Work Package is deemed to be complete when all relevant Work Package activities have a status of complete (CP) in CMS. Work Package activities may be software, hardware, or both. 4.11 CONFIGURATION MANAGEMENT SYSTEM (CMS) - the computer resident, relational data base program, used to support the simulator configuration control process. 4.12 CMS ADMINISTRATOR the individual primarily responsible for logging CMS TRs/ Source Documents, updating data base entities, assigning CCNs, an( generating CMS reports. , 5.0 RE51QNSIBILIT111 - 7"% 5.1 NUCLEAR LICENSED OPERATOR TRAINING MANAGER (NLOTM): () . The NLOTH is responsible for: The content and implementation of this procedure. Approval of all Source Document Final Evaluations. 5.2 NUCLEAR SlHULATOR TRAINING SUPERVISOR (NSTS) The NSTS is responsible for: Performing Initial and Final Evaluations of all Source Documents. 5.3 SlHULATOR HAINTENANCE AND ENGINEERING SUPERVISOR (SMES) i The SMES is responsible for: Assisting the NSTS in evaluating Source Documents. As required, distributing Source Documents to hardware and software j personnel for in-depth discinline evaluations, Reviewing hardware and software evaluations. Il Performing cost justifications when required. j G Reviewing the completed Source Document Evaluation Checklist.
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73 V ,m 9lM .N.5-NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE TDP 4 1 Rev 4 Page 6 of 19
$1MULATOR SOURCE DOCUMENT EVALUATION 1
5.4 SlHULATOR SYSTEMS ENGINEER (SSE): The SSE is responsible for conducting thorough, accurate discipline evaluations of Source Documents as directed by the SMES. 5.5 SIMULATOR COMPUTER CONTROLS SPECIALIST (SCCS): The SCCS is responsible for conducting thorough, accurate discipline evaluations of Source Documents as directed by the SMES. 5.6 NUCLEAR OPERATIONS INSTRUCTORS (N0ls): TheN0lsareresponsibleforconductingthorough,accuratediscipline! evaluations of Source Documents as directed by the NSTS. 6.0 ACTIONS 6.1 SOURCE DOCUMENT CLASSIFICATIONS , 6.1.1 tiodification Acoroval Records (MARS) 6.1.1.1 In general, MARS and associated FCNs should be considered a single Source Document. That is, the Source Document is the MAR itself; FCNs are considered revisions to the MAR. 6.1.1.2 For purposes of the Source Document evaluation process, MARS are classified as being either "in service" or "in progress". "In-service" MARS are reference plant modifications which have been accepted by Operations personnel. "In progress" MARS are reference plant modifications which have agi been accepted by Operations; the modification could-be anywhere from " drawing b>ard" to "in-work". 6.1.1.3 MAR /FCN drawings (interims) which have been generated in the reference plant prior to simulator " data freeze" are incorporated in the design data bese by the simulator vendor (CAE). 6.1.1.4 MARS and associated FCNs which are generated in the reference plant after simulator data freeze and *in-service" prior to simulator acceptance should be evaluated in accordance with this procedure. Final Evaluation of these documents and the implementation of required changes to the simulator shall be completed within 18 months of simulator acceptance.
TDb401 Rev 4 NUCLEAR OPERATIONS TRAINING A O 9 . .......... fTd DEPARTMENT PROCEDURE Page 7 of 19
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1 SlHULATOR SOURCE DOCUMENT EVALUATION 6.1.1.5 MARS and associated FCNs tsich are generated in the reference plant after simulator data freeze but nol "in-service" priur to simulator acceptance should be evaluated in accordance with this procedure. Final Evaluation of these documents shall be completed within 12 months of the "in service" date at the reference plant, implementation of required changes to the simulator shall be completed within 12 months of the "in-service" date or 12 months of the Final Evaluation date (whichever is later). 6.1.2 Drtwina Chance Notices (OCNs) 6.1.2.1 DCNs are the method by which the reference plant revises drawings to incorporate MAR Interim Drawings and correct drawing errors. That is,' the DCN documents the release of a new Master Drawing. Since all MARS are already evaluated for simulator impact, the DCNs (except those correcting drawing errors) are simply redundant with the Interim Drawing which has been explicitly reviewed during the MAR evaluation. Further, since the simulator design data base is a subset of all Master Drawings, only a ('sV) fraction of DCNs are applicable to the simulator's scope. Therefore, the . scope of DCNs which are revfewed as Source Documents should be limited to corrections of Master Drawings found in the simulator design data base. 6.1.3 Plant / Industry Ooerational Even.ta 6.1.3.1 Plant / industry Operational Events are reviewed in accordance with TOP-402. Those events which are evaluated as requiring simulator changes will be tracked as Source Documents. 6.1.4 Plant Procedgrff 6.1 4.1 The impact of plant procedure changes on the simulator are functionally evaluated through their use during training sessions and the conduct of
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annual and quadrennial performance testing. Plant procedures which require changes to the simulator will be tracked as Source Documents. 6.1.5 Simulator Chance Notices (SCNs) 6.1.5.1 The simulator organization may generate " internal" Source Documents known as Simulator Change Notices (SCNs). These are necessary to maintain traceability of elective enhancements. See step 6.6 of this procedure for information pertaining to generating a SCN.
IU"'# I "'" 4 9M 1iitt
.I NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE SIMULATOR SOURCE DOCUMENT EVALUATION Page 8 of 19 6.2 INITIAL EVALVATION 6.2.1 Upon receipt of a Source Document, the CMS Adminis Note, Source Documents which have been previously evaluated will already See 501 10; Configuration Management System, for have a CHS record.
guidance pertaining to data entry. 6.2.2 The NSTS shall evaluate the Source Document for simulator impact by The completing the Checklist Part A, Scope Assessment (see Attachment 2). NSTS should indicate an affirmative response to the " question" by marking ', . the box. 6.2.2.1 For Source Documents that have been previously evaluated, the Specifically, principle the f(STS areas of concern are the " field changes" (FCNs). e should verify the FCNs have not altered the scope of change, ie., the (") original scope " question" responses are still applicable. . 6.2.2.2 Depending on the type and scope of change addressed by the document, the NSTS may be unable to independently access the In impact, ie., there such cases, may be the NSTS Checklist " questions" the NSTS cannot resolve. should consult with the SMES on the specifics of the change. 6.2.2.3 Should the NSTS and SNES be unable to reach a consensus regarding the Source Document impact, the NSTS/SHES should designate one or moreOne, or simulator personnel to perform a detailed discipline evaluation. both hardware (SCCS) and software (SSr) oersonnel may be required to accurately access the change. The NSTS/ShE5 should indicate the request for discipline evaluations by marking the appropriate box. 6.2.2.4 Discipline evaluations should be conducted using the Discipline Evaluation The intent of the discipline evaluation is to Checklist (Attachment 3).
- 1) ascertain whether the Source Document will have impact, and 2) provide an indication regarding the magnitude of the required simulator change.
6.2.2.5 Ccmpleted discipline 9 valuations should be returned to the NSTS and appended to the Source Document Evaluation Checklist. 6.2.2.6 Based on information provided by the discipline evt.luation, the NSTS should complete any "open" Part A Checklist " questions". [_ v eia M
TDP401 Ru 4 (9 a SIM nta D[- NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE Page 9 of 19 SIMULATOR SOURCE DOCUMENT EVALUATION . 6.2.3 The NSTS shall determine if a Final Evaluation is permissible at this time by completing Part B of the Checklist. Depending on the Source Document classification and/or the nature of the change, it may not be necessary to delay final Evaluation. 6.2.3.1 For Source Documents classified as industry Event Reports, SCNs, DCNs, and Plant Procedures, the final scope of change is clearly identified. Such documents should receive an imediate Final Evaluation; discontinuc section 6.2 and continue with section 6.3. 6.2.3.2 Source Documents classified as MARS (either "in service" or "in progress") which involve gnh structural modifications, repairs, or exact one for one equipment changeouts in the reference plant cannot impact the simulator scope. Further, FCNs on such MARS should not revise the evaluation. MARS of this " type" should receive an imediate Final Evaluation; discontinue - G section 6.2 and continue with section 6.3. V 6.2.3.3 Source Documents which describe plant modifications to a system or . component which is clearly outside the scope of simulation cannot impact the simulator; neither should any related FCNs. Obviously, the evaluator is required to research the scope of simulation thoroughly to make an informed assessment. MARS of this " type" should receive an imediate Final Evaluation; discontinue section 6.2 and continue with section 6.3. 6.2.3.4 Source Documents classified as MARS which impact the simulator (as denoted by Part A responses) should receive an immediate Final Evaluation; Note, related FCNs discontinue section 6.2 and continue with section 6.3. which are generated up until the MAR is "in service", are considered integral to the Source Document itself. Any simulator work which is done in " parallel" with the reference plant is subject to revision as a result of- the FCNs. 6.2.3.5 Source Documents classified as "in service" MARS have a scope which is considered finalized. "In service" MARS should receive an imediate final Evaluation; discontinue section 6.2 and continue with section 6.3. Note, this is ultimately how *in progress" MARS which have been previously evaluated as having no simulator impact, get a Final Evaluation. The intent is to re examine the MAR after it is placed "in service" (this includes all the FCNs generated since the original evaluation). It is conceivable that a field change will alter the original impact evaluation.
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...........- DEPARTMENT PROCEDURE Page 10 of 19 SIMULATOR SOURCE DOCUMENT EVALUATION 6.2.4 Source Documents which cannot be classified as described in 6.2.3.1 thru 6.2.3.5 are, by default. "in progress" HARs that have been tentatively evaluated agl to impact the simulator. Because these MARS may ultimately generate FCNs which revise the evaluation, it is necessary to retain the MAR until the modification is "in service" at the reference plant. At that time the MAR will be re evaluated. The NSTS shall indicate a request to delay Final Evaluation by marking the appropriate Part B box and with a signature.
6.2.4.1 Route the Source Document and Checklist to the SMES. 6.2.5 The SMES shall review the entire Checklist for completeness and accuracy. In particular, the SMES should review any attached discipline evaluations (HW, SW) to ensure all facets of the change have been properly evaluated. l 6.2.5.1 The SMES shall indicate the review is complete by signature. - 6.2.5.2 Route the Source Document and Checklist to the CMS administrator. . 6.2.6 The CMS Administrator should update the Source Document record in CMS to reflect the evaluation results. See 501-10; Configuration Management System, for guidance pertaining to data entry. 6.2.6.1 The CMS Administrator should filt the Sourt* Document (and Checklist) until the MAR is "in service" at the reference plant. 6.2.7 As MAR related FCNs are received, they will be incorporated in the Source Document file. 6.2.8 The Source Document (MAR and associated FCNs) should only receive a Final Evaluation once a MAR status indicates the MAR is "in-service". This ensure FCNs have not altered the scope of the MAR from a simulator perspective. 6.3 FINAL EVALUATION OF SOURCE DOCUMENTS 6.3.1 The NSTS shall complete Checklist Part C. Final Evaluation. When appropriate, provide comments that clarify the evaluation process. 6.3.1.1 Route the Source Document and Checklist to the SMES. 6.3.2 The SMES shall review the entire Checklist for completeness and accuracy. /m\ In particular, the SMES should review any attached discipline evaluations O (HW, SW) to ensure all facets of the change have been properly evaluated. The SMES shall indicate the review is complete by signature. i ., m eu
TD 4 1 Rev 4 NUCLEAR OPERATIONS TRAINING o C 9..........- fl14 L DEPARTMENT PROCEDURE Page 11 of 19 31MULATOR SOURCE DOCUMENT EVALUATION 6.3.2.1 For each Source Document which involves a change to the Main Control Board Panels, Remote Shutdown Panel, or other reference plant hardware, the SMES should ensure that the Source Document contains provisions for procuring duplicate hardware for the simulator. Omission of duplicate hardware should be addressed by the SMES. 6.3.3 Should the NSTS/SMES determine that the changes involve extensive hardware / software modifications which have limited training value, the SMES should provide a detailed " Cost Analysis / Training impact Assessment". The intent of the " Cost Analysis / Training impact Assessment" is to 1) quantify the required changes (time / materials), and 2) qualify the merits of incorporating the change (from a training perspective). A relatively large differential between costs and value would provide the NLOTM with a basis to justify " relief" from the change. Attach the " Cost Analyses / Training Impact Assessment" to the Source Document Evaluation Checklist. - 6.3.3.1 The completed Checklist and Source Document should be submitted to the
- NLOTH for approval.
6.3.4 The NLOTH shall approve the Source Document evaluation process by signature. 6.3.4.1 Route the completed Checklist and Source Document to the CMS Administrator. 6.3.5 The CMS Administrator should update the Source Document record in CMS to reflect the evaluation results. See 501-10; Configuration Management System, for guidance pertaining to data entry. 6.3.5.1 For Source Documents determined to have a Final Evaluation of " CHANGE REQUIRED", the CMS Administrator should also assign a unique CCN, ie., create a new Change Control Record in CMS. The CMS Administrator should indicate the CCN in the appropriate field of the Checklist. 6.3.6 A ' copy of the completed Checklist (and any attachments) should be maintained in the CMS files; the original (Final Evaluation page only) shall be sent to Nuclear Records Management as a Quality Record. 6.3.6.1 For Source Documents determined to have a Final Evaluation of " CHANGE REQUIRE 0", the CMS Administrator should route the Source Document, Checklist, and any attachments to the SMES for initiation of a Simulator Change Record in accordance with TDP 403, Simulator Change Standards. (],,
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.E NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE Page 12 of 19 SIMULATOR SOURCE DOCUMENT EVALUATION 6.3.6.2 for Source Documents determined to have a Final Evaluation of "N0 CHANGE REQUIRE 0", the CMS Administrator can discard the Source Document and retain the Checklist in the CMS files.
6.4 SOURCE DOCUMENT RESOLUTIONS 6.4.1 All activities necessary to implement the Source Document shall be performed in full compliance with TOP 403, Simulator Change Standards. This will ensure Source Document resolutions are consistent with established standards regarding methods, testing, and documentation. i 6.5 SOURCE DOCUMENT CLOSURE 6.5.1 When all Work Package activities relevant to a particular Source Docu' ment have been COMPLETED (logged CP in CMS) adQ the change has been " Accepted", it should be routed to the SMET 'or closure approval. As described -in all configuration item changes should p} k TOP 403, Simulator Change Stanu. be logged CURRENT (CU) prior to Source Document closure. , 6.5.2 The SMES shall verify that the Work Package meets all guidelines regarding configuration control as specified in TOP-403. That is, the Work Package must contain a completed Work Package Checklist and a copy of each affected configuration item (clearly indicating the scope of change). Additionally the SMES should verify that each configuration item has been properly logged in CMS. 6.5.3 The SMES shall ;pprove the Source Document closure by signing the Simulater Change Record (SCR) and route the Work Package to the CMS Administrator. 6.5.4 The CMS Administrator should log the Source Document record CLOSED (CL) in CMS and file the Work Package sequentially in the CCN file. 6.6 GENERATION OF SIMULATOR CHANGE NOTICES (SCNs) 6.6.1 A Simulator Change Notice (SCN) is an " internally" generated Source Document which describes a proposed change to the simulator configuration. The SCN is treated as a Source Document and is evaluated in the same manner as other sources of change (MARS, DCNs, plant events, and plant procedures). That is, the SCN merely proposes a change; the decision to implement the change will result from the evaluation process as previously described in steps 6.2 and 6.3 of this procedure. (V3
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TDP 401 Rev 4 NUCLEAR OPERATIONS TRAINING ma 9 . .......... - DEPARTMENT PROCEDURE Page 13 of 19 S!MULATOR SOURCE DOCUMENT EVALUATION 6.6.2 An SCN should not be used in lieu of a Trouble Report (TR) where a TR will suffice. A TR describes a discrepancy in the simulator's performance / appearance as compared to the existing design data base. An SCN describes a proposed enhancement of the simulator. 6.6.3 Any individual associated with the simulator organization may initiate an SCN. This includes the NLOTM, NSTS, SNES, N0ls, SCCSs SSEs, and the CMS Administrator. See Attachment 4; Simulator Change Notice. 6.6.3.1 The originator should verbally inform the SMES that the SCN is forthcoming, e 6.6.3.2 The originator should fill out the header section of the SCN. Note, the - SCN number should be left blanks it will be sequentially assigned by the CMS Administrator. - 6.6.3.3 The originator should describe the proposed change in sufficient detail to facilitate proper evaluation. (]m 6.6.3.4 The originator should discuss the merits or benefits of implementing the proposed change. 6.6.3.5 The originator should discuss the consequences of Dn1 implementing the proposed change from a training or regulatory perspective. 6.6.3.6 The completed SCN should be routed to the CMS Administrator who will 1) assign the SCN number, and 2) log the SCN in CMS. 7.0 UTACHMENTS 7.1 ATTACHMENT 1 - Procedure Flow Chart 7.2 ATTACHMENT 2 Source Document Evaluation Checklist 7.3 ATTACliMENT 3 Discipline Evaluation Checklist 7.6 ATTACHMENT 4 - Simulator Change Notice O v., sa m
lM NUCLEAR OPERATIONS TRAINING TDP 401 Rev 4 1
. .5 DEPARTMENT PROCEDURE Page 14 of 19 O) ,,,a SIMULATOR SOURCE DOCUMENT EVALUATION 8.0 OUALITY RECORD) 8.1 ATTACHMENT 2 Source Document Evaluation Checklist 9.0 INTERPRETATION CONTACT 9.1 SIMULATOR MAINTENANCE AND ENGINEERING SUPERVISOR 10.0 REVISION HISTORY Revision 0 Date: 8/30/88 Original Release.
Revision 1 Date: 1/25/89 Clarification of review pronss and requirements for discipline evaluation; addition of Appendix A (CMS entries and (m Q updates). Revision 2 Date: 1/10/90 Clarification of review process; deletion of Appendix A. Revision 3 Date: 3/30/90 Total rewrite; clarification of review process; modify Evaluation Checklist; add Simulator Change Notice. Revision 4 Date: 10/12/90 Add permissives for immediate Final Evaluation to include plant repairs, exact one for one component changeouts, and changes to a system / components which is clearly outside the scope of simulation. Clarify DCN review requirements. Combine Discipline Checklists on a single form. 11.0 REYlEV HISTORY Ddt EEU 2n Conductina Review Comments v i.., ,,m
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TOP 401 Rev 4 9 SOURCE DOCUMENT 10: ATTACi; MENT 2 iQURCE DOCUMENT EVALUATION CHECKLIST DATE RECEIVED: Routing: Orig: CMS File DOCUMENT TITLE: A. SCOPE ASSESSMENT () The change will affect plant behavior / response in a manner that can be directly observed or inferred in the control room. [] The change will affect the manner in which information is presented to control room personnel (meters, recorders, alarms, status lights, CRT displays,computerpoints/ printouts,etc.). () The change will affect the appearance of the control room panels, hardware or the control room itself. () The change will affee. setpoints or calibration of a control system, i () The change will affect equipment power supplies. l [] The change will affect training evolutions or training materials. () The change will affect 1/F schematics. COMMENTS: _ () DISCIPLINE EVALVATIONS REQUIRED: [ ] HW [ ] SW B. INITIAL EVALVATION () The Source Document is an Industry Report, SCN, DCN, or Procedure. [] The Source Document describes a reference plant a) structural change, b) repair, c) exact one for one component changeout, or d) change to a systeg/ component which clearly is outside the " scope of simulation". () The Source Document describes a MAR which impacts the simulator. () TheSourceDbcumentdescribesan'in-service"MARwhichdoesgolimpact the simulator. IF 101 of the above four (4) responses are affirmative, complete Part C, Final Evaluation. [] OTHERWISE, the Source Document is being retained and will be re evaluated i when the MAR is 'in-service". Request of: DATE: , NSTS mo/ day /yr l Reviewed by: DATE: SMES mo/ day /yr
10P 401 Rev 4 Page 17 of 19 ATTACHMENT 2 (Pg. 2) Routing: ( 9"'""" $0URCE DOCUMENT EVALUATION- CHECKLlil Orig: Records Copy: CMS File I SOURCE DOCUMENT 10: DATE RECEIVED: I mo/ day /yr I DOCUMENT TITLE:. . 1 C. FINAL EVALUATION [] CHANGE REQUIRED: ()HW [ ] SW CCN: _ [] NO CHANGE REQUIRED: COMMENTS: e Od EVALUATED BY: NSTS DATE: mo/ day /yr REVIEWED BY: DATE: SMES mo/ day /yr APPROVED BY: DATE: NLOTH mo/ day /yr 4 O
TOP-401 Rev 4 Page 18 of 19 ATTACHMENT 3 Routing: DISCIPLINE EVALUATION CHECKLIST Orig: CMS File i DATE RECEIVED: mo/ day /yr SOURCE DOCUMENT 10: _ DOCUMENT TITLE:. _ _ _ _ _ _ A. SCOPE ASSESSMENT The change will affect plant behavior / response in a manner tha [] directly observed or inferred in the control room. l The change will affect the manner in which information is presen l [] control room personnel (meters, recorders, alarms, status lightsl displays, computer points / printouts, etc.). The change will affect the appearance of the control room pane [] or the control room itself. j The change will affect setpoints or calibration of a control ' system, l [] . The change will affect equipment power supplies. [] [j The change will affect training evolutions or training materials. ~ l The change will affect I/F schematics. ~ [] _ COMMENTS: _ -
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B. CONFIGURATION ITEMS AFFECTED [] Documentation Text [] SIMEX Files (SW Source Code) [] Engineering Drawings [] Documentation Drawings
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[ ] < month [ ] > month [ ] < day [ ] < week [ ] < 2 weeks DATE: __ EVALUATION COMPLETED BY: _
TDP 401 Rev 4 r] ATTACHMENT 4 Page 19 of 19 v l SIMULATOR CHANGE NOTICE O SCN i: ORIGINATOR: DATE: mo/ day /yr TITLE: Describe the proposed change: - 1 Discuss the merits of the change: O . Discuss the-consequences of no action: w - CMS LOGGED BY:-- DATE: CCN: mo/ day /yr
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REVIEW OF PLANT / INDUSTRY OPERATIONAL EVENTS i i 1 1 1 TABLE OF CONTENTS SECTIQH SUBJECT EAgg 1.0 Purpose ................................................... 3 2.0 Scope ..................................................... 3 3.0 References ................................................ 3 4.0 Defiriitions ............................................... 4 5.0 Responsibilities .......................................... 4 f s o.0 Action .................................................... 6 7.0 Attachments ............................................... 8 8.0 Quality Records ........................................... 8 9.0 Interpretation Contact .................................... 8 10.0 Revision History .......................................... 8 11.0 Review History ............................................ 8 Attachment 1 - Operational Event Evaluation Checklist. . . . . . 9 l L
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M NUCLEAR OPERATIONS TRAINING TOP 402 Rev 0 DEPARTMENT PROCEDURE Page 3 of 11 (O) wa REVIEW OF PLANT / INDUSTRY OPERATIOkAL EVENTS 1.0 PURPOSE This procedure provides instructions and defines responsibilities for the review and evaluation of plant / industry operational events for impact on the simulator in terms of scope of simulation, simulator modeling and simulator performance testing. 1.0 KQEE 2.1 The scope of work addressed in this procedure begins with the receipt of a plant / industry operational event and continues through the conduct and evaluation of test runs of the trarsient on the simulator (if required), resolution of corrective actions resulting from the test runs, and determination if the event should be incorporated in the simulator perfonnance test data base. 2.2 This procedure does not address the evaluation of operational events for incorporation in simulator training. Training evaluations of operational O; events should be made in accordance with TDP-105. D'
3.0 REFERENCES
3.1 10CFR55.45; " Operating Tests" 3.2 Regulatory Guide 1.149; Nuclear Power Plant Simulation Facilities for Use in Operator Licensing Examinations 3.3 ANSI /ANS-3.5(1985); American National Standard for Nuclear Power Plant Simulators for Use in Operator Training 3.4 NUREG 1258; Evaluation Procedures for Simulation Facilities Certified Under 10CFR 55 3.5 INPO TQ-505; Simulator Configuration Management System 3.6 TDP-401; Simulator Source Document Evaluation 3.7 TDP-403; Simulator Change Requests 3.8 TDP-405; Simulator Performance Testing i i 3.9 Simulator Operating Instructions 501-11; Simulator Trouble Reports l n 3.10 Simulator Operating Instructions S01-10; Configuration Management System V I l l
TO N 02 Rev 0 9M wa NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE REVIEW 0F PLANT / INDUSTRY OPERATluNAL EVENTS Page 4 of 11 4.0 DEF NITIOLG 4.1 PLANT / INDUSTRY OPERATIONAL EVENTS: There are four primary sources of information on plant / industry operational events: 1- LERs (Licensee Event Reports), UOERs (Unusual Operating Event Reports) and RECALL data for events that occur at Reliability Group and transmitted to Training. 2- TAP (Transient Assessment Program) Reports for events that occur at other B & W plants. These reports are generated by B & W and reviewed for CR-3 applicability by the Nuclear Safety 3- NRC Bulletins and Information Notices - these docume , [ applicable events are transmitted to Training. 4- SOERs (Significant Operating Event Reports) and by the Nuclear Safety and Reliability Group; applicable events are transmitted to Training. 5.0 RESP 0NSIBILITIES 5.1 NUCLEAR LICENSED OPERATOR TRAINING MANAGER (NLOTH): The NLOTM is responsible for the content and implementation of this 5.1.1 procedure, 5.2 SlHULATOR MAINTENANCE AND ENGINEERING SUPERVISOR ( The SMES is responsible for: 5.2.1 Assigning responsible SSEs to support test runs of operational events on the simulator. Reviewing, with the NSTS, results of test runs and determining if the 5.2.2 operational event should be incorporated in the simulator performance O test data base. O
%A . _ _ - - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ .
l TDP 402 Rev 0 NUCLEAR OPERATIONS TRAINING l
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Q.,)) "'" S DEPARTMENT PROCEDURE Page 5 of 11 REVIEW 0F PLANT / INDUSTRY OPERATIONAL EVENTS 5.3 SIMULATOR SYSTEMS ENGINEER (SSE): 5.3.1 The SSE is responsible for supporting test runs of operational events on the simulator. This support includes data collection and software evaluations of the test run results. 5.4 NVCLEAR SIMULATOR TRAINING SUPERVISOR (NSTS) The NSTS is responsible for: 5.4.1 Determining if operational events should be evaluated on the simulator. 5.4.2 Assigning responsible N0ls to test run events on the simulator, 5.4.3 Reviewing, with the SMES, results of test runs and determining if the , oper.tional event should be incorporated in the Certification Test Procedure data base, t p, 5.4.4 Updating the Certification Test Procedure data base as applicable. \ l Q 5.6 NUCLEAR OPERATIONS INSTRUCTOR (NOI): l 5.6.1 The N01 is responsible for conducting test runs of operational events on i the simulator. This includes test procedure scenario development, data collection, and operational / training evaluations of the test run results. l t l I l l O O l w mm
M NUCLEAR OPERATIONS TRAINING TDP 402 Rev 0 g1 . DEPARTMENT PROCEDURE Page 6 of 11 V nu REVIEW OF PLANT / INDUSTRY OPERATIONAL EVENTS 6.0 ACTION 6,1 Industry operating event reports will be received by the Nuclear Training Clerk and routed to the NSTS. 6.2 Upon receipt of operational event report, the NSTS (or his designee) should review the data using the Operational Event Evaluation checklist (Attachment 1) completing Part A. 6.3 If the initial review indicates that the event does not need to be run on the simulator, the NSTS and SMES should complete Part C of the Operational Event Evaluation Checklist, and submit it to the NLOTM for final review and approval. 6.4 The completed checklist (Parts A ar.d C) shoulo be retained in the Training Department files for a period of four (4) calendar years from the NLOTH approval date. n 6.5 (d.
' If the initial review indicates that the event should be run on the simulator, the NSTS and SMES should designate an NOI and SSE (as
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required) to conduct the test run and evaluate the test run results. 6.6 The designated N01 should develop the performance test procedure. The test procedure should incorporate a detailed sequence of events and expected simulator response. t.7 The designated N01 and SSE should determine the plant parameters to be recorded during the test run. This determination should be based on (1) the data collected during the &ctual transient, and (2) an examination of the data collected for similar plant performance tests. When similar transients do not exist, the data should include those parameters required to demonstrate the ability of the simulator to reproduce the plant transient. l 6.7.1 When practical, data collection / comparison should be accomplished using the Graphic Recorder. For CR-3 specific events, consideration should be given to the use of the VSA system for critical parameter verification. I 6.8 After completion of the test run, the results of the test should be evaluated against actual plant data by both the SSE and the NOI. Discrepancies between plant and simulator response should be resolved in one of three ways: 1 (", U i .., m.
M NUCLEAR OPERATIONS TRAINING IDP-402 Rev 0 [ DEPARTMENT PROCEDURE Page 7 of 11 f G wa i REVIEW 0F PLANT / INDUSTRY OPERATIONAL EVENTS 1- Differences in re;ponse which are due to differences in plant configuration / equipment, insufficient information or an inability to exactly replicate operator intervention should be documented on part B of the checkli:t. 2- Differences in response due to modeling considerations outside the existing scope of simulation should be identified on Attachment 1, Operational Event Evaluation Checklist, Part B. The Industry Event Report, the Test Procedure, and the Evaluation Checklist (Attachment 1) should be routed to the CMS Administrator. The CMS Administrator should log the " Industry Event" into CMS as a
" Simulator Source Document", in accordance with TDP-401, Simulator Source Document Evaluation.
3- Differences in response due to modeling considerations within the existing scope of simulation should be identified and a Trouble Report (TR) initiated in accordance with Simulator Operating Instructions 501-11. dO 6.9 Test runs requiring corrective action should be placed on hold until all required actions are completed in accordance with S01-11 or TDP-401, as applicable. After the associated CR or TR is " offered", the test should be re-run to verify correct simulator response. 6.10 If the test run is satisfactory, the completed test procedure and performance data evaluations should be submitted to the NSTS and SMES. 6.11 The NSTS and SMES should ensure the test package is complete and review the test results for possible incorporation into the Certification Test Procedure database. 6.12 The NSTS and SMES should indicate disposition of the operational event by completing Part C of the checklist. If the event will be incorporated into the Certification Test Procedure database this should be accomplished in accordance with TDP-405. 6.13 The completed checklist should be submitted to the NLOTH for review and approval. 6.14 The completed checklist, including test run results, should be retained in the Training Department files for a period of four (4) calendar years from the NLOTM approval date. O v s .r mm ___ _ _- -.__.__- -_ - _____-_-___--____-_._-_--._-_-_-_--_._-__-__--__ _ _ w
NUCt. EAR OPERATIONS TRAINING TDP 402 Rev 0 O
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DEPARTMENT PROCEDURE REVIEW 0F PLANT / INDUSTRY OPERATIONAL EVENTS Page 3 of 11 7.0 ATTACHMENTS 7.1 Attachment 1 Operational Event Evaluation Checklist 8.0 00AllTY RECORDS 8.1 None 9.0 INTERPRETATION CONTACT 9.1 Nuclear Simulator Training Supervisor 10.0 REVISION HISTORY Rev. 0 Original Release 11,0 REVIEW HISTORY l
TOP 402 Rev 0 Page 9 of 11
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REVIEW 0F PLANT / INDUSTRY OPERATIONAL EVENTS ATTACHMENT 1 (Page 1 of 3) OPERATIONAL EVENT EVALUATION CHECKLIST DOCUMENT NUMBER: DATE RECEIVED: ( ) RECALL DATA ( ) LER ( ) SER. [ ] SOER ( ) 00ER ( ) OTHER DOCUMENT TITLE: A. OPERATIONAL EVENT ASSESSMENT 1, [ ] Yes [ ] No The event can be initiated, monitored, controlled and mitigated through systems installed on Crystal River Unit 3. M . The event includes sufn.-tent documentation of operator actions, sequence of events, and plant / component responsa. AND The event is not similar to an event already included within the tested scope of simulation.
- 2. [ ]_Yes (- ] No FPC has comitted to incorporate the event in future operator training.
AND The event is not similar to an event already included within the tested scope of simulation. If I gr 2 above are checked YES, a performance test should be run to verify simulator response to the reference event. Performance Test Required (_] Yes [ ] No O l
TDP-402 Rev 0 Page 10 of 11 REVIEW OF PLANT / INDUSTRY OPERATIONAL EVENTS ATTACHMENT 1 (continued) (Page 2 of 3) OPERATIONAL EVENT EVALUATION CHECKLIST B. TEST RUN RESULTS/ EVALUATION: O - [] Test run satisfactory. No corrective action required. Attach test procedure and performance data and route to NSTS for closure. OR [] Trouble Report issued. TR # [] Operational Event logged into CMS as " Source Document". Scurce Document No. AND [] Corrective action completed. Test run satisfactory. Attach test procedure and performance data and route to NSTS Tor closure. EVALUATED BY: Nuclear Operation Instructor /Date Simulator Systems Engineer /Date
TDP-402 Rev 0
, Page 11 of 11 REVIEW 0F PLANT / INDUSTRY OPERATIONAL EVENTS ATTACHNENT 1 (continued)- (Page 3 of 3)
OPERATIONAL EVENT EVALUATION CHECKLIST C. FINAL EVALUATION ( ) Evaluation Complete. The Operational event does not require a simulator test run. [ ] Evaluation and test run complete. The operational event will not be incorporated into the simulator Certification Test Procedure database.
.( ') Evaluation and test run complete. The operational event W.il]. be incorporated into the simulator Certification Test Procedure database.
COMMENTS: EVALUATED BY: Nuclear. Simulator Training Supervisor /Date Simulator-Maintenance-and Engineering Supervisor /Date APPROVED BY: Nuclear Licensed Operator Training Manager /Date O
M NUCLEAR OPERATIONS TRAINING 03 [0 a bMd DEPARTMENT PROCEDURE Page: 2/$$/9 1 of 26 SIMULATOR CHANGE STANDARDS CONinGLLED n IDP SIGNATl)RE APPR OVERSHEET APPROYAL: liiLmjt QAlg [x] NTS: ff (d 3 /s4/11 [x] NTTH: o %/.:te/91 [x] NLOTH: [4 S/r > [x] DNOT' 1 [/'71 v\~ ) J \' ( ' [] VPNP: I [] OTHER: O - REVIEW / CONCURRENCE: Name QAlt [] NLOTS: [] NNLOTS: [] NTTS: [] SNST: - [x] NTAS: choJ/ b ?[4h/ [x] NTCS: c kmutmd ~$Aa'h'/ [x] NSTS: f d _ b r = --- e 3 h M q s [] DNP0: [] OTHER: 9
ND NUCLEAR OPERATIONS TRAINING TOP-403, Rev 0 1 DEPARTMENT PROCEDURE wa SIMULATOR CHANGE STANDARDS l!iEEl Sutdec t Ejtgg 1.0 PURPOSE 3 2.0 SCOPE 3
3.0 REFERENCES
3 l 4.0 DEFINITIONS 4 5.0 RESPONSIBILITIES 6 6.0 ACTIONS 8 6.1 GENERATION OF A SIMULATOR CHANGE RECORD (SCR) 8 h 6.2 PROGRESSION OF WORK 9 . 6.3 HOUSEKEEPING CHANGES 12 l 6.4 PLACING A SIMULATOR CHANGE ON OFFER 13 6.5 ACCEPTANCE OF A SIMULATOR CHANGE 17 6.6 COMPLETION OF THE WORK PACKAGE 19 6.7 DOCUMENT CONiROL & QUALITY RECORDS 21 6.8 DISK PACK BACK-UP AND STORAGE 23 7.0 ATTACHMENTS 23 8.0 QUALITY RECORDS 23 9.0 INTERPRETATION CONTACT 23 10.0 REVISION HISTORY 23 11.0 REVIEW HISTORY E4 ATTACHMENT 1 - SIMULATOR CHANGE RECORD T.5 ATTACHMENT 2 - WORK PACKAGE ACTIVITY CHECKLIST 26 l _'i_ _ _ --- ---- l*C
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9M 1 NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE SINULATOR CHANGE STANDARDS P sg;403,Reh0 9 eer 1.0 PURPOSE j This procedure provides instructions and defines responsibilities for implementing " Changes" to simulator software, hardware, and/or documentation. The intent is to utilize a modular modification approach (known as Work Packages) to provide essential traceability, implementation controls, and appropriate testing of all permanent simulator Changes. Simulator Changes resolve scope " deficiencies" as identified by the Source Document Evaluation process (TDP-401) and
" discrepancies" raised by the Simulator Trouble Report process (S01-1.1).
2.0 SCOPE The scope of work addressed by this procedure begins when :
- 1) a modification to the reference plant has been determined to impact q the simul 6 tor as evaluated by TDP-401.
'v/ OR,
- 2) a 1 rouble Report has been written as discussed in S01-11 and verified to be a valid discrepancy.
For either condition, a Simulator Change Record (SCR) is issued to track and control all simulator Changes required to resolve the deficiency / discrepancy. The resulting data and paperwork (including copies of all Configuration Item changes) are documented in a Work Package and thereafter maintained as a Quality Record. This procedure does not discuss simulator training materials, which are addressed in TOP-105.
3.0 REFERENCES
3.1 ANSI /ANS-3.5(1985); American National Standard for Nuclear Power Plant Simulators for Use in Operator Training 3.2 INPO TQ-504; Simulator Configuration Management System e,.m
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<Dd DEPARTMENT PROCEDURE kg;403, o s G wa SIMULATOR CHANGE STANDARDS 3.3 CAE Procedure CD98225.01.8.300; Deficiency Reporting 3.4 CAE Procedure C0109303.01.8.300; Software Programming Engineering Changa Request 3.5 TDP-405; Simulator Performance Testing 3.6 TDP-401; Simulator Source Document Evaluation.
3.7 TOF-109; Training Records Management. 3.8 Simulator Operating Instruction 501-07; Magnetic Media Backup / Restore 3.9 Simulator Operatirg Instruction S01-08; SIMEX and Foreground Management 3.10 Simulator Operating Instruction S01-10; Configuration Management System n 3.11 Simulator 08.erating Instructicn S01-11; Simulator Trouble Reports 3.12 Simulator Operating Instruction S01-12; Configuration Items and Design Data Base 3.13 Simulator Operating Instruction 501-13; CDB Update Logging 3.13 Simulator Operating Instruction 501-14; I/F Schematic Modification Logging 4.0 DEFINITIONS 4.1 REFERENCE PLANT - Crystal River Unit 3 4.2 DESIGN DATA BASE - those documents that describe the existing performance and physical appearance standards chosen as the simulator design boundar-ies; it encompasses a 5 .et of reference plant documents. 4.3 TROUBLE REPORT (TR) - a document which identifies a discrepancy between the simulator performance / appearance and the reference plant as defined by the existing simulator design data base. Resolution of a TR cannot result in a chance to the desian data base. For additional information, see S01-11, Simulator Trouble Reports. r k L var vs w
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- SIMULATOR CHANGE STANDARDS 4.4 SOURCE DOCUMENT - a document which 1) identifies or describes a change to the reference plant design or operation, 2) identifies a possible deficiency in the simulator design data base, or 3) an elective enhancement of the simulator capabilities which has no corresponding change in the reference plant. For additional information, see TOP 401, Simulator Source Document Evaluation.
4.5 SIMULATOR CONFIGURATION - those items that define the simulator's scope (software source code, documentation text, documentation crawings, engineering drawings, Handler code / text, malfunctions, and I/F schematics). 4.6 CHANGE CONTROL NUMBER (CCN) - A unique numeric designator assigned to track impact on the simulator configuration. 4.7 SIMULATOR CHANGE - the software, hardware and/or documentation modifications which are necessary to implement an " Initiating Document" g (Source Document or Trouble Report) on the simulator. ! I V 4.8 S!MULATOR CHANGE RECORD (SCR) - the "signoff" sheet associated with this procedure, which tracks the progression of any simulator Change due to a Source Document (new scope) or a Trouble Report (fix of old scope). The SCR is considered an integral piece of the Work Package. 4.9 WORK PACKAGE - the paperwork hardcopy which describes and documents the simulator Change. A Work Package is deemed to be complete when all relevant Work Package activities have a status of complete (CP) in CMS. Work Package activities may be software, hardware, or both. 4.10 CONFIGURATION MANAGEMENT SYSTEM (CMS) - the computer resident, relational data base program, used to support the simulator configuration control process. 4.11 CMS ADMINISTRA. TOR - the individual primarily responsible for logging CMS TRs/ Source Documents, assigning CCNs, and generating CMS reports. 4.12 HISTORY - a CAE written utility that performs a " smart" comparison between two FORTRAN source files. The purpose is to document all-changes to FORTRAN source code. O t<61 912 9%3
i NUCLEAR OPERATIONS TRAINING kg"j3'Re ('l 9 . ....... ... DEPARTMENT PROCEDURE o 26 ! l SIMULATOR CHANGE STANDARDS 4.13 COMPARE - a CAE written utility that performs file comparisons. This utility performs bit to bit comparisons. The purpose is to document all non-FORTRAN source code changes. 4.14 CDB - a Common Data Base file that contains the declarations for the various software labels in common blocks. This file is maintained separately. 4.15 ASSIGNEE - the individual (s) assigned resolution responsibility on the SCR. 4.16 0FFEREE - the person responsible for ensuring a Change is in a state conducive to acceptance testing - usually the offeree is one of the assignees. 4.17 ACCEPTEE - the simulator staff member responsible for testing the simulator Change to ensure the simulator's integrity is maintained. O 'd 4.18 SIMEX - a CAE written utility to control the simulator software
" foreground" configuration. The administrative controls pertaining to using SIMEX are fully described in 501-8, SIMEX and Foreground Management.
5.0 RESPONSIBILITIE$ 5.1 NUCLEAR LICENSED OPERATOR TRAINING HANAGER (NLOTM): 5.1.1 The NLOTM is responsible for the content and implementation of this procedure. 5.2 SIMULATOR MAINTENANCE AND ENGINEERING SUPERVISOR (SMES): The SMES is responsible for:
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5.2.1 Generating the Simulator Change Record (SCR) and assigning resolution responsibilities to one or more simulator software / hardware personnel. 5.2.2 Review of all simulator Work Packages to ensure compliance with this procedure. (3 u) 1Gr 0 J iP%3
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9 M h......,4.p NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS TDP-403, Rev 0 Page 7 of 26 l 5.3 NUCLEAR SIMULATOR TRAINING SUPERVISOR (NSTS): Tha NSTS is responsible for: 5.3.1 Assigning individuals for acceptance of Simulator Changes. 5.3.2 Optionally waiving the requirement for NOI acceptance of non-Housekeeping Simulator Changes and/or providing direction regarding the scope of acceptance testing. 5.4 SIMULATOR COMPUTER CONTROLS SPECIALIST (SCCS): l The SCCS is responsible for: 5.4.1 Implementing and documenting changes to simulator hardware in accordance with this procedure, 5.4.2 Ensuring that information contained in CMS is updated to reflect the g Work Package status and all changed Configuration Items. 5.5 SIMULATOR SYSTEMS ENGINEER (SSE): The SSE is responsible for: 5.5.1 Implementing and documenting changes to simulator software in accordance with this procedure. 5.5.2 Ensuring that information contained in CMS is updated to reflect the Work Package status and all changed Configuration Items. 5.6 NUCLEAR OPERATIONS INSTRUCTOR (NOI): 5.6.1 The NOI is responsible for acceptance of assigned Simulator Changes. l t l 9 147 02993
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9 gr. M u....... NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE TOP-403, Rev 0 Page 8 of 26 SIMULATOR CHANGE STANDARDS 6.0 ACTIONS 6.1 GENERATION OF A SIMULATOR CHANGE RECORD (SCR) 6.1.1 When a Source Document (TDP 401) or a Trouble Report (S01-11) has been determined to impact the simulator, a Simulator Change Record (SCR) is be issued. 6.1.2 The SMES shall initiate a Simulator Change Record (SCR), see Attachment
#1. On the top of the SCR, the SMES should log the Initiating Document, the Change Control Number (CCN), and the assigned individual (s). Note:
only one SCR is required for each Change. regardless of the number of assignees. 6.1.2.1 The Initiating Document number will simply be the Source Document number or the Trouble Report number (whichever is applicable). 6.1.2.2 S The Change Control Number (CCN) is auto-generated by CMS when the Source Document (anly for those w/ impact) or Trouble Report CMS record is originally logged. 6.1.2.3 The a; signee section should indicate the S/W and/or H/W assignees. Note, provisions are made on the SCR for re-assigning a Work Package activity if this becomes necessary (see step 6.2.7). 6.1.3 The SNES should enter the Work Package activity assignee (s) in CMS, indicating the target due date. The due date should be assigned by the SMES in accordance with priorities, training requirements, and the assignee work load (s). HQIll CMS automatically creates a record in the Work Package dBase with the associated CCN, Initiating Document number (SD or TR), and the CCN date. The SMES merely fills in the new data related to the assignee (s). For additional information pertaining to the CMS fields, see 501-10, Configuration Management System. 6.1.4 The SMES affixes the Source Document Evaluation Checklist (Attachment #2 from TDP-401) or the Simulator Trouble Report (Attachment #1 from 501-
- 11) to the SCR. This constitutes the creation of a Work Package.
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i S wa lO gp u......;o, NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS TDP-403, Rev 0 Page 9 of 26 6.1.4.1 It should be noted that although the Work Package and Source Document may be maintained together, they are two separate documents. 6.1.5 The SMES routes the Work Package and accompanying Source Document (if a;:plicable), to the relevant assignee. 6.1.5.1 In cases where a multi discipline resolution is required, the Work Package and the Source Document are generally ruuted to the H/W assignee first. l l 6.1.5.2 Once the H/W assignee has completed an initial analysis of the Work Package, enabling him/her to ensure correct parts are in stock or on
, order, the Work Package should be routed to the S/W assignee.
6.1.6 Periodic CMS reports will indicate the Work Packages assigned to an individual. Simulator Maintenance / Engineering personnel are responsible for all work assigned to them per these CMS status reports. In the event that th6 assignee determines that the scope of work clearly S exceeds the specified due date, he/she should immediately contact the SMES. Only the SMES has the authority to modify Work Package due dates. 6.1.7 The issuance of an SCR and associated activities described in this section may alternately be performed by SCCS/SSE personnel in order to l expedite the resolution of a TR (Housekeeping or otherwise). See 501-l 11, Simulator Trouble Reports. 6.2 PROGRESSION OF WORK 6.2.1 This section of the procedure provides instructions regarding the standards to be followed for all modifications to the simulator. 6.2.2 The work performed must meet or exceed the standards used by the Simulator Vendor (CAE Electronics). These standards are identified below:
- 1. CAE Hydraulic Models User's Guide (166701.01.J326)
- 2. An Introduction to COUPER Software for Reactor Modeling (CAE Modeling Primer)
- 3. An Introduction to Electrical System Simulation in Real Time (CAE Modeling Primer)
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- 4. An Introduction to Control System Modeling in Real Time (CAE Modeling Primer)
- 5. Documentation Requirements (Technical Directive #4)
(CD185403.04.8.300)
- 6. Documentation Requirements (Technical Directive #2)
(CD195198.02.8.300)
- 7. Application Software Documentation Standards (CD141749.01.8.800)
- 8. Technical Directive on Heat Exchanger Leaks (CD185407.08.8.300)
- 9. Power Simulators Non modeling Software Development Procedure (CD167619.01.8.800)
- 10. MTP (Module Text Processor) User's Guide h 11. FORT 77 SEL Reference Manual (Release 4.1) .
6.2.3 All changes to Configuration items must be documented according to actions specified in S01 12, Configuration Items and Design Data Base. Configuration Items include Software Source Code, Software Documentation (Text files), Handler Text / Code, Engineering Drawings, Documentation Drawings, I/F Schematics, and Halfunctions as defined by the Cause & Effect Reports. 6.2.4 Should the assignee require supplemental data and/or parts from outside sources before being able to proceed with the implementation of the Change, he/she should indicate so by updating the status field of the Work Package in CMS to (DR) or (PO) for Data Request and Parts on Order, respectively. This information will provide an indication of delays beyond the assignee's control. Again, if this delay will have an adverse effect on the existing Work Package due date, the SMES should be g noti fied. 6.2.5 Should the scope of the Change require modifications to any of the Configuration Items, the SCCS/SSE should verify the revision level of the current Configuration Item via CMS. The work should be performed on the " current" version. This is to ensure that the work performed is a continuation from previous modifications.
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SIMULATOR CHANGE STANDARDS 6.2.6 For TR based changes whose resolution will not be worked on or resolved
" tagging out" the affected componen)t.in short order (ie., within I week ,
The intent is to openly communicate that the problem is known such that duplicate TRs are not written. 6.2.6.1 Tagouts are performed using the yellow MAINTENANCE DEFICIENCY TAGS (MDT), available from the SCCS. 6.2.6.2 The TR number is placed on the tag in the field labeled "MR#". 6.2.6.3 The of theassignee's MDT. initials are placed in the "init" location on both halves 6.2.6.4 The the MDT. date of the TR is placed in the "date" location on both halves of 6.2.6.5 A very brief description of the problem should be provided in the " tag
#" field. (ie., punp won't auto start).
9 tag # field should reflect the TR number,On the tag's bottom half, the . 6.2.6.6 The top half of the Maintenance Deficiency Ttg is placed on the control room equipment most indicative of the TR's impacted component. This enables operators and instructors to determine that a problem exists and has been documented. 6.2.6.7 The i tselbottom f. Per portion of the Maintenance Deficiency Tag is placed on the TR step 6.5.7.1, the portion affixed to the panel will ultimately accepted. be removed and attached to the TR when the Change is 6.2.7 If an assignee determines that the Work Package activity assigned to him/her reassigned as follows: is not within his/her scope of responsibilities, it is to be 6.2.7.1 Consult assignedwith to. the person to whom the Work Package activity should be This ensures that both parties are in agreement. 6.2.7.2 Inform the SMES. I _ _ _ _ _ _ - - " " ^ ^ ^ ^
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SIMULATOR CHANGE STANDARDS 6.2.7.3 The SCR is modified by crossing out the present assignee's name and inserting the new assignee's name in the RE-ASSIGNEE column. 6.2.7.4 The CMS database for Work Packages should be updated by the original assignee. ~ 6.2.7.5 Forward the Work Package to the new assignee. 6.3 HOUSEKEEPING CHANGES 6.3.1 Although not technically discrepancies, the simulator staff incorporates various " Housekeeping Changes" to the simulator Configuration Items via TRs. This ensures all simulator Changes directly affect performance or appearance)(even those that have adequate do not traceability. 6.3.2 A Housekeeping Change is defined as a change in the simulator configuration (software, hardware, or documentation) which does not n affect the simulator from the perspective of either the instructor or t,") the operator. That is, a Housekeeping Change does not improve or degrade the fidelity of the simulator as compared to the reference plant. Typical Housekeeping Changes include:
- 1) Changes in documentation text or drawings (software or hardware) that are not accompanied by other changes which invalidate the Housekeeping definition. The release of a new revision to a Documentation or Engineering Drawing is typical of a Housekeeping Change.
- 2) FPC (FORTRAN precompiler) of any modules and any related "CP" block changes.
- 3) FORTRAN comment changes for clarity or cosmetic purposes.
- 4) ZPC (Hondler processor) of any module due to an approveo Handler change that does not necessitate any revision of the CALL statement arguments, ie., a " black-box" change.
- 5) Module Text Processor (HTP) delimiter changes.
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- 6) COB updates which make temporary spare labels unnecessary. The scope of change is limited to a) replacing the spare label in the "CP" block with the new permanent label, b) deleting any related EQUIVALENCE statements, and c) FPC (FORTRAN pre-compile) the module.
- 7) Removal of unused variables (CDB or internal) l A) Renaming variables (CDB or internal) for adherence to standards.
- 9) Modification to an I/F Schematic originating as a secondary result of a previous change.
6.3.3 The SMES must be consulted for any Housekeeping Changes which may be subject to interpretation. 6.3.4 By definition, Changes denoted as Housekeeping provide the SCCSs/SSEs with imediate authorization for acceptance testing. This will significantly expedite resolution of TRs that have no performance or - appearance impact, ie., the NOI acceptance is unessential. 6.3.5 If a TR can be classified as a Housekeeping Change, the SCR shall be marked accordingly by an "X" in the box marked " HOUSEKEEPING", see Attachment #1, 6.3.6 The declaration of a TR as a Housekeeping Change does not remove the traceability requirements, only the NOI concurrence. That is, all changes to Configuration Items must still be documented in accordance with 501-12, Configuration Items and Design Data Base. 6.3.7 The remaining work proceeds per the " normal" process except that the j offeree also functions as the acceptee. 6.4 PLACING A SIMULATOR CHANGE ON OFFER 6.4.1 Placing a Change on Offer indicates that the modification is functional; the Work Package need not be complete for the Change to be considered functio,al, i
@J1i This enables the Change to be available for acceptance testing and subsequent training in as short a time frame as possible.
It also avoids the possibility that the documentation work is performed on a Change that may not be acceptat,le. M
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SIMULATOR CHANGE STANDARDS 6.4.2 A Work Package is considered " functional" when the entire Work Package is operational, enabling a NOI to utilize the modifications of the Work Package for acceptance testing and subsequent training with minimal inconveniences. These inconveriences are discussed subsequently: 6.4.2,1 The Documentation need not be completed. This will not impact the functionality of the Change. 6.4.2.2 The CDB need not normally be updated for the Change to be offered. This sometimes results in CDB labels that have incorrect description fields. The SSEs must insure that CDB spare labels are used in accordance with S01-13, CDB Update File Logging. Should there be insufficient spare labels for proper operation of the simulator, the CDB must be updated prior to offering the Change, ie., store point dependant variables must reside in the CDB. 6.4.2.3 Instructor Controlled variables that are not accessible from the I/F or p are temporarily replaced by spare labels accessible from CTS i'j (Computerized Test System) are acceptable. This adds a slight - inconvenience to the NOI staff. However, if they are informed of the need to access the labels via CTS they will be able to utilize the Change. 6.4.2.4 A"y requirement to use RTD (Real Time Debugger) is totally unacceptable. 6.4.2.5 Hardware that is functional but does not have all nf the mimics er escutcheon plates available, can be considered functional. At the discretion of the SCCS, a temporary label can be used to identify a component. 6.4.3 The above itema (C.4.2.1 thru 6.4.2.5) are only acceptable should they be the " secondary" result of a Change and not the " primary" result. For example, if a TR is raised indicating that a T variable is not accessible from the I/F, the inability to access the failure is considered the primary problem. In such a case, item 6.4.2.3 discussed above is not an acceptable inconvenience and the TR cannot be considered functional by accessing the T variable from CTS. 6.4.4 The SMES should be consylted for any " inconveniences" which may be subject to interpretation prior to offering the Change. b) o vo' -
N dda TDb 403, Rev 0 gg NUCLEAR OPERATIONS TRAINING Page 15 of 26 a.....,4, DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS t 6.4.5 The offeree is the person responsible for ensuring that all aspects of the Work Package are completed to the " functional" status described above. The offeree for a Work Package will generally be an assignee. If a Change has both H/W and S/W activities, the S/W assignee sill typically be the offeree. This is simply due to the fact that the entire package can only be fully tested when the H/W is available for integrated testing, typically performed by the S/W assignee. 6.4.5.1 The offeree is responsible for compliance with S01-08, SIMEX and Foreground Management. This will ensure the simulator training and testing configurations are properly controlled. 6.4.5.2 For Simulator Changes which require modifications to the foreground configuration, the offeree is required to test the resolution prior to placing the Change on "0FFER", The offeree is responsible for determining the scope of testing required; where possible utili7ing existing simulator Exercise Guides, Performance Test Procedure:,, or o plant ops /SPs. Include test results where possible (graphic recorder ( o data, VSA overlays, PPC outputs, comparison photos with the plant, CTS test results/ dumps, etc.) in the Work Package. 6.4.5.3 Simulator Changes which are relatively minor in scope are to be tested in accordance with the above steps; however, due to their simplistic nature the test results mhy be excluded from the Work Package. The interpretation of " minor" is at the discretion of the offeree. The offer signature indicates that the Change has been tested to the satisfaction of the offeree. 6.4.5.4 The purpose of validation testing is to ensure 1) the simulator modification has proceeded correctly, and 2) the dynamic or logical behavior of the simulator has not been corrupted indirectly by the modification. 6.4.6 Should the S/W assignee wish to assume acceptance responsibilities for a non-Housekeeping Change, they should contact the NSTS for such "ethorization prior to the testing phase since the NSTS may provide
.eptance testing instructions that mio differ from those intended to le used. The SCCS/SSE should only seek this authorization when he/she are confident that the scope of acceptance testing does not exceed his/her abilities. Relative work load between Maintenance and Training should also be considered. See Section 6.5 for more information regarding the transfer of acceptance testing from NOI to SCCS/SSE personnel, m
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O ~9n.....3 DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS 6.4.7 The offeree should briefly describe the action taken to resolve the Change on the SCR, particularly any required IC Deposit Files, foregraund configuration changes, or special instructions that may be required during acceptance testing. See 501 08, SIMEX and Foreground Management for guidance as to the standard location of Deposit Files, or S01 13 CDB Update File Logging for guidance on CDD update files. 6.4.8 The SCCS/SSE should change the Activity Status in CMS for the Work Package recnrd to functional (FU) or complete (CP), whichever is applicable. Should the Work Package contain both S/W and H/W activities, the offeree is responsible for ensuring that both activities are log Change.ged either functional (FU) or complete (CP) prior to offering the
!!Q1L;, A Work Package CMS Activity Status of complete (CP) should be O reserved for those Instances where the assigned activity has V been completed in accordance with section 6.6 prior to offering the Change. That is, the assignee should not need to see the Work Package again. If the Change is accepted, the Initiating Document can be imediately closed (CL). The SCR indirectly indicates the activity's status is complete (CP) by the assignee's signature and date in the appropriate columns.
In such caser, the CMS Administrator will not route the Work Package back to the assignee after the Change is accepted. 6.4.9 The offeree signs the SCR in the "0FFER" location indicating that the Change has met the functional requirements of this precedure and has been tested satisfactorily. 6.4.10 Should the Change require a modification to the foregrout.d configuration software, 't HISTORY comparison (see 501-12) must be included in the Work Package at this time. 6.4.11 Place the Work Package in the *0N OFFER" tray in the lastructor's booth. k V
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NUCLEAR OPERATIONS TRAINING p!h*!3'o "fj O *9to..e... .. DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS 6.4.12 The CMS Administrator collects the offered Work Packages periodically, changing the status of the Initiating Document in CMS from open (0P) to offered (OF). 6.4.13 The Work Package is then routed to the NSTS for assignment of a N01 for acceptance testing. 6.5 ACCEPTANCE OF A SlHULATOR CHANGE 6.5.1 At the discretion of the NSTS, the responsibility for " accepting" the Chang' resolution may be transferred from the N01 to the SCCS/SSE. This shoLid be reserved for limited scope discrepancies where the Work Package testing methodology is trivial. 6.5.2 Should the SCCS/SSE determine that the scope of acceptance testing is within his/her abilities, he/she may seek the authorization for non N01 acceptance. See step 6.4.6. 6.5.3 Housekeeping Changes, as defined in section 6.3, need not be accepted by - the NSTS or a designee. In these cases the Simulator Change is both offered and accepted by the SCCS/SSE, and the box marked Housekeeping Change on the SCR is checked. Work then proceeds in the same manner as if the NSTS has given authority to accept the Initiating Document. 6.5.4 As indicau.d in step 6.4.6, authorization should be requested prior to the preliminary testing phase so as to minimize duplicate testing. 6.5.5 The NSTS shall indichte authorization fnr non N01 acceptance by checking the Box labeled " Authorization for non N01 acceptance" on the SCR. tlQIra, in cases where the NSTS elects not to transfer the acceptance responsibility, a N01 must accept the Change. 6.5.6 Should the NSTS wish to provide specific testing instructions, the box on the SCR marked " Acceptance Testing instructions Attached" should be checked and the test instructions placed in the Work Package. Unless the SCR explicitly specifies required testing, the N01 (or designee) is responsible for determining the scope of testing required; where possible utilizing exitting simulator Exercise Guides, Performance Test Procedures, or plant ops /SPs. O 4.,
(3 9M 5 NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS p!g"[!$,R 2l 6.5.6.1 Since only limited time is available to accept the Change, it is incumbent on the N01/ designee to identify a limited number of tests which challenge the Simulator Change to the maximum extent possible. The N01/ designee should select tests which:
... take into account the complexity of the Change's resolution.
... allow a true comparison between reference plant data (or operational experie.' ice) and the simulator.
... ensure the affected component / system is tested under multiple modes of operation
... verify the integrated response of the component / system.
... provide indications that the Work Package implementation has not indirectly introduced another discrepancy.
V If the Change implementation is acceptable, the N01 (or designee) shall indicate the Work Package resolution is " ACCEPTED" by signing the appropriate field of the SCR. Signature indicates that adequate performance testing was performed and tht the Change is considered fully functional and acceptable. The comment field on the SCR is available should the N01 (or designee) wish to add comments regarding the testing performed or response seen. 6.5.7.1 If a Maintenance and Deficiency Tag was issued, the portion attached to the H/W panels should be removed and attached to the TR by the acceptee. 6.5.7.2 The Work Package is then placed in the " ACCEPTED" tray in the instructor's booth. 6.5.8 If the Change's implementation is not acceptable (the Change is
" REJECTED"), the N01 or designee should provide coments indicating the reason for rejection. The SCR is signed and dated in the appropriate field and the WP is ) laced in tne " REJECTED" tray in the instructor's booth. Rejected Wor ( Packages will be re-worked until the implementation is satisfactory.
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$1MULATOR CHANGE STANDARDS 6.5.9 If " unrelated" discrepancies are noted, new TRs should be written to document the observed problem (s).
6.5.10 The CMS Administrator periodically collects the Work Packages from the Accep'ed and Rejected trays. 6.5.11 If the Work Package has been rejected, the CMS record for the corresponding Initiating Document is logged as rejected (RJ), indicating that work is still in progress. The Work Package is then routed back to the offeree, as indicated by the "0FFERED BY" signature. 6.5.11.1 The offeree is responsible for reworking, or coordinating with the other assignee (s), the Change until satisfactory results are obtained. Af terwards, the Change should again be placed on offer in accordance with stction 6.4. 6.5.12 If the Work Package has been accepted, the CMS record for the 'o corresponding Initiating Document is logged as accepted (AC). The Work Q Package is then routed to the assignee for completion. - E0lf i. If all assigned Work Package activities are already complete as indicated by SCR signature and date in the appropriate fields, the CMS Administrator should forward the Work Package to the SMES instead of the assignee. When the assignee receives a Work Package that has been accepted, he/she is responsible for completing the Work Package in accordance with section 6.6. 6.6 COMPLETION OF THE WORK PACKAGE 6.6.1 At this stage in the Change process, the Work Package will contain, as a minimum, the SCR and either the Source Document Evaluation Checklist or a Trouble Report. It may also contain test results and Ulstory comparisons for any changes to foreground configuration. 6.6.2 All changes to Configuration Items must be documented in accordance with 501-12, Configuration items and Design Data Base.
P NUCLEAR OPERATIONS TRAINING $g (3 V ,m 9e........ DEPARTMENT PROCEDURE
$1HULATOR CHANGE STANDARDS 6.6.3 Documents which are not identified as Configuration items should be modified in accordance with the existing format and level of detail.
The Work Package is to include the documentation of the change to these items at a level of detail comparable to a Configuration item. The scope of the inclusions in the Work Package should enable the Change to be fully reproduced from the Work Package contents. 6.6.4 Complete the Work Package Checklist. The intent of the checklist is to ensure all aspects of the Change have been properly evaluated. 6.6.4.1 The Checklist should document the updated Configuration Item and the revision level. There must be a one to one correspondence between this list and the entries in CMS. 6.6.5 The SCCS/SSE is responsible for updating the CMS record for any Configuration itsms that were affected by the Change, indicating the current status, revision level, and CCN. See 501-10, Configuration Manage'nent System, for more details. p! ( . 6.6.6 Once an individual assignee has completed the tasks relevant to his/her discipli s (H/W or S/W) the Work Package Activity status should be logged completed (CP) in CMS. 6.6.7 Once completed and logged in CMS, the assignee signs and dates the SCR in the relevant fields to indicate the completion of his/her assigned activity for the Change. 6.6.8 Any document which may assist the SCCS/SSE in recreating or understanding the Change documented by the Work Package should be included. This could simply be " scraps" of hand written notes or calculations. These miscellaneous items may be included in the Work Package at the discretion of the SCCS/SSE. 6.6.8.1 Examples of such ancillary items includes marked up drawings, installation instructions, parts specifications, design analysis, worksheets, costs, or expenditures. 6.6.9 If the Work Package has a single assigned activity, the Work Package is routed to the SMES for closure approval. (3 V L- . . -- --
t R NUCLEAR OPERATIONS TRAINING $g"[!!,o 2 o......n DEPARTMENT PROCEDURE
$1MULATOR CHANGE STANDARDS t 6.6.10 if the other assignee has not completed his/her cctivity, the Work Package should be routed to him/her for completion. When all activities are completed, the Work Package is routed to the SHES for closure approval.
6.6.11 The SMES shall approve the Work Package by 1) verifying that all components of the submitted Work Package conform to the applicable standards, and 2) the SCR has been properly completed. Note, the SMES cannot approve the CMS closure (Ct.) of the initiating Document until both H/W and S/W Work Package activities are CMS logged as completed (CP). Any questions or problems should be directed to either the Work Package assignee or offeree as appropriate. 6.6.12 Route the Work Package to the CMS Administrator. 6.6.13 Upon authorization from the SMES, the CMS Administrator should log the status of the Initiating Document (either Source Document or TR) as D closed (CL). 6.6.14 The Work Package is filed and maintained in accordance with section 6.7. 6.7 00CUMENT CONTROL & QUALITY RECORDS 6.7.1 This section provides instruction on the submittal of Work Packages to Records Management. 6.7.2 An initial baseline of Configuration Items shall be submitted to Quality Records and stored indefinitely within 3 months of declaring the simulator Ready-For-Training (RFT). This baseline will reflect the simulator as built by the vendor. 6.7.2.1 This baseline is effectively updated and transmitted to Records Management via Work Packages. 6.7.2.4 The baseline set for the Engineering and Documentation drawings will not explicitly be updated. However, when an Original drawing is updated and the Master re-issued, the drawing will be included in the Work Package. This will essentially " replace" the baseline version of the drawing. E. .,- - m
NUCLEAR OPERATIONS TRAINING $",4ll' o"f 2! 9 ....... .. DEPARTMENT PROCEDURE 9 SlHULATOR CHANGE STANDARDS 6.7.3 The Master drawings and Hylar Original Reproducible drawings should be stored in dedicated storage cabinets in a HALON protected room. 6.7.4 Once a Work Package has been completed (all Work Package activities logged CP in CMS), approved by the SHES, and the Initiating Document (Source Document or Trouble Report) logged as closed (CL) in CMS, the CMS Administrator shall file the Work Pachage in the Temporary Storage Cabinet, in accordance with TOP-109, Training Records Management. The Temporary Storage Cabinet shall be a one (1) hour rated fire proof cabinet. I 6.7.5 The CMS Administratur is responsible for ensuring that the Work Package is transmitted from temporary storage to Nuclear Operations Records Management for permanent storage within 90 days. Note, the Work Package should be maintained in temporary storage for a minimum of 60 days. This will ensure that the Work Package is readily accessible should a problem be encountered. 6.7.6 For each Work Package to be transmitted, the CMS Administrator should make a copy of the Simulator Change Record (SCR) and any Work Package Checklists. These copies should be retained in the Archived Work Package File sequentially by CCN. 6.7.7 The CMS Administrator forwards the Work Package to the Training Clerk for transmittal to the Nuclear Operations Records Manager in accordance with TDP-109, Training Records Management. 6.7.8 Once the Work Package is transmitted, the Training Clerk will forward a copy of the Transmittal form to the CMS Administrator. Upon receipt, the CMS Administrator should file the transmittil copy in the Archived Work Package File. l i I
I l R NUCLEAR OPERATIONS TRAINING $g7N, o 2
.......... DEPARTMENT PROCEDURE SIMULATOR CHANGE STANDARDS 6.8 DISK PACK BACK UP AND S10 RAGE 6.8.1 The simulator CDC disk drive packs are backed up on a periodic basis as directed by the SMES. The instructions and administrative controls pertaining to this process are described in 501 07; Magnetic Media Backup / Restore. By performing both incremental backups to tape, and full backups to disk, the integrity of the simulator magnetic media should be adequately protected.
6.8.1.1 As a minimum, three " sets" of disk packs should be rotated and maintained at the simulator facility (stored in the computer room).
- Offsite storage of disk packs should commence within 3 months of l declaring the simulator Ready for Training (RFT).
i 6.8.1.2 Additionally, a single " set" of disk packs should be stored at the CR-3 Nuclear Plant site in the Computer Services Department. This set should be rotated on a periodic basis not to exceed 6 months. Offsite storage is considered essential in the event of a major disaster at the simulator facility. 7.0 ATTACHMENTS 7.1 ATTACHMENT 1 - SIMULATOR CHANGE RECORD l l 7.2 ATTACHMENT 2 - WORK PACKAGE ACTIVITY CHECKLIST 8.0 OVALITY RECORDS WORK PACKAGE AND CONTENTS (including Simulator Change Record) 9.0 INTERPRETATION CONTACT SIMULATOR MAINTENANCE ENGINEERING SUPERVISOR I 10.0 REVISION HISTORY Rev. 0 Date: 02/01/91 Original Release
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.W1 NUCLEAR OPERATIONS TRAINING DEPARTMENT PROCEDURE
$1MULATOR CHANGE STAP0ARDS p$g";"$!,Ro 11.0 REVIEW HISTORY
_Date Person Conductina Revie t Comments l' . L__ _-_____L_--_--_------ - - - - - '~--
10P 403, Rev 0 Page 25 of 26 ) . DD 51MLeLOR_CliMELRECORD Routing: Ort Records
.M Attachment 1 Copy: WP file Initiating Document 8: CCN:
ASSIGNEE RE ASSIGNEE COMPLEi;0 BY COMPLETION DATE S/W: H/W: O Authorization for non N01 acceptance O Accaptance instructions attached DATE: N515 mo/ day /yr b Housekeeping Change only ACTION: I 0FFERED BY / DATE ACCEPTED BY / DATE REJECTED BY / DATE COMMENTS: NOTE: All Work Package activities must be CMS COMPLETE (CP) and include a Work Package Checklist before the Initiating Document can be submitted for Closure Approval. Further, all Configuration items should be logged CURRENT (CU) in CMS. CLOSURE APPROVED BY: DATE: I 3RES mo/ day /Vr INITIATING DOCUMENT LOGGED CLOSED BY: DATE: mo/ day /yr
/
10P 403, Rev 0 Page 26 of 26 WORK PACKAGE ACTIVITY CHECKLlil Routing: Orig: Records 9 CCN: INITIATING DOC: Attachment 2 Copy: WP file DATE: mo/ day /yr O Configuration items were modified and logged as CURRENT (CV) in CMS mark affected item types below with a (X) in circle.
" Redlines" of Cor. figuration items are in the Work Package.
O SIMEX Source Files 0 Documentation Text 0 Documentation Drawings O Engineering Drawings 0 Handler Code 0 Handler Text 0 1/T Schematic Pages 0 Numbered Malfunctions 0 The Change afrected the S;mulator Design Data Base. Provide details under COMMENTS or on a separate attachment. O roregro2nd Module (s) were put in configuration per S01 08. O initiai Condition (s) were modified and/or oeposit rile (s) were created to use the Change. Provide details under COMMENTS. 6 O Spare cob 1abels w2r used w/ Spare tabel tog updated, and/or - COB " update" files wire created per 501 13. O 1/F Schematic changes have been deferred and logged per S01 14. O INFACE was modified in accordance with 501-12 (SEL disk & CMS). O PROGEN was modified in accordance with 501 12 (hardcopy & CMS). O New Spare Parts, PM, and/or Maintenance Manuals are required. CONFIGURATION ITEMS AFFECTED: COMMENTS: I CHECKLIST COMPLETED BY: DATE: SCCS/SSE mo/ day /yr
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'V U Page so. 4 c4 CLOSED TRCUSLE REPORTS 09/09/91 TROUBLE REPORTS REtATED utMBER D.*.TE ORIGINATOR STAT PRORLEM IFACT PR!3t!TY DOC 3CET ATION CDI 00034 03/14/91 40L**ES, M. OP SOURCE RAMGE thDICATORS It t!/RPS CASS READ 800 CPS ALWAYS S C 910075 00032 03/13/91 GE NT R Y, 8. OP SCALES OF EF-98 & 99-Lt2 ARE 14 ERROR (REF DR-833) MS C C10073 00031 03/13/91 GENTRY, S. OP SCAllhG FOR BS-16 & 17-PI2 ARE Im ERROR (DR-E29) MS C 910072 00030 03/13/91 GENTRY, 8. OP SCALES OF EF-23,24,25, 1 26-F12 ARE IN ER*CR (04-823 $ 836) as C 910c71 00029 03/13/91 GENTRY, B. OP SCALES OF RC *58-P12 & 159-PI2 ARE tmCORRECT (REF 1,4-817) PS C 9'0070 00026 03/04/91 GENTRY, B. OP ST ATE MOT RINN PuCNE NEEDS TO BE RELOCATED (OR-683) 4 C 910068 00025 03/04/91 GENTRY, B. OF E#C PAhEL Atw JICOW COLOR DISCREP. (RET 02-477 & 478) M C 910067 00017 02/22/91 GauwSIR, R. Or ALRM PT. 798 40T FUNCTIONAL (SEC CYC COC NI) FROM SS-71-C!d S B 910C59 00014 01/25/91 AcBUTHCT, C. OF WJ4 MALF SC-0G3A PAS SEYE8AL PROCESS AuD I/F Au@ALIES S A PTm38/54sS253 910C56 00011 01/14/91 Sm!TM, D. OP 04 SEVERAL ES MALF., SevE*AL Sta. STAT. LTS. DID WOT Go CRJi S C PT=118/StA',-6134 910010 iME TOTAL NLMBER OF RTEDRDS FOR Tuts REPORT IS: 109 EIO OF REPORT
(h () FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR INITIAL SIMULATOR CERTIFICATION TABLE OF CONTENTS VOLUME 11 Performanen Test Abstracts BOOK 1 COMPUTER REAL TIME TEST PTC) - Computer Real-Time Test STEADY STATE AND NORMAL OPERATIONS TESTS PTN1 - Plant Heatup and Startup From Mode 5 to 100% Power PTH2 Reactor Trip and Recovery to Rated Power PTN3 - Load Changes PTN4 - 3 RCP Plant Shutdown and Startup PTN5 Plant Shutdown and Cooluown to Mode 5 PTN6 Core Performance Testing g PTN7 - Operator Conducted Surveillance Testing PTSI - Steady State Performance Test TRANSIENT TESTS PTil - Reactor Trip PTT2 Total loss of Feedwater (Main and Emergency) PT13 - Simultaneous Closure of all Main Stea., Isolation Valves PTT4 - Loss of 6900 VAC and 4160 VAC Unit Busses from Full Power PTTS - Reactor Coolant Pump Trip PTT6 - Main Turbine Trip PTT7 - Maximum Rate Power Ramp PTT8 Large Break LOCA(DBA) PTTQ - hain Steam Line Breck Inside Containment s Failure of PORV with HPl Inoperable MA' ,il0N TESTS PTM1A1 Steam Generator Tube Leak PTM1A2 - Steam Generator Tube Rupture PTMlB Letdown Line Leak Downstream of MUV-49 O ( . . J
' FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR INITIAL SIMULATOR CERTIFICATION TABLE OF CONTENTS (Cont'd) 10LUME 11 - Performance Test Abstrich fl0All MALFUNCTION TESTS PIMICI Repressurizing Small Break LOCA Inside Containment PIMIC2 - Small Break LOCA Inside Containment P1 HIC 3 - Small Break LOCA With Loss of liPI PIMID Pressurizer Safety Valve failure PTM2 - Loss of Instrument Air P1H3A Loss of Offsite Power (Low Decay lleat) PIM381 - Loss of Emergency Of f Site Source P1H3B2 - Loss of Emergency Off Site Source With A failure Of 1 EDG PTM3Cl Loss of NN1-Y i 24 VDC Power
,- P1M3C2 - Loss of NN1-Y AC Power PTM301 - Loss of ICS DC Power
/ ). \' P1M302 - Loss of ICS AC Power PIM3E - Loss of Vital Bus PTM3f - Loss of 480 VAC ES Bus A PIM3G - Loss of DC Distribution 73
T-Q TEST NUMBER / TITLE: PTCl/ COMPUTER REAL-TIME TEST ANSI /ANS-3.S(1985) CROSS
REFERENCE:
Appendix A, Section A3.1 DATE PERFORMED: 08/15/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100%, a large Break LDCA/ Loss of Offsite Power Transient is initiated. Using computer utilities, real-time operation of the simulator / simulator computers is verified through timings of all of the foreground simulation software modules. tLALFUNCTION IIILE QPl[QNS INITIAL CONDITIONS: Not applicable for this test. Simulator response to a large break LOCA i, verified in PT18. FINAL CONDITIONS: Not applicable for this test. Simulator response to a large break LOCA is verified in PIT 8. PLANT PROCEDURES USED DURING THE TEST: Not Applicable for this test. BASELINE DATA: Not applicable for ti.is test.
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.- . - . - _ - . - .. . _ _ - _ . . .- - . _ - . - . - . _ . . _ _ - . . - . _ - - = . -
PERFORMANCE TEST NUMBER: PTC) DATE: 08/15/91 TlTLE: COMPUTER REAL-TIME TES1 1 TEST RESULTS/ COMMENTS: Test results satisfactory. Average average and average instantaneous spare times for the test runs indicated 45% spare. Periodic low i1stantaneous spare times were noted for both CPUI and IPUl. These processors include the Plant Process Computer software, which is run as an asynchronous task. Low or zero instantaneous spare times on these processors will only effect completion of these tasks and will not impact real-time execution of plant systems simulation (synchronous) software. O l l l r _ m O
_ . . _ _ . _ _ . _ _ . - - ... _. _ _ _ . _ . _ _ ~ . _ _ __ _ _ . . _ _ . _ _ _ . .__
] TEST NUMBER / TITLE: PTC1/ COMPUTER REAL-TIME TESI TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
' Critical Parameter DEVICE CHART SPEED YESBICTR Avg Avg Spare Time Grfx Rcdr 0 - 100% 0.5 seconds YSCYCAVG 1) CPVO Avg Spare Time Grfx Rcdr 0 - 0.05 sec YSCYCAVG 2) IPU0 Avg Spare Time (s Grfx Rcdr 0 - 0.05 sec YSCYCAVG 3) CPUI Avg Spare Time s Grfx Rcdr 0 - 0.05 sec YSC)CAVG 4) IPUI Avg Spare Time s Grfx Rcdr 0 - 0.05 sec YSCYCAVG5) CPU 2AvgSpareTime s) Grfx Rcdr 0 - 0.05 sec YSCYCAVG(6) IPU2 Avg Spare Time s) Grfx Rcdr 0 0.05 sec YESAICTR Avg Inst Spare Time Grfx Rcdr 0 - 100%
YSTIMSPA(1) CPU 0 Int.tSpareTime Grfx Rcdr 0 - 0.05 sec YSTIMSPA(2) IPU0 inst Spare Time Grfx Rcdr 0 0.05 sec YSTIMSPA(3) CPUI Inst Spare Time Grfx Rcdr 0 - 0.05 sec YSTIMSPA(4) IPUI inst Spare Time Grfx Rcdr 0 0.05 sec YSTIMSPA(5) CPU 2 Inst Spare Time Grfx Rcdr 0 0.05 sec YSTIMSPA(6) IPU2 Inst Spare Time Grfx Rcdr 0 - 0.05 sec O l l
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PERFORMANCE TEST NilMBEr.: PTC) DATE: 08/15/91 TITLE: COMPUTER REAL-TIME TEST 1EST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed LEE
@ o ator fbe 2.0 All deficiencies have been resolved and this test has been completed.
O V NOT REQUIRP.D TPMcTtloordinator 7D' ate 3.0 Test results reviewed. Simulator performance t .tisfactory.
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APPROVED: M<i(, Osht-( tam DATE: Af'Lj_gj
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N arRf0RMANCE TESf NUMBER: PICl DATE: 08/15/91 (]} TITLE: COMPUTER REAl-TIME TEST STEP # TEST PROCEDURE INITIALS / TR HUMBER APPFyQ1X_6 PIREQfLMARGL1Dl_QE10 RIP _UM The simulation models are dis;ded into a finite number of tasks. The dispatcher follows a predefined sequence to activate a combination of these tasks every 50 milliseconds. Each combination has been structured so that it takes less than 50 milliseconds to execute in order to allow spare computer processing time until the activation of the next combination of tasks. Also, the predefined sequence determines the iteration rate of each individual model by activating different tasks at different frequencies; that is, models that run every 50 milliseconds are in tasks that are activated every iteration, models that run every 100 milliseconds are distributed in tasks that are activated on alternate iterations, and so on. Since the majority of the software models have a 200 millisecond iteration rate,-the majority of the models are distributed in tasks that execute once overy four 50 millisecond iterations gi.e., approximately one quarter of these models are activated each iteration).
,r3 The software models have been designed to reproduce CR-3 response in real time by v executing repetitively at a constant rate and utilizing the design time period between executions in any time-based calculations, in order to be consistent with its time period, a model must execute at its required frequency or iteration rate. It follows that as long as the models execute at their required frequency, their time based calculations will be valid and, therefore, real-time operation will be ensured.
The order of execution has been structured so as to provide spare computer processing time between iterations in order to allow the model execution times to ( increase over their normal values without affecting the start of the next iteration, in this way an individual model's actual execution rate is maintained within a few milliseconds of its designed rate, Any small deviation in an individual model's execution time for a given iteration is negligible compared to the design execution rate of the model and becomes totally meaningless over the course of a few model executions. If a processor takes longer than 50 milliseconds to complete a 50 millisecond iteration, a frame overrun is declared, if a frame overrun occurs, it will delay the start of the next 50 millisecond iteration. If the succeeding iteration or iterations contain enough spare time, the simulator may recover after-1 or 2 iterations and the time period between successive executions of individual models will remain constant and correct, llowever, if frame overruns occur continuously, in successive iterations, the time period between individual model executions i becomes longer than its design value and, therefore, real-time operation is lost.
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Q PERFORMANCE TEST NUMBER: PTC) DATE: 08/15/91 TITLE: COMPUTER REAL-TIME 1EST STEP # TEST PROCEDURE INITIALS / TR HUMBER APPENDIX A PJRFORMANCE TEST D[$CRIPTION Despite the fact that occasional frame overruns can occur without causing real-time degradation, the simulator has been designed to operate with no frame overruns. In order to discover potential problems, an alarm in the instructor booth actuates whenever a processor experiences more than 50 overruns. The CR3 simulator has a program called the Performance Utility (PFU) which provides comprehensive information on the execution of the software models. It can display the model order of execution, monitor and display the execution time of any model, task or combination of tasks, determine and display the spare time between successive iterations, detect and display frame overruns in any processor, and provide overall simulator summaries of all of the above information. PFU was used extensively to design the order of execution of the simulator models and is still used periodically to verify that the models are executing properly. O A computer real time test is used to verify the absence of frame overruns and to verify that the simulator responds in real time with sufficient spare time. It should be noted that background tasks (ie., model development) are run at much lower priorities than the foieground simulation modules. Therefore, background activities cannot affect real-time operation of the simulator. O
TEST NUMBER / TITLE: PIN 1/PLAf4T HEATUP Afl0 STARTUP fROM MODE 5 10 100% POWER ANSI /ANS-3.5(1985) CROSS
REFERENCE:
DATE PERF0P"ED: 01/06/91 - 01/08/91 TEST DURATION: 16 hours TEST DESCRIPTION: . e sirrulator is operating at Mode 5 conditions (<200 degrees) with the "B" DH Removal system :n operation. The sitaulator is maneuvered from Mode 5 to 100% full power using approved plant procedures. [LALFUNC110B llTM 0PT1.0RS INITIAL CONDITIONS: Mode 5, "B" DH Removal System in operation and all RCP's shutdown. FINAL CONDITIONS: 100% power, ICS in full automatic operation. PLANT PROCEDURES USED DURING THE TEST: OP-202 REV 84 PL ANT HEATUP OP-212 REV 3 OPERA 110NS lf4VOLVlf1G THE MSR OP-210 REV 29 REACTOR STARTUP SP-422 REV 19 RC SYSTEM H/V Af40 CD/D SURVElLLAf4CE OP-203 REV 61 PLANT STAR 10P OP-703 REV 34 PLANT DISTRIBUTION SYSTEM OP-605 REV 48 FEEDWATER SYSTEM OP-402 REV 66 MAKEUP Afl0 PURiflCAll0N SYSILM OP-204 REV 53 POWER OPERATIONS OP-404 REV 78 DECAY HEAT REMOVAL SYSl[M
^ OP-603 REV 37 CONDENSATE SYSTEM OP-502 REV 27 CONTROL ROD DRIVL SYSTEM
(, OP-302 REV 32 RC PUMP OPERATION SP-349A REV 13 EfP-1 OPERABILITY SURVEILLAf1CE OP-608 REV 36 OTSGs AND MAIN SP-349B REV 16 EfP-2 AND VALVE OP[RABILITY STEAM SYSTEM SURVEILLANCE SP-390 REV 14 STARTUP SURVEILLAf4CE LOG BASELINE DATA: Plant Operating and Surveillance procedures Plant performance data collected during power escalation 10/7/88 I (_ - . - _ - - - . _ . r
PERf0RMANCE TEST NUMBER: plN1 DATE: 01/08/91 l _ TITLE: PLANT HEATUP AND STARIUP FROM MODI 5 10 1000 p0WER TEST RESULTS/ COMMENTS: Simulator was maneuvered from -180 degrees on DH removt.1 to 100% full power using , approved plant procedures. Simulator response was excellent throughout the test. Data l was recorded -every 50 degrees during the heatup and -every 5% power during the power i escalation. Recorded parameters compared favorab!y to actual plant data . All : required local operator actions were tested and functioned correctly. All limitations imposed by simulator scope were evaluated and documented. One scope limitation was evaluated as unacceptable, it involves closing control board breaker 3392 (480V bus crosstie feeder). 3reaker 3392 was added as a plant modification under MAR 88-11 01 and is not currently simulated. Simulator r..odification tu incorporate this breaker is in progress and is being tracked under CCN# 900157. C 1 1 l
Test NUMBER /11TLE: P1N1/PLAHiHLAlVPANDSlARIUPFROMMobE 5 10 100% POWIR TEST DATA (Table 1) LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Paranoter DEVICC CilART SPEED DH REMOVAL OPERATION Hl:ll3 Pzr Level CTS HlJPR203 Pzr Temperature CIS L4JPR222 Loop A Pressure (NARROW) CTS L6JPR208 Loop A Pressure (WlDE) CIS PGJPE202 Unit 3 Generation CIS LlJPR212 Loop A Th (NARROW) CIS LlJPR220 Loop A Tc (WlDE) CIS LlJPR220 loop B ic (WIDE) C15 LlJPS287 SG-A S/V Level CTS LlJPS293 SG B S/U Level CTS LlJPS284 SG A OP Level C1s LlJPS336 SG-B OP Level CIS F3JPA320 "A" MS Pressure CTS F3JPA323 "B" HS Pressure CIS LlJP1228 "A" Header Pressure ClS L1JP1232 "B" Header Pressure CTS L1JPS329 S/V fW flow "A" Cls O LlJPS330 S/U fW flow "B" N2JPX202 Letdown flow CTS CTS N2JSA123 Makeup flow CIS N2JPX359 Makeup lank level CIS JPPAR230 RCS Total flow CIS R3JPP212 hi-3 1R Level CTS R301 NIM NI-l Counts CTS ~
LlJPS285 SG A full Range Level CIS L1JPS286 SG-B full Range Level CTS SRT01 Al Steam Line Temperature CTS SRT07 Bl Steam Line Temperature C15 L1014Afl RCS Loop "A" flow ClS L1014Bil RCS Loop "B" flow C15 ll0BBCORE RCS Boron CIS RRRXT Core Xenon Reactivity CIS N3:DIB DH-2-111 DH A Inject Temp CIS N3:Dif Dil-6-Til Dil Clr inlet Temp CIS N3:01D DH-2-112 DH B Inject Temp CTS N3:DlH Dil-6-T!2 Dil Clr Inlet Temp Cls N3:DI) DH-1 fil "A" DH flow C15 N3:013 DH-1-fl2 "B" DH flow CTS
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-- _ a TEST NUMBER / TITLE: P1N1/ PLANT HEA1UP AND STARTUP FROM MODE __5 10 100% POWER TEST DATA (Table 2)
LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED RCP STARTUP TO 532 f Hl ll3 Pzr level Grfx Rcdr 0 320 in 5.625 sec HlJPR203 Pzr Temperature Grfx Rcdr 0 700 f L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG L6JPR208 Loop A Pressure (WIDE) Grfx Rcdr 0 - 2500 PSIG PGJPE202 Unit 3 Generation Grfx Rcdr 0 - 1000 MWE LlJPR212 Loop A Th (NARROW) Grfx Rcdr 520 620 F LlJPR220 Loop A Tc (WIDE) Grfx Rcdr 50 - 650 f L1JPR220 Loop B Tc (WlDE) Grfx Rcdr 50 - 650 f LlJPS287 SG A S/U Level Grfa Rcdr 0 - 250 in L1JPS293 SG B S/V Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A OP Level Grfx Rcdr 0 - 100 %
LlJPS336 SG B OP Level Grfx Rcdr 0 100 % F3JPA320 "A" MS Pressure Grfx Rcdr 0 1200 PSIG f3JPA323 "B" MS Pressure Grfx Rcdr 0 - 1200 PSIG LlJPT228 "A" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPT232 "B" Heaoer Pressure Grfx Rcdr 600 1200 PSIG LlJPS329 S/V fW flow "A" Grfx Rcdr 0 - 1500 MLB/HR G LlJPS330 S/V FW flow "B" N2JPX202 Letdown flow Grfx Rcdr Grfx Rcdr 0 - 1500 MLB/HR 0 - 150 GPM N2JSAl23 Makeup flow Grfx Rcdr 0 - 150 GPM N2JPX359 Makeup lank level Grfx Rcdr 0 - 100 in JPPAR230 RCS Total flow Grfx Rcdr 0 - 160 MLB/HR R3JPP212 N! 3 1R Level Grfx Rcdr -11.0 - -3,0 log Amps R301 NIM N1-1 Counts Grfx Rcdr -1,0 6,0 log Cps llJPS285 SG-A full Range Level Grfx Rcdr 0 - 600 in LlJPS286 SG B full Range level Grfx Rcdr 0 600 in SRT01 Al Steam Line Temperature Grfx Rcdr 0 - 600 f SRT07 Bl Steam Line Temperature Grfx Rcdr 0 - 600 f L1014Afl RCS Loop "A" flow Grfx Rcdr 0 - 80 MLB/HR L1014Bfl RCS Loop "B" flow Grfx Rcdr 0 - 80 MLB/HR HQBBCORE RCS Boron Grfx Rcdr 500 - 1500 PPM RRRXT Core Xenon Reactivity Grfx Rcdr 0,0 - -0.06 dK/K
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TESTNUMBER/ TITLE: PTN1/PL AN1 lif A10P Af4D STARIUP F ROM MODI 510100% POWt R TES1 DATA (Table 3) LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CllART SPEED 532 F TO 100% POWER 111:113 Ptr level Grfx Rcdr 0 - 320 in 5.625 sec filJPR203 Pzr Temperature Grfx Rcdr 0 - 700 i L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG L6JPR208 Loop A Pressure (WIDE) .
Grfx Rcdr 0 - 2500 PSIG RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125 % PGJPE202 Unit 3 Generation Grfx Rcdr 0 - 1000 MWE LlJPR212 Loop A Th (NARROW) Grfx Rcdr 520 - 620 f IIC1 AVG RCS Average Temperature Grfx Rcdr 520 - 620 I LlJPR214 Loop A 1c (NARROW) Grfx Rcdr 520 - 620 f LlJPR215 Loop A Tc (NARROW) Grfx Rcdr 520 - 020 f LlJPR216 Loop B Tc (NARROW) Grfx Rcdr 520 620 f LlJPR217 Loop B Tc (NARROW) Grfx Rcdr 520 - 620 f LlJPS287 SG A S/V Level Grfx Rcdr 0 - 250 in LlJPS293 SG-8 S/V level Grfx Rcdr 0 - 250 in LlJPS284 SG-A OP Level Grfx Rcdr 0 100 % LlJPS336 SG-B OP Level Grfx Rcdr 0 - 100 % F3JPA320 "A" MS Pressure Grfx Rcdr 0 1200 PSIG f3JPA323 "B" MS Pressure Crfx Rcdr 0 - 1200 PSIG LlJP1228 "A" lleader Pressure Grfx Rcdr 600 1200 PSIG LlJP1232 "B" lleader Pressure Grfx Rcdr 600 - 1200 PSIG LlJPS301 Main iW flow "A" Grfx Rcdr 0 - 6000 KlB/llR LlJPS302 Main fW flow "B" Grfx Rcdr 0 6000 KiB/liR LlJPS329 S/U FW flow "A" Grfx Rcdr 0 - 1500 MlB/ lip LlJPS330 S/U f W flow "B" Grfx Rcdr 0 - 1500 MLB/IIR N2JPX202 Letdown flow Grfx Rcdr 0 - 150 GPM N2JSAl23 Makeup flow Grfx Rcdr 0 - 150 GPM N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in JPPAR230 RCS lotal flow Grfx Rcdr 0 - 160 MLB/llR R3JPP212 NI-3 1R level Grfx Rcdr -11.0 - -3.0 Log Amps R3JPP214 NI-5 Imbalance Grfx Rcdr -40.0 - 40.0 %
- R3JSA000 NI-5 Power Level Grfx Rcdr 0 100 %
R30lNIH NI-l Counts Grfx Rcdr 1.0 - 6.0 log Cps LlJPS285 SG-A full Range Level Grfx Rcdr 0 - 601 in LlJPS286 SG B full Range Level Grfx Rcdr 0 - 600 in SR101 Al Steam tine Temperature Grfx Rcdr 0 - 600 f SR107 B1 Steam Line Temperature Grfx Rcdr 0 - 600 f L1014Afl RCS Loop "A" flow Grfx Rcdr 0 - 80 MLB/llR L1014Bfl RCS Loop "B" flow Grfx Rcdr 0 80 ML B/llR ltQBBCORE RCS Baron Grfx Rcdr 500 - 1500 PPM RRRXT Core Xenon Reactivity Cr x Rcdr 0.0 - 0,06 dK/K _ ,. --n . . - = _ - - . . , = _ = -
G (4,)PERFORMANCETESTNUMBER: PTN1 DATE: 01/ca/91 TITLE: PLANT HEATUP AND STARTUP FROM MODE 5 T0 100% POWER TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none soindicate): TR # Priority Date Issued Date Closed NONE
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WCTest oordinator ate 2.0 All deficiencies have been resolved and this test has been completed. (~
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NOT REQUIRED TPC lest Coordinator /Date 3.0 Test results reviewed. Simulator performance is s isfactory, t ,k Pf)(- b
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{ TEST NUMBER / TITLE: PTN2/ REACTOR TRIP AND REC 0VERY 10 RATED POWER ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.l(4), Section 3.1.2(19) DATE PERFORMED: 01/29/91 TEST DURATION: 9 hours TEST DESCRIPTION: The simulator is operating at 100% power with all RCP's in operation when a spurious reactor trip occurs. The plant is stabilized in the normal post trip window using AP-580. A Reactor / Plant startup is performed and the plant is brought back to a full power conditions in accordance with applicable plant procedures. tiA_LLUNCT10N URE OPTIONS INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power conditions, all plant equipment in service with a xenon transient in progress. PLANT PROCEDURES USED DURING THE TEST: OP-203 REV 61 PLANT STARTUP OP-212 REV 2 OPERATIONS INVOLVING THE MSR OP-605 REV 48 FEEDWATER SYSTEM SP-422 REV 19 RC SYSTEM H/U AND C/D SURVEILLANCE (m OP-204 REV 53 POWER OPERATIONS OP-502 REV 27 CONTROL R0D DRIVE SYSTEM \'~' AP-580 REV 15 REACTOR TRIP OP-703 REV 34 PLANT DISTRIBUTION SYSTEM OP-210 REV 29 REACTOR STARTUP SP-390 REV 14 STARTUP SURV TLLANCE LOG OP-603 REV 37 CONDENSATE SYSTEM SP-312A REV 2 DALLY HEAT BALANCE POWER COMPARISON _ BASELINE DATA: Expected simulator response on a reactor trip based ';n the following data: RECALL DATA FILE 0070287 - Reactor Trip from 89% Power During Vital Bus Tran3fer 07/02/87 0 09:53:00 UOER 87-02 - Reactor Trip During Vital Bus Transfer Expected simulator response on a reactor trip recovery based on expert judgement, ability to conduct the test in accordance with plant procedures and known response of the plant during a reactor / turbine startup and power escalation. See test PTN1 for additional baseline information. J d
{ PERFORMANCE TEST NUMBER: Pill 2 DATE: 01/29/91 TITLE: REACTOR TRIP AfiD REC 0VERY TO RATED POWER TEST RESULTS/ COMMENTS: Simulator response on a reactor trip evaluated during performance test PTTI. Test results evaluation on reactor startup and recovery to rated power based on expert judgement, ability to conduct the test in accordance with plant procedures and known response of the plant during a reactor / turbine startup and power escalation. See test PINI for additional evaluation information. Test results satisfactory. No performance problems noted. One (1) Trouble Report written on Plant Process Computer Post Trip Review. TR does not require a rerun of the test. TR-221 cleared 08/27/91 c f% J
{ ST NUMBER / TITLE: PTN2/ REACTOR TRIP AND RECOVERY TO RATED POWER TEST DATA TABLE 1 LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5.625 sec *HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR200 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125 % *PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 10F *L1JPR212 Th loop-A Grfx Rcdr 52: 620F *HCTAVG RCS Average Tempera +ure Grfx Rcdr 52; 520F *LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 620F *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in *LlJPS284 SG-A Op level Grfx Rcdr 0 - 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi
/^ LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
, *LlJPS301 SG-A MFW Flow Grfx Rcdr 0 6K KLB/HR *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR N2JPX202 Letdown Flow Grfx Rcdr 0 - 150 GPM N2JSAl23 Makeup Flow Grfx Rcdr 0 150 GPM N2JPX359 Makeup Tank Level Grfx Rcdr 0- 100 in L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR R3JPP212 Intermediate Range Indication Grfx Rcdr -11.0 - -3.0 Amps R3JPP214 Power Range Imbalance Grfx Rcdr 40.0 - +40.0 %
R3JSA000 Power Range Indication Grfx Rcdr 0 - 100 % R30lNIM Source Range Indication Grfx Rcdr -1.0 - 6.0 log cps LlJPS285 SG-A Full Range Level Grfy Rcdr 0 600 in LlJPS286 SG-B Full Range Level Grfx Rcar 0 - 600 in SRT01 Main Steam Temperature Grfx Rcdr 0 - 600F SRT07 Main Steam Temperature Grfx Rcdr 0 - 600F L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr C - 80 MLB/HR HQBBCORE RCS Boron Concentration Crfx Rcdr f00 - 1500 PPM RRRXT Core Xenon Reactivity Grfx Rcdr C - -0.6 dK/K
5 i p PERFORMANCE TEST NUMBER: PTN2 DATE: 01/29/91 TITLE: REACTOR TRIP AND REC 0VERY TO RATED POWER TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed
,TH-221 C 07/31/91 08/27/91 E l'
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l 'lf FP est Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. f-O v -
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3.0 Test results reviewed. Simulator performance /is sat,isfactory.
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/Date APPROVED: 2 [
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( t TEST NUMBER / TITLE: PIN 3/ LOAD CliANGLS
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ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.l(6) DATE PERFORMED: 08/14/91 TEST DURATION: 6 1/2 hours TEST DESCRIPTION: With the simulator at 100% power, a load reduction to 700 MWe at a rate of 1%/ minute is performed with the ICS in full automatic. Onra the simulator is stabilized at 700 MWe, an RCS boratic; is performed to restore the control rods to a fully withdrawn (~95%) position. With the rods restored to a fully withdrawn position, an RCS dilution is performed to restore the siuulator to full power conditions. MALFUNCTION TITLE OPTIONS INITIAL CONDITIONS: 'ull power steady state conditions. all plant equipment in service. 4 FINAL CONDITIONS: Full power conditions with a xenon transient (burnout) in progress. PLANT PROCEbuRES USED DURING THE TEST:
,-s OP-201 REV 54 POWER OPERATIONS
( ,, 0?-304 REV 6 SOLUBLE POISON CONCENTRATION CONTROL BASELINE DATA: Expected simulator response based on expert judgement, known plant response to load changes using both control rods and boration/ dilution, ability to perform the test in accordance with the referenced procedures and plant data collected during - load changes on 10/07 and 10/08/88. I v
{ fPERFORMALCETESTNUMBER: Plf13 DATE: 08/14/91 TITLE: LOAD CilANGES TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement, known plant response to load changes using both control rods and boration/ dilution, ability to perform the test in accordance with the referenced procedures and piant data collected during load changes on 10/07 and 10/08/88. Test results satisfactory, fio performance problems noted. l 1 l l 1
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EST NUMBER / TITLE: PTN3/ LOAD CHANGES {]) TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / c' Critical Parameter DEVICE CHART SPEED Hl:ll3 Pzr level Grfx Rcdr 0 - 320 in 5.625 sec HlJPRE03 Pzr lemperature Grfx Rcdr 0 - 700 F L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG LEJPR208 Loop A Pressure (WIDE) Grfx Rcdr 0 - 2500 PSIG RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125 % PGJPE202 Unit 3 Generation Grfx Rcdr 0 - 1000 MWE L1:6ADT Delta-T Loop A Grfx Rcdr 0 - 50 F Ll:6BDT Delta-T Loop B Grfx Rcdr 0 - 50 F LlJPR212 Loop A Th (NARROW) Grfx Rcdr 520 - 620 F HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620 F LlJPR214 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR215 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPS287 SG-A S/U Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B S/U Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A 09 level Grfx Rcdr 0 - 100 % LlJPS336 SG-B OP Level Grfx Rcdr 0 - 100 % F3JFA320 "A" MS Pressure Grfx Rcdr 0 - 1200 PSIG F3JPA323 "B" MS Pressure Grfx Rcdr 0 - 1200 PSIG L1JPT228 "A" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPT232 "B" Header Pressure Grfx Rcdr 600 - 1200 PSIG C- LJJPS301 Main FW Flow "A" Grfx Rcdr 0 - 6000 KLB/HR LlJPS302 Main FW Flcw "B" Grfx Rcdr 0 - 6000 KLB/HR LlJPS329 S/U FW Flow "A" Grfx Rcdr 0 - 1.5 MLB/HR LlJPS330 S/U FW Flow "B" Grfx Rcdr 0 - 1.5 MLB/HR N2JPX202 Letdown Flow Grfx Rcdr 0 - 150 GPM N2JSAl23 Makeup Flow Grfx Rcdr 0 - 150 GPM N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in BMBMUTll Makeup Tank Boron Grfx Rcdr 0 - 2000 ppm L5G7AV Group 7 Position Grfx Rcdr 0 - 100% L508AV Group 3 Position Grfx Rcdr 0 - 100% R3JPP214 NI-5 Imbalance Grfx Rcdr -40.0 - 40.0 % R3JSA000 NI-5 Power level Grfx Rcdr 0 - 100 % HQBBCORE RCS Baron Grfx Rcdr 500 - 1500 PPM RAREAC Core Reactivity Grfx Rcdr . 01 .01 dK/K RRRX(22) Xenon Reactivity (Bottom) Grfx Rcdr .02 - .035 dK/K RRRX(24) Xenon Reactivity (Top) Grfx Rcdr 02 - .035 dK/K
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O U PERFORMANCE TEST NUMBER: PIN 3 DATE: 08/15/91 TITLE: LOAD CHANGES
-TEST STATUS:
1,0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _h0E
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[Date a 2.0 All deficiencies have been resolved and this test has been completed, O i
'd NOT REQUIRED FPC Test Coordinator /Date 30 Test results reviewed. Simulator performance.is satisfactory.
( / - h @blyh/ \ i LLLh Abb" Y s'Gh,/ _ Simulator nstructor /Date Operattons Representative /Date s
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I { f TEST NUMBER / TITLE: PTN4/3 RCP FLANT SHUTDOWN AND STARTUP AHSI/ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.l(7) DATE PERFORMED: 01/26/91 TEST DURATION: 5 hours TEST DESCRIPTION: The simulator is initialized at 75% power with the "D" RCP shutdown. A plant shutdown is initiated and reactor power is reduced to 12% with the main turbine of f line. Power is maintained at ~12% while the turbine coasts down to ~100 RPM. When main turbine is at -100 RPM a plant startup and power escalation to 75% power is performed. MALFUNCTION IJTLE OPTIONS INITIAL CONDITIONS: 75% power with the "D" RCP shutdown. FINAL CONDITIONS: 75% power with the "0" RCP shutdown. PLANT PROCEDURES USED DURING THE TEST: OP-203 REV 61 PLANT STARTUP OP-605 REV 48 FEEDWATER SYSTEM OP-204 REV 53 POWER OPERATIONS OP-212 REV 2 OPERATIONS INVOLVING THE MSR OP-603 REV 37 CONDENSATE SYSTEM SP-422 REV 19 RC SYSTEM H/V AND C/D SURVEILLANCE l OP-208 REV 36 PLANT SHUTDOWN OP-703 REV 34 PLANT DISTRIBUTION SYSTEM BASELINE DATA: Expected simulator response based on expert judgement extrapolation of normal plant power reduction and power escalation Md the ability to conduct the test in accordance with plant procedures. See performance tests PTN1 and PTN5 for additional baseline information. O J
PERFORMANCE TEST NUMBER: PTN4 DATE: 01/26/91 TITLE: 3 RCP PLANT SHUTDOWN AND STARlUP TEST RESULTS/ COMMENTS: Expected simulator response based on expert judgement extrapolation of normal plant power reduction and power escalation and the ability to conduct the test in accordance with plant procedures. See performance tests PlNI and PINS for additional baseline information. Test results satisfactory. No performance problems noted,
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TEST NUMBER / TITLE: PIN 4/3 RCP PLANT SHUTDOWN AND STARTUP TEST DATA TABLE 1 LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED Hl:ll3 Pzr Level Grfx Rcdr 0 - 320 in 5.625 sec HlJPR203 Pzr Temperature Grfx Rcdr 0 - 700 F L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG L6JPR208 Loop A Pressure (WIDE) Grfx Rcdr 0 - 2500 PSIG RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125 %
PGJPE202 Unit 3 Generation Grfx Rcdr 0 - 1000 MWE LlJPR212 Loop A Th (HARROW) Grfx Rcdr 520 - 620 F HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620 F LlJPR214 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR215 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LllPR216 Loop B Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR217 Loop B Tc (NARROW) Grfx Rcdr 520 - 620 F LldPS287 SG- A S/U Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B S/U Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A OP Level Grfx Rcdr 0 - 100 % L1JPS336 SG-B OP Level Grfx Rcdr 0 - 100 % F3JPA320 "A" MS Pressure Grfx Rcdr 0 - 1200 PSIG F3JPA323 "B" MS Pressure Grfx Rcdr 0 - 1200 PSIG LlJPT228 "A" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPT232 "B" Header Pressure Grfx Rcdr 600 - 1200 PSIG J LlJPS301 Main FW Flow "A" Grfx Rcdr 0 - 6000 KLB/HR LlJPS302 Main FW Flow "B" Grfx Rcdr 0 - 6000 KLB/HR LlJPS329 S/V FW Flow "A" Grfx Rcdr 0 - 1500 MLB/HR L1JPS330 S/V FW Flow "B" Grfx Rcdr 0 - 1500 MLB/HR N2JPX202 Letdown Flow Grfx Rcdr 0 - 150 GPM N2JSAl23 Makeup Flow Grfx Rcdr 0 - 150 GPM N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in JPPAR230 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR R3JPP212 N1-3 IR Level Grfx Rcdr -11.0 - -3.0 Log Amps R3JPP214 NI-5 Imbalance Grfx Rcdr -40,0 - 40.0 % R3JSA000 NI-5 Power Level Grfx Rcdr 0 - 100 % R30lNIM NI-l Counts Grfx Rcdr -1.0 - 6.0 Log Cps llJPS285 SG-A Full Range Level Grfx Rcdr 0 600 in LlJPS286 SG-B Full Range Level Grfx Rcdr 0 - 600 in SRT01 Al Steam Line Temperature Grfx Rcdr 0 - 600 F SRT07 B1 Steam Line Temperature Grfx Rcdr 0 - 600 F L1014AFI RCS Loop "A" Flow Grfx Rcdr 0 - 80 MLB/HR L1014BFI RCS Loop "B" Flow Grfx Rcdr 0 - 80 MLB/HR HQBBCORE RCS Boron Grfx Rcdr 500 - 1500 PPM RRRXT Core Xenon Peactivity Grfx Rcdr 0.0 - -0.06 dK/K O
PERFORMANCE TEST NUMBER: PTN4 DATE: 01/26/91 TITLE: 3 RCP PLANT SHU100WN AND STARTUP TEST STATUS: 1,0 This test is complete and has the following deficiencies noted (if none I so indicate): 1 TR # Priority Date issued Di.te Closed l NONI: _.
- 1
$h}kse'
/ ,,
L ) /f?{ j/ , [FiEFCoordinator FP /Date 2.0 All deficiencies have been resolved and this test has been completed, p V' _ NOT REQUlitED FPC Test Coordinator /Date i 3.0 Test results reviewed. Simulator performance is satisfactory,
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; ~g / 22 0 y_ . '[] r 'J [jj{~_ jl- ..g;,/y, imuTator Instructor /Date OperationsRepresentative/Da}e '
A . / / l' \ j 1 t. DdN Simulat'oYlnijingM APPROVED: /
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s,, f TEST NUMBER / TITLE: PTN5/ PLANT SHUTDOWN AND C00LDOWN TO MODE 5 b,m ANSI /ANS-3.5(1985) CROSS
REFERENCE:
DATE PERFORMED: 1/4/91 - 1/6/91 TEST DURATION: 16 hours TEST DESCRIPTION: The simulator is operating at 100% power, MOL, equilibrium xenon. The simulator will be maneuvered from 100% full power to Mode 5 (<200 degrees) using approved plant operating procedures. All plant procedure steps will be followed and signed off by the simulator operator. Administrative requirements and other non-applicable stcps will be signed off as Not Required "NR". MALFUNCTION llJ11 QPTIONS INITIAL CONDITIONS: Full pow, cuady state conditions, all plant equipment in service. FINAL CONDITIONS: Mode 5 on "B" DH Removal System with all RCP's shutdown. PLANT PROCEDURES USED DURING THE TEST: OP-204 REV 53 POWER OPERATIONS OP-209 REV 71 PLANT C00LDOWN n OP-211 REV 9 REACTOR SHUTDOWN OP-402 REV 66 MAKEUP AND PURIFICATION SYSTEM V OP-605 REV 48 FEEDWATER SYSTEM OP-603 REV 37 CONDENSATE SYSTEM OP-212 REV 3 OPERATIONS INVOLVING THE-MSR SP-422 REV 19 RC SYSTEM H/V AND C/D SURVEILLANCE OP-208 REV 36 PLANT SHUTDOWN OP-703 REV 34 PLANT DISTRIBUTION SYSTEM OP-404 REV 78 DHR SYSTEM OP-502 REV 27 CONTROL R0D DRIVE SYSTEM BASELINE DATA: Plant Operating and Surveillance procedures Plant performance data collecteu during power escalation 10/7/88 O
N i I
, PERFORMANCE TEST NUMBER: PTN5 DATE: 01/06/91 /
k TITLE: PLANT SHUTOOWN AND C00LDOWN TO MODE 5. _ TEST RESULTS/ COMMENTS: Simulator was maneuvered from 100% power to -180 degrees on DH removal using approved plant procedures. Simulator response was excellent throughout the test and no performance anomalies were noted. Data was recorded -every 5% power during the power reduction, and -every 50 degrees during the plant cooldown. Recorded parameters compared favorably.to actual plant data. The transition from OTSG cooling to DH removal went smoothly. All local operator actions (Remote functions) required to maneuver the plant from Mode 5 to full power were tested and functioned correctly. All procedure limitation imposed by simulator scope were evaluated and documented. Two scope limitations were evaluated as unacceptable. The first involved removing power from the power operated valves associated with the EfW System (0P-209, Step 4.2.4). A TR was generated and a simulator modification is in progress under CCN 910162 and CCN 910270. The second involves breaker manipulations associated with the newly installed Emergency Offsite Transformer (0P-703, Step 4.3.11). This transformer was installed under MAR 89-08-11-03, a simulator modification is in progress and being tracked under CCN 900148. l i t O l l
N. , r 7s TEST HUMBER/ TITLE: PTN5/ PLANT SHUTDOWN AND C00LOOWN TO MODE 5 TEST DATA (Table 1) LABEL DESCRIPTION RECORDIM RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED 100% POWFR TO 532 F Hl:ll3 Pzr Level Grfy Rcdr 0 - 320 in HlJPR203 Pzr Temperature Grfx Rcdr 0 - 700 F L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG L6JPR208 Loy A Pressure (WIDE) Grfx Rcdr 0 - 2500 PSIG RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125 %
PGJPE202 Unit 3 Generation Grfx Rcdr 0 - 1000 MWE LlJPR212 Loop A Th (NARROW) Grfx Rcdr 520 - 620 F HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620 F L1JPR214 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR215 Loop A Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR216 Loop B Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPR217 Loop B Tc (NARROW) Grfx Rcdr 520 - 620 F LlJPS287 SG-A S/U Level Grfx Rcdr 0 - 250 in LIJPS293 SG-B S/U Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A OP Levcl Grfx Rcdr 0 - 100 % L1JPS336 SG-B OP Level Grfx Rcdr 0 - 100 % F3JPA320 "A" HS Pressure Grfx Rcdr 0 - 1200 PSIG n F3JPA323 "B" HS Pressure Grfx Rcdr 0 - 1200 PSIG
!"/
L1JPT228 "A" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPT232 "B" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPS301 Main FW Flow "A" Grfx Rcdr 0 - 6000 KLB/HR LlJPS302 Main FW Flow "B" Grfx Rcdr 0 - 6000 KLB/HR LlJPS329 S/U FW Flow "A" Grfx Rcdr 0 - 1500 KLB/HR LlJPS330 S/U FW Flow "B" Grfx Rcdr 0 - 1500 KLB/HR t N2JPX202 1.etdown Flow Grfx Rcdr 0 - 150 GPM N2JSA123 Makeup Flow Grfx Rcdr 0 - 150 GPM N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in JPPAR230 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR R3JPP212 NI-3 IR Level Grfx Rcdr -11.0 - -3.0 Log Amps R3JPP214 NI-5 Imbalance Grfx Rcdr -40.0 - 40.0 % , R3JSA000 NI-5 Power Level Grfx Rcdr 0 - 100 % l R30lNIH NI-l Counts Grfx Rcdr -1.0 - 6.0 Log Cps L1JPS285 SG-A Full Range Level Grfx Rcdr 0 - 600 in i L1JPS286 SG-B Full Range Level Grfx Rcdr 0 - 600 in l SRT01 Al Steam Line Temperature Grfx Rcdr 0 - 600 F l SR107 B1 Steam Line Temperature Grfx Rcdr 0 - 600 F 1- L1014AFI RCS Loop "A" Flow Grfx Rcdr 0 - 80 MLB/HR L1014BFI RCS Loop "B" Flow Grfx Rcdr 0 - 80 MLB/HR HQBBCORE RCS Soron Grfx Rcdr 500 - 1500 PPM l RRRXT Core Xenon Reactivity Grfx Rcdr 0.0 - -0.6 dK/K l __ . ($ V
]
f TEST NUMBER / TITLE: PTN5 /. PLANT SHUTDOWN AND-C00LDOWN TO MODE 5 TEST DATA (Table 2) LA8EL DESCRIPTION RECORDING. RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED 532 F TO DH OPERATION Hl:113 Pzr Level __ Grfx Rcdr 0 - 320 in HlJPR203 Pzr Temperature - _Grfx Rcdr 0 - 700 F L4JPR222 Loop A Pressure (NARROW) Grfx Rcdr 1700 - 2500 PSIG .
L6JPR208LoopAPressure(WIDE) Grfx Rcdr 0 - 2500 PSIG PGJPE202 Unit 3 Generation - Grfx Rcdr 0 - 1000 MWE ! LIJPR212 Loop A-Th (NARROW) Grfx Rcdr 520 - 620 F LIJPR220. Loop:A Tc (WIDE) Grfx Rcdr 50 - 650 F .
-LlJPR220 Loop B Tc (WIDE) Grfx Rcdr 50 - 650 F LlJPS287 SG-A S/U Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B S/U Level Grfx Rcdr 0 - 250 in
.L1JPS284~SG A OP Level Grfx Rcdr 0 - 100 % ~
L1JPS336 SG-B OP Level Grfx Rcdr 0 - 100 % F3JPA320 "A" MS Pressure Grfx Rcdr 0 - 1200 PSIG F3JPA323 "B" MS Pressure Grfx Rcdr 0 1200 PSIG LlJPT228 "A" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPT232 "B" Header Pressure Grfx Rcdr 600 - 1200 PSIG LlJPS329_ S/U FW Flow "A" _ Grfx Rcdr 0 - 1500 KLB/HR LL1JPS330 S/U.FW Flow "B" Grfx Rcdr ,0 - 1500 KLB/HR O( N2JPX202' Letdown Flow N2JSA123 Makeup Flow Grfx Rcdr Grfx Rcdr 0 - 150 GPM 0 - 150 GPM. N2JPX359. Makeup Tank Level Grfx Rcdr 0 - 100 in
- JPPAR230 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR R3JPP212 NI-3 IR Level Grfx Rcdr -11.0 - -3.0 Log Amps R30lNIM NI-l Counts Grfx Rcdr -1.0 - 6.0 Log Cps
, LIJPS285 SG-A Full Range Level Grfx Rcdr 0 - 600.in LIJPS286 SG-B-Full Range Level . Grfx Rcdr. 0 - 600 in SRT01 -Al-Steam Line Temperature Grfx Rcdr 0 - 600 F SRT07 B1 Steam Line Temperature- Grfx_Rcdr 0 - 600.F L1014AFI RCS Loop "A" Flow Grfx Rcdr 0 - 80 MLB/HR L1014BFI RCS Loop "B" Flew Grfx Rcdr 0 - 80 MLB/HR HQBBCORE RCS Boron Grfx Rcdr 500 - 1500 PPM
-RRRXT Core Xenon Reactivity Crfx Rcdr 0.0 - -0.6 dK/K l
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'I '
TEST' NUMBER / TITLE:-PTN5/ PLANT SHUTDOWN AND C00LD0WN TO MODE 5 TEST DATA-(Table 3) > LAEEL DESCRIPTION- RECORDING RANGE FREQUENCY /-
- Critical Parameter- DEVICE CHART SPEED DH REMOVAL OPERATION H1:lI3 '.Pze level CTS HlJPR203 Pzr Temperature CTS L4JPR22? Loop-A Pressure-(NARROW) CTS L6JPR208 Loop A Pressure (WIDE) CTS PGJPE202 Unit 3 Generation CTS
[LlJPR212LoopATh(NARROW) CTS LlJPR220 Loop A lc-(WIDE) CTS LlJPR220 Loop _B-Tc-(WIDE) CTS LlJPS287 SG-A-S/U Level CTS
-LlJPS293 SG-B S/U Level CTS LlJPS284 SG-A OP Level CTS LlJPS336 SG-B-0P Level CTS
-F3JPA320="A" MS Pressure CTS F3JPA323."B" MS Pressure- CTS L1JPT228 "A" Header Pressure CTS LlJPT232 "B" Header Pressure CTS LlJPS320 S/U FW Flow "A"- CTS LlJPS330 S/U FW Flow "B" CTS
' 's N2JPX202' Letdown Flow CTS N2JSA123 Makeup Flow CTS N2dPX359 Makeup-_ Tank Level CTS
- JPPAR230.RCS Total Flow CTS R3JPP212 NI-3 IR Level CTS LR30lNIM NI-1 Counts CTS LlJPS285 SG-A Full-Range Level CTS LlJPS286 SG-B Full Range Level CTS SRT01: Al Steam Line Temperature CTS SRT07-- Bl: Steam Line Temperature _ CTS-L1014AFI RCS Loop "A" Flow CTS-L1014BFI RCS_ Loop _"B" Flow CTS HQBBCORE_RCS Boron CTS RRRXT. CTS' N3:DIB-. Core-Xenon DH-2-TIl DH Reactivity A Injec Temp -CTS N3:DIF- DH-6-Til DH Clr Inl Temp CTS N3:DID DH-2-TI2 DH B Inject Temp CTS N3:DIH DH-6-TI2 DH Clr Inlet' Temp- CTS
-N3:DIl- DH-1-FIl "A". DH Flow- CTS-N3:013 _ DH-1-FI2 "B" .DH Flow CTS l
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C \_/ PERFORMANCE TEST NUMBER: PTN5 DATE: 01/06/91 TITI.E: PLANT SHUTDOWN AND C00LDOWN TO MOCE 5 TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # .riority Date Issued Date Closed TR-121 C 04/23/91 07/16/91 TR-217 u 07/10/91_ i Co~ j% +149 I Q~ T l est Coordipatbr 70 te 2.0 All deficiencies have been resolved and this test has been completed. ,A *Closcout of TR-217 doca not prevent b &qM !Q /_ completion of t.his test. F VC iest Coordihator Q'/Date 3.0 Testresultsreviewed.Simulatorperformancyissatisfactory.
, c f brJ-Simulator Instrudtor 4 4 4 18 11
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\ '. O M0 # d. k 1)pera t i on s .. Rep r6 s on t a t i ve/Date l)$[h/
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?lln,;'e 1r SiniulatoIEntfin' eor '/O APPROVED: 2 DATE: (/// j_
, /HSTS/ L TM 4
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TEST NUMBER / TITLE: PTN6/ CORE PERFORMANCE IESTING ANSI /ANS-3.5(1985) CROSS REFr.RENCE: Section 3.1.l(9) DATE PERFORMED: 12/05/89 TEST DURATION: 12 hours TEST DESCRIPTION: Following a refueling outage, a reactor startup and the following zero power physics physics tests are performed: Initial Criticality, Critical Boron Concentration (all l regulating rods out), Isothermal Tempera *.ure Coef ficient, Integral Group Worths, and Differential Boron Worth. MALFUNCTION liTLE OPTIONS INITIAL CONDITIONS: Beginning of life (BOL), RCS temperatures at approximately 532F, RCS pressure at approximately 2155 psig. Pressurizer level is being automatically maintained at 60 inches. Turbine bypass valves are maintaining steam header pressureat 885 psig. OTSG levels are being maintained at minimumlevel by startup FW valves. Safety Group 1 is at its "out-limit". The core is xenon free. FINAL CONDIT10N5: BOL, Mode 2. PLANT PROCEDURES USED DURING THE TEST: PT-110 REV 13 CONTROLLING PROCEDURE FOR ZERO POWER PHYSICS TESTING PT-lll REV 8 HOT ZERO POWER ALL RODS OUT CRITICAL BORON TEST PT-ll2 REV 8 HOT ZERO POWER REGULATING GROUP WORTH AND DIFFERENTIAL BORON WORTH MEASUREMENT l PT-ll4 REV 13 MODERATOR AND TEMPERATURE COEFFICIENTS DETERMINATION AT HOT ZERO POWER OP-502 REV 22 CONTROL R0D DRIVE SYSTEM BASELINE DATA: FPC Cycle 7 Physics Test Manual l l l l l
PERFORMANCE TEST NUMBER: PTN6 DATE: 32fo3fg9 TITLE: CORE PERFORMANCE TESTING TEST RESULTS/ COMMENTS: Test results satisfactory. The calculated value of the hot full power moderator coefficient was 2.17 pcm/F. Although this value did not meet the zero power physics test acceptance criteria, it was determined through an evaluation of the power coefficient and the response of axial flux imbalance that the positive moderator coefficient would not noticeably effect core performance. All other test results met their physics test acceptance criteria. 1 A C ) [N l l
TEST NUMBER / TITLE: PTN6/ CORE PERf0RMANCE TESTING TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / CCritical Parameter DEVICE CHART SPEED RAREAC Core Reactivity 8 pen Rcdr .0005 .0005 dk/k 100 mm/sec HCTAVG RCS Tave 8 pen Rcdr 527 - 537 F RRBCONCP Core Boron Concentration 8 pen Rcdr 1600 - 2200 ppmb L5G7AV Grp 7 Avg Position 8 pen Rcdr 0 - 100 % L5G6AV Grp 6 Avg Position 8 pen Rcdr 0 - 100 % L5G5AV Grp 5 Avg Position 8 pen Rcdr 0 - 100 % L5G8AV Grp 8 Avg Position 8 pen Rcdr 0 - 100 % O l
)
g PERFORMANCE TEST NUMBER: PIN 6 DATE: 12/05/89 TITLE: CORE PERFORMANCE TESTING TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed
-@!NL .. -.- -- -
s yu. - _ m ay est Loordinator Date L-2.0 All deficiencies have been resolved and this test has been completed. p
.. .bb b @
fPC Test Coordinator _ _ . . - _ _/Date 3.0 Test results reviewed. Simulator performance is satisfactory, h v., k G u- . 3 .- 91i jM NOT R[OVlRLD
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$ inst rdctor /Date Operations Representative /Date
- /
l / j l TILb} l?~kt~] lt/R
,41'ulator Engineer /Date
( APPROVEU [ _ _ . _ _ DATE: _3l f_ O-
O TEST NUMBER / TITLE: PIN 7/0PERATOR CONDUCTED SURVEILLANCE TESTING ANSI /ANS-3.5(19CS) CROSS
REFERENCE:
Section 3.1.l(10) DATE PERFORMED: N/A TEST DURATION: N/A TEST DESCRIPTION: Surveillance procedure testing is accomplished by obtaining an approved copy of a plant surveillance procedure and performing those parts normally conducted by licensed operators from the main control board. This test procedure lists the operator conducted surveillance procedures that shall be tested, specifies the sections of the surveillance procedures are applicable, and indicates under what conditions the tests should be run, tim. FUNCTION TITLE OPTIONS INITIAL CONDITIONS: As specified in the individual Surveillance Procedures. FINAL CONDITIONS: As Specified in the individual Surveillance Procedures. PLANT PROCEDURES USED DURING THE TEST: O SP-130 REV 27 SP-300 REV 110 SP-301 REV 79 SP-312A REV 3 SP-317 REV 41 SP-321 REV 34 SP-323 REV 9 SP-325 REV 30 SP-332 REV 33 SP-333 REV 22 SP-343 REV 16 SP-349A REV 13 SP-349B REV 16 SP-354A REV 27 SP-356 REV 8 SP-358A REV 9 SP-358B REV 8 SP-358C REV 7 SP-390 REV 14 SP-442 REV 19 BASELINE DATA: Applicable acceptance criteria within the plant surveillance procedures will provide baseline data. j
g I PERFORMANCE TEST NUMBER: PTN7 DATE: 08/20/91 TITLE: OPERATOR CONDUClED SURVEILLANCE TESTING TEST RESULTS/ COMMENTS: Review individual Surveillance Test Procedures for test results. Actual acceptance criteria applies for plant surveillance procedures run on the simulator unless otherwise noted. NOTE 1: Due to recent changes to the Engineered Safeguards Actuation System logic current copies of the following ES System surveillance procedures cannot be performed successfully on the simulator: SP-130 SP-356 SP-358A SP-358B SP-358C The surveillance Procedures included with this transmittal were performed as part of the factory acceptance test, and are current with respect to the present scope of simulation. Upon completion of simulator modifications current controlled copies of ES System Surveillance Procedures will be used to verify simulator response. Six (6) Trouble Reports were generated during this test. TR-185 addresses a hardware problem with 2 of the RCP seal leakage integrators encountered during the conduct of an RC Systems Inventory Water Balance. TR-224 addresses an annunciator problem encountered during the conduct of the Evacuation and Fire Alarm Test. TR-225 f' addresses a problem with MSIV stroke times encountered during conduct cf MSIV Partial v Stroke Exercising. TR-2229, -230 and -231 address minor problems encountered during the conduct of monthly EDG surveillance testing. None of these TRs require a rerun of their associated surveillance test. TR-225 cleared 08/30/91 TR-229 cleared 08/23/91 TR-230 cleared 08/27/91 TR-231 cleared 08/29/91 C) o
TEST NUMBER / TITLE: PTN7/0PERATOR CONDUCTED SURVEILLANCE TESTING TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE COMPLETED SURVEILLANCE PROCEDURES C
PERFORMANCE TEST NUMBER: PTN7 DATE: 08/20/91 r' V TITLE: OPERATOR CONDUCTED SURVEILLANCE TESTING TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed TR-185 C 07/16/91 e TR-224 C ~657DT/91 TP-225 B 08/01/91 08/30/91_ TR-229 C 08/01/91 08/23/91 TR-230 C 08/01/91 08/27/91 TR-231 C _08/01/91 08/29/91 ( .. NI1/ FPC ___]estWoEd ~ nohi/Date 2.0 All deficiencies have been resolved and this test has been completed. p 5 -
*C30 scout of these THs does not
'L/ <- _. e- s wt [ / prevent cotapletion of this test
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FPYlest toofdihatoy ' /Date - 3.0 Test results reviewed. Simulator performance is 5slTsfactory. j I// I l s .n ' 1lel2
/Date N1lY bW b]?ht/
OperationsRepresentative/phte ' SimgTator In'structo
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A t. y'% l z/ h Simulator Emgip er p /Date APPROVED: _._ g , _ O DATE: ?[/ / S S/NL TM / /' 1 l l
10 TEST NUNBER/ TITLE: PTSl/ STEADY STATE PERFORMANCE TEST
-ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 4.1 DATE PERFORNLD:08/06/91- TEST DURATION: 6 hours TEST DESCRIPTION:- 1) With the simulator running at 100% power, critical parameters are monitored for one hour for stability.
- 2) During a plant load reduction Simulator Main Control Board parameters are compared with actual-plant data at 25%, 75% and 100%.
NALFUNCTION TITLE OPTIONS INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service, FINAL-CONDITIONS: 25% power, main feedwater pump FWP-2B secured. PLANT PROCEDURES USED DURING THE TEST: , OP-204 REVISION 37: OPERATIONS AT POWER SP-312A REVISION 3 DAILY HEAT BALANCE POWER COMPARISION SP-300 REVISION 110. OPERATING DAILY SURVEILLANCE LOG O BASELINE DATA: Plant Computer Data - Plant Process Computer Analo Point Summary at 27% power. u Plant Process Computer Analog Point Summary at 75% power. Plant Process Computer- Analog Point Summary-at 100% power. I O-a
m__ . _ _ . . _ _ _ _ . . _ _ _ .._ _ _ __.___. . _ ._ _ __ _ _ . _ - MI (] *ERFORMANCE TEST NUMBER: PTSI TITLE: STEADY STATE PERFORMANCE IT'iT DATEt 08/06/91 TEST RESilLTS/ COMMENTS: Test results satisfactory. No performance problems noted. The graphic recorder point rangos for the 1007, steady state test data represent a 227, band of plant full power stecdy state values. This allowed both the steady state drift crit 9ria and the 1007, power data comparisons / verifications to be accopmlished by , che<; king that all data remained i.i range for the duration c.' the test. The 257, and 75Y, comparisons / verifications were accomplished through comparisons of plant and simulator data.
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O ~ pERr0RanCc usi uusER, nsi OuE, TITLE, STEADV STATE PERf0RMAtiCE TEST OBJECT!VES: 1.0 Verify compilance with the steady state performance criteria of Af4Sl/AflS-3.5(1985) Section 4.1; Appendix A Section A3.2(1); and Appendix B Section 82.1. TEST ACCEPTANCE CRITERIA: 1.0 SPEC 1flC 1.1 The simulator computed values for steady state, full power operation with the reference plant control system configuration shall be stable and not vary more than A2% of the initial values over a 00 minute period. 1.2 The simulator computed values of critical parameters shall agree i2% of the reference plant parmeterc and shall not detract from training. 1.3 The calculated values of noncritical parameters pertinent to plant operation that are included on the simulator control room panels shall agree within 110% of the reference plant parameters and shall not detract from training. O , L O
(? TESTNUMBER/ TITLE: PTSl/51EADY SlAlf PERf0RMANCf 1[ v TEST DATA TABLE 1 LABEL DEECRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 216 - 224 in .375 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 635 - 661f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 2112 2198 psig
*l6JPR208 RCS Pressure (WR) Grfx Rcdr 2112 2198 psig
*RRSWTP Normalized Reactor Power Grtx Rcdr 98 - 102%
*PGJPE202 Gross Generation Grfx Rcdr 843 - 377 MWe Ll:6ADT delta 1 Loop A Grfx Rcdr 44.8 46.6f Ll:6BDT delta 1 Loop-B Grfx Rcdr 44.8 - 46.6f R3JPP208 Power Range Indication Grfx Rcdr 98 - 102%
R3JPP209 Power Range Indication Grfx Rcdr 98 102% R3JPP210 Power Range Indication Grfx Rcdr 98 102% R3JPP211 Power Range Indication Grfx Rcdr 98 - 102%
*LlJPR212 Th loop-A Grfx Rcdr 590 - 614F
*LlJPR213 Th loop-B Grfx Rcdr 590 - 614F
*HCTAVG RCS Average Temperature Grfx Rcdr 567 - 590f LITMARG4 Subcooling Margin Grfx Rcdr 45 - 47
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 545 - 567f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 545 567f
*LlJPR216 Tc Cold Leg Al Grix Rcdr 545 - 567f
*LlJPR217 Tc Cold Leg A2 Grfx Rcdr 545 - 567f n)
(" LlJPS287 SG A SV Level Grfx Rcdr 235 - 244 in LlJPS293 SG-B SV Level Grfx Rcdr 235 - 244 in LlJPS284 SG A Op level Grfx Rcdr 88.8 - 92.5 LlJPS336 SG B Op Level Grfx Rcdr 88.8 - 92.5 Grfx Rcdr 882 - 918 psig I
*f3JPA323 SG-A Press (ErlC)
*f3JPA320 SG B Press (EflC) Grfx Rcdr 882 918 psig LlJPT228 M5 Loop-A Hdr Prcss Grfx Rcdr 869 - 904 psig LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 869 - 904 psig
*LlJPS30! SG-A MfW flow Grfx Rcdr 5243 - 5457 KLB/hr
*llJPS302 SG-B MfW flow Crfx Rcdr 5243 - 5457 KlB/hr LlJPS329 SG A SU FW flow Grfx Rcdr .0245 .0255 KLB/hr LlJPS330 SG-B SU FW flow Grfx Rcdr .025 .026 KLB/hr L5GEAV Group 5 Avg Position Grfx Rcdr 98 - 102%
! L5G6AV Group 6 Avg Position Grfx Rcdr 98 - 102% L5G7AV Group 7 Avg Position Grfx Rcdr 86.8 - 90.4% HQBBCORE RCS Boron Grfx Rcdr 939 - 977 ppm Ll:RC61M Delta T-cold Grfx Rcdr .11 .13F l L4JPR234 RCS Loop A flow Grfx Rcdr 70.5 - 73.4 MLB/hr l L4JPR235 RCS Loop B flow Grfx Redr 70.5 - 73.4 MLB/hr l L4JPR233 RCS Totcl flow Grfx Rcdr 143 - 149 MLB/hr l tl:SP5AT feedwater Temperature Grfx Rcdr 447 465f l Ll:SP5BT feedwater Temperature Grfx Rcdr 447 - 465F l Ll:4AllR Steam Temperature Grfx Rcdr 580 - 604f I Ll:4BTIR Steam Temperature Grfx Rcdr 580 - 604f L- --
Q TEST NUMBER / TITLE: PTS)/ STEADY STATE PERFORMANCE TEST TEST DATA TABLE 2 LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parametu DEVICE CHART SPEED 100% POWER TO 25% POWER Hl:ll3 Pzr Level CTS 0 - 320 in N/A HlJPR203 Pzr Temperature CTS 0 - 700 f R3JPP208 Power Range Indic3 tion CTS 0 100%
R3JPP209 Power Range Indication CTS 0 - 100% R3JPP210 Power Range Indication CTS 0 - 100% R3JPP211-Power Range Indication CTS 0 - 100%
*LdJPR222 Loop A Pressure (NARROW) CTS 1700 - 2500 PSIG
*RRSWTP Normalized Reactor Power CTS 0 125 %
*PGJPE202 Unit 3 Generation CTS 0 1000 MWE
*LlJPR212 Loop A Th (NARROW) CTS 520 - 620 f HCTAVG RCS Average Temperature CTS 520 - 620 f
*LIJPR214 Loop A Tc (NARROW) CTS 520 - 620 f
*LlJPR215 Loop A Tc NARROW) CTS 520 - 620 F
*LlJPR216 Loop B 1c NARROW) CTS 520 - 620 f
*LlJPR217 Loop B Tc NARROW) CTS 520 - 620 F LlJPS287 SG A S/U Level CTS 0 250 in g LlJPS293 SG-B S/U Level CTS 0 250 in
'y LlJPS284 SG-A OP Level CTS 0 .100 %
L1JPS336 SG B OP Level C1S 0 100 % F3JPA320 "A" MS Pressure CTS 0 - 1200 PSIG r3JPA323 "B" MS Pressure CTS 0 1200 PSIG
*LlJP1228 "A" Header Pressure CTS 600 - 1200 PSlu
*LlJPT232 "B" Header Pressure CTS 600 - 1200 PSIG
*LlJPS301 Main fW flow "A" CTS 0 - 6000 KLB/HR
*LlJPS307. Main fW flow "B" CTS 0 - 6000 KLB/HR LlJPS329 S/U FW Flow "A" CTS 0 - 1500 KLB/HR LlJPS330 S/U FW flow "B" CTS 0 - 1500 KLB/HR N2JPX202 Letdown flow CTS 0 - 150 GPM N2JSA123 Makeup flow CTS 0 - 150 GPM N2JPX359 Makeup Tank Level CTS 0 - 100 in JPPAR230 RCS Total flow CTS 0 - 160 MLB/HR R3JPP212 N1-3 1R Level CTS -11.0 - -3.0 Log Amps llJPS285 SG A full Range level CTS 0 600 in LlJPS286 SG B full Range Level CTS 0 - 600 in SRT01 Al Steam Line Temperature CTS 0 600 f SRT07 Bl Steam Line Temperature CTS 0 600 f L1014Afl RCS Loop "A" flow CTS 0 - 80 MLB/HR L1014BFI RCS Loop "B" flow CTS 0 80 MLB/HR HOBBCORE RCS Boron CTS 500 - 1500 PPM RRRXT Core Xenon Reactivity CTS 0.0 - -0.6 dK/K O
DATE 08/08/93 O ecaronn^scE it51 noseEa, PlSi TITt.E STEADY STAIE PERFORMANCE TEST TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed rioun _ D . I ') w . E hil fl>f ~lest CoordTsthr,, /Date
'J a 2.0 All deficiencies have been resolved and thi test has been completed.
O IJoT 1<I:Quiltt:D FFC Test coordinator /Date 3.0 Test results reviewed. Simulator performanc if s satisfactory.
!l fnlatorl$(t'ruckfor a
/Date d1 J/M4 Operations Representative /Date
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Simula!.=.'ngineer
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/Date APFROVED M ___ DATE: _t/ _
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ll) TEST NUMBER / TITLE: PTTl/ REACTOR 1 RIP ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(19), Section B2.2(1) DATE PERFORMED: 01/09/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 89% power, steady state conditions when a reactor trip occurs. The simulator is stabilized in the normal post trip envelope in accordance with AP-580. tM.DlN0llE IIILE OPTIONS RP 012 Reactor Trip Module Trip Selectable to RPS Channel A, B, C and/or D Trip Module INITIAL CONDITIONS: 89% power, steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, simulator stabilized in the normal post trip envelope. PLANI PROCEDURES USED DURING THE TEST: AP 580 REVISION 15 REACTOR TRIP . Oa BASEllHE DATA: RECALL DATA FILE 0070287 - Reactor Trip from 89% Power During Vital Bus Transfer 07/02/87 @ 09:53:00 00ER 87-02 - Reactor Trip During Vital Bus Transfer O
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PERFORMANCE TEST NUMBER: P111 DATE: 01/09/91 TITLE: REACTOR 1 RIP TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. The post trip ref eed of the OTSGs that occurred at the plant was not repioduced on the simulator due to calibration differences / plant performance problems not incorporated on the simulator. This inability to replicate post trip refeeds has no training impact due to subsequent incorporation of the Rapid feedwater Reduction (RFR) circuit on both the plant and the simulator to prevent post trip refeeds from occurring. Excellent correlation between all other plant data and simulator response, ,n b
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. - - - - - .- . - _ = - . _
i l i { TEST NUMBER / TITLE: PTil/ REACTOR 1 RIP TEST DATA LABEL DESCRIPTION RECORDING RANGE TREQUENCY/
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 - 320 in .5 seconds
*HlJPR203 Piessurizer Temperature Grfx Rcdr 0 - 700f ;
*L4JPR221 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi Grfx Rcdr 0 2500 psi )
*L6JPR20B RCS Pressure (WR)
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125% !
PGJPE202 Gross Generation Grfx i<cdr 0 - 1000 MWe ! L5G6AV Grp 6 Avg Position Grfx R:dr 0 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th Loop-A Grfx Rcdr 520 620f
*HCTAVG RCS Average Temperature Grfx Rcdr $20 620f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f
*LlJPS287 SG A SV Level Grfx Rcdr 0 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 100%
*f3JPA323 SG-A Press (EflC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG-B Press (EflC) Grfx Rcur 0 1200 psi L1JPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT?32 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
" *LlJPS301 SG A MfW flow Grfx Rcdr 0 - 6K KLB/ilR
*LlJPS302 SG 8 MfW flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW ricw Grfx Rcdr 9 - 1500 MLB/HR
*L1JPS330 SG B SU FW flow Grfx Redr 0 - 1500 MI.8/HR l
l
{ PERFORMANCE TEST NUMBER: P111 DATE: 01/09/91 TITLE: RfAC10R1RJ_P TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed ' JGl__ _ _. [ tor f t 2.0 All deficiencies have been resolved and this test has been completed. O NOT REQU1HI:D ~ D F Test Coordinator /Date 3.0 ~ Test results reviewed. Simulator performance is satisfactory.
- 8 E' M I 7
5 ag[ tructor Jiae OperattnsRepresentative/Date
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Sin atoriWIdM e APPROVED: du , e M DATE: 9 3k'M l R$15/N(L.M a LOT k '
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TEST NUMBER / TITLE: PTT2/ TOTAL LOSS OF FEEDWATER (MAIN AND E ERGENCY) ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(10), Appendix B, Sectioti B2.2(2) DATE PERFORMED: 01/10/91 TEST DURATION: 1/2 hour TEST D:SCRIPTION: With the simulator ope : ting at 100% power, both main feedwater pumps trip, resulting in a reactor / turbine trip. EFP 1 fails to start; EfP 2 starts but immediately trips on overspeed. The steam generators boil dry; due to the loss of secondary heat sink, RCS temperature, pressurizar level and RCS pressure begin increasing, liAlf_VNCTIOJ IIILE QPTIONS FW 005 Main Feedwater Pump Trip Selectable to A or B Main FW Pump EF-003 Emergency FW Pump fails Selectable to EfP-1 or EFP-2 to Start EF-001 Emergency FW Pump Trip Selectable to EFP-1 or EfP 2 INITIAL CONDITIONS: Full power steady state canditions, all plant equipmant in service. FINAL CONDITIONS: RCS temperature increasing due to loss of secondary heat sink. Pressurizer full, RCS pressure cycling around the PORV lift point, saturation margin decreasing. PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP BASELINE DATA: CR-3 ATOG PART II, Volume 2, Appendix B Secticn 3.0 - Loss of Secondary Inventory Control l O l
PERf0RMANCE TEST NUMBER: P112 DATE: 01/10/91 TITLE: TOTALLOSSOffE[DWATER(MA14])fl0EMERGEldCY) TEST RESi'LTS/ COMMENTS: Test result s satisfactory. No performance problems noted. Excellent correlation between simulator response and the information provided in the ATOG discussion of loss of feedwater transients. This test verifies the fidelity of the simulator response to a total loss of feedwater to the point where an RCS heatup and respressurization i; in progress, the pressurizer has refilled and the RCS is cycling on the pressurizer PORV. Simulator response to a recovery from this accident (i.e., feed and bleed coeling, reestablishment of EfW/0TSG cooling and restoration of normal pressure control (pressurizer bubble)) is verified in performance test PIM10. O v h l
,Q _ , _ _ . . -
l
l { TEST NUMBER / TITLE: P112/T01AL LOSS Of FEEDWATER (MAIN AND [MERGEIK*) TEST DATA . LABEL DESCRIPTION RECORDING RAhGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*}{l:113 Pressurizer Level Grfx Rcdr 0 - 320 in .5 sec
*}{lJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700T
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi '
- RP.SW1 P Normalized Reactor Power Grfx Rcdr 0 125Y, PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 1007.
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 1007,
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 620F
*L1JPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620f
*LlJPS287 SS A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG A Op Level Grfx Redr 0 - 1007,
*L1JPS336 SG B Op Level Grfx Rcdr 0 - 1007,
*F3JPA323 SG A Press (EflC) Grfx Rcdr 0 - 1?00 psi
*f3JPA320 SG B Press (EflC) Grfx Rcdr 0 - 1200 psi LlJP1228 MS Loop A Hdr Press Grfx Rcdr ~00 - 1200 psi LlJPT232 MS Loop-B Hdr Prms Grfx Rcdr 600 - 1200 psi s *LlJPS301 SG- A MF W flow Grfx Rcdr 0 - 6K KLD/HR
*LlJPS302 SG-B MfW Flow Grfx Rcdr 0 - 6K KLB/HR
*L1JPS329 SG A SU FW flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS340 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R4JSA019 incore T/C Grfx Rcdr 0 - 2500f S R4JSA020 Incore T/C Grfx Rcdr 0 - 2500f R4JSA023 Incore T/C Grfx Rcdr 0 - 2500f R4JSA025 Incore-T/C Grfx Rcdr 0 - 2500F C --
{ PERFORMANCE TEST NUMBER: P112 DATE: 01/10/91 YlTLE: LOSS Of FEE 0 WATER (MAIN AND EMERGENCY) TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1 TR # Priority Date is a d Date Closed M1 tor 7.0 All deficiencies have been resolved and this test has t,een completed. O V NOT HI?QUIHisD TR~ Test Coordinator /DTte 3,0 Test results reviewed. Simulator performance is sati. factory.
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~ ~/6 ate APPROVED: -< c ut - , -4 DATE: _1 S(g_
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TEST NUMBER / TITLE: P113/SIMULTANE005 CLOSURE Of ALL MAlf4 STEAM 1SOLA110f4 VALVrs ANSI /ANS-3,5(1985) CROSS
REFERENCE:
Appendix B Section B2.2(3) DATE PERFORMED: 01/09/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 73% power, an inadvertant main steam isolation occurs, closing all four main steam isolation valves. A reactor / turbine trip on high RCS pressure occurs and the plant is stabilized in the normal post trip envelope. RALFUNCTION 1))LE OPILOM None. Transient initiated 'y mai,'al actuation of Main Steam Isolation on the Main Control Board. INITIAL CONDITIONS: Simulator at 73% power, B condensate and circulating water pumps secured. FINAL CONDITIONS: Reactor / turbine tripped, plant stabilized in the normal post trip envelope with steam generator levels being maintained by emergency feedwater. PLA:T PROCEDURES USED DURING THE TEST: C AP 580 REV 15 REACTOR 1 RIP BASELINE DATA: 00ER 82 Reactor Trip Initlated By inadvertant Main Steam Line Isolation Valve Closure. 06/20/82 @ 00:33:06 o_
~~~9.-* ,
Q PERFORMANCE TEST NUMBER: PTT3 DATE: 01/09/91 TITLE: SIMULTANEOUS CLOSURE OF All MAIN STEAM ISOLATION VALVES l TEST RESULTS/ COMMENTS: l 1est results satisfactory. No performance problems noted. l During *.he plant transient, both main feedwater pumps tripped. The exact cause of the trips is not known due to problems with the post trip reviewer and a subsequent loss I of secondary plant post trip data. Main feedwater pump trips did not not occur on ' simulator in the same time frame as in the plant; this difference in response cannot be evaluated due to the above mentioned lack of plant data. Excellent correlation between all other plant data and simulator response. O p l m
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. _ . . -~- - -_ . -.- .. - - - . _ - - . - . . - . . - - _ -
Y r { TEST NUMBER / TITLE: P113/SIMULTANE0VS CLOSURE OF ALL MAIN STEAM ISOLAT10N VALVES. TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 320 in .5 sec *HlJPR203 iretsuri7er Temperature Grfx Rcdr 0 - 700f *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 1257.
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 1007, L5G7AV Grp 7 Avg Position Grfx Rcdr 0 1007.
*LlJPR212 Th loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG A Op Level Grfx Rcdr 0 1007,
*LlJPS336 SG B Op Level Grfx Rcdr 0 1007,
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG B Press (EflC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 pst Grfx Rcdr 600 - 1200 psi
*LlJPT232 MS Loop B Hdr Press O tiJPS30i SG ^ Mrw riow
*LlJPS302 SG B MfW flow Grfx Rcdr Grfx Rcdr 0 6K KtB/HR 0 - 6K KLB/HR
*LlJPS329 SG A SU fW flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU IW flow Grfx Redr 0 - 1500 MLB/HR f3:23Fil EfW Flow S/G B Grfx Rcdr 0 - 1000 gpm f3:25fil EfW flow S/G A Grfx Rcdr 0 1000 gpm f3:26fil EfW flow S/G A Grfx Rcdr 0 1000 gpm F3:24Fil EfW Flow S/G B Grfx Rcdr 0 - 1000 gpm O
Q PERFORMANCE TEST NUNDER: PTT3 DATE: 01/09/91 TITLE: SlhULTANE005 CtOSURE Of All MAIN STft.M IS0lA110N VALVES TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed Md _ ._.__..._ . . - _ _ ____- _ _ N WWO 4 Iu sM4N 9=1m MWM-4 c
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2.0 All deficiencies have been resolved and this test has been completed. O NOT ltEOUlHED Iffliit Coordriiaior ~~7Difs 3.0 Test results reviewed. Simulator performance is satisfactory, y OT du U[peratl6kRepresentative/Dil~e^
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DATE: _13 ji[_ Nf /NLOlM[*"Q ' '- ' '
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TEST NUMBER / TITLE: PTT4/ LOSS Of 6900 VAC AND 4160 VAC UNIT BUSSES FROM FULL POWER ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3), Appendix 8. Section B2.2(4) DATE PERFORMED: 01/09/91 TEST DURATION: 1/2 hour ., TEST DESCRIPTION: With the simulator operating at 92% power, a turbine trip occurs, resulting in a loss of the auxiliary transformer. A phase differential on the startup transformer occurs, resulting in a loss of power to the 6900 VAC and 4160 VAC Unit busses. Using applicable APs, the sinulator is stabilized in mode 3 with natural circulation cooling of the RCS in prop-?ss. lLAlfjtNQ11M 111LE OPTIMS EH-001 Main Turbine Trip ED 001 Loss of the Startup Transformer INITIAL CONDITIONS: 92% power, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped. RCS heat removal being accomplished by natural circulation. OTSG 1evels being maintained at ES% by EFIC ES 4160 VAC busses being suppli.:d from the emergency offsite power source. PLANT PROCEDURES USED DURING THE TEST: AP 580 REV 15 REACTOR TRIP AP-450 REV 15 EMERGENCY FEEDWATER ACTUATION AP 530 REV 0 NATURAL CIRCULATION BASELINE DATA: U0ER 81 Reactor Trip and Loss of Offsite Power 6/16/81 CR-3 AT0G, PART 11, Volume 2, Appendix 0 - Loss of Offsite Power CR-3 AT0G, PART 11, Volume 1 Addendum B - Natural C rculation l lC l ' ? mr -=- 4,--,v w. ,,wy-m,. - - - . r , ,,,._,,.,,,m.,,__.. ,, , , _ . , , _ , , . _ . , , _ , , , , , _ , , , , . , _
PERFORMANCE TEST NUMBER: PIT 4 DATE: 01/09/91 TITLE: LOSS OF 6900 VAC AND 4160 VAC UNIT BUSSES fRCH FULL I'0WER f TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. During this transient, the plant experienced an excessive coaldown to control problems l with the startup feedwater control valves, resulting in excessive emergency feedwater
- flow (note that this transient occurred before installation of EFIC on the plant).
This overcocling was not reproduced on the simulator due to the fact that emergency feedwater control on the simulator utilizes EflC. Excellent correlation between all other plant data and simulator response. Note that simulator response to an overcooling while on natural circulation is verified in performance test PTM3A.
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g TEST NUMBER / TITLE: PTT4/ LOSS Of 6900 VAC AND 4160 VAC UNIT BUSSES FROM fl1LL POW [R TLST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 - 320 in .5 sec *HlJPR203 Pressurizer Tr.: 7erature Grfx Rcdr 0 700F *L4JPR222 RCS Pressure ".d) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 ;>si *RRSWTP Normalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe , L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% i L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th loop-A Grfx Rcdr 520 - 620f *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f I *LlJPR214 Tc Cold leg Al Grfx Rcdr 520 - 620f i *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f *LlJPS287 SG-A SV Level Grfx Rcdr 0 250 in *LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in *LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *f3JPA323 SG-A Press (ErlC) Grfx Rcdr 0 - 1200 psi *f3JPA320 SG-B Press (EflC) Grfx Rcdr 0 - 1200 psi LiJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi Grfx Rcdr 600 - 1200 psi 9 *LlJPT232 MS Loop B Hdr Press LlJPS301 SG-A MFW flow *LlJPS302 SG-B MfW flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR *LlJPS329 SG-A SU fW flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG-B SU FW flow Grfx Rcdr 0 - 1500 MLB/HR
, L4JPR233 Total RCS flow Grfx Rcdr 0 - 160 MLB/HR l Ll:6ADT Loop A Delta-T Grfx Rcdr 0 - 70f Ll:6BDT Loop B Delta-T Grfx Rcdr 0 70f F3:23ril EfW flow S/G B Grfx Rcdr 0 - 1000 gpm f3:25fil EfW flow S/G A Grfx Rcdr 0 - 1000 gpm f3:26fll EfW flow S/G A Grfx Rcdr 0 - 1000 gpm f3:?4fil EfW flow S/G B Grfx Rcdr 0 - 1000 gpm l l l l 1 m ,. - U
PERf0RMANCE TEST NUMBER: P114 DATE: 01/09/9 _ TITLE: LOSS Of 6900 VAC AND 4160 VI.C UNil BUSSES FROM FULL POWER TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Clcsed _llW1__ .___.._ _ _ _ . . _ . _ _. g4P tor Dt 2.0 All deficiencies have been resolved and this test has been completed, im O mrr su:ousia:n fPC~lest Coordinitor /Dati 3.0 Test results reviewed. Simulator performance is satisf actory, f),' k0 nTiructor //(Dat ehh, $O5 4 OperatiNsRepresentative/Date T ?.) S I
/ Simul 1
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s _ S ator Ingineer '/Date ( APPROVED: h ( 2: 1 w- , n DATE: _Sk$jfiL NST /NLOTM1 l
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O 'ts' "u"Soa/'i' ""5'at^c'on coo'^"' "u"" 'a'" ANSI /ANS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(4), Appendix B, Section 82.2(5) DATE PERFORMED: 01/10/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 907, power when RCP-1A trips. The plant is automatically run back to 75Y by the ICS and . stabilized in accordance with applicable plant procedures. l liALFUNCT108 1111I DELLQfil RC-007 RC Pump Trip Selectable to RCP-1A, IB, -10 and/or
-10 INITIAL CONDITIONS: 90Y. power, steady state conditions, all plant equipment in service.
FINAL CONDITIONS: Plant stabilized at ~757. power with RCP-1A tripped, i 1 PLANT PROCEDURES USED DURING THE TEST: i AP 545 REV 3 PLANT RUNBACK U0ER 82 Successful Runback on Loss of One Reactor Coolant O BASEllNE DATA: Pump. 3/27/82 l l I l n Q l l L l
g PERf0RMANCI TEST NUMBER: PT15 DATE: 01/10/91 TITLE: " LAC 10R COOLANT PUMP 1 RIP TEST RESULTS/COHHENTS: Test results satisfactory. No performance problems noted. Excellent correlation between available plant data and simulator response. p V O v m
l { TEST NUMBER / TITLE: P115/REAC10R COOL ANT PUMP 1 RIP TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / CCritical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in .5 seconds
*lilJPR203 Pressurizer Tamperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - ?500 psi
*RRSWTP Hormalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grix Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 50f 1.1:6BDT delta i Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop-A Grfx Rcdr 520 - 620f
*L1JPR213 1h loop B Grfx Rcdr 520 620f
*HCTAVG RCS Average lemperature Grfx Rcdr 520 620f LllMARG4 Subcooling Margin Grfx Rcdr -10 410f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620f
*LlJPR216 Tc Cold leg B1 Grfx Rcdr 520 620f
*LlJPR217 Tc Cold leg B2 Grfx Rcdr 520 - 620f
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 100%
bs *F3JPA323 SG-A Press (EflC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG-B Press (EflC) Grfx Rcdr 0 - 1200 pst LlJP1228 MS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS301 SG-A MfW flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 Sb B MfW flow Grfx Rcdr 0 6K KLB/IIR
*LldPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/ilR
*LlJPS330 SG B SU fW flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm Ll:RC61M Delta T Cold Grfx Rcdr +10f L4JPR234 RCS Loop A flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total flow Grfx Rcdr 0 160 MLB/HR n . - _
(v, PERFORMANCE TEST NUMBLA: plt 5 DATE: 01/10/< TITLE: REAC10R COOLANT PUMP 1 RIP TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed 3%. Y$ fPC les,t Coordinator J
. b+t e j .
2.0 All deficiencies have been resolved and this test has been comp 1cted. O V NOT RWUIRr.D FPC test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. fZ I inub$ or instructor r4/p
'/Date (LJ) L h 8 9 7 'r /
OperatiIsRepresentativc/Date
'Y ~
5TmulatorInTineer
^ ]/hN / ate C ..- .
APPROVED:
> k NlI gh2.f@~:f NST /NLO1M c 11 DATE: ,1 % 31 t ~ ~ ~'
O s
TEST NUMBER / TITLE: plt 6/ MAIN TURBillE 1 RIP ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(15), Appendix 0, Section 82.2(6) DATE PERFORMED: 01/10/91 TEST ')URATION: 1/2 hour TEST DESCRIPTION: The plant is operating at 40% power when a turbine trip occurs, lhe ICS automatically runs the plant back to 15% power. Cycling of the PORV carly in the transient may be required to prevent a reactor trip or high pressure; this is consistent with normal plant operating philosophy. tiALLUE110B IllLI Ofl10M EH-001 Main Turbine Trip INITIAL CONDITIONS: 40% power steady state conditions, one main feed pump running, main feedwater block valves closed. FINAL CONDITIONS: Plant stabilized at 15% power. Turbine tripped, header pressure being maintained at 885 psig by the turbine bypass valves, steam generator levels being maintained at low level limits by one main feedwater pump. fw PLANT PROCEDURES USED DURING THE TEST: AP 660 REV 6 TURBlNE TRIP BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. M O
DATE: 01/10/91 C PERFORMANCE TEST NUMBER: P116 TITLE: MAIN TURBlNE TRIP TEST RESULTS/COMMEH15: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. C; o -
. . . .W
(] ., TEST NUMBER / TITLE: P116/ MAIN TURBINE 1 RIP T EST DAT A LABEL DESCRIPTION RE;0RDlHG RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 - 320 in .5 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi s
*L6JPR208 RCS Pressure (WR) Crfx Rcdr 0 2500 psi ,
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th Loop A Grfx Rcdr 5?0 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f ',
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 620f
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620f -
*L3JPS287 SG-A SU Level Grfx Rcdr 0 250 in
*L1,1PS293 SG-B SU level Grfx Rcdr 0 - 250 in
*L11PS284 SG A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 100%
*f3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG B Press (EflC) Grfx Rcdr 0 - 1200 poi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi Grfx Rcdr 600 - 1200 psi C; *LlJPT232 MS Loop-B Hdr Press LlJPS301 SG A MfW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG B MFW Flow Grfx Rcdr 0 6K KLB/HR
*LlJPS329 5G-A SU FW flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW flow Grfr Rcdr 0 - 1500 MLB/HR
\ ,
N 3.
\
s O . - -
C PERFORMANCE TEST NUMBER: PTT6 DATE: 01/10/91 TITLE: MAIN TURBINE TRIP TEST STATUS: 1.0 This test is complete and has the following deficWeies noted (if none so injicate): TR # Priority Date issued Date Closed _NONE ik /
'Plehcoordinator /06te 2.0 Ai, deficiencies'have been resolved and this test has been completed.
NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator periormance is satisfactory. _, m / (LDTw EI 8' 17- 9 I
/Simula. tructor //D e Operatidhs Representative /Date
'/I ~' ?
SimuliithEsgineTr 7/9l71
' 7Date l
APPROVED: R kw
' \
w_ DATE: %.%k%t' iTS]/NLOTM i
u m. C TEST NUMBER / TITLE: PTT7/ MAXIMUM RATE POWER RAMP ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.l(6), Appendix 8. Section B2.2(7) DATE PERFORMED: 01/10/91 TEST DURATION: 1/2 hour 12ST DESCRIPTION: With the simulator operating at 100% power, a load decrease to 75% is initiated at 10% per minute with the ICS in full automatic. When the plant reaches 75%, a load increase to 100% is initiated at 10% per minute with the ICS in full automatic, tLALFUNCTION TITLE OPTION.S INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power conditions with a minor xenon transient in progress. runNT PROCEDURES USED DURING THE TEST: OP-204 REV 53 POWER OPERATIONS O U BASELINE DATA: There is no plan; data available for this transient. Expected L simulator response based on a combination of expert judgement and extrapolations of Known plant response during normal power changes. l O
O PERFORMANCE TEST NUMBER: PTT7 DATE: 01/10/91 TITLE: MAXIMUM RATE POWER RAMP TEST REStiLTS/ COMMENTS: Test results evaluation based on expert judgement and extrapolations of known plant response during noraml power changes. Test results satisfactory. No performance problems noted. P Ev l-l
\ ~
< v s
f TEST NUMBER / TITLE: PTT7/ MAXIMUM RA POWER RAMP f] TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Gefx Rcdr 0 - 320 in .5 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100%
-L5G7AV Grp 7 Avg "osition Grfx Rcdr 0 - 100%
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG_ RCS Average Temperature Gtfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcd. 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SC-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 HS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi p) t *LlJPS301 SG A MfW Flow Grfx Rcdr 0 - GK KLB/HR
*LldPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R3JPP210 Power Range Indicstion Grfx Rcdr 0 - 125%
R3JPP217 Power Range Imbalance Grfx Rcdr -62.5 - +62.5% RRRXT Xenon Reactivity Grfx Rcdr .025 - .026 dk/k HQBBCORE RCS Baron Grfx Rcdr 0 - 3000 ppm _a
DATE: 01/10/91 Q PERFORMANCE TEST NUMBER: PIT 7 TITLE: MAXIMUM RATE POWER RAMP TEST STATUS: 1.0 This test is complete anu nas the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _NONE __
=
l
- A /A
//,(q,j Tepf Coorilinator Iate 2.0 All deficiencies have been resolved and this test has been completed.
O NOT TmQUEED _FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
/) ,~ j (LO \gh %-9 M l
, f imul4(or 16structor te OperatioMs Representative /ifds
[ l s' s x lb kY Simulattir Erig;neer hll nh 'n
' / iTs APPROVLu. c't _ ^""l ' t DATE: 1 l
^
NST 4LOTM \ 1 0
C TEST-NUMBER / TITLE: PTT8/LARGE BREAK LOCA(DBA) ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(b)(c), Appendix B Section B2.2(8) DATE PERFORMED: 07/29/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when an 8.55 ft2 break l is initiated at the discharge of the B reactor coolant pump; a simultaneous loss of offsite power with a failure of one emergency diesel generator to start also occurs. MALFUNCTION llILE OPTIONS RC-033 RCS Cold leg Break Adjustable from 0 - 8.55 ft2 Selectable to A, B, C, or D RC Pump discharge EG-001 EDG Start failure Selectable to A or B EDG l ED-001 S/V Transformer Failure INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. ES 4160 V busses aligned to the S/U Transformer. FINAL CONDITIONS: RCS at ~20 psig, 260 F. One train of HPI, LPI and Building Spray in operation. , , _ _ l PLANT PROCEDURES USED DURING THE TEST: m (s AP-580 REV 16 AP-380 REV 18 REACTOR TRIP ENGINEERED SAFEGUARDS ACTUATION AP-770 REV 15 EMERGENCY DIESEL GENERATOR ACTUATION l I ( BASELINE DATA: CR-3 AT0G, Part II, Volume 2, Appendix F - Loss of Coolant Accidents L _ m -- -
\
l
'f': v PERFORMANCE TEST NUMBER: PTT8 DATE: 07/29/91 TITLE: LARGE BREAK LOCA(DBA) TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. Good correlation between simulator response and AT0G data / transient description. O v
{ TEST NUMBER / TITLE: PTT8/LARGE BREAK LOCA/DBA TEST DA1A LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 -2500 psi LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op Level Grfx Rcdr 0 100%
LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LldPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi l LlJPT232 MS Loup-B Hdr Press Grfx Rcdr 600 - 1200 psi LlJPS301 SG-A MFW Fleu Grfx Rcdr 0 - 6K KLB/HR LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 ML6/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R4JSA019 Incore T/C Grfx Rcdr 0 - 2500F R4JSA020 Incore T/C Grfx Rcdr 0 - 2500F
- R4JSA023 Incore T/C Grfx Rcdr 0 - 2500F l g R4JSA025 Incore T/C Grfx Rcdr 0 - 2500F
'o *ClJPP254 Containment Pressure Grfx Rcdr 0 - 70 psi
*CATAVG Avg Containment Temperature Grfx Rcdr 50 - 300F CALNS Normal Sump Level Grfx Rcdr 0 10 FT CALES Emergency Sump Level Grfx Rcdr 0 - 10 FT N2:M18 HPI Flow Al Grfx Rcdr 0 - 500 gpm N2:M14 HP1 Flow A2 Grfx Rcdr 0 - 500 gpm l N2:HI2 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 PPI Flow B2 Grfx Rcdr 0 - 500 gpm
, F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm E F3:25FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm l F3:24FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm L N3:Dil LPI Flow A Grfx Rcdr 0 - 5000 gpm l N1:Bil Building Spray Flow A Grfx Rcdr 0 - 1800 gpm l N3:013 LPI Flow B _ Grfx Rcdr 0 - 5000 gpm N1:BI2 Building Spray Flow B Grfx Rcdr 0 - 1800 gpm (
.=
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' PERFORMANCE TEST NUMBER: PTT8 DATE: 07/29/91 TITLE: LARGE BREAK LOCA(DBA)
TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed LGL .--
/ e DN 2.0 All deficiencies have been resolved and this test has been ompleted.
NOT REQUIRED FPC Test Coordinator /Date 3,0 Test results reviewed, Simulator performance is satisfactory, f[f ELD I qN T-27-S I imu atto nstructor /Date OperationsURepresentative/Date j
. -l If [
Simul a' tor -Engiyeer 9/9/7/
'/Date APPROVED: l') y - 9 % DATE: Rk R\
\
NSi,S/NLOTM ' > s 6 i O
C TEST NUM8ER/ TITLE: PTT9/ MAIN STEAM LINE BREAK INSIDE C0tTAINMENT ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(20), Appendix B, Section 82.2(9) DATE PERF0kMED: 01/10/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, a main steam line break of approximately 20E+06 lbm/hr is initiated inside containment A reactor / turbine trip occurs, followed by main steam isolation, main 'eed isolation and a safeguards actuation on high containment pressure, lhe RCS experiences a rapid cooldown and depressurization which is terminated when the affected OTSG boils dry, liAlf0NCHQN llRE OHLold MS-001 Main Steam line Break Inside Selectable to the Al, A2, B1 and/or Containment 82 Main Steam Line Adjustable from 0 - 20E+06 lbm/hr INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, affected 0TSG isolated and boiled dry. l RCS parameters stabilized in the normal post trip window, with the intact steam generator level control being maintained by A emergency feedwater. Safeguards systems secured. 1 PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-380 REV 17 ENGINEERED SAFEGUARDS ACTUAll0N AP-450 REV 15 EMEPSENCY FEEDWATER ACTUATION t.P-460 REV 8 STEAM GENERATOR ISOLATION ACTUATION , BASELINE DATA: There is no plant data for this transient. Expected simulator l_ response based on a combination of expert judgement and i extrapolation of FSAR data based on actual plant conditions vs l design b1 sis conditions.
?Q i
i
-e_._.., _
J
s .-. - mu (]) PERFORMANCE TEST NUMBER: PTT9 DATE: 01/10/91 TITLE: MAIN STEAM LINE BREAK INSIDE CONTAINMENT _ TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. A comparison of simulator response to FSAR data for a main steam line break shows a significant difference in the magnitude of the predicted vs actual RCS cooldown and depresurization resulting from the break. This difference appears to be primarily due to significant differences in assumed and actual RCS inventory and assumed and actual 0TSG secondary side inventories at the start of the transient. The higher initial RCS inventory and lower OTSG secondary side inventories in the simulator result in a smaller cooldown and depressurization than predicted in the FSAR. Test results satisfactory. No performance problems noted. (rm V l l i I i O
g TEST HUMBER/ TITLE: PTT9/ MAIN STEAM LINE BREAK INSIDE CONTAINMENT TEST DATA _ LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in .5 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi l
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LIJPR214 lc Cold Leg Al Grfx Rcdr 520 - 620F
*L1JPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi
cp B Flow Grfx Rcdr 0 - 80 MLB/HR
*R3JPP202 Source Range CPS Grfx Rcdr -1.0 - 6.0 cps
*R3JPP2CS Source Range CPS Grfx Rcdr - 1. 'J - 6.0 cps R3]PP212 intermediate Range Amps Grfx Rcdr -11.0 - -3.0 amps R3JPP213 1sterrediate Range Amps Grfx Rcdr -11.0 - -3,0 amps C UPP254 Containment Pressure Grfx Rcdr 0 - 70 psi CATAVG -Avg Containment temperature Grfx Rcdr 50 - 300F j CALNS Normal Sump Level Grfx Rcdr 0 - 10 FT cal.ES Emergency Sump level Grfx Rcdr 0 -10 FT HKilIQ- RCDT Level Grfx Rcdr 0 - 15 ft HKPGAS RCDT Pressure Grfx Rcdr 0 - 200 psi
*HKWPRZ Flow from Pzr to RCDT Grfx Rcdr 0 - 60 LB/SEC H17E30 RCV-10 Discharge Temp Grfx Rcdr 100 - 300F N2:M18 HPl Flow Al Grfx Rcdr 0 - 500 gpm N2:MI4 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:M12 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 HPI Flow B2 Grfx Rcdr 0 - 500 gpm o _ .__
i (~\
; TEST NUMBER / TITLE: PTTIO/ FAILURE OF PORV WITH HPI INOPERABLE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEilCE CHART SP,EED F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm O
w r O
DATE: 08/14/91 { PERFORMANCE TEST NUMBER: PTT10 TITLE: PORV FAILURE WITH HPI INOPERABLE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _L W S _H 08/14/91 _TR-0249 _B 08/14/91 0s/30/91 _[ ' _FPC Test Coardinator D
/Date 2.0 All defici ncies have been resolved and this test has been completed.
AQ *Closcout of TR-248 does not prevent D s f[
~
2J ff completion of thin test
. C Tep Coordinator /Date 3.0' Test results reviewed. Simulator performance is satisfactorv.
t 8.21k9t Odi ix) -
#-/7 '?/
Operatio'ns Representative /Date Simulator I nstructor 'Date [ / , 3lL))_] Simulator {ngi,nser /Date APPROVED: _ _
/ DATE:
f 7h r
l TEST NUMBER / TITLE: PTM1Al/ STEAM GENERATOR IVBE LEAK ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(a) DATE PERFORMED: 07/15/91 TEST DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating at full power steady state conditions when a 50 gpm tube leak is initiated in the B OTSG. Utilizing plant procedures, the plant is shutdown, the faulted steam generator is identified and isolated and the plant is cooled down and depressurized to 530F, 1700 psig in preparation for a cooldown to Mode 5. MALFUNCTION TITLE OPTIONS RC-004 OTSG Tube Leak Adjustable from 0 - 400 gpm Selectable to A and/or B OTSG INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. _ FINAL CONDITIONS: Plant shutdown, cooled down to 530F, depressurized to 1700 psig. B OTSG isolatad, with level being controlled by a combination of main feedwater flow, leak flow and steam flow to
- the condenser through the turbine bypass valves.
f .] PLANT PROCEDURES USED DURING THE TEST: OP-204 PEV 54 POWER SPERATIONS OP-208 REV 37 PLANT SHUTD0bN OP-209 REV 73 PLA'4T C00LDOWN EP-390 REV 12 STEAM GENERATOR TUBE LEAK AP-660 REV 6 TURBINE TRIP AF-580 REV 16 REACTOR TRIP BASELINE DA:A: There is no plant data for this transient. Expected simulator response based on expert judgement. m V
DATE: 07/15/91 ll)' PERFORMANCE TEST NUMBER: PTM1A1 TITLE: STEAM GENERATOR IVBE EEAK TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All procedure steps functioned normally. Plant was shutdown in accordance with OP-204, 0?-208 and OP-209. Simulator response was excellent. The following anomalier. existed:
- 1. When letdown flow was reduced , ar EP-390, the indicated flow oscillated ~15 gpm for about I minute. Oscillation was dampened by adjusting MUV-51 position.
- 2. When isolating 8 OTSG at <540f the B OTSG S/U Block Valve (fWV-33) was inadvertently closed for several minutes. Vlave was reopened with no impact on the test.
C v 6 I 1 O g _ .. _ _ _ _ _ _ _ _ _ . _ - . _ _ _ _ _ . _
TEST NUMBER / TITLE: PTM1Al/ STEAM GENERATOR TUBE LEAK -] TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec *HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F $L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Fressure (WR) Grfx Rcdr 0 - 2500 psi N2JPX202 Letdown flow Grfx Rcdr 0 - 160 gpm N2:M10 Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Injection Flow Grfx Rcdr 0 - 80 gpm N2JFX359 Makeup Tank Level Grfx Rcdr 0 - 120 in '*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe l Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50F L1:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Log A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr J - 100%
O F3JeA323 SG-^ eress (EFIC)
*F3JPA320 SG B Presa (EFIC)
Grfx Rcdr Grfx Rcdr 0 1200 Psi 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT23? MS long d Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LIJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS3?0 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5GSAV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcar- 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm CMCRMG26 RMG26 MSL B1 Rad Monitor Grfx Rcdr -i - 6 log mr/hr CHORMA12 RMA12 Noble Cas Monitor Grfx Rcdr 0 - 6 log cpm CMORMA02 RMA2 Noble Gas Monitor Grfx Rcdr 0 - 6 log cpm CMORMA04 RMA4 Noble Gas Monitor Grfx Rcdr 0 - 6 log cpm
l' PERFORMANCE TEST NUMBER: PTHIAl DATE: 07/15/91 V TITLE: STEAM GENERATOR TUBE LEAK TEST STATUS: 1.0 This test is coeplete and has the following deficiencies noted (if none so indicate): TR # w ity Date issued Date Closed JOM _ _ _
}s
- ~s 1 31
{ 1d._doordthe/ Test ator 2.0 All deficiencies have been resolved and this test has been completed, p ,, p TV ice Co NiStor di 3.0 Test results reviewed. Simulator performance is satisfactory. ii , L-d r=2Adn . 't 't I.ft L _I__ ._ In Operatioht Representatfve/Dtte b SS _.
$ltnulator InsUructor /Dath 4
/
y'.>/,f _ '! / v /, ~ilyj time 7korYn'glnNr /bte APPROVED:
- d%l/ .
DATE: _/
//
4 e J
TEST NUMBER / TITLE: PTM1A2/ STEAM G NERATOR TUBE RUPTURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(a) DATE PERFORMED: 07/16/91 TEST DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating at full power steady state conditions when a 320 gpm tube rupture is initiated in the A OTSG. Utilizing plant procedures, the plant is shutdown, the faulted steam generator is identified and isolated and an emergency cooldown and depressurization is initiated. tiALFUNGT10h TITLE OPTIONS RC-004 OTSG Tube '.eak Adjustable from 0 - 400 ypm Selectable to A and/or B OTSG INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Plant shutdown, cooled down to <500F, depressurized to 1000 psig. A OTSG isolated, with level being controlled by a combination of leak flow and steam flow to the condenser through the turbine bypass valves. RCS inventory being maintained by HPI. PLANT PROCEDURES USED DURING THE TEST: OP-204 REV 54 POWER OPERATIONS OP-208 REV 37 PLANT SHUTDOWN OP-209 REV 73 DLANT C00LDOWN EP-390 REV 12 STEAM GENERATOR TUBE LEAK AP-660 REV 6 TURBINE TRIP AP-580 REV 16 REACTOR TRIP BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement.
g PERFORMANCE TEST NUMBER: PIM1A2 DAiE: 07/16/91 TITLE: STEAM GENERATOR TUBE RUPTURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert , judgement. All procedure steps functioned correctly. Noted small oscillationt in letdown flow while the bleed valve was selected to RCBT position. This has been previously identified as a problem. A steam generator level oscillation developed during power reduction. The oscillation was short-lived and did not impact the test scenario. All other simulator response was excellent. (m. V [ LJ
] TEST NUMBER / TITLE: PTHIA2/ STEAM GENERATOR TUBE RUPTURE TEST DATA LABEL DESCRIP110N RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi I N2JPX202 Letdown Flow Grfx Rcdr 0 - 160 gpm N2:MIO Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Injection Flow Grfx Rcdr 0 - 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 5?0 - 620F
*HCTAVG RCS Average Temperatere Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LldPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
C: F3Jen323 SG-A eress (EFiC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 --100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 --100% HQBBCORE RCS Baron Grfx Rcdr 0 - 3000 ppm CMORMG25 RMG25 MSL Al Rad Monitor Grfx Rcdr 6 log mr/hr i CHORMAl2 RMAl2 Noble Gas Monitor Grfx Rcdr 0 - 6 log cpm l CMORMA02 RMA2 Noble Gas Monitor Grfx Rcdr 0 - 6 log cpm CHORMA04 RMn4 Noble Gas Monitor Grfx Rcdr 0 - 6 log cpm
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- (] TEST NUMBER / TITLE: PTM1A2/ STEAM GENERATOR TUBE RUPTURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / ccritical Parameter DEVICE CHART SPEED N2:MI8 HPI Flow Al Grfx Rcdr 0 - 500 gpm
=N2:MI4 HPI Flow A2 Grfx Redr 0 - 500 gpm N2:MI2~ HPI flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 HPI Flow S2 Grfx Rcdr 0 - 500 gpm
~ F3:23FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:2?fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow S/G B Grfx Redr 0 - 1000 gpm n
f) v .
{ PERFORMANCE TEST NUMBER: PIMIA2 DATE: 07/16/91 TITLE: STEAM GENERATOR 10BE RUP1URE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _NONE _ a h' o Y/Q,i*/ TPC lesC N dinator t
/Dat i
i 2.0 All deficiencies have been resolved and this test has been completed, t~ .m 9!*Nf
/Date FPfTestCoordinhtor 3.0 Test results reviewed. Simulator performance is satisfactory.
w f~ m 9h/h (10 i N Y S I OperatidhsRepresentative[Date
/Date
_ Sip)dlatorInsttuctor_
,f ifal .. fy Simulator'Engyneer t./
.d l
Rf z/_'l_ L
/Date APPROVED: _ _Y M/f M DATE: _fyIgh jd IS/NL9)M/ '/
O
C TEST NUMBER / TITLE: PTMlB/ LETDOWN LINE LEAK DOWNSTREAM 0F MUV-49 ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(b)(c) DATE PERFORMED: 01/16/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100f. power when a 70 gpm leak is initiated in the letdown line downstream of MUV-49. Utilizing plant procedures, a plant shutdown is commenced and leak isolation is accomplished in accordance with OP-301. MALFUNCTION IIILI OPTIONS
, MU-007 Letdown Line Leak Adjustable from 0 - 1000 gpm INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service.
FINAL CONDITIONS: Plant shutdown In progress, letdown isolated and leak terminated. PLANT PROCEDURES USED DURING THE TEST: OP-204 REV 53 POWER OPERATIONS OP-301 REV 39 OPERATION OF THE REACTOR COOLANT SYSTEM C' BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on 'xpert iudgement. I
DATE: 01/16/91 Q PERFORMANCE TEST NUMBER: PTMlB , TITLE: LET00WN LINE LEAK DOWNSTREAM 0F MUV-49 TEST RESULTS/ COMMENTS: 1 Test results evaluation based on expert judgement. Test results satisf actory. One (1) Trouble Report written on improper response of radiation monitor RM-G13. TR-002 closed 07/05/91. l 1
)
O O
l
] TEST NUMBER / TITLE: PTMlB/ LETDOWN LINE LEAK DOWNSTREAM 0F MUV-49 TEST DATA LABEL DESCRIPTION RECORDlHG RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi N2JPX202 Letdown Flow Grfx Rcdr 0 160 gpm N2:M10 Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Injection Flow Grfx Rcdr 0 - 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe i Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F l
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F l
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in l
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in !
*LlJPS284 SG A Op Level Grfx Rcdr 0 - 100% i
*LlJPS336 SG B Op level Grfx Rcdr 0 - 100% ;
hg *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 p-i '
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 pci
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5GSAV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm CM:10A02 RMA2 lodine Monitor Grfx Rcdr 0 - 6 log cpm CM:N0A02 RMA2 Gas Monitor Grfx Rcdr 0 - 6 log cpm i CM:PAA02 RMA2 Particulate Monitor Grfx Rcdr 0 - 6 log cpm
- CM
- N0A08 RMA8 Gas Monitor Grfx Rcdr 0 - 6 log cpm l
! O lV l
PERFORMANCE TEST NUMBER: PTMlB DATE: 01/16/91
-]
TITLE: LETDOWN LINE LEAK DOWNSTREAM 0F MUV-49 TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): l l , TR # Priority Date Issued Date Closed i _TR0002 __C _01/16/91 _ _07/05/91_ j f4PC~Tesf' Coordinator lf
/Date ,
2.0 All deficiencies have been resolved and this test has been completed. l
-O ,GM) gs l [ PC Test Coordinator / Oste 3.0 Test results reviewed. Simulator performance is satisfactory.
l r J Dhb \$ lw Y % 2 T-9 I
/dmularor Instructor '/Date Operati0ns Representative /Date
, ' gj ^
;j 4 / ,. j },/till Simulat6r Engipeer '/Date I
( Ig cd M l i"\ DATE: 4/il/4( APPROVED: .NSTS/NLOTM'
\j l
O I I I
d C TEST NUMBER / TITLE: PTMIC1/REPRESSURIZING SMALL 8REAK LOCA INSIDE CONTAINMENT ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(b)(c) DATE PERFORMED: 08/11/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at 100% power when a .01 ft2 break is initiated at the discharge of the D Reactor Coolant Pump. When HPI initiates, the C makeup pump trips and the A makeup pump fails to start, limiting HPI flow to that delivered by the B makeup pump. RCPs are tripped when subcooling mayin is lost; due to inadequate HPl flow and resultant RCS voiding, single phase natural circulation is interrupted, causing the RCS to begin heating up and repressurizing. As OTSG levels are raised to 95% on the operating range, reflux boiling is established and a cooldown of the RCS is established. HALFUNCTION TITLE ;fmHS RC-002 RCS Cold leg Break Adjustable from 0 to 8.55 ft2 Selectabir to A, B, C, or D RC Pump discharce TFBV21C MVP-lC Shaft Siezure TFB2BIA MVP-1A Bkr Fails Open Q INITIAL CONDITIONS: Full power steady state conditions, all p'. ant equipment in service. FINAL CONDITIONS: Saturated plant cooldown to mode 5 in progress. PL"NT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-380 REV 17 ENGINEERED SAFEGUARDS ACTUATION l AP 530 REV 8 NATURAL CIRCULATION I l BASELINE DATA: CR-3 AT0G, PART II, Volume 2, Section F - Loss of Coolant Accidents 1 - O
PERFORMANCE TEST NUMBER: PTMICl DATE: 08/11/91 TITLE: REPRESSURIZING SMALL BREAK LOCA INSIDE CONTAINMENT TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. During the test, deviations between simulator response and the baseline data contained in the AT0G documents were noted, primarily in the area of the times required to achieve certain conditions (time to trip, time to safeguards actuation, time to interruption of natural circulation, etc.) and the magnitude of the saturated heatup and repressurization following the loss of natural circulation flow due to hot leg voiding. These deviations are based on differences between the simulator and ATOG assumptions on the amount of post trip decay heat and RCS inventory at the start of the transient. A greater initial simulator RCS inventory (pressurizer level at 220" vs 180") and lower simulator decay heat result in longer durations to key events and a smaller heatup and respressurization af ter the RCS becomes de-coupled from the OTSGs. O O
O TEST NUMBER / TITLE: PIMIC1/REPRESSUR121NG SMALL BREAK LOCA INSIDE CONTAINMENT U TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1,875 see HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 -2500 psi R3JPP208 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP209 Power _ Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125%
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F *LlJPR213 Th Loop-B Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +10F *LlJPR220 Tc Cold leg Al Grfx Rcdr 50 - 650F *L1JPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPR221 Tc Cold leg B1 Grfx Rcdr 50 - 650F *LlJPR217 Tc Cold Leg B2 Grfx Rcdr 520 - 620F -LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% p 0 - 1200 psi V F3JPA323 SG-A Press (EFIC) Grfx Rcdr F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LIJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R4JSA019 Incore T/C Grfx Rcdr 0 - 600F R4JSA020 Incore T/C Grfx Rcdr 0 - 600F R4JSA023 Incore T/C Grfx Rcdr 0 - 600F R4JSA025 Incore T/C Grfx Rcdr 0 - 600F Ll:RC6TM Delta Tcold Grfx Rcdr + 10F L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR
*ClJPP254 Containment Pressure Grfx Rcdr 0 - 70 psi *CATAVG Avg Containment Temperature Grfx Rcdr 50 - 300F CALNS Normal Sump Level Grfx Rcdr 0 - 10 FT CALES Emergency Sump Level Grfx Rcdr 0 - 10 FT O
. . .. - . =.. _. _- _ _ - . .
{} TEST NUMBER / TITLE: PTMIC1/REPRESSURIZING SMALL BREAK LOCA INSIDE CONTAINMENT TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED N2:MIB HPI Flow Al Grfx Rcdr 0 - 500 gpm N2:M14 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:M12 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:N16 HPI Flow B2 Grfx Rcdr 0 - 500 gpm F3:21Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm t
F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW-Flow S/G B Grfx Rcdr 0 - 1000 gpm O V f). u
h PERFORMANCE TEST NUMBER: PIMICl DATE: 08/11/91 TITLE: REPRESSURIZING SMALL BREAK LOCA INSIDE CONTAINMENT TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1 TR # Priority Date Issued Date Closed _NONE l l l l l
, I Jth??.
,f9Cfest"CooFdinator
[
/b[ ate $'
2.0 All deficiencies have been resolved and this test has been completed. n
.)
___NGI_Jn;0n11nat l FPC Test Coordinator /Date t i 3.0 Test results reviewed. Simulator performance is satisfactory. f.~
.fh/'Nh'4 SD \ &
l fd i T- L ) ',[__ l /Simulat6r IIstructor /Date Dperation Representative /Date j .. l
- }'j .
.;,.l.,,
STiiiulatoilngineer 70 ate t i , l APPROVED: _ i 1. - L.- DATE: Sig5g_ NSTS[NLOTM .} 9 l l l
TEST NUMBER / TITLE: PlMIC2/SMALL BREAK LOCA INSIDE CONTAINMEN1 ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(b)(c) DATE PERFORMED: 08/13//91 TEST DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating at 100% power when a .04 f t2 break is initiated at the discharge of the A Reactor Coolant Pump. RCS pressure initially de ceases rapidly, stabilizes at Psat of the steam generators, then begins decreasing as a result of reflux cooling, break flow cooling and decreasing decay heat. As the simulator continues to cooldown and depressurize the core flood tanks empty and low pressure injection is actuated. MALFUNQT LON TITLE OHLO!!S RC-002 RCS Cold Leg Break Adjustable from 0 to 8.55 f t2 Selectable to A, B, C, or D RC Pump discharge INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Plant cooldown to mode 5 in progress using the OTSGs in a reflux boiling mode. HPl and LPI in progress.
/ PLANT PROCEDURES USED DURING THE TEST:
AP-580 REV 17 REACTOR TRIP AP-380 REV 15 ENGINEERED SAFEGUARDS ACTUATION BASELINE DATA: CR-3 ATOG, PART 11, Volume 2, Section F - Loss of Coolant Accidents O
\
-Q PERFORMANCE TEST NUMBER: PTMIC2 DATE: 08/13/91 TITLE: SMALL BREAK LOCA INSIDE CONTAINMENl lTESTRESULTS/ COMMENTS:
Test results staisfactory. No performance problems noted. Note: Hot leg level indication response minimal during the test. Design of the system precludes accurate indication of levels with flow present in the loops. Due to the flow present because of reflux cooling, level indication (by design) was non-responsive. s
- During the test, deviations between simulator response and the baseline data contained in the ATOG documents were noted in the area of the times required to achieve specific
'; transient milestones (time to trip, time to safeguards actuation, and the amount of time before a slow RCS depressurization begins due to decreasing decay heat). These deviations are based on differences between the simulator and A10G assumptions on the amount of post trip decay heat and RCS inventory at the start of the transient. A greater initial simulator RCS inventory (pressurizer level at 220" vs 180") and lower simulator decay heat result in slower durations to certain key events and a shorter time until depressurization begins due to decreasing decay heat. O i i r O
l k TEST NUMBER /TITLEt PTMIC2/SMALL BREAK LOCA INSIDE CONTAINMENT TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter CHART SPEED DEVICE _
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi R3JPP208 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP209 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% - R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% '
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*LlJPR213 Th loop-B Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +120F
*LlJPR220 Tc Cold Leg Al Grfx Rcdr 50 - 650F -
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPR221 Tc Cold Leg B1 Grfx Rcdr 50 - 650F
*L1JPR217 Tc Co'.d leg B2 Grfx Rcdr 520 - 620F LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in llJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op ' evel Grfx Rcdr 0 - 100%
O tiJeS336 SG-8 oP tevel F3JPA323 SG-A Press (EFIC) Grrx Rcdr Grfx Rcdr o - 100% 0 - 1200 psi F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 1200 psi LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR ' LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR - LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R4JSA019 incore T/C Grfx Rcdr 0 - 2500F R4JSA020 Incore T/C Grfx Rcdr 0 - 2500F R4JSA023 Incore T/C Grfx Rcdr 0 - 2500F R4JSA025 Incore T/C Grfx Rcdr 0 - 2500F g Ll:RC6TM Delta Tcold Grfx Rcdr 4 10F L4JPR234 RCS Leop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR
*ClJPP254 Containment Pressure Grfx Rcdr 0 - 70 psi
*CATAVG Avg Containment Temperature Grfx Rcdr 50 - 300F i CALNS Normal Sump Level Grfx Rcdr 0 - 10 iT CALES Emergency Sump Level Grfx Rcdr 0 - 10 FT p
a m i
TEST NUMBER / TITLE: PTMIC2/SMALL BREAK-LOCA INSIDE CONTAINMENT TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED N2:M18 HPI Flow Al Grfx Rcdr 0 - 500 gpm N2:M14 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:M12 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 HPl Flow B2 Grfx Rcdr 0 - 500 gpm F3:23FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm CH:RMG16 Containment Radiation Grfx Rcdr 1 - 5 log MR/HR H14ALO Reactor Vessel Level Grfx Rcdr 0 - 171 in H14BLO Reactor Vessel Level Grfx Rcdr 0 - 171 in H19TE0 Surge Line Temperature Grfx Rcdr 100 - 700F O
01 .
C PERFORMANCE TEST NUMBER: P1 HIC 2 DATE: 08/13/91 TITLE: SMALL BREAK LOCf INSIDE CONTAINMENT TEST STATUS: 1.0 Tais test is complete and har the following deficiencies noted (if none so indicate): TR # Priority Date issued Data Closed
.JDlif
? b jkTea'tCoordinator g /Date 2.0 All deficiencies have been resolved and this test has been completed.
i^ NOT REQUIRED FPC Test Coordinator /Date 3.0 . Test results reviewed. Simulator performance is satisfactory. _ k N / / MT ' b2MI glinulpfor liistructor 40 ate Operat% N epresentative/Date
'))' ;.(y
~
c/ sjk,' ., l l
'ti; q _
Simulator Engineer /Date
,g APPROVED: i id ht- t N S/NLOTM' m DATE: gulE x
O
O TEST NUMBER / TITLE: PIMIC3/SMALL BRE" LOCA WITH LOSS OF HPI ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(c)(d) DATE PERFORMED: 08/1?/991 TES1 DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating at 100% power when a .02 f t2 break is initiated on the discharge of the A RCP. As a result of the LOCA the plant trips and begins depressurizing. RC Pumps are secured, pressure c'ecreases to the HPI actuation setpoint and HP1 actuates, Due to multiple mechanical failures, all HPl pumps are lost. As RCS inventory decreases, natural circulation is interrupted, cooling is lost and the RCS begins to heatup and repressurize along the saturation line. Eventually, core uncovery takes place, resulting in indications of superheating of the RCS. Once inadequate core cooling indications are reached, two MU pumps are restored to service, HPl is started and core cooling is established, MALFUNCTION TITLE q.PTIONS TFB2TBl_ Makeup Pump Bkr Fails Open Selectable to MVP-1A, -1B and/or -1C INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: RCS inventory restored, core cooling established using HPl. PLANT PROCEDURES USED DURING THE TEST: EP-290, REVISION INADEQUATE CORE COOLING AP-380, REVISION ENGINEERED SAFEGUARDS ACTUATION BASELINE DATA: Expected simulator response based on a combination of expert ju.dgement and the following information: CR-3 ATOG, PART 11, Volume 2, Section F - Loss of Coolant Accidents NSAC-1 Revised March 1980 - Analysis of Three Mile Island - Unit 2 Accident O i _ _ _ . - _ . . _ _ _ . - _ _ _ _ _ _ ~ _ , . . _ _ _ _ . -
4 PERFORMANCE TEST NUMBER: PTMIC3 DATE: 08/12/91 (d TITLE: SMALL BREAK LOCA WITH LOSS Of HPI TEST.RESULTS/ COMMENTS: Test results evaluation based on expert judgment extrapolation of baseline data from ATOG and the Three Mile Island Accident. Test results satisfactory. No performance problems noted. Note: Reactor vessel level indication on the Main Control Board did not provide meaningful information during this transient. This sptem has severe design limitations and has recently undergone a significant recalibration in the plant which has not yet been incorporated on the simulator. During the transient, vessel level was tracked on CTS by monitoring reactor vessel nodal void fractions; the core was un: overed approximately 50% at the time that HPl was initiated. t O O
r' TEST NUMBER / TITLE: PTHIC3/SMALL BREAK LOCA WITH LOSS OF HPl
-O V TEST DATA LABEL DESCRIPTION- RECORDING RANGE FREQUENCY /
- Critical Parameter DEYlCE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F
- HISTED Surge Line Temperature Grfx Rcdr 100 - 700F HCTAVG- RCS Average Temperature Grfx Rcdr 520 - 620F
*LITMARG4 Subcooling Margin Grfx Rcdr- 120F LlJPR214 Tc Cold Leg Al- Grfx Rcdr 520 - 620F LldPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F R4JSA019 Incore T/C Grfx Rcdr 0 - 2500F R4JSA020 Incore T/C Grfx Rcdr 0 - 2500F L1JPS287 SG-A SU Level Grfx Rcdr 0 - 250 in LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
LIJPS336 SG-B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG-A Press (EFICl Grfx Rcdr 0 - 1200 psi
-F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi LlJPS301 SG- A MFt' Flow Grfx Rcdr 0 - 6K KLB/HR
( LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 ML3/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*H14ALO Reactor Vessel level Grfx Rcdr 0 - 171 in
*H14BLO- Reactor Vessel Level Grfx Rcdr 0 - 171 in '
*L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR
*L4JPR245 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR
*R3JPP202 Source Range CPS Grfx Rcdr -1.0 - 6.0 cps
*R3JPP203 Source Range CPS Grfx Rcdr -1.0 - 6.0 cps R3JPP212 Intermediate Range Amps Grfx Rcdr -11.0 - -3.0 amps R3JPP213 Intermediate Range Amps Grfx Rcdr -11.0 -3.0 amps ClJPP254 Containment Pressure Grfx Rcdr 0 - 70 psi CATAVG Avg Containment Temperature Grfx Rcdr 50 - 300F CALNS Normal Sump Level Grfx Rcdr 0 - 10 FT CALES Emergency Sump Level Grfx Rcdr 0 -10 FT HKLLIQ RCDT Level Grfx Rcdr 0 - 15 ft HKPGAS RCDT Pressure Grfx Rcdr 0 - 200 psi l' *HKWPRZ Flow from Pzr to RCDT Grfx Rcdr 0 - 60 LB/SEC H17E30 RCV-10 Discharge Temp Grfx Rcdr 100 - 300F
-N2:M18 HPI- Flow Al Grfx Rcdr 0 - 500 gpm
.N2:M14 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:MI2 HPI-Flow B1 Grfx Rcdr 0 - 500 gpm N2:HI6 HPI Flow B2 Grfx Rcdr 0 - 500 gpm O
i O TEST NUMBER / TITLE: PTMIC3/SMALL BREAK LOCA WITH LOSS OF HP! V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Phrameter DEVICE CHART SPEED F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm I
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t PERFORMANCE TEST NUMBER: PTHIC3 DATE: 08/12/91 TITLE: SMALL BREAK LOCA WITH LOSS OF HPl l TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _NONE
\
A C'16&f Coordinator /Date r 2.0 All deficiencies have been resolved and this test has been completed. O NOT REQUIRED TPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
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,/Simulattfr Instructor /Date Operationt Representative /Date J/ /
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/Date APPROVED: c-< i WSm DATE:
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4 l TEST NUMBER / TITLE: PTMID/ PRESSURIZER SAFETY VALVE FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(1)(d) DATE PERFORMED: 01/14/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at 100% power when a pressurizer safety valve fails open. A Reactor / Turbine trip occu s, RC Pumps are secured when subcooling margin is lost, and High Pressure Injection is actuated. Due to the size and location of the leak, the plant goes to saturated conditions and continues to slowly cool down and depressurize due to break flow. MALFUNCIIDS IJJLE Qf]_LQld RC-006 Safety Valve Failure Selectable to RCV-8, -9 Severity level from closed to full open INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, RC Pumps secured, HPl actuated. Saturated cooldown in progress. O PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-380 REV 17 ENGINEERED SAFEGUARDS ACTUATION BASELINE DATA: CR-3 ATOG, PART 11, Volume 2, Section F - Loss of Coolant Accidents O
O gs PERFORMANCE TEST NUMBER: PIMID DATE: 01/14/91 V' TITLE: PRESSURIZER SAFETY VALVE FAltVRE TEST RESULTS/ COMMENTS: Test results satisfactory. No performance proulems noted. Excellent correlation between ATOG data and simulator response.
1 i 1'
\
.mr .
l g TEST NUMBER / TITLE: PTM U/ PRESSURIZER SAFETY VALVE FAILURE D TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi RRWSTP Normalized Reactor Power Grfx Rcdr 0 - 125%
L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR
*L1JPR212 Th Loop A Grfx Rcdr 520 - 6LOF
*L1JPR213 Th loop-B Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +10F
*LlJPR220 Tc Cold Leg Al Grfx Rcdr 50 - 650F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPR221 Tc Cold Leg B1 Grfx Rcdr 50 650F
*LlJPR217 Tc Cold Leg B2 Grfx Rcdr 520 - 620F LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in L1JPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% p F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi V F3JPA320 SG-B Press (EFIC) LlJPT228 MS Loop A Hdr Press Grfx Rcdr Grfx Rcdr 0 - 1200 pst 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 1200 psi L1JPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR R4JSA019 Incore T/C Grfx Rcdr 0 - 600F R4JSA020 Incore T/C Grfx Rcdr_ 0 - 600F R4JSA023 Incore T/C Grfx Rcdr C 600F R4JSA025 Incore T/C Grfx Rcdr 0 - 600F
*ClJPP254 Containment Pressure Grfx Rcdr 0 - 70 psi
*CATAVG Avg Containment Temperature Grfx Rcdr 50 - 300F CALNS Normal Sump Level Grfx Rcdr 0 - 10 FT l CALES Emergency Sump Level Grfx Rcdr 0 - 10 FT N2:MI8 HPI Flow Al Grfx Rcdr 0 - 500 gpm N2:MI4- HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:MI2 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 HPI Flow B2 Grfx Rcdr 0 - 500 gpm l F3:23FIl EFK Flow S/G B Grfx Rcdr 0 - 1000 gpm l
F3:25FIl EFW-Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm l F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm 1 - O
i 7 TEST NUMBER / TITLE: PTMID/ PRESSURIZER SAFETY VALVE FAILURE V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED R3JPP202 Source Range CPS Grfx Rcdr -1.0 - 6.0 log cps !
R3JPP203 Source Range CPS Grfx Rcdr -1.0 - 6.0 log cps ! R3JPP212 Intermediate Range Amps Grfx Rcdr -11.0 - -3,0 log amps R3JPP213 Intermediate Range Amps Grfx Rcdr -11.0 - -3.0 log amps HKLLIQ RCDT Level Grfx Rcdr 0 - 15 ft HKPGAS RCDT Pressure Gifx Rcdr 0 - 200 psig
*HKWPRZ Flow from Pzr to RCDT Grfx Rcdr 0 - 150 lb/sec H17E10 RCV-8 Discharge Temperature Grfx Rcdr 100 - 300 F r
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PERFORMANCE TEST NUMBER: PTMID DATE: 01/14/91 TITLE: PRESSURIZER SAFETY VALVE FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _NONE _
,n ,
Tgit 06ordinator //date 2.0 All deficiencies have been resolved and this test has been completed. O 's/ NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. JAJ / edg(
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OperationsRepresentative/Date Y- L M/ 4
/Simdiat6rlostructor
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7/$ ! / 7/ Simulator Enginpe'r /Date t , APPROVED- 'kyh u-c n DATE: 'i n iN~ fists /NLOTiff-
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3 i TEST NUMBER / TITLE: PTM2/ LOSS OF INSTRUMENT AIR O- ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(2) DATE PERFORMED: 01/17/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at 100% power when a non-isolable air leak develops on the instrument air header. Since the leak rate is larger than the capacity of the available air l compressors, IA pres:ure begins decreasing; as supplied equipment is lost the reactor is tripped, the plant is ! stabilized in accordance with applicable plant procedures and a cooldown is commenced. MALFUNCTION TITLE OPTIONS 1A-001 Non-isolable Instrument Air Adjustable from 0 - 2600 scfm Header leak l Note: Through the 1/F, isolable leaks may also be initiated in the turbine and auxiliary building IA headers and the turbine building SA header. INITIAL CONDITIONS: Full power steady state conditions. Berm Air Compressor ID secured for maintenance. 01 ot' r plant equipment in service. O FINAL CONDITIONS: Reactor / turbine tripped, plant cooldown in progress on V emergency feedwater and atmospheric dump valves. Instrument air pressure at 0 psig. Affected equipment secured as required. PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-470 REV 1 LOSS OF INSTRUMENT AIR OP-209 REV 73 PLANT C00LDOWN BASELINE DATA: There is no plant data available for this transient. Expected results based on expert judgement. The proper failure mode for every simulated air-operated valve was verified during factory acceptance testing of the simulator. The purpose of this test is to demonstrate integrated plant response to a loss of air and the ability to utilize tne applicable plant procedures to respond to this transient. O
4 PERFORMANCE TEST NUMBER: PTM2 DATE: 01/17/91 TITLE: LOSS OF INSTRUMENT AIR TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. Note: Four simulator scope limitations were identified during the conduct of this test. Two of these limitations are transparent to the operators. The other two require the addition of two valves to the air system model. One of these valves (IAV-10) has been added. Until the second valve (IAV-27) is added to the scope of simulation, the procedural step calling for closure of this valve is being satisfied by closing other valves in the system. This procedural deviation is transparent to the operators and results in the same ; indications / response as if IAV-27 had been closed. O O
i TEST NUMBER / TITLE: PTM2/ LOSS OF INSTRUMENT AIR _) TEST DATA LABEL DESCRIPTION PECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED <
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5,625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR22i RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi '
'L6JPR208RCSPressure(WR) Grfx Redr 0 2500 psi N2JPX202 Letdown Flow Grfx Rcdr 0 160 gpm JPPAM711 Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Injection Flow Grfx Rcdr 0 - 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 50F L1:6BDT delta T Loop B Grfx Rcdr 0 50F
*LldPR212 Th Loop A c,rfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature .rfx Rcdr 520 - 620F
*LlJPR214 Tc Cold leg Al Grfx Rcdr 520 - 620F
*LlJFR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG A Op Level Grfx Rcdr 0 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 100%
s *F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi y *F3JPA320 SG-B Press (EFIG) Grfx Rcdr 0 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR (
*LlJPS302 SG-P MFW Flow Grfx Redr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU FW Flow Grfx Rcdr 0 1500 MLB/HR KFP03 Instrument Air Hdr Pressure Grfx Rcdr 0 - 150 psi L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5GbAV Grp 6 Avg Position Grfx Rcdr 0 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 100% i l \ I O l
DATE: 01/17/91 Q PERFORMANCE TEST NUMBER: PIM2 TITLE: LOSS OF INSTRUMENT AIR TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date Issued Date Closed REk___.__ __ _ ______ e +me --#M.PS e .
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=-
1 ..- l fj ! ) L lll7/Ql ftl est Coo Minator /Date 2.0 All deficiencies have been resolved and this test has been completed. O ,n i /,. a r: 1r TCTest CrioWInator 7Dit'e 340 lest results riviewed. Simulator performance is satisfactory. a,f'h5-[r, n p,s a ., g, kliiiulator Instructor /0 ate clieFiiriETRepresedit ive7Dile s , A/ , 5 f$r'[ng hkN APPROVED DATE: p(( O
--.-.---..-.,______.m _.,,___m__,_ __ _ _ _ _ _
s, i TEST NUMBER / TITLE: PTH3A/ LOSS OF OFFSITE POWER (LOW DECAY HEAT) ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORNED: 01/23/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 12% power and a main turbine startup in progress, a loss of the station startup transformer occurs, resulting in a loss of offsite power. RCPs and MFW pumps are lost; both EDGs start and supply power to the ES 4160V busses. EFW is initiated and begins feeding both OTSGs to 65% on the operating range level. Due to leakage past the TBVs, high steam flows through several steam trap bypass valves and high initial emergency feedwater flow rates, an excessive RCS cooldown and depressurization of the OTSGs to app m ximately 600 psig occurs. At approximately two minutes into thi transient, emergency feedwater is placed in hand and throttle at approximately 12 minutes into the transient the H!ii; are closed. Due to the extremely low EFW feed rates ano little or no steaming of the OTSGs after closure of the HSIVs, natural circulation is interrupted in the A OTSG. HALFUNCTIOE UILE OPTIONS ED 001 Loss of S/U Transformer 7 INITIAL CONDITIONS: 12% power, low decay heat, main turbine startup in progress. i _. FINAL CONDITIONS: Reactor / turbine tripped, main feedwater pumps and RCPs tripped. HSIVs closed, 01SG 1evels being fed up to 65% at a rate of approximately 100 'm/ generator by emergency feedwater being controlled in manua). Natural circulation in progress on the B ! OTSG. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: U0ER-89 0-02 - Reactor Trip and Emergency Feedwater Actuation Due to a Degradation of Off site Power. 06/16/89 RECALL DATA FILE 0061689 - Loss of Off site Power from 12% Reactor Power. 06/16/89 O
, _ _ - - . _ . . . _ . - _ _ . ~ . , _ . . _ _ . _ . . . . . _ , . . _ _ . . . . _ _ _ _ . - _ - - - -
O I -, PERFORMANCE TEST NUMBER: P1H3A DATE: 01/23/91 TITLE: LOSS Of OffSITE POWER (LOW DECAY HEA1) TEST RESULTS/ COMMENTS: Test results satisfactory No performance problems noted. Excellent correlation between plant data and simulator. O O
_ _ __ _ _ _ . _ - _ _ _ _ . _ _ ~ . _ - - .- _. _ - . _ _ __ _ __ _ O nwnau w a m-. - TEST NUMBER / TITLE: PIM3A/ LOSS Of Off@gjjfx (LOW bfCAY HEA1) TEST DATA LABEL DESCRIPTION RfCORDING RANGE FREQUENCY /
' Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 3.75 sec HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 -2500 psi R3JPP200 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP209 Power Range Indication Grfx Rcdr 0 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Inoication Grix Rcdr 0 125%
*LlJPR212 Th Loop A Grfx Rcdr 520 620F
*LlJPR213 Th loop B Grfx Redr 520 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 620f LITMARG4 Subcooling Margin Grfx Rcdr 10 - +120f
*LlJPR220 Tc Cold leg Al Grfx Rcdr 50 - 650f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f
*LlJPR221 Tc Cold leg B1 Grfx Rcdr 50 - 650f
*LIJPR217 Tc Cold leg B2 Grfx Rcdr 520 620f LlJPS2P7 SG A SU Level .Grfx Rcdr 0 - 250 in LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
LlJPS336 SG B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG A Press (EflC) Grfx Rcdr 0 1200 psi O, f3JPA320 SG B Press (EflC) Grfx Rcdr 0 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi LlJPS301 SG-A MfW flow Grfx Rcdr 0 6K KLB/HR LlJPS302 SG B MfW flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW flow Grfx Rcdr 0 - 1500 MLB/HR LIJPS330 SG-B SU FW flow Grfx Rcdr 0 - 1500 MLB/HR Ll:RC6TH Delta Tcold Grfx Rcdr +10 f L4JPR234 RCS toop A flow Grfx Rcdr 0 - 80 MLD/HR L4JPR235 RCS Loop B flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total flow Grfx Rcdr 0 160 MLB/HR f3:23N i EfW flow S/G B Grfx Rcdr 0 1000 gpm f3:25fil EfW flow S/G A Grfx Rcdr 0 1000 gpm F3:26fil EfW flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24fil EfW flow S/G B Grfx Rcdr 0 - 1000 gpm B O
~ ,~ . -, . - . . . , - , . - _ . - - -- , . - - - - - , - .
O etarona^NCE TEST NUMBER: PIM3A DATE: 01/23/91 T11LE: LOSS OF 0FFSITE POWER (LOW DECAY HEAT) TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _.f&lR _ - .
/
(FClestCoordinaorLhhk /Date / kbi 2.0 All deficiencies have been resolved and this test has been completed. O ' TPNeYhoN[inator /Date 3.0 Test results reviewed, Simulator performance is satisfactory.
~
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t(' l 7 ^ l Simulator Instructof /Da'Ie / Operat ns kepresentative/Date
// f 9b7/11 SimulatIEnginest ^'/Date APPROVED- / DATE:
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TEST NUMBER / TITLE: PTM3Bl/ LOSS OF EMERGENCY Off-SITE SOURCE ANSI /ANS-3.S(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/24/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when an electricel fault causes a loss of the emergency off site power source. Power is lost to the 4160 VAC ES A and B buses. The emergency EDGs automatically start on undervoltage and restore power to the ES buses. A makeup pump is restarted and the power to the A 4160 VAC ES bus is transferred from the A EDG to the Startup Transformer. MLEUNCTI0B TITLE OPTIONS TCE2DUM1 feeder Breaker from Unit 1&2 Startup Transformer Rack in or Out INITIAL C0h0!T10NS: Full power steady state conditions, all plant equipment in service. F!hAL CONDITIONS: Full power steady state conditions with the 4160 VAC ES A bus being supplied from the startup transformer and the 4160 MC ES B bus supplied by EDG B. PLANT PROCEDURES USED DURING THE TEST: OP 700A REV 6 6900, 4160 AND 480 VAC BUSES AP-770 REV 13 EMERGENCY DIESEL GENERATOR ACTUATION BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert , judgement and the ability to restore power using applicable plant procedures. l l lO I e
,-- - . - . , - , - . , , , ~ , . .. - - - - - . - . - . , - . . - -
0 PERFORMANCE TEST NUMBER: PIM301 DATE: 01/24/91 O TITLE: LOSS Of EMERGENCY OFF-SITE SOURCE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and the ability to utilize 1 applicable plant procedures. Test results satisf actory. No performance problems noted.
~
O
0 TEST NUMBER / TITLE: PTH3Bl/ LOSS OF EMERGENCY Off-SITE SOURCE g V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT
\
I O
O v PERFORMANCE TEST NUMBER: PTH381 DATE: 01/24/91 TITLE: LOSS OF EMERGENCY Off SITE SOURCE TEST STATUSr 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed l _l!QE
, , . ]
, I th /*
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PC Test Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. O- IJOT HI:QUllmD Fit Test Coordinator 7Date l 3.0 Tut results reviewed, Simulator performance is satisfactory.
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.i/,ki.nier Tilor E /0 ate APPROVED DA1E: Jd,!_
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s, _ r TEST NUMBER / TITLE: PTH382/ LOSS OF EMERGENCY OFF-SITE SOURCE WITH A FAILURE OF 1 EDG () ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/24/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when an electrical fault causes a loss of the emergency off-site power source. Power is lost to the 4160 VAC ES A and B b'4ses. The B emergency EDG automatically starts on undervoltage tnd restores power to the B 4160 ES bus; the A EDG fails to start. A makeup pump is started and power to the A 4160 VAC ES bus is restored from the Startup Transformer. MALFUNCTION IllLE OPTIONS TCE2DUM1 Feeder Breaker from Unit 1&2 S/U Transformer - Rack in or Out EG-002 Emergency Diesel Generator Trip Selectable to the A or B EDG INITIAL CONDITIONS: Full power steady state ennditions, all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions with the 4160 VAC ES A bus being supplied from the startup transformer and the 4160 VAC ES B bus supplied by EDG B. CN PLANT PROCEDURES USED DURING THE TEST: OP-700A REV 6 6900, 4160 AND 480 VAC BUSES AP-770 REV 13 EMERGENCY DIESEL GENERATOR ACTUATION l BASELINE DATA: There is no plant data for this transient. Expected simulator l response based on expert judgement and the ability to restore l power using applicable plant procedures. m.
s,, 4 PERFORMANCE TEST NUMBER: PTH302 DATE: 01/24/91 TITLE: LOSS OF EMERGENCY OFF-SITE SOURCE Wi1H A FAILURE OF 1 EDG TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and the ability to utilize applicable plant procedures. Test results satisfactory. No performance problems noted. O q .. V
-s, f
e-TEST NUMBER / TITLE: PIM382/ LOSS Of EMERGENCY Off 511E SOURCE WITH A FAILURE Of I LDG TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical parameter DEVICE CHART SPEED ANALOG DATA COLLECTION HOT REQUIRED FOR THIS TEST SIMULATOR RESP 0HE AhEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINT 0UT O
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g PERFORMANCE TEST NUMBER: PIM3B2 DATE: 01/24/94 TITLE: l0SS Of IMERGINCY Off-Sill SOURCl Wi1H A FAllVRI Of 1 (DG TEST STATUS: 1.0 This tes; is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Dite Closed
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9 j /y4', , G, , ' lllykl p,e -( /\i C, hKlillt UIoidT63f5r 70ats 2.0 All deficiencies have been resolved and this test has been completed. O s I. b !,..-PCUtest b :.fL Cobrd:inator. . - . , . , . . . . . ./Date 3.0 Test results reviewed. Simulator performance is satisfactory.
, l? 1('fY ff.27ljj ,
7 ,, sTm$ialtFGstructor /cate Operailind Repr esent at Iss/Dat 6
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vi, ly slinuTiliir TiiO'jIie~r~~ ~ 7l0at6 g ..;-? / APPROVlD: jfY,H:_,/;- DATE: fc,f*g
/
/515/ST(iM v
o () . TEST NUMBER / TITLE: PTH3Cl LOSS OF NNI-Y 24 VDC p0WER ANSI /ANS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 07/31//91 I TEST DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating at 100% power when the NNI-Y 124 VOC power is lost. This causes numerous instrument failures on the main control board. MALFUNCTION 111LE OPTIONS NN 001 NNI DC Bus failure Selectable to NNI X or NNI-Y Bus INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power, all plant equipment in service. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator O response based on expert judgement and plant design data on power supply dependencies. O
0 PERFORMANCE TEST NUMBER: PTH3Cl DATE: 07/31/91 TITLE: LOSS OF NNI-Y 24VOC POWER TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and plant design data on power supply dependencies lest results satisfactory. Two (2) irouble Reports written on failure values on 1 recorder point (suspected hardware problem) and 1 PPC point. These TRs do not require a rerun of the test. TR-239 (recorder point problem) cleared 08/27/91 O
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O TEST NUMBER / TITLE: PlH3C1/ LOSS Of NNI-Y f 24 VDC POWER V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION HOT REQUIRED TOR THIS TEST SlHULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE O
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$ PERFORMANCE TEST DER: PIM3Cl DATE: 07/31/91 TITLE: LOSS Of NN1-Y 1 24 VDC POSER YEST STATUS:
1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _IE00H2__ __ C __ _0B209L91__ an/27 /<> i _1E0_0242 ._L_ 09409Z21_ . _ _ _ me - = m--am A wa-s. () 1 1-, ,- , f,j/, 4h./ , f 1
-p'h/[r (Cf/</ !
f PFCresiloWilinator /liat e 2.0 A1. deficiercies have been resolved and this test has been completed.
-l, j , aclouure or vn-240 doen not preven t f] ,e _/g f f 3, 4., compt et ion of t 1,o t ent i
[IC[Tes CooIdinator /~/ hit'e 3.0 Test results reviewed. Simulator performance is satisfactory.
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(.. .;1/ hTiElator 1~nstructor ~~70 ate F} ylll zGTp y.zuI operati6is~Representatlve/Date
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IimularirTEiReer ~/tGie 6%-421 OATE: _gy;_j._:f_ APPROVED: f g __ i
o TEST NUHBER/ TITLE: PIM3C2/ LOSS Of NNI-Y AC POWER ANSI /AtlS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/22/91 TEST DURATION: 1/2 hour TEST DESCRIPf!0N: With the simulator operating at 407, power, the flN1-Y 118 VAC field supply is lost. Following applicable plant procedures, the reactor is tripped and main feedwater pumps are secured. The plant is stabilized in the normal post trip window with OTSG levels being maintained by emergency feedwater. Affected instrumentation and controls are verified to reflect a loss of power condition. MAlFUNCT193 IllLE QEllqM NN-003 Loss of NNI AC Power Selectable to NNI-X or Y INITIAL CONDITIONS: 401,' power, both main feedwater pumps in operation with the main feedwater block valves closed. FINAL CONDITIONS: Reactor turbine tripped, simulator stabilized in the normal post trip envelope with OTSG 1evels being controlled by Emergency feedwater. Affected instrumentation and controls in their loss of power position. PLANT PROCEDURES USED DURING THE TEST: AP-580, REVISION 15 REACTOR TRIP AP-582, REVISION 1 LOSS Of f4N1 Y BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement and paint design data on power supply dependencies. 6 ,F*M W~ -% gh em i O
0 L PERFORMANCE TEST NUMBERt PlH302 DATE: 01/22/91 V TITLE: LOSS OF NNI-Y AC POWER TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and plant design data on power supply dependencies. Test results satisfactory. No performance problems noted. One procedure step (NNI-Y TROUBLE Alarm) not completed due to scope limitation , SSL-P1H3C2-1. The MAR associated with this limitation had been previously identified and will be incorporated in accordance with applicable requirements. O O
N, TEST NUMBER / TITLE: P1M3C2/ LOSS OF NN1-Y AC POWER TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi N2JPX202 Letdown Flow Grfx Rcdr 0 - 160 gpm N2:M10 Makeup Flow Grfx Rcdr 0 200 gpm N2: Mil Total Seal Inj Flow Grfx Rcdr 0 - 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 -125%
PGJPE202 Gross Generation Arfx Rcdr 0 1000 MWe L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th Loop-A Grfx Rcdr 520 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LIJPS284 SG A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 100%
'T *F3JPA323 SG A Pret.s (EFIC) Grfx Rcdr 0 1200 psi (V *F3JPA320 SG 8 Press (EFIC)
LlJPT228 MS Loop-A Hdr Press Grfx Rcdr Grfx Redr 0 - 1200 psi 600 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 6K KLB/HR
*LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG A SU Fli Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU Fb Flow Grfx Rcdr 0 - 1500 MLB/HR F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm
PERFORMANCE TEST NL'HBER: PIM3C2 DATE: 01/22/91 TITLE: LOSS Of NNI Y AC POWIR _ TEST STATUS: 1.0 lhis test is complete and has the following deficiencies noted (if none, so indicate). 1R # Priority Date issued Date Closed SSL.P_lHKh1 _.__ ...
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N>ETelil~Coordinitiir ~ "/Dite 2.0 All deficiencies have been resolved and this test has been completed. v 3[ ,, , Her.olutlon of thin neope 1 imitation p - / __ [-Cj - . . ,., f ,.p, doen not pr event. completion of thin t e n t.
,JiiC"iisEo~oird inifoF~ ~/D'a'th i.j3.0 lest results reviewed. Simulator performance is satisfactory.
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Simul atora- f'n:!] gineer $l6 / ate APPROVED: [r ,f7Q" f7.s _ __ _ DATE: . R.'gf /
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i ___ TEST HUMBER/ TITLE: PTM3Dl/ LOSS OF ICS DC POWER ANSI /ANS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/24/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 1007. power, all DC power is lost to the ICS. Due to the power loss, the Main Feedwater Pumps go to minimum speed. The resultant rapid reduction in main feedwater flow causes a reactor trip on high pressure. In accordance with applicable procedures, the main feedwater pumps 7re tripped and the plant is stabilized in the normal post trip window with OTSG levels being controlled by emergency feedwater and header pressure being controlled by the atmospheric dump valves at 1025 psig. MALFUNCTIO8 IJJ1[ OPTIONS 10-017 Loss of ICS Power Selectable to E4 DC power or 118 VAC ICS field power supply INITIAL 00',0lT10NS: Full power steady state conditions, all plant equipment in service. TiNAL CONDITIONS: Reactor / turbine tripped. Main feedwater pumps tripped. OTSG levels being controlled by emergency feedwater, header pressure O being controlled at 1025 psig by the atmospheric dump valves. PLANT PROCEDURES USED DURING THE TEST: AP-580, REVISION 15 REACTOR TRIP AP 450, REVISION 15 EMERGENCY FEEDWATER ACTUATION BASELINE DATA: There is no plant data for this transient. Expeded simulator response based on expert judgement and plant design data. O
~. ,
PERFORMANCE TEST NUMBERt PIM3D1 DATE: 01/24/91 TITLE: LOSS OF ICS DC POWER l TEST RESULTS/CONNENTS: Test results evaluation based on exprt judgement and plant design data. Test results ! satisfactory. No performance problems noted. O 1 l O w . - - - - - _.
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l i TEST NUMBER / TITLE: P1H3Dl/ LOSS OF ICS DC POWER TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*H1:ll3 Pressurizer level Grfx Rcdr 0 - 320 in 1.B75 sec HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 pst
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi LlJPR212 Th Loop A Grfx Rcdr 520 - 620F HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LIJPR214 Tc Cold leg Al Grfx Rcdr 520 620F
.LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F LlJPS287 SG A SU Level Grfx Rcdr 0 250 in L1JPS293 SG B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG A Op Level Grfx Rcdr 0 100%
LlJPS336 SG B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi F3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 pst LlJPT228 HS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi , LlJPT232 MS Loop-B Hdr Press Grix Rcdr 600 - 1200 psi LlJPS301 SG A HFW Flow Grfx Rcdr 0 - 6K KLB/HR LlJPS302 SG B HFW Flow Grfx Rcdr 0 - 6K KLB/HR e LIJPS329 SG-A SU FW Flow Grf.: Rcdr (g ) LlJPS330 SG-B SU FW Flow F3:23Fil EFW Flow S/G A Grfx Rcdr Grfx Rcdr 0 0 1500 HLB /HR 1500 HLD/HR 0 - 1000 gpm F3:25Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm
l DATE: 01/24/91 { PERFORMANCE TEST NUMBER: PIM3D1 TITLE: LOSS OF ICS DC POWER TEST STA1US: 1.0 1his test is complete and hat, the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
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. . _ _ _ - - ~ _
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kNgalg [ , 6 . ~t ~' ll'l 'if Tes M o M nator /D~a~le 2.0 All deficiencies have been resolved and this test has been completed. NOT REQUIRl:D TPC lest Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
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,) l ' l g/77,/7/ py3 3 y. , g SimuTitor Instructor /Date DpeFati N RepresentatTie/Dif5
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simulat*o T69 ipeer /dite APPROVED: _ d 'M C__ _ DATE: _jh/ _
/et SIS /N
f j ___ TEST NUMBER / TITLE: PTH302/ LOSS OF ICS AC POWER ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/24/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power,118 VAC field power is lost to the ICS, Due to the power loss, the Main feedwater Block Valves and Low Low Block Valves close. The resultant rapid reduction in main feedwater flow causes a reactor trip on high pressure. In accordance with applicable proctdures, the main feedwater pumps are tripped and the plant is stabilized in the normal post trip window with OTSG levels being controlled by emergency feedwater. MALFUNCTION 11J1E Qfilplil 10-017 Loss of ICS Power Selectabic to 24 DC power or 118 VAC ICS field power supply INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped. Main feedwater pumps tripped. OTSG levels being controlled by emergency feedwater. PLANT PROCEDURES USED DURING THE TEST: AP-580, REVISION 15 REACTOR TRIP AP-450, REVISION 15 EMERGENCY FEEDWATER ACTUATION BASELINE DATA: There is no plant data for this transient. Expected simulator iesponse based on expert judgement and plant design data.
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4 PERFORMANCE TEST NUMBER: PIM3D2 DATE: 01/24/91 TITLE: LOSS OF ICS AC POWER TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and plant design data. Test results satisfactory No performance problems noted. O 4
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. TEST NUMBER / TITLE: PTM302/ LOSS OF ICS AC POWER TEST DATA-i LABEL - DESCRIPTION RECORDING- RANGE FREQUENCY /
- Critical - Parameter DEVICE CHART SPEED L
! *HlJPR310 Pressurizer Level Grfx Rcdr 0 --320 in 1.875 sec
-HlJPR203': Pressurizer Temperature Grfx Rcdr 0 - 700F
[ *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi l *L6JPR208 RCS Pressure-(WR) Grfx Rcdr 0 - 2500 psi j LlJPR212 Th Loop A Grfx Rcdr 520 620F 4 HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LIJPR214 Tc Cold Leg Al- Gefx Rcdr 520 - 620F !. LlJPR215 Tc Cold Leg A2_ Grfx Redr 520 620F 4 LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in [ L1JPS293 SG-B SU Level' Grfx Rcdr. 0 - 250 in + -LlJPS284 SG A Op-Level- Grfx Rcdr 0 - 100% LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG-A Press (EFIC)- Crfx Rcdr 0 - 1200 psi
- F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200-psi o LlJPT228 MS_ Loop-A Hdr Press Grfx Redr 600'- 1200 psi ;
i LlJPT232 .MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi , _LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
-LlJPS302 SG-B MFW Flow Grfx Redr 0 - 6K KLB/HR LlJPS329 SG-A SU Ft ' >w Grfx Rcdr 0 - 1500 MLB/HR
,-w g LIJPS330 .SG B SU . tiow- Grfx Rcdr 0 - 1500 MLB/HR
. ,j F3:23FIlt - EFW Flow.S/G /L Grfx Rcdr' 0 .1000 gpm-F3:25F11 EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm r_
_F3:26FIl EFW Flow S/G A. Grfx Rcdr 0 --1000 gpm F3:24Fil-.EFW Flow S/G-B Grfx Rcdr 0 - 1000 gpm I t bo Q 4
PERFORMANCE TEST NUMBER: PIM302 DATE: 01/24/91 TITLE: l0SS OF ICS A_C POWER TEST STATUS: 1,0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _E'Jlf _ __
' xy )'
//2 y'f?l FPC Test Coordinator / bate 2,0 All deficiencies have been resolved and this test has been completed, O NOT REQUIRED FPC Test Coordinator /Date 5.0 Test results reviewed. Simulator performance is satisfactory.
/48 l e E ) lf &l Operatio
- y. q s f Representative /Date simulator Instructor /Date
/
4 . f/. [,. ),'[f ed 7/k ' 51mulator Engineer /Date n . A APPROVED M DATE: _,-[,7j_ o
g TEST NUMBER / TITLE: PTM3E/ LOSS OF VITAL BUS U ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 01/21/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator i.; operating at 100% power when vital bus VBDP 3 is lost. Operation at power continues with A RPS thannel tripped and power lost to indication, control and monitoring equipment supplied by VBDP-3. HALFUN6TIQB llILE OPTIONS ED-Oll loss of Vital Bus Selectable to VBDP-1 through -15 INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power steady conditions with loads supplied by VBDP-3 exhibiting their loss of power values / conditions. PLANT PROCEDURES USED DURING THE TEST: OP-700D REV 9 120 VAC VITAL BUSSES i l BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement, plant design data on power dependencies and vital bus loads listed in the referenced procedure. l l 1 l l
w, , O ed PERFORMANCE TEST NUMBER: PTM3E TITLE:' LOSS OF VITAL BUS DATE: 01/21/91 TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement, plant design data on power supply dependencies and vital bus loads listed in the referenced procedure. Four (4) Trouble Reports were generated during this test. Three (3) have been voided. TR-004 (ES Lights) corrected 03/20/91. Test results satisfactory. No performance problems noted with the exception of the items discussed above. O 4 h
-.- J
x,,
,' l g TEST NUMBER /TITLEr PTM3E/ LOSS OF VITAL BUS V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED B
ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINT 0UT (aD l l l v-
i l l l PERFORMANCE TEST NUMBER: PIM3E DATE: 01/21/91 TITLE: LOSS OF VITAL BUS 1 TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1 TR # Priority Date Issued Date Closed i _TR-004 _ L. _plLZ3291_ _03/20/91_ _TR-005 __[ 01/21/91 _YQJD_
._TR-006 _C _QJ22JZ21_ __10JD_
TR-003 C _01/21/91_ __10]D__ _ 1 /-. y l [ xl,' ll?l l'f FPC Test Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. E f Y5I,t Coordinator (?h ?fif 7Date 3.0 Test results reviewed. Simulator performance is satisfactory.
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Simulator Instructor g%' 'f a g 7; w , m/Date Operations 0 Representative
/Da j l ; $
SimuliitirEngineer
/ c,,',,/n 76 ate
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APPROVED: h [ DATE: l l __
p TEST NUMBER / TITLE: PTM3F/ LOSS OF 480 '!AC ES BUS A _ _ _ k ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(3) DATE PERFORMED: 07/31/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, a fault develops on the 3A 480 VAC ES Bus, resulting in an overcurrent trip of the normal bus feeder breaker. The fault is cleared and power is restored to the bus from its alternate supply. MALFUNCTION TITLE OPTION} ED-005 Loss of 480 VAC ES Bus Selectable to 480 VAC ES Bus A or B INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions, 480 VAC ES Bus 3A power restored from its alternate supply. PLANT PROCEDURES USED DURING THE TEST: ( Q OP-700A, REVISION 6 OP-7708, REVISION IS 6900, 4160 and 480 VAC BUSES 480 VAC MOTOR CONTROL CENTERS BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement and Bus /MCC loads listed in the referenced plant procedures. O G
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t PERFORMANCE TEST NUMBER: PTM3F DATE: 07/31/91 TITLE: LOSS OF 480 VAC ES BUS A TEST RESULTS/ COMMENTS: Test results evaluation ased on expert judgement and Bus /MCC loads listed in referenced plant procedures. Test results satisfactory. No performance problems noted, f3 V m r -- (V
'N ,
o TEST NUMBER / TITLE: PTH3F/ LOSS OF 480 VAC ES BUS A V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /-
*Criticti Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT o
b (}
i i l g PERFORMANCE TEST NUMBER: PIM3F DATE: 07/31/91 l TITLE: LOSS OF 480 VAC ES BUS A 1 TEST STATUS: 1.0 This test is complete and has the .~ollowing deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _NONE ?
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~
FPC' Tes't Coordiiator 2.0 All deficiencies have been resolved and this test has been completed. NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performanca is satisfactory. {
/
(simulatorInstructor h[7 l _\LD Y Q]^7hf Operationi) Representative /Date
/Date i) l l Q ,l Simulato/TnFneer /Date
, f'-
APPROVED: _, ['$ NST5/NinjH DATE: _ _r_
~ , /
s, e' TEST NUMBER / TITLE: PTM3G/ LOSS OF DC DISTRIBUTION ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3_.l.2(3) DATE PERFORMED: 08/12/01 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, a fault causes a loss of DC Distribution Panel DPDP-1B, with resultant loss of 250/125 VDC to plant components. MALFUNCTION IJILE OPTIONS ED-019 Loss of DC Distribution Selectable to DPDB-1A and/or DPDP-1B INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power conditions with the components identified in the test procedure deenergized due to a loss of DPDP 18. PLANT PROCEDURES USED DURING THE TEST: OP-700E REV 11 125/250 VOC DISTRIBUTION PANELS BASELINE DATA: There is no plant data for this transient. Expected simulator
- O's ' response based on expert judgement, plant design data on power supply dependencies and the DC distribution panel loads listed in the referenced procedure.
O
s,, 4 n PERFORMANCE TEST NUMBER: PTM3G DATE: 08/12/91 V TITLE: LOSS OF DC DISTRIBUTION TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement, plant design data on power supply dependencies and the DC distribution panel loads listed in the referenced procedure. Test results satisfactory. No performance problems noted. (D V O
S _ ,m. TEST NUMBER / TITLE: PTM3G/ LOSS OF DC DISTR 1530T10N I V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Paramete* DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT r~%
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hI PERFORMANCE TEST NUMBER: PTM3G DAT : 08/12/91 TITLE: LOSS OF DC D'STRIBUTION TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none ; so indicate): TR # Priority Date Issued Date Closed _NONE l l m
/7 )
2 /$ Y FPC lest/ coordinator
$b2 (
/Date l
l l 2.0 All deficiencies have been resolved and this test has been completed. o NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. k,CAX') ek)k GT ww sI
$imulatpr Instructor /Date OperatS _ Representative /Date
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- 1 t ;l:ll Ji '_. ,p
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i Simulator Engineer f 5al'e
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APPROVED: [n,Y,~ DATE: r -' NIS)$f6TM
, FLORIDA POWER CORPORATION-CRYSTAL-RIVER UNIT-3 SIMULATOR
- {P)C1 INITIAL SIMULATOR CERTIFICATION TABLE OF CONTENTS (Cont'd)
VOLUME-II - Performance Test Abstracts
-BOOK 2 MALFUNCTION TESTS PTM17E - Turbine Bypass Valve Failure PTM17F - Continuous _ Rod Withdrawal Accident PTH17G - Dilution Accident PTM18A - Makeup Pump Trip PTM188 - Pressurizer Level Control Valve failure PTMA8C-- Letdown Isolation Valve Failuro PTM180 - Pressurizer Heater Controller Failure PTM18E - Pressurizer Spray Valve Failure PTM18F - RCS Narrow Range Pressure Transmitter Failure PTM20A --Main Steam Line Break Outside Containment PTM208 - Main Feedwater Line' Break Inside Containment PTM20C - Main Feedwater Line Break Outside Containment PTM21A..- Power: Range Channel failure PTM21B - Intermediate Range Channel failure PTM21C - Source Range Channel Failure s
PTM22A - RCS Thot Transmitter Failure PTM22B .- RCS Tcold Transmitter Failure PTM22C - Main Feedwater Flow Transmitter Failure
-PTM220-- 0TSG Operating Range Level Transmitter Failure
& PTM22E-- ICS Runback Failure PTM22F - ICS Meader Pressure Bias Failure
-PTM22G - Main.Feedwater Temperature Transmitter Failure PTM22H-- Failure of-ICS to Transfer to Tracking Mode PTM221 - Reactor Demand Signal Failure PTM22J - Selected TAVE' Failure PTM23A - Main Steam Isolation Actuation Failure PTM23B - Main Feedwater Isolation Actuation Failure PTM23C - ES Channel Failure to Actuate HPI PTM23D - ES-Channel Failure to Actuate RB Isolation and Cooling.
PTM23E - EFIC Auto Actuation Failurc
-PTM24A - ATWS (No Operator Intervention)
PTM24B-- ATWS (0perator Intervention) s O q )
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 SIMULATOR INITIAL SIMULATOR CERTIFICATION TABLE OF CONTENTS VOLUME II - Performance Test Abstra 11 BOOK 2 MALFUNCTION TESTS PTM4A - Reactor Coolant Pump Shaf t Shear PTM4B - Less of 2 Reactor Coolant Pumps PTM4C - Reactor Coolant Pump Seal Failure PTM4D - Reactor Coolant Pump Locked Rotor PTM5A - Loss of Condenser Vacuum Due to Air In-Leakage PTM5B - Loss of Condenser Level Control PTM6A - Loss of Nuclear Services Raw Water PTM6B - Loss of Circulating Water PTM6C - Loss of Decay Heat Raw Water PTM6D - Partial loss of Circulating Water PTM7 - Loss of Decey Heat Removal PTM8A - Loss of Nuclear Services Closed Cycle Cooling 7, PTM8B - Loss of Secondary Services Closed Cycle
'J
_ Cooling PTM8C - Loss of Decay Heat Closed Cycle Cooling PTH9A - Main Feedwater Pump Trip and Runback PTM98 - Loss of Both Main Feedwater Pumps PTM9C - 0TSG Overfill PTM9D - Loss of Both Condensate Pumps PTM10 - Total Loss of Feedwater (Normal and Emergency) , PTMllA - Failure of RPS -15V Cabinet Power Supply PTMllB - Failure of ES Cabinet Power Supply PTM12A - Mechanically Stuck Rod , PTM12B - Uncoupled Rod PTM12C - Degraded Rod Motion PlM12D - Dropped Control Pod l PTM13 - Inability to Drive Control Rods PTM14A - High RCS Activity PTM148 - Waste Gas Decay Tank Leak PTM16 - Main Generator Trip PTM17A - Control Rod Out Inhibit Failure PTM178 - Control Rod Group Command Enable Failure PTM17C - Failure of the Main Turbine to Trip PTM17D - Atmospheric Dump Valve failure ( -
N, TEST NUNBER/ TITLE: PTH4A/ REACTOR COOLANT PUMP SHAFT SHEAR _ ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(4) , DATE PERFORNED: 01/13/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at approximately 92% power when the shaft shears on RC Punip-1A. The plant trips (Power / Flow) and stabilizes in the normal post trip window with three RC Pumps in operation. BALFUNCTION TITLE OPl!OKS RC-008 RC Pump Shaft Shear Selectable to the A B, C and/or D RC Pump INITIAL CONDITIONS: Simulator operating at 92% power, Letdown Orifice Block isolation Valve (MVV-50) closed, all other plant equipment in service. FINAL CONDITIONS:- Reactor tripped, plant stabilized in the normal post trip envelope with three Reactor Coolant Pumps in operation. PLANT PR0r.EDURES USED DURING THE TEST:
- AP-580 REV 15 REACTOR TRIP - (- s 'u BASLLINE DATA: .U0ER 86 Failure of RCP-3A1. 01/01/86 RECALL DATA f!LE 0123185 - RCP Shaft Shear from 91% Power L
4 O
o PERFORMANCE TEST NUMBER: PTM4A DATE: 01/13/91 V TITLE: REACTOR COOLANT PUMP SHAFT SHEAR TEST RESU!.TS/ COMMENTS: Test results satisfactory. No performa:'ce problems noted. Excellent correlation between plant data and simulator. i V k I l-
- .bd
TEST NUMBER / TITLE: PTM4A/ REACTOR COOLANT PUMP SHAFT SHEAR TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Paramete,r DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grix Rcdr 0 - 320 in 1.875 sec *HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe 1.1:6ADT delta T Loop-A Grfx Rcdr 0 - 50F LI:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LldPR212 Th Loop-A Grfx Rcdr 520 - 620F *LlJPR213 Th Loop-B Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +120F *LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPR216 Tc Cold leg B1 Grfx Rcdr 520 - 620F *LlJPR217 Tc Cold Leg B2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in *LlJPS284 SG A Op level Grfx Rcdr 0 - 100% *L1JPS336 SG B Op level Grfx Rcdr 0 - 100%
Grfx Rcdr 0 - 1200 psi C **F3JPA323 SG-A Press (EFIC) F3JPA320 SG-B Press (EFIC) LlJPT228 MS Loop-A Hdr Press Grfx Rcdr Grfx Rcdr 0 - 1200 psi 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR *LlJPS329 SG A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR Ll:RC6TM Delta T-Cold Grfx Rcdr +10F L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR O
I:
- x. /
PERFORMANCE TEST NUMBER: PTM4A DATE: 01/13/91 TITLE: REACTOR COOLANT PUMP SHAFT SHEAR TEST STATUS: 1.0 This test is complete and has the following defic-lencies noted (if none so indicate): TR # Priority Date issued Date Closed NONE 1 _fhhd ~ FPC Test < Coordinator f:
/Date 2.0 All deficiencies have been resolved and this test has been completed. I o
NOT REQUIRED l FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. i l b
-Simulattir 1nstructor Y skf
/Date kS& Y Tib I Operati6hs Representative /Date 1
, ,,.4 di ~, 5 I 9/n f;l Siinulator Eiigip6er *
/Date APPROVED: 0C trx. DATE: _3 0 it ifs ~1,$/NLOTMf-1 V
N, p TEST NUMBER / TITLE: PTM4B/ LOSS OF 2 REACTOR COOLANT PUMPS O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(4) DATE PERFORMED: C,l/23/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at approximat 'v 93% power when a fault on the D RCP feeder breaker causes a loss of the B 6900 VAC Bus. The reactor trips due a loss of 2 RCPs (Pump Power Monitor). The plant is stabilized in the normal post trip window with OTSG levels being controlled at low level limits by emergency feedwater. MALFUNCTION TITLE OPTIONS ED 002 Loss of 6900 VAC Bus Selectable to the A and/or B Bus INITIAL CONDITIONS: Plant operating at -93% power, all plant equipment in service. FINAL CONDITIONS: Reactor tripped, plant in the normaqi post trip envelope with the A and C reactor coolant pumps in operation. OTSG 1evels being maintained at low level setpoint by emergency feedwater. PLANT PROCEDURES USED DURING THE TEST: A 'y) AP-580 REV 15 REACTOR TRIP BASELINE DATA: U0ER 85 Reactor Trip. 12/03/85 RECALL DATA FILE 0120385 - Reactor Trip / Loss of B 6900 VAC Bus from 93% Power, 12/03/85
N. ' PERFORMANCE TEST NUMBER: PTM4B DATE: 01/23/91 TITLE: LOSS OF 2 REACTOR COOLANT PUMPS TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. The post trip refeed of the OTSGs and A MFW pump control problems experienced at the plant were not replicated on the simulator. The post trip minimum pressurizer level at the plant was much less than on the simulator due to different initial values (220" on the simulator vs -200" on the plant), the effects of post trip feedwater control on the plant, and steam leakage past one of the plant main steam safety valves following the trip. Correlation between all other plant data and simulator response was excellent. O o
TEST NUMBER / TITLE: PTM4B/ LOSS OF 2 REACTOR COOLANT PUMPS V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter uCV at CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 620F
*LlJPR213 Th loop-B Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +120F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPR216 Tc Cold Leg B1 Grfx Rcdr 520 - 620F
*LIJPR217 Tc Cold leg B2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
fm *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi V *F3JPA320 SG-B Press (EFIC) LlJPT228 MS Loop-A Hdr Press Grfx Rcdr Grfx Rcdr 0 - 1200 psi 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*L1JPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LldPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L1:RC6TM Delta T-Cold Grfx Rcdr +10F L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/HR L4JPR233 RCS Total Flow Grfx Rcdr 0 - 160 MLB/HR F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm
.W - ._- - s i
PERFORMANCE TEST NUMBER: PIM4B DATE: 01/23/91 TITLE: LOSS OF 2 REACTOR COOLANT PUMPS TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): j TR # Priority Date issued Date Closed NONE
- p. r&.
l? l tic Test Coordinator /Datel 2.0 All deficiencies have been resolved and this test has been completed. 'v NOT REQUIRI:D TIC Test Coordinator /DiT5 3.0 Test results reviewed. Simulator performance is satisfactory. (hl, f ?hf;bil G\
< $b . 'i)
C5i irIiilator Instructor -~/Date UFiriti s'eiir~sienYitive/Date A
//l Y _fl' !
_Simulat6r Ehgipeer f/bijj LI
/Date n.,0/ /
APPROVED: < r??/,A-s ' DATE: _.1/A_
'NS S/NLOTM o
YY
N, , p TEST NUMBER / TITLE: PTM4C/ REACTOR COOLANT PUMP SEAL FAILURE b' N ANSI /A'S-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(4) DATE PERFORMED: 07/16/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, the first and second stage seals for RCP-1A fail. Plant load is reduced and RCP-1A is secured. MALFUNCTION TITLE OPTIONS RC-010 RCP First Stage Seal failure Selectable to RCP-1A. -18, -10 and/or -10 Adjustable from 0 - 100% degradation RC-Oll RCP Second Stage Seal failure Selectable to RCP-1 A, 18, -1C and/or -10 Adjustable from 0 - 100% degradation INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: 60% load, RCP-1A secured. PLANT PLCEDURES USED DURING THE TEST: OP-302, REVISION 33 RC PUMP OPERATION BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement.
N. , i 1 PERFORMANCE TEST NUMBER: PTM4C DATE: 07/16/91 O TITLE: REACTOR COOLANT PUMP SEAL FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test procedure steps functioned normally, Simulator response acceptable. O o- \
. - - _ . - - _ .. .- .~ - - . . _ .
e' TEST NUMBER / TITLE: PTM4C/ REACTOR COOLANT PUMP SEAL FAILURE 'V TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED HNP01' RCP-1A lst Stage Seal Pressure Grfx Rcdr 0 2500 psig 3.75 sec HNP02 RCP-1A 2nd Stage Seal Pressure Grfx Rcdr 0 - 2500 psig HNP03 RCP-1A 3rd Stage Seal Pressure Grfx Rcdr 0 - 2500 psig BMW7226 RCP-1A Controlled Bleedoff Flow Grfx Rcdr 0 - 5 gpm O
G
g PERFORt%NCE TEST NUMBER: PTM4C DATE: 07/16/91 TITLE: REACTOR C0OLANT PUMP SEAL FAILURE , TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed i NONE l 1 w b ? uj.3! T -t m
.f f a"estl CooYdir4 tor /) ate 2.0 All deficiencies have been resolved and this test has been completed.
s U NOT REQUIRED FPC Test Coordinator /Date 1 3.0 Test results reviewed. Simulator performance is satisfactory, 6Mle
*) % id i M
'peratioh_eprescr.LOS/
Simulator Instruttor /ba".e O 'ativa/Date l u ,
, .- /
f! w/ l [ ll Arl Simulat'&_Ehgineer or
?!'!GL
'/date l
APPROVED: fM M IS/NLOTM DATE: _/ Mh[__ l l l n u l l
x, a 1 I TEST NUMBER / TITLE: PTM4D/ REACTOR COOLANT PUMP LOCKED ROTOR O- ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(4) DATE PERFORMED: 07/17/91 TEST DURATIONt 1/2 hour Tbi DESCRIPTION: With the simulator operating at 100*, power, the rotor on RCP 1A ' locks, resulting in an abrupt decrease in RCS 1oop A flow,1he l sudden 1 ss of flow causes a rear .or trip on flux / flow / delta ; flux. l 9 1 D E ll!Rf Im1 EIlatil RC-009 Reactor Coolant Pump Locked P.otor Selectable to RCP lA, -10. 10 l and/or 10 l INITIAL CONDITIONS: Full power steady state concitions, all plant equipment in service. FINAL COH0!TIONS: Reactor turbine tripped. Plant stabilized in the normal post trip envelope with OTSG 1evels being controlled at low level limits by main feedwater. l PLANT PROCEDURES USED DURING THE TEST: AP 580, REVISION 16 REACTOR TRIP BASELINE DATA: There is no p' ant data available for this transient. Expected simulator response based on expert judgement extrapolation cf , FSAR analysis, i O
f _ TEST NUMBER / TITLE: PIM40/ REACTOR COOLANT PUMP LOCKED R010R v TEST _ DATA. LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 1.875 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 pst
*RRSWTP Normal 17ed Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Ganeration Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50f L1:6BDT delte T Loop B Grfx Rcdr 0 50f
*LlJPR212 Th loop A Grfx Rcdr 520 6.
*LlJPR213 Th loop B Grfx Rcdr 520 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f LITMARG4 Subcooling Margin Grfx Rcdr 10 +120f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr $20 - 620f
*LlJPR216 Tc Cold Leg 01 Grfx Rcdr 520 - 620F
*LlJPR217 Tc Cold leg B2 Grfx Rcdr 520 020f
*LlJPS287 SG A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 250 in
*L1JPS284 SG A Op Level Grfx Rcdr 0 100",
'llJPS336 SG B 3p Level Grfx Rcdr 0 100%
*f3JPA323 SG-A Press (EflC) Grfx Rcdr 0 - 1200 psi
'O *f3JPA320 SG B Press (EFIC)
LlJPT228 MS Loop-A Hdr Press Grfx Rcdr Grfx Rcdr 0 600 1200 pst 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS?01 SG-A MfW flow Grfx Rcdr 0 6K KLB/HR
*LlJPS302 SG B MfW flow Grfx Rcdr 0 6K KLB/HR
*LlJPS329 SG A SU FW flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW flow Grfx Rcdr 0 1500 MLB/HR Ll:RC6TM Delta T-Cold Grfx Rcdr -10 +10f L4JPR234 RCS Loop A flow Grfx Rcdr 0 - 80 MLB/HR L4JPR235 RCS Loop B flow Grfx Rcdr 0 - 80 MLD/HR L4JPR233 RCS Total flow Grfx Rcdr 0 160 MLB/HR O
0
.wm -
PERFORMANCE TEST NUMBER: P1M40 DATE: 01/17/91 TITLE: REACTOR COOLANT PUMP LOCKE0 ROTOR TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All procedure steps funct;oned normally. Simitlator response extclient. i O
~ '
O
p y PERFORMANCE TEST NUMBER: PIM4D DATE: 07/17/91 TITLE: Rf ACTOR COOL ANT PUMP LOCKf D R010R TES1 STATUS: , 1.0 lhis test is complete and has it:e f 0110 wing deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed _h0hl _ _ . . _...__ _ _ . _ _ _ . _ . m e ggsage-. Ae-h; A a r. b ewem .a4J.,queenMe v,me,n.se__.haens .h-.a.=maa FP SIfToofhI$to) /a 2.0 All deficiencies have been resolved and this test has been completed. C- NOT IW(M Al:D ITC TeTt fooFdTnator 7Dite 3.0 Test results reviewed. Simulator performance is satisfactory.
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Ild) k O '5' 2. r S / 3 w .( k- *J LTniilator Th, s k4,s'{iu {dr j'daq't ep Operati3nsRepresentative/Date w h ,, ,$p, BTmulii ITn iiiiiEr / Dais
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APPROVED , /]bf DATE:
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l
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w., O I TEST NUMBER / TITLE: P1MSA/ LOSS OF CONDENSER VACUUM DUE 10 AIR IN LEAKAGE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 1.1.2(5) l DATE PERFORMED: 01/21/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at 100% power when a vacuum lock develops which is within the capacity of the air removal system. Vacuum initially decreases until the standby Air Removal Pump starts and stabilizes vacuum. At this point, the leak is increased to a severity level greater than air removal system capacity, causing a turbine trip on low vacuum and a resultant recctor trip, tiAlf0NCT10N Illi,I QPTIONS CD 007 Condenser Air in-leakage Adjustable from 0 - 11,000 SCFM INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in ; service. l FINAL CONDITIONS: Turbine and reactor tripped, plant stabilized in the normal post trip window with S/G pressure being controlled by the atmospheric dump valves. PLANT PROCEDURES USED DURING THE TEST: Ap-580 REV 15 REACTOR TRIP OP 607 REV 13 CONDENSER VACUUM SYSTEM BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
_ _ _ _ _______m. _ p PERFORMANCE TEST NUMBER: PlH5A DATE: 01/21/91 TITLE: LOSS Of CONDENSER VACUUM OUE TO AIR IN LEAKAGE TEST RESULTS/ COMMENTS: Test results satisfactory. One (1) TR written on the automatic response of ARV-28 and 29 on decreasing vacuum. Valves were opened manually by the operator during the test. All other siraulator response excellent. TR does not require rerun of the test. O i i ( l l l I O
s, , 4 mih-mer TEST NUMBER / TITLE: PT_H5A/LO,SS Of C000ENSER VACUUM 000 TO AIR IN LEAKAGE Os TEST DATA LABEL DESCRIPTION RECORulNG RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED :
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 1.875 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:61.0T delta T Loop A Grfx Rcdr 0 50f Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50f
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 620f
*LlJPR214 Tc Cold leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620f
*LlJPS287 SG-A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG B SV Level Grfx Rcdr 0 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 100%
*f3JPA323 SG-A Press (EflC) Grfx Rcdr 0 1200 psi
*f3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi '
LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS301 SG A MfW flow Grfx Rcdr 0 - 6K KLB/HR s *LlJPS302 SG B MfW flow Grfx Rcdr 0 6K KLB/HR
*L1JPS329 SG A SU FW flow Grfx Rcdr 0 - 1500 MLB/HR
*L1JPS330 SG-B SU FW flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 100% L5G7AV Grp 7 Avg Position Grfx Redr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm DCP3AINV Main Condenser Vacuum Grfx Rcdr 0 - 30" Hg DCT3AT0T Main Condenser Steam / Liq Temp Grfx Rcdr 100 - 200 f DCT3AllQ Main Condenser Liquid Temp Grfx Rcdr 50 150 f O
4 PERFORMANCE TEST NUMBER: PIHSA DATE 01/a'1/9 11TLE: LOSS Of CONDO 4SER VACUUM DUE TO AIR IN-LEAKAGE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _ltL-oo7 _ C __. 9342ld_L_ (, >N..g.321 1
. <. g /[//i/
hPTf{estToordinator
~
70ite 2.0 All deficiencies have been resolved and this test has been completed.
, *ClotAure of Th-oo? doen not O a.
U Ffest Coordinajor {s r-~ qm 1.
/hlat'l6 l s
ptevent Completion of t.h i n tont s) 3.0 Test results reviewed. Simulator performance is satisfactory,
, V f f, Q cp/ g./'/
/
f, t W ggy gyg a 9. py Simulator Instructor 7DTte Operatisifs~ Representative /Date
,/ L*'. *l' o,,,,,
Simulator GgTiiler 7Ete g' /= APPROVED: +f 'f. C/. . DATE:~ m ~ ~ A
'REThAW1V l
g
4
. . . _ , _ _ . . , _ _ .__ ._ . _ _ _ . . - - . -3 l
TEST NUMBER / TITLE: PlMSB/ LOSS Of C0fiDENSER LEVEL CONTROL ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(E) DATE PERFORMED: 01/18/91 TEST DURATION: 1 bour TEST DESCRIPTION: With the simulator operating at 100% power, a large leak develops on the hotwell crossover piping. Since the leak is beyond the capacity of makeup from the condensate storage tank, hotwell level begins decreasing. Eventually, condenser level and vacuum, condensate feedwater booster and main feedwater pumps are Icst. tiALFUNCTION ILT11 REllMi CD 001 Hotwell Crossover Pipe Break Adjustable from 0 3000 lbm/sec INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, condensate and main feedwater lost. Plant stalilized at hot standby with OTSG levels beint., maintained by EfW and DTSG/ header pressures controlling on the ADVs. PLANT PROCE9VRES USED DURING THE TEST) BASELINE DATA: There is nc "lant data for this transient. Expected simulator results Nsed on expert judgement. O
s, i PERFORMU4CE TEST NUMBER: PTH58 OATE: 01/18/91 V TITLE: LOSS Of C0f4 DENSER LEVEL C0fliROL _ TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. One (1) Trouble Report written on sump responses during the leak. TR does not require rerun of the test. All other simulator response excellent. 1R 008 cleared 03/19/91. O O
N. l TEST NUMBER / TITLE: PTM58/ LOSS OF CONDENSFR ' ;EVEL CONTk01 TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 50F Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SV Level Grfx Rcdr 0 - 250 in
*LIJPS284 SG A Op Level Grfx Rcdr 0 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 1200 psi LlJPT228 HS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 1200 psi Grfx Rcdr 0 6K KLB/HR O **LlJPS301 LlJPS302 SG-B
*LlJPS329 SG A SU FW Flow SGMFW A MFWFlow Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 1500 MLB/HR DCWPLKL Liquid Leak Flow Out Grfx Rcdr 0 - 1500 LB/SEC DCWPLKA Air leak flow in Grfx Rcdr 0 - -150 LB/SEC
-DCl3A Condenser Hotwell Level Grfx Rcdr 0 - 10 ft DCL3B Condenser Hotwell level Grfx Rcdr 0 - 10 ft <
FAP65 CDP-1A Discharge Pressure Grfx Rcdr 0 - 200 psi FAP66 CDP-1B Discharge Pressure Grfx Rcdr 0 - 200 psi FAW3268 CDP 1A Flow Grfx Rcdr 0 - 1500 LB/SEC
'FAW3267 CDP-1B Flow Grfx Rcdr 0 1500 LB/SEC DCP3AINV Condenser Vacuum Grfx Rcdr 0 - 30" Hg DCT3ATOT Main Condenser Stm/ Liq Temp Grfx Rcdr 50 - 200F SVAMS10 TBV MSV-10 Port Area Grfx Rcdr 01 F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 1000 gpm F3:2GFil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3::4Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm O
. .~
. . _ . . - - . , - - - ,.~:.....
PERFORMANCE TEST NUMBER: P1M5B DATE: 01/16/91 TITLE: LOSS Of CONDENSER LEVEL CONTROL TEST STATUS: 1 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _TR 0.qB _J _ __ _ Q1/16Ml_ _0MISMl_ l l;f _ I,, f ' b 'l T'C~lest Coordinator /Dile t 2.0 All deficiencies have been resolved and this test has been completed. ( 'f
.. .c 'llt}fj F K fest cosrainator /Dats 3.0 Test results reviewed. Simulator performance is satisfactory.
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APPROVED: a G;, A - DATE: _g'_4 NSIS/NLTM
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I myu_,y w_e . _ . - _ ST NLlMBER[1111Ej PiH6A/ LOSS Of IJUCL EAR SERVICES RAW WAl[R ANSl]ANS-3.5(191Q,CR055 S _
REFERENCE:
Section 3.1.2(6) DATE PERFORMCD: 0?/?S/91 TEST DURATION: 1 1/2 hours TEST DESCRIP110N: With the simulator operating at 100', power, RWP 1 trips. RWP-2B automatically starts, but subsequently trips. Due to the loss i of heat sink to f4SCCC, temperatures on flSCCC-supplied components begin increasing. The SW system is allowed to heat up to approxin,ately llSf: RWP 1 is restarted and temperatures on effected components decrease and stabilite at normal values. m -___m__ mEmcum mu Omm RWP Supply Breaker Trip Selectable to RWP-1, -?A, --2B, -3A RW-001 and/or 3B INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power stecJy state conditions. RWP-1 restarted, temperatures on f4SCCC supplied components stabilized. PLANT PROCEDURES USED DURING THE TEST: AP-330 REV 4 EOSS Of liUCLEAR SERVICES Cl0Sl0 CYCLE C00Lil1G BASELINE DATA: Thcre is no plant data for this transient. Expected simulator response based on expert judgement. _m_.__ __ _. _ _ _ O
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~~
l PERFORMANCE TEST NUMBER: P1M6A DATE: 01/25/91 \ ! TITLE: LOSS Of NUCLEAR SERVICES RAW WAl[R TEST RESULTS/ COMMENTS Test results evaluation based on expert Judgcment. Test results satisfactory. No performs.. .e problems noted. i O
\r TESf NUMBER / TITLE: PIM6A/ LOSS Of NUCLEAR SERVICES RAW WATER TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5.625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Gefx Rcdr 0 - 1000 MWe LI:6ADT delta T Loop-A Grfx Rcdr 0 - 50f L1:6BDT delt T Loop B Grfx Rcdr 0 50f
*LlJPR212 Th loop A Grfx Rcdr 520 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f
*LldPR214 Tc Cold Leg Al Grfx Pcdr 520 - 620f
*l!JPR215 Tc Cold Leg A2 Grfx dcdr 520 - 620f
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 250 in
*LlJPS284 SG A Op Level Grix P.cdr v 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*f3JPA323 SG A Press (EflC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG B Press (EflC) Grfx Redr 0 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 asi N *L1JPS301 SG A MfW flow Grfx Rcdr 0 - 6K KlB/iR
\ *LlJPS302 SG B MfW flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG A SU FW flow Grfx Rcdr 0 1500 HLB /HR
*LlJPS330 SG B SU FW flow Grfx Rcdr 0 1500 MLB/HR KBLTANK SWT 1 Level Grfx Rcdr 0 - 12 ft BMT06 Letdown Temperature Grfx Rcdr 50 150 f BMi22 Seal Rtrn Clr Outlet Temp Grfx Rcdr 50 - 150 f HQBBCORE RCS Boron Concentration Grfx Rcdr 0 3000 ppm JPPAX329 MVP 10 Stator Temp Grfx Rcdr 50 250 f JPPAX361 MVP-1B Mtr 0.B. Temp Grfx Rcdr 50 - 250 f JPPAX364 MVP-1B Mtr I.B. Temp Grfx Rcdr 50 - 250 f JPPAX221 thru CRD Mtr Stator Temps Grfx Rcdr 50 - 250 f JPPAX279 JPPAX337 hSCCC Cooler inlet Temp Grfx Rcdr 50 - 250 f JPPAX338 SWP 1A Mtr I.B. lemp Grfx Rcdr 50 250 f JPPAX339 SWP-1A Mtr 0.B. Temp Grfx Rcdr 50 - 250 f JPPAX340 SWP-1A Mtr Disch. Air Grfx Rcdr 50 250 f
-JPPAX341 SWP-1B Mtr I.B. Temp Grfx Rcdr 50 - 250 f JPPAX342 SWP-1B Mtr 0.B. Temp Grfx Rcdr 50 250 F JPPAX343 SWP-1B Mtr Disch. Air Grfx Rcdr 50 250 t
m . amas,sm me --- m__ lO l
PERFORMANCE TEST NUMBER: PTH6A DATE: 01/25/91 L TITLE: LOSS Of NUCLEAR SERVICES RAW WA1ER TEST STATUS: - 1.0 This test is complete and has the fellowing deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _lEL . y ed/ ' d I/?Sh{ FFC~ Test coordinator /bai's 2.0 All deficiencies have been resolved and this test has been completed. NOT REQu1HEr WClest CooFdinator /Date 3.0 Test results reviewed Simulator performance is satisfactory, yn (b f
/
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l w.. q TEST NUMBER / TITLE: PTM6B/ LOSS Of CIRCULATING WA1ER i ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1,.2(6) : DATE PERFORMED: 01/20/91 TEST _ DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when circulating water pump CWP 10 trips, resulting in a decrease ir, condenser vacuum. A load reduction is commenced; when power reaches 85%, CWP-lA. -10 and 10 trip. Vacuum decreases rapidly and a turbine / reactor trip occurs. Due to the total loss of CW, secondary services closed cycle cooling begins heating up, with a corresponding rise in temperature on SSCC cooled components. MLFUNCT10B 111LE 91LIAtti CW 001 Circulating Water Pump Trio Selectable to the A, B, C and/or D Circulating water pump INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped. Ileader pressure being controlled at 1025 psig by the atmospheric dump valves. SSCC temperature increasing, temperatures on operating /enercized SSCC cooled component increasing. PLANT PROCEDURES USED DURING TiiE TEST: AP-580 REV 15 REACTOR TRIP BASELINE DATA: There is no plant aata for this transient. Expected simulator response based on expert judgement. O
T s , i PERFORMANCE TEST NUMBER: P1H68 DATE: 01/28/91 TITLE: LOSS OF CIRCUtATING WATER TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. One (1) Trouble Report written on response of ARV-54 and -55. SC Surge Tank level increase i verified by main control board indication (Step M60100(d); graphic recorder minimum value for this parameter is -2' above normal operating level. TR 012 cleared 03/12/91. O v i l i l l l l_ __ _;_-,_..__-.-___- _ _ - - - _ _ , _ _ . , . - - --- - . _ _ ~ - --.-_.---. --- ~ - - . - - - - . -
1 i
# i TEST NUMBER / TITLE: PlM6B/ LOSS Of CIRCULATING WATER l V TEST D_ATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter UEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 3.75 sec ,
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi ,
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - l > 5',
PGJPE202 Gross Generation P S Rcdr 0 - 1000 MWe , Ll:6ADT delta T Loop A Grfx Redr 0 - 50f LI:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop A Grfx Rcdr $20 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 620F
*LlJPR214 Tc Cold leg Al Grfx Rcdr 520 620'
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620f
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 100%
*f3JPA323 SG A Press (EllC) Grfx Rcd; O - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJP1232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS30) SG A MfW flow Grfx Rcdr 0 - 6K KLB/HR Os *LlJPS302 SG-B MfW flow Grfx Rcdr 0 - 6K KLB/HR
*LIJPS329 SG A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LIJPS330 SG B SU FW Flow Grfx Rcdr 0 1500 MLB/HR DCP3AINV Condenser Vacuum Grfx Rcdr 0 - 30" Hg KDLTK SCT 1 Level Grfx Rcdr 10 15 ft KDT02 SCP Common Suction Hdr Temp Grfx Rcdr 50 150f U3JPT310 MLO Cooler Outlet Temp Grfx Rcdr 50 250f U3JPT311 MLO Cooler Outlet Temp Grfx Rcdr 50 250f U3JPA268 FWP-1A Thrust Brg Temp Grfx Rcdr 50 - 250f U3JPA272 FWP 1B Thrust Brg Temp Grfx Rcdr 50 - 250F F1JPA258. CDP A Mtr Inb Brg Temp Grfx Rcdr 50 150f flJPA259 CDP A Mtr Outb Brg Temp Grfx Rcdr 50 - 150f FlJPA260 CDP A Cpl Up Thrst Brg Temp Grfx Rcdr 50 - 150f flJPA261 CDP A Cpl Dn Thrst Brg Temp Grfx Rcdr 50 150f FlJPA262 CDP B Mtr Inb Brg Temp Grfx Rcdr 50 - 150f F1JPA263 CDP B Mtr Outb Brg Temp Grfx Rcdr 50 - 150f flJPA264 CDP B Cpl Up Thrst Brg Temp Grfx Rcdr 50 - 150f flJPA265 CDP B Cpl Dn Thrst Brg Temp Grfx Rcdr 50 150f l O
. . - _ _ - _ . _ - . - . . . _ = . - . . _ - - _ . . ~ - . - . -
I PERFORMANCE TEST NUMBER: PIM6B DATE: 01/28/91 TITLE: LOSS OF CIRCULAllNG WATER TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.1RM2 U_ 31/121%. 33L12]M- l
//
l 8b! Rll/ I est Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. w{/ fTestCoordinator i
~
2/ir/?/ TPC /0'aT6 3.0 Test results reviewed. Simulator performance is satisfactory. u/ 4 h _?lt il Simulator i}nstructor /Date
. b Y" O 'l Operation)S*kepresentative7Dat e 3
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'/(fiEE APPROVED: _
DATE: k O
s., TEST NUMBER / TITLE: PIM6C/ LOSS OF DECAY HEAT RAW WATER ANSI /ANS_-3. 5(1985) CROS$
REFERENCE:
Section 3.1.2(6) DATE PERFORMED: 01/28/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating in mode 5 with core heat removal being provided by the B decay heat train. Decay heat raw water pump RWP-3B trips, resulting in a loss of heat sink for the B decay heat removal train. RCS temperature and pressure begin increasing until decay heat removal capability is restored by starting RWP-3A and the A decay heat removal train, tLALFUNCTION li1LE QEl10M RW 001 Raw Water Pump Supply Bkr Trip Selectable to RWP 1, -2A, 28. -3A and/or -3B INITIAL CONDITIONS: Mode 5, B decay heat removal train in operation. FINAL CONDITIONS: Mode 5 operation with decay heat removal being provided by the A decay heat removal train. PLANT PROCEDURES USED DURING THE TEST: OP 404 REV 78 DECAY HEA1 REMOVAL SYSTEM O BASELINE DATA: Expected simulator response based on expert judgement and the following limited plant data: 00ER 86 Diving Operations Lead to Diver fatalities and Loss of Heat Sink for Decay Heat Removal System. l i
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1 4 PERFORMANCE TEST NUMBER: PTM6C DATE: 01/28/91 O TITLE: LOSS Of DECAY HEAT RAW WATER TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and limited plant data from U0ER 86-2. Test results satisfactory. No performance problems noted. RCS heatup rate resulting from loss of Decay Heat Removal System heat sink shows good correlation with plant data from referenced incident. O l l
0 TEST NUMBER / TITLE: PTM6C/ LOSS OF DECAY HEAT RAW WATER TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED L6JPR20B RCS Pressure Grfx Rcdr 0 - 200 psig 3.75 see Hl:113 Pressurizer Level Grfx Rcdr 0 320 in.
Bl:00T31 DH/LPI Flow Loop A Grfx Rcdr 0 5000 gpm Bl:0VT41 DH/LPI Flow Loop B Grfx Rcdr 0 - $000 gpm N3: DIE DH Clr A in Temp Grfx Rcdr 0 - 300 F N3: DIG DH Clr B In Temp Grfx Rcdr 0 300 F N3JPX209 DH Clr A Out Temp Grfx Rcdr 0 - 300 F N3JPX210 DH Clr B Out Temp Grfx Rcdr 0 - 300 F KELTKB DHCCC Tk B Level Grfx Rcdr 0 - 16 feet KET06 DCHE A Hx Tin Grfx Rcdr 50 200 F KET10A DCHE-A Hx Tout Grfx Rcdr 50 200 F KET12 DCHE B Hx Tin Grfx Rcdr 50 - 200 F KET10B DCHE B Hx Tout Grfx Rcdr 50 - 200 F R4JSA019 Core Exit Temp H B Grfx Rcdr 0 400 F R4JSA020 Core Exit Temp F-8 Grfx Rcdr 0 400 F R4JSA023 r.. . Exit Temp G 9 Grfx Rcdr 0 - 400 F
?
i l g PERFORMANCE TEST NUMBER: PIM6C DAT E:ol/28/91 TITLE: LOSS OF DECAY HEAT RAW WATER 1 l TEST STATUS: 1 1.0 This test is complete and has the following deficiencies noted (if none l so indicate): TR # Priority Date issued Date Closed _R0f!E__ .
,. . . . . . ~ . - . - . _ _ _ -.-.
I? A -- . _1 t/ 60 lest Coordinator / hate 2.0 All deficiencies have been resolved and this test has been completed. p Nor 1mou11mu _ ._. IFC Test Coordinator /D_at e 3.0 Test results reviewed. Simulator performanco'is satisfactory. in ' t ?Y b 1LI' bper.ations Representative /Date L!I .? i
/
mulatorinstructor /Date L
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APPROVED: [_ _ _ _ DATE: .g?q 4
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lap pi N, TEST NUMBER / TITLE: PTM6D/ PARTIAL LOSS Of CIRCULATING WATER ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(6)
,DATE PERFORMED: 08/13/91 TEST LURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 50% power, 350 MWe with the "B" and "C" circulating water pumps and their associated waterboxes secured for maintenance. A trip of the A circulating water pump occurs, resulting in a loss of cooling to one of the condensers and a loss of Secondary Services Closed Cycle Cooling. A load reduction is commenced, and the main turbine is tripped at approximately 35% due to increasing turbine temperatures. As a result of the turbine trip and reduced bypass valve / atmospheric dump valve capacity (due to waterboxes being secured), 01SG pressure rapidly increases, causing a reduction in feedwater flow to the 01SGs. AMSAC actuates, initiating emergency feedwater. The reactor is manually tripped, emergency feedwater is secured and the plant is stabilized in the normal post trip envelope, tiA1 M C11 M IIILE QPl10M CW-001 Circulating Water Pump Trip Selectable to CWP 1A, 1B, 10 and/or -1D INITIAL CONDITIONS: 50% power, 350 MWe, CWP 1B, -10 secured for maintenance.
O MSV 26, 9 and -14 isolated. SSCCC lined up to SCHE 1A being supplied by CWP-1A. Letdown approximately 80 gpm. Group 7 rods at 94% withdrawn. _ FINAL CONDITIONS: Reactor / turbine tripped. Plant in the normal post trip envelope with OTSG levels at - 50". PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: Draft V0ER 91- - Reactor Trip and Emergency feedwater Actuation Due to a Trip of a Main Circulating Water Pump. 04/20/91 RECALL DATA flLE 0042091 Reactor 1 rip and EfW Actuation Due to a Partial loss of Circ Water. 04/20/91
~ ~ ~
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s, i e- - _ _ PERFORMANCE TEST NUMBER: PTM60 DATE: 08/13/91 TITLE: PARTIAL LOSS OF CIRCULATING WATER TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. A r.omparison of simulator and plant data indicates a generally excellent correlation. Plant data shows a more rapid post trip couldown than the simulator. This may be attributed to a larger sost trip main feedwater overfeed on the plant than on the simulator, combined wit 1 a lower plant decay heat level due to extended pre transient operation at reduced power levels. I o O
5., 0
-3 TEST NUMBER / TITLE: P1M6D/ PARTIAL LOSS OF CIRCULATING WATER TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE E CHAi(SP_EED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 5.625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop A Grfx Rcdr 0 50F Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50F '
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LIJPR214 Tc Cold Leg Al Grfx Rcdr 520 620F tLlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620F
*LlJPS287 SG-A SU Level Crfx Rcdr 0 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LIJPS284 SG A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG A Press EFIC) Grfx Rcdr 0 1200 psi
*F3JP/320 SG B Press EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop A H r Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi q *LlJPS301 SG A MFW Flow Grfx Rcdr 0 6K KLB/HR Q **LlJPS302 LilPS329 SG-ASG-B SU FW MFW Flow Flow Grfx Rcdr Grfx Rcdr 0 6K KLB/HR 0 - 1500 HLB /HR
*LlJPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR F3:23Fil EFW Flow S/G A Grfx Rcdr 0 1000 gpm F3:25FI) EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpn.
F3:24Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm UJTGEN Generator Gas Temperature Grfx Rcdr 50 - 250 F UJTACOUT Exciter Cooler Outlet Temp Grfx Rcdr 50 - 250 F l l n U
l {- PERFORMANCE TEST NUMBER: PIM60 DATE: 08/13/91 TITLE: PARTIAL LOSS Of CIRCULAllNG WATER TEST STATUS: 1.0 This test is complete and has ?he following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed NONE
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e . k f'Ckf~905rdinator
@l/.5/jf
/tisTe 2.0 All deficiencies have been resolved and this test has been completed.
O NOT HI;QU1Hl:D I W Test Coordinator /Date 3.0 Test results reviewed. Simulator performance i, satisf actory. Yk 31mulatof fnstructor
@~hCfy (N W Operation I %MM epresentative7Date
/Date Y / / }l,:!;j
}laforChgineer Simu "[Dili APPROVED: (
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DATE: l .L atb[
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TEST NUMBER / TITLE: PTM7/tJSS OF DECAY HEAT REMOVAL O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(7) DATE PERFORMED: 01/27/91 TEST DURATION: 1 1/2 hours TEST DESCRIPTION: The simulator is operating in mode 5 with the A decay heat removal train in service. Due to a failure of a pressure switch in the ACIS system (3RC3A-PS8), DHV-3 closes, resulting in the loss of the suction flowpath to the Decay Heat Removal pumps. The operating DHR pump DHP-1A is secured; the plant is allowed to heat up until until indicatiors of natural circulation and core heat rem' val by the OTSGs are established. MALFUNCTION lilLE OPTIONS DH-004 Decay Heat Removal Suction Selectable to DHV-3, -4 and/or -41 Valve Failure INITIAL CONDITIONS: Mode S, A decay heat removal train in serei e. FINAL CONDITIONS: Mode 4, natural circulation F. OTSG cooling established. PLANT PROCEDURES USED DURING THE TEST: AP-360 REV 3 LCLS OF DECAY HEA' : Mar OP-202 REV 84 PLANT HEATUP BASELINE DATA: Tl.ere is no plant data for this transient. Exp -ted simulator response based on expert judgement, s l O
4 PERFORMANCE TEST NUMBER: PTM7 DATE: 01/27/91 TITLE: LOSS OF DECAY HEAT REMOVAL TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement, Test results satisfactory. No performance problems noted. [v C/
4
-s TEST NUMBER / TITLE: PTM7/ LOSS OF DECAY HEAT REMOVAL
\s_/ TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED L6JPR208 RCS Pressure Grfx Rcdr 0 - 200 psig 3.75 sec Hl:ll3 Pressurizer level Grfx Rcdr 0 320 in.
Bl:0VT31 DH/LPI Flow Loop A Grfx Rcdr 0 - 5000 gpm N3: DIE DH Clr A In Temp Grfx Rcdr 0 - 300 F N3JPX209 DH Clr A Out Temp Grfx Rcdr 0 - 300 F LlJPR220 Tc Cold Leg Al Grfx Rcdr 50 - 650 F LlJPR221 Tc Cold leg B1 Grfx Rcdr 50 650 F F3JPA323 SG-A Pressure (EFIC) Grfx Rcdr 0 - 1200 psig F3JPA320 SG-B Pressure (EFIC) Grfx Rcdr 0 - 1200 psig R4JSA019 Core Exit Temp H 8 ' Grfx Rcdr 0 - 400 F R4JSA020 Core Exit Temp F-8 Grfx Rcdr 0 - 400 F R4JSA023 Core Exit Temp G-9 Grfx Rcdr 0 - 400 F R4JSA025 Core Exit Temp E-7 Grfx Rcdr 0 - 400 F O v ( s_-
I f c. PERFORMANCE TEST NUMBER: PTM7 DATF: 01/27/91 TITLE: LOSS OF DECAY HEAT REMOVAL TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed
.NDE I
s th7l9/ hbe[st' Coordinator
/Date 2.0 All deficiencies have been resolved and this test has been completed.
C NOT REQUIRED FPC Test Coordinator /Date 3.0- Test results reviewed. Simulator performance is satisfactory, a , A/ / b[4P/9( Y. ?_ Y S ) mulator Instructor /Date Op epresentative/Dite ' s
~ .k $ ).*
' Simulator Enginger l 1}/Date hll'tI APPROVED: i 0. <I t Mu FISTS /NLOTM DATE: ~ "'
9[II/T/ l
.a O
f TEST NUMBER / TITLE: PTM8A/ LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING k- ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(8) DATE PERFORMED: 01/28/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, a large leak is initiated on the SW pump common discharge header. The size and location of the leak results in a rapid depletion of the SW system inventory and a loss of cooling to all SW supplied components. The plant is tripped, reactor coolant pumps are secured and a natural circulation cooldown is started. MALFUNCTION TITLE OPTIONS SW-001 SW System Leak in the Aux Bldg Adjustable from 0 - 700 lbm' ec INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in i service. FINAL CONDITIONS: Reactor turbine tripped, RC Pumps secured, natural circulation cooldown in progress with no nuclear services closed cycle cooling available to SW supplied components. I PLANT PROCEDURES USED DURING THE TEST: 1 AP-580 REV 15 REACTOR TRIP AP-530 REV 8 NATURAL CIRCULATION AP-330 REV 4 LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
. . - . =.
r PERFORMANCE TEST NUMBER: PTM8A DATE: 01/28/91 TITLE: LOSS OF NUCLEAR, SERVICES CLOSED CYCLE COOLING TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted, b U O
TEST NUMBER / TITLE: PTM8A/ LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING C __ TEST DATA LABEL DESCRIP110N RECORDINt RANGE FREQUENCY /
- Critical Parameter DEVICE _ CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5.625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR20B RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi N2JPX202 Letdown Flow Grfx Rcdr 0 - 160 gpm JPPAM711 Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Inj Flow Grfx Rcdr 0 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in
*RRSWTP Normalized Reactor Pcwer Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop-A Grfx Redr 0 - 50F LI:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
(~ *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi ( *F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi l LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi l LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi l *L1JPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG-B HFW Flow Grfx Rcdr 0 - 6K KLB/HR l *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
, *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR KBLTANK SWT-1 Level Grfx Rcdr 0 - 12 ft KBP01 SWP Common Discharge Press Grfx Rcdr 0 - 150 psi KBW1920 SWP-1A Discharge Flow Grfx Rcdr 0 - 1500 LBM/SEC KBW19203 SWP-1B Discharge Flow Grfx Rcdr 0 - 1500 LSM/SEC KBW19205 SWP-lc Discharge Flow Grfx Rcdr 0 - 1500 LBM/SEC BMT06 Letdown Temperature Grfx Rcdr 50 - 150F BMT22 Seal Rtrn Clr Outlet Temp Grfx Rcdr 50 - 150F B2JPX329 MVP-1B Stator Temp Grfx Rcdr 50 - 150F B2JPX361 MVP-1B Mtr 0 B. Temp Grfx Rcdr 100 - 200F B2JPX364 MVP-1B Mtr I.B. Temp Grfx Rcdr 100 - 200F
- LIJPX221 CRD 10-0 Mtr Stator Temp Grfx Rcdr 50 - 250F l LIJPX223 CRD 3-E Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX225 CRD 7-E Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX227 CRD ll-F Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX229 CRD 2-F Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX231 CRD 6-F Mtr Stator Temp Grfx Rcdr 50 - 250F J
4 l TEST NUMBER / TITLE: PTM8A/ LOSS OF NUCLEAR SERVICES CLOSED CYCLE COOLING . k TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED LIJPX233 CRD 10-F Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX235 CRD 14-F Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX237 CRD 5-G Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX241 CRD 13 G Mtr Stator Temp Grfx Rcdr 50 - 250F LIJPX243 CRD 4-H Mtr Stator Temp Grfx Rcdr 50 250F K9JPX338 SWP-1A Mtr I.B. Temp Grfx Rcdr 100 - 200F K9JPX339 SWP-1A Mtr 0.B. Temp Grfx Rcdr 100 - 200F K9JPX340 SWP-1A Mtr Disch. Air Grfx Rcdr 100 - 200f K9JPX341 SWP-1B Mtr I.B. Temp Grfx Rcdr 100 - 200F K9JPX342 SWP-1B Mtr 0 B. Temp Grfx Rcdr 100 - 200F K9JPX343 SWP-1B Mtr Disch. Air Grfx Rcdr 100 - 200f K9JPX344 SWP-lC Mtr I.B. Temp Grfx Rcdr 100 200F K9JPX345 SWP-lc Mtr 0 B. Temp Grfx Rcdr 100 - 200F K9JPX346 SWP-IC Mtr Disch. Air Grfx Rcdr 100 - 200F KlJPX205 RWP-1 Mtr U.B. Temp Grfx Rcdr 100 - 200F KlJPX206 RWP-1 Mtr L.B. Temp Grfx Rcdr 100 - 200F KlJPX207 RWP-1.Mtr Disch. Air Grfx Rcdr 100 - 200F
-H1JPX288 RCP A Mtr Air in Temp Grfx Rcdr 100 - 200F HIJPX293 RCP C Mtr Air in Temp Grfx Rcdr 100 - 200F HlJPX298 RCP B Mtr Air in Temp Grfx Rcdr 100 - 200F
-HlJPX306 RCP D Mtr Air in Temp Grfx Rcdr 100 - 200F
, HlJPX290 RCP A Mtr Up Thrst Brg Temp Grfx Rcdr 100 - 200F HlJPX295 RCP C Mtr Up Thrst Brg Temp Grfx Rcdr 100 - 200F HIJPX300 RCP B Mtr Up Thrst Brg Temp Grfx Rcdr 100 - 200F HlJPX308 RCP D Mtr Up Thrst Brg Temp Crfx Rcdr 100 - 200F l
i O
} .= O etaron"^"ct test ausacn: e'"8^ o^to: 02/28/91 TITLE: LOSS OF NUCLEAR SERVICES Cl0 SED CYCLE COOLING TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.30E _ _ _ . _ _ _ _ . _
,u(
~
i < m/d/l1 FPC 1est Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. O NOT REQUIRI:D FPFiest Coordinator /0Hi 3.0 Test results reviewed. Simulator performance is satisfactory.
~
'}){ '!,j O(9 ad~~~' %'~iff/n>
v 3LsJ - >f -h / j, -
- 1. y Operations Representative /Date imulator Instructor /Date
//
cyl kt . Simulat'or Engine 6r "~/DiTE _2 L;{3 APPROVED: ..._'ig-(r a., _ DATE: NSTS/Nt01M ' O
4 i TEST NUMBER / TITLE: PTMS8/ LOSS Of SECONDARY SERVICES CLOSED CYCLE COOLING
ANSI /AHS-3.S(1985) CROSS
REFERENCE:
Section 3.1.2(8) DATE PERFORMED: 01/28/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, the running SSCC pump trips on overcurrer,t: the backup SSCCC pump autostarts, then also trips. The resultant loss of cooling to the SSCCC-supplii.i components causes increasing temperatures on these componeit'. In order to prevent equipment damage the plant is tripper and all unnecessary heat loads are secured. SCP-1B is subt.quently restored, and temperatures on effected components are stabilized. 11ALFUNCTLQN IIILE OPTIONS SC-002 StP Motor Trip Selectable to SCP-1 A and/or -1B
.4-INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service.
FINAL CONDITIONS: Reactor / turbine tripped. Main feedwater and condensate secured; plant stabilized in the normal post trip envelope with OTSG levels being controlled by emergency feedwater. SCP-1B in operation: temperatures of SSCCC-supplied components ('T V stabilize PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. O O
r l _ _ _ PERFORMANCE TEST NUMBER: PTM8B DATE: 01/28/91 O TITLE: LOSS OF SECONDARY SERVICES CLOSED CYCLE COOLING TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. O O
f n TEST NUMBER / TITLE: PTM8B/ LOSS OF SECONDARY SERVICES CLOSED CYCLF COOLING l k) TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / ,
- Critical Parameter DEVICE CHART SPEED '
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (VR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe L1:6ADT delta T Loop A Grfx Rcdr 0 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LIJPR212 Th loop-A Grfx Rcdr ' 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3Ji'320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi p)
(
*LlJPS301 SG-A MFW Flow
*LlJPS302 SG-B MFW Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR
*LIJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LIJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR KDLTK SCT-1 Level Grfx Rcdr 10 - 15 ft KDT02 SCP Common Suction Hdr Temp Grfx Rcdr 50 150F UJTGEN Avg Generator Gas Temp Grfx Rcdr 50 - 200F l UJTAC00T Exciter Cooler Outlet Temp Grfx Rcdr 50 - 200F i
USTSSFA Air Side S0 Cooler Out Temp Grfx Rcdr 50 - 150F USTSSFH H2 Side S0 Cooler Out . Temp Grfx Rcdr 50 - 150F V3JPT310 MLO Cooler Outlet Temp Grfx Rcdr 100 - 200F V3JPT311 MLO Cooler Outlet Temp Grfx Rcdr 100 - 200F V3JPA268 FWP-1A Thrust Brg Temp Grfx Rcdr 100 - 200F V3JA272 FWP-1B Thrust Brg~ Temp Grfx Rcdr 100 - 200F FlJPA258 CDP'A Mtr Inb Brg Temp Grfx Rcdr 50 - 150F F1JPA259 CDP A Mtr Outb Brg Temp Grfx Rcdr 50 - 150F F1JPA260 CDP A Cpl Up Thrust Brg Temp Grfx Rcdr 50 - 150F F1JPA261 CDP A Cpl Dn Thrust Brg Temp Grfx Rcdr 50 - 150F F1JPA262 CDP B Mtr Inb Brg Temp Grfx Rcdr 50 - 150F FlJPA263 CDP B Mtr Outb Brg-Temp Grfx Rcdr 50 - 150F FlJPA264 CDP B Cpl Up Thrust Brg Temp Grfx Rcdr 50 - IS0F FlJPA265 CDP B Cpl Dn Thrust Brg Temp Grfx Rcdr 50 - 150F p L)
. - = . _ - . ._ _
( PERFORMANCE TEST NUMBER: P1M8B TITLE: LOSS OF SECONDARY SERVICES CLOSED CYCLE COOLING DATE: 01/28/91 TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed RitL [M .b> b? $ 4 9 ' k/ T IC Test Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. bkl < Ob / (& l Ac 9. LXhl Simulator Instructor /Date Operations 0 Representative /Date
'ff .)
Simulator Engineer [1. :l
/Date APPROVED: t fu s DATE: ift2'
*t l ifST]/NLOTM \'
O
<w TEST NUMBER / TITLE: PTM8C/ LOSS OF DECAY HEAT CLOSED CYCLE COOLING l ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(8) _ DATE PERFORMED: OS/13/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating in mode 5 with the A Decay heat removal train in operation when DHCCC Pump DCP-1A trips. Due to the loss of heat sink to the decay heat removal system, RCS temperature and pressure begin increasing until the B DHCCC train and the B DHR train are placed in service. ! MALFUNCTION TITLE OPT 10tLS 00-001 DCP Motor Trip Selectable to DCP-1A and/or -1B INITIAL CONDITIONS: Mode 5, A Decay Heat Removal Train in service. FINAL CONDITIONS: Mode 5 operation with decay heat removal being provided by the B decay heat removal train. PLANT PR';CEDURES USED DURING THE TEST: OP-404, REV 78 DECAY HEAT REMOVAL SYSTEM (D U BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
/
n PERFORMANCE TEST NUMBER: PIM8C DATE: 08/13/91 l !v,) TITLE: LOSS OF DECAY HEAT CLOSED CYCLE COOLING TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted.
)
1 o v
w 3 TEST NUMBER / TITLE: PTM8C/ LOSS OF DECAY HEAT CLOSED CYCLE _ COOLING (_/ TEST DATA _ LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
*Critichl Parameter DEVICE CHART SPEED L6JPR208 RCS Pressure Grfx Rcdr 0 - 50 psig 5.625 sec Hl:ll? Pressurizer Level Grfx Rcdr 100 - 150 in Bl:0VT31 DH/LPI Flow Loop A Grfx Rcdr 0 - 5000 gpm l B1:0VT41 DH/LPI Flow Loop B Grfx Rcdr 0 - 5000 gpm N3: DIE DH Clr A In Temp Grfx Rcdr 50 - 150 F N3: DIG DH Clr B In Temp Grfx Rcdr 50 - 150 F N3JPX209 DH Clr A Out Temp Grfx Rcdr 50 - 150 F N3JPX210 DH Clr B Out Temp Grfx Rcdr 50 - 150 F KELTKA DHCCC Tk A Level Grfx Rcdr 5 - 15 feet KET06 DCHE-A Hx Tin Grfx Rcdr 50 - 100 F KET10A DCHE-A Hx Tout Grfx Rcdr 50 - 100 F KET12 DCHE-B Hx Tin Grfx Redr 50 - 100 F KET10B DCHE-B Hx Tout Grfx Rcdr 50 - 100 F KEWTKAI DHCCC TK A Hakeup Flow Grfx Rcdr 0 - 100 gpm KEWO708 DCP-1A Flow Grix Rcdr 0 - 4000 gpm R4JSA019 Core Exit T/C H-8 Grfx Rcdr 100 - 150 F R4JSA020 Core Exit T/C F-8 Grfx Rcdr 100 - 150 F R4JSA023 Core Exit T/C G-9 Grfx Rcdr 100 - 150 F BlJPR250 DHP-1A Brg Temp Grfx Rcdr 50 - 100 F BlJPX318 DHP-1A Mtr Inbd Brg Temp Grfx Rcdr 50 - 100 F BlJPX319 DHP-1A Mtr Outbd Brg Temp Grfx Rcdr 50 - 100 F
( , l
( PERFORMANCE TEST NUMBER: PTM8C DATE: 08/13/91 TITLE: LOSS OF DECAY HEAT CLOSED CYCLE COOLING TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _EE __ - . . _ _ . ( l 'u t (f/)l}/ FPC Test Coordinator /Date 2,0 All deficiencies have been resolved and this test has been completed. 'O v
.__8OTAB@lMD FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
(c { 5.E Y!// )/ (b l-s~'~e % L 0 Sj_ Simulator Instructof /Date Operation 8 Representative /Date ff') fi Simulatof'EngiSeer 7,,,;jh;
/Date c'
Y't/9/ APPROVED: \ cm.(tDi t- W CATE: NSff/NLOTM M O
TEST NUMBER / TITLE: PTM9A/ MAIN FEEDWATER PUMP TRIP AND RUNBACK ANSI /ANS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(9) DATE PERFORMED: 01/11/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 97% power, the suction valve for the A Main Feedwater Pump closes, resulting in an automatic trip of the pump. The ICS automatically runs the plant back to 55% power at a rate of 50% per minute. During the runback, the PORV is cycled twice to prevent a reactor trip on high RCS pressure. MALFUNCTION IllLE OPTIONS TVFWSF14 FWV-14 Fail to Specified Position Adjustable from 0 - 100% INITIAL CONDITIONS: 97% power, all plant equipment in service. FINAL CONDITIONS: Plant stabilized at ~53% power, with the A Main Feedwater Pump secured. Q PLANT PROCEDURES USED DURING THE TEST: b AP-545 REV 3 PLANT RUNBACK l BASELINE DATA: U0ER 87-01 "A" MFWP Trip and Successful Runback. 03/07/87 RECAlt. DATA FILE 0030787 - MFP Trip / Runback from 97% Power 03/07/87 l l l l O V l l
( es PERFORMANCE TEST NUMBER: PlM9A DATE: 01/11/91 ! /}
' TITLE: MAIN FEE 0 WATER PUMP TRIP AND RUNBACK TEST RESULTS/ COMMENTS:
Test results satisfactory. No performance problems noted. Excellent correlation i between plant data and simulator response. l 1 ?) b O
~
E-e TEST NUMBER / TITLE: PTM9A/ MAIN FEEDWATER PUMP TRIP AND RUNBACK TEST DATA-LABEL DESCRIPTION- RECORDING- RANGE FREQUENCY /
- Critical Parameter ' DEVICE CHART SPEED 4
*H1:ll3 Pressurizer Level Grfx Rcdr- 0 - 320 in 1.125 sec *HldPR203_ Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr- 1700 - 2500 psi *L6JPR208 RCS Pressure-(WR) Grfx Rcdr 0 - 2500 psi 1 *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta-T Loop A- Grfx Rcdr 0 - 50F Ll:6BDT- delta T Loop-B Grfx Rcdr _0 - 50F
*LIJPR212 Th Loop-A Grfx Rcdr 520 - 620F ' *HCTAVG- RCS Average Temperature. Grfx Rcdr 520 - 620F *LlJiR214 Tc Cold leg Al- Grfx Rcdr 520 - 620F *L1JPR215 Tc Cold leg A2 Grfx Rcdr-- 520 - 620F *LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG B SU Level Grfx-Redr 0 - 250 in *LlJPS284.SG-A Op Level Grfx Rcdr 0 - 100% *L1JPS336 SG-B Op Level- Grfx Rcdr 0_- 100% *F3JPA323-SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320'SG-B Press (EFIC)- Grfx Rcdr 0 - 1200 psi L1JPT228 MS_ Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS toop-B Hdr Press. Grfx Rcdr 600.- 1200 psi p
V **LlJPS301 SG-A MFW LIJPS302 SG-B-MFW Flow F1ow Grfx Rcdr-Grfx Rcdr 0 6K KLB/HR 0 - 6K~KLB/HR
*L1JPS329 SG-A SU_FW Flow Grfx Rcdr- 0 - 1500 MLB/HR *LlJPS330'SG-B SU'FW Flow Grfx Rcdr 0 - 1500 MLB/HR
- GGG03N Main Turbine Speed Grfx Rcdr 0 - 2000 rpm L5G6AV Grp 6 Avg Position Grfx_Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 -_3000_ ppm O
[ y PERFORMANCE TEST NUMBER: PTM9A DATE: 01/11/91 _ TITLE: MAIN FEE 0 WATER PUMP TRIP AND RUNBACK TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1 TR # Priority Date issued Date Closed 5 l I j.
$b
,A PC Test Coordinator k/
'/Date
! 2.0 All deficiencies have been resolved and this test has been completed. C V NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. ll. iniulhtoF Instructor .' /Date Y \nD 9$ OperationPRepresentative/Date l 9f{. cf </s .- '.n 'jg l Simulator Engineer '/Date APPROVED: -e < . DATE: ci taly l iis7S/NL0m ^w s i
\
x) L A L) l
1 f Q TEST NUMBER / TITLE: PTM98/ LOSS OF BOTH MAIN FEEDWATER PUMPS ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(9) DATE PERFORMED: 01/12/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The plant is operating at 21% power in preparation for synchronization of the main generator with the grid When the generator output breaker is closed, the generator immediately picks up 135 MWe do to a problem in the EHC system. As a result of picking up the excassive load, OTSG pressures rapidly decrease to approximately 600 psig, resulting in main steam isolation, trip of the main feedwater pumps and a reactor / turbine trip. The platt is stabilized in the normal post trip window with 0TSG levels being maintained by emergency feedwater. tLALFUNG03 TITLE 0_Pl10NS I TAL71VS1 Turbine Governor Valves - Initial Adjus table from 0 - 100% Valve Position , INITIAL CONDITIONS: 21% power, main generator ready for synchronizatlon. A MFW Pump I control in manual, B MFW Pump running on the low speed stop. ! q FINAL CONDITIONS: Reac+or/ turbine tripped, both MfW Pumps tripped, MSiks closed. v Head 3 pressure being controlled by the Atmospheric Dump Valves, OTSG levels being maintained by Emergency Feedwater. RCa temperature, pressure and pressurizer level stabilized in the normal post trip envelope. PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-460 REV 8 STEAM GENERATOR ISOLATION ACTUATION AP-450 REV 15 EMERGENCY FEE 0 WATER ACTUATION OP-203 REV 61 PLANT STARTUP BASELINE DATA: U0ER 88 Reactor Trip Due to Excessive Steam Generator Depressurization Following Output Breaker Closure. 10/28/88 RECALL DATA FILE D102888 - Reactor Trip /MS Jsols: ion During Startup. 10/28/83 O-
{} PERFORMANCE TEST NUMBER: PTM9B DATE: 01/12/91 TITLE: LOSS OF BOTH MAIN FEE 0 WATER PUMPS TEST RESULTS/ COMMENTS: Test results satisfactory. No performance problems noted. Excellent correlation between plant data and simulator. i l t O v l l f
~
l
O v TEST NUMBER / TITLE: PIM98/ LOSS OF BOTH MAIN FEEDWATER PUMPS TEST DATA LABEL DESCRIPTIGN RECORDING RANGE FREQUENCY /
- Critical Para.neter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in ,5 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LIJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average remperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Redr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi h) *LlJPS301 SG-A MFW Flow
*LlJPS302 SG-B MFW Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR
*LlJPS329 SG-A SV FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR GGG03N Main Turbine Speed Grfx Rcdr 0 - 2000 rpm L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm l i \
DATE: 01/12/91 Q PERFORMANCE TEST NUMBER: PTM9B TITLE: LOSS OF BOTH MAIN FEE 0 WATER PUMPS
=
TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed NONE
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2,0 All deficiencies have been resolved and this test has been completed. O- MOT REQUIRED TR Test toordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. xd[hfb
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REFERENCE:
Section 3.1.2(9) DATE PERFORMED: 01/11/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at approximately 73% power with the C RC Pump secured, when a failure of the A MFW pump governor causes the fW pump speed to increase to 4950 rpm. The increase in feed flow to the A OTSG causes an incease in generator level, a reduction in superheat, and a corresponding reduction in RCS temperature, pressure and pressurizer level, in order to terminate the overfeed the A MfW pump is tripped, resulting in a reactor / turbine trip on high RCS pressure. tiALEMC110H IJ1LE OPTIONS TVSXPGV Feed Pump Governor Valve - Selectable to A or B FW Pump Fail to Specified Position Adjustable from 0 - 100% INITIAL CONDITIONS: 73*, power, C Reactor Coolant Pump secured. FINAL CONDITIONS: Reactor / turbine tripped, plant stabilized in the normal post trip k'ndow with the C RC pump and A MFW pump secured. O PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR Trip BASELINE DATA: U0ER 83 Reactor Trip on High RCS Pressure Following a Main ~ feedwater Pump Trip. 08/26/83 RECALL DATA FILE 0062683 - 0TSG Overfill / Reactor Trip on liigh RCS Pressure. 08/26/83 0
- - -. .= . - - . = . . ..
. O; PERFORMANCE TEST NUMBER: PIM9C DATE: 01/11/91 v
-TITLE: OTSG OVERFILL TEST RESULTS/ COMMENTS:
Test results satisfactory. No performance problems noted. Excellent correlation between simulator and actual plant data. Note: Test rerun 08/11/91 to obtain second graphic recorder output. Bad software labei (HIJPR310) used on first test.
!est rerun 09/09/91 to obtain graphic recorder output for plant / simulator data comperison utility run. Rescaled Main and Startup feedwater flows to match 1ECALL data scaling, added main steam temperatures.
O O
p TEST NUMBER / TITLE: PTM9C/0TSG OVERFILL TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:llS Pressurizer Level Grfx Rcdr 0 - 320 in .9375 sec *HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F 9.375 sec* *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi S.75 sec** *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta 1 Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F *LlJPR214 Tc Cold leg Al. Grfx Rcdr 520 - 620F *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in *LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi Grfx Rcdr 600 - 1200 psi *LlJPT232 MS Loop B Hdr Press C, LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - GK KLB/HR *LIJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR LlJPS288 S/G A Downcomer Temperature Grfx Rcdr 0 - 600 F LlJPS289 S/G B Downtomer Tcmperature Grfx Rcdr 0 - 600 F FTT60 FWHE 3A Outlet Temperature Grfx Rcdr 0 - 500 F FTT62 FWHE 3B Outlet Temperature Grfx Rcdr 0 - 500 F *LlJSA100 SG-A MFW Flow Grfx Rcdr 0 - 6 MLB/HR** *LlJSA10 3G-L MFW Flow Grfx Rcdr 0 - 6 MLP/HR** *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1.5 MLB/HR** *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1.5 MLB/HR**
SRT01 A-MS Temperature Grfx Rcdr 100 - 650 F** SRT07 B-MS Temperature Grfx Rcdr 100 - 650 F** Notes: *Second graphic recorder file run (Replacement of bad pressurizer level variable)
** Third graphic recorder file run. Feedwater flows rescaled, Main Steam _line temperatures added for plant / simulator data comparison plots.
t 0
4 n. >c PERFORMANCE TEST NUMBER: PTM9C DATE: 01/11/91 ( TITLE: OTSG DVERFILL TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none soindicate): TR # Priority Date issued Date Closed
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/Date 2.0 All deficiencies have been resolved and this test has been completed.
NOT EQUIMD FPC Test Coordint,or"~' /Date 3.0 Test results reviewed. $1mulator performance is satisfactory.
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t l TEST NUMBER / TITLE: PlM9D/ LOSS OF B01H CONDENSAlE PUMPS l ANSl/ANS-3.5(1985) CROSS REFERF.NCE: Section 3.1.2(9) DATE PERFORMED: 07/02/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, deaerator level limit switch FW-LS-4 fails, resulting in the generation of a high-high level signal to the condensate pump control system. As a result, both condensate pumps are tripped and actual deacrator level begins decreasing. When the deacrator level reaches 1 foot, both main feedwater booster pumps autmatically trip, resulting in a trip of both main feedwater pumps and a loss of main feedwater. tLALEVlt011M IU11 9PIlpEi 1FF14CAC FW-Lb 4 Failure (Fail Closed) INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor turbine tripped. Both condensate pumps, both feedwater booster pumps and both main feedwater pumps tripped. Simulator stabilized in the normal post trip window with OISG levels being maintained at setpoint by emergency feedwater. PLANT PROCEDURt:S USED DURING THE TEST: AP-580, REVISION 15 REACTOR TRIP AP 450, REVISION 15 EMERGENCY FEEDWATER ACTUATION BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. I O
4 p PERFORMANCE TEST NUMBER: PTH90 DATE: 07/02/91 TITLE: l0SS Of BOTH C0fdDEf4 SATE PUMPS TEST RESULTS/ COMMENT $t Test results evaluation based on expert judgement. All test steps functioned correctly. Simulator response good. O
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l TEST NUMBER / TITLE: P1M90/ LOSS Of B0lH CONDENSA1E PUMPS TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
' Critical Parameter DEVICE CHART SPEED *Hl:ll3 Pressurizer Level Erfx Rcdr 0 - 320 in 1.875 sec *HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700F *L4JPR222 RCS F,* essure (NR) Grfx Rcdr 1700 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Redr 0 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 Mwe Ll:6AD1 delta T loop A Grfx Rcdr 0 50f Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50f
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620f 'HCTAVG RCS Average Temperature Grfx Rcdr 520 620f *LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - G20f *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620f *LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SU Level Grfx Rcdr 0 250 in *LlJPS284 SG-A Op Level Grfx Rcdr 0 100% *LlJPS336 SG B Op Level Grfx Rcdr 0 100% *f3JPA323 SG A Press (EFIC) Grfx Rcdr 0 1200 psi *F3JPA320 SG-B Press (EflC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJPT221 MS Loop-B Hdr Press Grfx Rcdr 600 1200 psi
( \
*LlJPS301 SG A MFW Flow *L1JPS302 SG B MFW flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 6K KLB/HR *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 1500 MLB/HR *LlJPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/IIR FAW2867 CDP-1A flow Grfx Rcdr 0 - 1600 LB/HR FfW3651 FWBP A Flow Grfx Rcdr 0 - 1600 LB/HR FfW3851 FWBP B Flow Grfx Rcdr 0 - 1600 LB/HR FTLHTil Deaerator Level Grfx Rcdr 0 - 20 ft O
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C eEnr0anAnCc 1cs1 nusata, ein90 DATti 07/0efei TITLE: LOSS OF BOTH CONDEf4 SATE PUMPS TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _HOM l l j s
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N h p- . . a ps TPCih.\c es{Toordiiiak.y or /dite (I 2.0 All deficiencies have been resolved and this test has been completed. O , or i,,mui,,,m TPCTest Coordinator /D' ate 3.0 Test results reviewed. Simulator performance is satisfactory. a
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4 TEST NUMBER / TITLE: PIM10/ TOTAL LOSS Of FEEDWATER(NORMAL AND [MERGENCY) O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(10), Appendix B. Section B2.2(2) DATE PERFORMED: 08/15/91 TFST DURATION: 3 1/2 hours TEST DESCRIPTION: With the simulator operating at 100% power, both main feedwater pumps trip, resulting in a reactor / turbine trip. EfP 1 fails to start: EFP 2 starts but immediately trips or, overspeed. HPl/PORV cooling is established, ifW capability is subsequently restored. OTSG cooling of the RCS is established, and a bubble is drawn in the pressurizer. tiA_LILNC.110H 111Lf, QEUM) FW 005 Main feedwater Pump Trip Selectable to A or B Main FW Pump EF 003 Emergency FW Pump f ails Selectable to EfP 1 or EfP-2 to Start Ef 001 Emergency FW Pump Trip Selectable to EfP 1 or Ef P-2 INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. flNAL CONDITIONS: Plant cooldown in progress, RCS cooling by the OTSGs with forced circulation, normal RCS pressure control. PLA"T PROCEDURES USED DURING THE TEST: AP-580 REV 16 REACTOR TRIP AP-450 REV 16 EMERGENCY FEEDWATER ACTUATION AP-380 REV 18 ENGINEERED SAFEGUARDS ACTUATION OP-305 REV 12 OPERATION Of THE PRESSURIZER BASELINE DATA: There is no plant data for this transient. Expected simulator reseponse based on a combination of expert judgement and the following information: CR-3 ATOG Part !!, Volume 2 Appendix B Section 3.0 Loss of Secondary Inventory Control O
PERFORMANCE TEST NUMBER: PlH10 DATE: 08/15/91 TITLE: TOTAL LOSS Of FEEDWATER(140RMAl Af4D [MERGiliCY) TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and the referenced baseline data. Test results satisfactory. With the exception of minor problems with RC Drain Tank responso (already identifed in performance test P112), no performance problems noted. O
~
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f i TEST NUMBER / TITLE: PTH10/10TAL LOSS OF FEEDWATER(NORMAL AND [HERGENCY) O TEST DATA LABEL DESCRIPTION RECOR0lHG RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEE0
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 5.625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe L5G6AV Group 6 Avg Position Grfx Redr 0 - 100% L5G7AV Group 7 Avg Position Grfx Rcdr 0 - 100%
*LlJPR212 Th loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS-Average Temperature Grfx Rcdr 520 - 620F LlJPR220 Tc Cold leg Al Grfx Rcdr 50 - 650F LlJPR221 Tc Cold Leg A2 Grfx Rcdr 50 - 650F L1JPS287 SG A SU Level Grfx Rcdr 0 - 250 in L1JPS293 SG B SU Level Grfx Rcdr 0 - 250 in LlJPS284 SG A Op Level Grfx Rcdr 0 - 100%
LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 1200 psi F3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR C] LlJPS302 SG-B MFW Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW Flow 0 1.5 MLB/HR LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1.5 MLB/HR R4JSA019 Incore T/C Grfx Rcdr 0 - 700F R4JSA020 Incore T/C Grfx Rcdr 0 - 700F R4JSA023 Incore T/C Grfx Rcdr 0 - 700F R4JSA025 Incore T/C Grfx Rcdr 0 - 700F N2:M18 HPI Flow Al Grfx Rcdr 0 - 500 gpm N2:M14 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:M12 HPl Flow B1 Grfx Rcdr 0 - 500 ppm N2:M16 HPI Flow B2 Grfx Rcdr 0 - 500 gpm F3:23Fil EFW Flow S/G B Grfx Rcdr 0 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm HKLLIQ RCOT Level Grfx Rcdr 0 - 15 ft r. HKPGAS RCOT Pressure Grfx Rcdr 0 200 psi
- HKWPRZ Flow from Pzr to RCDT Grfx Rcdr 0 - 60 LB/SEC l H17E10 RCV-10 Discharge Temp Grfx Rcdr 100 - 300F O
PERFORMANCE TEST NUMBER: P1M10 DATE: 08/15/91 TITLE: TOTAL LOSS Of IELOWAllR(NORMAL AND [MERGlNCY) TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _MI. _.. _..._ _. _ . .. _ . _ _ _.
.uss an.e-e s sue- w<w - w . .-==---w* W h 4.sm'%7* *----w em gi e w e g- --,em--. w-.._cs__ m_._. 4: a aeN 5
b f. _ . _ . IC1,s/Coordi.. nator /Date 2.0 All deficiencies have been resolved and this test has been completed,
' war nr:ourni:n I C Test Coordinator /Dif6 3.0 Test results reviewed Simulator performance is satisfactory.
N^ {? b ' h(I? ' WlY'Y
$I5iiilatprTRifructor ~ ~/6afe operail60v-{ Representative /Date
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0 TEST NUMBER / TITLE: PIM11A/ FAILURE OF RPS -15V CABINET POWER SUPPLY O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(11) DATE PERFORMED: 01/14/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, the -15V cabinet power supply for the A RPS cabinet fails to OV. This results in the in a trip of the A RPS channel and all process and nuclear instrumentation signals for the channel to fail low. Pressurizer heaters energize in response to the low indicated RCS pressure. _ MLfUNCTION 111LE 02 11.03 5 RP 009 RPS 15V Cabinet Power Selectable to the A, B, C or 0 Supply Failure RPS Channel INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in ' service. FINAL CONDITIONS: Plant at full power steady state conditions with the A RPS channel tripped and the RPS in a 1 out of-3 trip condition. The narrow range RC pressure plug in the RPS cabinet is transferred to Subassembly B. PLANT PROCEDURES USED DURING THE TEST: OP-501 REV 14 REACTOR NON-NUCLEAR INSTRUMENTATION . BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement and plant design data on power supply dependencies. 1 i O I t .
_ =- j PERFORMANCE TEST NUMBER: PIMllA DATE: 01/14/91 TITLE: FAILURE Of THE RPS -15V CABillET POWER SUPPLY TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and plant design data on power supply dependencies. Test results were satisfactory with the exception of two (2) Trouble Reports generated for incorrect PPC point response and incorrect Event Point response. TR-009 and -010 cleared 04/26/91. O
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1 o l TEST NUMBER / TITLE: PTHilA/ FAILURE Of RPS -15V CABINET POWER SUPPLY TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED I
ANALOG DATA COLLECTION NOT REQUIRED FOR THis TEST SIMblATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT i l O l
. . _ __ . . _ . - _ _ - __. _ .. - ~ . - _ - - - , . - - - . - - - - - - _ - --- - - - - - -
l PERFORMANCE TEST NUMBER: PlHilA DATE: 01/14/91 TITLE: FAILURE Of RPS -15V CABINEl POWER SUPPLY TEST STATUS: 1.0 lhis test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed IR00'J __. __ C _.0.1 LIE 91_ _0E26491 1E010 _L _01/14/9.1_ _ M/10Z21_
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TP flest Coordinator /Dite 2.0 Alldeficiencgs_havebeenresolvedandthistesthasbeencompleted. j ,- h O l(,4f,,%p f bu t- -- J'f;ih/ IR lest (Yordliiator /Dile 3.0 Test results reviewed. Simulator performance is satisfactory. ks ..
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TEST NUMBER / TITLE: PTMllB/ FAILURE OF ES CABINET POWER SUPPLY O ANSI /ANS-3.S(1985) CROSS
REFERENCE:
Section 3.1.2(11) DATE PERFORMED: 01/14/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, the cabinet power supply for ES Channel 1 fails to OV. ES Channel 1 trips, but no ES actuation occurs. MALFUNCTIM li1LE OPTIONE ES 004 CS Cabinet Power Supply Selectable to ES Channel 1, 2 or 3 Failure IN!TIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Simulator at 100% power, steady state conditions with ES Channel 1 tripped ES actuation logic in a 1 out-of-2 condition. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement and plant design data on >O power supply dependencies.
i PERFORMANCE TEST NUMBER: PTHil8 DATE: 01/14/91 O TITLE: FAILURLOFESCABINETPOWERSUPPLY TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement and plant design data on power supply dependencies. Test results satisfactory with the exception of one (1) Trouble Report generated due to improper response of breaker status indications on the Vital Bus Status Monitor Panels. O l 9
TEST NUMBER / TITLE: P1HilB/ FAILURE OF ES CABINET F0WER SUPPLY O.
^
TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
( ANALOG DATA COLLECTION HOT REQUIRED FOR THis TEST
$1MULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT O
l lO - 1
g PERFORMANCE TEST NUMBER: PIM11B DATE: 01/14/91 _QTLE: FAILURE Of ES CABINET POWER SUPPLY TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
._RQll __.C __ _01/14/91_
, r
^
, gb'/ lll$/l?l i'C~Tes[~ Coordinator / Dale 2.0 All deficiencies have been resolved and this test has been completed.
p
- Closure of TR-oll doen not prevent (i e,, ' (, Mn plet son of thin tent
,/Yf>Cle oordinator / 2D' ate ,
3.0 Test results reviewed. Simulator performance is satisfactory.
} j, > rm l ,, f Sim ator' Instructor 7C f
/Date gj \ .- Q J'. y{ h (
Operatiohitepresentative/Date
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C TEST NUMBER / TITLE: PlH12A 4ECHANICALLY STUCK ROD ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(12) DATE PERFORMED: 01/27/91 TEST DURAil0N: I hour TEST DESCRIPTION: The simulator is operating at 100% power when a power reduction J to 75% is started with the ICS in full automatic. Due to mechanical binding rod 4 in group 7_does not insert with the rest of the group. As the rest of the group 7 rods are inserted, assymetric alarm and assymetric f. ult indications are received. The plant is stabilized at 75% pover, where an unsuccessful attempt is made to align the f aulted rod using the auxiliary power supply. tLALFUNCTIM IllLE LPl]LN$ TVLISR_ Stuck Control Rod Selectable to any rod or combination of rods INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Simulator stabilized at 75% power with an assymetric rod condition present. PiANT PROCED'JRES USED DURING THE TEST: O oe sor REv 27 con > Rot rod oRivE SvSits BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. l l 1 O l l l l l., _ . . . . _ _ . . _ . . _ . . _ _ _ _ - _ _ , _ . . _ . _ _ . . - _ . _ _ _ _ . _ _ _ . - - . . _ _ _ . _ . . . _ _ _ _ _ _ . _
C roaron"*"cc tost wo"8cai einiz- oatti oi/27/9i TITLE: MECilANICALLY STUCK ROD - - TEST RESULTS/ COMMENTS: lest results evaluation based on expert judgement. lest results satisfactory. No simulator performance problems noted. 1 V 1 l l m -.
1 I i C ust ==/ti'cti einira/ston^"ic^ttv siock noo TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in j SM sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 b-
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADI delta T Loop A Grfx Rcdr 0 - 50f ' Ll:6BDT delta 1 Loop B Grfx Rcdr 0 50f
*LlJPR212 Th loop A Grfx Rcdr 520 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f
*LlJPR214 Tc Cold leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 250 in .
*LlJPS284 SG A Op Level Grfx Rcdr 0 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 100% ,
*f3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG B Press (EflC) Grfx Rcdr 0 - 1200 pst LlJP1228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi m LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi U *LlJPS301 SG A MfW flow Grfx Redr 0 6K KLB/ilR
*LlJPS302 SG B MfW flow Grfx Rcdr 0 6K KLB/ilR
*LlJPS329 SG A SU FW flow Grfx Rcdr 0 1500 MLB/llR
*LlJPS330 SG B SU FW flow Grfx Rcdr 0 - 1500 MLB/liR L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% LIZ37 Grp'7 Rod 3 API Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm P.3JPP208 Power Range Indication Grfx Rcdr 0 - 125% R3JPP209 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Redr 0 - 125% R3JPP214 Power Range imbalance Grfx Rcdr -62,5 +62.5% R3JPP215 Power Range imbalance Grfx Rcdr -62.5 - 462.5% R3JPP216 Power Range imbalance Grfx Rcdr 62.5 462.5% R3JPP217 Power Range lmbalance Grfx Rcdr -62.5 - +62.5% O
i $ PERFORMANCE TEST NUMBER: P1H12A DATE: 01/27/91 TITLE: MECHANICALLY S1VCK ROD TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicato): TR # Priority Date issued Date Closed UDM ___._ _ _ _ _ _ _ _ . _ ____
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_ _ _ _ _ _ _ _ ~.. .. _. _ _ _ .._ _. _ ._ _ __.
$lN J&PC Test CoordTnator
hb
'/63te 2.0 All deficiencies have been resolved and this test has been completed.
O NOT Hl:QU l itl'.D TWliit Coordinator /liali 3.0 Test results reviewed. Simulator performance is satisfactory.
,/jp ::' ( g,,;,' } Q 3 <g , 7 7 c, j
- fi faiiIr'Miructor ~/[i TI a OperatiksRepresentative/Date V w Simulator EngTneer _ . l'/cate S l ,
APPROVED: I l 3 -- t , n" t% DATE: 6 9 fil li$1%)lE01\f~' \ u
- - . . y.m._- wp
O 'Est "UseER/Ti1i.E riNi28,0NCOUetED ROD ANSI /ANS-3.5(1985) CROS3
REFERENCE:
Section 3.1.2(12) DATE PERFORMED: 01/12/91 TEST DURATION: 3.5 hours TEST DESCRIPTION: A reactor startup is performed with rod 1 group $ uncoupled. ' Criticality is achieved at a rod index higher than predicted. As power is increased into the power range, indications of a quadrant power tilt develop. Due to the nature of the fault, the rod position indication system is not able to detect the problem. MALEMGilM lilli OPT 10fl5 CR-004 Uncoupled Rod Selectable to any rod or combination of rods INITIAL CONDITIONS: Plant in mode 3, ready for a reactor startup. FINAL CONDITIONS: Reactor at 15% power with indications of a quadrant power tilt. PLANT PROCEDURES USED DURING THE TEST: OP- PtANT STARTUP O oe so2 203arvREV 27 61 coataot noo oaivt svsic" 0P-210 REV 29 REACTOR STAR 10P BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
l l PERFORMANCE TEST NilHBER: PIM120 DAT E: 01/12/91 TITLE: UNCOUPLED ROD TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. lest results satisfactory. No performance problems noted. O O
TEST NUMBER /TITLEt PIM12B/UNCOUPLID ROD TEST DATA LABEL DESCRIPTION RECORDING RANGL FREQUENCY /
- Critical Parameter DEVICE CHART SPEED R3JPP208 Power Range Indication Grfx Rcdr 0 - 125% 5.625 sec R3JPP209 Power Range Indication Grix Rcdr 0 - 125%
R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 125% R3JPP202 Source Range CPS Grfx Rtdr -1.0 - 6.0 log cps R3JPP203 Source Range CPS Grfx Rcdr 1.0 - 6.v log cps R3JPP212 Intermediate Range Amps Grfx Rcdr -11.0 3.0 log amps R3JPP213 Intermediate Range Amps Grfx Rcdr -11.0 - -3.0 log amps L5G5AV Group 5 Avg Position Grfx Rtdr 0 - 100% LEG 6AV Group 6 Avg Position Grfx Rcdr 0 100% L5G7AV Group 7 Avg Position Grfx Rcdr 0 100% HCTAVG RCS Tave Grfx Rcdr 520 - 620f O
~
O
. PERFORMANCE TEST NUMBER: PlM12B DATE: 01/12/91 TITLE: UNC0VPLED ROD TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none 50 indicate): 1R # Priority Date issued Date Closed _h0hl_ _
.f, $hy;ff,'l ,
/>-
TFCkeit(Tordinator /Dife 2.0 All deficiencies have been resolved and this test has been completed. O NOT RI:QU1RI:D ff Fiest Coordinator /DaTi 3.0 Test results reviewed. Simulator performance is satisfactory. An __
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dImulil r nitructor P/Date Operatiofis RepresentatW e~/Date
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v9 hor E 5'n /0 t APPROVED: _. l '
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ry DATE: Nidj[ NSIS/llt IM a l O
4
,........:.-........,..~._.-............_........_..._....._._.-._.............-..~.. __
)
TEST NUMBER / TITLE: PTM120/ DEGRADED kOD HOTION ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(12) i DATE PERFORMED: 01/13/91 TEST DURATION: !/2 hour TEST DESCRIPTION: The simulator is operating at 75% power when a power increase to 100% is started with the ICS in full automatic. Group 7 begins withdrawing; due to mechanical binding, rod 2 in group 7 withdraws at a slower rate than the other rods in the group. When the assymetric rod alarm is received for rod 7-2, the load increase is terminated. MALFUNCTION TITLE OPTIONS CR 003 Control Rod Mechanical Selectable to any rod or combination Binding of rods INITIAL CONDITIONS: 75% power, all plant equipment in service. FINAL CONDITIONS: Simulator stabilized at approximately 85% power, with an assymetric alarm condition for rod 7-2 present. PLANT PROCEDURES USED DURING THE TEST: BASEllHE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. i O
- "zaamm Msommene' PERFORMANCE TEST NUMBER: PTH12C DATO 01/17/91 f
s' TITLE: DEGRADED R0D MOTION TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. l l l 1 n V-l
. - -, m ., , ,-. _ , . -
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r m.. ... . . . , . . . . . . - - c TEST NUMBER / TITLE: PIM120/ DEGRADED ROD MOT 10f4 TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 3,75 sec
*HlJPR203 Pressurizer Temperature Grfx Redr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Hormalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 50F L1:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR218 1c Cold leg Al Grfx Rcdr 520 - 62^F
*L1JPR215 Tc Cold Leg A2 Grfx Ccdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 250 in
*LlJPS284 SG-A Op Level Grfx htdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG A Press (EFIC) C fx Rcdr 0 - 1200 psi h
*F3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 psi L1JPT228 MS Loop-A Hdr Press Grfx Rcdr 600 1200 psi LlJPT237. MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi (7 *LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
(_,1 *LlJPS302 SG B MFW Flow Grfx Rcdr 0 - LK KLD/HR
*LlJPS329 SG-A SU TW Flow Grfx Rcdr 0 1500 MLB/HR
*L1JPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5GEAV Grp 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% LIZ10 Grp 7 Rod 2 API Grfx Pcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm R3JPP208 Power Range Indication Grfx Rcdr 0 - 125% R3JPP205 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% R3JPP214 Power Range imbalance Grfx F,cdr -62.5 - +62.5% R3JPP215 Power Range Imbalance Grfx Rcdr -62.5 - +62.5% R3JPP216 Power Range imbalance Grfx Rcdr -62.5 - +62.5% R3JPP217 Power Range Imbalance Grfx Rcdr 62.S - +62.5% (] V' a
e' PERFORMANCE TEST NUMBER: PTM12C O b DATE: ol/13/91 TITLE: UEGRADED ROD MOTION TEST STATUS: 1.0 This test is complete and has the 'following deficiencies noted (if none, so indicate): TR # Priority Date issued Date Closed NONI: Ab /7,d 4 C Test Coordinator /Dato 2.0 All deficiencies have been resolved and this test has been completed.
- l
/FPC T4st Coordinator
//rb
./Date~
3.0 Test results reviewed. Simulator performance is satisfactory, p/.: hip { \ \ nk T US~D l OperationsJRepresentative/Date Simulator' Instructor //Da'te 7: uAn Gd., -- / 5 9/ gime'noYEngineer //Date APPROVED: n m bhy a DATE: 9(Itlqi NS{.NLOTH
\
'V
( _ TEST NUMBER / TITLE: PTH12D/ DROPPED CONTROL ROD ANSI / @S 5(1985) CROSS
REFERENCE:
Section 3.1.2(12) DATE P Q WHED: 01/11/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 98% power, control rod 3-2 drops into the core. An assymetric rod runback to 60% power is initiated by the ICS. After the plant is stabilized, the rod is recovered in accordance with applicable plant operating procedures. MALFUNCTION III1E QPTIONS CR-002 Dropped Control Rod Selectable to any rod or combination of rods-INITIAL CCNDITIONS: 98% power, all plant equipment in service. FINAL COULITIONS: Simulator stabilized at approximately 601, power, rod 3-2 restored to a 100% withdrawn position. PLANT PROCEDURES USED DURING THE TEST: AP-545 REV 3 PLANT RUNBACK OP-502 REV 27 CONTROL ROD DRIVE SYSTEM BASELINE DATA: RECALL DATA FILE Dll3089 - Dropped Control Rod 3-2 from 98% Power l G V
e PERFORMANCE TEST NUMBER: PTH12D DATE: 01/11/91 TITLE: OROPPED CONTROL ROD TEST RESULTS/ COMMENTS: Test results satisfactory. Excellent correlation between simulator and plant data. No performance problems noted. O O
a 1
/
g TEST NUMBER / TITLE: PTM120/ DROPPED CONTROL R0D D _ TEST DATA LABEL DESCRIPTION: RECORDING- RANGE FREQUENCY /
- Critical Parameter - DEVICE CHART SPEED
*Hl:lI3. Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr' 0 -1000 MWe L1:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A- Grfx Rcdr 520 - 620F
-*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F-
*LlJPR214 Tc Cold Leg.Al Grfx Rcdr 520 - 620F
-*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU. Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr- 0 - 100%
*L1JPS336 SG B Op Level Grfx Rcdr- 0 --100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
, *L1JPS301.SG-A HFW Flow Grfx Rcdr 0 - 6K KLB/HR s *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329-SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330. SG-B SU FW Flow Grfx Rcdr- 0 --1500 MLB/HR L5G5AV Grp 5 Avg _ Position Grfx Rcdr 100%-
L5G6AV _Grp 6 Avg Position _ Grfx Rcdr 0 - 100% L5G7AV Grp 7. Avg Position- Grfx Rcdr- -0 -100% HQBBCORE,RCS Boron Grfx Rcdr 0 - 3000 ppm R3JPP208 Power Range-Indication -Grfx-Rcdr 0 - 125% ' R3JPP209 Power Range Indication Grfx Rcdr 0-- 125%- R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power' Range Indication Grfx Rcdr 0 - 125%
.R3JPP214 Power Range Imbalance - -Grfx Rcdr -62.5 +62.5%
R3JPP215 Power Range Imbalance Grfx Rcdr --62.5 - +62.5% R3JPP216 Power Range Imbalance .Grfx Rcdr -62.5 - +62.5%
'R3JPP217. Power Range Imbalance Grfx Rcdr-- -62.5 - +62.5%
, .- -, - - , - , , - . , _ . . - r x _, -- _ - ,
) C'! etaronnance Test aux 8ta: etsizo o^Tt: oi/it/92 TITLE: OROPPED CONTROL R00 TEST STATUS: 1,0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _tLONJ ah (
,4PC Test Coordinator
//6bt
~/Uile 2.0 All deficiencies have been resolved and this test has been completed.
NOT REQU1 RED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. y h /b7h \Y kNU Operationu Representative /Date
'tiL W l
-Simulatof instructor
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- a% A N b -r 7 , J/
'imul ator ' Engineer / att
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APPROVED: _$ a _hp N
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DATE: ; 12{g_
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/
TEST NUMBER / TITLE: PTM13/ INABILITY TO DRIVE CONTROL RODS O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(13) DATE PERFORMED: 01/14/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, a load reduction to 700 MWe at 2% per minute is initiated with the ICS in full automatic. Due to a malfunction in the control rod drive system, the control rods fail to insert as load is reduced, resulting in an ICS cross limit. After unsuccessful attempts to insert grcup 7 rods in manual on both the normal and auxiliary power supply the simulator is stabilized with feedwater controlling RCS Tave. MALFUNCTIOE TITLE pfJ10NS CR-009 Block Control Rod Insert / Withdraw Commands INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Plant stabilized at approximately full power conditions. Control rods in manual, reactor demand in mini-track, feedwater controlling RCS Tave, PLANT PROCEDURES USED DURING THE TEST: r] \J OP-502 REV 27 CONTROL R00 DRIVE SYSTEM BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. Nr
n 4
~^^
IE PERFORMANCE TEST NUMBER: PTM13 DATE: 01/14/91 TITLE: INABILITY TO ORIVE CONTROL RODS TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. O
~
O
6 4,
/
TEST NUMBER / TITLE: PTM13/ INABILITY TO DRIVE CONTROL RODS C/ TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*H1JPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop-A Grfx Rcdr 520 - 6'd .'
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Te Cold Leg A2 Grix Rcdr 520 - 620F
'L1JPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJP5336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR O)
( *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR l
*L1JPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*L1JPS330 SG-8 SU FW Flow Grfx Rcdr 0 - 1500 HLB /HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm R3JPP208 Power Range Indication Grfx Rcdr 0 - 125% R3JPP209 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% O
(] PERFORMANCE TEST NUMBER: PTM13 DATE: 01/14/91 TITLE: INABILITY TO DRIVE CONTROL RODS TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed 10E ___
,,C!?h[dh? $
JPC Test Coordinator /6alte 2.0 All deficiencies have been resolved and this test has been completed. O NOTREQUIRED DC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. 9 f ,
;ll}WA ~
' Simulator Instructor
,hk b-)\ S 4 'b L5 ' ' l OperatioM Representative /Date
/Date
/
{ ILnuh - L 2,w-
,.<imulator Engineer 93 /
/Date
/
APPROVED: ic - r mA DATE: # 34 NSjS/NLOTM ' ) l l l (D C/ { l l
,f-- TEST NUMBER / TITLE: PTM14A/HIGH RCS ACTIVITY \-
ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(14) DATE PERFORMED: 07/29/91 TEST DURATION: 1 1/2 HOURS TEST DESCRIPTION: With the simulator operating at 100% power, a power reduction to 70% load is commenced at 3%/ minute with the ICS in full automatic. Due to the rapid power reduction, fuel cladding failure occurs, resulting in an increase in RCS activity. This increased activity is reflected in RM-L1 readings and RCS samples. MALFUNCTION IllLE OPTIONS TVH0 FAIL Fuel Cladding Failure INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Simulatt e operating at 70% power with elevated RCS activity levels. RM-L1 in High Alarm. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. l l O
( sa PERFORMANCE TEST NuMBERt PTH14A DATE: 07/29/91 TITLE: HIGH RCS ACTIVITY TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Te:t results satisfactory. No performance problems noted. Note: The CR-3 simulator dynamically models radiation transport of Xenon (gaseous activity), lodine and Cesium (particulate activity), including production and decay. The severity of the fuel cladding failure malfunction for this test was adjusted to produce gaseous, iodine and particulate activity levels -equal to those listed in the FSAR for a 1% fuel cladding failure.
~
Simulator activity levels are calculated in uc/lbm Division by 450 will provide approximate values expressed in uc/cc. O 1 O~
( TEST NUMBER / TITLE: PTM14A/HIGH RCS ACTIVITY \- TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED HQRICORE RCS Activity - lodine Grfx Rcdr 0 - 10000 uc/lbm 3.75 sec HQRMCORE RCS Activity - Particulate Grfx Rcdr 0 - 25,000 uc/lbm HQRLCORE RCS Activity - Gas Grfx Rcdr 0 - 1.2E05 uc/lbm BER108 Sample Activity - lodine Grfx Rcdr 0 - 5000 uc/lbm BERLO8 Sample Activity - Particulate Grfx Rcdr 0 - 25,000 uc/lbm BERM 08 Sample Activity - Noble Gas Grfx Rcdr 0 - 1.2E05 uc/lbm CHORML1 RM-L1 Indication Grfx Rcdr 0 - 6 log cpm Os V
(UT
g PERFORMANCE TEST NUMBER: PIM14A DATE: 07/29/91 TITLE: HIGH RCS ACTIVITY TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.lGE.
1 l Y?h'?TC
-IPC Te Coordinator 7{N !
~/Date i
l l l 2.0 All deficiencies have been resolved and this test has been completed. NOT REQUIRED l TR~fest Coordinator /Date 1 3.0 lest results reviewed. Simulator performance is satisfactory. l
,d b f!J i _[& I Y % T 4'Q{
l . Sim'ulat'or Instructor /Date' OperationdRepresentative/Date_ t , c/ <;
// , ,} -
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\ i.: a. Simulator Eiikneer /Date APPROVED: e- ( ii' v"- DATE: d ! t
>O NgS/NL0lM l
l' i d I l l
TEST NUMBER / TITLE: PTM14B/ WASTE GAS DECAY TANK LEAK ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(14) DATE PERFORMED: 08/14/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at full power, a large leak develops on Waste Gas Decay Tank 3A. Radiation levels begin to increase in the Auxiliary Building ventilation system, resulting in isolation of the waste gas decay tank room and automatic tripping of the Auxiliay Building supply fans. Due to the extremely high curie content of the gas in the Auxiliary Building, numerous area radiation monitors (RM-Gs) go into alarm. MALFUNCTION TITLE OPTIONS Waste Gas Decay Tank Leak Selectable to WGDT-3A, -3B and/or - TVAALK_ 3C Adjustable from 0 to ~1700 SCFM at a tank pressure of 80 psig INITIAL CONDIT10NSr Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power conditions. RM- A3 in high alarm, Waste Gas Decay Tank Room ventilation isolated, Auxiliary Building Supply fans "T tripped, elevated radiation levels indicated on RM-A8, RM A2 (V and Auxiliary Building area gamma monitors. PLANT PROCEDURES USED DURING THE TEST: AP-250 REV 4 RADIATION MONITOR ACTUATION BASELINE DATA: _ There is no plant data for this transient. Expected simulator response based on expert judgement extrapolation of FSAR analyses, and Emergency Plan dose assessment information. Initial Tank Activity levels set to Technical Specification limit of 39,000 curies of noble gas (Xenon). O
i DATE: 08/14/91 C PERFORMANCE TEST NUMBER: PIM14B TITLE: WASTE GAS DECAY TANK LEAK TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement extrapolation of FSAR analysesand Emergency Plan dose assessment information. Test results satisfactory. tio performance problems noted, Note: Step M14B060 calls for decreasing readings on RM-A's after isolation of the Waste Gas Decay Tank Room ventilation. Readings began decreasing on RM A2; Radiation levels at RM-A3 and RM-A8 began decreasiag (as indicated in CIS), but were still too high to result in a decrease on the monitors. Due to the relatively long decay time for Xenon, it was decided to terminate the test before on-scale indications were obtained. O O
(]) TEST NUMBER / TITLE: PTM14B/ WASTE GAS DECAY TANK LEAK TEST DATA ) LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CliART SPEED !
CMORMA03 RM-A3 Radiation Grfx Rcdr 0 - 6 log cpm 1.875 sec CMORMA08 RM-A8 Radiation Grfx Rcdr 0 - 6 log cpm CMORMA02 RM-A2 Radiation Grfx Rcdr 0 - 6 log cpm CMORMG05 RM-G5 Radiation Grfx Rcdr 0 - 5 log mr/hr i
$a-BM
-(s
( PERFORMANCE TEST NUMBER: PTM14B DATE: 08/14/91 TITLE: WASTE GAS DECAY TANK LEAK TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _NONE I Coordinator J!f ,
,, 7p* /Date 2.0 All deficiencies have been resolved and this test has been completed.
r~N b ' NOT REQUIRED TPC fest Coordinator /Date 3.0 Test results reviewed. Simulator perfcrmance is satisfactory. f[
/ Simulator Instructor ffh0 / MId( 3 Zb 7/
Operations Representative /Date
/Date a fl
^
wh b Q [~f mulator* ng neer /Da e APPROVED: , l- Co , m DATE: 9[It/4/' Fqstkaic i (') .
TEST NUMBER / TITLE: PTM16/ MAIN GENERATOR 1 RIP ANS!/ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(16) DATE PERFORMED: 07/17/91 TEST DURATION: 1/2 hou- _ TEST DESCRIPTION: The simulator is operating at -40% power when the main generator output breakers open. The ICS automatically runs the reactor and feedwater back and ntabilizes the plant at 20% power, with steam generators on low level limits and header pressure being controlled at 885 psig by the turbine bypass valves. Manual operation of the PORV during the early part of this transient my be necessary to prevent a reactor trip on high RCS pressure. tLALDLNCJJ0!{ UILE OPll0!i$ TfP8X61T Bkr 1661 Fail Open TFP8X621 Bkr 1662 Fail Open INITIAL CONDITIONS: Reactor at 40% power, 6900V Busses powered from the S/U Transformer FINA'. CONDITIONS: Plant stabilized with the reactor at 20% power, steam generators on low level limits, header pressure being controlled at 885 psig by the turbine bypass valves. O etAN1 eROCEouRES USEo oUaiNo 1NE 1ES1: AP-545 REV 3 PLANT RUNBACK BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. V
$ PERFORMANCE TEST NUMBER: PIM16 DATE: 07/17/91 TITLE: MAIN CENERATOR TRIP TEST RESULTS/ COMMENTS:
Test results evaluation based on expert judgement. All test procedure steps functioned properly. Simulator response excellent.
,o i
I I L_,1a
n TEST NUMBER / TITLE: PTM16/ MAIN GENERATOR TRIP C' TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in .9375 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
' *L6JPR20B RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe L1:6ADT delta T Loop-A Grfx Rcdr 0 - 50F L1:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520.- 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op level Grfx Rcdr 0 - 100%
*LIJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi L1JPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi O *LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR d *LlJPS302 SG-B MFW Flow
*L1JPS329 SG-A SU FW Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500.MLB/HR GGG03N Main Turbine Speed Grfx Rcdr 0 - 2000 rpm L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm
O etaroan^"ct test aussta: etnis o^'t: o7/i7/91 TITI.E MAIN GENERATOR TRIP TEST STATUS: i 1.0- This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _LGL ,-
, . .c ' q r +1 9 'l,ft1Y est Coordinatoh /Date k
2.0 All deficiencies have been resolved and this test has been completed. O d NoTHIkkkHim FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. LLLv 4/Dhte v MA Sr. w9 / Operation 3 Representative /Dat6 SimglatorInstructor oltLU W W Ebg)/ pimulatorEngineer / ate APPROVED: /$ [b
,,R$TS/NLdTM DATE:
7'ub_ p - .-. V s
l i l i ( TEST NUMBER / TITLE: PTM17A/ CONTROL ROD OUT INHlBli FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) DATE PERFORMED: 07/02/91 TEST DURATION: I hour TEST DESCRIPTV)N: With the simulator operating at 100% power, a group 6 control rod drops into the core, resulting in an asymmetric rod runback. Due to relay failures within the control rod drive system, the out inhibit signal which would normally exist under these conditions does not occur. MLFUNCTION TITLE OPTIONS 1 CR-008 Block Control Rod Out Inhibit INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in se rv i ce.. FINAL CONDITIONS: Reactor power at <60%, with a droppped rod condition in effect. PLANT PROCEDURES USED DURING THE TEST: AP-545 REV 3 PLANT RUNBACK BASELINE DATA: There is no plant data for this transient. Expected simulator tC response based on expert judgement. O
DATE: 07/02/91 Q PERFORMANCE TEST NUMBER: PTM17A TITLE: CONTROL R00 OUT If1HIBIT FAILURE TEST RESULTS/C0MMENTS: Test results evaluation based on expert judgement. All steps functioned correctly. A problem was noted with step M17A020(c) "ASSYMETRIC ROD rut 4BACK" Alarm. Alarm point cycled in arid out continously as the runback limit was reached. ULO was cycling 15 - 10%. All other simulator response war. good. Trouble Report written on cyclic alarm problem,1R does not require rerun of the test. TR-0196 cleared 08/14/91 O O
TEST NUMBER / TITLE: PTM17A/ CONTROL R00 OUT INHIBIT FAILURE T(]} TEST DATA 1 LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter - DEVICE CHART-SPEED R3JDP208 Power Range Indication Grfx Rcdr 0 - 125% 3:75 see L 3R3JPP209 Power Range Indication Grfx Rcdr .0 - 125%
R3JPP210 Power' Range Indication Grfx Rcdr 0 - 125% p R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% HCTAVG RCS Average, Temperature Grfx Rcdr 520 - 620F L L5G7AV Rod Group 7 API Grfx Rcdr 0 - 100%
'L3:!CS25 Neutron Error Grfx Rcdr + 10 L3:ULMM Megawatt Demand Grfx Rcdr 0 - 1000 MWe V
O i
O PERFORMANCE TEST NUMBER: PTM17A DATE: 07/02/91 TITLE: CONTROL ROD OUT INHIBIT FAILURE TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _TR00.116 B __07/02/91_ 08/14/91 s
!}
N_ . o, ~
'l lM' T TestC6erqinator /Date 2.0 All eficiencies have been resolved and this test has been completed.
O -
'( m ...dc S-N S& '1\ Rbl TPC Test Coordinator /date 3.0 Test results reviewed. Simulator performance is satisfactory.
G N- & t tfyl ,q'/ \* N YW I Spulator Instruch_ /Dat Operations epresentative/Date
)
Ga2Ud} iCbA F/4F/v./ !! Sisulator Engineer /Date V -, , APPROVED: 5 f
/ DATE: ((
O l
TEST NUMBER / TITLE: PTM178/ CONTROL ROD GROUP COMMAND ENABLE FAILURE V ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) DATE PERFORMED: 07/02/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the situulator operating at 100% pcwer, a load decrease to 90% power is initiated with the ICS in full automatic. During the load decrease, the command enable for control rod group 7 is lost, resulting in an inability of group 7 to respond to either automatic or manual commands. MALFUNCTION TJJig OPTIONS CR 005 Loss of Control Rod Group Command Selectable to Group 5, 6, 7, 3 Enable and/or the auxiliary programmer INITIAL CONDITIONS: Full p er steady state conditions, all plant equipment in service. FINAL CONDITIONS: 90% power with group 7 transferred and operating on the auxiliary power supply. PLANT PROCEDURES USED DURING THE TEST: OP-502 REV 27 CONTROL R00 DRIVE SYSTEM
\
BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. V
1 i l f- s PERFORMANCE TEST NUMBER: PIM17B DATE: 07/02/91 i
\ TITLE: CONTROL ROD GROUP COMMAND ENABLE FAILURE TEST RESULTS/ COMMENTS: y Test results evaluation based on expert judgement. Test results satisfactory. No $
pei'ormance problems noted. Annunciator Sequence of Events typer was off-line during this test. All annunciator alarms functioned properly during the test. Rochester Event Points were verified via Rochester CRT display. O i I v
TEST NUMBER / TITLE: P1H170/Co*: TROL R00 GR0VI' COMMAhD ENABL E FAILURE O TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SlHULATOR RESPONSE ADEQUATELY DOCUMENTED BY TEST PROCEDURE O
1
g PERFORMANCE TEST NUMBER: P1H17B DATE: 07/02/91 TITLE: CONTROL RUD GROUP COMMAND ENABLE FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if .one so indicate): TR # Priority Date issued Date Closed kOE _ __ k., d .s a. 7h / TFCDEstco$_rdibator /llH(e 2.0 All deficiencies have been resolved and this test has been completed, b NOT Rr.QL t 1) TPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory, i lw.3 /) E l ] $ k -ped 1.Y
- tor Instructor SiniuQ(la 7 bate Operation Representative /Date m,
'd' C ete=de _
APPROVED: Mic# s/NLO M,' 2 / DATE: _['d O
C! TEST NUMBER / TITLE: PTM17C/ FAILURE Of THE MAIN 10RBINE 101 RIP ANS!]ANS3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) _DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, a spurious reactor trip occurs. The main turbine fails to automatically trip as a result of the reactor trip, and cannot be tripped manually from the main control board, in accordance with the applicable Abnormal Procedure, the Main Steam isolation Valves are closed to prevent an excessive post trip cooldown, GV's are closed by using the f ast close PB on the EHC Panel. ljAlf0!iCJ193 IJ1LE O_PTIONS EH 002 Main Turbine Trip failure Selectable to a failure of the automatic and/or manual trip circuitry _ INITIAL CONDITIONS: Full _ power steady state conditions, all plant equiprent in service. FINAL CONDITIONS: Reactor tripped, MSIVs closed, Plant stabilized in the normal post trip window with header presture being controlled by the atmospheric dump valves. Turbine governor valves closed and the main turbine output breakers open. PLANT PRJCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR 1 RIP BASELINE DATA: There is no plant data for this transient. Expected simula or response based on expert judgement,
1 PERFORMANCE TEST NUMBER: PTM17C DATE: 07/03/91 TITLE: FAILURE Of THE MAIN 10RBlf4E TO TRIP ,, TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned normally. Simulatur response excellent. O O
1 l I I TEST NUMBER / TITLE: P1M17C/ FAILURE Of 1HE MAIN TURBlNE TO TRIP TEST DATA LADEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in .9375 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 pst
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125% l PGJPE202 Gross Generation Grfx Rcdr 0 1000 Hwe ll:6ADT delta T Loop A Grfx Rcdr 0 - 50f Ll:6BDT delta 1 Loop B Grfx Rcdr 0 - 50f
*LlJPR212 1h loop-A Grfx Rcdr 520 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr $20 - 620f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 620f
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*f3JPA323 SG A Press (ErlC) Grfx Rcdr 0 - 12C3 psi
*f3JPA320 SG 0 Press (EflC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
/'S *LlJPS301 SG-A MfW flow Grfx Rcdr 0 - 6K KLD/HR V *LlJPS302 SG-B MfW flow Grfx Rcdr 0 - 6K KLD/HR
*LlJFS329 SG-A SU fW flow Grfx Rcdr 0 1500 MLB/HR
*LIJPS330 SG B SU FW flow Grfx Rcdr 0 - 1500 MLD/HR L7:1070 GV-1 Position Grfx Rcdr 0 - 100%
L7:1080 GV-2 Position Grfx Rcdr 0 - 100% L7:1090 GV-3 Position Grfx Rcdr 0 - 100% L7:10A0 GV-4 Positinn Grfx Rcdr 0 - 100% l l l V
DATE: 07/03/91 C PERFORMANCE TEST NUMBER: PIM17C TITLE: FAILURE OF THE MAIN 1URBINE TO 1 RIP TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _flDE
. it ,_ .4 4 TR31Etcoordinator / d[e<t(
2.0 All deficiencies have been resolved and this test has been completed. O NOT IU:QU mr.D TPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
.3S-n t Q alcy (Ld l h .< T It.-9l OperaIT5ns liepresentitive/DiT5
~
b ulator insttuctor / bate s b $/ h 9l' jz{kt imu ator lngineer /Da e syf-A, , APPROVED: #2' b ~ DATE: 9 af", .
,efiSTS/NL1H-
/
i- -2E'
p TEST NUMBER / TITLE: PTM17D/ ATMOSPHERIC DUMP VALVE FAILURE V ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) DATE PERFORMED: 07/05/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, the B OTSG Atmospheric Dump Valve, MSV 26, fails open. The resultant increase in steam demand causes and increase in the feedwater and reactor demand signals generated by the 105. Since the increased steam demand (approximately 3.75%) cannot be fully met, turbine header pressure, generated megawatts and RCS Tave decrease until equilibrium conditions are met. MLFUNCTIM 11J11 OPTIONS MS-005 Atmospheric Dump Valve r a ils to Selectable to MSV-25 and or -26 Position Adjustable from 0 100% position INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Simulator at 103% power, feedwater limited by S/G High Level Limits. Generated megawatts and RCS Tave stabilized at less than normal full powar values. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator ' response based on expert judgement. l l l l l l 1 l l O l 1
, ,- - , - -- r
PERFORMANCE TEST NUMBER: PIM170 DATE: 07/05/91 O TITLE: A1HOSPHERIC OUMP VALVE FAILURE _ TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgernent. All test steps functioned normally. Simulator response excellent. RPS Channel B High flux Trip B1 stable tripped due to low setpoint. All other responses normal.
~
O
~ ~ ~
O
g TEST NU_MBER/ TITLE: P1M170/A1MOSPHERIC DUMP VALVE FAILURE ( TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Redr 0 - 320 in .9375 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Redr 1700 2500 psi
*L6JPR208RCSPre.isure(WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grfx Redr 0 - 1000 Mwo ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 1200 psi 1.lJPT228 MS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS30) SG-A MFW Flow Grfx Rcdr 0 6K KLB/HR O> *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU FW Flow Grfx Rcdr 0 1500 MLB/IIR L5G5AV Group 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Group 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Group 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 3000 ppm O '
l l C PERFORMANCE TEST NUMBER: PlH17D DATE: 07/05/91 TITLE: ATHOSPHERIC DUMP VAtVE FAltuRE __ TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed Jintil m 4 2hjfi/ IP{fesN(~bdhator oor /Date 2.0 All deficiencies have been resolved and this test has been completed. O - mmomo UC lest Coordinator 7 bate 3.0 Test results reviewed. Simulator parformance is satisf actory.
\ ,
( ~.JC , -1 9fo/4/ j)_.6 l t.[ _ '6- ( $ D$ 5Iihulator Instructor /Date Bperations)hepresentative/Date
,m'I
, y-h /
- e g- ullI ,
APPROVED: DATE: c'Y w . Nilc -bh
~~
O
p TEST NUMBER / TITLE: PIM17E/ TURBINE BYPASS VALVE FAILURE V ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when a reactor trip occurs. During the transient, turbine bypass valve MSV-9 on the OTSG fatir closed. The plant is stabilized in the normal post-trip envelope with header pressure being maintained by the operable bypass valves. MALFUNCTLQH IIILE OPfl0NS MS-006 Turbine Bypass Valve Fails to Selectable to MSV-9, -10, 11 and/or Position -14 Adjustable from 0 100% position lHITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, simulator stabilized in the normal post trip envelope with header pressure being maintained at 1010 psig by the operable TBVs and 0TSG levels being maintained at low level limits by main feedwater. PLANT PROCEDURES USED DURING THE TEST: AP-580, REVISION 15 REACTOR TRIP BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O b
n, PERFORMANCE TEST NUMBER: PTH17E DATE: 07/03/91 TITLE: 1URBINE BYPASS VALVE FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned normally. Simulator response excellent. O O
l l 1 TEST NUMBER / TITLE: PTM17E/1URBINE DYPASS VALVE FAILURE TEST DATA ) LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grix Rcdr 0 - 320 in .5 sec *HlJPR203 Pressurizer Temperature Grfx Redr 0 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE201 Gross Generation Grfx Rcdr 0 1000 Mwe Ll:6ADT delta T Loop A Grfx Rcdr 0 50F L1:6BDT oe.ta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop A Grfx Rcdr 520 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F *LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG B SU Level Grfx Rcdr 0 250 in *LlJPS284 SG A Op Level Grfx Rcdr 0 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi O *LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR V *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 1500 MLB/HR *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 1500 HLB /HR SAW41Cll MSV-10 Flow Grfx Rcdr 0 - 150 LB/SEC SAW41Cl2 M3V-9 Flow Grfx Rcdr 0 - 150 LB/SEC SAW40C51 MSV-ll Flow Grfx Rcdr 0 - 150 LB/SEC SAW40CS2 MSV 14 Flow Grfx Rcdr 0 150 lb/SEC n ~
g PERFORMANCE TEST NUMBER: P1H17E DATE: 07/03/91 TITLE: TURBlfjlBYPASSVALVEFAILURE TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
._l10!E _ _ _ _
k_. .g m TTC)estToordindtor /Da}te7f3 3/ 2.0 All deficiencies have been resolved and this test has been completed. P 'b/ NOT Hl: QUI RI:D FPC Test Coordinator /Olite' 3.0 Test results reviewed. Simulator performance is satisfactory, e
.( br~
. <n 'ib t k'1 '4 b 4 b' N "H
_Twilator Si Instrudtor /Dath Operations) Representative /Date ( / j' l ator a EngiiFeer a w n r[a 'fdate b (./ om: .dj/ez Ageamo:ggg;gv70
O 'ts' "u" Bra /'it'c' "'">7r/co"'i'iu0uS R0D wiTnDRnwnt ACCiDtri1 ANSI /ANS-3-5(1985) CROSS
REFERENCE:
Section 3.1.?(17) DATE PERFORNED: 07/16/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: During a startup with the reactor at 10E-08 amps, control rods are withdrawn to commence a power escalation. Due to t failure in the control rod drive system, a continuous rod witharawal occurs. As reactor power increases above the point of adding heat, a rapid RCS heatup and resultant pressurizer insurge and RCS pressure increase causes an reactor trip on high RCS prcssure. iiAlf0HCT10!i 11111 OPTIONS XHOCD0VL Panel Override - f all Diamond Panel Insert / Withdraw Switch in the Withdraw Position INITIAL CONDITIONS: Reactor stable at 10E 08 amps, plant startup in progress. FINAL CONDITIONS: Reactor trippeo, simulator stabilized in the normal post trip envelope. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement extrapolation of FSAR analyses. O 1 1
PERFORMANCE TEST NUMBER: pTH17F DATE: 07/16/91 TITLE: CONTil4UOUS ROD Wi1HDRAWAL ACC10EN1 TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement extrapolation of FSAR analyses. All test procedure steps functioned normally Simulator response excellent. Graphic recorder menu 5 had a scaling error. S/U FW flows were scaled at 0 1500 MLB/HR. They should have been 0 - 1.5 MLB/HR. Graphic recorder shows S/U flows at 0; flows actually increased (as required by the procedure) as indicated by OTSG level responses on on frame 2; feedwater flow increase also verified during the test by main control board indications. O
--- ap g O
TEST NUMBER / TITLE: PTH17F/CONilNU005 ROD WITHDRAWAL ACCIDENT TEST DATA _ LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 - 320 in .9375 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*RRSWTP Normalized Reactor Power Crfx Rcdr 0 - 125%
R3JPP212 Intermediate Range Amps Grfx Rcdr -11.0 - 3.0 amps R3JPP213 Intermediatt Range Amps Grfx Rcdr 11.0 - 3.0 amps R3SURN13 Intermediate Range SUR Grfx Rcdr +5 DPM R35 URN 14 Intermediate Range SUR Grfx Redr +5 DNi
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 620f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f
*LlJPR216 Tc Cold Leg B1 Grfx Rcdr 520 620f
*LlJPS287 SG A SV Level Grfx Rcdr 0 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 HS Loop A Hdr Press Grfx Rcdr 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
*LlJPS329 SG-A SU FW Flow Grfx Redr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR v
O
{ PERFORMANCE TEST NUMBER: P1H17F TITLE: CONTINUOUS R00 WITHDRAWAL ACCIO[NT , DATE: 07/16/91 TEST STATUS. l 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed l M i L l
. - _ j
()
.. l be TR Tiest Coordingtor m~ 'llP IY 7 bate ,
u 2.0 All deficiencies have been resolved and this test has been completed. NOT REQUlHED TPFTest Coordinator /D~ ATE 3.0 Test results reviewed. Simulator performance is satisfactory. O s1 1 S,- u,N MTp %. m s/ SjmulatorInstrudtor / bate' 'perationfRepresentatlve/Date O r \\ - fI -y ~~ / inu tor Engineer e APPROVED:
,lifisTNL6H /
s/ DATE: g/ 'j_ O
i ((( TEST NUMBER / TITLE: PIM17G/DlLUT10N ACCIDENT ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(17) DATE PERf0RHED: 08/13/91 TEST DURATION: 2 hours TEST DESCRIPTION: With the rimulator operating at 100% power, seat leakage develops past DWV-128, MVV-103 and MUV-108. The resultant flow of domineralized water to the RCS results in a continuous dilution, causing reactor power to increase. This increase re suits in the development of a neutron error in the ICS, causing the control rods to periodically insert to maintain reattor l power at setpoint. I KALEMMall0H IIILE OPTIONS l TVBVL108 MJV-108 Seat Leakage Adjustable from closed to full open TVBVL103 MUV-103 Seat Leakage Adjustable from closed to full open TVK1128L DWV 128 Seat Leakage Adjustable from closed to full npen I INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in i service. FINAL CONDITIONS: Full power conditions with the dilution terminated. RC5 boron lower than normal MOL value; Control rod positions lower than rs normal for full power conditions. V PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement extrapolation of FSAR analyses. O __
{' PERFORMANCE TEST NUMBER: PTH17G DATE: 08/13/91 TITLE: DILUTION ACCIDENT TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement extrapolation of FSAR enalyses. Test results satisfactory. No performance problems noted. A U i l C '
q p TEST NUMBER /TITI.Es PIM17G/DilVT10N ACCIDENT TEST DATA LABEL DESCRIPTION RECORDitlG RANGE FREQUENCY / 0 Critical Parameter DEVICE CHART SPEED HQBBCORE RCS Boron Concentration Grfx Rcdr 925 - 965 ppm 5.025 sec BMBMUllL Makeup lank Boron Grfx Rcdr 300 1000 ppm N2JPX359 Makeup Tank Level Grfx Rcdr 0 100 in BMW6315 DW flow to the Makeup lank Grfx Rcdr 0 - 8 gpm RRWSTP Normalized Reactor Power Grfx Rcdr 95 - 105% L3:lCS25 Neutron Error Grfx Rcdr -10 4 10 L5G7AV Control Rod Group 7 Position Grfx Rcdr 0 - 100% RAREAC Core Reactivity Grfx Rcdr .01 4.01 dK/K C
. _ _ --.x=
C PERFORMANCE TEST NUMBER: PIM17G DATE: 08/13/91 TITLE: DILUTION ACCIDENT TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed llDEL -. m l$
@ll3ht/
flblest~ oordinator ' /DiifE 2.0 All deficiencies have been resolved and this test has been completed. O D NOT RI:QUIRI:D TPCTist Girdinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
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Siiiiilat9 Tnstructor /D' ate OperatF6FRepresentatN/Date 3 < gj$eddir; %F APPROVED: DATE: ~3[tt / gj~ j Eip7 Relit ~M*\ o~h Ar- -[ 9 O a
TEST NUMBER / TITLE: PIM18A/ MAKEUP PUMP TRIP O' ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2t M) . DATE PERFORMED: 01/15/91 TEST DURA 1,UN: I hour TEST DESCRIPTION: With the simulator operating at full power, the normal running makeup pump MVP-1B motor trips causing a loss of RC makeup and RCP seal injection. Makeup tank level begins increasing and pressurizer level begins decreasing due to the loss of the pump. MVP-1A is started and seal injection and makeup are restored. MALFUNCTION LLILE OPTIONS MU-001 Hakeup Pump Trip Selectable to MVP-1 A, -10 or -1C I INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. l l FINAL CONDITIONS: Full power steady state conditions, MVP 1B secured, MVP-1A in I operation. { PLANT PROCEDURES USED DURING THE TEST: OP-402 REV 66 O OP-302 REV 32 MAKEUP AND PURIFICATION SYSTEM RC PUMP OPERATION BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. O
. - - - - - _ . . ~ . _ . - . . - . . . . , . . . . _ - , - - - . - . - - . . . _ - -
j PERFORMANCE TEST NUMBER: PIM18A DATE: 01/15/91 O _ TITLE: MAKEUP PUMP TRIP TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement, lest results satisfactory. No performance problems noted. O O
e TEST NUMBER / TITLE: PTH18A/ MAKEUP PUMP TRIP TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec *HlJPR203 Pressurizer Temperature Grfx Rrdr 0 - 700f *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 25J10 ;,si N2JPX202 Letdown flow Grfx Rcdr 0 160 gpm N2:M10 Makeup flow Grfx Rcdr 0 200 gpm N2:Hil Total Seal Inj flow Grfx Rcdr 6 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 120 in t
t G O
O PERFORMANCE TEST NUMBER: PTH18A DATE: 01/15/91 TITLE: MAKEUP PUMP TRIP TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed
.30RE jf:f,[Zh kh, ff N" lest Coordinafor /Dati 2.0 All deficient.ies have been resolved and this test has been completed.
D .d NOT REQUIRID TPETest Coordinator /UW6 3.0 Test results reviewed. Simulator performanceis satisfactory.
/
jperations uih1 Jhe 'l 0A Tichresentative/Date / p,d mu a Q Qstructor
, '/fater&
,./l S mu)latorcerh5 ?2 f E& 2Ub /Date APPROVED: em kr DATE: 4 i g 'i]
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1 _ TEST NUMBER / TITLE: PIM18B/ PRESSURIZER LLVfL CONTROL VALVE FAILURE ANSl/ANS 3.5(1985) CROSS
REFERENCE:
Section 3.1.2(18) DATE PERFORMED: 07/02/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, MUV 31 fails closed. The resultant reduction in makeup flow to the RCS l causes pressurizer level to begin decreasing and makeup tank level to begin increasing. Letdown is reduced and MUV 30 l (MUV 31 manual bypass) is adjusted to reestablish pressurizer and makeup tank levels. M EU30llDB lilLI QPTIONS MU 003 MVV 31 Fail to Position Adjustable from 0 - 100% INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. MUV 51 positioned for 75 gpm letdown flow. FINAL CONDITIONS: Full power steady state conditions. MUV-30 positioned to balance makeup flow with letdown flow. Pressurizer level and makeup tank level stabilized at normal opercting values. PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: Ihere is no plant data for this transient. Expected simulator response based on expert judgement. O
PERFORMANCE TEST NUMBER: P1H18B DATE: 07/02/91 TITLE: PRESSURIZER LEVEL CONTROL VALVE FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All steps functioned properly. Simulator response excellent. No discrepancies noted. O
TEST NUMBER / TITLE: PlM188/ PRES $URIZER LEVEL CONTROL VALVE FAILURE
\ !
TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5,625 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi l
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi -
N2JPX202 Letdown flow Grfx Rcdr 0 - 160 gpm N2:M10 Makeup Flow Grfx Rcdr 0 200 gpm N2:Mll Total Seal Inj flow Grfx Rcdr 0 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in l I o _
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P V PERFORHANCE TEST NUMBER: PTM18B OATE: 07/0E/91 TITLE: PRESSURIZER LEVEL CONTROL VALVE FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issuea Date Closed _hDE _ B yt$ \_;+-.<n h 7lCzl FPC Test Cobtdindtor /Dat6 u 2.0 All deficiencies have been resolved and this test has been completed. (3 V NOT REQUIRED FPC Test Coordinator /Date
/
3,0 Test results reviewed. Simulator performance,4 satt actory. o at 1 v{ i phd
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Operations Representative /Date YijJlator Instructor /Date_ [ wv2d &- 9//xIf
/Dat'
(,5~hiulator Tngineer APPROVED: i DATE: 4
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,N565/NLO Ff O'
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l TEST NUMBER / TITLE: PTM18C/ LETDOWN ISOLATION VALVE FAILURE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(18) DATE PERFORMED: 07/02/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operatint, it 100% power, temperature switch MU-5-TS fails, resulting in en erroneous high letdown temperature signal being generated. Letdown isolation valve MUV-49 closes in response to false high temperature signal. The High Temperature Bypass Switch is used to overide the signal and letdown is restored. MALFUNC1108 TITLE QPTIONS MU-002 Letdown isolation Valve Closure Due to a Failure of MU-5-TS INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions, letdown restored. PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: There is no plant data for this transient. Expected sim"lator response based on expert judgement. O
PERFORMANCE TEST NUMBER: PTM18C DATE: 07/02/9; TITLE: LETDOWN ISOLATION VALVE FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All steps functioned properly. Simulator response excellent. No discrepancies noted.
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O O
I 1 TEST NUMBER / TITLE: PTM18C/ LETDOWN ISOLATION VALVE FAILURE TEST DATA LABEL DESCRIP1 ION RECORDING RANGE FREQUENCY / OCrit+:al Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in 2.8125 sec oHlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F SL4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi OL6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi N2JPX202 Letdown Flow Grfx Rcdr 0 - 160 gpm N2:M10 Makeup Flow Grfx Rcdr 0 - 200 gpm N2:Mll Total Seal Inj Flow Grfx Rcdr 0 - 80 gpm N2JPX359 Makeup Tank Level Grfx Rcdr 0 - 100 in l
O PERFORMANCE TEST NUMBER: PTH18C DATE: 07/02/91 TITLE: LETDOWN ISOLATION VALVE FAILURE TEST STATUS: 1,0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed NONE j p(
.-1 4 Je n-, y llet/
FPCLTest Coordinator / Cath a 2,0 All deficiencies have been resolved and this test has been completed, NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed, Simulator performance,is satisfactory. C ww C' L2f'q l ~ 0ifll! -fl0[~ Operations Representative /Date l' f/ iulator 1nstructor
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S / (% Mnid 9 lg A-pimulato?Engifieer '/Dafte APPROVED: _ N [ DATE: [gh[
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j kSTS/NLOTM O
,q TEST NUMBER / TITLE: PTM180/ PRESSURIZER HEATER CONTROLLER FAILURE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(18) DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, a failure in the pressurizer heater controller RC-3-Plc occurs, resulting in a constant full demand to heater banks A, B and C. As a result, RCS pressure increases until the actuation setpoint for the pressurizer spray valve is reached. RCS pressure begins to cycle between the open and close setpoints for the spray valve. MALFUNCTION TITLE OPTIONS RC-014 Pressurizer Heater Controller Adjustable from 0 - 100% demand on Failure heater banks A, B and C INITIAL COND!TIONS: Full power steady state conditions, all plant equipment in . service. FINAL CONDITIONS: Full power conditions with pressurizer heater banks A, B and C full on and RCS pressure cycling between the open and close setpoints for the pressurizer spray valve. Q PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. s i ( U
PERFORMANCE TEST NUMBER: PTM18D DATE: 07/03/91 TITLE: PRESSURIZER HEATER CONTROLLER FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned correctly. Simulator response excellent. O O
TEST-NUNCER/ TITLE: PTM18D/PRESSURIZEi! HEATER CONTROLLER FAILURE (,/ TEST DATA
-LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
*Criti:al Parameter DEVICE CHART SPEED L4JPR222 RCS A Loop Pressure Grfx Rcdr 1700 - 2500 psig 1.875 sec HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700 F HHWHIX(24) PRZ Spray Flow Grfx Rcdr 0 - 50 lbm/s O
PERFORMANCE TEST NUMBER: PIM18D DATE: 07/03/91 TITLE: PRESSURIZER HEATER CONTROLLER FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed __liDUf
.c.f e p~s A c qMS/
F est Co6rdina' tor /Date 2.0 All deficiencies have been resolved and this test has been completed. woT anoutaen TPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance )[ satisf actory.
/
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Si LE 5~ lato'r in$truttor s
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Gwe(tu M<A 6A/ Operations Representative /p'ite knno C k () / Sli '/Date h nulator'Engilieer APPROVED: 2 DATE: 7 __ l 0
TEST NUMBER / TITLE: PTH18E/ PRESSURIZER SPRAY VALVE FAILURE
\~ ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(18) ,_ DATE PERFORMED: 07/03/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 1007. power, the pressurizer I spray valve fails open. Due to the failure, RCS pressure begins decreasing. Due to a failure of the pressurizer spray block valve, isolation of the spray valve cannot be performed. A rapid load reduction to 75Y. power is performed and the B RCP pump is secured in order to reduce the motive force for pressuirizer spray flow. During the transient the plant may trip on variable temperature / pressure. MALFUNCTION IJJig OPTIONS RC-021 Pressurizer Spray Valve Failure Adjustable from 0 -1007. Valve Position INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. l FINAL CONDITIONS: 1) Plant at 757, power with the B RCP secured. Spray flow reduced a a flow rate which can be matched by the pressurizer heaters.
- 2) Reactor / turbine tripped, plant in the process of being stabilized in the normal post trip envelope.
G V PLANT PROCEDURES USED DURING THE TEST: l l BASELINE DATA: There is no plant data for this transient. Expected simulator I response based on expert judgement. i l l l l l l V
=.
PERFORMANCE TEST NUMBER: PTH18E DATE: 07/03/91 TITLE: PRESSURIZER SPRAY VALVE FAILURE TEST RESULTS/ COMMENTS: All test steps functioned correctly. Simulator response excellent. Reactor trip on Variable Temperature / Pressure occurred at 55%. Although this was the alternative expected final condition for the test, a reactor trip under these conditions is not an unexpected / incorrect response. f$ U
1 1 o TEST NUMBER / TITLE: PTM18E/ PRESSURIZER SPRAY VALVE FAILURE t U TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 5,625 sec
*HIJPR203 Pressyrizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi HHWMIX24 Pressurizer Spray Flow Grfx Rcdr 0 - 200 lbm/sec
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125% !
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe i L4JPR234 RCS Loop A Flow Grfx Rcdr 0 - 80 MLB/hr ; L4JPR235 RCS Loop B Flow Grfx Rcdr 0 - 80 MLB/hr !
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Redr 520 - 620F
*LIJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LIJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LldPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in I
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100% l l *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi l LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi L1pT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
. /] *L; C S301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR V *LluPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR l *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5GSAV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm 1 0v
(T v
; PERFORMANCE TEST NUMBER: PTH18E DATE: 07/03/91 TITLE: PRESSURIZER SPRAY VALVE FAILURE .
TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _NONE nc w +^ ll03 il 1
/Dhte FPC(TestCoordihator 2.0 All deficiencies have been resolved and this test has been completed, o
NOTREQdhED FPC Test Coordinator /Date
,1 3.0 Test results reviewed. Simulator performance if satisfactory. l
,/ /
m *\ G q % ,I(I i 1 I ' /._ ,lllh Ji i/ S alatorInstruegor /Date l Operations Representath e/Djtte
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3 mulatot*'Eng'tnedr 9kb
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APPROVED: DATE: /
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M STS/NLOTM / O
TEST NL'MBER/ TITLE: PTM18F/RCS NARROW RANGE PRESSURE TRANSMITTER FAILURE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(18) DATE PERFORMED: 01/14/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at 100% power when RCS Harrow Range Pressure Transmitter RC-3A-PT1 fails high (2500 psig). Due to the erroneous indication the pressurizer heater banks turn off, the spray valve opens to its 40% position in automatic and the PORV opens. As pressure begins decreasing, the spray valve and PORV are placed in manual and closed. The patch cord in the A RPS cabinet is shifted to the B side and the heaters, spray and PORV are restored to aut.omatic. MALFUNCTION 111LE OPTIONS RC-016 RCS Narrow Range Pressure Selectable to RC-3A-PT1, -PT2, Transmitter Failure RC-3B-PTl or -PT2 Adjustable from 1700 to 2500 psig INITIAL CONDITIONS: Full power stea 'y state conditions, all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions with RCS pressure control restored to full automatic and the A RPS cabinet patch cord selected to B. PLANT PROCEDURES USED DURING THE TEST: OP-501, REVISION 14 REACTOR NON-NUCLEAR INSTRUMENTATION BASELINE DATA: - There is no plant data for this transient. Expected simulator response based on expert judgement. O l I
PERFORMANCE TEST NUMBER: PTH18F DATE: 01/14/91 O TITLE: RCSNARROWRANGEPRESSURETRANSMITTER[AILURE TEST RESULTS/ COMMENTS: Test results evaluation based cn expert judgement. Test results satisfactory. No performance problems noted. O
/
O l
p TEST NUMBER / TITLE:PTM18F/RCS NARROW RANGE PRESSURE TRANSMITTER FAILURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED i
ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST S1HULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT O l l L 1
DATE: 01/14/91 Q PERFORMANCE TEST NUMBER: PTM18F TITLE: RCS NARROW RANGE PRESSURE TRANSMITTER FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _NONE
./
. _ ; /' agi j eft Coordinator /0 ate 2.0 All deficiencies have been resolved and this test has been completed, woT anouinno ~
FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is _. satisfactory.
/ _
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/5iinulator Instructor //Cate OperationsRepresentatife/Date o gg
- ic03 ' hl m. 9 z. /
/ lator'Tngiiieer '/Date APPROVED: b_q~p-n b
DATE: kl5 il NSTgNL0lM l ,
TEST NUMBER / TITLE::PTM20A/ MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(20), Appendix B, Section B2.2(9) DATE PERFORMED:'01/16/91- -TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at' 90% power, a steam leak is initiated on the A-2 main steam'line outside containment upstream of MSV-411. The inital break is equivalent to 1% steam flow, and increases to 5% steam flow over a five minute period. After ten minutes, the break size increases to 100% steam- flow. A reactor / turbine trip occurs, followed by main steam
. isolation and main feed isolation-of the affected OTSG. The RCS experiences a rapid cooldown and-depressurization which is terminated when the affected 0TSG boils dry.
MALFUNCTIO.fi TITLE OPTIONS MS-002 Main Steam Line Break Selectable to the A1, A2,-B1 or B2-Outside Containment Main Steam Line Adjustable from 0 - 20E+06 lbm/hr INITIAL-CONDITIONS: 90% power, all plant equipment in service. FINAL CONDITIONS: Reactor turbine tripped,- affected 0TSG isolated and boiled dry. RCS parameters stabilized in the normal post trip window, with 1 O -- the intact steam generator level control being maintained by emergency feedwater. PLANT! PROCEDURES USED DURING THE TEST: j AP-580 REV-15 REACTOR TRIP-AP-450.REV 15 EMERGENCY FEEDWATER ACTUATION l
.AP-460 REV.8 STEAM GENERATOR ISOLATION ACTUATION l- BASELINE DATA:- There -is'no plant data for 'this . transient. Expected ' simulator response based on a combination of expert judgement and extrapolation of FSAR data based on actual plant conditions vs design basis conditions.
1 O
PERFORMANCE TEST NUMBER: PTM20A DATE: 01/16/91 q TITLE: MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. A comparison of simulator response to FSAR data for a main steam line break shows a significant difference in the magnitude of the predicted vs actual RCS cooldown and depresurization resulting from the break. This difference appears to be primarily due to significant differences in assumed and actual RCS inventory and assumed and actual OTSG secondary side inventories at the start of the transient. The higher initial RCS inventory and lower OTSG secondary side inventories in the simulator result in a smaller cooldown and depressurization than predicted in the FSAR. Test results satisfactory. No performance problems noted, O O
g TEST NUMBER / TITLE: PTH20A/ MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT V TEST DATA LABEL' DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in 5.625 sec *HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi R3JPP208 Power Range Indication Crfx Rcdr 0 - 125%
R3JPP209 Power Range Indication Grfx Redr 0 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125%
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F *LldPR214 Tc Colo Leg Al Grfx Rcdr 520 - 620F *LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in *L1JPS293 SG-B SU Level Grfx Rcdr 0 - 250 in *L1JPS284 SG-A Op Level Grfx Rcdr 0 - 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi e *L1JPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
( *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LIJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR N2:MI8 HPI Flow Al- Grfx Rcdr 0 - 500 gpm N2:MI4 HPI Flow A2 Grfx Rcdr 0 - 500 gpm N2:MI2 HPI Flow B1 Grfx Rcdr 0 - 500 gpm N2:MI6 HPI Flow B2 Grfx Rcdr 0 - 500 gpm F3:23FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:25FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm O
V
. - , .- n, g PERFORMANCE TEST NUMBER: PTM20A DATE: 01/16/91 TITLE: MAlii STEAM LINE BREAK OUTSIDE C0f4TAINMEt4T TEST STATUS:
1,0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed l L{0_NL_ __ _ ._.__ e>_sw*-me-w=w+- l l l
' bbJ f f C Tes Coorifinator ~7 bate 2.0 All deficiencies have been resolved and this test has been completed.
I l NOT REQUIRED FPC lest Coordinator /Date 3.0 Test results reviewed. Simulator performance is .satisf actory.
- / ,
/5Tm//pr~1nstructor Sk l, \ ! l f f' ~ 1 t JIh l' Operations RepYiGentativ//bTte Y! l dlii ' /ba t'e' f "
ub./sf fator fndlh.;s eer e9Iu
/Da e h 4 i APPROVED: k m DATE: #t (L (l_L_
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TEST NUMBER / TITLE: PTM20B/ MAIN FEEDWATER LINE BREAK INSIDE CONTAINMENT V ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(20) DATE PERFORMED: 01/17/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 1007. power when a 3E+06 lbm/hr break is initiated on the main feedwater line to the A-0TSG inside containment. A-0TSG level begins rapidly decreasing and the resultant loss of heatsink causes a reactor trip on high RCS pressure. EFIC is actuated on A-0TSG on low level; EFW and main feedwater feed A-0TSG in an effort to restore level until terminated by main feedwater isolation. Safeguards actuation is initiated by high containment pressure. MALFUNCTION TITLE OPTIONS FW-001 Main Feedwater Line Break Adjustable from 0 to 5.0E+6 lbm/hr Inside Containment Selectable to the A or B Feedwater Line INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, A-0TSG isolated and depressurized. Plant stabilized in the normal post trip window with heat removal being supplied by the B-0TSG. Safeguards systams secured. PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-380 REV 17 ENGINEERED SAFEGUARDS ACTUATION AP-450 REV 15 EMERGENCY FEEDWATER ACTUATION AP-460 REV 8 STEAM GENERATOR ISOLATION ACTUATION BASELINE DATA: There is no plant data for this transient. Expected simulator response based on exoert judgement extrapolation of FSAR analyses ano AT0G discussion of Loss of Feedwater Accidents. O
PERFORMANCE TEST NUMBER: PTM208 DATE: 01/17/91 O TITLE: MAIN FEEDWATER LINE BREAK INSIDE CONTAINMENT TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement extrapolation of FSAR analyses and AT0G discussion of loss of feedwater accidents. Test results satisfactory, No performance problems noted. O O-
~
l TEST NUMBER / TITLE: PTM20B/ MAIN FEEDWATER LINE BREAK INSIDE CONTAINMENT TEST DATA-
. LABEL- DESCRIPTION- RECORDING RANGE- FREQUENCY / ,
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HlJPR203 Pressurizar Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi -
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0.- 2500 psi R3JPP208 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP209 Power _ Range Indication: Grfx Rcdr 0 - 125%
'R3JPP210 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP211 Power Range Indication ' Grfx Rcdr 0 - 125% -,
*LlJPR212 Th Loop-A -Grfx Rcdr 520 - 620F l L
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 --620F
*L1JPR215 Tc_ Cold Leg A2- Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 250 in L *L1JPS293~SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level' Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 -_100%
l *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi , l
*F3JPA320 SG-B Press (EFIC)- Grfx-Rcdr 0 - 1200 psi i LIJPT228 MS Loop-A-Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Kdr Press Grfx-Rcdr 600 - 1200. psi
.O
- Grfx Rcdr .0-- 6K KLB/HR V *L1JPS301-LIJPS302 SG-B
*LlJPS329 SG-A SU FW F1ow SG-A MFWMFW-Flow - Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 1500 MLB/HR :
L *LlJPS330 SG-B SU FW Flow Grfx kcor 0 - 1500 MLB/HR j ClJPP254(Containment Pressure .Grfx Rcdr 0 - 70 psi l Containment Temperature Grfx Rcdr 50 - 300F CATAVG ' CALNS Normal Sump Level. Grfx.Rcdr 10-FT-CALES- Emergency Sump. Level- Grfx Rcdr 0 - 10 FT N2:MI8 HPI Flow A1 Grfx Rcdr. 0 - 500 gpm l N2:MI4 HPI Flow A2 Grfx Rcdr 0 -.500 gpm L N2:MI2 -HPI F1ow Bl. Grfx Rcdr 0 --500.gpm i' N2:MI6-iHPI Flow B2 Grfx Rcdr _0 - 500 gpm F3:23Fil EFW Flow S/G A - Grfx Rcdr 0 - 1000 gpm F3:25FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A. Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow.S/G B Grfx Rcdr 0 - 1000 gpm O r
g PERFORMANCE TEST NUMBER: PTM20B DATE: 01/17/91 TITLE: MAIN FEE 0 WATER LINE BREAK INSIDE CONTAINMENT TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
,_NONE qe..hh em.ee.e w
._ A)2 ' A7 '
,F C Tes Coordinator T/date 2.0 All deficiencies have been asolved and this test has been completed.
(_/ NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performange is satisfactory. _
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6fk Aim'ulator instructor r ,
/Date
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i 6 TEST NUMBER / TITLE: PTM20C/ MAIN FEEDWATER LINE BREAK OUTSIDE CONTAINMENT U ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(20) DATE PERFORMED: 01/16/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100Y. power when a 4E+06 lbm/hr break is initiated on the main feedwater line to the A-0TSG outside containment upstream of check valve FWV-45. A-0TSG level begins rapidly decreasing and the resultant loss of heatsink causes a reactor trip on high RCS pressure. EFIC is actuated on A-0TSG on low level. As level and pressure in A-0TSG are restored by emergency feedwater, FWV-45 closes. Main feedwater to A-0TSG is manually isolated in accorda':ce with applicable APs. MALFUNCTION UIjl OPTIONS FW-002 Main Feedwater Line Break Adjust ule from 0 to 5.0 E6 lbm/hr Outside Containment Selectable to A or B Main Feedwater Line INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped. Main feedw ter to A-0TSG secured, maintaining normal post trip pressure and level in A-0TSG with emergency feedwater. V PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-450 REV 15 EMERGENCY FEEDWATER ACTUATION AP-460 REV 8 STEAM GENERATOR ISOLATION ACTUATION BASELINE-DATA: There is no plant data for this transient. Expected simulator response based on expert judgement extrapolatien of FSAR analyses and AT0G discussion of Loss of Feedwater Accidents. l O
PERFORMANCE TEST NUMBER: PTM20C DATE: 01/16/91 V TITLE: MAIN FEEDWATER LINE BREAK OUTSIDE CONTAINMENT TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement extrapolation of FSAR analyses and AT0G discussion of loss of feedwater accidents. 'est results satisfactory. No performance problems noted. i 1
n TESTNUMBER/T!TjE: PTM20C/ MAIN FEE 0 WATER LINE BREAK OUTSIDE CONTAINMENT b TEST DATA LABEL DESCRIPTION hiCORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi R3JPP208 Power Range Indication Grfx Rcdr 0 - 125%
R3JPP209 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125%
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*L1JPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*L1JPS293 SG-B SV Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR O)
\ *LIJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*L1JPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR F3:23FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:25FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:26FIl EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm
()
PERFORMANCE TEST NUMBER: PTM200 DATE: 01/16/91 TITLE: MAIN FEEDWATER lil4E BREAK OUTSIDE C0l41All4MElfi TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if nt,nc so indicate): TR # Priority Date issued Date Closed lf0!ll _ _ _ _ . f ator 9$ 2.0 All deficiencies have been resolved and this test has been completed, o D NOT REQUllmD TIiC~ Tis ~t Coordinator /Date 3.0 Test results reviewed. Simulator performance is . satisfactory. l ( {Y{L?d fll 'N )'l:"l N ~/(L(' ] > f/ e' S 'mu i(tor Jnstructor ' (^/bli,ts 6perations flepresentativb/ 5 e l ,/
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p TEST NUMBER / TITLE: P1H21A/ POWER RANGE CHANNEL FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(21) DATE PERFORMED: 01/13/91 TEST DURATION: I huur TEST DESCRIPTION: The simulator is operating at 100Y. power when the summing amplifier for power range instrument N10 fails high, resulting in a 125Y. power output signal from than char.nel. The resultant neutron error seen by the ICS causes the control rods to begin inserting and main feedwater to increase due to a cross limit condition. The feedwater increase is limited when the steam generators reach their high level limit. The rod insertion is terminated when the neutron error is cancelled by the Tave-error resulting from increased feedwater flow and reduced power. The overfeed and low Tave also combine te reduce the amount of main steam superheat; the turbine governor valves 90 full open to compensate for the resultant loss of megawatts. HAlfuNCT10ff IL1M QP110NS RP-006 Power Range Summing Amp Adjustable from 0 - 1257. power failure Selectable to N! 5. -6. -7. or -8 INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. g f!NAL CONDITIONS: Plant stabilized at -807. power. Tave -562f. Steam generator s levels at high limit. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement.
PERf 0PdANCE TEST 1: UMBER: PlH21A DATE: 01/13/91 TITLE: POWER RANGE CilAf4NEL FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgenent. Test results satisfactory. No performance problems noted. O
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=
1 g TEST NUMBER / TITLE: PTM21A/ POWER RANGE CHANNEL FAILURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY / ,
- Critical Parameter DEVICE CHART SPEED l
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
'H1JPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll: 6ADT delta-T Loop A Grfx Rcdr 0 - 50 f L1:6BDT delta-T Loop B Grfx Rcdr 0 - 50 f
*LlJPR212 Th loop A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f l
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f l
*L1JPS287 SG A SU Level Grfx Rcdr 0 - 250 in l
*LlJPS293 SG-8 SU Level Grfx Rcdr 0 - 250 in ;
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 1200 psi LlJPT228 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301-SG A HFW Flow Grfx Rcdr 0 - 6K KLB/HR Os *LlJPS302 SG B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU FW flow Grfx Rcdr 0 - 1500 MLB/HR L5GSAV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm R3JPP208 Power Range Indication Grfx Rcdr 0 - 125% R3JPP209 Power Range Indication Grfx Rcdr 0 - 125% R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% O - -- w ,y-ww w w
{ PERFORMANCE TEST NUMBER: PlH21A DATE: 01/13/91 TITLE: POWER RANGE CHANNEL FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _lML
/ W/) 9 6[
lPfl)s(Co6'rdinator ff e //D' ate f 2.0 All deficiencies have been resolved and this test has been completed, t NOT lli:QUII(ED DC Test Coordinator /Date 1 3.0 Test results reviewed. Simulator performance is' satisfactory'.
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TEST NUMBER / TITLE: PIM218/ INTERMEDIATE RANGE CHANNEL FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(u ) DATE PERFORMED: 01/13/91 TEST FeJRATION: 1/2 hour TEST DESCRIPTION: The simulat r is operating at 100% power when a reactor trip occurs. Due to a failure of the A Intermediate Range compensating voltage power supply, the detector output is undercompensated. This results in the intermediate range post trip indications tracking together until approximately E-09 amps decreasing. At this point, the A channel indication begins decreasing at a slower rate than the B channel; this is also reflected in a corresponding difference in the SUR indications for the two intermediate range channels. MALFUNCTION lilLE OPTIONS RP-003 IR Compensation Power Supply Adjustable from 0 to 100% of the Failure nominal power supply output Selectable to the A or B detector INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Plant stabilized in the normal post trip window. Power in the source range and decreasing. B intermediate range indication at E-Il amps, A indication still on scale and decreasing. PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert ju!gement.
. . - - - . . - _ . . _ -. .-- -. . . - . . = _ - . - - . . . -
PERFORMANCE TEST NUMBER: PIM21B DATE: 01/13/31 ' TITLE: INTERMEDIATE RANGE CHANNEL FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. O v O
g TEST NUMBER / TITLE: PTM218/INTERME01A1E RANGE CilANNEL FAILURE U _ TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CilART SPEED R3JPP208 Power Range Indication Grfx Rcdr 0 - 125% 1.875 sec R3JPP209 Power Range Indication Grfx Redr 0 125%
R3JPP210 Power Range Indication Grfx Rcdr 0 - 125% R3JPP211 Power Range Indication Grfx Rcdr 0 - 125% R3JPP212 Intermediate Range Amps Grfx Rcdr 11.0 -3.0 amps R3JPP213 Intermediate Range Amps Grfx Rcdr 11.0 - 3.0 amps R3SURNI3 Intermediate Range SUR Grfx Rcdr +1 DPM I R3SURN14 Intermediate Range SUR Grfx Rcdr 1 41 DPH R3JPP202 Source Range CPS Grfx Redr -1.0 - 6.0 cps R3JPP203 Source Range CPS Grfx Rcdr -1.0 - 6.0 cps R3SURNil Source Range SUR Grfx Rcdr +1 DPM R3SURN12 Source Range SUR Grfx Rcdr -1 + +1 DPM G V P O
i DATE: 01/13/91 O PERFORMANCE TEST NUMBER: PIM21B TITLE: INTERMEDIATE RANGE CHANNEL FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.HQRE 7, ,
/'
f -/ '/ I CEdinator 7 big ff 2.0 All deficiencies have been resolved and this test has been completed. O NOT Hl QUllWD , TFClest coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. r- \; Y ' \ ; YT / 1/h l$ )9 II)6
/TimuTattr7iistructor Operations RepresentdW/LiaU '
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TEST NUMBER / TITLE: PTM210/ SOURCE RANGE CHANNEL FAILURE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(21) DATE PERFORMED: 01/13/91 TEST DURATION: 1/2 hour , TEST DESCRIPTION: While performing a reactor startup, the A source range detector high voltage power supply fails to 0 volts. MALFUNCTION llJig OPTIONS RP-002 Source Range Detector High Selectable to the A and/or B Source Volts Power Supply Failure Range Detector INITIAL CONDITIONS: Reactor Startup in progress, power at 1000 cps on the source range. FINAL CONDITIONS: Reactor stabilized at 100 cps on the source range with the A source range detector failed to 0 cps. PLANT PROCEDURES USED DURING THE TEST: OP-210 REV 29 7. ACTOR STARIUP BASE'.INE DATA: There is no plant data for this transient. Expected simulator
- response based on expert judgement.
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PERFORMANCE icST NUMBER: PTH21C DATE: 01/13/91 TIILE: SOURCE RAtlGE CHAfitiEL FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performanca problems noted. 4 O O-
TEST NUMBER / TITLE: P1H21C/ SOURCE RANGE CHANNEL FAILURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED t
R3JPP202 Source Range A Indication Grfx Rcdr +6 log cps 3.75 sec R3JPP203 Source Range B Indication Grfx Rcdr 46 log cps R3SURNil Source Range A SUR Grfx Rcdr +1 DPM R3SURN12 Source Range B SUR Grfx Rcdr +1 DPH A g . i i i l i O l
PERFORMANCE TEST NUMBER: PlH21C DATE: 01/13/91 {; TITLE: SOURCE RANGE CHANNEL FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.30E E7S' #uA
/fTesfCoordinator 10kte 2.0 All deficiencies have been resolved and this test has been completed.
m U NOT R10001 RI'.D TPf Test Coordinator /Date 3.0 Testresultsreviewed.SimulatorperformanceIIsatisfactory.
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l TEST NUMBER / TITLE: PTH22A/RCS Thot 1RANSHITTER FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 01/23/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% aower, the selected A loop Thot transmitter fails high (620f). T1e resuh ant increase in indicated Tave causes automatic control rod insertion in an effort to reduce Tave to setpoint (579f). The power reduction causes the ICS to reduce turbine load and feedwater. The reactor and feedwater control stations are placed in hand to terminate the transient, the non-fculted Thot indication is selected and Tave is restored to setpoint. MALFUNCTION TITLE QPRMS RC-018 Thot Transmitter failure Selectable to RC-4A-TEl, TE2, -TE3 or TE4 l Adjustable from 520f - 620F l INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in ! service. FINAL CONDITIONS: Non-faulted Thot signal selected, ICS restored to full automatic with plant parameters at normal values for the existing power level / turbine load. Load increase in progress. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. <O
PERFORMANCE TEST NUMBER: PTM22A DATE: 01/23/91 O TITLE: RCS Thot TRANSHITTER FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisf actory. No performance problems noted. 4 L O O
TEST NUMBER / TITLE: P1M22A/RCS Thot TRANSMITTER FAILURE TEST DATA
~
LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPELD
*Hl:ll3 o
-ressurizer Level Grfx Rcdr 0 - 320 in 1.875 sec
*HlJPR203 Pressurizer Temperaturo Grfx Rcdr 0 700f
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 125Y, PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50f Ll:6BDT delta T Loop B Grfx Redr 0 50f
*LlJPR212 Th Loop-A Grfx Redr 520 - 620f
*LlJPR213 Th loop B Grfx Rcdr 520 620F <
*HCTAVG RCS Average Temperature Grfx Rcdr 520 620f L1RCSDTC RCS delta-Tcold Grfx Rcdr 410f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr $20 - 620f
*LlJPR215 Tc Cold Leg A2- Grfx Rcdr 520 - 620f
*LlJ'R216 Tc Cold Leg 81 Grfx Rcdr 520 620f
*LlJPR217 Tc Cold leg B2 Grfx Rcdr $20 - 620f
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 Su-B SU Level Grfx Redr 0 250 in
*LlJPS284 SG A Op Level Grfx Redr 0 1007.
*LlJPS336 SG B Op Level Grfx Rcdr 0 O
1007. f3JPA323 SG-A Press ErlC) Grfx Rcdr 0 - 1200 psi
**f3JPA320 SG-B Press EflC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop A 11 r Press Grfx Rcdr 600 - 1200 psi LlJP1232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
*LIJPS301 SG A MfW flow Grfx Rcdr 0 6K KLB/HR
*LlJPS302 SG B MfW flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU fW flow Grfx Rcdr 0 1500 MLB/HR
*L1JPS330 SG-B SU FW flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 1007, L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 1007.
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 1007, l O u
_ _ . __..__..._..._m__.~___. , g PERf0RMANCf. TEST NUMBER: PIM22A DATE: 01/23/91 TITLE: RCS 1 hot IRAllSHITilR fAllVRL TEST STATUS: 1.0 1his test is complete and has the following deficiencies noted (if none so indicate): 1R # Priority Date issued Date Closed NONI1 ___ ,- , da.*' verm*-.1n-----a
~ e- --+ - "-*=--"*----*a--
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.-. . +-_'*ea-f, f (g [.yiWO ~
a/: s/gi IPC Iest Coordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. O x NOT HiiGUlitICD
~
I PC 1est Coordinai~or ~~~~7D5te i 3.0 lest results reviewed. Simulator performance is sat isf actory, f 3 9,!/ lj.7. -- *=-4g,i - p,, y 4, Simulator Instructor /Date Operat ions 'ReiiFe~sentWiV6/DaTU l ( li
\~/ ,/Symuf7t(Ad$ ator lngineer i% */Date L/
U , APPROVED: , p fn __.. DATE: f(. NSI/lillTk
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l I p TEST NUMBER / TITLE: PTH228/RCS Tcold TRANSMITTER FAILURr l V ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 01/23/91 TEST DURAT10N: I hour TEST DESCRIPTION: With the simulator operating at 100% power, the selected B loop Tcold transmitter fails mid-scale (570F). The resultant increase in indicated Tave causes automatic control rod insertion in an effort to reduce Tave to setpoint (579F). The power reduction causes the ICS to reduce turbine load and feedwater and re-ratio feedwater in response to the indicated delta-Tcold. The reactor and feedwater control stations are placed in hand to terminate the transient, the non-faulted Tcold indication is selected and Tave is restored to setpoint. l MALFUNCTION TITLE OPTIONS RC-019 Tcold Transmitter failure Selectable to RC-5A-TT1, -TT3, ' RC-58 TTl or -TT3. Adjustable from 520F - 620F INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Non faulted Tcold signal selected, ICS restored to full G automatic with plant parameters at normal values for the V existing power level / turbine load Load increase in progress. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. I O
PERFORMANCE TEST NUMBER: PTH22B DATE: 01/23/91 TITLE: RCS Tcold TRAN5 HITTER FAllURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. O O i
TEST NUMBER / TITLE: PTM228/RCS Tcold TRANSMITTER FAILURE TES1 DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*H1113 Pressurizer Level Grfx Rcdr 0 - 320 in 1.875 SEC *H1JPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:CADT delta T Loop A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop B Grfx Rcdr 0 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F *LlJPR213 Th Loop-B Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LlRCSDTC RCS delta-Tcold Grfx Rcdr 10 - +10f *LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F *L4JPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F *LlJPR216 Tc Cold Leg B1 Grfx Rcdr 520 - 620F *LlJPR217 Tc Cold leg B2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in *LIJPS284 SG-A Op Level Grfx Rcdr 0 - 100% *L1JPS336 SG-B Op Level Grfx Rcdr 0 - 100%
~
*F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LidPS330 SG-B SU FW F1ow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
LM6AV Grp 6 Avg Position Grfx Rcdr 0 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% O - l
DATE: 01/23/91 [ PERFORMANCE TEST NUMBER: P1M22B TITLE: RCS Tcold 1RANSMITTER FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date closed
.Hote
,,g g$ ~> c1/2 Suih FPC Test Coordinator /date 2.0 All deficiencies have been resolved and this test has been completed.
I NOT Itt:QUllti:D ITC Test CoorTinator /Date l 1 1 3.0 Test results reviewed. Simulator performance is satisfactory. g I fi27/7/ l /} l !( [ g/ .f / Simulator Instructor
/ ate (T(p~erations 4fepresentat1.ve -
Date /f f l 733tMnhhr*=#$' APPROVED: up ggcowdP~:\W- ed DATE: m 15 4/-- 31 l l \ i
q TEST NUMBER / TITLE: PTM22C/HAIN FEEDWATER FLOW TRANSMIT' :R FAILURE b ANS!/ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 01/24/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 100% power when main feedwater flow transmitter 3P-80 FTl fails low. The low flow signal causes the ICS to increase B main feedwater pump speed to increase flow to the B OTSG. As actual feedwater flow inct cses B OTSG level increases. When the B OTSG reaches its high level limit (95% on the operating range) the ICS terminates the feedwater flow increase. Feedwater is taken to manual control, a non faulted transmitter is selected for innut to the ICS, B main feedwater flow and 0TSG level are restored to normal values and _the ICS is_ restored to full __ automatic control. MALFUNCTION lilLE OPTIONS FW-009 Main Feedwater Flow Transmitter Selectable to SP-8 AFT 1, -FT2, Failure SP-88-FTl or -FT2 Adjustable from 0 - 6E+6 lbm/hr INITIAL CONDITIONS: Full power steady state conditions. all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions, with the B main feedwater flow signal to the ICS selected to a non faulted transmitter. (] . PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
PERFORMANCE TEST NUMBER: PIM22C DATE: 01/24/91 TITLE: MAIN FEEDWATER FLOW TRANSMITTER FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted, i O O
g TEST NUMBER / TITLE: PTM22C/ MAIN FEE 0 WATER FLOW TRANSMITTER FAILURE O TEST DATA LABEL DESCRIPTION RECORDING RANGr FREQUENCY /
- Critical Parameter ,
DEVICE , CHART SPEED
*Hl:ll3 Pressurizor Level Grfx Rcdr 0 120 in 1.875 sec
*H)JPR203 Pressurizer Temperature GrTx kcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe LI:6ADT delta T Loop-A Grfx Rcdr 0 - 50F L1:6BDT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop A Grfx Rcdr 520 - 620F
*LlJPR213 Th Loop B Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F LITMARG4 Subcooling Margin Grfx Rcdr +10F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPR216 Tc Cold Leg B1 Grfx Rcdr 520 - 620F
*L1JPR217 Tc Cold Leg B2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SV Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LIJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
( *F3JPA323 SG-A Press (EFIC) Grfx Redr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi L1JPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 3G-A HFW Flow Grfx Rcdr 0 6K KLB/HR
*LiJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
- L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100% l HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm l l
=
lO l \- _ _ . . - - - . - . - - . - -- . ., -
] PERFORMANCE TEST-NUMBERt PTH22C DATE: 01/24/91 TITLE: MAIN FEEDWATER FLOW 1RANSM111[R FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.ll0E
- 4 7
~nf Y )bTte; y'h/
WC lest Coordinator / 2.0 All deficiencies have been resolved and this test has been completed. (G d NOT HEQUIHHD TPC Test Coordinator /Date 3.0 Testresultsreviewed.Simulatorperformanceiksatisfactory.
?/2yf]/ % ]() L 1 /L f,/
la,,ato, Inst _to, e,e,a mn s,(.f,,'.,entate
,eate Q huator a h nneer 4, ,,
/Da e AerR m o: om. <usu M(/iKOTHL N5T
'\
TEST NUMBER / TITLE: PIM22D/0TSG OPERAT!!iG RANGE LEVEL TRAf1SMITTER FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORNLD: 01/31/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating at 1007, power when the selected 0 0150 operating range level transmitter fails high. The resultant level error causes the ICS to begin reducing feedwater flow to the B 01SG, causing an increase in RCS temperature, pressu,re and pressurizer level. EA. LFUitCT10B IllLE REILQMS
.fW 018 015G Operating Range Level Selectable to SP 1A LT2, LT3, Transmitter failure SP 18 LT2, LT3 Adjustable from 0 - 1007.
lHITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power steady state conditions, B 015G operating range level selected to the alternate levol transmitter. PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
-) PERFORMANCE TEST NUMBER: PTH220 DATE: 01/31/91 O TITLE: OTSG OPERATING RANGE LEVEL TRANSMITTER FAILURE _ TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted, b O
1 TEST NUMBER /TITLEt PIM22D/0TSG OPERATING RANGE LEVEL TRANSMITTER FAILURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 320 in 1.875 sec *HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop A Grfx Rcdr 520 - 620F
*LlJPR213 Th Loop-B Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Gefx Rcdr 520 620F Ll:RC6TM Delta T Cold Grfx ".," 410F
*LlJPR214 Tc Cold leg Al Grix Rcdr 520 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPR216 Tc Cold Leg B1 Grfx Rcdr 520 - 620F
*LlJPR217 Tc Cold Leg B2 Grfx Rcdr 520 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SU Level Grix Rcdr 0 - 250 in
*LlJPS284 SG A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
Grfx Rcdr 0 - 1200 psi O **F3JPA323 F3JPA320 SGSG-A Press B Press (EFIC)(EFIC) LlJPT228 HS Loop-A Hdr Press Grfx Rcdr Grfx Rcdr 0 - 1200 psi 600 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG A MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG B HFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR O ,
PEPFORMANCE TEST NUMBER: PTM22D DATE: 01/31/91 TITLE: 0TSG OPERATING RANGE LEVEL TRANSMITTER FAILURE TEST STATUS: 1.0 This test it complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Datc Clesed _NONE b lt FPC Test Coordinator
/f3/ l
/Date 2.0 All deficiencies have been resolved and this test has been completed.
NOT REQUIRED TPC Test Coordinator /DaTe
/
3.0 Test results reviewed. Simulator parformance is satjsfactory, et ur l 4k3f}/ \ i U{ !^ l - Y! $ }{/ Simulator Instructor /Date Operations Representative /Date OHfDJ nR ?h/ mulator Engin e'r /Dat'e v c, APPROVED: ,,3
%, , DATE: S ih al
' 'l NSTS iLOTM j' s o
Lj
. __ m . . _ . _ _ . . _ - _ _ . . _ . . . _ _ _ - . . . _ . _ _ _ _ - -. _ _ ._ _ _ . _ _ .
g -TEST NUNBER/ TITLE: PTM22E/ICS RUNBACK FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORNEDt- 01/30/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: The simulator is operating a 100% power when rod 1, group 7 drops. Due to a failure in the ICS, the plant does not automatically run back to 60% power. A load reduction to 60% power is performed at 5%/ minute with the ICS in order to stabilize the plant. tLA.LFUNCTION TITLE OPT 10NS IC 003 Block ICS Runback Capability INITIAL CONDITIONS: Full power steady state conditions, all plant equipment-in service. i FINAL-CONDITIONS: Plant at 60% power, all systems in automatic. Rod 1 group 7 fully inserted into the core. PLANT PROCEDURES USED DURING-THE TEST: AP-545, REVISION 3 PLANT RUNBACK BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. mma LO
. - - . _ . - _ _ _ . . . ~ - - - - . . _ _ _ _ __
PERFORMANCE TEST NUMBER: PTM22E DATE: 01/30/91 TITLE: ICS RUNBACK FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. k O O
--TEST NUMBER / TITLE: PTM22E/ICS RUNBACK FAILURE O
w./ TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
-* Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320-in 3.75 sec-
*HIJPR203 Pressurizer Temperature Grfx Redr 0 -'700F
*L4JPR222 RCS Pressure-(NR)' Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP -Normalized Reactor Power Grfx Rcdr 0 - 125%
-PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe l L1:6ADT delta T Loop-A- Grfx Rcdr 0 - 50F L1:6BDT delta T Loop-B Grfx Rcdr 0 - 50F ,
*LlJPR212 Th Loop-A_ Grfx Rcdr 520 - 620F
*HCTAVG -RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214>Tc Cold leg Al Grfx Rcdr 520 - 620F ,
*LIJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F l
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in l
*L1JPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op level Grfx Rcdr 0 - 100%
'*F3JPA323-SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
'*F3JPA320 SG-B Press (EFIC) .Grfx Rcdr 0 - 1200 psi LIJPT228 MS Loop-A Hdr Press Grfx Rcdr- 600 - 1200 psi LlJPT232.MS. Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi
*LlJPS301 SG-A MFW Flow- Grfx Rcdr O, *LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR
-*LlJPS329 SG-A SU FW Fle' Grfx Rcdr 1500 MLB/HR
*L1JPS330~SG-B SU FW F'
- Grfx Rcdr 0 - 1500 MLB/HR 1,5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV- Grp 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Grp 7 Avg Position Grfx Rcdr 0 - 100%- HQBBCORE RCS Boron- - Grfx Rcdr-- 0 - 3000 ppm l'3JPP208 Power Range Indication Grfx Rcdr 0 - 125%-
~R3JPP200 Power Range-Indication Grfx Rcdr 0 - 125%
R3JPP210 Power Range Indication
- Grfx R:dr 0 J125%
R3JPP211 Power Range Indication Grfx Rcue 0 -1125%. I v i:
(~ 3 PERFCRMANCE TEST NUMBER: PlM22E DATE: 01/30/91 TITLE: ICS RUNBACK FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _N0NE I is fl F C Test ~toordinator /Date 2.0 All deficiencies have been resolved and this test has been completed. m b) IlOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance ,is satisfactory.
/ ,
s
} l$ Y \; ' ;
Y ill ,l Simulator fnstructor /Date Operations Representative /Date i { L walk 9L Websk/ inul ator Engineer '/Date Aree _ us~ NS}S/NLDTM l om. c ' u' O _ O i
TEST NUMBER / TITLE: PTM22F/ICS HEADER PRESSURE BIAS FAILURE O ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 01/29/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 1007. power, a reactor trip occurs. Due to'a failure of circuit 1 in module 4-3-3, the ICS does not transfer to a 125 psi pressure bias following the trip, causing the turbine bypass valves to control post trip header pressure at 885 psig vs 1010 psig. As a result, Tave decreases to approximately 537'F. after the trip instead of tha normal post trip Tave of 555'F. The additional post trip cooldown causes a larger than normal post trip decrease in both pressurizer level and RCS pressure. Pressurizer level is stabilized through the use of one of the HPI nozzles in addition to normal makeup flow, t!ALFUNCTION ILILE OPTIONS 10-004 ICS Header Pressure Bias Failure Selectable to either the 50 or 125 psi bias signal. INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, Tave stabilized at approximately 535, pressurizer level and RCS pressure stabilized at normal post l trip values. l PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. I O l l - . - .
l ,ss
- PERFORMANCE TEST NUMBER: PTM22F DATE: 01/29/91 ' TITLE: ICS HEADER PRESSURE BIAS FAILURE TEST RESULTS/ COMMENTS:
Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted, O (O f3 V
m g TEST NUMBER / TITLE: PTM22F/ICS HEADER PRESSURE BIAS FAILURE TEST DATA LABEL DESCRIPTION RECORDlHG RANGE FREQUENCY /
- Critical - Parameter DEVICE CHART SPEED
*Hl:113 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec *HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F *L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi *L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi *RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F *HCTAVG RCS Average Temperature Crfx Rcdr 520 - 620F *L1JPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F *LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F *LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in *LlJPS293 SG-B SU Level C-fx Rcdr 0 - 250 in *LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100% *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100% *F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi *F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Lcop-A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi Q
V
*L1JPS301 SG-A MFW Flow *L1JPS302 SG B MFW flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR *LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR *LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV Group 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Group 6 Avg Position Grfx Rcdr 0 - 100% L5G7AV Group 7 Avg Position Grfx Rcdr 0 - 100% N2:M18 HPI Flow Al Grfx Rcdr 0 - 500 gpm O v
$ PERFORMANCE TCST NUMBER: PTH22F DATE: 01/29/91 TITLE: ICS HLADER PRESSURE BIAS FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.1%E f obh/
FPC Test Coordinator t!?hki
/Date 2.0 All deficiencies have been resolved and this test has been completed.
Ct V NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance'is satisfactory. LS / fi . 1 '7[/f / w 'N l !/ [ dh )$/ 'h/ S'fmulator Instructbr /Date Operations __ Representative /%te' I Y,/dws/d t N 3NatorEngiliEer /Date APPROVED: / DATE: o,/ph'
/sTS/difiM
~ a L)
e TEST NUMBER / TITLE: PTM22G/ MAIN FEEDWATER TEMPERATURE TRANSMITTER FAILURE l 1 V %NSI/ANS-3,5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, the selected Loop-A main feedwater temperature transmitter fails low. Due to the failure, the ICS automatically reduces feedwater flow in the affected loop. The simulator is stabilized, and the alternate Loop-A transmitter is selected for indication and control. MALFUNCTION TITLE OPTIONS FW-010 Main Feedwater Temperature Selectable to SP-5A-TEl, -TE2, Transmitter Failure SP-5B-TEl and/or -TE2 - Adjustable from 0 - 500F INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Full power conditions with feedwater flow restored to normal full power values and the alternate main feedwater temperature transmitter selected for service. PLANT PROCEDURES USED DURING THE TEST: 3 (G BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement.
g PERFORMANCE TEST NUMBER: PTM22G DATE: 07/03/91 TITLE: MAIN FEE 0 WATER TEMPERATURE TRANSMITTER FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps completed satisfactorily. Simulator response excellent. Step M22G030(b): A loop flow indication increase due to density compensation was visible on MCB flow indicators. Effact not apparent on graphic recorder due to relatively low sampling frequency, i l l l l l l n _ U l I
( TEST NUMBER / TITLE: PTM22G/ MAIN FEEDWATER TEMPERATURE TRANSMITTER FAILURE TEST DATA
-LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*H1:113 Pressurizer Level Grfx Rcdr 0 - 320 in 3.75 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F I
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcor 0 - 50F
*L1JPR212 Th Loop-A Grfx Rcdr 520 -- 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LIJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LIJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi O LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi V **LlJPS301 SG-A L1JPS302 SG-8 HFW MFW Flow Flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*L1JPS330 SG-B SU FW Flcw Grfx Rcdr 0 - 1500 MLB/HR LI:SP5AT Loop A MFW Temperature Grfx Rcdr 0 - 500F Ll:SP5BT Loop B MFW Temperature Grfx Rcd 500F L5G6AV Group 6 Avg Position Grfx Rcdr 0 - 100%
L5G7AV Group 7 Avg Position Grfx Rcdr 0 - 100%
'M, w
U
i i
""^"c' ^: 7' 3'
O "'"' S' """": "'"22 TITLE: MAIN FEEDWATER TEMPERATURE TRANSMITTER FAILURE TEST STATUS: 1.0 This test is complete and has the folicwing deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed
,JONE _
u k N,rm,, lhl.s-FPCTjstCoordin{ tor /Dat;e 2.0 All deficiencies have been resolved and this test has been completed. O ee exoo1 mo FPC Test Coordinator /Date ,
/ i 3.0 Test results reviewed, Simulator performance is satis. factory.
/ ,.
, , _ kr ., c C,[ v .y Ltjy 1(-l,){fj Simulato'r lhst uctor / e OperationsRepresentative/phte
\
c' /) wtIO) n d LA 9 z / imulator'EngT uer '/Date APPROVED: k DATE: 9 O 1
TEST NUMBER / TITLE: PTM22H/ FAILURE OF ICS TO TRANSFER TO TRACKING MODE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORMEU: 08/12/91 TEST DURATION: I hour TEST DESCRIPTION: With the simulator operating at 100% power, a load reduction to 40% power (-345) hWe is commenced with the control rod drive system in manual. Due to a failure within the ICS, the ICS does not revert to the TRACKING mode of operation when the Diamond Control Station is placed in manual. MALFUNCTION TITLE OPTIONS _IC-006 Block ICS Tracking function INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor power at -40%. Cenerated megawatts, feedwater flows and 0TSG leve'is at higher than expected values for the existing power level; RCS Tave at a value below its normal setpoint. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
i PERFORMANCE TEST NUMBER: PTH22H DATE: 08/12/91 TITLE: FAILURE OF ICS TO TRANSFER TO TRACKING MODE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted. O O'
TEST NUMBER / TITLE: PTM22H/ FAILURE OF ICS TO TRANSFER TO TRACKING MODE TEST DATA LABEL CESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in .375 sec
*HlJPR203 Pressurizer Temperature Grfx Redr 0 - 700F
*L4JPR222 RCS Pressure _(NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr- 0 -2500 psi
*RRSWTP Normalized Reactor Power Grfx Redr 0 -125%
' PGJPE102 GrossLGeneration Grfx Rcdr 0 - 1000 MWe L3:ULMM Megawatt Demand Grfx Rcdr 0 - 100%
L5G7AV Grp 7-Avg Position Grfx Rcdr 0 - 100% 1*LlJPR212 Th loop-A . _ Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature- Grfx Recr 520 - 620F
*LlJPR214-Tc Cold Leg-Al Grfx Rcdr 520 620F
*L1JPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*L1JPS284 SG-A Op: Level Grfx Rcdr 0 - 100%
*L1JPS336 SG-B Op Level Grfx Rcdr 0 - 100%-
*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi-
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi u llJPT228 MS Loop.A Hdr Press Grfx Rcdr 600 - 1200 psi l LlJPT232 MS Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi ;
*LlJPS301 SG-A.MFW Flow- Grfx Rcdr 0 - 6K KLB/HR l
'*LlJPS302 SG-B MFW Flow Grfx Rcdr 0 - 6K KLB/HR O
$ PERFORMANCE TEST NUMBER: PTM22il DATE: 08/12/91 TITLE: FAllVRE Of ICS TO TRANSFER 10 1 RACKING MODE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed EtfE S $ . ,^
/
,Q[0f&TW
/ X' '
,- Sh1/ll TffTEtUoordinator /Date 2.0 All deficiencies have been resolved and this test has been completed, o
'O NoT ImOUlltED UC~ Test Coordinator /Dati 3.0 Test results reviewed. Simulator performance is satisfactory. 3 i . . , I ( ! f
, ,bs(? ,t
'y f/hifif y _$ ; i(Y(' / jf,(L 'y' ,
31mulatorTnstructor /Date Operations Representative /05te
' - Y l l/
u f En'gtneer - /Da e APPROVED: DATE: _7 /_
/
_.] g V
t-p TEST NUMBER / TITLE: PTM221/ REACTOR DEMAND SIGNAL FAILURE D , TNSI/ANS-3.S(1985) CROSS
REFERENCE:
Section 3.1.2(22) DATE PERFORHED: 07/03/91 TEST DU_ RATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, the ICS Reactor demand signal fails to 0 volts, corresponding to a demand of 62.5% reactor power. The resultant ncutron error causes the rods to insert to reduce power to 62.5%, places the ICS in TRACK due to cross limits and causes feedwater flows in both loops to increase, liALFUNCTION TITLE OPTIONS IC-007 Reactor Deamnd Signal Failure to 0 volts INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor power at approximately 62.5%, generated megawatts at approximately 540 MWe, feedwater flow at approximately 7E06 lbm/hr. PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: There is no plant data for this u ansient. Expected simulator response based on expert judgement. O
~
i PERFORMANCE TEST NUhBER: PTM22I DATE: 07/03/91 TITLE: REACTOR DEMAND SIGNAL FAILURE _ TEST RESULTS/ COMMENTS: l Test results evaluation based on expert judgement. All test steps functioned correctly. Simulator response excellent. O O-
TEST NUMBER / TITLE: PTM221/ REACTOR DEMAND SIGNAL FAILURE TEST DATA , LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
.*Hl:113 Pressurizer Level Grfx Rcdr 0 - 320 in 5.625 sec
-*HlJPR203 Pressurizer _ Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr- 0 - 125%
PGJPE202 Gross-Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*L1JPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*L1JPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG-B-Op Level- Grfx=Rcdr 100%
-*F3JPA323 SG-A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi
;L1JPT228 MS Loop-A Hdr Press- Grfx Rcdr 600 - 1200 psi L1JPT232 MS: Loop-B Hdr Press Grfx Rcdr 600 - 1200 psi Grfx Rcdr O **L1JPS301-SG-A L1JPS302 SG-B MFWMFW Flow- Flow
*LlJPS329 SG-A SU FW Flow Gefx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 --6K KLB/HR 0 - 1500 MLB/HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5G5AV -_ Group 5 Avg-Position Grfx Rcdr 0 - 100%
L5G6AV Group 6 Avg Position- Grfx Rcdr 0 - 100%
- L5G7AV Group 7 Avg _ Position Grfx Rcdr 0 _100%
HQBBCOP.E RCS Boron Concentration Grfx Rcdr 0 - 3000 ppm-
- {} PERFORMANCE TEST NUMBER
- PTH221 DATE: 07/03/91 TITLE: REACTOR DEMAND SIGNAL FAILURE TEST STATUS:
1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # rriority Date issued Date Closed _ LEE
' ~
WC es[kodrdiS'or J /6dk 2.0 All deficiencies have been resolved and this test has been completed. 10
'sj NOT REQUIRED
'FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory. ,
at , ~ . , - 'i k 'l Al u ~
! l' i Ih:Yh9 f,i Siihulat'or 'fhstquctor /Date Operations RepresentativpfDate L W dw'/Dite inulat6r'Eng'in6er APPROVED
- DATE: 7 /
o
t TEST NUMBER / TITLE: PTM22J/ SELECTED TAVE FAILURE-
--ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(22)- DATE PERFORNED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 75% power, the selected Tave signal fails to 0 volts, corresponding to an indicated controlling Tave of 570F, The resultant Tave error causes a rod withdrawal signal to be generated; the large neutron error also results in a cross limit condition (Rx Pwr > Demand) which causes the ICS to revert to the tracking mode of operation- . The Crosslimit causes FW Demand / flow in A & B loop to decrease. With FW decreasing and Rx Pwr Increasing RCS pressure rapidily increases. A reactor trip occurrs on high RCS Pressure. post trip plant responso following the trip is normal. MALFUNCTI0ti TITLE 9f11111 NN-004 Selected-Tave Failure Adjustable from 520 - 620F INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Simulator stable at approximately 103% reactor power and load. Feedwater limited by- high OTSG levels. Indicated Tave at 570F, actual Tave stable at a higher than normal value. PLANT PROCEDURES USED DURING THE TEST: BASELINE DATA: There is no plant data for this transient. Expected simulator response-based on expert judgement. 10
i PERFORMANCE TEST NUMBER: PTM22J DATE:.07/03/91 C. TITLE: SELECTED TAVE FAILURE TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned correctly. Simulator response excellent. A momentary increase in MWe generation was noted during the transient. This was caused by a rapid decrease in Main FW flow and a resultant decrease in aspirating steam flow in the OTSG. This effect is short-lived and not normally observable on MCB instrumentation. Grafic recorder frame 1 menu 2 shows MWe increasing for -10 seconds. O O
q TEST NUMBER / TITLE: PTM22J/ SELECTED TAVE FAILURE TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:lI3 Pressurizer Level Grfx Rcdr 0 - 320 in .5 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 Mwe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F L1:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th loop-A Grfx Rcdr 520 - 620F
*HCT?VG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPR215 Tc Cold leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 250 in
*L1JPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*L1JPS284 SG-A Op Level Grfx Rcdr 0 - 100%
! *LlJPS336 SG-B Op Level Grfx Rcdr 0 - 100%
*F3JPA323 SG-A Press (FFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG-B Press (EFIC) Grfx Rcdr 0 - 1200 psi
, LlJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi l L1JPT232 MS Loop B Hdr Press Grfx Rcdr 600 - 1200 psi l O
- Grfx Rcdr 0 - 6K KLB/HR i
V *LlJPS301 SG-A LlJPS302 SG-B MFWHFWFlowFlow Grfx Rcdr 0 - 6K KLB/HR
*L1JPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LIJPS330 SG-8 SU FW Flow Grfx Rcdr 0 - 1500 HLB /HR L5G5AV Group 5 Avg Position Grfx Rcdr 0 - 100%
- L5G6AV Group 6 Avg Position Grfx Rcdr 0 - 100%
l L5G7AV Group 7 Avg Position Grfx Rcdr 0 - 100% Grfx Rcdr HQBBCORE RCS Boron 0 - 3000 ppm O y-l
O etaron"^"ce 'Est " UMBER: PTM22J DATE: 07/03/9] TITLE: SELECTED TAVE FAILURE TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed _NONE [-. FTC(Test Coordi .tor '/Date 2.0 All deficiencies have been resolved and this test has been completed. l .O i U NOT REQUIRED FPC Test Coordinator /Date 3,0 Test results reviewed, Simulator performancef satisfactory,
- I p uk msn
'l likij 'k!' M / E((' /$1?9 /
SimlatorIn'truttor s /Dat'e Operations Representative /Date
>,bNy1/rN9f7 k; h- 9 2 /
S mulator E'ngineer / ate l APPROVED: / DATE: 7-(M/ ffSTS/NLOTK' l n l L) l
1 p TEST NUMBER / TITLE: PTM23A/ MAIN STEAM ISOLATION ACTUATION FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(23) DATE PERFORMED: 07/15/91 TEST DURATION: 1/2 hour l TEST DESCRIPTION: With the simulator operating at 100% power, a main steam line break on the B OTSG occurs outside the containment downstream of main steam isolation valve MSV-414. Due to multiple relay malfunctions in the EFIC system neither EFIC train initiates a Main Steam Isolation signal when the B OTSG depressurizes to less than 600 psig. B OTSG isolation is accomplished by manually closing Main Steam Isolation Valves MSV-413 and -414. MALFUNCTION llILE OPTIONS EF-007 MS Isolation Actuation Failure Selectable to EFlc A OTSG A, EFIC B OTSG A, EFIC A OTSG B, EFIC B OTSG B INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, OTSG B isolated. PLANT PROCEDURES USED DURING THE TEST: A U BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement.
PERFORMANCE TEST NUMBER: PTM23A DATE: 07/15/91 TITLE: MAIN STEAM ISOLATION ACTUATION FAILURE , TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned r.ormally, Simulator response excellent. O O
j TEST NUMBER / TITLE: PTM23A/ MAIN STEAM ISOLATION ACTUATION FAILURE TEST DATA LABEL oESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIhULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT O
O
d C ' PERFORMANCE TEST NUMBEA: P1Hl3A DATE: 07/15/91 TITLE: MAIN STEAM ISOLATION ""ATION FAILURE TEST STATU$s 1.0 This test is com,
- 4i.. has the following deficielcies noted (If none so indicat ):
TR # Priority Date issued Date Closed i _NONE h
.._._ 1 P ' \~n .-, 1 !T
/Dpte -
TPC]\>estC60rdigator 2.0 All deficiencies have been resolvti and this test has been completeu NOT R1*.QUIrI:D FR lest Coordinator /Date
/
3.0 Test results reviewed. Simulator performance is satisfactory. ( ,
- i. l .
WW l !! l 5 !!U[ w/ h/ / IT. uatorInsti}ctor /Datd Dperations Representative /Date f,/l
/ '
hi) 2 Lb~ l - Y iriaer Date-()imulator N'N DATE: f/
'7/h~~
APPROVED [:bS/NOfif l
- O l- -
l l
e-- - , TEST NUMBER / TITLE: PTM23B/ MAIN FEt0 WATER ISOLATION ACTUATION FAILURE ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(23) UATE PERFORMED: 07/15/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, a main steam line break on the A OTSG occurs outside the containment upstream of main steam isolation valve MSV 412. Due to a malfunction of the EFIC A OTSG A Trip module, EFIC Train A does not initiate a Main feedwater Isolation signal for the A OTSG Since EFIC T:ain B is operable and initiates an MF1 signal for the A OTSG, isolation of the A OTSG occurs. MALFUNCTION TITLE OPTIONS EF-006 MFW ! solation Actuation failure Selectable to EFIC A OTSG A, EFIC B OTSG A, Ef10 A OTSG B, EFIC B OTSG B INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor' turbine tripped, OTSG A depressurized and isolated. PLANT PROCEDURES USED DURING THE TEST: i O BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. 1 i l i i (
._ a O
t
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~ . _ _ - _ _ . _ _ . . . . _ _ . _ - _ - - _ . _ . _ . - _ _ _ - _ . _ _ _ _ _ _ . _ . _ . _ _ _ _ _
(s 1 PERFORMANCE TEST NUMBER: P1H23B DATE: 07/15/91 TITLE: MAIN FEEDWATER ISOLATION ACTUATION FAllVRE TEST RESULTS/ COMMENTS: Test reruits evaluation based on expert judgement. All test steps functioned normally. Simulator response excellent. i l l l i i
TEST NUMBER / TITLE: P1H23B/ MAIN FEEDWATER ISOLAT10fl ACTUATION FAILURE O TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Paramoter DEVICE CilART SPEED ANALOG DA1A COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT O
O
C etarona^wct itSt ausatai 918238 o^tti o7/is/92 TITLE: MAIN FEEDWATER ISOLATION ACTUATION FAILURE , TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none soindicate): 1R 6 Priority Date issued Date Closed NONE h
~~
TFQ $If drd$ tor D le 2.0 All deficiencies have been resolved and this test has been completed. { NOT Hl QUIRim TPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performance,is satisfactory. o f ,
.n hl* Q % Il'blll gjhfA '$di h IWu. ator"I
~
fm uct r' 7
/Datej Opprations kepresentative/fjete/ )
- /
.(Wn}]'L}s.
^ imulatlor Enginee,.
_C c7 r L : _/ bate b/ 1 n / APPROVED: 2C [m DATE: _Ar m V
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, TEST NUMBER / TITLE: P1H23C/ES C,HANNEL FAILURE TO ACTUATE HPI ANSI /ANS-3.5(1985) CROSS REFEREliCE: Section 3.1.2(23) DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 1007. power, a LOCA occurs. Due to failures of bistables in two of the three ES HPI actuation channels, HPl does not automatically actuate when RCS pressure reaches 1500 psig. HPI is manually actuated at the ES panel. MALFUNCTI0ff IJ1LE OPTIONS ES 007 ES Channel Failure HPI Selectable to Channel A (B/S Bil coils K1 and K2) B (B/S BT2 coils K1 and K2 and/or C (B/S BT3 coils K1 and K2 INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, HPl actuated. PLANT PROCEDURES USED DURING THE TEST: 1 O BASELVNE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
_ . . _ _ _ _ . _ _ . _ _ . . _ . - _ - _ . _ . - _ _ _ _ _ = _ . _ _ - . . _ _ - _ _ _ _ _ . _ - _ _ _ _ _ . _ _ l PERFORMANCE TEST NUMBER: PTH230 DA1E 07/03/91 O TITLE: ES CliANNEL FAILURE TO ACTUATE 11PI TEST RESULTS/ COMMENTS: Test results evaluation bast.d on expert judgement. All test steps functioned normally. Simulator response excellent. O I
~
l l O
.-__. - - - . .-_. - ~.-... ..--..--._ _ - .- -....._ - .
TEST NUMBER / TITLE: PTM23C/ES CHANNEL FAILURE TO ACTUATE HPI TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEYlCE CHART SPEED i ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SIMULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPER PRINTOUT O
- O
_ _ . . _ . . _ _ _ _ _.___ _ _ .,. ~ _ _ _ _ _ _. _., . . _ _
i I O PERFORMANCE TEST NUMBER: P1M230 DATE: )7/03/91 TITLE: ES CHANNEL FAILURE TO AC10 ATE HPl _. TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed M fi \
.- t d i~ g 1 LN lest Coordinator
/lJate 2.0 All deficiencies have been resolved and this test has been completed.
NOT Iti: QUILTED TPC Test Coordinator /Date
/
3.0 Test results reviewed. Simulator performance is satisfactory.
/
bdk tor 4 p f(
/Date N
/2 :/// . .s l.lh Y Jf /
OperationsRepresentativpf0 ate Simu(atorIni?ru_s c'_ ,(y , / . ED & C W ' W 'liz;b/ S m~u1ator Engineer Date d -, , DATE: _;f i, APPROVED: f:k.J72N 6e
.R$T$/NLO1f(
. _ _ . - . _ _ . = _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ . _ _ _ . . _ . _ _
TEST NUMBER / TITLE: PIM23D/ES CHANNEL FAILURE 10 ACTUATE RB ISOLATION AND COOLING ANSI /ANS .s.5(1985) CROSS
REFERENCE:
Section 3.1.2(23) DATE PERFORMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100Y. power, a LOCA occurs. Due to failures of multiple pressure switches, Channel 2 of RBl&C does not actuate when RB pressure reaches 4 psig. Due to the trips of the other two channels, RBl&C is automatically initiated. MALFUNH LQN III1E OPTIONS ES 009 ES Channel failure RBl&C Selectable to Channel A (BS 24-PS and BS 27 PS), B (BS 25-PS and BS-28-PS) and/or C (BS-26-PS and and BS 29 PS) IN1f!AL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, HPI, LPI and RBl&C actuated. PLANT PROCEDURES USED DURING THE TEST: O BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. O iw, e--vmenw ~ - - c __o ,.-w.-gw w g-m-.a- .---n- y--mperg .y ,e,.,,w.w-., , - ,p ,-w.w-+,..i%.-e,.fy.,,-,m,, yp gg i_m,, ,er,%,-,r.- .-%,.9..,p. --yf,-up. . ..m 7--4g 9.g ,g.e.-
. _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ = _ . _ _ . _ _ _ __ . = - _ _ _ _ _ _ . _ _ _ . _ _ . _ .
PERFORMANCE TEST NUMBER: P1M23D DATE: 07/03/91 TITLE: ES CHANNEL FAILURE 10 ACTUATE RB ISOLATION AND COOLING TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. All test steps functioned normally. Test results satisfactory. O l l l O -
- - - - . - - . _ - ._- - - - - - - ._ _ _ - - - . - ~ -
TEST NUMBER / TITLE: P1M230/ES CHANNEL FAILURE TO ACTUATE RB ISOLATION AND COOLING TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED l 1
I i ANALOG DATA COLLECTION NOT REQUIRED FOR THIS TEST SlHULATOR RESPONSE ADEQUATELY DOCUMENTED BY THE TEST PROCEDURE AND ALARM TYPEh PRINTOUT O 1 1
~ . , . - - .cw , - ,
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C: etaronninCc fCsi uusicas ein23o oArc, 07/03,9i T!TLE: ES CHANNEL FAILURE TO ACTUATE RB ISOLATION AND COOLING TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
.LGL -_
TFCk ($7dl% tor / e 2.0 All deficiencies have been resolved and this test has been completed. O NOT Hl40UlHliD f [ lest Coordinator /Date 3.0 Test results reviewed. Simulator performance is satisfactory.
, ?
~
5Trariiddds%udtor /oatI Operations riebp7etentatur/ hate ' i
/,
{ / mu or hgtneer ae 4D72I '7/bd gjg7dm -d DATE: -- APPROVED: ,
1 TEST NUMBER / TITLE: PTM23E/Ef!C AUTO ACTUA110N FAILURE O ANSI /ANS-3. 5(1985) CROSS
REFERENCE:
Section 3.1.2(23) DATE PERF0GMED: 07/03/91 TEST DURATION: 1/2 hour TEST DESCRIPTION: With the simulator operating at 100% power, both main feedwater pumps trip, resulting a reactor / turbine trip. Due to the failure of Channel A EfW Bus 1 Trip Module, EfW Train A does not automtically actuate. OTSG 1evels are maintained at setpoint by B EfW train componenets. MALFUNCTION J11LE OPTIONS EF-005A EFIC Automatic Actuation failure Selectable to A EfW Bus 1, A EFW Bus 2. B EFW Bus 1 and/or B EFW Bus 2 Trip Modules FW-005 Main Feedwater Pump Trip Selectable to A and/or B MFW Pump INITIAL CONDITIONS: -Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor / turbine tripped, simulator stabilized in the normal post trip window with OTSG levels being maintained by B EFW train components. ,_ PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR TRIP AP-450 REV 15 EMERGENCY FEEDWATER ACTUATION BASELINE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. O
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,. . . . . . . , + . ,.. ....e, ,...r- : ..w-->y...-....-I q PERFORMANCE TEST NUMBER: PIM23E DATE: 07/03/91 TITLE: Efl0 AUTO ACTUATION FAILURE TES1 RESULTS/ COMMENTS:
Test results evaluation based on expert judgement. All test staps functioned correctly. Simulator response excellent. O
i TEST NUMBER / TITLE: PIM23f/Efl0 AUTO ACTUAT1011 FAILURE O TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPFE0 LlJPS287 SG A SU Level Grfx Rcdr 0 250 in 2.8125 see LlJPS293 SG-B SU Level Grfx Rcdr 0 250 in LlJPS284 SG A Op Level Grfx Rcdr 0 1007.
LlJPS336 SG-B Oa leven Grfx Rcdr 0 - 100'/, LlJPS301 SG A MfW flow Grfx Rcdr 0 - 6K KLB/HR LlJPS302 SG-B tifW flow Grfx Rcdr 0 - 6K KLB/HR LlJPS329 SG-A SU FW flow Grfx Rcdr 0 - 1500 KLB/HR I.1JPS330 SG B SU FW flow Grfx Rcdr 0 - 1500 KLB/HR F3:23fil EfW flow S/G B Grfx Rcdr 0 - 1000 gpm f3:25fil EfW flow S/G A Grfx Rcdr 0 1000 gpm F3:26fil EfW flow S/G A Grfx Rcdr 0 1000 gpm F3:24fil EfW flow S/G B Grfx Rcdr 0 1000 gpm O O
PERFORMANCE TEST NUMBER: PIM23E DATE: 07/03/91 TITLE: EFIC AUTO ACTUATION Fall 0RE TEST STATUS: 1 10 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed _11Q111 os k '.n -1 A 'llifil ITC~fpst Coofdinhtor /Date 2.0 All deficiencies have been resolved and this test has been completed. NOT Hi:0UIRI:D TPC Test Coordinator /Date s 3.0 Test results reviewed. Simulator performance is' satisfactory. > s( h iatdr' Ins uttbr 4 a, et
/DT
(( Id([I > [-[' Sc)/ Operat" ions Representative /Date
/
Timq O' r I JllmuN S)itiulator (ng neef [ % 9bb/ /Date
^'
/
APPROVED: . c/_f 9 DATE: '#fM'/
.f H~ST37N M /
V
l l TEST NUMBER / TITLE: PTM24A/ATWS (N0 OPERATOR li4TERVENTION) ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(24) DATE PERFORMED: 01/15/91 TEST DURATION: 1 1/2 hours _ TEST DESCRIPTION: With the plant operating at 1007, power, both main feedwater pumps trip. The main turbine trips, but due to a failure of the Reactcf Protection System, the reactor remains on line. The ICS automatically runs the plant back to 107, power, and emergency feedwater restores steam gentrator levels to low level limits. MALFUNCT10!f TITLE OPTI0fd RP Oll Reactor Trip Module Fails Selectable to A. 8, C, or D RPS to Trip Channels INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Reactor at 107, power. Turbine tripped. S/G levels being maintained at low level limits by emergency feedwater. Tave at 560F, normal RCS pressure, pressurizer level at approximately l 150 in. PLANT PROCEDURES USED DURING THE TEST: l AP-580 REV 15 REACTGR TRIP l BASELINE DATA: There is no plant data available for this transient. Expected simulator response based on expert judgement. l l [ O
PERFORMANCE TEST NUMBER: PTM24A DATE: 01/15/91 TITLE: ATWS (N0 pr rRATOR INTERVENTION) TEST RESULTS/COMHC4TS: Test results evaluation based on expert judgement. Test results satisfactory. No performance problems noted, c 4 O O 4
F G TEST NUMBER / TITLE: PTH24A/ATWS (N0 OPERATOR INTERVENTION) U TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Parameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in 0375 sec
*HIJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700F
*L4JPR222 RCS Pressure (NR) brfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 1257 PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50F Ll:6 BOT delta T Loop B Grfx Rcdr 0 - 50F
*LlJPR212 1h loop-A Grfx Rcdr 520 - 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LIJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620F
*LlJPF,215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG-A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG B SV Level Grfx Redr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100Y,
*LlJPS336 SG B Op Level Grfx Rcdr 0 - 1007,
*F3JPA323 SG A Press (EFIG) Grfx Rcdr 0 1200 psi
*F3JPA320 Sc B Press (EFIC) i Grfx Rcdr 0 - 1200 psi LlJPT2E8 MS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS loop B Hdr Press Grfx Rcdr 600 - 1200 psi
(_ *LlJPS301 SG A MFW Flow Grfx (Icdr 0 - 6K KLB/HR
\
*LlJPS302 SG B MFW Flow Grfx Rcdr 0 - 6K KLB/HR
*L1JPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS330 SG B SU FW Flow Grfx Rcdr 0 - 1500 MLB/HR L5GSAV Grp 5 Avg Position Grfx Rcdr 0 - 100Y, L5G6AV Grp 6 Avg Position Grfx Rcdr 0 - 100Y, L5G7AV Grp 7 Avg Position Grfx Rcdr 0 100%
HQBBCORE RCS Boron Grfx Rcdr 0 3000 ppm F3:23FIl EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:24Fil EFW Flow S/G B- Grfx Rcdr 0 - 1000 gpm i { l O l
P PERFORMANCE TEST NUMBER: P1H24A DATE: 01/15/91 TITLE: ATWS (N0 OPERATOR INTERVENTION) TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date Issued Date Closed noni:
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2.0 All deficiencies have been resolved and this test has been completed. O NOT REQUIRED TPT Test Coordinator /Date 3.0 Test results reviewed. Simulator performance'is satisfactory. J,. fNf yl, , % id.i Aj' Dperations. Representative /Date
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/Simulatbr,1fistructor /Date
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DATE: dl3ly_. ' EST(/;NLcTM' ) O
TEST NUMBER / TITLE: PIM24B/ATWS (OPERATOR INTERVENTION) ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(24) DATE PERFORMED: 01/15/91 TEST DURATION: 1/2 hour TE*T DESCRIPTION: With the plant operating at 100% power, both main feedwater pumps trip. The main turbine trips, but due to a failure of the Reactor Protection System, the reactor remains on line in accordance with plant procedures, the control ods are deenergized by opening 480V breakers 3305 and 3312. The reactor trips and the plant is stabilized in the normal post trip window with steam generator levels being controlled by emergency feedwater. MALFUNCTION 1111.E OPTIONS RP-011 Reactor Trip Module Fails Selectable to A. B, C, or D RPS to Trip Channels INITIAL CONDITIONS: Full power steady state conditions, all plant equipment in service. FINAL CONDITIONS: Peactor/ Turbine tripped. Pisant stabilized in the normal post trip window with steam generator levels being maintained by i emergency feedwater. PLANT PROCEDURES USED DURING THE TEST: AP 580 REV 15 REACTOR TRIP BASEllHE DATA: There is no plant data for this transient. Expected simulator response based on expert judgement. I 1 i 1 O 1 i _
PERFORMANCE TEST NUMBER: PTH748 DATE: 01/15/91 TITLE: ATWS (OPERATOR INTERVENTION) TEST RESULTS/ COMMENTS: Test results evaluation based on expert judgement. Test results satisfactory. One (1) Trouble Rerport written on the improper operation of the undervoltage shunt trip feature on the CRD Breakers. Until corrected this problem can be worked around through the use of conditional malfunctions to open the CRD breakers when their 480 V feeder breakers are opened. TR does not require a rerun of the test. No other performance problems noted. O v O
c TEST NUMBER / TITLE: PTH24B/ATWS (OPERATOR INTERVENTION) TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Paramotor DEVICE CHART SPEED
*Hl:ll3 Pressurizer Level Grfx Rcdr 0 - 320 in .9375 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 700F
*L4JPR222 RCS Pressure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 - 2500 psi
*RR5WTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 1000 MWe Ll:6ADT delta T Loop A Grfx Rcdr 0 - 50F Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50F
*LlJPR212 Th Loop-A Grfx Rcdr 520 620F
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620F
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 620F
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620F
*LlJPS287 SG A SU Level Grfx Rcdr 0 250 in
*LlJPS293 SG-B SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG A Op level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Grfx Rcdr 0 100%
*F3JPA323 SG A Press (EFIC) Grfx Rcdr 0 - 1200 psi
*F3JPA320 SG B Press (EFIC) Grfx Rcdr 0 - 1200 psi LlJPT228 HS Loop A Hdr Press Grfx Rcdr 600 - 1200 psi LlJPT232 MS Loop B Hdr Press Grfx Rcdr 600 1200 psi
(
*LIJPS301 SG A HFW Flow Grfx Rcdr 0 - 6K KLB/HR
*LlJPS302 SG-B HFW Flow Grfx Rcdr 0 - 6K KLB/HR *
*LlJPS329 SG-A SU FW Flow Grfx Rcdr 0 - 1500 HLB /HR
*LlJPS330 SG-B SU FW Flow Grfx Rcdr 0 1500 MLB/HR L5G5AV Grp 5 Avg Position Grfx Rcdr 0 - 100%
L5G6AV Grp 6 Avg Position Grfx Rcdr 0 100% L5G7AV Grp 7 Avg Position Crfx Rcdr 0 - 100% HQBBCORE RCS Boron Grfx Rcdr 0 - 3000 ppm F3:23Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm F3:25Fil EFW Flow S/G A Grfx Rcdr 0 - 1000 gpm F3:26Fil EFW Flow S/G A Grfx Rcdr 0 1000 gpm F3:24Fil EFW Flow S/G B Grfx Rcdr 0 - 1000 gpm O O f
C PERFORMANCE TEST NUMBER: PTM248 DATE: 01/15/91 TITLE: ATWS (OPERATOR INTERVENTION) TEST STATUS: 1.0 This test is complote and has the following deficiencies noted (if none soindicate): I 1R # Priority Date issued Date Closed 189923B _lL _ _ML.%L9L > ,
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mA . DC Test Coo'rdinator -/DTte, 2.0 All deficiencies have been resolved and this test has been completed. i ,f'.f -['
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- Closure of TR-23H does riot prevent
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~ completion of thio tent FPC lese Coordinator /DTtd 3.0 Test results reviewed. Simulator performance ls satisfactory, i
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{ l TEST NUMBER / TITLE: P1M9C/0TSG OVERflLL ANSI /ANS-3.5(1985) CROSS
REFERENCE:
Section 3.1.2(9) DATE PERFORMED: 01/11/91 TEST DURATION: I hour TEST DESCRIPTION: The simulator is operating at approximately 73% power with the C RC Pump secured, when a failure of the A MfW pump governor causes the FW pump speed to increase to 4950 rpm. The increase in feed flow to the A 015G causes an incease in generator level, a reduction in superheat, and a corresponding reduction in RCS temperature, pressure and pressurizer level. In order to tarminate the overfeed the A MFW pump is tripped, resulting in reactor / turbine trip on high RCS pressure. MALLURCJJON TITLE 0_PTJ0tn 1VSXPGV feed Pump Governor Valve - Selectable to A or B FW Pump fail to Specified Position Adjustable from 0 - 100% INITIAL CONDITIONS: 73% power, C Peactor Coolant Pump secured. FINAL CONDITIONS: Reactor / turbine tripped, plant stabilized in the normal post trip window with the C RC pun.p and A MfW pump secured. ( PLANT PROCEDURES USED DURING THE TEST: AP-580 REV 15 REACTOR 1 RIP BASELINE DATA: UOER 83 Reactor Trip on High RCS Pressure following a Main feedwater Pumo Trip. 08/26/83 - RECAl.L DATA flLE D082683 - 01SG Overfill / Reactor Trip on High RCS Pressure. 08/26/83 4 b'
g PERFORMANCE TEST NUMBER: plH9C DATE: 01/11/91 TITLE: OlSG OVERflLL TEST RESULTS/ COMMENTS: lest results satisfactory. No performance problems noted. Excellent correlation between simulator and actual plant data. Note: lest rerun 08/11/91 to obtain second graphic recorder output. liad software label
" JPR310) used on first test.
Test rerun 09/13/91 to obtain graphic recorder output for plant / simulator data comparison utility run. Added main steam temperatures. ,o D
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TEST NUMBER / TITLE: PTM9C/0TSG OVERf!LL TEST DATA LABEL DESCRIPTION RECORDING RANGE FREQUENCY /
- Critical Pavameter DEVICE CHART SPEED
*Hl:ll3 Pressurizer level Grfx Rcdr 0 - 320 in* 3.75 sec
*HlJPR203 Pressurizer Temperature Grfx Rcdr 0 - 700f
*L4JPR222 RCS Pre.,sure (NR) Grfx Rcdr 1700 - 2500 psi
*L6JPR208 RCS Pressure (WR) Grfx Rcdr 0 2500 psi
*RRSWTP Normalized Reactor Power Grfx Rcdr 0 - 125%
PGJPE202 Gross Generation Grfx Rcdr 0 - 1000 MWe Ll:6ADT delta T Loop-A Grfx Rcdr 0 - 50f Ll:6BDT delta T Loop-B Grfx Rcdr 0 - 50f
*LlJPR212 Th Loop-A Grfx Rcdr 520 - 620f
*HCTAVG RCS Average Temperature Grfx Rcdr 520 - 620f
*LlJPR214 Tc Cold Leg Al Grfx Rcdr 520 - 620f
*LlJPR215 Tc Cold Leg A2 Grfx Rcdr 520 - 620f
*LlJPS287 SG A SU Level Grfx Rcdr 0 - 250 in
*LlJPS293 SG 8 SU Level Grfx Rcdr 0 - 250 in
*LlJPS284 SG-A Op Level Grfx Rcdr 0 - 100%
*LlJPS336 SG B Op Level Crfx Rcdr 0 - 100%
*f3JPA323 SG-A Press (EflC) Grfx Rcdr 0 - 1200 psi
*f3JPA320 SG B Press (EflC) Grfx Rcdr 0 - 1200 psi LIJPT228 MS Loop-A Hdr Press Grfx Rcdr 600 - 1200 psi Grfx Rcdr 600 1200 psi
$ **LlJPT232 LlJPS301 SG A MfWMS Loop-B Hdr Press Flow LlJPS302 SG B MfW flow Grfx Rcdr Grfx Rcdr 0 - 6K KLB/HR 0 - 6K KLB/HR
*L1JPS329 SG-A SU fW Flow Grfx Rcdr 0 - 1500 MLB/HR
*LlJPS.130 SG 8 SU FW flow Grfx Rcdr 0 - 1500 MLB/IIR LlJPS288 S/G A Downcomer Temperature Grfx Rcdr 0 - 600 F L1JPS289 S/G B Downcomer Temperature Grfx Rcdr 0 - 600 F FTT60 FWHE-3A Outlet Temperature Grfx Rcdr 0 500 F FTT62 FWHE-3B Outlet Temperature Grfx Rcdr 0 - 500 f SRT01 A-MS Temperature Grfx Rcdr 100 - 650 f**
SRT07 B MS Temperature Grfx Rcdr 100 - 650 f** Notes: *Second graphic recorder file run (Replacement of bad pressurizer level variable)
** Third graphic recorder file run. Main Steam temperatures added for
! plant / simulator datc comparison plots.
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{I PERFORMANCE T T NUMBER: PTM9C TITLE: OTSG OVERFILL DATE: 01/11/91 OBJECTIVES: 1.0 \erify proper control and instrumentation system response to an OTSG overfill. 2.0 Verify proper thermal / hydraulic response to an OTSG overfill end subsequent main
'eedwater pump trip and reactor / turbine trip.
TESf ACCEPTANCE CRITERIA: 1.0 GEN (PAL 1.1 The observtble changes in the simulated parameters correspond in direction to tho:e expectea and do not violate the physical laws of nature. 1.2 The simJlator does nel 1. to Cause an alarm or trip the reference plant woulo i. e caused an alarm or trip, and conversely,ti.e simulator does not al t s. or trip if the plant would not have alarmed or tripped. 1.3 The response of sim;1ator cor. trol room instrumentation is realistic to the extent that within the criteria listed in 1.1 ab'. , the operators do not observe a difference between the response of the i simulator control room instrumentation and the reference plaht. 1.4 The degrce of simula+ ion of .ontrol room instrumentation and controls and plant. ystems, including eq :i pment and controls operated outsi: che control room 'a sJficient to perform the required evolution. 2.0 SPECIFIC 2.1 Specific acceptance test criteria for this test are identified within the procedure.
- _ - - . . ~ _ _ _ _ _ -
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DATE: 01/11/91 PERFORMANCE TEST HUMBER: PTM9C TITLE: 0TSG OVERF!' L INITIALS / S1EP # TEST PROCEDURE TR HUMBER M9C010 Reduce load to 73% (595 MWej at a rate of 2%/ minute with the ICS _ in full autom tic. When the plant reaches the desired load, secure tha C Reactor M90015 Coolant pump. _pi
/
M9CO20 Reduce pressurizer level to 200 inches. _ M90025 Borate the RCS to establish an initial group 7 rod po.ition of c
-85%.
M9C030 Set a conditional malfunction to block CRD Group 7 withdraw commands when Group 7 rod position exceeds 88%. (This simulates the Group 7 90% rod withdrawal stop which was in effect at the / time of the transient) M9C035 When feedwater flows have stabilized at the proper values for three RCP operation, place both MFW pump ICS stations in HAND and increase A MFW pump speed to 4950 rpm (1070 psig pump 8 discharge pressure) at the I/F by failing iss GV to 0.32 _f M9C040 Verify the following expected responses: a) A Main Feedwater flow increases to > 6 E06 lbm/hr, [V b) A OTSG operating range level increases to approximately 100%. _p mom GENERATOR A LEVEL HIGH/ LOW Alarm (K-7-2) _f c) RCS temperatures decrease, with Tave stabilizing at V approximate'y 570F. _ d) Pressurizer level decreases 'o a miniinum value of approximately 165", then oegins increasing towards setpoint as makeup flow increases. _ RCS pressure decreases approximately 60 psig before e) being restored to setpoint by the pressurizer heaters. _ [<_ y/ , f) Reactor power increases i mm 73% to approximately 75%. _fl g) A OTSG st<sm temperature decreases by approximately 50F, gy ' O ,
8 PERFORMANCE TEST NUMBER: PIM9C
=.
DATE: 01/11/91 TITLE: OTSG OVERFILL STEP # TEST PROCEDURE INITIALS / TR NUMBER h) A OTSG downcomer temperature decreases by approximately , 60F, _ e M9C050 At twelve minutes after the start of the transient, manually trip the A MFW pump. _ d_ M9C060 Verify that the reduction in feedwater flow caused by the pump trip causes the following responses: a) RLo temperatures, pressurizer level and RCS pressurizer begin rapidly increasing. , 6, b) A OTSG level begins decreasing. _[/ c) A OTSG steam temperature and lower downcomer tempera'ure begin increasing. _[ Reactor power begins decreasing. d) _[/ e) A reactor / turbine trip occurs on hign RCS pressure. __ f-M9C070 Verify that the plant stabilizes in the normal post trip window. / THIS CONCLUDES TEST PTM9C L f.m., L
l C PERFORMANCE TEST f: UMBER: PTM9C DATE: 01/11/91 TITLE: 0TSG OVERFILL TEST STATUS: 1.0 This test is complete and has the following deficiencies noted (if none so indicate): TR # Priority Date issued Date Closed
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f fPC Nst"C'oordinator [ Dan 2.0 All deficiencies have been resolved and this test has been completed. O NOT REQUIRED FPC Test Coordinator /Date 3.0 Test results reviewed. Simulator performa, ice is satisfactory.
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l UNPl,UdNr.D OPtJLATING r. VENT REPORT &1-1I Reacter Trtp on High Reactor Caetant Prenews Fo11ewteg a Main Peedwater hany Tet, August 26.1983 4 SLAufTTED
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EXECUTTyt
SUMMARY
CRYSTAL RIVER UNIT 3- AUCU5T 24,1983
SUMMARY
Ors Augue; 26, 1911, Crysta! River Urdt I wu operating at 73% reactor power generating 640 M *e. The lategrated Control Srst*m (ICS) was in full autornatic. Reactor Coolant Puenp lC was tecuted dse to a laulty leal. At appfontmately 0819 binding in the gr.ernoe lar* age f or the A dam Feedwate. Pump The operator (MFIi caused t;w pump speed to hang up and overleed the A steam generator. attempted to control the A MFP speed manua.ly. Wever, the pump would ret respJewt. MFP A was manually tripped at approstenstely 0907 be. tune of rugh level in the A steam smer a tor. The rendt6ng uniderf eed trentient caused . reactor trip on high Reacto< Cootet Systein (RC5) pseuu*e at 0908. ROOT CAU5_t Tre govert ce leaage on MFP A .at out et alignmeit alto.ing bmomg twt een the voke and thef t, The lidmage then'htng up .ith the governor valves in an overleed positiurs PJRPOkMANCE ANOMAltf.5 j e. - The revernor control linkage f or MF6' A =an out of alignment and bmcing. A lip on the constrol shalt then caught on the yt*e shach prevented the governor valves f rom lX; moving m the close direction 47 eesuiicd X1i' . vv a,n gene,aio, i,,ei 45 ,aimi .imve the ti,e of ine spirating pu is. r h., in s.stue.ied stcain rond nons in Main Steam piping (nr appe nsimairle us emnotet hine tagged due to overloed
- Aporats:nately 10 mmutes of annWlator events
- eve S {- >i the annunciator butter, 1-m v
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i IYTMI AIM 3H ENT 4 f _ g, A. StOUtNCE OF EYtMT1 The sequenew of events wu detwmited from the Arveunciatw Erwits Printout. h
/ Computw Post Trip' Review, Computee Alarm Summary, Recall $retem, Opweta-L oss, and Interviews. _
Over to minutes of ennunciator events wete not time tagged due to ovefload of the shriunctatur bufler. A pattial listing of sigrulicant events f or that transient as shuen taelo . M TILF. " VENT 50VRCE q _ Otondo Mf P A speed irm reasing. H [ - 08:49 0? \ team Generatur \ high le.cl alarm. C
.y Mr P A and in .n m.ariual. R 05:19645
) A 04:19:19 64C Pre ssuraiet level low. HC Cold Leg Loop A & B Temp Ditterential A 09:00:18 High(aTc). MFP A tripped. A,C 09:06:11 RC Pressurtter level high. A,C 09 07:17 5 team Gerwestor A Ngh level alarm clear. A,C 09:07:31 __ High R eactor Coolant Pressure. A
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09:08i % Reactoe Trip on Ngh RC pressure m A.C channelB & D. Main Turbane Tr6p. A. Aruwnw:lator Alarm TrPef - C. Plant Computet Data R- Recall System Date ll. Pt ANTT PERPORMANCE -
- 1. Pre-Trio Review Crystal Rivw Urdt 3 was operating at approalmately 73% poww and sweating 640 MWe with the ICS in full automatic. Three Reactor Coolant Pumps were operating with EtCP IC secured ihm to a gump seal f ailure.
- MPP A was applyy most of the RCS host removal due to the RC pump configwation. Level in Steam Generator A was approntmately g3% on the operats range.
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- 2. Initietina Event
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At omrostmately 0859 the speed of MPP A incteased from 437) r, n to 4930 This
' rpen, and pump diaAarge presaure Irw.reased trorn SM pstg to 1970 psip i resulted in high oil sole reedkes en the A eperate end startup level Instruments and approximately 400 inches an the full range indication (refw to Figure ik i Due to the high generator !* vel,' the septratlig ports in Steam Conwater A wwe
! covered which resulted in a la.,ge drop in downcorner temperature and reduction in steam temperature to saturation (Figure 2). The trcreased heat removal csused RC cold leg temperature (Tc) to drop from SMcF to 34)of' Wigure 1) and a sesequent drop In RL avnage temperature IT.,,) : toF.
The operator attempted to control the overleed by takmg both MFP hadry IC5 stations to Wi ar.d manually reduced the output of the A station. The MFP did not resp nd. A furthee attempt to control MP2 A by closmg the low speed governor valve also f ailed, in order to stop the overleed, ttu operator t tw*n tripped MFP A. This erwtisted an ICS plant runbach, the operator manually increased MFP h speed and initiated pressuriser spe a r but .as smable to compensate for the loss of heat removal wtuch resulted f rom trippmg MI'P A. The reactor trapped nn high RC pressure at 0901. The cause of the increase in MPP A speed oss subsequently found to be a sitrk mechanical linkage from the l'.HC servomotor to the main 'eed ater turbine governor valves. The linkage had apparently tuen slightly bent and misaligne<t with enaimmance duttrg the ref ueling outage a month earlier. The lituage then vibrated f urther out of abgnment during operation until binding vet.cred.
- 3. Post Irio R espoyn The resonse of the RCS to the reactor trip is illustrated in Figures M. Figure s shows RCS tempesature and preuure f ollowmg the trip, Response + 4s as espected eith the plant returning properly to the post tri,a wirwow. Preuuriser level is illustrated in Figure 3. The minimum post trip pressortsee level was 91 inches, theref ore, the use of HPl to maintain lesel was not required. The higher than normal post trip pressuriser level may be attributed to the escessive makeup during the tranuent prior to the trip plus the tact that sec ondar y Q resporue .as proper with no f eedeater overleed or saf ety valve stisch opem Figure 6 allustrates secondary response to the reactor trip, Response was as especteo eith steam generator preuure rising te 1010 png and steam generatur d
levels dropping to M inches (now level limits),
- s. Operstor Adions aM Procv4vral Adequacy Tbe operstor actions seere appeoptlata for the MFP overfced, howeyer, plant processes did ret edequately address the event. AP401, ICS I Annunciator Response 1 1 1 and 14-1, lists actions to be performed for an uncontrolled high level in a steam generator. Triming the MPP is not listed as one of the actions to take, however, and Ls appopriate f or this situatiers The procedire to folle' attw the trip of the IAFP h c& tressed in AP441,l.oes of IAala Feed Pump Runbeck. The proceeure and operator remporwe was correct for this event, however, the actlers vers not enegh to prevent a reactor trip.
HfP A was carrytrq a s1pl0 cant amount of the Ltdtlat heat remoesi load and 4 : *., ,. y ,:g- .,.g .'
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s. 3 . ' , . , s6..,. .. t^ i* tripping the putnp placed the unit in e large tratudent. If the plant had been Y*e l-epe et in a b41stcod state pelar to the tranalent, the opeestar actlans may havee ded a reactor trip. -
/
i AP480, Reactor Protection System Actuation, addresses the actions to be taken iollowing a reactor trip. The procedure and operator actlans f or the toector trip a were appropelate and adequat . C. S_AfTVY CONiltERATtONS
- 1. H eactivit y Control Ttw reacto, .as shutoo n eith all control root m the bottom ibe to ste trip signal from the Heattur Protection system,
- 7. T her mal Contr al A dequa te H Ch tuhrtuling mar gin was enannt ained t*st oughout the trentient.
Preuuriser level remained on scale eithout the use of any Entirwered %af eguard j~ systein. T he electro natic reliet valve and RC5 safet, valves .ev e not 7 t'h.slienged to lemst NCS pressure. Secondary heat removal eat maintamed eith the f eedw a ter syste n and MFP li, emergency f eed = a t e, actuation wat int r eqmt evt. The ovefleed of Steam Genef ator A resulted in a level of a00 arches for approaimately 5 minutes. This esceeded the Techn. cal Specificat6on limit of %0 inc he s. Tet:haical Specif 6 cations allow tartain parameters to enceed their analysed range f or short petiods of time, ov ed on the amount of enk involved. F or tieam genec ator level, the man 6 mum level n t>ased on the initial conditicni anumed f or a gleein line bream accident arul Technical Specifications alto. 8 operation for up to 6 hourt be. ore the unit must he sruitocen. Tha event uoes not, theref ore, constitute an unreviewed saf ety question erwt the i minutes %f the leve! = at outlede the limit e en within the time f rame allo *ed.
- 1. RMioactive Inventory Control Due to a small amount of steam genef ator tube leakage, a radioactive releue
. was made vla the main s.eam Saf ety valves and the atmosphefic dump valves.
Ttus release did not cause ar y radioactive release limits to be exceeded. 4 Safety Eculement Aesilability No safety related equipment was damaged or ot*wrvloe made inoperable daring this event. Reactor Coolant Pump IC was secured due to a f aulty seal. This resulted in RC loop 6 pressure instrument reading H pang higher than loop A I presaure. Pressuriser sprey stui heater control was selectea to tenp B per 09 y' 204, Paarer Operation, therefore the plant was not initially operating closer to the high pressure trip setpo&nt. 4
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This trannlent was caused trr mechanical b&rwilag eif t.*w linkage between the MFP A EHC servenotar and the governer valee actuator. The mecharitcal binding cawed an l oeerteed of steem generator A and edeequerst overcoolin6 of the Reector Coolant 5yatem. MFP A was tripped in order ta stop the overcootleg, and the ersoulng tranalent resulted in an urdercooling arid reactor trip on Ngh RCS pressure. The post trip primary and secondary resporsse was normal E. CORRECTIVE ACT10N5 The problern associated with the events recurder butter overic *d at Lewig ad@*sted br REl 10*ll*20 Money has been silocated to commeece a study f or s..
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A proceArc change has been submitted f or AP 301, ICS I Annurriator Resp. *
. 'n or der to incts.de ac tiant to be taken tar an overspeed of a MFP daring normal opef dthorb 4
The control linkage im vre 'A' . repaired and realigned. A procedure change het been subonitted for OP40%, Fecit*4ter syste.n, ihn rhange / mill adat J 4tep (a cheeb the \tfP f tmtrol lnkage f or propet alig8Hneul .tnd ;sper dt hm e hen t he gue n p n s t a r t e.t. i' h
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