ML20070C881
| ML20070C881 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/15/1991 |
| From: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | |
| Shared Package | |
| ML20070C879 | List: |
| References | |
| NUDOCS 9102270013 | |
| Download: ML20070C881 (317) | |
Text
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. U l NVC48aR RtGVLAtORY CO.ensaseeo IM'.". W ' SIMULATION FACIU( CERTIFICATION eh t e ,, o%.
1 INSTRUCTIONS, t Th4 form is u tie fiied 'or inst.4 certif. cation, recertification Of reowiredt and for any w change to e sem ietion f acihty a perfo mede ef ef snitiet lut>msttel oft wCh a Dien Provice the follo.mg s6formet on and CheCh the SDDropriate t>Ca to ind#Cete f eelon for tobmettel i , e ACiut, ( jooattNvveta Seabrook Station Unit 1 1 g 443 LIC E N5i t lDAft
'Public Service Company of New Hampshire et. al. I 2/15/91 3, .e* e i ., - w . , . ,
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NAW( ter since W,"dsersons AND LOCatsON 08 Sivstaf *ge r 8 AciLity y l $iuvLAf s0N 8 ACittiv PE R80 AM ANCt it st A nstm AC15 A f tACMt 0 stor sertermence resta tena.riar . cae ps,.ev eaa ,y v4ta tae se*e e o m e s ,,0N o, ,. . . o.. A NC t 18,1 tNo Co ,L i r t o m e,,_,-r, .,. e,. .o ..,, - .,_re .., -
.'3ee attached documentation.
Erwut AiiON f ACitsiv PE A* e V re= e,w,ao e,ta cae cute ro ra rerti8we'e O awANC4 4 Y(51.N4 5CatOvLt A T T AcalO ts r tae reasucr et approsaeter th of parr rmoare rarts per y.or e o r r rae rfou yew ser>ea 04 5C#s'f10N 08 ,8 m8C AM ANCE t tSTING TO BE CoNDucfID Mereca eer,eae<aepts; es accamasey ata ,oentify tae seem meer, pre emn, coate wi See Appendix G of the attached documentation. t V l #t R80mw ANCE TitTING PLAN CHANGE.f#er ear med>6cer-oa fe o e serv mente terting seen swe,rtev en e preview w certa steJ 045Ca TION Of Al R8OmWANCE illTING PL AN CM ANGE lAtrera edartte<W ##pta! es accesasry, sad #@pardy tne seem assersprecwe es sup tearsaved/ I atCin t ne ICA1K;; Wo.crece correctove act.one te.ea arraca r arts or temporar p.rtormente toetsnp na esserence w,ta 10 Cf n 4 $$ esosou6ari l A neca eeteswi se.etsi es i.resapey sad " spar tr rate afwa esserrar+oa ee.ae coarm oed 8 y' ) , Aa w u.i.a c., s
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.a iooai c.. ne u. .,aa.. p.wi, e. , m.u n. ..,m.i..a ,a m.i SicN - WoyED R ESENT ,#
l'" Executive Director of 0' O e(n accordance urp
'/Ag y Nuclear Production l# # I-Cs H 4 55 5. Commw n atio is, this form snett tw suomitted to the N HC si tonows
,.l , BT MAIL ADDOS$ED irector, Office of Nucneer Reactor Regulation 8Y DELIVERY IN PERSON
{ U.S. Nucteer Reguletory Commisiaen TO THE NRC OFFICE AT, 7920 Norfolk Avenue weenington.oC 20sss s.the.oe uD ' l 9102270013 910220 PDR ADOCK 05000443 P PDR
r y: INTRODUCTION
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-The initial certification of the Seabrook Station Unit I simulator is being
-submitted to the Nuclear Regulatory Commission in accord with the requirements of 10CFR55.45. The certification document has been organized as outlined in Appendix A of ANSI /ANS-3.5, " Guide for Documenting Simulator Performance".
Abstracts for each performance test conducted have been written following the guidelines promulgated by Mr. Neal K. Hunemuller of the Nuclear Regulatory
-Commission in his paper, " Simulation Facility Evaluation Program". Exceptions taken to the requirements of ANSI /ANS-3.5 are compiled in Appendix G of the certification document.
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INTRODUCTION
,,_ / 1.0 SIMULATOR INFORMATiON 1 1.1 General Information 1 1.2 Control Room 1 1.2.1 Control Room Physical Arrangement 1 1.2.2 Panels / Equipment 2
-1.2.3 Systems 2 1.2.3.1 Simulated Systems 2 1.2.3.2 Main Plant Computer System 3 1.2.4 Simulator Control Room Environment 3 1.2.4.1 Communications 3 1.2.4.2 Lighting 4 1.2.4.3 Aural Simulation 4 1.2.4.3.1 Annunciators and Alarms 4 1.2.4.3.2 Steam Noise 4
.1.2.4.3.3 Rod Step Counters 5- 1 1.2.4.3.4 Ventilation Noise' S
'1.3 Instructor- Interface 5
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~
1.3.1 Initial Conditions 5 x^ 1.3.2 Malfunctions 5 1.3.3 Controls Provided for Items Outside of the Control Rocm 7 1.3.4 Additional Special Instructor / Training Features Available 7
'1.3.4.1 Simulator.0perating Limits 7 1.3.4.2 Input / Output (1/0) Override 7 1.4- Operating Precedures for Reference Plant 8 2.0 SIMULATOR DESIGN DATA 9 3.0~ SIMULATOR TESTS 10 3.1; Computer Real Time Test 10-3.2 Steady State and Normal Operations Tests ' 10-3.2.1 Steady State Stability 10 3.2.2 Steady State Accuracy- 10 TABLE 1A: Exceptions (CriticalParameters) 12 TABLE 18: Exceptions'(Non-Critical. Parameters) 15 ,~ 'd i l
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i- >= TABLE 2A: Discrepancies (Critical Parameters) 17 TABLE 28: Discrepancies (Non-Critical Parameters) 18 Connent and Justification Code 20 3.2.3 Normal Operations Tests 21 3.2.3.1 Startup and Shutdown: Cold Shutdown to Hot Standby 21 3.2.3.2 Startup and Shutdown: Hot Standby to Minimum Load 22 3.2.3.3 Startup and Shutdown: Minimum Load to Full Power (Powerincrease) 23 3.2.3.4 Startup and Shutdown: Full Power to Minimum Load 24 3.2.3.5 Startup and Shutdown: Minimum Load to Hot Standby 24 3.2.3.6 Startup and Shutdown: Plant Cooldown 25 3.2.3.7 Shutdown with less than Full Reactor Coolant Flow 26 3.2.3.8 Reector Trip and Recovery 26 3.2.3.9 Core Performance Test 27 3.2.3.10 Operator Conducted Surveillance Testing 28 Main Turbine Overspeed Test 28 51 Quarterly and 18 Month Pump Flow Valve Test 29 PCCW Train 'A' Quarterly Operability Test x : and 18 Month Valve Position Indication 30 PCCW Train 'B' Quarterly Operability Test and 18 Month Valve Position-Indication 31 RHR Quarterly Flow and Valve Stroke Test and 18 Month Valve Stroke Observaticq 32 DG 1A Operability Surveillance 33 OG 18 Operability Surveillance 34 MS Isolation valve Quarterly Test 35 EFW 18 Month Auto Actuation Surveillance ~ 36 Quarterly Flow and Valve Stroke Test and Monthly Remote Position Indication Verification 37 3.3 Transient Test 38 3.3.1 Annual Transient Tests 38 Reactor Trip 39 Simultaneous Trip of all Feedwater Pumps .40 Closure of all MSIVs 41 Trip of all Reactor Coolant Pumps 42 Trip of any Single Reactor Coolant Pump 43 Main Turbine Trip 44 Maximum Rate Power Ramp _45 RCS Rupture with Loss of Offsite Power 46 ii
Main Steam Line Rupture 47 y Slow Primary Depressurization 48 j] 3.3.2 Transient Occurrence in the Plant 49 3.3.2.1 Natural Circulation Event (LER 89-08) 49 3.4 Malfunction Tests 50 Auto Rod Control Fails In The "IN" Direction 51 Auto Rod Control Fails In The "OUT" Direction 52 Dropped Rod : RCCA 08 53 Dropped Rod : RCCA F8 54 Dropped Rod : RCCA H2 55 Dropped Rod : RCCAS H2 & H8 56 RCCA H8 Failure To Move On Demand 57 Rod Position Indication (DRPI) Failure For RCCA H6 58 Failure OF Automatic And Manual Rod Cc.ntrol 59 Failed Fuel Element (RCS Activity increase) 60 Failure Of The SSPS To Automatically " rip The Reactor 62 Simultaneous Trip Of Both Main Feedwater Pumps 63 , PZR Pressure Instrument PT-456 Fails High 64 Pressurizer _ Safety Valve (RCV-116) Leakaoe 65 Pressurizer Spray Valve (PCV-455B) Fails'0 pen 66 Pressurizer Heater Control Failure 67 Failure Of Reactor Coolant Pump "D" #1 Seal 68 RCS Manifold Leak 69
]'v Reactor Vessel Flange Leak 70 Rod' Ejection . 71 ,
Reactor Coolant System Cold Leg LOCA 72 1 Tube Rupture To "C" S/G 73 , Reactor Coolant Pump "A" Overcurrent Trip 74 ! Reactor Coolant Pump "0" Locked Rotor Trip 75 Reactor Coolant Pump "C" Overcurrent Trip 76 Reactor _. Coolant Pump "B" High Oil Temperature 77 , Reactor Coolant Pump "C" PCCW To Oil Cooler Leakage 78 i Reactor Coolant Pump "A" 011-Reservoir Leak 79 Reactor Coolant Pump "A" High Vibration 80 Reactor Coolant Pump "0" High Vibration 81
. Reactor Coolant Pump "C" Loss-Of Seal Wator 82-Main. Steam Line Break 83
-Main Steam Line "A" Safety Valve Fails Open 84 Main Steam Line "B". Rupture Inside Containment 85 Turbine Control Valve #4 Fails Closed 86 Turbine Control Valve #3 Fails Open 87 Simultaneous Closure Of All Main Steam Isolation Valves 88 Feedwater Regulating Valve To "A" Steam Generator Fails Open 89- a '
Loss Of Main.Feedwater Pump "A" 90 Extraction Steam Valve To.High Pressure Feedwater Heater 26B Closes 91 V iii l ! i
L yb. a Loss Of Condensate-Pump:"A" 92 W Loss Of Condensate Pump "B" 93 if Loss Of Heater Drain Pump "A" High Pressure feedwater Heater 26A Tube Rupture 94 95 5J , c Loss Of' Heater Drain Pump "B" . 96 High Level Condenser Dump Valve Fails Open 97 g Low Level! Condenser Makeup Valve Fails Open 98 ,s Loss Of Mechanical Vacuum Pump "A" 99 100 ;
-Condenser Tube Leak --
E' Excessive Leakage Through FRV To "C" S/G 101 i Excessive Leakage Through feedwater Regulating Valve To_S/G
- 'D' .
102
- Loss Of Main Condenser. Vacuum 103 4 l Loss 0f Startup feedwater Pump 104-- !
High Pressure Feedwater Heater Bypass Valve fails Open 105 y MSR High load Valve Fails To'Close On Turbine Trip 106- @ Heater Drain' Tank High Level Dump Velve Fails Open 107 , 1 Low Pressure Feedwater Heater Bypass Valve fails Open 108 e Main Steam Reheater-Tube Rupture 109 Failure Of-Seal Water To Main feedwater Pump "A" 110-m High Pressure Feedwater Heater.High Level Control Valve j 4 Fails Open _ . 111 _ {
+ Loss Of Circulating Water Pump "A" 112-Loss Of Circulating Water Pump."B" 113 j!
Loss Of Circulating Water Pump "C" 114- i
- Loss 0f1 Condenser l Water Box Priming Pump 115 l Loss Of-SCCW To Turbine 011 Coolers- 116-
[l]S - Loss Of SCCW To Generate Hydrogen Coolers Loss Of PCCW To CVCS Letdown Heat Exchanger 117 118 ) Loss Of PCCW To Centrifugal Charging Pump "A" 119- J" Loss Of PCCW To BTRS Chiller Package: 120 Leak In lne A" PCCW Loop 121 g Total-Loss Of:"B" PCCW Loop
~
122 , Loss Of-Service' Water Pump "A"_ __ . 123 1
-Loss Ofl Service Water To:The""B" PCCW Heat Exchanger 124 1 Loss:0f Servoce Water To The "A" PCCW Heat Exchanger 125 ;
"B" Thermal Barrier Leak To~TBCCW 126. _: ;
cReactor' Coolant Loss Of_PCCWLTo ThePump'B' RHR; Heat Exchanger-127-Loss Of PCCW To Thes'A' RHR' Heat Exchanger 128-Fouling Of Steam Generator-BL0wdown Flash Tank Heat , Exchanger "A" , _
. '.129 1 j Fouling:0fLSteam Generator Blowdown Flash Tank' Heat i ExchangerE"B"- . _ -- _
- 130 Loss.0f Seal. Injection To All: Reactor Coolant Pumps.. 131 132i l
< -LetdownIsolation1 ValveValve
_Back1 Pressure' Regulating (V-150)(FailsClosedPC-V131)' Fails 133Open: i
.Back Pressure Regulating Valve (PC-V131) Fails Closed 134. .:
* ; Temperature Control Valve (TCV-129). Failure To VCT Position 135 Letdown Line: Leak - 136' Ti Auto Makeup Controller Malfunctions 137
= Makeup Controller Fails In Borate Mode 138- i VCT.High' Level Divert Valve Fails In VCT' Position 139 4 '
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g' g i E h w j p VCT High Level Divert Valve To Return To VCT After WWTk Diverting- _ BTRS Dilute Mode Failure-140 141 j
1 W. . Seal Flow Control Valve HCV-182 Fails Closed 142-
'F4 .
- Charging-Header Leak 143 Auxiliary. Spray Valve Fails Open 144 0
e , Power Range Channel 41 Fails High;
- Power Range Channel 42= Fails High 145
'146 9 Power Range Channel 43 Fails High 147
. Power Range Channel 44 Fails High 148 9 F ^ Source Range Channel 31 Sluggish _ Response 149 i "3"
Loss Of Power To Source Range 32 Power Range Channel 41 Upper Detector Fails Low
-150-151 i ;, Power Range Channel 42. Upper Det7ctor' Fails Low . 152' l Power Range Channel 43 Upper Detector _ Fails Low 153 j Power Range Channel,44 Upper Detector fails Low 154 4 Intermediate Range Channel 35 Overcompensated 155'
_ Intermediate Range Channel 36 Overcompensated 156 A - Intermediate RangelChannel 35 Undercompensated- 157
>$ Intermediate Range Channel 36 Undercompensated .158- 4 Z Source Range Channel:31 Fails lo Deenergize 159_ l Q",
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Intermediate Range Channel-35 Loss Of Detector Voltage Intermediate Range Channel 36. Loss Of' Detector. Voltage-
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160-161 y<, Total Loss Of Offsite, Power 162 "m Loss Of.-VAT Feed To 13.8 KV Bus,1 .163-
~
= Loss:0f RAT-Feed To 4.16 KV Bus E-6 164-0 Loss:0f'125 VOC: Bus 11A .
-165
/~y - l Failure Of'"A" Emergency Diesei Generator To Auto Start. 1664 1
' Loss Of: Einergencyl 0.iesel Generator- 1B ' 167! !
b( f Loss Of MCC 231 .
- Loss.Of Motor Control Center-111
-168L 169 +
i High Turbine Vibration-- 170 : 1.oss Of> Automatic-Load Contrcl Unit: 171
- 6. . & -Loss Of-AutomaticiSpeed Control Unit' -172:
# Automatic Voltage Regulator Fa' lure 1731 >
Main Generator Seal-011 Systen. Milu_re . , 174 e 4,
. Loss Of Electro-Hydraulic Suppb Pump-"A"l 175- A Loss Of Electro-Hydraulic Supply Pump "B" L176:
' Main Turning Gear: Trip 177- o Partial Load Rejection. . -
178,
-Failure Of Hotwell Level Transmitter, 179- -
4
+ Steam Generator "C" Level Channel LT-537 Fails Low -180
- Steam Generator:"0" Level Channel LT-549 Fails low 181:
Failure Of ~ Main Turbine Protection' System To Auto Trip LThe
~
' Turbine- - L182 m
.S/G "B"JLevel' Transmitter LT-529 Fails Highi '183- <
.'RCS-Wide Range Pressure Channel PT-403 Fails To 1000 PSIG, 184; +
,a RCS Loop A' NR TC Fails 'High 185a J 4 Pressurizer Level Channel LT-459 Fails Low ~ .186- 1 Pressurizer Level Channe? LT-549 Fails High -187- j Turbine 1mpulse Pressure PT-505 Fails Low- 188:
Reactor Coolant. Loop NR T n Fails Lo'w-189.
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Main Steam Header Pressure Instrument PT-507 Fails High 190 Loss Of Fire Protection System Jockey Pump 191 Failure To The Waste Gas Outlet Radiation Monitor- 192 Failure Of Containment Atmosphere Radiation Monitor 193 Inadvertent Containment Isolation Train "B" 194 Inadvertent Safety injection S gnal Train "A" 195 Loss Of Instrument Air 195 Loss Of Main Plant Computer 197 Loss Of Flow To Spent Fuel Pump A 198 Loss Of UPS Inverter 1B 199 Loss Of Normal Feedwater 200 Feedline Break Inside Containment On Feedline 'A' 201 Loss Of Stator Cooling 202 Failure Of Manual Reactor Trip Switches 203 Inadvertent Reactor Trip 204 MS 1 solation Valves All fail Open 205 Total loss Of Emergency Feedwater 206 MS Line B Safety Valve Fails Open 207 Main Steam Line 'C' Safety Valve Fails Open 208 Main Steam Line 'D' Safety Valve Fails Open 209 Steam Generator Tube Rupture To "A" Steam Generator 210 Steam Generator "B" Tube Rupture 211 Steam Generator "D" Tube Rupture 212 Main Steam Line 'A' Steam Flow Channel 1 Fails High 213 Main Steam Line 'A' Steam Flow Channel 1 Fails low 214 Turbine Stop Valve #2 Stuck As Is 215 Rod: Drive Motor Generator "A" Breaker Trip 216 g' Rod Drive Motor Generator "B" Breaker Trip PORV 456A Fails Closed 217 219 PORV 456B Fails Closed . 220 Stuck Rods (RCCA H10 AND RCCA K8) 221 Stuck Rods (RCCA DB) 222 Main Steam Pressure Transmitter "A" PT-3001 Failure 223 Main Steam Pressure Transmitter "B" PT-3002 Failure 224 Main Steam Pressure Transmitter 'C' PT-3003 Failure 225
' Main Steam Pressure Transmitter "D" PT-3004 Failure 226 StuckRod(RCCAD12) 227 Stuck Rod (RCCA'H3) 228 DroppedRod(RCCAH8) 229 3.5 Simulator Design Changes / Enhancements 230 3.5.1 Steam Generator Model Upgrade- 23C 3.5.2 Simulator Computer Complex Replacement 230 3.5.3 '18 Node Reactor Coolant System 230 3.5.4 Radiation Data Monitoring System Upgrade 231 3.5.5 . Reactor Vessel Level Indication System 231 3.6 . Physical Fidelity Audit 232 h vi
,, 4
.i 4.0l SIMULATOR DISCREPANCY RESOLUTION AND UPGP,ADING 235 J3-\~,/:
4.1 -Identifyin0. logging, ccrrecting, and testing reported simulator-discrepancies 235 4.2' Tracking _ design changes incorporated into the reference plant 235 4.3 Outstanding Simulator Change Requests 236 , 4.4. Overdue Plant Design Changes 236 5.0 MISCELLANE0US- 237 5.1 Schedule For Testing 237 i 5.2- Simulator Review Committee 237 t 5.3 Operational Experience Review Program 237
-5. 4 Major Simulator Upgrades -237-5.5 Paneltof Experts 238 5.6 Plotting of Parameters 238 Appendices 239
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( SEABROOK STATION-SIMULATOR CERTIFICATION-(
. 1.0 SIMULATOR INFORMATION I 1.1 General-Informatio'n-This is the-initial-certification submittal for the Seabrook Station-Unit 1 g controlLroom simulator. The!Seabrook Station control room simulator was
- manufactured by The Singer Company, Link Simulation Systems Division, it was delivered and: declared ready _for training.in 1980. The simulator.is owned by-the. Joint Owners.ofLSeabrookiStation and operated by.the New Hampshire Yankee-
-Division of Public Serv.ico Company of New Hampshire. The reference plant'is a LWestinghouse-lfouriloop pressurized water reactor. rated at 1150_'MWe. .The y turbineigenerator.was supplied by the. General Electric Corporation, and the' .
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balance lofthe_plantjwasdesignedbyUnitedEngineersandConstructors, , n Incorporated.- jp) s; ' ' , .. .
- 1i2 Control Room' ,
h lj2.1,ControIRoomPhysicaliArrangement -
; :The~physicaliarrangement of the-simulator. control: panels.A I;and-the rear-lconsoleL(designatedCP-295)lareidentical.tothat'of.thecontrolroom.-_lFigur.e T, Control: Room [ Arrangement,andfigure2,S'imulatorRoomArrangement','are attached ~ .PanelslA -;I have been verified to-beidimensionallyLidentical to'
'their,. counterparts-in.the control;r'oom. The reariconsole is six? inches ;
ifurthersback from panels'A' .l'in.the' simulator than-in the control _ room.a3Thc. W , zthree desks within i the horseshoe-area are positioned.the same with respect to s
- panelsLAL--I. ,
$ Outside: ofithe horseshoe,L the simulator room and the' control: room differ. : [ L0nly the backs of the:p_anels A'- C:and G - l'are.simu. ted. CP-16L(NI
. Cabinet), CP-65 (Rod Drop Disconnect Switches), and CP-180A and CP-180B are i
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- N V 1e SIMULATOR F_00R 3LA\ !
+ g i FIGURE 2: _-_ _ - --------_ - -- _ -- --_ _ _ . - - -. _. . .. . 3-
fully simulated. In the simulator, CP-16 is located 145 feet closer to panels
- A - I than in the control room. CP-65 is located three feet above the floor b . facing south in the control room, but is at eye level facing west in the '
simulator. In the simulator, CP-180A and CP-180B are shifted laterally four feet in the direction of CP-16, the support columns just behind the rear console are missing, and the ceiling'is six inches higher _than in the control room. .The Simulator Review Committee (SRC) has evaluated these identified differences and has determined that they do not detract from training. 1.2.2 Panels / Equipment The panels identified in Figure 2 are fully simulated. All instrumentation, except as noted in the abstract for the simulator physical fidelity audit, has been verif_ied to be identical in appearance and location to its control room , counterpart. The following panels are called for in operating procedures, but are not simulated: seismicmonitoringpanel(CP-58),B0P_processcontrol cabinets:(CP-152A,_1528, 153, 175, 244, 297A and 297B), process I&C control cabinets-(CP-1, 2, 3, 4, 5, 6, 7 and 8), ATWS mitigation panel (CP-519), and
. the vibration monitoring panel (CP-299). The Simulator Review Committee has evaluated the. impact of not simulating these panels and has determined that their omission does not detract from training.
The Solid State Protection System (SSPS) cabinets (CP-12, 13, 14 and 15) are not presently simulated. The Simulator Review Committee had determined that one train of SSPS is required for training. -A Simulator Change Request'(SCR) has been initiated (SCR #90-183) and will be completed by December 31, 1992, 1.2.3 Systems 1.2.3.1 Simulat'ed Systems All systems that the operators interact with via the Main Control Board (MCB) in the control room are simulated, without exception, p a 2 1
l.2.3.2 Main Plant Computer System C') The Main Plant Computer System (MPCS) is stimulated by the simulation computer. The Safety Parameter Display System (SPDS) is included as a subset of the MPCS. All displays available to the operator in the con.'rol room are available on the simulator. Display devices include the four Video Alarm System (VAS) displays, the four operater selectable CRT displays, the two multifunction CRT displays, the STA station, four two-pen recorders, the megawatt thermal display, and the control room loggers. 1.2.4 Simulator Control Room Environment 1.2.4.1 Communications All control room communications systems are simulated with the following exceptions: The keypad on each of the telephone sets has been disabled to prevent interaction with phones outside the simulator room. Consequently, the Q student can only use the speed dial feature of these phones. Each speed
'~
dial button is labeled identically to its counterpari in the control
-room. Depressing the speed dial causes an extension in the instructor, booth to be rung, with an indicator shcwing the party whom the student is calling.
The Motorola radio system is configured as an intercom which cortmunicates with a station in the instructor booth. This implementation prevents interaction with the plant system, however the operation is identical to that in the control room. The power fail phone is implemented on the simulator. This phone system is used in the control room only when power is lost to the Dimension 2000 PBX. The simulated system is fully available, but it does not switch over to an outside line upon loss of power to the Dimension 2000. n () 3 i
The Simulator Review Comittee has evaluated each of these deviations ar.d has
- determined that they do not detract from training.
g gq U' .l.2.4.2 Lighting The lighting in-the simulator room is laid out in the same physical arrangement as.in the control room. Electrically, each lighting string can be deenergized upon loss of the corresponding electrical bus. The level of illumination in the control room and simulator (iffer because the control room lighting is attached to dimmers which are adjusted based upon crew preference. This has been evaluated by the Simulator Review Committee and determined not
.'to detract from training.
The battery powered emergency lights in the control room are not installed on the simulator. These lights illuminated the passageway behind the horseshoe upon-loss of power. The Simulator Review Committee evaluated the need for these lights on the simulator and determined their omission does not detract i
'from training, i 7 1.2.4.3 Aural Simulation 1g.):
1.2.4.3.1 Annunciators and Alarms ,
-Since the same. annunciators and alarm tone generators are used on the
' simulator as in the control room, the sound heard by the operators is the same.
1.2.4.3.2 Steam Noise Y, } -The Seabrock simulator has a. sound generator for steam noise. This is ! -activated.by any malfunction resulting in a steam leak. 'The speaker is ' positioned such that the sound originates from the proper direction. , ib 4
i 1.2.4.3.3 Rod Step Counters
) Mechanical rod step counters are used in both the control room and the
.s imu la tor.~ -
1.2.4.3.4 Ventilation' Noise The sound resulting from the control room ventilation system differs from that of the simulator room. The difference has been evaluated by the Simulator Review Committee and has been determined not to detract from training. 1.3 Instructor Interface The instructor station complies with the requirements of ANS 3.5 with no exception. The following sections that follow describe additional and special features that are available. 1.3.1 Initial Conditions (T The instructor station has the capability of storing fifty initial condition (IC) sets, twenty of'which are standard 1Cs available for instruction. Appendix A contains:a detailed listing of the twenty standard ICs. - 1.3.2 Malfunctions Up to sixteen simultaneous, nonconflicting malfunctions may.be inserted at one time. Sequential malfunctions may be inserted'also. 'The simulator has
. sufficient memorly and processor time to implement new malfunctions identified for_ training through the Training-Systen Development process. Malfunctions areaddressedin<theirrespectiveabstracts(seesection3.4, Malfunction:
Tests)._ i All FSAR malfunctions,.except those listed below, are simulated,
- a. Uncontrolled rod cluster control assembly _ bank withdrawal from a
.subcriticalorlowpowerstartup_cond_ition(FSARsection15.0.1.2[k]).
( s 5
.,.,-y
This malfunction will be implemented on the simulator (SCR #90-104) no later than 12/31/91.
") b. Control rod misalignment - dropped full length assembly bank (FSAR section IS.0.1.2[m]). The Simulator Review Committee determined that this FSAR F
malfunction was not required for training because the operator response
-could be achieved through other malfunctions.
- c. Control rod misalignment - single rod cluster control assembly withdrawal atfullpower(FSARsection15.0.1.3[c]). The Simulator Review Committee has determined this FSAR transient not appropriate for training,
- d. Inadvertent loading and operation of a fuel assembly in an improper position (FSAR section 15.0.1.3(d)). The SRC has determined this malfunction not appropriate for simulator training.
- e. Waste gas system failure (FSAR section 15.0.1.3[f]). This malfunction will be implemented (SCR #90-107) no later than 12/30/91.
- f. Liquidcontainingtankfailure(FSARsection15.0.1.3[h]). An SCR
(!90-185) has been submitted for implementation of an RWST failure by December 31, 1991.
- g. Spent fuel cask drop accidents (FSAR section 15.0.1.3(i]). The SRC has determined this malfunction not appropriate for simulator training.
O h. Reactor coolant pump shaft break (FSAR section 15.0.1.4[d]). An SCR (#90-186) has been submitted to implement this malfunction by December 31, 1991,
- i. Fuelhandlingaccidents(FSARsection15.0.1.4[g]). The SRC has determined this malfunction is not appropriate for simulator training.
- j. Startup of an inactive reactor coolant pump at power (FSAR section 15.0.1.2[h]). There is no malfunction which causes an inactive reactor coolant pump to start. The condition can be simulated; however, it is_not allowed by procedure and therefore considered not appropriate for training,
- k. Inadvertent operation of the emergency core cooling system during power operation (FSARsection15.0.1.2[p]). An SCR (SCR #90-187) has been submitted to correct this problem by no later than 12/31/91.
The following malfunctions are required per ANSI /ANS section 3.1.2 and are not presently simulated, o V 6
.- ~.- _ . _. . . . . . . _ _ _ .- . _ _ _ _ _ _ .-
3 ilt Contr61 rod failure to include uncoupled rudsEand: drifting rods. These i
' ~
malfunctionswillbeimplementedl(SCR#90-101)no..laterthan-12/30/91~. s,- q im.uGenerator; trip. This malfunctionLwill'be implemented (SCR #90-102)Lno 1
~
- later than 12/30/91 -
)
'T n. Main fded line break outside'the containment; structure. ;This malfunction
' - willbeimplemented:(SCR490-103) no later than 12/30/91.
[ 1.3.3 . Controls Provided for-Items Outside of the_ Control Room lThe adeq'uacy of. controls provided for . items outside the control room (remote
. functions) was' evaluated via-the simulator performance testing program. The abili_ty to perform each step'of control _ led copies'of'the; plant procedures was verif,ied. Remots functions that'were required by_ procedure but not simuleted. I
~
were. evaluated by the SimulatorfReview Committee _for_ training impact, i Specifi.c details are; included within-the'applicablestest' abstract. t L1.3.4-: Additional'Special; Instructor / Training Features Available
.1.3.4.1L Simulator Operating Limits' !
- [\
4
/ . _. . . . .
i ;When!a simulatorJoperating' limit ~is reached, an audible and visual alarm is activated at1the instructor station. The simulator continues to run. The' -
' ; instructor can determine which operating limit has been exceeded by selecting; 'I the operating limit tableau at the4 instructor station. The instructor?
; determines whether or not the: operating limit exceeded will-affect'the scenario.in' progress and thereforeidetermines whether,or not to terminate the 2 O
simulation. i0perating limits-have been. implemented and-tested for'RCS
' ~
pressure' hicjh/-low, core exitLtemperature high/ low, containment pressure 'l[ h.igh/ low, andisubcooling marginihigh/ low. I
. l1.3.452 2 Input / Output _ (I/0): Override :
=Theiinstructor station has;the_ capability to override the status:of all
-meters,l switches andrindicator lights on the_ simulator. Overriding the status c'f meters and licjhts:provides only an erroneous indication at the control-board,-the mathematical;modeling is unaffected. t0verriding the position of a a ,
4 +
-.eg i ,o,,e va ,qr--..m .,=w.4 w a
+ r -ww.- rm-
switch _causes the system model=to respond as if the switch had actually been repositioned. Only a limited subset of the available I/O overrides is used-
.,~
j J for1 simulator training. The correct operation of these overrides is verified Eby;the instructor during preparation of.the simulator scenario. 1.4~ Operating Procedures for Reference Plant Simulator training is conducted using controlled copies of Seabrook Station procedures. All controlled copies are maintained up to date in accord with station procedures. Performance testing of the simulator is conducted using < controlled copies of these procedures. All steps that can not be performed 4 are identified and evaluated for simulator training impact by the Simulator g_ Review Committee. LSpecific details of discrepancies and exceptions are described in the individua1Ltest abstracts. sn t v -
-b 8
2.0 SIHilLATOR DESIGN DATA '^) A database listing simulator design data is maintained on the Simulator
'~'
infermation Management System (SIMS). All simulator change requests identify the reference data used as a basis for the change. Upon completion of a simulator change request,-the simulator design data base is updated to include the new references. (~N
=,s' U 9
l 1. 3.0 SIMULATOR TESTS 3.1 Computer Real Time Test [] The Computer Real Time Test (NT-3735) was performed in October 22, 1990. The test demonstrated that the simulator was running in real time under both steady state and severe transient conditions. The test also measured the available spare time. Measured spare time was less than the 15% guideline used to signal the need for additional computer power. An additional node has been procured and will be placed in operation by June 30, 1991. 3.2 Steady State and Normal Operations Tests 3.2.1 Steady State Stability A Steady State Value Stability Test was performed on July 10, 1990. This is an ANSI /ANS-3.5 requirement of Section 4.1(2) Princip!e mass and energy,
... full power operation with the reference plant control system configuration shall be stable and not vary more than 1 2% of the initial values over a 60 0 =i""te Perioe"-
The simulatt,. as reset to a 100% power IC, steady state. An HPCS " snapshot" of all computer points was then taken. Simulation was run with no operator intervention for 63 minutes, and another MPCS snapshot was taken. An MPCS data comparison program, written specifically for this test, was used to make the before and after comparison.- A computer generated printout of all MPCS points, iticir valu2s, and value deviation was made. Result, of the test showed that principle mass and energy balance parameters did not drift. No deficiencies were noted and no exceptions are taken. 3.2.2 Steady State Accuracy On September 29, 1990 a test was conducted to determine simulator accuracies as related to full power and interim power reference plant values. Plant data was collected at 30% and 47% and 100% power. The simulator was configured to match the plant as closely as possible at these three power levels, and n U 10 I
I
~
t 1 simulator data was then collected. A comparison was made between simulator and plant values. -The percent. deviation'was calculated using the following h) equation: (SimulatorValue-PlantValue)*100.- Plant Value Meter'. inaccuracies were not factored in. The results of the test are included in Table 3.2-1. Tables 3.2-1A and 3.2-1B list exceptions taken to ANSI /ANS-3.5-Section 4.1, with justification. All exceptions taken were evaluated by the-Simulator Review Committee and determined to not detract from training. Tables 3.2-2A and 3.2-28 list all discrepancies in which exceptions are not. being taken. These discrepancies will be corrected by no later than 12/31/91. AEsecondary heat balance was perforned at each power. level and no
' discrepancies-were noted. This steady state accuracy test was the first to use' update design data from the operating plant.
I v r
-)
( 11
U STEADY STATE
-COMPARIS0N RESULTS ,
TABLE 1A: Exceptior.s(CriticalParameters) U Power Acceptance
' Level- _ Parameters Deviation -Criteria Justification Pressurizer Steam 100% Temperature 2.6% 1 2.0% 2 100% S/G "0" Wide Range Flow 2.5% 1 2.0% 3 100% S/G "A" Feedwater Flow 2.5% i 2.0% 3 100%. S/G "A" Steam Flow 2,4%- 1 2.0% 3 100% S/G "C" Feedwater Flow 2.8% 1 2.0% 3 100% S/G "C" Steam Flow 3.1% 12.0'4 3 m
100% SR DET 1 Log Pwr 100%+ 1 2.0% 4 {} 100% SR DET 2 Low Pwr 100%+ 1 2.0% 4 47% S/G "0" Feedwater Flow 11.1% 1 2.0% 3 47% S/G "0" Steam Flow 12.9'4 1 2.0% , 3 47% S/G "D" Wide Range Level 7.6% 1 2. 0%. 3-47%' S/G "A" Feedwater. Flow 7.3% i 2.0% 3 47% S/G "A" Steam Flow 9.1% 1 2.0% -3 47%- S/G "A" Wide Range Level 5. 0's 1 2.0% 3 4'% S/G "B" Feedwater Flow 9.7% i 2.0% 3
, 1 of 8
/ 12 V
. - -._ . = .
STEADY STATE COMPARISON RESULTS TABLE 1A: Exceptions (CriticalParameters)(Cont.)
' )?
-. ' Power- _
Acceptance _ Level Parameters Deviation Criteria Justification 47% S/G "B" Steam Flow- 11.0% i 2.0% 3 47% S/G "B" Wide Range Level 7.0% 1 2.0% 3 47% S/G "C" Feedwater Flow 8.4% 1 2.0% 3 47% S/G "C" Steam Flow - 10.4% i 2.0% 3 47% S/G "C" Wide Range Level 5.4% 1 2.0% 3 47% Power Range Channel 1- 3.4% 1 2.0% -3 47% Power Range Channel 2 2.5% 1 2.0% 3 ( )' 47% SR DET 1 Log Pwr 100%+ 1'2.0% 4-47% SR DET 2 Log Pwr- 100%+- 1 2,0% 4 47%' Pressurizer Level 5.6% 1 2.0% -3 30% S/G "D" Feedwater_ Flow 5.2% 1 2.0% 3 30%- $/G "D" Wide Range Level 6.2% 1 2.0% 3 k 30% S/G "A" Steam Flow- 4.3% i 2.0% 3
. 30% S/G "A" Wide Range Level 4.0% 1 2.0% 3 30% S/G "B" Feedwater Flow 2.3% 1 2.0% 3 2 of 8 l
(a. 13 l l' YI A e
STEADY STATE COMPARISON RESULTS TABLE 1A - Exceptions (Critical Parameters) (Cont.)
?%4 Power .
Acceptance Level Parameters Deviation Criteria Justification , + 30% S/G "B" Wide Range Level 5.0% 1.2.0% 3 30% S/G 'C" Steum Flow 3.2% 1 2.0ft 3
. . . 30's S/G "C" Wide Range level 4.5% 1 2.0% 3 p .
30% SR DET 1 Log Pwr. 100fs+ 1 2.0% 4-30% SR DET 2 Log Pwr 100b 1 2.0% 4 , 30%- Pressurizer Level .9.2% 1 2.0% 3 (..
~
h i i 3 of 8 14 L L
STEADY STATE COMPARIS0N RESULTS , TABLE IB: . Exceptions (NonCriticalParameters) ( (3._) Power Acceptance i' Level Parameters Daviation Criteria Justification Pressurizer Relief Tank 100% Pressure 53.2% 1 10% 3 100% Letdown flow 11.3% 1 10% 3 100% VCT Pressure 15.9% 1 10% 3 100% Charging Pump Flow 11.0% 1 10% 3 Condenser "C" Hotwell 100% '.e v e 1 117.1% 1 10% 3 47% PCCW Head Tank "A" Level 11.1% 1 10% 3 47% PCCW Loop "B" Flow 19.3% i 10% 3 7 Pressurizer Relief Tank ( 47% Pressure 52.4% 1 10% 3 47% Letdown Flow 12.2% 1 10% 3 47% RCP "A" Seal Water Flow. 15.5% 1 10% 3 47% RCP "B" Seal Water Flow 21.6% 1 10% 3 47% . RCP "C" Seal Water Flow 24.5% 1 10% 3 4 */% : RCP "0" Seal Water flow 14.5% 1 10% 3 47%' Charging Pump Flow 24.0% 1 10% 3 Condensate Storage Tank 47% Leve1 12.1% 1 10% 3 4 of 8 (3 Q- 15
<-aw .g-- - - - , .s, y -
STEADY STATE COMPARISON RESULTS
,c3 TABLE IBt Exceptions (Non.CriticalParameters)(Cont.) l,
; )
'/ -
Power Acceptance
- Level Parameters Deviation Criteria Justification 47% Generator H2 Pressure 13.2% 1 10% 3 Condenser "C" Hotwell 47% Level 111.14 1 10% 3 L
30% PCCW Head Tank "A" Level 16.7% 1 10% 3 30%. PCCW Head Tank "B" Level 10.3% 1 10% 3 30% PCCW Loop "B" Flow 19.4% 1 10% 3 Pressurizer Relief Tank 30% Pressure 47.6% 1 10% 3 30% Letdown Flow 10.8% 1 10% 3
,ry -
V 30% VCT Pressure 20.3% 1 109 3 l l 30% RCP "B" Seal Water Flow 11.9% 1 10% 3 [ 30% RCP "C" Seal Water Flow 11.9% 1 10% 3 30% RCP "3" Seal Water Flow 13.1%~ 1 10% 3 30% Generator H2 Pressure 14.2% 1 10% Condenser "C" Hotwell > 50% Level 108.9% 1 10% 3 5 of 8 o b: 16 l
t 51EADY STATE COMPARISON RESULTS , g TABLE 2A: Discrepancies (CriticalParameters) { U Power Acceptance Level Parameters Deviation. Criteria f Comment 100% S/G "D" Pressure 2.6% 1 2.0% 1 100% S/G "A Pressure 2.6% 1 2.0% 1 100% S/G *B" Pressure 2.3% 1 2.0% 1 100% S/G "C" Pressure 2.3% 1 2.0% 1 100% IR DET 1 Log Pwr 40.9% 2 2.0% 1 l 100% IR DET-2 Log Pwr 37.6% 1 2.0%- 1 47% IR DET 1 Log Pwr 48.2% i 2.0% 1 47%- IR DET 2 Log Pwr 44.3% 1 2.0% 1 47% Core Thermal Power 6.6% 1 2.0% 1 30% IR DET 2 Log Pwr 38.8% i 2.0% 1 30%- IR DET 1 Log Pwr 43.2% 1 2.0% 1 30%- Core Thermal Power 5.3% 1 2.0% 1 ( L 6 of 8 17
STEADY STATE COMPARISON RESULTS TABLE 2B: Discrepancies (NonCriticalParameters) Acceptance ()
\/
Power Level Parameters Deviation Criteria Conment 100% Containment Delta-P 233.9% 1 10% 1 , 1004 S/G "A* Blowdown Flow 70.9% 1 10% 1 100% S/G "B" Blowdown flow 71.2% 1 10% 1 100% S/G "C" Blowdown Flow 71.1% 1 10% 1 100% S/G "0" Blowdown Flow. 70.7% 1 10% 1 100% Hydraulic fluid Pressure 99.1% 1 10% 1 : SW Pump "A" & "C" Disch. 100% Hdr Pressure 30.6% 1 10% 1 Turbine Generator ! f3 100%. Vibration Bearing #1 21.0% 1 10% 1 SW Pump "B" &:"0" Disch. 100% Hdr Pressure 33.8% i 10% 1 , 47% S/G "A" Blowdown Flow . 69.7% i 10% 1 47%- S/G "C" Blowdown Flow 68.2% 1 10% 1 1 47% S/G-."C" Blowdown Flow 70.7% i 10% 1 47% S/G "0" Blowdown Flow. 69.9% 1 10% 1 47% Hydraulic Fluid Pressure 99.0% i 104 1 1 SW Pump "A" & "C" Disch. 47% Hdr Pressure 27.2% 1 10% 1. 7 of 8 18 t fs L d- . a
+eN.' p-^-p,r-wwyy, y-g g yp, - - = + y e-w y-- A'Q, -,-en - . ,megw., y as,.p, W w m.%e.vqw-8 *-+g'-f
STEADY STATE COMPARISON RESVLTS Discrepancies (Non-CriticalParameters)(Cont.)
. TABLE 28: ,
N Power Acceptance Level Parameters Deviation Criteria Comment SW Pump "B" & "D" Disch. 47% Hdr Pressure 32.8% 1 104 1
-4 Containment 1A Hdr "B" 47% Pressure 14.0% 1 10% 1 30% Containment Delta P 279.0% 1 10% 1 30% S/G "A" Blowdown Flow 75.2% 1 10% 1 30% S/G "B" Blowdown Flow 74.4% 1 10% 1 30% S/G "C" Blowdown Flow 76.3% 1 10% 1 30%. S/G "0" Blowdown Flow 74.2% 1 10% 1 n
V, 30% Hydraulic Fluid Pressure 99.1% 1 10% 1 SW Pump "A" & "D" Disch. 30% Hdr Pressure 33.6% 1'10%- 1 SW Pump "C" & "D" Disch. 30% Hdr Pressure 39.6% 1 10% 1 Containment IA Hdr "A" 30% Pressure 10.1% 1 10%. 1 Containment IA Hdr "B" 30% Pressure 11.8% 1 10% 1 Turbine Generator 30% Vibration Bearing #1 24.3% 1 10% 1 8 of 8 19
l COMMENT AND JUSTIFICAf10N CODE
- 1. An SCR to correct the deviation has been submitted. The SCR will be completed by December 31, 1991.
- 2. Plant data appears to be inconsistent and/or simulator value is correct based on other parameter values and the laws of physics.
- 3. Deviation is attributable to differences in the plant and simulator configuration.
- 4. Deviation is attributable to dividing by zero.
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3.2.3 Normal Operations Tests l 3.2.3.1 .Startup and Shutdown: Cold Shutdown to Hot Stantiby The Cold Shutdown to Hot Standby section of the Startup and Shutdown Test was conducted on April 11 and 12, 1989. This is an ANSI /ANS-3.5 Requirement of Section3.1.1(1),Plantstartup-coldshutdowntohotstandby,and3.1.1(5), j Operations at hot standby, The test began from cold shutdown conditions with l the RCS solid and temperature at 145'F. During the test a bubble was formed in the pressurizer. RCS temperature was raised to 558'r (T,,,) and pressure ; to 2250 psig. The following operating procedures were used: 0S1000.01, Heatup from Cold Shutdown to Hot Standby 051001.06, PressurizerBubbleFormation(Section6.1) 051002.02, OperationofLetdownChargingandSealinjection(Sections 6.3 and6.8) 0S1002.04, OperationoftheLetdownDegassifier(Section6.2) q V 0S1023.68,ContainmentPurgeSysterOperation(Section6.2) 0$1023.67, Containment Ventilation System Operation (Sections 6.2, 6.6 and 6.8) 051013.06, ResidualHeatRemovalTrain"B" Shutdown (Section6.3) ON1035.10, Main feed Pump Standby and Startup Operation (Sections 6.2 and 6.3) 0S1001.05, Reactor Coolant Pump Operation'(Section 6.1 and 6.2) 151015.19, ContainmentCloseoutProcedure(Section6.2) 051015.18, Setting Containment Integrity for Mode IV Entry (Section 6.2) OX1456.02, ECCSMonthlyValveVerification(Section8.2) 051006.04, OperationofContainmentSpraySystem(Section6.2) ON1033.02, TurbineSteamSealSystemOperation(Section6.2) ON1034.03, Condensate System Operation (Sections 6.1, 6.3 and 6.4) 051030.01~, Main Steam System Operation (Section 6.1 and 6.2) ON1033.01, Operation of Mechanical vacuum Pumps (Sections 6.1, 6.2, and ! 6.3)-
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ON1038.01, Circulating Water Pump Startup (Section 6.1 and 6.3) l 051036.01, Aligning The EFW System for Automatic initiation (Section 6.1) I 0x1430.02, Main Steam 1 solation Signal 18 Month Stroke Test (Section 8.0) OX1430.02, Main Steam Isolation Valve Quarterly Test (Section 8.1 and 8.2) j 051030.01, Main Steam System Operation (Section 6.1 and 6.2) OX1436.02, Turbine Driven EFW Pump Monthly, Quarterly and 18C Month Surveillance (Section8.1,8.2and8.3) l 051021'.01, SteamGeneratorBlowdownSystemOperation(Section6.2) ) I ON1035.02, Startup Feed Pump Operation (Section 6.1)
"A" point trends for related parameters were saved, and ten-minute plots were J made for the fo'ilowing: loops 1, 2, 3, and 4 pressurizer pressure; pressurizer level; and SG level and pressure. One deficiency was noted.
Letdown pressure and flow reduced unrealistically during the heatup. While in auto control, PCV-131 cycled closed causing a loss of letdown flow. This problem had been previously noted and will be corrected under SCR #89-022 by , no later than 6/30/91. All other simulator responses were in accord with Seabrook Station Startup Test Data. No exceptions are taken. 3.2.3.2 Startup and Shutdown: Hot Standby to Minimum Load The Hot Standby to Minimum Load section of the Startup and Shutdown Test was conducted on April 16, 1990.- This is an ANSI /ANS-3.5 requirement of Sections 3.1.1(2)Nuclearstartupfromhotstandbytoratedpower,and3.1.1(3) , Turbine startup and. generator synchronization. The initial conditions were plant in hot standby with the reactor subcritical. The final conditions of the test were the turbine generator breaker closed supplying approximately 170 MW, steam dumps open, and feedwater in manual. The following reference procedures were used: 051000.07, Approach to Criticality ON1031.02, Starting and Phasing the Turbine Generator (Sections 6.2 and 6.3) ON1035.10, Main feed Pump Standby and Startup Operation (Sections 6.4 and-6.6) 051035.02, Startup Feed Pump Operution (Sections 6.2 and 6.3) O 22
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ON1032.01, ExtractionSteamSystemOperation(Section6.2) 051071.01, Steam Generator Blowdown System Operation (Section 6.3) ( . ON1035.06, Aux Steam to 26A and 26B Heaters (Section 6.3) 051000.02, Plant Startup from Hot Standby to Minimum Load One deficiency was noted. Procedure 051000.02, step 7.1.6, requires the operator to open the bypass valves around the moisture separator reheater steam supply valves MS V249, V250, V251 and V252. These bypass valves are not sittulated. AnSCR(#90-097)hasbeensubmittedtoaddthisenhancementbyno later than 12/31/91. All other simulator responses were in accord with the Seabrook Station Startup lest results. No exceptions are taken. 3.2.3.3 Startup and Shutdown: MinimumLoadtoFullPower(Powerincrease) The Minimum Load to full Power section of the Startup and Shutdown Test was conducted on April 6, 1989. This is an ANSl/ANS 3.5 requirement of Section 3.1.1(c),t.oadchanges. The initial conditions for the test were the turbine generator on line supplying approximately 15% rated output, the steam dumps open, and feedwater control in manual. The final conditions of the test were the plant operating at-100% power and all control systems in automatic. Ten-minute plots of major parameters were made. Simulator response was in accord with Seabrook Station Startup Test data. No deficiencies were noted and no exceptions are taken. The following reference procedures were used: 051000.05, Power increase 051008.01, Chemical and Volume Control System Makeup Operation (SectioN 6.5'and6.6) ON1034.03, Condensate System Operation (Section 6.2) ON1040.04, Operation of the Heater Orain Pumps (Sections 6.1 and 6.2) ON1038.01, CirculatingWaterSystemPumpStartup(Section6.4) ON1035.10, MainFeedPumpStandbyandStartupOperation(Sections 6.5and
-6.7).
n
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3.2.3.4 Startup and Shutdown: Full Power to Minimum Load O The Full Power to Minimum Load section of the Startup and Shutdown Test was conducted on April 5, 1989. This is an ANSI /ANS-3.5 requirement of Section 3.1.1(8), Plant shutdown from rated power to hot standby and cooldown to cold shutdown conditions. The initial conditions for the test were 100% reactor power with major control systems in automatic. Reactor power was reduced to approximately 19%. The following reference procedures were used: 051000.06, Power Decrease 051008.01, Chemical and Volume Control System Mateup Operations (Section , 6.3) ON1038.02, Operation of the Heater Drain Pumos (Section 6.4) ON1034.03, Condensate System Operation (Section 6.5) ON)035.11,MainfeedPumpReturntoStandbyandShutdown(Section-6.2) MPCS "A" point trends for related parameters and primarily calorimetric data were saved. Ten-minute plots were made for the following generator power; ( core thermal power; loops 1, 2, 3, and 4 T,,,; pressurizer pressure and level; and SG level and pressure. Simulator response was in accord with the Seabrook Station Startup Test Data. No deficiencies were noted and no exceptions are taken. 3.2.3.5 Startup and Shutdown: Minimum Load to Hot Standby The Minimum Load to Hot Standby section of-the Startup and Shutdown Test was conducted on April 5, 1989. This is an ANSI /ANS-3.5 requirement of Section 3.1.1(8),Plantshutdownfromratedpowertohotstandbyandcooldowntocold shutdown conditions. During tho test, the plant was taken from 19% reactor a power to hot standby conditions (reactor subcritical and T,y, at 557'F). The following reference procedures were used: 051000.03, Plant Shutdown from Minimum Load to Hot Standby 051035.02, StartupFeedPumpOperation(Section6.1) ON1031.03, Turbine Generator Shutdown O 24
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l ON1035.11, Main Feed Pump Return to Standby and Shutdown (Section 6.2) ON1035.06, Auxiliary Steam to 26A and 26B Heaters (Section 6.1) {} MPCS "A" point trends and primary plant calorimetric data were saved. Ten-minute plots were made for the followings generator power; core thermal , power; pressurizer pressure and level; loops 1, 2, 3, and 4 T,,,; and SG level l and pressure. Two deficiencies were noted. Feedwater heater outlet temperatures are excessively low at low power levels. This problem had been previously noted and is to be corrected under SCR #88-160, by no later than 12/31/91. Also, startup feed pump discharge pressures are at times erroneous. This problem also was previously noted and will be corrected under SCR
#88-089 by no later than 12/31/91. All other simulator responses were in accord with the Seabrook Station Startup Test Oata, No exceptions are taken.
3.2.3.6 Startup and Shutdown: Plant Cooldown The Plant Cooldown section of the Startup and Shutdown Test was conducted on 4/8/89. This is an ANSI /ANS-3.5 requirement of Section 3.1.1(8), Plant shutdown from rated power to hot standby and cooldown to cold shutdown conditions. During the test, the plant was taken from hot standby to an RCS temperature of <350'F-and a pressure of <400 psig, with the RHR system in service. The following reference procedures were used: 051000.04, Plant Cooldown OX1456.08, 51/ Charging Pump Monthly Inoperability Check 051013.04, RHR Train B Startup and Operation (Sections 6.2 and 6.4) OX1408.05, Boron Thermal Regeneration System inoperability Surveillance 0X1405.05, Monthly Accumulator Valve and Breaker Checks 051008.01, CVCSMakeupOperations(Section6.6) MPCS "A" point trends were saved, and ten-minute plots were made for the following: pressurizer prer:sure and level; loops 1, 2, 3, and 4 cold and hot leg temperatures; and SG 1evel and pressure. One deficiency was noted. RHR l flow would not rise above 3000 GPM with the temperature control valves and l- heat exchanger bypasses fully open. This resulted in an insufficient cooldown rate. This problem has been corrected (SCR #89-028, closed on 8/6/90). All 3 ~ N(v 25 4 r - ,-- ,. ..e--,
other simulator responses were in accord with the Seabrook Station Startup test results. No exceptions are taken. O 3.2.3.7 Shutdown with Less than full Reactor Coolant flow The Shutdown with less than full Reactor Coolant flow test was conducted on , March 27, 1989. This is an ANSI /ANS 3.5 requirement of section 3.1.1(7), , Startup, shutdown and power operations with less than full-reactor coolant I flow. Seabrook Station operational procedures do not allow startup or s s'ained power operations with less than full reactor coolant flow; the/efore, only the shutdown sith less than full reactor coolant flow was ttsted. This is an exception to the ANS!/ANS 3.5 requirement. The Simulator R.tview Committee has evaluated this exception and determined that it does not detract from training. From 45% reactor power, reactor coolant pump "A" was tripped. Once stabilized, the plant was shutdown. The test took approximately-1.5 hours. The MPCS alarm summary was saved. Simulator response was in accord with the Millstone 3 Startup Test Report and the Seabrook Station FSAR. No deficiencies were noted and no exceptions are taken. A 3.2.3.8 Reactor Trip and Recovery The Reactor Trip and Recovery Test was conducted on August 3, 1988, This is an ANS!/ANS 3 requirement of section 3.1.1.(4), Reactor trip followed by ecov6cy to rated power. The initial conditions for the test were 100% reactor power, BOL. A manual reactor trip was inserted and the plant was then brought-back to 100% power. Operating startup procedures were used as a guideline to return the plant to full power. Ten-minute plots were made for the followings loop 1 Tm; loop 1 cold and hot leg temperatures; SG "A" feed flow, steam flow, level and pressure; pressurizer liquid and vapor temperatures; pressurizer level and pressure; and instrument range power level. The reactor trip response was in accord with the Seabrook Station FSAR and the Millstone 3-Startup Test Reports. No discrepancies were noted. There were no discrepancies unique to the trip recovery noted when returning the plant to full power. No exceptions are taken, 0 26 l l
3.2.3.9 Core Performance Test O The Core Performance Test was performed in April 1990. Reference data used throughout the testing included the preliminary results of startup tests ST. 18, Isothermal Temperature Coefficient; ST-20, Control Rod Worth Measurement; and WCAP-10982, Westinghouse Nuclear Design Report. The tests identified several deficiencies. The moderator temperature coefficient is more pr-'tive than that measured in the plant. The doppler only cefect and the total power defect are lower than predicted in the Nuclear Design Report. Differential boron worth at beginning of life conditions is larger than predicted. Total control bank worth is 4.3% higher than observed while shutdown bank worth is 3.8% higher than observed. Shutdown margins between the plant and the simulator differ by 7.7%. Each of these deficiencies will be corrected when the core model is replaced. The replacement is scheduled to be completed by the third quarter of 1991. O O 27
3.2.3.10 Operator Conducted Surveillance Testing i SURVEILLANCE #0N1431.08 MAIN TURBINE OVERSPEED TEST The
- Main Turbine Overspeed Test" was conducted on June 4, 1990. ,
The objective of the test is to verify the proper operation of the Turbine Overspeed Trip System. One acceptance criterion was not met. The turbine should trip between 1980 and 1998 RPM, however it tripped at 2192 RPM. Thisproblemhasbeencorrected(SCR#90043, closed 8/6/90). No other deficiencies were noted and no exceptions are taken. ! O - 28 L
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i SURVEILLANCE OX1405.07 q S1 QUARTERLY AND 18 MONTH PUMP FLOW VALVE TEST V The " Safety injection Quarterly and 18 Month Pump flow and Valve Test" was 1 conducted on May 9, 1990. This is an ANSI /ANS-3.5 requirement of section 3.1.1, Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment. The initial conditions for the test were 100% reactor power, BOL. In this test, pump operability is demonstrated by running each pump and verifying pump parameters are within specifications. Also certain power operated valves are tested by . verifying acceptable stroke times and accurate position indication and status lights. Local pump parameters and local valve position indication are not available. However, these verifications can be simulated by the instructor. RWST temperature is not available in the control room and must be obtained locally (as in the plant). A plant design change has been planned to add an 'A' point to the main plant computer which will allow the operators to monitor RWST temperature from the control room. The-'A' point will also be added to the simulator MPCS under SCR #90-075 by no later than December 31, 1990. A simulator deficiency was noted. O Although flow rate is verified locally, a data pool variable (SIFV93) can be used to check recirc flow. A reading of 6.61 lbm/second (approximately 49 gpm) was taken during the test. The acceptance criterion ranges from 40.5 gpm to 43.2 gpm. This problem has been corrected (SCR #90-057, completed 8/6/90). All other acceptance criteria were met and no exceptions are taken. O 29
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SURVEILLANCE OX1412.01 PCCW TRAIN 'A' QUARTERLY OPERABILITY TEST AND 18 MONTH VALVE POSITION INDICATION I v
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The " Primary Component Cooling Water Train 'A' Operability Test and 18 Month Valve Position Indication" was conducted on May 10, 1990. This is an ANSI /ANS-3.S requirement of section 3.1.1, Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment or Systems. The initial conditions for the test were 100% reactor power with PCCW Loop 'A' in service (pump 11A running). In this test, pump operability is confirmed by running the pump and verifying pump parameters are within acceptable range. Power operated valves are demonstrated to meet the required stroke times and position indication. Local pump parameters and local valve indications are not available. This is not considered a deficiency because the control room operator receives verbal verification of these indications (which can be simulated by the instructor). One acceptance criterion could not be met due to a simulator deficiency. The acceptance criterion requires a combined total flow rate of greater than 11,000 gpm. The MPCS graphic display of individual PCCW header flow rates is inaccurate due tc a ur.its-conversion problem, d Therefore, the total indicated flow is only 7500 gpm. The display problem was corrected under SCR #90-045 on August 6, 1990. No other deficiencies were noted and all other acceptance criteria were met. No exceptions are taken, n
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_ _ - _ _ _ _ _ _ _ _ _ -._ __-_ _ _ _ . _ ___-__m.__-__
SURVEILLANCE 0x1412.02 n PCCW TRAIN 'B' QUARTERLY OPERABILITY TEST () AND 18 MONTH VALVE POSITION INDICATION The " Primary Component Cooling Water Train 'B' Operability Test and 18 Month Valve Position Indication" was conducted on May 10, 1990. This is an ANSI /ANS-3.5 requirement of section 3.1.1, Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment or Systems. The initial conditions for the test were 100% reactor power withPCCWLoop'B' inservice (pump 118 running), in this test, pump operability is confirmed by running the pump and verifying pump parameters are within acceptable range. Power operated valves are demonstrated to meet the required stroke times and position indication. Local pump parameters and local valve indications are not available. This is not considered a deficiency because the control room operator receives verbal-verificationoftheseindications(whichcanbesimulatedbythe instructor). One acceptance criterion could not be met due to a simulator deficiency. The acceptance criterion ~ requires a combined total flow rate of greater than 11,000 gpm. The MPCS graphic display of-individual PCCW. h- header flow rates is inaccurate due to a units conversion problem. Therefore,.the total indicated flow is only 7500 gpm. The display problem , was corrected under SCR #90-045 on August 6, 1990. No other deficiencies were noted and all'other acceptance criteria were met. No exceptions are taken. 31 '
l A SURVE!LLANCE #0X1413.01 V' RHR QUARTERLY FLOW AND VALVE STROKE TEST AND 18 MONTH VALVE STROKE OBSERVATION i l The "ResiJ ual Heat Removal Quarterly Flow and Vrive Stroke Test, and 18 Month valve Stroke Observation" surveillance was conducted on May 30, ; 1990. This is an ANSI /ANS-3.5 requirement of section 3.1.1, normal plant ) evolutions (10),OperatorConductedSurveillanceTestingonSafety- ; Related Equipment or Systems. The initial conditions for the test were j mode 5 with RHR Train 'A' in service and RHR Train 'B' in standby, in this test, the operability of residual heat removal Trains 'A' and 'B' are verified. Section 8.1 and 8.2 confirm pump operability by running each pump on miniflow using the RWST as a suction head tank and verifying pump parameters are within acceptable range. Valve cperability is confirmed by stroking each valve, verifying that its maximum stroke time is not exceeded, and verifying position indication. Sections 8.3 and 8.4 verify pump operability by running the pump on recirc with the suction aligned to the RCS. Although the surveillance requirement allows performing sections (] 8.1 and 8.2 in lieu of sections 8.3 and 8.4, all 4 sections were performed for this test.- Local pump parameters and valve indications are not available. This is not a deficiency because the control room operator receives verbal indication of these parameters (which can be simulated by the instructor). The acceptance criteria were met and no exceptions are taken.
).
32
SURVEll1AN;E OX1426.01 7-sj DG 1A OPERABILITY SURVEILLANCE U The " Diesel Generator 1A Operability" surveillance test was performed on May 9, 1990. This is an ANSl/Abb ?.5 requirement of sectien 3.1.1 Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety. Related Equipment or Systems. The initial conditions for the test were , 100% reactor power, with the diesel generator lined up for automatic start as per procedure 051026.07. In this test, operability of the Olesel Generator is confirmed by starting the diesel arid verifying that the generator reaches required voltage and frequency within 10 seconds. The DG is then placed on the bus and loaded. There were no deficiencies noted during the test. Step 8.10 requires a verification of the fuel transfer system. This can be accomplished by step changing the DG fuel tank level from the tank level page on the instructor station. All other local verifications can be simulated by the instructor. The acceptance criteria were met and no exceptions are taken. R . V r O 33
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SVRVElLLANCE 0x1426.05
! DG 10 OPERABILITY SURVEILLANCE The 'Diesei Generator 1B Operability" surveillance test was performed on May 9, 1990.- This is an ANS!/ANS-3.5 requirement of section 3.1.1 Normal Plan Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment or Systems. 1he initial conditions for the test were 100% reactor power, with the diesel generator lined up for automatic start as per procedure 051026.07. In this test, operability of the Diesel Generator is confirmed by starting the diesel and verifying that the generator reaches required voltage and frequency within 10 seconds. The DG is then placed on the bus and loaded. There were no deficiencies noted during the test. Step 8.10 requires a verification of the fuel transfer system. This can be accomplished by step changing the DG fuel tank level from the tank level page on the instructor station. All other local verifications can be simulated by the instructor. The acceptance criteria was net and no exceptions are taken.
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SURVEILLANCE #0X1430.28 k'_)s- MS ISOLATION VALVE QUARTERLY TEST The " Main Steam Isolation Valve Quarterly Test" was performed on May 14, 1990. This is an ANSI /ANS-3.5 requirement of section 3.1.1 Normal Plant Evolutions (10),OperatorConductedSurveillanceTestingonSafety-Related Equipment or Systems. The initial conditions for the test were ; 100% reactor power. The test verified MSIV operability by cycling each valve through 10% of valve travel. The acceptance criteria were met and no deficiencies were noted. No exceptions are taken. L i T ^ 35-
SURVEILLANCE OX1436.04 EFW 18 MONTH AUTO ACTUATION SURVEILLANCE The surveillance " Emergency feedwater 18 Month Automatic Actuation Test" 4 was performed on May 29, 1990. This is an ANSI /ANS-3.5 requirement of 1 section 3.1.1 Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment.or Systems. The initial conditions for the test were as required in the procedure: Mode 3 with all control rods fully inserted. The test verified that the steam admission valves to the EFW pump turbine open, and the pump starts upon receipt of an EFW actuation signal. Some portions of the procedure could not be performed as written. Step 6.3 requires a calibrated recorder to be installed on the safeguards test cabinet and that I&C initiate (and at ; completion, remove)anEFWactuationsignal. The safeguard test cabinets are not simulated; therefore, the EFW signal was generated by reducing feedwater flow to all 4 SG's and allowing the SG 1evel to drop below the low-low level setpoint. Step 8.3 is annther noted deficiency that requires the operator to unlock and close FW-V-65, the EFW pump discharge , valve. This valve is not operated from the MCB and must be operated locally. There is no remote function to operate this valve. AnSCR(#86-252) has been submitted to correct this deficiency during the feedwater system upgrade, to be corrected no later than December 30, 1991 . Also, i some local pump parameters are not available. The safeguard test cabinets and local pump parameters do not need to be simulated because control room operators do not perform any function nor do they train on this equipment, and local indications can be provided by the instructor. The acceptance criteria were met _with starting conditions taken into account. O 36
SURVEILLANCE 0x1456.01 QUARTERLY FLOW AND VALVE STR0rE TEST AND O sont"'t atsott eosi'io" isotcatio" vtairic^'io" The " Charging Pump 'A' and 'B' Quarterly flow and Valve Stroke Test and 18 Month Remote Position Indication Verification" was conducted on May 11, 1990. This is an ANSI /ANS-3.5 requirement of section 3.1.1, Normal Plant Evolutions (10), Operator Conducted Surveillance Testing on Safety-Related Equipment. The initial conditions for the test were 100% reactor power, BOL. This test verifies the operability of centrifugal charging pumps P-2A and P-2B and certain valves in both their normal and emergency flow paths. Pump operability is confirmed by running the pumps on miniflow and verifying pump parameters are within an acceptable range. Power operated valve operability is confirmed by stroking the valves and verifying valve stroke times and position indication. Check valve operability is also checked by stroking the miniflow check valves, partially stroking the charging pump discharge valves, and verifying that recirc flow is within required range. Local pump parameters and valve indications are not available. This is not a deficiency because the O control room oPeretor rece4ves verbei iadicat4oa of these veremeters (which can be simulated by the instructor). Some deficiencies were noted, however, several power operated valves did not meet the stroke time acceptance criteria. This will be corrected under SCR #90-096 by no later than 12/31/90. The procedures require closing valves CS-V210 and CS-V220 (CCP-2A and 'cB, Manual Discharge Valves). There are no remote functions to operate these valves. The remote functions will be added under SCR
#90-114 by no later than 6/30/91. The acceptance criteria were met and no exceptions are taken.
O l 37
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1 3.3 Transient Test g- 3.3.1 Annual Transient Tests k The following transients are annual testing requirements of ANSI /ANS-3.a- l 1985 that address Section 5.4.2, simulator Operability Test, and Section B.2.2, Transient Performance. Specific references are found in the ! respective abstract for each transient. i
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L a ( 38
TRANSIENT #1 REACTOR TRIP t r ()N The " Reactor trip" test was performed on July 31, 1990. This is an annual ; testing requirement addressing Section 5.4.2(3), verify simulator performance against the transient criteria of 4.2 for a benchmark set of ! transients, and Section B.2.2(1), Manual reactor trip. The initial' conditions were 100% reactor power, steady state. Malfunction i
#156 " Reactor Trip" was activated and simulation was allowed to run for ten minutes with no operator action, while data was recorded. The MPCS .
alarm summary and primary calorimetric data were s3ved. Ten-minute plots were made for power range, loop T ,,, loop hot leg and cold leg
' temperatures, pressurizer steam and liquid temperatures, pressurizer pressure and level, SG "A" pressure and level, and SG "A" feedwater and steam flow. The simulator test results met the acceptance criteria of the plant " Unit Trip" test (ST-38) included in the startup test program. <
Specifically, pressurizer safety valves did not lift, steam generator safety valves did not lift, safety injection did not initiate, and the overall RTO response time was less than 6.7 seconds. One discrepancy was h noted when simulator plots.were compared to the plant transient recorder plotsi SG levels took too long to recover on the simulator. This will be ; corrected-under.SCR #90-152 by no later than 12/31/91. Other plotted t i parameters were similar to plant response. No exceptions are taken.
+
TRANSIENT #2 SIMULTANE0US 1 RIP Of ALL FEEDWA1ER PUMPS The " Simultaneous Trip of All feedwater Pumps" test was performed on O December 12. 1990. This is an annual testing requirement addressing Section 5.4.2(3), Verify simulator performance against the transient criteria of 4.2 for a benchmark set of transients, and Section B.2.2(2), Simultaneous trip of all feedwater pumps. From 100% reactor power, steady state, a total loss of feedwater was estabiished by placing the s tor driven emergency feed pump and the startup feed pump in PULL-TO-LOCK, closing the steam supply valve to the steam driven emergency feed pump and simultaneously tripping the main feedwater pumps using Malfunction #14. No operator actinn was taken, and data was collected for ten minutes. The test concluded with the steam generator levels at approximately 15%. T'ne MPCS alarm data and chart recordings of pertinent parameters were saved. Ten minute plots were made for power level; average temperature; pressurizer pressure, level and temperature; steam and feedwater #10ws; hot and cold leg temperatures; and SG pressure and level. The results of this test were reviewed by a panel of experts and they concluded that parameters did trend in the correct O direction and that values were reasonable. No discrepancies were noted and no exceptions taken. v O ao
TRANSIENT #3 CLOSURE Of ALL MSIVs The " Closure of all Main Steam Isolation Valves" test was completed on Q 8/1/90. This is an annual testing requirement addressing Section 5.4.2(3), Verify simulator performance against the transient criteria of 4.2 for a benchmark set of transients, and Section B.2.2 (3), Simultaneous closure of all main steam isolation valves. Malfunction #40, " Simultaneous Closure of all MSlvs," was activated from 100% reactor power, BOL, steady state. Simulation was run and data was collected for ten minutes. No operator action was taken. MPCS and primary calorimetric data were saved. Ten-minute plots were made for power level, loop Tu ,, loop hot and cold leg tereperatures, pressurizer 5 pressure and level, i.ressurizer liquid and vapor temperature, SG pressure and level, feed flow, and steam flow. Apanelofexperts(describedin section 5.5) reviewed the results of this test including a comparison to data of a similar event at Calloway. The panel concluded that parameters did trend in the correct direction and values were reasonable for starting conditions. No discrepancies were noted and no exceptions are taken. O u, O 41
l TRANSIENT #4 I TRIP Of ALL REACTOR COOLANT PUMPS I 1 The simultaneous Trip of all RCPs* was tested on 7/31/90. This is ., annual testing requirement addressing Section 5.4.2(3), verify simulator performance against the transient criteria of 4.2 for a benchmark set of j transients, and Section B.2.2(4), Simultaneous trip of all reactcr coolant pumps. From 100% reactor power, steady state, all four reactor coolant pump breakers were simultaneously racked out. Data was collected for ten minutes and no operator action was taken. The MPC5 alarm summary and primary calorimetric data were saved. Ten minute plots were made for r pressurizer level, temperature and pressure; power level; loop Tm; loop 4 hot and cold leg temperatures; SG level and pressure; steam flow; and _; feed flow. A panel of experts reviewed the results of the test including making a comparison to Plant Startup Test ST-22, " Natural Circulation Test." The panel concluded that parameters did trend in the same , direction and values were reasonable based on the different starting conditions of the tests. No exceptions are taken. O h 42
TRANSIENT #5 TRIP 0F ANY SINGLE REACTOR COOLANT PUMP The " Trip of any Single Reactor. Coolant Pump" test was conducted on ('] 8/1/90. This is an annual testing requirement Section 5.4.2(3), Verify simulator perfornance against the transient criteria of 4.2 for a benchmark set of transients, and Section B.2.2(5), Trip of any single reactor coolart pump. Malfunction #26, " Trip of "A" Reactor Coolant Pump" was activated from 100% reactor pexer, steady state. Simulation was run for twenty minutes with no operator action, and data was collected during the initial ten minutes following malfunction activation. The MPCS-alarm summary and the primary calorimetric data were saved. Ten-minute plots we m me;ie for power level, SG level, SG feed flow, feedwater flow, hot and co;d leg temperatures, aind SG pressure. The results of the test were reviewed by ( panel of experts. The panel concluded that parameters did trend in the correct-direction and that values were reasonable. No deficiencies were noted and no exceptions are taken. O - 4: Oa 43 ~)
TRANSIENT #6 MAIN TURBINE TRIP
- (~'I _The . Main Turbine Trip" test was performed on Decmber 3,1990. This is an annual testin, requirement addressing Section 5.4.2(3), Verify simulator performance against the transient criteria of 4.2 for a benchmark set of transients, and Section B.2.2(6), Main turbine trip (maximum power level which does-not result in immediate reactor trip).
The test was. initiated fror.: 16% reactor power with feed water control in auteratic on the bypass valves (20% reactor power is the lowest power lewi that would result in a reactor trip), control rods in manual, a slightly positive MTC, and steam dumps maintaining a +3 Tue/T,,, deviation. The turbine tv;p wn initiated by activating Malfunction #122, "High Tarbine vibratin " .at 100% severity (approximately 18 mils). No operator action was taken, 4rd data was collected for ten 'ninutes. The duration of the test was apprcximate!v twenty-five minutes and terminated with the steam dumps controlling SG pressure. Primary calorimetric data and the MPCS alarm summary were saved. Ten-minute plots were made of neutron flux; averageftemperature; pressurizer pressure, level and temperature; , 4% steam and feedwater flows,-hot and cold leg temperatures; and steam generator pressure and level. The results of this test were reviewed by a panel of experts. The-panel concluded that the parameters did trend in the correct direction and valves were reasonable. .No discrepancies were noted,-and no exceptions are taken. l g V. 44 L lt ' L 1 1
- i as ! TRANSIENT-17 , MAXIMUM RATE POWER RAMP Ths'* Maximum _ Rate P N er Remp": test'was conducte'd on"8/1/90. .This is an N annual:testingrequirementof-Section.5.4.2(3),,verifysimulator A./' performanceiagainst-the transient criteria of 4.2 for a benchmark set of - transients,fand Section B.2.2(7), Maximum rate power ramp.
!The; initial. conditions were 100t reactor power, BOL. .The turbine load i decrease function was used to rapidly decrease turbine load to Lapproximately 800 MW. -The plant was then allowed to stabilize.- Turbine j
- load lwas then rapidly: increased, using the 34 per minute turbine-load- 7 increase function, to full rated load _~Again, the plant was allowed to l stabilize.- Data was collected for'the first ten minutes'of both the down j
; load and up-load; transients with no additional operator action taken. The ';
.MPCS alarm summary and-primary calorimetric data were saved. Ten-minute plots were made ofLpower level;~ pressurizer level, pressure._and y ,
, . temperature; steam and-feed flows;-hot and cold _ leg. temperatures;fand SG pressure and: level. The'results of.the-test were reviewed'by~atpanel of.
experts. : Included within the review was a comparison-to the plant-startup- 1 test:S.T-35, "Large Load Reduction,"'and ST-34, " Load Swing Test."~ The-panel concluded that parametersidid trend-^ comparatively, and that any' . differences ~in values were due to the differences-between the plant and: , simulator te'st procedures.--All plant acceptance test criteria were met.- No deficiencies were noted and=no exceptions are taken._ . l j t l i 1 i il L {,/
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TRANSIENT #8-RCS RUPTURE WITH LOSS OF OfFSITE POWFR The " Maximum Size Reactor Coolant System Rupture ConUned with Loss of All 1(') - Offsite Power" test was performed on 8/1/90. This is an annual testing i
'~~'
Section'5.4.2(3) verify simulator performance against the transient criteria of 4.2 for a benchmark set of transients, and Section B.2.2(8), Maximum size reactor coolant system rupture combined with loss of all offsite power. To create the scenario, two malfunctions were activated simultaneously from 100% reactor power, BOL. The malfunctions used were #24, " Reactor Coolant System. Cold Leg LOCA at 100% Severity" (equal to a double ended shear),and#114,"TotalLossofAllOffsitePower." Once activated, no operator action was taken while data was collected'for ten minutes. The MPCS alarm summary was saved, and ten-minute plots were made for pressurizer: pressure and level, and containment pressure and temperature. The results of-the test were reviewed by a panel of experts, The panel concluded that parameters did trend in the correct direction and that values _were reasonable. No discrepancies other than noted in the
~
individual malfunction-tests were noted. No exceptions are taken.
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o (f 46
, TRANSIENT #9 MAIN STEAM LINE RUPTURE ( b]'- The " Main Steam Line Rupture" transient test was conducted on S/1/90. ThisisanannualtestingrequirementaddressingSection5.4.2(3), Verify simulator performance against the transient criteria M 4.2 for a benchmark set of transients, and Section 8.2.2(9), Maximum size unisolable main steam line rupture. Malfunction #37, " Main Steam Line 'B' Ruptured inside Containment" was activated from 100% reactor power, MOL. No operator action was taken and data was collected for ten minutes. The MPCS alarm summary and primary calorimetric data were saved. Ten-minute plots were made for pressurizer pressure and level,-and containment pressure and temperature. The results of the-test were reviewed by a panel of experts, and they concluded that the only discrepancy was noted in the individual malfunction test abstract. Parameters did trend in the correct direction and values were reasonable. No excepti..3 are taken.
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(). V 47
, TRANSIENT #10 SLOW PRIMARY DEPRESSURIZATION The " Slow Primary Depressurization" transient test was conducted on 8/1/90. This is an annual testing requirement addressing Section 5.4.2(3), verify simulator performance against the transient-criteria of 4.2 for a-benchmark set of transients, Section B.2.2(10), Slow primary depressurization to saturated condition using pressurizer relief or safety valve stuck open. The initir.i conditions for the test were 100% reactor power, BOL with the safety itjection pumps disabled, and the associated suction valves closed. Malfunction 16 " Pressurizer Safety Valve Leakage" was activated at 30% severity (equal to approximately 30% open). No other operator action was taken. Simulation was frozen after approximately twenty minutes with the reactor coolant system saturated and the reactor coolant pumps cavitating. Data was collected during the first ten minutes of the transient. The MPCS alarm summary and primary calorimetric data were saved. Ten-minute plots ware made for pressurizar oressure, temperature and level; loop flow rates; reactor level; saturat-an margin; and hot leg and surge line (} temperatures. The results of the test were reviewed by a panel of experts. The panel concluded that the saturation condition was realistically demonstrated, all parameters graphed did trend in the correct direction, and values were reasonable. No discrepancies other than noted on the individual malfunction abstract were found. No discrepancies concerning the demonstration of saturation conditions were found. No exceptions are taken. O 48 l
3.3.2 -Transient Occurrence in the Plant 3.3.2.1 NaturalCirculationEvent(LER89-08) f/~/'% A. ! Following the. natural circulation test on June 22, 1989 at Saabrook
'Stati;ai benchmark' tests were conducted on the simulator. Plan, data from
. the~GETARS database was used for comparison. Testing, conducted in July.
1989, identified two minor discrepancies on the simulator: reactor coolant flow coastdown time was too rapid, and the narrow range T,y, , indication response to natural circulation was incorrect. Each problem was corrected and the test'was_ rerun with no discrepancies noted.
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3.4 Malfunction Tests I MALFUNCTION #1 (] v FAILURE OF MANUAL R0D CONTROL
.On October 21, 1988, the malfunction titled " Failure of Manual Rod Control" was tested. This malfunction satisfies the r_equirement of ANS/ ANSI-3.5, Section 3.1.2(13), Inability to Drive Control Rods, and (23)
Passive Malfunctions. The malfunction was tested twice, and both times was activated at 100% reactor power and deactivated ten minutes later. There is no severity option. During the first test, it was verified that rods would respond.to automatic control by shedding turbine load and allowing auto rod control to bring Tm back to program. Abnormal Operating Procedure 051210.02, Failure of Control Bank to Move, can be followed in its entirety. During the second test, rod control was placed in manual, and rod movement in both directions was attempted. No rod motion was observed. This was verified by creating and evaluating-ten-minute plots of the following
,q. parameters: core' thermal power, generator power, Power Range Channels N-41 and N-42, loop 1 CL/HL temps, NR pressurizer pressure, pressurizer hat calibrated level. SG "A" WR level,. SG A" steam pressure, SG "A" FW temp, SG "A" FW pressure, and. Control Rod Position Bank "0", Group 2, Rod 012.
At the conclusion of the second test, the malfunction was deactivated, and it was determined that the control rods would respond to the inanual in-hold-out switch. Simulator response was in accord with rod control
. schematic diagrams and the Seabrook Station rod worth. curves. No discrepancies were noted, and no exceptions are taken.
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i MALFUNCTION #2 O- AUTO ROD CONTROL FAILS IN THE "IN" DIRECTION The malfunction, " Auto Rod Control fails in the 'In' Direction" was tested on October 31, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(17), Failure in Automatic Control Systems that Affect Reactivity. The malfunction has no severity option. It was actuated at 10054 power, steady state, BOL, Bank "D" control rods at 186 steps, and deactivated 15 minutes later. Bank "D" rods began a continuous insertion at seven steps / minute due to a faulty SCR. Power range indications dropped from 100% to 85% and core thermal power dropped from 3400 MW to 2900 MW within ten minutes. It was verified after ten minutes that Bank "D" would respond to manual rod control. Abnormal Procedure OS1210.03, Continuous Rod Insertion Control, could be followed without deviation. Ten-minute plots were made of the following parameters: IR level N-35, PR Level N-43, generator power, core thermal power, Loop 1 and 2 NR Tm, SG "A" WR O- level, SG "B" steam pressure, SG "B" MS flow, SG "B" FW flow, WR PZR pressure and pressurizer saturation liquid temp. Also, the main plant computer alarm summary was saved. The ten-minute plots reflect a decrease in net reactivity, and subsequent primary and secondary thermal hydraulic conditions, consistent with the integral and differential rod worth curves provided by Westinghouse (WCAP-10982), "The Nuclear Design and Core Physics Characteristics of Seabrook Station, Unit 1, Cycle 1." There were no deficiencies noted and no exceptions taken. O 51
l l / MALFUNCTION #3 \]_/ AUTO R00 CONTROL FAILS IN THE "OUT" DIRECTION The malfunction, " Auto Rod Control fails in the Out Direction" was tested on Octcber 31, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(17), Failure in Automatic Control Systems that Affect Reactivity. There is no severity optiot). The malfunction was activated at 50% reactor power, steady state, MOL, and deactivated 15 minutes later. No operator action was taken for the first ten minutes. Bank "0" control rods began a continuous seven-step / minute withdrawal due to a faulty relay. Power range channels increased from 505s to 55%, and core thermal power went from 1730 MW to 1850 MW within ten minutes. After ten minutes, it was verified that Bank D would respond to manual rod control. Abnormal Procedure 0S1210.04, Continuous Red Withdrawal, could be followed without deviation. Ten-minute plots were made of the following parameters: PR level N-43/N-44, core thermal power, n generator power, loop 3 CL/HL temps, control rod D12 position, boron / \ V concentration,~NR pressurizer pressure, pressurizer saturation liquid temp, SG "C" WR level, SG "C" steam pressure, SG "C" MS flow, SG "C" FW flow. The main plant computer alarm summary was saved. The ten-minute plots reflect.an increase in net reactivity, and subsequent primary and secondary thermal. hydraulic _ conditions, consistent with the integral and differential.rodworthcurvesprovidedbyWestinghouse(WCAP-10982),"The
~ Nuclear Design and Core Physics Characteristics of Seabrook Station, Unit 1, Cycle 1." There were no deficiencies noted and no exceptions taken.
/m. (_) 52
E 6 fT' MJ' -MALFUNCTION #4 DROPPED R00 : RCCA 08 ,
" Dropped Rod :-RCCA 08" was~ tested on October 31, 1988. There is no severity option.. This malfunction satisfies the requirement of ANSI /ANS- I L3 5, Section?3.1.2(12),-Control Rod Failure Including Rod Drops.
. Starting l conditions were 76% Rx power, steady state, MOL, rod control in manual.: _ Activation to deactivation.was ten minutes with-no operator-action fol. lowed by the recovery of.the. dropped rod. Power range detector--
N_-43 dropped from 76% to 67% and recovered to 72%, and core thermal power-
~ dropped.from 2600 MW to'2250 MW and recovered to 2400 MW, within the ten-minute' period witti no operator action.-- Abnormal Procedure 0S1210.05, Dropped-Rod,;could-belused in its entirety.. Ten-minute plots were made-of
-the following parameters:1 PR level N-41, IR level, core thermal power, l generator power, Loop 1:CL/HL temp, NR pressurizer' pressure, pressurizer-saturationfliquid temp, SG;"A" WR level, SG'.'?A": steam pressure, SG' "A" MS-O v
- flow,DSG "A" FW= flow ~, SG "A"oFW. temp / pressure.= The following..information was'providedLbyithe main p1'nt a computer: initial and final. conditions
/ axial flux difference report', primary plant calorimetric, rod deviation report,_ quadrant power tilts report,?and alarm summary. All: data reflect ;
- net' reactivity-change.and' subsequent l primary and secondary thermal-J\ .nydraulicconditions~consistentwith(integralanddifferentialrod' worth ,
<curvesprovidedbyWestinghouse;(WCAP-10982) , "ThetNu'clearcDesignLand Core;
- Physics ' Characterist icslofiSeabrook Station,iVnit '1, Cycle.1.." LThere 'were ;:
no' deficiencies _noted;tandino_ exceptions;are_taken. J 4 53 l
W MALFUNCTION #5 OROPPE0 R00 : RCCA F8
" Dropped Rod : RCCA F8" was tested on October 31, 1988. There is no severity option. This malfunction satisfies the requirement of ANSl/ANS-3.5, Section 3.1.2(12), Control Rod Failure including Rod Drops.
Starting conditions were 76% Rx power, steady state, MOL, rod control in auto. Activation to deactivation was ten minutes with no operator action followed by the recovery of the dropped rod. Power range detector N-43 dropped from 76% to 65% and recovered to 71%, and core thermal power dropped from 2600 MW to 2280 MW and recovered to 2480 MW, within the ten-minute period with no operator action. Abnormal Procedure 051210.05, Dropped Rod, could be used in its entirety. Ten-minute plots were made of the following parameters: power level N-41, IR level N-36, generator power, core thermal power, SG "B" MS flow, SG "B" FW flow, Loop 2 CL/HL temp, SG "B" FW temp / pressure, Loop 2/3 NR T ,,, pressurizer NR em _ pressure / temp, SG "B" WR level / pressure. The following information was 'j provided by the main plant computer: initial and final conditions axial flux difference, primary calorimetric data, and MPCS alarm summary. All data reflects a net reactivity change and subsequent primary and secondary thermal hydraulic conditions consistent with integral and differential rod worth curves provided by Westinghouse (WCAP-10982), "The Nuclear Design and Core Physics Characteristics of Seabrook Station Unit 1, Cycle 1." There were no deficiencies noted, and no exceptions are taken, N 54
I 1
'd MALFUNCTION #6 vROPPED POD : RCCA H2 L.
" Dropped Rod : RCCA H2" was tested on October 31, 1988. There is no !
severity option. This malfunction satisfies the requirement of ANSI /ANS-3.5, Section 3.1.2(12), Control Rod Failure including Rod Drops. Starting. conditions were 50% Rx power, steady state, MOL, rod control in auto. Activation to deactivation was ten minutes with no operator action followed by the recovery of the dropped rod. Power range detector N-43 dropped from 51% to-43% and recovered to 48%, and core thermal power
- dropped frorr 1700 MW to 1520 MW and recovered to 1680 MW within the ten-
. minute-period with no operator action. Abnormal Procedure 0S1210.05, Dropped Rod, could be used.in its entirety. Ten-minute plots were made of the following parameters: PR Level N-43, IR Level N-35, core thermal
-power, generator power, Loop 3'CL/HL temp,-NR pressurizer pressure,
- pressurizer saturation liquid temp, SG "C" WR level / pressure / steam flow /FW j flow /FW temp /FW pressure. The following.information was provided by the -
main plant computer: initial-and final condition primary plant
= calorimetric-data, initial condition and final condition axial flux difference, initial condition and final condition quadrant power tilt report. All data reflects a net reactivity change and subsequent primary
.and. secondary thermal hydraulic conditions consistent with integral and
- differential' rod worth curves provided by Westinghouse l(WCAP-10982), "The
.)
Nuclear Design and Core Physics Characteristics of Seabrook'5tation, Unit 1, Cycle-1." There were no deficiencies noted and.no exceptions taken. w . 55 ,
~
l q MALFUNCTION #7 Os DROPPED R0D': RCCAS.H2 & H8-
" Dropped Rod : RCCAs.H2 and F8" was tested on October 31, 1988. There is no severity option.x-This malfunction satisfies the roquirement of ANSI /ANS-3.5, Section 3.1.2(12), Control Rod Failure Including Rod Drops.
Starting. conditions were 50% Rx power,. steady state, M0L. _ The test was , runLfor'approximatels 30 minutes with no operator action for the first ten- q minutes followed byja reactor trip response, Emergency Procedures E-0, Rx. 9 _Tri/p Sa f ety Injection (Steps 1-4), andes-0.1~,ReactorTripResponse,were used without deviation. Rod bottom lights for RCCAs H2 and F8 came on- ,
>immediately. Power level took a prompt drop followed by a reactor-trip'on !
pressurizer low l pressure. Ten-minute plots were made of the following parameters: PR level-N-41, IR level N-35, core thermal power,. generator power, Loop 1/2 NR T ,l pressurizer NR pressure / level, pressurizer steam / liquid temp, SG "B" WR level /pr_ essure,-SG "B"-MS/FW-flow, Loop;2- a
.CL/HL' temp,1 SG,"B" FW t temp / pressure. :The following data was collected
,I from the main plant computer: ' initial:and final primary plant 1 calorimetric data, axial-flux difference, and quadrant power tilt. The neti reactivity-drop and subsequent thermal hydraulic response was consistent withLthe--integral and differeat41 rod worth curves provided by
~
1 Westinghouse _(WCAP-10982)~,"TheNuclearDesignandCore_ Physics
~
Characteristics of'Seabrook? Station, Unit-1,' Cycle 1." There were no ideficiencies noted,(and no exceptions areitaken.. 40
.56
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~ / 's MALFUNCTION #8 V RCCA H8 FAILURE TO MOVE ON DEMAND "PCCA H8' Failure to Move on Demand" was tested on November 14, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirements of Section 3.1.2(12),
Control Rod Failure including Stuck Rods and (23) Passive Malfunctions. There is no severity option, The initial conditions were 100% Rx power, steady state, BOL;_and the duration of the test was approximately 30 minutes. During the test, rod control was placed in manual and rods were driven in to verify H8 did not 4 move. Rod alignment was restored, and rod control was returned to ' automatic. Turbine load was reduced, and H8 remained stationary. Rod 3 alignment was again restored. Turt,ine load was then increased and boron i concentration was-increased. Controlling rods stepped out, however, H8 did not move. Procedure 0S1210.06, Misaligned Rod (s), could be used without deviation. Ten-minute plots were made of key parameters. The n follcwing main plant computer data was collected: rod deviation reports, (,I quadrant power tilt reports, primary plant calorimetric data, and axial flux difference. All pls nd MPCS data reflect a net reactivity effect and subsequent thermal hydraulic conditions due to rod H8 stuck, consistent with the integ'ral and differential rod worth curves provided by
^
Westinghouse (WCAP-10982), "The Nuclear Design and' Core Physics Characteristics of Seabrook Station, Unit'1, Cycle ~1." There were no deficiencies:noted, and no exceptions are taken. O v , 57
q i G MALFUNCTION #9 V R00 POSITION INDICATION (DRPI) FAILORE FOR-RCCA H6 Rod' Position-Indication (DRPI) Failure for RCCA H6" was tested on November. 14.-1988. The initial conditions were 100% reactor power, BOL, and there is no d
/ severityoption./Activatingthemalfunctioncauseda'lossofindication - -!
for Rod H6; Also the urgent alarm'on the RPI panel,:the s,ata'A'and Data B Elights, the rod general: warning: alarm, and the_rodLbottom LED lights all-
': began flashing. The correct main plant computer?"0" point _ alarms,
- according:to_the RPIsloops and logics were activated. : Plant dynamics'were notiaffected :and.this was verified by. making ten-minute plots of several: -
; dynamic' parameters. Ten minutes'after the malfunction was activated, the-
-1 malfunction was removed. All RPI indications returned to normal,'and'the "0"poihtsalarms? cleared.D.WhileperformingProcedure 051210,07, RPI (Malfunction, Lit'wasnoted_that'theoperatorcannot-switchaccuracyLmodes y for DRPI because this: panel is not simulated.-- Although this would not- 4 chave hadian affect on this malfunction, it was noted.as_a. deficiency; and
~-the Simulatop Review Committee decided to install-a-simulated switch'as ldescribedLinnSRC' meeting: minutes #89-01.:- No exceptions are-taken.
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N -V MALFUNCTION #10 FAILURE OF AUTOMATIC AND MANUAL ROD CONTROL
" Failure of Automatic and Manual Rod Control" was tested on November 14, 1988. This malfunctiori satisfies the ANSI /ANS-3.5 requirement of Sections
-3.1.2(17), failure in automatic control systems that affect reactivity and core heat removal. (13) Inability to drive control rods, and (23) Passive
-malfunctions.
The initial, conditions were 45% Rx power, 80L. The malfunction was activated and rod control placed in manual. Rods would not respond to manual control, by group or individually. Rod control was then placed into automatic; turbine load was reduced and a short time later increasad. No rod-motion took piece. Ten-minute plots of key dynamic parameters were made. The changes in T,,, due to changes in turbine load reflect expected thermodynamic response based on steam tables. Core ~ thermal power response to the changes in Tg, was consistent with-all rods out moderator q temperature coefficient curves provided by Westinghouse'(WCAP-10982), "The
- k. J Nuclear Design and Core Physics Characteristics of the Seabrook Station, Unit 1 Cycle 1". The malfunction was removed _approximately 25 minutes after actuation. By again changing turbine load, it was verified that automatic rod control was operational as rods responded '- bring T,,, into
-program band. There were no deficiencies noted, and no exceptions are taken..
59
J~N MALFUNCTION 111 d .FAILE0FUELELEMENT(RCSACTIVITYINCREASE)
" Failed Fuel Element" was tested on November 15, 1988. This malfunction satisfiestheANSI/ANS-3.5requirementofSection3,1.2(14), Fuel
.. cladding failure resulting in high activity in reactor coolant or off gas and the associated high radiation alarms.
The malfunction has a severity option of 0-100% = 0-1% failed fuel. It was activated at 100% Rx power, BOL at 100% severity; and deactivated approximately 20 minutes later. RCS activity immediately began increasing when the malf -: tion'was activated. CCP area monitor indications also began increas...g. Procedure 051252.01, Process or Effluent High Radiation, was used and letdown was ist, lated. Letdown isolation caused CCP area-activity to begin decreasing, however, RCS activity continued to increase until malfunction deactivation. Activity levels were consistent with a 1% fuel failure as predicted in the Seabrook' Station FSAR accident j analysis. No deficiencies were noted and no exceptions are taken. i.,1
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60
I: L p, -; MALFUNCTION #12 $l h ,
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61
c m MALFUNCTION #13
)t FAILURE OF THE SSPS: TO AUTOMATICALLY TRIP THE REACTOR
" Failure of the Solid State Protection System to Automatically Trip the Reactor" was testea en January 10, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(24) Failure of the automatic reactor trip system and (23) Passive malfunctions.
The malfunction has no' severity option. It was tested twice at 100% Rx-power, HOL; and for both' tests, the malfunction was active for approximately 45 minutes for both tests. At the conclusion of the first test, nine Rx trip conditions and a manual Rx trip had been established. On the second test, five additional Rx trip conditions had been established. During the first test, the following Rx trip conditions were established: b-lo SG 1evel, pressurizer pressure lo-lo, pressurizer pressure hi-hi,-pressurizer level hi-hi, over-temperature delta temperature, overpressure delta temperature, RCS loop low flow, turbine trip, pressurizer pressure low SI, and steamline pressure low SI. The Rx trip first out lights for these conditions activated, however, the Rx failed to trip. Finally, a manual Rx trip was initiated and the Rx.did
. trip. During the second test, the following Rx trip conditions were established:- power range positive / negative flux rate high, neutron level high, RCP undervoltage/underfrequency,.and RCS loop low flow and o'vertemperature delta temperature. Again, the Rx trip.first out lights activated,_but the Rx did not trip. The malfunction was deactivated and the Rx tripped. Two procedures were used without deviation, E-0, Rx Trip.
or,SI', and ER-5.1, Response to ATWS. Ten-minute plots were made of several key dynamic parameters, and the main plant computer alarm summary l- was kept. All response data indicates that the plant did not undergo the Rx trip. There were no deficiencies noted and no exceptions are taken, b 62
k b- MALFUNCTION #14 XM SIMULTANE0US TRIP 0F BOTH MAIN FEEDHATER PUMPS
" Simultaneous Trip of Poth Main Feedwater Pumps" was tested on November 28, 1988. This malfunction satisfies the ANSI /ANS 3.S requirement of Section 3.1.2(9), Loss of normal feedwater or normal feedwater system failure.
The initial conditions for the test were 100% react $r power, MOL. Once activated, both main feedwater pumps tripped. As feed flow dropped, primary temperature, pressure, and subsequently, steam generator pressure began increasing._ Control rods began moving in due to the T,y,/T.,,. deviation. Steam generator level, dropping due to the 1 css of feedwater, swelled momentarily because of the steam dumps opening. The plant tripped on steam generator low-low level. Strip charts of key primary and secondary parameters and the main plant computer alt.rm summary were saved. Also, ten-minute plots of major primary and secondary parameters were g made. A panel of experts evaluated the following test result data to V determine simulator accuracy: SG "B" feedwater temperature and pressure, SG "B" steam flow, SG "B" level and pressure, pressurizer pressure and level, MCB strip chart recordings of pressurizer level, RCS pressure, SG flR level, feedwater flow, main steam flow, SG pressure and SG WR level, and the main plant computer alarm summary. A discrepancy was noticed on tne alarm summary. The "D" points for steam dump activation were inverted.- This was corrected under SCR # 90-044 on 8/6/90. The panel of experts concluded that the response of the key parameters evaluated was in the proper direction and was the proper magnitude. No exceptions are-taken. V(3 63
i (7. MALFUNCTION #15 A_) PZR PRESSURE INSTRUMENT PT-456 FAILS HIGH
" Pressurizer Pressure Instrument PT-456 Fails High" was tested on January 10, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2 (22), Process instrumentation, alarms, and control system failures,and(11)Lossofprotectivesystemchannel.
The malfunction har no severity option and was tested from 100% Rx power, MOL. The malfunction was active for aporoximately 15 minutes. The pressure _ channel immediately pegged high on the MCB indication and main plant computer graphic display. In accord with the NHY Precautions, Limitations, and Setpoints Manual, and the Solid State _ Protection System electrical schematics, the pressurizer pressure high' bistable trip lit and the plant tripped. No other effects were seen. Once the malfunction was removed, PT-456 returned to a normal reading. _ Ten-minute plots were made for_several key dynamic parameters and main plant computer trends for the ex four pressurizer pressure channels. The main plant computer alarm summary ' b, was. saved. All response data is consistent with benchmark data noted above. There were no deficiencies noted, and no exceptions arc taken, t '0 64
G MALFUNCTION #16 PRESSUR12ER SAFETY VALVE (RCV-116) LEAKAGE
" Pressurizer Safety Valve (RCV-116) Leakage" was tested on November 15, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of section '
3.1.2(1d), Loss of coolant including f ailu- of safety and relief valves. The initial conditions were 100% reactor power, MOL. The malfunction was activated at 10% severity where 0-100% severity is equal to 0-100% full open. Once the malfunction was activated, taiipice temperatures began increasing, pressurizer pressure began decreasing and "0" point 5791 (RCV-116 open) actuated. The reactor tripped within three minutes on over-temperature Delta temperature. The test was run for a I total of ten minutes to collect data. The malfunction results were i reviewed by a panel of experts. This data included the MPCS Alarm Summary, and 10 minute plots for subcooled temperature, pressurizer relief flow, pressurizer pressure, level and temperature, core thermal power, generator power, SG "C" steam flow level and pressure, and charging pump discharge pressure an'd flow. The panel concluded that the plotted parameters supported by the MPCS Alarm Summary did trend in the correct direction and values were reasonable based on the starting conditions. One deficiency was noted during the test: The loose parts monitor is not simulated. Therefore, the operator must determine which safety has lifted based on the actuated "0" point and the temperature indications. It was determined by the Simulator Review Committee (as per meeting minutes 89-
- 01) that the loose parts monitoring panel need not be simulated but can be trained on in the Control Room. This is an exception to the ANS1/ANS-3.5 ,
requirement of section 3.2.1 (the simulator shall contain.suffici- ' operational panels to provide the control instrumentation, alarm and other man-machine interfaces to conduct the normal plant evolutions of 3.1.1 and respond to the malfunctions of 3.1.2). A second deficiency was noted. During a training exercise on 8/8/89, the containment particulate monitor went into alarm even though it had been isolated for 15 minutes by a T signal. This will be corrected under SCR # 89-152 by 1/31/90. -No other exceptions are taken. 65 1
4
/ MALFUNCTION #17
()T PRES $URIZER SPRAY VALVE (PCV-455B) FAILS OPEN
" Pressurizer Spray Valve PCV-455B Fails Open" was tested on November 15, 1988. .This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(18), Failure of reactor-coolant pressure and volume control systems.
The malfunction was activated from 100% reactor power, MOL, at 100% severity where 0-100% severity is equal to 0-100% open spray valve. The malfunction was activated and MCB indications showed that the spray valve was open. The spray valve did not respond to an attempt to close it. The reactor tripped on over-temperature Delta temperature within two minutes of activation, however, the test was run for ten minutes to collect data. A panel of experts reviewed the following data: The MPCS alarm summary, ten-minute plots for pressurizer level and pressure, core thermal ~ power, generator power, loop 4 cold and hot leg temperatures, pressurizer steam and liquid temperatures, and SG "D" level and pressure. The panel n concluded that all plotted parameters supported by events listed on the U.. alarm summary did trend in.the correct direction and values were reasonable for the starting conditions.- No deficiencies were noted and no exceptions are taken. n - 66
O V MALFUNCTION #18 PRESSURIZER HEATER CONTROL FAILURE-
" Pressurizer Heater Control failure" was tested on December 12, 1988.
This malfunct_ ion satisfies the ANSI /ANS-3.5 requirements of section 3.1.2 (18), Failure of reactor coolant pressure and. volume control systems. The initial conditions for the test were 16% reactor power, BOL, with the color graphic system displaying pressurizer heater status. The malfunction was activated and the pressurizer pressure began slowly decreasing, until the backup heaters came on and restored pressure. Graphic displays and bus amps reflected that backup ceaters were- I controlling pressur'e. The malfunction was removed and the-controlling heaters came on. The'MPCS alarm summary, primary calorimetric data and "A" point trends were saved and ten-minute plots for pressurizer level pressure and temperature were.made. Simulator response was-in accord with
. pressurizer heater electrical schematics and the Westinghouse System
- n. Description. No deficiencies were noted and'no exceptions are taken.
L) t a N 67
4 f.- r j7 MALFUNCTION #19;and MALFUNCTION #20 l Q^; FAILURE OF-REACTOR' COOLANT PUMP "D" #1 SEAL "0" #2 SEAL FAILURE _0Fl REACTOR COOLANT _ PUMP Malfunctions'#19 and #20L'were. tested simultaneously'on December 1, 1988. 1These malfunctions satisfy the ANS!/ANS-3.5 requirement of Section 3.1.2(1),-Loss of coolant, and 3.1.2(4), Loss of forced coolant flow.. They are variable rate malfunctions, 0-100% severity'is equal to a 0-100%
- failed sol. 4 The malfunctions were tested at-100% reactor power,.M0L.- Malfunction #19,-
1" Failure ofzReactor Coolant Pump.:'D' #1-Seal", was activated first at 100% severity.: Graphic displays and MCB indications. responded correctly showinglincreased reactor, cool' ant-pump leakoff flow and decreasing pressurizer' level._ As per the' instructions in_ Procedure 0S1201.01,-RCP
- Malfunction,-the seal leakoff valve for "D" reactor coolant pe.np m s closed. - Malfunction #20;:"Failur,e:of Reactor' Coolant: Pump .'D f2 feal",.
nwas: then activated at L100% severity. Reactor. coolant pump'"[" beacing itemperature' began increasing, reactor coolant pump."D" differentia;-,eal
-l pressure began increasing,-shaft 1 frame vibration began' increasing, and iseal~ supply flow pegged 1 high. Containment pressure and': radiation' levels
.also' began < increasing'. The reactor then-tripped:on low pressurizer
- j pressure...MPCS containment radiation' level. trends,7and the MPCS' alarm'
~
isummarfwassaved.o_Also, ten-minute _plotsweremadeoften; key [ dynamic-parameters. ?-Simulator.fresponse'is'in accord with the Seabrook' Station FSAR and~the_Seabrook. Station Probabilistic-Safety Assessment 1 Technical' Summary Report provided'by Pickard, Lowe, and Garrick-Inc. There were no deficiencies noted,:and no except_ ions are-taken. ' u i i O 68
(N MALFUNCTION #21
' ( ,/- RCS MANIFOLD LEAK "The RCS Manifold Leak" is a variable severity malfunction in-which 0-100%.is equal to a 0-10,000 GPM leak from the Reactor Coolant System RTO Manifold, it satisfies the ANSI /ANS-3.5 requirements of Section 3.1.2
-(Ic), Loss of coolant, large and small reactor ccolant breaks including demonstration of saturation conditions.
The malfunction was tested at 10% severity from 100% reactor power, MOL. Testdatawascollectedandanalyzed(byapanelofexperts). The panel reviewed the following response datu: the MPCS alarm summary, MCB chart I recordings of pressurizer level and pressure, containment pressure, and ten-minute plots of thermal power, generator power, loop 1 cold leg ano hot leg temperatures, SG "B" level and pressure, reactor level, sub-
-cooled temperature, and SG'"B" steam-flow. Two discrepancies were noted.
The'"D" point-logic for the steam dump valves is inverted. This was 7-, corrected by SCR 890-044 on 8/6/90. Also,the"F" point (F4599), ' (s,) " Charging Pump 'A' Fails to Start" had activated even though the charging pump.had been running. This problem was corrected under SCR #89-051 on April 17, 1989. All dynamic parameters-did trend in the appropriate direction and values were reasonable for the starting conditions. No deficiencies? other than those discussed, were noted. 'No exceptions are taken; L , i k A
^O 69
~
s .; [V\ __ MALFUNCT10N #22 REACTOR VESSEL FLANGE LEAK ._, 7 4 tReactor Vessel Flange Leak" was tested on. November 16,'1988. .It satisfies the ANS/ ANSI-3.5 requirement of Section 3.1.2(1c), Large and d smallireactor coolant breaks. This malfunction simulates a reactor vessel 3 flange leak which can range in severity fr_om 0 to 25 gpm. It was activated _at 80% severity (a 20 gpm leakrate)~ _The in_itial conditions : were 100% reactor power, BOL. '
-As;soonasthemalfunctionwasactivated,RC-TI-401(Rxvesselflangeleak off_ temperature)beganincreasing. Ten-minute plots verified a slight PZR -
= level decrease consistent _with a.20 gpm leak. 'The malfunction was removed '
, :after approximately sixty minutes and the leak stopped. While using the JAbnormal Operating Procedure 051201.05, Loss-of_ Reactor Vessel Flange Seal,: a' deficiency was noted. The procedure requires an ouxiliary._ y operator-to close RC-V145, inner seal lockoff valve, and to.open RC-V146,: ~
outer scal valve. These valves are not simulated. The Simulator Review.
~
' Committee determined these values are not required for training. This is L an. exception to'ANS/ ANSI-3.5, Section 3.0, "The extent of simulation shall.
be such that the operator is required toLtake the same action on the. simulator to conduct an evolution as on the reference' plant using similar procedures" fand also to Section 3.1.2, "Where applicable to the
-c " malfunction, the simulator shall provide.-to the operatornthe capability of:
taking. action to recover the plant, mitigate the consequences, or:boti.."'
- 'As-described in meeting minutes #89-01', the exception is justified because there'is no control or-indication for valves.RC-V145 or RC-V146 on the MCB,yand plant recovery can be achieved by removing the malfunction. No other deficiencies were noted, and no~other exceptions are taken, s .
1 c A;
;(f 70
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)
MALFUNCTION #23 R0D CJECTION The " Rod Ejection" mr1' unction was testet on ovember 14, 1988. The malfunction satisfics the ANSl/ANS-3.2 ret,. ament of section 3.1.2(1), Loss of coolant (C) Large and small ret tor coolant $reaks including saturation condition. The malfunction was activatea from 100% reactor power, MOL. Once 6ctivated the rod position for RCCA E11 went full scale and then to zero. Pressurizer pressure decreased rapidly and the pressurizer heaters cycled on and off, due to the low level cutout. Safety injection began on low pressurizer pressure and pressure stabilized at approximately 700 psig. Cold leg temperatures decreased causing SG pressures to decrease, and the MS!V's finally . vent closed on low SG pressure. SG levels began shrinking from the turbine trip and were recovered bv emergency feedwater. Containment temperature and pressure increased. The duration of the test was approximately 35 minutes. No operator action was taken for the first 10 minutes. After the initial 10 minutes, the following procedures were s used: E-0 " Reactor Trip or Safety injection", E-1 " Loss of Reactor or - Secondary Coolant" and ES-1.2 " Post-LOCA Cooldown and Depressurization". The MPCS alarm summary and chart recordings of major parameters were saved. Ten minute plots were made of reactor power, core thermal power reactor level, subcooled temperature, containment pressure and temperature, SG "0" steam flow, level and pressure, loop 4 hot leg and cold leg temperatures and pressurfrer level and pressure. A panel of experts reviewed the results of the test and concluded that parameters did trend in the correct direction and that values were reasonable. No discrepancies were noted and exceptions are taken. O 71
l MALFUNCTION #24 REACTOR COOLANT SYSTEM COLD LEG LOCA i V The " Reactor Coolant System Cold ie0 LOCA" malfunction was tested on November 16,.1988. This malfunction satisfies the ANSI /ANS-3.5 requirementofsection3.1.2(1),Lossofcoolant. , it is a variable severity malfunction in which 0-100% severity is equal to ; 0-100% pipe break. It was activated from 100% reactor power, MOL at 100% severity. Opera or action taken during the twenty minute duration of the test consisted of tripping the reactor coolant pumps three seconds after malfunction activatici, and performing the steps of Emergency Procedure ES-1.4 " Transfer To Hot Leg Recirc". The MPCS alarm logic summary and radiation dcita summary were saved.- Ten-minute plots were made of core thermal power; generator power; loops 1-4 T,,,, charging flow and pressure; pressurizer pressure, level and temperature; RCS flow; reactor vessel level and subcooled temperature. The data was analyzed by a panel of experts and the panel concluded that parameters did trend in the g correct direction and values were reasonable for the starting conditions. () - The only discrepancy found 6.as that the RDMS trochle alarm did not activate when "B" SG isolated on high emergency feedwater flow. This problem has been corrected. No exceptions are taken. l 72
O MAlfUNC110N #25 )
- Q TUBE RUPTURE TO "C" S/G l "Tute Rupture to "C" S/G* is a variable severity malfunction in which 0-100% severity is equal to a 0-2000 GPM leak. This malfunction satisfies l
the ANSI /ANS 3.5 requirement of Section 3.1.2(1), Loss of coolant (a) l Steam generator leaks. The test, run on November 14, 1988, was activated at 2.5% severity from j 100% reactor power, BOL. Data was collected for ten minutes with no ; operator action. At the completion of the data collection, action was taken to locate and stop the leak in accord with procedure 0S1227.02, -
"Steaa Generator Tube Leak". The duration of the test was approximately
. thirty minutes. The following response data was collected: The MPCS alarm summary, mair steam line and condenser air evacuation radiation l level trends, and ten-minute plots of core thermal power, pressurizer pressure and level, loop 3 hot and cold leg temperatures, SG'"C" FW temperature and pressure, steam flow ard level, and SG "A" level and ;
pressure.- A-panel of experts reviewed the data and concluded ~that dynamic parameters did trend in the correct direction and values were reasonable for the starting conditions. No deficiencies were noted and no exceptions are taken. U 73
I i l l MALFUNCTION #26 REACTOR COOLANT PUMP "A" OVERCURRENT Trip
" Reactor Coolant Pump 'A' Overcurrent Trip" was tested on December 1, The cause of the malfunction is a faulty undervoltagt elay. This 1988.
malfunction satisfies the ANS!/ANS-3.5 requirement of Section 3.1.2(4), 3 Loss 'of forced coolant flow. This malfunction was tested at 100% reactor power, MOL, and the reactor tripped immediately on single loop loss of flow. Graphic displays of system status, SPDS, and RVLIS'showed correct pump configuration. The pump would not respond'to a restart attempt until after malfunction removal. Stripcharts for several parameters and the MPCS alarm summary ' were saved. Also, ten minute plots were made of twelve key parameters. Simulator response was in accord with the Seabrook Station FSAR; the Seabrook Station Startup Test 1-ST-11. " Reactor Coolant System riow Measurt: ment"; and the Seabrook Startup Test 1-ST-12, " Reactor Coolant System Flow Coastdown". There were no deficiencies noted, and no exceptions are taken. U O 74
(
\
MALFUNCTION #27 REACTOR COOLANT PUMP "D" LOCKED ROTOR TRIP
" Reactor Coolant Pump 'D' Locked Rotor Trip" was tested on December 1, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(4), Loss of forced coolant flow.
There is no severity option, and it was tested at 100% reactor power, MOL. The reactor tripped immediately following malfunction activation on one , single loor, loss of flow. Graphic displays of system status, SPDS, and RVLIS showed correct pump configuration. Pressure in "D" reactor coolant loop dropped low enough to actuate safety injection and letdown isolation. The pump would not respond to a restart attempt until after malfunction removal. The MPCS alarm summary was saved, and ten-minute plots of 12 key dynamic parameters were made. Simulator response was in accord with the Seabrook Station'FSAR and the Solid State Protection System loops and logics. There were no deficiencies noted, and no exceptions ire taken. O o 75 ,
i I I MALFUNCTION #28 i REACTOR COOLANT PUMP "C" OVERCURRENT TRIP l
" Reactor Coolant Pump 'C' Overcurrent Trip" was tested on December 2, 1988. The trip is caused by a faulty undervoltage relay. This I malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(4), i Loss of forced coolant flow.
This malfunction was tested at 454 reactor pewer, BOL. Due to a loss of loop "C", the "C" SG level dropped below the 10-10 level reactor trip setpoint. Pressurizer pressure dropped low enough to actuate letdown isolation. The pump would not respond to a restart attempt until after , the malfunction was removed. Graphic displays of system status, SPDS, and RVLIS showed corrLct pump configuration. The MPCS alarm summary was saved, and ten-minute plots of 12 key dynamic parameters were made. Simulator response was iri accord with the Seabrook Station FSAR;. the Seabrook Station Startup Test 1-ST-11, " Reactor Coolant System Flow Measurement"; and the Seabrook Station Startup Test 1-ST-12, " Reactor O Ceelent Sxstem Flow Ceastdewn". There were ne deficienc4es netee eee ee exceptions are taken. b e O 76
- . . ~ ~ . . , - - - - . _ - . __.
A MAlf0NCT10N #29 i REACTOR COOLANT PUMP "B" HIGH O!L TEMPERATURE
- Reactor Coolant Pump 'B' High Oil Temperature" was tested on December 2, 1988. There is a severity option, 0-100% is equal to the time to reach maximum temperature (approximately 0 40 degrees per hour). This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(4),
Loss of forced reactor coolant flow. Initial plant conditions were at 100% reactor power, MOL, with the 1 malfunction activated at 100% severity. Upon activation, pump vibration, amps, and oil temperature began to rise. Procedure 051201.01, RCP Malfunction, was used in its entirety. The malfunction was removed approximately 20 minetes after activation, and pump amps, vibration, and oil temperature slowly returned to normal. The MFCS alarm summary was saved. Simulator response was in accord with the Seabrook Station FSAR and the Westinghouse Reactor Coolant Pump System Description. There were no deficiencies'noted,.and no exceptions are taken. O 1 4 77
/ MALFUNCTION #30 \ REACTOR COOLANT PUMP "C" PCCW TO OIL COOLER LEAKAGE
" Reactor Coolant Pump 'C' PCCW to Oil Cooler Leakage" was tested on December 2, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(4), Loss of forced core coolant flow due to singic or multiple pump failure, and 3.1.2(8), Loss of component cooling system or cooling to individual components.
There is no severity option, and it was tested from 100% reactor power, MOL. Once activated, reactor coolant pump "C" amps, vibration, and bearing temperatures increased. PCCW head tank "B" level decreased an amount appropriate to the leak rate as per the tank volume curves.
' Procedure 051201.01, RCP Malfunction, was used. The reactor tripped on low low flow due to reactor coolant pump "C" tripping on overcurrent ;
approximately 12 minutes following activation. An MPCS group trend record of reactor coolant pump "C' related parameters and the MPCS alarm summary were saved. Simulator response was in accord with the Seabrook Station \ . FSAR and the technical manual provided by Westinghouse. There were no deficiencies noted, and no exceptions are taken, i O 78
)
i i MALFUNCTION #31
\ REACTOR COOLANT PUMP "A" 0!L RESERVOIR LEAK i i " Reactor Coolant Pump 'A' 011 Reservoir Leak" was tested on December 2, i 1988. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section i 3.1.7(4),Lossofforcedcorecoolantduetosingleormultiplepump ,
failure. There is a severity option in which 0-100% is equal to a 0-80 gallon loss. The malfunction was first activated at 704 severity (approximately 56 gallons) from 100% reactor power, MOL. Reactor coolant pump "A" oil reservoir level decreased and pump vibrations increased. Approximately 20 minutes later, the malfunction was removed and conditions remained "as , ts." The malfunction was activated again at 100% severity. Reactor coolant pump "A" bearing temperatures rose sharply and the pump tripped on overcurrent. Subsequently, the reactor tripped. Reference procedure 051201.01, RCP Malfunction, was used with no deviation. An MPCS group trend of reactor coolant pump "A" related parameters and the MPCS alarm O. summary were saved. Simulator response is in ucord with the " Reactor Coolant Pump Technical Manual" provided by Westinghouse. There were no deficiencies noted and no exceptions are taken. I l 79
/^ MALFUNCTION #32 i
-REACTOR COOLANT PUMP "A" HIGH VIBRATION " Reactor Coolant Pump 'A' high Vibration" was tested on December 2, 1988.
This malfunction satisfies the ANSl/ANS 3.5 requirement of Section 3.1.2 (4), Loss of forced core coolant due to single or multiple pump failure. The severity option is 0-100% equal to 0-30 mils added to base vibration. The malfunction was first activated at 33% severity from 1000 reactor power, MOL. Reactor coolant pump "A" vibration increased in, mediately, and bearing temperatures began increasing. The malfunction was rer.oved approximately ten minutes after activation and pump vibration returned to normal. It was activated again at 100% severity. Vibrations increased immediately and temperatures began increasing. Reference procedure 051201.01,. RCP Malfunction, was used with no riecessary deviations. An MPCS tN.nd of reactor coolant pump "A" related parameters and the MPCS alarm summary were saved. Simulator response is in accord with the Reactor Coo.lant Pump Technical Manual provided by Westinghouse. There , were no deficiencies noted, and no exceptions are taken, m V 80
. -- ..______ . -_ -- - - -.-_= .
MALFUNCTION #33 O' REACTOR COOLANT PUMP "0" HIGH VIBRATION
" Reactor Coolant Pump 'D' High Vibration" was tested on December 2, 1988.
This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (4),Lossofforcedcorecoolantduetosingleormultiplepumpfailure. The severity option of 0-100's is equal to 0-30 mils added to base vibration. The malfunction was activated at 100% severity from 100% reactor power, MOL. Reactor coolant pump "0" vibration increased immediately, and temperatures began slowly increasing. The reference procedure used was 051201.01, RCP Malfunction. The malfunction was deactivated approximately 20 minutes after activation. Pump vibration returned to normal and temperatures began decreasing. An MPCS group trend of reactor coolant pump "D" related parameters and the MPCS alarm summary were saved. Simulator response is in accord with the " Reactor Coolant Pump Technical Manual" provided by Westinghouse. There were no deficiencies noted, and no exceptions are.taken. v 81
(9 V MALFUNCTION #34 REACTOR COOLANT PUMP "C" LOSS OF SEAL WATER
" Reactor Coolant Pump 'C' Loss of Seal Water" was tested on December 2, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(1), Loss of coolant (c) large and small reactor coolant breaks.
There is no severity option, and the cause is an anomalous closure of reactor coolant pt.mp "C" seal water supply valve CS-V-158. It was activate ( from 1004 reactor power, MOL. Lower bearing temperature, shaft vibration, and seal _ return temperature began increasing. A slight decrease in pressuriter level and a corresponding increase in charging flow was observed. The malfunction was removed approximately 20 minutes after activation. CS-V-158 was opened and reactor coolant pump "C" seal water was returned to normal. The reference procedure used was 051201.01, RCP Malfunction. An MPCS trend record of reactor coolant pump "C" related parameters and the MPCS alarm summary were saved. Simulator response is in accord with the "Reactcr Coolant Pump Technical Manual" provided by (/ Westinghouse. There were no deficiencies noted, and no exceptions are taken. O 82
MALFUNCTION #35 MAIN STEAM LINE BREAK The malfunction " Main Steam Line Break" was tested on November 29, 1988, at 80% severity (in which 0-100% severity is equal to a 0-100% header break), from 100% reactor power, MOL. This malfunction satisifes the ANSI /ANS-3.5 requirement of section 3.1.2(20), Main steam line break outside containment. The break caused turbine impulse pressure to drop, T,,, decreased and rods began moving in because of the T,,,/T,,, deviation. An OP/ Delta-1 runback occurred due to the increase in steam demand. After an initial SG swell, the combination of decreasing RCS temperature and decreasing steam presst.re caused the steam generators to shrink, tripping the reactor on low-low level. Safety injection and main steam isolation occurred due to the rate circuit on steam generator pressure decrease. Once the main steam isolation valves closed, the RCS temperature increased almost to the atmospheric steam dump setpoint, however steam flow to the emergency feedwater pump was sufficient for heat removal. Reference procedures useri O ere t-o. aeecter Tr4n er sere.x iesec14e"". es o i. neecter trin Response" and ES-1,1, "SI Termination". Strip charts were saved, and 10 minute plots were made of major parameters. These graphs along with the MPCS alarm summary were reviewed by a panel of experts. The panel concluded that major parameters did trend in the correct direction and that their values were reasonable. No discrepancies were noted and no exceptions are taken. O l 83
MALFUNCTION #36 MAIN STEAM LINE "A" SAFETY VALVE FAILS OPEN
- Main Steam Line "A" Safety Valve fai s Open" was tested on November 21, 1988. This malfunction satisifies the AN51/ANS-3.5 requirement of section 3.1.2(20), Main ste6m line and main feed line break (outside containment).
The malfunction was activated from 100% power. MOL. Once activated the safety valve went open allowing SG pressure to decrease and SG level to swell. Feed and steam flow to the affect"d SG increased, and pressurizer pressure decreased as a result of the drop in loop "A" cold leg temperature. The duration of the test was 20 minutes. The MCB chart recordings of major parameters and the MPCS alarm summary were saved. Ten-minute plots of SG "A" and "C" steam flow, level and pressure; power level; and loop hot and cold leg temperatures were made. Test results were analyzed by a panel of experts. The panel concluded that the trends were in the correct direction and values were reasonable. During the test, a deficiency was noted. VAS procedure D5788 instructs the operator to check the acoustic monitoring panel. This panel is not simulated. it wasdeterminedbytheSimulatorReviewCommittee(aspermeetingminutes 89-01)thatthispanelneednotbesimulated,butcanbetrainedoninthe control room. This is an exception to the ANS!/ANS-3.5 requirement of - section 3.1.2: "The simulator shall contain sufficient operational panels to provide the control instrumentation, alarms and other man-machine interfaces to conduct the normal plant evolutions of 3.1.1 and respond to the malfunctions of 3,1.2". No other deficiencies were noted and no other exceptions are taken, O 84
MAlf"JNCTION #37 l g MAIN STEAM LINE "B" RUPTURE INSIDE CONTAINMENT
]
b-The " Main Steam Line "B" Rupture Inside Containment" malfunction was l tested on November 24, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(20), Main steam line and main feed line break (insidecontainment). The malfunction was activated from 100% reactor power, MOL. Once activated, "B" SG began blowing down into containment causing containment pressure to increase. The reactor tripped, safety injection initiated and ! main steam isolated on high containment pressure. Despite isolation of the "B" SG as per reference procedure E-2 (faulted SG isolation), "B" SG blew dry into containment. "B" SG pressure finally equalized with containment pressure, and began decreasing due to the activation of containment building spray._ The duration of the test was 20 minutes. The MPCS alarm summary was saved. Ten-minute plots-of the following were mader SG "B" and "C" steam flow, level and pressure; pressurizer pressure ant' level; loop 1 and 2 Tm, loop 2 hot and cold leg temperatures; core thermal power, generator power, and containment temperature and pressure. ' The: test results were reviewed by a panel of experts. The panel concluded that the parameters did trend in the correct direction and values were reasonable. No other deficiencies were noted and no exceptions are taken, t k V-85
i MALFUNCTION $38 s- ) TURBINE CONTROL VALVE #4 FAILS CLOSED "Turbire Control Valve #4 Fails Closed" was tested on December 13, 1988. The me.lfunction was tested from 100% reactor power, BOL. Once activated, turbine control valve #4 went closed, The loss of load caused the steam dumps to open and an overtemperature-delta temperature runback to occur with turbina load settling out at approximately 830 MW. The malfunction was removed approximately 20 minutes later, and the turbine control valve returned to its previous open position. Before, during, and after the malfunction, primary plant calorimetric data and the MpCS alarm summary were saved. Also, ten-minute plots were made of major related parameters. S',mulator response was in accord with the Seabrook Station P&lDs, the Seabrook Station FSAR, and the results of Millstone 3 load reduction (performed during their startup test program). There were no deficiencies noted and no exceptions are taken. E J 86
l
.C' MALFUNCTION #39 i
's . TURBINE CONTROL VALVE #3 FAILS OPEN ;
i
" Turbine Control Valve #3 Fails Open" was tested on December 13, 1988.
The malfunction was tested from 50% reacto? power, BOL. There is a severity option where 0-100% severity is equal 10-0-100% open. The malfunction was activated at 100% severit,", and the turbine control valve went full open. No operator action was tAken, allowing a plant cooldown > and a subsequent reactor trip on steam g?nera';or low level. The MPCS I alarm summary was saved. Also, ten-minu';c pitsts of major parametei s were made. Simulator response was in accord with the steam tables and the L
.Seabrook Station FSAR, and was similar to th) results of the 10% load swing performed for the Hillstone 3 pow'sr af tersion test (when dif ferences
^
-in initiating conditions are taken into accrunt). There were no discrepancies tested, and no exception!, are talen.
i L 87-
1 l 1 f~^ - MALFUNCTION #40
\, -
SIMULTANEOUS CLOSURE OF ALL MAIN STEAM ISOLATION VALVES
" Simultaneous Closure of All Main Steam Isolation Valves" was tested on November- 30, 1988. !
The initial conditions were 100% reactor power, MOL. The malfunction was activated and data was collected for approximately 10 minutes. A panel of experts' reviewed the following response data The MPCS alarm summary, and ten-minute plots of core thermal power; generator power; loop 1 cold and hot leg temperatures; loop 4 cold and hot leg temperatures; pressuriter pressure and level; SG "A" steam flow, pressure and level and SG "D" level and pressure. Results were compared to the Millstone 3 startup test results, taking into account the differences in starting conditions. The panel noted that the steam dump status logic was inverted. This has been corrected by SCR f 90-044 on 8/6/90. The panel concluded that all parameters did trend in the correct direction and their values were
, reasonable. No other deficiencies were noted and no exception are taken.
'( r I
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l f s MALFUNCTION #41 FEEDWATER REGULATING VALVE 10 "A" STEAM GENERATOR FAILS OPEN "feedwater Regulating Valve to 'A' Steam Generator fails Open" was tested on November 28, 1988. This malfunction satisfies the ANSI /ANS 3.5 , requirement of Section 3.1.2(9), Lot.s of normal feedwater or normal l feedwater system failure. The initial conditions for the test were 100% reactor power, MOL, with feedwater control in automatic. Once activated, the feed regulating valve to "A" SG failed open. No operator action was taken, and data was collecteofortenminutes._Indicationwasasexpected(i.e.,ULstatus lights for "A" MFRV indicated open, feedwater isolated, turbine tripped, both main feed pumps tripped, and the start-up feed pump did not start due totheP14 interlock), for this test, the MPCS alarm summary was saved and ten minute plots of major parameters were made. Results were reviewed by a panel of experts. A second test was run from the same starting conditions. However, manual control of "A" SG level was taken. and "A" SG dp level was restored. The malfunction was then removed, and "A" SG level control could be restored. The test result data reviewed by the panel included ten minute plots of SG "A" and "C" steam flow, SG "A" and "C" level, SG "A" and "C" steam pressure, SG "A" and "C" feedwater pressure, loop 1 cold and hot leg temperatures, loop 3 and 4 Tm, pressurizer pressure and level, core thermal power and generator power, the MPCS alarm summary. The panel concluded that the directions and values of dynamic simulator parameters were reasonable, and no discrepancies were noted. No exceptions are taken. , O v 89
_ _._ _ _.. ___ _ _ _ _ . _ _ .~. _ _ _ _ (~ MALFUNCTION #42 LOSS Of MAIN FEEDWATER PUMP "A" i
" Loss of Main feedwater Pump 'A'" was tested on November 28, 1988. This l malfunctionsatisfiestheANS!/ANS3.5requirementofSection3.1.2(9),
Loss of' normal feedwater or normal feedwater system failure. , The initial conditions for the test were 75% reactor power, BOL. Once , activated, "A" MFP tripped causing a turbine setback. Reference Procedure l 0:11231.03, Turbine Runback / Setback, was used for plant recovery. Color i graphics and MCB indication responded as expected. The malfunction was ; removed approximately 20 minutes after activation allowing the "A" MfP'to be restarted. Ter,-minute plots of major parameters were made and were reviewed h a panel of experts. These plots included SG "A" and "0" steam flow, core thermal power, generator power SG "A" and "0" level, loop 1 and loop 4 hot and cold leg temperatures, pressurizer pressure and level, and SG "A" and "D" steam pressure. The panel concluded that the direction and magnitude of change of the major parameters were reasonable Nd were l V similar to the FSAR graphs when the differences in assumptions and starting conditions were factored in. No discrepancies were noted, and no exceptions-are taken. O > 90 _._ _.. _.~ _ _ _ . _ _ . _ _ _ _ . _ _ _ . _ ._.. _ . _ ..-__._ _ _ _ . _
l J r~ x MAlfUNC110N #43 j
\j EXTRACTION STEAM VALVE TO HIGH PRESSURE FEEDWATER HEATER 26B CLOSES
" Extraction Steam Valve To High Pressure feedwater Heater 26B Closes" was tested on April 17. 1990. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2 (9), Loss of normal feedwater or normal j feedwater system failure. I 4
The initial conditions were 100% reactor power, MOL. Once activated the extractionsteamvalve(MOV-4)wentclosed. The decreasing feedwater i temperature from the 26B heater caused a slight decrease in 1,,, and : slight reactor power increase. The n,alfunction was removed approximately 30 minutes after activation and MOV-4 was open0d from the MCB. This restored extraction steam to the 268 feedwater heater and allowed plant conditions to return to the initial conditions. The MPCS alarm summary 6 was saved. Simulator response was in accord with the feedwater System P&lD and the Seabrook Station FSAR. No deficf ncies i were noted and no ' exceptions are taken.
-I 1
t 91
MALFUNCTION #44
-- - LOSS OF CONDENSATE PUMP "A"
- Loss of Condensate Pump 'A'" was tested on November 28, 1988. This malfunction satisfles the ANS!/ANS 3.5 requirement of Section 3.1.2(9),
Loss of normal feedwater or normal feedwater system failure. This malfunction was tested from 100% reactor power, MOL Once activated, i the *A" condensate pump tripped and the standby pump started, eliminating any sustained effects. After seven minutes, Malfunction #45, Loss of Condensate Pump "B" was also activated. Turbine load began shedding (ReferenceProcedure 051231.03 wasused),controlrodsbegandrivingin, both main feed pumps tripped on low suction pressure, and finally, the reactor tripped on 10-10 SG level. It was also verified that with Malfunction #44 active, "A" condensate pump could not be started. The test results were reviewed by a panel of experts'. The panels concluded that the malfunction performed correctly. The following test result data was available for review ten minute plots of power level, SG "B"
. - feedwater temperature and pressure, SG *B" steam flow, SG "B" level, SG-
"B" steam pressure, loop 2 hot and cold leg temperatures, pressurizer pressure and level, and the MPCS alarm sumary. All dynamic parameters trended in'the correct direction and values were reasonable for the-starting conditions. No discrepancies were noted and no exceptions are
-taken.
I I 92
(^ x MALFUNCTION #45 LOSS OF CONDENSATE PUMP "B"
" Loss of Condensate pump 'B'" was tested on November 28, 1988. This malfunction satisfies the ANSl/ANS 3.5 requirement of Section 3.1.2(9),
Loss of normal feedwater or normal feedwater system Failure. This malfunction was tested from 75% reactor power, BOL. Once activated, the "B" condensate pump tripped and the standby pump started, eliminating any sustained effects. After-seven minutes, Malfunction #44, Loss of Condensate Pump "A," was also activated. Turbine load began shedding (Reference Procedure 051231.03 wasused),controlrodsbegandrivingin, both main feed pumps tripped on low suction pressure, and finally, the o reactor tripped on low-low SG 1evel. It was also verified that with Malfunction #45 active, "B" condensate pump could not be started. A panel of experts reviewed the results of this test and concluded that there were no deficiencies during this test. Test result data included: ten-minute plots of SG "C" feeawater temperature and pressure, SG "C" steam flow, SG "C" level, SG "C" steam pressure, pressurizer pressure and !evel, loop 3 cold and hot leg temperatures, core thermal power, generator powe and the MPCS alarm summary. Dynamic parameters responded in the expected direction, and values were reasonable, it had been previously noted that with the 'A' condensate pump racked out and Malfunction #45 active, a feed and condensate oscillation occurs. This problem will be corrected under SCR # 88-176, as part of the new steam feed and condensate system models
.due to be completed by 12/30/91. No exceptions are taken.
N V 93 l
Q V MALFUNCTION #46 LOSS OF HEATER DRAIN PUMP "A"
" Loss of Heater Drain Pump 'A'" was tested on December 12, 1988. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(9),
Loss of r.ormal feedwater or normal feedwater system failure. 4 This malfunction was tested from 10058 reactor power, BOL. The malfunction , was activated, and the "A" heater drain pump tripped causing an OPDT runback. The malfunction was removed approximately 15 minutes later, allowing the pump to be restarted. The reference procedure used was OS123a.03, Turbine Runback / Setback. A panel of experts reviewed the following test result datat ten-minuteplotsofpowerlevel(N!-42); core thermal power; SG "C" and "D" feedwater temperatures; SG "A" and "B" steam flow, pressurizer pressure and level; M "A" and "B" feedwater pressure; SG "C" level and pressure, SG "A" and "B' feedwater temperatures, and Loop 3 cold leg and hot leg temperatures. The HPCS alarm summary and the pre-and post-test primary calorimetrics were also reviewed. There ns a concern that the change in feiedwater temperature was too great and unrealistic. This was also true when comparing the results to predicted data in the FSAR even when the differences in starting conditions were taken into-account. A generic problem that feedwater temperatures were too low at low power levels had been previously noted. This is being corrected as part of the new steam, feed and condensate systems models due to be completed by 12/30/91. All other dynamic parameters trended in the correct direction and values are reasonable for the starting conditions. No exceptions are taken. i O 94
MALFUNCTION #47
._HIGH PRESSURE FEEDWATER HEATER 26A TUBE RUPTURE This malfunction is not certified for training.
1 i-l' 4 n 4 I 1 (. i-O 4 95
i O ( MALFUNCTION #48 LOSS OF HEATER ORAIN PUMP "B"
" Loss of Heater Drain Pump 'B'" was tested oi December 13, 1988. This I malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(9),
Loss of normal feedwater or normal feedwater system failure. The initial conditions for the test were 50% reactor power, BOL. The malfunction was activated, and the "B" heater drain pump tripped. The second heater drain pump was then tripped. Feedwater control began slowly
. degrading, and thermal megawatts increased. The test was run for approximately 30 minutes and ended with the malfunction still active. The L MPCS alarm summary and pre and post test primary calorimetric data were saved. Also, ten-minute plots of major dynamic parameters were made. The results of this test were reviewed by a panel of experts, and they concluded that there were no deficiencies. This was based on a review of 'f the following test results: ten-minute plots of pressurizer pressure and (l level; core thermal and generator power; loop "A" hot and cold leg temperatures; SG "A" wide range level; pressure and steam flow; and as noted above, pre and post primary calorimetric data and'the MPCS alarm summary. All paratneters responded in the expected direction, and values were reasonable. No exceptions are taken.
l 96
-- _ - - -_= __ - _ - - _ ---. -
t'^ . MALFUNCTION #49 HIGH LEvC'. CONDENSER DUMP VALVE FAILS OPEN "High 1evel Condenser Dump Valve Fails Open" was tested on March 6, 1989. The malfunction was activated from 100% Reactor Power, MOL. The High Level Dump Valve (49149) immediately went full open, condenser level began decreasing, condensate storage tank level increased, and the condenser make up valve CO-4014A opened. The malfunction was removed approximately 15 minutes after activation; 4014B closed and condenser level began increasing. It was noted that the manual isolation valve (V 67) which would isolate the dump valve is not simulated. AnSCR(#90-098)hasbeen submitted to' correct the deficiency by no later than December 30, 1991. Simulator response was in accord with the condensatt systen, Pal 0's and the fant Volume Curves. No exceptions are taken. (:). ? E' s 5 f r o o 97 l
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MALFUNCT10N #SO ' LOW LEVEL CONDENSER MAKE9P VALVE FAILS OPEN O On March 6, 1989, " Low Level Condenser ML;eup Valve Fails Open" was tested. This malfunction satisfies the ANNANSt.-3.5 requirement of Section 3.1J(5), Loss of condenser vacuum inc'ading loss of condenser
- level control. There are no exceptions taken.
j, This malfutittfon simulates Valve 4014A f ailing open, allowing the condensate :.Nr, ge tank to gravity drain into the conyenter. There is no severity opt mn 7h' malfunction was activated at 100's rert tor power and - 3 deactivated te' imnutes later, Plant computer "A" points for condensate storage tank leve, ine-three hotoell levels, and CO P-30A/B/C discharge header pressure vc e gr'up trended. The hotwell levels increased by a cc mnding an 1 t that the CST decreased,-based on plant tank volume um ,so, the condenser spill valve 4014B opened to briq condenser level within band. When the malfunction was deactivated, 4(;4 :. hut and C51 and hotwell levels remained consttat. Although there is oc ts aormal operating proc'<inre which addresses this taalfunction, a simulator y O.- deficiency m notc*t while attempting recovery. There are two manu'ei ( isola Hon va Ne: (C0- 148 and CO-V-149) on either side of valve 40iu which can be shrt by $ e nuxiliary operator to isolate the som ce of condenser overfill. '.iese valves are not simulated. Thi', deficiency w.ss
; discussed by the Siruiator Review Cornittee as described in the meeting minutes #89-02. Tho committee decided that the training impact did warrant the addition of these valves to tt.e scope of simu btion. A simulator enbr.ncewnt Yequest to provide a remote function to open and 6- close v> W . O. 4-148 and C0-V 149 (SCR B9-174) was completed on 2/16/90, t
n O 98
MALFUNCT10N #'l LOSS OF MECHANICAL VACib PUMP "A" "Lars of Mechanical Vacuum Pump 'A'" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(5), Loss of condenser vacuuc including loss of condenser level control. There it ne severity option. The malfunction was tested from-100% reactor power, BOL. Once activated, mechanical vacuum pump "A" tripped and condenser vacuum decreased-from 28" to ro", and turbine load dropped approximately 25 MW. The standby vacuum pump started and returned vacuum to 28 " Procedure ON1233.01, Loss of Vacuum, was used to recover once the malfunction was removed, The str',pchart for t e iccndenser vacuum recorder was saved, and ten-minn a plots af generator Nwer,-core thermal power, pressurizer hot calibrated level, and pressurizer pressure were made. Simulator response was in accord w +h the vacuum purp control loop-logic diagram and the UE&C design descriptions. There we.re no deficiencies noted, and no exceptions are taken.
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W HAlfUNCTION #52 l , (/ L CONDENSER TUBE LEAK l i
' ?he " Condenser Tube Leak" is a variable rate malfunction in which 0-100%
severity is equal to a 0-200 GPM leak rate. j 1 The malfunction was tested on April 17, 1990, from 100% Reactor Power, MOL ! at 100% severity. Once activated, condenser conductivity beoan increasing l and the High Conductivity Alarm came in approximately 3 minutes after _[ activation. No problems were encountered using reference procedures j C01234.02, Condensate Tube Leak to respond to the malfunction. The malfunction was deactivated approximately 20 minutes after activation, and condenser conductivity returned to normal values. The MPCS condenser ] performance report and alarm summary were saved. Simulator response was in accord the condensate system P&ID's and logic diagrams. No
- deficiencie> were noted and no exceptions are taken.
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/O ' MALFUNCTION iS3 U EXCESSIVE LEAKAGE THROUGH FRV TO "C' S/G
" Excessive leakage through_the feedwater Regulating Valve to "C" S/G" was tested on November 16, 1988. This malfunction satisfies the A.NSI/ANS-3.5 requirementofsection3.1.2(9),Lossofnormalfeedwaterorfeeet.ater j system failure.
The starting conditions for the test were 45% reactor power, BOL. The malfunction'is a variable rate where 0-100% severity is equal to 0-30% of normal feedwater flow. The malfunction was activated at 100% severity which resulted in an immediate level deviation on SG "C". The SG "C" feedwater regulating valve closed to about 13% to maintain level. The malfunction was removed approximately 20 minutes later. A panel of experts reviewed the following data: ten-minute plots of core thermal power; loop'3 hot and cold leg temperatures; pressurizer pressurer level; SG '_'A", "B", "C", & "D" Level, SG _" A" and "C" steam pressure, SG "A" and "C" feedwater pressure and the MPCS alarm summary. The panel concluded V that the dynamic parameters did trend in the correct direction, and values were reasonable for the starting conditions. No ceficiencies were noted and no exceptions are taken. i e O. J 101
k MALFUNCTION #54' EXCESSIVE LEAKAGE THROUGH FEE 0 WATER REGULATING VALVE TO S/G '0' ( i The malfunction " Excessive Leakage Through Feedwater Regulating Valve to S/G 'O'" was tested on April 19, 1990. This malfunction satisfies the ANSl/ANS-3.5 requirement of section 3.1.2(9), Loss of normal feedwater or normal feedwater system failure. This is a variable severity malfunction in which 0-100% severity is equal to 0-30% of full feedwater flow. It was tested from 45% reactor power, BCL at 100% severity. Once activated, feedwater flow to SG "D" went up causing SG "D" ,evel to increase. Sensing the level increase,-the feedwater regulating valve closed down to return SG level to program band.
-The malfunction was removed 20 minutes after~ activation and the plant returned to the initial conditions. Primary' plant calorimetric data was saved and ten-minute plots of SG level and feed flow were made. Simulator.
response was in' accord with the Feedwater System P&l0s and the Seabrook Station FSAR. No deficiencies were.noted and no exceptions are taken. - 5 x y i i i O 102 ry
MALFUNCTION #55 O 4 LOSS OF MAIN CONDENSER VACUUM
" Loss of Main Condenser Vacuum" was tested on January 9, 1989. This mdlfunCtion satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(5),
Loss of condenser vacuum including loss of condenser level control. This is a variable rate malfunction where 0-100% severity is equal to 0-50% leakage past the vacuum breaker (leakage is the equivalent of opening thevacuumbreaker0-50%open). It was activated at 50% severity with o three-minute ramp from 100% reactor power, MOL. Once activated, vacuum began decreasing at a rate consistent with the severity and ramp time option. The standby vacuum pump started at 26" vacuum, and the steam dumps blocked at 25" vacuum. Severity was increased to 75% with a two-minute ramp time. The low vacuum alarms came in at 24.9". The severity was increased in steps until 100% severity was reached. Plant efficiency continued to decrease until a turbine trip and reactor trip occurred at 22.4" vacuum. The reference procedure used was ON1233.01, Loss of vacuum, d The simulator was reset to 100% reactor power and the malfunction was activated at 70% severity. The malfunction was removed five minutes later to verify that vacuum could be restored. An MPCS calorimetric data alarm summary, and trend record of main condenser related parameters were saved. Ten-minute plots of major parameters were made. Simulator response was in accord with the Seabrook Station FSAR,-UE&C Design Descriptions (condenser and evacuation), and vacuum pump and turbine trip logics. There were no identified deficiencies, and no exceptions are taken. O D 103
9
'f f , ~. MALFUNCTION #56 o
LOSS OF STARTUP_FEEDWATER PUMP
" Loss of Startup Feedwater Pump" was tested on 4/23/90 from 34 reactor power with the startup feed pump supplying feedwater._-This malfunction
- satisfied the ANSI /ANS-3.5 requirement of section 3.1.2(9), Loss of normal feedwater. .
Once activated, the startup;feedwaterl pump tripped and feedwater flow
- dropped to:zero.- SG levels decreased until the reactor tripped _on low SG level. The duration of the test was 20 minutes. The.HPCS alarm summary and primary calorimetric ' data were saved. Ten-minute plots were madeJof' core thermal power;. pressurizer pressure; SG "A", "B" and "C" level; -
feedwater flow; SG_"0" pressure and feedwater flow.' A-panel of experts' reviewed the results of.the-test and concluded that-parameters did' trend
- in'the correct directionEand values were reasonable. One-deficiency-was noted.- Theilock;out" alarm occurs if the malfunction is activated when.
. the startup feedwater; pump istshutdown. .In'this situation the-alarm will .
- inform the operatorJthat1the malfunction is active. This is a correct
-indication based on the cause of the_ malfunction, however, the training ,
value-is-reduced. lTherefore; a note was'added to the cause and effeet descriptionfand an enhancement request (SCR #89-179) was submitted to add-
- an additional malfunction which would rot alert the operator.to the 11mpending_ malfunction =when.the startup feedwater pump i_s.' shutdown. .No exceptions are taken. i 1
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T N- MALFUNCTION #57 d HIGH PRESSURE FEEDWATER HEATER BYPASS VALVE FAILS OPEN "High Pressure feedwater Heater Bypass Valve Fails Open" was tested on April 17, 1990. It is a variable severity malfunction in which 0-100% severity is equal to the Bypass Valve going 0-100% open. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(9), Loss of normal feedwater or normal feedwater system failure. This malfunction was tested from 100% reactor power, MOL at 100% severity. Once activated the 26A and 26B feedwater bypass valve opened, causing the-SG feedwater inlet temperature to decrease. -T,, decreased and reactor power increased slightly. The bypass valve was closed by lessening the severity in small increments over a 20 minute period. The plant 4 conditions returned to their state prior to the malfunction activation. Feedwater heater performance data and plant efficiency data were collected both before and after the test, and the MPCS alarm summary was saved.
,q , Also ten-minute plots of thermal power, pressurizer pressure, SG level, V- and feedwater flow were made. Simulator response was in accord with the feedwater-System P& ids and the Seabrook Station FSAR. No deficiencies were noted and no exceptions are taken, r
u 105 l
i MALFUNCTION #58 MSR HIGH LOAD VALVE FAILS TO CLOSE ON TURblNE TRIP .
" Moisture Separator Reheater (MSR) High Load Valve fails to Close on Turbine Trip" was tested on December 13, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(23), Passive malfunction.
It was tested from 100% reactor power, BOL. The malfunction was activated
' simultaneously with Malfunction #156, " Inadvertent Reactor Trip". The MSR high load valve #4 remained open following the reactor trip / turbine trip; however, the steam supply valves to the MSRs go shut; therefore, the effect is minimal. The malfunction was removed approximately 15 minutes later, and the MSR high load valve went shut. The MPCS alarm summary was.
saved. Also, ten-minute plots were made of major parameters. Simulator response is in accord with the steam tables and Seabrook Station P&lDs. There were no deficiencies noted, and no exceptions are taken.-
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105
MALFUNCTION #59 O HEATER DRAIN TANK HIGH LEVEL DUMP VALVE FAILS OPEN
" Heater Drain Tank High Level Dump Valve Fails Open" was tested on April 17, 1990. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2 (9), " Loss of normal feedwater or normal feedwater system failure".
The initial conditions for the test were 100% reactor power, MOL. Once - activated, the heater drain tank high level dump valve went open causing a low level in the heater drain tank. This resulted in the heater drain pump tripping on low level. The loss of feedwater heating caused the feedwater temperature to decrease and subsequently lowered T ,,, The drop in T,y, increased reactor power, resulting in an overpower / Delta-T [ runback. Removing the malfunction allowed the dump valve to go closed. The duration of the test was approximately 30 minutes. The MPCS alarm summary, feedwater heater performance monitoring data and primary calorimetric data were saved. Also 10 minute plots of feedwater O 1emneretere. <>ew. thermal ee er eed 1.. ere made. S4meieter reenense was in accord with the Seabrook Station FSAR and the Feedwater System Logic Diagrams. No deficiencies were noted and no exceptions are taken.
. O
MALFUNCTION.#60 p LOW PRESSVRE'FEEDWATER HEATER BYPASS VALVE FAILS OPEN Q The malfunction " Low Pressure Feedwater Heater Bypass Valve Fails Open"- was tested on April 17, 1990. This malfunction satisfies the ANSI /ANS 3.5 requirement of section 3.1.2(9), Loss of normal feedwater or normal feedwater system failure. This is a variable rate malfunction in which 0-100'4 severity is equal to the bypass valve 0-100% open. It was tested from 100% reactor power, M0L at 100% severity. Once activated the 25A and 25B heater bypass valve went open causing feedwater inlet temperature to decrease. T ,, dropped slightly, causing reactor power to increase-slightly. The malfunction was removed approximately 25 minutes after activation, and the low pressure feedwater heater bypass valve went closed, allowirgi the plant to return to the initial conditions. Ten-minute plots of core thermal power, pressurizer pressure and level, feedwater temperature and T , were made. Simulator response was in accord with the.feedwater system P&lDs and.the g Seabrook' Station FSAR,
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v/9 108
N MALFUNCTION #61 (/ ). MAIN STEAM REHEATER TUBE RUPTURE ,
" Main Steam Reheater Tube Rupture" was tested on April 19, 1990.- It has a variable. severity in which 0-100% severity is equal to 0-1500 pounds mass /'
hour leak rate. The malfunction was activated at 100%' severity from 100% Reactor Power, M0L. There is no MCB indication of this malfunction. MPCS Plant Efficiency.and Calorimetric Data Reports were generated prior to malfunction activation and after malfunction removal, approximately 15 minutes later. As expected, primary plant efficiency decreased slightly. This in accord with the Seabrook Station Startup Test Report ST-26,
' " Thermal Power Measurement and State Point Data". No. deficiencies were noted and no exceptions are taken.
s O 109
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MALFUNCTION #62 M
- FAILURE:0F SEAL WATER T0 MAIN FEEDWATER PUMP "A"' 1 4
" Failure of deal Water to Main Feedwater Pump 'A'_" was. tested on. December 15; 1988. ,
This isla variable severity' malfunction where 0-100% severity is equal to 100%closureofPCV-4092(scalwatersupplyvalve). The initial -;
,,[ conditions'were 75% reactor power,zBOL, and-the. initial malfunction l
~ '
;' . sever;ity was 75%. The malfunction was activated and MCB' vibration'-
+ indication' began increasing. =The severity was increased in steps'up to
-1100%. While increasing: severity 11t was noted that-the.MCB vibration
'm ; indication was inaccurate.1 The MCB indication was read.in.g approximately 4twiceas'muchastherecorder(therecorderreadingwasthesameas'the
,, "severityflevel). This deficiency was coprected.on' April _28,.1989?under !
SCR1189-004. Also,Ti_t had.been previously notedLthat'the; main feedwaterJ I l
'pumpIb' earing temperatures do not increase, this deficienep will: be (corrected under SCR 187-56 as: part of the' steam feel andhcondensate'
'(replacement to be completed by 12/30/91. - The duration of the test was L
approximately 30 minutes. The MPCSLalarm summary-and' trend' records of_ ' QN. . 'rel'ated parameters were saved. Simulator response was'in accord with the
~
gf General.: Electric vendor manual::and-the; feed.acer pump specification. No- 1 N['
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9 MALFUNCTION #63 HIGH PRESSURE FEE 0 WATER HEATER HIGH LEVEL CONTROL VALVE FAILS OPEN "High Pressure feedwater Heater High Level Control Valve Fails Open" was tested on April 19, 1990. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(9), Loss of normal feedwater or normal feedwater system failure. 3 This malfunction was tested from 100% Reactor Power, MOL. Once activated feedwater heater 260 drain valves opened. The loss of extraction steam flow due to the heater not condensing steam caused a small increase in system pressure which was felt by the SG. T,, increased slightly and thermal megawatts decreased slightly. The malfunction was removed approximately thirty minutes after activation. The drain valve went closed and systems returned to original conditions. MPCS feedwater heater performance monitoring data, plan efficiency data, and primary plant calorimetric data were saved. Simulator response was in accord with the feedwater system P&lDs and logic di grams and the Seabrook Station FSAR. No deficiencies were noted and no exceptions are taken. O 111
O d MALFUNCTION #64 LOSS OF CIRCULATING WATER PUMP "A"
" Loss of Circulating Water Pump 'A'" was tested on January 10, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (5),
Loss of condenser vacuum including loss of condenser level control. The cause of the pump trip is a faulty electrical protection relay (overcurrent). It was activated at 100% reactor power, MOL. Once activated, the pump tripped immediately, and its associated _ discharge valve went shut. The pump would not respond to a restart. This is ir accord with the-circulating water pump loops and logics diagram. Condenser vacuum and-plant efficiency decreased, circulating water Delta-T increased, and the vacuum breakers opened. The following procedures were used to recover the circulating water system: ON1233.01, Partial Loss of Vacuum; ON1038.04, Circ Water System Two Pump Operation; and ON1038.01, Circ Water System Startup. The MPCS trend records of key: p dynamic parameters were saved, and ten-minute plots of several of the same b parameters were made. Also, the MPCS alarm summary was saved. Simulator respo'nse was in acr rd with the " Analytical Studies of the Transients in a Once Through Cooling. Water System" provided by the Alden Research Laboratory of the Worcester Polytechnic Institute. The malfunction was removed approximately 30 minutes after activation, and it was verified that circulating water pump!"A" would restart. There were no deficiencies noted,- and no exceptions are taken. O v l 112
f VTe MALFUNCTION #65 (s L: -LOSS OF CIRCULATING WATER PUMP "B" r
" Loss of-Circulating Water Pump 'B'" was tested on Januar. 10, 1989. This malfunction satisfies' the ANSI /ANS-3.5 requiren. ant of Section 3.1.2(5), .
Loss of condenser vacuum including loss of condenser level control. The cause of the pump trip is a faulty protection relay (overcurrent). LThe malfunction was-activated at 100% reactor power, MOL.- Once activated, the pump tripped immediately, and its associated discharge valve went shut. The pump would not respond to a restart attempt. This is in accord
- o. with the circulating water pump loops and logics diagram. Condenser vacuum and plant efficiency decreased, circulating water DT increased, and the vacuum breakers opened. The following procedures were used to recover the circulating water system: ON1233.01, Partial loss of Vacuum; ON1038.04, Circ Water System Two Pump Operation; and ON1038.01, Circ Water System Startup. The MPCS trend records of key dynamic parameters were qe -~ ; saved, and ten-minute' plots of several of the same parameters were made, k_/ Also, the MPCS alarm summary was saved. Dynamic simulator response was in accord with the " Analytical Studies of the Transients in a 0nce Through OL Cooling Water System" provided by the Worcester Polytechnic Institute.
The malfunction was removed approximately thirty minutes after activation, 1 and it was verified that circulating water pump "B" would restart. There Ml i- ~were no deficiencies noted, and no exceptions are taken. - i i + t
<v 113 m
(7 MALFUNCTION #66 V LOSS Of CIRCULATING WATER PUMP "C"
" Loss of Circulating Water Pump 'C'" was tested on January 10, 1989. The cause is a high O/P on traveling screen C and a subsequent motor overturrent condition. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(5), Loss of condenser ve lum including loss of condenser level control.
The malfunction was activated at 100% reactor power, MOL. Once the malfunction was activated, pump amps increased, and then the pump tripped and the discharge valve went shut. The pump would restart; however, amps would again increase and the pump would trip. This is in accord with the circulating water pump loops and logic diagrams and electrical schematics. Condenser vacuum and plant efficiency decreased, circulating water DT increased, and the vacuum breakers opened. The following procedures were used to recover the circulating water system: ON1233.01, Partial Loss of f Vacuum; ON1038.04, Circulating Water System Two Pump Operation; and k ON1038.01, Circulating Water System Startup. The MPCS trend records of key dynamic parameters and the MPCS alarm summary were saved, and ten-minute plots of several of the same parameters were made. Simulator response was in accord with the " Analytical Studies of the Transients in a Once Through Cooling Water System" provided by the Alden Research Laboratory of the Worcester Polytechnic Institute. The malfunction was removed approximately thirty minutes af ter actuation, and it was verified that circulating water pump "C" would restart. There were no deficiencies noted, and no exceptions are taken. 114
MALFUNCTILN #67 O- LOSS OF CONDfN5ER WATER BOX PRIMING PUMP
" Loss of Condenser Water lox Priming Pump" was tested on January 10, 1989.
This malfunct!' satisf4 s the ANSI /ANS 3.5 requirement of section 3.1.2(5)Lossofcondenservacuum. This nialfunction was tested from 100% reactor power, MOL. The malfunction was activated and although there were no visible changes on the MCB, the MPCS "A" point trend for condenser vacuum decreased about .01 inch. Approximately 35 minurew later the " Condenser Water Box Vacuum Low" alarm came on. The malfunction was removed and vacuum restored itself to initial condition values. The MPCS Alarm Summary and "A" point trends for related parameters were saved. Simulator response was in accord with the condensate system PFIDs and logic diagrams. No deficiencies were noted and no exceptions are taken. O
}
O l 115
MALFUNCTION #68 o LOSS OF SCCW TO TURDINE OIL COOLERS CI
" Loss of Secondary Component Closed Cooling Water to the Turbine Oil Coolers" was tested on January 4, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(8), Loss of component cooling system or cooling to individual cComponents.
There is no severity option. The malfunction was tested from 100% reactor power, BOL, Once activated, turbine bearing oil drain and metal temperatures increased, and turbine vibration increased. High bearing temperature and high turbine vibration alarms were received, As per the instructions in Reference Procedure ON1231.01, High Turbine vibration, the turbine lube oil cooler setpoint was adjusted but had no effect. This is in accord with the cause of the malfunction--SCCW outlet valve from the lube oil cooler goes shut. The malfunction was removed apprcximately 20 minutes later, and the turbine oil temperatures and vibrations returned to normal. -An MPCS trend record of turbine related parameters and the MPCS alarm summary were saved. Simulator response was in accord with the SCCW 4 V UE&C drawings and foreign prints, the turbine lube oil UE&C drawings, lube oil P&l0s, the G.E. vendor manuals -("G.E. Steam Turbine Generator" - GEK49986 and "G.E. Turbine Operation Manual" - CG246), and the UESC " Lube Oil l Purification and Transfer Description." There were no deficiencies noted, and no exceptions are taken.
.A 116
O MALFUNCTION #69 < LOSS OF SCCW TO GENERATE-HYOROGEN COOLERS-
-This malfunction is not certified for training.
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?4 . MALFUNCTION 00-- '
1 $ LOSS OF PCCW.TO CVCS IET00WN HEAT EXCHANGER
'" Loss.of. Primary Component Cooling Water to the ChembAl Volume and Control' System Letdown Heat Exchanger" was' tested on January 4,L1989; ;
, This malfunction satisfies the ANSI /ANS-3.5. requirement of Section'3.1.2 y -(8), Loss ofl component cooling system or cooling to individual components,- !
and 3.1.2(18), Failure of reactor coolant' pressure-and volume control' isystems. . 1 o .E (Thereis;noseverity, option..-Thefunctionwastestedfrom100% reactor-n power..80l.1 The cause of the malfunction is TCV-130 failingLshut. . When the malfunction was activated,'TCV'129 diverted letdown flow from the'
;demineralizers the' letdown temperature control-CS-TK-130 output went to
-' -100%, and l'etdown heat _ exchanger temperature indication went..high..~.The-malfunctio'n.was removed'appr6ximately twenty minutes after activation and
- ;the system returned.to normal.- :There is no' reference procedure for'this-
~
<: U l malfunction. An MPCS trend record of' letdown temperatures'and flow,4 and
- the MPCS alarm summary were saved. Simulator response was ins accord Yith ithe SeabrookLStation FSAR,rthe CVCS P& ids, the CS-TK-1. feed and level-loop.
[' Idiagrams fand CS-TCV-129:logicidlagram.1- There were no deficiencies noted,- ps Land no exceptions;are taken.= J U g c f a ir i ? ( '[' ,
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[N ' m/ MALFUNCTION #71 LOSS OF PCCW TO CENTRIFUGAL CHARGING PUMP "A"
" Loss of Primary Component Cooling Water to Centrifugal Charging Pump 'A'"
was_ tested on January 4, 1989. This malfunction satisfies the ANS!/ANS-3.5 requirement of Section-3.1.2(8), Loss of component cooling system or cooling to individual components. l There.is no severity option. -The malfunction was tested from 100% reactor power, BOL. The cause of the malfunction is a dropped disc on-PCCW supply-th valve CC-V-31. Once activated, centrifugal charging pump "A" bearing oil
-temperatures and gear box temperatures began-increasing._ " Low PCCW Flow" E andl"CVCS Inop" alarms were received. The malfunction was removed ten minutes after activation, and the system returned to normal.- .There is no reference procedure for this malfunction. Simulator response was in
' accord with'the CVCS_PalDs, the charging pump _ logic diagram, and the'
, -- . Westinghouse vendor manual, "Chargifig Pump Operating and Maintenance Manual"(546AAV-236900-BPE). There were no deficiencies noted, and no exceptions are taken. 4 5
s O. 119
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9[\ M ' MALFUNCTION #72 (A . , LOSS OF-PCCW-TO BTRS CHILLER PACKAGE
'{i' c" Loss'of' Primary Component Cooling Water to the BTRS Chiller Package" was
, _ tested'on February;28, 1989, i _. ; y , There is'no-l severity option. This malfunction ~was! tested from 75% reactor j T - power, MOL,--with-the BTRS 1.ined-up for dilution through demineralizers "A"
,% :and."B".--Once activated, the chiller package outlet temperature' increased. 3
- 'and the demand for PCCW to the chiller increased to 100%. Rate of boron- U n,
~g (concentration decreased:to almost zero. The reference procedure used was 1 i ,VAS 4664, BTRS' Chiller 1 Trouble. The malfunction was removed approximately
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1..y _ fifteenminutesfafterlactivation..Thechillertemperaturebegan j . J.ll[ , Ldecreasing',iand the control rods began movingiin. response to the dilution. 7 An'MPCS, trend record of BTRSLtemperatures and-flows and the MPCSialarm VI Lj. summary were saved. Simulator response'was in accord'with the CVCS P&lDs. A n Cs , iahdjthe!BTRS= logic. diagrams.:Therewerenodeficiencies:noted,and:no- ,
- Lexceptions are takeni s
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MALFUNCTION #73 LEAK IN THE "A" PCCW LOOP
" Leak in the 'A' PCCW Loop" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(8),
Loss of component cooling system or cooling to individual components. There is no severity option. The malfunction was tested from 100% reactor ] power, BOL. Once activated, the "A" loop PCCW surge tank level began decreasing. PCCW to waste processing building isolated in nine minutes, and PCCW to containment isolated in eleven minutes. This is supported by the NHY tank level curves and the PCCW to WPB and the PCCW to containment logic diagrams. The reference procedure used was 051212.01, PCCW System Malfunction. The malfunction was removed approximately twenty minutes after activation, and the PCCW "A" head tank was refilled using the remote function. The MPCS trend record of PCCW related parameters and the MPCS alarm summary were saved. There were no deficiencies noted, and no exceptions are taken. O 121
(
\_)Y-MALFUNCTION #74 TOTAL LOSS OF "B" PCCW LOOP
" Total Loss of the 'B' Primary Component Cooling Water Loop" was. tested on January 5, 1989. This malfunction sa.tisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(8), Loss of component cooling system or cooling to individualcomponents,and(7)Lossofshutdowncooling.
There is no severity option. The malfunction was tested from 100% reactor power, BOL. Once activated,_the "B" PCCW pump tripped and the back-up pump failed to start resulting-in a complete loss of PCCW to "B" loop supplied equipment. The reference procedure used was OS1212.01, PCCW System Malfunction. An MPCS trend record of PCCW flows, temperatures, and head tank levels, and the MPCS alarm summary were saved. Simulator response was in accord with the Seabrook Station FSAR, the P&lDs for "B" loop supplied equipment, and the PCCW-pump loops and logic diagram. There were no deficiencies noted, and no exceptions are taken. [v ') s f
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1 122
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MALFUNCTION 0 5 O.. LOSS OF SERVICE WATER PUMP "A" l
" Loss of Service Water Pump 'A'" was tested on January 5, l'989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(6).
Loss of service water or cooling to individual components. There is no severity option. The malfunction was tested from 100% reacte-r ( power, BOL. The cause of the malfunction is a faulty thermal overload. Once activated, service water pump "A" tripped, and service water pump "C" started as per the service water loops and logic diagram. Service water flow and pressure decreased and then returned to normal following the start of service water pump "C." The reference procedures used were 051216.01, Degraded Ultimate Heat Sink, and 051016.03, "A" SW Train Operation. An MPCS trend record of service water pump temperatures and pressure was saved. A deficiency was noted from the trend record. Pump temperatures remained constant throughout the transient and did not reflect the pump status. AnSCRhasbeensubmitted(#89-020)andis scheduled for completion no later than December 31, 1991 . Excluding ttis deficiency, simulator response was in accord with the service water P&Ds, the Seabrook Station FSAR, and the service water pump vendor manual provided by the Johnson Pump Co. There are no exceptions taken. l l O 123
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~ ' / 'N MALFUNCTION D6
() - LOSS OF SERVICE WATER TO THE "B" PCCW HEAT EXCHANGER q
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" Loss of Service Water to the 'B' Primary Component Cooling Water Heat n Exchanger" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(6), 'oss of service water or cooling to individual components.
There is no severity option. The malfunction was tested from 100% reactor , , power, BOL. Once activated, "B" SW pressure increased, "B" PCCW heat exchanger temperature began increasing, and temperatures began increasing on equipment cooled by "B" PCCW. Procedure OS1216.01, Oegraded Ultimate ' Heat Sink, was used; however, "B" PCCW could not be restored due to the cause of the malfunction--a clogged throttle valve. The malfunction was ) removed approximately twenty minutes after activation, with the "B" PCCW supplied component temperatures continuing to increase. Service water to the "B" PCCW heat exchanger was then restored, and temperatures returned to normal. The MPCS alarm summary and an MPCS record of "B" PCCW related (g) parameters, including' component temperatures, were saved. Simulator response was in accord with the service water and PCCW P& ids and the Seabrook Station FSAR. There were no deficie,1cies noted, and no exceptions are taken. 4 2 124 l 1
/G MALFUNCl30N #77 d LOSS OF SERVICE WATER TO THE "A* PCCW HEAT EXCHANGER
" Loss of Service Water to the 'A' PritMey Component Cooling Water Heat Exchanger" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(6), Loss of service water or cooling to individual components.
There is no severity option. The malfunctior, was tested from 100% reactor power, BOL. Once activated, 'A" SW pressure increated, "B" PCCW heat exchanger-temperature began increasing, and teneratores began increasing on equipment cooled by A" PCCW, Procedure 0 9 216.01, Degreded Ultimate Heat Sink, was used; however., "A" PCCW could not be restored due to the cause of the taalfunction--a clogged throttle valve. TLe malfunction was removed approximately twenty minu.tes after activeion with the "A" PCCW supplied-component temperatures continuing to increase. Service water to the "A"'PCCW heat exchanger was then restored, and temperaturet returned ;
.g. to normal.- An MPCS record of "A" PCCW related parameters-including d component temperatures and the MPCS alarm summary were saved. Simulator response was in accord with the service water and PCCW P& ids and the Seabrook Station FSAR. There were no deficiencies noted, and no exceptions are taken.
i [- i r / L 125 L 1
9 MAlfUfW110h #70 R/E.TGK 00LANT PUMP "B" THERFAL BARD (R LEAK TO TBCCW "Macter tw ' cat Pump 'O' Thermal Barrier Leck to IBCCW" was tested on Jct.imber 12, 1904. This is a variable severity malfunction in which 0-100) severity is equal to a 0-120 gpm leak, it was activated at 50% severity from 100% reactor power, 80L. Once activated, the thermal barrier head tank level and the reactor coa ' ant pump "B" thermal barrier inlet temperature bt.gan increasing. The dialfunction was removed and the leak stopped. Xt was then activated at 90% saverity after resetting the simulator to 100% reactor power, BOL. Once activated, automatic TBCCW is91ation occurre. ' The re'iarence procedure used was OS12U.01, PCCW Nfunction. An MPC'. trend record of reatt r coolant pump "B' related parameters and the McCS alarm summary were sa 'ed. With the exception of one noted de icienc'c, simulator response was in accord wit:1 UE&C component cooHng cirdsinge, foreign prints, and the Reactor Coolant Pump Technica! Manual" prov ided by Westirp:)ouse. The deficiency fcund was the reactor coolant pump HPCS color . graphic display and usociated 9 points did not respond to the low flow conditions once TBCCh' was isolater!, An SCR has been written (#89-007) ard was completed on January 26,1Rd. There are no exceptions taken, 126 . 1
, s f MAlf0NCT10N #79 LOSS OF PCCW TO THE 'B' RHR HEAT EXCHANGER
" Loss of PCCW to the 'B' RHR Heat Exchanger" was tested on April 23. 1990.
It is a variable severity malfunction in which 0-100% severity is equal to a 0-100% flow blockage to the heat exchanger, and it satisfies the ; ANSI /ANS-3.5 requirement of Section 3.1.2(8), loss of component cooling or cooling to individual components. The malfunction was tested from Mode 4, train "B" RHR in service at 100% . severity. Once activated, the "B RHR heat exchanger outlet temperature begar, increasing, inlet and outlet temperatures slowly equalizeds and RCS temperature stopped decreasing. The test was run for approximately one
- hour. MPCS RHR analog _ temperature point data was saved. Simulator
. response was in accord with the PCCW System P&lDs. No deficiencies were noted, and no exceptions are taken.
O l U l 127 1 l-
MALFUNCTION #80 LOSS OF PCCh TO THE 'A' RHR HEAT EXCHANGER
" Loss of PCCW to the 'A' RHR Heat Exchanger" satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(u), Loss of component cooling or cooling to individual components.
Malfunction #80 is a variable severity malfunction in which 0-100% severity is equal to a 0-100% Flow Blockage (to the heat exchanger). It was tested on Apri.1 23, 1990 from Mode 4, RHR Train "A" in service, at 100% severity. Once activated, the "A" "HR heat exchanger outlet temperature began increasing, inlet and outlet temperatures slowly equalized, and RCS temperature itopped decreasing. The test was run for-aporoximately 45 minutes. MPCS RyR analog temperature point data was saved. Simulator response was in accord with the PCCW system P&lD's. No deficiencies were noted, and no exceptions are taken. 128 L- _ _ _ _ _ _ - - _ - _ _ _ - - _ _ _ .
s - - MALFUNCTION #81
-s FOULING OF STEAM GENERATOR BLOWOOWN FLASH TANK HEAT EXCHANGER "A" / \
() -
" Fouling of the S/G Blowdown Flash Tank Heat fxchanger 'A'" was tested on March 8, 1989.
It was tested from 45% reactor power, BOL, at 50% severity where the severity option is equal to a 0 85% decrease in heat transfer coefficient. , Once activated, flash tank temperature increased, heat exchanger i temperature spiked and then decreased due to valve 6519 closing. This caused blowdown flash tank level to increase. At high level, the inboard isolation valves closed. The malfunction was removed approximately twenty minutes later, and-using Reference Procedure 051227.01, " Recovery from Steam Gener6 tor Blowdown System Isolation," blowdown was restored. During the test, deficiencies were noted. The color graphic display and MPCS "0" points for the inboard isolation valves did not change when the valves ' charged. position. Also; 0" points for flash tank high level and high pressure were not received. An SCR was womitted (#89-061) and was
- c. corrected on July 17, 1989. The MPCS SG blowdown digital point status, the "A" point trends for SG blowdown related parameters, and the alarm summary were saved. Other than the deficiencies noted, simulator response was in accord with the Seabrook Station FSAR and the UE&C System Det gn Description. There are no exceptions taken. ,
l l( 129
. -- . . - _ _ - -. . - - - . - _ . .~
1 O MALFUNCTION #82 V FOULING OF STEAM GENERATOR BLOWDOWN FLASil TANK HEAT EXC61 ANGER "B"
" Fouling of the S/G Blowdown Flash Tank Heat Exchanger 'B'" was tested on March 9, 1989.
It was tested from 45% reactor power, BOL, at 75% severity where the severity option is equal 1. 0 85% decrease in the heat transfer coefficient. Once activated, flash tank temperature increased, heat exchanger temperature spiked and then decreased due to valve 6519 closing. This caused blowdown flash tank level to increase. At high level, the inboard isolation valves closed. The malfunction was removed approximately twenty minutes later, and using Reference Procedure 051227.01, " Recovery from Steam Generator Blowdown System isolation," blowdown was restored. During the test, deficiencies were noted. The color graphic display and MPCS "D" points for inboard isolation valves did not change when the valves changed positdon. Also, "D" points for flash
, tank high level and high pressure were not received. An SCR was submitted (3) -(189-061) and was corrected on July 17, 1989. The MPCS "A" point trend for SG blowdown related parameters was saved. Other than the deficiencies '
noted, simulator response was in accord with the Seabrook Station FSAR and the UE&C System Design Description. There are no exceptions taken. I 130
t'~\ MALFUNCTION #83 (): LOSS OF SEAL INJECTION TO ALL REACTOR COOLANT PUMPS
" Loss of Seal Injection to all Reactor Coolant Pumps" was tested on January 3, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(18), failure of reactor coolant pressure and volume control systems.
The malfunction was tested at 75% severity where the severity option is equal to 0-100% loss of seal injection due to a clogged seal injection filter. The initial conditions for the test were 45% reactor power, BOL. Once activated, color graphic displays and MCD indication of seal injection parameters showed the drop in seal injection flow. Reference procedure guidelines from 04620, Seal Injection Filter "A" DP High, and 051003.03, Operation of Seal Injection Filters, were-followed, and restoration of seal injection was attempted by adjusting V-182. Malfunction severity was then increased to 90%, and the high filter D/P alarm activated. Approximately 15 minutes after malfunction activation, a remote function was used to simulate changing out the filters, and flow was restored. The MPCS alarm summary and the "A" point trend record for related parameters was saved. Simulator response was in accord with the
-seal injection P&l0s and the Westinghouse RCP Technical Manual. No deficiencies were noted, and no exceptions are taken.
4 f O 131
I r MALFUNCTION #84 't LETDOWNlSOLATIONVALVE(V-IS0)FAILSCLOSED
" Letdown Isolation Valve (V-150) Fails Closed" was tested on January 3, ,
1989. This malfunction satisfies the ANS!/ANS 3.5 requirement of Section 3.1.2(18), failure of reactor coolant pressure and volume control systems. The malfunction was tested from 45% reactor power, BOL. The malfunction i was activated. V-150 went shut and letdown flow decreased to zero. The system relief valve opened, and the pressurizer relief tank temperature, , pressure, and level began increasing. Reference Procedures 051202.01, Loss of Letdown, and 051001.08, PRT Operation, were used. An unsuccessful attempt to open V-150 was made. The malfunction was removed approximately fifteen minutes after activation and letdown was restored. The MPCS alarm summary was saved, and ten-minute plots of letdown flow and pressurizer level were raade. Simulator response was in accord with the Chemical and Volume Control System P&l0s and the Seabrook Station Precautior,s, Limitations, and Setpoints Manual. There were no deficiencies noted, and no exceptions are taken. U 132
i 1 i i f' MALFUNCTION #85 , BACK PRESSURE REGULATING VALVE (PC-V131) FAILS OPEN !
"Back Pressure Regulating Valve (PC-V131) Fails Open" was tested on January 3, 1989. This malfunction satisfies W ANSI /ANS 3.5 requirement ofSection3.1.2(18),Failureofreactorcoolantpressureandvolume control systems. '
The malfunction was tested from 45% reactor power, BOL. Once activated, PC-V131 went shut due to a faulty signal from pressure transmitter PT-131. Letdown flow increased causing a slow decrease in pressurizer level. The malfunction was removed approximately fifteen minutes after , activation, and PC-V131 was returned to automatic operation. The MPCS j alarm summary was saved, and ten-minute plots of letdown flow and letdown pressure were made. Simulator response was in accord with the Chemical and Volume Control System P&lDs and logic diagrams. No deficiencies were found and no exceptions are taken, i l l O 133 i
V O MALFUNCTION #86 BACK PMSSURE REGULATING VALVE (PC-V131) FAILS CLOSED "Back Pressure Regulating Valve (PL-V131) Fails Closed" was tested on January 3, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(18), failure of reactor coolant pressure and volume control systems. The initial conditions were 45% reactor power, BOL. Once activated, PC-V131 went closed, and letdown flow dropped to zero. Letdown flow was restored using manual control in accord with Procedure 051202.01, Loss of / Letdown. The malfunction was removed and PC-V131 was returned to automatic operation._ The MPCS alarm summary was saved, and ten minute plots of letdown flow and letdown pressure were made. SimulMor response \ was in accord with the Chemical and Volume Control System P&lDs and logic diagrams. -There were no deficiencies noted and no exceptions are taken. 134
n () MALFUNCTION #87 TEMPERATURE CONTROL VALVE (TCV-129) FAILURE TO VCT POSITION
" Temperature Control Valve (TCV-129) Failure to VCT Position" was tested on February 28, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(18), failure of reactor coolant pressure and volume control systems.
The initial conditions were 75% reactor power, MOL, with the BTRS aligned for dilution through "A" and "B" demineralizers Once activated, TCV-129 switched to the bypast demins position due to a faulty high temperature signal, and BTRS flow decreased to zero. An attempt was made to reposition TCV-129 to the demins position; however, the valve would not respond. The malfunction was removed approximately ten minutes after activation, and TCV-129 was returned to the domin position. The reference procedure used was 04695, Letdown Heat Exchanger and Demins Bypass. The } / MPCS alarm summary was saved, and simulator response was in accord with the Chemical and Vo' .a Control System P&lDs and Logic Diagrams. There () were no deficiencies noted and no exceptions are taken, s o\ 135
f r MALFUNCT10N-#88 (,,)y_ LETDOWN LINE LEAK
" Letdown Line Leak" is a variable severity malfunction where 0-100% is equal 0-120 gpm leak rate. It was tested on January 3, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(18), t Failure of reactor coolant pressure and volume control systems, and Section3.1.2(1),Losscfcoolant.
The initial conditions were 45% reactor power, BOL. It was activated at 50% severity (approximately60gpm). Once activated, PAB radiation alarms including radiogas, ventilation, and air particulate were activated. VCT leval decreased approximately 1.5% per minute which correlates to a 60 gpm leak rate according to the tank level curves. The test was concluded in approxiraately fif teen minutes. The MPCS alarm summary was saved, and ten-minute plots were made of pressurizer level and letdown flow. Response was in accord with the Chemical and Volume Control System P&lDs and the
~~ Seabrook Station FSAR. No deficiencies were noted and no exceptions are (s ,/ taken.
r b O 136
O (/ MALFUNCTION #89 AUTO MAKEUP CONTROLLER MALFUNCTIONS
" Auto Makeup Controller Malfunction" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(i8),
Failure of reactor coolant pressure and volume control systems, and 3.1.2(17), failure in automatic control system that affect reactivity. For this malfunction, the boration valve controller (FX-110) will not open theborationvalve(FCV-110A)inautoormanual. Boration can only be accomplishedatmaxrate(approximately43gpm)byeithermanuallygoing to open on FV-110 or via the emergency boration flow path. The starting I conditions for the test were 100% reactor power. BOL, with a 100 gallon manual makeup with current boron concentration (811.9 ppm) set to begin. i The malfunction was activated, and makeup was started by way of V-1108 to ! charging pump suction. V-110A did not open. When the make-up was complete, RCS boron concentration had dropped to 810.7 ppm. V-110A would not change position until it was opened manually using the control switch, g/ L The malfunction was deactivated approximately twenty minutes after i activation, and FCV 110A opened when manual make-up to the RCS was started. The MPCS alarm summary was saved, and ten-minute plots of boron concentration and pressurizer leve' were made. The simulator response was in accord with the Chemical and Volume Control System P&lDs and logic diagrams and with the Westinghouse Boron Concentration Make-un and Dilution Curves. There were no deficiencies noted and no r ' ptions are , taken. O 137 l
if MALFUNCTION #90 MAC UP CONTROLLER FAILS IN BORATE MODE
" Makeup Controller Fails in the Borate Mode" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(17), Failure in automatic control systems that affect reactivity, ,
and Section 3.1.2(18), Failure of reactor coolant pressure and volume control systems. The malfunction was tested from 100% reactor power, BOL. The malfunction was activated, and blended makeup was started through FCV-1108. FCV-110A
~
would not respond'to controller operation and was manually closed using the control switch after 109 gallons of boric acid was added. Makeup was then set for an additional 13 gallons, and the malfunction was deactivated. FCV-110A opened, and RCS makeup began.- The MPCS alarm summary was saved, and simulator response was in accord with the Chemical and Volume Control System P&lDs and logic diagrams. No problems were encountered using Reference Procedure 0S1008.01, CVCS Operation. There O e e ne eeficiencies note 8 ane ne excePt4 ens ere teken. { 138
gm MALTUNCTION #91 VCT HIGH LEVEL DIVERT VALVE FAILS IN VCT POSITION p {} The "VCT High Level Divert Valva fails in the VC1 Position" was tested on January 4, 1989. This malfunctica satisfies the ANSI /ANS-3.S requirement I of Section 3.1.2(18), f ailure of reactor coolant pressure ano volume control systems. The malfunction was tested from 100% reactor power BOL, with the divert controller at 40% and letdown diverting to the primary drain tank. The malfunction was activated nine minutes after running the simulator in these conditions. Thedivertvalve(LCV112A)wenttoVCT. LCV 112A would not respond to MCB control. The malfunction was removed and LCV-l 112A went to divert and at the correct setpoint returned to VCT. Observed simulator response was in accord with the Chemical and Volume Control System P&lDs and the Precautions, Limitations, and Setpoints manual. No discrepancies were noted and no exceptions are taken, u) 139
>%LFUNCTION #92 O- VCT-HIGH LEVEL DIVERT VALVE TO RETURN TO VCT AFTER DIVERTING "VCT High Level Divert Valve Fails to Return To VCT After Diverting" was !
tested on January 4, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirementofSection3.1.2(18),Failureofreactorcoolantpressureand volume control systems. The malfunction was tested from 100% reactor power, BOL, with the VCT Divert Controller at 40%. The malfunction was activated and when VCT level reached 40% LCV-112A continued to divert. At 30% VCT level, auto makeup began and level returned to 50%. The malfunction was removed approximately fifteen minutes later and LCV-112A returned to the VCT at the correct setpoint. The malfunction was inserted again to verify that LCV-112A would change position. The observed response was in accord with the Chemical and Volume Control P&lDs and the Precautions, Limitations, and Setpoints manual. There were no deficiencies noted and no exceptions are taken. i-0 140
- = - . .-
i f^ MALFUNCTION #93
\ BTRS DILUTE MODE FAlLURE "BTRS Dilute Mode failure" was tested on March 1, 1989. This malfunction satisfies the ANS!/ANS-3.5 requirement of section 3.1.2(18), failure of ,
reactor coolant pressure and volume control systems. ! The initial conditions for the test were 75% reactor power, MOL, with BTRS aligned for dilution through "Aa and "B" demineralizers. The malfunction was activated and V-301 failed closed, the dilute light went out, and dilute flow went to zero. The BTRS selection switch was then placed in off. The malfunction was removed approximately ten minutes after activation and dilute flon remained at zero. The MPCS Alarm Summary and the "A" point trend for demineralizer temperatures and flow were saved. Simulator response was in accord with the Chemical and Volume Control System P&lDs and loops and logic diagrams. No deficiencies were noted and no exceptions are taken. 141
/^)
V MALFUNCTION #94 SEAL FLOW CON 1ROL VALVE HCV-182 FAILS CLOSED
" Seal Flow Control Valve MCV-182 Fails Closed" was tested on January 5, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(18), failure of reactor coolant pressure and volume control systems. ,
it was tested from 100% reactor power, BOL. Once activated charging flow decreasedandsealinjectionflowincreased,andHCV-182wouldnotrespond to manual control. . Reference procedure 051202.02, Loss of Charging, was-used to recover. The malfunction was removed approximately fifteen minutes after activation and. charging flow returned to normal. The MPCS Alarm Summary and "A" point trends for related parameters were saved. Simulator response was in accord with the Chemical and Volume Control System P& ids. No deficiencies were noted and no exceptions are taken.
.\
O 142 l .-- -_ - - - , _ - ,
MALFUNCTION #95 O CHARGING HEADER LEAK
" Charging Header Leak" was tested on January 4, 1989. This malfunction satisfies the ANSl/ANS-3.5 requirement of section 3.1.2(18), failure of reactor coolant pressure and volume control systems.
The malfunction has a variable severity leak rate, where 0-100% severity is equal to a 0-100 gallon per minute leak rate. The malfunction was activated at 75% severity. Following activation, pressurizer level began decreasing and PAB radiation levels began increasir.g. In accordance with reference procedure 051002.02, Charging Halfunction, 'etrown was isolated. Charging pump "A" was stopped and HCV-182 closed. This stopped the leak, and volume control tank level began increasing due to seal return flow. The following test results were reviewed by a panel of experts: the MPCS alarm summary and "A" point trends, and ten-minute plots of pressurizer level, pressure, charging pump flow, and discharge pressure. The panel concluded that all dynamic parameters did trend in the expected direction and that values were reasonable for the leak rate. No deficiencies were noted and no exceptions are taken. O 143
MALFUNCTION #96 AUXILIARY SPRAY VALVE FAILS OPEN
- Auxiliary Spray Valve fails Open" was tested on January 4, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of section 3.1.2(18),
failure of reactor coolant pressure and volume control systems. The initial conditions were 1004 reactor power, BOL. Once activated pressurizer pressure began slowly decreasing wr,til tne backup heaters came on restoring pressure. The malfunction was removed approximately ten minutes later, and the auxiliary spray valve went closed. The MPCS alarm summary and "A" point trends for related parameters were saved. Also ten-minute plots of pressurizer pressure, level, and temperature were made. l Simulator response was in accord with the Chemical and Volume Control I System and Reactor Coolant System P&l0's and the Westinghouse System Description. No deficiencies were noted and no exceptions are taken. O O 144
MALFUNCTION #97 ['b} POWER RANGE CHANNEL 41 FAILS HIGH
" Power Range Channel 41 Fails High" was tested on December 15, 1988. This malfunctionsatisfiestheANSI/ANS-3.5requirementofSection3.1.2(21),
Nuclear instrumentation failure. There is no severity option. The malfunction was activated from 100% reactor power, ROL, control rods in automatic. Once activated, Channel 41 indicction immediately went high. Channel I bistables tripped. Rods began moving in until power stabilized at 96% and T,,, stabilized at 580'F. The malfunction was removed approximately twenty minutes after activation, and control rods moved out to restore T,,,. Channel 41 returned to normal. Reference procedure 051211.04, Power Range NI Instrument failure, was used with no necessary deviations. An MPCS primary plant calorimetric and the MPCS alarm scmmary were saved. Also, ten-minute plots of core thermal power, pressurizer pressure, and power i q range channels N-41 and N-42 were made. Simulator response was in accord D with the Westinghouse functional Block Diagram, the Westinghouse Technical ManualfortheNuclearInstrumentationSystem(W-120-32)andtheSeabrook Precautions, limitations, and Setpoints manual. There were no deficien-cies noted, and ne exceptions are taken. O 145
(
' MALFUNCTION #98 POWER RANGE CHANNEL 42 FAILS HIGH *
" Power Range Channel 42 Fails High" was tested on December 15, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(21),
Nuclear instrumentation failure. There is no severity option. This malfunction was activated from 45% reactor power,~BOL, control rods in automatic. Once activated, Channel 42 indication immediately went high, and the power mismatch caused the control rods to begin moving in. Approximately r.inety seconds af ter activation, control rods were placed in inanual as per Procedure OS1211.04, Power Range NI Instrument failure. The reactor tripped approximately thirty seconds later on low pressurizer pressure with Channel 42 indication still pegged high, and the P-10 bistable tripped. The MPCS alarm summary was saved and ten-minute plots were made of core thermal - power, pressurizer pressure, and power range channels NI-42 and NI-43. Simulator response was in accord with the Westinghouse Functional Block
- Diagram, the Westinghouse technical manual for the Nuclear Instrumentation System (W-120-32),andtheSeabrookPrecautions, limitations,and Setpoints manual. There were no deficiencies noted and no exceptions are taken.-
4 146
MALFUNCTION #99 hr^' . POWER RANGE CHANNEL 43 F AIL $ HIGH
" Power Range Channel 43 fails High" was tested on December 15, 1988. This italfunction satisfies the AM51/AUS-3.5 requirement of Section 3.1.2(21),
Nuclear instrumentation failure.
.There is no severity option. The malfunction was tested from 16's reactor
_ power, BOL, control rods in manual . Once activateo, channel 43 indication , pegged high; and the high flux. trip, rod stop, and P8 bistables were activated. No rod motion occurred due to control rods.being in manual.
;The malfunction was removed approximately twenty minutes after activation, ;
and channel 43 indication was consistent with the other nuclear ' instrumentation channels. The reference procedure used was 051211.04, , Power Range N1 instrument failure. An MPCS primary calorimetric and the MPCS alarm summary were saved. Also, ten minute plots of channels 43 and
!44 were made. Simulator response was in accord with the-Westinghouse ,
7 Tuncticnal Block Diagram,-the Westinghouse Technical Manual for the NuclearInstrumentationSystem(W-120-32),andtheSeabrookPrecautions, limitations, and Setpoints manual. There were no-deficiencies noted, and no exceptions are taken, O 147 _ - . - . -- . . _ ~ _ -.
MALFUNCTION #100 O POWER RAr4GE CHANNEL 44 FAILS HIGH 15, 1988. This
Power Range Channel 44 Fails High" was tested on December rnal' unction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(21),
Nuclear instrumentation failure. There is no severity option. The malfunction was tested from 75% reactor Once activated, Channel 44 indication power, BOL, control rods in manual. No rod motion occurred due pegged high, and Channel 44 bistables tripped. The malfunction was removed approximately twenty to rods in manual. The minutes after activation, and the failed channel returned .o normal. reference procedure used was 051211.04, Power Range NI instrument Failure. An HPCS primary plant calorimetric and an HPCS alarm summary were saved. Also, ten-minute plots of core thermal power, pressurizer level, and power Simulator response was in accord with range Channels 43 and 44 were made. the Westinghouse Functional Block Diagram, the Westinghouse technical manualfortheNuclearInstrumentationSystem(W-120-32),andtheSeabrook There were no Precautions, Limitations, and Setpoints manual. hficiencies noted and no exceptions are taken. O 148
- - - - - - - _ _ _ ~ ~ ^ ^ ' ~ - - - - - - _ _ - _ __
MALFUNCTION #101 SOURCE RANGE CHANNEL 31 SLUGGISH RESPONSE
- Source Range Channel 31 Sluggish Response" was tested on December 14, 1988. This malfunction satisfies the ANS!/ANS 3.5 requirement of Section 3.1.2(21),Nuclearinstrumentationfailure.
The initial conditions for the test were 0% reactcr power, BOL, in the source range. The malfunction was activated, and a reactor startup was performed using Procedure 051000.07, Approach to Criticality. Source range channel 31 remained approximately one decade below source range channel 32. Power was driven above ni Delow the P-6Ynterlock to verify reenergizing of both source range cheisis. Source range channel 31 reenergized approximately one decade below source range 32. The malfunction was removed approximately forty minutes after activation and channel 31 reading matched channel 32. Several MPCS primary calorimetric printouts made during the startup and the MPCS alarm summary were saved. Also, ten-minute plots of several primary plant parameters and source
\
range channels 31 and 32 were made. Simulator response was in accord with the Westinghouse Functional Block Diagram, the Westinghouse Technical ManualfortheNuclearInstrumentationSystem(W-120-32),andtheSeabrook Precautions. Limitations, and Setpoints manual. There were no deficiencies noted and no exceptions are taken. O 149
.( l
~t MALFUNCTION #102 b LOSS OF POWER TO SOURCE RANGE 32 J
, " Loss of Power to Source Range 32" was tested on December 14, 1988. This , malfunction satisfies the ANS!/ANS 3.5 requirement of Section 3.1.2(21), Nuclear Hstrumentation failure. The initial conditions for the test were 0% reactor power, BOL, in the source range. The malfunction was activated, and the reactor immediately , tripped on source range high flux. Channel 32 indication read downscale . low, and the channel indicated deenergized on the NI cabinett. The malfunction was removed approximately twenty minutes after activation, and channel 32 returned to operation. The MPCS alarm summary was saved, and ten-minute plots were , made of core thermal power, intermediate range channel 35, and source range channels 31 and 32. From the plots, it was discovered that the "A" point for channel 32 remained "as is" even though channel 32 indication dropped to zero. This deficiency has been submitted p (SCR#90-176)forcorrectionnolaterthanJanuary 31, 1991. There were no other deficiencies noted and no exceptions are taken. n - . 150
,_ .~
(') V MALFUNCTION #103 POWER RANGE CHANNEL 41 UPPER DETECTOR FAILS LOW
" Power Range Channel 41 Upper Detector fails low" was tested on April 16, 1990. This malfunction satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(21), Nuclear instrumentation failure.
The initial condition 3 for the test were 100% reacter power, M0L. The malfunction was activated and NI-418 and power range recorder chat:nel PRI went to 50%. NI-41C(0 elta-l)peggedlow. On the NIS cabinet N t1A went to 50% and the " Negative Rate Trip Distable" was on. The other three power range detectors had normal readings. Reference procedure OS1211.04,
" Power Range NI instrument Failure" was used without deviation. "he malfunction was removed approximately thirty minutes after activation and PR-41 returned to normal. The MPCS alarm summary was saved. Simulator response was in accord with the Nuclear Instrumentation System lot;ic diagrams. No deficiencies were noted and no exceptions are taken.
O - 0 151
MALFUNCTION .104 POWER RANGE CHANNEL 42 UPPER DETECTOR FAILS LOW
" Power Range Channel 42 Upper Detector fails Low" was tested on December 15, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(21), Nuclear instrumentation failure.
The malfunction was tested at 4Sfs reactor power, BOL, Once activated, power range channel 42 indication and the axial flux difference reflected the failed detector. The malfunction was removed apptoximately fifteen minutes after activation and the failed charnel was restored. The MPCS axial flux difference report, primary plant calorimetric, and alarm summary were saved. Also, ten-minute plots were made for several primary plant parameters and power range channels 41 and 42. Simulator response was in accord with the Westinghouse Technical Manual for the Nuclear Instrumentation System. No deficiencies were noted and no exceptions are taken. O O 152
.. .. . - . . - - - ~ - . _ - . , . _ - . - . . . - - - _ . . . - . - - . . - .
5 MALFUNCTION #105
-s p0WER RANGE CHANNEL 43 UPPER DETECTOR FA.LS LOW d
" Power Range Channel 43 Upper Detector f ai)1 b w" was tested on December
- 15, 1988. This malfunction Atisfies the ANS'/ANS-3,5 requirement of Section 3.1.2(21), Nuclear istrumentation fi1Jre. j The malfunction was tested at 16% reactor power, BOL. Once activated, power range channel 43 indication and the axial flux difference reflected [ the failed detector. The malfunction was removed approximately fifteen ' minutes after activation arid the failed channel was restored. The MPCS . axial flux difference report, primary plant calorimetric, and clarm summary were saved, Ten-minute plots were made for the followings core ther.41 power, pressurizer level, and power range channels 43 and 44 Simulator response was in accord with the Westingnouse technical manual - for the Nuclear Instrumentation System. There were no deficiencies noted and no exceptions are taken. J i s t t u 153
. . -- _ . _ . _ . _ , ~ _. _ _ . _ _ . _ . _ . . _ . .- . .. . ~.
.k.
MAtrVHCTION /105
, POWER RANGE CHANNEL 44 09PEC DERCTOR FAILS LOW
-p
" Power Range Channel 44 Upper Detector Fails Low" was taited on Cecamber 15, 1988. This malfunction satisfies the ANSI /tNS-3.5 requirement of Section 3.1.2(21), Nuclear instrumentation failure.
T5e malfunct4on was tested at 10% reactor power, BOL, with control rods in auto. Once activated, power range channel 44 indication and the axial flux difference reflected the failed detector. The malfunction was re 'vved apprt>ximately fif teen minutes af ter activation and the f ailed i chainel was restored to ne' mal operation. The MPCS axial flux difference ter, ort, primary plant calorimetric, and alarm summary were saved. Ten-
.ivinute plots were mad 4 of the following: thermal power, pressurizer level, and power range channels NR-43 and NR-44. Simulator response tvas in accord with the Westinghouse technical manual for the Nuclear Instrumentation System. There were no deficiencies noted and no exceptions are taken.
h O 154 C ____._ .
.,___ - _ . _ . _ _ _ _ _ _ . ~ _ _ _ .
I MALFUNCTION #107~
/7, INTERMEDIATE RANGE-CHANNEL 35 OVER?OMPENSATED-Af In_termediate Range' Channel 35 Overcoipensated" was tested on December 14,~
1988, l his malfunction satisfies the ANSI /ANS-3.5 requirement of'Section-3.1.2(21),Nuclearinstrumentatiorfailure. L The initial-c'onditions fer the test were reactor startup in progress with counts in the source range, BOL. As soon as the-malfunction was activated, intermediate range channel 35 pegged low.- Power.was then raised by withdrawing control rods. Channel 35 indication remained low until just before channel 36 eached the P-6 interlock setpoint. As power-increased; the-two intermediate-range channels began' reading c' loser to.one another. The reactor-was then manually. tripped. . Channel 35 reached the P-6 reset point before channel 36 did and settled on tuttom scale. Channel 36 remained higher much longer. The malfunction was then removed
- an'd both intermediate-range channt.ls began reading approximately the same.-
Primary plant calorimetric data and the MPCS alarm summary were saved. D 1 Ten-minute plots were made of the following: _ intermediate range channels
'35 and 36, source range channel 31, and core thermal power.- Simulator response wa' -in accord with the Westinghouse functional block diaaram, the
- Sestinghouse technical mar # for the Nuclear Instrumentation System (W- 3 (120 .i2),s and- theLSeabrook Precautions, . Limitations, and Setpoints
- t. cument. There were no deficiencies noted and no exceptions are taken. 'l p
155
1 ;' MALFUNCTION #108
-D i ~1NTERMEDIATE RANGE CHANNEL 36 OVERCOMPENSATED-
" Intermediate Range Channel 36 Overcompensated" was tested on December 14, 1988. --This malfunction satisfies the_ ANSI /ANS-3.5 requirement of_Section j 3.1.2(21), Nuclear instrumentation failure.
The initial conditlons for the test were reactor startup in progress with a counts in the source range, BOL. As soon as the malfunction was-iactivatec,-intermediate range channel 36 indication pegged low. Reactor power was then increased byl withdrawing control rods. Just as channel 35 approached the P-6 interlock setpoint, channel 36 indication came-on
- scale.: The malfunction was then-deactivated and the two-intermediate
-range channels were reading approximately the_same. The malfunction was ;
then reactivated and the reactor _was manually tripped. Channel.36 began i reading lower than channel 35 as levels decreased. Channel 36 reached the ! P-63 interlock' reset point and finally pegged low at about the same time
= channel 35Lreached the P-6 interlock. reset point. _ Primary calorimetric i cdata and;the MPCS'ala'rm summary were. saved. Ten-minute plots were made of
- intermediate channels 35 and:36. Simulator response was in accord with '
the Westinghouse functional block diagram,'.the Westinghouse' technical-manualfonthe:duclearInstrumentation' System (W-120-32),andthe'Seabrook Precautions, Limitations, and Setpoints document.- There were no
= deficiencies-noted and-no exceptions are taken~.- .i q
J 156
I _ MALFUNCTION #109 INTERMEDIATE RANGE CHANNEL 35 UNDERCOMPENSATED
" Intermediate Rarge Channel 35 Undercompensated" was tested on Decemter 14, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirements of Section3.1.2(21), Nuclear.instrumentationfailure.-
The initial conditions for the test were reactor power in the intermediate range with control rods in manual. The malfunction was activated, and reactor power was decreased to below the P-6 interlock. However, source range high voltage did not energize due to channel 35 lagging behind channel 36. The malfunction was then deactivated. Channel-35 indication- , dropped due to the return of normal compensation, and source range high i voltage energized. Primary plant calorimetric data and the MPCS alarm ., summary were saved. Ten-minute plots of core thermal power, source range channel 31, and. intermediate range channels 35 and 36 we.re made. Simulator response was in accord with the Westinghouse technical manual ' for the Nuclear Instrumentation System (W-120-32), and the Precautions, limitations, and Setpoints document. There were no deficiencies noted and no exceptions are taken. 157 t
("') b/ MALFUNCTION #110 INTERMEDIATE RANGE CHANNEL 36 UNDERCOMPENSATED
Intermediate Range Channel 36 Undercompensated" was tested on December 14, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section3.1.2(21),Nuclearinstrumentationfailure.
The initial conditions for the test were reactor power in the intermediate range with control rods in manual. The malfunction was activated and reactor power was increased and then lowered. Channel 36 indication remained higher than channel 35 through the various power levels with the effect being more dominant at lower power. The malfunction was deactisated after approximately twemty minutes and the two intermediate range caannels began reading the same, Primary plant calorimetric data and the MPCS alarm summary were saved. Ten-minute plots were made of source range channel 31 and intermediate range channels 35 and 36.
-Simulator response was in accord with the Westinghouse technical manual q fortheNuclearInstrumentationSystem(W-120-32). There were no V deficiencies noted and no exceptions are taken.
} 1.'3
r MALFUNCT10N #111-
-SOURCE RANGE LCHANNEL 31 FAILS TO DEENERGlZE
" Source Range Channel 31 Fails to'Deenergize" was tested on December 14,=
1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section1-
--3.1.P.(21),,Nuclearinstrumentationfailure.
The' initial donditions'for the test were= reactor power in the_ intermediate range!above.thePh6Linterlock. The malfunction was-activated.and,..as-- expected, no effects were observed. The simulator-was then reset to
- l reactor' power'in the source range ~below the P-6 interlock. The
- malfunction was;again activated and reactor power was increased to above 1 the_P-6 permissive setpoint-by withdrawing control rods. Source range: ,
9 channel 31lwould notfdeenergize allowing the plant to trip on.high source-
- range flux ~. . The MPCS alarm summary was saved. Ten-minute plots of core
-thermal-power,-intermediate range channel 35,Jand source range channels 31 and'32 were made. Simulatcr response _was-in accord with the-Seabrook-
- Station Precautions, Limitat. ions, and Set' points document,:and the
-Westinghousetechnicalmanualfor-theNuclear1nstrumentation-System-(W-
"120-32)s- There were nocdeficiencies noted and-no exceptions _are taken.-
'N _.
2 l
;}
. _[
I; 4 159 l m
(~) - MALFUNCTION #112
.( f INTERMEDIATE RANGE CHANNEL 35 LOSS OF DETECTOR VOLTAGE
" Intermediate Range 35 Less of Detector Voltage" was tested on December 14, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section3,1.2(21),Nuclearinstrumentationfailure.
The malfunction was activated from 3% reactor power, BOL. Intermediate range channel 35 indication ' ailed low, and SUR for channel 35 pegged downscale and then returned to zero. No other effects on plant operation were observed. The malfunction was deactivated and the intermediate range. channel indications were the same value. The malfunction was then reactivated and the reactor tripped to verify that both source range channels energized at the P-6 setpoint, which they did. Primary calorimetric data and the MPCS alarm summary were saved. Ten-minute plots of core thermal power, source range channel 31, and intermediate range channels 35 and 36 were made. Simulator response was in accord with the g (^ k Seabrook Precautions, Limitations, and Setpoints document and the Westinghouse technical manual for the Nuclear Instrumentation System (W-120-32). There were no deficiencies noted and no exceptions are taken - O 160
~
I i 12"'s; ,
' 1'"'L MALFUNCTION'#113 - .
~!
- INTERMEDIATE-RANGE CHANNEL 36 LOSS OF DETECTOR VOLTAGE
" Intermediate Range Channel-36 Loss of Detector Voltage" was-tested on December 14, 3988.- This malfunction _ satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(21), Nuclear instrumentation. failure. '
~
The malfunction was actrotted with reactor power.in the source _ range, BOL. Reactor power was increased by withdrawing control rods above the P : interlock setpoint.- Intermediate range channe1~36 indication pegged low and remained there,-and:SUR for channel 36 pegged downscale and returned-
;to!zero. The malfunction was deactivated-and the two-intermediate range-channels began reading the same. Primary plant calorimetric data and the-
?MPCS alarm summary' were saved.: Ten-minute plots of core thermal power,--
source range channel 31 1, and = intermediate range channels 35 and 36 :were:
~~
made._ Simulator response wasiin accord with the Seabrook Precautions, ,
' Limitations,- and Setpoints document and the Westinghouse _technica'l manual
~
)k - for the' Nuclear Instrumentation. Systemt(W-120-32). There were no-
- deficiencies noted and no: exceptions are taken.
i
?
b I t i I
- , 161 4
9- MALFUNCTION #114 TOTAL LOSS OF 0FFSITE POWER
" Total Loss of Offsite Power" was tested on February 22, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(3),
Loss or degraded electrical power to the station including loss of offsite power. The malfunction was activated from 100% Rx power, MOL. Once activated, all offsite buses deenergized, and buses E-5 and E-6 reenergized powered by the emergency diesel generators. The malfunction was removed approximately 35 minutes later, and buses 1, 2, 3, and 4 reenergized. The MPCS alarm summary and the "A" point trend for major electrical parameters were saved. Some deficiencies were noted. The air removal pump status logic and dynamic parameters did not respond correctly. Thisproblemwassubmitted(SCR#89-057)andhasbeen corrected. Also, the containment cooling units did not reenergize when bus power was restored. SCR (#90-099) has been submitted for correction by no later than 1/31/91. The secondary component cooling water pump S status indication logic and dynamic parameters did not respond correctly. SCR (#89-059) was submitted and has been corrected. Control room lighting does not respond as in the plant once power is restored. There was an outstanding SCR (#86-161) to correct this problem at the time of the test, and has been completed. Other than these deficiencies, simulator response was in acccrd with the Seabrook Station electrical diagrams. There are no exceptions taken. O 162
, MALFUNCTION #115 r -
LOSS OF UAT FEED TO 13.8 KV BUS 1
" Loss of UAT Feed to 13.8 KV Bus 1" was tested on February 21, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(3),
Loss or degraded electrical power to the station including loss of electrical power to the plant's electrical distribution buses. The malfunction was tested from 75% reactor power, MOL. When activated, the UAT breaker to bus 1 opened, and the RAT breaker to bus I closed. A failed attempt to reclose the UAT breaker was made. The malfunction was deactivated and the VAT breaker to bus 1 was returned to service. The RAT to bus I was placed in pull-to-lock, and the malfunction reactivated to verify a sustained loss of bus 1. The MPCS alarm summary and "A" point trend for 13.8 KV parameters were saved. Two deficiencies were noted. The air removal and secondary component cooling water pump logic (as noted in Malfunction #114) did not respond correctly. Otherwise, simulator response was in accord with Seabrook Station electrical diagrams. No I exceptions are taken. 4 o i 163
A MALFUNCTION #116 - (. ) LOSS OF RAT FEED TO 4.16 KV BUS E-6
" Loss of RAT Feed to 4.16 KV Bus E-6" was tested on February 22, 1989.
This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (3), Loss or degraded electrical power to the station including loss of , electrical power to the plant's electrical distribution buses. The malfunction wa; tested from 100% reactor power, MOL with bus E-6 receiving power from the RAT. Once activated, all RAT power was lost due to a fault in the 34S KV switchyard. The emergency diesel generator started and reenergized bus E-6. Although UAT power is available, there is no automatic switchover to the UAT. Removing the malfunction allowed shutting the RAT feed breaker to bus E-6. The MPCS alarm summary and "A" pointtrendformajorrelat'edelectricalparametersweresaved. Simulator
~
response was in accord with the Seabrook Station electrical diagrams. No deficiencies noted and no exceptions are taken. IT .L) p. V 164
i
-MALFUNCTION #117:
LOSS OF:125 VDC BUS 11A
. -'" Loss of?l25 VDC= Bus-11A" was tested on February 23, 1989. This
-malfunction ~ satisfies the ANSI /ANS-3.5 requirement of-Section 3.1.2(3),
Loss of degraded ~ electrical power to the station, including loss of' electrical distribution buses:and loss of power to the~ individual
, instrumentation _ buses (ACaswellLasDC)thatprovidepowerto_ control room indication or plant control functions affecting the plant's' response. ;
1The malfunction'was--tested from 100% reactor-power, MOL, Once activated, Lbusl11A deenergized which resulted in a reactor-turbi_ne trip. 'The reactor-
- trip was/due'-to: low steam generator level caused by-the feedwater regulating' valves going shut on loss of power. Safety _ injection:was l manually initiated to verify.that "A"' emergency diesel generator would not start. Aniattempt was-made to stop the running charging pump from the-MCB' .q Lwith no(effect ~The malfunction'was removed approximately one hour _and i L-fifty' minutes later allowing power restoration to bus 11A. The MPC alarm 4
e summa ard~the " po t end record majo 2 DC_ rela pa te s-e ideenergize.--'IncludedwereEAH-5A'(SCRif89_056)-whichhasbeencorrected, I remergenc'ydiesel. generator /servicewater,coolingvalues(SCR#89-055) m Ewh_ich was completed on 8/6/90, land the undervoltage and underfrequency 4 reactoritripirelays(SCRif89-054)isubmittedfor-.correctionnoLlaterthan-- d
- 6/30/91.?Otherwise,$.simulatorresponse-was-inaccordLwith'Seabrook- [
LStation electrical. diagrams. ;There are no exceptions takent j q
.i 165' l
_M
' MALFUNCTION 8118 o FAILURE OF "A" EMERGENCY DIESEL GENERATOR TO AUTO START
]
" Failure of 'A' Emergency Diesel Generator to Auto Start" was tested on February-22, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(3), Loss of degraded electrical power to the Station including loss of emergency power and loss of emergency generators, and Section 3.1.2(23), Passive malfunctions.
The malfunction was tested from 100% reactor power, MOL. .The malfunction was activated followed by the activation of Malfunction #114, " Total loss of Offsite Power". Emergency diesel generator 18 started and 1A did not. Safety injection was manually. initiated and emergency diesel generator IA did not start. It was then started manually. The MPCS alarm summary was saved. Simulator response was in accord with the Seabrook Station electrical diagrams. There were no deficiencies noted and no excetations are taken. A-n 166
C MALFUNCTION #119 ()T . LOSS OF EMERGENCY DIESEL GENERATOR IB
" Loss of Emergency Diesel Generator 1B" was tested on February 22, 1989.
This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (3), Loss or' degraded electrical power to the station, including loss of emergency generators. The malfunction was tested from 100% reactor power, MOL. The malfunction
-was activated and safety injection initiated. Emergency diesel generator 18 started and tripped on low lube oil. The MPCS alarm summary and the "A" point trend-for major electrical parameters were saved. Simulator i response was in~ accord with the Seabrook Station Electrical Diagrams.and ,
the UE'& C Design Description 50-10, Diesel Generator Mechanical Systems. -l There were no deficiencies noted, and no exceptions are taken.
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I [T L) 167
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MALFUNCTION #120
't) LOSS OF MCC 231 .
" Loss of MCC 231" was tested on February 28, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (3), Loss or degraded electrical power to the station including loss of power to the plant's electrical distribution buses.
The malfunction was tested from 100% reactor power, MOL with "A" and "C" reactor coolant pump oil lift pumps running. The malfunction was activated; and "C" reactor coolant pump oil lift pump stopped, "A" continued running, and indication for "0" was lost. Indication for the
-steam generator b!owdown motor operated valves SGBD-V-189 and 195 was also lost. The malfunction was removed approximately ten minutes later, and the lift-pumps reenergized and indication was-restored to all components.
The MPCS alarm summary was saved and simulator response was in accord with the Seabrook Station electrical diagrams. There were no deficiencies noted and no exceptions are taken. I m
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168
Ii MALFUNCTION #121 (~ / LOSS OF MOTOR CONTROL CENTER 111
" Loss of Motor Control Center 111" was tested on February 23, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(3),
loss or degraded electrical power to the station including loss of power , to the plant's electrical distribution system. The malfunction was tested from 100% reactor power, MOL. The malfunction was inserted resulting in a loss of power and indication to the "A" and "B" reactor coolant pump lift oil pumps and "B" and "C" steam generator blowdown valves SGBD-y-191 and SGBD-V-193. The malfunction was' removed approximately ten minutes later and power was restored to the lift pumps and steam generator blowdown valves. Simulator response was in accord with the Seabrook Station electrical diagrams. There were no deficiencies noted and no exceptions are taken, gp
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169
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MALFUNCTION #122 1 3 HIGH TURBINE VIBRATION s
"_High Turbine' Vibration" was tested on March 6, 1989. This malfunction satisfiestheANSI/ANS-3.5,Section3.1.2(15),Turbinetrip. '
ThereLisTa severity' option where 0-100% severity is equal to 0-18 mils. ;
-It was' activated from '50% reactor power, BOL, at _45% severity (appr'oximately:10 mils)rampedovertenminutes. Once activated, turbine bearing #6 and the-adjacent bearing vibration began to increase.until the turbine-tripped on high vibration. It was noted-thatithe turbine trip and
--th'e turbine _high' vibration alarm did not occur at the correct values. An SCR-(#89-065)'was:submittedandcorrectedonJune 28, 1989. Reference t -Procedure ON1231~.01,JTurbine Generator High Vibration, was used with no' .;
simulator limitations- The simulator was then reset:to 50% reactor power,- -
; BOL,- and the; m'alfunction- reactivated at 30% severity. The malfunction'was 1
-removed approximately five minutes later to verify that turbine-vibtstion returned to'the initial value.
~
p The;MPCS alarm summary and the "A" point.
- C : trend. record for the turbine bearing vibration values were: saved. Ten- .i
- minute-plots of.-core thermal power and: generator power were made. All other simulator response wasiin accord.with the Seabrook Station -
Precautions,-? Limitations, and Setpoint= document'and the G.E. Turbine . Generator manual. NoLexceptionstare-taken. l 4 l
-l 1
( 170 j 1 i l a
' () - MALFUNCTION #123-() LOSS OF AUTOMATIC LOAD CONTROL UNIT
" Loss of Automatic Load Control Unit" was tested on December 15, 1988.
This malfunction satisfies the ANS!/ANS-3.5 requirement of Section 3.1.2 (17), Failure in-automatic control systems that affect reactivity and core heat removal. The malfunction was tested from 100% reactor power, MOL. The malfunction was activated and the turbine load decrease button was depressed. Indicated load ran down to 1000 MW, however, actual load did not decrease (primaryandsecondaryparametersdidnotchange). The same test was performed in the increase load direction with the same results. The generator stator cooling pump was then tripped. Normally, this would result in a turbine setback, but the malfunction prevents the setback from-occurring. The malfunction was removed and the turbine shed load as expected.- The simulator was reset to 100% reactor power, MOL, and "A" -j
, w) - main feedwater pump was manually tripped to verify a turbine setback would i
(/ occur. -The simulator was reset to 100% reactor power, MOL, and the-malfunction was again activated. Malfunction #99 was used to fail high the power range channel 43. Control rods began driving in, but no turbine load was shed. Standby turbine control was then verified operational. The MPCS alarm summary was saved, and ten-minute plots were made of core , thermal' power, generator power, and steam _ generator A level and steam generator pressure. Simulator response was in accord with the Seabrook
-Station Precautions, limitations, and Setpoints documant and the G. E.
-Turbine Generator manual. There were no response discrepancies noted and no exceptions are taken.
O O 171
I i l O MALFUNCTION #124 V- LOSS OF AUTOMATIC SPEED CONTROL UNIT
" Loss of Automatic Speed Control Unit" was tested on April 23, 1990 from 14% power, turbine just placed on line. This malfunction satisfies the ANSI /ANS 3.5 requirement of section 3.1.2(23) Passive malfunctions.
Once activated the turbine would not respond to an increase in speed demand. The malfunction was deactivated ten minutes after activation. The speed control unit then responded normally. Simulator response was in accord with the turbine generator vendor manual. No deficiencies were noted and no exceptions are taken. .O I l 172 l
' [T MALFUNCTION #125
'V AUTOMATIC VOLTAGE REGULATOR FAILURE
" Automatic Voltage Regulator failure" was tested on December 16. 1988. It was tested from 100% reactor power, MOL. The cause of the malfunction (blown fuse-in the sensing circuit) causes the voltage r Mulator to shift to manual.- The auto and manual voltage adjustments were set to a 2.5 volt mismatch with auto being higher than manual. The malfunction was then activated. Turbine generator output voltage, MVARs, and generator excitation amps all assumed the voltage set by the manual voltage regulator. The malfunction was removed approximately twenty minutes later, and the generator switched back to automatic voltage regulation.
The MPCS alarm summary and the "A" point trend for generator related - parameters were. saved. Simulator response was in accord with the G. E. Tnrbir.a Generator-manual. -No deficiencies were noted and no exceptions are taken. I j'*%
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{ 173
N (j ' MALFUNCTION #126 MAIN GENERATOR SEAL OIL SYSTEM FAILURE 4,
" Main Generator Seal Oil System Failure" was tested on December 16, 1988.
4 The malfunction was tested from 100% reactor power, MOL. Once activated, both the main and emergency seal oil pumps fail. The hydrogen pressure dropped to 25 psig (at which point the main lube oil system begins providing-seal oil pressure), and hydrogen temperature increased slightly. The malfunction was removed and remote functions were used to restart the seal oil pump and restore hydrogen pressure to the generator. Reference
' Procedures 0N1031.02, Starting and Phasing the Turbine, and ON1039.04, ,
Startup and Shutdown of the Shaft Seal Oil System,,were verified to be usable on the simulator. The MPCS alarm summary and the "A" point trend for related parameters were saved. Simulator response was in accord with the G.E. Turbine Generator manual. No deficiencies were noted and no ,S , exceptions are taken. L) b 174
I'T - MALFUNCTION #127 k) - LOSS OF ELECTR0-HYDRAULIC SUPPLY PUMP "A"
" loss of the Electro Hydraulic Supply-Pump" was tested on December 16, 1988.
The malfunction was tested from 100% reactor power, MOL, with the "A" EHC pump _in service and the "B" EHC pump in standby. The malfunction was activated, and the "A" electro-hydraulic supply pump tripped. -EHC pressure dropped slowly, and at 1300 psig, the standby pump started and restored pressure. -The malfunction was removed, and the "A" pump restarted. The MPCS alarm summary was saved. Simulator response was in accord with the pump logic diagrams. No deficiencies were noted and no exceptions were taken.
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v 175
'w MALFUNCTION #128
[f 1._, LOSS OF ELECTRO-HYDRAULIC SUPPLY PUMP "B"
" Loss of-Electro-Hydraulic Supply Pump 'B'" was tested on December 16,
-1988.
The malfunction was tested from 100% reactor power, MOL, with "B" EHC pump in service and "A" EHC pump in stanoby. The malfunction was activated, ' and the "B" EHC pump tripped. EHC pressure dropped slowly and at 1300 psig the standby pump started and restored pressure. The malfunction was removed, and the "B" pump restarted. The MPCS' alarm summary was saved. Simulator response was-in accord with the EHC logic diagrams. No deficiencies were noted and no exceptions are taken. (3, ,
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176
O MALFUNCTION #129 MAIN TURNING GEAR TRIP
" Main Turning Gear Trip" was tested on March 7, 1989.
The malfunction was tested from 15% reactor power, MOL, with the turbine generator removed from service and the turning gear engaged. The malfunction was activated and the turning gear tripped and turning gear motor amps went to zero. The malf' nction was removed and the turning gear restarted. Eccentricity readings had increased due to the turbine being hot and off the turning gear. The MPCS alarm summary and the-MPCS "A" point trend for turbine eccentricity and vibration readings were saved. Simulator response was in accord with the main turbine turning gear electrical schematics and the G. E. Turbine Generator manual. No deficiencies were noted and no exceptions are taken. O m
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MALFUNCTION 1130
- PARTIAL' LOAD REJECTION 4 ,
i
" Partial Load' Rejection" was tested on December 15, 1988. This 41 malfunctionsatisfiestheANSI/ANS3.5requirementof-Section3.1.2(17); -l
, Failure in automatic control system (s) that affect' reactivity.and core- ,
' heat: removal.- '
1 The ' malfunction 'wasitested -from'100%' reactor power, BOL, with control rods: fin automatic. The malfunction.was activated, and an approximate 60%: load: rejection 1 occurred. settling outEat approximately 450 MW. The reference. 4 : procedurelused was OS12319 03,; Turbine Runback / Setback. /The: primary and
- secondary plant responded'as expected 1to the decrease in load ~ demand =. a This was_ verified'by comparing;the primary and secondary parameter strip ch'rt(recordings,"A"and"C"pointtrends,theMPCSalarmsummary,and' a
- ten-minute plots to the-projected 1results in the Scabrook, Station.FSAR-and M
;theresultsofjtheMillstone13 power!ascensiontest(large'loadreduction g f( s tr'ansient).- There-were:no: deficiencies noted and no' exception:'are-taken.
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*I 178 or C, .- , _. . _ . . . . . ._ , - , . .
/~'j MALFUNCTION #131
(_ '1 FAILURE OF HOTWELL LEVEL TRANSMITTER
" Failure of Hotwell Level Transmitter" was tested on March 6, 1989. This malfunction satisfies the ANSI /ANS 3.5 re.tuirement of Section 3.1.2(22),
Process instrumentation alarms and control system failure. Initial conditions for the test were 100% reactor power, MOL. Once activated, level control valve LCV-4014A went open responding to a faulty s,ignal from LT-4014. Condenser level increased and condensate storage tank level decreased. The failed transmitter was then deselected and the spill valve opened returning hotwell level to normal. The MPCS alarm summary was saved. . Simulator response was in accord with the Precautions, ; Limitations, and Setpoints manual and the Seabrook Tank Volume Curves. No ' deficiencies were noted and no exceptions-are taken. (
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o a 179
9 MALFUNCTION #132 STEAM GENERATOR "C" LEVEL CHANNEL L1-537 FAILS LOW .-
.m
" Steam Generator "C" Level Channel LT-537 Fails Low" was tested on January 10, 1989. This malfunction satisfies the ANSI /ANS-3.5 requiruent t,f Section 3.1.2(11), Loss of protective system channel, and section 1
3.1.2(22), Process instrumentation, alarms, and control system failure. The malfunction has no severity option and was tested at 100% Rx power, MOL. The malfunction was active for approximately fifteen minutes. The level channel immediately failed low on the MCB indicetion. The MPCS "A" I point graphic display, however, did not change, and this was noted as a discrepancy. Because channel LT-537 does not have any control besides a , bistable trip input to a two-out-of-four logic stheme, and an SG lo 'o level alarm, there is no other effect on plant operatien. Operational a Procedure 0S1201.06, PZR Pressure Instrument Failure, could be used { without deviation. When the malfunction was removed t.. LT-537 MCB , indication returned to normal. Records of MPCS trends for SG "C" NR and WR level as well as the alarm summary were savei The simulator response, except for the noted discrepancy, is in accord with the NHY Precautions, ' Limitations, and Setpoints document provided by Westir,ghouse and the Solid State Protection System electrical schematics. AnSCRwassubmitted(#89-030) to correct the deficiencies and was completed on February 9, 1990. There are no exceptions taken. O 1 180
l l
!7 MALFUNCTION #133 y V- STEAM GENERATOR "D" LEVEL CHANNEL LT-549 FAILS LOW :
j
" Steam Generator "D" Level Channel LT-549 Fails Low" was tested on January ;
10, 1989.- This malfunction satisfies the ANSI /ANS-3.5 requiremert of I section.3.1.2(11) Loss of protective system channel, and Section 3.1.2 ; (22), Process instrumentation, alarms, and control system failure, q f The malfunction has no severity. option and w s 'ested at 100% Rx power, ! MOL. The malfunction was tested twiur, During te first test no operator action was taken. - Onca the malfunction .i y inse' ced, LT-549 MCB indication immediately dropped low, and because LT-549 is the feed flow controlling channel, feed flow began increasing. Subsequently, SG "D" level, increased causing 'a-turbit e trip and reactor trip. During the second test, operator action was taken to recover SG "0" level using-Abnormal Operating Procedere 0S1235.03, SG Level Instrument failure, no Rx trip occurred.- Once the malfunction was removed, the level channel 3n returned to service, the MCB indication was restored and feed flow responded to the proper SG 1evel input. During both tests a discrepancy was noted. The MPCS graphic display for the LT-549 A-point did not change when the malfunction was active. An SCR-has been submitted (89-030), and the deficiency was corrected on February 9, 1990. MPCS chart recordings fcr SG NR le.e and feed flow, the SG "0" level channel A point trend and the MPCS alarm summary were saved. The response of the-simulator was
. consistent with the NHY Precautions, Limitations, and Setpoints document and the Solid State Proteccion System electrical schematics. No
, exceptions are taken. ( 181
Y (d MALFUNCTION #134 FAILURE OF MAIN TURBINE PROTECTION SYSTEM TO AUTO TRIP THE TURBINE
" Failure of the Main Turbine Protection System to Auto Trip the. Turbine" ,
was tested on November 11, 1939. This malfunction satisfies the ANSI /ANS , 3.S requirement of Section 3.1.2(22), Process instrumentation, alarms, and l control system failures, Section 3.1.2(11). Los3 of protective system . channel, and Section 3.1.2(23)-Passive malfunctions. The malfunction was activated from 100% Rx power, MOL, and was active for approximately thirty minutes. The malfunction was activated, anci all actions taken resulted in no turbine trip. Malfunction #122, high-turbine
-vibration (15. mils),wasactivatedanddeactivated. SG "C" feed regulating valve was opened causing SG "C" level to increase resulting in a FW isolation signal. The reactor trippad on SG lo-lo level and SI was then initiated. Finally, the turbine did trip on a manual trip.
Reference Procedure E-t was used. Simulator response was consistent with _o; the NHY' Precautions, Limitations, and Setpoints document and the Solid (_) State Protection System electrical schematics with the exception of one noted deficiency. Once the-turbine is tripped manually, the automatic actions that should occur (i.e., turbine drains opening extraction steam isolation,etc.)-do-not. An SCR (#89-118) was written, and was completed on_ February 8,'1990. 'There are no exceptions taken. A L) 182 l
MALFUNCTION #135 V S/G "B" LE C TRANSMITTER LT 529 FAILS HIGH "S/G 'B Level Transmitter LT-529 Fails High" was tested on January 11, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2 (22), Process instruma tation alarms and control system failure, and Section 3.1.2(11), Loss of protective system channel. The malfunction was activated at 100% Rx power, MOL, and the duration of the test was approximately forty five minutes. As soon as the malfunction was inserted, MCB indication failed high, and because LT 529 was the SG level controlling channel, feed flow decreased. Subsequently, SG "B" level began decreasing. Operator action was taken in accord with Procedure 051235.03, "SG Level Instrument failure", to remove SG *B" level which includes switching controlling channels. Oce SG "B" level was restored to normal, ;he malfunction was removed and LT-529 was returned as the controlling channel. MCB indication and FW control returned to o normal. The malfunction was then inserted again, however, no operator U action was taken. The Rx tripped on SG 10-10 level. MPCS SG level "A" point trends, and the MPCS alem summary were saved. The simulator response was consistent witt i information in the NHY Precautions, Limitations, and Setpoints de aments and the Solid State Protection System electrical schematics with the exception of one noted discrepancy. When the malfunction was active, the HPCS graphic display of the associated A point continued to read correctly. AnSCRhasbeensubmitted(#89-030) and was corrected on February 9, 1990. There are no exceptions taken. O V 183
MALFUNCTION #136
.s f RCS W10E kANCf: PRESSURE CHANNEL PT-403 FAILS TO 1000 PSIG "PT-403 Fails to 1000 psig" was tested on May 7, 1990. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(22), Process ,
instrumentation, alarms and control system failures. The initial conditions for the test were 100% reactor, BOL. Once activated RCS wide range pressure channel PT-403 went to 1000 psig. PT-403 did not respond ;o changes in RCS pressure. Pressure changes were accomplished by placing spray in manual and activating the pressurizer heaters, and then shuttin.j off the heaters and putting spray back into auto. 'The malfunction was removed approximately twenty minutes after activation and PT-403 returned to a normal reading. An "A" point trend of Loop 1 and Loop 4 pressure was saved. Simulator response was in accord with the RCS P&l0's. No deficiencies were noted and no exceptions were taken. J 4 O 184 li i ii
MALFUNCTION #137 c RCS LOOP A NR TC FAILS HIGH
RCS Loop A NR Tc rails High" was tested on January 11, 1989. This malf unction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (22)
Process instrumentation, alarms, and control system failure. It was activated at 100% Rx power, MOL, and the duration of the test was l approximatelv thirty minutes. Once the malfunction was inserted, MCB indication for f ailed high, and Loop 1 T,,, indicated 630'F. This caused a T u ,/T,,, mismatch which in turn caused the control rods to begin moving in. OT Delta-T_and Delta T indications also reflected the faulty input signal. Channel 1 was defeated in accord with Procedure 051201.08, T,,,/ Delta-T Instrument failure. Control rods responded to the correct T,,, / T,,e ratio. The malfunction was removed and all MCB indications returned to normal. Ten-minute plots were made of cold and hot leg
. temperatures. An MPCS trend record of Loop 1 NR Tc, Delta-T, OP Delta-7, OT Delta-T, and T,,,; and the MPCS alarm summary were saved. The simulator response is in accord with the NHY Precautions, Limitations, and Setpoints document provided by Westinghouse and the Solid State Protection System electrical schematics. There were no deficiencies noted and no exceptions are taken.
A 4 185
r MALFUNCTION #138 t (V3 PRESSURIZER LEVEL CHANNEL LT-459 FAILS LW ;
" Pressurizer Level Channel LT-459 Fails Low" was tested on January 10, 1989. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(22),Processinstrumentationalarmsandcontrolsystemfailure,and Section3.1.2(11),Lossofprotectionsystemchannel.
The malfunction was activated at 100% reactor power, MOL. Once activated, MCB indication failed to zero, letdown isolated, the pressurizer heaters cutoff, and charging was reduced. in accord with Procedure 051201.07, Pressurizer Level Channel Failure, the failed channel was deselected, letdown restored..and pressurizer heaters came on. The malfunction was ' removed approximately thirty minutes after activation, and the channel returned to operation. Ten-minute plots of pressurizer level and pressure were made. MPCS A point trends for pressurizer level (three channels), I charging flow, letdown flow, and the MPCS alarm summary were saved. One deficiency was noted. The MPCS "A" point graphic display for LT 459 did not fail. AnSCRhas"beensubmitted(f89-030)andwascorrectedon February 9, 1990. With the exception of this deficiency, simulator s response was in accord with the NHY Precautions, Limitations, and
-Setpoints manual provided by Westinghouse and the Solid State Protection
. System electrical schematics. There are no exceptions taken.
LO 186
/7 MALFUNCTION #139 C/ PRESSURIZER LEVEL CHANNEL LT-549 FAILS HIGH ,
" Pressurizer Level Channel LT-549 Fails High" was tested on January 10, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(22), Process instrumentation alarms and control system failure, and Section 3.1.2(11), Loss of protection system channel.
The malfunction was activated at 100% reactor power, MOL. Once activated, MCB indication failed high, charging decreased and the reactor tripped on high pressurizer level. The trip caused actual pressurizer level to begin decreasing. In accord with Procedure 051201.07, Pressurizer Level Channel failure, the failed channel was deselected. The pressurizer high level reactor trip bistable cleared and charging returned to normal. The malfunction was removed approximately thirty minutes after activation and the MCB indication was restored. Ten-minute plots of pressurizer level and pressure were made. MPCS "A" point trends for pressurizer level (threechannels),chargingpumpflowandletdownflowandtheHPCSalarm , summary were saved. Simulator response was in accord with the NHY Precautions, limitations, and Setpoints document provided by Westinghouse and the Solid State Protection System electrical scheniatics with the { exception of one noted deficiency. The HPCS "A" point graphic display for LT-459 did not fail. AnSCRwaswritten(89-030)andcorrectedon February 9, 1990. There are no exceptions taken. 187
i MALFUNCTION #140 O TURBINE IMPULSE PRESSURE PT-505 FAILS LOW
" Turbine Impulse Pressure PT-505 Fails Low" was tested on January 12, 1989. This malfunction satisiies the ANSl/ANS-3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms, and control system failures.
The malfunction was activated at 100% reactor power, MOL, with control rods in automatic. There is no severity option. Once it was activated, the MCB indication failed low, and the control rods began moving in due to the faulty input to T,,,. Control rods were placed in manual and MCB indications verified. Control rods were returned to automatic. Corrective action was taken in accord with Procedure 0S1035.05, Turbine Impulse Pressure PT-505, 506 Instrument Failure. The malfunction was removed in approximately thirty minutes. The MCB indication returned to actual, and the control rods responded to a T,,, which was now reflective of turbine load. MPCS A point trends for T,,, and turbine impulse pressure and the MPCS alarm summary were saved. Simulator response was in accord O w4th 18e NNv ereceetiees, t4 41et40es, emd SetPe4ets eecemeet Prev 4eee 8, Westinghouse and the Solid State Protection System electrical schematics. There were no deficiencies found and no exceptions are taken. O 188
(' MALFUNCTION #141 REACTOP. COOLANT LOOP NRgT FAILS LOW
" Reactor Coolant Loop NRgT fails Low" was tested on January 11, 1989.
This malfunction satisfies the ANSI /ANS-3.5 requirement of, Section 3.1.2 (22), Process instrumentation, alarms, and control system failures. There is no severity option. i The malfunction was tested from 100% reactor power, MOL; and the duration of the test was approximately thirty minutes. Once the malfunction was , inserted, Loop C T,y,, Delta-T, and 0T Delta-T responded to the failed input channel. Bistables for f*edwater isolation, C-16 interlock (this interlock prevents turbine loading and defeats remote dispatching), and steam dump block (all Channel 3) lit. Malfunction Procedure 051201.08, T,y,/DT Instrument Failure, was used to recover the plant. The malfunction was removed, and the failed channel and all affected systems returned to normal. Ten-minute plots of ger m tor power, core thermal power, Loop 1 and 3 T,y,, and hot leg temperatures were made. MPCS "A" point trends for Loop 3 Delta-T, NR cold leg temp, OP Delta-T, OT Delta-T, and T,y,, and-the MPCS alarm summary were saved. Simulator response is in accord with the Precautions, Limitations, and Setpoints document provided by Westinghouse and the Solid State Protection System electrical schematics. There were no deficiencies found, and no exceptions are taken. l A V 189
[] C/ MALFUNCTION #142 MAIN STEAM HEADER PRESSURE INSTRUMENT PT-507 FAILS HIGH
" Main Steam Header Pressure Instrument PT-507 Fails High" was tested on February 21, 1989. This malfunction satisfies the ANSl/ANS-3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms, and control aystem failures. There is no severity option.
The malfunction was tested from 100% reactor power, MOL, with steam dumps in the T ,, mode. Once the malfunction was activated, MCB indication immediately failed high, and feedwater flow began increasing until the feedwater regulating valves closed down to recover SG level. Steam dump control was then placed in steam pressure mode, and the steam dumps opened fully. Steam dump control was returned to T ,, mode, and the steam dumps closed. Procedure ON1230.01, Steam Header Pressure PT-507 Instrument failure, was used to recover the plant. Ten-minute plots of several key dynamic parameters were made, and the MPCS alarm sr,mmary was saved.
,q Simulator response was in accord with the NHY Prer.autions, Limitations, U and Setpoints document provided by Westinghouse and the Solid State Protection System electrical schenatics. There were no deficiencies found and no excentions are taken.
O LJ 190 t .
(9 V MALFL OTION #143 ' LOSS OF FIRE PROTEL '10N SYSTEM JOCKEY PUMP
- Loss of Fire Protection Sjstem Jockey Pump" was tested on March 6. 1989.
This malfunction simulates a rupture in the common discharge of the jockey pump. It was activated from 1004 Reactor, MOL. Once activated, the fire main pressure dropped below 100 psig and the main fire pumps started. The reference procedure used was 050243.02, Fire Main Break. The procedure calls for the rupture to be isolated. This is accomplished by removing the malfunction and allowing fire main pressure to be restnred and the main fire pumps to stop. Simulator response was in accord with the Fire Water System P&l0s and Fire Pump logic diagrams, No deficiencies were noted and no exceptions are taken.
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P e O 191
MALFUNCTION #144 0
-) FAILURE TO THE WASTE GAS 0UTLET RADIATION MONITOR l
The " Power failure to the Waste Gas Outlet Monitor" was tested on March 1 - 1990, from 100% Reactor Power, BOL. As per the Radiation Monitoring System logic diagrams and operating instructions, the ROMS CRT Horn activated and the Waste Gas Monitor Activity and effluent changes (IGM 632 and IGM 634) on Grid _2 changed to magenta and began flashing. The Waste Gas Release valve WP V-1602 closed. The malfunction was removed and the alarms cleared and gas release valve WP V-1602 was opened. :No deficiencies were noted and no. exceptions are taken. O
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O 192
['N MALFUNCTION #145 V FAILURE OF CONTAINMENT ATMOSPHERE RADIATION MONITOR
- Failure of Containment Atmosphere Radiation Monitor" was tested on May 1, 1990, from 100% Reactor Power, BOL.
Once the malfunction was activated, the RDMS CRT Horn sounded. The containment monitors RM-6526-1 and RM-6526-2 (particulate), and gas channels IAP111 and ING101 on Grid 3, began flashing in blue, indicating a loss of sample flow. The malfunction was removed approximately ten minutes later and sample flow was restored. Simulator response was in accord with the radiation monitoring system log" d' x " and operating instructions. No deficiencies were noted and n' .. are taken. o 193
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MALFUNCTION #146 INADVERTENT CONTAINMENT ISOLATION TRAIN "B"
" Inadvertent Containment Isolation Train 'B'" was tested on January 11, 1989. This malfuretion satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms, and control system failures, There is no severity option, and the malfunction was activated at 100%
reactor puwer, E0L. Once activated, a "T" signal on "B" train activated, ani "B" train's associated equipment responded to the isolation signal. '
"A" train equipment was not affected. There was no resp 3nse to a reset of the containment ventilation isolation and the "T" signal. The malfunction was removed _approxint.tely twenty minutes after activation, and the containment ventilation isolation and "T" signal were reset. The "B" train equipment was returned to service, and the plant was recovered and stabilized using Procedure 051205.01, inadvertent Phase "A" Containment Isolation. The HPCS alarm summary was saved. Simulator response was in c accord with de NHY Precautions, Limitations, and Setpoints document and l
( the Solid State Protection electrical schenatics. There were no deficiencies and no exceptions are taken. T , 4 l
)
O 194 L I
O MALFUNCTION #147 INADVERTENT SAFETY INJECTION SIGNAL TRAIN "A"
" Inadvertent Safety injection Signal Train 'A'" wat tested on January 11, i 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(22),Processinstrumentation, alarms,andcontrolsystemfailure.
l The malfunction was activated at 100% reactor power, E0L. Once activated, "A" train safety inject'on actuated, the reactor tripped, and "A" train components started. it was verified that "B" train did not activate, and "B" train components did not start. 51 was then manually activated from the MCB, and "B" train components received an 51 actuation signal. The reference procedure _ used was E-0,- Rx Trip or Safety injection. The malfunction was removed approximately thirty minutes after activation, and the plant was lef t in an "$1 actuated" condition. Ten-minute plots of pressurizer level, core thermal power, and SG 'evel were made; and the MPCS alt.rm summary was saved. Simulator respot . was in accord with-the NHY g Precautions, Limitations, and Setpoints document provided by Westinghouse, U and with the Solid State Protection System electrical schematics. There were no deficiencies noted and no exceptions are taken, e v 195-
O MALFUNCTION #148 LOSS OF INSTRUMENT AIR
" Loss of Instrument Air" was tested on March 1, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(2), Loss of instrument air, to'the extent that the whole system or individual headers can lose pressure and effect the plant's static or dynamic performance.
The malfunction is of-variable severity where 0-100% severity is equal to a 0-1200 SCFM Leak Rate. The starting conditions for the test were 100% i reactor power, BOL, and the malfunction was activated at 50% severity, ramped over ten minutes. The compressed air header depressurized until the backup compressor started, stabilizing pressure-at 80 psig. About fifteen minutes into the transient, the lead compressor tripped on overtemperature. Headar pressure began decreasing again allowing the feedwater regulating valves to go closed. Primary temperature and pressure began increasing until the plant finally tripped on SG low level, and the test was stopped.
, The MPCS alarm summary and Air Header Pressure "A" point trends were saved.
Alarms and pressure trends were in accord with the Instrument Air Systems P&lDs and logic diagrams. Thereferenceprocedureused(0N1242.01, Loss of Instrument Air) called for the operator to start the centrifugal air compressor to restore system pressure; however, the centrifugal air compressor is not simulated. There is an outstanding SCR (#86-359) to correct this deficiency by no later than January 31, 1991. No exceptions are taken. i i O 196 l _ _ _- _--_-______-__ __-_____-_-__A
f MALFUNCTION #149 LOSS OF MAIN PLANT COMPUTER
" Loss of Main Plant Computer" wn tested on March 1,1989. This malfunction satisfies the ANSl/ANS-3.5 requirement of Section 3.1.2(22), process instrumentation, alarms, and control system failures.
The malfunction was activated from 7% reactor power, MOL. Once activated, trends and display screens did not update. Power level was reduced and although control rods moved in, the Thermal MW Display and Control Rod "D" , point did not change. The reference procedure used was ON1251.01, Loss of Plant Computer. The malfunction was removed approximately twenty minutes after activation; however, conditions did not change. It is necessary to reset the simulator or retoot the plant computer to recover from this malfunction. No deficiencies were noted and no exceptions are taken. O 197 ) y
T Mt.LFUNCTION #150 (l/ LOSS Of FLOW TO SPENT FUEL PUMP A
" Loss of Flow to Spent Fuol Pump A" was tested on March 8, 1989 from 45%
reactor power, BOL. 1 The malfunction was inserted and the spent fuel pump trioped. An attempt was made to restart the pump from the HCB but the pump would not react until the malfunction was removed. The reference procedure used was 051014.0* Spent Fuel Cooling and Purification. No problems were encountered using , this procedure. Itwasnotedthatthecauseofthemalfunction(closingof SV-2) should lead the pump bearin'g tempere'ures to increase until the pump finally overheats and trips. The MPCS alarm summary and the "A" point trend for the pump temperatures were saved and analyzed. They confirm that pump temperatures did not increase. An SCR (#89-063) has been submitted to correct this deficiency by no later than 12/30/91. Simulator response was , in accord with the Spent fuel System P&lDs and the spent fuel pump
,q electrical schematic. No exceptions are taken.
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MALFUNCTION #151 LOSS Of UPS INVERTER 1B p/ C
" Loss of Uninterruptible Power Supply Inverter IB" was tested on February 23, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of ,
Section 2.1.2(3), Loss or degraded electrical power to the station including loss of power to the individual instrumentation buses (AC as well as 00) that provide power to control room indication. The malfunction was tested from 100% reactor power, MOL. Once activated, , all MCB indication powered by UPS inverter 1B was lost, as verified by Seabrook Station electrical diagrams, included were "A" and "D" steam j generator level control channels which caused a high steam generator level trip. It was verified that the following reference procedures could be used without deviation 051247.01, Loss of 120 VAC Vital Instrumentation Panel; OS1047.01, Vital Inverter Operations; and OS1042.02, Transferring 120 VAC Vital Instrumentation Bus Power Supplies. The malfunction was removed approximately twenty minutes af ter activation, and power was restored to the instrumentation buses. The MPCS alarm summary was saved and no deficiencies ; O were noted during this test. No exceptions are taken. 199
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MALFUNCTION #152 LOSS OF NORMAL FEEDWATER
" Loss of Normal Feedwater" is caused by a closure of the main feedwater isolation valves. The malfunction was tested on 5/1/90 from 100% reactor power, BOL. This malfunction satisfies the ANSl/ANS-3.5 requirement of Section 3.1.2(9), Loss of normal feedwater or normal feeqwater system failure.
Once activated the isolation valves FW-V-30, 39, 48, and 57 went closed. The SGs rapidly steamed below the 10-10 level trip setpoint. Following the trip, pressurizer leve) initially decreased rapidly and then followed the decrease in T,,,. The duration of the test was twenty-five minutes. The MPCS alarm summary and primsry calorimetric data were saved. Also plots were made of core thermal power, loop 1 T,,,, pressurizer level and pressure, SG "A" feedwater flow, and feedwater pressure, level and steam pressure. A panel of experts reviewed the results of the test. The panel concluded that parameters did trend in the correct direction and values were reasonable. No deficiencies were noted and no exceptions are taken. V
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MALFUNCTION #153 p v FCE0LINE BREAK INSIDE CONTAINMENT ON FEEDLINE 'A' "feedline Break inside Containment" was tested on 5/1/90. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(20) Main steam line and main feed line break (inside containment). The malfunction is o, variable severity in which 0 100% severity is equal to a 0 100% line break. It was activated at 100% severity from 100% reactor - power, BOL. Once activated, SG 1evel dropped rapidly and the reactor tripped on low SG 1evel. Pressurizer pressure spiked up as the heat sink was lost and then decreased when the reactor tripped. Pressurizer level decreased rapidly- due to the trip, and containment pressure increased (shutting the MSIV's) until the SG was blown dry and then decreased. SG pressure decreased initially due to the break then spiked high due to the MSIV closure and then decreased to zero on the SG wh: the break. SG 1evel dropped to zero on the SG with the break. The duration of the test was 25 minutes. "A" point trend data, the MPCS alarm summary and primary p calorimetric data were saved. Also, plots were made of major parameters. V- Test results were reviewed by a panel of experts. The panel concluded that i parameters did trend in the correct-direction and values were reasonable. No deficiencies were noted and no exceptions are taken. (} 201
/] V MALFUNCTION #154 LOSS OF STATOR COOLING
" Loss of Stator Cooling" is caused by the pressure control valve (PCV-63) failing closed. The malfunction was tested on December 16,1989, f rom 100%
reactor power, MOL. Once the malfunction was activated, a turbine runback to about 22% occurred within a few seconds. The transient caused the steam generators to shrink enough to trip the plant on lo-lo level, after about five minutes. There were no problems encountered using the reference procedure 051231.03,
" Turbine Runback / Setback". The MPCS alarm summary was saved. Systems responded correctly to the loss of stator cooling as per the Stator Cooling System P&lDs and the Precautions, Limitations, and Setpoints document.
Observed response of the turbine runback was as expected. A more detailed review of simulator response to a runback was done for malfunction #130, LoadRejection. No deficiencies were noted an no exceptions are taken. l 202
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MALFUNCTION #155 FAILURE OF MANUAL REACTOR TRIP SWITCHES
" Failure of Manual Reactor Trip Switches" was tested on January 11, 1989.
This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(23), Passive malfunctions in systems such as engineered safety features, and emergency feedwater systems. The malfunction was activated at 100% reactor power, EOL, and deactivated approximately ten minutes later. Once activated, the reactor trip switches would not trip the reactor. The malfunction was removed and the reactor trip switches responded. The MPCS alarm summary was saved and no , discrepancies were noted. Simulator response was in accord with the Solid State Protection System electrical schematics. No exceptions are taken. O .g \_/ 203 r , - . - - , - ,, , . , , , . . - - , . . .av .- +
tQ MALFUNCTION #156 (_/ INADVERTENT REACTOR TRIP , ~
" Inadvertent Reactor Trip" was tested on January 4, 1989. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(19), Reactor trip.
The cause is a spurious turbine trip. Tne malfunction was activated at 100% reactor power, DOL, and the duration of the test was approximately twenty minutos. Once activated, the reactor tripped immediately, and the first out alarm was turbine trip. Procedures E-0, Reactor Trip or Safety injection, and ES-0.1, Reactor Trip Response, were used to verify conditions and to stabilize the plant. Ten-minute plots of key dynamic parameters were made and compared to the Hillstone Unit 3 . Station Report (generator trip from 100% power test). Simulator response was similar to the Millstone Unit 3 test results. There were no deficiencies noted, and no exceptions are taken. O l (:) - 204
O V MALFUNCTION #157 MS ISOLATION VALVES ALL Fall OPEN
" Main Steam Isolation Yalves All Fail Open" was tested on November 30, 1988.
l The malfunction was activated from 100% Reactor Power, MOL. The MSIVs would not respond to an attempt to close them from the MCB, usir,9 the MSiv control ' switches and the main steam isolation switches. Data collection was then started and malfunction #156, " Spurious Reactor Trip" and melfunction #36,
"'A' SG Safety Valve Fails Open" were activated. Also, "A" SG atmosphere steam dump valve was manually opened. The HSIVs remained open everi though a main steam isolation signal was being generated by low SG pressure. The ,
malfunction was removed approximately twenty minutes after activation and > all MSIVs went closed. The following response data was reviewed by a panel of experts: the MPCS altrm summary and ten-minute plots of loop i and loop 2 T , pressurizer pressure and level, SG "D" pressure, level and steam n flow, and SG "S" level and steam flow. Two logic deficiencies were noted. O "D" point logic for the steam dump valve position is inverted. This problem wascorrected(SCR#90-044)on8/6/90. "F" poi'nt F5434, "RH Train 'B' Hot Leg Suction and RWST Isolation Valve Open" spuriously activated. The problem was previously noted and was corrected under SCR #88-136 on' August 2, 1989. D4097 and 04098 "RHR Pump ' A' ('B') Flow Lo-Lo" and 04676 - 04679 "RCP ' A' 'D' Seal Injection Flow Low" have insuf ficient dead bands. The PHR "0" point deficiency was corrected under SCR #90-071 on 8/6/90. The seal injection "0" point deficiency will'be corrected under SCR #90-106 by no
.later than 1/31/91. The-panel concluded that all dynamic parameters that were reviewed did trend in the correct direction and values were reesonable for the starting conditions. No exceptions are taken.
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MALFUNCTION #158 I r TOTAL LOSS OF EMERGENCY FEEDWATER (
" Total Loss of Emergency feedwater" was tested on November 29, 1988. The initial conditions for the test were 100% Reactor Power, MOL.
Malfunction #158 was activated along with malfunction #156, " Inadvertent Reactor Trip". The trip resulted in SG levels shrinking to below the emergency feedwater actuation setpoint. Approximately six minutes into the test, the turbine driven emergency feedwater pump tripped and approximately two minutes later the motor driven emergency feedwater pump tripped. Neither pump would respond to a restart attempt. Fifteen minutes into the cost, malfunction #158 was removed and the emergency feedwater pumps were aarted. Emergency procedure FR-H.1 was used to verify the ability of the condensate system to supply emergency feedwater. RCS pressure was reduced to 1900 psig and *D" SG was depressurized to less than 500 psig. "A" and "B" r,ondensate pumps were then aligned to the emergency feedwater system using a remote function to open bypass valve FW V-103. It was noted that,
,3 although not required for this malfunction, the simulator lacks the
(> capability to line up firewater to the feedwater system if both emergency feedwater and condensate are unavailable. As per meeting minutes 89-02, the Simulator Review Committee decided that this limitation is not acceptable.
'An SCR (#89-180) was submitted to correct this deficiency. This will be completed as part of the steam, feed, and condensate upgrade due December 30, 1991. A panel of experts reviewed the MPCS alarm summary and ten-minute plots of the following loop 1 cold and hot leg temperatures; pressurizer pressure and level; SG "A" and "C" steam flow, pressure and level; and core thermal power and generator power. The panel concluded that the parameters responded in the correct direction and values were reasonable for the- starting conditions. No other deficiencies were noted and no exceptions are taken.
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MALFUNCTION #159 i MS LINE B SAFETY VALVE FAILS OPEN i bq
- Main Steam Line 'B' Safety Valve f ails Open" satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(20), Main steam line break outside containment.
The malfunction was tested on December 1,1989 from 45% reactor power, BDL. i Upon malfunction activation power level shot up and then stabilized with a T,,,/T,,, mismatch and the control rods at the C 11 setpoint. The malfunction was removed approximately twenty minutes later. Data was collected for the first ten minutes of the test. During the test, a deficiency was noted. VAS procedure (D5788) requires the operator to check the loose parts monitoring panel to determine if a safety valve has opened. This panel is not simulated. As per the Simulator Review Committee Meeting Hinutes (89-01), the committee decided that the loose parts monitoring panel is not required to be simulated but can be trained on in the control room. This is an exception to the ANS1/ANS 3.5 requirement of Section 3.2.1, Degree of Panel Simulation, "The simulator shall contain operational panels to provide [ (q) the controls, instrumentation, alarms, and other man-machine interfaces to respond. to the malfunctions of section 3.1.2. (Plant Malfunctions)." A panel. of- experts reviewed the following data the MPCS alarm summary, primary plant calorimetric data, MCB chart recordings of SG "B" narrow range level, wide range level, pressure, feed flow and steam flow, T,,, / T,,, , pressurizer level, and RCS pressure. Also reviewed were ten-minute plots of core thermal power, generator power, loop 2 and 4 hot and cold leg temperatures, pressurizer pressure and level, and SG "B" and "D" level and steam flow. The panel's conclusion was that dynamic parameters did trend in the correct direction and values were reasonable for starting conditions. No other deficiencies were noted or and no exceptions are taken. 1 ( .- 207
MALFUNCTION #160 MAIN STEAM LINE 'C' SAFETY VALVE FAILS OPEN
- Main Steam Line 'C' Safety Valve Fails Open" was tested on May 7, 1990.
This malfunction satisfies the ANSI /ANS 3.5 requirement of Section " 3.1.2(20), Main steam line break. The initial conditions for the test were 27% reactor power, BOL with rod control in manual. Once activated, main steam line "C" safety valve went open. Steam flow and feed flow to "C" SG increased. "C" SG pressure decreased and the SG level increased from swell. T,,, decreased causing a T,y, /T,,, do h at ion. The control rods were placed in auto and began stepring in because of the T,y, / T,,, deviation. The malfunction was removed approximately twenty-five minutes af ter activation and the plant returned to the original conditions except for the control rods, which were left in auto. The MPCS alarm summary and primary plant calorimetric data were saved. Simulator response was in accord with the Reactor Coolant System P&lDs and the Seabrook Station FSAR. One deficiency was noted during the test. The VAS procedure requires the operator to check the loose parts 4 ( monitoring panel to determine if a safety valve is open. This panel is not simulated. The Simulator Review Committee determined (as per meeting minutes 89-01) that this panel need not be simulated but can be trained on in the control room. This is an exception to the ANSI /ANS-3.5 requirement of Section 3.2.1, Degree of Panel Simulation, "The simulator shall contain sufficient operational panels to provide the control instrumentation alarms and other man-machine interfaces to respond to the malfunctions of 3.1.2 (PlantMalfunctions)." No other deficiencies were noted and no exceptions are taken. 208
Ok/ MALFUNCTION #161 MAIN STEAM LINE 'O' SAFETY VALVE FAILS OPEN The malfunction " Main Steam Line 'O' Safety Valve fails Open" was tested on May 7, 1990. This malfunction satisfies the ANS!/ANS-3.5 requirement of Section3.1.2(20),Mainsteamlinebreak. The initial conditior.s. for the test were 27% reactor power, BOL. Once activated, steam generator pressure dropped and lerel swelled. Feed flow , and steam flow increased and core thermal power i;:reased. T,,, and _ pressurizer pressure decreased. The plant stabilized out at approximately 37% reactor power. The duration of the test was twenty _ minutes. After the malfunction was removed, the safety valve went closed and the plant returned.
- to the initial conditions. The MPCS alarm summary and primary plant calorimetric data were saved. Ten < minute plots of pressurizer pressure and level, SG pressure and level, SG steam and FW flow, T,y,, and core thermal power were made. One deficiency was noted. The VAS procedure requires the operator to check the loose parts monitoring panel to determine if a safety A- valve is open. This panel is not simulated. The Simulator Review Committee deterrnined (per meeting minutes 89-01)_ that this panel need not ,
be simulated but can i>e trained on in the Control Room. This is an . exception to the ANSI /ANS-3.5 requirement of Section 3.2.1, Degree of Panel Simulation,.*The simulator shall contain sufficient operational panels to provide the control instrumentation, alarms and other man-machine interf aces to respond to the malfunctions of 3.1.2 (Plant Malfunctions)." Simulator response was in accord with the primary and secondary plant p&l0s and'the Seabrook Station FSAR. No other deficiencies were noted and no other exceptions are taken. ( _Q/ > 209
i 1 l l l MALFUNCTION #162 l STEAM GENERATOR TUBE RUPTURE T0 "A" STEAM GENERATOR O.
" Steam Generator Tube Rupture to "A" Steam Generator" was tested on l 11/17/88. This malfunction satisfies the ANSI /ANS-3.5 requirement of I Section 3.1.2(1) Loss of coolant (a) significant SG leaks. i i
The malfunction was activated at 10% severity (0100% is equal to 0-2000 gpm i leak), from 100% reactor power, MOL. The malfunction was run for twenty 1 a minutes with no operator action taken. The MPCS alarm summary was saved, and ten minute plots were made of the following: core thermal power; ; generator power; pressurizer pressure and level; loop 1 cold'and hot leg temperature; letdown and charging flow; and Sn "A" and "B" steam flow, pressure, and level, A panel of experts reviewed the test results and concluded that the parameters did trend in the correct direction and that values were reasonable. No deficiencies were noted and no exceptions are , taken, O G 9 O 210
d' i
,. MALFUNCTION #163 J Steam' Generator "B" TUBE RUPTURE-
"SG "B" Tube Rupture" was tested on November 28, 1988 at 50% severity (0-
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1004 severity-is equal to: 0-2000 gpm leak), from 100% reactor power,_MOL.- This me,1 function satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (1),Lossofcoolant-(a)'significantSGleaks. The malfunction was activated'and data was collected for-ten minutes. Once-the reactor tripped on OT Delta-T, emergency feedwater was cutback to "A",
"C" and "D" SG's. Emergency.feedwater was isolated to "B" SG once narrow range _ level was greater than 5%. .Becense subcooling:was never. lost, the-reactor coolant pumps were kept- running for the entire transient. The following data was collected-'and reviewed by_a panel of experts:- The MPCS ;
ialarm summary, . main steam line 2 and condenser air evacuation radiation monitor trends, and te'n-minute plots of- the followin. core therma'i power, generator power, pressurizer pressure and -level,. letdown flow, charging
; flow, and SGs-"A" and "B" wide range level and pressure. Also reviewed were:
MCB stripchart recordings of pressurizer level and pressure, loop 1 and'2 hot leg temperatures, and SG "B" narrow range level. The panel _ concluded .
~that all_ dynamic _ parameters did trend in the correct' direction and values were reasonable for starting conditions. No deficiencies weref aoted and no- '
exceptions are taken. L
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A MALFUNCTION #164 V STEAM GENERATOR "D" TURE RUPTURE S/G "D" Tube Rupture satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2 (1), Loss of coolant (a) significant ro leaks. The malfunction is a variable severity malfunction where 0-100% severity is equal to a 0-2000 GPM leak, it was tested at 100% severity from 100% reactor power, MOL on November 28, 1988. 4 The malfunction was activated and the following data was collected for ten minutes with no operator action: the MPCS alarm summary; condenser and main steam line radiation monitor trends; MCB chart recordings of pressurizer level, RCS pressure, SG "D" narrow range level; and ten-minute plots of core thermal power, generator power, pressurizer pressure and level, SG "C" steam flow and feed flow, and SG "0" steam flow and feed flow. The results were j reviewed by a panel of experts. The panel concluded that all dynamic
. parameter did trend in the correct direction and values were reasonable for p starting cotiditions. - No deficiencies were noted and no exceptions are V taken.
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.O v MALFUNCTION #163
' MAIN STEAM LINEA' STEAM FLOW CHANNEL 1 FAILS HIGH i
> 1" Main Steam Line AE Steam Flow Channel 1 Fails High" was tested on December 1,1988. - This malfunction satisfies the ANSI /ANS-3.5 requirement of Section s
3.1.2(22), Process 11nstrumentation,alarmsandcontrolsystemfailures. Three tests were conducted. The first test was_ run-from 100% Reactor Power, MOL with -no . operator? action to collect data. Once activated,-steam flow indicatio'n-(FI-512A): failed high. Actual fecd flow: tracked the indicated-steam flow causing'SG level to increase. The level error caused the feed y flow to decrease and theLSG. level to decrease to the original value, ' The seconditest was run from 45% Reactor Power, BOL. In thi_s case the h reactor tripped before the -level error could respond.- The cause and effect 4 description did_not state that at-lower power levelsLthe reactor would trip
-on SG;hi-hi- level due to: the overfeeding. - This response was verified
/ correct against. Millstone 3 test data. The cause and effect description was-updated?to reflect this.
The third test was again run from 45% Reactor Power, BOL. . However, for this test- operator action was taken -to - recover . from the malfunction. using-procedure 0S1235.04, SG - Feed . or -- Steam Flow ? Instrument - Failure. - Data , collected for the-first' test included MCB-chart recordings <of feed: flow, steam' flow', and:SG "A" level;=the MCB alarm summary; and ten-minute plotst ofl core thermal power, Loop, T ,-SG'"A" and'"B" level, steam flow, and SG "C" Land "D" level. Simulator response was in accord with system PalDs and steam and.-_ feed flow logic diagrams.' ' No deficiencies were -'noted and no'
= exceptions are taken._ 1
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MALFUNCTION #166 eq MAIN STEAM LINE 'A' STEAM FLOW CHANNEL 1 FAILS LOW L)
" Main Steam Line 'A' Steam Flow Channel 1 (F1-512A) Fails Low" was tested on December 1, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms and control failures.
Three tests were conducted. The first test was run from 45% Reactor Power, BOL. Once activated, indicated steam flow failed low causing actual feed flow to decrease. SG level began to decrease. The level error caused feed flow to increate and level was restored. Data was taken during this test. A second test was run to observe simulator response when the malfunction is activated from 100'S Reactor Power, MOL with no operator action taken. Once activated, the SG level quickly dropped below the SG 10-10 level trip. For the third test, the simulator was again reset to 100% Reactor Power, M0L, Once the malfunction was activated, operator action was taken in accord with O,.m procedure OS1235.04, SG Feed or Steam Flow Instrument Failure. No problems were encountered using this procedure, and the reactor trip was prevented. The malfunction-was then removed and proper channel operation was verified. Data collected during the first test included the MPCS alarm summary, MCB chart recordings of SG "A" level, and ten-minute plots of the following: core thermal power, pressurizer pressure, loop 1 and loop 2 T,,, and SG "A" and "C" level, pressure, steam flow and feedwater pressure. Simulator response was in accord with the system P&lDs and the Main Steam and Feedwater Systems' logic diagrams. No deficiencies were noted and no exceptions are taken. LJ I 214 l
(~\g - MALFUNCTION #167 O TURBINE STOP VALVE #2 STUCK AS IS
" Turbine Stop Valve #2 Stuck As Is" was tested on December 16, 1988. This malfunction satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(23),
Passive malfunctions. The malfunction.was activated at 100% Reactor Power, MOL. The weekly stop valve test ON1431.02, was performed. SV-1, 3, and 4 tested satisfactorily, o but SV-2 would not change position. The reactor was tripped and both EHC
' pumps stopped. SV-2 remained open. The malfunction was then removed and SV-2 closed. The duration of the test was approximately thirty minutes.
.The MPCS al arm summary was saved . Simulator response was in accord with the main steam system P&lDs and the G. E. Turbine Manual. No deficiencies were noted and no exceptions are taken.
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MALFUNCTION #168 ROD DRIVE MOTOR GENERATOR "A" BREAKER TRIP
" Rod Drive Motor Generator "A" Breaker Trip" was tested on May 1, 1990.
The malfunction was tested from 100's reactor power, BOL. Once activated the "RDMG ' A' or 'B' Trouble" and "RDMG ' A' Output Breaker Open" alarms came in.- The MPCS color graphic display indicated that the breaker was open. The malfunction was removed ten minutes later; the alarms cleared and the RDMG breaker showed closed. Simulator response was in accord with the Rod Drive System logic diagrams. No deficiencies were noted and no exceptions-are taken. O g 216
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MALFUNCTION #169 R0D ORIVE MOTOR GENERATOR "B" BREAKER TRIP
" Rod Drive Motor Generator "B" Breaker Trip" was tested on May 1, 1990.
The malfunction was tested from 10056 reactor power, BOL. Once activated the "RDMG ' A' or 'B' Trouble" and "RDMG 'B' Output Breaker 0 pen" alarms came in. The MPCS color graphic display indicated that the breaker was open. The malfunction was removed ten minutes later. The alarms cleared and the RDMG breaker showed closed. Malfunctions 168 and 169 were then activated
-simultaneously. Both RDMG outbreaker breakers opened and the control rods dropped. The rapid negative reactivity insertion caused an OT Delta-T trip to occur. Simulator response was in accord with the Rod Drive System logic diagrams, and the Seabrook Station Startup Testing Rod Worth Curves. No deficiencies were noted and no exceptions are taken. ,ry .
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MALFUNCTION #170/171/172/173 O Malfunctions #170,171,172, and 173 are RHR malfuncti ins in development and not-available for training. No tests were performed on these malfunctions. O O i 218
N MALFUNCTION #174
[d PORV 456A FAILS CLOSED
" Power Operated Relief Valve 456A Fails Closed" was tested on November 16, 1988. Because this malfunction causes PORV 456A ,o f ail closed wheti the setpoint is exceeded,'it satisfies the ANSI /ANS-3.5 requirement of section 3.1.2(23), Passive malfunctions.
The malfunction was activated from 100% reactor power, MOL with malfunction
#13, " Failure of SSPS to Trip the Reactor", active. The pressurizer spray valves we e placed in manual and the back up heaters placed on. PORV 456A did not respond to an attempt to manually open the valve. At 2385 psig, PORV 456B opened and it was manually closed. Primary pressure was allowed to increase to just less than the Code Safety Relief Valves Setpoint prior to removing the malfunction. Once removed PORV 456A opened immediately.
The MPCS alarm summary was saved. Simulator response was in accord with the Seabrook Station Precautions, Limitations, and Setpoints document and the o primary- system P& ids. No deficiencies were noted and no exceptions are b taken. O 219
l i r~N- MALFUNCTION #175 (f PORY 4568 FAILS CLOSED 4 "PORV 4568 Fails Closed" was tested on November 16, 1988. This malfunction ; satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(23), Passive I malfunctions. The initial conditions for this malfunction were 10S% Reactor Power, BOL with malfunction #12 (Failure of SSPS to Trip the Reactor) activated and ! the backup heaters eneigized. RCS pressure was increased to the PORV-setpoint. PORV 456A lif ted and PORV 456B did not. 456A was closed manually ! and RCS pressure was increased well beyond the lift setpoint; however, 456R- i would not open automatically or manually. The malfunction was removed and 3 4568 opened-- automatically. The duration of the test was approximately fifteen minutes.- Simulator response was in accord with the Reactor Coolant ; System P& ids and the Seabrook Station Precautions, Limitations, and Setpoints document. No deficiencies were noted and no exceptions are taken. ! r'm i
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i [\ l 'O l 220 l
1 MALFUNCTION #176 r~ STUCK RODS (RCCA H10 AND RCCA K8) b} The maifunction, " Stuck Rods (RCCA H10 and RCCA K8)" was tested on May 2, 1990. It satisfies the ANSI /ANS-3.5 requirement of Sections 3.1.2(12), Control rod failure including stuck rods, and (23), Passive malfunctions. The initial conditions were 100% reactor power, 80L. Once activated, control rods H10 and K8 would not respond to manual or individual group rod control or automatic rod control when turbine load was reduced. The malfunction was deactivated, and it was verified that the rod would respond to both manual and automatic rod control. The malfunction was then activated and a manual reactar trip inserted. Rods RCCA H10 and K8 stuck, as verified by position indication and the total reactivity worth inserted by the control rods. No problems were encountered using reference procedure 0S1210.06, " Misaligned Rods". The duration of the test was approximately thirty minutes. Simulator response was in accord with the Rod Control Logic Diagrams and the Seabrook Station Startup Test Rod Worth Curves. O v O , l 221
- /~'T MALFUNCTION #177
(_.) STUCK R005 (RCCA 08) The malfunction " Stuck Rods (RCCA 08)" was tested on May 2, 1990, it satisfies the ANSI /ANS-3.5 requirement of Sections 3.1.2(12), Control rod failure including stuck rods, and (23), Passive malfunctions.
-The initial conditions of the malfunction were 100% reactor power, BOL.
Once activated, control rod 08 would not respond to manual or individual group rod control or automatic rod control when turbine load was reduceo. The malfunction was deactivated, and it was verified that the rod would respond to both manual and automatic rod control. The malfunction was then activated-and a manual reactor. trip inserted. RCCA 08 stuck, as verified by position indication and the total reactivity worth inserted by the control rods. No problems were encountered using -reference procedure 051210.06, " Misaligned Rods". The duration of the test was approximately L thirty minutes. Simulator response was in accord with the Rod Control logic
- diagrams and the Seabrook Station Startup Test Rod Worth Curves.
1 222 1 m-
T'T MALFUNCTION #178 d MAIN-STEAMPRESSURETRANSMITTER"A"(PT-3001) FAILURE
" Main -Steam- Pressure Transmitter 'A' (PR-3001) Failure" is a variable severity malfunction in which 0-100% severity is equal to 0-1500 psig transmitted value. The malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(22), Process instrumentation alarms, and control system failures.
The malfunction was tested on. April 16,1990 at S0% severity from 100% reactor power, MOL. -Once activated PT-3001 failed to 750 psig. The "A" atmospheric steam. dump valve which had been closed, opened. PT-3001L continued- to read 750 psig. The malfunction was removed approximately
' fifteen minutes later and PT-3001 returned to normal. Simulator response was in accord with the Mt Steam System P&l0s. No deficiencies were noted and no exceptions.are taken.
T'\ , V O v-223
MALFUNCTION #179 (N); MAIN STEAM PRESSURE TRANSMITTER "B" (PT-3002) FAILURE _
" Main . Steam Pressure Transmitter 'B' (PR-3002) Failure" is a variable severity malfunction in which 0-100% severity is equal to 0-1500 psig transmitted value. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms and control system failures.
The malfunction was tested on April 16, 1990 at 100% severity-from 100% reactor power, MOL. Once activated, PT-3002 failed to 1500 psig, causing. the "B" atmospheric steam dump valve to open. The tiPCS color graphic display showed the "B" ASDV open and an increase in "B" SG steam flow. The "B"-ASDV was clo.eed from the MCB and then reopened when the control switch was returned to the " Modulate" position. The malfunction was- removed approximately twenty. minutes af ter activation, and PT-3002 returned to normal. Simulator response was in accord with the Main Steam System P&lDs
-., and logic diagrams. No deficiencies were noted and no exceptions are taken, n
U 224 l
-__-_-_____ - _____-____ ___ _ ________ - ____-______-_-__-___-_ - A
MALFUNCTION #180 MAIN STEAM PRESSURE TRANSMITTER 'C' (PT-3003) FAILURE "Mair Steam Pressure Transmitter 'C' Failure" is a variable severity malfunction in 'which 0-100% severity is equal to 0-1500 psig transmitted value. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(22), Process instrumentation, alarms, and control system f ailures. i The malfunction was tested on April 16,1990 at 100% severity f rom 100% reactor power, MOL. Once activated, PT-3003 went to 1500 psig, causing the "C" atmospheric steam dump valve to open. The MPCS color graphics display showed the "C" ASOV open and an increase in "C" SG Steam Flow. The "C" ASOV was manually closed from the MCB. It once reopened when the control switch was returned to the " Modulate" position. The malfunction was removed approximately twenty minutes af ter activation and PT-3003 returned to normal. Simulator response was in accord with the Main Steam System P& ids and logic diagrams. No deficiencies were noted and no exceptions are taken. 225
'i - M*d.IUNCTION #181-
[O MAIN STEAM PRESSURE TRANSMITTER "0" 1.PT-3004) FAILURE'
" Main Steam Pressure Transmitter 'O' Failure" is a variable severity malfunction- in which 0-100% severity is equal to 0-1500 psig transmitted value. This malfunction satisfies the ANSI /ANS 3.5 requirement of Section 3.1.2(22), Process instrumentation, _ alarms, and control system f ailures.
The malfunction was tested on April 16, 1990 at 100% severity from 100% reactor power,-MOL. Once activated PT-3004 went to 1100 psig, causing the "0" atmospheric steam dump valve to partially open. The ASOV was closed from the MCB and then reopened when the control switch was returned to the
" Modulate" position. The malfunction was removed approximately thirty minutes af ter activation and PT-3004 returned to normal. Simulator response was in accord with the Main Steam System P&l0s and logic diagrams. No 1 deficiencies were noted and no exceptions are taken. / .
NJ .
%)
226
(7f/ MALFUNCTION #182 STUCK R0D-(RCCA D12)
- The malfunction _ " Stuck Rod (RCCA 012)" was tested on May 2, 1990. It satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(12), Control rod failure including stuck rods, and (23), Passive malfunctions.
The initial conditions were 10086 reactor power, BOL. Once activated, . control rod D12 would not respond to manual rod control, individual group, or-automatic control when turbine load was decreased. The malfunction was deactivated, and it was_ verified that the rod would respond to both manual and automatic rod . control. The malfunction was then reactivated and a manual reactor trip _. inserted. RCCA 012 stuck, as verified by position indication,-_ and the total reactivity worth inserted by the control rods. No- problems were encountered. using reference procedure 0S1210.06,
" Misaligned Rods." The duration of the test 'was -approximately thirty minutes.- Simulator response was in accord with the Rod Control logic
-- diagrams and the Seabrook Station Startup Test Rod Worth Curves. G' 4 227
i (7
\_f MALFUNCTION #183 STUCKROD(RCCAH3)
The malfunction " Stuck Rod- (RCCA H3)" was tested on May 2, 1990.- it satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(12), Control rod failure including stuck rods, and (23), Passive malfunctions. The initial-conditions for this malfunction were 100% reactor power, BOL. Once activated, control rod H3 would not respond to manual, individual, group, or automatic control when turbine load was reduced. The malfunction was deactivated, and it was verified that the rod would respond to both manual and automatic rod control. The malfunction was then activated and s a manual reactor trip inserted. RCCA H3 stuck, as verified by position indication,'and.the total reactivity worth inserted by.the control rods. No problems were_ encountered using reference procedure. 0S1210.06,
" Misaligned Rods." The duration of the test was approximately thirty minutes. ' Simulator response was in accord with the Rod Control logic -7 diagrams and the Seabrook Station Startup Test Rod Worth Curves.
n J 4 228
3 [b MALFUNCTION #184 DROPPED ROD (RCCA H8) i The malfunction " Dropped Rod (RCCA H8)" was tested on May 2,1990. It ; satisfies the ANSI /ANS-3.5 requirement of Section 3.1.2(12), Control rod failure, including rod drops. The malfiunction was activated from 100% reactor power, BOL. The Rod Bottom Light (RCCA H8), and Rod Deviation Alarm immediately activated. A T,,,- T,,, Deviation caused rods to begin stepping out. Pressurizor pressure, l level, T,,, and SG Pressure decreased. An attempt was made to retrieve the t dropped rod; however, the rod could not be recovered until the malfunction was removed. Reference procedure 0S1210.05, Dropped Rod, was used. The duration of the test was approximately fif teen minutes. "Before" and "af ter" primary plant calorimetric data was taken and the MPC5 alarm summary was saved. Simulator response-was. in accord with the Rod Control System i logic diagrams and the Seabrook Station Rod Worth Curves. No deficiencies
- /^'t were noted and no exceptions are taken, i V
j i l l 229 L
]v 3.5 Simulator Design Changes / Enhancements 3.5.1 Steam Generator Model Upgrade The Steam Generator Model was upgraded in January of 1988. A non-equilibrium thermodynamics model was implemented to properly account for the steam generator dynamics during abnormal conditions. This new model was validated in September of 1988 by benchmarking simulator response for several transients to actual transient data from North Anna and Lof tran calculations, provided by Westinghouse. Following validation, the model was accepted for training use. 3.5.2 Simulator Computer Complex Replacement In July 1988, the simulator computer complex was replaced with a Gould 32/67420 SelPAC computer. In conjunction with the replacement, the simulutor executive sof tware was rewritten to provide additional memory and to utilize the reflective memory capability of the computers. The (] performance of the simulator was benchmarked by performing the following on the old computer system: a plant startup, a plant shutdown, and the ten annual operability tests of ANS 3.5, Appendix B. The tests were then repeated with the simulation software running on a single 32/67420 computer and also with two 32/67420 computers linked via reflective memory. Results of the test were benchmarked by plotting the response of critical parameters on the old ccmputer system versus the single 32/67420 configuration and the old computer system versus the reflective memory configuration on the same axis. Results from each series of test demonstrated that simulator response had not been affected by the changes to the executive software or the new computer system. Tests were completed in September 1988. 3.5.3 18 Node Reactor Coolant System In- February 1989, the Reactor Coolant System model was upgraded to an eighteen-node model. The new model's noding scheme consists of two nodes
'(] in the pressurizer, four in the reactor vessel (core, lower and upper plenum and upper head), and three in each loop (hot and cold leg and steam generator tube-side). Imolementation of the pressurizer component of the 230
l RCS model was completed in May of 1986. The eighteen node model was to r3 " big" to execute in real time on the Gould 32/77 computer being used at that d time; for that reason, a nine-node model was incorporated into the training load in April 1987. The eighteen node RCS model was incorporated into the training load in March of 1989 on the new Gould 32/67 computers. A thorough validation test was performed prior to using the new model for training. The test consisted of 8 FSAR transients, a natural circulation test, a three-loop controlled shutdown, and a complete plant startup and shutdown using Seabrook Station operating procedures. Results of the test were compared to real plant data from similar plants (wherever possible) and to the Seabrook Station FSAR. The test proved that the new RCS model did accurately reflect expected reactor coolant system response. 3.5.4 Radiation Data Monitoring System Upgrade The Radiation Data Monitoring System Model (ROMS) was upgraded in October p 1989. The-model was, tested utilizing a special test procedure. Several V problems were noted during the conduct of the test; all were corrected prior to accepting the model for training. All test criteria established by the test procedure were satisfied by the enhanced model. 3.5.5 Reactor Vessel Level Indication System The Reactor Vessel Level Indication System (RVLIS) was installed on the simulator in January 1990 by Westinghouse. A'special test was written to verify the proper operation of the system. Several minor problems were identified, but they were immediately correctea by the Westinghuse representative. All tests required by the test procedure were successfully completed, and the system was accepted for use in January 1990.
,m 231
3.6 Physical F.idelity Audit'. l D i An audit of the_ simulaterls phys _ical fidelity was last' performed in November l 1989 by comparing photographs of the main control board with the simuloior. I
- All discrepancies identified have either been-corrected or are exceptions that ' have- been determined not to detract from training by_ the Simulator Review Committee. These exceptions are listed below, u
'1. CP-180A The. spacing of the letters on the cabinet label for CP-180A is: different. The label-has four screws on the simulator, it has two in the control room.
- 2. -CP-180A A filler strip on the simulator's CP-180A does not-have=
notches. ' 3.- CP-180A. The lettering of;the simulator label 1-RR-6528-1 is larger than that of.the-control room.-
-4. ~CP-180A The ' lettering of the' simulator label 1-RR-6528-2_ is larger ,
- than that of the control room..
5.. CP-180A- -The lettering of the simulator label 1-RR-6506-A is larger q than that of the control room.- 67 CP-180A: .The lettering of the simulator label-1-RR-6530-is larger than that of the control room.- 3
- 7. CP-1808' The spacing of the' letters on'the_ cabinet label lfor CP-
~1808.'is different; The label'on:the simulator has four' -
screws, it has two screws in the control-room. , , 8k CP-180B The : filler strip on1 the > simulator's CP-1808 ~'does not contain notches.- '
- 9. :CP-180B The = lettering- of. _ label RM-RK-6576-B is larger - ino the - -
simulator than inithe control room.
- 10. ~CP-180B- The. lettering' of- label 1-RR-6507-8 is larger inL the simulator than'in the control room.
' - 11. CP-16 The spacing- between SR'and N31 on the channel selector w ,
j escutcheon is closer on the simulator than in the control n room.
* : 12. H Rear- The vendor name engraving in the lower left corner of the shaft voltage monitor is-missing on the simulator.
L 13. H. Rear- -The shaft voltage monitoring panel is located about two inches lower on the simulator than in the control room.
- 14 . : H Rear The demarcation for ground detection and vital-DC is about 232 a --%. -.- , '5.,,,,,..
. , , . . i.,w ., ,, i ,y - y _~+em , 5. .x , -
ri. . ., , y
one inch closer together in the plant. [ ~' ')( 15. H Rear- The ground detection meters are closer together in. the vertical direction in the simulator than in the control room. 16.- G Rear lhe recombiner A panel on the simu stor is a single pnel, witn no vendor information plate in the upper lett corner. The panel in the control room is comprised of three panels and has a vendor information plate.
- 17. G Rear The lettering on the RCP vibration panel on the simulator is larger than that in the control room.
18.- G Rear. The recombiner B panel on the simulator is a single panel, with no vendor information plate in the upper left corner. The panel in the control room is comprised of three panels and has a vendor information plate.
- 19. -G Rear Hydrogen analyzer 173A on the simulator has three screws on the sides; the one iri the control room has four.
- 20. G Rear The scales for recorder CS-LR-102 on the simulator have
" GAL 103 at the top; in the control room it is "KGAL".
- 21. G Rear Hydrogen analyzer 174A on the simulator has three screws on the sides; the one in the control room has-four.
- 22. C_ Rear The control room has a hole cover plate between the lights
. for FAH-F-41; the sim:llator does not.
L23; C Rear The.. lettering for CAH-DP-314 is larger in the simulator than in the control room.
- 24. C Rear The instruments along the left side of panel C rear:are closer together in the simulator ~than in the control room.
'25.: C Rear The control room has a hole cover plate between the lights
~for FAH-F-74; the 5,imulator does not.
- 26. B' Rear The mimic along the bottom of panel B rear' is closer together than in the plant.
- 27. B' Rear The hole _ cover . plates above PCV-126 are square in the simulator; in the control room, they are round.
- 28. . B Front The location of panel VA50 & S1 is slightly different.
'29. B Front' The lettering on label CBS-LR-2385 on the simulator is smaller.
n (_ 233
'l
-30. E Front The label for VL-1 is immediately below the cutout on the
(~'Y simulator.vice one1 inch below the cutout in the control room.
- 31. E Front The label for VR-1 is-immediately below the cutout in the control room.
- 32. E Front The-label for VL-2 is immediately below the cutout on the simulator vice one inch below the cutout in the control-Toom. ,
- 33. E Front The label-for VR-2 is immediately below the cutout on the simulator vice one inch below the cutout in the control room.
- 34 .~ G Front The mimic at the bottom of panel G front on the simulator
-is compressed.
l 11 f
- (_./
234
-i
?-- !
j 4.0 SIMULATOR DISCREPANCY RESOLUTION AND UPGRADING
- 4. 1 Identifying,. logging, correcting, and testing reported simulator discrepancies Nuclear Training Procedure.NT-3734 establishes the requirements for logging, correcting and testing simulator discrepancies. 'The following is a summary of the procedure. Discrepancies may be noted by any simulator user. New 3 discrepancies are . logged into_ the Simulator Information Management System d and forwarded to the following people- for review: Design Control Coordinator, Simulator Engineer, Hardware _ Technician, .0perations : Training . '
-Supervi.cor,.and Simulator; Systems Department Supervisor. -The priority for complet' ion of the Simulator Change Request is established by the Operations Training Supervisor 1as part'of the review process. Once the-review cycle [
-is complete, the SCR iis placed ' in the work queue based 'on its priority.
- Changes l to the. simulator software are first made' and -tested on the development . computer system. Once the assigned _ sof tware engineer _is y
- satisfied that the problem has been corrected, the change is loaded on :the .
J p ;simulatorJcomputer system where it _ is tested by the Simulator ~ Engineer. 3
\ 'Upon successful completion-of this initial test, the change is1 tested by the. 7 1 Simulator Support Instructor.- Once: the Simulator- Support _ Instructor is- 1 satisfied that the Lchange is correct,- a- new- training load, which-e incorporatesithe change, is created.: Any applicable simulator documentation ,
-affected bytthe-changefis then updated and:the. Simulator Change-Request is !
Tclosed.1/A cop'y of the' procedure is included in Append _ix B.
'4.21 Tracking design changes-incorporated-into the reference plant 4
The JTraining Group 'is on controlled distribution ofiall._ plant ~ Design
;CoordinationReports.(DCRs), Minor = Modifications (MMODs)andDocumentation- ;< LClange Reports (DRRs).~ The Design Control-Coordinator reviews-each of the-
[ ~
. above change d'ocuments :for- simulator training impact. If the: change affects-simulator - configuration,' at Simulator Change Request is initiated to simplement the change.
A y 235
- 4,3:-OutstandingSimuIatorChangeRequests
,f a
^ - A. list-of the open Simulator Change Requests-(SCRs') is contained in_ Appendix C. The status' codes areas follows:
R in initial review I- .A Deficiency awaiting tap'.ementation-E An enhancement awaiti ng implementation TR- The change:has been implemented and.is awaiting testing , mi
- A distinction has been made between thoe SCRs that represent -deficiencies
'a'nd those. that- represent- enhancements.
A _ deficiency is the incorrect-operation orcresponse of an item that is presently simulated. or~an ANSL3.5 : ) 1 requirement that is-not' simulated. An enhancement is the additian of--an
= item that.has not been previously simulated, but has been identified through;
- aL training: needs L assessment, q H
H E4.4t: Overdue Plant Design C.hanges g
, r~~N ' There areitwo reference uan modifications which have exceeded lthe one year
; period'between being-implemented in'the plant:and on the simulator. This-Lisp an .' exception :to-- the TANSI/ANS-3.5 requirement- of section 5.31(The h
[g
~
simulator Eshallf be modified - as required withinL12imonths following the 1annu'al: establishment of the: simulator update' design data...).- , l
,. LDCR.'~88-0125:(SCR#88-213)updatesthelelectricaltload-calculationsforthe :!
- l'25l VDC batteries, chargers, _ motor o starters - and= uninterruptiblef power - -i
= supplies.
4
- DCR 88-114 (SCRif88-216) added a- 15 minute' shunt trip which initie^es 'af ter :
iUPS:E0-112B has shifted toithe DC bus.- y ; s .1 W' Each of these changes-has been assigned priority 4 (the problem has'little: 7 gy or no1 adverse impact on -training). The Simulator Review Committee has M -evaluated the-impact of delaying-the implementation-of each of these design 1 changesiand-has determined that the-delay.will not detract from training. h 236
-i 5.0 MISCELLANEOUS ;
i V 5.1 Schedule For Testing The four year schedule for testing is included in Appendix D. 5.2 Simulator Review Committee The Simulator Review Committee (SRC) is responsible for defining the capabilities of the Seabrook Station Simulator based on identified training objectives. The SRC includes representatives from the Operations Department, the Operations Training Department, and the Simulator Systems Department. The SRC meets at least once per quarter. The SRC charter is include.1 as Appendix E. 5.3 Operational Experience Review Program C New Hampshire Yankee has establisned an Operational Experience Review
'D Program (0ERP) which provides a method for evaluating regulatory j L correspondence, inCastry operating experience, and Seabrook Station {
operating experience (New Hampshire Yankee procedure 12910)'. If an event is identified which has potential training' impact, a Training Development Recommendation (TDR) is initiated (New Hampshire Yankee procedure 18700). The' TDR is forwarded .to the Training Group where it is logged into a databaseandevaluatedbyteecognizantmanager(s). Items identified which affect the simulator result in the initiation of a Simulator Change Request. a The implementation of the change is then tracked in accord with Procedure NT-3734, " Simulator Change Control". 5.4 Major Simulator Upgrades Description Planned Completion > Install new core model Third Quarter 1991 h Instal1 new feed, condensate and main steam model Fourth Quarter 1991 (*)
%)
Tune critical paeameters to the plant values Fourth Quarter 1991 P 237 1
5.5 . Panel of Experts LThe panel of experts is responsible to perform a table top analysis of those transients which plant reference data ia not available. The acceptance criteria of Section 4.2 of ANSI /ANS-3.S is~ used. The panel of experts consists of the Simulator Systems Department Supervisor, the Simulator Engineer, the Design Control Coordinator, and the Senior Simulator Engineer. The resumes of each panel member are inclujed in Appendix F. 5.6 Plotting of Parameters The simulator is capable of plotting critical parameters. The time ' resolution of the plots is one second. This is less than the 0.5 cvond resolution specified in Appendix B of ANS-3.5. - The Simulator a.iew Committee evaluated this exception and deter mined that the one second resolution was sufficient for comparison to plant trends.
-y V
238
0 l Anoendices i i; s/ ~ A Initial Conditions (1.3.1) B Simulator Configuration Control ; C Outstanding Siula+0r Chaage Reques,- "$t) (4.3) D Simulator Performance 6 tid Qualification 7enting Schedule (5.1) E Simulator Review Committee Charter (5.2) F Resumes of the Panel of Experts (5.5) G Consolidated List of Exceptions
-1 1
i f e t O 4 e 'k i O 239
. . _ , . _ . . _ _ . . _ _ _ . _ . . - . _ _ _ . _ . _ _ - . _ . . -x- . _ . . ~ . _ . . _;_____. .._. _ _
IC Numbert i uant Updated: 07/24/90 Description MODE 5, ON RHR, BOL, READY TO DRAW A BL'BBLEl Primary Parameters Peactor Power ..... 0% RCS Tave . . . . . . . . . . . . . 150.9 ? Core Life .......... BOL RCS Pressure ........ 144 psig MW Thertal ........ 7 MWth PZR Level ............ 1004 RCS Boron ......... 2000 ppm AFD .................. N/A Xenon Conc ........ 32 pcm Control Rods ......... ARI Equilibrium Secor.dary Parameters MWe ................ O MWe Turbine Temperature .. Cold S/G Blowdown Flow .. O gpm Steam imp Mode . . . . . STM. PRESS L AUTO Running Equipment
,} Primary: Charging Pumps ......... A PCCW Pumpst ............. A B Reactor Coolant Pumps ..
RHR Pumpst .............. A Secondary: Main Feed Pumps: .... .. Condensate Pumps ....... Heater Drain Pumps ..... SCCW Pumpst ............. A B Cire Water Pumpst ....... Vacuum Pumpst ........... Service Water Pumps: .... A B
'^
Notes:
'._); Check T.S. for Mode 5 equipment tagout. Enter IC at step 6.1.3 of PZR s
bubble formation. 'A' RER in service, holding RCS temperature. 'B' RER in standby. SG 1evels between 30% and 40% in preparation for heatup. 1 ---
IC Numb 2rs 2 Lost Updated: 07/24/90
Description:
ON RRR, 2 RCPs RUNNING FOR Rx HEATUP / \ ( ) . . - . Js ,/ Primary Parameters Reactor Power ...... On RCS Tave ............. 307 r Core Life .......... BOL RCS Pressure ......... 359 psig MW Thermal ........ 6 MWth PER Level ............ 55% RCS Boron ......... 1355 ppm AFD .................. N/A Xenon "9nc ........ 16 pcm Control Rods ......... All Rods In Equilibrium Secondary Parameters Mwe .... ........... O MWe Turbine Temperature .. Hot S/G Blowdown Flow .. O gpm Steam Dump Mode ...... Steam Pressure Manual Running Equipment f Primary: Charging Pumps: ......... A (%x_)' PCCW Pumps: ...... 3.. ... A B Reactor Coolant Pumps ..AC RER Pumps: .............. A Secondary - Main Feed Pumps: ........ Condensete Pumps ....... A Heater Drain Pumps: ..... SCCW Pumps: ...........P. 8 Cire Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B
^
[ \ Notes: Enter IC at step 7.2 of MPE OS1000.01, Heatup From Cold Shutdown to 's_/ Hotstandby. Check T.S. for any equipment inoperability, i.e. ECCS equipment. 'A' train of RER maintaining RCS temp. 'B' train of RHR in standby. Turbine hot. vacuum in main condenser. 2
IC Number s 3 Last Updated: 07/18/90 _s Des:riptions BOL, READY Fors A REACTOR STARTUP 1
/ i I ~ , '
l Primary Parameters 1 Reactor Power ...... On RCS Tave ............. 558.2 F Cerg Life .......... BOL RCS Pressure ........ 2234 psig , 1 MW Thermal ........ 6.9 MWth PZR Level ........... 26.5% RCS Boron .......... 1066 ppm AFD .................. N/A Xenon Conc . . . . . . . . 13 pcm Control Rods ......... ALL SD RODS OUT Secondary Parameters , MWe ................ O MWe Turbine Temperature .. WARM S/G Blowdown Flov . 19 gpm per Steam Dump Mode ...... STM PRESS. AUTO Running Equipment Primary: v i j Charging Pumps: ......... A PCCW Pumps:~... ......... A B > Reactor Coolant Pumps: .. A B C D RIIR Pu m p s : . . . . . . . . . . . . . . Secondary ' Main Feed Pumps ........ SUFP Condensate Pumps: ....... A Heater Drain Pumps: ..... SCCW Pumps: ............. A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B l \- 11otes: Critical Data Bank D at 11 steps. Very early in core life, small or
\- negligible MTC. Feedwater in auto, SUFP to the feed reg bypass valves. Keep feed temp 200 - 230 using remotes 216 6 217 (EX STM to 26 FW heaters), remote value is 0.02.
3-
IC Numbert 4 Last Updated 07/17/90
Description:
REACTOR C.7ITICAL AT 1xE-8 i )
. <J Prime.ry Parameters Reactor Power ...... lxE-8 RCS Tave ............. $$8.2 F j i
Core Life .......... BOL RCS Pressure ........ 2237 psig MW Thermal ........ 7 MWth PZ'. Level ...........
. 26%
RCS Boron ......... 1066 ppm AFD .................. N/A Xenon Conc ........ 13 pcm Control Rods ......... D at 11 Secondary Parameters MWe ................ O MWe Turbane Temperature .. COL 3 S/G Blowdown Flow .. 10 gpm per Steam Dump Mode ...... STM. FRESS. AUTO Running Equipment
-[ GV; Primary: Charging Pumps ......... A PCCW Pumpst ............. A B Reactor Coolant Pumps .. A 3 C D RHR Pumps ..... ........
Secondary: Mair Feed Pumps ........ JDnP
-Condensate Pumps ....... A Heater Drain Pumps ..... -
SCCW Pumps ............. A B Circ Water Pumps ....... A B Vacuum Pumps ........... A B Service Water Pumpst .... A B {N Notest Little or no MTC value. Feedwater in auto with the SUFP supplying , y feed thru feedreg bypass valves. Remote function 216/217 for ex stm to 26 heaters are at 0.025. 1 4
IC Numbers $ Last Updated: 09/01/90 p_, Description BOL, LOW PO4ER, READY FOR FIRST MFP ( )_ -
% ./
Primary Parameters Reactor Power ..... 3% RCS Tave ............ 559.1 F Core Life .......... BOL RCS Pressure ........ 2234 psig MW Thermal ........ 100 MWth PZR Level ........... 27.4% RC3 Boron ......... 1066 ppm AFD .................. N/A Xenon Conc ........ 24.6 pcm Control Rods ......... D at 10 increasing Secondary Parameters HWe ................ N/A Turbine Temperatute .. Warm S/G Blowdown Tiow .._19 gpm per Steam Dump Mode ...... Stm. Press Auto Running Equipment +
./'~h ,
( ). Primary: Charging Pumps: ......... A- l PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RER Pumps: ............... Secondary: Main Feed Pumps: ........ SUFP
' Condensate Pumps .......A.
Heater Drain Pumps: ..... SCCW Pumps: ............. A B Circ Water Pumpa s : . . . . . . . A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B
-l# Notes: Ready to start-a MFP, with the SUFP maintaining 3/G 1evels thru the feed reg bypass valves. Maintain feedwater temperature 220 - 230 F by using aux stm to the 26A & B heaters with remotes 216,217. Enter IC at step 7.1.12 of OS1000.02.
5 ,_, __~ _ _ _ _ _ _ _ _ _, - ,
.. _ . _ _ . ~ _ . _ . _ . _ _ _ _
TC Cumber 6 Last Updated: 07/18/90
Description:
BOL, Turbine on line at 120 MWe
',\._/1 Primary Parameters Reactor Power ..... 15% RCS Tave ............ 562.6 F Core Life .......... BOL RCS Pressure ........ 2237 psig MW Thermal ........ 512 MWth PZR Level ........... 31.6%
RCS Boron ......... 1100 ppm AFD .................. -5.6% Xenon Conc ........ 39 pcm Control Rods ......... D at 100 Increasing Secondary Parameters ; HWe ............... 116.3 MWe Turbine Temperature .. HOT S/G Blowdown Flow . 18 gpm per Steam Dump Mode ...... Stm Press ! Auto Running Equipment -[' . Primary . Charging Pumps: ......... A
\,'
PCCW Pumps: ............. A B Reactor. Coolant Pumps: .. A B C D-RHR Pumpst .............. Secondary: ' Main Feed Pumps: ........ A ; Condensate Pumps: ....... A s Heater Drain Pumps: ..... SCCW Pumps: ............. A B 1 Circ Water Pumps ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B j Notes: Steam dumps maintaining a +3 Tave/ Tref mismatch. Enter IC at step J(_,) . 7.2.8 of OS1000.02, PLANT STARTUP FROM HOT STANDBY TO MINIMUM LOAD. Feed control is in AUTO, on the main feed reg bypass valves. MTC is slightly positive. Blowdown flash stm to 23C. 6
t IC Cumber 7 Last Updated: 07/17/90 ,
Description:
30% POWER ONE HEATER DRAIN PUMP ON i } N./ 3 Primary Parameters Reactor Power ......'29.9% RCS Tave ............. 564.1 T Core Life .......... BOL RCS Pressure ........ 2255 psig MW Thermal ......... 1011 MWth PER Level ........... 32.3% RCS Boren ....... .. 109Bppm AFD .................. -4.7 Xenon Conc ........ 132 pcm Control Rods ......... D at 162 steps , INCREASING 1 Secondary Parameters MWe ............... 291 MWe Turbine Temperature .. HOT S/G Blowdown Flow .. 15 gpm per Steam Dump Mode ...... Tave AUTO Running Equipment
~ *
[v~) Primary: Charging Pumps: ......... A PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RHR Pumps: .............. Secondary: Main Feud Pumps: ........ A Condensate Pumps: ....... A Heater Drain Pumps ..... A SCCW Pumps: ............. A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B ["w.'}/ Notes: Enter IC at step 7.1.11.2 of OS1000.05, Power Increase. Two backup heater groups in service for boron equalization. 7
IC Nutber: 8 Last Updated 07/16/90
Description:
BOL, 45% Power, Equilibrium XE Primary Parameters Reactor Power ..... 4S.5% RCS Tave ............. $69.6 F Core Life .......... BOL RCS Pressure ........ 2236 psig MW Thermal ......... l$41.3 MWth PZR Level ........... 39.2% RCS Boron ......... 917 ppm AFD .................. -2.98 Xenon Conc ........ 2133 pcm Control Rods ......... O at 177 Equilibrium Secondary Parameters MWe ................ 470 h.de Turbine Temperature .. Hot S/G Blowdown Flow .. 19 gpm per Steam Dump Mode ...... Tave Auto Running Equipment Primary: Charging Pumps: ......... A PCCW Pumps: ............. A B Reactor Coolant Pumps .. A B C D RER P9mps: .............. Secondary: Main Feed Pumps: ........ A Condensate Pumps: ....... A B Heater Drain Pumps: ..... A B SCCW Pumps: ............. A B Cire Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B Notes: B MFP in standby. Enter IC at step 7.2.8.3 of MPE OS1000.0a, Power Increase. 8
IC Numb]rt 9 tast Updated: 07/17/90
Description:
BOL, 75% POWER, EQUILIBRIUM XE ry . Primary Parameters Reactor Power ..... 75.5% RCS Tave ............. 579.5 P , Core Life .......... BOL RCS Pressure ........ 2236 psig MW Thermal ........ 2559 MWth PZR Level ........... 50.6% RCS Boron ......... 851 ppm AFD .................. -5.14 Xenon Conc ........ 2581 pcm Control Rods ......... D at 212 Equilibrium Secondary Parameters MWe ............... 852 MWe Turbine Temperature .. Hot S/G Blowdown Flow .. 21 gpm per Steam Dump Mode ...... Tave Auto Running Equipment Primary: Charging Pumps: ......... A N'.)) PCCW Pumps: ........ 4... A B Reactor Coolant Pumpa .. A B C D RHR Pumps: .............. Secondaryt Main Feed Pumps: ........ A B Condensate Pumps: ....... A B Heater Drain Pumps: ..... A B r SCCW Pumps: ............. A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B ri Notes: Enter IC at step 7.4 of OS1000.05 Power Increase t 1.
'us'
!= 9 i
IC Numbers 10 Last Updated: 07/10/90
Description:
BOL, 1004 POWER, EQUILIBRIUM XE
'] Primacy Parameters
\'
Reactor Power ..... 100% RCS Tave ............. 588.4 F Core Life .......... BOL RCS Pressure ........ 2234.6 psig MW Thermal ......... 3410.9 MWth PZR Level ........... 61.5% RCS Boron .......... 815 ppm AFD .................. -7.0 Xenon Conc ........ 2710 pcm Control Rods ......... D at 210 steps-Equilibrium 4 Secondary Parameters MWe ................ 1190 MWe Turbine Teeperature .. Hot S/G Blowdown Flow .. 21 gpm per Steam Dump Mode ...... rrave Auto Running Equipment Primary: Charging Pumps: ......... A [n%.)} PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RER Pumps: .............. Secondary: Main Feed Pumps: ........ A B -
' Condensate Pumps: ....... A B Heater Drain Pumps: ..... A B SCCW Pumps: ............. A B 1
Cire Water Pumps: ....... A B C Vacuum Pumps ........... A B Service Water. Pumps: .... A B I
/ Notes: Refer to MPE 0S1000.10, Operation at Power.
c)' 10
IC Numbers 11 Last Updated: 07/20/90
Description:
STABLE, CLOSE THE RER, 10 F COOLDOWN
, .g
- \ .
.]'
Primary Parameters Reactor Powet ...... 0% RCS Tave ............. 393.4 F
. Core Life .......... MOL RCS Pressure ........ 683 psig MW Thermal ........ 12 MWth PZR Level ........... 40%
RCS Boecn ......... 1104 ppm AFD .................. N/A Xenon Conc ......... N/A Control Rods ......... ARI ! Secondary Parameters ' r N' MWe ................ O MWe Turbine Temperature .. Cold S/G Slowdown Flow .. 10 gpm per Steam Dump Mode ...... Stm Press Auto Running Equipment
.rm.
Primary: f, i Charging Pumps: ......... A L/ PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A C RER Pumps: .............. Secondary: Main Feed Pumps: ........ SUFP Condensate Pumps: .......-A Heater Drain Pumps: ..... SCCW Pumps: ............. A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B
/T Notes: Prior to coming out of FREEZE one person is required to override the
( ,) _ Stm Dump blocks on low Tave or else the dumps will close as soon as the sim placed in RUN. Enter IC at step 7.5 of OS1000.04, Plant Cooldown from Hot Standby to Cold S/D. 11
IC Number 12 Last Updatedt 07/19/90.
Description:
MOL_ REACTJR STARTUP, SD RODS OUT, ECP D AT D=100 i
\
Primary Parameters Reactor Power ...... O tJS Teve ............ 558.3 P Core Life .......... MOL RCS Pressute ........ 2238 psig MW Thermal ........ 12.4 MWtn PZR Level ........... 26.4% RCS Boron ......... 744 ppm NPD..................N/A Xenon Conc ........ 3 pcm -Control Rods ......... SD RODS OUT Secondary Parameters MWe ................ O MWe Turbine Temperature .. WARM S/G Blowdown Flow . 21 gpm per Steam Dump Mode ...... Stm Press AUTO Running Equipment (Gs ,/ m
) -Primary: Charging Pumps: ......... P128 PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RER Pumps: ..............
Secondary: Main reed Pumps: ........ SUFP Condensate Pumps: ....... A lleater Drain-Pumps: ..... SCCW Pumps: ............. A B Circ Water Pumps: ....... A B Vacuum ? umps: ........... A B Service Water Pumps: .... A B i l L '(~h Notes: Shut down rods withdrawn, reactor goes critical at D at 100 steps. ( . l 1 12
IC Cumber: 13 Last Updated: 08/01/90
Description:
MOL, LOW POWER, READY TO START FIRST MFP {
'-- Primary Parameters '
Reactor Power ..... 2.9% RCS Tave . . .......... 559.6 r ' Core Life .......... MOL RCS Pressure ........ 2232 psig MW Thermal ........ 111 MWth PZR Level ........... 21.9% RCS Boron ......... 744 ppm AFD ........ ......... N/A Xenon Conc ......... 39 pcm Cont rol Rods . . . . . . . . . D a t 119 s t eps Increasing , MM - Secondary Parameters MWe ................ O MWe Turbine Temperature .. COLD S/G Slowdown Flow .. 21 gpm per Steam Dump Mode ...... Stm Press AUTO Running Equipment
- .[~T Primary: Charging-Pumps: . ........ A PCCW Pumps: ............. A B Reactor Coolant Pumps .. A B C D RER Pumps ..............
-Secondary: Main Feed Pumps: -. ....... SUPP
. Condensate Pumps: ....... A Heater Drain Pumps: .....
SCCW Pumps .......... 4 .A B Circ Water Pumps: ....... A B Vacuum Pumps . . . ........ A- B Service Water Pumps: .... A B I l- , g Notes Enter IC at step 7.1.5 of OS1000.02, Plant Startup from Hot Standby l t I to Minimum Load. Backup heaters C & D on for boron equalization. L PK-507'at 83.11. . 1 13
. a.
IC Cumber 14 Last Updated: 97/23/30 Descriptions TURBINE ON LINE t.T 120 MWe, r3 ( ) Primary Parameters Reactor Prwer ..... 14% RC3 Tave ............ 563.1 F Co r e L i '. e . . . . . . . . . . MOL RCS Pressure ........, 2239 psig MW Thermal ........ 503 HWth PZRLevel........... 3 2 . 4 '+ RCS Soron ......... 734 ppm AFD ................. 0 Xenon Conc . . . . . . . 96 pcm Control Rods ......... D at 136 steps Increasing Secondary Parameters MWe ............... 124 MWe. Turbine Temperature .. HOT S/G Blowdown. Flow .. 18 gpm per Steam Dump Mode ...... Stm Press. Auto Running Equipment [) Primary: Charging Pumps: ......... A
.%/
PCCW Pumps: ............. A B Reactor Coolant Pumpst .. A B CD RHR Pumpst .............. Secondary: Main Feed Pumpst ........ A
. Condensate Pumps: ....... A 1- Heater Drain Pumps: .....
SCCW Pumps ............. A B Circ Water Pumps: ....... A B Vacuum Pumpst ........... A B Service Water Pumpst .... A B g"'N Notes: Turbine at 120 MWe, Steam Generator control in manual. Maintain feedwater temperature above 250 F. Enter IC at step 7.2.9 of MPE OS1000.02, Plant Startup to Minimum Load. Tave/ Tref +6. 14
i IC Cumber: 15 Lest Updated: 07/17/90
Description:
50s POWER, 1 HOUR PACT PEAK XENot! TRANSIENT
. ,m
( -
. 's _.-)
Primary Parameters Reactor Power ..... 52% RCS Tave ............ 572.7% Core Life .......... MOL RCS Pressure ........ 2236 psig MW Thermal ........ 1774 MWth PER Level ........... 44.6n RCS Boron ......... 427 ppm AFD .................. -2.0 . Xenon Conc ........ 3405 pcm Control Rods ......... D at 165 steps Decreasing Secondary Parameters MWe ................ $81.0 MWe Turbine Temperature .. HOT S/G Blowdown Flow . 20 gpm per Steam Dump Mode ...... TAVE AUTO R'unning Equipment
.p\=
i Primary: Charging Pumps: ......... A
'v[
PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RER Pumpst .............. Secondary: Main Feed Pumps: ........ A B Condensate Pumps: ....... A B Heater Drain Pumps: ..... A B SCCW Pumps: ......-....... A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B. Service Water Pumps: .... A B Notes: Xenon transient past peak, both MFP's running. L 15
IC Number: 16 Last Updcted: 07/16/90 Descriptions- 50% POWER, EQUILIBRIUM XENON
,m
~ Primary Parameters Reactor Power ..... 51% RCS Tave ............. 571.8 F Core Life - ..... MOL RCS Pressure ........ 2235 psig MW Thermal ...... 1740.1 MWth P:R Level ............ 41.1% RCS Boron ......... 521 ppm AFD .................. -3.96 Xenon Conc ........ 2348 pcm Control Rods ......... D at 156 steps Equilibrium Secondary Parameters MWe ............... 575 MWe Turbine Temperature .. Hot S/G Blowdown Flow . 20 gpm per Steam Dump Mode ...... Tave , Auto Running Equipment
- ,~} Primary: Charging Pumps: ......... A I
PCCW Pumps: ............. A B Reactor Coolant Pumpsi .. A B C D RHR Pumps: .............. Secondary: Main Feed Pumps: ........ A B
-Condensate Pumps: ....... A B Heater Drain Pumps: ..... A B SCCW Pumpst ............. A B Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B Notes: Enter IC'at step 7.2.8.3 of MPE OS1000.05, Power Increase.
()T 16
IC Nutber 17 Last Updated: 07/17/90
Description:
75% POWER, EQUILIBRIUM XENON Primary Parameters Peactor Power ...... 77 % RCS Tave ............ 579.7 F Core Life .......... MOL RCS Pressure ........ 2233 psig MW Thermal ........ 2613 MWth PZR Level ........... 50.8 % RCS Boron .......... 484 ppm AFD .................. -4.7 Xenon Conc ........ 2527 pcm Control Rods ......... D at 202 steps Equilibttum Secondary Parameters MWe ............... 869 MWe Turbine Temperature .. Hot S/G Blowdown Flow .. 18 gpm per Steam Dump Mode ...... Tave Auto Running Equipsent Primary: Charging Pumps: ......... 4 PCCW Pumps: ............. A B Reactor Coolant Pumps: ..A B C D RHR Pumps .............. Secondary: Main Feed Pumps ........ A B Condensate Pumpst ....... A B Heater Drain Pumps: ..... A B SCCW Pumps: ............. A B
, Circ Water Pumps: ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B Notes: Enter IC at step 7.4 of MPE OS1000.05, Power Increase.
17
IC Number 18 Last tipdated: 07/20/90
Description:
100% POWER. MOL. EQUILIBRIUM XENON ! rN . i ) %J - l Primary Parameters Peactor Power ...... 100% .7CS Tave ............. SR8.4 F l Core Life .......... MOL RCS Pressure ........ 2236 peig HW Thermal ........ 3409.1 MWth PZR Level ........... 61.2 % . i RCS Boron ......... 436 ppm AFD .................. -3.4 r Xenon Conc ........ 2764 pcm Control Rodr ......... D at 215 steps Equilibrium Secondary Parameters MWe ............... 1187 MWe Turbine Temperature .. Hot . S/G Blowdown Flow .. 18 gpm per Steam Dump Mode ...... Tave , Auto Running Equipment p) ( s_/ Primary . Charging Pumps: ......... A PCCW pumps: ............. A B Reactor Coolant Pumps: .. A B C D RHR Pumps .............. r Secondary ' Main Feed Pumpsi......... A B ,
' Condensate pumps: ....... A B Heater Drain Pumps: ..... A B SCCW Pumps: ............. A B Circ Water Pumps: ....... A B C Vacuum Pumps: ........... A B Service Water pumps: .... A B
( \, Notes Refer to MFE OS1000.10, Operation at Power. 18
4 IC Numbers 19 Last Updated: 07/19/90 Description MOL REACTOR STARTUP WITH SD E NOT YET WITHDRAWN ir,) . ~ Primary Parameters Reactor Power ..... 0% RCS Tave ............. 558.3 F Core Life .......... MOL RCS Tressure ........ 2237 psig MW The rmal . . . . . . . . 12. 4 MWt h P2R Level ........... 26.4% RCS Boron ......... 744 ppm AFD .................. N/A Xenon Conc . . . . . . . . 11 pcm Control Rods ......... SD E INSERTED N/A Secondary Parameters MWe ................ O MWe Turbine Temperature .. WARM S/G Blowdown Flow . 21 gpm per Steam Dump Mode . . . . . . STM. PRESSURE AUTO Running Equipment (J,~'t~ Primary: Charging Pumps: ......... A PCCW Pumps: ............. A B Reactor Coolant Pumps: ..'A B C D RRR Pumps: .............. Secondary: flain Feed Pumps: . . . . . . . . S'!F P Condensate Pumps: ....... A Heater Drain Pumps: ..... SCCW Pumps ............. A B Circ Water Pumps ....... A B Vacuum Pumps: ........... A B Service Water Pumps: .... A B (N Notes: This is IC 12 with shutdown bank E inserted. U 19
IC Cumbar: 20 Last Updated: 07/16/90
Description:
100% POWER, EQUILIBRIUM XENON, EOL 3
)
Primary Parameters Reactor Power ...... 1001 RCS Tave ............ 538.2 r Core Life .......... EOL RCS Pressure ........ 2234 psig MW Thermal ........ 3402 MWth PZD Level ........... 61.3% RCS Boron ......... 101 ppm AFD .................. -0.61 Xenon Conc ........ 2770 Control Rods ......... D at 210 EQUILIBRIUM Secondary Parameters MWe .............. 1194 MWe Turbine Temperature .. HOT S/G Blowcown Flow . 20 gpm per Steam Demp Mode ...... Tave AUTO Running Equipment
/,T. Primary: Charging Pumps: ......... A %)
PCCW Pumps: ............. A B Reactor Coolant Pumps: .. A B C D RER Pumps .............. Secos.4ary: Main Feed Pumps: ........ A B Condensate Pumps ....... A B Heater Drain Pumps 1...... A B SCCW Pumps: ............. A B Circ Water Pumps: ....... A B C Vacuum Pumps: ........... A B Service Water Pumps: .... A B Notes - Refer to OS1000.10, Operation at Power. AFD moving negative. (V"'} 20
ruisem New amosiire Yankee B60RMATION ONL Y l SEABROOK NUCLEAR TRAINifMi PROCEDURE uter Identification Number NT 3734 Revision 1 Title Simulator Chance Control Originator Charlie Swinton Revised By Charlie Swinton Approval and Implementation Theattaciledprocedureisherebyapprovedandeffectiveonthedate anW W Training Gfoup Hanager'
- \
Apprbveh 'Date Form No. MT-1000-1 Rev. No. 1
Procedure No. N_T-3734 Page 1 of 23 Revisicn No. 1 l SIMULATOR CHANGE CONTROL
}
TABLE OF CONTENTS Section Tjtle Pigg 1.0 PURPOSE 2 2.0 APPLICABILITY 2
3.0 REFERENCES
2 4.0 O!FINIT10NS 2 e 5.0 RESPONSislLITIES 3 6.0 INSTRUCTIONS 5 i 6.1 Initiating a Simulator Change Request 5 6.2 Simulator Change Request Review 6 6.3 Simulator Change Request Information 8 6.4 Verification and Validation of Simulator 10 ( 'N Change Request b)~~ 6.5 Simulator Change Request Final Closecut 11 4 7.0 ATTACHMENTS .11 8.0 FORMS 15 ha [ v
Procedure No. NT-3734 Page 2 of 23 Revision No. 1 : -( 1.0 PURPOSE 1.1 To identify simulator deficiencies, enhancements, and plant design changes that affect simulator configuration and fidelity. t 1.2 - To . maintain an accurate status of the d*ficiencies, enhancements, and plant design changes as they are incorporated into the simulator design. 2.0 APPLICABILITY
- l This proctdure appites to all- simulator users.
3.0 BfffjlfKfji [ 3.1 NHY Nuclear Production Design Control Manual 3.2 Nuclear Production Rec ~ds Management Manual ' 3.3 Regulatory- Guide 1.149, " Nuclear Power Plant Simulation ,
' Facilities for Use in License Examinations". ,
C- 3.4 ANS!/ANS 3.5, " Nuclear Power Plant Simulators for Use in
.d Operstor Training".
3.5' NUREG 1258, " Evaluation Procedure for Simulation Facilities Certified Under 10 CFR 55". 3.6- 10 CFR 55, " Operators' Licenses". 4.0' OEFINITIONS J 4.1- .CurrentDesign_ Specification _(C0S) 1 The current: design specification is used to document the scope of simulation of a- modeled system. It is a living document maintained on the Simulator Information Management System. 4.2- DesignCoordinationReport-(DCR)
.The- documentation used by Seabrook Station - to initiate and '
track design changes to the physical plant. 4.3 . Simulator Information Management System (SIMS) A database used to store information related to the simulator. The Simulator Change Request tracking application of SIMS is O used throughout this procedure. 1;
Procedure No. NT.M 34 Page 3 of 21 Revision No. 1 4.4 Plant Design Change Documents ( Plant design change documents include DCRs. SPRs, or any other document which could potentially impact simulator ; configuration. 1
- 4. 5 SimulatorChangeRequest(SCR)
)
A work package used to track the implementation of a change to the simulator. l 1 4.6 Simulator Deficiency A-deviation in simulator performance or appearance that does
- not meet applicable tolerances and/or detracts from training fidelity.
4.7 Simulator Feedback Report (SFR) This form is used to document any identified problems with simulator response, to request improvement in the scope of-
- simulation of a system, or to request a change to simulator documentation.-
4.8 Simulator impact h V
- Simulator-impact refers to the need to update the simulator hardware and/or sof tware in order to maintain fidelity with the plant.
, 4.9 SystemProblemReport(SPR)
The- documentation used to initiate and track changes to the Main Plant Computer System. i 4.10 MinorM5dification(MM00) Minor modifications -are utilized to authorize minor, simple i changes to ' plant systems,- structures and . components. The total anticipated. expense for design and implementation of'an
- MMOD:is less than or equal to $10,000.
4.11. Document Revision Report (DRR) w Document revision ' reports are used to correct or update : engineering documents to resolve dist:repancies, m:Nr cer9rs,. l or inconsistencies. 5.0 RESP 0llSIBIL ITIES 5.1 Operations Manager Ah Ensures that a method exists to maintain and document V' simulator.compliar e with 10 CFR 55.44(b).
+ + + r 1 de . s V d'est- f ,=(pm *9 q "n 99 'i4 9 % M
- e- 9 g -*W4-= '* **+9 -s W-=s ---f3-= .*Wr"* v t
- s** qu 'e-
Procedure No. NT.3734 page.~4 of 23 Revision No. 1 5.2 Simulator Systems Department Supervisor Develops, implements, and maintains a program to identify plant design changes, simulator deficiencies, and simulator enhancements 50 -that the simulator complies with 10 CFR 55.45(b) and supports the needs of the Operaticas Training Department. 5.3. Operations Training Supervisor Designates a simulator support instructor and ensures that
- simulator users document known simulator deficiencies and/or simulator enhancements.
5.4- Simulator Support Instructor Evaluates completed simulator change requests to determine if the. SCR has been correctly implemented. Also, responsible to- : accept a simulator training load as being ready for. training. 5.5 Simulator Engineer : Schedules and tracks the implementation of scftware impacti.. simulator change requests. G . 6 _. Design Control Coordinator I O. ' .
' Maintains.the-procedures used to ensure that-the configuration
- and performance of the simulator meet the requirements of 10 .i
.. CFR55.44(b). -i
. 5.7 Administrative Aide l- ' Maintains the SCR files and performs SIMS data entry.
- 5.8 . Originator '
- Initiates a simulator feedback report to start the. simulator change process.
5.9 Assigned Engineer
- Makes"the sof tware changes necessary to correct a deficiency-
. or incorporate an enhancement.
L l+ ._ O r'armr-
"A4 W $ a vgw* 1 % +-
(- Procedure tio. RI, j ,2)! Page 5 of 23 j O Revision No. 1 4 ( 5.10 Simulator Technician j Makes the necessary modifications to the simulator hardware to ensure physical fidelity with the control room. 5.11 SimulatorReviewCommittee(SCR) Establish overall simulator priorities and determines whether
- an enhancement request wit; be implemented for training. ;
6.0 INSTRUCIM!Lii - 6.1 Initiating a Simulator Change Request ! A Simulator Change Request- (SCR) shall be initiated in response. t; a Simniator Feedback Report (SFR) or a plant design change document. . ' 6.1.1 Simulator feedback Reports I desired 6.1.1.1- Upon enhancement noting aorsimulator the need deficiency,ision for a rev to : 4 simulator Scumentation, the originator ' shall compic the information in Part I of . f form NT-3734 1. Atta*hment 7.1 -provides 1 instructions for completing Part I of this form. 6.1.1.2 The Design Control Coordinator shall review ,
.the' SFR. If the 'SFR identifies a ,
deficiency or documentation change, the DesignL Control Coordinator will complete-Part !! of form NT 3734 1. Attachment 7.1
. provides instructions ' for =' completing - Part J 11 of the . form. - ;If rthe SFR requests an "
. enhancement to,the scope of simulation the-request will be forwarded to the Simulator 1 Review Committee (SRC). The SRC will
' decide +hether or not to ' implement-- the e
enhancemont request based on a training l needs evaluation. If the SRC decides-- to '
. implement the enhancement request'. the' Desi n' Control Coordinator shall initiate- -
the CR. i 6.1.2 Plant Design Change Documents-f Plant _ Design Change Documents include DCRs, 2005, J ORRs, SPRs and all other documentation that. could > g impact simulator configuration.
~V- .
a _ . u .. , .;_._-_..._..~ .._....__,_..-.-._._-.__.__._.--_.._._4..._-..._
Procedure No. NT-3734 Page 6 of 23 Revision No. 1 6.1.2.1 Upon receipt of a Plant Design Change [V'l document, the Design Control Coordinator shall review the document for simulator impact and corriete form NT-3734-2, Plant Design Change Review, as appropriate. If the document does not have simulator impact, both the document and form NT-3734-2 shall be forwarded to the Administrative Aide for filing. If the document has simulator impact, the Design Control Coordinator shall check and indicate if the charige document is associated with an open SCR, and if so, indicate its number on form NT-3734-2. If there is ta open associated SCR, the Design Control Coordinator will assign the next sequential SCR number and enter the number on form NT-3734-2. Form NT-3734-2 shall be forwarded to the Administrative Aide for filing. 6.1.2.2 The' Administrative Aide will update SIMS to reflect whether each plant Design Change document has simulator impact or not, A using the information provided on form NT-L ('" ) 3734-2. If the Plant Design Change Document has simulator impact, the associated SCR number will be entered into the database. Two files shall be maintained. The first is the plant design change documents. These shall be filed sec"entially by number. The second file wiil be forms NT-3)14-2, filed in sequential order. 6.2 Simulator Change Request Review 6.2.1 The Design Control Coordinator shall assemble the SCR package. All forms shall be included within the package, and will be in -the following order:
, Simulator Change Request (fit-3734-3) . Simulator Change Control Design Data (NT-3734-7), Simulator Chenge Control Hardware (NT-3734-5), Simulator Change Control Software (HT-3734-6), Simulator Change Control Test (NT-3734-4), and either initiating EFR or the Plant Design Change document.
6.2.2 The Design Control Coordinator shall enter the following information into SIMS: g) ('~ 1.SCRHumber(assignnextsequentialnumber)
- 2. Date u______________________ . _ _ _ _
_ _J
Procedure No. 3T 3 4_ page 7 of 23 Revision No. 1
- 3. Source of change (STR, DCR, etc.)
~
~
- 4. Description of the change.
5.. Status (R for_. review) P The assembled SCR package will then be forwarded to the Simulator Engineer for 5- tware review and the status board will be updated (using a dry erase marker) to indicate the location of the SCR package. 6.2.3 The Simulator Engineer sheli review the SCR for software and hardware imin c t and update the appropriate information- on form NT-3734-3 'and NT-3734-6. If the SCR has no hardware impact, a line will be drawn across form dT-03734-5 and "No Hardware Impact" written in; then forward the - SCR package to the Design Control Coordinator. If the SCR has hardware impact, forward the SCR package tt the Simulator Technician &nd update the_ status-board to indicate the location e' the SCR package. If
~
the SCR was initiated in response to a design 1 change, the Simulator Engineer .will copy any
. appropriate _ pages from the- design change. - -(These; shall be - marked "Information Only". ) The _ design change itself will _be forwarded to the.
e 6.2.4
.Administr?tive Aide to be filed, The Simulator Technician shall review the SCR packag and complete the ' appropriate information on form hN3734-3 and NT-3734-5, Upon completing the appropriate - section of f - form, the Simulator Technician shall . forward e 'CR package to _ the Design. Control ~ Coordinator .d update the status
-board.
6.2.5 If the SCR - has no imp;ct, the Simulator Engineer 1 .shall forward the SCR package to the Design Control < Coordinator. The Design- Control Caort.:nator will indicate no -test is' required on ~ form MT-3734-4 ande
- forward the SCR package -to the Simulator Systems Department Supervisor for final review and closeout.
6.2.6 The Design Control' Coordinator shall review the documentation = attached to the - SCR package. The Design Control Coordinator will enter the-estimated
- r. umber of man - hcurs ' required to write and conduct the test procedure on form NT-3734-3.
Procedure No. HT-3734 Page 8----.of *) Revision No. 1
/ 6.2.7 Information necessary for the retest shall be
\ .. collected by the Design Control bordinator. The i 1 retest should be- written usi g form NT-3734-4, i
) Simulator Change Control Test. ~he retest procedure } may be written any timo prior to performing the acceptance test on the simulator. The Design Control Coordinator shall forward the SCR package to ;' the Simulator Systems Department Supervisor and-update _the SCR status board. ,, [ 6.2.8 The Simulator Systems Department Supervisor shall i perform a final review of the SCR package to ensure [ all required information has been entered. The SCR package will then be forwarded to the Operations-Training Supervisor to be reviewed and prioritized, c 6.2.9 The Operations Training Supervisor shall- review the s SCR package, assign the priority and return the i package. to the Simulator Systems Department-j Supervisor. Priorities shall be assigned based on
, the following criteria.
Priority 1: The problem greatly hinders or limits t"e ability- to conduct r, training.
..- Priority 2: The problem adversely affects the ability to conduct a reliable:
~
scenario on a given procedure, event or system. - Priority 3: The problem has a . but definite impact. Priority 4: The problem has little or no adverse.e Tect on training.
- 6. P .10 - The Simulator Systems Department - Supervisor shall' -
forward the SCR package to the Administrative Aids. 6.2.11 The Administrative Aide shall stamp the SCR package
" Implementation Package", and update the SCR tracking sof tware status to 'E' for an enhancement 7 -or 'I'.for any other SCR._ A tile shall be created .
for the status-board which has the SCR number and a-' short description. The color of the tile will indicate the priority of the SCR.. Tr.e tile shall be
) laced in the implementation column on- the status 4 '
aoard. Finally, the SCR pa_c: age will be filed:in the active work section of-tF.e filing cabinet. 6.3 Simulator-Change Request Information _ -6.3.1 The Simulator Systems Department Supervisor shall 9 set the simule. tor work priorities. The status board wil) be updated to reflect the implenentation status
'l y*
. Procedure No. NT-3734 P 6- i_of23l a Revision No. 1-o
-)
-s and-wi,ll also inJicate who is in--possession of each: W i l5CR package throughout the implementation process.
ce 6.3.2 The . Simulator- Engineer shall -coordinate the
-implementation of SCRs that indicate hardware and 3
, software impact.. .,
t' 6.3.3 For- hardware . impacting -SCRs, _the_ Simulator. t 1 Technician shall complete the Instrument Changes J'
&' section~of form NT-3734-5,. Simulator Change Control Hardware,-by completing.-the following steps: ,
6.3.3.1 Enter information as necessary:.in. the simulator- instrumentation, wiring T;>d- ,
, comments.section.
g" _6.3.3.2 Complete the Cosmetic Changes section.-- ;
- 6~.3.3.3 In the Comment section, explain any-detail: 1 that requires-further attention.
6'.3.3.4: Complete the. Documentation section 25
'necestrry.
m 6.3.3.5_ '. Update ~ the C eseout .-section of form NT-3734'-3 indicating actual ' man h o u r s ,- ' t! . fN - simulator hours, initial and date. .TJ a 6.3.3.6 The data which-' formed. the ' basis for the- '. change 1shall: be: listed on form NT-3734-7, 4 Simulator-Change Design" Data. (IndicatoNA
, y. . . ,
..if.notapplicable.) s,
, 6.3.4 The; Simulator Engineer shall assign an engineer-to?
a,
. make Ethe L sof tware : changesf required and update d the p Estatus . board to: reflectCwho'. possesses ithe SCR -
package..
, 4 i 6 '. 3. 5 ' For sof tware; impacting LSCRs,- the- assigned engineer 5 '
~w il1 complete, forr NT-3734-6,. Simulator ~ Change
, Control Software, by~ comp M ii.; the following steps.. ,
'f ~.6.3.5.1 Describe' the: changes 'made to the. modelin . .y the Change Description section. ' y^ . 9< 6.3.5.2 ' Indicate = the .. completion ~ dates for. the' 1* g + affected utilities. L6.3.5.3 Complete the 'Datapool and ' Computer. Points - section as applicable,
- 6.3.5.a . Update all affected documeatation and
- (99); ir.dicate the completien date, o u 1 s i OL1s , i - nt , . a
Procedure No. NT-373a Page 10 of 23_ ! Revision No. 1 D V NOTE: If' copies of controlled documents are included with- the SFR, - the copies shall be stamped "Information Only" in accordance with the Nuclear -l' Production Records Management Manual. 6.3;6 The assigned engineer shall forward the SCR package to the Simulator Engineer and update the status , board.. 6.3.7 The Simulator Engineer shall: 1 6.3.7.1 Ensure that the appropriate utility programs and data files are updated-6.3.7.2 - Ensure form' NT-3734-4 is complete and ) perform an initial test of the SCR. , 6.3.7.3 If- additional software / hardware work is necessary, return the SCR package to the j responsible engineer / Simulator Technician, , 6.3.7.4 Upon satisfactory completion of the initial; test, update the SCR tracking program' to indicate the SCR:is in test (TR). 6.4 Verification and Validation of Simulator Change Request-6.4.1- One the Simulator Engineer has assembled one or ; more Simulator Change Requests which are ready for i final-acceptance, a unique: software development load , will be built that contains- only the changes ready a for acceptance.- The SCR packages for -those changes will be given to the- Simulator Support _ Instructor , for acceptance testing. 6.4.2- .The Sim, tor Support Instructor - will retest the SCRs usin3 the attached -retest procedure. The Simulator upport . Instructor may perform any
- l
-additional- sts for each SCR that are -deemed appropriate, a description -of all such . tests shall.
be attached to the affected SCR package, if L the- 3
-retest-of the SCR is satisfactory,-the closecut and I
training : review sections shall be updated and the ". SCR package returned to the Simulator Engineer. If the retest is: not satisfactory, the ' SCR package shall be; returned to i the Simulator Engineer for correction. --
- 6. 3 If all SCRs in the development load. are -_ accepted, the Simulator Support Instructor will verify the
. integrity of the load has not been compromised by performing the annual operability tests described in
9% ~ .
, q-O <
' Procedure.No. NT.3734~ Page 11 of-23l N .
Revision-No. 1 + i
?
.' t 1 procedure.NT-3738. Upon satisfactory completion of 7~ y' the' annual operability t e s ,t.s ' :all -initial
<(% 1/? conditions shall be updated. Once the ICs have been .
--updated,. the load will be saved. into the training:
m ~
' load location.
)
6.5 _ Simulator Change Request Final Closeout o 6 '. S .1 - The Simulator Engineer shall ensure 'all-information 'i has been updated- on' fo ms NT-3734-3' and NT-37346a ~ ' and.6b. Upon turnover for training, form NT-3734-3 shall be completed. The SCR package shall-then = be- , forwarded ~?to the Simulator Systems Department Supervisor.for final? review and the'SCR status board
,M '
shall-be updated. ! l 6_.S.2 ' The Simulator Systems - Department Supervisor- shall ensure that; thef SCR package has been completed. Satisf actory - completion will . be indicated- by initial.ing and : dating the Final Review section- of a form NT-3734-3. e 6.5.37 The' Administrative Aide. shall'ensu'e that all-pages i F .in- the SCR Jpackage are M/abered correctlyL update - SIMS'to showLthat the SCR _.., closed, remove'the tile T. u- ' from the status board =and! file theiclosed .SCR f ;:N package in-the simulator. change." closed?. file. 1
;ywJ.l
- 7.0 2ATTAC}MENTS w.
]# < L 7.= 1 - SimulatorFeedback-Report (SFR)' Instructions
[.' . f 7.2 SimulatorChangeRequest(SCR) Instructions mm j!n ' ' '
, -i
'f -
4 b b $ y I ( 4 I
, , , W , ,
i nm %[-;.-. 1 I' -- -
o Procedure No.-NT-3734 Page 12 of 23 , Revision No.- 1 1 SIMULATOR FEE 0BACK REPORT (SFR) L(~% INSTRUCTIONS V RESPONSIBILITY- ITEM TASK PART I > Originator a. Fill in name, date, and telepi ne extension
- b. Fill in the starting IC and lesss plan (if _
applicable) and briefly describe the. transient, evolution or procedure,
- c. Describe in detail the' deficiency, enhancement j or desired documentation change. If the problem was stored in a temporary snapshot, including it's number.
l
- d. Describe the expected results.
4
- e. Include the document number of the data used as
-a reference for tne requested change.
NOTD If-controlled documents are included with thr: SFR, the' copies shall be stamped "INFORMATION ONLY" in accordance with the Nuclear Porduction Records Management Manual. * @).-
- t
- f. Place the SFR in the box-by the computer room
-door.
r PART 11 r
~
Design Control- a. ' Review and determine if the SFR represents a ,
; Coordinator deficiency, enhancement, or. document change check the appropriate. box.
- b. For a deficiency that has an associated open SCR, check "Outstandir,g SCR".and indicate the associated-SCR number. If the deficiency is newi. check."New SCR" and indicate-the,SCR numo r.
- c. Provide -detailed consnents under the Disposition - j section.
- d. If unabie to determine the impact of the SCR,_
forward to the Simulator Review Committee for eva'luation,
- e. Forward a copy of the completed SFR to the w originator.
V( \- ATTACHMENT 7.1
.;r
Procedure No. NT-3734 Page 13 of 23-Revision No. 1 y SIMULATOR CHANGE REQUEST (SCR) INSTRUCTIONS < RESPONSIBILITY ITEM TAlf Design Control- a. Assign the.SCR number. Coordinator
- b. Enter the current date-in the space provided,
- c. Enter the source of.the change, i.e. SFR, DCR, SPR.
NOTE: If the~ source of change is.a Plant Design Change, enter the associated number. _(Not applicable for SFRs.). 'i d.. For DCRs only, use-the-OCR Status Report _toJ determine status (Plant Impleme-tation Date)
'of the OCR. -Indicate:this information, if available', " Plant Implementation Date".
1l' e. Update SIMS. , f.- Forward the_SCR package to_the Simulator Engineer.
- g. Update the status board.
E -Simulator ~ a. Review and determine if there is software ori a'
-Engineer hardware impact.
i
- b. Enter-all 7information reauired in the software-review section.
4 ," , c. : Enter any.amplifiying comments as d'eemed- l necessary. d.- If there'is.no softw?re-impact, enter N,; 1
~
initial and date.-
- e. If there is no hai&:e impact, check "N" in the approp)iateLspases, initial and date.
,~ [ f. Update SIMS.
(" .g.- _If.' software impact, forward to the Design; Control Cordinator. y:
- h. For h vdware impact'-, forward the SCR to the 4 Simulatar Tech for review.
B l .])- s'V L ATTACHMENT-7.2' F h j '
Procedure No. NT-3734 Page 14 of 23 Revision No. 1 s SIMULATOR CHANGE REQUEST (SCR) +
) INSTRUCTIONS s;
RESPONSIBILITY ITEM _ TASK Simulator i. If the SCR has no impact, forward the SCR Engineer package to the Simulator Systems Department (continued) Supervisor for final review and closecut.
- j. Update the status board to show who is in possession of the SCR package. '
Simulator a. Enter all information required in the software Technician review section.
- b. Enter amplifying comments as deemed necessary.
- c. Forward the SCR to the Design Control Coordinator.
- d. Updete the status board to show who is in possession of the SCR package.
r
%,J M% ;
i ! ATTACHMENT 7.2
'#~
(Continued)
Procedure No, NT-3734 Page 15 of 23 Revision No. 1
-8,0 FORMS LIST OF ASSOCIATED FORMS Form Number Title Revision Number NT-3734 1 Simulator Feedback Report 1 NT-3734-2 Plant Document Review Form 1 NT-3734-3 Simulater Change Request NT-3734-4 Simuli. tor Change Control Test 0 NT-3734-5 Jimulator Change Control Hardware 1 NT-3734-6a Simulatcr Change Control Software 1 (Page1)
NT-3734-6b Simulatot Change Control Software 1 (P Je 2) NT-a/34-7 Simulator Change Control Design Data 0 g-i
g x -o , cAv
- Procedure lNo. NT-3734: Page 16 of-23!
Revision No. )
' <.-/"
t SIMULATOR FEEDB/,CK REPORT 1 PART-I-i
-f Originator.- =A)-Name: Date. Ext.:
.3) State of'theTsimulators- Starting IC.
Lesson Plan #- Plant System Transient, evolution. etci 1 I i t, 1 s C); Deficiency ,' Enhancement , ' Correr. ant , ete: Temporary Snapshot i ' l
~i i
~ ,.
't 1 l
1 t . . .. m-D' ) -:... Expected'or' Required:Results: .i a '
.5
.4
...m# -. .
(E)Ll References (i.e.,"Ref. Doc. f. Procedudre i, Other) -
~
- {
f s1 - PARTlII[ {'. Desig'n' Control- Coordinator - 1Date Reviewed: 'SCR No.:: I Thisl1sYati De ficisney .- Enhancement-~ :Other Out s tar. ding : SCR:- ,
- New-SCR ;
- !- ' Dispositions; '
i .
,. 4.
g;1 ,
,n .. \; > s >
l j,ib l Copy returned to originator - ai Form No. NT-3734 N ( p)) 99 Rev. No. 1 p ' i f o .
) g
_, _ _, l} l .
, a.
Procedure No. NT-3734 Page 17 of 23 Revision No. 1 Pt. ANT DOCUMENT REVIEW FORM
)
./
l-
; Document Type:
OCR NUMBER C/A MM00 NUMBER REVISION SPR NUMBER _ REVISION DRR NUMBER REVISION Review for Simul 6 tor Impact: i IMPACT NO IMPACT OUTSTANDING SCR ATTACH TO NO.-
^'), , -, , . . - , _ . , _ . NEW SCR NIJMBER i Initial Date NOTES:
)
I I l, x_'- Form No. NT-3734-2
,'. ) l Rev. No. 1
Procedura'Mo. NT-3134. Pago 18 of 23 Revision No. 1 gs i w.J = SIMULATOR CHANGE REQUEST SCR No. Date Plant System Source of Change Number (NA for SFRs) Priority Plant !mplementation Date _(NA for SFRs) Initials _ Description of Change Date _ 1 Hardware Review Software Review Affected Impact 'Y/N Impact Sim. Syst(s) Parts Required Y/N Schedular Y/N PCM Y/N Interface Design Y/N Syst. Data Files Y/N DORT YlN Switchcheck YlN Executives Y/N Wirelist Y/N Handshake Y/N I/O Override Y/N
~s Total: Datapool Y/N
; ). Affected Remote-
'~'
Estimated Manhours TOTAL.: fusction number ,_ Estimated Manhours Affected Malfunction Initial Date Initial Date , Number Test Review: Estimated manhours Initial Dato SSD ' Review Initial Date COMMENTS' d CLOSEOUT: l Actual Manhnurs In1Hal Date Software Test _ Simulator Support Instructor lAcceptedforTraining_ Initial Date _ F(SAL REVIEW Simulator Systems Department Supervisor Initial Date J Form No. NT-3734-3 Rev. No. 1 i
Procedure No. NT-3734 Page 19 of 23 Revision No. 1 SIMULATOR CHANGE CONTROL TEll SCR No. Page of Utilities Not Required Required Date Complete OORT Switchcheck ~~ I/O Override Retest Requirements j l
=-
v Retest Completed Satisfactory Date Documentation Cause and Effects Not Required Date Completed Form No. NT-3734-4 Rev. No. O
Procedur3 No. NT-3734 Page 20 of 23 Revisi n No. 1 3 I SIMULATOR CHANC. CONTROL HARDWARE SCR Ho. Page ___ of Procurement Information MPR Date Date Number Description Quantity Ordered Received interface Device Not Required _____ Required Complete Simulator Instrumentation Add / Delete / Repair Device No, Action Taken/ Comments Simulator Wiring i t/ I/O Sub- Half-C2 ete Unit # Classis # Connector # Dec # Type Contr # Word Bit Cosmetic Charges Not Rcquired Date Completed COMMENTS Mimic Labels . Escutcheons , , Annunciator Windows __ Ronan Lense Painting Cover Plates Documentation Update Panel Drawings Not Required Date Completed
' ~) Form No.
Rev. No. HT-3743-5 I
Procedura No. NT-3734 Page 21 of 23 Revisio'. No. 1
) SIMULATOR CHANGE CONTROL - SOFTWARE PAGE 1 l REVIEW & SCOPE MODULES:
l l
Description:
i l IMPLEMENTATION Assigned Personnel: Change
Description:
1
-s Utilities Updated: )
Not Date Not Date Required Complete Required Cenplete Scheduler PCM System Data Files DORT Executive (s) Switchcheck Handshake Wirelist Data Pool . 1/0 Override . l Data Poci A.D.M. Variable Ref Offset Type / Size Module Description
< ,'i
-/ Form No. NT-3743-6a i
Rev. No. 1 _
4 Procedure No. NT-3734 Page 22 of 23 / Revision No. 1 . O-SIMULATOR CHANGE CONTROL - SOFTWAF.E Implementatioa (con't) Computer Points l l A.D.M. Variable Ref Offset Type / Size Module Description l h3 O - 00CUENTATION: Not Date Required Completed Simulator Scope Drawings Current Design Specs SIMS - Design Spec 3 Design Data Instructor section Device List / wire list Process Comptier Points Data Pool REVIEW Initial Date Sections above complete: Accepted in a development lead Into training lead , 0l Form No. NT-3734-6b Rev. No. 1
; [
- i Procedure No. NT-3734 Page-23 -of'23-Revi_s ion _ No.- 1-I
. SIMULATOR CHANGE CONTROL _ DESIGN DATA y
SCR-No.-- Page __ of- , Document- Document Revilion - Associated--
; -Type Number Revision ,,.0.,f t e Document-Tit 1.g Sim. System;
'l
._ _ i 1
a
.q
)
1
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-i 0 ..
9(6 E, N/ .
)
A DESIGE Data EXA'MPLES ej q
- Typt - piscRIPT10N 'sPMAfi - JJ,1, ; DESCRIPfl0N FORMAT
--:, : w
-..; d
. AT . Acceptance Test 1AT.9h,99L ( Mi,c _ Miscellaneous- '!
I, ~ iCategory _
~
??-
CAL- ' . Calculation =CX 1 9999 30- . Plant Performance-Data -PD YY-9999- _CWD _ Control Wiring P it, Piping & Instr.. Otagram' CWD-9 XXX.999 Otagram .P10-1-XX X9999 LOCR- Detign foirdination. Report- f99 9999 CA99 !P? ~-PreoperatWaal- 1 PT-99.99; FF _ Foreign Print- FP.99999 ' Test
.SCHM Schematic Olegram'1.MHY.9999f9'-
,, - GET .Getars Data GET-YY.9999 SPR . . System Problem RPT.YYXX9999 ;
.-LOGIC -Logic diagrami 1-NHY.999999 TEL- 'Teleconference TEL-YY-9999 LOOP ' loop Olagram 1.NHY.99999 -
1
-1 1
Gm -- Form nc MT-3734-7
- -\ . yj 'Rev. No..
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, 'SCR TRACKING:"0 PEN:SCRS, ST;$CR NUMBER ' SORT PARAMETERS:'5TATUS,5CRacM' - . aEP0er s: Sta-a52' s SC R , SIM-
#98!gE! - (HHHy T ?)HlM - }HTEM - 91 SHIPT 10m 89-149 4' ;D TC iMAlm TUa81NE ROLLS OFf 0F~TuenthG GEAR 100 EASILY DURIh4 PROPEptf Sche:
SMELL" WARMING.. 154._ 3 8' CR~ =GET POSITIVE-RATE TRIP 04 920PP14G RCS H-M FROM MOL /1001 s ~ EOL 1001. . 89-159. L4' S1 C4 'A88 EFFECTS OF CCAE WOISING:AND'ACCOMPAh'_I*:G INCREASED FAST NEufach FLUE.AT: SOURCE'mAh6E'SETECTORS. 59-160. 3 .D SC Ass'MORE MODCOMP NIST0af FILES'(E.G. LEAEAATE) .To Pat LOAp #00 TINES Oh'SIMOLATOR RESST.~ 49-168' '3 B AC M0eULE 30M6 sU4ING INADEeuETE CORE C00 Lima WITM 2 PMASE F L 0ts. 09-177~ -4"' D' SG ASD WARIASLE-MALFS:TO FAIL PT 455, 45e ,-457 5 459. 39-179- 4 S Fw ADD MALF~10 PREVENT SUFP BRER($) # ROM CLOSING - OPEN IN SE4IES Cat FCs ;
'52m:CcIL. ~(MALFumCTIOm'A5e).E Jl i4 89-180 4 e Fh. -Att REMOTE FueCTION TO SUPPLT FIRE WATEa TO EFh PUPP:A SUCTIO9.
49-187- .4 5- MC9 GAITROWICS ALARM PAkEL 04 CP-295 BOES NOT MATCH PLANT.. 89-191 3 9 31 . ADS:"A" POINTS FOR 31/CS CMECE"VALwE LEAKAGE MONITORING.. 89-195 4 ' D .. TC aEMOWE' MAIN GFNER4TCR FIELS TEMPERATURE IhDICATOP ED-TI-9902 C #EPLACE , IT*S'FubCTION h/A' CALCULATION TO DRIDE C1100 $ 3559% Ib MotCOPP - MCB -; SEMakC ATION/MINIC : CNANGE af eUIRE D. ' 89-194 3 D Es Aes REMOTE: (MCB 4 ESCC),SaKR CLOSE- CohTROL OF SV SAKas 163,169,632,692 4 941. 4 89-200 '4 D TC me GSC u!Gm TEMPERATURE RUWWAtt AND GEhERATOR TEMetaAfuaE SCAhMEa ALARMS ON MALF 69 On GTAER CAUSES OF GSC NIGH TEMPERATURE.- 89-203 4 9 MS ~ REVISE MAIN STEAM ATMOSPhEAIC STEAM FLod 201SL 10 !#ClupE PNTSICAL
- LOCATION OF NJISE SOURCES..
89-207 3 e FW Ass'aEMOTE FueCTIomS To MANUALLY REPOSITICN EFW PuhP RECIRCS FW-V-346 .; 8 347.. .-t 49-212 4. 9 ES MOsEL EPS'SEautaCEa($) .s* POINTS AS WEEDES. 89-213 3 e 'CV ' sew!SE CMAAGIh6 WALWES*' ADMITTANCES TO plant I ts F O. - 90-006 .3 e FW MODIFT WSP DISCNARGE WALWE CONTROL. 90-007 4' D' NCS- CORRECT 9EFICIEmCIES FOUNs IN #0VEM&sR#3ECEM8Ea 1989 HARDWARE AUDIT. 90-008 3 D Fw' MosIFT Co-FV-4042 Ano -ITS CONTa0LS FOR IMemowEs STAeILITV ama mEcuCED h A T E R - es AMM E R. 90-013 4: DL CW spa 89PP0015 ADS A POIhTS PEA #Cs 89-0012 (SEE Sta 69-191 FOA 3CR). 90-019 4 3 St. iWALUATE HEED FOR ASSITIONAL PRELOADIhG OF MotCOMP MIST 0ay ARRAFS,
^
, SETERMIhE SCOPC AND IMPLEMEFT~AS NEECES - SAC MT6. AUG. 24, 1949 90-022 3 e an- 1Asa sActuP-ConTAlhMEmi GAsious Moh1 TOR. d
.90-024 4 -e- .CN. Ass 3ACKte METEOROLOGICAL MonITCA!nG STSTER. ,
9C-027 .3 s' CD REPLACE'UPS-48-2-4. . . 90-038 ~3 s- AM cCh asMS TREhD B14PL A WS - BM-6509 (CN A#mE L 1622) S now GREEN FOR LAFEST eATA'dAa:bui AES FOR ALL "0LD BATA* bass.
'90-049 4 > SC. COL 3h GaAPNIC.1Fv1 h0T' CURRENT b!TM PLANT *$ WEaSIJN (kEED TO INCLUDE i CGS In FlaST 1990 COMPLETE MeeCOMP UPSATE)..
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' REPORT s: $CR-R$2 sCR SIM gg3al! EB12111I '11AIVI 111115 'tiSERIPTIch __
90-074 -4 9 Cu CW STSTER TEMPERATURE RECORDER. 90-083 4 8 MS ADD REMOTE FU'CTom m & EATRACTIONS STM NON-RETURN VALUES TESTING. SAC
-OK 90D430.'
90-087 3 9 ED 'D' EDG BREAKERS DID NOT ' 0 PEN JN A TAAIN I'89V. A SI (MALT 8114) WITH bY-LDP. 90-058 4 D RM *B' 56 IhDICATED 8 tows 0WN ACTIVITT MUCN LOhER THAN CTMER !. 90-991 4 9 ED 95301 30E5 WOT SET Ch Loss CF INST. P&C 1 3 (PALF 131).. 90-092' 4- D FW ADD REMOTE FUhCTION FOR CONTROL OF Fk-W-o$. 90-097.' 4 9 MS ADD SVPh$$ WALVES FOR M$R ISCLATION M3v's TO SCOPE VIA REPCTE F0hCTIchs. 90-098 4 9 CO ADD AER0TE FUNCTICh FOR CD-V-67 (ISCLATIch FOR CD-LV-40144 Du=P 10 CST).
- 90-107- 4 9 MS - DUR!as M sIW STROKE YEST, $1MGLATOR REACTIchs TIME FOR *1UE C t*
LAMP !$ *MbCM FA3fER* fnAN IN THE PLANT. 90-118 3 D Fw RESCALE EFk FLOW INDIC ATION AhD CHAhEE E Fw AUTO 15CL ATION SETPOINT. 90-121 3 p SG AaD SPRAT CONDEh5ATION EFFECTS TC SG'S. 90-12; 3 9 Fw ADJUST FEED PUMP SPEED CONTROL. 90-128 ~3 9 TU . ADO TUa8thE TulF 10 MANUAL GENERATOR TRIP. 90-129 3 D CO ADD Ncf wEh! PAld. 90-130 3 0 ' SC' .vAs HTPASS/INOP LOGIC CHANGES. 90-131 4 D as 899 4LOhDokN FLA5M 1ANK SUSC00 LING INJECTION L INE. 90-112 4 0 WP RESCALE SPEnf itFL POOL LEWEL GAUSE. 90-13h 4 0 56- ADD LAG FCR STE39 FLOh/ FEED Flow MISMATCn In 5/G LEVEL ComTROL. 90-134 4 D CO lMODIFV RhBT'MI LEWEL DUMP WAltE INDICATION. 90-135 4 D . CC -CHANGE'PCCW LOW FLOW ALARM SETPCINYS. 90-136 3 D C0' HEAfER DRAIN SYSTEM ENMANCEPERTS. 90-137 t D Nt. CHANGE 80Roh DILUTION Moh! TOR READIh6 10 MATCn YnE PLA=T. 90-140 4
- MCS INSTALL SCREWS 10 mI'4AMEPLATES.
90-146 4 D Fu CCRRECT SUFP' TIME DELAY kNEM STARTIh& On aus 5. 90-152 3 D- 56 CHAhEE 55 DTNAMICS 53 LEVEL RESPJMD5 AS IN TME PLANT FOLLOWIhG A RN TRIP. 90-156 4 D CC. MO91FT IhPUT 1 SETPGINT SCALES FOR CC-TK-2171 & 2172. 90-157 4 D BD MODIFf thPUT 4 SETPOINT SCALET FOR 10-tK-1912/1913. 90-158 4 'D tv MCDIFY ImPUT & SETPOINT $CALES FOR CS-FE-375. 90-160 4 c. IA ikSUFi!CIEhi CCOLIkG CAUSING'3A-C-10 TO TRIP. 90-161 4 .0 PCM CORRECT PCM DCSCRIPTION of MALFs38. 90-163 4 D MCB ChahEE huCLEAR' ALERT PHof.2 FACM AUTO RING TO DIAL rih 6. 90-165 4 D SC . REVISE POST MORTEP ACTIWATION POINT LIST. 90-166 3 9 SC MCDIFT SPOS C0hTAINPEhT !$0LATION'TO TAKE VALWE $14705'$ INTO ACCouhf. 90-167 4 D SW ~REWISE.CTP'3ISCnARGE VAlbE STROKE' TIME. 90-tas 3 0 TC CHANGE TURothE 6EhERATOR WIeRATICM TRIP AND ALARM 3ETPOIhTS. 90-169 4 D FW CORRICT. Flow CALCULATION FCR'nD-FI-14518 & 4519 Ago SCALES Foa Tutst FI*S. 90-1F3 3 'O Ms thTPI PRESS TCO L0d 04 UhtSOLATED-(451b*5 07EN) STEAM 3REAK In ChTMT (MALI'$ 37 & 131 SIPULTAhECUSLT). e - .s v -w
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!~ REPORT 8: SCR-A52 i 56RT PARAMETERS: STATUS,5CENUM
- 5CR SIM -
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filts;tlign HLSDE! gR12EIII 11AIU1 111115 ____ - _ - - _ 90-174 4 D TU ADr *A* POINTS A1325 - A1334 (TSI RECORDER) TO SCOPE of SIPULATION. 90-151 4 9- CV C94:08 WCT VENT SETPOINT.
&$-041 3 E FW ENNANCEsUmi YO FAIL StG CONTR/ STEAR FEED FLCW C n A%hEL.
85-042 3 E ~RC" ENHANCEME4T TO FAIL 1 0F 4 PRES $URIZER PEESSURE CHANhELS MIGM. 85-142 3 E CO ADD M AL F ueCT ION FOR LEAE Id CONDENSED DISCHARCE MEADER. 85-187 3 E RM ADD MALfURCTION TO FAIL ADMS COMPUTER SIMULAT10h. l 85-227 3 E CV ADO VARIABLE 3PENING OF AMw-4-36. 86-128 3 E CW ADD MALfumCTION TO COLLAP5E DISCMARGE TUNNEL. ! 86-229 3 E CW AFD Mat t cNCTIch TO COLLAPSE INTAKE TUNNEL. 86-230 3 E SW ADD MALFUNCTION TO SMEAR SW PUP / SHAFTS. 86-235 3 E RM ADD ROMS COMPUTER FAILURE ALARM D PCINT. 86-238 3 E WP UFGRADE SEC0hDARY CHEMISTRY & SAMPLIhG MODELS. 86-248 3 E CE A05 Tbc NEW PCC LEAK MALTUhCTIONS (CONTAINRENT & wPS). E6-379 3 E RD Ast ONE OR 7wo DRPI DATA FAILURE MAFUhCTIoks. 87-001 3 E MISC DEVELOP PUMP & MOTOR WIhDIhG t EEARING TEMPERATURE SUBROUTINES. 87-005 3 E IA ADD IkDIVIDUAL du!LDINGS - INST 8vMENT AIR MEADERS. 87-0D9 3 E MISC GENERIC CAVITAIch EFFECTS ON PUMP TEMPERATURES. l 87-098 3 E CV CCP GEARIkG TEMP A POINTS kOT MCFELES. l 87-160 3 E CW IMPROPER CEhTRIFUGAL CHARGING PUMP SEARING TEMPS ON NO PCCW FLOW. l 87-195 3 E ED ENMANCE MALFbMLTION s116*5 EFFECTS WITH ADDITIONAL ALARMS TO MAKE IT ! IORE REALISTIC. 87-292 3 E- ED ACD WARIABLE GRID FREeUENCY TO ELECTAICAL MODELS. 87-307 3 E CC ADD 8 REMOTE FUNCTIONS FOR LOCAL CONTACL OF PCCh PUMPS. 87-3o6 3 E PCM ADD STEP CobkTER *O ON RESET
- UWERRIDE JPTION TO PCM.
88-110 3 E PCM ADD " MISMATCHED DEVICE" FIELD TO PCM PAGE *AA*. 68-197 4 E SC BUILD MODCOMP A POIkT FILES FCR SCHEGULER USE. 89-063 4 E WP MODEL A 7014T5 FOR 5F-P-10-A 4 9 MOTOR RADIAL 6 TMRUST dEARING5. 4 E PCM AFTER SEWERAL UNSUCCESSFUL ATTEMPS " TOC MANT TDF" MESSAGE 04 TSN. 89-122 ; 89-161 3 E RC ADP'S MALF(s) FOR P2R SPAAT WALbE. LEAKAGE BOTM TWE MAIN C Aus SPEAV ' bALUES. 85-175 4 E NI MAKE h1 TOP DCT. FAILURES (MALFS 103 -> 10e) bARIABLE. 4 SG ADD WARIABLE MALFS TO FAIL PT 505, SO4 & 507 TO Amy FRACTION 9F R4h6E. 89-174 E 89-178 4 E .hI mEPLACE DISCRETE NI SUMMING AMP FAIL MIGH PALFS (97-100) b/WARIASLE MALFS. 90-079 3 E RC ADD VARIABLE LEG BREAK MALFS FOR PIR L T /PT. 90-080 3 .E RD CHANGE RJD EJECTION MALFUNCTf9N TO ONE WITN VARIA8LE SEWERITY. 90-082 3. E MS ADD PAlfuuCTIch 1C MECeAh! CAL.Y JIND MSIV"5 (01 = FULL CLOSER, 1001'= futt OPEN). 90-109 4 E IA ADD MALF(5) GR ENHAN(L MALF 148 TO CAUSE A LOSS OF IA TO THE EXTENT THAT kMOLE SYSTEM OR INDIVIDUAL HEADERS L0b PRESSURE AFFECTS PLANT PERf0RMANCE.. 90-147 3 E. SC A03 REh0N WAtUi TREhD C AP ASILITY. 70-175 3 E RD ADD DRPI ACCURACV.SELECTCR Sw!TCn (MARDwARE ChLT). a
m .,+ - u rc- _ _ _ _ o o o I ~l SIMIN_AT!R PERFGtMMCE AIO CERTIFICATION PROGRAM SCIEDULE r 4 h r Plan l { , l i 1st Year 2nd Year 3rt! Year 4th Year t-Fiuc M rs F. &d.Tc Freauency Subsection itst Ralf 2nd Half 1st Half 2nd Half 1st Ha!I: 2rd Itali. Ist Half 2nd flatfl Ntaber Title Section Steulator Data Entire ongoing Not Applicable lN/A N/A i N/A , N/A N/A N/A N,'A { N/A l 3731 ! l I Bi-Annuei Not Applicable
- I ! X l- l-
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SIMULATOR REVIEW CODMI'1"rE8
/j s CHARTER
\
[/
% I' unction 1.0 The Simulator Review Committee (SRC) is responsible for defining the capabilities of the Seabrook Station simulator based on the identified training objectives. The SRC serves as a forum for communication between the Simulator Systems Department and the simulator users.
Compos i t ion 2.0 The SRC will consist of the following regular memberst
- a. Simulator Systems Department Supervisor, Chairman
- b. Simulator Design Control Coordinator, Vice Chairman
- c. Operations Training Supervisor / Manager (normally Supervisor)
- d. Requal training program team leader (optional)
- e. New license training team leader (optional)
- f. Simulator Suppott Instructor
- g. Operations Department staff member.
, 2.1 The Operations Training Manager will attend meetings as necessary. He will be responsible to approve all simulator improvements which require an expenditure in excess of $5000. 2.2 In addition to the regular members, the SRC can include, but la not limited to, the following person as a special member io provide expertise on topics undes considerations
- a. The Simulator Engineer.
Alternates 3.0 If a member can not ate;.ac, he may designate an alternate. Secretary 4.0 The SRC Chairman will appoint a secretary for the committee. Meetinos 5.0 The SRC will meet at least once each quarter and as convened by the chairman. 5.1 Prior to each meeting, the Chairman will distribute to the members a memorandum that gives the time, location, and agenda for the meeting.
,s 1
. . . _ . .- .~.
m
, 5.2 SRC members may invite persons who are n it members of the committee to rS attend meetings to provide expertise on topics under review.
(2 )! Quorum 6.0, A quorum of the SRC will consist oft
- the Chairman
- the'Vice Chairman
- the Operations Department Member
- one other committee member.
If mutual agreement can not be reached by the committee, the issue will be forwarded to higher authority. Decisions 7.0 Committee decisions, including approvals, will be based on mutual agreement reached through discussion and will be verified by the Chaltman's signature or initials on the meeting minutes. Responsibilities 8.0; The committee is responsible to ensure the simulator supports the training objectives of Seabrook Station. The consittee's specific i .. responsibilities include the actions' listed in Section 8.1.
,_s 8.1 Cosetttee' Actions:
- / %._
\- ~ -a. '
Review simulator capabilities versus ANSI /ANS 3.5 requirements to define the scope'of simulation wherever the phrase "as appropriate for training"'is used in ANS 3.5.
- b. Review / initiate and approve all er.hancements to the scope of simulation.
- c. Review and approve implementation of simulator operating limits as required by_ANS 3.5.
- d. Review and approve deviations in physical fidelity based on training impact analysis.
- e. Provide input to prioratization. of simulator systems department work.
Authority 9.0 75, SRC receives it's authority from the Operations Manager and the Cgerations Training Manager. Records
~10.0 The SRC secretary will maintain written minutes for each meeting. The minutes will' be sent to all members of the SRC, the Operations Manager,
/' the Assistant Operations Manager, the Operations Training Manager, and
,,, each Shif t Superintendent.
2
[)f
~,
CHARLES W. SWINTON 111 EXPERl_ENCE 1986 Present New Hampshirs Ye.nkee Division of Public Service Company of New Hampshire Simulator Systems Deoartment Suoervisor Responsible for the development and installation of software and hardware enhancements to the Seabrook Station control room simulator. THese efforts have included the development and installation of state of the art models of the steam generator and reactor coolant systems, development of a configuration control system and development of a performance testing program. Daily responsibilities include the overall planning, scheduling and ' supervising of department activities, preparation of the adherence to department budgets, and personnel management.
/]
LJ 1982 1988 The Singer Company, Link Simulation Systems Division Prolect Enaineer Successfully directed a team of twenty engineers and designers in the technical development, manuf acture and software checkout of two coal fired power plant simulators valued at over ten million dollars. This included seven months at a remote overseas location where I was the senior Singer employee on site responsible for direction all aspects of the final software and hardware , development and acceptance. Responsibilities included preparation of technical l proposals. approval of engineering and mechanical design, scheduling of manpower and equipment, customer liaison and ensuring compliance of the simulator with specification requirements. 1 !- 1979 1982 United States Navy Designated as engineer of a navat nuclear propulsion plant by Naval Reactors. Experience includes positions assistant engineer officer, chemistry / radiological
- controls assistant, niain propulsion assistant, communicator, and reactor
! controls assistant. Directly supervised from five to fifteen men. Qualified in , submarines and as officer of the deck, inport duty officer, diving officer of the watch and engineering officer of the watch. l l_ 7 l V) l I
.]
~
A; ,-~ _ A, j CHARLES W; SWir.' TON lli g. EDUCATION o Candidate,, Masters Degree in Business Administration : (65%' cf.,mpleteli Northeastern University. Masters in Sci (nce, major in Computer Science, The John Hopkins University.
~
- May 1985.
Bachelor of- Science 'in Mechanical- Engineering, .The United States Naval-
- Academyi June 1977.
Navy . Nuclear. Power School, Nuclear- Power Training Unit and Submarine
. Schooli United States NaW.
1 n u
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.. .__ _ _ . _ _ _ _ _ . _. ~._
M ARLIN P. BOYLE y .
\J.
g EXPERIENJE 1979 Present _ New Hampshire Yankee . Division of Public Service Company of New , Hampshire Simulator Enoineer Coordinated simulator software changes for several modification and upgrades . including: -
- Pressurizer model replacement.
- RCS and Steam Generator model replacement
- RDMS model replacement
- Computer complex upgrade.
- RW.lS/ICCM installation Simulator Desian Control Coordinator Provided operational and engineering expertise to the simulator support group on a full time basis. -
a Simulator Instructor / Senior Simulator Instructor I D Assisted with writing and conducting factory and post delivery tests of the Seabrook simulator, Prepared design change packages, revit.. ' test procedures and coordinated testing for major simulator modifications. Other non simalator projects and tasks included:
- Support of the Nuclear Reactor Fundamentals and Shif t Technical Advisor
. Programs.
- Development of Detailed System texts and program,
,1977-1979 NUS Corporation Trainino Enaineer
- Developed material for and presented a systems and procedures program for cold license candidates at a boiling water reactor.
w
~ f Marlin P. Bovig
~ continued v,
1971 1977 Casolina Power and Ught, H. B. Robinron Steam Electric Plant Trainino InstrucigI Conducted classes and programs for general employee training, reactor operator and senior reactor operator initial training and requalification. , Cantrol Room Ooerator Operated all plant systems and coordinated auxiliary operators during all modes of plant operation. Auxiliarv Ooerator P Operated equipment and systems outside the control room including several months on fossil fueled Unit 1.- 1965 1971 United States Navy
, _ Served on two nuclear submarines as reactor operator and prerequisite watch stations. Received temporary assignment as staff instructor at completion.of
. training at S3G prototype. Served on a heavy cruiser as general duty electronics technician' while awaiting assignment to nuclear power training school.
ERUCATION Bachelor of Science. University of The State of New York.
- U.S. Navy Nuclear Power Training.
U.S. Navy Electronics Technician .*A* School.
~
- LICENSES / CERTIFICATES Certified as having completed Shift Technical Advisor Training for Seabrook Unit 1.
Senior Reactor Operator License for H. B. Robinson, Unit 2. l 1
%< Reactor Operator License for H. B. Robinson, Unit 2.
'l l
l 1 1
g]
! ) KEVIN J. THIBODEAU QUAllF! CATION SUMM ARY Over 15 years of experience in the nuclear power industry, including seven years in the nuclear navy, two years in operations, and six years in nuclear simulation.
EXPERIENCE 1985 present New Hampshire Yankee, Division of Public Service Company of New Hampshire Desian Control Coordinator (Seabrook Trainino) Develop and maintain a program to certify the Seabrook simulator to the NRC requirements of 10CFR55. Respcasibilities at sociated with this include development of procedures for verification and testing, maintaining the design control of the simulator, ensure that the simulator is an exact mechanical and dynamic replica of the plant control room.- Evaluate all piant design changes for simulator impact, and associated documentation. Assist the testing of the []/ C simulator, coordinate with plant engineering in the collection of benchmark data, create plots, compare simulator response, and ensure that all design data is accounted for. Simulator Instructor JSeabrook Train _ingl Develop and administer instruction to license and non license operators. 1982 1985 Singer Link Simulation Systems Division Senior Test Ooeratqr. Savannah River Project Technical liaison between Capont and Singer LSSD engineerin0 team. Responsibilities included; writing and running the Acceptance Test Program, receiving and interpreting plant data for hardware and software design and testing, writing and subsequent follow up of all discrepancies between design specifications and actual simulator performance. 1980 1982 Three Mile Island Nuclear Generating Station, Unit 11 Licensed Control Rcom Ooerator Operation of systems and equipment for plant start up and shut down. () . i
<- KEVIN 1 THIBODEAU EXPERIENCE (continued)
United States Navy Nuclear Electrician involved in the final construction phase of a nuclear powered guided missile cruiser. My responsibilities included reactor installation, initial fill of the RCS, pre criticality and initial criticality programs. Also involved in the testing of all electrical equipment such as turbine generators, diesel generuors, reactor electrical circuits, and secondary electrical systems. As a work center supervisor i supervised 12 men in electrical maintenance. Radioloaical Controls Shif t Suoervis.gI Responsible for maintaining radon safety and exposure control during repair to a nuclear submarine. Also respensible for the transfer and receipt of radioactive 4 materials. Qualified as a quality assurance inspector. EDUCATION Bachelor of Science (General Engineering), State University of New York, O Marine Electricity, Navy Nuclear Power School of Engineering Navy Management School, Quality Assurance School United States Navy. Kepner Trego Management Training, Seabrook Training. O 1,
SHlH PING KAQ
.Lf)/
EXPERIENCE 1985 - Present New Hampshire Yankee . Division of Public Service Company of New Hampshire Egn!Pr Enoineer Full scope Pressurized Water Reactor training simulator realtime modeling and validation:
- Replaced equilibrium Reactor Coolant System, pressurizer, and steam generator thermaLhydraulic models with nonequilibrium, two ohase flow models developed by the author.
- Modeling plant instrumentation and entroller dynamic response.
- Simulator model validation against licensing codes, plant data, and operating procedures for normal operation and severe accident transients.
O O 1986 1987 Massachusetts Institute of Technology, Cambridge MA Visitina Enoineer Light water reactor Mnovation design project sponsored by the Electric Power Research Institute. Model development and analysis for an once-through horizontal steam generator model for PWR application, 1984 1985 Engineering Planning and Management, Inc, Framingham, MA Senior Enoineer PWR transient analysis and assessment for nuclear utilities:
- Station blackout and decay heat removal analysis.
- Operating procedure evaluation.
- OA verification for engineering computer programs.
O V I
1 O SHIH PING KAO continued COMPUTER EXPERIENCE Languages: FORTRAN, C, PASCAL, and GOULD Assembly. Systems: GOULD SEL 32/77 (MPX), DEC VAX 11/780 (VMS), Honeywell DPSS 11/80 (MULTICS), and IBM PC (DOS). EDUCATION Massachusetts Institute of Tachnology, Cambridge, MA Ph.D in Nuclear Engineering (7/1984). Thesis titled, " A Multiple Primary System Model for Pressurized Water Reactor Plant Sensor Validation:" frater than realtime PWR Reactor Coolant System modeling, simulation, and control.
- Draper Fellow, (1982 - 1984)
Masters of Science in Nuclear Engineering (6/1981). Thesis titled, " Dispersed Flow Heat Transfer and Prediction of Critical Heat Flux in Light Water Reactors". Georgia institute of Technology, Atlanta, GA Bachelors of Science in Nuclear Engineering (6/1979).
- Graduated with highest honor
- Engineer in Training (1979)
PROFESSIONAL SOCIETY: Member of the American Nuclear Society. 1 h O
4 i
. -l SH'lH PING KAO continued ' 4 PUBLICATIONS i
. S.P. Kao, "Seabrook Simulator Model Upgrade: Implementation and Validation
, of Two Phase, Nonequilibrium RCS and Steam Generator Model," EPRI 1988 :
E. Conference on Power Plant Training Simulators and Modeling, June 1988. S.P. Kao and J.E. Meyer, *A Plant Computer Based Pressurized Water Reactor ; Primary System Thermal. Hydraulic Model," ANS Transaction, Vol; 50, Nov.. 1985, pp. 637 38. S.P. Kao and J.E. Meyer, " A Two Fluid Pressurizer Digital- Model,' e.NS Transaction, Vol. 49, June 1985, pp. 469 70. S.P. Kao, "A Multiple Loop Primary System Model for Pressurized Water Reactor Plant Sensor Validation," C.S. Draper Laboratory Technical Report No. f- CSDL T P75, July 1984. S.P. Kao and M.S. Kazimi, " Critical Heat Flux Predictions in Rod Bundles,* ' - Nuclear Technoloov, Vol; 60. No 1,-January 1983. M. Massoud,-S.P. Kao, and N.E. Todreas,
- Evaluation of Horizontal Steam-Generator to PWR Application," The Third international Topical Meeting on Nuclear Power Plant -Thermal Hydraulics and Operations, Seoul, Korea.
November 1988. P J.E. Kelly,- S.P. Kao, and M.S. Kazimi, "THERMIT 2: A Two Fluid Model for . Light Water Reactor Subchannel Transient Analysis," M.I.T. Energy Laboratory Report No. MIT EL 81014, April 1981. 4 MOL
1
' Appendix G:-
List'of' Exceptions i _. lI 1This ds-;a. consolidated list of exceptions.: The justification for each {> -l exception is_ contained in the-referenced section'of the body of the ;
, certification 1 submittal.
. Abstract Hol -Section Description
~
{
- l. 1.2.1 :The simulator room is not laid out identically to;the--
control room.- -
- 2. 1.2.1 .The ceiling-in the simulator room is higher than the control
-room. Obstructions behind the rear horseshoe are not inLthe.
simulator room.
- 3. 1.2.1 The location of panels-CP-16-and CP-180A&B are siightly different than in the control room.
4.- - 1. 2 .1, -The location of the Rod; Drop Disconnect Panel is different than in the control room.- 5 .1 :1.2.1 : Panels CP-12,'13, 14, 15 and the back of -panels D, E and F: are not simulated. :I gj
- 6.: 1.2.2 -The Seismic Monitoring Panel (CP-58),: BOP Instrument: I L
Cabinets (CP-152A, 152B, 153, 175, 244,-297A and 297B),;
- Pro, cess Control Cabinets (CP-1, 2, 3, 4, 5, 6,.7 and-8),
t\ l ATWS Mitigation Panel (CP-519) and Vibration Monitoring.
' Panel:(CP-299) are not simulated.
l 7. ' 1'. 2 . 4 .1 .The'keypads-on the ECTS telephones.have been disabled, i
- 8. 1.2.4.1 The Motorolafradio system has been configured as an-intercom.
i 19. i.2.4.11 The power fall-phone-willLnot-switch-over to an outs /de line
-upon-loss =of power.
- 10. : 11.2.4.2- :-Lighting-levels-in the simulator' room do not match those in-
'the: control room.
11.'- 1.2.4.2- '.The battery powered emergency lights-in the control room are 3 ' Enot installed in'the simulator room.- ' k 12.- 1;2.4.3.4 -The: ventilation-noise in the control room is different from that in-the elmulator room.
- 13. 1.3.2 The FSAR accident'for " Dropped Full Length Assembly Bank" is) not required for training.
'14. 1.3.2 The FSXR accident for " Single Rod Cluster Control Assembly
-E i Withdrawal at Full Power" is not required for training.
. s.-
3 Appendix C List of Exceptions
~
/ T 15. 1.3.2 The PSAR accident of inadvertent loading and operation of a \m / fuel assembly in an improper position is not simulated.
- 16. 1.3.2 The PSAR accident of spent fuel cask drops is not simulated.
- 17. 1.3.2 The FSAR fuel handling accidents are not simulated.
- 18. 3.2.2- Several parameters are not within their steady state tolerance criteria at three points in the power range.
- 19. 3.2.3.7 Only shutdown operations with less than full reactor coolant flow are performed.
- 20. - 3. 4 ( 16 ) - The loose parts monitoring panel is not simulated.
- 21. 3.4(22) Valves RC-V145 and RC-V146 are not simulated.
- 22. -3. 6 The-spacing of the letters on the cabinet label for CP-1 BOA is different. The label on the simulator has four screws, in the control-room, it has two screws.
- 23. 3.6 A filler strip on the simulator's CP-180A does not have notches.
24, 3.6- The lettering on the simulator label 1-RR-6528-1 is larger, f~' (). '25. 3.6 The lettering on the simulator label 1-RR-6528-2 is larger.
'26. 3.6 .The lettering on the simulator label 1-RR-6506-A is larger.
- 27. 3.6 The lettering on the simulator label 1-RR-6530 is larger.
- 28. 3.'6 -The spacing of the letters on the cabinet label for CP-180B' is different. The label on the simulator has four screws, in the control room, it has two screws.
- 29. 3.6 The filler strip on the simulator's CP-1808 does not contain notches.
- 30. :3.6 The lettering on the simulator label RM-RK-6576-B is: larger.
- 31. -3.6 The lettering on U e simulator label 1-RR-6507-B is larger.
- 32. 3.6 The spacing between SR and N31 on the-channel selector escutcheon is closer on the simulator than in the control' room.
- 33. 3.6. Vendor name engraving in the lower left corner of the-shaft voltage monitor is missing on the simulator.
-. 3 4 , 3.6 The shaft voltage monitoring panel is located about two 1[ T /
inches lower on the simulator than in the control room. 2
Appendix C ; List of Exceptions i 35, 3.6- The demarcation for ground detection and vital DC is about [/~$
\~- one inch closer together in the plant.
- 36. 1.6 The ground detectSon meters are closer together in the vertical direction than in the control room.
- 37. 3.6 The recombiner A panel on the simulator is a single panel, with no vendor information plate in the upper left corner.
The panel in the control room is comprised of three panels and has a vendor information plate.
- 38. 3.6 The lettering on the RCp vibration panel on the simulator is larger.
39, 3.6 The recombiner B panel on the simulator is a single panel, with no vendor information plate in the upper left corner. The panel in the control room is comprised of three panels ! and has a vendor information plate. 1
- 40. 3.6 Hydrogen analyrer 173A on the simulator has three gray screws on the sides, the one in the control room has four !
black screws. l
- 41. 3.6 The scales for recorder CS-LR-102 on the simulator have 3
" GAL 10 " at the top, in the control room it is "KGAL". ;( )' 42. 3.6- Hydrogen analyzer 174A on the simulator has three gray screws on the sides, the one in the control room has four b?ack screws.
- 43. 3.6 The control room has a hole cover plate between the lights for FAH-F-41,-the simulator does not.
- 44. -3.6 The lettering on the-simulator for CAH-DP-314 is larger.
- 45. 3.6 The instruments along tot left side of panel C rear are closer together than an the control room.
- 46. 3.6 The control room has a hole cover plate between the lights for fAH-F-74.
- 47. 3.6 The mimic along the bottom of panel.D rear is closer together than in the plant.
- 48. 3.6 The hole cover plates on the simulator above PCV-126 are square, in the control room, they are round.
- 49. 3.6 The location of panel UA50 & 51 is slightly different.
- 50. 3.6 The lettering on label CBS-LR-2385 on the simulator is g smaller.
b 3
Appendix C-List of Exceptions
-(n) ~
- 51. 3.6- The label for VL-1 is immediately below the cutout.on the simulator vice one inch below the cutout in the control room.
52, 3.6 The label for-VR-1 is immediately below the cutout-on the simulator vice one inch below the cutout in the control room.. j
- 53. 3.6 The label for VL-2 is immediately belo'w the cutout on the simulator vice one inch below the cutout in the control room.
- 54. 3.6 The label for VR-2 is immediately below the cutout on the simulator vice one inch below the cutout in the control room.
- 55. 3.6 .The mimic at the bottom of panel G front'on the simulator is compressed.
1
- 56. 4.4 Two plant design changes have exceeded the one year
~
requirement for simulator implementation. .j
- 57. 5.6 The simulator system is only capable of plotting parameters at a one second resolution.
i
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