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| {{#Wiki_filter:,- | | {{#Wiki_filter:}} |
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| # 8"8kg UNITED STATES
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| ;[,- h, NUCLEAR REGULATORY COMMISSION
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| , M ,) . $ REGION IV -
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| 411 RYAN PLAZA DRIVE, SUITE 400 k'+,' '*
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| #[ AR LINGTON, T EXA$ 760118064 December 3, 1997
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| . NOTE TO: NRCDocumentControlDisk -
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| Mail Stop 0 5 D 24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region'IV
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| | |
| ==SUBJECT:==
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| OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON JULY 28 THROUGH AUGUST 1, 1997, AT RIVER BEND STATION, UNIT 1 DOCKET #50 458 On July 28 through August 1, 1997, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the followiiig information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
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| Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070.
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| b) As given operating examination, designated for distribution under RIDS Code A070.
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| Item #2 - Examination Report with the as given written exeminatior attached, designated for distribution under RIDS Code IE42.
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| ~ If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860 8253.
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| ~
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| 9712090307 971203 '
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| A O ((}ggg hil
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| 1 1
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| i Ta le of Contents 2 SRO
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| -r,
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| ; e
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| - Written Exam-Key a
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| \'n ten Exam
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| \
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| i RO l Written Exam-Key l
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| j RO l
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| ! Written Exam !
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| l M ll m> I M I m I m MD a 58
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| <1 li l
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| .a r
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| L >
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| '-g cSwd
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| - . - - - - . _ - . . . _ .- . - - . - . _ . - ~ - _ _ - .- - . - ~ - ..-- - ... - . - - - - - . - .
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| 0S-401 IlWR SRO fxemination Oudine Fonn ES-40t 1 4
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| Facihty: River Bend Station Date of Exam: 7/28/97 Exam Level: SRO K/A Category Points Tier Group K 1
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| K 2
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| K 3
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| K 4
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| K 5
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| K 6
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| A 1
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| A 2
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| AlAlG 3 4 Point Total W y; N b] C4 1
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| 1 4 4 A 'O 10 2 d!* '. 26 Emergency & 2 ?
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| W" 3 'Je t , q9 y'a Abnoimal 3 4 3 $@9' "
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| 3 4 s E 17 Plant Tier pgj pi fy Evolutions. Totals igj [$[yQ[ J O .W 4 7 8 9 iD4 Y E'. 13 6 %@' % 43 1
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| j 3 1 1 5 3 2 1 5 1 1 23 2 2 Plant 5 1 3 2 2 13 Systems 3 2 1 1 4 Tier Totals 10 1 3 9 3 4 3 5 1 1 40 _
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| Cat 1 Cat 2 Cat 3 Cat 4
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| : 3. Generic Knowledge and Abilities 7 l5 1( 4 17 Note: Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.
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| Actual point totals must match those specified in the table.
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| Select topics from many systems; avoid selecting more than two or three KIA topics from a given system unless they relate to plant-specific priorities.
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| Systems / evolutions within each group are identified on the associated outline.
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| The shaded areas are not applicable to the category / tier.
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| NUREG lo21 Intermin Rev. 8, knuary 1997 3
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| t ES-401 BWR SRO Examinaton Outline Emergency and Abnormal Plant Evolutons - Ter 1/ Group 1 Form ES-401-1 E/A PE#/Name/ Safety Function K1 K2 K3 A1 A2 G K/A Topics '
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| imp. Points 295003 Partal or Complete Loss of AC PwrM 03 Abhty to operateimorutor systems required for safe shutdown 4.4 1.00 295003 Partial or Complete Loss of AC PwrM 02 Abhty to operate Emergency Generators 4.3 1.00 295006 SCRAM / I . 06 Reason for Rectre Pump Speed Reduction 3.3 1.00 295007 High Reactor Pressuren11 04 Abhty to operate /morutor SRVs dunng H Rx Press. transients 4.1 1.00 295010 High Drywell PressureN 05 Relatonship between Dw11 Coohng and H: Drywell Pressure 38 1.00 295013 High Suppression Pool TempN 02 Reasons for Irmitng heat additons to the SP on hgh temp. 3.8 1.00 Relationshrp between void concentration and inadvert React 295014 Inadvertant Reactivity Additonn 07 adddion 4.1 1.00 295015 incomplete SCRAMA 04 Relatonship between RPS and imcomplete SCRAM 40 1.00 295016 Control Room AbandonmentN!I 03 Reason for disabling rWol room controts cunng evacuation 4.1 1.00
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| , 295016 Control Room AbandornnentMI 09 Abhty to control /morutur Rx water level from RSS panel 4.0 1.00 295017 High Off-site Release RatenX 02 Response of plant ventilation on High Rad. 3.5 1.00 295023 Refueling Accidents Cochng ModeMll 04 Abhty to determine the occurrence of a Fuel Handhng Acenden1 4.1 1.00 295024 High Drywell PressureN 03 Knowtedge of Containment Ventng dunng hgh CTMT pressur 4.1 1.00 2S5026 Suppression Pool High Water TempN 06 Relatonshtp between SP temperature and SP level. 3.7 1.00 295027 High Containment TemperatureN 02 Effects of ngh CTMT temperature of Rx levelindicaton 3.2 1.00 195030 Low Suppression Pool Water LevetN 08 Requirements of SRV discharge submergence and SP level 3.8 1.00 295031 Reactor Low Water Levetal 01 Abhty to determine Rx water level when RPV water level es lo 4.6 1.00 295037 SCRAM Cond. bon Present and Power Al:ove APRM Dowscale or Unknownn 04 Abikty to operate / monitor SLC under ATWS conditions 4.5 1.00 295038)iigh Off-site Release RatenX 06 Opeiate/morutor plant ventiation dunng H:gh Offsite Release 3.6 1.00
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| . 50CD00 High Containment Hydrogen ConcW 01 Abhty to monttor Rx water level dunng low RPV level conditions 295023 Refuehng Accidents Cooling ModeMll 01 ,
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| L Radiatron exposure hazards dunng refueling ops 4.1 1.00 ;
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| i 295031 Reactor Low Water Levetal 01 Know1 edge of adequate core cool!ng 4.7 1 00 j i 295031 Reactor Low Water Levetni 05 Reason for Emerg. Depress on Low RPV water level 4.3 1.00 295031 Reactor Low Water Levetal 08 Abbty to operate alt injection systems 3.9 1.00
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| ; 295037 SCRAM Conditon Present and Power Operate / monitor ARI oi ; ATWS conditions At:ove APRM Dowscale or Unknowna 03 4.0 1.00 295038 High Off-site Release RatenX 05 Operate /morutor PAMS dunng hgh offsite re recase 3.5 1.00 K/A Category Point Totals: 4 4 6 10 2 -
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| Group Potnt Total: 26.00 6
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| Page 4
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| t i
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| ES-401 BWR SRO Examinateon Outhne Emergency and Abnormal Plan! Evolutons - Tser 1/ Group 2 Form ES-401-1 E/A PE*/Name/ Safety functon K1 K2 K3 A1 A2 G K/A Topres imo Poents ,
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| 295001 Portal or Complete Loss of Forced Circutation / I &rv 01 Ability to monitor power /flow map dunng loss of Recire 3. 8 1.00 ;
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| 295001 F'artat or Complete Loss of Forced i Circulaton / i & IV 01 Ability to interpret the power to flow map 3.9 1 00 295002 Loss cf Main Cond. Vacuum I!!! 02 Knowledge of effects of loss of cond. vac. on marn turbtne 32 1.00 !
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| 295005 Main Turtxne Generator Tnp IIt! 04 Electncal dist status fottowing a tnp of the main generator 4.1 1.00 295008 Htgh Reactor Water LevelI!! 07 Relatonshp between HPCS and begh reactor water level 3.0 1.00 Operatonal empiteatens of a loss of CCW to systern -
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| 295018 Partal or Total Loss of CCW / Vit! 02 operatiot- 36 1.00 Operational empicatens of a loss of CCWto system ,
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| < 295018 Partal or Total Loss of CCW / VIII 02 operation 3.6 1.00 i 295020 Inadvr tant Cont. Isolabon / V & VII 06 Abthty to determine Rx water level applicable to inadv. CTMT 36 1.00 ,
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| 295021 Loss of Shutdown Cooling / IV 01 Abiltty to determine HtR1CDR following a loss of SDC 3.6 1.00 -
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| Reactor pressure vs control rod insertron capabihty on a loss 295022 Loss of CRD Pumps Ii 01 of CRD 34 1.00 l 295028 High Drywell Temparature / V 01 Effect of high drywell temp on RPV levelinstrumentaton 3.7 1.00 Imphcaton of high SP water level as it apphes to CTMT j 295029 H@h Suppression Pool Water Level / V 01 integrity 3.7 -1.00 .I i
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| 295032 High Secondary Containment Area ;
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| i Temperature / V 05 Abihty to operate systems affected by high SEC CTMT temp 3.9 1.00 295033 High Secondary Containment Ares Radsaton Levels / IX 03 Reason for isolating systems due to high SEC CTMT rad. Ms 39 1.00 295333 High Secondary Containment Area Radiaton Levels t IX 03 Reason for isolating systems due to high SEC CTMT rad Ms 3.9 1.00 ,
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| 295034 Htgh Secondary Containment Ventiaton [
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| Radiation Levels / IX 04 Abihty to monitor area rad monitors in the SEC CTMT 3.8 1,00 i 295035 High Secondary Containment Differental Pressure / IX 01 Knowledge of SEC CTMT integnty due to hi SEC CTMT D/P 42 1.00 l 295036 High Secondary Containment Sump / Area i Water Levels / IX L l >
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| ; 600000 Plant Fire On Site / VIII 5 [
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| e I l i l l !
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| K/A Category Point Totals: 3 4 3 3 4 Group Point Total: 17.00 NUREG-1021 Interim Rev 8. January 1997 l _ _ _ - - - . _ _ .
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| . - - - .~c - . . . .- .~ . .- --
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| . -)
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| n l
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| ES-401 Form ES-401-1 I l BWR SRO Exa nination Out!me Plant Systeg Tier 2/ Group :
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| System #/Name K1 K2 K3 K4 K5 KS A1 A2 A3 A4 G 1 K/A Topics ' tmp. Points' !
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| 201005 RCIS 05 Knowtedge :,f the concept of control red density 36 1.00 2G3000 RHR!LPCt: Injecten Mode 06 RHR mtertocks that prov*de adequate NPSH 3.5 1.00
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| ; 203000 RHR/LPCt: Injecton Mode 01 Abdtty to monitor RPV level dunng RHR mjecton $, . 3 1.00 [
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| 217000 RCIC 03 Effect of valve closures on RCIC system 3.3 1.00 ;
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| 209001 LPCS 08 Automate system wwtaton of LPCS 4.0 1.00 209002 HPCS 02 A_ffects of a loss cf CST will have on the HPCS system 34 1.00 211000 SLC 08 Knowledge of SLC system instaten 4.2 1.00 ' ;
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| 212000 RPS 14 Abthty to reset RPS following wutaton 38 1.00 [
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| 215004 Source Range Monttor 03 Predict system response to a stuck SRM detector 3.3 1.00 ,
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| 215005 APRM/LPRM 13 Relatonship between TIPS and APRMs 30 1.00 l l 217CD0 RCIC 01 Power supphes to RCIC motor operated valves 2.8 1.00 -
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| 218CD0 ADS 01 Power supply to ADS logic 3.8 1.00 l 223001 Primary CTMT and Auxiliaries 09 Predict effect of a maffuncton of a vacuum breaker 3.9 1.00
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| [
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| 123002 PCIS/NSSSS 08 Effects of defeating of interlocks dunng emergency svtuatons 37 1.00 t i 239002 SRVs 03 Abehty to mrtgate the consequences of a stuck open SRV 4.2 1.00 ,
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| 241000 Reactor /Turb. Press. Reg. 02 Know1 edge of the EHC regulator in the press regulatng mode 3.3 1.00 !
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| 259002 Reactor Water Level Control 02 Effect of a loss of Feedwater of the Rx Water Level Control Sy 3.7 1.00 231000 SGTS 01 Loss of AC pwr on the SGTS 3.0 1.00 [
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| l 01 Relatonship between AC Distnb. and the Emergency Generai 4.3 1.00 262001 AC Electncal Drstnbution l 264000 EDGs 10 Effects of a LOCA signal on the EDGs .. 4.2 1.00 l 2fD001 Secondary CTMT 04 Cause-effect relatonship SEC CTMr and SGTS 3.9 1.00 -[
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| 216000 Nuclear Boiier instrumentation 13 Operatonal emphcanons of reference leg flashing 3.6 1.00 7
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| 223002 PCIS/NSSSS t02 Instatons signal of the NSSS system, venfying valve closures 3.5 1.00 :
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| i i
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| i i
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| ' Group Point Total: 23
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| .K/A Category Point lotals: 3 1 1 5 3 2 1 5 ,1 1 l i
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| ? ?
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| 1 l
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| NUREG-1021 Interim Rev. 8. huary 1997 l i
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| i . . _ _ _ _ _ _ _ .
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| i t
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| b t
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| P ES-401 BWR SRO Examinaton Outine Plant Systems - Tier 2/ Group 2 Form ES-401-1 [
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| I System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topics imp. Pornts 201001 CRD Hydraulic 07 Relatonship betwe?n RPS and CRDH 3.4 1.00
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| * 201002 RMCS 04 Know1 edge of ircerlocks whch provide " single notch w/d' 3.7 1.00 !
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| 202001 Recircutabon - 09 Ablity to predict changes in Reenc Pump Seal pressures 3.3 1.00 ;
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| 204000 RWCU 15 Relatonshtp between RWCU and Leak Detecten Sys. 3.2 1.00 205000 Shutdown Cooling 01 Effect of a loss of AC pwr on RHR SDC operaten 3.4 1.00 295003 IRM 05 Scram and Rod Block setpoints of the IRMs 3.9 1.00 234000 Fuel Handitng Equrp. 01 Interlocks preventng core alts dunng CR movement 4.1 1.00 i-245000 Main Turtxne Gen / Aux. 06 Main generator operations and limitations 3.2 1.00 259001 Reactor Feedwater 05 Retabonship between Feedwater and condensate syste 3.2 1.00 263000 DC Electrical Dist 03 Effects of a loss of DC power on Feed Level Control 3.8 1.00 ;
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| 271000 offgas 07 Relatonship between Offgas and Plant Air Systems 2.7 1.00 29CD03 Control Room HVAC 04 Cause-effect relatronship CR HVAC and NSSSS 3.3 1.00 i 245000 Main Turbine Gen / Aux. 05 Interlocks which provide protechon of the Main Turbne 3.0 1.00 E
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| i KiA Category Point Totals: 5 1 3 2 2 Group Point Tctat 13 00 I
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| i l
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| ~
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| [
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| t i
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| 4 l NUREG-1021 Interim Rev. 8. January 1997
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| * l E5-40) Ocneric Knowledge and Abilities Outline (Tier 3) I'orm 05-4015 j l
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| Facility: River Bend Sta Date of Exarn: 7/28/97 Exam Level: SRO Category K/A # Topic imp. Points 2.1.4 Knowledge of shift staffing requirements 3.4 1.00 2.1.33 Knowledge of operator responsibil; ties during all modes of operation 4.0 1.00 Conduct of 2.1.11 Knowledge of less than one hour tech spec action statements 3.8 1.00 Operations 2.1.28 Knowledge of purpose and function of mag.r system components 3.3 1.00 2.1.32 Ability to explain and apply system limits and precautions 3.8 1.00 Ability to recognize indications for systems which are entry conditions 2.1.33 for technical specifications 4.0 1.00 2.1.7 Ability to evaluate plant performance , 4.4 1.00 2.1 Total 7.00 Knowledge of the process for determining if the proposed change, 2.2.8 test, or experiment involves an unreviewed safety question. 3.3 1.00 2.2.12 Knowledga of surveillance procedures 3.4 1.00 Equipment 2.2.13 Knowledge of tagging and clearance procedures 3.8 1.00 Knowledge of bases in technical specifications for limiting conditions Controf 2.2.25 for operations and safety limits. 37 1.00 2.2.13 Knowledge of tagging and clearance procedures 3.8 1.00 2.2 2.2 Total 5.00 Knowledge of radiation exposure limits and contamination control, 2.3.4 including permissible levels in excess of those authorizcd, 3.1 1.00 Radiation 2.3 Control 2.3 2.3 2.3 2.3 Total 1.00 2.4.16 Knowledge of EOP implementation heirarchy 4.0 1.00 Emergency 2.4.4 Ability to recognize EOP/AOP entry conditions 4.3 1.00 hocedures 2.4.6 Knowledge of symptom based EOP strategies 4.0 1.00 Ability to perform with reference to procedures those actions that and Plan 2.4.49 require immediate operation of system components and controls. 4.0 1.00 2.4 2.4 2.4 Total 4.00 Tier 3 Target Point Total (RO/SRO) 17.00 NUREG-102 I interim Rev. 8. knuary 1997
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| C'
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| . NRC SRD KEY 6/27/97 SRO Exam I RO-Exam l Enembank l Answer 001 016 243 B 002 017 378 C 003 002 390 C 005 019 373 D UO6 004 314 0 007 005 402 A 008 020 251 0 009 n21 280 C
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| ~
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| 010 415 D 011 034 380 8 012 018 A 013 022 D 014 091 D 015 029 023 8 016 007 224 8 017 416 D 018 032 311 D
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| "~
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| 019 389 0 020 008 230 8
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| ,)
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| 021 437 C 022 009 417 0 623 010 404 8
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| ' ~ ~
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| 024 013 387 D 025 418 A 026 014 044 B 027 041 419 0 028 015 364 A
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| ~~~~
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| 029 001 381 A 030 63iI' 327 C 031 022 241 0 032 V86 054 C 033 033 369 D 034 024
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| * 368 A 035 027 081 A 036 028 388 8 Page1
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| 9 NRC SRO KEY 6/27/97 SRD-Enem i R0-Exam l Enembank l Answer 037- 035 376 B 038 227 0 039 U56 - 312 A 040 031 420 C 041 0'36 377 C' 042 040 013 D 043 042 140 A 044 145 B 045 6I4 010 D 046 045 247 C 047 074 D
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| ~
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| 048 097 C 049 200 A 050 053 C EST 051 089 B 052 196 C 053 054 143 D 054 055 217 C 055 092 212 0
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| ~
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| 056 056 U32 C J
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| 057 057 028 C
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| ~
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| 058 060 348 8 059 061 272 B 060 076 261 A 061 062 199 C 062 080 082 C UEi~ 065 244 .D 064 232 D 065 079 D 066 039 122 0 067 006 C 068 069 197 C 069 071 422 .B 070 011 D 071 084 004 C 672 Q75 3b6 A Page 2
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| 4 4
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| NRCSROKEY 6/27/97 SRO-E n em l HOEmam l Enembank l Answer 073 085 A 074 077 42; 6'~
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| ~ ~
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| 075 079. 201 A 07b 081 167 0 077 074 071 C 078 063 049 0 079 083 438 0 080 073 190 B 081 215 0 082 023 038 0 083 430 B 084 088 433 D 085 326 C 086 061 B 087 089 411 A 088 100 37% B
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| ~~
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| 089 090 42'3 A 090 091 375 0 091 413 A 093 412 A 094 094 414 A 095 425 A 096 095 313 C 097 408 8 098 409 0 099 097 432 C 100 439 D 8
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| +
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| Page 3
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| h
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| ?(' SRO EXAM KEY
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| -(
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| Exam Number NRC SRO - Rev. O Exam Title NRC SRO EXAM i De following conditions exist:
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| The plant has experienced a station blackout.
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| The Div 3 Diesel Generator was started and is running normally.
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| Emergency use of Div 3 for decay heat removal and RPV level control is being implemented.
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| Which of the follow!ng describes the general flowpath for this cooling mechanism?
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| : a. CST !!PCS pump RilR "A" heat exchangers- RPV Shutdown cooling drains to Suppression pool.
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| : b. Suppression pool IIPCS pump RPV shutdown co('% to loop"A" RilR heat exchangers then test return to suppression pool.
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| : c. CST - IlPCS pump - RPV - shutdown cooling to loop "B" RilR heat exchanger then test return to Suppression pool.
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| : d. Suppression pool - IIPCS pump RilR "A" heat exchanger - RPV - shutdown cooling drains to suppression pool.
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| ANSWER:-
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| : b. Suppression pool IIPCS pump - RPV shutdown cooling to loop "A" RilR heat exchangers then test return to suppression pool.
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| pl IDNO: LP# OBJ #
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| 243 11L0-541 6 PROCEDURE NUMBER: OTHER:
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| AOP-0050 ATT 6 LEVEL 4
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| . . . . .. . ~
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| *295003 AA1.03 i 4.4' 4.4 _
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| COMMENTS: 7/97 new RO T1 G2 -
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| SRO T1 G1 ALSO OBJECTIVE 4
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| ATTACHMENT 6 $
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| PAGE 1 OF 4 EMERGENCY USE OF DIV III FOR DECAY HEAT REMOVAL AND RPV LEVEL CONTROL
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| - (^7ISCUSSION:
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| U This attachment provides an alternate method of decay heat removal when no Division I or II power is available for RHR Pumps. This lineup transfers decay heat from the core via the RHR A HX which is being cooled by Standby Service Water. This lineup also reduces the amount of heat energy transferred to the Suppression Pool.
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| Using the Division 3 Diesel Generator with the HPCS Pump, Standby Service Water Pump SWP-P2C, and RHR A HX, the following lineup is established:
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| Suppression Pool
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| * HPCS Pump -+ E22-MOVF004, HPCS INJECTION Valve -+ RPV
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| * E12 MOVF009, SHUTDOWN COOLING INBOARD ISOL Valve -+
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| E12 MOVF008, SHUTDOWN COOLING OUTBOARD ISOL Valve -+
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| E12-MOVF006A, SHUTDOWN COOLING TO A LOOP Valve -+ RHR A HX -+
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| E12 MOVF024A, RHR PUMP A TEST RTN TO SUP PL Valve -+ Suppression Pool INSTRUCTIONS:
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| CAUTION Prolonged operation of a DG without cooling water can lead to ' permanent damage to the DG.
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| Do nat allow a DG to run for more than one minute without cooling water.
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| NOTE A dr>vell entry might be needed to open E12-MOVF009 and complete this attachment.
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| I Verify SSW is in operation as follows:
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| 1.1 SWP-P2C, STANDBY SVCE WATER PUMP C is running.
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| 1.2 SWP MOV40C, PUMP DISCHARGE VALVE is open.
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| 1.3 SWP-AOV599, STBY CLG TWR INLET, STATION BLACKOUT RETURN TO STBY COOLING WTR is open.
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| O AOP n040 RFV 0 P AG7 7.1 OF .t2
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| 1 e SRO EXAM KEY O Esam Number NRC SRO . Rev. 0 . Esam Tit!c NRC SRO EXAM 2 A 1,oss of Offsite Power has occurred. The Division i Diesel generator is currently loaded to 2500 KW.
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| Which one of the following is the MAXIMUM allowed additional load that can be imposed en the generator 7
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| : c. 360 KW b.5s0KW
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| : c. 630 KW
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| : d. 730 KW ANSWER:
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| : c. 630 KW IDNO: LP# OBJ #
| |
| 378 ilLO 037 7 PROCEDURE NUMBER: OTHER:
| |
| AOP-0004 LEVEL 2 1295003 AA1.02 i 4.2i 4.3 .J COMMENTS: 7/97 new AOP 0004, Rev 9. Caution, p. '! of 39 ALSO AN CBJECTIVE OF HLO 523 ROT 1G2 SRO T1 G1 4
| |
| | |
| !. I l
| |
| CAUTION
| |
| ,g The operations in Section 5.10 add significant load to the Standby Diesel Generators. [
| |
| Excessive loading and a subsequent trip of a Standby Diesel Generator under these conditions !
| |
| will severely degrade plant recovery ability. Use Attachment 2, Division I and 11 Manually Started Loads Guide when loading. Do ng exceed 3130 KW on the Standby Diesel Generators. ;
| |
| 1 5.10 Operation of ManualIoads 5.10.1. IF a LOCA exists, THEN perform the following:
| |
| : 1. WHEN turbine has been at standstill for at least 15 minutes, THEN lockout the EBOP to reduce battery load. $
| |
| CAUTION j A high starting current will occur when starting Standby Cooling Tower Fans which could trip the associated EHS-MCC16A(B). Do ng start more than one ! .
| |
| bank of five Standby Cooling Tower Fans at a time. .
| |
| O
| |
| /O 2. Operate Standby Cooling Tower Fans as follows:
| |
| : 1) Stan the Standby Cooling Tower Fans necessary to maintain SSW header temperature less thar. 90*F.
| |
| C 2) Within one hour of the LOCA event, place all Standby Cooling Tower Fans in service.
| |
| : 3. Operate Fuel Pool Cooling as follows:
| |
| : 1) Locally monitor Fuel Storage / Area Pools and start a Fuel Pool Cooling Pump per SOP-0091, Fuel Pool Cooling and Cleanup to maintain pool temperature less than 140*F.
| |
| C 2) Within two hours of the LOCA event, cross connect SSW to RPCCW per SOP-0016, Reactor Plant Component Cooling Water System, and place the Fuel Pool Cooling System in service.
| |
| a0 .
| |
| AC' M REV - 14 PAGE 10 OF 43
| |
| | |
| n)
| |
| N
| |
| .SRO EXAM KEY Esam Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM -
| |
| 3 Which of the following states the overall low-low set system response of the SRVs for the RPV pressures given?
| |
| : a. At i113 psig, only one valve will be open. It recloses at 926 psig,
| |
| : b. At ii13 psig, only one valve will be open, it recloses at 936 psig.
| |
| c At i 103 psig, only two valves will be cqen. One recloses at 936 psig, tne other at 926 psig.
| |
| : d. A: 1103 psig, only eight valves will be open. Three reclose at 946 psig, three at 936 psig, and the last two at 926 psig.
| |
| ANSWER:
| |
| : e. At 1103 psig, only two valves will be open. One recloses at 936 psig, the other at 926 psig.
| |
| IDNO: LP# CILI #
| |
| 390 IILO-007 4
| |
| - PROCEDURE NUMBER: OTHER:
| |
| ARP-H13-P601 19.H08 LEVEL 3 ARP-H13M01 19-H11 TS 3,3 6 4 i 1295007 AA1.04 i 39 4.1 COMMENTS: 7/97 new RO T1 G1 SRO T1 G1 LOTM 24 6, p.6, and Table 2, p.21 of 26 4
| |
| Y 3 4
| |
| -4
| |
| | |
| SRV LOW LOW STPT I___ h ALARM NO. 2031 DIV I LOGIC IH13*P601 /19A / H08 SEALED IN <
| |
| INITIATING DEVICES SET POINTS
| |
| : 1. B21C'K72A Low Low Set Relay 1. Energized
| |
| : 2. B21C'K72E Low Low Set Relay 2. Energized AUTOMATIC ACTIONS
| |
| : 1. The lift and reset pressures for the relief valves change to the following:
| |
| LIEI RESET
| |
| : a. IB21*F051D 1033 926
| |
| : b. IB21*F051C 1073 936
| |
| : c. IB21*F047F 1113 946
| |
| : d. IB21*F051B 1113 946
| |
| : c. IB21*F051G 1113 946 DEERATOR ACTIONS
| |
| : 1. Refer to Eur, -
| |
| g 2. Observe pressure reliefs lift at reduced setpoints to verify Low-Low set is functional.
| |
| LONG TERM ACTIONS
| |
| : 1. The Low-Low setpoint logic can be reset by depressing IB21-S38A SRV LOW LOW SET LOGIC A RESET on 1H13*P601.
| |
| POSSIBLE CAUSES
| |
| : 1. High pressure sensed in the RPV by either:
| |
| : a. 1B21*N668A or E (1103 psig) -
| |
| : b. 1B21 *N669A or E (1113 pt.ig).
| |
| : c. 1B21*N670A or E (1123 psig). ,
| |
| REFERENCES
| |
| : 1. GSU Doc. 0222.250-000-092H
| |
| : 2. GE Dwg. 851E225AA Sh.16
| |
| ~
| |
| ?
| |
| l _
| |
| ARP-601-19 REV - 12A PAGE 92 OF 93 l
| |
| | |
| ...~ - - _ - . . _ _ =. _ _ .
| |
| SRO EXAM KEY O Esam Number NRC SRO Rev. O Exam Title NRC SRO EXAM 4 You have been instructed to control drywell temperature and pressure by operating all avaliable drywell cooling. While doing this, service water to the drywell unit coolers automatically isolates.
| |
| Which of the following caused the isolation?
| |
| : a. liigh drywell temperature (max, recorded 265 deg F)
| |
| : b. Low RPV water level (min. recorded 28")
| |
| : c. liigh drywell pressure (max. recorded 1.82 psid)
| |
| : d. Loss of 120 VAC power ANSWER:
| |
| : c. Drywell pressure 1.82 psid IDNO: LP# OILI #
| |
| 382 111.0-522 .3 PROCEDL%f NUMBER: OTHER:
| |
| O .- . _. --
| |
| AOP 0003 SIGNAL C LEVEL 3
| |
| '295010 AK2.05 i 3.7, 3.8 !
| |
| COMMENTS: 7/97 new LOTM 63 4, p.11 RO T1 G1 SRO T1 G1 O ,
| |
| | |
| A s i At.tmita a i PAGE 1 OF 4 SIGNAL TO ACTUATION / ISOLATION RELATIONSHIP q SIGNAL ACTUATIONS AND ISOLATIONS O A Reactor Vessel Water Level - Low: LEVEL 3: 9.7 inches
| |
| : 1. Group 5 and 14 valves isclate.
| |
| : 2. H13 P601 Post Accident Recorders anift to fast speed.
| |
| B Reactor Vessel Water Level- Low Low: LEVEL 2: -43 inches
| |
| : 1. Group 1,7,8,9,15, and 16 valves isolate.
| |
| : 2. Group 11,12, and 13 dampers isolate.
| |
| : 3. Standby Gas Treatment System auto initiates.
| |
| : 4. Annulus Mixing System auto initiates.
| |
| : 5. Fuel Bldg Exhaust Filtration Trains A & B auto initiate.
| |
| : 6. Containment Hydrogen Analyzer / Monitor System starts.
| |
| : 7. Control Room HVAC Emergency Mode auto initiate.
| |
| i C Reactor Vessel Water Level - Low Low Low: LEVEL 1: -143 inches
| |
| : 1. Group 6 and 10 valves isolate.
| |
| : 2. HVR-UCI A and UCIB, CONTAINMENT UNIT COOLERS auto start and are supplied by the Service Water System after a 1 minute time Os delay.
| |
| : n. SWP MOV502A and 502B, CONTMT COOL SPLY open.
| |
| : b. SWP MOV503A and 503B, CONTMT COOL RTN oren.
| |
| D Drywell Pressure - High: 1,68 psid
| |
| . 1. Group 1,3,8,10, and 14 valves isolate.
| |
| : 2. Group 11,12, and 13 dampers isolate.
| |
| : 3. Standby Gas Treatment System auto initiates.
| |
| : 4. Annulus Mixing System auto initiates.
| |
| : 5. Control Room HVAC Emergency Mode auto initiates.
| |
| : 6. Fuel Bldg Exhaust Ventilation Trains A & B auto initiate.
| |
| : 7. Containment Hydrogen Analyzer / Monitor System starts.
| |
| : 8. HVR-UCI A and UCIB, CONTAINMENT UNIT COOLERS auto start and are supplied by the Service Water System after a 1 minute time delay,
| |
| : a. SWP-MOV502A and'502B, CONTMT COOL SPLY open.
| |
| p b. SWP-MOV503 A and 503B, CONTMT COOL RTN open.
| |
| %)
| |
| AOP-0003 REV-10 PAGE 8 OF 17
| |
| | |
| A11 AC& WEN l~ 2 PAGE 1 OF 6 ISOLATION VALVE CHECKOFF SHEET IMXAIID DIVI &IV RENTOREN PANEL ISOIAIHY DIVII&IR RESIDRE.
| |
| IsmALS OLHBOARD INEALS IN H W S INBOARD INHIMS H13 P877 EJS ACB25 GROUP! EJS ACB77 SIGNALS EJS ACE 66 EJS-ACB09 B, D EJS ACB50 EJS-ACB49 H13-P870 SWP MOV4A GROUP 1 SWP MOV4B SWP MOV5B SIGNALS SWP MOVSA SFC-MOV121 SOP 0091 B. D SFC-MOVl39 SOP 0091 DFR AOV102 DFR AOV101 CNS MOV125 FPW MOV121 SAS MOV102 IAS MOV106 WCS-MOVl72 SOP-0090 WCS-MOVl78 SOP-0090 DER AOV127 DER AOV126 SFC MOVi19 SOP-0091 l SFC-MOV120 SOP 0091 SFC MOV122 SOP 0091
| |
| * Notify the OSS/CRS if CCP ha; been isolated to the Reactor Recire Pump.
| |
| (*) CCP-MOVl59 (*) (*) CCP-MOV158 (*)
| |
| (*) CCP-MOVl43 (*) (*) CCP-MOVl44 (*)
| |
| (*) CCP MOVl42 (*)
| |
| (*) CCP MOVl38 (*)
| |
| l i
| |
| app-0003 REV - 10 PAGE 12 OF 17 i
| |
| I
| |
| | |
| SRO EXAM KEY O Esam Number NRC SRO Rey, O Esam Title NRC SRO EXAM 5 EOP 2, " Primary Containment Control", requires the reactor to be scrammed before suppression pool temperature reaches 110 Degrees F. Which one of the following states the reason for this requirement?
| |
| : a. Assures that the containment design pressure will not be exceeded due to compression of the non condensable gasses due to the higher w ater ternperature,
| |
| : b. Assures that with the expected temperature rise of 70 Degrees F during the blowdown phase of an accident, that complete condensation of reactor coolant will occur.
| |
| : c. Assures the post LOCA suppression pool hydrodynamic forces are within the design limitation of containment.
| |
| : d. Assures a reactor shutdown occurs, to minimize heat rejected to the primary containment, if Emergency Depressurization is required.
| |
| ANSWER:
| |
| : d. Assures a reactor shutdown occurs, to minimize heat rejected to the primary containment, if Emergency Depressurization is required.
| |
| IDNO: LP# OR,I #
| |
| 373 IILO-514 5 PROCEDURE NUMBER: OTHER:
| |
| EPSTG*0002 LEVEL 4 I NRC KA: I RO: [TRUI J295026 EK2.05_ 3.3 1295013 AK3 02 _ _36_ . 34 _3.8_
| |
| COMMENTS: 7/97 00W EPSTG'0002, p. B 216 ROT 102 SF.O T1 G1 O ,
| |
| | |
| EOP 2 Primary Containment Control- SPT
| |
| (
| |
| ma. .
| |
| l'Nj[
| |
| ~
| |
| < STEP,SPT-4 l ._.
| |
| if suppression pool temperature continues to increase with all available suppression pool cooling in operation, further actions to reverse the increasing temperature trend must be directed.
| |
| These actions must be taken before suppression pool temperature reaches 1107. This will ensure that if emergency RPV depressurization is required later, the heat rejected to the primary containment will be minimized.
| |
| When it has been determined that suppression pool temperature cannot be maintained below that temperature requiring a reactor scram by Technical Specifications (1107), the operator will continue in this procedure. Subsequent steps (SPT-5 through SPT-7) will direct entry into EOJ-l and a reactor scram will be directed if one has not previously been initiated.
| |
| Additional actions will result in emergency RPV & pressurization if suppression pool temperature cannot be controlled.
| |
| * I EPSTG*0002 B - 169 Revision 3 l
| |
| | |
| l l
| |
| qV SRO EXAM KEY Esam Namber NRC-SRO Rev. 0 Ese~ Title NRC SRO EXAM 6 Which of the following will result in the addition of positive reactivity to the reactor 7 (Consider -
| |
| each case separately.)
| |
| : a. LPCS initiaiton during reactor STARTUP with reactor pressure at 500 psig.
| |
| : b. Sudden jet pump differential pressure reduction in one loop with the reactor in the
| |
| - RUN mode, c, Reduction in EllC pressure setpoint by 2 psig with reactor in the RUN mode.
| |
| - d. Initiation of the RCIC turbine during reactor STARTUP with reactor press;e at 150 psig.
| |
| ANSWER:
| |
| : d. Initiation of the RCIC turbine during reactor STARTUP with reactor pressure at 150 psig.
| |
| IDNO: LP# OBJ #
| |
| 314 11L0-318 2 PROCEDURE NUMBER: OTHER:
| |
| ~
| |
| GOP 0001 STEP 3.8 LEVEL 2 i
| |
| [~ NRC KA: I R0: l ~SUOS" 295014 AA1.07 I 4: 4.1 _
| |
| COMMENTS: 7/97 new RO T1 G1 SRO T1 G1
| |
| (
| |
| - \d ' 6
| |
| | |
| ATTACHMENT I PAGE 2 OF 48 PERFORMANCE PACKAGE
| |
| * 3;1.3 - 1In the event forced circulation is imavailable (Reactor Recirculation ptunp off)'
| |
| d and water level is at or below the Minimum Natural Ciiculation Levei lor any reason, then periodic monitoring of Vessel Metal temperatures above and below the intended water level should be initiated.
| |
| 3.2 During startups within 24 hours of previous high powi r operation the reactor xenon
| |
| - inventory will be high and this condition may result in unusually high control rod notch wonhs in certain regions of the reactor (particularly in the range of the 0-16 notch positions). When withdrawino control rods at or name criticality under these conditions.
| |
| exercise mAditional car,6nn to avoid unernected thnet neriods. If a short period is indicated, take prompt corrective action to terminate the indicated shon period. (Ref.
| |
| 2.29) ,
| |
| 3.3 ' If a substained reactor period ofless than 30 seconds is indicated take immediate action to correct this condition by inserting control rods until a reactor period of greater than 30 seconds is indicted. Contact the Operations Shift Superintendent prior to rewithdrawing the inserted control rod (s). (Ref. 2.29) 3.4 Control rod withdrawal shall not be conducted when reactor power is greater than the LPSP _when turbine bypass valves are open or when it is desired to establish significant ~
| |
| steam line drain flow. These conditions may affect turbine first stage , nssure and allow nonconservative control rod withdrawals.
| |
| 3.5 A Reactor Engineer will be used to assist with control rod movements when operating with a Limiting Control Rod Pattem or to approve any unplanned deviations from the rod
| |
| - sequence.
| |
| 3.6 When reactor power is between the low power setpoint and the high power setpoint, rod withdrawals are limited to 4 notches / rod. When reactor po.ver has been increased above high power setpoint, rod withdrawals are limited to 2 notches / rod (Technical Specifications 3.3.2.1).
| |
| 3.7 During power transients, the At-the-Controls Operator should nol take actions to add positive reactivity without the Operations Shift Superimendent's or CRS's pennission.
| |
| - (Ref. 2.37) 3.8 Extreme caution should be ex^ercised when performing operations which may affect
| |
| . reactor pressure and/or reactor coolant temperature to prevent inadvertent reactivity excursions. (Ref.2.38) 3.9 - During sub-critical operations, and until APRM down scale alarms are clear, the potential l
| |
| for cyOg bet;;;;a critical and sub-critical exists. Continuously monitor nuclear instrumentation for indication of criticality. (Ref. 2.38)
| |
| =
| |
| LO v
| |
| GOP-0001 REV - 20E - PAGE 7 OF 90 -
| |
| t
| |
| | |
| l l
| |
| SRO EXAM KEY: i O Exam Number NRC-SRO Rev. O,- Eram Title NRC SRO EXAM 7 .Which of the following methods for alternate control rod insertion during an ATWS REQUIRES
| |
| . the scram " signal" to be reset? !
| |
| Control rod insertion by:
| |
| : a. using the individual control rod scram test switches.
| |
| : b. venting the control rod mechanism over-piston volume, c maximizing CRD cooling water differential pressure,
| |
| : d. venting the scram air header, ANSWER:
| |
| : a. using the individual control rod scram test switches.
| |
| IDNO: LP# OILI #
| |
| 402 liLO-Sl3 4 PROCEDURE NUMBER: OTHER:
| |
| EOP-0005, End 13 LEVEL 3 (295016 AK2.04 1 47 4.1. , ,j COMMENTS: 7/97 New 7
| |
| , - - - ---,,,e-
| |
| | |
| ENCLOSURE 13 OPENING INDIVIDUAL SCRAM TEST SWITCHES O 1.0 1URPOSE To provide instructions for individually scramming con +rol rods with the scram test switches at the respective HCU.
| |
| 2.0 REOUIR FD TOOLS.EOUTPMEN*
| |
| 2.1 NONE 3.0 INSTRUCTIONS 3.1 IE necessary, THEN DEFEAT RPS M ARI logic uips per EOP-0005 ENCLOSURE 12. []
| |
| 3.2 RESET the reactor SCRAM. (lH13*P680) []
| |
| 3.3 At the respective HCU, PLACE the A M B scram test sw Jes to TEST.
| |
| (Rx Bldg EL 114 ft) []
| |
| 3.4 WHEN control rod w *. ion stops, IHEN RETURN tE wram test switches to NORM. (Rx Bldg EL 114 ft.) i]
| |
| | |
| ==4.0 REFERENCES==
| |
| | |
| 4.1 GE Elem Diag 762E429AA, Sh 6, Control Rod Drive Hyd System s
| |
| O l ENCLOSURE 13 l PAGE 1 OF 2 l EOP-0005 REV. 9 PAGE 41 OF 115 l
| |
| _______.-____m__._____m__ -- ___ . _ _ . _ . . - -+-- _
| |
| | |
| l J
| |
| l SRO EXAM KEY -
| |
| O Esam Number NRC.SRO Hev. O Esam Title NRC SRO EXAM .
| |
| t g The Remote Shutdown Panel emergency transfre switches (division I switch on C61'P001 and divsion ll switch on RSS'PNL102) for ADS'SRV 1121.I'0510 are in the EMERGENCY position.
| |
| l Which of the following Control Room handswitches can be used to manually open ADS'SRV 1121 l l'05107 !
| |
| : a. 110Til the Div i "A" and Div 11 *II" solenoid control switches.
| |
| l
| |
| : b. Div i "A" solenoid control switch only.
| |
| : c. Div 11 *11" solenoid control switch only. ,
| |
| : d. Control Room control switches are inoperable.
| |
| ANSWEN:
| |
| : d. Control Room control switches are inoperable.
| |
| IDNO: 1.P # OILI #
| |
| 251 IILO-066 6 PROCEDURE NUMSER: OTHER:
| |
| l AOP-0031 LEVEL 2 I 'NRC K'A!' l RO:' l,~1mi 296016 AK3 03 3 6' 3.7 ]
| |
| .3 COMMENTS: 7/97 new A0 T102
| |
| > SRO T10t f
| |
| O ,
| |
| E A . w.-- - - N,--.w- -
| |
| | |
| LIST OF PRINCIPAL INTERLOCKS BYPASSED BY OPERATION ENCLOSURE 28 FROM Ti1E DIVISION 1/2 SECrlON OF RSS PANEL IC61*P001 PAGE 1 OF 2 SYSTEM I.D. INTERLOCKS AND COMPONENT DESCRIPTION NUMBER CONDITIONS BYPASSED MSL C SAFETY / RELIEF B21*F051C a. Auto actuation for solenoid
| |
| _VALVE MSL D SAFETY / RELIEF B21*F051D a. Auto actuation for solenoid VALVE ,
| |
| MSL C SAFETY / RELIEF B21*F0510 a. Auto actuation for solenoid VALVE ( ADS VALVE) b. ADS actuation for solenoid RHR PUMP A Ef2*C002 N/A RHR '1UTDOWN COOLING E12*F009 a. Isolation signals from RPV water INBl ;OL VALVE level and pressure Rlil: .iUTDOWN COOLING E12*F008 a. Isolation signals from RPV water OU ) ISOL VALVE level and pressure RilR PUMP A SUP PL F12 F004A N/A SUCTION VALVE RHR PUMP A SDC SUCTION E12 F006A N/A VALVE RilR PUMP B SDC SUCTION E12 F006B N/A VALVE RilR HX A CNS FLUSH TO S12 F0ll A a. LOCA signals SUP PL RHR PUMP A TEST RTN TO E12 F024A a. LOCA signals SUP PL i PUMP A SDC E12 F053A a. LOCA signals l~ CTION VALVE Rl!R HX A OUTLET VALVE E12 F003A N/A RHR PUMP A LPCI INJECT E12 F042A a. LOCA signals .
| |
| ISOL VALVE b. Open permissive RHR PUMP A RX READ E12 F023 a. Isolation signals from NSSSS SPRAY VALVE RHR A TO RADWASTE DN E12 F040 a. Isolation signals from NSSSS STREAM ISOL VALVE RHR HX A BYPASS VALVE E12 F048A a. LOCA signals RHR HX A 'NLET VALVE E12 F047A N/A RilR A TO UPPER POOL FPC E12 F037A a. LOCA signals j O ASSIST l
| |
| AOP 0031 REV - 10 A PAGE 33 OF 107
| |
| | |
| SRO EXAM KEY Esam Number NRC.SRO Rev. O Esam Title NRC SRO EXAM ;
| |
| 1hc Control Room is uninhabitable and the Remote Shutdown Panels are being utilized to control '
| |
| 9 .
| |
| . the plant. Reactor level is 20 inches and lowering, and Reactor Pressure is 500 psig.
| |
| With present plant conditions, which of the following systems can be utilized to raise reactor level frorn the Memote Shutdo.vn Panels?
| |
| : a. LPCS
| |
| : a. RilR A
| |
| : c. RCIC >
| |
| : d. lilCS ANSWER:
| |
| : c. RCIC IDNO: LP# Olu #
| |
| 280 llLO-066 2 .
| |
| PROCEDURE NUMBER: OTHER:
| |
| AOP-0031 LEVEL 2 l' NRC KA: l RO: l SRO: !
| |
| 295016 AA1.09 ' 4t 4 .l COMMENTS: 7/97 new
| |
| % Q
| |
| | |
| w -_. .-. _
| |
| CHAPTER 21 -
| |
| 1 O REMOTE SHUTDOWN SYSTEM- .
| |
| I, GENERAL DFRCRIPTION A. System Purpose The Remote Shutdown System (RSS) provides remote manual control of safety related systems necessary for prompt shutdown and subsequent cooldown of the reactor from outside the main control room. These systems are provided with c*ontrols at either the Remote Shutdown Stations or local control stations and include the following (Figure 1):
| |
| : 1. Reactor Core Isolation Cooling (RCIC) 2 Residual Heat Removal (RHR)
| |
| : 3. Main Steam System (MSS)
| |
| : 4. Standby Service Water (SWP)
| |
| : 5. Emergency Diesel Generators (EGS)
| |
| : 6. Ventilation Systems (HVR, HVS, HVC, HVN, HVY, HVP)
| |
| : 7. Containment Atmosphere Monitoring (CMS)
| |
| B. Design Bases The RSS is designed to achieve and maintain hot reactor shutdown conditions, and to subsequently achieve cold shutdown from the following postulated conditions:
| |
| : l. The plant is at normal operating conditions; all plant personnel have been evacuated fmm the main control room and it is inaccessible. Cot trol room operations will not be regained for the duration of the event or until cold shutdown conditions are established.
| |
| : 2. The event that causes the main control room to be inaccessible is such that the reactor operator can manually scram the reactor before leaving the main control room. The reactor can also be scrammed by opening the output breakers of the Reactor Protective System (RPS) logic from outside the main control room, if requiral O
| |
| LOTM 2>7 Page 3 0f 30
| |
| | |
| [
| |
| i O SRO EXAM KEY Esam Number NRC SRO Rev. O Esam Title NRL SRO F. 3 10 Given the followir.' conditions:
| |
| 1he plant is operating at 100% power Offgas lluilding area radiation levels are rising Offgas lluilding and Main Plant Vent Stack effluent radiah M are rising Ofigan hydrogen concentration is 5%
| |
| . Actions are being directed to reduce hydrogen concentration in Offgas to less than 4%
| |
| Which of the following is the reason why the operator is NOT allowed to place the standby recombiner in service to re, luce hydrogen concentrations to less than 4%7 Placing the standby Offgas recombiner in service:
| |
| : a. may exceed the capacity of the preheater reducing temperatures to the tevel where recombination will stop,
| |
| : b. could reduce the flow rate w hen the service air purge is initiated.
| |
| : c. may cause a loss of main condenser vacuum resulting in an MSIV closure if the plant is still at power.
| |
| : d. could provide the ignition source for the hydrogen already present.
| |
| ANSWER:
| |
| : d. could provide the ignition source for the hydrogen already present.
| |
| IDNO: LP# OILI #
| |
| 415 IILO-047 13 PROOEDURE NUMBER: OTHER:
| |
| AOP-0039 LEVEL 3 I ~ SNC K Ai l Rd: I $Rd: ]
| |
| 1295017 AK3.02 i- 3.3 3.5 _j COMMENTS: 7/97 1/97 exam 10 i
| |
| | |
| l 5.1.2 System Leaks
| |
| ( WARNING Entry of personnel should be restripted whenever there is hydrogen downstream of the l condenser, or there is j indication of a leak.
| |
| : 1. Notify Radiation Protection to perform i surveys of the areas and obtain grab l samples as necessary.
| |
| : 2. If a leak in the offgas System is suspected, locate and isolate the leak. l
| |
| : a. Check offgas loop seals, vents, drain l valves, and gaskets that could be leaking in recombiner and air ejector rooms. !
| |
| 5.2 Hydrogen Concentration greater than 4%. !
| |
| CAtJTION After the hydrogan concentration exceeds four percent, any action
| |
| , taken to switch flow paths, bypass system components or to place standby components in service may cause a hydrogen ignition, possibly followed by combustion of the activated charcoal in the system.
| |
| 5.2.1 Honitor the following parameters; a l significant change in one or more of these parameters is indicative of ignition:
| |
| l
| |
| : 1. Profilter and afterfilter differential pressure
| |
| : 2. System Pressure ,
| |
| : 3. Charcoal bed and vault temperatures O AOP-0039 REV - 7 PAGE 4 OF 10 i
| |
| +
| |
| | |
| SRO EXAM KEY l Esans Number NRC SRO Rev. O Esam Title hRC SRO EXAM f 5
| |
| l] During refueling, the leakage rate of the Refueling Cavity has exceeded the capacity of the Drywell and Containment Equipment and floor Drain sumps. A fuel bundle is NOT in a safe storage !
| |
| location.
| |
| Which one of the following systems should be used for eniergency makeup to the Refueling Cavity?
| |
| : a. Control Rod Drive flydraulics
| |
| : b. Condensate
| |
| : c. Reactor Water Cleanup
| |
| : d. CNS service connection ANSWER:
| |
| : b. Condensate IDNO 1,l' # Olti # ,
| |
| 380 liLO-533 8 PROCEDURE NUMBER: OTHER:
| |
| AOP-0027 LEVEL 2 [
| |
| ! 'NRC N At l ' RO: l SRO:
| |
| '295023 AA7.04 4 3.4 ' 3.7 COMMENTS: 7/97 new AOP 0027, p. 5.
| |
| ROT 102 SRO T101 ,
| |
| e O ,,
| |
| | |
| 5.11.12 Reinstall the winch cow, being careful not to strike against the winch i
| |
| ~
| |
| O encoders or the cable load desices. Do not stand on or crush the encoder cables while installing the winch cover.
| |
| i i
| |
| 5.11.13 Store the winch handle, the (2) allen head cap screws, and the allen utench :
| |
| in a 31 ace designated to the Shift Su xrvisor. (Normally at ;
| |
| F42 ?HLN004A/B on Bldg north R 3186' el.)
| |
| . 5.12 For refueling cavity / upper containment pool water level problems, perfonn the following:
| |
| 5.12.1 Ifleakage rate is within the capacity of the dr>well and containment and floor drain sumps, add weer per SOP 0091 FUEL POOL r AND CLEANUP, 5.12.2 IfI rate exceeds the sump capacit ,
| |
| rest , Step 5.11.1), and a. fuel bundle /y or the water level ca '
| |
| consenauve storage location, use any of the following systems for makeup: !
| |
| NOIE I systems are in listedomier
| |
| : l. Condensate injection via feedwater lines O 2. necs i ; ctie-(csr ctie-)
| |
| : 3. Standby Senice Water (IE12*F094/F096) via RHR-B !
| |
| : 4. S Senice Water ausstic to SFC HXs (sia RPCCW) per SOP 1 FUEL POOL COOLING AND CLEANUP 5.12.3 Continue refueling cavity makeup until the fuel bundle or irrulid=1 device is placed in a safe, conservative position.
| |
| 5.12.4 Take actions necessary to repair the leak.
| |
| 5.13 For lower spmt fuel pool water level problems, perform the follcwing:
| |
| 5.13.1 Add water per SOP-0091 FUEL POOL COO 1.ING AND CLEANUP.
| |
| 5.13.2 Take actions necessary to isolate and repair the leak.
| |
| t
| |
| . 5.14 If possible, operse the fuel pool cleanup demineraliar at maximan flow to redum 4
| |
| water activity if fuel bundle damage occuned in Fuel Building spent fuel pooh 5.15 . Operate the reactor water charg filter de.is.er. liar at maximum flow to reduce
| |
| * water activity.
| |
| 1 5.15.1 -If fuel bundla damage is in the cuicim.iro, stop RWCU blowdown to f hotwell are o Radwaste.
| |
| AnIM1027 REV-10 PAM 6 & 10
| |
| | |
| 4 SRO EXAM KEY O Esam Number NRC SRO Hev. 0 Esam Title NRC SRO EXAM 12 Which of the following is the basis for venting primary containment inespective of offsite release rates at 20 psig per EOP 2 (Primary Containment Control)?
| |
| : a. 1his is done to vent the containment before the maximum containment pressure at w hich the primary containment vent valves can be opened and closed is reached.
| |
| : b. This is the maximum containment pressure at which the SRVs can be opened and closed.
| |
| c, This is the maximum pressure capability of the 1,rimary containment, where the most limiting component is the containment equipment hatch.
| |
| : d. 1his is the maximum containment pressure at w hich the Standby Gas Treatment system can be used to vent the containment.
| |
| ANSWER:
| |
| : a. This is done to vent the containment before the maximum containment pressure at u hich the primary containment vent valves can be opened and closed is reached.
| |
| IDNO: LP# Olli #
| |
| 18 ilLO-Si4 $
| |
| PROCEDURE NUMBER: OTHER:
| |
| EPSTG'0007 LEVEL 3 EOP-0002 NRC KAi ~ l ^ RO: l SRO: ~)
| |
| !?95024 iK3 03 t 36' 4.1 l COMMENTS: 7/97 new RO T1 G1 SRO T1 G1 b
| |
| G n
| |
| | |
| EOP-2 Primary Containment Control . CP 4.y,1,s -
| |
| i i,.
| |
| '~
| |
| 1 STEF CF-7-'
| |
| [i .
| |
| 7.
| |
| Step CP 7 is a hold step designed to initiate an emergency containment venting evolution before Containment pressure reaches the Primary Containment Pressure Limit (30 psig). The Primary Containment Pressure Limit is defined to be the maximum containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.
| |
| The basis for determining the Primary Containment Pressure Limit is described in Appendix A.
| |
| At the September 11,1985 ACRS meeting, GSU stated that emergency containment venting would be accomplished through the, Hydrogen Purge System at a pressure of 20 psig.
| |
| O 4
| |
| O EPSTG'0002 B - 162 Revision 3
| |
| | |
| SRO EXAM KEY O Eum Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM ,
| |
| 13 Which of the following conditions constitutes an unsafe condition for the containment?
| |
| RPV Suppression Pool Suppression Pool Pressure Level Temperature
| |
| : a. 1025 psig 17 A. 3 in. 120 degrees F
| |
| : b. 700 psig 16 A. 6 in. 131 degrees F
| |
| : c. 500 psig 18 A. 145 degrees F
| |
| : d. 400 psig 17 ft. 9 in. 154 degrees F ANSWER: ,
| |
| : d. 400 psig 17 A. 9 in. 154 degrees F IDNO: LP# OBJ #
| |
| r 22 IILO-514 8 PROCEDURE NUMBER: OTHER: >
| |
| EOP-0002 LEVEL 3
| |
| ;295026 iK2 06._ .
| |
| 3.6;_ _.3.7 10 2.1.25 28 3.1 _
| |
| L COMMENTS: 7/97 new EOP CURVES MUST BE GIVEN FOR THIS OVESTION. (HCLL & HTCLI RO T102 SRO T101 13
| |
| | |
| O '
| |
| IIEAT CAPACITY TE31PERATURE Llh11T The IIEAT CAPACITY TESIPERATURE LIS11T is defined to be the highest suppression pool temperature at which initiation of RPV depressurization will not result in exceeding either (1) the containment design temperature or (2) the PRIS1ARY CONTAIN5 TENT PRESSURE LIS11T before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent. This temperature is a function of RPV pressure, and the Limit is utilized to preclude failure of the containment or equipment necessary for the safe shutdown of the plant.
| |
| FIGURE 2 HEAT CAPACITY TEMPERATURE UMIT ,
| |
| HCTI.
| |
| 200- ,A : r r , T- r i t- T p 5' m:: 7 CI 7_EL7.~.1 ~"__]L-1 w
| |
| . -. ga_1.3_p.g.}gi_1 p. _ t- +.4_J
| |
| (*r) m. . -
| |
| .q + i_- pp .t.
| |
| . _3.__$ _g .g _ p a 170 i r- t - --
| |
| - 4 1 - F T -1 iso:~I. ,_
| |
| _ T4._ 93
| |
| . .F._T.''l m.
| |
| _i_+a_ +
| |
| _.t p.
| |
| -.4.J._.F+4_
| |
| _+J_C J_pa 144- i t -i m.
| |
| t.
| |
| 4 _4 psyg
| |
| : n. _IW- +. -
| |
| _ p . . n ._
| |
| n
| |
| .. - ~h c
| |
| _ .,. q p,g, p .,. ,. { . ._
| |
| o ' 260 ' 460 ' 660 ' suo ' iom '
| |
| RPV PRESSURE (PDC)
| |
| The llEAT CAPACITY TEN 1PERATURE LIA11T curve illustrated above cornprises two segments: A.B and B.C. For RPV pressure at or below the 511N151Uh1 RPV FLOODING PRESSURE, the rate of energy transfer from the RPV to the containment (with the 511NIh1U51 NUS1BER OF SRVS REQUIRED FOR E51ERGENCY DEPRESSURIZATION open)is, t,y definition of the 511NI51Uh1 RPV FLOODING PRESSURE, within the capacity of the containment vent; thus S:grnent A B is vertical at the A11Nih1Uh1 RPV FLOODING PRESSURE, and Point B is defined by this Pressure and the lower of either (1) the containment
| |
| , design temperature (RBS limiting factor) or (2) the containment temperature at which containment pressure will be at the PR131ARY CONTAIN51ENT PRESSURE Ll511T. For higher RPV pressure, suppression pool heatup during RPV depressurization is proponional to EPSTG'0002 A - 11 Revision 3
| |
| | |
| IIEAT CAPACITY LEVEL LIMIT
| |
| * The llEAT CAPACITY LEVEL LIS11T is dermed to be the higher of either (1) two feet above the elevation of the top of the Mark til horizontal vents or (2) the lowest suppr'ession pool water level at which initiation of RPV depressurization will not result in exceeding the llEAT CAPACITY TEStPERATURE LIS11T. This water levelis a function of the margin to llEAT CAPACITY TESIPERATURE LIMIT, and the Level Limit is utilized in conjunction with the Temperature Limit to preclude failure of the containment or equipment necessary for the safe shutdown of the plant and to preclude loss of the pressure suppression function of the containment F,1 CURE 4 HEAT CAPACITY LEVEL LIMIT - HCLL -
| |
| 20 Ar 1 1 T T rr- 1 r rr rrmcum
| |
| ,. ,4 g lj sArt Iliil o a- i
| |
| -%_ 2riL i U $,,:T,,11h, g,g;rrm@ ..
| |
| u n s errm
| |
| +r h-,4emH,+ttrrn.1,
| |
| : u. L L4 J 4 l 1 L Li-4 l .14 4. W-J .4
| |
| - +H ' WrHe!H+FH+1 e!n , 4 -n ! ' tr rm, io ' '
| |
| !'!'!' !': ' 4.1-4
| |
| * 0 3 4 6 0 ta it to 16 800 MHC N O cn. mu ire u . ,Q O .cw smu em. m.m.m . Q .
| |
| e,,c . O -@
| |
| c ,.e .
| |
| The liEAT CAPACITY LEVEL LIMIT curve illustrated above comprises two segments: A B anc'. B-C. For suppression pool water level at the low suppression pool water level LCO, no margin is required to preclude exceeding the IIEAT CAPACITY TEMPERATURE LIMIT following initiation of RPV depressurization; thus Point A is defined by zeio margin and the low suppression pool water level LCO, and Segment A B reflects a decreasing IIEAT CAPACITY LEVEL LIMIT with increasing margin. For high margin, two feet above the elevation of the top of the Mark 111 horizontal vents is limiting; thus Segment B _C p reflects a constant ilEAT CAPACITY LEVEL LIMIT, specifically two feet above the top of the Mark 111 horizontal vents, EPSTG'0002 A-9 Resision 3 2
| |
| | |
| SRO EXAM KEY :
| |
| Essen Nusuber NRC.SRO Rev. 0 Esam Title NRC SRO EXAM 14 During a major transient requiring use of the EOPs the following plant conditions exist:
| |
| Containment temp 119' 205 deg F f Drywelltemp 14$' 285 deg.F Reactor pressure 200 psig ,
| |
| Which one of the following indications is reliable RPV level information?
| |
| : a. Narrow range when octual RPV levelis +12"
| |
| : b. Upset range when actual RPV levelis +21"
| |
| : c. Wide range when actual RPV levelis 120" ;
| |
| : d. Shutdown range when actual RPV levelis + 46" A!.SWF.Rt
| |
| : d. Shutdown range when actual RPV levelis + 46" t
| |
| IDNO: LP# OILI #
| |
| O 91 ilLO 512 7 PROCEDURE NUMBER: CTHER:
| |
| EOP-0M1 Ceut 1 LEVEL 3 l296027 E K1.02i L ._ 3 g_ _ 3.2
| |
| '295028 iK1.01 3 6! 3.7 COMMENTSt 1/97 n9W requires EOP 1 Caution 1 ROT 102 '
| |
| SRO T101 k
| |
| - ,-----,,,,,.,--,--,.-,,...-,--,er n- . , , , - ~ , . , . - . . , - . . , , - - - - , ~ . - , - . , , - , w , a -- n., - - - - ----, , , , , - . - , ,, ,,-,w.w,,,,.,. ,.
| |
| | |
| l lpi .
| |
| l i .
| |
| CAUTIONS W ICM At0V18[
| |
| ? A,
| |
| , A rg c e,;,
| |
| a 'm8RtaCTOR O ~
| |
| ! CAUTION #1
| |
| 'fg< h d WIN DON Nt ron0w!NG CON 0!f!0NS ARE $Afl$rICD, 4 3., PJLM NAf INstuvutNT WAY pt sit 0 to CC't#WINC
| |
| ,., T t'. R*V nArtR LtWL
| |
| , e ;
| |
| .. Nt etW. NeiR Au ins,,vWtN1 R , Aa
| |
| , ,;, IN Nt SArt 20NE Or Nt Rp47 fr[0veg i j lt!Guat il l [ RPV SATURAT10N 4WatRAfvRt - RPv$t l i i i i g i > l i,i
| |
| ! 0 330-- , a - m g .; --g l
| |
| t i"lg'), om- - Ht-'5''41g-4 J L '
| |
| i m A=-, ,- ,#+ -s- H ,w- H ,s-i c ,, , r4-l m- - ' -.J pA L AJ.
| |
| i rtWP c*r>
| |
| t'"'" 3*- h-f- L,4hy'-Er Jrs q 4 -l- b l
| |
| 230 -
| |
| 200 : & 4 -L M' l--F r-4 -L I I I ' ' ' I I ' i 1 l 0 200 A00 600 Boo 1000 1200 l RPV PRESSVRC (P$10) 2 r04 ( Acw or Nt roLLow!No INsmuWtNis NC '
| |
| INSTRVWENT Rf AOS ABOW NC WIN!WuW !N0!CAft3 LEWL Ass 0CI ATED WIN 'HC HIG"(ST 'luptRatbat
| |
| , SECTION R0 ''''"''"""*'''"'"*''''''***"'''" l W A u 0* flup At CL 945 f t, 40c*r
| |
| _ rutt CON T A!NWINT NAeR0w w!00 20NC f!WplRATVRE y (e) r 200 -- - - - - - - - < - - - - - - - - - - - - -
| |
| Ci " ' "
| |
| W%!fM Cm 1 !N. -142 IN. -310 IN.
| |
| vtACfoR o.M.tR x , , -. .......- ----- - - - - - .
| |
| h 0 IN. -1$9 !N. -310 IN.
| |
| WR!ry Nt RC ActJR W0pt l SwltCH IN SHufDOWN J b 78 l
| |
| j
| |
| ' ~ ) Rv ten W paocr00*f\ /
| |
| AOP-0001, '
| |
| j i
| |
| j W Ax CTWY ftWP AT CL tig rf, 200*r i
| |
| ,y VPsti SHuf00nN f i
| |
| og IN, $2 IN,
| |
| ; **"[.',' rVRt y_. .______.._____ l CL $45 f f i 24 IN. 34 !N i a '4 200 -- . _ - - _ - - . _---_
| |
| l 9 !N. 21 !N j '32 i
| |
| | |
| i i
| |
| l i l
| |
| l t lN \
| |
| l /
| |
| st C AUTIONS . l gg $[Q't$(
| |
| l 3A RCAC704 i i as CANN0f ,
| |
| , i l / l
| |
| ' CAUTION #1 !
| |
| i 2
| |
| .1r a
| |
| v IECM Bom twt FOLLOW!NG CONDif10NS ARE $Affsrtt0, 1
| |
| 4, 2ht.N THAT INStRVWENT MAY DC V5t0 TO DCTERWINE q ,4, . .yt '. RPV WATER LEVEL
| |
| ~ E, t itWP NE AR ALL IN$fRVWENT RU t.N,Nn m - OrneRPm &NA+D' l
| |
| l Irtovat 11 g RPV $ATURAf!0N ftWPCRAfVRC a RPv57 i
| |
| l I I il 4 I
| |
| -.,y~ .p''pp-.a-y ~ _r u -l w' I 1
| |
| l i 0. ,,0 l
| |
| , gg7,sn.
| |
| 4 $0 - ,7 q L . q. q- q-
| |
| ~
| |
| ! CfWT 9~
| |
| L L
| |
| ! ftWP (*r) 350 -
| |
| '-d -t r D~brd~-p Mhr i t' " ' " 3=- , -r-20 u -u +r , r i ,d -W r-e -L i 200 ! ! ! ! -,! -Hr ' ! ! ! ! !!
| |
| l 0 200 400 000 b00 1000 1200
| |
| ) RPV PRES $uRE (P$!G) l 2 FOR CACH Or fHt FOLLOW 1NG INsfRVWtNis, fut
| |
| ; INSTRWitNT RCA01 ABOW TN! WIN!WVW IN0!CAft3
| |
| : LEWL Ass 0CIATED w!TH THE MICHCst itWPERATURt ,
| |
| L SECTION R0 "''""'''"""*'"'"'""'''''''"""C''""" '
| |
| i W AE Dw itWP At CL 145 Ft. 400'r
| |
| - IVEL !
| |
| CONTAINWCNT NARROW W10C ZONE l ftWP(R A TV4C j
| |
| (.F) 200 -- - - - - * - - - - - - - - - - - -
| |
| I j [L M 9 Ff i g gy, 142 IN. -310 IN.
| |
| WON!704 AND CONTROL
| |
| { ff(ACTOR PontR -----
| |
| 100 -- -- - - - - - - - - - - - - -
| |
| ! 4 0 IN. -159 IN. -310 IN.
| |
| 1 WRtrY THE RE ACTOR Woot
| |
| ; .yj
| |
| $W1?CH IN $HVfDonN i
| |
| 4 78 .
| |
| 4 x A0P-= t. x i
| |
| / ** F1' AW **0CCDutt / '
| |
| 1 4 M AX Ciut f!WP At Ct 119 Ft. 200*F t
| |
| ~
| |
| uPS[f l
| |
| ; 00 - . SHVfD_owN ._, ,
| |
| 4 49 IN, $2 IN. l ORT M j j .
| |
| 'Cuj.",jfuRt 33, __ ______ .._____ j l
| |
| CL 145 FT j f, 20 !N. 34 IN i .
| |
| j *N 300 _ ..._
| |
| i
| |
| ' {
| |
| l 4 9 !N. 21 IN. l
| |
| ,33 .
| |
| - +
| |
| - . - . _ - ~ . - - - . _ . . - . - - . - . . . . . - , . - . . . . . .
| |
| | |
| l l
| |
| l SRO EXAM KEY r Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM 15 in Emergency Depressurization, Step ED 3 asks,"Is Suppression Pool Level Above 13 ft? What is .
| |
| the significance of this level?
| |
| : a. It ensures a vortex will not be created when SRVs are opened.
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containment does not occur, 3
| |
| : c. It is required to present loss of NPSil to the RilR Pumps.
| |
| : d. It ensures there is enough water to cover the horizontal venta, ANSWER
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containment doec not occur.
| |
| - lDNO: LP# Olu #
| |
| 23 IIL O-512 5 PROCEDURE NUMBER: OTHER:
| |
| ' LEVEL 2 EOP-0001 EOP-0001 BASES
| |
| {~ NRC KAi
| |
| '295030 f K2.08 l~RO$ l ~5 MOI 3.6i 3.8 ]g COMMENTS: 7/97 new ROT 102 SRO Tt01 I
| |
| 5
| |
| \
| |
| 13
| |
| | |
| EOP 4 Contingencies . ED STEP ED 3 4 If suppression pool water is below 13 ft, it cannot be assured that the SRV discharge will be condensed and opening an SRV may cause pressurizatic.1 of the containment airspace. Since the extent of this pressurization cannot be predicted and may exceed the pressure capability of the primary containment. such operation is prohibited. The operator is sent to ED 6 to u:e means other than the SRVs for Emergency Depressurization if suppression poollevelis below 13 ft.
| |
| I .
| |
| O EPSTG'0002 B 269 Revision 3
| |
| | |
| i SRO EXAM KEY :
| |
| . Esse Number NRC SRO Rev. 0 Esam Title NRC SR0 EXAM 16 EOP-4 (Primary Containment I'looding) is executed to Good containment. 1
| |
| : 'the containment level band specified by EOP-4 is between 62 f1 and the Maximom Containment i Water Level Limit (MCWLL)of 85 ft.
| |
| i j Which of the following ranges corresponds to the core level band specified by E0P4' i j
| |
| : a. 143 to + 1331nches ;
| |
| : b. 162 to + i14 inches t
| |
| ; c. 193 to 451 inches
| |
| : d. 20$ to +71 inches t ANSWER:
| |
| : b. 162 to 4114 inches i
| |
| l IDNO: LP# Olti #
| |
| 224 IILO 512 5 PROCEDURE NUMBER: OTHER:
| |
| EOP-0004 LEVEL 3
| |
| ~ NRC KA: l RO$ l SRO:
| |
| '295031 I A2 01 .
| |
| ! 4 6- 46 COMMENTS: 7/97 new EOP BASES says MCWLL is 85 ft.RBS NRC 0195
| |
| = RO 1101 SRO T1 G1 l
| |
| l 16
| |
| -,,,.--,,,--,-.4 .c .m ,..m.a , . , , . .v 9 -t v*9 ,r w vv -r---r- -r--m.-rvr-re dw- *we -* e- -ma- w M -- *- .--+-. - .wew =m
| |
| | |
| e.
| |
| EOP4 Contingencies CF STEPS CF 8, CF 9 Steps CF 8 and CF 9 specify the primary containment water level range which assures that-
| |
| : 1. If the RPV is successfully vented. RPV water level is at least to the elevation of the top of the active fuel, and therefore, adequate core cooling is achieved.
| |
| : 2. Primary Containment water level does not exceed the Maximum containment Water Level Limit.
| |
| Once the primary containment water level is established within the specified range, it is appropriate to recirculate RPV injection flow, preferentially using systems which take a suction from outside sources only as needed to maintain primary containment water level at the elevation .-
| |
| corres inding to the top of the active fuel. ,
| |
| The same systems listed in step CF 2 are included with the following exceptions:
| |
| * No restriction regarding location of the suction source is required since flood up
| |
| - of the containment is essentially complete.
| |
| * RCIC system is not included because it is inoperable at such low RPV pressures and high suppression pool water level.
| |
| * Head spray is added as an alternate path to inject water directly inside the RPV.
| |
| * RHR A(B) SDC injection added as another path.
| |
| * LPCI and LPCS systems added.
| |
| EPSTG'0002 B - 300 Revision 3
| |
| | |
| SRO EXAM KEY !
| |
| t Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM :
| |
| I 17 A scram frorn full pow er has occurred as a result of high drywc!! pressure due to a leak. A'l automatic functions and isolations occurred per design, except that not all the control rods inserted a
| |
| fully. The Standby Liquid Control (SLC) pump "A" was initiated per applicable emergency ;
| |
| prcedures.
| |
| f As the operator monitors SLC parameters, SLC tank level indicates zero (0) inches.
| |
| Which one (1) of the following actions should be taken next?
| |
| l
| |
| : a. St .rt SLC Pump *B",
| |
| : b. Mestore power to the SLC tank level indication by restoring power to Nils.
| |
| MCC102A.
| |
| : c. Initiate actions to inject alternate SLC per Encl.'l5.
| |
| ; d. Install Encl.16 Bypassing CNTMT Instrument Air Isolation interlocks ANSWER:
| |
| : d. Install Encl.16 Bypassing CNTMT Instrument Air Isolation Interlocks O
| |
| IDNO: LP# Olti #
| |
| 416 11L0 513 4 PROCEDURE NUMBER: OTHER:
| |
| EOP.1 A Encl.16 Level 2 I' ~'NRC KAt l RO:" I SRO$
| |
| '296037 EA1,04 _ j . 4.5, . , 4. 5
| |
| !295037E A2.03 .] _ 4.3,. _ _4.4
| |
| '211000 K5 06 ! 3= 3.2 _
| |
| COMMENTS: 7/97 new a
| |
| P O n w r---y, ,,q-e v -r--' ,,y-- % - e , , - - ,-v-r--e-- m - s v e r w == 'rw +- w -ww v'-vw-v--- -,----v omv--=-mm-- = ww-ww-- --*
| |
| | |
| y 1
| |
| mi posas no 4404 te - ,
| |
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| m U DORON INACTION \
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| * f
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| 4 EXECUTE CONCURRENTLY
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| * 1r If 100 vat attri a=1. tertatino net toott
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| e FAlltet 10 Ot.tht#91N ACT!'JN% ?
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| |
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| |
| aps Lecte tages %o ng f.hTMl'f.'70"/L=
| |
| = M Act pelvtovat satam itst U
| |
| . 72c"4 L"','if."". Ek ' ws!LE ExtcuTING THE FOLLOWING STEPS ao*W " '#3 IF THEN se ta== .An, irm wie w wino n.c DeCP9 TO 9 4ALLOS PVWP
| |
| *G l v g Isact sanc3e swio twt arv snfw m.c e spur A gg g, put BC.I DOTH. DDI ATD80 CfW1 INlt# AIR I1(LA90N INTt4Loott (tiet it)10 EJP*LT Arn 101*C RC IM LtwL suaalit g aggJnett naa.y
| |
| .ac new 10. .
| |
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| u CIF F
| |
| thtt? DORON INTO M IPy u!1w ATLAhAft SLC ImKCTtch(Dut tE) g h04 3 v
| |
| l 2 I *""tr. a m a - - I n WHEN 110 LS (DetL.te10) or coman MA4 tifu
| |
| \ DtXCD IW10 b
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| mos.g a
| |
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| |
| "*b
| |
| | |
| 4 SRO EXAM KEY O Esam Number NRC-SRO Rev. 0 Etam T6tle NRC SRO IIXAM Ig EOP 3, Radirecth e Release Control, has been entered.
| |
| Which of the following is the reason that the operator is directed to ensure that the Turbine fluilding Ventilation fans are running?
| |
| : a. Reduce radioactive releases below General I'mergency levels
| |
| : b. Prevent radioactive releases from the Turbine Building.
| |
| : c. Filter radioactivity from the Turbine Ilullding atmosphere.
| |
| : d. Provide a monitored release point.
| |
| ANSWER:
| |
| : d. P ovide a monitored release point IDNO: LP# Olki #
| |
| 3ll llLO-515 4 PROCEDURE NUMBER: OTHER:
| |
| EOP 3 RR 1 LEVEL 2
| |
| , , g , ,,,
| |
| M6 18 r A1.06 i 3.6: 36 COMMENTSI 7/97 new s
| |
| | |
| EOP 3 Radioactive Release Control RR
| |
| . --_ _ . , _ STEP RR-1 .__._. . . .__
| |
| ~7
| |
| ==
| |
| Step RR l is an override statement which applies throughout this procedure. Continued personnel access to the turbine building may be essential for responding to emergency conditions or transients which may degrade into emergencies. Since the turbine building is not an air tight structure, a radicactive release inside the turbine building would not only limit personnel access but would eventually lead to an unmonitored ground level release.
| |
| Operation of the Turbine Building Ventil.ition System preserves accessibility of the turbine building for personnel access. Additionally, restaning the turbine building ventilation will assure that all or most of the radioactivity released to the turbine building will be discharged through a monitored release point. This will greatly aid in determining the overall release to the environment during an accident.
| |
| O 1
| |
| O EPSTG'0002 B - 231 Resision 3
| |
| | |
| SRO EXAM KEY O Etam Number ,NRC-SRO Rev. O Etam Title NRC SRO fiXAM l9 Which one of the following is the !! ASIS for maintaining the refueling cavity pool 23 feet above the top of the reactor pressure vessel flange during refueling?
| |
| : a. To provide adequate net positive suction head to the l'uel Pool Cooling Cleanup ,
| |
| Pump ,
| |
| : b. To maintais, a reservoir of water for suppression pool makeup.
| |
| : c. To provide spent f el decay heat removal for 7 days without makeup.
| |
| : d. To remove the lodine gap activity released from a fuel rupture.
| |
| ANSWER:
| |
| : d. To remove the lodine gap activity released from a fuel rupture.
| |
| IDNO: LP# OHJ#
| |
| 389 IILO-022 Sc PROCEDURE NUMBER: OTHER:
| |
| TS 3 9 6 LEVEL 2
| |
| ( ._ .- -. . . - . - - .-
| |
| ,0 2.3.10 _ . _ 4 _ 2. 9, , . 3.3
| |
| '295023 AK1.01 ! 36' 4 .1 _..,
| |
| COMMENTSt 7/97 new Technical Specificaten Bases 3.9.6 RO & SRO T4 03 ABILITY TO PERFORM PROC. TO REDUCE RAD LEVELS RO T1 G3 REFUELING ACCIDENTS SRO T101 19
| |
| | |
| RPV Water Level-Irradiated Fuel B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water level-Irradiated Fuel BASES BACKGROUND The movement of irradiated fuel assemblies within the RPV requires a minimum water level of 23 ft above the top of the RPV flange. During refueling, this maintains a sufficient water level in the upper conta'nment pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. I and2). Sufficient iodine activity would be retained to limit offsite doses from the accident to < 25% of 10 CFR 100 limits, as provided by the guidance of Reference 3.
| |
| APPLICABLE During movemnt of irradiated fuel assemblies, the water SAFETY ANALYSES level in tiie RPV is an initial condition design parameter in
| |
| ~
| |
| the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 100 to be used in the accident analysis for iodine. This Os relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retait.ed by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine inventory (Ref. 1).
| |
| Analysis of the fuel handling accident inside containment is described in Referer.ce 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours prior tc fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequatelv captured by the water, and that offsite doses are maintained within allowable limits (Ref.f).
| |
| While the worst case assumptions include the dropping of the irradiated fuel assemoly being handled onto the reactor core, the >ossibility exi!ts of the dropped assembly striking t1e RPV flange and releasing fissio products.
| |
| Therefore, the minimum depth for water coverage to ensure acceptable radiological consequences is specified from the RPV flange. ,
| |
| (continued)
| |
| (
| |
| RIVER BEND B 3.9-19 Revision No. 0
| |
| | |
| ._ _ . _ _ . _ . . - . _ . . . _ . _ . .. . ~ . . . - _ _ . _ . _ _. __.
| |
| t SRO EXAM KEY Q Enem Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM <
| |
| i 20 Which of the following conditions still constitutes " Adequate Core Cooling"?
| |
| NOTE: Only the injection sources stated are injecting. Regard each situation separately,
| |
| : s. ATWS in progress, the feed system is maintaining level between 205 inches and 195 inches, MSIVs are open,
| |
| : b. All rods in, MSIV/ ADS valves are closed, RPV level is 200 inches and RPV pressure is 200 psig,
| |
| : c. All rods in, RCIC is injecting,1 ADS valve is open, RPV level at 210 inches and MSIVs are closci
| |
| ' d. ATWS in progress, CRD, RC;C and SLC (with lloron) are injecting, RPV level is 200 inches and MSIVs are open.
| |
| ANSWER: j
| |
| : b. All rods in, MSIV/ ADS valves are closed, RPV level is 200 inches and RPV pressure is 200 psig.
| |
| IDNO: LP# 08.1#
| |
| O 230 11L0-511 3 PROCEDURE NUMBER: OTHER:
| |
| EOP.1 A RLA 11 LEV::L 4 1295031 EK1.01 1 4.6, 4.7 _
| |
| COMMENTS: 7/97 fWW 7
| |
| J 0 :o
| |
| | |
| EOP-1 A RPV Control- ATWS RLA o ::
| |
| r _ ,
| |
| STEP RLA-11 ,
| |
| if RPV water level cannot Se maintained within the limits previously specified (-162 in. to 51 in ), an alternate control band with a lower limit is defined (-193 in. to 51 in.). The widened RPV water level control band provides added operational flexibility while still assuring adequate core cooling through steam cooling. By establishing this alternate RPV viater level control band, additional time may be available to place injection systems not yet operating into service.
| |
| The wider control band also accommodates controlling RPV water level without employing additional contingency actions for a condition where a break exists between the top of active fuel and the hiinimum Steam Cooling RPV Water Level and where injection flow cannot overcome break flow. ,
| |
| The Niinimum Steam Cooling RPV (Vater Level is defined to be the lowest water level at -
| |
| which the covered portion of the core will generate snHicient steam flow to preclude any cladding '
| |
| temperature in the uncovered portion of the core from exceeding 15007 This level corresponds to an RPV water level of-193 inches at River Bend.
| |
| The basis for determining the hiinimum Steam Cooling RPV Water level is described in Appendix A.
| |
| Figure B-8 illustrates the manner in which the hiinimum Steam Cooling RPV Water Level is determined.
| |
| * Curve A represents the steam flow required to maintain peak clad temperature less than 1500'F ten minutes after shutdown from rated power, assuming the most limiting top-peaked power shape prior to shutdown.
| |
| - Curve B represents the actual steam flow generated by the reactor core as a functiort of RPV water level.
| |
| The intersection of Curve A and Curve B defines the RPV water level at which the ste4m flow generated by a partially uncovered core exactly meets the steam flow required to cool the uncovered portion. This water level has been calculated to occur at a core elevation corresponding to fuel node 19 of 24 for all BWR cores, assuming water injected into the RPV is at a temperature of 70'F.
| |
| 3PSTG'0002 B 87 Revision 3
| |
| | |
| EOP I A RPV Control ATWS - RLA in comparis to the Minimum Zero. injection RPV Water Level (-205 in.), the Minimum
| |
| . Steam Cooling RPV Water Level is a slightly higher RPV water level. This can be attributed to two key factors:
| |
| : 1. Injection of subcooled water requires that part of the energy which would be used to generate tha steam for cooling the uncovered portion of the core must now be expended in heating subcooled water to saturation temperature. (Minimum Zero-Injection RPV Waitr Level is calculated assuming no injection into the RPV).
| |
| 4 2; More steam is required to maintain clad temperature below 1500'F as compared to
| |
| ' the 1800"F limit assumed for the Minimum Zero-Injection RPV Water Level calculation. i
| |
| : l l
| |
| 4
| |
| ;O s
| |
| i t
| |
| : i. .
| |
| l lO l
| |
| 3PSTG'0602 B - 88 Revision 3 t
| |
| l .
| |
| | |
| g Esem Number NRC SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 2l The following conditions exist:
| |
| r The reactor has been scrammed.
| |
| 4 control rods are not fully inserted.
| |
| - 7here is'a coolant leak into containment.
| |
| Conditions degrade and require implementation of Emergency RPV Depressurization per EOP-4A, Which of the following actions are required?
| |
| : a. Immediately open 7 ADS /SRVs,
| |
| : b. Rapidily depressurize the RPV using bypass valves an1 MSL drains
| |
| : c. Terminate and prevent injection from all sources except SLC, CRD, & RCIC then open 7 ADS /SRVs valves.
| |
| : d. Close the MSIVs, MSL Drains, and RCIC Steam isolation Valves then open 7 ADS /SRVs valves.
| |
| ANSWER:
| |
| : c. Terminate and prevent injection from all sources except SLC, CRD, & RCIC then open 7 ADS /SRVs valves.
| |
| IDNO: LP# OBJ #
| |
| 437 IILO-512 7 PROCEDURE NUMB 2R: OTHER:
| |
| EOP-0001 LEVEL 2 EOP-0004A
| |
| !295024 EK3.04 , 3.7! 4.1
| |
| !295031 N3.05 ~I ~ ~4.2~I 4.3 ,_,
| |
| COMMENTS: 7/97 new 21
| |
| | |
| EOP-4A Contingencies - ATWS - AED L
| |
| . .- STEP AED-1 Ifit cannot be determinui that the reactor will remain shutdown under all conditions without boron, action to depressurize the RPV must wait until confirmation that injection into the RPV is terminated and prevented. Failure to do so may result in the rapid injection of a large quantity of relatively cold, unborated water from low pressure systems as RPV pressure decreases below the shutoff head of these pumps. Such an occurrence could quickly dilute in-core boron concentration and reduce water temperature in the core region. Sufficient positive reactivity might be added in this way to induce a reactor power excursion large enough to damage the core.
| |
| Injection from CRD and boron injection systems is allowed to continue because operation of these systems may be required to achieve or maintain the reactor shutdown. Termination of RCIC System operation is not necessary becpse the injection flow rate from this system is small. ,
| |
| continued operation of the turbine aids in depressurizing the RPV, and operation during RPV . .
| |
| depressurization is not expected to result in significant injection flow rate variations. l Terminat;ng and preventing injection shall be accomplished per the following:
| |
| O kJ 1. Condensate /Feedwalg - Place the Master Controller in. Manual and drive all Feed Reg valves full closed; place the Startup Feed Reg valve in Manual and drive it full closed.
| |
| : 2. HPCS - While holding the control switch for the injection valve in the closed position, arm and depress the manualinitiation push button. When the pump has started and come to full speed, secure the pump at the SRO's direction. When available, dispatch an operator to monitor and secure the diesel.
| |
| 3 LPCS/LPCI - If an initiation signal is present, manually override the injection valves by taking the control switches to the closed position. The pumps shall remain running. If an initiation signal is not present, prevent pump start and injection valve opening by defeating the initiation logic. Enclosure 27 provides guidance.
| |
| : 4. ECCS Keeofill/ Condensate Transfer - Close the appropriate LPCS/LPCI injection valves
| |
| : 5. Service Water / Fire Water - Close the Containment Fleod valves, E12*F094 and E12*F096 O
| |
| v EPSTG'0002 B - 327 Revision 3
| |
| | |
| g Esam Number NRC-SRO SRO EXAM KEY Rey, O Exam Title NRC SRO EXAM 22 All high pressure injection has been lost following a Reactor scram and RPV water level transient.
| |
| RPV water level is 162 and lowcring slowly. RilR and LPCS are running on minimum flow. The CRS directs emergency depressurization. Why must at least 4 SRVs be opened to accomplish emergency depressurization under these conditions?
| |
| : a. Li'CS and RHR will be injecting prior to RPV level reaching the minimum steam cooling level.
| |
| : b. The level swell from four open SRVs will keep the core submerged until RilR and LPCS are injecting at rated Dow,
| |
| : c. LPCS alone can re0ood the core prior to the core uncovery time exceeding the maximum core uncovery time limit,
| |
| : d. Enough SRV steam Cow to cool the core will exist at a pressure that RilR can make up for the steam flow.
| |
| ANSWER:
| |
| : d. Enough SRV steam How to cool the core will exist at the pressure Rl!R can make up for the steam now.
| |
| IDNO: LP# OILI #
| |
| 417 IILO-512 7 PROCEDURE NUMBER: OTHER:
| |
| EOP-0004 ED-5 LEVEL 3 1295031 EA1.08 I 3.8' 3.9 _
| |
| COMMENTS: 7/97 new 22
| |
| | |
| i EOP-4 Contingencies ED O
| |
| O . .
| |
| ~
| |
| STER ED i TheofMinimum the greater either: Number of SRVs required for Emergency Depressurization is defined to b The least number of SRVs which, if opened, will remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow.
| |
| The least number of SRVs which correspond to a Minimum Alternate RPV Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate RPV Flooding Prevure.
| |
| This number is utilized to assure the RPV will depressurize and remain dep;essurized -
| |
| ~
| |
| when emergency RPV depressurization is required. The basis for determining the Minimum Number of SRVs required for Emergency Depressurization is described in Appendix A.
| |
| A k Figure B 4 illustrates the manner in which the Minimum Number of SRVs required for :
| |
| Emergency Depressurizatio-is determined.
| |
| Curve A is a plot oflow pressure ECCS flow as a function of RPV pressure.
| |
| Curve B is a plot of the steam flow that corresponds to the removal of core decay heat by boiling ten minutes after reactor shutdown from related power.
| |
| Curve C is the steam flow defining the Minimum Alternate RPV Flooding Pressure.
| |
| Curve Di through DX are plots of steam flow through one to X number of open SRVs as a function of RPV pressure i
| |
| b) v l
| |
| ]
| |
| EPSTG'0002 B-271 l
| |
| Resision 3 i'
| |
| | |
| .- -. .- - - = - . . - - - - - . . . . . . - . .. .
| |
| SRO EXAM KEY I] Etam Number NRC-SRO Rev. O Eram Title NRC SRO EXAM ,
| |
| 23 Regarding the flydrogen DeGagration Overpressure Limit (llDOL) curve, as containment pressure [
| |
| increases, the maximum allow ed hydrogen concentration in percent (%) decreases.
| |
| Which of the following is the reason for this relationship?
| |
| : a. As containment pressure increases, the capabilities of the flydrogen Recombiners to remove hydrogen is decreased,
| |
| : b. This ensures a hydrogen deaagration at the limit combined with current pressure will not exceed containment overpressure failure limits.
| |
| : c. The containment hydrogen analyzer system response time is adversely affected as pressure increases.
| |
| : d. As containment pressure increases, the denagration pressure of hydrogen decreases requiring a lower concentration of hydrogen.
| |
| ANSWER:
| |
| b.1his ensures a hydrogen de0agration at the limit combined with current pressure will not exceed containment overpressure failure limits.
| |
| IDNO: LP# OBJ #
| |
| 404 IILO-514 2 PROCEDURE NUMBER: OTHER:
| |
| EPSTG'0002, App. A LEVEL 3 E O P-1 F gure 5 500000 EK1.01 1 3.3 3.9 i a
| |
| COMMENTS: 7/97 new RO T1 G1 SRO T1 G1 O
| |
| | |
| ~
| |
| O V CONTAINMENT HYDROGEN DEFLAGRATION OVERPRESSURE LIMIT (HDOL)
| |
| The purpose of the Containment HDOL is to assure that the postulated combustion of hydrogen and oxygen in the containment will not result in an overpressurization that will structurally fail the containment.
| |
| FIGURE 5 I
| |
| HYDROGEN DEFLACRATION OVERPRESSURE LIMIT HOOL t,2-3 -- -r i r T ' r T 'J_ L .'L J -
| |
| si- _ u L + J _' p +h - h' cmT to- T F r -i , Tr H-F+-4 H2 g. .q _ L .L y _ QgJ _ (, ,L .,J CON:ENTRATioN (x vot.) a-A L
| |
| Bj l Jl 7, , gj l l l l F7 l 7- -1 r -l- -i - F + -i g i
| |
| *-'TT WC T 5---t -4 r t -'-
| |
| ~ kh t- t -1 r1 4 l !: !!! : !0; :
| |
| O 5 10 1S 20 25 30 SS 40 45 50 55 cTWT PRESSURE (PSG)
| |
| The Containment HDOL curve illustrated above comprises three segments: A-B, B-C, and C-D. For Containment Hydrogen concentrations above 8.1%, the ultimate containment capacity could be exceeded as a result of a deflagration when the containment steam concentration is zero; thus segment A B is horizontal at 8.1%. For high containment pressure, containment peak pressure during a deflagration is decreasing HDOL with increasing containment pressure. For higher cor.tainment pressure, the pressure capability of the containment is limiting; thus Segment C-D is venical at 10 psig below the limiting pressure of the limiting location. (This 10 psig margin is provided to mitigate the consequences of a deflagration that could potentially occur at the maximum pressure as defmed by the HDOL).
| |
| The containment initial pressure ha., taken into account the increase in pressure as a result of assuming that the hydrogen has been added to a sealed containment. Since the hydrogen concentration is a percent of the volume and the densities of t'he enclosed gases are related to the initial pressure of the containment, the amount of hydrogen present increases as the initial Q pressure increases for equal volume percentages of hydrogen. Likewise, as initial containment EPSTG'0002 A.4 Revision 3
| |
| | |
| pressure increases, the margin to the failure pressure decreases. Both of these factors result in decreasing the allowable concentrations as the pre combustion containment pressure increases.
| |
| The pressure increase from a burn results from the heating of the atmosphere to high temperatures The maximum pressure occurs within 3 to 30 seconds, depending on hydrogen and steam concentrations. Following tLe burn, the atmosphere loses its energy to the heat sinks in the containment and the pressure decays to near the initial pre-burn pressure. The pressure decay occurs within a few minutes.
| |
| Containment Volume Assumptions:
| |
| : 1. The gases are uniformly mixed in containment when combustion occurs.
| |
| : 2. No fans are operating during combustion.
| |
| : 3. Containment is at the saturated atmosphere temperature (i e., relative humidity =
| |
| : 1) for all cases where the stearr concentration is greater than zero.
| |
| : 4. The Technical Specification Limiting Condition for Operation (LCO)is used for the initial containment temperature prior to the addition of steam and non-condensibles to the containn)ent. Initial containment pressure is one atmosphere. -
| |
| (14,7 psig) -
| |
| : 5. Combustion durations vary with hydrogen and steam concentrations. Combustion ,
| |
| times as a function of hydrogen concentration are based on the large volumetric burns conducted at the Nevada Test Site.
| |
| p 6. Before combustion initiation, the containment is assumed to be pressurized due to V a loss of coolant accident (LOCA) in the drywell by additional air equivalent to the
| |
| - drywell air mass at dmvell LCO temperature. The addition is above and beyond the air, steam, and hydrogen existing in containment at atmospheric conditions.
| |
| This containment atmosphere generates a more limiting pressure environment during a hydrogen burn.
| |
| T For cases with no steam in the atmosphere, hydrogen is assumed to enter the containment at 167 F, corresponding to the anticipated suppression pool temperature, and equilibrate with the air without heat transfer to the structures.
| |
| Plant-specific data required to calculate the Containment HDOL are as follows:
| |
| : 1. Best estimate of the actual maximum pressure the primary' containment can withstand -(53 psig).
| |
| 2 Drywell free volume -(286,467 fP).
| |
| : 3. Containment net free volume (wetwell, intermediate and upper containment volumes)-(1,264,953 ff).
| |
| : 4. Drywell LCO temperature -(145 F)
| |
| : 5. Containment LCO temperature - (90*F).
| |
| : 6. Normal operating drywell pressure - (0.0 psig).
| |
| : 7. Normal operating containment pressure - (0.0 psig).
| |
| : 8. Temperature of Hydrogen when released through the suppression pool. (Peak g suppression pool temperature during blowdown of a Design Basis Accident)'-
| |
| (167 F).
| |
| . EPSTG'0002 A-5 Revision 3
| |
| | |
| SRO EXAM KEY :
| |
| Esam Number NRC SRO Rev. O Exam Title NRC SRO EXAM 24 The plant was operating at 100% power when a scram signal was generated and the reactor failed to scram. EOP l A directs downshifting Recirc Pumps.
| |
| Which of the following describes the reason Recirc Pumps are down shifted prior to tripping?
| |
| ~
| |
| Tripping the Recire Pumps could result in:
| |
| : a. entering the region of thermal / hydraulic instability.
| |
| : b. an excessive feedwater temperature reduction rate that will cause power to increase rapidly,
| |
| : c. a large level shrink which could cause isolation signals complicating the event.
| |
| : d. a reactor level swell which could result in a main turbine trip.
| |
| ANSWER:
| |
| : d. a reactor level swell which could result in a main turbine trip.
| |
| ( IDNO: Ll' # Olki #
| |
| 387 IILO-512 5 PROCEDURE NUMBER: OTHER:
| |
| EOP-0001A LEVEL 3 EPSTG'0002
| |
| !295037 EA1.02_ ; 3.8;. _
| |
| ,4
| |
| !?95037 F A1.03 4.1' 4.1 _ _
| |
| COMMENTS: 7/97 new EOP 1 A, B 369 RO T1 G1 SRO T1 G1 O ,
| |
| | |
| EOP I A RPV Control- ATWS - RQA r ..
| |
| (
| |
| y.~~ .
| |
| .w--,_,_
| |
| . STEPS RQA-4, RQA-5 ~
| |
| Step RQA-4 represents the first of a series of steps designed to reduce reactor power to the lowest possible sta:e in a very controlled fashion. Briefly, steps RQA 4 through RQA 10 are designed to initiate a power reduction by reducing or securing reactor recirculation flow and directing boron injection if power is sufficiently high to warrant it. However, it takes into consideration the etTects on continued main turbine operation if the reactor recirc pumps were simply tripped from rated power conditions. Should this occur, a level " swell" of sufficient magnitude to trip the turbine on high level could possibly occur, causing additional problems from the lack of a sufficient heat sink for the power produced Steps RQA-4 and RQA-5 look at the effectiveness of the main turbine as a heat sink. If a steam flow path through open MSIVs is available (RQA-4) and the main turbine on-line (RQA-5), the remaining steps will methodically reduce power to the point where recire punips me. afely be tripped without causing a high water level turbine trip to occur. However, if both conditions are not satisfied in steps RQA 4 and ,
| |
| RQA-5, keeping the turbine on-line is not a concern In this case, the methodical steps to reduce ,
| |
| power are bypassed, since a turbine trip is not a concern.
| |
| (3 L)
| |
| EPSTG*0002 B - 122 Resision 3
| |
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| SRO EXAM KEY Q Esam Number NRC SRO Rey, 0- Exam Title NRC SRO EXAM
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| .25 - Which of the following is a sample point for the Containment Atmosphere Monitoring System?
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| : a. containment dome,
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| : b. RWCU pump room,
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| : c. reactor sampling sink area.
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| : d. CRD flow control station.
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| ANSWER:
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| : a. containment dome IDNO: LP# OBJ #
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| 418 11L0-014 2 PROCEDURE NUMBER: OTHER:
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| SOP-0084 LEVEt. 3
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| ~ . . . - - _ . . . . .
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| O !295032 EA1.01 1 3.6 3.7 COMMENTS: 7/97 new a n I
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| l
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| PAGE l OF 5 CONTROL BOARD LINEUP - CONTAINMENT ATMOSPHERE MONITORING SYSTEM PANEL ITEM PANEL ITEM POSITION INDICATION INITIALS ist 2nd TNE FOLLOWING ITEMS ARE LOCATED ON PANEL H13 P808 DIVISION I CONTMT ATMOS OFF N/A MON INOP SWITCH pAgy NORMAL AFTER STOP N/A ANALYZER CMS SOV33U, RWCU HEAT NONE DEPRESSED AFTER GREEN
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| . EXCHANGER ROOM OFF CMS-SOV33Y, DW NONE DEPRESSED AFTER GREEN PERSONNEL HATCH AREA OFF CMS SOV3LI SFC VALVE NONE DEPRESSED AFTER GREEN ROOM OFF CMS-SOV33AA, BENEATH MS NONE DEPRESSED AFTER GREEN TUNNEL FLOOR OFF CMS-SOV34A, DW DOME SELECT GREEN / WHITE AREA CMS-SOV33A, HVR FNI A NONE DEPRESSED AFTER GREEN AREA OFF ,
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| CMS SOV33W, ABOVE MAIN NONE DEPRESSED AFTER GREEN STEAM TUNNEL OFF H2 ANALYZER STANDBY N/A RED CMS-SOV33C, HVR-FNIC NONE DEPRESSED AFTER GREEN AREA OFF CMS-SOV33S, RWCU NONE DEPRESSED AFTER GREEN BACKWASH TANK ROOM OFF CMS-SOV34C, BELOW NONE DEPRESSED AFTER GREEN REFUEL SEAL AREA OFF CMS SOV31 A, ANALY2ER OPEN RED SPLY OUTBD ISOL CMS SOV35C, ANALYZER OPEN RED SPLY INBD ISOL YZER INLET CONTMT OR DRYWELL NONE LIT H2 ANALYZER AUTO AFTER OFF GREEN CMS-SOV33E, CONTAINMENT GREEN / WHITE SELECT DOME AREA CMS-SOV330, RWCU VALVE NONE DEPRESSED AFTER GREEN NEST ROOM OFF CMS-SOV31C, ANALYZER '
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| OPEN RED RTN OUTBD ISOL CMS-SOV35A, ANALYZER OPEN RED SOP-0084 REV-6 PAGE 31 OF 33
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| ATTACHMENT 4r PAGE 2 OF 5 CONTROL BOARD LINEUP. CONTAINMENT ATMOSPHERE MONITORING SYSTEM PANEL ITEM PANEL ITEM POSITION INDICATION INITIALS O RTN INBD ISOL lst 2nd H2 ANALYZER CONT RM NT
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| .g E T IS LIT H2 ANALYZER AUTO WHITE CMS-P7A, H2 SAMPLE PUMP A GREEN N/A CMS SOV310, DW NEUTRAL AFTER OPEN RED RADIATION SAMPLE RTN A OPEN RED R i O SAh PLE SUCT DIV 2 CONTMT ATMOS MON opp gfA INOP SWITCH Q
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| ANALYZER y NEUTRAL AFTER STOP N/A CMS SOV332, DW NONE DEPRESSED AFTER GREEN PERSONNEL HATCH AREA OFF CMS SOV33BB, BENEATH NONE DEPRESSED AFTER GREEN MAIN STEAM TUNNEL OFF '
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| A HT OOM hp CMS SOV33X, ABOVE MAIN NONE DEPRESSED AFTER GREEN STEAM TUNNEL OFF -
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| CMS-SOV34B, DW DOME SELECT GREEN / WHITE AREA CMS-SOV33B, lHVR FNID NONE DEPRESSED AFTER GREEN AREA opp CMS SOV33V, RWCU HEAT NONE DEPRESSED AFTER GREEN EXCHANGER ROOM OFF H2 ANALYZER STANDBY N/A RED CMS-SOV33D, HVR FNIC NONE DEPRESSED AFTER GREEN AREA OFF CMS-SOV33K SFC VALVE NONE DEPRESSED AFTER GREEN ROOM OFF CMS-SOV34D, BELOW NONE DEPRESSED AFTER GREEN REFUEL SEAL AREA OFF CMS-SOV31B, ANALYZER OPEN RED SPLY OUTBD ISOL CMS-SOV35D, ANALYZER
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| * OPEN RED
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| ; . SPLY INBD ISOL p H2 ANALYZERINLET CONTMT OR DRYWELL NONE LIT d SAMPLE ,
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| SOP-0084 REV-6 PAGE 32 OF 38
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| . - - = - - . - . - - - . .
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| ATTACHMENT 4A PAGE 3 OF 5 CONTROL BOARD LINEUP . CONTAINMENT ATMOSPHERE MONITORING SYSTEM PANEL ITEM PANEL ITEM POSITION INDICATION INITIALS O ist 2nd f H2 ANALYZER AUTO AFTER OFF GREEN l CMS SOV33F CONTAINMENT SELECT GREEN / WHITE DOME AREA CMS SOV33H, RWCU VALVE NONE DEPRESSED AFTER GREEN NEST ROOM OFF CMS SOV31D, ANALYZER RED OPEN RTN OUTBD ISOL CMS SOV35B, ANALYZER OPEN RED RTN INBD ISOL .
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| H2 ANALYZER CONT RM "
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| df!EL T IS LIT H2 ANALYZER AUTO WHITE CMS P7B, H2 SAMPLE PUMP N/A GREEN THE FOLLOWING ITEMS ARE LOCATED ON PANEL CMS PNL 12A, AB, EL 141', WEST ANALYZER ON/OFF ANALYZER ON NONE POWER ON N/A RED HEATER ON N/A N/A REMOTE SELECTOR NOT DEPRESSED N/A l
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| FUNCTION SELECTOR SAMPLE NONE
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| , MAN / AUTO SELECTOR AUTO MODE NONE STBY SELECTOR SW STBY RED SAMPLE PT SELECTOR SW DRYWELL OR NONE CONTAINMENT HIGH HYDROGEN N/A NONE COMMON ALARM N/A NONE ANALYZER CELL Fall N/A . NONE
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| ! LOW ANALYZER CELL FLOW N/A NONE I
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| LOW ANALYZER TEMP N/A NONE LOW GAS PRESS N/A NONE THE FOLLOWING ITEMS ARE LOCATED ON PANEL CMS-PNL12B, AB, EL 141', EAST ANALYZER ON/OFF ANALYZER ON NONE POWER ON N/A -
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| RED HEATER ON N/A N/A REMOTE SELECTOR NOT DEPRESSED N/A SOP-0084 REV-6 PAGE 33 OF 38 1 e
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| g Esam Number NRC SRO SRO EXAM KEY Rev. O Esam Title NRC SRO EXAM 26 During operation at 100% power with a rod line of 100%, the "A" Recirc Pump inadvertently trips to off. About 20 seconds later, the ''B" Recirc Pump trips to slow speed, resulting in the following steady state plant conditions:
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| 1hermal power 50%
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| Calculated core flow 34 %
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| What is the required operator action?
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| : a. Immediately SCRAM the reactor,
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| : b. Reduce thermal pow er to less than 40% by inserting control rods.
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| : c. Raise core flow by upshining the B recire pump to FAST.
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| : d. Raise core flow by starting the A recirc pump in SLOW.
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| ANSWER:
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| : b. Reduce thermal power to less than 40% by inserting control rods.
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| g, IDNO: LP# OBJ #
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| 44 I(1.0-534 15 PROCEDURE NUMBER: OTHER:
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| AOP-0024 LEVEL 3 i
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| ~ -" N'RC K Ai l' OI I SROi l
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| !295001 AA2.01 1 3.$i 3.8 __ . ]
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| COMMENTS: 7/97 new AOP.24 N!F Map RO T1 G2 SRO Tl 02 O ,
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| 9
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| l 5 SUBSEOUENT OPERATOR ACTIONS P
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| NOTE Perform steps 3.1 and S.2 concurrently.
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| 5.1 Refer to Attachment 1, or 2 to determine power / flow region.
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| N.QIE.
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| Entry into Region A is only allowed to protect thefuel and other vitalplant equipment, provided an IMMEDIA TE MANUAL SCRAMis executedfollowing the entry. Entry into Region B is only allowed durbgplant transients, provided the region is immediately exited.
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| Increasing recirculationflow by starting a recirculation pump or shiftingfrom low to high speed is NOIallowed while within Region B.
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| Process computer point B33NA0l Yshould be used to p determine coreflow when one Recirc pump is OFF.
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| O B33-R613 TOTAL CORE FLOW (Red Pen) may be inaccurate in this configuration and should nat be usect l
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| 5.1.1. Manually SCRAM the reactor if core flow is less than 33.8 x .
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| 106 lbm/hr (40%) AND rod line is greater than 100%.
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| (Attachment 1 or 2, Region A) 5.1.2. Take immediate action to exit Region B by INSERTING control rods using th: Shutdown Control Rod Sequence -
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| Package, or by raising recirculation flow by opening recire FCVs only.
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| O AOP-0024 REV - 14 PAGE 6 OF 14
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| s j o ATTACIlalEP1 i PAGEI i RBS SINGLE LOOP OPER ATION ?OWER/FLOF %1AP 1
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| RBS SINGLE LOOP OPERATION POWER / FLOW MAP
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| _ Maximum Rod-Line .; _ ;. .;
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| : q. . .,
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| _ l.
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| ;. .I i Normal 104.2% _; *
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| ; ,( _
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| __,__(_ f
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| * Maumum SW Com Thennel MELI.A 120.29% :
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| l ; ; Iwer = 2023 MW(th) po%)
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| l l
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| ' 15 + P-0.15 Wy l l l l l
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| ; l MELIA Repan 2000. _ RL - 100 _. __ __ _
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| ) ,
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| . . [100% RLl .,........
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| .....__7._
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| . ,. _,.__y___7._. ,. . g 4 , , ,__ _g. .. ., .g. , .,.. 3...
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| l p._
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| 1500.___p_ p __ y , _ ___ p __ .
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| , , _f_.
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| _Iug gtI . i t 3 _. ., ., . , . . , _ . . . . . _ . , . . .. ..
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| I E . . . J. . . . . . . . . .
| |
| . ,_ . , _ . , _ . ,_ _ _ ___., _;. _. 7_ .,.
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| < M , . . . . . . . . . . . .
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| tu .;. .. _. _ . _ _. . . . . ..
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| l ._ , . , _ _ . . . .. __ .
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| ; g . . . . . .
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| ; - ~;
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| ] J-- L o 1000 ;- ' - - - -- -
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| : n. ; ;
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| ses hwap cavitaama nesian h sw %%w
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| . . _ _ . ; l g g.j.;.
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| 3
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| . ._ .;.__,___y. . , ;. . ;.__i,
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| ; _ . ,t _
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| ____ _;. _;___;. . . ; . _ ; _ _ ; _ _ .:. __;_ ..;...___;__ _..;. .:. . _ 4 500 *
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| . .' . . . . A Scram
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| ._...__J.._&.
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| _L
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| .J_
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| . 'L .
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| .L
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| __.._f
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| .. .4
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| . . . . . . . . . . C Controlled Entry
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| . . . . . . . . . . . . . (FCBS 51. exit on transions )
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| . . : . . 4 . . . ; . _ ;. . . ___;__ ;. -
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| slo-neo.on u bound.,y tor sto i
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| -'-- '-- 1- ' -- -- '- *- - -- ^- - - - - > - -
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| n ! ! ! ! $ ! ! 5 0 10 20 30 40 50 FLOW (MibmIhr)
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| AOP-0024 REV - 14 PAGE 11 OF 14 l i
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| i
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| ATTACllMI-
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| * PAGEI RBS DUAL LOOP OPERATION POWER / FLOW MAP
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| +
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| RBS DUAL LOOP OPERATION POWER / FLOW MAP 2000 .
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| 1 . . . . . . . . . . . . . . . . . . . . l . . , . f . . . l . . . .
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| Maximum Rod-Line r -r *- -
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| r v '- * -
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| + mai 104.2*/i. * - * - - * -
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| MM %m .
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| 3100% RL
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| - > i - + *
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| * l
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| * MELIA 120.29 % - -
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| - 4. 4 -
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| (LATER) _*
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| t . <.t . . .. ..... . - _.. ... ... . . - . .. .. .
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| . 15 + p - 0.15 . w, . . . . . . . . . . .. . . . . . . . . . . .
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| ,...__a.
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| 2500.- p. RL - 100 45 + 0.55 Ws ___s _ a . 2_ _t a . i ..
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| t_ i _1._ _ t _a_ A . t . _, _ .
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| * * * * * ' ' =
| |
| : u. * * < * - * ' *
| |
| .=
| |
| 8
| |
| . .
| |
| * _ t. .
| |
| ___.__t t
| |
| l . . . . . . . . . . . . . . . . . . . . . .
| |
| ~i ~F E*i- i*1 . . . e . . . . . . . . . . . . . . . . . . . .
| |
| l . . . . . . . .
| |
| : 1. ,. . .
| |
| , , ... . - ^
| |
| , . , . 7. _ .g.. ,
| |
| 9 33 e
| |
| : 3. , p ., . . 3 - _q. . . .,. ".~r , '
| |
| ' $% Q . 1 , - 1' .* .-
| |
| , . r, . , , , ,-,-i- , , , - _ . - - - . - _ - ,4,- -
| |
| _..'.._..'_ .'_.'_4_
| |
| 2MC.._ ,*_
| |
| * __,'__,'_ .._ _', ' _ _ , ' _ '
| |
| l
| |
| ....,_. .... . 4 . . 4 4 . _. _= ._ 4 ._4 4 a _...s_ _v.. . . . . . . , . . . 4_ 4 i
| |
| i . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
| |
| , n . .. . . . . . . > . .... .. . ,_ a_> > .. . . . . . . . . . . .. . . . . . .. . . . .
| |
| 5 . . . .
| |
| .. . .1 J.
| |
| .1 . ..
| |
| J. _ J.
| |
| t.
| |
| t
| |
| . . i . . . .
| |
| .L . . . . . . . . .
| |
| I, 2 i . . . . . . . . . . . . . . . ~ . ' . . .~ 60% RL
| |
| ~
| |
| g 1500__'. '. .' '.. *
| |
| , m .
| |
| i .
| |
| i3-,_
| |
| , y . .
| |
| r . ., - .. , . . .
| |
| r
| |
| -, r ,..
| |
| i O , . , . , . . , , . , . . , . . . .
| |
| gL. . .,.. . . . , . , , . . _ . _ _ . . . _. . . -,-......
| |
| ' ., . ., . . -.-. . . 4 _ - . . . _ .-+ Dual loop operauon i .. . . . . . . . . . . . . .
| |
| 1000 4-.--
| |
| * _ - . - . m u_ _4. Cavatice Region i . . . . . . . . . . . j . . I l
| |
| .s.......
| |
| .t .L .
| |
| : 4. 8
| |
| *. L- ..
| |
| . ._s._n.._,
| |
| L _i l
| |
| , _ t. . . . .2.. . $. . J. . .. . _ i i i ;i ;;.-.- , .-.-.- -.- r r r- ; t i i-i
| |
| ., _7 . _ .,..,_ ,
| |
| ,._._3, 3
| |
| 500 r-- -- ,' v' ,' e' i . . . .-- . . r' - . . .
| |
| . . . B Exit
| |
| . . . . . . . . . . . . . . . . . . l CMW 4
| |
| . . . . . . . . (FCBB < 1. emit on transient)
| |
| ; .. . .. -.. ..... .a. .4. . . .. .m _t .._ ..-
| |
| * + ; SLC %--:--. Il boundary for SLO
| |
| ! .. .. ... _2 ., 2. ,_ _ . . __. _.._.o . .... ..4_ :
| |
| n . . . . . . . . . . . . . . . . . . .
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| 0 10 20 30 40 50 60 70 to 90 100 i
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| FLOW (Mibm/hr) l A OP-0024 REV - 14 PAGE 10 OF 14
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| | |
| SRO EXAM KEY (3 Essa Number NRC-SRO Rev. 0 Enam Title NRC SRO EXAhj 27 The plant was initially operating at 100% power. A transient occurred resulting in the following conditions:
| |
| . RPVlevelis 35 inches and stable Reactor power is 73% and stable _
| |
| - Total core flow is 51.$ E6 lbm/hr and stable -
| |
| The caue of this plant configuration was the receipt of a signal from the:
| |
| : a. EOC RPT logic.
| |
| : b. ATWS/ARIlogic.
| |
| : c. recirculation pump cavitation interlock circuitry,
| |
| : d. recirculation How control valve runback logic.
| |
| ANSWER:
| |
| : d. recirculation now control valve runback logic.
| |
| IDNO: LP# OILI #
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| 419 STM-053 2c PROCEDURE NUMBER: OTHER:
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| AOP@24 LEVEL 3 l202002 A2.01 1 3.4' 3.4 l f 29'3001 AK2.OI Y 3 52I 3I3 .
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| COMMENTS: 1/97 exam (modified values 6n stem & re-ordered answers)
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| ./
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| 27
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| 1.7 - - ABNORMAL OPERATION
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| ! l.7.1 - AOP-0024 Thermal Hydraulics Stability Controls This procedure was written to guide the operators in proper corrective action to be
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| : - taken following entry into any restricted operating region of the Power to Flow Map
| |
| - (figure 14). The restricted areas correspond to parameters where core instability has been observed or are predicted Specific directions are given depending upon what region of the Po ver to FMw Map the operation of the plent corrvsponds to. Since one of the axis e f the mg is core flow and the reactor recirculation system is the primary method of changing this parameter, the majority of the directions of this procedure are Gwcted at the reactor recirculation system. When conditions require entry into thls procedure, the main concem is to monitor and suppress thermal -
| |
| hydraulic instabilities which may occur.
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| 1.7.2 Flow Control Valve Runback !
| |
| Should reactor water level decrease to the Low Level Alarm setpoint, Level 4 (+
| |
| l n 30.8") and less than three feedwater pumps are running, the FCVs " runback" to the 60% drive flow position. The system does this to insert negative reactivity into the core in anticipation of a reactor level problem with insufficient feedwater flow. If the runback occurs from 100% power the resulting core flow is about 61% and
| |
| - reactor power about 73%.
| |
| The feedwater pumps are determined to be operating as sensed by greater than 3300 gpm flow at the suction of the pump. A Flow Control Valve Runback causes all flow controllers to automatically shift to the manual mode.
| |
| " NOTE: A RFP on minimum flow is NOT sufficient flow to prevent a runback.
| |
| A Flow Control Valve Runback also causes the FCV "A"("B") RUNBACK RFP TRIP annunciator (H13-PNLP680) to alarm. The Alarm Response Procedure (ARP) for these annunciators provides the operator cctions required to recover from this condition.
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| 1.7.3. FCV Motion Inhibit
| |
| 'Ihis interlock is initiated by any of the following conditions:
| |
| e HPU failure e Control circuit failure e High drywell pressure (1.68 psid)
| |
| O - n eu m is* teew r sse << s RBS-t-STM-GPST-A0053.01 PAGE 45 OF 69
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| ., , ,m y- g-
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| | |
| rw J
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| SRO EXAM KEY
| |
| - Esam Number NRC SRO - Rev. O Esam Title NRC SRO EXAM 2g A loss of condenser vacuum has occurred, vacuum is currently 18.5" lig. Which of the following automatic actions shouki have occurred?
| |
| a.. Turbine trip only,
| |
| : b. Turbin- trip and bypass valve closure.
| |
| : c. Turbine trip and MSIV isolation.
| |
| : d. Turbine trip, bypass valve closure and MSIV isolation.
| |
| ANSWER:
| |
| : a. Turbine trip only.
| |
| IDNO: LP# OBI #
| |
| 364 IILO-524 01 PROCEDURE NUMBER: OTHER:
| |
| AOP 0005 LEVEL 2 f .. _ . . - - - .
| |
| !295002 AK2.02 6 3.16 3.2 COMMENTS: 7/97 new ROT 102 SRO T1 G2
| |
| '%, 25
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| | |
| O 1 , PURPOSE / DISCUSSION 1.1 The purpose of this procedure is to provide instructions in the event of lowering Condenser vachum or trip of a Circulating Water Pump.
| |
| 1.2 A loss of Condenser vacuum can be gradual or rapid. If the condition can not be corrected quickly, a Turbine trip and Reactor scram will result. Eventually, MSIV closure and loss of Bypass Valve capability can occur. RCIC and Safety Relief Valve operation may then be required to maintain reactor pressure.
| |
| 2 SYMPTOMS 2.1 Lowering Condenser vacuum 2.2 Circulating Water Pumps or Cooling Tower Fans tripped 2.3 Loss of Gland Seal Steam pressure 2.4 Raising Offgas System flow, due to condenser air inleakage 2.5 Blockage in Offgas System 2.6 Circulating Water pipe rupture 2.7 Failure or isolation of SJAE 2.8 Reduction in generator load 3 AUTOMATIC ACTIONS
| |
| . Turbine Trip 22.3 in Hg e MSIV Closure 8.5 in Hg e Main Steam Bypass Valves Closure 8.5 in Hg O
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| AOP-0005 REV - 10 PAGE 3 OF 7
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| | |
| OG.
| |
| SRO EXAM KEY Esam Number NRC.SRO Rev. O Esam Title NRC SRO EXAM 29 The Plant was operating at 100% of rated power when a reactor scram occurred. Plant conditions are as follows:
| |
| - Reactor power is on range 2 of the IRMs and decreasing
| |
| - The Main Turbine is tripped
| |
| - The Main Generator Ouptut breakers are closed
| |
| - The Main Generator Exciter Field breaker is open
| |
| - Megawatt load on the Main Generator indicates -10 MWe (steady for 2 minutes)
| |
| - MVARs on the Main Generator indicate 50 MVARs leading (VARs in)
| |
| - Station loads are being supplied through the Preferr~ l Station Transformers
| |
| - Main Condenser Vacuum is 28"lig and steady Which one (1) of the following describes the action (s) required by AOP-0002, Main Turbine and Generator Trips?
| |
| : a. Immediately trip the Main Generator Output breakers.
| |
| : b. Immediately initiate a reverse power trip of the Main Generator Output breakers by decreasing generator VARs to zero,
| |
| : c. No action is required as long as Main Condenser Vacuum remains above 26" lig.
| |
| b' d. Within 5 minutes, initiate a reverse power trip of the Main Generator Output
| |
| (
| |
| breakers by decreasing generator VARs to zero.
| |
| ANSWER:
| |
| : a. Immediately trip the Main Generator Output breakers.
| |
| IDNO: LP# 05LI #
| |
| 381 IILO-521 8 PROCEDURE NUMBER: OTHER:
| |
| AOP-0002 Sect.5 Note LEVEL 3
| |
| !295005 AA1.G4 i 2.7i 2.8 COMMENTS: 7/97-new AOP 0002, p. 7 RO T1 G1 SRO T1 G2 AOP-0002 note on pago 7 require that the Main Generator Output breakers be immediately tripped if the Exciter Field Breaker has tripped, t'
| |
| ~$
| |
| 29
| |
| | |
| Q SRO EXAM KEY
| |
| -Q/
| |
| Esam Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM 30 A small-break LOta has occurred. Reactor level initially fell to -47 inches, then llPCS initiated -
| |
| and filled the reactor to a maximum of +55 inches, level is now steady at +40 inches. Which of the following describes the current status of E22 MOV F004, the liPCS injection isolation valve?
| |
| : a. F004 will open en a liigh Drywcli Pressure initiation signal even if the llPCS 111G11 WATER LEVEL 8 RESET pushhutton has not been depressed and the 11PCS 111G11 WATER LEVEL 8 RESET pushbutton must be depressed before the valve can be opened manually.
| |
| : b. F004 can be opened manually even if the llPCS lilGil WATER LEVEL 8 RESET pushhutton has not been depressed and the llPCS 111G11 WATER LEVEL 8 RbSET pushhutton must be depressed before the valve will open on a liigh Drywell initiation signal,
| |
| : c. F004 can NOT be opened manually and it will NOT open on a liigh Drywell Pressure initiation signal until the llPCS 111G11 WATER LEVEL 8 RESET pushhutton is depressed.
| |
| : d. FON can be opened manually after the llPCS INITIATION RESET pushbutton is depressed.
| |
| ANSWER:
| |
| O) q - c. F004 can NOT be opened manually and it will NOT open on a liigh Drywell Pressure initiation signal until the llPCS 111G11 WATER LEVEL. 8 RESET pushhutton is deressed, IDNO: LP# OILI #
| |
| 327 IILO-019 4 PROCEDURE NUMBER: OTHER:
| |
| SOP 0030 LEVEL 3 1295008 AK2.07 2.9 3_ _j COMMENTS: 7/97 new
| |
| '. SOP-0030, Rev 8, page 2 of 34,2.5 LOTM 3 4, TaNo 8, p. 26 of 31 ROT 1G2 SRO Tt G2 30
| |
| | |
| 1 PURPOSE 1.1 To provide instructions for operation of the High Pressure Core Spray System.
| |
| 2 PRECAUTIONS AND LIMITATIONS 2.1 To avoid impeller wear, minimize the time the HPCS Pump is operated on minimum flow with E22-F004, HPCS INJECT ISOL VALVE closed.
| |
| 2.2 Starting the HPCS Pump with an injection line low pressure condition may cause damage due to water hammer. If the HPCS Line Fill Pump is to be shutdown for an extended period, me HPCS Pump Motor breaker should be racked out.
| |
| C 2.3 When performing tests or system lineups, especially if components are operated in a random order, the potential exists for water hammer to occur if the system has not been properly filled and vented.
| |
| 2.4 Technical Specifications 3.5.1 and 3.5.2 contain HPCS operability requirenients.
| |
| 2.5 If E22-F004, HPCS INJECT ISOL VALVE is closed using its handswitch with a HPCS initiation signal present, it will not reopen automatically. The reinstatement of the auto open function requires depressing E22A-S7, HPCS INITIATION RESET Pushbutton after the reactor low level and drywell high pressure signals have cleared.
| |
| 2.6 An automatic closure of E22 F004, HPCS INJECT ISOL VALVE on reactor high water level will inhibit opening of the valve except on a subsequent level 2 initiation signal. To allow operation of the valve on dr/well high pressure or manual operation, E22A-S6, HPCS HIGH WATER LEVEL 8 RESET Pushbutton must be depressed after level has decreased below the high level point.
| |
| 2.7 Placing E22-F015, HPCS PUMP SL? PL SUCTION VALVE Switch to CLOSE will override all open signals from $e CST low level or Suppression Pool high level signals. Reinstatement of the automatic function requires both level signals to be clear.
| |
| O SOP-0030 REV - 16 PAGE 3 OF 31
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| | |
| SRO EXAM KEY (3 Exam Number NRC-SRO Rev. 0 Exam Title NRC SRO EXAM 31 With the reactor at 100% power, a loss of all Reactor Plant Component Cooling Water occurs.
| |
| What are the required oper.. tor actions?
| |
| : a. Monitor and reduce system heat loads as necessary to, sntinue plant operations.
| |
| : b. Insert a reactor scram and shift both recirculation pumps to slow speed.
| |
| : c. Commence a reactor shutdown per GOP-0002, Plant Shutdown.
| |
| : d. Insert a reactor scram and trip and isolate both recirculation pumps.
| |
| ANSWER:
| |
| : d. Insert a reactor scram and trip and isolate both recirculation pumps.
| |
| IDNO: LP# Olks :, .
| |
| 241 IILO-530 6 PROCEDURE NUMBER: OTHER:
| |
| AOP-0011 LEVEL 2
| |
| [295018 AK2 02_! _3.4j._3.6
| |
| . l i29501e AA102 i 3.3' 3.4 _]
| |
| COMMEN TS: 7/97 new ROT 1G2 500 T1 G2
| |
| ^(
| |
| | |
| -4 B1 MEDIATE OPERATOR ACTIONfz 4.1 E a total loss of CCP occurs, THEN perform the following:
| |
| 4.1.1. Manually scram the Reactoi, E .,
| |
| 4.1.2. Trip and isolate both Recire Pumps.
| |
| 5 Sli&SEOUENT OPERATOR ACTIONS i
| |
| 5.1 E the following conditions exist, THEN place the Mode Switch in SHUTDOWN:
| |
| l
| |
| * Reactor Steam Dome pressute less than 600 psig Q . No CRD Pump is running b
| |
| m e CRD Accumulator associated with a withdrawn control rod is inoperable.
| |
| 5.2 E the following conditions exist, THEN place the Mode Switch in SHUTDOWN:
| |
| . Reactor Steam Dome pressure greater than or equal to 600 psig M
| |
| * CRD Charging Water Header pressure less than 1520 psig M
| |
| + More than ore CRD Accumulator is inoperable E .
| |
| CRD Charging Water Header pressure can nR1 be restored and maintained within 20 minutes AQf-0011 REV-9 PAGE'S OF 7
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| | |
| - - - - -. .. . - . = ... .
| |
| g ~ Esam Number NRC-SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 32 .With the Division 1 Diesel rear air start compressor motor seized. Which of the following actions are required?
| |
| : a. Declare Division 1 DieselINOPERABLE. Both starting air compressors are required for an OPERABLE air start system.
| |
| : b. Take no action since only one staning air system is necessary to start the diesel and crosstying the unaffected air system with the affected air system would render the diesel INOPERABLE.
| |
| : c. Maintain pressure in the normal bar d of the reciever associated with the seized compressor by intermittently using a high pessure hose connected between the operable forward and rear system air dryer outlets. The Division i Diesel Generator will remain OPERABLE.
| |
| : d. Start and load the Division i Diesel Generator. With the diesel running. the starting air system is not required for OPERABILITY of the diesel.
| |
| ANSWEh:
| |
| : c. Maintain pressure in the normal band of the reciever associated with the seized compressor by intermittently using a high pressure hose connected between the A operable forward and rear system air dryer outlets. The Division i Diesel Generator IDNO: LP# OBJ #
| |
| $4 IlLO-037 9 PROCEDURE NUMBER: OTHER:
| |
| SOP-0053 LEVEL 4 TS 1.1 I ~NRC K5 l~'RIE l~SfiO[ l
| |
| [2 64000_ K 1.06_ _. 3.2 p , 3.2
| |
| ;G 2.1.33 =
| |
| 3.4 4 ,
| |
| COMMENTS: 7/97 new (must understand TS/TRM 1.1 defirwtion of OPERABLE / OPERABILITY and know that the compressors are not safety-related
| |
| \ 32
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| | |
| 5.7 Cross Connecang Divison I and II Air Receivers
| |
| /]
| |
| NOIE Tlw open ble dr compessor(s) sividd be opented mandly when presstoi:ing two -
| |
| cirreceiverinitu. Contirneta monitoring of the hose cordittom is recommerded touil they cre isolaed ad removed NOIE Controls cre loc?Ned d 1EGS*PNL3A, or iEGS*PNL3B.
| |
| 5.7.1 Place the control switch for the operable compressor to OFF.
| |
| 5.7.2 Install one end of an air hose to the operable air receiver at the AIR DRYER OUILET HEADER QUICK DISCONNECT ISOL VALVE.
| |
| For Div I Rear IEGA-V3009 / Fonwd IEGA-V3010 For Div 11 Rear lEGA-V30ll / Fonvard IEGA-V3012 5.7.3 Attach the other end of the hose to the Inoperable Air Receiver at the AIR O- DRYER OUllET HEADER QUICK DISCONNECT ISOL VALVE.
| |
| For Div I Rear IEGA-V3009 / Fonvard IEGA-V3010.
| |
| For DIV 11 Rear IEGA-V30ll / Fonvard IEGA V3012.
| |
| 5.7.4 gi the selected dryer outlet connection valves (from Steps 5.7.2 and NOIE The opentorshould remdn in the vicinity of the hose ad compressor totil tin evolution is complete.
| |
| 5.7.5 Operate the air compressor until air receiver train is at the desired prescure.
| |
| 5.7.6 Place the control switch for the operable compressor to OFF.
| |
| 5.7.7 At the israble air receiver close the AIR DRYER OUllET LEADER QUICK DISCONNECT ISOL VALVES valve opened in Step 5.7.2, and 5.7.3.
| |
| For Div i Rear IEGA V3009 / Fonvard IEGA-V3010 b
| |
| v For Div 11 Rear lEGA-V30ll / Forward IEGA-V3012 I
| |
| SOP-0053 REV - 16B PAM 29 OF 88
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| | |
| SRO EXAM KEY Esam Number NRC SRO Rev. O Esam Title NRC SRO IIXAM 33 Tollowing a complete loss of shutdown cooling, temperature readings indicate a i degree F iner.ase in bulk w ater temperature every 10 minutes. Assume the reactor vessel head is on, no other parameters change, and current temperature is 124 deg. F.
| |
| Which of the following is the minimum amt at of time before primary containment MUST be established?
| |
| : a. 160 minutes
| |
| : b. $60 minutes
| |
| : c. 580 minutes
| |
| : d. - 760 minutes ANSWER:
| |
| : d. 760 minutes ll)NO: LP# OHJ#
| |
| 369 11L0-013 9 PROCEDURE NUMBER: OTHER:
| |
| TS 3 61.2 LEVEL 2
| |
| ~
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| i NRC KAt l RO: l $ROi ]
| |
| :296021 AA2 01 < 36' 3.6 .._ !
| |
| COMMENTS 7/97 new T.S. 3.6.1.2 and TatAo 1.2, Operational Conditions ROT 1G2 SHOT 1G2 d
| |
| 33 .
| |
| r-,--- ----w' n- ,ar,,, ,,,r~, ,-,,,-www,r_N,,,w-- ,.,--,-,---,-,---e,--,-n-, - , , - , - n ,,,--mm-, .,+n, , .,--,,,,.-,w-nmn,,n-m,-n,
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| | |
| Dofinitions TR 1.1 Table 1.1-1 (page 1 of 11 O MODES MODE TITLE REACTOR MODE AVERAGE REACTOR SWITCH POSITION COOLANT TEMPERATURE
| |
| ('r) 1 Power Operation Run NA 2 Startup Refuel (a) or Startup/ Hot NA Standby 3 Hot shutdown (s) Shutdown > 200 4 Cold Shutdown (a) Shutdown s 200 5 Refueling (b) Shutdown or Refuel NA
| |
| , (a) All reactor vessel head closure bolts fully tensioned.
| |
| (b) One or more reactor vessel head closure bolts less than fully tensioned.
| |
| O O
| |
| RIVER BEND TR 1-11 Revision 5
| |
| | |
| 1.
| |
| ?
| |
| SRO EXAM KEY i Esam Number NRC.SRO Rev 0 Esam Title NRC SRO liXAM 1 34 A plant starup is in progress. Reactor pow er is on IRM Range 7 and reactor pressure is 450 psig w hen the "A" CRD pump trips. %e *B" CRD pump will not start.
| |
| A reactor scram is required if;
| |
| : s. A control rod receives an llCU accumulator fault and cannot be inserted.
| |
| : b. More than one CRD high temperature alarm is received
| |
| : c. No CRD pumps can be restarted within $ minutes.
| |
| : d. Two or hiore accumulator faults exist.
| |
| ANSWERt
| |
| : a. A control rod receives an llCU accumulator fault and cannot be inserted. i IDNOt LP# OILI #
| |
| 368 STM 052 11 PROCEDURE NUMBER: OTHER:
| |
| AOP 0011 LEVEL 4 TS 3.1.5 d ARP 60122 AO1
| |
| ~
| |
| 'NRC K Ai'~ 'I'~O['l R SRO:~ l 296022 AK3 01 3.7i 39 _}
| |
| COMMENTS: 7/97 new ROT 102 SROT1G2 i
| |
| NOTEDifficult question, need to pull together different references.
| |
| | |
| g 4. Settle Function kh The settle function is provided in order to prevent a hydraulic lock from occurring, to allow the drive to settle in the latched position, and to provide a flow path for drive water to be exhausted ouring rod settling. The settle function acts to latch the rod in either rod direction. For insenion, drive is insened a little past the desired notch position, and then the insert signal and drive water is removed. If the rod is being withdrawn, the withdrawal signal is removed before the desired notch is reached. As the control rod settles downward, water below the drive piston is exhausted out the insert port. This allows the drive to drill down to the intended notch and become latched. During titis time, the pressure above the drive piston decreases momentarily and then return to normal due to seal leakage.
| |
| 1.5.1.5 Control Rods Control rods are withdrawn from the core to achieve reactor criticality by removal of negative reactivity from the core. The withdrawal sequence is pre planned to approach the estimated point of criticality while minimizing neutron flux peaking in the core.
| |
| During reactor power operation, the control rods are positioned to flatten power in the core.
| |
| Large power transients, generally a change of 25% or greater, require manipulation of the control rods to maintain reactor criticality and proper neutron flux distribution in the core.
| |
| (~ Over core life, the desired power level is maintained by modifying control rod positions to compensate for depletion of fuel.
| |
| The control rods are inserted into the core at normal rod drive speed to effectively shut down the reactor by inserting negative reactivity into the core.
| |
| During an abnormal or casualty situation, the rods are rapidly inserted into the core via a manual or automatic scram. This provides for a rapid insertion of negative reactivity into the core, shutting down the reactor in an expeditious manner.
| |
| 1.5.2 Abnormal Operation 1.5.2.1 Control Rod Drive System if a CRD Pump trips the operator must start the standby pump. If neither pump can be staned the plant procedures require that the plant be shutdown while there is adequate energy in the'llCU's / reactor vessel to ensure that all withdrawn control rods can be fully insened.
| |
| See ARP 60122(A01) for details.
| |
| RBS ISTM GPST-A0052.00 PAGE 46 OF 65
| |
| | |
| e O
| |
| Q 4 IMMEDIA1E OPERATOR ACTIONF 4.1 E a total loss of CCP occurs, THEN perform the following:
| |
| 4.1.1. Manually scram the Reactor.
| |
| t M ..
| |
| 4.1.2. Trip and isolate both Recirc Pumps.
| |
| 5 SUBSEOUENT OPERATOR ACTIONS 5.1 E the following conditions exist, THEN place the Mode Switch in SHUTDOWN:
| |
| * Reactor Steam Dome pressure less than 600 psig M
| |
| i e No CRD Pump is running m
| |
| e CRD Accumulator associated with a withdrawn control rod is inoperable.
| |
| 5.2 E the following conditions exist, THEN place the Mode Switch in SHUTDOWN:
| |
| * Reactor Steam Dome pressure greater than or equal to 600 psig M
| |
| e CRD Charging Water Header pressure less than 1520 psig M
| |
| e More than one CRD Accumulator is inoperable .
| |
| M CRD Charging Water Header pressure can nat be restored and
| |
| , maintained within 20 minutes
| |
| ""-' "^ " "
| |
| _ _ ^ "!i )
| |
| | |
| I
| |
| - CRD PUMP Q# ) ALARM NO. 2120 A OR B lill3*P601 / 22A / A01 g
| |
| AUTO TRIP INITIATING DEVICES SET POINTS
| |
| : 1. ICIl PSN001 A&B 1. 25"lig abs 2"
| |
| : 2. IRDS PS3A&B 2. 3 psig
| |
| : 3. Device 86 3. N/A
| |
| : 4. Device 94A&94B 4. N/A
| |
| : 5. CCP PSI A thru lli 5. 56 psig AUTOMATIC ACTIONS NOTE With no CRD pumps running, the Reactor Retire pump sealpurge supply is lost.
| |
| : 1. CRD Pump Cll C001 A or C001B trips.
| |
| OPERATOR ACTIONS
| |
| : 1. Start the standby CRD Pump, at panel 1fil3*P601 as follows:
| |
| 9 Start standby CRD Pump Oil Cl1-C001 AP(BP).
| |
| Place Flow Controller 1Cl1 F002 to MANUAL and close.
| |
| : c. Verify White Control Power Light on for CRD Pump to be started.
| |
| : d. Start Standby Pump Cl1 C001 A(B) CRD PUMP. j
| |
| : e. Verify amps return to normal (< 45 amps).
| |
| : f. Slowly open Flow Controller Cl1-F002 to achieve 45 gpm, as indicated on Cl1 R606 COOLING WATER FLOW.
| |
| : g. Place Cl1-F002 Flow Controller to AUTO.
| |
| : 2. If neither CRD Pump can be restarted:
| |
| : a. if reactor pressure <600 psig, and one or mc,re control rod accumulator faults exist for withdrawn control rod (s) which cannot be inserted, immediatelv place the reactor mode switch to SliUTDOWN per LCO 3.1.5 CONDil10N D.
| |
| : b. If reactor pressure is 2 600 psig, and two (2) or more accumulator faults exist, restore charging water header pressure >l 520 psig within 20 minutes or fully insert the rods associated with the accumulator alarms. If unable to fully insert affected rods, immediately place the reactor mode switch in the SliUTDOWN position per LCO 3.1.5 CONDITION D. .
| |
| c~.
| |
| LJ ARP-601-22 REV - 7A PAGE 3 OF 32
| |
| | |
| Control Rod Scraa Accu:ulators 3.1.5
| |
| () ACTIONS (continued)
| |
| , CONDITION REQUIRED ACTION COMPLETION TIME B. Two or more control B.1 Restore charging 20 minutes from rod scram accumulators water header pressure discovery of inoperable with to at 1520 psig. Condition B reactor steam dome concurrent with pressure a: 600 psig. charging water header pressure
| |
| < 1520 psig 8!iQ B.2.1 -------NOTE -- --
| |
| Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance.
| |
| I Declare the I hour associated control rod scram time
| |
| " slow."
| |
| QB B.2.2 Declare the 1 hour associated control rod inoperable.
| |
| C. One or more control C.1 Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water reactor steam dome accumulators are header pressure pressure < 600 psig, fully inserted. < 1520 psig 811Q (continued) 4 O
| |
| RIVER BEND 3.1 16 Amendment No. 81 4
| |
| | |
| Control Rod Scram Accumulators 3.1.5 O ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME L
| |
| C. (continued) C.2 Declare the I hour !
| |
| associated control rod inoperable.
| |
| J 3
| |
| D. Required Action and D.1 --- NOTE- -- --
| |
| associated Completion Not applicable if all Time of Required inoperable control Action B.1 or C.1 not rod scram met, &ccumulators are associated with fully inserted control rods.
| |
| Place the reactor immediately mode switch in the shutdown position.
| |
| ; SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is a: 1520 psig.
| |
| 9 O
| |
| RIVER BEND 3.1-17 Amendment No. 81
| |
| | |
| I SRO EXAM KEY Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM i
| |
| 33 Which one of the following describes a cause and the expected inaccurate response of reactor level ,
| |
| ' instrumentation indications when in the UNSAFE region of the RPV Saturation Curve?
| |
| : s. liigh containment temperatures will result in boiling of the reference legs causing an enoncously high level indication,
| |
| : b. liigh reactor pressure will result in boiling of the reference legs causing an erroneously low levelindication.
| |
| : c. Low reactor pressure u ill result in boiling of the reference legs causing an erroncously low levelindication.
| |
| : d. liigh drywell temperatures will result in boiling of the variable legs causing an
| |
| ! erroneously low levelindication.
| |
| ANSWER:
| |
| : a. liigh containment temperatures will result in boiling of the reference legs causing an erroneously high level indication.
| |
| IDNO: LP# OILI #
| |
| 81 IILO-512 5 PROCEDURE NUMBER: OTHER:
| |
| EOP 0001 LEVEL 4 EPSTG'0002
| |
| [ NRC K A' I 'RO: ~l SRhi
| |
| '296028 ( K 1.01 +
| |
| 3 5' 3.7 COMMENTS: 7/91 new ROT 102 SRO T1 G2 m, - + -- - - - - ,-- .--ee_,-----,.-_-.--c- ee,.., ,
| |
| | |
| e EOP Usage and Cautions o ::
| |
| Under extreme conditions, a high and increasing drywell or containment temperature can decrease the density of water in the reference leg vertical mns such that the instrument falsely indicates an on scale and steadily increasing RPV water level even though the actual RPV water level may be below the elevation of the instrument variable leg tap. This phenomenon could lead the operator to a false belief that water level trend is increasing, when in fact it may be decreasing.
| |
| .This could lead to inadvertent core uncovery and core damage.
| |
| RPV Water LevelInstrument Calibration Data INSTRUhtENT RPV DW CThtT JET PUhtP OTHER PRESSURE TENtP TEh1P FLOW 0
| |
| Narrow Range 1025 psig 135 F 80"F NA 21.5 in.
| |
| O to 60 in. accuracy Wide Range 1025 psig 135"F 80'F NO 16 in,
| |
| -160 to 60 in accuracy Fuel Zone 0 psig 21.2"F 80 F NO + 6 in.
| |
| 310 to 110 in. . accuracy Upset Range 1025 psig 135*F 80 F NA NA -
| |
| O to 180 in.
| |
| Shutdown Range 0 psig 100 F 80 F NA 120*F 0 to 400 in. water temp Caution 41 defines the bounding conditions under which the respective water level instruments may be relied upon to determine water level. The caution it divided into two separate and distinct parts, both of which must be satisfied in order to use the instruments to determine water level and trends. It should be noted that the.information in this caution is not simply an accommodation for inaccuracies in RPV water level indication which occur when plant conditions are ditTerent from the instrument calibration conditions.
| |
| Pan 1 of Caution #1 identifies the limiting conditions beyond which the water in the instrument legs may boil. Water in the RPV water level instrument legs is maintained in a liquid state by the cooling action of the surrounding atmosphere and the pressure in the reactor vessel.
| |
| The water in the instrument legs will boil, however, ifits temperature exceeds the saturation temperature for the existing RPV pressure.
| |
| Boiling is a concem in both horizontal and vertical reference and variable instmment leg runs Boil otT from the reference leg reduces the reference leg head of water. This decreases the sensed AP, resulting in an erroneously high indicated RPV water level. Boilig in the instrument variable leg exerts increased pressure on the variable leg side of the differential pressure ~
| |
| transmitter (Figure B 2). This results in a lower sensed @ and an erroneously high indicated RPV water level.
| |
| O EPSTG'0002 8 - 24 Revision 3
| |
| | |
| EOP Usage and Cautions NN. ORWCLL
| |
| ~
| |
| - CRYWELL #ALL p f RPV
| |
| -w
| |
| /
| |
| +
| |
| N P
| |
| /O
| |
| [ '
| |
| ;j 0 ch Ys on 0 . I REFERENCE C oo i7
| |
| - LEG -
| |
| oc
| |
| \ o'o :
| |
| -VARI ABLE
| |
| * ce <
| |
| LEG o 4
| |
| ( ,
| |
| 7 0 0
| |
| do 8a oo, o, Q Ceil o o n
| |
| Pi_ N_
| |
| "'l TEMPERATURE CAUSING B0(UNG j
| |
| NO 90 fung: N0 FLOW AND P1 - PO + PElGHT OF INSTRUWENT LEG S01UNG: R.0W AND P1 > PO + HEIGHT OF iNSTRUWENT LEG Figure B-2
| |
| ) Boiling in RPV Water Level Instrument Variable Leg EPSTG'0002 B 25 Revision 3
| |
| | |
| i i
| |
| SRO EXAM KEY O Esam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM 36 Suppression pool level is of fscale high. t i'
| |
| Which one of the followiny, describes the efTect on indicated containment or drywcli pressure?
| |
| : a. Indicated containment pressure is less than actual. >
| |
| i
| |
| : b. Indicated containment pressure is greater than actaal. l 2 :
| |
| : c. Indicated drywell pressure is less than actual,
| |
| : d. Indicated drywell pressure is greater than actual.
| |
| 6 ANSWER:
| |
| : b. Containment pressure is greater than actual, r
| |
| IDNO: LP# OILI #
| |
| 388 IILO 514 $
| |
| PROCEDURE NUMBER: OTHER:
| |
| EOP-0001 LEVEL 2 O EPSTG'0002 Cauten 8
| |
| . ~ . . . . . - - _ . . ..
| |
| :296029 E K1.01 1 34 3.7 COMMENTS: 7/97 new ROT 1G2 SRO T102 0 ,,
| |
| | |
| EOP Usage and Cautions o ::
| |
| CAUTION #8 Caution u8 is necessary to ensure an accurate containment pressure is available under all plant conditions The pressure taps for containment pressure instruments are located 56 feet from the bottom of the suppression pool. If suppression poollevelis off scale high, the containment pressuie instrument taps are assumed to be covered with water. In this situation, containment pressure indications would be a combination of air space pressure plus the hydrostatic head of water abose the instrument tap. This situation would lead to falsely high containment pressure readings The operator is directed to determine containment pressure and suppression pool water lesel as directed by an EOP enclosure This.will accurately reflect containment pressure and
| |
| * suppression pool level using indirect measurements from other systems. ,
| |
| O O
| |
| EPSTG'0002 B 36 Revision 3
| |
| | |
| l SRO EXAM KEY l
| |
| \
| |
| Esem Number NRC SRO Hev. O Esam Title NRC SRO EXAM i
| |
| : 37. Given the following conditions:
| |
| . A failure to scram has occurred.
| |
| . Reactor power is 20% with control rods being inserted manually.
| |
| EOP 3,'' Secondary Containment Control' has been entered due to ilVAC cooler high differential temperatures caused by a fire in the Auxiliary Building. t
| |
| . MSIVs have closed, ,
| |
| . Condensate /feedwater is maintaining water level. t Which of the following systems should be isolated?
| |
| ?
| |
| : a. Feedwater
| |
| : b. Reactor Water Cleanup
| |
| : c. Control Rod Drive i
| |
| : d. Fire Suppression systems ANSWEN:
| |
| : b. Reactor Water Cleanup
| |
| , O ,
| |
| IDNO: LP# Olli #
| |
| 376 ilLO.$15 4 PROCEDURE NUMBER: OTHER: l EOPM3 SC-12 LEVEL 3
| |
| :296032 i AS.05 4 3.7: 3.9 COMMENTS 1/97 new HLO-614 obj 4 & 6 37 f
| |
| 1
| |
| --.y - ,-- %,m.. m, ._.-r-- - , - . - - - , - . . . re.- - , , , , , ,- ,# .- , . . , . , - ,, . , , , , - , , , , . . _ - . . - . . . - . - . - . , - , - - - , . -
| |
| | |
| EOP.3 Secondary Containment Control- SC q -
| |
| O STEP SC-12 Step SC 12 is common to all three sections of Secondary Containment Control.
| |
| Additionally, the actions taken from this point in the procedure are generic and common to all sections. Therefore, all three sectionsjoin at this step and operator actions are directed to mitigate the specific problem.
| |
| Should the operator be directed to this step, it means that actions to control either area temperatures, water levels, or radiation levels have been inetTective. At this point, t'urther actions must be taken in order to assure continued equipment operability, personnel access, and plant safety. .
| |
| The actions of this step are designed to terminate any leak or water addition into the .
| |
| containment by isolating all systems discharging into the area, in order to prevent the problem from escalating.
| |
| b V However, systems being used to assure adequate core cooling, shut down the reactor, to suppress a tire, or protect primary containment integrity at: not to be isolated. Preventing core damage, maintaining primary containmem integrity and ensuring personnel safety have priority over secondary containment concerns.
| |
| O
| |
| 'V EPSTG'0002 B - 222 Revision 3
| |
| | |
| ; SRO EXAM KEY '
| |
| Esam Numt.er NRC.SRO Rev. 0 Esam Title NRC SRO EXAM ,
| |
| 3g Isolation of a primary system leak is required by EOP.3, Secondary Contatnment and Radioactive ;
| |
| Release Control, in order to limit radioactive discharge.
| |
| l 4
| |
| 11y definition, the term ''Prirnary System" refers to any system:
| |
| i
| |
| : a. for which the ASME *N" stamp is issued.
| |
| I
| |
| : b. required to shutdown and cooldown the reactor. :
| |
| 1
| |
| : c. required to maintain Primary Containment Integrity. j
| |
| : d. connected directly to the RPV that has a reduced leak rate if RPV pressure is lowered.
| |
| ANSWER:
| |
| : d. connected directly to the RPV that has a reduced leak rate if RPV pressure is towered.
| |
| I
| |
| < ll)NO: LP# OBJ #
| |
| i 227 llLO 515 6 1 PROCEDURE NUMBER: OTHER:
| |
| EOP 0003 RR.2 LEVEL 2
| |
| ' NMC K5i 'l Ad$ l~SMOI
| |
| '295033 EK3 03 i 3.8' 3.9 4
| |
| COMMENTS: 7/97 00 W l l
| |
| i 38
| |
| | |
| EOP.3 Itadioactive Release Control RR STEP RR 2 -
| |
| m .
| |
| Primary systems comprise the pipes, valves and other equipment connected directly to the RPV such that a reduction in RPV pressure will result in a decrease in the flow of steam or water through an unisolated break in the system. Isolating primary systems that are discharging into areas outside the primary and secondary containments povides the most direct and appropriate action for terminating offsite radioactivity releases isolation of those primary systems required to assure adequate core cooling or shut down the reactor is not appropriate. Continued operation at power or a failure to cool the core may cause significant core dainage This could ultimately result in much larger release rates.
| |
| O 4
| |
| n v
| |
| EPSTO'0002 B 232 Revision 3
| |
| | |
| EOP Usage and Cautions O- Monitor ;
| |
| Observe or evaluate at a frequency sutlicient to remain apprised of the yalues, trends, and rate of change of the identified plant operating parameter. Plant conditions, the rate of change of !
| |
| the parameter, and its proximity to specific limits will dictate the required frequency of observation Present Take the necessary steps to preclude a specific action. This includes closing valves, tripping pumps, jumpering (or opening) contacts in the control logic of system components, de-energizing equipment, and overriding automatic signals in order to accomplish this action.
| |
| Primary Containment The airtight space immediately adjacent to and surrounding the RPV. For RBS, the containment.
| |
| l ,
| |
| Primary System '
| |
| The pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will decrease the steam or water being discharged through an unisolated break in the system Purge Force tiow through an enclosed volume. This includes establishing both an influent (driving) and effluent (exhaust) flowpath similar to that of a " feed and bleed" process.
| |
| Restore Take the appropriate action required to return the value of an identified parameter within applicable limits.
| |
| Secondary Containment The airtight spaces immediately adjacent to or surrounding the primary containment. For RBS. the Auxiliary Building Fuel Building, and the Shield Building (Annulus).
| |
| r Shutdown As it applies to the reactor, the condit' ion of having all'available SRM's inserted with steadily decreasing countrates (Ref. RXE 92 032, dated 7/9/92).
| |
| EPSTG'0002 B 19 Revision 3
| |
| | |
| !t SRO EXAM , LEY ;
| |
| Enem Neuber NRC.SRO Rev. 0 Esasu Title H.'jiGliNf M !
| |
| t 39- Whkh of the following unisolable system fallutes eam nutsde of pt0 m.7 (Mwvun n % [
| |
| constitute a piimary system for purposes of EDIV nudu) CmterenKe 2 ;
| |
| : a. Main steam drain lines in the Main Steam Line iwm! ;
| |
| i
| |
| : b. Containment vent line to drywell !
| |
| : c. Senice Water supply to drywell coolers f
| |
| : d. LPCS suction line from the suppression f ANSWER:
| |
| : a. Main steam drain lines in the Main ' ca ulne Tunnel i
| |
| 1 t-ll)NO: 1.P W OBJ # .
| |
| 312 111,0 51 $
| |
| E PROCEDURE NUMBER: OTHER:
| |
| EOP.003 LEVEL 2 ,
| |
| t .'T0'0002 I NRMt l 'Mu ')
| |
| g5033 f k3 61 ! 3 (
| |
| COMMENTS: 7/97 hDW ROT 102 SRO T102 t
| |
| -39 t
| |
| --,.-....,-.--,n.,----.-,.--,n - ,ww.. ,,n-n---,,,-.,.,,.m,,,..- , , , - , - - - - , - . , >~-,,,n.,-- ,.,--mn,,s .
| |
| | |
| l
| |
| - EOP.3 Secondary Containment Control . SC STEP SC.12 .
| |
| Step SC 12 is common to all three sections of Secondary Containment Control.
| |
| Additionally, the actions taken from this point in the procedure are generic and common to all sections Therefore, all three sectionsjoin at this step and operator actions are directed to mitigate the specific problem.
| |
| Should the operator be directed to this step, it means that actions to control either area temperatures, water levels, or radiation levels have been ineffective. At this point, further actions must be taken in order to assure continued equipment operability, personnel access, and plant safety. ,
| |
| e The actions of this step are designed to terminate any leak or water addition into the ',
| |
| containment by isolating all systems discharging into the area, in order to prevent the problem from escalating.
| |
| Ilowever. systems being used to assure adequate core cooling, shut down the reactor, to suppress a fire, or protect primary containment integrity are not to be isolated. Preventing core damage, maintaining primary containment integrity and ensuring personnel safety have priority over secondary containment concerns.
| |
| EPSTG'0002 B . 222 Revision 3
| |
| | |
| O SRO EXAM KEY l Esam Number ,NRC SRO Rev. 0 Esam Title NRC SRO EXAM !
| |
| 40 A high radiation alarm exists on the Annulus ventilation system (RMS'REllD). You are monitoring the CRT bar chart display for RMS'REll A to vandate the alarm condition on RMS*REllil. The 10 minute trend data for RMS*REll A is colored " light blue". Which of the following describes the status of RMS'REll A data' readings?
| |
| : a. RMS'REll A is reading within IV.of RMS'RElID.
| |
| : b. RMS' REIL A is in an Alert condition.
| |
| : c. RMS'REll A data is " questionable",
| |
| : d. communication has been lost between RM 80 and RM 23.
| |
| ANSWER:
| |
| : c. RMS'REli A data is " questionable".
| |
| IDNO: LP# Olti #
| |
| 420 llLO-069 7 PROCEDURE NUMBER: OTHER:
| |
| SOP @86 LEVEL 3 i 'NRC EA: ' I RO5 l SRO:' l 296034 E A1.01 3.8- 38_
| |
| COMMENTS: 7/97 new LOTM 66 T
| |
| \ 40
| |
| . , , . . ., , . . . . - , . , . _ - . - - _ . - - - - , . . - . - . . . . ~ . . . .-~-,. .
| |
| | |
| O The trend history data bar chart is color-coded to provide the operator with status information. Light blue, cyan,is used to mark questionable data. Half-intensity
| |
| (") white, appears gray,is used to indicate tt: occurrence of an RM-80 power failure during the history time interval associated with a history data bar. Red, yellow, and green are used solely to indicate that the scaled values fall above the current high limits, alert alarm limits, or fall below them, respectively. Consequently, red, yellow, or green color in the history data portion of the trend display is not intended to provide a history of actual alarm condition, especially if alarm limits have changed.
| |
| NOTE: The cyan and half-inte tsity white-colored indications override the red, yellow, and gre m-colored indications. For example, if a history value is marked as questionable, and that value exceeds the current alert and high alarm limits, the bar value is still c'isplayed in light blue.
| |
| The operator ins'tuction partition area of the CRT screen is used to provide the operator with a rend color--code key which shows the status represented by each color that may r.ppear on the trend display.
| |
| The operator can select one of three trend displays for the currently selected channel by depressing the associated trend display function key, e TREND 10 MIN f'~'s V e TREND IlOURLY e TREND DAILY The first key, TREND 10. MIN, shows channel activity over the previous 4 hours, as a set of twenty-four 10-min averages. The second key, TREND HOURLY, shows channel activity over the past day, as a set of 24 hourly averages. The third key, TREND DAILY shows channel activity over the past 28 days, as a set of 28 daily averages.
| |
| : 4. Monitor items The monitor data base items display is or541 red for the selected channel by depressing the MONITOR ITEMS function' key. This display shows channel chameteristics which maybe shared with or may pertain to other channels asse lated with the RM-80 monitor for the currently selected channel.
| |
| : 5. Channel items Depression of the CHANNEL ITEMS function key will order a display for the
| |
| ' selected channel which shows channel chara:teristics that are specific to a single channel and are independent of charactedstics of other channels associated with the RM-80 monitor for the cu Tently selected ch'annel. These items include such values as hiah and alert alarm limits, entered background information, conyt.rsion factors, (3) etc.
| |
| LOTM-65-4 PAGE 7 OF 69
| |
| | |
| I L
| |
| SRO EXAM KEY Esam Number NRC.*iRO Rev. 0 Esam Title NRC SRO EXAM 41 EOP 3, Secondary Containment and Radioactivity Release Control, must be entered if the I Secondary Containment differential pressure is above the maxinum normal operating difTerential l pressure.
| |
| Which one of the following is the reason for this entry condition?
| |
| : a. A significant steam leak into the secondary containment is indicated.
| |
| : b. A significant water leak from primary system may be discharging radioactivity directly to the secondary containment.
| |
| : c. A potential for the loss of secondary containment is indicated that could result in uncontrolled radioactive releases,
| |
| : d. An increase in the unmonitored ground level radioactive releases due to leakage through secondary containtnent is indicated.
| |
| ANSWER:
| |
| : c. A pote.o!al for the loss of secondary containment is indicated that could r-suit in uncontrolled radioactive releases.
| |
| IDNO: LP# OILI #
| |
| 377 IILO 515 5 PROCEDURE NUMBER: 0?HER:
| |
| EPSTG*0002 i EVEL 2
| |
| [295035 EK1.01 1 3.9~ 4.2 , _
| |
| COMMENTS: 7/97 new EPSTO'OOO21, Appendm B, p. 252 OF 269 RO T1 G2 SRO T102 O' '
| |
| 41
| |
| | |
| EOP 3 Secondary Containment Control i
| |
| .. Entry Conditions j Diferential pressures above the Maximum Normal Operating Value [i A high secondary containment differential pressure indicates a potential loss of
| |
| ; secondary containment structural integrity and could result in uncontrolled release of radioactivity to the environment. l I
| |
| 1 4
| |
| 4 J
| |
| EPSTG'0002 B 210 - Revision 3
| |
| ,,,...,,,.,.,....,--.y....,,,,.r v n- - wy *--* i-~ * -- r=''e--e-<r*--wm e m em e - -i--wevm'*'*m- *v- vem w=- - * * - * = = - - -+-v-'v--- ==-----=~e---==- - * - * * * = - * =
| |
| | |
| SRO EXAM KEY Esam Nmber NRC SRO Rev. 0 Esam Title NRC SRO EXAM 42 If one of the *ll" Reactor Recirculation System Flow Converter fails (resulting in zero output) with the reactor operating at 100% power, which one of the following describes what will be generated in APRM Channel 117
| |
| : s. A Downscale Alar i and a Rod Illock.
| |
| : b. A Rod Ilkxk only,
| |
| : c. A flalf Scrarn signal only,
| |
| : d. A Rod likxk and a llalf Scrata signal.
| |
| ANSWI:Rt
| |
| : d. A Rod Iliock and a llalf Scram signal.
| |
| lilNO: LP# OBJ #
| |
| 13 IILO 061 7 PROCEDURF NL'"BER: OTHER:
| |
| p SOP 0074 ATT 6 LEVEL 3 I .._......_._I......l...SRO:
| |
| NRC KA: RO: '
| |
| :20100b K5.05 1 3.6' 3.6 l i ?16005 K5.05 ' i~ 36I 36 _j COMMENTS: 7/97 new (Flow biased setpoint decreases, and average thermal power exceeds it.)
| |
| RO T2 01 SRO T2 G1 O c
| |
| | |
| ATTACHMENT 6 PAGE 1 OF 1 APRM SCRAM TRIPS / CONTROL ROD BLOCKS t
| |
| SCRAM TRIPS O SETPOINTS APRM UPSC TliERM 1. a. Two Recirculation Loop Operation s 0.66 W + 48% max 111.0% rated thermal power
| |
| : b. Single Recirculation Loop Operation :
| |
| s 0.66 W + 42.7% max 111.0% rateJ # ermal power APRM UPSC NBUT 2. a. RUN Mode - s 118% rated thermal power
| |
| : b. NOT in RUN mode - 515% rated thermal power APRM INOP 3. a. Switch ng in OPERATE
| |
| : b. < 11 LPRM inputs
| |
| : c. Flow converter inop
| |
| : d. Module unplugged CONTROL ROD BLOCKS SETPOINTS APRM UPSCALE 1. Setpoints for all instruments is as follows:
| |
| : a. Two Recirculation Loop Operation 50.66 W + 42%
| |
| O nted thermal power,
| |
| : b. Single Recirculation Loop Operation 50.66 W + 36.7%
| |
| rated thermal power
| |
| : c. s 12% rated thermal power NOT "4 RUN mode.
| |
| SETPOINT APRM DNSCALE 1. Setpoint for all instruments is as follows:
| |
| : a. s 5% rated thermal power O
| |
| SOP-0074 REV-6 PAGE 18 OF 35
| |
| | |
| SRO EXAM KEY Q Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM 43 During valse time testing on RilR System A. IEl2'MOVF004A, RilR pump A Suppression Pool Suction Valve, is closed with all other valvevswitches in their normal standby position when a valid LOCA signal occurs. in this condition, RilR pump A breaker will:
| |
| : a. Close and immediately trip because of the IE12*MOVr004A contacts in the breaker
| |
| 'tip circuit.
| |
| : b. Not close because of the IEl2'MOYF004A contacts in the breaker close pennissive circuit.
| |
| : c. Close after IE12'MOVr004A opens automatically,
| |
| : d. Close and remain closed, while IE12*MOVF004A remains closed.
| |
| ANSWER:
| |
| : a. Close and immediately trip because of the IE12*MOVI 004 A conmets in the baaker trip circuit.
| |
| II)NO: LP# Olki #
| |
| 140 11L0-021 - 6 PROCEDURE NUMBER: OTHER:
| |
| 828E534AA (sys204) LEVEL 4
| |
| ~
| |
| ! ' NRC K A: l "MO: l SRO:'
| |
| i203000 K4.06 i 3.5' 3.5 COMMENTS: 7/97 new RO T2 01 SRO T2 01 1
| |
| 4 43
| |
| | |
| l The redundancy of the RilR System provides assurance that LPCI capacity will be adequate i
| |
| to support ECCS requirements in the event of a LOCA, regardin m nitial system lineup (V3 or a failure in one of the loops.
| |
| : 11. SYSTEM DETAILS A. Component Descriptions NOTE- Unless otherwise stated, all valves referenced have the prefix E12.
| |
| : 1. RHR Pumps The Ril pump are multi stage, deepwell, vertically mounted pumps. Each pump is t . 2d at 5165 gpm at a head of 293 ft (130 psig). Maximum runout flow is 6060 gpm. Maxirnum shutoff head is less than 760 fl. (339 psi). The design pressures for the pumps and piping are 500 psig discharge pressure and 200 psig suction pressure.
| |
| A minimum flow line is provided for each pump to protect it froin overheating at low flow rates. The R11R pump A (B)(C) minimum flow valve, MOVF064A (B) (C) automatically opens on a low flow condition ofless than i100 gpm, as sensed by Flow Transmitter N052A B) (C), after a pump start, as sensed by RIIR pump breaker closed (in either the test or fully racked in positions). The minimum flow valves in loops A and B, MOVF064A (B), have an eight second time delay prior to opening on a low flow condition. This time delay is provided f] to allow the operator time to establish an RHR pump flowpath in the moder if
| |
| ' operation other than LPCI, prior to flow being sent to the suppression pool. This time delay reduces the chance of puniping reactor water to the suppression pool.
| |
| The RHR pumps A and B are interlocked to ensure that a suction source is present. 'Inc pumps will AUTO trip if Shutdown Cooling inboard Isolation Valve MOVF009, Outboard Isolation Valve MOVF008, or RHR Pamp A(B)
| |
| Shutdown Cooling Suction Valve MOVF006A(B) is not f:'!y c when suppression pool suction valve MOVF004A(B) and Fuel Pod Golg Pool Suction Valve MOVF066A(B) are not fully open (See Figure 12).
| |
| If MOVF064A(B) is open in conjunction with MOVF006A(B), an alarm will annunciate on panel H13-PNLP601 in the main control room. In this situation, if MOVF008 and MOVF009 were open, a path would exist to drain the reactor vessel to the suppression pool.
| |
| There are two conditions which will cause an RHR pump to trip:
| |
| : a. electrical fault, (>l50 amps)
| |
| : b. no fully open suction path Figure 2 shows the pump logic for RHR pump PC002B. RHR pumps PC002A and PC002C are similar. See Figure 3 for the pumps power supplies.
| |
| A i i V
| |
| LOTM-19-7 Page 4 of 37
| |
| | |
| c vn SRO EXAM KEY Enam Number NRC SRO Rev. O Exam Title NRC SRO EXAM 44 Following a Loss of Coolant Accident the following plant parameters exist:
| |
| _ Rea: tor pressure is 460 psig Ves:el level is -80 inches Drywd: pressure is 2.2 psid
| |
| . Conta nmet pressure is normal and steady.
| |
| Whi:h one of the following describes the Low Pressure Coolant injection mode of the Residual llea Rcmovalsystem7
| |
| : a. Pumps have started, but are not injecting because the injection valves, F042A, B, and C have not opened.
| |
| : b. Pumps have started, injection valves F042A, B, and C have opened, but reactor pressure is too high for injection.
| |
| : c. Pumps have not started, but injection valves F042A, B and C have opened.
| |
| : d. Pumps have started, injection valves F042A, B and C have opened, and injection has started.
| |
| p ANSWER:
| |
| .V h. Pumps have started, injection valves F042A, B, and C have opened, but reactor pressure is too high for injection.
| |
| IDNO: LP# OBJ #
| |
| 145 IILO-021 9 PROCEDURE NUMBER: OTHER:
| |
| SOP 4031 LEVEL 2 l203000_A.101_ j_ 4.2L 1 __4,3 j203000_K1.17 , ;_ __ 4 L ____4
| |
| :203000 A1.02
| |
| * 3.9 4 .__
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 01 O 44
| |
| | |
| e 4.1.8 Verify the, green and white indicating lights are on for all 3 RHR pump control switches.
| |
| 4.1.9 Verify all RHR System status lights off.
| |
| 4.2 Manual LPCI Startup NOTE All controls and indications are located on panel IH)3*P601 unless noted ofherMise.
| |
| NOIE Ifneeded the LPCI mode ofRIM may
| |
| * be manually initiated using the LPCS! RIB DIV1M4NURL INITIATION switchfor L.PCI A loop, or the RHR DIY 2 MANUAL INITIATIONsuitchfor the
| |
| . LPCIB & Cloop. Dtis uill cause the associated Div I or 11 Diesel Generator to start and initiate the associated Div I
| |
| \ or 11 Load Shed and Sequencing.
| |
| 4.2.1 Verify the system is in standby per Section 4.1.
| |
| 4.2.2 Start IE12*C002A(BXC) RHR PUMP A(BXC) in the desired loop.
| |
| 4.2.3 Verify pump amps less than or equal to 91 amps.
| |
| 4.2.4 Establish one of the following flowpaths:
| |
| : 1. Ifreactor is less than 450 psig, open IE12*FN2A(BXC) RHR PUMP XC) LPCI INJECT ISOL VALVE to establish flow to RPV.
| |
| CAUI1ON Do not operate the Ll'CS pump and RHR A Pump in the test return to suppression pool mode simultaneoisly.
| |
| : 2. If reactor pressure is er than 450 pig, open IE12*F024A(B)
| |
| (IE12*F021) RHR P A (BXC) TEST RTN TO SUP PL to establish a suppression pool to suppression pool loop.
| |
| 4.2.5 Verify 1E12*F064A(BXC) RHR PUMP A(BXC) MIN FLOW TO SUP PL <
| |
| closes when flow exceeds 1100 GPM.
| |
| O SOP-0031 REV - 17B PAGF 12 OF 107
| |
| | |
| 4.2.6 At approximately 250 psig reactor establish RHR flow to the RPV if desired by the perfomung of the llowing:
| |
| : 1. Verify open IE12*F(M2A(BXC) RHR PUMP A(BXC) LPCI INJECT ISOL VALVE (S).
| |
| : 2. Verify closed IE12*F024A(B)(IE12*F021) RHR PUMP A(BXC)
| |
| TEST RTN TO SUP PL.
| |
| 4.3 Shutdown Cooling Flush, Wannup and Startup CAU110N Use extreme caution wten . Refer to Precaution and Umitation 2.2. (Ref. 7.35) positioning the Shutdown Cooling Vahw CAUIlON During system warmu) or shutdown cooling operation, do not exceed 100 F/hr temp changes to the reactor water or R3R pump, u
| |
| CAU110N
| |
| , Prior to initiati Shutdown Cooling the leads lifted per AMR 894013 SHALL be relanded to O' wmide oved
| |
| $3D ISOL VALVE.
| |
| loss of power annunciation for IE12*MOVFD09 RHR SHUTDOWN COOUNG NOTE Ifperfonnirg this procedwefrom the remote shutdown pawls, E12*F006A(B)
| |
| RlB PUMP A(B) SDCSUCTION VAL VE(S) will opertte m throttle vdves.
| |
| 4.3.1 Shutdown Cooling Flush
| |
| : 1. Close IE12*VF085A(B) LPCS FILL PUMP STOP CHECK TO RHR A DISCH (DISCH FILL PUMP STOP CHECK TO RHR B DISCH).
| |
| : 2. Rack out IENS*SWGIA ACB03 (IENS*SWGlB ACB23) RHR PUMP A(B) Breaker.
| |
| : 3. Close IE12*F064A(B) RHR PUMP A(B) MIN FLOW TO SUP PL and close IE12*F004A(B) RHR PUMP A(B) SUP PL SUCTION VALVE.
| |
| : 4. Perfam the following- .
| |
| : a. Verify E12-MOVF008 ENABLE / DISABLE switch in ENABLE O (located at IC61-PNL001 in Div 1 RSS Room).
| |
| FOP-0031 REV - 17B PAGF 13 OF 107
| |
| | |
| . = . . . . . .....__ _ - - . - . - . =. - .. . .. ..-
| |
| SRO EXAM KEY O Exam Number NRC-SRO Rev, O. Exam Title NRC SRO EXAM 45 The RCIC system is in Standby Lineup, but the RCIC TURillNE EXilAUST SilUTOFF valve, E51-F068, is closed for a valve stroke test. A loss of feedwater causes a low reactor water level (level 2).
| |
| Select the statement w hich describes how the RCIC system will respond.
| |
| : a. RCIC TURillNE EXilAUST SilVTOFF valve, E51 F068, automatically opens; RCIC system initiates and injects water into the RPV.
| |
| : b. The RCIC turbine will start and trip on high RCIC turbine exhaust pressure at 25 psig.
| |
| : c. RCIC starts and the RCIC system exhaust rupture diaphrams will rupture initiating a RCIC system isolation at 10 psig exhaust diaphram pressure.
| |
| : d. RCIC turbine does not start. RCIC TURBINE EXilAUST SilVTOFF valve, E51 F068, must be open for RCIC STEAM SilUTOFF valve, E51.F045, to open.
| |
| ANSWER:
| |
| : d. RCIC turbine does not start. RCIC TURBINE EX11AUST SilUTOFF valve.
| |
| E51 F068, must be open for RCIC STEAM SilUTOFF valve, E51 F045, to open.
| |
| O IDNO: LP# OBJ #
| |
| 10 llLO-017 5 PROCEDURE NUMBER: OTHER:
| |
| SOP-0035 LEVEL 4 AOP-0031 ESKICS05 ESKICS06
| |
| [ NRC KA I ROi l 'SROI }
| |
| -217000 A2.03 3.4 3.3 !
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1
| |
| ( n
| |
| | |
| RCIC VALVE CONTR AND INTERLOCKS VALVE VALVE AUTO CONTROLS & INTERLOCKS NUMBER NAME POWER SOURCE F063 Turbine Steam Supply Inboard 480VAC 1 Ells *MCC2D/lEJS*SWG2B Shuts on Div 2 Isolation isolation F064 Turbine Steam Supply Otbd Isol 480VAC IEIIS*MCC2D/IEJS*SWG2B Shuts on Div 2 Isolation F076 Turbine Steam Supply Warmup 480VAC IEIIS*MCC2D/IEJS*SWG2B Shuts on Div 21 solation F025 Stean. Supply Drain Line 125VDC 1ENB'PNLO2AllENB*SWG01 A Shuts if F045 not shut -
| |
| Upstream Isolation F026 Steam Supply Drain Line 125VDC 1ENB'PFLO2A/IENB*SWG01 A Shuts if F045 not shut F054 Steam Supply Drain Line Trap 125VDC 1ENB'PNLO2AllENB*SWG01 A Cycles open when high level in drain trap exists Bypass F045 Steam Supply to Turbine 125VDC 1ENB*MCC1/1ENB*SWG01 A Opens on RCIC initiation if F%8 is open.
| |
| Closes on reactor high level (+51")
| |
| C002 Turbine Tripffhrottle 125VDC 1ENB*MCC1/1ENB*SWG01 A Cannot open if overspeed trip is actuated.
| |
| NOTE: This also supplies the trip solenoid Shuts for turbine trips, must be reset locally F004 Turbine Exhaust Drain Pot 125VDC 1ENB*PNLO2B/IENB*SWG01B Shuts if F045 not fully shut ;
| |
| Upstream Isolation F005 Turbine Exhaust Drain Pots 125VDC IENB*PNLO2A/IENB*SWG01 A Cycles on high trap level Downstream Isolation Opens if F045 fully shut .;
| |
| F068 Turbine Exhaust to 125VDC 1 Ells *MCC1/1ENB*SWG01 A None F077 Vacuum Breaker Isolation 480VAC 1 Ells *MCC2G/1EJS*SWG2A Shuts if high drywell pressure and lowstream Turbine Exhaust line supply pressure exit 110-01 PageIof2 IILO-017
| |
| | |
| - - ~ =- .- - -- . -- . - - - . - . ...
| |
| l SRO EXAM KEY O Esam Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM 46 The plant is operating at 100 % power, steady state. The Control Room Operator is performing LPCS Quarterly Pump Surveillance. ne LPCS pump is running in the test return to the suppression pool mode. A steam leak in the Drywell caused Drywell pressure to increase to 1.72 psid. Reactor pressure is being maintained at 950 psig by the bypass valves.
| |
| Which of the following statements describes the response of the LPCS system 7
| |
| : a. The LPCS Pump will load shed then remain in standby.
| |
| The E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21 F0ll (LPCS MIN FLO TO SUPP POOL) opens.
| |
| : b. The LPCS Pump will continue running.
| |
| The E21 F012 (LPCS TEST RTN TO SUPP POOL) remains open.
| |
| The E21 F0ll (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| : c. The LPCS Pump will continue running.
| |
| ne E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21 F011 (LPCS MIN FLO TO SUPP POOL) opens.
| |
| : d. The LPCS Pump will load shed then remain in standby.
| |
| The E21 F012 (1.PCS TEST RTN TO SUPP POOL) closes.
| |
| O De E21 F0ll (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| ANSWER:
| |
| : 5. The LPCS Pump will continue running.
| |
| He E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21 F01I (1.PCS MIN FLO TO SUPP POOL) opens.
| |
| IDNO: LP# OllJ #
| |
| 247 LOTM-17 9 PROCEDURE NUMBER: OTHER:
| |
| SOP 0032 LEVEL 2
| |
| ' ~NRC Ui '~l 30i l'SRO[
| |
| !209nOI K4.08 3.8 4 COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 46
| |
| | |
| l l
| |
| CAUTION Simultaneously operating LPCS and RHR A in Test Return Mode can exceed the design of the common return line and cause piping damage. Do ny simultaneously operate LPCS and RIIR A in the Test Return Mode to Suppression Pool.
| |
| NOTE E21-F005, LPCSINJECTISOL VALVE opens only ifthe pressure downstream ofthe valve is less than 550psig.
| |
| 4.2.3. Establish a discharge flow path per one of the following:
| |
| . LPCS flow to RPV, open E21-F005, LPCS INJECT ISOL VALVE e LPCS flow to Suppression Pool, open E21 F012, LPCS TEST RETURN VLV TO SUPPRESSION POOL.
| |
| 4.2.4. WHEN flow rises above 875 gpm, THEN verify E21-F011, LPCS MIN FLOW VLV TO SUPPRESSION POOL closes.
| |
| 4.3 ManualInitiation NOTE Thefollowing step also initiates RHR A LPCIMode and starts the Div IDiesel Generator.
| |
| 4.3.1. Arm and depress the LPCS/RHR DIV 1 MANUAL INITIATION Pushbutton.
| |
| 4.3.2. Oo To Section 5.1.
| |
| 5 SYSTEM OPERATION 5.1 Operation from Auto Initiation 5.1.1. Perform the following:
| |
| : 1. Verify E21-C001, LPCS PUMP started.
| |
| O SOP-0032 REV - 16 PAGE 8 CF 27
| |
| | |
| a
| |
| : 2. Check E21-C001 LPCS MOTOR AMP less than 175 amps as indicated on -
| |
| E21 C001, LPCS MOTOR AMPS.
| |
| : 3. LF reactor pressure is below 487 psig, THEN verify 521-F005, LPCS INJECT ISOL VALVE opens.
| |
| ;4. E E21-F012, LPCS TEST RETURN VLV TO SUPPRESSION POOL is open, THEN verify E21 F012 closes.
| |
| : 5. WHEN LPCS Pump flow rises above 875 gpm THEN verify E21 F011, LPCS MIN FLOW VLV TO SUPPRESSION POOL closes.
| |
| 5.1.2. WHEN it has been verified by 2 independent indications that LPCS flow is ng needed, AND it is desired to leave the LPCS Pump running, THEN perform the following:
| |
| : 1. Close E21-F005, LPCS INJECT ISOL VALVE.
| |
| : 2. Open E21-F012, LPCS TF.ST RETURN VLV TO SUPPRESSION .
| |
| POOL.
| |
| s 6 SYSTEM SHUTDOWN
| |
| ~h (J 6.1 Verify adequate core cooling is assured by two indepedent indications.
| |
| 6.2 LF LPCS was manually iaitiated, THEN verify the LPCS/RHR DIV 1 MANUAL INITIATIOI4 collar is in the DISARM position.
| |
| NOTE
| |
| , ifthe LPCS initiation signal is unable to be reset using the LPCS/RHR DIV 1 INITM TION RESET Pushbutton, the LPCS Pump can be overridden andstopped This will prevent an automatic start ofthe LPCS Pump until the initiation signals clear and the LPCS/RHR DIV 1 INITM TION RESET Pushbutton is depressed 6.3 g an automatic initiation signal is sealed in, THEN depress the LPCS/RHR DIV 1 INITIATION RESET Pushbutton and check the white light goes off.- .
| |
| 6.4 Verify the following valves are closed:
| |
| 6.4.1. E21 F012, LPCS TEST RETURN VLV TO SUPPRESSION POOL 6.4.2. E21-F005, LPCS INJECT ISOL VALVE SOP-0032 REV - 16 PAGE 9 OF 27
| |
| | |
| i TABLE 1 (continued) n.
| |
| ~- LOW PRESSURE CORE SPRAY CONTROLS P601-21C NOMENCLATURE CONTROL FUNCTION LPCS Pump Sup PL CLOSE
| |
| * Closes valve E21*F001.
| |
| Suction Valve Switch can be taken to close E21A-S1 position only with key (keyed switch) inserted.
| |
| OPEN
| |
| * Opens valves E21*F001. Key can only be removed from
| |
| . switch in the open position.
| |
| LPCS Pump STOP
| |
| * Opens LPCS pump circuit (w/ flag indicator) breaker at ENS *SWG1A, Spring return stopping pump. Green flag to center indicator indicates " Auto-after-stop" condition ''
| |
| START
| |
| * Closes LPCS pump circuit (w/ flag indicator) breake,r at ENS *SWG1A, starting pump. Red flag indication indicates " Auto
| |
| (~}
| |
| (,j -after-start" condition LPCS/RHR Div I STOP
| |
| * Stops discharge line fill Line Fill Pump (w/ indicator flag) pump C002. Green indicator Spring return to flag indicates pump center contacts open.
| |
| START
| |
| * Starts discharge fill (w/ indicator flag) pump C002.- Red indicator flag indicates pump contacts .
| |
| shut.
| |
| LPCS Min Flow VLV CLOSE
| |
| * Closes valve E21*F011 to sup pool AUTO
| |
| * Allows valves E21*F011 to E21A-S4 (spring return to) be controlled by LPCS pump flow transmitter PIS-N651 when LPCS pump is running OPEN
| |
| * Opens valve E21*F011 r~s s LOTM-17-6 Page 12 of 22
| |
| | |
| . TABLE 1 (continued)
| |
| U O'. -
| |
| LOW PRESSURE CORE SPRAY CONTROLS P601-21C NOMENCLATURE CONTROL EIDiCTION LPCS test return CLOSE
| |
| * Closes valve E21*F012 VLV tc sup pool AUTO
| |
| * Valve E21*F012 will E21A-S5 (spring return to) automatically close upon ;
| |
| LPCS initiation signal OPEN
| |
| * Opens valve E21*F012 LPCS Inject Isol CLOSE '' Closes valve E21*F005 Valve E21A-S2 AUTO
| |
| * Valve E21*F005 will (spring return to) automatically open upon LPCS Initiation when reactor pressure is less than 487 psig .
| |
| OPEN
| |
| * Will open E21*F005 providing low pressure permissive l (550 paid) signal from PIS-N650 is present or 1
| |
| ("' initiation signal present and
| |
| <487 psig.
| |
| l
| |
| 's I l
| |
| LPCS testable check TEST (PB)
| |
| * The actuator has been Valve Actuator removed, however the :
| |
| E21A-S3 pushbutton remains.
| |
| I s
| |
| 4 a LOTM-17-6 Page 13 of 22
| |
| | |
| TABLE 4 VALVE INTERLOCKS AND AUTOMATIC FUNCTIONS Y& LYE INTERLOCK / FUNCTION i
| |
| E21*F001 None E21*F005
| |
| * Auto open ift ,
| |
| - LPCS Initiation Signal present,
| |
| - Pwr available on Div I Bus, and
| |
| - Rx Vessel Pressure <487 psig
| |
| * Manually opened using control switch:
| |
| - Without LOCA Initiation Signal present:
| |
| * only if pressure between E21*F005 and E21*F006 is <550 psig (this low pressure permissivg_ starts a 15 minute time delay to allow F005 to be opened regardless of pressure for 15 minutes)
| |
| With LOCA Initiation Signal present F005 can be manually opened with Rx Vessel Pressure >487 psig, b E21*FO 1
| |
| * Auto OPEN with LPCS pump running and flow less than 875 gpm.
| |
| * Auto CLOSE when flow exceeds 875 gpm.
| |
| E21*F012
| |
| * Auto CLOSE upon receipt of LPCS initiation signal.
| |
| (~\
| |
| 'd LOTM-17-6 Page 20 of 22
| |
| | |
| t SRO EXAM KEY
| |
| 'C -
| |
| Enem Number NRC SRO Rev. 0 Exam Title NRC SRO EXAM 47 'A LOCA has resulted in the automatic start of the llPCS system and injection into the vessel.
| |
| During the transient, the operator closed llPCS Injection Valve (E22'F004) using the control switch when the following conditions existed:
| |
| Reactor water level - +50 inches - :
| |
| Condensate Storage Tank Level = 100 feet (Mean Sea Level)
| |
| Suppression Pool Level = 19 feet 4 inches Which one of the following descaibes the automatic response of the llPCS Suppression Pool Suction isolation Valve (E22'F015) and the llPCS Injection Valve (E22'F004) when reactor water level dectuses to -$0 inches and Suppression Pool level increases to 20' 6"? (Assume NO operator action.)
| |
| SUPPRESSION POOL SUCTION INJECTION VALVE ISOLATION VALVE (F015) (F004)
| |
| : a. Opens Opens
| |
| : b. Remains closed Opens c Remains closed Remains closed
| |
| : d. Opens Remains closed ANSWER:
| |
| : d. Opens Remains closed IDNO: LP# OBJ #
| |
| 74 IILO-019 9 PROCEDURE NUMBER: OTHER:
| |
| ARP-601 16-C05 LEVEL 4 ESK 6CSH01 ESK 6CSH03
| |
| [209002 K4 05 ___j_.,2Ai2.4 1
| |
| l (209002_ K 6.02 _ J _3.4( _3,4
| |
| ?209002 A2.03 i 3.21 3.4 _j f
| |
| I i
| |
| l- . _ _ _ _ . . . _ . - . _ - _ _ _ _ _ -, , ,-
| |
| | |
| HPCS SUCTION XFER SUP PL LEVEL HIGH g\LARM NO. 2336 IH13"P60dl6A/C05 INITIATING DEVICES SETPOINTS
| |
| : 1. E22*LISN655C or G Suppression Pool Level Indicating Switches 1. 20'4" AUTOMATIC ACTIONS
| |
| : 1. E22*F015 HPCS PUMP SUP PL SUCTION VALVE opens.
| |
| : 2. E22*F001 HPCS PUMP CST SUCTION VALVE closes when E22*F015 is full open.
| |
| : 3. - E22*F010 HPCS TEST BYPASS VLV TO CST and E22*F0ll HPCS TEST RETURN VALVE TO CST close if open.
| |
| OPERATOR ACTIONS
| |
| : 1. Verify Automatic Actions occur.
| |
| : 2. If HPCS is in operation, verify pump flow is inormal.
| |
| : 3. Refer to the EOP's.
| |
| : 4. Verify level on CMS *L123A and B SUPPRESSION POOL LEVEL H13*P808.
| |
| LONG TERM ACTIONS NOTE Minimi:e the amount ofin)ection timefrom Suppression Pool water to the Reactor Vessel.
| |
| : 1. Refer to Technical Specification 3/4.5.3.
| |
| : 2. Restore suppression pool level to normal using SOP-0031 RESIDUAL HEAT REMOVAL,
| |
| : 3. Return HPCS sudon to CST when level is normal.
| |
| ~
| |
| POSSIBLE CAUSES
| |
| : 1. HPCS or RCIC running or minimum flow.
| |
| : 2. Leakage from water sources in containment.
| |
| : 3. Relief valve operation.
| |
| REFERENCES
| |
| : 1. 828E536AA Sh 9
| |
| : 2. 1.lLCSH.0ll
| |
| : 3. 1.lLCSH.012 ARP.601-16 REV9 PAGE 19 OF 49
| |
| | |
| LIST OF ERINCIPAL INTERLOCKS BYPASSED BY OPERATION ENCLOSURE 2A FROM 111E DIVISION 1 SECTION OF RSS PANEL IC61*P001 PAGE 1 OF 2 O
| |
| V SYSTEM I.D. INTERLOCKS AND COMPONENT DESCRIPTION NUMBER CONDITIONS BYPASSED RCIC STEAM SUPPLY INBD E51-F063 a. RCIC. isolation signals ISOL VALVE RCIC PUMP CST SUCTION E51 F010 a. Auto open at RPV water level 2 VALVE b. Open interlock with E51 F031 open RCIC TEST BYPASS VLV TO E51 F022 a. Auto close with E51-F031 open CST b. Auto close with RPV water level 2 RCIC TEST RETURN VLV TO E51 F059 a. Auto close with E51 F031 open CST b. Auto close with RPV water level 2 RCIC PUMP SUP PL SUCTION E51 F031 a. Open interlock with suction and VALVE test return to condensate storage tank
| |
| : b. Auto open with:
| |
| : 1. no isolation signals
| |
| : 2. condensate storage tank low
| |
| : 3. suppression pool level high
| |
| : c. Auto close with isolation signal RCIC STEAM SUPPLY OUTBD E51-F064 a. RCIC isolation signals ISOL VALVE RCIC TURBINE EXHAUST TO E51-F068 N/A SUPPRESSION POOL RCIC TRIP & THROTTLE E51 F002 N/A
| |
| _V ALVE OPERATOR RCIC STEAM SUPPLY E51-F045 a. Auto open with RPV water level 2 TURBINE STOP VALVE and IE51*MOVF068 open
| |
| : b. Open interlock with lE51*MOVF068 closed
| |
| : c. Auto close with RPV water level 8 RCIC TURBINE LUBE OIL CLG E51-F046 a. Auto open with RPV water level 2 WATER SUPPLY VLV GLAND SEAL COMPRESSOR E51-PC002 a. Auto start with initiation signals RCIC TURB EXH VAC BRYa E51 F078 a. Open interlocks with no high UP STREAM ISOL VALVE drywell pressure and no low steam supply pressure
| |
| : b. Auto close with high drywell pressurs and low steam supply RCIC TURB EXH VAC BRKR E51-F077 pressure rq DN STREAM ISOL V
| |
| A OP-0031 REV - 10 A PAGE 31 OF 107
| |
| | |
| e l
| |
| l SRO EXAM KEY l' Esam Number NRC.SRO Rev. O Exam Title NRC SRO EXAM l COMMENTS: 7/97 ww
| |
| ! RO Tk G1 l SRO T2 G1 l
| |
| 5 i
| |
| i
| |
| ! i I
| |
| k l
| |
| 1 l'
| |
| d I
| |
| a i
| |
| i i
| |
| 1 1
| |
| i 1
| |
| a 4
| |
| 1 1
| |
| I i
| |
| 5-i 48 i
| |
| lJ J
| |
| -w ww -w.. w ww , , - w-- ..w- e-.-- . , , . . . - ww- - , ,- - . ,..-.,,--.~,w.ru . - . . , . . - - - -, . - - - , , - - - . . , . ~ . , . - , .
| |
| | |
| SRO EXAM KEY
| |
| -(] Esam Number NRC.SRO Rev. O Esam Title NRC SRO EXAM 43- SLC System A is in a normal STANDBY lineup but with the TEST TANK OUTLET VALVE (C41'F031) OPEN. Which of the following most accurately describes the effects on the STORAGE TANK OUTLET Vi.'.VE (F001) and SLC PUMP A by placing the SLC Keylock Control Switch for Pump A to START 7
| |
| : a. Valve F001 Opens, SLC Pump A Starts after the valve reaches its Full Open position.
| |
| : b. Valve F001 Opens, SLC Pump A starts concurrently with the valve opening.
| |
| : c. Valve F001 does Not Open, SLC Pump A Starts
| |
| : d. Valve F001 does Not Open, SLC Pump A does Not Stan ANSWER:
| |
| : c. Valve F001 does Not Open, SLC Pump A Starts IDNO: LP# OIU #
| |
| 97 IILO-016 4 PROCEDURE NUMBER: OTHER:
| |
| 828E234AA LEVEL 2 ESK 6SLS01 a11000 K4.08 4.2- 4.2 ,j COMMENTS: Exam bank RO T2 G1 SRO T2 G1
| |
| % 49
| |
| | |
| - T*cHNIOUES - VII/ SUILIECT MATTER
| |
| : 4. -System In*.erlocks OBJ #4 4.1 The SLC pump Suction MOV-(1C41*MOV-F001A/B) will not open if the Test Tank Outlet valve (1C41*VF031)~ is open. '
| |
| 4.2 The SLC pump will not start when started l from the Main. Control Room without a suctica path.
| |
| 4.2.1 The Pump Suction MOV (1C41*MOV-F001A/B) QR the Test Tank Outlet (1C41*VF031) must be open.
| |
| 4.3 To prevent removing the aodium pentaborate from the vessel water the RWCU system is interlocked with the SLC system.
| |
| 4.3.1 When the SLC pump start switch on the 1H13-P601 panel is taken to run, the RWCU pump's Suction Containment Oi Isolation Valves (1G33*F001/4) get a close signal.
| |
| 4.4 The signal is divisionalized
| |
| * Starting the "A" SLC pump causes the Div 1 valve (1G33*FG04,-RWCU pump suction outboard Containment Isolation) to shut.
| |
| * Starting the #B" SLC pump causes the Div 2 valve (1G33*F001, RWCU pump Suction Inbcard Containment Isolation) to shut.
| |
| * Either valve going shut will cause the RWCU pumps to-trip on low flow.
| |
| : 5. System Controls and Indications hlLO 016 5 PAGE 16 of 29 2 .
| |
| :j -
| |
| , , ,_ . _ . , . ~
| |
| | |
| SRO EXAM KEY O Esam Number NRC-SRO Hev. O Exam Title NRC SRO EXAM 49 in ATWS has occurred and the following conditions exist: >
| |
| Reactor power 12%
| |
| Ructor water level 13.4 inches Drywell pressure 1.4 psig All scram valves are open Scre.m discharge volume is full.
| |
| All discharge volume vent and drain valves are shut Mode switch is in SilVTDOWN Which one of the following conditions will allo y a resetting of the scram signal? '
| |
| : a. The scram can be reset using the " SCRAM RESET" switches aner placing the CRD Scram Discharge Volume keylock switches to bypass.
| |
| - b. The scram can be reset by placing the Mode switch in "RUN" and placing the
| |
| " SCRAM RESE1" switches to reset aAct placing the CRD Scram Discharge Volume keylock switches to bypass.
| |
| : c. The scram can not be reset due to drywell pressure being elevated above the alarm setpoint.
| |
| : d. The scram can not be reset due to reactor power being above the alarna setpoint'.
| |
| ANSWER:
| |
| : a. The scram can be reset using the " SCRAM RESET" switches after placing the CRD Scram Discharge Volume keylock switches to bypass IDNO: LP# Olk)#
| |
| 260 LOTM-15 5 PROCEDURE NUMBER: OTHER:
| |
| AOP-0001 LEVEL 2 ARP-P680-6A
| |
| ;212000 A4.14 ! 3.8 3. 8 _.
| |
| COMMEN iS: 7/97 new RO T2 G1 SRO T2 G1
| |
| ,~
| |
| , so
| |
| | |
| SRO EXAM KEY Q Esam Number NRC SRO Rey, O Esam Title NRC SRO EXAM O
| |
| l t
| |
| (
| |
| O ,,
| |
| l 1 --.
| |
| | |
| 5.6 E both Recirculation Pumps are off,IHFS start at least one Pump in slow speed.
| |
| 5.7 Initiate STP-050-0700, RCS Pressure / Temperature Limits Verification.
| |
| 5.8 Monitor Botto.m Head Drain temperature on B21-R643, RX VESSEL TEMP MONITORING point 4 or Process Computer Point G33NA001 and take the following actions to prevent an excessive temperature change:
| |
| * Reset any FCV runback per ARP-P680-04A-A03 and ARP P680-04A-A09.
| |
| * Reset the scram per Step 5.9.
| |
| WAkNING Resetting ARI or the scram when fuel failuie is present can result in the transfer of highly radioactive fuel particles or fission products to plant areas t!.at have normal access. This can result in overerposure to plant personnel. Do nat reset ARI or the scram until a qualified technician has surveyed the Scram Discharge Volume for abnormal radiation levels.
| |
| C 5.9 Resetting the Reactor Scram 5.9.1. E ARI was initiated, THEN reset as follows:
| |
| : 1. Check Annunciator, P680-07A E01, ARI READY TO RESET is alarmed.
| |
| : 2. At H13-P680 or H!3 P632, depress ARI INITIATION RESET Pushbutton.
| |
| : 3. Place both ClIC-SI A and B, ARI CHANNEL A and B MANUAL INITIATION Pushbutton Collars to the DISARM position.
| |
| 5.9.2. Verify minimum operable channels per trip system for Nuclear Instrumentati'o n prior to resetting both divisions of RPS.
| |
| 5.9.3. Place the following Switches to BYPASS:
| |
| * C71 A-S4A, CRD SCRAM DISCH VOL HI WTR VLV BYPASS
| |
| * C71 A-S4B, CRD SCRAM DISCH VOL HI WTR VLV BYPASS
| |
| * C71 A S4C, CRD SCRAM DISCH VOL HI WTR VLV BYPASS I e C71 A-S4D, CRD SCRAM DISCH VOL HI WTR VLV BYPASS D\
| |
| AOP-0001 REV - 12 PAGE 6 OF 8 1
| |
| I
| |
| | |
| 5.9.4. Verify Reactor scram initiation conditions have cleared and all Reactor parameters are normal and stabilized.
| |
| 5.9.5. Reset the Reactor scram by placing the following Switches to the RESET position:
| |
| * C71 A-SSA, SCRAM RESET LOGIC A ~
| |
| * C71 A SSB, SCRAM RESET LOGIC B e C71 A 55C, SCRAM RESET LOGIC C
| |
| * C71 A SSD, SCRAM RESET LOGIC D
| |
| : 1. Check the following white lights on:
| |
| * CRI A and CRIB, RPS SCRAM SOV e CR2A and CR2B,RPS SCRAM SOV e CR3A and CR3B,RPS SCRAM SOV e CR4A and CR4B,RPS SCRAM SOV
| |
| : 2. Check SCRAM DISCH VOL VENT & DRN VLV POSN lights indicate red light only.
| |
| 5.9.6. WHEN Charging Water Header pressure returns to normal,IHEN check all HCU Accumulator faults clear.
| |
| 5.9.7. WHEN the Scram Discharge Volume has drained, THEN check:
| |
| : 1. Annunciator, P680-06A-A08, CRD SCRAM DISCH VOL HIGH WATER LEVEL is clear.
| |
| : 2. Annunciator, P680-06A-C08, SCRAM DISCH VOL NOT DRAINED is clear.
| |
| AOP-0001 REV - 12 PAGE 7 OF 8
| |
| ._a -
| |
| | |
| SRO EXAM KEY O Exam Number NRC-SRO Rey, O Exam Title NRC SRO EXAM 50 During a reactor startup you have been withdrawing SRM detectors per GOP-0001. All SRMs except A indicate full out. SRM A has an upscale high and an upscale high-high trip indicated and is reading off-scale high. The P680 indications show the detector is " driving out".
| |
| You should:
| |
| . a. Immediaiely insert control rods to return SRM A readings on-scale.
| |
| : b. Insert a Div i half scram and continue with the plant sta tup.
| |
| : c. Check the SRM A drive power fuses, if the problem is not corrested, obtain reactor engineering assistance.
| |
| : d. Since the other drive OUT lights are on, SRM A drive has power therefore contact I&C for assistance.
| |
| ANSWER:
| |
| : c. Check the SRM A drive power fuses, if the problem is not corrected, obtain reactor engineering assistance.
| |
| IDNO: LP# Olu #
| |
| $3 IILO-051 4 PROCEDURE NUMBER: OTHER:
| |
| SOP-0074 LEVEL 2 i___2.8,_ _.2.9 l215004_K4.04 3 3.3_j
| |
| :216004 A2.03 ;
| |
| COMMENTS: 7/97 new
| |
| $2
| |
| | |
| - -a.m - - - - - - -
| |
| i
| |
| ,, 5.2.13.
| |
| ( ) IF any SRM/IRM detector fails to retract, THEN attempt to move detector pe
| |
| _, Section 5.3.
| |
| 5.3 Failure of SRM/IRM Detector to Retract 5.3.1.
| |
| Bypass the failed detector to prevent the rod block and scram signals.
| |
| 5.3.2.
| |
| Attempt to withdraw the affected detector per Section 5.2 and observe the following:
| |
| 1.
| |
| DRIVING OUT, RETRACT PERMIT and POWER ON lights on.
| |
| 2.
| |
| SRM/lRM power level indications to verify movement of the detectors.
| |
| 5.3.3.
| |
| IF the detector can not be withdrawn, THEN perform the following:
| |
| 1.
| |
| At Aux Bldg,114 ft el, east, verify the following breakers ire ON:
| |
| * SCA PNL2Cl BKR 3, H22-P008 INSTRUMENT CKTS
| |
| * NHS MCC2C BKR ICT, H22 P008 SRM IRM DRIVE CONTROL CABINET l
| |
| ,,., 2.
| |
| Verify the following fuses are installed and not blown.
| |
| +
| |
| '' s
| |
| * C51 S001 A SRM A, Fuse No. Fl A, F2A, F3A
| |
| * C51-S001B SRM B, Fuse No. FlB, F2B, F3B
| |
| )
| |
| * C51-S001C SRM C, Fuse No. F1C, F2C, F3C
| |
| * C51-S001D SRM D, Fuse No. FID, F2D, F3D
| |
| + C51 S001E IRM A Fuse No. F1E, F2E, F3E
| |
| * C51 S001F 1RM B Fuse No. elf, F2F, F3F
| |
| = C51-S00101RM C, Fuse No. F1G, F2G, F3G
| |
| * C51-S001H 1RM D, Fuse No. F1H F2H, F3H t
| |
| * C51-S001J IRM E, Fuse No. F1J, F2J, F3J
| |
| = C51 S001K 1P31 F, Fuse No. FlK, F2K, F3K
| |
| = C51-S001L IRM G, Fuse No. F1L, F2L, F3L
| |
| ,7 g
| |
| )
| |
| a
| |
| * C51 S001M IRM H. Fuse No. F1M, F2M, F3M SOP-0074 REV-6 PACE 7 OF 35
| |
| : 3. At the SRM/lRM cabinet, turn off power to the SRM/lRM detector.
| |
| O 4. Declare the de-energized channel inoperable.
| |
| : 5. Notify Reactor Engineering.
| |
| 5.3.4. E the detector is successfully withdrawn, THEN return the BYPASS Switch to NORMAL.
| |
| 6 SYSTEM SHUTDOWN NOT" The Neutron Monitoring System is not normally shutdown completely.
| |
| 6.1 E portions of the system are required to be shutdown, THEN Refer To the ^11owing to deenergin system components:-
| |
| * Attachn.c,1, Electrical Lineup - Neutron Monitoring System
| |
| . Power supply listing in Steps 5.3.3.i and 5.3.3.2 7 REFERENCES I
| |
| t-7.1 3224.I10-000-002A 7.2 dE Drawing 828E235AA #001 through 007 7.3 CBD-SCA2Cl #001 and 002 7.4 EE-6BB 7.5 EE-ICC 8 RECORDS ,
| |
| 8.1 Record disposition shall be in accordance with ADM-0022, Conduct of Operations and ADM 0006, Control of Plant Records.
| |
| 1 SOP-0074 REV-6 PAGE 8 OF 35
| |
| | |
| g Esam Number NRC-SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 51 The reactor has been operating near rated power for 200 days. Which one of the following degribes the change in the indicated LPRM output signal from day 1 to day 200 and the method used to calibrate the LPRMs?
| |
| INDICATED LPRM POWER METilOD OF LPRM CALIBRATION
| |
| : a. Decreases Core lleat Balance
| |
| : b. Decreases TIP System Trace
| |
| : c. Increases Core lleat Balance
| |
| : d. Increases TIP System Trace ANSWER:
| |
| : b. Decreases TIP System Trace IDNO: LP# . OBJ #
| |
| 89 STM 503 9 O PROCEDURE NUMBER:
| |
| SOP-0074 OTHER:
| |
| LEVEL 2 I ~~NRC'KA: l RO: l "$ROi~
| |
| 215001 K101_ .j.
| |
| 1 2.5; . 2.8 216005 K1.13 1 2.6 3_
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 RO T2 G3 (TIPS)
| |
| SRO T2 03 (TIPS) 33
| |
| | |
| 1.0 SYSTEM DESCRIPTION 73 ]
| |
| 1.1 GENERAL INFORMATION 1.1.1 Purpose -
| |
| The Reactor Neutron Monitoring Systems are designed to provide the operator with a continuous indication of reactor power and the rate of power level changes from the condition of reactor shutdown to above full rated power. Due to the wide range of neutron flux which must be measured, three neutron monitoring systems are used: 1) Source Range Monitors (SRMs); 2)
| |
| Intermediate Range Monitors (IRMs); and 3) Power Range Monitors. The power range monitors are divided into two subsystems: Local Power Range Monitors (LPRMs) and Average Power Range Monitors (APRMs). The range of each of the neutron monitoring systems is shown in Figure 1.
| |
| The Transver:ing Incore Probe (TIP) System serves as a reference for calibrating the LPRMs to compensate for changes in detector sensitivity as the fissionable material in the detector is depleted.
| |
| This text discusses the principles of neutron detection used by each of these systems.
| |
| 1.1.2 Design Basis The Reactor Neutron Monitoring Systems were designed to provide the following
| |
| * SRMs provide neutron flux informraion during reactor startup and low flux level orerations.
| |
| The SRM is also used to monitor neutron flux level during n. fueling operations and initiates protective signals during fuel loading and physics testing.
| |
| e IRMs provide neutron flux information during reauar startup, heatup and power ascent. In addition to providing the operator with neutron flux indication, the IRM initiates trip signals to prevent damage to the fuel from abnormal operational transients while operating within the intermediate range of power.
| |
| * LPRMs provide signals proportional to local thermal neutron flux at various radial and axial locations within the reactor core. 'Ihe neutron flux signal developed by the LPRM system is utilized by the Plant Process Computer (PPC), the APRM System, and Rod Control and Information System (RC&IS) to ensure protection of the fuel cladding and aid the operator in evaluating the nuclear and thermal-hydraulic performance of the reactor core.
| |
| tO V
| |
| RBS-1-STM-GPST-A0503.00 Page 5 of 100
| |
| | |
| e APRMs provide signals representative of the average thermal power production of the reactor
| |
| ('
| |
| V] core. These signals are used by various plant protection systems (RC&lS and Reactor Protection System (RPS)) to prevent damage to reactor fuel in the event of unanticipated power transients. The APRM system also provides indication to the operator ofcore thermal power through control room recorders and the PPC.
| |
| TIP provides a means to calibrate the LPRMs to compensate for changes in detector sensitivity as the fissionable mater,al in the detector is depleted. Tips can also plot the axial flux distribution at any LPRM string location.
| |
| 1.2 PERFORMANCE CIIARACTERISTICS The Neutron Monitoring Systems, developed by the General Electric Company, were designed to meet the following requirements:
| |
| 1.2.1 Source Range Monitors
| |
| * Sensitivity Nominal: 1.2 x 10'' cps /nv Minimum: 5 x 10" cps /nv Maximum: 2.5 x 10'3 cps /nv n
| |
| * Neutron Flux U -
| |
| Operating Range: 1 x 102 to lx 10* ny nomina!
| |
| Storage Range: 5 x 10' nv maximum
| |
| . Operating Gamma Flux: 2.5 x 10' R/hr maximum
| |
| . Design Temperature: 575*F e Design Pressure Extemal to detector atmospheric Intemal to dry tube: atmospheric Extemal to dry tube: 1050 psig operation and 552*F 1250 psig design and 575*F 1375 psig maximum under upset conditions and 583 F The detector ic designed to be continuously retractable as the flux in the reactor is increased. It is possible to c c7 the detector at any point. The equipment is capable of monitoring the neutron flux in this manner to a level equivalent to 10 % reactor power.
| |
| h V
| |
| RBS-1-STM GPST-A0503.00 Page 6 of 100
| |
| | |
| i SRO EXAM KEY :
| |
| Esam Number NRC.SRO Rev. 0, Esamic.c NRC SRO!!XAM 32 Given the following conditions:
| |
| . RCIC is operating in the Test Retum Mode for IST testing. '
| |
| . The plant experiences a Station tilackout.
| |
| . RCIC is th:n manually initiated. .
| |
| Which of the following valves will stroke CLOSE under these conditions? ,
| |
| : n. RCIC Steam Supply Valve,l'04$.
| |
| : b. RCIC Steam Supply Inboard Isolation Valve T063.
| |
| : c. 7 CIC Test Return to CST Valve, l'0$9.
| |
| : d. RCIC Suppression Pool Suction Valve,0031. ,
| |
| ANSWER: ;
| |
| : c. RCIC Test Return to CST Valve,00$9. l II)NO: LPW OILI #
| |
| O 196 ilLO-017 5 PROCEDURE NUMBER: OTHER:
| |
| SOP 003$ LEVEL 4 AOP 0003 217000 K2.01 i 2.0 2.8 COMMENTS: (7/97 new HO T2 01 '
| |
| SRO T2 01 AL.50 ODJECTIVE 13 I
| |
| t r
| |
| | |
| NITACl!.%1ENT 2 PAGE 2 OF 6 ISOLATION VALVE CilECKOFF SilEET IVXATIIY IXVI&IV RINTORUY PANIl 190tATHY DIVH&III R1570REIY INTI1AIS OUlilOARD N11AIS N ilAIS N10ARD N11AIS lil3 P863 liVN MOV127 GROUPI liVN MOV129 SIGNALS INN MOV102 liVN MOV128 BD IIVN-MOVl30 liVR AOV125 N/A INR AOVl47 N/A liVR.AOV126 N/A liVR AOVl48 N/A lil3 P808 RCS MOV61A GROUPl RCS-MOV61B RCS MOV60A SIGNALS RCS-MOV60B RCS MOV59A B. D RCS-MOV59B
| |
| ^
| |
| RCS MOV58A RCS-MOV588 END ACB563 SOP 0049 END ACB583 SOP 0049 til3 P601 122-MOVR123 GROUP 1 SIGNAIS B,D lil3 P601 E51 F064 SOP 0035 GROUP 2 E51 F063 SOP 0035 E51.F031 SOP 0035 SIGNALS E51 F076 SOP 0035 IlQAS,T,V,X II13-P601 GROUP 3 E51 F077 SOP 0035 SIGNALS E51 F078 SOP 0035 BOTil D, R 1113 P655 til3 P654 E33 F007 l GROUP 4 E33 F028 E33 F008 E33 F027
| |
| , .{ SIGNAL DD O
| |
| AOP-0003 REV - 10 PAGE 13 OF 17 i
| |
| | |
| -~ ~ ~
| |
| PAGE 1 OF 2 ELECTRICAL LINEUP REACTOR CORE ISOLATION COOLING (SAFETY RELATED) i O EQUIPMENT NUMBER EQUIPMENT DESCRIPTION POWER SUPPLY REQ'D POSITION INITIALS ist 2nd E51 C00*, RCic TURB IRIP THROI ENB MCCI ON VALVE BKR1B RCic PUMP SUCT VV ENB MCCI E51 F010 CNDS STOR TK BKRIC~ ON RCic INJ SHUTOFF ENB MCCI E51 F013 VALVE BKR 4C ON RCIC MIN FLOW VV ENB MCCI E51 F019 SUPPR POOL BKRSB ON E51 F022 RCIC TEST FCV TO CNDS ENB MCCI ON STOR BKR 2B E51 F031 RCIC PUMP SUCT SUPP ENB MCCl ON POOL VALVE BKR 2C RCIC STEAM TO TURB ENB MCCI E51 F045 VALVE BKR 4B ON RCIC TEST RTN CNDS ENB MCCl E51 F059 STO" VALVE BKR 3C ON RCIC TURB EXH VV TO ENB MCCI E51 F068 SUPPR POOL BKRSC ON -
| |
| RCIC & RHR STEAM EHS MCC2D E51 MOVF063 SUPPLY VALVE BKR 3C ON RCIC FILL SUB SYS PUMP EHS MCC2E E51 C003 MOTOR BKR1F ON
| |
| ~
| |
| RCIC STM SPl,Y LINE O E51 F064 OUTBD ISOL VALVE (MST 121 FT EL, N.W. CORNER)
| |
| EHS MCC2L BKR4B ON RCIC STEAM SUPPLY EHS MCC2D ON E51 F076 BYPASS VALVE BKR4A ICS VaC BRK ISOL EHS MCC2O ON E51 MOVF077 VALVE OUT BKR 4C ICS VACUUM BREAKER EHS MCC2H E51 MOVF078 ISOLATION VALVE BKR2B ON E51 C002C BYS INVolB RELIABLE BYS SWOO1B CLOSED BUS INVERTER & BKR 529 E51 C002C P.S.
| |
| RCIC STEAM SUPPLY E51 AOVF054 DRAIN POT BYPASS RCIC EX'i DRAIN POT ENB PNL02A ON E51 AOVF005 ISOL BKR 6 RCIC STEAM SUPPLY E51 AO'VF026 DRAIN POT TURBINE SUPERVISORY N/A CONTROL ENB. LO2A N/A RELAY AND TRIP LOGIC ON ENB- 2A E5! K603 RCIC POWER SUPPLY ON O
| |
| SOP 0035 REV 14A PAGE 36 OF 40
| |
| | |
| SRO EXAM KEY O Esam Number NRC SRO Res. O Esam Title NRC SRO EXAJ 33 Following a valid ADS initletion, the operator is directed to close the ADS valves with the initiating (
| |
| signals still present. Which one of the follow ing operator actions will cause the ADS valves to ;
| |
| close?
| |
| : a. Place the control switches on lil3 P601 and fil3 P631 for the ADS valves to the
| |
| *0FF" position.
| |
| : b. Place the ADS inhibit switches on lil3P-601 to the
| |
| * NORMAL" position.
| |
| : c. Stop all low pressure ECCS pumps in both Div. I and Div. 2.
| |
| : d. Depress both ' ADS Time:/l.evel 3 Seal in Reset" pushbuttons, S13 A(11)
| |
| ANSWE
| |
| : d. i , >ress both " ADS Timer / Level 3 Seal in Reset" pushbuttons, S13 A(!!)
| |
| IDNO: 1.P W OILI #
| |
| 143 IILO-064 2 PROCEDURE NUMBER: OTHER:
| |
| SOP C311 LEVEL 3 HLO-064 [
| |
| ~
| |
| i NRC KAt l ROI l SRO:
| |
| 218000 K6 01 e 38 38 COMMENTS: 1/97 new RO T2 G1 SRO T2 G1
| |
| - ~
| |
| 1 33
| |
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| _ __ ._ =__.._. _ .__ _ ._._ _ ___ _ _ _ ____ _.- _ _ . _ _ _ _ _ _ . _ _ . _ _ _ .
| |
| i 3
| |
| SRO EXAM KEY !
| |
| O Esem Number NRC.SRO Mev. O Esam Title NRC SRO EXAM y- A Safety Relief Valve (SRV) tailpipe vacuum breaker w as failed in the open position w hen the ,
| |
| SRV opened. Which of the following is the result 7 ,
| |
| : a. Containment pressure increased.
| |
| : b. Steam bypassed the quenchers with a direct discharge path into the suppression pool.
| |
| : c. Drywell to containment differential pressure increased.
| |
| i
| |
| : d. Suppression pool water will be drawn up into the SRV discharge line after the SRV is closed.
| |
| ANSWER:
| |
| : c. Drywell to containment differential pressure increased.
| |
| LP# Onj #
| |
| II)NO: ,
| |
| 217 IILO-007 3 PROCEDURE NUMBER: OTHER:
| |
| O C Pat 0 S1B LEVEL 2 NRC KA l RO: l SRO: l
| |
| '223001 A2.09 4 3.4F 36 ,j COMMENTS: 7/97 new RO T2 01
| |
| $RO T2 01 O n
| |
| -. ..,.,r. .c_,e .,m,..;... ...m..p,,,_.,-,___.__.,,,_ ,m. ,,__.y . - _ _ - . _ _ , . , . , . _ . , , , . _ ~
| |
| | |
| Guard pipes aie used around the main steam lines and the m:in steam drain line piping run between the dr> well and the shield building. The function of the guard pipes is to prevent pressurizatiotrof the containment volume in the event of a steam line break by containing escaping steam and directing it back into the drywell(Figure 3).
| |
| IL Elbows Taps Each steam line contains an elbow tap between the vessel and the relief valves. These flow devices measure steam line flow by sensing the differential pressure in the pipe elbow. The measured flow is sent to the Feedwater Level Control System and to the control room for indication.
| |
| C. Safety Relief Valses 16 Safety Relief Valves (SRVs) are located between the Reactor Vessel and the inboard MSIVs.
| |
| Each SRV has a rated capacity of 925,000 lb/hr.
| |
| The discharge of each SRV is directed into the suppression pool between the ,lrywell and the containment, with the discharge line terminatimfbelow the minimum suppression pool operating water level (Figure 4). Discharge qttenchers direct the flow of steam so , not to impinge directly ,
| |
| on the floor of the suppression pool (Figures 5 and 6).
| |
| Two parallel check valves (vacumn breakers) on each SRV discharge line (tailpiece) relieve any vacuum following the closure of an SRV before any pipe damage can occur. As the residual p
| |
| d stearn condenses, the tailpipe pressure decreases. The vacuum breakers open to admit drywell atmosphere if condensing steam forms a vacuum. This prevents siphoning water up into the tailpiece. Water in the line more than a few feet above the suppression pool level would cause excessive pressure (jet forces) on the tailpipe and discharge quenchers uhen the valve is again operated. The vacuum breakers are designed to pass full flow (2 scfm) within 0.2 see of opening.
| |
| There are no alarms or direct indications of vacuum breaker position. However, if SRV's are leaking and vacuum breakers do not seal tightly, unidentified Drywell leakage increases.
| |
| Each SRV t Figure 7) is a spring loaded, scaled bonnet, angle globe valve with an externally attached pneumatic operating cylinder. The arrangement of the operating cylinder is such that a malfunction cannot prevent opening of the valve to satisfy the safety valve function. Each SRV can operate mechanically as a safety valve and automatically as a (electro pneumatic pressure relief valve). (Refer to LOTM 21-4).
| |
| The SRVs are divided into three groups, each group with a different relief setpoint. Table 2 lists the SRVs and their setpoints. At least one SRV on each main steamline can be actuated by the Automatic Depressunzation System (ADS). There are seven SRVs assigned to ADS (Figure 8).
| |
| As ADS valves, the SRVs operate in the same manner as in the pressure relief mode (pneumatically).
| |
| Q v
| |
| Each SRV has a control air accumulator which allows limited relief actuation if the air supply is lost. The control air supply line to each SRV accumulator contains a check valve to isolate it if supply air pressure is lost.
| |
| LOTM 24 6 page 4 of 26
| |
| | |
| . . - . - _ - - _ . . _ . . . . . . , - . . . , . - ~ _ . . - .
| |
| 10900010 AIR SUPPLY ACCUMULATOR p SOLENOIDS SHOWN A 8 k k # VENT
| |
| .... 3......
| |
| l RMS l RMS P601 ; P631 l
| |
| ' e PRESS ADS SIGNAL APPLICABLE
| |
| * ADS TO SEVEN (7) SRVS SMTCH ,
| |
| O
| |
| ~~
| |
| SRV ACOUSTIC MAIN STEAM LINE (MONITOR VACUUM BREAKERS N
| |
| SUPPRESSION POOL IJ I') '
| |
| 4 O
| |
| QUENCHER O
| |
| SRV ARRANGEMENT RBS 1-STM4PSTA0109 00 FIGURE 4
| |
| | |
| . - . - . . . .. . . . ._ ~ _ _
| |
| . _ _ - . ~ ~ . - . - _ . - - . - . - _ . - . - - - - . . . _ .- - - . . . _ .
| |
| SRO EXAM KEY Etam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM
| |
| $$ When EOP-4, Emergency RPV Depressurization, permits defeating isolation interlocks in order to rapidly depressurize without SRVs. Which of the following MSIV isolation signals may be '
| |
| bypassed?
| |
| : a. Only the RPV low level I signal.
| |
| : b. Only the RPV low level I and low main steam line pressure signal.
| |
| : c. All M51V isolation signals except for Main condenser low vacuum.
| |
| : d. All MSIV automatic isolation signals.
| |
| ANSWER:
| |
| : d. All MSIV automatic isolation signals.
| |
| IDNO: LPW OBJ #
| |
| 212 IILO-516 9 s PROCEDURE NUM3ER: OTHER:
| |
| EOP-0004 LEVEL 2 O EOP-005 ENCL 9 f NRC KAi l Ri): l SRO:
| |
| (223002 K4.08 i 3.3- 3.7 COMMLNTSt 7/07 new RO T2 G1 SRO T2 G1
| |
| | |
| ENCLOSURE 9 DEFEATING MSIVs M MSL DRAINS ISOLATION INTERLOCKS (3 l 1.0 PURPOSE To provide instructions for defesting isolations that are used to close the MSIVs and MSL drains. :
| |
| Placing switches in bypass prevents automatic closure of the MSIVs and MSL drains M will allow reopening of the MSIVs and MSL drains with any. isolation signal present. i 2.0 REOUIRED TOOLS EOUIPMENT 2.1 EOP 0005 ENCL 9 keys, four (4)*-
| |
| 3.0 3TRUCTIONS DEFEAT Containment Instrument Air isolation interlocks per EOP-0005 ENCLOSURE 16, to supply air to the inboard MSIVs. []
| |
| 3.2 OBTAIN EOP 0005 ENCL 9 keys, four(4). []
| |
| O O
| |
| 3.3 PLACE the following switches in the EMERGENCY position m VERIFY the red lights illuminate:
| |
| Switch No. Panel No.
| |
| 3.3.1 B21H S77A MSIV AND MSL DR ISOL INTLK BYP lH13*P691 ()
| |
| 3.3.2 B21H S77B MSIV AND MSL DRISOL INTLK BYP IH13*P692 []
| |
| 3.3.4 B21H S77C MSIV AND MSL DR ISOL INTLK BYP IH13*P693 ()
| |
| 3.3.5 B21H S77D MSIV AND MSL DR ISOL INTLK BYP JH13*P694 []
| |
| 3.4 OPEN MSIVs and MSL as directed by the CRS.
| |
| =
| |
| | |
| ==4.0 REFERENCES==
| |
| | |
| 4.1 GE Elem Diag 828E445AA ,
| |
| Nuc Steam Supply Shutoff System s
| |
| ENCLOSURE 9 PAGE 1 OF 3 l EOP 0005 REV. 9 PAGE 30 OF 115 i
| |
| | |
| -- - __. _ _ . . . ._.~ -._.-. . . - . __~-. . - -._. -- -.
| |
| I SRO EXAM KEY O Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM 3(, Select the statement that describes an operator action required for the following plant conditions:
| |
| Reactor power: 75 %
| |
| Suppression pooltemperature: 105 degrees F and rising Suppression poollevel: 19 feet 8 inches SRV lil21.F04711: failed open
| |
| : a. If the SRY cannot be closed within five minutes, place the resctor mode switch in !
| |
| SilU11X)WN.
| |
| : b. If suppression pool temperature exceeds 120 degrees F., aim and depress the manual scram pushbuttons.
| |
| : c. Place the reactor mode switch in SilUT!X)WN,
| |
| : d. Reduce suppression pool temperature to less than 100 degrees F within I hour.
| |
| ANSWER:
| |
| : c. Place the reactor mode switch in SilUTDOWN.
| |
| IDNO: LP# OBJ #
| |
| 32 IILO 538 9 PROCEDURE NUMBER: OTHER:
| |
| TRM 3 4 4 COND B LEVEL 3 AOP 0035 I NRC K'Ai ~ l R05 l' SROI l
| |
| '239002 A2.o3 1 4.1! 4.2 .l COMMENTSt 7/97 new RO T2 G1 SRO T2 01 (NOTE: looks and sounds iike Tech Specs are rmquired, but should know the 105 degree limet)
| |
| O ,
| |
| 1
| |
| | |
| t 9
| |
| NOTE Ifthe Reactor is scrammed a cooldoun rate ofgreater than 100 *F/ hour can resultfrom the SORV.
| |
| 1 S.5.3. IF the SRV can not be closed before Suppression Pool temperature reaches 10$'F, THEN place C71 A S1, REACTOR SYSTEM MODE SWITCH in SHUTDOWN and enter AOP-0001, Reactor Scram.
| |
| 5.6 WHEN the SRV is closed, THEN verify plant conditions stabilize.
| |
| 5.7 Refer To AOP-0007, Loss of Feedwater Heating.
| |
| 6 REFERENCES 6.1 SIL No.106, Rev. 2 Suppression Pool Temperature Monitoring And Control 6.2 EA EM 92-0579, Feedwater Temperature Reduction on Stuck Open Safety Relief Valve 4
| |
| O i
| |
| l l AOP-0035 REV-9 PAGE5OF8 l '
| |
| | |
| t 4
| |
| t i
| |
| SRO EXAM KEY I Ensas Number NRC SRO Rev. O Esam Title NRC SRO EXAM ;
| |
| 57 Consider the following plant conditions:
| |
| * l Reactor power: 45 %
| |
| Generator load: 410 MWe ,
| |
| Recirculation flow control: Loop Manual 1
| |
| . SELECT the plant response to a continuous runback of the load set demand signal to zero in the Electro llydraulic Control (EllC) system.
| |
| : a. Turbine control valves (TCVs) throttle closed,1 bypass valves (BPVs) remain closed, reactor pressure increases, reactor scrams on high pressure or high neutron flux. ;
| |
| : b. Bypass valves (BPVs) throttle open, reactor pressure decreases, MSIVs isolate on low steam line pressure, l reactor serams on the MSIV closure, P
| |
| : c. Turbine control valves (TCVs) throttle closed.
| |
| bypass valves (BPVs) throttle open to compensate, once the BPVs are fully open reactor pressure increases causirg a scram on high pressure or neutron flux.
| |
| : d. Turbine control velves (TCVs) throttle closed, bypass valves (BPW) throttle open, reactor pressure temains fairly constant, reactor power increases slightly due to reduced feedwater heating.
| |
| . ANSWER:
| |
| : c. Turbine control valves (TCVs) throttle closed, bypass valves (BpVs) throttle open to compensate, once the BpVs are fully open re,ictor pressure increases causing a OBJ #
| |
| 4 IDNO: LP #
| |
| 28 IILO-059 7
| |
| # PROCEDURE NUMBER: OTHER:
| |
| SOP-00a0 LEVEL 4 F NRC KA: l RO: SMO[
| |
| 3
| |
| ?241000 K4.02 ' 3.3' 3.3 _
| |
| h I
| |
| h i
| |
| ~,- - , . , - , - , . - . , , , - - _ , - , - - - , , , , , - , - , , , , , ,, . , . . ~ , , - - - . , , . - ~ - - . - - - - - , , , - . - , ~ - , , --l-~. .-
| |
| | |
| I I
| |
| ; f
| |
| .I .
| |
| SRO EXAM KEY Esem Number NRC SRO Rev. 0 - Esam Title NRC SRO EXAM. j COMMENTS: 7/97 rww RO T2 01 SRO T2 01 !
| |
| i l
| |
| . i i >
| |
| I h
| |
| i I
| |
| d .
| |
| t P
| |
| 1 r
| |
| 5 t
| |
| i I
| |
| I d
| |
| f 5
| |
| l i.
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| r l
| |
| l l
| |
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| |
| 60 I i
| |
| _ . .. . . . _ . . . _ - . . ._.._..U__...... - - - , - ~ . . . , . . . ~ _ - . . . _ _ _ , _ _ _ _ _ . . _ _ _ _ _ _ , . _ . - . - _ - - _ .
| |
| | |
| In standby control, the load limit circuit is removed from the EllC circuitry and precautions should be taken to avoid overloading the turbine generator.
| |
| D. System Malfunctions
| |
| : 1. Load Set Runback A " runback" refers to repositioning of the load set motor operated potentiometer to reduce the load reference signal to the speed / load summer.
| |
| One condition that causes a load set runback is 1800 RPM (or synchronous speed) not selected on the SPEED SET section of the Turb.ne Control Panel. Loading the turbine at lower speed could cause turbine damage t y putting excessive stress on the last stage buckets of the turbine, to prevent this from happening with the generator on line, an interlock prevents selecting any speeds other than 1800 RPM. If, however, the
| |
| ' speed select circuitry fails in such a manner as to cause 1800 RPM to become deselected with the generator on line, a load set runback will occur.
| |
| O Power / Load Unbalance (PLU) also causes a runback of the load set pot. If turbine power is above 40% and the comparison of generator load (sensed by stator current) to turbine power (sensed via llP turbine exhaust pressure) > ields a difference of greater than or equal to 40% (Turbine Power:2 Generator Load), the Power / Load Unbalance circuit runs the load set motor back towards 2% until the condition clears. This circuit also initiates fast closure of the Turbine Control Valves and the control valve fast closure initiates a reactor scram. The load set runback and turbine control valve fast closure prevent the drastic overspeed that could accompany a load rejection of this magnitude.
| |
| If the PLU circuit loses the intermediate (cross around) pressure signal to its circuitry, the circuit is disabled and a light, PWR PRESS SIG LOST, illuminates on the Turbine Test Panel.
| |
| Loss of the Stator Coeling System (when generator Load >24.5%) as sensed by high stator cooling water exit temperature or low stator cooling water pressure or low stator cooling water flow causes a runback of the load set pot. If the output of the generator is not at or below 80% load within 2 minutes or less than 24.5% in 3.5 minutes after the stator cooling A
| |
| LOTM 27-9 Page 7s of Y6
| |
| | |
| t loss is sensed, the turbine will trip. The 24.5% limit is based on the heat removal capability of the generator with no stator cooling (i.e. self cooled).
| |
| : 2. Pressure Regulator Failure The SB and PR pressure regulators can fail in three different ways: 1) high (increasing valve position demand),2) low (decreasing valve position demand), and 3) "as is" (constant valve position demand). In addition, if a pressure regulator is placed in Test, the fault detection logic for the SB and PR system treats it as a failed regulator and does not transfer to that regulator in the event of a fault in the operating regulator.
| |
| If a regulator fails high or low, operation of the turbine is unaffected because the fault detection circuits of the SB and PR system switch regulation to the non. failed regulator whenever a fault of sufficient magnitude is detected. The ROLTR ERROR light on the SB and PR panel
| |
| . illuminates. In addition, if the failed regulator was the one initially in operation, the select light does not agree with the regulator in service.
| |
| As stated previously, if one regulator is in test the fault detection circuits do not allow transfer to that regulator. If the operating regulator fails low, with the non-operating regulator in test, it will decrease the signal to the
| |
| 'N control valves to point where they close. A scram results from the (d pressure increase due to high neutron flux or high pressure unless action is taken to prevent it.
| |
| Alternately, if the operating regulator fails high with the non-operating regulator in test, the load limit prevents the output signal of the Pressure / Load LVG from exceeding its setpoint. The bypass valves open to pass the remainder of the steam which the pressure regulator is calling for (limited by the Maximum Combined Flow Limiter Setting). The plant depressurizes due to excess steam flow until the MSIV's close on low steamline pressure (Reactor Mode Switch in RUN).
| |
| If the selected pressure regulator fails "as is", there is no immediate etTect on the plant if the plant is in steady state operations. That is the real danger of this malfunction. There are no alarms, the plant does not deviate from its present state and therefore the operator is unaware that a failure has occurred. Ilowever if Rx power is changed, pressure regulator control will swap to the standby regulator when a preset deviation between the two regulators has been exceeded.
| |
| 1.OTM 27 9 Page 52 of 76
| |
| | |
| {N {% [\
| |
| rROneSPEED W2N m s'Em 8+AS 1 300061 2U SELECTOR +
| |
| + esTERCEPT SPEE D + .
| |
| '" l y g l "
| |
| watwE (mf SELECT
| |
| $ $E g Q *. gg ,,,,g
| |
| + - l l "+
| |
| TuRem SPEED 7. [
| |
| -yK m .E n
| |
| [ Sa> RATE
| |
| * 25 "D*
| |
| + g % LD
| |
| = ~E wT LVG ' SPEED E RROR 4 4 3 -
| |
| _1 al (O c REF y TUR98sE SPEED COssTROL ,,4 i U f " " " %F -- I
| |
| [ e POwE*tttOADLPsanLANCED
| |
| ~
| |
| SE + 0- ! vesC ED aso? SELECTED TER wgw t .. : usuaCE flout DE MAND E RROR RECSC COssT 1 M TO REACTOR RECW. NSS SP N ADAr$TER 10 % WC SsGsent FENORWARD l @ EOe4ATCwm
| |
| + o e GAASTE R OWSERS
| |
| = PRE SS SP , o RATE g3, p ,__, , gg ,g ,$ _
| |
| ROW OEMaso RECIRC CossT tanTsu TOOTwR , l E5W WC ,,
| |
| mt P" cart (AasP
| |
| + ' '+ +
| |
| SEtF CCesTROL cwCuss HVG K LVG M vAtwe BUHERS
| |
| ]t
| |
| ; DE MANO TTLE P E - ---> SLRC PE PRESSURE A g STEAM t esE LV[G ,
| |
| I
| |
| . RESOfeANCE (SLRC) * #
| |
| k COesPEWSATOR OPE.,
| |
| \a 1P g naaJt
| |
| * 7AULT TOAD L estY TO BYPASS & CND E TECTOP CIPCutTS + FLOW LOGsC
| |
| ~*
| |
| l gy + - BUFFER PRESSURE SET tmT ,,gggy ,E NS g dL FtOveLesT "
| |
| ~
| |
| t mTER I DEAD 8AMO Jk DEC 3r-TO SPV gyp ,$g 7 y y ,gyg
| |
| ,fjG LvG 1r DE n4AMO 1, SneAtt ag-STEAns gp CLOSE TNROTTt E :[E _ --> StpC LVG saS _ l 4' PRES $URE B\ 4+ PE STEAAA tlNE _
| |
| evPASS JACR esAA COass
| |
| + y 6
| |
| I-[
| |
| l ,
| |
| PESONANCE mPC) row testT f pong g g ''
| |
| f ROas SE TPOWT COnsPE *sSATOR f MT DE T g
| |
| ADJUST CIRCUtf CIRCurf , 37 4
| |
| : EHC & SB/PR SMatt I
| |
| 'vS C04 "sSER wACuuu l LOTM-27 "
| |
| i ass _1.s1uPsmse w MAJOR SUBSYSTEMS LS PASS A cR nouacu
| |
| | |
| \
| |
| SRO EXAM-KEY :
| |
| F. sam Number NRC SRO Hev. O I: sam Title NRC SRO I:XAM [
| |
| t 1
| |
| 3 A plant startup is in progress. Reactor power is being held at 1% power for the 900 psig Dryw ell w alldown when the Startup Feedwater Reg. Valve drifts fully open.
| |
| i t
| |
| ' Which of the following actions / signals will occur as a result of this failure?
| |
| (NO111: Assume no operator action.)
| |
| : a. Reactor scrams on high reactor water level.
| |
| : b. Reactor feedwater pumps trip on high reactor level.
| |
| : c. Reactor water level remains unchanged due to compensation by the Long Cycle i
| |
| Cleanup Level Controller (CNM 104).
| |
| : d. - Reactor water level stabilizes at a new higher level.
| |
| ANSWl:Rf
| |
| : b. Reactor feedwater pumps trip on high reactor level. ,
| |
| IDNO: 1.P# OILI #
| |
| 348 STM.107 10 PROCEDURE NUMBER: OTHER:
| |
| AOP 0000 LEVEL 3 SOP-0009 259002 K3.02 1 3,71 3.7
| |
| [259002 K3 01 l' "38 '38 COMMENTS: 7/97 new LOf M 34 6 p.11 of 13.
| |
| LOTM 3 4, p. 26 of 31, Tat >le 8 RO T2 G1 SRO T2 01 61
| |
| | |
| l I
| |
| 1 60100012 s
| |
| k nEACT0AltWL llC STEAM FLO :, IlC FEEDWAftR Ft0W ?-l
| |
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| A b C D P MO 4 Eih4 heatC etet Id d U M D'd I t.A 11 1 gg a3%%%5"
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| autvyr,,
| |
| V Y ' '
| |
| wm u ,cm.,. L ,e , cm., .
| |
| RSS-1 STM<,PST A0107.00 FIGURE 1 FEEDWATER LEVEL CONTROL SYSTEM
| |
| | |
| -t b
| |
| SRO EXAM KEY l F. sam Number NRC SRO Rev. O Esem Title NRC SRO EX %M 59 Olven the following plent conditions: ,
| |
| p A LOCA has occurred. '
| |
| Reactor water level is .50 inches.
| |
| Drywell Pressure is 1.12 psid.
| |
| All radiation monitors are indicating normal for plant conditions.
| |
| "A" Standby Oas Treatment System (50 is)is running and the "It" SilOT system has been shutdown and placed in STANDilY.
| |
| Which of the following will be the status of Sil0T systems "A" and ''11" five (5) minutes after ENS- l SWO! A receives a degraded !!us Undervoltage Signal? (All associated systems respond normally)
| |
| : a. Iloth systems are shutdown.
| |
| : b. Iloth systems are running.
| |
| : c. "A" system is shutdown and *II" system is running.
| |
| A "A" s) stem is running and *II" system is shutdown.
| |
| ANSWER:
| |
| : b. Iloth systems are running.
| |
| IDNO: LI' # Olti #
| |
| 272 IILO-033 12 PROCEDURE NUMBER: OTHER:
| |
| SOP-0043 LEVEL 4 SBGT DWGS
| |
| ! NRC KAI l RO: l~SROI
| |
| !201000 K6 01 1 29 3 COMMENTS: 7/97 new RO T2 Gt SRO T2 Gt ~
| |
| (to see for D/G,30 sec for sequencing re eng from D/G LOCA still present B slu due to low flow in Al
| |
| ~~
| |
| O c
| |
| , , , , , ,, ,nr-- -w- - ,- , -~ ,
| |
| | |
| l l PURPOSE i l
| |
| I Provide instructions for operation of the Standby Gas Treatment (OTS) System.
| |
| l l
| |
| 2 PRECAUTIONS AND LIS11TATIONS 2.1 When Standby Gas Treatment is aligned with suction from the Auxiliary Building, personnel accer4 may be difficult due to the excessive vacuum drawn. OTS AOD22A(D), SGT FILTER A(B) RECIRC may be opened to allow personnel access.
| |
| 2.2 With an Auto Start signal present for GTS FN1 A(B) SGT EXH FAN A(B) or HVR FN11 A(B), ANNULUS MIXING FAN, the fan which is stopped will automatically restart upon receiving a low flow signal from the running train, whether abnormal conditions still exist or not. When !
| |
| auto start signals are cleared, before stopping the n' rating fan, it is necessary to press the STOP Pushbutton of k non running fan to prevent an unnecessary automatic restart of that fan.
| |
| (d 2.3 All controls and indications are located on Control Room Panel lil3 P863 unless otherwise indicated.
| |
| 3 PREREOUISITES FOR STARTUP AND OPERAT1QE 3.1 Check the following systems are in operation:
| |
| 3.1.1. Fire Water Protection Systern per SOP 0037 Fire Protection -
| |
| Water System.
| |
| 3.1.2. Instrument Air System per SOP-0022, Instrument Air System.
| |
| 3.1.3. Containment HVAC per SOP 0059, Containment Building HVAC.
| |
| 3.1.4. Auxiliary Building HVAC per SOP 0065, Auxiliary Building HVAC.
| |
| 3.1.5. Floor and Equipment Drains System per SOP-0104, Floor &
| |
| Equipment Drains.
| |
| SOP-0043 REV - 8C PAGE 3 OF 20
| |
| | |
| ATTAC115 NTI h3I PAG JF3 DIVISION I AND II TIM SEQUENCED LOND GUIDE EQUIPMENT DESCRIPTION DIVISION I DIVISION II SEQUENCE KW RATING COMPONENT COMPONENT NUMBER NUMBER DIESEL EXCITER CLG FAN IIVP-FN6A r;'/P-FN6B BKR 1.58/2.6 CLOSURE DIESEL FUEL TRANSF PUMP EGF-PI A EGF-PIB BKR 0.61 CLOSURE LPCS PI"4P E21-C001 N/A 12 SEC 917.2 RIIR A/B PliMP E12-C002A E12-C002B 17 SEC 509.2 R11R " SUMP N/A E12-C002C 12 SEC 509.2 STFLlY 7AS1REAT FAN GTS-FN1 A GTS-FNIB 30 SEC 41.6 STL GASTREATllTR GTS-IIIA GTS-lilB 30 SEC 85 ANNULUS MXG FAN IIVR-FNil A IIVR-FNIIB 34 SEC 109 DIESEL ROOM VENT FAN HVP-FN2A IIVP-FN2B 34 SEC 73.2 STBY SERV WTR PUMP SWP-P2A SWP-P2D 70 SEC 328.5 STBY SERV WTR PUMP N/A SWP-P2B 40 SEC 328.5 CONTMT UNIT COOLER IIVR-UCIA IIVR-UCIB 610 SEC 68 LEAKAGE CONTROL AIR COMPR LSV-C3A LSV-C3B 600 SEC 43 NOTE Heaters cycle on andofperiodically.
| |
| CONTROL ROOM HEATER IIVC-Clil A IIVC-ClliB UKR 65 CLOSURE 1
| |
| AOP-0004 REV - 14 PAGE 21 OF 43
| |
| | |
| (3 wJ SRO EXAM KEY Esam Number NRC-SRO Rev. O Esam Title NRC SRO EXAM 60 An MSIV closure resuhed in a reactor scram. The pressure transient caused a small steam leak in the drywell. The following conditions exist:
| |
| . Reactor pressure is at 900 psig.
| |
| Reactor Levelis at 80 inches wide range
| |
| - Drywcil pressure is 2.1 psid
| |
| . Containment pressure is 0.3 psig
| |
| . Lowest recorded ENS'SWG1 A Bus voltage was 3952 volts.
| |
| Which one of the following would be in service as indicated?
| |
| - (NO OPERATOR ACTION TAKEN)
| |
| : a. DIV I D!G running unloaded.
| |
| : b. DIV 11 SSW with flow through the *B" Containment Unit Cooler.
| |
| : c. - Drywell units coolers running with no cooling flow.
| |
| : d. LPCS injecting to the RPV.
| |
| ANSWER:
| |
| : a. DIV I D'G runnmg unloaded.
| |
| II)NO: LP# OBJ #
| |
| 261 IILO-037 4 PROCEDURE NUMBER: OTHER:
| |
| SOP @53 LEVEL 4 ARP-877 32-H03 f 262001 K1.01 8 3.8 4.3 j COMMENTS: ~b97 new RO T2 02 SRO T2 01 THIS OUESTION IS TO DETERMINE IF THE CANDIDATE UNDERSTANDS A SEQUENCE OF EVENTS IN A LOCA.
| |
| | |
| 4 ALARM NO. 0214 ENS *SWGlB IH13*P877 ! 32A i H03 SUSTAINED OR !
| |
| DEGRADED UV INITIATING DEVICES SET POINTS e
| |
| : 1. 62.lX lEGSB03 1 <72''. bus voltage for 3 Sec
| |
| : 2. 62 6X lEGSB04 2. <91% bus voltage for 3 Sec
| |
| : 3. 62 2X lEGSB03 3 <91% bus voltage for 60 See NOTE Auto actions will occur at less than 72% for 3 Sec. QB less than 91%
| |
| for 3 Sec. with o LOCA QB less than 91r'.for 60 Sec.
| |
| AUTOMATIC ACTIONS s
| |
| : 1. Breaker IENS*ACB24 trips.
| |
| : 2. Breaker IENS*ACB26 trips.
| |
| : 3. All motor loads on Bus IENS'SWGlB tnp.
| |
| : 4. LEGS *EGlB STANDBY GENERATOR auto starts, comes to rated speed and voltage, and ties to Bus LENS *SWGIB by the auto closure of breaker IENS* ACB27
| |
| : 5. One or more of the followmg alarms actuates on panel IHl3*P877
| |
| : a. 4160V Standby Bus Distr Breakers inoperative
| |
| : b. 4160V Standby Bus Distr Breakers Auto Trip.
| |
| : c. 480V Standby Bus Distr Breakers Auto Trip.
| |
| : d. 480V Standby Bus IB or 2B Sustained Undervoltage.
| |
| : e. Diesel Generator IB Prot Circuit Loss of Control Power.
| |
| : f. . 4160V Standby Bus IB PT Fuse Blowout.
| |
| , g. DG Room B Emergency Exhaust Fan Trouble.
| |
| 6- Safety related automatic load sequence for Bus IENS'SWGIB commences O
| |
| . ARP 877 32 REV 6B PAGE 45 OF 49
| |
| | |
| _ . . _ - . _ . . _ _ -_ - ~ _ . . - _ - _ . .
| |
| i SRO EXAM KEY O Esam Number NRC-SRO Rey, O Esam Title NRC SRO EXAM 61 The following conditions exist:
| |
| l The Div 2 standby dietel gene.ator is loaded and in parallel with bus IENS'SWGIB through the :
| |
| riormal breaker. A LOCA signal occurs. l Which of the following describes the efree of the standby diesel generator and bus IENS'SWGID? I
| |
| : a. The normal bus supply breaker M vin , d the diesel generator will supply bus loads,
| |
| : b. The normal bus supply breaker anc %re, generator output breaker will open, then after loads are stripped, the diesel ger.crator output breaker will reclose ,
| |
| : c. The diesel generator output breaker will open and cannot be closed as long as bus voltage is supplied by the nonnal or alternate feeders until the LOCA signal is reset.
| |
| : d. The diesel generator output breaker will remain closed in parallel operation with the bus.
| |
| ANSWER:
| |
| : c. 1he diesel generator output breaker will open and canr.at be closed as long as bus voltage is supplied by the normal or alternate feeders until the LOCA signal is reset IDNO: LP# OBJ #
| |
| 199 IILO 037 4 PROCEDURE NUMBER: OTHER:
| |
| SOP @53 LEVEL 4 ,
| |
| 1264000 A7.10 1 3.9' 4.2 COMMENTS: 7/97 new RO T2 01 SRO T2 Ot 63
| |
| | |
| IC ') 2.8 Do not operate the diesel generator without air pressure. All shutdown functions are inhibited with a loss of control air to the pneumatic logic. Following control air to the control panel, when air pressure meiswillrestored, start if the e in the standby mode. !
| |
| 2.9 If the diesel is run for one hour or greater, check and dmin from the day tank any ax.znulated water via IEGF*Vil(V41) DAY TANK 1K2A(2B) DRAIN. Refer to ,
| |
| 1 fechnical Specification 3.8.1.1 and 3.8.1.2.
| |
| 2.10 Lube oil must be added only through the fill connection on the sump. Do not overfill the sump. (Ref. 7.35) 2.11 If IEGS*EGI A(B) STANDBY DIESEL GENERATOR is declared inopemble, refer to Technical Specification 3.6.1.1 and 3.8.1.2.
| |
| 2.12 Never have 2 synchroscopes in the same division on at the same time.
| |
| 2.13 If the diesel generator is paralleled with the standby bus nonnal or altemate breaker and a LOCA signal occurs, the diesel cenerator output bn:aker will open.1he diesel genemtor breaker cannot be closed as fong as bus voltage is being supplied by the normal or attemate supply and the LOCA signal still exists.
| |
| 2.14 Sustained operation of the eng)ine at critical speeds of 190, 285, 350 an nould be avoided. (Ref. 7.25 m
| |
| If a diesel start signal is activated while the diesel is not available, the signal will I) 2.15 remain scaled in. 'If the diesel is thea made available, the diesel engine will auto start. To_ prevent this, before retuming the diesel to Operational Mode, the EMERGENCY START RESET switch on lH13*P877 should be depressed.
| |
| 2.16 Short dumtion runs and light load (less than 20%) operation shottld be avoided.
| |
| When hour.Ref. (possible, 7.26, 7.36) the diesel should be loaded to greater than 2300 KW 1.17 To minimize crankshaft torsional stresses, continuous engine operation at critical speeds shall not be allowed. Minimize the time the engine is opemted between 453 and 457 RPM (60.4 to 60.9 Hz). (Ref. 7.25) 2.18 Encine cylinder exhaust gas temperatu.e should be within 50 F of the average for all cylinders. Any cylinder temperature exceeding this limit should be investigated by mamtenance.
| |
| 2.19 Pirlube of the engine should be performed before all non-emergency starts. (Ref.
| |
| 7.32) p V
| |
| SOP-0053 REV - 16B PAGE 5 OF 88
| |
| | |
| ._ - ._ ._.- __ ___ ~ . ___ _. _ _ _ ._.. ~ _.._._ _ _ - .__-._._ _-_...._ . - _ . _
| |
| l SRO EXAM KEY Q Esam Number NRC-SRO Rev. O Exam Title b'RC SRO EXAM 62 Which one of the following maintains a negative pressure in the annulus following a LOCA7
| |
| : s. SinGT system starts and ta'(es a suction on the Annulus Pressure Control System.
| |
| : b. Annutes mixing fans start and discharge to the Annulus Pressure Control System.
| |
| c Annulus mixing fans start and discharge to the SilGT system. r
| |
| : d. SilGT system starts and takes a suction directly from the Annulus. i ANSWERt
| |
| - c. Annulus mixing fans start and discharge to the SilOTS.
| |
| IDNO: LP# OILI #
| |
| 82 IILO-038 2d PROCEDURE NUMBER: OTHER:
| |
| SOP-0059 LEVEL 2 SOP-0043 ,
| |
| t
| |
| !?90001 K1.04 i 3.7i 3. 9 _.. ,
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 01 i 66 l
| |
| . - - = - . - . - . , , . . , - - . . _ _ _ - . - - - - . - . -
| |
| | |
| TECHNIOUES VI. SUBJECT MATTER e System Shutdown (6.0) ;
| |
| * APCS (6.1) 6.4 Abnormal Operation
| |
| ~
| |
| 6.4.1 Loss of Offsite Power ,
| |
| e Loss of offsite pcwer will cause both APCS fans to lose power.
| |
| * AOD67A(B) will auto close when the associated fan de-energizes.
| |
| * AOD23 A(B),161,142 snd 261 will auto close on low flow after a time delay to allow fan start.
| |
| OBJ-#4e e To restore the fans to service, the associated dampers (AOD23 A(B), 262,142, 261) must be re-opened and the fan control switches must be taken first to STOP and then to
| |
| ' START or AUTO.
| |
| 64.2 Loss of Coolant Accident (LOCA) e The APCS response to a LOCA is the same as for a loss of offsite power.
| |
| -TP-05 6.4.3 Annulusliigh Radiation OBJ #6c, 11a e Iligh-liigh radiation n the annulus as sensed by ,
| |
| RE-I l A(B) will cause the fb . wing to occur:
| |
| * AOD-23 A(B),161,14e and 261 will auto close.
| |
| . AOD-22A(B) will auto open OBJ #3a e Annulus Mixing Fans FN11 A(B) will auto start
| |
| . The SGTS will auto start OBJ #7 64.4 Alignment of the annulus to the SGTS can be accomplished using the ANNULUS MIX SPLY TO SGT AOD 22A(B) RESET / MANUAL INITIATE switches located on P863.
| |
| \
| |
| ULO 038-4 PAGE 18 of 27 9- , y 'e#1 FW --v're-e w+ 4 -- --- +e-- m -w f M c- - --+- 4 e'~r
| |
| | |
| IECHNIOUES VI. SUBJECT MATTER e Placing switch in $ianual Initiate position will cause the following to occur:
| |
| * AOD-161,142 and 261 shut
| |
| . The APCS fans auto stop
| |
| * The Annulus Mixing fans (FN11 A,B) auto start.
| |
| e Dampers AOD-22A,B open.
| |
| 6.4.5 INSTRUCTOR NOTE: Ensure students understand that these switches merely align the annulus to the SGTS. They do not directly auto initiate the SGTS, however SGTS initiates on APCS low flow due to the damper isolation,
| |
| : 7. Annulus Mixing System 7.1 Description 7.1.1 The Annulus Mixing System is designed to perform two functions:
| |
| * To provide thorough mixing of any leakage ofiodine and noble gases from the Containment.
| |
| . To maintain the annulus at a negative pressure of 0.50 inches W.G. by exhausting to the SGTS following a LOCA.
| |
| 7.1.2 The sptem is provided with (2) 100 percent centrifugal fans with associated dampers and ductwork.
| |
| OBJ #2b 7.1.3 The duct work takes suction on the annulus at a lower elevation and discharges at a higher elevation.
| |
| OBJ #3b 7.1.4 Annulus Mixing Fan, lHVR*FN11 A(B), is automatically started when all of the following conditions are present.
| |
| * Breaker static inverse time and instantaneous overcurrent reset
| |
| * LOCKOUT switch ir RESET positwa e Load sequence start permissive and e manual or automatic containment isolation g signal or CJLO O38-4 PAGE 19 of 27
| |
| | |
| . _ .__. _ _ _ _ __ _ __ ~_ _ _ . . _ _
| |
| TECHNIOUES VI. SUBJECT MATTER e annulus exhaust radiation high-high or o' APCS system air flow low.
| |
| * Fan lHVR*FNI 1 A(B) has not been stopped with the above auto start signals present. (FN11 A(B) will auto restart with low flow on FN11B(A) discharge flow switch when stopped with an auto stan signal present) 7.2 Controls and Indications OBJ #7 INSTRUCTOR NOTE: Utilize Table 4, Controls and Indications (LOThi Chapter 63) and the indicated transparencies to cover the controls available at panel P863 in the hiain Control Room.
| |
| 7.3 Noxal System Operation During Normal Plant Operation, the system is not operating:
| |
| 7.4 Operating Procedures INSTRUCTOR NOTE: Utilize the following procedures to cover the indicated steps:
| |
| J #8 7,4.1 SOP-0059, Containment HVAC System Precautions and Limitations (2.0)
| |
| . Prerequisites for Startup and Operation (3.0)
| |
| * System Startup (4.0)
| |
| * Annulus hiixing System (4.4)
| |
| . System Operation (5.0) ,
| |
| e Annulus hiixing System (5.4)
| |
| * System Shutdown (6.0)
| |
| . Annulus hiixing System (6.4) 7.5 Abnormal Operation 7.5,1 Loss of Offsite Power
| |
| . Loss of offsite power will cause a secondary containment isolation signal and low flow condition in the APCS.
| |
| v iL O M 8 PAGE 20 of 27 I-
| |
| | |
| TECHNIOUES VI. SUBJECT M/.TTER BJ #3a e The Annulus Mixing Fans will both AUTO START aner 30 second load sequencing time delays.
| |
| * Dampers AOD53 A(B) will open when the '
| |
| respective fans start.
| |
| * The SGTS will auto start.
| |
| * AOD 22A(B) will auto open.
| |
| . The Annulus Mixing System will exhaust to the SGTS to maintain the annulus negative pressure.
| |
| 7.6 Loss of Coolant Accident (LOCA) / Annulus high radiation /
| |
| APCS low flow OBJ #4a e This system response is the same as for a loss of offsite power except there are no load sequence time delays.
| |
| : 8. Containment and Drywell Purge System TP-06 8.1 Description 8.1.1 This system is designed to purge either the Containment or Drywell.
| |
| * During normal operation only the Containment is purged.
| |
| * This may be done to maintain radionuclide concentrations in the Containment less than 25 percent of maximum allowable concentrations, control containment pressure, control containment atmospheric contaminants, or for surveillance testing.
| |
| * The Drywell is purged to reduce airborne levels in the Drywell during shutdown conditions and to eliminate atmospheric contaminants.
| |
| 8.1.2 The system has several modes of operation:
| |
| * Containment Purge to Plant Stack ,
| |
| o Containment Recirculation OBJ #11a- e Containment Purge to SGTS
| |
| * Drywell Purge to Plant Stack SILO 038-4 PAGE 21 of 27 l
| |
| | |
| SRO EXAM KEY O Exam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM 63 The plant is operating at 100% power. Both Recirc Flow Control Valves are in Flux Manual (Loop Auto) at 67% valve position, A leak in the Drywell has caused Drywell Pressure to increase to approximately 1.75 psid. Following the high drywell pressure signal, the "B" Reactor Feed Pump Trips and level decreases to + 14.5 inches and stabilizes. Which of the following describes the response of the Recirc Flow Control Valve?
| |
| Flow Control Valves will:
| |
| : a. runback to 22 % valve position.
| |
| : b. go to " min" position.
| |
| : c. move to a position to provide 60 % core flow,
| |
| : d. remain at 67 % valve position.
| |
| ANSWER:
| |
| : d. remain at 67 % valve position.
| |
| II)NO: LP# OILI #
| |
| O 244 IILO-005 4 OTHER:
| |
| PROCEDURE NUMBER:
| |
| 4
| |
| - ARP P680-4-803 LEVEL 4 ARP-P680-4 809 ARP-P680-4-C04 ARP-P6804010 (202002 A2.08 3.3: 3.3 ..l COMMENTS: i 7/97 new RO T2 G2 SRO T2 G1 O
| |
| .d 67
| |
| | |
| FCV A MOTION INHIBIT
| |
| .LARM NO. 2053 IH13*P680/04A/B03 INITIATING DEVICES SETPOINTS
| |
| : 1. Logic Relay IB33 Kl63A 1. N/A AUTOMATIC ACTIONS
| |
| : 1. IB33 F060A LOOP "A" FLOW CONTROL VALVE locked in position.
| |
| : 2. Both subloop pumps on HPU A shutdown.
| |
| OPERATOR ACTIONS
| |
| : 1. Verify Automatic Actions have taken place. -
| |
| : 2. Adjust Loop B flow in manual to maintain reactor power and flow within flow map constraints.
| |
| : 3. Check drywell pressure and enter Emergency Procedures if necessary.
| |
| LONG TERM ACTIONS
| |
| : 1. Check the subloop control panel on lH13 P614 to determine if the problem is mechanical,
| |
| : a. Mechanical problems mdicated if the " MAINTENANCE" pushbuttons are ON on both pumps accompanied by one of the following status lights:
| |
| : 1) TANK EMPTY
| |
| { ) 2) HOT OIL
| |
| : 3) OVERCURRENT
| |
| : 4) UNDERVOLTAGE
| |
| : 5) PRESSURIZED light OFF
| |
| : 2. Maintain reactor power and flow stable using manual control of Loop B FCV.
| |
| : 3. Ifinhibit is due to excessive rate of changes in valve position, reset B33-K637A in H13-P634,
| |
| : 4. When the cause of the problem has been corrected, restore the FCV to operation as follows:
| |
| : a. Return the HPU to operation on lH13-P614.
| |
| : 1) DEPRESS the " READY" pushbutton for Subloop 1 and Subloop 2.
| |
| : 2) DEPRESS the PUMP / FAN MOTOR RUNNING pushbutton for the Subloop selected as " LEAD".
| |
| : 3) Verify the " PRESSURIZED" status lights comes ON.
| |
| NOTE Servo error instrumentation becomes saturated at +(-) 12% Placing theflow controller servo error at zero or slightly negative ensures that any valve movement would be in the closed direction and therefore not addposition reactivity.
| |
| : b. DEPRESS the " MANUAL" pushbutton on the Loop A flow controller and adjest the servo error to zero or slightly negative. (ISEG OER 89-004).
| |
| l I ARP-680-04 REV 12 PAGE 21 OF 79
| |
| | |
| i FCV A MOTION INHEIT (Continued)
| |
| LARM NO.2053 - lH13*P680/04A/B03 LONG TERM ACTIONS (Continued)
| |
| : c. DEPRESS the MOTION INHIBIT RESET A pushbutton on lH13*P680. Verify the FCV does not move.
| |
| : d. Verify the " OPERATIONAL" status lights on 1H13 P614 comes ON.
| |
| : c. Manually adjust flow on Loop A, up and back do vn to verify valve responds satisfactorily to the controller.
| |
| : 5. Refer to Technical Specification 3.4.1.
| |
| POSSIBLE CAUSES
| |
| : 1. Drywell pressure 1,68 psig.
| |
| : 2. HPU Shutdown Buttons depressed.
| |
| : 3. Flow controller problems:
| |
| : a. Excessive rate of change in valve position.
| |
| : b. Demand signal out oflimits.
| |
| : c. Large velocity osci!!ations.
| |
| : d. Valve speed error.
| |
| { l HPU reservoir problems:
| |
| Hot oil (150'F)
| |
| : b. Tank empty (60 gallons)
| |
| : c. No pumps available.
| |
| : 1) Low pressure
| |
| : 2) Undervoltage/overcurrent on 2 pumps REFERENCES
| |
| : 1. GE DWG No. 828E446 Sh. 8
| |
| : 2. ESK 10lHA104
| |
| : 3. CR 96-0002 ARP-680-04 REV 12 PAGE 22 OF 79
| |
| | |
| q U
| |
| SRO EXAM KEY Esam Number NRC-SRO Rev. 0 Eram Title NRC SRO EXAM (a During power ascension the following plant conditions are noted to occur over a 3 minute period.
| |
| Reactor pressure decreased to 800 psig, now stable.
| |
| Reactor Water Level +25" and rising.
| |
| Reactor power decreased 5%, now stable at 50%
| |
| Generator output decreased to $50 Mwe from 600 Mwe.
| |
| No SCRAM No RPS actuations have occurred.
| |
| Which of the following is required?
| |
| : a. Increase power with recirculation flow,
| |
| : b. IMMEDI ATELY shut the MSIVs only,
| |
| : c. Insert a manual scram only.
| |
| : d. Scram and shut the MSIVs.
| |
| ANSWER:
| |
| : d. Scram and shut the MSIVs.
| |
| O IDNO: LP# Olu #
| |
| 232 IILO-007 9 PROCEDURE NUMBER: OTHER:
| |
| AOP-0003 LEVEL 3 I223002 A2.09 1 3.6' 3.7 _
| |
| COMMENTS: 7/97 new RO T2 Ot SRO T2 G1 (ADM 022 does not say *lMMEDIATELY", but requires actuation of the safety system tMSIV isolation). Intentional manual initiation of a scram condition (* plant conditions are approaching an unsafe condtion") should always be preceeded by a manual scram per ADM-022 and AOP-0001.1
| |
| .% 68
| |
| | |
| ATI'ACHMENT 1 PAGE 2 OF 4 SIGNAL TO ACTUATION / ISOLATION RELATIONSHIP SIGNAL ACTUATIONS AND ISOLATIONS E Main Steam Line Radiation - High: 3 X Normal Full Power Background.
| |
| : 1. Group 9 valves isolate.
| |
| : 2. Condenser Air Removal Pumps trip an.d isolate.
| |
| F Main Steam Line Pressure - Low: 849 psig with Reactor Mode Switch in RUN
| |
| : 1. Group 6 valves isolate.
| |
| G Main Steam Line Flow - High: 140% Flow
| |
| : 1. Group 6 valves isolate.
| |
| H Main Steam Line Tunnel Temperature - High: 141*F
| |
| : 1. Group 2,6,7,15, and 16 valves isolate.
| |
| J Main Steam Line Area Temperature - High: 142*F,95 ft el Steam Tunnel Area 142*F,114 ft el Steam Tunnel Area 108*F, Turbine Shield Wall Area 126*F, MSR Area
| |
| : 1. Group 6 valves isolate.
| |
| K Condenser Vacuum - Low: 8.5 inches Hg VAC.
| |
| : 1. Group 6 valves isolate.
| |
| L RWCU Equipment Area Temperature - High: 104.5 F in Heat Exchanger Rc.om 165*F in Pump Rooms 110 F in Valve Nest Room i10*F in Demineralizer Rooms 110*F in Receiving Tank Room
| |
| : 1. Group 7,15, and 16 valves isolate.
| |
| N RWCU Differential Flow - High: 55 gpm after a 45 second time delay -
| |
| : 1. Group 7,15, and 16 valves isolate.
| |
| O Standby Liquid Control System Initiation - Pump A
| |
| : 1. Group 7 valve isolates. .
| |
| O AOP-0003 REV - 10 PAGE 9 OF 17 l
| |
| ( *
| |
| | |
| (~) SRO EXAM KEY
| |
| . Esam Number NRC SRO Rev. O Exam Title NRC SRO EXAM 65 Following the receipt of an automatic reactor scram signal,10 control rods remained partially withdrawn. The plant conditions are as follows The scram valves on the 10 control rods indicate open Reactor pressure is 950 psig ARI has been manually initiated Which one of the following actions would be required to insert the 10 partially withdrawn control rods?
| |
| : a. Pull the RPS scram fuses to de energize the llCU .noids.
| |
| : b. Pull the control power fuses for the Backup Scrarn Valves,
| |
| : c. Reset ARI and vent the scram air header.
| |
| , d. Vent the CRD over piston volumes.
| |
| ANSWER:
| |
| : d. Vent the CRD over piston volumes.
| |
| IDNO: LP # Old#
| |
| 79 IILO-516 14 PROCEDURE NUMBER: OTHER:
| |
| EOP@5 ENCL 26 LEVEL 3
| |
| [201001 K1.07_. , J . 3 4;_ 3.4 {
| |
| !295037 EK3.07 i 4.2! 4.3, i COMMENTS: 7/97 new RO T1 G1 INCOMPLETE SCRAM SRO T1 G1 RO T2 G1 CRD HYD RELATIONSHIPS SRO T2 G2 69
| |
| | |
| EOP I A RPV Control- ATWS - RQA A
| |
| 2~ ' '
| |
| TGP RQA-13 Several alternate methods for inserting control rods are presented in step RQA-13. These methods are grouped within the plant symptoms indicatir g the potential problem. Ohce the potential problem is identified, tne actions listed should have the most effect at inserting centrol rods The SRO/STA should use EOP 0005 Enclosure 26 to determine the method (s) to be implemented and their priority. The operators should provide information regarding the status of the scram solenoids and the scram air header pressure to aid in this determination. Th: operators should not take any action without the direction of the SRO as other plant conditions may require that other legs of the EOPs receive higher priority than rod insertion.
| |
| A discussion of each of the alternatdcontrol rod insertion methods fol!ows. This includes - -
| |
| the various methods grouped into likely plant conditions causing the failure of rods to insert.
| |
| * Failure to De-energize Actions:
| |
| De-energize Scram Solenoids
| |
| . (l
| |
| %J This method is best employed if the scram relays fail to open. To be effective, it must be accomplished before the scram discharge volume pressurizes sutTiciently (from bypass leakage on any control rods which may have scrammed) to prevent further rod movement. If all scram valves are already open, this method is not erTective.
| |
| Enclosure 10 is utilized to pedorm the necessary actions to effect rod movement.
| |
| Maximize CRD Cooling Water AP .
| |
| If the scram failed, but control rods are otherwise not stuck, this method may be effective due to the increased pressure on the underside of the CRD drive piston. This increased differential pressure may result in some or all control rods slowly inserting into the core.
| |
| Drive Control Rods This method is best applied when only'a few control rods cannot be inserted, the scram cannot be reset, or individual control rod scrams are not g effective. Additionally, it serves as a very effective interim measure to insert (j negative reactivity while other options to insert rods are being pursued.
| |
| B - 127 Revision 3 EPSTG'0002
| |
| | |
| f EOP-I A RPV Control- ATWS RQA o
| |
| Enclosure 14 gives specific operator d.rections to accomplish this step.
| |
| * Fnilu're of A8r Header to Vent Actions: .
| |
| Vent the Scram Air Header This method is effective where one or more scram valves did not open and the hcl) area is accessible. While this method will not open valves that are mechanically prevented from c pening, it will allow those to open which are still being held closed by air pressure.
| |
| Enclosure 11 gives specific operator directions to accomplisn this step.
| |
| . Hydraulic Lock Actions:
| |
| Reset the Scram and Initia,6e a Manual Scram This method may be effective where control rods are stuck, reactor ,.
| |
| pressure and accumulator pressure are not sufficient to effect a full control rod scram, or the scram system functioned but did not result in full control rod insertions. A reactor scram is repeated as long as control rod movement occurs by
| |
| ~T (V startir.g with a drained scram discharge volume and charged accumulators. Scram signals may exist due to other olant conditions. These must be defeated to allow scram reset.
| |
| Enclosure 12 gives specificiperator directions to perform this action.
| |
| Scram Individual Control Rods This method acts on only a single control rod at a time, but can be repeated quickly for many rods. If the scram can be reset, this method may be more effective than a full core scram because the total available differential pressure of the CRD hydraulic system is applied to the single selected control rod. This results in the maximum differential pressure that can be applied over the full travcl of the control rod.
| |
| Enclosure 13 gives specific operator directions to perform this action.
| |
| Vent CRD Overpiston Volume
| |
| . This action maximizes the differential pressure across the drive piston and is a normal practice when control rod drive maintenarce is required. However, during an operational event necessitating use of this r.tethod, the discharged liquid EPSTG'0002 B - 128 Revision 3
| |
| | |
| i EOP 1 A R"Y Control- ATWS - RQA O
| |
| ( could be hot and radioactive. Access to the HCU area is required to perform these actions.
| |
| 1 Enclosure 17 gives specific operator directions to perform this action.
| |
| l t
| |
| 1
| |
| * i
| |
| +
| |
| 1 O i EPSTG'0002 8-l29 Revision 3
| |
| | |
| -. - -- _ . - =. . .- -. ..
| |
| O GI SRO EXAM KEY Exam Number NRC-SRO Rev. 0 . Exam Title NRC SRO EXAM '
| |
| 66 A reactor startup is in progress and reactor pressure is 800 psig. A loss of both CRD pumps has resulted in the receipt of the CRD ACCUMULATOR TROUBLE alarm. The nitrogen pressure on one of the CRD llCUs indicates 400 psig. Which one of the following describes the effect of this condition on the CRDM when a scram is initiated?
| |
| : a. Accumulator pressure alone will drive the rod in.
| |
| : b. Reactor pressure alone will drive the rod in,
| |
| : c. Both reactor pressure and accumulator pressure must be combined to drive the rod in.
| |
| : d. Both reactor pressure and accumulator pressure combined are inadequate to drive the rod in.
| |
| ANSWER:
| |
| : b. Reactor pressure alone will drive the rod in.
| |
| IDNO: LP# OBJ #
| |
| ( 122 IILO-003 5 PROCEDURE NUMBER: OTHER:
| |
| TS 31.5 A1&2 LEVEL 3 TS 3.1.5 C1&2 i NRC KA: I kdi T M bi ]
| |
| ;201002 K4.06 . _.j_ _ 3.5; _ _ 3.5
| |
| '201003 K1.01 2.9 3
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G3 O 70 t
| |
| l
| |
| | |
| Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS B.l. B.2.1, and 8.2.2 (continued) would already be considered " slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control rod is declared inoperable (Required Action B 2.2) and LCO 3.1.3 entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3.
| |
| The allowed Completion Time of I hour is considered reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.
| |
| C.1 and C.2 With one or more control rod scram accumulators inoperable and the reactor steam dome pressure < 600 psig, the pressure Q supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure < 600 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely-degraded during a depressurization event or at low reactor pressures. Therefore, immediately upon discovery of charging water header pressure < 1520 psig, concurrent with Condition C, all control rods associated with inoperable accumulators must be verified to be fully inserted.
| |
| Withdrawn control rods with inoperable scram accumulators may fail to scram under these low pressure conditions. The associated control rods must also be declared inoperable within 1 hour. The allowed Completion Time of I hour is reasonable for Required Action C.2, considering the low probability of a DBA or transient occurring during the time the accumulator is inoperable.
| |
| (continued)
| |
| RIVER BEND B 3.1-30 Revision No. 0
| |
| | |
| _- - - , .~. . .. , -
| |
| Q Esam Number NRC-SRO SRO EXAM KEY Rev. 0 Exa.a Title NRC SRO EXAM 67 Which one of the following conditions would result from the failure of Reactor Recirculation Pump
| |
| #1 (inner) seal assembly at rated conditions?
| |
| : a. A decrease in #1 seal cavity pressure from approximately 1000 psig to about 500 psig.
| |
| : b. An increase in #1 seal cavity pressure from approximately 500 psig to about 1000 psig.
| |
| : c. An increase in #2 seal cavity pressure from approximately 500 psig to about 1000 psig.
| |
| : d. A decrease in #2 seal cavity pressure from approximately 1000 psig to about 500 psig.
| |
| ANSWER:
| |
| : c. An increase in #2 seal cavity pressure from approximately 500 psig to about 1(XK) psig.
| |
| IDNO: LP# OBJ #
| |
| 6 STM-053 15 PROCEDURE NUMBER: OTHER:
| |
| SOP-053 LEVEL 2 ARP-680-4-E05 I202001 A1.09 ! 3.3' 3. 3 .. _
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 (D~ n
| |
| | |
| i l
| |
| RECIRC PUMP A SEAL STAGING HIGH/ LOW FLOW .
| |
| l
| |
| . 1 (LARM NO. 2069 IH13*P680/04A/E05 INITIATINO DEVICES ~ SETPOINTS -
| |
| i
| |
| : 1. Flow Switch 1B33-FISN007A 1. 1.2 gpm (Hi) 0.7 gpm (Low)
| |
| AUTOMATIC ACTIONS
| |
| : 1. None !
| |
| OPERATOR ACTIONS
| |
| : 1. Verify #1 and #2 seal cavity pressure. '.
| |
| : 2. Monitor seal staging flow temperatures.
| |
| : 3. Monitor drywell equipment drains leakage for excessive flow.
| |
| : 4. If seal purge flow to Recirc Pump A cannot be restored and Recire Pump A is idle with reactor coolant system temperature greater than 200'F, depress AUTO for A on B33 FVF075 A/B SEAL STAGING LINE SHUTOFF VALVE switch to close B33-FVF075A.
| |
| LONG TERM ACTIONS
| |
| : 1. Verify seal purge flow 3-10 gpm (local indicator).
| |
| I NOTE As long as seal cooling isprovided recircpump op' ration without sealpurgefor up to 14 days is not expected to reduce the reliable service life ofthe recircpump seals.
| |
| : 2. If seal purge flow is completely lost, restore per SOP-0003 REACTOR RECIRCULATION SYSTEM. l POSSIBLE CAUSES
| |
| : 1. #1 seal figure (high flow):
| |
| : a. #2 seal pressure approaches #1 seal pressure.
| |
| : 2. #2 seal failure (low flow):
| |
| l a. Low #2 seal pressure.
| |
| : b. RECIRC PUMP A HI OUTER SEAL LEAKAGE alarm.
| |
| : 3. Seal staging restricting labyrinth plugged (low flow).
| |
| l
| |
| , 4. Seal purge flow regulator imprope.ly set.
| |
| l S. IB33-FV-F075A SEAL STAGING FLOW VALVE isolated (lohr flow).
| |
| ; I I ARP-680-04 REV 12 PAGE 70 OF 79 i
| |
| l
| |
| | |
| SRO EXAM KEY O Exam Number NRC SRO Rev. 0- Exam Title NRC SRO EXAM 6g Given the following conditions: ,
| |
| %e Reactor Water Cleanup (RWCU) system is operating in the normal mode.
| |
| The RWCU lsolation Bypass Switches (E31 SI A,B) on P632 and P642 have been placed in "Hypass".
| |
| Select the expected effect on the RWCU system.
| |
| : a. De RWCU system isolation on hign non-regenerative heat exchanger outlet temperature is defeated.
| |
| : b. The RWCU system isolation from high area temperature ONLY are defeated.
| |
| c, The RWCU system isolation from high differential flow AND high area temperature are defeated.
| |
| : d. All RWCU system %1ation signals are defeated.
| |
| ANSWER:
| |
| c The RWCU system isolation from high difTerential flow AND high area O
| |
| temperature are defeated.
| |
| %J IDNO: LP# OIU #
| |
| 197 IILO-006 6 PROCEDURE NUMBER: OTHER:
| |
| 851E602AA l.EVEL 2 828E445AA
| |
| -. . . . . ~ . _
| |
| f 204000 K1.15 6 3.1l 3.2 _
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 O
| |
| 72
| |
| | |
| _ _ _ _ ~ _ _
| |
| ""'CHNIOUES VII. SUBJECT MATTER RWCU Filter-Demin B Room Vent. AT -
| |
| 46 degrees F.
| |
| RWCU Valve Nest Room Vent. AT - 46 degrees F.
| |
| Backwash Receiving Tank Room Vent.
| |
| AT - 46 degrees F.
| |
| i e RWCU hoist area temperature high - 141 degrees F.
| |
| OBJ #10 e RWCU Area Temperatures and Areas Ventilation AT High all annunciate alarms, at the listed value, on lH13*P680 OBJ #7c 3.4.2 Area temperature, ventilation AT and system differential flow isolations may be manually bypassed with key lock switches on panels IH13*P632 and P642.
| |
| Normally bypassed when placing the O system in operation and then taken out of bypass when system flows and temperatures have stabilized.
| |
| OBJ #4b 3.4.3 Affect on System Isolation Valves e All isolation signals except high NRHX efIluent temperature or Standby Liquid Control initiation, are indications of system line break (Leak Detection) and will result in the automatic closure of:
| |
| IG33*F001,F004,F028,F034,F039, F040, F053 and F054.
| |
| * High NRHX effluent temperature isolation is provided to prevent damage to the resin in the Filter-Demins and automatically closes 1G33*F004 only.
| |
| OBJ #10 -
| |
| Alann: F/D Inlet Temp High 140 F.
| |
| O H LO-006-6 PAGE 26 of 37
| |
| | |
| (g~' SRO EXAM KEY Esam Number NRC.SRO Rev. 0 Esam Title NRC SRO EXAM 69 The IOIR S/D Cooling isolation Valve Enable / Disable switch on the local panel (P001) has two positions," Enable / Disable" Which of the i'ollowing describes when the switch is REQUIRED to be in " Disable" and the effect on the operation of Shutdown Cooling when it is in this position?
| |
| The Rilk S/D Cooling isolation Valve Enable / Disable switch is placed in " Disable" when:
| |
| : a. reactor pressure is greater than 135 psig and prevents cperation of the RilR Shutdown Cooling inboard Isolation Valve (F009) from the Main Control Room.
| |
| : b. reactor pressure is greater than 135 psig and prevents operation of the RIIR Shutdown Cooling Outboard Isolation Valve (F008) from the Main Control Room, c, evacuating the Main Control Room to allow local operation of the IUIR Shutdown Cooling inboard Isolation Valve (F009).
| |
| : d. evacuating the Main Control Room to allow local operation of the RHR Shutdown Cooling Outboard Isolation Valve (F008).
| |
| ANSWER:
| |
| : b. reactor pressure is greater than 135 psig and prevents operation of the RHR Shutdown Cooling Outboard Isolation Valve (F008) from t:.e Main Control Room.
| |
| IDNO: LP# OllJ #
| |
| 422 IILO-021 6b PROCEDURE NUMBER: OTHER:
| |
| SOP-0031 LEVEL 4 4205000 KO 01 '
| |
| 3.3 3.4 _j COMMENTS: 1/97 NRC exam 4
| |
| | |
| The logic looks for F004 or F066 or F005 and F008 f.~s and F009 to be open.
| |
| ()
| |
| 4.2.2 Once initiated automatically, the pumps may stopped even though an initiation signal is present. This is the manual override condition. The pump will not start automatically again until the initiation signal (s) is(are) clear and the initiation reset pushbutton is depressed.
| |
| Obj. #6, 7 4.3 SDC suction valves (F008/009) 4.3.1 Shuts on RHR Equipment area high temperature 1170 F 4.3.2 Shuts on RHR Equipment area differential high temperature 290 F 4.3.3 Shuts on Low reactor water level 3 4.3.4 Shuts if reactor pressure is above 135 psig (Based on the design temperature of the pumps) 4.3.5 Manual CRVICS isolation Obj. #6,7 4.4 Pump Suppression Pool Suction Valves (F004A,B & F105C) 4.4.1 No Automatic functions.
| |
| 4.4.2 Will not open if pump SDC suctions (F006A,B) are open.
| |
| 4.4.3 Normally open when the system is in standby Obj. #6, 7 4.5 Pump SDC Suction Valves (F006A,B) 4.5.1 No automatic functions.
| |
| 4.5.2 Will not open if F004 A,B or FO24A/B valves are open to prevent draining the reactor vessel to the suppression pool.
| |
| NOTE: F024 can be opened if F006 is already open, thus providing a direct path to the suppression pool from the reactor.
| |
| An annup;iator will alarm if both valves are open at the same time. l Obj. #6, 7 4.6 RHR Pump Minimum Flow Valve (F064A,B,C).
| |
| 1 4.6.1 Auto Open )
| |
| e Pump breaker closed (in test or fully racked in)
| |
| O V e After 8 second time delay l l
| |
| H LO-021-7 PAGE 13 of 23 I
| |
| | |
| 2.3 - RHR Valves -
| |
| 0 2.3.1 If at any time the E12-F024NB) RHR PUMP A(B) TEST RTN TO SUP PL or E12-7021 RHR PUMP C TEST R1N TO SUP PL are either partially or fully open while the associated RHR puma is shutdown, the loop should be declared inoperable, the pump breaker rac ced out, and the system filled and vented 2.3.2 Do not allow E12-F048A(B) RHR A(B) HX BYPASS VALVE and E12-F003A(B) RHR A(B) HX OlJfLET VALVE to be simultaneously closed while the RHR pump is in service. If this occurs, the RHR pump should be
| |
| - shutdown prior to opening the valves. (Ref. 7.10) 2.3.3 To operate E12-F008 RHR SHUTDOWN COOLING OlJfBD ISOL VALVE from H13-P601, the ENABLE / DISABLE control switch on C61-PNLP001 must be in the ENABLE position. This switch shall be in the DISABLE position anytime the Reactor pressure is greater than 135 psig. (Ref. 7.22) 2.3.4 E12-FN0 RHR A TO RADWASTE DN STREAM ISOL VALVE and E12-FN9 RHR A TO RADWASTE UP STEAM ISOL VALVE can be if the Steam Condensing Mode of RHR is not utilized. (Ref.
| |
| 2.3.5 If performing this procedure from the remote shutdown panels, E12-F006A 4
| |
| throttlevalves.
| |
| (B) RHR pump A(B) SDC SUCITON VALVE (S) will operate 2.3.6 Valves E12-MOVF087A and B, E12-MOV026A and B, and E12-PVF051 A
| |
| . and B have been electrically disconnected per MR 88 0208. (Ref. 7.20 &
| |
| 7.40)
| |
| : 2.3.7 RHR LPCI function cannot be performed effectively in the event of a LOCA if the following valves are open. (Ref. 7.50)
| |
| : 1. E12-VF099A(B) RHR A(B) RETURN TO FUEL STORAGE AREA.
| |
| : 2. E12-VF044(B) RHR A(B) RETURN TO UPPER POOL FILL ,
| |
| 2.4 Operations 2.4.1 The RHR Shutdown Cooling Isolation at Ixvel 3 is automatically by3assed if contml of E12-F008 RHR SHIJIDOWN COOLING OUTBD ISOL VALVE and E12-F009 RHR SHtJIDOWN COOLING INBD ISOL VALVE is transferred to the Remote Shutdown Panel. Use extreme caution if control of these valves is transferred to the Remote Shutdown Panel. (Ref.
| |
| 7.45)
| |
| O SOP 4031 REV - 17B PAG 5' 6 OF 107
| |
| | |
| ATTACHMENT 1 PAGE 30 OF 48 PERFORMANCE PACKAGE !
| |
| ETEP INITIAL l I CAUTION Prior to exceeding 135 psig reactor pressure, the RIIR Shutdown Cooling Suction valves must be closed and disabled.
| |
| : 21. a. Verify 1E12*F008 RHR SHUTDOWN COOLINO OUTBD ISOL VALVE closed.
| |
| : b. Verify 1E12*F009 RHR SHUTDOWN COOLING INBD ISOL VALVE closed.
| |
| : c. Verify 1E12*F008 ENABLE / DISABLE switch is in the DISABLE position,(located on IC61*PNLP001 in DIV 1 RSS room).
| |
| : 22. Upon reaching 135 psig reactor pressure verify the following valves are closed with their respective feed breakers opened and locked. (Ref. 2.19)
| |
| INITIALS INITIALS : ,
| |
| VALVE CLOSED BRK
| |
| ()'"
| |
| LOCKED OPEN
| |
| {
| |
| IB21-MOVF001 INHS-MCC2A cub. IC 1B21-MOVF002 INHS-MCC2B cub. 3C 1E12*MOVF009 1EHS*MCC2K cub. 4A
| |
| : 23. When the turbine inlet pressure reaches the pressure regulator setpoint and if condenser vacuum is greater than 8.5" Hg, observe the bypass valve (s) open and heatup rate decrease to approximately 0.
| |
| O GOP-0001 REV - 20E PAGE 35 OF 90
| |
| | |
| (~) SRO EXAM KEY Esam Number NRC-SRO Rev 0 Esam Title NRC SRO EXAM 70' Given the following plant conditions:-
| |
| . Reactor shutdown in progress.
| |
| - IRM "C" indicating 75/125 on P.ange 6.
| |
| SELECT the statement that best describes the response of the plant if IRM C is ranged down by the operator depressing the down range pushbutton.
| |
| : a. Control rod movements can continue as normal.
| |
| : b. Only a rod block will be initiated.
| |
| - c. Only a half-scram will be initiated.
| |
| : d. A rod block and half-scram will be initiated.
| |
| ANSWER:
| |
| : d. A rod block and half-scram will be initiated.
| |
| . IDNO: LP# OBJ #
| |
| r il STM-503 13 PROCEDURE NUMBER: OTHER:
| |
| SOP-0074 LEVEL 2 i NRC MA: 7 'RUI~I 5RU[ ]
| |
| !215003 A1.05 1 3.9 3.9 _.]
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G2 m
| |
| 74
| |
| | |
| ^ ' ^ ~ ~
| |
| ATTACHMENT (
| |
| * PAGE 1 OF 1 IRM SCRAM TRIPS / CONTROL ROD BLOCKS SCRAM TRIPS v
| |
| TRIP SETPOINT BYPASSED IRM Upscale Trip Greater than 120/125 of scale Rector Mode Switch in RUN IRM Inoperative 1. Low High Voltage Reactor Mode Switch in RUN
| |
| : 2. Module Unplugged
| |
| : 3. Mode / rest Switch not in OPERATE CONTROL ROD BLOCKS TRIP SETPOINT BYPASSED i / '.! Downscale Less than 5/125 of scale Reactor Mode Switch in RUN or IRM on Range i IRM Upscale Alarm Greater than 108/125 of scale Reactor Mode Switch in RUN IRM Inoperative 1. Module unplugged Reactor Mode Switch in RUN
| |
| : 2. Low High Voltage O 3. I'RM Mode Switch not in OPERATE 9
| |
| 4 SOP-0074 REV-6 PAGE 15 OF 35
| |
| | |
| I-g Exam Number NRC-SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 71 Given the following plant conditions:
| |
| - The Refue! klatform is over the core.
| |
| The V. ode Switch is in REFUEL.
| |
| Which of the following will cause a control rod block?
| |
| : a. Withdraw a control rod.
| |
| : b. Loading the Auxiliary Platform hoist.
| |
| : c. Loading the Refuel Platform main hoist.
| |
| d Loading the Refuel Platform monorail hoist.
| |
| ANSWER:
| |
| : c. Loading the Refuel Platform main hoist.
| |
| IDNO: LP# OBJ #
| |
| 4 HLO-022 2 PROCEDURE NUMBER: OTHER:
| |
| FHP 0003 LEVEL 3 I~54RUEU l RO: l SRO: ~l
| |
| !234000 K4.01 1 3.3 4.1 COMMENTS 7/97 new RO T2 G3 SRO T2 G2 i
| |
| O 73
| |
| | |
| 5.4.7 . GRAPPLE ENGAGE pushbutton is covered to prevent inadvertent operation W is used to close the air-operated grapple fitted to the auxiliary hoist cable.
| |
| 5.4.8 GRAPPLE RELEASE pushbutton is covered to prevent inadvertent operation and is used to open the air operated grapple iltted to me auxiliary hoist cable.
| |
| 5.5 Interlock Status Display 5.5.1 BACK UP HOIST LIMIT is lit when: g,
| |
| : 1. The normal maximum up limit fails and the hoist is stopped by the maximum up limit switch.
| |
| : 2. The HOIST OVERRIDE pushbutton is depressed and hoist is rdsed above the normal up position.
| |
| 5.5.2 FUEL HOIST INTERLOCK is lit whenever the platform is over the reactor and a control rod is withdrawn, and a load greater than 550 lbs is on the grapple. When lit the fuel hoist will be inoperative.
| |
| NOTE Safety Tmvelintedock does not annunciate the TnasferPoolintedock.
| |
| O' 5.5.3 SAFETY TRAVEL INTERLOCK comes from the 7ane computer and when lit indicates the bridge, trolley, and hoist are in a res.ricted zone and blocks bridge trolley or hoist travel until corrected to prevent running the grapple into a wall or obstruction.
| |
| 5.5.4 BRIDGE COLLISION INTERLOCK, when lit, indicates that bridge motion has been blocked to stop the bridge when the auxiliary platform is approached.
| |
| 5.5.5 ROD BLOCK INTERLOCK NO. I indicates a platform-generated signal that is sent to the control room to block control rod withdrawal when the fuel hoist is loaded greater than 550 lbs and the platform is over the reactor.
| |
| 5.5.6 ROD BLOCK INTERLOCK NO. 2 indicates the same condition as No. I except it is initiated by a redundant switch.
| |
| 5.5.7 BRIDGE REVERSE STOP NO.1, when lit, indicates a condition which prohioits bridge travel towards the reactor if a control rod is withdrawn, the platform is near the core, and the fuel hoist is loaded, greater than 550 lbs.
| |
| 5.5.8 BRIDGE REVERSE STOP NO. 2, when lit, indicaree that bridge travel towards the core is prohibited because the Mode Switch is in STARTUP, and the platform is near the core.
| |
| O FHP-0003 REV - 7 PAGE 15 OF 26
| |
| | |
| g Esam Number NRC SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 72 A startup of the Main Turbine is being performed. The Main Turbine is at 60 percent of rated speed, when a loss of 125 VDC Trip Circuit Power is experienced. WillCll ONE (1) of the following describes the required operator action (s)?
| |
| : a. Enter AOP-0002, Main Turbine and Generator Trips, due to ' rip of the Turbine.
| |
| : b. ierify that 24 VDC ETS power is available and continue the startup of the Main Turbine IAW SOP-0080, otherwise manually trip the Main Turbine.
| |
| : c. Allow the Main Turbine to accelerate to greater than 90 percent of rated speed, at which time the 125 VDC Trip Circuit is no longer required because the PMG is supplying the trip circuitry.
| |
| : d. The start-up of the Main Turbine may continue, but at least one 125 VDC bus must t,e restored prior to synchornizing the generator to the grid.
| |
| ANSWER:
| |
| : a. Enter AOP-0002, Main Turbine and Generator Trips, due to trip of the Turbine.
| |
| IDNO: ;7a Olki #
| |
| f 356 STP ' 10 26 PROCEDURE NUMBER: OTHER:
| |
| AOP-0002 LEVEL 3
| |
| !245000 K6.06 ' 3 3.2 ,j COMMENTS: 7/97 new LOTM 27 5, p. 50 of 63 AOP 0002 2,1,12 Rev 7 1
| |
| l
| |
| .J 16 i
| |
| l
| |
| _ _ ._ , _- . .~
| |
| | |
| 4 e
| |
| b 1 PURPOSF3 DISCUSSION 1.1 The purpose of this procedure is to provide the operator with the instmetions for turbine / generator trips. It should be entered when a turbine / generator trip setpoint is exceeded, or when a reverse power trip is anticipated following a reactor scram.
| |
| 1.2 The Operator At The Controls will manually initiate a turbine trip whenever a turbine trip setpoint is exceeded and an aatomatic turbine trip has not occurred.
| |
| 2 SYMPTOMS 2.1 Turbine Trips:
| |
| NOTE Low ETSpressure of 400 psig is not a turbine trip signal, but will cause the Main Stop Valves. Cont ~ol Valves and CombinedIntercept Valves to close.~ Restoration ofETS pressure will cause the valves to reopen sfa turbine trip signalis not present.
| |
| 2.1.1. Reactor water high level, Level 8 2.1.2. High vibration,10 mils; when vibretion trip is disabled, a MANUAL TRIP is required at 12 mils sustained 2.1.3. Condenser low vacuum,22.3 in. Hg vac 2.1.4. Turbine bearing low oil pressure,12 psig 2.1.5. Turbine shaft oil pump discharge low pressure, less than 100 psig with turbine = peed greater than 1350 RPM 2.1.6. EHC fluid supply low pressure,1100 psig 2.1.7. Exhaust hood high temperature,225'F g 2.1.8. Active thrust bearing wear, approximately -40 mils U] -2.1.9. Inactive thrust bearing wear, approximately -40 mils AOP-0002 REV - 12 PAGE 3 OF 10
| |
| | |
| 2.1.10. Turbine overspeed 1980 RPhi,2000 RPhi backup 2.1.11. Loss of primary and backup speed signals, with turbine speed greater than 99 RPhi 2.1.12. Loss of 125VDC trip power with turbine less than 75% rated speed,1350 RPhi 2.1.13. hioisture separator A or B extreme high level 2.1.14. Stator water cooling runback falls to reduce generator load to less than 23,982 stator amps within 2 minutes, or less than 7401 stator amps within 3.5 minutes !
| |
| i 2.1.15. Automatic or manual RCIC System initiation signal with a 15 second time delay 2.1.16. Any generator trip j
| |
| 2.2 Generator Trips:
| |
| l 2.2.1. Generator reverse power f 2.2.2. Generator differential ]
| |
| 2.2.3. Generator negative phase sequence 2.2.4. Generator loss of field 2.2.5. Generator volts /liz 2.2.6. Generator network ground 2.2.7. Generator low frequency 2.2.8. Generator distance relay 2.2.9. Exciter differential a
| |
| 2.2.10. Exciter maximum excitation limiter 2.2.11. Pilot wire relay 2.2.12. Anti motoring device 2.2.13. Transfer trip from 230KV swhchyard O 2.2.14. . hiain transformer ground ovemurrent AQL-0002 REV .12 PAGE 4 OF 10
| |
| | |
| [
| |
| f I
| |
| SRO EXAM KEY
| |
| ~
| |
| O Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAh1 73 1he unit is operating at 100% rated power w hen a complete loss ofinstrument air occurs. Which one of the following indirectly causes the Reactor feedwater Pumps to TRIP?
| |
| : a. The condensate minimum now and heater drain pumps recirc valves fail open.
| |
| : b. The feedwater now conuoi valves lock up and then drin closed.
| |
| c, lhe feedwater pump suction valve drins closed.
| |
| d The feedw ster pump motor cooler TPCCW supply fails closed.
| |
| i ANSWERt
| |
| : a. The conciensate minimum now and heater drain pumps recirc valves fall open.
| |
| it) hot LP# Olu #
| |
| 85 STM 107 10 PROCEDURE NUMBER: OTHER:
| |
| AOP 0008 LEVEL 3 e
| |
| f ~ NRC kAt " I' 'RO: l 550:'l (269001 K1.06,___ . j . 3.2j . _ _.3.2
| |
| '269001 Ki 06 .
| |
| 29' 3 COMMENTS: 7/97 new RO T2 Ot SHO T2 G2 t
| |
| 0 ,,
| |
| | |
| 1 PURPOSE / DISCUSSION l.1 The purpose of this procedure is to provide guidance to the operators in the event that Instrument Air System air pressure is lowering or lost.
| |
| 1.2 A total loss ofinstrument air pressure may be caused by a break in the instrument air header, or by a loss of all air compressors. A multitude of actions occur as a result oflow instrument air supply pressure. These actions occur at various times depending upon tl>
| |
| rate of the instrument air pressure drop.
| |
| 2 SYMPTOMS 2.1 Lowering Instrument Air Header Pressure.
| |
| 2.2 Amber indicating lights for compressor C2A, C2B and/or C2C.
| |
| 2.3 Various AOVs fall in a random manner per Attachment 1, Actions To Mitigate Loss Of Air.
| |
| 3 AUTOMATIC ACTIONS NOTE These actions are listed in decreasing order ofsignificance to the Nuclear Steam Supply System.
| |
| 3.1 The Ccatrol Rod Drive System responds as follows:
| |
| 3.1.1. Control rods individually scram as the scram valves open.
| |
| 3.1.2. The Scram Discharge Volume vent and drain valves fait closed.
| |
| 3.1.3. CRD flow control valves fail closed.
| |
| 3.2 Condensate and heater drain pump recire valves open and starve the Resctor Feed Pumps; causing them to trip ,
| |
| 3.3 The Feedwater Reg Valves lock up and fail as is on low air pressure around 85 psig.
| |
| 3.4 Normal IWAC fails when AODs go closed on loss of air.
| |
| AOP-0008 REV - 10 PAGE 3 OF 17
| |
| ___-L _ . . _ _ _ _ _ _ _ _ _ _ _ . _ . - . _ . .
| |
| | |
| I g Esam Number NRC SRO SRO EXAM KEY Rev, O Esam Title NRC SRO EXAM i
| |
| 74 With the plant at 100% pow ct, a loss of VHN PNL0lBl has resulted in a loss of pow er to the feedwatir Level Control System giving a Feed P g Valve control signal failure. ,
| |
| lhe power loss also caused both Reactor Recirc pumps to shift from fast speed to slow speed and the 11 Recirc Ilow control valve to lockup. Which plant response would result from these failures?
| |
| (Assume no operator actions.)
| |
| : a. The "11" feed Reg Vahe would fait closed and the "A" and "C" Feed Reg Valves !
| |
| would AUTO OPEN to compensate. Reactor power will stabilize at a lower power level with both Recirc pumps in slow speed.
| |
| : b. All 3 Feed Reg valves will fail open. RPV level will raise to 51" which will initiate a reactor scram, Turbine trip, and feedwater pump trip.
| |
| : c. All 3 Feed Reg valves will fail closed. Reactor power willlower when Recirc pumps down shift and RPV level will lower to 9.7" w hich will initiate a reactor scram. IIPCS and RCIC will initiate at Level 2 and restore RPV level .
| |
| : d. All 3 I ced Reg valves will fail"as is". Reactor power will lower when Recire pumps down shift and RPV level will raise to $1" w hich will initiate reactor scram, Turbine trip, auf 1. dwater pump trip. j ANSWER: f
| |
| : d. All 3 Feed Reg sahes will fail"as is". Reactor power willlower when Recire pumps down shift and RPV level will raise to $1" w hich will initiate reactor scram, Turbine trip, and feedw ater pump trip.
| |
| IDNO: Lp# OllJ #
| |
| 421 IILO-532 7 PROC.EDURE NUMBER: OTHER:
| |
| AOP-0042 LEVEL 4 ARP-680-3-A06 I~ ~ NRC K Ai I'RO: 'fSRO:I
| |
| ;2t>9002 K6,02 L _ 3.3,.
| |
| ~
| |
| 3.4
| |
| * 1263000 K3 03 _ 3 di 38 COMMENTS: 1/97 new -. ,
| |
| AOP-0042 Rev 7 page 64 of 90 ARP/P680/03A/A06 i,
| |
| Jo l
| |
| . - , - - - .,-_,,-..,_,~--m,._, ,m.._ - . , . . . . . _ _ _ _ _ _ . . , _ _ _ . _ - . . . . _ , . . - _ _ _ . . - , _ _ , , . . . . _ . . _ , - - . - . _ . ,
| |
| | |
| ALARM NO. 2133 FW FIG VLV IHl3*P680 / 03A / A06
| |
| ()' -
| |
| CONTROL SIGNAL FAILURE INITIATING DEVICE 3 SET POINTS I C33 R607A, B, or C .
| |
| : 1. De. energized AtJTOMATIC ACTIONS
| |
| : 1. C33.LVF00l>'5,(C) FEEDWATER LEVEL CONTROL VALVE fails as is.
| |
| OPERATOR ACTIONS
| |
| : 1. Check the positions of the feedwater level control valves on lH13*P680 and verify they are stable.
| |
| : 2. Monitor reactor water level to verify the feedwater system is maintaining level.
| |
| LONG TERM ACTIONS
| |
| : 1. Determine which valve is affected by observing amber VALVE CONT SIGNAL FAILURE A(B)(C) light.
| |
| I
| |
| .. Closely monitor reactor level until problem is corrected.
| |
| 3 Investigate cause of alarm and initiate corrective action.
| |
| CAUTION Insure that the control signal failure condition is cleared prior to resetting the valve lockup. If the valve is reset while the control signal failure exists, the valve will fait closed. Also, insure the valve's position demand signal matches the present valve position.
| |
| 4 Once the control signal failure is cleared (Alarm 2133), the valve lockup may be reset by depressing the reset pushbutton associated with the affected valve on lH13'P680.
| |
| I l l ARP 680 03 REV 11 PAGE 6 0F 70
| |
| | |
| ALARM NO. 2133 FW REG VLV llil3*P680 / 03A / A06 CONTROL SIGNAL h FAILURE (Continued)
| |
| POSSIBLE CAUSEX
| |
| : l. Power Supply IVDN PNL0lDI CB 4 open.
| |
| : 3. Power Supply C33 K612 (K611)(K613) failure.
| |
| : 3. Fuse C33 Fl(F7)(F8)in lil3 P612 blown.
| |
| : 4. Fuse C33.F2A(D)(C) or F3 A(D)(C) in lill3 P612 blown.
| |
| 5 Controller demand less than or equal to 0%.
| |
| REFERENCES
| |
| : 1. S&W Doc. # 0247.230-000 021L
| |
| 'Il I I
| |
| , A5'd 680 03 REV - 11 PAGE 7 OF 70
| |
| | |
| i k
| |
| /
| |
| ( SRO EXAM KEY Esam Number NRC SRO Rev. O Esam Title ,NRC SRO EXAM ,
| |
| 75 1hc following conditions exist:
| |
| The reactor is at 100% power.
| |
| The Off Gas Post Treatment 111 lit til radiation alarm (P601/22A/A03) has occurred.
| |
| The Offgas System automatically isolated,(IN64 l'060 Off Gas Discharge to Vent valve is closed).
| |
| Which of the following actions is PROlllBITED?
| |
| : a. Purge the Off ..as system with service air.
| |
| : b. Shift to the Standby OfTGas Component.
| |
| : c. Reduce power as necessary to maintain condenser vacuum.
| |
| : d. Reduce power to below 60%.
| |
| ANSWER:
| |
| : a. Purge the Off Gas system with service air.
| |
| IIINO: 1.P# Olki # ,
| |
| 201 IILO-047 9 I PROCEDURE NUMBER: OTHER:
| |
| SOP-0092 LEVEL 2 L
| |
| -an- we e- 4-< em , ---* e. +
| |
| l271000 K1.07 1 2.7I 2.7 _j COMMUITS: 7/97 new RO T2 O2 SRO T2 O2 J
| |
| f
| |
| ~
| |
| u y
| |
| | |
| n n . - -- . ,
| |
| 1 EURPOSE 1.1 To provide instructions for the operation of the Offgas System and its subsystems.
| |
| 2 PRECAUTIONS AND LIMITATIONS 2.1 Any time the flowpath through the entire Offgas System is isolated, failure to close N64.VF005, AIR PURGE SUPPLY ISOL VLV may result in overpressurization of the Offgas System.
| |
| 2.2 Loss ofloop seal water or failure to fill loop seals prior to startup of the Offgas System will result in high airbome activity or a possibility of a buildup of hydrogen in unventilated areas.
| |
| C 2.3 To prevent fires or explosions, operations above 4% hydrogen concentration should an1 be allowed, and shifts to standby components should be made prior to the hydrogen concentration reaching 4%.
| |
| I 2.4 Leaks in the Ofigas System or Hydrogen Sampling System may result in the buildup of explosive hydrogen gas or radioactive gas in unventilated areas.
| |
| 2.5 To prevent wetting of the charcoal adsorbers, the charcoal adsorber beds should be bypassed during initial Ofigas System startup at reactor power levels below 10%.
| |
| 2.6 To minimize the chance of water entering the H2 Analyzers, the H2 Analyzers should un!
| |
| be placed in service until the SJAE is in operation and system flow has stabilized.
| |
| 2.7 To ensure operation, the Refrigeration Machine Selector Switch inside the local glycol panel should remain in AUTO unless one glycol refrigeration machine is inoperative.
| |
| 2.8 If the SJAE is in operation and condenser air in leakage is so low that the OfTgas System flow is less than 7 SCFM per Adsorber Train, the following valves may be opened as required to clear the low flow alarm:
| |
| * N64-F003A, RECOMB A AIR PURGE VLV e N643003B, RECOMB B AIR PURGE VLV e N64 F028A, RECOMB A PURGE FLOW VLV e N64 F028B RECOMB B PURGE FLOW VLV SOP-0092 REV - 16A PAGE 4 OF 84
| |
| | |
| SRO EXAM KFsY Q Esam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM f 76 Given the following conditions:
| |
| Reactor water level is 90 inches and lowering.
| |
| Drywell pressure is 2.2 psig and raising. l An outside fire has caused smoke in the Control Room.
| |
| The operator has attempted to manually place the Control Room ventilation in the smoke removal mode.
| |
| Under these conditions the Control Room Smoke Removal Damper (AOD 107/108) will:
| |
| : a. open and the Smoke Removal Fan will start.
| |
| b, open but the Smoke Removal Fan will be interlocked off. I c remain closed and the Smoke Removal Fan will run on recirc.
| |
| : d. remain closed and the Smoke Removal Fan will be interlocked oft.
| |
| ANSWER:
| |
| : d. remain closed and the Smoke Removal Fan will be interlocked off.
| |
| IDNO: LP# Olu #
| |
| 167 LOTM-61 11 PROCEDURE NUMBER: OTHER:
| |
| SOP @58 LEVEL 4
| |
| !290003 K1.04 1 3.2' 3.3 _j COMMENTS 1 7/97 new RO T2 02 SRO T2 02 80 l
| |
| l l
| |
| | |
| m -m .
| |
| ..s
| |
| /
| |
| 5.6 OxTation of 1HVC-FN10 MISCELLANEOUS CONTROL BLDG AREA SMOKE REMOVAL SYSTEM NOIE 1hefollowing steps are verfonned at the Fire Protection Control
| |
| * Console IH13-P861 in tie Control Room unless >
| |
| othensise noted NOIE 1he Control Room Smoke Renunal 5 stem cannot be started wlwn a Lewi 2
| |
| /1.68 prig LOCA, High Radiation, or manual isolation sigtnlis present.
| |
| NOTE Operation of the Snwke Renxnalfans may cause Control Bldg Ventilanon to trip on Low Flow.
| |
| 5.6.1 Smoke Removal from Miscellaneous Control Bldg Areas.
| |
| : 1. Shutdown Control Building chilled water loop as follows:
| |
| : a. Lockout the standby chiller
| |
| : b. Place the standby chilled water pump to stop.
| |
| : c. Lockout and stop the running chiller and verify the Otiller Recire SWP stops and its suction MOV closes.
| |
| NOIE 1he chilled unterpump uill trip automatically after 120 sec.
| |
| : d. After the running chiller has been stopped for thirty seconds, stop the running chilled water pump and verify the followmg:
| |
| : 1) The chilled water pump discharge vahe closes.
| |
| : 2) The running AHUs and fans stop, a) If necessary reset the trips on the AHUs and fans.
| |
| : 2. Start IHVC-FN13 SMOKE REMOVAL FAN.
| |
| ,q LJ SOP-0058 REV - 11A PAGE 15 OF 56
| |
| | |
| A i 1 As rtMENI' 2 '
| |
| PAGE 6 OF 6 ISOLATION VALVE CliECKOFF SHEET MXATHY DIVI &IV IGNIOREIY l'.tNEL EOIATHY DIVIIA E RINIORD LNTI1ALS OUIBOARD NIIALS NIIAIS INBOARD MITALS lil3 P863 INGAOD19C SOP 0058 liVC INGAOD19D SOP 0058 INGAOD19E SOP 0058 ISOLATION INGAOD19F SOP 0058 IWC-AODIA SOP 0058 SIGNALS INGAODlB SOP 0058 INC AOD108 SOP 0058 B,D.FF IWGAOD107 SOP 0058 INGAOD51A SOP 0058 IWC AODSIB SOP 0058 INC AOD52A SOP 0058 IWC-AOD52B SOP 0058
| |
| !!VK-hDV10A Place IWK-MOV10B Place switch in switch in CLOSE CLOSE and and release release Ranarks:
| |
| O Performed By: /
| |
| Signature KCN Initials Dateffime
| |
| /
| |
| Signature KCN Initials Date/ Time
| |
| /
| |
| Signature KCN Initials Date/ Time Reviewed By: /
| |
| OSS/CRS KCN ,
| |
| Date/ Time Second Review: /
| |
| Operations Management KCN Date/ Time AOP-0003 REV - 10 PAGE 17 OF 17
| |
| | |
| _ _ - - . .. . ~ . _ - _ . __ . _ _ - _ . - _ - . _ . . . - - _ - - - - . . . _ . . . - . . - . . . . .
| |
| (
| |
| SRO EXAM KEY Enam Number NRC.SRO Rev 0 Esam Title NRC SRO EXAM 77 The plant was operating at 100% power when a large steam leak on the MSR A reheat steam line required reheat steam to be isolated. Five minutes later the TURBINE 111G11 VBRATlON annunciator on P870 alarms. You check the recorder and observe bearing #5 reading 9 mils, bearing # 6 reading 13 mils, and bearing # 4 reading 5 mils. These readings appear to be steady.
| |
| You should:
| |
| : a. Continue to monitor the vibrations and initiate a reactor scram and trip the turbine if 15 mils is reached.
| |
| : b. Commence a rapid load reduction then take the turbine offline (scram and turbine trip)if vibrations cannot be reduced below 10 mits within 14 minutes.
| |
| : c. Immediately scram the reactor and trip the turbine.
| |
| : d. Monitor bearmg temperatures and scram the reactor and trip the turbine if bearing temperature exceeds 240 degrees F.
| |
| ANSWER:
| |
| : c. Immediately scram the reactor and trip the turbine.
| |
| IDNO: 1.P # 010 #
| |
| 71 IILO-025 9 PROCEDURE NUMBER: OTHER:
| |
| AOP-0002 LEVEL 3 ARP-870 54 D08 i ~ "NRC K Al ^ I'RO R ~SRU[ ~l (245000 K4.05 i 2.9 3 j COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 O n
| |
| | |
| v, , , ,
| |
| 1 PURPOSE /DISCUSSIQN 1.1 The purpose of this procedure is to provide the operator with the instructions for turbine / generator trips. It should be entered when a turbir.e/ generator trip setpoint is exec.:ded, or when a reverse power trip is anticipated following a reactor scram.
| |
| 1.2 The Operator At The Controls, will manually initiate a turbine trip whenever a turbine trip setpoint is exceeded and an automatic turbine trip has not occurred.
| |
| 2 SYMPTOMS 2.1 Turbine Trips:
| |
| NOTE ,
| |
| Low ETSpressure of 400psig is not a turbine trip signal, but will cause the Main Stop Valves, Control Valves, and CombinedIntercept Valves to close. Restoration ofETS pressure will cause the valves to reopen ifa turbine trip signalis notpresent.
| |
| 2.1.1. Reactor water high level, Level 8 2.1.2. High vibration,10 mils; when vibration trip is disabled, a MANUAL TRIP is required at 12 mils sustained -
| |
| 2.1.3. Condenser low vacuum,22.3 in. Hg vac 2.1.4. Turbine bearing low oil pressure,12 psig 2.1.5. Turbine shaft oil pump discharge low pressure, less than 100 psig with turbine speed greater than 1350 RPM 2.1.6. EHC fluid supply low pressure,1100 psig 2.1.7. Exhaust hood high temperature,225'F i
| |
| I 11.8. Active thrust bearing wear, approximately -40 mils 2.1.9. Inactive thmst bearing wear, approximately -40 mils 1
| |
| A OP-0002 REV - 12 PAGE 3 OF 10
| |
| | |
| 4.5.22. Monitor turbine vibration and take action as follows:
| |
| LIMITS FOR SUBSYNCHRONOUS TURBINE VfBRATION Trip After Trip Vibration 1 Any Immediately Level Journal if Journal ~ Acceptable Speed Vibration Vibration For RPM Exceed Exceeds Continued Operation less than N/A 8 mils N/A 800 800 1400 10 mils for 14 mils 7 mits 2 min 1400- 10 mils for 12 mils 5 mils running 15 min speed 4.5.23. WHEN the turbine reaches 900 RPM. THEN stop the bearing lift pumps.
| |
| ' NOTE Achieving I300 RPM with lube oilpressure less than 100psig results in a turbine trip.
| |
| 4.5.24. Check main shaft oil pump discharge pressure, as indicated on TML PI EPRI, MAIN SHAFT OIL PUMP DISCH PRESS is greater than 100 psig prior to turbine speed reaching 1300 RPM.
| |
| 4.5.25. WHEN 1500 RPM is reached, THEN perform the following:
| |
| : 1. Check SPEED INCREASING light goes off and SET SPEED light comes on.
| |
| : 2. Check oil supply temperature is greater than or equal to 100*F.
| |
| (
| |
| SOP-0080 REV - 12 PAGE 20 OF $7
| |
| | |
| SRO EXAM KEY O Esam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM 73 1hc plant is operating at 60% power for rod pattern adjustment. Rod 22-43 is to be inserted from .
| |
| notch 48 to notch 42. A few seconds after the rod is insened to notch 42 a ROD DRilT annunciator is recieved. While conducting your immediate actions, you observe rod 22 43 passing '
| |
| notch 46 and observe it to stop at notch 48.
| |
| The Control Rod Movement Sequence withdraw limit is notch 48. Reactor ,
| |
| power has returned to 60%. The correct actions will be:
| |
| : a. Place the Mode Switch in SilUTDOWN.
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whether the rod I has a stuck collet.
| |
| : c. Since power is below ilPSP the Rod Withdrsw Error analysis is affected and the rod should be inserted to notch 42 without delay.
| |
| : d. Notify a Reactor Engineer, declare rod 22 43 inoperable, and adjust the pattern as needed for flux shaping with rod 22-43 full out since it will not remain inserted.
| |
| ANSWER: ,
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whether the rod has a stuck collet.
| |
| i II)NO: LP# Otti #
| |
| 49 STM 052 4 :
| |
| PROCEDURE NUMBER: OTHER: ,
| |
| ARP-680-07 B02 LEVEL 3 f' NRC'K AI l" RO$ l 'SRO:' l ,
| |
| ;201003 K4.07 . l _ _ . 3. 2, . 3.2 j 4201003 A2 03 3.4' 3.7 COMMENTS: 7/97 new RO T2 G2 SRO T2 G3
| |
| [Requuod to determine TS 3,1.3 Condition A epphcabhtyl
| |
| .O n
| |
| -a-wW-- e,w . =*- e grwyP' e----s--g --ge*ye- ge ewa p- -rm +-&-++ 9- -
| |
| 3-t--- - -
| |
| y a= rN .
| |
| ye
| |
| | |
| CONTROL ROD DRIFT ALARM NO. 2115 1H13*P680 / 07A / B02 INITINITNG DEVICES SETPOINTS
| |
| : 1. PIP Switches 1. ODD Switch Opened and Not Driving Rods AUTOMATIC ACTIONS
| |
| : 1. None OPERATOR ACTIONS (Commitment 05609,07729)
| |
| : 1. Determine which rod (s) drified.
| |
| 1.1 Depn:ss ROD DRIFT pushbutton on 1H13*P680 and obsen'e red lights on ROD POSITION DISPLAY.
| |
| : 2. Select the drining rod (s) and apply a continuous insert signal.
| |
| I l
| |
| : 3. Verify cooling water pn:ssure is appmximately 20 psid.
| |
| : 4. If the red continues to withdraw from full in reinsert with continuous insert signal and hold until CRD drive pressure can be n:duced or an Operator isolates the HCU by closing Drive wuter (V103) and Exhaust water (V105) isolation valves.
| |
| 4.1 If the CRD remains inserted, then Dinxtional Contml Valve w.') failure is suspected.
| |
| NOTE A stuck collet piston could occur as a result offoreign material uedging in the close clearances around the collet piston, and is most likely to occsefollmving a semrn preceded by an ettended reactor rurt (GESIL 310) 4.2 If the CRD continues to drift out, then a stuck collet piston is suspected. Re-open the HCU Drive Water (V103) and Exhaust Water (V105) isolation valves, if closed, and reinsert the CRD with a continuous insert signal.
| |
| l 4.2.1 Do not unlatch other CRD mechanisms if a stuck collet pison is suspected as the cause of the rod drift.
| |
| ARP-680 07 REV-9A PAGE 16 OF 61
| |
| | |
| CONTROL ROD DRIFT ALARM NO. 2115 (Continued) lill3*P680 / 07A / B02 LONG TUtM ACTIONS
| |
| : 1. Contact Reactor Engineering as soon as possible for assistance.
| |
| 1.1 Observe CRD temperature recorder to assure adequate cooling flow to all rods.
| |
| : 2. If the rod drift was determined to be a failure of Directional Control Valve (122), then valve out the ,
| |
| itCU per SOP-0002 for valve repair. Refer to Tech Spec LCO 3.1.3.
| |
| : 3. If the rod was detennined to be a stuck collet piston, then perform the following:
| |
| 3.1 If reactor pressure is >500 psig, attempt to flush the collet area by performing an individual rtxl scram. Release the continuous insert signal after the individual scram to determine if the CRD is latched.
| |
| 3.2 If the individual scram did not latch the CRD:
| |
| 3.2.1 Reapply the continuous insert signal to maintain the CRD fully inserted.
| |
| I l NOIE lle reactor and rod control status s!wuld be such that the CRD can be allom'd to pasition 48.
| |
| 3.2.2 Obtain permission from the OSS/CRS and concum:nce of the Reactor Engineer te release the continuous insert signal.
| |
| 3.2.3 Drive out the CRD to position 48.
| |
| 3.2.4 Raise the CRD Drive Water pressure to appmximately 450-500 psid.
| |
| 3.2.5 Give the IICU a continuous withdrawal signal for approximately two minutes.
| |
| 3.2.6 Notch the CRD in to check collet operation.
| |
| 3.2.7 Repeat Steps 5 and 6 until the CRD latches and unlatches reliably. .
| |
| ARP 68M7 REV-9A PAGF 17 OF_fi2
| |
| | |
| SRO EXAM KEY O Esam Number NRC SRO Rey, O Esam Title NRC SRO EXAM 79 The control room operator is about to startup RilR in the fuel pool cooling assist mode. All prerequisites have been completed and the operator starts RilR pump 'A'. lie then starts to throttle open the llX OUTLET VIN (E12 FG 3A) Watching the valve position indication the operator ;
| |
| obsen es there is no change, also flow is 700 gpm. Fuel pool level: (choose one)
| |
| : a. INCREASE because Suppression Pool water is being diverted to the fuel Pool.
| |
| : b. REMAIN Tile SAME because the RilR system is currently recircing 700 gpm from!!o the Suppression Pool.
| |
| : c. REMAIN Tile SAME because the RilR system is currently recircing 700 rpm from'to the Fuel Pool.
| |
| : d. DECREASE because water from the Fuel Pool is being diverted to the Suppression Pool.
| |
| ANSWER:
| |
| : d. DECREASE because w ater from the Fuel Pool is being diverted to the Suppression Pool IDNO: LP# Olu #
| |
| 438 IlLO-021 08 PROCEDURE NUMBER: OTHER:
| |
| SOP 0031 LEVEL 3 TASK NUMHER:
| |
| 20$005001001 205013001001 I ~ NRC KAi 'I RO: l'SROi }
| |
| +233000 K3 02 4 3.1 f 32 _l COMMENTS: 7/97 NEW O ,,
| |
| | |
| CAtmON Failure to establish greater than 1100 mm within 8 secomb of pung start will cause 1E12*fM4A(B) RHR PUMP A(B) NIN FIDW TO SUP PL to open, dumpirg reactorwater to the suppression pool.
| |
| NOTE 1he respectiw torit cooler (s) shall be in service prior to starting RHRpenp(s)for sinadoun cooling. (Ref 7.17)
| |
| CAlmON
| |
| %e initiation of Shutdown Cooling may result in more pmnotared notching of imlicated reactor level. Monitor reactorlevel tremh. (reference 7.54)
| |
| : 10. Stan RHR PUMP NB) and IMMFDIATFI Y thmttle open IE12*F048A(B) RHR A(B) HX BYPASS VALVE to obtain 2000 -
| |
| 3000 gpm.
| |
| I1. Establish a stable Pow of approximately 4000 - 5000 gpm by throttling IE12*FN8A(B) RHR A(B) HX BYPASS VALVE. (Re~ f. 7.31) n V
| |
| 12.appmx.10 Throttle op% indication.en IE12*F003A(B) RHR A(B) HX OUT CAlmON RhR HX tempendure imlicatiorn E12-R601 PT 1(2) do not show actual water tenveratures with 1E12-FD03A(B) fully closed. With E12-FD03A(B) fully closed, tse E12-R60 L FF 5(6) for coolant tenverature indication.
| |
| : 13. Establish a cool down mte ofless than 100 DEFGhr as follows:
| |
| : a. Slowly jog open 1E12*F003A(B) RHR A(B) HX OU'II.ET VALVE and monitor the cooldown rate.
| |
| : b. nrottle 1E12*F003A(B) RHR A(B) HX OUTLET VALVE and IE12*FN8A(B) RHR A(B) HX BYPASS VALVE to obtain the desired cooldown rate while maintaining a constant RHR loop flow.
| |
| O SOP-0031 REV - 17B PAGE 27 OF 107
| |
| | |
| l SRO EXAM KEY ,
| |
| Esam Number NRC.SRO Rev. 0 Esam Title NRC SRO EXAM 80 The plant is operating at 75% power. The Control Room Operator places the Outboard MSIV Positive Leakage Control System swPch to OPERATE. Which of the following will prevent the Outboard MSIV Positive Leakage Control System from initiating? i
| |
| : n. A LOCA signal on either high drywell pressure or low reactor water level is not present.
| |
| : b. The required main steam line pressure and reactor pressure requirements have not been met.
| |
| : c. The post LOCA 20 minute timer has not timed out. l
| |
| : d. All Main Steam isolation Valves have not been fully closed.
| |
| ANSWERt
| |
| : b. The required main steam line pressure and reactor pressure requirements have not been met.
| |
| 10NO: LP# OltJ #
| |
| 190 LOTM 8 4 PROCEDURE NUMBER: OTHER:
| |
| ARP401 17-G05 LEVEL 3 ARP-601 17-G06 SOP-0034
| |
| !' ' NRd K A: l HOI l srb:
| |
| ;239001 K1.13 j _2.6, 2.8
| |
| ;239003 K1.01 ; 3.3; 3.4 239003 K4 01 2.9 32 COMMENTS: 7/97 new RO T2 G3 SRO T2 G2 O ,
| |
| | |
| NOTES Appropriate timeframefor manual activation ofsystem is 20 minutesfollowing a LOCA.
| |
| The switch in thefollowing step is a keylock switch.
| |
| 4.2 Take E33A S1 A(SIB), OPERATE INBOARD (OUTBOARD) MSIV PLCS Switch to OPERATE.
| |
| N_OTES The pressure permissives are annunciated at P601-17A G03(G06), PERhilSSIVE TO OPERATE INBOARD (OUTBOARD)MSIVPLCS. ,
| |
| 4.3 WHEN air supply pressure increases above 50 psig AND reactor pressure decreases below 25 psig, THEN check the following:
| |
| 4.3.1. E33 F014(F034), INBD(OUTBD) BYPASS VALVE opens 4.3.2. E33 F006(F026), DRAIN VALVE closes
| |
| * NOTE When E33 F006(F026), DRAIN VALVE closes, a 3 minute timer is activated which bypasses the high .gstemflow and ,
| |
| low & trips. This allows timefor the Main Steam Lines to .
| |
| pressurize and reach operating equHibrium.
| |
| 4.4 WHEN E33 F006(F026), DRAIN VALVE closes, THEN check E33-F005(F025),
| |
| INJECTION VALVE opens. >
| |
| 4.5 WHEN E33 F006(F026), DRAIN VALVE closes AND Main Steam Line pressure is less than 35 psig, THEN check de following valves open:
| |
| 4.5.I. E33 F007(F027), ISOLATION VALVE 4.5.2. E33-F008(F028), ISOLATION VALVE 4.6 WHEN at least 5 minutes have elapsed since E33 F006(F026), DRAIN VALVE has closed, THEN check E33 F0'4(F034), INBD(OUTBD) BYPASS VALVE closes.
| |
| 4.7 Check E33 R603(R623), PRESS CONTROLLER operates to control inboard (outboard) system differential pressure at 8.5 psid.
| |
| SOP-0034 REV-7 PAGE 5 OF 20 I
| |
| | |
| m s]
| |
| , i ,,.# .5
| |
| ., PERMISSIVE TO OPERATE ,
| |
| ALARM NO. 2432 INBOARD MSIV PLCS IH13*P601 / 17A / GOS INITIATING DEVICES SET POINTS
| |
| : 1. lE33'PTN001 and IE33*PISN602 l'. ;t45 psig air and s:25 psig steam l
| |
| AUTOMATIC ACTIONS l i
| |
| : 1. None DPERATOR ACTIONS
| |
| : 1. Manually initiate system per SOP-0034 MSIV SNALING SYSTEM (POSITIVE LEAKAGE CONTROL)if required.
| |
| LONG TERM ACTIONS
| |
| : 1. Monitor system operation per SOP-0034 MSIV SEALING SYSTEM (POSITIVE LEAKAGE CONTROL.)
| |
| POSSIBLE CAUSES *
| |
| : 1. Air supply pressure above low pressure setpoint, and Rx pressure is below setpoint.
| |
| REFERENCES
| |
| : 1. 1.LSMSI.013
| |
| : 2. 1.LSMSI.021 -
| |
| : 3. GE 793E922AA ARP-601-17 REV-5 PAGE 49 OF 63
| |
| | |
| SRO EXAM KEY O Esem Number NRC SRO Rev. O Esam Title NRC SRO EXAM gi 1he following conditions exist:
| |
| . The reactor is in cold Shutdown per GOP 0002. '
| |
| . Reactor recirculation pump A and RilR A shutdown cooling loop are in operation.
| |
| . RPV level is 35 inches.
| |
| Subsequently, Reactor Recirculation pump A trips and RPV level is being raised to greater than 75 ;
| |
| inchet Select the LOWEST vessel level which will result in water flow into the steam lines?
| |
| a, 75 inches
| |
| : b. Il0 inches e, I40 inches
| |
| : d. 196 inches ANSWERt
| |
| : b. Il0 inches O
| |
| IDNO: LP # OILI #
| |
| 215 LOTM 2 2 PROCEDURE NUMBER: OTHER:
| |
| OSP-0041 LEVEL 2
| |
| !290002 K1.01 + 32 3.2]
| |
| COMMENTS 7/97 new RO 12 03 SRO T2 03 G n
| |
| _ _ , _ _ . . . . . . - . . . , . . . . _ _ - . . . - . . - . . . . , . , , . _ . . _ . . . . = . . . _ . , . , _ z.-,m.
| |
| | |
| 5.16 Precautions Particular to MSL Flooding 5.16.1. To preclude brittle fracture concerns, maintain reactor temperature greater than 120*F and reactor pressure less than 700 psig. Temperature can be monitored at the following locations:
| |
| : 1. Bottom head drain when RWCU is in service
| |
| : 2. RHR Pump discharges when RHR A(B) is in service
| |
| : 3. Reactor Recirc Loops when Recirc Pumps are operating 5.16.2. During MSL flooding operations, the reactor shall remain depressurized. Refer to Attachment 8. MSL Alternate Decay Heat Removal Equipment Failures for guidance on RPV pressure increases.
| |
| 5.16.3. Radiation levels in the Turbine Building will increase during the implementation of this procedure. Appropriate precautions and posting of areas should be performed.
| |
| 4 5.16.4. The elevation of the Main Steam Lines'is approximately 105".
| |
| If RPV level is allowed to decrease below this level, RPV O, Cooling via this method will be lost.
| |
| I e
| |
| e S
| |
| O OSP 0041 REV-0 PAGE R OF 61 4
| |
| | |
| NOTE
| |
| ' A decrease in RPVtemperature is to be expected when raising water level to)111 the Main Steam Lines.
| |
| : 14. Using the Condensate and Feedwater system with the i Startup Level Control Valve in manual per SOP 0009, Reactor Feedwater System, Section 4.5, raise RPV level to greater than 110" but less than 180" using the Shutdown Range Level Instmment. If desired, bypass Feedwater Heaters to minimize areas affected by flow of Reactor Coolant Water as follows:
| |
| (Initials)
| |
| : 1) Open CNM MOVl36, LP HTRS BYP (Initials)
| |
| : 2) Close CNM MOV33A, LP HTR STO A TNL (inillais)
| |
| J) Close CNM MOV33B, LP HTR STO B INL (Initials)
| |
| : 4) Close CNM MOV32A, LP HTR STO A OUTL (inillais)
| |
| : 5) Close CNM MOV32B, LP HTR STO B OUTL (Imtials)
| |
| : 15. Maintain Hotwell Level at approximately 74'6" using the Condensate Transfer System.
| |
| (Initials) .
| |
| : 16. Commence monitoring Hotwell Level at least once per shift and record in the Control Room Log.
| |
| (Initials) 4 1
| |
| 4 0 -
| |
| OSP-0041 REV-0 PACE 40 OF 61 y 9--, --__,7- -,m -e-- --g,_7, --
| |
| | |
| SRO EXAM KEY O Exam Number NRC-SRO Rev. O Esam Title NRC SRO EXAhJt l
| |
| 32 The plant is conducting a startup with reactor pow er at 22%, w hen an unisolable rupture in the Turbine Plant Component Cooling Water (TPCCW) suction header causes a complete loss of TPCCW. What are the required immediate operato actions?
| |
| : a. Conduct an orderly reactor shutdown per GOP-0002. Plant Shutdown.
| |
| : b. Initiate RCIC and shutdown the Feedwater pumps.
| |
| : c. Reduce reactor power to within Ilypass Valve capacity then trip the Main Turbine.
| |
| : d. Manually scram the reactor.
| |
| ANSWER:
| |
| : d. Manually scram the reactor.
| |
| i IDNO: 1.P # OBJ #
| |
| 38 IILO 531 5 PROCEDURE NUMSER: OTHER:
| |
| AOP@12 LEVEL 2 .
| |
| r NRC KAi I ROi l'SRO:
| |
| 296018 AK2.02 _.q.
| |
| e 3.4: ..
| |
| - --3.6 COMMENTS: 7/97 new ROT 102 SRO T1 G2
| |
| --..--....,-,.-.4, - . . . 4 e - - - - -e.- --e-.e--- - - _ . . . -,.r-, ,,c.-y, , . . - ----m- y , 7y 9 yw--,- -- - . , , - - , . - , y,,
| |
| | |
| $'s 3 l e 1 2 SYMPTOMS 2.1 CCS Pump trip 2.2 Lowering CCS system pressure 2.3 Raising CCS temperature
| |
| ^
| |
| 2.4 CCS Surge Tank low level 2.5 Raising temperatures on equipmcat cooled by CCS :
| |
| 3 AUTOMATIC ACTIONS 3.1 Standby CCS Pump starts on low pressure or trip of a running pump.
| |
| 4 131 MEDIATE OPERATOR ACTIONS 4.1 Attempt to start at least one CCS Pump.
| |
| 4.2 E CCS flow can N.QI be re-established AND condensate /feedwater is required to maintain RPV level, THEN manually scram the Reactor.
| |
| 5 SUBSEOUENT OPERATOR ACTIONS 5.1 E CCS can HQI be restored within several minutes, THEN shut down the following equipment as necessary:
| |
| * Off Gu Vault Refrigeration Compressors
| |
| . Off Gas Glycol Refrigeration Units
| |
| * Generator Stator Cooling Pumps e Condenser Air Removal Pumps
| |
| * Feedwater Pumps .
| |
| * Hantar Drain Pumpa e Condensate Pumps AOP-0012 REV-8 PAGE 4 OF 5 1
| |
| | |
| i SRO EXAM KEY O Esam Number NRC SRO Rev. O Esam Title NRC SRO EXAM g3 A plant startup is in progress in accordance with GOP 0001. Plant conditions are as follows:
| |
| - 1he reactor is suberitical with control rod withdrawal in progress.
| |
| - Reactor power is 1 x 10-4 cps on SRMs. !
| |
| Recirculation loop temperatures are 180 deg F. 1 Whid, one of the following statements is correct conceming the administrative Control Room Supersvisor7
| |
| : a. The Admin CRS is required to be stationed at this time and remain stationed until criticality is achieved.
| |
| : b. "the Admin CRS is required to be stationed from criticality until the last FWREO valve is in service.
| |
| c.. The Admin CRS is required to be stationed from criticality until the first Main Turbine Dypass valve is 50% open,
| |
| : d. The Admin CRS is required to be stationed at this time and remain stationed until the last FWREO valva is in service.
| |
| ANSWER:
| |
| : b. The Admin CRS is required to be stationed from criticality until the last FWREG valve is in service.
| |
| IDNO: 1.P # 010 #
| |
| 436 ilLO 206 3 PROCEDURE NUMBER: OTHER:
| |
| ADM 0022 LEVEL 3
| |
| 'O 2.1 4 2.3- 3.4 g COMMENTS: 7/97 f4W 87
| |
| | |
| 1
| |
| . Additional Operations Section personnel may be required on 4.1.6.
| |
| O- shift because of unusual plant conditions or operational needs.
| |
| The OSS shall obtain the additional personnel as necessary.
| |
| Activities requiring additional personnel shall not be undertaken until the required personnel are available. ,
| |
| : 1. _ During Startups and Shutdowns an additional SRO'shall be utilized as Administrative Control Room Supervisor to relieve the on shift CRS of his administrative duties, and an extra NEO/NCO may also be used for additional support.-
| |
| : 2. The Admin bR'S must have a current SRO license; but the .
| |
| license is not required to be active (proficient). The
| |
| - Admin CRS, if not active (proficient), may not direct licensed operators in the manipulation of controls, nor may he manipulate the controls.
| |
| : 3. For startups the Admin CRS should be stationed from criticality until the last FWREG Valve is placed in service, and the ' extra NFO/NCO from criticality until 25% power. If the last FWREG Valve cannot be placed_
| |
| in service due to equipment problems, the Admin CRS
| |
| ,O- may be secured at the discretion of the Superintendent-Operations.
| |
| : 4. On normal planned shutdowns the Admin CRS should be stationed from 80% until the reactor is tripped.
| |
| 4.1.7. Personnel who are relieving the shift should review all pertinent operating data back to the time that they had' previously been on watch.
| |
| 4.1.8. When an Operations department or IST section STP is compl:ted on shift, the OSS/CRS will review the procedure and sign as the section supervisor signifying review completion per ADM-0015 STATION SURVEILLANCE TEST PROGRAM. The STA on the following crew will independently review the STP as a second-check that the acceptance criteria have been met and the STP has been completed satisfactorily.L (Ref. 2.77)
| |
| : O .
| |
| Fm;E 19 OF 53 I ' ADM-0622 REV - 19 i___-___. _ _ . . - - - - - - - - - - _ - - - .. -. - , ,
| |
| | |
| SRO EXAM KEY O Esam Number NRC SRO Rey, O Exam Title NRC SRO EXAM g4 While the plant is at power, a leak develops in an area that is acessible, but now radiologically contaminated. The OSS has directed that an investigation be performed immediately. What documentation must be generated before various personnel are allow ed entry into the area for the investigation?
| |
| : a. A daughter RWP to the Getieral RWP for that area must be generated.
| |
| : b. None, a General RWP already exists for this type of event.
| |
| : c. A Specific RWP must be generated,
| |
| : d. None, a RWP may be completed after the entry provid~i it is donc under continous RP coverage.
| |
| ANSWER:
| |
| : d. None, a RWP may be completed aller the entry provided it is donc under continous RP coverage.
| |
| IDNO: LP# On!J #
| |
| 433 GET-022 27 O. PROCEDURE NUMBER:
| |
| RSP-0200 OTHER:
| |
| LEVEL 2 TASK NUMilER:
| |
| 300157003002 G 2.3.4 i 2.5- 3.1 COMMENTS: 7/97 nGW O' 88
| |
| | |
| O SRO EXAM KEY Esam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM 85 SOP-0031," Residual lleat Removal" cautions the operator NOT to simul taneously close the RilR 11X BYPASS valve (El2-MOVF0480) and the RilR A IIX OUTLET valve (E12 MOVF003B) while the RilR pump is in service.
| |
| Which one of the following actions should be taken if both of these valves are inadvertently closed in RilR loop B while in suppression pool cooling?
| |
| : a. Open the F048B firr < a differential pressure across the F003B then throttle open F003B
| |
| : b. Open the F0031) imn- vh ato:s "sh flow through the RilP. l{X.
| |
| : c. Shutdown the RilR puinp , 2 'o c, T either the F003B or the F048B.
| |
| : d. Verify that the minimum flu. IB) is fully open and reopen either F003B or F00488.
| |
| ANSWER:
| |
| : c. Shutdown the RilR pump prior to opening either the F003B or the F048B.
| |
| IDNO: LP# Olti #
| |
| 326 IILO-021 8 PROCEDURE NUMBER: OTHER:
| |
| SOP-0031 LEVEL 3 40 2.1.32 3.4* 4 COMMENTS: 7/97 new RO & SRO T4 G1
| |
| | |
| .73 ,
| |
| l F
| |
| 2.3 RHR Valves -
| |
| 2.3.1 If at an time the E12-F024A(B) RHR PUMP TEST RTN TO'SUP PL or E12 h021 RHR PUMP C TEST RTN are L TOeither S partially or fully open while the associated RHR pum) is shutdown, the loop should be declared inoperable, the pump breaker rac ced out, and the system filled and vented.
| |
| 2.3.2 Do not allow E12-F048A(B) RHR A(B) HX BYPASS VALVE and E12-F003A(B RHR A(B) HX OUTLET VALVE to be simultaneously closed while the)RHR pump is in senice. If this occurs, the RHR pump s shutdown prior to opening the valves. (Ref. 7.10) 2.3.3 To operate E12-F008 RHR SHUIDOWN COOLING OUTBD ISOL VALVE fmm H13-P601, the ENABLE / DISABLE control switch on C61-PNLP001 must be in the ENABLE position. This switch shall be in i the DISABLE position anytime the Reactor pressure is greater than 135 psig. (Ref. 7.22) 2.3.4 E12-F040 RHR A TO RADWASTE DN STREAM ISOL VALVE and E12-2 F049 RHR A TO RADWASTE UP STEAM ISOL VALVE can be if the Steam Condensing Mode of RHR is not utilized. (Ref. ,
| |
| 2.3.5 If perfonning this procedure fmm the remote shutdown panels, E12-F006A RHR pump A(B) SDC SUCTION VALVE (S) will operate as throttle ves. (
| |
| 2.3.6 Valves E12-MOVF087A and B, E12-MOV026A and B, and E12-PVF051A and B have been electrically disconnected per MR 88-0208. (Ref. 7.20 &
| |
| 7.40) 2.3.7 RHR LPCI function cannot be performed effectively in the event of a LOCA if the following valves are open. (Ref. 7.50) 1
| |
| : 1. E12-VF099A(B) RHR A(B) RETURN TO FUEL STORAGE AREA.
| |
| : 2. E12-VF044(B) RHR A(B) RETURN TO UPPER POOL FILL 2.4 Operations 2.4.1 The RHR Shutdown Cooling Isolation at Ixvel 3 is automatically bysassed if control of E12-F008 RHR SHUTDOWN COOLING OUTBD ISOL VALVE and E12-F009 RHR SHUIDOWN COOLING INBD ISOL VALVE is transferred to the Remote Shutdown Panel. Use extreme caution if contml of these valves is transferred to the Remote Shutdown Panel. (Ref.
| |
| 7.45)
| |
| O SOP 0031 REV - 17B PAGF' 6 & 107 -
| |
| | |
| ( SRO EXAM KEY Esam Number NRC-SRO Rev. 0 Exam Title NRC SRO EXAM 86 A LOCA event inside primary containment is in progress. Level has been restored per the EOPs.
| |
| The llPCS diesel is running unloaded and the local operator reports thick smoke coming from the diesel. The CRS directs the U.O. to immediately shut down the diesel. Which of the following actions will successfully shut down the diesel ,
| |
| : n. Take the DIESEL ENGINE CONTROL switch to STOP.
| |
| : b. Depress the EMERGENCY STOP pushbutton,
| |
| : c. Reset the llPCS initiation signal then take the DIESEL ENGINE CONTROL switch to STOP.
| |
| : d. Place the local ENGINE MODE CONTROL switch in MAINTENANCE.
| |
| ANSWER:
| |
| : b. Depress the EMERGENCY STO: ;w outton.
| |
| IDNO: LP# OILI #
| |
| 61 ilLO-075 10 PROCEDURE NUMBER: OTHER:
| |
| SOP 0052 LEVEL 2 I NkdKd! I''ROh l SRO:
| |
| !264000 K4 02 4' 4.2 COMMENTS: 7/97 new b go
| |
| | |
| ... __ 4. Generator reverse power.
| |
| : 5. Generator overcurrent
| |
| : 6. Engine low lube oil pressure
| |
| : 7. Engine high temperature (jacket cooling water)
| |
| : 8. Engi2e overcrank cycle 220 seconds with no start 2.1.21 Anytime the diesel engine has tripped automatically, the operator is required to depress the "STOP" pushbutton on IE22'PNLS001 engine panei or by using the control switch on lH13*P601, (depending on where control is).
| |
| This will drop out the K1 relay which seals in upon a start. Fai'urs to do so will result in an automatic restart of the engine when the trip condition clears.
| |
| 2.1.22 Depressing the local or remote EMERGENCY STOP pushbutton will cause a diesel trip even with an emergency start signal present. This pushbutton should orily be depressed under extreme emergencies. For normal operation the local or remote STOP pushbutton should be used.
| |
| 2.1.23 The IDTM'V62 TURBOCHARGER AIR BOX DRAIN VALVE must be maintained thiottled open during all operations involving running of the diesel engine.
| |
| i f 2.1.24 When the diesel is running in parallel with the grid a fault on the grid could cause a loss of the bus associated with the diesel concurrent with a trip!!ockout of the diesel. To reduce the chances of this occurring minimize the time spent with the diesel paralleled to the grid.
| |
| 2.1.25 Operating data pertaining to diesel generator start attempts shall be obtained per PEP-0026. DIESEL GENERATOR TRENDING AND FAILURE REPORTING.
| |
| 2.1.26 During diesel fuel oil unloading, a fire watch shall be stationed at diu unloading area and two (2) 150 pound dry chemical ext inguishers placed near unloading area.
| |
| 2.1.27 If the diesel receives an inadvertent auto start signal that is determined not to represent a valid emergency start condition, immediately depress the local or remote EMERGENCY STOP pushbutton then place the local ENGINE CONTROL switch in the MAINT position. After the signal source has been identified and the auto start circuits reset, depress the local or remote STOP pushbutton prior to placing the local ENGINE CONTROL switch in the AUTO podtion when retummg the diesel generator to a standby mode.
| |
| Valid auto start signals are indicated as follows:
| |
| : 1. An ECCS auto start signal is identified at IH13*P601 by the white indicating light at the HPCS INITIATION RESET pu6 button being illummated and at IE22'PNLS001 by the white indicating light above the LOCKOUT RELAY being off.
| |
| SOP-0052 REV - 12 PAGE 6 OF 58
| |
| | |
| 9
| |
| .SRO EXAM KEY Q Exam Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM 87 . Given the following conditions:
| |
| - 1he plant is shutdown making preparations for Shudown Cooling (SDC) using the "A" loop of RilR Both Recirculation Pumps are shutdown with their discharge valves closed Which of the following describes how the "A" RIIR Pump that is being staryd for SDC is protected frota damage due to no flow?
| |
| : a. The operator is required to establish a pump discharge flow-path to the reactor as soon as
| |
| - possible after starting the pump.
| |
| : b. The pump minimum flow valve (F064 A) will open to provide flow until the RilR fleat Exchanger Bypass Valve (F048 A) can be opened.
| |
| c.1he operator will open the minimum flow valve (F064 A) until shutdown cooling flow is greater than 500 gpm.
| |
| : d. The pumr. will automatically trip on low suction pressure if flow / pressure is not adequate for pump sectio: .
| |
| ANSWER:
| |
| \ a. The operator is required to establish a pump discharge flow-path to the reactor as soon as pmsible aller stamng the pump.
| |
| IDNO: LP# OBJ #
| |
| 411 HLO-021 6 PROCEDURE NUMBER: OTHER:
| |
| SOP-0031 LEVEL 3
| |
| ;G 2.1.2 3 4 _j COMMENTS: 7/97 new also HLO-021, OBJ. 8 O n
| |
| -wm w,m , - w- > -+ -- - - - - e rp
| |
| | |
| CAtmON Failure to establish greater than 1100 unn within 8 secorxk of pump start will catse 1E12*FD64A(B) RI.R PUMP A(B) NIN FLOW TO SUP PL to open, dumping reactorwater to the suppression pool.
| |
| NOIE The respective rmit cooler (s) shall be in service prior to starting RHRpmip(s)for sinttdomr cooling. (Ref 7.17)
| |
| CAlmON z The initiation of Shutdown Cooling nmy result in more pronounced notching of indicated reactor level. Monitor reactorlevel treruk. (reference 7.54)
| |
| : 10. Start RHR PUMP A(B) and IMMmlATELY throttle open IE12*FM8A(B) RHR A(B) HX BYPASS VALVE to obtain 2000 -
| |
| 3000 gpm.
| |
| I1. Establish a stable flow of - ximately 4000 - 5000 gpm by throttling 1E12*FM8A(B) RHR A(B BYPASS VALVE. (Ref. 7.31) n 12. Throttle open IE12*F003A(B) RHR A(B) HX OUfLET VALVE to V approx.10% indication.
| |
| CMLilOti RIIR HX temperature indicatiors E12-R601 FT 1(2) do not show actual water temperatures with IS12-FUO3A(B) fully closed. Wth E12-FD03A(B) fully closed, tse E12-R60'l PT 5(6) for coolant temperature indiration.
| |
| : 13. Establish a cool down rate ofless than 100 DEFGhr as follows:
| |
| : a. Slowly _.j'og open IE12*F003A(B) RHR A(B) HX OUTLET VALVE and monitor the cooldown rate.
| |
| : b. Throttle IE12*F003A(B) RHR A(B) HX OUILET VALVE and IE12*F048A(B) RHR A(B) HX BYPASS VALVE to obtain the desired cooldown rate while maintaining a constant RHR loop flow.
| |
| O SOP-0031 REV - 17B PAGE 27 OF 107
| |
| | |
| g Esam Number NRC-SRO SRO EXAM KEY Esam Title NRC SRO EXAM
| |
| -Rev. O_
| |
| 88 Plant conditions are as follows:
| |
| Suppression pool temperature 87 degrees F
| |
| - Suppression pool level 20.5 fl Drywcli temperature 125 degrees F
| |
| - Reactor level 10.5 inches
| |
| - Aux. Bldg. pressure -0.25 -
| |
| Which one of the following EOP selections should be entered?
| |
| : a. EOP 1 and EOP 2
| |
| : b. EOP 2 only
| |
| - c. EOP 2 and EOP 3
| |
| : d. EOP 3 only ANSWER:
| |
| : b. EOP 2 only O IDNO: LP# OBJ #
| |
| 374 IILO-514 3 PROCEDURE NUMBER: OTHER:
| |
| EOPM2 LEVEL 2
| |
| ~.- . - - . . - . . . .
| |
| 10 2.4.16 3 4 j COMMENTS: 7/97 new EOP-0002 RO & SRO T4 04
| |
| . I-G 92
| |
| | |
| EOP-2 Primary Containment Control O - Entry Conditions U ,
| |
| Ky 3
| |
| ~
| |
| * ENTRY CONDITIONS :
| |
| When any of the parameters exceed the values listed in the entry conditions, E0P.2 must be entered Each enty condition is explained below.
| |
| - Drywell temperature above 145' F This entry condition addresses the ability to control drywell temperature to prevent exceeding the drywell design temperature of 330*F. Ntaintaining drywell temperature within acceptable values will alsa minimize errors in RPV water level indications, caused by instru;nent reference leg density changes. The value of 145'F was chosen as it is cas,ily identifiable and is the Technical Specification LCO.
| |
| for drywell temperature. ,
| |
| - Containment temperature above 90'F s OV This entry condition, like the drywell temperature entry condition, addresses the ability to control containment temperature to prevent exceeding the containment temperature d, sign limit of 185'F. Ntaintaining containment temperatures will also minimize RPV water level instmment errors, as instrument runs are also located inside the containment. The setpoint of 90*F is the LCO for containment temperature.
| |
| - Containment pressure above 0.3 psig This entry conditior addresses the ability to control containment pressure in order to prevent the loss of primary containment integrity due to overpressure conditions. The value of 0.3 psig was chosen since it is the LCO for containment pressure, and is sufficiently below the pressure at which systems used to control containment pressure isolate.
| |
| - Suppression pool temperature above 100*F This entry condition addresses control of suppression pool temperature in order to maintain the pressure suppression function of the primary containment, provide adequate NPSH for pumps taking a suction on the suppression pool, and to prevent exceeding the design temperature limits of the suppression pool. The value O
| |
| v of 100*F was chosen since it is the LCO for suppression pool temperature EPSTG-0002 B - 139 Revision 3 d
| |
| | |
| EOP 2 Primary Containment Control Entry Conditions
| |
| ((
| |
| C . Suppression pool water level below 19 ft. 6 in.
| |
| This entry condition addresses control of suppression pool water level to prevent the failure of the primary containment due to direct pressurization when the pressure suppression function of the suppression chamber is degraded or lost. This limit also assures adequate NPSH for pumps which take a suction on the suppression pool. The value of 19 ft. 6 in. was chosen since it is the LCO for minimum suppression pool water level.
| |
| . Suppression pool water level above 20 ft.
| |
| This entry condition addresses control of high suppression poc ; ater level to prevent primary containment failure due to static and dynamic ! .as and to prevent covering primary containment vent paths. The setpoint of 20 ft. was chosen as it is the LCO for maximum suppression pool water level.
| |
| O '
| |
| w.
| |
| G O
| |
| B 'A 0 Revision 3 EPSTG'0002
| |
| | |
| I O SRO EXAM KEY G-Exam Number NRC-SRO Rev. 0 Esam Title NRC SRO EXAM 89 Which of the following permission' notification requirements must be met for an INTENTIONAL entry into Tech Spec 3.0.37 Permission must be obtained from the:
| |
| : a. Operations Superintendent and the NRC Resident inspector notified.
| |
| : b. General Manager - Plant Operations and the NRC Resident inspector notified.
| |
| : c. Manager Operations and a 1-hour report made to the NRC.
| |
| : d. General Manager - Plant Operations and a 4-hour report made to the NRC.
| |
| ANSWER:
| |
| : a. Operations Superintendent and the NRC Resident inspector notified.
| |
| IDNO: LP# OBJ #
| |
| 423 ilLO-206 2 PROCEDURE NUMBER: OTHER:
| |
| (
| |
| i ADM-0022 LEVEL 2 fi ~NRC KAi' I~ RO- 'I SROT G 2.* 17 > 3.4l 3.8 _
| |
| COMMENTS: 1/97 NRC exam
| |
| ?
| |
| %s' 93 2
| |
| | |
| 5.0 GENERAL ,
| |
| ()-
| |
| V NONE~
| |
| 6.0 PROCEDURE NOTE -
| |
| The actions oj thisprocedure may be completedin any sequence, hon ever, the sequencepresentedis recommended 6.1 Upon declaration of an emergency classification, the Recovery Manager should:
| |
| NOTE Nonpcation must be made within 15 minutes ofa declaration of an emergency, emergency exalation. or -
| |
| Protective Action Recommendation (PAR) change.
| |
| 6.1.1 Direct the designated communicator to report to his/her position.
| |
| O V
| |
| 6.1.2 Complete the short Notification Message Form (NhE) for each new emergency classification or changes in PARS using Attachment 1.
| |
| 6.1.3 Direct the Communicator to notify both States and local authorities using the short NhE.
| |
| 6.1.4 Direct the initiation of the activation and augmentation of the Emergency l Response Organization.
| |
| 6.1.5 Direct the notification of the Nu:: lear Regulatory Commission (NRC) Operations Center as soon as possible after notifying the State and local authorities and not later than one hour after declaring the emergency using the Emergency Notification System (ENS).
| |
| 6.1.6 Ensure that follow up notifications are made to offsite authorities approximately every 60 minutes or anytime significant changes in emergency conditions occur using a long Nh6 (Attachment 1). Refer to Attachment 2 for directions on how to fill out the long NMF Significant emergency condition changes include, but are not limited to, onset of radioactive release, loss of cooling to the reactor vessel, major plant fire, security threat, etc.
| |
| 6.1.7 Upon termination of the emergency, ensure that notifications are made to the offsite authorities, using the Long Noti 6 cation Message Form..
| |
| EIP-2 006 REV. - 21 PAGE 3 OF 18 l
| |
| | |
| 4.2.4 authorize information for the media and the grneral public prior to release from the Joint Information Center (JIC), and
| |
| 's - 4.2. 5 terminate the emergency.
| |
| 4.3 Emergency Director:
| |
| 4.3.1 assess and classify emergency conditions, 43.2 authorize doses in excess of 10CFR20 limits, and 4.3.3 direct onsite protective and corrective actions.
| |
| 5.0 GENFRAL 5.1 For the Notification of Unuaaal Event classification, the emergency response can usually be handled by shift personnel without additional support or activation of emergency response facilities.
| |
| 5.2 Prompt notification means yithin approximately 15 minutes. The time is measured from the time at which the Operations Shift ,.
| |
| SupMntendent recognizes that events have occurred which make declaration of an emergency class appropriate. .
| |
| O O
| |
| EIP-2-002 REV - 15 PAGE 3 OF 15 l
| |
| | |
| 6.0 PROCEDhRE NOTE The actions of this procedure may be completed in any sequence, however, the sequence presented is recommended -
| |
| NOTE If the Control Room is evacuated the Operations Shift Superintendent, designated Communicator, and Chemistry l Technician shall report to the TSC toperform the actions ofthisprocedine. .
| |
| 6.1 Initial Actions - The Emergency Director should use Attachment 1 - 3 as a guideline and:
| |
| 6.1.1 Merge the Page, Party /Gaitroides System and inform plant personnel of the emergency. -
| |
| 6.1.2 Direct the activation of the pagers in accordance with EIP-2-006.
| |
| ~
| |
| 6.1.3 Direct prompt notifications to offsite authorities and the Nuclear Regulatory Commission in accordance with EIP-2-006. Recovery Mraager performs O. offsite notifications when EOF is operational.
| |
| 6.1.4 If necessary, direct evacuations in accordance with EIP-2-026.
| |
| AT AN ALERT EMERGENCY OR HIGHER CLASSIFICATION 6.1.5 Direct the Security Shift Supenisor to:
| |
| : 1. send one Security Officer to the EOF and one to the IIC for access control, and
| |
| : 2. activate the card reader in the Operations Support Center and Technical Support Center, if not previously done.
| |
| 6.1.6 Direct a Chemistry Technician to activate the Emergency Response Data System (ERDS) located in the TSC Computer room, if not previously done.
| |
| ERDS should be activated as soon as possible within I hour of an Alert or higher emergency classification level.
| |
| 6.1.7 Dispatch appropriate personnel to implement EIP 2-014, as necessary.
| |
| EIP-2-002 REV - 1S PAGE 4 OF 15 l
| |
| | |
| SRO EXAM KEY O Etam Number NRC.SRO Rev. O. Esam Title NRC SRO EXAM 90 ne Plant is operating at 100% reactor power when a loss of feedwater heating occurs. Which one of the following is a required IMMEDIATE action for this loss of feedwater heating?
| |
| : a. Reduce reactor power by 40 MWE with core now, then reduce another 110 MWE with core now and rod insertion.
| |
| : b. Reduce power to less than or equal to 100% rated thermal power using sore Cow,
| |
| : c. If failed fuel exists in the reactor, reduce reactor power by 495 to 500 MWE.
| |
| : d. Insert control rods in reverse order to get below the 80% rod line.
| |
| ANSWER:
| |
| b Reduce power to less than or equal to 100% rated thermal power using core now.
| |
| IDNO: LP# OILI #
| |
| 375 llLO-526 4 PROCEDURE NUMBER: OTHEst:
| |
| AOP-0007 LEVEt. 2
| |
| ( . -.
| |
| -G 2.4.49 4! 4 __l COMMENTS: 7/97 new AOP 0007, Rev 8, p. 2B of 6 l RO & SRO T3 G4
| |
| -_ - - - ~ _ _
| |
| 'J 94
| |
| | |
| e IFWR-FV2A(B)(C) RX FWP PI A(B)(C) MIN FLOW opening, e Open or stuck open SRVs and/or Turbine Bypass Valves.
| |
| * Loss of 1HDL-PI A(B)(C)(D), HTR DR PUMP 1 A(B)(C)(D).
| |
| 2.3 Reactor power rise with no change in recirculation flow or control rod positions.
| |
| 3 AUTOMATIC ACTIONS 3.1 Possible control rod block and scram due to high neutron flux.
| |
| 4 IMMFDIATE OPERATOR ACTIONS 4.1 Maintain reactor power less than or equal to 100% rated thermal power using core flow.
| |
| 4.2 Reduce core flow to less than or equal to 100% of rated core flow (84.5 Mlbs/hr).
| |
| 5 SUBSEOUENT OPERATOR ACTIONS 5.1 Monitor feedwater temperature using any of the following:
| |
| 5.1.1. Process computer video service No.3.
| |
| D 5.1.2. Balance of Plant Parameter or Process points:
| |
| * B21NA002 e B21NA003
| |
| = B21NA004 e B21NA005 AOP-0007 REV - 11B PAGE 4 OF S
| |
| | |
| r
| |
| \
| |
| SRO EXAM KEY Enam Number NRC SRO Rev. O Exam Title NRC SRO EXAM 9l The RCIC System has just been declared inoperable, the reactor is operating at 00% power.
| |
| 8 Which system must be demonstrated to be operable to continue operation in thi; condition?
| |
| : a. IIICS
| |
| : b. ADS
| |
| : c. LPCS
| |
| : d. RilR LPCI Division 11 ANSWER:
| |
| : a. IIPCS IDNO: LP# OILI #
| |
| 413 ilLO-017 18 PROCEDURE NUMBER: OTHER:
| |
| Tech. Spec. 3.5.3 LEVEL 3
| |
| ~
| |
| I NRC kAi l ~ RO: I' ~SROl~
| |
| 3.8
| |
| ]
| |
| 'O 2.1.11 ! 3- j COMMENTS: 7/97 new With RCIC system inoperable the HPCS system must be demonstrated operable by administrative means within one hour.
| |
| l O.- 93
| |
| | |
| RCIC System 3.5.3 p
| |
| 5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOL (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS CONDITION REQUIRED ACTION
| |
| . COMPLETION TIME A. RCIC System A.1 ' Verify by 1 hour inoperable, administrative means High Pressure Core Spray System is OPERABLE.
| |
| AND A.2 Restore RCIC System 14 days to OPERABLE status.
| |
| B. Required Action and 8.1- Be in MODE 3. 12 hours
| |
| , associated Completion Time not met. 6@
| |
| B.2 Reduce reactor steam 36 hours dome pressure to s 150 psig.
| |
| v)
| |
| RIVER BEND 3.5-10 Amendment No. 81 l1
| |
| | |
| l l
| |
| l SRO EXAM KEY l Exam Number NRC SRO Rev. 0 Esam Title NRC SRO EXAM 92 Prior to reactor startup the following SRM count rates are recorded:
| |
| l SRM A 25 cps SRM B 30 cps 4
| |
| SRM C 35 cps SRM D 15 cps At what SRM reading should the operator consult the Reactor Engineer for continued withdrawal i vommendations7
| |
| : a. 2500 cps on SRM A
| |
| : b. ' 3000 cps on SRM D
| |
| : c. 30') cps on SRM C
| |
| : d. 240 cps on SRM D ANSWER:
| |
| : d. 240 cps on SRM D O
| |
| IDNO: LP# OHJ #
| |
| 57 STM 503 07 PROCEDURE NUMBER: OTHER:
| |
| GOP 0001 LEVEL 3
| |
| 'O 2.1.7 3 4 __ ,
| |
| COMMENTS: 7/97 new Requirement to contact RE if af ter 4 count rate doublings (from initial) the reactor is not critical.
| |
| | |
| PAGE 22 OF 48 PERFORMANCE PACKAGE 4 STEP INITIAi C)
| |
| CAUTION Do not attempt to recouple a control rod by scramming If an out of sequence rod should be withdrawn, suspension of all control rod motion, notification of the Operations Shift Superintendent and reference to REP-0051 is required.
| |
| Extreme caution should be exercised when performing operations which may affect reactor pressure and/or reactor coolant temperature to prevent inadvertent reactivity excursions. (Ref. 2.3)
| |
| Normally the first rod in a group has a potential for high rod worth. (Ref. 2.29)
| |
| High xenon concentration can alter normal rod worth configurations, especially in the peripheral regions. (Ref. 2.29)
| |
| The operator should monitor all SRMs closely during the iO approach to criticality so as to have the best possible data at his disposal.
| |
| NOTE When control rod (s) reaches (reach) position 48, a coupling check shall be performed. When attempting to withdraw the rod (s)pastposition 48 the CONTROL ROD OVERTRA VEL (P680-07A-C02) alarm should not annunciate. (SR 3.1.3.5) (Ref 2.10)
| |
| As each control rod is withdrawn, observe discernible response ofNuclear Instrumentation to check that control rods are coupled (Ref 2.1G)
| |
| Activities that can distract operators and supervisors involved with the reactor startup, such as shift turnover and surveillance testing during the approach to criticalityshouldbe avoided (Ref 2.20)
| |
| : 9. Begin pulling control rods in sequence as specified by the Rod Movement Sequence Package.
| |
| : a. When the first SRM reaches the value calculated in Section C step 7.b notify the Reactor Engineer that Four Count Rate Doublings have been reached.
| |
| Obtain recommendations from the Reactor Engineer on continued rod movement.
| |
| GOP-0001 REV - 20E PAGE 27 OF 90
| |
| | |
| .. - -,- = - - - . - --_ . .-. . . - . .
| |
| g Exam Number NRC-SRO SRO EXAM KEY Rev. 0 Esam Title NRC SRO EXAM 93 The plant is starting up following a refueling outage. "Ihe reactor hasjust acheived criticality.
| |
| Which one of the following statements is true regarding the requirement for Shutdown Margin (SDM) determination?
| |
| : a. SDM must be determined within frer hours of criticality,
| |
| : b. SDM must be determined before proceeding further with the startup.
| |
| : c. SDM need not be determined if no control rods were replacsi
| |
| : d. SDM need not be determined ifit was determined analytically following the last fuel movement.
| |
| ANSWER:
| |
| : a. SDM must be determined within four hours of criticality.
| |
| IDNO: LP# OBJ #
| |
| 412 IILO-412 1 PROCEDURE NUMBER: OTHER:
| |
| TS 3.1.1 LEVEL 4
| |
| 'O 2.2.12 3 4 COMMENTS: 7/97 new T.S. 3.1.1, SR 3.1.1.1
| |
| .t N- 97
| |
| | |
| . . ~ - . - -- - _ - - - _ _ - . _ - . - . . . . - . . . . . . - . . . . _-. -
| |
| SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.4 --------NOTE---------
| |
| Entry and exit is permissible under administrative control.
| |
| Initiate action to I hour close one door in each primary containment air lock.
| |
| ,- SURVEILLANCE REQUIREMENTS SURVEILLANCE O
| |
| FREQUENCY SR 3.1.1.1 Verify SDM is: Prior to each in-vessel fuel
| |
| : a. m 0.38% Ak/k with the highest worth movement during control rod analytically determined; fuel loading or sequence
| |
| : b. m 0.28% Ak/k with the highest worth AND control rod determined by test.
| |
| Once within 4 hours after criticality following fuel movement within the reactor pressure vessel
| |
| . or control rod replacement f
| |
| : O
| |
| , RIVER BEND 3.1-3 Amendment No. 81
| |
| : b. , SRO EXAM KEY Esam Number NRC-SRO Rev. 0 Exam Title NRC SRO EXAM 94 The basis for Surveillance Requirement 3.1.5.1 for each control rod scram accumulator pressure to be verified > 1520 psig every 7 days is to ensure adequate accumulator pressure exists to:
| |
| : a. provide sufficient scram force.
| |
| : b. drive control rods on a loss of CRD pumps.
| |
| : c. maintain indication in the readable range on the gauge.
| |
| : d. verify accuracy of control room IICU pressure indications.
| |
| ANSWER:
| |
| : a. provide sufficient scram force.
| |
| IDNO: LP# OBJ #
| |
| 414 STM 052 8 PROCEDURE NUMBER: OTHER:
| |
| TS 3.1.5 LEVEL 2
| |
| '2.2.25 . 2.51 3.7 COMMENTS: 7/97 new SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator operability is the accumulator pressure.
| |
| 98
| |
| | |
| Control Rod Scram Accumulators-B 3.1.5 BASES ACTIONS {L1 (continued)
| |
| The reactor mode switch must be immedi'ately placed in the shutdown position if either Required Action and associated Completion Time associated with the loss of the CRD pump (Required Actions B.1 and C.1) cannot be met. This ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the. active function (i.e., scram) of the control rods. This Required Action is modified by a Note stating that the Required Action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.
| |
| SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the O,
| |
| capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 1520 psig is well below the expected pressure of 1750 psig (Ref. 2).
| |
| Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.
| |
| REFERENCES 1. USAR, Section 4.3.2.5.5.
| |
| : 2. USAR, Section 4.6.1.1.2.5.3.
| |
| : 3. USAR, Section 5.2.2.2.3.
| |
| : 4. USAR, Section 15.4.1.
| |
| RIVER BEND B 3.1-31 Revision No. O
| |
| | |
| -- .- . - , - - - _ - ... -- - - . . . . =
| |
| 0 SRO EXAM KEY Esam Number NRC-SRO Rev 0 Eram Title NRC SRO EXAM 95 in which of the following conditions is the Operation Shift Superintendent allowed to utilize Concurrent Verification as a substitute for independent Veri 6 cation?
| |
| : a. A Danger lloid tag is being removed from a Spent Fuel Pool Cooling & Cleanup Valve located in a contamination area. The valve's restoration position is " Locked Closed"
| |
| : b. Completing an SOP Vlave Lineup during a refueling outage on an instrument Air valve located in Containment. The valve's position is "Open".
| |
| : c. Conducting a routine OSP-0017 Control Board Lineup afler running RilR "A" in Suppression Pool Cooling and rejecting pool water to Radwaste.
| |
| : d. An SNEO trainee is perfroming an STP on the Diesel Driven Fire Pump. The trainee signs tha "perfromed by" steps and the on-shift SNEO signs the
| |
| " independent verifier" steps.
| |
| ANSWER:
| |
| : a. A Danger lloid tag is being removed from a Spent Fuel Pool Cooling & Cleanup Valve located in a contamination area. The valve's restoration position is " Locked Closed".
| |
| k IDNO: LP# OBJ #
| |
| 425 llLO-201 2 PROCEDURE NUMBER: OTHER:
| |
| ADM@76 LEVr.L 2
| |
| !G 2.2.13 3.6i 3.8 _
| |
| COMMENTS: 7/97 new O 99
| |
| | |
| P 6.7 d Concurrent' Verification Method 6.7.1.
| |
| The Concurrent Verifier should maintain contact with the Perforter 'an be in a position to observe the manipulations or indications of manipulations as '
| |
| they are perfonned.
| |
| 6.7.2. ,
| |
| The Performer shall then complete the intended component manipulation while ;
| |
| the Concurrent Verifier observes the manipulation er indications of manipulations.
| |
| 6.7.3. t Actions requiring Concurrent Verification are to be signed off by the Performer and the Concurrent Verifier prior to proceeding to the next step in the procedure. '
| |
| 6.7.4.
| |
| The following additional requirements exist if the concurrent verification is being used to avoid a plant transient (i.e. ESF actuation, RPS trip, turbine trip, etc.).
| |
| 1.
| |
| The Performer shall establish contact with the device to be manipulated, and announce his intentions (open this switch, select A & B position, etc.)
| |
| to the Concurrent Vezifier.
| |
| (D G'
| |
| 2.
| |
| The Concurrent Verifier shall, after reviewing the document requiring the component to be repositioned, concur with the component selection and planned manipulation.
| |
| 6.7.5.
| |
| Concurrent Verification may be utilized as a substitute for Independent Verification under the following conditions:
| |
| 1.
| |
| l The manipulations are located in areas where a two-person ru!c is in effect for safety considerations.
| |
| l 2.
| |
| Manipulations are located in high dose areas or surface contrminadon areas where exposure would be minimimi by use of Concurrent Verification.
| |
| 3.
| |
| Manipulations are located in areas ofdifficult or restricted access where the physical movements n~"my to remain out of visual and verbd contact are not practical and may prove more hazardous or less reliable than rernmining in contact.
| |
| ! 4.
| |
| I The Verifier would .have to repeat the action thus negating the action of the Perfonner.
| |
| ; O. S.
| |
| V A comment shall be entered in the document or Control Room Log to indicate which verifications were substituted and why.
| |
| ADM-0076 REV-2 PAGE 13 OF 19
| |
| | |
| vO SRO EXAM KEY Esam Number NRC SRO Rev 0 Esam Title NRC SRO EXAM 96 in EOP-I A, ATWS RPV Control, if SRVs are cycling, the operator is directed to manually open SRVs until RPV pretsure drops to 930 psig.
| |
| Which of the following is the reason for stopping the reactor pressure reduction at 930 psig?
| |
| : a. To ensure the turbine bypass valves do not have the oppprtunity to stick closed
| |
| : b. To prevent MSIVs from closing on low main steam line pressure
| |
| : c. To minimize the amount of steam that is sent to the suppiession pool
| |
| : d. To prevent excessive loss of reactor coolant inventory ANSWER:
| |
| : c. To minimize the amount of steam that is sent to the suppression pool IDNO: LP# Olki#
| |
| 313 IILO 513 4 PROCEDURE NUMBER: OTHER:
| |
| O EOP 1A EPSTG'0002 LEVEL 3 i NRC KA: l Rdi'l SRU[
| |
| }295037 EK2.1_0_. J.__ 3.8; _ 4.1 j 3.1i 4 j
| |
| :G 2.4.6 COMMENTS: 7/97 08W 100
| |
| | |
| EOP-I A RPV Control . ATWS . RPA o ..
| |
| STEP RPA.6 Step RPA 6 directs the operator to stabilire pressure between 1060 psig and 930 psig.
| |
| l This step reduces or eliminates energy deposition to the suppression pool by ensuring pressure is l
| |
| below that point where SRVs will open It also assures that SRVs will not continuously cycle i open on their lift pressure, then shut on their reset pressure.
| |
| Subsequent steps will direct pressure control actions to assure the minimum amount of heat is rejected to the suppression pool.
| |
| O ,
| |
| f t
| |
| d O
| |
| EPSTG'0002 B - 105 Resision 3
| |
| | |
| SRO EXAM KEY !
| |
| I:sa a Number NRC SRO Rev. O I:sim Title NRC SRO EXAM 97 W plant was operating at 100% rated power when a main turbine trip occurred. Plant conditions j are as follows I
| |
| RPV wat .t level is + 18" and steady RPV Pressure is 950 psig and beis g maintained by the bypass valves between 800 1000 psig.
| |
| Reactor Recirculation Pumps are running in slow speed after an aute natic transfer.
| |
| Which of the following identifies the signal which caused the transfer of the recirculation pumps to slow 7
| |
| : a. RPV water level 4 coincident with with a trip of a running RI'P.
| |
| : b. Turbine Stop Valve closure (< 95% open) or Turbine Control Valve closure '
| |
| (Low ETS pressure)
| |
| : c. RPV water level 2 ( 43")
| |
| : d. liigh RPV pressure (I127 psig) ;
| |
| ANSWi:Rt ,
| |
| : b. Turbine Stop Valve closure (< 95% open) or Turbine Control Valve closure (Low ETS pressure)
| |
| IDNO: 1.P # 01L1 #
| |
| 408' STM 053 2b ,
| |
| PROCEDURE NUMBER: OTHER:
| |
| A OP 0024 LEVEL 3 f ' ' NRC MAi 'l RO: l SR0[
| |
| !?96006 AK3 o6 i 3 21 3.3 COMMENTS: 7/97 new LOTM 16 N 101
| |
| .a* w- -+ y - - - + - -
| |
| | |
| p, 3. 'B33 F067A(B) RECIRC PL'MP A(B) DISCH VALVE less than 90% opea V
| |
| : 4. LFMG Lockout Relay (86 0; operation
| |
| : 5. CB 2 trip push button trip
| |
| : 6. CB 1 open
| |
| : 7. LFMG volt regulator contacts open 2.1.3. Shift to slow speed
| |
| : 1. Total teed flow less than 25% for 15 seconds.
| |
| : 2. Reactor water level at i evel 3.
| |
| : 3. End of Cycle Recirc Pump Trip. Turbine stop valve, control valve closure with reactor power greater than 40%.
| |
| : 4. Steam dome temperature to loop suction temperature differentialless than 8'F for 15 seconds. (For affected loop if CB 5 in other loop is not closed.)
| |
| 2.1.4. Decrease in reactor power / generator load with a potential for reactor scram /turbme trip.
| |
| 3 AUTOMATIC ACTIONS None 4 iMMFDIATE OPERATOR ACTIONS 4.1 Manually SCRAM the reactor if MQ recirculation pumps are operating with the MODE SWITCH in RUN.
| |
| AOP-0024 REV - 14 PAGE 5 OF 14
| |
| | |
| _ . - . . . - . . . . - . . _ - - . - . . . , . . - - ~ _ . . - _ _ _ _ - - . . _ . - - - - . . . . - . . . .
| |
| l SRO EXAM KEY ;
| |
| Esam Number NRC.SRO Rev. 0 Esem Title NRC SRO EXAM i
| |
| es A plant transient has occurred =t!ag e complete isolation of the RWCU systern.
| |
| . RWCU inboard and outboard isolation valves are closed.
| |
| (G33 F001/FM4/F028/l034'F039/F040/F053/F0$4)
| |
| Which of the following conditions caused this RWCU isolation?
| |
| : a. liigh Drywell Pressure initiation signal
| |
| : b. Low Reactor Water Level 2
| |
| : c. liigh Main Steam Tunnel Differen% mnperature
| |
| : d. Initiation of SLC "A" system ANSWl:Rt
| |
| : b. Low Reactor Water Level 2 IDNO: LP# 010 # !
| |
| ~
| |
| 409 11L0-062 4 !
| |
| PROCEDURE NUMBER: OTHER:
| |
| AOPEK)3 LFVEL 3 i
| |
| F ~ 'NRC K Ai~ l~R0i l"SR0i i 296020 AA2.06 '
| |
| 3.4; 3.8
| |
| !2.4.4 4I~ ~ 4.3 COMMENTS: 7/97 new E
| |
| 102
| |
| ._ . . . , . _ _ . . . , _ . . . . _ - - , _ - _ _ _ - - _ _ . . _ . . . _ . . _ .___ .- _. _ -_ _ ___ _._. _ . . _ . ~ . .. _ . . _ _ _ -
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| | |
| PAGE I OF 4 SIGNALTO ACTUATION / ISOLATION RELATIONSHIP SIGNAL ACTUATIONS AND ISOLATIONS O' A Reactor Vessel Water Level Low: LEVEL 3: 9.7 inches I
| |
| l
| |
| : 1. Group 5 and 14 valves isolate. l
| |
| : 2. H13.P601 Post Accident Recorders shift to fast speed. l B Reactor Vessel Water Level Low Low: LEVEL 2: -43 inches )
| |
| : 1. Group 1,7,8,9,15, and 16 valves isolate. 1
| |
| : 2. Group 11,12, and 13 dampers isolate.
| |
| : 3. Standby Gas Treatment System auto initiates.
| |
| : 4. Annulus Mixing System auto initiates.
| |
| : 5. Fuel Bldg Exhaust Filtration Trains A & B auto initiate.
| |
| : 6. Containment Hydrogen Analyzer / Monitor System starts.
| |
| : 7. Control Room HVAC Emergency Mode auto initiate.
| |
| C Reactor Vessel Water Level - Low Low Low: LEVEL 1: 143 inches
| |
| : 1. Group 6 and 10 valves isolate.
| |
| : 2. HVR UCI A and UCIB, CONTAINMENT UNIT COOLERS auto start and are supplied by the Service Water System after a 1 minute time delay,
| |
| : a. SWP MOV502A and 502B, CONTMT COOL SPLY open.
| |
| : b. SWP MOV503A and 503B, CONTMT COOL RTN open.
| |
| D Drywell Pressure High: 1.68 psid
| |
| : 1. Group 1,3,8,10, and 14 valves isolate.
| |
| : 2. Group 11,12, and 13 dampers isolate.
| |
| : 3. Standby Oas Treatment System auto initiates.
| |
| : 4. Annulus Mixing System auto initiates.
| |
| : 5. Control Room HVAC Emergency Mode auto initiates.
| |
| : 6. Fuel Bldg Exhaust Ventilation Trains A & B auto initiate.
| |
| : 7. Containment Hydrogen Analyzer / Monitor System starts.
| |
| : 8. HVR UCI A and UCIB, CONTAINMENT UNIT COOLERS auto start and are supplied by the Service Water System after a 1 minute time delay.
| |
| : a. SWP MOV502A and'502B, CONTMT COOL SPLY open.
| |
| : b. SWP MOV503A and 503B, CONTMT COOL RTN open.
| |
| O AOP-0003 REV - 10 PAGE 8 OF 17
| |
| | |
| A !TACHMENT t'-
| |
| PAGE 2 0F 4 SIGNAL TO ACTUATION / ISOLATION RELATIONSHIP SIGNAL ACTUATIONS AND ISOLATIONS E Main Steam Line Radiation High: 3 X Normal Full Power Background.
| |
| : 1. Group 9 valves isolate.
| |
| : 2. Condenser Air Removal Pumps trip and isolate.
| |
| F hiain Steam Line Pres:;ure - Low: 849 psig with Reactor hiode Switch in RUN
| |
| : 1. Group 6 valves isolate.
| |
| G hiain Steam Line Flow High: 140% Flow
| |
| : 1. Group 6 valves isolate.
| |
| ~
| |
| H Main Steam Line Tunnel Temperature High: 141*F
| |
| : 1. Group 2,6,7,15, and 16 valves isolate, J Main Steam Line Area Temperature High: 142'F,95 ft el Steam Tunnel Area 142*F, i14 ft el Steam Tunnel Area 108'F, Turbine Shield Wall Area-126*F, MSR Area
| |
| : l. Group 6 valves isolate.
| |
| K Condenser Vacuum - Low: 8.5 inches Hg VAC.
| |
| : 1. Group 6 valves isolate.
| |
| L RWCU Equipment Area Temperature High: 104.5'F in Heat Exchanger Room 165'F in Pump Rooms 110*F in Valve Nest Room 110'F in Demineralizer Rooms 110*F in Rec;iving Tank Room
| |
| : 1. Group 7,15, and 16 valves isolate.
| |
| N RWCU Differential Flow High: 55 gpm after a 45 second time delay
| |
| : 1. Group 7,15, and 16 valves isolate.
| |
| O Standby Liquid Control System initiation Pump A
| |
| : 1. Group 7 valve isolates. ,
| |
| N AOP 0003 REV - 10 PAGE 9 OF 17
| |
| | |
| .7 PAGE 5 OF 6 ISOLATION VALVE CIIECKOFF SIIEET ISOLA W DNI&lV RESTORfD PANFL ISOLA 11W DWH&IH RINIORID NilAIS OUIBOARD INmALS INmAIS INBOARD Nil 4IS lil3 P863 HVF AOD122 SOP 0062 GROUP 13 HVF-AOD101 SOP 0062 IIVF.AOD102 SOP 0062 SIGNALS HVF AODil2 SOP-0062 HVF.AOD104 SOP 0062 B,D AA HVR AOD137 SOP 0062 lil3 P601 E12 F037A SOP 0031 GROUP 14 E12 F037E SOP 0031 E12 F075A SOP 0031 SIGNALS E12 F060A SOP 0031 E12 F075B SOP 0031 A.D,V E12 F060B SOP 0031 E12 F040 SOP 0031 E12 F049 SOP 0031 Ill3 P601 G33 F004 SOP 0090 GROUP 15 G33 F001 SOP 0090 G33 F054 SOP 0090 SIGNALS G33 F053 SOP 0090 G33 F039 SOP 0090 B, H, L, N G33 F040 SOP 0090 G33 F034 SOP 0090 G33 F028 SOP-0090
| |
| , til3 P601 V GROUP 16 SIGNALS G33 F001 SOP 0090 B,H,L,N,P lil3 P863 INN 40V127 GROUPl7 HVN40V102 INNA0V128 SIGNAL EE HVN40V129 HVN40V130 HVR AOV166 N/A HVR-AOV128 N/A lil3 P870 SWP-MOV502A GROUPl7 S%P4 W 50:B SWP40V503A SIGNAL EE SWP4 W 503B AQP-0003 REV - 10 PAGE 16 OF 17
| |
| | |
| SRO EXAM KEY O Esam Number NRC.SRO Rev. 0 Esam Title NRC SRO EXAM 99 The plant is shutdown for a maintenance outage. Work is being performed on a portion of the l'eedu ater system by Me:hanical Maintenance. The I&C foreman has received a Clearanc e Receipt to uork within the feedwater system tagout boundary to calibrate an instrument. Upon completion of work, the Mechanical Maintenance foreman wishes to release his clearance and restore the system, but the instrument calibration is still taking place.
| |
| What actions (s), if any, must be taken to ensure the safety of the personnel performing the calibration?
| |
| t a.1hc Clearance form is transfened to the l&C foreman.
| |
| : b. The Clearance can be released with verbal permission from the I&C fortman,
| |
| : c. The l&C foreman must retum his Clearance Receipt to the tagging omelal prior to releasing the clearance.
| |
| : d. The Mechanical Maintenance foreman may clear all tags that penain to the I&C work ANSWER:
| |
| c.1he I&C foreman must retum his Clearance Receipt to the tagging omelat prior to releasing the clearance.
| |
| O ll>NO: 1.I' # Olti #
| |
| 432 ilLO 201 02 PROCEDURE NUMBER: OTHER:
| |
| ADM.00 gr7 LEVEL 3 TASK NUMilER:
| |
| 300101003001 NRC KAi l ROIL'SRO': }
| |
| G 2.2.13 3 34 j i COMMENTS: 7/97 new O 103 y _. . , ,. _
| |
| __,.__,__,_.y s_ ,__ ._ , g9
| |
| | |
| 7.10 Release of CLEARANCES 7.10.1. Release of Clearances for hiodification Requests;
| |
| : 1. Unless Authorized by the Operations Shift Superintendent, Clearances associated with hiodification implementation will not be removed and the equipment returned to sersice until one of the following criteria is met;
| |
| : a. The OSS/CRS has reviewed the installation documents (hiAl/Wl/lR) and has determined that there is no effect to plant equipment required for plant operations, and the modified system / component is not being returned to an operational status. This is usually dor'e to facilitate construction testing as described in PhiC 22 004.
| |
| OR
| |
| : b. A DCIP (Design Change Implementation Package) has been processed by the Responsible Field Engineer to release the affected component for testing in O accordance with PhtC 22 004 and the OSS/CRS has signed the DCIP in the " Release for Testing" block.
| |
| OR
| |
| : c. The hiodification has been completed for a Full, Divisional, Independent Operating Component (IOC) ~
| |
| or a Phase Installation and a DCIP has been prepared in accordance with PhiC-22 004 by the RFE and the DCIP has been signed by the OSS/CRS, authoriung operability of the affected system / component.
| |
| OR
| |
| : d. The OSS/CRS has reviewed the work documents and has determined that no change to permanent plant equipment has occurred. (e.g. A clearance was established for modification work, the modification work was postponed and a retum to service is required). This review will be annotated on any Clearances which may have been established for hiodification implementation.
| |
| ADM-0027 REY .16 PAGE 18 OF 31 1.
| |
| | |
| / 10.2. CLEARANCE Holder Requirements for Release;
| |
| (
| |
| : 1. Assure work is complete and perform a thorough investigation of the jobsite to ensure it is safe to remove DANGER Hold tags 2 Assure Work packages are complete to permit removal of CLEARANCE Receipt from the work package.
| |
| ED_TI Systems and'or components that are not operated by the Operations Department (eg. Sewage Treatment Plant) and do not have Systeru Operating Procedures (SOPS) will be returned to service per the direction of the responsible Department. 7 hey willprovide in uriting the required POSITIONING AC110N.
| |
| : 3. CLEARNCE Holder will then enter his CLEARANCE Holder " PIN" number on the Computerized Holder Form, if the Computer is unavailable, the CLEARANCE Holder will sign the " CLEARANCE Holder Release" block on the q
| |
| Q CLEARANCE Holder Sheet.
| |
| 39.lE 1he on-shift supervisor of the same department m charge of the u ork who has control of a CLEARANCE may request a CLL4RANCE Release held by a previous shift. In the absence of the CLEARANCE Holder, the Holder's s"pervisor or the duty maintenanceforeman may assume all the responsibilities of the CLE4RANCE Holder and writ comply with all CLEARANCE release requirements of this procedure. Telephone contact with the CLEARANCE Holder should be attempted to obtain his concurrence. 1he CLE4RANCE Receipt will be surrendered to the Tagging
| |
| '9cial.
| |
| 7.10.3. Tagging OfTicial Requirernents for Release:
| |
| : 1. Determine existence of additional CLEARANCE Holders.
| |
| CLEARANCE Receipts tumed in to the Tagging Official should be placed with the CLEARANCE Authorization to aid in accountability of all CLEARANCE Receipts. Once
| |
| (~~)N
| |
| ( accountability of CLEARANCE Receipts has been assured, the receipts may be discarded.
| |
| ADM-0027 REV - 16 PAGE 19 OF 31
| |
| | |
| ^
| |
| Prior to release of a Global Clearance, verify that all O 2.
| |
| Supplemental Clearances, within the Global, have been released.
| |
| SD.II The OSS CRS may authon:e the use of the SOP valve line-up or partial valve line-up to retunt the qstem to normal.
| |
| The partial line-up should be attached to the LLEAIMNCE. Ifan entire line-up is performed then it uill be placedin the syswin statusfile.
| |
| : 3. Consult the SOP to obtain proper component position.
| |
| Include all valves within the clearance boundary, in the tag i removal section, showing the restoration position.
| |
| (Commitment # 1484,3115,7118,7401,8840,12398 NRC IN 91-42, CR 92-0089A and INPO SOER 85 02).
| |
| : 4. Consult P& ids and enter components, and sequence to the Tag Removal Sheet Consider the effect on instrument reference legs of drained system. I&C will perform needed instrument line-up/ fill and vent, (Commitment #
| |
| 3115,7401 and INPO Finding (OP.3 2) 1992, Item b.).
| |
| 5 Determine verification requirements and check the appropriate block on the CLEARANCE Authorization Sheet. Verification requirements will be determined by consulting the SOP lineups. ( Commitment No.1484, 5727,3092,7118 and INPO Finding (OP. 3 2) 1992, item d.).
| |
| ED.fL 1he OSS'CRS may unive the Independent venfication and may consider an alterna rs means of 230 mrem exposure uouldoccur. IfIhe OSSCRSuairesIndependent l'enfication, : hen he uilt initialin the place of the venfier.
| |
| : 6. Sign removal approval on the CLEARANCE Authorization / Installation / Removal Sheet, obtain OSS/CRS/WCS signature for Authorization.
| |
| (Commitment No, 7405, NRC IN 91-42 and INPO SOER 85-02)
| |
| : 7. Direct the Designated Operator to restore the system in accordance whh the Tag Removal Section.
| |
| ADM-0027 REV - 16 PAGE 20 OF 31
| |
| | |
| i 7.10.4. The Designated Operator Responsibilities;
| |
| : 1. Have a full understanding of the Tag Removal Section, (component positioning, sequence and Special Clearance .
| |
| Instructions) before beginning the tag removal process.
| |
| : 2. Initial and date each Special Operating Instruction as it is !'
| |
| completed,if applicable,(CR 94 045l).
| |
| : 3. Maintain possession of a copy of the Clearance -
| |
| Authorization / Installation / Removal Sheet and initial ;
| |
| each step while performing DANGER Hold tag removal and component positioning in the sequence shown. When copy is used , transfer data to the original as soon as possible. (Commitment # 1484,7403 and SOER 85-02)
| |
| : 4. Sign in the "Fina! Release" section of the CLEARANCE Authorization / Installation / Pemoval Sheet when tag removalis complete ;
| |
| 7.10.5. When appropriate, the Independent Verifier will verify restoration of the CLEARANCE and initial each step.
| |
| ~'N (Commitment # 1484, 5727, 8840)
| |
| (G
| |
| : l. Sign in the " Final Release" section of the Clearance Authorization / Installation / Removal sheet when Indeper dent Verification is complete.
| |
| t 7.10.6. Tagging OfE :ial requirements following restoration :
| |
| : 1. Verify the following sections of the Clearance Authorization / Installation / Removal Sheet are complete; (CR 92-0089A).
| |
| : a. Final release section
| |
| : b. Irdependent Verification (signed off or N/A). ,
| |
| : c. Completion of Special Clearance Instruction (CR 0451)
| |
| : 2. Notify the OSS/CRS for removal of the CLEARANCE from LCOs as necessary.
| |
| : 3. Document Tag Removal on the Index Sheet, place the CLEARANCE in the inactive section of the CLEARANCE File.
| |
| ADM-0027 REV - 16 PAGE 21 OF 31
| |
| _ _ _ _ _ ~ _ _ _ _ . . . _ .
| |
| | |
| SRO EXAM KEY Esam Number NRC SRO Hev. 0 Esam Title NRC SRO EXAM lo0 While assigned in the Work Management Center as the WMC Supervisor, the following procedure CNs are submitted for your approval. Which one of the following procedure changes requires a formal 10CFR50.59 screening be completed?
| |
| : a. A step in an SOP was inadvertantly omitted between procedure revisions and needs to be reincorporated in order to complete an evolution whlch is in progress. ,
| |
| : b. An in progress IST STP requires use of service air. The valve specified has a broken bradwheel ,
| |
| and cannot &c used '!he CN to the procedure requests use of a different functional service air valve,
| |
| : c. While restoring from a bus outage IAW with OSP 0019. the procedure was found to contain an incorrect power supply. The CN correcting the power supply informadon must be added to the procedure to complete the bus restoration.
| |
| : d. A plant start up is in progress per GOP-0001. The annunciator " Turbine Hypass Valve Open" will not reset. A CN requests deletion of the step in GOp 0001 that verifies the annunciator is reset in order to continue the start up. The annunciator will be repaired when paits are available.
| |
| ANSWERt
| |
| : d. A plant start-up is in progress per GOP-000). The annunciator " Turbine Dypass Valve Open" will not re,et. A CN request deletion of the step in GOp-0001 that verifies the annuaciator is reset in order to continue the start-up. The annunciator will be repaired when parts are r v allable 4
| |
| IDNO: LP# OHJ #
| |
| 439 IILO-202 04 PROCEDURE NUMBER: OTHER:
| |
| RBNP-0001, page 5 LEVE8
| |
| * GOP-0001, Step 12 I'''NRC A$ l'ROi l~SROI ~l O 2 2.8 i 1.8= 3.3 _ j ;-
| |
| COMMENTS: 7/97 new ,
| |
| i O 104 yrg -yep--- y --w-- vs me .a 44 . ,9-., ,..e-,., w., , ._4w,e_. m., -. -&.-,3- ,., , ,, -- ,-,,--. qp. -9_y-r.-3 .g--.wy-y
| |
| | |
| NOTE The supervisor and duty OSSCRS st'ork Afanagement Supervisor iIt'AIS) mm. approve CNs nahout aformal30 39 screentngprovided that. In Ihearjudgment. Ihe CN Joes not involve a:
| |
| * change to a Tech SpectTRAf
| |
| * change to the plant describedin the S4R 1
| |
| * change to a procedure or test descrsbed on the SAR
| |
| * test or experament not described in the 54R j ;
| |
| 5.1.3. Preparer obtain supervisor approval. If unavailable another member of plant management staff kne.*ledgeable in the area of the affected procedure may approve changt. Telecon i approvalis acceptable. If not approved evaluate other change processes. ,
| |
| e if telecon approval, then indicate who approved change on each affected page.
| |
| e If not telecon appioval, then approvers place initials /KCN/date on each ufected page.
| |
| t 5.1.4. Preparer obtain duty OSS'CRS/WMS approval. Telecon approvalis acceptable. If not approved, evaluate other change processes.
| |
| * If telecon approval, then indicate who approved change on each alTected page.
| |
| e if not telecon approval, then approvers place initials /KCN/date on each affected page.
| |
| 5.1.5. Preparer may continue work. ,
| |
| i
| |
| . 5.1.6. Preparer should initiate PAR and attach copy of change (s). ,
| |
| : l. "Other" change notice should be designated on the PAR if the CN is to be used only one j
| |
| time or is temporary and has a specified expiration date.
| |
| : 2. "Other" change notices are nj incorporated into the procedure .
| |
| 5.1,7. Preparer obtain signature of:
| |
| : 1. Supervisor / Member of management providing temporary approval.
| |
| : 2. OSS/CRS/WMS providing temporary approval.
| |
| 5.1,8. Preparer provide OSS/CRS/WMS with a copy of the temporary approved change.
| |
| Y
| |
| - RBNP-001 REV - 15 = PAGF 5 OF 24 m
| |
| v.94 q- y+w-- s-+- # -y.,-#n.,,,,,yy.---W .s.e w e-ww w w. y.,ywra, w,--rwmm _,y-,g -t viert er w.:,-y g.-me=''5----* m- w' . + " ' - **t *""---f"W*-N---t P47rW'FT2'M T f WIh*W -"F'$W''"'"""3-'8FIE"""7'"*-'.
| |
| Y ~
| |
| | |
| I RIVER BEND STATION ;
| |
| SENIOR REACTOR OPERATOR .
| |
| NRC WRnTEN EXAMINATION ''
| |
| .I RESTRICTED INFORMATION OFFICIAL USE ONLY '
| |
| f 4
| |
| 0
| |
| ,,r-. , - , *, --v",--
| |
| | |
| RIVER BEND SENIOR REACTOR OPERATOR NRC EXAMINATION FACILITY: River Bend Station REACTOR TYPE: BWR DATE ADMINISTERED: 07/25/97 CANDIDATE:
| |
| IN, STRUCTIONS TO CANDIDATE:
| |
| Use the supplied answer sheet for documentation of your answers..there are 100 multiple choice questions on this examination, each worth 1.00 points. Passing grade for this examination consists of an overall score of 80%. Examination papers will be picked up four (4) hours after the examination begins.
| |
| 100 TOTAL POINTS CANDIDATES SCORE PERCENT All work on this examination is my own. I have neither given nor received aid.
| |
| Candidate's Signature
| |
| | |
| i i ne following conditions exist:
| |
| ne plant has experienced a station blackout.
| |
| ne Div 3 Diesel Gencrator was started and is running normally.
| |
| Emergency use of Div 3 for decay heat removal and RPV level control is being implemented.
| |
| Which of the following describes the general flowpath for this cooling mechanism?
| |
| : a. CST . IIPCS pump . RilR "A" heat exchangers . RPV . Shutdown cooling drains to Suppression pool.
| |
| : b. Suppression pool . IIPCS pump . RPV shutdown cooling to loop "A" RilR heat exchangers then test return to suppression pool.
| |
| : c. CST . IIPCS pump . RPV . Shutdown cooling to loop *B" RilR heat exchanger then test return to Suppression pool,
| |
| : d. Suppression pool . IIPCS pump . RilR "A" heat exchanger . RPV . ahutdown cooling drains to suppression pool.
| |
| 2 A Loss of Offsite Power has occurred. De Division i Diesel generator is currently loaded to 2500 KW.
| |
| Which one of the following is the MAXIMUM allowed additionalload that can be imposed on the generator 7
| |
| : a. 360 KW
| |
| , )
| |
| : b. 580 KW
| |
| : c. 630 KW
| |
| : d. 730 KW 3 Which of the following states the overall low low set system response of the SRVs for the RPV pressures given?
| |
| : a. At il13 psig, only one valve will be open. It recloses at 926 p j.
| |
| : b. At 1113 psig, only one valve will be open, it recloses at 936 psig.
| |
| : c. At 1103 psig, only two valves will be open. One recloses at 936 psig, the other at 926 psig.
| |
| : d. At i103 psig, only eight valves will be open. nree reclose at 946 psig, three at 936 psig, and the last two at 926 psig.
| |
| 2
| |
| , - , - - , - - _ y
| |
| | |
| _ _ ~ . _ _ _ _ _ . . _ _ _ ___ _ ____ _ _ _ _._-. . _ _ . _ _ , _ . . . . _ _ _ _ . _
| |
| s 4 You have been instructed to control drywell temperature and pressure by operating all avaliable drywell cooling. While doing this, service water to the drywell unit coolers automatically isolates.
| |
| Which of the following caused the isolation? l
| |
| : a. liigh drywc!! temperature (max. recorded 265 deg F)
| |
| : b. Low RPV water level (min. recorded 28")
| |
| : c. liigh drywell pressure (max. recorded 1.82 psid)
| |
| : d. Loss of 120 VAC power 5 EOP 2 " Primary Containment Control", requires the reactor to be scrammed before suppression pool temperature reaches 110 Degrees F. Which one of the following states the reason for this requirement? ,
| |
| : s. Assures that the containment design pressure wili not be exceeded due to I compression of the non condensable gasses due to the higher water temperature.
| |
| : b. Assures that with the expected temperature rise of 70 Degrees F during the blowdown phase of an accident, that complete condensation of reactor coolant will oC4-
| |
| : c. Assures the post LOCA suppression pool hydrodynamic forces are within the design limitation of. containment. ,
| |
| : d. Assures a reactor shutdown occurs, to minimize heat .)
| |
| rejected to the primary containment, if Emergency Depressurization is required.
| |
| 6 Which of the following will result in the addition of positive reactivity to the reactor? (Consider each case separately.)
| |
| : a. LPCS initiaiton during reactor STARTUP with reactor pressure at 300 psig.
| |
| : b. Sudden jet pump differential pressure reduction in one loop with the reactor in the RUN mode.
| |
| c, - Reduction in EllC pressure setpoint by 2 psig with reactor in the RUN mode,
| |
| : d. Initiation of me RCIC turbine during reactor STARTUP with reactor pressure at I50 psig.
| |
| | |
| 7 Which of the following methods for attemate control rod insertion during an ATWS REQUIRES the scram " signal" to be reset?
| |
| Control rod insertion by:
| |
| : a. using the individual control rod scram test switches.
| |
| : b. venting the control rod mechanism over piston volume.
| |
| c, maximizing CRD cooling water differential pressure.
| |
| : d. venting the scram air header.
| |
| 8 The Remote Shutdown Panel emergency transfer switches (division I rwitch on C61*P001 and divsion 11 switch on RSS'PNL102) for ADS /SRV B21.F0510 are in the EMERGENCY position.
| |
| Which of the tallowing Control Room handswitches can be used to manutlly open ADS /SRV D21 0051G7 .
| |
| : a. BOTil the Div i"A" and Div II *B" solenoid control switches.
| |
| : b. Div i "A" solenoid control switch only.
| |
| : c. Div !! "B" solenoid control switch only.
| |
| : d. Control Room control switches are inoperable.
| |
| J 9 The Control Room is uninhabitable and the Remote Shutdown Panels are being utilized to control the plant. Reactor levelis 20 inches and lowering, and Reactor Pressure is 500 psig.
| |
| With present plant conditions, whlch of the following systems can be utilized to raise reactor level from the Remote Shutdown Panels?
| |
| : a. LICS
| |
| : a. RitR A
| |
| : c. RCIC
| |
| : d. l{PCS
| |
| | |
| l 10 Given the following conditions:
| |
| The plant is operating at 100% power Offgas Building area radiation levels are rising j OfTgas Building and Main Plant Vent Stack effluent radiation levels are rising '
| |
| Ofigas hydrogen concentration is 5%
| |
| . Actions are being directed to reduce hydrogen concentration in Offgas to less than 4%
| |
| Which of the following is the reason why the operator is NOT allowed w place the standby recombiner in service to reduce hydrogen concentrations to less than 4%7 Placing the standby Offgas recombiner in service:
| |
| : a. may exceed the capacity of the preheater reducing temperatures to the level where recombination will stop.
| |
| : b. could reduce the flow rate when the service air purge is initiated.
| |
| : c. may cause a loss of main condenser vacuum resulting in an MSIV closure if the plant is still at power.
| |
| : d. could provide the ignition source for the hydrogen already present.
| |
| 1I Juing refueling, the leakage rate of the Refueling Cavity has exceeded the capacity of the Drywell and Containment Equipment and Floor Drain sumps. A fuel bundle is NOT in a safe storage ,
| |
| location.
| |
| J Which one of the following systems should be used for emergency makeup to the Refueling Cavity?
| |
| : a. Control Rod Drive liydraulics
| |
| : b. Condensate
| |
| : c. Reactor Water Cleanup
| |
| : d. CNS service connection
| |
| | |
| i 12 Which of the following is the basis for venting primary containment irrespective of offsite release rates at 20 psig per EOP 2 (Primary Containment Control)7
| |
| : a. This is done to vent the containment before the maximum containment pressure at which the primary containment vent 5 alves can be opened and closed is reached. l
| |
| : b. His is the maximum containment pressure at which the SRVs can be opened and i closed.
| |
| : i. This is the maximum pressure capability of the primary containment, where the most limiting component is the containment equipment batch.
| |
| : d. This is the maximum containment pressure at which the Standby Ges Treatment system can be used to vent the containment.
| |
| 13 Which of the following conditions constitutes an unsafe condition for the containment?
| |
| RPV Suppression Pool Suppression Pool Pressure Level Temperature
| |
| : a. 1025 psig 17 ft 3 in. 120 degrees F
| |
| : b. 700 psig 16 ft. 6 in. 131 degrees F
| |
| : c. 500 psig 18 ft 145 degrees F
| |
| : d. '400 psig 17 ft 9 in. 154 degrees F J
| |
| 14 During a major transient requiring use of the EOPs the following plant conditions exist:
| |
| Containment temp i19' 205 degy Drywelltemp 145' 285 deg.F Reactor pressure 200 psig Which one of the following indications is reliable RPV levelinformation?
| |
| : a. Narrow range when actual RPV level is +12"
| |
| : b. Upset range when actual RPV levelis +21"
| |
| : c. Wide range when actual RPV levelis 120"
| |
| : d. Shutdown range when actual RPV level is + 46" 6
| |
| - - , , - - - - . , , ,,- -y, ,,- . ~ , ,- -,r, w v.-
| |
| | |
| 15 in Emergency Depressurization, Step ED-3 asks,"Is Suppression Pool Level Above 13 ft? What is the significance of this level?
| |
| : a. It ensures a vortex will not 'ae created when SRVs are opened.
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containmem does not occur.
| |
| : c. It is required to prevent loss of NPSil to the RilR Pumps.
| |
| ^
| |
| : d. It ensures there is enough water to cover the horizontal vents.
| |
| 16 EOP 4 (Primary Containment Flooding)is executed to flood containment.
| |
| The containment level band specified by EOP-4 is between 62 ft and the Maximum Containment Water Level Limit (MCWLL)of 85 ft.
| |
| Which of the following ranges corresponds to the core level band specified by EOP 47
| |
| : a. 143 to +133 inches
| |
| : b. 162 to +114 inches
| |
| : c. 193 to +$1 inches
| |
| : d. 205 to +71 inches 17 A scran from full power has occurred as a result of high drywell pressur$ due to a leak. All automatic functions and isolations occurred per design, except that not all the control rods inserted fully. 'the Standby Liquid Control (SLC) pump "A" was initiated per applicable emergency prcedures.
| |
| As the operator monitors SLC parameters, SLC tank levelindicates zero (0) inches.
| |
| Which one (1) of the following actions should be taken next?
| |
| : a. Start SLC Pump "B".
| |
| : b. Restore power to the SLC tank level indication by restoring pow er to Nils.
| |
| MCC102A.
| |
| : c. Initiate actions to inject alternate SLC per Encl.15.
| |
| : d. Install Encl.16 Bypassing CNTMT Instrument Air Isolation Interlocks 7
| |
| | |
| 18 EOP 3, Radioactive Release Control, has been entered.
| |
| Which of the following is the reason that the operator is directed to ensure that the Turbine Building Ventilation fans are running?
| |
| : a. Reduce radioactive releases below General Emergency levels
| |
| : b. Prevent radioactive releases from the Turbine Buildhg.
| |
| f the Turbine Building atmosphere.
| |
| : c. Filter radioactivit/ rom
| |
| : d. Provide a monitored release point.
| |
| 19 Which one of the following is the BASIS for maintaining the refueling cavity pool 23 feet above the top of the reactor pressure vessel Bange during tefueling?
| |
| : a. To provide adequate net positive suction head to the fuel Pool Cooling Cleanup Pumps. ,
| |
| : b. To maintain a reservoir of water for suppression pool makeup.
| |
| : c. To provide spent fuel decay heat removal for 7 days without makeup,
| |
| : d. To remove the iodine gap activity released from a fuel rupture.
| |
| 20 Which of the following conditions still constitutes " Adequate Core Cooljng"?
| |
| NOTE: Only the injection sources stated are injecting. Regard each situation separately.
| |
| : a. ATWS in progress, the feed system is maintaining level between 465 inches and 195 inches, MSIVs are open.
| |
| : b. All rods in, MSIV/ ADS valves are closed, RPV level is 200 inches and RPV pressure is 200 psig.
| |
| : c. All rods in, RCIC is injecting,1 ADS valve is open, RPV level at 210 inches and MSIVs are closed.
| |
| : d. ATWS in progress, CRD, RCIC and SLC (with Boron) are injecting, RPV level is 200 inches and MSIVs are open.
| |
| 8
| |
| | |
| 21 ne following conditions exist:
| |
| . nc reactor has been scrammed.
| |
| 4 control rods are not fully inserted.
| |
| . Here is a coolant leak into containment.
| |
| Conditions degrade and require implementation of Emergency RPV Depressuriation per EOP-4A.
| |
| Which of the following actions are required?
| |
| : a. Immediately open 7 ADS /SRVs.
| |
| b Rapidity depressurize the RPV using bypass valves and MSL drains
| |
| : c. Terminate and prevent injection from all sources except SLC, CRD, & RCIC then open 7 ADS /SRVs valves.
| |
| : d. Close the MSivs, MSL Drains, and RCIC Steam Isolation Valves then open /
| |
| ADS /SRVs valves.
| |
| 22 All high pressure injection has been lost following a Reactor scram and RPV water level transient.
| |
| RPV water level is .162 and lowering slowly. RilR and LPCS are running on minimum flow. De CRS directs emergency depressurization. Why must at least 4 SRVs be opened to accomplish emergency depressurization under these conditions?
| |
| : a. LPCS and RilR will be injecting prior to RPV level reaching the minimum steam cooling level. -
| |
| : b. De level swell from four open SRVs will keep the core submerg#d until RilR and LPCS are injecting at rated flow.
| |
| : c. LPCS alone can reflood the core prior to the core uncovery time cxceeding the maximum core uncovery time limit.
| |
| : d. Enough SRV steam flow to cool the core will exist at a pressure that RilR can make up for the steam flow.
| |
| 9
| |
| | |
| 23 tegarding the Hydrogen Deflagration Overpressure Limit (ilDOL) curs e, as containment pressure
| |
| : i. *.reases, the maximum allow ed hydrogen concentration in percent (%) decreases.
| |
| Which of the following is the reason for this relationship?
| |
| : a. As containment pressure increases, the capabilities of the flydrogen Recombiners to remove hydrogen is decreased,
| |
| : b. His ensures a hydrogen deflagration at the limit combined with current pressure will not exceed containment overpressure failure limits,
| |
| : c. De containment hydrogen analyzer system response time is adversely affected es pressure inetcases.
| |
| : d. As containment pressure increases, the deflagration pressure of hydrogen dectcases requiring a lower concentration of hydrogen.
| |
| 24 The plant was operating at 100% power when a scram signal was generated and the reactor failed to scram. EOP.I A directs downshifting Recire Pumps.
| |
| Which of the following describes the reason Recirc Pumps are down shifted prior to tripping?
| |
| Tripping the Recirc Pumps could result in:
| |
| : a. entering the region of thermal /hydraulle instability,
| |
| : b. an excessive feedwater temperature reduction rate that will cause power to increase rapidly.
| |
| ,j
| |
| : c. a large level shrink which could cause isolation signals complicating the event.
| |
| : d. a reactor level swell which could result in a main turbine trip.
| |
| 25 Which of the following is a sample point for the Containment Atmosphere Monitoring System?
| |
| : a. containment dome,
| |
| : b. RWCU pump room.
| |
| : c. reactor sampling sink area.
| |
| : d. CRD flow control station.
| |
| 10 l
| |
| l l
| |
| | |
| 26 During operation at 100% power with a rod line of 100%, the "A" Recirc Pump inadvertently trips to off. About 20 seconds later, the *B" Recirc Pump trips to slow speed, resulting in the following steady state plant conditions:
| |
| Thermal power 50 %
| |
| Calculated core Dow 34 %
| |
| What is the required operator action?
| |
| : a. Immediately SCRAM the reactor,
| |
| : b. Reduce thermal power to less than 40% by inserting control rods,
| |
| : c. Raise core flow by upshifting the D recire pump to FAST.
| |
| : d. Raise core flow by starting the A recirc pump in SLOW.
| |
| 27 'ihe plant was initially operating at 100% power. A transient occurred resulting in the following conditions:
| |
| - RpV levelis 35 inches and stable Reactor power is 73% and stable
| |
| . Total core now is 51.5 E6 lbm/hr. and stable The cause of this plant c figuration was the receipt of a signal from the:
| |
| : a. EOC RPT logic.
| |
| A
| |
| : b. ATWS/ARIlogic,
| |
| : c. recirculation pump cavitation interlock circuitry.
| |
| : d. recirculation flow control valve runback logic.
| |
| 28 A loss of condenser vacuum has occurred, vacuum is currently I8.5" Hg. Which of the following automatic actions should have occurred?
| |
| : a. Turbine trip only.
| |
| : b. Turbine trip and bypass valve closure.
| |
| : c. Turbine trip and MSIV isolation.
| |
| : d. Turbine trip, bypass valve closure and M51V isolation.
| |
| II
| |
| | |
| _ _ . . . . _ . ~ . _ _ . _ . _ . _ _ _ _ . . _ _ . - _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| l l
| |
| 1 I
| |
| 29 De Plant was operating at 100% of rated power when a reactor scram occurred. Plant conditions l j
| |
| are as follows:
| |
| \
| |
| ~ Reactor power is on range 2 of the IRMs and decreasing
| |
| - De Main Turbine is tripped
| |
| - De Main Generator Ouptut breakers are closed i
| |
| - De Main Generator Exciter Field breaker is open
| |
| - Megawatt load on the Main Generator indicates .10 MWe (steady for 2 minutes)
| |
| - MVARs on the Main Generator indicate 50 MVARs leading (VARs in)
| |
| - Station loads are being supplied through the Prefened Station Transformer:
| |
| - Main Condenser Vacuum is 28"lig and steady ,
| |
| Which one (1) of the follr v eg describes the action (s) required by AOP-0002, Main Turbine and Generator Trips?
| |
| : a. Immediately trip the Main Generator Output breakers.
| |
| : b. Immediately initiate a reverse power trir of the Main Generator Output breakers by decreasing generator VARs to zero.
| |
| : c. No action is required as long as Main Condenser Vacuum remains above 26" IIg.
| |
| : d. Within 5 minute:,, initiate a revert.e power trip of the Main Generator Output breakers by decreasing generator VARs to zero.
| |
| 30 A small break LOCA has occurred. Reactor levelinitially fell to -47 inches, then HPCS initiated ,
| |
| and filled the reactor to a maximum of +55 inches, level is now steady at +40 inches. Which of the following describes the current status of E22 MOV F004, the HPCS injedion isolation valve?
| |
| : a. F004 will open on a liigh Drywell Pressure initiation signal even if the HPCS 111G11 WATER LEVEL 8 RESET pushhutton has not been depressed and the HPCS 111G11 WATER LEVEL 8 RESET pushbutton must be depressed before the valve ,
| |
| can be opened manually,
| |
| : b. F004 can be opened manually even if the HPCS lilGH WATER LEVEL 8 RESET pushhutton has not been depressed and the HPCS 111011 WATER LEVEL 8 RESET pushhutton must be depressed before the valve will open on a liigh Drywell initiation signal .
| |
| : c. F004 can NOT be opened manually and it will NOT open on a High Drywell Pressure initiation signal until the itPCS 111G11 WATER LEVEL 8 RESET pushhutton is depressed.
| |
| : d. F004 can be opened manually after the IIPCS INITIATION RESET pushbutton is depressed.
| |
| 12 P
| |
| ---..-...---,,,,,~...-,.----w-.._~E-, - - - -e --- , - - ,, .% -, --. -et-o - 4 , .- - -.m ---
| |
| w-- w -- - -- --
| |
| ,,wmr m .e . -->.e-. .--w-- - my----m St~~ r*
| |
| | |
| i l
| |
| 1 1
| |
| 31 With the reactor at 100% power, a loss of all Reactor Plant Component Cooling Water occurs. I What are the required operator actions? l
| |
| : a. Monitor and reduce system heat loads as necessary to continue plant operations.
| |
| : b. Insert a reactor scram and shift both recirculation pumps to slow speed,
| |
| : c. Commence a reactor shutdown per GOP-0002 Plant Shutdown,
| |
| : d. Insert a reactor scram and trip and isolate both recirculation pumps.
| |
| 32 With the Division i Diesel rear air start compressor motor seized. Which of the following actions are requhed?
| |
| : a. Declare Division i DieselINOPERABLE. Both starting air compressors are required for an OPERABLE air start system,
| |
| : b. Take no action since only one starting' air system is necessary to start the diesel and crosstying the unaffected air system with the affected air system would render the dieselINOPER/BLE.
| |
| : c. Maintain pressure in the normal band of the reciever associated with the seized compressor by intermittently using a high pressure hose connected between the operable forward and rear system air dr, " outlets. The Division 1 Diesel ('enerator willremain OPERABLE.
| |
| : d. Start and load the Division i Diesel Generator. With the diesel runabg. the starting air system is not required for OPERADILITY of the diesel.
| |
| 33 Following a complete loss of shutdown cooling, temperature readings indicate a i degree F increase in bulk water temperature every 10 minutes. Assume the reactor vessel head is on, no other parameters change, and current temperature is 124 deg. F.
| |
| Which of the following is the minimum amount of time before primary containment MUST be established?
| |
| : a. 160 minutes
| |
| : b. 560 minutes
| |
| : c. 580 minutes
| |
| : d. 760 minutes I3
| |
| | |
| 4' 34 A plant starup is in progress. Reactor power is on IRM Ra a;c '' and reactor pressure is 450 psig when the "A" CRD pump trips. The "B" CRD pump will not start.
| |
| A reactor scram is required if:
| |
| : a. A control rod recGves an llCU accumulator fault and cannot be inserted.
| |
| : b. More than one CRD high temperature alarm is received 4, No ClO pumps can be restarted within 5 minutes,
| |
| : d. Two or more accumulator faults exist.
| |
| 35 Which one of the following describes a cause and the expected inaccurate response of reactor level
| |
| :nstrumentation indications when in the UNSAFE region. of the RPV Saturation Curve?
| |
| : a. liigh containment temperatures will result in boiling of the reference legs causbg an errnneously high level indication.
| |
| : b. liigh reactor pressure will result in boiling of the reference legs causing an erroneously low level indication, c Low reactor pressure will result in boiling of the reference legs causing an erroneously low levelindication. <
| |
| : d. liigh drywell temperatures will result in boiling of the variable legs causing an erroneously low level indication.
| |
| J 36 Suppression pool level is offscale high.
| |
| Which one of the following describes the effect on indicated containment or drywell pressure?
| |
| a
| |
| : a. Indicated containment pressure is less than actual.
| |
| : b. Indicated containment pressure is greater than actual.
| |
| : c. Indicated drywell pressure is less than actual
| |
| : d. Indicated drywell pressure is greater than actual.
| |
| 14
| |
| | |
| m
| |
| +..
| |
| 37 Given the following conditions:
| |
| A failure to scram has occurred.
| |
| . Reactor power is 20% with control rods being inserted manually.
| |
| . EOP 3, " Secondary Containment Control" has been entered due to HVAC cooler high differential temperatures caused by a fire in the Auxiliary Building.
| |
| MSIVs have closed.
| |
| ~ . Condensate /feedwater is maintaining water level.
| |
| Which of the following systems should be isolated? _
| |
| a.' Feedwater
| |
| : b. Reactor Water Cleanup
| |
| : c. Control Rod Drive di Fire Suppression systems 38 Isolation of a primary system leak is required by EOP 3, Secondary Contatnment and Radioactive Release Control, in order to limit radioactive discharge.
| |
| By definition, the term '" Primary System" refers to any system:
| |
| : a. for which the ASME "N" stamp is issued. .
| |
| ^ ~
| |
| : b. required to shutdown and cooldown the reactor.
| |
| J
| |
| : c. required to maintain Primary Containment integrity.
| |
| : d. connected directly to the RPV that has a reduced leak rate if RPV pressure is lowered.
| |
| 39 Which of the following unisolable sys:em failures (ruptures outside of primary conta'mment) would constitute a primary system for purposes of EOP 3, Secondary Containment Control?
| |
| : a. . Main steam drain lines in the Main Steam Line Tunnel
| |
| : b. Containment vent line to drywell '
| |
| : c. Service Water supply to drywell coolers
| |
| : d. LPCS suction,line from the suppression pool-
| |
| < 13
| |
| | |
| 40 - A high radiation alarm exists on the Annulus ventilation system (RMS*REllB). You are monitoring the CRT bar chart display for RMS* REll A to validate the alarm condition on RMS'REllB. The 10 minute trend data for RMS'REll A is colored " light blue". Which of the following describes the status of RMS'REll A datfreadings?
| |
| : a. RMS'REll A is reading within 1% of RMS'RElIB.
| |
| : b. RMS'RElI A is in an Alert condition.
| |
| : c. RMS'REl1 A data is " questionable",
| |
| : d. communication has been lost between RM 80 and RM 23.
| |
| 41 EOP 3, Secondary Containment and Radioactivity Release Control, must be entered if the Secondary Containment differential pressure is above the maximum nortnal operating differential pressure.
| |
| Which on. of the followir.g is the reasor for this entry condition?
| |
| : a. A significant steac. leak into the seccadary containreent is indicated.
| |
| : b. A significant water leak from primary system may ee discharging radioactivity directly to the secondary conta'mment,
| |
| : c. A potential for the loss of secondary containment is indicated that could result in uncontrolled radioactive releases,
| |
| : d. An increase in the unmonitored ground level radioactive releases due) leakage '
| |
| through secondary containment is indicated.
| |
| 42 If one of the "B" Reactor Recirculation System Flow Converter fails (resulting in zero output) with the reactor operating at 100% power, which one of the following describes what will be generated in APRM Channel B?
| |
| : a. A Downscale Alarm and a Rod Block.
| |
| : b. A Rod Block only.
| |
| : c. A Half Scram signalonly.
| |
| : d. A Rod Block and a Half Scram signal.
| |
| 16
| |
| | |
| .- .-- - -... -. - .- , - .= - . . . - _ . - . . . - - . . - - - . _ . - . . - . .
| |
| - 43 During valve time testing on RHR System A,IE12*MOVF004A RHR pump A Suppression Pool Suction Valve, is closed with all other valves / switches in their normal standby position when a valid LOCA signal occurs. In this condition, RHR pump A breaker will:
| |
| : a. Close and immediately trip because of the IE12'MOVF004A contacts in the breaker trip circuit,
| |
| : b. Not close because of the IE12'MOVF004 A contacts in the breaker close permissive 5 circuit.
| |
| : c. Close after IE12'MOVF004A opens automatically,
| |
| : d. Close and remain closed, while IEl2'MOVF004 A remains closed.
| |
| 44 . Following a Losa of Coolant Accident the following plant parameters exist:
| |
| Reactor pressure is 460 psig Vessellevelis-80 inches -
| |
| Drywellpressure is 2.2 psid Containment pressure is normal and steady.
| |
| Which one of the following describes the Low Pressure Coolant injection mode of the Residual Heat Removalsystem?
| |
| : a. Pumps nave started, but are not injecting because the injection valves, F042A, B, and C have not opened. .
| |
| : b. Ihmps have started, irdection valves F042A, B, and C have opened.,lput reactor
| |
| * pressure is too high for injection.
| |
| i
| |
| : c. Pumps have not started, but injection valves F042A, B and C have opened,
| |
| : d. Pumps have started, injection valves F042A, B, and C have opened, and injection has started.
| |
| 4 1
| |
| 0 i7 i
| |
| | |
| 45 De RCIC system is in Standby Lincup, but the RCIC T1JRBINE EXilAUST SHUTOFF valve, E51 F068, is closed for a valve stroke test. A loss of feedwater causes a low reactor water level (level 2).
| |
| Select the statement which describes how the RCIC system will respond.
| |
| : a. RCIC TURBINE EXilAUST SilUTOFF valve, E51-F068, autc.matically opens; RCIC system initiates and injects water into the RPV.
| |
| : b. 1 he RCIC turbine will start and trip on high RCIC turbine exhaust pressure at 25 psig.
| |
| : c. RCIC starts and the RCIC system exhaust rupture diaphrams will rupture initiating a RCIC system isolation at 10 psig exhaust diaphram pressure.
| |
| : d. RCIC turbine does not stan. RCIC TURBINE EXilAUST SIUJTOFF valve, E51 F068, must be open for RC!C STEAM SHUTOFF valve, E51 F045, to open.
| |
| 46 De plant is operating at 100 % power, steady state. The Control Room Operator is performing LPCS Quarterly Pump Surveillance. The LPCS pump is running in the test return to the suppression pool mode. A steam leak in the Drywell caused Drywell pressure to increase to 1.72 psid. Reactor pressure is being maintained at 950 psig by the bypass valves.
| |
| Which of the following statements describes the response of the LPCS system 7
| |
| : a. The LPCS Pump willload shed then remair. in standby, The E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| De E21 F011 (LPCS MIN FLO TO SUPP POOL) opens. ,)
| |
| )
| |
| : b. The LPCS Pump will continue running.
| |
| The E21 F012 (LPCS TEST RTN TO SUPP POOL) remains open.
| |
| The E21 F011 (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| : c. The LPCS Pump will continue running.
| |
| ne E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21 F011 (LPCS MIN FLO TO SUP_PPOOL)' opens.
| |
| : d. The LPCS Pump will load shed then remain in standby, ne E21-F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| De E21 F011 (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| 18
| |
| | |
| 47 A LOCA has resulted in the automatic start of tr.c llPCS system and injection into the vessel.
| |
| During the transient, the operator closed ilPCS Injection Valve (E22'F004) using the control siritch when the following conditions existed:
| |
| Reactor water level = +50 inches Condensate Storage Tank Level = 100 feet (Mean Sea Level)
| |
| Suppression Pool Level = 19 feet 4 inches Which one of the following describes the automatic response of the HPCS Suppression Pool Suction isolation Valve (E22* F015) and the llPCS Injection Valve (E22'F004) when reactor water level decreases to -50 inches and Suppression Pool level increases to 20' 6"? (Assume NO operator action.)
| |
| SUPPRESSION POOL SUCTION INJECTION VALVE ISOLATION VALVE (F015) (F004)
| |
| : a. Opens Opens
| |
| : b. Remains closed Opens
| |
| : c. Remains clowd " Remains closed
| |
| : d. Opens Remains closed 45 SLC System A is in a normal STANDBY lineup but with the TEST TANK OUTLET VALVE ,
| |
| (C41* F031) OPEN. Which of the following most accurately describes the effects on the STORAGE TANK OUTLET VALVE (F001) and SLC PUMP A by placing the SLC Keylock ControlSwitch for Pump A to START 7 ,j
| |
| : a. Vahe FMI Opens, SLC Pump A Starts after the valve reaches its Full Open position.
| |
| : b. Valve F001 Opens, SLC Pamp A starts concurrently wi*.h the valve opening.
| |
| : c. Valve F001 does Not Open, SLC Pump A Starts
| |
| : d. Valve F001 does Not Open, SLC Pump A does Not Stut 19
| |
| | |
| ~
| |
| 49 An ATWS has occurred and the following conditions exist:
| |
| Reactor power 12%
| |
| Reactor water level 13.4 inches Drywell pressure 1.4 psig All scram valves are open Scram discharge volume is full.
| |
| All discharge volume vent and drain valves are shut Mode switch is in SHUTDOWN Which one of the following conditions will allow a resetting of the scram signal?
| |
| : a. The scram can be reset using the " SCRAM RESET" switches after placing the CRD Scram Discharge Volume keylock switches to bypass.
| |
| : b. De scram can be reset by placing the Mode switch in "RUN" and placing the
| |
| " SCRAM RESET" switches to reset after placing the CRD Scro Discharge Volume keylock switches to bypass,
| |
| : c. De scram can not be reset due to dryvfell pressure being elevated above th.: alarm <
| |
| setpoint.
| |
| : d. De scram can not be reset due to reactor power being above the alarm setpoint.
| |
| 50 During a reactor startup you have been withdrawing SRM detectors per GOP 0001. All SRMs
| |
| . except A indicate full out. SRM A has an upscale high and an upscale high-high trip indicated and is reading off scale high. De P680 indications show the detector is " driving cat".
| |
| J You should:
| |
| : a. Immediately insert control rods to return SRM A readings on-scale,
| |
| : b. Insert a Div i half scram and continue with the plant startup,
| |
| : c. Check the SRM A drive power fuses, if the problem is not corrected, obtain reactor engineering assistance,
| |
| : d. Since the other drive OUT lights are on, SRM A drive has power therefore contact IAC for assistance.
| |
| O J .
| |
| 20
| |
| +
| |
| u
| |
| | |
| . .~ . - - - . . - . . - __ .~. - - .-. . - _ . - - - . . __ _ .-- -.
| |
| 51 The reactor has been operating near rated power for 200 days. Which one of the following describes the change in the indicated LPRM output signal from day I to day 200 ar.d the method used to calibrate the LPRMs?
| |
| INDICATED LPRM POWER METiiOD OF LPRM CALIBRATION
| |
| : a. Decreases Core ficat Balance
| |
| : b. Decreases TIP System Trace
| |
| : c. Increases Core Heat Balance
| |
| : d. Increases TIP System Trace
| |
| $2 Given the following conditions:
| |
| RCIC is operating in the Test Return Mode for IST testing.
| |
| - The plant experiences a Station Blackout.
| |
| RCIC is then manually initiated. .
| |
| Which of the following valves will stroke CLOSE under these conditions?
| |
| : a. RCIC Steam Supply Valve, F045.
| |
| : b. RCIC Steam Supply Inboard Isolation Valve, F063.
| |
| : c. RCIC Test Return to CST Valve, F0$9.
| |
| : d. RCIC Suppression Pool Suction Valve, F031.
| |
| 53 Following a valid ADS initiation, the operator is directed to close the ADS valves with the initiating r8gnals still present. Which one of the following operator actions will cause the ADS valves to close?
| |
| : a. Place the control switches on H13 P601 and H13 P631 for the ADS valves to the "OFF" position.
| |
| : b. Place the ADS inhibit switches on Hl3P-601 to the " NORMAL" position.
| |
| : c. Stop all low pressure ECCS pumps in both Div. I and Div. 2.
| |
| : d. Depress both " ADS Timer / Level 3 Seal In Reset" pushbuttons, S13 A(B) 21
| |
| | |
| 54 A Safety Relief Valve (SRV) tailpipe vacuum breaker wes failed in the open position when the SRV opened. Which of the following is the result?
| |
| : a. Containment pressure increased.
| |
| : b. Steam bypassed the quenchers with a direct discharge path into the suppression pool.
| |
| : c. Drywell to containment differential pressure increaxd.
| |
| : d. Suppression pool water will be drawn up into the SRV discharge line after the SRV is closed.
| |
| 55 When EOP 4, Emergency RPV Depressurization, permits defeating isolation interlocks in order to rapidly depressurize without SRVs. Which of the following MSIV isolation signals may be bypassed?
| |
| : a. Only the RPV low level I signal.
| |
| : b. Only the RPV low level 1 and low main steam line pressure signal.
| |
| : c. All MSIV isolation signals except for Main condenser low vacuum.
| |
| : d. All MSIV automatic isolation signals.
| |
| 56 Select the statement that describes an operator action required fu the following plant conditions:
| |
| J Reactor power: 75 %
| |
| Suppression pool temperature: 105 degrees F and rising Suppression poollevel: 19 feet 8 inches SRV IB21 F047B: Failed open
| |
| : a. If the SRV cannot be closed within five minutes, place the reactor mode switch in SilUTDOWN.
| |
| : b. If suppression pool temperature exceeds 120 degrees F., arm and depress the manual scram pushbuttons.
| |
| : c. Place the reactor mode switch in SHUTDOWN.
| |
| : d. Reduce suppression pool temperature to less than 100 degrees F within I hour.
| |
| 22 y
| |
| | |
| ST Consider the following plant conditions:
| |
| Reactor power: -45%
| |
| Generator load: 410 MWe Recirculation flow control:- Loop Manual SELECT the plant response to a continuous runback of the load set demand signal to zero in the Electro-llydraulic Control (EHC) system.
| |
| : a. Turbine control valves (TCVs) throttle closed, bypass valves (BPVs) remain closed, reactor pressure mcreases, reactor scrams on high pressure or high neutron flux.
| |
| : b. - Bypass valves (BPVs) throttle open, reactor pressure decreases, MSIVs isolate on low steam line pressure, l
| |
| reactor scrams on the MSIV closure.
| |
| : c. Turbine control valves (TCVs) throttle closed, bypass yalves (BPVs) throttle open to compensate, once the BPVs are fully open reactor pressure insteases causing a scram on high pressure or neutron flux.
| |
| : d. Turbine control valves (TCVs) throttle closed,
| |
| ' bypass valves (BPVs) throttle open, reactor pressure remains fairly constant, reactor power increases slightly due to reduced feedwater heating.
| |
| ,)
| |
| 58 A plant startup is in progress. Reactor power is being held at 1% power for the 900 psig Drywell walkdown when the Startup Feedwater Reg. Valve drifts fully open.
| |
| Which of the following actions / signals will occur as a result of this failure?
| |
| (NOTE: Assume no operator action.)
| |
| : a. Reactor scrams on high reactor water level.
| |
| : b. Reactor feedwater pumps trip on high reactor level,
| |
| : c. Reactor water level remains unchanged due to compensation by the Long Cycle -
| |
| Cleanup Level Controller (CNM-104).
| |
| # de Reactor water level stabilizes at a new higher level.
| |
| 23
| |
| | |
| -59 Given the following plant conditions:
| |
| A LOCA has occurred. >
| |
| Reactor waterleset is 50 inches.
| |
| Drywell Pressure is 1,12 psid.
| |
| All radiation monitors are indicating normal for plant conditions.
| |
| "A" Standby Gas Treatment System (SGTS) is running and the "B" SBGT system has been shutdown and placed in STANDBY.
| |
| Which of the following will be the status of SBOT systems "A" and "B" five (5) minutes sfter ENS- -
| |
| SWGI A receives a degraded Bus Undervoltage Signal? (All associated systems respond normally)
| |
| : a. Both systems are shutdown.
| |
| : 14. Both systems are runnina,.
| |
| : c. "A" system is shutdown and "B" system is running,
| |
| : d. "A" system is running and "B" systen} is shutdown.
| |
| 60 An MSIV closure rr .alted in a reactor scram. The pressure transient caused a small steam leak in the drywell. The lollowing conditions exist:
| |
| Reactor pressure is at 900 psig.
| |
| Reactor Level is at 80 inches wide range -
| |
| Drywellpressure is 2.1 psid Containment pressure is 0.3 psig ,j Lowest recorded ENS *SWGl A Bus voltage was 3952 volts.
| |
| Which one of the following would be in service as indicated?
| |
| (NO OPERATOR ACTION TAKEN)
| |
| : a. DIV I D/G running unloaded.
| |
| : b. DIV 11 SSW with now through the ''B" Containment Unit Cooler,
| |
| : c. Drywell units coolers running with no cooling flow.
| |
| : d. LPCS injecting to the RPV, 24
| |
| | |
| 61 ne following conditions exist:
| |
| ne Div 2 standby diesel generator is loaded and in parallel with bus IENS'SWGIB throua.h the normal breaker. A LOCA signal occurs.
| |
| Which of the following describes the effect of the standby diesel generator and bus LENS'SWGlB?
| |
| : a. The normal bus supply breaker will open and the diesel generator will supply bus loads,
| |
| : b. De normal bus supply breaker and diesel generator output breaker will open, then after loads are stripped, the diesel generator output breaker will reclose .
| |
| : c. He diesel eneratur output breaker will open and cannot be closed as long as bus voltage is supplied by the normal or attemate feeders until the LOCA signal is reset.
| |
| : d. De diesel generator output breaker will remain closed in parallel operati a with the bus.
| |
| 67 Which one of the following maintains a n'egative pressure in the annulus following a LOCA?
| |
| : a. SBGT system starts and takes a section on the Annulus Pressure Control System.
| |
| : b. Annulus mixing fans start and discharge to the Annulus Pressure Control Systern.
| |
| : c. Annulus mixing fans start and discharge to the SBGT system. ,
| |
| : d. SBGT system starts and takes a suction directly from the Annulus. d 63 The plant is operating at 100% power. Both Recirc Flow Control Valves are in Flux Manual (Loop Auto) at 67% valve position. A leak in the Deywell has caused Drywell Pressure to increase to approximately 1.75 psid. Following the high dry vell pres.aa signal, the *B" Reactor Feed Pump Trips and level decreases to + 14.5 inches and stabilizes. Which of the following describes the response of the Recirc Flow Control Valve?
| |
| Flow Control Valves will:
| |
| : a. runback to 22 % valve position.
| |
| : b. go to " min" position.
| |
| : c. move to a position to provide 60 % core flow.
| |
| d remain at 67 % valve position.
| |
| 25 l
| |
| | |
| 64 During power ascension the following plant conditions are noted to occur over a 3 minute perioc -
| |
| Reactor pressure decreased to 800 psig, now stable.
| |
| Reactor Water Level +25" and rising.
| |
| Reactor power decreased 5%, now stable at 50%
| |
| Generator output decreased to 550 Mwe from 600 Mwe.
| |
| No SCRAM No PPS actuations have occurred.-
| |
| Which of the following is required?
| |
| : a. Increase power with recirculation flow,
| |
| : b. IMMEDIATELY shut the MSIVs only,
| |
| : c. Insent a manual scram only.
| |
| : d. Scram and shutthe MSIVs. '
| |
| ,. ?
| |
| 65 Following the receipt of an automatic reactor scram signal,10 control rods remained partially withdrawr.. ne plant conditions are as follows ne scram valves on the 10 control rods indicate open I Reactor pressure is 950 psig ARI has been manually initiated Which one of the following actions would be required to insert the 10 parlially withdrawn control rods?
| |
| : a. Pull the RPS scram fuses to de energize the IICU scram solenoids,
| |
| : b. Pull the control pcwcr fuses for the Backup Scram Valves.
| |
| : c. Reset ARI and vent the scram air header,
| |
| : d. Vent the CRD over piston volumes.
| |
| /
| |
| = .=
| |
| 0 26
| |
| | |
| 66 A reactor startup is in progress and reactor pressure is 800 psig. A loss of ooth CRD pumps has resulted in the receipt of the CRD ACCUMULATOR TROUBLE alarm. The nitrogen pressure on one of the CRD HCOs indicates 400 psig. Which one of the following describes the effect of this condition on the CRDM when a scram is initiated?
| |
| : a. Accumulator pressure alone will drive the rod in.
| |
| : b. Reactor pressure alone will drive the rod in.
| |
| : c. Both reactor pressure and accumulator pressure must be combined to drive the rod in.
| |
| : d. Both reactor pressure and accumulator pressure combined are inadequate to drive the rod in.
| |
| 67 Whict. one of the following conditions would result from the failure of Reactor Recirculation Pump
| |
| #1 (inner) seal assembly at rated conditions?
| |
| : a. A decrease in #1 seal cavity pressure from approximately 1000 psig to about 500 psig.
| |
| : b. An increase in #1 seal cavity pressure from approximately 500 psig to about 1000 psig.
| |
| : c. An increase in #2 seal cavity pressure from approximately 500 psig to about 1000 psig. .
| |
| : d. A decrease in #2 seat cavity pressure from approximately 1000 psig g about 500 psig.
| |
| 27
| |
| | |
| _ __ _ _ _ _. _ _.. _ _ . _ _.__.______......_m..-. _ _ _ _ _ . _ _ _ . _ _ . _ . _
| |
| 68' Given the following conditions:
| |
| - He Reactor Water Cleanup (RWCU) system is operating in the normal mode, ne RWCU holation Bypass Switches (E31 Si A,B) on P632 and P642 have been placed in
| |
| " Bypass".
| |
| -7 Select the expected c!Tect on the RWCU system.
| |
| : a. He RWCU system isolation on hign non-regenerative heat exchanger outlet temperature is defeated,
| |
| : b. He RWCU system isolation from high area temperature ONLY are defeated.
| |
| - c. De RWCU system isolation from high differential flow A ND high area temperature are d feated.
| |
| : d. All RWCU system isolation signals are defeated.
| |
| 69 The RHR S/D Cooling Isolstion Valve Enable / Disable switch on the local panel (P001) has two positions," Enable / Disable".
| |
| Which of the following describes when the switch is REQUIRED to be in " Disable" and the etTect on the operation of Shutdown Cooling when it is in this position?
| |
| He RHR SiD CoolingIsolation Valve Enable / Disable switch is placed in " Disable" when:
| |
| : a. reactor pressure is greater than 135 psig and prevents operation of thy RHR Shutdown Cooling inboard Isolation Valve (F009) from the Main Control Room.
| |
| : b. reacter pressure is greater than 135 psig and prevents operation of the RHR Shutdown Cooling Outboard Isolation Valve (F008) from the Main Control Room,
| |
| : c. evacuating the Main Contrel Room to allow local operation of the RHR Shutdown Cooling Inboard Isolation Valve (F009).
| |
| : d. evacuating the Main Control Room to allow local operation of the RHR Shutdown Cooling Outboard Isolation Valve (F008).
| |
| l 28
| |
| | |
| ~ 70 Given the following plant conditions:
| |
| Reactor shutdown in progress.
| |
| IRM "C" indicating 75/125 on Range 6.
| |
| SELECT the statement that best describes the response of the plant if IRM C is ranged down by the operator depressing the down range pushbetton,
| |
| : a. Control rod movements can continue as normal.
| |
| : b. Only a rod block will be initiated.
| |
| : c. Only a half-scram will be initiated.
| |
| : d. A rod block and half scram will be initiated.
| |
| 71 Given the following plant conditions: .
| |
| The Refuel Platform is over the cbre.
| |
| - The Mode Switc's is in REFUSL.
| |
| Which of the following will cause a control rod block?
| |
| : a. Withdraw a control rod.
| |
| : b. Loading the Auxiliary Platform hoist.
| |
| : c. Loading the Refuel Platform main hoist.
| |
| ,)
| |
| : d. Loadmg the Refuel Platform monorail hoist.
| |
| 72 A startup of the Main Turbine is being performed. The Main Turbine is at 60 percent of rated speed, when a loss of 125 VDC Trip Circuit Power is experienced. WHICH ONE (1) of the following describes the required operator action (s)?
| |
| : a. Enter AOP-0002, Main Turbine and Generator Trips, due to trip of the Turbine,
| |
| : b. Verify that 24 VDC ETS power is available ad continue the startue of the
| |
| . Main Turbine IAW SOP 0080, otherwise manually trip the Main Turbine.
| |
| : c. Allow the Main Turbine to accelerate to greater then 90 percent of rated speed, at which time the 125 VDC Trip Circuit is no longer required because the PMG is supplying the trip circuitry,
| |
| : d. The start up of the Main Turbine may continue, but at least one 125 VDC bus
| |
| * must be restored prior to synchomizing the generator to the grid.
| |
| T 29
| |
| _.- .- _ _ - - - - - .~ - - - . -
| |
| | |
| 73 De unit is operating at 100% rated power when r complete loss ofinstrument air occurs. Which one of the following indirectly causes the Reactor feedwater Pumps to TRIP?
| |
| a, ne condensate minimum now and heater drain pumps re . ire valves fail open.
| |
| : b. De feedwater Dow control valves lock up and then drift closed.
| |
| : c. De feedwater pump suction valve drifts closed.
| |
| : d. He feedwater pump motor cooler TPCCW supply fails closed.
| |
| 74 With the plant at 100% power, a loss of VBN PNL01B1 has resulted in a loss of power to the Feedwater Level Control System giving a Feed Reg Valve control signal failure.
| |
| He power loss also caused both Reactor Recirc pumps to shift from fast speed to slow speed and the B Recirc Flow control valve to lockup. Which plant response would result from these failures?
| |
| (Assume no operator actions.)
| |
| : a. He "B" Feed Reg Valve would fail closed and the "A" and "C" Feed Reg Valves would AUTO OPEN to compensate. Reactor power will stabilize at a lower power level with both Recirc pumps in s'ow speed.
| |
| : b. All 3 Feed Reg valves will fall open. RPV level will raise to 51" which willinitiate a reactor scram, Turbine trip, and Feedwater pump trip.
| |
| : c. All 3 Feed Reg valves will fail closed. Reactor power willlower when Recirc pumps
| |
| * down shift and RPV level will lower to 9.7" which will initiate a reactor scram. HPCS and RCIC will initiate at Level 2 and restore RPV level . ,)
| |
| : d. All 3 Feed Reg valves will fail"as is". Reactor power willlower when Recirc pumps down shift and RPV level will raise to $1" which will initiate reactor scram, Turbine trip, and Feedwater pump trip.
| |
| 30
| |
| | |
| 75 ' De following conditions exist:
| |
| De reactoris at 100% power.
| |
| De Off Gas Post Treatment HI Hi HI radiation alarm (P601/22A/A03) has occurred, ne Offgas System automatically isolated,(IN64 F060 Off Gas Discharge to Vent valve is closed).
| |
| Which of the following actions is PROHIBITED?
| |
| : a. Purge the Off Gas system with service air.
| |
| : b. Shift to the Standby Off Gas Component.
| |
| : c. Reduce power as necessary to maintain condenser vacuum,
| |
| : d. Reduce power to below 60%.
| |
| 76 Given the following conditions:
| |
| Reactor water level is -90 inches and lowering.
| |
| Drywell pressure is 2.2 psig and raising.
| |
| An outside fire has caused smoke in the Control Room.
| |
| De operator has attempted to manually place the Control Room ventilation in the smoke removal mode.
| |
| Under these conditions the Control Room Smoke Removal Damper (AOD 107/108) will:
| |
| : a. open and the Smoke Removal Fan will start.
| |
| J
| |
| : b. open but the Smoke Removal Fan will be interlocked off.
| |
| : c. remain closed and the Smoke Removal Fan will run on recire, ri. remain closed and the Smoke Removal Fan will be interlocked off, C
| |
| 31
| |
| | |
| 77- ne plant was operating at 100% power when a large steam leak on the MSR A reheat steam line
| |
| - required reheat steam to be isolated. Five minutes later the 'IUkBINE HIGH VBRATION annunciator on P870 alarms. You check the recorder and observe bearing #5 reading 9 mils, bearing # 6 reading 13 mils, and bearing # 4 reading 5 mils. These read' mss appear to be steady.
| |
| You should:
| |
| : a. Continue to monitor the vibrations and initiate a reactor scram and trip the turbine if 15 mils is reached.
| |
| : b. Commence a rapid load reduction then take the turbine offline (scram and turbine trip) if vibrations cannot be reduced below 10 mils within 14 minutes.
| |
| : c. Immediately scram the reactor and trip the t"rbine.
| |
| : d. Monitor bearing temperatures and scram the reactor and trip the turbine if bearing temperature exceeds 240 degrees F.
| |
| 78 ne plant is operating at 60% power for rod pattern adjustment. Rod 22-43 is to be inserted from notch 48 to notch 42. A few seconds after the rod is inserted to notch 42 a ROD DPJFT annunciator is recieved. While conducting your immediate actions, you observe rod 22-43 passing notch 46 and ol, serve it to stop at no:ch 48.
| |
| De Control Rod Movement Sequence withdraw limit is notch 48. Reactor power has retumed to 60%. He .:onect actions will be:
| |
| : a. Place the Mode Switch in SHUTDOWN. ,
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whdher the rod has a stuck collet.
| |
| : c. Since power is below HPSP the Rod Withdraw Error analysis is affected and the rod -
| |
| should be inserted to notch 42 without delay.
| |
| : d. Notify a Reactor Engineer, declare rod 22-43 inoperable, and adjust the pattern as .
| |
| needed for flux shaping with rod 22-43 full out since it will not remain inserted.
| |
| O 32
| |
| | |
| i 79 'R* control room operator is about to startup RilR ir, the fuel pool cooling assist mode. All prerequisites have been completed and the operator st sts RllR pump ' A'. He then starts to throttle open the HX OUTLET VLV (E12-F003 A). Watching the valve position indication the operator observes there is no change, also flow is 700 gpm. Fuel poollevel:(choose one)
| |
| : a. INCREASE because Suppression Pool water is being diverted to the Fuel Pool.
| |
| : b. REMAIN TIIE SAME because the RilR system is currently recircing 700 gpm from/to the Suppression Pool.
| |
| 3
| |
| : c. REMAIN Tile SAME because the RilR system is currently recircing 700 gpm from/to the fuel Pool.
| |
| : d. DECREASE because water from the Fuel Pool is being diverted to the Suppression Pool.
| |
| 80 he plant is operating at 75% power, ne Control Room Operator places the Outboard MSIV Positive Leakage Control System switch to OPERATE. Which of the following will prevent the Outboard MSIV Positive Leakage Control System from initiating?
| |
| : a. A LOCA signal on either high drywell pressure or low reactor water level is not j
| |
| j present.
| |
| : b. De required main steam line pressure and reactor pressure requirements have not been inct.
| |
| : c. He post LOCA 20 minute timer has not timed out. J
| |
| : d. All Main Steam isolation Valves have not been fully closed.
| |
| 33
| |
| | |
| 81 ~ ne following conditions exist:
| |
| We reactor is in cold shutdown per GOP-0002.
| |
| . Reactor recirew lon pump A and RilR A shutdown cooling loop are in operation.
| |
| RPV levelis 35 inches.
| |
| Subsequently, Reactor Recirculation pump A trips and RPV level is being raised to greater than 75 inches.
| |
| Select the LOWEST vessel level which will result in water flow into the steam lines?
| |
| : a. 75 inches
| |
| : b. I10 inches
| |
| : c. 140 inches
| |
| - d. 196 inches 82- ne plant is conduct'.ng a startup with reactor power at 22%, when an unisolable rupture in the Turbine Plant Component Cooling Water (TPCCW) suction header causes a complete loss of TPCCW. . What a e the required immediate operator actions?
| |
| : a. Conduct an orderly reactor shutdown per GOP-0002, Plant Shutdown.
| |
| : b. Initiate RCIC and shutdown the Feedwater pumps. .
| |
| : c. Reduce reactor power to within Bypass Valve capacity then trip the gain Turbine.
| |
| : d. Manually scram the reactor, b
| |
| /
| |
| | |
| 83 A plaat startup is in progress iri accordance with GOP-0001. Plant conditions are as follows:
| |
| - De reactor is suberitical with control rod withdrawal in progress,
| |
| - Reactor power is 1 x 10-4 cps on SRMs.
| |
| - Recirculation loop temperatures are 180 deg F.
| |
| Which one of the following statements is correct concerning the administrative Control Room 3
| |
| : Supersvisor?
| |
| : a. He Admin CRS is required to be stationed at this time and remain stationed until criticality is achieved.
| |
| : b. De Admin CPJS is required to be stationed from criticality until the last FWREG valve is in service.
| |
| : c. De Admin CRS is required to be stationed from criticality until the first Main Turbine Bypass valve is 50% open.
| |
| : d. The Admin CRS is required to be stationed at this time and remain stationed until the last FWREG valve is in service. ,
| |
| 84 While the plant is at pow .eak develops in an area that is acessible, but now radiologically contaminated. De OSS lu. directed that an investigation be performed immediately. What documentation must be generated before various personnel are allowed entry into the area for the investigation?
| |
| : a. A daughter RWP to the General RWP for that area me be generated.
| |
| : b. None, a General RWP already exists for this type of event.
| |
| : c. A Specific RWP must be generated,
| |
| : d. None, a RWP may be completed atler the entry provided it is donc under continous RP coverage.
| |
| 35
| |
| | |
| 85 SOP 0031," Residual lleat Removal" cautions the operator NOT to simultaneously close the Ri!R llX BYPASS valve (E12 MOVF048B) and the RilR A liX OUTLET valve (E12 MOVF003B) while the RilR pump is in service.
| |
| Which one of the following actions should be taken if both of these valves are midvertently closed in RilR loop B !.ne in suppression pool cooling?
| |
| : a. Open the F048B first to reduce the ditTerential pressure across the F003B then throttle open F003B.
| |
| : b. Open the F003B immediately to reestablish flow through the RilR IIX.
| |
| : c. Shutdown the RilR pump prior to opening either the F003B or the F048B.
| |
| : d. Verify that the minimum flow valve (F064B)is fully open and reopen either F003B or F0048B.
| |
| 86 A LOCA event inside primary containr. ent is in progress. Level has been restored per the EOPs.
| |
| The llPCS diesel is running unloaded and the local operator reports thick smoke coming from the diesel. The CRS directs the U.O. to immediately shut down the diesel. Which of the following actions will successfully shut down the diesel .
| |
| : a. Take the DIESEL ENGINE CONTROL switch to STOP.
| |
| : b. Depress the EMERGENCY STOP pushbutton.
| |
| : c. Reset the IIPCS initiation signal then take the DIESEL ENGINE CONTROL switch to STOP. ,J
| |
| - d. Place the local ENGINE MODE CONTROL switch in MAINTENANCE, 36
| |
| | |
| 87 Given the following conditions:
| |
| - The plant is shutdown making preparations for Shutdown Cooling (SDC) using the "A" loop of Rl!R
| |
| - Both Recirculation Pumps are shutdown with their discharge valves closed Which of the following describes how the "A" RilR Pump that is being started for SDC is protected from damage due to no flow?
| |
| : a. He operator is required to establish a pump discharge flow path to the reactor as soon as possible after starting the pumo.
| |
| : b. The pump minimum flow valve (F064 A) will open to provide flow until the RHR Heat Exchanger Bypass Valve (F048A) can be opened.
| |
| : c. He operator will open the minimum flow valve (F064 A) until shutdown cooling flow is greater than 500 gpm.
| |
| : d. He pump will automatically trip on low suction pressure if flow / pressure is not adequate for pump suction. ,
| |
| 88 Plant conditions are as follows:
| |
| - Suppression pool temperature 87 degrees F Suppression poollevel 20.5 ft.
| |
| Drywelltemperature 125 degrees F -
| |
| Reactor level 10.5 inches
| |
| - Aux. Bldg. pressure -0.25 j Which one of the following EOP selections should be entered?
| |
| : a. EOP I and EOP 2
| |
| : b. EOP 2 only
| |
| : c. EOP 2 and EOP 3
| |
| : d. EOP 3 only we 37
| |
| | |
| - - _ - - - _ _ ^ --__
| |
| l 1
| |
| 89 Which of the following permission / notification requirrenents mu>t be met for an INTENTIONAL entry into Tech Spec 3.0.37 Permission must be obtained from the:
| |
| : a. Operations Superintendent and the NRC Resident inspector nott'ied.
| |
| : b. General Manager - Plant Operations and the NRC Resident Inspector notified.
| |
| : c. Manager - Operations and a 1 hour report made to the NRC.
| |
| : d. General Manager - Plant Operations and a 4-hour report made to the NRC.
| |
| 90 The Plant is operating at 100% reactor power when a loss of feedwater heating occurs. Which one of the following is a required IMMEDIATE action for this loss of feedwater heating?
| |
| : a. Reduce reactor power by 40 MWE with core flow, then reduce another 110 MWE with core flow and rod inse tion.
| |
| : b. Reduce power to less than or equal to 10$% rated thermal power using core flow.
| |
| : c. If failed fuel exists in the reactor, reduce reactor power by 495 to 500 MWE.
| |
| : d. Insert control rods in reverse order to get below the 80% tod line.
| |
| 100% power.
| |
| 91 The RCIC System has just been declared inoperable, the reactor is operating a)is condition?
| |
| Which system must be demonstrated to be operable to continue operation in th
| |
| : a. HPCS
| |
| : b. ADS c.LPCS
| |
| : d. RilR LPCI Division 11 i
| |
| 38 f
| |
| {
| |
| | |
| 92 Prior to reactor startup the following SRM count rates are recorded:
| |
| SRM A 25 cps SRM B 30 cps SRM C 35 cps SRM D 15 cps At what SRM reading should the operator consult the Reactor Engineer for continued withdrawal recommendations?
| |
| h
| |
| : a. 2500 cps on SRM A
| |
| : b. 3000 cps on SRM B
| |
| : c. 300 cps on SRM C
| |
| : d. 240 cps on SRM D 93 ne plant is starting up following a refueling outage, ne reactor has just acheived criticality.
| |
| Which on: of the following statements is true regarding the requirement for Shutdown Margin (SDM) determination?
| |
| : a. SDM must be determined within four hours of criticality,
| |
| : b. SDM must be determined before proceeding further withthe startup,
| |
| : c. SDM need not be determined if no control rods were replaced. J
| |
| : d. SDM need not be determined if it was determined analytically following the last fuel movement.
| |
| 94 De basis for Surveillance Requirement 3.1.5.1 for each control rod scram accumulator pressure to be verified > 1520 psig every 7 days is to ensure adequate accumulator pressure exists to:
| |
| : a. provide sufficient scram force,
| |
| : b. drive control rods on a loss of CRD pumps.
| |
| : c. maintain indication in the readable range on the gauge.
| |
| : d. verify accuracy of control room flCU pressure indications.
| |
| ' 39
| |
| | |
| \,
| |
| 95 ' in which of the following conditions is the Operation Shi? Superintendent allowed to utilize ,
| |
| Concurrent Verification as a substitute for independent verification?
| |
| l.
| |
| : a. A Danger liold tag is being removed from a Spent Fuel Pool Cooling & Cleanup '
| |
| Valve located in a contamination area. The valve's restoration position is " Locked Closed".
| |
| : b. Completing an SOP Vlave Lineup during a refueling outage on an Instrument Air valve located in Containment. The valve's position is "Open".
| |
| : c. Conducting a routine OSP-0017 Control Board Lineup after running RHR "A" in Suppression Pool Cooling and rejecting pool water to Radwaste.
| |
| : d. An SNEO trainee is perfroming an STP on the Diesel Driven Fire Pump. The trainee signs the "perfromed by" steps and the on-shift SNEO signs the
| |
| " independent verifier" steps.
| |
| 96 in EOP.1 A, ATWS RPV Control, if SRVs are cycling, the operator is directed to manually open SRVs until RPV pressure drops to 930 psig.
| |
| Which of the following is the reason for stopping the reactor pressure reduction at 930 psig?
| |
| : a. To ensure the turbine bypass valves do not have the oppprtunity to stick closed
| |
| : b. To prevent MSIVs from closing on low main steam line pressure
| |
| : c. To minimize the amount of steam that is sent to the suppression pool A
| |
| : d. To prevent excessive loss of reactor coolant inventory 40
| |
| | |
| )
| |
| 97 'ihe plant was operating at 100% rated power when a main turbine trip occurred. Plant cenditions are as follows:
| |
| RPV water level is +18" and steady RPV Pressure is 950 psig and being maintained by the bypass valves between 800 - 1000 psig.
| |
| Reactor Recirculation Pumps are running in slow speed after an automatic transfer.
| |
| Which of the following identifies the signal which caused 'he transfer of the reci-culation pumps to slow?
| |
| : a. RPV water level 4 coincident with with a trip of a running RFP.
| |
| : b. Turbine Stop Valve closure (< 95% open) or Turbine Control Valve closure (Low ETS pressure)
| |
| : c. RPV water level 2 (-43")
| |
| : d. liigh RPV pressure (1127 psig) 98 A plant trans!:nt has occurred causing a complete isolation of the RWCU system.
| |
| - RWCU inboard and outboard isolation valves are closed.
| |
| (G 33-F001/F004T028/F034/F039/F040/F053/F054)
| |
| Which of the following conditions caused this RWCU isolation?
| |
| : a. liigh Drywell Pressure initiation signal ,3
| |
| : b. Low kenctor Water Level 2
| |
| : c. High Main Steam Tunnel Differential Temperature
| |
| : d. Initiation of SLC "A" system 41 I
| |
| | |
| p A _
| |
| a __ - - _
| |
| 99 De plant is shutdown for a maintenance outage. Work is being performed on a portion of the l'cedwater system by Mechanical Maintenance. The I&C foreman has received a Clearance Receipt to work within the feedwater system tagout boundary to calibrate an instrument. Ifpon completion of work, the htxhanical Maintenance foreman wishes to relase his clearance and restore the system, but the instrument calibration is still taking place.
| |
| What actions (s), if any, must be taken to ensure the safety olthe personnel performing the calibration?
| |
| : a. The Clearance form is transferred to the l&C foreman.
| |
| : b. De Clearance can be released with verbal permission from the I&C foreman.
| |
| : c. De I&C foreman must return his Clearance Receipt to the tagging official prior to releasing the clearance.
| |
| : d. De Mechanical Maintenance foreman may clear all tags that pertain to die I&C work.
| |
| 100 While assigned in the Work Management, Center as the WMC Supervisor, the following procedure CNs are sutimitted for your approval. Which one of the following procedure changes requires a formal 10CTR$0.59 screening be completed?
| |
| : a. A step in an SOP was inadvertantly omitted between procedure revisions and needs to be reincorporated in oider to complete an evolution which is in progress.
| |
| : b. An in progress IST STP requises use of service air, ne valve specified has a broken handwheel and cannot be used. ne CN to the procedure requests use of a different functional service air valve.,
| |
| : c. While eestoring from a bus outage IAW with OSP-0019, the procedur/ was found to contain an incorrect power supply, ne CN correcting the power supply information must be added to the procedure to complete the bus restoration,
| |
| : d. A plant start up is in progress per GOP4001, ne annunciatot ' Turbine Bypa:.s Valve Open" will not reset. A CN requests deletion of the step in GOP 0001 that verifies the annunciator is reset in order to continue the start-up. De annunciator will be wpaired when parts are available.
| |
| q
| |
| | |
| DWR RO l'.xandnation Outime l'enn l'.S-401 2 IA401 Date of Exam: 7/28/9T Exam level: RO Facitrty: River Bend Station K/A Category Points Group K K K K K K /, A A A G Point Total Tier 1 2 3 4 5 6 1 2 3 4 l
| |
| 1 ;.
| |
| )
| |
| 4 6 1 19 Emergency & 2 6 Abnormal 3 2 4 Plant i 1 Evolutions Tier ''
| |
| Totals 5 14 4 36 S 8 1
| |
| 6 4 2 0 6 _1 3 28 2 1 2 2 4 2 2 1 2 19 Plant 7 1 Systems 3 1 4 3
| |
| ^
| |
| Tier Totals 13 6 4 9 1 3 51 10 2 2 1 Cat 1 Cat 2 Cat 3 Cat 4
| |
| : 3. Generic Knowledge and Abilities 4 3 2') 4 13 Note: Attempt to distribute topics among all K/A categories; select at least one tort trom every K/A category within each tier.
| |
| Actual point totals must match those specified in the table.
| |
| Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
| |
| Systems / evolutions within each gmup are identified on the associated outline.
| |
| The shaded areas are not applicable to the category / tier.
| |
| y
| |
| * J Intennin Rev. 8, January 1997 NURlul021
| |
| | |
| !ES-401 BWR RO Examinaton Out:ne Emergency and Abnormal Plant Evolutons - Tier 1/ Group 1 Form ES 401-2 E/A PES /Name/ Safety Functon K1 K2 K3 A1 A2 G K/A Topics imp. Points 295005 Min Turbine Generator Trip ila 04 Electrical dist status foDowing a tnp of the main generator 32 1.00 295007 High Reactor PressurenII 04 Abiirty to operate / monitor SRVs dunng Hi Rx Press. L onsnts 3.9 1.00 295010 *iigh Drywell PressureN 05 Relationship between Drywell Cooling and Hi Drywell Pressure 3.7 1.00 Relationship between void concentrabon and inadvert. React.
| |
| 295014 Inadvertant Reactvrty AdC.tonn 07 addition 4.0 1.00 295015 Incornolete SCRAMA 04 Relatiorthip between RPS and imcomplete SCRAM 4.0 1.00 293724 High Drywell PressureN 19 Ability t3 operate / monitor CTMT Atmosphere Cont System 3.3 1.00 295031 Reactor Low Water Leveln1 ,
| |
| 01 Abilsty to determine Rx waier level when RPV water level is low 4.6 1.00 295031 Reactor Low Water Levetni 01 Knowledge of adequate core cooling 4.6 1.00 295031 Reactor Low Water LeveIn1 08 Ability to monitor RPV water level dunng low water tevel cond. 4.6 1.00 3CD00^ High Containment Hydrogen Conec.traton 01 Reason for Emergency Depressurization due to high M2 Conc. 3.1 1.00 293G37 SCRAM CondrDon Present and Power Above APRM Dowscale or Unknownn l04 Ability to operate / monitor SLC under ATWS co,ditions 4.5 1.00 293037 SCRAM Condrbon Present and Power
| |
| ! Above APRM Dowscale or Unknownn 03 Reactor water 'evel effects on reactor power during ATWS 4.1 1.00 i 293037 SCRAM Conditon Present and Power l Above APRM Dowscale or Unknownn 03
| |
| * Operate / monitor ARI during ATWS conditions 4.1 1.00 l
| |
| u l
| |
| I lK/A Category Point Totals: 2 2 1 7 1 Group Point Total: 13 00 l WUREG-1021 Intenm Rev. 8. January 1997
| |
| | |
| t l
| |
| [ES-401 BVJi RO Examinaton Outrne Eireigency and Abnormal Plant Evolutons - Tier 1/ Group 2 Form ES-401-2 l fE/A PES /Name/ Safety Functon K1 K2 K3 A1 A2 G K/A Topres imp. Pc.nts 2950C1 Partat or Complete Loss of Forced Crculaton ;
| |
| iaV 01 Abinty to monitor power!50w map during loss of Recire 3.5 1.00 295002 Loss of Main Cond VacuumI til 02 Knowledge of effects of loss of cond. vac. on main turt>ine 3.1 1.00 295003 Partal or Compfete Less of AC Pwr/Vi 03 Ability to operate / monitor systems required for safe shutdown 4.4 1.00 295003 Partal r.r Ccmplete Loss of AC Pwr/V1 02 Abihty to operate Emergency Generators 42 1.00 295005 High Reactor Wa*er Level 111 07 Relatonshtp between HPCS and high reactor water level 2.9 1.00 l 3.6 1.00 293013 High Suppressgen Pool Terrolv 02 Reasons for lirrutng heat additons to the SP on high temp.
| |
| 293016 Control Room Abandonment /Vil 03 Reason for d:sabling control room controls dunng evewaton 3.5 1.00 295016 Control Room Abandonmer?JVil 06 Abihty to control / monitor Rx water level from RSS panel 4.0 1 00 Operatonalimplications of a loss of CCW to system !
| |
| 295018 Partal or Total Loss of CCW / Vit! 02 operation 3.5 1.00 l Operationalimplications of a loss of CCWto system 3.5 1.00 2950 i8 Partal or To+si Loss of CCW1 V!11 02 operafirn tteactor pe essure vs. control rod insertion capabihty on a loss i 29*.,022 Loss of CRO Pumps Ii 01 of CRD 3.3 1.00 1295029 High Suppression Pool High We'er Temperatur 06 Relationship between SP termture and SP level 3.5 1.00
| |
| < 295027 High Contmnment Temperature /V 03 Relationship of CTMT cooling and CTMT temp. 3.5 1.00 295028 High Drywell Temperature / V 01 Effect of high drywell temp on RPV leve! instrumentation 3.5 1.00 implication of high SP water icvel as it applies to CTMT 295029 High Suppression F'ool Water LevellV 01 integrity 3.4 1.00 295030 Lov Suppression Pool Water LeveUV 08 Requirements of SRV drscharge submergence and SP level 3.5 1.00 ,
| |
| 295033 High Secondary Coritainment Area Rad:aton Levels / IX 03 Reason for isolating systems due to high SEC CTMT rad. lvts 3.8 1.00 295034 High Secondary Containment Ventiaton Radiaton Levels IlX 01 Ability to monitor area tad monitors in the SEC CTMT 3.8 1.00 295038 High Of'-site Refease Rate /1X 06 Operate / monitor plant ventilation dunng High Offsite Release 3.5 1.00 1 l
| |
| t
| |
| -4 i lK/A Category Point Totals: 2 6 4 6 1 Group Potnt Total: 19.00 ;
| |
| NUREG-1021 Intanm Rev. 8. January 1997
| |
| | |
| ! ' ![rr l ;:! ;[ ;!l i
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| 6 I
| |
| l ES-401 BWR RO Examenaton Outhne Plant Systems -Tier 2/ Group 1 Form ES401-2
| |
| & i System &"Name K1 K2 K3 K4 KE K6 A1 A2 A3 A4 G KIA Tocncs tmp Pcmts j 201001 CRO Hydraulic 03 Efect of a loss of Plant At on CRDH 30 1 00 201001 CRD Hydraunc 05 Knowledge of sys'em design which provides for md. Scram 38 1.00 ,
| |
| 201005 RCIS ' _
| |
| 05 Know. edge cf the concept cf control rod density 2.7 1 00 ,
| |
| 202002 Reorculaton Flow Control 06 Deterfrene the empact of a loss of AC on Recre Flew Control 2.9 1.00 203000 RHRiLPCI: Injechon Mooe 06 RHR an;ertocks that provide adeouste NPSH 35 1.00 i 203000 RHR/LPCI: Injecton Mode 05 Ab6hty to manualty operate RHR!LPCI manual mrtsten 4.3 1 00 !
| |
| . 217000 RCIC 03 EMect cf vatve closures on RCIC system 34 1 00 209001 LPCS 08 Automate system ntaten of LPCS 38 1 00 l 209002 HPCS 12 Effects of fugh SP level en HPCS cpe aton 3.3 1.00 211000 SLC 08 Abihty to monrtor SLC upon system rytaton 4.2 1.00 f 212000 RDS 01 Design feature to prevent suppfymg RPS from two supphes 30 1.00 j 215003 IRM 07 Venficaten of proper IRM operaton 36 '.A 215004 Source Range Montor 01 SRM rod withdrawal blocks 3.7 1.00 ;
| |
| 215005 APRM/LPRM 13 Relatonship between TIPS and APRMs 26 1.00 '
| |
| ! 216000 Nuclear Boiler instrumentatxm 13 Operatonal emphcatons of reference leg flashmg 35 1.C0 i
| |
| ) 217000 RCtc 02 Abst:ty to manua!!y operate RCIC tnp; thrott!e valve 39 1 00 .
| |
| I 210000 ADS 01 Know! edge of ADS logic cperaton 3.8 1 00 ;
| |
| ; 223001 Pnmary CTMT and Aurdianes 09 Predct the impact of madequate CTMT roem coolmg 26 1.00 223002 PCIS/NSSSS 08 Knowtedge of when et as permessrbie to defeat CTMT esel. 3,3 1.00 I i 239002 SRVs 03 Ab6hty to mitgate the conseouences of a stuck open SRV 4.1 1.00 f 241000 Reactor /Turb. Press Reg 02 Knowiedge of the EHC regulator m the press regutatny mode 3.3 1 1 00 259001 Reactor Feedwater 04 Predict th 2 effect of a loss of condensate pumps 36 1.00 '
| |
| 259001 Reactor Feedwater OS Connectons between RWLCS and feedwater System 36 1.00 259002 Reactor Water Level Control 02 , Effect of a loss of Feederater of the Rx Water Level Control Sy 37 1 00
| |
| ~
| |
| ; 261000 SGTS 03 Less of AC pwr on the SGTS ntaton log c 2.9 1.00 264000 EDGs 10 j Effects of a LOCA signal on the EDGs 3.5 1.00 215005 APRM/LPRM 05 Ef'ects on APRM Setponits due to a loss of recre 36 1.00 202002 Rececu!aten Flow Contret OS Impact on Rcre flow control w/ FCV lockup 33 1 00 i
| |
| ! ;K/A Category Point Totals: 3 2 1 6 4 2 0 6 1 3 0 Group Pomt Total. 28 00 l I
| |
| i i
| |
| i NURFO-1021 Intenm Rev. 3 January 1997 l l
| |
| i
| |
| | |
| {' <
| |
| , i 1 . :
| |
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| |
| i i
| |
| i BWR RO Exammaton Outrne Plant Systems - Teer 2/ Group 2 Form ES-401-2 j JES-401 i Syotems/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topics !mo - Poeres I j 201003 Centrol Rod Drhre Moch. 03 CRDM deogn which maintarns CR st a geren W:ston 33 1.00 i i 209003 Control RM Drhre beh. 01 Abihty to predict the impacts of a stuck control rod 34 1.00 ;
| |
| 201003 Control Rod Debre Mech. 04 Cause effect relationship between CRDM and Rx wate 29 1.00 t 203001 Recwcularion 09 Atphty to predict changes in Recnc Pump Seet pressur 33 1.00 l 203001 Recyculet6en 10 Pump Start gm.....__.s for Recarc Purnpo 3.3 1 00 ;
| |
| 2C1001 Racirculation 01 Effect of a loss of the jet pumps win have onRoore Sys 3.5 1 00 204000 RWCU 15 Relationship between RWCU and Leek Detection Sys. 3.1 1.00 204000 RWCU 02 Physscal connectrons between RWCU and Recaculatio 29 1.00 [
| |
| ; %5000 Shutdown Cochng 01 Effect of a loss of AC pwr on RHR SDC operaton 3.3 1.00 L j 205000 Shutdown Cooling 02 Knowledge of vaM operatons on shutdewn coolmg 2.8 1 00 ,
| |
| 239001Masn and Rehest Steem 13 Relationship between Man Steam and MSLCS 26 1 00 l l
| |
| 1 245000 Mowi Turtwo Gen /Aum. 05 Interlocks which providervMb,of the Man Turtune 2.9 1 00 g 245000 Mavi Turbine GerdAux. 07 10am g=neriy cperatons and berutatons 26 1 00 -
| |
| 207001 AC Electreat Detribution 01 Relatonship between AC Destnb. and the Emergency 38 1.0_0 !
| |
| 263000 DC Electncat Deut. 03 Effects of a lose cf DC power on Feed Level Control 34 1.00 i
| |
| , 283000 DC Electncal Drst. 02 DC electnca! distnbution LOCA anteriocks 31 1 00 271000 cegas 07 Relationship between Offges and Plant Asr Systems 2.7 1.00 390001 secondary CTMT 04 Cause effect re ee.nou SEC CTMT and SGTS 3.7 1 00 -
| |
| 39@03 Con *rol Room HVAC 04 CausqHrffect relatonship CR HVAC and NSSSS 32 1,00 t .
| |
| l !
| |
| l I i i
| |
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| |
| I l
| |
| 3 ibA Category Point s'otals: 7 1 4 2 2 1 2 Group Pomt Tetal: 19 00 !
| |
| I !
| |
| l l
| |
| ?
| |
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| |
| ; NUR EG-1021 !nterwn Rev. 8. January 1997 l .
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| A 0 ne m 1 a A x
| |
| E O E K R
| |
| R 5 W K B .
| |
| 4 5 6 1 3 K 0 0 0 ,
| |
| 3 K
| |
| 2 K
| |
| 1 K
| |
| . e n m s
| |
| p a p t
| |
| r u le s u n te t a
| |
| e C a i o E /
| |
| E le T v o lo g s t
| |
| . n s n C e i .
| |
| 9 J d V o 1 e a lo . n r P 2 a
| |
| m H Po 0
| |
| H to y 1 r
| |
| a c o G N le le le ae g s Fu Fu F R
| |
| / u e E n
| |
| t a R 1
| |
| r 0 C 0 2 U 0
| |
| 4 te 00 X 0 0 v u _ G N S y 34 s = 4 u A
| |
| /
| |
| 3 9 K E S 2 7* 2 2 ie :I; -
| |
| 4 a ;' '- I ;ll4!- 1i~ < ~ lil
| |
| - ii
| |
| | |
| Es-401 Generic Knowledge end Abihties Outhne (Tice 3) f orm ES401 $
| |
| i l
| |
| l Facihty: River Bend Sta Date of Exam: 7/28/97 Eram Level: RO Category K/A # 5 Topic imp. ' Points 2.1.2 Knowledge of operator responsibihties during all modes of operation 3.0 1 ')0 Cmduct 2.1.28 KnoMedge of purpose and function of major system cornponents 3.2 1.00 Abihty to recoDntre indications for systems which are entry conditions of 2.1.33 for technical specifications 3.4 1.00 Operations 2.1.7 Abihty to evaluate performance 3.7 1.00 2.1 Total 4.00 1 2.2.13 Know! edge of tagging and clearance procedures 3.6 1.00 l Knowledge of bases in technical specificabons for limiting con.itions )
| |
| Equipment 2.2.25 for operations and safety limits. 2.5 1.00 Control 2.2.32 Knowledge of RO duties during refuehng operations 3.5 1.00 2.2 ,
| |
| 2.2 )
| |
| Total 3.00 i Knowledge of radiation exposure hmits and contamination control, 2.3.4 including permissible levels in excess of those authorized. 2.5 1.00 2.3.2 Knowiedge of facikty ALARA program 2.5 1.00 i
| |
| 2.3 Radiation 2.3 Control 2.3 2.3 ,j 2.3 Total 2.00 Abihty to recognize abnormalindications for system operating parameters which are entry level conditions for emergency and 24.4 abnormal operating procedures 4.0 1.00 Emergency 2.4.17 Knowledge of EOP terms and definitions. 3.1 1.00 Abihty to perform with reference to procedures those actions that Procedures 2.4.49 require immediate operation of system components and controls. 4.0 1.00 KnorAedge of system setpoints, interlocks, and automatic initiatior.s and Plan 2.4.2 associated with EOP entry conditions 3.9 1.00 2.4 2.4 Total 4.00 Tier 3 Target Point Total (RO/SRO) 13.00 x
| |
| NURT.0-1021 Intenm Rev. 8, January 1997
| |
| | |
| ,6 r O l
| |
| NRC-RO KEY 6/27/37 I_M01:em I sRO-Esem T Esembenk i Answw 001 029 381 A fr.
| |
| 002 003 390 C 003 604 38T C 004 006 314 D 605 507 ~402 A 000 118 81 007 016 224 B /L 008 020 "2TO B 010 023 404 0 011 427 C 013 024 387 0
| |
| ~
| |
| 014 026 044 0 015 028 364 A 016 001 243 B 017 5'02 378 C 018 030 327 C 1 019 005 "37 3 D 020 008 251 D l d212 009 280 C
| |
| ~ ~
| |
| 022 03I 241 D /2, 023 082 038 0 ft.
| |
| 024 034 368 A /2.
| |
| 025 136 B 026 214 C
| |
| ~
| |
| 027 035 061 A /:.
| |
| 028 036 388 0 029 016 023 8 630 039 312 A
| |
| ~ _ . . , _ _ _ .
| |
| 032 018 311 0 $
| |
| 033 033 369 0 034 011 380 B 035 037 376 0 036 041 377 C~~ f,,
| |
| b37 042 B 4
| |
| 038 107 B g 039 066 122 B 040 03 2 013 D A 641 027 419 D 042 043 "140 A 043 278 D-5'4 4 045 010 D p.,
| |
| Page 1
| |
| | |
| _. - _ . _ _ _ _ . _ _ . . ~_ _ _ . _ __.
| |
| ..___1 vp l
| |
| e NRC RO KEY 6/27/97 ROfaem 55Usem Uembank [ Answer 045 046 247 .C p.
| |
| 046 255 'D 4 047 249 C f, 048 237 A /t 049 029 A 050 358 C f.
| |
| 051 051 089 8 4 052 424 D 1 j
| |
| 054 053 143 D 055 054 217 C /2 036~ 056 032 C 057 057 028 C
| |
| ~
| |
| 658 040 C 059 304 A
| |
| ~ '
| |
| 060 058 34'd 8 061 059 272 8 062 061
| |
| ~
| |
| 199 ~C 063 078 049 8 064 0$$ C 065 063 244 D 006 268 A
| |
| ~ ~
| |
| 067 027 A 068 095 A
| |
| ~"
| |
| gg '
| |
| 068 197 d f.
| |
| 070 352 8 071 OP9 422 8 072 182 C @
| |
| ~
| |
| 073 '000 i90 5 074 077 071 C 075 072 3L6 A 076 060 261 A UYI 074 421 0 078 077 A 079 075 201 A /!,
| |
| 080 062 082 C /,
| |
| 082 065 A 083 079 438 0 084 071 '004 C 086 032 0$4 C-
| |
| ~'
| |
| 087 005 A C88 084 433 D~ A Page 2
| |
| | |
| )
| |
| e NRC-RO KEY 6/27/97 ROfaem f SRO fkam ~frembank I Answer 089 087 411 A
| |
| ~
| |
| 090 089 423 A ft 091 ' j090 376 0 p.
| |
| ~ ~~ ~~ S g 037 joi,t,'~ ~~2i2 O 093 j092 057 0
| |
| ~
| |
| ~~~~
| |
| 094 ~094 414 A [L 096 096
| |
| ~
| |
| 313 U~
| |
| 096 431 A
| |
| ~
| |
| 097 099 432 C 100 088 374 fe 4
| |
| .I I
| |
| ' Page .5 i
| |
| l
| |
| | |
| 1 i
| |
| RO EXAM KEY :
| |
| Esam Number NRC-RO Rev. 0 Esaus Title NRC RO EXAM i 1he plant was operating at 100% of rated power when a reactor scram occurred. Plant conditions are as follows:
| |
| - Reactor power is on range 2 of the IRMs and decreasing 4 - The Main Turbine is tripped
| |
| - The Main Generator Ouptut breakers are closed
| |
| - Ihe Main Generator Exciter Field breaker is open
| |
| - Megawatt load on the Main Generator indicates 10 MWe (steady for 2 minutes)
| |
| - MVARs on the Main Generator indicate 50 MVARs leading (VARs in)
| |
| - Station loads are being supplied through the P:crened Station Transformers
| |
| - Main Condenser Vacuum is 28"lig and steady Which one (1) of the following describes the action (s) required by AOp-0002, Main Turbine and Generator Trips? -
| |
| : a. Immediately trip the Main Generator Output breakers.
| |
| : b. Immediately initiate a reverse power trip of the Main Generator Output breakers by decreasing generator VARs to zero.
| |
| : c. ' a action is required as long as Main Condenser Vacuum remains above 26" lig.
| |
| : d. Within 5 minutes, initiate a reserse power trip of the Main Generator Output breakers by decreasing generator VARs to zero.
| |
| ANSWER:
| |
| : a. Immediately trin % ' 'ain Generator Output breakers.
| |
| IDNO: LP# OILI #
| |
| 381 IILO-521 8 PROCEDURE NUMBER: OTHER:
| |
| AOP-0002 Sect.5 Note LEVEL 3
| |
| ~
| |
| NRC K ; } RO: I SRO:
| |
| 295005 AA1.04 l 2.7l 2.8 COMMENTS: 7/97 r4W AOP 0002, p. 7 RO T1 G1 SRO T1 G2 AOP 0002 note on page 7 require that the Main Generator Output breakers be immediately
| |
| , tripped if the Exciter Field Breaker has tripped.
| |
| t' t
| |
| I
| |
| | |
| l RO EXAM KEY Exam Number NRC-RO Rev, O Exam Title NRC RO EXAM 2
| |
| Which of the following states the overall low low set system response of the SRVs for the RPV i pressures given?
| |
| : a. At 1813 psig, only one valve will be open. It recloses at 926 psig. l l
| |
| : b. At 1113 psig, only one valve will be open, it recioses at 936 psig.
| |
| l
| |
| : c. At 1103 psig, only two valves will be open. One recloses at 936 psig, die other at 926 psig.
| |
| : d. At 1103 psig, only eight valves will be open. Three reclose at 946 psig, three at 936 psig, and the last two at 926 psig.
| |
| ANSWER:
| |
| : c. At 1103 psig, only two valves will be open. One recloses at 936 psig, the other at
| |
| ; 926 psig.
| |
| IDNO: LP# OILI #
| |
| 390 IILO-007 4 PROCEDURE NUMBER: OTHER:
| |
| ARP-lH3-P601 19-H08 LEVEL 3 ARP-H13-P601 19-H11 'j TS 3.3.6 4
| |
| ~
| |
| NRC N E ~l ROi T S507 296007 AA1.o4 } 3.9! 4.1 _
| |
| COMMENTS: 7/97 new RO T1 G1 SRO T1 G*
| |
| tOTM 24 6, p 6, and Table 2, p.21 of 26 2
| |
| y.-.. g - - .*-e v 5 ,,, - -- .-my - -a--g- wr, -
| |
| | |
| t RO EX.AM KEY Essa Number NRC RO Rev. 0 Esam Title NRC RO EXAM 3
| |
| You have been instructed to control drywell temperature and p.7ssure by operating ell avaliable drywell cooling. While doing this, service water to the drywell unit coolers automatically isolates.
| |
| 1 Which of the following caused the isolation? l
| |
| : a. liigh drywell temperature (max. recorded 265 deg F)
| |
| : b. Low RPV water level (min recorded 28")
| |
| : c. liigh drywell pressure (max. recorded 1.82 psid)
| |
| : d. Loss of 120 VAC power 4
| |
| ANSWER:
| |
| : c. Drywell pressure 1.82 psid IDNO, LP# OILI #
| |
| 382 IILO-522 3 PROCEDURE NUMBER: OTHER:
| |
| AOP-0003 31GNAL 0 LEVEL 3
| |
| 'j
| |
| ~
| |
| NRC KA: I RO: I SROI 295010 AK2.05 i 37l 3. 8 .__
| |
| COMMENTS: 7/97 riew LOTM 63 4, p.11 RO T101 1 SRO T101 4
| |
| a O
| |
| 3
| |
| | |
| . _ . _ - . _ . _ _ _ . - _ _ _ _ . - _ _ _ . ~ . _ _ _ . . . _ _ _ _ _ _ - _ . . . _ . . _ _ _ _ . _ . _ _ - . . . . _ . . ._.. .____
| |
| RO EXAM KEY Exam Number NRC RO Rev. O Esam Title NRC RO EXAM 4 Which of the following will result in the addition of positive reactivity to the reactor? (Consider each case separately.)
| |
| : a. LPCS initiaiton during reactor STARTUP widt reactor pressure at 500 p;ig.
| |
| 4
| |
| : b. Sudden jet pump differential pressure reduction in one loop with the reactor in the
| |
| - RUN mode,
| |
| : c. Reduction in EllC pressure setpoint by 2 psig with reactor in the RUN mode,
| |
| : d. Initiation of the RC IC turbine during reactor STARTUP with reactor pressure at i50 psig.
| |
| ANSWER:
| |
| d, Initiation of the RCIC turbine during reactor STARTUP with reactor pressure at 150 psig.
| |
| - IDNO: LPW OBJ #
| |
| 314 11L0-318 2 PROCEDURri NUMBER: OTHER:
| |
| GOP-0001 S'EP 3 8 LEVEL 2 I NRC KAt ~ R57'l SRO:
| |
| (296014 AA1.07 4I 4.1 _
| |
| COMMENTSt 7/97 flew ROT 1On SRO T1 G1 i
| |
| e 4
| |
| ~~~ o e ,.. , n -- -.-,-.,--,--,-..--re,-em,---ps,, .---m.mo, .r-=,w n-- -
| |
| ,ae - m m g w ,
| |
| | |
| r RO EXAM KEY t Esam Number NRC RO Rev. 0 Exam Title NRC RO EXAM j Which of the following methods for alternate control rod insertion during an ATWS REQUIRES the scram " signal" ta be reset?
| |
| Control rod insertion by;
| |
| : a. using the individual control rod scram test switches.
| |
| : b. venting the control rod mechanism over-piston volume.
| |
| : c. maximizing CRD cooling water differential pressure.
| |
| : d. venting the scram air header.
| |
| ANSWER:
| |
| : a. using the individual control rod scram test switches.
| |
| IDNO: Ll' # Olu #
| |
| 402 IILO 513 4 PROCEDURE NUMBER: OTHER:
| |
| EOP 0005, Etcl 13 LEVEt 3
| |
| ~~
| |
| ~MC K A: l RO:_.} SRO: ' ,)
| |
| 295015 AA2.04 1- 4! 4.1 COMMENTS: 7/97 New 5
| |
| | |
| - _ ._ _ _ _ ~ . _ . . - _ . . . . . _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ -
| |
| I 1
| |
| RO EXAM KEY i
| |
| Exam Number NRC.RO Rev. 0 Eram Title NRC RO EXAM l 6 EOP 2 requires containment to be vented when containment pressure reaches 20 psig. Which one of l the following describes the flow path for emergency venting of the containment?
| |
| : s. Drywell/ Containment purge system takes a suction on containment and discharges through filter train #6 to the main stack.
| |
| : b. Ilydrogen purge discharges to the annulus and the annulus mixing system is in operation discharging to SGTS.
| |
| 3
| |
| : c. Drywell/ Containment purge fan takes a suction on containment and discharges through the purge exhaust fans to the main stack,
| |
| : d. Ilydrogen purge discharges through the Drywell/Conta8 ament purge system filter train 86 to the main stack.
| |
| ANSWEP.:
| |
| : b. Ilydrogen purge discharges to the annulus and the annulus mixing system is in 4 operation discharging to SGTS.
| |
| IDNO: LP# Olki #
| |
| l18 IILO-516 21 PROCEDURE NUMBER: OTHER:
| |
| COP-0002 CP-8 LEyEL 4 EOPM5 ENCL 21 I! SROt ]
| |
| ~
| |
| 295024 NRC kAT~
| |
| E A1.19 I '3 l dd'.3!
| |
| 3.4 J COMMENTS: 7/97 new RO T101 SRO T1 G1 I
| |
| t O
| |
| 6 4
| |
| .r., ..m- .-,_..-.3 - - , . - ,- - , - , _ , . . , - -e--- --..,.--~w. wee-w r- rr--- v-, -
| |
| n w m. w ,, -
| |
| | |
| ENCLOSURE 21 EMERGENCY CONTAINMENT VENTINO AND. DEFEATING CONTAINMENT VENT PATli ISOLATION INTERLOCKS j 3.6 VENT Primary Containment as follows:
| |
| 3.5.1 OBTAIN $0P 0005 ENCL 21 key, one (1) for CPP PNL102. []
| |
| 3.C.2 VEIUFY the Annulus Mixing System in operation with flow to SGTS. []
| |
| 3.6.3 h SGTS in operation with flow to the rnain plant exhaust duct.
| |
| VERIFY t e []
| |
| 3.6.4 VERIFY llAS'MOV107 INST AIR SHUTOFF VALVE idlD llAS*MOV106 INST AIR OUTBD ISOL are open (IH13*P870). []
| |
| 3.6.5 V ERIFY IHVR* AOV128 CONTMT RTN INBD ISOL is open (IH13*P863). []
| |
| 3.6.6 OPEN IHVR AOD127 CONTMT PURGE RTN ISOL (IH13'P863). []
| |
| 3.6.7 OPEN 1CPP'MOV105 H1 PURGE FAN DISCH VALVE TO ANNULUS at 1CPP PNL102 (171 fLAux Bldg East Side, Containment Purge FLT 6/HVR FAN 14 Room). []
| |
| 3.6.8 lVHEN directed to secure Containment venting, TIIEN SlIUT 1CPP*MOV105 at ICPP PNL102. []
| |
| | |
| ==4.0 REFERENCES==
| |
| | |
| 4.1 PID 22.IB 4.2 PID 27-21 A 4.3 ESK-7HVR10 4.4 ESK-11HVR03 4.5 LCR 1.lLHVR.123 4.6 Conn. Diag. 914E582, Sh. 87, Sh. 91 PAbE3OF5 REV.9 PAGE 74 OF 115 ENCLOSURE 21 [ EOP 0005 l
| |
| L __ _ _ _ _ _ __
| |
| | |
| i RO EXAM KEY Esam Numhet NRC-RO Res. 0 Exam Title NRC RO EXAM 7 EOP-4 (I% nary Containment Flooding) is executed to flood containment.
| |
| The coreeinment level band specified by EOP 4 is between 62 ft and the Maximum Containment
| |
| , Water Level Limit (MCWLL)of 85 ft.
| |
| WI..'ch of the following ranges corresponds to the core level band specified by EOP 47 l
| |
| : a. 143 to +133 Inches
| |
| - b. 162 to +114 inches
| |
| : c. 193 to +51 inches
| |
| : d. 205 to +7I inches ANSWER:
| |
| i
| |
| : b. 162 to + 114 inches r
| |
| IDNO: LP # Olki #
| |
| 224 IILO-512 5 PROCEDURE NUMBER: OTHER:
| |
| EOP m LEVEL 3
| |
| ~ ~
| |
| NRC KAr ITO7I JR,L R
| |
| 295031 E A2.01 I 4.6i 4.6 _
| |
| COMMENTS: 7/97 new COP BASES says MCWLL is 85 ft.RBS NRC 01-95 RO T1 G1 SRO T1 G1 9
| |
| 7 7
| |
| -n - , - - . . . ..,m-- ,,. e e .a , . , . . . . . - -, . ,._,. ..-,..,.._---.m . , , , - , . ..,_.,.,,,n ,n.,w~ , , - , - - - , . - - ,
| |
| | |
| . . - . . . . _ - - . - - . . . _ _ . .- . - , . . - . . . . - . - . ~ - . . . . . . -
| |
| RO EXAM KEY Esam Number NRC-RO Rev. O Esam Title NRC RO EXAM g Which of the following conditions still constitutes " Adequate Core Cooling"?
| |
| NOTTI: Only the injection sources stated are injecting. Regard each situation separately.
| |
| : a. ATWS in progress, the feed system is maintaining level between -205 inches and -195 inches, MSIVs are open.
| |
| : b. Alli o3s in, MSIV/ ADS valves are closed, RPV level is 200 inches and RPV pressure is 200 psig.
| |
| : c. All rods in, RCIC is injecting,1 ADS valve is open, RPV level at 210 inches and MSIVs are closed.
| |
| : d. ATWS in progress, CRD, RCIC and SLC (with Boron) are injecting, RPV level is 200 inches and MSIVs are open.
| |
| ANSWER:
| |
| : b. All rods in, MSIV/ ADS valves are closed, RPV level is 200 inche3 and RPV pressure is 200 psig.
| |
| IDNO: LP# Olu #
| |
| 230 llLO Sil 3 PROCEDURE NUMBER: J OTHER:
| |
| EOP 1 A RLA-11 LEVEL 4 NRC KAt ~E07 SRO:
| |
| 296031 EK1.01 4.6 4.7 _
| |
| COMMENTS: 7/97 new 8
| |
| | |
| i RO EXAM KEY Esam Number NRC RO Rev. O Esam Title NRC RO EXAM 9 All high pressure injection has been lost following a Reactor scram and RPV water level transient.
| |
| RPV water level is 162 and lowering slowly. RilR and LPCS are running on minimum flow. The CRS directs emergency depressurization. Why must at least 4 SRVs be opened to accomplish emergency depressurization under these conditions?
| |
| - m. LPCS and RilR will be injecting prior to RPV level reaching the minimum steam cooling level.
| |
| : b. The level swell from four open SRVs will keep the core submerged until RilR and LPCS at injecting at rated Dow.
| |
| 9
| |
| : c. LPCS alone can reflood the core prior to the core uncovery time exceeding the maximum core uncovery time limit.
| |
| : d. Enough SRV steam flow to cool the core will exist at a '
| |
| pressure that RilR can make up for the steam flow.
| |
| ANSWER:
| |
| : d. Encugh SRV steam flow to cool the core will exist at the pressure RilR can make t'p for the steam flow.
| |
| IDNO: LP# OEJ #
| |
| 417 IILO 512 7 J OTHER:
| |
| PROCEDURE NUMBER:
| |
| EOP4)04 ED-5 LEVEL 3 295031 E A1.08 { 3.BI 3.9 _
| |
| COMMENTS: 7/97 new 9
| |
| | |
| l RO EXAM KEY Exam Number NRC.RO Rev. O Exam Title NRC RO EXAM 10 Regarding the Ilydrog,n DeGagration Overpressure Limit (llDOL) curve, as containment pressure '
| |
| increases, the maximum allowed hydrogen concentration in percent (%) decreases.
| |
| Which of the following is the reason for this relationship?
| |
| : a. As containment pressure increases, the capabilities of the flydrogen Recombiners to remove hydrogen is decreased.
| |
| b.1his ensures a hydrogen dcDagration at the limit combined with current pressure will not exceed containment overpressure failure limits.
| |
| : c. The containment hydrogen analyzer system response time is adversely affected as pressure increases.
| |
| : d. As contdnment pressure increases, the dcDagration pressure of hydrogen decreases requiring a lower concentration of hydrogen, ANSWER:
| |
| : b. This ensures a hydrogen denagration at the limit combined with current pressure will not exceed containment overpressure failure limits.
| |
| II)NO: LPts OBJ #
| |
| 404 llLO-514 2
| |
| .l PROCEDURE NUMBER: OTHER:
| |
| EPSTGW2, App. A LEVEL 3 EOP 1 F6gure 5 I NRC KA: I RO: I SRO:
| |
| (600000 EK1.01 i 3.3! 3.9 _
| |
| COMMENTS: 7/97 new RO T101 SRO T1 G1 10
| |
| | |
| _. ____m. _ _ . _ _ _ __ _ _ _ . . _ . . . _ _ .___ - . _ _ _ _ . _ . _ _ . . . _ _ _ _ . . . _ . _.
| |
| RO EXAM KEY Esam Number NRC RO Rev. O Esam Title NRC RO EXAM i1 EOP 1 A. "RPV Control ATWS," directs the operator to inject boron into the RPV with SLC pump "A" or "B" but not both.
| |
| Which of the following is the reason that EOP 1 A "RPV Control ATWS," specifically prohibits starting both Standby Liquid Control Pumps?
| |
| : a. The pumps are interlocked,"B" will not start if"A" is running.
| |
| : b. The *B" system explosive valve will not fire if the "A" explosive valve has been fired.
| |
| : c. Excess discharge pressure will lift both pump reliefs thus reducing Boron injection flow.
| |
| : d. Using two pumps would inject Boron at an excessive rate, preventing adequate mixing in the reactor.
| |
| ANSWER:
| |
| : c. Excess discharge pressure will lift both pump reliefs thus reducing Boron injection flow.
| |
| IDNO: 1,P# OILI#
| |
| 427 IILO-016 3 J OTHER:
| |
| PROCEDURE NUMBER:
| |
| EOP.1 A RQA.17 LEVEL 3
| |
| [,,,f!RC KAt l RO: I SRO:
| |
| 12950037f A104 4! A COMMENTS: 7/97 new RO 1101 SAO T1 G1 Simdar to 229
| |
| +
| |
| 1I
| |
| | |
| _ _ .- . . . - . _. - - _- _ _ = - _ _ - - - .. . .. -
| |
| - EOP I A RPV Control- ATWS RQA t S& . < : ~~
| |
| ... , , .n : ::: T:,
| |
| STEPS RQA 18, RQA-19, RQA-20[ 7 , (,,;-
| |
| .,L
| |
| =
| |
| Steps RQA 18 through RQA 21 give specific directions for injecting boron into the RPV.
| |
| The normal method of boron injection willinvolve use of one of the SLC pumps (RQA 18).
| |
| Although the SLC system is highly reliable, a number of shared components (such as the storage tank and injection lines) make SLC susceptible to a single failure. In cases where the SLC system cannot be used for boron injection, alternative methods must be employed. Step RQA 20 addresses these actions.
| |
| o B - 134 Revision 3 FPSTG*0002
| |
| | |
| TECHNIQUES VIL_ SUILIECT MATTFE 2.3.1 Each pump has its own Motor Operated i (1C41*F001A/B) and manual isolation i suction valves (1C41*VF002A).
| |
| 1 2.3.2 . Pump suctions are cross connected between Manual and Motor Operated suction valve to allow use of either
| |
| - pump with the opposite suction.
| |
| 2.3.3 Test tank (1C41-TKA002) outlet taps into pump suction on the cross connect line.
| |
| i e This provides a means of testing the pumps without using SLC ,
| |
| solution.
| |
| * This connection, through a restricting orifice (C41*RO-D001), also provides Condensate Transfer to the suction lines of the pump to maintain a positive pressure so,that the SLC solution vill not enter the pump suction lines. .
| |
| 2.3A Also between suction valves are taps for system draining.
| |
| 2A Each pump has an associated suction -
| |
| strainer (SLS*STRT1A/B) 2.5 The SLC pumps each discharge through an ,
| |
| associsted discharge check valve (1C41*VF033A/B) and manual isolation valve (1C41*VF003A/B).
| |
| 2.5.1 Between pump and check is a relief valve which discharges back to pump suction to. protect the pump on an over
| |
| , pressure condition.
| |
| HLO4165 PAGE 11 of 29 q- m -g.p- mi.- ,.a.=, g 6--g.-i , , ,g ,y,er,y,,wmg-e -s,yy.ggeppsyyw.-y-,,,.-,eq. py,g,,,,.w,.ym, p.y.-ry.w.y ets.yayfeywr.-.-a,-ww-.,- 4em p g ---m, wg wgge & gg 3-utg--r
| |
| | |
| l TECHNIOUES vfl SUILIECT MATTFD 2.52 Each pump has an associated Explosive Valve in its discharge line (1C41* VEX F004A/B) 2.6 Located between the discharge manual isolation and the Explosive valve is a cross connect to allow either pump to be used with the other flowpath.-
| |
| 2.6.1 There is also a connection for pump discharge to the test tank.
| |
| )
| |
| 2.62 On each pump discharge line is a connection for system draining.
| |
| l 2.7 Pump's discharge combine and pass through a check valve (1C41*VF006) prior to passing into the drywell. '
| |
| 2.8 After passing through the drywell penetration there is another check valve (1C41*VF007) and an injection line manual
| |
| , isolation valve (1C41*VF008).
| |
| L
| |
| .)
| |
| 2.9 After the injection line manual isolation valve (1C41*VF008) the line enters the RX vessel below core plate.
| |
| : 3. Component Description OBJ #3 3.1 Pumps 3.1.1 Two 100% capacity, Triplex plunger, positive displacement pumps-
| |
| * Rated Flow of 43 gpm from 0-1400 psig.
| |
| * Overpressure protection is provided by relief valve set at 1325 psig.
| |
| HLO-016 5 PAGE 12 of 29
| |
| | |
| B. Component Description :
| |
| : 1. Pumps Bere are two pumps associated with the Standby Lipid Control System. Each is a 100% capacity, triplex plunger, positive displacement pump. The pumps are designed to deliver a rated flow of 43 gpm against a back pressure of 0-1400 psig.
| |
| De SLC pump is capable ofinjecting the net contents of the storage tank (with a sodium pentaborate enrichment of 65 atom % Boron 10) into the reactor vessel in not less than 35 minutes and not more than 125 minutes against zero psig up to the initial setpoint of the reactor safety relief valves.
| |
| Normal operation of the pumps is from panel P601 in the control room. Local switches are also provided at H22 P011 in the reactor building for system testing.
| |
| The local switches, unlike the remote switches, do not activ.ite the explosive valves. He piping between the pump outlet and discharge valve has a relief valve which relieves at 1400 psig back to the pump suction line. It prbtects the pump from overpressurization in the event that the discharge path to.the reactor becomes blocked.
| |
| 2.. Storage Tank .
| |
| ne SLC storage tank is a stainless steel, atmosphe'ric tank,12.6 feet in height and 9 feet in diameter, with a maximum capacity of 5150 gallons.
| |
| Two immersion heaters are installed near the bottomj of the tank to elevate the temperature to between 135'F - 150*F during chenhcal mixing to increase the solubility of the sodium pentaborate in water.
| |
| Instrument air is supplied to a perforated sparger located in the bottom of the tank.
| |
| The 30 psig instrument air is used during poison solution mixing. Air is also supplied, via flow regulator R004, to a dip tube'for tank level indication. Service air is supplied for mixing during chemical addition. .
| |
| He pump suction lines are located on the side of the tank to minimize plugging from foreign material or from precipitation of the poison.
| |
| : 3. Pump Suetion Valves ne suction valves, F001A and B. are motor operated glob: valves with handwheels for manual operation. Each valve is located in.the pump suction line near the storage tank outlet. A cross-connect between the two suction lines located dowiistream of the suction valves allows either pump to draw a suction through either valve.
| |
| He opening of each valve is controlled by the pump start switch located in the main control room or the valve control switch on the locsl panel. Local switches LOTM 16-5 Page 5 of18
| |
| | |
| l l
| |
| l RO EXAM KEY Enam Number NRC-RO Rev. O Exam Title NRC RO EXAM i2 SELECT the reason why terminating and pre nting injection during a failure-to-scram (ATWS) tiansient results in a power reduction.
| |
| Terminating and preventing injection:
| |
| : a. results in increased core inlet subcooling as feed preheating is decreased.
| |
| : b. increases the void fraction by a reduction in core natural circulation flow.
| |
| : c. increases water temperature as level is decreased.
| |
| : d. results in an increase in fuel temperatures as core steaming rate increases.
| |
| ANSWER:
| |
| b increases the void fraction by a reduction in core natural circulation flov>.
| |
| IDNO: LP# OBJ #
| |
| 428 IILO-512 4 PROCEDURE Nty > 'n- OTHER:
| |
| EOP 1A icsEL3
| |
| ^~kRC KAi I RO: l SRO: 'j 295037 EK1.02 4 4 295037 .K3.0i di 4_
| |
| COMMENTS: 7/97 similar to 235 12
| |
| | |
| EOP-4 A Contingencies - ATWS - LP 9,
| |
| STEP LP-4 Step LP-4 directs the user to deliberately lower RPV water level to effect a reduction in h .
| |
| reactor power. Lowering RPV water level reduces the natural circulation drMng head and core flow, thereby reducing reactor power and the heat rate to the suppression pool. This process c curs as follows:
| |
| )- 1. The reactor is in a natural circulation mode following recirculation pump trip (EOP-1 A step RQA 9). The natural circulation driving head is a function of the
| |
| } fluid density difference between the regions inside and outside the core shroud J (void fraction directly affects the fluid density inside the core shroud) and the height of fluid columns (RP\Hvater level).
| |
| : 2. As RPV water level is lowered, the height of fluid columns is reduced, thereby ~
| |
| 3 reducing the natural circulation driving head.
| |
| : 3. As the natural circulation driving head is ref L the natural circulation flow through the core is reduced. ,
| |
| : 4. The reduced core flow results in a redus .d rate of steam femoval from the core.
| |
| : 5. The reduced rate of steam removal results in and increased void fraction inside the core shroud.
| |
| : 6. The increased void frac: ion adds negative reactivity to the reactor.
| |
| : 7. The negative rea:tivity drives the reactor slightly suberitical and power begins to dectease.
| |
| S. The reduced power results in a reduced steam generation rate.
| |
| : 9. The reduced steam generation rate results in a reduced void fraction.
| |
| : 10. When the void fraction drops to its original value (with some slight adjustment to account for reduced Doppler reactivity), the wactor returns to criticality at a lower power. .
| |
| These intetrelationships between RPV water level, natural circuiation core flow, and reactor power (graphically illustrated in Figures B-10, B-11, and B-12) have been observed in c.. rating BWRs with RPV water level in or near the normal operating band. Computer analyses and scale modpl test have confirmed the continued validity of these fundamental thermal-E
| |
| | |
| ~
| |
| i EOP-4 A Contingencies - ATWS . LP hydraulics and react r physics principles for RPV water levels at or below the elevation of the steam separators.
| |
| Lowering RPV water level is accomplished by terminating and preventing all injection int?
| |
| the RPV, except from boron injection systems and CRD since these two cystems are low flow and may be needed to establish and maintain reactor shutdown conditions. With essentially no makeup of reactor coolant, RPV water level then decreases by boil off.
| |
| Terminating and preventing injection shall be accomphshed per the following:
| |
| : 1. Condensate /Feedwater - Place the Master Controller in Manual and drive all Feed Reg valves full closed; place the Startup Feed Reg valve in Manual and drive it full closed.
| |
| : 1. HPCS - While holding the control switch for the injection valve in the closed position, arm and dqoress the manualinitiation push button. When the pump has started and come to full speed, secure the pump at the SRO's direction. When available, dispatch an operator to monitor and secure the diesel. .
| |
| RCIC - Depress the manual trip push button ,-
| |
| 3.
| |
| < LPCS/LPCI - If an initiation signal is present, manually override the injection valves by taking the control switches to the closed position. The pumps shall remrin running. If an s initiation signal is not present, prevent pump start and injection valve opening by defeating the initiation logic. Enclosure 27 provides guidance.
| |
| ) ,
| |
| J
| |
| : 5. ECCS Keeofill/Condensat;Inmfe_I- Close the appropriate LPCS/ LPCI injection valves.
| |
| : 6. Service Water / Fire Water - Close the Containment Flood valves, E12*F094 and E12*F096.
| |
| Power oscillations may occur when RPV water level is lowe,ed significantly below the normai operating range with the reactor all at power. The:;e oscillations have been analyzed and determined to result in thermal tranients well within the design capabilities of the fuel. The oscillations are noted at this point tc. indicate to operators that they are expected, and were considered in developing the step which require deliberately lowering RPV water level with the restor at power.
| |
| EPSTG'0002 B - 307 Revision 3
| |
| | |
| RO EXAM KEY Eram Number NRC.RO Rev. 0 Exam Title NRC RO EXAM 13 The plant was operating at 100% power when a scram signal was generated and the rtactor failed to scram. EOP-1 A directs downshifting Recirc Pumps.
| |
| Which of the followmg describes the reason Recirc Pumps are down shifted prior to tripping?
| |
| - Tripping the Recirc Pumps could result in:
| |
| : a. entering the region of thermal / hydraulic instability,
| |
| : b. an excessive feedwater teinperature reduction rate that will cause power to increase rapidly,
| |
| : c. a large level shrink which could cause isolation signals complicating the event.
| |
| : d. a reactor level swell which could result in a main turbine trip.
| |
| ANSWER:
| |
| : d. a reactor level swell which could result in a main turbine trip.
| |
| IDNO: LP# OfU #
| |
| 387 ilLO-512 5 b
| |
| PROCEDURE NUMBER: OTHER:
| |
| EOP 0001A LEVEL 3 EPSTG'0002 NRC KA: I RO: I SRO:
| |
| 295037 EA1.02 3.0! 4 295037 [ A1.03 4.1I 41 COMMENTS: 7/97 new EOP 1 A. B 369 RO T1 G1 SRO T1 G1 O
| |
| 13
| |
| | |
| RO EXAM KEY .
| |
| Exam Number NRC-RO Rev. O Eram Title NRC RO EXAM 14 During operation at 100% power with a rod line of 100%, the "A" Recirc Pump inadvertently trips to off. About 20 seconds itter, the "B" Recirc Pump trips to slow speed, resulting m the following .
| |
| steady state plant conditions:
| |
| Thermal power 50%
| |
| Calculated core flow 34 %
| |
| What is the required operator action?
| |
| : a. Immediately SCRAM the reactor, L Reduce thermal power to less than 40% by inseding control rods.
| |
| i
| |
| : c. Raise core flow by upshifting the B recirt, pump to FAST.
| |
| : d. Raise core flow by starting te A recirc pump in SLOW.
| |
| ANSWER:
| |
| : b. Reduce thermal power to less than 40% by inserting control rods.
| |
| IDNO: LP ## OBJ #
| |
| 44 IILO-534 15
| |
| .)
| |
| PROCEDURE NUMBER: OTHER:
| |
| AOP-0024 LEVEL 3
| |
| [ NRC KA: 3d[ l SRO:
| |
| 1295001 AA2.01 3 51 3.8 _
| |
| COMiAENTS: 59[ Is'.
| |
| AOP-24 N/F Map RO T1 G2 SRO T1 G2 14 l
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. O Exam Title NRC RO EX AM 15
| |
| /. loss of condenser vacuum has occurred, vacuum is currently 18.5" Hg. Which of the following automatic actions should have occurred?
| |
| : a. Turbine trip only,
| |
| : b. Turbine trip and bypass valve closure.
| |
| : c. Turbine tfip and MSIV isolation.
| |
| : d. Turbine trip, bypass valve closure and MSIV isolation.
| |
| ANSWER:
| |
| : a. Turbine trip or.ly.
| |
| i.
| |
| IDNO: LP# OBJ #
| |
| 364 IILO 524 01 PROCEDURE NUMBER: OTHER:
| |
| AOP 0005 LEVEL 2 .
| |
| NRC KA: I RO: I SRO:
| |
| 295002 AK2.02 l 3.11 3.2 COMMENTS: 7/97 new ROT 1G2 SRO T1 G2 S
| |
| e IS
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. O Exam Title NRC RO EXAM 16 The following conditions exist:
| |
| The plant has experienced a station blackout.
| |
| The Div 3 Diesel Generator was started and is runninr, normally.
| |
| Emergency use of Div 3 for decay heat removal and RPV level control is being imple:nented.
| |
| Which of the following describes the general flowpath for this cooling mechanism?
| |
| : a. CST IIPCS pump RilR "A" heat exchangers - RPV Shutdown cooling drsins to Suppression pool.
| |
| : b. Suppression pool IIPCS pump RPV - shutdown cooling to loop "A" RilR heat exchangers then test retum to suppression pool.
| |
| : c. CST - IIPCS pump - RPV - shutdown cooling to loop "B" RilR heat exchanger then test retum 'o Suppression pool.
| |
| : d. Suppression pool HPCS pump RilR"A" heat exchang:r - RPV shutdown cooling drains to suppression pool.
| |
| ANSWER:
| |
| : b. Suppression pool- HPCS pump RPV - shutdown cooling to loop "A" RI.R heat exchangers then test retum to suppression pool.
| |
| IDNO: LP# OBJ # ,)
| |
| 243 IILO-541 6 PROCEOURE NUMBER: OTHER:
| |
| AOP@50 ATT 6 LEVEL 4 NRC K A: I- RO: I SRO:
| |
| 295o03 AM.03 1 4.41 4.4 _
| |
| COMMENTS: 7/97 fl0W ROT 1G2 SRO T1 G1 ALSO OBJECTIVE 4 16
| |
| | |
| RO EXAM KEY Eram Number NRC-RO'. Rev. 0 Exam Title NRC RO EXAM 17 A Loss of Offsite Power has occuTed. '!he Division I Diesel generator is currently loaded to 2500 KW.
| |
| Which one of the following is the MAXIMUM sllowed additional load that can be imposed on the gen <.rator?
| |
| : a. 360 KW
| |
| : b. 580 KW
| |
| < c. 630 KW
| |
| : d. 730 KW ANSWER:
| |
| : c. 630 KW IDNO: LP it OBJ #
| |
| 378 IILO-037 7 PROCEDURE NUMBER: OTHER:
| |
| AOP-0004 LEVEL 2 A
| |
| - NRC KA: RO: I SRO:
| |
| 295003 AA1.02 4.21 4.3 ,
| |
| COMMENTS: 7/97 new AOP 0004, Rev 9, Caution, p. 3 of 39 ALSO AN OBJECTIVE OF HLO-523 RO T1 G2 SRO T1 G1 17
| |
| | |
| RO EXAM KEY Esim Number NRC-RO Rev. 0 Exam Title NRC RO EXAM ig A small-break LOCA has occurred. Reactor level initially fell to -47 inches, then llPCS initiated and filled the reactor to a maximum of +55 inches, level is now steady at +40 inches. Which of the following describes the current status of E22-MOV F004, the llPCS injection isolation valve?
| |
| a F004 will open on a liigh Drywell Presture initiation signal even if the llPCS 111G11 WATER LEVEL 8 RESET pushhutton has not been depressed and the llPCS lilGil WATER LEVEL 8 RESET pushbutton must le depressed before the valve can be opened manually,
| |
| : b. F004 can be opened manually even if the llPCS lilGil WATER LEVEL 8 RESET peshhutton has not been depressed and the llPCS 111G11 WATER LEVEL 8 RESET pushhutton must be depressed before the valve will open on a liigh Drywell initiation signal ,
| |
| : c. F004 can NOT be opened manually and it will NOT open on a liigh Drywell Pressure initiation signal until the llPCS 111G11 WATER LEVEL 8 RESET pushhutton is depressed.
| |
| : d. F004 can be opened manually after the llPCS INITIATION RESET pe#>utton is depressed.
| |
| ANSWER:
| |
| : c. F004 can NOT be opened manually a .d it will NOT open on a Iligh Drywell Pressure initiation signal until the llPCS lilGil WATER LEVEL 8 RESET pgshhutton is depressed. .1 IDNO: LP# OBJ #
| |
| 327 IILO-019 4 PROCEDURE NUMBER: OTHER:
| |
| SOP @30 LEVEL 3 NRC KA: l RO: I SRO:
| |
| 295008 AK2.07 I 2.91 3 COMMENTS: 7/97 new SOP-0030, Rev 8, page 2 of 34,2,5 LOTM 3 4. Table 8, p. 26 of 31 RO T1 G2 SRO T1 G2 18
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev 0 Exam Title NRC RO EXAM 19 EOP 2, " Primary Containment Control", requires the reactor to be scrammed before suppression pool temperature reaches 110 Degrees F. Which one of the following states the reason for this o requirement?
| |
| : a. Assures that the containment design pressure will not be exceeded due to compression of the non condensable gasses due to the higher water temperature.
| |
| : b. Assures that with the expected temperature rise of 7v Degrees F during the blowdown phase of an accident, that complete condensation of reactor coolant .<ill occur.
| |
| : c. Assures the post LOCA suppression pool hydrodynamic forces are within the design limitation of containment.
| |
| : d. Assures a reactor shutdown occurs, to minimize heat rejected to the primary containment, if Emergency Depressurization is required.
| |
| ANSWER:
| |
| : d. Assures a reactor shutdown occurs, to minimtze heat rejected to the primary containment, if Emergency Depressurization is required.
| |
| IDNO: LP# OBJ #
| |
| 373 IILO-514 5 PROCEDURE NUMBER: OTHER:
| |
| EPSTG*0002 LEVEL 4 NRC KA: I RO: SRO:
| |
| 295026 EK2.05 1 3 3.3 295013 AK3 02 l 3.6 3.8 COMMENTS: 7/97 new EPSTG'0)02, p. B 216 RO T1 G2 SRO T1 G1 19
| |
| | |
| RO EXAM KEY -
| |
| Eram Number NRC-RO Rev. 0 Exam Title NRC RO EXAM 20 The Remote Shutdown Panel emergency transfer switches (division I switch on C61'P001 and divsion 11 switch ca RSS*PNL102) for ADS /SRV B21 F0510 are in the EMERGENCY position.
| |
| Which of the following Control Room handswitches can be used to msnually open ADS /SRV B21 F051G7 ,
| |
| : n. BOTil the Div i "A" and Div !! "B" solenoid control switches.
| |
| : b. Div i "A" solenoid control switch only,
| |
| : c. Div 11 "B" solenoid control switch caly,
| |
| : d. Control P.ocm control switches are inoperable.
| |
| ANSWER:
| |
| : d. Control Room control switches are inoperable.
| |
| IDNO: LP# OBJ #
| |
| 251 IILO-066 6 PROCEDURE NUMBER: OTHER:
| |
| AOP 0031 LEVEL 2
| |
| % _ _ J NRC KA: I RO: SRO:. . .
| |
| 295016 AK3.03 1 3.5 3.7 _
| |
| COMMENTih 7/97 new hO T1 G2 SFtO T1 G1 g
| |
| 20 f
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. O Exam Title NRC RO EXAM 21 The Control Roon' is v, inhabitable and the Remote Shutdown Panels are being utilized to control the plant. Reactor leves is 20 inches and lowering, and Reactor Pressure is 500 psig.
| |
| With present plant conditions, which of the follov ing systems can be utilized to raise reactor level from the Remote Shutdown Panels?
| |
| )
| |
| : a. LPCS
| |
| : a. RilR A c, RCIC
| |
| : d. IIPCS ANSWER:
| |
| : c. RCIC
| |
| . IDNO: LP# OILI#
| |
| 280 llLO-066 2 PROCEDURE NUMBER: OTHER:
| |
| AOP-003) N L2
| |
| {295016 AA1.09 i 4! 4_.
| |
| COMMENTS: 7/97 ,
| |
| new E
| |
| e 21 m
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. 0 Esam Title NRC RO EXAM 22 . With the reactor at 100*A power, a loss of all Reactor Plant Component Cooling Water ocmts.
| |
| What are the required operator actions? ,
| |
| : n. Monitor and reduce system heat loads as necessary to continue plant operations.
| |
| : b. Insert a reactor scram and shin both recirculation pumps to slow speed.
| |
| : c. Comraence a reactor shutdown per GOP-0002, Plant Shutdown.
| |
| : d. Insert a reactor scram and trip and isolate both recirculation pumps.
| |
| ANSWER:
| |
| : d. Insert a reactor scram and trip and isolate both recirculation pumps.
| |
| IDNO: LP# OBJ #
| |
| 241 1ILO-530 6 PROCEDURE NUMBER: OTHER:
| |
| AOP-0011 LEVEL 2 NRC KA: l RO: I SRO:
| |
| 295018 AK2.02 3.41 3.'> 'j 295018 AA1.02 3.3! 3.4 _
| |
| COMMENTS: 7/97 new ROT 1G2 SRO T1 G2 c ,
| |
| t 22
| |
| | |
| RO EXAM KEY Enam Number NRC-RO Rev. 0 Exam Title NRC RO EXAM 23 'Ihe plant is conducting a startup with reactor power at 22%, when an unisolable rupture in the Turbine Plant Component Cooling Water (TPCCW) suction header causes a complete loss of TPCCW. What are the required immediate operator actions?
| |
| : a. Conduct an orderly reactor shutdown per GOP-0002, Plant Shutdowm.
| |
| : b. Initiate RCIC and shutdown the Feedwater pumps,
| |
| : c. Reduce reactor power to within Bypass Valve capacity then trip the Ma:n Turbine,
| |
| : d. Manually scram the reactor.
| |
| ANSWER:
| |
| : d. Manually scram the reacwr.
| |
| IDNO: LP# OBJ #
| |
| 38 IILO-53 6 5 PROCEDURE NUMBER: OTHER:
| |
| AOP412 LEVEL 2 NRC KA: I RO: SRO:
| |
| 29501B AK2.02 3.4 3.0 ,j G 2.4.49 4 4 _
| |
| COMMENTS: 7/97 new RO T1 G2 SRO TI G2 23 w
| |
| | |
| I RO EXAM KEY
| |
| ! Rey, O Esam Title NRC RO EXAM Esam Number NRC-RO M Range 7 and reactor pressure is 450 psig A plant starup is in progress. Reactor power is on IR" Cha pump will not st 24 when the "A" CRD pump trips. The"B A reactor ceramis required if:
| |
| l t r fault and cannot be inserted,
| |
| : a. A control rod receivesi eceived an llCU accumu a o
| |
| : b. More than one CRD high temperature alarm s r 5 mmutes.
| |
| c, No CRD pumps can H restarted within
| |
| : d. Two or more accc nulator faults exist.
| |
| ANSWER:
| |
| fl nd cannot be inserted.
| |
| : a. A control rod receives an llCU accumulator au t a Olki #
| |
| LP #
| |
| IDNO: 11 OTHER:
| |
| STM 032 368 PROCEDURE NUMBER: LEVEL 4 AOP-0011 TS 3.1.5 d ,j ARP 60122 A01 IT/97 new COMMENTS: ROT 1G2 h r difforent references.
| |
| SRO T1 G2 NOTE:Datficutt question, need to pull toget e I-24
| |
| ~~~ ' ~~
| |
| | |
| RO EXA.M KEY Exam Number NRC-RO Rev. O Eram Title NRC RO EXAM 24 A plant starup is :n progress. Reactor power is on IRM Range 7 and reactor pressur e is 450 psig when the "A" CRD pump trips. The "B" CRD pump will not start.
| |
| A reactor scram is required if:
| |
| : a. A control rod receives an llCU accumulator fault and cannot be inserted.
| |
| : b. More than one CRD high temperature alarm is received
| |
| : c. No CRD pumps can be restarted within 5 minutes.
| |
| : d. Two or more accumulator faults nist. .
| |
| ANSWER:
| |
| f ---
| |
| a.
| |
| IDNO:
| |
| A control rod receives an 11CU accumulator fault an2 cannot be inserted.
| |
| LP# OBJ #
| |
| _ 368 STM-057 11 PROCEDURE NUMBER: OTHER:
| |
| AOP-0011 LEVEL 4 TS 3.1.5 d ARP 60122-A01 ,j I NRC KA: } RO: l SRO:
| |
| 1295022 AK3.01 ! 3.71 3.9 _
| |
| COMMENTS: 7/97 new RO T1 G2 SRO T1 G2
| |
| * NOTE: Difficult question, need to pull together different references.
| |
| 1 .
| |
| 0 24 L --- _____-______ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
| |
| | |
| l i-I.
| |
| RO EXAM KEY Enam Numoer NRC-RO Rev. 0 Eram Title NRC RO EXAM 25 Following a small break LOCA, indicated wide range reactor level is 20" and slowly increasing due to RCIC injection from the CST. Other plant parameters are as follows:
| |
| RPV pressure 550 psig Suppression poolump 140 deg. F Contair ment pressure 4.0psig ,
| |
| The MINIMUM suppression pool 'evel which will assure adequate heat capacity is:
| |
| a.15.4 ft b.16.3 ft c.18.3 ft d.19.3 ft ANSWER:
| |
| b.16.3 fl IDNO: LP# OllJ #
| |
| 136 IILO-514 8 al PROCEDURE NUMBER: OTHER:
| |
| EOP-0002 LEVEL 3 I' NRC MA: I RO: I SRO:
| |
| I295026 EK2.06 i 3.5I 3.7 r
| |
| COMMENT 3: 7/97 new ROT 1G2 SRO T1 G1 CANDIDATE WILL NEED EOP FIGURES . (HCLL and HCTL) 23
| |
| | |
| f
| |
| ~
| |
| J IIEAT CAPACITY TEMPERATURE LIMIT The IIEAT CAPACITY TEMPERATURE LIMIT is defined to be the highest suppression pool temperature at which initiation of RPV depressurization will not result m exceeding either (1) the containment design temperature or (2) the PRIMARY CONTAINMENT PRESSURE LIMIT before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent. This temperature is a function of RPV pressure, and the Limit is utilized to preclude failure of the containment or equipment necessary for the safe shutdown of the plant.
| |
| FIGURE 2 HEAT CAPACITY TEMPERATURE UMlY HCT1.
| |
| 200 .A i T- r 1 1- r i rt
| |
| ~
| |
| 8 * '. 4 N
| |
| iso- ha- H I- J
| |
| ('O --
| |
| + -i 1 4 - Fr-1 Tr t - l -+t--4 170 i rt3 i - -
| |
| 1-1 Tj i.e _:. '
| |
| T 4_ t_ ,p 1._F_T.J
| |
| _ _ . _ a_ _ .t _
| |
| : m. _ 4. 3__r- , 4 _ _ _ g.
| |
| _+J_C +_4_pa 140 - -l
| |
| : m. _ gArg I +- i-.P-r ht
| |
| _. t. . .4 t- q'--H
| |
| [-
| |
| g.._ _-t.C
| |
| *o,'7 r,i r - i 260 ' 460 ' e60 ' suo ' tom '
| |
| i RPV PRESSURE (PS3C)
| |
| The HEAT CAPACITY TEMPERATURE LIMIT curve illustrated above comprises two segmentm A.B and B.C. For RPV pressure at or b: low the MINIMUM RPV FLOODING PRESSURE, the rate of energy transfer from the RPV to the containment (with the MINIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY DEPRESSURIZATION open) is, by defmitton of the MINIMUM RPV FLOODING PRESSURE, within the capacity of the containment vent; thus Segment A.B is ve:tical at the MINIMUM RPV FLOODING PRESSURE, and Point B is defined by this Pressure and the lower of either (1) the containment
| |
| , design temperature (RBS limiting factor) or (2) the conttinment temperature at which containment pressure will be at the PRIMARY CONTAINMENT PRESSURE LIMIT. For higher RPV pressure, suppression pool heatup during RPV depressurization is pioportional to EPSTG'0002 A 1I Revic.on 3
| |
| | |
| r
| |
| . HEAT C PACITY LEVEL Loirr "'' -
| |
| The IIEAT CAPACITY LEVEL L1511T is defined to be the itigher of either (1) two feet above the elevation of the top of the Mark III horizontal vents or (2) the lowest suppr'ession pool water level at which initiation of RPV depressurization will not esult in exceeding the IIEAT CAPACITY TEMPERATURE LIMIT. This water level is a function of the margin to IIEAT CAPACITY TESU i'RATURE LIMIT, and the Level Limit is utilize : conjunction with the Temperature Limit to preclude failure of the containment or equi;,nent necessary for the safe shutdown of the plant and to preclude loss of the pressure suppression function of the containment. g FjGURE 4 HEAT CAPACITY LEVEL LIMIT - HCLL -
| |
| 20 - Ar is .
| |
| 1 J T T'rr ri 1 r rr rr,,,
| |
| -4 .t4t H-i - 4 ' 'l IitiI li 8 J
| |
| ' * - ~i
| |
| ('l - Ti T rrl1 h'a l$h-E@ j_g_lj s' Art ii kIU'* ~-
| |
| 7'I u I INYTl l lYll
| |
| ~ ~
| |
| i'-
| |
| r t 1- t-t-t 1-4 -r r r rr i 1 1 is .-h 4tJ.A .L LLl_aJ.L 4 4 y_l 4 4 t2 - -h, hpI,J, 4 h g f,- H rh t t-r-t ;-i T 77 rri 17 to o a 4
| |
| '!', !'!'!'!'!' Aion s a to is is is ATwe M O cs mu v== u ._Q
| |
| @ ac w su m a.a., coa. m u m ._q e,,c. e -@
| |
| ar e.
| |
| The HEAT CAPACITY LEVEL LIMIT curve illustrated above comprises two segments: A-3 and B-C. For suppression pool water level at the low suppression pool water level LCO, v a margin is required to preclude exceeding the HEAT CAPACITY TEMPERATURE LutIT fo'bwing initiation of RPV depressurizmion; thus Point A is defmed by zero margin and ,the low :;uppression pool water level LCO, and Segment A-B reflects a decreasing HEAT CAPACITY LEVEL LBilT with increating margin. For high margin, two feet above the elevation of the top of the Mark III horizontal vents is limiting; thus Segment B_C reflects a constant HEAT CAPACITY LEVEL LBIIT, specifically two feet above the top of the Mark III horizontal vents.
| |
| EPSTG*0002 A-9 Revision 3
| |
| | |
| l-RO EXAM KEY l Esam Number NRC-RO Rev. O Exam Title NRC RO EXAM 26 Which of the following plant conditions are ALL of the reactor water leve' indicators invalid?
| |
| a RPV pressure 60 psig Drywell temperature EL I45' 200 degrees f
| |
| : b. RPV pressure 90 psig Drywcil temperature EL I45' 300 degrees F
| |
| : c. RpV pressure 100 ps',
| |
| Containment temperature EL 119' 360 degrees F
| |
| : d. RPV pressure 1000 psig Containment temperature EL 119' 180 degrees F ANSWER:
| |
| : e. RpV pressure 100 psig Containment temperature EL i19' 360 degrees F IDNO: LP # OBJ #
| |
| 214 IILO-5 t l 6 PROCEDURE NUMBER: OTHER:
| |
| EOP 0001 LEpL3 Cauten 1 NRC KAt l RO: I 5RO:
| |
| 295028 EK2.03 3.61 3,8 295027 FK1.02 3 3$2 -
| |
| COMMENTS: 7/97 new RO li G2 SRO T1 G1 Caution 1 required T
| |
| 26
| |
| | |
| p --
| |
| 1
| |
| ,; CAUTIONS '
| |
| .c., Rt-t A> M B Rhn h 6 RtACTOR Potet- as a CMNQI b(
| |
| Of1D W M D A,
| |
| g j
| |
| /
| |
| fj CAUTION #1 i 1r y IDEM BON THE FouCwtNO CON 0!TIONS ARE $AT!$FI[0, 4, . .
| |
| ,., ' ,j '{
| |
| HIM THAT INSTWVWENT WAY BC V5ED TO OCTERWINE RPV WATER LEVEL
| |
| : 1. Tkt TEWp NEAR Au INSTRuwtNT RgNhAD
| |
| ,4 IN THE 5AFI ZONE or THE RPWT tricuet il Ir1Cuat il
| |
| . RPV SATURAT10N TEWPERATURC - RPVST l 1 i i i l 1 i i i Ow 550 - - t* -i - 3 "
| |
| .. co"
| |
| . aw' r'' p, r 4.*- w t.
| |
| - 450 -.- i- q.- - .
| |
| r ,-w r s-s
| |
| .- +
| |
| o, c,, :,0 --
| |
| -.a-w,-s _wa_w.
| |
| e a "' " >=- ,- H H T 250 -
| |
| 2M 4 -I-L F+5 'b 7 4 H + F
| |
| ' ! I ' ' '
| |
| l p-t-h I I ''I O 200 ' 450 ' 6bo 850 ' 1000 '1200 RPY PRC3SURE (P$!0)
| |
| : 2. }FOR 1TRUMENT (ACH RCACS OrA80W THC THCFOLL0wtNC W'N!WWW IN0!CATED INSTRUWENTS.
| |
| LEWL Ass 0CIATED wifH fHC HJ.''EST TEWACRATVRC SECTION RO '''"''''''"*'"'"""'""""**'**
| |
| WAX Ow TEMP At CL I4s rt, 400*r
| |
| ~~~
| |
| , CONTA!NWCNT NARROW w1OC 2 TEWPERATVRC 37 fr) 200 - - - - - - - - - - - - - - - - - - - - - -
| |
| WONtf0R AMQ CONTROL 1 . -142 N. - 310 W.
| |
| . 7EACTOR 50w(R , g ,,. , , , , , , , , , , _ _ , , , , , , , , , , , , , , _ , , _ _ , , , ,
| |
| +
| |
| 0 IN. -t59 !N. -310 IN.
| |
| g SW1TCH IN SHVfDowM f
| |
| ''' AOP-000f.
| |
| )RtscRAWPROCEDumf' WAX CfMT TEWp AT CL I'9 FT. 200*F
| |
| ~
| |
| UPSET SHUTDOWN 49 IN. 52 IN.
| |
| "'[.",",'" sm . --------------
| |
| EL 145 Ff
| |
| ..4,. 26 IN. 34 IN.
| |
| N 200 - - - - - - - - - - -- - - - - - - - -
| |
| i -
| |
| 9 IN. 21 IN.
| |
| 'J2
| |
| | |
| tuf =
| |
| t i ., CAUTIONS .
| |
| elOs RE3J1NC
| |
| ?
| |
| m m.y~RCACT,OR 0 e ,m, E. #.
| |
| 9 CAUTION #1
| |
| '- L'
| |
| 'U q, ELHQi 80N mt FOLLOW!NC CONDITIONS ARC $AT!$f!ED.
| |
| ,, M NAT INSTRVWENT WAY BC v5t0 70 DCitRWINC 4 p RPV WATER L[ VEL
| |
| ,'y, 1. THE TEWP NCAR ALL INSTRUWENT R IN THE SAFE ZO4C OF THE RPv$f IFICURE il
| |
| , .g RPV $ATURAT!0N TEWPCRA1VRE - RPV$T l I i 4 4 l l 4 i ow sso. r .e t- A 4 -r 4. J' -l_gC
| |
| ** (*O
| |
| ' " " 300
| |
| : 1. R,5.^.','. Jg - 4. .J L 430 -y L ..q - .
| |
| '" "~1 +7~h b CfWT frap (*O 350 - - .. J L4.J T*1 J Ti~b p.L J.
| |
| m.
| |
| -t--
| |
| s -u+.,-s -h i -._whr .,-s 2x !!!!: !r!!r : ! !! .
| |
| O 200 400 800 800 1000 1200 RPY PRES $URC (P$!G,
| |
| : 2. FOR L'A04 0F THE FOLLOWINC !NSTRUWENT3. THC
| |
| ,b$fRUWENT READS A80W THE WIN!WUW IN0!CAftD JVEL ASSOCIAftD w!m NC HICHC$7 TEMPERAfuRC SECTION R0 "''""''**'****'"""***''*"""'''*
| |
| WAX Ow TEMP AT (L 14s FT. 600*r FUCL C&TAINWENT NARROW w!OC ZONC
| |
| *'[?f," m . .. ---------------.
| |
| EL M9 FT 1 IN. -142 IN. -310 IN.
| |
| retsCTOR PontR Yno - ..------ - - - -.- -----.
| |
| +
| |
| 0 IM. -159 !N. -310 IN WRlFY mt REACTOR WOOC
| |
| .g SWITCH IN SHUfDoubt b
| |
| AOP-0001, \
| |
| '*>RW$CRAWPRoctDUR(/
| |
| WAX CfWT TEWP AT CL,119 FT. 200*F
| |
| [, UP$ET SHufDowN
| |
| ~
| |
| tt IN. 52 !N.
| |
| "'[?l,* a - -------..-----
| |
| .. EL tas FT 4.
| |
| 28 IM. 34 (N.
| |
| 200 - .------..-----
| |
| : 9 !N. 21 IN.
| |
| !J2
| |
| | |
| t RO EXAM KEY Esam Number NRC-RO Rev. 0 Exam Title NRC RO EXAM 27 Which one of the following describes a cause and the expected inaccurate response of reactor level instrumentation indications when in the UNSAFE region of the RPV Saturation Curve?
| |
| : a. liigh cor.tainment temperatures will result in boiling of the reference legs causing an erroneously high level indication.
| |
| : b. liigh reactor pressure will result in boiling of the refe.cnce legs causing an #
| |
| erroneously low level indication.
| |
| : c. Low reactor pressure will result in boiling of the reference legs causing an erroneously low level indication.
| |
| : d. liigh drywell temperatures will result in boiling of the variable legs causing an erroneously low level indication.
| |
| ANSWER: ,
| |
| : a. liigh containment temperatures will result in boiling of the reference legs causing an erroneously high leve' indication.
| |
| IDNO: LP# OBJ #
| |
| 81 IILO-512 5 PROCEDURE NUMBER: OTHER:
| |
| EOP-0001 LEpL 4 EPSTG'0002 NRC KA: I RO II SRO:
| |
| !295028 EK1.01 1 3.Si 3.7 _
| |
| COMMENTS: 7/97 new RO T1 G2 SRO T1 G2 27
| |
| | |
| RO EXAM KEY 2 Esam Number NRC.RO Rey, O Esam Title NRC RO EXAM 28 Suppression pool level is offscale hi;;h.
| |
| Which one of the following describes the efTect on indicated containment or drywell pressure?
| |
| : a. Indicated containment pressure is less than actual.
| |
| : b. Indicated containment pressure is greater than actual.
| |
| : c. Indicated drywell pressure is less than actual,
| |
| : d. Indicated drywell pressure is greater than actual.
| |
| ANSWER:
| |
| : b. Containment pressure is greater than actual.
| |
| IDNO: LP# OBJ #
| |
| 388 IILO-514 5 PROCEDURE NUMBER: OTHER:
| |
| EOP.0001 LEVEL 2 EPSTG'0002 Cauten 8
| |
| .h NRC KA: I RO: SRO:
| |
| 295029 FK1.01 1 3.4 3.7 _
| |
| CCMMENTS: 7/97 new ROT 1G2 SRO T1 G2 28
| |
| | |
| r RO EXAM KEY Esam Number NRC-RO Rev. 0 Exam Title NRC RO EXAM 29 in Emergency Depressurization, Step ED-3 asks,"Is Suppression Pool Level Abov- 13 fl? What is the significance of this level?
| |
| : a. It ensures a vortex will not be created when SRVs are opened.
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containment does not occur.
| |
| : c. It is required to prevent loss of NPS11 to the RilR Pumps.
| |
| : d. It ensures there is enough water to cover the horizontal vents.
| |
| ANSWER:
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containment does not occur.
| |
| IDNO: LP # OILI #
| |
| IILO-512 5 23 PROCEDURE NUMBER: OTHER:
| |
| ECP-0001 LEVEL 2 EOP@01 BASES I NRC KA: l RO: l SRO:
| |
| {295030 EK2.oB l 3.5! 3.8 COMMENTS: 7/97 new ROT 1G2 SRO T1 G1 e
| |
| e 29 l
| |
| l
| |
| | |
| i I
| |
| RO EXAM KEY Esam Number NRC RO Rev. 0 Exam Title NRC RO EXAM 30 Which of the following unisolable system failures (ruptures outside of pnmary containment) would constitute a primary system for purposes of EOP-3, Secondary Containment Control?
| |
| : a. Main steam drain lines in the Main Steam Line Tunnel
| |
| : b. Containment vent line to drywell
| |
| : c. Service Water supply to drywell coolers
| |
| : d. LPCS suction lin- from the suppression pool ANSWER:
| |
| : a. Main steam drain lines in the Main Steam Line Tunnel IDNG: LP# OIkl #
| |
| 312 IILO-Sil 5 PROCEDURE NUMBER- OTHER:
| |
| EOP@3 LEVEL 2 EPSTG'0002 c- -- - gyp - - - -
| |
| 295033 EK3 03 3.81 3.9 _
| |
| COMMENTS: 7/97 new RO T1 G2 SRO T1 G2 30 l
| |
| | |
| l RO EXAM KEY-t Exam Number NRC-RO Rev. O ' Exam Title NRC RO EXAM 31 A high radiation alarm exists on the Annulus ventilation system (RMS'REl10). You are monitoring the CRT bar chart display for RMS*REll A to validate the alarm condition on RMS'REllB. The 10 minute trend data for RMS'REll A is colored " light blue". Which of the {
| |
| following describes the status of RMS*REll A data' readings? j
| |
| : a. RMS'REll A is reading within 1% of RMS*REIID.
| |
| : b. RMS*REll A is in an Alert condition.
| |
| : c. RMS'REll A data is " questionable",
| |
| : d. communication has been lost between RM-80 and RM-23.
| |
| ANSWER:
| |
| : c. RMS'REl1 A data is " questionable".
| |
| IDNO: LP# OBJ #
| |
| 420 IILO-069 7 PROCEDURE NUMBER: OTHER:
| |
| SOP-0086 LEVEL 3
| |
| ['idiTAI I I RO: I SRO: ,j i 1293a)3
| |
| * EA1.01 i 3.8! 3.8 _
| |
| COMMENTS: 7/97 new L OTM-65 4 __
| |
| 31
| |
| | |
| RO EXAM KEY Exam Number NRC RO Rev. 0 Exam Title NRC RO EXAM i
| |
| j EOP 3, Radioactive Release Control, has been entered.
| |
| 32 l 1
| |
| Which of the following is the r:ason that the operator is directed to ensure that the Turbine Building Ventilation fans are running?
| |
| : a. Reduce radioactive releases below General Emergency levels
| |
| : b. Prevent radioactive releases from the Turbine Building.
| |
| i
| |
| : c. Filter radioactivity from the Turbine Building atmosphere.
| |
| : d. Provide a monitored release point.
| |
| ANSWER:
| |
| : d. Provide a monitored release point IDNO: LP# OBJ #
| |
| 311 IILO-515 4 PROCEDURE NUMBER: OTHER:
| |
| EOP-3 RR-1 LEVEL 2
| |
| ,)
| |
| ~
| |
| NRC KA: l 30:~l SRO:
| |
| 295038 EA1.06 3.51 3.6 _
| |
| COMMENTS: 7/97 fieW l .
| |
| R
| |
| --- .w. ,-
| |
| | |
| l RO EXAM KEY Esam Number NRC RO Rey, 0, Exam Title NRC RO EXAM 33 Following a complete loss of shutdown cooling, temperature readings indicate a i degree F increase in bulk water temperature every 10 minutes. Assume the reactor vessel head is on, no other parameters change, and current temperature is 124 deg. F.
| |
| Which of the following is the minimum amount of time before primary containruent MUST be established?
| |
| : a. 160 minutes
| |
| : b. 560 minutes
| |
| : c. 580 minutes
| |
| : d. 760 minutes ANSWER:
| |
| : d. 760 minutes IDNO: LP# OILI #
| |
| 369 IILO-013 9 PROCEDURE NUMBER: OTHER:
| |
| TS 3.6.1.2 LE,yEL 2 NRC KA: T E0[l SRO:
| |
| 295021 AA7 01 1 3.51 3.6 COMMENTS: 7/97 new T.S. 3.6.1.2 and Table 1.2, Operational Conditions RO T1 G2 SRO T1 G2 e
| |
| 33 ,
| |
| | |
| RO EXAM KEY Eram Number NRC-RO Rev. O Esam Title NRC RO EXAM 34 During refueling, the leakage rate of the Refueling Cavity has exceeded the capacity of the Drywell and Containment Equipment and Floor Drain sumps. A fuel buncle is NOT in a safe storage location.
| |
| Which one of the following systems should be used for emergency makeup to the Befueling Cavity?
| |
| : a. Control Rod Drive flydraulics
| |
| : b. Condensate
| |
| : c. Reactor Water Cleanup
| |
| : d. CNS service connection ANSWER:
| |
| : b. Condensate IDNO: LP# Olkl #
| |
| 380 llLO 535 8 PROCEDURE NUMBER: OTHER:
| |
| AOP-0027 LEyEL 2 295023 AA2.04 I 3.4 3.7 _j COMMENTS: 7/97 new AOP 0027, p. 5.
| |
| ROT 1G2 SRO T1 G1 34
| |
| | |
| . . . _. .. . - ~ _ _ . _ . -
| |
| RO EXAM KEY Exam Number NRC-RO Rev. O Esam Title NRC RO EXAM 35 Given the following conditions:
| |
| - A failure to scram has occurred.
| |
| - Reactor power is 20% with control rods being inserted manually. l
| |
| - EOP-3, " Secondary Containment Control" has been entered due to ilVAC cooler l high differential temperatures caused by a fire in the Auxiliary Building.
| |
| - MSIVs have closed.
| |
| - Condensate /feedwater is maintaining water level.
| |
| Which of'ac following systems should be isolated?
| |
| : a. Feedwater
| |
| : b. Reactor Water Cleanup
| |
| : c. Control Rod Drive
| |
| : d. Fire Suppression systems f ANSWER:
| |
| : b. Reactor Water Cleanup IDNO: LP# OBJ # j 376 IILO-SIS 4 PROCEDURE NUMBER: OTHER:
| |
| EOP-0003 SC-12 LEVEL 3 NRC KA; I RO: ~ l SRO:
| |
| 295032 EA1.05 3.71 3.9 _
| |
| COMMENTS: 7/97 new HLO-514 obi 4 & 6 i
| |
| e 35
| |
| | |
| RO EXAM KEY Exam Number NRC RO Rev. 0 Esam Title NRC RO EXAM 36 EOP 3. Secondary Containment and Radioactivity Release Control, must be entered if the Secondary Conta'.nment differential pressure is above the maximum normal operating differential pressure.
| |
| Which one of the following is the reason for this entry condition?
| |
| : a. A significant steam leak into the secondary containment is indicated.
| |
| : b. A significant water leak from primary system may be discharging radioactivity directly to the secondary containment.
| |
| : c. A potential for the loss of secondary containment is indicated that could result in uncontrolled radioactive releases,
| |
| : d. An increase in the unmonitored ground level radioactive releases due to leakage through secondary containment is indicated.
| |
| ANSWER:
| |
| : c. A potential for the loss of secondary containment is indict.ted that could result in uncontrolled radioactive releases.
| |
| A IDNO: LP# OBJ #
| |
| 377 IILO-515 5 J
| |
| PROCEDURE NUkSER: OTHER:
| |
| EPSTG'0002 LEVEL 2 NRC KAt I RO: l SRO:
| |
| 295035 [K1.01 1 3.91 4.2 3
| |
| COMMENTS: 7/97 new EPSTG'00021, Appenix B, p. *.52 OF 269 ROT 1G2 SRO T1 G2 O
| |
| 36
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rey, 0, Exam Title NRC RO EXAM 37 Which one of the following would be indicative of a loss of air to the on line Control Rod Drive -
| |
| (CRD) Dow control valve?
| |
| : a. CRD cooling water high dp.
| |
| : b. CRD high temperature.
| |
| : c. IICU accumulator low pressure,
| |
| : d. Suction filter high dp.
| |
| ANSWER:
| |
| b, CRD high temperature.
| |
| IDNO: LP# OBJ #
| |
| 42 STM-052 9a PROCEDURE NUMBER: OTHER:
| |
| AOP-0008 LEVEL 2
| |
| ' NRC KA: I RO: I~SRO:
| |
| 201001 K6.03 1 3I 2.9 4
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G2
| |
| * O 37
| |
| - . - . - - - , . - ,,r- .
| |
| | |
| F -
| |
| O R
| |
| 3-I
| |
| 'tjg .
| |
| - < -: ~~
| |
| ,g, n -,
| |
| _g_g, :
| |
| :: a _ (Fc)
| |
| - "3 ,, , I m :
| |
| + :,. O-oo- 7 S P=H I o
| |
| ,2 rev
| |
| _f
| |
| -*< ;v O-oo- = =
| |
| '63 X x
| |
| : = :: =i =
| |
| N ?4 ;V ?O /.
| |
| i r essoumet FILig33 l q ." . ,5 2B = RECIRC
| |
| == neo me ._
| |
| ll CET I PAL }-
| |
| -- Auto / ! 251RamspsT ACGJM.
| |
| A
| |
| - cmanna
| |
| \'
| |
| CDOLJNO "t^
| |
| * I
| |
| ~
| |
| AED :"* ,= = _
| |
| AP i
| |
| /.
| |
| ee n trzneG VEWS coaums 1, 0-358 P53D
| |
| {
| |
| WATUt [A W I M(
| |
| P
| |
| ~~~*
| |
| L
| |
| ^
| |
| T m,c% 1 y
| |
| es ream -
| |
| saum DEXT -* ,
| |
| = w w w :: .
| |
| M ,
| |
| ~
| |
| P w
| |
| P~
| |
| w w "~ @
| |
| . - = =
| |
| L 1, "
| |
| d
| |
| .s===== ori..9 = i "' m J43
| |
| . 1 Yws 8'8"2T
| |
| "'" contasmeer st v
| |
| q
| |
| .o
| |
| < , 2Fe = ==l-" o m c= or n h b (
| |
| ce --
| |
| E. _ . . . _
| |
| " ==w -*
| |
| . CONTROL ROD HYDRAULIC '.~ SYSTEM
| |
| | |
| l 1
| |
| .) i
| |
| - D. Flow Control Station (Figures 1,2,3)
| |
| {
| |
| The Flow Control Station is an. electro-pneumatic: flow control network -which maintains uenstant flow through the CRD Hydraulic System regardless of variations in reactor system pressure.
| |
| Flow element Cl1-FE N003 senses system flow as a differential pressure across a flow venturi. This pressure signal is converted to an electrical signal by flow transmitter Cll-FT-N004 and sent to local flow indicator Cll-FI-R019, remote flow indicator
| |
| . Cl1-FI-R606 (H13 P601) Figure 5 flow controller Cl1 FC-R600 (H13-P601), and to .
| |
| the process computer.
| |
| Flow indicator Cl1 FI R606 does not indicate total CRD pump discharge flow, only drive water and cooling water flow. Charging water flow will also contribute to the t'otalindicated flow while tife accumulators are being charged. CRD pump minimum flow, recirculation pump seal purge flow, reactor water cleanup pump seal purge flow, i
| |
| reference leg backfill system flow, and the sample flow (0.5 gpm) are not indicated since they come off the pump discharge upstream of the flow element. Due to the _
| |
| range of Cl1-FI-R606,0-100 gpm, it will read off scale whenever the accumulators are charging. The high flow will close the Flow Control Valves resulting in all the CRD~ '
| |
| water being pumped to the accumulators. .
| |
| i
| |
| - Flow controller Cll-FC-R600 provides the operator the means for manual and
| |
| ' automatic flow control. In the manual mode, the control signal from R600 is changed by means of the open and close push buttons. The push buttons have double detents to allow for slow or fast output change. Pressing the buttors to the first detent yields a slow response, 50 seconds for 100% output change. 'Wessing the buttons to the second detent yields a fast response, 5 seconds for 100% output change- .
| |
| The current signal from R600 goes to the local current / pneumatic converter, K001.
| |
| K001 cotverts the 4-20 MA DC current signal to a 6-30 psig air signal, using air from the 30 psig instrument air supply, through needle valve F039.- The air signal from K001 goes to the local M/A station, D009A/B, through normally open gate valve F125A/B.
| |
| D009A/B each have a three way valve, a pressure indicator used during R600/K001 operation, a pressure indicator used durmg local manual operation, and a Pressure Control . Valve (PCV) used for local manual operations. The three-way valve is positioned to either receive the controlling air signal from R600/K001, or to receive the
| |
| . air supply through the manual loading PCV. The PCVs of both M/A stations _ are normally closed, but their supply gate valves, F126A/B, and the common needle valve F038 are normally open.
| |
| From the'three-way valve on the M/A station corresponding to ,the on line FCV, the air
| |
| - signal goes to the valve positioner of the FCV. The valve positioner uses 75-80 'psig instrument air as the control air to position the FCV. The 75-80 psig air goes to the FCV valve positioners through normally open gate valves F127A/B.
| |
| ; LOTM-5-6 PAGE 5 OF 33
| |
| _ _-- - _ _ _ . _ _ _...u._. -~-
| |
| | |
| Due to the location of the charging water line with respect to flow element Cll-FE-N003, the flow control valve will close during a scram or accumulator charging to divert system flow to the charging water header (See Figure 1).
| |
| E. Pressure Control Station (Figure 1)
| |
| The ability to accurately position the control rods is imponant to reactor control and safety. For precise control, the speed of control rod movemer wt be constant. In a hydraulic system, such as the CRD Hydraulic System, the differei.:d pressure across the working surface of the CRDM must be constant. The purpose of the pressure control station is to maintain system pressure at a constant pressure above reactor pressure regardless of reactor pressure variations.
| |
| In a system with a constant resistance to flow, if the flow rate through the system is maintained constant, the pressure drop across the system will remain constant.
| |
| The Flow Control Station and the Pressure Control Station work in conjunction with each other usini; this principle to maintain a constant drive water differential pressure.
| |
| The Flow Control Station, as previously discussed, maintains the desired constant flow while the Pressure Control Station maintains the desired resistance to flow.
| |
| i The Pressure Control Station consists of the pressure comrol valve, C11-MOV-F003)[
| |
| a local manual bypass valve (Cl1-HCV-F004), two manual isolation valves, and f'our sets of stabilizing valves (C11-FCV-F007A,B,C, and D).
| |
| F. Stabilizing Valves (Figure 1)
| |
| The Flow Control Station delivers a constant flow of wJter to the Pressure Control Station and all of this water normally flows to the cooling water header. The stabilizing valves nonnally diven a total of 16 gpm around the pressure control salve to the cooling water header. However, duing control rod movement, when there is a demand for water flow through the drive water header, one (for withdraw) or two (for insert) of the stabilizing valves close in order to supply the drive water header with the flow necessary to effect CRDM movement. The flow through F003 to the cooling water header remains unchanged. Therefore, the pressure differential across valve F003 remains unchanged and the drive water header / reactor vessel differential pressure remains unchanged at 250 psid.
| |
| Each set of stabilizing valves consists of two normally open solenoid valves, each passing 2 gpm around Cll-MOV-F003 to the cooling water header. The purpose of the stabilizing valves is to maintain drive water header pressure during constant control rod movement. Control of the stabilizing vaWes is from the Rod Control and Information System (RCIS).
| |
| LOTM-5-6 '
| |
| PAGE 6 0F 33
| |
| | |
| ARI System Testing (Figure 16) !
| |
| -1
| |
| ' The ARI system is functionally divided into two separate channels for testing purposes.- 3 Channel A.(Outboard)is associated with ARI valves 162B,162D, and 164A.. Char M ,
| |
| B (Inboard) .is - associated with . ARI valves 162A, 162C,160, and ' 164B.' : is -
| |
| arrangement is provided to allow actually energizing and changing positions of the ARI valves without depressurizing the scram air header.
| |
| A channel test selector switch is provided to ensure that both channels cannot be tested simultaneously.' Once the channel to be tested is selected, a common test pushbutton is depressed to simulate an initiation signal to the selected channel.
| |
| Testing the B channel is different than testing the A channel in that the F160 valve blocks offInstrument Air to the entire scram air header. To prevent depressuring the header during this test, F164B opens to maintain pressure in the header.
| |
| Either a test or an initiation of the A channel prevents F164B from opening. l An ARI IN TEST annunciator comes in when either channel is selected for test. An ARI INITIATED annunciator comes in when the TEST pushbutton is depressed regardless of which ARI channel is being tested. , ,
| |
| i ARI Power Supplies The ARI system is diverse, physically separated, and electricallyain p= lent from the RPS. Diversity is achieved through the use of energize-to-trip circuits and DC versus
| |
| ; AC power supplies. DC power is supplied frem BYS-PNLO2A2.
| |
| J Use of non-divisional, non-interruptible DC power precludes the need for isolation devices and provides for ARI capability durmg station blackout conditions.
| |
| The system will auto-initiate upon receipt of RPV level 2 or RPV pressure of 1127 psig. These are the same sensors and setpoints utili=1 for the ATWS trips of the recirculation pumps, Manual initiation is effected by simultaneous depression of two arm-and-depress push buttons on panel P680.
| |
| Ils INSTRUMENTATION Refer to Tables 1 through 3 (Figures 14 and 15)
| |
| . A. CRD Temperature The only temperature that is monitored is the CRDM temperature. A chromel-alumel
| |
| - thermocouple at the top of the position indication probe monitors the effectiveness of CRD system cooling water flow to the CRDMs. All 145 thermocouples feed into temperature recorder Cll-TR-R018 on local panel P007 (Aux Building 141' elev.).
| |
| Typicallp temperatures mn in the range of 220'F, with a high temperature alarm of 250*F, which annunciates on H13'-P6ft0.
| |
| B; CRD Drive Water Flow LOTM-5-6 PAGE 16 OF 33
| |
| | |
| RO EXAM KEY -
| |
| Eram Number NRC-RO Rev. 0 ' Exam Title NRC RO EXAM l
| |
| 3g Which one of the following describes the operation of the CRDM Individual Rod Scram Test switches when placed in the TEST position?
| |
| : a. Either test switch deenergius a single test solenoid to allow the air to vent from the scram inlet and outlet valves.
| |
| : b. One test switch deenergizes RPS A pilot solenoid and the other test switch i deenergizes the RPS B pilot solenoid to allow the air to tent from the pilot valve. j
| |
| : c. One test switch deenergizes a test solenoid to allow the scram inlet valve to vent and the other test switch deenergizes a test solenoid to allow the scram outlet valve to vent.
| |
| : d. Either test switch deenergizes both RPS A and leS B pilot solenoids to allow the air to vent from the scram inlet and outlet valves.
| |
| ANSWER:
| |
| : b. One test switch deenergizes RPS A pilot solenoid and the other test switch deenergizes the RPS B pilot solenoid to allow the air to vent from the pilot valve.
| |
| IDNO: LP# 011J #
| |
| 107 IILO-516 13 PROCEDURE NUMBER: ,hTHER:
| |
| EOP-0005 ENCL 13 LEVF.L 2 NRC KA l ~RO: SRO:
| |
| 201001 K4.05 3.8 3.8 l 212000 K4.10 3.3 3.6 -
| |
| l l COMMENTS: 7/97 new RO T2 01 SRO T2 02 I
| |
| l 38 l
| |
| i
| |
| | |
| period of greater than 30 seconds is indicated. Contact the Shift Superintendent prior to rewithdrawing the inserted control rod (s).
| |
| (3) A Reactor Engineer will be used to assist with control rod movements when operating with a limiting control rod pattern or to approve of any unplanned deviations from the rod sequence.
| |
| (4) If an out of sequence rod is withdrawn suspend all control rod motion, notify the Shift Superintendent and refer to Tech Spec 4.1.4.2.a.1.
| |
| : d. The individual rod scram TEST switches are located adjacent to the Hydraulic Control Unit (HCU) for each control rod in the Contairanent. Taking a TEST Switch to the TEST position deenergizes the associated rod scram pilot solenoid.
| |
| There are two TEST switches for each rod, TEST switch A and TEST switch B.
| |
| Both switches must be placed in TEST to effect a scram of the associated rod.
| |
| When both TEST switches are placed in the TEST posiuon, the associated rod scram pilot solenoid valves will open, bleeding air from the actuators of the associated scram valves. The scram valves will then open, effecting a scram of the associated rod.
| |
| The individual rod scram TEST switches are only allowed to be used in an attempt to scram the rod during an Anticipated Transient Without Scram (ATWS) condition or during individual rod scram time testing.
| |
| V. REFERENCES
| |
| : 1. River Bend Final Safety Analysis Report J
| |
| : 2. GEK 83350, Volume III, Part 4, Rod Control and Information System
| |
| : 3. ARP 680-07,
| |
| : 4. SOER-84-2, Control Rod Mispositioning ,
| |
| : 5. SER 13-86, Control Rod Misoperation, Peach Bottom Unit 3
| |
| : 6. RBS Technical Specifications e
| |
| o e
| |
| O LOTM 6 - 6 PAGE 26 OF 31
| |
| | |
| ENCLOSURE 13 OPENING INDIVIDUAL SCRAM TEST SWITCHES 1.0 EIJEEQSE To provide instructions forindividualb ramming control rods with the scram test switches at the respective HCU.
| |
| 2.0 REOUIRFn TOOLS-EOUIPMENT 2.1 NONE ..,
| |
| e 3.0 INSTRUCTIONS 3.1 IE necessary, THEN DEFEAT RPS M ARI logic trips per EOP-0005 ENCLOSURE 12. []
| |
| 3.2 RESET the reactor SCRAM (IH13*P680) []
| |
| 6 3.3 At the respective HCU, PLACE the A M B scram test switches to TEST.
| |
| (Rx Bldg EL 114 ft.) []
| |
| 3.4 WHEN control rod motion stops, J THEN RETURN the scram test switches to NORM. (Rx Bldg EL i14 ft.) []
| |
| | |
| ==4.0 REFERENCES==
| |
| | |
| 4.1 O I Elem Diag 762E429AA, Sh 6, Control Rod D;ive Hyd Sycia .
| |
| 4
| |
| ^
| |
| PAGE I OF 2 REV.9 PAGE 41 CF 115 ENCLOSURE 13 l EOP-0005 1
| |
| | |
| RO EXAM KEY Exam Number NRC RO Rev. O Exam Title NRC RO EXAM ;
| |
| I 39 A reactor startup is in progress and reactor pressure is 800 psig. A loss of both CRD pumps has resulted in the receipt of the CRD ACCUMULATOR TROUl1LE alarm. The nitrogen pmssure on one of the CRD llCUs indicates 400 psig. Which one of the following desc.ibes the efrect of this condition on the CRDM when a scram is initiated?
| |
| : a. Accumulator pressure alone will drive the rod in.
| |
| : b. Reactor pressure alone will drive the rod in,
| |
| : c. Both reactor pressure and accumulator pressure must be combined to drive the rod in,
| |
| : d. . Both reactor pressure and accumulator pressure combined are inadequate to drive the rod in. 3 ANSWER:
| |
| : b. Reactor pressure alone will drive the rod in.
| |
| IDNO: LP# ORJ #
| |
| 122 IILO-003 5 PROCEDURE NUMBER: OTHER:
| |
| TS 3.1.5 A1&2 L%fEL 3 l TS 3.1.5 C1&2 NRC KA: iOT SRO:
| |
| 201002 K4.06 3.5 3.5 201003 K1,01 2.9 3
| |
| ' COMMENTS: 7/97 new RO T2 G2 SRO T2 G3 e
| |
| 39
| |
| | |
| _ _ _ _ ._, _ . _ _ . _ _ _ _ _ _ . _ ... _ . . ____ ___ ~ _ _. . . . . . . . . .
| |
| l l
| |
| RO EXAM KEY Exam Number NRC.RO Rev. 0 Exam Title NRC RO EXAM 40 if one of the "B" Reactor Recirculation System Flow Converter fails (resulting in zero output) with the reactor operating at 100% power, which one of the following describes what will be gent.ated in APRM Channel B7
| |
| : n. A Downscale Alarm and a Rod Block,
| |
| : b. A Rod Block only.
| |
| : c. A flalf Scram signal only,
| |
| : d. A Rod Block and a llalf Suam signal.
| |
| ANSWER:
| |
| : d. A Rod Black and a IIalf Scram signal.
| |
| IDNO: LP# OHJ #
| |
| 13 IILO461 7 PROCEDURE NUMBER: OTHER:
| |
| ~/)P 0074 ATT e LEVEL 3 NRC KA: } HO: l SRO:
| |
| 201005 K5.05 3.6! 3.6 'j l 215005i5$05 ~~3.61 3.6 CCMMENTS: 7/97 new (Flow biased setpoint decreases, and avera0e thermal power exceeds it.)
| |
| RO T2 G1 SRO T2 G1 i
| |
| i 4
| |
| 40
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. 0 Exam Title NRC RO EXAM 41 The plant was initially operating at 100% power, A transient occurred resulting in the following conditions:
| |
| - RPV levelis 55 inches and stable
| |
| . Reactor power is D% ar.d stable
| |
| . Total core flow is 51.5 E6 lbm/hr. and stable The cause of this plant configuration was c.he receipt of a signal from the:
| |
| : a. EOC-RPT logic.
| |
| : b. ATWS/ARIlogic.
| |
| : c. recirculation pump cavitation interlock c!rcuitry.
| |
| : d. recirculation flow control valve runback logic.
| |
| ANSWER:
| |
| : d. recirculation flow control valve runback logic.
| |
| l IDNO: LP# OBJ #
| |
| A19 STM-053 2c j PROCEDURE NUMBER: OTHER:
| |
| AOP4024 LEVEL 3 NRC KA! l ~ RO: I SRO-20200;LA.2 AL 3.4 3.4 295001 AK2.02 32 ,,,3.3 COMMENTS: 1/97 exam (modified values in stem & te-ordered answers) i l
| |
| 4E
| |
| | |
| .- . = . - . .
| |
| -. . - _ - . -.. . ~ --
| |
| RO EXAM KEY Exara Tiember NRC-RO Rey, O Exam Title NRC RO EXAM 42 During valve time testing on RilR System A,IE12'MOVF004A. RilR pump A Suppression Pool Suction Valve, is closed with all ether vahevswitches in their normal standby position when a valid l LOCA cignal occurs. In this condition, RilR pump A breaker will: l
| |
| : a. Close and immediately trip because of the IE12*MOVF004A contacts in the bicaker trip circuit.
| |
| : b. Not close because of the IE12* MOVF004 A contacts in the breaker close permissive circuit.
| |
| : c. Close after IE12'MOVF004A opens automatically,
| |
| : d. Close and remain closed, while IE12'MOVF004A remains closed.
| |
| ANSWER:
| |
| : a. Close and immediately trip because of the IE12'MOVF004 A contacts in the breaker trip circuit.
| |
| IDNO: LP# OBJ #
| |
| 140 IILO-021 6 PROCEDURE NUMBER: OTHER:
| |
| 82SE5%AA (sys204) LEVEL 4 NRC KA: RO: ' SRO:
| |
| 2_7000 K4.06 3.5 3.6 _
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1
| |
| )
| |
| AI e
| |
| 42
| |
| | |
| RO EXAM KEY
| |
| - Esam Number NRC RO Rev. O_ Esam Title NRC RO EXAM 43 The following plant conditions exist:
| |
| - The reactor is in cold shutdown.
| |
| - RilR "A" is in shutdown cooling.
| |
| ENS'SWG IB is deenergized for maintenance.
| |
| A RPV water level transient occurs resulting in RPV water level lowering to 120" Which of the following actions will result in LPCI "A" injecting into the RPV7
| |
| : a. Close the SDC suction valve F006A, open suction valve F004A from the suppression pool and restart the RilR A pump.
| |
| : b. Close the SDC suction valve F008, open suction valve F004A from the suppression pool, manually open F027A and F042A, and restart the RHR A pump.
| |
| : c. Close the SDC suction valve F006A, then arm and depress Div i LPCI initiation pushbutton.
| |
| : d. Close the SDC suction valve F006A, open the suction valve F004A from the suppression pool, then arm and depress Div I LPCI initiation pushbutton, ANSWER:
| |
| : d. Close the SDC suction valve F006A, open the suction valve F004A from the i suppression pool, then arm and depress Div I LPCI initiation 'pushbugton.
| |
| I IDNO: LP# OBJ #
| |
| 278 IILO-021 9 PROCEDURE NUMBER: OTHER:
| |
| SOP 4031 LEVEL 3 NRC KA: I' RO: I SRO:
| |
| 203000 A4.oS } 4.3l 4.1 COMMENTS: 7/97 new RO T2 G1 SRO T2 G1
| |
| +
| |
| 43 l
| |
| | |
| _-. _ .. . . - . - - .. -- . - . - . . -, ~- -
| |
| * 480VAC System per SOP-0047. >
| |
| . 120VAC System per SOP-0048. ,
| |
| . 125VDC System per SOP-0049.
| |
| 3.3 Instmment Air system operable per SOP-0022.
| |
| 3.4 Condenaate Storage, Makeup, and Transfer is operable per SOP-0008.
| |
| 3.5 Respective unit cooler (s) operable.
| |
| 4.0 SYSTEM STARIUP 4.1 Placing RHR in LPCI Standby Mode ;
| |
| i 4.1.1 P rform the Valve Lineup per Attachment IA, B, or C as applicable.
| |
| 4.1.2 Perform the Instmment and Valve Lineup per Attachment 2.
| |
| 4.1.3 Perform the Electrical Lineup per Attachment 3.
| |
| 4.1.4 Perform the Control Board Lineup per Attachment 4A, B or C as applicable.
| |
| 4.1.5 Fill and vent the desired loop as follows'(Ref. 7.42):
| |
| : 1. Rack out the associated RHR pump breaker for loop to be vented.
| |
| : a. LENS *SWGIA ACB03 RHR Pung A
| |
| : b. LENS *SWOlB ACB23 RHR Pump B
| |
| : c. 1 ENS *SWGlB ACB28 RHR Pump C
| |
| : 2. To fill and vent RHR A(B) loop, perfonn the following steps:
| |
| gar
| |
| : a. fy 1E12*F004A(B) RHR PUMP A(B) SUP PL
| |
| : b. IE12*VF012A(B) and IE12*VF013A(B) RHR A(B) PUMP S ON.VEN'IE. Close when air is vented off,
| |
| : c. IE12*VF015A(B) and IE12*VF016A(B) RHR A(B) PUMP S VI.NIS. Close when air is vented off.
| |
| : d. O f .m or veri lE12*VF085 ) LPCS FILL PUMP STOP CriECK TO A DISCH (DI FILL PUMP STOP CHECK IO RHR B DISCH). ,
| |
| : c. Start or verify running IE21*C002(IE12*C003) LPCS/RHR DIV 1 i
| |
| LINE FILL PUMP (RHR DIV 2 UNE FILL PUMP).
| |
| SGWQ31 REV - 17B PAM 9 OF 107
| |
| | |
| < f. Open IE12*VF020 SHUTDOWN COOLING SIK310N FILL.
| |
| : g. Verify closed IE12*F009 RHR SHUTDOWN COOLING INBD ISOL VALVE and open IE12*F008 RHR SHUIDOWN COOLING OUTBDISOL VALVE.
| |
| h.' Op.en IRHS*V155 SHU1DOWN COOLING INLET VENT LMC CONN and close when all air is vented off,
| |
| : i. Close IE12*F008 RHR SHUIDOWN COOLING OUTBD ISOL VALVE and IE12*VF020 SHUTDOWN COOLING SUCTION FILL.
| |
| J. Oxn 1RHS*V162 RHR A FUEL POOL COOLING SI'C110N VENT VALVE and close when all air is vented off.
| |
| : k. Unlock and open IE12*VF063A(B) SHUIDOWN COOLING A(B)
| |
| RETURN LINE FILL.
| |
| : 1. Open IE12*F042A(B) RHR PUMP A(B) LPCI INJECT ISOL VALVE.
| |
| : m. If ne reduce injection line pressure by operu'ng IRHS?V376 A IN ON LINE HIGH POINT VENT INSIDE JRYWELL (AZ 15) or 1RHS*V375 RHR B INJECHON LINE VENT. !
| |
| : n. Open IRHS*V376 A INJECHON LINE HIGH POINT VENT INSIDE DRYWELL (AZ 15) or 1RHS*V375'RHR B INJECHON LINE VENT, then close wb:n air is vented off.
| |
| J
| |
| : o. Close IE12*F042A(B) RHR PUMP A(B) LPCI INJECT ISOL VALVE.
| |
| : p. Open the following vent valves, then close when air is vented:
| |
| 1RHS*VI10(V92) LPCS FILL PUMP TO RHR A DISCH
| |
| : 1) VENT (DISCH FILL PUMP DISCH LINE VE
| |
| : 2) VENT LINE VENT) arxl HIRHS*VI19(V VENT (RHR D HX VENT).
| |
| ~
| |
| 1RHS*V3010(V3009) RHR A LPCI INJECITON HDR
| |
| : 3) OUIBOARD HIGHPOINT VENT VLV (LPCI INJE VENT)
| |
| : q. Close and lock IE12*VF063A(B) SHUIDOWN COOLING A(B)
| |
| REIURN LINE FILL SOP-0031 REV - 17B PAG 10 OF 107
| |
| | |
| +
| |
| : 3. To fill RHR loop C, perform the following-CAlmON Do not open E12-MOVF105, RHR PUMP C SUP PL SUCDON VALVE if E12-VFD67 is open (Review M* ion & Umitation 2.2.2) a.
| |
| g verify open IE12*F105 RHR PUMP C SUP PL SUCTION
| |
| : b. Oxn IE12*VF012C and IE12*VF013C RHR C PUMP SUCTION VENTS then close when air is vented off.
| |
| : c. Oxn IE12*VF015C and IE12*VF016C RHR C PUh9 SEAL VENTS then close when air is vented off.
| |
| : d. Verify oxn IE12*VF085C DISCH FILL PUMP STOP CHECK TO RHR C DISCH.
| |
| : e. Verify running IE12*C003 RHR DIV 2 LINE FILL PUhF.
| |
| : f. Unlock and open IE12*VF063C RHR C F1L VALVE if necessary l
| |
| to assist in filling RHR C.
| |
| : g. Open IRHS*V3000 RHR B/C NE FILL PUMP DISCH VENT and close when air is vented off,
| |
| : h. Open IE12*F042C RHR PUMP C y INJECT ISOL VALVE.
| |
| ' N N E uN$VE I I C
| |
| : j. Open IRHS*Vl45 RHR C INJ LINE VENT LMC CONN, and close when air is vented off.
| |
| : k. Close IE12*F042C RHR PUMP C LPCI INJECf ISOL VALVE.
| |
| : 1. Open IRHS*V142 RHR C INJ LINE VALVE VENT LMC CONN then close when air is vented off. .
| |
| : m. Verify closed and locked IE12*VF063C RHR C FILL VALVE.
| |
| 4.1.6 Verify the associated RHR A (B) (C) DISCH PRESS ABNORMAL alarm resets.
| |
| 4.1.7 Rack in the associated RHR Pump Bn:aker for loop vented.
| |
| ~
| |
| : 1. LENS *SWGIA ACB03 RHR PUMP A
| |
| : 2. LENS *SWGlB ACB23 RHR Pyh9 B
| |
| : 3. LENS *SWGlB ACB28 RHR PUMP C SOP-0031 REV - 17B PAGE 11 OF 107
| |
| | |
| -1 i
| |
| 4.1.8 Verify the green and white indicating lights are on for all 3 RHR pump control switches.
| |
| 4 4.1.9 Verify all RHR System suitus lights off..
| |
| 4.2 Manual LPCI Startup NOIE All controls and indications are located onpar. I 1H13*P601 unless noted otherwise.
| |
| NOIE
| |
| ((needed the LPCImode ofRRR may LPC9'R I YI b1NITIA110N evitchfor LPCI A loop, or the RHR DIV
| |
| . 2 MANUAL 1NITIATIONswitchfor the LPCIB & Cloop. 7his will cause the associated Div I or 11 Diesel Generator to start and kiitiate the associated Div 1 or11 Load Shed and Sequencirg.
| |
| 4.2.1 Verify the system is in standby per Section 4.1.
| |
| 4.2.2 Start IE12*C002A(BXC) RHR PUMP A(BXC) in the desired loop.
| |
| 4.2.3 Verify pump amps less than or equal to 91 Amps.
| |
| [
| |
| 4.2.4 Establish one of the folicwing flowpaths:
| |
| : 1. If reactorpressure is less than 450 psi 3 _open IEl?*F042A(BXC) RHR PUMP A(BXC) LPCI INJECT ISOL VALVE to establish flow to RPV, L
| |
| CAlmON Do not opersee the LFC!i pump and RHR A Ptunp in the test return to suppression pool mode simulmW. :
| |
| : 2. If reactor pressure is than 450 i open IE12*F024A(B)
| |
| (IE12*F021) RHR P A (BXC) RTN TO SUP PL to establish a suppression pool to suppression pool loop.
| |
| ! 4.2.5 Verify lE12*F064A(BXC) RHR PUMP A(BXC) MIN FLOW TO SUP PL
| |
| . closes when flow exceeds 1100 GPM.
| |
| l* .
| |
| S &.0031 REV - 17B PAM 12 & 107
| |
| | |
| l 9
| |
| i 4.2.6 At approximately 250 psig n: actor pressure, establish RHR flow to the RPV if destred by the performmg of the following-
| |
| : 1. Vetify open IE12*F042A(BXC) RHR PUMP A(B)(C) LPCI INJECT ISOL VALVE (S).
| |
| : 2. Verify closed IE12*F024A(B)(IE12*F021) RHR PUMP A(DXC)
| |
| 'IIST R'IN. TO SUP PL 4.3 Shutdown Cooling Flush, Warmup wi Startup CArmON Use extreme caution when positioaing the Shutdown Cooling Valves. Refer to Precaution and Unitation 2.2. (Ref. 7.35)
| |
| J l CArmON During system warmu > or shindown cooling operation, do not exceed 100 F/hr t mp changes to l the reactormderorRiR pam
| |
| - CAUI1ON l Priorto initiati Shutdown Cooling the leads lified per AhR 89-0013 SHAIL be relanded to -
| |
| !nvide overt loss of power annunciation for 1E12*MOVFD09 RHR SHUIDOWN COOUNG
| |
| .NBD ISOL VALVE. .,
| |
| ' NOIE ;
| |
| Ifperfonning thisprocedurefrom the remote shutdown is, E12*R)06A RHR PUMP A SDCSUCTION (B)
| |
| VAL VE(S) wi opente as throttle vdves.
| |
| 4.3.1 Shutdown Cooling Flush
| |
| : 1. Close IE12*VF085A(B) LPCS FILL PUMP STOP CIECK TO RHR A DISCH (DISCH FILL PUMP STOP CHECK TO RHR B DISCH).
| |
| : 2. Rack out LENS *SWGIA ACB03 (IENS*SWGlB ACB23) RHR PUMP
| |
| ' A(B) Breaker.
| |
| : 3. Close IE12*F064A(B) RHR PUMP A(B) MIN FLOW TO SUP PL ard close 1E12*F004A(B) RHR PUMP A(B) SUP PL SUCHONVALVE.
| |
| : 4. Perfo m the following-
| |
| : a. Verify E12-MOVF008 ENABLE / DISABLE switch in ENABLE (located at IC61-PNL001. in Div 1 RSS Room).
| |
| SOP-0031 REV - 17B PAGE 13 OF 107
| |
| | |
| The logic looks for F0(M or F066 or F006 and F008 and F009 to be open.
| |
| 4.2.2 Once initiated automatically, the pumps may stopped even though an initiation signal is present. This is the manua; override cor.dition. The pump will not start automatically again until the initiation signal (s) is(are) clear and the initiation reset pushbutton is depressed Obj. #6, 7 4.3 SDC suction valves (F008/009) 4.3.1 Shuts on R'iR Equipment area high temperature 1170 F 4.3.2 Shuts on RHR Equipment area differential high temperature 290 F 4.3.3 Shut 'n Low reactor water level 3 4.3.4 Shuts if reactor pressure is above 135 psig. (Based on the design temperature of the pumps)-
| |
| 4.3.5 Manual CRVICS isolation Obj. #6, 7 4.4 Pump Suppression Pool Suction. Valves (F004A,B & F105C) 4.4.1 No Automatic functions.
| |
| 4.4.2 Will not open if pump SDC suctiong (F006A,B) are open.
| |
| 4.4.3 Normally open when the system is in standby Obj. #6, 7 4.5 Pump SDC Suction Valves (F006A,B) 4.5.1 No automatic functions.
| |
| 4.5.2 Will not open if F004A,B or FO24A/3 valves are open.to prevent draining the n: actor vessel to the suppression pool.
| |
| NOTE: F024 can be opened if F006 is already.open, thus providing a direct path to the suppression pool from the reactor.
| |
| An annunciator will alarm if bot valves are open at the same time.
| |
| Obj. #6, 7 4.6 RIIR Pump Minimum Flow Valve (F064A,B,C):
| |
| , 4.6.1 Auto Open l
| |
| * Pump breaker closed (in' test or fully racked in)
| |
| I
| |
| * After 8 second time delay i
| |
| l 11L0-021-7 PAGE 13 of 23 1
| |
| | |
| 1
| |
| * Pump flow less than 1100 gpm 4.6.2 Auto close when pump flow is greater than 1100 gpm.
| |
| Obj. #6,7 4.7 IIeat Exchanger Bypass Valves (F048A,B) 4.7.1 Auto open following an initiation signal and receive an open signal for 10 minutes.
| |
| 4.7.2 Can be closed manually 10 minutes after an initiation signal.
| |
| 4.7.3 Normally open in the standby mode Obj. #6, 7 4.8 Suppression Pool Test Return Valves (F024A,B) and F021(C) 4.8.1 Auto close upon an initiation signal.
| |
| 4.8.2 F024A,8 Can be opened manually following an initiation
| |
| , signal.
| |
| . Will cause the RHR Test Return Valve In Man Override annunciator to alarm Obj. #6, 7 4.9 RilR Pump LPCI Injection Valv'es (F042A,B,C):
| |
| 4.9.1 Auto open when LPCI initiation signal present AND reactor pressure less than 487 psig. J 4.9.2 Can be opened manually when reactor pressure is less than 480 psig, e A 15 minute timer starts when pressure decreases be:ow 480 psig. If pressure rises above 480 psig, the valve can still be stroked open unti; the fifteen minute timer expires.
| |
| 4.9.3 Can be closed manually following an initiation signal.
| |
| * RHR Injection Valve in Man Override annunciator will alarm o Amber light above valve's control switch will energize
| |
| * If in manual override, the valve will not operate automatically until either one of the following conditions is met:
| |
| - The initiation signal is clear and the initiation reset pushbutton has been depressed.
| |
| "r;= ;ic:; dcv 487 pdg.
| |
| l 111.0 -021-7 PAGE 14 of 23
| |
| | |
| RO EXAM KEY Esam Number NRC RO Rev. 0 Esam Title NRC RO EXAM 4.t 1he RCIC :yitem is in Standby Lineup, but the RCIC TURBINE EXilAUST SilOTOFF valve, E51 I O68, is closed for a valve stnke test. A loss of feedwater causes a low reac,ar water level (level 2).
| |
| Select the statement which describes how the RCIC system will respond.
| |
| : a. RCIC TURillNE EXilAUST SilVTOIT valve, E51 1068, automatically opens: RCIC system initiates and injects water into the RpV.
| |
| : b. The RCIC turbine will start and trip on high RCIC turbine exhaust pressure at 25 psig.
| |
| : c. RCIC starts and the RCIC lystcm exhaust rupture diaphrams will rupture initiating a PCIC system isolation at 10 psig exhaust diaphram pressure,
| |
| : d. RCIC turbine does not start. RCIC TURDINE EXil AUST SilUTOFF valve, E51 I'068, must be open ms RCIC STEAM SilVTOFF valve, E51 F045, to open.
| |
| ANSWER:
| |
| d RCIC turbine does not start. RCIC TURillNE EXilAUST SilVTOFF valve, s
| |
| E51 1068, must be open for RCIC STEAM SilUTOFF valve, E51 F045, to open.
| |
| IDNO: LPW 010 #
| |
| ~j 10 llLO 017 5 PRDCCDURE NUMBER: OTHER:
| |
| SOP 0035 LEVEL 4 AOP-0031 ESKICS05 ESK ICS06 I217000 A2.03 i 34I 3.3 COMMENTS; I/91 new RO T2 G1 SRO T2 G1
| |
| | |
| f RO EXAM KEY Esam Number NRC-RO Rev. 0 Esam Title NRC RO EXAM 45 ne plant is operating at 100 % power, steady state. De Control Room Operator is performing LPCS Quarterly Pump Surveillance. De LPCS pump is running in the test return to ll.e suppression pool mode. A steam leak in the Drywell caused Drywell pressure to increase to 1.72 psid. Reactor pressure is being maintained at 950 psig by the bypass valves.
| |
| Which of the following statements describes the response of the LPCS system 7
| |
| : a. ne LPCS Pump willload shed then remain in standby.
| |
| We L21 F012 (LI CS TEST RTN TO SUPP POOL) closes.
| |
| De E21 F0ll (LPCS hilN FLO TO SUPP POOL) opens.
| |
| : b. We LICS Pump will continue running.
| |
| We E21 F012 (LPCS TEST RTN TO SUPP POOL) remains open.
| |
| He E21 F0ll (LPCS hilN FLO TO SUPP l@L) remains closed.
| |
| : c. De LPCS Pump will continue running.
| |
| *lhe E21 l'012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| ne E21 F0ll (LPCS hilN FLO TO SUPP POOL) opens.
| |
| : d. %c LPCS Pump will load shed then remain in standby, he F.21 F012 (LICS TEST RTN TO SUPP POOL) closes.
| |
| De E21 F0ll (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| ANSWER: ,j
| |
| : c. De LPCS Pump will continue running.
| |
| He E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| ne E21.F0ll (LICS MIN FLO TO SUPP POOL) opens.
| |
| IDNO: LP # Olu #
| |
| 247 LOTM 17 9 PROCEDURE NUMBER: OTHER:
| |
| SOP 4032 LE'/EL 2 209001 K4.08 I 3.Bi 4 __
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 43
| |
| | |
| l RO EXAM KEY Esam Number NRC RO Rev 0 Esam Title NRC RO EXAM 46 1he llPCS system is running in test return to the CST. A Suppression' Pool high level occurs.
| |
| Which one of the following describes the flow path of the !!PCS system?
| |
| : a. IIICS pump suction from the suppression pool, discharge to the CST through the test return line.
| |
| : b. IIPCS pump suction from the CST, discharge to the CST through the test return line.
| |
| : c. IIPCS pump suction from the CST, discharge to the suppression pool through the min flow line.
| |
| : d. IIPCS pump suction from the suppression pool, discharge to the suppression pool through the min 110 v line.
| |
| ANSWERt
| |
| : d. IIPCS pump suction from the suppression pool, discharge to the suppression pool through the min flow line.
| |
| IDNO: LP # OllJ #
| |
| 255 llLO-Ol9 9 PROCEDURE NUMBER: OTHER:
| |
| LOTM-18 LdEL 2 SOP.0030 200o02 A2.12 1 3.3t 3 f> _
| |
| COMMENTS: 7/97 new RO T2 01 SRO T2 G1 SYSTEM 203 ESK 4
| |
| 46
| |
| * Core submergence RPV level at or above the top of active fuel
| |
| * Steam cooling with injection into the RPV.
| |
| * Steam cooling without injection into the RPV.
| |
| : 3. Motor Operated Valves All system MOVs receive power from 480 VAC Division ill i!SF bus E22*S002 and are remotely controlled from panel lill3 P601. HPCS MOVs may also be positioned locally by taking manual control of their limitorque actuators.
| |
| The llPCS injection valve (MOVF004) is a normally closed 10" gate valve which will 4 i
| |
| auto-open to provide flow to the vessel upon receipt of a HPCS initiation signal (manual, 1.68 psid Drywell or Reactor water level 2 -43"). F004 will close to prevent over filling 1
| |
| the vessel in the event that a reactor water level high signal (level 8, +51") is received, and will automatically reopen at Level 2. In the presence of a auto-initiation signal, F004 may be manually overridden by placing its control switch in CLOSE. If F004 is manually overridden closed, it will not automatically open/ reopen at Level 2.
| |
| The HPCS CST suction valve (MOVF001) is 16" gate valve. Even though the valveis
| |
| - normally open to provide water to the suction of the HPCS pump,it receives an open signal upon HPCS initiation. F001 is slaved to F015 (suppression pool suction valve). If F001 is open and F015 subsequently opens, F001 will close when F015 reaches full open.
| |
| HPCS suppression pool suction valve (MOVF015) is a ngrmally closed 20" gate valve.
| |
| Upon the receipt of either a low CST level (98 %") or a high suppression pool level (20'4") sign 01, F015 will open to provide the attemate suction supply to the HPCS pump.
| |
| When F015 n.sches full open, F001 will close. F015 may be manually overridden by placing its control switch to CLOSE in the presence of an auto-open signal. His will allow the CST suction valve MOV F001 to be reopened.
| |
| Minimurn flow control valve MOVF012 protects the HPCS pump from overheating due to low flow conditions. This normally closed 4" gate valve is automatically positioned in respc.nse to pump discharge pressure and flow signals. F012 will open when pump discharge pressure exceeds 300 psig and pump flow is less than 750 gpm. F012 will close when pump flow is greater than 750 gpm, or discharge pressure is less than 300 psig.
| |
| He HPCS test bypass valve to CST (MOVF010) and test retum valve to CST (MOVF011) provide a discharge flowpath for pump testing. Both are normally closed 10" globe valves. Globe valves are utilized because they can be throttled without the risk of valv'e disc or seat damage and provide good throttle characteristics; this allows pump testing at various flow conditions. If a HPCS initiation signalis received during system testing, these valves will automatically close to provide HPCS flow to the RPV. F010 and F011 also close if an F015 open signal is received (valve moves cffit's closed seat) l to prevent returning poor quality water to the CST.
| |
| LOTM 18-7 Page 8 of 29
| |
| | |
| + IfllPCS system flow is greater than the RPV blow down rate, F004 will cycle I open and closed between Level 2 and Level 8 until the LOCA condition signal is reset, or the valve is taken to manual override. ,
| |
| n ,
| |
| + F004 is manually ovenidden by taking its control switch to CLOSE with a llPCT Initiation signal present. Manual override will cause F004 to CLOSE, the IIPCS ,
| |
| ~a INJECTION VALVE E22'F004 MANUAL OVERRIDE annunciation to alann.
| |
| and the AMBER light above the F004 control switch to illuminate.
| |
| + If F004 is manually overridden, it WILL NOI reopen automaticdly.
| |
| Reinstatement of the auto open function requires depressing the IIPCS INITIATION RESET pushbutton after the LOCA signal has cleared
| |
| ' Die following apply to manual operation of F004:
| |
| . Taking the control switch to OPEN will open the valve providing no auto close signal (Level 8) is present. If F004 is closed automatical'y due to a high level (level 8) signal it may be manually reopened before level drops to level 2 by depressing the llPCS 111G11 WATER LEVEL RESET PUSilBUTTON when level drops below Level 8. When the reset pushbutton is depressed, the WlilTE indicating light (above the reset pushbutton) will extinguish. ,
| |
| Position indicator lights for F004 are provided as follows:
| |
| + GREEN - valve closed J
| |
| - AMBER - valve in manual override closed condition
| |
| - RED valve open ~
| |
| 5, llPCS CST Suction Valve (F001)
| |
| F001 is controlled by a 3 position, CLOSE-(spring return to) AUTO-OPEN, switch on P601-16C, In AUTO the valve will open upon HPCS initiation provided that the suppression pool suction valve (F015)is not fully open. F001 is slaved to F015 such that if F001 is open and F015 subsequently opens, F001 will shut when F015 reaches full open.
| |
| - Valve position is indicated by RED (valve open) and GREEN (valve closed) position indicating lights, a .F001 may be manually opened if F015 is not fully open.
| |
| Page 12 of 29 LOTM 18-7
| |
| | |
| l s ,
| |
| l 6. HPCS Suppression Pool Suction Valve (F015)
| |
| - F015 is controlled by a 3 position, CLOSE-(spring retum to) AUTO-OPEN, cwitch on i P601 16C. In AUTO the valve will open upon the receipt of either CST low level (98'6")
| |
| l or suppression pool high level (20'4") si nals. F015 can be manually operated by use of its control switch. . ;
| |
| + The auto-open signal on CST low level is time delayed by 2 seconds to prevent
| |
| 'l inadvertent valve cycling upon a pressure transient caused by the starting of either j the HPCS or RCIC pump. 1 1
| |
| + F015 may be manually overridden by placing its control switch to CLOSE in the l presence of a F015 auto open signal. The valve will close and the HPCS SUP PL SUCTION VALVE E22'F015 MANUAL OVERRIDE annunciator will alarm.
| |
| F015 will remain closed until the switch is taken to OPEN. Dolag this will allow !
| |
| F001 to be reopened.
| |
| . F015 position is indicated by RED (valve open) and OREEN (valve closed) '
| |
| indicator lights.
| |
| . HPCS initiation logic has na input to F015. F015 will only auto open on lowi CST or high suppression poollevel. . ,
| |
| : 7. HPCS Minimum Flow Control Valve (F012)
| |
| F012 is provided with a CLOSE -(spring return to AUTO)- OPEN control switch. In '
| |
| AUTO, the valve's position is controlled by signals from fiPCS pump discharge pressure i
| |
| transmitter PT-N051 and flow transmitter FT-N056.
| |
| When in AUTO, with a signal of greater than 300 psig from PT N051 (providing i indication of HPCS pump running), F012 will open upon a flow signal ofless than 750 '
| |
| l gpm as sened by FT-N056. When flow rises above 750 spm, F012 will close. Should HPCS pump discharge pressure drop to less than 300 psig (indicating pump stopped) i F012 will close.
| |
| j-
| |
| * F012 position is indicated by RED (valve open) and OREEN (valve closed) j -- indicator lights.
| |
| 1.
| |
| i
| |
| : 8. HPCS Test Bypass Valve To CST (F010) and HPCS Test Return Valve To CST (F0l l)
| |
| Both valves are provided with CLOSE- (spring return to) AUTO - OPEN control switches. In AUTO the valve control circuitry is established such that the valves will ;
| |
| automatically close upon receipt of a HPCS initiation signal. Additionally, both valves l l
| |
| . will receive a close signal whenever F015 opens. The auto close feature associated with an F015 open signal gevents the introduction of suppression pool water into the CST. l LOTM 18-7 Page 13 of 29. ;
| |
| .---.-r--- v.--,wmr--,-.-2-- r-. --n-.-,-. ,-m--., %+rv---r,- n-m.- -+ . , , - - . . ,p-. ---. , - - , + - .e.,r,. ,~ .
| |
| * m.-.r-4.-- e-- - w--r evet ev- -=w w-1
| |
| | |
| RO EXAM KEY ,
| |
| Esam Number NRC RO Rev. O Esam Title NRC RO EXAM 47 "Ihe SLC system is in its standby readiness lineup when an ATWS occurs. The UO is directed to initiate SLC "A" and verify it is injecting into the vessel. As the UO attempts to do this, which of the following statements identifica the EXPECTED sequence of events within the SLC system 7
| |
| : a. Suction valve opens fully, both squib valves fire, pump starts,
| |
| : b. Suction valve opens fully,"A" squib valve fires, the pump starts.
| |
| : c. The "A" squib valve fires, suction valve opens fully, pump starts,
| |
| : d. Both squib valves fire, both suction valves open fully, putop starts.
| |
| J ANSWER:
| |
| : c. The "A" squib valve fires, suction valve opens fully, pump starts.
| |
| I IDNO: Lp# Oit! #
| |
| 249 llLO-016 3 PROCEDURE NUMBER: OTHER:
| |
| SOP 0028 LEVEL 2 211000 A3.08 i 4.2i 4,2
| |
| ,j COMMENTS:T H97new RO T2 G1 SRO T2 G1
| |
| - - - , . - .- -. . . - . - . , . . . - . . . - _ - . . - - , - - . - . - = _ - . . - - . . . . . . - - - . - . . - -
| |
| | |
| 5.2.2. Unlock and open C41 VF010, MAKEUP TO SLC STORAGE TANK VALVE and add water to the SLC Storage Tank to a level determined by Chemistry.
| |
| 5.2.3. WHEN the desired level is reached in the SLC Storage Tank, HIES close and lock C41 VF010, MAKEUP TO SLC FTORAGE TANK VALVE.
| |
| 5.2.4. Notify Chemistry water addition to the SLC Storage Tank is complete.
| |
| I 5.3 SLC Injection NOTE Thefollowing controls andIndications are on H13 P601.
| |
| 5.3.1. Check C41 F031, TEST TK OUTLET VLV is closed.
| |
| 5.3.2. Place C41 SI A(B), SLC PUMP A(B) to RUN. .
| |
| 5.3.3. Check C41 F004A(B) SQUIB CONTINUITY light goes off and Annunciator P601 19A F05(F06), STANDBY LIQUID SYSTEM "A"("B") INOPERATIVE alarms.
| |
| 5.3.4. Verify G33 F004(F001), RWCU PUMPS 9UTBD(INBD) -
| |
| SUCTION VALVE closes.
| |
| 5.3.5. Check C41-F001 A(B), SLC PUMP A(B) SUCT VLV opens and SLC Pump A(B) starts.
| |
| 5.3.6. LE the selected pump or pump suction valve fails to function, THEN perform the following:
| |
| : 1. Place the selected switch to STOP.
| |
| 1 .
| |
| : 2. Repeat Steps 5.3.2 through 5.3.5 for the alternate pump.
| |
| 5,3.7. Check system is injecting into RPV by checking the following:
| |
| * SLC Storage Tank levellowering
| |
| * Reactor power lowering ,
| |
| 1 *
| |
| * System pressure greater than Reactor pressn.:.
| |
| SOP-0018 REV-9A PAGE 6 OE.18
| |
| | |
| l 1
| |
| l l
| |
| RO EXAM KEY ,
| |
| l Esasa Number NRC.RO Rev. 0 Esam Title NRC RO EXAM 4g 'the plant was operating at 100% power with the prefened AC power sources lired up, when the Prefened Station Service Transfonner 1RTX XSRIE (R.S.S #l Leads) experienced a sudden pressure lockout trip. Which of the following is the expected status of RPS power source?
| |
| I
| |
| : a. RPS A is de-energized; RPS 11 is energized from its nonnat supply. j i
| |
| : b. RPS A is de-energized; RPS 11is de-energized.
| |
| : c. RPS A is energized from its normal supply; RPS D is energized from its alternate .
| |
| I supply.
| |
| : d. RPS A is energized from its alternate supply; RPS Ilis de-energized. ;
| |
| ANSWER:
| |
| : a. RPS A is de-energized; RPS 11 is energized from its normal supply.
| |
| IDNO: LP# Olu #
| |
| 297 IILO 529 2 PROCEDURE NUMBER: OTHER:
| |
| AOP-0010 LEVEL 2 SOP 0079 J
| |
| NRC KA: ' RO: IshdI 212000 K4.03 31 3.1 212000 K6.01 3.6f~ 38 ._
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 01 0
| |
| 48
| |
| | |
| 506029-1 13.8KV (Norm) 4160 VAC (DIV 1) 4160 VAC (DIV ll) 13.8KV (Norm) 1NPS SWG1A 1 ENS'SWG1A 1 ENS *SWG1B 1NPS-SWG1B t
| |
| ee 13.8KV he 4.16KV em 4.1SKV em 13.8KV w m 480 VAC em 480 VAC mw 480 VAC w m 480 VAC t 450 017 041 062 1NJS SWG1U 1EJS*SWG1 A 1EJS*SWG1B 1NJS-SWG1D 451 08 48 59 1NHS-MCC10A2 1EHS*MCC14A 1EHS*MCC14B 1NHS-MCC108
| |
| -- -- 400 --
| |
| 400 125 125
| |
| ~~ ~~
| |
| PHA-B PHC-A RPS M 25HP 480VAC 480VAC RPS MG 25HP SET hM SET 1RPS*XRC10A1 RPS*XRC1081 A 18.75KVA ' ^
| |
| 120V 120V MG OUTPUT ,J MG OUTPUT 3E '3G BKR U BKR
| |
| *$A '3F '3H '3B fo KEY OPERATED
| |
| *3D
| |
| /0
| |
| '3C
| |
| \o SELECTOR \o 120VAC / SWITCHES --
| |
| 120VAC NORMAL --[ NORMAL SE SE
| |
| * EPA BREAKERS
| |
| * EPA BREAKERS 1C71 P001 RPS BUS A ' 1C71 P002 RPS BUS B l I NSSSS
| |
| - NEUTRON MONITORING
| |
| - RPS MAIN STEAMLINE RADIATION MONITORING
| |
| ~ 5i-s =
| |
| * s m $a
| |
| * RPS POWER SUPPLIES mum
| |
| | |
| .-_~ - __ _ . _ . _ _ _ . . - - _ . . . . _ . - . ._ _ . _ . . __ _ __ _ _ _ . _ .____ _ _ _.
| |
| RO EXAM KEY Esam Number NRC-RO Rev 0 Esem Title NRC RO EXAM 49 Given the following plant conditions:
| |
| IRM *0"is bypaned l lRM *11" is bypassed l Reactor Mode switch is in START /IlOT STANDilY '
| |
| All operable IRMs are reading 65/125 on Range 9 During the troubleshooting of IRM "G", I&C requests that the operator at the controls withdraw l IRM "G". The NCO withdraws IRM "C" by mistake. Ilow does the plant respond?
| |
| : a. A " detector not fu'l-in":od block is generated in RC&lS. ,
| |
| : b. A downscale rod block is generated when the detector gets full-out. ,
| |
| : c. Nothing, at this time all rod blocks and scrams are bypassed.
| |
| : d. An INOP rod block and 1/2 scram signal are generated when the detector leaves the " full in" position.
| |
| . ANSWER:
| |
| ]
| |
| : a. A " detector not full In" rod bk>ck is generated in RC&lS.
| |
| J IDNO: LP# OBJ # ,j 29 IILO-052 4 PROCEDURE NUMBER: OTHER:
| |
| SOP-0074 ATT4 LEVEL 3 NRC KAt l RO: [SMor 29003 K4.01 _ 3. 7 3.7 2]5003_K3.03 ,_3 71 3.7 216003 A4.07 3.6 36 _
| |
| COMMENTS: t/97 now RO T2 01 SRO T2 Gt 4
| |
| : m. _ _ - _ _ _ _ , _ , _ _ _ _ . , . _ . _ _ _.___-_ _..__. _ . _. _ _ _ _ _
| |
| | |
| D. System Interlocks and Alarms The IRM Dovmscale trip indicates that reactor power level is below the IRM range of indication based on the selected IRM range or that there is a possible malfunction of l
| |
| <he IRM ehannel. The trip setpoint is 5/125 of scale. The IRM Downscale trip results in an annunciator actuation, trip status light actuation on the IRM electronics drawer front panel and a trip signal to the RCIS system which will initiate a rod withdrawal i block.
| |
| The IRM Downscale trip to the RCIS will be bypassed with the Range Switch on the indicating channel selected to Range 1. With the Reactor Mode Switch in the "RUN" position, the IRM Dowmcale trip will not be functional and all annunciators, status light indication and control rod withdrawal blocks are bypassed except for the trip unit status lights on the IRM drawers The IRM Upscale Alarm trip indicates a neutron flux level in excess of that allowed for a specific plant operation condition. The trip setpoint is 108/125 of scale. The IRM Upscale trip results in an annunciator actuation, status light indication on the P680 panel, trip status light actuation on the IRM electronics drawer front panel and a trip signal to the RCIS system which will initiate a control rod withdrawal block. The IRM Upscale Alarm trip with the RCIS will be bypassed when the IRM channel affected is bypassed.
| |
| The IRM Upscale rod block trip is automatically bypassed in the RCIS when the Reactor Mode Switch is placed in the "RUN" position.
| |
| The IRM Upscale trip indicates that reactor power is ouJside the bounds of detector range or that there is a possible malfunction causing indication to fail upscale. The trip setpoint is 120/125 of scale. The Upscale Trip results in annunciator actuation, trip status light actuation on the P680 panel, trip unit status light actuation on the IRM electronics drawer front panel and a trip signal to RPS resulting in a Chalf scram" if not in the run mode. Failure to clear the upscale trip condition will ca .se a reactor scram if the Reactor Mode Switch is in the "STARTUP" position and a second IRM channel from the other RPS Division trips. For example, suppose IRM Channel A fails upscale and the upscale trip is initiated. As a result, RPS Channel A is tripped (half scram). If any IRMassociated with RPS Channels B or D should trip (i.e., IRMs B, D, F or H),
| |
| the reactor will scram.
| |
| The IRM laoperative trip indicates that the affected IRM channel is inoperative. The inoperativv trip will actuate when: 1) the high voltage to the fission chamber is less than 80 VDC,2) a drawer module is unplugged, or 3) the Mode /fest switch on the IRM drawer is not in the OOperate" position. This trip results .n i annunciator actuation on the P680 panel, trip unit status actuation on the IRM electronics drawer front panel, and a trip signal to RPS. Failure to clear the inoperative condition will cause a reactor scram if the Reactor Mode Switch is in the " START-UP" position and a second IRM channel not in the same RPS Division becomes inoperative or trips Upscale (120/125 of scale).
| |
| LOTM 10-5 Page 11 of 19
| |
| | |
| - A rod withdrawal block trip to 6 KCIS is initiated when an IRM Detector is in the wrong position. Detector Wrong Position is defined as the detector not fhil in and the Reactor Mode Switch not in the "RUN" position. There are no annunciators associated with this trip.
| |
| The only indication available to the operator in the event of an IRM detector in the wrong position will be a " ROD BLOCK" annunciator and the absence of an IN iight on
| |
| - the affected IRM Drive selec. matrix with Drive Power available. The appropriate 3
| |
| i operator action to correct the situation is to reposition the detector to the full in position, if the IRM detector will not move (insert), this may indicate that: 1) there is a ;
| |
| < malfunction in the Motor Drive Module or 2) the full in limit switch on the Motor ,
| |
| Drive Module is malfunctioning.
| |
| IV. SYSTEM INTERRELATIONS.
| |
| Information concerning system interrelations is contained throughout this chapter. For more i
| |
| information see the following LOTM chapters:
| |
| Reactor Protection System (Chapter 15) ,
| |
| Rod Control and Information System (Chapter 6) ;
| |
| . V. SYSTEM OFERATION :
| |
| l The IRM system is normally used during reactor startups and shutdowns when reactor
| |
| ; power level is within the range of the detection (nominally u,0001% to 20% rated reactor
| |
| . core thermal power), During a normal reactor staqup, all IRM detectors are fully inserted r;
| |
| into the reactor core (15 inches above core centarline) with the.Aange I selected. All IRM channels should be operable and each channel should be veri 8ed for operability prior to startup by surveillance tests. The Moderrest switches on the IRM electronics drawer should- ,
| |
| [
| |
| be in the " OPERATE" position for each IRM channel and all divisions of the neutron l-monitoring should be in service (not bypassed). Any trip status lights on the IRM l
| |
| electronics drawer should be reset prior to reactor startup.
| |
| [
| |
| Initially, during reactor startup, the approach to criticality arxi monitoring oflow reactor f
| |
| power will be monitored by the SRM system. When the SRM count rate is between 10 and 105 cps the IRM system should be 'on scale" on Range 1. This will indicate power overlap of the IRM SRM Systems. As soon as the IRM System is on scale, it will be used i
| |
| to monitor' reactor power and the SRM detectors will be increaHy withdrawn to ,
| |
| maintain a count rate between lx103and lx10 _5 eps, l As reactor power increases and IRM channels exceed 75/125 of scale, the Upscale status light on the affected IRM channel will illumhmte to remind the operator that the Range switch should,be incremented to the next higher rangc. While ranging IRM channgis, the IRM reading should be maintained between 15/125 and 75/125 of scale for each IRM channel to avoid withdrawal blocks (upscale or downscale) from being imposed. Improper ranging can result in unnecessary control rod withdrawal blocks or a reactor scram. -In the IRM range during operation with a positive period, the reactor power is displayed on a linear scale but reactor power increases exponentially. This results in the indication appearing to .
| |
| LOTM 10 5 Page 12 of 19
| |
| - . - .- - = - - - - - - - - _ . . _ _ _ _ - . _ _ _ . - - . - _ - -
| |
| | |
| - _ _ . . _ _ _ _ _ _ . . . - _ _ _ . m. - . ,_ _ . . _ . _ . - - _
| |
| P RO EXAM KEY t 1: sam Number NRC.RO Rev. O t. sam Title NRC RO EXAM ,
| |
| 50 A reactor startup is underway, with the MODE SWITCll in STAR 111P, SRMs are being withdrawn to maintain count rate per procedure, and power in the intermediate ange. Which of the following conditions would generate a rod block?
| |
| : a. SRM "A" fails low (pegged downseale), all other IRM's are on range 3.
| |
| : b. SRM *C" fails high (pegged upscale), IRM *0" is on range 8, all other IRM's are on '
| |
| range 9.
| |
| : c. SRM '!!" fails low (pegged downscale), IRM 'F" is on range 2, all other IRM's are on range 3.
| |
| : d. SRM *D" fails high (pegged upscale), all other IRM's are on range 9. P ANSWER:
| |
| : c. SRM *II" falls low (pegged downscale), IRM 'T" is on range 2, all other IRM's at:
| |
| on range 3.
| |
| IDNO: LP# Olk! #
| |
| 358 STM 503 4 PROCEDURE NUMSER: THER:
| |
| SOP 4074 L- EL2 ARNiS0-5-C05 NMC KAt RO l SROi 215004 K4.01 3.71 3,7 COMMENTS: 7/97 new LOTM 9 4. Tablo 1 Table 3. p.18 & 20 RO T2 01 SRO T2 G1 30
| |
| | |
| A l 1 AlttMt.N i 2 )
| |
| PAGE 1 OF 1 SRM SCRAM TRIPS / CONTROL ROD BLOCKS SCRAM TRIPS TRIP SETPOINT BYPASSED t
| |
| SRM Flux Upscale Trip 2 x 10'eps 1) Shortir.g links installed SRM Inoperative 1) Module Unplugged 1) Shorting links installed
| |
| : 2) Low High Voltage ofless than 95 %
| |
| : 3) SRM Mode Switch Not in OPEP ATE CONTROL ROD BLOCKS TRIP SETPOINT BYPASSED SRM Downscale Alarm 3 cps 1)IRM Range 3 or above
| |
| , 2) Mode Switch in RUN 5
| |
| SRM Upscale Alarm 1 x 10 cps 1)IRM Range 8 or above
| |
| : 2) Mode Switch in RUN SRM Inoperative 1) Module Unplugged j I)IRM Range 8 or above
| |
| : 2) Low Illgh Voltage ofless thad 2) Mode Switch in RUN 95 %
| |
| : 3) SRM Mode Switch Not in OPERATE SRM Detector Retract Permit 100 cps and Detectors Not Fully 1)IRM Range 3 or above Inserted 2) Mode Switch in RUN e
| |
| e SOP 0074 REV-6 PAGE 11 OF 35
| |
| | |
| SRM ,
| |
| ALARM NO. 2153 UPSCALE lill3*P680 / OSA / C05 OR INOPERATIVE INITIATING DEVICE SET POINT
| |
| : 1. Til3 P671 K59 1. 1 X 10' cps AUTOMATIC ACTIONS
| |
| : 1. An SRM UPSC or INOP initiates rod withdrawal block ifIRMs not on Range 8 or above in Op.
| |
| Condition 2.
| |
| : 2. With RPS shoning links removed, a reactor scram will occur on SRM Inop.
| |
| OPERATOR ACTIONS
| |
| : 1. Stop control rod pulls and allow count rate to stabilize.
| |
| : 2. If scram occurred, refer to AOP-0001 REACTOR SCRAM,
| |
| : 3. Ifin Op. Condition 5, stop all Core Alterations and remove personnel from line-of-site of reactor.
| |
| LONG TERM ACTIONS ,
| |
| : l. Insert Control rods to decrease count rate.
| |
| : 2. Determine cause of positive reactivity increase.
| |
| : 3. Withdraw the SRM detectors from the core to decrease the coupt rate.
| |
| : 4. Bypass the SRM channel if proven to be defective.
| |
| : 5. Refer to SOP 0074 NEUTRON MONITORING SYSTEht POSSIBLE CAUSES
| |
| : 1. Excessive control rod withdrawal. ,
| |
| : 2. Drifling control rod.
| |
| : 3. Reactivity addition from source other than rods.
| |
| : 4. Power level too high for SRM detectors to be fully or partially insened in the core.
| |
| : 5. Ifin Op. Condition 5, Core Alteration error.
| |
| : 6. Defective SRM channel.
| |
| : 7. Failure of SRM detectors to retract.
| |
| REFERENCES - ,
| |
| : 1. GE Dwg. 851E88flAA Sh.17 ARP-680-05 REV-6 PAGE 24 OF 29
| |
| | |
| RO EXAM KEY-Esem Number NRC RO Rev. O Esam Title NRC RO EXAh!
| |
| 51 The reactor has been operating near rated power for 200 days. Which one of the following ,
| |
| describes the change in the indicated LPRM output signal from day I to day 200 and the method l I
| |
| used to calibrate the LPRMs?
| |
| INDICATED LPRM POWER METilOD OF LPRM CAllBRATION :
| |
| : a. Decreases Core lleat Dalance 4
| |
| : b. Decreases TIP System Trace
| |
| : c. Increases Core Ileat Balance
| |
| : d. Increases TIP System Trace ANSWER:
| |
| E. Decreases TIP System Trace IDNO: LP# OILI W 89 STM 503 9 PROCEDURE NUMBER: OTHER:
| |
| SOP-0074 LEVEL 2 i _ _. . .-_ . . . _v ... .._
| |
| ,j 21600(K1.0L_i_2.6, ._ _2 : 8 215006 K1.13 l 2.6 3_
| |
| COMMENTS: 7/97 new RO 12 G1 SRO T2 G1 RO T2 G3 (TIPS)
| |
| Sr.O T2 03 (TIPS) 0 31
| |
| | |
| i i
| |
| I l
| |
| I
| |
| ^
| |
| l RO EXAM KEY Esam Number NRC RO Rev. 0 Esam Title NRC RO EXAM
| |
| $2 A transient has occurred requiring inserting a manual reactor scram. ARI was required to be initiated by the ATC operator to insert all control rods.
| |
| The following current plant conditions have been steady for the last 5 minutes:
| |
| All rods are inserted.
| |
| Reactor power is 0%.
| |
| RpV pressure is 1100 psig.
| |
| RPV water levelis 50" and steady, Given these conditions, when can ARl be reset?
| |
| : a. 32 seconds after ARIinitiation.
| |
| : b. Immediately by depressing the ARI Initiation Reset pushbutton.
| |
| : c. When reactor pressure is lowered to 1060 psig.
| |
| : d. When RPV water level is restored above -43".
| |
| ANSWER:
| |
| : d. When RPV water level is restored above 43".
| |
| IDNO: LP # 011J#
| |
| 424 STM-052 4 PROCEDURE NUMSER: OTHER:
| |
| LOTM5 LEVEL 3 216000 K4.11 1 4! 4 _
| |
| COMMENTS: 7/97 new LOTM6 32
| |
| | |
| Described in Section L, Alternate Rod Insertion L. Alternate Rod Insertion (Figures 12,15)
| |
| ARI is a non safety r'etated system that consists of seven solenoid valves located in the scram air header. Six of these solenoid valves provide an attemate means of bleeding air from the scram air headet. One ATJ valve is only used during ARI valve testing.
| |
| The ARI valves se:ve a function very similar to that of the backup scram valves. The main difference between the ARI valves and the backup scram valves is their initiating condition. The backup scram valves will vent the scram air header automatically upon a reactor scram signal. The ARI valves will vent the scram air headu automatically upon Level 2 or 1127 psig reactor pressure. The backup scram valves cannot be operated manually. The ARI valves may be initiated manually to bleed down the scram air header; they may also be operated for testing in such a way that no air escapes the scram air header.
| |
| Three ARI valves are located just downstream of the backup scram valves before the i scram air header branches off to the SDV. During normal operation, the air passes !
| |
| through the three way valve F160. All other ARI valves are normally closed. When F160 shifts to the initiate (energize) position, it blocks air from entering the entire. . i scram air header; in addition, it directs air from the header to F164A. F164A, if open, discharges the air to atmosphere. F164B is only used to bypass F160 to maintain the 1
| |
| scram air header pressurized whenever F160 is tested.
| |
| The F162A(B,C,D) valves are located in pairs in close proximity to the HCU banks.
| |
| They can be seen in the overhead above the SDIVs. Both valves in a pair must open to bleed down the header.
| |
| Power is provided to the ARI valves from BYS-PNL 02A2 (located in the Main Control Room). A DC power supply allows the system to function even under station blackout conditions.
| |
| ARI Controh (Figure 16)
| |
| The ARI htWUAL INITIATE ann-and-depress push buttons (SIA, SIB) are located on panel P680. An INITIATION RESET push button (S2) is prw;ded on the P680To illow resetting of the ARI logic once the 32 second seal in sigra has expired.
| |
| Panel P632 contains the following controls:
| |
| Keylock Test Switch (53) - this keylock switch will be aligned to either OTEST A" or DTEST B" positions to allow testing of the associated ARI l
| |
| va'ves. .
| |
| ARI TEST A (D) push buttons (S4A,B) - depression of these push buttons l _
| |
| l .
| |
| will cause the associated ARI valves to energize, prmiding the keylock switch is in the corresponding TEST pos; tion.
| |
| l 1.OTM-5-6 PAGE 14 OF 33 l
| |
| ~ -
| |
| y
| |
| | |
| ARI RESET (SS) - this push button resets the ARI logic. It serves exactly the same funedon as the ARI reset pushbutton on the P680, but will n,ormally,b, used during testing.
| |
| ARIIndicatiom (Figure 16,17)
| |
| Various indicating lights are also provided on panel P632, providing system status information. All indicating lights will be de-energized during normal plant operation, except for the green valve position indications. The ARI CONTACT STATUS lights are specifically provided to alert personnel that a specific initiating contact has closed on either channel (light will energize).
| |
| ARI Logic (Figure 15)
| |
| ARI consists of channel A and B trip systems. Each trip system monitors 2 reactor water level transmitters and 2 reactor pressure transmitters. Both water level transmitters or both pressure transmitters in either trip system are required for a full ARIIrdtlation. This is a two-out-of two logic scheme.
| |
| A manuti initiation reouires simultaneous depression of the ARI MANUAL INITIATE push buttons (SI A and SIB) at Fn0 to satisfy the logic. Arming and depressing one, ,
| |
| initiation pushbutton does not cause any ARI valves to change position.
| |
| Once the logic is satisfied, a scal in time delay circuit maintains the initiation signal for 32 seconds. The number and size of the ARI valves have been designed to allow insertion of all control rods within 15 seconds, with all control rods reaching their full-in position within 25 seconds. The 32. econd sealin pignal ensures that the ARI solenoid valves are energized for a suf5cient period of tirhe for all control rods to fully insert. Initial testing has shown that all control rods actually reach full-in position in less than 3 seconds.
| |
| A Ready To Reset annunciator comes in 32 seconds after an initiation signal.
| |
| The initiation logic for channels A and B is identical. In other worda, if the initiation logic for one group of valves is satisfied, then it will be satisSed for the other group as well.
| |
| Defeatina RPS and ARI Lonic Trips Keylock switches with NORMAL and EMERGENCY positions are located in the control room backpanels (Ill3 P691A, P692A, P693B, and P694B). When the switches are taken to EMERGENCY a red light by each switch illuminates. This bypasses all automatic RPS trips. Defeating ARI initiation signals requires the removal of relays in H13 *P632 (See EOP Enclosure f12). .
| |
| - This ca;iability is provided to permit the cperator,to reset the scram, allow the scram discharge volume to drain and then manually scram the reactor again. If the control rods ever become hydraulically locked in the withdrawn positions, this capability allows the hydraulic lock to be released, even if a scram signal is present.
| |
| LOTM-5-6 PAGE 15 OF 33
| |
| | |
| 4 RO EXAM KEY Esam Number NRC.RO Rev. 0 Esam Title NRC RO I!XAM 53 Shortly aner initiation, the RCIC turbine tripped on high steam exhaust pressure. The RCIC turbine trip throttle valve may be reset:
| |
| : a. using the control rcem trip throttle valve hand switch aner the trip condition is cleared.
| |
| b locally at the turUne aner the trip condition is cleared. !
| |
| : c. locally at the turbine at any time even if the condition is not cleared.
| |
| : d. In the control room using the trip reset push button on panel P601, once the condition is clear.
| |
| ANSWERt t.. using the control room trip throttle valve hand switch aner the trip condition is cleared.
| |
| II)NO: LPW OltJ#
| |
| 257 IILO-017 5 PROCEDURE NUMBER: OTHER: '
| |
| ARP-P60121-C03 LEVEL 2 SOP-0035 J
| |
| ~kiC KAt RO: $RO:
| |
| 217000 A4.02 3.9 3.9 _
| |
| COMMENTSt 7/97 new RO T2 G1 SRO T2 G1
| |
| ._ - , . - . . . _ _ - . . . ~ ~ . _ . . _ _ _ -- __. _ _ _ . _ _ _ - _ .
| |
| | |
| l, l
| |
| i DIV I RCIC 150L TURB EXH PRESS RIGH i t
| |
| HI3 P601/21A/C03
| |
| : u. ARM NO. 244i INITIATING DEVICES SETPOINTS
| |
| : 1. E51 PISN656A or E, RCIC Turbine Exhaust Pressure Switch 1. 25 psig j AUTOMATIC ACTIONS ,
| |
| : 1. RCIC Turbine trips. ;
| |
| : 2. E51.F013, RCIC INJECT ISOL VALVE closes, l OPERATOR ACTIONS 1, Verify reactor level is being maintained by an alternate source.
| |
| 2.' Verify Automatic Actions occur. l
| |
| : 3. Verify E51 F068, RCIC TURBINE EXHAUST TO SUPPRESSION POOL is open. ,
| |
| : 4. If RCIC System was running to maintain reactor level, then Refer To the. EOPs.
| |
| NOTE t
| |
| 1his method ofrestart reinserts the ramp generator to control speedincrease, i
| |
| *i 5.' Ifit is necessary to restart RCIC, then pe form the following: l
| |
| : a. Close E51 F045, RCIC STEAM SUPPLY TURBINE STOP VALVE.
| |
| : b. - Roset turbine trip by fully closing E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR and >
| |
| then reopening
| |
| : c. Verify Turbine Trip alaan has cleared. *
| |
| : d. Open E51 F045 and verify ramp generator controlling RCIC turbine speed. l
| |
| : e. A4ust Sow to desired dowrate.
| |
| : 6. Check ERIS Compuwr Point E51EA008.
| |
| . LONG TERM ACTIONS ~ '
| |
| : l. Determine cause of RCIC turbine exhaust pressure high and correct as necesury.
| |
| : 2. If RCIC is Dg required for level control, then shut down the RCIC System per SOP-0035, Reactor Core
| |
| - Isolation Cooling System.
| |
| 1 L If RCIC is declared inoperable, then Refer To Technical Speci8 cations Section 3.7.3.
| |
| ~
| |
| ' ARP-601-21 REV9- PAGE 29 OF 85 ,
| |
| . an. . _ . ._ n ._._ _ _ _ . _ _ _ _ _ _ . _ . . - . _ . _ __ _ ,..,.. _. _ --._. . _ _ , _ , . . _ _ _ -
| |
| | |
| 1 PURPOSE 1.1 To provide instructions for the operation of the Reactor Core Isolation Cooling (RCIC) System. .
| |
| 2 PRECAUTIONS AND f 1MITATIONS 2.1 The Main Turbine will trip after a 15 second Time Delay if RCIC is initiated automatically from an RPV Level 2 or manually using ESI A-S37, RCIC M ANUAL INITIATION Pushbutton, if E51 MOVC002 RCIC TRIP it THROTTLE VALVE is notfull closed.
| |
| I 2.2 If RCIC is started manually and aligned to inject into the RPV, the Main Turbine taip from RCIC initiation will not occur and moisture carryover .
| |
| f may result with the Main Turbine on line. ,
| |
| I f
| |
| 2.3 Starting the RCIC Pump with an injection line low pressure condition may cause damage due to water hammer. If the RCIC System Line Fill Pump is to be shutdown for an extended period, the RCIC Pump should not be allowed to start.
| |
| C 2.4 When performing tests or system lineups, the potenti# exi:ts for water hammer to occur if the system has not beer; properly filled and vented.
| |
| 2.5 f f narrow range RPV water level reaches Level 8 during RCIC operation, E51 F045, RCIC STEAM SUPPLY TURBINE STOP VALVE and ESI-F013, RCIC INJECT ISOL VALVE will automatically close which shuts down the RCIC Turbine. The RCIC Turbine will restart automatically at RPV Level 2.
| |
| 2.6 Operation of the RCIC kbine below 2300 RPM may cause Turbine Exhaust Check Valve damage due to chattering.
| |
| 2.7
| |
| * Operation of the RCIC Turbine below 1700 RPM may result in insufficient pump flow for cooling of pump intemals.
| |
| 2.8 in the event of a RCIC Turbine overspeed or local manual trip, the Turbine Trip and Throttle Valve must be reset locally. All other trips can be reset remotely from Control Room. ,
| |
| SOP-0035 REV - 14A PAGE 3 OF 40
| |
| | |
| 4.1.13. Verify E51 R600, RCIC PUMP FLOW FLOW CONTROLLER HYVC002 in Auto at 600 gpm.
| |
| 4.1.14. Close E51-C002. RCIC TRIP & THROTTLE VALVE OPERATOR.
| |
| NOTE b After a local manual trip or mechanical overspeed.the Turbine Trip Throttle Valve must be manually reset at the turbine.
| |
| . 4.1.15. IE a RCIC turbine overspeed or local manual trip exist IHEN perform the following: ,
| |
| : 1. Locally verify E51-MOVC002, RCIC TRIP &
| |
| THROTTLE VALVE is closed using the handwheel.
| |
| NOTE .
| |
| r ifdificulty is experienced while manually resetting the i
| |
| RCIC Turbine, a slight manual adjustment ef tk tappet nut >
| |
| might be required ,
| |
| : 2. At the turbine, pull the emergency trip rod towards the Trip / Throttle Valve. j
| |
| : 3. LIA up on the manual ovenpeed trip lever and verify the ,
| |
| trip lever engages the trip plunger and remair.s horizontal.
| |
| t
| |
| : 4. Verify the mr.aual trip level engages the trip plunger.
| |
| : 5. On the Trip /Ihrottle Valve, verify the trip hook lever is engaged.
| |
| 6.- Release the emergency trip rod and verify it does not '
| |
| move and the overspeed limit switch on the valve is clear.
| |
| : 7. Inform the Control Room the RCIC trip mechanism is -
| |
| reset.
| |
| L >
| |
| 4.1.16. Open E51-C002, RCIC TRIP & THROTTLE VALVE -
| |
| OPERATOR and verify RCIC TRIP & THROTTLE VAI,VE . -
| |
| POSITION red light is on. ,
| |
| i e
| |
| SOP-00M REV - 14A PAGE 8 OF 40 k' .- .. - .. - - . - ____ _ _
| |
| | |
| ._ . _ . .-. _...__ _ _ __ _ .._ , _ _ m _. _ _ _ . _ . _ _ _ _ . . _ _ _ . . -_ _ _ _ . _ _ . _ , -_m_
| |
| \
| |
| RO EXAM KEY rsam Number NRC RO Rev. g Esam Title NRC ROIIXAM
| |
| $4 following a valid ADS initiation, the operator is directed to close the ADS valves with the initiating signals still present. Which one of the following operator actions will cause the ADS valves to close?
| |
| : a. Place the control switches on lil3 P601 and til3 P631 for the ADS valves to the
| |
| *OFF" position.
| |
| I
| |
| : b. Place the ADS inhibit switches on lil3P-601 to the " NORMAL
| |
| * position
| |
| : c. Stop allicw pressure IICCS pumps in both Div. I and Div. 2.
| |
| : d. Depress both " ADS Timer / Level 3 Scal in Reset" pushbuttons, Sl3 A(D)
| |
| ANSWER:
| |
| : d. Depress both " ADS Timer /tevel 3 Seal in Reset" pushburtons, Sl3 A(II) lDNO: LP# OBJ #
| |
| 143 IILO-064 2 PROCEDUKE N'JMBER: OTHER:
| |
| SOP 0011 LEVEL 3 HLO-064 A
| |
| NRQ %At I RO: SHO:
| |
| 218000 K6.01 1 38 3.8 _
| |
| COMMENTS: 7/97 new RO T2 01 SRO T2 G1
| |
| [
| |
| l
| |
| | |
| ..-_.__=_- . _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ . _ . . _ . _ . _ - _ . _ . . _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
| |
| RO EXAM KEY Esam Nember NRC RO Rev. [ Esam Title NRC RO EXAM
| |
| $$ A Safety Relief Valve (SRV) tailpipe vacuum breaker was failed in the open position when the SRV opened. Which of the following is the result?
| |
| : a. Containment pressure increased. ;
| |
| : b. Steam trypassed the quenchers with a direct discharge path into the suppression pool.
| |
| : c. Drywell to containment differential pressure increased.
| |
| : d. Supt..ession pool water will be drawn up into the SRV discharge line after the SRV is closed.
| |
| ANSWER
| |
| : c. Drywell to containment differential pressure increased.
| |
| IDNO: I.P # OHJ#
| |
| 217 IILO 007 3 PROCEDURE NUMBER: OTHER:
| |
| P&l0 31B LEVEL 2 NRC KAt l' 505'l SRO: ')
| |
| 273001 A2.09 { 3 41 3.6 _
| |
| COMMLN15: 7/97 new RO T2 G1 SRO T2 G1 l
| |
| l l
| |
| l
| |
| | |
| i RO EXAM KEY .
| |
| I Enam Number NRC RO Rev u Esam Title NRC RO EXAM Select the statement that describes an operator action required for the following plant conditions:
| |
| 56 Reactor power: 75 % '
| |
| Suppession pooltempemture: 105 degrees F and rising Suppression poollevel: 19 feet 8 inches SRV IB21 F047D: Failed open
| |
| : a. If the SRV cannot be closed within five minu place the reactor mode switch in SilOTD0%H.
| |
| : b. If suppression pool temperature exceMs 120 segrees F., arm and depress the manual scram pushbuttons..
| |
| : c. Place the reactor mode switch in SIIUTDOWN.
| |
| : d. Reduce suppression pool tempera:ure to less than 100 degrees F within I hour.
| |
| ANSWERt
| |
| : c. Place the reactor mode swit(.h in SIIUTDOWN.
| |
| IDNO: LP # OILI#
| |
| 32 IlLO 538 9 J
| |
| PROCEDURE NUMBER: OTHER:
| |
| TRM 3 4 4 COND B LEVEL 3 AOP@35 NRC KAt RO: I SHct 239002 A2.03 4.1l 4.2 COMMENTS: 7/97 new RO T2 G1 SRO T2 01 (NOTE: looks and sounds like Tech Specs are required, but should know the 105 degree limit)
| |
| T 56
| |
| ~._ .
| |
| | |
| RO EXAM KEY Esem Number NRC-RO Rev. 0 Exam Title NRC RO EXAM
| |
| +
| |
| $7 Consider the following plant conditions:
| |
| Resctor power: 45%. _l Generatorload: 410 MWe l Recirculation flow control: Loop Manual i
| |
| SELECT the plant response to a continuous rui.oack of the load ret demand signal to zero in the Electro-llydraulic Control (EllC) system.
| |
| : a. Turbine control valves (TCVs) throttle closed, '
| |
| bypass valves (BPVs) remain closed, reactor pressure increases, ,
| |
| reactor scrams on high pressure or high neutron flux.
| |
| : b. Bypass valves (BPVs) throttle open, reactor pressure decreases, MSIVs isolate on low steam line pressure, reactor scrams (,n the MSIV closure.
| |
| < c. Turbine control <alves (TCVs) throttle closed, bypass valv(s (BPVs) throttle open to compensate, once the BPVs are fully open reactor pressure increases causing a scram on high pressure or neutron flux.
| |
| d, Turbine control valves (TCVs) throttle cloud, ,j bypass valves (BPVs) throttle open, reactor pressure remains fairly cc,nstant, reactor power increases slightly due to reduced feedwater heating.
| |
| ANSWER:
| |
| : c. Turbine control valves (TCVs) throttle closed, bypass . valves (BPVs) throttle open to compensate, once the BPVs are fully open reactor pressure increases causing a IDNO: LP # OBJ #
| |
| 28 IILO-059 7 PROCEDURE NUMBER: OTHER:
| |
| COP @80 LEVEL 4 NRC KA: ' RO: SRO:
| |
| 741000 K4.02 3.3 3.3
| |
| [-
| |
| 37 l
| |
| l
| |
| | |
| RO EXAM KEY Eram Number NRC-RO Rev. O Etam Title NRC RO EXAM COMMENTS: 7/97 new RO T2 G1 SRO T2 ul
| |
| .h i
| |
| 38
| |
| | |
| RO EXAM KEY Esam Number NRC-RO Rev 0 Exam T.tle NkC RO EXAM Which of the following condit ons/ i actions involving the Feedw:ter Level Control System, would 58 cause actual vessel level to lower.
| |
| : n. Loss of one feed flow input at 100% power and in automatic three element control.
| |
| : b. Loss of a feed flow input at 5% power and in automatic startup level control.
| |
| : c. Switching from the "A" level instrument (indicating 36") to the *II" instrument (indicating 39") when in three element automatic control at 100% power.
| |
| : d. At 100% power, placing one FRV in manual control and opening it manually, with the other FRVs in three element, automatic control.
| |
| ANSWER:
| |
| : c. Switching from the '' A" level instrument (indicating 36") to the "B" instrument (indicating 39") when in three element, automatic control at 100% power.
| |
| IDNO: LP # 011J #
| |
| 40 llLO-060 8 PROCEDURE NUMSt.R: OTHER:
| |
| AOP 0006 LEVEL 3 SOP-0007 LOTM-34 ,j 259001 K1.08 i 3.6i 3.7 _
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 59 i
| |
| | |
| l I
| |
| A. Detailed System Flowpath The Feedwater Level Control System (Figure 1) uses either a single element or three element control circuitry to position the three Feedwater Regulating (FW REG) Valves (C33-FOOIA, FOOlB, and FOOIC) during all modes of power operation. The single element control circuit uses only the input from the selected reactor vessel water level instrument. Single element is used at low Rx power (<l5%) because steam flow and feed flow instrumentation is not very accurate at low flow rates. The three element control circuit uses total steam flow, total feedwater flow, and Rx water level signals as inputs. The Single Element /Three Element Select push button located on Panel H13*P680 allows the operator to select single element or three element control.
| |
| During low flow conditions such as a plant startup, a smaller line (12") containing the Start-up FW REG Valve (C33 F002) is used to feed the reactor vessel. Due to its size and design, the startup FW REG valve is more precise at controlling low flow than the larger FW Reg. valves.
| |
| Single element control input to the Startup Controller is used to modulate the Startup FW REG Valve.
| |
| B. Design Features The following paragraphs describe the inputs into the Feedwater Control System.
| |
| : 1. Reactor Water Level (Figure 2) g The reactor water level signal for the Feedwater Level Control System is selected from either of two narrow range reactor vessellevel transmitters (A or B). The level transmitters send a current signal (4-20 mA) from the field to a signal resistor unit (SRU) in the Main Control Room (voltage is not used due to line losses). The SRU converts the 4-20 mA signal to a 1-5 volt signal. Switch C33-S1 on H13*P680 is provided to select either the Channel A or B level instrument as an input to the feedwater control circuitry (Figure 1). Narrow range level signals are indicated on meters and the selected level is indicated on a recorder on Panel H13
| |
| * P680(figure 10).
| |
| The third level transmitter, Channel C, is provided for indication and input to the Main Turbine and Reactor Feed Pump High Water Level 8 Trip (two-out of three coincidence) circuit. It also provides input into three differential alarm units (A/B, B/C, C/A). A differential of >6" between any 2 of the level input signals initiates both of the following:
| |
| * RX LEVEL CONTROL SIGNAL FAILURE annunciator (H13*P680-03-C08) e LEVEL SIGNAL DEVIATION amber indicator lamp on H13*P680-3C.
| |
| LOTM-34-6 Page 4 of 15
| |
| | |
| . Component Description
| |
| : 1. Feed Flow / Steam Flow Summer (Figure 1 - K602)
| |
| The total feed water flow and steam flow signals are input into the feed flow / steam flow summer.
| |
| Its purpose is to anticipate Rx level changes by comparing the two inputs. The feedwater flow '
| |
| input is a negative input while the steam flow input is positive. If steam flow equals steam flow, the two signals cancel each other out and the output of the summer has no effect on Rx level control if feedwater flow is greater than steam flow, the output from the summer will be negative. This will reduce feedwater flow (as discussed in the next section). If steam flow is greater than feedwater flow, the output will be positive and feedwater flow is be increased. The output of the feed flow / steam flow summer is input to the total error summer.
| |
| : 2. Total Error Summer (Figure 1 - K616)
| |
| The two inputs to the total error summer are: 1) the selected Rx level signal and 2) the output signal from the feed flow / steam flow summer. The purpose of this summer is to develop a signal to adjust FW REG valve position based on Rx level and any mismatch between feed and steam flow. Assuming stable plant conditions, the misrhatch between feed and steam will be zero and the output will be based only on Rx level. If steam flow is greater than feed flow, this positive signal will be added to the Rx level signal, resulting in the FW REG valves opening more to provide additional fecdwater, if feed flow is greater than steam flow, the negative signal will be summed with the Rx level signal, causing the FW REG valved to move in the closed direction to reduce feedwater flow. Since the feed flow / steam flow input signal is normally small in comparison to the Rx level signal, the summer is referred to as level dominant. The output of the total error summer is input to the dynamic compensator.
| |
| : 3. Dynamic Compensator (Figure 1)
| |
| The purpose of the dynamic compensator is to " smooth" the control signal such that the feedwater control system is not too sensitive to plant conditions. A loss of this component would likely cause the FW REG valves to be frequently "hundng". The output of the dynamic compensator is input to the single /three element control switch.
| |
| : 4. Single /Three Element Control Switch The operator uses this control switch, located on H13 *P680(insert 3C-see Figure 11), to select either single element or three element level control. During low power operations (<l5% power),
| |
| the instmmentation for steam flow and feedwater flow is not very accurate. Therefore, single element control is used, because it uses only the selected Rx level as the input. With power
| |
| >l5%, three element control is selected. Three element control uses the feed flow, steam flow, and selected Rx level inputs as described previously. The output of the single /three e!-ment control switch is the input to the master flow controller.
| |
| l l
| |
| LOTM-34-6 Page 7 of15
| |
| : 5. Startup FW REG Valve Flow Controller (Figure S&10)
| |
| The Startup FW REG Valve Flow Controller (C33 R602)is used to control the position of the startup FW REG valve, thereby controlling reactor water level during conditions oflow feedwater flow (<10% power). The controller can be operated in either manual or automatic.
| |
| When operated in automatic, the thumb wheel is utilized to set the desired water level (normally set at 36 inches). This desired level is compared to the selected water level channel (A or B), and a signal is developed to position the Startup FW REG Valve (C33-LVF002).
| |
| When the Startup FW REG Valve Flow Controller is operated in the manual mode, the operator utilizes the OPEN/CLOSE push buttons to control the valve position.
| |
| The controller output (demand) signal is indicated on the horizontal meter above the OPEN/CLOSE push buttons. The manual or automatic control mode is selected by depressing the asse,ciated M or A push button on the controller.
| |
| Before exceeding 10% power, one of the FW REG valves will be placed in service and the startup FW REG valve will be closed.
| |
| : 6. Master Flow Controller (Figure 6 & 10)
| |
| The Master Flow Controller is used to control reactor water u .'el by controlling the position of the three FW Reg. valves. The face of the controller contains a vertical setpoint scale, index, and setpoint adjusting thumb wheel. j The controller setpoint thumbwheel allows the operator to adjust the reactor water level setpoint.
| |
| The setpoint is read in inches (0-60") on a drum which is rotated behind an index hairline using the adjusting thumb wheel.
| |
| The controller deviation meter indicates the difference between the level setpoint and the current reactor water level. During normal operation, the reactor water level will be maintained at its setpoint and the meter will indicate zero deviation (meter point centered).
| |
| The output meter (horizontal) indicates the controller's output signal being sent to the FW REG Valve Flow Controllers. When the FW Reg. Valve Controllers are in automatic, the master controller controls their output, thereby controlling FW Reg. valve position.
| |
| The master flow controller is made up of five separate components as follows:
| |
| : 1) Setpoint M/A Station
| |
| : 2) Logic Unit
| |
| : 3) Manual Unit
| |
| : 4) Control Unit
| |
| : 5) Proportional Delay Unit LOTM-344 Page 8 of 15 i
| |
| : 2. Reactor Water High Level ;
| |
| The signals from all three water level legs are monitored for level in excess of 51 inches (Level 8). Level 8 will cause a trip of the main turbine and feedwater pumps to prevent the water _ from
| |
| . flooding the steam lines. In addition, during an ATWS condition, the level 8 trip helps prevent positive reactivity addition caused by dilution of berated water with excessive feedwater flow.
| |
| The trip coincidence for these high level trips is a two-out-of-three configuration.
| |
| : 3. Instrument Losses
| |
| : a. The complete loss of one feedwater flow signal will cause the feedwater control system to see 50% -less feedwater flow, though actual feedwater flow has not changed. This results in a iarge steam flow / feed flow mismatch, ne Feedwater Level Control System will try to increase feed flow (by opening the FW !
| |
| REO Valves) to reduce the steam flow / feed flow mismatch. This will cause reactor water level to increase. Since reactor level takes precedence in this logic system over steam flow / feed flow mismatch (level dominant), the level detectors will sense this increase in l ,
| |
| level, compare it to the level setpoint and create an error signal which will begin to close the FW REO Valves back to their original position.
| |
| t r
| |
| However, due to the gross mismatch between steam flow and feed flow, the error signal from the level circuit cannot compensate quickly eno@h to preclude reactor water level from reaching 51 inches. When this level is reached, the main turbine will trip, the feedwater pumps will trip, and a reactor scram will occur.
| |
| 'b. Loss of a Steam Flow Signal A loss of one steam flow signal will cause the Feedwater Level Control System to sense a 25% reduction in steam flow (though actual steam flow has not changed). The steam flow / feed flow summer will see this mismatch and call for a reduction in feed flow.
| |
| The FW REO Valves will start to close. Steam flow has not actually changed, so reactor water level will start to decrease. Since reactor level is dominant the level circuit will now sense the lower level and produce an error signal which will tend to reopen the valves. Reactor water level will stabilize at a lower level as the level error signal compensates for the steam flow / feed flow mismatch. .
| |
| A total loss of steam flow signal will result in a reactor scram on low reactor water level.
| |
| : c. Level Control Signal Failure .
| |
| m Failure of a level detector will cause the Rx FW Level Control Signal Failure annunciator (H13*P680-03A-B08) to alarm when the difference between level signals (A, B, C) exceeds 6 inches. The affects on the feedwater control system will depend upon whether LOTM-34-6 ~ Page 12 of15
| |
| | |
| the failure is a gross failure or simply a greater than 6 inch mismatch, which level detector has failed, and whether or not the failed detector was selected (A or B). A summary oflevel signal failure consequences is as follows:
| |
| * No automatic actions (other than the alarm) occur for failure of Level Channel
| |
| . No automatic actions will occur for failure of Channel A or B ifit is not selected as the level input to the Feedwater Level Control Logic.
| |
| . If the level channel selected for the Feedwater Level Control logic fails, the following will occur:
| |
| - Signal fails high (greater than 6 inch mismatch but not a gross failure): RPV level will decrease and may settle out at a lower level without any automatic protective actions occurring on low level.
| |
| - Signal fails high (gross failure): RPV level will continue to decrease until operator action is taken, or until low level trip setpoints are reached.
| |
| - - Signal fails low (greater than 6 inch mismatch but not a gross failure): RPV level will increase and may settle out at a higher level prior to reaching level 8.
| |
| - Signal fails low (gross failure): RPV level Sill continue to increase until operator action is taken, or until level 8 is reached.
| |
| Upon receipt of abnormal level indications and failure alarms, the operator should note which level channel has failed. If the failed channel is the selected channel, the operator should select the alternate channel by depressing the appropriate A/B SELECT push button on panel H13*P680. When this is done, the failed channel will have no further affect on the feedwater control system.
| |
| Instrument bus VBN-PNLOlB1 provides power to the valve control signal failure circuitry. A loss of this instrument bus will result in a Reactor Scram due to the loss of Feedwater Level Control. See AOP-0042 (Loss ofInstrument Bus) for additional details.
| |
| Ill. CONTROL AND INDICATIONS .
| |
| A. Controls and Indications Summary Unless otherwise indicated, all controls and indications are located on Panel H13*P680(Figures 10&11).
| |
| LOTM-34-6 Page 13 of 15
| |
| | |
| . . - ~ . .-
| |
| RO EXAM KEY Esam Number NRC-RO Rev. O Exam Tit!< NRC RO EXAM 59 The plant is at 80% pov';.. Two condensate pumps (A and B), and two feedwater pumps (B and C) are in service. If a loss of both condensate pumps occurs, FEEDWATER PUMP
| |
| : a. "B" trips 15 seconds after suction pressure decreases to 270 psig.
| |
| : b. *B" trips 40 seconds after suction pressure decreases to 270 psig.
| |
| : c. "C" continues to run indefinately on heater drain pump discharge pressure,
| |
| : d. "C" trips 60 seconds after suction pressure decreases to 270 psig.
| |
| ANSWER:
| |
| : a. "B" trips 15 seconds after suction pressure decreases to 270 psig.
| |
| IDNO: LP# OBJ #
| |
| 304 IlLO-012 4 PROCEDURE NUMBER: OTHER:
| |
| ARP-P680-3-B03 LEVEL 2 NRC KA: 1 RO: [SRO:
| |
| 259001 A2A3 ._ 3.6 3M J 256000 K1.02 3.3 3.3 209001 K1.05 3.2I 3.2,_
| |
| COMMENTS: 7/97 new RO T2 G1 FEEDWATER SRO 12 G2 l RO T2 G2 CONDENSATE
| |
| ( SRO T2 G3 1
| |
| 60'
| |
| | |
| ~
| |
| -b ~ ALARM NO,- 0106 RX FW PUMPS IH13*P680 / 03 A /- B03 !
| |
| LOW SUCTION PRESS 1
| |
| INITIATING DEVICES SET POINTS
| |
| : 1. Feedwater Pump Suction Transmitter 1. 295 psig ICNM PT70A(B)(C) .
| |
| - AUTOMATIC ACTIONS
| |
| : 1. None NOTE ,
| |
| A RFP trip will result if suction pressure decreases to 270 psig. (A, 10 sec. TD; B,15 sec. TD; C,20 sec. TD) i OPERATOR ACTIONS j
| |
| : 1. Verify one condensate pump in service for each running RFP, If no't, start a condensate pump per SOP-0007 CONDENSATE SYSTEM.
| |
| ; 2. Verify thelCNM FV114 CNDS MIN RECIRC is closed. If not, place tpe ICNM-H/Al14 MAIN '
| |
| CONDENSER RECIRC in MANUAL and close valve.
| |
| : 3. Venfy the IFWS MOV109 RX FWP BYP V FROM is closed.
| |
| 2 4 If necessary to reduce feedwater flow requirements, reduce reactor power with Recirc Flow Control.
| |
| LONG TERM ACTIONS
| |
| : 1. Verify the RFP A,B & C Manual Suction Valves are fully open; (CNM-V208, CNM-V198, CNM-V189).
| |
| 2.- If RFP Suction Relief Valve is stuck open, attempt to close.
| |
| -POSSIBLE CAUSES
| |
| : 1. Insufficient number of Condensate Pumps running.
| |
| : 3. Manual Suction Valve het fully open. ,
| |
| 1
| |
| : 3. RFP Suction Relief Valve lifting
| |
| : 4. Flow cantrol and bypass' malfunction in Condensate Demineralizer System.
| |
| ARP - 680 - 03 REV - 11 PAGE 20 OF 70 4
| |
| | |
| _ - _ . --- . . - . - . _ . - = .
| |
| RO EXAM KEY Esam Number NRC RO Rev. 0 Esam Title NRC RO EXAM 60 A plant startup is in progress. Reactor power is being held at 1% power for the 900 psig Drywell walkdown when the Startup Feedwater Reg. Valve drifts fully open.
| |
| Which of the following actions / signals will occur as a result of this failure?
| |
| (NOT11: Assume no operator action.)
| |
| : a. Reactor scrams on high reactor water level.
| |
| : b. Reactor feedwater pumps trip on high reactor level.
| |
| : c. Reactor water level remains unchanged due to compensation by the Long Cycle Cleanup Level Controller (CNM 104).
| |
| : d. Reactor water level stabilizes at a new higher level.
| |
| ANSWER:
| |
| : b. Reactor feedwater pumps trip on high reactor level, i
| |
| IDNO: LP# Olu #
| |
| 348 STM 107 10 PROCEDURE NUMBER: OTHER:
| |
| AOP-0006 LEyL 3 SOP 4X)09 NRC KAt l RO: 1 SRO:
| |
| 259002 K3.02,___ __3.7. _ _3. 7 259002 K3.01 3.8 3.8 _j COMMENTS: 7/97 new LOTM 34 5 p.11 of 13, LOTM 3 4, p. 26 of 31, Table 8 RO T2 G1 SRO T2 G1 61
| |
| | |
| -RO EXAM KEY-Esam Number NRC-RO Rev. 0. Esam Title NRC RO EXAM 61 Given the following plas.t conditions:
| |
| A LOCA has occurred.
| |
| Reactor water level is 50 inches.
| |
| Drywell Pressure is 1.12 psid.
| |
| All radiation monitors are indicating normal for plant conditions.
| |
| "A" Standby Gas Treatment System (SGTS) is running and the "B" SBGT system has been shutdown and placed in STANDBY.
| |
| Whici. of the following will be the statu. of SBOT systems "A" and "B" five (5) minutes after ENS-SWGI A receives a degraded Bus Undervoltage Signal? (All associated systems respond normally)
| |
| : a. Both systems are shutdown.
| |
| : b. Both systems are running.
| |
| : c. "A" system is shutdow's and "B" system is running.
| |
| : d. "A" system is running and "B" system is shutdown.
| |
| ANSWER:
| |
| : b. Both systems are running.
| |
| .)
| |
| IDNO: LP # OBJ #
| |
| 272 IILO-033 12 PROCEDURE NUMBER: OTHER:
| |
| SOP.0043 LEVEL 4 SBGT DWGS
| |
| ~~URC KAt NO: l SRO:
| |
| 261000 K6.01 2.9 3 COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 (10 sec for O/G 30 sec for sequencing re-eng from D/G - LOCA still present B s/u due to low flow in Al 62
| |
| | |
| RO EXAM KEY Enam Number NRC-RO Rev. O Exam Title NRC RO EXAM 62 The following conditions exist:
| |
| The Div 2 standby diesel generator is loaded and lu parallel with bus IENS'SWGlB through the normal breaker, A LOCA signal occurs.
| |
| Which of the following describes the effect of the standby diesel generator and bus IENS'SWGIB7
| |
| : a. He normal bus supply breaker will open and the <tice.cl generator will supply bus loads.
| |
| : b. De normal bus supply breaker and diesel generator (,utput breaker will open, then after loads are stripped, the diesel generator output breaker will reclose .
| |
| : c. ne diesel generator output breaker will open and cannot be closed as long as bus voltage is supplied by the normal or attemate feedeu until the LOCA signal is reset.
| |
| : d. ne diesel generator output breaker will remain closed in parallel operation with the bus.
| |
| I ANSWER:
| |
| : c. He diesel generator output breaker will open and cannot be closed as long as bus voltage is supplied by the normal or alternate feeders until the LOCA signal is reset J
| |
| IDNO: LP# Olu #
| |
| 199 IILO-037 4 PROCEDURE NUMBER: OTHER:
| |
| SOP-0053 'EVEL 4 NRC KAt L HO: l SRO:
| |
| 264000 A2.10 3.9 4.2 _
| |
| COMMENTS: 7/97 new RO T2 G1 SRO T2 G1 i
| |
| e 63 t
| |
| | |
| . . _ . . ~ - . - -_ ._ - _ _ . - -. - . ._. _ . - - - - . -
| |
| RO EXAM KEY Esam Number NRC-RO Hev, 0_ Exam Title NRC RO EXAM 63 The plant is operating at 60% power for rod pattern adjustment. Rod 22-43 is to be inserted from notch 48 to notch 42. A few seccnds after the rod is insened to notch 42 a ROD DRIFT annunciator is recieved. While conducting your immediate actions, you observe rod 22-43 passing notch 46 and observe it to stop at notch 48.
| |
| The Control Rod Movement Sequence withdraw limit is notch 48. Reactor power has returried to 60%. The correct actions will be:
| |
| : a. Place the Mode Switch in SilUTDOWN.
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whether the rod ,
| |
| has a stuck collet.
| |
| : c. Since power is below IIPSP the Rod Withdraw Error analysis is affected and the rod should be inserted to notch 42 without delay,
| |
| : d. Notify a Reactor Engineer, declare rod 22 43 inoperable, and adjust the pattern as needed for flux shaping with rod 22-43 full out since it will not remain insened.
| |
| ANSWER:
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whether the rod has a stuck collet.
| |
| IDNO: LP # OBJ # ,)
| |
| 49 STM 052 4 PROCEDURE NUMCER: OTHER:
| |
| ARP 680-07-802 LEVEL 3 NRC KA: RO:"I SRO:
| |
| 201003 K4.07 3.2 , 3.2 201003 A2.03 3.4 3.7 COMMENTS: 7/97 new RO T2 G2 SRO T2 G3
| |
| [ Required to determine TS 3.1.3 Condition A applicablityl 64
| |
| | |
| I RO EXAM KEY Etam Num: r NRC-RO Rev. 0 Esam Title NRC RO EXAM 64 During a weekly control rod operability test, the ATC operator is unable to insert rod 10-47 from position 48. CRD parameters are normal and no rod blocks are present. Which of the following actions should be taken:
| |
| : a. Depress "IN TIMER SKIP" to insert control rod 10-47 to the required position.
| |
| : b. Attempt the double clutching method to free the rod; if unsuccessful, raise drive water pressure to a maximum of 350 psid to move control rod 10 47.
| |
| : c. Insert control rod 10-47 to the required position by incrementally raising drive pressure to a maximum of 350 psid.
| |
| : d. Declare rod 10-47 INOPERAllLE, SCRAM and disarm it then verify compliance with the remaining constraints of Technical Specifications regarding inoperable contr01 rods.
| |
| ANSWER:
| |
| : c. Insert control rod 10-47 to the required position by incrementally raising drive pressure to a maximum of 350 psid.
| |
| IDNO: LP# Olki#
| |
| 55 STM-052 2 PROCEDURE NUMBER: ,joTHER:
| |
| SOP-007t SECT 5.7 LEVEL 3 201003 A2.01 3.41 3.6 _.
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G3 65
| |
| | |
| - l H
| |
| 5.6.2. Perform High Drive Water Pressure Method as follows:
| |
| NOTE ,
| |
| The drive water headerpressure control valve is in _
| |
| parallel with an inline reliefvalve. The manual -
| |
| adjustment ofhigh drive waterpressure must be ;
| |
| below the reliefvalve setting. approximately SSO psid t
| |
| : 1. Raise drive water header pressure to 300 psid.
| |
| : 2. Notch out the Control Rod per Step 5.1.1.
| |
| : 3. IF the Control Rod notch withdrawal was not successful, THEN perform the Double Clutching method per Step 5.6.1.
| |
| : 4. E necessary, THEN raise drive water pressure to 350 psid and repeat Steps 5.6.2.2 through 5.6.2.3.
| |
| : 5. WHEN the Control Rod has moved OR the attempt has been made at 350 psid, THEN return drive water header pressure to the normal operating range.
| |
| 5.7 Notching Control Rods Out of Positions Other Than '00'
| |
| ')
| |
| NOTE The drive water headerpressure control valve is in parallel with an inline reliefvalve. The manual adjustment ofhigh drive waterpressure must be below the reliefvalve setting (approximately SSO psid). .
| |
| 5.7.1. Raise drive water header pressure to 300 psid.
| |
| 5.7.2. Notch out the Control Rod per Step 5.1.1.
| |
| 5.7.3. E the Control Rod notch withdrawal was ng successful, THEN raise drive water header pressure to 350 psid.
| |
| 5.7.4. Notch out the Control Rod per Step 5.1.1.
| |
| 5.7.5. WHEN the Control Rod has moved'OR the attempt has been made at 350 psid, THEN retum drive water header pressure to the normal operating range.
| |
| SOP-0071 REV.9 PAGE 13 OF 26
| |
| | |
| RO EXAM KEY Etam Number NRC-RO Rev. O Exam Title NRC RO EXAM 65 The plant is operating at 100% power. Both Recirc Flow Control Valves are in Flux Manual (Loop Auto) at 67% valve position. A leak in the Drywell has caused Drywell Pressure to increase to approximately 1.75 psid. Following t1 high drywell pressure signal, the ''B" Reactor Feed Pump Trips and level decreases to + 14.3 inches and stabilizes. Which of the following describes the response of the Recirc Flow Cont.o1 Valve?
| |
| Flow Control Valves will:
| |
| : a. runback to 22 % valve position,
| |
| : b. go to " min" position.
| |
| : c. move to a position to provide 60 % core flow,
| |
| : d. remain at 67 % valve position.
| |
| ANSWER:
| |
| : d. remain at 67 % valve position.
| |
| IDNO: LP # OBJ #
| |
| 244 ilLO-005 4 PROCEDURE NUMBER: ,jdTHER:
| |
| ARP P680-4-803 LEVEL 4 ARP-P680-4-B09 ARP-P680-4404 ARP-P6804C10 NRC KAt RO: l'MO:
| |
| 202002 A2.o8 3.3! 3.3 _
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G1 66
| |
| | |
| I I
| |
| RO EXAM KEY Exam Number NRC RO Rev. O Exam Title NRC RO EXAM The Division til diesel has run for 30 minutes unloaded following an inadvertent start of the diesel.
| |
| 66 What is the correct method to conduct a normal diesel shutdown?
| |
| : a. Parallel the diesel with offsite, load it to > 1950 KW for > 30 minutes, then unload and cool it down for at least 4 minutes and stop the engine.
| |
| : b. shut down the engine. It is adequately cooled down,
| |
| : c. Parallel the diesel with offsite, load it to 2500 - 2600KW for I hour, then trip the output breaker and shut down the engine,
| |
| : d. Parallel the diesel with offsit load it to > 1950 KW for > 30 minutes, then unload and stop the engine.
| |
| ANSWER:
| |
| : a. Parallel the diesel with of' site, load it tn > 1950 KW for > 30 minutes, then unload and cool it down for at least 4 minutes and stop the engine IDNO: LP# OILI#
| |
| 268 IILO437 07 PROCEDURE NUMBER: OTHER:
| |
| SOP-0052 LEVEL 3 NRC KA: EO:I~SRO: '
| |
| 264000 A4.04 3.71 3.7 _
| |
| COMMENTS: 7/97 new SOP 0052. Precautions and Limitations 67
| |
| | |
| l 1.0 PURPOSE _ ,
| |
| 1.1 To provide instructions to River Bend personnel for the operation of the HPCS Diesel Generator and its auxiliaries.
| |
| 3.0 PRECAUTIONS AND IIMrrATIONS 2.1 Diesel Engine 2.1.1 If an engine alarm results from high crankcase pressure, DO NOT open any hand hole or top deck cover to make an inspection until the engine has been stopped and allowed to cool off for at least 2 hours.
| |
| 2.1.2 Do not attempt to restart an engine following a shutdown due to a cranker.:e pressure alarm until the cause of the trip has been determined.
| |
| 2.1.3 After unloading the diesel generator allow the engine to run unloaded for 4 minutes prior to shutdown anytime the engine has been run at rated temperature.
| |
| 2.1.4 Under normal conditibns, the engine should not be run at less than 1950 KW for longer than 10 minutes. Anytime this time limit is exceeded the diesel generator shall be loaded to at least 1950 KW for 30 minutes priof to shutting down the engine.
| |
| 2.1.5 Under normal conditions the engine shall be barred over one complete revolution with the cylinder test valves open before starting. The engine' need not be barred over if the start is within 4 hours of the last shutdown.
| |
| If any fluid discharge is observed from any cyljnder, find the cause and make the necessary repairs prior to operating the engine.
| |
| 2.1.6 If a start failure occurs, the failure must be corrected and the SAFETY RESET (S7) pushbutton depressed in order to reset the start circuits.
| |
| 2.1.7 Except for emergencies, including automatic starting, the engine should not be started unless:
| |
| l_. The engine has been operated in standby with the circulating oil pump, soakback (turbo) oil pump and jacket water immersion heater in operation for at least I hour;
| |
| : 2. Lube oil level is between the two round sightglasses (lower sightglass full, upper empty);
| |
| : 3. Lube oil and jacket water temperatures are at least 85*F. Where possible, it is preferable to have operated in standby for at least 24 hours (24 hour normal standby fluid temperatures are approximately
| |
| - 100*F lube oil and 125'F jacket water).
| |
| 2.1.8 The A.C. Circulating Oil Pump may be taken out of service as long as the D.C. Circulating Oil Pump is operating as back-up.
| |
| SOP-0052 REV - 12 PAGE 4 OF 58
| |
| | |
| RO EXAM KEY Eram Number NRC-RO Rev. O Exam Title NRC RO EXAM 67 River Bend is operating at 90% power with recite in Dux manual when the discharge isolation valve for the "11" recirculation pump commences to stroke closed. Which of the following statements describes the expected response of the recirculation system to this event?
| |
| The Recirc Pump *11":
| |
| : a. trips to off.
| |
| : b. transfers to slow speed.
| |
| : c. contiues to run in fast speed; core Dow and power decrease.
| |
| : d. hydraulic power unit locks up.
| |
| ANSWEll:
| |
| : a. trips to off.
| |
| IDNO: LP # Olu #
| |
| 27 IILO-005 4 PROCED'JRE NUMBER: OTHER:
| |
| SOP @03 LEVEL 2 AOP-0024 )
| |
| 202001 K4.10_ _ _3.3 _ _'I.4 l 1 202001 A3 07 3.J 3.3 _j COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 63
| |
| | |
| 1.8 During Single Loop Operation, the flow boundary of region B is reduced as identified on the SLO power / flow map. Power operation to the right of the region B line is pennitted during Single Loop Operation. Refer to the Power / Flow map shown in Attachment 2. ,
| |
| 2 SYMPTOMS 2.1 Reactor Recirc Pump Trips 2.1,1. Fart to off ..,
| |
| : 1. ATWS Signal reactor water level at Level 2, or reactor pressure at 1127 psig
| |
| : 2. 1833 F023A(B) RECIRC PUMP A(B) SUCTION VALVE less than 90% open
| |
| '3. IB33-F067A(B) RECIRC PUMP A(B) DISCH VALVE less than 90% open s
| |
| : 4. CB-5 " Push to lock" push-button depressed
| |
| : 5. ~ CB 3 or 4 open
| |
| : 6. Incomplete Sequence Relay operation
| |
| : 7. Loss of control power during start sequence
| |
| : 8. CB-5 "Stop" push-button depressed
| |
| : 9. Motor protection relays
| |
| : 10. Steam Dome /Recirc Loop Suction differential '
| |
| i Temperature low, less than 8'F for 15 seconds. (For
| |
| ; affected loop if other loop is still running in fast speed.)
| |
| l 2.1.2 ? Slow to off
| |
| : 1. ATWS Signal-reactor water level at Level 2 or reactor pressure at 1127 psig -
| |
| : 2. 1B33-F023A(B) RECIRC PUMP A(B) SUCTION VALVE less than 90% open l t A O P-0024 REV .14 P"E 4 OF 14
| |
| | |
| l RO EXAM KEY l
| |
| Exam Number NRC-RO Rev. O Exam Title NRC RO EXAM Which one of the following INDICATIONS would you expect to see as a result of a " Jet Pump 68 Riser Failure"? Assume RECIRC FLOW CONTROL is in FLUX MANU AL.
| |
| 4 A DECREASE in reactor (APRM) power, a DECREASE in the Failed Jet Pump Flow, and a DECREASE in Core DifTerential Picssure.
| |
| : b. A DECREASt!in core difrcrential pressure, an INCPIASE in Reactor Power, and an n!CR PASE in Indicated Core Flow.
| |
| : c. A DECREASE in failed Jet Pump flow, an INCREASE in indicated Core Flow, and a DECREASE in Core Differential Pressure.
| |
| : d. An INCREASE in in.l;cated core flow, an INCREASE in Failed Jet Pump Flow, and an INCREASE in reactor Power.
| |
| ANSWER:
| |
| : a. A DECREASE in reactor (APRM) power, a DECREASE in the Failed Jet Pump Flow, and a PECREASE in Core Differential Pressure.
| |
| IDNO: LP# OB.I #
| |
| IILO-005 11 95 PROCEDURE NUMBER: OTHER:
| |
| STP-053-3001 L(yEL 3 LOT'A-2
| |
| ~
| |
| I NRC K A: } R6I~! SRO:
| |
| 1202001 K6.01 1 3.51 37 __
| |
| COMMENTS: 7/97 exam Mnk RO T2 02 SRO T2 G2 O
| |
| 69
| |
| | |
| 1 1.. Jet Pumps The jet pumpt provide forced flow of coolant through the reactor vessel to yield a higher reactor power output than would be possible with natural circulation. The 20 jet pumps are located in two semicircular groups in the downcomer region, with two jet pumps and a common inlet header combined to form ajet pump ass mbly (Figure 8).
| |
| Each jet pump assembly consists of the following:
| |
| e one inlet riser and thermal sleeve, e one transition piece welded to the top of the inlet riser, e two nozzle sections, e two mixing sections, e one bracket and restrainer assembly, and
| |
| . two diffuser sections.
| |
| A thermal sleeve is welded into the vessel penetration for each inlet riser.
| |
| The thermal sleeve prevents overstressing the reactor vessel nozzle due to differences in temperature between the inlet water, vessel wall and inlet nozzle. Inlet risers are used for each jet pump assembly to permit lowering the reactor recirculation inlet nozzles to below the active fuel region.
| |
| Riser brace arms provide lateral support for the upper end of the riser and yet allow for the vertical difTerentiaf expansion between the riser and reactor vessel during heatup and cooldown. J The upper end of the jet pump is the nozzle assembly, or " rams head",
| |
| which is bolted to the upper end of the riser. The lower end of the nozzle is secured to the upper end of the mixing section by a clamp. The nozzle assembly changes the direction of the driving flow by 180 degrees, before entering the jet pump nozzle. Each jet pump nozzle actually consists of five (5) smaller nozzles which demonstrate improved efficiency over the older design single nozzle performance.
| |
| The mixer section fits into the jet pump diffuser, with four guide vanes of the diffuser mating with the lower end of the mixer section. This slip fit makes it possible to change out the mixer sections during refueling outages if required, and allows for differenti4 expansion between the riser, diffuser, and reactor vessel.
| |
| The mixer assemblies are braced to the inlet riser pipe just above the slip fit into the diffuser. The bra e ring is welded to the riser and holds the mixing assembly in place with a sliding, tapered wedge and restrainer assembly.
| |
| The jet pump diffusers are welded to adapters that are first welded to the shroud support plate. Welding the smaller adapter to the shroud support plate first aids in the alignment and welding of the diffusers. i LOTH 2-7 PAGE 9 of 30
| |
| | |
| During operation of the Reactor Recirculation System, the reactor recirculation pumps draw coolant from the downcomer annulus and discharge into the inlet risers for the jet pumps. The inlet risers direct the flow of coolant through the jet pump nozzles and into the mixing assemblies. The flow of coolant from the nozzles entrains additional coclant.from the downcomer annulus, (feedwater inlet flow and water separated in the steam separators and dryers) thus providing an increased volume of flow through the mixing assemblies. The driving flow enters the jet pump nozzles at a high pressure and is accelerated to a high velocity at the nozzle outlet. The suction flow from the downcomer annulus (known as the driven flow) enters the suction inlet at a lower pressure. This low pressure is further reduced as the flow is accelerated through the converging inlet nozzle.
| |
| The two flows merge in the mixing section where a pressure rise occu" due to a change in the flow pattem caused by the mixing action. Near the end of the mixing section, the rate of pressure rise decreases as the stream begins expanding into the control diffuser section. The diffuser slows the high velocity mixed stream and converts the dynamic head into static pressure (Figure 9).
| |
| The 20 jet pumps discharge through the baffle plate into the lower vessel l
| |
| plenmn. After mixing in the lower plenum, the flow procer '.s upward through the core, it is important to note that total flow through the jet pmnps is composed ofjet pump drive flow and flow induced by the drive flow (driven flow). Total jet pump flow is equal to jet pump nozzle flow from the recirculation system, combined with)the induced feedwater flow and rejected water from the steam separator and dryers in the downcomer region. The composite of these flows is termed total core flow. For complete accuracy, recirculation pump seal flow from the Control Rod Drive System (10 gpm), avj the CRD mechanism cooling flow (45 gpm) must be included in a calculation of total core flow.
| |
| NOTE:
| |
| During full power operation, the Jet pump driving flow comprises approximately IB of total corepow. The remaining 2B is induced by theJetpumps. Thatis a 2: 1 ratio ofdrivenpow to drivingpow and 3:
| |
| 1 ratio of total corefo w to recirc systempow.
| |
| : 3. Core Shroud The core shroud (Figures 10 through 12) is an approximately two-inch thick, cylindrical type 304 stainless steel assembly that surrounds the core.
| |
| The shroud provides a barrier to separate or divide the upward core flow from the downward flow in the downcomer annulus. The shroud also provides vertical and lateral support f'or the core plate and top guide. In addition the shroud serves to provide a floodable volume in the event of a LOCA (Figure 36).
| |
| LOTM 2-7 PAGE 10 Of 30
| |
| | |
| I RO EXAM KEY Esam Number NRC-RO Rev. 0 Eram Title NRC RO EXAM 69 Given the following cond.tions:
| |
| The Reactor Water Cleanup (RWCU) system is operating in the normal mode.
| |
| The RWCU isolation Bypass Switches (E31 SI A,B) on P632 and P642 have been placed in
| |
| " Bypass".
| |
| Select the expected efTect on the RWCU system.
| |
| : a. He RWCU system isolation on hign non-regenerative heat exchanger outlet temperature is defeated.
| |
| : b. He RWCU syste n isolation from high area temperature ONLY are defeated.
| |
| : c. He RWCU system isolation from high differential flow AND high area temperature are defeated.
| |
| : d. All RWCU system isolation signals are defeated.
| |
| ANSWER:
| |
| c ne RWCU system isolation from high differential now AND high area temperature are defeated.
| |
| IDNO: LP # OBJ # ,j 197 IILO-006 6 PROCEDURE NUMBER: OTHER:
| |
| 851ECO2AA LEVEL 2 828E445AA NRC KA: } kU['' kRO:
| |
| 204000 K1.15 1 3.1 3.2 COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 70
| |
| | |
| . - . -_. - -- _- -- . ~. . _ _ = - .
| |
| RO EXAM KEY Esam Number NRC-RO Rey, O Esam Title NRC RO EXAM 70 Which of the following describes the optimal lineup for RWCU suction flow during normal full power operation?
| |
| : a. Flow should be half from each Recirc line the bonom head suction should be closed when the reactor is pressurized,
| |
| : b. Bottom head drain line flow should not exceed I/3 of total RWCU flow, if the recire loop suctions are available,
| |
| : c. Flow should be entirely from the bottom head drain line ifit is available,
| |
| : d. Suction should be from the recirc lines if recire pumps are in slow speed, and from the bottom head drain line if recirc pumps are in fast speed.
| |
| ANSWER:
| |
| : b. Bottom head drain line flow should not exceed 1/3 of total RWCU flow, if the recirc loop suctions are available.
| |
| IDNO: LP# OILI #
| |
| l 352 IILO-006 2 PROCEDURE NUMBER: OTHER:
| |
| SOP 6 LEVEL 3 NRC KA: l RO: l SRO: ~)
| |
| 204000 K1.07 l 2.91 3 _j COMMENTS: 7/97 new SOP-0090, Rev 10, p. 6 of 83,2.27 I
| |
| l 71
| |
| | |
| 2.11 Ensuring G33 F101, RWCU BOTTOM HEAD DRAIN remains open at all times, within -
| |
| the limitations of Step 2.12, provides adequate flow for RPV Bottom Head cleanup and temperature detection at the Bottom Head Drain Line.-
| |
| 2.12 With the Reactor pressurized, RWCU Bottom Head Drain Flow should not exceed approximately 1/3 of RWCU total flow when both the Bottom Head and the Recire loops are suction paths. This will assure optimization of Rx chemistry when the F/D(s) are in service. ,
| |
| 2.13 Under some conditions, such as the Reactor de pressurized, an adequate source of s ster to the RWCU Pump suction may not be available if suction is only from the Bottom Head Drain. If proper flow can not be achieved, or there are any indications of pump cavitation, the RWCU Pump should be secured.
| |
| 2.14 When the Reactor is de-pressurized, indication of Bottom Head Drain Flow, as indicated on ERIS and H13 P680, will be erroneously low. His is due to the difference in elevation of the taps for the flow instmmentation.-
| |
| 2.15 With a flow path available through at least one RECIRC Loop and the Reactor de pressurized, flow through'the Bottom Head Drain will be approximately 0 gpm.
| |
| ' - 2.16 Isolating the RWCU System by closing tim containment isolation valves may result in difficulty re-establishing system lineup if the Reactor is at temperature and pressure. His I occurs because the piping downcream of the containment isolation valves cools down and depressurizes. Systems Engineering support may p necessary for resolution, -
| |
| situation is experienced.
| |
| 2.17 Temporary hoses or piping should not be cross connected with other plant systems except I as authorized by approved plant procedures.
| |
| l l 2.18 HVR UC2 Aux Bldg Unit Cooler should be operable prior to startup of a RWCU pump.
| |
| 2.19 Backwashing filter demins as soon as possible after they are taken out of service I minimizes the release of gaseous radioactive waste products into containment and the annulus.
| |
| [
| |
| i l
| |
| SOP-0090 REV .19 PAGE 5 OF 113
| |
| , s.+i . ,v.,--.m3- + ne x , -- _ ,,%-,- --. . .----m ., --- ,m,w ,%-#n-- .- < au- wwg=
| |
| | |
| l RO EXAM KEY Exam Number NRC-RO Rev. O Exam Title NRC RO EXAM
| |
| (
| |
| 71 The RilR S/D Cooling isolation Valve Enable / Disable switch on the local panel (P001) has two positions," Enable / Disable".
| |
| Which of the following describes when the switch is REQUIRED to be in " Disable" and the effect on the operation of Shutdown Cooling when it is in this position?
| |
| The RilR S/D Cooling isolation Valve Enable / Disable switch is placed in " Disable" when:
| |
| : a. reactor pressure is greater than 135 twig .nd prevents operation of the RilR Shutdown Cooling Inboard Isolaaon Valve (F009) from the Main Control Room.
| |
| : b. reactor pressure is greater than 135 psig and prevents operation of the RilR Shutdown Cooling Outboard Isolation Valve (F00h) from the Main Control Roe
| |
| : c. evacuating the Main Control Room to allow local opereL. of the RilR Shutdown Cooling Inboard Isolation Valve (F009).
| |
| : d. evacuating the Main Cemol Room to allow local operation of the RHR Shutdown Cooling Outboard Isolauon Valve (F008).
| |
| ANSWER:
| |
| : b. reactor pressure is greater than 135 psig and prevents operation of the RilR Shutdown Cooling Oetboard Isolation Valve (F008) from the Main Control Room.
| |
| J IDNO: LP# OBJ #
| |
| 422 IILO-021 6b PROCEDURE NUMBER: OTHER:
| |
| SOP 0031 LEVEL 4 NRC K Ai l RO: SR E 205000 K6.01 l 3.3 3.4 _
| |
| COMMENTS: 1/97 NRC exam 4
| |
| | |
| - . ~ ~ . - .. . _
| |
| RO EXAM KEY Essm Number NRC-RO Rev. O Exam Title NRC RO EXAM 72 The following conditions exist:
| |
| RilR shutdown cooling loop A is in operation. Reactor water level is 75 inches.
| |
| Under these conditions, misoperation of which of the following valves from the main control room has the potential to inadvertantly drain the reactor vessel?
| |
| : a. Suppression Pool Suction Valve IE12'F004A.
| |
| : b. Shutdown Cooling Suction Valve IE12*F006A.
| |
| : c. Test Return To Suppression Pool Valve IE12'F024A.
| |
| : d. Shutdown Cooling Outboard Isolation Valve IE12'F008.
| |
| ANSWER:
| |
| : c. Test Return To Suppression Pool Valve IE12'F024A.
| |
| IDNO: LP# OBJ #
| |
| 182 llLO-021 6 PROCEDURE NUMBER: OTHER:
| |
| SOP-0031 LEKL 2
| |
| ~
| |
| I NRC KA! l R.0: I SRO:
| |
| l205o00K5.o2 1 2.8! 2.9 _
| |
| COMMENTS: 7/97 new RO T2 G2 SRO T2 G2 E
| |
| 73
| |
| | |
| i 2.1.10 ~Ihe respective unit cooler (s) shall be in senice Mor to starting RHR pump (s) for shutdown cooling. (Ref. 7.16 & 7.: 7) 2.1.11 To prevent overpressurization of the RHR Heat Exchanger Senice Water channel in the event of a tube rupture, do not isolate the sen' ice water (i.e.,
| |
| inlet and outlets closed with no dram or vent path established) without racking out the associated RIIR Pump Breaker. (Ref. 7.21) 2.1.12 Minimize operation of RHR Pump A w'th . system now less than 3500 gpm and E12-MOVF(M8A RHR A HX BYPASS VALVE less than 5% open due to unacceptable vibration levels.
| |
| 2.2 Opemtions with a potential for dmining the Reactor Vessel (OPDRV) 2.2.1 To prevent draining the xactor to the suppression pool. ENSUPE E12-F064A(B), RHR PUMP A(B) MIN FLOW TO SUP PL and E12-F024A(B)
| |
| TEST R1N 1D SUP PL are CLOSED before o RHR PUMP E12 F006A(B) RH A(B) R PUMP A(B) SDC SUC110N7.6VALVE. & (Ref. pe 7.35).
| |
| 2.2.2 With E12-VF067, RHR PUMP C SDC SUGON ISOL VLV o a tential exists to drain the RPV via E12 MOVF021, RHR P C 'IEST p>IN I TO SUP PL or E12-MOVF105, RHR PUMP C SUP PL SUC110N VALVE. ENSURE F021 and F105 are closed prior to opening F%7.
| |
| 2.2.3 1he following RHR shutdown cooling interlocks shall not be bypassed without senior plant management review and appmval. Contingency methods to supply sufIicient makeup water if a draining event occurs while the interlocks are bypassed must be in place., (Ref. 7.51)
| |
| NOIE Only one isolation whv (El2-inJ08 or E12 FD09) is r? pared to be operable.
| |
| I n flea $on o x ed pwcedures to pment sptem7 sol and loss of SDC diring rnainte.unce cr testirgispermissible. (Ref 7.52)
| |
| : 1. Low reactor water level isolation of the S C suction valves (E12 F008 RHR SHUIDOWN COOLING OUTBD ISOL VALVE and E12-F009 SHUTDOWN COOLINO INBD ISOL VALVE). (Ref. 7.52)
| |
| : 2. Interlocks beween the RHR suction valves (E12-F004 RHR PUMP SUP PL SUCITON AL and IE12'F006 RHR PUMP SDC SUCTION VALVE). (Ref. 7.52) l SOP-0031 REV - 17B PAGE 5 OF 1Q2
| |
| | |
| RO EXAM KEY Esam Number NRC.RO Rev. 0 Etam Title NRC RO EXAM 73 1he plant is operating at 75% power.1he Control Room Operator places the Outboard MSIV positive Leakage Control System switch to OPERATE. Which of the following will prevent the Outboard MStV Positive Leakage Control System from initiating 7
| |
| : a. A LOCA signal on either high drywell pressure or low reactor water levelis not present.
| |
| : b. The required main steam line pressure and reactor pressure requirements have not been met.
| |
| : c. The post LOCA 20 minute timer has not timed out.
| |
| : d. All Main Steam isolation Valves have not been fully closed.
| |
| ANSWER:
| |
| : b. The required main steam line pressure and reactor pressure requirements have not been met.
| |
| IDNO: LP# 011J #
| |
| 190 LOTMS 4 PROCEbORE NUMBER: OTHER:
| |
| ARP40117 G05 LEVEL 3 ARP401 17.G06 j SOP 4034 F ~NRQlAi' I ~' HOI h 'SRE:
| |
| 239001. K 13 13 __, _ _2,6 ,, _ _ 2 ._8 239003 K1.01._ ._ _ _ 3 . 3,, _ _ _ 3.4 239003 K4 03 2.94 32 g COMMENTS: 7/97 new RO 12 03 SRO T2 G2 N
| |
| | |
| RO EXAM KEY rsam Number NRC.RO lley. 0 1: sam Title NRC RO EXAM 74 1he plant was operating at 100% power when a large steam leak on the MSR A reheat steam line required reheat steam to be isolated. I ive minutes later the TURBINE 111011 V11 RATION annunciator on P870 alanns. You check the recorder and observe bearing #$ readmg 9 mils, bearing # 6 reading 13 mils, and bearing # 4 reading 5 mits. These readings appear to be steady.
| |
| You should:
| |
| : a. Continue to monitor the vibrations and initiate a reactor scram and trip the turbine if 15 mils is reached.
| |
| : b. Commence a rapid load reduction then take the turbine offline (scram and turbine trip)if vibrations cannot be reduced below 10 mits within 14 minutes.
| |
| : c. Immediately scram the reactor and trip the turbine.
| |
| : d. Monitor bearing temperatures and scram the reactor and trip the turbine if bearing temperature exceeds 240 degrees F.
| |
| ANSWl:lt:
| |
| : c. Immediately scram the reactor and trip the turbine.
| |
| IDNO: LP # Olti #
| |
| IILO.025 9 71 y PROCEDURE NUMBER: OTHER:
| |
| AOP 0002 LEVEL 3 ARP 870-54 D06 NRC K Ai I ~ RO: ! SROi 245000 K4.05 i 2.9! 3 ,j]
| |
| COMMLNTS: 7/97 new RO T2 G2 SRO T2 G2 75
| |
| | |
| t l
| |
| RO EXAM KEY Esam Number NRC RO Rev. O Esam Title NRC RO EXAM 73 A startup of the Main Turbine is being performed. 'ihe Main Turbine is at 60 percent of rated speed, when a loss of 125 VDC Trip Circuit Power is experienced. WillCil ONE (1) of the following describes the required operator action (s)?
| |
| : a. Enter AOp 0002, Main Turbine and Generator Trips, due to trip of the Turbine.
| |
| : b. Verity that 24 VDC ETS power is available and continue the startup of the Main Turbiae I AW sop.0080, otherwise manually trip the Main Turbine.
| |
| : c. Allow the Main Turbine to accelerate to greater than 90 percent of rated speed,
| |
| ; at which time the 125 VDC Trip Circuit is no longer required because the PMG is supplying the trip circuitry,
| |
| : d. The start up of the Main Turbine may continue, but at least one 125 VDC bus must be restored prior to synchornizing the generator to the grid.
| |
| ANSWER:
| |
| : a. Enter AOp 0002, Main Turbine and Generator Trips, due to trip of the Turbine.
| |
| IDNO: LP# OltJ N 356 STM 110 26 PROCEDURE NUMBER: ,j oTHER:
| |
| AOP-0002 LEVEL 3 245000 K6.06 3 3.2 _
| |
| COMMENTS: 7/97 DOW LOTM 27 6, p. 60 of 63 AOP 0002 2,1,12 Rev 7 e
| |
| 76
| |
| | |
| RO EXAM KEY Exam Numtwr NRC.RO Rev. 0 Esam Title NRC RO EXAM 7(, An MSIV closure resulted in a reactor scram 1hc pressure transient caused a small steam leak in the drywell. De following conditions exist:
| |
| . Reactor pressuie is at 900 psig.
| |
| . Reactor Level is at .80 inches wide range Drywcll pressure is 2.1 psid
| |
| . Containment pressure is 0.3 psig
| |
| . lowest recorded ENS'SWGI A Bus voltage was 3952 volts.
| |
| Which one of the following would be in service as indicated?
| |
| (NO OPERATOR ACTION TAKEN) i
| |
| : a. DIV I D/G running utf.oaded.
| |
| : b. DIV ll SSW with flow through the "B" Containment Unit Cooler.
| |
| : c. Drywell units coolers running with no cooling flow,
| |
| : d. LICS injecting to the RPV.
| |
| ANSWER:
| |
| : a. DIV I D/G running unloaded.
| |
| J lDNO: LP# OILI #
| |
| l 261 IILO-037 4 PROCEDURE NUMBER: OTHER:
| |
| SOP-0053 LEVEL 4 f ARP.877 32-H03 262001 K1.01 1 3 BI 4.3 _
| |
| COMMENTS: 7/97 new RO T2 G2 l
| |
| SRO T2 G1 THIS OUESTION IS TO OETERMINE IF THE CANDIDATE UNDERSTANDS A SEQUENCE OF EVENTS IN A LOCA.
| |
| i
| |
| ~
| |
| l l
| |
| l 77 l
| |
| | |
| i l
| |
| 1 RO EXAM KEY !
| |
| Esam Number NRC.RO Rev. 0 Einm Title NRC RO EXAM J
| |
| 78
| |
| | |
| . ._ _ __._ ._.__ _ __._._. _ _ . . ______._m___.. .___ ..._ .._ ._. _ m _._ _ . _ _ . _ _ __
| |
| l I
| |
| RO EXAM KEY Enam Number ilRC RO Rev, O Easm Title NRC RO EXAM i
| |
| 77 With the plant at .00Y. power, a loss of Vi!N PNL0lBl has resulted in a loss of power to the Feedwater Level Control System giving a Feed Reg Valve control signal failure.
| |
| 'Ihe power loss also caused both Reactor Recite pumps to shift from fast speed to slow speed and the B Recite Flow control valve to lockup. Which plant response would result from these failures?
| |
| (Assume no operator actions.)
| |
| : a. 7he "B" Feed Reg Valve would fail closed and the "A" and"C" Feed Reg Valves would AUTO OPEN to compensate. Reactor power will stabilize at a lower power level with both Recirc pumps in slow speed.
| |
| : b. All 3 Feed Reg valves will fail open. RPV level will raise to 51" which will initiate a reactor scram, Turbine trip, and Feedwater pump trip.
| |
| I
| |
| : c. All 3 feed Reg valves will fait closed. Reactor power willlower when Recirc pumps down shift and RPV level will lower to 9.7" which will initiate a reactor scram. IIPCS and RCIC will initiate at Level 2 and restore RPV level ,
| |
| : d. All 3 feed Reg valves will fall"as is". Reactor power willlower when Recirc pumps down shill and RPV level will raise to $ 1" which will initiate reactor scram, Turbine trip, and Feedwater pump trip.
| |
| 4 ANSWER:
| |
| : d. All 3 Feed Reg valves will fail"as is". Reactor power willlower whpn Recirc pumps down shift and RPV level will raise to $1" which will initiate reactorhcram, Turbine 4
| |
| trip, and Feedwater pump trip.
| |
| IDNO: LP# OILI #
| |
| 421 ilLO-532 7 PROCEDURE NUMBER: OTHER:
| |
| AOP 0042 LEVEL 4 ARP-680-3-A06
| |
| ..7.-
| |
| 259002 K6.02 3.3[ 3.4 263000 K3 03 3.4l 3.8 _
| |
| COMMENTS: 1/97 new AOP-0042 Rev 7 page 64 of 90 ARP/P680/03A/A06 79
| |
| .-..m..,-....---=-----,.-----.w--.--=- . - - , - . - ,- ,,,-m... ,- ~ ~ ,~s.- - . - - , . , -.,.--,v.,. --~ . . , - -- -v s e
| |
| | |
| i
| |
| ?
| |
| i RO EXAM KEY Esam Number NRC RO Rev. 0 Esam Title NRC RO EXAM 73 The llPCS 125 VDC Switchgear E22 S001 is being supplied by the Backup Battery Charger BYS- ,
| |
| CllGRID. Which one of the following describes the llPCS 125 VDC Switchgear status if a LOCA occurs?
| |
| : a. The backup battery charger supply breaker will trip leaving the llPCS battery as the only power source.
| |
| : b. The backup battery charger will continue to supply the llPCS DC system loads and the battery will act as a backup.
| |
| : c. The backup battery charger will be load shed from its non safety related power supply and must be manually aligned to a 480 VAC Standby bus.
| |
| : d. The llPCS DC system will be supplied by the llPCS battery with the backup "
| |
| battery charger available if a low battery voltage condition is present.
| |
| ANSWER:
| |
| : a. The backup battery charger supply breaker will trip leaving the llPCS battery as the only power source.
| |
| II)NO: LP# Olu # J !
| |
| 77 IILO-035 4 PROCEDURE NUMBER: OTHER:
| |
| AOP 0003 LESTL 3 t OTM-57 NRC K At~ Y id["I SRO: l~
| |
| 263000 K4.02 3.1I 3.6i_
| |
| COMMENis: 7/97 new RO T2 02 SRO T2 G2 80 4
| |
| l
| |
| _m , , . . . . . ._..-...,m,m.._, .m,... ,_..,,..,r. , _ _ _ . , , _ , . , _ _ , , _ ---. .
| |
| | |
| -- _ _ _ _ - _ _ ___ . _ _ _ = _ _ _ _ ___ . _ - _ _ _ _ _
| |
| _ _ - _ _ _ _ . _ _ __ - = _ - . . . - _
| |
| ATFACllMENT 2' PAGE 2 OF 6 '
| |
| ISOLATION VALVE CilECKOFF S!!EET IsotA111Y DIVI &IV RISIORi1Y PANEL ISOLA 111Y DIVII&III RESIORilY NnA13 OUDIOARD N IIAIS NI1AIS NOARD L\TIIAIS lil3 P863 liVN MOV127 GROUPI liVN MOV129 SIGNALS IIVN MOV102 liVN MOV128 BD liVN MOVl30 liVR AOV125 N/A liVR AOVl47 N/A liVR AOV126 N/A liVR AOV148 N/A fil3 P808 RCS MOV61 A GROUP 1 RCS.MOV61B RCS-MOV60A SIGNALS RCS MOV60B RCS-MOV59A B, D RCS-MOV598 RCS-MOV58A RCS-MOV58B END ACB563 SOP 0049 ENB ACB583 SOP-0049 Ill3 P601 ,
| |
| E22-h0VIU23 GROUP 1 SIGNALSB,D lil3 P601 ,
| |
| E51 F064 SOP-0035 GROUP 2 E51.F063 SOP 0035 E51 F031 SOP 0035 SIGNALS E51 F076 SOP-0035 11Q,R S.T,V,X lil3 P601 GROUP 3 E51 F077 SOP 0035 SIGNALS E51 F078 SOP 0035 BOTH D R lii3 P655 Ill3 P654 E33 F007 GROUP 4 E33 F028 E33 F008 SIGNAL DD E33 F027
| |
| ~
| |
| A OP-0003 REV - 10 PAGE 13 OF 17
| |
| | |
| i B. Abnormal Operation
| |
| : 1. Backup Charger Operation I
| |
| ne backup charger can provide power to any of the 125 VDC buses except IBYS- ,
| |
| PNL0lrIBYS PNLO4, and IBYS SWOO6. The breakers to the standby buses (ACB :
| |
| 563, ACB 583, and ACB623) will automatically trip on a LOCA signal.
| |
| During normal operation, the backup charger will not be in service and the backup :
| |
| t charger feeder breaker to the DC bus is located in its spare cubicle. In order to place the backup charger in service, the normal charger is first taken out of service, and its breaker put into the backup charger oreaker cubicle. Den the backup charger feeder breaker is installed in the appropriate cubicle, ne backup charger is then placed in service. The operator should minimize the time for switching the chargers since the battery supplies !
| |
| the DC bus during switching.
| |
| 2.- Load Testing ;
| |
| Periodic load testing is performed to demonstrate battery capacity. his is performed by ,
| |
| removing the battery from service and allowing the normal battery charger to handle the DC loads. . .
| |
| When the load test is complete, the portable tester is removed and battery is retumed to service. De battery will require a complete recharge prior to being relied upon for emergencies.
| |
| l L
| |
| i
| |
| | |
| ==REFERENCES:==
| |
| | |
| SD-305, DC Distribution SOP-0049,125 VDC System SOP-0050,48 VDC Distribution System AOP-0014, Lossoi.125 VDC System SOER 83 5 CR 90-0553 l - STP-000-0001 l'
| |
| l l'
| |
| - LOTM-574 - Page 8 of 27.
| |
| L. - .
| |
| = -- . _ . - - - - . - .-
| |
| | |
| .-. .- -. -- . . . - . . - ~ .. . - . . _ - - . - - . - .
| |
| 1 RO EXAM KEY 1 Esam Numtwr NRC.RO Rev. 0 Esam Title NRC RO EXAM ,
| |
| 79 The following conditions exist:
| |
| i 1he reactor is at 100% power, lhe Off Gas Post Treatment til 111111 radiation alarm (P601/22A/A03) has occurred.
| |
| The Ofigas System automatica'.ly isolated,(IN64 F060 Off Gas Discharge to Vent valve is closed).
| |
| Which of the following actions is PROlllBITED? I
| |
| : a. Purge the Off Gas system with service air.
| |
| : b. Shift to the Standby Off Gas Component.
| |
| : c. Reduce power r,s necessary to maintain condenser vacuum.
| |
| f
| |
| : d. Reduce power to below 60%.
| |
| ANSWER:
| |
| : a. Purge the Off Gas system with service air.
| |
| IDNO: LPW OllJ #
| |
| 201 IILO-047 9 PROCEDURE NUMBER: ,j OTHER:
| |
| SOP-0092 LEVEL 2 1
| |
| NRC KAt RO: ~ SRO:
| |
| 271000 K1.07 2.7 2.7 COMMENTS: 7/97 new RO T2 02 SRO T2 02 l .
| |
| 81
| |
| - -. - - - ~. - . . .-. . - - - , - . . . . - . . _ -.- . - .
| |
| | |
| t RO EXAM KEY .6 Esam Number NRC RO Rev. O Esam Title NRC RO EXAM i
| |
| 80 Which one of the following maintains a negative pressure in the annulus following a LOCA? ,
| |
| : s. SBOT system starts and takes a suction on the Annulus Pressure Control System,
| |
| : b. Annulus mixing fans start and dischvge to the Annulus Pressure Contiol System. ;
| |
| : c. Annulus mixing fans start and discharge to the SBGT system.
| |
| : d. SBGT system starts and takes a suction directly from the Annulus.
| |
| ANSWER:
| |
| : c. Annulus mixing fans start and discharge to the SBOTS.
| |
| IDNO: LP# OHJ# ,
| |
| 82 IILO-038 2d PROCEDURE NUMBER: OTHER:
| |
| SOP 0059 LEVEL 2 SOP 0043
| |
| ~
| |
| NRC KAt I RO: I SRO '
| |
| ~
| |
| 'j 290001 K1.04 i 3.7j 3.9 _
| |
| COMMENTSt 7/97 new RO T2 G2 SRO T2 01 S
| |
| 4 82
| |
| | |
| i i
| |
| i RO EXAM KEY Esam Number NRC.RO Rev. O F. sam Title NRC RO EXAM g1 Olven the following conditions:
| |
| \
| |
| Reactor water level is .90 inches and lowering.
| |
| Drywell pressure is 2.2 psig and raising.
| |
| An outside fire has caused smoke in the Control Room.
| |
| He operator has attempted to manually place the Control Room ventilation in the smoke removal mode.
| |
| Under these conditions the Control Room Smoke Removal Damper (AOD 107/108) will:
| |
| : a. open and the Smoke Removal Fan will start.
| |
| : b. open but the Smoke Removal Fan will be interlocked off. ,
| |
| : c. remain closed and the Smoke Removal Fan will run on recire,
| |
| : d. remain closed and the Smoke Removal Fan will be interlocked off.
| |
| ANSWER:
| |
| i remain closed and the Smoke Removal Fan will be interlocked off.
| |
| IDNO: LP# OBJ #
| |
| 167 LOTM-61 11 PROCEDURE NUMBER: OTHER:
| |
| SOP 0058 LEVEL 4 ,
| |
| ~~ NkikAi'l RO':1 SMO' 290003 K1.o4 I 3.21 3.3 _
| |
| COMM(.NTS: 7/97 new RO T2 G2 SRO T2 G2 4
| |
| 83
| |
| - r. ., -..,,n., - . - . , , . , . - . , , . . . . . . , . , , , , , - , _ , . . . . . , , , . , . . , . . . . . . . _ . , ,
| |
| , = . .. , .n- . ...-.., .
| |
| | |
| RO EXAM KEY ,
| |
| Esam Naimber NRC RO Rev. O l'ssm Title NRC RO EXAM 82 With the plant operating at 100% power, a small stator water cooling leak has required addition of makeup water w eekly for the past few weeks. "Ihe Turbine Building operator is adding makeup water per SOP-0020 when artnunciator P680-15A.A06, MN GEN ll2/ STATOR CLG/ SEAL OIL !
| |
| TRIIL, alarms 'Ihe Turbine Ikilding operatc: investigates and reports that a high conductivity alarm has been received. Chemistry confirms the conductivity reading > 10.2 microSiemens/cm.
| |
| Which of the following actions is required?
| |
| : a. Immediately scram the reactor and trip the turbine.
| |
| l
| |
| : b. Redue stator amps to <l650 within 20 minutes,
| |
| : c. llave the operator verify proper operation of TPCCW cooling for the Stator Water Cooler,
| |
| : d. Commenee a normal plant shutdown.
| |
| ANSWER:
| |
| : a. immediately r im the reac;or and trip the turbine.
| |
| IDNO: LP# OILI #
| |
| 65 IILO-087 4
| |
| . i PROCEDt>RE NUMBER: OTHER:
| |
| SOP @20 LEVEL 4
| |
| ~
| |
| NRC KA! RO: SUI 245000 K5.07 2.6 2.9 _,
| |
| COMMENTS: 7/97 new RO T2 02 SRO T2 G2 84 i
| |
| - - . - , . ..-..,r. . . . . , . . . . , . , ,- . - . ~ . . . , , _ .,. .
| |
| --.-c . , . _ . . , , _ , . . . . , _..-._,,~_,..m. . .-
| |
| | |
| 9 i PURPOSE 1.1 To provide instructions for the operation of the Generator Stator Cooling Water System.
| |
| 2 PRECAUTIONS AND LIMITATIONS 2.1 If adjustments are to be made to flow or instrumentation on the Stator Cooling System when on line, consideration should be given to loweimg generator load to less than 282,000 KVA until adjustments are complete.
| |
| 2.2 When generator or rectifier cooling water filter differential pressure reaches 8 psid, the filter should be changed at the next outage.
| |
| 2.3 Gne microsiemen/cm is equal to I micromho/cm.
| |
| 2.4 Deionizer resins should be changed:
| |
| * When pressure drop across Delonizer is 15 psi'or greater.
| |
| . When water conductivity can nE be maintained below 0.5 microSiemens/cm.
| |
| 2.5 If stator cooling water conductivity exceeds 9.9 micr6Siemens/cm, the generator must be immediately removed frorn service.
| |
| 2.6 Without circulation through the Stator Cooling System, the quality of the cooling water l
| |
| will deteriorate and conductivity will increase. ARP 870-54 should be referred to for required actions on a loss of stator cooling water.
| |
| 2.7 Water entering the Deionizer should tjv exceed 47'C to prevent deterioration of the resin.
| |
| 2.8 GE TIL 1098 3R2 identifies a potential for Generator Stator damage due to Stator Cooling System inleakage when generator hydrogen pressure is less than Stator Cooling System pressure. Therefore, generstor hydrogen pressure should be greater than or equ to 55 psig while Stator Cooling System is in service. 50 psig is allowed for startup of the Main Generator.
| |
| 3 PREREdUISITES FOR STARTUP AND OPERATION 3.1 Check 480VAC is ope.ating and lined up per SOP-0047,480 VAC System.
| |
| SOP-0020 REV 3 PAGE30F18
| |
| | |
| RO EXAM KEY Essa Number NRC RO Rev. O Esam Title NRC RO!!XAM g3 The contsof room operator is about to startup RilR in the fuel pool cooling assist mode. All prerequisites have been completed and the operator starts RilR pump
| |
| * A'. lie then starts to throttle open the llX OUTLET VLV (El2 F003A). Watching the valve position indication the operator observes there is no change, also flow is 700 gpm. Fuel pool level: (choc se or.e)
| |
| : a. INCREASE because Suppression Pool water is being diverted to the Fuel Pool.
| |
| : b. REMAIN Tile SAME because the RilR system is currently recircing 700 gpm from/to the Suppression Pool,
| |
| : c. REMAIN TilB SAME because the RilR systeni is currently recircing 700 gpm from/to the Fuel Pool.
| |
| : d. DECREASE because water from the Fuel Pool is being diverted to the Supptession Pool.
| |
| ANSWER
| |
| : d. DECREASE because water from the Fuel Pool is being diverted to the Suppression Pool IDNO: LPW OllJ #
| |
| 438 IILO-021 08
| |
| .I PROCEDURE NUMBER: OTHER:
| |
| SOP 403t LEVEL 3 TASK NUMilER:
| |
| 205005001001 205013001001 233000 K3 02 3.1i 32 _
| |
| COMMLNTS: 7/97 NCW 85 t I
| |
| | |
| RO EXAM KEY Esem Number NRC RO Rev. 0 Esam Title NRC RO EXAM g4 Given the following plant conditions: l
| |
| . The Refuel Platform is over the core. !
| |
| . The Mode Switch is in REFUEL.
| |
| Which of the following will cause a control rod block?
| |
| r.. Withdraw a control rod.
| |
| : b. Loading the Auxiliary Platform hoist,
| |
| : c. Loading the Refuel Platform main hoist.
| |
| : d. Loading the Refuel Platform monorail holst.
| |
| ANSWER:
| |
| : c. Loading the Refuel PINform main holst.
| |
| IDNO: LP # OllJ #
| |
| 4 IILO-022 2 PROCEDURE NUMBER: OTHER: ,
| |
| FHPM3 LEyEL 3 NRC KA! RO: 3RE 234000 K4.01 3.3 4.1 _
| |
| COMMENTS: 7/97 new RO T2 03 SRO T2 G2 86
| |
| | |
| RO EXAM KEY !
| |
| Esam Number NRC.RO Rev 0 Esam Title NRC RO EXAM g5 A loss of all AC power occurred. The following plant conditions exist:
| |
| . Reactor shutdown (all roits inserted).
| |
| RPV pressure 1030 psig at 0700.
| |
| . RCIC systern is manually started.
| |
| . llue is a small coolant leak into the contalnment.
| |
| The CRS directs an RpV cooldown at 0700. Which of the following is the lowest pressurc
| |
| - permissible at 08007
| |
| : n. $00 psig
| |
| : b. 440 psig
| |
| . c. 400 psig
| |
| : d. O psig ANSWER:
| |
| : b. 440 psig II)NO: LPW OILI #
| |
| I 216 11L0 501 10 PROCEDURE NUMBER: OTHER:
| |
| GOP 0002 LEVEL 3 STPW700 TS 3 4.11
| |
| ~
| |
| NRC KA': IiUIl S00:
| |
| 3.71 4.1
| |
| [C0002 A2.04 ,_
| |
| CCMMENTS: 7/W7 new RO T2 03 SRO 12 03 37
| |
| | |
| i RCS P/T Limits 3.4.11 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. - --- ---NOTE- - - -
| |
| C.1 Initiate action to immediately Required Action C.2 restore parameter (s) shall be compleced if to within limits, tSis Condition is entered. 6NQ C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2
| |
| .LC0 not met in other operation. or 3 than MODES 1, 2, and 3.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ---- --- ---- --- NOTF ------ --------F---
| |
| Only required to be performed during RtS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.
| |
| Verify: 30 minutes
| |
| : a. RCS pressure and RCS temperature are within the limits of Figure 3.4.11-1, and
| |
| : b. RCS heatup and cooldown rates are l s 100*F in any one hour period.
| |
| (continued)
| |
| RIVER BEND 3.4-28 Amendment No. 81
| |
| | |
| RO EXAM KEY :
| |
| Esam N9mber NRC RO Rey, O Esam Title NRC RO EX AM i
| |
| 86 With the Division i Diesel rear air start compressor motor selred Which of the following actiou are required?
| |
| : a. Declaro Division i DieselINOPERABLE Both starting air compressors are required for an OPERABLE air start system.
| |
| i
| |
| : b. Take no action since only one starting air system is necessary to start the dicsci ,
| |
| and crosstying the unaffected air system with the affected alt system would render the dieselINOPERABLE.
| |
| : c. Maintain pressure in the normal band of the reclever associated with the selred compressor by intennittently using a high pressure hose coected between the operable forward and rear system air dryer outlets. The Division i Diesci Generator will remain OPERAIILE.
| |
| r
| |
| : d. Start and load the Division i Diesel denerator. With the dieselomning, the starting air system is not required for OPEit ABILITY of the diesel.
| |
| At(SWER:
| |
| : c. Maintain pressure in the nonnal band of the reciever associated with the seired compressor by intermittently using a high pressure hose connected between th operable forward and rent system air dryer outlets. The Division i Diesel C data IDNO: LP# OltJ #
| |
| ~l 54 IILO-037 9 PROCEDURE NUMBER: OTHER SOP 0053 !EVEL 4 1S 1.1 NnC mat RO: ' $ Rot .
| |
| 264000A1;00,_,. 32 3._2 0 2,1.33 3.4 4 ,
| |
| ^
| |
| COMMLNTS: 7/97 new Imust understand TS/TRM 1.1 definition of OPERABLE / OPERABILITY and know that the comrwessoes are not safety < elated l
| |
| . I l 88 l.
| |
| -M
| |
| * p pT-=i- ,7 y- g r7= -+ w ,<1mve9 y- -
| |
| -7g-- q y-- y e- *g- gp---- p gy.w>y -m-- p---- g7m+' ~oW'---Yd77 f F+*F h
| |
| | |
| i RO EXAM KEY j l'. sam Number NRC-RO Rev. O Esam Title NRC RO EXAM 37 Given the following initial conditions for the inclined fuel Transfer (ll'IS) System:
| |
| - Tube full
| |
| . Upper upender vertical
| |
| * e
| |
| ( -
| |
| Carriage at upper terminal 1.ower upender inclined
| |
| - System powered up and neither bridge in the li'IS area SELECT the conect statement regarding IITS operation,
| |
| : n. The refueling bildge can enter the IITS area.
| |
| b, lhe fuel handling bridge can entet the IFTS area in the l'uct fluilding.
| |
| : c. 1he transfer tube can be drained.
| |
| : d. The winch can be lowered using the " lower" pushbutton on the upper control panel.
| |
| ANSWER:
| |
| : a. The tefueling bridge can enter the li'IS area.
| |
| 011J# ,)
| |
| IDNO: If #
| |
| 5 llLO-228 03 PROCEDURE FUMBER: OTHER:
| |
| FHP-000$ LEVEL 3
| |
| ' NnCI' At ~l R0i~l $f50i 234000 K4 05 _I ?l 3.8 COMMENTS: 7/97 new 7 89
| |
| | |
| _ _ . . _ . _ _ _ . _ _ _ . - . _ ~ - _ _ _ _ . _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _
| |
| RO EXAM KEY Esam Number NRC RO Rev. 0 Esam Title NRC RO EXAM j I
| |
| gg While the plant is at power, a leak develops in an ana that is acessible, but now radiologleally contaminated. The OSS has directed that an investigation be performed immediately. What documentation must be generated before various personnel are allowed entry into the area for the investigation?
| |
| : a. A daughter RWP to the General RWP for that area must be generated.
| |
| b, None, a General RWP already exists for this type of event.
| |
| : c. A Specific RWP must be generated.
| |
| : d. None, a RWP may be completed after the entry provided it is done under continous RP coverage.
| |
| ANSWER:
| |
| : d. None a RWP may be completed after the entry provided it is done under continous RP coverage.
| |
| 1 IDNO: LP# OILI #
| |
| 433 GET 022 27 i
| |
| PROCEDURE NUMBER: OTHER:
| |
| RSP 0200 LEVEL 2 ,
| |
| TASK NUMllER *j >
| |
| 300157003002 G 2.3.4 I 2 Si 3.1 _
| |
| COMMENTS: 7/97 new 90 l
| |
| l f ,. __ _ _ __ _, ___ _ _ _ _ .. __
| |
| | |
| l RO EXAM KEY Esam Number NRC-RO Rey, O Esam Tit!* MRC RO EXAM
| |
| ~
| |
| l 39 Given the following conditions:
| |
| . He plant is shutdown making preparations for Shutdown Cooling (SDC) using the "A" loop of RilR
| |
| . Iloth Recirculation pumps are shutdown with their discharge valves closed Which of the following describes how the "A" RilR pump that is being started for SDC is protected from damage due to no flow?
| |
| : a. He operator is required to establish a pump discharge flow path to the reactor as soon as possible aner starting the pump.
| |
| : b. He pump minimum flow valve (f064 A) wili open to provide fbw until the RilR 51 eat Exchanger !!ypass Valve (F048A) can be opened.
| |
| F
| |
| : c. He operator will open the minimum flow valve (F064 A) until shutdown cc ating flow ls greater than 500 gpm.
| |
| : d. The pump will automatically trip on low suction pressure if flow / pressure is not adequate for pump suction.
| |
| ANSWER:
| |
| : a. %e operator is required to establish a pump discharge flow path to the reactor as soon as possible after starting the pump.
| |
| J IDNO: LP# Olu #
| |
| 411 !!LO-021 6 PROCEDURE NUMBER: OTHER:
| |
| SOP 403t LEVEL 3 G 2.1.2 3) 4_
| |
| COMMENT $t 7/97 new also HLO42 t, OBJ. 8 91
| |
| .-- - - - . - - . - . - .-. . . - - .-:~..
| |
| | |
| RO EXAM KEY Esam Number NRC-RO Rev. O Esam Title NRC RO EXAM 90 Which of the following pennission/ notification requirements i ust be met for an INTENTIONAL entry into Tech Spec 3.0.37 permissic 1 must be obtained from the:
| |
| : a. Operations Superintendent and the NRC Resident inspector notified,
| |
| : b. General Manager . plant Operations and the NRC Resident inspector notified,
| |
| : c. Manager Operations and a 1 hour report made to the NRC,
| |
| : d. General Manager . plant Operations and a 4 hout report made to the NRC.
| |
| ANSWER:
| |
| : a. Operations Superintendent and the NRC Resident inspector notified.
| |
| IDNO: LP # OBJ#
| |
| 423 IILO-206 2 PROCEDURE NUMBER: OTHER:
| |
| ADM@22 LEVEL 2 l
| |
| NRC KAt l RO: SRO: J G 2.1.17 3.4 3.8 _
| |
| COMMENTSt 1/97 NRC eham 0
| |
| l 92 l
| |
| l I
| |
| | |
| ... . . . . _ . _ - _ , - - - - _ . _ - - . - . . . - . . - . . _ - _ _ _ . _ _ ~ . _ - _ _ _ -__
| |
| RO EXAM KEY Esam Nusuber NRC-RO Rev. O Esam Title 14RC RO EXAM 91 1he Plant is operating at 100% reactor power when a loss of feedwater heating occurs. Which one of the following is a required IMMEDIATE action for this loss of feedwater heating?
| |
| : a. Reduce reactor power by 40 MWE with core flow, then reduce another i10 MWE with core flow and rod insertion.
| |
| : b. Reduce power to less than or equal to 100% rated thermal power using core flow,
| |
| : c. If failed fuel exists in the reactor, reduce reactor power by 495 to $00 MWE.
| |
| : d. Insert contro' rods in reverse order to get below the 80% rod line.
| |
| ANSWER:
| |
| : b. Reduce power to less than or equal to 100% rated thermal power using core flow.
| |
| IDNO: LP# 010 #
| |
| 375 llLO 526 4 PROCEDURE NUMBER: OTHER:
| |
| AOP 0007 LEVEL 2 NRC KA: 'I RO: l SRO:
| |
| G 2.4.49 i 4 4_ J COMMENTS: 7/97 new AOP 0007, Rev 8, p. 20 of 6 no & SRO T3 04 0
| |
| | |
| RO EXAM KEY Esem Number NRC.RO Rev. O Esam Title NRC RO EXAM, 92 When EOP-4, Emergency RPV Depressurization, permits defeating isolation interlocks in order to rapidly depressurize witho:it SRVs. Which of the following MSIV isolation signals may be bypassed?
| |
| : a. Only the RPV low level I sirpal.
| |
| : b. Only the RPV low level I and low main steam line pressure signi .
| |
| : c. All MSiv isolation signals except for Main condenser low vacuum.
| |
| : d. All MSIV automatic isolation signals.
| |
| ANSWER:
| |
| : d. All MSIV automatic isolation signals.
| |
| IDNO: LP# OIU #
| |
| 212 IILO-516 9 PROCEDURE NUMBER: OTHER:
| |
| EOP-0004 LEVEL 2 EOP-005 ENCL 9 NRC KAt RO: l~55UI 223002 K4.08 3.3! 3.7 _
| |
| COMMENTS: 7/97 new RO T2 01 SRO T2 01 l
| |
| l l
| |
| L 94
| |
| | |
| RO EXAM-KEY t Esam Number NRC RO Rev. O Esam Title NRC RO EXAM 93 Prior to reactor startup the f ellowing SRM count rates are recorded:
| |
| i SRM A 25 cps SRM B 30 cps i
| |
| SRM C 35 cps SRM D 1$ eps At what SRM reading should the operator consult the Reactor Engineer for continued withdrawal recommendations?
| |
| : a. 2500 cps on SRM A
| |
| : b. 3000 cps on SRM B
| |
| : c. 300 cps on SRM C
| |
| : d. 240 eps on SRM D ANSWER:
| |
| : d. 240 cps on SRM D IDNO: LP# OBJ # ,j
| |
| $7 STM 503 07 PROCEDURE NUMBER: OTHER:
| |
| GOP 0001 LEVEL 3 1
| |
| NRC KAt RO: I SRO:
| |
| 0 2.1.7 3I 4_
| |
| COMMENTS: 7/97 new Requirement to contact RE if e'ter 4 co;n' sto doubhngs (from irvtial) the reactor is not critical, 4
| |
| e 93
| |
| | |
| _ = _ _ _ _ _ _ . . _ . _ . _ _ _ _. . _ . . _ . ._ _ . _ _ . , _ _ _ _ _ . _ _ _ . . _ _ _ _ _ . _ . _ . . _
| |
| RO EXAM KEY Esam Number NRC.RO_ Rev. 0 Esam Title NRC RO EXAM 94 nc basis for Surveillance Requirement 3.1.5.1 for each control rod xram accumulator pressure to be verified > 1520 psig every 7 days is to ensure adeqeate aaumulator pressure exists to:
| |
| : a. provide sumcient scram force.
| |
| : b. drive control rods on a loss of CRD pumps.
| |
| : c. maintain mdication in the readable range c? : tuge.
| |
| : d. verify accuracy of control room IICU pressure indications.
| |
| ANSWER:
| |
| : a. provide sumclent scram force.
| |
| IDNO: Li' # OILI #
| |
| 414 STM-052 8 PROCEDURE NGMBER: OTHER:
| |
| TS 3.1.5 LEVEL 2
| |
| ,2,2.25 NRC KA! l~ 50:T sio:-
| |
| l 15I 3.7
| |
| -]
| |
| J e
| |
| COMMENTS: 7/97 new SR 3.1.6.1 requires that the accumulator pre?,sure be checked every 7 days to ensure adequate accumulator pressure exists to provede sufficient scram force. The primary indicator of accumulator operability is the accumulator pressure.
| |
| 96
| |
| _ . . _ - . . . ~ _ _ . . _ _ . _ _ . _ . ._ _ _ _ _._ .. _ . _ . . ~ .. _ -
| |
| | |
| k RO EXAM KEY Esam Number NRC-RO Rev 0 Exam Title NRC RO EXAM 95 in EOP-1 A, ATWS RPV Control, if SRVs are cycling, the operator is directed to manually open SRVs until RPV pressure drops to 930 psig.
| |
| Which of the following is the reason for stopping the reactor presso eductio. at 930 psig?
| |
| : a. To ensure the turbine bypass valves do not have the oppprtunity to stick closed
| |
| : b. To p sent MSIVs from closing on low main steam line pressure
| |
| : c. To minimize the amount of steam that is sent to the suppreasion pool
| |
| : d. To prevent excessive loss of reactor coolant inventory ANSWER:
| |
| : c. To minimize the amount of steam that is sent to the suppression pool IDNO: LP# OILI #
| |
| 313 IILO-513 4 PROCEDURE NUMBER: OTHER:
| |
| EOP 1A LEVEL 3 EPSTG*0002 J
| |
| q 295037 EKJ e G 2.4 6 }__3.8:3.1 f4 _ _4.1 COMMENTS: 7/97 00W 97
| |
| | |
| RO EXAM KEY Exam Number NRC-RO Rev. 0 Eram Title NRC RO EXAM 96 The plant is operating at power and a fuel handling 'erm is wking in the Fuel Building, when an
| |
| !rradiated fuel bundle is dropped and is lying on tht tottom c : the spent fuel pool.
| |
| Which of the following actions should be taken?
| |
| : a. Evacuate the Fuel Building
| |
| : b. Start the second charcoal filter train of the Fuel Building ilVAC.
| |
| : c. Manually scram the reactor,
| |
| : d. Dire:t the team to attempt to recover the bundle with the bridge grapple.
| |
| ANSWER:
| |
| : a. Evacuate the Fuel Building IDNO: Ll' # OBJ #
| |
| 431 IILO-535 4 PROCEDURE NUMBER: OTHER:
| |
| AOP-0027 LEVEL 3 TASK NUMBER:
| |
| 400043004001 J
| |
| NRC KA: I RO: SRO:
| |
| G 2.2.32 I 3.5 3.3 _
| |
| COMMENTS: 7/97 n.:w 98
| |
| | |
| [
| |
| i L
| |
| : 1.0 PURPOSFlDISGSSION I 1.1 The purpose of this procedure is to provide instructions to the operator in the event
| |
| : of a nushap causing damage to the nuclear fi:e! resulting from a dmpped irradiated fuel bundle or the indication of inadvsent .:riticality or refueling cavity / upper conttinment pool or lower fuel pool water level problems. Minor bumpmg of a fuel bundle does not require implementation of this procedure.
| |
| 1.2 The major emphasis is on minimizing radiation e. pure to plant personnel and off-site radioactivity releases.
| |
| will be covered by 1.3 Recovery from the eventat( the . retrieval Special Procedure (s) devel time of theof a dropped event, if deeme fuel bundle)d necessary Emergency Plan Recovery ger.
| |
| 1.4 - The ition, duties and responsibilities of the Fuel Handling Crew are specified in REP 10 and FHPs (Fuel Ifandling Procedures).
| |
| 2.0 SYMPIME 2.1 Observation of fuel bundle dropping or striking a fixed object while moving.
| |
| 2.2 - High ahbome activity detected in the Fuel Building or Containment.
| |
| 2.3 Visual observation of a slow, rapid or steady decrease in refueling cavity or lower fuel pool water level during refueling operations.
| |
| 2.4 Visual observation of leakage from the refueling cavity seal assembly fiom the drywell side during refuel,ing operations. .)
| |
| 2.5 Malfunction of Inclined Fuel Transfer System (IFIS) with irradiated fuel loaded in the carrier.
| |
| 2.6 Observation of steady increasing count rate with a measunible period.
| |
| 3,0 AUIOMNI1C ACI1ONS 3.1 1RMS*REllA B) ANNULUS EXHAUST HIGH RADIA'ITON will auto start the Annulus Mixin(g System and Standby Gas Treatment System if aligned operation. Not required _in modes 4 and 5.
| |
| 3.2 1RMS*RESA B) FUEL BUILDING EFFLUENT HIGH RADIAllON will isolate Fuel Bldg nor(mal1 and exhaust ventilation (if running) and au+o initiate Fuel Bldg exhaust ilter trains A & B.
| |
| 4.0 - IMMEDINIE OPERATOR ACI1ONS 4.1 Manually insert a reactor scram if inadvertent criticality is indicated.
| |
| 4.2 Stop ah refueling operations and move fuel bundle / irradiated device to a safe
| |
| - posittors 4.3 Evacuate the Fuel Building, Contahunent and Drywell as necessary.
| |
| AOP-0027 - REV - 10 PAGE 2 OF 10
| |
| | |
| l 1
| |
| RO EXAM KEY Esam Number NRC RO Rev. O Esarn Title NRC RO EXAM 97 The plant is shutdown for a maintenance outage. Work is being performed on a portion of the Feedwater system by Mechanical Maintenance. The I&C foreman has received a Clearance Receipt to work within the feedwater system tagout boundary to calibrate an instrument. Upon completion of work, the Mechanical Maintenance foreman wishes to release his clearance and restore the system, but the instrument ca!!bration is still taking place.
| |
| What actions (s), if any, must be taken to ensure the safety of the personnel performing the calibration?
| |
| : a. The Clearance form is transferred to the I&C foreman,
| |
| : b. The Clearance can be released with verbal permission from the I&C foreman.
| |
| c.1he I&C foreman must return his Clearance Receipt to the tagging omcial prior to releasing the clearance.
| |
| d.1hc Mechanical Maintenance foreman may clear all tags that pertain to the I&C work.
| |
| ANSWER:
| |
| : c. The I&C foreman must return his C'earance Receipt to the tagging omcial prior to releasing the clearance.
| |
| IDNO: LP # OllJ #
| |
| 432 ilLO-201 02 PROCEDURE NUMBER: OTHER:
| |
| ADM.00m M3 TASK NUM11ER:
| |
| 300101003001 NRC KA: ~I RO: [SRO:
| |
| G 2.2.13 ! 3! 3.4g]
| |
| COMMENTS: 7/97 new 99
| |
| | |
| .- .. _ . . - .- .. . _ . . . . - . - . . = -. . . . . . - . - _. - . .
| |
| RO EXAM KEY Exam Number NRC-RO Rev 0 Exam Title NRC RO EXAM An ATWS event has occurred and SLC injection has been initiated. EOP-I A "RPV Control 98 ATWS," directs RPV water level be lowered by terminating and preventing feedwater injection into the RPV What are the required actions by the ATC operator to terminate and prevent feedwater injection into the RPV7
| |
| : a. Shut the feedwater inboard isolation valves IB21'F065A(B).
| |
| : b. Dial the Master Feedwater Level Flow Controller setpoint down to 0".
| |
| : c. Place the Maste. Feedwater Level Flow Controller in manual and drive all Feedwater Regulating Valves full closed.
| |
| : d. Trip all 3 Reactor Feedwater pumps.
| |
| ANSWER:
| |
| : c. Place the Master Feedwater Level Flow Controller in manual and drive all Feedwater Regulating Valves full closed.
| |
| IDNO: LP# OlU #
| |
| 429 ilLO 512 4 PROCEDURE NUMBER: OTHER:
| |
| EOP 1A LEVEL 3
| |
| .h NRC KA: RO: SRO:
| |
| 295037 EK1.02 4 4 295037 EK3 03 4 4_
| |
| COMMENTS: 7/97 new l
| |
| l 100
| |
| | |
| EOP-4A Contingencies - ATWS - LP hydraulics and reactor physics principles for RPV water levels at or below the elevation of the steam separators.
| |
| Lowering RPV water level is accomplished by terminating and preventing all injection into the RPV, except from boron injection sys.tems and CRD since these two systems are low flow and may be needed to establish and maintain reactor shutdown conditions. With essentially no makeup of reactor coolant, RPV water level then decreases by boil-off.
| |
| Terminating and preventing injection shall be accomplisheo per the following:
| |
| : 1. Condensate /Feedwater Place the Master Controller in Manual and drive all Feed Reg valves full closed; place the Startup Feed Reg valve in Manual and drive it full closed.
| |
| : 2. HPCS - While holding the control switch for the injection valve in the closed position, arTn and depress the manualinitiation push button. When the pump has started and come to full speed, secure the pump at the SRO's direction. When ave.ilable, dispatch an operator to monitor and secure the diesel. ,' .
| |
| : 3. EClC - Depress the manual trip push button ,"
| |
| : 4. LPCS/LPCI - If an initiation signal is present, manually override the injection valves by taking the control switches to the closed position. The pumps shall remain running. If an initiation signal is not present, prevent pump start and injection valve opening by defeating the initiation logic. Enclosure 27 provides guidance. ,)
| |
| : 5. ECCS Keeofill/ Condensate Transfer - Close the appropriate LPCS/ LPCI injection valves.
| |
| : 6. Service Water / Fire Water - Close the Containment Flood valves, E12*F094 and E12*F096.
| |
| Power oscillations may occur when RPV water level is lowered significantly below the normal operating ranet with the reactor still at power. These oscillations have been analyzed and determined to result in thermal transients well within the design capabilities of the fuel. The oscillations are noted at this point to indicate to operators that they are expected, and were considered in developing the steps which require deliberately lowering RPV water level with the reactor at power.
| |
| EPSTG*0002 B - 307 Revision 3
| |
| | |
| l RO EXAM KEY Esam Number NRC-RO Rev. 0 Esam Titk NRC RO EXAM 99 A 28 year old nuclear worker has been contracted to work during the upcoming refuel outage. His NRC Form 4 is current, and his accumulated whole body dose to date is 48.25 rem. what is the MAXIMUM radiation dose he can be authorized to receive during the next quarter I AW regulatory limits?
| |
| : a. 0.75 rem
| |
| : b. 1.00 rem c.1.25 rem
| |
| : d. 3.00 rem ANSWER:
| |
| : d. 3.00 rem IDNO: LP# 010 #
| |
| 434 GET-022 30 Pr.OCEDURE NUMBER: OTHER:
| |
| RSP-0200 LEVEL 2 NRC KA: I RO: I SRO: ,j G 2.3.2 l 2,6l 2.9 _
| |
| COMMENTS: 7/97 new l
| |
| i l
| |
| l 101
| |
| | |
| . . _ __ . -- ~ -
| |
| l RO EXAM KEY Exam Number NRC-RO - Rev. 0 Exam Title NRC RO EXAM 100 Plant conditions are as follows:
| |
| - Suppression pool temperature 87 degrees F Suppression pool level 20.5 ft.
| |
| Drywelltemperature 123 degrees F Reactor level 10.5 inches
| |
| . Aux. Bldg. pressure 0.25 Which one of the following EOP selections should be entered?
| |
| : a. EOP I and EOP 2
| |
| : b. EOP 2 only
| |
| : c. EOP 2 and EOP 3
| |
| : d. EOP 3 only ANSWERt
| |
| : b. EOP 2 only IDNO: LP# Olki #
| |
| 374 IILO-514 3 PROCEDURE NUMBER: OTHER:
| |
| EOP-0002 LEVEL 2 NRC KAt I RO: I SRO:
| |
| O 2.4.16 31 4_
| |
| COMMENTS: 7/97 new EOP-OOO2 RO & SRO T4 G4 102
| |
| | |
| i
| |
| '1 RIVER BEND STATION REACTOR OPERATOR NRC WRITTEN EXAMINATION 4
| |
| d RESTRICTED INFORMATION OFFICIAL USE ONLY
| |
| | |
| RIVER BEND REACTOR OPERATOR NRC EXAMINATION FACILITY: River Bend Station
| |
| ' REACTOR TYPE: BWR DATE ADMINISTERED: 07/25/97 CANDIDATE:
| |
| I 1
| |
| (
| |
| INSTRUCTIONS TO CANDIDATE:
| |
| i l Use the supplied answer sheet for documentation of your answers.,flhere are 100 multiple choice questions on this examination, each worth 1.00 points. Passing grade for this examination consists of an overall score of 80%. Examination papers will be picked up four (4) hours aller the examination begins.
| |
| l
| |
| : l. 100 l
| |
| TOTAL POINTS CANDIDATES SCORE PERCENT All work on this examination is my own. I have neither given nor received aid.
| |
| Candidate's Signature
| |
| , -> , ~, , , , . . - . , , . - . . - -, , - - , - -.
| |
| | |
| i 1- ne Plant was operating at 100% of rated power when a reactor scram occurred. Plant conditions are as followsi
| |
| - Reactor power is on range 2 of the IRMs and decreasing
| |
| - De Main Turbine is tripped
| |
| - De Main Generator Ouptut breakers are closed
| |
| - ne Main Generator Exciter Field breaker is opeu
| |
| - Megawatt load on the Main Generator ir.dicates 10 MWe (steady for 2 minutes)
| |
| - MVARs on the Main Generator indicate 50 MVARs leading (VARs in)
| |
| - - Station loads are being supplied through th: Preferred Station Transformers
| |
| - Main Condenaer Vacuum is 28" Hg and steady Which one (1) of the following describes the action (s) required by AOP 0002, Main Turbine and Generator Trips?
| |
| : a. Immediately trip the Main Generator Output breakers.
| |
| ~ b. Immediately initiate a reverse power trip of the Main Genetutor Output breaken by decreasing generator VARs to zero,
| |
| : c. No action is required as long as Main Conderser Vacuum remains above 26" Hg.
| |
| : d. Within 5 minutes, initiate a reverse power trip of the Main Generator Output breakers by decreasing generator VARs to zero, 2 Whl Sh of the following states the overall low low set system response of the SRVs for the RPV ,
| |
| pressures given?
| |
| : a. At 1113 psig, only one valve will be open. It recloses at 926 psig.
| |
| : b. At 1113 psig, only one valve will be open, it recloses at 936 psig.
| |
| : c. At 1103 psig, only two valves will be open. One recloses at 936 psig, the other at 926 psig.-
| |
| : d. At 1103 psig, only eight valves will be open. Bree reclose at $46 psig, three at 936 psig, and the last two at 926 psig.
| |
| 2
| |
| | |
| 3 You have been instructed to control drywel! temperature and pressure by operating all avaliable drywell coohng. While doing this, service water to the drywell unit coolers automatically isolates.
| |
| Which of the following caused the isolation?
| |
| : a. High drywell temperature (max. recorded 265 deg F)
| |
| : b. Low RPV water level (min. recorded 28")
| |
| : c. liigh drywell pressure (max. recorded 1.82 psid)
| |
| : d. Loss of 120 VAC power 4 Which of the following will result in the addition of positive reactivity to the reactor? (Consider each case separately.)
| |
| : a. LPCS initiaiton during reactor STARTUP with reactor pressure at 500 psig.
| |
| : b. Sudden jet pump differential pressure reduction in one loop with the reactor in the RUN mode,
| |
| : c. Reduction in EHC pressure setpoint by 2 psig with reactor in the RUN mode.
| |
| : d. Initiation of the RCIC turbine during reactor STARTUP with reactor pressure at 150 psig. .
| |
| J l 5 Which of the following methods for alternate control rod insertion during an ATWS REQUIRES the scram " signal" so be reset?
| |
| Control rod insertion by:
| |
| : a. t sing the individual control rod scram test switches.
| |
| : b. venting the control rod mechanism over piston volume.
| |
| : c. maximizing CRD cooling water differential pressure.
| |
| J. venting the scram air header.
| |
| l l
| |
| 3
| |
| | |
| 6 EOP-2 requires containment to be vented when containment pressure reaches 20 psig. Which one of the following describes the flow path for emergency venting of the containment?
| |
| : a. Drywell/ Containment pirge system takes a suction on containment and discharges through filter train #6 t a the main stack,
| |
| : b. Ilydrogen purge discharges to the annulus and the annulus mixing system is in operation discharging to SGTS.
| |
| : c. Drywell/ Containment purge fan takes a suction on containment and discharges through the purge exhaust fans to the main stack.
| |
| : d. Ilydrogen purge discharges through the Drywell/ Containment purge system filter train #6 to the main stack.
| |
| 7 EOP-4 (Primary Containtnent Floodirg) is executed to flood containment.
| |
| The containment level band specified by EOP-4 is between 62 ft and the Maximum Containment Water Level Limit (MCWLL)of 85 ft ,
| |
| Which of the following ranges corresponds to the core level band specified by EOP-47
| |
| : a. -143 to +133 inches
| |
| : b. 162 to +114 inches
| |
| : c. 193 to +51 inches 205 to +71 inches J
| |
| ( d.
| |
| t 8 Which of the following conditions still constitutes " Adequate Core Cooling"?
| |
| NOTE: Only the injection tources stated are injecting. Regard each situation separately.
| |
| : a. ATWS in progress, the feed system is maintaining level between 205 inches and -195 inches, MSIVs are open.
| |
| : b. All rods in, MSIV/ ADS valves are closed, RPV level is 200 inches and RPV pressure is 200 psig.
| |
| : c. All rods in, RCIC is injecting,1 ADS valve is open, RPV level at -210 inches and MSIVs are closed.
| |
| : d. ATWS in progress, CRD, RCIC and SLC (with Boron) are injecting, RPV level is 200 inches and MSIVs are open.
| |
| 4
| |
| | |
| i 9 All high pressure injection has been lost following a Reactor scram and RPV water level transient.
| |
| RPV water level is .162 and lowering slowly. Ri!R and LPCS are running on minimum flow. The CRS directs emergency depressurization. Why must at least 4 SRVs be opened to accomplish emergency depressurization under these conditions?
| |
| : a. LPCS and RilR will be injecting prior to RPV level reaching the minimum steam cooling level,
| |
| : b. The level swell from four open SRVs will keep the core submerged until RilR and LPCS are injecting at rated flow,
| |
| : c. LPCS alone can reflood the core prior to the core uncovery time exceeding the maximum core uncovery time limit.
| |
| : d. Enough SRV steam flow to cool the core will exist at a pressure that RilR can make up for the steam flow.
| |
| 10 Regarding the liydrogen Deflagration Overpressure Limit (HDOL) curve, as containment pressure increases, the maxtmum allowed hydroge'n concentration in percent (%) decreases.
| |
| Which of the following is the reason for this relationship?
| |
| : a. As containment pressure increases, the capr.bilities of the liydrogen Recombiners to remove hydrogen is decreased.
| |
| : b. This ensures a hydrogen de0agt : tion at the limit combined with current pressure ,
| |
| will not exceed containment overpressure failure limits.
| |
| I
| |
| : c. Tbc containment hydrogen analyzer system response time is adversely affected as pressure increases,
| |
| : d. As containment pressure increases, the deflagts e_ gessure of hydrogen decreases requiring a lower concentration of hydrogen.
| |
| | |
| .- - = _ -- - . . - . . . - - . .- -
| |
| 1 11 EOP 1 A. "RPV Control ATWS," directs.the of erator to inject boron into the RPV with SLC pump
| |
| _ "A" or "B" but not both.
| |
| Which of the following is the reason that EOP 1 A. "RPV Control ATWS," specifically prohibits ,
| |
| starting both Standby Liquid Control Pumps? t
| |
| : a. De pumps are interlocked,"B" will not start if "A" is running.
| |
| : b. De "B" system explosive valve will not fire if the "A" explosive valve has been
| |
| . fired.
| |
| : c. Excess discharge pressure will lift both pump reliefs thus reducing Boron irdection flow.
| |
| : d. 81 sing two pumps would inject Boron at an excessive rate, preventing adequate mixing in the reactor.
| |
| 12 SELECT the reason why terminating and* preventing injection during a failure-to-scram (ATWS) transient results in a power reduction.
| |
| Terminating and preventing injection:
| |
| : a. results in increased core inlet subcooling as feed preheating is decreased.
| |
| : b. increases the void fraction by a reduction in core natural circulation flow.
| |
| : c. iricreases water temperature as level is decreased.
| |
| . }
| |
| : d. results in an increase in fuel temperatures as core steaming rate increases.
| |
| 13 he plant was operating at 100% power when a scram signal was generated and the reactor failed to scram. EOP.1 A directs downshifting Recirc Pumps.
| |
| Which of the following describes the reason Recire Pumps are down shifted prior to tripping?
| |
| Tripping the Recire Pumps could result in:
| |
| : a. entering the region of thermal / hydraulic instability,
| |
| : b. an excessive feedwater temperature reduction rate that will cause power to increase rapidly.
| |
| ' c. a large level shrink which could cause isolation signals complicating the event,
| |
| : d. a reactor icvel swell which could result in a main turbine trip 6
| |
| .-. , .,. ,..--m. . , - - , .
| |
| | |
| l l
| |
| 14 During operation at 100% po'ver with a rod line of 100%, the "A" Recirc Pump inadvertently trips to off. Abo;t 20 seconds later, the "B" Recire Pump trips to slow speed, resulting in the following 1 steady state plant conditions:
| |
| Dermal power 50 %
| |
| Calculated core flow 34 %
| |
| What is the required operator action?
| |
| . a. Immediately SCRAM the reactor.
| |
| ' b. Reduce therms.1 power to less than 40% by inserting control rods.
| |
| : c. Raise core flow by upshifting the B recirc pump to FAST.
| |
| : d. Raise core flow by starting the A recirc pump in SLOW.
| |
| 15 A loss of condenser vacuum has occurred, vacuum is currently 18.5" Hg. Which of the following automatic actions should have occurred?
| |
| : a. Turbine trip only,
| |
| : b. Turbine trip and bypass valve closure.
| |
| : c. Turbine trip and MSIV isolation.
| |
| : d. T'urbine trip, bypass valve closure and MSIV isolation.
| |
| .h 16 The following conditions exist:
| |
| The plant has experienced a station blackout.
| |
| The Div 3 Diesel Generator was started and is running normally..
| |
| Emergency use of Div 3 for decay heat removal and RPV level controlis being implemented.
| |
| Which of the following describes the general flowpath for this cooling mechanism?
| |
| : a. CST - IIPCS pump - RilR "A" heat exchangers - RPV Shutdown cooling
| |
| - drains to Suppression pool.
| |
| ' b. Suppression pool- HPCS pump - RPV - shutdown coohng to loop "A" ,
| |
| RHR heat exchange
| |
| * s then test return to suppression pool.-
| |
| : c. CST -IIN'S pump RPV - shutdown cooling to loop "B" RHR heat exchanger then test retern to Suppresalon pool.
| |
| : d. Suppression pool ilPCS rump RilR "A" heat exchanger - RPV shutdown cooling drains to suppression pool. _,
| |
| 7
| |
| | |
| I i
| |
| i 17 A Loss of Offsite Power has occurred. De Division 1 Diesel generator is currently loaded to 2500 KW.
| |
| Which one of the following is the MAXIMUM allowed additional load that can be imposed on the generator?
| |
| : a. 360 KW
| |
| : b. $80 KW
| |
| : c. 630 KW
| |
| : d. 730 KW 18 A small break LOCA has occurred. Reactor level initially fell to -47 inches, then HPCS initiated and filled the reactor to a maximum of +55 inches, level is now steady at +40 inches. Which of the following describes the current status of E22 MOV F004, the llPCS injection isolation valve?
| |
| : a. F004 will open on a High Drywell Pressure initiation signal even if the HPCS lilGH WATER LEVEL 8 RESET pushhutton has not been depressed and the HPCS HIGH WATER LEVEL 8 RESET pushbutton must be depressed before the valve can be opened manually,
| |
| : b. F004 can be openei manually even if the HPCS HIGH WATER LEVEL 8 RESET pushhutton has not been depressed and the HPCS HIGH WAlliR LEVEL 8 RESET pushhutton must be depressed before the valve will open on a High Drywell initiation signal . j
| |
| : c. F004 can NOT be opened manually M lt will NOT open on a High Drywell Pressure initiation signal until the Hf 's > PH WATER LEVEL 8 RESET pusthutton is depressed.
| |
| : d. F004 can be opened manually after the HPCS INITIATION RESET pushbutton is depressed.
| |
| 8
| |
| | |
| ._ _ _ _ . . _ _ .._ _.= _ . . _ . _ _ _ _ _- - __ _. . _ _ _ _ _ _ . - _ _ _
| |
| l 19 EOP 2,_* Primary Containment Control", requires the reactor to he scrammed before suppression pool temperature reaches 110 Degrees F. Which one of the following states the reason for this requirement?
| |
| : a. Assures tha: the containment design pressure will not be exceeded due to compression of the non condensable r, asses due to the higher water temperature,
| |
| : b. Assures that with the expected temperat'c.re rise of 70 Degrees F during the ,
| |
| blowdown phaa of an accident, that complete condensation of reactor coolant will occur.
| |
| c, Assures the post LOJA suppression pool hydrodynamic forces are within the design limitation o'' minment,
| |
| : d. Assures a reactor shutdown ocuas. O minimize heat rejected to the primary containment, T.". . rpn y Depressurization is required.
| |
| 20 De Remote Shutdown Panel emergency transfer switches (division i switch on C61'P001 and divsion 11 switch on RSS*PNL102) for ADS /SRV B21 F05lG are in the EMERGENCY position.
| |
| Which of the following Control Room handswitches can be used to manually open ADS /SRV B21 F051G7
| |
| : a. BOTH the Div i "A" and Div II "B" solenoid control switches.
| |
| - b. Div I "A" solenoid control switch only.
| |
| : c. Div 11 "B" solenoid control switch only, d
| |
| : d. Control Room control switches are inoperable.
| |
| 21 ne Control Room is uninhabitable and the Remote Sh'utdown Panels are being utilized to control the plant. Reactor levelis 20 inches and lowering, and Reactor Pressure is 500 psig.
| |
| With present plant conditions, which of the following systems can be utilized to raise reactor level from the Remote Shutdown Panels?
| |
| : a. LPCS
| |
| : a. RHR A
| |
| : c. RCIC
| |
| : d. IIPCS
| |
| ~
| |
| 9
| |
| | |
| _. - ._. .. . - __. - - . .. = -_
| |
| 22 With the reactor at 100% power, a loss of all Reactor Plant Component Cooling Water occurs.
| |
| What are 'he required opera'or actions?
| |
| : a. Monitor and reduce system heat loads as neceswy to continue plant operations.
| |
| : b. Insert a reactor scrrers and shift both recirculation pumps to slow speed.
| |
| : c. Commence a reactor shutdown per GOP-0002, Plant Shutdown.
| |
| : d. Insett a reactor scram and trip and isolate both recirculation pumps.
| |
| 23 ne plant is conducting a startup with reactor power at 22%, when an unisolable rupture in the Terbine Plant Component Cooling Wate-(TPCCW) suction header causes a complete loss of TPCCW. What are the required immediate operator actions?
| |
| : a. Conduct an orderly reactor shutdown per GOP-0002. Plant Shutdown,
| |
| : b. Initiate RCIC and shutdown the Feed' water pumps.
| |
| : c. Reduce reactor power to within Bypass Valve capacity then trip the Main Turt,ine,
| |
| : d. IV mually scram the reactor.
| |
| 24 A plant starup is in progress. R'eactor power is on IRM Range 7 and reactor pressure is 450 psig when the "A" CRD pump trips. De "B" CRD pump will not start. ,j A reactor scram is required if:
| |
| : a. A control rod receives an IICU accumulator fault and cannot be inserted.
| |
| : b. More than one CRD high tempenture alarm is received
| |
| : c. No CRD pumps can be restarted with:n 5 minutes.
| |
| : d. Two or more accumulator faults exist.
| |
| 10
| |
| | |
| 25 Following a small break L.OCA, indicated wide range reactor level is 20" and slowly increasing due to RCIC injection from the CST. Other plant parameters are as follows:
| |
| - RPV pressurc $50 psig
| |
| . Suppression pool temp 140 deg. F Containment pressure 4.0 psig The MINIMUM suppression pool level which will anure adequate heat capacity is:
| |
| a.15.4 ft b.16.3 ft c,18.3 ft d.19.3 ft 26 Which of the following plant conditions are ALL of the reactor water level indicators invalid?
| |
| : a. RPV pressure 60 psig Dry.< ell temperature EL 145' 200 degrees F
| |
| : b. RPV pressure 90 psig Drywelltemperature EL I45' 300 degrees F
| |
| : c. RPV pressure 100 psig Containment temperature EL 119' 360 degrees F d
| |
| : d. RPV pressure 1000 psig Containment temperature EL i19'
| |
| * 180 degrees F 27 Which one of the following describes a cause and the exp;cted inaccurate response of reactor level f instrumentation indications when in the UNSAFE region of the RPV Saturation Curve?
| |
| : a. liigh containment temperatures will result in boiling of the reference legs causing l
| |
| an erroneously high level indication.
| |
| I
| |
| : b. liigh reactor pressure will result in boiling of the reference leg.$ causing an erroneously low levelindication.
| |
| : c. Low reactor pressure will result in boiling of the reference legs causing an erroneously low levelindication.
| |
| : d. liigh dry,well temperatures will result in boiling of the variable legs causing an erroneously low level indication.
| |
| I1
| |
| | |
| 28 Suppression pori tu:1 is offscale high.
| |
| %W or.e of the followin; describes the efrect on indicated containment or drywc!! pressure?
| |
| : a. Indicated containment pressure is less than actual.
| |
| : b. Indicated containment pressure is greater than actual.
| |
| : c. Indicater drywell pressure is less than actual.
| |
| : d. Indica' rd drywell pressure is greater than actual.
| |
| 29 In Emergency Depressurization, Step ED-3 asks,"Is Suppression Pool Level Above 13 P' What is the significance of this level?
| |
| : a. It ensures e vortex will not be created when SRVs are opened.
| |
| s
| |
| : b. It ensures the SRV discharge quencher is covered, so direct pressurization of containment &cs not occur.
| |
| : c. It is required to prevent loss of NPSi{ to the RIIR Pumps.
| |
| : d. It ensures there is enough water to cover the horizontal vents.
| |
| s 30 Which of the following unisolable system failures (taptures outside of primary containment) would '
| |
| * constitute a primary system for purposes of EOP 3, Secondary Containntdtt Control?
| |
| : a. Main steam drain lines in the Main Steam Line Tunnel a
| |
| : b. Containment vent line to drywell
| |
| ' c. Cervice Water supply to drywell coolers
| |
| : d. LPCS s.'.. ion line from the suppression pool 12
| |
| | |
| 1 l
| |
| 31 A high radiation alarm exists on the Annulus ventilation system (RMS*REllB). You are monitoring the CRT bar chart display for RMS'REll A to validate the alarm condition on RMS'REllB. The 10 minute trend data for RMS'REll A is colored " light blue". Which of the '
| |
| following describes the status of RMS'REll A data' readings?
| |
| : a. PMS* REll A is reading within 1% of RMS'REllB.
| |
| : b. RMS' REIL A is in an Alert condition.
| |
| : c. RMS'REll A data is " questionable".
| |
| : d. communication has been lost between RM 80 and RM 23.
| |
| 32 EOP-3, Radioactive Release Control, has been entered.
| |
| Which of the following is the tenson that the operator is directed to ensure that the Turbine Building Ventilation fans are running?
| |
| : a. Reduce radioactive releases below Ge'neral Emergency levels
| |
| : b. Prevent radioactive releases from the Turbine Building.
| |
| : c. Filter radioactivity from the Turbine Building atmosphere,
| |
| : d. Provide a monitored release point.
| |
| J 33 Following a complete loss of shutdown cooling, temperature readings indicate a 1 degree F increase .
| |
| in bulk water temperature every 10 minutes. Assume the reactor vessel head is on, no other parameters change, and current temperature is 124 deg. F.
| |
| Which of the following is the minimum amount of time before primsry containment MUST be established?
| |
| : a. 160 minutes
| |
| : b. 560 minutes
| |
| : c. 580 minutes
| |
| : d. 760 minutes 13
| |
| | |
| - . . . - . _ _ -. ~ _ . _ _ . - . _ _ _ _ _ _ _ . _ _.._._.___ _._ _ _ _._ _ _ _._ _.
| |
| 1 5
| |
| i 34 During refueling, the leakage rate of the Refueling Cavity has exceeded the capacity of the Drywell '
| |
| and Containment Equipment ud Floor Drain sumps. A fuel bundle is NOT in a safc storage :
| |
| location. ,
| |
| Which one of the following systems should be used for emergency makeup to the Refueling Cavity?
| |
| a.; Control Rod Drive Hydraulics
| |
| : b. ~ Condensate -
| |
| : c. Reactor Water Cleanup __
| |
| : d. CNS service connection
| |
| '35 - Given the following conditions:
| |
| - A failure to scram has occurred.
| |
| Reactor power is 20% with control rods being inserted manually.
| |
| - EOP-3, " Secondary Containmen'. Control" has been entered due to HVAC cooler high differential temperatures caused by a fire in the Auxiliary Building.
| |
| MSIVs have closed.
| |
| Condensate /feedwater is maintaining water level.
| |
| Which of the'following systems should be isolated? .
| |
| : a. Feedwater -
| |
| l J
| |
| : b. Reactor Water Cleanup #
| |
| ! .. c, Control Rod Drive
| |
| : d. Fire Suppression systems l
| |
| i I
| |
| i N
| |
| H
| |
| ._.a . - . _ , . _ . - . . - - - _ . . . _ _ . , - - _ . _ - - _ _ _ . - _ _ - , . _ - - - . . . . _ .
| |
| | |
| i i
| |
| . 36 EOP 3, Secondary Containment and Radioactivity Release Control, must be entered if the
| |
| - Secondary Containment differential pressure is above the maximum normal operating dJfferential
| |
| - pressure.
| |
| Which one of the following is the reason for this entry condition?
| |
| : a. A significant steam leak into the secondary containment is indicated.
| |
| : b. A significant water leak from primary system may be discharging radioactivity directly to the secondary containment.
| |
| : c. A potential for the loss of secondary containment is indicated that could result in uncontrolled radioactive releases.
| |
| : d. An increase in the unmonitored ground level radioactive releases due to leakage through secondary containment is indicated.
| |
| 37 Which one of the following would be indicative of a loss of air to the on-line Control Re d Drive (CRD) flow control valve?
| |
| : a. CRD cooling water high dp.
| |
| : b. CRD high temperature,
| |
| : c. IICU accumulator low pressure,
| |
| : d. Suction filter high dp. '
| |
| J 38 Which one of the following describes the operation of the CRDM Individual Rod Scram Test switches when placed in the TEST position?
| |
| : a. Either test switch deenergizes a single test solenoid to allow the air to vent from the scram inlet and outlet valves.
| |
| : b. One test switch deenergizes RPS A pilot solenoid and the other test switch deenergizes the RPS B pilot soler.oid to allow the air to vent from the pilot valve.
| |
| : c. One test switch deenergizes a test solenoid to allow the scram inlet valve to vent and the other test switch deenergizes a test solenoid to allow the scram outlet valve to vent.
| |
| : d. Either test switch deenergizes both RPS A and RPS B pilot solenoids to allow the air to vent from the scram inlet and outlet valves.
| |
| 15 t
| |
| R
| |
| | |
| . .. , . . . - . - . . - . _ ~ , . - - .~ . .- -~.-.-. . - - . . - . - - - - - . -.
| |
| : 39. - A reactor startup is in progress and reactor pressure is 800 psig. A loss of both CRD pumps has resulted in the receipt of the CRD ACCUMULATOR TROUBLE alarm. The nitrogen pressure on one of the CRD HCUs indicates 400 psig. Which one of the following describes the effect of this condition on the CRDM when a scrarn is initiated?
| |
| : a. Accumulator pressure alone will drive the rod in,
| |
| : b. Reactor pressure alone will drive the rod in,
| |
| : c. Both reactor pressure and accumulator pressure must be combined to drive the rod in.
| |
| : d. Both reacto. pressure and accumulator pressure combined are inadeqt. ate to drive the rod in.
| |
| 40 if one of the "B" Reactor Recirculation System Flow Converter fails (resuking in zero output) with the reactor operating at 100% power, which one of the following describes what will be generated in APRM Channel B7 .
| |
| : a. A Downscale Alarm and a Rod Block.
| |
| : b. A Rod Block only.
| |
| : c. A flatf Scram signal only.
| |
| : d. A Rod Block and a Half Scram signal.
| |
| . J 41 The plant was ini'ially operating at 100% po ver A transient occurred resulting in the following conditions: ,
| |
| - RPV levelis 35 inches and stable
| |
| - Reactor power is 73% and stable Total core flow is $1.5 E6 lbm/hr. and sL%
| |
| The cause of this plant configuration was the receipt of a signal from the:
| |
| : a. EOC RPT logic,
| |
| : b. ATWS/ARIlogic,
| |
| : c. recirculation pump cavitation interlock circuitry.
| |
| : d. recirculation flow control valve runback logic.
| |
| W 16
| |
| ,r ,
| |
| : r. . .-i, .-,..w.-
| |
| ,. - , - v..-+-r, , ,~ ,. _ - _ _ _ _ . _ _ _ _ _ _ _ - - - __ _ _ _ _ _ _ _ _ _ _ _ _ _ - --
| |
| | |
| I
| |
| ~42 During va!ve time testing on RHR System A,IEl2'MOVF004A, RHR pump A Suppression Pool Suction Valve, is closed with all other valves / switches in their normal standby position when a valid LOCA signal occuts. in this condition, RHR pump A breaker will:
| |
| : a. Close and immediateiy trip because of the IE12'MOVF004A contacts in the breaker-trip circuit.-
| |
| : b. Not close because of the IE12'MOVF004A contacts in the breaker close permissive circuit.
| |
| : c. Close after IE12'MOVF004A opens automatically,
| |
| : d. Close and remain closed, while IE12'MOVF004A remains closed.
| |
| 43 The following plant conditions exist:
| |
| The reactor is in cold shutdown.
| |
| - RHR "A"is in shutdown cooling.
| |
| - ENS *SWGl E. is deenergized for mainienance.
| |
| A RPV water level transient occurs resulting in RPV water level lowering to 120". Which of the following actions will result in LPCI "A" injecting into the RPY'.
| |
| : a. Close the SDC suction valve F006A, open suction valve F004A from the suppression pool and restart the RHR A pump.
| |
| . b. Close the SDC suction valve F008, op n suction valve F004A from the suppression pool, manually open F027A and F042A, and restart the idiR A pump.
| |
| : c. Close the SDC suction valve F006A, then arm and depress Div I LPCI initiation pushbutton.
| |
| : d. Close the SDC suction valvo F006A, open the suction valve F004A from the suppression pool, then arm and depress Div i LPCI initiation pushbutton.
| |
| e
| |
| - 17
| |
| | |
| 44 The RCIC system is in Standby Lineup, but the RCIC TURBINE EXHAUST SHUTOFF valve, E51 F068, is closed for a valve stroke test. A loss of feedwater causes a low reactor water level (level 2).
| |
| Select the statement which describes how the RCIC system will respond.
| |
| : a. RCIC TURBINE EXHAUST SHUTOFF valve, E51 F068, automatically opens; RCIC system initiates and injects water into the RPV.
| |
| : b. The RCIC turbine will start and trip on high RCIC turbine exhaust pressure at 25 psig.
| |
| : c. RCIC stans and the RCIC system exhaust rupture diaphrams will rupture initiating a RCIC system isolation at 10 psig exhaust diaphram pressure.
| |
| : d. RCIC turbine does not start. RCIC TURBINE EXHAUST SHUTOFF valve, E51 F068, must be open for RCIC STEAM SHUTOFF valve, E51-F045, to open.
| |
| 45 The plant is operating at 100 % power, st'eady state. %c Control Room Operator is performing LPCS Quarterly Pump Surveillance. He LPCS pump is running in the test return to the suppression pool mode. A steam leak in the Drywell caused Drywell pressure to increase to 1.72 psid. Reactor pressure is being maintained at 950 psig by the bypass valves.
| |
| Which of the following statements describes the response of the LPCS system ?
| |
| : a. The LPCS Pump will load shed then remain in standby.
| |
| The E21 F012 (LPCS TEST RTN TO SUPP POOL) closes. j The E21 F011 (LPCS MfN FLO TO SUPP POOL) opens.
| |
| : b. The LPCS Pump will continue running.
| |
| De E21 F012 (LPCS TEST RTN TO SUPP POOL) remains open.
| |
| The E21.F011 (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| : c. He LPCS Pump will continue running.
| |
| He E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21 F011 (LPCS MIN FLO TO SUPP POOL) opens.
| |
| : d. The LPCS Fump willload shed then remain in standby.
| |
| The E21 F012 (LPCS TEST RTN TO SUPP POOL) closes.
| |
| The E21-F011 (LPCS MIN FLO TO SUPP POOL) remains closed.
| |
| 18
| |
| | |
| t 1
| |
| 46 ne llPCS system is running in test return to the CST. A Suppression Pool high level occurs. '
| |
| Which one of the following describes the Dow path of the IIPCS system?
| |
| j ~
| |
| : a. IIPCS pump suction from the suppression pool, discharge to the CST through the test retum line.
| |
| : b. IIPCS pump suction from the CST, discharg,e to the CST through the test return line.
| |
| : c. IIPCS pump suction from the CST, discharge to the suppression pool through the min now line.
| |
| : d. IIPCS pump suction from the suppression pool, discharge to the suppression pool through the min Cow line.
| |
| 47 The SLC system is in its standby readiness lineup when an ATWS occurs. nc UO is directed to initiate SLC "A" and verify it is injecting into the vessel. As the UO attempts to do this, which of the following statements identines the EXPECTED sequence of events wi:hin the SLC system?
| |
| : a. Suction valve opens fully, both squib valves fire, pump starts,
| |
| : b. Suction valve opens fully,"A" squib valve fires, the pump starts,
| |
| : c. De "A" squib valve Dres, suction valve opens fully, pump starts,
| |
| : d. Both squib valves fire, both uction
| |
| . valves open fully, pump starts.
| |
| ,j 48 ne plant was operating at 100% power with the preferred AC power sources lined up, when the Preferred Station Service Transformer IRTX XSRIE (R.S.S #1 Leads) experienced a sudden pressure lockout trip. Which of the following is the expected status of RPS power source?
| |
| : a. RPS A is de-energized; RPS B is energized from its normal supply.
| |
| : b. RPS A is de-energized; RPS B is de-energized.
| |
| : c. RPS A is energized frcm its normal supply; RPS B is energized from its attemate supply.
| |
| : d. RPS A is energized from its alternate supply; RPS B is de-energized.
| |
| 19
| |
| | |
| ~
| |
| O Given the following plant conditions:
| |
| IRM "G* is bypassed IRM "II" is typassed Reactor Mode switch is in START /ilOT STANDBY All operable IRMs are reading 65/125 on Rar:ge 9 During the troubleshooting of IRM *G", l&C requests that the operator at the controls withdraw 1RM *G". The NCO withdraws IRM *C" by mistake. Ilow does the plant respond?
| |
| : a. A " detector not full in" rod block is generated in RC&lS.
| |
| : b. A downscale rod block is generated when the detector gets full-out.
| |
| : c. Nothing, at this time all rod blocks and scrams are cypassed,
| |
| : d. An INOP rod block and 1/2 scram signal are generated when the detector leaves the ' full in" position.
| |
| I P, 50 A reactor startup is underway, with the $10DE SWITCll in STARTUP. SRMs are being withdrawn to maintain count rate per procedure, and power in the intermediate range. Which of the following conditions would generate a rod block?
| |
| i n. SRM *A" fails low (pegged downscale), all other 1RM's are on range 3.
| |
| : b. SRM "C" fails high (pegged upscale), IRM "G" is on range 8, all other IRM's are en i range 9. ,j
| |
| : c. SRM "B" fails low (pegged downscale), IRM "F" is on range 2, all other IRM's are on range 3.
| |
| i SRM "D" fails high (pegged upscale), all other IRM's are on reng2 9.
| |
| .=
| |
| 4 20
| |
| | |
| 51 ne reactor has been operating near rated power for 200 days. Which one of the following describes the change in :he indicated LPRM output signal from day I to day 200 and the method used to calibrate the LPRMs?
| |
| INDICATED LPRM POWER METiiOD OF LPRM CALIBRATlON
| |
| : a. Decreases Core ticat Balance
| |
| : b. Decreases TIP System Trace
| |
| : c. Increases Core liest flataace
| |
| : d. Increases TIP System Traco 52 A transient has occurred requiring inserting a manu)1 reactor scram. ARI was required to be initiated by the ATC operator to insert all control rods.
| |
| De following current plant conditions have been steady for the last 5 minutes:
| |
| . All rods are inserted.
| |
| . Reactor power is 0%.
| |
| . RPV pressure is 1100 psig.
| |
| RPV water level is .50" and steady.
| |
| Given these conditions, when can ARI be reset?
| |
| : a. 32 seconds after ARIinitiation.
| |
| J
| |
| : b. Immediately by depressing the ARI initiation Reset pushbutton.
| |
| : c. When reactor pressure is lowered to 1060 psig.
| |
| : d. When RPV water levelis restored above 43".
| |
| 53 Shortly after initiation, the RCIC turbine tripped on high steam exhaust pressure. ne RCIC turbine trip throttle valve may be reset:
| |
| : a. using the control room trip throttle valve hand switch after the tn,.
| |
| condition is cleared.
| |
| : b. locally at the turbine after the trip condition is cleared.
| |
| : c. locally at the turbine at any time even if the condition is not cleared.
| |
| : d. in the control room using the trip reset push button on panel P601, i once the condition is clear.
| |
| 21 l
| |
| t
| |
| | |
| ~ .. -. _
| |
| 54 Following a valid ADS initiation, the operator is directed to close the ADS valves with the initiating signals still present. Which one of the following operator actions will ceuse the ADS valves to close?
| |
| : a. place the control switches on lil3 P601 and lil3.P631 for the ADS valves to the l "OFF" position. l
| |
| : b. Place the ADS inhibit switches on lil3P-601 to the " NORMAL" position.
| |
| : c. Stop all low pressure FCCS pumps in both Div. I and Div. 2.
| |
| : d. Depress both " ADS Timsr/ Level 3 Seal in Reset" pushbuttons, Sl3 A(B) 55 A Safety Relief Valve (SRV) tallpipe vacuum breaker was failed in the open position when the SRV opened. Which of the following is the result?
| |
| : a. Containment pressure increased.
| |
| : b. Steam bypassed the quenchers with a direct discharge path into the suppression pool.
| |
| : c. Drywell to containment differential pres;;re increased.
| |
| : d. Suppression pool water will be drawn up into the SRV discharge line after the SRV is closed.
| |
| 56 Select the statement that describes an operator action required for the following plant conditions:
| |
| Reactor power: 75 %
| |
| Suppression pooltemperature: 105 degrees F and rising Suppression poollevel: 19 feet 8 inches SRV IB21 F047B: Failed open
| |
| : a. If the SRV cannot be closed within five minutes, place the reactor mode switch in SilUTDOWN.
| |
| : b. If sappression pool temperature exceeds 120 degrees F., arm and depress the manual scram pushbuttons.
| |
| : c. place uie reactor mode switch in SilVTDOWN.
| |
| : d. Reduce suppression pool temperature to less than 100 degrees F within 1 hour.
| |
| 22
| |
| | |
| i 57 Consider the following plant conditions:
| |
| Reactor power: 45%
| |
| Generator load: 410 MWe
| |
| . Recirculation now control: Loop Manual SELECT the plant response to a continuous runback of the load set demand sigaal to zero Electro.llydraulic Control (EllC) system.
| |
| : a. Turbine control valves (TCVs) throttle closed, bypass valves (DPVs) remain closed, reactor pressure increa:es, reactor scrams on high pressure or high neutron Dux.
| |
| : b. 1)ypass valves (DPVs) throttle open, reactor pressure decreases, MSIVs isolate on low steam line pressure, ,
| |
| reactor scrams on the MSIV closure.
| |
| : c. Turbine control valves (TCW) throttle closed, bypass valves (13PVs) throttle open to compensate, once the DPVs are fully open reactor pressure increases causing a scram on high pressure or neutron Dux.
| |
| : d. Turbine control valves (TCVs) throttle closed, bypass valves (IIPVs) throttle open, reactor pressure remains fairly constant, reactor power increases slightly due to reduced feedwater heating.
| |
| J 58 Which of the following conditions / actions involving the Feedwater Level Control System, w cause actual vessellevel to lower.
| |
| : n. Loss of one feed Dow input at 100% power and in automatic three element control.
| |
| : b. Loss of a feed now input at 5% power and in automatic startup level control.
| |
| : c. Switching from the "A" level instrument (indicating 36") to the "B" instmment (indicating 39") when in three element, automatic control at 100% power.
| |
| : d. At 100% power, placing one FRV in manual control and opening it manually, with the other FRVs in three element, automatic control.
| |
| 23 i
| |
| | |
| .. . .~ . . - . . . - . . _ _ _ _ - - - - . - - - - . _ - _ - _ - . . - - _ _ . - _ _ _ _
| |
| 4 59 1he plant is at 80% power. Two condensate pumps (A and B), and two feedwater pumps (B and C) are in service, if a loss of both condensate pumps occurs, FEEDWATER PUMP ,
| |
| 1 l
| |
| : a. *B" trips 15 seconds after suction pressure decreases to 270 psig. j
| |
| : b. *D" trips 40 seconds after suction pressure decreases to 270 psig.
| |
| : c. "C" continues to run indefinately on heater drain pump discharge pressure,
| |
| : d. "C" trips 60 seconds after suction pressure decreases to 270 psig.
| |
| 1 60 A plant startup is in progress. Reactor power is being held at 1% power for the 900 psig Drywell walkdown when the Startup Feedwater Reg. Valve drifts fully open.
| |
| Which of the following actions / signals will occur as a result of this failure?
| |
| (NOTE: Assume no operator action.)
| |
| : a. Reactor scrams on high reactor water' level.
| |
| : b. Reactor feedwater pumps trip on high reactor level.
| |
| : c. Rea, or water level remains unchanged due to compensation by the Long Cycle Cleanup LevelController (CNM 104).
| |
| : d. Reactor water level stabilizes at a new higher level.
| |
| J 24
| |
| | |
| 61 Olven the following plant conditions:
| |
| A LOCA has occuned.
| |
| Reactor water levelis.50 inches.
| |
| Dryuell Pressure is 1.12 psid.
| |
| All radiation monitors are indicating normal for plant conditions.
| |
| "A" Standby Gas Treatment System (SOTS) is running and the "B" SBOT system has been shutdown and placed in STANDDY.
| |
| Which of the following will be the status of SDGT systems "A" and *B" five (5) minutes aner ENS.
| |
| SWOI A receives a degraded Bus Undervoltage Signal? (All associated systems respond normally)
| |
| : a. Both systems are shutdown.
| |
| : b. Both systems are running.
| |
| : c. I "A" system is shutdown and "B" system is running.
| |
| )
| |
| : d. "A" systerr % W; .% ud *B" system is shutdown.
| |
| 62 ne following conditions exist:
| |
| ne Div 2 standby diesel generator is loaded and in parallel with bus IENS*SWOl8 through the l normal breaker, A LOCA signal occurs.
| |
| Which of the following describes the effect of the standby diesel generator and bus LENS *SWOID7
| |
| : a. He normal bus supply breaker will open and th'e diesel generator wil} supply bus loads,
| |
| : b. ne nonnal bus supply breaker and diesel generator output breaker will opan, then aner loads are stripped, the diesel generator output breaker will reelose ,
| |
| : c. ne diesel generator output breaker will open and cannot be closed as long as bus voltage is supplied by the normal or attemate feeders until the LOCA signal is reset.
| |
| : d. He diesel generator output breaker will remain closed in parallel operation with the bus.
| |
| l l
| |
| 25
| |
| | |
| - - _ . __ - .- _. .. - . - _ - _ _ - - = ~ _ _ - _ _ . - _ . . - _ _
| |
| 63 The plant is operating at 60% pow er for rod pattern adjustment. Rod 22 43 is to be inserted from notch 48 to notch 32. A few seconds after the rod is inserted to notch 42 a ROD DRIIT annunciator is recieved. While conducting your immediate actions, you observe rod 22 43 passing notch 46 and observe it to stop at notch 48.
| |
| The Control Rod Movement Sequence withdraw limit is notch 48. Reactor power has retumed to 60%. The correct actions will be:
| |
| : a. Place the Mode Switch in SifUTDOWN.
| |
| : b. Notify a Reactor Engineer and fully insert the rod and determine whether the rod has a stuck collet.
| |
| : c. Since pow er is below ilPSP the Rod Wkhdraw Error analysis is affected and the rod should be inserted to notch 42 w!thout delay.
| |
| : d. Notify a Reactor Engineer, declare rod 22 43 inoperable, and adjust the pattern as needed for flux shaping with rod 22 43 full out since it will not remain inserted.
| |
| 64 During a weekly control rod operability test, the ATC operator is unable to insert rod 10-47 from position 48. CRD parameters are normal and no rod blocks are present. Which of the following actions should be takem
| |
| : a. Depress ''IN TIMER SKIP" to insert control rod 10-47 to the required position.
| |
| : b. Attempt the double clutching method to free the rod; if unsuccessful, re.lse drive ,
| |
| w: ster pressure to a maximun of 350 psid to move control rod 10-47.j
| |
| : c. Insert control rod 10-47 to the required position by incrementally raising drive pressure to a maximum of 350 psid.
| |
| : d. Declare rod 10-47 INOPERABLE, SCRAM and disarm it then verify compliance with the remaining constraints of Technical Specifications regarding inoperable controf rods.
| |
| 26
| |
| | |
| , 65 The plant is operating at 100% pow er. Both Recirc Flow Control Valves are in Flux M i Auto) at 67Malve position. A leak in the Drywell has caused Dryw ell Pressure to increase to approximately 1.75 psid. Following the high drywell pressure signal, the ''B" Reactor Fee Trips and level decreases to + 14.5 inches and stabilizes. Which of the following describes the response of the Recite Flow Control Valve?
| |
| Flow Control Valves will:
| |
| : a. tunback to 22 % valve position.
| |
| : b. go to " min" position.
| |
| : c. move to a position to provide 60 % core flow,
| |
| : d. remain at 67 % valve position.
| |
| 66 The Division 111 diesel has run for 30 minutes unloaded following an inadvertent start of the diesel.
| |
| What is the cettect method to conduct a normal diesel shutdowti?
| |
| : a. Parallel the diesel with offsite, load it to > 1930 KW for > 30 minutes, then unload and cool it down for at least 4 minutes and stop the engine.
| |
| : b. Shut down the engine. It is adequately cooled down.
| |
| : c. - Parallel the diesel with offsite, load it to 2500 2600KW for I hour, then trip the output breaker and shut down the engine,
| |
| : d. and Parallel stop thethe diesel with ofTsite, load it to > 1930 KW for > 30 minute} then unload engine. '
| |
| 67 River Bend is operating at 90% power with recirc in flux manual when the discharge isolation valv for the *B" recirculation pump commences to stroke closed. Which of the following statements describes the expected response of the recirculation system to this event?
| |
| 1he Recirc Pump "B":
| |
| : a. trips to ofi.
| |
| : b. transfers to slow speed.
| |
| : c. contiues to run in fast speed; core flow and power decrease.
| |
| : d. hydraulic power unit locks up.
| |
| 27 s
| |
| | |
| _. . . - - ._ .__. . - _ . - . _ _ . . ~ = _ _ _ _ - - - . ,_. . _ = _.. .
| |
| , 68 Which one of the following INDICATIONS vould you expect to see as a result o Riser Failure"? Assume RECIRC FLOW CONTROL is in FLUX MANUAL.
| |
| : a. A DECREASE in reactor (APRM) power, a DECREASE in the Failed Jet Pump Flow, and a DECREASE in Core Differential Pressure,
| |
| : b. A DECREASE in core differential pressure, an INCREASE in Reactor Power, and an INCREASE in Indicated Core Flow.
| |
| : c. A DECREASE in failed Jet Pump flow, an INCREASE in indicated Core Flow, and a DECREASE in Core Differential Pressure,
| |
| : d. An INCREASE in indicated core flow, an INCREASE in Failed Jet Pump Flow, and an INCREASE in Reactor Power.
| |
| 69 Given the follnwing conditions:
| |
| The Reactor Water Cleanup (RWCU) system is operating in the normal mode.
| |
| De RWCU lsolation Bypass Switches (E31 S! A,B) on P632 and P642 have been placed in
| |
| * Bypass".
| |
| Select the expected effect on the RWCU system.
| |
| : a. He RWCU system isolation on hign non regenerative heat exchanger outlet te;aperature is defeated.
| |
| ^
| |
| : b. De RWCU system isolation from high area temperature ONLY are defeated.
| |
| ,j c.
| |
| He RWCU system iso!ation from high differential flow AND high area temperature are defeated,
| |
| : d. All RWCU systein isolation signals are defeated.
| |
| 28 3,
| |
| -.e ,__m--
| |
| | |
| l 70 Which of the following describes the optimal lineup for RWCU suction flow during normal full j power operation?
| |
| : a. Flow should be half from each Recirc line the bottom head ;ction should be closed when the reactor is pressurized.
| |
| : b. Bottom head drain line flow should not exceed 1/3 of total RWCU flow, if the recite loop suctions are available,
| |
| : c. Flow should be entirely from the bottom head drain line ifit is available.
| |
| : d. Suction should be from the recire lines if recirc pumps are in slow speed, and from the bottom head drain line if recirc pumps are in fast speed.
| |
| 71 The RilR S/D Cooling Isolation Valve Enable / Disable switch on the local panel (P001) has two positions," Enable / Disable".
| |
| Which of the following describes when t}}e switch is REQUIRED to be in " Disable" and the effect on the operation of Sht.tdown Cooling when it is in this position?
| |
| The RilR S/D Cooling Isolation Valve Enable / Disable switch is placed in "Disab!c" when:
| |
| : a. reactor pressure is greater than 135 psig and prevents operation of the RIIR Shutdown Cooling Inboard Isolation Valve (F009) from the Main Control Room.
| |
| : b. reactor pressure is greater than 135 psig and prevents operation of the RiiR Shutdown Cooling Outboard Isolation Valve (F008) from the Main Cpntrol Room.
| |
| : c. evacuating the Main Control Room to allow local operation of the RifR Shutdown Cooling inboard Isolation Valve (F009).
| |
| : d. evacuating the Main Control Room to allow local operation of the RIIR Shutdown Cooling Outboard Isolation Valve (F008).
| |
| 29
| |
| | |
| .. - - - - - - - _ . - - . . - . - - - . - . _ . ~ . - - _ _ - - - - .
| |
| 72 ne following conditions exist:
| |
| RilR shutdown cooling loop A is in operation. Reactor water levelis 75 inches.
| |
| Under these conditions, misoperation of which of the following valves from the main control room has the potential to inadvertantly drain the reactor vessel?
| |
| : a. Suppression Pool Suction Valve IE!2'F004A.
| |
| : b. Shutdown Cooling Suction Valve IEl2*F006A.
| |
| : c. Test Retum 's o Suppression Pool Valve IE12'F024 A.
| |
| : d. Shutdown Cooling Outboard Isolation Valve IE12*F008.
| |
| -73 ne plant is operating at 75% power. The Control Room Operator places the Outboard MSIV .
| |
| Positive Leakage Control System switch to OPERATE. Which of the following will prevent the :
| |
| Outboard MSIV Positive Leakage Control System from initiating?
| |
| : s. A LOCA signal on either high drywell pressure or low reactor water level is not present.
| |
| i
| |
| : b. De required main steam line pressure and reactor pressure requirements have not been met. .
| |
| : c. .The post LOCA 20 minute timer has not timed out.
| |
| d, All Main Steam isolation Valves have not been fully closed.
| |
| 74 De plant was operating at 100% power when a large steam leak on the MSR A reheat steam line required reheat steam to be isolated. Five minutes later the 'IURBINE HIGH VBRATION annunciator on P870 alarms. You check the recorder and observe bearing #5 reading 9 mils, bearing # 6 reading 13 mils, and bearing # 4 reading 5 mils, nese readings appear to be steady.
| |
| You should:
| |
| : a. Continue to monitor the vibrations and initiate a reactor scram and tr;p the turbine if 15 mits is reached.
| |
| : b. - Commence a rapid load reduction then take the turbine offline (scram and turbine trip) If vibrations cannot be reduced below 10 mils within 14 minutes,
| |
| : c. Immediately scram the reactor and trip the turbine.
| |
| : d. Monitor bearing temperatures and scram the reactor and trip the turbine if bearing temperature exceeds 240 degrees F.
| |
| 30 L
| |
| ,.--.,---a,_ - . . - . _ . - , _ . - ~ . _ _ . - -. . - - _ - -- _ . - . - - . - . .
| |
| | |
| 7$ A startup of the Main Turbine is being perfonned. He Main Turbine is at 60 percent of rated speed, when a loss of 125 VDC Trip Circuit Pow er is experienced. Wl(ICli ONE (1) of the following describes the required operator action (s)?
| |
| : a. Enter AOP-0002, Main Turbine and Generator Trips, due to trip of the Turbine.
| |
| : b. Verify that 24 VDC ETS power is available and continue the startup of the Main Turbine I AW SOP-0080, otherwise manually trip the Main Turbine.
| |
| - c. Allow the Main Turbine to accelerate to greater than 90 percent of rated speed, at which time the 12$ VDC Trip Circuit is no longer required because the PMG is supplying the trip circuitry.
| |
| : d. The start-up of the Main Turbine may continue, but at least one 125 VDC bus must be restored prior to synchornir.ing the generator to the grid.
| |
| 76 An MSIV closure resulted in a reactor scram, ne pressure transient caused a small steam leak in the drywell. De following conditions exist:
| |
| Reactor pressure is at 900 psig.
| |
| kcactor Level is at .80 inches wide range Drywell pressure is 2.1 psid
| |
| . Containment pressure is 0.3 psig Lowest recorded ENS'SW0l A Ilus voltage was 3952 volts.
| |
| Which one oithe following would be in service as indicated? ,
| |
| (NO OPERATOR ACTICN TAKEN) J
| |
| : a. DIV I D/O running unloaded.
| |
| : b. DIV 11 SSW with now through the "ll" Containment Unit Cooler,
| |
| : c. Drywell units coolers running with no cooling now.
| |
| : d. LPCS injecting to the RPV.
| |
| 31
| |
| | |
| 77 With the plant at 100% power, a loss of VDN.PNL0lBI has resulted in a loss of power to the feedwater Level Control System giving a Feed Reg Valve control signal failure.
| |
| The power loss also caused both Reactor Recirc pumps to shiR from fast speed to slow speed and the B Recire Flow control valve to lockup. Which plant response would result fiom these failures?
| |
| (Assume no operator actions.)
| |
| : a. The "D" Feed Reg Valve would fait closed and the "A" and "C" Feed Reg Valves would AUTO OPEN to compensate. Reactor powei will stabilize at a lower power level with both Recite pumps in slow speed.
| |
| : b. All 3 Feed Reg valves will fail open. RPV level will raise to $1" which will initiate a reactor scram, Turbine trip, and Feedwater pump trip,
| |
| : c. All 3 Feed Reg valves will fail closed. Reactor power willlower when Recirc pumps down shiR and RPV level will lower to 9.7" which will initiate a reactor scram. iiPCS and RCIC will initiate at Level 2 and restore RPV level ,
| |
| : d. All 3 Feed Reg valves will fall"as is". Reactor power will lower when Recirc pumps down shift and RPV level will raise 19 $ 1" which will initiate reactor scram, Turbine trip, and Feedwater pump trip.
| |
| 78 1he llPCS 125 VDC Switchgear E22.S001 is being supplied by the Backup Battery Charger BYS-CilGRID. Which one of the following describes the llPCS 125 VDC Switchges status if a LOCA occurs?
| |
| : a. The backup battery charger supply breaker will trip leaving the IIPCS battery as the
| |
| ,)
| |
| only power soufre.
| |
| : b. The backup battery charger will continue to supply the IIPCS DC system loads and the battery will act as a backup.
| |
| : c. The backup battery charger will be load shed from its non safety related power l
| |
| supply and must be manually aligned to a 480 VAC Standby bus.
| |
| l
| |
| : d. The llPCS DC system will be supplied by the !!PCS battery with the backup battery charger available if a low battery voltage condition is present.
| |
| l l
| |
| 32 l
| |
| t
| |
| | |
| __.__._.,-..____m.____..._ _ . . _ _ . . . . . _ _ . . _ _ . _ _ _ . - _ . . _ . _ _ . . _ -
| |
| l l
| |
| l 79 ne following conditions exist:
| |
| [
| |
| The reactor is at 100% power. ,
| |
| ne Off Ons Post Treatment Hi Hi til radiation alarm (P601/22A/A03) has occuned.
| |
| nc Offgas System automatically isolated. (IN64 F060 Off Gas Discharge to Vent valve is closed). !
| |
| Which of the following actions is PROHIBITED 7
| |
| : a. Purge the Off Gas system with service air,
| |
| : b. Shift to the Standby Off Gas Component. ,
| |
| I
| |
| : c. Reduce power as necessary to maintain condenser vacuum.
| |
| : d. Reduce power to below 60%.
| |
| Which one of the following maintains a negative pressure in the annulus 80 !
| |
| following a LOCA? .
| |
| : s. SBOT system starts and takes a suction on the Annulus Pressure Control System.
| |
| i
| |
| : b. Annulus mixing fans start and discharge to the Annulus Pressure Control System. l
| |
| : c. Annulus mixing fans start and discharge to the SBGT system. >
| |
| : d. SBOT system starts and takes a suction directly from the Annulus.
| |
| J 81 Given the following conditions:
| |
| Reactor water level is 90 inches and lowering.
| |
| Drywell pressure is 2.2 psig and raising.
| |
| l An outside fire has caused smoke in the Control Room.
| |
| De operator has attempted to manually place the Control Room ventilation in the smoke removal mode.
| |
| ' Under these conditions the Control Room Smoke Removal Damper (AOD 107/!08) will; a open and the Smoke Removal Fan will start.
| |
| : b. open but the Smoke Removal Fan will be interlocked off.
| |
| i
| |
| : c. remain closed and the Smoke Removal Fan will run on recire.
| |
| L
| |
| : d. remain closed and the Smoke Removal Fan will be inte 'xked off.
| |
| l
| |
| --~._ . _ . . _ _ _ _ . _ _ _ . , , _ _ _ , ,
| |
| | |
| l l
| |
| I I
| |
| 82 With the plant operating at 100% power, a small stator water cooling leak has required addition of makeup water weekly for the past few weeks. 'Ihe Turbine !!uilding operator is adding makeup l
| |
| water per SOP-0020 when annunciator p68015A A06, htN GEN ll2/ STATOR CLG/ SEAL OIL '
| |
| TR13L, alarms. The Turbine fluilding operator inve stigates and reports that a high coreductivity alarm has been received. Chemi:.P, .onfirms the conductivity reading > 10.2 microSiement'c:n. ,
| |
| Which of the followirig actions is required?
| |
| : a. Immediately scram the reactor and trip the turbir c.
| |
| : b. Reduce stator amps to <l650 within 20 minutcs,
| |
| : c. llave the operator verify proper operation of TPCCW cooling for the Stator Water Cooler,
| |
| : d. Commence a normal plant shutdown.
| |
| 83 'Ihe control room operwtor is about to startup RilR in the fuel pool cooling assist mode. All prerequisites have been completed and the operator starts RilR pump ' A'. lie then starts to thrott open the llX OUTLET VLV (El2 F003A). Watching the valve position indication the operator observes there is no change, also 110w is 700 gpm. Fuel pool level: (choose one)
| |
| : a. INCREASE because Suppression pool water is being diverted to the fuel Pool.
| |
| * b. REhtAIN 111E SAhtE because the RilR system is currently recircing ,
| |
| 700 gpm from/to the Suppression Pool,
| |
| : c. REhtAIN Tile SAhtE because the RilR system is currently recircing 700 gpm from/to the Fuel Pool.
| |
| : d. DECREASE because water from the fuel Pool is being diverted to the Suppression Pool.
| |
| i l
| |
| 34
| |
| | |
| I I
| |
| l 84 Given the following plant conditions:
| |
| . The Refuel Plationn is over the core.
| |
| . The Mode switch is in REFUEL.
| |
| I Which of the following will cause a control rod block?
| |
| : a. Withdraw a control rod. l l
| |
| l
| |
| : b. Loading the Auxiliary Platform hoist.
| |
| : c. Loading the Refuel Platform main hoist.
| |
| : d. Loading the Refuel Platronn monorail holst.
| |
| 85 A loss of all AC power occurred. The following plant conditions exist:
| |
| - Reactor shutdown (all rods inserted). '
| |
| . RFV pressure 1030 psig at 0700.
| |
| . RCIC system is manually started.
| |
| . There is a small cool *.nt leak into the containment.
| |
| The CRS directs an RPV cooldown at 0700. Which of the following is the lowest pressure permissible at 08007
| |
| : a. 500 psig
| |
| : b. 440 psig
| |
| : c. 400 psig
| |
| : d. O psig 33
| |
| | |
| 1 86 With the Division 1 Diesel rear air start compressor motor selnd. Which of the following actions are required?
| |
| : a. Declare Division i DieselINOPERABLE. Both starting air compressors are required for an OPERABLE air start system,
| |
| : b. Take no action since only one starting air system is necessary to start the diesel and crosstying the unaffected air system with the affected air system would render the dieselINOPERABLE.
| |
| : c. Maintain pressure in the normal band of the reciever associated with the selnd compressor by intermittently using a high pressure hose connected between the operable forward and rear system air dryer outlets. ne Division i Diesel Generator will remain OPERABLE.
| |
| : d. Start and load the Division 1 Diesel Generator. With the diesel running, the starting air system is not required for OPERABILITY of the diesel.
| |
| 87 Given the following initial conditions for'the inclined Fuel Transfer (IFTS) System:
| |
| . Tube full
| |
| . Upper opender vertical
| |
| . Carriage at upper terminal
| |
| . Lower upender inclined
| |
| . System powered up and neither bridge in the IFTS area SELECT the correct statement regarding IFTS q tration'. j
| |
| : a. He tefueling bridge can enter the IFTS area.
| |
| : b. ne fuel handling bridge can enter the IFTS area in the Fuel Building.
| |
| : c. De transfer tube can be drained,
| |
| : d. De winch can be lowered using the '' lower" pushbutton on the upper control panel.
| |
| 36
| |
| | |
| l 88 While the plant is at power, a leak develops in a. area that is acessible, but now radiologically contaminated. The OSS has directed that an investigation be performed immediately, What documentation must be generated before various personnel are allowed entry into the area for the f investigation? j t
| |
| : a. A daughter RWP to the General RWP for that area must be generated. ;
| |
| b, None, a General RWP already exists for this type of event.
| |
| c A Specific RWP must be generated. !
| |
| s
| |
| : d. None, a RWP may be completed after the entry provided it is done under continous Rp coverage, l
| |
| t 4
| |
| 89 Given the following conditions: l I
| |
| . The plant is shutdown making preparations for Shutdown Cooling (SDC) using the " A" loop of i
| |
| RilR
| |
| . Both Recirculation Pumps are shutdowt with their discharge valves closed Which of the following describes how the "A" RliR Pump that is being started for SDC is l protected from damage due to no flow? ,
| |
| s.1he operator is requi:ed to establish a pump discharge flow path to the reactor as soon as possible after starting the pump.
| |
| : b. The pump minimuni flow valve (F064 A) will open to provide flow uryll the RilR liest
| |
| - i Exchanger Bypass Valve (F048 A) can be opened.
| |
| : c. The operator will open the minimt.m flow valve (F064 A) until shutdown cooling flow is greater than 500 gpm.
| |
| : d. The pump will automatically trip on low suction pressure if flow / pressure is not adequate for ,
| |
| pump suction.
| |
| l i
| |
| I-
| |
| +
| |
| e 37 i
| |
| , , . . , , - , ~ ,wm--. . . - - - n,,..-w',-,,rv .r. . ., , #- . - - . . _ , -- . , - , - . . . .---....-,,.--,,ws-c.cc,r,-,-,%-.. ,---,.n,, , , _ _m,.w,- - . , ,-..-ww m.
| |
| | |
| 90 Which of the following permission'notincation requirements must be met for an INTENTIONAL.
| |
| entry into Tech Spec 3.0.37 Pennission must be obtained from the:
| |
| : a. Operations Superintendent and the NRC Resident inspector notined.
| |
| : b. General Manager . Plant Operations and the NRC Resident inspector notined,
| |
| : c. ManaEer . Operations and a 1 hour report made to the NRC.
| |
| : d. General ManaScr . Plant Operations and a 4-hour report made to the NRC.
| |
| 91 The Plant is operating at 100% reactor power when a loss of feedwater heating occurs. Which one of the fotlowing is a required IMMEDIA11I action for this loss of feedwater heating?
| |
| : a. Reduce resetar power by 40 MWE with co e inuw, then reduce another 110 MWE with core now and rod insertion. .
| |
| : b. Reduce power to less than or equal to 100% rated thermal power ming core Gow.
| |
| : c. If failed fuel exists in the reactor, reduce reactor power by 495 to 500 MWE.
| |
| : d. Insert control rods in reverse order to get below the 80% rod line.
| |
| 92 When EOP-4, Emergency RPV Depressurization, permits defeating isol,dion interlocks in order to rapidly depressurize without SRVs. Which of the following MSIV isolation signals may be bypassed?
| |
| : a. Only the RPV low level I signal.
| |
| : b. Only the RPV low level I and low main steam line pressure signal.
| |
| : c. All MSIV isolation signals except for Main condenser low vacuum.
| |
| : d. All MSIV automatic isolation signals.
| |
| 38 1
| |
| | |
| l 93 Prior to reWor startup the following SRM count rates are recorded:
| |
| SRM A 2$ cps SRM B 30 cps SRMC 35 cps SRM D 15 cps At what SRM reading should the operator consult the Reactor Engineer for continued withdrawal recommendations? .
| |
| : a. 2500 cps on SRM A i
| |
| : b. 3000 cps on SRM B
| |
| : c. 300 cps on SRM C
| |
| : d. 240 cps on SRM D 94 The basis for Surveillance Requirement 3.1.5.1 for each control rod scram accumulator pressure to be verified > 1520 psig every 7 days is to ensure adequate accumulator pressure exists to: ,
| |
| : a. provide sufficient scram force. l
| |
| : b. drive control rods on a loss of CRD pumps.
| |
| : c. maintain indication in the readable range on the gauge.
| |
| J
| |
| : d. verify accuracy of control room HCU pressure indications.
| |
| 95 In EOP 1 A. ATWS RPV Control, if SRVs are cycling, the operator is directed to manually open SRVs until RPV pressure drops to 930 psig.
| |
| Which of the following is the reason for stopping the reactor pressure reduction at 930 psig?
| |
| : a. To ensure the turbine bypass valves do not have the oppprtunity to stick closed
| |
| : b. To prevent MSIVs from_ closing on low main steam line pressure
| |
| : c. To minimize the amount of steam that is sent to the suppression pool
| |
| : d. To prevent excessive loss of reactor coolant inventory
| |
| | |
| 96 De plant is operating at power and a fuel handling team is working in the Fuel Building, w hen an inndiated fuel bundle is dropped and is lying on the bottom of the spent fuel pool.
| |
| Which of the following actions should be taken?
| |
| : a. Evacuate the fuel Building j
| |
| : b. Start the second charcoal filter train of the Fuel Building HVAC.
| |
| : c. anually scram the reactor.
| |
| : d. Direct the tenm to attempt to recover the bundle with the bridge grapple.
| |
| 97 ne plant is shutdown for a maintenance outage. Work is being performed on a portion of the feedwater system by Mechanical Maintenance. De I&C foreman has received a Clevance Receipt to work within the feedwater syster i tagout boundary to calibrate an instrument. Upon i completion of work, the Mechanical Maintenance foreman wishes to release his clearance and restore the system, but the instrument calibration is still taking place.
| |
| What actions (s), if any, must be taken to hnsure the safety of the personnel performing the calibration?
| |
| : a. De Clearance form is transferred to the IAC foreman.
| |
| : b. De Clearance can be released with verbal permission from the l&C foreman.
| |
| : c. The I&C foreman must return his Clearance Receipt to the tagging ofYicial prior to releasing the clearance.
| |
| d
| |
| : d. De Mechanical Maintenance foreman may clear all tags that pertain to the I&C work.
| |
| 98 An ATWS event has occurred and SLC injection has been initiated. EOP I A, "RPV Control ATWS," directs RPV water level be lowered by terminating and pieventing feedwater injection into the RPV. What are the required actions by the ATC operator to terminate and prevent feedwater injection into the RPV?
| |
| : a. Shut the feedwater inboard isolation valves IB21*F065A(B).
| |
| : b. Dial the Master Feedwater Level Flow Controller setpoint down to 0".
| |
| : c. Place the Master Feedwater Level Flow Controller in manual and drive all Feedwater Regulating Valves full closed,
| |
| : d. Trip all 3 Reactor Feedwater pumps.
| |
| 40 w...-. .~r ~.~, . - - ,- . , . . - w---,,.m--.---.---.----.-,,---..-.---,-n----,.--,,,---.m, .--r-,-.--.,-,_,w y v. ..--
| |
| | |
| . _____ _ _ . _ . ._ _ _ _ _ _ _ . _ _ _ ___..__.___________________._._.._.___m..__._..__
| |
| s 99 A 28 year old nuclear worker has been contracted to work during the upcoming refuel outage. His .
| |
| NRC Form 4 is urant, and his accumulated whole body dose to date is 48.25 rem, what is the '
| |
| MAXIMUM radiation dose he can be authorized to receive during the next quarter IAW regulatory limits?
| |
| 3 ;
| |
| : a. 0.75 rem i l
| |
| : b. 1.00 tem
| |
| : c. l.25 rem
| |
| : d. 3.00 tem !
| |
| 100 Plant conditions are as follows:
| |
| . SuPriession pool temperature 87 degrees F f
| |
| . Suppression poollevel 20.5 A.
| |
| . Drywell temperature 125 degrees F
| |
| . Reactor level 10.5 inches .
| |
| . Aux. Bldg.gwessure 0.25 [
| |
| i Which one of the following EOP selections should be entered? ,j
| |
| : a. COP 1 and EOP 2
| |
| : b. EOP 2 only
| |
| ~
| |
| : c. f[OP 2 and EOP 3 j
| |
| I
| |
| : d. EOP 3 only j 4
| |
| l i
| |
| i f
| |
| I l
| |
| O 0
| |
| 41
| |
| + . _ ~ ..._ . . . _ _ _ . _ _ _ _ _ _ _ . - _ _ . _ _ _ _ _ _ _ _ _ . _ . . _ . _ . _ _ ., ..~. _ ._. _ _ _ . _ _ _ . _ _ . . . _ . , . . -
| |
| : u. , ., e. w . <
| |
| ,; ,t
| |
| ; nv. $q;7;h.:e, :ljRO and SRO (I) pl
| |
| [j[Nd hfdhyidi Job Performance Measures qf,;4dsf6 l
| |
| iig.n.M$._w.,fj nf i l
| |
| b Pc ormance Measures i
| |
| I' RO [
| |
| Administrative Topics !
| |
| t 1
| |
| 4 i
| |
| SRO I
| |
| Administrative Topics l
| |
| l Administrative l.
| |
| Job Performance Measures i R
| |
| e l M I M
| |
| l I M
| |
| I, 4
| |
| IIM ll
| |
| | |
| 1 ES 301 Individual Walk through Test Outline Form ES 301-2 Facility: River Bend fExamination: 7/28/97 Exam Level (Circle One): RO/ SRO ) / SRO (U) Operating Test No.:
| |
| System / JPM / Type Safety Planned Follow-up Questions:
| |
| Codes
| |
| * Function K / A / O -Importance - Description i 1.JPM 053 06, Single a. K/A 202001 K4.16 Rating 3.3 / 3.6 Recirculation Pump i Knowledge of Recirculation System design
| |
| . Shutdown and feature (s) and/or interlocks which provide for Recirculation Loop recirculation pump downshift.
| |
| Isolation at Power
| |
| : b. K/A 202001 K4.15 Rating 3.1/3.4 Knowledge of Recirculation System design feature (s) and/or interlocks which provide for N/S slow speed pump start.
| |
| : 2. JPM 110-07, n. K/A: 245000 K4,09 Rating: 3.1/3.2 Turbine Overspeed 111 Knowledge of the effect that a loss or malfunction Protection System of the Main Turbine will have on the Operability Test for Reactor / Turbine Pressure control system.
| |
| ^
| |
| Turbine Control Valve #1
| |
| .i
| |
| : b. K/A: 241000 Kl.02 Rating: 3.9/4.1 Knowledge of the physical connections and/or cause-effect relationships between the Pressure N/S Regulating System and Reactor Pressure
| |
| : 3. JPM-200-04, a. K/A: 203000 K4.14 Rating: 3.6/3.7 Manually start RilR 11 Knowledge of RilR/LPCI: Injection mode design "A"in LPCI mode feature (s) and/or interlocks which provide for from the Remote operation from the remote shutdown panel.
| |
| Shutdown Panel
| |
| : b. K/A: 203000 K4.03 Rating: 3.2/3.3 Knowledge of RHR/LPCI: Injection mode design
| |
| . feature (s) and/or interlocks which provide for M/P pump minimum flow protection.
| |
| I'
| |
| (
| |
| NUREG 1021 Page 1 of 4 (Rev.1) Interim Rev. 8, January 1997
| |
| | |
| ES 301 Individual Wolk through Test Outline Form ES-3012 i
| |
| Sy. stem / JPhi / Type Safety Planned Follow-up Questions: !
| |
| Codes
| |
| * Function K / A / G -Importance - Description !
| |
| : 4. JPM 203 02, a. K/A 2.1.24 Rating: 2.8/3.1 Shutdown the liigh IV Ability to obtain and interpret station electrical Pressure Core Spray and mechanical drawings. ;
| |
| Pump aller an inadvertent Automatic Initiation with a failure of the llPCS minimum flow valve to automatically open.
| |
| : b. K/A: 209002 A2.01 Rating: 3575.'8 '
| |
| Ability to predict the impacts of a system D/S/A . initiation on the liigh Pressure Core Spray (last NRC exam) System.
| |
| : 5. JPM 309 03, a. K/A: 264000 K4.08 Rating: 3.8/3.7 Perfonn a Non. VI Knowledge of Emergency Generators design Emergency Start, feature (s) and/or interlocks which provide for Load and Parallel of automatic startup.
| |
| the Division 1(11)
| |
| Emergency Diesel J Generator (Locally)
| |
| : b. K/A: 2.1.12 Rating: 2.9/4.0 Ability to apply technical specifications for a l M / P (last NRC exam) system.
| |
| : 6. JPM 403-03, a. K/A: 223001 A2.06 Rating: 4.1/4.1 Sta: tup Drywell IX Ability to predict the impact of high containment Purge using Standby pressure and based on these predictions, use Gas Treatment A procedures to correct, control, or mitigate the l
| |
| (GTS*FN1 A) consequences of these abnonnal conditions or operations,
| |
| : b. K/A: 261000 K4.01 Rating: 3.7/3.8 Knowledge of Standby Gas Treatment design feature (s) and/or interlocks which provide for M/S automatic system initiation.
| |
| NUREG 1021 Page 2 of 4 (Rev.1) Interim Rev. 8, January 1997
| |
| | |
| ES 301 Individual Wolk through Test Outline Form ES-3012 System / JPM / Type Safety Planned Fo: low-up Questions:
| |
| Codes
| |
| * Function K / A / G -Importance - Description
| |
| : 7. JPM 501-01, a. K/A: 259002 K4.04 Rating: 2.9/2.9 Transfer from 11 Knowledge of Reactor Water Level Control Startup Feedwater System design feature (s) and/or interlocks which Level Controller to provide for reactor water level setpoint setdown the Master following a reactor scram.
| |
| Feedwater Level Controller
| |
| : b. K/A: 259002 Kl.03 Rating: 3.8/3.9 Knowledge of the physical connections and/or ,
| |
| M/S/L cause-effect relationships between Reactor Water (last NRC exam) Level Control System and reactor water level.
| |
| : 8. JPM 80011, a. K/A: 295037 EK1.02 Rating: 4.1/4.3 Vent the Scram Air Vil Knowledge of the operational implication of licader reactor water level effects on reacto power.
| |
| : b. K/A: 2d5037 EA2.03 Rating: 4.3/4.4 Ability to determine and/or interpret Standby M/P/R Liquid Control System tank level.
| |
| .)
| |
| : 9. JPM 800-14, a. K/A: 201005 K5.10 Rating: 3.2/3.3 Defeat the RC&lS Vil Knowledge of operational implications of the Rod Interkv;ks Withdrawal Limiter as it applies to the Rod Control and Information System (RC&lS).
| |
| : b. K/A: 201005 K4.01 Rating: 3.2/3.2 Knowledge of the Rod Control and Information System (RC&lS) design feature (s) and/or interlocks which provide for limiting the effecu of a control rod accident.
| |
| M/C NUREG-1021 Page 3 of 4 (Rey,1) _ interim Rev. 8, January 1997
| |
| | |
| ES 301 Individual Walk through Test Outline Fomi ES 3012 4
| |
| ,= , u System / JPM / Type Safety . Planned Follow-up Questions:
| |
| Codes
| |
| * Function K / A / O Importance - Description !
| |
| : 10. ',PM 800-28, a. K/A: 295035 EKl.01 Rating: 3.9/4.2 Defeating liigh DW V Knowledge of the operational implications of the Pressure and Low secondary containment integrity.
| |
| RPV Water Level Containment Vent and Purge Isolation Interlocks
| |
| : b. K/A: 500000 EK2.03 Rating: 3.3/3.4 Knowledge of the interrelations between high con ainment hydrogen concentration and M/C containment atmosphere control systems.
| |
| * Type Codes:(D) direct from bank,(M) modified from bank,(N) new,(A) alternate path, (C) control room, (S) simulator, (L) low-power, (P) plant, (R) RCA
| |
| .I NUREG 1021 Page 4 of 4 (Rev.1) Interim Rev. 8, January 1997
| |
| | |
| o ES 301 Individual Walk-through Test Outline Form ES 3012 Facility: RiverBend Date of E. 'on: '//28/97 Exam Level (Circle One): RO/ SRO (1) / SRO (U Operating Test No.:
| |
| System / JPM / Type Safety Planned Follow-up Questions:
| |
| Codes
| |
| * Function K / A / G - Importance - Description
| |
| : 1. JPM 053-06, Single a. K/A 202001 K4.16 Rating 3.3 / 3.6 Recirculation Pump i Knowledge of Recirculation System design Shutdown and feature (s) and/or interlocks which provide for Recirculation Lwp recirculation pump downshift.
| |
| Isolation at Power
| |
| : b. K/A 202001 K4.15 Rating 3.1/3.4 Knowledge of Recirculation System design feature (s) and/or interlocks which provide for N/S slow speed pump start.
| |
| : 2. JPM 501-02, - a. K/A: 259002 K4.04 Rating: 2.9/2.9 Tro.tsfer from 11 Knowledge of Reactor Water Level Control Startup Feedwater System design feature (s) and/or interlocks which Level Controller to provide for reactor water level setpoint setdown the Master following a reactor scram.
| |
| Feedwater Level Controller with a ,J failure of Feed Reg.
| |
| Valve "A"
| |
| : b. K/A: 259002 Kl.03 Rating: 3.8/3.9 Knowledge of the physical connections and/or N/S/L/A cause-effect relationships between Reactor Water Level Control System and reactor water level
| |
| : 3. JPM 309-03, a. K/A: 264000 K4.08 Rating: 3.8/3.7 Perfomi a Non- VI Knowledge of Emergency Generators design Emergency Start, feature (s) and/or interlocks which provide for Load and Parallel of automatic startup.
| |
| the Division 1(II)
| |
| Emergency Diesel Generator (Locally)
| |
| : b. K/A: 2.1.12 Rating: 2.9/4.0 Ability to apply technical specifications for a M / P (last NRC exam) system.
| |
| NUREG 1021 PageIof2 (Rev.1) Interim Rev. 8, January 1997
| |
| | |
| ES 301 Individual Walk through Test Outline Form ES 301-2 r
| |
| l l
| |
| System / JPM / Type Safety Planned Follow-up Questions:
| |
| Codes
| |
| * Function K / A / G - Importance - Description 1 JPM-800-ll, a. K/A: 295037 EKl.02 Rating: 4.1/4.3 Vent the Scram Air Vil Knowledge of the operational implication of lleader reactor water level effects on reactor power,
| |
| : b. K/A: 295037 EA2.03 Rating: 4.3/4.4 Ability to determine and/or interpret Standby M/P/R Liquid Control System tank level.
| |
| : 5. JPM 800-28, a. K/A: 295035 EKl.01 Rating: 3.9/4.2 Defeating Iligh DW V Knowledge of the operational implications of the Pressure and Low secondary containment integrity.
| |
| RPV Water Level Containment Vent and Purge Isolation Interlocks
| |
| : b. K/A: 500000 EK2.03 Rating: 3.3/3.4 Knowledge of the interrelations between high contair.r.ient hydrogen concentration and M/C containment atmosphere control systems.
| |
| J
| |
| * Type Codes: (D) direct from bank, (M) modified from bank, (N) new (A) attemate path, (C) control room, (S) simulator, (L) low power, (P) plant, (R) RCA NUREG-1021 Page 2 (T2 (Rev.1) Interim Rev. 8. January 1997
| |
| | |
| t -
| |
| i RIIS JOll PERFORMANCE MEASURE l
| |
| JPM NUMBER: JPM 053 06 Revision 0 TASK DESCRIPTION: Recirculation Pump Shutdown and Recirculation Loop Isolation at Power K/A REFERENCE & RATING: 202001 A4.01 3.7/3.7 A4.02 3.56.4 202001 A47t 3.7/3.7 TASK
| |
| | |
| ==REFERENCE:==
| |
| 202008001001 Actual Performance: X TESTING METIlOD: Simulate Performance:
| |
| Control Room: X In _ Simulator:
| |
| Plant:
| |
| COMPLETION 'flME: 10 minutes MAX. TIME: N/A JOB LEVEL: RO/SRO TIME CRITICAL: No EIP CLASSIFICATION REQUIRED: No J
| |
| PRA RISK DOMLIATE: No ALTERNATE PATII(FAULTED): No SAFETY FUNCTION GROUP: 1 Prepared by: D.E. Dietzel / 1248 Date: 6/5/97 KCN Ops Validation: B. Carver / 1191 Date: 6/11/97 KCN Approved by: . D.E. Dietal / 1248 Date: _ 6/13/97 KCN JPM-053-06 Rev. O Page 1 of 9
| |
| | |
| t RBS JOB PERFORMANCE MEASURE SIMULATOR SETUP SHEET Task
| |
| | |
| == Description:==
| |
| Recirculation Pump Shutdown and Recirculation Loop Isolation at Power Required Power: IC-11 (60 - 65% power)
| |
| I C N o.: IC-11 Notes: Set up conditions reflecting a seal failure on RCP " A"
| |
| : 1) *erify reactor power is ~ 60-65%.
| |
| : 2) ior -AO B33-R602A-M 300.00 R0 494.51 680-04 B33 R602A+M (60) R'ecire Loop A Seal #2 Cavity Pressure meter
| |
| : 3) ior AO B33 R603A-M 600.00 R01003.78 680-04 B33 R603A+M (60) Recirc Loop A Seal #1 Cavity Pressure meter
| |
| : 4) ior -ANN xall680_4a_d_5 ON d 680-04-D05 RECIRC PUMP A OUTER SEAL HIGH LEAKAGE
| |
| : 5) ior -ANN xall680_4a e_5 ON 680-04-E05 RECIRC PUMP A SEAL STAGING HIGIVLOW FLOW JPM-053-06 Rey, O Page 2 of 9
| |
| | |
| -s
| |
| . RBS JOB PERFORMANCE MEASURE DATA SIIEET References fe' Development: SOP-0003, Reactor Recirculation Required Materials: SOP-0003, Reactor Recirculation Required Plant Condition: Reactor power is < 70%, Both RCP operating in Fast Speed..
| |
| Applicable Objectives: liLO-005, Obj 3 9 and 11 11LO-058, Obj 2 Safety Related Task: (If K/A less than 3.0) i Control Manipulations: N/A Items marked with an "*" are required to be performed, and are Critical Steps, failure to successfully complete a Critical Step requires the JPM to be evaluated as
| |
| " Unsatisfactory". Comments describing the reason for failure are required in the comments section of the Verification of Completion sheet. 'I Items marked with an "^" are required to be performed in the sequence described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
| |
| i l
| |
| JPM-053-06 Rev. 0 Page 3 of 9 l
| |
| I
| |
| | |
| RBS JOB PERFORMANCE MEASURE -
| |
| IfIn-Plant or In de Control Room:
| |
| Caution the Operator NOT to MANIPULATE the controls, but make clear what they would do if this were not a simulated situation.
| |
| head to the Operator:
| |
| I will explain the initial conditions, and provide initiating cues, I may provide cues during the perfonnance of this JPM, I may.ask follow-up questions as part of this IPM. When you complete the task successfully, the objective for this JPM will be satisfied, you should inform me '
| |
| when you have comp eted the task.
| |
| Initi;I Condiuons: ' A loss of CCP to the recire pump seal coolers remited in overheating the "A" Recire Pump seals. After CCP was restored, the "A" Recire Pump seals failed.
| |
| Initiating Cue: The CRS bas directed you to secure the "A" Recire Pump and isolate the " A" Recire Loop in accordance with SOP-0003, section 6.4.
| |
| PERFORMANCE STEP l STANDARD S/U l COMMLNTS
| |
| : 1. Transfer flow con'rol from flux manual to Both pushbutton depressed.
| |
| loop manual by depressing MAN Pushbutton on both B33-K603A and B33-K603B,'RECIRC LOOP A and B FLOW CONTROL M/A Stations.
| |
| * 2. Depress the STOP pushbutton for B33- STOP pushbuttN depressed.
| |
| C001 A RECIRC PUMP A MOTOR BREAKER 5A.
| |
| JPM-053-06 Rev. O Page 4 of 9
| |
| | |
| l RBS JOB PERFOkmANCE MEASURE '
| |
| l l PERFORMANCE STEP l STANDARD l S/U l COMMENTS l
| |
| - 3. Verify B33-C001 A RECIRC PUMP.4 B33-C001 A RECIRC PUMP A MOTOR MOTOR BREAKER SA opens and pump BREAKER SA opens and pump coasts down to coasts down to 0% speed. 0% speed. ,
| |
| : 4. . Open NJS-ACB305 by depressing the TRIP pushbutton depressed for B33-S001 A TRIP pushbutton for B33-S001 A LFMG A LFMG A MOT BRKR 1 A, red light off, green MOT BRKR 1 A. light on.
| |
| * 5. Close B33-F067A, RECIRC PUMP A B33-FM7A, RECIRC PUMP A DISCII VLV DISCII VLV. Closed, red light off, greu light on.
| |
| * 6. Close B33-F023A, RECIRC PUMP A B33-F023A, RECIRC PUMP A SUCTION VLV SUCTION VLV. Closed, red light off, green light on.
| |
| * 7. Close G33-MOVF100, RWCU RECIRC A G33-MOVF100, RWCU RECIRC A SUCT.
| |
| SUCT. Closed, red light off, green light on.
| |
| * 8. Place B33-IIYVF060A, Flow Control Valve B33-11YVF060X, Flow Control Valve is in CUE: The CRS has directed in minimum position. minimum position. another operator to monitor parameters and shutdown the llPU A.
| |
| Terminating Criteria: The "A" Recirculation Pump is secured and the "A" Recire Loop is isolated in accordance with SOP-0003, Section 6.4.
| |
| JPM-053-06 Rev. O Page 5 of 9
| |
| | |
| RBS JOB PERFORMANCE MEASURE JPM QUESTIONS Question i K/A: 202002 K4.16 Rating: 3.3/3.6
| |
| | |
| ==Reference:==
| |
| OPEN Specific power and flow conditions allow bypassing the Recirculation System Steam Dome / Pump Suction 8'F AT interlock? What could be an adverse affect on the Recirculation System,if these power and flow conditions are met and the 8*F AT interlocks were not bypassed?
| |
| Answer:
| |
| A spurious Recirculation Pump down shift to slow speed could occur.
| |
| Response / Comments:
| |
| 9 SAT UNSAT
| |
| | |
| ==References:==
| |
| LOTM Chapter 8, Reactor Recirculation Flow Control Question 2 K/A: 202001 K4.15 Rating: 3.1/3.4
| |
| | |
| ==Reference:==
| |
| OPEN Preparations are being made to re-stan Recirculation Pump A. The following plant conditions exist:
| |
| J e Reactor Steam Dome Temp: 545*
| |
| * Rea~ctor Vessel Drain Temp: 520'
| |
| * Recirculation Loop A Water Temp: 500*
| |
| * Recirculation Loop B Water Temp: 520'
| |
| * Recirculation Loop B Flow: 55% of rated flow For the above plant conditions, what condition must be met to permit re-starting Recirculation Pump A7 Answer:
| |
| Recirculation Loop B flow must be reduced to < 50% of rated flow.
| |
| Response / Comments:
| |
| SAT UNSAT
| |
| | |
| ==References:==
| |
| LOTM Chapter 8, Reactor Recire Flow Control, GOP-0004, Single Loop Ops.
| |
| JPM-053-06 Rev. O Page 6 of 9
| |
| | |
| RBS. JOB PERFORMANCE MEASURE VERIFICATION OF COMPLETION Operator: SSN:
| |
| Evaluator: KCN:
| |
| Date: License (Circle one): RO/SRO No. of Attempts:
| |
| Follow-up Questions:
| |
| Follow-up Question Response:
| |
| Time to complete JPM: minutes Comments / Feedback:
| |
| J f RESULT: Satisfactory / Unsatisfactory Note: An " Unsatisfactory" requires comments and remedial training.
| |
| l Evaluator's Signature: Date:
| |
| {'
| |
| JPM-053-06 Rev. O Page 7 of 9
| |
| | |
| RBS JOB PERFORMANCE MEASURE JPM QUESTIONS (OPERATOR COPY)
| |
| Question 1 (tills IS AN OPEN REFERENCE QUESTION)
| |
| Specific power and flow conc.. dons allow bypassing the Recirculation System Steam Dome / Pump Suction 8*F AT interlock? What could be an adverse affect on the Recirculation System,if these power and flow conditions are met and the 8 F AT interlocks were not bypassed?
| |
| Question 2 (tills IS Ah OPEN REFERENCE QUESTION)
| |
| Preparations are being made to re-start Recirculation Pump A. The following plant conditions exist:
| |
| . Reactor Steam Dome Temp: 545
| |
| '3
| |
| . Reactor Vessel Drain Temp: 520
| |
| . Recirculation Loop A Water Temp: 500*
| |
| . Recirculation Loop B Water Temp: 520
| |
| . Recirculation Loop B Flow: 55% of rated flow For the above plant conditions, what condition must be met to permit re-starting Recirculation Pump A?
| |
| JPM-053-06 Rev. O Page 8 of 9
| |
| | |
| RBS JOB PERFORMANCE MEASURE JPM Task Conditions / Cues (Operator Copy)
| |
| Initial Conditions: A loss of CCP to the recire pump seal coolers resulted in overheating the "A" Recirc Pump seals. After CCP was restored, the "A" Recirc Pump seals failed.
| |
| Initiating Cues: The CRS has directed you to secure the "A" Recirc Pump and isolate the "A" Recire Loop in accordance with SOP-0003, Section 6.4.
| |
| l-6 JPM-053-06 Rev. O Page 9 of 9
| |
| | |
| R l rg i
| |
| * o RBS JOB PERFORMANCE MEASURE t JPM NUMBER: JPM-110-07, Revision 0 TASK DESCRIPTION: Turbine Overspeed Protection System Operability Test for Turbine Control Valve #1 K/A REFERENCE & RATING: 241000 Kl.05 3.5/3.6 241000 Kl.08 3.6/3.7 241000 K3.08 3.7/3.7 241000 K4.05 3.7/3.8 241000 A2.04 3.7/3.8 241000 A4.08 3.5/3.4 TASK
| |
| | |
| ==REFERENCE:==
| |
| | |
| TESTING METIIOD: Simulate Performance: Actual Performance: X Control Room: Simulator: X In-Plant:
| |
| COMPLETION TIME: 10 minutes MAX. TIME: N/A JOB LEVEL: RO/SRO TIME CRITICAL: No J
| |
| EIP CLASSIFICATION REQUIRED: No l'RA RISK DOMINATE: No ALTERNATE PATII (FAULTED): No SAFETY FUNCTION GROUP: III Prepared by: D. E. Dietzel / 1248 :?;e: 6/5/97 KCN Ops Validation: B. Carver / 1191 Date: 6/12/97 KCN Approved by: D.E. Dietzel / 1248 Date: 6/1297 KCN JPM-110-07 Rev. 0 Page 1 of 10
| |
| | |
| RBS JOB PERFORMANCE MEASURE SIMULATOR SETUP SHEET Task
| |
| | |
| == Description:==
| |
| Turbine Overspeed Protection System Operability Test for Turbine ConJol Valve #1 Required Power: 70% Reactor Power I C N o.: IC-12 Notes:
| |
| : 1. Lower power to 70% using Recire Flow.
| |
| : 2. Ensure the Process Computer and printer are operational.
| |
| : 3. Verify Remote Function, TG,115, Turbine Hi Vibration Trip Bypass is in " Bypass".
| |
| : 4. ior-ANN xall870_54a_c_8 ON
| |
| " Turbine Trips Manually Bypassed" annunciator.
| |
| J JPM-110-07 Rev. O Page 2 of 10
| |
| | |
| RBS JOB PERFORMANCE MEASURE DATA SHEET References for Development: STP-110-0101, Turbine Overspeed Protection System Operability Test Required N aterials: STP-110-0101, Turbine Overspeed Protection System Operability Test Required Plant Condition: 70% Reactor Power Applicable Objectives: STM-110, Obj. 2 Safety Related Task: (If K/A less than 3.0)
| |
| - Control Manipulations: N/A i
| |
| l Items marked with an "*" are required to be performed, and are Critical Steps, failure to successfully complete a Critical Step requires the JPM to be evaluated as
| |
| " Unsatisfactory". Comments describing the re? son for failure are qquired in the comments section of the Verification of Con:r': tion sheet, items marked with an "^" are required to be performed in the sequence described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
| |
| l l
| |
| JPM-110-07 Rev. O Page 3 of 10 l
| |
| | |
| .~
| |
| RBS JOB. PERFORamNCE MEASURE ' .
| |
| IfIn-Plant or In the Control Room:
| |
| Caution the Operator NOT to MANIPULATE the controls, but make clear what they would do if this were not a simulated situation. ,
| |
| ! Read to the Operator:
| |
| I I will explain the initial conditions, and provide initiating cues, I may provide cues during the performance of this JPM, I will ask follow-up i
| |
| .. questions as part of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied, you should inform me i j when you have completed the task. ;
| |
| Initi:1 Conditions: Reactor power at approximately 70%.
| |
| Inititting Cue: The CRS directs you to perform Turbine Overspeed Protection System Operability Test for Turbine Control Valve #1 in accordance with STP-110-0101, Section 7.5 (High Pressure Turbine control Valve (CV) Testing). The procedure is complete up through step 7.4.4. The evaluator will act as a second qualified individual to monitor non-tested TCVs.
| |
| t l PIRFORMANCE STEP - STANDARD S/U COMMENTS j l 1. Verify reactor power is less than or equal to Reactor power verified to be less than or equal to Performance of this procedure at 84% rated. 84% using indication from the APRM recorders power levels greater than 84% ;
| |
| or process computer critical parameters screen. could result in pressure excursions and possibly a reactor scram due to f high pressure. t
| |
| : 2. Verify Reactor steam dome pressure is less Steam dome pressure verified to be less than 1020 Performance of this procedure at than 1020 psig. psig as indicated 4y: steam dome pressures greater than ' ,
| |
| * 1H13*P680, C33-R609,TURB I sT STO & 1020 psig could result in pressure RX PRESS red pen OR excursions and possibly a reactor
| |
| * ERIS Computer points B21EA008 through scram due to high pressure.' ,
| |
| B21EA013.
| |
| : 3. Verify initial valve position of all Control Control Valve position indication for TCVs 1-4 Valves is less than 35 PERCENT. verified to be less than 35% at 1H13*P680.
| |
| a JPM-110-07 Rev. 0 Page 4 of 10 t
| |
| . __ _ . ~ ~
| |
| | |
| RBS JOB PERFORMANCE MEASURE ,
| |
| PERFORMANCE STEP l STANDARD l SIU l COMMENTS
| |
| : 4. Check that the Control Valve Disc Dump Verify position of the Control Valve Disc Lunp CUE: Report as the I&C '
| |
| Test Switches are OPEN. Test Switches by contacting I&C to verify nc technician, the Control Valve continuity at the terminals indicated in Step 7 i.2 Dise Dump Test switches for of STP-110-0101. TCVs 1-4 have been verifwd open per step 7.5.2 of STP -110-0101 and you will return to initial those steps of the STP.
| |
| : 5. At H13-P822 Turbine Supervisory Candidate should indicate where H13-P822 is CUE: Inform the operator that Instruments Panel, verify the turbine located in relation to H13*P680 to verify the the turbine vibration trip is in vibration TRIP is in DISABLED. turbine vibration trip is disabled. DISABLED at H13-P822 Turbine Supervisory Instruments Panel.
| |
| : 6. Observe the following valves pretest Candidate visually observes the positions of positions: TCVs 1-4 by looking at the CONTROL VLV
| |
| : a. CV-1 using CONTROL VLV-1 VLV POS meters on H13*P680 prior to performing the POS test.
| |
| : b. CV-2 using CONTROL VLV-2 VLV POS
| |
| : c. CV-3 using CONTROL VLV-3 VLV POS
| |
| : c. CV-4 using CONTROL VLV-4 VLV '
| |
| POS
| |
| * 7. Depress and hold CV-1 TEST pushbutton CV-1 pushbutton depressed and held, CV-1 and check CV-1 slow closes using closing at moderate speed. .
| |
| CONTROL VLV-1 VLV POS.
| |
| JPM-110-07 Rev. O Page 5 of 10
| |
| | |
| RBS JOB PERFORMANCE MEASURE- .
| |
| [ PERFORMANCE STEP l STANDARD l S/U l COMMENTS l
| |
| : 8. Observe CV-2, CV-3, CV-4 opening to Candidate observes CV-2, CV-3, CV-4 opening -
| |
| compensate for CV-1 closure using as CV-1 closes.
| |
| CONTROL VLV-2,-3,-4, VLV POS.
| |
| : 9. Check for proper operation of the Fast Candidate observes CV-1 fast close when valve Acting Valve as CV-1 closes. position indicates approximately 10%.
| |
| * 10. When CV-1 has indicated closed for 10 to Candidate releases CV-1 TEST pushbutton when -
| |
| 15 seconds, then release CV-1 TEST valve has indicated fully closed for at least 10 pushbutton. seconds, and CV-1 is observed opening.
| |
| 1
| |
| : 11. Verify the following valves return to Candidate verifies CV-1 through CV-4 return to pretest positions: approximately the same position noted prior to
| |
| : a. CV-1 using CONTROL VLV-1 VLV performance of the test.
| |
| FvS
| |
| : b. CV-2 using CONTROL VLV-2 VLV
| |
| , POS
| |
| : c. CV-3 using CGNTROL VLV-3 VLV POS
| |
| , d. CV-4 using CONTROL VLV-4 VLV POS
| |
| : 12. Reset the half Scram. Half scram re5ct.
| |
| : 13. Verify Independent Vedfication Steps for Cue: I (as the evaluator) will Section 7.5.6 have been cot.,pleted. complete the Independent Verification.
| |
| Tcrmination Criteria: Turbine Overspeed Protection System Operability Test for Turbine Control Valve #1 is complete.
| |
| JPM-Il0-07 Rev. 0 Page 6 of10
| |
| | |
| RBS JOB PERFORMANCE MEASURE JPM QUESTIONS Question 1 K/A: 245000 K4.09 Rating: 3.1/3.2
| |
| | |
| ==Reference:==
| |
| OPEN The plant is returning to power following a forced outage. Turbine first stage shell temperature is 170*F and pressure is 60 psig. HP Turbine Shell warming hasjust been initiated Ww long should the turbine be heat soaked?
| |
| Answer:
| |
| 3 hours (+/-10 minutes)
| |
| Response / Comments:
| |
| SAT UNSAT Refereneca: SOP-0080, Enclosure 2, Main Turbine i
| |
| l Question 2 K/A: 24'.000 Kl.02 Rating: 3.9/4.1 ,)
| |
| | |
| ==Reference:==
| |
| CLOSED While operating at 90% p)wer the "A" pressure regulator fails causing the switch to the "B" pressure regulator. The s(tpoint of the "B" pressure regulator causes pressure to increase by 20 psig. What effect will this 20 psig pressure increase initially have on reactor power and why?
| |
| Answer:
| |
| Reactor power will increase due to the reduced void fraction caused by the increase in pressure.
| |
| Response / Comments:
| |
| SAT UNSAT
| |
| | |
| ==References:==
| |
| GEk-64892, Steam Turbine-Generator, EHC and TSI Section JPM-110-07 Rev. 0 Page 7 of 10
| |
| | |
| PREWARMING HRST STAGE SHELL BY PRESSUluZAllON ENCIJOSURE 2 ,
| |
| PAGE1 OF1 l
| |
| )
| |
| Prewanning on Tuming Gear is undenaken by pressurizing the turbine up to the intercept l eare to the saruration temperanne of the stearn.
| |
| valves with steam, The pressure nonnally specified is 60- thus raising the terIO0 psin which cwmponds to a twe4we oj least 250*F. neefore. nreatention should r ned whenew the fmt-stny shell is below
| |
| ~
| |
| 2$0*E. .
| |
| 'Ihe time required to heat soak the turbine after the casing, pressure has reached a minimum
| |
| ! of 60 psig is given. Note, that this chart gives the pressunzing :xewarming time when the ,
| |
| first stage shell temperature is less than 250*F. Every attempt s wuld be made to increase so that the Ttatine will not roll off tuming the pressure and tempensure in the shell slowiy,f rubbing occurs.
| |
| gear, as this could cause damage to the turbine t 4.0 l A m
| |
| 3e N 5E os .g 3.0 W ,,
| |
| h (
| |
| wo 39 ra i W (
| |
| gN -2.0 m
| |
| 33
| |
| =2 gw ,
| |
| w f-E I.0 .
| |
| M
| |
| ' I I
| |
| O I
| |
| . 70 90 11 0 130 150 170 190 210 230 250l OR I Lg33 FIRST STAGE SHELL TEMPERATURE (F) i SEFORE MTIAL PRESSURIZING l sop.nnan REV 11 PArF M OF51
| |
| | |
| L
| |
| .RBS JOB PERFORMANCE MEASURE VERIFICATION OF COMPLETION Operator: SSN:
| |
| Evaluator: KCN:
| |
| Date: License (Circle one): RO / SRO No. of Attempts:
| |
| Follow-up Questions:
| |
| Follow-up Question Response:
| |
| l Time to complete JPM: minutes Comments / Feedback:
| |
| J RESULT: Satisfactory / Unsatisfactory Note: An " Unsatisfactory" requires comments and remedial training.
| |
| Evaluator's Signature: Date:
| |
| JPM-110-07 Rev. O Page 8 of 10
| |
| | |
| 4 RIlS JOll PERFORMANCE MEASURE j JPM QUESTIONS (OPERATOR COPY)
| |
| Question 1 (Tills IS AN OPEN REFERENCE QUESTION)
| |
| The phun is returning to power following a forced outage. Turbine first stage shell temperature is 170'F and pressure is 60 psig. IIP Turbine Shell warming has just been initiated, llow long should the turbine be heat soaked?
| |
| Question 2 (tills IS A CLOSED REFERENCE QUESTION)
| |
| While operating at 90% power the " A" pressure regulator fails causing the switch to the "B" pressure regulator. The setpoint of the "B" pressure regulator causes pressure to increase by 20 psig. What effect will this 20 psig pressure increase initially have on reactor power and why?
| |
| J JPM-110-07 Rev. O Page 9 of 10
| |
| | |
| RBS JOB PERFORMANCE MEASURE JPM Task Conditions /Cucs (Operator Copy)
| |
| Initial Conditions: Reactor power at approximately 70%.
| |
| Initiating Cues: The CRS directs you to perform Turbine Overspeed Protection System Operability Test for Turbine Control Valve #1 in accordance with STP-110-0101, Section 7.5 (High Pressure Turbine control Valve (CV)
| |
| Testing). The procedure is complete up through step 7.4.4. The evaluator will act as a second qualified individual to monitor non-tested TCVs.
| |
| .i JPM-110-07 Rev.0 Page 10 of 10
| |
| | |
| o i
| |
| e RBS JOB PERFORMANCE MEASURE JPM NUMBER: JPM 200-04, Revision 3 TASK DESCRIPTION: Manually Startup RilR " A" in the LPCI Mode hom the Remote Shutdown Panel K/A REFERENCE & RATING: 20300 Kl.17 4.0/4.0 K4.03 3.2/3.3 20300 K4.14 3.6/3.7 A1.01 4.2/4.3 20300 Al.03 3.8/3.7 TASK
| |
| | |
| ==REFERENCE:==
| |
| 2003310401 TESTING METIIOD: Simulate Performance: X Actual Performance:
| |
| Control Room: Simulator: In-Plant: X COMPLETION TIME: 10 minutes M AX. TIME: N/A JOB LEVEL: RO/SRO TIME CRITICAL: No EIP CLASSIFICATION REQUIRED: No PRA RISK DOMINATE: No ALTERNATE PATII(FAULTED): No SAFETY FUNCTION GROUP: 11 Prepared by: D. E. Dietzel / 1248 Date: 6/5/97 KCN Ops Validation: D. Felps / 0394 Date: 6/12/97 KCN Approved by: D.E. Dietzel / 1248 Date: 6/13/97 KCN JPM-200-04 Rev. 3 Page 1 of 9
| |
| | |
| g RBS JOB PERFORMANCE MEASURE 4 SIMULATOR SETUP SHEET
| |
| . Task
| |
| | |
| == Description:==
| |
| . Manually Startup RHR "A" in the LPCI Mode from the Remote -
| |
| Shutdown Panel Required Power: N/A IC No.: N/A Notes: NONE J
| |
| l l
| |
| I l'
| |
| l JPM-200-04 Rev.3 Page 2 of 9
| |
| | |
| i 1
| |
| RBS JOB PERFORMANCE MEASURE i
| |
| DATA SIIEET References for Development: AOP-0031, Shutdown From Outside the Main Control Room Required Materials: AOP-0031, Shutdown From Outside the Main Control R.o om Required Plant Condition: N/A Applicable Objectives: liLO-021, Obj. 2 liLO-066, Obj. 2 Safety Related Task: (If K/A less than 3.0)
| |
| Control Manipulations: N/A Items marked with an "*" are required to be performed, and are Critical Steps, failure to successfully complete a Critical Step requires the JPM to be evaluated as
| |
| " Unsatisfactory", Comments describing the reason for failure are idqui. d in the comments section of the Verification of Completion sheet.
| |
| Items marked with an "^" are required to be performed in the sequence described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
| |
| JPM-200-04 Rev.3 Page 3 of 9
| |
| | |
| RBS JOB PERFORMANCE MEASURE .
| |
| 4 IfIn-Plant or In the Control Room: '
| |
| Caution the Operator NOT to MANIPULATE the controls, but make clear what they would do if this were not a simulated ;
| |
| situation.
| |
| i Read to the Operator:
| |
| l I will explain the initial conditions, and provide initiating cues, I may provide cues during the performance of this JPM, I will ask follow-up questions as part of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied, you should inform me when you have completed the task.
| |
| I Initial Conditions: The control room is evacuated, the reactor is in hot shutdown and control has been established at the Remote SI utdown
| |
| . Panel. ' Reactor pressure is approximately 300 psig.
| |
| 9 L Initi ting Cue: The CRS at h: Remote Shutdown Panel has directed you to initiate injection with RHR "A" in LPCI mode.
| |
| l PERFORMANCE STEP - l STANDARD l S/U COMMENTS
| |
| : 1. Manually open an SRV(s) as necessary to SRVs are cycled and RPV pressure lowered to Cue: Shortly after SRVs are decrease RPV pressure to approximately ~225 psig opened, pressure is 225 psig 225 psig and maintain for duration of (indicated on C61*PIR011)
| |
| , LPCI mode operation.
| |
| * 2. Open E12*F003A RHR HX A OUTLET E12*F003A RHR HX A OUTLET VALVE Cue: Ped light on, green light off.
| |
| VALVE Opened.
| |
| u ,
| |
| : 3. Open E12*F048A RHR HX A BYPASS E12*F048A RHR HX A BYPASS VALVE Cue: Red light on, green light off.
| |
| VALVE. Opened.
| |
| JPM-200-04 Rev.3 Page 4 of 9 I
| |
| | |
| . .. l RBS JOB PERFORhnANCE MEASURE 1
| |
| l S/U l COMMENTS l l PERFORMANCE STEP l STANDARD Cue: Red light on, green light off.
| |
| * 4. Start E12*C002 A RHR PUMP A. E12*C002 A RHR PUMP A Started.
| |
| Cue: Red light on, green light off.
| |
| : 5. Open E12*F064A RHR PUMP A MIN E12*F064A RHR PUMP MIN FLOW TO SUP PL. Opened.
| |
| FLOW TO SUP PL.
| |
| Cue: Red light on, green light off.
| |
| * 6. Open E12*F042A A'IR PUMP A LPCI E12*FO42A RHR PUMP A LPCIINJECT ISOL VALVE Opened.
| |
| INJECT ISOL VALVE.
| |
| RHR A Heat Exchanger Bypass Valve Cue: Red light on, green light on.
| |
| * 7. Control injection flow and RPV level by Valves are being throttled to first throttling closed E12*F048A RHR (E12*MOVF048A) and RHR HX A OUTLET VALVE (E12*F003A ) throttled to maintain maintain RPV water level.
| |
| HX A BYPASS VALVE and then E12*F003A RHR HX A OUTLET RPV level.
| |
| VALVE.
| |
| Termination Criteria: RHR A injecting in LPCI mode.
| |
| i l
| |
| l L
| |
| l l
| |
| l l
| |
| i Page 5 of 9 JPM-200-04 Rev.3
| |
| ^~
| |
| | |
| 4 h
| |
| RBS JOB PERFORMANCE MEASURE JPM QUESTIONS Question 1 K/A: 203000 K4.14 Rating: 3.6/3.7
| |
| | |
| ==Reference:==
| |
| OPEN In addition to an Annunciator, what other control room indications will alert the Main Control Rocm operators that control for RHR "A" has been transferred to the Remote Shutdown Panel.
| |
| Answer:
| |
| indicating lights for the associated components (pumps / valves) on P601 will be de-energized (OFF).
| |
| Response / Comments:
| |
| SAT UNSAT
| |
| | |
| ==References:==
| |
| LOTM 22, AOP-0031, and SOP-0027.
| |
| J Question 2 K/A: 20300 K4.03 Rating: 3.2/3.3
| |
| | |
| ==Reference:==
| |
| OPEN Why is the Minimum Flow Valve for the RHR A pump left open during LPCI mode of operation from the Remote Shutdown Panel?
| |
| Answer:
| |
| To provide minimum flow protection for the pump due to the automatic functions being disabl-d when operated from the Remote Shutdown Panel.
| |
| Response / Comments:
| |
| SAT UNSAT
| |
| | |
| ==References:==
| |
| LOTM-22, AOP-0031, and SOP-0027.
| |
| JPM-200-04 Rev.3 Page 6 of 9
| |
| | |
| RBS JOB PERFORMANCE MEASURE VERIFICATION OF COMPLETION Operator: SSN:
| |
| Evaluator: KCN:
| |
| Date: License (Circle one): RO/SRO No. of Attempts:
| |
| Follow-up Questions:
| |
| Follow-up Question Response:
| |
| Time to complete JPM: minutes Comments / Feedback:
| |
| J RESULT: Satisfactory / Unsatisfactory Note: An " Unsatisfactory" requires comments and remedial training.
| |
| Evaluator's Signature: Date:
| |
| JPM-200-04 Rev. 3 Page 7 of 9
| |
| | |
| ) 1111S JOll PERFORMANCE MEASUllE JPM QUESTIONS (OPERATOR COPY)
| |
| Question 1 (Tills IS AN OPEN ItEFERENCE QUESTION)
| |
| In addition to an Annunciator, what other control room indications will alert the Main Control '
| |
| Room operators that control for RllR " A" has been transferred to the Remote Shutdown Panel, Questlos 2 (Tills IS AN OPEN REFERENCE QtJESTION)
| |
| Why is the Minimum Flow Valve for the RilR A pump left open during LPCI mode of operation from th: Remote Shutdown Pancl? g JPM 2f,0-04 Rev.3 Page 8 of 9
| |
| | |
| RIIS Jolt PERFORMANCE MEASURE JPM Task Conditions / Cues (Operator Copy)
| |
| Initial Conditions: The control room is evacuated, the reactor is in hot shutdown and control has been established at the Remote Shutdcwn Panel. Reactor pressure is approximately 300 psig.
| |
| Initiating Cues: The CRS at the Remote Shutdown Panel has directed you to initiate injection with RIIR "A" in LPCI mode.
| |
| 9 J
| |
| JPM-200-04 Rev. 3 Page 9 of 9 I
| |
| | |
| I'
| |
| , RIIS JOll PERFORMANCE MEASURE l
| |
| JPM NUMIlER: JPM.203 02 Revision 0 TASK DESCRIPTION: Shutdown the liigh Pressure Core Spray Pump after an inadvertent Automatic initiation with a failure of the llPCS ,
| |
| minimum flow valve to automatically open.
| |
| K/A REFERENCE & RATING: 209002 Al.01 3.6/3.7 Al.02 3.4/3.6 i 209002 Al.08 3.1/3.3 A2.08 3.1/3.2 209002 A3.01 3.3/3.3 A3.03 3.6/3.6 209002 A3.04 3.7/3.7 A4.01 3.7/3.7 209002 A4.03 3.8/3.8 A4.04 3.1/3.1 209002 A4.15 3.6/3.6 TASK
| |
| | |
| ==REFERENCE:==
| |
| 206016001001 TESTING METIIOD: Simulate Performance: Actual Performance: X Control Room: Simulator: X In Plant:
| |
| COMPLETION TIME: 10 minutes MAX. TIME N/A
| |
| . JOB LEVEL: RO/SRO J
| |
| TIME CRITICAL: No EIP CLASSIFICATION REQUIRED: No PRA RISK DOMINATE: No ALTERNATE PATil(FAULTED): Yes SAFE TY FUNCTION GROUP: IV Prepared by: Michael K. Cantrell / 1116 Date: 1/7/97 KCN Ops Validation. F.R. Godwin / 1409 Date: 1/21/97 KCN Approved by: D.E. Dietrel / 1248 Date: 1/21/97 KCN l
| |
| l JPM.203 02 Rev. 0 Page 1 of 9
| |
| | |
| .. =. - . - . . _ . - . . - - - - . _ _ - - _ _ . _ _ _ _ - . _ _ _-- _ _ _ - _ - - _ -
| |
| 4 s
| |
| RBS JOB PERFORMANCE MEASURE SIMULATOR SETUP SilEET Task
| |
| | |
| == Description:==
| |
| Shutdown the liigh Pressure Core Spray Pump after an inadvertent ;
| |
| Automatic initiatio.: with a failure of the llPCS minimum flow valve to automatically open.
| |
| j Required Power: Reactor Power < 100%
| |
| i I C N o.: IC 12 (less than 100% reactor power)
| |
| Notes: '
| |
| : 1. Enter MF $4, Failure ofilPCS D/O to start.
| |
| : 2. ManuallyinitiatellPCS.
| |
| : 3. Disarm the manualinitiation collar.
| |
| : 4. Verify RPV water level is below level 8.
| |
| j 5. Insert Override 601,16C, E12,4 4ge 6 of 7
| |
| : a. E22 F012 + 0 ON (green light on)
| |
| : b. E22 F012 + R OFF (red light off)
| |
| : 1. When the operator manually opens E22-{012, remove the overrides input in the previous step in the following order:
| |
| : a. remove E22 F012 + R
| |
| : b. remove E22-F012 + 0 k
| |
| -JPM 203-02 Rev 0 - Page 2 of 9
| |
| | |
| ._- - _ _ . . _ - _ _ - _ - - - - _ - - _ _ - _ = _ - _ _ _ . . - -
| |
| 4
| |
| (
| |
| RBS JOB PERFORMANCE MEASURE l
| |
| DATA SilEET l l
| |
| References for Development: SOP-0030, High Pressure Core Spray Required Materials: SOP-0030,liigh Pressure Core Spray Required Plant Condition: Reactor Power < 100%
| |
| Applicable Objectives: liLO-019-06, Objectives 3,4,7, and 11 Safety Related Task: (If K/A less than 3.0)
| |
| Control Manipulations: N/A Items marked with an "*" are required to be performed, and are Critical Steps, failure to successfully complete a Critical Step requires the JPM to be evaluated as
| |
| " Unsatisfactory". Comments describing the reason for failure are required in the comments section of the Verification of Completion sheet, items marked with an "^" are required to be performed in the sequlace described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
| |
| l l
| |
| l
| |
| .IPM.203-02 Rev. 0 Page 3 of 9 l _ _ _ __ _. _
| |
| | |
| RBS JOJ PERFOkalANCE MEASURE .
| |
| IfIn-Plant or In the Control Room:
| |
| Caution the Operator NOT to MAN 1PUIATE the controls, but make clear what they would do if this were not a simulated situation.
| |
| Read to the Operator:
| |
| I will explain the initial conditions, and provide initiating cues, I may provide cues during the performance of this JPM, I may ask follow-up questions as part of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied, you should inform me when you have completed the task.
| |
| Initial Conditions: The plant is operating at 75% power after an inadvertent HPCS in tiation. Reactor water level is in the normal range.
| |
| The HPCS diesel generator has been shutdowm.
| |
| Initiating Cue: The C"S has directed you to shutdown the High Pressure Core Spray Pump using SC ' 0030, High Pressure Core Spray, Section 6.
| |
| l PERFORMANCE STEP l STANDARD l S/U l COMMENTS l
| |
| : 1. Verify F.22A-S2, HPCS MANUAL Manual initiation pushbutton collar rotated to the INITIA1:ON collarisin the DISARM DISARM position.
| |
| position.
| |
| * 2. Depress E22A-S7, HPCS INITIATION HPCS INITIATION RESET pushbutton RESET pushbutton and check the white depressed and white light is off.
| |
| light goes off.
| |
| * 3. Verify closed the following valves: The following valves are closed with the green
| |
| : a. E22-F023, HPCS TEST RETURN light on and red light i off:
| |
| VLV TO SUPPRESSION POOL a. E22-F023, HPCS TEST RETURN VLV TO
| |
| : b. E22-F010, HPCS TEST BYPASS VLV SUPPRESSION POOL TO CST b. E22-F010, HPCS TEST BYPASS VLV TO
| |
| : c. E22-F011, HPCS TEST RETURN CST VALVE TO CST c. E22-F011. HPCS TEST RETURN VALVE
| |
| : d. E22-F004, HPCS INJECT ISOL TO CST VALVE d. E22-F004, HPCS INJECT ISOL VALVE JPM-203-02 Rev. O Page 4 of 9
| |
| | |
| n.
| |
| RBS JO3 PERFOktaANCE MEASURE ,
| |
| I PERFORMANCE STEP STANDARD S/U l COMMENTS i
| |
| !
| |
| * 4. When flov fowers below 63 gpm on E22- Candidate recognizes the failure of E22-F012 to CUE: As CRS ackaewledge the R603, HPCS FLOW, verify E22-F012, open automatically. failmet of E22-F012 te open and HPCS MIN FLOW VALVE TO Manually opens E22-F012, verifying red light on if regeested, direct the gan l SUPPRESSION POOL opens. and green light off or trips the HPCS pump. to open E22-F012. l NOTE: As a conservative action, Candidate should also inform the CRS of the failure of E22-F012 to automatically open when an inumeediate pesop trip immy be ;
| |
| pump flow falls below 625 gpm. initiated. Resmaining steps should be perfornoed following j i the trip. i
| |
| '
| |
| * 5. If E22-PC003, HPCS LINE FILL PUMP is E22-PC003 running witn red light on, green light not running, then start E22-PC003. off. i
| |
| * 6. Trip E22-ACB02, HPCS PUMP SUPPLY E22-ACB02 open, green light on, red light off. l BRKR. I
| |
| : 7. When HPCS Pump discharge pressure Valve verified to auto close when rw mc drops !
| |
| . lowers below 300 psig on E22-R601, below 300 psig by observing green light on, red i HPCS PUMP DISCH PRESSURE,then light off. l verify E22-F012, HPCS MIN FLOW
| |
| . VALVE TO SUPPRESSION POOL i closes. I i t i j- 8. If the HPCS DG is operating, then shut 'Ihe HPCS DG is shutdown. CUE: If requested,infern ;
| |
| ] down the DG per SOP-0052, HPCS Diesel candidate that abe HPCS DG l l Generator. was shutdown by othes- !
| |
| i l eperators.
| |
| i . !
| |
| j Terminating Criteria: High Pressure Core Spray Pump is shutdown. l
| |
| ?
| |
| I JPM-203-02 Rev. O Page 5 of 9 l l
| |
| r
| |
| _ _ ___ _ - _ _ _j
| |
| | |
| 4 1
| |
| . RBS JOB PERFORMANCE MEASURE 1
| |
| JPM QUESTIONS Question 1 K/A: 2.1.24 Rating: 2.8/3.1
| |
| | |
| ==Reference:==
| |
| OPEN Note: Print Reading Exercise A valid llPCS initiation signal is received and all llPCS components respond properly but the !
| |
| white "IIPCS INITI ATION RESET" light fails to illuminate. He bulb is checked and is good. l What is the most likely cause of this failure?
| |
| Answer:
| |
| Failure of contacts Ml Tl associated with relay E22 K9.
| |
| Response / Comments:
| |
| SAT UNSAT Refercuees: GE Elementary Diagram 828E536AA, sheets 4 and 5.
| |
| J Question 2 K/A: 209002 A2.01 Rating: 3.8/3.8
| |
| | |
| ==Reference:==
| |
| OPEN llow would the liigh Pressure Core Spray system respond to an automatic initiation signal if both the suppression pool suction valve and CST suction valves were shut?
| |
| Answer:
| |
| ne CST suction valve would open.
| |
| Response / Comments:
| |
| SAT UNSAT
| |
| | |
| ==References:==
| |
| GE Elementary Diagram 828E536AA, sheet 5.
| |
| - JPM 203-02 Rev.O Page 6 of 9
| |
| | |
| s 1
| |
| . RBS JOB PERFORMANCE MEASURE VERIFICATION OF COMPLETION j Operator: ,
| |
| SSN:
| |
| Evaluator: KCN:
| |
| l Date: License (Circle one): RO / SRO No. of Attempts:
| |
| l l
| |
| Follow up Questions:
| |
| i Follow up Question Response:
| |
| Time to complete JPM: minutes Comments / Feedback:
| |
| J RESULT: Satisfactory / Unsatisfactory l
| |
| Note: An " Unsatisfactory" requires comments and remedial training.
| |
| Evaluator's Signature: Date:
| |
| JPM-203-02 Rev. O Page 7 of 9
| |
| | |
| 4 RilS JOB PERFORMANCE MEASURE l JPM QUESTIONS (OPERATOR COPY)
| |
| Question 1 TIIIS IS AN OPEN REFERENCE QUESTION A valid IIPCS initiation signal is received and all llPCS components respond properly but the white "IIPCS INITIATION RESET" light fails to illuminate. The bulb is checked and is good.
| |
| Mat is the most likely cause of this failure?
| |
| E Question 2 Tills IS AN OPEN REFERENCE QUESTION llow would the liigh Pressure Core Spray system respond to an automatic initiation signal if both the suppression pool suction valve and CST suction valves wye shut?
| |
| JPM 203-02 Rev.O Page 8 of 9
| |
| | |
| . ._. -. - ~ . - - = . __ - - - . ~ _. -
| |
| RBS JOB PERFORMANCE MEASURE JPM Task Conditions / Cues (Operator Copy)
| |
| Initial Condit!ons: The plant is operating at 75% power after an inadvertent ilPCS initiation. Reactor water level is in the normal range. The llPCS diesel generator has been shutdown.
| |
| Initiating Cues: 'Ihe CRS has directed you to shutdown the liigh Pressure Core Spray i Pump using SOP-0030, liigh Pressure Core Spray, Section 6.
| |
| l l
| |
| l J
| |
| JPM 203-02 Rev.O Page 9 of 9
| |
| ,, _._.- , .. __ -_ _. .~.
| |
| | |
| li* !
| |
| RBS JOB PERFORMANCE MEASURE i 1
| |
| JPM NUMHER: JPM 309-03, Revision 0 l TASK DESCRIPTION: Perform a Non-Emergency Start, Load, and Parallel of the Division 1(11) Emergency Diesel Generator (Locally)
| |
| K/A REFERENCE & RATING: 264000 A1.09 3.0/3.1 A2.01 3.5/3.6 264000 A2.05 3.6/3.6 K4.07 3.3/3.4 TASK
| |
| | |
| ==REFERENCE:==
| |
| 264007001004 TESTING METilOD: Simulate Performance: X Actual Performance:
| |
| Control Room: Simulator: In Plant: X COMPLETib ! TIME: 30 minutes MAX. TIME: N/A JOB LEVEL: RO/SRO TIME CRITICAL: No l
| |
| EIP CLASSIFICATION REQUIRED: No J
| |
| l PRA RISK DOMINATE: No ALTERNATE PATil(FAULTED): No SAFETY FUNCTION GROUP: VI
| |
| ! Prepared by: M.K. Cantrell / I116 Date: 1/7/97 KCN Ops Validation: D. Felps / 0394 Date: 6/12/97
| |
| ! KCN Approved by: D.E. Dietzel / 1248 Date: 6/13/97 KCN JPM-309-03 Rev. 0 Page 1 of 15
| |
| | |
| RBS JOB PERFORMANCE MEASURE SIMULATOR SETUP SilEET Task
| |
| | |
| == Description:==
| |
| Perform a Non-Emergency sta", load and parallel of the Division 1(11) Emergency Diesel Generator (Locally).
| |
| Required Power: N/A IC No.: N/A Notes: None
| |
| .j JPM-309-03 Rev. O Page 2 of 15
| |
| | |
| RHS JOB PERFORMANCE MEASURE ,
| |
| DATA SiiEET i References for Development: SOP-0053, Standby Diesel Generator and Auxiliaries
| |
| )
| |
| Required Materials: SOP-0053, Standby Diesel Generator and Auxiliaries, Section 4.3, Non Emergency Starting, I oading and Paralleling of the Standby Diesel Generator l SOP-0053, Attaciunent 6, KW vs. KVAR l Required Plant Condition: Any Applicable Objectives: NEO-015 Objectives 6,7,8,9, and 10 Safety Related Task: (if K/A less than 3.0)
| |
| Control Manipulations: N/A Items marked with an "'" are required to be performed, and are Criitical Steps, failure to successfully complete a Critical Step requires the JPM to be evafuated as
| |
| " Unsatisfactory". Comments describing the reason for failure are required in the comments section of the Verification of Completion sheet, items marked with an "^" are required to be performed in une sequence described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
| |
| JPM 309 03 Rev.O Page 3 of 15
| |
| | |
| l ~
| |
| 1 RBS JOB PERFOkmaNCE MEASURE .l t
| |
| ! . IfIn-Plant or In the Control Room: Caution the Operator NOT to MANIPULATE the controls, but make clear what they would de if this w:n Mot a simulated situation. Performance of this JPM on Division I is preferred, but if Division I is not accessible, then direct the ,
| |
| ! er .. tor to perform this JPM on Division II.
| |
| t l
| |
| [ Read to the Operator:
| |
| 1 will explain the initial conditions, and provide initiating cues, I may provide cues during the performance of this JPM, I may ask follow-up t
| |
| questions as part of this JPM. When you ::omplete the task successfully, the objective for this JPM will be satisfied, you should inform me j when you have, completed the task.
| |
| r l Initi:1 Conditions: The Division I (II) Standby Diesel Generator has been placed in a standby lineup per SOP-0053 from a previous D/G run ~ 3
| |
| [ hours ago.
| |
| i 2
| |
| Initiating Cue: The CRS has directed you to start, parallel and load to 3100 KW the Division I (II) Emergency Diesel Generator from the l Engine Control Panel in accordance with SOP-0053, Section 4.3.
| |
| ! 1 PERFORMANCE STEP l STANDARD S/U COMMENTS f- 1. . Start the Diesel Generator Building Diesel Generator Building Ventilation verified to CUE: Inform the operator that the Ventilation per SOP-0061. be in operation in accordance with SOP-0061. Diesel Generator Building Ventilation is in operation in accordance with SOP-0061.
| |
| 4 1
| |
| : 2. Verify the diesel is in STANIYBY per Diesel verified to be in a standby lineup per CUE: Inform the operator that the Section 4.1 of this procedure. section 4.1. Dieselis in STANDBY per section i' . 4.1.
| |
| i
| |
| : 3. Bar or air roll the engine per Section 4.2 of Determination from Precaution 2.6 that a bar or CUE: When asked, inform the air roll of the diesel is not required. operator that the diesel has been
| |
| ~
| |
| this procedure, if required per Precaution
| |
| , 2.6. shutdown forapproximately 3 i hours from the previous run.
| |
| l.
| |
| : l. 4. Open the Turbocharger Prelube Valve and Turbocharger Prelube Valve open and 2 minutes i ' wait 2 minutes before proceeding, have clapsed.
| |
| ! i 1 JPM-309-03 Rev. 0 - Page 4 of15 l
| |
| i i !
| |
| l'
| |
| | |
| RBS JOB PERFORMANCE MEASURE .
| |
| l PERFORMANCE STEP l STANDARD l S/U l CO31MENTS l
| |
| * 5. St.m the engine by depressing the Diesel running after NORMAL START CUE: Provide the following NORMAL START pushbutton and venfy pushbutton depressed, with the following parameters when asked:
| |
| the following parameters are in the parameters verified in the specified bands:
| |
| indicated range: a. 1 EGO-PISA (B) LUBE OIL PRESSURE 50- a. 1 EGO-PI5A(B) LUBE OIL
| |
| : a. 1 EGO-PISA (B) LUBE OIL 65 psig PRESSURE 56 psig PRESSURE 50-65 psig b. IEGO-PI10A(B) TURBO OIL PRESSURE b. 1 EGO-PI10A(B) TURBO OIL
| |
| : b. 1 EGO-PII0A(B) TURBO OIL 25-35 psig PRESSURE 34 psig PRESSURE 25-35 psig c. 1EGF-PI27A(B) FUEL OIL PRESSURE 30- c. 1EGF-PI27A(B) FUEL OIL
| |
| : c. 1EGF-PI27A(B) FUEL OIL 40 psig PRESSURE 36 psig PRESSURE 30-40 psig d. IEGF-PDI29A(B) FUEL OIL FILTER DIFF d. 1EGF-PDI29A(B) FUEL OIL
| |
| : d. 1EGF-PDI29A(B) FUEL OIL FILTER PRESS less than 20 psid FILTER DIFF PRESS 4 psid DIFF PRESS less than 20 psid c. 1 EGO-PDI7A(B) LUBE OIL FILTER DIFF e. 1 EGO-PDI7A(B) LUBE OIL
| |
| : e. 1 EGO-PDl7A(B) LUBE OIL FILTER PRESS less than 20 psid FILTER DIFF PRESS 12 psid DIFF PRESS less than 20 psid f. 1EGT-PI4A(B) JACKET WATER f. IEGT-PI4A(B) JACKET ,
| |
| : f. 1EGT-PI4A(B) JACKET WATER PRESSURE 12-30 psig WATER PRESSURE 13.5 PRESSURE 12-30 psig g. 1EGS-PI6A(B) CRANKCASE PRESSURE O psig
| |
| : g. 1EGS-PI6 A(B) CRANKCASE .5 in Hg. g. IEGS-P16A(B) CRANKCASE PRESSURE 0 .5 in Hg. PRESSURE .25 in lig.
| |
| : 6. Close the Turbocharger Prelube Valve. Turbocharger Prelube Valve closed. CUE- Turbocharger Prelube Valve closed.
| |
| : 7. Verify air start valves are fully closed by Air lines upstream of the air start valves verified CUE: When candidate goes to ensuring the air start lines upstream of the not hot to the touch. verify the piping upstream of the valves are not hot to touch. air start valves, tell then they are not hot to the touch.
| |
| : 8. Perform the requirements of PEP-0026, Applicable data has been recorded on PEP-0026 CUE: Inform the operator that all diesel Generator Trending and Failure and is within the specified ranges. applicable data has been recorded Reporting. on PEP-0026 and is within the specified ranges.
| |
| JPM-309-03 Rev. O Page5 of15
| |
| | |
| RBS JOJ PERFORMANCE MLASURE .
| |
| l PERFORMANCE STEP l STANDARD l S/U l COMMENTS l
| |
| : 9. Select the phase of bus and generator BUS VOLTS and GENERATOR VOLTS _ _ _
| |
| CUE BUS VOLTS and voltage to be monitored on the BUS voltmeters selected to either 1-2,2-3 or 3-1. GENERATOR VOLTS voltmeters selected to 1-2.
| |
| VOLTS and GENERATOR VOLTS vollmCter.
| |
| : 10. Verify the blue REMOTE . REMOTE SYNCHRONIZING SELECTOR CUE- REMOTE SYNCIIRONIZING SELECTOR SWITCH OFF blue light ON. SYNCIIRONIZING SELECTOR SWITCH OFF light is ON. SWITCH OFF blue light ON.
| |
| * 11. Place the SYNCHRONIZING CONTROL S(NCHRONIZING CONTROL in the GEN CUE: SYNCHRONIZING position. CONTROL in the GEN pcsition.
| |
| to GEN.
| |
| * 12. Adjust ShANDBY DIESEL EGI A(B) STANDBY DIESEL EGI A(B) INCOMING NOTE: Running Volts-120 Volts.
| |
| INCOMING VOLTS to about 1-2 volts VOLTS 1-2 volts above RUNNING VOLTS.
| |
| above RUNNING VOLTS using the VOLTAGE REGULATOR CONTROL
| |
| * 13. Adjust the STANDBY DIESEL EGI A(B) Synchroscope rotating slowing in the FAST CAUTION: To avoid diesel speed, using the GOVERNOR CONTROL direction. engine crankshan critical speed, so that the synchroscope is rotating slowly continuous operation between 453 in the FAST direction. and 457 RPM (60.4 to 60.9 HZ) shall not be permitted.
| |
| * 14. When the synchroscope indicator is IEGS-1EGI A(I}) GENERATOR to 1 ENS- NOTE: Syne check moving slowly in the FAST direction AND SWG1 A(B) ST8Y BUS breaker taken to close instrumentation requires the output the synchroscope indicator is 5 minutes to position when the synchroscope indicates breaker switch to be held in the 2 minutes befc te the 12 o' clock position, between 5 minutes and 2 minutes till 12 o' clock closed position until breaker THEN close the 1EGS-EGI A(B) and moving in the FAST direction. Breaker closure or symchroscope needle GENERATOR to 1 ENS *SWG1 A(B) control switch held in the close position until passes through 12 o' clock.
| |
| STBY BUS breaker. Verify the red breaker closure or the symchroscope needle passes breaker lights are ON. through 12 o' clock, breaker red light ON, green light OFF.
| |
| JPM-309-33 Rev. O Page 6 of IS E -__- - _.-___-- - _ _ _ _ - - - - - - _ _ - - - - - - _ _ - - - - - - _ - -
| |
| | |
| j RBS JO3 PERFORMANCE MEASURE .
| |
| l STANDARD S/U / COMMENTS l
| |
| [ PERFORMANCE STEP
| |
| : 15. Increase generator load to approximately Generator load approximately 175 KW. JUE: Generator load 175 KW using the GOVERNOR approximately 175 KW.
| |
| CONTROL
| |
| : 16. As soon as diesel generator minimum load SYNCHRONIZING CONTROL to OFF. CUE- SYNCllRONIZING has stabilized,' place the CONTROL to OFF.
| |
| SYNCHRONIZING CONTROL to OFF.
| |
| : 17. Increase diesel generator load with the Diesel generator load at 3100 KW and diesel NOTE- Loading limitation -
| |
| . GOVERNOR CONTROL and adjust generator VARS at approximately 2100 to <2300 guideline in the table above VARS using the VOLTAGE KVARS. procedure step 4.3.17 simuld be REGULATORCONTROL Use followed when increasing Attachment 6 as a guide to verify the generator load. ;
| |
| generator is not operated at less than a .8 power factor. ;
| |
| Terminating Criteria: Emergency Diesel Generator I A(B) operating and loaded to 3100 KW.
| |
| L L
| |
| ' JPM-309-03 Rev. O Page 7 of15 t
| |
| | |
| 4 RBS JOB PERFORMANCE MEASURE JPM QUESTIONS I
| |
| Question i K/A: 264000 K4.08 Rating: 3.8/3.7
| |
| | |
| ==Reference:==
| |
| CLOSED If a high drywell pressure signal is received while the Division I Diesel Generator is out of service, what actions are required to prevent an automatic start upon returning the diesel generator to service? ,
| |
| Answer:
| |
| . 'Ihe high drywell pressure signal must be cleared and reset.
| |
| * Depress the EMERGENCY START RESET pushbutton on lill3*P877.
| |
| Response / Comments:
| |
| SAT . UNSAT
| |
| | |
| ==References:==
| |
| SOP-0053 and LO1 1 58 ESK-11EGA02 ESK-7EGA03 _ ,
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| l JPM-309-03 Rev. O Page 8 of 15
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| l RHS JOB PERFORMANCE MEASURE JPM QIJESTIONS ;
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| K/A: 2.1.12 Rating: 2.9/4.0
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| ==Reference:==
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| OPEN Question 2 (RO Candidates]
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| Given the following condition:
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| * Div. I D/G is in nonnat standby lineup.
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| As the Control Bldg. Operator, while monitoring the Div. I D/G parameters, you note that the Div. I D/G Fuel Oil Day Tank level indicator is indicating ~40%.
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| What are your required actions based on this D/G Fuel Oil Tank Level?
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| Answer:
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| Notify the Main Control Room and report that the Div. I D/G Fuel Oil Day Tank is out-of- spec.
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| Iow. (40% = 210 gallons; T.S. .$ 45% or 5 316.3 gallons)
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| [ NOTE: Tech. Spec actions are not required as part of this question.]
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| Response / Comments:
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| SAT UNSAT
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| ==References:==
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| LOTM 58, Standby Diesel Generators SOP-0053, Standby Diesel Generators and Auxiliaries OSP-0028, Daily Operating Log T.S. 3.8.1 SR 3.8.1.4 JPM-309-03 Rev. O Page 9 of 15
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| * 1 I
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| RHS JOB PERFORMANCE MEASURE PM QUESTIONS 1
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| Question 2 K/A: 2.1.12 Rating: 2.9/4.0
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| ==Reference:==
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| OPEN I
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| [SRO Candidates]
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| Given the following condition: ,
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| e Div. I D/G is in normal standby lineup.
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| As the Control Room Supervisor, the Control Bldg. Operator notifics the Main Control Room that the Div. I D/G Fuel Oil Day Tank level indicator is indicating ~40%.
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| What actions are required based on this D/G Fuel Oil Tank Level?
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| Answer:
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| Declare the Div. I D/G INOPERABLE and take the required actions per T.S. 3.8.1.
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| [ NOTE: Tech. Spec, actions are not required as part of this question.]
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| J Response / Comments:
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| SAT UNSAT l
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| ==References:==
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| LOTM 58, Standby Diesel Generators SOP-0053, Standby Diesel Generators and Auxiliaries T.S. 3.8.1 and SR 3.8.1.4 OSP-0028, Daily Operating Log.
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| JPM 309-03 Rev.O Page 10 of 15
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| ~-
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| 4 RilS JOll PERFORMANCE MEASURE VERIFICATION OF COMPLETION Operator: SSN:
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| Evaluator: KCN:
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| Date: License (Circle one): RO/SRO No. of Attempts:
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| Follow up Questions:
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| Follow up Question Response:
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| Time to complete JPM: minutes Comments / Feedback:
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| J RESULT: Satisfactory / Unsatisfactory Note: An " Unsatisfactory" requires comments and remedial training.
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| EvaluatoA Signature: Date:
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| JPM-309-03 Rev. 0 Page 11 of 15
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| 4 4
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| RilS JOll PERFORMANCE MEASURE JPM QUESTIONS (OPERATOR COPY)
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| Question 1 TIIIS IS A CLOSED REFERENCE QUESTION ,
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| if a high drywell pressure signal is received while the Division I Diesel Generator is out of service, what actions are required to prevent an automatic start upon returning the diesel generator to service?
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| i JPM-309-03 Rev. O Page 12 of 15
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| ._ _ - -.. .______ .-. .- -_ - - _ _ = _ - _ _ .
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| 1 RilS Jolt PERFORMANCE MEASURE !
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| JPM QUESTIONS (OPERATOR COPY)
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| Question 2 Tills IS AN OPEN REFERENCE QUESTION
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| [RO Candidates]
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| Given the following condition:
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| * Div.1 D/O is in nonnal standby lineup, As the Control Bldg. Operator, while monitoring the Div. I D/O parameters, you note that the Div. I D/O Fuel Oil Day Tank level indicator is indicating ~40%.
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| What are your required actions based on this D/G Fuel Oil Tank Level?
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| l JPM 309-03 Rev.O Page 13 of 15
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| RilS JOll PERFORMANCE MEASURE JPM QUESTIONS (OPERATOR COPY)
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| Question 2 Tills IS AN OPEN REFERENCE QUESTION lSRO Candidates]
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| Olven the following condition:
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| . Div. I D/G is in normal standby lineup.
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| As the Control Room Supervisor, the Control Bldg. Operator notifies the Main Control Room that the Div. I D/G Fuel Oil Day Tank level indicator is indicating -40%.
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| What actions are required based on this D/G Fuel Oil Tank Level? .
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| JPM-309-03 Rev. O Page 14 of 15
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| i RHS JOB PERFORMANCE MEASURE I JPM Task Conditions / Cues (Operator Copy) ,
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| initial Conditions: The Division I (II) Standby Diesel Generator has been placed in a l standby lineup per SOP-0053 from a previous D/O run < 3 hours ago. >
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| Initiating Cues: The CRS has directed you to start, parallel and load to 3100 KW the Division 1 (II) Emergency Diesel Generator from the Engine Control Panel in accordance with SOP-0053, Section 4.3. ,
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| JPM-309-03 Rev. O Page 15 of 15
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| .- RBS JOB PERFORMANCE MEASURE JPM NUMUER: JPM-403-03 Revision 1 TASK DESCRIPTION: Stanup Drywell Purge Using Standby Gas Treatment A f i
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| (GTS*FNIA)
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| K/A REFERENCE & RATING: 261000 Kl.01 3.4/3.6 A4.04 3.3/3.4 288000 A2.04 3.7/3.8 A4.01 3.1/2.9 272000 A1.01 3.2/3.2 TASK
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| ==REFERENCE:==
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| 2223120101 TESTING METiiOD: Simulate Performance: Actual Performance: X Control Room: Simulator: X In-Plant:
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| COMPLETION TIME: 8 minu.es MAX. TIME: N/A JOB LEVEL: RO/SRO TIME CRITICAL: No EIP CLASSIFICATION REQUIRED: No I
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| PRA RISK DOMINATE: No ALTERNATE PATil(FAULTED): No SAFETY FUNCTION GROUP: IX Prepared by: D. E. Dietzel / 1248 Date: 6/6/97 KCN Ops Validation: H. Carver / 1191 Date: 6/12/97 KCN Approved by: D.E. Dietzel / 1248, Date- 6/13/97 KCN s
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| JPM-403-03 Rev.1 Page 1 of 10
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| RBS JOB PERFORMANCE MEASURE i
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| l SIMULATOR SETUP SilEET l Task
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| == Description:==
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| Startup Drywell Purge Using Standby Gas Treatment A (GTS*FNIA)
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| Required Power: Cold Shutdown or Refuel Modes i
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| IC No.: ANY (IC 1) I Notes:
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| : 1. Remove Tags from dampers on panel P863.
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| : 2. imf RMS-RE103, 2.00 E-7
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| : 3. Caution statement does not allow operation of the Drywell Purge valves in Modes 1,2 or 3.
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| J JPM-403-03 Rev.1 Page 2 of 10
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| l RBS JOB PERFORMANCE MEASURE DATA SilEET References for Development: SOP-0059, Containment HVAC System (SYS#403)
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| Required Materials: SOP-0059, Containment HVAC System (SYS#403)
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| Required Piant Condition: Cold Shutdown or Refuel Modes Applicable Objectives: HLO-038, Objectives 2 Safety Related Task: (If K/A less than 3.0)
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| Contiol Manipulations: N/A Items marked with an "*" are required to be performed, and are Critical Steps, failure to successfully complete a Critical Sten requires the JPM to be evaluated as
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| " Unsatisfactory". Comments describing the reason for failure are required in the ,
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| comments section of the Verification of Completion sheet.
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| items marked with an "^" are required to be perfoi.nM in the sequdice described, if not performed in the sequence described, appropriate cues other than described in the body of the JPM may be required to provide proper feedback.
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| JPM-403-03 Rev.1 Page 3 of 10
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| ! I RBS JOB PERFORMANCE MEASURE IfIn-Plant or In the Control Room:
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| Caution the Operster NOT to MANIPULATE the contals, but make clear what they would do if this were not a simula Read to the Operator:
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| I will explain the initial conditions, and provide initiating cues, I may provide cues during the performance of this questions as part of this JPM. When you complete the task successfully, the o'. itctive for this JPM will be satis when you have completed the task.
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| I iti:1 Conditions: Drywell purge is required prior to opening the Drywell for maintenance access. The plant is shutdown and de Initiating Cue: The CRS has directed you to start a Drywell Purge using Standby Gas Treatment A (GTS*FNIA) IAW SOP-0059. Step 5.6.
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| l S/U l COMMENTS l l STANDARD l PERFORMANCE STEP Both}}
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