ML20217M052
| ML20217M052 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/14/1997 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20210K753 | List: |
| References | |
| NUDOCS 9708200056 | |
| Download: ML20217M052 (425) | |
Text
{{#Wiki_filter:t >' M 004 UNITED ST ATES
. ,$ g NUCLEAR REGULATORY COMMfSSION f
h ..k . ), c R EC,10N IV 611 RY AN PLAZA DRIVE, SUITE 400
, AR LINGTON, TEXAS 760110064 August 14, 1997 NOTE T0: NRC Document Control Desk Mail Stop 0 5 D 24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV
SUBJECT:
OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON APRIL 14 THROUGH
-MAY 6, 1997, AT FORT CALHOUN STATION, UNIT 1 DOCKET #50 285 On April 14 through May 6,1997, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070. b) As given operating examinaticn, designated for distribution under RIDS Code A070. Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42. _ If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860 8253. para mag _
% 1 d
Facility: forLCathgsg Esam Da.e: 04/14/91 Knowledge and Ability Record Form COUNT MATRIX Summarizing Counts by K/A Group for PWR Senior Reactor Operator Total Plant Wide Gencrics 17 Kl K2 K3 K4 K5 K6 Al A2 A3 A4 SG Plant S.s stems ! 1 2 1 3 1 1 1 3 1 2 3 19 Plant Systems li I ! 3 4 I i 1 1 1 1 2 17 Plant Systems lil 1 0 1 0 0 0 0 0 1 0 1 4 Emergency /Abn i 1 3 2 - - -- 10 3 - - 5 24 Emergency /Abn 11 1 1 2 -- - -- 5 4 - - - - 3 16 Emergency /Abn ill 0 0 0 - - - 1 2 - - 0 3 9 7 = = = Totals 5 7 2 2 18 13 3 3 14 Model Total 100 I Fo o
.. - - - . . _- - ..- --~ ~. . ._- - . . .-
PWR -SRO Pl ANT SWI EMS - 40% -Group 1 puup I - 19% Kl K2 K3 K4 K5 K6 Al A2 A3 A4 SG 001 Control Rod Drn e Splem 11 04 04 001 Reactor Coolant Pump Splem 01
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004 Chemical and Volun.c Control Sytcm 05 07 011 Enginected Safet) Fcamrcs Actuation Sutem 02 014 Rod Position Indication Sutem 06 015 Nuclear Instrumentation S3stem 05 017 in-Core Temperature Monitor Sptem . 06 022 Containment Cooling Sutem 02 04 ; 025 Ice Condenser Splem No Ice Condenser Splem at Fort Calhoun 026 Containment Spras Sptem l l l l l l l l l 04 l l 056 Condenwte Ss stem No SRO importance factors 3.0 or greater 059 _ Main Feedaater Splem 02 001 Ausiliary/ Emergency Feedwater Sutem 15 061 DC Electrical Dntribution Splem 01 06M 1.iquid Raduaste Splem 04 071 Waste Gas Disposal Splem 09 072 Area Radiation Monitorine Sptem 09 Totals l Group l=19 1 2 1 3 1 l1 1 3 1 2 3 Su Mode l K/Aff item l Imp. 001 (H N) K2 01 Knowledge of powcr supplies to the RCPs. ~ 3.1 061 000 GliN-15 Abilit3 to recognite abtiormal indications for sptem operating panuncters which are entry-level 4.1 conditions for emergenc3 and abnormal operating procedurcs. i 001 (MN) K201 Knowledge of th: power supplies to the maior dc loads. 3.1 001 010 A2 04 Predict the impact and mitigate the consequences of a station blackout upon the Wration of 4.0 CRDS 071 tKM) A4.09 Ability to manually operate and monitor the uaste gas release rad monitors 3.5 tm4 020 K602 Knowledge of pressure control methods during solid plant operations. 4.1 015 (MN) K4.05 Knowledge of the NIS design features and interlocks which provide reactor inp. 38 022 (KM1 K3.02 Knowledge of the effect that a loss of the CCS will has e on containment instrumentation 3.3 readings. 013 000 K1.02 Knowledge of the relationship betwccn the ESF and RCP. 3.6 tm4 000 A2.07 Ability to predict and mitigate the consequences of isolation of leidow n/ makeup. 3.7 014 (NM) K4 06 Knowledge of RPIS design features which provide for individual and group misalignment 3.7 indication 072 (KM) GiiN4)9 Ability to locate and operate components: including local controls 3.0 026 020 A4.04 Ability to numually operate and monitor in the control room the containment spray reset suitches. 3.5 059 tMK) A302 Ability to monitor the automatic operation of the programmed levels of the S/G. 3.1 022 (MM) A1.04 Ability to predict and/or monitor clumges in cooling uater flow due to operating CCS controls 3.3 tml oto K5 13 Knowledge of reactivity balance and that withdrawal of shutdown banks must precede dilution. 36 017 00() GEN-06 Knowledge of bases in tech specs for limiting conditions for operations and safety limits. 3.4 (m1 010 K604 Location and operation of CRDS fault detection and resets including rod control anunciators. 3.2 003 000 A2.04 Ability to predict and mitigate the consequences of failure of automatic isolation 3.3
PWR SRO Pl. ANT SYS'I D1S 40% -Group 2: Groog 17". , Ki K2 K3 K4 K5'K6 Al A2 A3 A4 SG l on2 Reactor Coolant SSstem 02 l tH Mi I;mergency Core Cool;ng S3 stem 04 010 Pressuri/cr Pressure Control Ss stem ~ 081- Pressuri/cr Level Control S) stem 01 012 Reactor Protection S3 stem 02 02 Olh Non-Nuclear Instrumentation System 03 027 Containment lodine Removal Sy stem ; 02n Ih drogen Recombiner and Purge Control S3 stem 03 029 Containment Purge System 04 011 Spent Fuct Pool Cooling Sptem 01 014 Fuel Handling liquipment Splem 015 Steam Generator System 019 Main and Rcheat Steam S3stem 05 , 1 55 Condenser Air Removal S3 stem No SRO importance factors 3 0 or greater 062 AC filectrical Distribution S) stem 01 15 (44 timergency Diesci Generator Sy stem 0M e6 073 Process Radiation Monitoring System 01 075 Circulating Water Sy stem 079 Station Air Sptem 086 Firc Protection Splem 01 101 Containment Ssstem 02
'lotals Group 2-17 l 1 1 3 4 1 1 I I i 1 '2 Sss Mode K/A# item imp.
002 (MR) K4.02 Knowledge of RCS design leatures u hich provide for monitoring reactor s essel level. 3.M 006 030 K4 04 Knowledge of the !!CCS design features u hich relate to vah c positioning on a safety 4.1 inlection signal 01I (Mio K2.01 Knowledge of bus power supplies to the charging pumps. 3.2 012 (M K) A202 Ability to predict and mitigate the consequences of a loss of instrument pow cr. 3.9 012 (MM) K1.02 Knowledge of the clTect that a loss of the RPS will have on the T/G. 3.3 016 (M M) K 103 Knowledge of the relatianship betw ccn the NNIS and the Steam Dump System. 3.2 029 000 GliN-04 Knowledge of sptem purpose and function 3.0 033 (MM) K4 01 Knowledge of SFPCS design features and interlocks u hich provide for the maintenance of 3.2 spent fuellesel 02M (XM) K5 03 Knowledge of the sources of h3drogen within containment. 36 039 000 K3.05 Knowledge of the effect that a loss of the MRSS will have on the RCS. 3,7 062 (MM) A4.01 Ability to manually opcmte and'or monitor all breakers (including available switchyard). 3.1 062 (MM) GliN 15 Ability to recognize abnormal indications for sptem operating parameters w hich are entry- 3.8 level conditions for emergency and abnormal operating procedures. tw4 000 A306 Ability to monitor automatic operation of the lid /G system; including start and stop. 34 084 000 K6 08 Knowitdge of the performance and design attributes of fuel oil storage tanks. 3.3 086 000 K4.01 Knowledts of design features u hich provide for adequate supply of water for FPS. 3.7 07.3 OtM) Al .01 Ability to predict and monitor changes in radiation levels. 3.5 201 (M10 K102 Knowledge of the ellect a loss of the containment sptem will have on the loss of 4.2 containment integrity under normal operations.
PWR -SRO Pl. ANT WSTEMS - 40% -Group 3 Group 3 4% Kl K2 K3 K4 K5 K6 Al A2 A3 A4 SG 005 Residual licat Remosal System 01 tm7 Preuuriier Relief Tank / Quench Tank S3 stem (Rut Component Cooling Water System 01 011 Sicam Dump Ss stem 045 Main I'mbine Generator ~ 076 Sersice Water S) stem 06 078- Insinunent Air Ssstcm 02 Iotais- Group 1 = 4 i i l i < Total SRO Plant Ssstemsr40 3 3 5 7 2 2 2 4 3 3 6 S)s. Mode K/A# item imp (WM 030 A3 01 Ability to monitor automatic operation of the CC System; including control of the clectrically 3.1 operated, automatic isolation valves in the CC System. 00$ (HH1 K I 01 Knowledge of the relationship between illlR and CC System 34 07n too K302 Knowledge of the effect that a loss of the I AS will has e on sy stems having pneumatic valves 3.0 and controls 076 (MN) GliN4ki Knowledge of bases in technical specifications for limiting conditions for operations and safety 3.3 limits I l l l
PWR - SRO E51ERGENCY Pl. ANT EVOI.UTIONS - 43*b -Group I Group I 24"i Kl K2 K3 Al A2 SG _
. txN) 001 Contmucus Rod Withdrawal 05 000 tun Dropped Control Rod Oo 10 INNI 005 Inoperabic/ Stuck Control Rod 03 03 't N U 011 1.arge llreak LOCA 10 (H:0 015- RCP Motor Malfunction 01 10 (M N) 024 Einergency Doration 23 000 016 Lou of Component Cooling Water 02 01 txu 029 Anticipated Transient Without Scram ( ATWS) 07 (NKI 040 Steam 1.ine Rupture 01 000 OSI l.ou of Condenser Vacuum (MH1 055 Lou of Ofhile and Onslic Powcr 01 (Nx) 057 1.oss of Vital AC Electrical Instnunent Bus 04 14 (MN) 059 Accidental Liquid Radioacth e Waste Release (Mjo 067 Plant Fire on Site 05 000 06M Control Room Evacuation 01 23 03 (xxi 069 Lou of Containment Integrits 01 000 074 Inadecpiate Core Cooling 19 000 076 liigh Reactor Coolant Acthitt 01 01 lotah - Group l=24 1 3 2 10 3 5 Ss stem Mode K/Ad item imp.
txHi 011 E A l.10 Ability to operate AFW and SWS pumps during large break LOCA. 3M O(x) (M15 GEN-03 Knowledge of the limiting conditions for operations and safety limits. 3.6 000 057 EAl04 Abihty to operate RWST and VCT valves. 3.6 000 076 GEN-01 Knowledge of sptem status criteria which require the notification of plant personnel. 3.6 (MKl 003 GEN 10 Ability to perform those actions; w ithout reference to procedure; for all casualtics 3.8 w hich require immediate operation of system components or controls. tMM) 06M liA 1.23 Ability to operate and monitor nuumal trip of reactor and turbinc. 4.4 (RHi 024 EAI 23 Ability to operate and monitor the CVCS centrifugal charging pump switches and 3.3 indicators. OtN) 029 EA2 07 Ability to monitor reactor trip breaker indicating lights during an ATWS. 4.3 l Otui 06M GEN 03 Knowledge of limiting conditions for operations and safety limits 3.6
'MMI 057 EA214 Abilit) to detennine that substitute power sources have come on line or. a loss of initial 3.6 AC.
000 074 li A l .19 Ability to monitor AFW supply tank levelindicators. 3.8 (xM) 026 EK3.02 Knowledge of the bases fc.r the automatic actions (alignments) within the CC System 3.9 resulting from the actuation of the ESF. 000 015 E A2.10 Ability to detennine w hen to securc RCPs on loss of cooling or scal injection. 3.7 000 003 EA l.06 Ability to monitor RCS pressure and temperature. 4.1 WM1 015 EAl.03 Ability to operate and monitor the reactor trip alanus; switches and indicators. 3.8 (x10 026 EA101 Ability to operate and monitor CC/ Cooling Water temperature indications. 3.1 000 005 EK2.03 Know ledge of the metroscope as it applies to the inoperabic/ stuck control nod. 3.3 (x10 069 E A 1.01 Ability to operate and monitor isolation valves; dampers and clectro-pneumatic 3.7 desices following a loss of contahunent integrity. h)0 055 EK 1.01 Knowledge of effect of battery discharge rates on capacity as is applies to station 3.7 blackout (K10 001 GEN-05 Knowledge of the annunciator alanus and indications; and use of the response 3.6 Instnictions during a Continuous Rod Withdraw al emergency . tr10 076 EK201 Knowledge of process radiation monitors. 3.0 000 007 E A 1.05 Ability to operate and monitor plant and control room ventilatica s)sicms. 3.1 (x)0 040 EK301 Knowledge of the bases for operation of steam line isolation s ah cs during a steam line 4.5 rupture, (M M) 06M EK2 01 Knowledge of controllers and positioncts as applied to control room evacuation 3.1
PWR- SRO ESTERGENG Pi ANT EVOl.UTIONS - 43% -Group ' liroup 2 - Id'. Ki K2 K3 Al A2 SG l 0 00 007 Reactor Trip 04 0 EU wJ Preuuri/cr Vapor Space Accident 06 0 (Q) 009 Small lircak LOCA 01 37 0 (HI- 022 Loss of Reactor Coolant hiakeup 01 O fc) 025- Loss of Residual Heat Remosal Ss stem . 03 OtI O (RA 027 Pressuri/cr Preuure Control S) stem Mallanction 05 0 00 - 012 Lou of Source Range Nuclear Instrumentation 01 0 ta) 033 Loss of Intennediate Range Nuclear Instrumentation 01 0 00 017 Steam Generator Tube Leak 01 0 tu) 01M Stcam Generator Tube Rupture 12 0,00 054 - Loss of Main Facduater 01 0 (Mi 0$M Loss of DC Power 01 0 00 060 Accidental Gascot.s Waste Release 03 0 (c) 061 Area Radiation Monitoring S) stem Alanus 0 00 065 Lou ofinstrument Air 03
~lotals Group 2 = lb i 1 2 5 4 ?
Ss stem . Mode KN.# item linp. Otm 054 EA 1.01 Ability to operate AFW controls includmg'.hc use of alternate AFW sources during a 4.4 loss of main feedwater. tMm 037 EAl 01 Ability to monitor maximum controlled depressurization rate fo > s ed S/G. 3.6 iKm 007 EA2 04 Ability to manually trip the reactor and carry out actions in ATWs 1 ,P if reactor 4.6 should has e tripped but has not done so. (Mni OU EK301 Knowledge of the basis for tennination of startup following loss of intermediate-rtmge 3.6 instrumentation. (ux) 03K E A 1.12 Ability to monitor S/G blowdown line radiation rnonitors. 4.3 OtN) 032 EK2.01 Knowledge of pow er supplies; including proper switch positions as applied to a loss of 3.I d source range nuclear instrumentation. (M K) 027 GEN-05 Knowledge of the annunciator alanus and indications; and use of the response 3,3 instructions. (M N) 009 EA101 Ability to monitor RCS pressure and temperature during small break LOCA. 4.3 ink) 00M GEN 4W> Ability to locate and operate components; including local controls. 3.H (MM) 009 EA2.37 Ability to detennine the existence of adequate natural circulation during small break 4.5 LOCA. (MR) 025 GEN-06 Ability to locate and operate components; including local controls 3.6 (Run 022 E A2.01 Ability to determine whether charging line leak exists. 3.8 tax) (Mio EK3 01 Knowledge of the basis for actions contained in EOP for accidental gaseous w aste 4.2 release-(MD) 025 EA2 03 Ability to determine or interpret increasing reactor building sump level. 3.8 (nm 058 EK 1.01 Knowledge of battery charge: equipment and instrumentation as applied 10 a loss of 3.1 DC pow cr. txx) 065 EAl03 Ability to monitor restoration of sy stems served by instrument air w hen pressure is 3.1 regained
.. - = .. -
PWR. SRO ESIERGENCY PIANT EVOLUTIONS - 43% -Group 3 Group 3 .3% - Kl K2 K3 Al A2 SG tNM) . 02M Prewuri/cr Les el Malfunction 13 iWx) 0% I uct flandling incident 01 (WM) 0% Loss of Olisite Powcr 50 Totals Group 3=3 0 0- 0 1 2 0 Total SRO lignergency Plant Lvolutions=43 2 4 4 16 9 8 Spicm Mode K/A# itern Irnp. (NMI 028 E A2.13 Ability to deternune actual P4R level; given uncompensated level with an appropriate 3.2 graph tx)0 056 EA2.50 Ability to determine that load and VAR limits; alarm setpoints; frequency and voltage 3.1 limits for ED'Os ase not being exceeded. (H)0 0% l EA 1.01 Ability to operate and monitor the containment purge ventilation system. 3.8
. .- - . . . . .. - , . _. -_ . .- - . - . _ . . ~ .. .
PW R 5 HO h. ANT-Mil >E GENERIC RE5PONSillit.lTIES 17% K/A81 Imp.
- A i 11, Abilit) to take actions called for in the Facihty timergenc3 Plar' including (if requiredl supporting or 4.4 acting as the Emergencv Coordinator, Ai 01 Abihty to locate and use procedures and station directh es related to shlit stalling and aethitics 3.4 Ki 14 Knowledge of salet) procedures related to contined spaces 36 Kl.07 Knonledge of safety procedures related to electrical equipinent. 3.7 Kl 02 Knowledge of tagging and clearance procedures. 4.I' Kl lb kreouledge of facility pratection requirements; including fire brigade and portable fire-fighting equipment - 4.2 i utige.
Ki OM Knowledge of r.afety procedures related to high temperature 3.4 Al.lu Ability to coordinate persoluiel acth ities outside the control room. 3.9 A 1.11 Ability to locate control room suitches, contmis: and indications; and to determine that they are correctly 4.1 reflecting the desired plant lineup. r K I 01 Knowledge of to Cl R 20 and related f acility radiation control requirements. 3.4 K1.15 Knowledge of safety procedures related to hydrogen 38 A 1.0M Ability to obtain and interpret station reference material such as graphs; monographs and tables uhich 3.1 contain s) stem perfonnance data. A1.09 Ability to coordinate penonnel acth itics inside the control room. 3.9 A1.02 Ability to esecute procedural steps. 3.9 K 1.05 Knowledge of facility requirements for controlling access to vital / control areas. 3.4 Kl.09 Knowledge of safety procedures related to high pressure. 3.4 Kl i1 Knowledge of safety procedures related to chlorine. 3.5
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MS.lN ' indiildual WalkdhMudLIV5LQMilit!q FpnnfS10J 2 Esamination Level _ SRO(U) t'acility; [Mt GilhPM)- Week of Esamination: ()4/J 4/97 Esaminer's Name (print). Systerr / JPM - Safet) Planned Follow up Questions: Function K/A/G // Importance // Description
- 1. RCP - Reactor Coolant IV liow is RCP seal bleed olr temperature monitored and Pump Start controlled?
(alternate path, plant 003.tXK)/A1.02//2.9/2.9// Abilly to monitor changes in parameters shutdow n) associated with contiots including RCP pump temperatures What is the basis for restriction on pressuri/cr level or S/G - primary / secondary temperature difference prior to starting RCP7 002/020/K$.08/D.8/4.1// Know ledge of relationships betu ccn effects in the primary and secondary coolant system.
- 2. CSS / 0335 Containment VI Why is the pump discharge valic closed before starting the Spray Pump Operability containment spray pump?
Test 026AMX)/A3.01//4.3/4.5// Ability to monitor pump starts and (cnginected safety feature) conect valve positioning. What cifcct would an hudvertant CSAS hae on containment spray valves while pump was running during test? 026AXX)/A2.03/D.9/4 2// Ability to predict impact of failure of ESF on CSS operation Simulator JPMs abos e this line. The following JPM may be performed in the Simulator or in the Plant.
- 3. AFW/new Control Room Vill Also(11, What clTect will the transfer to Al 185 have on the sciccted lead
, Evacuation til, V) ,, charging pump? (new, AOP, plant shutdown) 004/000/K4 04/D.20.1// Knowledge of CVCS design features uhich provide for mamial/ automatic trcnsfers of control. Ilow do you transfer DC contati power to Al.179 to the alternate powcr supply? 061AX)0/A2.03/D.lO 4// Ability to mitigate a loss of DC control power to the AFW system. Plant JPMs below this line.
- 4. DC/0306A Alternating Vil What safety precaution must be taken w hen a switchgcar room and Securing battery roll up dooris opened?
clurgers 063/000/ GENI //3. lD.2// Knowledge of operator responsibilities during all modes of plant operation. Ilow would consequences ditTer if voltage regulator on an in sen ice battery charger failed high or low? 00XXX)/K1.03//2.9/3.5// Knowledge of cause/cffect relationships betw een DC clectrical system and battery charger and battery.
- 5. WGDS Transfer waste XI What actions would be required if Al l10 (gas analizer) was gas from vent inoperabic during this evolution 7 header to inservice decay 071 AXXVK4.06//2.7D,5// Knowledge of design features for tank monitoring of waste gas release tanks.
(new JPM, RCA entry) Why is there a caution not to drain the vent header while collecting a Vti gas sample? 071ANK)/A4.29//3.08.6// Ability to mamially operate sampling gas concentrations in WDGS decay tank.
. Examiner: Chief Examiner;
) .- j l
ES 101 Adruinhtnttive TopictQpt. ling Fonn ES401 1 Esamination Level: SRO Facility:E91LCathqua Week of Esamination: 04/14/97 Esaminer's Name (printt Administrative Describe method of evaluation: Topic / Subject 1. ONE Administrative JPhl, OR Description 2. TWO Adctinistrative Questions A.! Temporary Temporary Procedure Change, Change of Intent Detennination JPM rnodifications oIpr0Ceditfes '
- Plant JPht . Resicw shutdown ruargin calculation Parameter Verification A.2 Sun cillance JPM - Review of OP ST-SHIFf4K)]
4 Testing h A.3 Radiation RCA Entry and Esit JPM , control i A.4 Emerpcacy Event Classitication JPM Plan Esamincr: Chief Examiner
t K5901_ sggnariglvents _ fann.Es-3013 Simulation Facility : l'ortf;tl heurt Scenario No : SIM 971 Esaminers: . Applicants: . Ini'la! Conditions : Kt% power Turnover: 1 W 54. AC-3 A tagged out I:sent Malf. 1:s ent I)cncri l, tion [ Esent No. No. ' T,s pc
- I i N1 WR Channel D fails 2 i Controling I'ZR pressure ch:umel fails high 1 C 4n gpm S/G Tube leak ( A S/G) 4 C llCV-978 does not isolate 5 R/N Emergency Shutdown a
6 I Condenser vacuum su itch 952A indicates low vacuum (high pressu d
.._I 7 M S/G Tube Rupture Shaded entries are tc be initiated by a cue from examinct.
* (R)cactivity, (C)omponent, (M)ajor (N)ormal. (1)nstrument.
Examiner. Chief Examincr:
FS 101 _SqqnarigJhentL FontLES 30L2 , Simulation l'acility; Egnfalhourt Scenario No : SIM 97-2 Esaminers: Applicants: Initial Conditions : 100% pow er Ttirnover: FW-54. AC 3A tagged out Esent Malf. Esent Esent flescription No. No. T3 pc' 1 C Dropped rod 2 R/N Reduce Power to 70% 3 i T 2987 (letdown ilX CCW outlet temperature) falls low 4 i S/G les el LT-903X fails high 5 C 1 A4 bus lault/rx trip
~6 M loss of of fsite power 7 C RCP RC 3C breaker does not open (D/G output breaker will not closc)
Shaded entrics are to be initiated by .uc from examiner.
* (N)ormal, (R)cactivity, (1)nstrument, (C)omponent, (M)ajor Examiner:
Chief Examincr:
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- l J S:lR L _ _ _ Scenariofignis fonp,gs.39b3 Siinulation :'acility: EgtLCalheun Scenario No,: SIM 97-3 l I
Esarnincts- Applicants , Initial Conditions : ItN)*; pow er Tiirnos cr. FW-4 A. DG I tagged out Esent Slalf. Es ent Ei ent I)c>cription N o. No. Tspe' I C RC 3C lower seal fails 2 1 Pressurifer level channel fails high 3 C RC-3C middle scal falls 4 R/N Emergency Shutdown 5 1 S/G steam flow channel fails high (i C 2 stuck rods 7 M Steam line break in contaimnent 8 i CPilS fails to actuate Shaded entries are to be initiated by a cue from c.saminct.
* (N)ormal, (R)cactivity. (1)nstrument. (C)omponent. (M)ajor Euuniner:
Chief Euuniner-
i l J5dfll_ SSc!!arighenu __ fprm ES 3013 - Simutaflon Facility: for1Calltgyn_ Scenario No.: SIM 97-4 thaminers: _ Applicants:
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I initial Conditions : 1(mapowcr Turnour Pow cr range Ni channel "C" tagged out, bistables tripped. CPR in progress Esent M alf. Esent Esent Description No. No. T,s pc
- I I Loss of Ni pon er range channel "B" 2 IUN Poner reduction to 70"i
' Irr-910 f alls high 3 I
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4 l VCT level channel fails low 5 M Loss ofload
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6 C/M IORY lails open 7 C IORV block valve will not close 8 i VI AS does not actuate Shaded entries are to be initiated by a cue from examinct.
* (N)ormal, (R)cactivity, (1)nstnament. (C)omponent, (M)ajor Examincr; Chief Examiner
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j I S9t!l,-_ _ _ Sqenariolien;5 [ornLliSQit! 3 > Simulation Facility forifillhoort Scciario No :SIM 97 5 (Sparc1 thaminers. Applicants: initial Conditions : 50*6 power Turnos cr: Continue powcr increase Esent M alf. Es ent Esent Description No. No. T,s pc
- I R/N Raise power 2 i FW 110w transmitter fails low (affects XC-105) 3 i PT-210 fails low 4 C Instrurent Air leak
,'. M Manual reactor trip required 6 C Generator field breaker fails open 7 M CRDM ejection following trip i
Shaded entries are to be initiated by a cue from examiner. (N)ornul. (R)cactivity. (1)nstrument, (C)omponent. (M)ajor Examincr: _ Chief Examincr:
l l 3/12/9] l
-To: Ryan Lantz and John Pellet The operating test is attached.
The Administrative section of the exam consists of 5 JPMs. These are the same as described on form ES-301-1 in the initial outline. The Walk-Through test section of the exam is as descnbed on form ES-301-2 in the initial outline. Each JPM includes at least one new question and 2 of the JPMs are new. Five scenarios are provided for the simulator portion of the exam. It is expected that 4 will be used. One is provided as a spare. All of the scenarios are new. Some minor changes were made to the scenarios from those provided in the initial outline:
. An inadvertent AFAS actuation was substituted for a Wide Range Nuclear Instrumentation failure in the first scenario. The WR NIS was recently modified and the simulator has not been fully updated.
. In the 4th scenario, the failure of HPSI pumps to automatically start was substituted for failure of VIAS actuation. We felt that the failure of VIAS actuation was too similar to the failure of CPHS actuation in the 3rd scenario.
The selected scenarios reflect high risk events and risk significant operator actions from the Fort Calhoun IPE. For example:
. The failure of reactor coolant pump seals is included in the 3rd scenario. This event is a contributor to risk in the FCS IPE.
. The failure of a reactor coolant pump breaker to open preveniing the diesel generator breaker from closing on the bus is an event where credit was taken for operator action in the FCS IPE.
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-- Revision 0 February 25,1997 Fort Calhoun ' Station _ Operator Training-JOB PERFORMANCE MEASURE JPM No: JPM OP-ST SHIFT-0001
' JPM
Title:
Required Shift Surveillance _; Approximate Time: 8 minutes Actual Time: ._._._. i- : Reference (s): 1) OP-ST-SHIFT-0001 (rev 51)
- 2) NRC K/A 001010,A4.04,RO3.5/SRO4,1
- Verify current reference revisions match those listed above Operator's Name: SS #:
All Critical Steps (*). must be-performed or-simulated in accordance with the standards contained in %is JPM. The operators performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluators .. Signature: Date:
- Reason, if unsatisfactory:
( e, ._- -.,,--,-3 m ,, m , _ - . _ . _ _ _ _ _ . _ _ _ _ _ _ - _ _
_ . _ __._ __ . . _ _ _ _ ___ _ _ _ .___ _. ___ w - _ _ _
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- Rovision 0-February 25,1997 Fort Calhoun Station- Operator Training
-JOB PERFORMANCE MEASURE-JPM No: -JPM OP ST SHIFT 0001 JPM
Title:
- Required -- Shilt Surveillance Initiating Cue: Complete the Shift Supervisor review for Wednesday of the attached ; portion - of OP ST-SHIFT-0001. 4 STANDARD:
- RM/057 counts have doubled, therefore sampling per S.O. G-105 must be started.
i I M T- * *
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Initiating Cue: Com,nlete the Shift- Supervisor review for Wednesday of the. attached i portion of OP ST-SHIFT 0001, J
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-~ ,. - ,
i Fort Calhoun Station Unit No.1 I OP-ST SHIFT-0001 SURVElLLANCE TEST
Title:
OPERATIONS TECHNICAL SPECIFICATION REQUIRED SHIFT SURVEILLANCE FC-68 Number: 48030 R:ason for Change: This Procedure change to page 9 incorporates the comparison / decision funcion of Tech. Spec. 3.1 Table 3-1 Item 1.b. adjustment of power rango safety channels to agree with heat Balance. Contact Person: J. R. Tucker ISSUED: 01-30-97 9:30 am R51
3 o ' FORT CALHOUN STATION OP ST-SHIFT-OOC 1 SURVEILLANCE TEST PAGE 1 OF 48 OPEPATIONS TECHNICAL SPECIFICATION
. REQUIRED SHIFT SURVEILLANCE
- SAFETY RELATED 1 _ PURPOSE- u To satisfy Technical Specification requirements for all Operations Shiftly and Daily-Surveillance Checks.
- 2. REFERENCES / COMMITMENT DOCUMENTS
- 2. 1' --_ Applicable Technical Specifications are listed on the appropriate Shift Data Sheets.
- 2.2 Applicable procedures are listed on the appropnate Shift Data Sheets.
2.3 Ongoing Commitments l
*- CID 882063, LER-88-013, Step 5.1
*- CID 780026, LER-78-02, Page 5
- e- .CID 910569, LER-91-14, Pages 14,15 and 16 2,4 Standing Order S0-G-23, Surveillance Test Program
- 3. DEFINITIONS _,.
3.1 Surveillance interval (Technical Specification 3.0.2) e- Shift - at least once per eight (8) hours,
- e. Daily - at least once per 24 hours.
3.2 Specified Time Interval (Technical Specification 3.0.1)
- e. Each Surveillance Requirement shall be performed within the specified .
Surveillance Interval with a maximum allowable extension NOT TO EXCEED 25% of the Surveillance Interval x_, R51
y.,. .. . - - .- - - . - .- . . - - . - . - - FdRT CALHOUN STATION - OP-ST-SHIFT-0001 PAGE 2 OF 48 SURVEILLANCE TEST. 1
- 4. EOUIPMENT LIST l
_ None? c 5.L PRECAUTIONS AND LIMITATIONS - , [5.1]' All anomalies'and deficiencies shall be reported immediately to the Shift Supervisor , and noted in tho' Remarks Section. An immediate check shall be made-to verify ,
' Limiting Cornditions for Operation,' per Technical Specifications, have not been
- exceeded.
5.2 - A Condition Report shall be initiated in accordance with Standing Order SO-R-2 to'
- report any anomalies or deficiencies. The Condition Report number shall be
.= recorded on the Comment Sheet / Chronological Log. -
-5.3 -
A Maintenance Work Request (MWR) shall be initiated to correct any reported deficiency and the MWR number shall be referenced in the Remarks Section; 5.4 No maintenance shall be conducted within this Surveillance Test other than that specifically directed by this procedure. 5.5 Test data shall be evaluated by the Shift Technical Advisor (STA) AND reviewed by the Shift Supervisor for acceptability pri::t to the end of the normal operating ahlft in ' which the data was taken. 5.6 The System Engineer shall be notified within 24 hours of the completion of this test " of any marginal, unexpected, or unacceptable results, d 57' This Surveillance Test shall be performed at approximately the same time overy shift. This will preclude the possibility exceeding the siloi.vable Surveillance Test perfonnance interval of 8 hours (+25%). 5.8 The use of N/A (not applicable) in this procedure shall be in accordance with the >
- requirements listed in Stending Order S0-G-23.
The completed Surveillance Test procedure and all applicable Attachments shall to 5.9
' signed and dated by the person (s) who actually performed the test.
l' R51 7
.5--.< E . , - . , . - , . . . - . . . , . . . , . [ s. - - - . ~ , ,, - - , - - , - e---- , , , ,w v. o-b - - , ,m.. ,-.c--. . -~
FORT CALHOUN STATION OP-ST-SHIFT-0001 SURVEILLANCE TEST PAGE 3 OF 48 INITIALS /DATE
.. INITIAL CONDITIOtiS 6.1 Procedure revision verification:
Master Revision No. 5' / />e497 62 An RWP has been issued, if required. RwP No. ? ? - 00 o .2-l > As-fr
- 7. PROCEDURE CAUTION All data required to be taken for the applicable mode aust shift mull be completed within the first two (2) hours of the applicable normal operating shift.
NOTE: The Surveillance Test Signature Sheet only needs to be completed once each week for each individual recording, evaluating or reviewing data on the Shift Data Sheets. 7.1 Complete the Surveillance Test Signature Sheet.
~
7.2 Record all data required to be taken duriilg the applicablo mode ans[ operating shift as scheduled on each Shift Data Sheet. 7.3 WHEN all applicable data has been recorded for the shift, THEN the Control Room Operator must sign the " Completed by" section of the Shift Review Sheet for the appropriste Day and Shift. REMARKS %er R51
[~..: LFORT CALHOUN STATION: = OP-ST-SHIFT-0001) LSURVEILLANCE TEST; PAGE 4 OF 48:
- 8. . RE STOR.ATION
!None- q ' 9. - ACCEPTANCE CRITERIA Acceptance Criteria are specified on the applicable Shift Data Sheets.
10 ~ TEST RECORQ-
. This entire procedure AND a copy of any additional procedures or attachments required to -
be performed to complete this surveillance.
.11. REVIEW -
The Manager-Operations is responsible for ensuring the completed-Surveillance Test is reviewed in a timely manner and forwarded in accordance'with Standing Order S0-G-23.
' Test data shall be evaluated by the STA and reviewed by the Shift Supervisor for ,
acceptability prior to the end of the normal operating shift in which the data was taken. : 4 M a E R51
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i 1 i 1 1 i OF ST-SHtF T 4Kiot ! FORT CALHOUN STATOtt PAGE 19 OF 48
';;;:rT DATA SHEET WEEK ENDING- AREA RADIATION MONITORS
^ TIME INITIALS APPLICABLE MQDEL MSTRUMENT ._ __
Modes 1. 2. 3. 4 and 5 j 2 mom PROCEQ11BE BEEEEEtLCE; N Wam S P. BlHr 2500 _ps O O wH _ _ _ _ _ _ . _ _ - ._. OP-ST-RM 000t Meter Rooding M 4/ 4/ 2330 4 730
$ TECH.SEEC m REEEEEHCR :
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- 2 2: Tat:!e 2-10, item 1 T
- 31. Table 3-3. Item 3 a T Meter Reedwig R/Hr 4. ; 4/ 2330 0730 U N Wam S P. R/Hr ygg3 .2fE3 N/f E ACMEE CRREM W
- Meter readings are d tal displayed or t W Meter Reading RJHr /_/ 4/ 23h4730 E are on scam and wss .an m >
E - o wam/ Alert setposits. or the SNt
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Revision 0 February 25,1997 Fort Calhoun Station Operator Training JOB PERFORMANCE MEASURE l JPM No: JPM SDM JPM
Title:
Shutdown Margin Calculation Approximate Time: 8 minutes Actual Time: - Reference (s): 1) TDB-V.9 (rev 28)
- 2) NRC K/A 001010,A4.04,RO3.5/SRO4.1 Verify current reference revisions match those listed above Operator's Name: SS #:
All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. The operators performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluators Signature: Date: Heason, if unsatisfactory:
-e w p T +P =
m - ey y p-
Rovision 0 February 25,1997
.6 4 Calhoun Station. Operator Training JOB PERFORMANCE MEASURE
- -. IP! !k- JPM-SDM
- JPM
Title:
Shutdown Margin Calculation
, Initiating Cue:
A RCS cooldown to Mode 4 is in progress. RCS Teoidi s 505'F and lowering. Current core burnup is 7800 MWD /MTU. Review the attached shutdown margin calculation for approval to continue the cooldown. STANDARD: 1) Step 6, correct value should be - 665 ppm. (incorrect curve was used)
- 2) Step 9, correct difference should be ~100 ppm.
(Carryover eror from step 6)
- 3) Step 10, RCS boration required therefore RCS
, cooldown must not continue until the boron
, concentration is corrected.
, , , - - - - - ,- , ,,r- -- ,a ..-,--,-,-,.,.,,n+p-,. - - - . y . ,. ,,, - - - , , - , , . - . .,c ..m , , . -,
- . _ . _ = _ - - - .. __ . - . _- . _-_. . -__ -. - . - . ._ . . . . --- .
l$ Initiating Cue: A RCS cooldown to Mode 4 is in progress. RCS Teoie is 505'F and I lowering. Current core burnup is 7800 MWD /MTU. Review the attached shutdown rnargin calculation for approval to continue the cooldown. I A d 'l
- ,. -. y-a -
,. , . - . , , ,_.,,.v. . . . , , . . ,. _ . ,. , _ ,....-- _ _ , , _ ,, ~ , , . - , , y,, , , _ , . ., -, ,,,, , , . ,_,,,, , , ,. . . - . _.r.._,s.-,,
T Fon Calhoun Station Unit No.1 TDB-V.9 TECHNICAL DATA BOOK
Title:
SHUTDOWN MARGIN WORKSHEEi FC-68 Number: 46726 Reason for Change: Revise for Cycle 17 TDB figure changes, clarify source for RCS loop temperature, add N/A to step 7 of Part til if in Mode 5. Contact Person: C, Stafford , ISSUED: 1122-96 2:00 pm R28
FORT CALHOUN STATION TDB V.9 TECHNICAL DATA BOOK - PAGE 7 OF 16 PART 111 - RCS Temperature s 515'F Conditions INITIALS /DATE
- 1. Date/ Time: EJyAWo /
- 2. Burnup: 7J'<V MWD /MTU /
l
- 3. Verify that the Reactor Coolant System indicated loop !
temperature is s 515' F using a, or b. below. I
- a. if on shutdown cooling, use TE - 346Y (TR-346, RED Pen = Outlet temperatu.*e) l
= N /A 'F
' b. If not on shutdown cooling, then use the lowest valid RCS loop temperature.
= @ S~ 'F / .
- 4. Reactor C5olont System Boron Concentration (Boron Analysis must have been performed within the past >
24 hours or more recently if boration or dilution has occurred.) W, 9 opm
- 5. Verify that all regulating and shutdown CEks are inserted to at least the Lower Electrical Limit (LEL) /
- 6. Determine the required boron concentration by using the applicable TDB Figure ll.A.3, based on RCS temperature '
(3.a or 3.b) and core burnup (2). If in mode 5 enter the refueling boron concentration from the COLR. kV opm 6 r p . R28 , i t
, ., ,_.r_7., 4.s-r,--...,- -....v..,.- . _ . ,e_. . . . . . , . ~ . _ , . . , _ _ .- . .-_. w.- _ #--..--.-- s.-. _ - - _ . -
FORT CALHOUN STATION TDB V.9 TECHNICAL DATA BOOK PAGE 8 OF 16
- 7. Determine the deviation between actual and predicted critical a
boron (N/A if in Mode 6). l
- a. Reactor Coolant System boron concentration prior to shutdown or trip. (N/A if in 4 . Mode 5) {i 6*&O opm 7.a 6
i
- b. Reactor power level before shutdown or trip (% of 1500 MWth) lod %
7.b i l '
- c. Using the burnup in step 2 and the power from Stop 7.b, c'etermine the critical boron concentration (use TDB Figures ll.A.2.a through II.A.2.a G&C opm 7.c
- d. Calculate the deviation between prooided and actual boron concentrations. If the value of the deviation is negative, entw zero, ch.O . 6M = 0 . .,_ ppm ?
7.a 7.c 7.c ,
- 8. Adjust the required boron by adding the value of the boron deviation, if in Mode 5, enter the refueling boron concentration on line 8.
4 400 + 0 = 40 0 oprn . 6 7.d 8
- 9. Calculate the difference between actual and adjusted required boron.
i '%9 . tcw a -H opm 4 8 9 9 i 1 4 R28 o,-__ ,wv.+=v--,,,+,aww,---r+--*www.v-,--,-,e -----v- , -h -w ' - - - rc = , y- --w--ww,--w,--v.---,+,- - - , , - - --%,-,.==w.-+-,--*e
- . - =__ _. . . - _ . . - . .-,
l FORT CALHOUN STATION TDB-V.9 TECHNICAL DATA POOK PAGE 9 OF 16 SolubleEnton Concentration 10 a. IF 9 is greator than or equal to zero, the boron concentration is adequate,
- b. IF 9 is less than zero, use Ol ERFCS-1, Procedure 32 or manual calculations and borato the reactor coolant system to the concentration given in 8.
REMARKS Completed by Datomme / 9 e R28
FORT CALHOUN STATION TDB-V.9 TECHNICAL DATA BOOK - PAGE 7 OF 16. , PART 111 - RCS Temperature s 515'F : i Conditions INITIALS /DATE
- 1. Date/Timo: /
- 2. Bumup: Wad MWD /MTU /
- 3. Verify that the Reactor Coolant Sys'em indicated loop ,
temperature is s 515' F using a. cr b. below. ~
- a. if on shutdown cooling, use TE - 346Y (TR 348, RED Pen = Outlet temperature) wa .p
~
- b. If not on shutdown cooling, then use the lowest valid RCS loop temperature.
. co r *F /
- 4. Reactor Coolant System Boron Concentration (Boron Analysis must have been performed within the past 24 hours or more recently if boration or dilution has occurred.)
4 s' opm
- 5. Verify that all regulating and shutdown CEAs are inserted to at least the Lower Electrical Limit (LEL) / >
6.- Determine the required boron concentration by using the applicable TDB Figure ll.A.3, based on RCS temperature , (3.a or 3.b) and core bumup (2). If in mode 5 enter the refueling boron concentration from the
'COLR.
(l4N Iwo opm 6
/
,.,7 q ,4 , , . . . + . . , . , - . . . . , . . , , . , ,,,,m,,,,,, ,,,,,,,w mg__,.._,,.,,,,,.,,,,,..,,,..,n,..,,,, n,_.n,,.y._,,.,-..=.,,,,, .,-,w__.--,
i FORT CALHOUN STATION TDB V.9 i 1 TECHNICAL DATA BOOK PAGE 8 OF 16 l I
- 7. Determine the deviation between actual and predicteo critical boron (N/A if in Mode 5). l ;
- a. Reactor Coolant System boron concentration prior to r,hutdown or trip. (N/A if in Mode 5) l l
d a opm c i 7.a
- b. Reactor power level before shutdown or trip (% of 1500 MWth) !
na % i 7.b
- c. Using the burnup in step 2 and the power from Step 7.b, determine the predicted entical boron concentration (use TDB Figures ll.A.2.a through II.A.2.d)
'*4 0 opm 7.c
- d. Calculate the deviation between predicted arid actual boron concentrations. If the value nf the deviation is negative, enter zero. ,
Lc . _ G .a _ ,. _,, = 0 opm
- 7.a 7.c 7.d ;
- 2. Adjust the required boron by adding the value of the boron deviation. If in Mode 5, enter the refueling boron concentration on line 8.
! /, u + a = Ino opni , 6 7.d 8 - j fus) (@S) . 9. Calculate the difference between actual and adjusted required boron.
~ t. c' . wo a -35 opm 4 8 9
(&cs) (~'0*b R28. ,
FORT CALHOUN STATION TDB-V.9 TECHNICAL DATA BOOK PAGE 9 OF 16 Soluble Boron Concentration 10 a. IF 9 is greater than or equal to zero, the boron concentration is adequate.
- b. IF 9 is less than zero, use Ol ERFCS-1, Procedure 32 or manual calculations and borate the reactor coolant system to the concentration given in 8.
REMARKS _ _ _ , , . - Completed by Date/Timo / 4 6 G R28
.. _ . =- _-
FORT CALHOUN STATION TDB.ll TECHNICAL DATA BOOK PAGE 8 OF 37 O O O O O O O O O O O O O O D O D O D O D O D o D O C 4 4 4 f M M N N n - O O
- - n - n - e e - e e n e O
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v g .ap o . l l l l l l l l
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--i n Figure ll.A.2.b $@
Cycle 17 Critical Boron Concentration 'vs. Power E2 o> (ARO, Equilibrium Xenon) >r rr - 2 to 6 GWD/MTU ao P C
-4 Z 1550 g...>g. ..t . . .
.t. ..i .. . .
g ... g... ; . . . g_. . . . -: 1550$$
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Figure li.A.2.c n1 O Cycle 17 Critical Boron Concentrotion vs. Power D Zo (ARO. Equilibrium Xenon) o>
>r rI 6 to 10 GWD/MTU CO C,
1250 . - - . g- - -
=1 1 1' - -
g == 1 g - - . g
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[ Figure ll.A 3.c $@ --' I Cycie 17 Boren Concentration for z_ o
- i. 4% Shutdown Mar 9i n vs. BurnuP 9A cz l
(ARI. Zero Power - 500 Degrees F. Equilibrium Somcrium) y@ ,
-4 z cn i
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Figure ll.A.3.d oay i I; Cycle 17 Boron Concentration for i_ O O '> 1 - 4%. Shutdown Margin vs. Bumup jeg ! l L (AR1, Zero' Power - 475 Degrees F. Equilibrium Samarium) SO
-4 z c
l 13M
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~
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a l Burnup (GWD/MTU) . ! 4 I b 3 f v.. . ,-- . -., , , . , _ , , , . - -
. - , , - , . .- - - . ~ - , . , , - - - , - . , , , . - _ , , - . -
i9 + am Fi9ure ll.A.3.e - mo ox -> Cycle 17 Boron Concentration for T_oH i 4% Shutdown Morgin vs. Burnup 9A t- I i 1-(ARI. Zero Power - 450 Degrees F. Equilibrium Somorium) oO yC ! -i z 1300 : ''t -l l - 1300 g, $ _'..'g'.'
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o 200 200 > !- ! _ - - " t. - " " '." - " " *. " " ~ "i. " " " t. " " " t. " " " 1. - " - " t. " " ~ t. " " " I. " " - " . " " " t. " " " t. " " * " - - - g m j jg ,,,,... . p...-..J- *.... .......p......p...... ...............p "..-.p...... -*.....a....... ......s....... ....- 100 $ ,
. . . . . . . . . . . . 1 i
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. . . . I....I.a ..I.a e a Ie .a a Ia e a a la a a e Ia . a . Ia a a a I.a . a la a a a Ie a . .Ia a a . I....l....~
5 6 7 8 9 10 11- 12 15 N l- -^ 0 1 2 3 4 13 14 i 1 Burnup (GWD/MTU) - 1
?
4 -
h .i sm Figure li.A.3.f mO l a :n
- Cycle 17. Boron Concentration for i__do 1
4% Shutdown Margin vs. Burnup 9# r- 1 l- (ARI. Zero Power - 400 Degrees F. Equilibrium Samarium) oO
>C az i i 1400 _. . . . g . . . I ' ' ' .1 ' ' ' .l ' ' ' ' l ' ' ' ' .l ' ' ' ' .l ' ' ' ' .l ' ' ' ' .l ' ' ' ' .l ' ' ' ' .l ' ' ' .l ' ' ' 'T. ' ' ' ' l ' ' ' ' - 1400 $ $
i O O >d [ 1300 xo i 1300 ~ " - - - + - " " i. " " - " .! - " - " 4. " " " t. " " " I. " " " i. " " " +. - " " "*" " " - P " - " 4. " " " +. " " " i. " " - ~ P " " 7
. . . . Z r
1 4
. . . . . . 3 1200 1200 [-~---------f.----
'-"-~~~I.------t."----f.--"---!.------+.------.l-"----!.-""-4.----"Y.------T.--"---!---~--
l, 4 m - ! E a 1100 ---~ " ?,-" j"-~ ~: -""! " "i GQ i.WouT! - """?. " " " i." - " " i. "" " !. " " " i. - " - - l.- ~ ~ - " .i " -" 7
'100 t v
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,u _
t C _ 8 800 "* - " t. " " " P. " " " I " " " t. " " " P. - " - " P. - - - - - 800-4 C 1" - " t. " -- - t. " " - " I. " " " t. " " " t. - " - " i. " " " 4. "
* * * *
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'E 500 500
-> 1 -- -- t. - - - - t." " " - :. - " " "I " " " t. " " " P. " " " . " " " 4. " " " i." " - " !. " " " . " " t - " " - P. . " " " :. " - " m 2--
1 3 400 400 l 2 1 - - - " t. " - - " '. " " - " .* " " " 1. " - - " t. " " " '. " " " 1. " " " t. " " " t. " " - P. " " " ! " - " . ""'.""-"P.""-
.g 1----- 300. y i-0 _
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g [ i
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r 200 200 - - - - - - t.
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y o,g 4
....i....r....i....i....i........i....i............i........i........- ,gg j b
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, _ _ , _ _ _ ._ ,_ _. _ -_ _. _ . . . _ . _ _ _ __F
E e c . g-
- --i m Figure ll.A.3.g .
l Cycle 17 Boron Concentration for M' . 41 Shutdown Margin vs. Burnup D r- r p (ARI. Zero Power - 350 Degrees F. Equilibrium Samarium) g@Z
-4
- 14C3
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. Z 4
3 1200 1200 l
---------T.-----.~~~--~4.~~~~~-+.-"-"I.-~~~~~~I-"'--->~~"~~~I-"--~~I.---'i."~'--t.-~~~~-I.-'--'-!"---- 't n . . . . . . . . . . . _
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~
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800
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. . . . . . . . . O m
200 200 ^ t , 1 - " " t " " " '. " " " .* " " - ".i " " " t. " " " I. " " ' " 2 " " " ?. " " " i." ~ ~ .~ " P " " " " " " t. " " " P. " " " P. " " .- om O 4-n 100
...1....t....t....t....t....t...1....t....t...1....t....t....t....t....- 100 nm g 2 3 4 5 6 7 8 9 10 11 12 -13 14 15 (d .
N= ~ _. 0 1 Burnup (GWD/MTU) , l i l
1 FORT CALHOUN STATION TDBIl : TECHNICAL DATA BOOK PAGE 19 OF 37 . O O O 8e O n O N O
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I FORT CALHOUN STATION TDBoll ; TECHNICAL DATA BOOK PAGE 20 OF 37 f t O O 8. o 8 O O o 6 o 5 O 5 O o O 5 m o o O N - O c o 5 c O O i
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- mO '
Fi9ure ILA 3.J o 2. -
-a "1-Cycle 17 Boron . Concentration for z 'o o>
3% Shutdown Margin vs. Burnup. M OO (ARI. Zero Power - 68 Degrees F. Equilibrium Samarium) >c
-4 Z
> tn 1300 . _ . . . g . ' . ' .l ' ' ' * .I * ' ' = 1 ' ' ' .l ' ' ' ' .l ' ' ' ' .l ' ' ' ' l ' ' ' ' .l ' ' ' ' . I * ' ' .l ' ' ' ' 1 * ' = 1
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Revision 0 March .3,1997 4 Fort. Calhoun Station Operator Training JOB PERFORMANCE MEASURE
- JPM No: JPM PC JPM
Title:
Review of a proposed procedure change for approval. Approximate Time: 8 minutes Actual Time: Reference (s): 1) FC-688 .
- 2) NRC K/A 194001 GEN A1.01,RO3.3/SRO3.4 Verify _ current reference revisions match those listed above Operator's Name: SS #:
All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. t. The operator's performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluators Signature: Date: Reason, if unsatisfactory: 4
# e -- .-- -
Rovision 0 Merch 3,1997 Fort Calhoun Station Operator Training JOB PERFORMANCE MEASURE JPM No: JPM-PC JPM
Title:
Review of a proposed procedure change for approval.
~
Initiating Cue: Review the attached Temporary Procedure Change for approval as the Plant Supervisory Staff Member w/SRO License. STANDARD:
- Thin Temp. Procedure change .results in a change of intent and therefore cannot be approved as a Temp. Change.
3 1
-4
-c --
9 4 a u
Initiating . Cue: Review the attached Temporary Procedure Change for approval as the Plant Supervisory Staff Member w/SRO License. i. E
i
! CIO/!R/ CON RPT NumD3r PORT CAIJt00H STAT!ted r;*.ga g TD(PORARJ TROCEDURA CHANGR REQUEST R13 NCPTE ! If any questicos en f0*68C 70hanQe c! Ptenti are answere3 YE3. a Iemporary Chance can NOT be stagg, NOTE: 7emocrary Changes can t.:)? te moos tc E0P, AOP, AERP anc S3, Sect 10th ! - Ituttator Use Ortly FROCIDURE REVISICH VERITICATION
=ceaute a ver .' Y ~ G A A ~ M '4 Master Revision No, /d ntie: 1 f / b,e 6 b Y h-,4*p r m.4 dad, / N 95 7 Signature 7 .>#,.-v N 9
/& sv5rTime E.
' O,u f r ,-.u rn Late neseontPutpose W t r'A asv.h' -_n N 4UOO 4 k %' &
- C'/t e=*r A ' d r *' (*' V- jfN n > dik) ,M< hem.Gv A . tm,~ o / AMsr6,./.i.r%-d & . =w c.
e PA.y enast t.e returned to Docuament Control wtthin 72 hours. h u c' Ja , ? . ~_ ) ' / ' U" A ' hW Hod. No. Preterer Name Insttaae Date Extenelon Interim oppsoval sagrSlies that the Intent of the original procedure is NOT changed and that Training is not requated prior to imple.nenting the TPC,
/ / /
Plant suoe rviso rv St a f f womner ' ate Time Plant suee rvt so rv Sta f f womee r w/ SRO 1.t cense Cate Time Raview IPyl ( ) CR I 1 PRC ' a.s e t on ed P DH / pac M* sbe r Oate Assioned P rt rna rv Qua l i f i ed Peviewer teetlas !!
- Fes itC/9ealtfled 8stower Use Daly Seeties ZZZ
- Per Plc Sevtew Use caly Paccamenced Paynew By: 4 } Quorum i ) subccessattee teseevenete into operating Manualt t > Yes t > No savtow by goerin (maalaum of stal changes Made chtags affects associatea format { } Yes ( ) No Dug 1Ag P8WASW (If yes, attaen FC-164)
( ) Yes ( ) No ( ) Chases ettects an Alignment chectilst? t 1 yes ( ) No Manager = upstat&cna wate , {t! yes, attach it.es!) ( ' ' Yes ( ) No ( ) Rsqunsts 5AAC (N3RG) revtewt i 1 Yes ( ) No Manages - system sagt Gate Squarts PED-8E!.9 change 7 t ) Yes ( l No ( l ' Yes ( ) No ( ) Manager - Maintenance este quires 1000 settes sury fest reviewt ( ) Yes ( ) No ( ) Yes ( ) No() tt proceoute changed) ( ) Yes ( ) No Manager
- Rad Protectacts Late Could change tapact field wort? ( ) Yes ( i uo (
Yes ( ) No ( ) (!! yss, notify _ Manager - Gnomastry Cate ( 1 ies ( ) No ( ) Manager
- Nuc k&cens&ng ate
( _ Yes ( ) No ( ) Manages - Iranaang L..e C sco-oneeselleesy/ Crees reestiend/Othee Soview ( l Yes t i No ( , Yes ( ) No ( ) Manages - muc tro1ects Late Print Name $1gnature/Cate ( 1 Yes ( ) No() Reactor sagtneet Late ( 1 Yes ( ) No ( ) Asstatant taaat Manages case ( l ' Yes ( ) No ( ) otner: Late Seetie. ZY = hattew by 8" ^ttee Unaatmous Concurrence Dy. Subccessattee Meepers Yes ( ) No ( ) b r.ccam1Lt.ae Maesang r awer Eaccameneed let Approvalt i ) Yes ( ) No i c onis.es e ac we.ee r , e r sw.t v vua u f t eo h av i ewe r cete l toveo Syt # # Cffective Date: ' isnt vansaar:>asransicie i nart>ent beaa .ata rectne ina m ment ;chtroll ST pewtow SAPC (N3ftd) Pevgew typtng/ Revtew WP froofread By Pav. No. italt/04tet {!ntt/Gatet ilant Dates (Init/Catel ; 1a n t/ Date s Catet W/PC1
FORT CALHOUN STATION FC-68C FORM R4 CHANGE OF INTENT DETERMINATION Procedure Change No.: Procedure No.: Yes No t is a new procedure being initiated? l
- 2. Does the revision add, remove or modify acceptance critena inat affects couipment operability?
- 3. Does the revision modify the specified operational condition of a system or component?
L Does the revision reduce or p.avide less conservative ' Hold Point" cnteria?
- 5. Does the revision modify any portion of the procedure involving a known commitment? Answer 'NO' if a review is conducted and it is determined that the commitment will be satisfied?
- 6. Would the revision result in a reduction of personnel or equipment safety?
- 7. Does the revision change the original purpose of the procedure?
- 8. Does the revision violate any requirements of current Technical Specifications?
The following sections were reviewed:
- 9. Does the revision deviate from the USAR description? l The following sections were reviewed:
A Change of Intent is involved if any checklist item is checked "YES* [] Change of Intent is involved (if yes, complete, FC-154). [] Change of Intent is NOT invo!ved. COMMENTS: NOTE: Preparer must be 50.59 qualified. Completed by (Preparer): Date:
g
. Fon Calhoun Station -
Unit No.1-i OP-ST-AFW-0004 4 - SURVEILLANCE TEST l
Title:
' AUXILIARY FEEDWATER PUMP OPERABILITY TEST k I' 4
- -- FC-68 Number: 48432 Reason for Change: Reversing Steps 7.2.5 and 7.2.6 in order to relieve pressure to PI-1379 before securing pump; specifying closing and reopening of equalizing valve for DPI-1038.
-Contact Person: Laurence P. Lees Documentable Error (a) Page 8 4-
^
>y-ISSOED: 11-06-96 2:00 am R12
'w. _m___.__--_
l- FORT CALHOUN STATION OP-ST-AFW-0004 f SURVEILLANCE TEST PAGE 1 OF 9 AUXILIARY FEEDWATER PUMP OPERABILITY TEST SAFETY RELATED
- 1. PURPOSE 1.1 To satisfy, monthly, the requirements of Technical Speerfications 3.9(2) .
1.2 To satisfy, following cold shutdown and prior to raising reactor coolant temperature above 300*F, the requirements of Technical Specification 3.9(4). 1.3 To satisfy, in part, on a refueling frequency, the requirements of Technical Spectiication 3.9(5) for FW-10. This requirement, per Technical Specification Interpretation TSI-96-05, applies to all cold shutdowns that are longer than 30 days. 1.4 To satisfy, monthly, the requirements of LER-82-12.
- 2. REFERENCES 2.1 P&lD 11405-M-252, Flow Diagram - Steam (File No.10458) 2.2 PalD 11405-M-253, Flow Diagram - Steam Generator Feedwater and Blowdown (File No.10459) 2.3 P&lC 11405-M-254, Flow Diagram - Condensate (File No.10460) 2.4 Drawing 11405-EM-1039 (File No.15770) 2.5 Drawing 11405-EM-1038 (File No. 55146) 2.6 Technical Specifications, Sections 2.5 and 3.9 2.7 Technical Specification Interpretation TSI-96-05 2.8 USAR, Section 9.4, Auxiliary Feedwater System 2.9 LER 82-12 2.10 Standing Order G-23, Surveillance Test Program ic 2.11 EAR 93-030 Setpoint Changes
- 3. DEFINITIONS None R12-
I FORT CALHOUN STATION OP-ST-AFW-0004 SURVElLLANCE TEST PAGE 2 OF 9 j
- 4. ECMJPMENT LIST- ..m i L4.1 . Flashlight l
4.2 Pressure Test Gauge l 0 - 1500 PSIG, %% (0-3000 PSIG, %% gauge may be used until 0-1500 PSIG gauges become available). , 1
- 4.3 TEST EQUIPMENT REQUIRED /USED:
l NOTE: During the course of this procedure, some portable or temporary test equipment may be used to prove operability of a plant component. All such equipment must be entered into the appropriate l&C Instrument Log. CAllBRATION EQUIPMENT OPPD NO./ DUE DATE INITIALS /DATE. PI-916 PI- / '
/ /
PI-1379 PI- / / / PI-1380 / / / DPI-1038 / / /
- / / /
~
- 5. PRECAUTIONS AND LIMITATIONS 5.1 All anomalies and deficiencies shall be reported immediately to the Shift Supervisor and noted in the Remarks Section. An immediate check shall be made to verify Limiting Conditions for Operation, per Technical Specifications, have not been
-exceeded.
5.2 A Maintenance Work Request (MWR) shall be initiated to correct any reported deficiency and the MWR number shall be referenced in the Remarks Section. 5.3 No maintenance shall be conducted within this Surveillance Test, other than that specifically directed by this procedure. 5.4 When repair or routine servicing of valves or traps may affect reference pressure values, new values shall be determined or the previous values confirmed, prior to or within 96 hours of retuming the affected component to service. 5.5 Test data shall be evaluated by the Shift Technical Advisor (STA) and 'reviewed by the Shift Supervisor for acceptability within 24 hours of the completion of this test. ss R12
FORT CALHOUN STATION OP-ST-AFW-0004 SURVEILLANCE TEST PAGE 3 OF 9 5.6 The System Engineer shall be notified within 24 hours of the completion of this test of any marginal, unexpected, or unacceptable results. 5.7 A Condition Report shall be initiated, in accordance with SO-R-2, to report any anomalies or deficiencies. The Condition Report number shall be recorded in the Remarks Section. 5.8 The use of N/A (not applicable) in this procedure shall be in accordance with the requirements listed in Standing Order G-23. 5.9 The completed Surveillance Test procedure and all applicable attachments shall be signed and dated by the person (s) who actually performed the test. 5.10 Caution should be taken to prevent personal injury around hot surfaces and potential steam line breaks. 5.11 Personnel panicipating in this test have completed the Surveillance Test Verification Sheet. 5.12 Caution should bo taken in the vicinity of safety related equipment within Rooms 19 and 81, to prevent damage and/or inadvertent operation. 5.13 Temperatures above 120'F on TI-1382 or TI-1383 may indicate check valve leakage and should be investigated to minimize the possibility of steam binding the pump.
- 6. INITIAL CONDITIONS INITIALS /DATE 6.1 Procedure revision verification:
Master Revision No. / 6.2 No other test is in progress which could potentially affect this test, or if this test were performed, could have an effect on any other test. / 6.3 Main Steam is available up to YCV-1045. / 6.4 The AFW System is aligned for Normal Operation in accordance with OI-AFW-1, Auxiliary Feedwater Actuation System Normal Operation. .-
/
6.5 The Shift Supervisor authorizes performance of this test: Shift Supervisor Date/ Time / R12
-FORT CALHOUN STATION' .,
OP-ST AFW-0004 SURVEILLANCE TEST = PAGE 4 OF 9
- 7.~ PROCEDURE f% ,
' ~ ' NOTE:- Sections 7'.1 and 7.2 may be performed individually. -
- 7,,1 : . AFW Pumo FW-10 Ooerability Test
- 7.1.1e .
TI-1383 i's s 120'F - 2 F / e 7.1.2 Prior _to starting FW-10,' inspect the steam supply piping in Room 19 up to YCV-1045 for any signs ~of leakage.
- Check for visible traces of steam and for any water puddles which may result from condensed steam.- /
31 CAUTION 1 l4 - DPI-1038 should be valved-in aftet starting FW-10 to .
- avoid subjecting it to a large reverse AP due to warm-up i steam pressure.
'7.1.3 Start FW-10 per OI-AFW-4. /
7.1.4 1 With FW-10 in operation, inspect the FW-10 steam supply .... , piping in Room 19 between YCV-1045 and the FW-10 4
- steam chest for any signs of leakage. Check for visible traces of steam and for any waterpuddles which may result from condensed steam. Document leakage in Remarks. /
b 7.1.5 Valve in'DPI-1038 by opening MS-DPI-1038B and
- FW-DPI-10388, and closing MS-DPI-1038-E. / l I&C NOTE: DPI-1038 will normally be used to indicate the AP P across FW-10. . If it is found out of tolerance, the AP should be verified by Steps 7.1.8A through 7.1.8D before declaring an As Found failure.
7.1.6 - After a ten (10) minute warmup, record the following data: FIC-1369 aom (At Least 85 GPM)- DPl-1038 osid - (130 to 150) _ / v R12
-- r
' FORT CAI.HOUN STATION OP-ST-AFW-0004 1 SURVEILLANCE TEST PAGE 5 OF 9 7,1 =.7 - IF the DPI-1038 indication recorded in Step 7.1.6 is ,
140
- 10 psid (130 to 150), THEN mark all of Step 7.1.8 and 7.1.9 N/A Aud go to Step 7.1.10. OTHERWISE l
- continue below. ~/
7,1.8 i - Adjust the pump-turbine AP as follows: A. Open MS-Pl-916-B1_(PI 916 isolathn Valve.) /
; 8. - Open Valve FW-176. /
i C = Record: f . : PI-1380 osig , PI-916 osig / NOTE: If the calculated AP in Step 7.1.8.D below is less than 130 psid, an As Found failure has occurred and a Condition Report should be written. D. Calculate AP = 4 osig - osig = osid / 1 , PI-1380 PI-916 ... E. !E the AP calculated in Step 7.1.8.D is 140 i10 psid (130 to 150) M no improvement is desired, THEN mark the remainder of Step 7.1.8 N/A E go to Step 7;1.9 OTHERWISE continue below. / F. Adjust the setpoint (black knob on top of DPC-1039) until the AP Indicated by DPI-1038 is 140 1.10 , psid (130 to 150). / i
- G. Record the following:
PI-1380 osig PI-916 osig
~ DPI-1038 osid >
R12 i
- .a . . , - - , .
y i-FORT CALHOUN STATION OP-ST-AFW-0004 SURVEILLANCE TEST PAGE 6 OF 9 7,1.8 H. Calculate AP = _ , , osig- osig = osid PI-1380 PI-916 /
- l. Count and record the number of exposed threads on '
the shaft of the setpoint adjust (black knob on top of DPC-1039).
, # of exposed threads / ,
7.1.9 CLOSE the following valves: A. Close MS-PI-916-B1 / Independent infication / B. Close FW-176 / Independent Venfication ..
/
7.1.10 Record As Left AP and indicate from which step of this procedure it was taken. M As Left AP psid From Step 7.1.6 (DPI-1038) (Same as As Found) . From Step 7.1.8D (AP Prior to adjustment) From Step 7.1.8H (AP After Adjustment) / CAUTION DPI-1038 should be valveds.sut before stopping FW-10 to avoid subjecting it to a large reverst. AP due to warm-up steam. 7.1.11 Valve out DPI-1038 by closing MS-DPI-10388 and FW-DPI-10388, and opening MS-DPI-1038-E. / l l&C Independent Verification / 7.1.12 Stop FW-10 per OI-Al'W-4.
- /
R12
' FORT CALHOUN STATION OP.ST-AFW-0004 .
SURVEILLANCE TEST , PAGE 7 OF 9
- 7.2 AFW Pumo FW.6 Ooerability Test .
7.2.1 TI-1382 is s 120'F
*F /
7.2.2 Sta*f ! W4 per O!-APN4 andha-10minut:"/ 'mup. / Oper,Velv PW-176- / 7.2.3_ g 7-2:+ --Reeerd-the fc!!cdrg doi:: d if l osig
/ h 4~
PI-1379 (>1135) p
- 7. FIC-1368 apm 4' f p ' ~ /
(At Least 55 CPM) b -7.2.5 Sicp PN S pc.- O! .^PN i. / 7.2.0 O'c;; Velve PM-17&--- ' / Independent Verification / REMARKS ,_ , Completed by. Datomme /
- 8. RESTORATION 8.1 Shift Supervisor notified this test is completed.
Shift Supervisor Datomme_. /- 4 R12
y .- - . - - .. . - _- . . - . . . . - - - _ - - t
. FORT CALHOUN STATION OP-ST-AFW-0004 A SURVEILLANCE TEST PAGE 8 OF 9
- 9. 6.CCEPTANCE CRIT.EHl6 .s
~ 9.1 The As Left differential pressure of F'W-10 is 140 i 10 nsi (130 to 150 psi).-
. (Step 7.1,10) 9.2 The suctir.v1 flow (FIC-1369) of FW-10 is at least 85 gpm. (Step 7.1.6) -
- 9. 3 . The pump discharge pressure (Pl-1379) of FW-6 is at least 1135 psig. (Step 7.2.4) 9,4 The suction flow (FIC-1368) of FW-6 is at least 55 gpm. _(Step 7.2.4) ;
9.5 - The indicated temperature (TI-1383) for FW-10 is s 120*F (Step 7.1.1). i 9.6 The indicated temperature (TI-1382) for FW-6 is s 120*F (Step 7.2.1).
- 10. IEST RECORD This entire procedure.
, 11. REVIEW 11.1 The Manager-Operations is responsible for ensuring this completed surveillance is l reviewed in a timely manner and forwarded in accordance with SO-G-23.
Test data shall be evaluated by tne Shift Technical Advisor and reviewed by the Shift Supervisor for acceptability within 24 hours of the completion of this test. Evaluated by Date/ Time - /
- STA Reviewed by Date/ Time /
Shift Supervisor 11.2 The System Engineer shall be notified within 24 hours of the completion of this test of any marginal, unexpected, or unacceptable results. REMARKS A l-
- System Engineer : Date/ Time /
"'niW" R12.
. : .. . _ - .. -. _ = - - , - .
b l FORT CALHOUN STATION OP-ST-AFW-0004 h i SURVEILLANCE TEST PAGE 9 OF 9 9 Surveillance Test Signature Sheet All p:rsons participating in the performance of this test shall enter their printed name, signature, cod initials below. NAME (PRINTi SIGNATURE INITIALS N R12
- . -_ - .- . - - . - _ . ._ - .. - ~.-
FORT CALHOUN STATION OP ST AFW-0004 SURVEILLANCE TEST PAGE 8 OF 9 CAUTION
- DPI 1038 should be valved out before stopping FW 10 to avoid subjecting it to a large reverse AP due to l warm up steam. ;
l 7.1.11 Valve out DPI 1038 by closing MS DPI 1038B and FW DPl 10388, and opening MS DPl 1038 E. / l&C Independent Verification / J 7.1.12 Stop FW-10 per Ol AFW-4. / 7.2 AFW Pumo FW-6 Ooerability Test 7.2.1 T11382 is s 120 F Y / 4 7,MI. Sted "?! S per O! ^Pf! ' :nd 2!!cw a 10 Mute / We88MPe l 7.2.2 Close FW-171(FW 6 Discharge Valve)
- 7.2.3 Start FW-6 per OI AFW-4 and allow a 10 minute warmup with the discharge valve closed.
7.2.4 Open Valve FW 175. / l 7.2.5 Record the following data: PI1379 osig (>1135) F)C-1368 arn (At least 55 GPM) / 7.2.6 Stop FW-6 per OI AFW-4. / _7.2.7 Open FW 171(FW-6 Discharge Valve) / l l Independent Verification 7.2.8 Start FW-6 per OI AFW-4 / l R12.
' FORT CALHOUN STATION OP ST AFW 0004 SURVEILLANCE TEST : PAGE 9 OF 9 - l 7.2.9 Record the following data:- PI1379= osig (>1135) RC-1368 ' com /'
-(At least 55 gpm) 7.2.10 Stop FW 6 per Ol AFW 4 /
7.2.11-' Close Valve FW 175. / Independent Venfication . / REMARKS Completed by Datamme / __
- 8. RESTORATION B.1 Shift Supervisor notified this test is completed.
Shift Supervisor Datamme / R12~
FORT CALHOUN STATION OP ST AFW 0004 SURVEILLANCE TEST PAGE 10 OF 9
- 9. ACCEPTANCE CRITERIA 9.1 The As Left differential pressure of FW 10 is 140 10 psi (130 to 150 psi),
(Step 7.1.10) 9.2 The suction flow (FIC-1369) of FW 10 is at least 85 gpm. (Step 7,1.6) 9.3 Tbs pump discharge pressure (PI 1379) of FW 6 is et least 1135 psig. (Step 7.2.4) 9.4 The r,uction flow (FIC-1368) of FW 6 is at least 55 gpm. (Step 7.2.4) 9.5 The indicated temperatura (TI 1383) for FW 10 is s 120*F (Step 7.1.1). 9.6 The indicated temperature (TI 1382) for FW 6 is s 120 F (Step 7.2.1). l 9.7 The indicated recire flow (FIC 1368) should not change between steps 7.2.5 and 7.2.9
- 10. TEST RECORD This entire procedure.
11, REVIEW 11.1 The Manager-Operations is responsible for ensuring this completed surveillance is ieviewed in a timely manner and forwarded in accordance with SO-G 23. Test data shall be evaluated by the Shift Technical Advisor and reviewed by the Shift
- Supervisor for acceptability within 24 hours of the completion of this test.
Evaluated by Datemme / STA Reviewed by Datsmme / Shift Supervisor 11.2 The System Engineer shall be notified within 24 hours of the completion of this test of any marginal, unexpected, or unacceptable results. REMARKS System Engineer Datemme / R12
FORT CALHOUN STATION S0-G-30 STANDING ORDER PAGE 17 OF 41 5.4.15 E. 'f the PAP will change the procedure number, a search shall be made to identify all other procedures that reference the original procedure number. (1) Submit a PAP for procedures that reference the original procedurc number revising the reference to the new number. F. Include original plus one copy of the FC-68 and changes in the PAP. Additional copy is not required for one-time uso procedures. G. Inform Document Contiol if regular / blanket revisions of procedure changes need to be issued concurrently, if applicable. 5.4.16 PAP packages submitted to Document Control will be processed as fo: lows: A. Log and track as appropriate. B. Complete the review assignment by consulting FCS Procedure Review and Approval Authority Assignments, Attachment 1. C. The original PAP shall be fonvarded to the Responsible Department Head for Qualified Reviewer assignment or PRC Member for review and approval according to SO-G#5 or S0-G-5, as applicable. The PAP will be retumed to Document Coarol when the asc!9nment/ review has been made. D. Route the original PAP to the appropriate reviewers and update the tracking system as necessary to reflect the current review status of the PAP. 5.5 Temporary Procedure Change (TPC) 5.5.1 Temporary changes shall NOT be implemented for Standing Orders, Abnormal ] Operating Procedures, Emergency Operating Procedures, or Radiological l Emergency Response Plan. 5.5.2 The TPC becomes void if Section I, FC-68B including Interim Approvals is not completed and retumed to Document Control within 72 hours of initial checkout. Process through Step 5.5.10 of this procedure must be completed. 5,5.3 Revision package submitted as a TPC will have expedited processing to ensure approval within the 14 day requirement specified in FCS Technical Specification 5.8.3. 5.5.4 Obtain a TPC PAP from Document Control. R70
l FORT CALHOUN STATION SO-G-30 STANDING ORDER I' AGE 18 OF 41 [5.5.5) Initial, date and incorporate the changes in the body of the procedure at j appropriate locations. i 5.5.6 Ensure a Change of Intent Determination (FC480) is completed, 5.5.7 Ensure that Section I, FC4BB is completed incluaing TPC approval by two . members of the Pf art Supervisory Staff, at least or e of whom holos an SRO ) License, j 5.5.8 Approval of Section I, FC488 signdies that: A. The intent of the original document is not changed. Any question on the FC48C, Change of Intent Determination, answered yes, prohibits a TPC. B. Training is not required prior to implementing the 'i PC. 5.5.0 Following TPC approval, the oreparer shall proceed as follows: A. Changes to Procedures (1) Attach a copy of the TPC Form and one copy of all attachments to the working copy of the procedura. B. Changes to the Technical Data Book (1) Attach a copy of the TPC Form and a copy of all attachments to the appropriate TDB figure (s) in the Control Room copy. The TDB Section is then considered to be the working copy. (2) If the change affects both the TDB and a procedure, then attach a copy of the TPC form and a copy of all attachments to the appropriate TDB figure (s) and the working copy of the procedure. 5.5.10 Upon completion of activities required above, submit original plus one copy of the entire PAP to Document Control. Additional copy is not required for one-time use procedures. 5.5.11 Document Control shall complete review assignment by consulting FCS Procedure Review and Approval Authority Assignments, Attechment 1. R70
FORT cat.HOUN STATION SO-G-30 STANDING ORDER PAGE 19 OF 41 5.5.12 If any changes are mado during the review process, the changes will be marked in red, initialed and dated. A. A Change of Intent Determination (FC-68C) will be completed for the new changes. B. If a change of intent is identified, the new changes can not be made under the TPC and a regular procedure change will have to be cubmitted cnce the TPC is issued. 5.5.13 At the completion of the 72 hour requilament, if the TPC has not been retumed to Document Control for review, Document Control shall issue a Condition Report (CR) as applicable. 5.5.14 in order to meet the 14 day review and approval time frame specified in the Technical Specification, upon receipt Document Control shall: A. Verify review assignment by consulting FCS Procedure Review and Approval Authority Assignments, Attachment 1. B. Enter the TPC PAP information into the tracking system and start 14 day completion tracking. C. Dispense PAP according to SO-G-95. Notify the Responsible Department Head of the 14 day review requirements. 5.5.15 At the completion of the 14 day requirement, if the TPC has not undergone Qualified Review and been approved by the Responsible Department Head, Document Control shall issue a Condition Report as applicable. 5.6 Procedure Correction Correction of typographical / documentable errors will not change the current procedure revision number. However, the page containing the error will be re-issued with a thick line in the right hand margin. The procedure coversheet will contain the tyRographical/ documentable error, alpha character (for Document Control use) and pages to be issued. 5.6.1 Typographical Error A. To correct a typographical error, the preparer sha!!:
- Complete an FC-68D, Procedure Correction form o Answer all questions as applicable
- Attach marked up pages from the original procedure change with the typos marked in red
- Include procedure cover page R70
FORT cal.HOUN STATION S0-G-30 STANDING ORDER PAGE 20 OF 41 5.6.1 D. Typographical error correccions will then be approved by Supervisor Document Control, or designee or P: ant Supervisory Staff Member. C. Submit Procedure Correction Form to Document Control. D. Document Control may re-issue a typographical error at their discretion unless otherwise specified by the preparer on FC-68D. E. After the correction has been made, route the change to the preparer, or designee, for typing review. 5.6.2 Documentable Error A. To correct a documentable error, the preparer shall:
- Complete an FC-68D, Procedure Correction form
- Answer all questions as applicable e include procedure cover page
- Attach marked up pages from the original procedure with the documentable errors marked in red
- Attach all documentation needed to support the correction
- Obtain Plant Supervisory Staff Member Signature
- Submit Procedure Correction Fomi to Document Control B. After the correction has been made, route the change to the preparer, or designee, for typing review.
5.7 Deleting or Superseding Procedures 5.7.1 Deletion of procedure changes in progress: A, For a PAP that has not entered the review process, the preparer shall draw a line diagonally across the entire page, write ' VOID", initial, date and return to Document Control. B. Deletion of a procedure / procedure change in the review process requires the Department Head, PRC Member or Qualified Reviewer (as appropriate) to draw a line diagonally across the FC-68 Form, write
'VOlD", initial, date and retum to Document Control.
m R70
~ Revision: 0
- October 23.1995 - +
Fort Calhoun Station . Operations Wir:ing
' JOB Pi:RFORMANCE MEASURE -
JPM Ns: JPM . RCA RCA Entry and Esit~ i- ' i L Topic ts) : RCA Entry and Exit :
- 1,ocation(s) RCA Access Control-
' Approximate Time: 10 rnhutes Actual The 1_ _
Reference ash' (1) . GET Radiation Worker Training. (2) - NRC K/A 194001. K1.03 (RO2.8/SRO3A) . 0): NRC K/A 194001. K1.04 - (RO3.3!SRO3.5) , Verify current reference revisions match those listed above
~
l Open. tor's Name: _ SS # : All C4itical Steps ( * ) must be performed or simulated in ac,ordance with the standards contained in this
- . jpM.~
4 I - The opsrator's performance was evaluated as: SATISFACi'ORY UNSATISFACTORY Evaluator's Signature : Date : Reason, if unsatisfactory : Operator's review : Date: 1 Tools A Equipment : None Safety Considentions: Ihis JPM requires entry into the RCA. l Comments : i h" he._. + , .er wo- we---..-v -, ,*-rum tvt-'ze t -- ,-----rw---- -t-
, n t--- r-- --we-
P
. Revision: 0 October 25,1995 i Fon Calhoun Station - Operations training JOB PERFORMANCE MEASURE ,
JPM No: JPM RCA
- RCA Entrij and Esit 1
Irdtladng Cue: An entry into the RCA is required as part of the operating exam. . START CRITICAL STEP * ' ELEMENT STANDARD-
- Review RWP. Read RWP Check survey maps Check survey maps for radiological conditions in arecs to be entered
- Obtam dosimetry Venfy TLD attached to secunty badge. Obtain ALNOR
- Slgn m on appropnate RWP lasert ALNOR into reader, enter PID and RWP number.
Inform RP Personnel about the nature of Tell RP at access control w here you are going your entry and what you will be doing. Ecter RCA RCA entered
- Comply with all postmg within RCA No violation of posted requirements
~
- Momtor f or personnel contanunauon pnor - Momtor for contanunauon usmg PCM 1 or
. to exiting frisker.
- Sign out of RCA Insert ALNOR in reader. Enter PID number,
; conarm dose and place ALNOR la charpng rack.
l 4 s
. , . - . . , . - , , . + . - . . .
Fort Calhoun Station Unit No.1 l SO-G-101 STANDING ORDER
Title:
RADIATION WORKER PRACTICES FC-6FA Number: 48435 Reason for Change: To provide clearer guidance to only workers who have been issued a TLD who are undergoing or have had nuclear medicine treatments. Contact Person: Chuck Anderson Documentable Error (a): Page 4 ISSUED: 11-08-96 2:00 am RB
FORT CALHOUN STATION SO-G-101 STANDING ORDER PAGE 1 OF 13 RADIATION WORKER PRACTICES NON. SAFETY RELATED [1.] PURPOSE AND SCOPE 1.1 Purpose This procedure provides guidelines to be followed while in a Radiologically Controlled Area (RCA), for the prevention of contamination spread and reduction of personnel dose. 1.2 Scope The guidance provided in this procedure is to be followed, howtsver, it is not all inclusive. Direction provided on Radiation Work Permits (RWP) or by Radiation Protection personnel is also to be followed, and shall take precedent over this procedure. '
- 2. STATEMENT OF APPLICABILITY This procedure applies to all personnel who enter a Radiologically Controlled Area (RCA).
- 3. DEFINITIONS 3.1 Radiation - Energy in the form of particles (alpha, beta, neutron) cr electromagnetic waves (gamma).
3.2 - Contaminated Area - An area where the loose surface activity is 11000 dpm/100 cm2beta-gamma or 220 dpm/100 cm2 long lived alpha. 3.3 Radioactive Material - Items which are made radioactive by exposure to contamination or to neutron radiation. 3.4 Doss - The quantity of radiation absorbed by the tissue of concern. 3.5 REM - A unit of radiation dose defined as "that amount of any type radiation which will cause damage equivalent to the deposition of 100 ergs of gamma radiation in 1 gram of tissue." 3.6 Posting - A sign or label bearing a magenta or black radiation symbol on a yellow background, the word " Caution,"" Danger," or " Grave Danger," and supplemental information identifying the type or quantity of radiation / radioactive materials and the protective measures required to minimize personnel exposures. R8
FORT CALHOUN STATION S0-G-101 STANDING ORDER PAGE 2 OF 13 3.7 Boundary - an established line which defines an area where personnel may be exposed to radiation or radioactive material. 3.8 Radiologically Controlled Area - an area controlled for the purpose of protecting personnel from radiation or radioactive materials. 3.9 Radiation Work Permit - A permit issued by the Radiation Protection Department to control work evolutions which involve personnel exposures to radiation or radioactive materials. 3.10 ALARA (As Low As Reasonably Achievable)- Making every reasonable effort to - maintain exposures to radiation as low as possible. 3.11 TEDE (Total Effective Dose Equivalent)- The sum of the deep dose equivalent (for extemal exposures) and the committed effective dose equivalent (for internal exposures). This represents the combined dose to a worker. 3.12 TODE (Total Organ Dose Equivalent)- The sum of the deep dose equivalent (for external exposures) and the committed dose equivalent (for internal exposures). + This represents the combined dose to a worker's organ. 3.13 SDE (Shallow Dose Equivalent)- The external exposure of the skin at a depth of
.007 cm.
3.14 LDE (Lens Dose Equivalent) - The external exposure to the lens of the eye at a depth of 0.3 cm. 3.15 Low Background Area - An area where the background count rate is less than 300 cpm.
- 4. RESPONSIBILITIES 4.1 Work Supervisors are responsible for.
4.1.1 The conduct of workers in the RCA. 4.1.2 Requesting Radiation Work Permits (RWP) and providing adequate description of the work to be performed. 4.1.3 Ensuring worker compliance with this procedure and other guidance provided by Radiation Protection Personnel. 4.1.4 Notifying dosimetry personnel of special dosimetry needs per the RWP at least one shift before needed. R8
FORT CALHOUN STATION S0-G-101 STANDING ORDER PAGE 3 OF 13 4.2 Persons entering the RCA are responsible for: 4.2.1 Compliance with this procedure and other guidance provided by Radiation Protection personnel. 4.2.2 Adhering to the requirements listed on the RWP they are using. 4.2.3 Reporting to Radiation Protection personnel radiological problems or questions.
- 5. PROCEDURE 5.1 As Low As Reasonably Achievable (ALARA) 5.1.1 Work Supervisors or Crew Leaders should:
A. Ensure that workers under their direction have been adequately instructed in proper radiological work practices to perform their duties. B. Schedule and assign work so exposures are individually and collectively maintained ALARA. C. Enforce proper work practices and adherence to RWP requirements. D. Remain knowledgeable of the radiological conditions of work areas where work under their supervision is being performed. E. Remain knowledgeable of their department exposure goals and the exposure status of their assigned personnel. F. Notify Radiation Protection personnel of radinlogical problems encountered by workers. G. Making recommendations for exposure reduction by submitting an ALARA Suggestion form. 5.1.2 Personnel working in a RCA shall: A. . Remain knowledgeable of.their. exposure and margin. B. Maintain their exposure ALARA by using proper work practices and adhering to the requirements of their RWP. C. Notify Radiation Protection of items they recognize that could reduce personnel exposure. ALARA Suggestion forms are available. R8
t i FORT CALHOUN STATION S0 G 109 STANDING ORDER PAGE 4 OF 13 5.1.2 Remain knowledgeable of work area dose rates. D. 5.2 Fort Calhoun Station (FCS) Exposure Limits ; 5.2.1 The following are the administrative exposure limits for FCS: A. Total Effective Dose Equivalent (TEDE) 4.0 rem / year from all f
- occupational sources. ;
4 B. Total Effective Dose Equivalent (TEDE) 1.0 rem / year from l occupational sources at FCS. : i
- C. Total Effective Dose Equivalent (TEDE) 0.3 rem / pregnancy for !
! Declared Pregnant Women. ; D. Total Effective Dose Equivalent (TEDE) 0.25 rem /yssr for Temporary ! Radiation Workers, unless the Manager Radiation Protection l l authorizes a higher limit, not to exceed 0.5 rem / year. i 5.2.2 Regulatory Limits l l 1 A. Total Effective Dose Equivalent (TEDE) 5.0 rern/ year. ! , B. Total Organ Dose Equivalent (TODE) 50,0 rem /v' to any individual i organ, other than the lens.of the eye. 4 C. Shallow Dose Equivalent to the skin or extremities (SDEw. or SDE ) 50.0 rem / year, D. Eye Dose Equivalent (LDE) 15 remlysar.
- 5.3 Personnel Monitoring 5'.3.1 Individuals who have been issued a TLD shall notify Radiation Protection if t they are undergoing nuclear medicina treatment or have been
=
administered radioactive materials as part of a medical procedure. 5.3.2 Personnel who enter Radiologically Controlled Areas shall: A. - Know their dose limits and current exposure. i B. Report for an exit whole body cour.t and complete a Termination Data Sheet when they terminate or no longer require dosimetry. w/ R8 .
FORT cat.HOUN STATION S0-G 101 i STANDING ORDER PAGE 5 OF 13 , 5.3.2 C. Notify dosimetry personnel before receiving occupational exposure : from another licensee while being monitored at Fort Calhoun Station. , D. Ensure that initial, annual and termination whole body counts are done as required. E. Provide bioassay samples as requested by the Manager-Radiation Protection. l F. Be mor,itored for rixternal radiation exposure by use of a ! thermoluminescent dosimeter (TLD) and a direct rsading or electronic , dosimeter. G. The TLD and other dosimeter will be worn on the trunk of the body, between the waist and neck in close proximity (i.e. within 8 inches) to each other. When wearing protective clothing, the ALNOR will be , worn on the outside of the clothing. H. If any of the dosimeters prescribed by the RWP are lost or off scale, the individual will immediately exit the RCA and notify the Shift Radiation Protection Technician.
- l. Women who are, or .may be pregnant and desire to declare their .
pregnancy or anticipated pregnancy are responsible for notifying the Manager Radiation Protection in writing by completing the applicable forms in procedure RP-602. 5.4 Radiation Work Permit (RWP) Generation and Use 5A1 RWPs are required when: A. Entering a RCA. B. Entering any other area which is posted as " Radiation Work Permit 4 Required." C. Performing radiography. D. Moving radioactive material between RCAs. E. Preparing a radioactive shipment. i 5.4.2 RWPs may be requested by any work group needing an RWP to perform a ; task. The "RWP Request Worksheet " /C-RP AD-200-2, is available. l R8 4
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l t FORT CALHOUN STATION S0-G-101 ! STANDING ORDER PAGE 6 OF 13 l 1 5.4.3 RWP requests should bo prepared in accordance with the following lead j time allowances: : i A. One working day for jobs < 1 person rem. B. One week for jobs t i person-rem. l C. Exceptions to the above RWP preparation schedule may be made as ! e follows: l (1) in the form of an emergency waiver as determined by the Shift t Supervisor Operativns. (2) By the Supervisor-Radiological Operations. 5.4.4 A delay in the start of a job will be reported by the work supervisor to the ALARA group. A representative from the ALARA group will evaluate any 4 changes in radiological conditions prior to revising the start and/or expiration dates of an RWP. . S.4.5 If the work scope changes from the job description originally indicated on the RWP, the job supervisor is responsible for notifying the ^ Supervisor-Radiological Operations or ALARA Group. 5.4.6 Personnel signed in on an RWP shall: A. Read the RWP dally when used. , B. Adhere to the requirements and instructions listed on the RWP. C. Review applicable surveys and understand the radiological conditions , in the work area. D. Contact a member of the Radiation Protection Department to clarify any information which is not understood. E. Comply with radiological work practices as established by the i Radiation Protection Department and subsequently presented in the
.- General Employee Training Program.
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FORT CALHOUN STATION S0-G 101 STANDING ORDER PAGE 7 OF 13 5.5 Access Control 5.5.1 All personnel should log in and out of the access control system for each RCA entry, except as follows: A. Radiation Protection, Operations, Chemistry and Security personnel, working on the appropriate " General RWP" may log in and out once par shift. These personnel are responsible for properly signing out at access control at the end of their shift. 5.5.2 Persons wishing to enter the RCA shall complete the following: NOTE: Steps 5.S.2A through 5.5.2F may be performed in any order. A. Obtain an understanding of tho information provided and the requirements of the appropriate RWP. B. Become familiar with radiological conditions in the work area. Methods for accomplishing this include; reviewing radiological survey maps, attending ALARA or RP j;.b briefings or discussions with RP personnel at access control, C. Obtain electronic dosimeter from rack. D. Proceed to the ALNOR Reader: I (1) Insert ALNOR into reader. (2) Enter PID number, via keypad, then press OK; or enter PlO number using scanner, then press OK. (3) Confirm your name and your PID number and press OK NOTE: By entering RWP number you signify that you have read and understand the requirements of the RWP. (4) Enter the last four digits of the RWP, via keypad, then press OK, or enter RWP using scanner, then press OK (5) Confirm your available dose and alarm setpoints and press OK. (6) Remove ALNOR when prompted by display. R8
i 4
; FORT CALHOUN STATION S0-G 101 -
i STANDING ORDER PAGE 8 OF 13 i i 5.5.2 E. Confirm that the electronic dosimeter is on and reading zero. F. Speak with RP personnel at access control. Provide RP personne; 4 with the following information: (1) Where you are going to be working. : (2) What you are going to be doing. ' 4
! (3) What methods you will use.
M) Which RWP you will be using, , t G. Con'irm that you meet all the requirements to enter the RCA then j proceed to the RCA entrance. 5.5.3 Upon exiting the RCA: A. Monitor yourself for contamination. B. Proceed to the dLNOR Reader: (1) Insert ALNOR into reader. (2) Enter PID number, via keypad, then press OK; or bater PID l number using scanner, then press OK. L (3) Confirm your name and your PID and press OK (4) Confirm your dose and press OK l (5) Remove ALNOR when prompted by display. C. Place electronic dosimeter in the charging rack. , b
^
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FORT CALHOUN STATION SO-G 101 STANDING ORDER PAGE 9 OF 13 5.6 , Workor Practices While !n The RCA. 5.6.1 General Rules of Conduct. A. Eating, drinking, smoking and chewing are prohibited in the RCA. B. Promptly obey Stop Work or evacu6 tion instructions from radiation protection personnol. C. If an Area Radiation Monitor or Continuous Air Monitor alarms, immediately leave the brea and notify the Control Room. D. Report all wounds or injuries received in the RCA to the Control Room and the Shift RP Technician. E. Become familiar with the area to be entered prior to entry. Methods to accomplish this include reviewing posted survey maps, RWP information/ instructions and speaking with RP personnel. F. Radiological postings, boundaries, or barricades must not be moved or degraded. G. Areas in the RCA which are greater than six feet off the floor are not routinely surveyed. Contact RP prior to entering these areas. H. Personnel shall log in and out on Airborne Radioactivity Area Entry Logs upon entering and exiting Altborne Radioactivity Areas. 1
- 1. Permission shall be obtelned from a Radiation Protection Technician prior to moving radioactive materials beyond the boundaries of an RCA.
J. Handle dosimetry devices with care. Throwing, dropping or other abuse may result in damage to the instruments. 5.6.2 Contamination Control Work Practices. ; A. Abide by all contamination control postings, boundaries and barriers, i B. Do not use fans, blowers, or compressed air in contaminated or I potentially contaminated areas, without approval of radiation protection personnel. C. When working in contaminated areas, do not touch personal items (e.g. safety glasses) with contaminated gloves. R8 I _.-.\
f FORT CALHOUN STATION S0-G-101 ! STANDING ORDER PAGE 10 OF 13 , 5.6.2 D. Reaching / working across Contaminated Area boundaries is permitted only when specifically authorized by a RWP governing the task (s) . performance. [
! t E. Prior to leaving a contaminated work area, bag all tools, equipment i and trash. Contact RP for a survey of the material, prior to removing r it from the area.
4 i F. Hoses and cords that must cross a contamination boundary should be secured such that they do not move in and out of the contaminated 1 area. ! 4 G. wunimize as much as possible the amount o' consumable items taken into the RCA. H. Remove protective clothing slowly and in the order of highest potential contamination first.
- l. Place protective clothing gently into the proper container. ,
J. Do not compress contents of containers, while performing Step 5.6.21. K. Contact RP personnel as containers approach full, f' 5.6.3 Radiation Exposure Work Practices A. Minimize your exposure to ionizing ra'dlation by conducting yourself in - accordance with radiological worker practices and A1 ARA principles when in an RCA. B. Check dosimeters frequently when in the RCA, except when RP technicians are providing time keeping for exposure control. C. Promptly report to RP personnel any unsafe radiological conditions or . practices. - D. If a method of exposure reduction becomes apparent, notify RP and submit an ALARA Suggestion Form. t i- _ R8 t
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FORT CAlllOUN STATION S0-G-101 STANDING ORDER PAGE 11 OF 13 5.7 Contamination Monitoring 5.7.1 Personnel leaving a contaminated area will: A. Perform a minimum of a hand and foot frisk at the nearest frisker location. This activity should be performed in a low background area. B. Take measures to prevent the spread of contamination. C. Notify RP personnelif any contamination above background is indicated. D. Not attempt self-decontamination until RP has evaluated the contamination. 5,7.2 Personnel leaving an RCA with a PCM present will: A. Enter the PCM and initiate the count process. B. At the end of the count, exit the PCM and leave the area if no alarm has occurred. C. If an alarm occurs, recount. D. If the second count does not result in an alarm, the initial count may
, be considered a false alarm and you may leave.
~
E. If two consecutive alarms are indicated, the individual is presumed to-be contaminated, and the following actions will be taken: (1) Take measures to prevent the spread of contamination. (2) Immediately notify RP personnel. l (3) Make no attempt to perform self-decontamination until RP has evaluated the contamination. F. Small personal items (i.e., pens, keys, wallet, etc.) carried into an RCA in an individuals pocket, may be monitored with the individual in the RCA exit contamination monitor and released in the same manner as personnel. G, All other items (i.e., tools, equipment, parts, logs, notebooks, etc.) or any item picked up while in the RCA, must be surveyed and released by qualified RP personnel. R8
. i l
FORT CALHOUN STATION S0 G 101 STANDING ORDER PAGE 12 OF 13 ; 5.7.3 Personnelleaving an RCA without a PCM present will:
^
A. Perform a whole body frisk using a portable frisker. This activity f j should be performed in a low background area. If any contamination above background is indicated: i B. (1) Take measures to prevent the spread of contamination. l (2) Immediately notity RP personnel, i !I - (3) Make no attempt to perform self-decontamination until RP has evaluated the contamination. i C. If no contamination is indicated proceed to the nearest PCM and perform monitoring per 5.7.2. 5.7.4 Control of Routine Samples / Sample Containers A. Routine samples.(i.e., RP and Chemistry samples) suspected of being radioactive are required to be labeled as radioactive material unless any one of the following conditions exist. (1) The sample container is attached to or contained within equipment dedicated to the collection or analysis of samples. (2) The sample / sample container is continuously attended by a person qualified in the handling of radioactive material. (3) The sample is stored in a location designated for the storage of radioactive rnatorials. (4) The sample / sample container has been analyzed and . determined not to contain activation or fission product nuclides. B. Samples / samples containers required to be labeled as per Section 5.7.4.A shall be labeled 'CAUT10N RADIOACTIVE MATERIAL.' C. A RP Technician shall escort samples with dose rates expected or .' determined to exceed 100 mrem / hour at 30 centimeters from the external surface of the sample. D. These samples (described in 5.7.4C) shall be stored in an area designated as a High Radiation Area. v 1 R8 E a gm.g>.----,ver ,,-rwur , e, - - - . r-w,.~cw-s~w--,-.,,----,we ev -rr ,- .v.,, . --wo-,--.%-.e.-,re- , .,, -. , e e, -ww-,%.-2.-.-. c<r.----- ,, - - . - . . , -
FORT CALHOUN STATION S0 G 101 STANDING ORDER PAGE 13 OF 13
- 6. REFERENCES / COMMITMENT DOCUMENTS 6.1 RP AD 200," Radiation Protection Surveillance program Administrative Procedures" 6.2 RP 207, " Personnel Monitonng and Decontamination" 6.3 RP AD-300, "ALARA Program" 6.4 RP-AD#a00, " Dosimetry Program" 6.5 INPO Guideline 91-014, " Guidelines for Radiological Protection at Nuclear Power Stations."
6.6 Title 10, Code of Federal Regulations, Part 20, " Standards for Protection Against Radiation" 6.7 Commitment Documents
- CID 931132/01, LIC-93 0283, Step 1
- 7. ATTACHMENTS None R8
. .- .. _- . _ . . .. . - _ . - _ - -=- .
l Revision 0 j February 10,1997 1
]
Fort Calhoun Station. Operator Training JOB PERFORMANCE MEASURE , JPM No: JPM EPIP 4 JPM
Title:
- Ernergency Plan Classification Approximate Time: 5 minutes Actual Time: Reference (s): (1) EPIP OSC 1 (rev 26) (2) NRC K/A: 194001.A1.1.6, RO3.1/SR04.4 Verify current reference revisions match those listed above Operator's Name: SS #: All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. i The operators performance was evaluated as: t SATISFACTORY UNSATISFACTORY Evaluators Signature: Date: Reason, if unsatisfactory:
Rovision 0 Februery 10,1997 Fort Calhoun Station . Operator Training JOB PERFORMANCE MEASURE , JPM No: JPM EPIP
- JPM
Title:
Emergency Plan Classification Inillating Cue: While repairing leaking fuel bundles during a refueling outage, a fuel pin was broken in half while being removad from a bundle. All fuel movement A. repairs were immediately stopped. The following indications were available in the Control Room:
- RM/062 in High alarm
- CRHS/ VIAS actuated
- RM/081 25 mr/hr
- RM/082 35 mr/hr
- RM/087 250 mr/hr
- RM/088 150 mr/hr Classify the event.
- Answer => Alert per EAL 7,1 t
V
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i i Initiating Cue: l < While repairing leaking fuel bundles during a refueling outage, a fuel , pin was broken in hall while being removed from a bundle. ; All fuel inovement & repairs were immediately stopped. l The following Indications were available in the Control Room: : 4
- RM/062 in High alarm
- CRHS/ VIAS actuated
- R M/081 25 mr/hr
- RM/082 35 mr/hr
- RM/087 250 mr/hr
- RM/088 150 mr/hr Classify the event. l l
4 i I 1 I f L i i
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- Fort C:lhoun St: tion !
Unit No.1 l r i i l EPIP OSC 1 l EMERGENCY PLAN IMPLEMENTING PROCEDURE i + ,
Title:
EMERGENCY CLASSlFICATION I i , IN ACCORDANCE WITH 10 CFR 50.54 (q), THIS REVISION j DOES NOT REDUCE THE EFFECTIVENESS OF THE FCS RERP. ; i REVIEWED per EPDM 6: i . .nunk.(M %A3 . L Manager . gerQPlanning (signature r d prior to distribution) FC 68 Number: 47302 i Reason for Change: Delete turbine failure from the title of EAL 11.11. Revise EALs 8.7 and 8.9 to clarify different dose and dose rate levels resulting in a 'SAE" or 'GE". Revise EAL 1.13 to remove 20*F subcooled curve criteria. Match EAL { titles / attachment 6.2 titles for all EALs. Corrected typo in EAL 1.2. Contact Person: Mike Christensen i _ ISSUED:. 0718 95 __4:00 pm . R26 - T
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EPIP OSC 1 l FORT CALHOUN STATION PAGE 1 OF 97 EMERGENCY PLAN IMPLEMENTING PROCEDilRE EMERGENCY CLASSIFICATION l 1.0 2.UBEQSE 1.1 This procedure establishes crrteria for classification of abnormal events into one of the four standard emergency classifications. These classifications are consistent with guidance found in NUREG-0654/ FEMA REP 1, Rev.1. l l 2.0 PREREQUISITES 2.1 There are no specific prerequisites for this procedure. Any abnormal or offnor.nal event is cause for referring to this Emergency Plan implementing Procedure.
3.0 REFERENCES
/ COMMITMENT DOCUMENTS
- 3. '. Radiological Emergency Response Plan 3.2 Emergency Plan implementing Procedures 3.3 NUREG-0654/ FEMA REP 1, Rev.1 3.4 10CFR50 3.5 Commitment Documents IMPLEMENTING COMMITMENT SOURCE
_ STEP fiUMDER (CID) DOCUMENT EAL 3.2 883055 UC 88-0165 EAL 3.3 883055 UC-88-0165 EAL 3.4 883055 UC-88-0165 EAL 11.5 883055 UC-88 0165 Attachment 8.2 883055 UC-88 0165 R26 FC/EPIP
FORT CALHOlJN STATION EPIP OSC 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE2 OF97 4.0 DEFINITIONS 4.1 - EMERGENCY ACTION LEVEL (EAL) Alarms, instrument readings or visual sightings that have exceeded predetermined limds which would categorize the situation into an g initiating condition of one of the four emergency classifications.
-4.2 EMERGENCY CLASSIFICATION One of the following classifications:
4.2.1 NOTIFICATION OF UNUSUAL EVENT (NOUE) Unusual events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs, furoose: (1) to assure that the first step in any response later found to be necessary has been carried out, (2) bring the operating staff to a state of readiness and (3) provide systematic handling of unusual events information and decision making, ' 4 2.2 ALERT Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Purpose:
(1) to assure thet emergency personnel are readily available to respond if the situatbn becomes more serious or to perform confirmatory radiation monitoring if required and (2) provide offsite authorities current status information. 4.2.3 SITE AREA EMERGENCY Events are in progress or have occurred which involve actual or likely major failures of the plant functions needed for protection of the public. Any releases are not expected to exceed EPA Emergency Action Guideline exposure levels exuept near the site boundary. Puroose: (1) to assure that response centers are manned, (2) assure that monitoring teams are dispatched, (3) assurs that offsite personnel required for evacuation of near site areas are at duty stations if situation becomes more serious, (4) provide consultation with offsite authorities and (5) provide updates for the public through offsite authoritles.
..V FC/EPIP R26 1
. -- ._-_ - . . _ . - _ _ _ - __ _ =_-
FORT CALHOUN STATION EPIP OSC 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 3 OF 97 4.2.4 GENERAL EMERGENCY livents are in progress or have occurred which involve actual or imminent substantial core degradation or molting with the potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure lovels offsite. fucqQM: (1) to initiate predetermined protective actions for the public, (2) provide continuous assessments of information from licensee and offsite organization measurements, (3) initiate additiond measures as Indicated by actual or potential releases, (4) provide consultation with offsite authorities and (5) provide updates for the public through offsite authorities. 4.3 EMERGENCY COMMAND AND CONTROL - Overall direction of licensee response which must include the non delegable responsibilities for the decision to notify and recommend protective actions to the state and counties and other authorities responsibl6 for offsite emergency measures. The direction of licensee operations to mitigate accident consequences remains with a qualified command and control position. 4.4 ENGINEERED SAFETY FEATURES (ESF) The basic features of engineered safety systems, intended to mitigate the consequences of design basis accidents and beyond design basis LOCA. - 4.5 EXCLUSION AREA The area surrounding the nuclear power plant in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from that area. The term is synonymous with "onsite'. 4.6 FAILED FISSION PRODUCT BARRIER Ble fission product barrier is incapable of sufficiently retaining radioactive materials to protect the public. 4.7 FISSION PRODUCT BARRIER The fuel cladding, reactor coolant system boundary, or the containment building. 4.8 INTACT The fission product barrier retains the ability to protect the public from a harmful release of radioactive materials. 4.9 MODES OF OPERATION One of the following classified plant conditions: 4.9.1 POWER OPERATION CONDITION (MODE 1) The reactor is in the power operation condition when it is entical and the neutron flux power range instrumentation indicates greater than 2% of rated power. R26 FC/EPIP
l FORT CALHOUN STATION EPIP OSC 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 4 OF 97 1 4.9.2 HOT STANDBY CONDITION (MODE 2) The reactor is considered to be in a , hot standby conddion if the average temperature of the reactor cooiant (Tave) is greater than 515'F, the reactor is critical, and neutron flux power range , instrumentation indicates less then 2% of rated power. 4.9.3 HOT SHUTDOWN CONDITION (MODE 3) The reactor is in a hot shutdown , condition if the overage temperature of the reactor coolant (Tave) is greater then 515'F and the reactor is subcritical by at least the amount defined in j Technical Specification paragraph 2.10.2. 4.9.4 COLD SHUTDOWN CONDITION (MODE 4) The reactor co >lant temperature (T cold) is less than 210'F and the reactor coolant is at shutdown boron ! concentration. 4.9.5 REFUELING SHUTDOWN CONDITION (MODE 5) The reactor coolant is at e refueling boron concentration and reactor coolant temperature (Tcold) is less then 210'F. 4.10 OFFSITE Those areas not within the exclusion area boundary. 4.11 ONSITE The area surrounding the nucioar power plant in wh!ch the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from that area. The term is synonymous with
- Exclusion Area Boundary'.
4.12 VERIFICATION CRITERIA The plant or site condition by wh!ch the decision may be based for classifying the emergency. v FC/EPIP R26
l EPIP OSC 1 FORT CALHOUN STATION PAGE 5 OF 97 EMERGENCY PLAN IMPLEMENTING PROCEDURE
%0 PBDGEDDBE I
NOTE , Th3 highest emergency classification for which an Emergency Action Levelis currently met should l be declared. !! an action level for a higher classification was exceeded but has since abated or otherwise been resolved prior to offsite reporting, REPORT the higher classification to the states (& counties if necessary) and the Nuclear Regulatory Commission, but do not DECLARE the classification. The notification must indicate the current classification, the period (s) of time that th3 higher classification existed, and the mitigating condrtions that caused the emergancy ci ssification. An explanation should be given in the remarks section of the FC 1188. NOTE Report any undeclared NOUE events as instructed by EPIP OSC 2. NOTE The Emergency Action Levels described in this procedure are not intended to be used during approved maintenance and/or testing situations where abnormal temperature, pressure, equipment status, etc., is expected. NOTE Ecch EAL contains information on the modes of operation during which it is applicable. If it appears that different classifications could be made for the current plant conditions, the highest classification indicated should be declared. 5.1 Venty the indications of the off normal event or reported sighting. 5.2 Ensure the immediate actions (use of Emergency and Abnormal Operating Procedures) are taken for the safe and proper operation of the plant. R26 FC/EPIP
FORT CALHOUN STATION EPIP OSC 1 ! EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 6 OF 97 5.3 Compare the abnormal conditions with the EAL's listed on Attachment 6.2. Choose : the appropriate EAL 5.4 Turn to the selectred EAL page in Attachment 6.1 and verify the EAL sgainst the ' verrfication criteria and applicable modes. 5.4.1 It verification is made; Declare the__Emeragacy Classification Indicate 1 5.4.2 If venfication is not made; repeat Steps 5.3 and 5.4 and evaluate other rc!ated EAL's as necessary. 5.5 Monitor response activities and plant conditions and adjust classifications as necessary. 6.0 ATTACHMENTS 0.1 Emergency Action Level Verification Criteria 6.2 Emergency Action Levels (EAL's) 0.3 Three Fission Product Barrier Griteria s 4 yd FC/EPIP R26
EPIP OSC 1 ! FORT CALHOUN STATION EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 57 OF 97 EAL 7.1 EAL 7.1 { t Attachment G.1 3 (Continued) IRRADIATED FUEL ACCIDENT Page Si of 88 , VERIFICATION CRITERIA:
- 1. Damage to an irradlated fuel assembly in the plant.
AME
- 2. One or more of the following alarms occur concurrently: t A. Area radiation monitor high alarm (RM-070 through RM-089). ,
- 8. Any of the following process' monitor high alarms:
(a) RM 050 Containment Particulate RM 052 If monitoring containment) (b) RM-051 Containment Gas * (c) RM-0$2 Stack Gas (or RM 0,2 If monitoring stack) , C. Ventilation isolation Actuation Elgnal (VIAS). D. Containment Radiation High Signal (CRHS). APPLICASLE MODES: . 15 EMERMNCY CLASSIFICATION: Alert R26
. FC/EPIP 1
FORT CALHOUN STATION EPIP.OSC 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 58 OF 97 EAL 7.2 ; l EALL2 l Attachment 6,1 (Continued) MAJOR IRRADIATED FUEL ACCIDENT Page 52 of 88 VERIFICATION CRITEBLt
- 1. Major damage (large object damages fuel or water loss below fuel level) to an irradiated fuel assembly in the plant.
AND
- 2. Any area radiation monitor Indicating.1DQO TIMES normal as listed in the TDB Figure IV.8.
APPLICABLE MODES: 15 EMERGENCY CLASSIFICATION: Site Area Emergency R26 FC/EPIP
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- 19.4 shuteense por Inhnice4 1 specl*lcessens.
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4 EPIP-03C-1 FORT L ROUM OTATION pacg 97 or 97 EMERGENCY PLANT IMPLEMENTING PROCEDUCE ATTACHMENT 6.3 THREE FISSION PRODUCT BARRIER CRITERIA I FUEL CtADDING REACTOR COOLANT SYSTEM CONTAINMENT ] i l 1 RCS Dose Equivalent Iodine 131 1. Reactor Coolant System leak greater 1. Any failure of the Containment, its; t 1. sample is greater than 300 pci/gm. than 40 gpm. penetrations, isolation valves, connections and piping extensions '
- 2. Any valid CET Temperature >1000*F. 2. Reactor Coolant System pressure up to the outer isolation valve AND greater than 2350 psia. a release pathway exists to the j
< 3. Representative CET temperatures environeent.
>950*F and rising. 3. Containment pressure greater than 5 psig AND any valid containment Area 2. Containment Hydrogen Concentration
- 4. Failure of the reactor protective Radiation Monitor indication of greater than 3 percent.
system to trip the reactor upon 1000 mr/hr above the normal values ' reaching a limiting safety system listed in TDB Figure IV.S. 3. Containment pressure greater than setpoint. 50 psig. ; S. Pressure-Temperature limits of TDB
- 5. Failure of the Manual Trip to bring III.25 OR EOP Attachment 2 are not 4. Containment pressure ElsING at a reactor power subcritical (reactor met. rate that will exceed 60 psig before corrective action halt or l power less than 10 E-04% AND A situation exists which will cause reverse the pressure increase.
constant or decreasing). 5. , the failure of the Reactor Coolant !
- 6. RVLMs Indicates 0.01 level. System within a predictable time 5. Containment Integrity as defined by L unless successfully corrective Technical Specification is not !
- 7. Figure 1-1 of Technical action occurs- present during an unplanned Specification 1.1, Safety Limits is 6. An event has occurred which has a transient AND the potential exists for a loss of the Fuel Clad #ing or exceeded.
high probability of having damaged the Reactor Coolant System i
- 8. A situation exists which will cause the Reactor Coolant System barrier, Barriers.
the failure of the Feel Clad within but tine has not yet permitted a predictable time unless verification. 6. A situation exists which will cause successful corrective action the failure of the Containment within a predictable time unless occurs. successful corrective action
- 9. An event has occurred whic'3 has a occurs. i high probability of having damaged n the Fuel Clad barrier, but time has 7. An event has occurred which has a ;
not yet permitted verification. high probability of having damaged the containment barrier, but time I has not yet permitted verification. ; I i r R26 FC/EPsP ,
1
]
AOP-07 Page 13 of 57 { Section 1 - Plant _ to Hot Shutdown ) INSTRUCTIONS CONTINGENCY ACTIONS
- 10. (continued)
- d. Throttle the AFW lsolation Valves d.1 IF the AFW isolation Valves have using BOTH of the following failed open, controllers (Al-179): THEN manually throttle BOTH !
- HCV-11078, " AUX FW TO S/G AFW lsolation Valves (Room 81):
RC 2A ISOL VLV e HCV-11078, " STEAM l CONTROLLER GENERATOR RC 2A
^ X' ' Y FEEDWATER ;
- HCV-11088, " AUX FW TO S/G NLE ALV RC 2B ISOL VLV CONTROLLER" e HCV-11088, " STEAM GENERATOR RC-2A AUXILIARY FEEDWATER INLET VALVE"
- 11. IF Aux Feedwater is in service, THEN dite.clthe EONT to stop all of the Feed Pumps, FW-4A/B/C (Turbine Building Basement).
'a Part i R3.2
AOP 07 i Page 12 of 57 l
- Section 1 - Plant to Hot Shutd2E0 INSTRUCTIONS - CQBIJNGENCY ACTIONS ,
E
- 10. (continued) 4
- b. Open ALL of the following valves (Al 179):
i * . HCV-1107A, AFW lsolation Valve
- HCV 1108A, AFW lsolation -
Valve
- YCV 1045, FW-10 Steam inlet Valve
- YCV 1045A, RC 2A to FW-10 Isolation Valve '
o YCV 1045B, RC 2B to FW isolation Valve 9
- c. Place BOTH of the following switches in ' THROTTLE" (Al-179):
- HCV-11078, " AUX FW TO SIG
. RC-2A ISOLATION VALVE" e . HCV-11088, " AUX FW TO S/G RC-2B ISOLATION VALVE" 4 (continue) , P
+
b I Part i R3.2
AOP-07 Page 11 of 57 Section 1 - Plant to Hot Shutdown INSTRUCTIONS CONTINGENCY ACTIONS
- 10. Maintain S/G levels 85 95% NR (94-98% WR) by performing the following steps:
- a. Verfy HC%1384," MAIN AND a.1 IF HCV-1384 is NOT closed, AUXILIARY FEEDWATER THEN plagg HCV-1384 by CROSS-CONNECT VALVE", is performing the following steps:
closed (Room 81).
- 1) QpfD MCC-4C1-E03, "HCV-1384 FW AND AUX FEED WATER CROSS CONNECTION VALVE" (MCC-4C1).
(continue) 2) Class HCV-1384," MAIN AND AUXILIARY FEEDWATER CROSS-CONNECT VALVE" (Room 81). Part 1,R3.2
AOP-07 Page 10 of 57 - Section I- Plant to Hot Shutdgyg1 e b 'HSTRUCTIONS CONTINGENCY ACTIONS
..... ...... ......... 4................................. .......................................
CAUTION v 5 levels less than 85% NR (94% WR) may cause undesired thermal cycles or v/ir r
%tmer.
...... . ............... ........................... ............. ..... 4....... ..... .
RQIE HCV-1107B and HCV-1108B can be positioned when Instrument Air is available using the Hend Load Controllers on panel Al-179 when their control switches are in throttle. If these switches are in open, or Instrument Air is iost, the valves will remain open. 4
- 9. Direct the EON t' to align the Feedwater System by performing the following steps (Turbine Building Basement):
- a. IF more than one Feed Pump is running, ;
THEN 81gg ONE Feed Pump, FW-4A/B/C. !
- b. . Sign all of the Heater Drain :
1 Pumps, l i
- c. IP more than one Condensate l Pump is running, THEN atog ONE Condensate i
Pump, FW-2A/B/C. l Part 1 R3.2 i
L - S ide m(s):f Alternate- Shutdown Panels . I -t . .. . . y . .
' Location (s):. Upper: Electrical Penetration Room (Panel Al 179 & Al 185) t
" Approximate' Time: 10 minutes-Actual T:me:- !
. 1 Reference (s): (1) AOP 07 (R 9) (. _. (P.) NRC K/A 000061,K1.01 (RO4.1/SRO4.2) ' (3) NRC_ K/A 000061,A1.01 (RO3.9/SRO4.2)
~ Verify current reference revisions match those listed above T zOperators Name: SS #: ,
All Critical Steps (*) must be-performed or simulated in
- accordance with the standards contained in this JPM. .
The. operators performance was evaluated as: SATISFACTORY- UNSATISFACTOHY Evaluatars d% nature: Date:
- Reason, if unsatisfactory:
L:: i
,7 f
f
< Operater's - reviewed: . Date:
L (-
,. , ,,,2. -, . . . . , - , , . . , , , . , - .. , , _ . , - . , , , . . . . . n ...
c . - .. -. - - _ - . . =_ . - _ - Initiating cue: A bomb has been found in the shift supervisor's office. The Shift Supervisor has directed an evacuation of the Control Room. All of the actions of step 1 of AOP-07 have been completed. All feedwater pumps are secured. You are directed to establish control at the alternate Shutdown panels. START 1 wd l l
1 Revicion: 0 January 30,1997 Fort Calhoun Station - Operations Training 1 JOB PERFORMANCE MEASURE JPM No.: JPM AOP 07 JPM
Title:
- Forced Evacuation of the Control Room l
- Tools & Equipment: None,
, Safety Considerations: None. l Comments: THIS JPM WILL NORMALLY BE CONDUCTED AS A STATIC JPM IN THE PLANT. THE EVALUATOR MAY ELECT TO PERFORM THIS JPM AS A DYNAMIC JPM USING THE SIMULATOR ALTERNATE SHUTDOWN PANEL t 9 4
i
' Revision: 0 January 30,1997 Fort Calhoun Station - Operations Training JOB PERFORMANCE MEASURE t
-JPM No.: JPM AOP-07 JPM
Title:
Forced Evacuation of the Control Room Initiating cue: A bomb has been found in the shift supervisor's office. The Shift-Supervisor has directed an evacuation of the Control Room. All of the actions of step 1 of AOP 07 have been completed. All feedwater pumps are secured. You are directed to establish control at the Alternate Shutdown panels. START
- CRITICAL ELEMENT STANDARD STEP-I
- 1. Establish control at Al-185 Al-185 Place " REMOTE LOCAL TRANSFER SWITCH 43"In LOCAL Al- 179
- 2. Establish control at Al-179 Place both "AFW CONTROLS TRANSFER SWITCHES,43/RC-2A/B" in LOCAL Al-212
- 3. Start Wide Range Channel NR-004 switch to ON "D" recorder 4
l l i. Revision: 0 January 30,1997 Fort Calhoun Station Operations Training : JOB PERFORMANCE MEASURE
'JPM No.: JPM-AOP-07 JPM
Title:
Forced Evacuation of the Control Room
- CRITICAL ELEMENT STANDARD STEP 4
Al-185
- 4. Maintain Pzr Level 45 60% Operate CH-1B as necessary CUE: PRZ Ievel is 50%
- 5. Maintain RCS pressure MCC-4C1/Al-179 2050 2150 psia. Operate Pzr Backup heaters, Bank 4 4 groups 10,11, & 12 as necessary, CUE: RCS _ pressure is 2100 psia
Rovision: 0 January 30,1997 Fort Calhoun Station - Oporations Training JOB PERFORMANCE MEASURE JPM No.: JPM-AOP JPM
Title:
Forced Evacuation of the Cqntrol Room =====--- ===- w m _ ___ -- __ -=- _;s=
- CRITICAL ELEMENT STANDARD STEP
_ - _ _ = _ _ -== -
- 6. Maintain S/G levels 35 85%
NR by: _ a. Verify HCV-1384 closed a. CUE: EONT reports HCV-1384 is closed ___b. Open all of following : b. Al- 17 9
+ HCV-1107A YCV-1045 control switch to
- HCV-1108A open
*
- YCV-1045 AND
+ YCV-1045A verify red lights on for all
- YCV-10458 valves.
- c. Control feed flow via-
- c. HCV-1107B/1108B control HCV-1108A/B switches to THROTTLE AND Regulate air loader for HCV-1107B/1108B for desired flow. STOP Termination criteria: Control has been established at the alternate shutdown panels
Revision: 0 January 30,1997 Fort Calhouu Station - Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM-AOP-07 JPM
Title:
Forced Evacuation of the Control Room QUESTION: JPM AOP 07-O-1 (new) During this event, what affect will the transfer to Al-185 have on the selected lead charging pump? ANSWER: 1. If the selected pump is CH-1B then the control switch on Al 185 will determine its operation,-
- 2. If the selected pump is CH-1 A/C then the pump will continue to operated until the associated breaker is opened locally,
REFERENCE:
NRC K/A: 004-000 K4.04 (3.2/3.1) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS: l I
l Revision: 0 _ January 30,1997 Fort Calhoun Station - Operations Training , JOB PERFORMANCE MEASURE l JPM No.: JPM-AOP 07 JPM
Title:
Forced Evacuation of the Control Room QUESTION: JPM-AOP-07 O-2 (new) How do you transfer DC control power to Al-179 to the alternate power supply? ANSWER: In-the back of Al-179. ( via a two position switch)
REFERENCE:
NRC K/A: 061-000 A2.03 (3.1/3.4) The operator's response to this question was: SATISFACTORY UNSATISFACTORY . COMMENTS:
. ._ _ _ ~. ~_. ._ _ _ _ . . _ . _ _ _ . . _ . . . .
AOP-07 Page 8 of 57. l Section I - Plant to Hpt Shutdown l l lNSTRUCTIONS CONTINGENCY ACTIONS l l 6
]
RoIE Valve indication becomes available at Al-185 for TCV-202, HCV-240, HCV-249 HCV-248, HCV-239, PCV-102-2, PCV-103-1 and PCV-103 2. i
- 2. Establish control at Al-185,
" ALTERNATE SHUTDOWN PANEL",
(Upper Electrical Penetration Room), by performing the following steps:
- a. Place " REMOTE-LOCAL TRANSFER SWITCH 43"in
" LOCAL".
- 3. Establish control at Al-179, Auxiliary Feedwater Panel (Upper Electrical Penetration Room), by placing both "AFW CONTROLS TRANSFER SWITCHES", 43/RC-2A/8, in " LOCAL".
- 4. jilg1 NR 004, " WIDE RANGE NEUTRON FLUX CHAN "D" RECORDER"(Al-212).
- 5. Monitor Reactor power on NR-004 and Ni 004," WIDE RANGE NEUTRON FLUX CHAN "D" SIGNAL PROCESSOR" (Al-212).
Part 1 R3.2
AOP-07
- Page 9 of 57' Section 1 - Plant to Hot Shutdown JNSTRUCTIONS CONTINGENCY ACTIONS
. CAUTION Charging to the RCS may cause overpressurization due to the isolation of Letdown and RCS Heatup.
- 6. Maintain PZR level (45560%) by operating CH-18, " CHARGING PUMP 18"(Al-185).
- 7. Maintain a record of Charging Pump run time to support estimates of RCS boron concentration.
s . - 8. - Maintain RCS pressure 2050-2150 psia by control of PZR Back-up Heater Bank 4 Groups 10,11 and 12 (MCC-4C1). Part 1 R3.2
AOP-07 Page 13 of 57 Section I - Plant to Hot Shutdown JNSTRUCT!ONS CONTINGENCY ACTIONS
- 10. (continued) . ,
- d. Throttle the AFW lsolation Valves d.1 IF the AFW lsolation Valves have using BOTH of the following failed open, controllers (Al-179): THEN manually throttle BOTH
- HCV-11078, " AUX FW TO S/G AFW isolation Valves (Room 81):
RC-2A ISOL VLV e HCV-11078, " STEAM ^ CONTROLLER GENERATOR RC-2A RY FEEDWATER
- HCV-11088, " AUX FW TO SIG hUXI RC-28 ISOL VLV CONTROLLER" e HCV-11088, " STEAM GENERATOR RC-2A AUXILIARY FEEDWATER INLET VALVE" 11, IF Aux Feodwater is in service, THEN ditesithe EONT to stop all of the Feeo Pumps, FW-4A/B/C (Turbine Building Basement).
m. Part i R12
. . - . - - -. . . . -. . -_ - -- - -- - - . ~.. .- - -
l l AOP-07-- l Page 12 of 57 '- - Section I- Plant to Hot Shutdown.-- INSTRUCTIONS - CONTINGENCY ACTIONS 10; (continued) 1
-b. Open ALL of the following valves (Al-179):
e- HCV-1107A, AFW isolation Valve e, HCV-1108A, AFW lsolation. Valve o YCV-1045, FW-10 Steam inlet Valve 4 e YCV-1045A, RC-2A to FW-10 Isolation Valve e YCV-10458, RC-2B to FW-10 ' Isolation Valve
- c. Place BOTH of the following
, switches in ' THROTTLE" (Al-179):
- HCV-1107B, " AUX FW TO S/G RC-2A ISOLATION VALVE"
,
- HCV-11088, " AUX FW TO S/G RC-2B ISOLATION VALVE" i
i [ (continue) J i I . Part 1- R3.2 I y 4 , , , , _ , , - - - , - - - - - . , -,--_ ,y. ~ ma,r ,. ..~.. -_ . - , . . . -_,4 _______ _ --
_. _. _. _ . _ . - . . . . . . _ .. _ - - . _ - _ . _ .. . . ~ _ _ _ _ _ _ _ . . . _ 4 AOP-07 Page 11 of 57- .] I Section 1 - Plant to Hot Shutdown ,
- INSTRUCTIONS CONTINGENCY ACTIONS i- _.. :
- 10. Maintain S/Glevels 85-95% NR (34-98% WR) by performing the
- following steps:
4 a; Venfv HCV-1384," MAIN AND a.1 IF HCV-1384 is NOT closed, AUXILIARY FEEDWATER THEN gig.at HCV-1384 by CROSS-CONNECT VALVE", is performing the following steps: l-closed (Room 81).
- 1) Qg2D MCC-4C1-E03, "HCV-1384 FW AND AUX l
FEED WATER CROSS CONNECTION VALVE" (MCC-4C1). (contince) 2) Close HCV-1384," MAIN AND AUXILIARY FEEDWATER CROSS-CONNECT VALVE" (Room 81).- i .. t J 1 f {, -%d Part 1- R3.2
AOP-07 Page 10 of 57 - Section 1 - Plant to Hot Shutdown INSTRUCT!ONS CONTINGENCY ACTIONS GAUTION S/G lovels less than 85% NR (94% WR) may cause undesired thermal cycles or water hammer. NOTE HCV-11078 and HCV-11088 can be positioned when Instrument Air is available using the Hand Load Controllers on panel Al-179 when their control switches are in throttle, if these switches are in open, or Instrument Air is lost, the valves will remain open.
. 9. Qkagt the EONT to align the Feedwater System by performing the following steps (Turbine Building Basement);
- a. IF more than one Feed Pump is running, THEN gigg ONE Feed Pump, FW-4A/BIC.
- b. Sigg all of the Heater Drain Pumps.
- c. !F more than one Condensate Pump is running, THEN 11gg ONE Condensate Pump, FW-2A/B/C.
Part 1 R3.2
Revision: 5 Jan. 23,1997 Fort Calhoun Station - Operations Training JOB PERFORMANCE MEASURE
' J PM 11o. : JPM-0335 (OLD No:- SI-21)
JPM
Title:
Containraent . Spray Pump Operability Test System (s): Safety Injection Location (s): Control Room Approximate Time: 10 minutes Actual Time: Reference (s): (1) OI-CS-1-6.2 (R-11) (2) K/A 026000A401 (4.5/4.3) 5 Verify current reference revisions match those listed above Operator's Name: SS #: All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. The operator's perfortnance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluator's Signature Date: Reason, if unsatisfactory: Operator's reviewed: Date:
i
+
t
--(
- Initiating Cue: ,
Containment spray pump SI-3C was tagged out for breaker. ' maintenance. The electricians have removed all tags anf' the Shift Supervisor directs you the LO, to perform an
- operability check of the pump.- The breaker is racked in and the 69 switch is red flagged. - All prerequisites are met.-
START. i
-?
?
8 e i h 2 0 d' f i 4 5 i k.u< wwy awwe yw =- wwrw,+vy --yyyy yw,w--- ,ma e-----w-y- r ,-r Twy-'rr C r-t wr g-q- noe ryet vt - --+ emy-- -- - Tp ~ w7
Revision: 5 Jan. 23,1997-Fort Calhoun Station - Operations Training JOB PERFORMANCE IT.ASURE JPM !Jo.: JPM-0335 (OLD No: SI-21) JPM
Title:
Containment Spray Pump Operabjlity Test Tools &. Equipment: None. Safety Considerations: Inoperable pumps affect Technical Specification Requirement. Comments: THIS JPM WILL BE PERFORMED AS A DYNAMIC JPM ON THE SIMULATOR. SIMULA'IOR IC SET (S) :
Revision: 5 Jan. 23,1997-Fort Calhoun Station -' Operations Training
-JOB PERFORMANCE MEASURE JPM No'.": JPM-0335 (OLD No: SI-21)
JPM - Title : Containment Spray Pump Operability Test m=================================================================
- CRITICAL ELEMENT STANDARD STEP '
==================================================================
Initiating Cue
-Containment spray pump SI-3C was tagged out for breaker maintenance. The electricians have removed all tags and the Shift Supervisor directs you the LO, to perform an operability check of the pump. The breaker is racked in and '
the 69 switch is red flagged. All prerequisites are met. START. AI-30A/B
- 1. Place Containment HCV-344 in TEST Spray Valve test AND switches in test. HCV-345 in TEST l
- 2. Verify HCV-344/345 A33-1 ,
SET SPRAY PUMP TEST H5 and A34-1 H3 ALARM lights i Permit annunciators -lit. . are in ALARM. 1
- 3. Close the discharge AI-128 valve HCV-2978 for HCV-2978 in CLOSE.
j SI-3C. GREEN light lit. l
- 4. Check marual . recirc Direct EONA to verify SI-152 valve open (SI-152) OPEN.
Cue: . BONA reports valve is OPEN. I Mr d
Revision: 5
'Jan. 23,1997-Fort Calhoun_ Station --Operations Training JOB PERFORMANCE MEA.SURE - JPM ' f fo . : JPM-0335 (OLD Not SI-21)
JPM Tit lra t L M inmant Sorav - P"am Ocarability Tant
==e===============================================================
- CRITICAL ELEMENT STANDARD
< STEP =======================================================n==========
- 5. SIRWT recirc valves AI-30A/B HCV-?85 and HCV-386 OPEN. Verify HCV-385 &386 in OPEN AND RED lights lit.
- 6. Start SI-3C and run 'AI-30B pump for minimum of SI-3C C.S. to START 5 minutes. AND RED light lit.
Cue 5 minutes are have passed.
- 7. Secure SI-3C. AI-30B SI-3C C.S. to SE P AND GREEN light lit.
- 8. OPEN SI-3C AI-128 discharge valve HCV-2978 in OPEN HCV-2978. RED light lit.
- 9. Close and lock Direct EONA to close and lock SI-152. SI-152.
Cues' SI-152 is shut and locked.
. - -- - ... .. . ~. . . . . . - . . - . - . . - - . . . . - . . _ . . - - - - . . - . . - 5 h Revision': 5: Jan.,23,1997 I Fort Calhoun-Station - Operations Training.. , JOB ~ ; PERFORMANCE MEASURE.
; JPH tio. .JPM-0335 (OLD No's SI-21).
JPM : Titl'e r Containme'nt Spray Pump Operability Test
================================================================== ,
- CRITICAL- ELEMENT STANDARD 'l STEP-
============================================================= ====
- 10. Return HCV-344 & AI-30A/B -
345 test switches to'OFF. Test switches to OFF.
.11. Verify annunciators A33-1 H5 and A34-1 H3 alarms CLEAR. RESET.
Termination Criteria: Pump operability check has been completed and all controls have been returned to normal. P F N' 'M 9 P+mi- m.-.1im-p m, _ , _ _, ,_
Revision: S Jan.-23,1997-Fort Calhoun Station - Operations Training. JOE : PERFORMANCE . MEASURE VPM No.: JPM-0335- (OLD No: SI-21). JPM
Title:
Containment Spray - Pump - Operability Test QUESTION: JPM-0335-Q-4 Why is the pump discharge valve closed before starting the containment spray pump?: ANSWER: Closing the discharge valve is a precautionary measure taken to prevent leakage in case either of the spray valves are not closed.
REFERENCE:
imC K/A: 026-000-A3.01 (4.3/4.5) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
l l l Revision: 5 Jan. 23,1997- ) i Fort Calhoun Station - Operations-Training JOE PEkFORMANCE MEASURE I JPM No.t JPM-0335 (OLD No - SI-21)
. JPM
Title:
Containment Spray Pump _ Operability Test j QUESTION: JPM-0335-0-7 (new) What' affect, if any, would an inadvertent CSAS have on the containment spray valves if the actuation occurred during the 5 minutes that the pump was running? ANSWER: HCV-344 would remain closed and HCV-345 would open.
REFERENCE:
NRC K/A: 026-000-A2.03 :(3.9/4.2) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS: 4 9 4
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r
" J Fort Calhoun' Station -
- Unit No.~ 1 .
- .t 4
h o Ol-CS-1 OPERATING INSTRUCTION 1 a .e
Title:
CONTAINMENT SPRAY NORMAL OPERATION I 1 FC 68 Number: 46971 Reason for Change; Reformat per SO-G-73A. Remove Step 6,1. i l . Contact Person: R. Amold i F c-ISSUED: '02-2'7 -97 9:30 am R11 , I
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- FORT CALHOUN STATION 01-C 3-1 --
LOPERATING INSTRUCTION - PAGE 1 OF 94 .; CONTAiFMENT SPRAY - NORMAL OPERATION . > 4 SAFETY RELATED n PURPOSE , PAGE
,1 - Containment Spray Pump Operability Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -
t REFERENCES / COMMITMENT DOCUMENTS
- 1. Technical Specifications l e: 2.3, Emergency Core Cooling System - !
- 2.4, Containment Cooling .
- 2. USAR
;e 6.3, Containment Spray System
*2 7.3 2.4, Containment Spray Actuation Signal (CSAS) e 14.15, Loss-of-coolant Accident e 14.16; Containment Pressure Analysis
- 3. P&lD .
E-23866-2'lo-130, Safety injection and Containment Spray System Flow Diagram + t APPENDICES
- O l - C S C L-A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4
4 R11
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FORT CALHOUN STATION Ol-CS-1 OPERATING INSTRUCTION PAGE 2 OF 14 c Contnuous Use Attachment 1 - Containment Spray Pump Operability Test PREREQUISITES f/_1 . INIT.
- 1. Procedure Revision Verification Master Revision Number Date:
- 2. Electrical buses 183C,1848,1838, and MCCs 3A2 and 4C2 are energized.
- 3. Instrument Air System is in service PER Ol-CA-1.
- 4. The Component Cooling Water System is in operation PER OI-CC-1.
- 5. The SIRWT has been filled to normal operating level with at least Refueling Boron Concentration PER OI-CH-4.
PROCEDURE
- 1. Place the following switches to TEST:
- CNTMT Spray Valve HCV-344 Test Switch
- CNTMT Spray Valve HCV-345' Test Switch
- 2. Verify the following annunciators are in alarm:
- A33-1/ H-5, HCV-344/345 SET SPRAY PUMPS TEST PERMIT
- A34-1/ H-3, HCV-344/345 SET SPRAY PUMPS TEST PERMIT CAUTION Closing the discharge valve on a Containment Spray Pump renders the pump inoperable PER Technical Specification 2.4.
- 3. Review Technical Specification 2.4 requirements, and log into the appropriate LCO.
LSO
- 4. Close the following valve for the pump to be tested:
- HCV-2958, C Spray Pump SI-3A Disch _
e HCV-2968, C Spray Pump SI-3B Disch _
- HCV-2978, C Spray Pump SI-3C Disch _
- 5. Ensure the following valve for the pump to be tested is unlocked and open: '
- Sl-138, Containment Spray Pump SI-3A Minimum Recirc Iso Valve _
- SI-146, Containment Spray Pump SI-3B Minimum Recirc iso Valve _
- SI-152, Containment Spray Pump SI-3C Minimum Recire Iso Valve _
R11
FORkbALHbUN STNTION! 1 biAb-1[ d OPERATING INSTRUCTION: - PAGE 3 OF 14- ; l .e t
-k continuous use -
l Attachment 1 - Containment Spray Pump _Operabili_ty Test l
- PROCEDURE (continu,ed): (2.} l INIT:
6;: Verify 't he following valves are open. . e: HCV-385l SIRWT-Tank Recirculation Valve - e HCV-386, SIRWT: Tank Retirculation Valve . q
- 11. Start the pump to be tested:
ei SI 3A; CNTMT Spray Pump _
- SI-38, CNTMT Spray Pump _.
e' SI-SC, CNTMT Spray Pump :
?
'8. WHEN 'at least five minutes has elapsed,-
THEN stop the pump. ]
- 9. Open the discharge valve for the pump tested:
- *- HCV-2958,- C Spray Pump SI-3A Disch e _ HCV-2968, C Spray Pump SI-38 Disch e HCV-2978, C Spray Pump SI-3C Disch
- 10. Close and lock the Minimum Recirc iso Valve for the pump tested.
. .: SI-138, Containment Spray Pump SI-3A Minimum Recirc iso Valve ,
- SI-146, Containment Spray Pump SI-3B Minimum Recire Iso Valve -
*- SI-152, Containment Spray Pump SI-3C Minimum Recirc Iso Valve ._._
ind Ver
- 11. Place the following switches _to OFF: ,
e= CNTMT Spray Valve HCV-344 Test Switch q .* CNTMT Spray Valve HCV-345 Test Switch
- 12. Verify the following annunciators are clear:
-e ~A33-1/ H-5, HCV-344/345 SET SPRAY PUMPS TEST PERMIT
-* - A34-1/ H-3, HCV-344/345 SET SPRAY PL'iv DS TEST PERMIT:
- 13. - Log out of the Technical Specification 2.4 LCO.
'50L
, Completed by' - Date/ Time._ I
~
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_ . . . . . . _ . _ _ . . . _ - _ . . . _ - ~ . . _ _ _ . _ _._ FORT CALHOUN STATION . Ol-CS OPERATING INSTRUCTION PAGE 4 OF 14
- System Containment Spray .
- Checklist: Ol-CS-1-CL-A -
2 Page 1 of 11
Reference:
P&lD E 23866-210-130 Position Valve Not Descriotion Desired b_G1 Mal Rm 21 lA LCV-383-1-B LCV-383-1 Instrument Air Supply isolation Valve Qatta_ BA-LCV-383-1-V . LCV-383-1 Instrument Air Supply Vent Valve Closed LCV-383-1 SIRWT SI 5 Outlet Header Level Control Valve HJR Locked SI-156 Containment Spray SIRWT SI-5 Sample Valve Closed IA-LCV-383-2-8 LCV-383-2 Instrument Air Supply isolation Valve Open IA LCV-383-2-V LCV-383-2 Instrument.Alr Supply Vent Valve Closed LCV-383-2 SIRWT SI-5 Outlet Header Level Control Valve HJR IA-HCV-2957-B HCV-2957 Instrument Air Supply Isolation Valve Open HCV-2957 Containment Spray Pump SI-3A Suction Valve HJR _ _ , _ 4 SI-137 Containment Spray Pump SI-3A Disch Drain to Locked Waste Disposal isolation Valve Closed t SI-138 Containment Spray Pump SI-3A Minimum Recire Locked Isolation Valve Closed SI-228 Containment Spray Pump SI-3A Disch Press Indicator Root Valve Closed
' SI-331 Containment Spray Pump SI-3C Casing Vent Valve Closed DW-185 Containment Spray Pump SI-3A Domin Water Locked
, Isolation Valve ' Closed . lA-HCV-2958-B HCV-2958 Instrument Air Supply isolation Valve Open 'k Completed by Time /Date: / s R11
, - - - , , m - - - - - . - - >~ c
~~-= - - - -
Revision: 3 September 23,1996
+
Fort Calhoun Station - Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM 0612F (Faulted) JPM
Title:
Start a Reactor Coolant Pump System (s): Reactor Coolant Location l's): Con'=' Room Approximate Time: 5 minutes Actual Time: Referenos(s): (1) OI RC G Att.1 (R 26) (2) NRC K/A 003000G013 (3.6/3.7) Verify current reference revisions match those listed above Opstator's Name: _ SS #: All Critical Step (') must be performed or simulated in accordance with the standards contained in this JPM. Tha operators performance was evaluated as: SATISFACTORY UNSATISFACTORY Ev luators Signature: Date: ROCson, if unsatisfactory: v Operator's reviewed: Date:
. _- . _ _ _ . . . _. - . _ = _ - - - _ . _ . - . - . _ - - . . Initiating Cue: i A Plant startup is in progress, RC 3A, RC 38, and RC 3C are running. You are directed to place RC 3D in service. All prequisites are met. START.
Rsvision: 3 September 23,1996 Fort Calhoun Station Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM 0612F (Faulted) JPM
Title:
- Start a Reactor Coolant Pump Tools & Equipment: None. 3 i Sti;ty Considerations: None. 4
- Comments: THIS JPM WILL BE PERFORMED AS A DYNAMIC JPM ON THE SIMULATOR. j
' _ SIMULATOR 10 SET (S): Setup; file zrepjpm l I i B w
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Rovision: 3 September 23,1996 Fort Calhoun Station Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM 0612F (Faulted) JPM
Title:
Start a Reactor Coolant Pump
- CRITICAL ELEMENT STANDARD STEP
- - _ - . . . . . . = . . . . . . . . . .
Initiating Cue: A Plant startup is in progress, RC 3A, RC-30, and RC 3C are running. You are directed to place RC 3D in service. All prequisites are met. START.
- 1. Set up ERF Display Screen for EFF RC 3D. Type (443), Press (DSP).
AI270
- 2. Station Operator to respond to CUE: Operator is at Al 270 Vib. Alarms Room 57
- 3. Locally verify Controlled Bleed EONT reports Control Bleed Off Off Flow. Flow is 1.0 gpm.
CB 4 or Al 31
- 4. Ensure RCP NPSH.- PT-115 or PT 105 and TDB lll.25
Revision: 3 September 23,1996 Fort Calhoun Station Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM 0612F (Faulted) JPM
Title:
Start a Reactor Coolant Pump
-======-----------============================_-===================-=====
- CRITICAL- ELEMENT STANDARD STEP
=====.__ ; . . _ _ _.._ _------ -_._ = = == == =-
. _ , ================= _.-___=====
- 5. Ensure 86/RC 3D Reset. CB 1/2/3 86/RC 3D RED light lit.
- 6. RCP reverse rotation CB 1/2/3 alarm is cleared. ANN, A 6 D 5 clear.
- 7. START RC 3D oil lift 08 1/2/3 pump. - RC-3D 1 in START AND RED light lit.
C B- 1/2/3
- 8. Start the selected RCP:
- a. wait 2 minutes
_ a. Run oil litt pump b. RC-3D in AFTER START
,__ b. Start RC 3D. AND Red light lit.
R3 vision: 3 September 23,1996 Fori Calhoun Station Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM 0012F (Faulted) JPM
Title:
Start a Reactor Coolant Pump
========== = =
- CRITICAL ELEMENT STANDARD STEP
------ -- ==========u==___ _ =========================
a.. ._
- 9. Verify the following: CB 1/2/3
- a. Ensure oil pump stops. RC 3D 1 Observe GREEN light lit.
.__ b. Monitor amps. Observe Ammeter on CB 3 drops below 425 amps within 17 seconds.
- 10. Trip RC 3D RC 3D C.S. to After Stop and Loen light Ilt.
CUE: Pump amps drop to 0. i TCrmination Criteria: RC-3D is Shutdown w
- . - - - = -. - ..-.----.-_-.--__--_. - -.-- - -.
t Revision: 3 ; September 23,1996 ; Fort Calhoun Station Operations Training ; JOB PERFORMANCE MEASURE l JPM No.: JPM 0612F (Faulted) JPM
Title:
Start a Reactor Coolant Pump GUESTION: JPM 0612 O 1 The recommended seal bleed off temperature when at 100% 3 full power is 100140'F:
- 1) How la this temperature monitored?
- 2) How is this temperature controlled?
ANSWER: 1) Temperatures are monitored (and alarmed) on ERF Computer
- 2) Ternperatures are controlled via CCW outlet valves operated on Al 45.
REFERENCE:
K/A 003000A102 (2,9/2.9) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
___ _ _ _ _ . _ . _ _ _ . _ _ _ _ ~ __ _ _ ___._._ _ .___ _ _ _.__ _ __ I Revision: 3 September 23,1996 Fort Calhoun Station Operations Training f JOB PERFORMANCE MEASURE-JPM No.: JPM 0612F (Faulted) JPM
Title:
Start a Reactor Coolant Pump l CUESTION: JPM 0012 0 2 (new) > When starting the first RCP, there is a restriction on either pressurizer level or the AT between the steam generator secondary side and Teold. What is the basis of this temperature limit?
' i NSWER: It ensures (that a single PORV will provide) overpressure protection from the resultant insurge into the pressurizer,
REFERENCE:
K/A 002 020 K5.08 (3.8/4.1) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
FORT CALHOUN STATION Ol RC-9 OPERATING INSTRUCTION PAGE 5 OF 57 REFERENCES / COMMITMENT DOCUMENTS (continued)
- 7. SDFRP 5, Attachment 30, isolation of Dilution Flowpaths
- 8. PED SYE 96-197, Table 1,2,3 and 4, RCP Normal Operating Parameters (Note)
- 9. Commitments IMPLEMENTING COMMITMENT SOURCE STEP NO. HUMBER (ClQ1 DOCUMENT Attachment 1 Caution 1 950785 SER 24-95 APPENDICE_S
- 1. Ol RC-9-CL-A, Reactor Coolant Pumps
- 2. Table 1,2,3 and 4, RCP Normal Operating Parameters v
R30
FORT CALHOUN STATION Ol RC-9 OPERATING INSTRUCTION PAGE 6 OF 57 c Conbnuous Use Attachment 1 - Starting Reactor Coolant Pumps (Coupled) PREREQUISITES $1L NOTE Ol RC-9, Attachment 7 provides recovery guidance from the possibility of a slug of water with reduced boron concentration in an idle RCS loop.
- 1. Procedure Revision Verification Master Revision Number Date:
- 2. The Reactor is Shutdown (Mode 3, Mode 4 or Mode 5).
- 3. Communications have been established between the Control Room and locally at each Reactor Coolant Pump, as needed, prior to Reactor Coolant Pump start.
- 4. Applicable sections of Checklist OI-RC-9-CL-A have been completed for each RCP to be run.
5, A Minimum Reactor Coolant System pressure of 155 psia, OR minimum pressure per Technical Data Book (TDB 111.25) has been established to ensure adequate Reactor Coolant Pump Net Positive Suction Head (NPSH).
- 6. The Component Cooling Water System is in operation to each Reactor Coolant Pump to be run PER OI-CC-1. _
- 7. The Red Motor Heater Lights for each non-operating Reactor Coolant Pump are energized.
- 8. The ERF Computer is operable per Ol-ERFCS-1.
- 9. If the RCP has been idle for a long outage OR for maintenance, a general visual inspection of the RCP should be pmformed prior to starting.
- 10. The Reactor Coolant Pump Motor upper and lower oli reservoirs are filled to a level between 70-110%.
- 11. Instrumentation and alarms associated with RCPs to be run are operable.
R30
FORT CALHOUN STATION OI RC 9 OPERATING INSTRUCTION PAGE 7 OF 57 c Conbnuous Use Attachment 1 - Starting Reactor Coolant Pumps (Coupled) PREREQUISITES (Continued) .60. Ih0L
- 12. The Reactor Coolant System has been filled PER OI RC-2A.
- 13. 4.16 KV Buses for RCPs to be run (1 A1,1 A2,1 A3 and 1 A4) are energized.
- 14. RCPs to be operated are couoled PER MM-RR RC-0008.
PROCEDURE QAUTIONS [1.] IF the Steam Generator secondary side temperature is >30* F above the RCS cold leg temperature, the effects of heat input from the secondary side, and the subsequent effects on heatup rate, must be evaluated / considered.
- 2. A Reactor Coolant Pump shall NOT be started unless at least Qat of the following conditions is satisfied:
e
- Actual PZA The Steam level issecondary Generator s50% (TDB side lli.1a) temperature is <30* F above that of the RCS Cold Log temperature
- 3. Each RCP shall be monitored continuously on the ERF Computer throughout each starting period.
NQIE RCPs must be started in the listed order. Any deviation from this order must be approved by the Manager-Operations. High vibration alarms should be anticipated when starting RCPs.
- 1. Set up one ERF Display Screen to monitor the selected Reactor Coolant Pump and Motor parameters as follows:
- RC-3C '442' DSPL
- RC 3B "441' 'DSP' J
- RC-3A 440' 'DSP'
- RC-3D ;443] ;DSP]
R30 Y
1 FORT CALHOUN STATION ._ OI RC 9 ' OPERATING INSTRUCTION PAGE 8 OF 57 e contnuous use Attachment 1 - Starting Reactor Coolant Pumps (Coupled) PROCEDURE (Continued) U) ltdL
- 2. Station an operator at Al 270 (Room 57) to respond to vibration alarms.
Oi RC-13 provides alarm responso guidance. NOTE At low RCS Pressure, verification of positive Controlled Bleedoff Flow m :y NOT be possible.
- 3. Locally verify positive Controlled Bleedoff Flow for the selected RCP:
- RC 3C FIA-3155 (RM. 57)
- RC-38 FIA 3135 (RM. 57)
- RC 3A FIA-3115 (RM 57) .
- RC 3D FIA-3175 (RM. 57)
- 4. Ensure RCP Pressure is greater than minimum NPSH Pressure PER TDB lll.25.
8 R30
FORT CALHOUN STATION OleRC-9 OPERATING INSTRUCTION PAGE 9 OF 57 c Continuous Use Attachment 1 - Starting Reactor Coolant Pumps (Coupled) PROCEDURE (Continued) M INIL NOTE IF Lockout Relays are NOT RESET, THEN Electrical Maintenance shall be contacted for an inspection of the affected relay (s)
- 5. Ensure the Lockout Relay Switch for the selected RCP is in RESET and the Amber indicating light is ON:
- 86/RC-3C e 86/RC-38
- 86/RC-3A e 86/RC-3D
- 6. REACTOR COOLANT PUMP REVERSE ROTATION Annunciator is CLEAR (Ann, A-6):
- RC 3C C-5 e RC-3B B-5
- RC-3A A-5
- RC-3D D-5 CAURON -
Prior to starting a RCP when personnel are in Containment, the Containment Coordinator sha!! be notified, OR an announcement made on the Geitronics. __
- 7. Start the Oil Lift Pump for the selected RCP:
- RC-3C-1
- RC-3B-1
- RC-3A-1
- RC-3D 1
- 8. Verify adequate ARRD Lube Oil Flow for the selected RCP:
- RC-3C F3187 (ERF Page 342
- RC-3B F3184 ERF Page 342
- RC-3A F3181 ERF Page 342
- RC-3D F3190 ERF Page 342 R30
FORT CALHOUN STATION Ol RC 9 : OPERATING INSTRUCTION PAGE 10 OF 57 i e Continuous Use Attachment 1 - Staning Reactor Coolant Pumps (Coupled) P_RQHEDJ)RE (Continued) u) 181L CAUTIONS WHEN jogging RCPs to sweep Steam Generator U-Tubes, Pressurizer Lcvel will drop, if jogging RCPs to sweep Steam Generator U-Tubes, only run RCPs for two minutes. Do NOT exceed five minutes.
- 9. Startup sequence for selected RCP:
- a. Run the oil lift pump for the selected RCP a minimum of two minutes,
- b. Start the selected RCP:
- RC-3C e RC-38
- RC-3A
= RC-3D
- 10. Verify the following for the selected RCP:
- a. Oil Lift Pump stops (Green indicating light ON):
- RC 3C-1
- RC-3B-1
- RC-3A-1
- RC 3D-1 v
R30
FORT CALHOUN STATION OI RC 9 OPERATING INSTRUCTION PAGE 11 OF 57 c Continuous Use i Attachment 1 Starting Reactor Coolant Pumps (Coupled) PROCEDURE (Continued) .U.l. INIL
- 10. b. Ammeter drops below 425 amps within seventeen seconds:
e RC-3C !
- RC 38 e RC-3A i e RC 3D i
- c. REACTOR COOLANT PUMP REVERSE ROTATION Annunciator is CLEAR (Ann. A-6):
e RC-3C C-5
- RC-3B B-5 e RC#JA A-5
- RC 3D D5 MO.IE RCP Vibration Hi annunciator may be erratic until pump speed / flow stabillzes.
- d. REACTOR COOLANT PUMP VIBRATION HI Annunciator is CLEAR (Ann. A 6):
- RC 3C C-4 e RC 3B B-4
- RC 3A A-4
- RC-3D D-4 1
k w R30
FORT CALHOUN STATION Ol RC-9 OPERATING INSTRUCTION e PAGE 12 OF 57 : continuous use j At;achment 1 - Starting Reactor Coolant Pumps (Coupled) f 4 PROCEDURE (Continued) . . 10. 1NJL i NOTE , At low RCS Pressure, verification of positive Controlled Bleedoff Flow i mcy NOT be possible. , 11. Verify positive Controlled Bleedoff flow for the selected RCP:
- RC 3C F3155 (ERF Page 342) e - RC 3B F3135 (ERF Page 342) e RC 3A F3115 (ERF Page 342) ,
o RC-3D F3175 (ERF Page 342) i Monitor the ERF Computer and verify all rarameters are normal for the
- 12. l ,
selected RCP:
- RC-3C e RC48 '
i e RC-3A e RC-3D 13, IF the RCP was started (jogged) for sweeping Steam Generator U-Tubes-AND has run for two minutes, - THEN shutdown the RCP per Attachment 2. , ,
- 14. IF other RCPs are to be started, THEN repeat Steps 1 through 12 for each RCP to be started; Completed by Date/ Time I ,
4 *e a R30
FORT CALHOUN STATION Ol EE 3 OPERATING INSTRUCTION PAGE 11 OF 20 e Conbnuous Use Attachment 3 - Dattery Charger No. 3 Operation (l.) INIT. PROCEDUllE (continued)
- 3. IF placing EE-8E,12SV DC Battery Charger Number 3 in Equalize Mode, THEN perform the following:
- a. Vonfy Charger No. 3 is in service,
- b. Rotate the equalizing voltage potentiometer counterclockwise (CCW) to its minimum voltage $6iting.
EM HQIE When equalizing mode is selected, Annunciators A15 C3, DC BUS #1 LOW VOLTAGE or A19 C1, DC BUS #2 LOW VOLTAGE and A15-C5, BATTERY CHARGER #3 TROUBLE may alarm. The alarm (s) will clear when the equalizing voltage is adjusted
- c. IF desired to use the equalizing Timer (up to 120 hours),
THEN set the Timer for the number of equalizing hours required.
- d. IF desired to use the equalizer switch, THEN place the Voltage Selector in EQUALIZE,
- e. Adjust the battery charger for a terminal voltage of 135.5 to 136.5 (136.0 VDC is normal).
EM Completed by Date/ Time / R12
FORT CALHOUN STATION Ol-EE 3 OPERATIN3 INSTRUCTION PACE 12 OF 20 System: Plant Electrical Distribution Checklist: Ol EE 3-CL-A Page 1 of
Reference:
P&lD Figure 8.1 1 Breaker Position No Qtscriotion QtikDA 6 Glut! Eqom 56 (East Switchoegt) EE4R Battory Number 1, EE-8A Discharge Test Breaker Locked Ooen EE-8F 125V DC Number 1 Main Distribution Panel EE 8F-CB1 Battery Number 1 EE-8A Main Breaker Locked Closed - EE-8F CB2 Batt Charger 3, EE-8E .0oen EE-8F-CB3 Battery Charger 1, EE-8C .C.l.asad. EE 8F-CD4 Emerg Bearing Oil Pump LO 4 Closed EE-8F-CBS Inverter '1' EE-8P .Clated. EE-8F Batterv Number i 125V DC Bus Numter.1 EE4F-CBSA Spare .Qpen ..... EE-8F-CBS Spare Ooen EE-8F-CB7 REMOVED N/A EE-8F-CB8 Man. Trans. Switch ATD D2 Emerg. Power Closed EE-8F-CD9 Man. Trans. Switch ATD-D1 Normal Power Closed - EE-8F-CB10 Man. Trans. Switch Al-41B-MTS Emerg. Power Closed EE 8F-CB11 Aux Bldg Emgy Lighting Panel ELP-1 Closed-EE-8F CB12 Man. Trans. Switch 183C-40-MTS Normal Power Closed EE-8F-CB13 Man. Trans. Switch 1B3848-MTS Emerg. Power Closed EE-8F CB14 Man. Trans. Switch 183A-4A-MTS Normal Power Closed EE 8F-CB15 Man. Trans. Swtich 1 A2-1 A4-MTS Emerg. Power - Closed Completed by Date/ Time / J R12
- . - . . - . ~. - . . _ - . _ = - - . ~ . - - . . - - - . - - - . - . - . - - - . - - . - -
5 l i Revision: OL Jan. 24,1997 ] . Fort Calhoun Station - Operations Training ! JOB PERFORMANCE - MEASURE ! 4
! JPM llo.: JM4-WDG J JPM Title Waste gas transfer from the vent header to the i gao decay tank.
System (s)l: Waste Disposal Gas Location (s): Auxiliary Building Controlled Area ; Approximate Time: 10 minutes' Actual Time ' I 1 Reference (s): (1) OI-WDG-1 (R-18) (2) K/A 071-000 A4.29 (3.0/3.6) l 4 Verify current reference revisions match those listed above
'P
, Operator's 11amos SS #1 All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. The operator's performance was evaluated as ! i SATISFACTORY UNSATISFACTORY l t Evaluator's Signatures Dates l Reason, if unsatisfactory: , Operator's reviewed: Dates l e q
, ,.-y .-,c..-n.#- .,.re.---,,,ho,y-,y,y,v,.. ,y,-. ,.,,,--,,,y,--,-,,,,h,..,, -1,,, - , - , y,_,..,-, , . , , , , _ , . . , , - .,m . , ,.-,, .,- -#.u-,,,,,
t i i Initiating Cuoi t l Vent header pressure is at 2 psig and you are directed to !' pump the vent header to.the in service gas decay tank using
- WD-28A until vent header pressure has been reduced to 1 psig,
- - AI-110 is operable and has been sampling the in service gas decay tank for the past 20 minutes.
All prorequisites are met. START l l 1 l t I f ii 1 h i a 1 ! 1 b h. t
, _ i j
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i t Revision: 0 Jan. 24,1997 l Fort Calhoun Station - Operations Training , JOB PERFORMANCE HEASURE i J PM No . : JTM-WDG JPM Titles Waste gas transfer from the vent header to the gas decay tank. Tools & Equipment: None. Safety Considerations: OBSERVE REQUIREMENTS OF T!!E RWP FOR ENTRY IITIO Ti!E CO!TTROLLED AREA OF Tl!E AUXILILARY BUILDING. ; Consnents : TilIS JPM WILL BE PERFORMED AS AN ORAL EXAM IN Tile , CO!TTROLLED AREA OF T11E AUXILIARY BUILDING. (AI-100 AND ROOM 16) i
~a. . - ~ ~ - -- -. . - , . _ .
Revision: 0 Jan. 24,1997-Fort Calhoun Station - Operations Training JOE PERFORMANCE MEASURE j l JPM 11o. JFH-WDG l JTH Title Waste gas transfer from the vent header to the gas decay tank.
======================================================
- CRITICAL- ELEMENT- STANDARD STEP
======================================================
Initiating Cues Vent header pressure is at 2 psig and you are directed to pump the vent header to the in service gas decay tank using
'WD-28A until vent header pressure has been reduced to 1 psig.
AI-110 is operable and has been sampling the in service gas decay tank for the past 20 minutes. All prerequisites are' met. START AI-110
- 1. Verify VCT gas WD-242 and WD-1080 indicate sample is secured. Closed.
- 2. Ensure that the Vent CUE: The vent header in Header is drained.- room 13 has been drained.
Room 15
- 3. Ensure that the gas verity DW-156 is open compressor is primed and verify water level in-WD-28A is visible but below centerline.
- 4.Punp the vent header AI-100 WD-28A control switch to start, red light on
i Rovision: 0 Jan. 24,1997 Fort Calhoun Station - Operations Training d JOE PERFORMANCE MEASURE JPM tio.: JPM-WDG JPM Title Waste gas transfer from the vent header to the i gas decay tank. ,
- 1. !
n , ammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmans
- CRITICAL ELEMENT STANDARD i I
STEP I sammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmen. \ ^ I AI-110
- 5. Monitor the CUE: Ha m 0.1%
following: 03 m 0.0% a . !!2 and 02 AI-100 conentrations. WDGT = 65-PSIG ,
- b. WGDT pressuro VENT HDR = 1.0 PSIG a
- c. Vent Hdr Pressure- ,
k t
- 6. Secure gas transfer AI-100 f WD-28A control switch to-stop and I-- green light on CUE: STOP
- Termination Criteria: Vent header pressure has been reduced to 1.0 ;
psig. i
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Revision: 0
-Jan. 24,1997' -
]
Fort Calhoun Station - Operations Training I JOB PERFORMANCE - MEASURE i
- JTH !!o, JIH-WDG JiH Titles Waste gas transfer from the Vent heador to the gas decay-' tank. l t
\ i i QUESTION: JPM-WDG-Q-1 What actions,if any, would be required if AI-110 was . 4 inoperable during this evolution? , e I ANSWERt The chemists Would collect grab samples of the in-service gas decay tank. ; NRC K/A: 071-000 K4.06 (SRO-3.5) ~ EEFERENCE: The operator's response to this question was: - SATISFACTORY UNSATISFACTORY COMMENTS: I r f b
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i Revision: 0 Jan. 24,1997 Fort Calhoun Station - Operations Training JOB PERFORMANCE NEASURE JPM No.: JPM-WDG JPM
Title:
Waste gas transfer from the vent header to the gas decay tank. QUESTION: JPH-WDG-Q-2 Why is there a caution MQT to drain the vent header while collecting a-VCT gas sample?
-AMsWER: The vr c header will contain the same gas mixture _as the VCT anu if that concentration was allowed into the drain header it would contaminate the Asxiliary Building.
(concept)
REFERENCE:
tmC K/A: 071-000-A4.29 (3.0/3.6) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
FORT CAUiOUN STATION OI WDG 1 OPERATING INSTRUCTION g PAGE 5 OF 38 Reference Use Attachment 2 Removing Wasto Gas Decay Tank, WD 29A/B/C/D, from Service PROCEDURE (Continued) - Vi JNIL
- 4. Notify the following plant personnel of the isolated WGDT data:
- Shift Supervisor e Shift Chemist
- 5. GO TO Attachment 1, Placing Waste Gas Decay Tank in Service.
i Completec' by Daternme / R18
FORT CALHOUN STATION Ol WDG 1 l OPERATilJG INSTRUCTIOf1 g M 6 & 38 Reference Use i Attachment 3 Waste Gas Transfer from the Vent Header to a Waste Gas Decay Tank ff.EHEOUISITES hd JML
- 1. Procedure Revision Vanfication Master Revision Number Date:
- 2. A Waste Gas Decay Tank (WGDT) is in service per Attachment 1.
- 3. Technical Specification 2.9 has been reviewed.
- 4. Waste Gas Analyzer Panel (Al 110) is in service per Attachment 4 to monur WGDT H, and O, concentrations OR Chemistry has been notif.ed to obta!n greb samples per Technical Specification 2.9.
- 5. Uquid Waste System is in service per OI WDL 1,
- 6. Deaerated/ Demineralized Water System is in service per Ol DW-4.
- 7. Nttrogen Gas System is in service per Ol NG 1. _
ff0CEDURE
- 1. IF it is cesired to pump the vent header below 1.0 psig, THEN perform Steps 1 through 4 of Attachment 7.
CAUTION To prevent an uncontrolled release of radioactive gas to the Auxillary Building and a possible Ventilation isolation Actuation Signal (VIAS), the Waste Gas Vent Header shall NOT be drained while a VCT Gas Sample is being drawn.
- 2. Ensure the following VCT Gas Space Sampling valves are closed (Al 110):
- WD 242
- WD 1080 R18
FORT CALHOUN STATION Ol WDG 1 OPERAT1NG INSTRUCTION g PAGE 7 OF 38 Reference Use Attachment 3 i Waste Gas Transfer from the Vent Header to a Waste Gas Decay Tank q PROCEDURE (Continued) ML E
- 3. Verify the Waste Gas Vent Header is vold of water by performing the '
following steps (Room 13):
- a. Slowly open WD 948, Vent Header L/G Inlet Valve.
- b. Open WD-949, Vent Header L/G Outlet Valve.
- c. IF liquid is visible in LG-608, THEN open WD 747, Vent Header L/G Drain Valve.
- d. WHEN the Vent Header is drained, THEN close the following Vent Header L/G isolation Valves:
- WD 747 -
e WD 949 e WD 948 R18
4 FORT cal.HOUN STATION Ol WDG-1:
'OPERpTING INSTRUCTION l PAGE 8 OF 38'. .i R
Reference Use
.' Attachment 3 Waste Gas Transfer from the Vent Header to a War'ie Gas Decay Tank 4
fBQCEDURE (Continued) .td. JglT,, , CAUTION - ] To provent pump rotor fsilure, a Waste Gas Compressor must NOT be toersted until seal water is aligned and proper Moisture Separator.- Tw.k level is established. 4 Ensure the selected Waste Gas Compressor is operable and primed by
- performing the following steps (Room 16)
J
- a. Ensure the Seal Water Isolation Valve to the Waste Gas
. Compre:Mer Skid is open:
- WD-28A, DW-156 e WD 28B, DW-157 -
- b. Verify Moisture Separator Tank level, relative to the pump casing, L is above the casing bottom AND below the pump rotor centerlir?
(CJ: e- WD 28A, Visible belc y C t e WD 288, Visible below Ct
- c. IF Moisture Separator Tank is at or above the pump rotor centerline, THEN drain the Moisture Separator Tank unth level is just below the pump rotor centerline: 1 e WD 28A, WD 216 e WD-288, WD-217 m
n . R18
.~ . _, _ _ _ . __ . . - _ _- , _ , __ __ .. _
-' FORT CALHOUN STATION OI WDG 1 t PAGE 9 OF 38 I LOPERATING INSTRUCTION g
= Reference Use Attachment 3 ,
Waste Gas Transfer from the Vent Header to a Waste Gas Decay Tank - l PROCEDURE (Continued) M JNIL 1: i CAUTlDH H, AND O, Concentrations in the WGDT must NOT exceed 3%. l
. 5. = Commence Waste Gas Transfer by performing the following steps: ,
ai IF Al-110 is operable, THEN ve.ify the in Service WGDT has been sampled for at least
- 15 mint es.
. b. Start the selected Waste Gas Compressor: o - WD-28A e WD-28B
- c. MonMor the following parameters: ,
e WGDT H, AND O2 Concentrations e WGDT Pressure ____ e Vent Header Pressure i e { d 4 4
+
R18
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FORT CALHOUN STATION OI WDG 1 OPERATING INSTRUCTION PAGE 10 OF 38 R Reference Use Attachment 3 Waste Gas Transfer from the Vent Header to a Waste Gas Decay Tank PROCEDURE (Continued) M) JML NOTE HI H, AND O, CONCENTRATION Annunciator (A-50, B-4) will be IN ALARM (Al.100) AND the white lights for HI H, and F'l O2 (Al-110) will be ON, when H, AND O, Concentrations exce3d 3%
- 6. IF H, AND O, Concentrations exceed 3%,
THEN secure the Waste Gas Transfer by performing the following steps:
- a. Stop the selected Waste Gas Compressor,
- b. Notify the Shift Supervisor,
- c. Notify the Shift Chemist.
- d. GO TO Attachment 5 and purge the in Service WGDT with Nitrogen.
1101F. Waste Gas Compressors automatically shutdown on a Low Suction Pressure of approximately 0.5 psig. CAUIlQM The Vent Header 9 ould not be pumped down beyond atmospheric pressure. Placing a vacuum on the vent header could cause damage to the system and/or system components. This header was not designed to be under a vacuum.
- 7. WHEN the Vent Header pressure is approximately 1.0 psig (PI 526),
in Service WGDT pressure reaches full capacity (95-100 psig), OR Degassing Operations are completed, OR Vent Header is at desired pressure below 1.0 psig (Pl-526), THEN stop the selected Waste Gas Compressor:
- WD-28A
- WD-288 v R18
- FORT CALHOUN STATION UI.wou 1 ;
OPERATING INSTRUCTION g PAGE 11 OF 38 Reference Use E Attachment 3 [ Waste. Gas Transfer from the Vent Header to a Waste Gas Decay Tank . j 4 PROCEDURE .-(Continued) V) .[NLL ,
- 8. IF Al-110 is operable,
' THEN 5 minutes after the Waste Gas Transfer is secured, a.~ -Record the H,'AND O, Concentrations from the analyzers:
-e YlA-628 -
H, mbar divided by 10 = H, % e= YlA 627 O, %
- b. Secure Al 110 per Attachment 4.
- c. Transfer the H2 AND 0, Concentration readings to the following:
e FC-143, Auxiliary Building Log o OP ST SHIFT-0001 I 9. IF Al 110 is inoperable,-
- - THEN notify the following
- Shift Chemist to obtain grab samples.
e Shift Supervisor of 24 hour LCO per Technical Specification 2.9. NOTE Steps 10 and 11 may be completed concurrently or in any order. 10.- IF Attachment 7 was performed to pump Vent Header below 1.0 psig, c THEN GO_TO Attachment 7, Step 5. -
- 11. = IF WGDT is full (95-100 psig),-
WEN GO TO Attachment 2.
/
,w.
Completed by Date/ Time
. R18
FORT CALHOUN STATION Ol-WDG 1 OPERATING INSTRUCTION g PAGC 12 OF 38 Reference Use [ Attachment 4]-
)
Sampling Waste Gas Decay Tanks Using Al 110 l EREREQUISITES M E
- 1. Procedure Revision Verification l Master Revision Number Date: ,
- 2. AC Power is available to Panel Al-110.
- 3. Instrumentation and alarms on Panel Al 110 are operable.
- 4. A Spent Regen Tank is in service per OI WDL-1.
NOTE OI WDG-3-CL-A, OI WDG-3-CL-8, and Ol WDG-3-CL-C, are maintained in OP-1 File after the first performance of OI WDG-3 following a Refueling Outage.
- 5. IF not in OP-1 File for this Fuel Cycle, THEN complete the following checklists AND place in OP-1 File:
- OI-WDG 3-CL-A e Ol WDG-3-CL-B e OI-WDG-3-CL-C l
l
. R18
115:321 JgnxipJsr3ti_ Form ES 1013 Simulation Facility: EorLQt!hourt Scenario No ; SIM 971 Euminerv Applicants: initial Conditions : Moabpower Turnover: FW-54, AC 3A tagged out Etent Malf. Esent Egent Description No. N o, Typc* 1 MAL AFW5C 1 Inadvertant AFAS 2 XMT RCS96 i Controlling ZR p pressure chamiel fails high 3 MAL SGNI A C 40 gpm S/G Tube leak'( A S/G) 1.3% 4 VI,V MISS C llCV 978 does not isolate (preset) 5 R/N Emergency Shutdown 6 !!ST CND9 i Condenser vacuum switch 952A indicates low vacuum (high pressure) 7 MAL SGNI A M S/G Tube Rupture (10%) Shaded entries are to be initiated by a cue from examiner.
* (N)ormal, (R)cactivity, (1)nstrument, (C)omponent, (M)ajor Examincr:
Chief Examiner:
ES 301 - __ Operator Actions _ Form _F,S 301-4 Scenario No : FCS-97-1 Event No.: 1 Eage1 of Z Event
Description:
Inadvertant AFAS actuation 1ime Position Applicant's Actions or Behavior RO/ BOP Identify AFAS actuation from alarms SRO Enter AOP-28 4 SRO Direct BOP to isolate AFW to the Steam Generators BOP Close HCVs-1107A,8 1108A,B SRO Direct BOP to contact EONT to bypass affected AFAS channel SRO Direct BOP to secure AFW pumps BOP Close YCV-1045,1045A,1045B(override /close) and place FW-6 switch in pull-to-lock SRL Refer to tech specs 2.5 and 2.15 SRO Determine plar.t is in a 6 hour LCO with no c aerable AFW pumps, an 8 hour LCO to bypass the a"fected channel and a 48 hour LCO with an inoperable AFAS channel. SRO When the affected channel is bypassed, direct the BOP to return the AFAS system to automatic BOP Place the control switches for HCVs-1107A/B,1108 Alt; and YCV-1045 to the RESET then to the AUTO position. Place the FW-6 control s.vitch in after stop and place the YCV-1045A/B override switches to normal.
- . . . . . . _ . - _ - =
ES:30L OperatorActioc_s FomLES-301-4 Scenario No.: FCS-97-1 Event No.: 2 P_ age 2 of Z Event
Description:
Controlling pressurizer pressure channel fails high Time Position Applicant's Actions or Behavior RO Identify and report deviation between pressurizer pressure channels RO identify and report high indication on controlling channel and lowering pressure on other channel SRO Direct RO to transfer control to other channel or take manual control of pressurizer pressure RO Transfer control channels or take manual control as directed _ RO Monitor and maintain proper pressurizer pressure BOP Monitor and control secondary parameters 4 .p
. - . .- .-... - - . - . . . . .. -. - .,. ._.. - - ... - - . _ - - - - - ....- ~ .. - . - - .
Y i JS-301 OpeLatotAgtions i Form ES-30_1-4 4. Scenario No.: FCS 97-1. Event No.: 3 -- Page 3 of Z_ 1 4 Event
Description:
Steam generator tube leak Time Position Applicant's Actions or Behavior RO Identify and report charging / letdown mismatch RO identify and report condenser off-gas radiation alarm (RM-057) , SRO Enter AOP-22 SRO Direct RO to control pressurizer level RO Control pressurizer level i BOP Determine and report that RCV-978 (supply to aux steam) did not close. (Event 4) SRO initiate Emergency Shutdown (AOP-5) (event 5) I SRO Identify affected steam generator (A) I -SRO Direct RO or BOP to place steamline radiation monitor in
- service i RO or BOP Place steamline radiation monitor in service ,
SRO- Direct RO or BOP to have EONT swap blowdown sample flow to waste RO or BOP Direct EONT to swap blowdown sample flow to waste SRO May direct BOP to place YCV-1045A in_ override and - close i
- BOP
- If directed, place YCV-1045A in override and close d-k:
4 4 4
- , ,, , . , ~ , , , , , - . . . , , . , , . . - , , , ,--.----c , . . , . - , , - . . - . , . - , - ,
3 ES101 OperatotActions FoLm_E.S-3.01 -4 . Scenario No.: FCS 97-1 Event No.: 4 Page 4 of Z Event
Description:
RCV-978 (Supply to Aux Steam) does not close Time Position Applicant's Actions or Behavior BOP Determine and report that RCV-978 did not close SRO Direct BOP to close RCV-978 BOP Take action to close RCV-978 00P Determine and report that RCV-978 did not close SRO Direct EONT to close RCV-978
.m T
e ~ - _ . - .
ES90t_.______,__. Operator Actjons _. __Egrm_ES-30H Scenario No : FCS 97-1 Event No.: 5 Page 5 of Z Event
Description:
Emergency Shutdown Time Position Applicant's Actions or Behavior SRO Enter AOP-05 (Emergency Shutdown)- Direct Emergency Shutdown _ SRO Notify System Operations of Power Decrease SRO Direct RO to begin boration using SIRWT RO Switch charging pump suction from the VCT to the SIRWT SRO Direct BOP to control RCS cold leg temperature by reducing turbine load BOP Reduco turbine load to control cold leg temperature SRO Direct RO to operate control rods to control ASI RO Operate Control Rods to control ASI RO Monitor and control primary parameters BOP Monitor and control RCS cold leg temperature and secondary parameters SRO Continue to coordinate RO and BOP actions during power reduction
ES-301 - .__ Operator. Actions ____Ecrm_ E S-301 -4
. Scenario No.: FCS-97-1 Event No : 6 Eage 6 of I l
Event
Description:
Condenser vacuum switch 952A fails to prevent operation of steam i dump and bypass valves l Time Position Applicant's Actions or Behavior BOP Determine that RCS temperatures are high following reactor inp BOP identify and report failure of steam dump and bypass valves to open following trip SRO Direct BOP to control RCS temperature using HCV-1040, or MS-291 and 292 BOP Control RCS tem aerature using HCV-1040 and/or MS-291 and 292 as cirected by the SRO BOP Do not open MS-291 following isolation of steam generator A t f
- - . - - . , . - - - ,s, 4 - - --
ES-301_ _ _ _ __ Operator Actions _ _ _ _ _ _ Eorm _E S-301-4 Scenaiio No.: FCS-97-1 Event Noc 7 Eage Z of Z Event
Description:
Tube Rupture - Steam Generator A [ Time Position Applicant's Actions or Behavior g RO identify and report RCS inventory loss SRO May direct reactor trip SRO Following manual or auto reactor trip, direct standard } post trip actions RO Perform primary standard post trip actions BOP Perform secondary standard post trip actions SRO Diagnose tube rupture - enter EOP-04 or EOP-20 L SRO Direct RCS cooldown - Tw less than 510 F BOP Cooldown RCS Tw to less than 510oF RO Identify and verify PPLS SRO/ BOP Identify most affected steam generator (A) SRO Direct BOP to isolate steam generater A BOP isolate steam generator A SRO Direct RO to depressurize RCS to less than 1000 psia RO Depressurize the RCS RO Maintain subcooling BOP Monitor and control secondary parameters RO Monitor and control primary parameters E Scenario ends when steam generator A is isolated and RCS pressure is less than 1000 psia. mm i sieimum-- -- -
Revision:3 January 27,1997 Fort Calhoun Station - Operations Training JOB' PERFORMANCE MEASURE JPM No.: VPM-0306A (OLD'No N/A) JPM
Title:
Alternating and Securing Battery chargers-System (s): Electrical Distribut' ion Lrcation (s) : Switchgear Room Approximate Time: 15 minutes Actual Time: Reference (s):- (1) OI-EE-3 (R-10) Attachment 3 1.0 Attachment 2 2.0 (2) NRC K/A 000062,K1.03 (RO3.5/SRO4.0) 1 (3) NRC K/A 000063,K3.02 (RO3.5/SRO3.7) Verify current reference revisions match those listed above Operator's Name: SS N: All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. t The operator's performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluator's Signature: Date: Reason, if unsatisfactory: Operator's. reviewed: ._ Date:
~g- -
s t--r
- Initiating Cues .
'You are directed to place #3 Battery Charger in service and
. secure #2-Battery Charger. START.-
3 l
+
i 1 1 e h I
-er- .,--,e -
-n y,,-, < a , ,,, -
Revision:3 January 27,1997 Fort Calhoun Station - Operations Training JOB PERFORMANCE HEASURE J PM !!o . : JPM-0306A (OLD 110: !!/A) JPM
Title:
Alternating and Securing Battery Chargers Tools & Equipment: Keys to rollup door in Switchgear Room. Safety Considerations: Improper alignment can result in a degraded DC Distribution System. Comments: Starting a battery charger is covered under a separate JPM. THIS JPM WILL BE PERFORMED AS A STATIC JPM IN THE PLANT.
.. . . . - . . - . . . .- . . - - . . .- - . _-. - - - - - . ~ . . - - - - ..- _. .~ . .-
Revision: 3 January 27,1997-f Fort-Calhoun Station - Operations Training , JOB. PERFORMANCE MEASURE , r JPM-No.: JPM-0306A (OLD No: N/A) JPM
Title:
Alternating and Securing Battery Chargers-- p .t
==================================================================
- CRITICAL ELEMENT ,
STANDARD
- STEP
=======================================================rs========== .;
i-Initiating Cue: You are directed to place #3 Battery Charger in service and secure #2 Batter'/ Charger. START. . E
- 1. Place Charger No. 2 EE-8D in FLOAT. SW-2 to FLOAT.
t
- 2. Check CB-1 OPEN EE-8E Check CB-2 OPEN Check Breakers in OPEN position.
- 3. Place Charger No. 3 EE-8E in FLOAT. SW-2 to FLOAT.
- 4. Check Charger No. 3 MCC-3C1 AC supply Closed. Breaker closed ?
- 5. Place the Charger EE-8E Alarm Normal / Inhibit Place Switch to NORMAL. position.
Switch in the NORMAL position.
,,.g -o = y ,.. - vw,-y- _- y . ,,w , y -,
l Revision 3-January 27,1997: Fort Calhoun Station - Operations Training JOE PERFORMANCE NEASURE JPM No.: JPM-0306A (OLD Not N/A) JPM
Title:
Alternating and Securing Eattery Chargers i
==================================================================
- CRITICAL ELEMENT STANDARD STEP
==================================================================
4
- 6. Close AC input to EE-8E Charger No. 3. Close CB-1.
Cue #3 Charger is at 130 VDC.
- 7. Close DC output from EE-8E
- Charger No. 3. Close CB-2.
- 8. Connect Charger No. EE-8G 3 to Bus. Close DC3-2 EE-8G
- 9. Remove Charger No. 2 Open DC2-1.
from Bus. EE-8D
- 10.Open DC output from Open CB-2.
Charger No. 2. EE-8D
- 11.Open AC Input to Open CB-1.
Charger No. 2.
...n ._ _ - . - ,
. .. . - . . . , . . _ . . . - - - -- . - ~ . - - - . . - . . . . - .
- Revision:3 ,
- January 27,1997- ,
Fort Calhoun Station --Operations Training ! JOB PERFORMANCE MEASURE -)
- JPM No. JPM-0306A- (OLD No:-N/A)
JPM - Title : Alternating and Securing Battery Chargers
==================================================================
- CRITICAL EtwnT - STANDARD STEP
==================================================================
1
- 12. Place Charger .
EE-8D Normal / Inhibit Switch in Place Switch to INHIBIT position. i the INHIBIT position.
- Termination Criteria: #3 Charger in service on #2 DC Bus and #2 Charger is secured.
4
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, .m_. m. ,,_ ,
. ~ . . __ __
Revision:3 i January 27,1997 > Fort Calhoun Station - Operations Training JOB PERFORMANCE MEASURE JPM No.: JPM-0306A (OLD Not N/A) JPM
Title:
Alternating and Securing sattery Chargers-QUESTION: JPM-0306-Q-2 What safety precaution must be taken when a switch
. gear roll-up door is opened?
ANSWER: A continuous fire watch must be established and it must be logged in the Control Room Log.
REFERENCE:
NRC K/A: 063-000-GEN 1 (3.1/3.2) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
Revision:3 January ~27,1997 Fort Calhoun Station - Operations Training JOE PERFORMANCE - MEASURE - JPM No.: JPM-0306A (OLD No: 11/A) JPM
Title:
Alternating and Securing Battery Chargers
- QUEGTION: JPM-0306-0-3(NEW)
!!ow would the consequences differ if the voltage regulator on an in service battery charger failed high or low?
ANSWER: (Both would result in a charger trouble alarm, DC bus low voltage alarm, and the battery supplying the DC bus.) The main difference being that the charger will trip on high voltage but not on low voltage.
REFERENCE:
NRC K/A: 063-000-Kl.03 (2.9/3.5) 063-000-K4.02 (2.9/3.2) The operator's response to this question was: SATISFACTORY UNSATISFACTORY COMMENTS:
FORT CALHOUN STATION OI-EE-3 OPERATING INSTRUCTION PAGE 5 OF 20 e Con 6nuous Use Attachment 2 - Battery Charger No. 2 Operation PREREQUISITES ((} INIT.
- 1. Procedure Revision Verification Master Revision Number Date:
- 2. Panel CB-20 Annunciator A15-85, BATTER CHARGER # 2 TROUBLE alarm is clear.
- 3. MCC 4A1 is in service PER Ol EE 2.
- 4. Checkilst OI-EE-3-CL-A has been completed as required by OP-1.
- 5. Switchgear control power feeders are in their normal position.
PROCEDURE
- 1. IF placing EE-8D,125V DC Battery Charger Number 2 in service, THEN perform the following;
- a. Ensure EE-8E,125V DC Battery Charger Number 3 Voltage Selector is in FLOAT.
- b. Ensure the following Battery Charger Number 2 breakers are off:
- EE 8D-CB1 Battery Charger Number 2 AC Input Breaker e EE-8D-CB2 Battery Charger Number 2 DC Output Breaker
- c. Ensure Battery Charger No. 2 Voltage Selector is in FLOAT.
- d. Ensure MCC-4A1-C02, EE-8D Battery Charger Number 2 is on.
- e. At Battery Charger Number 2 place 69/EE-8D, Alarm Permissive in NOFsMAL NOTE Charger No. 2 DC output voltage should be within 0.5 volts of the 125 VDC Bus No. 2.
- f. Place EE-8D-CB1 in on.
- g. Place EE-8D-CB2 in on.
. h. At EE-8G.125V DC Number 2 Main Distribution Panel, Ensure EE-8G-CB1, Batt Charger 2, EE-D is on.
R12
4T CALHOUN STATION ) RATING INSTRUCTION [ c 4
.J
[ continuous use j i Attachment 2 - Battery Charger No. 2 Operation M.) _INIT _ LOCEDURE i nected to i. IF EE-8G,125V DC Number 2 Main Distribution h nt 3. Panel s con Battery Charger No. 3,THEN remove Battery Charger No. 3 from se . NOTE dj sting Charger No. 2 DC ammeter must be less than 380 amps while a u voltage No. 2 DC l) ~ J. If necessary, notify Electrical Maintenance EM to adj 2 from service, 2. IF removing EE-8D,125V DC Battery Charger Number _ THEN perform the following: i NOTE l b fore ! g Battery Charger No. 3 should be placed in service, if possib e, e removing Charger No. 2 from service ilits DC
- c. IF Electrical Maintenance is available,THEN EM lower Battery Ammeter stops decreasing (unloaded).
At EE-8G,125V DC Number 2 Main Distribution Panel place
- b. EE-8G-CB1, Batt. Charger 2, EE-8D, in off. _
k r in off. c. At EE-80 place EE-80-CB2, Batt Charger No. 2 DC 2 DC Output Brea e al).
- d. If necessary, notify Electrical Maintenance to a _EM i
)
R12 l
FORT CALHOUN STATION OI-EE-3 , OPERATING INSTRUCTION PAGE 7 OF 20 l c Continuous Use Attachment 2 - Battery Charger No. 2 Operation (/_) INIT. PROCEDURE (continued)
- 2. e. At EE-8D place EE-8D-CB1, Batt Charger 2 AC Input Breaker in off.
- f. Place 69/EE-8D, Alarm Permissive in INHIBIT.
- g. IF removing Battery Charger No. 2 for maintenance, THEN place MCC 4A1-C02, EE-8D Battery Charger Number 2 in off.
- 3. IF placing EE-8D,125V DC Battery Charger Number 2 in Equalize Mode, THEN perform the following;
- a. Verify Battery Charger No. 2 is in service.
- b. Rotate the equalizing voltage potentiometer counterclockwise (CCW) to its minimum voltage setting.
EM NOTE I When equalizing mode is selected, Annunciators A19-C1, DC BUS #2 LOW VOLTAGE and A15-BF DATTERY CHARGER #2 TROUBLE may alarm. The alarm (s) will clear when the equalizing voltage is adjusted.
- c. IF desired to use the equalizing Timer (up to 120 hours),
THEN set the Timer for the number of equalizing hours required.
- d. IF desired to use the equalizing switch, THEN place the Voltage Selector in EQUALIZE.
- e. Adjust the battery charger for a terminal voltage of 135.5 to 136.5 VDC (136.0 VDC is normal).
EM Completed by Date/ Time / R12
~~~~
m,
,~ " m o vi ;
0 RT CALHOUN STATION 3 c JERATING INSTRUCTION
-t Continuous Use Attachment 3 - Battery Charger No. 3 Operation (4) _ INIT.
LflEREQUISITES Procedure Revision Verification Date: Manter Revision Number. 2. P nel CB-20 Annunciator A15-CS, BATTER CHARGER # 3 TROUBLE alarrn is clear.
- 3. MCC 3C1 is in service PER OI EE-2.
4. Checklist OI-EE-3-CL-A has been completed as required by OP-1. _ 5. Switchgear control power feeders are in their normal position. PROCEDURE 1. IF placing EE-8E,125V DC Battery Charger Number 3 in service, THEN perform the following; ' c. IF EE-8F,125V DC Number 1 Main Distribution Panel, Breaker ; EE-8F-CB3, is on, THEN ensure EE-8C,125V DC Battery Charger Number i Voltage Selector is in FLOAT. b. IF EE-8G,125V DC Number 2 Main Distribution Panel, Breaker EE-8G-CB1, is on, THEN ensure EE-BD,125V DC Battery Charger Number 2 Voltage Selector is in FLOAT.
- c. Complete the Following:
Ensure the following Battery Charger Number 3 breakers are off:
- 1) _
- EE-8E-CB1 Batt Charger 3 AC Input Breaker e EE-8E-CB2 Batt Charger 3 DC Output Breaker Ensure Battery Charger Number 3 Voltage Selector is in FLOAT. ___
2) Ensure MCC-3C1-A2L, EE-8E Battery Charger No. 3 is on. 3) At Battery Charger Number 3 place 69/EE BE, Alarm Permissive 4) in NORMAL I
- 5) Place EE-8E-CB1 in on. _
- 6) Place EE-8E-CB2 in on. R12 ,
l
FORT CALHOUN STATION Ol-EE-3 OPERATING INSTRUCTION PAGE 9 OF 20 c Continuous Use Attachment 3 - Battery Charger No. 3 Operation (2) INIT. PROCEDURE NOTE Charger No. 3 DC output voltage should be 'vithin 0.5 volts of the - 125 VDC Bus to which it is being connected.
- 1. d. IF connecting EE-8E Battery Charger Number 3 to EE-8F,125V DC Main Distribution Panel NO.1, THEN perform the following:
NOTE Breaker EE-8F-CB2 operation will recuire using a Kirk Key.
- 1) At EE-8F place Breaker EE-8F-CB2 in on.
- 2) IF Bus EE-8F is connected to Battery Charger No.1.
THEN remove Battery Charger No.1 from service PER Attachment 1.
- e. IF connecting EE-8E Battery Charger Number 3 to EE-8G,125V DC Main Distribution Panel No. 2, THEN perform the following:
NOTE Breaker EE-8G-CB2 operation wf!! require using a Kirk Key.
- 1) At EE-8G place Breaker EE-8G-CB2 in on.
- 2) IF Bus EE-8G is connected to Battery Charger No. 2, THEN remove Battery Charger No. 2 from service PER Attachment 2. NOTE Charger No. 3 ammeter must be less than 380 amps while adjusting voltage _
- f. If necessary, notify Electrical Maintenance to adjust Charger No. 3 DC Output Voltage to maintain 130.0 (129.5-130.5 VDC.
EM R12
e. (OUN STATION G INSTRUCTION continuous Use No. 3 Operction INIT . {L} Attachment 3 Battery Charger fJBE ger Number 3 from service, V DC Main cmoving EE-8E, Battery Char 3 is connected to EE (EN perform the fo llowing: i service before IF Battery Charger NumberDistribution D C M ain Panel No.1, THEN,if possible,placercmoving Battery Charger Num IF Battery Charger Number 3 s Charper No. 2in service before A Distribution Panel No. 2,THEN,if possible, place Batt removing Battery Charger ilablo, Voltage untilits DC Ammeter EM tput
- c. IF ElectricalMaintenanceis avaTHEN lower Ch stops decreasing (unloaded). .
ff
. d. At EE-8F ensureff EE-8F-CB2is o .
- o. At EE-8G ensure EE-8G-CB2is3DC ol) . EE-8E-CB2 i At Battery Charger Number 3 place Maintenance to adjustEMCharger N f.
- g. If necessary, notify ElectricalOutput Voltage t l ce EE-8E-CB1 in off.
At Battery Charger Number 3pa in INHIBIT. h. Place 69/EE-8E, er Alarm NumberPermissive 3 for maintenance, ber 3 in off,
. i. 8E Battery Charger Num J. -IF removing Batter ChargTHEN place MCC-3C1-A t
4 R12
- f3 del _ SEt Mtfj(Lib rpts _ Form 1i54013 Simulation l'acility: EortCalhourt Scenario No : SIM 97 2 Examiners' Applicants:
Initial Conditions : 100% power Tunun cr: FW 54. AC-3 A tagged out Esent M alf, Esent Esent Description No. No. T3pc* 1 MAL CRD6 C Dropped rod (rod #1) 2 R/N Reduce Powcr to 70% ~3 XMT CVC23 i T 29M7 (letdown llX CCW outict temperature)t' ails low 4 XMT SGN27 i S/G level LT-903X fails high 5 MAL EDSID C 1 A4 bus' fault /rx tnp 6 MAL EDSI til M loss of of fsite powcr Iboth 161 and 345 KV) MAL EDSil A (preset on trip) 7 LOA EDS7i C RCP RC-3C breaker does not open (D/G output breaker will not close) (preset) Shaded entries are to be initiated by a cue from examinct.
* (N)ormal, (R)cactivity, (1)nstrument, (C)omponent. (M)ajor Examiner.
Chief Examiner
ES 301 Op_ptator Actions FomLES 3_Old
*m ja Scenario No.: P '2 -
Event No.; 1 Page 1 of 7 Event
Description:
Dropped Control Rod r-Time f'osition Applicant'c Actions or Behavior RO Identify event from alarme RO Determine only one rod has dropped SRO Enter AOP-02 (CEDM) Malfunction) GRO Direct BOP to adjust turbine load to match reactor power BOP Adjust turbine load to match reactor power SRO Direct RO to control pressurizer pressure and level RO Monitor Pressurizer pressure and level SRO Notify Reactor Engineer SRO Consu'l Tech Sec 2.10.2. (Note: Requirements of this Tech Spec are covered in the actions required by AOP-02
.O Inform RO and BOP that Tech Specs require a power reduction to less than 70% within one hour SRO Notify system Operations of impending po_ wor reduction J
v . -- ~ . - - . , .-. - . ,_
liS 3QL._, Operatotections ._ ForrTLES-301d i Sconario No.: FCS 97 2 Event No.:2 Page 2 of 7 Event Dosenption: Power Roduction to 70% within ono hour Timo Position Applicant's Actions or Behavior SRO Direct RO and BOP to commence power reduction SRO Direct RO on method of boration to use. (Options are normal boration, shifting charging pump suction to the SIRWT, or entering AOP-051morgency Sn ,tdown) RO. Begin boration BOP Reduce turbino load to control RCS Tc. RO Monitor and control primary parameters during power reduction . BOP Monitor and control secondary parameters during power reduction. 4
-e 0 -- .- , - , - - - . - -,. - - + . --r,.,. ., n ,,
l 1 l fiS-30L_ Operatot Actions formfSa0B Scenario No.: FCS 97 2 Event No.:3 Page 3 of 7 Event Desenption: T-2987 (letdown HX CCW Outlet temp) fails low Time Position Applicant's Actions or Behavior RO identify high letdown temperature condition from alarms RO Venfy that TCV 211-2 has repositioned to bypass demineralizers RO Determine high temperature due to reduced CCW flow to letdown heat exchanger SRO Direct RO to manually control CCW flow to letdown HX RO Manually control CCW flow to restore letdown temperature SRO May direct RO to reposition TCV 211-2 RO Reposition TCV 211-2 if directed RO Monitor primary parameters BOP Monitor Secondary Parameters We
t E S-301_. - - - _ Operator. Actions _ Formes 101-4 Scenario No.: FCS-97-2 Event No.:4 Page 4 of 7 4 Event
Description:
Steam Generator Level (LT-903X) Fails Hip ,
' h Faseru 'ades.w Time Positior. Applicant's Actions or Behavior BOP Determine and report failure of steam generator level channel SRO Direct BOP to take manual control of steam generator level BOP Take manual control of FW Reg Valve. Restore normal level BOP Monitor and control steam generator level RO Monitor primary parameters -r,- - , -- --e - - - - - .+-.y- ,v-- ----
. __ ._. - - - _ - . . . _ - = - - _ - -. - . _ _ . . _ _ _ - . - -
P ES. 301 . _0poratorAct!ons Fonp_ES 301-4 ; Sconario No.: FCS 97 2 Event No.:5 Page 5 of 7 Event Description : 1 A4 bus fault / Reactor trip, loss of offsite power Timo Position Applicant's Actions or Behavior SRO Direct standard Post-trip actions BOP Report No Power to buses 1 A3 and 1 A4 RO or BOP _ Report both diesels running at 900 RPM RO or BOP Report that D/G breakers did not close SRO Ensure that foodwater be restored BOP Rostore feedwater using FW 10 or FW-6 (if power restored to bus 1 A3) , t RO Perform remainder of SPTA's , BOP Perform remainder of SPTA's
, SRO Venfy com;.!etion of SPTA's w
ES:30L._ ___ _ Operator Actpos _ ForrrtES-30j-4 Scenario No : FCS 97-2 Event No.6: Page 6 of 7 Event Descriptiort RCP RC 3C Breaker does not open Time Position Applicant's Actio".s or Behavior RO identify failure of breaker to open RO Report failure of breaker to open SRO Direct RO to direct EONT to manually trip breaker RO Direct EONT to manually trip breaker EONT Open bre_aker SRO Direct RO/ BOP to verify restoration of power to bus 1 A3 RO/ BOP Verify power to bus 1 A3 _ b
ES-301_ _ _ DperatorActtoos Form fiS 30j-4 Scenario No. FCS 97-2 Event No.:7 Page 7 of 7 Event Desenption: Loss of offsite power Time Position Applicant's Actions or Behavior SRO Diagnoso event and enter EOP-7 (if power has not been restored to bus 1 A3 or EOP-02 (if power has been restored to bus 1 A3 SRO Direct RO to venfy Natural Circulation RO Venfy Natural Circulation SRO Direct RO to monitor and control primary parameters SRO Direct RO and BOP to ensure Raw water, CCW Containment Cooling, instrument Air and Charg,ng i Pumps are restored RO and BOP Ensure that all items listed above are restored. Scenario ends when power has been restored to bus I A3 and all safety functions are satisfied.
_ I35 ML .__.-__..SWMtiElh ynts l'eni1JiS-MM Sirnulation l'acility [pnfallstyL Scenario No.: Siht 97 3 lisaminers' __ __ Applicants-Initial Conditions : 100% pow ct Turnover. I W-4 A, DG.I tagged out
~
IMent St af f, 1:s ent I:sent Descfiption No. No. , 'I ,$ pc
- I AtAl. RCP9D C RC 3C lowcr seal falls 2 Xhti RCS97 i Pressuri/cr levci channel 10lX falls low
~~
1 h1AL RCPinD C RC-1C ruiddic scal fails 4 R/N limergency Shutdown
~
5 XN1l' SGNio 1 S/G steam flow channel 1T-908 lalls high 6 h1AL htSSill h1 Stearn line break in containment 20% ramp over 1200 sec 7 htAL CRDSD C 2 stuck rods h1AL CRD511 (preset) M Ni AL liSI'2 A 1 CPilS f alls to actuate h1AL llSF2il (preset) Studed entries are to be loitiated by a cue from examiner.
* (N)onnat, (R)eactivity, (1)nstnunent, (C)omponent, (ht)ajor lisamincr:
Chief IIxamincr=
ESQQ1_ Oper.atoLActions _ FomtES&01-4 Scenario No.: FCS-97-3 Event No.: 1 Page 1 of 8 Eunt
Description:
RC-3C lower seal fails Time Position Applicant's Actions or Behavior RO Identify and communicate high seat leakage from alarms SRO Enter alarm response procedure RO Monitor RCP seal pressures and determine that the lower seal on RCP C has failed SRO Determine that operation can continue with one seal failure SRO Direct RO to continue to monitor RCP seals BOP Monitor secondary parameters
ES 301 , OperatorJctions - FormESiload Scenano No.: FCS 97-3 Event No.: 2 Page 2 of 8 Event Description' Pressurizer level channel 10iX fails low Time Position Applicant's Actions or Behavior RO identify and report failure of pressurizer levelinstrument SRO Direct RO to transfer control channels or take manual control of controlling channel RO Transfer control channel or take manual control of level RO Monitor and control pressurizer level SRO Direct LO to select Y channel on the low level heater cutout switch RO Select Y channel on low level heater cutout switch BOP Monitor secondary parameters 9 y w ,.w * , --
.g
ES-301 OperatoLActicas__. ForfrLES 3014 Scenario No FCS 97 3 Event No. 3 Page 3 of 8 Event Descripticn: RC-3C middle seat fails Time Position Applicant's Actions or Behavior RO Monitor RCP seal parameters RO identify and communicate additional seal failure SRO Enter alarm response procedure SRO Direct Emergency Shutdown and enter AOP-05 (Emergency Shutdown) BOP Monitor secondary parameters de
t l E S-301 ._ ._ ._ _ . _ __ _._. . Operator Actions __.. ._ ,__-. .____.Eorrn E S 301-4 Scenario No : FCS 97-3 Event No. 4 Page 4 of 8 Event
Description:
Emergency Shutdown Time Position Applicant's Actions or Behavior SRO Direct RO and BOP to commence Emergency Shutdown SRO Notify System Operations of power decreaso SRO_ Direct RO to begin boration urjng SIRWT RO Switch charging pump suction from VCT to SIRWT - SRO Direct BOP to control RCS T-cold by reducing turbine load BOP Reduce turbine load to centrol T-cold SRO Direct RO to operate control rods as required to control ASI RO Operate control rods to control ASI RO Monitor and control primary parameters BOP Monitor and control RCS T-cold and secondary parameters 1
,.a y ~yw-w- w-y- wa a - -
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y - c *i,- en--. w- e r < - - . , -------+--,--yw-----+---- --,w+--_-- - t
\
\
ES130j __Qperator. Actjpps_ Form _ES-301-4 Scenano No FCS 97-3
. Event No.: 5 Page 5 of 8 Event
Description:
Steam flow transmitter (FT 908) fails high Time Position Applicant's Actions or Behavior ; BOP Identify and communicate lowering FW flow and level in S_/G "B" SRO Direct BOP to take rnanual control of feedwater BOP Take manual control and restore fesowater level BOP Identify FT-908 as failed instrument , _ SRO Inform l&C of failure of FT-908 BOP Continue to monitor and control S/G levni RO Monitor primary parameters l
4 ES 301 . .
. _Qperator Actioris __._ . FormfS-Sold Scenano No; FCS 97-3 Event No. 6 Page 6 of 8 9
Event Description. Steam line break in containment , , Time Position Applicant's Actions or Behavior f BOP Identify and communicate lowering RCS T-cold , RO Identify and communicate lowering oressurizer pressure and level SRO May direct RO to manually trip the reactor RO If directed, trip the reactor SRO Direct the RO and BOP to perform standard post trip actions RO Perform primary post trip actions BOP Perform secondary post trip actions SRO Direct BOP to verify isolations following Steam Generator Isolation Signal BOP Verify SGIS actuation
- \
1 s
- ,.+ _-a+-.-.-+ge.n aw-.% ...w.-- w, . ,,.-%., ,,,% e,m,f, , , , , , ,,.,g- % -wq,- ry..-,w- ,w,p w ,-9.m,., s,.,%&9, , - ,y ee.y.- .. g.ymy yp ge_m-.e7 9 -t,.r etW r
- . _ . ~ - - _ . . - - _ . . _ - . . . . . .
ES 301 Operator Actioris ... EonnfS:391-4 Scenario No.: FCS 97-3 Event Noc 7 Page 7 of 8 ; Event
Description:
Two controt rods stick on trip Time Position Applicant's Actions or Behavior RO Identify and report failure of two control rods to insert following reactor trip SRO Direct RO to initiate emergency boration RO initiate Emergency boration SRO Direct BOP to manually throttle feedwater flow BOP Manually control feedwater flow __ SRO Diagnoso event and enter EOP-05 or EOP-20 l'
--,r-m_. - . - . - - , , - w, <- , . , - ...w.. . . . . , - y,-
re- . , , .._,e..,,y,. . , - ., . , , . 7.m-,,v,-,..
_ __l I ES:401 __OperatotAetions _ EpmLIIS:M-A 1 i Scenano No.: FCS 97-3 Event No 8 Page 8 of 8 i i Event
Description:
CPHS fails to actuate Time Position Applicant's Actions or Behavior RO Monitor containment pressure and determine that CPHS did not actuate at setpoint SRO Direct RO to rnanually actuate CPHS RO Manually actuato CPHS and venfy containment spray flow SRO Direct BOP to establish steam flow from intact steam generator prior to dryout of faulted steam generator BOP Establish steam flow from intact steam generator and control RCS temperature. SRO Direct RO to monitor subcooling and pressurizer level to determine when HPSI "stop and throttle" cnteria are met RO Monitor subcooling and when "stop and throttle" pressurizer cnteria are met level and report SRO Direct primary operator to throttle and/or stcp HPSI flow RO , Thr_ottle and/or stop HPSI flow SRO Direct RO to monitor and control RCS pressure to maintain subcooling between 20 and 200*F RO - Monitor and control pressure to maintain subcooling between 20* and 200*F i
; Scenario ends when "stop and throttle criteria have been met and HPSI flow has been reduced.
. . _ _ .- , -.. - - _ . . . . . _ _ _ - _ _ . _ . - - - . _ . . - . - . - _ . - . _ = _
i I
._ lierut(1001 1 !
_ _ 1400 t._ _._ __ _. .Junario 13 s oli Sittnil.ttlun I"atility . (M1Lillmtut SLclurio No : Slhi 97 4 lhaininers' _ _ _ _ _ _ _ _ _ _ APplicant$ _ _ _ _ _ . _ _ _ _ . _
.. . . _ . . _ . . _ . . - . . _,_ _ _ . _ _ . . _ . . , . . ~ . . . _ . _ _ . _ .
.- ~ . ~ . - . . - . . - . . _ . - . - . ~ - . -
initial Conditions . 100*e pow er
'lustun er- Pou rt r uige N1 clunnel T" tamul out, tilp units 1.9 and 12 bypassed. CPI ( in progress 13 cnt Malf, IA rni 13 rnt liricription Nn. N n._ 'l}ye
- I M Al,' NIS711 1 Ni pow or iange (hannci "ll" pow er supply f allutc 2 1l51 CVC11 1 VCl Ic5 el thannelit.CS 2lMt.l.1.)lails low (bistable trip)
.1 It/N l'ower reduction to 70"i, 4 XM I' ItitS t l Plivl0 lails high t itNN) psia)
$ 3WI tillCil)V M 1,oss of load ?
6 YLV l(CSlH C/M 14)l(V POV.lH21 lalls open C JitPSitXTP _ IPl.CSC'l. . 7 VI.V 110S17 C ..14)ltV block s ah c ilCV.151 ulli not close
~
8 l'il.li /. 4201 C 111'S1 Punips fall to start Shaded entries are to be inillaicd by a cue Iroin esaminer.
* (locactivity, (linstruinent, tC)oniponent. (Miajar (N)ormal.
lhaminct; _ _ _ Chief thaminer. 6
l ES:30L._. ._ ._ _,.__OperatoLActions_ _ _ __Ectrn_ES:30M l l
- l Scenario Noa FCS-97-4 Event No :1 Page 1 of 8 Event Dosenption Loss of NI power rango channel *0' Timo Position Applicant's Actions or Behavior f
=
RO identify the failure from alarms i SRO Referenco AOP-15 SRO Determine the nood to placo "B" channel trip units 1,9 & 12 in the tripped condition,1 hour LCO, and 48 hour LCO for repair of one channel SRO Direct the RO to pull the 1,9 & 12 trip units on "B" RPS channel or place test switch off normal RO Pull 1,9 and 12 trip units (simulator operator action) or place lost switch off normal SRO Dolormino nood to reduca power to 70% or loss _ wee -r.w e.< w.i--+ r y .tw'-v T= w *r r-+ ~ T rs w-e r- -w---eres v+*'--~e- -m-- -~-v e w7--
- r '< m - ,
3
)
E S-30 L .._ _. __._.._.-__ _ Operator Acttopo_ EntntES901-4 l l l Scenario No.: FCS 97-4 Event No ;2 Page 2 of 8 i Event
Description:
VCT level channel fails low Time Position Applicant's Actions or Behavior RO identif transmitter failure and operation of LCV 218 2 and L V 218-3 l SRO Direct RO to manually open LCV-218-2 and close LCV- i 218-3 , RO_ Monitor and control primary parameters BOP Monitor and control secondary parameters Deup
ES40t_ _.-._ __ _ __ _ Operator Actions.__ ___.___.f orm_ES.-301-4 Scenano NoJ FCS 97-4 Event Nox 3 Page 3 of 8 Event Description. Power reduction to 70% Time Position Applicant's Actions or Behavior SRO Notify System Operations of power decrease i SRO Direct RO & BOP to commence power reduction SRO Dject RO on method of boration to use RO Begin bor_ation as directed BOP Reduce turbine load to control T cold RO Monitor and control pnmary parameters during power reduction BOP Monitor and control secondary parameters during power reduction SRO Coordinate RO and BOP actions during power reduction 1 manw
& b
l ES 301_. ____ .__ ._ ___ Operator Actions ____ Form _ES 30h4 Scenario No : FCS 97 4 Event NoJ4 Page 4 of 8 Event
Description:
PT-910 fails high l Time Position Applicant's Actions or Behavior BOP Identify rapid decrease in RCS T cold BOP Determine cause as turbine bypass valve being open SRO Direct BOP to take manual control of PCV 910 and close valve BOP Take manual control of PCV-910 and close it BOP Monfor RCS Tc RO Monitor and control RCS parameters SRO Notify I & C of failure _
,. -e , , . . . . , --,-,,----,,.-,.-,-,-.,nr - ,- . . , - , -.. . - , - .
.i i
ES-30L__..____ _._.. Operator Actions . Fos _ESd301-d Scenario No.: FCS 97-4 Event No :5 Page 5 of 8 Event
Description:
Loss of load l Time Position Applicant's Actions or Behavior SRO Direct RO and BOP to perform SPTAs RO Perform primary post trip actions BOP Perform secondary post trip actions r l
~--,.,<-r-u .- . , , . ,,r ,w--., -v.r,e-n- -w N w, -w--= e-,,,-- - , - , , , , - , , , , , , , - - - . , - - - - - - , . - - , - ~ e,-
fiS:301_ __._______ Operator Actions _ _ Eorrn_ES-301d j 1 Scenario No. FCS 97-4 Event No.:6 Page 6 of 8 l i Event
Description:
PORV fails open Time Position Applicant's Actions or Behavior RO Identify failure of PORV to close SRO May direct the RO to close the PORV block valves RO Report that HCV-151 will not close SRO Verify completion of SPTAs by both operators s i 8
,-r - ~ , - ~ , , , , 4,,n ,, ,w , -- ,n,e a g - -,,,. , ,- ,,--
[ ES 301_.__.__.___ Opera!or Actions __._. Form (iS-30h4 i Scenario No.: FCS 97-4 Event No.:7 Page 7 of 8 Event Description' PORV block volve will not close Time Position Applicant's Actions or Behavior RO _ Identify failure of HCV 151 to close SRO Direct RO to complete SPTAs RO Perform remainder of SPTAs BOP Perform remainder of SPTAs RO Trip one RCP in each loop when pressure drops to 1350 Psia _ _ , , RO_ Trip remaining RCPs when NPSH lost SRO Diagnoso event and enter EOP-03 2
-. m 1
-- . .m- , -,- ,. - . .
. - , w ,
.-,4- . , u _, .
- . . _ - _ - _ . _ _ = _ - _ . _ - -. _ _ - _ - . -.
l l i ES-30L___._._.___operatotActions . - ___Eorm E S:30B li Scenano No : FCS 97-4 Event No.:8 Page 8 of a ! Event
Description:
Failure of HPSI pumps to auto start following PPLS Time Position Applicant's Actions or Behavior RO or BOP ldontify failure of HPSI pumps to start SRO Direct all HPSI pumps manually started RO or BOP Start all HPSI pumps , RO or BOP Venfy adequate SI flow SRO Verify all safety functions are satisfied Scenario ends when HPSI flow has been restored and all safety functions are satisfied. s
i liS9!!__. _ _ __., _ ._S(cri;irly in en1L_,_, _ . _ __lernLl;Sdl!)d i i l Sirnulation lincility: fptt_[tijlmitt Scenario No SIM 97 3 (Spare) l Ih:iininers _ _ _ _ ___ Applicants __ Inillal Conditions : 50% powu i Turnos cr: Continue power inctcase I:s ent Malf. I:sent 1:sent Dentipuun Nn. No. T}pc* , I IUN f(alse pow er
? XM I l'DW29 1 1 W flow transtuitter 171101 falls low ,
.1 XM l~ CVC10 l iT 210 lails low) psig) 4 h1AL CAS3C C Instrmnent Air leak (10"a) 5 M Manual reactor trip ruluirc<1 O MAL Gl:N4 C Generator field breaker tails to open (preset) 7 M Al. Cl(DM M CitDM # 30 ejection following trip Sht,Jed entrics are to be initiated by a cue from csaminct.
* (N)oimal, (lutstrument, (C)omponent, (M)ajor til)cacthith lhamincr: ,
Chief thaminct;
'e-- 4-- - -.,..p..- - -
- ., - . - - - , . . , .%.._ ~
l f ES 301__ __._ Operator Actions __._-_ _Eorm ES:301-4 , Scenario No : FCS- 97-5 Event No,: 1 Page 1 of 7 Event Description Raise Power Time Position Applicant's Actions or Dohavior SRO Direct RO on RCS dilution and ASI control. SRO Direct DOP on maintaining Tc during power chcnge RO Dilute RCS and adjust ASI to achieve power change BOP Monitor and control secondary parameters I _. ._ _ . , . . - - . _ . . . . . . _ , , _ - . . + . . _ , , ,. m. . , _ , y_ _ _ . _ , . _ , _.
ES301___..___. _Dperator Actions ___ . Form 1S-301d Scenario No.: FCS 97 5 Event No.: 2 Page 2 of 7 impur- h Event Description Feedwater flow transmitter fails low Time Position Applicant's Actions or Benavior Identify and communicate lowering feed flow and steam BOP ponerator levels SRO Direct BOP to take manual control of feedwater DOP Take manual control of feedwater and restore feedwater level SRO Inform I & C of the failure RO Monitor primary parameters A $ i
ES-301 ..___._. Operator Actions _F_orm ES-301-4 Scenario No.: FCS-97-5 Event No : 3 Page 3 of 7 Event
Description:
Letdown backpressure Instrument (PCV-21C) fails low - letdown 1 isolates Tin.e Position Applicant's Actions or Behavior RO Identify and communicate loss of letdown , RO Determine and communicate ioss of letdown due to PCV 210 being closed SRO Direct RO to take manual control of PCV-210 and reestablish letdown flow. RO Take manual control of PCV-210 and reestablish letdown flow. SRO Notify I & C of PT-210 failure RO_ Monitor and controlletdown BOP Monitor secondary parameters
.M
. .. ._ .. _. _ __ 4 _._.. _ _ . _ .__ _ __ _ _ ..._ _._
t r
;(
ES 301 dpetatorActions ~ ._Eomlisa0_1s. !
~
i I Scenario No.: FCS 97-5 Event No;; 4 - Page 4 of 7 *
- Event
Description:
Instrument Air Leak , Time - Position Applicant's Actions or Behavior I BOP Identify and communicate lowering air pressure SRO Direct BOP to verify STBY compressor _ operation SRO Enter AOP-17 i SRO Caution BOP of possible feed reg. Valve failure 3 SRO Direct EONT to verify compressor and air dryer operations, locate the leak
- SRO- Direct EONT to verify PCV-1753 closure when i instrument air pressure drops to less than 80 psig 1' g SRO Direct EONT to verify PCV-1752 opens when instrument air pressure drops to less than 78 psig SRO Direct BOP to isolate instrument air to containment
?. BOP Close PCV-1849A/B and after verifying air pressure continues to drop,. reopen PCV-1849A/B
-RO Monitor primary parameters SRO Direct manual trip of RX when air pressure drops to 50 psig.
RO or BOP Manually trip reactor , I >g 1 o
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ES-301.__ _ Operator Actions Form ES-3_01_-4 Scenano No.: FCS- 97-5 Event No.:5 Page 5 of 7 Event Description Manual Reactor Trip Time Position Applicant's Actions or Behavior
.SRO Direct RO and BOP to perform standard post trip actions RO Perform primary standard post trip actions BOP Perform secondary standard post trip actions including ^
control of feedwater to both steam generators including securing all feedwater pump _s if necessary _ SRO Diagnose events and enter EOP-20 W esusutD 6
ESr301_ OperatoLActions ForLEfi-3_01-4 Scenario No.: FCS-97 -5 Event No.: 6 Page G of 7 Event
Description:
Generator Field breaker fails to open Time Position Applicant's Actions or Behavior BOP Report that generator field breaker did not open _ BOP Attempt to open field breaker BOP Report that generator field breaker still did not open SRO Direct EONT to manually open field breaker (EONT) Locally open generator field breaker e 0
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I ES 301 ._ __ Operator Actio_Os__ Form ES.-301-4 Sceneno No.: FCS-97-5 Event No.: 7 Page 7 of 7 Event
Description:
CEA # 30 ejection Time Position Applicant's Actions or Behavior RO Identify lowering RCS pressure SRO Direct RO to verify reactivity control RO Verify only one rod not inserted and power lowering RO Secure 1 RCP in each loop at 1350 psia RO Secure remaining RCPs on loss of NPSH SRO Direct RO to verify Si flow and all available charging pumps RO Verify Si flow and all available charging pumps SRO Direct BOP to verify at least one S/G has level or level is being restored BOP Verify at least one S/G has level or level is being restored SRO Direct RO to verify spray flow (when containment pressure is above 5 psig) RO Verify spray flow when directed SRO Enter EOP-20, MVA-IA Scenario ends when all safety funcions have been verified and SRO has entered MVA-IA. l
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. ,,.s- 03/12/97-Toi Ryan Lantz and John Pellet.
The proposed Fort Calhoun SRO written exarn is_ attached. The breakdown of - the questions is as follows:
. 38 questions are from the FCS exam bank. Of these exam bank questions, _8 were seen by the candidates as part of their SRO upgrade training. The date
_ _each of these questions was last used is provided.
-. 7 questions are significantly modified bank questions. The original bank question is provided for each of these questions.
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- 8 questions require the.use of attachments by the candidates. 6 questions '
require FCS figures.or procedures. 2 questions require the use of steam tables.- In addition to the K/A information, the Fort Calhoun lesson plan objective most - , closely related to the question is provided. 4' Several of the K/A's were changed from the original outline. The changes and
-the reason for the changes are included in the comments for the questions. In -
most cases the changes were made to reduce overlap between the written and operating portions of the exam.
- A coinpi lati on o f supporti ngdocumentation or f the written exam questions is included in a separate binder.--
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- Questiojl.; 1 Systern; Mod _ej Mitem; RO_Impj SRO Impj 005 000 K2.C1 3.1 3.1 Syste_rt; Reactor Coolant Pump System Roslej Generic Dgscription; Knowledge of power supplies to the RCPs.
Que.stitrt; 1 The reactor trips from a normal full power condition. Following the trip, all fast transfer to 161 KV fails. What combination or teactor coolant pumps would continue to operate in this situation 7 A. RC-3A and RC 3B B. RC-3B and RC-3C C. RC-3C and RC-3D D. RC 3A and RC-3D 3 w Answer: CIA level: question source: 8ttachment: C Yes New Question none LP numberj _Obiective #: 07-11 20. 01.07c Obloctive: LIST the power supplies for the RCPs.
Reference:
LP 07-1120 Comments; C is the only correct combination, other selections include at least one pump that is powered from a bus that is normally suppl led by 345KV and will lose power if fast transfer does not occur.
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Qugstionj 2 Systemj Modej KA item: RO Impj SRO Impj 061 000 GEN 15 3.8 4.1 Systggu Auxiliary / Emergency Feedwater Modej Genenc System Reicdnt1911; Ability to recognize abnormalindications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. Queslign; A reading of 300*F on the discharge piping of an idle AFW pump is an indication of which one of the following? A. The pump is properly warmed up for rapid start. B. The pump may be steam bound.
- q. C. The pump discharDe vove is open.
D. AFW turbine steam seals are leaking. W3
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4 Answer: CIA level: Question source:
Attachment:
B No Bank Quettbn - not seen none LPJimnben .O_tti,ctive #: 07-11 01 01.09 Obiective: . EXPLAIN symptoms of AFW pump steam binding and appropriate corrective action. fteforencq; LP 07-11-01
. Q9g! stents: bank question 0711-01,1.9 003 (and 005) v'
. Questio!Li 3 System; M_qdel KA iterrt; 80 Impj S_R_OlrDpl 063- '000 K2.01 2.9 3.1 S.ystenti DC Electrical Distribution Modej Genenc System Dg.sqIjpflom Knowledge of the power supplies to the major dc loads.
QueEll9El Within 15 minutes following a r,tation blackout, DC loads were minimized in accordance with EOP attachments. The following breakers were placed in *OFF': EE 8G,CB12, *400 CYCLE INVERTER EE 21* EE-8G-CBS,
- EMERGENCY 1.!GHTING PNL ELP-2 TRANSFER SWITCH
- EE 8F CB11,
- EMERGENCY LIGHTING PNL ELP 1 TRANSFER SWITCH
- BKR 15 EMERG. LIGHTING PANEL NO. 5 (DC-PNL-1 125 VDC PANEL)
Which one of the following instruments will not provide indication in this condition? A.. Emergency Feedwater Storage Tank Level. B. Primary Rod Position Indication. r C. Pressurtzer Steam Space Temperature. D. Volume Control Tank Pressure. A_qspen C/A level: Question source; attacnmentj B No New Question none Objective #: J.P number: 07-12-26 01.02a ODjective: Describe the interf ace / interaction between the CRDS and the following systems / components: ElectricalDistribution System. Reference _J LP-07-12-2S Comments: DC power breaker to the 400 HZ inverter used for primary rod position indication is opened.
Quest [o_rtj 4 S s,t*m; hdm IMite_mj RO InJg; SRO Imp; J ._ 001 010 A2.04 3.8 4.0 Sys19mj Control Rod Drive System Mo_dej Startup/ Shutdown D_31gdpflom Predict the impact and mitigate the consequences of a station blackout upon the operation of CRDS Qugf1!gn; Which one of the following statements is true conceming the response of the CEDMs following a loss of power to all 4160 V busses A. The CEDMs will fallimmediately due to a loss of power to the drive motors. B. The CEDMs will fallimmediately due to a loss of power to the clutches. C. The CEDMs will not fall until a valid RPS signalis generated and power to the drive motors is interrupted. D. The CEDMs will not fall until a valid RPS signalis generated and power to the clutches is interrupted. Sg w' Answer: CIA levej; Question source: 8t_tachment: D Yes New Question none l.P number: Obiective #: 07-12-26 01.02d Obiectiv_eJ Describe the interface / interaction between the CROS and the following systemsicomponents: Reactor Protective System.
!Lefgrejige_1 LP 07-12 26 gggtments: B is inncorrect because power io the clutches is provided by instrument busses backed up by battery power. Tl4e rods will not fall until a valid RPS trip signal is generated. A&C are incorrect t ecause the drive motor is not involved in a reactor trip.
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Qugsti.on,J 5- SyslerD; Mgdel KA item: RO_lmpj SHO impj 071 000 A4.09 3.3 3.5 SysterJL; Waste Gas Disposal System MoAej o Generic DisMplioE; Ability to manually operate and monitor the waste gas release rad monitors. Questio_nj s Under which one of the following conditions is it acceptable to release a waste gas decay tank with RM 062 inoperable? A. RM 052 is operab!c and lined up to the stack. B. RM-057 is operable and condenser offgas and the gas decay tank are both lined up to the hydrogen purge filters. C. The PASS nionitors are operable snd the release directed through the PASS system. D. The Waste gas decay tank being released has been isolated and all of it's contents allowed to decay for 7 days,
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.t Answer: C/A level: Question source; Attac_llment: Yes New Question none A LP number: _0_biective #: 07-11-31 03.01 Objective: COMPLETE applicable portions of a dummy FC-213, waste gas release permit, and EXPLAIN the sections that are reviewed by the Shift Supervisor, (SRO only) Refegencel a LP 07-11-31 Commentsj SRO level question 55.43(b)(4)
QutStiOR.; 6 Systern; M_ ode; KA item: 80 ImS; SRO IfDPJ ,
- 004 020 K6.02 3.8 4.1 l Systegt; Chemical and Volume Control Mode _; NormalOperations System i
MHf Pil911; i Knowledge of pressure control methods during sold plant operations. QuestLoJ1; l During sold plant operation, PT 115 fails such that it provides a high indication. How will operation of the LTOP system be affected? A. One PORV will open immediately following the instrument failure. B. Both PORV's will open immediately following the instrument failure. C. One PORV will open if the actual system pressure and temperature rea:,h the LTOP setpoint. D. Both PORV's will open if the actual system pressure and temperature reach the LTOP setpoint. C/A level: Cluestion source;
Attachment:
Answer: Yes New Question none B LF number: .O_bitctive #: 07 11-20 03.00 Obiective: EXPLAIN how the RCS is started up and shutdown, using applicable Operating Instructions as a guide for major steps, prerequisites and precautions. Refefencti LP 07-11-20 Comments: b -
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Questionj . 7 System; Kodej KA.jtem; ROlmnpj r SfLO_Irnpj 015 000 K4.05 4.3 4.5 Systernj Nuclear instrurnentation System Modej Generic Desgription; Knowledge of the NI.3 design features and interlocks which provide reactor trip. Questjonj Two days following the failure of Ni power range channel *D*, l & C technicians placed the bistables for channel *D* inp units 1, 9 and 12 in the ' tripped" condition. They would hke to calibrate NI power range channel *A*, The cahbration procedure involves operation of the cahbration knob on the chantiel'A' drawer. How will RPS operability be affected ifinp units 1,9 and 12 on channel *A* are not bypassed during this calibration procedure? A. There will be no effect on RPS operability. B. The reactor will trip during this evolution. C. The trip logic will be altered to 1 of 2 operable channels during the calibration. D. The trip logic will be altered to 2 of 2 operable channels during the calibration. Ajnw_en CIA level: Ottestion source;
Attachment:
New Question none 8 Yes L.P num_ ben _Obiective #: 07 12 19 01.12 O_blective.j Explain the difference in the resultant coincidence if one channel in a 2 of 4 logic configuration is bypassed or de-energized. Gefere_nge; LP 07-12-19 Cpmmentsj SRO level question- 55.43(b)(2). SRO authortzes test. Power trip test interlock will place channel A bistables in trip when calibration knob taken out of operate position,
91HtstiorLt 8 gystem; Modej KAitem: RO_lmpj SRO Impj 022 000 K3.02 3.0 3.3 Syatem : Containment Coohng System Mode; Generic Qticl!Rtioni Knowledge of the effect that a loss of the CCS will have on containment instrumentation readings.
- 939stloji; What effect will a loss of the detector well cooling system have on the operation of nuclear instrumentation systems?
A. The Power Range NI detectors will begin to provide erratic indication. B. The Wde Range NI detectors will begin .to provide erratic indication. C. The in-Core detectors will begin to provide erratic indication. D. There will be no effect because all NI systems are designed to operate without detector well cooling. qs 6_rtsweg CIA level: Qtwst!on source; ettachment; No New Question none A LP number: Obiectlyel 14-05 01.01 Oblecitygj State the purpose of the Detector Well Cooling System.
Reference:
LP 07-14-05 90mments;
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a QWgsti.9nj s 9 S sie3n Modej KA iteJm RO Impj !.fLO Impj l 013 000 K1.02 3.2 3.6 7< gyste3L; Engineered Safety Features Modej Generic Actuation System Qgg.gIfpilpff Knowledge of the relationship between the ESF and RCP. 9Mtst! Rill Which one of the following conditions will resutt in isolation of component cooling water to the reactor coolant pumps? A. CIAS and low component cooling water pressure. O. CIAS and high component cooling water radiatica levels. C. CRHS and low component cooling water pressure. D. CRHS and high component cooling w.eter radiation lewis. I i Answer: C/A level: Question source: Machment: A No New Question none LP_pumber: - oblectivip; 07-11-06 01.04 Obiective: EXPLAIN standby operation of CCW pumps in terms of switch positions and automatic actions. Baf9fyncJ; LP 0711-06 99mmestt41 \
QuestJgnj 10 Systen1; Mgdej KA itemj RO Imp; SROImpj 004 000 A2.07 3.4 3.7 Bystenjj Chemical and Volume Control Mo_ del Generic System pesqfiptionj Abiltty to preosci and mitigate the consequences of isolation of letdown / makeup. Question; How would VCT level respond if charging and letdown were both secured? A. VCT level will slowly decrease. B. VCT level will remain constant. C. VCT level will slowly increase. D. VCT level will osci! ate. ,sg a CIA level: Question source _; Aggchment; Answer: C Yes BankQuestion - seen none L.P tiumt!gn _O._kjectlye #:
.07-11-02 04.02 QDActive: EXPLAIN the effects of isolating letdown and charging if reator coolant pumps are operating, fleference; LP 0711-02 C_oH1me_ntsj bank question 4.2,N 001 last used on 07/19/03 y
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QgestjoJLJ 11 System; MMej KA_!tts RQJrnpj SLOlfDPi 014 000 K4,06 3.4 3.7 SEtem : Rod Position Indication System Mgdej Genenc Qescrjgtiqq; Knowledge of RPIS design features which provide for individual and group misalignment indication. QMt1119D1 Wnich one of the following statements conceming the Secondary CEA Position Indication Systern (SCEAPIS) is correct 7 A.- The SCEAPIS can display the position of all the rods simultaneously. B. The SCEAPIS is more accurate than the primary CEAPIS. C. The SCEAPIS provides sequencing of the regulating CEA groups. D. The SCEAPIS activates a ' Dropped CEA* annunciator if any CEA is inserted below 2 inches. 4i %._ 1 Answer: CIA level: Question source: ettschment: A No BankQuestion - seen none LP number: __ Objective #:
-07 12 26 01.07 Obioctive: Describe the methods of control rod position indication. Include the readouts and displays associated with each method. (CID No.
931191/02) EqfeEng_ej LP 07-12-26 Comments: bank question 1.7 001
. fast used 08/02/96 v
l Qugtsthoj 12 Snterm Modej KAitem; ROImpj S&O_fmp; I 072 000 GEN-09 2.8 ' 3.0 Sygigmj Area Radiation Mondonng M_odej Generic System posgdpjlofu Ability to locate and operate components; including local controls. f Que!!igI1; 1 Where are the area radiation monMor alert and high level setpoints listed? j A. Operating procedure, Ol-RM-1. B. Offsite Dose Calculation Manual. C. TechnicalData Book. D. Technical Spec.ilcations. C/A level: Question source: Attachment; A_nswer: C No BankQuestion - seen none LP number: Obiective #: 07 12-03 02.02 Obiective: STATE the two alarm setpoints for most monitors and EXPLAIN what each setpoint designates.
Reference:
LP 0712-03 Q2fpments; Bank question 2.2 002 one distractor changed SRO 55.43(b)(5)
- last used 07/26/96 v
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- Questio!L; 13 System: Modej KA ite_m_; RO Imju SJlo Iml; 026 020 A4.04 3.5 - 3.5 Rygte_rIL; Containment Spray System Mod _e; NormalOperations pesstiptio[1; Ability to manually operate and monitor in the control room the containment spray reset switches.
Qugtio[1;
. Which one of the following statements is true concoming the Containment Pressure High Signal (CPHS) lockout relays?
A. They normally reset automatically and only require manual reset after a loss of voltage. B. They allow manual reset so that containment spray actuation can be overridden with a CPHS signal present. C. They allow manual reset of the CPHS signal after containment pressure drops below the CPHS setpoint. D. They allow the CPHS signal to be
- locked out* during containment spray pump testing.
Answer: CIA level; Question source:
Attachment:
C No Modified Bank Question none LP number: - Objective #: 07 12-14 01.01b Objej; tin; EXPLAIN the operation of the following devices: Lockout relays Referencjg LP 07-12-14 Commeltsj Question 1.1B 001 significantly modifM onginalquestion seen 08/02/96 original question on next page.
- -- - -_ - . - . - _ . - . . . - - . . _ - _ _ - . ~ . .
owging n rw.a. 30.Im etonAM 5.0 RO/SRO r.e= 6 171214,1,1 A 002 KEY WORDS: Keyword 1 Keyword 2 Keyword 3 Keyword 4 Keyword 5 Keyword 6 Keyword 7 Keyword 8 i i l l I I I I I I D ATES: Modited: Fri. Aug 30, 1996 Used: 171214.1.18 001 Wrucn one of the following accurately casences tne operanon of locnout resays used in Safeguards drcutts? [1.0pt) A. Lockout relays are normally self.reseting and only required manual reset after a loss of voltage. B. Lockout relays have an AC coil which remairis fully energtzed after the relay trips. / _ W. Lockout relays have a DC cost which is de. energized after the relay inps. D. Lockout relays have to be manually
- locked out' In order to cause the designed system actuation. ,
d The correct answer le C [1.0 pt] 1 Ent. 8/19/92 l
Reference:
i T. Gurtis 4
*** KEY WORDS:
Keyword 1 Keyword 2 Keyword 3 Keyword 4 Keyword 6 Keyword 6 Keyword 7 Keyword 8
. I I I I I I I I I
' 0ATES: Modified:Wed, Aug 2t, t996 Umst . . 17 12 14.1.18 002 Which of tne foisowing reiays inp as a result of an e6ectncal signal and stay in that condition until manually reset. [1.0pt)
- 4. Lockout relays B. Supervisor relays C. Self reseting relays
- 0. Blocking relays The correct enewer is A (1.0 pt)
Ent. 1/13194 ' KEY WORDS: Keyword i Keyword 2 Keyword 3 keyword 4 Keyword S Keyword 6 KevWord 7 Keyword 8 I I I I I I I I I DATES: Modined:Frt. Aug 30. 1996 Unst
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i QuestiorL; 14 Systern; Mode; KA item; RO Impi SRO_ Imp;
-s 059 000 A3.02 2.9 3.1 System 1 Main Feedwater System Mgdej Generic Description; Ability to monitor the automatic operation of the programmed levels of the S/G.
Questionj Which one of the following instrumentation failures would cause steam generator level to increase wrlh the rnain feedwater regulating valves closed and the bypass valves in automatic control? A. Steam flow instrument fails high. B. Steam Generator pressure instrument fails high. C. Feedwater flow instrument fails low. D. Steam Generator level instrument fails low. sy, 9/A level; Question source; Attachstent; A_nspen Yes New Question none D LE_num_ ben y e #j _0_bjectl 07-11-11 04.03 _ Objective: EXPLAIN the various modes for steam generator water level contrci per Ol-FW 3. Refe_rencel LP 07-11-11 Com_ments; D is correct because a low levelinput will cause FW flow to increase raising level. A,B and C are incorrect because the bypass controllers are single-element using only level for input.
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QuestLonj 15 System: Modej KA item Ho im_p; SRO Impj 022 000 A1.04 3.2 3.3 SysterDj Containment Cooling System Modej Generic pgsgription; Ability to predict and/or monitor changes in cooling water flow due to operating CCS controls QuejilloJl.; Which one of the following conditions will cause the containment cooler lulet and outlet valves to close following receipt of a CIAS signal? A. CCW pump discharge pressure is less than 40 psig. B. A containment presure high signal (CPHS)is not present. C. The fan associated with the containment cooling unit failed to start. D. There is lo.v CCW flow from the cooler outlet. ?.m 1 Answer: CIA level; Question source:
Attachment:
D No Bankouestion - seen none LP numbel Objective #: 07-11-06 01.05 Obloctive: EXPLAIN the response of the CCW System to signats from the Engineered Safeguards Control System.
Reference:
LP 07-1106 Cpmments_,J bank question 1.5 003 slightly modified to update distractor last used on 7/12/96 h
outstion; 18 systami Modej KA item _j RO Imp; SROimpj 001 010 K5.13 3.1 3.6 gyyjamj Control Rod Drive System Mo_dej Startup/ Shutdown Qtscf plio0; i Knowledge of reactivity balance and that withdrawal of shutdown banks must precede dilution. Rutstlo!11 Preparations are being made to perform a reactor startup by CEA withdrawal. According to the ECC calculation, the boron concentration should be reduced by 250 ppm prior to the startup. According to OP 2A, which one of the following must be done prior to dilution? A. The non-tsippable CEAs must be fully inserted. B. The non-trippable CEAs must be fu!!y withdrawn. C, The shutdown bank CEAs must be fuliy inserted. D. The shutdown bank CEAs must be fully withdrawn. m C/A level: Question source:
Attachment:
Answe_n D No New Question none l,.P numble Obiective #: 08 12 11 01.01 Objectivel Demonstrate proper administrative and supervisory skills by giving correct and timely directions to shift personnel as to actions required and proper procedures to follow. Referengel LP 08-1211, OP-2A Commentsj SRO- 55.43(b)(6)
Queslipfi,; 17 Sytes Mo_dgj KMt!E ROJr_m SRO Impj 017 000 GEN 06 2.2 3.4 Systemj In-Core Temperature Monitor Mode _; Generic System Dgscriptjo_nj Knowledge of bases in tech specs for limiting conditions for operations and safety limits. Quesilm What is the basis for the technical specification operability requi!ements for the core exit thennoucouples? ) A. Operability requirements are based on using the CETs to monitor core peaking factors during power operation. B. Operability requirements arts based on using the CETs to adjust the TMLP trip setpoint. C. Operability requirements are based on using the CETs to calibrate the reactor vessellevel monitoring system. D. Operability requirements are based on using the CETs to monitor for inadequate core cooling.
,s Answer: C/A level: Question source:
Attachment:
D' _No New Question none LP number: _Qblective #: 07-12-20 01.05 Qblegttypj EXPLAIN the basis of the Technical specification limits on incore NI System.
Reference:
LP 0712 20, Technical specifications Comments: SRO 55.43(b)(2)
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system _; Mode; taitean 80Jmp; Saq1mp;
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ouestion.; 18 001 010 K6.04 2.9 3.2 Systg.gij Control Rod Drive System Mode: Startup/ Shutdown - .Descfipflon; Location and operation of CRDS fault detection and resets including rod control anunciators. Questlon; While conducting rounds, the EONT infoims you of the following conditions: The voltmeters on EE 22 indicate 30 and 90 VAC. Both hghts on EE 22 are dimly it. What would you conclude from these indications? A. A ground is developing in the rod drive cire'Jits. B. A ground is developing in the core mimic bus. C. A ground is developing which affect's both the rod drive circuits and the core mimic ' bus. D. There is no indication of a problem. 8Ds_wey C/A level: Question source: 8ttachment: A Yes New Question none W numboS Obiective #: 07 12-26 01,10 Objective: Describe the operation and purpose of the Control Rod Drive and CEA Mimic Bus ground detection system. Beforence: LP 07-12 26 Commentsj SRO This information would be reported to the SRO by the EONT 55.43(b)(5)
Questiomj 19 Systent; M_ods KA item: RO Imp; SEp_I_rnpj 068- 000 A2.04 3.3 3.3 Systemj Liquid Radwaste System Modej Genenc Desnip11911; Ability to predict and mitigate the consequences of failure of automatic isolation. Q1Hlts!!oJ11 An alarm on effluent radiation monitor, RM-055, will cause which one of the following to happen? A. HCV-691 and HCV 692 (overboard discharge control valves) will close and the monitor tank pumps will trip. B. HCV-672 and HCV-678 (monitor tank inlet valves) will close and the monitor tank pumps will trip. C. HCV 691 and HCV-692 (overboard riischarge control valves) will close and the monitor tank pumps will continue to operate.. D. HCV-672 and HCV 678 (monitor tank inlet valves) will close and the monitor tank pumps will continue to operate. A3sygn C/A level: gpestion sourc_ej Attad! ment: A No Bank Question - not seen none LP number: Objective #: 07-11 32 01.02 Objective; Identify the alarms, controls and indications available in the Main Control Room that are associated with the WDLS, Beierungej LP 07-11-32 Comments; bank question 1.2 003 distractors modified. 55.43(b)(4)
I Questionj ~ 20 System: Modej _KA itefn; BAl.mJj S((O_Im_p; 002 000 K4.02 3.5 3.8 SnteMJ Reactor Coolant System Mod.e; Generic Dendpijon] Knowledge of RCS design features which provide for monitonng reactor vessel level. Questioni RVLMS must be 43% or greater prior to performing HPSI stop and throttle during a LOCA. Maintaining RVLMS 43% or greater assures that: A. any steam void in the reactor vessel head will not be large enough to csuse a significant increase in pressurizer level. B. the core exit thermocouple are covered with water and will respond property. C. the reactor vesselievelis above the entical vessel welds reducing the likelyhood of Pressurized Thermal Shock. D. The hot legs are covered provding sufficient inventory to support natural circulation if the S/G's are provding a heat sink.. w CIA level; Question source: Attachment; Answen D Yes New Question none LP number: _Obkctive #: 07 18 13 01.08 oblective: GIVEN a copy of the Safety Function Status Checklist and a set of plant conditions, DETERMINE whether or not each Safety Function meets the acceptance cnteria listed. Referenc_ej LP 07-18-13, CEN-152 rev 04 Commentsj SRO level question - basis for EOP 55.43(b)(5)
Question.1 21 Dyttemi Mode; KAJtemi RQJmp; sRQJmp; 006 030 K4.04 3.9 4.1 Synternj Emergency Core Cooling Mytjgj Mods Change System QgscylptioDj Knowledge of the ECCS design features which relate to valve positioning on a safeN injection signal. Apostion; r A SIAS actuation occurs followin0 a loss of coolant accident. 30 minutes into the accident, the water lovel in the SIRWT lowers to the Sn.S setpoint. Assuming all systems operate as deshned, which of the following valves will close. A. The LPSIinjection valves, HCV 327,329,331,333. D. The SIRWT recirculation valves, HCV 385,388 C. The SIT loop isolution valves, HCV 2'd14,2934,2954,2974. D. The containment spray header discharge valves, HCV 344,345. o M CIA level; RgentlpfLigyIge; Agath!ntDS
&Dfwen B- No New Question none LP_.0Mmk95 .Qb}tfilYiff 07 11 22 01.06c QDiffilytj Explain overall system response to actuation of automatic engineered safeguards signals: Recirculation Actuation Signal (RAS),
RefgInngej LP 071122 99mmentti w y e -, ,e - - . -
Chsest!pnj 22 SYM9fD; Medvi K<9m; f{0jmpj $80_impj
- 011 000 K2.01 3.1 3.2 Syptemj Pressutt.er LevelControl Mqde; Genenc System Ptspfjplion; Knowledge of bus power supplies to the charging pumps.
Questioni The plant experienced a loss of all offsite power. D/G #1 started normally but D/G # 2 failed to start. Assuming that no action is taken to cross tie busses, what is the maximum available charging flow capaciti in this condition? A. Oopm. D. 40gpm. C, 80 ppm. i D, 120 gpm. I 8DfY!tG G8_ItY*li RuffLl9n_fouLgg; 811Mhatilti B Yes New Question nont LP_tlumhtc _QbitfilY111 07 11 02 01.03 9)ltitlYoj EXPLAIN the automatic and manual controls associated with the charging pumps and boric acid piamps, Bfitifatti LP 071102 E9.01mfatt; $5.43(b)(5) Only CH 1 A has power. Each charging pump has a 40 opm capacity. v
Apestion; 23 sygernt Mod *J KAJ1tml fiOJmPJ S.89]mRI 012 000 A2.02 3.6 3.9 Systerrtj Reactor Protection System Mode; Generic Destriptio.0; Abtidy to piedict and rnnigate the consequences of a loss of instrument power, Qufstign; Which of the following is true regarding the Diverse Scram System upon a loss of the 120 VAC power supply to either Al 190 or Al 1987 A. To bypass the failed channel, you must place the TEST / BYPASS switch on the affected panel to the bypass position. B. To bypass the failed channel, you must place the TEST / BYPASS swrtches on both panels to the bypass position. C. It is not posible to bypass a single sensor channel on bss of 120 VAC power. D. It is not necessary to bypass a chtanel upon loss of the 120 VAC power because the DSS requires power to generate a inp signal. 99ffil91UE9 VIE 1; <achment; ADtWell GlAJtYtli No Bank Question not seen none C LPnumhtfi .Q> Jest.tvi!1 07 12 25 05.04 Ob}titlyel EXPLAIN the function and STATE the location of the Diverse Scram System test and trip switches. Eoittentti LP 0712 25 E9mmtatsj from 10/89 NRC exam v
l l QW,ptigai 24 Symfpin; Mede; KA_litml ROJmpj AROJmp; ' ..__ 012 000 K3.02 3.2 3.3 i Snternj Reactor Protection System M2d ; ceneric pgsJtiptjgp] Knowledge of the effect that a loss of the RPS will have on the T/G. RVestign; MT If one turbine stop valve and one turbine control valve remain open following a reactor tripiurbine trip, which one Jf the following automatic actions will be delayed? A. Closure of the combined-intermediate valves. B. Ramp down of the feedwater regulating valves. C. Loss of load trip generation by the RPG. D. Opening of breakers 34514 and 34515. 8DtWifj Cj/ Lle_ytJ; Question source: AttalhtHtD11 D Yes New Question none LP_.oitinhen .pblestive #: 07 12 29 03,01 Qbjettlym EXPLAIN the interfaces that the Turbine Protection System has with other plant systems. Egittyngtj LP 0712-29 comments s
ouest191ti 25 systemi Modt1 EA.ittfril RQ. trop; sag. trop; 016 000 K1.03 3.2 3.2 Sysitmj Non-Nuclear Instrumentation Mg{tej Generic System prict.jplioDj Knowledge of the telationship between the NNIS and the Steam Dump System. 49epilorll Which one of the following desenbes the response of the steam dump and bypass valves to an uncomplicated reactor trip if PT 910 falls such that it Indicates 'O' psig? Assume all systems are in automatic. A. All steam dump and bypass valves will opan. RCS average temperature will be restored to and maintained between 530*F and 535'F. B. Only the steam dump valves (TCV 909-14) will open. RCS average temperature will be restored to and maintained between 530*F and 535'F. C. All steam dump and bypass valvos will open. RCS average temperature will be resored to and rnaintained between 535'F and $40'F. D. Only the steam dump valves (TCV 90914) will open. RCS average temperature will be resored to and maintained between 535'F and 540*F, ansytfi C_!A.Je_ytli Question sourcej attachment: C Yes New Question none LP_nMmhtii ShitEllY1!1 07 12 31 02.03-Q)Jgitlyg; STATE the indications and points of control available to the operator for RRS functions.
!Lettrencei LP 07.t2 31
.Cemments; m
.-..a ,,..e... . . - _ .
Augstion; 26 System; Modt1 KAitem; Rolms; gaolenpj 029 000 GEN-04 2.9 3.0 Systemj Containment Purge System Moctej Generic pe$cr.jptlpn; Knowledge of system purpose and function. quettloDI What is the purpose of the mini-purge fans in the containment purge system? A. To provide reduced flow operation of the containment purge system during power operations. B. To provde direct purge of the containment area inside the biological shlekt. C. To allow operation of the containment purge system during specific electrical bus outages. D. To allow operation of the containrnent purge system while minimizing the potential for airbome contarnination. AutWiff CLA_{tyti; QggetioJLsau.riej s AttaGbmeilli D No New Question none LP_numhts .Rtzlestlytt OM 444 01.04 pkjestlyel STATE the function of each major component of the Containment Purge System. Btf*_MnE9J LP 0714-04 Cd2HLmintsi Knowledge level but addesses objec0ve directly.
QVtflioft; 27 Syltem; Niodf1 K8Jtem; ROlmpf SR0jmp; 033 000 K4.01 2.9 3.2 Systent; Spent Fuel Poul Cooling System Mode; Generic Qtscriptign; Knowledge of SFPCS design fe%res and interiocks which provide for the maintenance of spent fuellevel. 09ettJ9Di Makeup to the spent fuel pool for normal evaporative losses is: A. From the SIRWT via the Fuel Transfer Canel Drain Pumps. D. From the SIRWT via the Reactor Coolant Drain Tank Pump. C. From the CVCS using the Boric Acid Pumps Demin Water Pumps and the blending 100. D. From the Safety injection System via the LPSI pumps. C/A leyel; Question source:
Attachment:
811tWeG A No Bank Question . not seen none LP_nuaLben _Qbjettiy*_J!i 07 11 24 01.03g Objectival STATE the function of each of the following major components of the Spent Fuel Pool Cooling System: Fuel transfer canal drain pumps (AC 13A and 138) Battfintt; LP 07-1124 Comments; one objective modified, bank question 071124,1.2 002 SRO 55.43(b)(7) v y
OL tttignj 28 System; Mpdei IMitem; RQlmp; Sf{Q. imp; 028 000 K5.03 2.9 3.6 Systemj Hydrogen Recombinar and Mode; Gsneric Purge Control System pesgtlptlpry Knowledge of the sources of hydrogen within containment. Quettign; Hydrogen concentration within containment was found to be 2.6%, four hours after a srnalt break loss of coolant accident with sustained core uncovery. Which one of the following was responsible for production of the majority of the hydrogen? A. The Zircalloy-Water reaction. D. Corrosion of aluminum. C. Radiolysts. D. De-gassing of the RCS coolant. Angmn CIA levelj Question source: Alta_chment: D No New Question none L.P_nwnken
. .9b_Jective #:
07 15 28 01.13 QA}tcllyyj EXPLAIN the generaticn of hydrogen in an accident scenario. Etttangtj 1.P 0715-28 99mments;
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09t1119f1) 29 System; Modt; KAitefm K9J!Dej SROImpj 039 000 K3.05 3.6 3.7 Systemj Main and Reheat Steam System Mqd_tj Genenc Qes.gffpligm Knowledge of the effect that a loss of the MRSS will have on the RCS. questLom Which one of the following wil act to mitigate a RCS high temperature condition following a complete loss of load accident without reactor trip 7 A. The steam generator safety valves. D. The main steam isolation valves. C. The emergency boration system. D. The emergency core cooling system. rM[h 9/A-level; Ruttilgnigpree: A_ttachmelt; entwin New ouestion none A No LP_itufBhtG Sk}tEllYffl 07 11 17 01.05 Qhitstlyt; Given the Technical Specification Manual, INTERPRET the limiting conditbns for operation that apply to the Main Steam System. Retamisti LP 0711 17, Technical specification bases 9.RfD! Dints; SRO $5.43(b)(2) s.-
r Quettiprt; 30 Syttem; Model RAjtem; 80_Lrnp; MQ.jgip; 002 000 A4.01 3.3 3.1 Syytern.J AC Electncal Distnbution Modej Genenc System ResGilptiom Ability to manually operate and/or monitor all breakers (includin0 available switchyard). Questipm Assurning all fast transfer permissives are made up, which of the following will result in a fast bus transfer of Bus 1 A3 from the 101KV to the 22 KV source? A. OPLS lock out relay actuation. O. Manual trip of the 69 switch on breaker 1 A33. C. Overcurrent lockout of breaker 1 A33. D. Undervoltage on bus 1 A3. G!%.JtYil; QugitLon.nnutte_; Attachment; 8Aswer: B Yes BankQuestion seen none LP_numhqE QDiective #: 07 13-02 01.09 Qbjt.q.tlygj e Explain how the system responds automatically to malfunctions. BeffieAqgi LP 0713-02 CJNDmentsj bank question 1.9 003 slightly modifd, last used on 07/03/96
Qpes1lQp.; 31 Syftem; Modej KAj!em; RO_lrppi SRO_ imp; 062 000 GEN-15 3.4 3.8 Systernj AC Electrical Distribution M9de; Generic Gystem pets 11ption; Abilsty to recognize abnormalindications for system operating parameters which are entry level conditions for emergency and abnormal operating procedures. Quettjgn; i Which one of the following conditions is addresed by AOP 32,
- LOSS OF 4160 VOLT OR 480 VOLT DUS POWER *?
A. Loss of MCC-4A2. B. Loss of instrument bus #1. C. Loss of busses 1 A3 and 1 A4. D. Loss of DC bus #2. .y w 80Ht9G GlA ItYRll QWilll9A.91VlS11 AttacArntnti No New Question none A
. LP__numhtt; _Q#1tstivtfi 07 17 32 01.03 Qbittily11 Describe the major recovery actions of this AOP.
Refetynttj LP 0717 32, AOP-32 99mmentti SRO question 55.43(b)(5) s
Qu9sti9ft_i 32 SYttem; Mpdg1 KAftem; ROjmpf RB_O.jmp; 064 000 , A3,00 33 3.4 Systemj Emergency DieselGenerator Mpdej Genenc System postflatig[1; Abildy to monitor automrtic operation of the ED/G system; including start and stop. QM91tl9BI The reactor is cperating at 0% power dunng a plant startup. Turbine speed is 1500 ' rpm and increasing when an inadveriant turbine trip occurs. The reactor does not trip. What is the expected response of the D/G's to this event? A. The D/G's will tornain shutdown. D. The D/G's will start and come to idle speed. C. The D/G's will start and go to full speed. The D/G breaker will close. D. The D!G's will start and go to full speed. The D/G breaker will not close. rs Answer; 9/A level; Qgestion sourgg; AttMtifntilii A Yes New Question none ig_[LuntHgj _Oklective #: 07 13-05 01.10a Qhhtsilysi Explain an emergency start of the EDO. Inc!ude in your explanation the following: The conditions that will cause an auto start. Smitttaggi LP 0713-05 Gemments; h
Questionj 33 System; Mpde; tQJtemi liq 301PJ SEQ _trDpj 034 000 K6.01 2.1 3.0 Systeral Fuel Handling Equipment Modej Genenc Systern pescription] Knowledge of performance and design attnbutes of the fuel handling equipment. Qugstlon; What is the purpose of waiting a minimum of 72 hours prior to initiating fuel (novernent after a shutdown from a power level greater than 2%7 A. It allows tirne for the decay of short lived fission products and allows for any failed fuel to purge itself. B. It allows time for the containment ventilation system to clean up the containment atmosphere to prevent a release in excess of technical specification limits. C. It allows time for fllling the reactor cavity with approximately 250,000 gallons of borated water to maintain a sufficient shutdown margin. D. It allows time to cooldown and depressurize the RCG to assure no steam will be formed and minimizes the potential for a containment pressure buildup should a RCS rupture occur. u AJMWfE GlA_by_ ell Question souffeJ 6Eafhmtuti A No Bank Question not seen none LP_nUmbeE -Q539CilYifi 07 11 13 02,01 Q)Jeltlye.j Discuss the prerequisites and precautions associated with fuel handling equipment and the refueling machine. Refttentfl LP 0711 13 9Enstatti K/A changed from 064 000 K6.08 to improve sampling of 55.43 Items. bank question 0711 13,2.1001 55,43(b)(7) f
l Questiort; 34 Systemi Model 6/Litemi RQJrpp; sRolrpp; 086 000 K4.01 3.1 3.7 Systemj Fire Protection System MQIfej Generic post [lptign; Knowledge of design features which provide for adequate supply of water for FPG, Quotilort; A fire has been discovered in the main transformer. T.1, What would be the correct sequence for automatic starting of the fire pump (s)? Assume the fire protection system is fully operable. A. Deluge valve opening would cause only the electnc fire pump to start on low system pressure. D. Deluge valve opening wouki cause both fire pumps to start on low system pressure. C, Actuation of the T 1 deluge system will generate a direct start signal to both fire pumps. D. A T 1 fire will not result in automatic starting of the fire pumps until the pull stations are operated in the service building.' 60sytel; Q!A letel; Qugstigits.gutce; Maemgnti C Yes New Question none LP_numhtn Rk191tittJ!1 07 11 12 01.00 Qbitit.lya; When given specific plant conditions, be able to APPLY operating principles to diagnose Fire Protection System response. Ettetyngq; LP 0711 12 commentsj
l i Question; 35 Hy).tomj M_Edtj M1t.emi 80 lmpj SRo lmp; 073 000 A1.01 3.2 3.5 Sytternj Process Radiation Monitonng Model Generic Systern QtMElplign; Ability to predict and monitor changes in radiation levels. Qutst.ippj Which one of tho following sets of symptorns would be indicative of a Reactor Coolant Pump Seal Cooler leak during full power operation? A. Increasing CCW surgo tank level and increcsing radiation on RM-053. D. Increasing CCW surge tank level and decreasing radiation on RM-053. C. Dectensing CCW surge tank level and increasing radiation on RM 053. D. Decreasing CCW surge tank lovel and decreasing radiation on RM 053. es Angyref; Q]A_leytli Qygstion soutqti Atlas 11tntat; A Yes New Question none LP_.numhtn .9hittityt !i 07 11 06 06.01 Qbjtsilyti EXPLAIN conditions that indicate leakage in or out of the CCW System. Stitfintf; LP 071106 G9mmtnisj SRO 55.43(b)(4) RCP seal cooler leak identified as risk significant in FCS IPE
Questi9n_; 36 Systemi Modej KMtem; RO.Jmp; SHOJmpi 103 000 K3.02 3.8 4.2 Sy$ternj Conta6 ament Systern Mgtl ej Genenc Qescription; Knowledge of the effect a loss of the containment system will havt, on the loss of containment integrfty under normal operations. Question; The plant is at 100% power when during maintenance activities a containment electrical penetration is damaged to the point that noticeable air is leaking from it. What action should be taken in accordaniv with AOP 12,' Loss of Containment integray*? A. Trip the reactor and enter EOP 00. B. Terminato any positive reactivtty changes and allow reactor power to coast while e repairs can be inntated. C. Intilate a cor'tainment pressure reduction to decrease the pressure differential suoss the penetration. D. Ensure all containment cooling and filtering units are in operation to provkle maximum filtration of the containtnent atmosphere. Antwen CIA level: Russ119Agours,; Mc1Lrntnti B Yes Bank Question not seen none LP_.aVmhtG Sb3tMlYt).1 07 11-08 02.03 Qkjeg.tlytj Briefly DESCRIBE actions necessary if containment inte0rtty is violated as per AOP-12 and Tech Spec 2.6, Riffitnsti LP 071106, AOP 12 E9mmin111 question 071108,2.3 003 SRO level 55.43(b)(2,4 and 5)
OHest!grLi 37 Syst*mi Modei KAllemi BQ.lmPi SRQ)mVi l - 008 030 A3.01 3,0 3.1 l Hyptern; Component Cooling Water Modpf Mode Change System p9pgrjptiqui At>ildy to monitor automatic operation of the CC System; including control of the electrical'y operated, automatic isolation valves in the CC System. Qu9stioDi Which one of the followiwJ functions will occur because of a CIAS? A. Containment air coodng coil (VA 8A) valves HCV 402NU/C/D open.
- 8. RCP & CEDM cooler containment isolation valves HCV 43BA/B/C/D open, C. Detector well cooling containment isolation valves HCV 407NB/C/D open.
D. Storage pool heat exchanger outlet valve HCV 478 opens. i .-n 6fitW911 GlA_ltYtli Q991119A19H199) <till!Dinti A No Bank Question . not seen none LP_DWifthtfi .QDjtS11Y11U 07 11 06 01.05 Qbjtgilygj EXPLAIN the response of the CCW System to signals from the Engineered Safeguards Control System. Baffnngp2 LP 071106 99fnflit11tgj bank question 0711-06,1.5 001
Questignj 38 Systemi Mode; KAHem; ROJmpj SEQlrnp; 005 000 K1.01 3.2 3.4 Syyternj Residual Heat Removal System Modej Generic Desqdptigni Knowledge of the relat onship tietween RHR and CC System. Quellio0i During which one of the following conditions is the CCW heat load the greatest? A. Dunng full power operation. D. While shutting down the reactor from 100% to 0% power. C. While cooling down the RCS from 500 F to 400 'F. D. WhUe cooling down the RCS from 300'F to 200 F. n, C!A_!tytti gutstigungte_t1 Attachment; Antwen Yes Bank Question - not seen none O L.P numhtn _Qhlettiytt 07 11 06 01.01 QDjetityt; LIST the components cooled by CCW. Reititatti LP 071106 poinments; bank question 071106,1.2 001 one distractor changed. 4
Quest!gD_; 39 Sys1910; Mgdel BAtittml ROJmpi 489Jmei 078 000 K3.02 3.4 3.6 Sygttr!Li instrument Air System M_qdej Generic l QtsEf!Rti90; Knowledge of the effect that a loss of the IAS will have on systems having pneumatic valves and controts. Questioni What effect will a totalloss of instrument air header pressure have on SIRWT level indication? Assume that there is no actual change in SIRWT level and that the loss of pressure is of short enough duration that local air accumulators maintain pressure. A. Low level wdl a9j be indicated, low level alarms will ngj be received, STLS actuation will noj occur. B. Low level will be iridicated, low level alarms will be received. STLS actuation will 091 occur. C. Low level will 0g1 be indicated, low level alarms will 09] be roccived, STLS actuation will occur. D. Low level will be indicated, low level alarms will be received STLS actuation will occur. G/A_ltYalj Quetilon_s93rc,; Attachmertt; Answeri B Yes New Question none LP number: _Qhitttive #: 07 17 17 01.02 Qblectivgi Describe how the plant responds to a loss of instrument air in terms of how specific equipment is affected and how it affects overall plant operation and reliability. Befgtggqgi LP 071717, AOP 17 9.emmentti
Questlofij 40 Systerpj Mode; KAt itemi RQJmpi SRO impj 070 000 GEN 06 2.1 3.3 Systetuj Service water System Modej Generic ptperiptiont Knowledge of bases in technical specifications for limrling conditions for operations and safety limits. QVestiotti Tecnical Specification seguirements for operability of the raw water pumps, valves and piping are based on which one of the following? A. RCS and CVCS cooling loads during normal operation. D. RCS and CVCS cooling loads during accident conditions. C. Containment cooling loads during normal operation. D. Containment cooling loads during accident condnions. Mayttn CIA leygjj Question soumej AllMDRfilli No New Question none D LP_.D.umben _qpjective #: 07 11 19 01.02 OD.plygj Given the Technical Specification Manual, EXPLAIN the Technical Specification and bases associated with the Raw Water System. Btf9RDffj LP 0711 19, Tech Spec 2.4 basis 99mmtDisj SRO 55.43(b)(2)
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l Quegypaj 41 SYiitm; Mgdt; KA_ltem] ROjmpj SRQJimpi < 000 011 GEN-03 3.3 3.g ! Systentj Emergency Plant Evolution Modo; large Break LOCA Despllpflon; Knowledge of limiting conditions for operations and safety krnfts. Queglionj Evaluate the information below for each safety injection tank and determine what action is required prior to taking the plant from (node 3 to mode 2. LEWI, eaESWEE noBoM SI-6A 67.2 % 245 psg 2115 ppm St6B 69.5% 241 psy 2125 ppm SI6C 72.2 % 265 psig 2150 ppm SI-6D 74.6 % 258 pstg 2250 ppm A. Increase the levelin SI-6A B. Increase the pressure in SI-6B C. Decrease the pressure in SI-6C D. Decrease the levelin SI-6D Ansyren 9/A levtl; Quff.U9n_94WLGel Atta9htntD3 D Yes Bank Question - not seen none LP__numhts .pbittilyt_t1 07 11 22 01.13 Qhleltlytj Given a current copy of Ol St 1, explain the major steps, prerequisites and precautions for filling a SIT. Retettnqvi LP 071122 Technicalspecification Q.gtnmentp1 KA changed from 000 011 GEN-03 because components mentioned do not play a major role in large LOCA. Replacement K/A more SRO level,
$5 43(b)(2) v
=- . - - _ _ _ - -- __ . . _ - - - - . - -
9999tioDi 42 Syftefnj MQdt; K8Jtemi RQ.lmR1 SRO3P1 000 005 GEN 03 3.1 3.6 System.J Err +rgency Plant Evolution Modt; inoperable / Stuck ControlRod 093gIjptigm Knowledge of the hmeting conditions for operations and safety limits. Question; While operating at 100% power, a group 4 CEA dropped into the core. While stabi!izing plant conditions, an instrurnent malfunction caused RCS pressure to decrease rapidly. When the plant was stabilized a review of the charts indicated that the following conditions existed at one point during the transient:
- Reactor power was 95%
e RCC pressure was 1750 psia e RCf Tm was $51*F Based on the parameters above, which of the following actions should be taken w# thin i hour? l A. The plant should be placed in hot shutdown. I I D. Reactor power should be lowered to 70%
- 0. The dropped CEA should be fully withdrawn.
i D. RCS To houlds be ioduced to 542*F. Anac.6mnt; 8DtMG 9/A 19xtli RutilleILstuts,j Yes New Question Tech spec fgure 1.1 A 1,P_t19mt!au Qtth51tyi.!!i 07-12 25 01.04 QDjetilytj EXPLAIN the bases for each reactor trip. R9fe3nstj LP 0712-25, Tech spec foure 1.1 E9m'AtDill SRG $5.43(b)(2)
' 4 590- -
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1 p. - 2400 psia 1 w 560 - d w 2250 psia W i
,< z 4 2075 psia 6
o 5 50 - t i u i O- ; w , 8 u - . 540 - t 1750 psia ; 530 ' ' ' I 70 80 - 90 100 - 110 120 CORE POWER - (% OF RATED POWER) 1 c . 4 g ThermalMargin/LowPressureSafety OmahaPublicPowerDistrict figure
.v limits 4Pum000eration fortCalhounStation-UnitNo,i 1-1 Amenament no. A/,/p,_-t y ,92 ; ,, ~,,w--- --yr-,-,v. v +,ve e e g , e ,, e w- .w w ~w-r-- <.-e w--ern- , -w -w, v--,- -n -w .- , .-e, w .--,,s-~ e
l I Question ; 43 System; Mode; KAltem; ROlmp; SROJmp) m 000 057 EA1.04 3.5 30 System _; Emergency Plant Evolution Mode; Loss of Vital AC Electncal Instrument Bus pescription; Abddy to operate RWST and VCT valves. Question; Wdh the reactor shutdown, power was lost to bus 1 A3. If bus 1 A3 remains deenergized, which one of the following statements concoming power supply to the charging pump VCT suction MOV (LCV 218 2) and the charging pump SIRWT suction MOV (LCV 218 3)is conect7 Assume normal algnment of 480V breakers. A. Power will not be available to either LCV 218 2 or LCV 218-3. B. Power will be available to LCV 218-2 but not to LCV 218 3. C. Power will not be availablo to LCV 218 2 but will be available to LCV 218-3. D. Power will be available to both LCV 218-2 and LCV 218 3. a C!A_jtyel; QueJtioitsougej &Hac_hmtill; 80tW9G A Yes New Question none L.P numhtG _Obhcjtye #; 07 11-02 01.03 Q#}e_cjyoj EXPLAIN the automatic and manual controls associated with the charging pumps and boric acid pumps. Reference; LP 071102 90mmentti
l 4 QV**ll0R1 44 SYA19m] Mpde; KAJtt!n; ftOjmpj SfiQ_lmaj 000 078 GEN-04 2.1 3.7 3.yptgr!L; Emergency Plant Evolution Medpj High Reactor Coolant Activny Dessuptjgn; Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. Qutgijoilj I The technical specification limiting condrtions for operation on RCS actMty are based on the analysis of which one of the following accidents? A. Steam generator tube rupture. B. Reactor coolant pump siezed rotor. C. Doron dilution acedent. D. Loss of coolant accident. an ADRWWG CIA ltYtli Qu'ition soytqq; 8t!Aihmt011 No New Question none A LP_nufnktfi _QbJtElly931 07 15 33 03.04 Qhjegilyt; EXPLAIN how the steam Qenerator tube rupture analysis is used in determining the basis for Technical Specification requirements. Btf9BDS_91 LP 0715 33, technical specification bases 99famfatfl KA changed from 000 070 GEN 01. Task fitting original KA described chemistry and reactor eng6neer responsibilities at FCS. SRO 55.43(b)(2) v
Questfort j 45 System; Mode; KAJtertt; 80jmp; SBOlmp; 000 003 GEN 10 3.9 3.8 Syst*m; Ernergency Plant Evolution Mode; Dropped Control Rod pescription; Abildy to perform those actions; without reference to procedure; for all l casualties which require immediate operation of system components or ! controls. l t Question; l l The reactor is being diluted to criticality for the initial startup following a refueling outage. The t aurce range count rato is 8 times the baseline count rate but a positivo sus'.alned startup rate has not been achieved. Dunny this evolution CEA #1 falls into the core. Which one of the following directions should you give to the RO? A. Terminate the boron dilution until the dropped CEA can be recovered.
- 8. Begin emergency boration.
C. Manually reinsert all trtppablo CEAs. D. Manually inp the reactor. +, An1W,n CJAjtytt; queviton_tquite; Macil!ntilli D No New Question none LP_numbeti Rbhtily*Ji 07 17 02 01.03 Obhfityti Desenbe the major recovery actions of this AOP. Rettience; LP 0717 02 99fnm9Bt1; SRO $5.43(b)(6)
Quentloft; 46 Rystem; Modej Mllemi RRlmp; Sft_QJmp; 000 068 EA1.23 4.3 4.4 Syttem_; Emergent.y Plant Evolution Mode; ControlRoom Evacuation Desgrjplign; Ability to operate and rnonstor manual trip of reactor and turbino. QVeSiloni What actions are taken to inp the reactor and turbine dunng an AOP 06 control room evacuation? A. Prior to evacuation, the reactor and the main generator are manually tripped. The turbine is tripped frorr, the front standard. D. Frier to evacuation, the reactor is tripped and the clutch power supply breakers are opened. The turbine is tripped from the front standard. C. Prior to evacuation, the reactor is manually tripped and the EHC pumps are placed in pulbout. D. Prior to evacuation, the reactor is tripped and the MSIVs and the MSIV bypass valves are closed. 4 8tiffttil C_lA_lty,11 QWttugfLiqpicej Atta9hmtDil B No New Question none LP_n9mtMtti .Qhititirt_!; 07 17 06 01.03 Q)]ttilyt; Describe the major recovery actions of this AOP, Refttinsti AOP 00 G9mmtniti SRO level 55.43(b)(5)
l I Question; 47 system; Mode; KAjtem; RO_ImPJ REOjmPJ 000 024 EA1.23 3.3 3.3 Sytticij Emergency Plant Evolution Mo#ej Emergency Boration I p,5ct pMop; l Abihty to operate and monitor the CVCS centnfugal charging pump switches and indicators. Queptlpn; A laigo break LOCA has occurred. All safeguards components are operating as required. The primary operator informs you that he has received 1010 level alarms on both boric acid tanks. SIAS actuated 26 minutes ago and the present SIRWT levelis 52 inches. What action should you direct the primary RO to take? A. Place two charging pumps in pull stop, open LCV 218 2, close LCV 218 3, close HCV's 265,268 and 258. B. Place all charging pumps in pull stop, close HCV 238,239,240, and 249, open HCV 247,248 and 308. C. Closo HCV's 205,258 and 208, open HCV 308 or HCV 2988. D. Continue Emergency Boration until at least 30 minutes have elapsed. Ansmta CIA level; Rug 1MoJ1L1.ogr$9] &llaihfRtDli B Yes Bank Question not seen none LP numben _Q)JtEtly!JJ 07 18 13 03.19 93dtillYt; STATE from memory the two criteria listed in the Alignment of Charging Pump Suction to SIRWI floating step, either of which satisfies the require.Tient for termination of emergency boration. BefitfRE9] LI 7*1813, EOP 03 E93Dmintsj t question LR EOP 03 RO 004 slightly nmdified. Sb.43(b)(5)
pyestign; 48 Systerm Mqde; EAJ1erm RO_Jr!!R; S301rppj 000 029 EK3,12 4.4 4.7 Systemj Emergency Plant Evoiution Mode; Anticipated Transient Without Scram (ATWS) ppicfR119DJi Knowledge of the basis for actions contained in EOP for ATWS. 9Reslins What is the purpose of closing the VCT suction valve (LCV 218-2) durira an ATWS event? A. To prevent gas binding of the charging pumps. B. To allow Gravrty feed from the boric acid tanks. C. To prevent overpressurization of the VCT. D. To prevent VCT low level from opening LCV-218-3. 1 CIA levell Question source; Anaghmejit; Answen B Yes New Question none LP numbec _O_bjective !; 07 18 10 01.15 Objective; LIST the required operator actions during an Anticipated Transient Without a Scram condltion (ATWS).
Reference:
LP 0718-10 Cogigtents; K/A change from 000 029 EA2.07 because there are no trip breaker lights at FCS SRO level 55.43(b)(5) i
94es!!9n.] 49 Sy. stem; M9Ao; KA itemi 80 Impj S.AO_lrDRi 000-- 068 GEN-03 3.1 3.6
-SE tee j Einergency Plant Evolution Flode: Control Room Evacuation pes _SIjPil9B; Knowledge of limiting conditions for operations and safety limits.
9Ms1L9E 11 is 1000 on April 14,1997. A plant startup is in progress. Reactor power is 17% The turtaine generator was synchronized to the grid at 0940. Both fission chambers on the 'D' wide range log channel and the power supply for the 'B' wide range log channel have just failed, How will continued plant operation be affected if these channels can not be repaired 7 A. Power operations may contnue indefinitely but reactor power must remain above 15 %
- 0. Power operations may continue indefinitely but reactor power must remain below 70 %
C. The reactor must be in hot shutdown by 1000 on April 16,1997 D. The reactor must be in hot shutdow'n by 1000 on Apr!! 21,1997.
, m, f
C/A level: Question source;
Attachment:
Answe_n D Yes Modified Bank Question none LP number; Objective #: 07-17-15 01.06 9 hits 1Lve; Describe the Technical Specification LCO chahenged by this AOP. Reference; LP 07-17-15, AOP 15, Technical specircation 2.15 h att; significant modification of LR-AOP 15-RO 007 SRO 55.43(b)(2) original question not seen original question on next page
-e _,.
.. .. -,,. - . . , . . . . . _ - _. . . . - . . . ~ - . . - .-- - - .. . - . . .. . - . .,
*' LR AOP 15.RO 007 Oki si/> <[-
it is t000 on May 1,1992. A piant stcrtup is in progress. Reecer power is t 7%. N-The turtune generator was synchror@ed to the god at 0945. *D* Wide Range !
# Log Channet failed at 2200 on April 30,1992 dus to a f ailed fission chamber .
- *B' channel wide iange log channel has just f ailed low. WJw will this affect continued plant operations?
[1.0 pts) A. Power opwabone may connnue but Reactor power must remain grooter than 15W
- 59. The Reactor must be in Hot Shutdown by 1000 May 8,1992.
7 C. The Reactor must be tripped and EDP 00 implemented. D. The Reactor must be in Hot Shutdown by 2200 May 3,1992.
- The correct enewer le S (1.0 Pte]
. . ORIG:
- TECH 1
+ STYLE: NOTE: De NOT use with LR.AOP 15 RO 001, 003, 004 & 00s Rev.- 3/1/95; 7 min-Estimated
Reference:
71715,1.2;1.3;1.4;1.6; - AOP 15
'^'
K/A 015000 A2.02 RO3.1/SRO3.5 Memoran 5/27/92 1 I s 4 4 m.se , 4 __ ' .,,--.e.- . _ - + -y -- g-' -- - e.- * %, = --w y- y- y-
Qu,stion; 50 .S.ystem; Mode; KA itemj _R_Q_lmp; SBO_lmp; 000 057 EA2.14 3.2 3.6 Systent; Emergency Plant Evolution Mode; Loss of Vital AC Electrical Instrument Bus Descriptionf Abil i ty to determine that substitute power sources have come on line on a loss of initial AC. Questlo!1; With the plant operating at 25% power, the " Inverter A trouble
- alarm annunciated.
Prior to the alarm, the ERF indicated inverter "A" was at 117 volts and 25 amps. Which one of the following combination of indications would confirm that Al-40A is being supplied by the bypass transformer? A. Inverter output voltage zero. inverter output current zero, instrument bus current zero. B. Inverter output voltage zero, inverter output current 25 amps, instrument bus volta 0e 120 V. C. Inverter output voltage 117 V, inverter output cunent zero, instrument bus voltage 120 V.
- D. Inverter output voltage 117 V, instrument bus voltage zero, instrument bus current zero, i
A_nswer; C/A level; Qu_estion source. Attachment; C Yes New Question none I.E nemben ,O_bjective #: 07 13-04 01.04 Objective: Explain the Control Room indications for the systems and list the normal values for these indications. Bef_e.rpace; LP 07-13-04 gomments; 4
. =
QuestlA0J SI- S_yptemJ Modej KA item: fLO l m_pj S R O l m.pj
, 000 074 EA1.19 3.7 3.8 ily. stent; Emergency Plant Evolution M__st d ej inadequate Core Cooling Qes_qdption: Abildy to monitor AFW supply tank level indicators.
Qugs. tion; 5 A loss of offsite power occured with the plant in mode 3. The LSO noted the following conditions: D/G 1 and D/G 2 will not start FW-54 is not available FW 10 is supplying 50 gpm to each S/G EFWST levelis 100% If the present rate of feed is maintained to both S/G's, how long willit take to empty the EFWST7 , I A. 7 hours j B. 8 hours C. 9 hours D. 10 hours l Answer: CIA level; Question sourcej
Attachment:
D Yes Bank Question - not seen AOP-30 attachment A LP numben _Obiective #: 07-17 30 01.00 Obiective: Use the Emergency Fill of EFWST Procedure to makeup to the tank if it goes below Technical Specification levels and normal makeup is not avaliable. Beforencgj LP 07-17-30, AOP-30 9_q!Dments: bank question LR-AOP-30-RO-001 with minor wordirig change. 55.43(b)(5)
AOP-30 Page 22 of 23 Attachment A EFWST EmnNina Characteristics EFWST EMPTYING CHARACTERISTICS FOR VARIOUS INITIAL TANK LEVELS
^
100.00
4 I I 1 O
B "39.81 @ l ! l 5 s
*15.05 \h .
E 6.31 NN %%%- m
-3 O 2'5
' a 1007. (160~)
o N. NiN x N . +% , . 1.58 - - - t 707. (l 12'.) g 1.00 . , o 507. (80 ) g 0.63 .g % a 307, (48-)
% i g 0.40 0.25
% x 207. (32")
p - - y g o7, ( 3 g-) 0.16 [ ' ~ "' O 50 100 150 200 250 300 AUXILIARY- FEEDWATER FLOWRATE (GPM)
, End of Attachment A Part 2 R4
Questjp_rij '52 Systemi MoJi el E_A itemj RO_Lrnpj SP tmE 000 051 EA2.02 3.9 4.1 SISleEj Emergency Plant Evolution Mode; Loss of Condenser Vacuum QescriptJom Abiltty to determine conditions requiring reactor and/or turbine trip during a loss of condenser vacuum. Question: Which one of the following conditions could result due to operating with a vacuum less than or equal to 23.85 inches Hg with reactor power less than 30%? A. Overpressurizing the turbine casing. B. Overheating the final stages of the turbine blading, C Overpressurizing the condenser hotwell. D. Inducing vibration problems in the turbino. Atiswer: C/A lev _e]; Question source:
Attachment:
8 Yes Bank Question - not seen none L.P number: _Obiective #: 07-17 26 01.05 Objective: Given the caution statements and/or notes listed in this AOP, explain the reason for each.
Reference:
LP 07-17 26, AOP-26 Gomments.] KIA changed from 000 026 EK3.02 to prevent over sampling K//is associated with auto operation of CCW valves, bank question 07-17 26 a
Questiorij $3 Systemj Modej KA item: RO Impj SROImp; 000 015 EA2.10 3.7 3.7 Sytemj Emergency Plant Evolution Model RCP Motor Malfunctions p3sgjptiort; - Ability to determine when to secure RCPs on loss of cooling or sealinjection. Questionj The plant is operating at 65% pcwcr and all systems are aligned for normal operation. The "CCW Surge Tank Hi/ Low" alarm is received and the Auxiliary Building operator
- reports that he hears a loud rumbling noise from AC 3A. The pnmary RO reports fluctuating current and discharge pressure for AC-3A (The operating CCW pump)
The CCW surge tank levelis 8 inches and decreasing. Which one of the following actions should you direct the primary RO to perform in addition to the standard post trip actions once the reactor is tripped? A. Attemately mn one RCP at a time until cooling water is restored. B. Estabhsh raw water backup cooling to the RCP's within 5 minutes. C. Shutdown all RCP's within 5 minutes. D. Trip one RCP in each loop initially. Trip the remaining RCP's when high temperature alarms are received. Answer; CIA __ leveJJ Question sourgg,;
Attachment:
C Yes Bank Question - not seen none LP number: Objective #: 07-17 11 01.02 Obiective: Describe how the plant responds to a Loss of Component Cooling Water in terms of how specific equipment is affected and how it affects overall plant operation and reliabiltty.
Reference:
LP 07-1711, AOP-11 C9mments; bank question LR-AOP 11 RO 001
Que1Mofu 54 Snte,m_; lLodej KA itenji RO Impj SRO Im.pj 000 003 EK1.11 2.5 3.5 Syltemj Emergency Plant Evolution Mode: Dropped Control Rod pesgjp1Lom Knowledge of the long-range effects of core quadrant power tilt. ggestionj The plant has been operating at 50% power for the last 10 days. A group A CEA drops into the core. The reactor continues to operate. Which one of the following statements is correct conceming the Azimuthal Power Tilt factor (quadrant tilt) due to this event? A. The Azimuthal Power Tilt willincrease immediately following the CEA drop and will continue to increase for several hours. B. The Azimuthal Power Tilt willincrease imraediately following the CEA drop and then will decrease for several hours. C. The Azimuthal Power Tilt will decrease immediately following the CEA drop and will continue to decrease for several hours. D. The Azimuthal Power Tilt will decrease immediately following the CEA drop and then will increase for several hours. A_ttachmen._t; Anss n C/A level: Question source: Yes New Question none A LP number: Otpjective #: 07-15-32 02.03 Obiective: A CEA drop event Referencjt; LP 0715-32,0712-20, tech spec definitions Comments: K/A change from 000 003 EK1.11 to reduce overlap with question #42 and simulator scenario.
Qugstion;' 55 System 1 Mod *J taitem; ito impi sao imp;
., 000 015 EK1.01 4.4 4.6 SyMem ; Emergency Plant Evolution Mode.; RCP Motor Malfunctions pescdP319.0j Knowledge of natural circulation as applied to RCP malfunctions.
OMtstlogi During natural circulation, which one of the following temperature indications should be used with the 20"F subcooled and saturation curves of attachment 2, RCS Pressure-Temperature Limits. A. Tw B. Ta C. CETs D. Tm w Answe_n C/A level: Question source:
Attachment:
-C Yes Bank Question - not seen none LP number: Oblective #:
07 18-12 02.01 Obiective: GIVEN a copy of Attachment 2, EXPLAIN tts use to monitor RCS pressure and temperature limits.
Reference:
LP 0718-12, EOP attachment 2 Comments: K/A changed from 000 015 EA1.03 to reduce oversampling of EA1 K/A's
Rugslign.J 56 Sys_temj Mode; L9LLtegij R O Im pj SEQ.impj
-, 000 059 EA2.02 2.9 3.9 Syltemj F.mergency Plant Evolution Mode _; AccidentalLiquid Radioactive Release (LesGJRtJgm Ability to interpret 'ho permit for liquid radioactive waste release.
Que_sJlon; With the reactor in Mode 3, hot shutdown, the following conditions exist:
. Circulating Water Pump, CW 1 A is running
= Circulating Water Pumps, CW 1B & 1C are secured e Raw water Pump, AC-10A is operating
= Waste Monitor Tank *B'is being released Which one of the following actions should you direct to be taken if CW-1 A trips?
A. One of the other Circulating Water Pumps should be started imrrediately. B. The Waste Monitor Tank Release should be terminated imediately. C. An addnional Raw water Pump should be statted immediately. D. No immediate action is required as long as one Raw Water Pump remains in
^ service.
=
Ansvygn CIA level; Question source: Attachmenti B Yes Bank Question - not seen none L.P numben _ Objective #: 07 11-03 01.06 Oblective: Using the Offisite Dose Calculation Manual (ODCM) and Ol's as reference, explain the minimum circulating water requirements during radioactive liquid effluent releases.
Reference:
LP 07-1103, Ol-WDL-3 Comments; K/A changed from 000 026 EA1.01 to reduce overlap with simulator scenarios bank question LR WDL-SRO 002 SRO level 55.43(b)(4)
Question; 57- Systeml Modej EAltemJ Ro._ Jap; sRO Impj
- 000 005 EK2.03 3.1 3.3 Syttenij Emergency Plant Evolution Modej Inoperable / Stuck Control Rod Qes_cfjption; Knowledge of the metroscope as it applies to the inoperable / stuck control rod.
Ruts _tton; A CEA position deviation will be indicated on the SCEAPIS CRT display by: A. Steady, green bar graph indication of the group with the deviation.
- 8. Flashing, green bar graph indication of the group with the deviation.
C. Steady, red bar graph indication of the group with the deviation. D. Flashing red bar graph indication of the group with the indication. . m
Attachment:
&nswen C/A level1 Question source:
No New Question none C LP numben _O_ _biective #: 07-12-26 01.07 Oblective: Describe the methods of control rod position indication. Include the readouts and displays associated with each method. (CID No. 931191/02) Reh!trence; LP 07-12 26 Comments: M4 ______-.m. . _ _ _ _ _ _ ____________._______-_.m._____.___.____-______m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
Qugstiggj 58 Syst*!DJ Mode; KA_iltm.j ftO Impj SROimpj ; 4 000 069 EA1.01 3.5 3.7 l 1 Systenij Emergency Plant Evolution Modej Loss of Containment Integnty Descriptjonin Ability to operate and monitor isolation valves; dampers and electro-pneumatic devices following a loss of containment integrity. J Quq111o_Dj During power operations, which one of the following valves may be opened intermittently under administrative control 7 i A. Containment Purge Valve, PCV 742A. D. Hydrogen Purge valve, VA-280. C. Containment Purge Valve, PCV 742D as long as PCV-742C remains closed. D. Hydrogen Purge Valve, HCV-882. 5 w AnsweG CIA level: Question source; ettschment: D' No Bank Question - not seen none LP number: _0_bjective #: 07-14-03 01.09 Qldtetlye.] Given a current copy of the Technical Specifications, LOCATE the apphcable specifications and EXPLAIN the bases for the specifications. Beforence: LP 07-14-03,04 TS 2.6.1(a),2.0.3(a) CommejitJJ bank question LR-CO-RO 002
1 I QRet@D.] 59 Systeln; Modej KA iterm HQJ!ap; .SJOjrppj 000 055 EK1.01 3.3 3.7 l Systemj Emergency Plant Evolution d el Loss of Offsite and Onsite M_st Power ptsgIiR11901 Knowledge of effect of battery discharge rates on capacity as is applies to f station blackout. I Qge_stio_01 What is the basis for minimizing DC loads during a stat;on blackout? A. To assure that instrumentation and controls required to monitor plant safety functions operate for the required time period. B. To assure that instrumentation and controls required to start FW-54 operate for the required time period. C. To assure that it will be possible to utilize a source of c!fsite power that becomes available. D. To assure that it will be possible to start a diesel generator that becomes availab!a. Answe_g CIA level; Question sougej 6ttachment; A Yes New Question none 1,.P number: Objective #: 07-18-17 02.03 _ O_bjective: GIVEN a copy of Attachment 6, EXPLAIN the steps necessary to minimize DC loads. Beforensej LP 07-18-17, AOP/EOP attachment 6 Commentsj SRO 55.43(b)(5)
O_uestionj 60 System! Mode; KA_ item; RO_Lmp; SRO impj
~ 000 001 GEN-05 3.4 3.6 Systemj Emergency Plant Evolution Mode; Continuous Rod Withdrawat Descdption; Knowledge of the annunciator alarms and indications; and use of the response instructions during a Continuous Rod Withdrawal emergency.
Question; During a reactor startup, CEA group 4 rods are being raised above 50*. Motion of all group 4 CEAs continues when the RO releases the IN-OUT-HOLD switch. Which one of the following alarms would be expected during the event? A. "P DIL* , B. ' Continuous Rod Motion
- C. " Rod Position Deviation Low limit
- D.
- Rod Position Deviation Reed Switch" s
Ansyrer; CIA level: Questiortsource; Attachmen_t; B No New Question none LP numbef; _obhetJve #: 07-17-02 01.02 Obhet.Jv_ej Describe how the plant responds to a CEA or Control System malfunction in terms of how specific equipment is affected and how it affects overall plant operation and reliability. l Refe3nce; LP 0717-02, AOP-02 Comtnentsj 55.43(b)(6) i i v i l
Questiggj 61 Syptem; Mode; het.lem;i RQ.jmpj SHO_lmpj 000 076 EK2.01 2.6 30 Syntegt; Emergency Plant Evolution Modej High Reactor Coolant Activity Dg59tiption; Knowledge of process radiation mondors. Qugstion]
- Dunng a source check of a process radiation monitor performed in the control room per OP ST-RM-0002:
A. The monitor is considered operable due to the short time that it takes to do a source check. B. The monitor is considered inoperable, however no physicallogging of inoperability is required provided the operator is stationed at the monitor during the check. C. The monitor is considered operable provided the operator is stationed at the monitor during the check. D. The monitor is considered inoperable and formallogging of monitor inoperability is required when performing this check.
~.
C16 lty.el; Qggsil0J159u[ggj 6ttacim9nt; 6R9EtG Yes BankQuestion seen none B Qnuq1 ben pbjgc!Lve>_1 07-12-03 06.00 Q)]ggilyoj DISCUSS the watch standing requirements for the Radiation Monitoring dystem as they pertain to log keeping and surveillance requirements. BettitDit;- LP 0712-03 OP-ST RM-0002 Qormnty; bank question 07-12-03 001 Minor wording changes. last used on 07/26/96 4
Qugstlon; 62 Sys_ s tem; Model IM.ite_rn; Bo imp; SBp Imp; - 000 007 EA1.05 3.0 3.1 Systeml . Emergency Plant Evolution Mode; Plant Fire on Site Qqscf i ptlgn; Ability to operate and monitor plant and control roorn ventilation systems. QueAllo!1; 1 Which one of the following sets of signals will be sent to the control room HVAC fans if a fire detection sgnalis generated? A. All fans (VA-46A, VA-460, VA-63A, and VA-63B) will receive a start signal. B. VA-46A and VA-46B will receive a start signal. VA-63A and VA-63B will receive a trip signal. C. VA-46A and VA-468 will receive a trip signal. VA-63A and VA-63B will receive a start signal. D. All fans (VA-46A, VA-46B, VA-63A, and VA 638) will receive a trip signal. m ' .s CIA level: 9.uestion source: Attachment; Answan D Yes New Question none LP numben . Objective #.; 07-14-06 01.01d ggjective: State the functional relationships between the Control Room Ventilation System and the following systems: Fire Protection / Detection System. Bgittence: LP 0714-06 9_ofDfDtDtgj 4
- ep.
l l QugiliSIL; 03 Systemj Mode; KA item.; 80 Impj SRO Impj 000 040 EK3.01 4.2 4.5 Sy_ stem; Emergency Plant Evolution Mode _; Steam line Rupture pe!cIlptLon; Knowledge of the bases for operaton of steam line iso lation valves during a sieam line rupture. Qugsilojl; The technical spedfication requirement f r maximum allowable closure time of the main steam isolation valves is based on the analysis of; A. A steam generator tube rupture from full power, B. A steam generator tube rupture from zero power. C. A steam line break inside containment. D, A steam line break outside containment. 6Jigye!tn C/A level: Question soutce_; c
Attachment:
D Yes New Question none l,.P number: _Oklective #: 07 15-20 03.05 Obiective: EXPLAIN how the steam line break is used in determining the basis for Technical Specification requirenvr?s, Bettgejiggi LP 07-15-20, tech spec bases Commeritsj SRO 55,43(b)(2) v
8 Outitlo;1; 64 System: Mode; KA item _; RO_Jmpj SRO Impj 000 068 EK2.03 2.9 3.1 Systernj Emergency Plant Evolution Modej Control Room Evacuation
- Deschation: Knowledge of controllers and positioners as applied to control room evacuation.
QuMil9E Control has been established at Al-185 and controlis to be retumed to the control roorn. Which one of the following desenbes the proper sequence of actions to be taken? A. Each relay 43A,432,43C and 43D are reset first. Then the 43 transfer switch is retumed to REMOTE. B. Each relay 43A. 43B,43C and 43D are reset first. Then the 43 transfer switch is retumed to LOCAL. C. The 43 transfer switch must be retumed to the REMOTE position first. Then each relay 43A,438,43C and 43D are reset. D. The 43 transfer switch must be retumed to the LOCAL position first. Then each relay 43A,438,43C and 43D are reset An_swer: CIA levth Question soulce;
Attachment:
C No BankQuestion - seen none lf_D4GLb_en , Objective #: 07 12 02 01.05 Obiective: EXPLAIN the operation of the transu l switch 43 and the lockout relays 43A/B/C/D. Referencej LP 0712-02 LomR*_Q10 bank question 07 ~.2-02,1.5 001 order change, minor wording change. last used on 07/19/96
Quettfortj 65 System; M9de; KAltem; R0_lmpi SK0Jmp;
.-- - . 000 0 54 EA1.01 4.5 4.4 System,j Ernergency Plant Evolution M9dt; Loss of Main Feedwater DeScfjptioni Abiltty to operate AFW controts includin0 the use of altemate AFW sources during a loss of main feedwater.-
Que#tign; To locally start on0ino4 riven AFW pump FW 54, you must: A. Place the RuidStop switch to Run first. B. Place the Run Glop sw!!ch to Stop first. C. Place the Local-Renet-Remote switch to local first. D. Place the Local-Roset Remote switch to Reset fkst.
- ag AntlWifi 9/A_lfY311 QM99tl9RJ93tGt; AU4GhfD9flti D No BankQuestion seen none LP. IIMfDktf1 _QD}991lY211 07 11 01 01.11 Q h t[ygj EXPLAIN the actions necessary for a local start of the diesel driven AFW pump, FW 54.
Rtft19n53; LP 071101 f Qgg}mentti bank question 071101,1,11001 order changed. last used on 07/t2/96
OpostjoDj 66 Systeml Mode: MA itenu RO_jmpj SfLO impJ
.~ 000 037 EA1.01 3.7 3.6 Sysiemj s Emer0ency Plant Evolution Model Steam Generator Tube Leak pescripflo.0J Ability to monitor maximum controlled depressurization rate for affected S/G.
@teSil9BI Of the following, which response applies to the cooldown indicated on the attached chart?
A. All adowable RCS pressure / temperature combinations were met between 1100-1200. B. Low Temperature Overpressure Protection was in effect between 10001100. C. The RCP NPSH requirements were violated. D. The allowable RCS cooldown rate was exceeded. ." ;;a
&Demn CIA level: Question source; 6ttachment:
Yes Bank Question not seen P105/T123 chart, EOP attachm D LP number: Obloctive #: 07 11 20 02.02 Qblectigj EXPT.AIN the basis for the RCS heatup and cooldown curves and STATE the limits.
Reference:
LP 07-1120, EOP attachment 2 99mm*J31sj bank question LR-TS-RC-RO 005 Remove all reference to which instrument to use from attachment 2 to prevent look up for question 55. SRO 55.43(b)(2) w
(
- LR.TS RC RO 005 Of the sonoveng,wnica response appues to tne cocoown inoicateo on tne anoched chart? (1.0 p*]
O TSO 996 "O *Jo '?So '%0 97 2 2000 ??SO ??00 4 6 i ! i i ! ! l i. 4 P10E. Pressursaar Peginurn 0 2500 asia .
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t I 4 I i 6 ) ! g ..a un in 2.a as ma un un su a A. AR enowable RCS pressure / temperature combinanons were met between 1100 1200. B. Low Temperature Overpressure Protecuon wee in effect between 1000 1100. C. The RCP NPSH requirements were voleted.
- 4. The snowable RCS cooldown rate has been exceeded.
The correct snower is 0 [1.0 pt] OnlG: TECH: STYLE:
EOP/AOP ATTACHMENTS Page 4 of 101-Attachment 2 RCS Pressure-Temoerature Umits 2500 i ; , , , , , i i,ii , , i,, ,;;i
- I -
~ f-LOWEST SERVICE / ~
/ TEMPERATURE /
I ' 2000 l d - 100'/HR I 200* - f/J COOLDOWN I SUBC00 LED ' v
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M 1500 /f /> M j / U ~ PORV - /
~
/
M TRIP /
/ N PORY / -
/ PRE-TRIP /
{ - g 1000 f g - SHUTDOWN / RCP er,: , N _ C00UNG / NPSH _ E -
/
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~
_- / [ j S BC00 LED _ 250- _- g. ,, sg. SuecOOLED - 0
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' '~ m (SATURATION) i 100 200 300 400 500 600 RCS Tc . TEMPERATURE (*F)
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Quesjio!L; 67 Systemi _M_odej }%jie_m; RO Imp; Sff0_ Imp; 000 007 EA2.04 4.4 4.6 Systern; Emergency Plant Evolution Mqdej Reactor Trip DescIjpt{on; Abiltty to manually trip the reactor and carry out actions in ATWS EOP if reador shoud have ldpped but has nol done so. Qutallollj Charging pump CH-1 A is disassembled for replacement of the block. Charging pump CH 1B is disassembled for packing and plunger replacement. A reactor trip occurs and control rods 2-25 and 2 26 remain at 126 inches withdrawn. All attempts to insert these two rods have been unsuccessful. Charging pump CH 1C is running. Which one of the following actions should you direct the RO to perform? A. Rapdly depressurize the RCS to below HPSI shutoff head to allow more rapid injection of boric acid. B. Emergency borate with CH-1C until reactor power is less than 10"% power. C. Emergency borate with CH 10 until the reactor has achieved a calculated 4% shutdown margin. D. Emergency borate with CH 1C until 20 minutes have elapsed. Ansvyen CIA level: Question source: attachment: C Yes Bank Question not seen none 1.P numtrle Objective #: 07 16-10 01.05a Obiective: GIVEN a set of plant conditions and a copy of the EOP resource Assessment Trees, DETERMINE the correct success path for any of the following safety functions: Reactivity Control fteferencel EOP-00 Comraents: bank question LR-EOP-00-RO 010
Questi.oDj 68 System; Mqdej KAJem_j RO lm.pj SRO lmp: 000 033 E.K3.01 3.2 3.6 SY!1tfD.1 - Emergency Plant Evolution Modej Loss ofintermediate Range Nuclear Instrumentation D9sgriipiloEi r Knowledge of the basis for termination of startup following loss of intermediate-range instrumentation. 9LatSfloRI A plant startup is in progress. Criticality has just been achieved at 5 x 10'5 %.24 hours prior to startup, 'A' wide range was removed due to a failed fission chamber.
*B' and *D' wide range channels have just failed low due to a voltage spike on DC bus #2.
The reactor must be tripped because; A. 'D' channel must be operable for Al-212 during critical operations. B. SCEAPIS requires at least 2 wide range channels greater than 10" % to enable the rod block function. C. Startup rato protection will be inadoquate with 3 of the 4 wide range channels failed low. d '" D. APD trip requires at least 2 wide range channels greater than 10 to be enabled. Answen CIA level: Question source: A_ttachment: C Yes New Question none LP numben Objective #: 07-12-25 04.01 OEtctive: Using the Technical Specif6 cations as a reference, EXPLAIN the time limitations associated with placing an RPG trip unit in a tripped or bypass condition. Refelencej LP 0712-25 gotDmental SRO level 55.43(b)(2)
I 1 i QuestioD.; 69 Syst*m; Mode: MAL [terD; RO Imp; $E0_ltnpl
- 000 038 EK3.01 4.1 4.3 Systemj Ernergency Plant Evolution Mode _; Steam Generator Tube Rupture Deigdatlan; Knowledge of the reasons for equalizing pressure on primary and secondary sides of ruptured S/G, quest [qpj A potential consequence of reducing RCS pressure far below the ruptured steam generator pressure following a steam generator tube rupture is:
A. Pressurtzed thermal shock of the reactor vessel. B. Increased release of radioactivity to the environment, C. Rupture of a tube in the intact steam generator. D. Reduction of the reactor coolant boron concentration. m Answer: C/A level: Question source:
Attachment:
D Yes New Question none LP_rwmber: _Qtitissitve #: 07-15 33 02.05 Obiectlye_; EXPT.AIN the operator actions required to mitigate a steam Generator tube rupture event. Refelanse.J LP 07-15-33 Comments: K/A changed from 000 038 EA1.12 to reduce overlap between simulator and written exams. SRO 55.43(b)(5)
l Ques!lRD.] 70 Systerq; Mode; KA itemi 80 Impj SHO__Impj 000 032 EK2.01 2.7 3.1 Systgmj Emergency Plant Evolution Modej Loss of Source Range Nuclear Instrumentation peJLEIlptioDi Knowledge of power supphes; including proper switch positions as applied to a loss of source range nuclear instrumentation. Questign; Which one of the following statements is true conceming operation of the source range portion of the WR Ni system? A. The fission chambers ur each channel are located one above the other in the detector well. Only one detector is powered above 1000 CPS. B. The fission chambers for each channel are located one above the other in the Cetector well. Both detectors are powered above 1000 CPS. C. The fission chambers for each channel are located side by side in the detector well. Only one detector is powered above 1000 CPS. D. The fission chambers for each channel are located side by side in the detector well. Both detectors are powered above 1000 CPS. Answer: CIA level: Question.jiource.
Attachment:
D No New Question none LP number; O)Jactive #: 07 12-18 02.03 Obiective: State the function and explain how each major component affects the operation of the WR NIS channel. Beforence; LP 07-1218 Qommentsj WR NIS was modified last outage.
l 9Vestigill 71 S.ystem; Model t%,itemj RO_Jmp; SRO_JmPJ 000 027 GEN-05 3.3 3.3 Syttemj Emergency Plant Evolution Mode; Pressurizer Pressure Control System Malfunction ptsqrip.tlon; Knowledge of the annunciator alarms and indications; and use of the response instructions. Qu_t31ioJ11 The reactor inpped 20 minutes ago. T he following conditions are observed:
. ' PRESSURIZER PRESSt;RE OFF NORMAL Hi-LO' channel X and Y are in alarm.
. PRC 103X (controlling channel) indicates 2160 psia and stable
. All backup heater in auto and energized
. LRC-101Y (controlling channel) indicates 60% and stable
. LRC-101X indicates 43% and increasing slowly
. Ll-106 indicates 28%
. Letdown flow is 26 gpm
. One charging pump is running
. T indicates 533'F, Tw indicates 534*F, both are stable Select the probable cause and the action you should direct the RO to take to restore RCS pressure:
A. Low level on LRC-101X is maintaining the B/U heaters on, place the pressurizer heater cutout channel select switch in chanel Y. B. The bistable for the B/U heaters needs to be reset, place the control switches for all B/U heaters to reset and back to auto. C. LRC-101Y has malfunctioned causing the B/U heaters to remain on, place LRC-101X in service. D. PRC 103X has malfunctioned causing the B/U heaters to remain on, place PRC-103Y in service. Answer: CIA level: Question source;
Attachment:
Yes Bank Question - not seen none C LP nugtben Oblective #: 07 11 20 05.04 Objective: Given a current copy of ARP, EXPLAIN the alarms associated with the RCS Instrumentation System and the required actions.
Reference:
LP 07-1120 Comments: bank question LR-TDB-RO 024
-_- - - - ______________m__ _ _ _ _ _ _ _ _ _ _
gytetiorij 72 System; Mode; K/LLtem; ftQJrnp] SfiQ_lmPi 000 009 EA1.01 4.4 4.3 Systemj Emergency Plant Evolution Modej Small Break LOCA Qtsstintippj Abthty to rnonttor RCS pressure and temperature during small break LOCA. Question: Following a 10 week run at 100% power, the reactor has tripped due to a small break LOCA. Power has been lost to instrurnent bus 'B'. RCS pressure has stabilized at 1350 psia arid HPSI fluw is 65 gpm.15 minutes have elapsed since the innlation of the event. CPHS occurred 4 minutes ago. RCS Ta is 495'F. How would you expect Tm to respond over the next 30 minutes with no operator action? A. Tu willincrease because the heat removal by the break is less than the decay heat being produced. D. Tu will stabilize due to a balance between decay heat and heat removal by the break. C. Ta will decrease because the heat removal by the break is greater than the decay heat bein0 produced. D. There will be no indication of Ta due to the Instrument bus failure. m- c t Cjultytjj QWgglignpyfiej MaqllstDt AtitERG A Yo1 Modified Bank Questbn none LPL!19]DhtG _QhltEllYf_fi 07 15 23 01.02 QDj2EllYf3 EXPLAIN how the decay heat removal capacity of the break affects plant response. fitttfynsgj LP 0715-23, EOP 20 E9mmentti Modification of LR EOP.20-RO 006 SRO 55.43(b)(5) onginal question was not seen originalquestion attached
. , . , a c - m
+ LR EOP.20 RO 008 % / 3 f /) f roio.n, a io..... tun ai ious. m.., in. . actor na. vippeo ou. io a i.o.a oi Otinste Power and a Small Break LOCA. RCS possure has statalized at 1350 psis ard HPSI flow is 85 gpm. 15 trunutes have elapsed since the irutistion of the I { event. CPHS occurred 4 minutes ago. RCS T,g i s 495'F. l Of the foHoeng, how would you espect Tcold to testond vor the next 30 rrenutet, with no oporstor action? [1.0ps) ti. T gog mil continue to trend (bwn because the steam being dumped to the condenser comtaned with the HPSI flow cut the trenk is enough to contmus to coc' the teactor (kwn D. Tgg mil continue to trend cbwn because the flow out the brott is arough to remove the arrount of decay heat that is in the reactor cr te C. T,g mil begm to statalize because the enthalpy at 495'F ard the given hole eJze mil relieve the heat produced by the reactor
- 4. T,g el! tegm to increase because the t.OCA size is too small to totieve the amount of heat being produced by the teactor The correct answer le O (1.0 pte]
\-
Rev. 3/1/95; 4 min Estimated
Reference:
4-46-06S4t 71818A4 EOP 20; HR3 K/A 000074,EK1.03,RO4.5/SHO4.9 Ward 4/3/91 _ 0 .we w - .- - , _ _ ..m _.--. .,,. _ _... ,_ ,
09etil9Aj 73 SV7temt Medei KA.ittini R0_lmpj SjiO_imPi _ 000 008 EA1.01 4.2 4.0 Systemj Emergency Plant Evolution Model Pressurizer Vapor Space Accident ptscLipil00j Abihty to operate and monitor PZR spray block valve and PORV block valve. QJestipfli Due to excess leakage of PORV PCV 1021, HCV.151 was closed. Which one of the following statements is true conceming operabikty of HCV.1517 A. HCV.151 is considered operable anytime that it is closed. D. HCV.151 is considered operable as long as power is available to its motor operator. C. HCV.151 is considered inoperable when closed due to potential boric acid buildup. D HCV 151 is considered inoperable when closed due to potential thermal binding. 99ttiLqttfource: Attaghment: 6Rt.LvtG 9/A levell D No New Question none LP numbtG .Q.bititlytJj 07 11 20 02.00 QblicilYti DISCUSS the Technical Specifications limiting conditions for operation that apply to the RCS. Referenitj LP 071120 Q2mmtDigi K/A changed from 000 008 GEN-06 to reduce overiap with the simulator scenarios.
qutstlottj 74 Sy11emj Mode; LAdlem; [1Qimpi $JLQJ!pp;
- 000 009 EA2.37 4.2 4.5 Sygtemj Emergency Plant Evolution Mode; Small Break LOCA pesstjpilon; Ability to deterrnine the existence of adequate natural circulation during small break LOCA.
QM9stion; A LOCA has been diagnosed and the proper procedure steps are being performed. Plant conditions are as follows: TIME 0000 0010 0020 0030 PZR pressure, psia 2100 1800 1000 1400 CETs, 'F $52 555 558 562 Tw,'F 552 555 558 562 Ta 'F 532 535 538 538 S/G pressures, psla 900 930 950 950 S/G WR levels, % 70% 60% 50 % 50 % Has subcooled natural circulation been developed? A. No, RCS subcooling cnteria has not been met. D. No, RCS temperature trends enteria has not been met. C. No, S/G level crtteria have not been met. D, Yes, all criteria for subcooled natural circulation have been met. Gmestion soufce; Attachment; Ansatu CIA le v_eJJ Yes Bank Question - not seen steam tables D WJWHLbjn ,Q)JgitiveFj 07 18 13 03 05 QbititlYej STATE from memory the four indications used to venfy the development of Subcooled Natural Circulation. liefeEnS_ej LP 071613, EOP 03 floating steps Q3fnments.j bank question LR EOP-03 RO 009 some wolding changes "4
I i Questionj 75 Svstem; Mode; KAJtemi fiq1mp; SRQlmp; U.10 025 GEN 06 3.7 36 Systern; Emergency Plant Evolute Mode; Loss of ResdualHeat Removal System Descripflon; Abakty to locate and operate components; including local controls Questiori; Which one of the folowing stattments correctly describes the cochng flowpath to be used in a loss of shutdown cookng caused as a result of an inoperable LPSI header downstream of FCV 320 A. Charging pumps take a suction from the RCS loop and discharge through the shutdown cooling heat exchanger back to the RCS. D. HPSI pumps take a suction from the RCS loop and discharge through the shutdown cooling heat exchanger back to the RCS. C. Both a containment spray purnp and a HPSI pump take a suction on the LPSI pump suction The containment spray pump discharges through the shutdown cooling heat exchanger hack to the HPSI pump suction The HPSI pump discharge flows back to the RCS. D. Both a containment spray pump and a HPSI pump take a suction on the LPSI pump suction. The HPSI pump discharges through the shutdown coohng heat exchanger hack to the containment spray pump suction. The containment spray pump discharge flows back to the RCS. Cl&jtygli QuestloILisoggej AHa_qtent; Af1tW9G C Yes Bank Question not seen none kP_,numlMG ,Qb}fGilYR_#.J 07.t 719 01.02 Q)]egtlyoj Desenbe how the plant responds to a Loss of Shutdown Cooling in terms of how specific equipment is affected and how it affects overall plant operation and reliability, Reitrentyj LP 071719, AOP 19 CpmmgDisj bank question LR AOP 19-RO 007
1 l Otjestiort; 76 System; Mode; KAJtem; 80 imp; SROJmp; m 000 022 EA2.01 3.2 3.8 Systernj Emergency Plant Evolution Mode; Loss of Reactor Coolant makeup p9tGription; Ability to determine whether charging line leak exists. Que_stlogj Given the following plant conditions, what is the most probable cause? e RCS pressuro is 2050 psia and decreasing
- PZR levelis $6% and decreasing
- Letdown flow is 26 gptn
* - Charging flow h 120 gpm
. Reactor power is 99.5% and constant
- Ta is $42*F and constant
+ Tw is 594*F and constant
- RM-054A/B indicate 500 cpm / 700 cpm both are constant
- Containment sump is 18' and constant
- VCT levelis 47% and decreasing A. An uncontrolled heat extraction.
B. An RCS leak inside cor.tainment.
+
C. A S/G tube leak in tho 'B' S/G. D. A charging header teak in room 13. Ariswet; C!A It.Y991; Rue, tion sourge; att49hrtit!!!; D Yes Modified Bank Question none LEllWilhem _Okjt_ctlyo_#j 07 17 22 01.02 9)]tqtlytj Describe how the plant responds to a Reactor Coolant Leak in terms of how specific equipment is affected and how it affects overall plant operation and reliability. [tjitittitej LP 0717 22, AOP 22 Lomm_ints; bank question LR AOP 22-RO 005 modified. SRO 55.43(b)(5) y originalquestion not seen original question attached
LR AOP 22.RO OO5 won ine sonomrq p. ant conoitions, anst is ine most procao2e cause r [1.0ptj e RCS Pressure 2050 psia coeressing
- PZR Level 56 % decreasing Letdown Flow 26 gpm
- Charging Flow 120 gpm
- Char 0tng pump amps 49 amps [a!!]
- Reactor Power 99.5%
- T coid $42'F constant
- T hat $94*F constarit
- RM.064 A 500 com constant
- RM.054 0 700 cpm constant
- Contalnment Sump > 36' constant a VCT Level 47 % decreasing A. An uncontrolled heat extraction, because the Tcold and That are too low for this power level.
4
- 8. A RCS leak some what greater than 90 gpm, because of the mismaten between Charging and Latesown.
C, A S/G tube leak in *0* S/G, because of the high countrate on its Radiation Monitor, vil A leak in the charging header inside the containment, because the Charging pump amps are too low. The correct answer is D [1.0 pt) ORIO: - TECHI STYLE: NOTE: Do NOT use with questions LR.AOP 22 RO 001, 002, , 003, 004, 006 REV 7/9/96; 4 min Estimated
Reference:
71102,1.0;4.1; 71722,1.2; AOP 22 K/A 011000,K1.01,RO3.6/SRO3.9 Ward.3/7/91
-yv- - - + - w y ea- - -
Quest!qaj 77 pyttem; Medej K6Jtem; B0Japa GRO_lmp;
. 000 060 EK3.03 3.8 4.2 Sygtemj Emergency Plant Evolution Mqdej Accidentalgaseous waste release pescI!pjlopj Knowledge of the basis for actions contained in EOP for accdental gaseous waste release, putstjoD1 Why does AOP-09 direct that only one Auxiliary BuilJing exhaurt fan and no Auxiliary Building supply fans be operated when high radioactivity is detected in the Auxiliary Building 7 A. To achieve the maximum available negative pressure in the Auxiliary Building.
B. To minimize the radiation release to the environment. C. To prevent exceeding the flow capacity of VA 60. D. To assure a negative pressure while allowing doors in and out of the Auxiliary Building to be opened. A_DIYt9_G CIA level: Questioayo!Atku 6.tleib!OAD1 0 Yes New Question none LPJMmh*E Q)]ective #: 07 17-09 01.03 Qhigillygj Describe the ma}or recovery actions of this AOP.
Reference:
LP 0717-09, AOP 09 TBD C_ommeatsj SRO 55.43(b)(4) M'
QW9ftl9n.1 78 SYstemi M951*i Mall *mi RQlmp; SRQlrpp; ,.= 000 025 EA2.03 3.6 3.8 Syilemj Emergency Plant Evolution M_ ode; Loss of ResidualHeat Removal System Desgjplion; Ab;hty to determine or interpret increasing reactor buikling sump level. Quettlon; Which one of the following events would produce a SIAS actuation with a decrease in SIRWT level but would not produce a significant increase in containment sump level. A. A rupture of the pressurtzer surge line. D. A rupture of a main feedwater kne,just upstrearn of one of the steam generators. C. Significant leakage through reactor vessel head vent valves., D. Significant leakage through SDC loop suction valves HCV 347 and HCV 348. <c% 8 DIE *f; 9/6 levtl; Qqqstion sourct; 8.ttachment: Yes New Question none D LP numil'I; _O D itlYf.fi 07 15-23 02.09 QMettlyt; EXPLAIN the operator actions taken to mitigate an interfacing system LOCA. EtfpIgnggi LP 07-15-23 Cjmwn*31sj 4
QuestlQrtj 79 Systemj M9dtj KMitsj flojmpj Sflo_Lmp; 000 058 EK1.01 2.8 3.1 Systern; Emergency Plant Evolution .Modej Loss of DC Power Dgscription; Knowledge of battery charger equipment and instrumentation as applied to a loss of DC power. Quest 19p] The plant tripped due to a loss of 161KV and failure of fast transfer. At the time of the trip, CA 10 was the only major equipment tagged out. Following the performance of the standard post trip actions, the following condittons exist:
. D-1 failed to start.
- D-2 supplying bus 1 A4 e instrument air pressure is 80 psig and dropping e #1 Battery charger troublo alarm, in alarm e All other safety functions are satisfied Which one of the following actions should you, as LSO, take?
A. Enter EOP-07 to restore power to the #3 battery charger. B. Enter EOP 20, MVA-DC to restore a battery charger to DC bus #1. C. Enter EOP 20, MVA lA to restoro an air compressor. D. Enter EOP 02, energize bus 1 A3 to restore an air compressor. C/A leyeli QuestLgfLsELcej Altschment; AusE G C Yes Bank Question - not seen none LP_numhts .9#1titive #: 07 18-10 01.06 O jegityt; GIVEN a set of plant conditions, DETERMINE if the Standard Post J Trip Actions (SPTA's), the Optimal Recovery Guidelines or the Functional Recovery Guideline (FGR) should be used. Refelongej LP 071810, EOP-02. EOP 20 qqsuttentt; bank question LR-EOP 02-RO 002
Questignj 80 System 1 Mpdtj KAt jtem; RQJmp; SRQlmpj
.- 000 065 EA1.03 2.9 3.1 Systernj Emergency Plant Evolution Modej Loss of Instrument Air Pg!Cliptiggj Ability to rnonitor restoration of systems served by instrument air when pressure is regained.
QV95169m The plant expenenced a loss of instrument air pressure. Air to the operators for the foodwater regulating valves, FCV 1101 and 1102, was automatically isolated. What action,6f any,is required to reestabhsh automatic operation of the regulating valves following restoration of normalinstrument air pressure? A. No action is required. Automatic opetation wil be regained when pressure is restored. B. Automatic operation can be restored by operation of the reset switch. C Automatic operation can be restored by placing the SPEC 200 controller in manual and then back to auto. D. Automatic operation can be restored by placing the auxiliary controllor in manual and then back to auto.
,r, CIA level; Glue _ption sourcej AttaghfDent;
&DsytG B No New Question none LP_n9mtria .9)ltstive #:
07 11 11 02.03 Qtije_qtlytj EXPLAIN the automatic features and interlocks associated with the feedwater components, Etfir_engel LP 0711 11
%g8*
--+,
l Quettfort.j 81 Systemj Modej MAJ1emi RQ.lmni DRolmp; 0C') 028 EA2.13 2.9 3.2 Syyjernj Emergency Plant Evolution Mode; Pressurizer Level Mattunction pescrip!1oni Ability to deterrruno actual PZR level; given uncompensated level wit 5 an appropriate graph. ) Question] The reactor coolant system is being cooled down and depressurtzed for a refueling outage, l&C technicians have placed a dummy signal on channel 101X . The following indications exist: Li 101Y indicates 45% Ll 101X indicates 48% LI 106 indicates 33% TI 108 (pressunzer 14uid) indicates 300 F. Which one of the following statements is true: A. Channst Ll 101Y and Channel Lt 100 both indicate that the pressurtzer levelis above the heater cutoff level. B. Channel LI 101Y indicates that levehs above the heater cutoff level but channel LI-100 indicates level below the cutoff level. C. Channel LI 101Y indicates that levelis below the heater cutoff level but channel Ll. 106 indicates level above the cutoff level. D. Channel Ll 101Y and Channel Ll 100 both indicate that the level is below the heater cutoff level. 8_01Een 9/B.J8Ytl; Qu.esilon source;
Attachment:
New Question TDD sections lit.1a,111.2 D Yes LP_Q9mbtn .Q)Jegtlyel 07 11 20 04.00 Q)]tfilYrj When given specific plant conditions, EXPLAIN operating principles to predict response of Reactor Coolant System (RCS) Instrumentation. Referenew: LP 071120 TDB sections 111.1a,111.2 99!!LfD9atti
. + - _ gw ..
?
Fort Calhoun Station i Unit No. 1
?
i TDBill.1.a > TECHNICAL DATA BOOK PROCEDURE , 4
Title:
TEMPERATURE CORRECTION FOR PRESSURIZER LEVEL
!- INDICATORS LI 101X/Y 1
Setpoint/ Procedure Form Number (FC-68): 39172 ,- erS s/ Reason for Change: To annotate the TDB Figures-to indicate that they are related to ongoing commitment 790101. 4 Contact Person: Deb Matthews T 4f+ .<
- v 09-14-92.-4:00 pet R3 ISSUED:
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5 Fort Calhoun Station : Unit No. 1 l t I TDB lli.1.b i - TECHNICAL DATA BOOK PROCEDURE ,
; Titles- PRESSURIZER LEVEL CORRECTION (DENSITY EFFECTS ,
CORRECTION) l p Satpoint/ Procedure Form. Number (FC-68): 39172 4m, Reason for Change: To annotate the TDB Figures to indicate that they are related to ongoing commitment 790101. 4 Contact Person: Deb Matthews i t 8 1 4
%d ISSUED: 09-14 '4300 pm R3 . - - ,.n.. - ~w r r,. ,v. .w , ~. . .,,w,r ar , .,,e, .,-w ,a.,,.,,~- -,,,--.,-,+-,nn n.,, --,--..nne- ---e-.,.n.,n.,.,-n-m-,
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i FORT CALIIOUt? STATION TDB-IZZ.1.b TECIUlICAL DATA BOOK PROCEDURE PAGE 1 OF & , PRESSURIZER LEVEL CORRECTION 1 (DENSITY EFFECTS CORRECTION) L-101 100 ' j' l!il . l l 8
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* (TDB9204E.WPG) ,
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^ Fort Calhoun Station
. Unit No. 1 TDB lll.1.c . TECIUi1 CAL DATA BOOK PROCEDURE o 4
-Title: PRESSURIZER LEVEL CORRECTION (DENSITY EFFECTS CORRECTION)
Setpoint/ Procedure Fom Number (FC 58) : 39172 _<e'r , T I Raasan for Change: To annotate the TDB Figures to indicate that they are related to ongoing conmitment 790101. a l Contact Person: Deb Matthews
~
l i l . l . c, !
- ISSUED:. 09-14 92 l4:00 put R3
. - -,2__ - - . . - . . _ . ~ . _ . . . , - - . _ -
I I FORT CAL 110UN STATION TDB-III.1.c TECHNICAL DATA BOOK PROCEDURE PAGE 1 OF 1 PRESSURIZER LEVEL CORRECTION- i
- (DENSITY EFFECTS CORRECTION) L-101 2 100 , ,
l 2 i- ! I i l. i 1, . 9 l i i l AMBittii TEMPERATURE = '2007 ii { , NOTE: temperature obioined from - 90
- I l (RT coints T-887, BBS, 889,1 . I ! ! . .
"d '
[' 85
! bOMMITMElli 10#790101 l !2100 psio-[ l l 80 ' '
3 ! l ! i l! ' l i
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- n. 35 :
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i a : i 4 , 2 30 ;, 7 U 4 25 ' 20 i /7 to' S 3 ' 0' / _i 0 10 20 30 40 50 60 70 80 90 100 5 15 25 35 45 55 65 75 85 95 INDICATED PRESSURIZER LEVEL (". Top Span) (TDB9204P.WPG) FC/TDB R3 4
. . , , , -- - - - , _ . , , . , ...,,-n-,-,,-nn .cn...- , , . . . - , , , , , - ,,w,,n.. ,-,,--.,n.,., , . - , , .,_,.,,_--.n.,c.. _ - - . - , -
- . . . . . . _ . - .- ..- -. - - . - ...-. ~ .. - - -. _ .-. _ - ..-- -.- - . - .-. . -
,m Port Calhoun Station l Unit No. 1 4
)
1 TDBlli.1.d j TECHNICAL DATA BOOK PROCEDURE
Title:
PRESSURIZER LEVEL CORRECTION (DENSITY EFFECTS CORRECTION) d ' Satpoint/ Procedure l Form Number (FC 68) : 39172
- r;h Reason for Change: To annotate the TDB Figures to l indicate that they are related to 1 ongoing commitment 790101. l Contact Persont Deb Matthews l
l l 1 4 w. ISSUED: 09-14-92 4:00 pm R3 l
FORT CALil0UN STAT 20N TDB II2.1.d TECID12 CAL DATA BOOK PROCEDURE PAGE 1 OF 1
- PRESSUklZER LEVEL CORRECTION (DENSITY EFFECTS CORRECTION) L-101
'0
- , i ! i i : I' '
i i 4 1 i l ! j ! l ' ! l l I 95 I AWBIENT TEMPERATURE = 250T 1 l l ! ll i NOTE: femperoture cotoined from 90 -; [RT points T-887, 888. 889 , ! i ' j ' I j , 3 end 890. ! I I i I I l COMMifMENT 10/790101 L 2100 psia-] !' !
,[
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( ; Illl1050pslo-(l v) ;;; ,7 l
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a 1 Fort Calhoun Station Unit-No. 1 TDB lll.1.e TECitNICAL DATA BOOK PROCEDURE Titles PRESSURIZER LEVEL CORRECTION (DENSITY EFFECTS CORRECTION) Setpoint/ Procedure Fom Number (FC 68) : 39172 Reason for Change: To annotate the TDB Figure to indicate that they are related to ongoing conunitment 790101. Contact , Person: Deb Matthews v-
- ISSUED: :
09-14-92 -4:00 pa R3
FORT CALHOUN STATION TDB III 1.e TECID18 CAL DATA BOOK PROCEDURE PAGE 1 OF 1 l 1 I PRESSURIZER LEVEL CORRECTION (DENSITY EFFECTS CORRECTION) L-101 ! l 100 ; 95 " AWBlENT T,EMPl ERA,1VRE , 30,0T 2 . NOTE: temperature obtained from. l 90 4 EFF points T-887, 888, B89, 4 85-. ; Y , . 80
; COMMITMEH1 10#790101 2100 psio m s [
. 4 7 v>
{ 75 j 70 , 1000 psia- [
- o. : N /
j l 300 psia [ y' j j , 55 y %< m 5 3
- ) )7 /
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g ;
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0 -- 0 10 20 30 40 50 60 70 80 90 100 5 15 25 35 45 55 65 75 85 95 INDICATED PRESSURIZER LEVEL. (!' Top Span) (TDB9204H.WPG) R3 FC/TDB
. . - . ...... - - =. - - . - - - - - - .. . - . - - - . . . . = . ~ = - . - . . - - . . - . -.
l l l 1 Fort Calhoun Station ; 4
--Unit No. 1 i k
i TDB lli.2 ) TECHNICAL DATA BOOK PROCEDURE l t
Title:
ACTUAL LEVEL IN PRESSURIZER VS. INDICATED > LEVEL IN PRESSURIZER
~
1-Setpoint/ Procedure Form Number (FC 68): 39172 Reason for Change: To annotate the TDB Figures to j indicate that they are related to ongoing commitment'790101. i T Contact-Person: Deb Matthews i i-l 4 x
. i v '.
.09-14-92 '4:00 pm'- R3
. ISSUED:
.,.:.-, ._-..,A._., .-, . - . _ - . . . , , , . - - - - , , , . , _ , , . - - , . . , . . . - ,. . , _ _ _ . . , _ . . . . , . . , , . , . _ , , , _ . - . - . . . , . . - . ~ , . _ , . _ - . . , . .
- _- . . . - - - - ~ _ _ . - _ _ - _ - - . . - . - - - - _ . .__
FORT CAL 1!OUN STATION TDB-III.2 TECIDlICAL DATA BOOK PROCEDURE PAGE 1 OF 1 ACTUAL LEVEL IN PRESSURIZER .- 1200 riio "#@ VS. 5500 pg,o-i 1000 raio 50psio7 INDICATED LEVEL IN PRESSURIZER i soo ,,i, ~ i oo ,,,, rio ,,io , 1,\ 4 44 , t4 = ,,., 7,/,
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Il r 5% [ COW TE D7901dl 0"- 0% - f 10% 20% 30% 40% 50% 60% 70% 80% 90% 10,0% 0,% 0" 47" 94" 141" 188" LI- 106 INDICATED LEVEL (". AND INCHES) v (TDB9204J.WPG) R3 FC/TDB
, , - . , - - , , -- - ,- ,.-e-n.-,----.,
Questiort ; 82 Systern; NIode; KAJtemt 80_ Imp; SRO Imp.; 000 056 EA2.50 2.8 3.1 Systern ; Emergency Plant Evolution Mode; t.oss of Offsde Power Description; Abihty to determine that load and VAR hmns; alarm setpoints; frequency and voltage hmrts for ED/Gs are not being exceeded. Questior); The reactor has tripped 15 minutes ago. The operators completed EOP-00 and are . now in EOP 20. DG1 is carrying 2375 KW and 395 amps and is the only available AC power source. Outside temperature is 102*F. A bearing water pump has been restarted arid it is now desired to restart an air comp;ssor. Which one of the following actions should you direct to be taken? A Do not start the cornpressor because the current draw for the comptes or would overload DG 1 by exceeding its 2500 KW rating. B. Do not start the compressor because the starting current will exceed the maximum instantaneous amp rating on DG 1. C. Start the compressor, because the running load is within the acceptable range on the DG's oading curves. D. Start the compressor, because the totalload with the compressor running is below the maximum load rating of 2625 KW for the DG's. An1 Wet; C%_leyel; Qugglioitsputee; 6_Ma@ ment C Yes Bank Question not seen TDB Ill,26A LP_flumber; .OMectlygJJ - 07 13-05 01.06 Qbjedlye; Given TDB Figure Ill.26.A, be able to pmdict if dieselloading hmits will be exceeded when loads are restarted following a loss of off-site power. Referente; LP 0713-05, TDB-lil,26A Commepitj bank question LR EOP 20-RO 010
\, I Fort Calhoun Station Unit No.1 TOB Ill.26 TECHNICAL DATA BOOK PROCEDURE
Title:
DIESEL GENERATOR CAPABILITY CURVE (4160 VOLTS)
"' FC 68 Number:
47741
- 7 Reason for Change: Update / redraw the curve using calculations developed under FC06501.
Contact Person: K Boston
+wg ISSUED: 0547-96 4:00 pm R2
FORT CALHOUN STATION TDB Ill.26 TECHNICAL DATA BOOK PROCEDURE PAGE 1 OF 1 Diesel Generator Capability Curve (4160 Volts) 3 00 l l l' I 1.0 _ l> ! ! . . l , ; lii - j f i 3000 l* l i/1 i 0.9 - l +
/! ,
l : i , $ _ / l [ 0.85 l, . 4
' ' je ; [l I /f j _l, '
l, ,i i , i -
/t g{/j 0.8 g 2500 j' ;i I I IiI j.
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lli'i j;e, t ,! ii litt t!!/ l i/lfl'/{l i 0.75 'b . e + ll!Ii !!!. I !> !!I !Ill ld [l/ / I!
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Fort Calhoun Station Unit No.1 , TDB-lli.26.A TECHNICAL DATA BOOK PROCEDURE
Title:
DIESEL GENERATOR LOADING CURVE FC 68 Number: 44386
- Reason for Change: To incorporate new TDB curves.
Contact Person: Douglas Molzer
~'
ISSUED: 06 22-95 diOO pm ' RT
FoBT( a.HOUN STATION '; B-lil.26.A-TECHNICAL DATA BOOK PROCEDURE . PAGE 1 OF 2 Figura 1 DG-1 OtTITilT POWER R ATING 'i illt De;. F Manlonim Ambient Ilevit) WEATieER TOWER Tuir. #N DEG. F (Nat'l Weasher Service may be und K westie tem r amar.uir.tel - 93 99 100 101 10. IJ3 104 105 toe. not it;a 109 les
- 2. m v*
'2.627 2.aso 2.600 " - - -
' 2000 hi157:$~ , -- .-
2.uns .
/ "
is y ' ' m Ar.6ima Tciup ; 2 2.550 - - - - - - - - - - - -
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e 2.500
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- 2. San o ~
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)
- p Denmaml vs. Time - --
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l
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a=-- . . , , . _ . . , , , g ESA
- 2,ynn .j o 5 In 15 20 25 30 35 4n 45 50 3$ gi EJhylene Glycol Coolant Swe: DG-1 can be cnnsedered operabic esth anesess scanycraseres less than as er l es 110 deg 3 i
R7 .. = . . - . - .
iii. - d-ill 26~A FORTl _HOUN STATION PAGE 2 OF 2 TECHNICAL DATA BOOK PROCEDURE Figu.m - DG-2 OtrITtiT POWER RATING (114 Deg 'Ihlastr.usm Ambient IJmist WEATitER 10MER IEntr. IN pkG. t (Natl Weseber Service may be ese11 if weather temer uma,aal. Lact 93 92 >+ 96 95 ((al It22 lui i+4 toe llo 182 184 2.o w 2.650 2,627 _.__ _ _ g ,,
- 2. tot)
,$ ~--
235 r-2,59 - k '
- / 2S" M 2.500 --
2000 hr Rating 3 E vs.
' ~
~ #**
2.450 Ambiern Temp ; s 2.4 a) 2.se % l gg
- - - -- - - - - -- ------ --- ~
2.)* 2.350 O 2.11* y 2y _ _ - - .- . - . . 2.srn 5 as -
# V$' T8" - - - - - - - - - -
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- - - - - - -- - - S' -
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2.u % 2.050 su 15 20 25 30 35 4n 45 sa 55 8 0 $ Time ime. Esces (siial lihylene G'ycol Ceciaan
% g)G-2 can tic consadered egenatde emh entiens eennpersaires less aba, w egiat en 111 des I-R7
Questionj 83 System; M_o.de; KA,iteny RO Imp; SRO Impj 000 036 EA1.01 3.3 3.8 Systemj Emergency Plant Evolution Mode; Fuel haru ling Incident Descriptiom Abildy to operate and monitor the containment purge ventilation system. Questlom The plant is in a refueling outage with the containment purge system operating. Fuel inspection is in progress in the spent fuel pool. A spent fuel bundle has been dropped and damaged resulting in RM 062 going into high alarm Which one of the following statements correctly describes the response of the containment purge isolation valves to this event? A. The inboard containment isolation valves only would close. B. The outbcard containment isolation valves only would close. C. Both the inboard and the outboard containment isolation valves would close. D. Neither the inboard nor the outboard containment isolation valves would close since the event did not occur in containment. u ensweI; Q!A level: Quet, tion source:
Attachment:
C Yes Modified Bank Question none LP numbef; ,_Oblective #: 07 14-04 01.08 Obje_g.tive. EXPLAIN the principles of Emergency operation of the Containment Purge System in terms of flowpaths, major parameters, (temperature, pressure, flow, etc.) alarms and control devices. Beference; LP 07-14-04 Cu_mmentsJ bank question 07-14-04,1.8 002 significantly modified. Original question attached bank question was not seen. SRO 55.43(b)(7) 1
l N. M it. im o 1.s6 tu 5.0 ROISRO / Pgc n 6714 04.1.8001 now odes tne Containment Purge System respono to a GRH5?
. -(lih pt)
/
Cantainment isolation valves close and f ans trip [on a VIAS]. [1.0 pt) Ent. 7/17/89 /
Reference:
p 71404,1,8 G t.EY WORDS: , MayWord 1 Keyword 2 Keyword 3 Keyword 4 Keyword 5 keyword 6 KeyWoro 7 Keyword 8 I I I I I I I I I C ATES: Modified: Fri Jun 10 t994 Daxt I7 14-04.1.8 002 OV/$/ /)r/ 23 Tne piant is in Goid Shutoown for a retuonng outage with the containment purge system operating in the low purge ra;e nxade with f an VA 77 A malfunctiort 6n radiaton morvtor RM 061 [ stack gas morwtor] causes the morvtor to exceso ..ie high alarnt setpoint. Which of the followtrq statements correctly desenbes the response of the containment purge system to this event? [1.0pt) A. The inboard containment isolation valves only would close and the low purge rate fan would trip. B. The outboard containment isolation valves only would close and the low purge rate f an would trip.
-C. Both the inboard and the cutboard isolation valves would close and the low purge f amwould tnp.
D. No automatic action would occur since there must be at least 2 out of f,P radiation morntors alarming in order for a containment purge system isolation to occur. The correct answer is C [1.0 pt) Ent. 11/15/89
Reference:
K/A 029000,K4.03,RO3.2/SRO3.5 71404,1.8 New Ridenoure KEY WORDS: Keyword 1 Keyword 2 Keyword 3 Keyword 4 Keyword J KeyWoro 6 Keyword 7 KeyWrwd 8 I I I I I I I I I D ATESt Mod 0eo: Frt. Jun 10. 1964 Usect l
QugitJgnj 84 System _; Mgdej KA itemi RO ImPJ SJ_O Impj 194 001A A1.16 3.1 4.4 System; Plant Wde Generic Mode; Plant Wde Genenc Responsibilities Qtsgip!tioli Abildy to take actions called for in the Facility Emergency Plan; includirg (if required) supporting or acting as the Emergency Coordinator. Question:. A large break LOCA has occurred. Containment pressure is 55 psig and increasing slowly. RCS subcooling has been zero for the past 10 minutes. RMO-91 A & B are reading 25,000 R/hr. Containment integrity can not be venfied and the shift supervisor has declared a General Emergency. Which one of the following Protective Action Recommendations should be made? A. S: letter 2 mile radius, Shelter 5 miles downwind sector. B. Evacuate 2 m;io radius, Evacuate 5 miles downwind sector. C. Evacuate 2 mile radius, Sheher 5 miles downwind sector. D. Shelter 2 mile radius, Evacuate 5 miles downwind sector C/A levelj Question source: Attachment; Answen B Yes Bank Question - not seen EPIP EOF-7 1.P_ number: _ObiectIvo #: 08-13-06 07.00 Obiective: SRO students will be able to recognize that an E-plan event is in progress, classify the event, and complete required sith personnel and plant management notifications.
Reference:
EPIP-EOF-7 Commentsl bank question LR-EPIP-SRO 017 SRO level question 55.43(b)(4)
l Fort Calhoun Station Unit No. 1 EPIP-EOF-7 EMERGENCY PLAN IMPLEMENTING PROCEDURE
Title:
PROTECTIVE ACTION GUIDELINES IN ACCORDANCE WITH 10 CFR 50.54 (q), THIS REVISION DOES NOT REDUCE THE EFFECTIVENESS OF THE FCS RERP. REVIEWED per EPDM-6: N i 11.uxt D Supervi Emerg cy Planning (signature r edpri@ortodistribution) FC-68 Number: 45297 Reason for Change: Grammatical changes to text. Changed "DDE" term to "SDE" par EPA 400-R-92-001. Contact Person: Mike Christensen ISSUED: 09-01-94 4:00 pm R12 l
FORT CALHOUN STATION EPIPoEOFJ EMERGENCY PLAN IMPLEMENTING' PROCEDURE . PAGE 1 OF 5. m- PROTECTIVE ACTION GUIDELINES
'9.0 PURPOSE 1.1 - - This procedure outlines a method for determining Protective Action Recommendations (PARS). l 2.0 : PREREQUISITE 2.1 An emergency has been declared per EPIP OSC-1.
3.0 BEFERENCES 3.1 ' Manual of Protective Action Guides and Protective Actions for Nuclear Incidents",- EPA 400 R 92401, May 1992 3.2 - NUREG/BR 0150. " Response Technical Manual" USNRC, Vol 1, Rev. 2, October,1992. 4.0 DEFINmONS NONE r,*
.-ty,,
.~% Y_ R12
FORT CALHOUN STATION EPIP EOF-7 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 2 OF 5 5.0 PROCEDURE NOTES Any incident requiring a FC-1188, " Emergency Notification Form" to be issued to the states also: requires a PAR to be issued. The PAR may be "NONE' " Evacuate", " Shelter" or a combination-thereof. Concurrence of states is NOT required prior to issuing Protective Action Recommendations. 5.1 Determine appropnate PAR as follows: NOTES DO NOT delay issuance of PAR notification while awaiting field team results or dose assessment projections. For Notification of Unusual Event and Alert classifications, the PAR is typically "NONE'; For General Emergencies, there is a minimum PAR specified in Attachment 6.1. 5.1.1 Refer to Attachment 6.1 and follow flowchart to determine appropriate PAR. A. If no PAR is used from Attachment 6.1, and there is no release offsite, the appropriate PAR will be "NONE". 5.1.2 Refer to Attachment 6.2 when dose assessment / field survey data become available to determine PAR (s) based on this data. 5.2 Record PAR (s) On: 5.2.1 Form FC-1188 for initial notification, or when used for updates, j 5.2.2 EAGLE output, if EAGLE is used for updates. A. If EAGLE distribution to the states is not functioning property, use a blank FC-1188 form to insert the applicable information and distribute (via FAX) to the state locations. R12 1
l ( FORT CALHOUN STATION EPlP EOF-7 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 3 OF 5 m 5.3 PAR Review and Approval: 5.3.1 Control Room A. PAR (s) based on plant conditions are normally prepared by Oper ations personnel. B. PAR (s) based on radiological concerns are normally prepared by the position performing dose assessment. C. Reviewed and approved by the Command and Control position. 5.3.2 Technical di..pport Center A. PAR (s) based on plant conditions are normally prepared by TSC Operations Uaison or Site Director. B. PAR (s) based on radiological concerns are normally prepared by the TSC Protective Measures Coordinator Group. C. Normally reviewed by the Protective Measures Coordinator. = D. Approved by the Command and Control positica
~
5.3.3 Emergency Operations Facility A. PAR (s) are normally prepared by the Protective Measures Manager's group with operational input from the EOF Operations Uaison. B. Normally reviewed by the Protective Measures Manager. C. Approved by ine Command and Control position. NOTE u; ~ . Stateiand/or County _ officials are responsible for making the final decision.on PAR (s).to bec . 2rsi$ss'ed;to therpublier ' ~ - 5.4 Transmit PARS to local and/or state agencies per EPIP-OSC-2 or EPIP-EOF 6.
. 6.0 ATTACHMENTS 6.1 Protective Action Recommendations Process Flowchart 6.2 Protective Action Recommendations Based on Radiological Data R12
i FORT CALHOUN STATION EPIP EOF-7 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 4 OF 5 ATTACHMENT 6.1 Page 1 of 1 PROTECTIVE ACTION RECOMMENDATIONS PROCESS FLOWCHART PROTECTIVE ACTION RECOMMENDATIONS PROCESS FLOWCHART
* " m ,o
-- o- aiYuas b y
ns n
' ,J4"" "2SlL no estamN cata y i.t'n-,' ,".!.?
u ers ns g eaa.s N0 fv.cuan DCEID DCvtLor P. mas a non . 2 tale.t.a.e.dut . ,,;gagn = =,,, g e ansa. .
- a. e.n no .c m/a fossT vts
%T u "1'
" toum"cs or starSt:
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...,e... *Taaf m ( taa ser en cvacuan : ins amann -*-
n_"*"c' an LEC DTD t u .
~. ,
CCT - co2E Enr TMutuocoL7tz CC - CDetitAL EMutcDecY LOCA - Loss or coctANT Acc:Dct? PAR
- Pft0TECT!YE &cTIOld IttconOdDtDAT1001 PAG = Pft0TECT:YE ACDON CUIDE R12
FORT CALHOUN STATION EPIP EOF-7 EMERGENCY PLAN IMPLEMENTING PROCEDURE PAGE 5 OF 5 ATTACHMENT 6.2 PROTECTIVE ACTION RECOMMENDATIONS BASED ON RADIOLOGICAL DATA' Early Phase (PLUME Phase): PROTECTIVE ACTION PROTECTIVE ACTION GUIDES RECOMMENDATIONS Types of (Projected or (PARS) Exposure Actual Dose) TEDE < 1 REM TEDE *
- NO PAR REQUIRED *
- or Continue to monitor
< 5 REM CDE environmental radiation CDE (Thyroid) levels.
(Thyroid) However, at > 10% of limit: (> 100 mrem TEDE or >500 mrem CDE (Thyroid)) 2 Consider shelterino and/or eartv evacuation of children and creonant women. m TEOE a 1 REM TEDE *
- EVACUATE *
- or ' Shelter, if it will provide protection 2 5 REM CDE equal to or greater than evacuation,
. CDE (Thyroid) up (Thyroid) to 10 REM. CDE a 25 REM CDE *
- EVACUATE * *
(Thyroid) (Thyroid) Recommand administration of stable iodine (KI) to OPPD emergency workers. SDE a 50 REM SDE *
- EVACUATE * *
(Skin) (Skin) l
' Based upon guidance in chapter 2, EPA-400-R-92-001, " Manual of Protective Action Guides And Protective Actions For Nuclear incidents", May,1992.
2 Sheltering may be preferable to evacuation as a protective action in some situat.ons. Because of the higher risk associated with evacuation of some special groups in the population (e.g. those who are not readily mobile), sheltering may be the preferred attemative for such groups as a protective action at projected doses up to 5 rem. In addition, under unusually hazardous environmental conditions, use of sheltering at projected doses up to 5
^) rem to the general population (and up to 10 rem to special grcups) may become justified.
V R12
l ( Questio!U 85 Sy_s_ tem; Mode; KA item: RO Imp; SRO Imp; 194 001A A1.03 2.5 3.4 SysterDJ Plant Wde Generic Mode; Plant Wde Generic Responsibihties Descrip_ tion; Abihty to locate and use proceou.es and station directives related to shift staffing and activities. Question; With the plant in hot shutdown, the Shift Technical Advisor slipped on a stairway and injured his ankle. An off-duty Shift Supervisor leaving the site has taken him to the emergency room. Which one of the following actions should be taken due te this event? A. Make a four hour report to the NRC. B. Have a System Engineer who is training to be an STA fill the position. C. Have a qualified STA on-site within two hours. D. Declare a Notification of Unusual Event based on transport of an injured person. Answen CIA level: Question source: Attachmentl C Yes Bank Question - not seen none 1.P numben Objective #: 07 62-08 02.06 Objective: Administrative Controls
Reference:
LP 07-62-08 Comm_e_nts_; bank question LR-ADMIN-SRO-010 minor wording change
Questi_oru 86 Eystem: Mode; KA item: RO Imp; SRO Imp; _ 194 - 001A K1.14 3.3 3.6 n Syste_m : Plant Wde Generic - Mode: Plant Wde Generic Responsibilities Descriptio3; Knowledge of safety procedures related to confined spaces. Questlof!.; While touring the turbine building during an outage, you observe an unconscious person 3 feet inside the upper manway of the main condenser. Which one of the following action should you take? A. Puti the pemon out and give mouth to mouth recesitation. B. Locate an SCBA, Don it and pull the person out. C. Notify secunty immediately. I D. Notify the control room immediately.
.m
~
Answer: C/A level: Question source:
Attachment:
D No New Question none LP number: Objective #: 10-27 12 - 01.08 Objective; identify FCS policies, precautions and equipment used for entering asbestos restricted areas, confined space areas ami potential heat stress areas.
Reference:
LP 10-27-12, FCS Safety Manual section VI-4 Comments: Y
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~
QueJtionj 87 Sysjerm Mode; KAijem: RO Imp; SEO. Imp; 194 001A K1.07 3.6 3.7 System : Plant Wde Generic Modej Plant Wde Generic ResponsibAties pescriplips Knowledge of safety procedures related to e5ctrical equipment. Questiom Which one of the following is a requirement for OPPD Electncal Operations Departn:ent (EOD) to enter into the FCS switchyard per NOD-QP-36,' Control of Switchyard Activities at Fort Calhoun Station.' A. EOD pctsonnel must not take company vehicles into the FCS switc1 yard. B. EOD personnel must be accompanied by FCS security wMn entering the FCS switchyard. C. EOD personnel must have scheduled the swithyard entry at least 7 days in advance. D. EOD personnel must notify the FCS controt room of their activities. CIA leyel; gyestion souse; ettachment; Answen D No Modified Bank Question ncne L.P numbgn _ Objective #: 07-13 01.06e O_bjgctive: Given a copy of the procedure NOD-QP 36, Control of SWYD Activities at FCS, the student will be able to perform the following: Describe how access to the switchyard is obtained for scheduled and non-scheduled activities. Bgiqaftc.ej LP 07'13-01, NOD-QP-36 Comments; bank question 1.6E 001 significantly modified. SRO level original questions attached, but not seen
~
l I *,' l. m.M :. em o 12.4: rw - 5.0 RO/SRO hem n
^
4-713 01.16L 001 Whch one of the following is NOT an acuon requireo for pwoonnel to erWor into the FCS swnchyrd per NOD.OP 36," Control of Switchyard Activities at Fort Calhoun Station?* [1.0 pt) -
'A. EOO pwoonnoi entwing the swtchyard rnust notify the Une Dispatcher to report who they are and the activities to be performed.
- 4. EOO pwoonnel must be secompanied by FCS Securrty when entering the FCS switchyard.
C. . EOO personnel must notif y the FCS Control Room to report who they are and the acavities to be performed.
?
D. The FCS Shit Supervisor will call the Production Operations Division System Operator when personnel are inside the switchyard. The correct enewer is B [1.0 pt) Ent. 7/21/94 KEY WORDS: Keyword 1 KeyWoed 2 Keyword 3 Keyword 4 Keyword 5 KeyWoed 8 Keyword 7 Keyword 8 I I I I I I I I I DATE3: Modsthed:Thu, Sep 1,1994 Unast 1713-01,1.6E 002 Which one of the followng is NOT an accon required for personnel to exit the FCS swichystd per NOD OP 36,' Control of Swichyard ActMties at Fort Calhoun Station?"
- [1.0 pt)
A. EOO personnel shall notdy the Une Dispatcher when they exit the FCS swichyard. B. Upon completion of Work for the day on the swtichyard, the FC3 Shift Supervisor shall have SeoAsty venty switchyard gates are locked.
- 4. Upon completion of work for the day on the switchyard, EOO pwoonnel shall be socorted from the outchyard by Security.
D. Upon completion of swichyard work activities, EOD pwsonnel shan notify the FCS Shift Supervisor, w 1he correct answer le C [1.0 pt) Ent. 7/21/94 KEY WOftOS:
- Keyword i KeyWood 2 Keyword 3 KeyWood 4 Keyword 5 keyword 8 Keyword 7 Keyword 8 I I I I I I I I I DATEst ModHied:Thu, Jul 21,19e4 Used:
1 I 1
- e
. - n s.:is,imo nei rM 5.0 RO/SRO ress: 2s ;
F713 01.1.6F 001 WNch one of the fosoeng is NOT true concorrung use of vehicses in the ownchyard per NOD OP.36 Controlof Switchyard Achvities at Fort Calhoun Sta%on?" [1.0 pt)
-e
. 4. - AE vehicles rnust be parked within 10 feet of any 161KV or 345KV structure.
B. A5 vehicles must be operated at a speed of 10 rnph or less.
- - C. A vehicle may be left unattended if it is turned off.
' O. Safety rules for operahng vehicles apply whenever vehicles are in the omschyard.
The correct answer le A [1.0 pt) , Ent. 7/21/94 f KEY WORDS . KeyWeed 1 Keyword 2 KeyWoed 3 Keyword 4 Keyword 6 Keyword 8 KeyWoed 7 Keyword 8 I I I I I I I I I OATE5:MedWied:Thu. Jul 2t 1994 Uume l 713 01,1.6F 002 WNch one of the foRomng is NOT true concorrung pertung of vehdee in 4% the switchyard per NOD.QP.36, *Corwol of Switchyard Acevhies at Fort Calhoun Station?' [1.0 pt] A. Aceve vehicles are to be parked in the areas adjacent to eithet satchyard control buildng or where necessary to perform acthrition.
' B. Nowactive vehicles are to be parked in the southeast corner of the FCS switchyard.
*C, Keys to reactive vehicles are to be left inside the vehicle hanging on the escrage location of the driver's side sunvisor.
D.- ' Parked vehicles may be left running if placed it
- park' and eithw the wheels cocked or the outriggers extended and set.
The correct answer is C [1.0 pt)
- Ent. 7/21/94 Kav Womes:
KeyWeni 1 KeyWesd 2 KeyWood 3 KeyWosd 4 KeyWeed 6 KeyWoed 8 KeyWoed 7 KeyWoed 8
-1 I i 1- l l -I 7 OATEs:Mossned:Thu, Jul 21, 1994 Umst -
Ag~
- - _- - . _ _ _ _ _ _ _ _ _ _ E
i nu.hito.iwee :4:rw 5.0 RO/SRO r 26
~
l./-13 01.1.f,D 00t Wnecn one of the loiiowing is NOT true concorrung periocic inspections of the setchyard per NOD.OP.36,' Control of Switchyard Activities et Fort Calhoun Stabon?* (1.0 pt] A. Electncal Operatons Division shall provide a list of required Mspectons to the FCS Plant Manager. B. The inspeccon listing shall be maintained by the FCS Shift Supervisor in the Control Room.
*C. The inspection listng shall be posted on the bulletin board at the entrance to the Control Room.
D. The inspection listing shall be reviewed by System Engineering to ensure trurumal riska exists. The correct enewer is C [1.0 pt] Ent. 7/21/94 KEY WORDS: Keyword 1 Keyword 2 Keyword 3 Keyword 4 Keyword 5 XeyWord 6 Keyword 7 Keyword 8 I I I I I I I I I D ATE S: Moefied:Tue. Oct 11,1994 unut 17 13 01.1.00 002 Which one of the foliomng is NOT true concorrung emergent worn on tne switchyard por NOD.OP 36,' Control of Sutchyud Activities at Fort Calhoun Station?* (1.0 pt] A. The EOD field Supervisor or Une Dispatcher shall notty the FCS Shift Supervisor of the emergent Work prior to dispatching a crew to the switchyard. B. The neufication of the emergent work should occur at least one (1) day in advance to permit addition of the worx to the Plan of the Day, C. The FCS Shift Supervisor shall consult with System Engineering for aseistance in resolving questione and evaluating risks of the emergent work.
- 4. A liet of emergent work items shall be maintained by the Division Manager of Electrio Operations Divianon.
The correct answer is D [1.0 pt] Ent. 7/21/94 KEY WOHDS: Keyword 1 Keyword 2 Keyword 3 Keyword 4 Keyword 5 keyword 6 Keyword 7 Keyword 8 I I I I I I I I I D ATESt Modaed:Thu. Jul 21,1994 unsct li - -- _ _ - _ _ - - _ - _ - _ - - - - - - _ _ _ _ _ _ _ _ - -
999stiptu 88 Sy. stem; Mgde, KAjtemi RO_Ir0p; sBo imp; 194 001A K1.02 3.7 4.1 System _; Plant Wde Ger erio . Medej Plant Wde Genenc Responsibilities Qgenipilpn; Knowledge of tagging and clearance procedures. Ruestipal; AC-3A has t>een tagged out for repairs. A Temporary / Testing Release was done to allow testing. Following the testing, it was determined that additional repairs were required. Identify the proper sequence for removing the Temporary / Testing Release. A. Remove the Temporary / Testing Release tags, independently venfy equipment positions, and rehang the Danger tags, D. Remove the Temporary / Testing Release tags and rehang the Danger tags. C. Remove the Temporary / Testing Release tags, reposition equipment as necessary, Rehang Danger tags, and independently verify. D. Remove the Temporary / Testing Release tags and independently venfy equipment positions. Answer: CIA level: Question sourcej 6ttachrnegt; C No Bank Question - not seen none LP number: _0_Djective #; 07-62-01 01.00 QDlective: STATE the major sections of the Standing Orders. Reference; LP 07 02-01, SO G 20A 9.9mm*Dipl bank question LR-ADMIN-RO-020 updated
QuestioRJ 89 System; Model KA iterm RO Iring; SLO Imp; 194 001A K1.16 3.5 4.2 Sygtem.; Plant Wde Generic Modej Plant Wde Generic Responsibilities Res_cijpttgan; Knowledge of facility protection requirements; including fire brigade and portable fire-fighting equipment usage. 991stign; A fire has been discovered in the drum storage area of the radwaste building. All personnel have been evacuated. The Blair fire department has been requested to respond to help fight the fire, What personnel should make up the fire brigade responding to the fire? A. LO", EONA. EONT, Securny(2), and RP Tech. B. LSO, LO*, EONA, Secunty(2), and Shift Chemist. C, LO*, EONA, AON, and Secunty(2). D. LO*, EONA EONT, and Secunty(3). Ansm C/A levej; Question source:
Attachment:
A No Bank Question - not seen none LP number: .O_hjective #: 07-62-01 01.00 Obloctive: STATE the maior sections of the Standing Orders.
Reference:
SO G-28 Comments; bank question LR-ADMIN-RO 006 l I l I
r Queytl9nj 90 Systemi M9odel KAIJern; Ro imp; Sgojmp;
,. 194 001A K1.08 3.5 3.4 Systent; Plant Wde Generic M_odei Plant Wde Generic Responsibilities R9s.Gilpil9B; Knowledge of safety procedures related to high temperature, Wesil9n; D
Which one of the following sets of heat stress symptoms is indicative of heat etroke? s A. Dry skin and small pupils 0, Heavy sweating and thirst C. Headache and coolskin D. Nausea and dilated pupils s Ansyrgy; CIA leyel; Question soulcej ejta91LMt011 A No New Question none , LP nugt,gn b ,Q)Je_qtive #: 10-27-12 01.09 Obiective: Identify the first aid that sheund be administered when a worker suffers from heat stress. Bgigence; LP 10 27-12, FCS safety manual section Il-4 Comments.j Heat stroke requires immediate medical attention. B & C are incorrect because heat stroke syympoms include hot dry skin, not sweating or cool skin. D is _ incorrect because sma!! pupils, not dilated pupils, is a symptom of heat stroke.
~.
l QuestLon; 91 syfte._; m Modej $A itemj SQ_1m._pj S_RO ImE
. - , 194 001A A1.10 2.9 3.9 Sy!stemj Plant Wde Generic Modej Plant Wde Generic Responsibildics pgsciplioD; Ability to coordinate personnel activities outside the control room.
Questjo_rJ1 Which one of the fotowing is a responsibility of the OCC LSOs accoroing to OPD 167 A. Schedule and review all surveillence tests. B. Attend all ALARA briefings for scheduled work. C. Perform the weekly tagout audit. D. Approve alltemporary procedure changes. p. Answen C/A level: Question source: Atta.chment: C No New Question none LP numben .Qb_lective #: 07 67-05 01.00b Obloctive: Operation Department Organization Referencel OPD-4-16 Comments: SRO question 55.43(b)(3) t
Questi_9nj 92- Systety; Modej KA item: RO Imp; SED Imp: 194 001A A1.13 4.3 4.1 Sygemj Plant Wde Generic Motiej Plant Wde Generic Responsibilities QesLcijpfon; Ability to locate control room switches; controls; and indications; and to determine that they are correctly reflecting the desired plant lineup. Qu91tig_gi The plant is in mode 4. Which one of the following will be designated by the use of placards to ensure RCS makeup capability as a prerequisite for initiating shutdown cooling? A. One LPSI pump with 6 backup power supply. B. _One Boric Acid pump with a backup power supply. C. One Charing pump with a backup power supply. D. One Containment Sprav pump with a backup power supply.. m C/A leveh guestion source:
Attachment:
Answer: D No New Question none LP number: Obiective #: 07-07 42 10.00 QDjeglyve.; e LIST all the shutdown risk related success paths for. Referensej OI-SC 1 Commentgj
?
1 l Question,j 93 Systejg; Mo_dej 5A_ item; BO Impj SRO imA 194 001A K1.03 2.8 3.4 System ; Plant Wde Genenc Mpdej Plant Wde Generic Responsibiltties DngriptioE Knowledge of 10 CFR 20 and re ated facility radiation control requirements. GLugjiori; During a refueling outage, an individual working on the S/G tube inspection has exceeded his stay time inside the S/G. After reading his TLD, it was fourd that he received a dose equivalent of 26 rem. Which one of the following actions is required? A. Make a one hour notification to the NRC. B. Make a 24 hour notification to the NRC. C. Declare a Notification of Unusual Event. D. Declare an Alert. . 1 7 Answer: CIA leveji Question source:
Attachment:
A Yes Bank Question - not seen none LP numble Obiective #: 07 62-08 02.06 Obiective_; Administrative Controls Refef9nce; LP 07-62-06, SO-R 11, TS section S 9_omments; bank question LR ADMIN-SRO 021 distractor changed SRO SS.43(b)(2)
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i Questiort; 94 Systerm Mode; KA_Jttml RO Imp; SRO impj 194 001A K1.15 3.4 3.8 Sylte_mj Plant Wde Genenc Modej Plant Wide Genenc Responsibilities pescript.lom Knowledge of safety procedures related to hydrogen. Quettlon; Which one of the following operations is used to fill the main generator with hydrogen following a refueling outage? A. The generator is filled wrlh cerbon dioxide through the top header as air is vented through the bottom header. Then hydrogen is introduced through the top header as carbon dioxide is vented fron the bottom header. B. The generator is filled with carbon dioxide through the top header as air is vented through the bottom header. Then hydrogen is introduced through the bottom header as carbon dioxide is vented fron the top header. C. The generator is filled with carbon dioxide through tred bottom header as air is vented through the top header. Then hydrogen is intr. uced through the top header as carbon dioxide is vented fron the^ bottom header. D. The generator is filled with carbon dioxide through the bottom header as air is vented through the top header. Then hydrogen is introduced through the bottom header as carbon dioxide is vented fron the top header. Ansvyen CIA level: Question sourcej 6ttachment: C No New Quesi.*on none LP numben Objective #: 07-13-06 03.00 Rbjective: EXPl.AtN the principles of operation of the Main Generator Hydrogen and Carbon Dioxide Cooling System. Beforence: LP 07-13-06 Comments: . Tests candidate's knowledg1 of procedure for safety filling the main generator with hydrogen. l
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l l. Questjon_; 95 S_ystepli Mp(tej M itemi Bo_Lmp; SjLO Imp; tX 001A A138 2.6 3.1
, gy.ste.mj Plant Wide Generic M_ofa, Plant Wide 09neric Responsibihties l pescripj{gn; Abihty to obtain and interpret station reference matenal sucn as graphs; monographs; and tables which contain system performance data.
REHiO!11 A fire has occurred in the cable spreading room The LSO has established control at the attemate shutdown p1nel. RCS temperatura is 536*F and RCS pressure is 2200 psla. What is the subcooling margin? A. 84*F. B 94*F C. 104*F D. 114*F. o,
~s i
Answer: Q/A lovsl: Question source: 6ttachment: D Yes Bank Question - not seen steam tables LP_twmtper; Obiective #; 07-17 00 01.02b QldestLvg; Describe how the plant may respond to a fire in the following locations: Cable Spread Room.
Reference:
LP 0717-06, AOP-06, steam tables Comments; . bank question LRmOP-06-RO 001, distractors changed
I Questigrt; 96 Systemj Mgdel KAjtem; BOJmp; SB_0_Irppj 194 001A A1.09 2.7 3.9 Sy.s_te.rnj Plant Wde Genenc Modej Plant Wde Generic l Responsibtisties Descfjpf[QB; Ability to coordinate personnel activities inside the control room. 1 Questio01 The plant has been shutdown for a refueling outage. The RCS is in a water solid condrtion. You are the LSO in charDe of the control room. Why must you asign an operator full time to monitor RCS pressure when the RCS is solid? A.- To ensure RCP NPSH requirements are met. B. To ensure a flowpath of bonc acid to the core. C, To ensure pressure-temperature hmits are not exceeded. D. To ensure PIC-210 does not isolate letdown due to a pressure spike. Ques _tiorLsource; 8.tta9hmentj A_rlswet;
._ CIA level:
C No Modified Bank Question none LP_numbeti .9ble.ctive #: 07-11 20 02.08
. Qblegtly_e; STATE the overpressure protection that shall be operable, for mode 1 operations.
fteforence: Ol-RC 3 precautions 1 Cor.pmoqis; bank question LR RC-RO 016 stem and one distractor modified SRO 55.43(b)(5) originalquestion not seen onginal question follows .(
.- - - . . - . ~. . . - _ .. . . _ ~ .. -
LR-RC RO O16 The paara has oeen snutoown for a refueung outage. The RG5 is fil6ed to ,
~
50% in the Presounzer. You the 96 are directed to,ectemtthe Reactor i N r.oolaru Systern startup. 13g Jvpp p/g l Why, when the RCS is solid, rnust an operator be assigned full time to rnorator RCS prosaure? [1,0 pts) A. To ensure shutdown cooling is ret lost B. To erwure a flowpath of boric mod to the core ) l
- 4. To onours pressure. temperature lirnits are not !
exceeded O. To ensure PIC 210 does not isolate letdown due to a pressure spike , l l 1 The correct answer le C [1.0 pts) ORIG: TECH: STYLE: 9 Rev. 3/2/95; 4 min Verified ,4
Reference:
71120,2.8; 4.1; 4.ta; . OI RC-3 Precautions K/A 002000,K4.10,RO4.2/SRO4.4 '- Rennerfeldt.7/10/89; -ttG GREBEEO,90 " ^SS SRo )cvel fuest!M
- Questionj 97 Sy_s.t.em_j Mode; KAltem; ROJm_f; SJ_O Impf 194 001A A1.02 4.1 3.9 Systemj Plant Wide Generic Mode; Plant Wide Generic l Responsibilities ;
pescrigtigm Ability to execute procedu al steps. Que_st.lgn; You are supervising a turbine sta 1up in accordance with OI-ST-2. Which one of the following statements is correct conceming required sequence of I prerequisites and procedural steps? A. Prerequisites and procedural steps must be satisfied or completed in sequence unless the procedure states otherwise. B. Prerequisites must be satisfied in sequence but procedure steps are not required to be completed in sequence unless specified in the procedure. C. Prerequisites are not required to be se'isfied in sequence but procedure steps must be completed in sequence unless the procedure states otherwise. D. Prerequidtes and procedure steps are not required to be satisfied or completed in sequence unless the precedure states otherwise. CIA level; Question source;
Attachment:
6Jswer: C No Bank Question - not seen none LP number: .O_biective #: 07-67 06 02.03 Obiectivej DEMONSTRATE the FCS policies for proper procedure usage and compliance. RgteroDge; LP 07-67 06, SO O-16 C_ommentsj bank question LR-ADMIN-RO-009 order change
Qugstiga_; 98 Syste!n; s Modej KA itenu RO Im.pj SRO Imp: 194 001A K1,05 3.1 3.4 Rysternj Plant Wde Generic M_gdej Plant Wide Generic , Responsibilities _DesidRti_9B; Knowledge of facility requirements for controlling access to vital / control areas. Quest!p_nj Channel 7 news has arrived for a tour of the plant. The group consists of 5 newspeuple and 2 cameramen. You have been assigned as their escort. Which one of the following areas of the plant will require additional escorts? A. Warehouse. B. Switchyard, C. Merlanine level of the Turbine Building. D. Room 81,
) -s Answer; CIA leveji Question source:
Attachment:
D Yes New Question none LP numbel; Obiective #: 10-27-06 01,06 Obiectivej Identify the visitor to escort ratio in different areas of FCS, and the escori's responsibilities. Beforenct; LP 10-27-08 9_9!Reents_; W l l
Questlof1J 99 Syste_m_; hLodej KA ittm; RO Impj SRO Impj 194 001A K1.09 3.4 3.4 Sysitnij Plant Wde Generic Mpdej Plant Wide Generic Responsibilities Descripil_ ort; Knowledge of safety procedures related to high pressure. Question: Which one of the following practices conceming compressed gas cylinders violate Fort Calhoun safety procedures? A. Storing a compressed gas cylinder within 3 feet of another compressed gas cylinder. B, Storing a compressed gas cylinder in a horizontal position. j C. Storing a compressed gas cylinder outside. D. Storing a compressed gas cylinder in an area with a temperature greater than 90 F. n w# Answer: C/A level: Question sourcej
Attachment:
B No New Question nono LP number: objective #: 10 27 12 01.00 Oblecitye; Demonstrate basic knowledge of industrial safety requirements at Fort Calhoun Station. EgleIgI14.p; LP 10-2712, FCS safety manual section XI-3 Commentgj v l
Qugstjopj 100 ' System; Modet KA itemi ROJmp; sRqtmp; 104 001A K1.11 3.4 3.5 Sylterrlj- Plant Wde Generic Mqdej Plant Wde Generic Responsibiltties Descriptiori; Knowledge of safety procedures triated to chlorine. AyestloJ1; Following detection of a high chlorine concentration by the detectors in the control room ventilation system, which of the following statements is true? A. The contr01 room ventilation cystem will go into the ' Recirculation" mode of operation. Air conditioning units will continue to operate. . B. The control room ventilation system will go into the ' Recirculation
- mode of operation. The air conditioning units will trip and can not be restarted unless the chlorine signalis overndden.
C. The control room ventilation system will go into the ' Filtered-Air" mode of operation. Air conditioning units will continue to operate.
' O. The control room ventilation system will go into the ' Filtered Air" niode of operation.
The air conditioning units will trip and can not be restarted unless the chlorine signal
,, is overridden.
Answer: CIA level: 9_uestion source:
Attachment:
B No New Question none LP number: OblecJt ve #: 07-14-06 01.08 Obiective: Explain the principles of emergency operation of the Control Room Ventilation System in tenns of flow path, major parameters (temp., press., tiow, etc.), alanns and control devices. Reference.j Control room ventitation LP 07-14-06 Com_nier1111 A is wrong because the AC units trip. C&D are wrong because the system goes into Recirculation mode not Filtered-Air.
r ES-301 Administrative Topics Outline Form ES 301 1 Examination Lcyci:- SRO Facility:EolLCalh9 Lag Week of Examination: 04/14/97 Examiner's Name (print): _ Administrative Describe method of cyaluation: Topic / Subject 1, ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Temporary Temporary Procedure Change, Change of Intent Determination JPM modifications ofprocedures Plant JPM Review shutdown margin calculation Parameter Verification i A.2 Surveillance JPM - Review of OP ST SillFT-001 Testing A.3 Radiation RCA Entry and Exit JPM control A.4 Emergency Event Classification JPM Plan Examiner Chief Examiner. OC U N 9\\ 4090
ES 301 Individg;tt Walk-thrgu2h Test Outling Form ES 3012 Examination Level: SRO(U) Facility: EqllCdhquD Week of E amination: 04/14/97 Examincr's Name (print): _ ._ System / JPM Safety Planned Follow-up Questions: Function K/A/G // Importance // Description
- 1. RCP Reactor Coolant IV Can RCP be started if lift oil pump does not develop pressure?
Pump Start 003/000/Al.02//2.9/2.9// Abilly to monitor changes in parameters (plant shutdown) associated with operating controls including RCP pump and motor bearing temperatures. Consequenses of increasing powa with 3 RCPs operating? 003M0/K3.01//3.7/4.0// Knowledge that a loss of RCPs will have on the RCS.
- 2. CSS / 0335 Containment V1 Wluit efTect would an inadvertant CSAS have on containment Spray Pump Operability spray valves while pump was sunning during test?
Test 026/000/A3.01//4.3/4.5// Ability to monitor pump starts and (engineen:d sarcty feature) correct valve positioning. What effect would an tru*crtant CSAS have on containment spray valves if pump tripped during test? 026/000/A2.03//3.9/4.2// Ability to predict impact of failure of ESF on CSS operation. Simulator JPMs above this line. The following JPM may be performed in the Simulator or in the Plant. 3 AFW/new Control Room Vill Also(11 What actions are necessary to control pressurizer level from ASD Evacuation 111, V) pancl? (new,AOP, plant shutdown) 004/000/K4.04//3.2/3.1// Knowledge of CVCS design features which provide for manual / automatic transfers of control. Ilow do you transf er DC control power to Al 179 to the attemate power supply? 061/000/A2.03/D.1/3.4// Ability to mitigate a loss of DC control power to the AFW system. Plant JPMs below this line. 4, DCJ0306A - Alternating Vil Purpose of Kirk key interlock on #3 battery charger ? and Securing battery 063/000/ GEN 1//3.lD.2// Knowledge of opemtor responsibilities chargers during all modes of plant operation-llow would consequences differ if voltage regulator on an in service battery charger failed high or low? 063/000/K1.03//2.9/3.5// Knowledge of cause/cfTect relationships between DC clectrical system and battery charger and battery.
- 5. WGDS Tnmsfer waste XI Ilow can a waste gas decay tank release be tenninated from the gas froni vent control room?
header to truervice decay 071/000/A4.25/D.2D.2// Ability to manually operate autonctic tank functions of process radiation monitors (new JPM, RCA entry) Flowpath causing VI AS if vent header is drained while collecting a VCT gas sample? 071/000/A4.29//3 0/3 6// Ability to manually operate sampling gas concentrations in WDGS decay tank. l Exa niner Chief Examiner: u
1 ES 3J1 baa=4 Events Form ES4013 Simulation Facility: Fort Calhoun Scenario No.: SIM 971 - Examiners: Applicants:
~-
l Scena:lo objectivei To determine the ability of the candulates to mitigate a steam generator tube rupture event-without the use of steam dump and bypass valves. Initial Conditions : 80% power Turncver: }W-54, AC 3 A tagged out (Just tripped prior to shift change, TS not entered yet) Event Malf. Event Eneet Desedption
, No. - No.
- Type
- I MAL AFW3C I Inadvertent AFAS 2 XMT RCS96 I Controlling PZR pressure channel fails high 3 MAL SGN! A C 40 gpm S/G Tube leak (A S/G) 1.3%
4 VLV MIS 5 C- RCV-978 does not isolate (prese0 5 R/N Emergency Shutdown i - 6 BST CND9 I- Condenser vacuum switch 952A indicates low vacuum (high i pressure) 7 MAL SGNI A M S/G Tube Rupture (10%) 4 Shaded entries are to be initiated by a cue from examiner.
* (N)ormal, (R)eactkity, . (I)nstrument, (C)omponent, (M)ajor Examiner; _ _ _ _ ,
'ChiefExaminer: AdfG U lM
4-ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-1 Event No.: 1 Pace j ojZ Event
Description:
Inadvertant AFAS actuation Time Position Applicant's Actions or Behavior RO/ BOP ldentify AFAS actuation from alarms SRO Enter AOP-28 SRO Direct BOP to isolate AFW to the Steam Generators BOP Close HCVs-1107A,8 1108A,B SRO Direct BOP to contact EONT to bypass affected AFAS channel SRO Direct BOP to secure AFW pumps BOP Close YCV-1045,1045A 1045B(override /close) and place FW-6 switch in pull-to-lock SRO Refer to tech specs 2.5 and 2.15 _ SRO Determine plant is in a 6 hour 1 JO with no operable AFW pumps.
l. ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-1 Event No.: 2 Paos 2 pf Z Event
Description:
Controlling pressurizer pressure channel falls high Time Position Applicant's Actions or Behavior RO Identify and report deviation between pressurizer pressure channels RO Identify and report high indication on controlling channel and lowerino pressure on other channel SRO Direct RO to transfer control to other channel or take manual control of pressurizer pressure RO Transfer control channels or take manual contrcl as directed RO Monitor and maintain proper pressurizer pressure BOP Monitor and control secondary parameters
ES-301 Operator Actions Form ES;301-4 Scenario No.: FCS-97-1 Event No.: 3 Paae 3 of 7 Event
Description:
Steam generator tube leak Time Position Applicant's Actions or Behavior RO identify and report charging / letdown mismatch RO Identify and report condenser off-gas radiation alarm (RM-057) SRO Enter AOP-22 SRO Direct RO to control pressurizer level RO Control pressurizer level BOP Determine and report that RCV-978 (supply to aux steam) did not close. (Event 4) SRO initiate Emergency Shutdown (AOP-5) (event 5) SRO identify affected steam generator. (A) SRO Direct RO or BOP to place steamline radiation monitor in service RO or BOP Place steamline radiation monitor in service SRC Direct RO or BOP to have EONT swap blowdown sample flow to waste RO or BOP Direct EONT to swap blowdown sample flow to waste SRO May direct BOP to place YCV-1045A in override and close BOP If directed, place YCV-1045A in override end close
e ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-1 Event No.: 4 Pace 4 gf Z Event
Description:
RCV 978 (Supply to Aux Steam) does not close Time Position Applicant's Actions or Behavior BOP Determine and report that RCV-978 did not close SRO Direct BOP to close RCV 978 BOP Take action to close RCV-978 _ BOP Determine and report that RCV-978 did not close SRO Direct EONT to close RCV-978
-. ~ -
ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS 97-1 Event No.: 5 Pace 5gfI Event
Description:
Emergency Shutdown s Time Position Applicant's Actions or Behavior SRO Entcr AOP-05 (Emergency Shutdown) - Direct Emergency Shutdown SRO Notify System Operations of Power Decrease SRO Direct RO to begin boration using SIRWT RO Switch charging pump suction from the VCT to the SIRWT SRO Direct BOP to control RCS cold leg temperature by reducing turbine load , BOP Reduce turbine load to control cold leg temperature SRO Direct RO to operate control rods to control ASI . RO Operate Control Rods to control ASI RO Monitor and control primary parameters BOP Monitor and control RCS cold leg temperature and secondary parameters SRO Continue to coordinate RO and BOP actions during power reduction ame-- eh W - t
1 ES-301 Operator Actions Form ES-3014 Scenario No.: FCS-97-1 Event No.; 6 Paae Q o, fI Event
Description:
Condenser vacuum switch 952A fails to prevent operation of steam ' dump and bypass valves Time Position Applicant's Actions or Behavlur Determine that RCS temperatures are high following BOP reactor trip BOP Identify and report failure of steam dump and bypass valves to open following trip SRO Direct BOP to control RCS temperature using HCV-1040, or MS-291 and 292 BOP Control RCS tem arature using HCV-1040 and/or MS-291 and 292 as cirected by the SRO BOP Do not open MS-291 following isolation of steem generator A
ES-301 Operator Actions Form ES-301-4 l Scenario No.: FCS-97-1 Event No.: 7 Pace Z pfI , l Event
Description:
Tube Rupture - Steam Generator A ^ Time Position Applicant's Actions or Behavior RO identify and report RCS inventory loss SRO May direct reactor trip SRO Following manual or auto reactor trip, direct standard post trip actions RO Perform primary standard post trip actions BOP Perform secondary standard post trip actions _ SPO Diagnose tube rupture - enter EOP-04 or EOP-20 SRO Direct RCS cooldown - Tw less than 510*F BOP Cooldown RCS Tw to less than 510 F
- RO identify and verify PPLS SRO/ BOP Identify most.affected steam generator (A)
SRO Direct BOP to isolate steam generator A BOP isolate steam generator A SRO Direct R,0 to depressurize RCS to less than 1000 psia RO Depressurize the RCS RO Maintain subcooling BOP Monitor and control secondary parameters RO Monitor and control primary parameters t Scenario ends when steam generator A is isolated and RCS pressure is less than 1000 psia.
'4 ES 301 Scenario Events Form ES *l013 r
Simulation Facility: Fort Call.oun Scenario No.: SIM 97 2 Examiners: _ _ Applicants: Scenario objective: To evaluate the candidates' abilities to mitigate a station blackout due to a failure of RCP pump breakers to open, Initial Conditions : 100% power , Turnover: FW 54, AC-3A tagged out Event Malf. Event Fveet Description No. No. Type
- I MAL CRD6 C Dropped rod (to,1 #1) 2 R/N Reduce Power to 70%
3 XMT CVC23 J T 2897 (letdown HX CCW outlet temperature) fails low 4 XhtT SGN27 i S/O level LT-903X .?lls "as is" (preset) . 5 MAL EDSIC C 1 A3 bus fault /rx trip 6 MAL EDSIIB M loss of offsite power (both 161 and 345 KV) MAL EDSIl A (preset on trip) 7 LOA EDS83 C RCP RC 3D breaker does not open (D/G output breaker will not close) (preset) 8 MAL AFW1 C Turbine driven AFW pump (FW 10) fails 4 1 4
. Si.aded entries are to be initiated by a cte from exanunct.
* (N)ormal,- (R)eactivity, (T)nstrument, (C)omponent. (M)ajor
- Examiner; m ChiefExaminer: 1 O v ia
i i ES,-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-2 Event No.: 1 Page 1 of 8 Event
Description:
Dropped Control Rod Time Position Applicant's Actions or Behav!cr RO Identify event from alarms RO Determine only one rod has dropped SRO Enter AOP-02 (CEDM) Malfunction) SRO Direct BOP to adjust turbine load to match reactor power BOP Adjust turbine load to match ' ' actor power SRO Direct RO to control pressurizer pressure and level RO Monitor Pressurizer pressure and level SRO Notify Reactor Engineer SRO Consult Tech Sec 2.10.2. (Note: Requirements of this Tech Spec are covered in the actions required by AOP-02 SRO Inform RO and BOP that Tech Specs require a power reduction to less than 70% within one hour SRO Notify system Operations of impending power reduction I
t i ES-301 Operator Actions Form ES-301-4 I Scenario No.: FCS-97-2 Event No.:2 Page 2 of 8 Event
Description:
Power Reduction to 70% within one hour Time Position Applicant's Actions or Behavior SRO Direct RO and BOP to commence power reduction SRO Direct RO on method of boration to use. (Options are normal SIRWT, or boration,ing enter AOP-05 (Emergency Shutdown) shiftinj RO Begin boration ! BOP Reduce turbine load to control RCS Tc. , _ RO Monitor and control primary parameters $uring power reduction 1 BOP Monitor and control secondary parameters during power reduction
E_S-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-2 Event No.:3 Page 3 of 8 Event Descript!on: T-2897 (letdown HX CCW Outlet temp) falls low Time Position Applicant's Actions or Behavior RO Identify high letdown temperature condition from alarms RO Verify that TCV-211-2 has repositioned to bypass demineralizers RO Determine high temperature due to reduced CCW flow to letdown heat exchanger SRO Direct RO to manually control CCW flow to letdown HX RO Manually control CCW flow to restore letdown temperature SRO May direct RO to reposition TCV-211-2 __ RO Reposition TCV-211-2 if directed RO Monitor primary parameters BOP Monitor Secondary Parameters f
+
. i I
l l l ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-2 Event No.:4 Page 4 of 8 Event
Description:
Steam Generator Level (LT-903X) Fails "as is' Time Position Applicant's Actions or Behavior BOP Determine and report failure of steam generator level channel SRO Direct BOP to take manual control of steam generator level , BOP Take manual control of FW Reg Valve. Restore normal level BOP Monitor and mntrol steam generator level RO Monitor primary parameters d
ES-301 Operator Acuons Form ES-301-4 Scenario No.: FCS-97-2 Event No.:5 Pago 5 of 8 Event Description : 1 A3 bus fault / Reactor trip, loss of offsite power 4 Time Pos:Jon Applicant's Actions or Behavior SRO Direct standard Post-trip actions BOP Report No Power to buses 1 A3 and 1 A4 RO or BOP Report both diesels running at 900 RPM RO or BOP Report that D/G breakers did not close SRO Ensure that feedwater be restored BOP Restore feedwater usino FW-10 RO Perform remainder of SPTA's BOP Perform remainder of SPTA's SRO Verify completion of SPTA's 8
-- . , . . .= ... ~ -_ - .
l* ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-2 Event No.6: Page 6 of 8 Event
Description:
RCP RC-3D Breaker does not open Time Position Applicant's Actions or Behavior RO identify failure of breaker to open RO Report failure of breaker to open SRO Direct RO to direct EONT to manually trip breaker RO Direct EONT to manually trip breaker EONT Open breaker SRO Direct RO/ BOP to verify restoration of power to bus 1 A4 RO/ BOP Verify power to bus 1 A4
4 ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-2 Event No.:7 Page 7 of 8 Event
Description:
Loss of offsito power Time Position Applicant's Actions or Behavior SRO Diagnose restored event to bus 1 A4 and enter EOP-7 or EOP-02 (i (f power has beenif pow restored to bus 1 A4 SRO Direct RO to vyify Natural Circulation RO Verify Natural Circulation SRO Direct RO to monitor and control primary parameters SRO Direct RO and BOP to ensure Raw water, CCW Containment Cooling, Instrument Air and Charg,ng i Pumps are restored RO and BOP Ensure that all items listed above are restored. J
l dS-301 Operator Actions Form ES-301-4 I i Scenario No.: FCS-97-2 Event No.:8 Page 8 of 8 i Event
Description:
Turbine Driven AFW pump (fw-10) Fails Time Position Applicant's Actions or Behavior BOP Report loss of FW-10 SRO Direct BOP to monitor S!G level BOP Monitor and report S/G level SRO Enter EOP-20 due to a loss of all feedwater with a loss of offsite power SRO Determine that heat removal safety function is not being met. Scenario ends when SRO enters EOP-20 and determines heat removal safety function not met.
. I l ES401 Scenario Events Form ES-3013 Simulation Facility: Fort Calhoun - Scenario No.: SIM 97 3 Examiners: Applicants: Scenario objective: To evaluate the candidate's abilities to mitigate an excessive steam demand event complicated by the failure of automatic containment spray actuation. Initial Conditions : 100% powtr Turnover: FW-4 A tagged out, RC 3C lower seal failed, CPR in progress, NIS power range cimnnel C tagged out, Tus 1,9 and 12 bypassed. In AOP 15, entry actions taken. Event Malf. Event Event Description No. No. Type
- I MAL DSG6A C DG l Radiatorleak 2 XMT RCS97 I Pressurizer level channel 10lX fails low 3 MAL RCP10D C RC 3C lowtr and raiddle seals fail 4 R/N Emergency Shutdown 5 XMT SONIO I S/G steam flow channel FT 903 fails low 6 MAL MSSID M Steam line break in containment 20% rampover 1200 sec 7 MAL CRD5D C 2 rods fail to tully insert GF XMT CRD64 (preset) 8 MAL ESF2A I CPilS fails to actuate MAL ESF2D (preset) 4 Shaded entries are to be initiated by a cue from examiner.
* (N)ormal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor Examiner: _
Chief Examiner: C U io
ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-3 Event No.: 1 Page 1 of 8 Event
Description:
DG-1 Radiator leak Time Position Applicant's Actions or Behavior RO Identi event from
- Diesel Trouble" alarm and low water levelI ht.
SRO Enter alarm procedure. SRO Direct EONT to check diesel. (EONT) Report water on diesel room floor. SRO Direct RO to place DG 1 in "off auto" RQ Place DG-1 in "off auto" SRO Enter Tech Spec 2.7 (7 days) SRO Inform maintenance of problem
A ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS-97-3 Event No.: 2 Page 2 of 8 Event
Description:
Pressurizer level channel 101X falls low i Time Position Applicant's Actions or Behavior RO Identify and report failure of pressurizer level instrument SRO Direct RO to transfer control channels or take manual control of coittrollina channel RO Transfer control channel or take manual control of level RO Monito and control pressurizer level ORO Direct LO to select Y channel on the low level heater cutout switch RO Seloc; Y chennel on low level heater cutout switch BOP Monitor secondary parameters I
e ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS 97-3 Event No.: 3 Page 3 of 8 Event
Description:
RC 3C middle seals fall Time Position Applicant's Actions or Behavior RO Identify and communicate blah seal leakage from alarms SRO Enter alarm response procedure l RO Monitor RCP soal pressures and determine that the lower and middle sealsl on RCP O have failed SRO Direct Emergency Shutdown and enter AOP-05 (Emergency Shutdown) BOP Monitor secondary parameters p D
^%
a w -yw - e -se w r- -&_-q --rer- =-v-msv- y--. - .w' y -
ES 301 Operator Actions Form ES 301-4 Scenario No.: FCS 97-3 Event No.: 4 Page 4 of 8 J Event
Description:
Emergency Shutdown Time Position App!Icant's Actions or Behavior SRO Direct RO and BOP to commence Emergency Shutdown SRO Notify System Operations of power decrease SRO Direct RO to begin boration usina SIRWT RO S_wKeh chargina pump suction from VCT to SIRWT SRO Direct BOP to control RCS T cold by reducing turbine load BOP Reduce turbine load to control T-cold SRO Direct RO to operate control rods as required to controi ASI RO Operate control rods to control ASI RO Monitor and, control primary parameters BOP Monitor and control RCS T-cold and secondary parameters l
ES-301 Operator Actions Form ES 301-4 Scenario No.: FCS-97-3 Event No.: 5 Pa0e 5 of 8 Event
Description:
Steam flow transmitter (PT-908) falls low Time Position Applicant's Actions or Behavior BOP identify and communicate lowering FW flow and level in S/G 'B' _ SRO Direct BOP to take manual control of feedwater BOP Take manual control and re_ store feedwater level BOP Identify FT-908 as failed instrument SRO Inform 18C of failure of FT-908 BOP Continue to monitor and control S/G level RO Monitor primary parameters M
. . . - . . . - - . . - . ,,, ---- _.-.-- ..-.-.- . , , - .n. -. . . , . ,-, -- .
)
ES-301 Operator Actions Form ES-301-4 l Scenario No.: FCS 97-3 Event No.: 6 Page 6 of 8 Event
Description:
Steam line break in containment Time Position Applicant's Actions or Behavior BOP Identify and communicate lowering RCS T-cold RO Identif and communicate lowering pressurizer pressure and le el SRO May direct RO to manually trip the reactor RO If directed, trip the reactor SRO Direct the RO and BOP to perform standard post trip actions RO Perform primary post trip actions BOP Perform secondary post trip actions SRO Direct BOP to verify isolations following Steam Generator Isolation Signal BOP Verify SGIS actuation
. \
I ! l l l ffi fi Operator Actions Form ES-301-4 Scenario No.: FCS-97-3 Event No.: 7 Page 7 of 8 j Event
Description:
Two control rods fall to fully insert on trip l Time Position Applicant's Actions or Behavior i RO identify and report failure of two control rods to insert following reactor trip SRO Direct RO to initiate emergency boration RO_ initiate emergency boration SRO Direct BOP to manually throttle feedwater flow BOP Manually control feedwater flow SRO Diagnose event and enter EOP-05 or EOP-20
/
4 e r ES-301 Operator Actions Form ES 301-4 Scenario No.: FCS 97-3 Event No.: 8 Page 8 of 8 Event
Description:
CPHS falls to actuate Time Position Applicant's Actions or Behavior RO Monitor containment ressure and determine that CPHS did not actuate at set oint J SRO Direct RO to manually actuate CPHS RO Manually actuate CPHS and verify containment spray flow SRO Direct BOP to establish steam flow from intact steam generator prior to dryout of faulted steam generator BOP Establish steam flow from Intact steam generator and control RCS temperature. SRO Direct RO to monitor subcooling and 3ressurizer level to determine when HPSI'stop and throtule" criteria are met RO Monitor subcooling and when 'stop cnd throttle" pressurizer criteria are met level and report SRO Direct primary operator to throttle and/or stop HPSI flow RO Throttle and/or stop HPSI flow SRO Direct RO to monitcc and control RCS pressure to maintain subcooling between 20' and 200'F RO Monitor and control pressure to maintain subcooling between 20* and 200'F Scenario ends when 'stop and throttle criteria have been met and HPSI flow has been reduced,
- l l
l l ES 301 Sornario Event:- Form ES 3qb) Sirnulation Facility: Fort Calheurt. S enario No.: SIM 974 Er.1 miners: Applicants: Scenario objective: To evaluate the candidates' abilities to mitigate a LOCA due to a failed open IORV. Initial Conditions : 100% power , 1 Turnover: IT-4 A tagged out, Power range Ni clumnel"C" tased out, trip units 1,9 and 12 , bypnesod, CPR in progress, RC 3C lour seal failed Ewat Malf. Event Event Desenplion No. No. Type
- 1 13ST CVCl3 I VCT level channel (LCS 218LL L) fails low (bistable trip) 2 MAL NIS711 1 N1 power range channel *11" power supply failure 3 R/N Power reduction to 70%
4 XMT RRSI I PT 910 fails high ( 1000 psia)
$ SWI EllCIDV M Loss ofload 6 YLV RCSl8 C/M PORV PCV 1021 fails oixn C=)RPSRXTP (presci) 7 VLV RCS17 C PORY block valve llCV 151 will not close 8 FILE Z4201 C llPSIIMmps fail to start Shaded entries are to be initiated by a cue from examiner.
*- (N)ormal, (R)eactivity, (1)nstmment, (C)omponent, (M)ajor Examiner: _
Chief Examiner: r() [ (. iu
ES-301 Operator Actions Form ES 301-4 Scenario No.: FCS 97 4 Event No.:1 Page 1 of 8 Event
Description:
Loss of Ni power range channel'B' Time Position Applicant's Act!ons or Behavior RO Identify the failure from alarms SRO Reference AOP-15 SRO Determine the need to place "B' channel trip units 1,9 & 12 in the tripped condition,1 hour LCO, and 48 hour LCO for repair of one channel SRO Direct the RO to pull the 1, 9 & 12 trip units on 'B' RPS channel or place test switch off normal RO Pull 1,9 and 12 trip units place test switch off norma l (simulator operator action) or SRO D_etermine need to reduce power to 70% or less
. , , . , . , - , --,_,.y..,
ES-301 Operator Actions Form ES 301-4 1 _=- l Scenario No.: CCS 97-4 Event No.:2 Page 2 of 8 Event
Description:
VCT level channel falls low Time Position Applicant's Actions or Behavior RO Identify transmitter failure and operation of LCV 218-2 ' and LCV 218-3. SRO Direct RO to manually open LCV-218 2 and close LCV-218-3 RO Monitor and control primary parameters BOP Monitor and control secondary parameters = yfj-gygr --e--war og ys-, 7 + + y--------r y- g r--------- -
l I ES 301 Operator Actions Form ES-3014
- Scenario No.
- FCS 97-4 Event No.: 3 Page 3 of 8 Event
Description:
Power reduction to 70% Time Position Applicant's Actions or Behavior SRO Notify System Operations of power decrease SRO Direct RO & BOP to commence power reduction SRO Direct RO on method of boration to use RO Begin boration as directed BOP Reduce turbine load to control T cold RO Monitor and control primary parameters during power reduction BOP Monitor and control secondary parameters during power reduction SRO Coordinate RO and BOP actions during power reduction 5
- - - - n- ,-
,- , n-- , - . - . , - - - ,n. , , ,~.,~,--,.v---
a 4 a - ,a--,, _a- _ _ 4 _ __.,s ma dE._a.. ,-w-4 m. -_m~ A* **.- -m mea--+ 4 ES 301 Operator Actions Form ES-301-4 Scenario No.: FCS 97-4 Event No.:4 Page 4 of 8 1 Event
Description:
PT-910 fails high Time Position Applicant's Actions or Behavior BOP Identify rapid decrease in RCS T cold BOP Determine cause as, turbine bypass valve being open SRO Direct DOP to take manual control of PCV 910 and close valve BOP Take manual con, trol of PCV 910 and c'ose it BOP Monitor RCS Tc RO Monitor and control RCS parame_ters SRO Notify I & C of failure
- , , , . - , - , , , . - - , , . , --,.- y y-.y - --,-- ,-.. ,_ . . - ., --
ES-301 Operator Actions Form ES 301-4 Scenario No.: FCS 97-4 Event No.:5 Page 5 of 8 Event
Description:
I oss of load Tiine Position Applicant's Actions or Behavior SRO Direct RO and BOP to perform SPTAs RO Perform primary post trip actions BOP Perform secondary post trip actions im M J anse 9dum
- ~ , , , - - _ - . . . , , . . . + _ . _ _ _ . __ _ . . . . . , , , , . _ _ _ . , _ . _ . . _ - ,
. _ . ._. , - .-. . _ _ - _ - . .. . .-. . . . . -_ =-.. . . . - . . - -
ES 301 . Operator Actions Form ES-301-4 Scenario No.: FCS 97-4 Event No.:6 Page 6 cf 8 Event
Description:
PORV falls open Time Position Applicant's Actions or Behavior RO Identify failure of PORV to close SRO May, direct the RO to close the PORV block valves RO Report that HCV-151 will not close SRO Verify completion of SPTAs by both operators k 9 4
..-4, - ,# ..e . . , ,y.. ,. - . . ._ .. 7..,,,. ,,.% . _ _ . . , --
ES-301 Operator Actions Form ES-301-4 Scenario No.: FCS 97-4 Event No.:7 Page 7 of 8 Event
Description:
PORV block valve will not close Time Position Applicant's Actions or Behavior RO Identify failure of HCV-151 to close SRO Direct RO to complete SPTAs RO Perform remainder of SPTAs BOP Perform remainder of SPTAs RO Trip one RCP in each !oop when pressure drops to 1350 psfa RO Trip remaining RCPs when NPSH lost SRO Diaarose event and enter EOP-03
t ES-301 Operator Actions Form ES-3014 Scenario No.: FCS 97-4 Event No.:8 Page 8 of 8 Event
Description:
Failure of HPSI pumps to auto start following PPLS Time Position Applicant's Actions or Behavior RO or BOP identify failure of HPSI pumps to start SRO Direct all HPSI pumps manually started RO or BOP Start all HPSI pumps RO or BOP Verify adequate Si flow SRO Verify all safety functions are satisfied Scenario ends when HPSI flow has been restored and all safety functions are satisfied.
Topic (s): RCA Entry ard Exit tecation (s): RCA Acocas Cmtr01 Approximate Time: 10 minutes Actual Time : Reference (s): (1) Oltr.Redisticnt Workce Training
-(2) NRC K/A 194001, K1.03 (RO2.R/SRO3,4)
(3). NRC K/A 194001, Kl.04 (RO3.3/SRO3.$) Vertfy corrent refervoce revielsen snatch these listed above s en Operskt's Name: _ SS # : All Critical Steps ( * ) must be perismed or simulated in accordance with the standards contained in this JPM. 1hc operstor's petigmance was evaluated as: SNriSFACTORY UNSATISFACTORY Evaluator's Signature : Date : ,_ Reamm,if unsatisfactory : Operator's review : Date: Tools & Equipnent : None Safety Considerations: This JPM requires entryinto tiw RCA. Comments: i
leisteting Caos - As entry into the RCA is required as part of the operating esaan. START CRITICAL STEP
- EIEMENT STANDARD 1.* Review RWP. Read RWP
- 2. Check surwy maps check sevey maps f<r redelogical conditions in erses to be entered 3.* Obtain dasimetry Verify TLD attached u scourity badge. Obtain ALHOR d* Signin on appropriate RWP insert ALNOR into roadw enter PID and RWP number.
$. Inform RP Personml about the netwe of your Tell RP at sooens control where you are going and
~
entry what you willbe doing.
- 6. Enter RCA RCA entered 7.* Comply with RWP and all postings within Notify RP prior to operating WD.1080 during RCA perfonnance of waste gas JPM.
No violation ofposted requirements CUE What action should be taken if you sw a person uncono6ous and bleeding badly in a contasalmated aese? 9.- Answer Queticst Contact control roont !!nta cont.minated area to administer emergency first sid. 10.* Monitor for personnel contamination raior to Monitor for cantamination using PCM. exitins , CUE: PCM Indicates esetandmaties on right hand.
- 11. Reenter PCM and recount PCM reentered.
CUE: PCM Indicates contamilnat6en se right hand. 12.* Contact RP candidiate indicates that he would inform RP toch of repeated PCM indications of guitaminatica
- 13. Sign out of RCA - Innert ALNOR in reader, Enter PID mimber, ..
conftrm dose and place ALNOR in charging rock. Teralmat6en Criteria RCA has been esited. _ __..,__._ _ _ _ _ a
Revision 0 February 25,1997 Fort Oalhoun Station Operator Training JOB PERFORMANCE MEASURE JPM No: JPM OP ST SHIFT 0001 JPM
Title:
Required Shift Surveillance Approximate Time: 8 minutes Actual Time: i Reference (s): 1) OP ST SHIFT 0001 (rev 51)
- 2) NRC K/A 001010,A4.04,RO3.5/SRO4.1 Verify current referonce revisions match those listed above Operator's Name: _ _ SS #:
All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. The operators performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluator's Signature: . Date: Reason, if unsatisfactory: J 4
4 Revistor O February 25,1997 Fort Calhoun Station. Operator Training JOB PERFORMANCE MEASURE JPM No: JPM OP 8T SHIFT 0001 JPM
Title:
Required Shift Surveillance initiating Cue: Complete the Shift Supervisor review for Wednesday of the attached portion of OP-ST SHIFT 0001. STANDARD:
- GM/057 counts have doubled, therefore sampling per S.O. G-105 muet be started.
5 1 -
i t f Initiating Cue: , Complete the Shift Supervisor review for Wednesday of the attached portion of OP ST SHIFT 0001. l F r h k f f i I f e 4 9
=
1
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4 Fort Calhoun Station
- Unit No.1 OP ST-SHIFT-0001 SURVElLLANCE TEST
Title:
OPERATIONS TECHNICAL SPECIFICATION REQUIRED SHIFT SURVEILLANCE FC 68 Number: 48410 Reason for Change: Add remarks for dolng source checks of rad monitors, C/LT note deleted, and change reference for S/G Blowdown flow to ODCM. Change Diesel Fuel Inventory to Match Technical Specification. Contact Person: D Hochstein ISSUED: 02-27 97 9:30 am RS2
OP-ST-SHIFT-0001 FORT CALHOUN STATION PAGE 1 OF 49 SURVEILLANCE TEST OPERATIONS TECHNICAL SPECIFICATION REQUIRED SHIFT SURVEILLANCE SAFETY RELATED
- 1. PURPOSE To satisfy Technical Specification requirements for all Operations Shiftly and Daily Surveillance Checks.
- 2. EffERENCES/ COMMITMENT DOCUMENTS 2.1 Applicable Technical Specifications are listed on the appropriate Shift Data Sheets.
2.2 Applicable procedures are listed on the appropriate Shift Data Sheets. 2.3 Ongoing Commitments
- CID 882063, LER-88-013, Stop 5.1
- CID 780026, LER-78-02,.Page 5
= CID 910569, LER-91-14, Pages 14,15 and 16 2.4 Standing Order 80 G-23, Surveillance Test Program
- 3. DEFINITIONS 3.1 Surveillance Interval (Technical Specification 3.0.2)
- Shift - at least once per o!ght (8) hours.
- Daily - at least once per 24 hours.
3.2 Specified Time interval (Technical Specid.2 tion 3.0.1)
- Each Surveillance Requirement shall be performed within the specified Surveillanco interval with a maximum allowable extension NOT TO EXCEED 25% of the Surveillance Interval RS2
l OP-ST-SHIFT 4001 l - FORT CALHOUN STATION PAGE 3 OF 49 SURVElLLANCE TEST INITIALS /DATE
- 3. INITIAL CONDITIONS l
6.1 Procedure, revision verification: Master Revision No, if 2 b / J-* V> 6.2 An RWP has been issued, if required. RWP No. 97 - N#4 O/u-ry '
- 7. PROCEDURE CAUTION All data required to be taken for the applicable mode Bad shift muni be completed within the first two (2) hours of the applicable normal operating shift.
NOTE: The Surveillance Test Signature Sheet only needs to be completed once each week for each individual recording, evaluating gI reviewing data on the Shift Data Sheets. 7.1 Complete the Surveillance Test Signature Sheet. 7.2 Record all data required to be taken during the applicable , mode BDd Operating shift as scheduled on each Shift Data Sheet. 7.3 WHEN all applicable data has been recorded for the shift, THEN the Control Room Operator must sign the " Completed by" section of the Shift Review Sheet for the appropriate Day and Shift. REMARKS b RS2
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OP-ST-SHIFT-0001 ' FORT CALHOUN STATION PAGE 19 OF 49 - SURVEILLANCE TEST SHIFT DATA SHEE7 WEEK ENDING: AREA RADIATION MONITORS RM-091A RM-091B (NSTRUMENT TIME INITIALS APPUCABLE MODES: S Modes 1,2,3. 4 and 5 S Meter Reading R/Hr 4f 4f N PROCEDURE
REFERENCE:
N Wam S.P. R/Hr xyg ;gg ,,g M OFMm1 M Meter Reedog R/Hr 4f 4/ O 'A g2 2330-0730 O N TECH. SPEC. REFERENCE N Wam S.P. R/Hr yfg3 ;gg3 mo - e 2.21 Table 2-10, item 1 T Meter Reading R/Hr T e 3.1 Table 3-3, Pem 3.a 4, f gj 2330-0730 0 U ACCEPTANCE CRITERIA: s E Warn S..P. R/Hr yyg3 p.fc3 ufg E
~
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- Meter are onreadings se and are lessdig. displayed or E
4f <f 2330-0730 E the D am D Warn S P. R/Hr ypa jygg opy . {uper/viAlert s notified T Meter Reading R/Hr T e Observed on scale meter response to 3 -7 y WamIAlert Sc^ point Check U Warn S.P. R/Hr REMARKS F Meter Reading R/Hr F R 2330-0730 R I Warn S.P. R/Hr I l S Meter Reading R/Hr S 6 A 2330-0730 A T Wam S.P. R/Hr T 1 RS2'
FORT CALHOUN STATION OP ST-SHIFT-0001 SURVEILLANCE TEST PAGE 48 OF 48 Surveillance Test Signature Sheet All aersons partlcipating in the performance of this test shall enter their printed name, signature anc initials below, WEEK ENDING: 8. 2 2 W NAME (PRINT) SIGNATURE INITIALS rluiML 0 /%9 V/WS $95
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t Revision 0 March 3,1997 Fort Calhoun Station- Operator Training JOB PERFORMANCE MEASURE JPM No: JPM-PC JPM
Title:
Review of a proposed procedure change for approval. Approximate Time: 8 minutes Actual Time: _ _ Reference (s): 1) FC 68B
- 2) NRC K/A 194001 GEN A1.01,RO3.3/SRO3.4 Verify current reference revisions match those listed above Operator's Name: SS #:
All Critical Steps (*) must be performed or simulated in accordance with the standards contained in this JPM. The operators performance was evaluated as: SATISFACTORY UNSATISFACTORY Evaluators Signature: Date: Reason, if unsatisfactory:
- . - .-- . - . .. . . - - . . .- - . _ . - . . - ~ - _ -
i Revision 0 March 3,1997 Fod Calhoun Station- Operator Training i JOB PERFORMANCE MEASURE JPM No: JPM-PC JPM
Title:
Review of a proposed procedure change for approval. Initiating Cue: Review the attached Temporary Procedure Change for approval as the Plant Supervisory Staff Member w/SRO License. STANDARD:
- This Temp. Procedure change results in a change of intent and therefore cannot be approved as a Temp. Change.
4 4 w
FORT CALHOUN STATION FC-68C FORM R5 STA A/D/ RD CHANGE OF INTENT DETERMINATION Procedure Change No.: Procedure No.: Yes No
- 1. Is a new procedure being initiated? /
- 2. Does the revision add, remove or modify acceptance criteria that y affects equipment operability? -
- 3. Does the revision modify the specified operational condition of a system or component? Answer "NO" if change is due to modification /
or ECN and a 10 CFR 50.59 has been completed for Mod /ECN.
- 4. Does the revision reduce or provide less conservative " Hold Point" y criteria?
- 5. Does the revision modify any portion of the procedure involving a known commitment? Answer "NO" if a review is conducted and it is V
determined that the commitment will be satisfied?
- 6. Would the revision result in a reduction of personnel or equipment y safety?
- 7. Does the revision change the original purpose of the procedure? /
- 8. Oces the revision violate any requirements of current Technical y Specifications?
The following sections were reviewed:
- 9. Does the revision deviate from the USAR description? V The following sections were reviewed:
A Change of intent is involved if any checklist item is checked "YES". N [] Change of intent is involved (if yes, complete, FC-154). Change of Intent is NOT involved. COMMENTS: NOTE: Preparer must be 50.59 qualified.
- Completed by (Preparer)
- Date:
l i
. _ . _ . _ . . _ _ . _ ._ . . . . _ ._ . ._ . _ ~ . __ _ . . _ . _ _
. initiating C,ue:
-Review the attached. Temporary Procedure Change for approval as the Plant Supervisory Staff Member w/SRO License.
I 4 i
~ _ . . _. ._ _ ____
. FC-68C FORT CALHOUN STATION ,
FORM , R5 i CHANGE OF INTENT DETERMINATION Procedure Change No.: Procedure No.: Yes No 1, is a new procedure being initiated? .,
- 2. Does the revision add, remove or modify acceptance criteria that affects equipment operability?
- 3. Does the revision modify the specified operational condition of a system or component? Answer "NO' if change is due to modification or ECN and a 10 CFR 50.69 has been completed for Mod /ECN.
- 4. Does the revision reduce or provide less conservative ' Hold Point" criteria?
- 5. . Does the revision modify any portion of the procedure involving a known commitment? Answer 'NO' if a review is conducted and it is determined that the commitment will be satisfied?
- 6. Would the revision result in a reduction of personnel or equipment safety?
- 7. Does the revision change the original purpose of the procedure?
- 8. Does the revision violate any requirements of current Technical Specifications? .
The folloviing sections were reviewed:
- 9. Does the revision deviate from the USAR description?
The following sections were reviewed: A Change of Intent is involved if any checklist item is checked "YES". [] Change of Intent is involved (if yes, complete, FC-154). [] Change of inton'.is NOT involved. COMMENTS: NOTE: Preparer must be 50.59 qualified. Completed by (Preparer): Date: l l l
i-CID/1R/ CON RPT Numoer F'ORF CAMOUH BTATIces rc.68B TDfroRAAT 51LocEDURE caA340s PJiu2UIs? R13 NOTE: If any questions on FC-68C (Change of Intent) are answerea YES, a Temporary Change can NOT t>e er. ace. HOTE: Temporary Changes can NOT be made to EOP, AOP, RERP ano so. Sectiert 1
- Iftitistar Use Ordy FROCEDURE REVISIOH VERIFICATICat cesaute n a . & - W- A M ~ n'Y' H Master Revision No. /d Title s d i r / Neht ** be'N" #+4 e*d M'dd I N I<1.7 '
signature 7 h ,. a / V 'r
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- r // 4 //t? d A l?/J P.,a a cfb ec*k <_ .s / r.J f* f/. /M'/
/o)A//W Adt mfcu ar, oh+ A m u ,,. . = / J.MM 4 ' A i. r!.fw/ _ er / ,=w - 6, e FAP anast be returded to Document control vnhan 72 hours.
hur dar,u) ! / r S* U* O / Je M s non. No. Precerer Naaw initials Dat.e Extension m Interim approval signfies that the intent of the original procedure is NOT changed and that Training is not required prior to taplementing the TPC.
/ / / /
tient suoerviserv staf f Memoer Date Time Plant Suee rviserv Sta f f Member w/ SRO I,1 cense Date Time
/
paview Wyi ( ) QR ( ) PRC Asstaned RDH/PRC M*,nber Date A.ssioned Primarv Qualified Revtewer see:Aes II
- ter rec / Qualified Mytower Use cair secties III
- Per tac Sartow Dee only Recosamenced Review Sys ( ) Quorum ( ) subcosaattee laseeporate into Operettag Manuali d ) Yes ( l No 9ewtow bT 9eeswa telaisum of skal Changes Hade
( ) Yes ( ) No During Review Chants af f ects associ6ted f orms) (If yes, attacA ft.160) ( ) Yes ( ) No ( ) Change affects en Alignment Choctinstt t ) Yes ( ) No Manager
- Operettons Late (If yes, attach TC-408)
( 1 Yes ( ) No ( ) Requases sand (NsRG) revnew) ( ) Yes ( ) No Manager
- 6ystem Eagt Late
*quares ttD ft!*t changet ( ) Yes ( ) No ( )
Yes ( ) No() Manager - Masatoaance Late
, quires 3000 series sury Yest svaswt ( ) Yes ( ) No 4 1 Yes ( ) No ( )
tt troceanare changedi ( ) Yes ( ) No Manager . Me trotection Date l Could change impact field wort? ( ) Yes ( ) No ( ) Yes ( ) No ( ) (If yes, notify Manager - chemistry Late l
/
( Yes ( ) No ( ) Manager
- Nuc L&CGGanaq Late t /
Yes ( ) No ( )
, Manager - Tranaang Late l #
' Creas-etesattamary/Creas rushstional/otJhes Devtew ( ) Yes ( ) No ( l Yes ( ) No ( ) Hamager = Nuc Projects Late Print Name signature /Date ! ( ) Yes ( ) No ( ) Reactor Lagnaeer Late ( ) Yes ( ) No() Assistant tiant namager case ( l ' Yes ( ) No() other: Data seetten IT - sectew by ashemmeutttee Unaatmous concurrence Dy, saccommattee steepers Yes ( ) No ( ) i
-emiu.e neeuna n.a.aer 1
Raccounenced f or Approvan t ( l Yes ( ) No 1 I cu ei n ant e nc M.ee r s e r iea rv oua t t r iea m.w s.w r case L , l f irovao sy / ' Effective Dates riant nanaoer nesocasam e icoartment b aa . ate cu ect ne cate enven t ca r.t roll j KT Reenew SAE (MSRG) Review YvDino Poview HP Proofread By Fev. No. (latu Datop (lett/Datet U nit / Dates (lait /Datel (latt/ Dates Dates N/tCs l l
( FORT CALHOUN STATION OP ST AFW 0004 GURVEILLANCE TEST PAGF,8 OF 0 CAUTION DPI 1038 should be valved out before stopping FW 10 to avoid subjecting it to a large reverse AP due to warm up steam. 7.1.11 Valve out DPI 1038 by clos;ng MS-DPI 1038B and FW DPI 10388, and opening MS DPI 1038 E. / I&C Independent Verification / 7.1.12 Stop FW 10 per Ol AFW 4. / 7.2 AFW Pumo FW 6 Ooerability Test 7.2.1 TI 1382 is 5120'F 9 /
- r,2,3 Sted Pf! S per O! ?Pf! ' and 2!!'"."
- 10 Mete /
":17~ u p.
7.2.2 Close FW-171(FW 6 Dischargo Valve) l 7.2.3 Start FW 6 per OI AFW 4 and allow a 10 minute warmup with the discharge valve closed. 7.2.4 Open Valve FW 175.}}