ML072960547
| ML072960547 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/2007 |
| From: | - No Known Affiliation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 50-324/07-301, 50-325/07-301, ES-301, ES-301-2, NUREG-1021 R9 50-324/07-301, 50-325/07-301 | |
| Download: ML072960547 (33) | |
Text
{{#Wiki_filter:Final Submittal (Blue Paper) BRUNSWICK JULY-AUG EXAM - 325,324/2007-301 FINAL JPMS FINAL JPMS
- 1. ADMINISTRATIVE JPMs
- 2. IN-PLANT JPMs
- 3. SIMULATOR JPMs (CONTROL ROOM)
ES-301. Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Brunswick Date of Examination: JULY / 2007 Exam Level (circle one): RO / 5RO (U) Operating Test No.: NRC 2007 Control Room Systems@ (8 for RO; 2 or 3 for SRO-U, including 1 ESF) Type Code* Safety System / J PM Title Function S-1 Uncoupled Control Rod During Startup N,A, L, S 1 5-2 RCIC Failure to Isolate N,A,S,E 5 S-3 Core Spray Pump Surveillance Min Flow Valve Failure N,A,S 2 5-4 Restore Shutdown Cooling following a spurious isolation lAW N,A, L, S 4 AOP-15 S-5 Primary Containment Venting During Personnel Entry. D, L, S 9 S-6 Manual Transfer of 4160 Emergency Bus Supply from the DG to N,S 6 the Normal Feeder lAW OOP-50.1 S-7 RWM failure to enforce rod blocks D,S 7 S-8 Re-Establish RBCCW For Drywell Cooling to the Blacked Out Unit D,E,S 8 Per AOP-36.2. (RO) In-Plant Systems@ (3 for RO; 3 or 2 for SRO-U) P-1 Station Blackout: Crosstie of 4KV E-Buses D,E 6 P-2 Control Room Evacuation lAW AOP-32, Placing the RHR Service D,R 8 Water System in Operation. P-3 Staging the Reactor Recire Pump Seals. D,R 1 NUREG-1 021, Revision 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes (A)lternate path 4-6/4-6/2-3
~ z (C)ontrol room (D)irect from bank (p~ 9 I ~ 8 I .2 4 (E)mergency or abnormal in-plant '~1/~1/~1 (L)ow-Power I Shutdown }~1/~1jl1 (N)ew or (M)odified from bank including 1(A) 5""~ 2 I ~ 2 I ~<"1 (P)revious 2 exams ~s 3 I S 31'S 2 (randomly selected)
(R)CA ~ 11 ~ 11 ~1 (S)imulator 2007 NRC Examination Summary Description of JPMs S-1 This is a new alternate path JPM in the Reactivity Control safety function area. The candidate will be pulling control rods for startup when a rod becomes uncoupled. Actions will be required to insert/re-couple the control rod. S-2 This is a new alternate path JPM in the Containment Integrity safety function area. The candidate will be placing RCIC in service when an exhaust diaphragm rupture occurs and RCIC will fail to isolate. Actions will be required to manually isolate RCIC . S-3 This is a new alternate path JPM in the Reactor Water Inventory Control safety function area. The candidate will be performing the Core Spray Operability Surveillance and the minimum flow valve will fail to function properly. This will require actions to prevent equipment damage. S~4 This a new alternate path JPM in the Heat Removal From Reactor Core Safety Function area. The candidate will be required to restore Shutdown Cooling following a spurious isolation signal lAW abnormal procedures and restart an RHRSW pump following a pump trip. S-5 This a bank JPM in the Radioactivity release safety function area. The candidate will be required to startup Primary Containment Ventilation during personnel entry, per 20P-24 using both purge exhaust fans. S-6 This a new JPM in the Electrical safety function area. The candidate will be required to manually transfer the 4160 Emergency Bus Supply from the DG to the Normal Feeder. S-7 This a bank JPM in the Instrumentation safety function area. The candidate will perform a portion of the RWM operability check. S-8 This a bank JPM in the Plant Service System safety function area. The candidate will continue re-establishing Drywell Cooling per AOP-36.2 P-1 This is a bank JPM in the Electrical safety function area. The candidate will be required to locally cross-tie the 4KV emergency buses following a station blackout. NUREG-1021, Revision 9
hNHL ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 P-2 This is a bank JPM in the Plant Service System safety function area. The candidate will be required to place RHRSW inservice following a control room evacuation. P-3 This is a bank JPM in the Reactivity Control safety function area. The candidate will be required to place the Reactor Recirc Pump Seals in service.
- NUREG-1021, Revision 9
ES-401 BWR Examination Outline Form ES-401-1 Facility: Brunswick NRC Date of Exam: JULY 2007 RO KIA Category Points SR-O-O-nly Points Tier Group K K K K K K A A A A G Tot A2 G* Total 1 2 3 4 5 6 1 2 3 4
- al
- 1. 1 5 4 1 2 4 4 20 3 4 7 Emergency
& 2 2 1 2 N/A 1 1 N/A 0 7 2 1 3 Abnormal Plant Tier Totals 7 5 3 3 5 4 27 5 5 10 Evolutions 1 4 3 1 5 1 2 *0 2 3 2 3 26 2 3 5 2.
Plant 2 1 0 1 2 2 1 2 0 1 2 0 12 0 1 2 3 Systems' Tier Totals 5 3 2 7 3 3 2 2 4 4 3 38 3 5 8
--- *--3-:--Geflefi*e-KA0wleags*-a-r-1d..**.. .1----- .---2.------- ---.---3--.------- - - - . . 4------. - .-. *"1--0---*-*-------*..*j --- .....~ __3 4.._.._--T--- 11"-" __' _. .
Abilities Categories 2 2 2 4 2221 Note: 1. Ensure that at least two topics from every KIA category are sampled within each tier of the RO outline (Le., the "Tier Totals" in each KIA category shall not be less than two). Refer to Section D.1.c for additional guidance regarding SRO sampling.
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Select topics from many systems and evolutions; avoid selecting more than two KIA topics from a given system or evolution unless they relate to plant-specific priorities.
- 4. Systems/evolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the category/tier.
6.* The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO KlAs must also be linked to 10 CFR 55.43 or an SRO-Ievel learning objective.
- 7. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals for each system and category. Enter the group and tier totals for each category in the table above; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled "K" and "A".* Use duplicate pages for RO and SR9-only exams.
- 8. For Tier 3, .enter the KIA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.
- 9. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate KIA statements.
NUREG-1021 1
ES-. Brun.NRC . Form 1-1 Written Examination Outline i Emergency and Abnormal Plant Evolutions - Tier r Group 1 i E/APE#/NameSafetyFunction ~ Number I KIA Topic(s) limp. ~I Ability to determine and/or interpret the following as 295001 Partial or Complete Loss of Forced Core Flow they flPply to PARTIAL OR COMPLETE LOSS OF Circulation I 1 & 4 X AA2.01 3.8 76 FORCED CORE FLOW CIRCULATION: Power/flow map+...................................... 295003 Partial or Complete Loss of AC / 6 X Conduct of Operations: Knowledge of system status 2.1.14 3.3 77 criteria, which require notification of plant personnel. I Emergency Procedures / Plan Knowledge *of 295024 High Drywell Pressure I 5 X 2.4.31 annuhciators alarms and indications, and use of the 3.4 78 resp6nse instructions. Emetgency Procedures/Plan: Knowledge of symptom 295025 High Reactor Pressure / 3 X 2.4.6 "2.9 79 base~ EOP mitigation strategies 295028 High Drywell Temperature / ~ AbilitY to determine and/or interpret the following as X EA2.01 they :apply to HIGH DRYWELL TEMPERATURE: 4.1 80 D~ell Temperature Abili~ to determine and/or interpret the following as they :apply to LOW SUPPRESSION POOL WATER 295030 Low Suppression Pool Water Levell 5 X EA2.02 3.9 81 LEVEL: Suppression pool
- tem~erature ..........................
Conduct of Operations: Ability to perform specific 295038 High Off-site Release Rate / 9 X 2.1.23 syst~m and integrated plant procedures during all 4.0 *82 mod~s of plant operation. Ability to determine and/or interpret the following as 295001 Partial or Complete Loss of Forced Core Flow they~apply to PARTIAL OR COMPLETE LOSS OF X AA2.05 FOR1CED CORE FLOW CIRCULATION: Jet pump 3.1 39 Circulation / 1 & 4 operrbility: Not-BWR-1 &2.................... Kno~ledge of the operational implications of the follo~ing concepts as they apply to PARTIAL OR 295003 Partial or Complete Loss of AC I 6 X AK1.03 COMPLETE LOSS OF A.C. POWER: Under 2.9 40 volt~ge/degraded voltage effects on electrical load~ ................................ I Con~uct of Operations: Ability to locate and operate 3.9 41 295004 Partial or Total Loss of DC Pwr 16 X 2.1.30 com~onents, t including local controls. Knotedge of the interrelations between MAl N 3.8 42 295005 Main Turbine Generator Trip / 3 X AK2.01 TUR j INE GENERATOR TRIP and the fo,lowing: RPS Abilt to determine and/or interpret the following as 295006 SCRAM / 1 X AA2.01 they apply to SCRAM : Reactor 4.5 43 pow,r......................................... NUREG-1021 2
- ES-e Brun~NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier I
1 Group 1 Form 1-1 E/APE # I Name Safety Function Number KIA Topic(s) KnO~ledge of the interrelations between CONTROL 295016 Control Room Abandonment / 7 X AK2.01 ROOM ABANDONMENT and the following: Remote 4.4 44 shut~own panel: Plant-Specific................ I Ability to determine and/or interpret the following as 295018 Partial'or Total Loss of CCW /8 X they ~pply to PARTIAL OR COMPLETE LOSS OF AA2.04 2.9 45 CO~PONENT C?OLING WATER: System flow* 1*********......***..........**..**..***** I Ability to operate and/or monitor the following as they 295019 Partial or Total Loss of Inst. Air I 8 X AA1.01 apply to PARTIAL OR COMPLETE LOSS OF 3.5 46 INSljRUMENT AIR: Backup Air supply Knowledge of the interrelations between LOSS OF 295021 Loss of Shutdown Cooling 14 X AK2.03 SHUTDOWN COOLING and the following: 3.6 47 RHRrshutdown cooling.................................. Kno~ledge of the interrelations between REFUELING 295023 Refueling Acc Cooling Mode I 8 X AK2.03 ACClDENTS and the following: Radiation monitoring 3.4 48 equipment........................ Equipment Control Knowledge of bases in technical 295024 High Drywell Pressure I 5 X 2.2.25 specifications for limiting conditions for operations and 2.5 49 safe~y limits. Conduct of Operations: Ability to perform specific 295025 High Reactor Pressure / 3 X 2.1.23 syst~m and integrated plant procedures during 3.9 50 diffe~ent modes of plant operation 295026 Suppression Pool High Water Temp. 15 X 2.1.2 Con~uct of Operations: Knowledge of operator 3.0 51 resppnsibilities during all modes of plant operation. Kno~ledge of the reasons for the following responses 295028 High Containment Temperature / 5 X EK3.01 as tHey apply to HIGH DRYWELL TEMPERATURE: 3.6 52 Em~rgency depressurization Ability to operate and/or monitor the following as they 295028 High Drywell Temperature I 5 X EA1.04 apply to HIGH DRYWELL TEMPERATURE: Drywell 3.9 53 presfure...................................... Knofledge of the operational implications of the following concepts as they apply to LOW 295030 Low Suppression Pool Water Levell 5 X EK1.01 3.8 54 SURPRESSION POOL WATER LEVEL: Steam condensation .................................... I Kno~ledge of the operational implications of the folloWing concepts as they apply to REACTOR LOW 295031 Reactor Low Water Levell 2 X EK1.02 3.8 55 WAt~.R LEVEL: Natural circulation: Plant-Speplflc.................. I NUREG-1021 3
ES-.* Brun.NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier Group 1 Form 1-1 E/APE # I Name Safety Function Number KIA Topic(s) Abilit~ to determine and/or interpret the following as they ~pply to SCRAM CONDITION PRESENT AND 295037 SCRAM Condition Present and Power Above X EA2.06 REAGTOR POWER ABOVE APRM DOWNSCALE 4.0 56 APRM Downscale or Unknown /1 OR yNKNOWN : Reactor pres~ure ...................................... Kn041edge of the operational implications of the 295038 High Off-site Release Rate / 9 X EK1.02 folloVl(ing concepts as they apply to HIGH OFF-SITE 4.2 57 REL~ASE RATE: Protection of the general public Knowledge of the operation applications of the 600000 Plant Fire On-site / 8 X AK1.02 following concepts as they apply to Plant Fire On Site: 2.9 58 Fire Fighting KIA Category Point Totals: 4/4 5 4 1 2 4/3 Group Point Total: I; 20/7 NUREG-1021 4
ES-4 Brun~NRC Form 1-1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier Group 2 II E/APE#/NameSafetyFunction ~ Number I KIA Topic(s) limp. ~ Ability to determine and/or interpret the following as they 295008 High Reactor Water Levell 5 x AA2.01 appl~ to HIGH REACTOR WATER LEVEL: Reactor Level 3.9 83 I Conduct of Operations: Ability to explain and apply all 295029 High Suppression Pool Water Levell 5 x 2.1.32 systehl limits and precautions. 3.8 84 I Abilit~ to determine and/or interpret the following as they 295032 High Secondary Containment Area Temperature appl~ to HIGH SECONDARY CONTAINMENT AREA 15 x EA2.02 TEMPERATURE: Equipment 3.5 85 oper~bility . Knowledge of the interrelations between HIGH DRYWELL 295010 High Drywell Pressure I 5 x AK2.05 PRESSURE and the follOWing: Drywell cooling and 3.7 59 venti(ation . Knowledge of the operational implications of the following conc~pts as they apply to INCOMPLETE SCRAM: (CFR 295015 Incomplete SCRAM 11 x AK1.02 41.8 ~o 41.10) Cooldown effects on reactor 3.9 60 pow~r . Kno~ledge of the operational implications of the following conc6pts as they apply to INADVERTENT 295020 Inadvertent Cont. Isolation I 5 & 7 x AK1.01 CON~AINMENTISOLATION: Loss of normal heat 3.7 61 sink.~ . Abilit~ to operate and/or monitor the following as they 295022 Loss of CRD Pumps 11 x AA1.01 appl* to LOSS OF CRD PUMPS: CRD Hydraulic System 3.1 62 I AbilitY to determine and/or interpret the following as they 295029 High Suppression Pool Water Level / 5 x EA2.02 apply to HIGH SUPPRESSION POOL WATER LEVEL: 3.5 63 Reaqtor pressure...........*.......................... Kno~ledge of the reasons for the following responses as 295033 .High Secondary Containment Area Radiation they :apply to HIGH SECONDARY CONTAINMENT AREA Levels / 9 x EK3:04 RAD'IATION LEVELS: Personnel 4.0 64 evac~ation I KnoWledge of the reasons for the following responses as 295035 Secondary Containment High Differential they!apply to SECONDARY CONTAINMENT HIGH Pressure 15 x EK3.01 DIF1EREN.~IAL PRESSURE: Blow-out panel operation: 2.8 65 Plan I-Specific . KIA Category Point Total: 0/1 2 2 1/2 Group Point Total: -I i 7/3 NUREG-1021 5
ES-4 Brun~NRC Form 1-1 Written Examination Outline Plant Systems - Tier 2 Group 1 [ S-yst~~#/N~~~- - - Number I KIA Topics I Imp. [§!] Conduct of Operations: Ability to perform specific 203000 RHRlLPCI: Injection X 2.1.23 sys~em and integrated plant procedures during all 4.0 86 Mode mo?es of plant operation. Abi'ity to (a) predict the impacts of the following on the/REACTOR PROTECTION SYSTEM; and (b) ba~edon those predictions, use procedures to 212000 RPS X A2.02 3.9 87 corfect, control, or mitigate the consequence"s of those abnormal conditions or operations: RPS bus po~er supply failure Ability to (a) predict the impacts of the following on theilNTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use 2150031RM X A2.01 3.7 88 procedures to correct, control, or mitigate the consequences of those abnormal conditions or op~rations: Power supply degraded Emjergency Procedures / Plan Knowledge of 205000 Shutdown Cooling X 2.4.11 3.6 89 abryormal condition procedures Emlergency Procedures / Plan Knowledge symptom 300000 Instrument Air X 2.4.6 4.0 90 ba~ed EOP mitigation strategies. Abi'lity to monitor automatic operations of the 203000 RHRlLPCI: Injection X A3.09 RH'R/LPCI: INJECTION MODE (PLANT SPECIFIC) 3.6 1 Mode including: Emergency generator load sequencing
-I Knbwledge of electrical power supplies to the 205000 Shutdown Cooling X K2.02 2.5 2 follbwing: Motor operated valves i
Abi)ity to manually operate and/or monitor in the 205000 Shutdown Cooling X A4.05 3.2 3 control room: Minimum flow valves
~
Kn~wledge of HIGH PRESSURE COOLANT INJECTION SYSTEM design feature(s) and/or X K4.18 3.2 4 206000 HPCI int~rlocks which provide for the following: Pump mi~imum flow: BWR-2,3,4 Kn6wledge of electrical power supplies to the 217000 RCIC X K2.02 2.8 5 follbwing: RCIC Initiation logic I Knbwl~dge of the physical connections and/or caJ.se- effect relationships between LOW 209001 LPCS X K1.09 3.2 6 PRESSURE CORE SPRAY SYSTEM and the follbwing: Nuclear boiler instrumentation I Ability to manually operate and/or monitor in the X A4.09 3.2 7 2002002 Recirc Flow Control co~trol room: Core Flow NUREG-1021 6
ES-4 Brun~NRC Form 1-1 Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topics Abi(ity to (a) predict the impacts of the following on the/STANDBY LIQUID CONTROL SYSTEM; and 211000 SLC X A2.04 (b) based on those predictions, use procedures to cortect, control, or mitigate the consequences of 3.1 8 thoke abnormal conditions or operations: Inadequate sys!tem flow I 212000 RPS Kn~wledge of the effect that a loss or malfunction of X K6.05 the~ollOWing will have on the REACTOR 3.5 9 PR 1 TECTION SYSTEM: RPS Sensor inputs I Abi,lity to (a) predict the impacts of the following on thelREACTOR PROTECTION SYSTEM; and (b) 212000 RPS X A2.09 ba~ed on those predictions, use procedures to. correct, control, or mitigate the consequences of 4.1 10 tho$e abnormal conditions or operations: High corhainment I
/Drywell pressure Kn6wledge of the effect that a loss or malfunction of 2150031RM X K6.05 the: following will have on the INTERMEDIATE 3.1 11 R~NGE MONITOR (IRM) SYSTEM: Trip Units Kn~wledge of the operational implications of the 215004 Source Range Monitor X K5.01 follbwing concepts as they apply to SOURCE R~NGE MONITOR (SRM) SYSTEM: Detector 2.6 12 operation I
215005 APRM / LPRM X 2.1.20 Cohduct of operations: Ability to execute procedure st~ps 4.3 13 I Conduct of Operations: Ability to recognize 217000 RCIC X 2.1.33 indications for system operating parameters which 3.4 14 ar~ entry..level conditions for technical specifications. Knbwledge of AUTOMATIC DEPRESSURIZATION 218000 ADS X K4.02 S~STEM design feature(s) and/or interlocks which prdvide for the following: Allows manual initiation of 3.8 15 AqS logic Errlergency Procedures / Plan Ability to perform 218000 ADS X without reference to procedures those actions that 2.4.49 4.0 16 require immediate operation of system components anp controls. I Kn;owledge of PRIMARY CONTAINMENT 223002 PCIS/Nuclear Steam IS0LATION SYSTEM/NUCLEAR STEAM SUPPLY Supply Shutoff X K4.05 SjUT-OFF design feature(s) and/or interlocks which 2.9 17 pr ,vide for the following: Single failures will not impair the function ability of the system i NUREG-1021 7
ES-4~ Brun~NRC Form ~1-1 Written Examination Outline Plant Systems - Tier 2 Group 1 i System #/Name Number I KIA Topics I Imp. ~I i Knowledge of the physical connections and/or cause- effect relationships between 239002 SRVs X K1.07 3.6 18 RE~IEF/SAFETY VALVES and the following: Su~pression pool Kn6wledge I of REACTOR WATER LEVEL 259002 Reactor Water Level CqNTROL SYSTEM design feature(s) and/or Control X K4.10 3.4 19 int~rlocks which provide for the following: three element I control Knowledge of the physical connections and/or 261000 SGTS X K1.09 ca~se- effect relationships between SGTS and the 3.2 20 follpwing: PCIS 262001 AC Electrical Ability to monitor automatic operations of the A.C. Distribution X A3.03 ELECTRICAL DISTRIBUTION including: Load 3.4 21 sh~dding Knpwledge of UNINTERRUPTABLE POWER SUjPPLY (A.C.lD.C.) design feature(s) and/or 262002 UPS (AC/DC) X K4.01 3.1 22 int¢rlocks which provide for the following: Transfer fro~ preferred power to alternate power supplies 263000 DC Electrical X K2.01 Kn~wledge of electrical power supplies to the 3.1 23 Distribution follpwing: Major D.C. loads Kn~wledge of the physical connections and/or ca~se- effect relationships between EMERGENCY 264000 EDGs X K1.07 3.9 24 GENERATORS (DIESEUJET) and the following: En1ergency core cooling systems Knpwledge of the effect that a loss or malfunction of 300000 Instrument Air X K3.02 th~ Instrument Air System will have on the following: 3.3 25 Sy~tems having pneumatic valves or controls. I Ability to monitor automatic operations of the CCWS 400000 Component Cooling indluding: Setpoints on instrument signal levels for X A3.01 3.0 26 Water normal operations, warnings, and trips that are applicable to the CCWS K/A Category Point Totals: . 3/3 4 3 1 5 1 2 0 2/2 3 2 Group PointiTotal: I I 26/5 NUREG-1021 8
ES-4 Form E Bruns~RC Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KIA Topics 206000 HPCI X 2.4.7 Kndwledge of event based EOP mitigation stra~egies 3.8 91 Abil!ity to (a) predict the impacts of the following on the jSECONDARY CONTAINMENT; and (b) 290001 Secondary CTMT X A2.03 based on those predictions, use procedures to corf,ect, control, or mitigate the consequences of 3.6 92 those abnormal conditions or operations: High are~ radiation 226001 RHRlLPCI Containment X 2.1.9 Co~duct of Operations: Ability to apply tech specs Spray Mode for a system 4.0 93 201002 RMCS X A4.02 Ability to manually operate and/or monitor in the con'troI room: Emergency in/notc~ override switch 3.5 27 201003 Control Rod and Drive Kn~wledge of the operational im plications of the Mechanism X K5.01 foll6wing concepts as they apply to CONTROL 2.6 28 ROID DRIVE AND MECHANISM: Hydraulics Kn$wledge of the effect that a loss or malfunction 201006 RWM X of the following will have on the ROD WORTH K6.01 2.8 29 MI~IMIZER SYSTEM (RWM) (PLANT SPECIFIC)
- RY\'M power supply: P-Spec(Not-BWR6)
Kn6wledge of the effect that a loss or malfunction 202001 Recirculation X of the RECIRCULATION SYSTEM will have on K3.06 3.7 .30 follbwing: Low pressure coolant injection logic: Plaint-Specific Kn6wledge of RECIRCULATION FLOW 202002 Recirculation Flow CONTROL SYSTEM design feature(s) and/or Control X K4.02 3.0 31 intJrlocks which provide for the following: Rebirculation pump speed control: Plant-Specific Kn~wledge of the physical connections and/or 204000 RWCU X cause- effect relationships between REACTOR K5.05 2.6 32 W4TER CLEANUP and the following: Flow Cohtrollers Ability to predict and/or monitor changes in 241000 Reactor/Turbine partameters associated with operating the X A1.14 REIACTORlTURBINE PRESSURE REGULATING 3.4 33 Pressure Regulator SYSTEM controls including: Pressure se~pointlpressure demand Ability to predict and/or monitor changes in 202001 Recirculation X A1.09 p~ameters associated with operating the 3.3 34 R 'CIRCULATION SYSTEM controls including: Repirc Pump Seal Pressures NUREG-1021 9
ES-4~ Bruns~RC Form E Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name A1 A2 A3 A4 Number KIA Topics Q# Ability to manually operate and/or monitor in the 268000 Radwaste X A4.01 I 3.4 I 35 control room: Su.mp integrators Kndwledge of FIRE PROTECTION SYSTEM I des,gn feature(s) and/or interlocks which provide I I 36 286000 Fire Protection I I I I I X I I I I I I I K4.01 for the following: Adequate supply of water for the 3.4 fire )protection. I Abi~ity to monitor automatic operations of the 290001 Secondary CTMT I I I I I I I I I I X I I A3.02 I SE~ONDARY CONTAINMENT including: Normal I 3.5 I 37 buil~ing differential pressure: Plant-Specific I I Kn6wledge of the physical connections and/or I I I I I I I I I I I 214000 Rod Position K1 05 cau,se- effect relationships between ROD X
. POplTION INFORMATION SYSTEM and the I 3.3 I 38 Information System foll9wing: Full core di~play K/A Category Point Totals: I 0/2 I 2 I 0 I 1 I 2 I 1 I 1 I 2 I 0/1 I 1 I 2 I Group Point Totkl:
I I I 12/3 NUREG-1021 10
I ES-401 ' Generic Knowledge and Abilities Outline (Tier3) I Form ES-401-3 I Facility: I Brunswick NRC I Date of Exam: I 3/10/2007 RO SRO-Only Category KIA#. Topic IR Q# IR Q# 2.1.12 Ability to apply technical specifications for a system. 4.0 94 Ability to obtain and interpret station reference 2.1.25 materials such as graphs, monographs, and tables 3.1 95 which contain performance data. 1. Conduct of Operations Ability to explain and apply all system limits and 2.1.32 3.4 66 precautions. 2.1.1 Knowledge of conduct of operations requirements 3.7 67 Subtotal 2 2 Knowledge of the process for conducting tests or
--._--- --2.2-.:]---- --.-experiments.noLdescribedin lhe.safety_analysis._ .-.- ---_... _._ .._---. _____~_ _2__.. _____96 report.
Knowledge of bases in technical specifications for 2.2.25 3.7 97 limiting conditions for operations and safety limits. 2. Knowledge of the process for determining the internal Equipment Control 2.2.34 2.8 68 and external effects on core reactivity. (multi-unit) Knowledge of the design, procedural, and 2.2.3 3.1 69 operational differences between units. Subtotal 2 2 Knowledge of SRO responsibilities for auxiliary 2.3.3 systems that are outside the control room (e.g., waste 2.9 98 disposal and handling systems). Knowledge of radiation exposure limits and 2.3.4 contamination control, including permissible levels in 3.1 99 excess of those authorized. 3. Radiation Control 2.3.2 Knowledge of the facility ALARA program 2.6 70 Knowle.dge of 10 CFR: 20 and related facility radiation 2.3.1 2.6 71 control requirements Subtotal 2 2
- 4. Ability to recognize abnormal indications for system Emergency Procedures 2.4.4 operating parameters which are entry-level conditions 4.3 100
/ Plan for emerQency and abnormal operating procedures.
2.4.27 Knowledge of fire in the plant procedure. 3.0 72 NUREG-1021 11
I ES-401 Generic Knowledge and Abilities Outline (Tier3) I Form ES-401-3 , Knowledge of the parameters and logic used to assess the status of safety functions including:1 Reactivity 2.4.21 control 2. Core cooling and heat removal 3. Reactor 3.7 73 coolant system integrity 4. Containment conditions 5. Radioactivity release control. Knowledge of the bases for prioritizing emergency 2.4.23 procedure implementation during emergency 2.8 74 operations. Knowledge of communications procedures associated 2.4.15 3.0 75 with EOP implementation Subtotal 4 1 Tier 3 Point Total 10 7 NUREG-1021 12
I ES-401 Record of Rejected K/As I Form ES-401-41 Tier / Randomly* Reason for Rejection roup Selected KIA 295027 (0.80) 295028 EA2.01 randomly reselected.--MarkJ11-Containment_only, 1/ 1 does not apply to Brunswick. EA2.02 295005 (0.#42) AK2.01 randomly reselected. Initial selection does not apply to 1/ 1 Brunswick (BWR 2 only) AK2.09 295025 (0.#50) 2.1.23 randomly reselected. Operational valid, discriminating 1/ 1 question could not be written for the original topic select.ion. 2.1.14 295027 (0.# 52) 295028 EK3.01 randomly reselected. Mark III Containment 1/ 1 only, does not apply to Brunswick. EK3.01 295008 (0.#83) AA2.02 randomly reselected. A discriminating que~tion at the 1 / 2-AA2.03 SRO level could not be written 295022 (0.#62) AA1.01 randomly reselected. Initial selection does not apply to 1/ 2 Brunswick AA1.04 215004 (0.89) 2.4.11 randomly reselected. A discriminating question at the 2/1 SRO level could not be written for original topic. 2.4.49 --- - - - . - . _ - ------- 7-000--- -{Q:-#51-2-1-7-000-K--2--:-02--ra-n-cJo-m-ly-reseleet-ed--does*-not-na-ve --a-A-~ sela-tioA-2/1 Condenser (original selection). K2.01 261000 (0.#20) K1.09 randomly reselected. An operational valid question could 2/1 not be written for original selection 2.1.14 300000 (0.#25) K3.02 randomly reselected. A discriminating question could not 2/1 be written for original topic selectron. K2.01 204000 (0.#32) K5.05 randomly selected. A discriminating question could not 2/2 be written for original topic selection. K3.01 215003 (0.#11) K6.05 randomly selected. Insufficient documented technical 2/1 information existed for original topic selection. K6.02 201003 (0.#28) K5.01 randomly selected. Initial selection ~as on previous NRC 2/1 exam. A1.02 256000 (0.#34) 202001 A1.09_randomly selected. A discriminatin'g question 2/2 could not be written for original topic selection A1.09 286000, (0.#36) K4.01 randomly selected. Original selection was at a low level 2/2 of importance and was considered minutia. K2.02 290002 (0.#38) 214000 K1.05 randomly selected. Original selection was at a 2/2 low level of importance and was considered minutia. K1.05 295025 (0.#79) 2.4.6 randomly reselected. The topic originally selected was not 1/ 1 discriminating at the SRO level 2.1.27 295038 (0.#57) EK1.02 randomly selected by the NRC. A discriminating 1/ 1 question at the RO level could not be written for the original selection EK1.01 215004 (Q.#89).205000 randomly selected by the NRC. A discriminating 2/1 question at the SRO level could not be written for the original selection. 2.4.11 295008 (Q.#83) AA2.01 randomly selected by the NRC. A discriminating 1/2 question at the SRO level could not be written for the original selection AA2.02 NUREG-1021 13
Brunswick 2007 NRC Scenario 1 Page 1 of 2 Facility: BRUNSWICK Scenario No.:> 1 Op Test No.: 2007 NRC Examiners: Operators: (SRO) (RO) (BOP) InitialConditions: The plant is operating at 100% power, End of Cycle. No equipment is out of service Turnover: SwapRB Supply & Exhaust Fans from 2C to 2D for maintenance work. Maintenance personnel are standing by. Critical Task: See Scenario Summary Event Malt. No. Event Event Description No. Type* 1 N/A N-SRO Swap RB Supply and Exhaust Fans N-BOP 2 K4522A OFF C-SRO 2C RBCCW sheared shaft, 2B fails to auto start (AOP) C-BOP 3 RC026F C-SRO Runback of 2A Recirc Pump to Limiter #2. (TS) (AOP) C-RO 4 N/A R-SRO Increase power following the runback. R-RO 5 NB014F TS-SRO Instrument Penetration X49A Line Break - Remote Shutdown Instrumentation is lost (TS) 6 CN019F C-SRO. AOG fails to isolate on High H2 signal CN011F I-BOP 7 EEOO9F M-ALL Loss of Offsite Power, DG3 Differential Fault, Reactor Scram (AOPs, DG026F EOPs) 8 ES028F C-SRO HPCI injection valve fails to auto open C-RO' 9 NBOO6F M-ALL Steam Leak in Drywell, Emergency Depressurization (EOPs) 10 K1J36A C-SRO RHR Loop "B", drywell spray valve fails closed C-BOP 11 NB025F I-ALL Level instrument failure, Reference leg flashing, Reactor Flooding required
* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
- 1) 2)
S = Satisfactory; U - Unsatisfactory; N/O = Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* =Critical Task/Step
) ....
Brunswick 2007 NRC Scenario 1 Page 2 of 2
- SCENARIO DESCRIPTION Unit 2 is operating at maximum power, End Of Cycle.
A swap of RB Supply & Exhaust Fans will be' required to support maintenance activities. Following the swap of RB fans, RBCCW Pump 2C coupling fails and 2B RBCCW Pump will fail to Auto-Start on pressure, but will be able to be manually started. After restart of the 2B RBCCW Pump, Reactor Recirculation Pump 2A will runback to Limiter #2. After addressing the Technical Specifications and discussions with I&C, the 2A Recirculation Pump Limiter #2 signal will be reset and power returned to the pre-event level. A Reactor Instrument Penetration line break occurs (X49A) and the line is isolate'd affecting instrument N026S for Remote Shutdown Panel Level Indication R604BX. Technical Specifications must be addressed. After TS are addressed, AOG will receive a high-high H2
- -**--*---*-------sig-A-a-J-eue-te-a-fa*i*led-hydroger:l-detector:-a.nd~wi-l.l . fa_il-to-isolate--r:eq-uiri-rlg-man-uaLjsolation.. by the SOP.
Off-Site Power will be lost. DG4 will auto start and tie to E4. DG3 will auto start and briefly tie to E3, but will then trip on overcurrent and E3 will be unavailable. E1 and E3 cannot be cross-tied due to the overcurrent lockout. If the crew attempts to crosstie E7 and E8 the* cross-tie breaker at E8 will fail. The loss of E3/E7 results in loss of level transmitters N026A and N027A. HPCI and RCIC are available for RPV level control. SRVs are available for pressure control. The HPCI injection valve will fail to auto open but can be manually opened. Additionally, RBCCW cooling will be shifted to conventional service water. 120V Panels 2-AB, 2-AB-RX and 32AB will be transferred to alternate. A steam leak will occur in the drywell. Drywell coolers will trip and the RHR Loop "S" drywell spray valve (E11-F016B) cannot be opened causing drywell temperature to rise above 300°F requiring emergency depressurization (CRITICAL, TASK). Following emergency depressurization reactor pressure and drywell reference leg temperature will be in the 'unsafe region of the RPV saturation limit The only available level instruments (N004A, N004C, N036 and N027B) will begin to exhibit indications of reference leg flashing. With no valid indication of RPV level, the crew will enter the Reactor Flooding Procedure. The crew will increase available injection to maximum until at least 5 SRVs are open and Reactor pressure is at least 50 psig above suppression chamber pressure (Minimum Reactor Flooding Pressure) (CRITICAL TASK). Once these conditions are established the crew will throttle flow to maintain at least the required 50 psig differential but as low as possible. When RPV flooding conditions have been established, the scenario may be terminated .
- NOTES: 1) 2)
. S = Satisfactory;
* =Critical Task/Step U - Unsatisfactory; N/O = Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
Brunswick 2007 NRC Scenario 2 Page 1 of 3 Facility: BRUNSWICK Scenario No.: 2 Op Test No.: 2007 NRC Examiners: Operators: (SRO) (RO) (S'OP) Initial Conditions: The plant is operating at maximum power, Middle of Cycle. Turnover: Several LPRMs have failed and have been bypassed. CRD Pump 28 is under clearance to replace oil in the speeq changer and will be out of service for four hours.
- -------*--*----11----- - - - - - - - - - - - -~SW-P-l:Hllp-2-G-wa-s-tJ*Ader--Glea*r-a-J1Ge-ang-is-r-eady-to-be-pla.Ged--ir-l--ser-vice ...-
Once it has been placed in service. CSW pump 2A is to be removed from service. No other equipment is out of service. Reduce power to 90% for upcoming turbine valve testing .
- Critical Task:
Event No. Malf. No. See Scenario Summary Event Type* Event Description 1 N/A R-SRO Power reduction with recirc flow from 1Ooo~ to 90% R-RO 2 N1048M, 28- I-SRO An LPRM fails low and must be bypassed this will make APRM # 1 1'30 inoperative for only one LPRM operable for Level B. I-RO 3 N/A N-SRO Swap CSW pumps N-BOP 4 K4B39A Auto C-SRO After CSW Pump 2A is removed from service CSW pump 2C trips on Off overcurrent and CSW Pump 2A fails to auto start. (AOP) C-BOP CW027F 5 ES015F TS-SRO HPCI logic power failure (TS) 6 CW015F C-ALL CW intake screens progressively foul resulting in high screen ,dips and eventual total loss of CW, lowering vacuum (AOP)
- NOTES: 1) 2)
=
S Satisfactory; U - Unsatisfactory; All Unsatisfactory ratings require comments; a comment sheet is attached.
* =Critical Task/Step
=
N/O Not Observed
Brunswick 2007 NRC Scenario 2 Page 2 of 3 7 RP011 F M-ALL Group 1 isolation, ATWS, One SLC pump fails (EOPs) SL_IASLRB C-RO BKR OFF 8 RW016F C-SRO RWCU outboard isolation valve fails to automatically isolate on SLC initiation. C-RO 9 RI_IARIUNCP C-SRO Failure of RCIC turbine coupling C-BOP 10 K2213A C-SRO Failure of SDV vent and drain valves to OPEN following the scram reset. Isolate 11 K1507A Open C-SRO One ADS SRV fails to OPEN during Emergency depressurization (EOPs), Off Restoration of vessel level C-BOP
* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
- NOTES: 1) 2)
S =Satisfactory; U - Unsatisfacto*ry;* N/O = Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* =Critical Task/Step
Brunswick 2007 NRC Scenario 2 Page 3 of 3 SCENARIO DESCRIPTION BRUNSWICK 2007 NRC Scenario #2 The plant is operating at 100% power, Middle of Cycle with 2C CWIP and CRD Pump 2B under clearance. 2C CSW Pump has jiust been placed in standby following maintenance and needs to be placed in service. A downpower is required for upcoming turbine valve testing. While power is being reduced an LPRM will fail downscale causing a Technical Specification tracking LCO and inoperability of the #1 APRM on LPRM level inputs for the B level. The operators will start CSW pump 2C following maintenance work. Following the pump start, CSW Pump 2A will be removed from service and placed in standby. The 2C CSW pump will then trip on overcurrent and the 2A CSW pump will fail to auto start (it will start manually). The crew will respond per AOP-19.0. Once the CSW pump issue is addressed, a HPCI logic power ________________ failure will occur requiring a technical specification entry.
~----------=----=-------~-----=------~------- ----------------------------
The Circulating Water Intake Pumps (CWIPs) traveling screens will progressively plug with river silt resulting in CWIP pump trips. Condenser vacuum will initially slowly lower. The crew will enter AOP-37.1, Intake Structure Blockage. The RO will lower reactor power and the BOP operator will attempt to recover CWIPs. Eventually all CWIPs will trip and condenser vacuum will be lost causing a Group 1 isolation. If the Reactor had not been scrammed, a scram will occur. Most control rods will fail to insert on the scram. The crew will respond per 2-EOP LPC. When SLC is initiated, SLC pump 2B will trip on overcurrent. Additionally RWCU will fail to automatically isolate. When RPS is reset, the SDV vent and drain valves will fail to open. Initiating SLC and/or inserting control rods per LEP-02 is a Critical Task RPV level will be deliberately lowered to suppress power. 'When RCle initiates on LL2, the RCle Turbine coupling will break. Without' HPCI or RCIC, RPV level will drop below LL4 requiring Emergency Depressurization (Critical Task). The crew must terminate'and prevent injection prior to ED per 2EOP-01-LPC (Critical Task). When ADS valves are opened ADS Valve C fails to open. Low pressure ECCS must be overridden off to prevent uncontrolled injection during depressurization. When pressure drops below MARFP, injection may be recommenced to restore RPV level above LL4 (Critical Task). Condensate should be used for injection due to inability to throttle RHR flow for 5 minutes. When the rods are all inserted and/or hot shutdown boron weight has been injected and level is being restored to 170-200", the scenario may be terminated .
- NOTES: 1) 2)
S = Satisfactory; U - Unsatisfactory; N/O = Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* = Critical Task/Step
Brunswick 2007 NRC Scenario 3 Page 1 of 3 Facility: BRUNSWICK Scenario No.: 3 Op Test No.: 2007 NRC Examiners: Operators: (SRO) (RO) (BOP) Initial Conditions: The plant is operating at 94% power, End Of Cycle. RHR SW Pump 20 is under clearance for motor replacement and will remain out of service for two days. TBCCW Pump 2B is under clearance to investigate a high vibration. TBCCW Pump 2C has been placed in service on Unit 2. No other equipment 'is out of service. Turnover: --'-Swap Servlc,e-WateFPDmps-f6r m'aintenanceworR o-fi-lfiEfopeFafingJjUmp~- Raise power to 100% Critical Task: See Scenario Summary Event Malf. No. Event Event Description No. Type*
- 1 2
N/A N/A N-SRO N-BOP R-SRO R-RO Swap NSW pumps Power increase to 100% for rod pattern adjustment 3 MRC021F C-SRO Recirc Pump "A" scoop tube lockup C-RO 4 ZUA2162 ON TS-SRO EDG ,low starting air pressure (TS) 5 CW019F (A) C-SRO NSW pump trip(AOP) and standby pump fails to auto start K4821 A-Auto C-BOP Off 6 ES27F C-SRO RCIC Mechanical Overspeed Trip C-RO 7 K4403A Open C-ALL Partial Loss of FW heating, Power reduction required (AOPs) 30 sec 8 NBOO5F M-ALL Fuel Failure, Hi MSL Rads, MSIVs closed, Manual & Auto Scram Fail, RPOO5F (EOP)(AOP) Initiates ARI (CT) K2503A-AS IS 9 ES004F C-SRO SRV F sticks open C-BOP
- NOTES: 1) 2)
S = Satisfactory; U - Unsatisfactory; N/O = Not Observed All Unsatisfactory ratings. require comments; a comment sheet is attached.
* = Critical Task/Step
Brunswick 2007 NRC Scenario 3 Page 2 of 3
- 10 11 K1230A-AS IS RSIARHBYPB-Bypass CW071F (8)
C-SRO C-80P C-SRO RHR Loop A SW HX outlet valve fails, F0688 valve (RHR HX Service Water Outlet) will fails to auto close. RHR SW 28 pump trip, RHR leak into service water. (CT) CW013F C-80P 12 CA020F M-ALL SRV F tailpipe break, ED required (CT)
* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
------------_._-~----------------~_._-----_._---_ ...
- NOTES: 1) 2)
=
S Satisfactory; U - Unsatisfactory; = N/O Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* =Critical Task/Step
Brunswick 2007 NRC Scenario 3 Page 3 of 3 SCENARIO DESCRIPTION BRUNSWICK 2007 NRC Scenario #3 The plant is operating at 87% power, End Of Cycle with RHR SW Pump 20 and T8CCW Pump 28 under clearance. A swap of NSW pumps is required for upcoming maintenance on the operating pump. After swapping NSW pumps, reactor power will be raised to 100°/0. While power is being raised a scoop tube lockup will occur on the "A" Recirc MG Set. I&C will report a circuit breaker caused the problem and the operator can reset the scoop tube. Once the scoop tube has been res~t and recirc flows are matched, the #3 EDG will have a low starting air pressure requiring a technical' specification determination (TS). (The EDG #3 must be declared inoperable). Following the TS determination for the EDG (3.8.1.0), the NSW pump previously started will trip, requiring a restart of the NSW pump originally removed from service (TS 3.7.2.8). Once the NSW pump is restarted', a RCle overspeed trip will occur due to a field operator accidentally unlatching th~ mechanism (T.S.3.5.3.A). The RO will respond and re-Iatch the trip mechanism. Feedwater valve FW-V120 will partially open resulting in a loss of feedwater heating and rising reactor power. The crew will respond per A(JP--=-O-3~Oancrreauce reactor power. 1*t1eJ=W=\tr2u valve carrt5Ef' .' .N .** _ *. _ **_ ** __ .**. _ ****.**** _ manually closed by the operators. Fuel failure will occur that causes SJAE readings to rise and MSL Rad Hi to alarm. The crew will respond by entering AOP-05.0 and OEOP-04-RRCP. Power will be reduced to clear the MSL Rad Hi alarm. The fuel failure will get worse resulting in MSL Hi-Hi alarm along with rising Main Stack readings and alarms. Per the guidance of OEOP-04-RRCP, the crew will insert a manual reactor scram and close the Group 1 Isolation Valves ***(Critical task to Close the MSIVs and Drains). The manual scram switch for channel 8 will fail. The reactor can be scrammed by Mode switch or ARI initiation (Critical Task). When the MSIVs are closed SRVs will be required for pressure control. When SRV F is opened, it will stick open. Suppression pool temperature will rise requiring initiation of suppression pool cooling per OEOP-02-PCCP. If RHR Loop "A" is started for suppression pool cooling, the E11-F068A valve (RHR HX Service Water Outlet) will fail to open and RHR Loop "A" will be unavailable for suppression pool cooling. When RHR Loop 8 is started for suppression pool cooling, the RHR Heat Exchanger will develop a tube leak. The tube leak will initially result in leakage of service water into the RHR system and'RHR high conductivity alarm. RHR SW 800ster Pump 28 will then trip (RHR SW 20 is under clearance) and E11-F0688 will fail to auto close. Without an RHR Service Water 'pump in operation, RHR system water will now leak into service water. Service Water high radiation will alarm. The crew will respond to the service water release per EOP RRC.P by closing E11-F0688, shutting down RHR Loop 8 and isolating the heat exchanger (Critical Task). The F SRV tailpipe will fail and Emergency Depressurization will be required per OEOP-02 pecp when the safe region of Pressure Suppression Pressure (PSP) can not be maintained (Critical Task). When the reactor is depressurized by the Emergency Depressurization, the scenario may be terminated.
- NOTES: 1) 2)
S =Satisfactory; U - Unsatisfactory; N/O =Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* =Critical Task/Step
.~
Brunswick 2007 NRC Scenario 4 Page 1 of 3 Facility: BRUNSWICK Scenario No.: 4 Op Test No.: 2007 NRC Examiners: Operators: Initial Conditions: A plant startup is in progress. Reactor Power is approximately 4°ib GP-2 has been completed with the exception of steps 5.3.55, 5.3.60, 5.3.61, and 5.3.62 The "8" SJAE is in Full Load; the "A" SJAE is shut down Reactor Pressure is being held at 800 psig to support EHC electrical testing. Power increase by control rod withdrawal has been authorized to provide additional bypass valve opening. The Nuclear Engineer has been contacted and continuous rod withdrawal may be used., Turnover: Continue plant startup lAW GP-2 at step 5.3.55 Place the SJAEs in half load. Currently in 20P-30. Continue in GP-10 at Sequence A2X, step 24, Item 251 and raise power, using control rods, to achieve one bypass valve open. Do not exceed 8 % power. Critical Task: See Scenario Summary Event Malf. No. Event Event Description No. Type* 1 N/A N-SRO Place SJAEs in half load. N-BOP 2 N/A R-SRO Increase reactoF power R-RO 3 N1019F I-SRO IRM "C" Fails "Downscale" I-RO 4 Overrides TS-SRO HCU Alarm (TS) 5 C-SRO "A" SPE Fan Trips C-BOP 6 NI024F, I/C-SRO Seismic Event, spurious start of EDG, IRM upscale (TS) (AOP) Overrides I-RO C-BOP
- NOTES: 1) 2)
S = Satisfactory; U - Unsatisfactory; = N/O Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
* ='Critical Task/Step
Brunswick 2007 NRC Scenario 4 Page 2 of 3 7 NI018F M-ALL Seismic aftershock, ATWS, RWCU Unisolable Leak (EOP) (AOP) RP011F RW013F RW016F EE030M Overrides 8 CF035F C-SRO FW injection Valve FW-V120 or Startup Level Control Valve fails closed (dependent on which method of feed is chosen by the operator) C-BOP
* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
- NOTES: 1) 2)
=
S Satisfactory; U - Unsatisfactory; N/O = Not Observed All Unsatisfactory ratings require' comments; a comment sheet is attached.
* =Critical Task/Step
Brunswick 2007 NRC Scenario 4 Page 3 of 3 SCENARIO DESCRIPTION BRUNSWICK 2007 NRC Scenario #4 The plant is at approximately 4% power and a startup is in progress lAW GP-2. The SJAEs must be placed in half load. As control rods are withdrawn, IRM "C" will fails Downsale The operator must recogize that the IRM "C" is failed. When diagnosed, the SRO will consult TS (3.3.1-RPS) and declare IRM "C" inoperable and direct the RO to bypass IRM "C". Once the startup continues, Scram Accumulator 34-19 will alarm on low pressure. The crew will dispatch an AO to investigate/charge the HCU. The crew will reference Tech Spec 3.1.5.C.2. and declare the control rod accumulator inoperable if it cannot be recharged within 1 hour. After TS are addressed for the acc'umulator the "B" Steam Packing Exhauster will trip requiring _~t~~rti ng_of t.h~ "A" Stear1!_Eacking _~xha u~_~~~ .__ ._~_.~. ... __~_._ ...__.__. .... _ ._. .. _ A seismic event will occur. This will cause the spurious start of the EDG 3 and a high alarm trip of IRM F. A half scram will not occur and this must be diagnosed by the crew. AOP-13 will be entered and the EDG can be secured. lAW Te.ch Specs (3.3.1), a half scram must be inserted due to the IRM high alarm trip and it will be successful.- I&C will also be notified to pull fuses lAW 01-18.
- A seismic aftershock will then occur causing IRM A to trip on high alarm but RPS will fail and a manual scram must be inserted. Numerous control rods will fail to insert and 2-EOP-1-LPC must entered for the ATWS Control Rods must be inserted manually and/or SLC injected (Critical Task) to shut down the reactor. Additionally, a le.ak will develop in the RWCU system, RWCU will fail to isolate on the SLC initiation and power will be lost to MCC 2XC.
Attempts to isolate the RWCU leak will be ongoing but will fail. The crew will enter OEOP SCCP as temperatures and sump levels begin to rise in the Reactor Building., When room flood levels in two areas of the Rx Building exceed max safe levels, Emergency Depressurization will pe required (Critical Task). The crew must terminate and prevent injection prior to ED per 2EOP-01-LPC (Critical Task). NOTE: In the event the crew inserts control rods and enters 2EOP-01-RVCP prior to reflood, Terminate and Prevent would not be required and would not be a critical task. ", After the ED, a failure will occur on either the RFPT S/U level control valve or the 2FW-V120 (High Pressure Feedwater Heaters bypass valve) to restore and control vessel level following emergency depressurization. The failure will occur on whatever method is first used by the operator to feed the vessel. The scenario can be ~erminated once emergency depressurization occurs and reactor water level has recovered and is stable in the normal band
- NOTES: 1)
- 2) *
=
S Satisfactory;
=Critical Task/Step U - Unsatisfactory; =
N/O Not Observed All Unsatisfactory ratings require comments; a comment sheet is attached.
Final Submittal (Blue Paper) FINAL OUTLINES
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Brunswick Date of Examination: JULY 2007 Examination Level (circle one): RO ISRO Operating Test Number: NRC 2007 Administrative Topic Type Describe activity to be performed (see Note) Code* D Determine Primary Containment Water Level and Conduct of Operations Evaluate PCPL-A. N Perform a portion of Control Operator Daily Conduct of Operations Surveillance Report 201-03.2 and identify 4 OOS readings and appropriate TS entries. M Generate a Clearance for maintenance activities on Equipment Control the 2C TBCCW Pump. D Determine Off-Site Release Per PEP-03.4.7 and Complete Radiation Control Notification Form. N Evaluate plant conditions (includes security event) and Emergency Plan classify the event. Make PAR determination as required.
- NUREG-1021, Revision 9
ES-301 Administrative Topics Outline Form ES-301-1 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required .
.*Type Codes & Criteria: (C)ontrol room /'-;.,..,
(D)irect from bank (::;; 3 for ROs; (~_~)or SROs & RO retakes) (N)ew or (M)odified from bank (>1)1.- (P)revious 2 exams (:s; 1; randomly selected)ty (S)imulator . 2007 NRC Examination Summary Description of Admin Tasks A.1.a The candidate will determine Primary Containment Water Level and Evaluate PCPL-A This is a bank JPM. A.1.b The candidate will review a portion of the Control Operator Daily Surveillance Report 201-03.2 and identify 4 OOS readings and appropriate TS entries. This is a new JPM. A.2 The candidate will generate a clearance for maintenance activities on the 2C TBCCW. This is a modified JPM requiring sequence and dual unit power supply tagging requirements. A.3 The candidate will determine the offsite release rate and fill out appropriate forms. This is a bank JPM. A.4 The candidate will evaluate degraded plant conditions which include a security event and make an event classification and PAR as required. This is a new JPM.
- NUREG-1 021, Revision 9
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Brunswick Date of Examination: 2007 Examination Level (circle one): RO I SRO Operating Test Number: NRC 2007 Administrative Topic Type Describe activity to be performed (see Note) Code* D Hand Calculation Of APRM GAFs Per PT-01.8C Conduct of Operations N Perform a portion of Control Operator Daily Conduct of Operations Surveillance Report 201-03.2 and identify 4 OOS readings. M Generate a Clearance for maintenance activities on Equipment Control the 2C TBCCW pump. N Determine Stay Time and Radiological requirements Radiation Control for performing work in a High Radiation Area. N/A Emergency Plan
- NUREG-1021, Revision 9
ES-301 Administrative Topics Outline Form ES-301-1 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room i (D)irect from bank (8,for ROs; :s; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (> 1)2 (P)revious 2 exams (~ 1; randomly selected)p (S)imulator 2007 NRC Examination Summary Description of Admin Tasks A.1.a This is a bank JPM. The candidate will manually calculate APRM GAFs. A.1.b This is a new JPM. The candidate will perform a portion of Control Operator Daily Surveillance Report 201-03.2 and identify 4 OOS readings. A.2 The candidate will generate a clearance for maintenance activities on the 2C TBCCW. This is a modified JPM requiring sequence and dual unit power supply tagging requirements. A.3 This is a new JPM. The candidate will be required to determine stay time and radiological requirements for performing work in a High Radiation Area NUREG-1021, Revision 9}}