ML081190547

Examination Outline Submittal and Comments for the Fermi Initial Examination - January 2008

ML081190547
Person / Time
Site:	Fermi
Issue date:	03/25/2008
From:	Hironori Peterson Operations Branch III
To:	Jennifer Davis Detroit Edison
Shared Package
ML080860174	List: ML081190547 ML081190699 ML081190701 ML081190704 ML081190707 ML081200422 ML081200435 ML081200448 ML081200455
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50-341/08-301
Download: ML081190547 (36)

Text

EXAMINATION OUTLINE SUBMITTAL AND COMMENTS FOR THE FERMI INITIAL EXAMINATION -JANUARY 2008 DTE Energy October 16,2007 NRC-07-0054 Mr. Hironori Peterson Chief, Operations Branch Division of Reactor Safety U. S. Nuclear Regulatory Commission Region III Suite 210 2443 Warrenville Road Lisle, Illinois 60532-4352

Dear Mr. Peterson:

Enclosed please find the proposed examination outline submitted to the NRC in preparation for the upcoming Fermi 2 Initial License Examination scheduled during the week of January 28,2008: . . . . . . . . . . . . . Examination Outline Quality Checklist (Form ES-201-2)

Photocopies of Examination Security Agreements (Form ES-201-3) RO Administrative Topics Outline(s) (Form ES-301-1)

SRO Administrative Topics Outline(s) (Form ES-301-1)

RO Control Room / In-Plant Systems Outline (Form ES-301-2)

SRO Upgrade Control Room

/ In-Plant Systems Outline (Form ES-301-2)

SRO Instant Control Room/In-Plant Systems Outline (Form ES-301-2) Transient and Event Checklist (Form ES-301-5) Scenario Outlines (Form ES-D-I) RO BWR Examination Outline (Form ES-401-1)

SRO BWR Examination Outline (Form ES-401-1)

Generic WA Outline (Form ES-401-3)

Record of Rejected WAS (Form ES-401-4)

Hironori Peterson Chief, Operations Branch NRC-07-0054 October 16,2007 Page 2 The examination outline was developed using the appropriate guidance contained in Revision 9, NUREG 1021, Supplement 1. These materials shall be withheld from public disclosure until after the examinations are complete.

We look forward to working with you and your examination team during the examination development and administration process.

If you have any questions or comments regarding the contents of the items listed above, please contact Mr. Timothy P. Horan. General Supervisor, Operations Training at (734) 586-4961.

Sincerely, Enclosure cc: Chief, Reactor Operations Branch NRC Resident Office Document Control Desk Washington D.

C.

ES-201 Examination Outline Quality Chacklist F~rm ES-201-2 Fadmy: Date of Dramination:

of mal evolutions.

instwmenl and component failures.

teCmicBl specibtions.

and quantiIativ8 cMecia spe4tied on Form ES-3014 and dewbed in Appendix D. , ES-201, Page 26 of 28 ES-301 Administrative ToDics Outline Form ES-301-1 Iacility:

Fermi 2 Ixamination Level: RO SRO 0 Date of Examination:

01/28/2006 Operating Test Number:

2008-1 Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Type Code" Describe activity to be performed Perform Thermal Limit Verification (MAPRAT) 293009 Core Thermal Limits GENERIC 2.1.7 -Ability to evaluate plant performance and make operational judgments based on operating characteristics

/ reactor behavior I and instrument interpretation. RO 4.4 I SRO 4.7 Complete and Communicate a Nuclear Plant Technical Data Form (Alert) GENERIC 2.4.39 - Knowledge of the ROs responsibilities in emergency plan implementation. RO 3.91SRO 3.8 Perform 24.202.03, "HPCI System Piping Filled And Valve Position Verification" 206000 High Pressure Coolant Injection System GENERIC 2.2.12 - Knowledge of surveillance procedures.

RO 3.7 I SRO 4.1 Determine Dose Limit Will Be Exceeded and Initiate a Dose Extension.

GENERIC 2.3.4 - Knowledge of radiation exposure limits and contamination control

/ including permissible levels in excess of those authorized. RO 3.21SRO 3.7 4OTE: 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, (S)imulator, or Class(R)oom (D)irect from bank (5 3 for ROs; 5 4 for SROs & RO retakes) (N)ew or (M)odified from bank (> 1) (P)revious 2 exams (5 1; randomly selected)

ES 301, Page 22 of 27 ES-301 Administrative Tooics Outline Form ES-301-1 -acility:

Fermi 2 Ixamination Level: RO 0 SRO [XI Administrative Topic (See Note) Code' Conduct of Operations Conduct of Operations C,R,S.P Equipment Control SD Radiation Control Emergency Plan Date of Examination:

1/28/2008 Operating Test Number: 2008-1 Describe activity to be performed Perform Thermal Limit Verification (MAPRAT) 293009 Core Thermal Limits GENERIC 2.1.7 -Ability to evaluate plant performance and make operational judgments based on operating characteristics

/ reactor behavior

/ and instrument interpretation.

RO 4.4 I SRO 4.7 Knowledge of Shift Staffing Requirements GENERIC 2.1.4 - Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1OCFR55, etc. RO 3.31SRO 3.8 Perform 24.202.03, "HPCI System Piping Filled And Valve Position Verification" 206000 High Pressure Coolant Injection System GENERIC 2.2.12 - Knowledge of surveillance procedures.

RO 3.7 SRO 4.1 Determine Dose Limit and Complete a Dose Extension GENERIC 2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

RO 3.21SRO 3.7 Perform On-Site Protective Actions and Classification for Security Event (Alert) GENERIC 2.4.40 - Knowledge of the SROs responsibilities in emergency plan implementation RO 2.7 SRO 4.5 rlOTE: 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, (S)imulator, or Class(R)oom (D)irect from bank (5 3 for ROs; 54 for SROs & RO retakes) (N)ew or (M)odified from bank (z 1) (P)revious 2 exams (5 1; randomly selected)

ES 301, Page 22 of 27 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Fermi 2 Exam Level:

RO [XI SRO-l 0 SRO-U 0 Date of Examination:

1/28/2008 Operating Test Number: 2008-1 Control Room Systems' (8 for RO); (7 for SRO-I);

(2 or 3 for SRO-U, including 1 ESF) System / JPM Title a. Shift Operating CRD Pumps 201001 Control Rod Drive Hydraulic System A4. Ability to manually operate andlor monitor in the control room: A4.01 CRD DUmDS RO 3.1 I SRO 3.1 b. Manually Start the RClC System 217000 Reactor Core lsolatlon Cooling System A4. Ability to manually operate andlor monitor in the control room: A4.04 Manually initiated controls RO 3.6 I SRO 3.6 c. Manually Initiate Low Low Set (No Fault) 239002 RellefISafety Valves A4. Ability to manually operate and/or monitor in the control room: A4.01 SRVs RO 4.4 I SRO 4.4 d. SOP Run of Core Spray System in Test Mode 209001 Low Pressure Core Spray System A2. Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

AZ.06 Inadequate system flow RO 3.2 I SRO 3.2 e. Restore a RWCU Pump After Oil Sample (Plant Hot) With System Leak 223002 Primary Containment Isolation System /Nuclear Steam Supply Shut-Off A2. Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

and Failure to Isolate Type Code' DS D.S P,EN,D,A,S N.S A2.03 System logic failures RO 3.0 I SRO 3.3 f. Respond to Multiple Rod Drifts and RPS Failure DAS 212000 Reactor Protection System A2. Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

A2.16 Chanaina mode switch Dosition RO 4.0 I SRO 4.1 Safety Function 1 2 3 4 5 7

g. Respond to Refuel Floor High Radiation PS?A,L,S 9 272000 Radiation Monitoring System 'Type Codes -. A2. Ability to (a) predict the impacts of the following on the RADIATION MONITORING SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

A2.12 Refuel floor handling accidentsloperations RO 3.3 I SRO 4.0 Criteria for RO / SRO-I I SRO-U h. Manually Initiate Div 1 Emergency Equipment Cooling Water 400000 Component Cooling Water Systems (CCWS)

A4. Ability to manually operate and I or monitor in the control room: A4.01 CCW indications and control .RO 3.1 I SRO 3.0 (C)ontrol room (D)irect from bank (EN)gineered safety feature (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams WCA (S)imulator Implant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) - <91 (0 154 - >lI '1 1'1 I ' 1 (control room system) - >lI )1 121 - >21 '2 1'1 - < 3 I( 3 I 5 2 (randomly selected) - >lI '1 1'1 i. Shin In Service IAS Dryers 300000: Instrument Air System A2. Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

A2.01 Air dryer and filter malfunctions RO 2.9 I SRO 2.8 j. Defeat ARI Logic Trips 216000: Nuclear Boiler Instrumentation Generic 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.

RO 4.2 I SRO 4.1 k. Place ESF Battery Charger in Service 263000: DC Electrical Distribution Generic 2.1.30 Ability to locate and operate components, including local controls.

RO 4.4 I SRO 4.0 KI. Knowledge of the physical connections andlor cause effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following:

K1.02 Battery charger and battery RO 3.2 I SRO 3.3 0 D.R 0 7 6 ES-301, Page 23 of 27 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility:

Fermi-2 Exam Level:

RO 0 SRO-I SRO-U 0 Date of Examination:

01/28/2008 Operating Test Number: 2008-1 Control Room Systems" (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title

a. Manually Start the RClC System 217000 Reactor Core Isolation Cooling System A4. Ability to manually operate and/or monitor in the control room: A4.04 Manuallv initiated controls RO 3.6 I SRO 3.6 b. Manually Initiate Low Low Set (No Fault) 239002 ReliefISafety Valves A4. Ability to manually operate andlor monitor in the control room:

A4.01 SRVs RO 4.4 I SRO 4.4 c. SOP Run of Core Spray System in Test Mode 209001 Low Pressure Core Spray System AZ. Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

AZ.06 Inadequate system flow RO 3.2 I SRO 3.2 d. Restore a RWCU Pump After Oil Sample (Plant Hot) With System Leak 223002 Primary Containment Isolation System /Nuclear Steam Supply S h ut-Off AZ. Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.03 Svstem logic failures RO 3.0 I SRO 3.3 and Failure to Isolate e. Respond to Multiple Rod Drifts and RPS Failure 212000 Reactor Protection System A2. Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

A2.16 Chanaina mode switch Dosition RO 4.0 I SRO 4.1 Type Code' D.S D,S P,EN,D,A,S N.S Safety Function 2 5 7

f. Respond to Refuel Floor High Radiation 272000 Radiation Monitoring System A2. Ability to (a) predict the impacts of the following on the RADIATION MONITORING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.12 Refuel floor handling accidents/operations RO 3.3 I SRO 4.0 g. Manually Initiate Div 1 Emergency Equipment Cooling Water 400000 Component Cooling Water Systems (CCWS) A4. Ability to manually operate and I or monitor in the control room:

A4.01 CCW indications and control .RO 3.1 I SRO 3.0 h. In-Plant Systems" (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Shift In Service IAS Dryers 300000: instrument Air System A2. Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

A2.01 Air drver and filter malfunctions RO 2.9 I SRO 2.8 j. Defeat ARI Logic Trips 216000: Nuclear Boiler Instrumentation Generic 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.

RO 4.2 I SRO 4.1 k. Place ESF Battery Charger in Service 263000: DC Electrical Distribution Generic 2.1.30 Ability to locate and operate components, including local controls.

RO 4.4 I SRO 4.0 K1. Knowledge of the physical connections and/or cause effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following:

K1.02 Battery charger and battery RO 3.2 I SRO 3.3 D.R 9 8 8 7 6 @ 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 I Criteria for RO I SRO-I I SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (RFA 4-6 I 4-6 I 2 <9/ 58 154 - >11 ?I 1>1 -I - / > 1 (control room system) - >1/ 21 1>1 - >21 22 1>1 - < 3 15 3 I 5 2 (randomly selected) - >lI >1 />1 ES-301, Page 23 of 27 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility:

Fermi-2 Exam Level:

RO 0 SRO-I 0 SRO-U Date of Examination: 01/28/2008 Operating Test Number: 2008-1 Control Room Systems' (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title a. Temporary Removal and Restoration of SDC for I&C Surveillances 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) A4. Ability to manually operate andlor monitor in the A4.01 SDClRHR pumps RO 3.7 / SRO 3.7 A4.03 SDClRHR discharge valves RO 3.6 / SRO 3.5 A4.05 Minimum flow valves RO 3.2 I SRO 3.2 control room: b. Respond to Refuel Floor High Radiation 272000 Radiation Monitorlng System A2. Ability to (a) predict the impacts of the following on the RADIATION MONITORING SYSTEM

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

A2.12 Refuel floor handling accidentdoperations RO 3.3 I SRO 4.0 c. Manually Start the RClC System 217000 Reactor Core Isolation Cooling System A4. Ability to manually operate and/or monitor in the control room:

A4.04 Manually initiated controls RO 3.6 I SRO 3.6 Implant Systems" (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Place ESF Battery Charger in Service 263000: DC Electrical Distribution Generic 2.1.30 Ability to locate and operate components, including local controls. RO 4.4 I SRO 4.0 KI. Knowledge of the physical connections and/or cause effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following:

KI.02 Battery charger and battery RO 3.2 I SRO 3.3 j. Defeat ARI Logic Trips 216000: Nuclear Boiler Instrumentation Generic 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system imDlications.

RO 4.2 I SRO 4.1 Type Code' N,EN,A,S D.S D.R Safety Function 4 9 2 6 7 @ 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.

I Type Codes I Criteria for RO I SRO-I I SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (RFA jS)imulator ES-301, Page 23 of 27 4-6 I 4-6 I 2 e91 (a 154 - >I/ 21 />I I 2 1 (control room system) - >I1 21 />I - >2/ 22 121 - 3 15 3 I 5 2 (randomly selected) - >I1 >I 121 ES-401 BWR Examination Outline Form ES-401-I iacility: Fermi 2 Date of Exam: I /28/2008 qote: 1. 2. 3. 4. 5. 6. 7.' 8. 9. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not he less than two). 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 il 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. Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401.

for guidance regarding elimination of inappropriate WA statements. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall he selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

Select SRO topics for Tiers I and 2 from the shaded systems and WA categories. The generic (G) WAS in Tiers I and 2 shall be selected from Section 2 of the WA Catalog, but the topics must he relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable WA's On the following pages, enter the WA 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. If fuel handling equipment is sampled in other than Category A2 or G' on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note # I does not apply). Use duplicate pages for RO and SRO-only exams. For Tier 3. select tooics from Section 2 of the WA Cataloe. and enter the KIA numbers. descriotions. IRs. and ,, point totals (#)on Fbrm ES-401-3. Limit SRO selections;o WAS that are linked to IOCFR55.i3 ES-401 2 Form ES-401-1 I ES-401 I BWR Examination Outline Form ES401-1 Emergency and Abnormal Plant Evolutions -Tier I/Group 1 (RO / SRO) I E/APE # / Name I Safety Function I K 2WOOJ Partiid or 'Tutal lms oCIX I Pwrlb I ?9501 b Control Room Abnndoninent I 295025 Iligli Rcuctur Pressure 13 29?030 Low Suppression Pnnl Water L.crcl is 295038 High OW-sttc Rclcax Kate I 9 I 295001 Panial or Complete Loss of Forced Core Flow Circulation I 1 & 4 295003 Panial or Complete Loss of I ACl6 295004 Panial or Total Loss of E* Pwr 16 I I 295005 Main Turbine Generator Trip I l3 295006 SCRAM I I 295016 Control Room Abandonment I 295018PartialorTotal LassofCCW

/ 295019 Panial or Total Loss of Inst. Air I8 295021 Loss ofShutdown Cooling1 4 I following responses as they apply 10 PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION

Reduced Imp COMPLETE LOSS OF A.C. POWER : Bane I power i AK2 07 - Knowledae of the interrelations following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Increasing cooling water flow to heat PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR

Instrument air system ressure between LOSS OF SHUTDOWN COOLING and the fo owing Reactor rec rculation ES-401 2 Form ES-401-1 ES401 BWR Examination Outline Form ES401-1 Emergency and Abnormal Plant Evolutions -Tier 11Group 1 (RO I SRO) EIAPE #I Name I Safety Function 295023 Refueling Accidents I8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure 13 295026 Suppression Pool High Water Temp. I5 29503 I Reactor Low Water Level / 2 295028 High Drywell Temperature

/ 5 295030 Low Suppression Pool Water Level I5 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown I I 295038 High Off-site Release Rate 19 600000 Plant Fire On-site I8 700000 Generator Voltage and Electric 3id Disturbances WA Category Totals: impiications of the following concepts as D ELECTRIC - IR - 3.6 - 3.9 - 3.6 - 3.5 - 4.2 - 3.8 - 3.4 - 3.9 - 3.8 - 2.8 - 3.8 -

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401- Emergency and Abnormal Plant Evolutions - Tier 1IGroup 2 (RO I SRO) EIAPE #I Name I Safety Function KK 12 Z95008 I ligh Reaclar Wslrr Lcvrl 12 295012 High Ikywell 'Tcmpcraturc I5 295033 High Secondary COW .iinmeni Area Iladiation Lcvcis 19 295007 High Reactor Pressure 13 X 295010 High Drywell Pressure IS 295012 High Drywell Temperature IS 295032 High Secondary Containment Area Temperature I5 295033 High Secondary Containment Area Radiation Levels 19 X 295036 Secondary Containment High SumplArea Water Level I5 KIA Category Totals: 111' KIA Topic@)

the Following as they apply > f I(?>

ES-401 4 Form ES-401-1 is401 System # / Name 223002 ITlSRjuclear Stcain $upply SllUtOtT 21 8000 ADS 215003 IRM 215005 APRM/I.I'KM 203000 RHRLPCI: Injection Uode 205000 Shutdown Cooling 206000 HPCI 209001 LPCS 21 1000 SLC 212000 RPS H 4 BWR Examination Outline Plant Systems -Tier Z/Group 1 (RO / SRO) U 5 Form ES-401-1 K/A Topic@) the cmseqecnccs of those ohnormal conditions or opcrations:

ADS initiation POWFR RANGII MONI'IORII.OCAI.

POWER RANGF MONITOR SYSTI'M ; and (h) biwd on ~hosc. predictions.

~ise procedur~.~

10 correct. conlid, or rnitigitlc use plant cmnputers to cvaluatc system or connections andlar cause- effect relationships between RHRLPCI: INJECTION MODE (PLANT SPECIFIC) and the following: Condensate storage and chanaes in Darametem associated with operitmg tie SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including predictions.

use procedures to correct. control or mitiaate the conseauences of those abnorkal cond t ons or operations A C fat ares BWR-2 3.4 1 K3 03 - Knowledae of the eflen tnat a 1 loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: Emergency

\

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Plant Systems - Tier 2/Group 1 (RO / SRO) System # / Name 215003 IRM 2 I5003 IRM 215004 Source Range Monitor 215005 APRM / LPRM 217000 RClC 218000ADS 223002 PClSMuclear Steam Supply Shutoff 239002 SRVs 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 261000 SGTS < I KIA Topic@)

II 2 4 31 Knowledge of annunciator X alarms, indications, or response I I procedures I K2 01 - Knowledae of electrical Dower supplies to the foilowing: IRM channelddetectors RANGE MONITOR (SRM) SYSTEM design feature@) andlor interlocks which provide for the following: Rod withdrawal blocks implications of the following MnCepts as they apply to AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM LPRM I I detector location and core symmetry I 2 2 44 -Equipment Control Ability lo interpret wntrol room indications to verify the status and operation of a I XI svstem and understand how oDerator ahions and directives affect plant and system conditions I K6 04 - Knowledae of the effect that a I loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM

Air andlor monitor in the control room:

loss or malfLnction of tne IOIIOW ng w have on the RELIEFSAFETY VALVES I I : Discharge line vacuum breaker I K5 04 - Knowledae of the ODerational mplications of tne fol owing wncepts as they apply lo RELlEFlSAFETY VALVES Tail pipe temperature montlor n of the following on the REACTOR ana (b) Dased on those preoicbons use proceoures to correct. control or m mate the consea.ences of tnose WATER LEVEL CONTROL SYSTEM abnormal conditions or operations loss or malfunction ofthe STANDBY Form ES-401-1 IR # 4.2 7 2.5 8 3.7 9 2.9 IO 4.2 I I 3.6 12 3.6 13 3.0 14 3.3 I5 3.2 16 3.0 17 3.2 18 ES-401 4 Form ES-401-1 I ES401 Plant Systems 1 Tier ZGroup 1 (RO / SRO) I IR 1234 91 KIA Topic@) Major breakers and control power I 2 4 35 - Emergency Procedures I Plan Knowledge of local auxiliaty operator 90 I 'O I I I I I tasks during emergency and the I I I I I resutantoperatioialetiects I K4 06 -Know edge of EMERGENCY GENERATORS (DIESEUJET) design feature@) andlor interlocks which 26 I IIIII Drovide for the followina Governor I I I I I bontrol - I K4 03 - Knowledge of (INSTRUMENT AIR SYSTEM) desian featureb) and or curing of IAS upon loss of I I I I SYSTEM ml,d.ng Air temperature I I K2 01 - Know edge of e enrica power I _I ,, I,,, I I I I supplies tothe following CCW pumps I 26226 Group Poinl Total ES-401 5 Form ES-401-1 I System # / Name I 271000 ow-ga5 201002 RMCS 204000 RWCU 215002 REiM 234000 Fuel Handling Equipment 201001 Control Rod Drive 245000 Main Turbine Gen. I Aux. I 256000 Reactor Condensate 271000 Off-gas I BWR Examination Outline Form ES-40 Plant Systems -Tier 2/Group 2 (RO / SRO) 561234 Kl Kl A I A/ KIA Topic@)

ES-401 5 Form ES-401-1 ES-401 I BWR Examination Outline Plant Systems -Tier ZGroup 2 (RO / SRO) 1 Form ES-401 ES-401 Generic Knowledge and Abilities Outline (Tier3) Form ES-401-3

'acility:

Fermi-2 Category :onduct if Operations

quipment

ontrol Ladiation

ontrol Date of Exam: 1/28/2008 RO Topic IR Ability to use procedures to determine the effects on reactivity of plant changes.

such as KCS temperature.

secondary plant, fuel depletion, etc.

2. I .43 2.1.41 2.1.5 Knowledge of the refueling process.

Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. hbtotal Knowledge ofthe process for controlling equipment Ability to determine Technical Specification Mode of 3.0 Knowledge of the process used to track inoperable alarms. Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. 2.2.43 3.9 2.2.15 2.2.40 Abilitv to amlv technical specifications for a system. 3.4 , , . .,,, , , ,, ,, ,.>,I ,, j u b t o t a I 2.3.5 , ., , .,. . Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Knowledze of radiation exposure limits under normal . , - 1.3.4 1 or cmergency conditions. Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. Knowledge of radiation monitoring systems, such as 2.3.15 fixed radiation monitors and alarms, portable survey 2.9 instruments, personnel monitoring equipment, etc. 3.4 2'3'13 ,.. ., ,,, ,,.. Subtotal SRO-Onlv 4.3 I 94 -t 2 29 97 3 7 98 2 ES-401 Generic Knowledge and Abilities Outline (Tier3) Form ES-401-3 7 Category 1. Emergency

'rocedures

'Ian - 2.4.20 t 2.4.23 I 2.4.43 Topic I IY Knowledgc of opcrational implications of EOP ~orniiigr.

caJiioiis.

and iiol~>. I -1.. . . Kiior Idg: uI'iIic basus for prioritizing cmqenc) proccdureimplementation during emergency operations.

Knowledge of EOP entry conditions and immediate action steps. Knowledge of emergency communications systems and techniques.

4,6 73 Knowledge of "fire in the plant" procedures.

3.4 74 3,2 75 Cate gory 4.3 __ 4.4 __

Appendix D Scenario Outline FOITII ES-D-1 NIA 831 RFOOl5 E51MF0009 Facility:

Fermi 2 Scenario No. 1 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaqi Initial Conditions:

IC-18, MOL, 75% Rx. Power Turnover: The plant has been operatinq for 103 davs. Reactor Power is currently 75% of Rated Thermal Power with Control Rods at the 86% Rod Line followino rod pattern adiustment.

Seneral Service Water Pump

1. 4 is out of service for motor replacement with an expected return

o service in 2 davs. This shin will start the East Heater Feed Pump and raise Reactor Power to 35% of rated with Recirculation flow. Reactor Enoineerina will be readv in one hour to pull rods for another rod pattern adiustment. GOP 22.000.03, "Plant Operation 25%

to 100% to 25%" [Rev 73) actions for power increase are complete throunh Step 4.2.17.1.

NOTE: The crew's Pre-job Briefing for the reactor power increase is to be conducted prior to entering the simulator. (SI gested time 30 minutes prior to beginning the scenario.)

R (ATC) R (SRO) C (ATC) C (SRO) C (BOP) C (SRO) - Event No. 1. 2. 3. - - 4. - Malf. No. I Event I Event Description Start the Third HFP using SOP 23.107, "Reactor Feedwater and Condensate Svstems".

Section 5.5. Increase Reactor Power With Recirculation Flow per GOP 22.000.03,"Power Operation 25% To 100% To 25%". RRMG "B Walkaway Uncontrolled RRMG "B Speed Change (>IO%). crew trips the affected RRMG (Immediate Action per AOP 20.138.03, "Uncontrolled Recirc Flow Change"). Crew enters AOP 20.138.01, "Recirc Pump Trip", Condition C & D. NOTE: OPRMs are operable. CRS directs increased core monitoring for instability. He also directs increasing speed on the operating RRMG to raise core flow (>43%) and exit the ScramlExit Region of the Power to Flow Map.

CRS reviews TS 3.4.1.A, Recirc Loops Operating, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to declare loop inop and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to adjust RPS trip setpoints to single loop values. Spurious RClC Initiation Spurious start of RCIC, BOP verifies no valid actuation signal and trips RClC when directed. CRS reviews TS 3.5.3.A, RClC System, (Immediately verify HPCl operable and 14 day LCO). * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D, 38 of 39 Appendix D Scenario Outline Fo~ ES-D-1 Malf. No. R11 MFOOOl N20MF0023 E41 MF0009 E41 MF0005 N21MF0031 N21 RF0019 B31MF0066 age 1 Event I Event Type* I Description M (All) C (BOP) C (SRO) Facility:

Fermi 2 Scenario No. 1 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaai Initial Conditions:

IC-18, MOL, 75% Rx. Power.

NOTE: Continued from Event No. 5. - 6. 7. 8a. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Loss of Div 2 Offsite Power

/ EDGs start / Loss of all Heater Feed Pumps, and Reactor Feed Pumps. Loss of all Div 2 Buses.

EDGs 13 & 14 auto start and re- energize ESF buses only. CRS enters AOP 20.300.345kV, Mode Switch in SHUTDOWN, and performs AOP 20.000.21, "Reactor Scram". CRS enters EOP 29.100.01, "RPV Control", Sheet 1 (Level 3). HPCl Auto Start Failure HPCl fails to start on Level 2. The crew will identify and manually start HPCl using SOP 23.202, "HPCI System", Encl C. (Hard Card), but HPCl will isolate after about 1 min of oDeration. - C (BOP) c (SRO) SBFW Fool Fails As Is BOP will start SBFW Pump A, for level control, and identify a Fool failure to open. When he depresses the valve's Open pushbutton the valve will lose power. The BOP will direct an operator to investigate and restore. NOTE: SBFW Pump A is the only available pump. M (All) Recirc Loop A Rupture LOCA -A Recirculation leak will cause High Drywell Pressure and level to decrease.

EOP 29.100.01, RPV Control, Sheet 1 re-entry on high Drywell Pressure and EOP Primary Containment Control, Sheet 2 entry on High Drywell Pressure.

The crew will start all available high pressure injection systems. (SLC and re-starts RCIC by resetting the Trip Throttle Valve due to spurious initiation still in 23.206, RCIC, Sect 7.1 or Encl B Hard Card) At -100 RPV Water Level BOP reports water level cannot be maintained

=. TAF. I CRS briefs crew for Emergency Depressurization.

Appendix D, 38 of 39 Appendix D Scenario Outline Fo~ ES-D-1 Facility:

Fermi 2 Scenario No. 1 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Pilaaai Initial Conditions:

IC-18, MOL, 75% Rx. Power. NOTE: Continued fron - - Event No. - - 8b.. - 9. - 10. Malf. No NIA EOPRF0038 NIA age 2 Event Type' M (ALL) I (ATC) I (SRO) N (ATC) N (SRO) Event Description AT TAF Emergency Depressurizes (EOP C-2) EOP 29.100.01, RFIEDISC, Sheet 3 (CT). BOP opens 5 SRVs ADS preferred. Crew restores water level 173 - 214 with available High Pressure and Low Pressure ECCS Svstems. (CT) - Div 2 EECW Hi Drywell Pressure Lead Lifted. ATC verifies EECW actuation and isolation to the Drywell.

Determines Div 2 EECW is not isolated, and isolates by closina the P4400-F606B.

He then restores coolina to CRD. Crew sprays the Torus using SOP 23.205, "RHR System", Encl A. (Hard Card) Crew sprays the Drywell using SOP 23.205, "RHR System", Encl A. (Hard Card) (CT) * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D, 38 of 39 Awendix D Scenario Outline Form ES-D-1 M (ALL) c (ATC) C (SRO) R (ATC) R (SRO) Facility:

Fermi 2 Scenario No. 2 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaai Initial Conditions:

IC-20, MOL, 100% Rx. Power Turnover:

The plant has been operatina for 23 davs. Reactor Power is 100% of Rated East Condenser Air Leak CRS enters AOP 20.125.01, "Loss of Condenser Vacuum". BOP starts an additional SJAE and OG Ring Water Pump. ATC attempts Rapid Power Reduction, per SOP 23.623, Reactor Manual Control System, Section 9.7, and determines failure of Manual Runback. ATC takes individual manual control to reduce recirc speeds to 55- 60% core flow. Thermal Power. The South RBCCW Pump is out of service for motor replacement.

It is scheduled to be restored tomorrow. Plans are to shift from Division 1 CCHVAC to Division 2 4. to collect routine vibration data on Division 2 CCHVAC eauipment.

NOTE: The Pre-iob Briefing for the CCHVAC shift is to be conducted Drior to enterino the N61 MF0003 sirnulato;. (sug Event I Malf. No.

No. I NIA 1. C11 MF0445 C97MF1087 MF EBAORL TCTVSP 1 I * (N)ormal, (R)eactir P - IsTed time 30 minutes prior to beginning the scenario.) Event Event Type* Description I__ N (BOP) N (SRO) CCHVAC Crew shifts from Division 1 CCHVAC to Division 2 Appendix D, 38 of 39 Appendix D Scenario Outline Form ES-D-1 I Facility:

Fermi 2 Scenario No. 2 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore 6. Palaai Initial Conditions:

IC-20, MOL, 100% Rx. Power. NOTE: Continued from page

1. Event I Malf. No, No. C97MF1087 B21MF0059 821 MF0060 621 MF0073 621MF0037 I Event M (ALL) Type' M (ALL) I (ALL) C (BOP) C (SRO) ALL Event Description Reactor Scram Performs override action 20.125.01 and places Mode Switch in SHUTDOWN 5 2.5 psia. After Scram Reports, CRS enters EOP 29.100.01 RPV Control, Sheet 1 (Level 3) and directs entry into AOP 20.000.21, Reactor Scram.

BOP controls water level 173 - 214 inches. Aftershock Seismic Event

/ Event Trouble Alarm (6D69)

Loss of all level indication.

Div 1 and Div 2 Level Instrument Reference Leg Ruptures. Flood up Level Indication fails upscale high. (Level 8 trip on Main Turbine, RFPs, HPCI, RCIC, and SBFW.) CRS enters (EOP C-4) EOP RPV Flooding, 29.100.01, Sheet 3. (Level cannot be determined)

CRS enters EOP 29.100.01, Primary Containment Control, Sheet

2. (High Drywell Pressure) When BOP is directed to open 5 SRVs, SRV "R fails to open. BOP will select and open another SRV and report this to the CRS. (CT) The crew floods to the Main Steam Lines using Feedwater, SBFW, and Low Pressure ECCS Systems. (CT) (N)ormal. (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D, 38 of 39 Appendix D Scenario Outline FOWII ES-D-1 Facility:

Fermi 2 Scenario No. 3 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaai Initial Conditions:

IC-19, EOL. 100% Rx. Power Turnover:

The plant has been operating for 403 davs. Reactor Power is currentlv 100% of Rated Thermal Power with Control Rods at the 86% Rod Line followina rod pattern adiustment.

The N. Turbine Lube Oil Vapor Extractor is out of service for motor replacement.

APRM #I is bypassed due to a power SUDD~V failure and a trackina LCO is written for it. The plan for the shift is to remove the North TBCCW Pump from service for lubrication and outboard motor bearing replacement. The shift is also to perform 27.109.01, "Turbine Steam Valves Test. The crew will beain with the LPSV and LPCV portion of the test because I&C support will not be available to perform the on- line valve position calibration on #I LPlV later in the shift.

NOTE: The crew's Pre-job Briefing for the reactor power decrease and "Turbine Steam Valves Test" is to be conducted prior to entering the simulator. (Suggested time 30 minutes - - Event No. - - 1. 2. - - 3. 4. - xior to beainnina the scenario.) - Malf. No. NIA NIA N30MF0036 C51MF0001 C51MF0002

• (N)ormal. (R)eact - Event N (BOP N (SRO R (ATC: R (SRO Type' - C (BOP C (SRO I (ATC) I (SRO) Event Description Crew shifts TBCCW Pumps due to scheduled maintenance on North TBCCW Pump. Lower reactor power to 93% with Recirculation Flow per GOP 22.000.03, "Power Operation 25% to 100% to 25%", in preparation to perform 27.109.01, Turbine Steam Valves Test. Crew performs section 5.7 of 27.109.01.

Tests LPSV #I and closes LPlV #I for on-line valve position calibration.

When I&C reports complete BOP determines, #I LPlV fails to re-open. Crew stops test, informs Ops Management, and System Engineer of problem. #2 APRM fails upscale high. When directed, ATC removes APRM #I from bypass and bypasses #2 APRM per 23.605, "Average Power Range Monitoring (APRM) System", Sect 6.4. (APRM #1 downscale and rod block in alarm.) SROreviewsTS3.3.1.1.F(Mode2-6 hrs). :v. (l)nstrument, (C)omponent. (M)aior Appendix D, 38 of 39 Aooendix D Scenario Outline FOIIII ES-D-1 Facility:

Fermi 2 Scenario No. 3 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaai Initial Conditions:

IC-19, EOL. 100% Rx. Power.

NOTE: Continued from - Event No. - 5. - 6. - 7. - 8. 9 7 Malf. No. 821 MF0029 PO P603 - B087-1 N30MF0044 C71MF0006 CIIMF0001 C41MF0003 C41MF0004 NIA age 1 Event C (BOP) C (SRO) Type* C (ATC) C (SRO) M (ALL) C (ATC) C (SRO) C (ATC) C (SRO) M (ALL) Event Description SRV G fails open. BOP takes immediate actions per AOP 20.000.25, "Failed Safety Relief Valve (SRV)", by depressing the open and closed pushbuttons repeatedly.

The SRV will close. The CRS reviews TS 3.4.3.A, SRVs (tracking LCO), TS 3.6.1.6.A, Low - Low Set Valves (14 day LCO), and TS 3.6.1.8, Suppression Chamber-to-Drywell Vacuum Breakers (SR 3.6.1 B.2, 12 hrs to perform the vacuum breaker test after a SRV discharge to Torus). May direct removing fuses to prevent re-occurance per AOP 20.000.25, Enclosure A. CRD Flow Control Valve drifts close. ATC observes and reports. SRO enters AOP 20.106.03, "CRD Flow Control Valve Failure", Condition A. ATC takes manual control, opens, and adjusts CRD Pressure.

Turbine Trips I Failure to Scram (ATWS)

Failure to scram is reported and the crew enters the EOPs on Scram Condition with power > 3%. CRS directs actions from 29.100.01 Sheet IA, RPV Control-ATWS. ATC performs FSQ 1-8 actions (unsuccessful) and BOP performs inhibit ADS, bypasses Drywell Pneumatics, and directs 29.ESP.11.

~~ Initial SLC Pump selected trips. When ATC is ordered to inject SLC. He informs CRS and starts second SLC Pump. SLC is successfully injected. (CT) BOP lowers water level and maintains 0 - 50 wr (EOP C-5, Level I Power Control)

ATC performs manual rod insertion (CT).

When 29.ESP.10 (Defeat of ARI Logic Trips) is complete and ARI Trip Logic is reset, the scram discharge volume will drain (3D94 Clear).

The ATC will perform the Scram-Reset-Scram section of 29.ESP.03, "Alternate Control Rod Insertion", which will insert all rods.

• (N)ormal. (R)eactivity, (I)nstrument, (C)omponent. (M)ajor - Appendix D, 38 of 39 Appendix D Scenario Outline Form ES-D-1 Facility:

Fermi 2 Scenario No. 4 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaai Initial Conditions:

IC-20, MOL, 100% Rx. Power Turnover:

The plant has been operatina at 100% power for 400 davs. Reactor Power is currentlv 100% of Rated Thermal Power with Control Rods at the 86% Rod Line followina rod pattern adjustment.

The plan for the shift is to shift RBCCW Pumps for scheduled maintenanc on the center pump.

NOTE: The crew's Pre-job Briefing for the reactor power increase is to be conducted prior to entering the simulator. (Suggested time 30 minutes prior to beginning the scenario.)

3_ Event Malf. No. NO. I 1. N/A 2. 821 MF0044 N21 MF0029 3. 4. C51 MFOI 98 E41MF0007 EOPRF0022 E41MF0001 E41MF0008 5a. Event Event Type' Description - N (BOP) N (SRO) Section 6.1 Shift RBCCW Pumps in accordance with SOP 23.127, C (ATC) c (SRO) Jet Pump 5/6 Failure ATC observes Recirc Loop Flow changes and diagnoses Jet Pump failure.

The crew enters AOP 20.138.02, "Jet Pump Failure".

The SRO evaluates TS 3.3.1.1.C (1 hr LCO), TS 3.3.1.1.F(6hrstoMode2),TS3.4.1.A(2 hrLCO),andTS 3.4.2.A (12 hrs to Mode 3). The CRS declares JP

1. 5 inoperable and directs the ATC to monitor for thermal hydraulic instabilities.

Spurious N. RFP Trip / RRS Runback / Loss of Heater Crew enters AOP 20.107.01, "Loss of Feedwater or Feedwater Control".

The crew verifies RR runs back, starts SBFW and injects at 1200 gpm, and inserts the Cram Array to lower reactor power to 5 65%. (-15% reactivity change) R (ATC) R (SRO) Drains C (BOP) C (SRO) I (ATC) I (SRO) RBM B Fails High The crew bypasses RBM B per 23.607. "Rod Block Monitoring System", Section 5.1. The CRS declares RBM B inoperable and enters TS 3.3.2.1.A (24 hr LCO). C (BOP) C (SRO) E4150-F600 Thermal Overload HPCl Steam Leak & E4150-F002 Failure to Auto Isolate I Fire Alarm in the HPCl Quad, the crew may enter AOP 20.000.22, "Plant Fires". When 3D34, SEC CONTM TEMP HIGH - HIGH EOP ENTRY alarms the crew will enter EOP 29.100.01, SC/RR, Sheet 5 due to HPCl Area temperature greater than Max Normal Operating (MNO) 148 OF.. * (N)ormal, (R)eactivity, (1)nstrument. (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Appendix D, 38 of 39 Facility:

Fermi 2 Scenario No. 4 Op-Test No: 2008-1 Examiners:

M. Bielbv Operators:

C. Moore B. Palaqi Initial Conditions:

NOTE: Continued from page 1 - - Event No. - 5b. 6. 7. - 8. 9 NIA I M (ALL) B21MF0009 B21MF0054 M (ALL) NIA Event Description NOTE: 1D66, STEAM LEAK DETECTION AMBIENT TEMP HIGH alarms at 2 154 OF in HPCl area and an isolation signal is generated.

When BOP attempts to isolate and E4150-F002 will not close (failed as is) and when E4150-F600 valve close pushbutton is depressed or an isolation signal is received, the valve loses power. The leak cannot be isolated. Reactor Manual Scram (Mode Switch to Shutdown) Before the HPCl Area temperature reaches the Max Safe Operating Temperature (MSO) of 210 OF, the crew briefs and places the Mode Switch in Shutdown. (CT)

When Scram Reports are complete the crew enters EOP 29.100.01, RPV Control, Sheet 1 (Level

3) and AOP 20.000.21, Reactor Scram. RB Steam Tunnel Leak I MSlVs Fail to Isolate Crew observes increasing temperature in RB Steam Tunnel on IPCS. Crew should attempt to isolate the steam leak, by closing the MSIVs, when the area temperature is 2 160 OF (MNO), to isolate all systems discharging into the area. IF area temp is 2 200 OF and the MSlVs are not closed, then the crew should close them due to the auto isolation failure. NOTE: MSL C MSlVs do not fully close.

When Area Temperatures are

> MSO (210 OF) in 2 areas the EOP directs Emergency Depressurization, EOP 29.1 00.01 RPV Flood/ED, Sheet 3 (EOP Contingency 2). CRS directs BOP to open 5 SRVs (ADS preferred) (CT) and while the plant is depressurizing the crew briefs water level restoration and control.

The crew restores and maintains water level 173 - 214" (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D, 38 of 39 Outline Review 10/30/07 I COMMENT I RESPONSElRESOLUTlON I WRITTEN EXAM: No comments.

ADMIN JPMs: Suggestions:

I -On the RO and SRO ES 301-1 forms, assign I align I common letter designators to the admin JPMs. SYSTEM JPMs: Suggestions:

I- On the RO and SRO-I:-U ES 301-2 forms. alian similar .I JPMs to the same 'letter designatof -Pick the 5 SRO-U from the SRO-I JPMs (created 1 new JPM). 1. How did you select the JPMs from the previous 2 exams? 2. The Emergency or Abnormal inplant JPM KAs have to come from the 295- KAs vice the Generic 2.4.- section (SF 7, 216000. Defeat ARI Logic Trips). 3. Delete the reference to Generic 2.1.30 from the system JPM. DC Electrical Distribution, 263000. on ES-301-2.

I SCENARIOS:

1. No "low power" scenario (ref ES 301, D.5.c) I SCENARlOl:

1. Event 3, where should the RRMG walk-away occur to ensure entry into the scramlexit region?

2. Event 4, are there enough verifiable actions for an evaluation of the BOP. or does he only depress the trip pushbutton?

Need to look at taking more actions to achieve a better evaluation.

3. Event 7, if there is no success path for aligning SBFW A, how do we hold the BOP accountable if he decides not to investigate the SBFW A FOO1 failure, or does not identify it? Licensee followed their process for randomly selecting exam material (submitted copy with outline).

I Licensee will look to see if they can apply one of the 295- KAs to the existing JPM. Agreed. ES 301, D.5.c says "should." QA sheets, ES-301- 4 and -201-2 do not require a low power scenario.

Have used low power scenarios on previous exams and minimum number of events that can be used (don't want to be oredictable).

If RRMG speed is 57%. will enter the prohibited region. Will coordinate with examiner directing events and practice during validation.

BOP should runback RCIC, then trip the pump. Will look at providing more actions (ie. injection valve not dosing...).

Will include manual startup of SLC andlor RCIC as part of the evaluation success path to achieve adeauate RWL. Page 1 of 3 Outline Review 10130107 COMMENT RESPONSE 5. Event 9, what are consequences of failing to isolate P44- F606B? Suppose ATC fails, or decides not to perform the action, how do we hold him accountable?

6 Event 10, spraying the Torus and DW is not a "N" evolution It is an action directed by the EOPs when specific plant conditions require the action SCENARIO 2: 1. Event 2, why not have the rod driff out vice into, the core?

What are the ATC verifiable actions? 2. Event 3, are there enough verifiable actions for an evaluation of the BOP? This is not a "C" because there are no actions to take in response to the fan trip. 3. Event 4: -not a "M." -not a 'C" because starting SJAE and OG Ring Water pump does not provide a success path, still lose vacuum. -sufficient verifiable BOP actions for an evaluation?

How do we hold the BOP accountable if he decides not to start SJAE or OG RWP, or is unsuccessful? -why is ATC given credit for

'C? Only "R" for SRO and ATC. -may give 'C" for failure of the RMCS? 4. Event 6, this is "M-ALL" not "I-ALL," cannot restore level instrumentation. onlv take EOP directed actions for a flood. SCENARIO 3: 1. Event 3, not a "C." BOP cannot take any verifiable actions, cannot restore operability, can only stop test and make notifications.

2. Event 7: -what is the "C" and subsequent verifiable actions? What is the success path (manually insert rods)? -can't pet a

'C" and 'M" for the same event. 3. Event 8, what are the verifiable ATC actions for an evaluation (ie, what is critieria for being unsat)?

SCENARIO 4 Failure to isolate Div 2 EECW could result in an inner system LOCA due to heating of water in DW that causes piping to rupture. Isolation is required by EOP for PCP Sheet 2. Delete Event 10 as "N." The selected rod is full out, so will select a partially inserted rod, and allow it to drifl out. The operator will be required to fully insert rod. Power will exceed required power limit. Will delete event as "C." but will leave in for a TS call. Need to break up event and clarify who is getting what credit. The "C" for ATClSRO is for failure of the RRP Master Controller and taking individual manual control of RRPs (not starting the SJAE and OG RWP). The 'R" is for the Rapid Power Reduction. Delete the "I-ALL." Delete "C." Clarify that "C" is failure of the normal scram function, and manual insertion of control rods is the success path.

Clarify the success criteria for starting the second SLC pump, ie. before BIT exceeds 110 degrees F. Page 2 of 3 COMMENT 4. Event 7, good chance BOPISRO will not identify "failure of MSlVs to auto-isolate" if they manually close valves before exceeding 200 degrees F (would not get credit for "C." Will leave in, but not depend on RESPONSElRESOLUTlON Page 3 of 3 1. Event 2, what audible cues does the ATC receive? Suppose ATC doesn't identify the failure (BOP or SRO may unless we pull them back somehow).

2. Event 3, SRO cannot get both "R" and "c" for same event. 3. Event 5a and b, not a "C," because can't isolate HPCl steam leak. Don't have to take any action and get same result. Only provides a reason to scram.

High RWL on P-603. Need to verify during validation.

SRO will only get 'R." Delete "C."