ML072681202

Draft - Submittal of 2007 License Exam Outlines (Folder 2)

ML072681202
Person / Time
Site:	Ginna
Issue date:	07/17/2007
From:	Geckle M Constellation Energy Group
To:	Caruso J Division of Reactor Safety I
Sykes, Marvin D.
Shared Package
ML063560007	List: ML071310329 ML072340101 ML072400127 ML072640224 ML072681173 ML072681198 ML072681202 ML072690061 ML072690085 ML072690103 ML072690105 ML072690112 ML072690120 ML072690132 ML072690139 ML072690147 ML072690158 ML072780402
References
50-244/07-301 50-244/07-301
Download: ML072681202 (25)

Text

N 9,

June 8,2007 Mr. John Caruso Division of Reactor Safety U S . Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415

Subject:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 2007 License Exam Outline Submittal

Dear Mr. Caruso,

Enclosed please find the outline for the Ginna 2007 License Exam. The Ginna Training Department in conjunction with Western Technical Services constructed the outline utilizing the guidance of NUREG-1021. Rev. 9 and NUREG-1 122, Rev. 2. In accordance with 10CFR55.49 and NUREG-102 1 section ES-201 Attachment 1 these materials shall be withheld from public disclosure until after the examinations are completed.

If you have any questions or comments, please contact the General Supervisor, Operations Training, James Reid at (585) 771-5415 or via e-mail james.rcid ~!)cconstellatioii.coin.

Sincerely, /?

Mark Geckle Manager, Nuclear Training

R.E. Ginna 2007 NRC Initial License Written Examination Written Examination Outline Methodology The written examination outline was developed using a proprietary electronic random outline generator developed by Western Technical Services, Inc.

The software was designed to provide a written examination outline in accordance with the criteria contained in NUREG 1021, Revision 9.

The application was developed using Visual Basic code, relying on a true random function based on the PC system clock. The random generator selects topics in a Microsoft Access Database containing Revision 2 of the PWR K&A catalogue. The selected data is then written to a separate data table. The process for selection of topics is similar to the guidance in ES-401, Attachment 1.

The attached outline report and plant specific suppression profile report are written directly from the data tables created by the software. Electronic copies of the data tables are on file.

The process used to develop the outlines is as follows:

For Tier 1 and Tier 2 generic items, only the items required to be included in accordance with ES-401, Attachment 2 are included in the generation process.

Outline is generated for all topics with KA importance 22.5.

25 SRO topics are randomly selected from Tier 1 AA2 and required generic items, Tier 2 A2 and required generic items, (including all System 034 topics) and Tier 3 generic items (All with ties to 10CFR55.43). 75 RO topics are randomly selected to complete the outline, 100 topics total.

The exam report generated lists the topic (Question) number in the far right column. RO topics are numbered 1-75, and SRO topics are numbered 76-100.

The SRO topics are written in red ink for ease of identification.

Items that are rejected after the initial generation process are placed on the rejected items page. Replacement topics are either randomly generated or are inserted upon direction of the Chief Examiner.

Disposition of any item randomly selected but not included in the outline is documented and included.

2007 R.E. Ginna Initial License Examination Outline Submittal PSA Considerations The scenarios developed for the 2007 Ginna NRC license examination were constructed in consideration of the Ginna Plant Specific Analysis. (PSA)

Each scenario considered one or more of the following 3 factors:

1. Contribution to CDF by sequence type

2. Contribution to CDF by initiator

3. Contribution to CDF by system Component or instrument failures were chosen based upon the importance of the system to CDF. Service Water, Auxiliary Feedwater, and Electrical Distribution failures throughout the scenarios will all increase the likelihood of core damage in accordance with the Ginna PSA. Instrument failures affecting the operation of control systems such as Pressurizer Pressure Control are also part of important event sequences.

Major transients were developed based upon either sequence type or initiator.

Each of the Major events developed for the Ginna scenarios were selected for their importance either as an initiator (Loss of Feedwater, ATWS), or by sequence type (LOCA, Blackout)

Additionally, important operator actions for transfer to Cold Leg Recirculation are included in the examination For the JPM examination, JPMs were selected to evaluate important operator actions contributing to CDF:

1. Actions for restoration of CCW to RHR

2. Actions to depressurize the RCS following a SGTR In addition to PSA considerations, plant events resulting in 3 separate LERs have been incorporated into the examination.

ES-401 PWR Examination Outline Form ES-401-2

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Facility: Ginna Date of Exam: 8/27/2007 of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the 'Tier Totals" in each WA cateoorv shall not be less than two).

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 21 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. Systems/evolutions 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 ES-401, Attachment 2, for guidance regarding elimination of inappropriate W A statements.

4. 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.

5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.' The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the toDics must be relevant to the aoolicable evolution or svstem.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) 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 #1 does not apply). Use duplicate pages for RO and SRO-onlv exams.

9. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10CFR55.43 NUREG-I021 1

ES-401 Ginna Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 Ability to determine or interpret the following as they 008 I Small Break LOCA I 3 X EA2.14 apply to a small break LOCA: Actions to be taken if 4.4 76 PTS limits are violated

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022 I Loss of Reactor Coolant Makeup 12 Emergency Procedures I Plan Knowledge of 2,4,1 3.6 77 Abnormal Condition procedures.

027 I Pressurizer Pressure Control System Malfunction 029 I Anticipated Transient Without Scram (ATWS) I 1

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036 I Steam Generator Tube Rupture / 3 E12 I Uncontrolled Depressurizationof all Steam Generators 1 4 procedures and operation within the limitations in 008 I Pressurizer Vapor Space Accident I 3

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015 117 / Reactor Coolant Pump Malfunctions1 4 X 022 I Loss of Reactor Coolant Makeup 12 X I AA1.09 I ~ ~~

they apply'to the Loss of Reactor Coolant P i &

Makeup: RCP seal flows, temperatures, pressures, and vibrations Knowledge of the reasons for the following 025 I Loss of Residual Heat Removal System 1 4 X AK3.01 responses as they apply to the Loss of Residual 3.1 42 Heat Removal System: Shift to alternate flowpath 026 / Loss of Component Cooling Water I 8 2,1 ,30 Conduct of Operations: Ability to locate and operate components, including local controls. 3.9 43 027 I Pressudzer Pressure Control System Malfunction Knowledge of the interrelations between the 13 AK2.03 Pressurizer Pressure Control Malfunctions and the 2.6 44 following: Controllers and posilioners NUREG-1021

ES-401 Ginna Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 Conduct of Operations: Knowledge of system status 029 I Anticipated Transient Without Scram (ATWS) I 1 X 2.1.14 criteria which require the notification of plant 2.5 oersonnel.

038 I Steam Generator Tube Rupture I 3 4.5 040 I Steam Line Rupture 1 4 4.1 054 I Loss of Main Feedwater14 4.1

- 7 055 I Station Blackout 16 3.3 057 I Loss of Vital AC Electrical Instrument Bus 16 3.8 058 I Loss d DC Power I 6 4.0 062 I Loss of Nuclear Service. Water 14 2.6 7

065 I Loss Of Instrument Air1 8 2.9 E04 I LOCA Outside Containment1 3 3.5 NUREG-I021 3

ES-401 Ginna Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following:

Components, and functions of control and safety E05 / Loss of Secondaly Heat Sink / 4 EW.1 55 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Ability to operate and / or monitor the following as they apply to the (Loss of Emergency Coolant E l 1 I Loss of Emergency Coolant Recirculation1 4 EA1.3 56 Recirculation) Desired operating results during abnormal and emergency situations.

KIA Categoly Point Totals: 313 3 3 3 3 3/3 Group PointTotal:

NUREG-1021 4

ES-401 Ginna Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 032 I Loss of Source Range Nuclear Instrumentation 1 7 T- AA2.02 A~,lilyto oeferm ne ana nierprer tnc lo o h ny as lney apply to me 01 So-ice Range NLCear Instrmenlat on EKpectea cnanje n SUJCP ranqe curnl rate when roas are move3 39 82 4-Emergency Procea-res Pan ADI ty 10 recognirt 061 I Area Radiation Monitoring (ARM) System Alarms I aDnOrma meat ons lor syslem operating 2.4.4 4 3 83 7 parameiers wnicn are entry eve cona I#onslor emeraencv ana amorma8 opera1 nq p!ocedues

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Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: The effect 01 074 i Inadequate Core Cooling 1 4 EA2.08 turbine bypass valve operation on RCS temperature and pressure Conduct of Operations: Knowledge of operator E03 I LOCA Cooldown and Depressurization1 4 2.1.2 resoansibilitiesdurino all modes of Dtant ODeration.

AD. tiy IO determine ana nlerprel Ine lollo.vmq as tney app.y 10 tne Coni m o d s Roo W lhdrawa 001 I Continuous Rod Withdrawal i 1 AA2.03 45 57 Proper acl.on5 10 oe laken I aLtOmalc safely IJnclions have nor laKen place

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I Knowledge of the reasons for the following 032 I Loss of Source Range Nuclear Instrumentationi 7 I AK3.01 responses as they apply to the Loss of Source Range Nuclear Instrumentation: Startup termination on source-range loss Ability to operate and I or monitor the following as 3.2 58 033 I Loss of Intermediate Range Nuclear they apply to the Loss of Intermediate Range "

AAt.03 Instrumentationi 7 Nuclear Instrumentation: Manual restoration of power Knowledge of the interrelations between the 059 I Accidental Liquid RadWaste Release I 9 AK2.01 Accidental Liquid Radwaste Release and the followina: Radioactive-liouidmonitors ADMYto aetem ne and nlerprel tne lo ow ng as 060 I Accidental Gaseous RadWaste Release I 9 AA2.06 lney apply IO me Acc aenta Gaseous Radwaste 36 61 Value ne-p for re.ease 01 rao oactwe gases CondLcl 31 Opera! oris ADI 1). lo perlorni spec I c EO2 I SI Termination 1 3 2.123 sfstem ana mtegraled p ani procBaLres d Jring ad 39 62 rnoaes 01 pant operat on E03 I LOCA Cooldown and Oepressurization1 4 I EK1.1 Knowieuge 01 me opera! ona mp car ons ul lne lo owing concepls as lney apply 10 Ine (-OCA Coonaonn ana Depress-r zai on, Componenis 34 63 caDac IY ana ILncl on 01 erneraencv svslems NUREG-1021 5

ES-401 Ginna Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 Knowledge of the interrelations between the (Natural Circulation Operations) and the following:

Facility's heat removal systems, including primary EO9 / Natural Circulation Operations / 4 X EK2.2 coolant. emergency coolant, the decay heat removal 3.6 64 systems, and relations between the proper operation of these systems to the operation of the facility.

Knowledge of the operational implications of the following concepts as they apply to the (High E141 High Containment Pressure/ 5 X EK1.3 Containment Pressure) Annunciators and conditions 3.3 65 indicating signals, and remedial actions associated with the (Hiah Containment Pressure).

WA Category Point Total: 1/2 2 2 1 1 a2 Grouo Point Total: I I 914 1 NUREG-1021 6

ES-401 Ginna Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 1 A2.26 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions. use proceduresto I 3.0 I 86 correct, control. or mitigate the consequences of those malfunctions or operations: Low VCT pressure Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS. and (b) based on those predictions, use procedures to A2.04 2.9 87 correct. control, or mitigate the consequences of those malfunctions or operations: RHR valve malfunction

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Equipment Control Knowledge of bases in 2.2.25 technical specifications for limiting conditions for 3.7 88 operations and safely limits.

Conduct of Operations: Knowledge of the purpose 2.1.28 and function of major system components and 3.3 89 controls.

Conduct of Operations: Ability to recognize 2.1.33 indications for system operating parameters which are ently-level conditions for technical 4.0 90 specifications.

Knowledge of the physical connections andlor K1.03 1 Knowledge of the operational implications of the K6.17 following concepts as they apply to the CVCS: 4.4 2 Flow paths for emergency boration Knowledge of RHRS design feature@) andlor interlock(s) which provide or the following: Lineup K4.11 3.5 3 for low head recirculation mode (external and internal)

Emergency Procedures/ Plan Ability to perform 2.4.49 without reference to procedures those actions that require immediate operation of system 4,0 4 components and controls.

Knowledge of the effect of a loss or malfunction K6.02 on the following will have on the ECCS: Core flood 3.4 5 t a n k laccurnulators)

Knowledge of the operational implications of the K5.02 following concepts as the apply to PRTS: Method 3.1 6 of forming a steam bubble in the PZR NUREG-1021 7

ES-401 Ginna Form ES-401-2 Written Examination Outline Plant Systems - Tier 2 Group 1 I A4 I Number 1 I I l-3 KIA Topics imp. Q#

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS. and (b) based on those predictions, use proceduresto A2.08 correct. control. or mitigate the consequences of 2.5 t

those malfunctionsor operations: Effects of shutting (automatically or othetwise) the isolation valves of the letdown cooler Ability lo predict andlor monitor changes in parameters (to prevent exceeding design limits)

Al.08 3.2 associated with operating the PZR PCS wntrols includina: SDrav nozzle DT 010 Pressurizer Pressure Control 012 Reactor Protection 2.1.2 K5.01 CoriuLcl of Operal,ons Knon eoge of operator responslo I cs a d ng al mooes 01 p a01 operat on Knowledge of the operational implications of the I1 30 3.3 IollowinO conceDtS as the amlv to the RPS: DNB Knodeage 01 tne eHeclo1a loss or ma I .ncl on 013 Engineered Safety Features K6.01 on tne folloning VI 11 nave on me ESFAS Sensors 27 Actuation dno delectors Ab 11y IOpreo LI d i d 01 iiiuiiilor cnanges in 313 Engineered Safety Features paramelers (lo Prevent exceeding aesign limits)

Al.09 34 Acluation associalea with operating Ine ESFAS conlros nc La ng T not Know edge of CCS de5 gn teaidreis ana or 022 Containment Cooling K4.04 riter O C ~ ( S Iwn cn prov ae lor me lo oning 28 I I I Coo no of coniro roo or be motors 026 Containment Spray X EqL pmenl Control Knorv eoge of lam Img 2.2.22 34 conamons lor operatlons ana salery m IS 339 Main and Reheat Steam K3.04 Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: MFW DUmDS I 2.5 tinowleago 01 lne pnysica conneclions ana or 339 Main and Reheat Steam K1.01 case-etlecl Tela1onsh ps oelncen Ine MRSS an0 31 me lo Ivw ng systems S G 359 Main Feedwater I 1 x 1 L K1.04 Knowledge of the physical connections andlor cause-effect relationships between the MFW and the tollowing systems: S/GS water level control SYStPrn 1 3.4 NUREG-1021 8

Form ES-401-2 ES-401 Ginna Written Examination Outline Plant Svstems -Tier 2 Group 1 Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) 3.6 18 061 AuxillarylEmergency A1.05 associatedwith operating the AFW controls Feedwater including: AFW flowlmotor amps Ability to (a) predict the impacts of the following malfunctionsor operations on the A N , and (b) based on those predictions, use procedures to 19 061 AuxillalylEmergency A2.09 correct. control. or mitiaate the conseauences of Feedwater

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those ma ILnctons or ;perat ons Tola loss Of leedwaler \no Va e Lislw n KA CAT) 062 AC Electrical Distribution K2.01 Knowledge of bus power supplies to the following:

Maior svstem loads I 3.3 I 20 Ability to manually operate andlor monitor in the A4.07 control room: Synchronizingand paralleling of 062 AC Electrical Distribution different ac supplies Ability to manually operate andlor monitor in the A4.03 063 DC Electrical Distribution control room: Batterv discharue rate Knowledge of bus power supplies to the following: 3,2 23 K2.03 064 Emergency Diesel Generator I 1 Control power

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Knowlnage 01 the enect tnat a 05s or rnall~nclion K3.01 01 the PRM system WI I nave on me lo ow ny 36 24 073 Process Radiation Monitoring Ran oactive ell Lent releases 076 Service Water A3.02 Ability to monitor automatic operation of the SWS, inrludinn. Ernnrnencv heat loads 1 3.7 1 25

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Ability to manually operate andlor monitor in the 26 A4.01 3.1 078 Instrument Air control room: Pressure gauges Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those predictions, use A2 03 3,5 27 103 Containment procedures to correct, control, or mitigate the consequences of those malfunctions or operations Phase A and B isolation Ability to monitor automatic operation of the 103 Containment A3.01 containment system, including: Containment

- isolation I I Group lint Total: 2WS WA Categoly Point Totals:

NUREG-1021 9

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ES-401 Ginna Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 2 079 Station Air NUREG-102 1 11

Facilitv: I Ginna I I

Dateof Exam: II 812712007 RO SRO-Only Category KIA # Topic IR Q# IR Ability to obtain and interpret station reference 2.1.25 materials such as graphs, monographs, and 3.1 tables which contain performance data.

Ability to locate and use procedures and 2.1.5 3.4

1. - directives related to shift staffing and activities.

Conduct of 2.1.22 Operations - Ability to determine Mode of Operation. 2.8 66 Ability to perform specific system and integrated 2.1.23 plant procedures during all modes of plant 3.9 67

- operation.

Subtota 2 Knowledge of the process for determining if the 2.2.10 margin of safety, as defined in the basis of any technical specification is reduced by a proposed 3.3 cnange. test or experiment Knowleoge 01 new and spent fLel movement

2. 2.2.28 3.5 Equipment Control procedures.

2.2.27 Knowledoe of the refuelina orocess. 2.6 68

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2.2.12 Knowledge of surveillance Drocedures. 3.0 69 Subtotal 2 2.3.2 Knowledge of facility ALARA program. 2.9 Ability to perform procedures to reduce 2.3.10 excessive levels of radiation and ouard aaainst 2.9 70 3.

Radiation Control I personnel expowre Know edge of 10 CFR 20 ana relatea lac lity 2.6 71 radial on control requ remcnts -- -

Subtotal 2 2.4.44 Knowledge of emergency plan protective action recommendations. 4.0 Knowledge of RO tasks performed outside the main control room during emergency operations 2.4.34 3.6 including system geography and system 4.

implications.

2.4.27 3.0 72 Emergency - Knowledge of fire in the plant procedure.

ProceduresI Plan 2.4.3 3.5 73

- Ability to identify post-accident instrumentation.

2.4.17 Knowledge of EOP terms and definitions. 3.1 74 2.4.29 Knowledge of the emergency plan. 75 Subtotal 4 Tier 3 Point Total

- 10 NUREG-I021 12

her I Randomly Reason for Rejection Selected;r(/A No LCOs and Safety Limits associated with Circulating Water System.

Randomly reselected G2.1.32 Removed due to excessive overlap with 027 AK2.03. Since 027 had only 1 KP

>2.5 in that category and 008 had 3 topics, removed the 008 topic and randomly reselected AK2.02 Overlap with 045 Al.05, so removed 035 topic since it had a lower KA 035 K5.01 importance value. Manually added K5.03 because it was the only other K5 in the 035 topic area. (Same as random)

No immediate actions for APE selected, and SRO requirements for test item 022 G2.4.49 development could not be adhered to. Randomly reselected G2.4.11

+

+

NUREG-1021 13

ES-301 Administrative Topics Outline Form ES-301-1 Rev Final Facility: Ginna Date of Examination: 8/27/07 Examination Level (circle one): SRO Operating Test Number: N07-1-1 AdministrativeTopic Type Code' Describe activity to be performed (see Note) 2.1.7 (4.4) Ability to evaluate plant pelformance and make Conduct of Operations operationaljudgments based on opeiating characteristics, reactor behavior, and instrument M, S interpretation JPM: Calculate QPTR and take action as necessary 2.1.25 (3.1) Ability to obtain and interpret station reference Conduct of Operations materials such as graphs, monographs, and tables which contain performance data JPM: Verify Required Service Water to Emergency Diesel Generators 2.2.26 (3.7) Knowledge of RefuelingAdministrative Equipment Control Reauirements JPM: Determine if the Electrical System is aligned for Mode 6 operation 2.3.10 (3.3) Ability to perform procedures to reduce Radiation Control excessive levels of radiation and guard against N. R personnelexposure JPM: Respond to a contaminated injured person 2.4.44 (4.1) Knowledge of Emergency Plan Protective Action Emergency Plan Recommendations M. R JPM: Make Protective Action Recommendations during a General Emergency NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrativetopics, when 5 are required.

'Type Codes & Criteria: (C)ontrol room Class(R)oom (D)irect from bank ( 5 3 for ROs; <for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams (s 1; randomly selected)

(S)imulator NUREG-1021. Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 Rev Final SRO Admin JPM Summary A1a The operator will be required to calculate QPTR using 0-6.4. The initial conditions will be off normal in that one NIS PR Detector will be out of service and reactor power will be at 65%. QPTR will be high requiring the operator to take action in accordance with TS 3.2-4. This JPM will be accomplished on the Simulator. (Similar JPM exists in Facility Bank - JR015.001)

Alb The operator, as part of the Daily Surveillance (0-6.13) will be directed to verify sufficient Service Water flow to the Emergency Diesel Generators. The Controlotron flow instrumentation will be unavailable requiring the operator to use Attachment 12, and Lake Temperature will be 69OF requiring the use of Attachment 14. The operator should determine that the SW flow to the EDG is insufficient and declare the EDG inoperable in accordance with TS 3.8.1.

A2 While in Mode 5, the operator will be directed to verify that the electrical system will support operation in Mode 6. The operator will need to compare the given electrical distribution system conditions against the requirements of 0-15.1 (Step 6.1.6). The operator will determine that one full Electrical Distribution train is not operable in accordance with TS 3.8.10, and determine that the system must be made operable before Core Alterations can occur.

A3 While operating at power, the operator will receive a report of an employee who received serious steam burns while performing RCS sampling activities. The sampling system will be reported as stabilized and no radioactive releases are in progress. The operator will be expected to respond in accordance with A-7, and address the issue that the individual being transported to an off-site medical facility is contaminated. The operator will need to determine which off site notifications must be made.

A4 The operator will be placed in a post-accident condition with a Large Break LOCA in Containment with a release from the Containment. The Emergency Plan has been implemented. The operator will be given an emergency classification of General Emergency, the wind direction, and then asked to determine the PARS. Following the initial recommendation, wind speed will change and the operator will be asked to make a Secondary PAR. (Similar JPMs exist in Facility Bank - JS340.003, 013, 015)

NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Final I Facility: Gtnn'l Exam Level (circle one): SRO(I) I SRO (U)

Date of Examination:

Operating Test No.:

8/27/07 N07-1-1 I Control Room Systems" (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U. including 1 ESF)

Type Code' Safety System IJPM Title Function

a. 004 Chemical &Volume Control S, N, A All Rods Not Fully Inserted

b. 010 Pressurizer Pressure Control System S , M, A Depressurize RCS Using Pzr PONSon Nitrogen

c. 005 Residual Heat Removal System 4P Start Up the RHR System

d. 062 A.C. Electrical Distribution S, D 6 Transfer 1A Instrument Bus

e. 041 Steam Dump System S, N. A 4s Use Atmospheric Dump Valves to Cooldown in a SGTR

f. 008 Component Cooling Water System S, M, A. L 8 Restore CCW to RHR

g. 006 Emergency Core Cooling System S,D, A 2 Makeup to the "B' Accumulator From RWST In-PlantSystems" (3 for RO: 3 for SRO-I: 3 or 2 for SRO-U)

h. 004 Chemical &Volume Control 2

Take Local Manual Control of Charging Pump

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i. 039 Main and Reheat Steam System 4s Locally Operate the ARVs

j. 027 Containment Iodine Removal System D. E 5 Locally Open Charcoal Filter Dampers in Relay Room

• Type Codes Criteria for RO ISRO-l ISRO-U NUREG-1021, Revision 9

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Final (A)ltetnate path 4-6 (6)/2-3(3)

(Qontrol room (D)irect from bank 5 8 (5)I S 4 (3)

(Qmergency or abnormal in-plant '. 1 (3)/ '. 1 (2)

(L)ow-Power / Shutdown 1 1 ( 2 ) / 2 1 (1)

(N)ew or (M)odified from bank including 1(A) "

2 2 (5)/ 1 (2)

(P)revious 2 exams i 3 / 5 2 (randomly selected) - None (WCA 21 (1)lZl (1) lS)imulator NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Svstems Outline Form ES-301-2 JPM Summarq JPM A This is a New JPM. The operator will be placed in a Post-Reactor Trip situation and told that the crew has progressed from E-0 to ES-0.1. Upon verifying that all Control Rods are on the bottom with MRPI, it has been discovered that one rod did not drop into the core. The operator will be directed to borate to ensure Shutdown Margin in accordance with the RNO of Step 4 of ES-0.1. The reactor makeup system has failed to operate as designed rendering this JPM an Alternate path JPM. The operator will diagnose the failure and proceed to ER-CVCS.l and take action in accordance with steps 4.2.2.1-3.

JPM B This is a modified JPM that will use Bank JPM J010.003 as its basis. The modificationwill render the JPM an Alternate Path JPM. The plant will have experienced a SGTR, reactor trip and safety injection. The RCPs will have been tripped and Instrument Air will not be available to Containment. All required steps of E-0 have been completed, and E-3 has been completed up to Step 22. Since no instrument air is available to the containment, the operator must Depressurize the RCS per Attachment 12.0, Attachment N2 PORV's using PCV-431C. However, when the operator attempts to open SOV-86166 it will be discovered that it will not open. A decision will be made to use PCV-430, and the operator will depressurize in accordance with the E-3 step 22.

JPM C This is Bank JPM JR005.002. The plant will be performing a normal cooldown prior to refueling outage. The RCS is at approximately 350 degrees F and 360 psig. The Overpressure ProtectionSystem is in service.

The operator will be required to place RHR in service per 0-2.2, Step 6.4.21. A Simulator instructor may need to maintain RCS temperature.

JPM D This is Bank JPM JR062.024. The "A" Inverter has failed. Instrument Bus 'A' is to be placed on Maintenance supply in preparationfor transferring Bus 14 to 'A' DG. The operator will be directed to transfer Instrument Bus "A"from normal power supply to maintenance supply per ER-INST.3 INSTRUMENT BUS POWER RESTORATION, Step 4.2. The operator will be told that an A-52.4 for the "A" Instrument Bus has already been initiated, and that no other temperature and pressure channels are defeated.

JPM E This is a new JPM. The operator will be placed in a situation where a Steam Generator Tube Rupture has occurred. The crew has entered E-0 and transitioned to E-3, and is ready to cooldown the RCS in E-3. The operator will be directed to cooldown the RCS to target temperature using the Steam Dump System. Upon operating the Steam Dump System the operator will determine that the system is unavailable and use the Atmospheric Relief Valves. The will render this JPM an Alternate Path JPM.

JPM F This is a modified JPM that will use Bank JPM JR008.002 as its basis. In the modified JPM the Standby CCW Pump will fail to start manually as well as automatically. This will require the operator to apply step 1 RNO step band reset and attempt to start the originally running pump. This pump will start upon attempted restart and CCW will be restored. The operator may recognize entry into AP-RHR.2. If so, Steps 1-3 will be verified, and then transition will be made to Step 19. When the Step 19 RNO is applied, re-entry to AP-CCW.3 may be made and the operator must apply step 1 RNO step band reset and attempt to start the originally running pump.

JPM G This is Bank JPM JR006.001. Accumulator "B" level has decreased to 56% with a resulting low pressure.

The operator will be directed to fillit to 65% using the 'c" SI Pump per procedure S-16.13, RWST WATER MAKEUP TO THE ACCUMULATORS. This JPM is an Alternate Path JPM, and the Alternate path (Option

2) will be used. When the C SI Pump is started it will trip, and the B SI Pump will need to be started instead.

JPM H This is Bank JPM JR004.009. A fire in the Cable Tunnel has required the evacuation of the Control Room.

ER-FIRE.2 is being implemented. The operator will be directed to perform lineups of the Charging Pump in accordance with procedure ER-FIRE.2, Attachment 4, Section 6.0. Continue until charging flow is verified to the RCS. The operator will be told that all previous steps of the procedure, including transfer of Charging Pump 'A" to Local control per Step 3.0, have been completed. This will require the operator to enter the RCA.

JPM I This is a modified JPM that will use Bank JPM JC039.003 as its basis. The plant has experienced a Control Complex fire and the Shifl Manager has implemented ER-FIRE.1, ALTERNATIVE SHUTDOWN FOR CONTROLCOMPLEX FIRE. The reactor has been tripped, the MSlVs were closed. the Turbine Generator was verified tripped and the Reactor Coolant Pumps were tripped. The operator will be told that all the equipment from the Appendix "R" locker is available, and that all the steps of Attachment 3, HCO in ER-NUREG-1021. Revision 9

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Final FIRE.1, have been completed. The operator will be directed to locally operate the Atmosperic Relief Valve, V-3411, per Step 4.5.4 of ER-FIRE.l and P-15.2, and dump steam through the ARV's locally. Upon local operation of the ARV, the operator will be able to open, but not close the valve, rendering the JPM an Alternate Path JPM. Upon realization that the ARV cannot be closed, the operator will close the root isolation valve for the ARV.

JPM J This is Bank JPM JR027.001. The plant experienced a reactor trip and SI. During E-0, while verifying the automatic actions in accordance with Attachment 27, the Charcoal Filter Dampers were found closed. The operator will be directed to locally open the Charcoal Filter Dampers using Trip Relay Pushbuttons in the Relay Room.

NUREG-1021. Revision 9

ADDendix D Scenario Outline Form ES-D-1 Operators:

Initial Conditions: The plant is at approximately 48% power (BOL), with both A and B MFW Pumps operating. Load was reduced due to 'A'MRN pump vibration. Pump repairs have been made, the pump has been tested and restarted, and a Load Ascension is anticipated.

Turnover: cted back in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />), and been defeated per ER-(I 2 I Pzr05A

@2% 1 C - RO, SRO PORV Leakage on PCV-430 (TS) - SRO 3 E-MIS10 to TS - SRO Post Contingency Low Voltage Alarm occurs 1.O (60 sec ramp) 4 CVClOAto I-RO, VCT Level Transmitter (LT-112) Fails Low 0% SRO 5 EDSO4B (TS) - An Electrical Ground Fault occurs on Safeguards Bus 16 SRO 6 N/A R-RO Load Decrease at lO%/hour.

N - SRO 7 ROD02-J4 C-RO, Dropped Rod SRO 8 ROD02-JlO M Second Dropped Control Rod occur during load decrease C-RO, A Total Loss of Off-Site Power occurs SRO 1 I RPS07M lo a n d N IC Steam Admission Valves on TDAFW Pump Fail to Open Automatically 1 11 I N/A I NIA I The A EDG is restored to service

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor ADDendix D Scenario Outline Form E S - D - I

-acility: GINNA Scenario No.: 2 Op Test No.: NRC-07-1-2 Examiners: Operators:

Initial Conditions: The Plant is at 100% power Steady-State (EOL), and been for the last 9 days following Refueling Outage.

Turnover: The following equipment is Out-Of-Service: A MDAFW Pump (Expected back in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), Steam Flow channel FT-475 (The channel has been defeated per ER-INST.1) and MCB Annunciator J-8 has alarmed spuriously several times over the last hour (I&C is investigating).

- Event Malf. No. Event Event No. Type* Description

~

1 RMS02B (TS) - SRO R-12fails high.

2 SGN04E (TS) - SRO A Small SG Tube Leak (.4 gpm) develops on B SG

(.4 gpm) 3 NIA NIA Plant Shutdown to Hot Standby required by Tech Specs 4 RODl A-66 C - RO, Uncontrolled Automatic Rod motion occurs.

SRO 5 PZR03C I - RO, SRO Pzr level Fails High.

(100) 6 NIA The SG Tube Leak will degrade to about 10 gpm 7 R-RO Rapid Down Power.

N - SRO 8 SlSOl M Inadvertent Safety Injection occurs causing an ATWS, (0-A) n 9 RPS05 N B b Automatic Reactor TriD is blocked -Manual TriD Received.

n 10 FDWllA b E MDAFW Pump fails to auto start and then trips shortly after manual start.

11 FDW12 (0) NIA TDAFW Pump trips on overspeed, 12 FWD15 NIA SAFW Flow Transmitters are isolated resulting in pump run out of AJB S A W Pump. (Pumps trips c 60 seconds after start).

• (N)orrnal, (R)eactivity, (I)nstrurnent, (C)omponent, (M)ajor ionn c;aruso - tiinna t x a m rage iI From: John Caruso To: mdsl Date: 08/30/2007 3:44:17 PM

Subject:

Ginna Exam Marvin, Two things I wanted to share with you and the Branch regarding our Ginna exam administration this week:

1) When we were administering our 3rd scenario, the operator was a little slow at controlling a FRV controller failure and the plant tripped prematurely. This resulted in skipping over 3 planned malfunctions (i.e., the SRO applicant was short one TS and one malfunction as a result). We continued to run the scenario to the end since the malfunctions had no direct affect on the major transient and remainder of the sceanrio. Gil made a very good suggestion tor recovery to just go back and Stan the scenario over again with the same set-up and we just ran the 3 previously missed malfunctions. In essence, we ran the same scenario in two parts. There was no need to run the spare scenario.

2) Since we didn't use the spare scenario, I asked Fred Guenther if it would be appropriate to withhold it from the exam record in ADAMS, so that the licensee could re-use this on a future exam. Fred suggested it would be okay, if I put a note to file explaining the circumstances that the spare scenario was not used and the licensee wishes to use it fo future e q m . It had no bear'ng on the exam outcome and it saves the licensee future resources. +*9-A?y--%

cc: b c h l ; dms3; gxj; jcs5; jmd; pap3; rrm2; sxg; thf