ML080940413

From kanterella
Jump to navigation Jump to search
Draft - Written Examination Outlines (Folder 2)
ML080940413
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 03/31/2008
From:
NRC Region 1
To:
Hansell S
Shared Package
ML073040288 List:
References
ES-401
Download: ML080940413 (31)
v • d • e


Text

ES-401 Systems Tier Totals Written Examination Outline 4

3 3

3 4

4 3

3 3

4 4

38 3

5 8

Form ES-401-1 1

2

3. Generic Knowledge & Abilities Categories 2

3 4

1 2

3 4

3 3

2 2

2 1

2 10 7

Note:

1.
2.
3.
4.
5.
6.

7.*

8.
9.

Ensure that at least two topics from every applicable WA category are sampled within each tier 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 category shall not be 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 +I 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.

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 section D.l.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 be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.l.b of ES-401 for the applicable WAs 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 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 1 OCFR55.43

ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 &

4 295006 SCRAM I 1 29501 9 Partial or Total Loss of Inst. Air 1 8 295021 Loss of Shutdown Cooling 1 4 295025 High Reactor Pressure I 3 295026 Suppression Pool High Water Temp. I 5 700000 Generator Voltage and Electrical Grid Disturbances / 6

~

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 &

4 295003 Partial or Complete Loss of AC16 295004 Partial or Total Loss of DC Pwr I 6

295005 Main Turbine Generator Trip I 3 295006 SCRAM I 1

295016 Control Room Abandonment I 7 295018 Partial or Total Loss of ccw 1 a PARTIAL OR COMPLETE LOSS OF Ability to recognize abnormal indications for system operating parameters that are LOSS OF FORCED CORE FLOW CIRCULATION : Natural circulation 2.4.1 1 - Emergency Procedures / Plan:

of abnormal condition owing responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

MAIN TURBINE GENERATOR TRIP :

1 Feedwater temperature I AKI.01 - Knowledge of the operational

ES-40 1 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 owing respons PARTIAL OR CO INSTRUMENT AIR : Standby air 46 295019 Partial or Total Loss of Inst. Air / 8 3.5 3.5 I

I compressor operation I AA2.01 -Ability to determine and/or I,2:5021 Loss of Shutdown Cooling interpret the following as they apply to I I I 1 x 1 I LOSS OF SHUTDOWN COOLING :

47 4.1 -

4.2 48 -

49 295023 Refueling Acc I 8 295024 High Drywell Pressure I 5 295025 High Reactor Pressure I 3 I major system components and controls.

I I

I I

I 2.4.50 - Emergency Procedures I Plan:

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

EA1.03 - Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Safetvlrelief X

4.4 50 I

I I

I valves: Plant-Specific I

I 1 EA1.02 - Abilitv to oDerate and/or monitor the following & they apply to SUPPRESSION POOL HIGH WATER I I I x l I I TEMPERATURE: Suppression pool spray:

295026 Suppression Pool High Water Temp. / 5 51 52 3.6 -

3.8 3.5 -

4.2 Plant-Specific EK2.03 - Knowledge of the interrelations between HIGH DRYWELL X

TEMPERATURE and the followina:

295028 High Drywell Temperature 1 I Reactor water level indication EK3.04 - Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL:

HPCS operation: Plant-Specific 2.2.44 - Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions.

EKI.07 - Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Shutdown margin EK2.05 - Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Site emergency plan AKI.02 - Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting AA1.03 - Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Voltage regulator Controls Group Point Total:

295030 Low Suppression Pool Water Level I 5 53 -

54 I 295031 Reactor Low Water Level I 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown I 1 3.4 55 295038 High Off-site Release Rate I 9 3.7 -

2.9 56 -

57 600000 Plant Fire On-site I 8 700000 Generator Voltage and Electric Grid Disturbances 3.8 I

58 -

I WA Category Totals:

ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 11 EAPE # / Name Safety Function I K1 I K2 I K3 I A I I A.2 I G I WA Topic(s)

I Imp. I Q# I Vac / 3 295009 Low Reactor Water Level / 2 500000 High CTMT Hydrogen Cone.

1 5 I

l l

295008 High Reactor Water Level /

2 I

l l

295015 Incomplete SCRAM / 1 I x I I 295020 Inadvertent Cont. Isolation /

5 & 7 I I x 1 295029 High Suppression Pool Water Level / 5 I

I I

295033 High Secondary Containment Area Radiation Levels / 1 X I 1

9 I

I I

/

I

/

295035 Secondary Containment High Differential Pressure I 5 295036 Secondary Containment High SumpIArea Water Level I 5 WA Category Totals:

2.4.31 - Emergency Procedures / Plan:

Knowledge of annunciator alarms, indications, or response procedures.

AA2.02 -Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL : Steam flowlfeed flow mismatch 2.4.6 Emergency Procedures / Plan:

Knowledge of EOP mitigation strategies.

M2.05 - Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL : Swell 2.4.2 - Emergency Procedures / Plan:

Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

AKI.01 - Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION : Loss of normal heat sink EK2.03 - Knowledge of the interrelations between HIGH SUPPRESSION POOL WATER LEVEL and the following: HPCS:

Plant-Specific EKI.02 - Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : Personnel protection EA1.02 - Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL

~

PRESSURE: SBGT/FRVS EK3.04 - Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL : Pumping secondary containment sumps Group Point Total:

4.1 3.7 -

4.7 2.9 -

4.5 -

3.7 -

3.3 3.9 3.8 -

3.1

ES-401 System # I Name Form ES-401-1 Imp.

Q#

K K

K K

K K

A A

A A

G 1

2 3

4 5

6 1

2 3

4 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 2.2.40 - Equipment Control: Ability to apply Technical Specifications for a system.

A2.07 -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:

Core Spray Line Break 2.2.25 - Equipment Control:

Knowledge of bases in technical specifications for limiting conditions for 4.7 86 3,6 87 4.2 88 215003 IRM operations and safety limits.

A2.12 - Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Valve openings 2.2.12 - Equipment Control:

Knowledge of surveillance procedures.

A4.01 -Ability to manually operate andlor monitor in the control room:

Pumps A3.03 -Ability to monitor automatic operations of the SHUTDOWN SHUTDOWN COOLING MODE) including: Lights and alarms K5.02 - Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :

3.0 4.1 4.3 COOLING SYSTEM (RHR 3.5 2.8 209001 LPCS 89 90 1

2 3

21 1000 SLC loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM : ECCS room cooler(s)

K5.04 - Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS):

Adequate core cooling: BWR-5,6 K2.02 - Knowledge of electrical power supplies to the following: Explosive 21 7000 RCIC 2.8 4

3.8 5

3.1 6

239002 SRVs provide for the following: Select rod insertion: Plant-Specific 203000 RHRILPCI: Injection Mode feature(s) and/or interlocks which 3.0 205000 Shutdown Cooling 7

205000 Shutdown Cooling 209001 LPCS 209002 HPCS 21 1000 SLC 21 2000 RPS Valve operation K6.05 - Knowledoe of the effect that a I

I valves K4.06 - Knowledge of REACTOR PROTECTION SYSTEM desian

ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 K

K K

K K

K A

A A

A G

1 2

3 4

5 6

1 2

3 4

System ## I Name connections andlor cause-effect relationships between INTERMEDIATE 215003 IRM 3.0 2.6 215003 IRM

~

215004 Source Range Monitor 3.1 -

2.6 -

3.5 4.2 I including: Lights and alarms I K2.02 - Knowledge of electrical power 215005 APRM I LPRM supplies to the following: APRM channels loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Condensate storage and 217000 RCIC transfer system A3.01 -Ability to monitor automatic oDerations of the AUTOMATIC 21 8000 ADS 223002 PClSlNuclear Steam Supply Shutoff 3.6 239002 SRVs 3.6 relationships between 239002 SRVs 3.3 259002 Reactor Water Level Control 3.6 3.9 261 000 SGTS 262001 AC Electrical Distribution 3.4

ES-401 Form ES-401-1 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name

/I 262002 UPS (AC/DC) 262002 UPS (AC/DC) 263000 DC Electrical Distribution 264000 EDGs 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water WA Category Totals:

K K

K A

A A

A G

Imp.

Q#

4 5

6 1

2 3

4 K3.07 - Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Movement of control rods:

Plant-Specific K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE 2.8 21 POWER SUPPLY (A.C./D.C.) : D.C.

electrical power K4.02 - Knowledge of D.C.

ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Breaker interlocks, permissives, bypasses and cross ties: Plant-Specific K3.01 - Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS 4.2 23 (DIESELIJET) will have on following:

Emergency core cooling systems A4.04 -Ability to manually operate and/or monitor in the control room:

Manual start, loading, and stopping of emergency generator: Plant-Specific A2.01 - Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, 2.9 25 control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions K1.02 - Knowledge of the physical connections and /or cause-effect relationships between CCWS and the following: Loads cooled by CCWS 2.6 20 3.1 22 3.7 24 3.2 26 Group Point Total:

264

ES-40 1 A

A A

A G

Imp.

Q#

1 2

3 4

Form ES-401-1 3.9 4.2 4.0 2.8 2.8 2.8 3.4 2.7 2.9 2.6 4.3 3.6 NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 91 92 93 27 28 29 30 31 32 33 34 35 System # / Name 214000 RPlS 245000 Main Turbine Gen. I Aux.

202001 Recirculation 201001 CRD Hydraulic 21 5002 RBM 204000 RWCU

~

214000 RPlS 256000 Reactor Condensate 216000 Nuclear Boiler Inst.

226001 RHWLPCI: CTMT Spray Mode 234000 Fuel Handling Equipment 239001 Main and Reheat Steam Overtravellin-out 2.4.47 - Emergency Procedures I Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

2.4.41 - Conduct of operations: Ability to apply technical specifications for a A2 03 -Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM,

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

X X

system A4.06 -Ability to manually operate operations of the ROD POSITION INFORMATION SYSTEM including:

I Full core display I K2.01 - Knowledge of electrical power

ES-401 K

K K

K K

K A

A A

A G

1 2

3 4

5 6

1 2

3 4

System # I Name NMP2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 Imp.

Q#

Form ES-401-1 X

2 I It K4.11 - Knowledge of REACTOR FEEDWATER SYSTEM design feature(s) andlor interlocks which 3.5 36 provide for the following: Recirculation runbacks 2.1.32 - Conduct of Operations: Ability to explain and apply system limits and 3.4 37 precautions.

K3.01 - Knowledge of the effect that a loss or malfunction of the OFFGAS SYSTEM will have on following:

Condenser vacuum 3.5 38 Group Point Total:

124 259001 Reactor Feedwater 268000 Radwaste I 271000 off-gas KIA Category Totals: i

ES-401 Generic Knowledge and Abilities Outline (Tier3)

Form ES-401-3 Facility:

9 Mile Point Unit II Outline 1 Date:

March 2008 SRO-Only Category WA #

Topic Ability to interpret reference materials, such as graphs, curves, tables, etc.

Knowledge of fuel handling responsibilities for SROs 2.1'25 235 66 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, AtC.

I.

Conduct of Operations 2.1.12 3.3 67 2.1.41 Knowledge of the refueling process.

2.8

-t-2 Subtotal 2.2.43 2.2.23 Knowledge of the process used to track inoperable alarms.

Ability to track Technical Specification limiting conditions for operations.

2.

Equipment Control Knowledge of less than or equal to one hour Technical Specification action statements for 3.9 systems.

Knowledae of surveillance Drocedures.

3.7 2.2.39 68 69 -

2.2.12 I

Ability to manipulate-the console controls as required to operate the facility between shutdown and designated power levels.

4.6 2.2.2 Subtotal 2.3.1 1 Ability to control radiation releases.

I

3.

Radiation Control 2.3.4 71 3.7 Knowledge of radiation exposure limits under normal or emergency 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 fixed radiation monitors and alarms, portable survey instruments, personnel monitoring eauiDment. etc.

3.4 2.9 2.3.13 72 2.3.15 73

ES-401 2.4.21

4.

Emergency Procedures /

Plan Generic Knowledge and Abilities Outline (Tier3)

Subtotal 2.4.30 2.4.14 2.4.35 Subtotal Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Knowledge of events related to system operatiodstatus that must be reported to internal the State, the NRC, or the transmission system operator. Knowledge of events related to system operatiodstatus that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. Knowledge of events related to system operatiodstatus that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system organizations or external agencies, such as operator. Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Knowledge of general guidelines for EOP usage.

Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.

Tier 3 Point Total 3.8 3.8 Form ES-401-3 74 75

Record of Rejected KIAs Form ES-401-4 1

ES-401 212 239001 I AI.05 Randomly Selected UIA 1 Tier/Group 1 111 111 1 / 2 I

295021 12.4.8 700000 12.4.6 500000 12.2.38 295027 I EK2.02 3

1 212 I 2390031K4.01 2.4.45 1

211 I 215003lK1.03 21 8000 I A3.05 223002 I A2.07 I

211 261 000 I 2.1.25 I

212 201 004 I K6.04 212 21 5002 I K2.03 I

212 268000 I 2.1.30 I/

1 I 1 I 700000 lAA1.05 295029 I EK2.09 2.2.14 211 I 20500012.2.40 211 300000 I 2.2.12 Reason for Rejection

(#52) Mark Ill containment does not apply to NMP2.

Randomly selected 295028 EK2.03

(#36) MSlV Leakage system does not apply at NMP2.

Randomly selected 259001 K4.11

(#8) Topic does not apply at NMP2. Randomly selected K1.07

(#I 3) Overlap with other portions of the exam. Randomly reselected A3.01

(#14) Overlap with scenario exams. Kept same System 223002, randomly reselected A2.09.

(#I

8) Original selection not related to aspects of system operation. Kept same System 261 000, randomly reselected statement 2.2.42.

(#28) Turbine First Stage Pressure input to RSCS has been defeated at NMP2. RSCS has low operational impact.

Randomly reselected System 21 5002.

(#31) Double Jeopardy with Tier 2 Group 1 21 5005 K2.02.

Both topics are APRM Power Supply. Randomly reselected 256000 K2.01

(#35) MSL Radiation Monitors do not initiate Group 1 Isolations at NMP2. Randomly reselected statement AI.01

(#37) Could not write a discriminating RO Level question for local operation of Radwaste controls. Randomly replaced with statement 2.1.32

(#58) Could not write an operationally oriented RO Level question. Randomly reselected statement AA1.03

(#62) Topic does not apply at NMP-2. Randomly reselected EK2.03

(#69) Could not write a discriminating RO Level question.

Randomly reselected statement 2.2.12

(#79) Topic not used in conjunction with EOPs. Randomly reselected 2.4.9.

(#82) Topic not directly addressed in EOPs. Randomly reselected 2.4.4.

(#85) Could not write a discriminating SRO Level question.

Randomly reselected 2.4.6.

(#86) Double Jeopardy with (#79), Shutdown Cooling Technical Specifications. Other Shutdown Cooling topics are covered on RO #2, #3, and #47. Randomly reselected System 21 5003.

(#go) Could not write a discriminating SRO Level question pertaining to Instrument Air I Surveillance Procedures.

Randomly reselected System 239002.

(#I 00) Topic covered extensively in scenarios. Randomly reselected statement 2.4.30

ES-401 Record of Rejected WAs Form ES-401-4

ES-30 1 Administrative Topics Outline Form ES-301-1 Radiation Control Emergency Plan Facility:

NMP2-NRC Examination Level: RO 1

SRO D, R N,R I/

I Administrative Topic TY Pe (see Note)

Code*

Conduct of Operations Conduct of Operations

/I I

NR I

I Equipment Control ll I

NR U

I Date of Examination: March 08 Operating Test Number:

1 Describe activity to be performed Operability/Reportability Review for a CR Review a completed portion of control room daily logs. Identify deficiencies and TS requirements.

Review a tagging request for the A RHR pump. Address Tech Specs.

Generate and approve an Emergency Exposure Authorization Determine EAL and complete initial notification paperwork.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (53 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (21)

(P)revious 2 exams (11; randomly selected)

AIA The candidate will perform an Operability/Reportability Review for a CR. This is a bank JPM AI B The candidate will review a completed portion of control room daily logs. Identify deficiencies and TS requirements. This is a new JPM A2 The candidate will review a tagging request for the A RHR pump (or other piece of TS equipment).

The request will contain several errors and TS will be addressed. This is a new JPM.

A3 The candidate will generate and approve an Emergency Exposure Authorization. This is a bank JPM A4 The candidate will perform an EAL determination given plant conditions and complete initial notification paperwork. This is a time critical modified JPM.

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility:

NMP2 Date of Examination:

MARCH 2008 Exam Level (circle one):

SRO Operating Test No.:

1 Control Room Systems@ (8 for RO; 2 or 3 for SRO-U, including 1 ESF) SRO-U in BOLD #s S-4,6,7/P-1,2 System / JPM Title I1 S-I ReactodTurbine Pressure Regulating / Place Main Turbine Shell Warming in service IAW N2-OP-21 -Rev.8 WA 241000 A4.18 2.912.8 S-2 Primary Containment / Initiate Division I H2/02 monitoring to monitor sample path 4, post LOCA restart required.

WA 223001 A4.04 3.5/3.6, A4.05 3.6/3.6 S-3 Reactor Feedwater / Transfer Feedwater Level Control to FWS-LV55A at approximately 2% power IAW N2-OP-3.

WA 259001 A4.05 4.013.9 S-4 RClC / Place RClC in service due to a level transient, RClC fails to isolate on isolation signal.

WA 217000 A2.01 3.813.7 A4.01 3.713.7, A3.06 3.513.5 S-5 Standby Gas Treatment / Align SBGTS Train A to reduce Drywell pressure IAW N2-OP-61A KIA 295024 EA1.20 3.513.6 (02-OPS-SJE-261-2-02)

S-6 Emergency Diesel Generators / Manual start and load of the Division Ill EDG from Panel P-852 IAW N2-OP-100B. EDG overspeeds.

WA 264000 A4.04 3.713.7 (02-OPS-SJE-264-2-67)

S-7 SRMs / Insert SRMs during a shutdown IAW N2-02-92 WA 21 5004 A4.04 3.2/3.2 S-8 NIA In-Plant Systems@ (3 for RO; 3 or 2 for SRO-U)

Type Code*

Safety Function 3

9 2

4 5

6 7

P-I Instrument Air / Startup of Air Dryer 21AS -DRY1 B IAW N2-OP-19 WA 300000 A2.01 2.9/28 I

N.R 8

NUREG-I 021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 1

D,R,E 1

1 P-2 Standby Liquid Control / Boron Injection with Hydro Pump IAW 1

EOP-16, Att.15 WA 295037 EA1. I O 3.7/3.9 (02-OPS-PJE-211-2-01)

D.C. Distribution / Place Battery Charger 2BYS-CHGRIAI is WA 263000 AI.01 2.542.8 (02 -0PS-PJE-263-2-03 Revl) p-3 placed in service.

NUREG-1021, Revision 9

ES-301 Control Roomh-Plant Systems Outline Form ES-301-2

  • Type Codes (A)lternate path (C)ontrol room (D)irect from bank (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 W C A (S)imulator 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.

Criteria for RO I SRO-I I SRO-U 4-6 ! 4-6 12-3

91. 81. 4 I / 1 1. 1 I/ 1 1. 1 2 1 2 1 1 3 I 3 I 2 (randomly selected) 1 1 1 1 1 2008 NRC Examination Summary Description of JPMs S-I This is a new alternate path JPM in the Reactor Pressure Control safety function area. The candidate will place Main Turbine Chest Warming in service IAW N2-OP-21 -Rev.8, Section 3.0 and the Main Turbine will roll off the turning gear requiring additional operator actions.

S-2 This is a new JPM in the Radioactive Release safety function area. The candidate will initiate Division I H2102 monitoring to monitor sample path 4 (suppression chamber), post LOCA, IAW N2-OP-82. Section H.l.O.

S-3 This is a new JPM in the Reactor Water Inventory Control safety function area. The candidate will Transfer Feedwater Level Control to FWS-LV55A at approximately 2% power IAW N2-OP-3. Section E.3.16 S-4 This is a bank alternate path JPM in the Heat Removal From Reactor Core Safety Function area.

The candidate will be required to place RClC in service due to a level transient. A valid isolation signal will occur and RClC will fail to isolate. Operator action is required to isolate and trip RCIC.

S-5 This is a bank JPM in the Containment Integrity safety function area. The candidate will align SBGTS Train A to reduce Drywell pressure IAW N2-OP-61A, Section H.

S-6 This is a modified bank alternate path JPM in the Electrical safety function area. The candidate will perform a Manual start and load of the Division Ill EDG from Panel P-852 IAW N2-OP-I00B, Section F.2.0. EDG will overspeed but will not trip requiring operator actions to trip the EDG.

S-7 This is a new JPM in the Instrumentation safety function area. The candidate will insert SRMs during a shutdown IAW N2-OP-92, Section G.4.0.

P-I This is a new JPM in the Plant Systems safety function area. The candidate will startup Air Dryer 21AS -DRY1 B, IAW N2-OP-19, Section E.2.0.

NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 P-2 This is a bank JPM in the Reactivity Control safety function area. The candidate will establish a flowpath from SLS tank to the Reactor Vessel using a hydro pump and hoses staged, complete with an air supply to the pump and then commence Boron injection.

P-3 This is an alternate path bank JPM in the Electrical safety function area. The candidate will place Battery Charger 2BYS-CHGRI AI in service IAW N2-OP-73A Section E.4.0. Alternate actions will be required due to high charger current.

NUREG-1021, Revision 9

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Examiners:

Operators:

Initial Conditions: Simulator IC-241

1. Reactor Power 90%

Turnover:

1. All equipment operable.
2. Swap Service Water Pumps from the 2SWP*P1 B to the 2SWP*P1 F for normal equipment rotation. Pre-start checks have been completed and an A 0 is standing by at the F pump Scenario No.: NRC-01 Op-Test No.: March 2008 Event 1 Malf. No.

No.

I (N)ormal, (R)eac Event I

Event DescriDtion Power decrease to 85%.

PSP exceeded, RPV blowdown required (CT)

EOP-C2

vity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 1 March 2008

Facilitv: Nine Mile Point 2 ACTUAL TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

Events 2,3,4,5,6,8

2. Malfunctions after EOP entry (1-2)

Events 8

3. Abnormal events (2-4)

Event 2 SOP-I3 Event 3 SOP-34, Event 4 SOP-6,29

4. Major transients (1-2)

Event 7

5. EOPs enteredlrequiring substantive actions (1-2)

Events 7,8 EOP-RPV, EOP-PC

6. EOP contingencies requiring substantive actions (0-2)

Event 9 EOP-C2

7. Critical tasks (2-3)

CRITICAL TASK DESCRIPTIONS:

CT-1.O Initiate DW spray to control containment CT-2.0 Initiate RPV Blowdown when PSP is exceeded Event 5 SOP-8 pressure and DW spray established ATTRIBUTES 6

1 4

1 2

1 2

OD-Test No.: March 2008 NRC Scenario 1 March 2008

SCENARIO

SUMMARY

The scenario begins at 90% power. The BOP will be required to swap from the 2SWP*P1 B to the 2SWP*P1 F for normal equipment rotation. After completion of the pump swap, one of the running RBCLCW pumps will trip and the standby pump will fail to auto start. The standby pump can be manually started by the operators.

An ADS SRV will then go open but can be closed bycycling the control switch. When this occurs, a Drywell vacuum breaker pair will fail open. The crew will enter the SOP for Stuck Open SRV and the SRO will review Technical Specifications (TS) in regard to the vacuum breakers.

Once TS are addressed a trip will occur on one Feedwater Pump and only a partial Recirc Runback will occur. SOPS must entered to control feedwater level and address the reduction in power. Cram rods will need to be inserted and TS must be addressed due to loop flow mismatch.

When plant conditions stabilize, one control rod will drift out requiring an entry into the SOP for Unplanned Power Changes. The SOP will require that power be lowered and the RO will reduce recirculation flow IAW procedures. After power is lowered, another control rod will drift out requiring a reactor scram.

When the reactor scrams a steam leak will occur inside the drywell. LPCS and RHR A will fail to initiate and RHR B will trip when Drywell pressure exceeds 1.68 psig. However, both LPCS and RHR A can be started manually. RHR A must then be placed in Suppression Chamber sprays per EOP-PC. When Suppression Chamber pressure exceeds 10 psig, the crew must spray the Drywell with RHR A (Critical Task). Once DW sprays are in service and PSP is evaluated, it will be recognized that RPV Blowdown is required once above the PSP (Critical Task). The crew will blowdown the reactor and continue to control Containment pressure.

The scenario ends with the blowdown complete and containment pressure lowering.

NRC Scenario 1 March 2008

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Examiners:

Operators:

Initial Conditions: Simulator IC-I 7 or equivalent

1. Reactor Power 100%

Turnover:

1. All equipment operable.
2. Perform RHR Pump Operability Test IAW N2-OSP-RHS-QOO6 Scenario No.: NRC-02 Op-Test No.: March 2008 Event I Malf. No.

No.

I Event DescriDtion I

RHS B/C Water Leg Pump breaker trip (TS)

TS (SRO) be placed in service manually.

N2-SOP-23, EHC Press Reg Failure N2-SOP-IO1 D, Rapid Power Reduction.

ion FCV failure causes FCV to open.

7 (N)ormal, (R)eact ivi ty, (I)nst ru men t, (C)om ponen t, (M)ajor NRC Scenario 2 March 2008

Facility: Nine Mile Point 2 Scenario TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

Events 2,3,4,5,6,8,9

2. Malfunctions after EOP entry (1-2)

Events 8,9

3. Abnormal events (2-4)

Event 2 -SOP-19, Event 3SOP-101 D, SOP-23 Event 5 - SOP-6, Event 6 SOP-8

4. Major transients (1-2)

Event 7

5. EOPs enteredhequiring substantive actions (1-2)

Events 7,8,9 EOP-RPV, EOP-PC

6. EOP contingencies requiring substantive actions (0-2)

Event 7,8,9 EOP-C5,

7. Critical tasks (2-3)

CRITICAL TASK DESCRIPTIONS:

CT-1.O Place ADS inhibit switches to ON to prevent CT-2.0 restore & maintain RPV level above the injection under ATWS conditions MSCWL precluding the need to perform a RPV Blowdown.

CT-3.0 Inject SLC before exceeding HCTL CT4.0 RO inserts all control rods 0.:

NRC-02 ACTUAL ATTRIBUTES 7

2 4

1 2

1 4

Op-Test No.: March 2008 NRC Scenario 2 March 2008

SCENARIO

SUMMARY

The scenario begins at 100% power. The RO will perform the surveillance test for the C RHR Pump, N2-OSP-RHS-QOO6. While the pump is running, the breaker will trip for the RHS B/C Water Leg Pump RHS*P2 requiring a TS entry by the SRO (TS 3.5.1.C - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). Once TS are addressed, the in service Instrument Air Compressor will trip requiring operator action to manually start the C standby compressor. The B compressor will not start.

Oscillations of the in service EHC pressure regulator will occur and require actions to swap to the alternate regulator and lower reactor pressure. Additionally, the RO will be required to lower reactor power to 85% IAW the SOP-23.

Once conditions stabilize, a HPCS spurious start will occur requiring operator action to terminate the initiation. The SRO will address TS for HPCS inoperability and 2 other ECCS pumps inoperable (3.0.3). After addressing TS, the Feedwater Master Controller will then fail as-is. The crew will enter SOP-6 and control feedwater in manual.

Additionally, a failure of the Recirculation FCV will cause the FCV to open. Operator action will be required to control the FCV and reactor level. Cram rods may be inserted or Recirc flow lowered to lower reactor power to pre-transient levels.

The backup EHC pressure regulator will fail and result in a rapid RPV pressure rise.

The reactor will automatically scrams, however, all control rods will not fully insert and A and B reactor feed pumps will trip. EOPs RPV, EOP-Failure-To-Scram will be entered. The RO must inhibit ADS to prevent injection during the ATWS (CT).

The RClC turbine can be manually controlled after a controller malfunction. SLC pumps will fail to auto-start and must be manually started prior to exceeding the HCTL (CT). RPV level must be restored with C FW pump, RClC (or Condensate Booster Pumps with RPV pressure lowered) precluding the need to perform a RPV Blowdown (CT). The RO will implement actions to insert control rods until all rods are inserted (CT).

The scenario can be terminated when RPV level is being controlled in the required band and all control rods are inserted.

NRC Scenario 2 March 2008

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Examiners:

Operators:

Scenario No.: NRC-03 Op-Test No.: March 2008 Initial Conditions: Simulator IC-244

1. Plant startup is in progress IAW N2-OPIOIA @ Step 2.45.
2. RWM @ Step 16. Rod 34-1 1
3. Reactor Pressure is at approximately 900 psig.
4. One Bypass Valve is approximately 15%
5. Other operators will be performing SJAE startup later today.

Turnover:

1. Continue Power Increase to get one bypass valve open approximately 25%
2. Transfer Reboiler Steam Supply to Main Steam IAW N2-OP-25, Section 5.0, then continue startup Event I

I

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 3 March 2008

TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

Events 3,4,5,6,8,9

2. Malfunctions after EOP entry (1-2)

Events 8.9 3

ACTUAL ATTRIBUTES 6

2

3. Abnormal events (2-4)
5. EOPs enteredlrequiring substantive actions (1-2)

Events 7.9 EOP-RPV. EOP-PC.

7. Critical tasks (2-3)

I 2

6. EOP contingencies requiring substantive actions (0-2)

Event 10 EOP-C4 I

1 CRITICAL TASK DESCRIPTIONS:

CT-1.O initiate an RPV blowdown when level indication is lost or if the PSP is exceeded.

CT-2.0 flood the RPV to the elevation of the main steam lines IAW RPV flooding.

NRC Scenario 3 March 2008

SCENARIO

SUMMARY

The scenario begins with a plant startup in progress and reactor pressure at 900 psig.

Control rods will be withdrawn until one bypass valve is open 25%. When that occurs, the operators will transfer the Reboiler Steam Supply to Main Steam. After that occurs, the plant startup will continue with control rod withdrawal.

While increasing power an inop trip will occur on IRM channel A. The operators will bypass the affected IRM and reset the half scram. TS (TS 3.3.1. I ) will be referred to by the SRO, no LCO entry required.

Next, a loss of offsite power to Div I Switchgear occurs when breaker 16-2 fails open, the crew will take action per SOP-N2-3 to stabilize plant parameters and adjust Service Water flow. The SRO will be required to address TS (TS 3.8.1.A - 1 Hr for surveillance

- 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> SD LCO). When plant conditions are stabilized IRM G will have an upscale trip requiring a TS entry by the SRO (TS 3.3.1. I.A).

A seismic event will then occur causing a Condensate pump to trip and the standby pump must be started. A service water pump will also trip requiring entry to TS 3.7.1.E.

-72 hours). SOP-N2-90 will be entered to address the seismic event.

Once conditions are stabilized, a seismic aftershock will occur causing a rupture of the recirculation loop suction line. Initially drywell pressure rise slowly and the operators will manually scram the plant. The leak will get larger and entry into EOP-RPV and EOP-PC will be required. The event is complicated by a failure of Division 2 ECCS to initiate automatically. Operator action will be required to manually initiate Division 2 ECCS.

The crew must initiate an RPV blowdown when level indication is lost or if the PSP is exceeded (CT). Only 5 SRVs will initially open requiring operator action to open additional SRVs. The crew must flood the RPV to the elevation of the main steam lines IAW RPV flooding EOP-C4 (CT). Once level is RPV is flooded, containment parameters can be addressed.

Once level has recovered to the main steam lines as indicated by acoustic monitors, RPV pressure rising, or tailpipe temperatures lowering, the scenario may be terminated.

NRC Scenario 3 March 2008

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 2 Examiners:

Operators:

Initial Conditions: Simulator IC-I 7

1. Reactor Power 100%

Turnover:

1.
2.

Scenario No.: NRC-04 Op-Test No.: March 2008 Reduce power to 90% per LD for a rod line adjustment which will take place on the next shift.

Perform N2-OSP-RMC-S@001 Control Rod Movement and Position Verification Surveillance Test Event OSP-RMC-W@001 Control Rod Movement and I

I days.

I TS(SR0)

RB. Power reduction may be required.

gear 15, loss of one division of RPS solenoids ss of AC Power, N2-SOP-97 RPS Failures egraded CCP System, N2-SOP-60, Loss of DW ARI successful.

EOP-RPV. EOP-CS Failure-To-Scram I DG04A.B I I N2-SOP-3. N2-SOP-11. EOP-RPV. EOP-PC (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 4 March 2008

Facilitv: Nine Mile Point 2 Scenario No.: NRC-04 Op-Test No.: March 2008 ACTUAL

-~

TARGET QUANTITATIVE ATRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

Events 3,5,6,7,8,10

2. Malfunctions after EOP entry (1-2)

Events 10

3. Abnormal events (2-4)

Event 3-SOP-30, Event 6-SOP-68, Event 7-3,13,60,97,

4. Major transients (1-2)

Event 9

5. EOPs enteredhequiring substantive actions (1-2)
6. EOP contingencies requiring substantive actions (0-2)

EOP-Failure to Scram,

7. Critical tasks (2-3)

CRITICAL TASK DESCRIPTIONS:

CT 1.O - Upon Mode Switch and RPS PB Failure, Scram is accomplished with RRCS.

CT 2.0 - On EDG Auto-start failure, start the EDGs from the control room IAW SOP-03.

CT 3.0 - drywell spray is initiated prior to exceeding the PSP.

EOP-RPV, EOP-PC ATTRIBUTES I

2 NRC Scenario 4 March 2008

SCENARIO

SUMMARY

The scenario begins with the crew performing a power maneuver power will be lowered to 90% for a rod line adjustment which will take place on the next shift. The crew will perform surveillance N2-OSP-RMC-W@001 Control Rod Movement and Position Verification. Following the third rod tested, a trip of the running CRD pump will occur due to a clogged suction strainer. Actions must be taken to start the standby pump.

Once the standby pump is started, position indication for a control rod will be lost requiring the SRO to address Technical Specifications (TS 3.1.3.C.1 and 2).

One of the Control Room AC units will trip and the standby unit will fail to auto-start.

The SRO will enter TS 3.7.3.A. The running Stator Water Cooling pump then trips and the standby pump fails to start but can be manually started. If a bypass valve opens due to the pump trip and generator runback a power reduction will be required. When conditions stabilize, a loss of Switchgear 15 will occur which affects RPS and requires operator action to transfer RPS to its alternate supply. Additionally, Switchgear 15 may be re-powered.

Once conditions are stabilized, a small leak will develop in the drywell requiring a manual scram of the reactor. A failure of the mode switch and RPS manual scram pushbuttons will occur requiring the use of RRCS to insert control rods (CT). The SRO will enter EOPs and EOP contingencies.

Once the rods are inserted, a Loss of Offsite Power will occur. Additionally, the EDGs will fail to auto start. The operators will take actions to start the EDGs in the control room IAW SOP-O3(CT). The containment leak will get worse. RHR pump P I A will trip when placed in service and Suppression Chamber Spray must continue with RHR pump P I B. The operators will be expected to control containment pressure with Drywell Spray prior to exceeding the Pressure Suppression Pressure (CT).

The scenario may be terminated once containment pressure is decreasing and RPV level and pressure are being controlled.

NRC Scenario 4.

March 2008

Retrieved from "https://kanterella.com/w/index.php?title=ML080940413&oldid=5333011"