ML13169A183

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301 Draft Administrative Documents
ML13169A183
Person / Time
Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 06/18/2013
From:
NRC/RGN-II
To:
Duke Energy Carolinas
References
50-369/12-301, 50-370/12-301
Download: ML13169A183 (24)
v • d • e


Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: MCGUIRE Date of Exam: AUGUST 2012 RO KIA Category Points SAO-Only. Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • Total
1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

Abnormal 2 I 1 2 N/A 2 2 N/A I 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 2 2 2 2 2 3 3 3 3 3 3 28 3 2 5 2.

Plant 2 I I I I I I I I 1 I 0 10 0 2 I 3 Systems Tier Totals 3 3 3 3 3 4 4 4 4 4 3 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 2 I 2 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SAO-only outlines (i.e., except for one category in Tier 3 of the SAO-only outline, the "Tier Totals" in each KIA category 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 +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SAO-only exam must total25 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/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA 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 KIAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SAO ratings for the RO and SAO-only portions, respectively.

6. Select SAO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.

8. On the following pages, enter the KIA 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 SAO-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 SAO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Fonn ES-401-3. Limit SAO selections to KIAs that are linked to 10 CFR 55.43.

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 007EG2.4.49 Reactor Trip - Stabilization - Recovery 4.6 4.4 0 0 0 0 0 0 0 0 0 0 ~ Ability to perform without reference to procedures those 11 actions that require immediate operation of system components and controls.

008AK3.02 Pressurizer Vapor Space Accident I 3 3.6 4.1 O D ~ D D D D D D D D Why PORV or code safety exit temperature is below RCS or PZR temperature 009EA2.14 SmaiiBreakLOCAI3 3.8 4.4 0 0 0 0 0 0 0 ~ 0 0 0 Actions to be taken if PTS limits are violated 011EK2.02 LargeBreakLOCAI3 2.6 2.7 0 ~ 0 0 0 0 0 0 0 0 0 Pumps 022AG2.1.31 Loss of Ax Coolant Makeup I 2 4.6 4.3 0000000000~ Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.

025AK3.02 Loss of AHA System I 4 3.3 3.7 00~00000000 Isolation of AHA low-pressure piping prior to pressure increase above specified level 026AA1.03 Loss of Component Cooling Water I 8 3.6 3.6 000000~0000 SWS as a backup to the CCWS 027AK1.02 Pressurizer Pressure Control System 2.8 3.1 ~0000000000 Expansion of liquids as temperature increases Malfunction I 3 029EK1.01 ATWSI1 2.8 3.1 ~0000000000 Reactor nucleonics and thermo-hydraulics behavior 038EG2.4.8 Steam Gen. Tube Rupture 13 3.8 4.5 0000000000~ Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

040AA1.11 Steam Line Rupture - Excessive Heat 3.2 3.1 000000~0000 MFWsystem Transfer 14 Page 1 of 2 12/912011 8:52 AM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 054AA2.03 Loss of Main Feedwater I 4 4.1 4.2 D D D D D D D ~ D D D Conditions and reasons for AFW pump startup 055EK3.02 Station Blackout I 6 4.3 4.6 D D ~ D D D D D D D D Actions contained in EOP for loss of offsite and onsite power 056AK1.01 Loss of Off-site Power I 6 3.7 4.2 ~ D D D D D D D D D D Principle of cooling by natural convection 057AA 1.04 Loss of Vital AC lnst. Bus I 6 3.5 3.6 D D D D D D ~ D D D D RWST and VCT valves 062AA2.04 Loss of Nuclear Svc Water I 4 2.5 2.9 D D D D D D D ~ D D D The normal values and upper limits for the temperatures of the components cooled by SWS WE04EK2.2 LOCA Outside Containment I 3 3.8 4.o D ~ D D D D D D D D D Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

WE05EK2.2 Inadequate Heat Transfer - Loss of 3.9 4.2 D ~ D D D D D D D D D Facility's heat removal systems, including primary Secondary Heat Sink I 4 coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

Page 2 of 2 12/912011 8:52 AM

ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 003AG2.2.22 Dropped Control Rod I 1 4.o 4.7 D D D D D D D D D D ~ Knowledge of limiting conditions for operations and safety limits.

005AK2.02 Inoperable/Stuck Control Rod I 1 2.5 2.6 D ~ D D D D D D D D D Breakers, relays, disconnects and control room switches 024AA 1.18 Emergency Boration I 1 3.7 3.6 D D D D D D ~ D D D D Emergency boron flow meter 032AA2.05 Loss of Source Range Nl I 7 2.9 3.2 D D D D D D D ~ D D D Nature of abnormality, from rapid survey of control room data 033AK3.01 Loss of Intermediate Range Nl I 7 3.2 3.6 D D ~ D D D D D D D D Termination of startup following loss of intermediate-range instrumentation 061 AK1.01 ARM System Alarms I 7 2.5 2.9 ~ D D D D D D D D D D Detector limitations 067AA 1.01 Plant Fire On-site I 8 3.6 3.6 D D D D D D ~ D D D D Respirator air pack WE13EK3.4 Steam Generator Over-pressure I 4 3.1 3.3 D D ~ D D D D D D D D RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

WE16EA2.2 High Containment Radiation I 9 3.o 3.3 D D D D D D D ~ D D D Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Page 1 of 1 12/9/2011 8:52 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 003K5.04 Reactor Coolant Pump 3.2 3.5 DDDD~DDDDDD Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow and feed flow 003K6.04 Reactor Coolant Pump 2.8 3.1 DDDDD~DDDDD Containment isolation valves affecting RCP operation 004G2.2.4 Chemical and Volume Control 3.6 3.6 DDDDDDDDDD~ (multi-unit) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.

005K4.12 Residual Heat Removal 3.1 3.7 DDD~DDDDDDD Lineup for piggyback mode with CSS 006K2.02 Emergency Core Cooling 2.5 2.9 D~DDDDDDDDD Valve operators for accumulators 007K5.02 Pressurizer Relief/Quench Tank 3.1 3.4 DDDD~DDDDDD Method of forming a steam bubble in the PZR 008A1.02 Component Cooling Water 2.9 3.1 DDDDDD~DDDD CCW temperature 008A3.08 Component Cooling Water 3.6 3.7 DDDDDDDD~DD Automatic actions associated with the CCWS that occur as a result of a safety injection signal 010K4.01 Pressurizer Pressure Control 2.7 2.9 DDD~DDDDDDD Spray valve warm-up 012K1.06 Reactor Protection 3.1 3.1 ~DDDDDDDDDD T/G 012K3.01 Reactor Protection 3.9 4.0 DD~DDDDDDDD CADS Page 1 of 3 12/9/2011 8:52AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 013A1.06 Engineered Safety Features Actuation 3.6 3.9 D D D D D D ~ D D D D RWSTievel 013K6.01 Engineered Safety Features Actuation 2.7 3.1 D D D D D ~ D D D D D Sensors and detectors 022K3.02 Containment Cooling 3.o 3.3 D D ~ D D D D D D D D Containment instrumentation readings 025A2.03 Ice Condenser 3.o 3.2 D D D D D D D ~ D D D Opening of ice condenser doors 025A4.02 Ice Condenser 2.7 2.5 D D D D D D D D D ~ D Containment vent fans 026A4.01 Containment Spray 4.5 4.3 D D D D D D D D D ~ D CSS controls 039A3.02 Main and Reheat Steam 3.1 3.5 D D D D D D D D ~ D D Isolation of the MASS 059A1.07 Main Feedwater 2.5 2.6 D D D D D D ~ D D D D Feed Pump speed, including normal control speed for ICS 061K6.02 Auxiliary/Emergency Feedwater 2.6 2.7 D D D D D ~ D D D D D Pumps 062G2.4.1 AC Electrical Distribution 4.6 4.8 D D D D D D D D D D ~ Knowledge of EOP entry conditions and immediate action steps.

0621<2.01 AC Electrical Distribution 3.3 3.4 D ~ D D D D D D D D D Major system loads Page 2 of 3 12/9/2011 8:52 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 063K1.03 DC Electrical Distribution 2.9 3.5 ~ 0 0 0 0 0 0 0 0 0 0 Battery charger and battery 064G2.1.30 Emergency Diesel Generator 4.4 4.0 0 0 0 0 0 0 0 0 0 0 ~ Ability to locate and operate components, including local controls.

073A2.01 Process Radiation Monitoring 2.5 2.9 0 0 0 0 0 0 0 ~ 0 0 0 Erratic or failed power supply 076A3.02 Service Water 3.7 3.7 0 0 0 0 0 0 0 0 ~ 0 0 Emergency heat loads 078A4.01 Instrument Air 3.1 3.1 0 0 0 0 0 0 0 0 0 ~ 0 Pressure gauges 103A2.04 Containment 3.5 3.6 0 0 0 0 0 0 0 ~ 0 0 0 Containment evacuation (including recognition of the alarm)

Page3of3 12/9/2011 8:52 AM

ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 001A1.03 Control Rod Drive 3.6 3.7 0 0 0 0 0 0 ~ 0 0 0 0 S/G level and pressure 002K6.12 Reactor Coolant 3.o 3.5 D D D D D ~ D D D D D Code Safety valves 017A2.02 In-core Temperature Monitor 3.6 4.1 0 0 0 0 0 0 0 ~ 0 D 0 Core damage 027K2.01 Containment Iodine Removal 3.1 3.4 0 ~ D 0 0 0 D 0 0 0 0 Fans 029K4.02 Containment Purge 2.9 3.1 0 0 0 ~ 0 0 0 0 0 0 0 Negative pressure in containment 033K3.03 Spent Fuel Pool Cooling 3.0 3.3 0 0 ~ 0 0 0 0 0 0 0 0 Spent fuel temperature 034A4.01 Fuel Handling Equipment 3.3 3.7 0 0 0 0 0 0 0 0 0 ~ 0 Radiation levels 035A3.01 Steam Generator 4.0 3.9 0 0 0 0 0 0 0 0 ~ 0 0 S/ G water level control 068K5.04 Liquid Radwaste 3.2 3.5 0 0 0 0 ~ 0 0 0 D D 0 Biological hazards of radiation and the resulting goal of A LARA 079K1.01 Station Air 3.o 3.1 ~ 0 0 0 0 0 0 0 D 0 0 lAS Page 1 of 1 12/9/2011 8:52 AM

ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM Es-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO G2.1.29 Conduct of operations 4.1 4.o D D D D D D D D D D ~ Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

G2.1.5 Conduct of operations 2.9 3.9 D D D D D D D D D D ~ Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

G2.1.6 Conduct of operations 3.8 4.8 D D D D D D D D D D ~ Ability to manage the control room crew during plant transients.

G2.2.13 Equipment Control 4.1 4.3 D D D D D D D D D D ~ Knowledge of tagging and clearance procedures.

G2.2.2 Equipment Control 4.6 4.1 D D D D D D D D D D ~ Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

G2.3.5 Radiation Control 2.9 2.9 D D D D D D D D D D ~ Ability to use radiation monitoring systems G2.3.7 Radiation Control 3.5 3.6 D D D D D D D D D D ~ Ability to comply with radiation work permit requirements during normal or abnormal conditions G2.4.28 Emergency Procedures/Plans 3.2 4.1 D D D D D D D D D D ~ Knowledge of procedures relating to emergency response to sabotage.

G2.4.34 Emergency Procedures/Plans 4.2 4.1 D D D D D D D D D D ~ Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects G2.4.45 Emergency Procedures/Plans 4.1 4.3 D D D D D D D D D D ~ Ability to prioritize and interpret the significance of each annunciator or alarm.

Page 1 of 1 12/9/2011 8:52 AM

ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 015AA2.1 0 RCP Malfunctions I 4 3.7 3.7 D D D D D D D ~ D D D When to secure RCPs on loss of cooling or seal injection 022AA2.04 Loss of Rx Coolant Makeup I 2 2.9 3.8 D D D D D D D ~ D D D How long PZR level can be maintained within limits 029EG2.2.3 ATWS 11 3.8 3.9 D D D D D D D D D D ~ (multi-unit license) Knowledge of the design, procedural and operational differences between units.

038EA2.17 Steam Gen. Tube Rupture I 3 3.8 4.4 D D D D D D D ~ D D D RCP restart criteria 058AG2.4.47 Loss of DC Power I 6 4.2 4.2 D D D D D D D D D D ~ Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

we11 EG2.2.3 Loss of Emergency Coolant Recirc. 14 3.1 4.2 D D D D D D D D D D ~ Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations Page 1 of 1 12/912011 8:52AM

ES-401, REV 9 SRO T1 G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 037AA2.03 Steam Generator Tube Leak /3 3.4 3.9 D D D D D D D ~ D D D That the expected indication on main steam lines from the S/Gs should show increasing radiation levels 059AA2.03 Accidental Uquid RadWaste Rei. /9 3.1 3.6 D D D D D D D ~ D D D Failure modes, their symptoms and the causes of misleading indications on a radioactive-liquid monitor 067AG2.4.18 Plant Fire On-site I 8 3.3 4.o D D D D D D D D D D ~ Knowledge of the specific bases for EOPs.

we15EG2.4.30 Containment Flooding /5 2.7 4.1 D D D D D D D D D D ~ Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

Page 1 of 1 12/9/2011 8:52AM

ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 004A2.27 Chemical and Volume Control 3.5 4.2 0 0 0 0 0 0 0 ~ 0 0 0 Improper RWST boron concentration 006A2.02 Emergency Core Cooling 3.9 4.3 D D D D D D D ~ D D D Loss of flow path 022A2.02 Containment Cooling 2.3 2.6 0 0 0 0 D 0 0 ~ 0 0 0 Fan motor vibration 059G2.1.20 Main Feedwater 4.6 4.6 0 0 D 0 0 0 0 0 0 0 ~ Ability to execute procedure steps.

064G2.4.45 Emergency Diesel Generator 4.1 4.3 0 0 0 0 0 0 0 0 0 0 ~ Ability to prioritize and interpret the significance of each annunciator or alarm.

Page 1 of 1 12/9/2011 8:52 AM

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO 001 A2.11 Control Rod Drive 4.4 4.7 0 0 0 0 0 0 0 ~ 0 0 0 Situations requiring a reactor trip 015G2.4.20 Nuclear Instrumentation 3.8 4.3 0 0 0 0 0 0 0 0 0 0 ~ Knowledge of operational implications of EOP warnings, cautions and notes.

027A2.01 Containment Iodine Removal 3.0 3.3 0 0 0 0 0 0 0 ~ 0 0 0 High temperature in the filter system Page 1 of 1 1219/2011 8:52AM

ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SAO G2.1.34 Conduct of operations 2.7 3.5 DDDDDDDDDD~ Knowledge of primary and secondary chemistry limits G2.1.45 Conduct of operations 4.3 4.3 DDDDDDDDDD~ Ability to identify and interpret diverse indications to validate the response of another indication G2.2.37 Equipment Control 3.6 4.6 DDDDDDDDDD~ Ability to determine operability and/or availability of safety related equipment G2.3.12 Radiation Control 3.2 3.7 DDDDDDDDDD~ Knowledge of radiological safety principles pertaining to licensed operator duties G2.3.14 Radiation Control 3.4 3.8 DDDDDDDDDD~ Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities G2.4.43 Emergency Procedures/Plans 3.2 3.8 DDDDDDDDDD~ Knowledge of emergency communications systems and techniques.

G2.4.44 Emergency Procedures/Plans 2.4 4.4 DDDDDDDDDD~ Knowledge of emergency plan protective action recommendations.

Page 1 of 1 12/9/2011 8:52 AM

-WRITTEN OAJL'/ -

ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: McGuire Date of Examination: August 2012 Initials Item Task Description

1. a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401 .

w R b. Assess whether the outline was systematically and randomly prepared in accordance with I Section D.1 of ES-401 and whether aiii<JA categories are appropriately sampled.

T T c. Assess whether the outline v ...... - ..............~ ...... ~ any systems, evolutions, or generic topics.

E N d. Assess whether the justifications for deselected or rejected I<JA statements are appropriate.

2.

s I

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major transients. 7

\

M b. Assess whether there are enough scenario sets (and spares) to test the projected number u and mix of applicants in accordance with the expected crew composition and rotation schedule L without compromising exam integrity, and ensure that each applicant can be tested using A at least one new or. "'l:l' """"' ~*:r. modified scenario, that no scenarios are duplicated T from the applicants' audit test(s), and that scenarios will not be ' .................... on "' onC><>nl oont days.

0 R

3.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.
a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

\ I

!Vv\~

(1) the outline(s) contain(s) the required number of control room and in-plant tasks w distributed among the safety functions as specified on the form I (2) task repetition from the last two NRC examinations is within the limits specified on the form T (3) no tasks are duplicated from the applicants' audit test(s)

(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form.

  • I \
b. Verify that the administrative outline meets the criteria specified on Form ES-301 -1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is ~"'"'~"'u from the last two NRC ~"""~ 'l:l examinations I

Determine if there are enough different outlines to test the projected number and mix of ......,;:;;,.., ..~ and ensure that no items are rl**nl i,.,torl on ~***h~Anuent days. \

Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the "' exam sections.

Assess whether the 10 CFR 55.41/43 and 55.45 sampling is ........... v ....""'"'* ~~f21~ 1 t---E-ns-u-re-th_a_t_I<J_A_im-.....-v,..-""-""'-rati-ngs--(ex-cep-tfo_r__lii_C1'-----"_..-..,_....-',,'-'-,..,_rn....:....*es-_La_r_e_a_t-le-a-st-2-.5-.---..~....lfl'~, _"j Check for rl*' and overlap among exam sections. lf'.I[A ~~ A. ;j

~~C~h~ec~k_!!th~e~e~n~ti~re~e~x~a:!!m~f~o!..rb~a~la~n~c~e~o~f~~~,:...._-----------------!1 'JtM lJ Assess whether the exam fits the ........... job level (RO or SRO). --~ AfA ~ V

j*

v ........ , ...

a. Author
b. Facility Reviewer (*)

!fhru"CL lflti!k~"1~7m"£.H.a/f//,4 "' ~ /1 /l M

c. NRC Chief Examiner(#) t! ..... -
  • W l -* A. I ..!::1'-, .... r,f.Mt.,...l~ '*-
d. NRC Supervisor ul1 i"r.tt.-1 A- St..., I ~.LII - "1/tolz., rz.

I I I  ? I Note: # Independent NRC reviewer initial items in Column "c"; chief examiner concurrence required.

  • Not appucao1e for NRr.. ,......, examination outlines

- WR.11fEN fVIN t:J/VLY-

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev 032812)

Facility: McGuire Date of Examination: 8/20/12 Examination Level: RO Operating Test Number: N12-1 Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1 .7 (4.4) Ability to evaluate plant performance and Conduct of Operations make operational judgments based on operating characteristics, reactor behavior, D,R and instrument interpretation.

JPM: Calculate QPTR with an Inoperable Power Range Instrument 2.1.25 (3.9) Ability to interpret reference materials, such Conduct of Operations as graphs, curves, tables, etc.

M,R JPM: Calculate Boration Needed for a Specified Rod Change 2.2.12 (3. 7) Knowledge of Surveillance Procedures.

Equipment Control M,P,R JPM: Perform a Manual NC Leakage Calculation 2.3.14 (3.4) Knowledge of radiation or contamination Radiation Control hazards that may arise during normal, abnormal, or emergency conditions or N,R activities.

JPM: Predict Radiation Levels While Responding to a Damaged Spent Fuel Pool NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (0) (S)imulator, (0) or Class(R)oom (4)

(D)irect from bank(~ 3 for ROs; ~ 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank (~ 1) (3)

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

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev 032812)

RO Admin JPM Summary A1a This is a Bank JPM. With the plant at 74% power, the operator will be told that the Unit 1 OAC failed and is not operating, and that PR-41 has been removed from service. The operator will be directed to calculate QPTR in accordance with PT/1 IA/4600/21 A (Loss of Operator Aid Computer while in Mode 1). The operator will be expected to calculate QPTR for the three operable Power Range Detectors, and determine that Technical Specification 3.2.4, Quadrant Power Tilt Ratio, has been exceeded.

A1b This is a modified JPM using Bank JPM OP-MC-JPM-ADM-215 as its basis. The operator will be given a set of initial conditions and told that it is desired to withdrawal the Bank D Control Rods about 45 steps. The Operator will be given the Core Data Book and asked to manually determine the amount of Boric Acid that will be necessary to add, to complete the rod height adjustment.

A2 This is a modified JPM using Bank JPMs ADM-NRC-A2-05 and 12 as its basis.

The operator will be told that Unit 1 is at 100% power, the Unit 1 OAC point M1 L4554 is out of service, and that PT/1/A/4200/040 (Reactor Coolant Leakage Detection) has been completed showing that NCS Leakage is 1.6 gpm. The operator will be given Enclosure 13.2 (NC Leakage Determination Using Manual Calculations) of PT/1/A/4150/001B (Reactor Coolant Leakage Calculation) with the necessary raw data compiled on a Data Sheet; and directed to complete the calculations within the Enclosure. The operator will be expected to complete all calculations, and identify any Technical Specification Limits that have been exceeded. This is a modified version of a similar JPM on NRC Exam N10-1-1.

A3 This is a new JPM. The operator will be given a set of conditions reflecting a damaged and leaking Spent Fuel Pool with a full core off-loaded, where attempts of makeup have failed, but are expected to be successful within four hours. The operator will also be given a present Spent Fuel Pool leak rate and level. The operator will be directed to refer to Enclosure 13 (Spent Fuel Pool Radiation Level vs. Water level Above Fuel) of AP/1/A/5500/41 (Loss of Spent Fuel Pool Cooling or Level), and determine the expected radiation levels one hour, two hours, three hours and four hours from now, based on the last known leak rate.

The operator will be expected to determine the expected Dose rate within +/-50%.

NUREG-1 021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev 032812)

Facility: McGuire Date of Examination: 8/2012 Examination Level: SAO Operating Test Number: N12-1 Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1.20 (4.6) Ability to interpret and execute procedure Conduct of Operations steps.

M,R JPM: Perform a Manual Shutdown Margin Calculation 2.1.25 (4.2) Ability to interpret reference materials, such Conduct of Operations as graphs, curves, tables, etc.

M,R JPM: Manual AFD Calculation 2.2.12 (4.1) Knowledge of Surveillance Procedures.

Equipment Control M,P,R JPM: Perform/Review a Manual NC Leakage Calculation 2.3.14 (3.8) Knowledge of radiation or contamination Radiation Control hazards that may arise during normal, abnormal, or emergency conditions or activities.

N,R JPM: Calculate Spent Fuel Pool Boiloff rate and predict when Spent Fuel Pool Radiation levels will exceed 1Rem/Hour 2.4.30 (4.1) Knowledge of events related to system Emergency operation/status that must be reported to Procedures/Plan internal organizations or external agencies, M,R such as the State, the NRC, or the transmission system operator.

JPM: Determine Reportability Requirements NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (0) (S)imulator, (0) or Class(R)oom (5)

(D)irect from bank (~ 3 for ROs; ~ 4 for SROs & RO retakes) (0)

(N)ew or (M)odified from bank ~ 1) (5)

(P)revious 2 exams (~ 1; randomly selected) (0)

NU REG-1021 , Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev 032812)

SRO Admin JPM Summary A1a This is a modified JPM using Bank JPM ADM-NRC-A 1-011 as its basis. The operator will be told that Unit 1 was shutdown to Mode 3, 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ago, that the reactivity computer (REACT) is out of service, and that preparations are being made to commence a plant cooldown to 350°F, and that the OATC has determined that sufficient SDM exists to perform this cooldown. The operator will be directed to perform a Shutdown Margin Calculation per OP/0/A/61 00/006 Enclosure 4.5 (Shutdown Margin - Unit Shutdown, Modes 5, 4, or 3 Without Xenon Credit) to independently verify the conclusions of the OATC, and identify whether or not the cooldown can be initiated under the present conditions. The operator by completing Enclosure 4.5 of OP/1/A/6100/006 (Reactivity Balance Calculation) will determine the Shutdown Margin to be 1579.7+/-5 ppm, and identify that the NC System must be borated to this value before the cooldown can be initiated.

A1b This is a modified JPM using Bank JPM ADM-NRC-A 1-021 as its basis. The operator will be told that Unit 1 is at 100% power, the OAC has been out of service for 30 minutes, that PT/1/A/4600/021 A (Loss of Operator Aid Computer While in Mode 1) is being performed, and that the Main Control Board AFD meters are INOPERABLE. The operator will be given the present current values for the Power Range upper and lower detectors and directed to calculate AFD per PT/1/A/4600/021A (Loss of Operator Aid Computer while in Mode 1), Section 12.10 for current plant conditions and verify that AFD is within the limits specified in the COLR. The operator will be expected to manually calculate AFD, determine that the AFD calculated for N42 and N44 is in excess of the limits allowed by the COLR, and identify all required Technical Specification ACTION.

A2 This is a modified JPM using Bank JPMs ADM-NRC-A2-05 and 12 as its basis.

The operator will be told that Unit 1 is at 100% power, the Unit 1 OAC point M1 L4554 is out of service, and that PT/1/A/4200/040 (Reactor Coolant Leakage Detection) has been completed showing that NCS Leakage is 1.6 gpm. The operator will be given Enclosure 13.2 (NC Leakage Determination Using Manual Calculations) of PT/1/A/4150/001B (Reactor Coolant Leakage Calculation) with the necessary raw data compiled on a Data Sheet; and directed to complete the calculations within the Enclosure. The operator will be expected to complete all calculations in accordance with the provided Key, identify any Technical Specification Limits that have been exceeded, and (SRO aspect) identify with all Technical Specification ACTION.

A3 This is a new JPM. The operator will be given a set of conditions reflecting a loss of level and cooling to the Spent Fuel Pool with a full core off-loaded, where attempts of makeup have failed. The operator will also be given a present Spent Fuel Pool level and time since shutdown. The operator will be directed to use Enclosure 5 (Spent Fuel Pool Boiloff rate) of AP/1/A/5500/41 (Loss of Spent Fuel Pool Cooling or Level), and estimate how fast the Spent Fuel Pool level will go down as the pool boils; and then use this boiloff rate and Enclosure 13 (Spent NUREG-1021, Revision 9

....s ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev 032812)

Fuel Pool Radiation Level vs. Water level Above Fuel) of AP/1/A/5500/41 (Loss of Spent Fuel Pool Cooling or Level}, and predict when the radiation levels in the Spent Fuel Pool area will be greater than 1 Rem/hour if no makeup is added.

The operator will be expected to determine that the boiloff rate is 4.4 inches/hour, and that the expected time that the Dose Rate in the Spent Fuel Area exceeds 1 Rem/hour is 36+/-2 hours.

A4 This is a modified version of a Bank JPM. The operator will be told that Unit 1 is operating at 100% power, while Unit 2 is operating at 30% for a Chemistry hold; and then given a timeline of events that result in both Units entering in Technical Specification 3.0.3. The operator will be told that Unit 1 initiated a shutdown, but that Unit 2 is still holding at 30%; and that now the OSM believes that this event is reportable to the NRC. The operator will be directed to perform the Subsequent Actions of RP/0/A/5700/010 (NRC Immediate Notification Requirements), and to complete any necessary paperwork. The operator will be expected to identify that this condition requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification to the NRC for Units 1 and 2, and to complete Enclosure 4.2 (NRC Event Notification Worksheet) in accordance with the attached Key.

NU REG-1 021 , Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 032912)

Facility: McGuire Date of Examination: 8/2012 Exam Level (circle one): RO (only) I SRO(I) I SRO (U) Operating Test No.: N12-1 Control Room Systems@(8 for RO; 7 for SR0-1; 2 or 3 for SRO-U, including 1 ESF)

Type Code* Safety System I JPM Title Function A. 001 Control Rod Drive System S,N 1 Realign a Misaligned Control Rod B. 045 Main Turbine Generator System S,D,P,A 4S Perform the Main Turbine Overspeed Trip Test

c. 010 Pressurizer Pressure Control System S,D,P,A 3 Remove Pressurizer Heaters from Service D. 026 Containment Spray System S,D,EN,A 5 Align the Containment Spray System During Cold Leg Recirc E. APE 025 Loss of Residual Heat Removal System S,M,L 4P Midloop Makeup to the NCS F. APE 026 Loss of Component Cooling Water S,N,A 8 Respond to High VCT Temperature G. 064 Emergency Diesel Generators S,D,A 6 Perform Diesel Generator Operability Test H. APE 061 ARM System Alarms S,N 7 Control Room Air Intake High Radiation Alarms In-Plant Systems@ (3 for RO; 3 for SR0-1; 3 or 2 for SRO-U)

I. APE 058 Loss of DC Power D,R,E 6 Swap Battery Charger EVCA Power Supply from Unit 1 to Unit 2 J. APE 024 Emergency Boration D,R,E 1 Emergency Borate the NCS Locally Using 1NV-269 K. APE 065 Loss of Instrument Air D,A, E 8 Bypass A, B, and C VI Dryers following a Loss of Instrument Air NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 032912)

@ All RO and SR0-1 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 Criteria for RO I SAO-II SRO-U (A)Iternate path 4-6 (6) 14-6 (6) I 2-3 (2)

(C)ontrol room (D)irect from bank $ 9 (7) I$ 8 (7) I $ 4 (3)

(E)mergency or abnormal in-plant ~ 1 (3) I~ 1 (3) I ~ 1 (2)

(EN)gineered Safety Feature - I - I ~ 1 (1) (Control Room System)

(L)ow-Power I Shutdown ~ 1 (1)1~ 1 (1)1 ~ 1 (1)

(N)ew or (M)odified from bank including 1(A) ~ 2 (4) I ~ 2 (3) I ~ 1 (2)

(P)revious 2 exams $ 3 (2) I $ 3 (2)1 $ 2 (0) (Randomly Selected)

(R)CA ~ 1 (2 )I~ 1 (2) I ~ 1 (2)

(S)imulator JPM Summary JPM A This is a new JPM. The operator will be told that due to a plant transient, Control Rod M-4 is misaligned from its bank by 14 steps, and that Unit 1 has been stabilized at 45%

power for the past 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. The operator will also be told that the crew entered AP/1/A/5500/14 (Rod Control Malfunction), and completed Enclosure 1 (Response to Dropped or Misaligned Rod) through Step 16; and that IAE and RE has determined that rod realignment is permissible by moving the misaligned rod inward. The operator will be directed to realign Control Rod M-4 with Bank D by performing Steps 17-21 of Enclosure 1 (Response to Dropped or Misaligned Rod), of AP/1/A/5500/14 (Rod Control Malfunction). The operator will be expected to initiate this process by completing the actions through step 21.g of Enclosure 1.

JPM B This is a Bank JPM, previously used on N 11-1-1. The operator will be told that Unit 1 is starting up after a refueling outage, that the Turbine/Generator is off line and rolling at 1800 RPM in preparation for performing PT/1 IA/4250/004C (Turbine OPC and Mechanical Overspeed Trip Test), that all prerequisite conditions have been met, that two operators have been stationed at the Turbine as required, and that communications have been established with all involved. The operator will be directed to complete the Turbine OPC and Mechanical Overspeed Trip Test per PT/1/A/4250/004C (Turbine OPC and Mechanical Overspeed Trip Test), starting with Step 12.7. The operator will raise Turbine speed to OPC setpoint, and then raise speed until the Turbine Overspeed trip should be actuated (Alternate Path). The operator is expected to recognize the turbine has failed to trip at the expected setpoint and then manually trip the Turbine.

JPM C This is a Bank JPM, previously used on N10-1-1. The operator will be told that plant power has just been raised to 100% per OP/1/A/6100/003 (Controlling Procedure for Unit Operation). The operator will be directed to remove Pzr Heater Groups A, Band D from service per Enclosure 4.6 (Operation of Pzr Heaters) of OP/1/A/6100/003. The operator will be expected to remove the A, B and D Pzr Heater Groups from service in accordance with the Enclosure. After the Pzr Pressure Master has been placed in MANUAL and its output has been adjusted, the Pzr Variable Heaters (Group C) will fail (Alternate Path). The operator will be required to respond to MCB Annunciator 1AD6/D6 (PZR HTR CONTROLLER TROUBLE), and manually control pressure using NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 032912) the other heater groups. The operator will be expected to place at least one Pzr Heater Group in service in accordance with Step 3.3.1 (or equivalent) of Enclosure 4.6.

JPM D This is a bank JPM updated for recent plant modification/procedure update. The operator will be told that earlier in the shift, the Unit 1 reactor tripped due to a Large Break LOCA inside of containment, that EP/1/A/5000/ES-1 .3 (Transfer To Cold Leg Recirc) has been implemented and completed through step 7, that Containment pressure is approximately 4 psig, and that FWST Level is approximately 80 inches and lowering. The operator will be directed to perform Step 8 of EP/1/A/5000/ES-1.3 (Transfer To Cold Leg Recirc) and align NS for Recirc. The operator will attempt to place the A Train of NS in service manually, discover that it cannot be placed in service (Alternate Path), and then place the B Train in service as an alternative.

JPM E This is a modified version of a Bank JPM. The operator will be told that Unit 1 is in mode 5 during a refueling outage, that AP/1/A/5500/19 (Loss of ND or ND System Leakage) has been entered and completed up to Step 5; and that the CAS is considering makeup options for the NCS. The operator will be directed to use Enclosure 5 (Makeup Via NV Pumps Through S/1 Flowpath) of AP/1/A/5500/19 (Loss of ND or ND System leakage) to maintain NCS level greater than 10 inches. The operator will be expected to initiate make up to the NCS using Enclosure 5 of AP/1/A/5500/19.

JPM F This is a new JPM. The operator will told that Unit 1 was at 100% power when a leak developed in the KC System, that the crew entered AP/1/A/5500/21 (Loss of KC or KC System Leakage) and has completed the actions through Step 12. They will be told that MCB Annunciator 1AD-7, D1, VCT HI TEMP, has just alarmed, making Foldout Page item #5 applicable. The operator will be directed to perform the actions of Enclosure 4.6 of AP/1/A/5500/21 (Loss of KC or KC System Leakage), while the crew continues with the AOP. The operator will be expected to isolate Letdown, and attempt to start the PD Pump. Upon PD Pump start, the motor will fail (Alternate Path), and the operator will ultimately align the suction of the NV Pumps to the FWST.

JPM G This is a bank JPM. The operator will be told that Unit 1 is operating at 100% power, that a monthly test of the 1B Emergency Diesel Generator is required, and that the System Engineer wants to start and stop the Diesel from the Control Room. The operator will be directed to conduct a Slow Start of the 1B Emergency Diesel Generator using Enclosure-13.1 of PT/1/A/4350/002B, "Diesel Generator 1B Operability Test."

During the performance of Enclosure 13.1, a sudden loss of crankcase vacuum will be indicated together with a loss of engine speed (Alternate Path). The operator will be expected to stop the engine with the normal stop switch.

JPM H This is a new JPM. The operator will told that Units 1 and 2 are operating at 100%

power, that Annunciator 1RAD-2 B2, EMF 43B CR AIR INTAKE B HI RAD, alarmed 45 seconds ago; and that Annunciator 1RAD-1 B2, EMF 43A CR AIR INTAKE A HI RAD, alarmed 15 seconds ago. The operator will be directed to perform the immediate actions associated with the Annunciator Response Procedures for both alarms. The operator will be expected to determine that the Unit 2 intake presents a greater threat than Unit 1, and align the VC inlet to take suction on Unit 1 only; and then pressurize the Control Room from the A Train Outside Air Pressure Fan.

JPM I This is a Bank JPM. The operator will be told that Unit 1 has just experienced a Loss of Offsite Power, that the 1A D/G will not start, and that 1ETA is de-energized.

NUREG-1021 , Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 032912)

AP/1/A/5500/07, "Loss of Electrical Power," Case 1 has been implemented. The operator will be directed to swap power supplies to the EVCA Battery Charger from Unit 1 to Unit 2 in accordance with Enclosure 22 (Swapping Battery Charger Power Supplies) of AP/1/A/5500/07, (loss of Electrical Power). The operator will be expected to place Battery Charger EVCA in service with power being supplied from Unit 2 within 20 minutes of dispatch. This is a Time Critical JPM.

JPM J This is a Bank JPM. The operator will be told that Unit 2 was at 100% power when a Boron dilution event occurred, that AP/2/A/5500/38 (Emergency Boration) was entered, and that while attempting to open 2NV-265B (Boric Acid To NV Pumps), the RO discovered that 2NV-265B was de-energized. The operator will be directed to emergency borate the NC System by performing Step 12.d RNO of AP/2/A/5500/38 (Emergency Boration and Response to Inadvertent Dilution). The operator will be expected to locate and open 2NV-269 (simulated) within ten (1 0) minutes of dispatch.

This is a Time Critical JPM.

JPM K This is a Bank JPM. The operator will be told that Unit 1 is operating at 100% power, that AP/1/A/5500/22 (loss of VI) has been implemented due to decreasing VI Pressure, that VI Pressure is holding at 75 psig but the leak has not been isolated, and that several other operators have been dispatched for various tasks in response to the loss of VI.

The operator will be directed to bypass the VI dryers and isolate VS per Enclosure 5 (VI Dryer and VI to VS System Isolation) of AP/1/A/5500/22 (loss of VI) using the copy of the procedure located beside Service Building Lube Oil Station door. When the operator checks the position of the Dryer Bypass valve, it will be determined that the valve has failed closed (Alternate Path). The operator will be expected to fail open the Dryer Bypass Valve when it is recognized that the valve has failed closed, and then close the A, B, and C VI Dryers Inlet and Outlet Isolation valves.

NUREG-1 021, Revision 9

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