ML110980758

Initial Exam 2011-301 Draft RO Written Exam

ML110980758
Person / Time
Site:	Browns Ferry
Issue date:	04/04/2011
From:	NRC/RGN-II/DRS/OLB
To:	Tennessee Valley Authority
References
50-259/11-301, 50-260/11-301, 50-296/11-301
Download: ML110980758 (458)

Text

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Exami*nation Outline r

.r oss-reference: Level RO lete Loss of Forced Core Flow Circulation/i &

Tier # 1 102 i8TO411 Group 1 as theYFaP,tIRCULATlON: K/A 5001AK1.02 COR

• Power!fI o w distribution Importance Rating Proposed Question: #I Unit 1 is at 100% Reactor Power AND Core Flow is 92%. A trip of IA Recirc Pump results in Operation in Region II of the Core Power to Flow Map.

Which ONE of the following completes the statement below?

The required action(s) in accordance with 1-AOl-68-IA, Recirc Pump Trip I Core Flow Decrease, is (are) to IMMEDIATELY A. insert a Manual Reactor Scram B. raise Core Flow until Region II of the Power to Flow Map is exited C. insert Control Rods until Region II of the Power to Flow Map is exited D. insert Control Rods until Load Line is < 95.2%; then, raise Core Flow to > 45%

Proposed Answer: D Explanation A INCORRECT: Plausible in that IF both Recirc Pumps are tripped in Modes 1 (Optional): or 2, THEN 1-AOI-68-1A requires the Reactor to be Scrammed.

B INCORRECT: Plausible in that immediately raising core flow would be an expeditious method to exit instability regions. If load line was less than 95.2% following the Recirc Pump trip, this would be the correct answer.

C INCORRECT: Plausible in that Control Rod are required to be immediately inserted if in Region I or II but the crew will stop inserting Control Rods when Load Line is < 95.2%. That is, Control Rod insertion will stop prior to exiting the Region and raising core flow will complete the exit from Region II. If core flow was greater than 45% following the Recirc Pump Trip, this would be the correct answer.

D CORRECT: In accordance with 1-AOI-68-1A, IF Region I or II of the Power to Flow Map is entered due to a trip of a Recirc Pump, THEN IMMEDIATELY take actions to insert control rods to less than 95.2%

loadline. Then, RAISE core flow to greater than 45% in accordance withi 01-68.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because it tests candidates knowledge of operational implications of Reactor Power I Flow distribution with a partial loss of core circulation as a result of a Recirc Pump trip.

Question Cognitive Level:

Question rated as C/A because Candidates must process multiple pieces of data to determine correct actions in accordance with 1-AOI-68-1A. Candidate must recognize that with core flow of 92% at Reactor Power of 100% that Load Line is greater 100% and will remain greater than 100% following the Recirc Pump trip. Also, must recognize that following the trip, Core Flow will be less than 45% requiring increase in core flow also.

Technical Reference(s): 1-AOl-68-1A Rev 3 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.007 V.8.28 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Recirc Pump Trip/Core Flow Decrease 1-AOl-68-IA Unit I OPRMs Operable Rev. OOG3 Page 7 ot 12 4.2 Subsequent Actions Icontinued)

NOTE

1) Step 42(2] through Step 42[17.3J apply to any core flow Iowenng event 2> Power To Flow Map is maintained in 0-TI-248, Station Reactor Engineer and on ICS.

[2] IF a single Recirc Pump has tripped, THEN CLOSE tripped Recirc Pump discharge valve. []

[3] IF Region I or II of the Power to Flow Map is entered, THEN (OthenMse N/A)

IMMEDIATELY take actions to insert control reds to less than r 95.2% loadl[ne AND REFER TO 0-Tl-464, Reactivity Control Plan Development and Implementation. D

[4] RAISE core flow to greater than 45% in accordance with l-Ol-fi& I]

[5] INSERT control rods to exit regions if NOT already exited AND REFER TO 0-Tl-464, ReactM Control Plan Development and Implementation. U NOTE The remaining subsequent action steps apply to a single Reactor Recirc Pump trip.

6] MAINTAIN operating Recirc pump flow less than 46,600 gpm in accordance with 1-01-68. []

[7] !NERIC) WHEN plant conditions allow, THEN, (Otherwise NIA)

MAINTAIN operating jet pump loop flow greater than 41 x 10 Ibm/hr (1-Fl-68-.46 or 1-Fl-68-48). GESIL U

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT BFN Recirc Pump Trip(Core Flow Decrease 1-AOl-68-IA Unit I OPRMs Operable Rev. 0003 Page 6 of 12 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions

[lj IF both Recirc Pumps are tiipped in modes I or 2, THEN (Otherwise N/A)

[Ill SCRAM the Reactor U CAUTION

[NEW Failure to restart Reactor Recirculalion pumps in a timely manner may result in exceeding the differential temperature limit for pump start and subsequently require plant depressunzation to avoid exceeding pressure-temperature limits for the reactor vessel. (SER 3-OO5(

[I 2j RESTART affected Reactor Recirculation pumps Refer to 1-01-68 Section 80. [1

[1 -31 IF the T between the Rx vessel bottom head temperature and the moderator temperature precludes restart of a Recirc pump, OR forced Recirculation 110w CANNOT be established for any reason, THEN (Otherwise NA)

[1 3.i} INITIATE a plant cooldown to prevent exceeding the pressure limit for the Rx vessel bottom head temperature indicated on REACTOR VESSEL METAL TEMPERATURE, 1TR-56-4 pt. 10 (Panel 1-9-47) and based on Tech Specs Figure 3A9-1. []

[1 32] TN FORM the Unit Supervisor, Tech Spec 34 I requires the Reactor be placed in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. REFER TO 1-GOl-100-12A and Tech Specs &41.B. U]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILiTY SUPPORT If Load Line was < 952% following the Recirc Pump Trip, Distractor B would be the correct answer.

211 2 3* 3 4 4 U 3 &* 1* 1$) flfl

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILiTY SUPPORT If Core Flow was >45% following the Recire Pump Trip, Distractor C would be the correct answer.

Si S iS 15 O t4 55 lii 555 flU 115 PUICIWI RARJi CPUftI

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level 295003 Partial or Complete Loss of A.C. Power /6 Tier# RO!

G2.4.6 (IOCFR 55.41.10)

Knowledge of EOP mitigation strategies. Group#

KIA# 295003 G2.4.6 Importance Rating 3.7 Proposed Question: #2 A leak in the Unit 1 Drywell results in the following conditions:

• Drywell Temperature is 170° F and rising

• A Lockout occurs on 4kV Shutdown Board C

• Reactor Level is (+) 10 inches and stable

• Suppression Pool Level is 15 feet Which ONE of the following completes the statements below?

In accordance with 1-EOl-2, Primary Containment Control, Drywell Spray must be initiated before MAXIMUM Drywell Temperature of _(1)_. Assuming no manual electric board transfers are performed, RHR _(2) is (are) available for Drywell Spray from the control room.

A. (1)200°F (2) Loop I ONLY B. (1)200°F (2) Loop I AND Loop II C. (1)280°F (2) Loop I ONLY D. (1)280°F (2) Loop I AND Loop II Proposed Answer: C Explanation A INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See (Optional): Explanation C.

B INCORRECT: Part 1 incorrect Plausible in that Drywell Temperature of 200° F is a recognizable value of 1-EOI-2, Drywell Temp Leg requiring entry into EOI-1. Part 2 incorrect Plausible in that Unit 2 480 V Shutdown Board B is supplied from 4 kV SID Board D. On Unit 2 this would be the correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet C CORRECT: Part 1 correct 1-EOl-2 directs Drywell Spray prior to Drywell Temp of 2800 F. Part 2 correct Loop II Drywell Spray valves are powered from 480 RMOV Board B which is powered from 480 V S/D Board B. This Board is powered from 4 kV S/D Board C on Unit 1 which is locked out.

Although one pump is available on Loop 2, Spray Valves can not be opened from the control room.

D IN CORRECT: Part 1 correct See Explanation C. Part 2 incorrect See Explanation B.

KA Justification:

The KA is met because question tests knowledge of EOl mitigation strategies with partial loss of AC Power.

Question Cognitive Level:

This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Candidate must determine effect of a Lockout on 4kV Shutdown Board C on ability to Spray the Drywell.

Technical Reference(s): OPL171.036 Rev. 12/ 1-EOl-2 Rev. 1 (Attach if not previously provided)

OPL171.044 Rev. 17 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .044 V.B.19 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.044 Revision 17 Page 27 of 146 INSTRUCTOR NOTES (2) Tube side fluid RHRSW © 4500 gpm

d. Tube side pressure should be kept higher than shefl side if possible to minimize the potential leakage of RHR water to the RHRSW. This limits the potential for radioactive discharge to the environment. RHRSW discharge is monitored for radioactivity prior to discharge to the river. No automatic actions occur due to a high radioactivity condition in the RHRSW.

.4. Valves Obj V.B.8

a. Power supplies All RHR motor-operated valves

- Obj. V.E.6 F are powered from the 480V Reactor MDV Boards except as noted. The Reactor MDV Board power supplies are as follows. Note divisional separation maintained.

480 RMOV NORMAL DIV ALT POWER DIV VALVES BOARD POWER A 480V S/D A I 480V 5/0 B H RHR Sys I valves (except as noted)

B 480V SJD B H 480V Sf0 A I RHR Sys II valves (except as noted)

C 480V Sf0 B H 480V Sf0 A I none D ON MG Set I DA MG Set H 74-7 & 53 E EN MG Set H EN MG Set I 74-30 & 67 Unit 1 does not have RMOV Bd 0 or E.

The loads on D Bd are fed from 1A Bd and E are fed from lB Bd Outboard Shutdown Cooling Isolation Valve Obj. V.B.8 FCV-74-47 is powered from 250 VDC MDV Board A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .044 Revision 17 Appendix C Page 97 of 146 VALVE# VALVE NAME INTERLOCKS

. NOT a throttle valve normally closed

. Cannot open valve if Outboard Injection is open and Rx pressure is >4504

. Auto opens on [PCI initiation signal when Rx pressure is <4504. Remains open until LPCI initiation signal is clear and reset

. Auto close if Inboard and Outboard SDC Isolation Valves open and Group 2

[PCI Inboard isolation signal received. Seals in and must be reset with SOC Isolation reset

/ pushbutton. Valve will not open on LPCI initiation signal until reset.

Injection

. NORMAL/EMERGENCY switch in EMERGENCY bypasses >4504 and Outboard Injection valve open interlock Also prevents auto open and dose signal from logic and allows valve operation ONLY at breaker. Applies to 2/3-74-53 and 1-74-67.

. EMERGENCY OPEN switch bypasses all interlocks. Applies to 213-74-53 and 1-74-67.

e No auto open logic

. Cannot open Inboard valve nonraIIy unless Outboard valve is fully closed.

. Cannot open Outboard valve normally unless Inboard valve is fully closed.

e Auto closed on LPCI initiation signal 60161 Drywell

• The Sup. Pool/Chamber Isol. valve and LPCI iniliation signal interlocks can be (Containment) bypassed if Rx level >-i 83 AND Drywell pressure is >1.96 psig AND LPCI J

7

/

74 Spray initiation signal present AND CONTAINMENT SPRAY OVERRIDE switch is in SELECT

• Fix level and [PCI initiation signal can be overridden with 2/3 CORE HEIGHT OVERRIDE switch

• NORMAL/EMERGENCY switch in EMERGENCY bypasses all interlocks and allows valve operation ONLY at breaker. Applies to 213-74-60 and 1-74-74.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .044 Revision 17 Appendix D Page 104 of 146 TP 5 RHR SYSTEM SUPPRESSION POOL SPRAY FLOW DIAGRAM UNIT 2 SYSTEM II

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.036 Revision 12 Page 34 of 60

d. Synchronizing System (1) AU four breakers feeding the unit 1/2 shutdown boards require the use of synchroscope to parallel supplies or perform manual transfer.

(2) The SYNC switch must be on to complete the closing circuit for any board feeder unless the Board is dead as sensed by the Boards residual voltage relay.

I 48OVAC Standby Distribution Substations 480V Shutdown Boards

a. Each unit has two 480V Shutdown Boards, A Obj. V.B.6.e and B. Their nonnal and alternate power Obj V.D.5 supplies are from their associated 4kV Obj. V.D.6.e Shutdown Boards, as follows: Obj. V.C.1.e 480V Board 4kV Board Obj. V.B.6.f Obj. V.C,1,f Ui/U3 Obj. V.D.6.f A Normal A B Alternate B C B Normal r CD Alternate B C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17I .036 Revision 12 Page 35 of 60 Li All transfers are manual. The Board Obj. .B.8.e 1

\

may be transferred from the Control Obj. V.C.2,e Room by operating the transfer selector Obj. V.D.8.e switch on panel 9-8. Manual transfer at Obj. V.B.8.f the Shutdown Board is accomplished Obj. V.C.2.f by (1) placing the normallemergency Obj. \!.D.8.f switches (both normal and alternate breakers) in EMERGENCY, (2) placing the altemate breaker control switch in CLOSE and holding until (3) the normal breaker control switch is operated to TRIP. After the transfer operation, the normal/emergency switches should be returned to NORMAL so the breakers can be controlled from the Control Room.

c. The 480V Shutdown Boards feed Examples: SLC, safety-related loads, either directly or RWCU, RBCCW, via feeder breakers to MCC boards. (In & P20 general, motors rated between 40 and 200 hp are sewed directly.)

d. Supply breakers are provided with relay overcurrent protection which will trip and lockout the associated breaker and lockout its altemate.

2. 48W Diesel Auxiliary Boards

a. Diesel Auxiliary Boards A, B, 3EA, and Obj V.D.5 3EB principally serve loads associated with the operation of the diesel generators. Other essential small loads are also sewed from these boards.

Loss of any single diesel auxiliary board will not negate the effectiveness of standby core cooling. (Standby Gas Treatment System Trains A and B are served by Diesel Auxiliary Boards A and B. Train C is served by the 480V Standby Gas Treatment Board, which is connected through a transformer to 4kV Shutdown Board 3ED.)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DPL171 .036 Revision 12 Page 37 of 60

4. 480V Reactor MDV Boards Unit 2 Reactor MDV Boards A, B, C, 0, and E Dbj. V.B.6.g are discussed hera Unit 3 Boards are similar. Dbj. V.C.I.g Valves on D & E boards, on Unit I. were Dbj, V.D.6.g moved to A & B boards. Unit I does not have UNIT D & E boards. DIFFERENCE

a. Reactor MDV Boards serve the smaller Dbj V.0.5 480V loads that are important to plant safety. Each MDV board has two incoming sources, one from each 480V shutdown board. Reactor MDV Boards A and 0 feed normally from 480V Shutdown Board A and alternately from 480V Shutdown Board B The normal supply for Reactor MDV Boards B, C, and E is 480V Shutdown Board B with A being the alternate.

b. Boards D and E, the LPCI Valve Examples:

Boards, are fed through motor- Recirculation generator sets for both their normal and discharge valves, alternate supplies. Unit 1 LPCI MG sets LPCI inboard have been removed. Loads that were injection valves, &

on U-i DIE board are now on A & B RHR mm-flow valves boards (Unit difference)

IL Boards A, B, and C have manual Dbj. V.B.8.g transfer of power supplies. Boards 0 Dbj. V.C.2.g and E transfer automatically from Dbj, V.D.8.g normal to alternate on undervoltage; transfer back is manual.

LER 2-85-007

d. Selected feeder breakers have normallemergency selector switches to allow local operation of the associated component.

5. 480 volt board indications and controls

a. Panel 9-8 indications (I) 480V Shutdown Bd. B voltage (2) 480V Unit Boards voltage and amperage

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet L

/

L L

PRIMARY CONTAINMENT CONTROL urr i BROWNS FERRY NUCLEAR PLANT I

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet L

L

1. 2 PLP H1ELT L

L#aTEY 4J WPX 11 L

1o PRIMARY CONTAINMENT CONTROL UNIF I BROWNS FERRY NUCLEAR PLANT EV1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL17l .036 Revision 12 Page 34 of 60

d. Synchronizing System (1) All four breakers feeding the unit 1/2 shutdown boards require the use of synchroscope to parallel supplies or perform manual transfer (2) The SYNC switch must be on to complete the closing circuit for any board feeder unless the Board is dead as sensed by the Boards residual voltage relay.

J. 4BOVAC Standby Distribution Substations

1. 460V Shutdown Boards

a. Each unit has two 480V Shutdown Boards, A Obj. V.B.6.e and B. Their normal and alternate power Obj V.D.5 supplies are from their associated 4kV Obj. V.D.6.e Shutdown Boards, as follows: Obj. V.C.1,e 480V Board 4kV Board Obj. V.B.6.f Obj. V.C.1.f Ui/U3 U2 Obj. VD.6.f A Normal A B j Alternate B C A

B Normal CD p Alternate B C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT L

L A

j tWFT-6 L

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO 295004 Partial or Total Loss of DC Pwr / 6 Tier # 1 AAI 03 (1 OCFR 55 41 7)  :

Ability to operate and/or monitor the following as they apply to Group # 1 PARTIAL OR COMPLETE LOSS OF D.C. POWER: K/A # 295004AA1 .C

• A.C. electrical distribution .

Importance Rating 3.4 Proposed Question: #3 Unit 2 was operating at 100% Reactor Power.

A ground AND subsequent fire in Shutdown Board 250V DC Distribution Panel SB-B resulted in de-energization of the SB-B panel AND trip of 4kV Shutdown Board B Normal Feeder Breaker.

Which ONE of the following completes the statements below?

480V Shutdown Board 2B is (1)_.

4kV Shutdown Board B (2) automatically transfer to its alternate source.

A. (1)energized (2) will B. (1) de-energized (2) will C. (1) energized (2) will NOT D. (1) de-energized (2) will NOT Proposed Answer: C Explanation A INCORRECT: Part 1 correct See explanation C. Part 2 incorrect See (Optional): explanation B.

B INCORRECT: Part I incorrect 480v Shutdown Board 2B remains energized with the loss of 4kV Shutdown Board B. Plausibility based on misconception 480v Shutdown Board B normal power supply would be from 4kV Shutdown Board B. If this was Unit 1 480 V and 4Kv A Shutdown Boards, this would be the correct answer. Part 2 incorrect Each Shutdown Battery system supplies its respective 4KV Shutdown Board and 480V Shutdown Board. All control power transfers are manual. Plausible in that if control power transfer is automatic as board power supply is or control power was not from SB-B DC Distribution Panel, this would be the correct answer

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet c CORRECT: Part 1 correct 480v Shutdown Board 2B remains energized with the loss of 4kV Shutdown Board B. 4kV Shutdown Board D is the normal feeder to the 480v SID Bd 2B. Part 2 correct Each Shutdown Battery system supplies its respective 4KV Shutdown Board and 480V Shutdown Board. All control power transfers are manual. With the loss of control power, normal automatic transfer to alternate power supply will not occur.

D INCORRECT: Part 1 incorrect See explanation A. Part 2 correct See explanation D.

KA Justification:

The KA is met because to successfully answer this question, candidate must recognize the impact of partial loss of DC (SB-B Distribution Panel) will have on control power to 4 kV Shutdown Board B and the impact of loss of 4kV Shutdown Board B will have on 480v Shutdown Board 2B.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71 .036 Rev 12 (Attach if not previously provided)

OPL171.037 Rev 12 0-Ol-57B Rev 189 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .037 V.B.1 (As available)

OPL1 71 .036 V.B.6/8 Question Source:

Mod ified Bank # BEN 1006 #3 (Note changes or attach parent)

Question History:

Last NRC Exam Browns Ferry 2010 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Corn ments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .036 Revision 12 Page 35 of 60

b. All transfers are manual. The Board Obj. \1.B.8.e may be transferred from the Control Obj. V.C.Ze Room by operating the transfer selector Obj. V.D.8.e switch on panel 9-8. Manual transfer at Obj, \.B.8.f the Shutdown Board is accomplished Obj. V.C.2,f by (1) placing the normal/emergency Obj. V.D.8.f switches (both normal and alternate breakers) in EMERGENCY (2) placing the alternate breaker control switch in CLOSE and holding until (3) the normal breaker control switch is operated to TRIP. After the transfer operation, the normal/emergency switches should be returned to NORMAL so the breakers can be controlled from the Control Room.

c. The 480V Shutdown Boards feed Examples: SLC, safety-related loads, either directly or RWCU, RBCCW, via feeder breakers to MCC boards. (In & FPC general, motors rated between 40 and 200 hp are served directly.)

d. Supply breakers are provided with relay overcurrent protection which will trip and lockout the associated breaker and lockout its alternate.

2. 480/ Diesel Auxiliary Boards

a. Diesel Auxiliary Boards A, B, 3EA, and Obj V.D.5 3EB principally serve loads associated with the operation of the diesel generators. Other essential small loads are also served from these boards.

Loss of any single diesel auxiliary board will not negate the effectiveness of standby core cooling. (Standby Gas Treatment System Trains A and B are served by Diesel Auxiliary Boards A and B. Train C is served by the 480V Standby Gas Treatment Board, which is connected through a transformer to 4kV Shutdown Board 3ED.)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL1 71 .037 Rev 12 ci. Distribution Each Shutdown Battery system supplies its respective 4K\/ and 480V Shutdown Board All control power transfers are manual.

BFN 480V!240V AC Electrical System 0-Ol-57B Unit U Rév 0189 Page 106 of 112 Illustration I (Page 7 of 9)

Auxiliary Power Supplies and Bus Transfer ITEM BOARD AND(OR MAIN BUS NORMAL ALTERNATE I ALTERNATE 2 REMARKS 12 480V Turbine Building Vent Boards A. Board A (Unit 12,3) 480V Unit 480V Common Automatic transfer from normal to alternate source is initiated Board A 80 1 (Unit 1 by time-undervoltage on the normal source. Return to normal (Unit 1,2,3> only) 480-V source is automatic upon return of voltage to normal source.

Corn. 80 3 The normally closed, manually operated bus tie breaker (Unit 2 and 3) provides for maintenance on one bus section while keeping B. Board B (Unit 1,2,3) 480V Unit 480V Common the other bus section energized and in operation.

Board B Board 2 (Unit 1,2,3>

13 480V Shutdown Boards A. Unit 1, 480V Shutdown 4kV Shutdown 4kV Shutdown Transfer from normal to alternate source is manual.

80 1A Board A Board B Interlocking is provided to prevent manually transferring to a B. Unit I 480V Shutdown 4kV Shutdown 4kV Shutdown faulted board and to prevent paralleling two sources. 480V 80 18 So 4 C Load Shed Relay Time Delay Setting is set at 1.8 secs per Bo 8 DCN-W14030.

C. Unit 2, 480V Shutdown 4kV Board B 4kV Shutdown 8024 Board C

0. Unit 2, 480V Shutdown 4kV Shutdown 4kV Shutdown 8028 Board 0 Board C

8. Unit 3, 480V Shutdown 4kV Shutdown 4kV Shutdown 80 3A Board 3EA Board 388 F. Unit 3, 480V Shutdown 4kV Shutdown 4kV Shutdown 8038 Board 3EC Board 388

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN 480V!240V AC Electrical System 0.Ol.57B Unit 0 RevS 0189 Pagelo6of 112 IllustratIon I (Page 7 of 9)

Auxiliary Power Supplies and aus Transfer ITEM BOARD AND!OR MAIN BUS NORMAL ALTERNATE I ALTERNATE 2 REMARKS 12 480V Turbine Building Vent Boards A. Board A (Unit 1,2,3) 430V Unit 480V Common Automatic transfer from normal to alternate source is initiated Board A BD 1 (Unit 1 by time-undervollage on the normal source. Return to normal (Unit 12.3) only) 480-V source is automatic upon return of voltage to normal source.

Corn. BE) 3 The normally closed, manually operated bus tie breaker (Unit 2 and 3) provides for maintenance on one bus section while keeping B. Board B (Unit 1,2,3) 480V Unit 480V Common the other bus section energized and in operation.

Board B Board 2 (Unit 1,2,3) 13 480Y Shutdown Boards A. Unit 1, 480V Shutdown 4kV Shutdown 4kV Shutdown Transfer from normal to aitemate source is manual.

BE) 1A Board A Board B Interlocking is provided to prevent manually transferring to a

8. Unit I 480V Shutdown 4kV Shutdown 4kV Shutdown taulted board and In prevent paratteling Iwo sources. 480V BE) 18 Board C Board B Load Shed Relay Time Delay Setting is set at 1.8 secs per DCN-W14030.

C. Unit 2, 480V Shutdown 4kV Board B 4kV Shutdown BD2A Board C E). Unit 2, 480V Shutdown 4kV Shutdown 4kV Shutdown BE) 28 Board 0 Board C E. UnitS, 480V Shutdown 4kV Shutdown 4kV Shutdown BE) 3A Board 3EA Board 3EB F Unit 3, 480V Shutdown 4kV Shutdown 4kV Shutdown BE) 38 Board 3EC Board 3EB

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBIlITY SUPPORT BFN Switchyard and 4160V AC Electrical 0.Ol-57A UnIt 0 System RevS 0141 Page 188 of 201 Illustration I (Page 5 of 7)

Auxiliary Power Supplies and Bus Transfer Schemes REMARKS: Automatic delayed transfer from tIle normal to alternate 1 source is irstiated by unctervoltage on the normal source and automatic return Is Initiated by normal voltage on normal source. These transfers are blocked alter lime delay a lire presence of an accident signal. When an sOdded signal is present, alternate 1 7 source breakers are tripped. Also, on 4kV Shutdown Ed A, B, C and Eli, the common accident sIgnal auto trip from U3 bus tie breskers (Alternate 3), has been removed. All diesel generators are autornatlcaty started by air occident signal, loss of voltage on Its shutdown board for ill seconds or degraded voltage ror4 seconils on its Shutdown board. Alter five (5) seconds with no voltage on lite shutdown board, at its supply breakers arid all Its loads except 416O-18OV transformers are automatically tripped. Alternate 2 source Is then automatically connected. A second level voltage protection Is provided for each 4KV shutdown board which wit operate an undervoltage relay. If voltage reduces to that board and after 7.43 seconds (from the Initial lime zero) the feed to the board Is tripped, the auto transfer Is blocked arid motor breakers on the board are tapped. 1.36 seconds later the 00 breaker closes In on that shutdown board. Manual return to the normal auxiliary power system is permitted if normal aurritiery power system voltage returns and If a unit is NOT in early stage 01 accident. UnIts 1 and 2 shutdown boards can be manually lied to their respectIve 3 unIt shutdown board, When doing 11115, Unit 3s breaker must be dosed In on a stead line (Intettocked to prevent closing si on an energized line) then Units 1 and 2 respectIve shutdown breaker con be synchronized to tie tire two boards together. Provision is included for backfeedlng diesel generator power from the shutdown boards into the 4160V unit boards for hot standby shutdown cooling if alt plant power, other than diesel generator power, Is lost. For this purpose, means are provided to manually synchronize 4KV shutdown boards

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN 1006 #3 Unit 2 was operating at 100% Reactor Power.

A ground AND subsequent fire in Shutdown Board 250V DC Distribution Panel SB-B resulted in de-energization of the SB-B panel AND trip of 4kV Shutdown Board B Normal Feeder Breaker.

Which ONE of the following completes the statements?

480V Shutdown Board 2A is _(1).

4kV Shutdown Board B (2)_ automatically transfer to its alternate source.

A. (1)energized (2) will B. (1) de-energized (2) will C. (1) energized (2) will NOT D. (1) de-energized (2) will NOT Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect 480v Shutdown Board 2A is de-energized (Optional): with the loss of 4kV Shutdown Board B. The transfer to alternate power is manual. Plausible in that Unit 1 and 3 480v Shutdown Board A normal power supply is from 4kV Shutdown Board A. Part 2 incorrect Each-Shutdown Battery system supplies its respective 4KV Shutdown Board and 480V Shutdown Board. All control power transfers are manual. Plausible in that if control power transfer is automatic as board power supply is or control power was not from SB-B DC Distribution Panel, this would be the correct answer.

B INCORRECT: Part 1 correct See explanation D. Part 2 incorrect See explanation A.

c INCORRECT: Part 1 incorrect See explanation A. Part 2 correct See explanation D.

D CORRECT: Part 1 correct 480v Shutdown Board 2A is deenergized with the loss of 4kV Shutdown Board B. It is the normal feeder to the 480v S/D Bd 2A and the transfer to alternate power is manual. Part 2 correct Each Shutdown Battery system supplies its respective 4KV Shutdown Board and 480V Shutdown Board. All control power transfers are manual. With the loss of control power, normal automatic transfer to alternate power supply will not occur.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295005 Main Turbine Generator Trip I 3 Tier# 1 AKI.O1 (IOCFR 55.41.8 to 41.10) S Knowledge of the operational implications of the following concepts Group# 1 as they apply to MAIN TURBINE GENERATOR TRIP: K/A# 295005AK1.01

. Pressure effects on reactor power Importance Rating 4.0 Proposed Question: #4 Unit 3 is operating at 20% Reactor Power with the Main Turbine online when a pipe rupture results in loss of ALL EHC:

Which ONE of the following completes the statement below?

Reactor Pressure will _(1) AND the Reactor _(2)_ Scram.

A. (1) rise (2) will B. (1) lower (2) will C. (1) rise (2) will NOT D. (1) lower (2) will NOT Proposed Answer: A Explanation A CORRECT: With the failure of EHC, the Main Turbine Trips and Bypass (Optional): Valves will fail closed. Reactor Pressure will rise until the Reactor High Pressure Scram setpoint is reached.

B INCORRECT: Plausibility based on misconception that Bypass Valves fail open on loss of EHC and subsequent scram on MSIV closure. Failing open is plausible in that there are EHC failures which will result in Bypass Valves failing open. For example, with EHC Control System in HEADER PRESSURE CONTROL, a single Header Pressure input failing high would result in Main Turbine Control Valves and Bypass Valves opening in attempt lower Reactor Pressure. Additionally, 3-AOl-47-2, Turbine EHC Control System Malfunctions, addresses EHC System Failures which result in lowering Reactor Pressure.

C INCORRECT: Plausible in that if candidate considers only Main Turbine Trip actuation of RPS, this would be the correct answer since it is bypassed at this power level.

D INCORRECT: Plausibility based on misconceptions that Bypass Valves fail open on loss of EHC as discussed in detail above and subsequent scram on MSIV closure is bypassed at this power level or candidate considers only Main Turbine Trip actuation of RPS.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the operational implications of Pressure effects on reactor power as they apply to Main Turbine Generator Trip due to loss of EHC.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71.010, Rev. 12 (Attach if not previously provided) 3-01-99 Rev. 47 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .010, V.B.6 (As available)

OPL1 71 .010, V.B.23 Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of even, question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71.O1O Revision 12 Page 29 of 80 F. Turbine Bypass Valves (Nos. 1 through 9) TP-1 and TP-7,8 c3% per BPV

1. Purposes ObJ.V.B.6.d

a. Routes steam not needed by the turbine to Obj.V.C.2.d the condenser during the following Obj.V.E.27 conditions:

(1) Reactor Startup (2) Turbine Roll (3) Turbine Trips (4) Reactor cooldowri b; Works in conjunction with the turbine control valves to maintain a constant reactor pressure for a given reactor power level.

c. Provides the capability to prevent over pressurization of the reactor If the MSIVs are open.

2. Location The nine bypass valves are physically located above the turbine throttle in the moisture separator room near the main turbine stop and control valves.

3. Bypass Valve Design

a. Bypass valves are hydraulically operated, reverse seatIng globe valves.

b. The valves are positioned as required by a Control PAC and Servo-valves, Valves fail closed upon loss of hydraulics.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Protection System 3-OI..99 Unit 3 Rev. 0047 Page 63 of 80 Illustration 2 (Page 2 of 2)

Unit 3 Reactor Scram Initiation Signals Scram Setpoint Bypass J. OPRM TRIP Any one of the three algorithms, Reactor is NOT operating in period, growth, or amplitude for an the AUTO ENABLE Region of operable OPRM cell has the Power/Flow Map.

exceeded its trip value conditions:

K. Low RPV N/A Water Level (Level 3)

L. Hi RPV 1073 psig N/A Pressure M. HI DW 2.45 psig N/A Pressure N. MSIV closure 90% open (3 Main Steam Lines) NOT in RUN

0. Scram

• Thermal level switches Mode Switch in SHUTDOWN Discharge 49 gallons or REFUEL with keylock Instrument (LS-85-45A,B,G,H) switch in BYPASS Volume Hi Hi

• Float level switches 45 gallons (LS-85-45C,D,E,F)

P. TSV Closure 90% open (3 TSVs) <30% Rx Power ( lS4psig 1st stage pressure)(TR 3.3.1)

0. TCV Fast 40% mismatch (amps to <30% Rx Power (lS4pslg Closure (load cross-under pressure>: 850 psig 1st stage pressure)(TR 3.3.1) reject> EHC RETS at TCV (1 or 3) & (2 or 4)

R. LossofRPS N/A N/A Power S. Scram Channel Key-locked in AUTO N/A Test Switches Panels 3-9-15 & 3-9-17

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Turbine EHC Control System 3.AOl-47-2 Unit 3 Malfunctions Rev. 0006 Page 3 of 8 1.0 PURPOSE This abnormal operating instruction provides symptoms, automatic actions, and operator actions for malfunctions of the EHC Control System.

2.0 SYMPTOMS A. While in REACTOR PRESSURE CONTROL failed high or low reactor pressure input. The following symptoms may occur:

1. EHC/TSI SYSTEM TROUBLE annunciation, 3-XA-55-7A, Window 6, alarms.

2. On Panel 3-9-7, REACTOR PRESS A(B)(C)(D) BYPASS pushbutton backlight Illuminates.

B. While in HEADER PRESSURE CONTROL a single header pressure input signal fails low. The following symptoms may occur:

1. EHCITSI SYSTEM TROUBLE annunciation. 3-XA-55-7A. Window 6, alarms.

2. On Panel 3-9-7, HEADER PRESSURE A(B) BYPASS pushbutton backlight illuminates.

C. While In HEADER PRESSURE CONTROL a single header pressure input signal fails high.

1. The following symptoms may occur:

a. On Panel 3-9-7, HEADER PRESSURE A(S) BYPASS pushbutton backlight illuminates.

b. Turbine control valves open to position established by CV POSITION LIMIT setpolnt.

c. Turbine bypass valves open.

d. Feedwater/Steam flow mismatch.

e. Reactor pressure lowers.

f, Generator output rapidly lowers.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BPN Turbine EHC Control System 3-AOi-47-2 Unit 3 Malfunctions Rev 0006 Page 6 of 8 4.0 OPERATOR ACTIONS NOTE If necessary, place keeping marks may be made directly in the Control Room copy of this instruction. Management Services should be contacted for a replacement copy when time permits.

4.1 ImmedIate Actions

[1] IF Reactor Pressure lowers to or below 900 psig, THEN MANUALLY SCRAM the Reactor and CLOSE the MSIVs.

[PER 03-006187-4300]

4.2 Subsequent Actions (1] IF ANY EOI entry condition is met, THEN ENTER the appropriate EGI(s). C

[2] VERIFY Automatic Actions have occurred. C

[3) iF a Group 1 isolation has occurred, THEN PLACE EHC PUMP 3A and 38, 3-HS-47-1A and 3-HS-47-2A, to PULL TO LOCK. C

[3.1] WHEN the turbine bypass valves dose, THEN RESET the Group I PCIS isolation and OPEN MSIV5, as desired. REFER TO 3OI-1. C

[4] USE EHC WORKSTATION computer to aide in diagnosing the problem. C

[5] REQUEST assistance from Site Engineering. C

[6] IF necessary, THEN TROUBLESHOOT the EHC Control System. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Reactor Protection System 3-01-99 Unit 3 RevS 0047 Page 63 of 80 Illustration 2 (Page 2 of 2)

Unit 3 Reactor Scram Initiation Signals Scram Setpoint Bypass J. OPRM TRIP My one of the three algorithms, Reactor is NOT operating in period, growth, or amplitude for an the AUTO ENABLE Region of operable OPRM cell has the Power/Flow Map.

exceeded its trip value conditions:

K. Low RPV N/A Water Level (Level 3)

L, Hi RPV 1073 pslg N/A Pressure M. Hi DW 2.45 psig N/A Pressure N. MSIV closure 90% open (3 Main Steam Lines) NOT in RUN

0. Scram

• Thermal level switches Mode Switch in SHUTDOWN Discharge 49 gallons or REFUEL with keylock instrument (LS-85-45ABG,H) switch in BYPASS Volume Hi Hi

• Float level switches 45 gallons (LS-85-45C,DE,F)

P. TSV Closure 90% open (3 TSVs) <30% Rx Power ( 1 S4psig A

1st stage pressure)(TR 3.3.1)

0. TCV Fast 40% mismatch (amps to <30% Rx Power (1 S4psig Closure (load cross-under pressure); 850 psig 1st stage pressure)(TR 3.3.1) reject) EHCRETSatTCV(1 or3)&(2 or 4)

R, LossofRPS N/A N/A Power S. Scram Channel Key-locked in AUTO N/A Test Switches Panels 3-9-15 & 3-9-17

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295006 SCRAM / 1 Tier# 1 AAI.05 (10CFR 55.41.7)

Ability to operate and/or monitor the following as they apply to Group# 1 SCRAM: KIA# 295006AA1.05

. Neutron monitoring system Importance Rating 4.2 Proposed Question: #5 With Unit 2 in Mode 2, Intermediate Range Monitors (IRMs) indicate 29.1 on Range 3 AND Reactor Period is 90 seconds.

Which ONE of the following identifies approximately how long it will take to reach the IRM Scram setpoint?

A. 35 seconds B. 65 seconds C. 125 seconds D. 180 seconds Proposed Answer: C Explanation A INCORRECT: Plausible in that this would be half the time to the first (Optional): doubling.

B INCORRECT: Plausible in that this would be the time to the first doubling.

C CORRECT: C is correct as with a reactor period of 90 and 2 doubling times, (29.1-58.2 and 58.2-116.4). This time would be 62.28 seconds times 2. The scram setpoint would be reached in 124.56 seconds.

D INCORRECT: Plausible in that this would be twice the period.

KA Justification:

The KA is met because the question tests candidates ability to monitor IRMs as they apply to Scram.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Candidates must determine doubling time based on Reactor Period then calculate time to reach IRM Scram setpoint.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL1 71.020, Rev. 11 I 2-Ol-92A, Rev. 28 (Attach if not previously provided) 2-GOI-1 00-lA Rev. 145 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .020 V.B.7 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Monticello 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.020 Revision 1 1 Page 20 of 44 INSTRUCTOR NOTES E. Trips TP-10 I. Rod blocks Obj.V.D.7, V.B.5 Obj. V.C.3.,

Block Setpoint When Bypassed Downscale <7.5 Range I or RUN Obj. V.8.6.

Obj.V.C.4 Obj. V.B.5

>90/104.6 RUN Mode Unit Difference IRM high setpoint is 1NOP -HV low (<90v) RUN Mode o at Unit 2 and 104.6

-Module unplugged on Unit 1 and Unit 3

-Function switch not in OPERATE

-Loss of +/-24VDC Detector Wrong Detector RUN Mode Obj.V.B.13 Position Not Full IN

2. Scrams TP-11 Scrams Setpoint When Bypassed Obj. V.8.7.

Obj. V.C.5. Obj.V.D.8 High-High 116.4 In RUN Mode INOP -HV low (<90v) in RUN Mode

-Module unplugged

-Function switch not in OPERATE

-Loss of +/-24VDC F. Controls Provided

1. Panel 9-5

a. Recorder switches select between IRM channels, and APRM/RBM channels have been removed. All units now contain digital recorders, which do not require operation of selector switches. These switches have been removed.

b. Range switches allow operator to select appropriate 1PM range to maintain indications between 25 to 75 on 0-125 scale. 0-40 scale is no longer utilized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Intermediate Range Monitors 2-OI-92A Unit 2 Rev. 0028 Page 14 of 14 illustration I (Page 1 ofl)

IRM Trip Outputs TRIP SIGNAL SETPOINT ACTION IRM High > 90 ON 125 SCALE Rod block unless REACTOR MODE SWITCH in RUN IRM mop A. Module unplugged Rod block unless REACTOR MODE SWITCH in RUN B. Mode switch NOT in operate Reactor Scram unless C. HV power supply low REACTOR MODE SWITCH in voltage D. Loss of +1-24 vclc IRM Downscale < 7.5 on 125 SCALE Rod block unless lRMs on range 1 unless REACTOR MODE SWITCH in RUN

• IRM Detector Wrong detector NOT full in Rod block unless detector Position full-in, or REACTOR MODE SWITCH in RUN IRM High-High > 1 1S,4 ON 125 SCALE Reactor Scram unless REACTOR MODE SWITCH in RUN

ES-401 Sample Written Examination Form ES-401-5

• Question Worksheet BFN Unit Startup and Power Operation 2-GOt-I 00-IA Unit2 Rev.0145 Page 88 of 178 5.4 Withdrawal of Control Rods while in Mode 2 (continued)

. NOTE Period Is measured directly from IRMs, using one of the following methods:

{

1) MULTIPLY time for 10% power rise by 10.5.

2) MULTIPLY doubling time by 1.445,

3) DIVIDE time for decade rise by 2,3.

4) Directly, time for power to rise from 25 to 8.

[15.2] RECORD the following in the Narrative Log:

A. Critical Data Period D

• Time D a Rod Group

• Rod Number

• Rod Notch D (R)

Initials Date Time B. Recirc Pump 2A and 28 Temperatures using either of the following: (N/A indication for a pump that is 005 and in Single Loop Operation.)

• RECIRC PUMPS DISCH FLOW & TEMP PMP-2A (PMP-28).

CH 3 (CH 4) on 2-XR-68-2/5 on Panel 2-9-4.

• RECIRC PMP A (B> SUCT TEMP 68-6A (8-83A) on ICS.

• RECIRC PMP A (B) DISCHARGE TEMP 68-2 (68-78) on ICS.

2A LOOP F / 28 LOOP F (R)

Initials Date Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295016 Control Room Abandonment /7 Tier# 1 G2.1.28 (10CFR 55.41.7)

Knowledge of the purpose and function of major system Group# 1 components and controls. KIA# 295016G2.1.28 Importance Rating 4.1 Proposed Question: # 6 Which ONE of the following functions can be performed at Backup Control Panel 2-25-32?

A. Close ALL MSIVs B. Operate ALL ADS Valves C. Suppression Chamber Spray D. Control Reactor Level with HPCI Proposed Answer: A Explanation A CORRECT: BOTH Inboard and Outboard MSIVs can be closed from (Optional): Backup Control Panel 2-25-32.

B INCORRECT: Plausible in that Four ADS valves can be controlled from Panel 25-32. Six SRVs (Non-ADS) have disconnect switches at Panel 25-32.

C INCORRECT: Plausible in that indications for RHR are on 2-25-32 and 2-AOI-100-2, Control Room Abandonment, provides instruction for Suppression Pool Cooling and Shutdown Cooling.

D INCORRECT: Plausible in that Reactor Level can be controlled with RCIC at PnI 2-25-32.

KA Justification:

The KA is met because it tests the candidates knowledge of function of major system components associated with Control Room Abandonment procedure and the Backup Control Panel.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOl-100-2, Rev. 54 (Attach if not previously provided)

OPL1 71 .208, Rev. 5 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source:

(Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 2-AOl-I 00-2 Unit 2 Rev, 0054 Page 10 of 96 42 Unit 2 Subsequent Actions (continued)

CAUTION Failure to place control switch in desired position prior to transferring to emergency position may result in inadvertent actuation of the component.

[6] CLOSE MSIVs using the following switch sequence at Panel 2-25-32;

[6.1] PLACE control switch in CLOSE. C

[6.2] PLACE transfer switch in EMERG. C Control Required Transfer Required MSIV LINE Switch Position Switch Position A INBOARD 2-HS-1-14C CLOSE C 2-XS-1-14 EMERO C B INBOARD 2-HS-1-26C CLOSE C 2-XS-1-26 EMERG C C INBOARD 2-HS-1-37C CLOSE C 2-XS-1-37 EMERG C 0 INBOARD 2-HS-1-51C CLOSE C 2-XS-1-51 EMERG C A OUTBOARD 2-HS-1-15C CLOSE C 2-XS-1-15 EMERG ID B OUTBOARD 2-HS-1-27C CLOSE C 2-XS-1-27 EMERG ID C OUTBOARD 241S-1-38C CLOSE ID 2-XS-1-38 EMERG C 0 OUTBOARD 2-HS-1-52C CLOSE C 2-XS-1-52 EMERG C 4.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Control Room Abandonment 2-AOl-I 00-2 Unlt2 Rev. 0054 Page 8 of 96 4.2 Unit 2 Subsequent Actions

[1] IF ALL control rods were NOT fully inserted AND RPS failed to deenergize, THEN: (Otherwise N/A>

DIRECT an operator to Unit 2 Auxiliary Instrument Room to perform Attachment 11, D NOTES

1) The following transfers Reactor Pressure Control to Panel 2-25-32 to allow for pressure control while completing the Panel Checklist,

2) Attachment 9, Alarm Response Procedure Panel 2-25-32, provides for any alarms associated with this instruction.

CAUTION 13 Failure to place control switch in desired position prior to transferring to emergency position may result in inadvertent actuation of the component.

2) [NERICI Operation from Panel 2-25-32 bypasses logic and interlocks normally associated with the components. [GE SL 326,

[2] PLACE the following MSRV control switches in CLOSE/AUTO at Panel 2-25-32:

Switch No. Description 2-HS-1-22C MAIN STM LINE B RELIEF VALVE C 2-HS-1-SC MAIN STM LINE A RELIEF VALVE C 2-HS-1-30C MAIN STM LINE C RELIEF VALVE C 2-HS-1-34C MAIN STM LINE C RELIEF VALVE C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BPN Control Raom Abandonment 2-AOl-I 00-2 Unlt2 Rev. 0054 Page 9 of 96 4.2 Unit 2 Subsequent Actions (continued) t3] PLACE the following MSRV disconnect switches In DISCT at Panel 2-25-32:

Switch No, Description 2-XS-1-18 MAIN STM LINE B RELIEF VALVE DISCT 0 2-XS-1-4 MAIN STM LINE A RELIEF VALVE DISCT C 2-XS-1-42 MAIN STM LINE D RELIEF VALVE DISCY C 2-XS-1-23 MAIN STM LINE B RELIEF VALVE DISCT C 2-XS-I-41 MAIN STM LINE D RELIEF VALVE DISCT C 2-XS-i-180 MAIN STM LINE D RELIEF VALVE DISCT C (4] PLACE the following MSRV transfer switches in EMERG at Panel 2-25-32:

Switch No. Description 2-XS-1-22 MAIN STM LINE B RELIEF VALVE XFR C 2-XS-1-5 MAIN STM LINE A RELIEF VALVE XFR C 2-XS-1-30 MAIN STM LINE C RELIEF VALVE XFR C 2-XS-1-34 MAIN STM LINE C RELIEF VALVE XFR C

{ (5]

8.

C.

NOTE Use of the following sequence when opening MSRVS should distribute heat evenly in the Suppression Pool.

MAINTAIN Reactor Pressure between 800 and 1000 psig using the following sequence at Panel 2-25-32:

A. 2-HS-1-22C, MAIN STM LINE B RELIEF VALVE 2-HS-1-5C, MAIN STM LINE A RELIEF VALVE 2-HS-1-30C, MAIN STM LINE C RELIEF VALVE C

C a

D, 2-HS-1-34C, MAIN STM LINE C RELIEF VALVE C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI7I.208 Revision 5 Page 6 of 10

9. Trip reactor feed pumps as necessary to prevent tripping Obj. V.B.8 on high water level. Obj. V.C.5

10. Start the diesel generators. (9-8 Switch starts respective units DIG only)

11. Verify each EECW header has one pump in service.

12. Announce to all plant personnel that the Control Room is being evacuated and all operators are to report to their assigned backup control stations.

13. Obtain hand held radios from the control room.

14. Proceed to the Backup Control Panel (25-32)

F. Subsequent Actions See AOl-i 00-2 for details for actions HU Tools: Procedure Use If rods failed to fully insert and RPS did not deenergize, Obj V.C.2 an operator is directed to pull RPS fuses. However, this See AOI-100-2 is beyond the actual design bases. Attachment 11

2. Transfer reactor pressure control to Panel 25-32 to allow Note: System Status for pressure control while the rest of the panel checklist prior to abandonment is being completed. maintained by GOI-300-1 checklists.

3. Before any transfer switch is placed in EMERGENCY, its Obj. V.B.2 associated control switch must be verified to be in the proper Obj. V.B.3.

position. Placing a transfer switch in the EMERGENCY position enables the local control switch, and the device will assume the condition called for by the local control switch.

For example, if a transfer switch for an ADS valve is placed in EMERGENCY with the local control switch in OPEN, the ADS valve will open.

a. Place the transfer switches for the ADS valves, and TP-1 the disconnect switches for the non-ADS valves in Obj. V.B.7 EMERGENCY after making sure the control switches are in the AUTO position. This action disables the Control Room hand switches and the ADS function and is performed to prevent spurious blowdown of the primary system. The other 3 SRVs are disabled by opening their breakers on 25OVDC

{ RMOV board 2B(3B).

Four ADS valves can be controlled from Panel 25-32. Six SRVs (Non-ADS) have only disconnect switches at Panel 25-32.

Obj. V.B.8 Obj. V.B.7

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Control Room Abandonment 2-AOl-I 00-2 Uriit2 Rev.0054 Page 110ff 96 4.2 Unit 2 Subsequent Actions (continued)

NOTES

1) Attachment 1 provides normal backup control stations and available communications.

2) Attachment 10 provides PAX extensions and locations.

t71 ESTABLISH communication with the following personnel and DIRECT attachments be completed as follows:

• U-2 Unit Operator complete Attachment 2, Part A. C]

• U-2 Rx Bldg AUC complete Attachment 3, Part A. C]

• U-2 Turb Bldg AUO complete Attachment 4. Part A. C

[8] Upon completion of attachments, RE-ESTABLISH communication using the best available means and continue procedure. C]

{ 1) 2)

[91

[9.1]

[9.2]

CAUTION RCIC TURBINE STEAM SUPPLY VALVE, 2-FCV-715, transfer switch has been placed in EMERGENCY and will NOT trip on Reactor Water Level High (+51 inches).

Failure to maintain level below this value may result In equipment damage.

RCIC will still trip on low suction pressure, high turbine exhaust pressure, mechanical overspeed, and trip push button on pnl 25-32.

INITIATE RCIC as follows:

CHECK OPEN 2-FCV-71-9 RC1C TURB TRIPITH ROT VALVE RESET, 2-HS-71-9D At Panel 2-25-32. (Red Light above switch)

PLACE RCIC PUMP MIN FLOW VALVE EMER HAND SWITCH, 2-HS-071-0034C, in OPEN at 250V DC RMOV Bd 28, compt. 5D. (Unit 2 Turbine Building AUO)

C]

C

[9.3) PLACE RCIC TURB STM SUPPLY VALVE EMER HAND SWITCH, 2-HS-071-000BC, in OPEN at 250V DC RMOV Bd 2C, compt. 48. (Unit 2 Reactor Building AUO) C]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Control Room Abandonment 2-AOl-I 00-2 Unit 2 Rev. 0054 Page 12 of 96 4.2 Unit 2 Subsequent Actions (continued)

NOTE RCIC Turbine should start and flow should stabilize at 620 gpm.

f (9.4] CHECK turbine speed 2100 rpm or above using RCIC I TURBINE SPEED, 2-51-71 -428 at Panel 2-25-32. El I [9.5] PLACE RCIC PUMP MIN FLOW VALVE EMER HAND A SWITCH, 2-HS-071-0034C, in CLOSE at 250V DC J RMOV Bd 28, compt. 3D. (Unit 2 Turbine Building AUO) C

£9.61 ADJUST flowrate as necessary using RCIC SYSTEM I FLOW/CONTROL, 2-FIC-71-368 at Panel 2-25-32. C IA [9.7] MAINTAIN Reactor Water Level betveen +2 and

+50 Inches using RX WATER LEVEL A & B, 2-Ll-3-46A 1% & B at Panel 2-25-32. C NOTE The following step prevents HPCI operation and automatic opening of HPCI MAIN PUMP MINIMUM FLOW VALVE, 2-FCV-73-30, ElO] At 250V Reactor MOV Bd 2A, PERFORM the following:

.(10.1] VERIFY CLOSED HPCI STEAM SUPPLY VALVE TO TURB FCV-73-16 at Compt. 3D. (MO 23-14). El

[10.2] PLACE HPCI TURBINE STEAM SUP VLV TRANS, 2-XS-73-1 6, in EMERG at Compt. 3D. C

[10.3] IF desired to verify HPCI MIN FLOW BYPASS TO SUPPRESSION CHAMBER VALVE, 2-FCV-73-30, closed prior to opening breaker, THEN (Otherwise N/A>

DIRECT operator to verify locally. El

[10.4] PLACE HPCI MAIN PUMP MIN FLOW VLV FCV-73-30, breaker in OFF Compt. SD. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILLTY SUPPORT BFN Control Room Abandonment 2-AOl-I 00-2 Unit 2 Rev. 0054 Page 16 of 96 4.2 Unit 2 Subsequent Actions (continued)

[15] INITIATE RHR Suppression Pool Coaling as follows:

CAUTIONS

1) The RH RSW and EECW Systems are common to all three units. Coordination and communication between the Unit Operators on all three units is required whenever configuration changes to the RHRSW and/or the EECW Systems are made.

2> Communication between 41 60V Shutdown Sd A arid 48W RMOV Sd 2A is necessary for establishing RHRSW flow and to prevent exceedIng 53 amps on RHRSW Pump Ci

[15.1] PLACE RHRSW PUMP C2 MOTOR, O-HS-2312C, in CLOSE at 4160V Shutdown Sd B, compt. 15, to start RHR SERVICE WATER PUMP C2. D

[15.2] THROTTLE OPEN RHR HX 2C OUTLET VLV, 2-HS-023-0040C at 480V RMOV Sd 2A, compt. 1 8C. C

[15.3] WHEN between 48 arid 52 amps on RHR SERVICE WATER PUMP C2, THEN:

STOP throttling, RHR HX 2C OUTLET VLV, 2-HS-023-0040C. C

[15.4] VERIFY OPEN RHR SYSTEM I MINIMUM FLOW VALVE, 2-FCV-74-7, at either of the following:

• 480V RMOV Sd 2D, compt. SE, RHR SYSTEM I MINIMUM FLOW VLV, OR

• Rx Bldg SW Quad El 541 local control switch RHR SYSTEM I MINIMUM FLOW VALVE, 2-HS-74-7B. Q

[15.5] PLACE RHR PUMP 2C, 2-HS-074-0016C, in CLOSE to start RHR PUMP 2C at 4160V Shutdown Sd B, compt.

17. C

[15.6] PLACE RHR SYSTEM I SUPP POOL SPRAY/TEST ISOL VLV, 2-HS-74-57C, in OPEN at 480V RMOV Sd 2A, compt. 1IC. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Control Room Abandonment 2-AOl-I 00-2 Unit 2 Rev. 0054 Page 23 of 96 4.2 Unit 2 Subsequent Actions (continued)

NOTE Unit 2 and Unit 3 both align RHR Loop I for Shutdown Cooling. IF both Unit 2 and Unit 3 Control Rooms have been abandoned, THEN coordinate the initiation of Shutdown Cooling such that one unit uses RHR Pump A and the other Unit uses RHR Pump C. This will allow minimum flow protection to be maintained for the RHRSW Pumps and prevent flow adjustments on one Unit from affecting the opposing Units Cooldown rate.

[20] INITIATE RHR Shutdown Cooling as follows: (Otherwise N/A)

[20.1] VERIFY REACTOR PRESSURE B, 2-Pi-3-79, less than 50 psig, at Panel 225-32. C

[20.2] IF RHR pumps are operating in Suppression Pool Cooling or RHR LPCI, THEN PERFORM the following: (Otherwise N/A)

[20.2.13 PLACE RHR SYSTEM I TEST VLV, 2-HS-074-0059C, in CLOSE, at 480V RMOV Gd 2A, compi. 19C5. C

[20.2.2] PLACE RHR SYSTEM I SUPP POOL SPRAY/TEST ISOL VLV. 2-HS-74-57C, in CLOSE, at 480V RMOV Gd 2A, compt. 11 C. C (20,2.3] VERIFY CLOSED RHR SYSTEM I OUTBD INJECTION VLV, 2-HS-74-52C at 480V RMOV Gd 2A, Compt. 2G. C (20.2.4] PLACE RHR PUMP C, 2-HS-74-1SC, in TRIP to stop RHR PUMP 2C, at 4160V Shutdown Gd B, compt. 1. C

[20.2.53 PLACE RHR PUMP 2A, 2-HS-074-0005C, in TRIP, at4l6OV Shutdown Gd A, compt. 19, to stop RHR PUMP2A. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295018 Partial or Complete Loss of Component Cooling Water / 8 Tier#

1 AA2.O1 (1 0CFR 55.41 .10)

Ability to determine and/or interpret the following as they apply to Group# 1 PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING KJA# 295018 AA2.01 WATER:

Component temperatures Importance Rating 33 Proposed Question: #7 Unit 3 is operating at 100% Reactor Power when the following alarms AND indications are received:

• A Partial Loss of Reactor Building Closed Loop Cooling Water occurs.

• RWCU NON-REGENERATIVE HX DISCH TEMP HIGH, (3-9-4B, Window 17) is in alarm.

• RWCU Non- Regenerative Heat Exchanger Discharge Temperature is 140° F.

Which ONE of the following describes the effect of this condition, if any, on the operation of the Reactor Water Cleanup (RWCU) Pumps?

A. TRIP immediately due to isolation valve position.

B. TRIP directly due to the high temperature signal.

C. CONTINUE to operate since no trips are received.

D. TRIP after a low flow condition exists for 30 seconds.

Proposed Answer: A Explanation A CORRECT: RWCU Non- Regenerative Heat Exchanger Discharge (Optional): Temperature at 140° F isolates RWCU. When RWCU isolation valve FCV 69-br 2 Not Full Open, RWCU Pumps trip.

B INCORRECT: B is plausible; identifies misconception about RWCU Pump Trip directly from High Temperature signal. High Temperature initiates a PCIS Isolation. When the Isolation Valve is NOT FULLY OPEN, the RWCU Pump TRIPS. Also, Pump cooling water outlet high temperature (RBCCW) 140°F after a 30-second time delay trips RWCU Pumps.

C INCORRECT: Plausible in that the purpose of the 140°F isolation is to protect ion exchange resin from high temp damage but the F/Ds Design temperature of 150° F has not yet been reached.

D INCORRECT: Plausible in that System Low Flow of 56 gpm with a time delay of 30 seconds will trip RWCU Pumps. With the Isolation Trip coming with valves just off full open, they would cause the trip prior to low flow condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

This question satisfies the K/A statement by testing candidates ability to interpret RWCU Temperatures as they apply to Partial Loss of RBCCW. Partial loss of RBCCW results in RWCU Non- Regenerative Heat Exchanger Discharge Temperature at 1400 F which isolates RWCU. When RWCU isolation valve FCV 69-1 or 2 Not Full Open, RWCU Pumps trip.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): 3-01-69 Rev. 79 (Attach if not previously provided)

OPLI7I.013 Rev. 18 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .013 V.B.3 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Nine Mile 2 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BPN Reactor Water Cleanup System 3.01.69 Unit 3 Rev. 0079 Page 15 of 138 3.8 RWCU Isolation Signals A. Reactor water level low (LEVEL 3).

8, Non-regenerative heat exchanger outlet high temperature 140°F.

C. RWCU Pump Room 3A high temperature 148°F.

ID. RWCIJ Pump Room 38 hIgh temperature 148°F.

E. Main Steam Tunnel/RWCU Piping high temperature 197°F.

F. RWCU System Pipe Trench 131°F.

G. RWCU Heat Exchanger Room Pipe Chase Area high temperature 166°F.

H. RWCU Heat Exchanger Room high temperature 139°F I. Standby Liquid Control system initiation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CORRECT ANSWER AND DISTRACTOR PLAUSIBILITY SUPPORT BFN Reactor Water Cleanup System 3-01-69 Unit 3 Rev. 0079 Page 14 of 138 3.6 Pumps (continued)

C. RWCU is required to be operated with the following restrictions with reactor pressure 50 psig (modes 2 or 3) or any time the unit is in mode 4, mode 5. or defueled:

1. One pump in operation, pump can be operated to its maximum flow capacity.

2. Two pumps in operation, maximum flow limited to 100 gpm per pump (200 gpm total).

D. Leaving an idle pump pressurized can damage the seals by erosion paths across the seal faces.

E. If conditions listed below are satisfied, the Unit 3 RWCU pumps may be operated with 0 gpm seal water flow, after pump start. However, RWCU pump operation with 0 gpm seal water flow will reduce seal life and the seal wIll most likely need to be replaced before operating the seal at its design parameter.

1. CRD pumps are not available and RWCU seal water is supplied from the CS&S System.

2. Reactor Vessel is at atmospheric pressure.

3. RWCU seal flow is 1.8 to 2 gpm prior to pump start.

3.7 RWCU Pump Trip Signals A. Low flow 56 gpm (30 second time delay if control switch in NORMAL after start).

B. Cooling water high temperature 140°F (7 sec. time delay).

C. RWCLJ INBD SUCT ISOLATION VALVE, 3-FCV-69-1 not full open.

D. RWCU OUT5D SUCT ISOLATION VALVE, 3-FCV-69-2 not full open.

E. RWCU RETURN ISOLATION VALVE. 3-FCV-69-12 fully closed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CORRECT ANSWER AND DISTRACTOR PLAUSIBILITY SUPPORT OPL1 71.013 Revision 18 Page 29 of 47 (4) High temperature at the outlet of the Not a Safety NRHX (140°F, TIS-69-11) to protect function.

the ion exchange resin from damage due to high temperature. An alarm is provided on Panel 9-4 from TIC 10.

(5) Loss of RPS A will result in an inboard and outboard Group 3 (RWCU) isolation. Loss of RPS B will result in an outboard Group 3 (RWC(J) isolation.

b. RWCIJ Recirc Pumps Trips Obj. V.B.5 0bj V.D.6 (1) RWCU isolation valve FC/ 69-br 2 Obj. V.E.10 not full open.

(2) RWCU return isolation valve FCV UNIT 69-12 full closed. DIFFERENCE TACF 2-08-00 1-069 disables 2B pump trip on (3) Pump flow 56 gpm for 30 seconds 2-FCV-69-1 2 with the control switch in NORMAL- closure AFTER-START position.

(4) Pump cooling water outlet high Right unitltrainl temperature (RBCCW) 140°F after a component 30-second time delay, if the control switch is in the NORMAL-AFTER-START position.

(5) 480V Load Shed Logic

c. Filter!Demineralizers (1) Holding pump auto starts at4O gpm decreasing (2) Alarm on high differential pressure at 25 psid; alarm on resin trap differential pressure of 20 psid.

(3) Automatic isolation at 25 psid (F/D);

20 psid resin trap NOTE: Will only close the effluent valves

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI 71.013 Revision 18 Page 17 of 47

4. Filter/Demineralizers (FIDs) Obj. V.8.2 Obj. V.0.2.

a. Are used to maintain water purity by Obj. V.E.3.d mechanical and chemical filtration. They remove insoluble solid particles and dissolved solids from the water. Each of the units is of the pressure precoat type, which uses finely ground mixed ion exchange medium. The F/Ds operate in parallel at 50% of the total system capacity.

Design temperature (°F) is 150°F: Water to the F/Ds should be maintained less than Monitor Critical 130°F to maximize resin efficiency and to Plant Parameters avoid resin damage. Water Temperature of 150° 200°F reduces the impurity removal capacity. Temperature >200°F causes the Powdex to decompose. Resin traps are provided to prevent carryover into the reactor system of filter or resin material due to filter element failure.

b. Resin introduction into the reactor coolant See Plantllndustry can cause conductivity increase, pH Experience decrease, and sulfate concentration (Section X. 0) increase (which propagates pitting corrosion and intergranular stress corrosion cracking).

c. Holding pumps are provided to maintain the TP-3 filter charged until the unit is in service.

PCR 08003930 Pump will automatically start if flow through added steps to the filter drops to 40 gpm to prevent the rmove Demins precoat from dropping off the filter from service prior elements.

to RPS Bus xfer

d. A Flow control valve maintains constant UNIT flow rate through each FID for varying DIFFERENCE pressure drops. This is set at local FID control station. (NOTE: Flow control valves are normally operated in manual.). Flow should never exceed 135 gpmfunit (170 gpm/unit on Unit 1/3) Test results has shown that U-i will not exceed 160 gpm.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NINE MILE 22008 Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO Tier K/A Number Statement IR Ongin Source Question 451 295018 AA2.013.3jNNA LOKGrpIIOCFR55:41(b)7 LOD(1-5)RefeceDocunients H I  : ARP602319 to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLJNG WATER: Component temperatures QUESTION 45 The plant is operating at 100% power with the following:

• Loss of Reactor Building Closed Loop Cooling Water occurs

• Annunciator 602319, RWCU FILTER DEMIN INLET TEMP HIHl alarms Which one of the following describes the affect of this condition, if any, on the operation of the Reactor Water Cleanup Pumps?

A. CONTINUE to operate since no trips are received.

B. TRIP immediately due to isolation valve position.

C. TRIP directly due to the high temperature signal.

0. TRIP after a low flow condition exists for 15 minutes.

,Correct Answer: B When 602319 alarms, at WCS*MOVI 12, CLEANUP SUCT OUTBOARD ISOL VLV closes. When the valve is NOT FULLY OPEN, the WCS Pumps TRIP immediately.

Plausible Distractors:

A is plausible; would be true for RWCU F/D Inlet (NRHX Outlet> temperature below 1 40F.

C is plausible; identities misconception about RWCU Pump Trip directly from High Temperature signal. High Temperature initiates a PCIS Isolation. When the Isolation Valve is NOT FULLY OPEN, the RWCU Pump TRIPS, 0 is plausible; would be true for valve closures OTHER THAN WCS*MOV1 12. A Low Flow condition would develop which initiates a time delayed RWCU Pump TRIP.

Page 51 of 66

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RD SRO 295019 Partial or Total Loss of Inst. Air! 8 Tier # 1 AK3.03 (10CFR55.41.5)

Knowledge of the reasons for the following responses as they apply Group # 1 to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: K/A # 2951 09AK3.03

• Service air isolations: Plant-Specific Importance Rating 3.2 Proposed Question: #8 Control Air Header Pressure is lowering due to a rupture in the system.

Which ONE of the following identifies the HIGHEST Control Air Pressure that will result in Service Air Isolation Valve, 0-FCV-33-1, closing AND the reason?

A. 30 psig; To isolate non-essential Service Air loads.

B. 30 psig; Due to insufficient air pressure to keep the valve open.

C. 50 psig; To isolate non-essential Service Air loads.

D. 50 psig; Due to insufficient air pressure to keep the valve open.

Proposed Answer: B nd 2

Explanation A INCORRECT: l part correct See B Explanation. Part incorrect (Optional): See C Explanation.

B CORRECT: Service air supply valve from control air header (0-FCV-33-1).

The valve automatically opens if control air pressure falls to 85 psig and closes at 30 psig (due to insufficient air pressure to keep the valve open).

C INCORRECT: Recognizable pressure associated with loss of Control Air as the pressure that Condensate Demin Bypass Valve Fails open. Plausible in that it is logical to isolate non-essential Service Air loads with a loss of Control Air similar to RBCCW Sectionalizing Valve closing on low header pressure to isolate non-essential RBCCW loads.

D INCORRECT: 1 t part incorrect see C Explanation. nd 2

Part Correct See B Explanation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

This question satisfies the K/A statement by testing knowledge of the reason and the setpoint for Service air isolation Valve, 0-FCV-33-1, closing as a result of a rupture in the Control Air System and lowering pressure.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): 0-01-32 Rev 127 (Attach if not previously provided)

OPLI7I.054 Rev 15 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.054 V.B.4 (As available)

Question Source:

Modified Bank# (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OP Li 71 054 Revision 15 Page 27 of 69 (b) Service air supply valve from control air header (0-FCV-33-1). Can be operated from panel 1-9-20 and!or 3-9-20. The switch positions are CLOSE-AUTO-OPEN, with position indication lamps just above each control switch. The valve automatically opens if control air pressure falls to 85 psig and closes at 30 psig (due to insufficient air pressure to keep the valve open).

(C) A manual bypass valve can be utilized if 0-FCV-33-1 should fail to open.

c. System Annunciators ARP usage (1) AIR COMPRESSOR ABNORMAL alarm on PANEL 1-9-20 ONLY. Any alarm annunciated on panel 0-LPNL-925-01 18 (2) SERVICE AIR XTIE \!ALVE OPEN alarm PANEL 1-9-20 and 3-9-20 (PCV 33-i opens at 85 psig)

(3) CONTROL AIR PRESSURE LOW alarm on each units 9-20 panel at 70 psig (4) CONTROL AIR DEW POINT HIGH at -20F on 2-9-20 panel and -28.9CC on 1-9-20 panel.

(5 Two local alarms annunciate to indicate primary controller failure or backup controller failure due to loss of power to controller or software failure.

d. Control Room Indication Panel 9-20 Unit Control Air Header Pressure

e. G Air Compressor amps are indicated on panel 1-9-20.

f. A-D Control Air Compressor Normal Operating Parameters (1) Operator Setpoints screen (a) Lead Offline Pressure 60-128 psig (b) Lead Online pressure 50-118 psig (c) Lag Offset 0-45 psig (d) Load time delay 0-60 seconds (e) Condensate Interval 60-360 seconds (f) Condensate discharge time 2-20 seconds (g) Max. First State Temperature 300-440°F

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Air System O.Ol32 ijnitO Rev.0127 Page 67 of 113 Illustration I (Page 1 of 1)

Control Air System Pressure Spectrum

+ 132 Cornpreooor A 3, 2,3 Oisdargu RohufUalruo lint.

in its Control Ar Roceiner RntofVatuuo 15tCumprensor 0 Odarn R.ntnfVahe lint 33 C.oropro:oorr no, t,r2, irl LEPO Lint iriHL2.E on pnll.

F. C. 1itt.°C act 0 tlVicF C6

&4 .rpo;.: F..D cOc*F t,lO 04 jr Crcn 6 S&ic, 0 0.FCV.33.i, npnc to t.id<op Cornto4,°O 6ntnrn.

2 Mn Stn3rn RIVii.nIOt currotor Proojrc Lno ar ur.nitno, (Por.n 2.03D Un20Jnit3) C:AriCorcUnn FOn-o S&nct. 2. 3.PC63.65o otirt- Ntr ntrrn CL 76 Thrk Atn2.3-F41tr&i-2l.

73 Controt It Ernnrenci Coonpreonirotirto.

70 trot Pr Prnrscore icco annunda nn(Parrnl 1339-2Q0Of2I ant FarnI 0 L 01.2501 16A1.

2 65 5cram PitnotAir Ha4er Pccncure Lout annurciann(andui23.8-525.

65. rjrit2 tot Jnit3 Cr P,2.PCV.332.OO3Cti, cnna; (ccno-ron at&5 pni, Unit2(UrtS Stanton Bp20 cnntroiVatou. 2%FCO0362. Loin an to. Thu nalou cannot loot cotnollud tyPD3 orAl aoouoot tothorut to 0.55 por, 60 Unit2(U cit 3) Cornduncatot Outrun Orpaou n 22CV-2-120.loiIo opun 4 312 ir ocern, herrucu too Crosohe. O-FC\) I. cmr,on.

(Thin liotration ctepicA some utttro plant postru I In a dupartumafrnrrr normal Ccrrtrol tr SyoIutm I psor ( 1)0 punt) Rung aod ioucrin1 prenounen

\,_orn isdruatnd mnct

• C CA CorcrprOotcn F and CA Cnrrprunr or C- trip on unseal air pr000norO.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT BFN Control Air System 0-01-32 UnitO Rev. 0127 Page 67 of 113 Illustration I (Page 1 of 1)

Control Air System Pressure Spectrum

+ 122 Comprrnoorri, 9. S.D Diodrare Relief Vaborn lift.

115 Control Air Recwkrer Relict Valoca bot;Comprcon 6 Disotrargo Rebel Vain; lifts.

6 C(ropio ;;nro °, 1,.:, Dii 1600 LOu tJtl100 on

+ 86 Aii33rIP 0,2. 8 Ii 11001. 0. La.t rIL3F0 Ci poll.

• on CoropF;ocr: , 1.0 0 ii ;&i>id 1°C moo .OOloo°0 ill p;lj.

-4 86 SerAcoAi101ccntie,C.FC°.i323.1, tpenn tohadrop CoolAlSpotorn 82 Mino Steam lieu Vatooinir Accoirnlator Pro;orire Low arrro000iaboo, (Panel; 2,5838).

Unt 2 IJcoft2) CAE/Cootolinur Flow Soled, 2, 3.PC\OS3664 aroroc rintrogo o from CAt 4 75 lark Mo 2,3-F Srj-61-rn aod-2 1.

• 73 Control Bay BrcJordecocy Compresoor ntarfo.

700 ontrclOrrFrrn;reLwoanrnoodatioctanel12,59.2oJcOr52)aodParreIL?tlb26011Sin.

66 Scram PuiotAo H eater Preonore tow annorciateo rPaodeu2A4S2131.

66. Urit2toUrrrt2 Croccbe,2PCV032-58D1, oboe; (renoero; at&op;io3.

Ue62QJniti Stadop G3pas; corrtnotriatoe, 23FCu353. fail; -a; to. Tb; nato.;

cannot tie ;or,bolbed by liDS wrIt prfo$rire retoorc to 065 per5, 4 50 Urot 2(Ur,itS3 C orrtoocate Fremin Bypaos oato;,23-FCU-2130.tafo open.

4 50 6 open. Senor;; Ar Crooote. (r-FCV I. cbweo.

Trot tlroitration depict room; -,tt3o plant rrnpnroon to a do pa rtor e from normal C cord rot Air S-Atom pr0000ro& -IOU arØ. lining and bomerre poorororec are 4frrdioatotbytant rorrpectfoeby.

6 56 Ccorrprocser F arid CA Conropresoror C tip c-cr looooeal air pwo000re.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT OPL Ill .u47 Revision 12 Page 12of41

5. Motor Operated \/alves (MOV5)

a. The spare RBCCW Pump has $ MOVs Right Unit. Right which can be used to line it LI to Unit 1 2. train, Right or 3. Component (1) These MOVs a.re controlled from INPC) SER 30-05 Panel 9-4. Unit 1 Control Room. Attention to Detail and Intrusiveness (2) These MOVs are inierlocked to permit alignment of the spare pump tu only ROTORK valves have open, one unit at a time, to prevent cross-tying of the Unit RBCCW Systems. closed, and mid position. Mid does

b. FCV-70-48 controls the RBCCW supply to not indicate a Sb of the non-essential equipment loop. valve open or (Referred to as the SECTIONALIZING closed valve)

(I) Ul/2 ECV-70-48 automaticafly closes Obj. V 5.4 on:

(a) Initiation of LII /2 460V Load C)hj. Vii Shed Log ic,(Loss of normal AC (Phi. V.D.5 power with any Ui/2 diesel generator tied to a U1!2 4kV UNIT shutdown board as a sole DIFFERENCE source. in conjunction with an Unit 3. 70-48 accident signal) closes on low (CAS signal 2.45 psig OW press plessure as a with 4513 psig Rx press, or -122 result of the pump Level) trips from load shed.

0) All three units FCV-70-48 close on low RBCCW supply header pressure of 57 psig (corresponds to an actual header pressure of 50 psig)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295021 Loss of Shutdown Cooling /4 Tier # 1 AK2.O1 (10CFR 55.41.7)

Knowledge of the interrelations between LOSS OF SHUTDOWN Group # 1 COOLING and the following: K/A # 295021AK2.O1

• Reactor water temperature Importance Rating 36 Proposed Question: #9 Unit 3 is in Mode 4 with the following conditions:

• Reactor Level band is (+) 78 inches to support testing.

• ALL Reactor Recirc AND RWCU Pumps are isolated and tagged out.

• RHR Loop I in Shutdown Cooling experiences an inadvertent Group 2 Isolation AND can NOT be restored.

In accordance with 3-AOI-74-1, Loss of Shutdown Cooling, which ONE of the following completes the statements?

Accurate Reactor Water Temperature _(1 )_ available.

If Reactor Coolant Stratification occurs, it is indicated by _(2)_.

A. (1)is (2) Reactor pressure GREATER THAN 0 psig with any Reactor Coolant temperature indication AT OR BELOW 21 2°F B. (1)1sNOT (2) Reactor pressure GREATER THAN 0 psig with any Reactor Coolant temperature indication AT OR BELOW 21 2°F C. (1)is (2) Differential temperatures of 40°F between Reactor Vessel Bottom Head AND any Reactor Vessel Feedwater Nozzle D. (1)isNOT (2) Differential temperatures of 40°F between Reactor Vessel Bottom Head AND any Reactor Vessel Feedwater Nozzle Proposed Answer: B Explanation A INCORRECT: Part 1 incorrect Plausible in that Reactor Level is high (Optional): enough to establish natural circulation. Candidate may believe natural circulation is adequate to provide accurate level indication. Part 2 correct See explanation B.

B CORRECT: Part 1 correct In accordance with Loss of Shutdown Cooling, 3-AOl-74-1, accurate coolant temperatures will not be available if forced circulation is lost. Part 2 correct Reactor Coolant Stratification is indicated by Reactor pressure> 0 psig with any Reactor Coolant temperature indication <212°F

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet c INCORRECT: Part 1 incorrect See explanation A. Part 2 incorrect Plausible in that in accordance with Loss of Shutdown Cooling, 3-AOl 1, with the Reactor in Cold Shutdown Condition (Mode 4 or Mode 5) coolant stratification may be indicated by Differential temperatures of> 50°F between Reactor Vessel Bottom Head AND any Reactor Vessel Feedwater Nozzle.

D INCORRECT: Part 1 and 2 incorrect as explained above.

KA Justification:

The KA is met because to successfully answer the question, the candidate must demonstrate knowledge of the interrelationship between loss of shutdown cooling and Reactor Water Temp.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): OPL171.074 Rev 8 (Attach if not previously provided) 3-AOI-74-1 Rev 19 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.074 V.B.6 (As available)

Question Source: Bank #

Modified Bank # BFN 1006 #9 (Note changes or attach parent)

New Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of Shutdown Cooling 3.AOI74-1 Unit3 Rev.0O19 Page 9 of 26 4.2 Subsequent Actions (continued)

NOTES

1) With the Reactor in Cold Shutdown Condition (Mode 4 or Mode 5), reactor coolant stratification may be indicated by one of the following:

a Reactor pressure above 0 psig with any reactor coolant temperature indication reading at or below 212°F.

a Differential temperatures of 50°F or greater between either RX VESSEL BOTTOM HEAD (FLANGE DR LINE) 3-TE-56-29 (8) temperatures and RX VESSEL FW NOZZLE N48 END (N4B INBD)(N48 END)( 4D INBD) 3-TE-56-13(4)(15)(16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder, 3-TR-56-4.

With recirculation pumps and shutdown cooling out of service, a Feedwater sparger temperature of 200°F or greater on any RX VESSEL FW NOZZLE (N4B END (N4B INBD)(N4D END)(N4D INBD> 3-TE-56-13(14)(15)(16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder. 3-TR-56-4,

2) (NERIC: For purposes of thermal stratification monitoring, the bottom head drain line is more representative as long as there is flow in the line. [6E81L251 and 430]

[6] PLOT heatup/cooldown rate as necessary. REFER TO 3-SR-3.4.9.1(1).

[7] REQUEST the SRO to ESTIMATE the following times at least once per shift until a method of decay heat removal is restored:

[71) DETERMiNE the time since shutdown. D

[7.2] DETERMINE the current RPV heat-up rate from 3-SR-3.4.9.l1), or. if reactor coolant stratification is suspected, use Illustration 1.

[7.2.1] IF additional information is required to determine the heat-up rates, THEN NOTIFY Reactor Engineer. D

[7.3] DETERMINE the reactor coolant temperature or use the last valid reactor coolant temperature available. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of Shutdown Cooling 3-AOI-74-1 Unit3 Rev. 0019 Page 13 of 26 4.2 Subsequent Actions (continued)

CAUTION Accurate coolant temperatures will NOT be available if all forced circu[ation is lost.

[13] ric: IF forced circulation has been lost AND vessel cavity is less than 80 inches, THEN PERFORM the following: (Otherwise N/A)

[13.1] RAISE RPV water level to 80 inches as indicated on RX WTR LEVEL FLOOD-UP, 3-Ll-3-55.

[13.2] MAINTAIN RP\! water level between +70 inches to

+90 inches as indicated on RX WTR LE\JEL FLOOD-UP. 3-Ll-3-55.

[13.3] RAISE monitoring frequency of reactor coolant temperature and pressure. using multiple indications. D

[14] IF the affected loop of RHR cannot be placed back in Shutdown Cooling. THEN PLACE the alternate loop of RHR in Shutdown Cooling.

REFER TO 3-01-74. (Otherwise N/A) D

[15] IF no Unit 3 RHR loop can be placed in Shutdown Cooling.

THEN OBTAIN Shift Manager approval and PLACE Unit 2 RHR loop in service, CROSS-TIED with Unit 3, for Shutdown Cooling.

REFER TO 3-01-74. (Otherwise N/A)

[16] IF no RHR loops can be placed in service. THEN VERIFY a Recirculation Pump in service.

REFER TO 3-01-68. (Otherwise N/A)

[17] IF the Reactor is in a Cold Shutdown Condition (Mode 4 or Mode 5) AND the reactor vessel head studs are tensioned or head tensioning is in progress. THEN PERFORM 3-SR-3.4.9.5-7, RPV Head Temperature Monitoring. (Otherwise N/A) D

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet OPL1J1 .074 Revision S Page7of t6 INSTRUCTOR NOTES C. 1/2/3 AOl-IOU-i, Reactor Scram

1. The reactor scram AOl-i 00I pnvides guidance NOTE: Immediate regarding immediate operator actions required to Operator Actions are to stabilize the plant in the areos of controlling and be penormed from monitoring reactor power, level, and pressure. iliemory.

2. The subsequent actions provide guidance for long term stabilization and recoverj of both RP\/ and balance-of-plant parameters.

3. The following subsequent action sections should he studied in detail:

a) Actions to stabilize Reactor power. level, and pressure h) Veitfication of all rods fully inserted c Actions to secure the Main Generator and Turbine d) Resetting the scram and PCIS Obj. V.8.3 Obj. V.8.4 e) Scram Report (Attachments 1-3)

[I. i/2/3AOl74-i, Loss of Shutdown Cooling

1. This instruction provides the symptoms and operator actions for a Loss of Shutdown Cooling.

2. Accurate coolant temperatures will not he available if all forced circulation is lost.

3. Reactor vessel stratification may occur until Shutdown Cooling is restored or a Reactor Recirculation Pump is placed in senice.

4. With the reactor in Cold Shutclcnvn Condition (Mode 4 or Mode 5), reactor coolant stratification may he indicated by one of the following:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.074 Revision S Page Sof 16 iNSTRUCTOR NOTES a) Reactor pressure above 0 psig with any Obj. V.B.5 reactor coolant temperature indication reading at or below 212°F.

b) Differential temperatures of 50°F or greater Obj. \.B.S between either RX VESSEL BOTTOM HEAD (FLANGE DR LINE) 112r3-TE-56-20 (6) temperatures and RX VESSEL FW NOZZLE N4B END (N4B INBD) (N4B END) (N4D INBD) 1 2?3TE-bO-lJ(14) (15) (it) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder. i/2,3-TR-56-4.

c) With recircLilation pumps and sutdown coohng out of service, a Feedwater spa rger temperature of 200°F or greater on any RX VESSEL FW NOZZLE (N4B END) (N4B INBD) (N4D END) (N4D INBD) l!2/3-TE 13(14) (15) (16) temperatures from the REACTOR VESSEL METAL TEMPERATU RE recorder. 1 /2/3-TR-56-4.

5. For purposes of thermal stratification monitoring, the bottom head drain line is more representative as long as there s flow in the line. {GE SIL 251 and 430]

6. IF forced circulation has been lost and vessel cavity is less than 80 inches, THEN, RAISE RPV water level to 80 inches as indicated on 11213-Ll-3-55

7. Maintain RPV water level between ÷70 inches to ÷00 inches as indicated on RX WTR LEVEL FLOOD-IJP.

l2/3 3-55.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Loss of Shutdown Cooling 3.AOI.74.1 Uriit3 Rev. 0019 Page 9 of 26 42 Subsequent Actions (continued)

NOTES With the Reactor in Cold Shutdovn Condition (Mode 4 or Mode 5), reactor coolant stratification may be indicated by one of the following:

. Reactor pressure above 0 psig with any reactor coolant temperature indication reading at or below 212°F.

. Differential temperatures of 50°F or greater between either RX \/ESSEL BOTTOM HEAD (FLANGE DR LINE) 3-TE-56-29 (8) temperatures and RX VESSEL FW NOZZLE N4B END (N4B INBD)(N4B END)( 4D INBD) 3-TE-56-i 3(1 4)(1 5)(1 6) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder, 3-TR-56-4.

. With recirculation pumps and shutdown cooling out of service, a Feedwater sparger temperature of 200°F or greater on any RX VESSEL FW NOZZLE (N4B END (N46 INBD)(N4D END)(N4D INBD) 3-TE-56-13(14)(15)(16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder, 3-TR-56-4.

2 :NER!c] For purposes of thermal stratification monitoring, the bottom head drain line is more representative as long as there is flow in the line. [GESL 251 nG43OJ

[6) PLOT heatup/cooldown rate as necessary. REFER TO 3-SR-3 .4 .9. 1( 1). D

[7] REQUEST the SRO to ESTIMATE the following times at least once per shift until a method of decay heat removal is restored:

[7.1] DETERMINE the time since shutdown. C

[7.2) DETERMINE the current RP\/ heat-up rate from 3-SR-3.4.9.i(1), or. if reactor coolant stratification is suspected. use Illustration 1.

[7.2.1] IF additional information is required to determine the heat-up rates. THEN NOTIFY Reactor Engineer. C

[7.3] DETERMINE the reactor coolant temperature or use the last valid reactor coolant temperature available,

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL171 .074 Revision 3 PageS of 16 INSTRUCTOR NOTES a) Reactor pressure above 0 psig with any Obj. V.8.5 reactor coolant temperature indication reading at or below 212F.

b Differential temperatures of 50°F or greater Obj. V.8.6 between either RX VESSEL BOTTOM HEAD (FLANGE DR LINE) 1i2!3-TE-56-29 (8) temperatures and RX VESSEL FW NOZZLE N4B END (N4B INBD) (N4B END) (N40 JNBD) 11213-TE-56-13(14) (15) (16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder. i12/3-TR-56-4.

c) Vith i-ecirculation pumps and shutdown cooling out of service, a Feeciwater sparger temperature of 200°F or greater on any RX VESSEL EW NOZZLE (N4B END) (N4B INBD) (N4D END) (N4D INBD) 1!2/3-TE 13(14) (15) (16) temperatures from the REACTOR VESSEL METAL TEMPERATURE recorder. 112/3-TR-5h-4.

5. For purposes of thermal strat[fication monitoring, the bottom head drain line is more representative as long as there is flow in the line. {GE SIL 251 and 430]

6. IF forced circulation has been lost arid vessel cavity is less than 80 inches, THEN, RAISE RPV water level to 80 inches as indicated on 11213-Ll-3-55.

7. Maintain RPV water level between ÷70 inches to ÷90 inches as indicated on RX WTR LEVEL FLOOD-IJP.

1/213 3-55.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 NRC #9 Examination Outline Cross-reference: Level RO SRO 295021 Loss of Shutdown Cooling /

Tier # 1 AK2.O1 (100FR55.41.7)

Knowledge of the interrelations between LOSS OF SHUTDOWN Group # 1 COOLING and the following:

K/A # 295021AK2.O1

• Reactor water temperature Importance Rating 3.6 Proposed Question: #9 Unit 3 is in Mode 4 with the following conditions:

• Reactor Level band is (+) 70 to (+) 80 inches to support testing

• ALL Reactor Recirc AND RWCU Pumps are isolated and tagged

• RHR Loop I in Shutdown Cooling experiences an inadvertent Group 2 Isolation AND can NOT be restored In accordance with 3-AOl-74-1, Loss of Shutdown Cooling, which ONE of the following completes the statements?

Accurate Reactor Water Temperature (1) available.

If Reactor Coolant Stratification occurs, it is indicated by (2)

A. (1)is (2) Feedwater Sparger temperature GREATER THAN OR EQUAL TO 200°F on any Vessel Feedwater Nozzle indication B. (1)isNOT (2) Feedwater Sparger temperature GREATER THAN OR EQUAL TO 200°F on any Vessel Feedwater Nozzle indication C. (1)is (2) Reactor pressure GREATER THAN 0 psig with any Reactor Coolant temperature indication GREATER THAN 212°F D. (1)is NOT (2) Reactor pressure GREATER THAN 0 psig with any Reactor Coolant temperature indication GREATER THAN 212°F

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRQ 295023 Refueling Acc/ 8 Tier # 1 AAI.03 (10CFR55.41.7)

AbHity to operate and/or monitor the following as they apply to Group # 1 REFUELING ACCIDENTS: K/A # 295023AA1 .03

• Fuel handling equipment Importance Rating 3.3 oposed Question: # 10 Unit 1 is in a Refueling Outage. The Refueling Supervisor reports that an IRRADIATED fuel assembly has been seated in the WRONG location in the core. The grapple remains engaged on the bundle.

The following conditions are then noted:

• Rising count rates on SRMs

• SRM Period lights illuminated o Rising dose rates on the Refuel Floor Which ONE of the following describes an IMMEDIATE Operator action in accordance with Refueling AOIs?

A. Verify Secondary Containment is intact.

B. If any CRD Pump is in service stop the CRD Pump.

C. Raise the fuel bundle from the core location AND traverse to the area of the cattle chute.

D. If SLC is operable place SLC PUMP IA/I B, 1-HS-63-6A control switch in START A OR START B.

Proposed Answer: C Explanation A INCORRECT: This is plausible because it is a required subsequent action (Optional): of 1-AOI-79-1, Fuel Damage During Refueling.

B INCORRECT: This is plausible because it is a required subsequent action of 1-AOI-79-2, not immediate action.

C CORRECT: In order to answer this question correctly the candidate must determine the appropriate condition and Immediate Action required by 1-AO 1-79-2.

D INCORRECT: This is plausible because it is a required subsequent action of 1-AOI-79-2, not immediate action.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

This question satisfies the K/A statement by requiring the candidate to analyze specific plant conditions to determine appropriate actions to take with fuel handling equipment in response to inadvertent criticality.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Candidate must recognize that inadvertent criticality has occurred based on indications and select appropriate immediate actions.

Technical Reference(s): 1-AOl-79-2 Rev. 0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .060 V.B.3 (As available)

Question Source: # BEN 1006 #10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam BEN 2010 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CER Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Inadvertent Criticality During Incore lAOl-79-2 Unit I Fuel Movements Rev. 0000 Page 6 of 9 4.0 OPERATOR ACTIONS 4.1 lmmediate Actions

[1] IF unexpected criticality is observed following control rod withdrawal, THEN REINSERT the control rod. C

[2] IF all control rods can NOT be fully inserted, THEN MANUALLY SCRAM the Reactor. C

[3] IF unexpected criticality is observed following the insertion of a fuel assembly, THEN PERFORM the following:

[3.1] VERIFY fuel grapple latched onto the fuel assembly handle AND IMMEDIATELY REMOVE the fuel assembly from the Reactor core.

[3.2] IF the Reactor can be detennined to be subcritical AND no radiological hazard is apparent, THEN PLACE the. fuel assembly in a spent fuel storage pool location with the least possible number of surrounding fuel assemblies and LEAVE the fuel grapple latched to the fuel assembly handle.

[3.3] IF the Reactor can NQT be determined to he subcrtical OR adverse radiological conditions exist, THEN TRAVERSE the Refueling Bridge and fuel assembly away from the Reactor core, preferably to the area of the cattle chute and CONTINUE at Step 4.1 [4]. C

[4] IF the Reactor can NOT be determined to he subcritical OR adverse radiological conditions exist, THEN EVACUATE the refuel floor. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Fuel Damage During Refueling 1-AOl-79-1 Unit 1 Rev. 0000 Page 6 of 9 4.0 OPERATOR ACTIONS 4.1 Immediate Actions I] STOP all fuel handling.

[2] EVACUATE all non-essential personnel from Refuel Floor.

4.2 Subsequent Actions CAUTI ON The release of IODINE is of major concern. If gas bubbles are identified at any time. Iodine release should be assumed until RADCON determines otherwise.

[1] VERIFY Secondary Containment is intact. REFER TO Tech Spec 3.G.4.i.

[21 IF any EOl entry condition is met. THEN ENTER the appropriate ECI(s).

[3] VERIFY automatic actions. 0

[4] NOTIFY RADCON to perform the following:

EVALUATE the radiation levels.

• MAKE recommendntion for personnel access. 0

• MONiTOR around the Reactor Building Equipment Hatch at levels below the Refuel Floor for possible spread of the release.

151 REFER TO EPIP-i for proper notification. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Inadvertent Criticality During Incore 1-AOl-79-2 Unit I Fuel Movements Rev. 0000 Page 7 of 9 4.2 Subsequent Actions

[1] NOTIFY the Shift Manager and Reactor Engineer. D

[21 IF any EQi entry condition is met. THEN ENTER the appropriate ECIs.

[3] VERIFY all control rods are inseiled. 0

[4] IF criticality is still evident AND at the direction of the LI nit Supervisor. THEN PERFORM the following:

[4. I] IF the CRD pump is in operation, THEN STOP the CRD pump.

[4.2] IF the RWCU system is in service, THEN ISOLATE RWOLJ as follows.

[4.2.11 CLOSE 1-FCV-069-0001 using RWCLJ INBD SLICT ISOLATION \/ALVE. 1-HS-69-t C

[4.2 2] CLOSE l-FCV-069-0092 using RWCU OLJTBD SIJCT ISOLATION \fAL\!E. 1-HS-69-2A.

[4.3] IF SLC is operal:le, THEN UNLOCK and PLACE SLC PUMP 1AJ1B 1-HS-63-6A control switch in START A or START B. C

[5] NOTIFY RADCC)N to conduct surveys to determine rachation levels on Refuel Floor. C

[6] NOTIFY Chemistry to sample and analyze the Reactor water C

[7] REFER TO EPIP-1 for proper notifications. C

[8] NOTIFY NRC REFER TO SPP-3.5. C

[9] NOTIFY Plant Manager. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0610 NRC RO EXAM

49. RO 25u2.AK1 .02 n0 fA. GL702VJ>. .W20O2AK1 .02;ROSRO:MOD1HcD Fuel loading is in progress on Unit 1 when you notice an unexplained rise in Source Range Monitor (SRM) count rate and an indicated positive reactor period.

Which ONE of the following actions is an appropriate response?

A. Immediately EVACUATE all personnel from the refuel floor.

B. If unexpected criticality is observed following control rod withdrawal, manually SCRAM the reactor.

C. If the reactor cannot be determined to be subcritical, traverse the refueling bridge and fuel assembly away from the reactor core, preferably to the areu of the cattle chute.

9. If all rods are not inserted/cannot be inserted, verify the fuel grapple is latched onto the fuel assembly handle and immediately remove the fuel assembly from the reactor core.

K/A Statement:

295023 Refueling Acc Cooling Mode / 8 AK1 .02 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS; Shutdown margin K/A Justification: This question satisfies the K/A statement by requiring the candid ate to analyze specific plant conditions to determine a reduction in Shutdown Margin has occurred and the actions required to address that condition,

References:

1 -AOl-79-2 Levet of Knowledge Justification: This question is rated as C/A due to the requirement to assemble. sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

0610 NRC Exam MODIFIED FROM OPLI71.060 #1 FricJy, February 20. 2008 3:0t06 AM 104

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. The appropriate condition and Immediate Action required by iAOl-79-2 C correct:

A incorrect: This is plausible because the evacuation of the Refuel Floor MAY be directed, but other actions to mitigate the problem take precedence until personnel safety is compromised.

S incorrect: This is plausible because the condition is correct, but the action to scram is incorrect, Reinserting the control rod is required.

D incorrect: This is plausible because the required action is correct, hut the condition is NOT correct. This action is based on unexplained criticality following insertion of a fuel assembly.

rrfrv, Ferruey 20, 2008 301:O6 AM 105

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295024 High Drywell Pressure /

Tier # 1 EK3.08 (10CFR55.41.5)

Knowledge of the reasons for the following responses as they apply Group # 1 to HIGH DRYWELL PRESSURE: K/A # 295024EK3.08

• Containment spray Plant-Specific Importance Rating 37 Proposed Question: # 11 Unit 2 was at 100% Reactor Power when a spurious Group I Isolation occurred. The pressure transient caused a small-break LOCA to occur inside the Drywell.

Which ONE of the following describes the basis for actions with respect to 12 psig Suppression Chamber Pressure?

A. Drywell sprays must be initiated prior to this pressure to prevent opening the Suppression Chamber to Reactor Building vacuum breakers AND de-inerting the containment.

B. Drywell sprays must be initiated above this pressure because almost ALL of the nitrogen AND other non-condensable gases. in the drywell have been transferred to the torus AND chugging is possible..

C. Above this pressure indicates that almost ALL of the nitrogen AND other non-condensable gases in the torus have been transferred to the drywell air space AND Suppression Chamber Sprays will be ineffective.

D. Above this pressure indicates that almost ALL of the nitrogen AND other non-condensable gases in the drywell have been transferred to the torus so initiating Drywell Sprays may result in containment failure.

Proposed Answer: B Explanation A INCORRECT: This is plausible because initiation of DW sprays at high SC (Optional): pressure could reduce pressure low enough to open the Suppression Chamber to Reactor Building Vacuum Breakers. However, this is part of the bases for the Drywell Spray Initiation Pressure Limit Curve #5.

B CORRECT: Drywell sprays must be initiated above this pressure because almost all of the nitrogen AND other non-condensable gases in the drywell have been transferred to the torus AND chugging is possible. The basis for the Pressure Suppression Pressure Limit of 12 psig Suppression Chamber pressure.

C IN CORRECT: This is plausible if the LOCA occurred inside the Suppression Chamber and NOT the Drywell as given in the stem.

D INCORRECT: This is plausible because initiating SC sprays with high temperature non-condensable gases in the SC will result in evaporative cooling and a rapid pressure drop. However, the SC to DW vacuum relief system is capable of compensating for this pressure drop. This is also part of the bases for the Drywell Spray Initiation Pressure Limit Curve #5.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because it tests knowledge of the reasons for Drywell Spray as it applies to High Drywell Pressure.

Question Cognitive Level:

This question is rated as Memory due to the requirement to recall or recognize discrete bits of information.

Technical Reference(s): EOIPM Section 0-V-D Rev. 0 (Attach if not previously provided)

OPLI 71 .203 Rev. 7 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.203 V.B.5 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Corn ments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.203 Revision 7 Page 25 of 7

1. Step PC/P-2

a. This decision step has the operator Questioning evaluate present and future performance Attitude of venting the drywell or suppression chamber using CAD and SGTS, in relation to the current value and trend of drywell and suppression chamber pressure, to determine if primary containment pressure can be maintained below the high drywell pressure scram setpoint.

b. If primary containment pressure can be maintained below the high drywell pressure scram setpoint, the operator returns to Step PC/P-i until EOl-2 can be exited or primary containment pressure cannot be maintained below the high drywell pressure scram setpoint.

c. If containment pressure control systems are unable to maintain primary containment pressure below the high drywell pressure scram setpoint, then further control actions, beginning at Step PC/P-3, need to be addressed.

2. Step PC/P-3 Obj.V.B.5, V.C.5

a. This before decision step has the operator evaluate present and future performance of venting the drywell orsuppression chamber using CAD and SGTS, in relation to the current value and trend of suppression chamber pressure, to determine if suppression chamber pressure can be maintained below 12 psig (Suppression Chamber Spray Initiation Pressure).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.203 Revision 7 Page 26 of 7

b. Engineering calculations have determined that if suppression chamber pressure exceeds 12 psig, Suppression Chamber Spray Initiation Pressure, there is no assurance that chugging will be prevented at downcomer openings of the drywell vents.

c. Suppression Chamber Spray Initiation Pressure is defined to be the lowest suppression chamber pressure that can occur when 95% of the noncondensables in the drywell have been transferred to the airspace of the suppression chamber.

d. Scale model tests have demonstrated that chugging will not occur so long as the drywell atmosphere contains at least I %

no nco ndensables.

e. To prevent the occurrence of conditions under which chugging may happen, the Suppression Chamber Spray Initiation Pressure is conservatively defined by specifying 5% noncondensables.

f. Chugging is the cyclic condensation of Obj.V.B.6a steam at downcomer openings of the Obj.V.C.6a drywell vents. Chugging occurs when SER 03-05 steam bubbles collapse at the exit of the downcomers. The rush of water that fills the void (some of which is drawn up into the downcomer pipe) induces a severe stress at the junction of the downcomer and vent header.

g. Repeated application of this stress can SER 03-05 cause these joints to experience fatigue failure, thereby creating a pathway that bypasses the pressure suppression function of primary containment.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.203 Revision 7 Page27 of 7

h. Subsequent steam that discharges through the downcomers would then exit through fatigued cracks and directly pressurize the suppression chamber air space, rather than discharging to and condensing in the suppression pool.

3. Step PC/P-4

a. This action step directs the operator to EQI Appendix 17C manually place those pumps not required provides step-by-to assure adequate core cooling, in the step guidance for suppression chamber spray mode. operating RHR in Because this step is prioritized with a the suppression miniature before decision step PC/P-3 chamber spray symbol, this action must be performed mode.

before suppression chamber pressure reaches 12 psig, Suppression Chamber Spray Initiation Pressure.

b. This step only addresses initiation of suppression chamber sprays. Instructions for terminating suppression chamber spray operation, once initiated, are provided by Step PCC-2.

4. Step PC/P-5

a. This contingent action step requires the operator to wait until the stated condition has been met before continuing in EOI-2.

Subsequent actions in this section of EQI 2 will not be performed until suppression chamber pressure exceeds Suppression Chamber Spray Initiation Pressure.

b. Although operation of suppression chamber sprays by itself will not prevent chugging, the requirement to wait to initiate drywell sprays until reaching Suppression Chamber Spray Initiation Pressure assures that suppression chamber spray operation is attempted before operation of drywell sprays.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EOI-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O.V.D

• r DISCUSSJON: STEP PCIP-3 This before decision step has the operator evaluate present and future perfonnance of venting the drywell or suppression chamber using CAD and SGTS, in relation to the current value and trend ofsuppression chamber pressure, to detennine if suppression chamber pressure can be maintained below Suppression Chamber Spray Initiation Pressure. The before decision step requires that this determination and subsequent actions be performed before suppression chamber pressure reaches Suppression Chamber Spray Initiation Pressure.

Engineering calculations have determined that ifsuppression chamber pressure exceeds <A.65>,

Suppression Chamber Spray Initiation Pressure, there is no assurance that chugging will be prevented at downcomer openings of the drywdll vents. This value is rounded off in the BOl to use the closest, most conservative value that can be accurately determined on available instrumentation.

Suppression Chamber Spray Initiation Pressure is defined to be the lowest suppression chamber pressure that can occur when 95% of the noncondensables in the drywall have been transferred to the airspace ofthe suppression chamber. Scale model tests have demonstrated that chugging will not occur so long as the drywell atmosphere contains at least 1% noncondensables. To prevent the occurrence of conditions under which chugging may happen, the Suppression Chamber Spray

• Initiation Pressure is conservatively defined by specifying 5% noncondensables.

Chugging is the cyclic condensation of steam at downeomer openings ofthe drywell vents.

Chugging occurs when steam bubbles collapse at the exit of the downcomers. The rush of water that fills the void (some of which is drawn up into the downcomer pipe) induces a severe stress at the junction of the dowacomer and vent header. Repeated application of this stress can cause these joints to experience fatigue failure, thereby creating a pathway that bypasses the pressure suppression function of primary containment. Subsequent steam that discharges through the downcomers would then exit through fatigued cracks and directly pressurize the suppression chamber air space, rather than discharging to and condensing in the suppression pooi.

Although operation ofsuppression chamber sprays by itselfwill not prevent chugging, initiation before reaching the Suppression Chamber Spray Initiation Pressure assures that this method of primary containment pressure reduction is attempted before the operation of drywell sprays is directed in subsequent steps of the procedure.

Ifsuppression chamber pressure can be maintained below <A.5>, Suppression Chamber Spray Initiation Pressure, the operator returns to Step PC/P-I. If suppression chamber pressure cannot be maintained below Suppression Chamber Spray Initiation Pressure, the operator continues at Step PCIP-4 before suppression chamber pressure actually reaches Suppression Chamber Spray Initiation Pressure.

. REVISION 0 PAGE 37 OF 244 SECTION 04.0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Browns Ferry 0610 #62

1. R0295020AK3.08 1 Unit 2 was at 100% rated power when a spurious Group I Isolation occurred. The pressure transient caused a small-break LOCA to occur inside the Drywell.

Which ONE of the following describes the basis for actions with respect to 12 psig Suppression Chamber Pressure?

A. Drywell sprays must be initiated prior to this pressure to prevent opening the Suppression Chamber to Reactor Building vacuum breakers and de-inerting the containment.

B. Drywell sprays must be initiated above this pressure because almost all of the nitrogen and other non-condensible gases in the drywell have been transferred to the torus and chugging is possible.

C. Above this pressure indicates that almost all of the nitrogen and other non condensible gases in the torus have been transferred to the drywell air space and Suppression Chamber Sprays will be ineffective.

D. Above this pressure indicates that almost all of the nitrogen and other non condensible gases in the drywell have been transferred to the torus so initiating Drywell Sprays may result in containment failure.

Answer: B In order to answer this question correctly the candidate must determine the following:

1. The basis for the Pressure Suppression Pressure Limit of 12 psig Suppression Chamber pressure.

B correct:

A incorrect: This is plausible because initiation of DW sprays at high SC pressure could reduce pressure low enough to open the Suppression Chamber to Reactor Building Vacuum Breakers. However, this is part of the bases for the Drywell Spray Initiation Pressure Limit Curve #5.

C incorrect: This is plausible if the LOCA occurred inside the Suppression Chamber and NOT the Drywell as given in the stem.

D incorrect: This is plausible because initiating SC sprays with high temperature non condensable gases in the SC will result in evaporative cooling and a rapid pressure drop.

However, the SC to DW vacuum relief system is capable of compensating for this pressure drop. This is also part of the bases for the Drywell Spray Initiation Pressure Limit Curve #5.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295025 High Reactor Pressure /

Tier # 1 EAI.04 (IOCFR 55.41 .7)

Ability to operate and/or monitor the following as they apply to HIGH Group # 1 REACTOR PRESSURE: K/A # 295025EA1 .04

• HPCI: Plant-Specific Importance Rating 3.8 Proposed Question: # 12 Unit 1 HPCI is in operation in Pressure Control Mode per 1-EOl Appendix IIC, ALTERNATE RPV PRESSURE CONTROL SYSTEMS HPCI TEST MODE.

• Reactor Pressure is 1050 psig

• 1-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, is in Automatic Which ONE of the following completes the statement below?

To lower Reactor Pressure, the operator is required to use _(1 )_ AND _(2)_ in accordance with 1-EOI Appendix I1C.

A. (1) 1-FCV-73-36, HPCI/RCIC CST TEST VLV, (2) throttle it in the CLOSE direction B. (1) 1-FCV-73-36, HPCI/RCIC CST TEST VLV, (2) throttle it in the OPEN direction C. (1) 1-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, (2) LOWER the setpoint D. (1) 1-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, (2) RAISE the setpoint Proposed Answer: D Explanation A INCORRECT: Plausible in that 1-FCV-73-35, HPCI PUMP CST TEST VLV (Optional): is adjusted in accordance with 1-EOI Appendix 11C to control HPCI pump discharge pressure at or below 1100 psig.

B INCORRECT: See Explanation A.

c INCORRECT: Second Part is incorrect Plausibility based on misconception that lowering setpoint will result in lowering Reactor Pressure.

D CORRECT: Both parts are correct Per 1-EOI Appendix 1 1C, ADJUST 1-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller to control RPV pressure. Raising set point will lower reactor pressure, per the appendix..

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates ability to operate and monitor HPCI in pressure control mode as it applies to high Reactor Pressure.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 1-EOl Appendix 11C Rev. 1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .042 V.B.1O (As available)

Question Source: Hatch 09 #52 (Note changes or attach parent)

Question History: Last NRC Exam Hatch 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 )(

55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN ALTERNATE RPV PRESSURE CONTROL 1-EOI APPENDIX-IIC UNIT I SYSTEMS HPCI TEST MODE Rev. 1 Pace 3 of 3

6. VERIFY proper HPCI minimum flow valve operation as follows:

a. IF ......,......... HPCI flow is above 1200 gpm, THEN VERIFY CLOSED 1..FCV-73-30, HPCI PUMP MIN FLOW VALVE.

b IF HPCI flow is below 600 gpm, THEN VERIFY OPEN i.-FCV-73-30, HPCI PUMP MIN FLOW \ALVE.

7. THROTTLE i-FCV-73-35, HPCI PUMP CST TEST VLV, to control HPCI pump discharge pressure at or below 1100 psig.

8. ADJUST i-FlC-7333, HPCI SYSTEM FLOW/CONTROL, controller to control RPV pressure.

9. IF HPCI injection to the RPV becomes necessary, THEN ALIGN HPCI to the RPV as follows:

a. OPEN 1-FCV73-44, HPCI PUMP INJECTION VALVE.

b. THROTTLE i-FCV-73-35, HPCI PUMP CST TEST VLV, to control injection.

c. GO TO EOI Appendix 5D.

LAST PAGE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HATCH 09 #52 HLT 4 NRC Exam

52. 29502562.1.23 001 3 1EO-EOP-l 07-2, ALTERNATE RPV PRESSURE CONTROL is in progress.

o The HPCI system is being used to control reactor pressure.

o The 2E41-R612. HPCI flow controller$ is in automatic, with the setpoint at 3001) gpm.

1 To NCREASE the reactor cooldown rate (COR), the operator is required to use and 3 IEO-EOP-107-2.

A? 2E41-R612, HPCI flow controller, RAISE the setpoint B. 2E41-R612, HPCI flow controller, LOWER the setpoint C. 2E41-F0l 1, Test to CST VLV, throttle it in the CLOSE direction D. 2E41-F0l 1, Test to CST VLV, throttle it in the OPEN direction Description; While HPCI is in pressure control mode with the controller in automatic, per procedure the cooldown rate (CDR) is controlled by throttling 2E41-F008. Test to CST VLV.

3 1EO-EOP-l07-2 specifies that throttling F008 in the closed direction will increase the CDR if the controller is in auto. If the controller is in Manual, throttling FOOR will have minimal effect on CDR. In Manual the CDR is increased by increasing the controller output and decreased by reducing the controller output.

This concept has been difficult for some students to master (which direction to throttle the valve to increase CDR).

A. Correct; see description above.

B. Incorrect, 1st part is correct, 2nd is not correct, opening the valve will reduce the OR.

Plausible if the candidate assumes that opening the valve results in more water flow, which would require more steam flow.

C. Incorrect, 1st part is not correct (wrong valve). 2nd part is correct. Plausible if the student does not remember which valve is throttled to control CDR. The valves (FOOS & FOl 1) are in series and have the same name.

D. Incorrect. 1st part is not correct (wrong valve). 2nd part is not correct. Plausible if the candidate assumes that opening the valve results in more water flow, which would require more steam flow, Friday, May 01, 2009 8:37:21 AM 94

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO SRO 295026 Suppression Pooi High Water Temp /

Tier # 1 EK2 02 (1 OCFR 55 41 7)

Knowledge of the interrelations between SUPPRESSION POOL Group # 1 HIGH WATER TEMPERATURE and the following: K/A # 295026EK2.02 Suppression pool spray: Plant-Specific Importance Rating 3.6 Proposed Question: # 13 Unit 3 has experienced a LOCA AND the following conditions exist:

• Suppression Chamber Pressure is 5 psig

• Suppression Pool level is 14.5 feet

• Drywell Pressure is 7.5 psig

• Suppression Pool Temperature is 200° F

• BOTH RHR Loop I Pumps are in Suppression Chamber I Drywell Spray with Loop flow of 11,500 gpm

• Core Spray Pump 2A flow is 4000 gpm

• NO other ECCS Pumps are running Based on the above conditions, which ONE of the following identifies the EGGS Pump(s), if any, that has (have) sufficient NPSH for continued operation?

[REFERENCE PROVIDED]

A. NONE B. RHR Loop** I Pumps ONLY C. Core Spray Pump 2A ONLY D. Gore Spray Pump 2A AND RHR Loop I Pumps Proposed Answer: B Explanation A INCORRECT: Plausible in that If RHR is plotted for the loop flow and not (Optional): the pump flows, it would be in the unsafe region of Curve 2 making this the correct answer.

B CORRECT: Operating point for RHR Loop I Pumps is within the safe region of Curve 2.

C INCORRECT: Core Spray Pump 2A above the safe region of NPSH Limits Curve 1. Plausible in that if Drywell pressure is used to plot Curve 1, Pump would be operating in the safe region of curve 1 and if RHR is Plotted for Loop flow, it would be in the Unsafe of Curve 2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: RHR Loop I Pumps have adequate NPSH. However, CS Pump 2A does not. Plausible in that if Drywell pressure is used to plot both Curves, all Pumps would be operating in the safe regions and this would be the correct answer.

KA Justification:

The KA is met because the question tests the candidates knowledge of the interrelationship between High Suppression Pool Temperature and RHR Spray Operation.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question and use a reference to solve a problem.

Technical Reference(s): 3-EOl-1 Curve 1 I Curve 2 Rev. 8 (Attach if not previously provided)

OPL171.201 Rev. 7 Proposed references to be provided to applicants during examination: CS NPSH Limit Curve 1 RHR NPSH Limit Curve 2 Learning Objective: OPL171.201 V.B.13 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CURVE I Cs NPSH UMITS 250 230 15 PSIG SAFE

• 1 IOPSIGSAFE*

U 0 210 5PSIGSAFE*

190 0 PSIG SAF 170 0

150 130 110 90 500 1500 2500 3600 4500 CS PUMP FLOW (GPM) *SUPPR CHMSR PRESS CURVE 2 RHR NPSH LIMITS 245 235 225 IOPSKSAPE*

215 5LIGSAFE 205 195 opSr3SAE*

185 175 165 155 145 500 II, 2500 let 4500 Ill 6500 II, 8500 Ill

\

10500 12000 RHR PUMP FLOW IGPM) *$UPPR CHMSR PRESS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.201 Revision 7 Page 4 of 5

a. There are two items to note concerning this table: 1) Except for the Shutdown Floodup instrument, all drywell temperatures are not applicable, because there is very little vertical pipe run in the drywell. This means very little error can be caused by elevated dryweli temperatures (until boiling occurs), and 2)

The MAX SC RUN TEMP is the highest temperature reading which can be obtained from Table 6, Secondary Containment Instrument Runs.

2. Caution #2 See EOl flow Operation of RHR or CS with suction from the charts y Suppr p1 may result in equipment damage if:

• Pump flow is above the NPSH limit (curve I or 2) TP-17/18 OR

• Suppr P1 lvi is below the vortex limit (10 ft.)

a. The NPSH Limit is reached when available SER 03-05 NPSH (NPSHa) equals the NPSH required by the pump vendor (NPSHreq). For use in the EOls, it is helpful to express the NPSH Limit in terms that are recognizable and measurable by the control room operator.

Therefore, the NPSH Limit is calculated as a function of pump flow and suppression pool temperature for selected suppression chamber airspace pressures. To accommodate suppression pool water levels above the minimum LCO water level, suppression chamber airspace pressure is expressed as overpressure in the NPSH Limit. Overpressure is the sum of suppression chamber pressure and the hydrostatic head of water above the minimum LCO water level and must be determined by the operator when using the NPSH Limit.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN 1006 #15 Examination Outline Cross-reference: Level RO 295030 Low Suppression Pool Wtr Lvi Tier # 1 EK1.02 (10CFR 55.41 .8)

Knowledge of the operational implications of the following concepts Group # 1 as they apply to LOW SUPPRESSION POOL WATER LEVEL: K/A # 295030EK1 .02

. Pump NPSH Importance Rating 3.5 Proposed Question: # 15 Unit 3 has experienced a LOCA AND the following conditions exist:

. Suppression Pool Level is (-) 5.5 inches

. Suppression Chamber Pressure is 5 psig

. Drywell Pressure is 10 psig

. Suppression Pool Temperature is 200° F

. RHR Pump 2A flow is 11,500 gpm

. Core Spray Loop II flow is 4,000 gpm

. NO other ECCS Pumps are running Based on the above conditions, which ONE of the following identifies the ECCS Pump(s), if any, that has (have) sufficient NPSH for continued operation?

[REFERENCE PROVIDED]

A. NONE B. RHR Pump 2A ONLY C. Core Spray Loop II Pumps ONLY D. Core Spray Loop II Pumps AND RHR Pump 2A Proposed Answer: C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295028 High Drywell Temperature! 5 Tier # 1 EKI.O1 (10CFR 55.41 .8)

Knowledge of the operational implications of the following concepts Group # 1 as they apply to HIGH DRYWELL TEMPERATURE: K/A # 295028EK1 .01

• Reactor water level measurement Importance Rating 3.5 Proposed Question: # 14 Given the following Unit 2 plant conditions:

• Reactor pressure is being maintained at 50 psig

• Temperature near the water level instrument run in the Drywell is 2200 F

• The Shutdown Vessel Flooding Range Instrument (2-Ll-3-55) is reading (+) 35 inches Which ONE of the following is the HIGHEST Drywell Run Temperature at which the 2-Ll-3-55 reading (+) 35 inches is considered valid?

[REFERENCE PROVIDED]

A 200°F B 250°F C 270°F D. 300° F Proposed Answer: B Explanation A INCORRECT: This is plausible since 200°F is a valid indication; (Optional): however the question calls for the HIGHEST temperature.

B CORRECT: In order to answer this question correctly, the candidate must use EOI Caution #1 to determine operable RPV water level instruments.

C INCORRECT: This is plausible if the candidate interpolates the Caution

1. 1 table, however this is NOT permissible.

D INCORRECT: This is plausible if the candidate uses only Curve 8.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because it tests knowledge of the operational implications of Reactor water level measurement with High Drywell Temperature near the water level instruments runs.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): OPL1 71 .201 Rev 7 (Attach if not previously provided) 2-EOI-1 Rev 12 (Including version! revision number)

Proposed references to be provided to applicants during examination: 2-EOl Caution #1 and Curve 8 Learning Objective: OPL171.201 V.B.13 (As available)

Question Source: BEN 0610 #73 (Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 0610 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EOI-1 PAGE 1 OF 1 RPV CONTROL UNIT 2 BROWNS FERRY NUCLEAR PLANT REV: 12 CAUTIONS CAUTION #1

• AN RPV WATER LVI INSTRUMENT MAY BE USED TO DETERMINE OR TREND LVI LXWLIT READS ABOVE THE MINIMUM 1NDICATE1) LVI ASSOCIATEO WITH THE HIGHEST MAX DW OR SC RUN TEMP.

• IF OW TEMPS, OR SC AREATEMPS (TABIE 6), AS APPLICABLE, ARE OUTSIDE THE SAFE REGION OF CURVES.

THE ASSOCIATED INSTRUMENT 8E UNRELIABLE DUE TO BOILING IN THE RUN, MINIMUM MAX DW RUN TEMP MAX SC INSTRUMENT RANGE INDICATED (FROM XR-64-5O RUN TEMP LVL OR TI-84-52AB (FROM TABLE 6 ON SCALE N/A BELOW 150

,445 N/A 151 T0200 LI$4SA, B EMERGENCY N/A

.155T0 .60 .140 201 T0250

.130 N/A 251 TO 3(X)

.120 N/A 301 T0350 LI4.53 ON SCALE N/A BELOW 150 L1440 +5 N/A 151 TO 200 LI3206 NORMAL 0 TO .60

+15 N/A 201 TO 250 LI4253 +20 N/A 251 TO 300 L14.208A, B, 0, 0 .30 N/A 301 TO 350 L14$2 POST Li3.62A ACCIDENT ON SCALE N/A N/A

.25510 +32

.10 BELOW 1130 N/A

.15 IOOTO 150 N/A SHUTDOWN +20 151 TO 200 N/A LI.355 FL000IJP +30 201 TO 250 010.400 +40 N/A

.50 301 T0350 N/A 465

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CURVE 8 RPV SATURATION TEMP 400 380 U

ACTN 360 REQUIREO 340

,u 300 280 -

. SAFE 260 -

z 240 220

,nO 4 0 50 100 150 200 250 RPV PRESS (PSG) *C0N$t4TABJE 250 P510 TABLE6 SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT SC TEMP EL.EMNTS AND LOCATIONS EL 621 EL 593 EL 565 RWCU HXRM (74.0SF) (74-95C AND 0) (69435A ThRU 0) (69..29F, G. H) 113.56A CF Cf N/A Cf 11-3-588 CF N/A NA 11-3-53 CF CF N/A CF LI460 CF Cf N/A NA 1J4..206 CF CF N/A Cf 11-3-253 Cf CF N/A 114-52 Cf CF CF NfA L1-342A Cf CF CF NfA 11-346 Cf CF N/A NIA 1l4-20.08 Cf CF N/A Cf Ll4-20SCD F CF N/A N/A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.201 Revision 7 Page 32 of 7 Caution #1 See EOI flow charts

a. RPV water level instrument systems sense SER 03-05 liquid level in the vessel downcomer region CRDH injection by measuring differential pressure (dP) prevents between a variable leg water column and a reference leg reference leg water column. The reference notching which leg remains full of water from steam can occur if the condensing in the chamber located at the top reference legs are of the reference leg water column. Excess filled with non-condensate drains back into the RPV. To condensable gas ensure reference leg water remains gas free super saturated a trickle flow of CRDH water is continuously water, then injected into the 4 primary reference legs.

depressurized.

b. When water level in the reactor vessel SER 03-05 lowers, variable leg height of water decreases, sensed dP increases, and indicated RPV water level lowers. The converse occurs when water level in the reactor vessel increases; variable leg height of water increases, sensed dP decreases, and indicated RPV water level increases.

c. Changes in height or density of water in the SER 03-05 instrument reference leg can cause changes in indicated RPV water level. For example: if actual RPV water level is constant at some on-scale value and the instrument reference leg head of water (height and/or density) decreases, sensed dP decreases and indicated RPV water level increases. Under extreme conditions, a high and increasing drywell or containment temperature can decrease the density of water in the reference leg such that the instrument falsely indicates an on-scale and steadily increasing water level even though the actual RPV water level is decreasing and well below the elevation of the instrument variable leg tap.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71 .201 Revision 7 Page 33 of 7

d. It is important to note that the information SER 03-05 presented in Caution #1 is not just a simple accommodation for inaccuracies in RPV water level indication which occur when plant conditions are different from those for which the instruments are calibrated. Rather, the caution defines conditions under which the displayed value and the indicated trend of RPV water level cannot be relied upon.

e. Part B of Caution #1 identifies the limiting SER 03-05 conditions beyond which water in instrument legs may boil. Water in the RPV water level instrument legs is maintained in a liquid state by cooling action of the surrounding atmosphere and pressure in the reactor vessel. Water in the instrument legs will boil, however, if its temperature exceeds saturation temperature for the existing RPV pressure.

f. Boiling is a concern in both horizontal and SER 03-05 vertical reference and variable instrument leg runs. Boil-off from reference leg water inventory reduces the reference head of water, decreases dP sensed by the instrument, and results in an erroneously high indicated RPV water level. Boiling in the instruments variable leg exerts increased pressure on the variable leg side of the dP cell. This effect results in a lower sensed dP and an erroneously high indicated RPV water level.

g. Part B of Caution #1 references the RPV SER 03-05 Saturation Temperature Curve (Curve 8) The RPV Saturation Temperature Curve is generic, based simply on the properties of water. The axis for RPV pressure is plotted from atmospheric pressure to the pressure setpoint of the lowest lifting MSRV. Note that the temperature axis of the RPV Saturation Temperature Curve is not simply drywell temperature. Depending upon the relative location of instrument reference legs and variable legs, indications from monitors near instrument runs must be considered.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7I .201 Revision 7 Page 34 of 7

h. Because BEN does not have the capability of directly reading temperature indications near instrument runs located in secondary containment, the RPV Saturation Temperature Curve (Curve 8) is supplemented with Table 6, Secondary Containment Instrument Runs. Table 6 identifies the temperature elements and general locations for the instrument runs to each RPV water level instrument.

Caution I part B says instruments may be unreliable if Curve 8 is exceeded. This means instruments may continue to be used until and unless erratic indication is observed since momentary excursions (expected in some post LOCA situations) into curve 8 unsafe region will not result in boiling. If, however, indications of boiling are observed then that instrument is unusable until the instrument lines can be cooled and refilled.

j. Part A of Caution #1 allows the operator to The instrument determine if each indicated RPV water level will indicate high range is reliable by being above the Minimum by the amount of Indicated Level for each of a series of this offset instrument run temperature ranges. throughout its Engineering calculations have determined range.

that when indicated RPV water level is above the Minimum Indicated Level, the operator is assured that actual RPV water level is above the instrument variable leg tap, and trends are valid.

k. The Minimum Indicated Level is defined to be SER 03-05 the highest RPV water level instrument indication which results from off-calibration instrument run temperature conditions when RPV water level is actually at the elevation of the instrument variable leg tap. Separate levels are provided for each RPV water level instrument.

The table in Part A is structured to give a Minimum Indicated Level corresponding to several temperature ranges for each of the RPV water level instrument ranges. This yields more usable instrument range than would be available if single values were used.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO R 295030 Low Suppression Pool Wtr Lvi /

Tier # 1 G2 1 31 (10CFR 554110)

Ability to locate control room switches, controls, and indications, and G roup # 1 to determine that they correctly reflect the desired plant lineup. K/A # 295030G2.1 .31 Importance Rating 46 Proposed Question: # 15 Unit 3 was at 100% Reactor Power when a leak from the Torus resulted in Suppression Pool Level of 11.4 feet. Required actions of the EOls have been performed.

Which ONE of the following completes the statement below?

Two minutes after initiating required EOl actions, Wide Range Reactor Pressure Indication(s) available on Control Room Panel(s) _(l) will be (2)_.

A. (1)3-9-5 ONLY (2) stable B. (1)3-9-5ONLY (2) lowering C. (1) 3-9-3 AND 3-9-5 (2) stable D. (1) 3-9-3 AND 3-9-5 (2) lowering Proposed Answer: D Explanation A INCORRECT: Part I incorrect Plausible in that this would be the correct (Optional): answer if the question asked where Narrow Range Pressure indication is available. Part 2 incorrect Plausible in that in accordance with 3-EOI-2, reactor scram is required if Suppression Pool can not be maintained >11.5 feet. Two minutes after the scram, reactor pressure would be stable.

However, this is incorrect since 3-EOI-2 also required ED for this condition.

B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

C INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A.

D CORRECT: Part 1 correct Wide Range Pressure indication is available on both 3-9-3 and 3-9-5. Part 2 correct Per 3-EOI-2, if Suppression Pool Level can not be maintained> 11.5 feet, Reactor Scram and Emergency Depressurization are required.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests candidates ability to locate control room wide range pressure indications, and to determine that they correctly reflect the desired plant lineup which is lowering pressure due to requirement to ED on Low Suppression Pool Level.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): 3-EQ 1-1 Rev. 8 I 3-EQ 1-2 Rev. 8 (Attach if not previously provided)

OPL171.003 Rev. 19 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171 .203 V.B.13 (As available)

Question Source: Bank#:

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet L

V ES L EOI-, V CDNTFOL, L

SPIL-?

3-EQ 1-2 PAG.E 1 OF 1 PR[MARY CONTAINMENT CONTROL UNIT 3 BROWNS FERRY NUCLEAR PLANT REV: B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 7 1.003 Revision 19 Page :39 of 88 INSTRUCTOR NUT_S (a) Provides the high reactor vessel TP-6 pressure signal (1148 psig) to initiate an ATiNS

. Also see PiP-95-71.

ARl RPT. this PIP gives Inslr Racks, local panels, (b) Provides input for opening logic instr.#,

for all SRVs. Uses slave relays Master/Slave trip (4 per SRV) set at 1135, I 145, T\/A (GE), panel in or 1155 psig in a 2 of 2 once AIR, function, &

logic to open SRV. power supply.

(c) Provides input to EHC for A and C Reactor Pressure Control CR B and U (d) Provides pressure indication on the .ATU cabinets (9-83. 9-84.

9-85. 9-86).

te) Pressure input is from Steam Ohj. V.8.14, V.D.8, space V.E.4 (2) PT-3-22-AA, -BB, -C, -D (a) Provide the reactor vessel high pressure (1073 psig) signal to RPS for reactor sci-am.

(Li) Piovide reactor pressure indication on the ATLI cabinets (9-83, 9-84. 9-85. 9-86).

(c) Pressure input is from the Dbj. V.8.14. V.D.8, Steam space V E.4 (3) PlS-3-22A and B:

(a) Trip mechanical vacuum pumps if reactor pressure is >800 psig and condenser vacuum is >22 inches Hg.

(b) Pressure input is fnm Steam Chj. V.8.14, V.D.8, space.

V.E 4 (4) PT-3-54, -61, -207 (a) Provide pressure input to the FWLCS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CPL 17 L003 Revision 19 Page 40 of 66 1NSTRIJOTOR NOTES (b) Provides reactor pressure indication on recorder PR-353 (Panel 9-5) over a range of 0-1500 psig (average pressure).

Reactor high pressure alarm is actuated at 1058 psig.

PT-3-54. -61, -207 provide reactor pressure indication on Panel 9-5.

(c) Pressure input is from the Obj. yB. 4, VD.8.

Steam space V.E.4 (5) PT-3-74A!S. (reference columns)

PT-68-95!96 (SLC diffuser piping)

(a) Provides reactor pressure permissive signal (<450 psig) for opening Core Spray and LPCI admission valves.

(b) In conjunction with high drywell pressure provides Core Spray and LPCI automatic initiation signal.

(c) Provides recirc discha i-ge valve auto closure at 230 psig.

(d) Provides reactor pressure indication on Panel 93.

(e) Provides reactor pressure indicator on the MU cabinets (9-81. 9-82).

(f) PT-3-74 pressure input is from Obj. V.B. 14: V.D.8, Steam space. PT-68-95!96 pressure input is from liquid 54 space below core plate (6) PT-3-59 (a) Provides a narrow range (850-1100 psig) reactor oressure indication on Panel 9-5 recorder.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT L

3-EO[-2 PAGE 1 OF I PRIMARY CONTAINMENT CONTROL UNIT 3 BROWNS FERRY NUCLEAR PLANT REV: 8

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSB1L1TY SUPPORT SCR4 N X P3WZ A ED NRO:.

SCf-7. SCi-12 WHILE EXECUTING THIS PROCEDURE:

S.IG REGJEtETSD xr ci wz:s

&M H AS,iE CYJTRL IERroRM L

VRFY XSCRM 3-EOi-1 PAGE 1 QfZ 1 RPV CONTROL UNIT 3 BROVINS FERRY NUCLEAR PLANT REV: S

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSBILITY SUPPORT OPL1 71.003 Revision 19 Page 40 of 66 NSTRLJCTOR NOTES (b) Provides reactor pressure indication on recori:ler PR-3-53 (Panel 9L5) over a range of 0-1500 psig (average pressure).

Reactor hgh pressure alarm is actuated at 1058 psig.

PT-3-54. -61, -207 provide reactor pressure indication on Panel 9-5.

(c) Pressure input is from the Qhj. V.B. 4. V.D.8.

Steam Space V. E.4 (5) PT374A/B. (reference columns)

PT-68-95!96 (SLC diffuser piping)

(a) Provides reactor pressure permissive signal (< 450 5ii for opening Core Spray and LPCI admission valves.

(b) In conjunction with high drywell pressure provides Core Spray and LPCI automatic initiation signaL 1c) Provides recirc discha i-ge valve auto closure at 230 psig.

14) Provides reactor pressure indication on Panel 93.

(e) Provides reactor pi-essure indicator on the ATIJ cabinets (9-81. 9-82).

(f) PT-3-74 pressure input is from Ohj. V B. 14. V.D.8, Steam space. PT-68-95!96 V. F .4 pressure input is from liquid space below core plate.

(6) PT-3-59 (a) Provides a narrow range (850-1 100 psig) reactor pressure indication on Panel 9-5 recorder.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295031 Reactor Low Water Level Tier # 1 K3.O1 (CFR41.5)

Knowledge of the reasons for the following responses as they apply Group # 1 to REACTOR LOW WATER LEVEL: K/A # 295031 K3.01

• Automatic depressurization system actuation Importance Rating 3.9 Proposed Question: # 16 Given the following Unit 1 plant conditions:

• HPCI 12OVAC POWER FAILURE, (1-9-3F, Window 7) is in alarm

• Reactor Water Level is (-) 122 inches and lowering

• Drywell Pressure is 1 .8 psig and steady

• Assume NO operator action Which ONE of the following describes the time that must elapse before ADS automatically initiates AND the reason for this response?

ADS will initiate in (1 )_. This actuation is in response to a LOCA (2).

A. (1) 265 seconds (2) inside the Drywell B. (1)360 seconds (2) inside the Drywell C. (1) 265 seconds (2) outside the Drywell D. (1)360 seconds (2) outside the Drywell Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect - This time delay is associated with -122 (Optional): inches received without a high DW pressure (>2.45 psig), which is given in the stem. However, once this timer times out, if ECCS pumps are running, a 95 second timer initiates and must time out before ADS initiates. This makes the total time 360 seconds. Part 2 incorrect This is the basis for ADS initiation with BOTH high DW pressure AND low RPV level.

B INCORRECT: Part correct as stated in D. Part 2 incorrect as stated in A above.

C INCORRECT: Part 1 incorrect as stated in A above. Part 2 correct. ADS initiation in the absence of high DW pressure is due to decay heat boil-off following a LOCA outside the Drywell with MSIV isolation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: Part I correct Time delay associated with -122 inches received without a high DW pressure >2.45 psig (265 sec), plus the 95 second timer makes the total time 360 seconds. Part 2 correct. ADS initiation in the absence of high DW pressure is due to decay heat boil-off following a LOCA outside the Drywell with MSIV isolation.

KA Justification:

The KA is met because the question tests knowledge of the reason for Automatic Depressurization system actuation as it applies to Low Reactor Water Level.

Question Cognitive Leve:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): OPL171.043 Rev. 13 (Attach if not previously provided) 1-01-1 Rev. 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.043 V.B.4 (As available)

Question Source: Bank # BFN 0707 #54 Modified Bank # (Note changes or attach parent>

New Question History: Last NRC Exam Browns Ferry 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The addition of Assume NO operator action was added due to procedural guidance which would inhibit ADS initiation under this condition. In this condition, 1 -EOl-1 flowchart path RC/L would allow ADS to be inhibited below -100 inches. In addition, 1-EOl-C1 would be entered below approximately -120 inches and direct that ADS be inhibited. In fact, there are no foreseeable circumstances where ADS would be allowed to auto initiate by procedure.

The HPCI 12OVAC Power Failure annunciator is to provide realistic conditions where ADS would auto initiate. If HPCI were operable, ADS would not be required under these conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CPL1T 1.042 Revision 13 Page 12 of 30 INSTRUCTOR NOTES

d. EOl Appendix BC crossties CAD to DWCA PROCEDURE USE

& ADHERENCE 4 ADS systems controls TP-2

a. Consists of pressure and water level sensors arrançed in the trip systems that control a solenoid-operated pilot air valve

5. The solenoid-operated valve controls the pneumatic DCN 5110$

pressure applied to a diaphragm actuator which Cable & Switch controls the SRV directly configuration I modifications

c. Cables from sensors lead to the Control Room where logic arrangements are formed in cabinets ci. Control channels are separated to limit the effects of electrical failures

e. A twoposition control switch is provided in the Control Room for control of the ADS valves

1) Two positions are OPEN and AUTO HP Use SELF-CHECKING

2) ln OPEN, the switch energizes a DC solenoid which allows pneumatic pressure to be applied Pressure relief to the diaphragm actuator of the relief valve consists of actuation of NOTE: reactor Pressure The relief valves can he manually opened to provide a controlled on mternai pilot or nuclear system cooldown under conditions where the normal heat sink by electro is not available pneumatic operation via

3) In AUTO. the valves are controlled by the ADS pressure switches.

logic and pressure relief logic

f. Four of the six ADS valves may also he controlled from U N IT a backup control boani which is provided to facilitate DIFFERENCE.

plant shutdown and cooldown from outside the Control DCN 5110$ adds Room new panel 25-

$58 to Unit 1 Automatic Depressurization Initiation Logic

a. The following conditions must he met before automatic Obj. V.B.4 depressurization will occur Obj. \LC.2

1) TWO coincident signals of high drywell pressure Obj. V.D.3

(+/-2.45 psig) and low low low reactor vessel Obi. V.E.4

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.043 Revision 13 Page 13 of 30 INSTRUCTOR NOTES water level (-122)

OR

-122 for 265 sec. LT-3-SSA-0

2) A confirmatory low reactor vessel water level LT-3-l 84 signal (÷2) (Tech Spec Value 0) LT-3-185

3) Any one of the four RHR pumps or either A or B Obi V.C.4 and either C or C Core Spray pumps runnincj Obj. \.D.4 NOTE:

This signal comes from pressure switches on the discharge of the pumps which give permissives in the logic above a set pressure of 100 psig for RHR pumps and 185 psig for the Core Spray pumps.

RHR CS Associated PS-74-$A and SB P5-75-7 shutdown boards (Pump A) (Pump A) must be energized PS-74-3IA and 316 P5-75-35 for the respective (Pump B) (Pump B) pumps.

P5-74-WA and 196 P5-75-18 (Pump C) (Pump C)

PS-74-42A and 426 PS-75-44 (Pump 0) (Pump 0)

4) A 95-second timer must he timed out

h. The high drwell pressure signal seals in immediately Oh). V.C.4 UPOfl receipt of The signal Ob V.0.4

1) Must he manually reset after the signal has PS-64-57A-D cleared 2 Indicative of a breach in the process system HP ProcedLire use barrier inside the drywell and Adherence

c. The reactor vessel low water level signals (-122 and Obj. V.6.4

+2) indicate that fuel is in danqer of becoming Ohi. V C 3 overheated Obj. V.0.3 Obi. V.E.4

1) The 122 water level signal would not normally K 28, 29. & 30 occur unless the HPCI System had failed Ohi. \!.C.4

2) These signals do not seal Cl:I. V.[).4

3) The -122 water level initiation setpoint is selected to open the SRVs and depressunze the reactor vessel in time to allow fLiel cooling by the Core Spray and LPCI Systems following TP-3 a LOCA, in the event that the other makeup Oh]. V.C.4 systems (Feedwater, CRC Hydraulic, RCIC, Obj. V.0.4

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 171 94:3 Revision 13 Page 14 of 30 INSTRUCTOR NOTES and HPOI) fail to maintain vessel level Timer logic in

4) The -122 setpoint will also initiate 265 second ECCS ATU timers that seal in and will ILifl even if water drawing series level is restored to >-l 22 The timers can he 45E670.

reset (if Rx Level >-1 2Z) using pushbuttons in the auxiliary instrument room.

fr5) Once these timers have timed out, the drywell pressure contacts are bypassed. but other relays (that are not sealed in) must still sense reactor level <-122

6) If so. and the other conditions are met (+7 anti low pressure pumps running), the 95 second timers will start.

7) This feature is based on a LOCA outside of the divwell which has been isolated. Level is below 122 and inventory is boiling off due to decay heat.

6) General Electric calculations have deterniined that the core wUl remain covered for 15 minutes after the 122 level is reached. Our system will initiated within the 15 minutes calculated by GE

9) The ÷2 water level signal is a confirmatory low Oh). VB.4 level signal Obj. V.0.3 Obj V.0.4

d. The 95-second timer allows the primary high pressure Obi. V.0.3 ECCS srstem (HPCI) to function and relieve Obj. V.E.4 conditions that would require ADS

1) If during the 95-second timer run-out the water C)bi. V C 4 level signals clear, the tinier i-esets ObJ. V.0.4 automatically

2) The operator can use timer reset pushbuttons Obj. V.0.4 on Panel 9-3 to delay automatic opening of the Oh). V.D.4 SRVs

3) The operator can use the keylock inhibit Keylock XS-1 -159 switches (Panel 9-3) to prevent the initiation of A Logic the 95-second timers and thereby prevent an XS-1-161 B ADS actuation. Logic

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Main Steam System 1.01-1 Uniti Rev.0011 Page 12 of 63 3.4 Main Steam Relief Valve (MSRV / ADS)

A. Whenever both the acoustic monitor and the temperature indication on a relief valve fail to indicate in the Control Room, the Technical Specifications Section 3.3.3.1 should be consulted to determine what limiting conditions for operation apply.

B. In the event that a relief valve fails to function as designed and the cause of the malfunction is not clearly determined and then corrected, the valve should be considered inoperable and Technical Specifications. Sections 3.5.1 and 3.4.3.

should he consulted to determine what limiting conditions for operation apply.

C. ADS will initiate when ALL of the following conditions are met:

1. A confirmatory low reactor water level signal (+2.0 inches), REACTOR LEVEL LOW ADS BLOWDOWN PERMISSIVE, 1-XA-55-9-3C. Window 3.

2. Tvo coincident signals for each of the follov.fng parameters:

a. high drywell pressure (+2.45 psig) in conjunction with low-low-low reactor water level (-122 inches), ADS BLOWDOWN HIGH DRYWELL PRESS SEAL-IN, 1-XA-55-9-3C, Window 33 and RX WTR LVL LOW LOW LOW ECCS/ESF NIT 1-LA-3-58A, 1-XA-55-9-3C, Window 28 OR

b. low-low-low reactor water level (-122 inches). RX WTR LVL LOW LOW LOW ECCS!ESF INIT 1-LA-3-SSA, 1-XA-55-9-3C Window 28, for 265 seconds (High drywell pressure bypass). and

3. One RHR pump OR two Core Spray pumps (A or B and C or D) running, RHR OR CS PUMPS RUNNING ADS BLOWDOWN PERMISSIVE, 1-XA-55-9-3C, Window 10.

4. When ALL of the above logic is satisfied, then a 95 second timer starts and ADS BLOWDOWN TIMERS INITIATED, i-9-3C. Window 11, alarms, and the timer must be timed out to initiate ADS blowdown.

D. Depressing 1-XS-1-159 and -161 on Panel 1-9-3 resets the ADS Blowclown Timers. They also reset the ADS initiation, if the timers have timed out. ADS will re-initiate upon subsequent timing out of the timer provided the low level and pump logic signals still exist. The timer setpoint is 95 seconds, however setpoint tolerance allows it to be as low 77 seconds.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 0707 Examination Outline Cross-reference: Level RO SRO 295031EK3.O1 Tier# 1 Knowledge of the reasons for the following responses as they apply Grou p # 1 to Reactor Low Water Level: Automatic Depressurization System actuation. K/A # 295031 EK3.01 Importance Rating 3 4.2 oposed Question: RO # 54 Given the following Unit 1 plant conditions:

• HPCI 12OVAC POWER FAILURE (9-3F W7) is in alarm.

• A LOCA has occurred initiating a scram on Low Reactor Water Level.

• Reactor water level (-) 122 inches and lowering

• Drywell pressure 1.8 psig and steady

• A Pre-Accident Signal (PAS) has just been received and all ECCS equipment respond as designed.

• Assume NO operator actions.

Which ONE of the following describes the time that must elapse before ADS automatically initiates and the reason for this response?

ADS will initiate in (1) . This actuation is in response to a (2)

(1) (2)

A. 265 seconds LOCA inside the Drywell B. 360 seconds LOCA inside the Drywell C. 265 seconds LOCA outside the Drywell D. 360 seconds LOCA outside the Drywell

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Crossreference: Level RO 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown

/1 ier 1 EA2.06 (10CFR55.41.1O) Group# 1 Ability to determine and/or interpret the following as they apply to K/A # 295037EA2.06 SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

• Reactor pressure Importance Rating 4.0 Proposed Question: # 17 An ATWS has occurred on Unit I with the following time line AND conditions:

• At 1200 Reactor Power is 15%

• At 1210 SLC is initiated

• At 1235 SLC Storage Tank Level is 67%

• At 1300 SLC Storage Tank Level is 43%

Which ONE of the following completes the statements below?

In accordance with 1-EOl-1, RPV Control, (1) is the earliest time the crew must commence depressurizing the Reactor below the Shutdown Cooling Reactor Pressure interlock.

Cooldown rate of 1000 F per hour _(2) be exceeded.

A. (1)1235 (2) can B. (1)1235 (2) CANNOT C. (1) 1300 (2) can D. (1) 1300 (2) CANNOT Proposed Answer: D Explanation A INCORRECT: Part I incorrect Level must be 43% to commence (Optional): cooldown. Plausible in that 67% tank level is Hot Shutdown weight for SLC.

Part 2 incorrect Plausible in that under certain conditions in EOl-l, cooldown is performed irrespective of cooldown rates.

B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

c INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: Part 1 correct In accordance with 1-EOl-1, when SLC has been injected into the RPV to a tank level of 43%, depressurize the RPV below the shutdown cooling pressure interlock. Part 2 correct Must maintain cooldown rate < 1000 F per hour.

KA Justification:

The KA is met because the question tests the candidates ability to determine when Reactor Pressure is lowered in accordance with the EOls with an ATWS condition present.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemb le, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome Technical Reference(s): 1-EOl-1, Rev. 0 (Attach if not previously provided)

OPL171.202 Rev. 8 Proposed references to be provided to applicants during examination:

NONE Learning Objective: OPL171.039, V.B.6 (As available)

OPL171.202, V.B.9 Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from I -EO(-I, RPV Control, RC/P leg I I WHE N THE REACTOR WILL REMAIN SUBCRITIGAL

\_ _) WITHOUT BORON UNDER ALL CONDITIONS

SEE NOTEI OR SLC FL.S INLECTED INTO THE RPVIOA TANK LVL OF 43%

OR THE RX IS SUBCRITICAL AND ND BORON HAS BEEN IIsLECTED INTO THE RPV, THEN cONTINUE A L RCP-Th C AU T[ ON 3 ELE\AED SUPPR CHMBR PRESS MAY RIPRCIC

1. 6 HFCI OR RCIC SUCTION TEMP ABOVE iO.F L

EPRESSUkIZE THE RPi BELOW THE SHUTDOWN COOLNO RPM PRESS INTERsOCAWITHONE OR MORE OF THE FOLLOWING DEPRESSURIZATION SYSTEMS. MAINTAIN COCLDOWN RATE BELOW IOl°F,HR.

DEPRESSURIZATIO SYSTEM .PPx tN TURBBYPASSVLY1E1HE MAIN CONDENSER IS AVLELE4 USE APPX TO OPEN MSIV ES MSR ONX WHEN SUfPR LVL IS ASOES.5 FT IS .t. IN SFEAM BELIEF 1 Ci AC cu M PRESS iow ANNUNCIATOR E-SO-18) IS IN ALARM Iff5... MIMMZE MSRCCYCIJNCBVUSING SUSTANED OPENING FOR DEPRESSURUTiTION HPCIWITHCSTSUOTIONIF POSSIBLE RCIC, WITH CST SLJ Cl ION IF POSSIBLE B RFPT ON MN FLOW tAIN STE ISSTEM DRPJS

.STEAMSEALS SJAE INC OFF GAS RNCU N NOFORON HAS SEEN NJEGTED RCP-II L

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBLITY SUPPORT OPL171 .202 Revision 8 Page 30 of 6

a. The subsequent steps in this procedure depressurize and cool down the RPV to cold shutdown conditions. If no boron has been injected into the RPV, depressurization and cooldown may proceed as long as control rod insertion is sufficient to shut down the reactor. Such action is permitted even though the existing margin to criticality is small. The positive reactivity added during cooldown may return the reactor to criticality. Should this condition occur, the operator is directed to return to Step RC/P-10, to terminate the cooldown and stabilize RPV pressure, until the reactor can once again be made subcritical.

2. Step RC/P-13

a. This contingent actions step requires the operator to wait until one or more of the stated conditions have been met before continuing in this procedure.

b. After RPV pressure is stabilized, it is appropriate to ensure that the reactor is subcritical prior to performing a normal RPV depressurization and cooldown. Otherwise, the positive reactivity added during forced cooldown, below the saturation j temperature for low RPV pressure, may cause the I reactor to return to power. Any one of three conditions will ensure that the reactor is subcritical.

c. The first condition requires that the reactor will See Note 1 Bases in remain subcritical without boron under all OPL1 71.201 conditions.

d. The second condition requires that the SLC System Obj.V.B.4.c I has injected into the RPV to at least all but 43% of I the SLC tank level. This SLC tank level I corresponds to the Cold Shutdown Boron Weight of boron. The Cold Shutdown Boron Weight is 1 defined to be the least weight of soluble boron I which, if injected into the RPV and mixed I uniformly, will maintain the reactor shutdown L under all conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .202 Revision 8 Page 31 of6

e. The third condition allows RPV depressurization and cooldown to proceed as long as control rod insertion is sufficient to maintain the reactor subcritical under present conditions. As used in the EOIs, the term subcritical t means that reactor power is below the heating range and not increasing. This condition is applicable only if no boron has been injected into the RPV. Such action is permitted even though the existing margin to criticality may be small. A return to criticality under these conditions is acceptable because termination of the cooldown will stop the reactor power increase. Direction to terminate the cooldown is provided in Step RC/P-12.

3. Step RC/P-14

a. This action step directs the operator to use any of the depressurization sources listed in Steps RC/P 10 and RC/P-1 1 to depressurize the RPV.

b. Once it has been determined that the reactor is I subcritical, the operator is directed to depressurize the RPV ensuring that the Technical Specification cooldown rate of 100 °F/hour is observed to maintain RPV metal ductility limits. The cooldown rate is also controlled to avoid an inadvertent, rapid return to criticality, if the margin to subcriticality is small.

c. If MSRVs are being used to depressurize the RPV Obj.V.B.7 and the continuous pneumatic supply to the MSRV Obj .V.C.3 actuators is isolated or unavailable. Even though MSRV accumulators contain a reserve pneumatic supply, leakage through in-line valves, fittings, and actuators may deplete the reserve capacity. Thus, subsequent to loss of the continuous MSRV pneumatic supply, there is no assurance as to the number of MSRV operating cycles remaining.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295038 High Off-Site Release Rate EK2.1O (100FR 55.41 .7) Tier # 1 Knowledge of the interrelations between HIGH OFF-SITE RELEASE Group # 1 RATE and the following:

K/A # 295038EK2.1 0

• Condenser air removal system Importance Rating 3.2 Proposed Question: # 18 Unit 2 is in Start Up. Off Gas Treatment Select Switch, 2-XS-66-1 13, is in BYPASS. The following alarm/indication are received:

• OG POST-TREATMENT RADIATION HIGH, (2-9-4C, Window 33)

• Offgas Post-Treatment Radiation is 4 6.5x c ps lO Which ONE of the following identifies the impact of this condition on the Offgas System?

A. NO valves will reposition B. Adsorber Bypass Valve, 2-FCV-66-113B will close. NO other valves will reposition.

C. Adsorber Bypass Valve, 2-FCV-66-1 13B will close AND Adsorber Inlet Valve, 2-FCV 11 3A will open. NO other valves will reposition.

D. Adsorber Bypass Valve, 2-FCV-66-1 13B will close. Adsorber Inlet Valve, 2-FCV 113A AND Charcoal Adsorber Train 2 Inlet Valve, 2-FCV-66-118 will open.

Proposed Answer: A Explanation A CORRECT: With Off Gas Treatment Select Switch, 2-XS-66-1 13, not in (Optional): AUTO, the Radiation High will not result in automatic alignment of Offgas Charcoal Adsorbers.

B INCORRECT: Plausibility based on misconception that only Adsorber Bypass Valve, 2-FCV-66-1 13B will close on High Radiation and that the function remains in force with the Off Gas Treatment Select Switch, 2-XS-66-113, is in BYPASS.

C INCORRECT: If Off Gas Treatment Select Switch, 2-XS-66-1 13, was in AUTO, this would be the correct answer. Adsorber Bypass Valve (FCV 11 3B) will close, and Adsorber Inlet Valve (FCV-66-1 1 3A) will open when one channel reaches OG POST-TREATMENT RADIATION HIGH.

Plausible in that the 3 X High Radiation Offgas isolation will occur with the Off Gas Treatment Select Switch, 2-XS-66-113 in any position.

D INCORRECT: Plausibility based on misconception that Charcoal Adsorber Train 2 Inlet Valve, 2-FCV-66-118 will open on High Radiation and that the function remains in force with the Off Gas Treatment Select Switch, 2-XS-66-1 13, is in BYPASS. Plausible in that when aligning charcoal filters for parallel operation, 2-01-66 directs opening of this valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the interrelations between High Off-Site Release Rate as indicated by Offgas Post Treat Radiation High and the Condenser air removal system including the response of Adsorber Bypass Valve, FCV-66-1 I 3B, AND the Adsorber Inlet Valve, FCV-66-113A. Since there is no procedural guidance for operation with the Off Gas Treatment Select Switch, 2-XS-66-113, in AUTO in any conditions, the question is posed with the Select Switch in BYPASS for operational validity.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL171 .033 Rev. 13 (Attach if not previously provided)

OPL171.030 Rev. 18 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.033 V.B.4 (As available)

Question Source: Bank #

Modifid Bank# (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .033 Rev sian i3 Page 9 of 75 INSTRLICTOR NOTES (4) Uses same sample as pretreatment monitor (5) Used as an expanded scale device for Ohj. V.B.i ,3.b locating ruptured or failed fuel elements Ohj. V.0.1 3.b (6) One pen recorder (RM-90-i 60)

(7) No alami functions

c. Off-Gas post-treatment radiation monitoring TPs 8. 9 system (RM-90-265A/266A)

(1) Two Gamma sensitive scintillation Ohj. V.D.3.d detectors, powered from ÷ 24 VDC. Ot4. \,B.2 Neutron Monitoring Batteries. Ohj. V.0.2 (2) Post-treat Radiation monitors (RM-90-265A/266A) Feed a two-pen recorder on Control Room panel 92 (RR-90-265).

(3) Samples are drawn from Off-Gas flow just downstream of the charcoal beds and returns sample to inlet of charcoal beds (4) Alarm/trip signals conic from the Ohj. V.B.5 radiation monitors and the following Obj. V.0.5 alaniis and protective functions:

(a) OG POST-TREATMENT RADIATION HIGH (55-40-33) alarms at 6.2Xlocps.

(i) It the Off Gas Treatment Select Switch (XS-66-1 13) is in AUTO and the Hgh Radahon alarm is received k

(U) Adsorber Bypass Valve r (FCV-66-113B) will close, and Adsorber In let Valve (FC\/66t 1 3A) wUl open

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet DISTRACTOR PLAUSIBLITY SUPPORT OPL171 .033 Resion 3 Page 21 of 75 INSTRUCTOR NOTES (5) Off-Gas isolation is a twoout-oftwo Obj. V.B.4.b logic Ohj. V.0.4.0 (a) Downscale, Hi-Hi-Hi or INOP on RM-90-265A AND Downscale, Hi-Hi-Hi or INOP on RM-90-266A will automatically isolate the Off-Gas system alter a 5 second time delay.

(PCV-86-28 closes)

3. Stack-Gas Radiation Monitoring System (RM-9O- 147 Ohj. V.0.7

& 148) Obj. V.B.3.b Ohj. V.C.3.b

a. Purpose U) Used to indicate and record release rates from the stack during normal operation and to alam whenever limits at-c reached (2) To monitor the stack gas effluent, a Note: isokinetic sample is drawn through an isokinetic probe explained in probe which is located two-thirds of the- section 9 of this way up the stack lesson

b. The stack receives exhaust gases from following:

(1) Steam Jet Air Ejector (SJAE)

(2) Steam Packing Exhauster (SPE)

(3) Mechanical vacuum pump (4) Standby Gas Treatment (SGT)

(5) Stack Gas Analyzer Room Vent

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 17L033 Reviwon 13 PsIte 73 of 75 Appendix 2 - Monitor Summary TeeS Monitor Von:or ID No Spec Power Tpe Indications nnnctIcn Main Wean 1315 None RPS A Ion CR NUMAC diQEal display Ala-ms - DNSCL 12)

Uses 537 RPS A - -lION (ISa NFLE)

HSNI1INO :3 a NFLE)

Vc Pms Viva ac Pnps, Reccider IRR-93-1 SW I selector s45che Off Gas CR Icier Alnma - 00 ALG ANNUAL RELEASE Pre-Veatn-ent 157 None SC A lOS Recorder RR-90-157 LIMIT EXCEEDED

-03 9ETREATMENT FIGF-

- 05 PRETREATMENT DNSC

- 00 SAM°.E njyy ABNML Fink Thu 530 None NMS Ion CR Meter md cation only Recorder: One pen RR-90-15Q 315cc Gas 47 000Iot NMS Scwtiliation cdllcstot(U1 I Alarms - HIGI 148 11.2 RecorderiLli) - HIGH,HiGH

- DNSCLSNOP

- FLOW ABNORMAL Alamis Only-No Trips 05-Gas 255) 00CM NMS Scintillation lsdmcators(2) Ala-ms - HIGH Alrans cnarcoal Lees tin ALTO)

Fost Treatment 2315 11.2 Recorder RR-00-265 - HIGH-HIGH -Slain Only

- HIGH-HIGH-HIGHlNOPi2of2 isocttes Off-Gas: any combination)

- OOWNSCALE Turb/Rx!Ref,el 250 00CM RCA Scntitiation C.T.IU1I Alarms - HIGH Veed Slits 1.1.2 (For alt unts) -OOWNSCALE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.030 Revision 18 Page 44 of 74 INSTRUCTOR NOTES

d. Any Channel Hi F-IS66-113 is kept in TREAT to keep the Initiates charcoal adsorbers by opening adsorbers in service adsorber inlet valves (11 3A) (117) and when the unit is at closing adsorber bypass valve (113B), power. Major provided HS66-113 is in AUTO. system flow changes would cause a

5. Condenser Vacuum Low (. 25 Hg Vacuum) radiation spike

a. Initiates auto start of selected SJAE Auto swap inhibited by procedure on U2

b. The following valves should respond:

(1) The steam admission valve (motor- Auto swap capability and air-operated) for the standby removed on U3.

SJAE should open. DON 51323 (2) The condensate inlet and outlet valves (motor-operated) for the standby SJAE should open.

(3) The outlet valve for the standby SJAE should open when steam press> 173 psig if control switch is in OPEN.

(4) The steam admission valves for the running SJAE should shut (MOV5, PCV. and outlet valve).

6. Condenser Vacuum High ( 26 Hg Vacuum) PS between suction valve & pump

a. Prevents operation of condenser vacuum pump when it is improper to do so

b. Trips the condenser vacuum pump

7. Condenser Vacuum High ( 22 Hg Vacuum) with Reactor Pressure High ( 600 psig)

Prevents operation of condenser vacuum pump and isolates vacuum pump suction valve

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBLITY SUPPORT BFN Off-Gas System 2.01-66 Unit2 Rev.0099 Page 50 of 135 5.11 Aligning Charcoal Filters for Parallel Flow NOTE The charcoal beds can be aligned for either parallel or series flow, but normally parallel flow is preferred. Performing the following steps at Panel 9-53 aligns the charcoal beds for parallel flow. If series alignment is preferred, Section 8.10 is required to be performed in lieu of the following steps.

CAUTION The charcoal adsorhers are required to be aligned in the treatment mode prior to reaching 25% power.

[1] PLACE OFFGAS TREATMENT SELECT handswitch.

2-XS-66-113, in TREAT.

[21 OPEN CHARCOAL ADSORBER TRAIN 2 INLET VALVE.

[3]

using 2-HS-66-1 17.

OPEN CHARCOAL ADSORBER TRAIN I DISCH VALVE.

using 2-HS-66-1 18.

[4] CLOSE CHARCOAL ADSORBER TRAINS SERiES VLV.

using 2-HS-66-1 16.

[5] CHECK dewpoint temperature on OFFGAS REHEATER TEMPERATURE recorder, 2-TRS-66-108, indicates 45F or less (Blue Pen>.

[6] IF the Off-Gas System is intended to be operated with charcoal beds in parallel with the charcoal beds on another (shutdown) unit. THEN (Otherwise N/A.>

COMPLETE Section 8.11.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 600000 Plant Fire On Site / 8 Tier # 1 AA2.13 (10CFR 55.41.10)

Ability to determine and interpret the following as they apply to Group # 1 PLANT FIRE ON SITE: K/A # 600000AA2.13

• Need for emergency plant shutdown Importance Rating 3.2 Proposed Question: # 19 With ALL 3 Units operating at 100% Reactor Power, a fire at 4 kV Shutdown Board A has resulted in the following:

• Failure of Unit I RHR Pump IA AND Core Spray Pump IA

• Shift Manager has declared an Appendix R Fire In accordance with Safe Shutdown Instructions, which ONE of the following identifies which, if any, Reactor(s) is (are) required to be scrammed?

A. NO Reactor Scram is required B. Unit I ONLY C. Unit I AND Unit 2 ONLY D. ALL 3 Units Proposed Answer: D Explanation A INCORRECT: Plausible in that no conditions have been identified which (Optional): would require a Reactor Scram in accordance with AOIs (including 0-AOl-26-I, Response to Fires), EOls or Tech Specs. If candidate considers only these Abnormal / Emergency Procedures, this would be the correct answer.

B INCORRECT: Plausible in that ONLY Unit 1 has equipment that has been damaged by the fire.

C INCORRECT: Plausible in that 4 kV Shutdown Board A supplies loads on Unit 1 and Unit 2.

D CORRECT: Per Safe Shutdown Instructions, if SSls are entered for an Appendix R Fire, ALL 3 Units must be scrammed.

ES-401 Sample Wriften Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because it tests the candidates ability to determine need to emergency shutdown Units based plant fire on site.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL171.031 Rev 13 (Attach if not previously provided)

O-SSI-5 Rev. 7 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Unit 1, 4KV Electric Board Room IA 0-SSI4 UnitO Rev. 0007 Page 6 of 117 INITIALS 2.0 UNiT 2 CONTROL ROOM OPERATOR ACTIONS TBD-2

[1] DIRECT Unit 3 Unit Supervisor to perform Section 3.0 of F 0-SSI-5 to Scram Unit 3, AND PROCEED TO cold shutdown.

[2] DIRECT Unit 1 Unit Supervisor to perform Section 4.0 of 0-SSI-5 to Scram Unit 1, AND PROCEED TO cold shutdown.

NOTE The following instruments are those which have been credited for safe shutdown, and must be referenced when executing manual actions for this fire area:

2-Ll-3-58A and 2-Pl-3-74A for reactor level and pressure 2-Tl-64-52AB and 2-Pl-64-67B for drywell temperature and pressure TBD-81 2-Ll-64-1 59A and 2-Tl-64-161 for the suppression pool level and temperature 2-Ll-21 61A for Condensate Storage Tank 2 (0 Mm)

[3] DIRECT Unit 2 Operator to perform the following:

TBD-3 TBD-1

[3.1] VERIFY reactor Scram AND RECORD current time(SSI time of entry).

Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.031 Revision 13 Page 7 of 50 INSTRUCTOR NOTES B, Brief Overview of the Procedure (Generic)

The Shift Manager (SM) determines when the entry conditions are met and uses 0-SS1-001 to determine which subsection (fire area) to perform.

I.

SM initiates the procedure and confirms with the Unit 1, 2, T=0 at verification and 3 Operators that the unit is scrammed, the MSIVs are of each Reactor closed/checked closed, and a fire pump is verified running. trip (TBD-1)

Use of equipment fed from a shutdown board whose DIG is considered unreliable is allowed while executing the SSI; Conservative however, prompt action may be required to secure Decision equipment which is determined to be operating spuriously, Making or the board may be lost to a fault at any time.

Appropriate reliable diesels are started from the Control Rooms as required for the particular power system alignment for the subsection (Electrical Alignment Illustration).

HPCI and/or RCIC will be used to maintain reactor water Some areas also level and the MSRVs used to control reactor pressure. have I-IPCI1RCIC unavailable Unit Operator begins a rapid depressurization of the reactor using MSRVs for the affected unit. Aligns the V.B.4 electrical distribution per the sub instruction. The final TP-1 plant system lineup has RHR flooding the vessel with the flow path recirculating water into vessel out the MSRVs to the torus to the RHR pumps and RHR heat exchangers with RHRSW as the cooling medium.

Entry into the SSI, when the entry conditions are met, The SSI provide a CANNOT be delayed. The time-llnes associated with methodology to implementation are measured from the time the SSI is protect the health entered. As such, delay into entry could cause the and safety of the analysis to be invalidated. public during a fire

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO SRO 700000 Generator Voltage and Electric Grid Disturbances / 6 Tier # 1 AK207 (10CFR 55 41 5)

Knowledge of the interrelations between GENERATOR VOLTAGE Group # 1 AND ELECTRIC GRID DISTURBANCES and the following K/A # 700000AK2.07

• Turbine/generator control Importance Rating 3.6 Proposed Question: # 20 Unit 3 is operating at 80% Reactor Power AND the crew has entered 0-AOl-57-1 E, Grid Instability, due to the 530 kV system voltage being at 513 kV. The crew reaches the following step in the procedure:

• RAISE reactive power until voltage returns to 520 kV.

Which ONE of the following identifies how to raise reactive power AND the 161 kV Capacitor Bank Status that will restore the system voltage in accordance with 0-AOl-57-1 E?

A. Depress the EHC load set RAISE pushbutton, 3-HS-47-75C; Check the 161 kV Capacitor Banks are IN service.

B. Depress the EHC load set RAISE pushbutton, 3-HS-47-75C; Check the 161 kV Capacitor Banks are OUT of service C Place the Generator Field Voltage Auto Adjust (90P), 3-HS-57-26, to the RAISE position, check the 161 kV Capacitor Banks are IN service.

D. Place the Generator Field Voltage Auto Adjust (90P), 3-HS-57-26, to the RAISE position; check the 161 kV Capacitor Banks are OUT of service.

Proposed Answer: C Explanation A INCORRECT: Part 1 incorrect Depress the EHC load set RAISE (Optional): pushbutton will have no affect on load or voltage at current power levels.

Plausible in that raising load would aid in mitigating the grid low voltage condition. Part 2 is correct as required by 0-AOI-57-1 E B INCORRECT: Part 1 and 2 incorrect 161 kV Capacitor Banks out of service will not aid in restoring system voltage. Plausible in that it is an action directed under certain conditions for Grid Instability in 0-AOI-57-1 E C CORRECT: Part 1 correct Per O-AOl-57-1 E, RAISE reactive power to system voltage returns to 520KV OR UNTIL Generator Reactive Power reaches +200 MVAR, Per 3-01-47, To adjust GENERATOR MVAR, 3-EI 51, in the positive or lagging direction, PLACE GENERATOR FIELD VOLTAGE AUTO ADJUST (90P), 3-HS-57-26, in RAISE UNTIL desired MVAR is indicated. Part 2 correct Per 0-AOI-57-1 E, CHECK 161KV Cap Banks are In Service D INCORRECT: Part 1 is correct and Part 2 is incorrect.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the interrelations between low system voltage due to Grid Disturbance and Generator Field Voltage Auto Adjust (90P), 3-HS-57-26.

Question Cognitive Level:

This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 0-AOl-57-1 E Rev 7 (Attach if not previously provided) 3-01-47 Rev 91 (Including version I revision number)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL17t036 V.B.13 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Turbine-Generator System 3.01-47

• Unit3 Rev.0091 Page 87 of 241 6.1 Normal Operation (continued) tiOl MAINTAIN GENERATOR MVAR, 3-EI-57-51, 200MVAR outgoing, and those of Illustration 6, Generator KVAR Limitations, (Capability Curve), the above note arid as directed by the Transmission Operator as follows:

[10.1] To adjust GENERATOR MVAR, 3-El-57-51. in the positive or lagging direction, PLACE GENERATOR FIELD VOLTAGE AUTO ADJUST (SOP), 3-HS-57-26, in RAISE UNTIL desired MVAR is indicated. D

[10.2] To adjust GENERATOR MVAR, 3-El-57-51, in the negative or leading direction, PLACE GENERATOR FIELD VOLTAGE AUTO ADJUST (SOP), 3-HS-57-26, in LOWER UNTIL desired MVAR is indicated. D

[10.3] PERFORM the following to minimize generator heat load or check GENERATOR MVAR, 3-El-57-51, accuracy:

[10.3.1] ADJUST GENERATOR MVAR, 3-El-57-51, per Steps 6,1 [10.13 or 6.1[1O.2] for zero MVAR and MONITOR GENERATOR PHASE A(B)(C) amps, 3-EI-57-47(48)(49). D

[10.3.2] WHEN minimum amps are indicated on GENERATOR PHASE A(B)(C) amps, 3El-57-47(48)(49), THEN ZERO MVAR has been obtained. D

[10.43 ADJUST GENERATOR FIELD VOLTAGE MANUAL ADJUST (70P), 3-HS-57-25. UNTIL GEN TRANSFER VOLTS. 3-El-57-41, indicates zero. C 1 1] PERFORM Illustration 7, Turbine-Generator Bearing Metal Temperature, daily. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Grid Instability O-AOl57-1 E Unit 0 Rev. 0007 Page 7 of 18 4.2 Subsequent Action (continued)

[6] IF grid instability is characterized by system voltage being maintained outside the normal limits of 525 +5 KV, THEN PERFORM the following steps:

[6:1] IF system voltage is greater than 540KV, THEN

[6.1 .1] LOWER reactive power to system voltage returns to 530KV, OR UNTIL Generator Reactive power reaches -150 MVAR D

[6.1.2] CHECK 161KV Cap Banks are Out of Service and EVALUATE conditions to determine appropriate actions. REFER TO 0-001-300-4. D

[6.2] IF system voltage is lower than 515KV, THEN PERFORM the following:

[6.3] RAISE reactive power to system voltage returns to 520KV OR UNTIL Generator Reactive Power reaches

-i-200 MVAR, D

[6.4] CHECK 161KV Cap Banks are In Service and EVALUATE conditions to determine appropriate actions.

REFER TO 0-001-300-4. D

[6.5) EVALUATE as applicable, entry into Technical Specifications 3.8.1, 3.8.2, 3.8.7 and 3.8.8. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN 0801 #20 Unit 3 is operating at 80% Reactor.Power and the crew has entered 0-AOI 1 E, Grid Instability, due to the 530 kV system voltage being at 513 kV. The crew reaches the following step in the procedure:

RAISE reactive power until voltage returns to 520 kV Which ONE of the following identifies how to raise reactive power AND the 1.61 kV Capacitor Bank Status that will restore the system voltage in accordance with 0-AOI 1 E?

A. Depress the EHC load set RAISE pushbutton, 3-HS-47-75C; Check the 161 kV Capacitor Banks are IN service.

B. Depress the EHC load set RAISE pushbutton, 3-HS-47-75C; Check the 161 kV Capacitor Banks are OUT of service C. Place the Generator Field Voltage Auto Adjust (90P), 3-HS-57-26, to the RAISE position; check the 161 kV Capacitor Banks are IN service.

D. Place the Generator Field Voltage Auto Adjust (90P), 3-HS-57-26, to the RAISE position; check the 161 kV Capacitor Banks are OUT of service.

Answer: C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295002 Loss of Main Condenser Vac /

Tier # 1 AKI.03 (IOCFR55.41.10)

Knowledge of the operational implications of the following concepts Group # 2 as they apply to LOSS OF MAIN CONDENSER VACUUM: K/A # 295002AK1 .03

• Loss of heat sink Importance Rating 3.6 Proposed Question: # 21 Unit 3 is operating at 28% Reactor Power, when a lightning strike results in a loss of ALL Condenser Circulating Water Pumps. Immediate Actions of 3-AOl-I 00-1, Reactor Scram, are complete.

Which ONE of the following identifies the AUTOMATIC protective actions that will occur?

A. Reactor Feed Pump Turbine trip AND Main Turbine Bypass Valve closure ONLY B. MSIV Closure, Reactor Feed Pump Turbine trip AND Main Turbine Bypass Valve closure ONLY C. Main Turbine trip, Reactor Feed Pump Turbine trip AND Main Turbine Bypass Valve closure ONLY D. MSIV Closure, Main Turbine trip, Reactor Feed Pump Turbine trip AND Main Turbine Bypass Valve closure Proposed Answer: C Explanation A INCORRECT: Plausibility based on misconception that the Main Turbine (Optional): trip is bypassed at <30% Reactor Power. The subsequent Reactor Scram due to Turbine Trip is what is bypassed at < 30% Reactor Power.

B INCORRECT: Plausibility based on misconception that the Main Turbine trip is bypassed at <30% Reactor Power along with misconception that MSIV closure would result from loss of condenser vacuum. See discussion of MSIV Closure in D explanation.

C CORRECT: Main Turbine will trip at condenser vacuum of 21.8 Hg. Both Reactor Feed Pump Turbine Trip and Main Turbine Bypass Valve closure occur at 7 Hg Condenser Vacuum.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Plausibility based on misconception that MSIV closure would result from loss of condenser vacuum. The automatic functions associated with degrading condenser vacuum primarily exist to prevent condenser overpressurization. Even after all the automatic functions occur, the*

condenser is still vulnerable to overpressurization with the MSIVs open.

Therefore, it is very logical that an automatic isolation of MSIVs would occur under these conditions and thus removing all sources of Nuclear Steam to the condenser. To make a comparison, there are several examples that can be found on NRC exams that utilize MSIV closure in response to High-High MSL Radiation. One could not really even argue that it is plausible because it was a Group 1 isolation previously since most plants eliminated the function so long ago. However, It is plausible because it is logical that an automatic isolation of MSIVs would occur under these conditions.

Additionally, this was a distractor suggested by the chief on our previous NRC exam for a loss of condenser vacuum question. Plausibility also based on if Mode Switch is not taken to Shutdown, the MSIVs could close as a result of this transient due to Reactor Pressure < 850 psig with Mode Switch in Run.

KA Justification:

The KA is met because the question tests the candidates knowledge of the operational implications (Main Turbine trip I RFPT Trip I MT Bypass Valve closure) of loss of heat sink (all the Condenser Circ Water Pumps tripping) as it applies to Loss of Main Condenser Vacuum.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Candidate must recognize that 3-AOl-I 00-1 Immediate Actions require Operator to place the Mode Switch to Shutdown. Then, with Mode Switch in Shutdown, recognize MSIV closure at 850 psig is bypassed.

Technical Reference(s): 3-AOI-47-3, Rev. 11 (Attach if not previously provided) 3-01-47, Rev. 91 OPL171.010, Rev. 12 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.010 V.B.12/23 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

BFN Loss of Condenser Vacuum 3-401-47-3 lJnit3 Rev. 0011 Page 5 of 11 NOTES

1) Rising Off-Gas flow wou]d indicate condenser inleakage if the Off-Gas System is functioning propedy Low Off-Gas flow in conjunction with low condenser vacuum could be indicative of an Off-Gas problem.

2) During operations with vali:d CONDENSER A, B, OR C VACUUM LOW 3-PA-47-125 alarm, and condensate temperature of 138 F or greater at the inlet of the SJAE(ICS point 2-28), reduced SJAE First Stage performance (stalling) may occur. This condition will cause reduced Off Gas flow and a loss of vacuum/turbine trip.

[BFPER O2-18D1-DDo]

{

3.0 AUTOMATIC ACTIONS NOTE Turbine trip is expected around 243 inches Hg as indicated on 3-XR-2-2 due to differences between instrument taps for turbine trip and indicated vacuum. (PER 89506)

A. of the following will cause a turbine trip:

1. Condenser A, both 3-PS-047-072A and 728 at 21.8 Hg vacuum.

2. Condenser B, both 3-PS-047-073A 738 at 21 .8 Hg vacuum.

3. Condenser C, both 3-PS-047-074A 4 748 at 21.8 Hg vacuum.

B. RFP turbines trip and main turbine bypass valves closure occurs at -7 Hg hotwell pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Turbine-Generator System 3-0147 Unit3 Rev. 0091 Page 17 of 241 3.4 Turning Gear Operation (continued)

D. For relatively short outages where restart is expected, the Turbine must be maintained on turning gear as long as any shell temperature is above 500°F.

E. If it is necessary to discontinue turning gear operation while the rotors are still hot as indicated by shell metal temperature 500°F, oil flow to the bearings should be maintained to prevent bearing damage due to overheating.

F. If lube oil flow must be stopped with a shell metal temperature greater than 500°F. bearing temperatures should be monitored on THRUST/JOURNAL ERG TEMPERATURE,3-TR-47-23, to ensure Main Turbine bearing metal temperatures do NOT exceed 300°F.

0. Following any shutdown, turning gear operation may be discontinued indefinitely after shell metal temperatures are less than 500°FoK-5cJ H. When the turbine is to be removed from turning gear operation for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a WO should be completed for Electrical Maintenance to remove the main generator and exciter brushes.

3.5 Turbine Trips A. High Reactor Water Level Trip logic for the Main Turbine at +55 inches is taken from Narrow Range level instruments 3-Ll-3-208A, 3-Ll-3-208B. 3-Ll-3-208C, and 3-LI3-208D. The logic is arranged in two channels; Channel A Is fed from 3-Ll-3-208A and 3-Ll-3-208C. Channel B is fed from 3-Ll-3-208B and 3-Ll-3-208D. A trip of the Main Turbine and the RFPTs will occur if both instruments In Channel A, or Channel B sense reactor water level at

+55 inches.

B. Turbine trip on low main condenser vacuum is expected around an indicated 24.3 inches Hg, instead of the 21,8 inches currently stated in this procedure, due to differences between Instrument taps for turbine trip and indicated vacuum.

This condition was discovered during maintenance activities on Unit 3 when condenser vacuum was being monitored by operations using 3-XR-2 on Panel 3-9-6 that is fed by 3-PT-2-1. The instrument tap for 3-PT-2-1 is located just above the condenser tubes, which is the point of highest vacuum. The Instrument taps for the sensing lines feeding the turbine trip switches are located just below the LP turbines. Because of the Volumetric differences between the two locations of the taps, and the steam flow direction from top to bottom, the sensed vacuum is greater at the lower tap than at the higher tap.

(See PER 89506)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71. 010 Revision 12 Page 29 of BC F. Turbine Bypass Valves (Nos. I Through 9) iF-I and TP-7,8 n3% per BPV Purposes Obj.V.B.6.d

a. Routes steam not needed by the turbine to ObjN.C.2.d the condenser during the foIowing Obj.V.E.27 conditions:

(1) Reactor Startup (2) Turbine Roil (3) Turbine Trips (4) Reactor coddown

b. Works in conjunction with The turbine control valves to maintain a constant reactor pressure fore given reactor power level.

c. Provides the capability to prevent over pressurization of the reactor if The MSIVs are open.

2. Location The nine bypass vaes are physically located above the turbine throttle in The moisture separator room near the main turbine stop and control valves.

3. Bypass Valve Design

a. Bypass valves are hydraulically operated, reverse seating globe valves.

b. The valves are positioned as required by a Control PAC and Servo-valves.

c. Valves fail closed upon loss of hydraulics.

rage ai

d. Valves close automaticaliy on loss of vacuum, (1 mercury), to protect the Condenser from over pressurization.

e. The valves route steam from The main steam crosstie header directly to The main condenser.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet N.. Turbine Protection and Reactor Scram Instrumentation

1. Turbine Trips ObJ. VB12 0bj VC.5 Se.tpoint Reason for Thp Obj. VD4 0bj UE2Q

a. High reactor water +5S To prevent moisture level Level 8 carryover from the 213 logic reactor into the turbine

b. Low El-IC control I 1O psig Prevent loss of control oil pressure (FAS) 213 Logic of the turbine processor

p. Loss of condenser <2t.W Hg Indicative of loss of Indicated value vacuum heat sink The turbine will be 243 in is not designed to Hg when turbine operate at low vacuum frp See 01-47 conthtions. precaution &

limitation

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.010 Revision 12 Appendix E Page 67 of 80 Turbine Trip Setpoint Warning First-Out Alarm Overspeed Electrical 107% (1926 rpm) TURD TRIPPED Backup dec. 109% TRIP OVERSPEED (1962 rpm) XA-55-1-1 Generator and 86 devices TURD TRIPPED Transformer Faufts ELECTRICAL TROUBLE XA 1-2 Main Condenser Trip 21.8 inches Hg CONDENSER TURD TRIPPED Vacuum Low vacuum (indicated will A,B, OR C COND VAC LOW be 24.3 in) VACUUM LOW XA-55-1-3 XA-55-7B-1 7 Alarm @ 24.3 inches 1

actual Moisture Separator 11 ft above El 586 MOIST SEP TURD TRIPPED Drain Tank Level floor level LC RES MOIS SEP High LEVEL HIGH LEVEL HIGH XA-56-7C- XA-55-1-4 2,34,16,17,18 Stator Coolant 85 deg° C (810 C U- GEN STATOR TURD TRIPPED Failures 113), or COOL SYS STAT COOLANT 468 gpm (542 gpm U- ABNORMAL SYS FAILURE 1/3) >7726 stator X&557422 X&65 15 amps (70 sec TD)

TURBINE TRIP TIMER INITIATED XA-55-8A MSOP Discharge 106 psig >1300 rpm TURD TRIPPED Pressure Low 213 logic MN SHAFT OIL PUMP INOP XA-55-1 -6 Turbine Trip Annunciators (Cont)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295014 Inadvertent Reactivity Addition Tier# 1 G2.4.50 (10CFR55.41 .10)

Ability to verify system alarm setpoints and operate controls Group# 2 identified in the alarm response manual. KIA# 295014G2.4.50 Importance Rating 42 Proposed Question: # 22 Unit us performing a startup per 1-GOI-100-IA, Unit Startup. When the Operator At The Controls (OATC) placed the rod movement control switch to the single notch out position for the next control rod, the rod quickly moved 3 notches beyond its intended position. The following indications are received:

• SRM PERIOD, (1-9-5A, Window 20), in alarm

• SRM period indicates 25 seconds on I -XI-92-7/44A D -

Which ONE of the following completes the statement below?

The OATC is required to A. INSERT Control Rods until the Reactor is brought subcritical.

B. SHUT DOWN the Reactor until a thorough assessment has been performed.

C. REINSERT the last Control Rod withdrawn to obtain a stable period greater than 60 seconds.

D. STOP Control Rod withdrawal until a stable period of greater than 100 seconds is observed.

Proposed Answer: A Explanation A CORRECT: Per 1-ARP-9-5A and GOI-1 00-lA, IF withdrawing control rods (Optional): and a period less than 30 seconds is observed, THEN INSERT rods until subcriticality is observed.

B IN CORRECT: Plausible in that this is the correct action if a 5 second period indication is observed.

C IN CORRECT: Plausible in that this is the correct action for indication of < 60 but >30 second period.

D INCORRECT: Plausible in that 1-GOI-1 00-lA directs WITHDRAW control rods to maintain a period of 100 seconds or greater.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because to successfully answer this question Operator must be able to verify that the SRM Period alarm as a result of the inadvertent reactivity addition is valid based on period indication. Then, recognize the need to insert control rods until the reactor is subcritical in accordance with the ARP.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 1-ARP-9-5A Rev 16 (Attach if not previously provided) 1GOI-1 00-lA Rev 23 OPL171.059 Rev 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.059 V.B.5 (As available)

Question Source: Bank # 1006 Audit # 69 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-5 l-ARP-9-5A Unit I I-XA-55-5A Rev. 0016 Page 25 of 44 Sensor/Trip Point:

Reiy K21 30 seconds period PERIOD (Page 1 of 1)

Sensor Panel 1-9-12. MCR.

Location:

Probable A. Electrical noise.

Cause: B. Rx power rising on a period of 30 sec.

C. SI (or SR) in progress.

D. Malfunction of sensor.

Automatic None Action:

Operator A. CHECK reactor period meter reading and amber fndicating light Action: illuminated on Panel 195. D B. IF withdrawing control rods and a penod less than 30 seconds is observed. THEN INSERT rods until subcriticality is observed and OBTAIN Reactor Engineer. Reactivity Manager. and Shift Manager permission before pulling any more rods.

NOTE Periods less than S seconds are reportable to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. REFER TO l-AOl-79-2. if applicable. 0 D. REFER TO Tech. Spec. Sect. 3.3.1.2, Table 3.3. 1.2-1 TRM Tables 3.3.4-1 and 3.3.5-I.

References:

i-45E620-6-1 1-730E237-8

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN UnitStartup 1-GOI-100-IA Uniti Rev.0023 Page 79 of 171 5.0 INSTRUCTION STEPS (continued)

[24.4j VERIFY the following Panel 1-9-5 SRM display lights extinguished:

. HIGH HIGH.

. HIGHORINOP. D

. DNSCL,

. BYPASSED (Will be illuminated if channel bypassed.) D

. RETRACT PERMIT N!A if above setpoint.)

. PERIOD. D (R)

Initials Date Time N OTE The following steps apply for all Control Rod Withdrawals and do not require an Operator signoff for the steps. The actions should be reviewed by all personnel involved with withdrawing control rods.

[25] MONITOR Reactor Power during rod withdrawals and perform the following for the associated conditions.

[25.1] IF single-notch withdrawals result in a Reactor period of less than 60 seconds. THEN PERFORM the following:

[25.1.1] REINSERT the last control rod withdrawn to obtain a stable period greater than 60 seconds.

[25.1.2) OBTAIN Reactor Engineer, Reactivity Manager, and Shift Manager permission prior to subsequent control rod withdrawal.

[25.2] IF a Reactor period of less than 30 seconds is observed, THEN PERFORM the following:

[25.2.11 INSERT control rods in accordance with 1-SR-3.13.5(A).

[25.2.2] VERIFY Reactor subcritical.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSiBILITY SUPPORT OL?.O52 Rec sion II Fage 13c23 N2TR:OTC>R JOE2 E*atornanoe of ts hecuip an: c:occwn rate 2-S-3.43.1i:

nonitohng surve anoe is required 5 itnutes oortc bealup ac pre.ssunzator. Use tfVRon 03

3. ewew instruction steps E20 tncgh 5.2 a if a sinpte notch witnorawal res ts in a reacto ::::. V.E.&o ce:co of less Van 80 seconds, the ast control Or. V.C5.b or pulled ill be eserted Lrtil a period o orter than 80 seotfids is obta ec. the Reacor Sngineer, Reactivity Mrage, and SM acntvai 15 required to resume rot ithdraraL V if a eacior period of less than 30 seros is O.

orsensed control rots shall be itserted until Ihe O. V.0 5.o eaclor is subotioa, and obtain the Reactor 2-pincer, eacLvib Manager, ant SW approsal to esune rod wthdrawai

o. a reactor percc of less thanE seocncs S obseryed :he reactor shall be shut down and cannot be restastec unf I an assessnnent has caer.pertarmeo.

d Near end of core fe, cr ticalit\ may occur beoe 1 O,.

Thie doubilngs ore to a stronger top peak 9uk Or. V.0.4 snc the builc.o of piutonum.

e S npie notch wthdrawa rriist bean vnen the Oc. V.B.3 SRM count rate has increaser by a factor o 18 0.. VCS

fourdcublings

, rc nay be stc;:e: after Is t-orgh g stohing the heat range.

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Unit Startup 1-GOI-1 00-lA Unit I Rev. 0023 Page79ofl7I 5.0 INSTRUCTION STEPS (continued)

[24.4) VERIFY the following Panel 1-9-5 SRM display lights extinguished:

• HIGH HIGH.

• HIGHORINOP.

• DNSCL.

• BYPASSED (Will be illuminated if channel bypassed.)

• RETRACT PERMIT (N/A if above setpoint. C

• PERIOD.

(R)

Initials Date Time NOTE The following steps apply for all Control Rod Withdrawals and do not require an Operator signoff for the steps. The actions should be reviewed by all personnel invohed \:ith withdrawing control rods.

[25] MONITOR Reactor Power during rod withdrawals and perform the following for the associated conditions.

[25.1) IF single-notch withdrawals result in a Reactor period of less than 60 seconds, THEN PERFORM the following:

[25.1.1) REINSERT the last control rod withdrawn to obtain a stable period greater than 60 seconds.

[25.1.2) OBTAIN Reactor Engineer. Reactivity Manager, and Shift Manager permission prior to subsequent control rod withdrawal.

[25.2j IF a Reactor period of less than 30 seconds is observed. THEN PERFORM the following:

[25.2.1) INSERT control rods in accordance with 1-SR-3.1.3.5(A).

[25.2.21 VERIFY Reactor subcritical.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN UnitStartup 1..GO1-100..IA Uniti Rev.0023 Page8Oofl7l 5.0 INSTRUCTION STEPS (continued)

[25.2.3] OBTAIN Reactor Engineer, Reactivity Manager, and Shift Manager permission prior to subsequent control rod withdrawal.

[25.3] IF a Reactor period of less than 5 seconds is observed. THEN SHUT DOWN the Reactor until a thorough assessment has been performed. REFER TO 1-GOl-100-12A.

CAUTION Criticality should he expected at all times.

[25.4] COMMENCE rod withdrawal. REFER TO 1-01-85 and 1-SR-3.1.3.5(A).

(R)

Initials Date Time

[25.5] CHECK coupling integrity by performing i-SR-3.1.3.5(A) as each control rod is withdrawn.

(R)

Initials Date Time

[25.8) waic MONITOR SRM!IRM instrumentation closely during rod withdrawal while approaching criticality, pausing between rod withdrawals as needed for neutron level stabilization. [INPOSEROO6]

(R)

Initials Date Time

[25.7] CONTINUE withdrawing control rods in accordance with 1-SR-3.1 .3.5(A).

(R)

Initials Date Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSiBILITY SUPPORT BFN lJnitStartup 1-GOI-1001A Unit I Rev. 0023 Page 82 of 171 5.0 INSTRUCTION STEPS (continued)

CAUTIONS

1) Criticality should be expected at all times.

2) Extended operation close to The point of criticality could result in inadvertent criticality and must be avoided.

[27] WHEN in a configuration that is expected to be near critical AND Nuclear Instrument response is NOT as expected. THEN NOTIFY Reactor Engineer and Shift Manager.

Initials Date Time

[28] IF operation is to be suspended for greater than one hour near the point of criticality. THEN (Otherwise N/A)

PLACE the Reactor core sufficiently subcritical as directed by the Shift Manager and as advised by the Reactor Engineer, to avoid an inadvertent criticality.

Initials Date Time

[29] WITHDRAW control rods to maintain a period of 100 seconds or greater as indicated on the following indicators on Panel 1-9-5:

• CHANNELA PERIOD. 1-XI-92-7144A. C

• CHANNEL B PERIOD, 1-XI-92-7!44B. C

• CHANNEL C PERIOD, 1-XI-92-7/44C. C

• CHANNEL D PERIOD, 1-XI-92-7/44D. C (R)

Initials Date Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295022 Loss of CRD Pumps/I Tier # 1 AAI.O1 (IOCFR 55.41 .7)

Ability to operate and/or monitor the following as they apply to LOSS Group # 2 OF CRD PUMPS: K/A # 295022AA1 .01 CRD hydraulic system Importance Rating 3.1 Proposed Question: # 23 Unit I is at 100% Reactor Power when Control Rod Drive (CRD) Pump IA trips. During the start of CRD Pump I B, the following occurs:

• Control Rod 30-23 moves from position 16 to position 14

• Control Rod 38-31 moves from position 16 to position 12 Which ONE of the following identifies the required action(s) in accordance with CRD AOls?

A. Immediately Scram the Reactor.

B. Reduce Reactor Power to 90%

C. Reduce Core Flow to 60% AND then Scram the Reactor.

D. Insert Control Rods 30-23 AND 38-31 to position 00 using CONTINUOUS IN.

Proposed Answer: A Explanation A CORRECT: In accordance with 1-AOI-85-6, if more than 1 CR drifts, insert (Optional): a reactor Scram Immediately B INCORRECT: Plausible in that this is correct AOl actions for a single Control Rod Drifting out and unable to insert the control rod.

c INCORRECT: Plausible in that AOl action requiring a Reactor Scram are typically preempted with Core Flow reduction to 60%.

D INCORRECT: Plausible in that this is correct AOl actions for a single Control Rod Drifting out.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates ability to monitor the CRD hydraulic system as it applies to Loss of the in service CRD Pump. Trip of CRD pump requires start of the standby pump. During start of the standby Pump, the CRD Hydraulic system is susceptible to inadvertent control rod drift if flow is raised rapidly or there is significant seat leakage on the in service CRD flow control valve.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Operator must diagnose multiple control rod drifts based on indication and select appropriate action.

Technical Reference(s): 1-AOI-85-5 Rev. 1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.074 V.B.2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Rod Drift In 1-AO1-85-5 Unit I Rev. 0001 Page 5 of 10 4.0 OPERATOR ACTIONS

( 4.1 Immediate Actions

[1] IF multiple rods are drifting into core, THEN t MANUALLY SCRAM Reactor. REFER TO 1..AOl-i00-1.

4.2 Subsequent Actions

[1) IF the Control Rod is moving from its intended position without operator actions. THEN INSERT the Control Rod to position 00 using CONTINUOUS IN. (Otherwise NIA)

[2] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

[3] IF another Control Rod Drift occurs before Reactor Engineering provides a verbal or written evaluation. THEN MANUALLY SCRAM Reactor and enter AOl-100-1. D

[4] CHECK Thermal Limits on ICS by running OFFICIAL 3D.

[5) ADJUST control rod pattern as directed by Reactor Engineer and CHECK Thermal Limits on 105 (RUN OFFICIAL 3DI. D

[S) IF CRD Cooling Water Header DP is excessive and causing the control rod drift, THEN ADJUST CRD SYSTEM FLOW CONTROL. 1-FIC-85-11. and CRD DRIVE WATER PRESS CONTROL VLV, 1-HS-65-23A.

as required to establish the following: (Otherwise N/A)

• CRD DRIVE WTR HDR DP, 1-PDI-85-17A. between 250 and 270 psid

• CRD SYSTEM FLOW CONTROL, i-FIC-85-11, between 4oand65gpm U

• CRD CLG WTR HDR DP, 1-PD 1-85-iSA, at about 20 psid or as close as possible while maintaining flow and Header pressure. U

Sample Written Examination Form ES-401-5 ES-401 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT CR0 System Failure 1AOI85.3 BFN Unit I Rev. 0004 Page 9 of 12 4.2 Subsequent Actions (continued)

[2] IF Reactor Pressure is greater than or equal to 900 psc AND Charging Water Pressure can NOT be restored and niaintaned greater than 940 psig within 20 minutes, AND Two or more Scram accumulators are INOP with associated control rod NQT fully inserted, THEN PERFORM the following: (Otherwise N/A) 12.1] iF core flow is above 60%, THEN REDUCE core flow to between 50-60%.

[2.2] MANUALLY SCRAM Reactor and IMMEDIATELY PLACE the Reactor Mode Switch in the SHUTDOWN position. D

[2.3] REFER TO l-AOl-i00-1.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBLITY SUPPORT BFN Rod Drift Out 1.AOI-85-6 Unit I Rev. 0001 Page 5 of 9 4.0 OPERATOR ACTIONS 4.1 Immediate Actions

[1] IF mulUple Control Rod drifts are identified, THEN MANUALLY SCRAM the Reactor and enter 1-AOl-laO-i D 4.2 Subsequent Actions 1 [1] IF a Control Rod is moving from its intended position without operator actions, THEN I SELECT the drifting control rod and INSERT to the FULL IN (00) position. D CAUTION

[NRC?C] Operations outside of the allowable regions shown on the Recirculation System Operating Map could result in thermal-hydraulic power oscillations and subseq uent fuel damage. i-GOI-100-i2A should be referenced for required actions and monito ring to be performed during a power decrease. (NcO 940245010)

[2] IF Control Rod Drive does NOT respond to INSERT signal.

TH EN PERFORM the following: (Otherwise NIA)

[2.1] REDUCE Total Core Flow, as indicated on TOTAL CORE FLOW/CORE PRESS DROP. i-XR-8-50 on Panel 1-9-5, by approximately 10% to control possible power increase. D

[2.2] lNE!CJ IF drifting control rod is causing Reactor power to rapidly rise at a rate which can t4Q]E be controlled by reducing recirculation flow. THEN MANUALLY SCRAM the Reactor. (Otherwise N/A) lNPO SER 90-OlE) D

[3] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295029 High Suppression Pool Wtr Lvi! 5 Tier # 1 EK2.02 (100FR 55.41 .7)

Knowledge of the interrelations between HIGH SUPPRESSION Group # 2 POOL WATER LEVEL and the following: K/A # 295029EK2.02 HPCI: Plant-Specific Importance Rating 3.4 Proposed Question: # 24 Unit 1 Suppression Pool Level is (+) 7 inches.

Which ONE of the following completes the statements below?

HPCI Suction (1)_ automatically transfer to the Suppression Pool.

RCIC Suction (2) automatically transfer to the Suppression Pool.

A. (1) will (2) will B. (1) will (2) will NOT C. (1)will NOT (2) will D. (1)will NOT (2) will NOT Proposed Answer: B Explanation A INCORRECT: Part 1 correct See explanation B. Part 2 incorrect See (Optional): Explanation C.

B CORRECT: Part 1 correct HPCI Suction automatically swaps to suppression pool on high suppression pool level +5.25 or low CST level Elev <5526. Part 2 correct RCIC has no automatic transfer from CST to torus.

c INCORRECT: Part 1 incorrect Plausible in that this would be true for RCIC. Part 2 incorrect Plausible in that this would be true for H PCI.

D INCORRECT: Part 1 incorrect See explanation C. Part 2 correct See Explanation B.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the interrelations between High Suppression Pool Water Level and HPCI Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL171.040 Rev. 23/1-01-73 Rev. 17 (Attach if not previously provided)

OPL171.042 Rev. 20 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.042 V.B.1 (As available)

OPL171.040 V.B.6 Question Source:

Modified Bank # (Note changes or attach parent) x Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7t042 Revision 21)

Page II of 69 INSTRUCTOR NOTES (4) Exhaust through check valve to suppression pool c Water path TP-1 (1) Normal condensate path from OST to the HPCI pump, to the A Feedwater line and into the reactor vessel (2) Alternate suction path from suppression pool (3) Automatic swapover to suppression Approximately 7000 pool on high suppression pool level gallons left in (ST

+5.25 or low (ST leve Elev <5526 piping when auto swap occurs on low NOTE: There is a 5-second time delay GET Level for suction swap on high suppression pool level.

(4) Test flow path to GET (5) Test line orifice which provides a discharge head to simulate reactor pressure

d. Drain System (discussed in detail in section BA)

e. Turbine auxiliaries (discussed in detail in section B.2)

(1) Gland seal condenser for leakoffs

2) Cooling water for gland seal condenser (3) 0 land seal condenser blower (4) (3land seal condenser condensate p Li nip

ES-401 Sample Wriften Examination Form ES-401-5 Question Worksheet OPLI7I.040 Revision 23 Page 35 of 74 (3) Unit 3 power supply to the EGM Control Box is Div I ECCS Inverter.

(4) If Bus B fails, B channel trp logic and B channel isolation logic will he inoperative.

e. Steam line break Obj. V.B.1 l.a.

Obj. V.C.7.a RCIC is provided with two independent flow to detect high steam flow. High steam flow of :150% for 3 seconds measured on either one or both flow elements will isolate PCV 71-2 and 71-3.

f. Low CST level Obj. V.B.i lb.

Ohj. V.C.7.h RCIC has no automatic transfer from CBT to torus. 01- Ohj. V.E.13 71 directs transfer when HPCI auto transfers on low Ohj. V.E.14 CST level or high torus level and if RCIC trips on low suction pressure 10 Hg vacuum.

g. High suppression pool temperature Obj. \.B.ii.b Obi. \/.C:.7.h (1) RFIC is normally aligned to the CST for pump Obj. V.B.Il.c suction cooling water. Suppression pool temperature will adversely affect the pools capacity as a heat sink. While performing RCIC surveillance. pool temperature is monitored and pool cooling is directed at 95°F bulk temperature.

Calculations have shown a 1°F rise torus temperature for every 16 minutes of operation.

(2) When RCIC is operating on suppression pool Obj. V.D.lO suction and the following alarms are received: Obj. V.D.1 1 Obj. \/.B:12 RCIC OIL CLR OUTLET DISCH OIL HI TEMP (120°F)

RCIC GOVERNOR END BEARING HIGH TEMP (lOOt)

RCIC COUPL END BEARING TEMP HIGH ( 160°F)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BPN High Pressure Coolant Injection 1.01-73 Unit I System Rev. 0017 Page 10 of 78 3.0 PRECAUTIONS AND LIMITATIONS (continued)

E. When any of the following signals are received, HPCI SUPPR POOL OUTED SUCT \!LV, 1-FCV-073-0027, and HPCI SUPPR POOL INBD SUCT VALVE.

i-FCV-073-0026 automatically open, unless a I-IPCI isolation signal is present.

1. Suppression Pool Level High at +5.25 in.

2. HPCI Pump Suction Condensate Header Level Low at approximately 7000 gallons (El. 5526 on 1-LS-073-0056A and 1-LS-073-0056B)

F. When HPCI SUPPR POOL OUTBD SUCT VLV, 1-FC\/-073-0027 and HPCI SUPPR POOL INBD SUCT !LV, 1-FCV-073-0026 are fully open, HPCI CST SUCTION \!ALVE. 1 -FCV-073-0040 automatically closes.

G. When either HPCI SUPPR POOL OUTED SUCTVLV, i-FC\/-073-0027, or HPCI SUPPR POOL INBD SUCT \LV. 1-FCV-073-0026. is FULL OPEN, the HPCIIRCIC CST TEST VLV, 1-FCV-073-0036, and HPCI PUMP CST TEST VLV, 1-FCV-073-0035. close.

H. When the HPCI TURBINE STEAM SUPPLY VALVE, 1-FCV-073-0016, is opened, the following valves close:

1. HPCI HOTWELL PUMP INBD ISOLVLV. 1-FCV-073-0017A

2. HPCI HOTWELL PUMP OUTED ISOL VLV. 1-FCV-073-0017B

3. HPCI STM LINE INBD DRAIN \JLV. 1-FCV-073-000GA

4. HPCI STM LINE OUTED ISOL VLV. 1-FCV-073-000GB I. The HPCI PUMP MIN FLOW \/ALVE, 1-FCV-073-0030, automatically opens when system flow is at or below 900 gprn (lowering) if a system initiation signal is present. and automatically closes when system flow is at or above 1255 gpm (rising) regardless of presence of initiation signal.

J. HPCI PUMP MIN FLOW VALVE. 1-FCV-073-0030. opens on receipt of an initiation signal, even with HPCI Auxiliary Oil pump in PULL-TO-LOCK position.

resulting in slowly draining CST to Suppression Chamber.

K. When a HPCI System isolation signal is reset, the steam line isolation valves do not automatically open, and are required to be opened via hands\vitch operation, even if a system initiation signal is present.

L. HPCI turbine operation below 2.400 rpm should be minirnied to ensure adequate oil pressure from the turbine driven oil pump, to reduce system vibration, and prevent possible water hammer in the exhaust line.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO  :

295034 Secondary Containment Ventiiation High Radiation /

Tier #

EA2 02 (1 OCFR 55 41 10)

Ability to determine and/or interpret the following as they apply to Group # 2 SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: K/A # 295034EA2.02

. Cause of high radiation levels Importance Rating 31 Proposed Question: # 25 Unit 1 is at 100% Reactor Power with the following system line ups:

. Reactor Building Closed Cooling Water (RBCCW) Pumps 1A AND I B are in service

. Reactor Water Cleanup (RWCU) Pumps 1A AND I B are in service

. Fuel Pool Cooling and Cleanup (FPCC) Pump IA is in service Unit I Reactor Scrams AND the following alarms I indications are received:

. 480 V Shutdown Board IA is locked out

. RBCCW SURGE TANK LEVEL HIGH, (1-9-4C, Window 6)

. RBCCW EFFLUENT RADIATION HIGH, (1-9-3A, Window 17)

. RX BLDG,TURB BLDG, RE ZONE EXH RADIATION HIGH, (1-9-3A, Window 4)

Which ONE of the following is a potential cause of the alarms?

Leakage into RBCCW from A. Reactor Recirc Pump seal coolers B. Fuel Pool Cooling Heat Exchangers C. Reactor Water Cleanup Pump Seal Coolers D. Reactor Water Cleanup Non-Regenerative Heat Exchangers Proposed Answer: A Explanation A CORRECT: With the isolation of RWCU at (+) 2 inches due to the scram (Optional): and the loss of FPCC due to the lock out of Shutdown Board 1A, this remains the only choice that is not tripped and/or isolated. RBCCW Pump 1 B remains in service supplying Reactor Recirc Pump seal coolers.

Therefore, this is a potential source of inleakage into RBCCW and source of the high radiations alarms.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B INCORRECT: Fuel Pool Cooling Pump A power supply is from 480 V Shutdown Board 1A which is locked out. Any FPCC Heat Exchanger leakage would result in leakage into the FPCC system and not into RBCCW. Therefore, this could NOT be the source of the inleakage into RBCCW and the resulting high radiation alarms. Plausible in that FPCC is an RBCCW load and with the absence of the power loss, this could be the source of the Radiation Alarms.

c INCORRECT: With a Scram from 100% power, Reactor Level drops less than (+) 2 inches, RWCU is isolated and the Pumps are tripped. Therefore, this could NOT be the source of the inleakage into RBCCW and the resulting high radiation alarms. Plausible in that RWCU is an RBCCW load and with the absence of the isolation, this could be the source of the Radiation Alarms.

D INCORRECT: With a Scram from 100% power, Reactor Level drops less than (+) 2 inches, RWCU is isolated and the Pumps are tripped. Therefore, this could NOT be the source of the inleakage into RBCCW and the resulting high radiation alarms. Plausible in that RWCU is an RBCCW load and with the absence of the isolation, this could be the source of the Radiation Alarms.

KA Justification:

The KA is met because it tests the candidates ability to assess the status of RBCCW and its loads to determine the cause of high radiation levels indicated in Secondary Containment and Secondary Containment Ventilation.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71.047 Rev. 12 (Attach if not previously provided) 1- ARP-9-3A Rev. 40/1- ARP-9-4C Rev. 18 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171 .033 V.B.3 (As available)

Question Source: Bank#1 r Modified Bank# (Note changes or attach parent)

New X Question History: Last JI Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet FPCC LESSON PLAN OPL171 .052 Revision 10 Page 24 of 49 INSTRUCTOR NOTES

11. Circulating pumps

a. Purpose To provide forced circulation of water through the system and back to the pool (1) Quantity-2 (2) Type centrifugal horizontal (3) Capacity 600 gpm each (4) Electrical supplies Ohj. \,B.B.d Obj. V.C.2.d (a) Pump l from 480 V Shutdown Obj. V.D.7 Board IA (similar for Unit 2 & 3) Obj. V.E.7 (b) Pump 1 B from 480 V Shutdo.vn Board lB (Similar for Unit 2 & 3)

(5) Control of the pumps is from either the Procedural control room (panel 9-4) or the local directions use the panel by the pumps in the reactor MCR switch on building on elevation 621. panel 9-4.

(6) Operation under normal conditions

- both heat system flow will be 500 gpm (utilizing exchangers and one pump). To handle the maximum the D demin as normal heat load, both pumps are well.

required to be operating, each at 500 gpm flow. Pump disch pressure Procedure normally 140 psig. Adherence (7) System design flow rate is 600 gpm.

System maximum flow rate is 1200 gprn.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DPL 171 04 7 Revision 12 Page 10 0(41

d. Proper system flow operation is assured by Done Each Shift monitoring the system DP (pump discharge minus tumo suction).

I RBCCW Heat Loads

a. Essential loop toads Obj. V.B.2

• Drywell BlowersQlO) Qhj. V.D.2

• Reactor recirculation pump motor coolers (2)

• Reactor recirculation pump seal coolers (2)

• Drywell equipment drain sump heat exchanger (1)

b. Non-essential loop loads C)bj. V.B.3

• Reactor Building equipment drain Ohj. V.D.3 sump heat exchanger (1)

• Reactor water cleanup pump seal water coolers and bearing oil coolers (2)

• RWCU Non-regenerative heat exchangers (2)

• Fuel pool cooling heat exchangeis (2)

• Reactor recirculation pump discharge sample cooler (1)

3. RBCCW Heat Exchangers

a. These provide the means for heat removal DON 51195 from RBCCW by ROW with Emergency replaced HXlo. &

Equipment Cooling Water (EECW) as a IB, HX 10 NOT backup replaced

b. They are counter-flow type; 50% capacity OPL171 .051 each.

• RBCCW flow makes one pass through the shell side.

• ROW makes one pass through the tube side.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .013 Revision 16 Page 27 of 47

5. Significant Interlocks and Trip Logic Obj. V.8.3: V.D.4 Obj. \!.C.3: V.D.5

a. System isolation (closure of the inboard and Obj. V.8.5: V.E.6 outboard inlet isolation valves FCV 69-1 Obj. \!.C.4: V.E.9 and 69-2 and the return isolation valve FCV Note: 69-12 69-12) will occur on any of the following closes on isol.

signals or conditions. Signal but is not a (1) Low reactor water level (level 3) to protect the core in case of a break in PCIS valve.

Level 3 is Tech Spec terminology RVJCU System piping or equipment. (>528 above One-out-of-two taken twice logic. vessel zero). Also, Note: Tech Specs low level setpoint >0. stated as 0 Current actual setpoint is still +2. indicated level.

LT-3-203A thru D (2) High temperature in areas occupied by RWCU equipment and piping to Refer to ARPs for isolate system in case of a piping latest setpoints break.

(a) Hi Temp PCIS Isolation Logic is triggered by at least two of twenty-four temperature switches. These switches cause alarms on Panel 9-5.

These switches are installed in the following locations and can be read in the Aux Instrument Room.

. Main Steam Tunnel 834Athru D

• Pipe Trench 835 A thru D

• A Pump Room 836 Athru 0

• 8 Pump Room 837 Athru D

• East Wall Hx Room 838 A thru D

• West Wall Hx Room 839 A thru 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9.3 1-ARP-9-3A Unit I XA=55-3A Rev. 0040 Page 26 of 52 SensocTrip Point:

RBCCW EFFLUENT HI HI-HI RADIATION HIGH 1-RA-90-131A I-RM-90-13lD (Note I) (Note 1)

(1) ChemLab should he contacted for current selpoints per 0-Tl-45.

(Page 1 of 2)

Sensor I-RE-090-0131A (off-line) RBCCW HX, Rx Bldg, EL59Y, R-R2 Location:

Probable Hx tube leak into RBCCW system.

Cause:

Automatic None Action:

Operator A. DETERMINE cause of alarm by obseriing the following:

Action: I CHECK RBCCW EFFLUENT OFFLINE RAD MON.

1-RM-90-l:31D. Panel l-9-l0. C

8. NOTIFY Chemistry to sample RBCCW for total gamma activity to verity condition.

C. DETERMINE if source of leak is R\NCU Non-Reqenerative Heat Exchanger, Fuel Pool C:ooling, or Reactor Water Sample. Recirc Pump A or B Seal Water Heat Exchanger(s). C U. [N ER/C] CHECK the following for indication of Reciro Pump Seal Heat Exchanger leak:

• LOWERING in Reactor Recirculation Pump lA(18) NO. 1 or2 SEAL. I-Pl-68-64A or I-Pl-68-03A (l-Pl-68-JOA or i-Pl-68-75A) on Panel 1-9-4 C

• Temperature rise on CLG WTR FROM SEAL CLG i-TE-68-54, on RECIRC PMP MTR IA WINDING AND ERG TEMP temperature recorder I-TR-68-58. on Panel 1-9-21, C

• Temperature rise on CLG WTR FROM SEAL CLG l-TE-68-67.

on RECIRC PMP MTR I B WINDING AND ERG TEMP temperature recorder, l-TR-68-84, on Panel 1-9-2 1 C Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9.3 1.ARP.93A Unit I XA.55.3A Rev. 0040 Page 9 of 52 Sensol7Trip Point:

RX BLDG,TURB BLDG.

RF ZONE EXH i-RFvI-90-2F0 RADIATION HIGH 1 -RA.90-250A Gas HIGH ALRft1 6594 CPM ALERT 3297 CPM r

(Page 1 of 1)

Sensor E 664 Refuel Floor R-4 P-Line Location:

Probable A Doily source check.

Cause: B. High radiation in the Reactor Building, Turbine Building, Refuel Zone exhaust ventilation ducts.

C. On, Cask stora e activities in progress.

Automatic None Action:

Operator A. CHECK i-RM-90-250 on Panel I-9-2 (O-MON-90-361)and Action: MONITOR activity levels on recorder AIR PARTICLILATE MONITOR CONTROLLER i-MON-go-SO on Panel 1-9-2. C B. IF high activity is conformed, THEN NOTIFY RAD PRO. C C:. REQUEST Chernisti perform analysis to determine source. C D IF Dry Cask storaqe activities are in progress, THEN NOTIFY CASK Supervisor. C E. IF the TSC is NOT manned. THEN EVACUATE personnel from affected oreas C F. IF the TSC is manned. THEN REQUEST the TSC to evacuate non-essential personnel from affected areas.

G. REFERTO i-AOl-79-l or i-AOI-79-2 if applicable. C H. MONITOR release rate for 00CM compliance. C I. IF 00CM Limits are exceeded. THEN REFERTOEPIP-i. C J. IF Eherline is operable, THEN REFER TO 1-01-90. to reset alarms

References:

047W600$0 147E610901 45E6203 i-Slr\*ll-OOB TVACaIcN0000902005008/EDC63693

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-4 1-ARP94C Unit I 1-XA-55-4C Rev. 0018 Page 12 of 43 Sensor/Trip Point:

RECA SURGE TANK LEVEL HIGH i-LS-c70-0002A 4 Inches Above Center Line of Tank 1 -LA-70-2A (Page 1 of 2)

Sensor RBCCW surge tank on the foLlrth floor in the M-G -set room.

Location:

Probable A. Makeup valve 1-FCV-70-1 open.

Caus : B. Bypass valve 1-2-1369 leaking.

C. Leak into the system.

Automatic None Action:

Operator A. VERIFY make-up value 1-FCV-70-1 closed, using RBCCW SYS Action: SURGE TANK FILL VALVE, i-HS-70--1, on Panel 1-9-4. C B. CHECK RBCCW PUMP SUCTION HDR TEMP. l-TIS-70-3, indicates water temperature is 100F or less, on Panel 1-9-4. C C. DISPATCH personnel to verify high level, ensure bypass valve.

1-2-1369. is closed and observe sight glass leveL C D. OPEN surge tank drain valve, 1-70-609. then CLOSE valve when desired level is obtained. C E. REQUEST Chemistty to pull and analyze a san]ple for total gamma activity and attempt to qualify source of leak. C F. CHECK activity reading on RM-90-131D. C Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUS1BILITY SUPPORT OPL17l .047 Revision 12 Page 10 of 41

d. Proper system flow operation is assured by Done Each Shift monitoring the system DP (pump discharge minus punip suction).

2. RBCCW Heat Loads

a. Essential loop loads Obj. V.B.2

• D&we!l Blowers(10) Obi. V D.2

• Reactor recirculation pump motor coolers (2)

• Reactor recirculation pump seal coolers (2)

• Drywell equipment drain sump heat exchanger (1)

b. Non-essential loop loads Obj. V.B.3

• Reactor Building equipment drain Obj. V.D.3 sump heat exchanger (1)

• Reactor water cleanup pump seal water coolers and hearinq oil coolers (2)

• RWCU Non-regenerative heat r exchangers (2)

• Fuel pool cooling heat exchangers (2)

• Reactor recirculation pump discharge sample cooler (1)

3. RBCOW Heat Exchangers

a. These provide the means for heat removal DON 51105, froni RBCCW by ROW with Emergency replaced HX1A &

Equipment Cooling Water (EECW) as a IS, HX IC NOT backup. replaced.

b. They are counter-flow type, 50% capacity OPLI71 .051 each.

• RBCCW flow makes one pass through the shell side.

• ROW makes one pass through the tube side.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL17l .047 Revision 12 Page 11 of4l RBCCW flow is in the opposite direction to ROW flow.

c. The spare RBCCW heat exchanger has manual isolation valves which allow it to he lined up to Unit 1, 2, or 3.

4. Chemical Feeder

a. Provides for addition of chemicals

b. It is the bypass type, The RBCCW pumps provide the DP for chemical feed injection.

c. Sodium nitrite is injected as a rust inhibitor and for pH control.

5. Expansion Tank RB EL 639

a. Allows for water expansion from temperature and pressure changes within RBCCW System

b. Provides adequate NPSH to RBCCW pumps

c. Provides a place to add makeup v,ater to Pn! 9-4 RBCCW from demineralized water, through FCV-70-1. or through a manual bypass valve.

d. Provides an overflow to Reactor Building floor drain sump.

e. High and low levels in the expansion tank +/- 4 above/be low alarm in the MCR. This allows leaks to be centerline detected. A high level alarm with no high radiation alarm means RCW is probably leaking into the system. A high level alarm with a high radiation alarm indicates in-leakage from a heat load such as Fuel Pool Cooling or RWCU.

f. The tank is provided with a vent pipe at the top which prevents pressure buildup in the RBCCW System.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SI 295036 Secondary Containment High Sump/Area Water Level / 5 Tier# 1 EK3.O1 (1 OCFR 55.41.5)

Knowledge of the reasons for the following responses as they apply Group # 2 to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER KIA# 295036EK3.O1 LEVEL:

. Emergency depressurization Importance Rating 2.6 Proposed Question: # 26 A HPCI Steam Supply leak has resulted in elevated Secondary Containment temperatures AND area water levels. HPCI Steam Supply Isolation valves have failed to isolate AND CANNOT be manually closed. Two Secondary Containment Water Levels are above their Maximum Safe Value requiring Emergency Depressurization.

Which ONE of the following completes the statement below?

In accordance with EOl-3, Secondary Containment Control Bases, ALL of the following are reasons for requiring Emergency Depressurization with the EXCEPTION of A. to place the primary system in the lowest possible energy state B. to reject decay heat to the suppression pool, rather than secondary containment C. to reduce driving head and flow of primary systems that are unisolated and discharging into secondary containment D. to allow access into the Reactor Building by the Emergency Response Organization to locate and manually isolate the leak Proposed Answer: D Explanation A INCORRECT: This is one of the four reasons specified in EOIPM Section (Optional): O-V-E for Emergency Depressurizing with 2 or more area water levels above the Maximum Safe Operating Value with a Primary System discharging into Secondary CTMT.

B INCORRECT: This is one of the four reasons specified in EQ 1PM Section O-V-E for Emergency Depressurizing with 2 or more area water levels above the Maximum Safe Operating Value with a Primary System discharging into Secondary CTMT.

c INCORRECT: This is one of the four reasons specified in EQ 1PM Section O-V-E for Emergency Depressurizing with 2 or more area water levels above the Maximum Safe Operating Value with a Primary System discharging into Secondary CTMT.

D CORRECT: This is NOT one of the four reasons specified in EOIPM Section O-V-E for Emergency Depressurizing with 2 or more area water levels above the Maximum Safe Operating Value with a Primary System discharging into Secondary CTMT.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question test knowledge of the reasons for Emergency Depressurization as it applies to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVELS.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Justification:

Technical Reference(s): OPL 171 .204 Rev. 7 (Attach if not previously provided)

EOIPM 0-V-E Rev. 1 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the faculty since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EO1-3, SECONDARY CONTAINMENT CONTROL BASES EOl PROGRAM MANUAL SECTION O-V-E

[ DISCUSSION SC/L-14 This signal step informs the opcrator that actions to control RPV pressure must immediately change because of present plant conditions.

When ernergeacy RPV depressurization is required, the operator shall transfer RPV pressure control actions to C2, Emergency RPV Depressurization.

This step has been reached because water levels in two or more secondary containment areas have exceeded maximum safe operating value, and a direct threat exists relative to secondary containment integrity, equipment located in secondary containment, and continued safe operation ofthe plant. The RPV must be rapidly depressurized for the following reasons:

• To reducefprevent further increase in secondary containment levels.

• To place the primary system in the lowest possible energy state.

• To reduce driving head and flow of primary systems that are unisolated and discharging into secondary containment.

• To reject decay heat to the suppression pool, rather than secondary containment.

REVISION I PAGE 53 OF 73 SECTION O-V-E

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.204 Revision 7 Page 31 AND 32 of 52

o. SC/L-14 This signal step informs the operator that actions to control RPV pressure must immediately change because of present plant conditions.

When emergency RPV depressurization is required, the operator shall transfer RPV pressure control actions to C2, Emergency RPV Depressurization.

This step has been reached because water levels in Obj.V.B.7 two or more secondary containment areas have Obj.V.C.7 exceeded maximum safe operating value, and a direct threat exists relative to secondary containment integrity, equipment located in secondary containment, and continued safe operation of the plant. The RPV must be rapidly depressurized for the following reasons:

To reduce/prevent further rises in secondary containment levels.

To place the primary system in the lowest possible energy state.

• To reduce driving head and flow of primary systems that are unisolated and discharging into secondary containment.

• To reject decay heat to the suppression pool, rather than secondary containment.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 500000 High Containment Hydrogen Concentration Tier# 1 EK2.09 (IOCFR 55.41.7)

Knowledge of the interrelations between HIGH CONTAINMENT Group# 2 HYDROGEN CONCENTRATIONS the following: KIA# 500000EK2.09 Drywell nitrogen purge system Importance Rating 3.0 Proposed Question: # 27 Unit 2 was operating at 100% Reactor Power when a LOCA occurred. Plant conditions are as follows:

• Drywell H 2 is 3% increasing

• Drywell 02 is 4% increasing

• Suppression Chamber H 2 is 2% steady

• Suppression Chamber 02 is 3% steady Which ONE of the following completes the statement below?

Based on the above conditions, Nitrogen must be lined up to A. the Drywell B. the Suppression Chamber C. the Drywell AND Suppression Chamber D. NO primary containment area; the Primary Containment EOI entry condition for hydrogen concentration has NOT been exceeded Proposed Answer: A Explanation A CORRECT: 2-EOl-2 directs monitoring and controlling Drywell and (Optional): Suppression Chamber, H2 at or below 2.4% AND 02 at or below 3.3%. The Drywell is above both values. 3% H2 in the Drywell is greater than 2.3%, the minimum detectable value. 2-EOI Appendix 14A states to continue in the procedure when H2 or 02 concentration(s) are increasing. The stem states both are increasing in the Drywell. It then directs the operator to determine which area has the highest H2 or 02 concentrations and directs adding nitrogen to that area to reduce the concentration(s).

B INCORRECT: Suppression Chamber H2 and 02 are below the control limits, NO change is occurring, and lower than the Drywell. Plausible if the candidate doesnt know the control parameter values. H2 value is below the BFN mm detectable value of 2.3%.

C INCORRECT: You never add to both areas at once. Procedure adds to one area at a time. Plausible since both areas have elevated H2 and 02 concentrations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Procedure addresses correcting area before 3%. Drywell is above control parameters and increasing. Candidate may not know the EOl entry condition for primary, containment hydrogen concentration.

KA Justification:

The KA is met because the question tests knowledge of the interrelations between elevated Primary Containment Hydrogen levels and Nitrogen makeup to Containment.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. The RO has to know the primary containment entry condition for high hydrogen concentration and deduce which area has the worst degrading conditions based on that fact.

Technical Reference(s): 2-EOI-2 Rev 10, OPL1 71 .032 Rev 12 (Attach if not previously provided) 2-EOI-Appendix 14A Rev 7 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.3 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TON( SUIPR PLTEMP

(

j WHILE EXECUTiNG THIS PROCEDURE:

! I!!!!!

SASIOEMTRY IS RE JIREO ThE TSC EXITAU £01 W01IARTSAS)

SAM TEAM HASASSUO C0I AND PERIFORM SAEIQ ACTIONS AS DIERCTED CONTROL BY TIE SAM TEAM SARO EN1RY IS REQUIRED ENTER SPA, AT STEP SRL PVC-I CAU11O

1. 4 PCPRCSSVSPUMPIIFSH E1RESPECTIVE OP THE EN1RY COIANTIONS, EXECUTE DWT. PC!P. PC

.4, WHILE EXECUTING THE FOLLOWING STEPS:

I. RESEFAHALYZER ISOlATION, H2JD2ANALYZEI4S ARE IN STANSBY IF NECESSARY

2. $ELECT 17W OR SUPPRCBMBR AND NO PCISCROUPE ISOLATION EXETE MOMENTARILY PI.iU. OIJT SELECTEWITCH HANDLETO START SAAELEPUMPS I. PLACE ANALYZER ISOlATION BYPASS (SYI.OCI( SWITCHES TO BYPASS POlE GRCIJP SOLAUGN EXISTS 2 IECT Ol CR SUPPRCNMBR AND I E,NTARLYPLU. our SELECT SWITCH RANDL.E TO START SAMPLE PUMPS H2 AND O2EIONITORINGSYSTEM NOTIFY CIEM LAB TO SAMPLE FOR IS INOPERABLE 1 AND SUPPR CHUBA +42 AND 02 PCR4-MOIIITORAND CONTROL 17WANOSUPPRC24MBA IA2ATEL01WZA%

AND

• O2ATORBEI.0W3.3%

1351140 TIE NO MM(EUP SYSTEM (APPX 14A)

L

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EQI APPENDIX-14A Rev. 7 Page 1 of 3 2-EOl APPENDIX-14A NITROGEN MAKEUP TO PRiMARY CONTAINMENT LOCATION: Unft 2 Control Room ATTACJENTS: None, (\/)

1. IF......... PCIS Group $ Isolation signal exists, THEN PERFORM Appendix BE concurrently with this procedure.

2. MONITOR Drywell and Suppression Chamber Hydrogen and Oxygen concentrations with H2/02 CONCENTRATION recorders 2-XR-76-1 1OA or 2-XR-76-1 lOB (Panel 9-54 or 9-55).

3. IF Drywell or Suppression Chamber Hydrogen or Oxygen analyzers are or become inoperable, THEN NOTIFY Chem Lab to sample Drywell or Suppression Chamber for Hydrogen or Oxygen using CI-644.

WHEN Drywell or Suppression Chamber Hydrogen or Oxygen concentration is increasing, THEN CONTINUE in this procedure.

NOTE: This procedure assumes that the normal makeup is from Nitrogen Storage Tank B.

5. DISPATCH personnel to N 2 Storage Tank B to CHECK 0-PI-76-331, N2 STORAGE TANK B PRESSURE, between 160 and 190 psig.

DETERMINE area with highest Hydrogen or Oxygen concentrations and INFORM SRO.

7 VERIFY PCIS RESET (Panel 9-4).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EOl APPENDIX-14A Rev. 7 Page 2 of 3

8. IF.............. It is desired to makeup Nitrogen to the Suppression Chamber, THEN ....... CONTINUE in this procedure at Step 10.

9. CONTROL Drywell Hydrogen or Oxygen as follows:

a. OPEN the following valves to admit Nitrogen to Drywell (Panel 9-3):

• 2-FCV-76-18, DRYWELL N2 MAKEUP INBD ISOLATION VLV

• 2-FCV-76-17, PRI CTMT N2 MAKEUP OUTBD ISOLATION VLV,

b. SLOWLY ADJUST 2-PC-76-14, 0W/SUPPR CHBR N2 MU PRESS CONTROL (Panel 9-3), to maintain between 55 and 60 scfm, or as directed by SPO if SAMG execution is in progress.

c. VERIFY 2-XR-76-14, DW/SUPPR CHBR N2 MAKEUP FLOW/PRESS (Panel 9-3), indicates below 60 scfm on the red pen, or as directed by SRO if SAMG execution is in progress.

d. CONTINUE Nitrogen admission to the Drywell UNTIL Drywell Hydrogen and Oxygen are below desired values.

e. CONTINUE in this procedure at Step Ii.

10. CONTROL Suppression Chamber Hydrogen or Oxygen as follows:

a. OPEN the following valves to admit Nitrogen to the Suppression Chamber (Panel 9-3):

• 2FCV7619r SUPPR CHBR N2 MAKEUP INBD ISOLATION VLV

• 2-FCV-76-17, PRI CTMT N2 MAKEUP OUTBD ISOLATION VLV.

b. SLOWLY ADJUST 2-PC-76-14, DW/SUPPR CHBR N2 MU PRESS CONTROL (Panel 9-3), to maintain between 55 and 60 scfm, or as directed by SRO if SAMG execution is in progress.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 203000 RHRILPCI: Injection Mode (Plant Specific)

Tier # 2 K3.04 (CFR 41.7)

Knowledge of the effect that a loss or malfunction of the RHRILPCI: Group # 1 INJECTION MODE (PLANT SPECIFIC) will have on following: K/A # 203000K3.04 Adequate core cooling Importance Rating 4.6 Proposed Question: # 28 An accident occurred on Unit 2 AND resulted in the following conditions:

• Reactor water level indicates () 200 inches on Post Accident Range

• Reactor pressure is 400 psig

• ALL RHR I LPCI are lost

• ONLY ONE CRD Pump AND ONE Core Spray pump are running Which ONE of the following completes the statement below?

Adequate core cooling

[REFERENCE PROVIDED]

A. does NOT exist B. is provided by Spray Cooling C. is provided by Steam Cooling

0. is provided by Core Submergence Proposed Answer: D Explanation A INCORRECT: is incorrect because adequate core cooling exists. The (Optional): candidate that fails to correct fuel zone level would believe that the core is no longer adequately cooled.

B INCORRECT: is incorrect because reactor pressure is too high for CS to inject. Plausible in that candidate may fail to recognize reactor pressure greater than the shutoff head (330 psig) of the CS pump.

C INCORRECT: is incorrect because the core is submerged with actual level above top of active fuel.

D CORRECT: The indicated parameter place corrected water level above TAF. With water level above TAF, adequate core cooling is assured by submergence.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the affect of Loss of RHR I LPCI on adequate core cooling.

Question Cognitive Level:

Question is rated as C/A because it involves the multi-part mental process of assembling, sorting, and use reference to solve a problem.

Technical Reference(s): OPL1 71.201 Rev. 7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: 2-Ll-3-52/62 Correction Curve Learning Objective: OPL171.201 V.B.10 (As available)

Question Source: Bank # CNP 08 #17 (Note changes or attach parent)

Question History: Last NRC Exam Cooper 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71 .201 Revision 7 Page 3 of 6 A. Key Words and Terms Obj. V.B.10

1. Section I-C to the Program Manual (see Attachment

1) provides definitions for terms, phrases, and acronyms used in the EOls. The following terms/phrases are to be highlighted in this lesson:

a. Adequate Core Cooling Obj. V.B.10.a Any of the following conditions (1-4):

(1) Submergence: Reactor water level is verified at or above TAF, and based on present and past trends and plant conditions, is expected to remain above TAF.

(2) Spray Cooling: During the execution of Cl, the following conditions are met:

• The reactor can be determined to be shutdown without boron (note 1)

AND

• One Core Spray subsystem is One spray ring for injecting at or above 6250 gpm. design pattern AND

• RPV water level can be determined to be above -215 inches (2/3 core height)

(3) Steam Cooling With Injection:

• During execution of C5 and Cl, This will maintain RPV water level can be PCT < 1500 °F maintained above the lower water level band allowed by the procedure, [Minimum Steam Cooling Water Level (MSCWL).

180 inches].

OR

• Reactor pressure can be maintained above MARFP following reactor depressurization.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71 2O1 Revision 7 Appendix D Page 4 of 6 2-L!-3-52 & 62 CORRECTION CURVES

-I Ui w

-J 0

LU C

REALTOR PRESSURE (PSIG) PP.584 REV 15 TP-28 Figure 4 Ll-3-52 &-62 CORRECTION CURVES

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet INJECTION SOURCES Capacity Shutoff Head System Pumps (gpm) Motive Force (psig)

HPCI 5000 1 1240 Steam (150-1150 psig)

RCIC 600 1 1240 Steam (150-1150 psig)

CRD 2 98 each 1640 Motor Feedwater 3 11,200 each 1210 Steam Condensate 10,800 each Booster 410 Offsite Power (300 psig)

Core Spray 6250 per loop 2 loops 330 Motor (105 RHR (LPCI) 100O0 each 320 Motor 0 psig)

Condensate 10,830 each 3 130 Offsite Power (103 psig)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION: NRC RO 17 An accident occurred and resulted in the following conditions:

• Reactor water level is -2lkdicated FZ steady.

• ReactorpreureisOOg,stable) ,Fnaneakt

• Only one (I) Control Rod Osive Hydraulic Pump and one Cs pump are running.

• LPCI and Cs initiation signals are present.

What, if anything. ensures Adequate Core Cooling at this tinie!

a a, does not exist.

5. is provided by spray cooling.

c. is provided by core submergence. -

d. is provided by steam updraft through the core.

ANSWER: NRC RO 17

c. is provided by core submergence. Nw Rcn.in Explanation:

The indicated parameter place corrected water level at TAF. With wOtCr level at TAF adequate core cooling is assured.

Distractors:

L is incorrect because adequate core cooling exists. The candidate that fails to correct fuel zone level would believe that the core is no louger adequately cooled.

is incorrect because reactor pressure is to high for CS to inject the candidate that fails to recognize reactor pressure greater than the shutoff head of the CS pump.

d. is incorrect because the core is submerged with actual level at 5 inches above top of active fuel.

Provide EOP graph 14 Famiatted Fain (Defsi1t) Thais 36

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 205000 Shutdown Cooling Tier # 2 G2.2.22 (10CFR55.41.5)

Knowledge of limiting conditions for operations and safety limits. Group # 1 K/A # 205000G2.2.22 Importance Rating 4.0 Proposed Question: # 29 Unit 1 is in Mode 4 with RHR Pump lB in Shutdown Cooling.

Which ONE of the following completes the statements below?

In accordance with Tech Spec 3.5.2, ECCS - Shutdown, RHR Pump I B (1) Operable for the ECCS function.

The MAXIMUM allowed RCS cooldown rate per Tech Spec 3.4.9, RCS Pressure and Temperature (PIT) Limits, is _(2)_ in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

A. (1)is (2) 90° F B. (1)is (2) 100° F C. (1)is NOT (2) 900 F D. (I)is NOT (2) 100° F Proposed Answer: B Explanation A INCORRECT: Part 1 correct See explanation B. Part 2 incorrect (Optional):

See explanation C.

B CORRECT: Part 1 correct Per Tech Spec 3.5.2, A LPCI subsystem may be aligned for decay heat removal and considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local) to the LPCI mode. Part 2 correct per Tech Spec 3.4.9, RCS cooldown shall be 100° Fin any one hour C INCORRECT: Part 1 incorrect Plausible in that ECCS systems are normally required to start, align and inject in response to a system initiation signal to be considered operable. The provision to allow manual realignment is an exception for the conditions. Part 2 incorrect Plausible in that this is the administrative RCS Cooldown limit.

D INCORRECT: Part 1 incorrect See explanation A. Part 2 correct See explanation D.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of limiting conditions for operations associated with Shutdown Cooling.

Question Cognitive Level:

Question is rated as C/A because it involves the multi-part mental process of assembling, sorting, and using knowledge and its meaning to solve a problem.

Technical Reference(s): Ui TS 3.4-21 Amm 234 (Attach if not previously provided)

Ui TS 3.4-24 Amm 234/ 3.4-26 Amm 256 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank # -

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ECCS - Shutdown B 3.5.2 BASES (continued)

LCO Two low pressure ECCS injection/spray subsystems are required to be OPERABLE, The low pressure ECCS injection/spray subsystems include CS subsystems and LPCI subsystems. Each CS subsystem consists of one motor driven pump. piping. and valves to transfer water from the suppression pool to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping. and valves to transfer water from the suppression pool to the RPV.

The necessary portions of the Emergency Equipment Cooling Water System are also required to provide adequate cooling to each required ECCS subsystem.

An LPCI subsystem may be aligned for decay heat removal and considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local> to the LPCI mode and is not otherwise inoperable. Because of low pressure and low temperature conditions in MODES 4 and 5. sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery.

APPLICABILITY OPERABILITY of the low pressure ECCS injection/spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1. ECCS subsystems are not required to be OPERABLE during Fv1ODE 5 with the spent fuel storage pool gates removed and the water level maintained at 22 ft above the RPV flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown.

(continued)

BFN-UNIT 1 B 3.5-24 Revision 9--46 March 14, 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RCS PIT Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (PIT) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits.

APPLICABILITY: At all times.

ACTI ONS CONDITION REQUIRED ACTION COMPLETION TI ME A. NOTE A.i Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

he completed if this t::ondition is entered.

A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Requirements of the LCO acceptable for continued not met in MODE 1. 2, or operation.

3.

B. Required Action and B.i Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time or Condition A not NQ met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

BEN-UNIT 1 3.4-24 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RCS P/T Limits 3.4.9 SURVEILLANCE REQU IREMENTS SURVEILLANCE FREQIJENOY SR 3.41 NOTES

1. Only required to be performed during RCS heatup and cooldown operations or RCS inservice leak and hydrostatic testing when the vessel pressure is > 312 psig.

2. The limits oi Figure 3.4.9-2 may be applied during nonnuclear heatup and ambient loss cooldown associated with inservice leak and hydrostatic testing provided that the heatup and cooldown rates are i5E/hour.

3. The limits of Figures 3.4.9-1 and 3.4.9-2 do not apply when the tension from the reactor head flange bolting studs is removed.

Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits sI:iecified by Curves No. 1 and No. 2 of Figures 3.4.9-1 and 34.9-2:

and

b. RCS heatup and coo[down rates are I00F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 34.9.2 Verify ROS pressure and RCS temperature Once within 15 are within the criticality limits specified in minutes prior o Figure 3.4.91, Curve No. 3. control rod withdrawal for the purpose of achieving cilticality (continued)

SEN-UNIT *1 3.4-26 Amendment No. 2A, 256 July 26, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RHR-High Water Level 3.9.7

3.9 REPUELING OPERATIONS 3.9.7 Residual Heat Removal (RHR) - High Water Level LCO 3.9.7 One RHR shutdown cooling subsystem shall he OPERABLE and in operation.

NOTE The required RHR shutdown cooling subsystem may not be in operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 22 ft above the top of the RP\/ flange, ACTIONS CONDITION REQUIRED ACTION COMPLETION TI ME A. Required RHR shutdown A. I Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooling subsystem method of decay heat inoperable, removal is available. AiL Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

BPN-IJNIT 1 3.9-14 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBLITY SUPPORT OPL 171.044 Revision 17 Page 71 of 146 INSTRUCTOR NOTES

i. Core Spray System (1) Combines with at least two RHR pumps to meet ECCS cooling requirements on design basis LOCA.

(2) Shares divisional sepaiated electrical power supplies.

(3) Load shedding interlocks and time delays prevent overloading power supplies.

(4) The Keep-fill System from Core Spray keeps the LPCI injection path full from the pump discharge check valve to the inboard LPCI injection valve.

j. Automatic Depressurization System (ADS) Ohj. yB, IT Obj. V.E.l0 (1) Receives an input from RHR pump discharge pressure switches for an initiation permissive.

(2) Provides RPV depressurization on a small break LOCA to allow LPCI injection.

k. High Pressure Coolant Injection System (HPCI)

Provides small break depressurization makeup if LPCI is not needed.

F. Modes of Operation Shutdown Cooling all manual operation TP8

a. Normal cooldown from rated conditions Procedural Bypass steam to main condenser until SDC Rx compliance will pressure interlock is met (105 psig) then line up pieei RHR pump in the SDC mode. exceeding limitations.

Maximum cooldown rate 90°F/hr 100°F/hr Tech Spec Ix Refer to Ol-T4and EOI Appendix 17D for requirements for placing Shutdown Cooling in service.

2. Containment CoolingSpray TP-4, 5, and 6

a. When core is reflooded after a LOCA, flow can he Ohj. V.B.3 diverted to the containment spray header.

h. Control of valves is gained by placing a selector switch in SELECT position.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBLITY SUPPORT ECCS Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 RE QU REM ENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that.

with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI wi[l cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low-low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool.

The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience with these components supports performance of the Surveillance at the 24 month Frequency, which is based on the refueling cycle. Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RP\

is not required during the Surveillance.

(continued)

BFN-UNIT 1 B 3.5-18 Revision -43 January 17. 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 206000 High Pressure Coolant Injection System Tier # 2 A3.05 (CFR: 41.7)

Ability to monitor automatic operations of the HIGH Group # 1 PRESSURE COOLANT INJECTION SYSTEM including: K/A # 206000A3.05

• Reactor water level: Importance Rating 43*

Proposed Question: # 30 Unit 1 was operating at 100% Reactor Power when a LOCA occurred which resulted in the following conditions:

• RPV water level lowered to (-) 50 inches and is currently (+) 55 inches and slowly lowering.

Which ONE of the following is the FIRST condition that would cause an AUTOMATIC restart of HPCI?

A. Level lowers to (+) 27 inches.

B. Level lowers to (+) 2 inches.

C. Level lowers to (-) 45 inches.

D. Drywell Pressure greater than 2.45 psig.

Proposed Answer: C Explanation A INCORRECT: With level at (+) 27 inches, the level 8 (+) 51 inches signal (Optional): will be clear. However, the Level 8 Turbine Trip will still be sealed in, unless manually reset. The candidate may select this if he/she doesnt realize the turbine trip relay seals itself in, and needs to be manually reset. Also (+) 27 is a recognizable value in that it is set point for Reactor Level Low alarm.

B INCORRECT: HPCI does NOT initiate on a Level 3 signal, (+) 2 inches.

HPCI will NOT restart, if reactor water level lowers to this value, because of the sealed in Level 8 Turbine Trip. Level 3 is below Level 8 and the candidate may select this as a safe value. PCIS isolations and other events happen at Level 3. HPCI could be restarted with this condition, if the Level 8 reset pushbutton was depressed on the control room panel.

C CORRECT: HPCI will initiate on a Level 2 signal, (-) 45 inches, even though the Level 8 trip, (+) 51 inches, has NOT been manually reset. The Level 2 signal opens contacts that de-energize the Level 8 trip relay, which enables the HPCI Turbine to auto restart.

D INCORRECT: A drywell pressure of 2.45 psig is a normal HPCI initiation signal, and the signal seals in. However, the HPCI Turbine Trip is sealed in and will NOT reset on this initiation signal. Since this is an initiation signal, the candidate may think the HPCI Turbine will automatically restart.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A is matched because question is on the HPCI system and monitoring autom atic operation, based on water level conditions. The question asks what water level condition will allow HPCI to auto restart, based on the conditions of the stem.

Question Cognitive Level:

The candidate must know several facts: HPCI initiates on Level 2 level and on High Drywell Pressure (+) 2.45 psig. The stem also states level is

(-) 45 inches reactor water the candidate must determine that water level is above level 8 (+) 55 inches and

(+) 51 inches . The candidate must also know that the HPCI level 8 Turbine Trip Logic seals in and does not autom atically reset. Operator action is required to manually reset it, unless Level 2 is reached.

The sealed in Level 8 HPCI Turbine Trip will NOT allow the sealed in HPCI Initiation Signal Hi Drywe ll press to restart the system unless the Trip is manually reset or level again lowers to Level

2. Level 2 contacts will open and de-energize the L8 Turbine Trip relay, which will facilitate an automatic restart. To solve the problem posed by the question, the candidate must use a multi-p art mental process to assemble, sort, and integrate parts of the HPCI and HPCI Logic systems.

Technical Reference(s): OPL1 71.042 Rev 20 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.3.c (As available)

Question Source: Bank # Fermi 2 Modified Bank# (Note changes or attach parent)

New Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorou s review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: References attached.

m Cl) 0 I (t4- 1.hfl) 154t. UitZ Cl) 0)

H 3

0-c

-U C CD C,

-I 0_

-U z CD 0

Om H -Cx C

1 C, U. Mechani1 Ov÷rzpee 3.

Cl, H 0

-C D

-C r CLOSE 35 PSJ61 0 OPEN q PSIG CD C,

u>u0 U 13 CD (tQ). -

1 J 0

(.JS.. _S -C

>< 3 o M° AU: LPJt1P Oo m I4PCIAIJX PEFt WPJE (0 Cl) tilL PUMP 0

2:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .042 Revision 20 Appendix C Page 63 of 69 C) 0)

0

-J 0

cJ C) C:) *0 C)

0) Ii 0

c x >

0) 0<

r Cl) _J 0 0 ID U.)

ID Li.)

c.J TP-8: HPCI Initiation Logic

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 209001 Low Pressure Core Spray System Tier # 2 K1.07 (10 CFR 55.41.2 to 41.9)

Knowledge of the physical connections and/or cause effect Group # 1 relationships between LOW PRESSURE CORE SPRAY SYSTEM K/A # 209001 Ki .07 and the following:

• D.C. electrical power Importance Rating 2.5 Proposed Question: # 31 Unit 2 was operating at 100% Reactor Power, when a plant event resulted in a reactor scram AND loss of 250 VDC RMOV BD 2A. Degrading plant conditions have resulted in the following:

• Reactor Pressure is 325 psig and stable

• A few minutes later, Drywell Pressure is 2.8 psig Based on the above conditions, which ONE of the following predicts how Core Spray will be affected by the bus loss?

A. ALL Core Spray pumps will start AND ALL injection valves will open.

B. ONLY the Loop I Core Spray pumps will start AND Loop I injection valves will open.

C. ONLY the Loop 2 Core Spray pumps will start AND Loop 2 injection valves will open.

D. NO Core Spray pumps will start AND NO injection valves will open.

Proposed Answer: B Explanation A INCORRECT: Loop 2 pumps will not start and injection valves will not open.

(Optional): Candidate misconception that logic failure causes valves to fail open and pump start will NOT be affected.

B CORRECT: Loop 1 pumps will start and injection valves will open. SYS I Initiation Logic is still energized.

C INCORRECT: Loop 2 pumps will not start and injection valves will not open.

Candidate misconception that logic failure causes valves to fail open and pump start will NOT be affected. Candidate misconception that 250 VDC RMOV BD 2A is a division 1 feed and affects Loop 1 pumps and valves.

D INCORRECT: Loop 1 pumps will start and injection valves will open.

Candidate misconception that there is only one logic system for both loops of Core Spray so both would be affected and Loop 2 logic would be for UNIT 2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A requires cause effect relationship between Core Spray System and DC power. Question is about Core Spray system and the loss of DC power to one portion of its initiation logic and its effect on the system.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. The candidate must know the power supply to the Core Spray loop 2 logic and the effects of its loss. He/she must understand the system and logic interrelationships.

Technical Reference(s): OPL1 71.045 Rev 15 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: OPL171.045 Obi 4.d (As available)

Question Source: Bank #

Modified Bank # VY 2007 NRC Q6 (Note changes or attach parent)

New Question History: Vermont Yankee Last NRC Exam 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet VY 2007 NRC Exam A plant event has resulted in a reactor scram and loss of Bus DC-2C. Degrading containment conditions has resulted in the following:

• Reactor Pressure is at 325 psig

• Drywell Pressure is at 2.8 psig Based on the above conditions, how will Core Spray be affected by the bus loss?

A. All Core Spray pumps will start and All injection valves will open.

B. ONLY the Loop A Core Spray pump will start and its injection valves will open.

C. ONLY the Loop B Core Spray pump will start and its injection valves will open.

D. No Core Spray pumps will start and NO injection valves will open.

Proposed Answer: C Explanation (Optional):

A. Incorrect Loop A pumps and valves have lost initiation logic power.

B. Incorrect Bus DC-2C provides 125 VDC power to A loop pump and valve initiation Logic. Only B loop would have power.

C. Correct Only Loop 8 pumps and valves have initiation logic power.

D. Incorrect Loop B pumps and valves still have initiation logic power.

I,,

Co 0

14AF 8 P$-S-74[fl s--

on Low Close cc levi ReactrPras [

RoncAor S 251W DC DiV II iZSOyDC RMQVSd Lwa P78 MB C,)

w OW PRESS 3 Dv RHR LOGIC Snarod wtth CS) Sharad with OS) Ref. Lirawincu U, (SnareS with CS) -4 GEI3OESSG -I Co z3 5.

251W DC 0 DiV II a>ao

0) *0 0 tO *0) <

° 2 s Fm L..

H1c-frIDW PRESS v i 1 LOWR

. PACrCPLFVtL_ UNIT 1,23O!GSTAPT LO?ø SHED LOGIC CORE SPRAY I -

PUMP INITIAl TO r . cC 1<118 C

C inbity cEcil 71 StartS DOs CAS. PAS, Pr&ered 0

• 1 3

m C,)

h 0

m C,)

1X 51 3D C 416OSDBDD 1

-4 C) a m

C,, Cl) 3 CD z

-i

-1 0 Om z U)

I a

0 C) i--- t\O 0 K9s C,)

m m

H 0

-n C)

. I (DCD

-3: -

D NVA CD TD U- NITArWPREED CQF* SPRAY TD SYSTEM READY pUMPTRrANDaLocK PUMP 2D COROLS CD 1 frhro 0 7-f-St2D 3 m

U) 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 211000 Standby Liquid Control System Tier # 2 A2.07 (10CFR 55.41.5)

Ability to (a) predict the impacts of the following on the STANDBY Group # 1 LIQUID CONTROL SYSTEM ; and (b) based on those predictions, K/A # 211 000A2.07 use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

• Valve closures Importance Rating 2.9 Proposed Question: # 32 Unit I is executing I -EOl-1, RPV Control, due to a Scram AND an ATWS. The Unit Operator (UO) is directed to inject Standby Liquid Control (SLC) per 1-EOl-1 Appendix 3A, SLC Injection.

The UO places the SLC Pump control switch in the START-A position.

Given the following plant conditions:

• SLC SQUIB VALVE CONTINUITY LOST, (1-9-5B, Window 20) Extinguished

• SQUIB VALVE A and B CONTINUITY, blue lights on Panel 1-9-5 Illuminated

• SLC Pump IA red light Illuminated Which ONE of the following describes the status of SLC AND the correct action(s) to take?

A. ONE squib valve has fired; Place SLC Pump IA in Stop, start the SLC Pump I B, AND verify proper operation.

B. NO squib valves have fired; Place SLC Pump IA in Stop, start the SLC Pump IB, AND verify proper operation.

C. ONE squib valve has fired; Verify proper system operation by observing the SLC tank level lowering by -1 % per minute.

D. BOTH squib valves have fired; Verify proper system operation by observing the SLC tank level lowering by --1 % per minute.

Proposed Answer: B Explanation A INCORRECT: A pump did start by indication of RED light illuminated.

(Optional): Neither squib valve has fired; as indicated by the lack of the alarm and the blue lights are still lit. The squib valves are arranged in Parallel, so 1 firing would allow injection into RPV. Starting B would allow the squib valves to be fired from the other primer.

B CORRECT: A pump did start by indication of RED light illuminated.

Neither squib valve has fired as indicated by the lack of the alarm and the blue lights are still lit. Starting B would allow the squib valves to be fired from the other primer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet C INCORRECT: A pump did start by indication of RED light illuminated. It is not required in the EQ ls to dispatch personnel to the area. Starting B would allow the squib valves to be fired from the other primer.

D INCORRECT: A pump did start by indication of RED light illuminated.

Neither squib valve has fired as indicated by the lack of the alarm and the blue lights are still lit. No flow so tank level will not decrease.

KA Justification:

The KA is met because the question tests the ability to predict the impact of valve closures on the SLC System. Based on the indications provided, candidate must conclude that following system initiation both Squib Valves remain closed and recognize the impact on SLC Injection.

Based on the Squib Valves failing to open, the candidate must use 1-EOI-1 Appendix 3A to correct the consequences of this abnormal condition.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. Candidate must diagnose the system condition based on indications provided and then determine appropriate action to take to correct the abnormal condition.

Technical Reference(s): I -EOl Appendix 3A rev 0 (Attach if not previously provided)

OPL1 71 .039 rev 16 (Including version / revision number)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.4 I V.B.5 (As available)

Question Source: BFN 1#33 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Browns Ferry 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The A SLC pump has started and neither squib valve has fired as indicated by the lack of the alarm and the blue lights are still lit. The proper action iaw EOl-app 3A is to start the other pump and verify proper operation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I

BFN 1-EOI APPENDIX-3A UNITI SLC INJECTION Rev. 0 Page 1 of 2

{

LOCATION: Unit I Control Room ATTACHMENTS: None (____

UNLOCK and PLACE 1-HS-63-6A, SLC PUMP lA/i B, control switch in START-A or START-B position.

2. CHECK SLC System for injection by observing the fo[Iovring:

• Selected pump starts, as indicated by red light illuminated above pump control switch.

• Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,

• SLC SQUIB VALVE CONTINUITY LOST 1-EA-63-3 Annunciator in alarm on Panel 1-9-5 (1-XA-55-58, Window 20).

• 1-Pl-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

• System flow, as indicated by I -IL-63-1 1, SLC FLOW, red light illuminated on Panel 1-9-5,

• SLC INJECTION FLOW TO REACTOR 1-FA-63-11, Annunciator in alarm on Panel 1-9-5 (i-XA-55-5B, Window 14).

3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.

4. VERIFY RWCU isolation by observing the following:

• RWCU Pumps lAand lB tripped

• 1-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE closed

• i-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed

• 1-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71 .039 r16

1. Explosive Valves a) Two 100% capacity explosive (Squib) valves, FCV 63-8A and B, are installed in parallel.

b) Provide a zero leakage seal between the boron solution and the reactor c) Each valve contains two firing primers, powered by the 250V DC contro l power from the 480V Shutdown Boards A and B, (unit specific).

d) Either primer is capable of actuating the valve.

e) The primer is fired by taking the main control room handswitch, HS-63

-6A, to the START PUMP A or START PUMP B position. This forces the ram outward, which shears the end cap off the valve fitting, allowing flow to pass through the valve.

f) After firing, the ram remains extended. This prevents the sheared cap from obstructing flow through the valve.

g) The primer requires a minimum current of 2 amps to fire, and fires within 2

milliseconds after this circuit is applied. All the explosion by-products are retained in the trigger explosive chamber.

h) Each valves firing circuit continuity is monitored by a blue indicating light on Panel 9-5 and a current meter located in the back of Panel 9-5.

1. Main Control Room Instrumentation (Panel 9-5)

Parameter Device Range Normal Indication SLC Storage Tank Level Indicator 0 100%

61 - 69%

Level SLC Pump Disch. Pressure Indicator 0 2000 psig 0 psig with system Pressure in standby, -1250 psig with pump running HCV-63-12 Position Red Light ON when valve is ON open HCV-63-13 Position Green Light ON when valve is ON closed HCV-63-14 Position Green Light ON when valve is ON closed Squib valve firing circuit Blue Light ON when squib ON continuity (One for each squib) valve firing power is available and circuit continuity is maintained

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet System Flow Red Light ON when sensed OFF for normal flow downstream of system standby squibs is >40 gpm SLC Pump Red Light ON when pump is OFF for normal (One for each pump) running system standby SLC Pump Green Light ON when pump is ON for normal (One for each pump) stopped system standby

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 212000 Reactor Protection System Tier # 2 A4.03 (10 CFR 55.41.7)

Ability to manually operate and/or monitor in the control room: Group # 1

• Provide manual select rod insertion K/A # 212000A4.03 Importance Rating 3.9 posed Question: # 33 Unit 2 was operating at 100% Reactor Power, when the plant experienced a complete loss of the Control Air system. The following plant conditions exist:

• ALL eight Scram Solenoid Group A/B Logic Reset Lights are NOT lit

• Recirc Pumps are Tripped

• Reactor Power is 20%

You are the OATC and have been directed to perform 2-EQ I Appendix 1 D, Insert Control Rods Using Reactor Manual Control System (RMCS).

Based on the above conditions which ONE of the following responses contains the correct steps to manually insert AND determine when the control rods are inserted?

Verify CRD Pump operating, (1) , direct manually opening CRD Flow Control Valve (2-FCV-85-1 1A or B), verify Mode Switch in SHUTDOWN, bypass the Rod Worth Minimizer, CRD Power Switch ON, select control rod, AND place CRD (2)

A. (1) reset ART (2) Control Switch in ROD IN, until green 00 is lit, on the four rod display B. (1)resetARl (2) Notch Override Switch in EMERG IN, until the control rod stops moving inward C. (1) direct closure of CHARGING WATER SHUTOFF, 2-SHV-85.-586 (2) Control Switch in ROD IN, until the green 00 is lit, on the four rod display D. (1) direct closure of CHARGING WATER SHUTOFF, 2-SHV-85-586 (2) Notch Override Switch in EMERG IN, until the control rod stops moving inward Proposed Answer: D Explanation A INCORRECT: A loss of Control Air occurred, so scram and ARI cannot be (Optional): reset. Also CRD Notch Override Switch is placed in Emergency In, in an ATWS. Procedure directs insert until movement stops. Candidate misconception that scram and ARI can be reset with NO Control Air available. Also misconception that CRD Control Switch is used in an ATWS when driving rods. NOT used due to RMCS settle function requirements between rods, would delay rod insertion in this emergency.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B IN CORRECT: A loss of Control Air occurred, so ARI and scram cannot be reset. Part 2 is correct; the procedure directs insert until movement stops and use of Notch Override Switch in EMERGENCY IN until rod stops moving. Candidate misconception that scram and ARI can be reset with NO Control Air available.

C INCORRECT: A loss of Control Air occurred. Part 1 is correct because cannot reset scram or ARI. Candidate misconception that CRD Control Switch is used in an ATWS when driving rods. NOT used due to RMCS settle function requirements between rods, would delay rod insertion in this emergency. CRD Notch Override Switch is placed in Emergency In to insert the control rod, in an ATWS D CORRECT: A loss of Control Air occurred. Scram and AR! cannot be reset because no air pressure. Charging water shutoff valve needs to be closed to direct water from Charging header to Drive Water Header to move rods.

The CRD Flow Control Valve has lost air and needs to be manually opened to provide Drive Water Pressure to drive control rods. Emergency In is used to bypass the settle function on the Reactor Manual Control Sys, so the control rods can be inserted without waiting between rod selections, therefore taking less time to insert in the ATWS emergency.

KA Justification:

The K/A is matched because the question and K/A require how to manually select, insert, and determine (monitor) when the control rods are inserted.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. The candidate must deduce that an ATWS has occurred.

He/she must determine that the loss of Control Air caused the scram and Recirc Pump Trip. The loss of Control Air will not allow reset of the scram or ARI. It complicates control rod movement because of loss of air to the CRD Flow Control Valve. Because of the ATWS, control rod movement will be with the ROD Notch Override Switch instead of the CRD Control Switch.

Technical Reference(s): 2-EOI Appendix 1 D Rev 6 (Attach if not previously provided) 2-AOl-32-2 Rev 32 Proposed references to be provided to applicants during examination: None Learning Objective: V.B.9 (As available)

Question Source: Bank #

difjed Bnk  : (Note changes or attach parent)

New X Question History: Last NR Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Corn me nts:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EOI APPENDIX-iD INSERT CONTROL RODS USING REACTOR MANUAL CONTROL LOCATION: Unit 2 Control Room, Panel S AITACEL4ENTS: . Tools and Er tment

2. Core Position Nan NDTE: Thi £01 Ac-cendix may be ax :ted concurrently dth ED: Aopendix Ji or 13 at 3RD s discretiri when time and manpower sermit.

I. VERIFY at least one CR0 o:mc in service.

NDTE: hosing 235386, CHARCNG WATER 1500, vave may seduce. the effectiveness of ECI Apoendix IA or 13.

2 IF .....Reactor Scram or ARI CANNOT he reset, THEN .. . DISPATCH personnel to close 2-SHV83386,

is WATER SHUCO;FF (RB NE, El 385 fo)

3. VERIFY REACTOR NODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

REFER To Attachment 2 and. INSERT oontr: rods In area of highest tower as follows:

a. SELECT control rod.

PLAOE CR0 NOTCH DVERR13E switch in EMBED ROD IN oosition UNT13: control rod is NOT moving inward.

o. REPEAT Stis 5. a and 3 .b for each control rod to be.

inserted.

NOTE: A lldder may be recuireC to nerform the following step. REFER TO Tools and Ecp:ipiuent, Attachment I.

IF necessary, an alternate ladder is available at the ECU Modes, EAST and West banks, It is stored by the TED Charging Dart.

6. WHEN . . . NC forther control rod movement is tcssible or desired, THEN . . DISPATCH persorine. to verify onen 2SHV3558 5, CHARGNG NATE?. SHUTOFF (RB NE, El 355 ft)

END DR TEXT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT (Appendix normally performed for ATWS. Will not be effective due to loss of Control Air)

I I

BFN 1-Ed 1 APPENDIX-IF UNITI MANUAL SCRAM Rev. I I PageIof7 LOCATION: Unit 1 Control Room ATTACHMENTS: 1. Tools and Equipment

2. 1-PNLA-009-0015, Rear

3. 1-PNLA-009-0017, Rear I____

1 VERIFY Reactor Scram and ARI reset.

a. IF ARI CANNOT be reset, THEN EXECUTE EOI Appendix 2 concurrently with Step 1.b of this procedure.

b. IF Reactor Scram CANNOT be reset.

THEN DISPATCH personnel to Unit 1 Auxiliary Instrument Room to defeat ALL RPS logic trips as follows:

1) REFER to Attachment 1 and OBTAIN four 3-ft banana jack jumpers from ECI Equipment Storage Box.

2) REFER to Attachment 2 and JUMPER the following relay terminals in 1-PNLA-009-001 5. Rear:

a) Relay 5A-K1OA (DQ) Terminal 2 to Test Terminal 1 -TX-099-05A-K1 2E (Bay 1).

b) Relay 5A-K1OC (AT) Terminal 2 to Test Terminal 1-TX-099-05A-lKl2G (Bay 3).

3) REFER to Attachment 3 and JUMPER the folioing relay terminals in 1PNLA-O09-0017, Rear:

a) Relay 5A-K1 08 (DQ) Terminal 2 to Test Terminal 1-TX-099-05A-K12F (Bay 1).

b) Relay 5A-KIOD (AT) Terminal 2 to Test Terminal 1 -TX-099-05A-K1 2H (Bay 3).

2. WHEN RPS Logic has been defeated, THEN RESET Reactor Scram.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 215003 Intermediate Range Monitor (IRM) System Tier# 2 K2.01 (10 CFR 55.41.7)

Knowledge of electrical power supplies to the Group# 1 following: K/A# 21 5003K2.01

. IRM channels/detectors Importance Rating 2.5 Proposed Question: # 34 Unit 2 is performing a startup with the following conditions:

• Mode Switch is in STARTUP

• Reactor is critical

• IRMs are steady on Range 2 Which ONE of the following identifies the IRM power source AND the effect of a loss of power to a single IRM?

IRM Power Source Effect of Power Loss to IRM A. 24 VDC Battery Rod Block ONLY B. 24VDC Battery Rod Block AND Half Scram C. 250 VDC Battery Rod Block ONLY D. 250 VDC Battery Rod Block AND Half Scram Proposed Answer: B Explanation A INCORRECT: An INOP half scram is also processed, as well as a rod (Optional): block. Candidate misconception that scram function bypassed on range 2.

B CORRECT: 24 VDC supplies IRM detector voltage. With a loss of power, the detector will indicate downscale and receive an INOP trip. The INOP trip enforces both a rod block and a half scram on the corresponding RPS channel.

C INCORRECT: 24 VDC supplies IRM detector voltage. An INOP half scram is also processed. Candidate misconception that 250 VDC supplies IRMs. It does supply the neutron monitoring battery chargers.

D INCORRECT: 24 VDC supplies IRM detector voltage. Candidate misconception that 250 VDC supplies IRMs. It does supply the neutron monitoring battery chargers.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A is matched because the question asks for power supply to the IRM5 and affect of loss of the power supply. K/A asks for knowledge of electrical power supply to the lRMs channels/detectors.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL171.020 Rev 11 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.11 (As available)

Question Source: Bank #

Modified Bank# Nine Mile 2 !9? (Note changes or attach parent)

New Question History: Last NRC Exam Nine Mile 2 / 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Nine Mile 2 NRC 2008 Tier#2 Group # 1 K/A# 215003, K2.01 Importance Rating 2.5 (K&A Statement) Knowledge of electrical power supplies to the following: IRM channels/detectors Proposed Question: Common 23 The plant is performing a startup with the following conditions:

• Mode Switch in STARTUP

• Reactor critical

• IRMs steady on Range 2 Which one of the following will result from the failure of the 24 VDC Power Supply Fuses to a single IRM?

Rod Block Half Scram A. IRM INOP None B. IRM DOWNSCALE None C.IRMINOP IRMINOP D. IRM DOWNSCALE IRM DOWNSCALE Proposed Answer: C.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.020 Revision 11 Page 18 of 44 C. Power Supplies Obj.V.B.1 1

1. The IRM power supplies receive unregulated +24 VDC power from the neutron monitoring battery and convert it to regulated voltages of proper magnitude for use by the IRM detectors and logic circuits. A loss of 24VDC power will give an mop trip, additionally there will be a loss of IRM indication.

2. Neutron monitoring battery chargers are fed from its units 250V Battery Board, Panel 8, which in turn is fed from l&C A and B regulating transformers.

3. Detector Drives are from l&C A power supply. A loss of this power supply would result in an inability to move IRMs.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71,020 Revision 11 Page 20 of 44 INSTRUCTOR NOTES E. Trips TP-10

1. Rod blocks Obj.V.D.7. V.8.5 Obj. V.0.3.,

Block Setpoint When Bypassed Downscale <7.5 Range I or RUN Obj. \J.B.6.

Qbj V.04 Obj. \/.8.5 High > 90/104.6 RUN Mode Unit Difference IRM high setpoint is INOP -HV low (<90v RUN Mode 90 at Unit 2 and 104.6

-Module unplugged on Unit 1 and Unit 3

-Function switch not in OPERATE

-Loss of +24VD0 Detector Wrong Detector RUN Mode Obj.V.B. 13 Position Not Full [N Scrams TP-11 Scra rns Setpoint When Bypassed Obj. V.B.7.

Ohj. V.0.5. Oh.V.D.8 High-High > 116.4 In RUN Mode INOP -HV low (<90v) In RUN Mode

-Module unplugged

-Function switch not in OPERATE

-Loss of +24VDC F. Controls Provided

1. Panel9-5

a. Recorder switches select between IRM channels. and APRM/RBM channels have been removed. All units now contain digital recorders.

which do not require operation of selector switches. These switches have been removed.

b. Range switches allow operator to select appropriate IRM range to maintain indications between 25 to 75 on 0-125 scale. 0-40 scale is no longer utilized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBLITY SUPPORT OPL171 .020 Revision 11 Page 18 of 44 INSTRUCTOR NOTES C. Po..ver Supplies Obj.\/.B.1 1 1 The IRM power supplies receive unregulated +/-24 VDC power from the neutron monitoring battery and convert it to regulated voltages of proper magnitude for use by the IRM detectors and logic circuits. A loss of 24VDC power will give an mop trip.

additionally there will be a loss of IRM indication.

2 Neutron monitoring battery chargers are fed from its

+/- 24 VDC Neutron units 250V Battery Board, Panel 8. which in turn is fed Monitor Battery from l&C A and B regulating transformers. powers cabinets and detectors.

3. Detector Drives are from l&C A power supply. A loss of this power supply would result in an inability to move IRMs.

D. Instrumentation SER 03-05 Controlling Plant

1. Control Room Instrumentation Evolutions Precisely Item Monitor all available Device Range indications during Reactor Power Dual Recorders (4) 0 to 125 Located 9-5 reactivity changes Reactor Power 8 meters. dual scale 0 to 40 not used, located 9-12 0 to125 located 9-12

2. Annunciators, alarm indication Obj.\!.B.5/ V.5.7

a. Annunciators Annunciator Function/Remarks lRf1 Downscale Rod Block (Range 1-bypassed) lRt,4 High Rod block IRM High-High Scram (RPS Channel A/B) Obj.\/.D.3 or INOP b.. Alarms other than annunciators on Panel 9-5 Each IRM channel has a set of 4 lights above (1) Hi-Hi/mop (red) the IRM range sviitches (2) High (amber)

(3) Downscale (White)

(4) Bypassed (White)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 215004 Source Range Monitor (SRM) System 2!

Tier#

K5.01 (10 CFR 55.41.5)

Knowledge of the operational implications of the following concepts Group# 1 as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: KIA# 21 5004K5.01

• Detector operation Importance Rating 2.6 ------

Proposed Question: # 35 Which ONE of the following completes the statement below?

The applied voltage to the SRM detector is _(1)_ than the applied voltage used for the IRM detector AND the SRM electrode generates an electrical signal _(2)_ proportional to neutron flux in the core.

A. (1)lower (2) directly B. (1) higher (2) directly C. (1)lower (2) inversely D. (1) higher (2) inversely Proposed Answer: B Explanation A INCORRECT: SRM voltage is higher. Candidate misconception that SRM (Optional): detectors detect lower power therefore the voltage detector power requirement is lower.

B CORRECT: The SRM (IRM) detector is a fission chamber that has an applied voltage to the electrode of approximately 350 (100) volts. The operating chamber is pressurize with Argon to about 213 (17) psia. They generate an electrical signal proportional to the neutron flux level in the core.

C INCORRECT: SRM voltage is higher and the signal is not inversely proportional. Candidate misconception that SRM detectors detect lower power therefore the voltage detector power requirement is lower. Candidate misconception that Campbeling correction (square root effect) is used by the SRM, and this makes the signal inversely proportional.

D INCORRECT: the signal is not inversely proportional. Candidate misconception that Campbeling correction (square root effect) is used by the SRM, and this makes the signal inversely proportional.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A is met by question asking knowledge of the SRM detector operation.

RO knowledge Task.

Memory knowledge because RO must recall facts about SRM detector operation.

Question Cognitive Level:

The question tests for the total recall of discrete facts or bits of information, for a single system.

Technical Reference(s): OPL171.019 Rev 13 (Attach if not previously provided)

OPL171.020 Rev 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.D.2 (As available)

Question Source: Bank # Brunswick 07 #12 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Brunswick 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ES-401 Sample Written ENamination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KA# 215004 K5O1 Importance Rating Z6 ioweige of th operioiI imp1Datns of tho fooing cccps as cy ppIy o SOURCE RANGE M0NT0R

&M SYSTEM De1cor opfo Proposed Question:

Common 12 The Source Range Monitor (SRM) detectors are fission chambers that have an applied voltage to an electrode. The applied voltage to the SRM detector is A. higher than the applied voltage used for the IRM detector and the SRM electrode generates an electrical signal inversely proportional to neutron flux in the core.

B. lower than the applied voftage used for the iRM detector and the SRM electrode generates an electrical signal inversely proportional to neutron flux in the core.

C. higher than the applied voltage used for the IRM detector and the SRM electrode generates an electrical signal proportional to neutron flux in the core.

0. lower than the applied voltage used for the RM detector and the SRM electro de generates an electrical signal proportional to neutron flux in the core.

Proposed Answer: C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet A. Incorrect the signal is not inversely proportionaL B Incorrect SRM voltage is higher and the signal is not inversely proportional.

0, Incorrect SRM voltage is higher Technical Reference(s): _SD_09.1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.4i 55,43

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CPL1J 4.019 Revision 13 Page 10 of 51 Instructor Notes (4) For a core of <20,000 MwcI/T exposure, spontaneous fission of Cm -242. is the pilmary intrinsic neutron source.

(5) For a core of> 20,000 Mwd!T exposure, spontaneous fission of Cm -244 is the priniary intrinsic neutron source.

Detection Chamber

a. The purpose of the detection chamber is to cjenerate an electrical signal proportional to the neutron flux level in the core.

(3) Ionization chamber Obj. V.3.3 Obj. V.D.2 (a) The inner electrode of the ionization chamber is supplied with 350 VDC by a high voltage power supply.

OPL171 .020 Revision 11 Page 9 of 44 (4) Operating voltage is bOy DC (350V DC for SRM)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 215004 Source Range Monitor (SRM) System Tier# 2 K5.03 (10 CFR 55.41.5)

Knowledge of the operational implications of the Group # 1 following concepts as they apply to SOURCE RANGE KIA# 21 5004K5.03 MONITOR (SRM) SYSTEM:

Changing detector position Importance Rating 2.8 Proposed Question: # 36 A plant start up on Unit 3 is in progress. A control rod block has occurred. The following nuclear instrument indications are noted:

SRMA SRMB SRMC SRMD Position j__Full in Mid-position Mid-position I Full in IRMA IRMB IRMC IRMD IRME IRMF IRMG IRMH 25/125 15/125 35/125 55/125 75/125 75/125 30/125 25/125 Range 3 Range 2 Range 3 Range 3 Range 2 Range 2 Range 3 Range 3 Which ONE of the following identifies the MINIMUM action needed to clear the ROD WITHDRAWAL BLOCK?

A. Insert SRM B ONLY B. InsertSRM BAND SRMC C. Range up on IRM B AND IRM F to range 3 D. RangeuponlRME AND IRMFto range3 Proposed Answer: A Explanation A CORRECT: SRM RETRACT NOT PERMITTED will alarm and cause a rod (Optional): block with SRM counts <l45cps with associated IRMs Range 2 and the Detector not Full In.

B INCORRECT: Plausible in that with SRM C Not Full in and associated IRM E not on range 3, candidate may believe that it must also be inserted to clear the Rod Block. However, although SRM C is not full in, it is above the Rod Block set point of 145 cps so the Rod Block is bypassed.

C INCORRECT: Plausible in that it would clear the Control Rod Block from SRM B. However, it would result in IRM B causing a rod block due to IRM downscale.

D INCORRECT: Plausible in that ranging up IRM E and F would not result in an IRM downscale rod block. However, a rod block would remain with IRM B still on range 2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A is matched because in the question operational conditions/implications have arisen from the mis-positioning of the SRM detectors. The candidate must determine which detector is causing the conditions and based on his/her knowledge resolve the situation. Knowledge involves recognizing the interaction between the SRM/IRM systems, including consequences and implications.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 3-01-92 Rev. 14 (Attach if not previously provided)

OPL17I.019 Rev 13 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.019 V.B.8 (As available)

Question Source:

Modified Bank # BEN 1006 #37 (Note changes or attach parent)

Question History:

Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo /ess rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 21 5004 Source Range Monitor Tier # 2 G2.2.2 (10CFR 55.41.7)

Ability to manipulate the console controls as required to operate the Group # 1 facility between shutdown and designated power levels. K/A # 215004G2.2.2 Importance Rating 4.6 Proposed Question: # 37 A plant start up on Unit 3 is in progress. A control rod block has occurred. The following nuclear instrument indications are noted:

SRMA SRMB SRMC SRMD Position Full in Mid-position Mid-position Full in Counts (CPS) 3 9.5x10 95 80 3 8.0x10 IRMA IRMB IRMC IRMD IRME IRMF IRMG IRMH 25/125 15/125 35/125 55/125 75/125 75/125 30/125 25/125 Range 3 Range 2 Range 3 Range 3 Range 2 Range 2 Range 3 Range 3 Which ONE of the following identifies the MINIMUM action needed to clear the ROD WITHDRAWAL BLOCK?

A. Insert SRM B ONLY B. Insert SRM B AND SRM C C. Range up on IRM B AND IRM F to range 3

• D. RangeuponiRME AND IRMFt0 range3 Proposed Answer: B Explanation A INCORRECT: Plausible in that with IRM C on range 3, candidate may (Optional): believe SRM C Detector Not Full In Rod Block is bypassed. However, with any associated IRM (A, C, E or G) not on range 3, the trip remains in force.

B CORRECT: SRM RETRACT NOT PERMITTED will alarm and cause a rod block with SRM counts <l45cps with associated IRMs Range 2 and the Detector not Full In.

C INCORRECT: Plausible in that it would clear the Control Rod Block from SRM B. However, it would result in IRM B causing a rod block due to IRM downscale.

D INCORRECT: Plausible in that ranging up IRM E and F would not result in an IRM downscale rod block. However, a rod block would remain with IRM B still on range 2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Justification: Candidate must demonstrate ability to manipulate the console controls for SRMs as required to operate the facility between shutdown and designated power levels.

Technical Reference(s): 3-01-92 Rev. 13 (Attach if not previously provided)

OPL171.019 Rev. 13 Proposed references to be provided to applicants during examination:

Learning Objective: OPL1 71 .019 V.B.8 (As available)

Question Source:

Mod ified Bank # Perry 09 #37 (Note changes or attach parent)

Question History:

Last NRC Exam Perry 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 )(

55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Source Range Monitors 3-01-92 Unlt3 Rev.0014 Page l4of 14 Illustration I (Page 1 of 1)

SRM Trip Outputs TRIP SIGNAL SETPOINT ACTION SRM High 68 X i0 4 counts per second Rod block unless IRMs on Range 8 (or higher) or REACTOR MODE SWITCH in RUN SRM Inop A. Module unplugged Rod block unless 1RMs on Range B. Mode switch not In operate C. HV power supply low voltage D. Loss of +/-24 vdc SRM Downscale 5 counts per second Rod block unless IRMs on range 3 (or higher) or REACTOR MODE SWITCH in RUN II SRM Detector 145 counts per second Rod block unless detector full-in, Wrong Position lRMs on range 3 (or higher), or REACTOR MODE SWITCH in RUN SRM High-High = 2 x iO counts per second Scram if shorting links removed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7tO19 Revision 13 Page 22 of 51 Instructor Notes

b. Alarms. Interlocks, Trips and Annunciators ObJ.V.B8 ObjN.C.2N.D.5 Annunciator/Function Setpoint Bypassed SRM Hi(Alarm and Rod Block) 6.8 X io IRM range 8 or above (Panel 9-5A, Windowl3) OR in Run Mode INOP(Alami and Rod Block) IRM range 8 or above, (Panel 9-5A, Window 13) OR in Run Mode (1) module unplugged; Obj. V.B.5 (2) switch not in Operate Obj. VCI (3) HV Power supply voltage Low Obj. V.D.4 (4) Loss of 24 VDC power supply Loss of power gives Rod Block SRM DOWNSCALE <5cps IRM range 3 or in (Alarm and Rod Block) RUN Mode (Panel 9-5A, Window 6)

SRM SHORT PERIOD 30 seconds Never (Alarm only) (Panel 9-5A, Window 20)

h. SRM RETRACT NOT <l4Scps IRM range 3 OR in ObJVB7 PERMITTED RUN Mode OR F (Alarm and Rod Block) Detector Full-in.

c. Alarms other than annunciators on ObJVB.8 panel 9-5 Obj.VC2 (1) Each SRM channel has a set of four lights on the apron section:

(a) Hi Hi (red) Refer to 01-92 for (b) High/INOP (amber) current setpoints cps) 5 (2XlO (c) Downscale (white)

(d) Bypassed (white)

(2) Each SRM channel has a white Obj. V.B.6 RETRACT PERMISSIVE light above Set points in 01-its respective LCR meter. 92 minimum cps energizes the light.)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17I,019 Revision 13 Appendix C Page 50 of 51 Reactor Reactor Mode Switch Mode Switch in STARTUP in REFUEL L.+/-

Pod Bridga (x lO ops) 4 lM B 8 IM 0>81 1 INOP (SOO,MU,HVL)

IRM F 8 IRM H 8 IRM B IRM 0 3 I Downscale cpa)

IRMF3 Ii Full In CP retract permit IRM H >3 I

Trip Unit SRM lN0P Control (1) SOO mode switch out ol operate Rod Block (2) MU muduh unplujged (3) HVL high voltage low 2J30E321 (Partial)

TP-10: SRM ROD BLOCK DIAGRAM

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BEN Intermediate Range Monitors I 3-OI-92A Unit 3 IRev 0015

. Pagel5ofl5 Illustration I

<Page 1 ofl)

IRM Trip Outputs TRIP SIGNAL SETPOINT ACTION IRM High >104.6 ON 125 SCALE Rod block unless REACTOR MODE SWITCH in RUN IRM mop A. Module unplugged Rod block unless REACTOR B. Mode switch not in operate MODE SWITCH in RUN C. HVpower supply low Reactor Scram unless REACTOR MODE SWITCH in RUN D. Loss of +1-24 vctc IRM Downscale <7.5 on 125 SCALE Rod block unless IRMs on range I unless REACTOR MODE SWITCH in RUN IRM Detector Wrong detector not full in Rod block unless detector full-in, Position or REACTOR MODE SWITCH in RUN IRM High-High >1 16.4 ON 125 SCALE Reactor Scram unless REACTOR MODE SWITCH in RUN

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 215005 APRM I LPRM Tier # 2 A3.08 (1 OCFR 55.41.7)

Ability to monitor automatic operations of the AVERAGE POWER Group # 1 RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM K/A # 215005A3.08 -

including:

• Control rod block status Importance Rating 3.7 Proposed Question: # 37 Unit 2 APRM Channel 3 has a total of 18 LPRM inputs.

Which ONE of the following statements identifies the expected response to this condition?

A. The A* **** will produce a Rod Block signal ONLY.

B. NO Rod Block OR Reactor Scram signals are generated.

C. The APRM will produce a Rod Block signal AND a Scram signal input to EACH 2/4 logic voter module.

D. The APRM will produce a Rod Block signal AND a Scram signal input to ITS RESPECTIVE 2/4 logic voter module ONLY.

Proposed Answer: A Explanation A CORRECT: If the number of un-bypassed LPRM inputs exceeds the (Optional): minimum number required in the APRM average (<20 total or <3 per level),

an APRM INOP condition is applied. This results in a Rod Block only -

manual trip must be inserted for inoperable condition.

B INCORRECT: Plausibility based on misconception that since no Reactor Scram signal is generated with this mop condition, likewise, no Control Rod Block is generated. Also plausible that the candidate may believe the minimum number of LPRM inputs is still available and conditions are not met for Rod Block or Scram Signal.

C INCORRECT: Plausible in that < 20 LPRM inputs to an APRM results in INOP Condition. ALL other APRM Inop signals do result in an APRM Trip.

This would be the correct answer for any other APRM Inop Signal.

D INCORRECT: Plausibility based on the misconception that a Scram Signal would result with <20 LPRMs input into the APRM and that the resultant scram signal would input only into associated logic voter module.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests ability to monitor automatic operations of the Average Power Range Monitoring System including Control rod block status and scram signal input to voter logic given less than the required 20 LPRM inputs into an APRM.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL17I.148 Rev 12 (Attach if not previously provided) 2-Ol-92B Rev. 38 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPLI71.148 V.B.7/31 (As available)

Question Source: OPL1711 #58 (Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .148 Revision 12 Page 18 of 106 INSTRUCTOR NOTES (4) When the LPRM signal is 3%.

b. The LPRM signals can be manually bypassed from the APRM average flux Password entry is calculation. This operation can only required performed on panel 9-14.

c. An LPRM may be manually bypassed with Obj. V.C. I c either the high voltage applied (BYP/HV Indication ON) (Still have LPRM indication) OR with preserved the high voltage off (BYPIHV OFF). (No LPRM indication avail)

No Indication d, A bypassed LPRM will not be included in the APRM average and will not indicate downscale or upscale conditions.

e. For the APRM channel, the total number of Obj. V.B.7.b.(I)

LPRM inputs that may be bypassed is 23 Obj. V.C:I.b.(i) before reaching an INOP condition. More later I. (1) If the number of un-bypassed LPRM inputs exceeds the minimum number Rod Block only-required in the APRM average (<20 manual trip must be total or <3 per level), an APRM INOP inserted for condition is applied. inoperable condition.

f. Operability for all aspects of the PRNM system needs to be assessed when bypassing LPRMs.

9. Keylock Mode Switch

• One keylock switch per LPRMIAPRM instrument

• Has two positions aOPER and INOP.

a The key is removable in either position

10. LPRM alarms (Panel 9-5) Obj. V.8.5 a LPRM Upscale and LPRM Downscale Total Scale = 0 to 125%

(1) The upscale and downscale set point markers are displayed inside the bargraphs and a status indication is Downscale is less displayed above the bargraphs. The than or equal to solid box above the bargraph 3% of scale AND indicates that the set point marker is upscale is greater presently exceeded while a hollow than or equal to box indicates a past condition. 100% of scale

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Average Power Range Monitoring 2-Ot-92B Unit2 Rev. 0038 Page 22 of 30 Illustration I (Page 1 of 5)

APRM Trip Outputs APRM Trip Outputs TRIP SIGNAl.. SETPOINT ACTION APRM Downscale 5% t Rod Blocl if REACTOR MODE SWITCH in RUN.

• h.. APRM mop I APRM Chassis Made not in OPERATE 1. One Channel detected, no alarm or RPS (keylock to INOP). output signal.

2. Loss of Input Power to APRM. 2. Two Channels detected, RPS output signal to all four Voters (Full Reactor Scram).

3. Self Test detected Critical Fault in the APRM instrument.

4. Firmware Watchdog timer has timed out APRM mop Condition 1, <20 LPRMS in OPERATE, or < 3 per 1. <21] LPRMs total or <3 per level results in a P. level. Rod Block and a trouble alarm on the display panel. This does not eld an automatic APRM trip, but does, however, make the associated APRM NOR APRM High 1. OLO 1. Rod Block ii REACTOR MODE SWITCH in (0.66W ÷ 59%) RUN.

SLO (0.66W(W-10%) ÷ 59%)

[W = Total Recirc Drive Flow in %

ratedi.

2. Neutron Flux Clamp Rod Block 2, Rod Block in all REACTOR MODE 113% SWITCH positions except RUN.

3. 10%APRMFIux.

APRM High High 1. OLO 1. Scram.

(0.66W + 65%)

2. SLO (D.66(W10%) + 65%)

[W = Total Recirc Drive Flow in %

rated].

2. 119% APRM Flux. 2. Scram in all REACTOR MODE SWITCH positions except RUN.

3. 14% APRM Flux.

APRM Flow 1. 5% mismatch between APRM 1. Flow compare inverse video alarm.

Converter Channels.

2. 107% Flow monitor upscale. 2. Rod Block.

OPRM lnop <23 Operable Cells - Annunciation Only A cell is mop when it has < 2 operable LPRMs OPRM Fre-Tiip Any one of three algorithms, period, growth, Rod Block Condition or amplitude exceeds its pre-trip alarm setpoint for an operable OPRM cell.

OPRM Trip Any one of the three algorithms, period, 1. One Channel detected, no RPS output growth, or amplitude for an operable OPRM signal.

cell has exceeded its trip value: 2. Two Channels detected, RPS output signal to all four Voters_(Full Reactor Scram).

All OPRM setpoints are bypassed when the Reactor Mode Switch is not in RUN or the Reactor is not operating in the PowerlFlow region where instabilities can occur (25% Power

& <60% Recirc Drive Flow).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 217000 Reactor Core Isolation Cooling System (RCIC)

Tier # 2 K2.02 (10 CFR 55.41.7)

Knowledge of electrical power supplies to the Group # 1 following: K/A # 217000K2.02

• RCIC initiation signals (logic)

Importance Rating 2.8 Proposed Question: # 38 Which ONE of the following identifies the RCIC initiation logic power supply?

A. 250 VDC RMOV BD A B. 250 VDC RMOV BD B C. Div I ECCS ATU inverter D. Div 2 ECCS ATU inverter Proposed Answer: B Explanation A INCORRECT: This supplies the Channel/Bus B Isolation Logic. Easily (Optional): confused by candidates.

B CORRECT: This supplies the Initiation Logic and Channel/Bus A Isolation Logic.

C INCORRECT: This supplies 125 VAC to the RCIC Flow controller and various RCIC indicators. HPCI and RCIC system components and power supplies are easily confused by the examinees.

D INCORRECT: This supplies 125 VAC to the HPCI Flow controller and various HPCI indicators. HPCI and RCIC system components and power supplies are easily confused by the examinees.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A asks for knowledge of electrical power supply to RCIC initiation logic. Question is designed to ask directly for the RCIC initiation logic power supply.

Question Cognitive Level:

Question requires recall of discrete information and is therefore a memory or low cognitive question.

Technical Reference(s): OPL1 71.040 Rev 23 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.7 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .040 Revision 23 Page 34 of 74

8. Failure Modes

a. Loss of Power to the Flow Controller Obj. V.C.4 (1) Div I ECCS ATU Inverter h.

r (2) Loss of Power causes the controller to go to zero Reference P&L milliamp output and turbine speed would lower to 3.23 I 3.0.W minimum (-6O0 rpm). However, on Units 1&3, the Div 1 ECCS Inverter also powers to EGM Control Box which would result in overspeed on Units 1&

3 only.

b. Loss of control air Obj.V.B.7 Obj. V.C.4 (1) RCIC steam line steam trap bypass valve (FCV 71-5) fails closed (Unit 3)

(2) RCIC steam line condensate drain valves (FCV 71-GA and 6B) fail closed (3) RCIC condensate pump Clean Radwaste discharge valves (FCV-71-7A and 7B) fail closed

c. Loss of electrical power to valves Obj. V.B.7.

Obj. V.C.4.

All motor-operated isolation valves remain in the last position upon failure of valve power. Solenoid operated valve FCV 71-5 (Unit 2) fails closed.

d. Loss of Power to Relay Logic Obj. V.B.7.

Obj. V.CA (1) If Bus A fails, the automatic initiation circuit and UNIT turbine trip solenoid will not operate. Channel A DIFFERENCE isolation logic circuit is lost. Power is lost to EG-M control box and this causes FCV-71-10 trip governor valve to go wide open (if RCIC is operating). Unit 2 (Unit 3 EGM power is from DIV 1 Inverter)

(2) if power is lost to the EGM Control Box, Springs UNIT will re-position the 71-10 servo to fully open the DIFFERENCE governor valve (Unit 2 only).

m Cl) 0

-5 I 1OAKI9A Rctnr K?9I3 1<3 Lwlevel T° From HS7152.

I 1tJAKOA I*IR 1WK8OB Reset to EGM Lo9tcT Control Ebx 1 (Unit 2 Only)

-1 Opens 712,

[ Opens 7137 I to Manual RCIC 1<2 K2 Starts G.S 1<37? Isalation tD P

I

<?139 Clcses?1-3 Vacuum Pump Relay () XS7152 Cl) 0 RCIC INITIATION LOGIC 3 0

CD z Cl) 1 CD 0

(flu) 0.

-4 CD 0

0 1:1> yQ (1U

. r mm ox 1

0

-I 3

m ci) 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO 21 7000 Reactor Core Isolation Cooling System (RCIC)

Tier # 2 K2 04 (10CFR 5541 7)

Knowledge of electrical power supplies to the following: Group # 1

• Gland seal compressor (vacuum pump) KJA # 21 7000K2.04 Importance Rating 2.6 Proposed Question: # 39 Which ONE of the following completes the statement?

The power supply to the Unit 2 RCIC Vacuum Pump is A: 250 VDC RMOV BD 2A B. 250 VDC RMOV BD 2C C. 480 VAC RMOVBD2A D. 480 VAC RMOV BD 2B Proposed Answer: B Explanation A INCORRECT: This is, in fact a power supply to RCIC components; just not (Optional): the RCIC Vacuum Pump. Refer to attached PRESTARTUP REQUIREMENTS.

B CORRECT: 250 VDC RMOV BD 2C is the power supply to the RCIC Vacuum Pump. See Attached Electrical Lineup Checklist.

C INCORRECT: This is, in fact a power supply to RCIC components; just not the RCIC Vacuum Pump. Refer to attached PRESTARTUP REQUIREMENTS.

D INCORRECT: This is, in fact a power supply to RCIC components; just not the RCIC Vacuum Pump. Refer to attached PRESTARTUP REQUIREMENTS.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests candidate knowledge of power supplies to RCIC Vacuum Pump. Level of difficulty is compounded by the similarities of HPCI and RCIC in conjunction with the complex electrical distribution system at BFN. HPCI is a Div II System with B Logic as the primary logic; but it comes from an A Board. RCIC is the opposite A Logic from a B Board. This Often creates confusion between the power supplies for the two systems.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): 2-01-71, Rev. 61 / 2-01-71 Att. 3 Rev. 58 (Attach if not previously provided)

OPL1 71 .040 Rev. 23 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source:

(Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Reactor Core Isolation Cooling 2-01-71 UnIt 2 Rev. 0061 Page 12 of 73 4O PRESTARTUP!STANDBY READINESS REQUIREMENTS NOTE When Section 4.0 is required to be verified by subsequent Sections, Section 4.0 will be performed.

(1] VERIFY the following related system requirements are satisfied A. Oil level Is visible in RCIC turbine pedestal sight glass.

B. The following panels are energized.

(REFER TO 0-01-578, 0-Ol-57C, and 0-Ol-57D)

• 25OVDC Reactor MOV Board 2A C

• 25OVDC Reactor MOV Board 2C C

• 240V Lighting Board 2A C

• 480V Reactor MOV Board 2A C

• 480V Reactor MOV Board 28 C

• Panel 2-9-9, Cabinet 2 C

• Panel 2-9-9, Cabinet 3 C

• Panel 2-9-9, Cabinet 4 C

• Panel 2-9-9, CabInet 5 C

• 1 E ECCS ATU lnverter (Division I) C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1JI .040 Revision 23 Page 13 of 74 A

ii. Noncondensables are renioved by a C 25OVDC DC-powered vacuum pump RMOV Bd discharging to the suppression pool.

if the vacuum is excessive, a valve controlled by condenser pressure, in the vacuum pump discharge line, will automatically open and release noncondensables back to the condenser. The vacuum pump automatically starts on system initiation.

(c) During operation, liquid from the spray C 250 VDC and condensed steam is collected in a RMOV Bd receiver section of the barometric condenser and pumped by a DC powered condensate pump back to the suction of the RCIC pump.

i. Pump cycles on high and low level signals from the barometric condenser.

ii. Pump is rated at 3 hp.

- (d) During periods of system non-use, the barometric condenser is continually drained to Clean Radwaste through two air-operated valves in the condensate pump discharge line. The valves operate off level in the condenser (FCV-71 -7A and 7B). These valves automatically close when 71-8 is not fully closed.

(e) High pressure in the barometric condenser alarms at approximately 8 Hg (Only if normally shut steam supply valve FCV-71 -8) is not fully closed. (15 sec.

TD))

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Attachment 3 2-01-71 IATT-3 Unit 0 Reactor Core Isolation Cooling Rev. 0058 Electrical Lineup Checklist Page 7 of 7 4.0 ATTACHMENT DATA (continued)

PanellBreaker Initials Number Component Description Required Position lstllV Reactor Bldg. 250V RMOV Bd 2C El 565 8B 2-BKR-071-0017 ON t RCIC SUPPR POOL INBD SUCT

VALVE BREAKER (GE-13-41) 8D 2-BKR-O7i-0025 ON

. RCIC LUBE OIL COOLING WATER I VALVE BREAKER (GE-13-132) bE 2-BKR-071-0031 ON

RCIC TURB BAROMETRIC CNDRVACPUMPBREAKER Electric Board Room 2B 250V RMOV Bd 2B El 59Y 8E1 2BKR-071 -002B18E1 ON RCIC SYS LOGIC DIV 1-2 PNL 2-25-31 5D 2-BKR-071-0034 ON RCIC PUMP MIN FLOW VALVE BREAKER (GE-i 3-27) 5B 2-BKR-071-0003 ON RCIC STMLINE OUTBD ISOL VALVEBREAKER(GE-13-16)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 218000 ADS Tier# 2 G2.1.7 (10CFR 55.41.5)

Ability to evaluate plant performance and make operational Group# 1 judgments based on operating characteristics, reactor behavior, and KIA# 218000G2.1.7 instrument interpretation.

Importance Rating 4.4 Proposed Question: # 40 Unit 2 was operating at 100% Reactor Power with RHR Pump 2D tagged out of service. A Loss of Coolant Accident with a subsequent Loss of Off Site Power has resulted in the following plant conditions:

• Reactor Water Level is (-)1 25 inches

• Drywell Pressure is 4.1 psig

• A AND C 4KV Shutdown Boards are de-energized Which ONE of the following identifies the MINIMUM action(s), if any, that will prevent the Automatic Depressurization System (ADS) from an Auto-Initiation?

A. NO action is required B. Place ONLY ADS Logic Inhibit Switch A to INHIBIT C. Place ONLY ADS Logic Inhibit Switch B to INHIBIT D. Place BOTH ADS Logic Inhibit Switches A AND B to INHIBIT Proposed Answer: D Explanation A INCORRECT: RHR Pump C running meets Pump running permissive for (Optional): System 1 and 2 ADS logic. Any one of the four RHR pumps or either A or B and either C or D Core Spray pumps running is required. RHR C Pump is running and NO Core Spray Pumps are running.

B INCORRECT: Plausible in that different combinations of ECCS Pumps operating meet the pump running permissive for different ADS logic channels.

C INCORRECT: Plausible in that different combinations of ECCS Pumps operating meet the pump running permissive for different ADS logic channels.

D CORRECT: RHR Pump C running meets Pump running permissive for System 1 and 2 ADS logic. Any one of the four RHR pumps or either A or B and either C or D Core Spray pumps running is required. RHR C Pump is running and NO Core Spray Pumps are running

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests candidates ability to evaluate plant performance and make operational judgments for the ADS System based on operating characteristics, reactor behavior, and instrument interpretation including Reactor Level, Drywell Pressure and Electrical Distribution indications. Based on those indications, candidate must make operational judgment regarding the status of ADS logic.

Question Cognitive Level:

This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71 .043 Rev 13 (Attach if not previously provided) 2-Cl-i Rev. 47 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL17I.043 VB4 (As available)

Question Source:

Modified Bank # BFN 1006 #40 (Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.043 Revision 13 Page 12 of 30 INSTRUCTOR NOTES

d. EOI Appendix 8G crosslies CAD to DWCA PROCEDURE USE

& ADHERENCE

4. ADS systems controls TP-2

a. Consists of pressure and water level sensors arranged in the trip systems that control a solenoid-operated pilot air valve

b. The solenoid-operated valve controls the pneumatic DCN 51106 pressure applied to a diaphragm actuator which Cable & Switch controls the SRV directly configuration I modifications

c. Cables from sensors lead to the Control Room where logic arrangements are formed in cabinets

d. Control channels are separated to limit the effects of electrical failures

e. A two-position control switch is provided in the Control Room for control of the ADS valves l) Two positions are OPEN and AUTO HP Use SELF-CHECKING

2) In OPEN, the switch energizes a DC solenoid which allows pneumatic pressure to be applied Pressure relief to the diaphragm actuator of the relief valve consists of actuation of NOTE: reactor pressure The relief valves can be manually opened to provide a controlled on internal pilot or nuclear system cooldown under conditions where the normal heat sink by electro is not available pneumatic operation via

3) In AUTO, the valves are controlled by the ADS pressure switches.

logic and pressure relief logic f, Four of the six ADS valves may also be controlled from UNIT a backup control board which is provided to facilitate DIFFERENCE, plant shutdown and cooldown from outside the Control DCN 51106 adds Room new panel 25-658 to Unit I

{5. Automatic Depressurization Initiation Logic

a. The following conditions must be met before automatic Obj V.B.4 depressurization will occur Obj. V.C.3

1) Two coincident signals of high drywell pressure Obj. V.D.3

(+2.45 psig) and low low low reactor vessel Obj yEA

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71 .043 Revision 13 Page 13 of 30 INSTRUCTOR NOTES water level (-122)

OR

-122 for 265 sec. LT-3-58A-D

2) A confirmatory low reactor vessel water level LT-3-1 84 4 signal (+2) (Tech Spec Value 0) LT-3-1 85

3) Any one of the four RHR pumps or either A or B Obj, V.C.4 and either C or D Core Spray pumps running Obj. V.D.4 NOTE:

This signal comes from pressure switches on the discharge of the pumps which give permissives in the logic above a set pressure of 100 psig for RHR pumps and 185 psig for the Core Spray pumps.

RHR CS Associated PS-74-8A and 8B PS-75-7 shutdown boards (Pump A) (Pump A) must be energized PS-74-31A and 31B PS-75-35 for the respective (Pump B) (Pump B) pumps.

PS-74-19A and 19B PS-75-16 (Pump C) (Pump C)

PS-74-42A and 42B PS-75-44 (Pump D) (Pump D)

4) A 95-second timer must be timed out

b. The high drywell pressure signal seals in immediately Obj. V.C.4 upon receipt of the signal Obj. V.D.4

1) Must be manually reset after the signal has PS-64-57A-D cleared

2) Indicative of a breach in the process system HP Procedure Use barrier inside the drywell and Adherence

c. The reactor vessel low water level signals (-122 and Obj. V.B,4

÷2) indicate that fuel is in danger of becoming .Obj. V.C.3 overheated Obj. V.D.3 Obj. V.E.4

1) The -.122 water level signal would not normally K28, 29,&30 occur unless the HPCI System had failed Obj. V.C,4

2) These signals do not seal Obj. V.D.4

3) The -122 water level initiation setpoint is selected to open the SRVs and depressurize the reactor vessel in time to allow fuel cooling by the Core Spray and [PCI Systems following TP-3 a LOCA, in the event that the other makeup Obj. V.0.4 systems (Feedwater, CRD Hydraulic. RCIC, Obj. V.D.4

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Main Steam System 2-01-1 Unit 2 Rev. 0047 Page 12 of 64 3,4 Main Steam Relief Valve (MSRV I ADS)

A. Whenever both the acoustic monitor and the temperature indication on a relief valve fall to indicate In the Control Room, the Technical Specifications Section 3.3.3.1 should be consulted to determine what limiting conditions for operation apply.

B. In the event that a relief valve fails to function as designed and the cause of the malfunction is not clearly determined and then corrected, the valve should be considered inoperable and TechnIcal Specifications Section 3.5.1 and 3.4.3 should be consulted to determine what limiting conditions for operation apply.

C. ADS will initiate when ALL of the following conditions are met:

1. A confirmatory Low reactor water level signals (+2.0 inches), REACTOR LEVEL LOW ADS SLOWDOWN PERMISSIVE, 2-9-3C Window 3 2, Two coincident signals for each of the following parameters:

a, high drywell pressure (+2.45 psig) in conjunction with low low low reactor water level (-122 Inches), ADS SLOWDOWN HIGH DRYWELL PRESS SEAL-IN, 2-XA-55-9-3C Window 33 and RX WTR LVL LOW LOW LOW ECCSIESF INIT 2-LA-3-58A, 2-XA-55-9-3C Window 28 OR

b. low low low reactor water level (-122 inches), RX WTR LVL LOW LOW LOW ECCSIESF INIT 2-LA-3-58A, 2-XA-55-9-3C Window 28, for 265 seconds (High cirywell pressure bypass)

3. One RHR pump OR two Core Spray pumps (A or B and C or D) running, RHR OR CS PUMPS RUNNING ADS SLOWDOWN PERMISSIVE, 2-XA-55-9-SC Window 10.

4. When ALL of the above logic is satisfied, then a 95 second timer starts (ADS BLOWDOWN TIMERS INITIATED, 2-XA-55-9-3C, Window 11) and the timer must be timed out to initiate ADS blowdown.

D. Depressing 2-XS-1-169 and -161 on Panel 2-9-3 will reset the ADS Slowdown Timers. They also reset an ADS initiation, if the timers have timed out. ADS will re-initiate upon subsequent timing out of the timer provided the low level and pump logic signals still exist. The timer setpoint is 95 seconds, however setpoint tolerance allows it to be as low as 77 seconds.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 Proposed Question: # 40 Unit 2 was operating at 100% Reactor Power with RHR Pump 2B tagged. A Loss of Coolant Accident with a subsequent Loss of Off Site Power has resulted in the following plant conditions:

• Reactor Water Level is (-)1 25 inches

• Drywell Pressure is 4.1 psig

• B AND D 4KV Shutdown Boards are de-energized

• RHR Pump 2A tripped Which ONE of the following identifies the MINIMUM action, if any, that will prevent the Automatic Depressurization System (ADS) from an Auto-Initiation?

A. NO action is required B. Place ONLY ADS Logic Inhibit Switch A to INHIBIT C. Place ONLY ADS Logic Inhibit Switch B to INHIBIT D. Place BOTH ADS Logic Inhibit Switches A AND B to INHIBIT Proposed Answer:

Explanation INCORRECT: Pump running permissive is not met with only Core Spray A

(Optional): Pumps A and B running. It is the same permissive for System 1 and 2 ADS logic. Any one of the four RHR pumps or either A or B and either C or D Core Spray pumps running is required. No RHR Pumps are running and only A and B Core Spray Pumps are running.

B INCORRECT: Plausible in that channel A logic is made up. However channel C logic is not so there is no requirement to Inhibit System 1 logic.

c INCORRECT: Plausible in that channel B logic is made up. However channel D logic is not so there is no requirement to Inhibit System 1 logic.

D INCORRECT: Plausible in that if the right combination of Core Spray Pumps were running on any RHR Pump running, this would be correct answer.

Justification: To correctly answer this question, candidate must recognize condition not met for automatic initiation of ADS to determine no action is required to prevent inadvertent initiation of ADS logic. This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off Tier# 2 K4.02 (10 CFR 55.41.7)

Knowledge of PRIMARY CONTAINMENT ISOLATION Group# 1 SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) KIA# 223002K4.02 and/or interlocks which provide for the following:

Testability Importance Rating 2.7 Proposed Question: # 41 Which ONE of the following explains the response of the isolation logic for Reactor Water Cleanup Suction Isolation Valves?

A trip of BOTH division 1 (A, C) low level sensor relay(s) within a logic trip channel will cause a (1)_ isolation AND _(2)_ closure.

A. (1)half (2) NO valve B. (1)half (2) inboard valve C. (1)full (2) inboard valve D. (1)full (2) inboard AND outboard valve Proposed Answer: A Explanation A CORRECT: Typical PCIS logic is designed so each valve has 2 trip (Optional): channels, each containing 4 level sensor relays two from division 1 (A and C contacts in series) and two from division 2 (B and D contacts in series) with both sets of contacts in parallel. The trip of one or both division 1 low level sensor relays in a single channel will cause a half isolation on the lnbd and Obrd valves and no valve closure. The isolation is said to be half-cocked. A trip of one or both low level sensor relays in each division will cause a full isolation and valve closure. (lnbd and Obrd valves)

B INCORRECT: Half is correct, but no valve closure will occur. It would take a trip of a sensor relay in the other low level sensor division to affect closure.

C INCORRECT: Full is incorrect. There would be no valve closure for the conditions given. Misconception by candidate that a trip of any two sensor relays would cause valve closure.

D INCORRECT: Full is incorrect. Neither valve would move under the given conditions. Misconception of logic operation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The K/A is matched because the stem asks for knowledge of how a trip channel is tested and how an isolation does not occur. This is RO level knowledge because Instrument technicians test isolation instrumentation daily in the plant without isolations occurring. Knowledge is covered in lesson plan.

Question Cognitive Level:

Examinee must know discrete bits of information about the system.

Technical Reference(s): OPL1 71.017 Rev 15 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.3 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Nt Thmddm1b4n

• how th aenada ki the HZd nCP4ra*J St$

Channel A kgk Channel B logic Reset Seal 1(26 1(48 1 -

NRHX Hfgh Temp >

(Not PCIS)

Sensot Relays K6OB K60A Open on Flgh

• Area Temperatures 1(600 K60C RPS Bus KA K68 B Low Water Level Sensor Relay Contacts K8C K6D Valve Actuation Relays Cause Valve Closure *1<27 When Deenerglzed Inboard Valve (Ref 730t927 Sb-I 3)

RWCU Valves

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .017 Revision 15 Page 12 of 56 INSTRUCTOR NOTES C. Typical PCIS Isolation Logic A typical logic arrangement for the PCIS valves PCIS de-energizes (except MSIVs) is shown in TP-l. This figure shows to isolate (except that two separate trip channels (A and B) are each HPCI/RCIC) provided with two sensor relay contacts (A/C and BID). Obj.V.B.1 Obj. V.C.1

a. This arrangement creates trip subchannels A1IA2 and 81/82.

b. A trip of either sensor relay within a trip HPCIIRCIC are channel will cause opening of the associated energize to actuate contact and de-energization of the associated relay. This condition will create a half Obj. V.8.3 isolation signal within both logic channels Obj. V.C.3 but NO VALVE MOVEMENT.

c. Should a trip of either sensor relay in the other trip channel occur, conditions will exist to de-energize the valve actuation relays in each logic channel, causing isolation valves to close.

PCIS logic is arranged as follows:

Al OR A2 r AND = Inboard AND Outboard valve closure BI OR B2 Note: Most PCIS logic is assembled as above.

The MSL drains however are an exception.

The MSL drain logic is as follows:

Al AND 81 = I/B valve closure A2 AND B2 = 0/8 valve closure

2. The Channel A logic is powered from RPS Bus A, Al A2 and contains the valve actuation relay associated with the Inboard Valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .017 Revision 15 Page 13 of 56 INSTRUCTOR NOTES

3. The Channel B logic is powered from RPS Bus B, Bi B2 and contains the valve actuation relay associated with the Outboard Valve.

4. it is noteworthy to point out that while RPS A &B supplies power to their respective logic channels, a loss of A RPS would result in the closure of the 1/B MSL drain valve FCV-55, and a loss of B RPS would result in the closure of the 0/B MSL drain valve FCV56. This is due to a loss of power to their respective relays rather than satisfying the logic.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 223002 PCIS/Nuclear Steam Supply Shutoff Tier# 2 K4.05 (IOCFR 55.41.7)

Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM I Group# 1 NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) KIA# 2230 02 K4 05 and/or interlocks which provide for the following:

Single failures will not impair the function ability of the system Importance Rating 2.9 Proposed Question: # 42 Unit 2 is starting up following a refueling outage with Reactor Pressure at 80 psig.

RPS MG Set A has tripped. RPS Distribution Panel A has NOT yet been transferred to its alternate source.

The 3 X Low Reactor Water Level instrument providing input to PCIS Channel B2 fails downscale.

Which ONE of the following describes the response of MSIVs AND Main Steam Line Drains?

A. ONLY the Inboard Steam Line Drain valve AND ALL MSIV5 close.

B. ONLY the Outboard Steam Line Drain valve AND ALL MSIV5 close.

C. Inboard AND Outboard Steam Line Drain valves AND ALL MSIVs close.

D. Inboard AND Outboard Steam Line Drain valves close, AND ALL MSIVs remain open.

Proposed Answer: C Explanation A INCORRECT: Plausible in that Loss of RPS A will close MSL Inboard (Optional): Drain Valve AND deenergize MSIV AC solenoids. However with B2 failed downscale and RPS A deenergized, both A and B logic are made up to deenergize both AC and DC solenoids and provides an isolation signal to the outboard MSL drain. If Bi channel had failed, this would be the correct answer.

B INCORRECT: Plausibility based on misconception that only outboard will isolate as result of combination of logic power and failure of B2. The inboard valve will close as a result of loss of relay power with loss of RPS A.

If RPS B had failed, this would be the correct answer.

c CORRECT: Channel B2 tripped would give a Group 1 logic BID tripped, loss of RPS A would remove power from Group 1 logic A/C and result in a full MSIV isolation. A2 (Loss of RPS) and B2 closes outboard steam line drain. Loss of A logic power from RPS A will close the Inboard steam line drains.

D INCORRECT: Plausibility based on misconception that DC Pilot Solenoids would remain energized and therefore MSIVs remain open since either solenoid energized maintains the valves open. If B logic was also powered from 250 VDC, like the DC solenoids, this would be the correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests candidates knowledge of Primary Containment Isolation System design features and interlocks which provide for single failures not impairing the function ability of the system.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 2-Cl-i, Rev. 47 (Attach if not previously provided)

OPL17i.0i7, Rev.15 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .017 V.B.3 (As available)

Question Source: Bank # Brunswick 07 #17 (Note changes or attach parent)

Question History: Last NRC Exam Brunswick 2007 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .017 Revision 15 Page 12 of 56 INSTRUCTOR NOTES C. Typical PCIS Isolation Logic A typical logic arrangement for the PCIS valves PCIS de-energizes (except MSIV5) is shown in TP-1. This figure shows to isolate (except that two separate trip channels (A and B) are each HPCI/RCIC) provided with two sensor relay contacts (A/C and BID). Obj. V.B.1 Obj. V.C.1

a. This arrangement creates trip subchannels A1IA2 and B1/B2.

b. A trip of either sensor relay within a trip HPCIIRCIC are channel will cause opening of the associated energize to actuate contact and de-energization of the associated relay. This condition will create a °half Obj. V.B..3 isolations signal within both logic channels Ob). V.C.3 but NO VALVE MOVEMENT.

c. Should a trip of either sensor relay in the other trip channel occur, conditions will exist to de-energize the valve actuation relays in each logic channel, causing ith isolation valves to close.

PCIS logic is arranged as follows:

Al OR A2 AND = Inboard AND Outboard valve closure Bi OR B2 Note: Most PCIS logic is assembled as above.

The MSL drains however are an exception.

The MSL drain logic is as follows:

Al AND Bi = I/B valve closure A2 AND B2 = 0/B valve closure

2. The Channel A logic is powered from RPS Bus A, Al A2 and contains the valve actuation relay associated with the Inboard Valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.017 Revision 15 Page 13 of 56 INSTRUCTOR NOTES

3. The Channel B logic is powered from RPS Bus B, Bi B2 and contains the valve actuation relay associated with the Outboard Valve

4. It is noteworthy to point out that while RPS A &B supplies power to their respective logic channels, a loss of A RPS would result in the closure of the I/B MSL drain valve FCV-55, and a loss of B RPS would result in the closure of the 0/B MSL drain valve FCV-56. This is due to a loss of power to their respective relays rather than satisfying the logic.

D. Group 1 (MSIV) Isolation Logic TP-2 provides a simplified diagram of the isolation logic for the A main steamline inboard isolation valve (FCV-1-14). 2-730E927-10

2. The MSIV is provided with both an AC-powered pilot Obj. V.B.2 solenoid (FSV-1-14C) and a DC-powered pilot Obj. LC.2 solenoid (FSV-1-14B).

Both of these pilot solenoids must be de-energized to cause the MSIV to close.

3. With the control handswftch in the AUTO/OPEN position, the associated HS-1-14A contacts will be closed.

a. Should a Group I isolation signal exist, the K7A,B,C,D relays will de-energize (see TP 3), causing the associated contacts to open.

b. When these contacts open, the K13/K51 and K14/K77 relays de-energize, opening the associated contacts. This will cause the pilot solenoids to de-energize and the MSIV will close.

4. Further detail regarding the MSIV isolation and reset logic can be seen in TP-3. This is a simplified 2730E927L7 illustration of the Al isolation Sub-channel (relay K7A)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17tO17 Revision 15 Appendix C Page 54 of 56 NORMJEMERG sw in NORM HS 1-148 FSV 1-14A (Test Sol)

HS 1-14 contacts shown with the control switch In the CLOSED positIon. Contacts will be closed with the control switch in the AUTOIOPEN position Reference Drawina: 2-730E927 sheet 10)

TP-2 TYPICAL MSIV CONTROL CIRCUIT (FCV-1-14)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0PL17t017 Revision 15 Appendix C Page 55 of 56 ckses wtien Mode Switch m not m RUN Opens at> 85 psig M$L K4A IE *i_

pressu te Opens on x low RPV KIA IT I I water level Opens on steam tiamel high temperature Opens on MSL Ngh flow K2A K3A :

KM NP (sea)

(Reset) 120 VAC RPS Bus w

K13 K14

1: HS HS126*

KS i-ar K7A HP 1-61 T

<>K7A Contacts dosed when wiIch s h the CLOSE on (Reference 2?3OE927 SH7 TP4 GROUP I ISOLATION AND RESET LOGIC CHANNEL Al

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Main Steam System 2..OI-1 Unit 2 Rev. 0047 Page 9 of 64 3.2 Main Steam Isolation Valves (MS IV) 3.2.1 MSIV Closure A. The MSIVs should be fast closed when the reactor is shutdown and no steam flow, unless required to be slow closed by surveillance, test instruction, or an abnormal condition. [BFNPER 164499 B. When a MSIV is closed at power, the potential exists for an isolation of the Hydrogen Water Chemistry System to occur, This is due to the possibility of a hydrogen bubbe becoming entrained in the main steam line drains and subsequently being released when the main steam line drains reposition In response to a MSIV closure. This scenario can result in a small Off Gas System hydrogen spike of sufficient strength to cause a automatic isolation of the Hydrogen Water Chemistry System.

C. Closure of all MSlVs could cause turbine shaft damage if main condenser vacuum is maintained and seal steam supply is not established from the auxiliary boiler.

3.2.2 MSIV Isolation A. Main steam tunnel temperature should not be allowed to exceed 189°F to prevent MSIV Isolation.

B. Whenever reactor pressure is reduced to 852 pslg and the reactor mode switch Is in RUN position, the MSIVs will close.

C. The MSIVs will close if 250 Vdc and 120 Vac power to the MSIV control logic is de-energized.

D. Reactor power should be 66% prior to closing an MSIV greater than 15 percent during closure testing. This should prevent a high steam line flow MSIV closure and subsequent reactor scram.

E. Placing all MSIV Handswitches in the Close Position allows the PCIS group one trip logic to be reset. Leaving any Handswltch in the Open Position prevents resetting the group one logic.

F. The PCIS group one trip parameters do not exceed trip setpoints.

1. Reactor water level above -122 in.

2. MSL flow less than 135%.

3. MSL tunnel temperature less than 189°F.

4. MSL pressure greater than 852 psig If in Mode 1.

BRUNSWICK 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ES-401 Sample Written Examination Form ES..401 ..5 Question Worksheet Examination Outline Crossreference: Level RO SRO Tier# 2 Group# 1 K/A It 223002 K4.05 Importance Rating 2.9

)<ii,wiede of PRIMARY CONTAINMENT ISOLATION SYSTEMINUCLEAR STEAM SUPPLY $HUT.OFF design feature(s) andlo inIerkX* which provld iør the iiowin SLagie faures will not ipalr the function ebility of the system Proposed Question: Common 17 RPS MG Set A has tripped. RPS Distribution Panel A has NOT yet been transferred to its alternate source The LL3 instrument providing input to PCIS Channel 82 rails downscale.

Which of the following describes the response of MSIVs and Steam line Drains?

. A.

8.

C.

Only the Inboard Steam Line Drain valve and all MSIVs close.

Only the Outboard Steam line Drain valve and all MSIVs close.

Inboard and Outboard Steam Line Drain valves and all MSIVs close.

0. Inboard and Outboard Steam Line Drain valves close, and all MSIVs remain open.

Proposed Answer: C Explanation (Optional):

Channel B2 tripped would give a Group I logic B/D tripped, loss of RPS A would remove power from Group I logic A/C and result in a full MSIV isolation. A2 (Loss of RPS) and 82 closes outboard steam line drain, Loss of A logic power from RPS A will close the Inboard steam line drain.

See Figure 25.7 in SD-25 Technical Reference(s): 50-025 (Attach if not previously provided)

NUREG-1021, Revision

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 239002 SRVs Tier# 2 K1.O1 (IOCFR 55.41.3)

Knowledge of the physical connections and/or cause-effect Group# 1 relationships between RELIEF/SAFETY VALVES and the following: K/A# 239002K 1.01

. Nuclear boiler 4 Importance Rating 3.8 Proposed Question: # 43 During a transient on Unit I, Reactor Pressure reached 1150 psig.

Which ONE of the following identifies how many SRVs opened?

A. Four B. Eight C. Nine D. Thirteen Proposed Answer: B Explanation A INCORRECT: Plausible in that this would be the correct answer if Reactor (Optional): Pressure was between 1135 and 1145 psig.

B CORRECT: The first two groups open with Reactor Pressure> 1145 psig.

Each of these groups has 4 valves.

C INCORRECT: Plausible in that this would be the correct answer if group 2 had 5 SRVs instead of group 3 D INCORRECT: Plausible in that this would be the correct answer if Reactor Pressure was>1155 psig.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates knowledge of the cause-effect relationship between the Nuclear Boiler and SRVs.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL1 71 .009, Rev. 11 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71.009 V.B.2 (As available)

Question Source: OPL1 71.009 #3 (Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .009 Revision 11 Page 17 of 63 (e) This relief mode logic can be defeated by use of a switch on 9-3. This switch MSRV AUTO ACTUATION LOGIC INHIBIT (XS-1-202) also brings in an alarm on 9-3.

d. Valve setpoints for safety function Obj. VB.2 Qbj.V.C.1 (1) 4 valves @1135 psig÷3% Obj. V.D.1 Obj. V.E1 (2) 4 valves @ 1145 psig + 3%

(3) 5 valves © 1155 psig +/-3% TP-3

e. Blowdown path Qbj. V.B.2 Obj. V.C.1 (1) Individually piped to the suppression pool via the T-Quenchers below the minimum water leveL The T Quenchers enhance thermal mixing in the Suppression Pool (2) Each SRV has two vacuum breakers Obj. V.B.5 (one 10 inch and one 2 1/2 inch Obj. V.C.2 vacuum breaker on SRV tailpiece)in Obj. V.C.3 parallel. They are provided to allow Obj. VD.2 entry of drywell air into the relief line Obj. V.E.2 to prevent water from the suppression pool being pulled up into the relief line upon completion of blowdown. Without the vacuum breaker the steam in the relief line condenses and forms a vacuum in the relief line drawing water from the pool into the line. Subsequent reopening of the valve with its relief line partially filled with water could over pressurize the relief line, with a potential for tail pipe damage. If the vacuum breakers were to fail open, steam could be discharged directly to the drywell during SRV operation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EXAM BANK OPL171.009 3 During a transient, RPV pressure reached 1150 psig.

Assuming no operator action, how many SRVs opened?

A. Four B. Eight C. Nine D. Thirteen Answer: B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO SRO 259002 Reactor Water Level Control System Al 01(10 CFR 55 41 5) Tier # 2 Ability to predict and/or monitor changes in parameters associated Group # 1 with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: K/A # 259002A1.05

• Reactor water level Importance Rating 38 Proposed Question: # 44 Unit 2 Feedwater Level Control System (FWLCS) is operating in 3-Element Control with Narrow Range Level Instruments indicating as follows:

• 2-LT-3-53, LEVEL A, (+) 46 inches

• 2-LT-3-60, LEVEL B, (+) 32 inches

• 2-LT-3-206, LEVEL C, (+) 34 inches

• 2-LT-3-253, LEVEL D, (+) 33 inches Which ONE of the following completes the statement?

If 2-LT-3-60, LEVEL B, is manually bypassed, the FWLCS will control Reactor Water Level based on A. ONLY the 2-LT-3-206 instrument B. LOWEST of 2-LT-3-206 OR 2-LT-3-253 instruments C. AVERAGE of 2-LT-3-206 AND 2-LT-3-253 instruments D. AVERAGE of 2-LT-3-53, 2-LT-3-206, AND 2-LT-3-253 instruments Proposed Answer: C Explanation A INCORRECT: Plausible in that if FWLCS selected the middle of the 3 (Optional): remaining channels when one channel is bypassed, this would be the correct answer.

B INCORRECT: Plausible in that if FWLCS selected the lower of the channels not manually or automatically bypassed, this would be the correct answer.

C CORRECT: The average level value is used for the three element control logic. The algorithm validates each level signal by comparing them to the average. Level signals that deviate from the average by more than 8 inches are declared invalid, and are discarded from the average. LT-3-53 deviation is> 8 and is bypassed and LT-3-60 is manually bypassed. If two level signals are BAD or invalid, the algorithm will average the remaining two levels and will control on that value. In this instance the two remaining signals D INCORRECT: Plausible in that if candidate fails to recognize that 2-LT-3-53, LEVEL A is bypassed due to deviation >8 inches from average, this would be the correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question test candidates ability to predict and monitor changes in Reactor water level associated with operating the Reactor Water Level Control System.

Candidate must recognize that one level channel meets the criteria to be automatically bypassed. Then, when another channel is manually bypassed, candidate must predict how the level control logic will function to monitor for expected changes in Reactor Level.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPLI 71 .012 Rev 14 (Attach if not previously provided) 2-01-3 Rev 136 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.012 V.B.5 (As available)

Question Source:

(Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Although this question has been modified from its original bank form to meet the KA, it does not meet the criteria for a significantly modified question and is therefore designated as a Bank Question.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Feedwater System 2-01-3 Unit 2 Rev. 0136 Page 200 of 216 Illustration 8 (Page 1 of 7)

RFWCS Instrumentation 1.0 NARROW RANGE REACTOR WATER LEVEL 1.1 Components 2-Ll-3-53 2-LI-3-60 2-LI-3-206 2-Ll-3-253 1.2 Description The instruments are located on Panel 2-9-5 along with their corresponding bypass pushbuttons. These instruments provide two types of indication and ranges; analog (0 to 60 inches) and digital (-10 to 70 inches). Each instrument has an amber light which illuminates when the signal has been bypassed automatically by the RFW Control System or manually by the Unit Operator.

1.3 System Operation The RFW Control System will use a level signal provided the system determines the signal to be good and valid. A GOOD level signal is one that has not failed and is on scale. A VALID level signal is one that does not deviate from the average (or median) level by more than 8 inches.

The REW Control System validates each narrow range level signal by comparing them to the average. A level signal that deviates from the average by more than 8 inches is declared invalid and is bypassed. A level signal that is declared bad by the RFWCS will also be bypassed automatically.

To avoid individual on-scale but faulty level signals from skewing the average, a secondary validation process is used to compare the average level to the median of the valid signals. If the average value differs from the median value by more than 4 inches, the RFWCS will validate each level signal to the median value instead of the average. In this case, any level signal that varies by more than 8 inches from the median is declared invalid and bypassed by the system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 012 Revision 14 Page 15 of 85

4. Additional process measurements include: ObJ. V.8.1

a. Four feedwater ilne temperature RTD outputs.

b. Three reactor feedwater pump turbine speed signals output from the three Woodward Governors. Each has 2 MPUs but only one needed for speed indication to pnl. 96. Third MPU is for zero speed.

B. Component Description Reactor Water Level Obj. V.8.1 Obj. V.D.5

a. Four independent narrow range level transmitters (LT-3-53,-60,-206 and 253).

They are differential pressure transmitters connected to water reference condensing chambers. Digital readouts are spanned for a reactor level of -10 to + 70 inches but, analog range is still 0 to 60.

b. The control algorithm checks the signal quality. If BAD (failed or outof-range high or low), the signal is discarded. if GOOD, the signal is further processed.

C. Each level signal Is pressure compensated for density differences by the algorithm and the four signals are averaged.

d. The aigorithm validates each level signal by comparing them to the average. Level signals that deviate from the average by more than 8 inches are declared invalid, and are discarded from the average.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 012 Revision 14 Page 16 of 85

e. The algorithm applies a second validation GOOD in this process to prevent an individual GOOD, but case means on faulty level signal from degrading the scale but faulty.

average. The average is compared to the Example:

median level signal. Where there is an even variable leg leak number of signals, the median will select which causes a the higher of the two middle values. If the low level signal average and median values deviate by on two LIs..

more than 4 inches, the algorithm wilt 23,24,34,&

validate the individual level signals to the 35 ave. of 29.

median, instead of the average. In this case Median value of any individual level signal that deviates from 34 > 29 by 5.

the median by more than 8 inches is The 23 & 24 declared invalid and is discarded from the signals are calculations. discarded causing median to go back to an average of 34.5.

f. The average level value Is used for the single element and three element control logics.

(1) The individual density compensated Obj. V.8.1 levels are output to Control Room indicators.

(2) The average level is output to one pen of a twopen recorder.

(3) Reactor Vessel high and low level alarms are generated by comparing the average level to high (>39) and low (<27) setpoints.

(4) The average level is also used in the Obj. V.8.7 Recirculation pump runback level Ob). V.C,6 interlock logic within the algorithm.

g. If one level signal is BAD or invalid, the Obj. V.8.6 algorithm will calculate the average of the Obj. V.C.5 three remaining level signals and will control on that value.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7I XJ12 Revision 14 Page 17 of 85

h. If two level signals are BAD or invalid, the algorithm will average the remaining two levels and will control on that value. In this instance the two remaining signals are compared to each other. If they deviate by more than 8 inches, a process alarm will be generated, but neither will be declared invalid.

I. If three level signals are BAD or invalid, the algorithm will control on the remaining signal alone.

j. If all four level signals are BAD or Invalid, the algorithm will transfer the system to Manual control mode, and generate a process alarm.

Should not be able to manually bypass all 4 (using pushbuttons)

2. Main Steam Flow Obj, V.D.5

a. Four steam flow differential pressure Obj. V.B.1 transmitters provide square-rooted signals corresponding to 0 to 5 Mlblhr flow rates.

(actual 4.6 to 4.7Mlblhr)

b. The control algorithm checks the input signal quality and discards BAD data signals.

c. Each steam flow signal is adjusted for a flow nozzle adiabatic expansion factor, which is a function of the nozzle geometry and the ratio of the nozzle throat pressure to inlet pressure.

d. The algorithm calculates the average steam line flow and derives a total steam flow by multiplying the average by 4.

e. The total flow is further compensated for density based on the reactor pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 261000SGTS Tier# 2 K4.05 (IOCFR 55.41.7)

Knowledge of STANDBY GAS TREATMENT SYSTEM design Group # 1 feature(s) and/or interlocks which provide for the following:

K/A # 261 000K4.05

• Fission product gas removal Importance Rating 2.6 Proposed Question: # 45 Which ONE of the following completes the statement?

Standby Gas Treatment System _(1) are designed to remove a MAXIMUM of _(2) of elemental iodine.

A. (1) HEPA Filters (2) 70%

B. (1) Carbon Beds (2) 70%

C. (1)HEPAFilters (2) 99.9%

D. (1) Carbon Beds (2) 99.9%

Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausible in that HEPA filters function to (Optional): remove fine particulate matter. Part 2 incorrect Plausible in that this is a recognizable value associated with the performance of SGTS filter trains.

The electric heaters reduce Relative Humidity down to 70% which is part of the criteria for Carbon Bed iodine removal capability.

B INCORRECT: Part 1 correct See Explanation D. Part 1 incorrect See Explanation A.

C INCORRECT: Part 1 incorrect See Explanation A. Part 1 correct See Explanation D.

D CORRECT: Parts 1 and 2 correct Carbon Beds are designed to remove at least 99.9% of elemental iodine upon entering conditions of 70% relative humidity at 190°F.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests candidates knowledge of Standby Gas Treatment System Carbon Bed design criteria which provide for fission product gas removal.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL171.018 Rev. 10 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.018V.B.6 (As available)

Question Source: Bank #

Modified Banlc# (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DPL17 LOiS Revision ID Page 14 of 37 INSTRUCTOR NOTES

k. Decay heat removal crosstie valves Orosstie solenoid valves and control switches DON Wi 041 6A have been removed because of inability to meet EQ requirements. Line size has been increased and a manual, locked damper has been placed in the lines to ensure sufficient flow.

I, Cross-tie valve for Trains A and B (22) TP- I (1) Normally closed Review H0.2 (PIP-95-32)

(2) No automatic actions (3) Used for full cross-tie capability between Trains A and B.

(4) Normally powered from Dsl Aux Bd A, Obj. V.E.5 automatically trarsfers to Dsl Aux Sd B on loss of power to Board A.

I Moisture Separator Ohj. \/.B.6..a Obj. VOA.a

a. Reduces moisture content of incoming air Obj. V.D.4.a Obj. VE.2

b. Woven nylon mesh, traps water droplets

c. Moisture drains by gravity to SBOT sump and is then pumped to Radwaste.

4. Electric Heater Ohj. V.B.S.b r a. The relative humidity heater reduces relative Obj. V,C.4.h humidity to 70%. Obj. V.D4.b Obj. V.E.2

b. 40kW heaters for relative humidity control. SOT A and B powered from A and B 480v Dsl Aux Bds respectively. SOT C from the SOT board.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17I .018 Revision 10 Page 15of37 INSTRUCTOR NOTES

c. 15kW charcoal bed heaters formerly maintained LER 35-029-01 a 125°F charcoal bed temperature when SBGT was out of service. Heater control switches were spring-returntoneutral arid required resetting after SBGT operaflon.

Due to the Technical Specification requirement of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> t monthly operation with the relative humidity heaters in service, the charcoal bed heaters are no longer needed

d. The relative humidity heater is energized automatically on startup by the fan breaker closure and is dc-energized on shutdown by fan breaker opening.

e. The heater will also trip if ambient temperature reaches 180°F.

f. If 480 Volt Load Shed logic is initiated: the Train A and B relative humidity heaters will automatically trip. They will restart after 40 seconds. Train C is not affected by the 480 volt load shed logic.

g. Relative humidity heater control switches t12, 34.

60) in ON or OFF cause annunciation.

5. Prefilter Oh]. \/.B.6.c Obj. V.C.4.c Used to remove large particles (dust, dirt. lint) and to Obj. \/.D.4.c protect HEPA filter Obj. V.E.2

6. HEPA Filter Ohj. V.B.6.d Obj. V.C.4.d Removes 99.9% of 0.3 micron particles Oh]. V.D.4.d Oh] VE2

7. Carbon Bed (Adsorber Type) Obj, VB.6.e Ohj. V.C.4.e

a. Designed to remove at least 95% of iodine in the Obj. V.D.4.e form of methyl iodine (CH3I) and 99.9% of Obj. V.E.2 elemental iodine upon entering conditions of 70%

relative humidity at 1 90T

b. Made up of individual rectangular canisters of charcoal

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 262001 A.C. Electrical Distribution Tier # 2 K3.04 (IOCFR 55.41 .7)

Knowledge of the effect that a loss or malfunction of the A.C. Group # 1 ELECTRICAL DISTRIBUTION will have on following: K/A # 262001 K3.04

• Uninterruptible power supply Importance Rating 3.1 Proposed Question: # 46 The Unit 1 Unit Preferred Inverter is operating in a normal lineup, when a loss of off-site power AND a failure of DG A to start occurs.

Based ONLY on the above plant conditions, which ONE of the responses will identify the power source for Battery Board 1, panel 10?

The Unit Preferred Inverter is powered from A. 480V RMOV BD IA B. 250 VDC Battery Board 4 C. 250 VDC Battery Board 5 D. the Unit Preferred Transformer Proposed Answer: C Explanation A INCORRECT: The UPS Rectifier/Inverter is normally powered from the (Optional): 480V RMOV BD 1A, but it is NOT energized based on the conditions given.

Plausible because the candidate may believe that 480V RMOV BD supplied by auto transfer to DG B.

B INCORRECT: Battery Board 4 is the alternate DC supply to the invérter and would have to be manually shifted to supply it. Plausible because easily confused with Battery Board 5 and it is the normal supply to one of the MMGs. MMGs are also a Unit Preferred System.

C CORRECT: Loss of off-site power and a failure of DG A to start would result in no power to 4kV SD BD 1A, 480V SD BD 1A, and 480V RMOV BD 1A, which is the Normal supply to the Unit Preferred Rectifier/Inverter. The UPS would automatically shift to 250 VDC Battery Board 5 supplying the inverter, when the diode in the inverter is no longer reversed biased by the rectifier output.

D INCORRECT: The Unit Preferred Transformer is supplied by 480V RMOV BD 1A, which is also the normal supply to the Rectifier/Inverter. This RMOV Board has no power based on the given conditions. IF it were powered, it would have to be manually shifted to supply the static inverter. Plausible because candidate may believe it is powered from 480V RMOV Bd B.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the effects of loss of offsite power and failure of EDG A has on the Unit I Unit Preferred Inverter which is an uninte rruptible power supply.

Question Cognitive Level:

This question is low cognitive or memory question.

Technical Reference(s): OPL1 71 .102 Rev 7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.2.a (As available)

Question Source: Bank#

Modificd sni#* (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .102 Revision 7 Page 56 of 67 480V RMOV 8014 Norm 1JCfi1I Norm 25OVDC 25OVDC 6a:t lId 5 Osti Bd 4 BaN Bd 6 Alt NI 480V 25OVdc 250V do SD 3d BaIt lId S BaIt Sd 5 3A 1

Panel 101 JIOCI 1002 Panel 11 Bat Bdl t

L 480V RMOV lId 28 I

1 PanellO 460V RMOV Bd 3B Panel 1 I Ii BatlBd2 I L

F --

Panel lOt W03 F--

Panel 11 Salt Bd3 TP-2: POWER SUPPLIES TO UPS BATTERY BOARD CABINETS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.102 Revision 7 Page 19 of67 Instructor Notes (b) The INVERTER Unit (Unit 1> TP-2 Unit 1 is powered by an uninterruptibie power supply (inverter unit).Norrnal supply is 480 VAC to the rectifier!

inverter unit itself where it is converted to DC volts then back to a smooth 1201 24OVAC signal fed to Batt.

Bd 1 Panel 10/11. There is a backup 250 VDC backup pover supply fed to the inverter unit for a bumpless transfer in case of loss of AC power. Additionally there is a regulated AC alternate power to the inverter static switch for a continuation of power in case of inverter failure.

(c) Unit Preferred Transformer The alternate power source is the unit preferred transformer. This transformer receives power from the 480V portion of the standby AC power system.

Unit I transformer is from the 480 V RMOV Bd 1A Note that the UPS Unit 2 & 3 transformers are transformer is also povered from 480V RMOV the alternate RPS Board 2B & 3B. Transfers power supply for U2 to this source are done manually at battery hoard 2 panel 11.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO  : SRO 262002UPS(AC/DC)

Tier# 2 A2.02 (IOCFR 55.41.5)

Ability to (a) predict the impacts of the following on the Group # 1 UNINTERRUPTABLE POWER SUPPLY (AC/D.C.); and (b)

K/A # 262002A2.02 based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Overvoltage Importance Rating 2.5 Proposed Question: # 47 Which ONE of the following completes the statements?

The 1001 AND 1003 breaker from Unit 2 Unit Preferred System (UPS) Motor-Motor-Generator (MMG) set will trip on _(1)_ at the output of the MMG.

In accordance with 2-AOI-57-4, Loss of Unit Preferred, if UPS is lost, the crew must (2).

A. (1) underfrequencyONLY (2) take manual control of Master Feedwater Level Controller B. (1) underfrequency ONLY (2) verify Reactor Feedwater Control System is maintaining Reactor Water Level C. (1) underfrequency OR overvoltage (2) take manual control of Master Feedwater Level Controller D. (1) under frequency OR overvoltage (2) verify Reactor Feedwater Control System is maintaining Reactor Water Level Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausible in that under frequency ONLY at (Optional): the generator output will trip the DC Motor of the MMG set. Part 2 incorrect Plausible in that loss of UPS does impact Feedwater Level Control System. RFW Control System Panel Display Stations on Panel 2-9-5 is disabled. PDS Controls are inoperative and displays become blank. The RFW Control System continues to control system parameters according to water level setpoint.

B INCORRECT: Part 1 incorrect See explanation A. Part 2 correct - See explanation D.

C INCORRECT: Part 1 correct See explanation D. Part 2 incorrect See explanation A.

D CORRECT: Part 1 correct The 1001 and 1003 breakers from an MMG set will trip on overvoltage or under frequency at the output of the MMG.

Part 2 correct Per 2-AOI-57-4, Subsequent action 4.2[1], verify RFW Control System is maintaining Reactor Water Level. The RFW Control System continues to control system parameters according to water level setpoint.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the Candidates ability to predict the impacts of Over voltage on the Unit 2 Unit Preferred System MMG which is an uninterruptable power supply Then, assess impact of loss of UPS on FWLC to determine correct actions in accordance with 2-AOl-57-4.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meanin g

to predict the correct outcome. Candidate must predict impact of loss of UPS on FWLC to determine appropriate action to take.

Technical Reference(s): OPL17I.102 Rev. 7 (Attach if not previously provided) 2-AOI-57-4 Rev. 47 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL17I.102V.B.2 (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New X Question History: rr Last NRC Exam r

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71102 Revision 7 Page 20 of 67 Instructor Notes (d) Another lJnits MMG set The second alternate is from another uniVs MMG set OLitput. Unit 2 MMG is the second alternate for either Unit 1 or Unit 3; lJnit 3 is the second alternate for lJnt 2.

Transfers to this source are done manually at Battenj Board 2 panel 11

b. MMG Sets (Unit 2&3) Cbj. V.B.2.h TP-11 (1) The MMG is normally driven by the Obj.V.D .2 .c AC motor, powered from 480V Obj.\/. D.2.cb Shutdown Board A. Should this Obj V.E.2.c supply fail, the AC motor is Obj .V. E.2.d/i automatically disconnected and the Obj V.6.2.h DC motor starts powered from Obj V. C .3.e 250V Batter Board. The DC Obj .V. B .2.j motor has an alternate power Obj.V.E.2.i supply from another 250V Battery Board. Transfer to the alternate DC source is manual. Under-frequency on the generator output will trip the DC motor. Transfer of the MMG set back to the AC motor is manual.

(2) The 1001 and 1003 breakers from an MMG set will trip on overvoltage or underfrequency at the output of the MMG. Also Unit 2 MMG breakers are interlocked to prevent alternate power to unit I and 3 at the same time.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 17 I. 1 02 Revision 7 Page 21 of 67 Instructor Notes (3) When an under frequency or Obj. V.B.2.h overvoltage condition exists at the Obj. V.C.3.e Generator Output the following Obj. V.D,2.j occurs: Ohj. V.E2.i (a) BB panel IC breakers from the MMG set trip.

U2 1001 (U2) 1003 (tJl&3) r U3 1001 (U3) 1003 (U2)

(b) Excitation is lost and the MMG Set continues to run.

(The Hold to build up voltage switch must be depressed to restore voltage.)

(4) The starting sequence for the MMG is as follows:

(a) Stan the AC motor first. This Instructor:

is a larger motor and has Emphasis enough power to procedural compensate for the flywheel adherence load.

(b) Transfer to DC motor by stopping the AC motor. This automatically starts the DC motor allows the speed to he controlled for paralleling.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of Unit Preferred 2-AOl-57-4 Unit 2 Rev. 0041 Page 5 of 32 2.0 SYMPTOMS (continued)

F. Loss of RPIS. REFER TO 2-AOI-85-4.

S. PEW Control System Panel Display Stations on Panel 2-9-5 disabled. PDS Controls are inoperative and displays become blank. The RFW Control System continues to control system parameters according to water level setpoint.

H. The following RFW Control System annunciators in alarm on Panel 2-9-6:

t RFPT GO\!ERNOR POWER FAILURE OR GOVERNOR ABNORMAL (2-XA-55-6C, Window 12).

2 RFWCS TROUBLE :(2XA556C Window 28).

L The following EHC Control System annunciators in alarm on Panel 2-9-6:

I EHC POWER ABNORMAL (2-XA-55-7B Window 5) 2 EHCJTSI SYSTEM TROUBLE (2-XA-55-7B Window 6)

J. EHC Control System PLIJ 1 (power load unbalance) can bypass with a sustained loss of power to Panel 9-9 Cabinet 5. An uninterruptible power supply will keep the PLU energized for approximately l 5 minutes after normal power is lost.

K. EHC Control System HMI on Panel 2-9-31 may become blank if power is lost to Panel 9-9 Cabinet 6. An uninterruptible power supply will keel) this component energized for approximately 15 minutes after normal power is lost.

L. RECLRC FLOV/ SYSTEM TROUBLE ALARM (2-XA-55-4A, Window 23).

M. Loss of power to CRD Select Modules..

N. ANN: PNL 2-9-21 SYS LEAK DETECTION POWER FAILURE (2-XA-55-3D, Window 31)on loss of power to Panel 2-9-21 Steam Leak Detection Panel.

O TIP isolation signal when Cabinet 5 (Breaker 503) is dc-energized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of Unit Preferred 2-AOI-57.4 Unit 2 Rev. 0041 Page 8 of 32 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None NOTE The blanks to the side of steps contained in Section 4.0 Operator Actions are intended for place keeping only. Initials are NOT required. If necessary, place keeping marks may be made directly in the Control Room copy of this instruction CONTACT Management Services for a replacement copy when time perrnts 4.2 Subsequent Actions

[I] VERIFY the following:

• RFW Control System is maintaining Reactor Water Level.

• Recirc Flow Control System maintaining Recirc pump speeds.

• EHC Control System maintaining Reactor Pressure and Turbine control parameters. 0

• VERIFY TIP ISOLATION.

[2] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s). (Otheiwise N/A)

CAUTION While RPIS and the process computer are inoperable, control rod movement may only be performed by manual reactor scram.

[3] IF control rod movement is required while RPIS and The process computer are inoperable, THEN INSERT a MANUAL SCRAM. REFER TO 2-AOl-i 00-i.

(Otherwise N/A)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUS(BILITY SUPPORT QPLI7I.102 Revision 7 Page 20 of67 Instructor Notes (d) Another Units MMG set The second alternate is from another unlIrs MMG set output. Unit 2 MMG is the second alternate for either Unit I or Unit 3; Unit 3 is the second alternate for Unit 2.

Transfers to this source are done manually at Battery Board 2 panel 11.

b. MMG Sets (Unit 2&3) Obj. \.!.B.2.b TP-1 1 (1) T.he MMG is normally driven by the Ohj .V.D .2. c AC motor, powered from 480V o hj.V. D .2.dfj Shutdown Board A. Should this ObI V.E.2.c supply fail, the AC motor is Obj.V. E .2.dli automatically disconnected and the Obj V.B.2.h DC notor starts. powered from Obj .V .C .3 .e 250V Battery Board. The DC Ohj.V.D.2.j motor has an alternate power Obj.V. E .2 supply from another 250V Battery Board. Transfer to the alternate DC source is manual. Under-frequency on the generator output will trip the DC motor. Transfer of the MMG set back to the AC motor is manual.

(2) The 1001 and 1003 breakers from an MMG set will trip on overvoltage or underfrequency at the output of the MMG. Also Unit 2 MMG breakers are interlocked to prevent alternate power to unit 1 and 3 at the same time.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO 263000 DC Electrical Distribution K6.02 (100FR 55.41.7) Tier # 2 Knowledge of the effect that a loss or malfunction of the following Group # 1 will have on the D.C. ELECTRICAL DISTRIBUTION:

K/A # 263000K6.02

• Battery ventilation Importance Rating 2.5 Proposed Question: # 48 Which ONE of the following is a concern to plant operation if the Battery and Board Room Exhaust Fans are not operating properly?

A. The lead-calcium batteries tend to release toxic gas into the atmosphere above 90 °F.

B. The design limit for hydrogen concentration in the rooms may be reached when the batteries are being charged.

C. Electrical Maintenance will not be able to obtain accurate cell specific gravity readings if temperature is above 90 °F.

D. The quarterly battery SR frequency is lowered to weekly when temperatures are outside the 70 °F to 90 F temperature range.

Proposed Answer: B Explanation A INCORRECT: Lead-calcium batteries suffer degraded performance at high (Optional): temperatures but do not release toxic gas as a result.

B CORRECT: Battery Room ventilation is required to prevent buildup of explosive hydrogen concentration.

C INCORRECT: This would be correct for low temperatures.

D INCORRECT: Quarterly battery SR frequency is lowered if temperatures were below the temperature range, not above it.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates knowledge of the impacts of a loss I malfunction of battery ventilation on the DC Electrical Distribution System.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): 0-01-31, Rev. 136 (Attach if not previously provided)

OPL1 71 .037 Rev. 12 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.037 V.B.10 (As available)

Question Source: Bank # HLT 0707 # 23 Modified Bank # - -

(Note changes or attach parent)

New Question History: Last NRC Exam Browns Ferry 0707 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of ever, question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off-Gas Treatment 0-01-31 Unit 0 Building Air Conditioning System Rev. 0136 Page 129 of 286 7.11 Shutdown of Battery and Board Room Exhaust Fans

{

CAUTION Battery Room ventilation is required to prevent buildup of explosive hydrogen concentration.

[1] REVIEW all Precautions and Urnitations in Section 3.0.

[2] OBTAIN Unit Supervisors approval prior to shutting clown the fan(s).

[3] PERFORM the following at Panel 25-165 in Unit 1 Mechanical Equipment Room, El 617. to stop the running exhaust fan lA/i B(3A/3B):

[3.1] PLACE BATTERY & BOARD RM EXHAUST FAN 1A11B( A/3B), 0-HS-031-0074A(97A), in OFF.

[3.2] CHECK that green Oft light illuminates on upper left or right section of panel.

• Bat & Bd Rm Exhaust Fan IA( A)-upper left section of panel.

• Bat & Board Rm Exhaust Fan 1B(3B)-upper right section of panel. D

[4] REFER TO Section 8.15 for operation with ventilation out of service.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17 .037 Revision ii Page 23 of 69 INSTRUCTOR NOTES

7. Battery Room Ventilation Systems

{ a. Purpose The various battery room ventilation systems prov]de adequate room ventilation to prevent an explosive atmosphere due to hydrogen buildup from the Ohj. V.B.10 Ohj. V.0.10 Ohj. \/.D.8 batteries.

h. The Unit Battery Rooms 1, 2 and 3 and the Communications Battery Room are supplied air through the door ventilators. Air is exhausted with Battery and Board Room Exhaust Pans 1A and lB (Battery Room 1 and 2, and communications battery room), and Unit 3 Battery and Board Room Exhaust Pans 3A and 3B (Battery Room 3).

Plant/Station Battery Rooms are supplied air via an HVAC unit located outside the rooms to maintain an optimum temperature between 70 and 80 degrees P.

A small exhaust fan is located in the ceiling with a off and on switch located on the wall. (speed is variablet The purpose of the exhaust fan is to keep hydrogen concentration below 2%. With the exhaust fan off it will take over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to reach the design limit of 2% hydrogen. Upon loss of the exhaust fan.

a system abnormaV will alarm in the control room.

The ceiling also has vent pipes to exhaust the flow of air. Battery Room 4 also required the installation of a new bypass damper with the existing ventilation fan to maintain hydrogen concentration below the design limit. (Damper located by EHC tinit). The existing grille and damper between battery room 4 and the adjacent board room was blocked.

c, The 250V DC Shutdown Board Battery Rooms are supplied with supply and exhaust fans for each unit.

d. Each DG 125V DC battery has an exhaust fan that provides adequate ventilation in the battery area.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HLT 0707 NRC EXAM Question 23 Sample Written Exarninatien Form ES-40 3 Question Worksheet Examination Outline Cross-reference: Level RO CR0 2630001(5.01 Tier# 2 Knowledge of the operational implications of the lbllowing concept3 as they apply the DC Elechical Disbibution: Hydrogen GeneraUon r

cjIO 1

i-kP during battery charging.

IKJA# 263000K5.fl1 Importance Rating 2.6 2.9 Proposed OLlestion: RO # 23 Which ONE of the following is a concern to plant operation it the Plant/Station Battery Rooms HVAC units are not operating properly?

A. The design limit for hydrogen concentration in the rooms may be reached when the batteries are being charged.

B, Electrical Maintenance will not be able to obtain accurate Cell specific gravity readings if temperature is above 90°F.

C. The lead-calcium batteries tend to release toxic gas into the atmosphere above 90 °F, and access to the room would be limited.

0. The Quarterly Battery SR frequency is lowered to weekly when temperatures are above the 70 °F to 90°F temperature range.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 264000 Emergency Generators (Dies&/Jet)

Tier # 2 K5.05 (IOCFR 55.41 .5)

Knowledge of the operational implications of the following concepts Group # 1 as they apply to EMERGENCY GENERATORS (DIESEL/JET):

K/A # 264000K5.05

• Paralleling A.C. power sources Importance Rating 3.4 Proposed Question: # 49 Diesel Generator (DIG) A is synchronized to 4KV Shutdown Board A. The instrumentation readings for the DIG are as follows:

• Voltage=416OVAC

• Frequency 60 Hz

• Current = 280 amps

• Vars = 2200 Kvars

• Watts = 2600 Kw Which ONE of the following is the correct action to obtain a 0.8 lagging power factor?

Take the

[REFERENCE PROVIDED]

A. Governor control switch to RAISE.

B. Governor control switch to LOWER.

C. Voltage Regulator control switch to RAISE.

D. Voltage Regulator control switch to LOWER.

Proposed Answer: D Explanation A INCORRECT: The governor controls KW not KVAR. Candidate (Optional): misunderstanding of governor controlling speed and real load or KW.

B INCORRECT: The governor controls KW not KVAR. Candidate misunderstanding of governor controlling speed and real load or KW.

C INCORRECT: Taking the voltage regulator control switch to raise will increase generator excitation and raise KVAR. This will place the generator operating point farther away from the 0.8 power factor line. Candidate error in determining where the generator is operating in relationship to the 0.8 pf line.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: Need to lower KVARs by lowering generator excitation to lower reactive load. Desired operation at 2600 KW = a 1950 KVAR with a 0.8 lagging power factor.

KA Justification:

The KA is met because it tests knowledge of operational implications of paralleled AC sources design and how KW and KVAR are controlled to obtain optimum power factor on a DG Question Cognitive Level:

This question is rated as C/A due to the requirement to solve a problem using references. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 0-01-82 Rev 112 (Attach if not previously provided)

OPL171.038 Revl7 Proposed references to be provided to applicants during examination: 0-01-82 Illustration -1 Learning Objective: V.B.1 (As available)

Question Source: LXR TEST Bank# OPL171.038 #3 Last used BFN 1006 Audit Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Standby Diesel Generator System 0-01-82 lJnitO Rev,0112 Page 171 of 174 Illustration I (Page 1 of 1)

DG kW vs. kVAR Loading 3000 2600 2600 - OOW. *ts aATNa - - - -

r 2400 I

2200 2000 2000 XW 5 PEaCENT 0 CGMTNU0U I:

1800 - - - - - -

1400

-  ::  : if :

1200 - - -

- -I - -

DIAGONAL 1000 -

- I REPRSNtS soo

- I LAGGING OR OIJTGOtNG 0

VI 200 I

400 Ii If1 00 I

800 111 1000 1200 1400 I

100 I LLiI 1600 IiI 2000 2200 2400 kVAR (Lagging or Outgoing) kVAR

ES-.401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7I.038 Revmon i6 P,e IOci4 RSRUCT0R NOTES

s. Lesson Eody

.4. General Descnption Safety Objectives The safety ch;ective of the standby AC power system is to crovide a self-contained, high y r&iabe source of power as recLired for the engineered safeguard systems so that ro single credi.be event sari dsabie the core standby cooling run otions or their support ng auxiiaries.

6. Component Description Ob.1.Bi.

1. Diesel Generators (all 8)

a. Ratngs -4160 volt. 3 phase, 60 ha rated for 5eiewlNPO maximum loacing w,th 0. power factor lag: 83-01 (1) 2600r2553vKw continuous (>2 hours) ti-0-iL Emphasze i2 KW for 0-2 hours (Short t

256012&OO UDd2fl ci opera Tine Stesdy 5trie

3) 28.D/2S1 EKW ror 0c minutes cold toward Engine instantaneous) preventing Diese:

(4) 3050/3325r*Kwfor>3 miii. (Hot Genernftnres.

Engine Instantaneous)

15) 3575 KVA (short tine generator only) 0.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 16.) 3252 KVA (continuous) >2 hours. See 01-82 ;a:sr Note: Items t::2)(3)(4) & (6,i are rarrnac Engine ratings I:>, Cacaitie of fast starting and being ready to load within 10 seconds,

2. Reduced rating 1 & 2 (above) appy icr engine 0 -82 ?&L 3.2: 03 cad cooling water outlet temperature exceeding 190F in conjuncron wtth combustion air exceeding S0F.

aocu,r,uta:ion n the exhaiat system.

Reduced rating 38 (above: apply when 0b.LC.12 combustion ar exceeds SOCF regardles.s of engine cooling water outlet temperature. (For more detats see 01 62\

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 300000 Instrument Air System (lAS)

Tier # 2 K6.07 (IOCFR 55.41 .7)

Knowledge of the effect that a loss or malfunction of the following Group # 1 will have on the INSTRUMENT AIR SYSTEM: K/A # 300000K6.07

• Valves Importance Rating 2 5 -

Proposed Question: # 50 G Control Air Compressors microcontroller fails, causing the Compressor Inlet Flow Valve to throttle open and the Compressor Bypass Control Valve to fail fully open.

Which ONE of following completes the statement below?

G Air Compressors discharge pressure will A. decrease to LESS THAN 100 psig B. stabilize at 100 to 105 psig C. increase tol2O psig D. increase to 132 psig Proposed Answer: A Explanation A CORRECT: Pressure will decrease to the point of the compressor not (Optional): supplying compressed air (compressor running). Any air entering the compressor will be discharged through the Bypass Control Valve, to the Air Silencer, and back to atmosphere. The two selected lead air compressors start at 98 psig; the first lag at 96 psig and the second lag at 94 psig.

B INCORRECT: Pressure will decrease to the point of the compressor not supplying air (Unloaded with the compressor running). The two selected lead air compressors start at 98 psig. Plausible in that the normal pressure control band is 100-105 psig. The header will be at this pressure but the discharge of the G Compressor will be less than 100 psig.

C INCORRECT: Plausible if the candidate doesnt know what the Bypass Control Valve does. IF he/she believes the valve bypasses the normal pressure control. 120 psig is a recognizable value in that it is the rated pressure of G Control Air Compressor.

D INCORRECT: Compressor discharge pressure lowers. Plausible if the candidate doesnt know what the Bypass Control Valve does. IF he/she believes the valve bypasses the normal pressure control. Compressor Relief Valve setpoint is 132 psig.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

K/A asks for effect of a malfunction of a control air system valve. Question asks about the effect of failure of the Bypass Control Valve on the G Air Compressor and Control Air System.

Question Cognitive Level:

Answering the question involves the multi-part mental process of assembling, sorting, or integrating the parts, which also requires the candidate to predict an outcome from the valves failure.

Technical Reference(s): OPL1 71.054, Rev 15 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.9 (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

3. Control Air System Component Description

a. Four Rotary Screw Air Compressors A-D (2-stage Ingersoll-Rand H 125W and rotary screw type) are located El 565, U-i TP-28 Turbine Building. DCN 66433 (1) Suppy air to the control air receivers at 524 scfm each at a normal operating pressure of 94 -

108 psig.

(2) 480V. 60 Hz, 3-phase, drive motors (c) The primary controller normally controls the loading sequence for each control air compressor A (B, C, D) through the SEQUENCE SELECTOR svntch positions (automatic control positions 1, 2, 3, and 5)

(d) The controller contains the logic to load and unload the compressors automatically according to control air header pressure.

i. The compressors (2) in LEAD position has TP-5 a pressure range of 98 to 108 psig ii. As air pressure lowers to 98 psig, the compressors will go to full load.

iii. As air pressure rises to 108 psig, the first compressors wHI go to unload.

iv. First LAG compressor operating range is 96 to 106 psig.

v. Second LAG compressor operating range is 94 to 104 psig.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet (c) Relief valves on the compressors discharge set at 132 psig protects the compressor and piping.

G Air Compressor centrifugal type, two stage (a) Located 5651 EL Turbine Bldg., Unit I end.

Control Air Compressor G is the primary control air compressor and provides most of the control air needed for normal plant operation.

(b) Rated at 1445 SCFM @ 120 psig.

OPL171.054 Revision 15 Page 14 of 69

i. In the UNLOAD position, the compressor will run but not supply compressed air. The Compressor Inlet Flow Valve is throttled open and the Compressor Bypass Control Valve opens fully.

(a) Air is drawn through the Inlet Filter and through the Inlet Valve to the first stage impeller of the compressor.

(b) The compressed air discharged from the first stage impeller passes through internally finned copper tubes inside the intercooler, and then through a moisture separator.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QPL171.054 Revision 15 Page 17 of 69 (c) The second stage impeller takes suction from the intercooler, raises the pressure, and discharges to the after-cooler (similar to intercooler). Moisture is again removed from the compressed air by a moisture separator.

(d) The compressed air is then passed to the Control Air header viith some of the air being recirculated through the silencer via the bypass valve.

(e) Both the Inlet Valve and the Bypass Valve are positioned by the microcontroller to maintain the compressor discharge air at the desired pressure ( 100-1 05 psig).

(3) Control air receiver pressures should be between in pipe corrosion which 90 and 105 psig. G air compressor will normally led to failure of maintain receiver pressure >100 psig. When G air components.

compressor is not operating, then the primary/backup controllers will maintain 90 to 101 psig.

(4) Relief valves on the receivers set at 115 psig.

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet BEN Loss of Control and Service Air 0-AOl-32-1 Unit 0 Compressors Rev. 0039 Page 5 of 34 3.0 AUTOMATIC ACTIONS

• Service Air crosstie to Control Air valve, 0-FCV-33-1, opens at control air header pressure less than or equal to 85 psig.

• Control Air Compressors A,B,C,D will start as their on-line pressure setpoints are reached.

• The Emergency Control Say Air Compressor will start at Control Air Header pressure less than or equal to 73 psig.

• Unit 2 to Unit 3 Control Air Crosstie, 2-PCV-032-3901. will close when Control Air Header pressure reaches equal to or less than to 65 psig at the valve.

• Unit 1 to Unit 2 Control Air Crosstie, 1-PCV-032-3901, will close \hen Control Air Header pressure reaches equal to or less than to 65 psig at the valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 300000 Instrument Air System (lAS)

Tier # 2 K6.12 (IOCFR55.4t7)

Knowledge of the effect that a loss or malfunction of the following Group # 1 will have on the INSTRUMENT AIR SYSTEM:

K/A # 300000K6.12 Breakers, relays and disconnects Importance Rating 2.9 Proposed Question: # 51 Control Air Compressors A AND C are in service. A momentary loss of power to 480V Shutdown Board 1 B occurs. Three seconds later, normal voltage is restored.

Which ONE of the following describes the impact of this board loss on the Air System?

Control Air Compressor _(1)_ will trip AND _(2)_ automatically re-start when normal voltage is restored.

A. (1)A (2) will B. (1)C (2) will C. (1)A (2) will NOT D. (1)C (2) will NOT Proposed Answer: C Explanation A INCORRECT: A compressor is powered from 480V SD Bd 1 B, and will (Optional): therefore trip. The compressor will not auto start when normal voltage is restored. Plausible in that Control Air Compressor G does restart if voltage restored within 4 seconds.

B INCORRECT: C is powered from 480v Common Bd 1, which is not affected by this event. Plausible in that candidates could confuse 480V SD Bd 1 B which does supply A with 480 V Common Bd 1 which does not. If C power supply had been momentarily interrupted, the second part would NOT be true with voltage restored within 4 seconds.

C CORRECT: A compressor is powered from 480V SD Bd 1 B, which is affected by this event. It does NOT have auto restart capability for 4 sec power loss, like Control Air Compressor G.

D INCORRECT: C is powered from 480v Common Bd 1, which is not affected by this event. The G compressor power loss logic is set © 4 seconds on a loss of 480V RMOV Bd 2A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The effect of a breaker failure resulting in momentary loss of 480V Shutdown Board 1 B to the instrument air system (Control Air at BFN) agrees with the stated K/A.

Question Cognitive Level:

This question is high comprehension because the examinee must evaluate the situation and predict the effect on the instrument/control air system. This involves a multi-part mental process of assembling, sorting, and integrating the parts of the system.

Technical Reference(s): OPLI 71 .054 Rev 15 (Attach if not previously provided) 0-01-32 Rev 127 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.1 (As available)

Question Source: Bank #

Modified Bank # BEN 0801 #52 (Note changes or attach parent)

New Question History: Last NRC Exam Browns Ferry 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71 .054 r15

3. Control Air System Component Description

a. Four Reciprocating Air Compressori A-D (2-stage, double acting, Y-type) are located El 565, u-i Turbine Building.

(1) Supply air to the control air receivers a(610 scfm each ata normal operating pressure of 90- 101 psi g.

S0v60 Hi, a: ai:clhe mote (3) Power supplies A from 480V Shutdown Board lB DCN 17780 D from 480V Shutdown Board 2A B from 480V Common Board I C from 480V Common Board 1 (a) Control air compressors which are powered Obj. V.B1.

from the 480 VAt shutdown boards are Qbj. V.C1.

tripped automatically due to:

i. under voltage oh the shutdown boord.

ii. load shed logic during an accident signal concurrent with a loss of offsite power.

NOTE: The compressors must be restarted manually after power is restored to the board.

(b) units powered from c6mmon beards thiso trip due to under voltage.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet U

iv. The primary controller power is auctioneered from one of three sources:

• 480VAC Shutdown Board IA, Same power supply as

• 48OVAC Shutdown Board 2A for air compressors

• 48OVAC Common Board 1 (c) Each power supply has a 48OVAC to 12OVAC 0-45E76ti-5 transformer.

(d) The backup controller 0-PIC-032-0002 loads TP-41 and unloads the compressors at the same control air header pressure setpoints as the primary controller.

(e) When the backup is in control compressor A Independent of selector will run at full load, B, C, & D will load and switch position unload at 1 .5 psig increments, in alphabetical order as pressure falls and rises.

(1) The backup controller is powered from 250 VDC control power on 480 VAC Common Board I

e. G Air Compressor centrifugal type, two stage (1) Located 565 EL Turbine Bldg., Unit 1 end.

Control Air Compressor G is the primary control air compressor and provides most of the control air needed for normal plant operation.

(2) Rated at 1445 SCPM © 120 psig.

(3) Power Supply (a) 4 RV Shutdown Board B supplies power to the compressor motor.

(b) 480 V RMOV Bd. 2A SUpplies the follovinj:

• Pre lube pump .

• Oil reservoir heater

• Cooingwater pumps

• Panel(s) control power

• Auto Restart circuit (c) Except for short power interruptions on the 480v RMOV Bd, Loss of either of these two power supplies will result in a shutdown of the G air compressor.

(d) With the G air compressor AUTO START DCNF41321A selector switch in ON the compressorwill Power interruptions automatically restart if there is a momentary 4 seconds will lock out interruption of power (c 4 seconds) of the the Auto Restart circuit 4SOv RMOV board 2A. (see 0-01-32) and trip the compressor.

i. This feature was designed to maintain The feed from 4KV Control Air Compressor G operation during Shutdown Board B to board transfers and momentary the compressor motor is interruptions in power involving 480V not affected by the auto RMOV Board 2A. restart circuit.

ii. If power is NOT restored within 4 seconds, TP-7 the compressor will trip and must be manually restarted when power is restored.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

3. Component Description Obj. V.E.6

a. Compressors E and F (EL 565, U-S Turbine Building) Cbj. V.D4 are designated for service air.

b. The F air compressor is rated for approximately 630 SCEM © 105 psig, centrifugal type, 2 stages

c. The Jowiii supply for both compressors is 48OVAC Common Board 3.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Control Air System 0-01-32 UnitO Rev.0127 Page 9 of 113 3.0 PRECAUTIONS AND LIMITATIONS (continued)

E. Control Air Compressor G will automatically trip and remain tripped on any of the following conditions:

1. Vibration high, Stage I - 1.00 mu

2. \Jibration high, Stage 2 0.94 mil

3. Lube Oil Pressure low 16 psig

4. Lube Oil Temperature high - 125F

5. Lube Oil Temperature low 65F -

6. Air Temperature high, Stage I -

l25F

7. Discharge Air Temperature high - l25>F

8. Seal Air Pressure low 6 psig F. A loss of power to 4KV Shutdown Board B OR a sustained loss of power (greater than 4 seconds) to 480V RMOV Board 2A will result in a trip of Control Air Compressor G.

G. Control Air Compressor G has an auto restart circuit which will restart the compressor after a momentary power loss (up to 4 seconds) from 480V RMOV Board 2A. This feature was designed to maintain Control Air Compressor G operation during board transfers and momentary interruptions in power involving 480V RMOV Board 2A. The restart circuit is in place when COMPR G AUTO-RESTART ON-OFF SELECTOR switch, 0-HS-032-3087, is in the ON position AND the compressor is running. The auto restart circuit viill reset automatically after each restart attempt. thus enabling multiple restart attempts.

H. During a surge condition, Control Air Compressor G will alarm and automatically unload. The compressor will automatically reload after 6 seconds for the first 3 surges in ten minutes. If a fourth surge occurs within the 10 minute period, the compressor will remain unloaded until being acknowledged at the Microcontroller. Section 6.1 provides additional instruction on compressor surge.

Control Air Compressor G shall NOT be manually restarted until it has come to a complete rest.

J. When blowing down Control Air Compressor G raw cooling water strainer open RCW STRAINER BLOWDOWN VLV, 0-DRV-024-1 510 only halfway (45) open4pER 20749]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRQ 300000 Instrument Air Tier # 2 K2.O1 (100FR 5541.7)

Knowledge of electrical power supplies to the following: Group # 1

• Instrument air compressor K/A # 300000K2.O1 Importance Rating 2.8 Proposed Question: # 52 All Units are operating at 100% Reactor Power, when a momentary undervoltage condition occurs on 480V RMOV Board 2A. Five seconds later, normal voltage is restored.

Which ONE of the following describes the impact of this board loss on the Air System?

A. Control Air Compressors B AND C will trip. BOTH will need to be re-started locally.

B. Service Air Compressors E AND F will trip. BOTH will need to be re-started locally.

C. Control Air Compressor A will trip, AND will automatically re-start when normal voltage is sensed.

D. Control Air Compressor G will trip, AND will NOT automatically re-start when normal voltage is sensed.

Proposed Answer: D Explanation A INCORRECT: B & C compressors are powered from 480V Common Bd 1, (Optional): which is not affected by this event. Air Compressor G is powered from 4kV SD Bd B I 480V RMOV Bd 2A, which is affected by this event.

B INCORRECT: E & F compressors are powered from 480V Common Bd 3, which is not affected by this event. Air Compressor G is powered from 4kV SD Bd B I 480V RMOV Bd 2A, which is affected by this event.

C INCORRECT: A compressor is powered from 480V SD Bd 1 B, which is not affected by this event. Air Compressor G is powered from 4kV SD Bd B I 480V RMOV Bd 2A, which is affected by this event.

D CORRECT: Air Compressor G is powered from 4kV SD Bd B I 480V RMOV Bd 2A, which is affected by this event. The G compressor power loss logic is set @ 4 seconds on a loss of 480V RMOV Bd 2A. Transfer of the RMOV Bd is a Manual action, thus > 4 seconds will have transpired prior to the transfer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL17I.054 Rev 13 (Attach if not previously provided)

O-Ol-57B Rev 181 (Including version I revision number)

Proposed references to be provided to applicants during examination: NONE Learning Objective: VR1 (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Reviewed Rev 181 of 0-Ol-57B. The latest revision of this procedure has no impact on the question.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 400000 Component Cooling Water Tier# 2 A1.O1 (IOCFR 55.41.5)

Ability to predict and/or monitor changes in parameters associated Group# 1 with operating the COMPONENT COOLING WATER SYSTEM controls including:

K/A# 400000A1.O1

• __CCW_flo w_rate Importance Rating 2.8 Proposed Question: # 52 Which ONE of the following completes the statement below?

The Unit 2 Reactor Building Closed Cooling Water (RBCCW) Temperature Controller, 2-TIC-24-80, is located in Unit 2 Reactor Building at(1)_.

If the controller is placed in AUTO with the indications shown below, the Temperature Control Valve will modulate to a more (2)

A. (1) Panel 2-25-1 96, Elevation 565 (2) closed position B. (1) RBCCW Heat Exchanger area, Elevation 593 (2) close position C. (1) Panel 2-25-1 96, Elevation 565 (2) open position D. (1) RBCCW Heat Exchanger area, Elevation 593 (2) open position Proposed Answer: A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Explanation A CORRECT: Part 1 correct RBCCW Temp Controller, 2-TIC-24-80, is (Optional): located in Unit 2 Reactor Building at Panel 2-25-196, Elevation 565. Part 2 correct with the RED indicator (Set Point) higher than the BLACK needle, indicates that actual temperature is cooler than desired. The TCV will modulate CLOSED.

B INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A.

C INCORRECT: Part 1 correct See Explanation A. Part 2 incorrect See Explanation D.

D INCORRECT: Part 1 incorrect Plausible in that several RCW valves associated with RBCCW are located at the RBCCW Heat Exchanger area, Reactor Building Elevation 593. Part 2 incorrect Plausibility based on misconception that with the feedback signal less than the control set point that the TCV would modulate Open to remove the deviation or that the controller is bypassing flow rather than controlling cooling water flow through the heat exchanger.

KA Justification:

The KA is met because the question tests the ability to predict and monitor changes in CCW Heat Exchanger flow in response to operating CCW Temperature control valve from Auto to Manual with a deviation between set point and feedback signal.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 2-01-24 Rev 77 (Attach if not previously provided)

OPL171.048 Rev 14 Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank #

Mod ified Bank # BEN 0801 #53 (Note changes or attach parent)

Npw Question History: Last NRC Exam BEN 0801 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of eve,y question.)

Question Cognitive Level: Memory or Eundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Comments:

BEN Raw Cooling Water System 2-01-24 Unit 2 Rev. 0077 Page 58 of 58 Illustration I (Page 4 of 4)

Operation of 2 TIC 24 80(85) and I TIC 24 90 for RBCCW Temperature Control 3.0 RBCCW TEMPERATURE CONTROLLERS NULLMATIC MIP -

CONTROLLER NOTE Setp oint 27% scale = 8OF Maximum Allowed Setpoint: 36% scale = 95°F.

Range: 40°F to 190°F

\/alve is fully closed at 0% and fully open at 100%.

Re lricllcaI1n Por Sensei ThmpeWr S4pIit cae  % Thumbwt&

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 171 048 Revision 14 Page 14 of 35 INSTRUCTOR NOTES (2) Output from the temperature sensor is changed by the temperature modifier from a milivolt signal to an air signal which is proportional to the 3-15 psig typical temperature.

(3) The temperature indicating Some electronic TICs controller (TIC) compares the process the milivolt temperature from the TM to a signal from the desired temperature (or thermocouple directly set oint.) or from a TM.

(4) The TIC sends a signal to the Electronic TICs send valve positioner and TCV to an electrical output throttle open the valve if signal to a TM which temperature is higher than the develops the air signal setpoint, or throttle it closed if to the valve positioner.

below the desired temperature.

h. most TCVs are Control Air operated valves which throttle flow through individual loads for temperature control.

c. Nullmatic Temperature Controllers TP-5 (1) This is the type used on RBCCVV heat exchangers.

(2) The BLACK (center) pointer always indicates RBCCW temp.

at outlet of RB COW Heat See 0124 Illustration I Exchanger. for detailed information (3) The RED (peripheral) pointer is controlled by the Select Switch.

In VAL\/E position, it senses regulating air pressure to control valve air operator and displays in % valve is closed. In REG position, it senses controlling press. from air regulator (controlled by thumb wheel) and displays desired control point in same units as the BLACK pointer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Building Closed Cooling Water 2-01-70 Unit 2 System Rev. 0061 Page 29 of 63 8.3 Placing Spare Heat Exchanger in Service (continued) l4] VENT the RCW System side of the spare RSCCW Heat Exchanger UNTIL a solid stream of water is discharged, THEN CLOSE, using the following: D

. RBCCW CLR C RCW VENT, 1-VT\/-024-1089 (south end RBCCW Heat Exchanger area. El 596). 0

. RBCCW CLR C RCW VENT. 1-VTV-024-1090 (north end RBCCW Heat Exchanger area, El 593). 0

[5] VERIFY OPEN the following (south end Drywell equipment hatch area, El 565):

. TCV-24-90A INLET, 1-SHV-024-0724C. 0

. TCV-24-90B INLET, 1-SHV-024-0722C. 0

. TCV-24-90A OUTLET, 1-SHV-024-0725C. 0

. TCV-24-90B OUTLET, i-SHV-024-0723C. 0 1:6] DETERMINE position of temperature control valve on RBCCW Heat Exchanger to be taken out of service, RBCCW HX A(B)

TEMP CONT, 2-TIC-24-80(85), located at Panel 2-25-196, El 565. 0 1:7) PLACE RBCCW SECTIONALIZING VLV TRANSFER.

2-XS-70-48, in EMERG (480V Reactor MOV Board 2B.

compartment 5A). 0 NOTE The following action will lower RCW System pressure.

[8] PERFORM the following at Panel 1-25-1 96, El 565 and REFER TO Illustration 1:

VERIFY in service, RBCCW SPARE HX TEMP CONT.

1 -TIC-24-90.

• PLACE in MANUAL AND OPEN temperature control valve to same position as temperature control valve on RBCCW Heater Exchanger to be taken out of service. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Reactor Building Closed Cooling Water 2-01-70 Unit 2 System Rev, 0061 Page 28 of 63 8.3 Placing Spare Heat Exchanger in Service f 1) VERIFY the following initial conditions are satisfied:

. Unit 2 RBCCW System is in operation.

. Raw Cooling Water available to supply Spare RBCCW Heat Exchanger. REFER TO 2-01-24.

. Spare RBCCW Heat Exchanger available for use on Unit 2. D (1.1) [NRC/Cl WHEN the RCW or EECW supplied to any RBCCW heat exchanger is put into service or taken out of service, THEN NOTIFY Chemistry Shift Supervisor so any required sampling can be initiated or stopped. [NRC LER 259/88010! D NOTE All operations are performed locally unless noted otherwise.

f2] NOTIFY Unit 1 and Unit 3 that Unit 2 will be placing the spare heat exchanger in operation on Unit 2.

CAUTION Filling and venting the Raw Cooling Water side of the spare heat exchanger may cause a lowering in RCVV System pressure. resulting in an ESF actuation. Slow and cautious performance of any actions that may cause system pressure to lower will minimize this problem.

{3] VERIFY OPEN, RCW TO RBCCW HTX C. 1-SHV-024-07200 (north end RBCCW Heat Exchanger area, El 593, Chain operated valve in overhead). D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Reactor Building Closed Cooling Water 2-01-70 Unit 2 System Rev. 0061 Page 29 of 63 8.3 Placing Spare Heat Exchanger in Service (continued)

[4] VENT the RCW System side of the spare RBCCW Heat Exchanger UNTIL a solid stream of water is discharged, THEN CLOSE. using The following: 0

• RBCCW CLR C RCW VENT, 1-/TV-024-1089 (south end RBCCW Heat Exchanger area. El 596). 0

• RBCCW CLR C ROW VENT, 1-VTV-024-1 090 (north end RBCCW Heat Exchanger area. El 593). C

[5] VERiFY OPEN the following (south end Diy.vell equipment hatch area, El 565):

• TCV-24-90A INLET, 1-SHV-024-0724C. C

• TCV-24-90B INLET, 1-SRV-024-0722C. C

• TC\J-24-90A OUTLET, 1-S HV-024-0725C. C

• TC\J-24-90B OUTLET, 1-SHV-024-0723C. C

[6] DETERMINE position of temperature control valve on RBCCVJ Heat Exchanger to be taken out of service. RBCCVV HX A(S)

TEMP CONT, 2-TIC-24-80(85), located at Panel 2-25-196, El 565. C

[7] PLACE RBCCW SECTIONALIZING VLV TRANSFER.

2-XS-70-48, in EMERG (480V Reactor MOV Board 25, compartment 5A). C NOTE Z following action will lower RCW System pressure.

[8] PERFORM the following at Panel 1-25-196, El 566 and REFER TO Illustration 1:

• VERIFY in service, RBCCW SPARE HX TEMP CONT.

1 -TIC-24-90. C

• PLACE in MANUAL AND OPEN temperature control valve to same position as temperature control valve on RBCCW Heater Exchanger to be taken out of service. C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

53. At 10:00 a.m., the Unit 2 RBCCW Temperature Control Valve (TCV), (2-TIC-24-80(85)) was place in Manual as follows:

• REG was selected

• MAN was selected Before transferring from Manual back to Automatic, in order to NULL the controller, the RBAUO i required to place the (1)AND adjust the thumbwheel until the RED pointer lines up with the BLACK pointer.

At 11:00 a.m., the controller was transferred to auto. If the RBAUO observes the following indicati after the controller was transferred back to auto, this means that the TCV will modulate to a more (2)

After Transfer:

VALVE OPENS A. (1) Transfer Switch to SEAL.

(2) open position.

B. (1) Selector Switch to VALVE.

(2) closed position.

C. (1) Transfer Switch to SEAL.

(2) closed position.

D. (1) Selector Switch to VALVE.

(2) open position.

ANSWER: A BROWNS FERRY 0801

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 400000 Component Cooling Water A1.04 (100FR55.41.5) Tier # 2 Ability to predict and / or monitor changes in parameters associated Group # 1 with operating the CCWS controls including:

K/A # 400000A1 .04 Surge Tank Level Importance Rating 2.8 Proposed Question: # 53 Unit 2 RBCCW Heat Exchanger 2A is being filled and vented per 2-01-70, Reactor Building Closed Cooling Water System.

Which ONE of the following completes the statement?

While filling the Heat Exchanger, RBCCW Surge Tank Level will lower until RBCCW SYS SURGE TANK FILL VALVE, 2-FCV-70-1, A. is manually opened from the Control Room B. is manually opened locally at the Surge Tank C. automatically opens at 4 inches below the Surge Tank centerline D. automatically opens at 4 inches above the Surge Tank centerline Proposed Answer: A Explanation A CORRECT: RBCCW SYS SURGE TANK FILL VALVE, 2-FCV-70-1, is (Optional): operated remotely from Control Room Panel 2-9-4.

B INCORRECT: Plausible in that manual BYPASS VLV, 2-FCV-70-1, is LOCALLY operated at the surge tank.

C INCORRECT: Plausible in that it is logical to have automatic make up capability to the RBCCW Surge Tank and 4 inches below the Surge Tank centerline is the set point for the Surge Tank Level Low Alarm. Additionally other plant head tanks automatically fill on low level. Examples: Demin Water Head Tank / PSC Surge Tank.

D INCORRECT: Plausible in that it is logical to have automatic make up capability to the RBCCW Surge Tank and 4 inches above the Surge Tank centerline is a recognizable value as the set point for the Surge Tank Level High Alarm. Additionally other plant head tanks automatically fill on low level. Examples: Demin Water Head Tank I PSC Surge Tank.

KA Justification:

The KA is met because the question tests candidates ability to predict and monitor changes in Surge Tank Level associated with operating RBCCW controls to fill a Heat Exchanger.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-70 Rev. 61 (Attach if not previously provided) 2-ARP-9-4C Rev. 30 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank #

Mcadified Bank# (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet F BFN Unit 2 Reactor Building Closed Cooling Water 2-017O System Rev. 0061 Page 40 of 63 8.7 Filling And Venting Of An Individual Heat Exchanger.

NOTE Filling of RBCCW SYS SURGE TANK may be performed from Control Room (Step 8.7[2})

or locally at surge tank (Step 8.7[3]).

CAUTION If RBCCW SURGE TANK HIGH LEV EL (XA-55-4C, window 6) Annunciator alarms, and the fill valve is open, it should be closed immediately to prevent tank overflow.

[1] STATION personnel to monitor RBCCW surge tank level as the heat exchanger is filled. D 12] [F System Fill is to he performed from the Control Room THEN (Otherwise NiA)

[2.1] ESTABLISH direct communications between the Personnel at the RBCCW System Surge Tank. the heat exchanger to be filled and vented and the Control Room Operator.

[22] OPEN RBCCW SYS SURGE TANK FILL VALVE, 2-FCV-70-1, using 2-HS-70-1 (Panel 2-9-4). 0

[2.3) FILL system until RBCCW Surge Tank level is normal (4 inches below tank centerline to 4 inches above tank centerline), THEN C

[2.3.1] CLOSE RBCCW SYS SURGE TANK FILL \ALVE, 2-FCV-70-1, (Panel 2-9-4).

[2.3.2) MAINTAIN this range durng fill and vent.

[3] IF System Fill is to be performed locally at the RBCCW Surge Tank. THEN (Otherwise N/A)

[3.1] ESTABLISH direct communications between the Personnel at the RBCCW System Surge Tank, the heat exchanger to be filled and vented and the Control Room Operator. C

[32] OPEN FCV-70-i BYPASS VLV 2-BYV-002-1369 (locally at surge tank).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-4 2-AR P-9-4C Unit 2 2-XA-55-4C Rev. 0030 Page 20 of 44 Sensor/Trip Point:

RBCCW 2-LS-70-2B 4 inches below center line of tank SURGE TANK LEVEL LOW 2-LA-70-2B (Page 1 of i Sensor On the RBCCW surge tank in the MG Set Room. El 639 Location:

Probable A. Normal leakage.

Cause: B. Drain valves open.

C. Abnormal leakage.

Automatic None Action:

Operator A. ADD water to the RBCCW Surge Tank for approximately one minute Action: or until low level alarm resets using the following: C RBCCW SYS SURGE TANK F1LL VLV, 2-Fey-JO-I (Panel 2-9-4) OR FCV-70-1 BYPASS VLV, 2-HCV-2-1369 (locally.

B. IF alami does NOT resets THEN CHECK tank locally.

C. IF unable to maintain RBCCW Surge Tank level THEN REFER TO 2-AOl-70-1. C D. IF necessary to add water more than once per shift, THEN CHECK Diy.ve.ll floor drain system for excessive operation AND INSPECT system outside DrweFl for leakage.

References:

46N6l44 247E610701 45N6204 FSAR Sections 10.6.4 and 13.6.2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0030

. Pagel2of44 Sensor/Trip Point:

RBCCW 2-LS-70-2A 4 inches above center line of tank SIJRGE TANK LE\IEL HIGH 2-LA-70-2A (Page 1 of 2 Sensor RBCCW surge tank in the MG set room El 639.

Location:

Probable A, Makeup valve, 2-FCV-70-I, open.

Cause: B. Bypass valve 2-2-1369 leaking.

C. Leak into the system.

Automatic None Action:

Operator A. CHECK make-up valve 2-FCV-70-i, 2-HS-70-1, CLOSED on Action: Panel 2-9-4.

B. CHECK RBCCW system water leaving the RBCCW system heat exchangers is 100°F or less on 2-Ti-70-3, Panel 2-9-4. C C. DISPATCH personnel to verify high level and to ensure bypass valve. 2-HCV-2-1369, for 2-FCV-70-i is CLOSED. OBSERVE sight glass level. C D. OPEN surge tank drain valve, 2-70-609. CLOSE valve when desired level is obtained.

E. IF level continues to rise THEN REQUEST Chemistry to pull and analyze a sample for total gamma activity and attempt to qualify source of leak. C F. CHECK activity reading on 2-RM-90-131 (2-RR-90-134 Ch 3). C Continued on Next Pncje

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 9.20 1-ARP-9..20B Unit I 1-XA-55-20B Rev. 0028 Page 17 of 39 SensorTrip Point:

DEMIN WATER HEAD TANK LEVEL ABN High level:

0-LA-2-159 LS-2-i5SA 9 ft (Elevation 72T25) rising Low level:

LS-2-159C 7ft4 in (Elevation 725.58.lIowerin (Page 1 of 1)

Sensor Reactor Building Roof Location:

Probable A. Excessive demineralized water usage.

Cause: B. System leakage.

C. Level switch malfunction.

D. Demineralized water transfer pump malfunction.

Automatic A. Both demineralized water transfer pumps secure on high level.

Action: B. Second demineralized water transfer pumps stails on low level.

Operator A. CHECK Demineralized Water Transfer Pumps 0-HS 154A on Action: Panel 1-9-22. and 0-HS-2-l55A on Panel 1-9-20, in AUTO and operating status lights illuminated.

C B. CHECK 0-HS-2-159, DEMIN WTR HD TNKS 1 INLET VLV. on Panel 1-9-22, in OPEN and light illuminated.

C C. DISPATCH personnel to the demin water head tank to determine condition.

C D. iF level is high, THEN VERIFY both Demineralized Water Transfer Pumps are off. C E. IF level is low, THEN PERFORM the following:

• CHECK demineralized water storage tank level, on Panel 1-9-20. greater than 26 ft with Ll-2-153 and START both transfer pumps. C

• OPEN FCV-2-159 BYPASS VLV. 0-BYV-002-0926. to restore level (Elev. 565, T-4 M-line).

• CHECK system for leaks.

C

References:

047W49 13 [45E620l 22

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BPN Panel 9-3 1-ARP-9-3A Unit I XA-55.3A Rev. 0040 Page 39 of 52 SensorlTrip Point:

PSC HEAD TANK BETWEEN LEVEL LOW 1-LS-075-00780 644-il & 645-2 1/4 1 -LA-75-79 (Page 1 of 1 Sensor Rx Bldg, EL 639, R-i T-Line.

Location:

Probable A. Both pumps NOT running.

Cause: 1. Level switch malfunctioned.

2 Thermal overloads NOT reset, 480V Reactor MOV Board IC and 18. compartments I IC and 7A respectively.

3. FCV-75-57 and -58 closed (PCIS Group II isolation).

& Strainer AP push buttons NOT reset.

B. Both pumps running.

Pump discharge pressure < 60 psig. A minimum pressure of 55 psig is required to reach EL 645.

A Automatic Low level switch starts both pumps.

Action:

Operator A. VERIFY both pumps are running. D Action: 8. VERIFY power available to pumps.

C CHECK PSC FIJMP SUCTION INBD and OUTBD ISOL VALVES.

i-FCV-75-57 and 58 open. C D. IF the alarm does NOT reset within a few minutes. THEN DISPATCH personnel to CHECK pumps locally. C E. IF the PSC Head Tank Pumps will NOT maintain the RHR and Core Spray Systems charged above TRM Limits, THEN LINE UP the condensate transfer system to each loop. REFER TO 1-01-75. C F. REFER TO Tech Spec 3.5.1. 3.5.2.TRM Sections 3.3.3.1 & 3.5.4. C

References:

147W6l075I 145E7513 and-5 l-45E620-3 TRM 3.5.4 Tech Spec 3.5.1,3.5.2 TRM 3.33.l, 3.5.4

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 201001 CRD Hydraulic K5.05 (100FR 5541.5)

Tier # 2 Knowledge of the operational implications of the Group # 2 following concepts as they apply to CONTROL ROD DRIVE K/A # 201002K5.05 HYDRAULIC SYSTEM:

. lndicaUons of pump runout: Plant-Specific Importance Rating 2.7 Proposed Question: # 54 J Unit 1 Control Rod Drive System has ruptured on the Charging Water Header resulting in CRD Pump 1A operating at pump runout.

Which ONE of the following completes the statement?

This condition is indicated by CRD Pump IA motor amps _(1)_ than normal AND CRD Flow Control Valve FULL (2).

A. (1) LOWER (2) OPEN B. (I) HIGHER (2) OPEN C. (1) LOWER (2) CLOSED D. (1) HIGHER (2) CLOSED Proposed Answer: D Explanation A INCORRECT: Part 1 Correct Plausibility based on misconception that (Optional): pumping against backpressure of atmospheric as opposed to above Reactor Pressure would result in lower motor amps. Part 2 Correct Plausible in that if CRD flow elements providing feedback to CRD FCV were downstream of where Charging Water Header ties in, the TCV would see low flow and go full open.

B INCORRECT: Part 1 Correct See Explanation D. Part 2 Incorrect See Explanation A.

C INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

D CORRECT: Part 1 correct The increase in Pump flow associated with going from normal flow to runout conditions would result in CRD Pump 1A motor amps higher than normal. Part 2 correct CRD flow elements providing feedback to CRD FCV are upstream of where Charging Water Header ties in resulting in high flow sensed by the controller. The TCV would go full closed in response to the high flow condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of indication and operational implications of CRD Pump 1A at runout due to a break in the system on the Charging Water Header.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL17I.005 Rev. 17 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank # I Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17l .005 Revision 17 Page 19of79 hISTRuCTOR NOTES ii. The CRD pump will try to recharge all the accUmulators at once.

Flow through the charging header will cause the flow control valves to close.

in. To prevent pump runout and probable tripping of Q: What is a good the pump motor on indication of pump overcurrent, a runout?

restricting orifice is provided to limit the maximum rate of A: Indication of recharging to 179 gpm. pump runout is high (Maximum flow is with cunent on the the reactor at CRDH pumps atmospheric pressure).

SER-3-05 iv. A thnttle valve downstream of the restricting orifice is provided to l:rovide additional throttling if requirerL

v. The accumulators SER 3.05.

cannot be recharged until the scram is reset (with the scram inlet and Charging pressure outlet valves closed) will he due to drive seal approximately equal leakage being greater to reactor pressure than pump capacity.

(c) Charging water pressure is independent of reactor vessel pressure. and is set by manually positioning the pump discharge throttling valve between 1475-1500 psig per 01-85.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 201003 Control Rod and Drive Mechanism K1.O1 (IOCFR 55.41.2 to 41.9)

Tier # 2 Knowledge of the physical connections and/or cause effect Group # 2 relationships between CONTROL ROD AND DRIVE MECHANISM and the following: K/A # 201 003K1 .01

. Control rod drive hydraulic system Importance Rating 3.2 Proposed Question: # 55 During a UNIT 1 startup, a control rod drive mechanism is difficult to withdraw (will not move after several attempts to notch the rod out) and is stuck at position 00.

HCU hydraulic lines were vented and the problem is NOT believed to be air in the hydraulic system.

Which ONE of the actions, listed below, is the correct 1-01-85, Control Rod Drive System set of actions to be taken to address difficult to withdraw control rods, and to get the control rod to move?

GOTO A. ROD IN, then ROD OUT NOTCH with the CRD CONTROL SWITCH, release if rod moves B. ROD OUT NOTCH with the CRD CONTROL SWITCH, then NOTCH OVERRIDE with the CRD NOTCH OVERRIDE SWITCH, release switches if rod moves C. EMERGENCY IN with the CRD NOTCH OVERRIDE SWITCH, then simultaneously place the CRD CONTROL SWITCH in ROD OUT NOTCH, release switches if rod moves D. EMERGENCY IN, then NOTCH OVERRIDE with the CRD NOTCH OVERRIDE SWITCH, and then simultaneously place CRD CONTROL SWITCH in ROD OUT NOTCH, release switches if rod moves Proposed Answer: D Explanation A INCORRECT: This method may be used to vent some air from the CRDH (Optional): lines but stem gives NOT believed to be air. RMCS settle time will be enforced between in and out signals. This method does give a withdrawal signal. Candidate may believe this will unstick the rod because it does give a withdrawal signal.

B INCORRECT: Would still ONLY get a single rod out notch signal. IF rod wouldnt move with single rod out notch signal, it wont move now. IF went to notch override first, then rod out, at least you would get a continuous withdrawal signal and vent any air from the withdrawal header/lines.

Candidate misconception that notch override is giving a signal continuous withdrawal signal in this condition.

C INCORRECT: Drives rod in ONLY. Rod wont move out. It already has a continuous insert signal. May chose because of rod out notch signal.

Candidate confusion that this is giving a continuous withdrawal signal.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D CORRECT: This is procedurally correct per 1-01-85. The double clutch method is described.

KA Justification:

Question asks if relationship is understood between CRDH and CRDM and control room controls. It tests knowledge of double clutching a stuck rod to get it unstuck from the full in position. Candidate must understand RMCS, CRDH, and CRDM systems to determine how controls may be operated to unstick the control rod.

Question Cognitive Level:

This question has high cognitive value because; the candidate must recognize interaction between systems, including consequences and implications.

Technical Reference(s): 1-01-85 Rev 23 (Attach if not previously provided)

OPL171.005 Rev 17 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.26 (As available)

Question Source: FERMI 2 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Rod Drive System 1.01.85 Unit I Rev. 0023 Page 136 of 221 8.15 Control Rod Difficult to Withdraw

[1] VERIFY the control rod will NOT notch out and REFER Section 6.6. ci

[23 REVIEW all Precautions and Limitations in Section 3.0. 0 CAUTION

[NERIC] Never pull control rods except in a deliberate, carefully controlled manner, while closely monitoring the Reactor response. [NPO soER-96-OQ1

[33 [NRCIC] IF RWM is enforcing. THEN VERIFY RWM is operable and latched into the correct ROD GROUP. tNRC-R84-D2J C NOTES

1) Steps 8.15[4] through 8.15[6J should be used when the control rod is at position 00 while Step 8.15[7] should be used when the control rod is at or between positions 02 and 46.

2) Double clutching of a control rod at position 00 will place the rod at the overtravel in stop. independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet fingers from engaging the 00 notch.

3) Step 8.154] may be repeated as necessary until it is determined that this method will NOT free the control rod.

[4] IF the control rod problem is NOT believed to be air in the hydraulic system. THEN PERFORM the following to double clutch the control rod at position 00:

[4.13 PLACE AND HOLD CRD NOTCH OVERRIDE, 1-HS-85-47, in EMERG ROD IN. for several seconds.

[4.2] CHECK the control rod full in indication (double green dashes) on the Full Core Display for the associated control rod. C

[4.3] SIMULTANEOUSLY PLACE AND HOLD CRD NOTCH OVERRIDE. 1HS-85-47, in NOTCH OVERRIDE AND CRD CONTROL SWITCH, 1-HS-85-48, in ROD OUT NOTCH. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Rod Drive System 1.01-85 Unit I Rev. 0023 Page 137 of 221 8.15 Control Rod Difficult to Withdraw (continued) 4.4] WHEN EITHER of the following occur:

• Control rod begins to move. OR

• It is determined the rod \,ill NOT move. THEN RELEASE 1-HS-85-47 and 1-HS-85-48. D

[4.51 IF the control rod successfully notches out, THEN PROCEED to Section 6.6 and WITHDRAW the control rod to the appropriate position,

[4.6] IF desired, THEN REPEAT Steps 3.15[4.1) through 8.15[4.5) several times prior to raising drive water pressure in Step 8.15[5].

[5] IF double clutching the control rod was unsuccessful, THEN PERFORM the following to withdraw the control rod using elevated drive water pressure:

[5.1] RAISE drive water differential pressure to 300 psid as indicated on CRD DRR/E WTR HDR DP, 1-PDI-85-17A using CRD DRIVE WATER PRESS CONTROL VLV, 1 -HS-85-23A.

[5.2] PERFORM the following to double clutch the control rod at position 00 using elevated Control Rod Drive pressure:

[5.2.1) PLACE AND HOLD CRD NOTCH OVERRIDE, 1-HS-85-47, in EM ERG ROD IN, for several seconds. D

[5.2.2) CHECK the control rod full in indication (double green dashes) on the Full Core Display for the associated control rod. Q

[5.2.3] SIMULTANEOUSLY PLACE AND HOLD CRD NOTCH O\!ERRIDE, l-HS-85-47, in NOTCH OVERRIDE and CRD CONTROL SWITCH.

1-HS-85-48. in ROD OUT NOTCH.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 215001 Traversing In-core Probe Tier # 2 K4.O1 (IOCFR 55.41 .7)

Knowledge of TRAVERSING IN-CORE PROBE design feature(s) Group # 2 and/or interlocks which provide for the following:

K/A # 215001 K4.01 Primary containment isolation: Mark-I&ll(Not-BWRI)

Importance Rating 3.4 Proposed Question: # 56 Unit 1 is operating at 100% Reactor Power with the A Traversing In-Core Probe (TIP) inserted in the core. A transient occurs resulting in the following plant conditions:

• Reactor Level is (-) 20 inches

• Drywell pressure is 1 .5 psig Which ONE of the following completes the statement?

The A TIP will withdraw to the _(1)_ position AND the Ball Valve position will be _(2)_.

A. (1) PARKED (2) open B. (1)PARKED (2) closed C. (1) IN-SHIELD (2) open D. (1) IN-SHIELD (2) closed Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect The TIP is withdrawn to the in-shield. For (Optional): the ball valve to close, it must be in the in-shield position. Plausible in that there are TIP interlocks associated with the PARKED position. Part 2 incorrect, the Ball Valve will close. Plausible in that shear valve will not close.

B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

C INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A.

D CORRECT: Per 1-AOI-64-2E, on a Group 8 signal, an AUTO withdraw signal is actuated. The TIP is withdrawn to the in-shield position. Part 2 =

Once in the in shield position, the Ball Valve will automatically close

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of TIP design feature and interlocks which provide for Primary containment isolation.

Question Cognitive Level:

Candidate must recognize Reactor Level is less than the set point for a Group 8 isolation and predict the impact on the TIP System.

Technical Reference(s): OPL171.17 Rev 15, OPL171.023 Rev 6 (Attach if not previously provided) 1 -AOi-64-2E Rev 1 (Including version / revision number)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.023V.B.5 (As available)

Question Source: Bank# Hatch 09#12 Modified Bank#, (Note changes or attach parent)

Question History:

Last NRC Exam Hatch 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 )(

55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.017 Revision 15 Page 19 of 56 INSTRUCTOR NOTES

7. Group 8 Obj. V B.2 Obj. V.0.2 This group provides for isolation of the five Gives TIP auto Traversing Incore Probe (TIP) Guide Tubes via withdraw siqnal then solenoid operated hail valves. Signals which initiate doses the ball valve a Group 8 isolation are: Blue light on normal

• RPV Low Level (÷2 or Level 3)

• Op/well High Pressure (2.45 psig)

G. I nstru men tation

1. Sensors and logic are arranged such that no single Obj. V.5.3 failure will either initiate nor prevent an isolation. Qhj. V.0.3

2. The sensor arrangements used for the varous isolation signals are as follows:

a. RPV low level Obj. \&B.2 Obj. V.0.2

• Eight dp transmitters are used to produce low RPV level isolation signals. Four transmitters are used for the ÷2 (or Level 3) isolation:

the other four are used for the 122 (Level 1) isolation.

• The (÷2 or Level 3) transmitters al-c LIS 203A-D. while the (-122 or Level 1) transmitters are LlS-3-56A-D.

b. Main Steam Line Area High Temperature Obj. V 5.2 Gb]. V.0.2

• High temperature in the vicinity of the main 4 for each area total steam lines is detected by 16 bimetallic -l from each area in temperature switches located along the main Al, A2. 51,52 steam line between the drywall wall and the main turbine.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BEN Traversing Incore Probe Isolation 1-AOI-64-2e Unit I Rev. 0001 Page 3 of 7 1.0 PURPOSE This instruction provides syml:toms, automatic actions and operator actions for a Group 8. Travei-sing Incore Probe (TIP) Isolation and detection of a reactor coolant leak in a TIP guide tube.

2.0 SYMPTOMS NOTE A PCIS Group B isolation is initiated by either of the followinci:

• Reactor Vessel Water Level Low

• Dn,well High Pressure I) Any one or more of the following annunciators in alarm:

• RX VESSEL WTR LEVEL LOW HALE SCRAM (i-XA-55-4A. Window 2) in alarm (Group 8 Isolation).

• DRYWELL PRESSURE HIGH HALF SCRAM (I-XA-55-4A, Window 8)in alarm (Group 8 Isolation),

• AIR PARTICULATE MONITOR RADIATION HIGH I-RA-90-50A (i-XA-55-3A, Window 2) in alarm (indicative of TIP guide tube leak).

• RX BLDG AREA RADIATION HIGH i-RA-90-ID (l-XA-55-3A.

Window 22) in alarm (indicative of TIP guide tube leak).

3.0 AUTOMATIC ACTIONS i] IF a Group 8 isola.tion occurred, THEN the following are automatic actions:

• IF TIP probes are outside their shields, THEN TIP withdrawal initiated to IN-SHIELD position.

• TIP BaIl Valves receive a close signal, or close after TIP probes are withdrawn to their INSHIELD position.

• TIP Purge Valves closes (no indications provided).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17 1.023 Revision 6 Page 27 of 64 INSTRUCTOR NOTES (iv) For Automatic TIP operation from 3-D Monicore (and local ATCIJ), the TIP must be in the PARKED position prior to starting (Manually performed at NLIMAC unit in Control room)

(v) All TIP machines can be configured to run simultaneously or one or more channels may be excluded. (via 3-D Monicore or locally at the ATCU)

(vi) Once the scanning selection is made and the TIP is at the Parked position. the ATCU controls the actual scannincj function and interfacing with 3-D to download data.

(vii) Once initiated AUTO -TIP scan can be aborted at each individual NUfv1AC ATCU by pressing the ABORT AUTO-TIP soft key or via the 3-fl Monicore program (b) Manual Operation: Obj.V.8.5iV.C.5 (i) Location operation at the ATCLI penormed for:

• Exercising TIP drives for Rx startup

• Exercising Ball valves

• Selecting specific areas of the core to be monitored

• Used in conjunction with hand crank to Determine core top and bottom positions, parked position in shield position for setting the travel limits.

• Obtain torque data

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HATCH 2009 HLT4 NRC Exam ii 2:5001A3.05 001 Unit 1 is operating at 100% power with the A Traversing In-Core Probe (TIP) inserted in the core to perform 57CP-C5 1-010-Ct. TIP Flux Probing Monitor.

A transient occurs on Unit I with the following plant conditions:

Reactor pressure 900 psig and stable Reactor level (lowe st -20 inche.s and slowly increasing Drnvell pressure 1.5 psig and stable Drvvell temperature l29tF Which ONE of the following completes the statement below?

The A TIP will withdraw to the and the Bail Valve position will be A. Indexer (Parke rl) position:

open B. Indexer (Parked) position:

closed C. In Shield position:

open D In Shield position:

closed

Description:

The TIP receives a signal to withdraw to the in-shield posrtion upon receipt of a oroup 2 signal (1.85 psig OW press or +/-3 RPV water level. The ball valve auto closes when the probe is filly withdrawn.

The normal position for the TIP is in the Indexer with the ball valve open.

Friday, May31. 2009 8:37:14 AM 24

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 230000 RHR/LPCI: Torus/Suppression Pool Cooling Mode G2.4.31 (10CFR55.41.1O) Tier # 2 Knowledge of annunciator alarms, indications, or response Group # 2 procedures.

KIA# 219000G2.4.31 Importance Rating 4.2 Proposed Question: # 57 Unit 2 is at 100% Reactor Power with Residual Heat Removal (RHR) Loop II in Suppre ssion Pool Cooling mode. The following alarms are received on Unit 1:

• DRYWELL PRESSURE HIGH HALF SCRAM, (1-9-4A, Window 8)

• RX PRESS LOW CORE SPRAY/RHR PERMISSIVE, (1-9-30, Window 35)

Which ONE of the following describes the current status of Unit 2 RHR system AND what actions, if any, must be taken to restore Suppression Pool Cooling on Unit 2?

A. ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling IMMEDIATELY.

B. 2A AND 2C RHR Pumps are tripped. 2B AND 2D pumps are unaffected. NO additional action is required.

C. ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling after a 60-second time delay.

D. 2B AND 2D RHR Pumps are tripped. 2A AND 2C pumps are unaffected. Place RHR Loop I in Suppression Pool Cooling IMMEDIATELY.

Proposed Answer: C Explanation A INCORRECT: This is plausible because all four RHR pumps on Unit 2 will (OptIonal): trip, but they are locked out from manual start for 60 seconds based on Diesel Generator and/or Shutdown Board loading concerns.

B INCORRECT: This is plausible based on RHR Loop II being the preferred pumps for Unit 2.

C CORRECT: Candidate must determine that the combination of Unit 1 annunciators indicates a CAS initiation and the response of Unit 2 RHR pumps in Suppression Pool Cooling. Then, must recognize that Preferred and Non-preferred Emergency Core Cooling System (ECCS) Pumps do NOT apply with the given conditions. Unit 1 Preferred RHR pumps are 1A and 10. Unit 2 Preferred RHR pumps are 2B and 2D. LOCA signals are divided into two separate signals, one referred to as a Pre Accident Signal (PAS) and the other referred to as a Common Accident Signal (CAS). If a unit receives a CAS, then all its respective RHR and Core Spray pumps will sequence on based upon power source to the SD Boards. All RHR and Core Spray pumps on the non-affected unit will trip (if running) and will be blocked from manual starting for 60 seconds. After 60 seconds all RHR pumps on the non-affected unit may be manually started.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: This is plausible if taken from the perspective of Unit 1 operation, NOT Unit 2 operation.

KA Justification:

This question satisfies the KIA statement by requiring the candidate to use knowledge of annunciators for specific plant conditions to determine which RHR pumps can be used for Suppression Pool Cooling.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 1-ARP-9-3C Rev. 22/ OPL17I.044 R. 17 (Attach if not previously provided) 1-ARP-9-4A Rev. 18 / 2-Ol74 Rev. 152 Proposed references to be provided to applicants during examination:

Learning Objective: OPL1 71044 V.B.9/1 3 (As available)

Question Source: Bank # BEN 0610 #32 Modified Bank# (Note changes or attach parent)

New Question History: Last NRC Exam Browns Ferry 0610 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Question stem has been modified from original to meet KA. However, changes do not meet requirement of significantly modified question and is therefore identified as a Bank Question. Original attached.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 1-9-3 I -ARP-9-3C Unit I I -XA-55-3C Rev. 0022 Page 41 of 41 Sensor;Trip Point:

RX PRESS LOVV i-PIS-003-0074A 450 psi CORE SPRAY/RHR PERMISSIVE l-PIS-003-0074B 450 psi

- PA-3-74 l-PIS-0G8-0095 450 psi i-P[S-068-009t3 450 psi (Page I of 1)

Sensor IPIS-0030074A lPlS-003-0074B i-P1S0680095 1-PIS-06800913 Location: 1-PNLA-009-00$ 1 i-PNLA-009-0082 l-PNLA-009-COEii 1-PNLA-009-0082 ALJX. INST. Rm. AUX. INST. Rrn. AUX. INST. Rm. AUX. INST. Rrn.

EL 593 EL 593 EL 593 EL 593 Probable A. Reactor Pressure 450 psig.

Cause: B. Sensor Malfunction.

Automatic A. Switch No. 2 permits opening of Inboard Injection Valves for Core Spray Action: (1-FCV-75-25) and RHR(i-FCV-74-67).

B. In conjunction with High Drywell Pressure ( 2.45 psig) provides Auto Start Signal to Core Spray and RHR (LPCI).

NOTE Switch No. 1 (setpoint 230 psig) auto closes the Recirc Pump A Disch. Valve, i-FCV-S3-3.

Operator A. VERIFY RPV pressure by multiple indications.

Action: B. MONITOR drywell pressure.

References:

i-45E620-2-i i-47E61 0-3-i I -47W600-58 GE 0-730E930-3 and -9 Tech Spec 3.3.5.1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-4 1.ARP-9-4A Unit I 1.XA-55-4A Rev, 0018 Page 11 of 47 SensoriTrip Point:

DRYWELL PRESSURE HIGH i-PIS-064-0Ul56A 2.45 psig positive pressure in the HALF SCRAM I-PIS-064-00566 drywelL I -P15-064-00560 1 -PIS-064-0056D (Page 1 of 1)

Sensor 1 -PIS-064-0056A, 1 -PNLA-0Q9-0083, Auxiliary Instrument Room Location: 1-PIS-064-0056B, 1-PNLA-009-0084, Auxiliary Instrument Room I -P15-064-00560, 1 -PNLA-009-0085, Auxiliary Instrument Room I -PIS-084-0056D, 1 -PNLA-009-0086. Auxiliary Instrument Room Probable A. 2.45 psig in the drywell.

Cause: B. SI/SR in progress.

Automatic A. Half scram if one sensor actuates.

Action: B. Reactor scram if one sensor per channel actuates and group 2, 6 and 8 P015.

Operator A. VERIFY alarm by multiple indications. D Action: B. IF drywell pressure is 2.45 psig AND reactor has NOT scrammed, THEN MANUALLY SCRAM the reactor. ENTER 1-EOI-1 & 1-EOl-2 FLOWCHARTS. 0 C. DISPATCH personnel to the pressure switches to check for abnormal condition. C D. IF alarm is NOT valid. OR mitiating condition is corrected. THEN with SRO permission, RESET Half Scram. REFER TO 1-01-99. C

References:

l45E62051 IJ30E9l5i FSAR Sections 7.2.3.1. 7.2.3.5, 13.6.2

ES-401 Sample Written Examination Form ES-401 -5

{

Question Worksheet CPU 71.044 Revision 17 Page 51 of 146 INSTRUCTOR NOTES a Common Accident Signal anticipation of a

-12Z Rx water level (Level 1) GAS injection requirement.

OR PAS and GAS are 2.45 psig OW pressure initiated by any unit Core Spray AND logm.

<450 psig Rx pressure

2. If a unit receives an accident signal, then all its respective RHR and Gore. S pray pumps will sequence on based upon power Level I source to the S.D Boards. OR 245#AND<450 psig RPV

{

3. Affected, non-affected and All 8 DGs are preferred pump logic applies to started by any Units 1 & 2 because they share unit PAS signal.

DGs and SD boards. Unit 3 pumps are not affected by Unit 1/2 signals.

a. All RHR and Gore Spray pumps on The non-affected unit will trip (if running) and will he blocked from manual starting for 80 seconds.

After 60 seconds all RHR Operator diligence pumps on the non-affected required to unit may he manually started. prevent

c. The non-preferred pumps on overloading SD the non-affected unit are also boardsDUs prevented from automatically starting until the affected units accident signal is clear.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. 0152 Page 393 of 442 Appendix A (Page 2 of 7)

Unit I & 2 Core SprayJRHR Logic Discussion 2.2 ECCS Preferred Pump Logic Concurrent Accident Signals On Unit I and Unit 2 With normal power available, the starting and running of RHR pumps on a 4KV

{

Shutdown Board already loaded by the apposite units Core Spray, RHR pumps

, and RHRSW pumps could overload the affected 4KV Shutdown Boards and trip the normal feeder breaker. This would result in a temporary loss of power to the affected 4KV Shutdown Boards while the boards are being transferred to their diesels. To prevent this undesirable transient, Unit 2 RHR Pumps 2A and 2C are load shed on a Unit 1 accident signal and Unit I Pumps lB and 1D v.ill he load shed on a Unit 2 accident signal. Unit 2 Core Spray Pumps 2A and 2C are load shed on a Unit 1 accident signal and Unit I Core Spray Pumps lB and 1D v1ll be load shed on a Unit 2 accident signal. This makes the Preferred ECCS pumps Unit 1 Division I

Core Spray and RHR Pumps and Unit 2 Division 2 Core Spray and RHR Pumps Conversely, the Non-preferred ECCS pumps are Unit 1 Division 2 Core Spray and RHR Pumps and Unit 2 Division 1 Core Spray and RHR Pumps.

The preferred and non-preferred ECCS pumps are as follows:

UNiT I & 2 PREFERRED ECCS Pumps CS 1A. CS 1C, RHR 1A, RHR 1C CS 28, CS 2D, RI-fR 2B, RHR 2D NON-PREFERRED ECCS Pumps CS 1B, CS 1D. RI-fR 18, RHR 1D CS 2A. CS 2C. RHR 2A. RHR 2C UNIT 3 Unit 3 does not have ECCS Preferred!Non-Preferred Pump Logic.

Accident Signal On One Unit With an accident on one unit, ECCS Preferred pump logic trips all running RI-IR and r Core Spray pumps on the non-accident unit.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0610 NRC RO EXAM

32. i.o i9OoOK2 2 O2f/ROfSROBANK Given the following plant conditions:

Unit 2 is at 100% rated power with Residual Heat Removal (RHR) Loop II in Suppression Pool Cooling mode to support a High Pressure Coolant Injecfion (HPCI) Full Flow Test surveillance.

Unit I experiences a LOCA which results in a Common Accident Signal (CAS) initiation on Unit 1.

Which ONE of the following describes the current status of Unit 2 RHR system and what actions must be taken to restore Suppression Pool Cooling on Unit 2?

A. ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling IMMEDIATELY.

B. 2A and C RHR Pumps are tripped. 2B and 2D pumps are unaffected. NO additional action is required.

C ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling after a 60-second time delay.

0. 2B and 2D RHR Pumps are tripped. 2A and 2C pumps are unaffected. Place RHR Loop I in Suppression Pool Cooling IMMEDIATELY.

KIA Statement:

219000 RHR/LPC[: Torus/Pool Cooling Mode K2.02 Knowledge of electrical power supplies to the following: Pumps K1A Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to determine which RHR pumps can be used for Suppression Pool Cooling.

References 2-01-74, OPL171.044 Level of Know[edciJustification: This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

0610 NRC Exam Fndy, ebruarv 29, 2008 3:005 AM 69

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 230000 RHR/LPCI: Torus/Pool Spray Mode Tier # 2 K2.02 (1 OCFR 55.41.7)

Knowledge of electrical power supplies to the following: Group # 2 Pumps KIA# 230000K2.02 Importance Rating 2.8 Proposed Question: # 58 Unit 3 is operating at 100% Reactor Power with the Alternate Supply Breaker 1528 to 4 kV Unit Board 3B tagged out of service. An accident results in the following conditions:

• Unit Station Service Transformer 3B locks out

• Suppression Chamber Pressure reaches 3 psig

• 3A AND 3B RHR pumps are running in Suppression Chamber Spray Mode.

Which ONE of the following completes the statement?

The power supply for the 4 kV Shutdown Board to RHR Pump 3A is _(1)_ AND RHR Pump 3B is_(2)_.

A. (1) Common Station Service Transformer A (2) Common Station Service Transformer A B. (1) Common Station Service Transformer A (2) its associated Emergency Diesel Generator C. (1) its associated Emergency Diesel Generator (2) Common Station Service Transformer A D. (1) its associated Emergency Diesel Generator (2) its associated Emergency Diesel Generator Proposed Answer: B Explanation A INCORRECT: Part 1 correct See Explanation B. Part 2 incorrect See (Optional): Explanation C.

B CORRECT: 500 kV through USSTs is the normal supply to all U3 Unit Boards which in turn supply the 4kV Shutdown Boards. CSSTs are the alternate supply to the Unit Boards. EDGs are the emergency supply in case there is a loss of both normal and alternate supplies. Ordinarily the Unit Boards automatically transfer to alternate, however in this case the Unit Board 3B Alt is tagged out. So, when USST is lost, the 3C DIG will start and supply the 3EC 4 kV Shutdown Board which feeds RHR Pump 3B.

Unit Board 3A will transfer and be supplied power via the CSST A. Unit Board 3A feeds 4 kV Shutdown Board 3EA which feeds RHR Pump 3A.

C INCORRECT: Part 1 and 2 incorrect Plausible since the examinee must know which Unit Boards Supply which Shutdown Boards then RHR Pumps to eliminate these distractors.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Part 1 incorrect See Explanation C. Part 2 correct See Explanation B.

KA Justification:

The KA is met because it tests knowledge of electric power supplies to RHR Pumps.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71 .044 Rev. 17 (Attach if not previously provided)

OPL1 71 .036 Rev. 12 3-ARP-9-8B Rev. 14 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.036 V.B.8 (As available)

Question Source: Bank#

Modified Bank # Hatch 09 #22 (Note changes or attach parent)

NeW Question History: Last NRC Exam Hatch 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the in formation will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLiJi.036 Revision 12 Page i7 of 60 h.. Condensate pumps (3) 900hp (Unit 3), 1250 hp (for U-i and Unit 2)

c. Condensate booster pumps (3) i75Ohp (Unit 3), 3000 hp (each for Unit iand Unit 2)

d. Raw Cooling Water pumps (3) 300hp each a Control Rod Orve Water Pump A, 250 hp.(Units 1, 2, and 3; Board C)

f. 480V Unit Board transforniers (2) (Boards iA and 1B)

g. 460V Water Supply Board transformers (4kV B boards)

2. There are nine 4kV Unit Boards three per unit.

Refer to prints They are located in the turbine building on Elev. lSE-SOO series.

604 (A and C Boards) and Elev. 566 (B Boards).

The USSTs are the normal supply and start buses are the alternate.

a. USST A is the normal supply to 4k\ Unit Obj. V.B.6.d Board C and UBST B is the nonnal power Obj. V.C.l.d supply to 4kV Unit Boards A and B, (All Obj. V.D.6.d Units) b.. 4kV Start Bus 1A is the alternate power supply to 4kV Unit Boards iA, 2A, 2C, 3A, and 3C.

c. 4kV Start Bus iB is the alternate power supply to 4W Unit Boanls 1 B, I C, 2B, and 3B.

3. lJi and U2 4kV Unit Boards A and B supply Ohj. V.B.6.a power to 4k! Shutdown Buses I and 2 thereby Obj. V.B.6.c providing off-site power to the Standby AC Obj. V.B.7 Power System. 3A, 3B 4kV Unit Boards supply Chj. V.C.i.a power directly to the U3 4kV Shutdown Boards. Obj. V.C.i.c

4. Control Room Indications Oh]. V.D.6.a Oh]. V.D.6.c Obj. V.D.7

a. Voltmeter, 2 ammeters (one on each supply) on panel 9-8 from each 4kV Unit Board.

h. Ammeters in the Control Roorn for each

- No Amp Meters for of the boards pump motors. CRD Pumps

5. Indication of the 4kV Unit Boards voltages and

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .038 Revision 12 Page 16 of 60 amperages are available on panel 9-8. In Monitor redundan!

addition, each boards pump motor amps is also indications available. (except CRD pumps)

6. Transfer Schemes

a. General Operation 0-45E763-i, 2 The 4k! lJnit Boards are normally fed Obj. V.C.2.d from the Unit Station Service Obj. V.B.8.d Transformers with an alternate feed from Obj .. O8 cl the 4kV Start Buses. Illustration 1 0-01 -57A Transfer to the Start Buses may be manual or automatic but transfer back to Only 10, 20, the USST is manual only. All manual 3A/B/C Unit Board transfers and transformer trip-actuated have 30% slow transfers are fast transfers. Undervoltage relayactuated transfer is delayed until transfer. Removed from lA/B & 2AJB bus voltage has decreased to 30%

Unit Board.

normal. A voltage relay prevents automatic transfer to a dead bus. The breakers are electrically interlocked to prevent paralleling the Unit and Common transformers.

b. Automatic East transfer of Unit boards occur on Gen protective relaying or USST relaying.

To automatically fasttransfer from normal to alternate (l) nomial feed breaker tripped

ç2) 43 selector switch in AUTO (3) Alternate feed line-side voltage available 27SUX (4) Alternate feeder breaker closes, provided no lock-outs are present.

c. To automatically transfer from normal to LIV transfer only on altemate from undervoltage IC, 20, 3A. 3B. 30 (1) 43 transfer switch in AUTO (2) Alternate voltage available

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-8 3-ARP-9-8B Unit 3 3-XA-55-SB Rev. 0014 Page 13 of 38 Sensor/Trip Point:

4KV UNITBD 43 Switch in AUTO fXS-5741 Generator/Transformer

,LSUIQXFR and protective relays

- Alt feeder brk 1432 (52a contacts) or (Page 1 of 1) r closed undevoltage relay Sensor Unit Bd 3A Location: El 604, T-16 C-LINE TLJrb Bldg Probable A. Protective relay operation.

Cause:

• 66TX, 86TF, 86C (any relay causes high speed transfer).

27IUAX (time delayed transfer).

B. Fuse failure (metering potential transformer)

C. Relay malfunction.

Automatic Transfer to alternate feeder (Start Bus IA).

Action:

Operator A. VERIFY Unit in stable condition by checking:

Action:

• Condensate Pump 3A

• Condensate Booster Pump 3A C

• RCW Pump 3A C

• CCW Pump 3A C B. ON Panel 3-9-8, CHECK:

1. Alternate bkr to Unit Bd 3A closed (red light illuminated). C

2. Nomial bkr Unit Bd 3A open (green light illuminated). C

3. SELECT Unit Bd 3A with volt switch and CHECK voltage on meter (3-El-57-28). C C. CHECK Unit Bd 3A for abnormal conditions: relay targets; smoke, burned paint; bkr position; etc. C D. REFER TO 0-Ol-57A for board transfer. C

References:

345E721 045N7631 345N5201 I

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17l .036 Revision l2 Page 25 of 60 (7) CASx (CASA or CASS) accident signal -122 RxVL OR 2.45 (after 5 second delay via BBRX relay) DWP AND <450 RPV 4kV Shutdown Boards (Normal Power Seeking) Refer to prints 15E-500 seiles Key Diagram of STDBY Aux. Power System Power sources Obj. V.B.6.c

a. 4kV supplies to each U 1/2 Shutdown Board:

areasfoflows:

3! 1

\C.LC Board NORMAL Supply A Shutdown Bus 1 B Shutdown Bus 1 C Shutdown Bus 2 D Shutdown Bus 2 The first alternate is from the other Shutdown SBO Bus. The second alternate is from the diesel .,

.-2viaUu tie aenerator. The third alternate is from tne 1J3 iOc1i(

diesel generators via a U3 Shutdown Boani.

!4 2 via other SD Bus

b. There are two possible 4kV supplies to each U3 Shutdown Board:

Board NORMAL Supply 3EA lJnit Board 3A 3EB Unit Board 3A 3EC Unit Board 3B 3ED Unit Board 3B (1) The first alternate is from the diesel generators. The U112 diesel generators cannot supply power to the U3 Shutdown Boards alone. They may; however, be paralleled with the U3 diesel generators for backfeed operation. The tie breaker off the unit 3 Shutdown Board is interlocked as follows:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17 1.044 Revision 17 Page 26 of 14$

INSTRUCTOR NOTES Li 1/2 RHR B II Shutdown Board C i2 C pumps come off the C and 6 IJ1/2 RHR D II Shutdown Boani 0 1/2 0 shutdown boards respectively.

L13 RHR A 1 Shutdown Bd 3EA 3A BPN events have occurred due to U3 RHR C I Shutdown Bd 3E6 36 racking out the wrong breaker US RHR B II Shutdown Bd 3EC 30 which resulted in LOC 3.0.3 U3 RHR D II Shutdown Bd 3ED 3D BEPER-09720$

See OPL17I.045 for details.

TP-1 0, Ii and 12

c. Pump cooling Ohj. V.6.5 (I) Pump beahngs cooled by RHR pump discharge from seal heat exchanger (2) Seal heat exchanger normally by EECW North or South headers (3) EECW also cools the RHR room coolers

d. Check valves located on the discharge of the TP-1 and2 pumps (1) Prevents backflow through the pumps (2) Maintains a water leg in the discharge piping (3) VVater legs kept filled up to the injection Ohj. V.6.6 valves by the keep fill system. Obj. V.D.2 (4) This prevents water hammer on pump start and possible pipe/valve damage.

(5) Also enables water to reach the core in the shortest possible time in the event of a LOCA (6) Discharge piping kept pressurized (Tech. IRM 3.5.4 Requirements Manual limit)

3. RHR Heat Exchangers Ohj. V.6.7 Obj. V.E.5

a. Four vertical, shell and tube per unit Baffled at top.

b. Located in separate portions of Rx Bldg. RHRSW vents on top head (2)

c. Design data/conditions (1) Shell side fluid Rx water or S/P water (d 10000 gpm

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HATCH 2009 HLT4 NRC Exam

22. 226CO1K2.0 001 Unit I was operating at 100%. power with the Alternate Supply Breaker to 41 6OVAC bus 1E tagged out.

A loss of Startup Transformer (SAT) iD occurred.

o Torus Pressure reaches 3 psig during the transient.

o lA and lB RHR pumps are running in the Torus Spray Mode.

The power supply for the 4160 VAC bus to the 1A RHR Pump is (1) and to the lB RHR Pump is (2)

A. (1) SAT iC (2) SAT IC B (1) its associated EDG (2) SAT ic C. (1)SATiC (2) its associated EDG D. (1) its associated EDG (2) its associated E.DG Friday, May01. 2009 8:37.15 AM 41

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 234000 Fuel Handling Equipment Tier# 2 A4.02 (IOCFR 55.41 .7)

Ability to manually operate and/or monitor in the control room: Group #

• Control rod drive system KIA# 234000A4.02 Importance Rating 3.4 Proposed Question: # 59 Given the following:

• Unit I is in Mode 5

• The Refuel Platform is over the Spent Fuel Pool

• The Reactor Mode Switch is in START & HOT STBY for testing Which ONE of the following identifies when a rod block will occur?

A. When the Refuel Platform Fuel Grapple is lowered.

B. When a load is placed on the Refuel Platform Fuel Grapple.

C. When the Refuel Platform is driven near or over the core.

D. When the Refuel Platform starts moving towards the core.

Proposed Answer: C Explanation A INCORRECT: Plausible in that this would be the correct answer if the Mode (Optional): Switch was in Refuel and Platform near or over the core.

B INCORRECT: Plausible in that this is true if the service platform hoist is loaded.

C CORRECT: As the Refuel Platform is driven near the core with the Mode Switch in Startup, a rod block will occur.

D INCORRECT: The refuel platform can move towards the core but will be stopped when the platform starts to move over the core

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the ability to monitor Control Rod Drive system in the control room as it applies to Fuel Handling Equipment.

Question Cognitive Level:

This question is rated as Fundamental Knowledge Technical Reference(s): 0-GOl-100-3A Rev. 53 (Attach if not previously provided)

OPL1 71 .053 Rev. 18 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .053 V.B.5 (As available)

Question Source: Bank # Cooper 08 #59 (Note changes or attach parent)

Question History: Last NRC Exam Cooper 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Refueling Operations (In-Vessel 0-GOI-100-3A Unit 0 Operations) Rev. 0053 Page 18 of 175 3.3 Refueling Bridge Operation (continued)

C. When operating the refuel bridge in any speed other than JOG, ensure that the grapple or devices being transported have adequate clearance above items stored in the SFSP and Reactor Cavity.

ft Bridge travel toward the core will be stopped if any of the following conditions are met (except when interlocks are jumpered out by instruction in this procedure):

1. Any platform hoist loaded or main grapple NOT full up and all rods NOT full in with the platform near or over the core.

2. Platform near or over the core with the Mode Switch in other than REFUEL.

3. One rod withdrawn and when withdrawn rod is initially deselected with the Mode Switch in REFUEL. (As long as the rod that is withdrawn is never deselected bridge travel may continue and not be blocked by this interlock.)

E, The Associated Hoist operation will be stopped if any of the following exist.

1. Main Grapple position at full lower (46 ft.). Stops main hoist lower.

2. Main Grapple slack cable signal (<50 lb. tension on cable) stops main hoist lower.

3. Associated Hoist loaded with all rods NOT full in with the platform near or over the core. Stops raise.

4. Associated Hoist overloaded (> 1000 lb.). Stops hoist raise.

5. All rods NOT full in with Platform near or over the core. Stops main hoist raise or lower.

6. Associated hoist at full up. Stops raise.

F. A Rod Block will occur if any of the following conditions are met:

1. Any platform hoist loaded or main grapple NOT full up with the platform near or over the core with the Mode Switch in REFUEL.

2. Service platform dummy plug not installed.

3. One rod withdrawn and a second rod selected with the Mode Switch in REFUEL.

4. Platform near or over the core with the Mode Switch in STARTUP.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7I.053 Revision 18 Appendix C Page 43 of 56 TP-3: Refueling Rod Blocks and Refueling Interlocks

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet COOPER 2008 Question Newt Rev Revision Last Used Exam Bank Applicability Number Modified # Date Date or Bank NRC 110 Bank 00 07/28/1 999 01/30/2008 NRC Style 110: Y 59 1477 Question 8110: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 1 1 4 Multiple Choice Topic Area Description Ssteins COROO 12102001 100B Refueline Related Lessons C0110012 102 Refueling Related Objectives COROOI21O200IIOOB Given conditions associated with refueling activities determine if the following should occur: Refueling mast restrictions Related References 10CFRSS.41(b)7 Related Skills (K/A) 2340001(4.02 Knowledge of FUEL HANDLING EQUIPMENT design feature(s) and for interlocks which provide for the following: (CFR: 41.7) Prevention of control rod movement during core alterations (3.3 / 4.1) 122

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUESTION: NRC RO 5 Font *(Cfact)Tirns Giventhe following: .

• Core offload n in prngres.

• The refuel platform is over the fuel pool.

• The Reactor Mode Switch is placed in START & HOT STBY for testing.

If core ofuiosd activities conue,WHENwillarodblockoccur? onnattedForn Bed

a. When the refuel platform starts moving towards the core.

b. When the refuel platform is driven near or over the core.

c. When the Fuel Gsiipple is lowered.

& henaloadisplacedontheFuelGrapple, Fonnd:{DefA)Tfrres ANSWER: NRC RO 9 matredFou (cefk)Tsrs

b. When the refuel platform is driven near or over the core.

EXPLANATION OF ANSWER: l. correct, As a bundle is moved from the fuel pooi to the core the rod block will occur when the refuel bridge is driven over the core. a, The refuel platform can move towards the core but will be stopped when the platform starts to move over the core c

& d. the rod block would occur before this paint.

123

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO SRO4 259001 Reactor Feedwater System Tier # 2 K5 03 (IOCFR 55 41 5)

Knowledge of the operational implications of the following concepts Group # 2 as they apply to REACTOR FEEDWATER SYSTEM: K/A # 259001 K5.03

• Turbine operation: TDRFPs-Only Importance Rating 2.8 Proposed Question: # 60 RFPT IA OVERSPEED TEST TRIP LOCKOUT, 1-HS-3-109A, has just been placed in the ELEC position per 1-01-3, Reactor Feedwater System, Section 8.10, Overspeed Trip Exerciser Test, when RFPT IA experiences an ACTUAL over-speed condition.

Which ONE of the following describes the AUTOMATIC response of RFPT IA?

A. Trips as a result of the electrical trip solenoid.

B. Trips as a result of the mechanical trip mechanism.

C. Will ONLY trip when I -HS-3-1 09A is restored to the NORM position.

D. Ramps up due to the overspeed condition AND locks at a high speed stop.

Proposed Answer: B Explanation A INCORRECT: The mechanical trip solenoid is still active, and will actuate, (Optional): causing a trip of the RFPT.

B CORRECT: The test blocks the electrical device trip but leaves the mechanical trip system active.

C INCORRECT: Yes the RFPT will trip when restored to NORM; however, the mechanical trip system remains active even in ELEC.

D INCORRECT: Even though it will ramp up, there is no protective function short of the mechanical overspeed trip device.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KAJustification:

The KA is met because the question tests the candidates knowledge of the operational implications of Turbine operation as it applies to the Reactor Feedwater System.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL1 71.026 Rev. 15 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .026 V.B.5 (As available)

Question Source: Bank # BEN 1006 Audit #63 (Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI7I O26 Revision 15 Page 33 of 58 The Emergency Governor Lockout Valve Review trip provides a method to periodically test and system failure exercise the Trip Dump Valve and mechanical events described overspeed mechanism without tripping the in INPO 0 and turbine. To accomplish this it is moved up into MR 399.

position to block the pressure holding the Pressure Relay Valve open from being dumped by the Trip Dump Valve.

j. Placing the OVERSPEED TEST TRIP 01-3 sectIon 8.10 LOCKOUT Switch in the MECH position on panel 9-6 energizes the Lockout Solenoid Valve providing the oil pressure to move the Emergency Governor Lockout Valve to block trips. Emergency Trip Governor Valve position indication changes from Green (normal) to amber (lockout>. Electrical overspeed will deenergize the Lockout Solenoid Valve.

k. To provide continuous trip protection for the Note: the lockout RFPT during testing, the lockout oil pressure valve blocks ALL is also ported to the Electrical Trip Solenoid trips. Only Valve which will dump lockout pressure removal of the should a trip condition occur while testing the lockout will restore Trip Dump Valve and overspeed mechanism. trips. Electrical (The 1/8 orifice in the oil supply cannot overspeed maintain pressure with a trip dump.) Electrical removal of the overspeed will also deenerglze the lockout lockout occurs solenoid yielding earlier response to an actual before the actual overspeed condition, trip setpoint is reached. See 01-3 I. The OVERSPEED TEST pushbutton supplies 8.9.49.

oil pressure to move the overspeed plunger which trips the Trip Dump Valve to exercise the overspeed mechanism and Trip Dump Valve, The green normal indication extinguishes and the white trip light lights.

m. The OVERSPEED TEST RESET pushbutton performs the Trip Dump Valve reset for this test. The normal Trip Reset will not function because the Stop and Control Valves must be closed for a normal reset. The white trip light must be extinguished and the green reset light must come on before returning the OVERSPEED TEST TRIP LOCKOUT switch to normal to preclude an actual turbine trip.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .026 Revision 15 Page 34 of 58

n. The ELEC position of the OVERSPEED TEST 01-3 section 8.9 TRIP LOCKOUT switch removes electrical overspeed trip for testing. Al! other trips remain functional.

a. Amber light below tachometer on 9-6 and Unit difference locally will be lit when electrical overspeed condition is reached. (Unit 3 flashes, Unit 2 does not.)

p. Testing of the turbine stop valves is required but the high pressure stop valve can only be tested If the HP control valve is fully closed.

Depressing the pushbutton on 9-6 causes the valve to close until it reaches its fully closed position or the pushbutton is released.

q. The Low Pressure Stop Valve can be tested at any time. Depressing the pushbutton on 9-6 causes the valve to travel to the mid (50%)

position and remain until the pushbutton is released.

r. High Water Level Trip

1) High water level trip at 55 comes off of These are LS-3-208A. B, C, D. uncompensated indicators

2) Logic is such that it Is 2-out-of-2 taken once. For example, in order for a full turbine trip to occur, either 208A and 208C or 2088 and 208D must be picked up.

3) Trip Channel A is 208A & 208C: Trip Channel B is 2088 & 208D.

4) Two sets of indicating lIghts (red & green) are Installed on panel 9-5 and two reset switches. Normal condition Green Light on; Trip condition Red light on:

5) Ready to reset condition Green & Red lights on

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 271000 Offgas System Tier# - 2 K3.02 (10CFR 55.41.5)

Knowledge of the effect that a loss or malfunction of the OFFGAS Group# 2 SYSTEM will have on following: KJA# 271 000K3.02

• tOff-site radioactive release rate Importance Rating 3.3 Proposed Question: # 61 Unit 2 Offgas Post Treat Radiation Monitor, 2-RM-90-265A, has failed downscale.

Which ONE of the following identifies the impact of this failure?

If Offgas Post Treat Radiation Monitor, 2-RM-90-266A, reaches the High-High-High setpoint, Off-Gas System Isolation Valve, 2-FCV-66-28, _(1)_ close.

If Offgas Post Treat Radiation Monitor, 2-RM-90-266A, fails downscale, Off-Gas System Isolation Valve, 2-FCV-66-28, _(2)_ close.

A. (1)wiII (2) will B. (1)willNOT (2) will C. (1)will (2) will NOT D. (1)wiIl NOT (2) will NOT Proposed Answer: A Explanation A CORRECT: Parts 1 and 2 correct OG POST TREATMENT RAD (Optional): MONITOR DOWNSCALE (55-4C-32) alarms when signal is < 1 cps and sends a trip signal to the Off-Gas isolation logic. OG POST-TREATMENT OFF-GAS HI-HI-HI/INOP (55-4C-35) alarms at 6.2X105 cps sends a trip signal to the Off-Gas isolation logic. Off-Gas isolation is a two-out-of-two logic. Downscale, Hi-Hi-Hi or INOP on RM-90-265A AND Downscale, Hi-Hi-Hi or INOP on RM-90-266A will automatically isolate the Off-Gas system after a 5 second time delay. (FCV-66-28 closes).

B INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A.

c INCORRECT: Part I correct See Explanation A. Part 2 incorrect See Explanation D.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Part I incorrect Plausible in that two channels are required for an isolation signal to 2-FCV-66-28 to be generated. Some process radiation monitors do not combine downscale with high radiation to generate the trips signal. Example: this combination would not result in a actuation of trip logic for Rx Zone Rad Monitors. Part 2 incorrect Plausibility based on the misconception that the downscale does not result in a trip condition which is true of some process rad monitors. Example:

Downscale on MSL Rad Monitors does not result in actuation of associated trip logic.

KA Justification:

The KA is met because the question tests candidates knowledge of the effect that a malfunction of the OFFGAS SYSTEM Post Treatment Radiation Monitor will have on Offgas Isolation Valve 2-FCV-66-28 and therefore Off-site radioactive release rate.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71.033 Rev. 13 (Attach if not previously provided) 2-01-90 Rev. 79 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .033 V.B.3 (As available)

Question Source:

(Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Radiation Monitoring System 2-01-90 Unit2 Rev. 0079 Page lOofYO 3.0 PRECAUTIONS AND LIMITATIONS A. The following Radiation Monitoring subsystems initiate the hsted automatic actions and isolations on high radiation trip signals:

1. Main Steam Line (3 times normal full-load background radiation).

a. Mechanical Vacuum Pump trip and suction valve isolation.

2. Off-Gas Post-Treatment

a. High opens Adsorber Inlet Valve, 2-FCV-66-i 13A, and closes Adsorber Bypass Valve 2-FCV-66-i 13B, if 2-HS-66-1 13 is in AUTO, ii High-High Alarms only.

c. High-High-High sends a close signal to Off-Gas System Isolation Valve, 2-FCV-6&-28 (5-second time delay).

3. Refueling Zone Ventilation (72 mr/hr high radiation signal from 2 out of 2 taken once logic or downscale/inop signal from I out of 2 taken twice logic.

a. Standby Gas Treatment System auto start.

b. Refueling Zone Vent System isolation.

c. Control Room Emergency Ventilation auto start. (Normal Control Room Ventilation isolates.)

4. Reactor Zone Ventilation (72 mr/hr high radiation signal from 2 out of 2 taken once logic or downscale/inop signal from 1 out of 2 taken twice logic.

a. Group 6 Isolation.

b. Standby Gas Treatment System auto start.

c. Refueling Zone Ventilation isolation.

d. Control Room Emergency Ventilation auto start. (Normal Control Room Ventilation isolates.)

5. Control Room Ventilation Monitoring (221 cpm above background high activity or two channels downscale/inop)

a. Control Room Emergency Ventilation auto start. (Normal Control Room Ventilation isolates.)

B. Abnormal or significant rises in radiation levels are required to be reported to the Unit Supervisor/SRO.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Radiation Monitoring System 2-01-90 Unit 2 Rev. 0079 Page 40 of 70 Illustration I (Page 2 of 4)

Radiation Monitoring System Operational Summary NOTE Only the noble gas detectors are required by Technical Specifications.

Stack Gas Two radiation detectors monitor activity release rates From the Radiation Monitors Off-Gas stack. PNL 0-25-39 0-RE-90-1 47& 148 Off-Gas Pretreatment Two radiation detectors monitor radiation at the inlet of the Radiation Monitors 6-hour holdup volume. PNL 2-25-38 2-RE-90-157&160 Off-Gas Post-treatment Two radiation detectors monitor radiation downstream of the Radiation Monitors charcoal beds (adsorbers) If adsorber control switch is in 2-RE-90-265&266 AUTO, the detector High trip ensures Off-Gas flow is directed through the adsorbers by inserting a CLOSE signal to the adsorber bypass valve and an OPEN signal to the adsorber inlet valve. High-High gives alarm signal. When the High-High-High trip is actuated, the Off-Gas System isolation valve closes after a 5-second time delay. PNL 2-25-94 Main Steam Line Two detectors monitor the Main Steam Lines for high Radiation Monitors radiation.

2-RE-90-i36137 Process Liquid Radiation detectors monitor radiation in the following systems:

Radiation Monitors 2-RE-90-131A Reactor Building Closed Cooling Water (off-line), Pnl 2-25-339 2-RE-90-130 Radwaste Effluent Discharge (in-line only) 2-RE-90-133A & 134A RHR Service Water (off-line), Pnl 2-25-337 & 338 2-RE-90-1 32A Raw Cooling Water (off-line) Pnl 2-25-336

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71.033 Revision 13 Page 21 of 75 INSTRUCTOR NOTES (5) Off-Gas isolation is a two-out-of-two Obj. V.B4.b logic Obj. V.C.4a I

(a) Downscale, Hi-Hi-Hi or INOP on RM-90-265A AND Downscale, Hi-Hi-Hi or INOP on RM-90-266A will automatically isolate the Off-Gas system after a 5 second time delay.

(FC\/-66-28 closes)

I Stack-Gas Radiation Monitoring System (RM-99-147 Obj. V.D.7

& 148) Obj VB3b Obj V.C3t a Purpose (1) Used to indicate and record release rates from the stack during normal operation and to alarm whenever limits are reached (2) To monitor the stack gas effluent, a Note: isokinetic sample is drawn through an isokinetic probe explained in probe which is located two-thirds of the section 9 of this way up the stack lesson

b. The stack receives exhaust gases from following:

(1) Steam Jet Air Ejector (SJAE)

(2) Steam Packing Exhauster (SPE)

(3) Mechanical vacuum pump (4) Standby Gas Treatment (SGT)

(5) Stack Gas Analyzer Room Vent

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT OPL171 033 Revision 13 Page 16 of 75 INSTRUCTOR NOTES (2) High-High Radiation (a) MAIN STEAM LINE RAD HIGH-HIGH / INOP (55-3A-27) alarm at a radiation level of 3 times the Normal Full Power Background radiation level (b) RAD HIGH-HIGH / INOP Alarm signal is generated by MSL Rad Recorder (RR-90-i 35)

(3) Downscale (a) MAIN STEAM LINE F

DOWNSALE (55-3A-14) alarms when low detector output is sensed (b) During normal power operation this indicates instrument malfunction (c) This alarm is expected during conditions of very low Main Steam flow (d) DOWNSCALE Alarm signal is generated by NUMAC Log Radiation Monitor

e. Trip Obj. V.Bi Obj.V.Ci (1) Trip level MAIN STEAM LINE RAD

- Obj. VD.2 HIGH-HIGH / INOP 3 times normal full power background radiation from monitor or detector INOP (2) Closes condenser vacuum pump Obj. V.B4.b suction valves FCV-66-36 and 40 and Obj. V.C.4.b trips condenser mechanical vacuum pump

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT OPL171 .033 Revision 13 Page 29 of 75 NSTRUCT0R NOTES

{

h. (c) Trip logic for the refueling and the reactor zones is identical, and the following combinations will generate a trip:

Two high level trips in the same Two-out-of-two, once channel, (division)

-OR-One downscale trip in each One-out-of-two, twice channel (division)

-OR-One monitor INOP in each One-out-of-two, twice channel (division) Obj. V.B.3.f Obj. V.C.3.f

-OR-Loss of RPS power to either channel (2) Automatic actions Obj V.Bi,3.e Obj. VCI,3.e (a) Refuel Zone Trip Obj V.0.6 (i) Isolate Refuel Zone (ii) Starts Standby Gas Treatment System (iii) PCIS Group 6 isolation Obj. V.B.1,3.f 3.g Obj. V.C.1,3.f 3g (iv) Starts CREVs (b) Reactor Zone Trip (i) Isolate Control Room, Reactor Zone, and Refueling Zone ventilation (ii) Starts Standby Gas Treatment System (iii) Start CREV5 (iv) PCIS Group 6 isolation

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO 288000 Plant Ventilation Systems Tier # 2 A301 (IOCFR 55417)

Ability to monitor automatic operations of the PLANT Group # 2 VENTILATION SYSTEMS including: K/A # 288000A3.O1

• Isolation/initiation signals Importance Rating 3.8 Proposed Question: # 62 Given the following Control Room Emergency Ventilation (CREV) system conditions:

• CREV Train A was started to prove operability following maintenance on the charcoal trays using the STOP-AUTO-START switch on Panel 9-22.

• The SYSTEM PRIORITY SELECTOR SWITCH is selected for TRAIN-B.

Which ONE of the following describes the CREV system response should a valid CREV initiation signal be received?

CREV Train B would_(1)_ AND CREV Train A would _(2)_.

A. (1) initiate (2) shutdown B. (1) NOT initiate (2) shutdown C. (1) initiate (2) NOT shutdown D. (1)NOTinitiate (2) NOT shutdown Proposed Answer: C Explanation A INCORRECT Part 1 correct See explanation C. Part 2 incorrect See (Optional): Explanation B.

B INCORRECT: Part 1 incorrect Normally, when an auto initiation signal is received, the TRAIN selected for secondary begins its start sequence but will not finish if the Primary CREV train is running. This is sensed by looking at the P across the HEPA filter. Since Train B was selected as the Primary CREV unit, the start sequence does not look at the AP. Part 2 incorrect -

This would be correct if CREV Train A was started using the AUTO-INmATE TEST switch, as would be the case during the periodic surveillance test.

C CORRECT: Part 1 correct CREV Train B will initiate without a time delay since the CREV UNIT PRIMARY SELECTOR SWITCH is selected for TRAIN B. Part 2 correct CREV will not automatically shutdown with a valid initiation signal present.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D INCORRECT: Part 1 incorrect See explanation B. Part 2 correct See Explanation C.

KA Justification:

The KA is met because the question tests the ability to monitor automatic operation of Control Room Emergency Ventilation including system initiation signals for the given conditions.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): 0-01-31 Rev. 136 (Attach if not previously provided) 1-EOl-3 Rev 12, OPL1 71 .067 Rev 16 Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.2.g (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 0707 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off.Gas Treatment 0.01-31 Unit 0 Building Air Conditioning System Rev, 0136 Page 21 of 285 3.6 CREV and CREV instrumentation operability issues (continued)

B. The main control room boundary may be opened intermittently under administration controls. For openings other than normal entry and exit, these controls consist of stationing a dedicated individual at the opening who Is in continuous communication with the main control room and whose task is to close the opening when main control room isolation is indicated, With both CREVs Inoperable in Modes 1, 2 or 3, for other than a control room boundary issue, enter LCO 3.0.3 Immediately. With two CREV subsystems inoperable during OPDRVs, initiate action to suspend OPDRVs. Reference TS 3.7.3.

C. When there is an automatic actuation of CREVS, the following automatic isolation dampers and hatch is required to be closed for CREVS to be considered operable.

1. 0-FCO-31 -15DB, 0-FCO-3 1-1500, 0-FCO-31-1 50E, 0-FCO-31 -1 50F, 0-FCO-31 -150G.

2. Removable equipment hatch in U-3 Mechanical Equipment Room, floor Elevation 617.

0. The CREV system utilizes 15.45 kW Duct heaters to control moisture buildup in the charcoal adsorber. A malfunction of the 15.45 kW duct heater makes the applicable CREV unit inoperable. [Reference Functional Evaluation In PER 74959 and 75680]

E. One of the UNIT 1 & 2 Control Bay Supply Fans and one of the UNIT 3 Control Bay Supply Fans and their associated power and control circuits is required to be operable for CREVS instrumentation (control bay high radiation)to be considered operable. Reference Tech Spec 3.3.7.1.

r. CREV UNIT PRIMARY SELECTOR, 0-XSW-031-7214 may be placed in either the A or B position, depending on the operability status of the CREV trains, When a CREV train is inoperable, it will NOT be selected as lead. When both CREV trains are operable, the preferred position for CREV UNIT PRIMARY SELECTOR, 0-XSW-031-7214, is in TRAIN 1 A which makes the A CREV the lead train. In the event that A CREV is INOP, CREV UNIT PRIMARY SELECTOR, 0-XSW-031-7214 is required to be placed in the TRAIN B position so the SB CREV will Initiate, without a tIme delay, as the lead train.

3. When one of the CREV trains is inoperable for testing, the CREV UNIT PRIMARY SELECTOR SWITCH, 0-XSW-031-7214 is required to be aligned to the train which is NOT under testing conditions to ensure the non-test train will initiate under an actual initiation signal.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off-Gas Treatment 0-01-31 UnIt 0 BuildIng Air Conditioning System Rev. 0136 Page 146 of 285 7.20 Shutdown of Control Room Emergency Ventilation (CREV) Fans to Standby Readiness CAUTION In the event the pressurization units have initiated automatically on a Group Six Isolation signal or control room ventilation inlet duct high radiation, the Initiating condition should be removed or corrected prior to shutting down the units.

NOTES

1) cicj After an automatic Initiation, the CREV System is required to be manually SHUTDOWN from the control room by placing the CREV Train hanclswitches to STOP, which also resets the initiation logic. A local shutdown will NOT reset the seal-in logic. ELER 88-0351 2> Normally, the train selected by the CREV PRIMARY UNIT SELECTOR as the lead train starts on auto initiation and the other train remains idle, unless the lead train trips.

Upon restoring the system to standby, the handswitch for the Idle train is required to be turned to STOP first to prevent It from starting when the Lead train is stopped.

3) The charcoal adsorber resistance heaters will be automatically placed in operation to maintain the charcoal beds at 10 degrees F greater than ambient temperature, provided that fan A(B) power supply breakers 14C (13C2) on 480V Reactor MOV Board IA(3B) are closed.

4) if a CREV train is in service for testing, and an actuation signal is received, both trains will be running. In this case. ONLY the train uricer test will be required to be shutdown.

f 1] IF CREV was manually or automatically Initiated, AND conditions requiring the initiation are cleared, THEN STOP CREV train A(B) as follows:

[1.13 VERIFY CREV TRAIN A INIT/CB ISOL, O-HS-31-1 50A, and CREV TRAIN B INIT/CB ISOL, O-HS-31-150B, are in the AUTO position at Panel 2-9-22, D

[1.23 For the CREV TRAIN that is NOT running, PLACE CREV TRAIN A, O-HS-31-7214A, or CREV TRAIN B, O-HS-31-7213A, momentarily in STOP at Panel 2-9-22. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 0707 #38 Examination Outline Cross-reference Level RO SRO 290003K104 Tier # 2 Knowledge of the physical connections and/or cause-effect Group # 2 relationships between Control Room HVAC system and the KJA # 290003K1 .04 following: Nuclear Steam Supply Shut off System (NSSSS/PCIS). Importance Rating 3.2 3.3 Proposed Question: RO # 38 Given the following Control Room Emergency Ventilation (CREV) system conditions:

• CREV Train A was started to prove operability following maintenance on the charcoal trays using the STOP-AUTO-START switch on Panel 9-22.

• The SYSTEM PRIORITY SELECTOR SWITCH is selected for TRAIN-B.

Which ONE of the following describes the CREV system response should a valid CREV initiation signal be received?

On a valid initiation, CREV Train B would (1) and CREV Train A would (2)

(1) (2)

A. initiate shutdown.

B. initiate NOT shutdown.

C. NOT initiate shutdown.

D. NOT initiate NOT shutdown.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 290001 Secondary Containment Tier # 2 A1.O1 (IOCFR 55.41 .5)

Ability to predict and/or monitor changes in parameters associated Group # 2 with operating the SECONDARY CONTAINMENT controls K/A # 290001A1 .01 including:

System lineups Importance Rating 3.1 Proposed Question: # 63 On Unit 1, the Standby Gas Treatment System (SGTS) A Control Switch, 1 -HS-65-l 8A, on Panel 1-9-25 has been placed in the pull-to-lock position.

Which one of the following conditions would still cause SGTS A to start?

A. Unit 2 drywell pressure rises to 2.5 psig.

B. Unit 3 SGTS A start pushbutton is depressed.

C. The local (SGTS Building) SGTS A start pushbutton is depressed.

D. SGT TRAIN A INBD ISOL TEST SIG Keylock switch (HS-65-48A) is placed in the TEST position.

Proposed Answer: C Explanation A INCORRECT: With the SGTS A Control Switch in Pull to Lock, the system (Optional): will not auto start on 2.5 psig. Plausible in that this condition will normally cause SGTS A to start.

B INCORRECT: With the SGTS A Control Switch in Pull to Lock, the system will not start with the Unit 3 SGTS A Start Pushbutton. Plausibility based misconception that Unit Control Switch will not affect operation from Unit 3.

C CORRECT: With control switch in pulled-out (STOP) position, the blower can still be started locally.

D INCORRECT: With the SGTS A Control Switch in Pull to Lock, the system will not auto start with SOT TRAIN A INBD ISOL TEST SIG Keylock switch (HS-65-48A) placed in the TEST. Plausible in that this condition will normally cause SGTS A to start.

KA Justification:

The KA is met because the question tests the candidates ability to predict changes in the SGTS associated with operating the SGTS Control Switch.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. Candidate must be able to predict the effect of changing the Control Switch position from its normal line up on the operation of the system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL171.018 Rev 10 (Attach if not previously provided) 0-01-65 Rev 53 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank# OPLI7I.018#13 (Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17I.018 RevIsion 10 Appendix C Page 36 of 37 TP4 SGT A (B) CONTROL CiRCUIT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.018 Revision 10 Appendix C Page 37 of 37 CONTROL POWER 480V SBGT Ed 480!120V TP.3: SOT A (B) CONTROL CIRCUIT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.018 Revision 10 Page 22 of 37 INSTRUCTOR NOTES

3. Control Logic

a. Control switch must be in AUTO for auto-start signal to start train.

b. With control switch in pulled-out (STOP) position, the blower can still be started locally.

c. Cs is spring-return-to-AUTO, unless pulled out in STOP.

d. Switch can be pulled out in the STOP position only.

e. Inlet damper will auto-open when fan motor coil is TP4 energized, if in AUTO. LER 88-0 17

f. SGT A and B will trip on initiation of the 480V load-shed logic, but will auto-restart after forty seconds if initiation signal is present. SGT C is not affected by 480 volt load shed logic initiation.

g. LER 88-017 covers an event that occurred at LER 88-017 BFNP. With the supply breaker (48OVAC) open, an engineer directed Maintenance to change the Review 0-01-65 P&L for state of the latching relay (MCX) to the operate Relay Information state, The control switch was in the LOCKOUT position. When the supply breaker was closed, the fan started since the MCX contact in series with the MC coil was closed, (1) Be aware that the pull-to-lock logic will not always prevent equipment start. MCX relay has two states:

RESET (blue PB out)

OR ACTUATED (PB depressed),

(2) If not RESET SGT may start when power is restored.

4. Emergency Operation

a. CAD System operation after a LOCA

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Standby Gas Treatment System 0-01-65 UnitO Rev.0053 Page 10 of 41 3.0 PRECAUTIONS AND LIMITATIONS (continued)

T. [NRG/] If any relays are ACTUATED, Site Engineering SHALL be contacted prior to energizing the circuit. The pull-to-lock logic will NOT inhibit the SGT Blower from starting when the SGT Blower breaker is racked in and the MCX relay is actuated (blue contact position indicator retracted). INRSIER 88017]

U. Start relays, MAX and MBX for Standby Gas Treatment trains A 9 and 9B respectively, are of a different type than the MCX for train However, the same problem exist for these relays as does for the MCX relay. If the contacts are closed (pulled up) prior to the breaker being closed, the standby gas treatment train will start when the breaker is closed. FAILURE to have the contacts open (dropped down position) wilt result In the associated Standby Gas Treatment train starting when the breaker is closed.

V. The following signals on any unit will start all three SGT trains when the respective control switches are In AUTO:

1. High drywell pressure (2A5 psig).

2. Low Reactor Water Level (LEVEL 3).

3. High Rx Zone Ventilation Radiation (72 MR/hr).

4. High Refuel Zone Ventilation Radiation (72 MR/hr).

5. One out of two taken twice trip logic for Reactor Zone Ventilation Radiation downscale.

6. One out of two taken twice trip logic for Refuel Zone Ventilation Radiation downscale.

W. When the control room handswitch for an SGT Fan is in PULL-TO-LOCK, the fan may still be operated locally.

X. The following system valves fail open upon a loss of power (all other system valves fail closed):

1. SOT FILTER BANK C OUTLET DAMPER, O-DMP-065-0067

2. SOT FAN A INLET DAMPER, 0-DMP-065-0017

3. SOT FAN B INLET DAMPER, O-DMP-065-0039 Y. The SOT FILTER BANK A & B BYPASS DAMPER, 0-DMP-065-0022, is normally fed power from 480V Diesel Aux Bd A. Power to O-DMP-O65-0022 is automatically transferred to 480V Diesel Aux Bd B upon a loss of power from Aux Bd A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level* RO 290002 Reactor Vessel Internals Tier # 2 A2.O1 (IOCFR55.41.5)

Ability to (a) predict the impacts of the following on the REACTOR Group # 2 VESSEL INTERNALS and (b) based on those predictions, use K/A # 290002A2.O1 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

LOCA Importance Rating 3.7 Proposed Question: # 64 Which ONE of the following completes the statement?

Jet Pumps are designed such that following a DBA LOCA, a re-floodable core volume NO lower than _(1)_ is assured. Following a DBA LOCA with ALL ECCS available, Severe Accident Management Guidelines _(2) be required to be entered.

A. (1) (-) 180 inches (2) will B. (1) (-) 180 inches (2) will NOT C. (1) (-)215 inches (2) will

b. (1) (-)215 inches (2) will NOT Proposed Answer: D Explanation A INCORRECT: Part 1 incorrect Plausible in that (-) 180 inches is a (Optional): recognizable value associated with Low Reactor Water Level accident conditions and criteria for adequate core cooling. This is the minimum zero injection water level limit. Part 2 incorrect Plausible in that a severe accident has occurred in a DBA LOCA and candidate may have the misconception that under these conditions SAMG entry is required regardless of whether adequate core cooling is met or not.

B INCORRECT: Part 1 incorrect See Explanation A. Part 2 correct See Explanation D.

c INCORRECT: Part 1 correct See Explanation D. Part 2 incorrect See Explanation A.

D CORRECT: Part 1 correct - Jet Pumps are designed such that following a DBA LOCA a re-floodable core volume NO lower than two thirds core height is assured. Two thirds core height corresponds to (-) 215 inches. Part 2 correct - ECCS is designed such that adequate core cooling will be met following a LOCA, assuming the worst case single active component failure in the ECCS. With all ECCS available, adequate core cooling is assure.

Therefore, SAMGs are not required to be entered.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates ability to predict the impacts of a LOCA on the Reactor Vessel Internals and based on those predictions, use procedures to control or mitigate the consequences of those abnormal conditions or operations in that the candidate must utilize the applicable sections and steps of EOl-1, RPV Control, and EOl-C1, Alternate Level Control to determine that these procedures will not be exited for the SAMGs based on current plant conditions and predicted impact.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.

Technical Reference(s): OPL1 71 .212, Rev. 4 (Attach if not previously provided)

OPL171.201 Rev. 7/OPL171.002 Rev. 9 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPLI 71 .212 V.B.2 (As available)

Question Source:

(Note changes or attach parent)

Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71 .002 Revision 9 Page 42 of 82 Instructor Notes lx Flows out:

(1) Steam flow:

1415 x E+6 lbs/hr (2) Flow to Cleanup System:

013 x E÷6 lbs/hr (3) Total flow out:

1418 x E÷6 lbs/hr a Flows in:

(1) Feedwater flow:

1410 x E+6 lbs/hr (2) Control Rod Drive System:

0.05 x E+6 lbs/hr (3) RWCU System return water flow:

0.13 x E÷8 lbs/hr (4) Total flow in:

14.28 x E+6 lbs/hr

2. Core Floodability TP-26

a. Applicability Obj. V.B.7 (1) Applicable to a loss of coolant Obj. V.C.7 accident.

(2) The worst case loss of coolant accident is a 28 recirculation suction line break with the reactor at full power, steady state.

(3) In this case the core will become completely uncovered.

(4) This will be discussed in detail during the Emergency Core Cooling System presentation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.002 Revision 9 Page 43 of 82 Instructor Notes

b. Design features (1) The emergency core cooling systems and the reactor vessel design must be compatible so that following a loss of coolant accident the core can be adequately cooled.

(2) There are several systems that will Procedure Use:

provide water to the reactor following a s

1 EOl loss of coolant accident.

(3) One of these systems is the low Pressure Coolant Injection System (LPCI) mode of RHR.

(4) For simplification only the LPCI system will be discussed here.

(5) The LPCI system injects water into the reactor vessel using the RHR pumps via both recirculation inlet lines and down the 20 jet pumps.

(6) This flooding water then increases the UNIT water level in the reactor starting at the DIFFERENCE bottom of the vessel and working its way up into the core. Calculations in the FSAR show that leakage through slip fit (and unit I bolted accesses) into the downcomer will not exceed 964 gpm (unit 1) or 807 gpm (units 2 & 3) while level is being restored.

(7) When the water level reaches the top of the jet pump mixing sections, water will begin spilling out into the downcomer area and out of the vessel through the broken recirculation line.

h.

(8) This elevation where water begins to spill out of the jet pumps is 2/3 of the height of the active fuel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.002 Revision 9 Page 44 of 82 Instructor Notes (9) Calculations show that if flooding of

{

the reactor vessel is accomplished NOTE: These within a specified time frame & the Calculations are level maintained at the 2/3 point, the based on FSAR core will be adequately cooled and not BFNP EOl indefinitely and the integrity of the Program Manual.

fuel cladding maintained.

(a) Lower 213 of the core cooled because it is flooded with water.

(b) Upper 113 of the core (i) Vigorous boiling in the Fundamental:

lower 2/3 of the core What are the 3 provides a mixture of types of heat steam and water transfer and which which, upon flowing is prevalent during upward, cools the this condition?

upper 1/3 of the core.

1. Radiation (ii) Long term (after fuel decay heat has 2. Conduction lowered) there will be 3. Convection less boiling in the lower 2/3 of the core to provide the flow of steam and water to cool the upper 1/3 of the core.

(iii) Fuel clad temperature would raise with time.

However, it would still remain acceptable under these conditions.

(10) Under the assumed conditions, water would have to be continually made up to the vessel to accommodate for the following cooling losses:

(a) Boil off AND, (b) The aforementioned leakages.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.201 Revision 7 Page 27 of 8 A. KeyWordsandTerms Obj.V.B.10 Section I-C to the Program Manual (see Attachment

1) provides definitions for terms, phrases, and acronyms used in the EOIs. The following terms/phrases are to be highlighted in this lesson:

a. Adequate Core Cooling Obj. V.B.10.a Any of the following conditions (1-4):

(1) Submergence: Reactor water level is verified at or above TAF, and based on present and past trends and plant conditions, is expected to remain above TAF.

(2) Spray Cooling: During the execution of Cl, the following conditions are met:

• The reactor can be determined to be shutdown without boron (note 1)

AND

• One Core Spray subsystem is One spray ring for P

injecting at or above 6250 gpm. design pattern AND

• RPV water level can be P

determined to be above -215 inches (2/3 core height)

(3) Steam Cooling With Injection:

• During execution of C5 and Cl, This will maintain RPV water level can be PCT < 1500 °F maintained above the lower water level band allowed by the procedure, [Minimum Steam h.

Cooling Water Level (MSCWL) -

180 inches].

OR

• Reactor pressure can be maintained above MARFP following reactor depressurization.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.212 Revision 4 Page 7 of 8 X. Lesson Body A. EOl Transition into SAMG Loss of Coolable Geometry

1. The SAMGs are entered, then the core geometry is assumed to be changed and NOT coolable. The j EOl strategies are employed for accidents inside BEN design basis. When accidents progress to a L 2.

point where BEN design basis is exceeded, SAMG entry will be required.

SAMG entry is required, i.e., core geometry Obj. V.B.1 assumed to be lost. These are the specific EOl contingency points:

a. In EOl Step C1-25, ALTERNATE LEVEL CONTROL, when primary containment flooding is required and either one Core Spray ioop is not injecting at >6250 gpm, or RPV water level cannot be determined to be above -215 inches.

b. In EOI Step C4-14, RPV FLOODING, when the reactor is NOT assured of remaining sub critical under all conditions and the RPV pressure due to injection will not remain above MARFP with at least four MSRVs open.

c. In EQI Step C4-24 and 25, RPV FLOODING, when the reactor will remain sub critical under all conditions and the RPV pressure due to injection will not remain 70 psig over suppression chamber pressure with at least four MSRVs open.

d. In EOl C5-26, LEVEL/POWER CONTROL, with control rods out and unable to restore and maintain RPV water level above -180 inches.

3. At each of these specific points, we cannot assume a coolable geometry exists and SAMG entry is required.

4. Once the SAMGs are entered, the EOI flowcharts Obj. V.B.2.c no longer apply because the configuration of the core may no longer be amenable to adequate cooling. All EOI flowcharts will be exited and will not be referred to again. Any subsequent EOI entry condition which is received will NOT result in EOI entry.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI 71 .201 Revision 7 Page 27 of 8 B. KeyWordsandTerms Obj.V.B.1O Section I-C to the Program Manual (see Attachment

1) provides definitions for terms, phrases, and acronyms used in the EOls. The following terms/phrases are to be highlighted in this lesson:

a. Adequate Core Cooling Obj. V.B.10.a Any of the following conditions (1-4):

(1) Submergence: Reactor water level is verified at or above TAF, and based on present and past trends and plant conditions, is expected to remain above TAF.

(2) Spray Cooling: During the execution of Cl, the following conditions are met:

• The reactor can be determined to be shutdown without boron (note 1)

AND

• One Core Spray subsystem is One spray ring for injecting at or above 6250 gpm. design pattern AND

• RPV water level can be determined to be above -215 inches (2/3 core height)

(3) Steam Cooling With Iniection:

• During execution of C5 and Cl, This will maintain RPV water level can be PCT < 1500 °F maintained above the lower water level band allowed by the procedure, [Minimum Steam Cooling Water Level (MSCWL) -

180 inches].

OR

• Reactor pressure can be maintained above MARFP following reactor depressurization.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 290003 Control Room HVAC Tier# 2 K6.O1 (10CFR 55.41 .7)

Knowledge of the effect that a loss or malfunction of the following Group# 2$

will have on the CONTROL ROOM HVAC: KIA# 290003K6.01

. Electrical power Importance Rating 2.7 Proposed Question: # 65 Which ONE of the following combinations of electrical board losses would result in BOTH Control Room Emergency Ventilation Fans being de-energized? (Assume normal alignment)

A. 480V Shutdown Board 1 B; 4kV Shutdown Board 3EC B. 480V Shutdown Board IA; 480V Shutdown Board 2B C. 480V Shutdown Board 3B; 4kV Shutdown Board A D. 4kV Shutdown Board B; 4kV Shutdown Board 3EA Proposed Answer: C Explanation A INCORRECT: These do not meet the combination of power supplies for (Optional): the CREV trains.

B INCORRECT: These do not meet the combination of power supplies for the CREV trains.

C CORRECT: Correct since the power supplies are 480 VAC RMOV Board 3B for fan B and 480 VAC RMOV Board 1A for fan A which is supplied by 4KV Shutdown Board A D INCORRECT: These do not meet the combination of power supplies for the CREV trains.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests whether the candidate has knowledge of the effect that a loss or malfunction of Electrical power will have on Control Room Emergency Ventilation.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPLI 71 .067, Rev. 16 (Attach if not previously provided) 0-01-31 Att 3 Rev. 133 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL17I.067 V.B.2 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam BEN 2004-301 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Attachment 3 OOI-31lATT-3 Unit 0 Elactncal Lineup Checklist Rev. 0133 Page 6 of 22 4.0 ATTACHMENT DATA Performed On:

PanellBreaker Required Initials Number Component Description Position lstlIV Control Bay 4160V Shutdown Board B El 593 18 O-BKR-031.-2100 CLOSED 4KV SUPPLY FOR 1&2 j CO NTROL BAY CH LER A Control Bay 4160V Shutdown Board D El 593 12 1 O-BKR-O3i.22OQ CLOSED 4KV SUPPLY FOR 1&2 CONTROL BAY CHILLER B Control Bay 480V Reactor MOV Board IA El 621 1A SHUTDOWN BOARD ROOMS ON EXHAUST FAN IA OA i-BKR-031-2300 ON.

R9A I 250V SHUTDOWN BD ON BATTERY ROOM EXHAUST 1 FANIA R9B 250V SHUTDOWN BD ON BATTERY ROOM SUPPLY FAN IA R9D1 250V SHUTDOWN BD :ON BATTERY ROOM DUCT HEATER 14D AUXILIARY PRESSURIZATION I OFF 1

FANA 14C O-BKR--31-7214 CREVS FILTRATION UNITA I ON Leads are hfted at breaker per DCN W1752L Fan is thoperable.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Attachment 3 0-Ol-3IIATT-3 Unit 0 Electrical Lineup ChecklIst Rev. 0133 Page 10 of 22 4.0 ATTACHMENT DATA (continued)

Performed On:

PanellBreaker Required Initials Number Component Description Position lst/lV Control Bay 480V Reactor MOV Board 3B El 59Y lOB 3-BKR-031-720$ ON ELEC BD RM ACU 3B I F 13C2 0-BKR-031-7213 ON CREVS FILTRATION UNIT B 17A 3-BKR-O3lO139 ON UNIT 3 CONTROL BAY SUPPLY FAN3B -F R9A  : 3-BKR-031-0651 ON SDBR CHILLER 3A-2 R9B1 3-BKR-031-0667 ON SDBR CHILLER 3B-2 R9B2 3-BKR-031-0608  : ON DIVISION II DUCT HEATERS F FL SHUTDOWN BOARD ROOMS UNIT 3 F R9C 3-BKR-031-0645 CHILLED WATER CIRC PUMP 3A-2 ON SHUTDOWN BOARD ROOMS UNIT 3 1 CHILLED WATER CIRC PUMP 3B-2 SHUTDOWN BOARD ROOMS UNIT 3 R9E 3-BKR-031-061 1 ON AIR HANDLING UNIT 3A-2 SHUTDOWN BOARD ROOMS UNIT 3 R9F 3-BKR-031-0612 ON AIR HANDLING UNIT 3B-2 SHUTDOWN BOARD ROOMS UNIT 3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI 71.067 Revision 16 Page 5 of 6 Control Room Emergency Ventilation (CREV) is Tech. Spec. 3.7.3 designed to supply and process the outdoor air Obj.V.B.2/ V.C.6 needed for pressurization during isolated conditions. N.C.7 There are 2 CREV units rated at 3000 cfm each. A (Old CREV Units CREV unit consists of Motor-driven fan, (power abandoned in place as P. supply is from 480V RMOV Bd 1A for CREV Fan A; Auxiliary RMOV Bd 3B for CREV Fan B), HEPA filter Pressurization (common), charcoal filter assemblies located in the Systems)

CREVS Equipment Room, charcoal heater, and inlet TP-4 isolation damper and a backflow check outlet 2-47E2865-4 damper. They are designed to maintain a positive pressurization to 1/8 w.g. minimum to the control room.

a. A CREV may be started manually from control Red indicating lights room Panel 2-9-22 if local control switch is in on panel 3-9-2 1 to AUTO position via a 3 position, spring-return provide indication of to center switch. (STOP-AUTO-START). CREV Fan A and/or B Actuates only the CREVS unit & associated running on Unit 3.

damper, not the isolation dampers. Annunciators are on

b. There is also a 2 position maintained contact, panel 9-6 for all units.

one per train, AUTO-INITIATE! TEST switch which is used to perform system level actions for that train (primarily testing). It provides the same response as auto start.

c. Local start at local control station in Relay Room is done using a 2 position maintained, one per train, AUTO-TEST switch. Isolation dampers do not operate automatically if started from local panel.

d. Automatic start signals are: Obj. V.B.1N.B.2 (1) High radiation of 221 cpm above Obj. V.C.1 background + 2 Mm TD (270 cpm Obj. V.C.17 Tech Specs) in air inlet ducts to Control Room from (Radiation monitor T. S. 3.3.7.1 RM 90-259A Units 1 & 2, Radiation monitor RM 90-259B Unit 3). Either monitor starts selected CREV unit.

(2) Reactor zone ventilation systems radiation high ?72 MR/hr

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Browns Feny Nuclear Plant 2004-301 SRO Inital Exam

42. 2XOOOK6,Q OO1T2G2iNENTILATJONIMEM2 7/2 7iB/aFo43ojRcrcLc Which ONE of the following Combinatbns of electrical board losses would result in both CREV units being inoperable? (Assume normal alignment and no board transfers)

A. 480V Shutdown Board 18; 4kV Shutdown Board 3E0

, 480V Shutdown BoardlA; 480V Shutdown Board 28 C 480V Shutdown Board 38; 4kV Shutdown Board A

5. 4kv Shutdown Board 8; 4kV Shutdown Beard 3EA KIA 288000 K6A)1 Knowledge of the effect that a loss or malfunction of the following will have on the PLANT VENTILATION SYSTEMS: AC. electrical (2.712J)

References:

OPLI7I.067,Revil.Pg 28 cf 60 Learning Objective #B2 A, B, and D. Incorrect since these do not meet the combination of power supplies for the ORV trains.

C. Correct since the power supplies are 480 VAC RMOV 8oard38 for fan B and 480 VAC RMOV Board lAforfanAwhich is supplied by4KV Shutdown BoardA

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.1.25 (10CFR55.41.1O) Tier# 3 Ability to interpret reference materials, such as graphs, curves, tables, etc. Group#

KIA# G2.1 .25 Proposed Question: # 66 Importance Rating 3.9 CURVE 6 PRESS SUPPR PRESS M

T 0

30 REQUEJ 1_

J--

25 20 a

uJ a:, 15 7 -v--

j SAFE C) 10 5

0 4

1 4

I I

4 11 115 12 13 14 15 16 17 18 19 20 SUPPR PL LVL (Fr)

Which ONE of the following completes the statement?

In accordance with the EOl Program Manual derivation, Line on Curve 6, Pressure Suppression Pressure, above, corresponds to the Suppression Pool Water Level at which the A. Downcomer Vents become uncovered B. HPCI Turbine Exhaust opening becomes uncovered C. Safety Relief Valve (SRV) Tailpipe openings become uncovered D. Control Room Suppression Pool Water Narrow Range Level Indication goes off scale low Proposed Answer: A Explanation A CORRECT: (See attached excerpt) According to the EOl Program Manual, (Optional): 11.5 feet (or Line 4) is the Suppression Pool Water Level which corresponds to the elevation of the downcomer vent openings.

B INCORRECT: The HPCI Turbine Exhaust becomes uncovered in the range of but above this value (at 12.75 feet) and is a significant direct Suppression Chamber Air Space pressurization event if HPCI remains running. PSP would be quickly exceeded.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet c INCORRECT: SRV Tailpipes become uncovered around 5.5 feet. This is plausible because of the required ED at 11.5 feet. Normally, an ED on a parameter such as this is accomplished before you lose the ability to do so safely (within Safety Analyses assumptions).

D INCORRECT: Plausible because the X-Axis is based upon Suppression Pool Water Level and Narrow Range goes off-scale low at -25 inches which corresponds to approximately 13 feet.

KA Justification:

The KA is met because the question tests the candidates ability to interpret Pressure Suppression Pressure Curve bounding limitations on Suppression Chamber Pressure versus Suppression Pool Level.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

RO Level Justification: Tests Technical Reference(s): OPLI 71.201, Rev. 7 (Attach if not previously provided)

EOl Program Manual Sect. 2-Vl-H, Rev. 10 0-Tl-394, Rev. 4 Proposed references to be provided to applicants during examination: Embedded EOl Curve 6 PSP -

Learning Objective: OPL17I.201 V.B.12 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: This question was originally developed for an Audit Exam.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EOI PROGRAM MANUAL PRESSURE SUPPRESSION PRESSURE WORKSHEET 8 SECTION 2-Vt-H 5.0 CALCULATIONS The derivation of the PSP is shown graphically in Figure 2. Line 1 corresponds to the highest suppression chamber pressure which can occur without steam in the suppression chamber airspace. This pressure is determined by calculating the pressure that would exist as a flmction of suppression pool water level with all drywell noncondensibles purged to the suppression chamber and suppression pool temperature at the Heat Capacity Temperature Limit corresponding to the lowest SRV lift pressure. Higher suppression pool water levels result in higher pressures since the airspace volume is smaller.

Line 2 corresponds to the highest suppression chamber pressure from which an emergency depressurization will not raise suppression chamber pressure above Primar Containment Pressure Limit A before RPV pressure drops to the Minimum RPV Flooding Pressure. This curve is calculated by subtracting the rise in suppression chamber pressure during blowdown from Primary Containment Pressure Limit A. The calculation assumes the blowdown is initiated at the lowest SRV lift pressure and compensates for changes in suppression pool heat capacity with changes in suppression pool water level (as defined by the Heat Capacity Temperature Limit). As suppression pool water level increases, a larger heat sink is available to absorb blowdown energy.

Consequently, the difference in suppression pool temperature before and after the blowdown decreases, causing the rise in suppression chamber pressure to decrease. Since Primary Containment Pressure Limit A is constant in this range, Line 2 rises with increasing suppression pool water level.

Line 3 corresponds to the highest suppression chamber pressure at *hich SRVs can be opened without exceeding the suppression pool boundary design load. This curve is the suppression pool boundary design pressure less (I) the suppression pool boundary loads imposed by SRV actuation and (2) the hydrostatic head between the suppression pool water level and the level assumed in the desiRn calculation.

I.

Line 4 is the suppression pool water level corresponding to the elevation of the downcomer vent openings. If suppression pool water level is below this elevation, the RPV may not be kept in a pressurized state since steam discharged through the vents may not be condensed. The PSP is therefore vertical at this elevation.

Line .5 is the suppression pool water level corresponding to the Maximum Pressure Suppression Primary Containment Water Level, Above this elevation, the pressure suppression function of the containment cannot be assured. The PSP is therefore vertical at this elevation.

The PSP is thus the envelope defmed by Lines 4 and 5 and the most limiting values of Lines 1, 2. and 3. As shown in Figure 2, Line 1 is most limiting over the range of SECTION 2-Vt-H PAGE 10 OF 33 REVISION 10

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PRESSURE SUPPRESSION PRESSURE WORKSHEET 8 EOI PROGRAM MANUAL SECTION 2.VIH suppression pool water levels considered.

A personal coniputer running Microsoft ExceP is used to compute the results of this calculation as one might use a hand calculator. Three or hw significant figures are sufficient to obtain reasonable EPO/SAG Appendix C calculation results. The personal computer carries more significant figures and hence is more accurate. Since the results given in this calculation are based on the precision resident in the computer, any band calculations using the asdisplayed precision of the data shown herein may yield results which are less precise.

Tables 1 contains a list of abbreviations employed for parameter notation. Table 2 identifies the notation tued for the properties of water.

Figure 2 PSP flerivatlon Littel j___Liie2 H--Lüie4 I Suppresiau ?Gl Vater Leel (ft)

REVISiON 10 PAGE 11 OF 33 SECTION 2-V[H

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CPU 71.201 Revision 7 Page$8 o1117 NSTP.UCTOR NOTES

+ That iFilfialsiippression chamber pressure dF11Th1t111 which, if RPV depressurization was initiated and allovyd to continue until RPV pressure reaches the Minimum RPV Flooding Pressure (90180170 psig), would cause suppression chamber pressure to reach the Primary Containment Pressure Umit. This initial allowed pressure decreases with increasing suppression pool level due to the larger heat sink available.

OR

• That suppression chamber pressure which nfflThlling can be maintained without exceeding the suppression pool boundary design load if SRVs are opened. This pressure decreases with increasing suppression pool level.

Pressure Curve is to determine if the pressure suppression capability has been degraded and to preclude containment failure due to exceeding design loads and the primary containment pressure limit.

comprised ofthree segments:

rsanie-sionpooW suppression chamber downcomer openings (11.5ff) 1 the pressure suppression function ofthe suppression chamber cannot be assured. Any steam produced as a result of a leak or break would be directed through the downcomers and pressurize the suppression chamber directly. If suppression pool level is i found to be atthis level, Emergency L RPV depressuhzation is initiated.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Browns Ferry Nuclear Plant Unit 0 Technical Instruction O-Tl-394 Technical Support for Severe Accident Management Guidelines (SAMG)

Revision 0004 ILLUSTRATION I EXCERPT

{

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outhne Cross-reference Level RO SROI G2 1 27 (CFR 41 7) Tier # 3 Knowledge of system purpose and/or function Group #

K/A# G2.1.27 A

Importance Rating 39 Proposed Question: # 67 Which ONE of the following is a Design Basis of HPCI?

A. Maintain sufficient reactor water inventory so the fuel wont overheat when a reactor isolation AND loss of feedwater occurs.

B. Make up water to the vessel in the event of a loss of coolant situation that does NOT result in rapid vessel depressurization.

C. Assures that the reactor core is adequately cooled to limit fuel clad temperature to < 1800 °F in the event of a large break in the reactor coolant system.

D. Assures that the reactor core is adequately cooled to limit primary containment pressure in the event of a small break in the reactor coolant system Proposed Answer: B Explanation A INCORRECT: Maintains reactor water inventory so the fuel wont overheat (Optional): is true, but this statement is the design basis for RCIC. Candidate may confuse the basis for HPCI and RCIC because they are similar in many respects. HPCI can also supply water to the reactor when a MSIV isolation and a loss of feedwater occur.

B CORRECT: Provides Adequate Core Cooling (ACC) for all break sizes that do NOT result in rapid depressurization of the reactor vessel. Correct design basis statement.

C INCORRECT: ECCS general design criteria is to limit fuel clad temperatures < 2200 °F. 1800 °F is EOl MZIRWL fuel clad temperature.

Candidate may confuse EOI zero injection water level fuel clad temperature with ECCS design value.

D INCORRECT: HPCI design basis isnt about limiting primary containment pressure. Candidate may confuse primary containment design criteria with HPCI.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The question meets the K/A by asking the design basis of HPCI.

Question Cognitive Level:

This is a low cognitive question, It asks for recall of the basis of the system or discrete bits of information.

Technical Reference(s): OPL1 71.042 Rev 20 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: V.B.1 (As available)

Question Source: Bank # Quad Cities 98 (Note changes or attach parent)

Question History:

Last NRC Exam Quad Cities 1998 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .042 Revision 20 Page lOaf 69 INSTRUCTOR NOTES X. LESSON BODY:

A. General Description The High Pressure Coolant Injection System (HPCI) Obj. V.B.1 consists of a steam turbine-driven system driving a Obj. V.C.1 constant-flow pump assembly to inject either Condensate Storage Tank (CST) water or Suppression Pool (Torus) water into the reactor under emergency conditions at the rate of at least 5000 gpm over an 1174 -150 psi reactor pressure range.

1. System Design Basis

a. To provide adequate core cooling for all Obj. V.E.1 break sizes which do not result in rapid SER 3-05 depressurization of the reactor vessel

b. Ta function independent of off-site power SER 3-05 sources and diesel generators

2. Components Obj. V.0.1, Obj. V.E.2

a. Turbine

b. Main and booster pumps

c. Turbine auxiliaries

3. Flow Path TP-1 Obj. V.C.1

a. One 100% system Obj. V.B.1 Obj. V.E.10

b. Steam path (1) From B Main Steam line upstream of the flow restrictor (2) Through isolation valves (3) Through stop valve and control valves

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL1 71.040 Revision 23 Page 11 of 74 X. LESSON BODY A. General Description DCNs 51149, 51196, 51220, 51236 Make Ui, U2, U3 the same.

1. The purpose of the RCIC System is to provide a source TP-1 of high pressure coolant makeup to the reactor vessel In Obj. V.8.1.

case of a loss of feedwater flow. The system is used to Obj. V.E.1 maintain the reactor water ievel and for reactor pressure control under MSIV isolation conditions and lass of normal feedwater,

2. Safety Design Basis RCIC operates automatically to maintain sufficient coolant in the vessel so that the fuel will not overheat In the event of reactor isolation and loss of feedwater flow.

The system is a consequence limiting system rather than an ECCS system.

B. The RCIC System consists of: Obj. V.D.1 ObJ. V.E.2

1. Turbine-driven pump located in basement of Reactor Building (Elev. 519)

2. Turbine is driven by steam from Main Steam Line C and Obj. V.8.2.

exhausts to the suppression pool.

3. Pump Is normally lined up to take suction from the Obj. V.E.3 Condensate Storage Tank (CST), but can take suction from suppression pool (only done manually).

4. Pump discharges to reactor via feedwater line B.

a. Turbine Obj. V.8.1.

TP-1 & TP-2 (1) 100% capacIty (2) Delivers full pump design flow at reactor pressures of 150 to 1120 psig (3) 500 hp at 1200 psig to 80 hp at 225 psig

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI71 .201 Revision 7 Page 5 of 5 During C4 execution it has been verified that based on present and past trends and plant conditions, either Minimum Reactor Flooding Pressure (all rods in) or Minimum Alternate Reactor Flooding Pressure (all rods not in) can be maintained.

(1) Steam Coolinci Without Iniection: This will maintain During Cl execution RPV water level PCT < 1800 °F has not yet lowered to [Minimum Zero Injection Water Level (MZIWL) -200]

and reactor pressure is either at the lifting point of the MSRVs or is stabilized and not rising.

b. Augment Obj. V.B.10.b (1) To supplement the systems that are currently in use.

(2) Augment RPV water level with the following systems:

c. Verify Obj. V.B.10.c (1) To observe an expected characteristic or condition and, if not as expected, to take action to place it in the expected condition. Usually applied for response to automatic actions, but is not limited to only those actions.

(2) Verify recirc flow runback to minimum.

d. Injection Subsystem Obj. V.B.10.d Any of the following independent flow paths capable of delivering coolant to the RPV:

(1) Condensate System, with at least one Condensate pump and one Condensate Booster pump capable of delivering Coolant to the RPV.

(2) LPCI System I, or II, with at least one operable pump capable of delivering Coolant to the RPV is one Injection Subsystem.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference Level RO G2 1 28 (10CFR 5541 7) Tier# 3 Knowledge of the purpose and function of major system components and controls. Grou p #

KIA# G2.1.28 Importance Rating 41 Proposed QuesUon: # 68 Which ONE of the following defines the purpose of the Rod Worth Minimizer (RWM) in accordance with Technical Specifications?

A. Ensures that fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when Reactor Power is < 10%.

B. Ensures that fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when Reactor Power is > 27%.

C. Ensures that the Minimum Critical Power Ratio remains greater than 1.08, while withdrawing control rods, when Reactor Power is < 10%.

D. Ensures that the Minimum Critical Power Ratio remains greater than 1.08, while withdrawing control rods, when Reactor Power is > 27%. -

Proposed Answer: A Explanation A CORRECT: The purpose of the RWM system is to limit control rod worth (Optional): such that the fuel enthalpy limit of 280 cal/gm will not be exceeded during a Control Rod Drop Accident (CRDA). TS Table 3.3.2.1-1 requires the RWM to be operable in modes I and 2 with thermal power <10% RTP.

B INCORRECT: 1st part correct. 2nd part incorrect Plausible in that 27%

is the TS requirement for the RBM, and the candidate may confuse the requirements between the RBM and RWM.

C INCORRECT: 1st part is incorrect. Plausible because the RBM does provide rod blocks to prevent MCPR from being exceeded due to additional rod withdrawal. 2nd part is correct.

D INCORRECT: 1st part is incorrect. Plausible because the RBM does provide rod blocks to prevent MCPR from being exceeded. 2nd part is incorrect. Plausible because 27% is the TS requirement for the RBM, and the candidate may confuse the requirements between the RBM and RWM.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of the purpose of the Rod Worth Minimizer.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL171.024 Rev. 14 (Attach if not previously provided)

TS 3.1-20 Amm 253 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.024V.B.1 /3 (As available)

Question Source:

(Note changes or attach parent)

Question History: Last NRC Exam Hatch 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS),

I.

APPLICABILITY: MODES 1 and 2 with THERMAL POWER 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OPERABLE A.1 control rods not in Rod worth minimizer compliance with BPWS. (RWM) may be bypassed as allowed by LCO 3.3.2.1, ControL Rod Block Instrumentation.

Move associated control 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rod(s) to correct position.

OR A.2 Declare associated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control rod(s) inoperable.

(continued)

BEN-UNIT 2 3.1-20 Amendment No. 253

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71.024 Revision 14 Page 9 of 58 INSTRUCTOR NOTES X. Lesson Body A. Banked Position Withdrawal Sequence (BPWS) TP-1 Basis The BPWS is designed to ensure under all

- Obj. V.B. 1 operating conditions that control rod worths are SER 03 05 limited such that a control rod drop accident would result in peak fuel energy deposition of less than 280 cal/gram (safety basis). Restrictions on control rod patterns while at low power are required during both startup and shutdown.

2. Review Tech Spec and Bases Section 3.1.6.1 -

Note: This is required for both ROs and SROs.

3. Fuels group in Chattanooga designs the Control QUESTIONING Rod Withdrawal Sequence such that no single ATTITUDE control rod notch will cause less than a 60 second reactor period.

B. Sequence Types Obj. V.B.2 Sequence A This sequence results in the center

- SR-3.1 .3.5(A) control rod (30-31) being fully inserted when 50 percent control rod density (black and white pattern) is reached. Al versus A2 sequences differ only in which control rod groups beyond the black and white pattern will ultimately be deep and which will be shallow or fully withdrawn.

2. Sequence B This sequence results in the center control rod being fully withdrawn when 50 percent control rod density is reached. B1 versus B2 sequences differ as described above for Sequence A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPLI71.024 Revision 14 Page 11 of 58 tNSTRUCTOR NOTES

5. When all the control rods in Group I have been withdrawn to the Group 1 withdraw limit, the oparator proceeds to Group 2.

6. After Group 1 control rods have been withdrawn, a Example: TP-2 given control rod or set of control rods may Gps 2- 6 same rods comprise more than one group.

Gps 7 -12 same rods

7. In this case the withdraw limit for a control rod in a Gpsl3- 17 same given group will be the same as the insert limit for rods the next higher group in which the control rod appears.

D. Reduced Notch Worth Procedure (RNWP) TP- 1

1. Basis The RNWP is a conservative extension of the BPWS. and is designed to further lower notch NOTE: Single worth in order to reduce the chance of a scram on notch withdrawals short period during startup. Since this is not a concern during shutdown, RNWP procedures should never result need not be utilized except for startup pull sheets. in a reactor period faster than 60 sec.

2. A high notch worth control rod is designated by an 2-MINUTE asterisk on the control rod withdrawal sequence SITUATIONAL sheet. The designated high worth control rod must AWARENESS be withdrawn a single notch at a time within the indicated range of high notch worth.

E. Rod Worth Minimizer Purpose and Terms SOER 84-2 Recommendation 7d h 1. The RWM system design is based on Banked Position Withdrawal (BPWS) system design requirements.

2. The RWM, in conjunction with the control rod OBJ. V.B.3 velocity limiter, limits the amount of fuel damage OBJ. V.C.1 that could occur during a control rod drop accident.

a. The RWM acts to enforce of the TP-3 programmed control rod patterns and generates a rod block if significant deviation from the programmed sequence is detected.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT Control Rod B[ock Instrumentation 33.2.1 Tatee 3.3.2.1-1 (page 1 & 1)

Control Rod Block Insinimentation APPLICABLE MODES OR FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED CHANNELS REQUIREMENTS VAWE CONDITIONS

1. RdBlockMonitor I.

a. Low Power Range Upscale

- (a) 2 SR 1321.1 (e)

SR 3.32t4 SR 3.32.1.8

b. InteedisePowerRange-Upseale (b) 2 SR 3,321.1 (e)

SR 3321.4 SR 3.321,8

c. lligflPower Range-Upscale (f), (g) 2 SR 3.3.2.1.1 (e)

SR 3.321,4 SR 3.32.1.8

d. Inop (g),(h) 2 SR 3,3.2.1.1 NA

e. Downscale g).(h 2 SR 3,3.2.1.1 Ii)

SR 3,321.4

2. Rod Worth rAnimizer (cl,(c) 1 1 SR 3.3.2.1.2 NA SR 3.32.1.3 SR 3321.5 SR 3.32.1.7

3. Reactor Mode Switch - Shutdown Positron (d) 2 SR 3.3.2.1.6 NA I.

(a) ThERMAL POWER 27% and 62% RTP and MCPR lass than the value specified in the COLR.

(b) ThERMAL POWER> 62% and 82% RTP and MCPR less than the value specified in the COIR.

(c) With ThERMAL POWER 10% RTP.

(d) Reactor mode switch in the shutdown position.

(e) Less Than or equal to the Allowable Value specified to the COLR.

(f) ThERMAL POWER> 62% and <90% RTP and MCPR less than the value specified in the COLR.

ig) THERWI. POWER 90% Ri? and MCPR less than the value specified in the COLR (h) THERMAL POWER 27% and <90% RTP and MCPR less than the value specified in the COtR

{i) Greater than or equal to the Allowable Value specified in the COIR.

BEN-UNIT 2 3.3-21 Amendment No. 253

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT Control Rod Block Instrumentation B 3.3,2.1 B 3.3 INSTRUMENTATION B 3.3,2.1 Control Rod Block Instrumentation P.

BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch Shutdown Position Function ensure that all control rods remain inserted to prevent Inadvertent criticahties.

The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. it is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel Inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn.

(continued)

BFN-UN1T 1 B 3.3-57 Revision O-4C October 26, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet HATCH 2009 HLT 4 NRC Exam

66. 62.1,27 001 Which ONE of the following defines the purpose of the Rod Worth Minimizer (RWM) lAW Technical Specifications.

A. Ensures that fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when power is > 29%.

B? Ensures that fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when reactor power is < 10%.

C Ensures that the Minimum Critical Power Ratio remains nreater than 1.08, while withdrawing control rods, when power is 29W 1). Ensures that the Minimum Critical Power Ratio remains greater than 1.08, while withdrawing control rods, when reactor power is < 10%.

Friday, May 01, 2009 8:37:25 AM 115

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.2.2 (10CFR55.41.6) Tier# 3 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power Group levels. KIA# G2.2.2 Importance Rating 4.6 I Proposed Question: # 69 Unit 1 Plant Startup is in progress.

Which ONE of the following identifies the criteria specified in 1-GOl-100-1A,Unit Startup, for Control Rod single notch withdrawal?

Control Rod withdrawal is limited to single notch when the _(1)_ SRM count rate doubling is reached AND must continue until _(2)_.

A. (1) fourth (2) the Reactor is Critical B. (1) fifth (2) the Reactor is Critical C. (1) fourth (2) Reactor Power is in the heating range D. (1) fifth (2) Reactor Power is in the heating range Proposed Answer: C Explanation A INCORRECT: Part 1 correct See Explanation C. Part 2 incorrect See (Optional):

Explanation B.

B INCORRECT: Part 1 incorrect Plausible in that Calculations have shown that when the initial SRM count rate has doubled 5 times that the reactor is very near criticality. Part 2 incorrect Plausible in that 1-GOI-1 00-lA contains several cautions regarding the careful and controlled approach to criticality and the point of criticality is the trigger for several actions in the GOl.

C CORRECT: Part 1 correct In accordance with 1-GOI-1 00-IA, A review of startup data has revealed that when count rate doubles five times, criticality is imminent. As an added precaution, the fourth count rate doubling has been chosen as a starting point to limit rod withdrawal to single notch movement. Part 2 correct In accordance with 1-GOI-100-1A, once required, Control rod withdrawal is limited to single-notch withdrawal until Reactor power is in the heating range.

D INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See Explanation C.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the candidates ability to manipulate Control Rod console controls as required based on SRM response to operate the facility between shutdown and designated power levels.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): 1-GOl-1 00-lA Rev. 23 (Attach if not previously provided)

OPL171.059 Rev. 11 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.059V.B.3!4 (As available)

Question Source: Bank# I Modified Bank # Nine Mile 2 08 #70 (Note changes or attach parent)

New:

Question History: Last NRC Exam Nine Mile 2 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 )(

55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL17 1.059 Revision 11 Page 12 of 23 INSTRUCTOR NOTES

2. Review instruction steps from 5.2 through 5.29.

a. SRM reading are recorded prior to start up to Ob. V.8.4 determine the count rate at which s;ngle notch Ob. V.0.4 withdrawal should begin. Calculations have shown that when the initial SRM count rate has doubled 5 times that the reactor is very near criticality, so when the initial count rate on any SRM has increased by a factor of 16 (four doublings) single notch withdrawal shall begin.

Criticality should be expected at all times. Expect the unexpected.

b. IRM downsca]e functions are bypassed on Obj. V.5.3 Range 1, so to venfy the downscale function Obj. V.0.3 operabe the IRMs will have to be ranged to Range 2 or 3.

c. For control rods that are difficult to move from Always return the the full iii position, increased drive water drive water DP to pressure is allowed by 01-85. should be referred normal after the rod is to in this situation, moved. This will prevent double notching.

d. Surveillance requirements for RWM are Operations completed prior to withdrawing control rods for Management the purpose of making the reactor critical. Expectation.

Ths is a so required by the (301

e. Control rods shall not be pulled for startup if the Obj. V.8.4 Plant Control Air is supplying the Dryvell Control Obj. V.0.4 Air System.

f. The Unit Operator is responsible for controlling reactivity and should be alert for any conditions Obj. V.8.4 that might affect reactivity. Any activity that Obj. V.0.4 could affect reactivity should be coordinated with Conservative Decision the operator. These activities v,ould include Making and Follow wcrc control changes, addition ot feedwater, Procedures.

use of nuclear steam. It is vital that good corn munications are exercised during these evolutions. The operator should be aware that a startup following operation at high power and peak Xenon could result in extremely high notch worth.

g. All activites that can distract the operator and Obj. V.8.4 supervisors during the approach to criticality Obj. V.0.4 should be avoided. These activities could include shift turnover, surveillances, and excessive personnel in the control room.

h. Verify moderator temperature is greater than the SR 3.4.9.2 temperature required byTS 3.4.9-1 Curve 3 Ex12-SR-3.4.9.i(i) within 15 minutes prior to withdrawing control (which is also the rods to achieve critical. heatup monitoring SR).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 71 .059 Revi&on 11 Page 13 of 23 INSTRUCTOR NOTES Performance of the. heautp and cooldown rate 2-SR-3.4.S.1{1) monitoring surveilLance is required 15 minutes prior to heatup and pressurization. Use HUR on ICS

3. Review instruction steps 5.29 through 5.42 SRO in CR

a. If a single notch withdrawal results in a reactor Ob. V.3.5.b period of less than 60 seconds, the iasI control ObL V.C.5.b rod pulled will be reinserted unth a period of greater than 60 seconds is obtained, the Reactor Engineer, Reactivdy Manager, and SM approval is required to resume rod withdrawal.

b. If a reactor period of less than 33 seconds is Ohj. V.E.5.c observed, control rods shall be inserted until the Obj. V.C.5.c reactor is subcritical, and obtain the Reactor Engineer, Reactivity Manager. and SM approval to resume rod withdrawal.

c. If a reactor period of less than 5 seconds is Ohj. V,3.5.d observed, the reactor shall be shut down and Obi. V.C.5.d cannot be restarted until an assessment has been performed.

d. Near end of core life, criticality may occur before Obj. V.3.4 five doublings due to a stronger top peak flux ObI. V.C.4 and the buildup of plutoniuni.

e. Single notch withdrawal must begin when the Obj. V.3.3 r SRM count rate has increased by a factor of 16 Obj. V.C.3 (four doublings), and may he stopped after (e through g) reaching the heat range.

f. The operator should expect the reactor to go Withdraw CR to critical at ANY TIME while pulling control rods maintain 100 second for startup. period as indiceted on the oeriod meter.

g. Inadvertent criticality could result from extended operation close to the point of criticality.

h. GE SIL 316 cautions when rod movement ;s Obj. V.3.4 restricted to single notch withdrawal failure to Obi. V.C.4 stop at each notch position may result in high notch worth.

L r,hen the reactor is critical and the desired Obj. V.3.6 period is obtained, the time, rod group, rod number, rod notch, and Reactor water temperature shall be recorded on data sheet and in the NOMS Narrative Log.

j. Reactor periods may he calculated by: Obj. V.3.S.a Obj. V.C.5.a (1) muktiplying the time for a 10% power rise by 10.5

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Unit Startup 1-GCl100-1A Unit I Rev. 0023 Page 78 of 171 5.0 INSTRUCTION STEPS (continued)

NOTE

[NERICJ A review of startup data has revealed that when count rate doubles five times.

criticality is imminent. As an added precaution, the fourth count rate doubli ng has been chosen as a starting point to limit rod withdrawal to single notch movement.

This requirement along with close monitoring of neutron monitoring instrumentatio n should assure a SIO\. controlled approach to criticality. Criticality should be expect

[SO ER 88-120 2] ed at all times.

[24.21 CALCULATE SRM count rate at which notch withdrawal limitations shall be imposed by multiplying pre-startup count rate, recorded in Step 5.0[24.1], by a factor of 16. RECORD results below and at Step 5.0[26):

CHANNEL A LEVEL cps CHANNEL C LEVEL cps CHANNEL B LEVEL cps CHANNEL D LEVEL cps (R)

Initials Date Time 1st (R

Initials Date Time Reactor Engineer

[24.3) RECORD SRM BYPASS, 1-HS-92-7A/S3 joystick position. (N/A if NOT bypassed.)

CHANNEL BYPASSED (R) initials Date Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Unit Startup l-GOI-100-IA UnitI Rev.0023 Page 81 of 171 5,0 INSTRUCTION STEPS (continued)

CAUTIONS

1) Near end of core life, criticality may occur before five doublings due to a stronger top peak flux and buildup of plutonium.

2) NERJC3 When rod movement is restricted to notch withdrawal, failure to stop at each notch position may result in high notch worth. 1GEs1316)

NOTE Once required. Control rod withdrawal is limited to single-notch withdrawal until Reactor power is in the heating range.

[26] WHEN SRMs indicate the calculated values recorded below:

CHANNEL A LEVEL cps CHANNEL C LE!EL cps CHANNEL8LEVEL cps CHANNEL D LEVEL cps THEN START single-notch withdrawal of control rods.

(R)

Initials Date Time 1st (R)

Initials Date Time Reactor Engineer

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NINE MILE 2 2008 Nine Mile Point Unit 2 Reactor Operator Written Examination Draft Submittal RO I Tier K/A Number Statement IR Origin Source Question 70 3 Generic 2.2.2 4.6 B NMP-2 Bank

SySID 22775 LOk Grp 10 CFR 55.41(b) 10 LCD (1-5) Reference Documents F NA N2-OP-l0lARevl4 I

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

QUESTION 70 Plant startup is in progress with the following:

• Mode switch is in Start/Hot Standby.

• RSCS Group 2 rods are being withdrawn using Continuous Withdrawal

• Reactor is Subcritical.

Which one of the following describes the criteria for using SINGLE NOTCH WITHDRAWAL per N2-OP-IOIA, Plant Startup?

A. Starting with RSCS Group 4 until criticality is achieved.

B. Starting with RSCS with Group 5 after the Reactor is critical.

C. When TWO SRM5 approach 3 count rate doublings in RSCS group 4, D. When TWO SRMs approach 3 count rate doublings prior to RSCS group 3.

Correct Answer: D When TWO SRMs approach 3 count rate doublings prior to RSCS group 3, SINGLE NOTCH WITHDRAWAL is required per N2-OR.1OIA, Plant Startup.

Plausible Distractors:

A through C are plausible; and ALL of these answer choices invoke Single Rod Withdrawal requirements too late in the startup process to meet the requirements of N2-OP-1OIA Page 81 of 88

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.2.39 (100FR55.41.7) Tier# 3 Knowledge of less than or equal to one hour Technical Specification action statements for systems. Group #

K/A # G2.2.39 Importance Rating Proposed Question: # 70 Which ONE of the following completes the statement?

In accordance with Unit 2 Tech Spec 3.4.10, Reactor Steam Dome Pressure, if the MAXIMUM Reactor Steam Dome Pressure of(1)_ is exceeded, it must be restored within a MAXIMUM completion time of (2).

A. (1) 1050 psig (2) 15 minutes:

B. (1) 1050 psig (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) 1073 psig (2) 15 minutes D. (1) 1073 psig (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Proposed Answer: A Explanation A CORRECT: Part 1 correct In accordance with Unit 2 Tech Spec 3.4.10, (Optional): the reactor steam dome pressure shall be S 1050 psig. Part 2 correct In accordance with Unit 2 Tech Spec 3.4.10 Condition A, if Reactor steam dome pressure not within limit, it must be restored with completion time of 15 minutes.

B INCORRECT: Part 1 correct See Explanation A. Part 2 incorrect See Explanation D.

C INCORRECT: Part 1 incorrect See Explanation D. Part 2 correct See Explanation A.

D INCORRECT: Part 1 is incorrect Plausible in that this is a recognizable value associated with Reactor Pressure, i.e. EOI entry. Part 2 incorrect Plausible in that 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is common completion time in Tech Specs.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of less than or equal to one hour Technical Specification action statements for TS 3.4.10, Reactor Steam Dome Pressure.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): U2 TS 3.4-30 Amm 254 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank#

Modified Bnk# (Note changes or attach parent)

New X V

Question History: Last NRC Exam i I (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of eveiy question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Piessure LCO 3.4.10 The reactor steam dome pressure shall be 1050 psig.

APPLICABILITY: MODES 1 and 2.

ACT] ONS CONDITION REQUIRED ACTION COMPLETION 1I I .1 I II%H A. Reactor steam dome A. 1 Restore reactor steam 15 minutes pressure not w]thin limit. dome pressure to within I imit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

BEN-UNIT 2 3.4-30 Amendment No. 254 September08, 1998

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT RPV CONTROL t

2-EO-1 PAGE 1 OF 1 RPV CONTROL UNIT 2 BROWNS FERRY NUCLEAR PLANT REV: 12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PLAUSIBILITY SUPPORT RHR Shutdown Cooling System Cold Shutdown 3.4 8 3.4 REACTOR COOLANT SYSTEM (ROS) 3.4.8 Residual Heat Removal (RHR) Shutdown Coolina System .- Cold Shutdovn LCO 3.4.3 Two RHR shutdown cooling subsystems shall be OPERABLE, and.

with no recirculation pump in operation, at least one. RHR shutdown coohng subsystem shall be in operation.

--- NOTES

1. Both required RHR shutdown cooling subs stems and recirculation pumps may not be in operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for performance of Surveillances.

APPLICABILITY: MODE 4.

ACTIONS NOTE Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQIJIRED ACTION COMPLETION TI ME A. One or two required RHR Al Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown cooling method of decay heat subsystems inoperable, removal is available for AND each inoperable required RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

BFN-UNIT 2 3.4-21 Amendment No. 253

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.4.43 (10CFR55.41.10) Tier# 3 Knowledge of the process used to track inoperable alarms.

Group #

K/A # G2.4.43 Importance Rating 3.0 Proposed Question: # 71 Which ONE of the following describes the meaning of a BLUE magnetic border being installed on a Main Control Room panel annunciator?

This type of border indicates that the annunciator A. has ONE OR more alarm inputs disabled B. is NOT ABNORMAL for current plant conditions C. is associated with ongoing testing OR maintenance D. window is being relocated to a different window location Proposed Answer: A Explanation A CORRECT: In accordance with Annunciator Disablement, OPDP-4, a blue (Optional): magnetic border indicates that an alarm is out of service.

B INCORRECT: In accordance with Annunciator System, 0-01-55, a hot pink border indicates that an alarm is NOT ABNORMAL for current plant conditions.

C INCORRECT: In accordance with Annunciator Disablement, OPDP-4, a white magnetic border indicates that an alarm is out of service for TESTING or MAINTENANCE.

D INCORRECT: In accordance with Annunciator System, 0-01-55, section 8.5, a yellow border is used to signify that an annunciator window is being relocated.

KA Justification:

The KA is met because the question tests knowledge of Annunciator Disablement, OPDP-4, process for tracking inoperable alarms.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPDP-4 Rev. 4 (Attach if not previously provided) 0-01-55 Rev. 46 Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source: Bank#

Mod ified Bank # BEN 1006 # 75 (Note changes or attach parent) r eW ,

Question History: Last NRC Exam Browns Ferry 1006 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Annunciator Disablement OPDP-4 Department Rev. 0004 Procedure Page 11 of 21 5.0 DEFINITIONS Disabled In put indicator

• BFN -A blue magnetic border labeled Disabled Alarm Input.

• SON-A biue dot (sticker) attached to the window with the SER point written on it.

• WBN-An orange plastic lens cover labeled Disabled Alarm which snaps over the affected window and a blue plastic lens cover labeled Disabled Input.

Out-of-Service indicator

• BFN -A white magnetic border labeled TestingiMaintenance.

• SON-An orange sticker attached to the window.

• WBN-A green plastic lens cover labeled Maintenance which snaps over the affected window.

Maintenance Activities Activities that restore components to their as-designed condition, including activities that implement approved design changes. Maintenance activities are not subject to 10 CFR 50.59. Maintenance activities include troubleshooting. calibration, refurbishment, maintenance-related testing, identical replacements, housekeeping and similar activities that do not permanently alter the design, performance requirements, operation or control of equipment. Maintenance activities also include temporary alterations to the facility or procedures that directly relate to and are necessary to support the maintenance. Examples of temporary alterations that support maintenance include jumpering terminals, lifting leads; placing temporary lead shielding on pipes and equipment.

removal of barriers, and use of temporanj blocks, bypasses, scaffolding and supports.

Nuisance Alarm An alarm that comes in repetitively due to an instrumentation problem.

or maintenance activity that detracts from the operators ability to monitor and control the plant.

Valid Alarm An alan-n that is actuated when the monitored parameter exceeds the setpoint or meets the intent of a setpoint (e.g. if a high pressure alarm occurs at 580# and the alam setpoint is 600# but pressure is normally zero or close to zero, that is a valid alanii. In a simUar scenaro, if pressure is normally *550#, the alarm may not be valid).

6.0 REQUIREMENTS AND REFERENCES Requirements and References are contained in the OPDP-4 REQ & REF document.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Annunciator System 0-01-55 Unit 0 Rev. 0046 Page 21 of 46 8.3 Identification of Out of Service Annunciators

. REFERENCE OPDP-4, Annunciator Disablement NOTES

1) This Section applies to annunciators which alarm or are in alarm status due to the present plant conditions (i.e., Modifications, extended Maintenance, alarms due to plant Mode, etc.).

2 These borders signify THESE ILLUMINATED ALARMS ARE ILLUMINATED DUE TO THE PRESENT PLANT CONDITIONS. and no operator action is recluired.

3 The diagonal bar in the Hot Pink border means NOT ABNORMAL for current plant conditions.

8.4 Identification of Lit Annunciators for Normal Plant Conditions

[1] PLACE Hot Pink identification border on each applicable annunciator window.

[21 WHEN conditions of the plant change such that the annunciator will no longer remain illuminated as a normal condition, THEN REMOVE the Hot Pink identification border from each applicable annunciator window. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Annunciator System 0-01-55 Unit 0 RevS 0044 Pane 22 of 45

&5 identification of Nuisance Alarms 8.5i Short Term Nuisance Alarms 4 REFERENCE OPDP-4, Annunciator Disablement NOTES r) This section applies to annunciators which are modified or being modified) to the new annunciator systenL

2) A.1 anntnciator relccaion perfornied by this procedure is temporary and is performed in accordance with the \ork Order Process..

The Yellow borders for identification of relocated windows communicate to oersonne; the correct annunciator resoonse procedure for relocated annunciators and are required to meet the foilowinci criteria:

• YeLow in color,

• The temporary location is delineated on the top border,

• The correct ARP is referenced for response on the bottom border.

4) The new window annunciator iocatioWs) are updated to reflect the same descnption as used in the original annunciator window location(s).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT NPG Standard Annunciator Disablement OPDP-4 Department Rev. 0004 Procedure Page 11 of 21 5.0 DEFINITIONS Disabled Input Indicator

• BEN -A blue magnetic border labeled Disabled Alarm Input

• SON-A blue dot (sticker) attached to the window with the SER point written on it.

• WBN-An orange plastic lens cover labeled Disabled Alarm which snaps over the affected window and a blue plastic lens cover labeled Disabled Input.

Out-of-Service Indicator

• BEN -A white magnetic border labeled TestingfMaintenance

• SON-An orange sticker attached to the window.

• WBN-A green plastic lens cover labeled Maintenance which snaps over the affected window.

Maintenance Activities Activities that restore components to their as-designed condition, including activities that implement approved design changes. Maintenance activities are not subject to 10 CFR 50.59. Maintenance actMties include troubleshooting, calibration.

refurbishment, maintenance-related testing, identical replacements, housekeeping and similar activities that do not pemianently alter the design, performance requirements, operation or control of equipment. Maintenance activities also include temporary alterations to the facility or procedures that directly relate to and are necessary to support the maintenance. Examples of temporary alterations that support maintenance include jumpering terminals, lifting leads, placing temporary lead shielding on pipes and equipment, removal of barriers, and use of temporary blocks, bypasses, scaffolding and supports.

Nuisance Alarm An alarm that conies in repetitively due to an instrumentation problem, or maintenance activity that detracts from the operators ability to monitor and control the plant.

Valid Alarm An alarm that is actuated when the monitored parameter exceeds the setpoint or meets the intent of a setpoint (e.g. if a high pressure alarm occurs at 5S0# and the alarm setpoint is 600ff but pressure is normally zero or close to zero, that is a valid alarm. In a similar scenaro, if pressure is normally 550ff, the alarm may not be valid).

6.0 REQUIREMENTS AND REFERENCES Requirements and References are contained in the OPDP-4 REO & REF document.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY 1006 Examination Outline Cross-reference: Level RO SRO G2.4.45 (10CFR55.41.1O) Tier# 3 Ability to prioritize and interpret the significance of each annunciator or alarm. G roup K/A # G2.4.45 Importance Rating 4.1 Proposed Question: # 75 Which ONE of the following describes the meaning of a WHITE magnetic border being installed on a Main Control Room panel annunciator?

This type of border indicates that the annunciator A. has ONE OR more alarm inputs disabled B. is associated with ongoing testing OR maintenance C. is NOT ABNORMAL for current plant conditions D. window is being relocated to a different window location Proposed Answer: B Explanation A INCORRECT: In accordance with Annunciator Disablement, OPDP-4, a (Optional): blue magnetic border indicates that an alarm is out of service.

B CORRECT: In accordance with Annunciator Disablement, OPDP-4, a white magnetic border indicates that an alarm is out of service for TESTING or MAINTENANCE.

C INCORRECT: In accordance with Annunciator System, 0-01-55, a hot pink border indicates that an alarm is NOT ABNORMAL for current plant conditions.

D INCORRECT: In accordance with Annunciator System, 0-01-55, section 8.5, a yellow border is used to signify that an annunciator window is being relocated.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.3.13 (10CFR55.41.12) Tier#

Knowledge of radiological safety procedures pertaining to 3 licensed operator duties, such as response to radiation Group # ------

monitor alarms, containment entry requirements, fuel handling K/A # G2.3.13 responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance Rating 3.4 Proposed Question: # 72 A valve lineup is to be performed on valves with the following conditions:

• Area temperature is 105° F

• Area radiation is 40 mr/hr

• The valves are located 15 off the floor Independent Verification of this valve lineup is expected to take 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Which one of the following choices completes the statement below in accordance with SPP-1 0.3, Verification Program?

Based on the above conditions, Independent Verification of this lineup A. CANNOT be exempted B. may be exempted due to elevation C. may be exempted due to excessive dose D. may be exempted due extreme temperature Proposed Answer: C Explanation A INCORRECT: Plausible in that candidate may believe dose levels are not (Optional): high enough to warrant waiving IV. If the criteria for waiving IV was based on valve located in a High Radiation Area, this would be the correct answer.

B INCORRECT: Plausible in that there are multiple criteria in SPP-1O.3 for waiving Independent Verification. However, valve in a hazardous location due to elevation is not C CORRECT: Activities involving significant radiation exposure can be waived in accordance with SPP 10.3. As a guideline, an exposure greater than 10 mrem TEDE to perform verification would be considered excessive.

This verification would result in dose of 20 mrem.

D INCORRECT: Plausible in that there are multiple criteria in SPP-10.3 for waiving Independent Verification. However, extreme temperature is not one.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of radiological safety procedural requirements pertaining to licensed operator duties. Specifically, when the requirements for Independent Verification may be waived based on excessive dose.

Question Cognitive Level:

This question is rated as C/A due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowl edge and its meaning to predict the correct outcome. Candidate must determine dose to be accum ulated during the verification. Then, compare that to S PP-i 0.3 criteria for waiving IV to determ ine the correct answer.

Technical Reference(s): SPP-10.3 Rev. 2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: NONE Learning Objective: (As available)

Question Source:

Modified Bank # Brunswick 08 # 72 (Note changes or attach parent)

New Question History: Last NRC Exam Brunswick 2008 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Verification Program SPP-10.3 Programs and Rev. 0002 Processes Page 9 of 18 3.4.1 Clearance ActivIties A. /erfication is required for alt clearance (hold order) activities (excep t when verification during clearance release is waived as allowed by Section 3.4.3B

). IV or CV shall be used as specified in Section 3.4.4 or 3.4.5.

B. If authorized by the Cps 1 Manager CV may be used in lieu of IV for clearances which were prepared and reviewed prior to the issue date of Revision 2 of this SPP.

3.4.2 Verification Requirements for Other Activities For activities other than clearance activities, IV or C\f is require d for those systems listed in Appendix A and shall include the following as a minimum:

A. All valves, breakers, and other components in safety-related system s where an inappropriate positioning could adversely affect system/plant operati on or containment integrity.

B. All valves, breakers, and other components in fire protection system major flow paths, including fire fighting water supply and storage, carbon dioxide storage systems, fire protection systems, and all components necessary for the system to function and supply extinguishing media to the fire.

C. All valves, breakers, and other components in gaseous and liquid radioac tive waste handling and processing systems where an inappropriate positioning could result in radioactive material release to the environment.

3.4.3 Activities Exempt From Independent and Concurrent Verific ation Requirements The following items may be exempted from verification requirements.

A. Calculations performed by qualified computer software.

B. Activities for which verifications would he required and one or more of the following conditions exist. These exemptions shall NOT be applied during hold order placement.

1. Out-of-service systems/channels/components for which configuration control will not be maintained and will be verified to be in the proper config uration during the return to operable status.

2. Activities involving significant radiation exposure. As a guideline, an exposure greater than 10 mrem TEDE to perform verification would be consid ered excessive.

3. Activities occurring during emergency conditions (imminent danger to plant or personnel) requiring rapid personnel action.

4. Components located within locked/covered/controlled access areas provided access to the area has not occurred since the last documented verific ation.

For these instances, the decision not to perfonn a verification is to be documented on the procedure/instruction or work document.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BRUNSWICK 2008

74. A valve lineup is to be performed in an area that has the following conditions:

Area temperature 115° F Area radiation 40 mr/hr Independent verification of this valve lineup is expected to take 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Which one of the following choices completes the statement below in accordance with OPS-NGGCl303, Independent Verification?

Independent verification of this lineup, based on the above conditions, may be waived because of A. both extreme temperature and excessive dose BY excessive dose only

0. extreme temperature only

0. either extreme temperature or excessive dose

REFERENCE:

N000.1303 EXPLANATION:

IV may be waived if the dose tH be excessive (as a guideline 10 mrem is excessive) or if personnel safety issues exists (e,g. temperature is above 120° F). IV of this lineup would result in a dose of 20 mrern, CHOICE °A° incorrect. Would be allowed to be waived based on dose only.

CHOICE °B Correct answer, CHOICE °C Incorrect. Would be allowed to be waived based on dose not temperature, CHOICE D° Incorrect. Would be allowed to be waived based on dose only.

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry reqLurernents, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 45.9145.10) 1MPORTANCE RO 3.2 SRO 3.7 SOURCE: Bank LESSON PLAN/OBJECTIVE:

CLSLP2o1 C, Obj. lob. Describe the loilowing regarding OPS-NGGC-1 303 Exemptions from Independent Verification, COG LEVEL: High Page 100 of 147

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.3.5 (10CFR55.41.11/12)

Tier# 3 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, G roup personnel monitoring equipment, etc. K/A # G2.3.5 Importance Rating 2.9 Proposed Question: # 73 Which ONE of the following completes the statement?

The Wide Range Gaseous Effluent Radiation Monitor System (WRGERMS) consists of _(1) ranges, AND has (2).

A. (1)TWO (2) monitors in ALL three Units Control Rooms B. (1)THREE (2) monitors in ALL three Units Control Rooms C. (1)TWO (2) a monitor in Unit 2 Control Room ONLY D. (1)THREE (2) a monitor in Unit 2 Control Room ONLY Proposed Answer: D Explanation A INCORRECT: Part 1 = incorrect, Normal, Intermediate and high ranges are (Optional): supplied. Part 2 = incorrect, The only remote monitoring is from Unit 2.

Plausible in that Units 1 & 3 receive WRGRM alarms. 113-9-3A windows 6 &

13.

B INCORRECT: Part 1 = correct, Normal, Intermediate and high ranges are supplied. Part 2 incorrect, The only remote monitoring is from Unit 2.

Plausible in that Units 1 & 3 receive WRGRM alarms. 1/3-9-3A windows 6 &

13.

C INCORRECT: Part 1 = incorrect, Normal, Intermediate and high ranges are supplied. Part 2 correct, Units 1 & 3 only receive common alarms. 1/3 3A windows 6 & 13. The only remote monitoring is from Unit 2.

D CORRECT: Part 1 = correct, Normal, Intermediate and high ranges are supplied. Part 2 correct, Units 1 & 3 only receive common alarms. 1/3 3A windows 6 & 13. The only remote monitoring is from Unit 2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests the ability to use the Wide Range Gaseous Effluent Radiation Monitor System which is a fixed radiation monitor.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): OPL1 71 .033 Rev 13 (Attach if not previously provided) 2-01-90 Rev 79 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .033 V.B.2 (As available)

Question Source: -

Bank#

Modified Bank # (Note changes or attach parent)

New X Question History:

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Corn me nts:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL1 71033 Revs ion 13 PaQe 44 cf 75

[NSTRUCTOR NOTES (2) The objectives of this stack-gas radiation monitoring system are twofo ki:

(a) lndLcate and record release (ibj. VB:l 3 rates from the stack during )bi. V.U. 1.3 normal operation and alarm Obj.V.D3 when limits are reached (b) Indicate and recoi-d release rates from the stack dLlring accident conditions which could result in gross radiation release Wide Range Gaseous Effluent Radiation Monitor (ORM-9O-3O6)consists of the followinci:

(1) O-RE-90-093 Normal Range noble gas Normal range detector particulate and Iodine filters are abandoned in place 2i 0-RE-90-98A Intermediate Ranqe noble gas detector (3) O-RE-90-98B High range noble gas detector Remote Computer (1) Located in the unit 2 control room panel 2-9-10 (2) Displays stack release rate, statLtS Detailed operation is stack flow rate, and detector virtual contained in 2-01-90 instruments, and annunciators. Illustration 2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 20190 lilustration 2 g The WROERM System changes state and resets at the followinq values:

(1) Normal to Mid Range = 1.00E-02 COl) uCi/sec (2) Mid to High Range = 1 .OOE÷01 (10) uCisec (3) High to Mid Range = 5.OOE÷00 (5)

LICI/sec (4) Mid to Normal Range = I .OOE-03

(.001) uCi/sec BFN Radiation Monitoring System 2-01-90 unit 2 Rev 0079 Page 43 of 70 lflustration 2 (Page 1 of Ii)

Wide Range Gaseous Effluent Radiation Monitor Operation PMIEL LAYOUT REmore 14i.n Screen (Control Room Screen)

i

;1;n iThTii71

- -1EO ift Lfl in zSo.Ot ittsit_:I isct wrin tLtt -

y r- -

Ira

- LIE P4 3

1E3 I -IE-4 iat F I 1:;: I SISISISItJLWNI1II1INMWIJ rE.4o-n 4 GRL9tIwfl4 I RA.9SS 44 if 1aP1MAL f 4b RAisE 4 14t$FtR4HSE i .

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Unit 1 Panel 93 XA55-3A 1 -ARP-9-3A Rev, 0040 Page 3 of52 Annunciator Window Legend FUEL FOOL AIR PARTICULATE RHRSWIRCW FLOORAWEA AX ELDGTURB BLDG. 06 PRETREATMENT STACK GAS MONITOR I EFFLUENT MAIN STEAM LINE RADIATION HIGH RFZONEEXH RAD RADIATION RADIATION HIGH RADIATION HIGH RADIATiON HIGH RAOATICN 1-RA-SO-1A 1-RA-SO-EWA H_________ HIGH-HIGH HIGH l-RA-EG-132 1-RA-9G-255A 1-RA- l-RA-SS-147A I-AA-90-135A fl 2 AREA RADIATION PSC PUMP RX BLDG, TURB BLDG. 06 PRETREATMENT STACK GAS MONITOR SUCTION STNR MAIN STEAM LINE DOWNSCALE RF ZONE EXH RADIATION MONITOR I RADIATION DIFF PRESS HIGH RAD ATION MON:TOR 1-RA-ES-IC RAO MON DNSC DOW HIGH 1-POA-75-74 1-RA-9O-250B JOWNSCALE 1-RA-________ l-RS-9O-47B l-AA-9C-IASB r

RADWASTE BLDG RAD WASTE EFFLUENT RBCCW EFFLUENT TURS BLDG ROOF FEC HEAD TANK STACK GAS AREA RADIATION RADIATiON REACTOR ZONE RADIATION HIGH EXR VENT LE RADIATiON MONITOR HIGH HIGH RADIATION HIGH EXHAUST iRA-BE-iF E-RA-ES-130A 1-RA-AB-ISIA NSC!INOP RADIATION HIGH i-RA-9D-251A I- -,-7E l-RA-9O-147C I-RA-EC-142A H

AX BLDG AREA FADWASTE EFFL RBCCWIACW1RHRSW TURS BLDG ROOF FEC HEAD TANK MAIN STEALI LINE RADIATION RAD:ATiON EFFL RADIATION REFUELING ZONE EXH VENT RADIATION LEVEL LOW RADIATION HIGH MONITOR DOWNSCALE MONITOR ONSCI1NOP EXHAUST RADIATION MONITOR DOWNSCALB HIGH-GH 1-RA-Si-1D O-RA-EO--135C 1-RA-EB-IB1B MONITOR DDOKGOALE I-RA-9O-2E1B i-LA-75-7E i-RA-9O-1SEC l-RA-SE-I4EB TURBINE BLDG AREA RADIATION 05 PRETREATMENT REFUELING ZONE RX ZONE SAMPLE FLOW HIGH EXHAUST EXH RADIATION 5

iR5501 ABNORMAL R.ADIATICN HIGH MCNITOR DNSO 1-FA-SO-ISE i-RA-SS-140A i-RA-Sc-1429 r r

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT BFN Panel 94 3-ARP.9-3A Unit 3 34A.55.3A Rev. 0043 Page 3 of 51 Annunciator Window Legend FUEL POOL AIR PARTICULATE RHRSWIRCW EX BLDG. TURB BLOc-, 00 PRSTP T4T STACK GAS MAIN STEAM LINE FLOOR AREA MCINITOR EFFLUENT RF ZONE EXH RADI RADIATION RADIATION RADIATIONHIGH RADIATIONHIGH RADIATIONHIGH RADIATIONHIGH HL.. HIGH-HIGH HIGH 3-RA-90-IA 3-PA-SO-GSA 3-RA-90-132 3-RA-SII-250A 3-RA-1G-IE?A 3-RA-90-I47A 3-RA-FO-135A Il r rr 7 AREA RADIATION AIR PARTICULATE Pc-C PUMP AX BLDG. TURS BLOC-, 00 PRETREATMENT STACK GAS MAIN STEAM LNE MONITOR MONITOR SUCTION STNR RF ZONE EXN RADIATIGlOjTflR I RADIATION RAORTION ?.IONLTCR DOWNSCALE MALFUNCTION DIFF PRESS HIGH RAD MON DNSC DOW. HIGH DGNXSCALE 3-PA-SO-ic 3-RA-SO-50B 3-PDA-75-74 3-RA-90-2EE ERA- 3-RA-SO-1475 3-RA$c--I358 I C- I I 12 14 PAD WASTE BLDG RAOCASTE EFFLuENT RBCCW EFFLUENT TURB BLDG ROOF PSC HEAD REACTOR ZONE ITACK GAS AREA RADIAIDN RADIATION EXH VENT TA RAD1ATION MONITOR EXHAUST RADIATION HIGI HIGH RADIATION HIGH TuNSCIINOP RADIATION HIGH 3-PA-SO-iF O-RA-90-130A 3-RA-9O-131A 3-RA-90-251A 3-LA-75-78 r 3-RA-SC-147C 3-RA-EO-142A 20 AX BLDG AREA RLDW4STE EFFL RBGCWJRCWIRHRSW PSC HEAD TANK MAIN STEAM LINE TUBE BLDG REF1JC-JNG ZONE RADIATION RAOC-TION EFFL RADIATION AXH VENT RADIATION LEVEL LOW RADIATION EXNGJST RADIATION HIGH UONIIOR 000NSCALE MONITOR DNSc1INOP UONITDR 000NSCALE MONITOR OGANSCALE

-E--9S-I3SC HIGH-HIGH 3-PA-SO-ID 3-RA-SC-i318 3-RE-SO-DDE 3-LA-75-79 3-RA-9O-135C 23 24 25 20 TURBINE BLDG 00 PRETREATMENT REFUELING ZONE AX ZONE RADIATION (SPARE) Ic-PARE) (SPAR:)

SNCRML RADLATONHVTh MON[TORDNON 3-FR-SO-tEE 3-RA-9O-I40A 3-RA-OrDi42B 3-RA-GU-IE r r 13

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.4.42 (10CFR55.41.1O)

Tier# 3 Knowledge of emergency response facilities.

Group #

KJA # G2.4.42 Importance Rating 2.6 Proposed Question: # 74 A plant emergency is in progress that requires a declaration in accordance with EPIP-1, Emergency Plan Implementing Procedure. The plant emergency in progress is NOT a security threat to facility protection.

Which ONE of the following is the LOWEST classification level that requires the Technical Support Center (TSC) AND Operations Support Center (OSC) to be activated?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Proposed Answer: B Explanation A INCORRECT: Plausibility based on misconception that any declaration of (Optional): an event requires activation of OSC and TSC B CORRECT: The TSC and OSC are required to be activated at the Alert or higher emergency classification.

C INCORRECT: Plausible because some actions are first initiated at the Site Area Emergency level (e.g., State headquarters are established at the Morgan County Courthouse and Joint Information Center at Calhoun Community College is staffed.)

D INCORRECT: Plausible because some actions are first initiated at the General Emergency level (e.g., PARs are issued to the state).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the question tests knowledge of what Emergency Action Level Emergency Response Facilities, OSC and TSC, are required to be activated.

Question Cognitive Level:

This question is rated as Fundamental Knowledge.

Technical Reference(s): EPIP-6 Rev. 30 I EPIP-7 Rev. 27 (Attach if not previously provided)

OPL171.075 Rev. 25 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL1 71 .075 V.B.10 (As available)

Question Source: Bank# Quad CWesO9#75 Modified ank# (Note changes or attach parent)

New Question History: Last NRC Exam Quad Cities 2009 (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY ACTIVATION AND OPERATION OF THE OPE RATIONS SUPPORT CENTER E PIP7 I .0 INTRODUCTION 1.1 Purpose The purpose of this procedure is to describe the process for activation of the USC as well as define the activities and responsibilities of USC team members.

2.0 REFERENCES

21 hidustry Documents A. NUREG-Q65L, Criteria for Preparation and Evaluation of Radioogicai Emergency Response Plans and Preparedness in Support of Nuclear Power Plants B. 10 CFR 50.47, Code of Federal Regulations 2.2 Plant Instructions A. TVA Radiological Emergency Plan B. BFN EP1P 1, Emergency CIassification Procedure C. BEN EPIP 2. Notification of Unusual Event D. BFNEPiP-3.Alert E. BEN EPIP 4. Site Area Emergency F. BEN EPIP 5, General Emergency

0. BEN EPIP 15, Terni:nation and Recover, H. BEN Business Practice (BPi 319, Emergency Preparedness Guidetnes 3.0 INStRUCTIONS 3.1 Activation The USC is required to be activated at the Alert or higher emergency classification, however.

activation may occur at the discretion of the Shift Manager. Once an emergency classification has been declared, the Shift Manager (SM) becomes the Site Emergency Director (SED).

Depending upon the emergency classification declared, steps to activate the USC are specfied n the applicable EPIP for that emergency classification. Activation time for the USC is defined in the Radiological Emergency Plan.

3.2 Methods of Notification of Emergency Response Organization (ERU)

Notification of the USC personnel can be accomplished by one or more of the folloing methods:

• Activation of the Emergency Paging System (EPS) is the primary method.

• Manual call-out through utilization of the call-out list.

• Plant Public Address (PA) announcement.

• Activation of the Assembly and Accountability swen.

3.3 ERU Information SPP 1.9, Emergency Preparedness provides the ERU with information regarding duty assignments and response to emergency call-outs.

PAGE 1 OF $1 REVISION 0027

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY ACTIVATION AND OPERATION OF THE TECHNICAL SUPPORT CENTER E P1 P-6

1.0 INTRODUCTION

1.1 Purpose The purpose of this procedure is to describe activation of the Techni cal Support Center (TSC), define the TSC organization and provide for TSC operations by defining staff responsibilities.

2.0 REFERENCES

2.1 Industry Documents A. NUREG-0654, Criteria for Preparation and Evaluation of Radiol ogical Emergency Response Plans and Preparedness in Support of Nuclear Power Plants B. 10 CFR 50.47. Code of Federal Regulations 2.2 Plant Instructions A. TVA Radiological Emergency Plan B. Emergency Plan Implementing Procedure (EPIP) - 1, Emergency Classification Procedure C. EPIP 2. Notification of Unusual Event D. EPIP 3. Alert E. EPIP 4, Site Area Emergency F. EPIP 5, General Emergency G. EPIP 16, Termination and Recovenj H. EPIP-15, Emergency Exposures I. EPIP-i 1, Security and Access Control 3.0 INSTRUCTIONS 3.1 Activotion The TSC is required to he activated at the Alert or higher emerge ncy classification, however, activahon can occur at the discretion of the Shift Manager (SM).

Once an emergency classiflcation has been declared, the SM becomes the Site Emerg ency Director (SED).

Depending upon the emergency classification declared, steps to activat e

specified in the applicable EPIP for the emergency classification. When the TSC am the the on-call SED will obtain a turnover from the SMISED, ensure that minim TSC :s actrated, um staffing is met far the emergency center, and assume the responsibilities of the SED from the SM/SED.

Once the responsibilities of the SM!SED have been assumed by the an-call and control of the emergency response transfers to the TSC. TSC activat SED, command ion time is def,ned in the Radiological Emergency Plan.

3.2 Methods of Notification of Emergency Response Organization (ERD)

Notification of the TSC personnel can be accomplished by one or more of the following methods:

• Activation of the Emergency Paging System (EPS) is the primary method

• Manual call-out through utilization of the call-out list.

• Plant Public Address (PA) announcement

• Activation of the Assembly and Accountability Siren 3.3 ERO Information SPP 1.9, Emergency Preparedness provides the Emergency Respon se Organization (ERO) with information regarding duty assignments and response to emergency call-outs.

PAGE 1 OF 49 REVISION 0030

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .675 Revision 25 Page 16 of 50 INSTRUCTORS NOTES E. Alert, EPIP-3 Refer to EPIP-3 1 - Lowest classification during which emergency centers are required to he manned.

2. This classification assures that emergency personnel are readily available to respond if the situation becomes more serious. EPIP-3 contains the direction for activating the emergency response organization for the Alert.

3. Upon declaration of this class, the following Ohj. V. B.8 actions are perfornied:

a. Notification Requirements based on Required for Emergency Center-s staffed or not staffed. EPIP-3, 4. 5

h. The Operations Duty Specialist (ODS) should he notified by the SM within five minutes of the event classification. The ODS relays the information to the EDO, the State of Alabama. and the CECC Director.

The EDO keeps the CECC Director informed of the situation as necessary.

c. SM!SED completes Appendix A.

d. Site emergency response personnel, including the Plant Manager, are notified by the Unit -l operator using Appendix B.

e. Pax a copy of Attachment A to the ODS

f. A plant PA announcement is macIc.

g. The SM!SED notifies the NRC as soon as possible and within one hour of the event classification.

Note: Any notification may be delegated to other individuals.

h. If the situation warrants accountability, activate the Accountability Alarm in accordance with EPIP-8.

i. The CECC is staffed by the ODS.

j. The TSC and OSC are activated.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPL171 .075 Revision 25 Page 18 of 50 INSTRUCTORS NOTES F. Site Area Emergency, EPIP-4. Refer to EPIP-4 1 . Notification Requirements based on Emergency Centers Staffed or Not Staffed.

2. Upon declaration of this class, the actions Ohj. V.B.8 described in E.3. are performed. In addition:

a. A precautionary Accountability is initiated (if EPIP-4 makes not already performed) and then an this mandatory evacuation of non-emergency responders is initiated in accordance With EPIP8. Accountabilit then Evacuation

h. If appropriate, protective actions for the public are recommended to State agencies by the CECC (not required for SAE).

c. Also of interest at Site Area Emergency:

State headquarters are established at the Morgan County Courthouse and Joint Information Center at Calhoun Community College is staffed.

3. The initiating conditions and emergency action Review Appendix levels which require the Site Area Emergency are C explained in the Technical Basis. EPIP-4 directs a continuous mode of evaluation and reevaluation of changing conditions for the event using EPIP-i.

When those changes are recognized they are to be communicated to offsite agencies.

4. Discuss all sections of EPIP-4 and stress the following: 3.4 & Appendix A.

G. General Emergency, EPIP-5 Refer to EPIP-5

1. Notification Requirements based on Emergency Centers Staffed or Not Staffed.

2. This classification initiates predetermined Conservative protective action for the public, provides decision making continuous assessment of information and initiates .additional measures as required by releases of radioactivity.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DISTRACTOR PLAUSIBILITY SUPPORT OPLI71.075 Revision 25 Page 19 of 50 ENSTRUOTORS NOTES

3. EPIP-5 contains the directions for activating the Review: EPIP-5 emergency response for the General Emergency Attachment C for and the guidance for making protective action PARs recommendations.

4. The Site Emergency Director must make any Ohj. V.6.7 required recommendations until the CECC is staffed. This responsibility cannot be delegated until CECC is in operation. Recommendations are required at General Emergency.

5. If this is the initial classification, the SM notifies the SM has 5 nun ODS within 5 minutes, and the ODS notifies the local governmental agencies within 15 minutes, ODS has IS mm and recommends protective actions, If in a General Emergency and ODS cannot he contacted use phone numbers at bottom of page 2 of EPIP-5 to contact local counties directly and State of Alabama Rad Health Duty Officer.

6. The initiating conditions and emergency action levels which require the General Emergency are explained in the Technical Basis. EPIP-5 directs a Review Appendix continuous mode of evaluation and reevaluation of changing conditions for the event using EPIP.

When those changes are recognized, they are to be communicated to offsite agencies.

7. A plant evacuation of non-emergency responders, must be conducted in accordance with EPIP-8.

8. Discuss all sections of EPIP-5 and stress Ohj. V.6. 9 Protective Action Recommendations (Appendix C).

H. Emergency Organizations EPIP-6 & 7

1. The onsite organization is composed of the Site Ohj. V.610 Emergency Director and technical staff located in the Technical Support Center, the on-shift Operations personnel, and additional support NP REP Plan personnel in the Operations Support Center Appendix A

2. The Technical Support Center (TSC) is staffed EPI P-6 during an ALERT, SITE AREA EMERGENCY, or GENERAL EMERGENCY. TP-1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet QUAD CITIES 2009 EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2009 SRO Written Exam (Quad Cities) 75 ID: QDCJLT.15550 Points: 1,00 A plant emergency is in progress that requires a declaration in accordance with the Exelon Nuclear Emergency Plan (E-Plan).

The plant emergency in progress is NOT a security threat to facility protection.

Which one of the following states the lowest classification level that REQUIRES the Technica Support Center (TSC) and Operations Support Center (OSC) to be ACTIVATED?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: B Answer Explanation:

Answer: The TSC and DSC must be activated at an ALERT classification or higher whe.n NOT a security event.

Distractor 1 is incorrect: Plausible if the candidate assumes TSC is always activated at an Unusual Event (events other than a security event).

Distractor 2 is incorrect: P ausible because some actions are first initiated at the Site Area 1

Emergency level (e.g., Assembly/Accountability).

Distractor 3 is incorrect: Plausible because some actions are first initiated at the General Emergency level (e.g., PARs are issued to the state).

Reference:

G-1 / EP Overview Rev 7 Reference provided during examination: N/A Cognitive level: Memory Level (ROISRO): RO Tier: 3 Group: N/A Question Source: Catawba ILl Bank # 581 Question History: 2009 Catawba ILT NRC Exam 10 CFR Part 55 Content 41,10 Comments: Changed ansvier location (response to NRC comment).

Associated objective(s):

NGET Objective link (Refer to Non-Acredited Project for NGET/RWT objectives) 2.4.42 Knowledge of emergency response facilities. (R02.6 / SRO3.8j

DPS MLSIER STAXCALtDNE Pace: l7 c 198 £4 Noerraer 2tu

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO G2.4.47 Tier # 3 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Group # N/A K/A # G2.4.47 Importance Rating 4.2 Proposed Question: # 75 ALL High Pressure Injection has been lost on Unit 2.

• At 16:00:00, Reactor Water Level is (-) 110 inches

• At 16:02:00, Reactor Water Level is (-) 118 inches If level continues to lower at the same rate, which ONE of the following completes the statement?

A Common Accident Signal will be initiated by (1) Range level instruments AND the EARLIEST time that ALL Core Spray Pumps will have auto started is (2)

A. (1) Emergency (2) 16:03:07 B. (1) Post Accident (2)16:03:07 C. (1) Emergency (2) 16:03:21 D. (1) Post Accident (2) 16:03:21 Proposed Answer: C Explanation A INCORRECT: Part 1 correct See Explanation C. Part 2 incorrect (Optional):

See Explanation B.

B INCORRECT: (1) Incorrect, this instrument indicates (-)268 to (+)58 inches and initiates the Containment Spray Interlock. Candidate may select because instrument indication is within the desired range of Level 1. (2)

Time is incorrect. Plausible in that this would be the correct answer for DIG Voltage Available (DGVA) sequence. Since there is no loss of offsite power, a Normal Voltage Available (NVA) sequence will occur.

C CORRECT: 1) Correct instrument. Emergency Range is (-)155 to (+)60 inches. Initiates HPC), RCIC, RHR, CS and ADS. (2) Time is correct, level trend is 4 inches/mm. Three minutes to Level 1, and with Normal Voltage Available (NVA), the last Core Spray Pump will sequence on 21 seconds after the accident signal is received.

D INCORRECT: Part 1 incorrect See Explanation B. Part 2 correct See Explanation C.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet KA Justification:

The KA is met because the candidate must diagnose and determine trend and know correct control room instrument (range and function).

Question Cognitive Level:

This is higher cognitive because the examinee must know at what level Core Spray auto starts, calculate the time to the level, know the Core Spray sequence times based on the given plant conditions, and calculate the total time. The examinee must also know which type of instrumentation initiates the signal. He/she must use a multi-part mental process of assemb ling, sorting, or integrating parts of multiple systems to predict the outcome.

Technical Reference(s): OPL171 .038 Rev 17 (Attach if not previously provided)

OPL171.003 Rev 19 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171.038 V.B.9, V.B.11 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam -

(Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of evety question.)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171 .038 Revision 17 Page 41 of68 INSTRUCTOR NOTES (12) The redundant start may be canceled before the start circuit locks out by opening the logic breaker and pushing both engine stop push-buttons. Note that pulling the engine driven fuel pump shutoff plunger will not stop the diesel since the electric fuel pump \vill still be supplying fuel. (01-82)

3. Accident Operation

a. Accident signal received (CASx)

(1) Signals diesel generators to start.

(2) Opens diesel output breakers if shut. ObJ.V.8.9

b. If normal voltage is available, load will ObJ.V.C.6 sequence on as follows: (NVA)

Time After Accident SID Board S/D Board S!D Board SID Board A C B D 0 RHR/CS A 7 RHR/CS B 14 RHR/CS C 21 RHRJCS D 28 RHRSW* RHRSW* RHRSW* RHRSW*

• RHRSW pumps assigned for EECW autom atic start

c. If normal voltage is NOT available: (DGVA) ObJ.V.8.9 Obj.V.C.6 (1) After 5-second time delay, all 4kV Shutdo\.vn Board loads except 4160f480V transformer breakers are automatically tripped.

(2) Diesel generator output breaker closes when diesel is at speed.

(3) Loads sequence as indicated below Time After Accident SID Board SID Board SID Board SJD Board A B C D 0 RHR A RHR C RHR B RHR D 7 CSA CSC 086 CSD 14 RHRSW* RHRSW* RHRSW* RHRSW

• RHRSW pumps assigned for EECW automa tic start

d. Certain 480V loads are shed whenever an accident signal is received in conjunction with the diesel generator tied to the board. (see OPL171 .072)

b. Capable of fast starfing and being ready to load within ID seconds.

ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet fUtr kCC!UriIT kIb i,t, I:ti1 V:1L St.

I. Fc :c t4dT MLII FJSCT1D44

[ NIT N_h Lfl CDCL 5C. TCFL9PTFTr-. r cri L]2.LIL ftsIRActFcT2 Sc a +LI !M3 LQ FPCIFSCJ-Ti.CSSC Wt F L3 £2552 CCl1rIIlLrc-Ii1C- L2tY I I A fl c:i ir ij LIII%tfl c4I I IYI F ST I ) IT RIB CIT _OCr NIl. N& F!YIGI4 LVCL OaTRtc. FT 2Y5t S j lICk H ;ftI4./sL I- I F 114FI.?W LI .TtTi - :c Nt Pc 4-k Lj _c.+I cc-p PCI TI TIIT2I 12 lC4CITrCM TP-6 Reactor Vessel Level/Pressure Instrumentation