Text

Final RO & SRO Written
	ML021350097
Person / Time
Site:	Hope Creek
Issue date:	03/18/2002
From:	Diane Jackson; Public Service Enterprise Group
To:	Conte R; NRC/RGN-I/DRS/OSB
References
	05000354/2002-301 05000354/2002-301
	Download: ML021350097 (151)
	v • d • e

"ReactorOperator Answer Key f FA&-L.,A

1. d 26. c

2. b 27. a

3. a 28. b

4. e 29. d

5. b 30. d

6. b 31 a

7. a 32. a

8. b 33. c

9. d 34.A e~~

10. a 35. c

11. b *.

12. a 37. d

13. d 38. d

14. d 39. b

15. c 40. c

16. b 41. a

17. a 42. c

18. a 43. d

19. b 44. d

20. a 45. b

21. J56 C.L 46. b

22. d 47. c

23. a 48. d

24. c 49. a

25. b 50. b Page 1

Reactor Operator Answer Key 51 . b 76. d

52. c 77. b

53. d 78. d

54. a 79. d

55. 80. a

56. b 81 . c

57. b 82. c

58. d 83. d

59. A ,? ""-,,- 84. a

60. a 85. c

61. d 86. a

62. a 87

63. c 88. d

64. a 89. c

65. d 90. C

66. a 91 . C

92. ,a.-" * *r4.

67. c

68. d 93. c

69. b 94.*

70. c 95. b

71. b 96. b

72. c 97. b

73. a 98. c *a CL

74. c 99. b

75. a 100. P--

Page 2

Reactor Operator Answer Key - '

1. d 26. c

2. b 27. a

3. a 28. b

4. P Tc 29. d

5. b 30. d

6. b 31 . a

7. a 32. a

8. b 33. c 34*A* * (.4,I-.e,

9. d

10. a 35. c

11. b o

12. a 37. d

13. d 38. d

14. d 39. b

15. c 40.

16. b 41. a

17. a 42. c

18. a 43. d

19. b 44. d

20. a 45. b

21. J6"' CL 46. b

22. d 47. c

23. a 48. d

24. c 49. a

25. b 50. b Page 1

Reactor Operator Answer Key 51 . b 76. d

52. c 77. b

53. d 78. d

54. a 79. d

55. a 80. a

56. b 81 . c

57. b 82. c

58. d 83. d

59. ,e 72JeT-- 84. a

60. a 85. c 61 . d 86. a

62. a 87

63. c 88. d

64. a 89. c

65. d 90. C

66. a 91 . C c 92 . ,,a- , ' ,..

67.

68. d 93. c

69. b 94. a,-* ' t[:*-

70. c 95. b

71. b 96. b

72. c 97. b

73. a 98. C 00- q,

74. c 99. b

75. a 100. A- DA4*1,,*#

Page 2

SLIo/ Senior Reactor Operator Answer Key 7

26. d

1. d 27. C

2. b 28. ,,y'Xck

3. a 29. d

4. a 30. a

5. b 31. c

6. b 32. b

7. b 33. a

8. a 34. b

9. b 35. d

10. d 36. d

11. a 37. d

12. b 38. a

13. a 39. b

14. a 40. a

15. c 41. C

16. d 42. C

17. C

18. b 44. d

19. a 45. C

20. C 46. a

21. C 47. d

22. a 48. C

23. b 49. d

24. d 50. b

25. b Page 1

Key Senior Reactor Operator Answer

76. d

51. c 77. a

52. d 78. c

53. a 79. b

54. b 80. c

55. c 81 . d

56. b 82. a

57. b 83 ).

58. d 84. d

59. a 85. d

60. d,_ Ve['1

86. c 61 - a 87. c

62. d 88. a

63. a 89. C

64. c 90. d

65. c 91 . b

66. a 92. a

67. c 93. 6

68. b 94. a

69. c 95. d

70. a 96. b

71. b 97. co*- ok

72. c 98. b

73. b 99. b

74. d 00. 2

75. b Page 2

- 0 Senior Reactor Operator Answer Key

26. d

1. d 27. C

2. b 28. ,,

3. a 29. d

4. a 30. a

5. b 31 . c

6. b 32. b

7. b 33. a

8. a 34. b

9. b 35. d

10. d 36. d

11. a 37. d

12. b , N

38. a

13. a 39. b

14. a 40. a

15. C 41 . c

16. d 42. C

17. C

18. b 44. d

19. a 45. c

20. C 46. a

21. c 47. d

22. a 48. C

23. b 49. d

24. d 50. b

25. b Page 1

Key Senior Reactor Operator Answer

76. d 51 . C 77. a

52. d 78. c

53. a 79. b

54. b 80. c

55. C 81 . d

56. b 82. a

57. b 83 .,

58. d 84. d

59. a 85. d

60. d jV-(c 86. c 61 . a 87 c

62. d 88. a

63. a 89. c

64. c 9o. d

65. C 91 . b

66. a 92. a

67. c 93. V -*'*

68. b 94. a

69. c 95. d

70. a 96. b

71. b 97. co'- o*

72. c 98. b

73. b 99. b

74. d 00.

75. b Page 2

Question Cross Reference Record Exam KA Number Level ")R0 SRO _

295001 AA1.02 1 1 1 295002 AK1.04 2 2 2 295003 AA1.03 3 3 3 295003 2.4.9 4 4 295004 AA2.01 5 4 295004 AK3.03 6 5 5 295005 AA2.04 7 6 295006 2.1.28 8 6 7 295006 AKI.01 9 7 8 295007 AK2.05 10 8 9 295007 AK3.04 11 9 10 295008 AA1.01 12 10 11 295008 AK3.04 13 11 12 295009 AA2.01 14 12 13 295009 2.4.6 15 14 295010 AA1.02 16 15 295010 AA1.02 17 13 295012 AK1.01 18 14 16 295013 AK2.01 19 15 17 295014 AK2.04 20 16 18 295014 AK3.01 21 17 19 295015 2.3.4 22 20 295015 AK3.01 23 21 295016 AA1.02 24 18 22 295017 AA2.01 25 23 295018 AA2.03 26 24 295019 AA1.02 27 19 25 295019 AA2.01 28 20 295021 2.4.41 29 26 295022 2.4.48 30 27 295022 AK2.03 31 21 28 295023 AA1.02 32 22 29 295023 2.4.11 33 23 30 295024 EA1.10 34 24 31 Saturday, March 23, 2002 I 0f 4

Record Exam KA Number Level RO SRO 295024 2.1.6 35 B 25 32 295025 EA2.06 36 R 26 295025 EKI.05 37 B 27 33 295026 EK1.02 38 B 28 34 295028 EK1.02 39 B 29 35 295030 EA2.04 40 S 36 295030 EK2.03 41 B 30 37 295031 EK2.13 42 B 31 38 295034 2.4.30 43 S 39 295036 EK2.01 44 B 32 40 295036 EK3.01 45 B 33 41 295038 EA2.03 46 R 34 295038 EK1.02 47 B 35 42 500000 EK3.03 48 B 36 43 600000 2.4.25 49 R 37 201001 A3.05 50 B 38 44 201002 2.4.21 51 R 39 201002 K4.08 52 B 40 45 201003 K4.05 53 B 41 46 201006 2.1.12 54 S 47 201006 K3.01 55 B 42 48 202001 A4.04 56 R 43 202001 K3.07 57 B 44 49 202002 K6.04 58 R 45 203000 A4.07 59 B 46 50 203000 K1.14 60 B 47 51 204000 A2.14 61 B 48 52 206000 A1.06 62 B 49 53 206000 A3.07 63 R 50 209001 K1.10 64 B 51 54 209001 K2.02 65 B 52 55 211000 2.4.10 66 R 53 211000 K1.05 67 R 54 212000 2.1.23 68 R 55 212000 K5.02 69 B 56 56 Saturday,March 23, 2002 2 of 4

Record Exam KA Number Level RO SRO 215001 K1.05 70 B 57 57 215004 A3.03 71 B 58 58 215004 2.2.6 72 S 59 215005 K3.05 73 B 59 60 215005 K5.05 74 B 60 61 216000 A2.08 75 B 61 62 216000 K2.01 76 B 62 63 217000 A2.01 77 S 64 217000 K4.05 78 B 63 65 219000 A3.01 79 R 64 223001 K6.13 80 R 65 223002 K3.16 81 B 66 66 223002 K4.01 82 B 67 67 226001 A1.06 83 R 68 226001 A3.05 84 B 69 68 230000 K6.01 85 B 70 69 233000 2.1.7 86 S 70 234000 2.2.25 87 S 71 239002 A1.02 88 R 71 239002 A1.05 89 B 72 72 245000 K5.02 90 R 73 256000 A2.13 91 R 74 256000 K4.06 92 R 75 259002 2.4.32 93 S 73 261000 A4.07 94 B 76 74 261000 K6.03 95 B 77 75 262001 K2.01 96 R 78 262001 K6.01 97 B 79 76 263000 A1.01 98 B 80 77 263000 K2.01 99 B 81 78 264000 2.1.11 100 S 79 268000 A1.01 101 B 82 80 271000 K1.02 102 B 83 81 272000 K6.03 103 B 84 82 290002 A2.02 104 R 85 Saturday, March 23, 2002 3o0f4

Record Exam KA Number Level RO SRO 290002 K3.03 105 R 86 290003 K5.01 106 B 87 83 GENERIC 2.1.3 107 B 88 84 GENERIC 2.1.14 108 S 85 GENERIC 2.1.24 109 B 89 86 GENERIC 2.1.33 110 S 87 GENERIC 2.1.33 111 R 90 GENERIC 2.1.34 112 S 88 GENERIC 2.2.22 113 R 91 GENERIC 2.2.22 114 S 89 GENERIC 2.2.26 115 S 90 GENERIC 2.2.27 116 R 92 GENERIC 2.2.27 117 S 91 GENERIC 2.2.30 118 R 93 GENERIC 2.2.31 119 S 92 GENERIC 2.3.1 120 B 94 93 GENERIC 2.3.2 121 R 95 GENERIC 2.3.4 122 S 94 GENERIC 2.3.9 123 R 96 GENERIC 2.3.10 124 S 95 GENERIC 2.3.11 125 B 97 96 GENERIC 2.4.5 126 B 98 97 GENERIC 2.4.18 127 B 99 98 GENERIC 2.4.28 128 S 99 GENERIC 2.4.34 129 B 100 100 Saturday, March 23, 2002 4 of 4

Given the following:

- The plant is operating at 100% power

- A transient caused by a short in the reactor recirculation control circuitry occurs Immediately following the transient, the plant stabilizes with the following parameters:

- Reactor Power 50%

- "A" Recirc pump tripped

- "B" Recirc pump at 45% speed

- Loop "A" total jet pump flow is 10 Mlbmlhr

- Loop "B" total jet pump flow is 46 Mlbmlhr

- Total indicated core flow 36 Mlbm/hr What is actual core flow, and how will the loss of the "A" Recirc pump affect the APRM Scram setpoint?

[I 36 Mlbm/hr. Setpoint unaffected

[I 36 Mlbm/hr. Setpoint needs to be adjusted

[ 56 Mlbm/hr. Setpoint unaffected W 56 Mlbm/hr. Setpoint needs to be adjusted Anserd Level B Bxam CognitiveLevel Comprehension Facility Hope Creek 03/12/20022 0xamDate:

Tier*: Emergency and Abnormal Plant Evolutions RoGroup 2 SROGroup 2 295001A102 295001 Partial or Complete Loss of Forced Core Flow Circulation Record Number 1 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

AA1.02 RPS 3.3 3.3 pnation Of Below 48% running recirc loop speed, Jet pump loop flows are both positive and added together.

Answer Setpoints must be adjusted to single loop values within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Reference Title I HC.OP-AB.ZZ-0300 HC.OP-DL.ZZ-0026 Attach 3V TS 2.2.1 and 3.4.1 Learning Objectives OAB300ED03 (R) Discuss the operational implications of the abnormal indications/alarms for system operating parameters related to Reactor Power Oscillations, Abnormal Operating Procedure.

rMaterial Required for Examination IQuestionSuouýE] INPO Exam Bank [5-u lon Modific at Meon Significantly Modified Q t Srce Comments: INPO EXAM BANK QID# 17049 Susquehanna 1 09/30/1999 Saturday, March 23, 2002 10:47:27 AM Pagel1of 139

The plant has been operating at full power for several days.

- Operators notice that, over the last several hours, Main Condenser Vacuum has risen from 3.2"HgA to 4.0"HgA.

- Over this same period, Offgas system flow has increased from 25 scfm to 38 scfm.

- There have been NO ALARMS associated with this problem.

Which one of the following would cause these indications?

[i Cooling tower outlet temperature increase ti Reactor Feed Pump Turbine exhaust piping leak W Tube leak in #2A Feedwater Heater

] Resin intrusion from the Condensate Demineralizers b B Cognitiv Comprehension Hope Creek 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGoup 2 S roup 2 295002K104 295002 Loss of Main Condenser Vacuum Record-Number 2 AK1. Knowledge of the operational implications of the following concepts as they apply to LOSS OF MAIN CONDENSER VACUUM:

AK1 .04 Increased offgas flow 3.0 3.3 Explanationof Increase in air inleakage via the RFPT exhaust piping under vacuum into the main condenser will cause Answer Offgas outlet flows to increase. Cooling tower outlet temp increase would degrade vacuum but not change outlet flow. 2A Heaters are internal to the main condenser so no change in outlet flow. Resin intrusion causes offgas radiation levels to increase Reference Title HC.OP-AB.ZZ-0001 Learning Objectives 0AB208E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Main Condenser Low Vacuum, Abnormal Operating Procedure.

[Material Required for Examination

[Question Soure] INPO Exam Bank aQuestion Modification Method: Significantly Modified Question Source Comments: j INPO BANK QID# 647 Duane Arnold 05/25/1999 Saturday, March 23, 2002 10:47:27 AM Page 2 of 139

Given the following:

- The plant is operating at 100 percent power

- A severe electrical transient results in a loss of all offsite power

- 2 control rods are at position "48"

- Reactor power is less than 1 percent Which one of the following describes the equipment available to control reactor pressure and level?

and SRVs SHPCI and Main Steam Line Drains SHPCI WReactor Feed Pumps and SRVs and Main Steam Line Drains SRCIC a xmLevl B [cognitive Level Comprehension FciyHope Creek E-xa-m -Dat-e:. 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGrou 2 SRO Group 295003A103 295003 Partial or Complete Loss of A.C. Power Record Number 3 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER:

AA1.03 Systems necessary to assure safe plant shutdown 4.4 4.4 Explanation oi Although the reactor is not shutdown, power is less than 4%. The candidate must determine HPCI Answer I injection is allowed under EOP-101A PREFERRED ATWS INJECTION SYSTEMS TABLE 1 using EOP 322 as necessary. Loss of offsite power causes Group 1 isIn. MSL Drains will close if the valves have power. RFPT oil pumps can be restored from EDG backed busses, but condensate pumps are tripped.

Ii Reference Title HC.OP-AB.ZZ-0135 EOP-101A Learning Objectives OAB135E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Station Blackout/Loss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination J EOP Flowcharts without entry conditions t Source: New Qu~es-tion Modification Mehd Question Source Comments:

Saturday, March 23, 2002 10:47:28 AM Page 3 of 139

Given the following:

- The reactor is in Operational Condition 4

- "A" RHR Pump is in Shutdown Cooling at rated flow

- 10A404 4.16KV 1E Bus trips on bus differential overcurrent Which one of the following describes the effect the bus loss will have on Shutdown Cooling?

Ei The Shutdown Cooling common suction line isolates and CANNOT be reset El The AP228 Jockey pump trips causing Shutdown Cooling Loop "A" to lose keepfill W Both "A" and "B" Shutdown Cooling Loops lose ability to adjust flow F "B" Reactor Recirc Pump discharge valve automatically opens bypassing core flow n ] c Exam Level R Cognitive Level Comprehension Hope Creek 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 2 1 295003G409 295003 Partial or Complete Loss of A.C. Power Record Number 4 2.4 Emergency Procedures and Plan 2.4.9 Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation 3.3 3.9 strategies.

Explanation of A" RHR SDC Return valve F015A is powered from "D"Channel 1E 480VAC. Loss of D Bus fails this Answer . valve as is. Adjusting flow via RHR HX outlet valve and /or bypass valve is not proceduralized. AP228 provides keepfill to HPCI only. B RRP disch valve is controlled by NON 1E power.

Reference Title HC.OP-SO.BC-0002 HC.OP-SO.SM-0001 Learning Objectives 000028E008 (R) Given a system which physically connects to or is required to support the operation of the RHR System or components therein, explain the function of the supporting system, lAW the RHR System Lesson Plan.

Material Required for Examination Question Source: New lQuestion Modification Method:

lQuestion Source Comments:

Saturday, March 23, 2002 10:47:28 AM Page 4 of 139

Which one of the following conditions will automatically remove the 125 VDC battery charger from service per HC.OP-AB.ZZ-0150, 125VDC System Malfunction?

IK High output voltage

[i Equalize timer reaches zero PI Blown fuse in the battery transfer switch LI] Low battery terminal voltage Anwra S LCognitive Level Memory F Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 2 2 295004A201 295004 Partial or Complete Loss of D.C. Power Record Number 5 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

AA2.01 Cause of partial or complete loss of D.C. power 3.2 3.6 Explanation of SRO UNIQUE - RO LEVEL QUESTION Answer The following battery charger malfunctions will shutdown the battery charger:

High Voltage Shutdown Relay AC Input Breaker Open/Tripped DC Output Breaker Open/Tripped Loss of 120 VAC Supply Power CORRECT - High Voltage Shutdown Relay.

INCORRECT - Low battery terminal voltage. This will generate a Battery Monitor Alarm not a charger trip.

INCORRECT - High Voltage Shutdown Relay. This will generate a Battery Monitor Alarm not a charger trip.

INCORRECT - Blown fuse in the battery transfer switch. This will generate a Battery Monitor Alarm, not a charger trip.

Reference Title HC.OP-AB.ZZ-0150 Learning Objectives OAB150EO06 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of 125 VDC System Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination luestion Source: New Question Modification Methd:

IQuestion Source Comments:I Concept used from Vision Bank QID# Q61703 for 24 VDC chargers Saturday, March 23, 2002 10:47:28 AM Page 5 of 139

Given the following:

- The Reactor is in Operational Condition 4

- Plant startup operations are in progress

- The negative battery charger for the "A" +/-24 VDC System is out of service

- The positive battery charger for the "B" +/-24 VDC System is on an equalizing charge

- All other equipment is aligned for normal operation Which one of the following will occur if these conditions remain for a prolonged period of time?

An RPS trip will occur due to:

fi A and C SRMs fail upscale because of low voltage to the drawers C, E, and SA,G IRMs fail upscale because of low voltage to the drawers W B and D LPRMs fail upscale because of high voltage to the detectors WB, D, and F APRMs fail upscale because of high voltage to the detectors b Ex Leve B cognitiveLee Memory Facility Hope Creek Exam Date. 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 2 2 295004K303 295004 Partial or Complete Loss of D.C. Power Record Number 6 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

AK3.03 Reactor SCRAM: Plant-Specific 3.1 3.5 xplanation Of JUSTIFICATION:

Answer The negative charger only charges the negative battery while the positive charger only charges the positive battery. Even with the positive charger operating in the Equalizer mode, the negative battery will be discharged resulting in the loss of the DC bus.

CORRECT - IRMs upscale (1/2 scram). The loss of the -24VDC from the A +/- 24VDC System will cause IRM indications to rise (upscale). This will insert a 1/2 scram from RPS Channel A.

INCORRECT - SRMs upscale (Full Scram). SRM indications to lower (downscale)

INCORRECT - LPRMs upscale (Full Scram). LPRMs and APRMs are unaffected by the loss of -24VDC.

INCORRECT - APRMs upscale (1/2 scram). LPRMs and APRMs are unaffected by the loss of -24VDC.

Reference Title HC.OP-AB.ZZ-0151, Sections 2.1, 4.5 & 5.1 H.C. Incident Report 86-067, CD-1 826, PTS-1 826 Learning Objectives OAB151E003 (R) Discuss the operational implications of the abnormal indications/alarms for system operating parameters related to 24 VDC Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination I SourceE tbon Facility Exam Bank Method: I Editorially Modified MQestionModification IQuestion Source Comments: Vision Bank QID# Q61702 Saturday, March 23, 2002 10:47:28 AM Page 6 of 139

Given the following:

- The plant is operating at 29 percent power

- Overhead Annunciator C5-C2 TCV FAST CLOSURE & MSV TRIP BYP is ILLUMINATED Then the Main Turbine Generator trips

- All Turbine Bypass valves responded full open

- Overhead Annunciator B3-E5 RPV PRESSURE HI is ILLUMINATED

- Overhead Annunciators C5-A2 & B2 for TCV FAST CLOSURE and MAIN STOP VALVE CLOSURE are ILLUMINATED

- Overhead Annunciators C3-A2, A3, A4, & A5 for REACTOR SCRAM TRIP LOGIC Al, A2, BI,&

B2 are EXTINGUISHED Which one of the following actions is required?

iI Lock the Reactor Mode Switch in Shutdown immediately reactor SReduce pressure below the alarm point within 15 minutes HI Reduce reactor power by at least 4 percent within 30 minutes P Commence a normal shutdown within one hour Anwrb 5 [ExamgLevel conitive Comprehension Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RO~roup 1 2 295005A204 295005 Main Turbine Generator Trip Record Number 7 AA2. Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP:

AA2.04 Reactor pressure 3.7 3.8 Eixaation of RPV Pressure is above the Tech Spec LCO of 1020 psig but below the scram setpoint of 1037 psig. The Anwe required action is to lower RPV pressure below the TS 3.4.6.2 LCO value which coinsides with the alarm point (1020 psig) within 15 minutes.

Lock the MS in SD immediately - incorrect because RPS alarms are extinguished and all systems functioned properly. An ATWS does not exist.

Reduce power - incorrect because although reducing power will reduce RPV pressure, the time requirement of 15 minutes would not be met. Immediate operator action for AB-202 High RPV pressure is to reduce REACTOR POWER as necessary to clear the RPV PRESSURE HIGH alarm.

Commence a normal SD within one hour - incorrect. The TS action time is 15 minutes, not one hour.

Reference Title HC.OP-AB.ZZ-0202 Tech Spec 3.4.6.2 10CFR55.43(5)

Learning Objectives 0AB138E004 Explain the reasons for how plant/system parameters respond when implementing, Turbine Generator Trip/Malfunction, Abnormal Operating Procedure 000106E001 Given the following lists, summarize and explain both the initial response (goes up, down, stays the same) and the long term response of the parameters in List A to the plant transients in List B lAW the Student Handout.

List A Reactor Power (APRM)

Reactor Power (Surface Heat Flux)

Reactor Pressure (Dome)

Saturday, March 23, 2002 10:47:28 AM Page 7 of 139

Reactor Indicated Water Level Reactor Indicated Steam Flow Reactor Actual Steam Flow Reactor Feedwater Flow Reactor Core Flow Reactor Recirculation Loop Flow SRV Flow List B Loss of Feedwater Heating Feedwater Controller Failing to Maximum Demand EHC Pressure Sensor Failing High Generator Load Rejection with Bypass Valves Available Generator Load Rejection without Bypass Valves Available Turbine Trip with Bypass Valves Available Turbine Trip without Bypass Valves Available MSIV Closure Loss of Condenser Vacuum Loss of All Grid Connections Loss of Feedwater Flow Trip of One Recirculation Pump Trip of Both Recirculation Pumps Recirculation Flow Control Failure - Decreasing Flow Seizure of One Recirculation Pump Recirculation Flow Control Failure - Increasing flow IMaterial Required for Examination Tech Specs without Definitions, Safety Limits, and bases

[uesion ource: New lQuestion Modification Method:

Question Source Comments:

Saturday, March 23, 2002 10:47:29 AM Page 8 of 139

Which one of the following is the reason that the reactor operator must wait at least 10 seconds following a reactor scram before attempting a scram reset?

E] To allow reactor water level to recover above the scram setpoint FI To allow all the control rods to insert fully F] To allow the Scram Air header to repressurize PW To allow the Scram Discharge Volume vent and drain valves to cycle b Exam Level B Cognitive Level Memory Facilty Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROeroUp 1 SRO Group 1 295006G128 295006 SCRAM Record Number 8 2.1 Conduct of Operations 2.1.28 Knowledge of the purpose and function of major system components and controls. 3.2 3.3 Explanation of 10 Second time delay is to allow all control rods time to insert full in.

Answer I Reference Title Lesson Plan 0301-000.OOH-000022-19 Learning Objectives 000022E007 From memory, state the purpose of the time delay after a scram, lAW the Lesson Plan.

[Material Required for Examination I lQuestion INPO Exam Bank [Question Modification Methjod Editorially Modified IQuestion Source Comments:- INPO EXAM BANK QID# 7112 Duane Arnold 1 06/03/1996 Saturday, March 23, 2002 10:47:29 AM Page 9 of 139

Hope Creek requires an Emergency Depressurization after performing steam cooling in EOP-101 "Reactor Control". All actions required by EOP-202, "Emergency Depressurization", have been taken but only 4 Safety Relief Valves (SRV) can be opened and no other means of depressurization is available.

Which one of the following describes the consequences of this failure?

[i Steam removal rate from the core is NOT adequate to ensure adequate decay heat removal exists.

[i Steam removal rate during a LOCA is NOT adequate to prevent exceeding the Drywell design pressure.

EI The pressure reduction rate will NOT allow low pressure injection systems to inject soon enough to recover level before the core becomes uncovered.

- The pressure reduction rate will NOT allow low pressure injection systems to inject prior to reaching the Minimum Steam Cooling RPV Water Level.

a Exam Level B cognitiv Memory Hope Creek 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 1 1 295006K101 295006 SCRAM Record Number 9 AK1. Knowledge of the operational implications of the following concepts as they apply to SCRAM:

AK1.01 Decay heat generation and removal. 3.7 3.9 Explanation of Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) at Hope Creek is 5 Answer " SRVs is sufficient to remove all decay heat from the core.

Reference Title HC-EOP 202 Bases HC.OP-EO.ZZ-LIMITS-CONV Learning Objectives 000130E003 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by that step.

[Material Required for Examination EOP Flowcharts without entry conditions

[Question Source: INPO Exam Bank IQuestion Modification Method: I Editorially Modified IQuestion Source Comments: I QID# 14157 Peach Bottom 2 03/26/2001 Saturday, March 23, 2002 10:47:29 AM Page 10 of 139

Given the following:

- The plant is in Operational Condition 3

- Main Condenser vacuum is broken

- RHR Loop "B" is in Shutdown Cooling

- Reactor level is stable at +35 inches

- Reactor pressure is 50 psig and lowering

- "D" SSW Pump has just tripped

- "B" SSW Pump will NOT start Which one of the following describes the effect this will have on the plant?

(Assume no operator action)

El The RHR Shutdown Cooling Loop will isolate due to lowering reactor level El The RHR Shutdown Cooling Loop will isolate due to increasing reactor pressure El "B" RHR Pump Min-Flow valve will open due to lowering loop flow

" "B" RHR Pump Min-Flow valve will open due to reaching pump shutoff head Anwrb Ex Level B LCognitive LeveComprehension Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RO~roup 1 1 295007K205 295007 High Reactor Pressure Record Number 10 AK2. Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

AK2.05 Shutdown cooling: Plant-Specific 2.9 3.1 Explanation of Loss of cooling media to RHR HX will cause reactor pressure to increase until 82 psig setpoint for Answer NSSSS SDC isolation Reference TitleI HC.OP-SO.SM-0001 Learning Objectives 000028E008 (R) Given a system which physically connects to or is required to support the operation of the RHR System or components therein, explain the function of the supporting system, lAW the RHR System Lesson Plan.

IMaterial Required for Examination I FQuestion Source: New Question Modification Mehd:

Question Source Comments:

Saturday, March 23, 2002 10:47:29 AM Page 11 of 139

Following a reactor scram and Main Steam Isolation Valve closure, reactor steam dome pressure reaches 1050 psig causing the "H" and "P" Safety Relief Valves (SRV) to open.

Which one of the following lists the operating setpoints for subsequent openings of the "P' SRV?

[1 SRV "P" opens at 1017 psig and closes at 905 psig S SRV TP" opens at 1017 psig and closes at 935 psig Li SRV "P" opens at 1047 psig and closes at 905 psig

- SRV TP" opens at 1047 psig and closes at 935 psig Answer d Exam Level B cognitive Level Memory Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 1 SROeGoup 1 295007K304 295007 High Reactor Pressure Record Number 11 AK3. Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

AK3.04 Safety/relief valve operation: Plant-Specific 4.0 4.1

[Epaations o SRV "P' opens at 1047 psig and closes at 935 psig F _ Reference Title HC.OP-SO.SN-0001 Precautions 3.2.12 Learning Objectives 000046E003 (R) Concerning the safety relief valves; summarize, list or identify the following lAW the lesson plan.

a. The number and type of SRV's at Hope Creek.

b. Which SRV's have an ADS function.

c. Power supplies to the SRV solenoids.

d. Which SRV's can be operated remotely and the location from which each of these valves can be operated.

e. Purpose of the low-low set function and determine which SRVs are used for this function.

f. Determine the sequence of operation of the low-low set SRV's including arming setpoints, lift points and reclose setpoints.

Material Required for Examination I

[Question Source I Facility Exam Bank Question Modification Method: Direct From Source Question Source Comments: I QID #8451 Hope Creek 02/28/1998 Saturday, March 23, 2002 10:47:29 AM Page 12 of 139

The plant is at 62% power, recovering from an inadvertent trip of the "B" Reactor Recirc pump.

Shortly after the Recirc pump was started and power ascension commenced, annunciator C8-B5 "RPV LEVEL 7" is received. The NCO notes that actual level is 39" and rising.

At this time, the required operator action is to...

II place the reactor vessel water level control system in manual.

[ verify Hydrogen Water Chemical Injection trip.

W close the Main Steam Isolation Valves.

[I reduce reactor recirc flow to minimum.

Anwra ExamLevel B LCognitiveLevel Memory F Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions Rou 2 2 295008A101 295008 High Reactor Water Level Record Number 12 AA1. Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL:

AA1.01 Reactor water level control: Plant-Specific 3.7 3.7 IExplanaton of Immediate operator action from AB-200 Answer I Reference Title HC.OP-AB.ZZ-0200 Learning Objectives OAB200EO02 (R) From memory, recall the Immediate Operator Actions for Reactor Level Control Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination QuestiionSourceI1 Facility Exam Bank Q;ueston Modification Method: Editorially Modified Question Source Comments: I Vision QID# Q53987 Saturday, March 23, 2002 10:47:30 AM Page 13 of 139

A malfunction of the Digital Feedwater Level Controller has resulted in an INCREASING reactor water level. The Reactor Feedwater Pumps are automatically tripped on a high reactor water level signal to prevent:

FiI feed pump damage due to increasing pump discharge flow rate and head.

main turbine damage due to water impingement on turbine blades.

reactor vessel damage due to completely filling and overpressurizing the vessel.

W main steam line piping and hanger damage due to filling the main steam lines.

b Exam Level B CognitiveLevel Memory Facility Hope Creek JExam Date.: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ReGroup 2 sROGroup 2 295008K304 295008 High Reactor Water Level Record Number 13 AK3. Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL:

AK3.04 Reactor feed pump trip: Plant-Specific 3.3 3.5 Explanation of Feedpumps are tripped to prevent reactor overfill and damage to the main turbine.

Answer -1 F Reference Title TC Bases 3/4.3.9 Learning Objectives 000002E008 (R) Given a list of reactor vessel pressure and/or level setpoints determine the automatic action that occurs lAW the Lesson Plan.

IMaterial Required for Examination I

[Question Source:I INPO Exam Bank [Question Mification Method: I Editorially Modified IQuestion Source Comments: j QID #6574 Dresden 03/11/1996 Saturday, March 23, 2002 10:47:30 AM Page 14 of 139

Given the following:

- A plant start-up is in progress

- Reactor power is 1%

- Recirculation loop temperature is 300°F

- "RPV LEVEL 4" alarm is received What is the actual RPV water level?

I] 24 inches

[ 27 inches W 30 inches W 33 inches Answera Le B BExa Cognitive Level] Comprehension i Hope Creek {Exam Date: 03/12/2002 Tier- Emergency and Abnormal Plant Evolutions ROGroup 1 {SRO GroupJ 295009A201 295009 Low Reactor Water Level Record Number 14 AA2. Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL:

AA2.01 Reactor water level 4.2 4.2 Ex[laation of Provide a copy of HC.OP-IO.ZZ-0003, Attachment 6 Narrow Range Answer IJCORRECT - 24 inches. Value obtained from the 250°F INCORRECT - 27 inches. Value obtained from the 350'F lines.

INCORRECT - 30 inches. Value without temperature compensation.

INCORRECT - 33 inches. Value indicated level for actual level of 30" at 450 F.

I Reference Title HC.OP-IO.ZZ-0003, Attachment 6 Learning Objectives 001 12CE005 (R) Interpret charts, graphs and tables contained within the STARTUP FROM COLD SHUTDOWN TO RATED POWER Integrated Operating Procedure to maintain plant operations within specified limits.

IMaterial Required for Examination ] HC.OP-IO.ZZ-0003, Attachment 6 page 52 I Facility Exam Bank 1Question Modification Method: Significantly Modified IQuestion Source Comments: 1 VISION BANK QID# Q56518 Saturday, March 23, 2002 10:47:30 AM Page 15 of139

Given the following:

- A LOCA has occurred

- All rods are full in

- "A" and "B" RHR Pumps are NOT available

- HPCI AND RCIC are NOT available

- Reactor water level is -150 and steady

- Reactor Feedwater Pumps are flowing 12,000 gpm each

- Reactor pressure is 1000 psig

- Drywell pressure is 45 psig and rising at 10 psig per minute

- Suppression Chamber pressure is 45 psig and rising at 10 psig per minute The EOP mitigation strategy for this event is:

E Depressurize with SRVs; inject with sources internal to the containment SDepressurize with SRVs; inject with sources external to the containment I] Inhibit ADS and remain at pressure to conserve inventory; inject with sources internal to the containment W Inhibit ADS and remain at pressure to conserve inventory; inject with sources external to the containment Anwra LCognitive S Level Application F Hope Creek Ea Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RO Group 1 1 295009G406 295009 Low Reactor Water Level Record Number 15 2.4 Emergency Procedures and Plan 2.4.6 Knowledge symptom based EOP mitigation strategies. 3.1 4.0 Explanation of Conditions provided are symptoms of a Feedwater line break inside the drywell. Drywell pressure above Answer PSP requires emergency depressurization. If Drywell pressure cannot be maintained below 65 psig, then terminate RPV injection from sources outside containment not required for adequate core cooling.

Reference Title EOP 101 Step RC/L2 10CFR55.43(5)

Learning Objectives 00124AE006 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulation prescribed by that step.

Material Required for Examination EOP Flowcharts without entry conditions lQuestion Source: I New IQuestion Modification Method:

L . . . . . . . . . . .. . . . . . .. . . . . . . . . I Question Source Comments:

Saturday, March 23, 2002 10:47:30 AM Page 16 of 139

Given the following:

- The plant is operating at 100 percent power

- Equipment Drain Sump leakage has remained constant at 2.0 gpm for 8 weeks.

- Floor Drain Sump leakage has risen steadily over several days from 1.2 g to 1.8 gpm.

At 0800 this day and hourly thereafter, operators obtained the following readings on the Floor Drain Sump:

0800 1.8 0900 2.1 1000 2.5 1100 2.7 1200 3.1 1300 3.2 1400 3.7 1500 3.9 Has a Technical Specification operational leakage limit for the Reactor Coolant System been exceeded and what is the bases for your answer?

F1 No, because total leakage has remained less than 5 gpm El No, because unidentified leakage has remained at about 2 gpm El Yes, because unidentified leakage has increased by more than 2 gpm El Yes, because total leakage has increased to more than 5 gpm A c S LCognitive Level Application Facility Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RoGrouP 1 sROGroup 1 295010A102 295010 High Drywell Pressure Record Number 16 AA1. Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

AA1.02 Drywell floor and equipment drain sumps 3.6 3.6

,xplanation of] SRO UNIQUE - RO LEVEL QUESTION Answer J Floor drain leakage is Unidentified leakage. 2 gpm or more increase in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is an entry into TS 3.4.3.2 Reference Title TS 3.4.3.2 Learning Objectives 000033E007 (R) Given a copy of the Technical Specifications, choose those sections which are applicable to the Drywell Ventilation System lAW the Drywell Ventilation System Lesson Plan.

000221E006 Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Identify those sections which are applicable to the Radiation Monitoring System lAW the Radiation Monitoring System Lesson Plan.

b. Evaluate RMS operability and determine required actions associated with Radiation Monitoring System inoperability.

c. Explain the bases for those Technical Specification items associated with the Radiation Monitoring System. (SRO only)

Saturday, March 23, 2002 10:47:30 AM Page 17 of 139

IMateral Required for Examination I Tech Specs withou t Definitions, Safety Limits, and bases lusonSour: INPO Exam Bank [Quet Modification Method; Significantly Modified Quetion Surce Comments: j QID#628 Duane Arnold 05/25/1999 Saturday, March 23, 2002 10:47:30 AM Page 18 of 139

Given the following:

- The reactor has scrammed due to rising Drywell pressure

- Drywell Floor Drain Sump Pumps have stopped running

- Drywell pressure continues to increase Which one of the following describes the reason why the sump pumps have stopped?

EI The Drywell Leak Detection (DLD) Sump Monitoring goes offscale high F1 The Reactor Recirc Seal Staging flow is isolated El The sump pump suction screens are clogged F1 The Non-lE power source is shed Anwrd ExamLevel R CognitiveLevel Comprehension c Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROeroup 1 SROGroup 1 295010A102 295010 High Drywell Pressure Record Number 17 AA1. Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

AA1.02 Drywell floor and equipment drain sumps 3.6 3.6 Explanation of Drywell Floor Drain Sump pumps are powered from 10B252 and 262 MCC's which are shed on high w drywell pressure.

- DLD goes offscale high- incorrect. This occurs but is not the reason the pumps stopped

- RR seal staging flow leakoff goes to the DW Equipment sump

- Suctions screens could clog from debris, but the pumps would continue to run Reference Title HC.OP-SO.SM-0001 Table SM-20 Learning Objectives 000086E011 (R) From memory test/identify the conditions/signals that will cause the Drywell Equipment and Floor Drain Containment Isolation valves to automatically close, lAW the Lesson Plan.

[Material Required for Examination

[Question Sour New Ques-tio- Modification Method:

Question Source Comments: I Saturday, March 23, 2002 10:47:30 AM Page 19 of 139

Given the following conditions:

- A LOCA has occurred

- Drywell temperature is 300F

- Drywell pressure is 3.0 psig Which one of the following describes the plant response when one loop of Drywell Spray is initiated?

l] Reactor vessel level indications will be lost

[i SRV operation can no longer be assured H Running Drywell cooling fans will automatically trip WDrywell pressure will drop below the scram setpoint Sd ExamLevel B Icognitivev Comprehension a Hope Creek xam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 2 2 295012K101 295012 High Drywell Temperature Record Number 18 AKI. Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:

AK1.01 Pressure/temperature relationship 3.3 3.5

[xplanation of Bases of Drywell Spray Initiation Limit states "DWSIL is the highest drywell temperature at which Answer initiation of drywell sprays will not result in an evaporative cooling pressure drop to below either: The drywell-below-wetwell differential pressure capability, or - The high drywell pressure scram setpoint.

Since the parameters given are in the UNSAFE region of the DWSIL curve, Drywell pressure will drop below the scram setpoint.

-Drywell sprays cool the areas surrounding the RPV Level instrumentation reference legs, improving reliablity

- Drywell cooling fans trip on 1.68 psig or manually before; not because sprays are initiated

- SRV operation is limited by DW temps above 340F, not sprays initiation Reference Title EOP Caution 1 Learning Objectives 00124AE006 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulation prescribed by that step.

000002E014 (R) Given changes in the following parameters, evaluate the affect on each RPV level indication lAW the Lesson Plan.

a. Reactor Pressure

b. Drywell Temperature

c. Steam Flow IMaterial Required for Examination iQuestion Source:I INPO Exam Bank Question Modification Method: I Editorially Modified Question Source Comments: QID# 8632 Hope Creek Unit 08/10/1998 Saturday, March 23, 2002 10:47:31 AM Page 20 of 139

Which one of the following describes the bases for Suppression Pool Cooling being required to be in service as a prerequisite to starting HPCI for surveillance testing?

Fa To ensure adequate thermal mixing of the water in the Suppression Pool to limit stress on the torus shell due to differential thermal expansion.

El To allow the maximum average Suppression Pool water temperature limit to be increased to 1050F.

El To extend the operating time for HPCI testing before the maximum average temperature limit is reached and testing is required to be stopped.

STo ensure that heat added to the Suppression Pool does NOT increase Suppression Chamber air space pressure to the point where the Suppression Chamber to Drywell vacuum breakers cycle.

c C Exam Level B Cognitive Level Memory Facility Hope Creek Ex Dte] 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 2 SROGroup 1 295013K201 295013 High Suppression Pool Temperature Record Number 19 AK2. Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following:

AK2.01 Suppression pool cooling 3.6 3.7

[Explaion of Reason for prerequisite 2.1.9 of quarterly surveillance test Answer Reference Title HC.OP-IS.BJ-0001 Section 2.1.9 Learning Objectives 000026E014 (R) Given plant problems/industry events associated with the HPCI system:

a. Discuss the root cause of the plant problem/industry event lAW the HPCI System Lesson Plan.

b. Discuss the HCGS design and/or procedural guidelines that mitigate/reduce the likelihood of the problem/industry event at HCGS lAW the HPCI System Lesson Plan.

c. Discuss the "lessons learned" from this problem/event lAW the HPCI System Lesson Plan.

IMaterial Required for Examination

[QuestionSource: INPO Exam Bank [Question Modification Method: I Editorially Modified Question Source Comments: J INPO BANK QID# 8883 Saturday, March 23, 2002 10:47:31 AM Page 21 of 139

Given the following:

- The plant is operating at 100% power

- Main Steam Isolation Valve AB-HV-F022A inadvertently closes Which one of the following describes the response of the reactor?

Reactor power will:

[ drop initially due to a Reactor Recirc intermediate runback when RPV level reaches +30 inches. This increases the boiling boundary length which adds negative reactivity.

initiallySrise due to the reactor pressure rising. This causes a collapse of voids in the core which adds positive reactivity.

W rise initially due to a rising core water level caused by rising reactor pressure. Power will return to a slightly lower level in response to Reactor Water Level Control and Turbine Control Valve movement.

W drop initially due to the void boundary being pushed lower in the core. As the Turbine Control Valves respond to lower reactor pressure, power rises as the void boundary rises.

Anserb ExamLevel B Cognitive Level Comprehension F Hope Creek m 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROeroup 1 SROGroup 1 295014K204 295014 Inadvertent Reactivity Addition Record Number 20 AK2. Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following:

AK2.04 Void concentration 3.2 3.3 Explanation of When the MSIV closes, steam flow is reduced as 100 percent steam flow now passes through only 3

[Answer steamlines. Reactor pressure rises and voids collapse. Void collapse causes power to rise.

Reference Title HC.OP-AB.ZZ-0202 Learning Objectives 000228E024 Given a reactor power change analyze that power change and predict how the various reactivity coefficients respond.

Material Required for Examination

[ INPO Exam Bank Question Modification Method:*I Editorially Modified

ýQuestion Source Comments: INPO BANK QID# 16307 Grand Gulf 1 04/01/2000 Saturday, March 23, 2002 10:47:31 AM Page 22 of 139

Given the following:

- The plant is operating at 60% power

- Both Reactor Recirc Pump Speed Controllers are in AUTO (Master Manual)

Which one of the following would require the operator to immediately place the Reactor Mode Switch to Shutdown Ei SIC-R620 Master Speed Control Recirc Master Demand fails full upscale F1 SIC-R620 Master Speed Control Recirc Master Demand fails full downscale I] SIC-R621A Reactor Recirc pump speed demand fails full upscale SSIC-R621A Reactor Recirc pump speed demand fails full downscale e a Exam Level B Cognitive Level Memory F Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 1 SROGroup 1 295014K301 295014 Inadvertent Reactivity Addition Record Number 21 AK3. Knowledge of the reasons for the following responses as they apply to INADVERTENT REACTIVITY ADDITION:

AK3.01 Reactor SCRAM 4.1 4.1 Explanation o Immediate operator action for dual recirc pump runaway lAW HC.OP-AB.ZZ-0204 Answer Reference Title HC.OP-AB.ZZ-0204 Learning Objectives 0AB204E002 (R) From memory, recall the Immediate Operator Actions for Positive Reactivity Addition, Abnormal Operating Procedure.

Material Required for Examination

[QSoution soe: New lQuestion Modification Method:

Q*uestionSource Comments:

Saturday, March 23, 2002 10:47:31 AM Page 23 of 139

Given the following:

- An ATWS with fuel damage has occurred

- The Emergency Duty Officer (EDO) decides that it is necessary to send someone into the Reactor Building (with Radiation Protection) to individually scram rods What is the maximum allowable dose limit that the EDO may authorize for this evolution?

E 5 REM 10 REM 25 REM 75 REM A c xam Level CognitiveLeel Memory Facility Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RO*roup 1 SROGrouP 1 295015G304 295015 Incomplete SCRAM Record Number 22 2.3 Radiological Controls 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in 2.5 3.1 excess of those authorized.

[xplanation of] The EDO may authorize 25 REM per person for emergency actions to mitigate the consequences of an Answer accident.

Reference Title NC.EP-EP.ZZ-0304 Sect 5.2 10CFR55.43(4)

Learning Objectives IMaterial Required for Examination

[Question Soueiý INPO Exam Bank [Question Mdification Method: I Editorially Modified Question Source Comments: INPO EXAM BANK Question ID# 7986. Hatch 3/97 modified for Hope Creek. EP for Licensed Operators. Lesson Plan Saturday, March 23, 2002 10:47:31 AM Page 24 of 139

Given the following:

- The reactor scrammed from 100 percent power

- Reactor power is on the Source Range Monitors

- 3 rods remain at position "48"

- Scram air header reads 0 psig

- The scram CANNOT be reset lAW EOP Bases, which one of the following methods of achieving shutdown condition is best for these conditions?

[i Vent control rod over-piston areas to insert rods rE De-energize scram solenoids to insert rods Hi Defeat Rod Worth Minimizer to insert rods

[ Initiate Standby Liquid Control to inject boron c Exam Level S Cognitive Level I Comprehension a Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 1 1 295015K301 295015 Incomplete SCRAM Record Number 23 AK3. Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM:

AK3.01 Bypassing rod insertion blocks 3.4 3.7 Explanation of RC/Q-21 states "Drive control rods, defeat RWM interlocks if necessary. This method is best applied Answer ]when only a few control rods cannot be inserted". The scram cannot be reset. Scram solenoids are already de-energized. Power is less than 4% so SBLC should not be used.

Reference Title HC EOP Bases step RC/Q-21 10CFR55.43(2)

Learning Objectives 00124BE008 (R) Given any step of the procedure, explain the reason for performance of that step and/or evaluate the expected system response to control manipulations prescribed by that step.

IMaterial Required for Examination EOP Flowcharts without entry conditions Question Source: I New lQuestion Modification Method:

Question Source Comments:

Saturday, March 23, 2002 10:47:32 AM Page 25 of 139

Given the following:

- The plant was operating at 100% power

- Toxic gas concerns have required the Main Control Room to be evacuated

- The transfer of controls to the Remote Shutdown Panel have been completed Which of the following systems are available for reactor vessel pressure control from the Remote Shutdown Panel?

Ri SRV's F, H & M and RHR Shutdown Cooling E] Turbine Bypass Valves and Reactor Core Isolation Cooling I] Reactor Feed Pumps and Reactor Recirculation W High Pressure Coolant Injection and LO-LO SET SRVs a xam Level B Cognitive Level Memory Facility Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGrouP 2 SRO Group 1 295016A102 295016 Control Room Abandonment Record Number 24 AA1. Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

AA1.02 Reactor/turbine pressure regulating system 2.9 3.1 Explanation of 10-8 initiates rpv cooldown with SRV's F,H, & M until SDC can be established. Subsequent actions of AB Answer :130 Trip the main turbine and close the MSIVs. Recirc pumps are manually tripped and discharge valves closed. HPCI cannot be controlled from the RSP. LO LO SET SRVs are only controlled from the Control Room.

Reference Title HC.OP-IO.ZZ-0008 HC.OP-AB.ZZ-0130 Learning Objectives 00112HE006 (R) Analyze plant conditions and parameters to determine if plant operation is in accordance with the SHUTDOWN FROM OUTSIDE THE CONTROL ROOM Integrated Operating Procedure, supporting System Operating Procedures and Technical Specifications.

IMaterial Required for Examination lQuestion source INPO Exam Bank iQuestion Modification Method: Editorially Modified IQuestion Source Comments: INPO EXAM BANK QID# 18592 Peach Bottom 2 09/19/1997 Saturday, March 23, 2002 10:47:32 AM Page 26 of 139

Given the following:

- A plant shutdown is in progress

- North Plant Vent RMS is in HIGH alarm

- South Plant Vent RMS is reading 4.5 e+2 uCi/sec

- FRVS Vent RMS is reading 6.5 e-2 uCi/sec

- FRVS is NOT in service Which one of the following is the source of the high alarm?

EI Service Area Exhaust System IE Solid Radwaste Exhaust System I] Radwaste Area Exhaust System WTurbine Building Exhaust System b Exam Leve S lCognitive Level Comprehension a Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 2 SROGroup 1 295017A201 295017 High Off-Site Release Rate Record Number 25 AA2. Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:

AA2.01 Off-site release rate: Plant-Specific 2.9 4.2 Explanation of HC.OP-AB.ZZ-0126 Answe JSolid Radwaste Exhaust discharges to North Plant Vent Stack Others discharge to the South Plant Vent Reference Title HC.OP-AB.ZZ-0126 10CFR55.43(4)

Learning Objectives 000114E003 (R) Discuss the operational implications of the abnormal indications/alarms for system operating parameters related to a given Abnormal Operating Procedure.

IMaterial Required for Examination LQestion Source: Facility Exam Bank lQuestion Modification Method: Direct From Source IQuestion Source Comments: I VISION BANK QID# Q55943 Saturday, March 23, 2002 10:47:32 AM Page 27 of 139

Given the following:

- Marsh Grass intrusion has reduced the flow in Service Water Loops "A" & "B"

- The differential pressure across the "A" SSW Pump Strainer is being reduced to maximize strainer backwash operation Per HC.OP-AB.ZZ-0122, Service Water System Malfunction, why should the discharge valve of "A" SSW Pump be closed for no more than two minutes during this evolution?

i] All SSW flow from SSW Loop "A" to RACS and SACS will be lost

[I Lubricating water flow will be lost to SSW Pump "A".

WSpray Water Booster Pump "A" will remain stopped by interlock WBlockage problems could worsen on other SSW Pump strainers d EeL Memory Faciity Hope Creek Exam Date.] 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 2 sRo Group 2 295018A203 295018 Partial or Complete Loss of Component Cooling Water Record Number 26 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

AA2.03 Cause for partial or complete loss 3.2 3.5

[......atio..of SRO UNIQUE - RO LEVEL QUESTION Answer CORRECT - Blockage problems could worsen on other SSW Pump strainers. Per Caution 4.6.4.D of HC.OP-AB.ZZ-0122, pump operation should be limited to 2 minutes with the discharge valve closed.

Closing the discharge path on one pump may compound blockage problems on other pumps by increasing differential pressure for those strainers.

INCORRECT - Lubricating water flow will be lost to SSW Pump A. Lubricating water flow is supplied from the Lubrication Head Tanks upstream of the SSW Pump discharge valve.

INCORRECT - Spray Water Booster Pump A will remain stopped by interlock. The Spray Water Booster Pump is stopped when its own discharge valve is shut, not when the SSW Pump discharge valve is shut.

INCORRECT - All SSW flow from SSW Loop A to RACS and SACS will be lost. SSW Loop A flow can be maintained by operating the C SSW Pump while the SSW Pump A flow is stopped.

Reference Title HC.OP-AB.ZZ-0122, Caution 4.6.4.D Learning Objectives 0AB122E004 Explain the reasons for how plant/system parameters respond when implementing, Service Water System Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination I

[Question Source: I Facility Exam Bank Question Modification Method: Direct From Source IQuestion Source Comments: Vision Bank QID# Q61195 Saturday, March 23, 2002 10:47-32 AM Page 28 of 139

Given the following:

- A leak on the Instrument Air header has resulted in lowering header pressure.

- The "INST AIR HEADER A PRESSURE LO" annunciator alarm is received.

Which one of the following valves automatically open to restore header pressure and at what pressure?

Fj The Instrument Air Dryer 1A-F-104 outlet valve KB-HV-11416; 70 psig E The Instrument Air Dryer 1A-F-104 outlet valve KB-HV-11416; 85 psig Pl The Instrument Air Dryer 10-F-104 outlet valve KB-HV-7618; 70 psig FW The Instrument Air Dryer 10-F-1 04 outlet valve KB-HV-7618; 85 psig b Exam Level B Cognitive Level Memory Facility Hope Creek Exar Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 2 SROGroup 2 295019A102 295019 Partial or Complete Loss of Instrument Air Record Number 27 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

AA1.02 Instrument air system valves: Plant-Specific 3.3 3.1

[xplanation of Instrument Air Dryer AF-1 04 outlet valve will automatically open at 85 psig on lowering air pressure.

S..Reference Title HC-OP.AB-ZZ-0131 Learning Objectives OAB131E004 Explain the reasons for how plant/system parameters respond when implementing, Loss Of Instrument Air And/Or Service Air, Abnormal Operating Procedure.

IMaterial Required for Examination

[Question Souc: E INPO Exam Bank Question Modification Method: Significantly Modified Ques rce Comments: INPO Exam Bank QID #6871 Dresden 2 07/28/1999 Saturday, March 23, 2002 10:47:32 AM Page 29 of 139

Given the following:

- The plant is operating at 50 percent power during a startup

- Overhead alarms received - "MSIV CLOSURE"

- All 4 Outboard MSIV's OPEN and CLOSED indication lights are illuminated Which one of the following would cause the alarm condition?

Z Degrading Instrument Air header pressure

[ Degrading Instrument Gas header pressure ILoss of solenoid power to the MSIV 4-way "operator valves" W Loss of solenoid power to the MSIV "test valves" Anwra ExamLevel R JCognitiveevel Comprehension Hope Creek ExamnDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROGroup 2 2 295019A201 295019 Partial or Complete Loss of Instrument Air Record Number 28 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

AA2.01 Instrument air system pressure 3.5 3.6

[xplanation of Degrading air header pressure on the Reactor Bldg supply header which supplies the outboard MSIVs w would cause all four to slowly close. Loss of solenoid power to the test valves will have no effect because they are normally deenergized. "4 way valves" are air powered.

Reference Title HC.OP-AB.ZZ-0131 Attachment 1 Learning Objectives 000046E014 (R) Concerning the Main Steam Isolation Valves (MSIV's), summarize, list or identify the following lAW the lesson plan.

a. Assess the effect on a MSIV if loss of electric or loss of pneumatic supply occurs.

b. Determine the signals which will automatically close the MSIV's and when, if ever, certain isolations can be bypassed.

IMaterial Required for Examination

[Questiourc New Question Modification Method:

[Question Source Comments:

Saturday, March 23, 2002 10:47:32 AM Page 30 of 139

Given the following:

- The reactor has been in COLD SHUTDOWN for two (2) days following power operation

- Reactor vessel water level is +30 inches

- Neither Reactor Recirculation pump is available

- Shutdown Cooling has isolated and the Shutdown Cooling suction valves CANNOT be opened

- The highest RPV metal temperature is 190°F and rising

- HC.OP-AB.ZZ-0142, Loss of Shutdown Cooling has been entered Based on given information, which one of the following is the highest Reporting Requirement/ECG classification applicable?

[] 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report W Unusual Event W Alert A d Exam Level S CognitiveLevel Application j*Iil Hope Creek Exam Date 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 3 2 295021G441 295021 Loss of Shutdown Cooling Record Number 29 2.4 Emergency Procedures and Plan 2.4.41 Knowledge of the emergency action level thresholds and classifications. 2.3 4.1 Exp[anatinof ECG EAL 8.1.2 Inability to maintain the plant in Cold Shutdown Answer HCReference Title HO ECG EAL 8.1.2 10CFR55.43(5)

Learning Objectives IMaterial Required for Examination Question Source: J New lQuestion Modification Method:I Quftion Source Comments: EP Lesson Plan Saturday, March 23, 2002 10:47:33 AM Page 31 of 139

Given the following

- The plant is operating at 100 percent power

- "A" CRD Pump is C/T for maintenance

- CRD SYSTEM TROUBLE overhead alarm C6-F2 comes in

- CRD Cooling Water flow drops to zero gpm What actions are required and what is the bases for those actions?

El Scram the reactor upon the receipt of the second accumulator trouble alarm based on demonstrated Shutdown Margin IE Scram the reactor upon the receipt of the second accumulator trouble alarm based on average control rod scram times L] Scram the reactor within 20 minutes based on adequate time to place a CRD pump back in service W Scram the reactor within 20 minutes based on the ability for charging header pressure alone to fully insert all control rods C Exam LevelS Cognitiv el Memory I Hope Creek 03/12/2002 Tier* Emergency and Abnormal Plant Evolutions 2 2 295022G448 295022 Loss of CRD Pumps Record Number 30 2.4 Emergency Procedures and Plan 2.4.48 Ability to interpret control room indications to verify the status and operation of system, and 3.5 3.8 understand how operator action s and directives affect plant and system conditions.

xplanation of Tech spec bases 3/4 1.3. The question is based on the TS bases for an operator manual scram time Answer.. requirements on a loss of both CRD pumps.

Reference Title Tech spec bases 3/4 1.3 10CFR55.43(2)

Learning Objectives 000006E033 (R) Given a scenario of applicable operating conditions and access to Technical Specifications complete each of the following lAW Technical Specifications:

a- Select those sections applicable to the CRDH System.

b. Evaluate CRDH System operability and determine required actions and time limits associated with inoperable components

c. Explain the bases for those Technical Specification sections associated with the CRDH System. SRO ONLY Material Required for Examination Tech Specs without Definitions, Safety Limits, and bases I etionSourceK: New IQuestion Modification Method: I Quýstion Source Comments: I Saturday, March 23, 2002 10:47:33 AM Page 32 of 139

Given the following:

- The plant is at 37% power

- Both CRD pumps are tripped on low suction pressure

- The Reactor Building Operator is swapping CRD suction filters

- CRD ACCUM TROUBLE Overhead Annuciator C6-D4 is clear (Assume NO other operator actions)

Which one of the following describes the effect on gas pressure in the HCU Accumulators 2 minutes following the pump trip?

[I Stays the same because reactor pressure holds the charging water check valve closed SStays the same because accumulator pressure holds the charging water check valve closed H] Lowers because the reactor scrams WLowers because the accumulator piston moves when charging water header pressure is lost A b ExamLevel B Cognitive Level Comprehension l Hope Creek Exam Date 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROeroup 2 sRQ roup 2 295022K203 295022 Loss of CRD Pumps Record Number 31 AK2. Knowledge of the interrelations between LOSS OF CRD PUMPS and the following:

AK2.03 Accumulator pressures. 3.4 3.4 Explanaion of Charging water check valve 115 maintains water volume on a loss of charging pressure from the CRD Answer pumps. N2 gas pressure will remain the same as long as the check valve holds. If the check valve does not hold, the piston will stroke and N2 pressure will drop causing low accumulator pressure alarm.

Reference Title HC.OP-IS.BF-0103 Purpose Lesson Plan 00006 Learning Objectives 000006E017 (R) Given the appropriate procedure or access to the procedure, summarize the accumulator trouble alarms and their setpoints associated with each CRD HCU and how these problems may impact CRDH System Operation, lAW the Lesson Plan.

IMaterial Required for Examination lQuestion Source: New IQuestion Modification Method:

Queston Source Comments:

Saturday, March 23, 2002 10:47:33 AM Page 33 of 139

Given the following:

- "A" Fuel Pool Cooling Pump is tagged for maintenance

- "B" Fuel Pool Cooling Pump trips How does this affect the ability to monitor Fuel Pool temperature in the Control Room?

[1 Temperature recorder TR-4683 is unaffected because it monitors Skimmer Surge Tank temperature iThe High Temperature alarm to Fuel Pool System Trouble (D1-D5) is INVALID because it monitors Skimmer Surge Tank temperature

[ The High Temperature alarm to Fuel Pool System Trouble (D1-D5) is VALID because it monitors Fuel Pool Cooling Pump common discharge piping W Temperature recorder TR-4683 is INVALID because it monitors Fuel Pool Cooling Pump common discharge piping d ExaLevel B Cognitive Level] Comprehension a Hope Creek 03/12/2 002 Tier: Emergency and Abnormal Plant Evolutions 3 SROGroup 1 295023A102 295023 Refueling Accidents Record Number 32 AA1. Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS:

AA1.02 Fuel pool cooling and cleanup system 2.9 .1 Explanatio ofl Both Temp alarm and Recorder TR-4683 monitor the same parameter in the common discharge piping IAnswer Jof the FPCC pumps. With no flow, the piping will equalize with ambient air temperature, no longer valid indication of Fuel Pool Temperature.

I AReference Title HC.OP-AR.ZZ-0013 Attachment D5 Learning Objectives 000043E017 (R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the Control Room, identify the status of the FPCCS or its components by evaluation of the controls/instrumentation/alarms, lAW the Fuel Pool Cooling and Cleanup System (FPCCS) Lesson Plan.

Drawing of alarm window D1-D5. Drawing of TR-4683 IQFlaterial Required for Examination

uestion Soure: New lQuestion Modification Method:

Question Source Comments:

Saturday, March 23, 2002 10:47:33 AM Page 34 of 139

Given the following:

- Core offload is in progress

- A fuel bundle was removed from the reactor vessel, full up on the Fuel Hoist, and in the cattle chute heading for the Fuel Pool

- Fuel Pool Skimmer Surge Tank Level is lowering rapidly Which one of the following describes the operator actions required?

El Place the bundle into its original reactor core location Ei Place the bundle into the Fuel Prep Machine

[i Stop the bridge at its current location and leave the refueling floor W Stop the bridge at its current location and lower the bundle full down Anwra Exam Leve B lCognitive Level Memory a Hope Creek ExamDate: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 3 SROGrup 1 295023G411 295023 Refueling Accidents Record Number 33 2.4 Emergency Procedures and Plan 2.4.11 Knowledge of abnormal condition procedures. 3.4 3.6

[Explanation f Immediate operator action on loss of fuel pool inventory/cooling is to return the bundle to either the Answer reactor vessel or the fuel pool. Tech Spec definition of Core Alterations allows continued movement of a component to a safe location.

Reference Title HC.OP-AB.ZZ-0144 Learning Objectives 0AB144E002 (R) From memory, recall the Immediate Operator Actions for Loss Of Fuel Pool Inventory/Cooling, Abnormal Operating Procedure.

Material Required for Examination lQueStonSouýýI New Question Modification Method:

1Questin Source Comments:

Saturday, March 23, 2002 10:47:33 AM Page 35 of 139

Given the following:

- The reactor at rated power

- Hope Creek experiences a Loss of Offsite Power event and a reactor scram

- Approximately 13 sec into the event, Drywell pressure is 1.9 psig Which one of the following describes the operation of the LOCA and LOP sequencers?

Ra1 The LOP sequencer program will be in control of restoring the loads.

EJ The LOP sequencer will complete sequencing 2 minutes later, then the LOCA sequencer will start.

E. The LOCA sequencer program will be in control of restoring the loads.

W1 The LOCA sequencer will complete sequencing 2 minutes later, then the LOP sequencer will start.

Anserc ExamLevl B CogitveLevl Memory a Hope Creek EF-xam Date:J 03/12/2002

,Tier: Emergency and Abnormal Plant Evolutions ROeroup 1 [SRO GroupI 295024A110 295024 High Drywell Pressure Record Number 34 EA1. Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

EAI.10 A.C. distribution 3.4 3.6 Explanation of Justification Answer I The LOCA Sequencer takes priority over the LOP Sequencer to ensure that all systems required to protect the core and prevent radioactive release are sequenced on when required and that unnecessary loads do not overload the diesels I Reference Title HC.OP-SO.KJ-0001 Learning Objectives 000066E012 Summarize/identify the emergency load sequencer response for a LOP concurrent with a LOCA signal lAW Attachment 1 of the Lesson Plan.

IMateral Required for Examination l Facility Exam Bank lQuestion Modification Method: I Editorially Modified IQuestion Source Comments: j VISION BANK QID# Q53753 Saturday, March 23, 2002 10:47:34 AM Page 36 of 139

Given the following:

- A small Reactor Coolant leak in the Drywell occurs

- Drywell Leak Detection System alarms

- Drywell pressure is rising Which one of the following actions requires CRS authorization prior to performance?

I] Start an Emergency Diesel Generator following failure to start Primary Containment Instrument SRestore Gas following isolation LI Maximize Drywell cooling prior to high Drywell pressure alarm E] Terminate Drywell inerting if in progress b Exam Level B Cognitive Level Memory lFacilit Hope Creek Exam Date:] 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions ROQroup 1 SROGroup 1 295024G106 295024 High Drywell Pressure Record Number 35 2.1 Conduct of Operations 2.1.6 Ability to supervise and assume a management role during plant transients and upset conditions. 2.1 4.3

[Expanation o, Restoring PCIG following isolation requires overriding LOCA signals. Overriding Tech Spec required

[Answer isolation to Primary Containment Isolation Valves requires SRO authorization. This directed by the CRS through implementation of EOP-101 or 101A.

Start an EDG following failure - incorrect - Immediate operator action of AB-1 35 Maximize DW Cooling - incorrect - Immediate operator action of AB-201 Terminate DW inerting - incorrect - Immediate operator action of AB-201 Reference Title HC.OP-EO.ZZ-0101 step RC/P-5 Learning Objectives 000113E079 State the three (3) conditions when a facility must evaluate proposed actions.

IMaterial Required for Examination LQuestion Sou*e: j Facility Exam Bank [Question Modification Method: Editorially Modified ce Comments: VISION BANK QID# Q57081 Saturday, March 23, 2002 10:47:34 AM Page 37 of 139

Given the following:

- The reactor is operating at 100% power

- A spurious Main Turbine trip occurs

- The reactor scrams with all rods going full in

- Turbine Bypass valves fail to operate properly resulting in a reactor pressure excursion up to 1100 psig What is the impact on the Digital Feedwater Level Control System?

(Assume no operator action)

[i Operation of RFP Controllers in MANUAL is available after 55 seconds 1] Operation of all controllers is automatically restored in 12.5 minutes W Operation of RFP Controllers is available in MANUAL or AUTO until the RFP's trip WThe Master Level Controller will stay at its original demand signal for 10 seconds SC R cognitiveCLevl Comprehension a Hope Creek 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 1 SROGroup 1 295025A206 295025 High Reactor Pressure Record Number 36 EA2. Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:

EA2.06 Reactor water level 3.7 3.8 pAnsweof Reactor high pressure causes RRCS initiation. ARI initiates but Feedwater Runback requires APRMS Ane Inop or Not Downscale (ATWS). This question tests the operators ability to determine the post transient control of reactor level.

Reference Title HC.OP-SO.AE-0001 Learning Objectives 000059E015 (R) From memory, describe the three possible RFP runback signals including conditions, setpoints and time delays if applicable, lAW the Feedwater Control System Lesson Plan.

[Material Required for Examination lQuestion Source: INPO Exam Bank Question Modification Method: Significantly Modified Question Source Comments: INPO EXAM BANK QID# 2336 Perry 08/30/1997 Saturday, March 23, 2002 10:47:34 AM Page 38 of 139

Given the following:

- The plant was operating at 100% power

- A transient occurs

- RPV pressure reached 1330 psig before turning downward WHICH ONE of the following states the required action(s) for RPV pressure reaching 1330 psig?

El Prepare and submit a Safety Limit Violation Report within 30 days.

Ei Restore to within limits within 15 minutes or be in COLD SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . I

[i Restore to within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

[i Perform an engineering evaluation on the out-of-limits condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Ansera Exa Lee B Cognitive Level Application FacilitHope Creek Exam t 03/12/2002 Tier: -Emergency and Abnormal Plant Evolutions RO *oI -i 1 O Grop 1 295025K105 295025____j High Reactor Pressure Record Number 37 EKI. Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE:

. 4.4 4. 7 EKi .05 lExceeding safety limits F lantionof Action for Safety Limit Violation is specified in TS Admin controls section 6.7.1 .d F- Reference Title HC Tech Specs 6.7.1.d Learning Objectives 000110E002 (R) Given Technical Specifications, determine the administrative and operational actions that must be performed if a Safety Limit is violated.

Material Required for Examination Tech Specs without Definitions, Safety Limits, and bases INPO Exam Bank .Question Modification Method SSouce Modified I Significantly FQuestion source Comments: IINPO BANK QOD# 13957 Palo Verde 11/18/1996 Saturday, March 23, 2002 10:47:34 AM Page 39 of 139

Given the following:

- A LOCA has occurred

- Drywell temperature is 240'F

- Suppression Chamber pressure is 7.5 psig

- Suppression Pool temperature is 125 F and rising Which one of the following describes the bases for initiating Suppression Chamber Spray at this pressure?

iTo prevent exceeding the negative design pressure of the primary containment.

ITo reduce primary containment pressure by condensing steam which may be present in the Suppression Chamber airspace.

STo reduce accumulation of non-condensibles in tihe Suppression Chamber.

WTo prevent Drywell depressurization that exceeds the capacity of the Suppression Chamber to Drywell vacuum breakers.

Anrb Lel B-- Cognitive Level Comprehension F yope Creek Exam a03/12/2002 Tier: Emergency and Abnormal Plant Evolutions R...r.. p 2 Rru 1 295026K102 295026 Suppression Pool High Water Ternerature Record Number 38 EKi... Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

EKI.U02 Steam condensation 3.5_3.8 Explanation of Suppression chamber sprays are initiated below 9.5 psig to reduce primary containment pressure by condensing steam which may be present in the SC airspace.

Reference Title HC EOP Bases step DW/P-5 Learning Objectives 00126AE009 (R) Given plant conditions and access to EOPs, select the value of the Suppression Chamber Spray Initiation Pressure and explain the basis for this limit lAW the Primary Containment Control - Drywell Lesson Plan.

[Material Required for Examination

[Qustion Sourc INPO Exam Bank .Question Modification Method: Significantly Modified IQuestion Source Comments: ,INPO EXAM BANK QID# 8062Hope Creek Unit 09/28/1997 Saturday, March 23, 2002 10:47:34 AM Page 40 of 139

Given the following:

- The Reactor has scrammed

- A small break occurred on the RPV head vent line

- Drywell temperature is 330'F and rising

- Drywell sprays are NOT available Emergency Depressurization is required to prevent exceeding which one of the following?

F] Readable range of Dry well tempeerat ure inst rumentation

[i Maximum capacity of the Drywell CooIing system saturation temperature for the Drywell design pressure S

Environmental qualification temperature of safety related equipment in the Drywell Answer daciite d Eam Cne e Comprehension Hope Creek Exam Date 03/12/2002 Tier: Emergency And Abnormal Plant Evolutions REvoouti2osR..Grou. 2 295028K102 295028 High Drywell Temperature .Number R.ecord ! ---- 39 EKI. Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:

EK!.02 Equipment environmental qualification 2.9 3.1 lAW EOP 102 bases step DW/T-3, 340°F is the qualification limit for ADS as well as the Drywell design lxplanionof temp. ED before 340 so that ADS valves can be used.

Reference Title SEOP 102 bases step DW/T-3 and DW/T-5 Learning Objectives 00126AE007 (R) Given any step of the procedure, determine th~e reason for performance of that step and/or predict expected system response to control manipulations prescribed by that step lAW the Primary Containment Control - Drywell Lesson Plan.

Material Required for Examinaion EOP Flowcharts without entry conditions

[Queston Source] Facility Exam Bank lQuestion Modification Method: Significantly Modified.

Quetio Sorce Comments-Saturday, March 23, 2002 10:47:34 AM Page 41 of 139

Given the following:

- Reactor is scrammed

- Suppression Pool level is lowering If Suppression Pool level reaches 49 inches, which one of the following would occur?

I] Reactor Building to Suppression Chamber Vacuum Breakers close if open

[ Reactor Building to Suppression Ch-amber Vacuum-Brea-kers open if closed SDrywell to Suppression Chamber differential pressure increases SDrywell to Suppression Chamber differential pressure decreases d x L S Comprehension opeCreek Exm 03/12/2002 Tier.: Emergency and Abnormal Plant Evolutions ROGroup 2 RG 1 295030A204 295030 Low Suppression Pool Water Level Record Number 40 EA2. Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

EA2.04 Drywell/ suppression chamber differential pressure: Mark-l&IlI 3.5 3.71

[-panatonof SRO UNIQUE- RO LEVEL QUESTION At 55 inches in the SC, the Vent Header drain pipe uncovers causing differential pressure to equalize.

Reference Title iEOP. 102 step SPIL-5 bases Learning Objectives 00125AE009 (R) Given any step of the procedure, determine the reason for performance of that step and/or predict expected system response to control manipulations prescribed by that step lAW the Primary Containment Control - Suppression Pool Lesson Plan.

Material Required for Examinatioa leonSouý] New .Question Modification Method:

lQuestion Source Comments: -

Saturday, March 23, 2002 10:47:35 AM Page 42 of 139

Given the following:

- A plant shutdown is in progress

"- "A" RHR is tagged for motor replacement

"- "B" RHR is in Shutdown Cooling at 210°F

- Suppression Pool Level Low annunciator alarms

- The PO reports Suppression Pool level is lowering Which one of the following makeup sources must be used?

ISuppression Pool Makeup from HPCI using OP-EO.ZZ-0312 ISuppression Pool Makeup from RCIC using OP-EO.ZZ-0313 F1 Suppression PooI Makeup from Service Water Using OP-EO.ZZ-0314 R Suppression Pool Makeup from Core Spray using 0P-E0.ZZ-0315 d B Cognitive Level !Comprehension . Hope Creek Ex-am Date:] 03/12/2002 Tier- Emergency and Abnormal Plant Evolutions I Grup 2 ISROrUPI 1 295030K203 295030 Low Supppression Pool Water Level Record Number 41 EK2. Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following:

EK2.03 LPCS 3.8 3.91 E of The plant is in Op Cond 3. Core Spray must be used because it is the only source available for the given Aseconditions. HPCI and RCIC do not have steam to run. SSW cannot be used due to B RHR is in SDC mode.

IReference Title HC EOP step SP/L-4 Learning Objectives 00125AE009 (R) Given any step of the procedure, determine the reason for performance of that step and/or predict expected system response to control manipulations prescribed by that step lAW the Primary Containment Control - Suppression Pool Lesson I Plan.

jMaterial Required for Examination I EOP Flowcharts without entry conditions lQuestion Modification Method:

Saturday, March 23, 2002 10:47:35 AM Page 43 of 139

Given the following:

- I&C surveillance testing has caused an inadvertent RPS scram signal

- RPS actuates but some rods remain out with power at 2%

- RPV level lowers until RCIC and HPCI initiate

- Operators commence recovering level with Feedwater

- RPV level was below Level 2 for 15 seconds

- The Main Turbine is still on-line (Assume NO other operator actions)

Which one of the following describes the status of RRCS?

an ARI valves are energized and RPT breakers are open I RPT breakers are closed and ARI vaIves are de-energized.

F11 Feed pumps have runback to minimum and RPT breakers are closed EW ARI valves are energized and SLC pumps will initiate when 3.9 minute timer times out Anwera B Cognitive Leve Comprehension Facility Hope Creek Em Date: 03/12/2002 Tier* Emergency and Abnormal Plant Evolutions R. Gro 295031 K213 295031 Reactor Low Water Level Record Number 42 EK2. Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

EK2.13 ARI/RPT/ATWS: Plant-Specific 4.1 4.2

[Expanatiuonof SLC will not initiate, power < 4%

Answer Feedpumps will not runback, no 1071 psig signal.

ARI will actuate, energizing the valves.

RPT breakers will be open, level was < -38 for >9 seconds.

Reference Title HC.OP-SO.SA-0001 Learning Objectives 000024E007 (R) From memory, predict the sequence of events which occur within the Redundant Reactivity Control System upon:

a. Automatic initiation in response to a high reactor vessel pressure condition with or without the APRM permissive, lAW the Lesson Plan.

b. Automatic initiation in response to a low reactor vessel water level condition with or without the APRM permissive, lAW the Lesson Plan.

c. Manual initiation with or without the APRM permissive, lAW the Lesson Plan.

Material Required for Examination esoSou Facility Exam Bank Question Modification Method: Editorially Modified Queto uýrce Comments: Vision Bank OID# Q53742 Saturday, March 23, 2002 10:47:35 AM Page 44 of 139

Given the following:

- The plant is in Operational Condition 4 for vessel disassembly

- Due to mishandling of the Reactor Vessel head insulation package, all 3 channels of Refuel Floor Exhaust RMS unexpectedly alarm HIGH on the RM-1 1

- PCIS responds normally Which one of the following is the highest Reporting Requirement/ECG classification (if any) applicable?

[] NOT Reportable Eli 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report El Unusual Event R1 Alert Cognitive LeveAplication Hope Creek x Date: 03/12/2002 Tie-r: -Emergency and Abnormal Plant Evolutions 22 295034G430 295034 Secondary Containment Ventilation High _Radiatin Re.ord. Nmber 43 2.4 Emergency Procedures and Plan 2-4.30 Knowledge of which events related to system operations/status should be reported to outside 2.2 3.6

,agencies.

Explanaion of RAL 133Bases - Valid actuation of listed systems listed in the Tech Bases. The RFE RMS responded LAnse J to valid Hi radiation conditions from the radiography. The actuations were not part of a pre-planned test.

Therefore, the ESF is reportable.

Reference Title ECG Section 11.3.3 bases 10CFR55.43(5)

Learning Objectives Material Required for Examina ECG and ECG Technical Bases ESF Actuation Flow chart page 2 and 3 of 4 l[uston SouHeI] New Question Modification Method:

estiSource Comments: I EP Lesson Plan Saturday, March 23, 2002 10:47:35 AM Page 45 of 139

Given the following:

- The reactor is operating at 100% power

- Annunciator B1-B3 ( RCIC PUMP ROOM FLOODED) alarms with the following alarm message presented on the CRIDS display: D2887 RCIC PUMP RM 4151-1 LSH 4151-1 HI

- An investigation reveals that Reactor Building Floor Drain Sump pumps have been running continuously for 10 minutes

- The Reactor Building Operator reports the leak is coming from the CST suction line In addition to running the sump pumps, which of the following action(s), if any, is required by EOP 103/4?

I --- Isolate RCIC II -- Immediately commence a normal reactor shutdown III -- Runback reactor recirculation and manually scram the reactor IV - Emergency depressurize the reactor

[c-1I and 11 I , 111, and IV Awea xB cntiLelApplication Hope Creek mDae .. 03/12/2002 Tier.:, Emergency and Abnormal Plant Evolutions R..r.o. 3 SROGroup 2 295036K201 295036 Secondary Containment High Sump/Area Water Level Record Number 44 EK2. ýKnowledge of the interrelations between SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL and the following:

EK2.01 Secondary containment equipment and floor drain system 3.1 3.2 ExpaionOf The source of the leak is RCIC suction from the CST. Step RB 14 applies since RCIC is not required to Aw assure adequate core cooling, shutdown the reactor, protect primary containment integrity, or suppress a fire. Reactor coolant is not the source of the leak based on RBO report. RB-15 is answered NO. Only iI one area is affected therefore Step RB-22 is not reached.

Reference Title EOP 103/4 step RB-12 "LearningObjectives 000127E006 (R) Given any step in the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by the step.

Material Required for Examinaion EOP Flowcharts without entry conditions

[ uesti New Buan QiD Question Modification Method:

Qiuestion Source Comments: vision Bank QID# 056139 concept used Saturday, March 23, 2002 10:47:35 AM Page 46 of 139

HC.OP-EO.ZZ-103/4, "Reactor Building Control", requires an Emergency Depressurization of the RPV if the Maximum Safe Operating Limit is exceeded in 2 or more areas listed in Table 2 Column 2.

SELECT the BASES for this Emergency Depressurization of the RPV.

II To reduce the maximum Iodine release allowable during a MSL break accident ITo prevent release of fission products into the Reactor Building by preventing fuel damage El To reduce the driving head and, therefore, the flow of the unisolated leaking Primary System Fl To protect personnel from high temperature environments while operating equipment c B Cognitive Level Memory a Hope Creek Exam Date: 03/12/2002 Tier: Emergency and Abnormal Plant Evolutions RO~roup 3 R Grou 2 295036K301 295036 _Secondary Containment High Sump/Area Water Level Record Number 45 EK3. Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

EK3.01 Emergency depressurization 2.6 2.81 EOP Bases states "RPV depressurization places the primary system in its lowest energy state, rejects Lof heat to the suppression pool in preference to outside containment, and reduces the driving head and flow of the primary systems that are unisolated and discharging into the reactor building."

Reference Title EOP 103 Bases Step RB-19 Learning Objectives 000127E003 :1 R) Define the term "Maximum Safe Operating Temperature".

Material Required for Examination Facility Exam Bank IQuestion Modification Method: Editorially Modified Vision Bank QID# Q56112 Saturday, March 23, 2002 10:47:36 AM Page 47 of 139

Which one of the following gaseous radioactive release limits corresponds to the EOP-1 04 entry condition?

.:500 mRem to the Thyroid CEDE S5000 mRem to the Thyroid CEDE Fc 2 times 10CFR 20 Appendix B limits times 10CFR 20 Appendix B limits S200 Answer.d Exa, Lel R.Cognitive Leve" Mem.ory Hope Creek 03/12/ 2002 Tier: Emergency and Abnormal Plant Evolutions ROGop 2 SOGop I295038A20 3

.295038..... High Off-Site Release Rate Numbr . 46 E A2. Abbility to determin r and/or .ne inpterpret th.e following as they apply to HIGH OFF-SITE RELEASE RATE:

EA2.03 :Radiation levels . . . 13.5 4.3 Explanation of JuSTiFiCATIO N:..

Anwer CORRECT - lAW ECG Section 6 and Lesson plan 0302-000.00H-000127, the alert value is 200 times the IOCFR20 Appendix B value Reference Title ECG Section 6.0

!LP 0302-000.OOH-000127 Learning Objectives 000127E002 Given a set of plant conditions, analyze and determine if entry conditions into HC.OP-EO.ZZ-01 03/4 exists.

IQuestion Modification Method; Saturday, March 23, 2002 10:47:36 AM Page 48 of 139

Given the following:

- A LOCA outside primary containment and the Reactor Building has occurred

- AB-203 Main Steam Line High Radiation actions have been completed

- All control rods are full in

- Fuel cladding damage has occurred

- Release rates are above General Emergency levels

- Reactor level is -60 inches and rising slowly

- Reactor pressure is 100 psig Why is an Emergency Depressurization required?

To ensure primary containment integrity To allow low pressure ECCS to inject R] To reduce the release rates

[i To provide core steam cooling e LCognitive Level _Coprehension Hope Creek ate03/12/2002 Tier: Emergency and Abnormal Plant Evolutions 1295038K102 295038 High Off-Site Release Rate Record Number] 47 EKi. , Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE:

EK1i.02 ;Protection of the general public 4.2 4.4

-of 104 Bases for step RR-6 states an ED is required if release rates are above GE levels to reduce EOP Ane the radioactivity release rate. RX pressure is already low enough for low press ECCS to inject. The primary containment is already somehow bypassed. The ED is not driven by adequate core cooling requirements. The RPV water level is above TAF.

Reference Title EOP 103/4 step RR-6 through 8 Learning Objectives 000127E006 (R) Given any step in the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by the step.

[Material Requiredl for Examination EOP Flowcharts without entry conditions IQuetion SouHI Facility Exam Bank uestion Modification Methd Significantly Modified Iauestion Source Comments: I Vision Bank QI3D# Q56164 Saturday, March 23, 2002 10:47:36 AM Page 49 of 139

Given the following:

- A large break LOCA has occurred inside the Drywell

- Multiple equipment failures occurred

- Drywell pressure is 15 psig

- Steam cooling was required until water level was restored above TAF with Fire Water

- The Containment H2/02 Analyzers were placed in service

- The High Hydrogen alarms are clear Which one of the following actions is required lAW EOP-1 02?

vent the Drywell because Hydrogen c n cent ration is above 2%

V

[i Exit EOP-1 02 and enter SAG because Hydrogen concentration is above 2%

vent the Suppression Chamber becaus e Hy drogen concentration is below 2%

VI WPlace the Hydrogen Recombiners in service because Hydrogen concentration is below 2%

Answe b-E-am LevelB lCognitive Levl [ Creek [ 03/12/2002 Tir:[ Emergency and Abnormal Plant Evolutions .... u. SR- Grou 500000K303 500000 Hgh Containment Hydrogen Concentration Record Number 48 EK3. Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

EK3.03 Operation of hydrogen and oxygen recombiners 3.0 3.5 Explanation of High H2 alarms come in at 2% Hydrogen. Since the H2 concentration is above 2%, EOP-102 step Anwe PC/H1 directs exit from EOP-1 02 and enter SAG Reference Title

EOP-102 step PC/Hi Learning Objectives 00126AE004 Recall the reasons why the following are used for determining the entry condition and / or subsequent actions lAW the Primary Containment Control - Drywell Lesson Plan.

a. Drywell Pressure

b. Average Drywell Temperature

c. H2 and 02 concentrations in the drywell Material Required for Examination EOP Flowcharts without entry conditions l ore Nwon Questio Modification Method:

Saturday, March 23, 2002 10:47:36 AM - Page 50 -of 139

During a fire in the Turbine Building, a Fire Department Liaison is assigned by the Operations Superintendent.

Who, by title, can be assigned this role and what is their duty?

El Communicator #1. Advises the Fire Department on how to mitigate the fire.

Communicator S* #2. Advises the Operations Superintendent on what equipment needs to be removed from service.

F1 Shift Technical Advisor. Advises the Fire Department on how to mitigate the fire.

EW Work Control Supervisor. Advises the Operations Superintendent on what equipment needs to be removed from service.

d E R- Cognitive Level Memory [FHope Creek xa] a 03/12/2002

,te:

Tier: ,Emergency and Abnormal Plant Evolutions R o 2 G 2 600000G425 600000 :Plant Fire On Site Record Number 49 2.4 Emergency Procedures and Plan 2.4.25 Knowledge of fire protection procedures. 2.9 3.41 fpnatioef The Work Control Supervisor or a qualified Equipment operator with no other emergency responsibilities, shall function as the station fire brigade liaison. The liaison shall make recommendations to the OS what equipment needs to be removed from service to mitigate the fire and/or stabilize the plant.

Reference Title HC.FP-EO.ZZ-0001 note 2.13 NC.NA-AP.ZZ-0005 Attachment 9 Note 3 Learning Objectives 000113E01 1 a. Summarize the responsibilities of the following personnel:

Operations Superintendent Control Room Supervisor/ Field Supervisor Shift Technical Advisor Licensed Operators [RO/PO]

00O-O01 1-3-E0 a. D-e-ter-min-e t-h-e-f-oll-owing-:------

The level of licensing required for the OS, CRS, and RO/PO.

Minimum shift manning requirements for all plant conditions.

Normal shift staffing levels.

When a person can serve a dual role as CRS/STA or OS/STA IMaterial Required for Examination I uestion Modification Method:I Saturday, March 23, 2002 10:47:36 AM Page 51 of 139

Given the following:

- A plant startup is in progress.

- Reactor Power on range 4 of IRMs

- Reactor Level at + 46 inches

- Reactor Pressure at 0 psig

- Reactor Temperature at 180°F The operating Control Rod Drive Pump trips. The Control Room Operator attempted to start the standby CRD Pump and the pump failed to start. Control Rod movement has been suspended.

Which one of the following describes the response of reactor water level and why?

(ASSUME NO OPERATOR ACTION)

Reactor Water level will:

na- rise due to the reactor being at the point.of adding heat..

Sremain stable due to water expansion from heating overcoming any losses to ambient.

Sremain stable due to water expansion from heating overcoming any losses to RWCU.

Wdrop due to RWCU rejecting water for level control.

d Ex-a ICognitiveLev:el Commprehension Fac I[ity Hope creek 03/12/2002 Tier: Plant Systems R 1 sRO G F.r... 2 201001A305 201001 Control Rod Drive Hydraulic System R-e c'or dNumnbe r 50

ýA3. Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including:

A3.05 Reactor water level 2.8 2.8 Elaof RWCU is normally balanced to reject the 69 gpm makeup from CRD. Without the CRD pump running, Answer RWCU is rejecting at approximately the same rate. RPV level will lower.

Reference Title HC.OP-IO.ZZ-0003 Learning Objectives 000006E028 From memory, determine why a method of reactor water level control must be available prior to placing the CRDH System in service including the preferred method of level control, lAW the Lesson Plan.

[Material Required for Examinatio I

[uestionSou INPO Exam Bank uestion Modification Method: Editorially Modified Question source comments: 7 INPO EXAM BANK QID# 16316 Grand Gulf 1 04/01/2000 Saturday, March 23, 2002 10:47:36 AM Page 52 of 139

Given the following:

- The plant was manually scrammed due to prolonged loss of CRD

- A CRD pump has been restarted

- All surveillances are current

- All equipment is operable Which one of the following PREVENTS control rod withdrawals?

Rod Worth Minimizer insert and withdraw errors will result in a control rod withdrawal block signal ReactorSTheMode Switch in "Shutdown" inserts a continuous control rod withdrawal block signal 1] The Reactor Mode Switch in "Shutdown" maintains a scram signal on RPS until reset by the operator I] Rod Block Monitor "Downscale" inserts a control rod withdrawal signal until bypassed A rb R-ExaIevelR cognitive Level M.....

.ci.i...HopecreeE.. 03/12/2002 I i r:

Plant Systems . .. 2 201002G421 201002 Reactor Manual Control System Record Number 51 2.4 Emergency Procedures and Plan 2.4.21 iKnowledge of the parameters and logic used to assess the status of safety functions including: 3.7,4.3

1. Reactivity control

ý2.Core cooling and heat removal 13.Reactor coolant system integrity 4.Containment conditions

5. Radioactivity release control.

Eplanation of MS in Refuel, one rod withdrawn, second rod selected causes a rod out motion block.

F- Reference Title HC.OP-SO.KE-0001 Learning Objectives 000007E008 (R) From memory, explain the interrelationships between the Reactor Manual Control System and the following, lAW the Reactor Manual Control System Lesson Plan:

a. Rod Worth Minimizer

b. Neutron Monitoring System

c. Rod Block Monitor System

d. Mode Switch

e. Refueling System

1) Refueling Bridge

2) Refueling Grapple/Hoists

f. 120 VAC Uninterruptible Power Supply Required for Examination .Material l ~'urce: INPO Exam Bank Q.uestionModification Method: Editorially Modified Q t orce Comments: INPO BANK Q-ID#1 599 Peach Bottom 2 - 09/19/1997 Saturday, March 23, 2002 1-0:47:37 AM Page 53 of 139

Saturday, March 23, 2002 10:47:37 AM Page 54 of 139 The Control Room operator is moving control rods when a ROD DRIFT annunciator is received.

Which one of the following controls caused this annunciator?

SAn odd reed switch is passed while settling from Insert of the control rod one notch using the INSERT PB El An even reed switch is passed while settling from Withdrawal of the control rod one notch using the WITHDRAW PB El -An odd reed switch is passed while settling from Insert of the control rod two notches using the CONTINUOUS INSERT PB WAn even reed switch is passed while seitting from WithdrawaI of the control rod two notches using the CONTINUOUS WITHDRAW PB EmLeeB ConteLeeMeoyFcltHope Creek Examn Date:1 03/12/2002 Tier: Plant Systems RO*roup 1 SRO-G- u 2 201002K408 201002.. Reactor Manual Control System Reord Number 25 K4. Knowledge of REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.08 ýContinuous In rod insertion 3.2 3.21 E of Rod settle function is bypassed during Continuous Insert. A rod settling from continuous insert does not Answer . have a rod motion command. Any rod motion detected from odd or even reed switches without a rod motion command causes the Rod Drift alarm.

Reference Title HC.OP-AR.ZZ-00 11 Attachment E3 Learnilng Objectives 000007E003 (R) Given a labeled diagram/drawing of, or access to, the Reactor Manual Control System controls/indication bezel, summarize the following lAW the Reactor Manual Control System Lesson Plan:

a. The function of each indicator.

b. The condition which will cause the indicator to light or extinguish.

c. The effect of each control on the Reactor Manual Control System.

d. The conditions or permissives required for the control switches to perform their intended function.

IMaterial Required for Examitation Bank IQuestion Modification Method: iEditorially Modified INPO BANK QID# 17255 Columbia Gen Sta 03/08/1999 Saturday, March 23, 2002 10:47:37 AM Page 55 of 139

Control rod 30-31 is being inserted from position 12 to position 08.

The RO notes that during rod motion the following occur:

- Control Rod 30-31 position indicates "XX" on the 4-Rod display

- Control Rod 30-31 position indicates "XX" on CRIDS

- RPIS Status DATA FAULT light on 10C651 is lit WHICH ONE of the following describes the status of rod 30-31?

EIReed switch has failed IIScrammed WUncoupled S Disarmed Anse a Exam-Level B Cognitive Le.el Memory Facility Hope Creek E1x Date: 03/12/2002 Tier: Plant Systems .. 22....IRO ro 1 3 201003K405 201003 Control Rod and Drive Mechanism Record Number 53 K4. I Knowledge of CONTROL ROD AND DRIVE MECHANISM design feature(s) and/or interlocks which provide for the following:

K4.05 !Rod position indication 3.2 3.33 Eof Control rod indication with a failed open reed switch is XX.

Reference Title HC.OP-SO.SF-0001 Attachment 1 page 31 Learning Objectives 000007E004 (R) Given plant conditions and a drawing of the controls, instrumentation and/or alarms located in the Control Room, assess the status of the Reactor Manual Control System lAW the Reactor Manual Control System Lesson Plan.

Material Required for Examinat.on 1 INPO Exam Bank Qestion Modification Method: Editorially Modified IQuestion Source Comments: I INPO EXAM BANK QID# 12303 Limerick 01/20/1998 Saturday, March 23, 2002 10:47:37 AM Page 56 of 139

Given the following:

- An entire startup was performed with an inoperable RWM on 1/12/2001

- With the RWM still inoperable, the reactor scrams on 12/25/2001

- Today's date is 1/6/2002 What RWM requirements must be met to withdraw control rods per Technical Specifications?

Ei Startup is NOT allowed until 01/12/2002 IThe RWM must be restored to operability within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of withdrawal of the first rod SStartup may commence as soon as one licensed operator and one technically qualified member of the technical staff are present at the reactor control console until the first twelve control rods are fully withdrawn HiStartup may commence as soon as one licensed operator and one technically qualified member of the technical staff are present at the reactor control console until power is above 10%

rd xm LeCognitive evel ComprehensionFacility _HopeCreek WExamDate:

O3/12/2002 Tier: Plant Systems . .... R.. 2 2 201006G112 201006 Rod Worth Minimizer System (RWM _(Plant Spepefic) Record Number 54 2.1 Conduct of Operations 2.1.12 Ability to apply technical specifications for a system. .2.9 i4.06

[Explantonof TS 3.1.4.1 is limited by calendar year not rolling year. Startup may commence since this is the first

[Answer startup of the new year. Once >10 percent power, the LCO is no longer applicable.

Ii Reference Title HC TS 3.1.4.1 10CFR55.43(2)

Learning Objectives 000009E009 (R) Given plant conditions and access to Technical Specifications:

a. Select those sections, which are applicable to the Rod Worth Minimizer lAW HCGS Technical Specifications.

b. Evaluate Rod Worth Minimizer operability and determine required actions based upon system inoperability lAW HCGS Technical Specifications.

c. Explain the bases for those Technical Specifications associated with the Rod Worth Minimizer lAW HCGS Technical Specifications. (SRO ONLY).

[material Required for Examinato lQuestion Sourc. ] Facility Exam Bank [Q-uestion Modification Met: .Significantly Modified Question Source Comments: Vision Bank QID# Q56372 Saturday, March 23, 2002 10:47:37 AM Page 57 of 139

Given the following:

- A reactor startup is in progress

- Reactor power is 42% after completion of RWM Group Step 500

- The total steam flow signal output from Digital Feed fails to the equivalent of 15% power Which one of the following describes how control rod motion is effected by the Rod Worth Minimizer (RWM)?

Fa The RWM willNOT allow any control, riod..insertion or withdrawal.

K1 The RWM will allow all normal control rod motion until actual power is Iess than the LPSP.

EI The RWM will apply rod blocks in accordance with the loaded rod sequence.

W The RWM will allow continued control rod movement only by single notch increments.

Ansawe r L Cognitive Level Comprehension at pe Creek E 03/12/2002 Tier: Plant Systems .. 2....

2§ROGo 2 201006K301<30 201006 Rod Worth Minimizer System (RWM)(lant Specific) Number .ecord 55 K3. Knowledge of the effect that a loss or malfunction of the ROD WORTH MINIMIZER SYSTEM (RWM) will have on following:

K3.01 Reactor manual control system: P-Spec(Not-BWR6) 3.2 3.5.

Explanatonof "The RWM will not allow control rod withdrawals if any control rod is withdrawn past its withdraw limit."

LAnwer The following distractors are incorrect as follows:

"The RWM will allow all normal control rod motion until actual power is less than the LPSP."

Total steam flow from Digital Feed is the signal used by the RWM to determine the LPSP, not actual power.

"The RWM will NOT allow any control rod insertion or withdrawal."

The RWM will allow movements as long as they meet the required sequence programmed into the computer.

"The RWM will allow continued control rod movement only by single notch increments."

The RWM will allow movements as long as they meet the required sequence programmed into the computer.

Reference Title

HC.OP-SO.SF-0003 Learning Objectives 000009E004 (R) Given plant conditions, summarize the interrelationship(s) between the Rod Worth Minimizer and any of the following lAW the RWM Lesson Plan.

a. Rod Position Information System (RPIS)

b. Reactor Manual Control System (RMCS)

c. Feedwater Level Control System

d. Process Computer

e. 120 VAC Material Required for Exainatioa

[eton Source] Facility Exam Bank Question Modification Method: Direct From Source Saturday, March 23, 2002 10:47:37 AM Page 58 of 139

IQuestion Source Comments: Vision Bank QID# Q53252 Saturday, March 23, 2002 10:47:38 AM Page 59 of 139

Given the following:

- The reactor is operating at 80% power

- Core flow was 68.0 Mlbm/hour

- The "A" Recirculation Pump tripped

- Reactor power stabilized at 57%

- Total core flow stabilized at 43.0 Mlbm/hour

- No operator actions have been taken Based on plant conditions, which one of the following operator actions are required?

(AB-300 Attachment-1 is attached)

FI Reduce power by single rod scrams EI Reduce power by lowering recircuIation flow El Raise flow by restarting the "A" Recir culatio n Pu mp W Raise flow by raising the speed of the "B" RecircuIation Pump Answr d xam evelR ICognitive Level I Comprehensin Fclt oeCek Ea ae 311 2002 Tier: Plant Systems ROGroup 2 SRO .ro 2 . 202001A40, 202001 - Recirculation System Record Number 56 A4. Ability to manually operate and/or monitor in the control room:

A4.04 System flow 3.7 3.7 Explanation of Options are to raise recirc flow to exit or insert rods. Inserting rods by single rod scrams is not allowed.

Reference Title

ýHC.OP-AB.ZZ-0300 Learning Objectives OAB300EO05 (R) Interpret and apply charts, graphs and tables contained within the Reactor Power Oscillations, Abnormal Operating Procedure.

IMaterial Required forExamination I HC.OP-AB.ZZ-0300 Attachment 1 INPO Exam Bank Q6ue stion Modification Method: I significantly Modified niments. IINPO BANK QID# 15785 Vermont Yankee 01/22/1999 Saturday, March 23, 2002 10:47:38 AM Page 60 of 139

Given the following:

- The plant was operating at 100 percent power

- RWCU pump "A" is C/T for maintenance

- The plant scrammed following a dual recirc pump trip

- RPV level is stable at +30 inches

- RPV Pressure is stable at 920 psig Based on plant conditions, which one of the following is required?

I1Trip CRD pu mps Trip RWCU pumps R increase CRD cooling water flow E Reduce RWCU flow from the Recirc Loops d am Lel B Application y Hope Creek Dte 03/12/2002 Tier: Plant Systems .roUP R. 2 S rop 2 202001 K307 202001 Recirculation System Record Number 57 K3. Knowledge of the effect that a loss or malfunction of the RECIRCULATION SYSTEM will have on following:

K3.07 :Vessel bottom head drain temperature 2.9 2.9 of If 1 RWCU pump is running, maximize bottom head drain flow to prevent thermal stratification. This is accomplished by throttling down flow to the recirc loops using valves HV-F102 Reference Title HC.OP-AB.ZZ-0000 Step S-18 HC.OP-SO.BG-0001 Step 5.5.3.B Learning Objectives 000123E004 (R) Given any step of the procedure, determine the reason for performance of that step and/or evaluate expected system response to control manipulations prescribed by that step.

000021E013 Given any system that interrelates with the RWCU System, explain the purpose of that interface lAW the RWCU System Lesson Plan.

Material Required for Examinaton IQuestion Source: INew lQuestion Modification Method:

Question Source Comments:

Saturday, March 23, 2002 10:47:38 AM Page 61 of 139

Given the following:

- The plant is operating at 100 percent power

- Total Feedwater Flow signal from Digital Feed is lost Which one of the following describes the effect of the loss on the plant?

t] Recirc Pump Scoop Tube Lockup

[J Recirc Pump Speed Limiter Full run bac k 1] Reactor Scram on Low R PV level e El Reactor Feed Pumps Speed Limited to 2500 RPM Anwr xm e Rni Co rehension F tHope Creek E 03/1212002 Tier: Plant Systems 1 G EROJGr 1 202002K604 202002 Recirculation Flow Control System Record Number 58 K6... Knowledge of the effect that a loss or malfunction of the following wilI have on the RECIRCULATION FLOW CONTROL SYSTEM:

K6.04 Feedwater flow inputs: BWR-3, 4, 5,6 3.5 3.5 Exlntonof Loss of FW flow signal causes RR runback full due to FW flow <20%

Reference Title HC.OP-SO.BB-0002 HC.OP-SO.AE-0001 Learning Objectives 000019E013 (R) From memory, explain the purpose of each recirc pump runback and list signals which will generate each runback lAW the Lesson Plan.

Material Required for Examinati*on*

QusinSource:- New. Question Modification Method:

Saturday, March 23, 2002 10:47:38 AM Page 62 of 139

Given the following:

- A LOCA concurrent with a partial Station Blackout has occurred

"- "A" LPCI is being injected into the RPV

- Reactor Pressure is steady at 100 psig

- Reactor Bldg Temperature is steady at 105 0 F

- Drywell Temperature is increasing slowly at 285°F

- Fuel Zone indicators are reading -168 inches and steady Based on the above current conditions, adequate core cooling is ..........

I] assured, since actual RPV ievel is -150" I] assured, since actual RPV level is -159" E NOT assured, since actual RPV level is -170*

SNOT assured, since actual IeveI is -173" w a B-b ecognitive Levei] Application FatyHope Creek 03/12/2002 0ExamDa:

Tier: Plant Systems___ou SOGru 20300OA407 203000 RHR/LPCI: Injection Mode (Plant ific Record Number 59 A4. Ability to manually operate and/or monitor in the control room:

A4.07 Reactor water level 1; 4.5 4.5

[xplaationof Uncompensated level is -168".

Ase -RB Temp Correction: 1050 - 75' = 30'

-DW Temp Correction: 2850 - 1350 = 1500

-TAF curves shift upwards 6" for a 300 F increase in RX Bldg temp

-TAF curves shift down 3" for a 150°F increase in Drywell Temp

-The resulting TAF curve at 100 psig is shifted upwards 3".

-The TAF Curve at 100 psig is -173".

-Shifting upwards 3" places the Curve at -170".

-Indicated level of -168" is 2" above the TAF Curve.

Actual compensated level is therefore 2" above TAF or -159" and therefore Adequate Core Cooling is assured.

Reference Title

Station Aid OPA-92-039 Learning Objectives 00124AE006 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulation prescribed by that step.

Material Required for Examinaton Station Aid OPA-92-039 lu7 nour: Facility Exam Bank Question Modification Method: Significantly Modified IQuestion Source Comments: I vision Bank QID # Q56158 Saturday, March 23, 2002 10:47:38 AM Page 63 of 139

Given the following:

- "B" RHR pump is running in Shutdown Cooling (SDC)

- I&C error initiates "B" LPCI Initiation Logic on High Drywell Pressure signal Which one of the following describes the "B" RHR Pump, SDC Discharge Valve F015B, and LPCI Injection Valve F017B response?

" "B" RHR pump trips, F015B closes and F017B opens H "B" RHR pump trips, FOI5B closes and Foi7B remains closed 13"B" RHR pump remains running, F015B remains open and F017B opens

" "B" RHR pump remains running, F015B remains open and F017B remains closed nsrB Cognitive Level Compre hension Fcit Hope Creek x Date: 03/12/2002 Tier: Plant Systems F0GROp 1 S---u 1 203000K114 203000 RHR/LPCI: Injection Mode Plant Specific) Record Number. 60 K1. Knowledge of the physical connections and/or cause- effect relationships between RHR/LPCI: INJECTION MODE and the following:

K1.14 iShutdown cooling system: Plant-Specific 3.6 3.7 Expianton of F01 5B closes on Low RPV level or Hi RPV pressure. Neither is present so F015B stays open. F017B

.opens Answer on High Drywell Pressure or Low RPV pressure. Since High DW pressure initiation is given, F017B opens. There is no signal present to trip LPCI pump B so B RHR pump remains running.

Reference Title

HC.OP-SO.BC-0001

HC.OP-SO.BC-0002 Learning Objectives

--00-0028E0,1 4-.. Given a copy/mimic of the RHR System controls on I0C650A, predict proper RHR System response during the LPCI mode of operation to include the following, lAW the RHR System Lesson Plan:

a. From memory, state the two automatic initiation signals and setpoints for LPCI initiation, lAW the RHR System Lesson Plan.

b. Determine the pump starting sequence for the LPCI pumps with and without off-site power available, lAW the RHR System Lesson Plan.

c. Determine the actions required to override the LPCI initiation and stop the LPCI pump, lAW the RHR System Lesson Plan.

d. Determine the actions required to override the LPCI initiation and close the LPCI injection valve HV-F017, lAW the RHR System Lesson Plan.

e. Determine the operator actions required to initiate suppression pool cooling during LPCI mode of operation, lAW the RHR System Lesson Plan.

f. Determine the operator actions required to initiate Torus/containment spray during LPCI mode of operation, lAW the RHR System Lesson Plan.

IMaterial Required for Examnination

[uston Sour~ New Qutestion Modificationi Method:

QustonSorce Comments:I Saturday, March 23, 2002 10:47:38 AM Page 64 of 139

Given the following:

- The reactor is operating in STARTUP

- RCS temp is 190°F

- RWCU blowdown operation to the Liquid Radwaste System at 60 gpm

- The operator fully opens Blowdown Line Restricting Orifice Bypass Valve (HV-F031)

Which one of the following describes the operational effect of this high bypass flow and how does the operator adjust for the change?

I The Regenerative Heat Exchanger (RHX) RWCU outlet temperature will lower. Lower RA* S flow lAW OP-SO.BG-0001 I The Regenerative Heat Exchanger (RHX) RWCU outlet temperature will lower. Raise RACS flow lAW OP-SO.BG-0001 The Non-Regenerative Heat Exchanger (NRHX) RACS outlet temperature will rise. Lower Ti RACS flow lAW OP-SO.BG-0001 IIThe Non-Regenerative Heat Exchanger (NRHX) RACS outlet temperature will rise. Raise RACS flow lAW OP-SO.BG-0001 Answer d Exýam Lvei B Cognitive Level Comprehension i Hope Creek Exam..ate 03/12/2002 Tier: Plant Systems ROiroul 2 2 204000A214 204000 Reactor Water Cleanup SystemRe.... Number 61 A2. Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A214- System high temperature 3.2 3.22 of With increased blowdown flow, less RWCU return flow through the Regen HX causes temp to increase at the RWCU inlet to the NRHX. For the same RACS cooling flow, RACS outlet temp will increase. Step 5.9.4 throttles open 1-ED-V035 to increase RACS flow.

Reference Title

!HC.OP-SO.BG-0001 step 5.9.4 Learning Objectives 000021 E012 (R) From memory, summarize the effects of RWCU System blowdown operation on the RHX and NRHX's lAW the RWCU "System Lesson Plan.

Material Required for Examination QusinSuE Facility Exam Bank Question Modification Method: Sgnificantly Modified QustonSorce Comments: Vision Bank QID# Q56399 Saturday, March 23, 2002 10:47:39 AM Page 65 of 139

Given the following:

- The High Pressure Coolant Injection System (HPCI) is operating in Pressure Control alignment

- The HPCI flow controller is in "Automatic"

- HPCI turbine speed is 2450 rpm Which one of the following describes the response of HPCI turbine speed and system flow if the operator throttles the HPCI Test Bypass To CST Isolation Valve (FO08) in the "open" direction for the given conditions?

(Compare the conditions after they stabilize to before the valve was throttled.)

Fl -- HPCI turbine speed lowers

-- System flow returns to original value I1 -- HPCI turbine speed lowers

-- System flow goes down El -- HPCI turbine speed raises

-- System flow returns to original value El -- HPCI turbine speed raises

-- System flow goes up AnswCeromprehension F Hope Creek Exam Date: 03/12/2002 Tier: Plant Systems RO Gru 1 G 1 206000A106 206000 High Pressure Coolant Injection System Record Number 62 Al..... Ability to predict and/or monitor changes in parameters associated with operating the HIGH PRESSURE COOLANT INJECTION SYSTEM (HPCI) controls including:

A1.06 System flow: BWR-2, 3,4 3.8 3.7 fExplanation7f With flow controller in Auto opening, the CST valve reduces the resistance to flow (flow increases for tAnswer -that turbine speed), flow controller reduces turbine speed to return to flow setpoint.

Reference Title HC.OP-SO.1BJ-0001 Learning Objectives 000026E016 (R) Given plant conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, assess "thestatus of the HPCi System by evaluation of the controls/instrumentation/alarms, lAW the HPCI System Lesson Plan.

Material Required for Examinatio.

Question SouIc New "Ques..on Modification Method:

Qei Souirce Comments:

Saturday, March 23, 2002 10:47:39 AM !L _Page 66 of 139

Given the following:

- I&C is performing testing on HPCI TURBINE EXHAUST DIAPHRAGM RUPTURE transmitter PISH-N655A

- A ZERO psig signal is set on the calibration device

- The following Alarms/ Status lights from the testing are received in the Control Room:

- HPCI SYSTEM OUT OF SERVICE - LIT

- IN TEST STATUS on Logic Channel "A" - LIT

- TRIP UNIT IN CAL OR GROSS FAIL on Logic Channel "A" - LIT

- HPCI TURBINE EXHAUST DIAPHRAGM RUPTURED - Extinguished With this configuration, how will HPCI respond to an actual HPCI Initiation with a subsequent diaphragm rupture?

Fal "A" channel isolation vaIves only wilI isolate. HPCI Turbine will NOT trip.

FI "C" channel isolation valves only will isolate. HPCI Turbine will trip.

Ri "A" and "C" channel isolation valves will isolate. HPCI Turbine will trip.

EW "A" and "C" channel isolation valves will isolate. HPCI Turbine will NOT trip.

Answer b Exam Levl R lcognitive evel Application a Hope Creek Exam Date: 03/12/2002 "Tier: plant Systems RO. Group 1SRO Group I 1 206000A307 206000 High Pressure Coolant Injection System Record Number 63 A3. IAbility to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM (HPCI) including:

A3.07 Lights and alarms: BWR-2, 3, 4 3.9 3.8

[xplanationot "C" Channel transmitters PISH -655C & G will still respond properly to a valid diaphragm rupture. HPCI w Turbine will trip. The Channel "A" Logic will not trip due to the 655A transmitter is in test with a zero psig signal. 2 of 2 transmitters are required per logic channel to actuate.

Reference Title HC.OP-SO. BJ-0001 Learning Objectives 000026E016 (R) Given plant conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, assess the status of the HPCI System by evaluation of the controls/instrumentation/alarms, lAW the HPCI System Lesson Plan.

Material Required for Examination J-0650 of HPCI status lights and overhead alarm windows NewQuestion

............. .ification Method .

Iuestion Source Comments: INPO BANK QID# 17084 Susquehanna 1 09/30/1999 Concept Used Saturday, March 23, 2002 10:47:39 AM Page 67 of 139

Given the following:

- The plant is operating at 100% reactor power

- A small instrument line break LOCA occurs

- Drywell pressure is 2.5 psig increasing

- RPV water level reaches -30 inches and is rising

- Drywell pressure trip unit N694F to Core Spray (1 of 2) has failed to trip Which one of the following describes the response of the Emergency Diesel Generators?

F All Emergency Diesel Generato rs start and load onto their respective busses IA, C, & D Emergency Diesel Generators start but DO NOT load onto their respective busses A, B, & D Emergency Diesel Generators start and load onto their respective busses All Emergency Diesel Generators start but DO NOT load onto their respective busses Aseb Exam Level B I LComprehension i Hope Creek 03/12/2002 Tier, Plant Systems .. . . R.. 1r 2091001Kl10 209001 Low Pressure Core Sp raySystem ,Record Number 64 KI. Knowledge of the physical connections and/or cause- effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following:

K1.i0 Emergency generator 3.7 3.8

[Exilaation of] Both Hi DW Pressure trip units N694B & F must trip to cause an initiation signal to B Core Spray Loop Anse Jand its respective EDG. RPV level did not reach -129" necessary to cause Level 1 trip. The EDG does not load because LOP is not present.

IOReference Title HC. OP-SO. BE-QO0l HC.OP-SO.KJ-0001 Learning Objectives 000027E009 (R) From memory, summarize/identify the two Core Spray System initiation signals which will also cause an automatic start of the emergency diesel generators, lAW the Core Spray System Lesson Plan.

Material Required for Examination

[etonSource I Facility Exam Bank.Question Modification Method: Direct From Source Question Source Comments: I Vision Bank QID# Q56486 Saturday, March 23, 2002 10:47:39 AM Page 68 of 139

Given the following:

- A Loss of Offsite Power occurs followed by a LOCA

- "B" EDG fails to start

- Drywell pressure is 5.7 psig

- Reactor pressure is 440 psig decreasing Which one of the following describes the effect on the "D" Core Spray Pump and Injection Valve BE-HV-F005B?

"Ii "D"Core Spray Pump will NOT start but F005B opens

"[i"D" Core Spray Pump will NOT start and F005B will NOT open "Ii "D" Core Spray Pump starts but F005B will NOT open "I"D" Core Spray Pump starts and Foo5B opens e ....

c .. ExamLev`e B- [Cognitive Level CHope Creek - DI; 03/12/2002 Tier: Plant Systems ..... . roup SRG p 1 209001 K202 209001 Low Pressure Core Spray System Record Numb .. 65 K2. Knowledge of electrical power supplies to the following:

K2.02 Valve power 2.5 2.7 E of Correct answer based on F005B is a B channel valve which will not open in response to the LOCA.

Reference Title HC.OP-SO.BE-0001 P&ID M-52 Learning Objectives 000027E004 (R) From memory, summarize/identify the sequence of events following receipt of an automatic or manual Core Spray System initiation signal, lAW the Core Spray System Lesson Plan.

000027E005 (R) For a given set of plant conditions, from memory, summarize/identify the interrelationship between the Core Spray System and any of the following, lAW the Core Spray System Lesson Plan:

a. Residual Heat Removal (RHR) System

b. Torus Compartment

c. 4160 VAC Class 1E Distribution System

d. 480 VAC Class 1E Distribution System

e. 125 VDC Class 1E Distribution System

f. Nuclear Boiler

g. Liquid Radwaste System

h. Condensate Storage and Transfer System

i. Primary Containment Instrument Gas (PCIG) System

j. High Pressure Coolant Injection (HPCI) System

k. Condensate Storage Tank I. Automatic Depressurization System (ADS)

m. Emergency Diesel Generators (EDGs)

n. Nuclear Boiler Instrumentation System

o. Standby Liquid Control (SLC) System 000027E012 (R) Given a labeled diagram/drawing of the Core Spray System controls/indication bezel, lAW the Core Spray System Lesson Plan:

a. Explain the function of each indicator.

b. Assess plant conditions that will cause the indicators to light or extinguish.

c. Determine the effect of each control switch on the Core Spray System.

d. Assess plant conditions or permissives required for the control switches to perform their intended functions.

Material Required for Examination Saturday, March 23, 2002 10:47:39 AM Page 69 of 139

Facility Exam Bank jQuestion Modfction Method: Editorially Modified m I Vision Bank QiD# Q56228 Saturday, March 23, 2002 10:47:39 AM Page 70 of 139

Given the following:

- A plant shutdown is in progress lAW HC.OP-IO.ZZ-0004

- Both Standby Liquid Control pumps are inoperable

- A scram condition is reached and the reactor fails to scram When will the SLC/RRCS INITIATION FAILURE Overhead Alarm occur?

(Assume RPV level stabilizes at -50 inches and reactor power remains at 8%.)

[i When the RRCS POTENTIAL ATWS alarm occurs

[i When the RRCS CONFIRMED ATWS alarm occurs W30 seconds after the RRCS POTENTIAL ATWS alarm 30 seconds after the RRCS CONFIRMED ATWS alarm d Ea Le R cognitiveLevel Memory Date 03/12/2002 Tier: ýPlant Systems Ro roup I R G up -- 211000G410 21 000 -Standby Liquid Control System .cordNu.m.b 66 2.4 Emergency Procedures and Plan 2.4.10 Knowledge of annunciator response procedures. 3.0 13.1 Eo 30 seconds after the RRCS CONFIRMED ATWS alarm occurs.-Correct- lAW OHA C1-F1 Reference Title

-HC.OP-AR.ZZ-00_08 Attachment F1 Learning Objectives 000024E005 (R) Given a set of conditions and a drawing of the controls, instrumentation, and/or alarms located in the main control room, determine the status of the Redundant Reactivity Control System by evaluation of the controls/instrumentation/alarms, lAW the Lesson Plan.

Material Required for Examinati*on* *

[eion Souwe Facility Exam Bank . Q..

uestion Modification Method: Editorially Modified source Comments:

.Q.uestion

.7 Visi-o-n.Ba.nkQ.D..#Q56844 Saturday, March 23, 2002 10:47:39 AM Page 71 of 139

Given the following:

- Reactor power is 90%

- HC.OP-IS.BH-0001, Standby Liquid Control Pump AP208 In-service Test, will be performed to check flow rates during power operation.

How is the automatic Reactor Water Cleanup system isolation avoided during this test?

a. The Standby Liquid Control pump is started with the local control switch.

IThe RWCU system must be shutdown and the appropriate isolation valves closed.

IThe breakers for the appropriate RWCU-isolation valves must be opened.

IiThe fuses for the SLC squib valve firing circuitry must be removed.

[swj a Fxam Level RCognitive Level Memory [ Hope

°lii]Creek 03/12/2002 Tier: Pl,an-tSystems RO ýroup J1 S§RO_G~ro~u~p 2 1-000K1I05-211000 Standby Liquid Control System -r- Re- e.. .... 67 K1. Knowledge of the physical Connections and/or cause- effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:

K1.05 RWCU 3.4'3.6 Starting the Standby Liquid Control pump with the local control switch bypasses the RWCU isolation signal.

Reference Title HC.OP-SO.BH-0001 HC.OP-SO.BG-0001 Learning Objectives 000023E004 Given plant conditions, summarize/identify the interrelationship between the following Systems and the Standby Liquid Control System I.A.W. the Lesson Plan.

a. 480V 1E AC Distribution

b. Core Spray

c. Service Compressed Air

d. Demineralized Water Makeup, Storage & Transfer System

e. Redundant Reactivity Control System

f. Reactor Water Cleanup System

g. Standby Diesel Generator

h. Nuclear Steam Supply Shutoff System

i. Heat Trace IMaterial Required for Examination lE-stonSour Facility Exam Bank Question Modification Method: I Editorially Modified Question rce Comments: Vision Bank QID# Q56772 Saturday, March 23, 2002 10:47:40 AM Page 72 of 139

Which one of the following describes when the Reactor Mode Switch Shutdown position scram may be bypassed?

7a When moving the mode switch from REFUEL to SHUTDOWN F -When moving the mode switch from SHUTDOWN to REFUEL F] When testing the "One Rod Out Interlock" M- When-a control rod blade is being uncoupled a

w Eýx v R Cognitiv Memory Facility Hope creek E.am Date " 03/12/2002 Tier:, Plant Systems .. Group 1 G 1 212000G123 212000 Reactor Protection System Record Number 68 2.1 Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant 3.9 4.0 operation.

,Expl **aiof The Reactor Mode Switch Shutdown position scram may be bypassed to move the MS from refuel to Aw Shutdown when all control rods are fully inserted or the reactor is defueled.

Reference Tite HC.OP-SO.SB-0001 Prereq 2.6.2 Learning Objectives 000022E004 (R) From memory, identify the parameters which initiate a Reactor Scram, list the scram initiation setpoints for each identified parameter, and determine when the parameter is bypassed, lAW the Lesson Plan.

IMaterial Required for Examination

[Queston SourcýI New- Question Modification Method:

Queston Source Comments:

Saturday, March 23, 2002 10:47:40 AM Page 73 of 139

With the plant operating at rated power, the power supply fuse to a backup scram valve fails creating an open in the supply circuit.

Which one of the following identifies the response of the associated backup scram valve and scram response due to this failure?

F1 _Valve repositions to trip position but NO scram occurs Valve CANNOT reposition but redundant valves can effect scram if an RPS trip occurs

-I

[ Valve CANNOT repositi-on-andN- scram can-occurevenif--a-n--P-Stri-poccurs---- _--_

P1 Valve repositions to trip position and a full scram occurs b cognitive Comprehension Hope creek Exam Date: 03/1212002 Tier: Plant Systems .rou. R. G 1 212000K502 212000 Reactor Protection SYstem Record Number 69 K5. Knowledge of the operational implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM:

K5.02. Specific logic arrangements 3.3: 3.4ý Explanation*of Backup Scram valves are normally de-energized, energize to operate solenoid valves therefore the valve w cannot reposition. In conjunction with the valve piping arrangement, the other valve will complete the scram function if a full scram signal is received.

Reference Title HC.OP-SO.SB-0001 Learning Objectives 000022E009 (R) Given plant conditions, evaluate the response of RPS to an electrical failure, lAW the Lesson Plan.

material Required for Examination u INPO Exam Bank Question Modification Method: Editorially Modified Question Source Comments: INPO EXAM BANK QID# 11769 Nine Mile Point 1 01/20/1998 Saturday, March 23, 2002 10:47:40 AM Page 74 of 139

A TIP System trace is being taken when an I&C Technician error causes actuation of the NSSSS Channel "A" manual isolation switch.

Which one of the following describes the TIP system response?

R The TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube.

El The TIP detectors will automatically withdraw to their "in-shield" position and the TIP Guide Tube Ball Valves automatically close.

WThe TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube.

WNo automatic actions occur when only one NSSSS ch anneI manual isolation switch is actuated.

Anwrb ExmLvlB Cogniti e Level Comprehension ai Hope Creek Exam Date: 03112/2002 Tier: i Plant Systems IRISO°I 3 3 215001K105 215001 Traversing In-Core Probe Record Nu-mbe-r 70 Ki.1 Knowledge of the physical connections and/or cause- effect relationships between TRAVERSING IN CORE PROBE and the following:

K1.05 Primary containment isolation system: (Not-BWR1) 3.3 3.4, JUSTIFICATION:

The TIP detectors not in the "in-shield" position will automatically withdraw to their "in-shield" position and the TIP Guide Tube Ball Valves automatically close. Correct The TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube.

Incorrect - the Shear Valves must be manually initiated.

The TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube.

Incorrect - the Ball Valve will not close with the cable inside the valve.

No automatic actions occur when only one NSSSS channel manual isolation switch is actuated.

Incorrect - manual initiation of NSSSS Channel "A" will cause isolation of affected systems, including TIP.

fReference Title HC.OP-SO.SM-0001 Table SM-017 Learning Objectives 000018E006 (R) From memory explain the response of the TIP System following the receipt of an isolation signal from the Nuclear Steam "Supply Shutoff System, lAW the Lesson Plan.

laterial Required for Examination u Source: Facility Exam Bank Question Modification Method: Direct From Source uestion Source Comments: I Vision Bank QID# Q5371 0 Saturday, March 23, 2002 10:47:40 AM Page 75 of 139

Given the following:

- The reactor is in Operational Condition 5

- The RPS shorting links are removed

- SRM "A" fails upscale Which one of the following describes the resulting automatic action?

SRod block only S1/2 scram RPS-A only SRod block and 1/2 scram RPS-A only WFull scram d Ea Level B Cognitive L Comprehension

've lHope Creek _tED 03/12/2002 Tier: Plant Systems R 1 ROGroup 1 215004A303 215004 Source Range Monitor (SRM) System Record Number 71 A3. 'Ability to monitor automatic operations of the SOURCE RANGE MONITOR (SRM) SYSTEM including:

A3.03 RPS status 3.6 3.5, Explantion Installation of the Shorting links enables the SRM Hi-Hi rps scrams and changes the coincidence to 1 of 18 taken once Reference Title

ýHC.OP-SO.SB-0001 Learning Objectives 000022E014 Given labeled diagrams/drawings of the RPS trip logics, explain the coincidence requirements necessary to generate a reactor scram.

Material Required for Examination Q INPO Exam Bank QQuestion Modificatio2 Method: Editorially Modified

[Question Source Comments: I NPO BANK QID# 17225 Columbia Gen Sta 03/08/1999 Saturday, March 23, 2002 10:47:40 AM Page 76 of 139

Given the following:

- An I&C Technician is in the middle of SRM "A" Channel Functional Test

- The next section of his procedure contains several discrepancies Which one of the following changes is PROHIBITED as an "On The Spot Change" to the procedure?

a. Increasing the trip setpoint tolerance to reduce nuisance alarms El Minor alterations to a step to clarify tha t step FIChanging a step which returns the "B" SRM Mode Switch to the original position WAdding a supervisory review signoff aCognitive Comprehension Cevel ope Creek Exa_ Dat 03/12/2002 Tier: Plant Systems ISRO Group 215004G206 215004 Source Range Monitor (SRM) System Re~cord Number 72 2.2 Equipment Control 2.2.6 1 Knowledge of the process for making changes in procedures as described in the safety analysis 2.3 3.3i report.

Increasing the tolerance of the trip setpoint is a change of intent because it is not being performed to Iof align with Technical Specifications. Clarifying a step is permitted under Attachment 1. Changing the level of oversight is permitted IF the change results in increased oversight. "B" SRM Mode switch is a typo error because the Tech is performing "A" SRM channel testing.

Reference Title NC.NA-AP.ZZ-0001 Attachment 1 and Form 1 10CFR55.43(3)

Learning Objectives 00011 3E002 Describe what requirements must be satisfied to make an On-the-Spot change, and the required approval signatures.

Material Required for Examination Qusto SourcExa Bank Question Modification Method: -Significantly Mo -d ified Question Source Comets INPO EXAM BANK QID# 355 Saturday, March 23, 2002 10:47:40 AM Page 77 of 139

Given the following:

- Local Power Range Monitor (LPRM) detector 32-33-C has just failed downscale

- Subsequently, Control Rod 30-31 is selected Which one of the following describes the effect of the failure on the associated APRM and RBM channels?

The LPRM input:

1i will be automatically bypassed and removed from the APRM only. The APRM and RBM readings will be lower than actual power.

Will w be automatically bypassed and removed from both the APRM and RBM. The APRM and RBM readings will remain the same.

El will be automatically bypassed and removed from the APRM only. The APRM reading will remain the same and the RBM reading will be lower than actual power.

El will be automatically bypassed and removed from the RBM only. The APRM and the RBM readings will be lower than actual power.

Anserd Exam Lv B cognitiveLevel C--ompre-hension . ac.i. Hope Creek amDate 03/ 12/2002 Tier: Plant Systems.ROGroup- 1 FSRO-0o I 1 215005K:305 215005 Average Power Range Monitor/Local Power Range Monitor System Record Number 73 K3. Knowledge of the effect that a loss or malfunction of the APRM/LPRM will have on following:

K3.05 Reactor power indication 3.8 3.8 Explanationof The LPRM must be manually bypassed to remove from the APRM averaging circuit. The LPRM is Answer automatically bypassed in the RBM Count Circuit if the detector is reading <4%. Since the LPRM is still feeding the APRM avg, the indicated avg will be lower. Since the control rod is selected after the LPRM fails downscale so the gain change circuit will null to the now lower APRM reference signal.

Reference Title HC.O P-SO. SF-0002 Learning Objectives 000017E008 Given the applicable drawing, determine how the Rod Block Monitor (RBM) System interrelates with the following systems:

a. Local Power Range Monitoring (LPRM) System

b. Average Power Range Monitoring (APRM) System

c. Recirculation Flow Units

d. 120 VAC Instrument Power System

e. 120 VAC Un-interruptible Power Supply System

f. Reactor Manual Control System (RMCS)

IlAW the Rod Block Monitor (RBM) System Lesson Plan IMaterial Required for Examination Qusion SourI INPO Exam Bank Question Modification Method: Significantly Modified QetnSrce Comments: QID# 12556 Limerick 11/10/1995 Saturday, March 23, 2002 10:47:40 AM Page 78 of 139

Given the following:

The plant is operating at full power when the hold down assembly fails on Jet Pumps #1 & #2. This allows the Jet Pump nozzle assembly (Rams Head) to separate from the "B" Recirc Loop piping inside the RPV.

- Annunciators APRM/RBM FLOW REF OFF NORMAL and ROD OUT MOTION BLOCK are also received

- At 10C650, Recirc Pump Discharge Flow indicators are found to be reading 47,000 gpm for "A" Recirc and 54,000 gpm for "B" Recirc Which one of the following describes how the APRM Flow Units will respond in this situation?

FI Upscale trips from all four (4) Fiow units compare trips from only two (2) Flow units C1 C]

compare trips from all four (4) Flow units

'Upscale trips from only two (2) Flow units A a xB Cogniti Comprehension Hope ceek 03/12/2002 Tier: Plant Systems ... tp r Gop Ce 211035K2005 215005 Average Power Range Monitor/Local Power Range Monitor System Record Number 74 K5. Knowledge of the operational implications of the following concepts as they apply to APRM/LPRM:

K5.05 iCore flow effects on APRM trip setpo ints 3.6 3.6 E of Recirc Flow units are part of the APRMS. This topic was chosen because of the KA similarity to AwQuestion #1. Summed flows from both loops will be 101 Kgpm or >111 % rated flow upscale setpoint.

Rated flow is 45,200 gpm per loop. Comparator trips will not be in because each flow unit sums flow through both loops. They will read high but the same values between channels.

Reference Title HC.OP-SO.SE-0001 Table SE-001 Learning Objectives 000016E002 (R) Given a labeled diagram of, or access to, the APRMS/Flow Unit controls located on control room panels 10C608/10C651:

a. Explain the function of each indicator, lAW the Student Handout.

b. Assess the plant conditions that cause each indicator to light or extinguish, lAW the Student Handout.

c. Predict the effect of each control switch on the APRMS/Flow Units, lAW the Student Handout.

d. Select the conditions or permissives required for the control switches to perform their intended function, lAW the Student Handout.

000016Eo05 Given a basic diagram of the recirc flow units and an APRM Block Diagram evaluate how the flow signal is developed for use in determining flow biased setpoints, lAW the Student Handout.

Material Required for Examinato QsonSoure: INPO Exam Bank n if.caton Method: Editorially Modified Question SourceIomments: 1 INPO BANK QID# 675 Duane Arnold 1 05/25/1999 Saturday, March 23, 2002 10:47:41 AM Page 79 of 139

Given the following:

- A Small Break LOCA occurred

- Drywell temp is 450°F and rising

- RPV pressure is 275 psig

- RPV level indication is lost

- 28 control rods are full out

- Suppression Chamber pressure is 10 psig What is action is required to assure adequate core cooling?

Fa-- Enter HC.OP-EO.ZZ-0206, open SRVs until RPV pressure is below 60 psig K Enter HC.OP-EO.ZZ-0206, open at least 5 SRVs SEnter HC.OP-EO.ZZ-0206A, open SRVs until RPV pressure is below 275 psig WEnter HC.OP-EO.ZZ-0206A, open at Ieast 5 SRVs Answrd a B-- o e pplication Fa.i.ity Hope Creek xaDe 03/12 /2002 Tier: PIant Systems R... eup 1 fRO:roul 1 216000A2C)8 216000 Nuclear Boiler Instrumentation Record Number 75 A2. ,Abili to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.08 Elevated containment temperature 3.2 3.4 ExplaionOf 28 rods are out so EOP -206A is appropriate. 5 SRVs are required to assure adequate flow to assure adequate core cooling.

Reference Title EOP-206A step RF-5 Learning Objectives 000134E008 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by that step.

IMaterial Required for Examinatn JEOP Flowcharts without entry conditions QCýue-stion Modification Method:

Saturday, March 23, 2002 10:47:41 AM Page 80 of 139

Given the following:

-120 VAC UPS TROUBLE Annunciator D3-E3 alarms

- "B" channel EGOS Rosemount Trip Units lose power

- "B" channel analog RPV level indications fail downscale Which one of the following 120 VAC inverter malfunctions would cause this loss?

Eli 1BD-481 1iBD-483 1i 1BD-491 nama Level B cognitive Leveli Mmr IiJHpCreek EFxam Date. 03/12/2002 Tier: Plant Systems ROru roup up 121600OK201 216000 Nuclear Boiler Instrumentation Record Number 76 K2. Knowledge of electrical power supplies to the following:

K2.O01 Analog trip system: Plant-Specific 2.8 2.8 Explanation of] AB-1 36 Caution 4.9 states: "The 1(A-D)D481 Inverters power the ECCS Analog Trip Units and the 1(A Anser j D)ID484 Inverters power the Bailey 1E and Non 1E Logic Cabinets." 1BD483 Inverter powers the overhead annunciators which would prevent all overhead alarms from coming in. 1BD492 feeds the BOP Computer. 1BD491 is the B channel essential lighting inverter.

Reference Title HC.OP-AB.ZZ-0136 Caution 4.9 Learning Objectives 000066E018 From memory, summarize/identify the systems/components supplied by the Uninterruptable Power Supplies System, lAW Attachment 2 of the Lesson Plan.

IMaterial Required for Examinato QuestinSoue] New Question Modification Method:

P-U-estion source Comments:

Saturday, March 23, 2002 10:47:41 AM Pg 811oof 139 Page 3

Given the following:

- RPV level dropped until RCIC reached an automatic initiation setpoint

- RCIC failed to automatically initiate

- When armed and pressed, RCIC fails to initiate lAW HC.OP-AB.ZZ-0001 Transient Plant Conditions, which one of the following actions is taken FIRST to manually inject with RCIC?

1i Adjust FIC-R600 RCIC FLOW setpoint to zero %

1iPress and hold the FC-HV-F045 RCIC Steam Supply OPEN PB IEnsure OP-219 RcIC VACUUM PUMP is running El Ensure BD-HV-F046 Lube Oil Cooling is open A cCognitiv Memory Fa..ityHope Creek E 03/12/2002 Tier: , lant Systems ........ .oup...217000A201 1......

217000 Reactor Core Isolation Cooling System (RCIC) e

... ber7..7 A2. Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2,01 System initiation signal 3.8 3.71 Eof HC.OP-AB.ZZ-0001 Attachment 6 step 8.1. answer c. is first, followed by d. , b. then a.

Reference Title HC.OP-AB.ZZ-0001 Attachment 6 S10CFR55.43(5)

Learning Objectives 000030E022 (R) Given RCIC turbine control system failures, evaluate and determine the effect on the RCIC system, lAW the RCIC System Lesson Plan.

000030E023 (R) Given any of the following and appropriate control room reference material, evaluate and determine the effect on the RCIC system, lAW the RCIC System Lesson Plan:

a. A given valve opening or closure

b. Loss of DC or AC power supply

c. Inadequate system flow

d. An oil system malfunction

e. Failure of the RCIC Gland Seal Condenser Vacuum Pump

f. Loss of room cooling

g. Rupture disc failure on the RCIC exhaust

h. Steam line break

i. Low condensate storage tank level IMaterial Required for Examination

[~uesion Sour~ce New lQuestion ModificationMeh:

lQuestion Source Comments:

Saturday, March 23, 2002 10:47:41 AM Page 82 of 139

Given the following:

- A reactor scram occurred during a startup

- RCIC started and tripped

- RCIC Turbine Exhaust piping has ruptured

- Reactor pressure is 50 psig and lowering

- A small steam leak in the Drywell is causing Drywell pressure to rise Which one of the following valves will automatically close if Drywell pressure reaches HI-HI?

8. BD-HV-F031 Torus Suction Isolation Valve EI BD-HV-F013 Pump Discharge to Feedwater Isolation Valve H FC-HV-F062 Vacuum Breaker Isolation.Valve .

F1 FC-HV-F059 Exhaust Line Isolation Valve A c Ex B Icognitive, Lee C rehension Hp 1i70 e 03/12/2002 Tier-, Plant Systems . ..... SR Gou 1 217000K405 217000 Reactor Core Isolation Cooling System RCIC Record Number 78 K4. Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following:

K4.05 Prevents radioactivity release to auxiliary/reactor building 3.2 3.5!

Expl~anaionof RCIC Exhaust vacuum breaker isin valve will close to isolate the piping breach Reference Title

-HC.O-P-SO. BD-0001 Learning Objectives 000030E023 (R) Given any of the following and appropriate control room reference material, evaluate and determine the effect on the RCIC system, lAW the RCIC System Lesson Plan:

a. A given valve opening or closure LA ~ ~~~~c7...b. Loss of DC or AC power supply . . . . .. .... ..... ....... .. . . ... .

c. Inadequate system flow

d. An oil system malfunction

e. Failure of the RCIC Gland Seal Condenser Vacuum Pump

f. Loss of room cooling

g. Rupture disc failure on the RCIC exhaust

h. Steam line break

-i.-- Low-condensate storage tank level Material Required for Examinato IQuestion Source New [i~u-eson modification Method:

!i"i*ii*¸ *i* iii*iii* *ii k!_*L'!*'i_'i!*i!i_-"i i ii ;!*ii _"i!!_-i_-i"ii i*i Saturday, March 23, 2002 10

47:41 AM Page 83 of 139

Given the following:

- "A" RHR Pump is running in Suppression Pool Cooling mode

- A complete loss of offsite power occurs

- All Emergency Diesel Generators have automatically started and aligned to their respective busses Which one of the following describes the response of the "A" RHR Test Return Valve BC-HV F024A?

Fa Remains open until CLOSE PB is pressed FiRemains open until AUTO CLOSE OVERRIDE PB is pressed F11 Receives close signal 5 seconds after bus reenergized

- Receives close signal 10 seconds after bus reenergized a R Cognitive Comprehension Facility Hope-creek -- mDae 03/12/2002 Tier: Plant Systems r 2 R ro 2 219000A301 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode Record Number 79 A3. Ability to monitor automatic operations of the RHR/LPCI: TORUSISUPPRESSION POOL COOLING i MODE including:

A3.01 Valve operation 3.3 3.3,

[xplanaton ofOn LOP, F024A will remain as is. Once power is restored, valve will remain open because there is no

[Answer LOCA signal. 5 and 10 second delays are for pump start for normal or emergency power.

Reference Title HC.OP-SO. BC-0001 Learning Objectives 000028E012 Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, assess the status of the Residual Heat Removal System or its components by evaluation of the controls/instrumentation/alarms lAW the RHR System Lesson Plan.

[Material Required for Examination

[ on Sourý New .Question Modification Method:

Question Source Comments: INPO EXAM BANK QID# 12246 Limerick 01/20/1998 Concept Used Saturday, March 23, 2002 10:47:42 AM Page 84 of 139

Given the following:

- A LOCA has occurred

- The CRS directs the Suppression Chamber to be vented lAW HC.OP-EO.77-0318 Containment Venting

- Instrument air header pressure is 0 psig Which one of the following describes how the Hard Torus Vent path valves/dampers are operated JAW HC.OP-EO.ZZ-0318 under these conditions?

[i PCIG opens the inboard damper; the out bo:ard valve is motor operated tI PCIG opens the inboard damper; the outboard valve is manually operated IThe inboard damper is motor operated; the outboard-valve is motor operated W The inboard damper is manually operated; tthe outboarddvalve is manuaily operated Anwr d_! Exam LevelR Conitive -Level Comprehension IFaility Hope Creek Exam Date: 03/12/20 Tier: Plant Systems R1 G 1 2230011K613 223001 Primary Containment System and Auxiliaries Record Number . 80.

K6. Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES:

K6.13 Applicable plant air system/ nitrogen make-up system . 3.2 3.4 Elaionof Thehinoard damper is normally operated with Instrument Air. Since IA is zero psig, the only way to vent is using Hydraulic Manual operators on both valves.

Reference Title HC.OP-EO.ZZ-0318 Learning Objectives 000158E°04 From memory, describe any/all flow paths established by the performance of each of the 300 series Emergency Operating procedures.

Material Required for Examinato

[uston ource: New Question Modification Method:

lQuestion Source Comments:

Saturday, March 23, 2002 10:47:42 AM Page 85 of 139

The plant is in Cold Shutdown with Shutdown Cooling in service. A single transmitter fails causing a loss of Shutdown Cooling.

Which one of the following caused the trip?

1iN078B RPV Pressure transmitter fails upscale upscale F1 NO80A RPV Level transmitter fails E N080A RPV Level transmitter fails downscale K N078B RPV Pressure transmitter fails downscale nswer a a e Cogtive Level Mem0C F. ciitHpecreek Exam Date: 03/12/2002 Tier: Plant Systems 1 RG 1 223002K316 223002 Primary Containment Isolation System/Nuclear Steam Suppjy Shut-Off 'Record Number 81 K3. Knowledge of the effect that a loss or malfunction of the PCIS/NSSSS will have on following:

K3..16 Shutdown cooling system/RHR 3.2: 33 Explanationo*fI Justification: N080A RPV Level transmitter fails upscale-Incorrect- failure mode would be downscale Answer N080A RPV Level transmitter fails downscale-lncorrect- correct mode of failure but requires two detectors per channel to fail N078B RPV Pressure transmitter fails downscale-lncorrect- wrong failure mode needs to see high pressure not low N078B RPV Pressure transmitter fails upscale-Correct -Pressure transmitter upscale is single coincidence isolation Reference Title iHC.OP-SO.SM-0001 Learning Objectives 000045E010 I Given a malfunction of the NSSSS, which either isolates or fails to isolate a plant system, evaluate and explain the effects, if

- any, of that malfunction on each of the following lAW the NSSSS Lesson Plan.

a. Reactor Water Level

b. Fuel Cladding Temperatures

c. Inplant/Offsite Radiological Concerns

d. Reactor Pressure 000045E014 (R) Given a specific parameter, which initiates NSSSS, isolation signals, identify all valves isolated by that parameter and the setpoint at which the isolation signal is generated lAW the NSSSS Lesson Plan.

laterial Required for Examinati.n ustonSour: Facility Exam Bank Question Modification Method: Direct From Source tuestion Source Comments: I VISION BANK QID# Q56211 Saturday, March 23, 2002 10:47:42 AM Page 86 of 139

Given the following:

- The plant is operating at 100 percent power

- Main Steam Line (MSL) "A" Flow Transmitter PDT- N086A fails low Which one of the following describes how Main Steam Lines will be isolated if an actual high flow in the "A" MSL occurs?

(LIMIT YOUR RESPONSE TO MAIN STEAM LINE FLOW INSTRUMENTATION ONLY)

"A" and "C" NSSSS logic will trip.closing. Inboard MSIVs only

" "A" and "D" NSSSS logic will trip closing Outboard MSIVs only F "B"and "C" NSSSS logic will trip closing Inboard and Outboard MSIVs P "B" and "D"NSSSS logic will trip closing Inboard and Outboard MSIVs c - [Eam Level B Cogniive Level 1Comprehension a Hope Creek 03/12/2002 Tier: Plant Systems RO Iou 1 SRO 223002K401 22-3002 .... -Primary Containment Isolation System/Nuclear Steam Supply Shut-Off...... Record Number . 82 K4. I Knowledge of PCIS/NSSSS design feature(s) and/or interlocks which provide for the following:

K4.01 !Redundancy 3.0 3.2 MSL Flow transmitter failed low in A MSL will prevent A NSSSS logic from tripping. B, C, and D NSSSS Mpationf flow transmitters on the A MSL will trip in response to an actual high flow but only C and B or C and D can make the MSIVs go closed I Reference Title HC.OP-SO.SM-0001 HC Tech Specs 3.3.2 Learning Objectives 000045E005 Given a labeled diagram/drawing of NSSSS controls, identify/explain each of the following lAW the NSSSS Lesson Plan.

a. The function of each indicator.

b. The condition which will cause the indicator to light or extinguish.

c. The effect of each control on the NSSSS.

Material Required for ExaminaIP&D M-41 Sheet 1 Source: I New Modification Method: .u.estion Saturday, March 23, 2002 10:47:42 AM Page 87 of 139

Given the following:

- Drywell pressure is 13 psig

- "A" RHR pump is running

- The CRS orders Drywell Spray initiated on the "A" RHR loop

- The associated Drywell Spray Containment Isolation Valves are opened Which one of the following describes actions required to establish desired RHR flow?

I] Throttle BC-HV-F048A to obtain 540 gpm flow on FI-4461A 1] Throttle BC-HV-F048A to obtain 10,470 gpm flow on FR-R608A W Throttle BC-HV-FO03A to obtain 540 gpm flow on FI-4461A L] Throttle BCH V-FI0003A to obtain 10,470 gpm flow on FR-R608A n -swe-rExam Lee CognitiveLeve Comprehension iHope Creek Exam Date .. 03/12/2002 Tier: Plant Systems FRO.Grou.P 2ISRO r 1 226001A106 226001 RHR/LPCI: Containment Spray System Mode Record Num . .83 Al. Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE controls including:

A1.06 System flow :3.2.3.2i

[planationof lAW HC.OP-SO.BC-0001 and HC.OP-AB.ZZ-0001 Drywell Spray is throttled to maintain 10,470 gpm loop flow on FR-R608A in the control room.

FI-4461 is Suppression Chamber Spray Flow indication F048A valve is fully closed.

P&ID M-51-0 Sheet 2 is used by the student to determine which flow indicator monitors Drywell Spray Reference Title HC.OP-SO.BC-0001

ýM-51 Sheet 2 Learning Objectives 000028E011 Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 10C650:

a. Explain the function of each indicator lAW the RHR System Lesson Plan.

b. Assess plant conditions which will cause the indicators to light or extinguish lAW the RHR System Lesson Plan.

c. Determine the effect of each control on the RHR System lAW the RHR System Lesson Plan.

d. Assess plant conditions or permissives required for the control switches/pushbuttons to perform their intended functions lAW the RHR System Lesson Plan.

material Required for Examination P&ID M-51 Sheet 2 Question Source: New .i..uest.on ModificationMethod Q..u.es.tion Source Comments:...

Saturday, March 23, 2002 10:47:42 AM Page 88 of 139

In response to a steam leak in the Drywell, the "B" loop of RHR was placed in Suppression Chamber Spray and Suppression Pool Cooling.

The "A" loop of RHR was placed in Drywell Spray lAW EOP-102.

Select the automatic system response as Drywell pressure lowers below 1.68 psig. Assume no other operator action.

Drywell and Suppression Chamber sprays isolating.

Drywell and Suppression Chamber sprays continuing.

1i Drywell spray continuing and Suppression Chamber spray isoIating.

Drywell spray isolating and Suppression Chamber spray continuing.

A b ae B c e Memor atope Creek [xam Date: 03/122002 Tier: Plant Systems ROGrouP 2 SRO Group 1 226001A305 226001 RHR/LPCI: Containment Spray System Mode Record.Number 84 A3l Ability to monitor automatic operations of the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE

.including:

A3.05 Containment pressure 4.0 4.0 Explanation of Drwel spray valves need 1.68 psig permissive to open however once started open the valves will stay open. There is no interlock to close the valves on low Drywell pressure.

Reference Title HC.OP-SO.B3C-0001 Learning Objectives 000028E01 1 Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 10C650:

a. Explain the function of each indicator lAW the RHR System Lesson Plan.

b. Assess plant conditions which will cause the indicators to light or extinguish lAW the RHR System Lesson Plan.

c. Determine the effect of each control on the RHR System lAW the RHR System Lesson Plan.

d. Assess plant conditions or permissives required for the control switches/pushbuttons to perform their intended functions lAW the RHR System Lesson Plan.

laterial Required for Examinatin u oSoure: INPO Exam Bank Quest.on.Modification Method: Significantly Modified uestion SourceComments: INPO BANK QID# 18032 Pilgrim 1 10/16/1998 Saturday, March 23, 2002 10:47:43 AM Page 89 of 139

Given the following:

- Suppression Chamber pressure is elevated

- The CRS orders Suppression Chamber Sprays placed in service

- While opening the "B" RHR Suppression Chamber Spray Valve F027B, indications are as follows:

- Yellow OVLD/PWR FAIL light is FLASHING

- Green CLSD light is EXTINGUISHED

- Red OPEN light is LIT

- White OVERRIDDEN light is LIT Which one of the following describes the valve status?

Fa The valve breaker is tripped open. The valve is open with spray flow.

ElThe valve breaker is tripped open. The valve is closed.

El The valve overloads have tripped. The vaive is open with spray flow.

SThe valve overloads have tripped. The valve is closed.

Anwrc aB ICognitive Level- Comprehension Hope Creek Exam Da:6 03/12/2002 Tier: Plant Systems .. .r... 2 R.u 2 230000K601 230000 RHR/LPCI: Torus/Suppression Pool Spray Mode Record Number. 85 K6. Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI:

TORUS/SUPPRESSION POOL SPRAY MODE:

.K6.01 A.C. electrical .. 3 33.4 Exlaaionof Valve motor overloads have tripped causing the yellow flashing light. Red OPEN light Lit means the MOV still has power, therefore the breaker is not tripped.

Reference Title HC.OP-AR.ZZ-0005 Attachment B1 Learning Objectives 000028E0ll1 Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 10C650:

a. Explain the function of each indicator lAW the RHR System Lesson Plan.

b. Assess plant conditions which will cause the indicators to light or extinguish lAW the RHR System Lesson Plan.

c. Determine the effect of each control on the RHR System lAW the RHR System Lesson Plan.

d. Assess plant conditions or permissives required for the control switches/pushbuttons to perform their intended functions lAW the RHR System Lesson Plan.

IMaterial Required for Examination Q New Question Modification Method:

Saturday, March 23, 2002 10:47:43 AM Page 90 of 139

'Given the following:

- The plant is in Operational Condition 2 with a reactor startup in progress

- One Fuel Pool Cooling Pump, Heat Exchanger and demin are in service

- Fuel Pool inventory is slowly lowering

- Digital alarms and leak detection monitors do NOT identify the source of the leakage

-ALL sump pumps appear to be operating normally

- CST level is stable

- HC.OP-AB.ZZ-0144, Loss of Fuel Pool Inventory/Cooling is entered Which one of the following actions is required lAW HC.OP-AB.ZZ-0144 Attachment 1.

Ial Isolate FPCC Heat Exchanger.

IEnter the Drywell and check for leakage

[c] Check Torus Level and verify RHR alignment iisolate RWCU Non-Regenerative Heat Exchanger Aa Ecognitive Level Aplication Faciity Hope Creek a 03/12/2002 Tier: Plant systems . ...... 3 SGo 3 233000(G107 233000 -1 Fuel Pool Cooling and Clean-up Record Number 86 2.1 Conduct of Operations 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating 3.7 4.4 characteristics, reactor behavior, and instrument interpretation.

[xplanation of Justification Answe CORRECT - Isolate FPCC Heat Exchanger. Following the flow chart of Attachment 1 of HC.OP-AB.ZZ 0144: All conditions enter the first two decision blocks on excessive sump pump operations. The stem stipulates that all sump pumps are operating normally. The third decision block is in regards to the CST level. The stem stipulates that the CST level is normal. The fourth decision block is in regards to RPV head status. The stem stipulates OC2; hence the RPV head is installed. The fifth decision block asks whether FPCC or RHR FPCC Assist is in service. The stem stipulates that FPCC is in service.

Therefore, the action is to check for increasing SACS Head Tank Levels - isolate FPCC Hx.

I! Reference Title HC.OP-AB.ZZ-0144 Attachment 1 Learning Objectives 0AB144E005 (R) Interpret and apply charts, graphs and tables contained within the Loss Of Fuel Pool Inventory/Cooling, Abnormal Operating Procedure.

[material Required for Examination HC.OP-AB.ZZ-0144 Attachment 1

[ustinsouce Facility Exam Bank Question Modification Method. Significantly Modified lQuestion Source Comments: -Vision BANK-QI D#-Q61347 ....... .

Saturday, March 23, 2002 10:47:43 AM Page 91 of 139

Which one of the following describes the bases for the Refueling Platform Main Grapple weight limit interlocks?

I] Prevents release of activity in excess of that contained in a single fuel assembly SPrevents damage to core internals from excessive lifting force

[c] Prevents damage to hoist safety brake from excessive speed

-Preventsengaging more than one fuel assembly or control rod blade guide AS Cognitive Level mory Fac.lity Hope Creek x Date: 03,12/200 2 Tier: Plant Systems RO 3 SRO "Go 2 234000G225 234000 _Fuel Handling E ipment Record Number 87 2.2 Equipment Control 2.2.25 'Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. 2.5 3.7

[Explanationof I Tech Spec bases 3/4.9.6 Tech Spec bases 3/4.9.6 1(..F. . . . Learning

. . Objectives 000226E012 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Choose those sections which are applicable to the refueling platform and associated equipment lAW HCGS Technical Specifications.

b. Evaluate Refuel Platform operability and determine required actions based upon system operability lAW HCGS Technical Specifications.

c. Explain the basis for those Tech Spec items associated with the refuel platform lAW HCGS Technical Specifications.

(SRO only)

Materal Required for Examination J Tech Specs without Definitions, Safety Limits, and bases Question Source New l~eo Moiiaio ehd Quetio Sorce Comments:

Saturday, March 23, 2002 10:47:43 AM Page 92 of 139

Given the following:

- The plant is operating at 100 percent power

- SRV "B" has inadvertently opened

- Operators attempt to close the SRV lAW HC.OP-AB.ZZ-0121 "FAILED OPEN SRV/RELIEF VALVE", which one of the following is a positive indication that the SRV has CLOSED?

The "B" SRV...

a] "Sv ENRGZ"' iight extinguishes.

I Acoustic Monitor green light illuminates.

I] associated power fuse is pulled.

W tailpipe temperature stabilizes.

Anwerb- Exam e R Cognitive Level Memor a Hope Creek E1am Date. 03/12/2002 Tier: Plant Systems ... .. . . I SRO Group 1 239002A102 239002 Relief/Safety Valves Record Number 88 Al. Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:

A1.02 Acoustical monitor noise: Plant-Specific :3.7 3.8 Expanion o Acoustic Monitor green close light on is used to verify the SRV is closed Reference Title HC.OP-AB.ZZ-0121 step 4.5 Learning Objectives 0ABI21E001__ Recognize abnormal indications/alarms and/or procedural requirements for implementing, Failed Open Safety Relief Valve, Abnormal Operating Procedure.

OAB121EO04 Explain the reasons for how plant/system parameters respond when implementing, Failed Open Safety Relief Valve, Abnormal Operating Procedure.

OAB121EO06 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Failed Open Safety Relief Valve, Abnormal Operating Procedure.

Material Required for Examination Q New .Quest.on Modification Method:

! . . . . . . . . . . . .. ... .. .. .. .. ... .. .. .I Saturday, March 23, 2002 10:47:43 AM Page 93 of 139

The following plant conditions exist at T = 0:

- Reactor water level is -130 inches

- Reactor pressure is 900 psig

- Drywell pressure is 1.2 psig

-All ECCS pumps are running

- MSIV's are closed Based on plant conditions, which one of the following describes the response of ADS?

II ADS will initiate at T =.105 seconds

[i ADS will initiate at T= 300 seconds I] ADS will initiate at T = 405 secon ds W ADS will NOT initiate until Drywell pressure increases above 1.68 psig c B J-- gnitiv e Comprehension i Hope Creek x 03/12/2002 Tier:* Plant Systems

.... rou'p]Gro.p S.. 1 239002A105 239002 Relief/Safety Valves Record Num*ber 89

.Al. Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:

A1.05 Reactor water level 3.7 3.81 Explnatonof Justification: lAW HC.OP-SO.SN-0001

ADS will NOT initiate until Drywell pressure increases above 1.68 psig.-Incorrect- it will initiate based on the DW Press Bypass Timer

-ADS will initiate at T = 105 seconds.-Incorrect- With the plant conditions as stated, the ADS HIGH DRYWELL PRESSURE BYPASS TIMER will have initiated. This timer is 5 minutes (or 300 seconds).

Once this timer is timed out the ADS initiating timer starts. This has a time of 105 seconds. The total time required to reach initiation is 405 seconds

-ADS will initiate at T = 300 seconds. Incorrect- With the plant conditions as stated, the ADS HIGH DRYWELL PRESSURE BYPASS TIMER will have initiated. This timer is 5 minutes (or 300 seconds).

Once this timer is timed out the ADS initiating timer starts. This has a time of 105 seconds. The total time required to reach initiation is 405 seconds

-ADS will initiate at T = 405 seconds.-Correct Reference Title HC.OP-SO.SN-0001 section 3.3.1 Learning Objectives 000029EO1 0 (R) From memory, evaluate the interrelationship between the Automatic Depressurization System and the following, lAW the Automatic Depressurization System Lesson Plan:

a. Residual Heat Removal (RHR) and Core Spray Systems

b. Deleted

c. Primary Containment Instrument Gas (PCIG) System

d. 125 VDC Class IE Distribution System

e. 120 VAC Uninterruptible Power Supply (UPS) Instrumentation Material Required for Examination

[Question Sourc: Facility Exam Bank IQuestion Modification Method: Editorially Modified

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l Saturday, March 23, 2002 10:47:43 AM Page 94 of 139

Question Source Comments: I Vision Bank QID# Q53701 Saturday, March 23, 2002 10:47:44 AM Page 95 of 139

Given the following:

- With power at 22%, a loss of Stator Cooling occurred

- All automatic actions occurred as designed

- The turbine did NOT trip

- HC.OP-AB.ZZ-0138 MAIN TURBINE TRIP/MALFUNCTION has been entered

- There is no time estimate for restoration of Stator Cooling The decision if and when to trip the Main Turbine is based upon:

E] stator cooling water conductivity prior to the start of the transient.

F1 the rate of increase of stator temperatures after the runback is complete.

Sthe current plant location on the power to flow map.

Wfinal main generator field (amps) after the runback has gone to completion.

nswer -a Lel Remory Level Me.y.tope ___ope. CrekCognitive 03/12/2002 Tier: Plant Systems RF.G.Oup 2 SROG 2 245000 K502 245000 Main Turbine Generator and Auxiliary Systems Record N.mb.r 90

.K5. Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS:

K5.02.. Turbine operation and limitations 2.8 3.11 ationof Conductivity readings are not valid following loss of system flow. The conductivity reading prior to the event is key.

Reference Title HC.OP-AB.ZZ-0138 Step 4.4.8 Learning Objectives 0AB138E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Turbine Generator Trip/Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination

[~ueson Soure INPO Exam Bank Q..stio..Modif.cati.n..et....Editorially Modified lQu tion Source Comments: 7 INPO Exam Bank QID# 14179 Peach Bottom 2 03/26/2001 Saturday, March 23, 2002 10:47:44 AM Page 96 of 139

Given the following:

-The plant is operating at 100 percent power

-Instrument air is lost to the following valves:

- AD-LV-1 657-1 Condensate Makeup

- AD-LV-1 657-2 Condensate Reject

- AD-FV-1 677 SCP Suction Reject Bypass lAW HC.OP-AB.ZZ-01 31 "LOSS OF INSTRUMENT AIR AND/OR SERVICE AIR", which one of the following describes the Condensate System response and operator "Contingency Action" necessary to mitigate the event?

1~1Condensate Reject Valve fails open; Close Condensate Makeup Bypass Valve to restore Hotwell level 1iCondensate Reject Valve fails open; Open Condensa te M akeup Isol at ion Valve to res tore Hotwell level

~iCondensate Makeup Valve fails closed; Ope n Co nden sate Ma keup Bypas s Valve to restore Hotwell level I~Condensate Make up Valve fails closed; Close Co ndensate Makeup Isol ation Valve to restore Hotwell-level An sw-er- C [&-a mL-e-vel R lCognitive Level Copeeso aiiyHope Creek -Ea ae 03/12/2002 Tier: ýPlant Systems [RO-roupý 2 R-ru 256000A213 256000 Reactor Condensate System Riecord Number- 91

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

A2. 13- Loss of applicable plant air systems 2.9 3.0 Exlntoof Contingency action of AR3-I131. Condensate Makeup valve LV-1 657 fails closed. Makeup bypass valve Answer v091 is opened to raise level Reference Title

ýHC.OP-AB.ZZ-0131 Attachment 1 iM-05 sheet 3 Learning Objectives OAB1 31 E004 Explain the reasons for how plant/system parameters respond when implementing, Loss Of Instrument Air And/Or Service Air, Abnormal Operating Procedure.

IMaterial Required for Examination P&ID M-05 Sheet 3

'Questiorn Source: New__ lQuestion Modification etho~d:

Saturday, March 23, 2002 10:47:44 AM Pg 97 Page 7oof 139 3

Given the following:

- The plant is at 70% power

- The Main Turbine trips causing Hi Hi levels in the 1A, 2A heaters and 2A drain cooler Which one of the following describes the valves that isolate for the 1A, 2A heaters and 2A drain cooler?

C...Condensate side inlet and outlet valves SThe extraction steam isolation valves SThe High Level Dump valves WThe Startup and Operating vent valves a Ea .... el....CognitiveLe Memory _ii[Hope Creek E 03/12/2002 Tier: IPlant Systems F6_rýOup 2 G 3 256000K406 256000 Reactor Condensate System Record umber . 92 K4. i Knowledge of REACTOR CONDENSATE SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.06 Control of extraction steam 2.8 2.8 Explanation of JUSTIFICATION:

Condensate system inlet and outlet valves for the 1A, 2A heaters and 2A drain cooler.

Correct - lAW HC.OP-SO.AF-0001, Section 3.2, Limitations.

The extraction steam isolation valve for the 1A, 2A heaters and 2A drain cooler.

Incorrect - there is no extraction steam isolation valve for the 1A, 2A heaters and 2A drain cooler.

The level control valves for the 1A, 2A heaters and 2A drain cooler.

Incorrect - there are no level control valves for feedwater heater 1A; only the normal level control valve for feedwater heater 2A (and drain cooler 2A)--going to feedwater heater 1A--would close Condensate inlet valve for the 1A, 2A feedwater heaters, allowing condensate flow through the 2A drain cooler.

Incorrect - the condensate inlet and outlet valves close, isolating flow through feedwater heaters 1A and 2A and the drain cooler; they are in series.

Reference Title HC.OP-SO.AF-0001 Section 3.2.4 Learning Objectives OAB118E004 Explain the reasons for how plant/system parameters respond when implementing, Loss Of Feedwater Heaters, Abnormal Operating Procedure.

000055E008 (R) From memory, determine the automatic system response associated with the following abnormal conditions for all feedwater heaters, lAW the lesson plan.

a. Heater high level

b. Heater trip

c. Main turbine trip Material Required for Examinati.on-.. .... ..

[ son Sour: Facility Exam Bank Question Modification Method. Significantly Modified Saturday, March 23, 2002 10:47:44 AM Page 98 of 139

IQuestionSource Comments: I Vision QID# Q53238 Saturday, March 23, 2002 10:47:44 AM Page 99 of 139

Given the following:

- The plant is operating at 90 percent power

- Power ascension in progress

- 1BD483 120 VAC inverter output is lost In addition to entering HC.OP-AB.ZZ-0136 "Loss of 120 VAC Inverter", which other operating procedure must be entered for this condition and why?

Fa] HC.OP-EO.ZZ-0101 R"RPV ControI. to stabilize reactor pressure El HC.OP-AB.ZZ-0143 "Loss of Overhead Annunciators Loss0 of CRIDS" to stabilize RPV Level FJ HC.OP-EO.ZZ-01 01A "ATWS RPV Control" to respond to failure to scram S'HC.OP-AB.ZZ-0153 "Optic Isolator Panel Malfunction" to respond to single Recirc Pump trip Ab Exam.. Leve....Cognitivej evel ].Memory Facilit. Hope

a Creek 03112/2002 Tier: - Plant Systems FR r Group 1 259002G432 259002 Reactor Water Level Control System Record N-umber 93 2.4 Emergency Procedures and Plan 2.4.32 Knowledge of operator response to loss of all annunciators. 3.33.51 Explanation of Loss of BD 483 inverter causes loss of overhead annunciators and trip of B RFPT. AB-143 directs n operator to stabilize level if 1BD483 inverter is lost. The reactor should not scram.

Reference Title HC.OP-AB.ZZ-0143 10CFR55.43(5)

Learning Objectives 0AB143E004 Explain the reasons for how plant/system parameters respond when implementing, Loss Of The Overhead Annunciators/Loss of CRIDS, Abnormal Operating Procedure.

IMaterial Required for Examination Quesi~on Sou ] New QetoMoification Method.

IQuestion Source Comments:

Saturday, March 23, 2002 10:47:44 AM Page 100 of 139

Given the following:

- The plant is in Operational Condition 5

- All RBVS and RBVE fans are running

- FRVS is in a normal standby configuration

"- "B" and "D" Emergency Diesel Generators are tagged out for maintenance A radiological incident on the Refuel Floor causes Refuel Floor Exhaust Radiation to reach 3.1x10 3 uCi/ml.

Which one of the following describes total FRVS recirculation flow one minute after this event?

(Assume no operator actions)

L] 90,000 cfm 1120,000 cfm W180,000 cfm.

Answerd a tie Le Comprehension iHope Creek Ea 0311212002 Tier: Plant Systems . . .. .. roup 1 RGo 1 261000A407 261000 Standby Gas Treatment System Record Number. 94 A4. iAbility to manually operate and/or monitor in the control room:

A4.07 System flow .3.1.3.. 2 Epnion of RFE HIHi Start of FRVS is 2.0 E-2 uCi/ml. The value given is above this setpoint.

ARefuel Floor Exhaust HIHI starts all 6 fans at 30,000 scfm each, since normal power is available.

Reference Title IHC.OP-SO.GU-0001 section 3.1.1 Learning Objectives 000042EO006 I(R) Given plant conditions, distinguish between the automatic starts and stops associated with the Filtration Recirculation I Ventilation System (FRVS) Recirc Fans, lAW the Lesson Plan.

Material Required for Examination Facility Exam Bank Puestion Modification Editorially Modified mrnents: I Vision QID# Q60662 Saturday, March 23, 2002 10:47:44 AM Page 101 of 139

Given the following:

- The plant is operating at 100 percent power

- A Loss of Offsite power occurs

- Drywell pressure is 5 psig and rising

- "A" Emergency Diesel Generator fails to start Which one of the following describes the effect on FRVS after 3 minutes?

(Assume NO operator action)

I1 Only 3 Recirc Fans and one Vent Fan start SOnly 4 Recirc Fans and one Vent Fan start FI Only 3 Recirc Fans and NO Vent Fan start SOnly 4 Recirc Fans and No Vent Fan start AnSw b--er b B-omLeve B Cognitie mprehension [Hope Creek 03/12/2002 Tier: Plant Systems .r....1- R. Groui 1 261000K603 261000 -Standby Gas Treatment System Record Number 95 K61. Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM:

K6.03 [Emergency diesel generator system 3.0 3.1 Fof A EDG powers A and E FRVS Recirc and A Vent fan. The B,C, D and F Recirc Fans will start as well as Answer Jthe B Vent Fan after Low Flow from A Fans starts the Auto fan.

Reference Title IHC.OP-SO.GU-0001 Learning Objectives 000040E012 (R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, identify the status of the Secondary Containment by evaluation of the controls/instrumentation/alarms lAW the Secondary Containment Lesson Plan.

material Required for Examinato lQuestlon Modification Method:I Saturday, March 23, 2002 10:47:44 AM Page 102 of 139

Given the following:

- A station blackout has occurred

- All 500 KV Lines to Hope Creek are de-energized lAW HC.OP-AB.ZZ-01 35, which one of the following 500 KV lines is the first to be re-energized to restore power to Hope Creek 13 KV ring bus?

al Red Lion 5015 Line .......

D...Deans 5021 Line SNew Freedom 5023 Line SSalem 5037 Line Ai=**w d Exam Levelm R Cognitive Level Memory I! Hope Creek J~xan Dae_*J __03/12/2002 Tier: Plant Systems [O Group] 2Z .ro . 262001K201 262001 A.C. Electrical Distribution Number 96

.K2. Knowledge of electrical power supplies to the following:

K2.01 'Off-site sources of power 3.3 3.6 AB-1 35 power restoration strategy is to restore power via the Salem 5037 line and the Salem Gas Turbine Reference Title HC.OP-AB.ZZ-0135 LearlningObjectives 000065E01 5 - (R) Given plant problems/industry events associated with the Main Power System:

a. Discuss the root cause of the plant problem/industry event lAW the Main Power System Lesson Plan.

b. Discuss the HCGS design and /or procedural guidelines that mitigate/reduce the likelihood of the problem/event lAW the Main Power System Lesson Plan.

c. Discuss the "lesson learned" from this problem/event lAW the Main Power System Lesson Plan.

OAB135E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Station Blackout/Loss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination

[l~uestonSo New IQuestion Modification Method:I QustonSorce Comments-,

Saturday, March 23, 2002 10:47:45 AM Page 103 of 139

Given the following:

- The plant is operating at 100 percent power

- Several overhead annunciators alarm including:

- 4.16KV FDR TO USS XFMR BRKR MALF

- 4.16KV SYS INCOMING BRKR MALF

- Yellow INOP control bezels are flashing on 10A401 "A" 1E 4.16KV bus circuit breakers but the equipment does NOT change state

- Reactor power, pressure, and level remain stable Which one of the following caused the alarms?

[i Loss of power to the Optical Isolator Cabinets of inverter power to the "A" Channel 1 E Bailey Cabinets SLoss I] Loss of AC powe r t0o thle 10A401 bus Ii Loss of DC control power to the i 0A401 bus A d a B= Cognitive Comprehension a Hope Creek 03/12/2002 Tier, Plant Systems R9...u sRO Gop 262001K601 26200_1__..... A.C. Electrical Distribution Record Number 97 K6.1 Knowledge of the effect that a loss or malfunction of the following will have on the A.C. ELECTRICAL DISTRIBUTION:

K6.01 D.C. power 3.1 3.4 of A loss of DC control power prevents the breakers from tripping and causes flashing INOP alarm on Ane bezels. A loss of AC power would cause equipment to trip, specifically A RFPT would trip which would cause RPV level to change. Loss of inverter power to Bailey would cause all control room breaker bezels to go dark. Loss of power to the optical Isolation cabinets would not cause flashing INOP bezels.

Reference Title HC.OP-AR.ZZ-0016 Learning Objectives 000066E026 Given a set of conditions and a drawing of the controls, instrumentation, and/or alarms located in the main control room, assess the status of the 1 E AC Power Distribution by evaluation of the controls/instrumentation/alarms lAW the Lesson Plan.

000066E027 Given the loss of a portion of the DC distribution system, evaluate the affect on the 1 E AC distribution system lAW the Lesson Plan.

Material Required for Examination Question Source: INew, Question Modification Method:7 Question Source Comments:

Saturday, March 23, 2002 10:47:45 AM Page 104 of 139

Given the following:

- The plant is operating at 100 percent power

- HPCI 250 VDC battery has just completed deep discharge rate surveillance testing

- The HPCI Battery charger has been returned to service and associated fuse transfer switch closed

- Overhead annunciator 250 VDC TROUBLE alarm remains ILLUMINATED Which one of the following is recommended by HC.OP-AB.ZZ-0149 250 VDC MALFUNCTION prior to declaring the HPCI 250 VDC system operable?

El Perform the Maintenance Weekly Battery SurveiIlance EI Place the battery charger timer to the ZERO position El Verify the battery charger voltage is Iess than 268 volts F1 Verify charging current is less than 5 amps A-.sw.e.r _ntive B- e elMemory Fiy Hope Creek Exa0mDate: 03/12/2002 Tier: Plant Systems .. RO IG*ro*u p 263000A101 2ROGrup 2 263000 D.C. Electrical Distribution Record Number, 98

.Al. IAbility to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including:

A1.01 Battery charging/discharging rate 2.5 12.8[

of AB-149 recommends performing Maint surv HC.MD-ST.PJ-0001 250 Volt Weekly Battery Surveillance to verify battery operability following the battery discharge event. This is based on OE9182 - Battery Inoperable When Exiting LCO where the battery was declared operable before charging restored battery to operable category limits.

lIi Reference Title I 1

HC.OP-AB.ZZ-0149 Learning Objectives OAB149EO01 Recognize abnormal indications/alarms and/or procedural requirements for implementing, 250 VDC System Malfunction, Abnormal Operating Procedure.

0AB149E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of 250 VDC System Malfunction, Abnormal Operating Procedure.

IMaterial Required for Examination

[esWo New ..... ..

lQuestion Modification Method:I Q~uesi orce Comments:

Saturday, March 23, 2002 10:47:45 AM Page 105 of 139

Which one of the following describes the effect on the Class 1E AC Power Distribution System by a loss of Channel "A" 125VDC CLASS 1E Panel 1AD417?

FI Loss of switchgear 10B430 Normal Control Power Iq Loss of switchgear 10B440 Normal Control Power

[c Loss of switchgear 1-0B450 Alternate Control Power F Loss of switchgear 10B460 Alternate Control Power Answer c a L B cognitlvee Memo a Hope Creek Exa. ae: 03/12/2002 Tier: Plant Systems o 21 G 2 263000K201 263000

D.C.

.~~Electrical Distribution

~~~~ ....

~~~~~~~~~~ Record Number

.. 99 K2.1 Knowledge of electrical power supplies to the following:

K2.01 Major D.C. loads 3.1 _3.4 Exlntonof correct answer. 1OB450 is "A" Channel switchgear Alternate control power is fed from 1AD417 Reference Til

!HC.OP-SO.PK-0001 Learning Objectives 000069E019 (R) Given a D.C. electrical load and access to control room reference material, determine the power supply to the load lAW the DC Distribution System Lesson Plan.

MWaterial Required for Examination

[usio ouce Facility Exam Bank Question Modification Method: Significantly Modified IQuetionSource Comments: I Vision Bank QID# Q56195 Saturday, March 23, 2002 10:47:45 AM Page 106 of 139

Given the following:

- The plant is operating at 100 percent power

"- "D" SACS Pump is inoperable for scheduled maintenance

"-"B"Emergency Diesel Generator (EDG) becomes inoperable Which one of the following actions is required within one hour?

Ii Cross-tie m~~~~ the "D" EDG to the "A"..

~~~~~~.

~fr ..SACS Loop iAWIA.C...T.zooiHC.OP-SO.EG-0001 .iii ...

~IPerform AC Po~we~r .Distribut~i.o~n L~ineu~p .- .W.ee~kly lAW HC. P-ST.ZZ-.0001 Perform "B" SACS Pump In-service Test - Quarterly lAW HC.OP-IS.EG-0002 W Perform "A" EDG 0perability SurveilIance Test - Monthly lAW HC.OP-ST.KJ-o0001 S -

[j---w*~r M m r Hope Creek ..... 0311-2/2002 Anwrb ExmLvlS Cognitive Level MemoryFaliyHpCre EamDt: 01202 Tier: Plant Systems .roup-1S 1 26400OG111..

264000 Emergency Generators (Diesel/Jet) e u 100 2.1 iConduct of Operations 2.1.11 Knowledge of less than one hour technical specification action statements for systems. 30 3.8i

[xplaation Tech spec 3.8.1.1 action b. requires surveillance requirement 4.8.1.1.1.a. within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. This surveillance requirement is contained within HC.OP-ST.ZZ-0001 F_ Reference Title

!Tech spec 3.8.1.1.1 10CFR55.43(2)

Learning Objectives 000068E030 (R) Given a scenario of applicable conditions and access to technical specifications:

a. Choose those sections which are applicable to the Emergency Diesel Generators, lAW HCGS Technical Specifications.

b. Assess Emergency Diesel Generator operability and determine required actions associated with Diesel Generator inoperability, lAW HCGS Technical Specifications.

c. Explain the basis for those technical specification items associated with the Diesel Generators, lAW HCGS Technical Specifications. (SRO ONLY)

IMaterial Required for Examination lQuestion Modification Method: I Saturday, March 23, 2002 10:47:45 AM Page 107 of 139

Given the following:

- A discharge of the Equipment Drain Sample Tank is in progress to the River

- The Liquid Radwaste Discharge Isolation Valve to the Cooling Tower Blowdown automatically closes Which one of the following conditions would cause this termination?

(Assume no operator action)

El Cooling Tower Blowdown weir flow rate HI setpoint is reached F1 Liquid Radwaste Effluent sample flow rate HI setpoint is reached F1 Liquid Radwaste Effluent radiation HI setpoint is reached WThe Cooling Tower Blowdown RMS radiation HI setpoint is reached

u*~~t Fee *a~te., --**o2 A.swer c......Cognitive Level Comprehension Faciti Hope Creek E m .ate. 03/12/2002 Tier: Plant Systems . ...... 3 S Go 3 268000A101 268000 R aadwaste Nuber .eco. 101 Al. Ability to predict and/or monitor changes in parameters associated with operating the RADWASTE controls including:

A11.01 Radiation level 12.7 3.1 Epaaiono Of choices given, only Radwaste Effluent Radiation HI setpoint will cause release isolation and termination. Other answer choices cause alarms but not isolation.

Reference Title HC.OP-AR.SP-0001 Attachment 5 Learning Objectives 000086E005 (R) From memory list/identify the five conditions that will cause a liquid release to be automatically terminated, lAW the Lesson Plan.

Material Required for Examination QusioSou H] INPO Exam Bank Modification Method: Iestion

[Editoi Modified ue norce Comments: -INPO EXAM BANK QID# 16367 Grand Gulf 04/0112000 Saturday, March 23, 2002 10:47:45 AM Page 108 of 139

The Off-Gas Pre-Treatment High Radiation alarm on the RM-1 1 has just annunciated. In addition to a fuel element failure, which one of the following could cause the high offgas pre-treatment radiation condition?

El Fire in the offgas holdup pipe FI:Low offgas recombiner temperatures tIncreased Main Condenser air in-leakage El F1 'Condensate demineralizer resin intrusion Answe d..xam.Level.B.Cognitive Level Memory F Hope Creek .. ame Oa. 03/12/2002 Tier: Plant Systems .... ....up. 2 Grou..2 27100.K102 271000 Offgas System Record Number 102 K1. Knowledge of the physical connections and/or cause- effect relationships between OFFGAS SYSTEM and '

the following:

K1.02 P*rocess radiation monitoring system . 3.1 3.i3 Explanaionof Condensate demin resin intrusion into the RPV will cause increased rad levels at the Pre-Treatment Rad monitors. Others affect offgas flows and temperatures. Fire in the holdup pipe is downstream of offgas pretreatment RMS.

Reference Title HC.OP-AB.ZZ-01 00 Learning Objectives OAB100E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of High Reactor Coolant Activity, Abnormal Operating Procedure.

Material Required for Examination INPO Exam Bank Question Modification Method: Significantly Modified m*rents: ] INPO EXAM BANK QID# 6550 Dresden 0 Saturday, March 23, 2002 10:47:45 AM Page 109 of 139

Given the following:

- A plant startup is in progress

- The 'A' RPS Motor-Generator Voltage Regulator fails causing generator output voltage to decrease to approximately 1 0OVAC Which one of the following describes the effect of this condition on the Main Steam Line (MSL)

Radiation Monitors?

1i Power is lost to MSL Radiation Monitors RE-NO06A and RE- NO06C, resulting in an INOP trip

[i Power is lost to MSL Radiation Monitors RE-NO06A and RE-NO06C, resulting in a Hi-HI RAD trip I] The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-NO06A and RE-N006C SThe reduced voltage causes radiation levels for MSL Radiation Monitors RE-NO06A and RE N006B to indicate lower than actual a-Em L B Cognitive LeveIl Memory Facirity Hope Creek E-xam-Date: 03/12/2002 Tier: Plant Systems .oGrouIp -2. SRouI 2 272000K603 272000 Radiation Monitoring System Record Numbe 103 K6. Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORING SYSTEM:

K6.03 iAC. power 2.8 3..0 of Power is lost to MSL Radiation Monitors RE-NO06A and RE-N006C, resulting in an INOP trip.

Answer Correct - when 'A' RPS MG output is less than 108 VAC, the EPA Breakers on the MG output to the 'A' RPS Bus trip on undervoltage, causing a loss of the 'A' RPS bus. This results in an INOP trip of MSL Rad Monitors RE-NO06A & C since they are powered from RPS Bus 'A'.

Power is lost to MSL Radiation Monitors RE-NO06A and RE-NO06C, resulting in a HI-HI RAD trip.

Incorrect - an INOP trip occurs on a loss of power to the MSL Rad Monitors.

The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-NO06A and RE-NO06C.

Incorrect - any voltage reduction would be momentary due to the UV trip of the EPA Breakers; an INOP trip occurs on a loss of power to the MSL Rad Monitors.

The reduced voltage causes radiation levels for MSL Radiation Monitors RE-NO06A and RE-NO06B to indicate lower than actual.

Incorrect - any voltage reduction would be momentary due to the UV trip of the EPA Breakers; an INOP trip occurs on a loss of power to the MSL Rad Monitors.

Reference Title HC.OP-SO.SB-0001 HC.OP-SO.SP-0001 Learning Objectives 000221 E002 (R) Regarding the main steam line Radiation Monitoring System:

a. From memory, explain the setpoints/conditions associated with a high-high radiation or inoperative trip lAW the Radiation Monitoring System Lesson Plan.

b. Given normal Control Room references, determine the automatic plant actuations/trips which occur as a result of a high high radiation or inoperative trip lAW the Radiation Monitoring System Lesson Plan.

c. From memory, evaluate the effect of a loss of RPS power lAW the Radiation Monitoring System Lesson Plan.

Saturday, March 23, 2002 10:47:46 AM Page 110 of 139

Facility Exam Bank iQuestion

odif::ic**i Method: !Editorially Modified rnments: I VISION BANK QID# Q56950 Saturday, March 23, 2002 10:47:46 AM Page 111 of 139

Given the following:

- The plant is in Operational Condition 2 with a startup in progress

- Reactor pressure is 300 psig

- The lowest Reactor Vessel Metal Temperature thermocouple is reading 150°F Which one of the following actions is required?

(Use Technical Specification Figure 3.4.6.1-3 provided)

a. Hold reactor pressure at current value for at least 30 minutes I Raise reactor pressure at least 20 psig within the ne xt 30 minutes WLower reactor pressure at least 20 psig within the next 30 minutes WRaise reactor metal temperature a maximum of 20 0 F within thie next 30 minutes

- *r- ....

c RamLeU

Iuve LeelIv Application It'aciWtl IHope Creek ixam Date: 03/12/2002 Tier: Plant Systems. .. roup ---3 Gro .. 3 290002A202 290002 Reactor Vessel Internals Record Number 104 A2. Ability to (a) predict the impacts of the following on the REACTOR VESSEL INTERNALS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Overpressurization transient 3.6 3.91 Given conditions place the lowest metal temp to the left of the curve in figure 3.4.6.1-3. Lowering reactor pressure moves the operating plot to the right side of the curve. Metal temp must move 37 DegF to get to the right side of the curve Reference Title Tech spec 3.4.6.1 figure 3.4.6.1-3 Learning Objectives 001 12CE006 (R) Analyze plant conditions and parameters to determine if plant operation is in accordance with the STARTUP FROM COLD SHUTDOWN TO RATED POWER Integrated Operating Procedure, supporting System Operating Procedures and Technical Specifications.

aterial Required for Examinai Tech spec 3.4.6.1 figure 3.4.6.1-3 IM Question Sou rce I New Question Modification Method:

Saturday, March 23, 2002 10:47:46 AM Page 112 of 139

The plant is operating at 100 percent power Which one of the following describes the effect on the plant if a piece of foreign material blocked a fuel support piece flow orifice?

F Core thermal power would decrease SSteam quality exiting the reactor vessel will decrease SJet pump net positive suction head would increase PIindicated reactor water level will fluctuate

..... r.a.......e....R..o.nitive..e Memor .. . f. cility Hope creek m 03/12/2002 Tier: . P ant Systems 3..G 3 290002K303 290002 Reactor Vessel Internals ]! 105 K3. Knowledge of the effect that a loss or malfunction of the REACTOR VESSEL INTERNALS will have on following:

K3.03 Reactor power 3133.4

[xplantion of Low reactor coolant flow past a bundle will drastically increase voids in the channel. Reactor power will S..........................................

d ec rea s e.

I Reference Title LP 0301-000.OOH-000001-12 Learning Objectives 000228E024 Given a reactor power change analyze that power change and predict how the various reactivity coefficients respond.

000001EO08 (R) From memory, explain the reason for core orificing and how this is accomplished, lAW the Lesson Plan.

000001EO09 (R) Given plant problems/industry events associated with the Reactor Vessel and Internals:

a. Discuss the root cause of the plant problem/industry event lAW the plant/industry event.

b. Discuss the HCGS design and/or procedural guidelines that mitigate/reduce the likelihood of the problem/industry event at HCGS lAW the plant/industry event.

c. Discuss the "lessons learned" from this problem/event lAW the plant/industry event.

Material Required for Examination Question Source:] New IQuestion Modification Method:-

I.... ... . / / / / / . / / : ': . ; / :: i ¸ / '.............. .. II Saturday, March 23, 2002 10:47:46 AM Page 113 of 139

Given the following:

- The plant is operating at 100% power

"- "A" Control Room HVAC train and Chilled Water system is running

- A light haze with an acrid odor is noticed in the Main Control Room

- No alarms are received that could explain the origin of the haze and odor

- HC.OP-AB.ZZ-0129, High Radiation, Smoke or Toxic Gases in the Control Room Air Supply is entered Based on plant conditions, which one of the following is an immediate action lAW HC.OP-AB.ZZ 0129?

Verify that the C otrol Room Supply V ent ilation has automatically isolated V]

verify that the "A" Control Room Emergency FiIter Unit automatically started Vi El Press the CONTROL ROOM EMER FILTER UNIT A and BO*A pushbuttons E Press the CONTROL ROOM EMER FILTER UNIT A and B RECIRC MODE pushbuttons d EB Cognitive Leve Memory a Hope Creek E.mDae 03/12/2002 Tier* Plant Systems .u..2 R[.. R o 2- 290003K501 290003 Control Room HVAC Record Number 106 K5. Knowledge of the operational implications of the following concepts as they apply to CONTROL ROOM HVAC K5.01 ýAirborne contamination (e.g., radiological, toxic gas, smoke) control 3.2 3.5

[Explnationof Press the CONTROL ROOM EMER FILTER UNIT A and B RECIRC MODE pushbuttons. For a toxic gas in the Control Room Supply, isolate Control Room Ventilation and place CREF in the Recirc Mode.

INCORRECT - Press the CONTROL ROOM EMER FILTER UNIT A and B OA pushbuttons. CREF must be in the Recirc Mode for a toxic gas event.

INCORRECT - Verify that the Control Room Supply Ventilation has automatically isolated. Toxic gas will not automatically isolate Control Room Ventilation. Only high rad.

INCORRECT -Verify that the "A" Control Room Emergency Filter Unit automatically started. Does not automatically start on toxic gas, only high rads.

Reference Title HC.OP-AB.ZZ-0129 Lea rning Objectives 0AB129E002 (R) From memory, recall the Immediate Operator Actions for High Radiation, Smoke or Toxic Gases in the Contorl Room Air Supply, Abnormal Operating Procedure.

IMaterial Required for Examination

[uston oure Facility Exam Bank lQuestion Modification Method: Significantly Modified

[Question source Comments: I VISION BANK QID# Q61261 Saturday, March 23, 2002 10:47:46 AM Page 114 of 139

Given the following:

- The Oncoming Day Shift Reactor Operator (RO) is returning to shift after 4 days vacation

- Today is March 18, 2002 Which one of the following identifies the date of the earliest Control Room Narrative Log the RO is required to review PRIOR to assuming the watch today?

IMarch 10, 2002 IMarch 13, 2002 M...March 14, 2002

[iMarch 15, 2002 An..er.d Exm Lev.eI B cognitive Level Memory Fa...t IHope Creek 03/12/2002 Tier: Generic Knowledge and Abilities R r...

-. 1 R Gu 1 294001G103 GENERIC Record Number 107 22.1 Conduct of Operations 2.1.3 'Knowledge of shift turnover practices. 3.0 3.4

[xlnatoof The RO must review 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to assuming watch because it is shorter than the 4 days of vacation.

The RO must then review the previous 5 days after turnover. Distractors are based on different combinations of these requirements.

Reference Title SH.OP-AP.ZZ-0107 Section 5.3.1 Learning Objectives 000113E1 00 Summarize six items covered in a minimum Shift Turnover.

IMaterial Required for Examination IQlo*osion rce] Facility Exam Bank .Question Modification Method: Significantly Modified Question Source Comments:. INPO BANK QlD # 15027 Salem 2/22199 Saturday, March 23, 2002 10:47:46 AM Page 115 of 139

Given the following:

- The plant is in Operational Condition 5

- Core offload is in progress

- A spent fuel bundle is full up on the main hoist over the core

- The refuel bridge spotter notices the fuel bundle has unlatched and fallen free into the vessel What operator action is required?

1] Determine South Plant Vent RMS release rate.....

SDetermine the location of the dropped bundIe and inform the Reactor Engineer 1] Re-establish Seconda0ryContainment within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SEvacuate all unnecessary personnel from the Reactor Building Answer-d EmLevel; S CogitiveLevel Memory Creek FaityHope ........ CameDate x 03/12/2002 Tier: Generic Knowledge and Abilities R. Group 1SRO Grp 1 294001 G114 GENERIC Record Number 108 2.1 Conduct of Operations 2.1.14 Knowledge of system status criteria which require the notification of plant personnel. 2.513.3 Eof Subsequent operator action 4.2 of AB-1 01 AJustification: lAW HC.OP-AB.ZZ-0101 section 4.0

- Evacuate all unnecessary personnel from the Reactor Building.-Correct- lAW HC.OP-AB.ZZ-0101 step 4.2

- Determine South Plant Vent RMS release rate-Incorrect- if Rx Bldg or RF Floor rad levels are rising the FRVS system is placed in service which does not exhaust though the South Plant vent see step 4.5

- Re-establish Secondary Containment within 1 hour- Incorrect- Secondary containment is required to be in place during all fuel moves there is no time limit if lost, actions are to suspend irradiated fuel moves, CORE ALTERATIONS and operations with the potential for draining the vessel. See step 4.3 -Determine the location of the dropped bundle and inform the SRO-Incorrect- there are no actions to determine the location F_ Reference Title HC.OP-AB.ZZ-0101 section 4.0 10CFR55.43(5)(7)

Learning Objectives OAB1O1EO06 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Irradiated Fuel Damage, Abnormal Operating Procedure.

IMaterial Required for Examinato

[uson ourc Facility Exam Bank .uest1.ion Modification Method: Editorially Modified

[Question Source comments: Vision QID# Q60871 Saturday, March 23, 2002 10:47:46 AM Page 116 of 139

Using provided copy of P&ID M-51 Sheet 2, determine the computer point ID for "A" RHR Heat Exchanger Outlet Temperature.

SiIA2020 SIA2380 SA2381

[ A3132 Anw C. B Cognitive Level Comprehension F [Hope Creek Exam Date . 03/1 2/2002 Tier: Generic Knowledge and Abilities 1 R G . 294001G1 24 GENERIC] Rcd Number, 109 2.1 Conduct of Operations 2.1.24 Ability to obtain and interpret station electrical and mechanical drawings. 2.8 3.1

[~iaation A2020 HX Outlet Conductivity - Incorrect

[nwer A2380 HX Inlet Temperature - Incorrect A2381 HX Outlet Temperature - Correct A3132 HX Outlet Flow - Incorrect IRefteresice Title M-51 Sheet 2 Learning Objectives 000028E012 Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, assess the status of the Residual Heat Removal System or its components by evaluation of the controls/instrumentation/alarms lAW the RHR System Lesson Plan.

[Material Required for Examination l~u reeonSoi]cý New Que stion Modification Mehd lQuestion source comments:

Saturday, March 23, 2002 10:47:47 AM Page 117 of 139

Given the following:

- The reactor is operating at 75% power following a transient

- The reactor engineer reports that the MAXIMUM FRACTION OF LIMITING CRITICAL POWER RATIO (MFLCPR) is 1.001 Which one of the following describes the Technical Specifications required action(s)?

1] The reactor must be in H0T SHUTDOWN within two hours and the N RC notified within one hour.

II The reactor must be in STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

[ Corrective action be initiated within 15 minutes and the MCPR restored to within the Iimit within two hours or reduce thermal power to less than 25% of rated within the next four hours.

'An immediate reactor scram by placing the-Reactor Mode Switch in the SHUTDOWN position.

Answe r c Eam Lvl5 CognitiveLeveI Application a iHope Creek E Dat.: 03/12/2002 Tier: LGeneric Knowledge and Abilities R GroupI SRO Group 1 294001G133 GENERIC Record Number 110 2.1 Conduct of Operations 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for 3.4 4. 0 technical specifications.

[xplanion TECH SPEC 2.1.2 1 TECH SPEC 6.7.1 Justification:

The reactor to be in HOT SHUTDOWN within two hours and the NRC operations center notified as soon as possible and in all cases within one hour-Incorrect- exceed SL 2.1.2 MCPR limit of >1.10 MCPR, requires Hot shutdown within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 6.7.1 .a requires 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification No operator action since reactor pressure is greater than 785 psig and core flow is greater than 10% of rated flow.-1ncorrect- SL 2.1.2 exceed Corrective action be initiated within 15 minutes and the MCPR restored to within the limit within two hours or reduce thermal power to less than 25% of rated within the next four hours-Correct- A MFLCPR Value of 1.001 indicates the CPR. in the core is slightly exceeding the LCO Limit but below the SL. This is the action for MCPR Thermal Limit exceeding Tech Spec limit. The Safety Limit is not violated.

An immediate reactor scram by placing the Reactor Mode Switch in the SHUTDOWN position.-Incorrect Hot shutdown in 4 hrs does not requires immediate MSS to Shutdown Reference Title TS 2.1.2 10CFR55.43(2)

Learning Objectives 000110E008 (R) Given specific plant operating conditions and a copy of the Hope Creek Generating Station Technical Specifications, evaluate plant/system operability and determine required actions (if any) to be taken. (SRO Only)

Material Required for Examina Tech Specs without Definitions, Safety Limits, and bases lQuestion source: Facility Exam Bank Question Modification Method: ISignificantly Modified I............

...I Saturday, March 23, 2002 10:47:47 AM Page 118 of 139

oQurscn IComments Vision Bank QID# Q54864 Saturday, March 23, 2002 10:47:47 AM VPage 119 of 139

Given the following:

- The plant is operating at 90 percent power

- The #1 Main Turbine Stop Valve has slowly drifted closed

- All Turbine Bypass valves responded full open

- Reactor steam dome pressure stabilizes at 1025 psig

- All other equipment functions properly Which one of the following actions is required by Technical Specifications?

Fal Re-open the Turbine Stop Valve within one hour I Reduce reactor thermal power by at least 25 percent within 15 minutes SReduce reactor steam dome pressure by at least 6 psig within 15 minutes W Determine MCPR is less than or equal to the EOC-RPT inoperable limit within one hour An.wer.c.ExamLel Coiteve Application Hope Creek _..xam D.at.e: 03/12/2002 Tie:, Generic Knowledge and Abilities R..Group 1 S...G...p 294001G-133 GENERIC Record Number..1 2.1 Conduct of Operations 2.1.33. Ability to recognize indications for system operating parameters which are entry-level conditions for 3.4 4.0

ýtechnical specifications.

Explanation of Reactor steam dome pressure is above the LCO limit of 1020 psig. Reduce pressure to less than 1020 psig within 15 minutes or be in at least hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Reference Title HC Tech Specs 3.4.6.2 Learning Objectives 000051E017 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Select those sections applicable to the EHC Control Logic System, lAW HCGS Technical Specifications.

b. Evaluate EHC Control Logic System operability and determine required actions based upon system inoperability, lAW HCGS Technical Specifications.

c. Explain the bases for those Technical Specification sections associated with the EHC Logic System, lAW HCGS Technical Specifications. (SRO ONLY)

Material Required for Examination J Tech Specs without Definitions, Safety Limits, and bases lQuestion Source: I New Question Modification Method:

Saturday, March 23, 2002 10:47:47 AM Page 120 of 139

Given the following:

- A plant condition has resulted in a reactor power reduction

- Reactor power is now stable at 50% after the transient

- Chemistry reports that DOSE EQUIVALENT 1-131 is 3.0 microcuries/gram Which one of the following describes the bases that allows plant operation to continue for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months lAW Technical Specifications?

ITo allow for possible Iodine spiking phenomenon ITo allow for stable Reactor Coolant chemistry sample data ITo allow for decay of short lived isotopes ITo allow reasonable time to verify the initial s ample results Answe a Eam- Leve Level

.Cognitive Memory FHope Creek E Date: 03/12/2002 Tier: Generic Knowledge and Abilities R..Group.1.....Group.294001G134 GENERIC Recor Number' 112 2.1 Conduct of Operations 2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits. 2.3 2.9 ioi TS BASES 3.4.5 Allows for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months with a limit of 4 microcuries/gram to accommodate possible iodine spiking which may occur following changes in Thermal Power Reference Title Tech Spec bases 3/4.4.5 10CFR55.43(2)

Learning Objectives 000220E006 (R) Given a copy of the Chemistry Daily Summary, a scenario of applicable operating conditions and access to Technical Specifications:

a. Identify the sections, which are applicable to Chemistry Control lAW Technical Specifications. (SRO / STA only)

b. Evaluate the status of the applicable LCOs and summarize the actions required lAW Technical specifications.

c. Explain the bases for those Technical specification sections associated with Chemistry Control lAW Technical Specifications.

IMaterial Required for Examination I Bank IQuestion Modification Method: Significantly Modified INPO BANK QID# 1573 Palo Verde 03/24/1997 Saturday, March 23, 2002 10:47:47 AM Page 121 of 139

Given the following:

- Reactor power is 40%

- ALL Turbine Bypass Valves fail OPEN

- The MSIVs FAIL to automatically close Which one of the following combinations of reactor power and reactor pressure would indicate that a Safety Limit violation occurred?

II Reactor power is 10% and RPV pressure is 750 psig

[I Reactor power is 20% and RPV pressure is 770 psig

[c Reactor power is 30% and RPV pressure is 775 psig R1 Reactor power is 35% and RPV pressure is 810 psig Answer c EmLevel R _Cognitive Lev c rehension Hope Creek Eam IDate:t 03/12/2002 Tier: Generic Knowledge and Abilities R. GroI -- SRO rol 1 294001G222 GENERIC Rcr 113 2.2 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits. 3. 41411 Explanation of Reactor thermal power is greater than 25% with reactor pressure less than 800 psia Answer Reference Title Tech Spec 2.1.1 Learning Objectives 00011OE001 (R) From memory, state the four (4) Safety Limits in terms of conditions.

Material Required for Examination QustonSourE] INPO Exam Bank .Question Modification Method] Direct From Source I Source Comments: 1 INPO BANK QID# 6303 Dresden 2 09/26/1998 Saturday, March 23, 2002 10:47:47 AM Page 122 of 139

The reactor is operating at 100% power. During an operability check of the RCIC system it is discovered that the flow controller FIC-600 on 10C650B will NOT regulate RCIC flow in automatic, however, manual control does function properly.

Based on plant conditions, which one of the following actions is required?

If No action is required since RCIC flow can be manually controlled L-1Restore the controller to operable status within 7 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and have steam dome pressure less than 150 psig in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months L] Restore the controller to operable status within 14 d ays, or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and have steam dome pressure less than 150 psig in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W Restore the controller to operable status within 14 days, or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and have steam dome pressure less than 100 psig in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months A c mCognitive Level-Application Hope Creek Examioate 03/12/2002 Tier: Generic Knowiedge and Abilities. . . Group 1 R 1 294001 G222 GENERIC Record Number 114 2.2 Equipment Control 2.2.22 'Knowledge of limiting conditions for operations and safety limits . 3.4 4.1 Justification:

Jof T/S [amendment 126] 3.7.4 action with RCIC operable is a 14 day LCO with HOT SHUTDOWN within the next 12 hrs and reactor steam dome pressure is 150 psig within the following 24 hrs.

Reference Title T/S [amendment 126] 3.7.4 action S10CFR55.43(2)

Learning Objectives 000030E013 (R) Given plant conditions and access to Technical Specifications:

a. Select those sections which are applicable to the ROIC System, IAW HCGS Technical Specifications.

b. Evaluate RCIC System operability and determine required actions based upon system inoperability, IAW HCGS Technical Specifications. (SRO Only)

c. Explain the bases for those Technical Specification items associated with the ROIC System, IAW HOGS Technical Specifications.

Material Required for Examination QusinSu _ Facility Exam Bank---_ -- Question Modification Method: rect From Source QuinSrce Comments: I Vision Exam Bank QID# Q54171 Saturday, March 23, 2002 10:47:47 AM Page 123 of 139

Given the following:

- A complete core offload was completed at the beginning of the refueling outage

- Fuel reload is ready to commence lAW "Fuel Handling Control" Core Alteration Forms. [HC.OP FR.ZZ-0001]

- All SRM's are fully inserted with the following count rates:

"-"A"-5 cps

"- "B"- 2 cps

"-"C"- 6 cps

"- "D"- 1 cps Based on these conditions, which of the following actions is required lAW plant procedures?_

[-] Spiral Reload may commence witih no restrictions as long as any tw o SRM's are reading >

3 cps El A Movable SRM detector must be hooked up to the normaI SRM channel instrumentation and be placed in either "B" or "D" quadrant, indicating > 3 CPS prior to Spiral fuel reload commencement

[] spiral fuel reload may commence in "A"and "C" quadrants only, until either "B" or "D" quadrant SRM is reading > 3 cps at which time complete reload may be commenced E Spiral fuel re load may commence up to the first 16 bundles, at which time all four SRM's must read > 3 cps to perform a complete reload Aw *era Le 5 cognitive Level Application Facility Hope Creek Exam Datel 03/112/2002 Tier: Generic Knowledge and Abilities RO roup 1 SROGroup 1 294001G226 GENERIC RecordNumber 115

-2.2 Equipment Control -

2.2.26 Knowledge of refueling administrative requirements . 2.513.7 Epaationf Justifi-cation: IAW HC.OP-lOZZ-0009 step 5.2.10 directs toverify SRM counts > 3CPS after first 16 Anwer:

bundles when performing Spiral reload. This is to ensure compliance T.S.3.9.2.e.

Reference Title T.S.3.9.2.e.

HC.OP-IO.ZZ-0009 step 5.2.10 10CFR55.43(6)

Learning Objectives 001121E006 (R) Analyze plant conditions and parameters to determine if plant operation is in accordance with the REFUELING OPERATIONS Integrated Operating Procedure, supporting System Operating Procedures and Technical Specifications IMaterial Required for Examination Tech Specs without Definitions, Safety Limits, and bases ton Source: Facility Exam Bank Question Modification Method: Editorially Modified Question Source Comments: Vision Exam Bank QID# Q58929 Saturday, March 23, 2002 10:47:48 AM Page 124 of 139

Given the following:

- Tech Spec compliance has been verified lAW "Refueling Operations". [HC.OP-IO.ZZ-0009]

- Multiple Control Rod Drive Mechanisms are being removed lAW Technical Specification 3.9.10.2

- Spiral Fuel offload is in progress per directions of Reactor Engineers and Fuel Handling Control Core Alteration forms. [HC.RE-FR.ZZ-0001]

- 14 Fuel Assemblies are remaining in the Vessel Which one of the following conditions would require a formal declaration of Suspension of Core Alterations as described in plant procedures?

an Spent Fuel Storage Area Radiation Monitor in alarm while transporting LPRMS through -the Cattle Shute SRMs indicate between 2.1 & 2.6 cps I] Mode Switch position change from Shutdown to RefueI for Rod Speed adjustments per system operating procedure Bridge Platform surveillance identifies Frame Mounted hoist up travel stops are out of Technical Specification tolerance SR cognitive Leve Application [Facilty Hope Creek 03/12/2002, Tier: Generic Knowledge and Abilities .......... ou......G2271 GENERIC Recor*d Nu-mber 116 2.2 Equipment Control 2.2.27 Knowledge of the refueling process. 2.6 13.5

[xplaaionof Justification Ane HC.OP-IO.ZZ-0009, directs use of NC.NA-AP.ZZ-0049, for direction on formal suspension of fuel handling activities, adverse radiological conditions are one of the criteria.

Additionally, Refuel Radiation Area Alarms is an entry condition for HC.OP-AB.ZZ-0101 "Irradiated Fuel Damage" which directs suspension of all refueling operations.

Other choices are all within the Allowable Technical Specification boundaries for Core Alterations.

Reference Title iNC. NA-AP.ZZ-0049 Learning Objectives 001121E004 I (R) Apply Precautions, Limitations and Notes whlie executing the REFUELING OPERATIONS Integrated Operating Procedure IMaterial Required for Examination Sou.:

onsSo Facility Exam Bank Question Modification Method: Editorially Modified Quetio Sorce Comments: J1 Vision Bank QID# Q58930 Saturday, March 23, 2002 10:47:48 AM ,Page 125 of 139

Given the following:

- The plant has just completed a shutdown for refueling

- Vessel disassembly has commenced

- The I&C department has determined that IRM "A" and SRM "B" have bad detectors and are inoperable Which one of the following actions must be completed prior to full core offload?

Ii Shutdown margin must be demonstrated II SRM B will have to be replaced so offload can occur in that quadrant FrI Both instruments must be repIaced before any core alterations can begin W IRM A must be restored in order to meet t he rnminimum operable channeI requirements Answer Ibxam-[eve*l Cognitive Lev Co rehension aility Hope Creek Ixm e..- -- 03/12/2-002 Tier: Generic Knowledge and Abilities ROGop iSOGop 1294001 G227 GENRICRecord Number 117 2.2 Equipment Control 2.2.27 Knowledge of the refueling process. 2.613.5 Explaniono Core alterations may only be conducted in a quadrant with an operable SRM detector A n s w er Refe re nce Title I Tech Spec 3.9.2 NC. NA-AP.ZZ-0049 10CFR55.43(6)

Learning Objectives 000113E071 a. State the responsibilities of the following personnel:

Refueling SRO.(SRO ONLY)

Refueling Bridge Operator Control Room refuel Monitor material Required for Examination Tech Specs without Definitions, Safety Limits, and bases t[usiourceýI] INPO Exam Bank Question Modification Method: Significantly Modified-IQuestion Source Comments: --] INPO EXAM BANK QID# 16846 Quad Cities 03/19/1998 Saturday, March 23, 2002 10:47:48 AM Page 126 of 139

Given the following:

- The Core has been off-loaded to the Fuel Pool per NC.NA-AP.ZZ-0049, Conduct of Fuel Handling

- Five control rods are to be replaced with new ABB rods

- All plant conditions have been met for the control rod replacement In addition to the Refuel Bridge operator and RP Technician, which one of the following must be part of the minimum crew compliment required for the replacement of the control rods?

I] Refueling SRO - Only required to be on site

[i Refueling SRO - Required on the Refuel Floor IReactor Engineer - Required on the Refuel Floor PI Reactor Engineer - Only required to be on site c R cognitive Level mo Fait [Hope Creek Exa Date:* 03/12/2002 Tier: Generic Knowledge and Abilities R..Group 1 SOGoupC 294001G230 GENERIC Record Number 118 2.2 : Equipment Control 2.2.30 Knowledge of new and spent fuel movement procedures. .2.6 3.5:

Eof JUSTIFICATION:

With no fuel in the vessel, no component manipulation within the vessel is considered a Core Alteration JAW Technical Specifications 1.7, and NC.NA-AP.ZZ-0049 sections 5.1.2.A & 7.1 NC.NA-AP.ZZ-0049 stipulates that the minimum crew for non-core alteration fuel handling activities includes the Fuel Crane Operator, Radiation Protection Technician, Reactor Engineer and Spotter. The Reactor Engineer may fulfill the duties of the spotter; hence the minimum permissible crew is three.

Reference Title NC. NA.-AP.ZZ-0049 Learning Objectives 000113E073 State the minimum fuel handling crew requirement for non-core alteration non irradiated fuel handling.

Question Modification Method: Direct From.Source Vision Exam Bank QID# Q61057 Saturday, March 23, 2002 10:47:48 AM Page 127 of 139

Given the following:

- The plant is in Operational Condition 5

- You are the oncoming Refueling SRO

- The offgoing SRO briefs you of their activities Which one of the following would constitute a violation of Refuel SRO duties while core alterations are IN PROGRESS?

IPicking up a fuel bundle after Control Room communications are lost S5 hours of continuous fuel moves tc] Control rod blade removal from an unloaded fuel ceIll.....

L-1 Fuel movement with Fuel Pool water level 1 inch below wave scuppers ra a Le Memory MCognitiveLevel Facility Hope Creek a Date: 03/1 2/2002 Tier: Generic Knowledge and Abilities . FRO GrIoupi*

1 i Group i 1 294001G231 GENERIC Record Number 119 2.2 !'Equipment Control 2.2.31 Knowledge of SRO fuel handling responsibilities. 1.613.8

[iEpaaion Continuous communications must be established with the main control room. Core Alts must be suspended if continuous comms lost.

F- Reference Title

ýNC.NA-AP.ZZ-0049

ý10CFR55.43(7)

Learning Objectives 000113E071 a. State the responsibilities of the following personnel:

Refueling SRO.(SRO ONLY)

Refueling Bridge Operator 1 Control Room refuel Monitor Material Required for Examination ionSoue] Facility Exam Bank..Question Modification Method- Direct From Source

[duestion Source Comments: VISION BANK QID# Q56513 Saturday, March 23, 2002 10:47:49 AM Page 128 of 139

An operator has the following exposure history this year until today:

Deep Dose Equivalent (DDE) 210 mrem Committed Effective Dose Equivalent (CEDE) 45 mrem Shallow Dose Equivalent (SDE) 33 mrem Today, the operator was required to make two entries into the Drywell at 5 percent reactor power:

Entry 1: Gamma dose: 52 mrem; Neutron dose: 24 mrem Entry 2: Gamma dose: 124 mrem; Neutron dose: 54 mrem How much radiation exposure is available to the operator without extension if he has to make additional entries?

His available Non-Emergency margin for the year is...

1488 mrem m.. ..

Ii6 1521 mrem I1599 mrem S1712 mrem Anwrb xam LeelB cognitive Level] Comprehension Faci ity Hope Creek Ea ae 03/12/2002 Tier: Generic Knowledge and Abilities Roup 1S1RO Group 294001 G301 GENERIC Record Number 120 2.3 Radiological Controls 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 3.0 E of CORRECT ANSWER. Gamma and neutron dose are summed for DDE. DDE and CEDE are summed together to obtain TEDE. The Dose limit without extension is 2000 mrem/year TEDE Reference Title NC. NA-AP.ZZ-0024 Learning Objectives 000113E059 I a. Identify the personnel responsible for approval of the following dose extension:

Yearly Dose Extension Declared Pregnant Women Dose Extension Lifetime Dose Extension IMaterial Required for Examination I lQuestion Source: I INPO Exam Bank Question Modification Method: I Significantly Modified INPO EXAM BANK QUESTION ID#3324. Braidwood 1 09/14/1998 Saturday, March 23, 2002 10:47:49 AM Page 129 of 139

Given the following:

- A LPCI manual injection valve with remote indication requires an Independent Verification (IV)

- The valve is located 8 feet above the grating

- The valve is located in a 90 mrem/hr radiation area

- The temperature in the area is 90 F

- It is estimated that an individual will take 10 minutes to conduct the IV locally Based on these conditions, which one of the following describes when the "Hands On" IV requirement can be waived?

I] For climbing on equipment concerns I] For ALARA concerns C] For heat stress concerns Ed For fall protection concerns Aw b aLeR Cognitive Leve 1 Comprehension Hopecreek- xa a 03/12/2002 Tier: Generic Knowledge and Abilities R 294001G302 GENERIC Ro mb121 2.3 Radiological Controls 2.3.2 Knowledge of facility ALARA program. 2.5 2.91

[xplanionof lAW NC.NA-AP.ZZ-0005, Attachment 6, section 1.4 areas in which Independent Verification would receive in excess of 10 mrem dose can be verified by alternate means such as status/position indicators.

Distrators are based on safety concerns which would make the job longer or more difficult but are not allowed to waive Hands On IV.

Reference Title NC.NA-AP.ZZ-0005 Section 1.4 Learning Objectives 0001 13E01 5 Determine the require ments for Independent Vrfcto IMaterial Required for Examination Quston ource INPO Exam Bank QTuestion Modiication Method: Editorially Modified QustonSorce Comments: I NPO EXAM BANK QID# 11424 LaSalie 1 04/21/1 997 Saturday, March 23, 2002 10:47:49 AM Page 130 of 139

Given the following:

- The plant is in Operational Condition 4 for a short outage

- During a Drywell inspection, the operator notices some radiation barricade ropes in the area of RWCU Isolation valve BG-HV-F001

- A radiation sign on the ropes reads "Caution; High Radiation Area, RWP Required For Entry" and indicates a MAXIMUM radiation level of 1.10 Rem/hr inside the ropes Which one of the following additional posting requirements and /or controls are required for this area according to Technical Specifications?

[i The area requires a flashing light in the immediate area as a warning device I] The area is required to be fenced off and the Drywel Airlock shall be kept locked with the keys kept under the administrative control of the Operations Superintendent C. The area should be posted as a Very High Radiation Area with continuous electronic surveillance used to control access TlThe area requires a closed circuit TV monitor be installed to give radiation protection personnel continuous monitoring capability A ae Exam LevelS------ cognitive Lee Application Hope Creek ExamD-ate 03/12/2002

)

Tier: iGeneric Knowledge and Abilities R..Group 1 R G 294001 G304 GENERIC Record Number 122 2.3 Radiological Controls 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in 2.513.1 excess of those authorized.

E of TS 6.12 requires the area roped off, conspicuously posted and a flashing warning light.

F- Reference Title Tech Specs 6.12.2 10CFR55.43(4)

Learning Objectives 000113E057 a. State the definition of the following terms:

Contaminated Area High Radiation Area Locked High Radiation Area Radiation Area Restricted Area Very High Radiation Area Airborne Radioactivity Area Declared Pregnant Woman (DPW)

Total Effective Dose Equivalent (TEDE)

Material Required for Examination Tech Specs without Definitions, Safety Limits, and bases Isoure INPO Exam Bank Question Modification Method: I Editorially Modified IQiustion Source Comments: I iNPO EXAM BANK QID# 5484 Salem Unit 07/08/1996 Saturday, March 23, 2002 10:47:49 AM Page 131 of 139

Given the following:

- The plant is in Operational Condition 3 - Hot Shutdown, going to Cold Shutdown

- The reason for shutdown was excessive unidentified RCS leakage

- Reactor pressure is 920 psig

- Drywell Oxygen concentration is 2.5%

- Primary Containment Release permit has been obtained Which one of the following is required prior to purging the Primary Containment?

SA Drywell walkdown must be completed SA Valve Open Time permit must be initiatedd E The plant must be in Operationa ICondition 4 - Cold Sh utdown I Primary Containment Airlock Operabiiity Test must be performed Anserb am Level R- CognitiveLe-el .Memory Faiity Hope Creek Exam Datl 03/12/2002 Tier: Generic Knowledge and Abilities . . i ru i_ RO Gr-u 294001 G309 GENERIC Record Number 123 2.3... Radiological Controls 2.3.9 Knowledge of the process for performing a containment purge. 2.5 3.4 correct answer. A Valve open time permit must be prepared to track # of hours that purge valves are open.

Reference Title HC.OP-SO-GS.O001 HC.OP-AP.ZZ-0104 Learning Objectives 000032E015 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Select those sections which are applicable to the Containment Inerting and Purge System lAW the Lesson Plan.

b. Evaluate Containment Inerting and Purge System operability and determine required actions based upon system inoperability lAW the Lesson Plan. (SRO Only)

c. Explain the bases for those Technical Specification sections associated with the Containment Inerting a Purge System lAW the Lesson Plan.

Material Required for Examina n Tech Specs without Definitions, Safety Limits, and bases QuestionSouc Facility Exam Bank Question Modification Method: I IEditorially Modified

ýQusion ource Comments: Vision Bank QID# Q55956 Saturday, March 23, 2002 10:47:49 AM Page 132 of 139

Given the following:

Off Gas Radiation 9RX612 and 9RX622 parameters indicate yellow on the RM-1 1 terminal Chemistry has been directed to commence sampling Based on plant conditions, power level should be lowered ...

auntil the GAS RADW CHAR TRTMT PNL 00C367 alarm is clear.

to maintain Main Steam Line Rad Monitors less than the HIGH alarm setpoint.

iuntil North Plant Vent activity less t han tihe HIGH alarm setpoint.

P] to maintain .~Off~Gas ~ activity

~ ~ less ve~~ than ~....

the .RM-1

...............1 ALERT alarm setpoint e d [xamLe`vel JcognitiveLvel Memor i Hope Creek [e 03/1212002 Tier: Generic Knowledge and Abilities ROGRoup 1_ G 1 294001G310 GENERIC Rcord Number 124 2.3 Radiological Controls 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel 2.9 13.3 exposure.

Ea on Justification:

ASubsequent operator actions for Offgas system High radiation. Reduce reactor power as necessary to maintain the Offgas activity less than the alert alarm setpoint. Correct Reduce Power to maintain Main Steam Line Rad Monitors less than the high alarm setpoint.-lncorrect Main Steam Rad not listed in this procedure Reduce power until the GAS RADW CHAR TRTMT PNL 00367 alarm is clear-lncorrect-RM-1 1 is the entry not the Charcoal treatment.

Reduce Power to maintain North Plant Vent activity less than the high alarm setpoint. Incorrect-Off Gas RM-1 1 is the entry not the NPV.

I Reference Title HC.OP-AB.ZZ-0127 rev 5, Section 4.1 1 OCFR55.43(4)

Learning Objectives 0AB127E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Off-Gas System-High Radiation, Abnormal Operating Procedure.

MaNterialRequired for ExamiBa n l~uston ouýe: INPO Exam Bank Question Modification Mehd Editorially Modified IPuestnSorce Commients: -1Vision Bank QID# Q62046 Saturday, March 23, 2002 10:47:50 AM Page 133 of 139

Given the following conditions:

- The plant has been operating at 100% power for several weeks

- Main Steam Line (MSL) radiation levels have been averaging 120 mRem but are now slowly trending upwards

- Chemistry reports the higher radiation levels are due to failed fuel

- HC.OP-AB.ZZ-0203, Main Steam Line High Radiation is entered Based on plant conditions, which one of the following Immediate Operator Actions are required?

Ei Place additional Condensate Demineralizers in service if possible

[i Reduce reactor power to maintain MSL radiation levels less than 180 mRem SDirect Reactor Water Cleanup flow to the main condenser to reduce coolant activity IScram the reactor and close the Main Steam Isolation VaIves when MSL levels reach 180 mRem Anr -a Level B Conitve Level Memory iHope Creek -- a Da: 03/1 2/2002 Tier: Generic Knowledge and Abilities R...r....1.... G 1 294001G311 GENERIC Record Number 125 2.3 Radiological Controls 2.3.11 Ability to control radiation releases. 2.773.21 j~naow of JUSTIFICATION:

The immediate actions of HC.OP-AB.ZZ-0203 include:

Ensure all appropriate automatic actions have occurred Reduce reactor power to clear the MN STM LINE RADIATION HI alarm (1.5X)

Trip the H202 Injection System if Radmonitors reach 2.0 and notify Chemistry to verify the system is Shutdown If a valid MAIN STEAM LINE HI HI Radiation Condition exists, then SCRAM and shut the MSIVs and Drains CORRECT - Reduce reactor power to maintain MSL radiation levels less than 180 mrem. 180 mr is 1.5X the normal average value of 120 mr stated in the stem. Reducing power is IOA 2 above.

INCORRECT - Direct Reactor Water Cleanup flow to the main condenser to reduce coolant activity. This will not have an appreciable affect on coolant activity unless the RWCU demin is out of service of exhausted neither of which were stated in the stem.

INCORRECT - Scram the reactor and close the Main Steam Isolation Valves when MSL levels are greater than 120 mrem. This is not performed until the MSL HI HI Rad alarm is in at 3xNormal.

INCORRECT - Place additional Condensate Demineralizers in service if possible. While this could be done, it is not directed by AB-203.

Reference Title HC.OP-AB.ZZ-0203, Section 3.2 Learning Objectives OAB203E002 (R) From memory, recall the Immediate Operator Actions for Main Steam Line High Radiation, Abnormal Operating Procedure.

Material Required for Examination Saturday, March 23, 2002 10.47.50 AM Page 134 of 139

Facility Exam Bank lQuestion Mo ifkation Method: Editorially Modified rments:I Vision Bank QID# Q61774 Saturday, March 23, 2002 10:47:50 AM Page 135 of 139

Which one of the following describes organizational grouping of Abnormal Operating Procedures (ABs) lAW SH.OP-AP.ZZ-0102 "Use of Procedures".

El 100 series are operational transient procedures Ii 200 series address component failures 3I300 series apply at all times 000 series address fire and medical emergencies

-I e B B .Ia cognitiv vel Memory . a.ilit. Hope Creek E.a. Date: 03/12/, 2002 Tier: IGeneric Knowledge and Abilities SRO Group 1 294001G40, 5 GENERIC Record Number 126 2.4 Emergency Procedures and Plan 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and 2.9 3.6 1 emergency evolutions.

[paation of Justification IAW SH.OP-AP.ZZ-01 02, section 5.5.2 S - 0 Reference Title SH.OP-AP.ZZ-0102, section 5.5.2 Learning Objectives 00011-3E005 a. Summarize the guidelines for the use of the following types of procedures:

Abnormal Operating Procedures Emergency Operating Procedures Alarm Response procedures Material Required for Examination QetionSourE Facility Exam Bank Qustion Modification Method: i Direct From Source FQuestion Source Comments, I VISION BANK QID# Q57004 Saturday, March 23, 2002 10:47:50 AM Page 136 of 139

HC.OP-EO.ZZ-0206A is being implemented during an ATWS event.

Which one of the following describes why RCIC injection must be terminated prior to opening SRVs?

I1 RCIC is injecting cold water

[i RCIC Turbine damage may occur El The Boron concentration will be diluted 1RPV pressure may NOT be sufficient to drive the RCIC Turbine Answer b a Lee _B Memory_

n_ .CognitiveLevel. ct Hope Creek E03/12/2002 Tier: ,Generic Knowledge and Abilities R Group I S G 1 294001G418 GENERIC Record Number 127 2.4 Emergency Procedures and Plan 2.4.18 Knowledge of the specific bases for EOPs. 2.7 13.6 Ixpln¶toO]

Reference:

HC.OP-EO.ZZ-0206A, Step RF-15 Bases HC.OP-EO.ZZ-0206A, Step RF-15 Bases Learning Objectives 000134E008 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by that step.

I-acuity texam tsanK Question Modification Method: Editorially Modified rnts: I Vision Exam Bank QID # Q62115 Saturday, March 23, 2002 10:47:50 AM Page 137 of 139

Given the following:

- The Control Room receives a telephoned bomb threat

- The caller states that an explosive device is attached to a hydrogen trailer at the Hydrogen Water Chemical Injection offloading station in the yard south of the power block

- Security is implementing Contingency Procedures

- Security officers confirm the presence of a suspicious device

- No other suspicious activity is observed at this time Which one of the following describes the time requirement in which the NRC must be notified?

[i Within 15 minutes t~IWithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months R Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months jb xa Level - cognitive Cornrehension Fait Hope Creek E Date- 03/12/2002 Tier: Generic Knowledge and Abilities R. Group SG 1 294001iG428 GENERIC Record Number 128 2.4 Emergency Procedures and Plan 2.4.28 Knowledge of procedures relating to emergency response to sabotage. 2.3 3.3

[anauonof The event requires Unusual Event declaration lAW ECG Section 9.1.1. NRC notification is required within I one hour.

Reference Title HC ECG Section 9.1.1 10CFR55.43(1)

Learning Objectives material Required for Examination ECG without Usage Section 1 SINPO Exam Bank Question Modification Method: Editorially Modified lQuestionSource Comments: 7 INPO EXAM BANK QID# 1978 Palisades 1 06/14/1999 Saturday, March 23, 2002 10:47:50 AM Page 138 of 139

Which one of the following describes how a scram is verified in accordance with HC.OP-IO.ZZ 0008 Shutdown from Outside the Control Room?

El HCU nitrogen pressure verified to be less than 800 psig at each HCU F- Reactor vessel pressure verified less than 920 psig El RPS power distribution circuit breakers verified to be open I

rq Scram air header pressure verified to be less than 100 psig Answr a Exam L~eve B Memor .Hope Cognitive Level .a.iity Creek Exam at 03/12/2002 Tier' Generic Knowledge and Abilities R. G 1 SOGop1 294001G434 GENERIC 'Recorl Nu m-b , 129 2.4 Emergency Procedures and Plan 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations 3. 6 including system geography and system implications.

Explaaionof The scram is verified outside the control room via HCU Accumulator pressures < 800 psig at each HCU r Reference Title 1 HC. OP-I O.ZZ-0008 teaming Objectives 001 12HE004 (R) Apply Precautions, Limitations and Notes while executing the SHUTDOWN FROM OUTSIDE THE CONTROL ROOM S. Integrated Operating Procedure.

Material Required for Examination Son ur . Facility Exam Bank lQueon Modification Method: IDirect From Source IQuestion Source Comments: I Vision Exam Bank QID # Q54018 Saturday, March 23, 2002 10:47:51 AM Page 139 of 139

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