Text

2009 Initial Examination 05000327/2009301 and 05000328/2009/301, Draft SRO Written Examination
	ML090710653
Person / Time
Site:	Sequoyah
Issue date:	02/06/2009
From:	; - No Known Affiliation
To:	; NRC/RGN-II
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{{#Wiki_filter:( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 76. 008 AG2.4.2 076 Given the following: Unit 1 at 100% power when a pressurizer safety valve failed open. The operator manually tripped the reactor and initiated a safety injection. While performing the step to determine if the RHR spray should be placed in service in accordance with E-1, "Reactor Trip or Safety Injection", the crew determines the following: When pressurizer pressure dropped to 1280 psig, the safety valve reclosed and pressurizer pressure started to rise. Containment pressure rose to 2.6 psig, and began trending down. Pressurizer level is 100%. RCS subcooling is 43°F. -All four SG levels at 33% narrow range. Which ONE of the following identifies the correct procedure implementation and operation of the RCPs for the above conditions? A'I Transition from E-1 to ES-1.1, SI Termination; The RCPs will have remained running throughout the event. B. Transition from E-1 to ES-1.1, SI Termination; The RCPs would have been shutdown but will be restarted in ES-1.1, SI Termination. C. Continue E-1 until a transition is directed to ES-1.2, Post LOCA Cooldown; The RCPs will have remained running throughout the event. D. A transition will be made to ES-1.2, Post LOCA Cooldown; Page 1 The RCPs would have been shutdown but will be restarted in ES-1.2, Post LOCA Cooldown. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA C TOR ANAL YSIS: Page 2 A. CORRECT, with the safety valve reclosed and the conditions as identified in the stem, SI termination criteria is met. While the crew would be beyond the step in E-1 that first checks for SI termination and beyond the followup step for checking the criteria, the SI termination step is a continous action step and if the criteria is met the transition is to be made. Subcooling is greater than he 400F setpoint, pressurizer level above the 10% setpoint, heat sink is established and RCS pressure rising meet the entry conditions for ES-1.1. Containment pressure did not rise to the automatic initiation setpoint of 2.8 psig (Phase B) nor did the RCS pressure drop to the 1250 psig setpoint, so the RCP trip criteria was not met and the pumps remained in service. B. Incorrect, With the conditions identified in the stem, the SI termination criteria is met and a transition to ES-1.1 is required. The RCP trip criteria was not met and the pumps would have remained in service throughout the even. Plausible because the transition to ES-1.1 is the correct transition and if the RCPs had been stopped they would be restarted in ES-1.1. C. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions identified in the stem indicate SI termination criteria is met and a transition to ES-1.1 is required. The RCP trip criteria was not met and the pumps would have remained in service through out the event. Plausible because the transition to ES-1.2 would be the correct transition if the SI could not be terminated and the RCPs remaining in service through out the event is correct. D. Incorrect, While ES-1.2 would be entered if E-1 was continued, the conditions identified in the stem indicate SI termination criteria is met and a transition to ES-1. 1 is required. Because the RCP trip criteria was not met, the pumps would have remained in service through out the event. Plausible because the transition to ES-1.2 would be the correct transition if the SI could not be terminated and if the RCPs had been stopped they would be restarted in ES-1.2. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 76 Tier 1 Group 1 KIA 008 AG2.4.2 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. Importance Rating: 4.5 1 4.6 Technical Reference: E-1, Loss of Reactor Or Secondary Coolant, Rev 23 ES-1.1, SI Termination, Rev 10 ES-1.2, Post LOCA Cooldown, Rev 17 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 E-1 B.5 Question Source: Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Bank# ----Modified Bank # ----New X ---Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.7/45.7/45.8 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 3 . ' . .... "'. SQN LOSS OF REACTOR OR SECONDARY COOLANT FOLDOUT PAGE RCP TRIP CRITERIA IF any of the following conditions occurs: * RCS pressure less than 1250 psig AND at least one CCP or SI pump running OR * Phase B isolation, THEN STOP all RCPs. SI REINITIATION CRITERIA IF any of the following conditions occurs: * RCS subcooling based on core exit TICs less than 40°F OR * Pressurizer level CANNOT be maintained greater than 10% [20% ADV], THEN RAISE ECCS flow by performing one or both of the following as necessary:

. ESTABLISH

CCPIT flow USING Appendix C * START CCPs or SI pumps manually. EVENT DIAGNOSTICS

IF both trains of shutdown boards de-energized, THEN GO TO ECA-O.O, Loss of All AC Power. * IF any SIG pressure dropping in an uncontrolled

manner or less than 140 psig AND SIG NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation.

IF any S/G has level rising in uncontrolled

manner or has abnormal radiation, THEN: a. RAISE ECCS flow by performing one or both of the following as necessary:

ESTABLISH

CCPIT flow USING Appendix C * START CCPs or SI pumps manually. b. GO TO E-3, Steam Generator Tube Rupture. TANK SWITCHOVER SETPOINTS

IF CST level less than 5%, THEN ALIGN AFW suction to ERCW. * IF RWST level less than 27%, THEN GO TO ES-1.3, Transfer to RHR Containment

Sump. Page 1a of 26 E-1 Rev. 23 ( SQN LOSS OF REACTOR OR SECONDARY COOLANT E .. 1 Rev. 23 I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 7. MONITOR SI termination criteria: a. RCS subcooling based on core exit TICs greater than 40°F. b. Secondary heat sink: * Narrow range level in at least one Intact S/G greater than 10% [25% ADV]. OR * Total feed flow to Intact SIGs greater than 440 gpm. c. RCS pressure STABLE or RISING. d. Pressurizer level greater than 10% [20% ADV]. e. GO TO ES-1.1, SI Termination. ---.----a. GO TO StepS. b. GO TO Step S. c. GO TO Step S. L lKJ N 01 -fff),IJ! t'{1 ::a-.. d. ATTEMPT to stabilize RCS pressure with normal pressurizer spray. GO TO Step S. Page 10 of 26 ( OPL271 E-1 Revision 2 Page 3 of 85 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-1, "Loss of Reactor or Secondary Coolant" IV. LENGTH OF LESSON/COURSE: 2 hours V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of E-1, "Loss of Reactor or Secondary Coolant. B. Enabling Objectives O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with E-1, "Loss of Reactor or Secondary Coolant that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of E-1 . 2. Discuss the E-1 entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into E-1. 4. Describe the bases for all limits, notes, cautions, and steps of E-1. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use E-1 to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of E-1 conditions. ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 77. 009 EA2.37 077 Given the following: -Unit 1 is operating at 100% power when a loss of offsite power occurs. -The operators subsequently initiate Safety Injection due to a small break LOCA. -Thirty minutes after the Safety Injection, the following conditions exist: -E-1, "Loss of Reactor or Secondary Coolant" is being performed. -All 4 SG pressures are approximately 1010 psig and stable. -RCS pressure is 2230 psig and stable. -Thot is approximately 575°F in all 4 loops and lowering slowly. -Core Exit TCs indicate approximately 580°F and slowly rising. -T cold is approximately 560°F in all 4 loops and stable. Based on the above indications, which ONE of the following identifies the condition of the RCS and the procedure transition that will be made? A. Natural Circulation exists and a transition will be directed to ES-O.2, Natural Circulation Cooldown as the E-1 procedure is continued. B. Natural Circulation exists and a transition will be directed to ES-1.2, Post LOCA Cooldown as the E-1 procedure is continued. C. Natural Circulation does NOT exist and a transition will be directed to ES-0.2, Natural Circulation Cooldown as the E-1 procedure is continued. Natural Circulation does NOT exist and a transition will be directed to ES-1.2, Post LOCA Cooldown as the E-1 procedure is continued. Page 4 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted ( DIS TRA CTOR ANAL YSIS: Page 5 A. Incorrect, Natural Circulation does not exist. Tcold (5600F) is approximately 10 degrees higher than saturation temperature of all 4 SGs (-54ffJF). Although SG pressures are at approximately the Atmospheric dump valve pressure, they mayor may not be open. The only way to tell if natural circulation exists is by trending Tcold. The transition to ES-O.2 from E-1 is not correct. Plausible because Thot lowering and Tcold stable could exist with natural circulation and ES-0.2 is a natural circulation procedure. B. Incorrect, Natural circulation does not exist. Tcold (5600F) is approximately 10 degrees higher than saturation temperature of all 4 SGs (-54ffJF). The only way to tell if natural circulation exists is by trending Tcold. Steam dumps are unavailable due to the loss of off site power. The transition to ES-1.2 from E-1 is correct. Plausible because Thot lowering and Tcold stable could exist with natural circulation and ES-1.2 is the correct procedure. C. Incorrect, Natural circulation does not exist. Tcold (5600F) is approximately 10 degrees higher than saturation temperature of all 4 SGs (-54ffJF). The only way to tell if natural circulation exists is by trending Tcold. Steam dumps are unavailable due to the loss of offsite power. The transition to ES-0.2 from E-1 is not correct. Plausible because natural circulation does not exist (Core Exit temperatures rising) and ES-0.2 is a natural circulation procedure. D. CORRECT, Natural Circulation does not exist. Tcold (5600F) is approximately 10 degrees higher than saturation temperature of all 4 SGs (-54ffJF). Although SG pressures are at approximately the Atmospheric dump valve pressure, they mayor may not be open. The only way to tell if natural circulation exists is by trending Tcold. The transition to ES-1-1 is the correct transition from E-1 for the conditions stated. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 77 Tier 1 Group 1 KIA 009 EA2.37 Ability to determine or interpret the following as they apply to a small break LOCA: Existence of adequate natural circulation Importance Rating: 4.2 I 4.5 Technical Reference: E-1, Loss of Reactor or Secondary Coolant, Rev 23 Steam Tables EA-68-6, Monitoring NAtural Circulation Conditions, Rev 0 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 E-1 B.5 Question Source: Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Bank# ___ _ Modified Bank # X Question History: ---New ---SQN Bank question ES-O.2-B.6 002 modified to add SRO component Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 43.5 I 45.13 ) 10CFR55.43.b ( 5 ) Comments: SQN Bank question ES-0.2-B.6 002 modified to add SRO component SQN Question modified from Salem Unit 1 NRC exam dtd 11/04/2002 Page 6 SQN LOSS OF REACTOR OR SECONDARY COOLANT E-1 Rev. 23 I STEP I I ACTION/EXPECTED RESPONSE 15. c. MONITOR containment sump level less than 68%. d. NOTIFY TSC to initiate post-accident sampling as necessary. e. EVALUATE plant equipment status USING EA-0-4, Evaluation of Equipment Status. 16. DETERMINE ifRCS cooldown and depressurization is required: a. CHECKRCS pressure greater than 300 psig. b. GO TO ES-1.2, Post LOGA Cooldown and Depressurization. --.. 11----I I RESPONSE NOT OBTAINED c. NOTIFY TSG to evaluate containment sump level and actions of FR-Z.2, Containment Flooding. a. IF RHR injection flow greater than 1000 gpm, THEN GO TO Step 17. Page 19 of 26 SQN MONITORING NATURAL CIRCULATION CONDITIONS 1,2 4.2 Verification of Natural Circulation 2. DETERMINE parameter trends between monitoring intervals and EVALUATE if natural circulation is occurring. 3. IF natural circulation NOT verified, THEN NOTIFY ASOS. 4. GO TO Section 4.1, step in effect. END OF TEXT EA-68-6 Rev. 0 Page 4 of 5 D D D ES-O.2-B.6 002 Given the following: QUESTIONS REPORT for BANK SQN Questions -Unit 2 is operating at 100% RTP when a Loss of Off-Site power causes a reactor trip. Ten minutes after the trip, the following conditions exist: -SG #1 Pressure1 01 0 psig and stable SG #2 Pressure1005 psig and stable SG #3 Pressure 1015 psig and stable SG #4 Pressure1 01 0 psig and stable RCS Pressure is 2230 psig and stable Thot is approximately 575 of in all 4 loops and lowering slowly Core Exit TCs indicate approximately 580 OF Tcold is approximately 560 OF in all 4 loops and stable Based on the above indications, what is the condition of the RCS? A. Natural Circulation exists. S/G PORVs are maintaining heat removal. B. Natural Circulation exists. The steam dumps are maintaining heat removal. C. Natural Circulation does NOT exist. Heat removal may be established by opening the steam dumps. Natural Circulation does NOT exist. Heat removal may be established by opening the S/G PORVs. Monday, November 10, 2008 8:14:12 AM 1 OPL271E-1 Revision 2 Page 3 of 85 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-1, "Loss of Reactor or Secondary Coolant" IV. LENGTH OF LESSON/COURSE: 2 hours V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of E-1, "Loss of Reactor or Secondary Coolant. B. Enabling Objectives o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with E-1, "Loss of Reactor or Secondary Coolant that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of E-1 . 2. Discuss the E-1 entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into E-1. 4. Describe the bases for all limits, notes, cautions, and steps of E-1 . 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use E-1 to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of E-1 conditions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 78. 011 EA2.14 078 Given the following: Unit 1 is at 100% power. RHR Pump 1 B-8 is tagged for motor bearing maintenance. -A Safety Injection occurs due to a LOCA. RHR Pump 1 A-A trips on instantaneous overcurrent when it attempts to start. -As the crew transitions from E-1, Loss of Reactor or Secondary Coolant, to ECA-1.1, "Loss of RHR Sump Recirculation", the STA reports a RED path to FR-P.1, "Pressurized Thermal Shock". RCS pressure is currently 230 psig. RWST level is 54% and dropping. Which ONE of the following describes the required operator action? A. The transition to FR-P.1 should NOT be made due to the ECA-1.1 entry requirement. B. The transition to FR-P.1 should NOT be made until transfer to the containment sump is accomplished. C. The transition to FR-P.1 should be made but a transition back to ECA-1.1 will be directed if SI termination criteria is NOT met in FR-P.1. The transition to FR-P.1 should be made but a transition back to ECA-1.1 will be directed with no RHR pump running. Page 7 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DISTRAGTOR ANAL YSIS: Page 8 A. Incorrect, The transition to FR-P.1 is required to be made from EGA-1.1. Plausible because some other EGAs suspend the implementation of Status trees or direct the transition to FRGs not be made during the EGA performance. B. Incorrect, The transition to FR-P.1 is required to be made from EGA-1.1. Plausible because some other EGAs suspend the implementation of Status trees or direct the transition to FRGs not be made during the EGA performance until certain conditions are met and the establishment of sump recirculation is a priority for the given conditions. G. Incorrect, The transition to FR-P.1 is required to be made and the transition back to EGA-1.1 is required but not because to the status of SI termination criteria. Plausible because FR-P.1, Step 3 does direct a return to the procedure in effect prior to reaching the SI Termination check which would also direct the same transition back to the instruction in effect. D. GORREGT, The transition to FR-P.1 is required to be made and when FR-P.1 is entered, Step 3 will direct a return to the procedure in effect due to the RGS pressure being less than 300 psig along with both RHR pumps being stopped and sump recirculation capability loss. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 78 Tier 1 Group 1 KIA 011 EA2.14 Ability to determine or interpret the following as they apply to a Large Break LOCA: Actions to be taken if limits for PTS are violated Importance Rating: 3.6* 14.0 Technical Reference: ECA-1.1, Loss of RHR Sump Recirculation,Rev 11 FR-P.1, Pressurized Thermal Shock, Rev 13 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 FR-P.1 B.5 Question Source: Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Bank# ____ _ Modified Bank # ___ _ New --'X'-'--__ Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge ___ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (43.5/45.13 ) 10CFR55.43.b (5) Comments: New question for SQN 1/2009 exam Page 9 I SQN I PRESSURIZED THERMAL SHOCK FR-P.1 Rev. 13 I STEP II ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED 1. MONITOR RWST level greater than 27 %. 2. MONITOR CST level greater than 5%. 3. CHECK RCS pressure greater than 300 psig. IF RHR pumps aligned to RWST, THEN GO TO ES-1.3, Transfer to RHR Containment Sump. ALIGN AFW suction to ERCW USING EA-3-9, Establishing Turbine Driven AFW Flow, and EA-3-10, Establishing Motor Driven AFW Flow. IF, any of the following conditions exist: * RHR injection flow greater than 1000gpm OR * both RHR pumps STOPPED AND sump recire capability has been lost THEN RETURN TO procedure and step in effect. Page 3 of 25 OPL271 FR-P.1 Revision 1 Page 3 of 16 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: FR-P.1, PRESSURIZED THERMAL SHOCK IV. LENGTH OF LESSON/COURSE: 1 hours V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of the FR-P.1, Pressurized Thermal Shock. B. Enabling Objectives o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated FR-P.1, Pressurized Thermal Shock, that are rated;:::: 2.5 during Initial License Training and;:::: 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of FR-P .1. 2. Discuss the FR-P.1 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with FR-P.1 entry conditions. b. Describe the requirements associated with FR-P.1 entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into FR-P.1. 4. Describe the bases for all limits, notes, cautions, and steps of FR-P.1. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use FR-P.1 to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of FR-P.1 conditions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 79. 015 AA2.01 079 Given the following: Unit 1 is at 32% power. -The following annunciator windows alarm on 1-M-6: FS-68-48A REACTOR COOLANT LOOP 3 LOW FLOW RCP BUS UNDERFREQUENCY 1 UNDERVOL TAGE RCP #3 indications are: -Ammeter -'0' amps. -Green and white lights above the handswitch are lit. Flow -'0' on all 3 indicators. -The other RCPs remain in service and the reactor trip breakers remain closed. -The OATC manually trips the reactor as directed by the SRO. Which ONE of the following identifies the condition causing the RCP trip and the earliest NRC notification required by 1 OCFR50. 72? Condition causing RCP Trip Notification Requirement A. Under voltage 1 hour Under voltage 4 hours C. Under frequency 1 hour D. Under frequency 4 hours Page 10 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 11 The RCPs have protection from undervoltage and underfrequency via relays. Both conditions can cause trips of the RCPs and if conditions and logic is met, the reactor would automatically trip. A. Incorrect, The conditions in the stem describe the trip of RCP #3 from an undervoltage condition on the 6.9 Kv Unit Board feeding the RCP, however NO automatic trip would be initiated at the stated power level (below P8) and the emergency plan would not be implemented. (No 1 hour notification would be required). Plausible because the condition causing the RCP trip is correct and if the power level had been greater than P8, an automatic reactor trip should have occurred .. If the reactor had failed to trip automatically, a 1 hour report would be required due to implementation of the Emergency Plan. B. CORRECT, The conditions in the stem describe the trip of RCP #3 from an undervoltage condition on the 6.9 Kv Unit Board feeding the RCP. With the reactor power level below P-8, no automatic reactor trip would be initiated from low flow condition in a single loop. AOP-R.04, Malfunction of Reactor Coolant Pump, would be the controlling procedure and at the stated power level, a manual reactor trip would be directed. The reactor trip would require a 4 hour notification. C. Incorrect, The conditions in the stem describe the a trip of RCP #3 from an undervoltage condition on the 6.9 Kv Unit Board feeding the RCP not an underfrequency trip. The underfrequency trip of the RCP requires 214 logic to be made in the SSPS which then will trip the reactor and all RCPs. The logic is not met in the stem even though power is above the P7 permissive (10%), only one RCPs is involved and the logic requires 2 out of 4. NO automatic trip would be initiated at the stated power level (below P8) and the emergency plan would not be implemented. (No 1 hour notification would be required). Plausible because underfrequency can cause RCP trip with conditions different than stated in the stem and following the underfrequency initiation at greater than P7, a failure of the reactor to automatically trip would require a 1 hour due to the implementation of the Emergency Plan. D. Incorrect, The conditions in the stem describe the trip of RCP #3 from an undervoltage condition on the 6.9 Kv Unit Board feeding the RCP not an underfrequency trip. The underfrequency trip of the RCP requires 214 logic to be made in the SSPS which then will trip the reactor and all RCPs. The logic is not met in the stem even though power is above the P7 permissive (10%), only one RCPs is involved and the logic requires 2 out of 4. the 4 hour notification due to the manual reactor trip is correct. Plausible because underfrequency can cause RCP trip with conditions different than stated in the stem and the 4 hour notification due to the manual reactor trip is correct. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 79 Tier 1 Group 1 KIA 015 AA2.01 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Cause of RCP failure Importance Rating: 3.0/3.5* Technical Reference: 1-AR-M6-A Reactor Protection and Safeguards 1-XA-55-6A, Rev 15. 1,2-47N763-2 R20 1-47W611-99-6 R2 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.RCP S.4.i. & 5.c Question Source: Describe the following items for each major component in the Reactor Coolant Pump system as described in this lesson: i. Protective features (including setpoints) Describe the operation of the RCP system as it relates to the following: c. Alarms and Alarm Response Sank# -----Modified Sank # ----New _X __ Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 43.5 1 45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 12 ( 32 (E-4) Source Setpoint SER 460,467,476,477,478,482,1613, 1614 57Hz and -72% voltage (-5022 volts) Rep BUS UNDERFREQUENCYI UNDERVOLTAGE UV and UF relays OneUVand UF Relay on each Rep bus. 1/4 for alarm 2/4 for trip input Probable Causes NOTE 1 NOTE 2 Corrective Actions References 1. Loss of power to 6.9kVunit bds. 2. Channel malfunction or testing. 3. RCP Undervoltage or Underfrequency relay failure. Underfrequency condition on 2/4 RCPs will trip all four RCPs automatically. Undervoltage condition on individual Rep buss will block operation of associated UF relay. [1] IF no reactor trip occurs, THEN [a] CHECK 1.,.XX-55-6A Reactor Trip 81 status panel for bistables that may be tripped. [b] EVALUATE reactor trip criteria with SRO per AOP-R.04, Reactor Coolant Pump Malfunctions. [c] IF reactor trip should have occurred automatically, ntEN GO TO E-O, Reactor Trip or Safety Injection. [2] IF RCP trips, THEN GO TO AOP-R.04, Reactor Coo/ant Pump Malfunctions, [3] IF RCP undervoltage or underfrequency relay has failed, THEN GO TO AOP-I.10, RCP Undervoltage or underfrequency Instrument Malfunction. [4] EVALUA TE Technical Specifications, 45B655-06A-O, 47m10-68, 47W811-99, 45N721-1,45N763-2 SQN Page 37 of 40 1 1-AR-MS-A Rev. 16 Source SER46S SER 583 SER592 SER 596 1-Fs.6S-'4BA .1-FS-68-48S 1-FS-68-48D 1/3 for alann 2J3 for trip input Probable Causes Corrective Actions References 18 (C-4) Setpoint FS-68-48A REACTOR COOLANT LOOP 3 LOW FLOW Flow < 90% 1. Pump trip. 2. Flow instrument malfunction or testing. 3. Operator stopping pump. [1] IF no reactor trip occurs, THEN [a] CHECK 1-XX-55-6A Reactor Trip $1 status panel for blstables that may be tripped. [b) EVALUATE reactor trip criteria with SRO per AOP-R.04, Reactor Coo/ant Pump Malfunctions. [e] IF reactor trip should have occurred automatically, THEN TRIP the reactor, AND GO TO E-O, Reactor Trfp or Safety Injection. [d] CHECK ri-FI-68-48Al, ri-FI-68-48BJ, and r1-FI-68-48D] for channel failure. [e] IF a single flow channel has failed, THEN GO TO AOP-1.03, ReS Flow Instrument Malfunction. [2] IF Rep trips, THEN GO TO AOP-R.04, Reactor Coolant Pump Malfunctions. . [3] EVALUATE Technical Specifications. 45B655-06A-O, 47'M510-68, 47W611-99 I 1 SQN I i-AR-MS-A Page 23 of 40 Rev. is Trip Logic 1 Breaker Trip I---------- l HS "A" in STOP -:P Transfer Switch in NORMAL ---I I i HS 'C' in STOP I i I Breaker TriPi ___ J Transfer Switch ---. inAUX I I I I 1 6.9KV Bus 1 ____ _J I I RCP Mota, DC jn-----[ ! I RCP BUS UF r.D--J (Train A) RCP BUS UF ---(Train B) E04 x. LESSON BODY: J. Purpose/function of components COMPONENT PURPOSE / FUNCTION RCP Trip logic * A RCP may be stopped from its handswitch or at the switchgear. The (SD page 33) breaker will trip open on a motor protection fault (unit board UV, motor overcurrent) or an underfrequency trip of either reactor protection system train. * In addition to providing a trip to the RCP, an UV trip signal provides an input to the reactor protection system (RPS). The purpose of the RPS trip is to provide DNB protection to the reactor by anticipating a complete loss of flow condition.

This reactor trip requires an undervoltage

condition on two RCPs and is blocked below P-7 (10%). OPT200.RCP Rev.5 Page 28 of 56 RCP Bus 1A UF -.!r-'\ Test Switch in Test RCP Bus 1B UF -Test Switch in Test RCP Bus 1C UF :b J Test Switch in Test RCP Bus 10 UF --'!r-'\ Test Switch in Test Power> P7 _____ _ Under-Frequency Logic RCP Bus UF (Train A) E04 X. LESSON BODY: J. Purpose/function of components COMPONENT PURPOSE / FUNCTION RCPUF logic * An UF condition on two of the four RCP buses supplying the RCPs (SDpage 34) will initiate a reactor trip and trip all four RCP breakers. The purpose of the UF trip is to provide DNB protection to the reactor by anticipating a complete loss of flow condition in two or more loops. * This reactor trip requires an underfrequency condition Hz) on two RCP buses and is blocked below P-7 (10%). * A rapid decrease in electrical frequency can decelerate the RCPs faster than a complete loss of power. This trip, in conjunction with the rotational inertia provided by the RCP flywheel, ensures the proper flow coast-down time is maintained. OPT200.RCP Rev. 5 Page 29 of 56 A B c o E F G H K Z-£9LNGt-Z'1 2 3 I 30" :2: l-F'U2-158-IB }ro ReP SENSOR PANEl. (UV" UF RELAYS) tlOV AC METERIHC POTENTIAL ON UNITBD1A t 51 I-i1 L--I r-------<i L...-J DETAIL A2 :11 :'1 '9 fi 1-..1 r-------<i I-.-J DETAIL 82 DETAIL C2 I 30" 2 ,30'" 2 I 30 ... :2: REACTOR COOLANT PIMI" I REACTOR COOLANT PIMP :2: REACTOR COOLANT PIAIP J REACTOR COOLANT PUMP .. 4101/ REACTOR BLDG :1 }+-) &.!!:::!! ""-, t , XS-68-8 Alii( ..,5 ".; X9-SB-8+ I:' __ 81 START + AlO XS-68-84 START o 2 ... 2Nl HS-68-84A "'; "7 2 ... 2H ef2 1.6 XS-68-XS-88-14 + +!L-A8 XS-68-+ " 88 """"ii5ii:"" 4 82889 t RCl'I !-___ "_"_R*-.JDI2 B8 HOlI. fUl-68-SC MOIl" "ux COttY PENETRAUOH ReP lOlL un PU", l_FU:2:_IB_84A. a+ l-F'U3-88-84 Rep 2 OIL UFT PU"," l-F'U2-88-85A 85 I-FUJ-68-86 ReP JaIL UFT PUMP l-FU2-6B-86A, a6 l-FU3-,a.-86 ReP .. OIL LIfT PUll" l-FU2-68-17A a7 l-FUJ-&&-17 2 12 11 2A2Y REACTOR BLDG VENT BD WIRE PREFIX 1.\-" V,2 111-8 JE lA-A 782 .... ,-. ,-. M-O CONT srs HS-68-84Aa:C HS-68-8i5AaC HS-68-86AaC HS-III-87Aa.C XS-68-&4 XS-68-85 XS-68-86 5 5 6 TACF 2-08-008-068 t' )(5-158-1 .. t 'AnI" -""""l. NOR_ TACF 2-08-008-068 2711. 121,1, "'" 622,1, II[ 272B 622B 7 RaOIS ')0, 20r e y-. OJ .. " 1 TO ilEAC , Sf RCP UVa. (UMITS aliI. I 1121. I 7 8 tANH-PMl "r 1-11-72 fUSE SlOlN" AYS !IF , TEST SWITCHES J H UF IA HS-68-343 ,. IDII12B IA 208118 HS-68-346 " 2011118 IA 307118 HS<-II-347 HS-H-34B " 3071211 13. 409038 HS-68-3+8 40904B 8 IIA FUSE MO. (NOrEO lA/I IB/I WI ,./1 Ttl REAC TRIP LClCJC L'! .. -i _ -344 -81UlI[ I -f--I -PICA' J 2 P)(I. PilI, 10811al PICA 9 10 * 8111.)[ , MAINTAINED PTAIC-SEC l-FU2-S&-IH 2-fU2-S8-IH l-FU2-68-J1H 2-fU2-I58-J1H 2-FU2-B8-50H l-FU2-6S-7JH 2-FU2-S8-7JH 11 12 NOTES: L FOR UNIT 2. rUSE UlUDS OlAMGE UNIT OESICtlATION. EXAMPLE, l-F'U+-611-6A '-THIS IIJIIIBER WILL CHANCE TO 2, FOR UNIT 2. BOARDS. 2. RELAY TYPES ARE AS FOLLOWS: A -D.C. RELAY. toE TTPE 1AC77A B -D.C. RELAY. GE: TYPE IACII6.I C -DISTR RELAY.! TYPE KOIO l. 'IRE DESICNATIONS SHOWN ARE fOR Rep lOlL LIFT PUMP. WIRE DESIGNATIONS FOR !tel' 2. J 4 + OlL LIrT I'UtrS ARE SHOrtt 011 CONNECTIOH DIAGft,wS. REFERENCE Dltt,'IM!:;:SI 4SN721-f. -2 -----6800V UNIT eo SINCt,E LINE DtAGRAWS 4SN755 * ___________ 480V RUCTOR BLOC '/[1(1, BO SINGLE LINE DIAGRAMS +5"'721 ----------- ISOOV UNIT SO COHNECTIOI4 DIAGRAIoIS A B C o E

480V REACTOR BLOC VE.rtT so CONNECTION

DJAGRAIoIS F SYNEIOLS, .. --_ EQUIPMENT LOCATtD OM UNIT CONTROL &I) t -EQUIPMENT tlJUNTED OM &!IaDV SWnCHGEAf!: + " __ EQUll't.IEMT IIIlUHTED OM 480V ..:lTIlR ootnRO\. CENTER c---EOUIPt.lENT LOCATED ON TVA BOP RACKS III. AUX ItGTR ROCN * -_ EQUIPMENT t.OCATED ON HSSS RACKS IN AUX INSTR ROCIoI 20 TACY 2-OS-oDS-<l68 INC:TACf'2-Q8-0CI1Hl88 REV CHANGE REI" PREPARER OlECK(II SCALE: HONE TURBINE BUILDING UNITS 1 a:. 2 WIRING DIAGRAMS * , 0/1 APPROVED DATE' EXCEPT AS NOTED CATECORY 1 6900V UNIT AUXILIARY POWER SCHEMATIC DIAGRAMS SEQuovAH NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY ENGIIlEERIHC API'RO.fAL .20 G H 4-{Jul( .. /'fLL -.. ., 9-86-L L9MLjoo-L 2 OVERTEMPERATURE AT A 8 UNDERVOLTAGE (NOTE 2) c D oA-U E FLOW LOOP 1 F ...... G H 2 OVERPOWER 4T 'OVERPOWER loT REACTOR TR I P SHEET 1 4 5 UNDERFREQUENCY RCP-BUSSES (NOTE 2) Loo YQ"IOOD FLOW LOOP 2 FLOW LOOP :3 4 iA.-.2 6 lliillJ FLOW LOOP 4 \ ' 7 8 9 '0' POWER RANGE 10 12 TURBINE TRIP [SS,JED BYI P.G.TRUDEL PROCAD MAINTAINED DRAWING THI! ctlNTMlL DftA'II'INC [S BY 11-£ :!QN CIIIJ UNIT 0II1II1J lS I<<JII PAltr OF THE T'tA I"Rot:AM.! DATABASE. lURIHhI£ TRIP REAeTOA TlUFI NOTEs: 1, FOR Attn (liENEIiAl- $EE 3HEET 1. 1 ,2: sECONOs .altD A B ,0 s c D TIU£ E :5. O]GITAL II.HD ANALOC LOCIC SYliISOLS ARE USED ON LOG;IC [)lAGRAIIS 10 FUNCT10NALLY DEseRlBE THE PROCESS CONTROL. REFER TO THE ASSOCIATED IIIUNt: SCHEIdATIC FOR THE ELECTRICAL CO\oIPONIi:NTS USED TO UIPLEIdENT THE CONTROL SCHU.oIE. REFER TO TABLE .... t FDA ADDITIONAL DIGITAl COWPUTER POINTS (STEAM GENEAATOR HI-HI LEYEL TURBINE TI'Up gTATus). 6, MEFE" TO TASLE 1.5 SHEET 5 FOR AODITIOfI,&,L DI0ITAL COMflUTER PQ1NT$ t6900Y aHVTD01l'N aQARD eLACKOUT PCN MI15.!!.A (TY? ALL BACKCIRCLES) FmI ORIO'[t.JAL SIGNATURES IN TITlE BLOCK SEE REVIslQhI 0 \CICIiOFILWi '0(11'1'. Z I1CN l,t11.5sIIA N/A aGR INC: RD REV £eN/DeN CHG DOC DRiIl'N m*u{O APPD DATE NONE POWERHOUSE UNIT 1 MECHANICAL LOGIC DIAGRAM EXCEPT ,s,S NOTED CATEGORY 1 REACTOR PROTECTION SYSTEM SEQUOYAH NUCLEAR PLANT Q TENNESSEE VALLEY AUTHORITY NUIiLEAR ENF;]NEERING: DmICN ORMTERl L.8EASLEY IJE$lQHERl REVlt'WER: . __ K.R.SPINO DATE UUT]AL IssuE RO JS!UE flU: ENC"INEERINC" AP,.IIiC'lAL .. D.), -04, 1 L,BEASLt"'f Ii 2 ".8.N[8IUlT S 'W.R.StDLACII</L'IIA S-z;1 .. iQ 45 w CGO NO:1-47W611-99-6 R2 F I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: REACTOR COOLANT PUMP SYSTEM IV. LENGTH OF LESSON: 4 hour lecture; 2 hour simulator demonstration; 2 hour self-study/workshop V. TRAINING OBJECTIVES: A. Terminal Objective: OPT200.RCP Rev.5 Page 3 of 56 Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Reactor Coolant Pump system in the plant and on the simulator. B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with the Reactor Coolant Pump System that are rated:::: 2.5 during Initial License training for the appropriate license position as identified in Appendix A. 1. State the purpose/functions of the Reactor Coolant Pump System as described in the SQN FSAR. 2. State the design basis of the Reactor Coolant Pump System in accordance with the SQNFSAR. 3. Explain the purpose/function of each major component in the flow path of the Reactor Coolant Pump System as illustrated on the simplified system drawing. 4. Describe the following items for each major component in the Reactor Coolant Pump System as described in this lesson: a. Location b. Power supply (include control power as applicable) c. Support equipment and systems d. Normal operating parameters e. Component operation f. Controls g. Interlocks (including setpoints) h. Instrumentation and Indications i. Protective features (including setpoints) j. Failure modes k. Unit differences 1. Types of accidents for which the Reactor Coolant Pump System components are designed m. Location of controls and indications associated with the Reactor Coolant Pump System in the control room and auxiliary control room V. TRAINING OBJECTIVES (Cont'd): B. Enabling Objectives (Cont'd): 5. Describe the operation of the RCP system as it relates to the following: a. Precautions and limitations b. Major steps performed while placing the RCP system in service c. Alarms and alarm response d. How a component failure will affect system operation e. How a support system failure will affect RCP system operation f. How a instrument failure will affect system operation OPT200.RCP Rev.5 Page 4 of 56 6. Describe the administrative controls and limits for the RCP system as explained in this lesson: a. State Tech Specs/TRM LCOs that govern the RCPs b. State the::;1 hour action limit TS LCOs c. Given the conditions/status of the RCP system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required 7. Discuss related Industry Events: a. SQ961761PER; RCP#4 above 15mil vibration alarm b. SOER-82-5; Reactor Coolant Pump seal failure c. SER 20-86; RCP shaft failure at Crystal River VI. TRAINING AIDS: A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available) Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted SO. 022 AG2.1.28 080 Given the following: Unit 1 is in Mid-Loop Operation with RHR Train A in service during a refueling outage with the reactor core reloaded. SG nozzle dams are installed. SI pump 1 B-B is tagged and disassembled for impeller replacement CCP 1 A-A trips due to motor failure. RCS level begins to drop and the RHR pump 1 A-A is stopped due to cavitation. -Attempts to open 1-FCV-63-1, RHR Pump Suction from RWST, are unsuccessful. Core exit thermocouples have increased to 20SoF. Which ONE of the following identifies the procedure to be used and how SI pump IA-A will be aligned during performance of the procedure? A. AOP-R02, "Shutdown LOCA". SI pumps will be aligned for Hot Leg injection. B. AOP-R02, "Shutdown LOCA". SI pumps will be prevented from injecting due to L TOP requirements. AOP-R03, "RHR System Malfunction". SI pumps will be aligned for Hot Leg injection. D. AOP-R03, "RHR System Malfunction". SI pumps will be prevented from injecting due to L TOP requirements. Page 13 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 14 A. Incorrect, AOP-R03 is the correct procedure to be used not AOP-R02. The AOP-R03 does direct the SI pumps to be aligned in Hot Leg injection. Plausible because the conditions in the stem would exist if a shutdown LOCA was occurring and if AOP-R02 were entered a note at the start of the AOP directs the use of AOP-R03 if the plant is in Mid-Loop operation and the SI pumps alignment in Hot Leg injection is correct. B. Incorrect, AOP-R03 is the correct procedure to be used not AOP-R02 and the AOP directs the SI pumps to be aligned in Hot Leg injection. Plausible because the conditions in the stem would exist if a shutdown LOCA was occurring and if A OP-R 02 were entered a note at the start of the A OP directs the use of A OP-R 03 if the plant is in Mid-Loop operation and the SI pumps normal alignment with injection prevented due to the L TOPS requirement. C. CORRECT, when the charging pump trips (loss of RCS makeup) the vessel level will start to decrease. When the RHR pump is stopped due to cavitation, core cooling flow is terminated, the core starts heating up and with 1-FCV-63-1 unable to be opened, the AOP -R03 directs the SI pumps to be aligned for Hot Leg Injection. A note prior to the step to align the pumps states that core cooling takes priority over the L TOPS requirements for the SI pumps. The conditions would be the same if a shutdown LOCA was occurring and if AOP-R02 was entered there is a note at the start of the A OP directing the use of A OP-R 03 if the plant is in Mid-Loop operation. D. Incorrect, AOP-R03 is the correct procedure to be used but the AOP directs the SI pumps to be aligned in Hot Leg injection. Plausible because the procedure is the correct procedure to be used and and the SI pumps normal alignment is with injection prevented due to the L TOPS requirements. ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 80 Tier 1 Group 1 KIA 022 AG2.1.28 Loss of Reactor Coolant Makeup Knowledge of the purpose and function of major system components and controls. Importance Rating: 4.1 14.1 Technical Reference: AOP-R02, Shutdown LOCA, Rev 10 AOP-R03, RHR System Malfunction, Rev 20 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 AOP-R03 B.5 Question Source: Summarize the mitigating strategy for the failure that initiated entry into AOP-R03. Bank# ___ _ Modified Bank # X ---New ---Question History: SON 1/2009 exam, AOP-R03 006 modified Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.7) 10CFR55.43.b ( 5 ) Comments: SON question AOP-R03 006 modified Page 15 (' SQN RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 .1 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERA TOR ACTIONS NOTE: If this procedure is entered from AOP-P .05 or AOP-P.06 due to loss of shutdown board(s), then Section 2.3 is the applicable section. 1. DIAGNOSE the failure: GOTO IF ... SECTION PAGE RHR malfunctions due to low water level 'J 1 J1 during reduced inventory or mid-loop operations RHR overpressurization due to high RCS pressure 2.2 29 RHR pump(s) failure or trip 2.3 36 RHR system leak 2.4 41 Failure of RHR due to Loss of CCS 2.5 48 END OF SECTION / Page 3 of 97 ! ( SQN RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Ops CAUTION Changes in RCS pressure could result in inaccuracies in RCS level readings. 1. DETERMINE whether RHR pumps should be STOPPED: a. CHECK any RHR pump RUNNING. b. MONITOR RCS level c. CHECK RHR flow less than 2000 gpm. d. CHECK RHR pump CAVITATING. e. STOP RHR pumps and PLACE in PULL-TO-LOCK. a. GO TO Step 2. b. STOP RHR pumps and GO TO Step 2. c. REDUCE RHR flow to between 1000 gpm and 1500 gpm. [C.6] d. PERFORM the following: 1) RESTORE RCS level to normal band by adjusting charging and letdown. 2) IF RHRleak is suspected, THEN GO TO Section 2.4, RHR Leak. 3) GO TO appropriate procedure. Page 4 of 97 SQN RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Ops (continued) 2. CHECK RCS Vacuum Refill NOT in progress. 3. ISOLATE RCS letdown and drain paths: a. CLOSE FCV-62-83, RHR Letdown Flow Control Valve. b. CLOSE CVCS letdown isolation valves: * FCV-62-69

FCV-62-70

c. ISOLATE any known RCS drain paths. IF RCS vacuum refill in progress, THEN PERFORM the following: a. ENSURE vacuum break valve VB OPEN. [Vacuum Refill Skid] b. ENSURE vacuum pump STOPPED. [Vacuum Refill Skid] c. ENSURE charging flow has been raised. d. WHEN RCS is at atmospheric pressure, Incl4 PERFORM the following: 1) ENSURE PZR PORVs CLOSED. 2) ENSURE Rx Head Vent FSVs CLOSED. Page 5 of 97 SON RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Operations (continued) CAUTION Containment will become a harsh environment due to steam and potential airborne activity if boiling occurs. [C.S] 4. INITIATE actions to protect personnel in containment: a. ANNOUNCE over PA to evacuate personnel from containment. b. NOTIFY RADCON to evacuate personnel I fOrTI COflldlf IIIH:Jfll. c. NOTIFY RADCON to monitor containment radiation conditions. [C.10] 5. INITIATE actions to establish containment closure: a. NOTIFY WCC to initiate closure of all containment penetrations being tracked in Containment Closure Control. b. ENSURE all valves with Containment Closure tags on MCR bench boards are CLOSED. Page 6 of 97 SQN RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 I STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Operations (continued) 6. EVALUATE the following Tech Specs for applicability:

3.4.1.4, Reactor Coolant System Cold Shutdown * 3.9.8.1, Refueling

Operations Residual Heat Removal and Coolant Circulation

3.9.8.2, Refueling

Operations Low Water Level 7. EVALUATE EPIP-1, Emergency Plan Classification Matrix. Page 7 of 97 SON RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Operations (continued) CAUTION The following step may result in water spilling in lower containment if RCS is breached. . NOTE: Opening FCV-63-1 will raise RCS level and flush air from the high point in the RHR suction line using gravity fill from RWST. 8. RAISE RCS level to fill RHR suction line: a. IF any S/G primary side manways =n .*. Q'v'V 1-,', THEN CONTACT RADCON to verify personnel are clear of S/G platforms. b. DISPATCH operator to ensure power restored to FCV-63-1 USING Appendix A. c. ENSURE FCV-74-1 and FCV-74-2 OPEN. d. CLOSE FCV-74-3 and FCV-74-21. b. IF power CANNOT be restored to FCV-63-1, THEN DISPATCH operator with radio to operate FCV-63-1 locally. [AB el. 669 Pipe Chase] (Step continued on next page) Page 8 of 97 SQN RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Operations (continued) 8. e. OPEN FCV-63-1 USING one of the following methods: * handswitch in MCR * handswitch on Rx. MOV Bd * local handwheel [AB el. 669 pipe chase] f. WHEN FCV-63-1 has been THEN PERFORM the following: 1) CLOSE FCV-63-1. 2) OPEN FCV-74-3 and FCV-74-21. N<J1 NOTE: ERCW isolation valves for available upper and lower compartment coolers which were closed in step 5 may be reopened if ERCW piping is intact. 9. START available upper and lower compartment coolers USING Appendix B. Page 9 of 97 . / -.! SQN STEP RHR SYSTEM MALFUNCTION AOP-R.03 Rev. 20 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 RHR Malfunctions Due to Low Water Level During Reduced Inventory or Mid-Loop Operations (continued))

The following

step establishes ECCS flow for feed-and-bleed cooling if core boiling point is approached. Restoration of core cooling takes priority over L TOPS requirements for SI pumps. 10. MONITOR Core Exit TICs less than 200°F USING available TICs on Exosensor display. PERFORM the following: a. ESTABLISH SI pump flow to hot legs as follows: 1) ENSURE operator dispatched to perform App. C, Restoring Power to SI Pumps and Hot Leg Inj Valves. 2) ENSURE FCV-63-5, SI pump suction from RWST, OPEN. 3) ENSURE SI pump suction valves OPEN: * FCV-63-47, SI pump A suction * FCV-63-48, SI pump B suction 4) ENSURE SI pump cold leg injection flowpath isolated:

FCV-63-22

CLOSED OR * FCV-63-152 and FCV-63-153 CLOSED 5) OPEN SI pump hot leg injection valve for SI pump to be started: * FCV-63-156, SI pump A OR * FCV-63-157, SI pump B 6) ENSURE one SI pump RUNNING. (step continued on next page) Page 10 of 97 SQN SHUTDOWN LOCA lAO. P-R.02 Rev. 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS NOTE 1: AOP-R.03 should be used for LOCA while in reduced inventory or midloop. NOTE 2: This procedure has a foldout page. 1. MONITOR if RHR pumps should be stopped: a. CHECK RHR aligned for shutdown cooling: * FCV-74-1 OPEN * OPEN. b. CHECK the following:

Pressurizer

level less than 10% [20% ADV] OR * RGS subcooling based on core exit T/Cs less than 58°F c. STOP RHR pumps and PLACE in PULL-TO-LOCK. a. GO TO Step 2. b. IF Pressurizer level greater than 10% [20% ADV] AND RCS subcooling greater than 58°F, THEN PERFORM the following: 1) DISPATCH operator to ensure HCV-74-34 RHR Return to RWST CLOSED. [AB el. 690, RHR HX Room B] 2) GO TO Step 2. (step continued on next page) Page 3 of 90 SON 1.0 PURPOSE SHUTDOWN LOCA I AOP-R.02 Rev. 10 The procedure provides the actions necessary to mitigate the effects of a LOCA which exceeds normal charging capacity during Mode 4 or Mode 5 (with the exception of leaks during reduced inventory/midloop operation resulting in loss of RHR, which are addressed in AOP-R.03). [C.1] Page 2 of 90 AOP-R.03 006 QUESTIONS REPORT for BANK SQN Questions Unit 1 is preparing for a refueling outage, the unit is in mode 6 RCS is at Reduced Inventory level. Indications of a LOCA are observed. Which of the following procedures is applicable? A. AOP-R02 Shutdown LOCA B:' AOP-R03 RHR System Malfunction C. AOP-R05 RCS Leak & Leak Source Identification D. E-1 Loss of Reactor or Secondary Coolant 'j!\" incorrect: Note in AOP-R.02 states that AOP-R.03 should be used for LOCA when in reduced inventory or midloop. "B" correct: per note in AOP-R.02 "c" incorrect: Note in AOP states to use AOP-R.03 if in reduced inventory or mid/oop. "0" incorrect: Unit in mode 5 Reference: KIA: 2.4.4 (4.0 -4.3) 41.10/43.2/45.6 034 A 1.02 (2.9 -3.7) 41.5/45.5 OPL273C0611 obj 6 Friday, November 14, 2008 7:44:31 AM 1 OPL271 AOP-R.03 Revision 2 Page 3 of 39 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R03, RHR SYSTEM MALFUNCTION IV. LENGTH OF LESSON/COURSE: 2 hours V. TRAINING OBJECTIVES: o. 1. 2. 3. 4. 5. 6. A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-R03, RHR SYSTEM MALFUNCTION. B. Enabling Objectives Objectives Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with RHR SYSTEM MALFUNCTIONs that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A. State the purpose/goal of this AOP-R.03. Describe the AOP-R03 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-R03 entry conditions. b. Describe the ARP requirements associated with AOP-R03 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-R.03 entry conditions. d. Describe the plant parameters that may indicate an RHR System Malfunction. Describe the initial operator response to stabilize the plant upon entry into AOP-R.03. Upon entry into AOP-R03, diagnose the applicable condition and transition to the appropriate procedural section for response. Summarize the mitigating strategy for the failure that initiated entry into AOP-R03. Describe the bases for all limits, notes, cautions, and steps of AOP-R03. 7. 8. 9. 10. OPL271 AOP-R.03 Revision 2 Page 4 of 39 Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Given a set of initial plant conditions use AOP-R.03 to correctly: a. Recognize entry_ conditions. b. Identify required actions. c. Respond to Contingencies. d. Observe and Interpret Cautions and Notes. Describe the Tech Spec and TRM actions applicable during the performance of AOP-R.03. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition. / Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 81 . 077 AG2.4.5 081 Given the following: -Unit 2 operating at 100% power with the switchyard in normal alignment. -Generator operating at 24 Kv and +150 MVAR. -A disturbance occurs causing main generator voltage to spike upward and the following annunciators alarm: -on 0-XA-55-ECB6-A; "GEN 2 MVARABNORMAL OR MVAR RELAY FAILURE" "OSCILLOGRAPH OPERATION OR FAILURE" -on 0-XA-55-ECB6-B; "CC RELAY TEST OR OPERA TIOi'fV -on 2-XA-55-1A "GEN VOLT REGULATOR TRIP" -The BLUE light on161kV CONCORD LINE CARRIER RECEIVED is lit. -Reactive power stabilized at +230 MVARS following the disturbance and Concord line PCB in the switchyard remain closed. 3 Which ONE of the following statements describes additional required actions? A. Manually open the Concord line PCB. Notify the Southeast Area Load Dispatcher to evaluate offsite power status for determining operability. B. Manually open the Concord line PCB. Declare both trains of Offsite power inoperable until the Southeast Area Load Dispatcher completes evaluation of the status of the offsite power system. C!' Notify the Southeast Area Load Dispatcher to evaluate offsite power status for determining operability. Within 24 hours, notify the Operations Duty Specialist of the time period the unit was without automatic voltage control. D. Declare both trains of Offsite power inoperable until the Southeast Area Load Dispatcher completes evaluation of the status of the offsite power system. Within 24 hours, notify the Operations Duty Specialist of the time period the unit was without automatic voltage control. Page 16 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 17 A. Incorrect, The Concord Line PCB is not required to be opened with the stated conditions. The Southeast Area Load Dispatcher is required to be notified immediately to determine off site power voltage requirements to determine if offsite power supplies are operable. Plausible because conditions would require opening the PCB and if the PCBs had opened, the MOD s are required to be opened manually. Additionally, the dispatcher notification to detemine status of the off site power operability is required. B. Incorrect, The Concord Line PCB is not required to be opened with the stated conditions and the Southeast Area Load Dispatcher is required to be immediately notified so that an evaluation can be performed but the offsite power operability status is determined after the evaluation is complete. Plausible because conditions would require opening the PCB and if the PCBs had opened, the MOD s are required to be opened manually. Additionally, the dispatcher notification to detemine status of the offsite power is required so that an evaluation can be completed to determine the offsite power operability status. C. CORRECT, Operation of the unit without automatic voltage control requires the Southeast Area Load Dispatcher be notified immediately to determine offsite power voltage requirements to determine if off site power supplies are operable and the Operations Duty Specialist is required to be notified of the time interval without automatic voltage control within 24 hours. D. Incorrect, The Southeast Area Load Dispatcher is required to be immediately notified so that an evaluation can be performed but the offsite power operability status is determined after the evaluation is complete. The Operations Duty Specialist is required to be notified of the time interval without automatic voltage control within 24 hours. Plausible because the dispatcher notification is required so that an evaluation can be completed so that the offsite power operability status can be determined and the Operations Duty Specialist notification is required within 24 hours. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 81 Tier 1 Group 1 KIA 077 AG2.4.5 Generator Voltage and Electric Grid Disturbances Ability to prioritize and interpret the significance of each annunciator or alarm. Importance Rating: 4.1 14.3 Technical Reference: O-AR-ECB6-A, Electrical Control Board, Rev 38 2-AR-M1-A, Generator and Transformers, Rev 33 GOI-6, Apparatus Operation, Rev. 128 Proposed references to be provided to applicants during examination: None Learning Objective: No Learning Objective identified Question Source: Bank# ___ _ Modified Bank # ___ _ New X ---'------ Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 41.10/43.5/45.3 /45.12 ) 10CFR55.43.b ( 2,5 ) Comments: New question for Sequoyah 2009 exam Page 18 Source Setpoint SER 2743 N/A 1. Carrier Test Signal 27 CC RELAY TEST OR OPERATION (0-6) 2. Carrier Blocking Signal sent or received for any 161kV or 500kV PCB Probable Causes Corrective Actions References 1. Manual actuation of carrier test for any 161 kV on 500kV PCB. 2. Carrier blocking signal sent or received for any 161 kV or 500 kV line. (EM relays only for Bradley line) [1] OBSERVE Line Loading (Amps) and PCB position indicating lights to VERIFY affected PCB has not tripped. [2] IF Line PCB has tripped, THEN OPEN the related MODs. [3] NOTIFY Dispatcher to EVALUATE IF the line needs to De returned to service. [4] IF no PCB has tripped, THEN RESET the Carrier Signal indicating light (light and reset button located on apron of ECB or on Static Relay Panels in Relay Room). 55N651,45B655-ECB6-B SQN O-AR-EC86-8 Page 36 of 44 o Rev. 20 __ . __________ -L ______________ __________ ( 33 (E-5) Source Setpoint SER 3091: RELAY 274VH OP OR FAIL 220 MVAR increasing GEN 2 MVAR ABNORMAL OR MVAR RELAY FAILURE SER 3092: RELAY 274VL OP OR FAIL -100 MVAR decreasing Probable Causes CAUTION NOTE Corrective Actions References 1. High MVAR on Unit 2 main generator. 2. Low MVAR on Unit 2 main generator. 3. Failure of relay 274VH or 274VL. 4. Failure of supervisory MW relay 237W, at low power. Unit 2 main generator must be operated within the limits shown in the generator capability curve (TI-28, figure A.14). This alarm is automatically defeated when Unit 2 output < 240 MWE by relay 237W. [1] CHECK Unit 2 MVARs indicated on 2-EI-57-8 (2-M-1). [2] CHECK CRT SER point to determine which relay actuated. [3] IF Unit 2 MVARs out of limit, THEN [a] ADJUST voltage USING Unit 2 generator voltage adjust and/or intertie transformer tap changer to restore MVARs to within limit. [b) REFER to GOI-6 and 15E500 Sheet 3. [c] EVALUATE off-site power sources for Tech Spec LCO 3.8.1.1. [d] IF attempts to restore Unit 2 MVARs to within limit are unsuccessful, THEN CONTACT Dispatcher for assistance. [4] IF Unit 2 MVARs are within limits, THEN CONTACT Transmission Power Supply (TPS) to investigate suspected failure of relay 274 VH or 274 VL. [5] IF Unit 2 output < 240MWe, THEN CONTACT Transmission Power Supply (TPS) to investigate suspected failure of supervisory MW relay 237W. 55N652-1, 1,2-458655-ECBBA-O, 2-45W541, 55N715-1 SQN O-AR-ECB6-A Page 42 of 45 0 Rev. 38 Source SER 1545 94RB relay 260 relay Probable Causes Corrective Actions CAUTION 3 (A-3) Setpoint N/A GEN VOLT REGULATOR TRIP 1. Loss of 30 Regulating Potential. 2. Exciter Field Breaker (41) OPEN. 3. Volts-Hertz Relay Operation. 4. Generator overexcitation relay operation. [1] CONFIRM alarm by verifying [2-HS-57-20] Exciter Regulator Control Switch green light LIT. If the exciter field breaker opens with the generator tied to grid, rapid and excessive heating of the stator will occur. If this condition exists, the generator should trip by 221 GB relay operation. [2] IF Exciter Field Breaker is TRIPPED, AND PCB's CLOSED, AND Generator NOT TRIPPED, THEN PERFORM the following: [a] IF Reactor Power greater than 50% (P-9), THEN TRIP the Reactor, AND GO TO E-O, Reactor Trip or Safety Injection. [b] IF Reactor Power is less than 50% (P-9), THEN TRIP the turbine, AND GO TO AOP-S.06, Turbine Trip. [3] PLACE [2-HS-57 -20] Exciter Regulator Control Switch to OFF position. [4] IFWindow A-6, GENERATOR EXCITER FIELD OVERCURRENT is in alarm, THEN GO TO Window A-6 of this Instruction. [5] IF Window B-2, GENERATOR VOLTS PER CYCLE HIGH is in alarm, THEN GO TO Window B-2 of this Instruction. (step continued on next page) SQN 2 \2-AR-M1-A Page 6 of 53 Rev. 33 ( CONTINUED CAUTION Corrective Actions (Continued) References . 3 (A-3) GEN VOLT REGULATOR TRIP Operation without automatic voltage regulator may impact offsite power voltage requirements. [6] IF Unit 2 main generator remains in service without automatic voltage regulator, THEN PERFORM the following: [a] NOTIFY SELD to evaluate off-site power voltage requirements with Unit 2 voltage regulator in MANUAL. [b] MAINTAIN Unit 2 MVARs within limits specified in GOI-6, Apparatus Operations, Section E, Turbogenerator Operations. [7] CHECK for blown Voltage Regulating PT fuses with [2-HS-57 -15] Generator Voltmeter Selector Switch. [8] NOTIFY Operations Duty Specialist (ODS) within twenty four (24) hours of any time interval without automatic voltage control. 45B655-01A-O, 45N573-1, 45N551 SQN 2 !2-AR-M1-A Page 7 of 53 Rev. 33 ( \ SQN APPARATUS OPERATIONS GOI-6 Rev: 128 Page 48 of 174 SECTION E Page 2 of 4 3.0 MVAR LIMITS FOR GENERATOR STABILITY (REFERENCE USE) NOTE Operation of main generator without automatic voltage control could impact grid voltage requirements. SELD should be notified immediately if automatic voltage regulator is lost. Studies show that there is some risk of instability in the event of a fault at high side of SON 500/161 kV Intertie Transformer Bank plus a failure of a high side breaker to clear. Backup breakers would then take the entire 500kV bus section out of service. This double-fault event is not postulated to occur simultaneously with a LOCA and is therefore not a scenario used to determine nuclear offsite power adequacy. This is an issue related to grid reliability only and operating guidelines to ensure stability are included in this document for convenience. SON Units 1 and 2 must observe generation limits under certain grid conditions in order to ensure stability under the above double-fault scenario. Both Units are limited to a Maximum Outgoing Reactive Load of 240 Mvar. This limitation supports offsite power source qualification. Transmission Reliability Organization's SON Grid Operating Guide directs that the Transmission Operator will notify the SON Generator Operator of any Mvar Limits recommended. Grid stabilization following the loss of an element depends on the coordination of multiple changes including SON reactive loading. Real time information on the factors affecting grid stability is not available to SON Operators, therefore the Transmission Operator will coordinate the effort. The limits provided in the following tables are for information only. ( SQN APPARATUS OPERATIONS GOI-6 Rev: 128 Page 53 of 174 SECTION F Page 3 of 9 3.1 Offsite Power Source Requirements (continued) D. The plant will coordinate and communicate with the SELD for entry into and exit out of the alternate alignment so the transmission operators will know which criteria to use in monitoring Sequoyah Nuclear Plant offsite power adequacy. E. The SON 161 kV normal scheduled voltage is 165kV, +/-1 kV. The voltage may be increased up to 168kV if necessary due to light load conditions. F. The 161kV switchyard undervoltage relay is set to alarm at 164KV to ensure the minimum 161 kV grid level required to maintain a minimum of 6,560 volts at the 6,900 volt shutdown boards for design basis trips. G. The 500kV bus voltage should be maintained at a level of 525kV. This should be done by means of the Generator No.1 reactive control coordinated with the load tap changer for the SON 500/161 kV Intertie Transformer Bank while adhering to the 161 kV bus voltage schedule. During emergencies or abnormal conditions, the 500kV bus voltage may be raised as coordinated with the power system dispatcher, but it should not exceed 535kV. H. The load tap changer (L TC)on the high side winding of the intertie transformer, 161 KVcapacitor banks, and reactive output of the units shall be coordinated to maintain the published voltage schedule. [C.3] I. If, for any reason, the voltage schedule cannot be maintained, the Southeast Area Load Dispatcher (SELD) should be notified as soon as possible to evaluate offsite power operability. This notification shall include the time and date of the start of the inability to maintain the voltage schedule, an explanation of the problem and the time of anticipated return to compliance. J. Operation of main generator without automatic voltage control could impact grid voltage requirements. The Load Dispatcher should be notified immediately if generator is in service without automatic voltage regulator. Also, refer to Section E for Mvar limits. 29 (E-1) Source Setpoint SER 2683 * Any actuation signal for Oscillograph No.1 or No.2 NfA OSCILLOGRAPH OPERATION OR FAILURE * Loss of control power to Oscillograph No.1 or No.2 Probable Causes Corrective Actions References 1. Fault condition on a transmission line. 2. Fault condition at a remote yard or substation. 3. 500KV or 161 KV bus voltage drop. 4. Unit trip. 5. Loss of control power to either Oscillograph. 6. Oscillograph sending a fax. NOTE: If alarm asa result of Oscillograph sending a fax, NO operator action required. [1] CHECK for other annunciators lit or breaker.cJisagreement lights lit. [2] IF Reactor Trip, THEN GO TO E-O, Reactor Trip or Safety Injection. [3] DISPATCH operator to the relay room to check for any relay targets, AND WRITE down relay targets dropped. [4] NOTIFY Dispatcher of Oscillograph operation and targets noted in step [3]. 45B655-E CB6A-O 55N634-1,2 55N3763-1, 2,3,4 SQN 0 O-AR-ECB6-A Page 37 of 45 Rev. 38 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 82. 068 AG2.4.7 082 Given the following: Conditions require that the Main Control Room (MCR) be abandoned. While completing the MCR actions of AOP-C.04, "Shutdown From Auxiliary Control Room", the Unit 2 OATC notices that a Safety Injection (SI) has occurred. -All AOP-C.04 actions in the MCR were completed. -The crew establishes control in the Auxiliary Control Room and determines that the plant is to be cooled down to Mode 5. Which ONE of the following identifies ... (1) how the safety injection termination will be performed and (2) the instruments on 2-L-10 used to trend the RCS cooldown rate if the RCPs remain out of service? A. (1) AOP-C.04 directs the use of ES-1.1, SI Termination, to terminate the SI. (2) The Thot instruments on each loop. B. (1) AOP-C.04 directs the use of ES-1.1, SI Termination, to terminate the SI. (2) The SG Main Steam Pressure instrument on each loop. C. (1) SI termination steps are contained within AOP-C.04. (2) The Thot instruments on each loop. D!' (1) SI termination steps are contained within AOP-C.04. (2) The SG Main Steam Pressure instrument on each loop. Page 19 ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 20 A. Incorrect, The SI will be terminated using AOP-C.04 Appendix F, Terminating SI flow not by the use of ES-1.1, SI Termination and with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments which have a saturation temperature scale on them. Thot can only be used when the RCPs are in service. Plausible because procedure ES-1.1, SI Termination, is the procedure used for SI termination in many conditions and Thot could be used to monitor cooldown if the RCPs were in service. B. Incorrect, The SI will be terminated using AOP-C.04 Appendix F, Terminating SI flow not by the use of ES-1.1, SI Termination. Since all MCR actions have been completed, the reactor coolant pumps are shutdown resulting in the steam generator pressure instruments being used to trend the RCS cooldown to be correct. Plausible because procedure ES-1.1, SI Termination, is the procedure used for SI termination in many conditions and the steam generator pressure instruments are used to monitor cooldown. C. Incorrect, The SI will be terminated using AOP-C.04 Appendix F, Terminating SI flow but with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments which have a saturation temperature scale on them. Thot can only be used when the RCPs are in service. Plausible because the procedure used for SI termination is correct and Thot could be used if the RCPs were in service. D. CORRECT, The SI will be terminated using AOP-C.04 Appendix F, Terminating SI flow and with all MCR actions completed, the reactor coolant pumps are shutdown leaving the RCS cooldown to be trended by the SG pressure instruments which have a saturation temperature scale on them. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 82 Tier 1 Group 2 KIA 068 AG2.4.7 Control Room Evacuation Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Importance Rating: 4.2/4.2 Technical Reference: AOP-C.04, Shutdown From Auxiliary Control Room, Rev 16 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271-C.04 B.5 & 10.b Question Source: Describe the actions that must be taken prior to abandoning the main control room, including a basis for each actions. Describe actions per AOP-C.04, that are required to: b. Cooldown from Aux Control Room. Bank# ___ _ Modified Bank # ___ _ New X ---Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 41.10/43.5/45.12 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 21

Page 1 of 1 file:III:\Sequoyah\Control Room Photos\2-L-1O\2-L-1O-01.JPG 07/2312008 ( SQN SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 Rev. 16 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment NOTE EOPs are NOT applicable when evacuating MCR. 2. ENSURE reactor TRIPPED. [M-4] 3. ENSURE MSIVs and MSIV bypass valve handswitches in CLOSE. [M-4] 4. PLACE RCP handswitches

in STOP/PULL

TO LOCK. [M-5] 5. ENSURE one CCP placed in PULL TO LOCK. 6. IF MCR must be evacuated due to life-threatening conditions, THEN PERFORM the following: a. EVACUATE MCR on affected unit(s). h. NOTIFY AUOs of MeR evacuation using radio or PAsystem. c. GO TO Cautbn prior to Step 11. ._. ___ ._""' ... _ .. _ ....... ___ .... *_.-=_** _'-_______________ . Page 5 of 185 SQN STEP SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 Rev. 16 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd) NOTE 1 NOTE Accidents requiring containment spray or ECCS operation are outside scope of this procedure. If containment spray or ECCS is running during MCR abandonment event, it is assumed that these pumps are not needed and should be stopped to prevent RWST depletion and/or pressurizer overfill. Checklist 5 (Unit 1) or 6 (Unit 2) directs local operator to ensure CCPIT valves closed from Rx MOV Boards within 13 minutes. 21. CHECK SI signal NOT actuated: IF SI signal has actuated, THEN * NO indication of SI actuation prior to leaving MCR * NO reports of SI or RHR pump breakers CLOSED. PERFORM Appendix F, Terminating SI Flow. Page 13 of 185 ( SQN STEP SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 Rev. 16 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd) 24. CHECK unit in Mode 3 (greater than 350°F). IF unit in Mode 4, 5, or 6, THEN NOTE 1 NOTE 2 NOTE 3 GO TO Step 26. S/G pressure indicators on L-10 are scaled to correlate S/G pressure to T -sat. S/G T-sat indicates approximate RCS T-cold. Atmospheric relief controllers should be set for 85% in AUTO to maintain T-cold at approximately 547°F. Local control of S/G #1 and 4 atmospheric relief valves may be required if essential air has been lost. 25. CONTROL RCS temperature: a. ENSURE S/G atmospheric relief valves maintaining RCS T-cold at desired value (540-550°F following trip). b. GO TO Step 27. Page 18 of 185 a. OPERATE S/G #1 and 4 atmospheric relief valves locally as necessary: (60 minutes) 1 ) DISPATCH personnel to perform Appendix K, Local Control of S/G Atmospheric Reliefs. 2) PLACE S/G #1 and 4 atmospheric relief valve controllers in MANUAL and ADJUST controller output to zero. OPL271 AOP-C.04 Revision 2 Page 3 of 27 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-C.04, SHUTDOWN FROM AUXILIARY CONTROL ROOM IV. LENGTH OF LESSON/COURSE: 3 hour(s) V. TRAINING OBJECTIVES: O. 1. 2. 3. 4. 5. 6. 7. 8. A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-C.04, SHUTDOWN FROM AUXILIARY CONTROL ROOM. B. Enabling Objectives: Objectives Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with Shutdown from the Auxiliary Control Room that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A. State the purpose/goal of this AOP-C.04. Describe the AOP-C.04 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-C.04 entry conditions. b. Describe the ARP requirements associated with AOP-C.04 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-C.04 entry conditions. d. Describe the plant parameters that may indicate a Shutdown from the Auxiliary Control Room is required. Upon entry into AOP-C.04, diagnose the applicable condition arid transition to the appropriate procedural section for response. Summarize the mitigating strategy for the failure that initiated entry into AOP-C.04. Describe the actions that must be taken before abandoning the main control room, including a basis for each action. Explain the staffing requirements for unit abandonment per AOP-C.04. Describe the types of equipment that are on the various checklists associated with AOP-C.04 Describe the actions that may be necessary if procedure steps are taken before all checklists are complete. 9. 10. 11. 12. 13. 14. 15. Objectives Describe the bases for the limits, notes, cautions of AOP-C.04. Describe actions per AOP-C.04, that are required to: a. Maintain Plant in Hot Shutdown b. Cooldown plant form Aux. Control Room c. Return to Main Control Room OPL271 AOP-C.04 Revision 2 Page 4 of 27 Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Given a set of initial plant conditions use AOP-C.04 to correctly: a. Recognize entry conditions. b. Identify required actions. c. Respond to Contingencies. d. Observe and Interpret Cautions and Notes. Describe the Tech Spec and TRM actions applicable during the performance of AOP-C.04. Discuss the parameters to be considered by the SED when making a REP classification during a control room evacuation. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 83. 076 AA2.04 083 Given the following conditions: Unit 1 is in Mode 3 with RCS at normal operating pressure and temperature awaiting secondary plant equipment repair to continue the startup. -At 1300 on 01/25109 RCS Activity was determined to be 0.38 microcuries/gram DOSE EQUIVALENT 1-131. -At 1300 on 01/27/09 Chemistry reports that the RCS Activity has been on a continuous slow increase and is now 0.43 microcuries/gram DOSE EQUIVALENT 1-131. Which ONE of the following identifies actions that are required by 1900 on 01/27/09 and the bases for the RCS Specific Activity limit? A. Reduce RCS Tavg below 500°F to limit doses at the site boundary in the event of a LOCA in conjunction with 0.25La leakage from containment. Reduce RCS Tavg below 500°F to limit doses at the site boundary in the event of a SGTR in conjunction with steady state SG tube leakage of 1 gpm. ( C. Reduce RCS Tavg below 350°F to limit doses at the site boundary in the event of a LOCA in conjunction with 0.25La leakage from containment. D. Reduce RCS Tavg below 350°F to limit doses at the site boundary in the event of a SGTR in conjunction with steady state SG tube leakage of 1 gpm. Page 22 ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Page 23 DISTRACTOR ANAL YSIS: A. Incorrect, reducing Tavg below 500°F is correct but the bases is not due to a LOCA with assumed containment leakage. Plausible because the action stated is correct and a LOCA with leakage from containment could cause elevated doses at the site boundary. B. CORRECT, with the activity above the 0.35 microcurieslgram limit in the Tech Spec 3.4.8 for 48 continuous hours, Tavg is required to be reduced to less than 500° F within 6 hours in accordance with the Tech Spec. The TIS bases states that reducing Tavg below 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The limit on activity is based on the resulting 2-hour doses at the site boundary not exceeding a small fraction of the 10 CFR 100 limits following a SGTR in conjunction with an assumed steady state SG tube leak of 1 gpm. C. Incorrect, reducing Tavg below 350°F is not correct and the bases is not due to a LOCA with assumed containment leakage. Plausible because lowering Tavg to 350°F would mean changing to Mode 4 within the next 6 hours (which is a directed action in many TIS) and a LOCA with leakage from containment could cause elevated doses at the site boundary. D. Incorrect, reducing Tavg below 350°F is not correct but the bases is being to limit doses in the event of a SGTR is correct. Plausible because lowering Tavg to 3500F would mean changing to Mode 4 within the next 6 hours (which is a directed action in many TIS) and the bases is to limit doses at the site boundary during SGTR accident. ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 83 Tier 1 Group 2 KIA 076 AA2.04 Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: Corrective actions required for high fission product activity in RCS Importance Rating: 2.8 1 3.4 Technical Reference: Technical Specifications 3.4.8 and Bases, Amendments 301 and 305. Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-R06 B.9 Question Source: Describe the Tech Spec and TRM actions applicable during the performance of AOP-R06. Bank# ___ _ Modified Bank # X'--__ Question History: New __ _ SQN questions AOP-R06-B.2 001 and AOP-R06-B.9 001 combined and modified for SQN 1/2009 exam. Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 43.5 1 45.13 ) 10CFR55.43.b ( 2 ) Comments: SQN questions AOP-R06-B.2 002 and 0 AOP-R06-B.9 001 combined and modified. Page 24 ( REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: a. Less than or equal to 0.35 microcuries/gram DOSE EQUIVALENT 1-131, and b. Less than or equal to 100/E microcuries/gram. APPLICABILITY: MODES 1, 2, 3, 4 and 5 ACTION: MODES 1, 2 and 3* a. With the specific activity of the primary coolant greater than 0.35 microcuries/gram DOSE EQUIVALENT 1-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T avg less than 500°F within 6 hours. LCO 3.0.4.c is applicable. b. With the specific activity of the primary coolant greater than 100jE microcuries/gram, be in aUeast HOT STANDBY with Tavg less than 500°F within 6 hours. MODES 1, 2, 3, 4 and 5 a. With the specific activity of the primary coolant greater than 0.35 microcuries/gram DOSE EQUIVALENT 1-131 or greater thaif 100jE microcuriesfgram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. *With Tavg greater than or equal 500°F. SEQUOYAH -UNIT 1 3/44-19 April 11 , 2005 Amendment No. 36,117,237,301 ( REACTOR COOLANT SYSTEM BASES I 3/4A8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-ta-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Sequoyah Nuclear Plant site, such as site boundary and meteorological conditions, were not considered in this evaluation .. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.35 microcuriesigram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER Operation with specific activity levels exceeding 0.35 microcuriesigram DOSE EQUIVALENT 1-131 but within the limits shown on Figure 3.4-1 should be limited to no more than 800 hours per year since the activity levels allpwed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. A Note permits the use of the provisioris of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific. activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operations .. , ., Reducing Tavg to less than 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pres5'ureof the atmospheric steam reliefvalves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected insufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomefla. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. SEQUOYAH -UNIT 1 . December 28, 2005 Amendment No. 117, 237. 301 305 AOP-R.06-B.2 001 QUESTIONS REPORT for BANK SQN Questions Given the following plant conditions: Unit 1 has been at 90% power for 5 days . -At 1300 on 8/15/03 ReS Activity was determined to be 0.5 microcuries/gram DOSE EQUIVALENT 1-131 -At 1330 on 8/17/03 Chemistry informs you RCS Activity has increased to 55 microcuries/gram DOSE EQUIVALENT 1-131. Which ONE (1) of the following identifies the required action for this condition? A'! Immediately initiate a plant shutdown and reduce RCS T avg below 500 0 F by 1900 on B. Increase fre-quency of ReS sampling and analysis for RCS activity to once every 2 hours. C. If RCS activity remains unchanged, at 1300 on 8/17/03 initiate a plant shutdown to HOT STANDBY. D. Reduce power level to 50% of rated and have Chemistry re-sample the RCS for specific activity. KIA: 000076SG08 [2.8/3.5] Reference: Tech Spec LCO 3.4.8 Objective: OPL271 C370, B.2 History: i\lew Level: fI.nalysis Note: Provide a copy section 3/4.4.8 as an attachment to the exam. Note: Make SHO only question, developed replacement for RO exam -PEH 8/8/97 Wednesday, September 03, 20088:33:05 AM 3 ( AOP-R.06-B.9 00 I QUESTIONS REPORT for BANK SQN Questions Which one of the following describes the basis for LCO 3.4.8, RCS Specific Activity? Ensures that the resulting doses at the site boundary will not exceed a small fraction of 10 CFR 100 limits following a __ _ A. LOCA in conjunction with 0.25 La leakage from containment. B:t SGTR in conjunction with steady state S/G tube leakage of 1 gpm. C. steam line break in conjunction with steady state S/G tube leakage of 1 gpm. D. LOCA in conjunction with 0.25 La leakage from secondary containment enclosure boundary. Justification: A. Incorrect. is the bases for LCO 3.6.1, Primary Containment Integrity. Plausible because this is a steaming path that would affect dose rate at the site boundary. B. Correct. C. Incorrect. This is part of the assumptions (MSLB with S/G tube leakage) made for S/G Operational Leakage. Plausible because this is a steaming path that would affect dose rate at the site boundary. D. Incorrect.

(his is the bases for LCO 3.6.2, Secondary

Containment Bypass Leakage. Plausible because this is a steaming path that would affect dose rate at the site boundary. Notes: KIA: Ref: LP/Obj: History: Level: Est Time: Comment: 076A1<3.05 O?13AK3.06 076(32.2.25 [2.9/3.6] [3.2/3.8] [2.5/3.7]

PL271 AOP-R.06, Obj 9 9/07 -New min Wednesday, September

02 20n8 8:33:05 AM {41.5,41.10} {41.5,41.10} {41.5} 4 ( OPL271 AOP-R.06 Revision 0 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R.06 HIGH RCS ACTIVITY IV. LENGTH OF LESSON/COURSE: 1.0 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-R.06 HIGH RCS ACTIVITY. B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with High RCS Activity that are rated z 2.5 during Initial License Training and z 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of AOP-R.06. 2. Discuss the AOP-R.06 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-R06 entry conditions. b. Describe the ARP requirements associated with AOP-R06 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-R.06 entry conditions. d. Describe the Administrative conditions that require Turbine Trip/ Reactor trip due to Reactor Coolant Pump Malfunctions. 3. Describe the initial operator response to stabilize the plant upon entry into AOP-R06. 4. Upon entry into AOP-R06, diagnose the applicable condition and transition to the appropriate procedural section for response. 5. Summarize the mitigating strategy for the failure that initiated entry into AOP-R06. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-R.06. 7. 8. 9. 10. OPL271 AOP-R.06 Revision 0 Page 4 of 15 Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Given a set of initial plant conditions use AOP-R.06 to correctly: a. Recognize entry conditions b. Identify required actions c. Respond to Contingencies d. Observe and Interpret Cautions and Notes Describe the Tech Spec and TRM actions applicable during the performance of AOP-R.06. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 84. WIEOl EG2.4.11 084 Given the following: Unit 2 is operating at 100% power when a loss of Train A CCS occurs. -The crew enters AOP-M.03, "Loss of Component Cooling Water", and initiates a Reactor trip. -As the crew is performing the Immediate Operator Actions of E-O, "Reactor Trip or Safety Injection", an automatic Safety Injection occurs. -The crew performs E-O to the last step without identifying a transition. Which ONE of the following identifies both the correct use of the Emergency Procedure and the proper crew action relative to the use of the Abnormal Operating Procedure? Emergency Abnormal Operating Instructions Procedures A. Loop back in E-O to AOP-M.03 can NOT re-perform steps to be implemented in identify a transition. in parallel with E-O. B:' Loop back in E-O to AOP-M.03 can re-perform steps to be implemented in identify a transition. parallel with E-O. C. Transition to ES-O.O, AOP-M.03 can NOT be "Rediagnosis" ,and identify implemented in parallel the proper transition. with ES-O.O, "Rediagnosis". D. Transition to ES-O.O, AOP-M.03 can be identify implemented in parallel the proper transition. with ES-O.O, "Rediagnosis". Page 25 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 26 A. Incorrect, if E-O is performed to the last step, the step will direct a loop back to step 8 so that a transition can be identified either to an accident procedure or to the procedure to terminate the safety injection. EPM-4, User's Guide does permit AOP performance while in the EOP network with the conditions stated in the stem. Plausible because the loop back in E-O is correct and the EOPs do have priority over the AOPs except under certain conditions. B. CORRECT, if a transition is not made while E-O is performed the last step will direct a loop back to step 8 so that a transition can be identified either to an accident procedure or to the procedure to terminate the safety injection. EPM-4, User's Guide, does permit the AOP performance while in the EOP network with the conditions stated in the stem. C. Incorrect, while ES-O.O, Rediagnosis, can be used to determine the correct procedure, it is not applicable until a transition is made from E-O. EPM-4, User's Guide does permit AOP performance while in the EOP network with the conditions stated in the stem. Plausible because ES-O.O, Rediagnosis, can be used to determine the correct procedure under different conditions and the EOPs do have priority over the AOPs except under certain conditions. D. Incorrect, while ES-O.O, Rediagnosis, can be used to determine the correct procedure, it is not applicable until a transition is made from E-O. EPM-4, User's Guide does permit AOP performance while in the EOP network with the conditions stated in the stem. Plausible because ES-O.O, Rediagnosis, can be used to determine the correct procedure under different conditions and EPM-4, User's Guide, does permit AOP performance while in the EOP network with the conditions stated in the stem. ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 84 Tier 1 Group 2 KIA WE01 EG2.4.11 Rediagnosis Knowledge of abnormal condition procedures. Importance Rating: 4.0/4.2 Technical Reference: E-O, Reactor Trip or Safety Injection, Rev 30 EPM-4, User's Guide, Rev 20 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 EPM-4 B.7 & 8 Question Source: Given plant operating conditions, determine if EOP entry conditions have been met and state the resultant appropriate immediate action steps for those conditions. Given plant operating conditions, determine if AOP entry conditions have been met and state the resultant appropriate actions for those conditions. OPL271 EPM-4 B.2.b Discuss the ES-O.O entry conditions. Describe the requirements associated with ES-O.O entry conditions. Bank# ___ _ Modified Bank # ___ _ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for Sequoyah 2009 exam Page 27 SQN EOI EPM-4 PROGRAM USER'S GUIDE Rev. 20 MANUAL Page 63 of 97 3.11.4 Two-Column Pr;ocedure Walkthrough Demonstration (Continued) 3.11.5 K. Step 7 directs the operator to "CHECK SI termination criteria." 1. If all of the criteria in Substeps a through d are met, the operator remains in the AER column and transitions to ES-1.1 to terminate SI. 2. If any of the criteria in Substeps a through d are NOT met, the operator moves to the RNO column and proceeds to Step 8. L. directs the operator to "GO TO E-1, Loss of Reactor or Secondary Coolant." 1. This step ends performance of procedure E-2 with a transition to E-1. 2. The highlighted word "END" centered after the last step emphasizes that the action steps for E-2 are complete. Use of ES-O.O, Rediagnosis A. ES-O.O, Rediagnosis, is unique among the EOPs in that it has no specific transition into it. It is entered strictly based on operator judgment and is applicable only if SI is in progress and E-O has already been performed (diagnostic steps completed and!.!:ansition made to another procedure). ES-O.O should be used when the operator has any concern that he may not be in the right EOP based on plant conditions. This is most likely to happen if multiple accidents occur either simultaneously or sequentially. B. Once entered, ES-O.O will either transition the operator to ECA-2.1, E-1, E-2, or E-3, or will return him to the procedure and step in effect, depending on ,diagnostics done within the procedure. If ES-O.O determines that an operator should be in a certain series of procedures (e.g., E-1 or ECA-1 series), and he is, then he simply returns to the procedure and step in effect. If ES-O.O determines that an operator should be in a certain series of procedures (e.g., E-3 or ECA-3 series), and he is NOT, then he is sent to either E-1 (if he should be in E-1 or ECA-1 series) or E-3 (if he should be in E-3 or ECA-3 series) to enter the appropriate series at the beginning and work his way through the series normally from that point on. SQN EOI EPM-4 PROGRAM USER'S GUIDE Rev. 20 MANUAL Page 65 of 97 3.11.7 Use of AOPs Within the EOP Network A. EOPs have priority over AOPs at all times, except when a reactor trip or safety injection has occurred in conjunction with an Appendix R fire N.08), Control Room abandonment(AOP-C.04), or Loss of all ERCW capability (AOP-M.01). B. AOP performance while in the EOP network is allowable under the following two circumstances: [C.1] 1. AOPperformance is directed by EOPs in effect. 2. AOP performance is deemed necessary by the SM or US to address abnormal plant conditions NOT directly addressed by the EOPs but which have a significant impact on the ability of the EOPs to perform their function (e;g., loss of ERCW, CCS, off-site power, vital instrument power board, etc.) In this case, the following guidelines should be followed: a. Concurrent performance of the EOPs and the AOP should enhance, NOT degrade, the performance of EOPs in progress .. b. Manpower reSources are adequate to allow performing the EOPs and the AOP concurrently. c.* The AOP should be performed using the single perfomer method so the procedure reader remains dedicated to the EOPs in progress, which are mitigative in nature. The SM may elect to deviate from this requirement when in ES-O.1. d. Certain AOPs may be required to be performed concurrently with the EOPs in order for the EOPs to function as intended; for example, loss of CCS, loss of ERCW, loss of air or vital power to equipment important to safety--any of these could have a significant impact on the ability of the EOPs to achieve their goals. e. Upon transition to ES-O.1, the SM will designate the mitigating crew responsibilities as appropriate, based on the events in progress. Normally, the procedure reader and OATC will perform ES-O.1 while the CRO performs the AOP using the single perfomer method. -E-O SON REACTOR TRIP OR SAFETY INJECTION Rev. 30 .. I STEP I I ACTION/EXPECTED RESPONSE 25.' DETERMINE if diesel generators should be stopped: a. VERIFY shutdown boards ENERGIZED from start busses. b. STOP any unloaded diesel generators and PLACE in standby USING EA-82-1, Placing DIGs in Standby. 26. GO TO Step 8. END II ______ a. ATTEMPT to restore offsite power to shutdown boards USING EA-202-1, Restoring Off-Site Power to 6900 V Shutdown Boards. Page 19 of 21 ( SQN LOSS OF COMPONENT COOLING WATER AOP-M.03 Rev. 11 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.3 Train A CCS Header Failure CAUTION: During operation, the Containment Spray Pumps may experience bearing failure after 10 minutes of a loss of CCS cooling. NOTE 1: When the associated TRAIN of CCS is out of service the CCPs, SI Pumps, and RHR Pumps are INOPERABLE for ECCS purposes due to not being able to fulfill it's design function for sump recirculation. LCOs 3.5.2, 3.5.3, 3.6.2.1,3.7.3 should be evaluated and appropriately entered by the SRO. NOTE 2: When CCS is out of service to mechanical seal HXs ONLY, the affected CCPs, SI Pumps, and RHR Pumps have been evaluated to be OPERABLE and AVAILABLE. These pumps can run indefinitely without CCS cooling water to mechanical seal HXs (Ref: PER 72528, DCN Q-11452-A, and RIMS B38941123 802). 1. MONITOR REACTOR COOLANT PUMPS MOTOR THRUST BEARING TEMP HIGH annunciator DARK [M-5B, E-3]. PERFORM the following on the affected Unit: a. IF affected unit in Mode 1 or 2, THEN PERFORM the following: 1) TRIP reactor. 2) STOP RCPs. 3) GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing in this procedure. [C.2] b. ENSURE RCPs are TRIPPED. c. IF in Mode 4, 5, or 6, THEN STABILIZE RCS temperature USING RHR shutdown cooling. Page 14 of 64 PROGRAM: OPL271 EPM-4 Revision 1 Page 3 of 26 OPERATOR TRAINING -LICENSED I. COURSE: LICENSE TRAINING II. LESSON TITLE: III. LENGTH OF LESSON/COURSE: 4-6 hour(s) IV. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of HLC Procedures training, the participant shall be able to explain, using classroom evaluations and/or simulator scenarios, the requirements of EMP-4, EOP-E-O, "User's Guide". B. Enabling Objectives: 1. Determinelidentify the correct procedural application(s) based on the operating procedures network for normal, abnormal, and emergency evolutions. 2. Analyze an EOP layout and determine (according to EPM-4): a. correct procedural layout application; b. if the use of terms is correct (e.g.: Faulted Steam Generator, Shall, Lowering, etc per Appx. B); c. correct use of symbols and icons. 3. Define EOP warnings, cautions, and notes and, given an EOP condition, determine appropriate usage. 4. Compare and contrast event-based emergency/abnormal operating procedures used in parallel with the symptom-based EOPs. 5. Given an example, apply general guidelines, crew roles and responsibilities for EOP procedural use and determine: a. format and use of sequenced and non-sequenced sub steps; b. transition between Action/Expected Response column and the Response Not Obtained column; c. requirements for task completion prior to proceeding to the next action (and how any exceptions are identified); d. requirements for task completion still in progress following transition to another procedure or step; e. actions based on fold-out page use; f. actions based on hand-out page use; g. if EOP termination is appropriate based on given conditions. 6. Identify post-accident instrumentation and determine if its use is required. 7. Given plant operating conditions, determine if EOP entry conditions have been met and state the resultant appropriate immediate action steps for those conditions. OPL271 EPM-4 Revision 1 Page 4 of 26 8. Given plant operating conditions, determine if AOP entry conditions have been met and state the resultant appropriate actions for those conditions. 9. Identify general operating crew responsibilities during emergency operations including appropriate implementation of prudent operator actions. 10. Identify general operating crew responsibilities during emergency operations including requirements for actions outside Technical Specifications/plant licensed conditions (1 OCFR50.54x application). 11. Given a set of conditions, analyze the EOP/FRP implementation: a. identify the basis for the implementation; b. determine the correct implementation hierarchy; c. determine if Critical Safety Function Status Trees (CFSTs) implementation is required; d. identify the status tree colors by priority and summarize each tree's purpose; e. identify conditions which will allow a FRP to be exited once it is entered (a RED or ORANGE condition); f. state the monitoring frequency of CFSTs and when this can be relaxed; g. determine correct coordination with other support procedures h. identify conditions permissible to terminate CFSTs monitoring. 12. Given an operational Situation, analyze a crew brief and determine if it meets Management expectations. II OPL271 ES-O.O Revision 0 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: ES-O.O, "Rediagnosis" IV. LENGTH OF LESSON/COURSE: .5 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of this lesson and others presented, the student shall demonstrate an understanding of the EOP-ES-O.O, "Rediagnosis" by successfully completing a written examination with a score of 80 percent or greater. B. Enabling Objectives O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with ES-O.O, Rediagnosis, that are rated 2 2.5 during Initial License Training and 2 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of ES-O.O. 2. Discuss the ES-O.O entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with ES-O.O entry conditions. b. Describe the requirements associated with ES-O.O entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into ES-O.O. 4. Describe the bases for all limits, notes, cautions, and steps of ES-O.O. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use ES-O.O to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of ES-O.O conditions. II Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 85. W/E16 EA2.1 085 Given the following: Unit 1 is at 100% power. -A LOCA occurred inside containment. The crew has just implemented E-1, "Loss of Reactor or Secondary Coolant". The STA has completed the initial performance of the status trees and reports the highest priority path exists on the CONTAINMENT status tree. Containment conditions are as follows: Pressure is 2.6 psig and lowering. Upper containment Rad Monitors read 85 Rlhr. Lower containment Rad Monitors read 125 Rlhr . Containment Sump Level is 58%. Based on the above conditions, the Unit Supervisor ... A. is required to IMMEDIATELY implement and complete FR-Z.2, "Containment Flooding", then transition back to E-1. B. will acknowledge entry criteria for FR-Z.2, "Containment Flooding", is met but entry into the FR is optional. C. is required to IMMEDIATELY implement and complete FR-Z.3, "High Containment Radiation", then transition back to E-1. will acknowledge entry criteria for FR-Z.3, "High Containment Radiation", is met but entry into the FR is optional. Page 28 ( ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 29 A. Incorrect, the level is the sump is below the 68% level required to enter FR-Z. 2, Containment Flooding but if the level was high enough for entry the transition would be required. Plausible because the containment sump level is elevated and if the required level was present an Orange would be present and require immediate transition to FR-Z.2. B. Incorrect, the level is the sump is below the 68% level required to enter FR-Z. 2, Containment Flooding but if the level was high enough for entry the transition would not be optional. Plausible because the containment sump level is elevated and if the required level being met resulted in a yellow path, the entry would be optional. C. Incorrect, the radiation in lower containment is greater than the threshold level for entering FR-Z.3, High Containment Radiation, but the challenge is a Yellow path which allows the performance of the procedure to be optional. Only Red and Orange path challenges are required to be immediately implemented. Plausible because the entry conditions are met and if the challenge had been an Orange path, immediate transition would be required. D. CORRECT, the radiation in lower containment is greater than the threshold level for entering FR-Z.3, High Containment Radiation, and the challenge is a yellow path which allows the performance of the procedure to be optional. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 85 Tier 1 Group 2 KIA W/E16 EA2.1 High Containment Radiation Ability to determine and interpret the following as they apply to the (High Containment Radiation) Facility conditions and selection of appropriate procedures during abnormal and emergency operations. Importance Rating: 2.9 I 3.3 Technical Reference: EPM-4, User's Guide, Rev. 20 1-FR-0, Status trees, Rev. 1 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 EPM-4 B.11 Question Source: Given a set of conditions, analyze the EOP/FRP implementation. OPL271 FR-O B.6 Given a set of initial plant conditions use FR-O to correctly identify the: a. Identify required actions. Bank# ____ _ Modified Bank # X, __ _ New ___ _ Question History: Modified from Braidwood 12-2007 SRO exam Question Cognitive Level: Memory or fundamental knowledge ___ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 43.5/45.13 ) 10CFR55.43.b (5) Comments: Modified from Braidwood 12-2007 SRO exam Page 30 ( I SON EOI PROGRAM MANUAL I USER'S GUIDE EPM-4 Rev. 20 Page 48 of 97 3.10.5 Status Tree Rules of Usage (continued) 6. If any ORANGE challenge is encountered, the person monitoring status trees continues monitoring until all six status trees have been evaluated. This is necessary because a subsequent RED challenge has priority over any ORANGE challenge. If any RED is encountered, then Rule 3.10.5.0.4 applies. Otherwise, once it is determined that no RED challenges exist, then the person monitoring status trees informs the procedure reader of the highest priority ORANGE challenge. 7. RED or ORANGE challenges must be addressed immediately by implementing appropriate FRPs in order of priority and per the rules of usage. When the person monitoring status trees informs the procedure reader that a RED or ORANGE challenge exists, the procedure reader immediately suspends the ORP (or lower priority FRP) in progress and implements the appropriate FRP, as indicated at the terminus point of the CSF under challenge. 8. YELLOW challenges may be addressed by implementing appropriate FRPs if desired, but do not require immediate operator action. Addressing YELLOW challenges is optional since these are usually temporary, off-normal conditions that will be restored to normal status by actions already in progress. In other cases, the YELLOW path might provide an early indication of a developing RED or ORANGE condition. Following FRP implementation, a YELLOW might indicate a residual normal condition. When the person monitoring status trees informs the procedure reader that a YELLOW challenge exists, the procedure reader should evaluate if the YELLOW challenge FRP should be implemented. This decision will be based on the following:

Whether the procedures

in effect will address the challenge as a matter of course. * Whether the procedures in effect are more important at that time based upon available time and current plant conditions.

Whether the challenge

is of a nature that it will likely develop into an ORANGE or RED condition if action is not taken early. CONTAINMENT PRESSURE NO LESS THAN 12.0 PSIG YES CONTAINMENT PRESSURE LESS THAN 2.8 PSIG CONTAINMENT F-O.5 SQN 1-FR-O Rev. 1 R GOTO , , FR-Z.1 NO Ii!r'Ji,UI "'"

GOTO d FR-Z.1 YES NO Immlli CONTAINMENT

SUMP LEVEL LESS THAN 68% YES UPPER AND LOWER CONTAINMENT RADIATION MONITORS LESS THAN 100 RlHR Page 10 of 16 I'b'!lWi:mitI GOTO FR-Z.2 NO F=======I y I YES IIil Iii {oi. il1il CSF SAT Braidwood 12-2007 exam Change to LOCA Quest No: RO SRO: TIER: GROUP: Topic No: KA No: RO: SRO: Cog Level: 83 SRO 1 2 00WE14 00WE14EA2.1 3.33.8 High System/Evolution Name: Category Statement: High Containment Pressure Ability to determine and interpret the following as they apply to the High Containment Pressure: KA Statement: Facility conditions and selection of appropriate procedures during abnormal and emergency operations UserID: Topic Line: Question Stem: Given: -Unit 1 was at 100% power. -All systems were normally aligned. -A large steam break occurred inside containment. -The crew has just implemented 1BwEP-2, FAULTED STEAM GENERATOR ISOLATION. -The STA has completed the initial scan of the status trees and the following conditions exist: -An ORANGE path exists on the containment status tree. -Containment pressure is 26 psig and lowering. -The faulted SG has NOT been isolated. -ALL other CSFs are GREEN. Based on the above conditions, the Unit Supervisor will direct the crew to ... A IMMEDIATELY implement and complete 1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE, THEN transition back to 1BwEP-2. B remain in 1BwEP-2 and continue to monitor containment pressure. If containment pressure begins to rise, THEN implement 1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE. C remain in 1BwEP-2 until the faulted SG is isolated, THEN implement 1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE. D IMMEDIATELY implement 1BwFR-Z.1, RESPONSE TO HIGH CONTAINMENT PRESSURE, and remain in 1BwFR-Z.1 until the containment CSF is restored to GREEN OR YELLOW, THEN transition back to 1BwEP-2. Answer: Task No: S-FR-01S Question Source: Question Difficulty A Obj No: 7D.FR-00SA New Low Time: Cross Ref: 10CFRSS.43(b)(S) 1 Reference: No reference will be provided to examinee. ILT lesson plan Il-FR-XL-01, BwFR-Z BwAP 340-1, Use of Procedures for the Operating Department 1BwFR-Z.1, Response to High Containment Pressure Explanation: Question meets KA. Question requires examinee ability to determine and interpret facility conditions and select appropriate procedures during high containment pressure. With containment pressure> 20 psig, an orange path conditions exist on the containment status tree. With an orange path present, immediate transition is made to 1BwFR-Z.1 to restore the CSF. Once entered, 1BwFR-Z.1 is entirely completed, then transition is made back to procedure and step in effect at time of transition to 1BwFR-Z.1. A is correct, see explanation above. B is incorrect, even though containment pressure is lowering and nearing point at which CSF will change to yellow status, transition is made to 1BwFR-Z.1. C is incorrect, 1BwFR-Z.1 will perform faulted SG isolation sequence. D is incorrect, 1BwFR-Z.1 is completed and procedure is exited even if CSF is not restored. Date Written: 6/28/2007 Author: Darren Stiles ill App. Ref: none OPL271FR-O Revision 1 Page 3 of 42 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: FR-O, STATUS TREES IV. LENGTH OF LESSON/COURSE: 1 hours V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of FR-O Status Trees. B. Enabling Objectives O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with Status Trees that are rated 22.5 during Initial License Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of FR-O. 2. Explain the bases for prioritizing safety functions during emergency operations. 3. Summarize the mitigating strategy for the failure that initiated entry into FR-O 4. Discuss requirements for monitoring Status Trees. a. Describe the conditions when monitoring is to be initiated . b. Describe the required frequency for monitoring the status trees and how the frequency is determined c. Describe the conditions when monitoring can be terminated. 5. Describe the bases for all decision blocks, limits, notes, cautions, and steps of FR-O. 6. Given a set of initial plant conditions use FR-O to correctly identify the : a. Identify required actions b. Observe and Interpret Cautions and Notes c. Requirements when a RED or ORANGE Path is diagnosed 7. Apply GFE and system response concepts to the performance of FR-O PROGRAM: OPL271 EPM-4 Revision 1 Page 3 of 26 OPERATOR TRAINING -LICENSED I. COURSE: LICENSE TRAINING II. LESSON TITLE: III. LENGTH OF LESSON/COURSE: 4-6 hour(s) IV. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of HLC Procedures training, the participant shall be able to explain, using classroom evaluations and/or simulator scenarios, the requirements of EMP-4, EOP-E-O, "User's Guide". B. Enabling Objectives: 1. Determine/identify the correct procedural application(s) based on the operating procedures network for normal, abnormal, and emergency evolutions. 2. Analyze an EOP layout and determine (according to EPM-4): a. correct procedural layout application; b. if the use of terms is correct (e.g.: Faulted Steam Generator, Shall, Lowering, etc per Appx. B); c. correct use of symbols and icons. 3. Define EOP warnings, cautions, and notes and, given an EOP condition, determine appropriate usage. 4. Compare and contrast event-based emergency/abnormal operating procedures used in parallel with the symptom-based EOPs. 5. Given an example, apply general guidelines, crew roles and responsibilities for EOP procedural use and determine: a. format and use of sequenced and non-sequenced sub steps; b. transition between Action/Expected Response column and the Response Not Obtained column; c. requirements for task completion prior to proceeding to the next action (and how any exceptions are identified); d. requirements for task completion still in progress following transition to another procedure or step; e. actions based on fold-out page use; f. actions based on hand-out page use; g. if EOP termination is appropriate based on given conditions. 6. Identify post-accident instrumentation and determine if its use is required. 7. Given plant operating conditions, determine if EOP entry conditions have been met and state the resultant appropriate immediate action steps for those conditions. OPL271 EPM-4 Revision 1 Page 4 of 26 8. Given plant operating conditions, determine if AOP entry conditions have been met and state the resultant appropriate actions for those conditions. 9. Identify general operating crew responsibilities during emergency operations including appropriate implementation of prudent operator actions. 10. Identify general operating crew responsibilities during emergency operations including requirements for actions outside Technical Specifications/plant licensed conditions (1 OCFR50.54x application). 11. Given a set of conditions, analyze the EOP/FRP implementation: a. identify the basis for the implementation; b. correctimplementatio rl hierarchy; c. determine if Critical Safety Function Status Trees (CFSTs) implementation is required; d. identify the status tree colors by priority and summarize each tree's purpose; e. identify conditions which will allow a FRP to be exited once it is entered (a RED or ORANGE condition); f. state the monitoring frequency of CFSTs and when this can be relaxed; g. determine correct coordination with other support procedures h. identify conditions permissible to terminate CFSTs monitoring. 12. Given an operational situation, analyze a crew brief and determine if it meets Management expectations. ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 86. 003 A2.03 086 Given the following: Unit 1 in Mode 3 with RCS at normal operating temperature and pressure. RCPs #1, #2 and #3 are running. Following the start of RCP #4, the pump stabilizes as follows: -Motor current at 625 amps. -#1 sealleakoff at 1.1 gpm. -All pump and motor temperatures stable and within limits. -The operators implement AOP-R.04, "Reactor Coolant Pump Malfunctions". Which ONE of the following identifies why AOP-R.04 entry was required and the action directed by the AOP? A'I The AOP was entered due to the high motor current and the AOP will direct removal of the RCP from service. B. The AOP was entered due to the high motor current but the AOP will NOT direct removal of the RCP unless bearing or stator temperatures are increasing. C. The AOP was entered due to the low #1 seal leakoff flow and the AOP will direct removal of the RCP from service. D. The AOP was entered due to the low #1 sealleakoff flow but the AOP will NOT direct removal of the RCP unless lower bearing or seal water temperatures are increasing. Page 31 ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 32 A. CORRECT, the AOP entry was required due to the high current flow to the motor and the with the current flow greater than 608 amps, the procedure directs the RCP to be stopped. B. Incorrect, the AOP entry being required due to the high current flow to the motor is correct but with the current flow greater than 608 amps, the procedure does not require the bearing or stator temperatures to be increasing to direct the RCP to be stopped. Plausible because the high motor current is why the procedure entry was required and increasing temperatures do result in stopping the RCP with other conditions in the procedure. C. Incorrect, the AOP entry was not required due to low #1 sealleakoff flow as the leakoff is above the low flow alarm, and if the flow was low with the bearing and seal water temperature stable, the AOP would not require the pump be stopped unless the #1 sealleakoff flow was further reduced. Plausible because low seal leakoff flow would require the procedure to be entered and if the flow were low enough the stopping of the RCP would be required. D. Incorrect, the AOP entry was not required due to low #1 sealleakoff flow but if the seal was less than entry conditions, rising temperature would result in stopping the RCP. Plausible because low sealleakoff flow would require the procedure to be entered and if the lower bearing or seal water temperatures were increasing, the stopping of the RCP would be required. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 86 Tier 2 Group 1 KIA 003 A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems associated with RCP motors, including faulty motors and current, and winding and bearing temperature problems Importance Rating: 2.7/3.1 Technical Reference: AOP-R.04, Reactor Coolant Pump Malfunctions, Rev. 23 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-R.04 B.8.a & b Given a set of plant conditions use AOP-R.04 to correctly: a. Recognize entry conditions b. Identify required actions Question Source: Bank# ___ _ Modified Bank # ____ _ New _X __ Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.5/43.5/45.3/45/13 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 33

__ ---'-_A_O_P_-R_._04---1 L Rev. 23 [ STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS CAUTION: Exceeding the following limitations requires trip of the affected RCP, unless RCP operation is required by FR-C.1, Inadequate Core Cooling or FR-C.2, Degraded Core Cooling: * RCP #1 Seal than 220 psid * Rep #1 Seal Temperature greater than 225°F * RCP Lower Bearing Temperature greater than 225°F * RCP Upper Motor Bearing Temperature greater than 200°F * RCP Lower Motor Bearing Temperature greater than 200°F * RCP Motor Voltage less than 5940V or greater than 7260V * RCP Motor Amps greater than 608 amps * RCP Vibration greater than 20 mils on any axis (x and/or y) [C.3] NOTE 1: During plant startup following seal maintenance, the seal package should seat and operate normally following 24 hours of run time. NOTE 2: RCP trip criteria is also located in Appendix B. This appendix should be referred to throughout the performance of this procedure. 1. DIAGNOSE the failure: IF ... Reactor Coolant Pump(s) tripped or shutdown required RCP #1 Seal Leakoff high flow (high flow Alarm) RCP #1 Seal Leakoff low flow (low flow Alarm) RCP #2 Seal Leakoff high flow (high RCP standpipe level) RCP #3 Seal Leakoff high flow (low RCP standpipe level) RCP Motor Stator Temperature High Page 3 of 34 GOTO SECTION 2.1 2.2 2.3 2.4 2.5 2.6 PAGE 4 7 13 18 21 24 /' rr e SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required CAUTION: A rapid drop in level and steam flow on the affected loop S/G may occur when RCP is tripped. 1. CHECKunit in Mode 1 or 2. GO TO Step 3. NOTE: This procedure is intended to be performed concurrently with E-O, Reactor Trip or Safety Injection. 2. TRIP the reactor, and GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing in this procedure. ---.---3. STOP and LOCK OUT affected RCP(s). Page 4 of 34 ( ( SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required (cont'd) CAUTION: If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur. [C.2] 4. MONITOR Rep seal leakoff less than 8 gpm per pump: * FR-62-24 [RCP 1 & 2] * FR-62-50 [RCP 3 & 4] 5. PULL TO DEFEAT affected loop f.. T and T-avg: * XS-68-2D(f.. T) * XS-68-2M (T-avg) 6. CHECK RCPs 1 and 2 RUNNING. WHEN the RCP has coasted down (30 sec.), THEN CLOSE affected RCP seal return FCV: [C.2] * FCV-62-9 [RCP 1] * FCV-62-22 [RCP 2] * FCV-62-35 [RCP 3] * FCV-62-48 [RCP 4] CLOSE affected loop's pressurizer spray valve. Page 5 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE I* RESPONSE NOT OBTAINED 2.1 Reactor Coolant Pump Tripped or Shutdown Required (cont'd) CAUTION: Restoring seal water injection to a hot seal package could result in failure of the RCP seals. NOTE: The plant should be cooled down to reduce heat input into the pump seal package if RCP seal injection flow has been lost and cannot be restored prior to exceeding temperature limits. 7. IF RCP Seal Temperatures or Bearing Temperatures are increasing uncontrolled due to loss of Seal Injection, THEN EVALUATE initiating RCS cooldown. 8. EVALUATE EPIP-1, Emergency Plan Initiating Conditions Matrix. 9. EVALUATE the following Tech Specs for applicability:

3.2.5, DNB Parameters
3.4.1.1, Reactor Coolant Loops and Coolant Circulation

-Startup and Power Operation

3.4.1.2, Reactor Coolant Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational

Leakage 10. GO TO appropriate plant procedure. ---.---END OF SECTION , Page 6 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 RCP #1 Seal Leakoff High Flow * CAUTION: RCP bearing damage may occur if temperature exceeds 225°F . * CAUTION: If the RCP seal return flow control valve is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur. [C.2] 1. MONITOR #1 seal leakoff less than 6 gpm per pump: * FR-62-24 [Rep 1 & 2] * FR*62**50 [Rep 3 & 4] a. MONITOR Rep lower bearing temperature and seal temperature. IF Rep lower bearing temperature OR seal temperature are rising uncontrolled, THEN GO TO Section 2.1, Rep Tripped or Shutdown Required. [C.1] [C.2] IF lower bearing temperature AND seal temperature indication are NOT available for affected Rep, THEN GO TO Section 2.1, Rep Tripped or Shutdown Required. [C.1] (Step continued on next page.) Page 7 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R04 Rev. 23 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 1. (Continued) b. CHECK #1 seal leakoff flow: IF #1 seal leakoff flow greater than 8 gpm, THEN PERFORM the following: 1) INITIATE plant shutdown at 2-4% per minute USING AOP-C.03, Rapid Shutdown or Load Reduction. 2) WHEN reactor is tripped, THEN GO TO Section 2.1, RCP Tripped or Shutdown Required. [C.1] IF #1 seal leakoff flow less than 8 gpm, THEN PERFORM the following: 1) CONTROL RCP seal injection flow for the affected RCP greater than or equal to 9 gpm. 2) CONTACT Engineering for recommendations WHILE continuing with this procedure. (Step continued on next page.) Page 8 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXP.ECTED RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 1. (Continued) . 2. MONITOR RCP lower bearing and seal water temperatures less than 225°F. Page 9 of 34 RESPONSE NOT OBTAINED 3) IMPLEMENT Engineering recommendations to address specific RCP seal performance conditions. OR COMPLETE normal plant shutdown within 8 hours USING appropriate plant procedure. 4) WHEN reactor is shutdown or tripped, THEN GO TO Section 2.1, RCP Tripped or Shutdown Required. [C.1] ____ ::e--IF any of the following conditions met: * RCP lower bearing temperature or seal water temperature greater than 225°F OR * seal leakoff flow greater than 6 gpm AND lower bearing and seal temp NOT available for affected RCP THEN GO TO Section 2.1, RCP Tripped or Shutdown Required. [C.1] __ ..... ::e-- ( SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 3. MONITOR RCP #1 seal greater than 220 psid: * POI-62-8A

POI-62-21A
POI-62-34A
PDI-62-47A

4. ENSURE RCP seal water supply flow 6-10 gpm per pump: * FI-62-1A * FI-62-14A

FI-62-27A
FI-62-40A

5. CONTACT Engineering for recommendations WH I LE continuing with this procedure. 6. EVALUATE EPIP-1, Emergency Plan Initiating Conditions Matrix. RESPONSE NOT OBTAINED GO TO Section 2.1, RCP Tripped or Shutdown Required. [C.1] IF seal water supply flow is less than 6 gpm AND CANNOT be restored, THEN ENSURE CCS supply to thermal barriers less than 105°F on TR-70-161 [CCS HX 1A1/1A2 (2A1/2A2) Outlet Temp] Page 10 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.2 RCP #1 Seal Leakoff High Flow (cont'd) 7. EVALUATE the following Tech Specs for applicability:

3.2.5, DNB Parameters
3.4.1.1, Reactor Coolant Loops and Coolant Circulation

-Startup and Power Operation

3.4.1.2, Reactor Coolant System -Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational

Leakage RESPONSE NOT OBTAINED CAUTION: Slow and uniform temperature adjustments (approx. 50°F in one hour) will prevent thermal shock to the seals. 8. CHECK VCT outlet temperature less than 130°F [TI-62-131]. 9. ENSURE VCT pressure between 17 psig and 45 psig [PI-62-122]. ADJUST HIG-62 .. 78A to reduce VCT temperature to less than 130°F. Page 11 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.2 Rep #1 Seal Leakoff High Flow (cont'd) 10. CHECK RCPlower bearing and seal water temperature less than 180°F: 11. GO TO appropriate plant procedure. RESPONSE NOT OBTAINED IF any of the following conditions met: * affected RCP lower bearing or seal water temperature greater than 180°F OR * lower bearing and seal water temp indication NOT available for affected RCP, THEN GO TO Step 1. END OF SECTION Page 12 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow 1. CHECK #1 sealleakoff flow greater than 0.8 gpm per pump: * FR-62-23 [RCP 1 & 2] * FR-62-49 [RCP 3 & 4] 2. CHECK WI seal leakoff flow greater than 0.9 gpm per pump and NOT decreasing:

FR-62-23 [RCP 1 & 2] * FR-62-49 [RCP 3 & 4] 3. GO TO appropriate

plant procedure. 4. ENSURE RCP seal water supply flow between 6 gpm and 10 gpm per pump: * FI-'62-1A

FI-62-14A
FI-62-27A
FI-62-40A

GO TO Step 4. GO TO Step 4. IF seal water supply flow is less than 6gpm AND CANNOT be restored, THEN ENSURE CCS supply to thermal barriers is less than 1 05°F on TR-70-161. [CCS HX 1A1/1A2 (2A1/2A2) Outlet Temp] Page 13 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 5. CONTACT Engineering for recommendations WHILE continuing with this procedure. 6. ENSURE VCT pressure between 17 psig and 45 psig [PI-62-122J. 7. CHECK RCP standpipe level alarms DARK [M-5B, A-2, B-2, C-2, 0-2]. RESPONSE NOT OBTAINED MONITOR the following: a. RCDT parameters (O-L-2 AB, el. 669) * Level, U-77-1 * Pressure, PI-77-2 * Temperature, TI-77-21 b. Cntmt FI. & Sump Level rate of rise (ICS pt. U0969) Page 14 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 8. VERIFY RCP #2 sealleakoff less than or equal to 0.5 gpm USING Appendix A, RCDT Level Rate-of-Change. 9. MONITOR RCP lower bearing temperature and seal water temperature are stable and within limits (less than 225°F). RESPONSE NOT OBTAINED GO TO Section 2.4, RCP #2 Seal Leakoff High Flow. IF any of the following conditions met: * affected RCP lower bearing temp or seal water temp rising uncontrolled OR * affected RCP lower bearing temp or seal water temp greater than 225°F OR * affected RCP lower bearing temp and seal temp indication NOT available THEN GO TO Section 2.1, RCP Tripped or Shutdown Required. [C.1] Page 15 of 34 SQN REACTOR COOLANT PUMP MALFUNCTIONS AOP-R.04 Rev. 23 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) CAUTION: If low seal leakoff compensatory actions are NOT successful, seal failure may result as indicated bya sudden increase in seal leakoff flow (greater than 8 gpm). NOTE: Plant shutdown may be terminated if Seal Leakoff flow stabilizes at greater than 0.8 gpm with pump Lower Bearing temperature and Seal Water Temperature remaining stable (no indications of seal failure). 10. MONITOR Rep #1 seal leakoff flow greater than 0.8 gpm: * FR-62-23 [Rep 1 & 2] * FR-62-49 [Rep 3 & 4] 11. CHECK #1 sealleakoff flow greater than 0.9 gpm per pump and NOT

FR-62-23 [Rep 1 & 2] * FR-62-49 [Rep 3 & 4] INITIATE normal plant shutdown USING appropriate

plant procedures AND STOP affected Rep within 8 hours. IF Rep #1 seal leakoff flow reverts to high leakage (greater than 8.0 gpm): * FR-62-24 [Rep 1 & 2] * FR-62-50 [Rep 3 & 4] THEN GO TO Section 2.1, Rep Tripped or Shutdown Required. GO TO Step 1. ..... Page 16 of 34 [S_Q_N_--, ___ __ Rev. 23 [ STEP I ACTION/EXPECTED RESPONSE 2.3 RCP #1 Seal Leakoff Low Flow (cont'd) 12. EVALUATE EPIP-1, Emergency Plan Initiating Conditions Matrix. 13. EVALUATE the following Tech Specs for applicability:

3.2.5, DNB Parameters
3.4.1.1, Reactor Coolant Loops and Coolant Circulation

-Startup and Power Operation

3.4.1.2, Reactor Coolant System -Hot Standby * 3.4.1.3, Reactor Coolant System -Shutdown * 3.4.6.2, RCS Operational

Leakage 14. GO TO appropriate plant procedure. ---.. ---END OF SECTION Page 17 of 34 RESPONSE NOT OBTAINED Source SER 2123 SER 2122 SER 2121 SER2120 Probable Causes Corrective Actions CONTINUED 3 (A-3) Setpoint Pump 1 FS-62-tO Pump 2 FS-62-23 Pump 3 FS-62-36 Pump 4 FS-62-49 .9 gpm decreasing .9 gpm decreasing .9 gpm decreasing .9 gpm decreasing FS-62-10 REAC COOL PMPS SEAL LEAKOFF LOW FLOW 1. No.1 seal less than 275 psid. 2. No.1 seal damage. 3. No.2 seal failure. [1] VERIFY Low Leakoff flow condition on affected RCP(s) with the following instruments: Pump Leakoff Instrumentation RCP 1 1-FR-62-23 RCP 2 1-FR-62-23 RCP3 1-FR-62-49 RCP4 1-FR-62-49 [2] ENSURE No.1 Seal Return Isolation Valves OPEN Pump Valve RCP 1 1-FCV,..62-9 RCP2 1-FCV-62-22 RCP3 1-FCV-62-35 RCP4 1-FCV-62-48 [3] IF Unit 1 isin Mode 1 or 2, THEN GO TO AOP-R04, Reactor Coolant Pump Malfunctions. [4] IF Unit 1 is in Mode 3, 4,or 5, THEN PERFORM the following: [a] VERIFY No.1 Seal 220 psid AND No.1 seal leakoff greater than the minimum value shown in 1-S0-68-2 Appendix D. I SQN 1 1 1-AR':M5-S Page 60t 45 Rev. 36 3 A-3 CORRECTIVE ACTIONS (CONTINUED) References FS-62-10 REAC COOL PMPS SEAL LEAKOFF LOW FLOW [b] ENSURE RCP seal water supply flow is between 6 gpm and 10 gpm per pump. [c] IF No.1 Seal ilP OR No.1 sealleakoff is less than the minimum required values, THEN STOP the affected RCP USING 1-S0-68-2. [d] ENSURE VCT pressure is between 17 psig and 45 psig. [e] CONTACT Engineering for assistance. 458655-058-0,478601-62-2,4,7,9,47W610-68-1 SQN 1 Rev. 36 1 1-AR-M5-S J Page 7 of 45 OPL271 AOP-R.04 Revision 1 Page 3 of 26 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R.04, REACTOR COOLANT PUMP MALFUNCTIONS IV. LENGTH OF LESSON/COURSE: 2 hours V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-R.04, Reactor Coolant Pump Malfunctions. B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with , Reactor Coolant Pump Malfunctions that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of AOP-R.04. 2. Discuss the AOP-R.04 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-R.04 entry conditions. b. Describe the ARP requirements associated with AOP-R.04 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-R.04 entry conditions. d. Describe the Administrative conditions that require Turbine Trip/ Reactor trip due to Reactor Coolant Pump Malfunctions. 3. Describe the initial operator response to stabilize the plant upon entry into AOP-R.04. 4. Upon entry into AOP-R.04, diagnose the applicable condition and transition to the appropriate procedural section for response. 5. Summarize the mitigating strategy for the failure that initiated entry into AOP-R.04. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-R.04. 7. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 8. 9. 10. OPL271 AOP-R.04 Revision 1 Page 4 of 26 Given a set of initial plant conditions use AOP-R.04 to correctly: a. Recognize entry conditions b. Identify required actions c. Respond to Contingencies d. Observe and Interpret Cautions and Notes Describe the Tech Spec and TRM actions applicable during the performance of AOP-R.04. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 87. 026 G2.4.20 087 Given the following: Unit 1 at 100% power with Containment Spray Pump 1 A-A out of service and tagged. -A LOCA results in a Reactor Trip and Safety Injection. -After transferring to the containment sump the crew observes the amps fluctuating on Containment Spray Pump 1 B-B. In response, the crew stops the pump and transitions to ECA-1.3, "Containment Sump Blockage". -The ST A reports a RED path to FR-Z.1, "High Containment Pressure". Which ONE of the following identifies the condition that results in the restart of Containment Spray Pump 1 B-B and what would be the desired flow rate? A. Due to implementing FR-Z.1 , the pump will be restarted and the desired flow rate is the DESIGN spray flow due to the challenge to the Containment Barrier. B. Due to implementing FR-Z.1 , the pump will be restarted and the desired flow rate is the MINIMUM spray flow required to control containment pressure provided it does not cause the RHR pump to cavitate. C. ECA-1.3 will restart the spray pump after TSC evaluation and the desired flow rate is the DESIGN spray flow due to the challenge to the Containment Barrier. D!" ECA-1.3 will restart the spray pump after TSC evaluation and the desired flow rate is the MINIMUM spray flow required to control containment pressure provided it does not cause the RHR pump to cavitate. Page 34 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 35 A. Incorrect, when ECA-1.3 is entered FRGs are monitored for information only and the implementation of FRGs is suspended. Transition to FR-Z 1 will not be made and a note in the ECA Appendix E for throttling spray flow identifies the desired flow rate to be the MINIMUM to control containment pressure without causing cavitation of the RHR pump not the MAXIMUM without causing cavitation. Plausible because FR-Z 1 would restart the spray pump and with a RED path identified the Containment pressure is at or above the design pressure and a severe challenge to the barrier exists and establishing the maximum flow would cause a greater drop in containment pressure. B. Incorrect, when ECA-1.3 is entered FRGs are monitored for information only and the implementation of FRGs is suspended. Transition to FR-Z 1 will not be made. A note in the ECA Appendix E for throttling spray flow identifies the desired flow rate is the MINIMUM to control containment pressure without causing cavitation of the RHR pump. Plausible because FR-Z 1 would restart the spray pump and establishing the minimum flow to control containment pressure without causing cavitation of the RHR pump is correct for the conditions in the stem. C. Incorrect, The FRGs are monitored for information only so ECA-1.3 is continued and the ECA directs the TSC evaluation if containment pressure is greater than 9.5 psig and the pressure would be with a RED path present (12psig). A note in the ECA Appendix E for throttling spray flow identifies the desired flow rate to be the MINIMUM to control containment pressure without causing cavitation of the RHR pump not the MAXIMUM without causing cavitation. Plausible because the ECA is the procedure which restart and control the pump and with a RED path identified the Containment pressure is at or above the design pressure and a severe challenge to the barrier exists and establishing the maximum flow would cause a greater drop in containment pressure. D. CORRECT, A RED path on the Containment Status Tree occurs when the pressure is greater than 12 psig. The FRGs are monitored for information only so ECA-1.3 is continued. The ECA directs the TSC evaluation if containment pressure is greater than 9.5 psig and the pressure would be with a RED path present (12psig). A note in the ECA Appendix E for throttling spray flow identifies the desired flow rate is the MINIMUM to control containment pressure without causing cavitation of the RHR pump. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 87 Tier 2 Group 1 KIA 026 G2.4.20 Containment Spray Knowledge of the operational implications of EOP warnings, cautions, and notes. Ilmportance Rating: 3.8/4.3 Technical Reference: ECA-1.3, Containment Sump Blockage, Rev 1 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 ECA-1.3 B.4 Question Source: Describe the bases for all limits, notes, cautions, and steps of ECA-1.3 Bank# ___ _ Modified Bank # ___ _ New X, __ Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 36 ( SQN CONTAINMENT SUMP BLOCKAGE I STEP] I ACTION/EXPECTED RESPONSE 1. SUSPEND FRP implementation and MONITOR status trees for information only. II RESPONSE NOT OBTAINED Page 3 of63 ECA-1.3 Rev. 1 SQN CONTAINMENT SUMP BLOCKAGE ECA-1.3 Rev. 1 I S'TEP II ACTION/EXPECTED RESPONSE 9. NOTIFY TSC to determine optimum ECCSand containment spray alignment WHILE continuing in this procedure. 10. ENSURE makeup water being added to RWST USING EA-63-2, Refilling the RWST. II RESPONSE NOT OBTAINED CAUTION Re-establishing Containment Spray flow may result in RHR pump cavitation. 11. MONITOR containment pressure: a. CHECK containment pressure less than 9.5 psig. a. NOTIFY TSC to evaluate restarting Containment Spray USING Appendix E, Throttling Containment Spray Flow. Page 14 of63 ( SQN CONTAINMENT SUMP BLOCKAGE* ECA-1.3 Rev. 1 Page 1 of 3 NOTE 1 NOTE 2 APPENDIX E THROTTLING CONTAINMENT SPRAY FLOW This appendix assumes containment spray suction is still aligned to containment sump as specified in ES-1.3. Throttling containment spray flow is desired to allow controlling flow to prevent RHR pump cavitation. Flow rate should be established as directed by TSC. Desired flow is the minimum flow rate needed to control containment pressure without causing RHR cavitation. 1. RESET Containment Spray Signal. 2. STATION operator in communication with MeR at breaker for containment spray discharge valve (identified in Step 6). 3. ENSURE Containment Spray pump recirc valve for train to be started in PULL P-AUTO: * FCV-72-34 (Train A) OR * FCV-72-13 (Train B) 4. ENSURE discharge valve CLOSED for pump to be started: * FCV-72-39 (Train A) 'OR * FCV-72-2 (Train 8) 5. START one containment spray pump. Page 61 of 63 o o o o o o o ( OPL271 ECA-1.3 Revision 0 Page 3 of 22 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: EMERGENCY OPERATING PROCEDURE ECA-1.3, "Containment Sump Blockage" IV. LENGTH OF LESSON/COURSE: 1 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ECA-1.3, "Containment Sump Blockage" B. Enabling Objectives: B. Enabling Objectives: o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with Containment Sump Blockage that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of ECA-1.3. 2. Discuss the ECA-1 .3 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with ECA-1.3 entry conditions. b. Describe the requirements associated with ECA-1.3 entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into ECA-1.3. 4. Describe the bases for all limits, notes, cautions, and steps of ECA-1.3. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use ECA-1.3 to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of ECA-1.3 conditions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 88. 039 A2.01 088 Given the following: Unit 1 is at 100% power when a LOCA occurs. -The reactor trips when the containment pressure rise causes a Safety Injection. -The EOP network is entered. RCS pressure stabilizes at 1580 psig. -Containment pressure rises to 2.4 psig and stabilizes. -The crew is in the process of terminating Safety Injection. -When determining if..Jhe SI pumps should be stopped the following is noted: RCS pressure is now 1540 psig and trending down. RCS subcooling is 43°F. Pressurizer level is 19% and dropping. -SG pressures:

1 -590 psig and dropping.
2 -600 psig and dropping.
3 -580 psig and dropping.
4 -605 psig and dropping.

MSIVs are open. Which ONE of the following identifies the status of the MSIVs and the proper crew response to the conditions? A'I The MSIVs have failed to automatically close. SI Reinitiation Criteria does NOT exist, a transition should be made to E-2, Faulted Steam Generator Isolation. B. The MSIVs have failed to automatically close. SI Reinitiation Criteria exist, restart the CCP, establish CCPIT flow and Go To E-1, Loss of Reactor or Secondary Coolant. C. MSIV automatic closure Signal would NOT have been initiated. SI Reinitiation Criteria does NOT exist, a transition should be made to E-2, Faulted Steam Generator Isolation. D. MSIV automatic closure signal would NOT have been initiated. Page 37 SI Reinitiation Criteria exists, restart the CCP, establish CCPIT flow and Go to E-1, Loss of Reactor or Secondary Coolant. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 38 A. Correct, the MSIVs should have received a signal to automatically close due to SG pressure but the SI Reinitiation Criteria not being met is correct and a transition to E-2 should be made. B. Incorrect, the MSIVs should have received a signal to automatically close and the SI Reinitiation Criteria not being met is correct. Plausible because of the 2 parameters that cause an MSIV to automatically close, one (containment pressure) is identified below the setpoint and the other (SG pressure) is only slightly less than setpoint and if the SI Reinitiation Criteria were applicable, the actions listed and transition to E-1 are correct. C. Incorrect, the MSIVs should have received a signal to automatically close due to pressure in the steam generators. The transition to E-2 is correct. Plausible because of the 2 parameters that cause an MSIV to automatically close, one (containment pressure) is identified below the setpoint and the other (SG pressure) is only slightly less than setpoint and the transition to E-2 is correct. D. Incorrect, the MSIVs having not received a signal to automatically close is correct but the SI reinitiation criteria is not met. Plausible because of the 2 parameters that cause an MSIV to automatically close, one (containment pressure) is identified below the setpoint and the other (SG pressure) is only slightly less than setpoint and if the SI Reinitiation Criteria were applicable, the actions listed and transition to E-1 are correct. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 88 Tier 2 Group 1 KIA 039 A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Flow paths of steam during a LOCA Importance Rating: 3.1 / 3.2 Technical Reference: ES-1.1, SI Termination, Rev 10 1,2-47W611-1-1 R13 TI-28, Attachment 9, Unit 1 and 2 Cycle Data Sheet, Effective Date 06/28/2007 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200.MS B.4.e & i Question Source: Describe the following features for each major component in the Main Steam System as described in this lesson. e. Component operation i. Protective features (including setpoints) OPL271ES-1.1 B.5 Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Bank# ___ _ Modified Bank # ___ _ New X --Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 1 0 CFR Part 55 Content: ( 41.5/43.5/45.3/45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for SQN 1/2009 exam Page 39 A 8 c D E F G H L-L-LL9MLv-Z' L I r-----I I I I I I I I I I I I I I I I I I I I I P7 l-1C STEAl.! GENERATOR LOOP 1 TO SHAM GENERATOR ElLOlfOO\1N (oI\7nl1-'-3 COORO 0-') TRA '" 2 I FCV-'-15 FULLY CLOSED 1 2 P ) 1040 I'-AVTO :5 I 1 :5 4 I OpEN Tit a TEsT I-lS-1-4D})T\ ,----1 I .sAlliE ASI I F'CV-'-17 \ I NOTE 1 H1GH TEMP IN AV)(

j TVRtUNE FlOOI.t is-1-MA c HIGH HIotP IN AUK FEED'IITER

TVFlEUNE ROOIII TS-1-1aa 1 4 VENT I l 5 I 1 5 IN-UANVAL OflEN 6 I fr" ' 1 6 PULL P-AUTO ." CllOsE 1 TO AUXILIA.RY flUllifI TURBINE (4711'611-3-4, D-4, 01iD I ! HI TEt.\1I IN AU)!: FEEr:t1¥ATE/I TuRSIHE ROOIJi 1S-1-17A HI fEW" IN AVX FEEI::t1¥ATER lUMINE "00f.4 l.!1-'-17B 01iD I ! 9 CONTROL slY IN CLOSE HS-1 -UA ,ST tAU GEN LOOP 3 CONT!tOL s.r IN 'CLOSE HS-1 -.2:9A .sT EIW GEN LOOP "I sEE NOTE * fIIlOM STEAW LOOFI .2 -CONT!tOL SI IN ClOSE 1-iS-1-nA Sl EA\l GEN LOOP ! AS LOOP 1) CONTI NIJ ED ON 47WIJ"-'-,2 FRCt.! STEA'II! CENERATOR LOOfl ::5 tSAI.lE AS LOOP 1) -TYPICAL CONTROL ADMIN CHAN£lc ____ LOCIC lSSUED BY; P.G. TRUDEL ------------------_. DR"'I1Ml:"tEIt'lJlnS IRIlf PROCAD MAINTAINED DRAWING TH1S GONFlCUPlAHDN CtlNnaL DRAlI'llIC [! IlAINTMH[D IIY THE: 5QN CAD UNIT AND [5 PlAIU Of THE TVA ()ATAliA!!:, 47I1S"-O-1.2-----------LOGIG DIAGRA\I IND'EX &-SY'UBOLS 4711810-'-' ,2 * .!l.4--...... DIAGItAU 4711601-1---------------FLO'lf DIAGRA'III FOJ1 ORICINAL INFORMATION IN TITLE SLOCK SEE REVISION 0 MICfiOFIlU COPY. 13 AD\UN CHANGE I.1JB REV CHANGE FIEF PREPARER CHECKER APPROVED DATE NONE POWERHOUSE UNITS 1 &: 2 MECHANICAL LOGIC DIAGRAM EXCEPT AS NOTED CATEGORY 1 MAIN AND REHEAT STEAM SEQUOYAH NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY NUt:LENt ENI;[NEEIIHNC J A 8 c D E F ( SQN UNIT 1 & 2 CYCLE DATA SHEET TI-28 Att. 9 Unit 0 {FOR INFORMATION ONLy} Effective Date: 06-28-2007 Page 7 of 16 Signal Setpoint Logic Block/Permissive SAFETY INJECTION SIGNALS (SIS) 1. Containment Press Hi 1.54 psid 2/3 PTs None 2. Pressurizer Press. Low 1870 psig 2/3 PTs Manual Below P-11 3. Steamline Press Low 600 psig 2/3 PTs on 1/4 loops Manual Below P-11 4. Manual N/A 1/2 HS's SIRESET AFTER SIINITIATION, MUST WAIT FOR 60 SECOND TIMER TO RESET. THEN THE SI RESET PB FOR EACH TRAIN MUST BE ACTUATED. THIS WILL BLOCK ANY AUTOMATIC SI ACTUATION SIGNAL BUT MANUAL SIIS NOT BLOCKED. TO REMOVE AUTO SI BLOCK, THE RX TRIP BREAKERS MUST BE CYCLED TO REMOVE THE P-4 SEAL-IN SIGNAL. CONTAINMENT ISOLATION SIGNALS (CIS) Phase A 1. SIS Any Signal 2. Manual Phase AI CVI HS 1/2 CONTAINMENT ISOLATION SIGNALS (CIS) Phase B 1. Containment Press Hi-Hi 2.81 psid 2/4 2. Manual Phase B Handswitch 2/2 CONTAINMENT VENT ISOLATION SIGNALS (CVI) 1. RM-90-130 & 131 High Rad Signal 1/2 2. SIS Any Signal 3. Manual Phase B Phase B Handswitch 2/2 4. Manual Phase A Phase A Handswitch 1/2 CONTAINMENT SPRAY ACTUATION SIGNALS 1. Containment Press Hi-Hi 2.81 psid 2/4 2. Manual Phase B Handswitch 2/2 MAIN STEAMLINE ISOLATION SIGNALS 1. Containment Press Hi-Hi 2.81 psid 2/4 PTs 2. Steam line Press Low 600 psig 2/3 PTs on 1/4 loops Manual Below P-11 3. Steamline Press Negative 100 psig decreasing in 2/3 PTs on 1/4 loops Enabled only when Rate a 50 second time Steamline Press SI constant signal blocked. FEEDWATER ISOLATION SIGNALS 1. S/G Level Hi-Hi 81% (P-14) 2/3 L Ts on any S/G 2. Rx Trip (P-4) with Lo T ave Rx Trip Bkrs Open 550'F 2/4 loops 3. SIS Any signal I SQN SI TERMINATION I ES-1.1 . ________ ______ __________________________ I STEP II ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED NOTE RCS pressure may be slowly dropping due to spray bypass flow or slight cooling of the pressurizer; This should be considered "stable" pressure. 10. DETERMINE if SI pumps should be stopped: a. CHECK RCS pressure:

RCS pressure STABLE or RISING * RCS pressure greater than 1500 psig. b. STOP SI pumps, and PLACE in A-AUTO. a. IF NO S/G is Faulted, THEN GO TO ES-1.2, Post LOCA Cooldown and Depressurization.

---.---IF any S/G is Faulted, THEN PERFORM the following: 1) DO NOT CONTINUE this procedure UNTIL Faulted S/G depressurization stops OR criteria for stopping SI pumps are satisfied. 2) IF criteria for stopping SI pumps CANNOT be satisfied after Faulted S/G depressurization stops, THEN GO TO ES-1.2, Post LOCA Cooldown and Depressurization. ___ . ' b. IF pump(s) CANNOT be stopped in A-AUTO, THEN PLACE *affected SI pump(s) in PULL TO LOCK. Page 10 of 27 ( SQN SI TERMINATION FOLDOUT PAGE SIREINITIATION CRITERIA IF SI has been terminated (CCPIT isolated, 81 pumps stopped, and RHR pumps NOT running in ECCS mode) AND either of the following conditions occurs: * RCS subcooling based on core exit TICs less than 40°F OR * Pressurizer level CANNOT be maintained greater than 10% [20% ADV], THEN a. ESTABLISH ECCS flow by performing one or both of the following:

ESTABLISH

CCPIT flow as necessary USINGAppendix C * START CCPs or Sipumps manually as necessary. b. GO TO E-1, Loss of Reactor or Secondary Coolant. EVENT DIAGNOSTICS

IF both trains of shutdown boards de-energized, THEN GO TO ECA-O.O, Loss of All AC Power. I ES-1.1 Rev. 10 * IF any S/G pressure dropping in an uncontrolled

manner or less than 140 psig AND SIG NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation. TANK SWITCHOVER SETPOINTS

IF CSTlevelless

than 5%, THEN ALIGN AFW suction to ERCW. * IF RWST level less than 27%, THEN GO TO ES.:.1.3, Transfer to RHR Containment Sump. Page 1a of 27 OPL271 ES-1.1 Revision 1 Page 3 of 45 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: ES-1.1, "SI Termination" IV. LENGTH OF LESSON/COURSE: 1 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ES-1.1, SI Termination. B. Enabling Objectives: o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with SI Termination that are rated;:::: 2.5 during Initial License Training and;:::: 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of ES-1.1. 2. Discuss the ES-1.1 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with ES-1.1 entry conditions. b. Describe the requirements associated with ES-1.1 entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into ES-1.1. 4. Describe the bases for all limits, notes, cautions, and steps of ES-1.1. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use ES-1.1 to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply G FE and system response concepts to the performance of ES-1.1 conditions. I. PROGRAM: OPERA TOR TRAINING II. COURSE: SYSTEMS TRAINING III. LESSON TITLE: Main Steam IV. LENGTH OF LESSON: 1 112 HOURS V. TRAINING OBJECTIVES A. Terminal Objective: (included in slides/slide notes) OPT200.MS Rev. 3 Page 3 of 54 Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Main Steam System in the plant and on the simulator. B. Enabling Objectives: (included in slides/slide notes) Demonstrate an understanding ofNUREG 1122 knowledge's and abilities associated with the Main Steam System that are rated 2.5 during Initial License training for the appropriate license position as identified in appendix A. I.State the purpose/functions of the Main Steam System as described in the FSAR. 2.State the design basis ofthe Main Steam System in accordance with the SQNFSAR. 3.Explain the purpose/function of each major component in the flow path of the Main Steam System as illustrated on the simplified system drawing. ( V. TRAINING OBJECTIVES (continued) 4. Describe the following features for each major component in the Main Steam System as described in this lesson. a. Location b. Power supply (include control power as applicable) c. Support equipment and systems d. Normal operating parameters e. Component operation f. Controls g. Interlocks (including setpoints) h. Instrumentation and Indications 1. Protective features (including setpoints) J. Failure modes k. Unit differences 1. Types of accidents for which the Main Steam components are designed m. Location of controls and indications associated with the Main Steam in the control room and auxiliary control room. OPT200.MS Rev. 3 Page 4 of 54 ( Guidance for SRO-only Questions Rev 0 Figure 2: Screening for SRO-only linked to 10CFRSS.43(b)(S) (Procedures) Can question be answered by knowing "systems knowledge", i.e., how the system works, flowpath, Yes 1 logic, location, etc. RO question No 1"1 SJ V $1 <::1(1)17 h J B,.xt P ILC{J:'D +v pc., I /")17 I vv' 4 W',/q UJ 11/9111 \;;) (..() \ 4 N \It Can question be answered by knowing immediate Yes 1 operator actions? '\ RO question >:iNo Can question be answered by knowing entry conditions for major EOPs? Yes I t RO question I No Does the question involve one or more of the following?

* Assessing

plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed Recalling what strategy or action is written into a plant,)'es

1 procedure, including

when the strategy or action is required 1 No Question is not linked to 1 OCFR55.43(b)(5) for SRO-only Page 8 of 19 SRO-only question Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 89. 062 G2 4.18089 Given the following: Unit 1 is operating at 100% power Diesel Generator 1 B-B is out of service and is expected to return to service in 2 hours. Subsequently, the following events occur: -A loss of offsite power occurs. -The reactor trips and the crew enters the emergency procedures. -SI is NOT actuated. -The crew transitions to FR-H.1, "Loss of Secondary Heat Sink" due to a RED Path condition and is performing the first step. No other Status Tree RED paths are present. -A fault on Shutdown Board 1A-A results in the emergency supply breaker from Diesel Generator 1A-A tripping due a differential relay. Which ONE of the following identifies the correct action to be taken and the bases for the action? A. Transition to ECA-O .0, Loss of All AC Power because the ECA will direct actions to establish heat sink with the TD-AFW pump. B. Remain in FR-H.1 unless a higher priority RED path occurs because the FR will direct actions to establish heat sink with the TD-AFW pump. Transition to ECA-O .0, Loss of All AC Power because all other procedures in the EOP network assume a minimum of at least one 6.9kV Shutdown Board is available. D. Remain in FR-H.1 and initiate actions to manually restore 6.9kV Shutdown Board 1 B-B from 2B-B 6.9kV Shutdown Board maintenance breaker because restoring heat sink is the highest priority evolution in progress. Page 40 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA GTOR ANAL YSIS: Page 41 A. Incorrect, The transition to EGA-O.O is the correct action and the EGA will direct actions ensuring the TO AFW pump is in service but the reason is not because the EGA establishes the heat sink. Plausible because the transition is correct and EGA will direct actions to place the TO AFW pump in service and establish a heat sink. B. Incorrect, Remaining is FR-H.1 is not correct, a transition to EGA-O.O is required. Plausible because other EGAs do not take precedence over the FRGs and with a RED path there is a severe challenge to the Heat Sink function that would be addressed if the TO AFW pump were in service. G. GORREGT, The transition to EGA-O.O is the correct action and the reason is because all other procedures, including FR-H. 1, assume a minimum of at least one train of shutdown power is available. O. Incorrect, Remaining is FR-H.1 is not correct, a transition to EGA-O.O is required. Plausible because power could be restored to the 1 B-B board using the maintenance breaker and restoring Heat Sink is critical safety function being challenged. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 89 Tier 2 Group 1 KIA 062 G2 4.18 AC Electrical distribution Knowledge of the specific bases for EOPs. Importance Rating: 3.3/4.0 Technical Reference: 1,2-15E500-1 R26 EPM-3-ECAOO.0, Basis Document for Loss of All AC Power, Rev 10 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 ECA-O.O BA Question Source: Describe the bases for all limits, notes, cautions and steps of ECA-O.O. Bank# ___ _ Modified Bank # X __ _ New __ _ Question History: Question modified from VC SUMMER 2007 SRO exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X. __ 10 CFR Part 55 Content: (41.10/43.1/45.13 ) 10CFR55A3.b ( 5 ) Comments: VC SUMMER 2007 SRO exam question modified Page 42 ( 2007 ill NRC Exam ill SRO 100. Given the following plant conditions:

The plant is operating

at 100% power. * EDG "B" is out of service and is expected to return to service in two (2) hours. * Subsequently, the following events occur: * A loss of offsite power. * The reactor is tripped and the crew enters EOP-l.O, Reactor Trip or Safety Injection.

51 is NOT actuated.
The crew made a transition

to EOP-1S.0, Loss of Secondary Heat Sink, based on a CSFST RED Path. * EDG "A" output breaker subsequently trips on a differentia/lockout on Bus 1DA. Which ONE (1) of the following describes the actions that will be taken and its bases? A. Immediately transition to EOP-6 .0, Loss OfAlI ESF AC Power. All other procedures in the ERG network assume both 7.2 KV ESF busses are available. B. Immediately transition to EOP-6 .0, Loss OfAlI ESF AC Power. All other procedures in the ERG network assume a minimum of ONE (1) 7.2 KV ESF bus is available. C. Remain in EOP-1S.0 until feed is restored and the RED condition is cleared, and then transition to EOP-6.0 , Loss of All ESF AC Power. RED path Function Recovery procedures must be performed until the condition is cleared. D. Remain in EOP-1S.0 until directed to return to procedure in effect, and then transition to EOP-6 .0, Loss of All ESF AC Power. RED path Function Recovery procedures must be finished to completion. SON EOI BASIS DOCUMENT FOR ECA .. O.O EPM .. 3 .. ECA .. 0.0 PROGRAM LOSS OF ALL AC POWER Rev. 10 MANUAL Page 8 of 98 EOP Step Number: 1 SUSPEND FRP implementation and MONITOR status trees for information only. ERG Step Number: 1, Note 2 of 2 CSF Status Trees should be monitored for information only. FRGs should not be implemented. Purpose: To inform the operator that this guideline should not be exited to perform any FRP in response to an identified CSF challenge. ERG Basis: The guideline has priority over all FRGs and is written to implicitly monitor and maintain critical safety functions. This priority is necessary since all FRGs are written on the premise that at least one shutdown board is energized. Knowledge: Guideline ECA .. O.O has priority over the FRGs. EOP Basis: Same. Deviation: Converted note into a step. Justification: Converted note into a step since it contains a hidden action. SON EOP writer's guide disallows hidden actions in cautions and notes. Since the action is required upon entry to ECA .. O.O, it is made an immediate action step. Setpoint: None. / i I l. A B c o E F G H I( 1-00S3S I-Z' I TO 6.9KY co.tIlNBDB (15£500-2) 2 2 :3 :3 4 0) (411N732-1) ! QU!i§l AU?! '9 3M-" 4 5 6 TO 161KV SWl1'CHYARD HI.:52.NC)3DSg- UNIT 1624NC) i -+--'-'":12iZ-)= TO &.tKV -17":2-'!+-)--+";'" NC NC 4aOVUNITIiD UTILlTV BUS Ugh All '9 aM-" 5 6 '-___ TO eel COOL TOWiR f * TRAItSFOlNtR A (45N50I)1 . (15[500*2) '¥b ccu.rJN STATION stRVICE - .1 * X Y 4/32/40 L C !ii6' '=' NC)414 HC)!!12 NO) d l TURS. BLOC fl .-:t ) r-------I I "-" I I I 'XFRS i J 'n'L AUXBO IB1_B HC) Y us'" '0) (45N732-Z1 ! gInn All! R!l !,u_A r --------, -, I s:.!. y) NOTES; 1. 3. INCREASE THE YOND BE 4. X: ONLY ONE BREAI(ER REQUIRED rOR PANELS

.. e. THESE CABLES ARE NOT URhlINATrn

TO THE DIESEL CENERATOR EXC1TER SUPI'OAT[D UMOEIUtEATH THEIR CABLE TRAY 7. THESE CABLES ARE NOT TERhltNAT[O TO THE SHUTDOWN BOARDS. THEY ARE COILED UP 1M NAIiiOLE NUtMI[RS lJA. 13B. 14" ANO 14B * .. 9. CABLE LEADS FROM BREAKER 1120SB TO TIWISFROMER Ct*! HAVE BEtN DISCONNECTED AND ARE HOT TO BE RECONNECTED WITHOUT M.E. APPROVAL. ID. GENERAL CATEGORY 1 KEY DIAGRAM STATION AUX POWER SYSTEM SEQUOYAH NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY OPL271 ECA-O.O Revision 1 Page 3 of 21 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: EMERGENCY OPERATING PROCEDURE ECA-O.O, LOSS OF ALL AC POWER IV. LENGTH OF LESSON/COURSE: 1 hour V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ECA-O.O, LOSS OF ALL AC POWER. B. Enabling Objectives O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with ECA-O.O, LOSS OF ALL AC POWER that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of ECA-O.O. 2. Discuss the ECA-O.O entry conditions. 3. Summarize the mitigating strategy for the failure that initiated entry into ECA-O.O. 4. Describe the bases for all limits, notes, cautions, and steps of ECA-O.O. 5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 6. Given a set of initial plant conditions use ECA-O.O to correctly: a. Identify required actions b. Respond to Contingencies c. Observe and Interpret Cautions and Notes 7. Apply GFE and system response concepts to the performance of ECA-O.O conditions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 90. 076 A2.02 090 Given the following plant conditions: Both Units operating at 100% power. ERCW system in normal alignment. -The following annunciators are LIT on Essential Raw Cooling Water 0-XA-55-27A panel on 1-M-27: -"UNIT 1 HEADER A PRESSURE LOW". -"UNIT 2 HEADER A PRESSURE LOW". -"PUMP J-A DISCH PRESS LOW". -"PUMP Q-A DISCH PRESS LOW". -The following annunciator status on Miscellaneous 1-XA-55-15B panel on 1-M-15: -"ERCW DECK SUMP PUMP A RUNNING" is LIT. ERCW headers 1A and 2A indicating LOW flow. -The crew implements AOP-M.01, "Loss of Essential Raw Cooling Water". Which ONE of the following identifies the correct section of AOP-M.01 to be implemented for the conditions and a mitigating action directed to be taken in response to the conditions? A. Section 2.7, "Supply Header 1A/2A Failure in the Yard"; Start additional ERCW pumps to maintain pressure. B. Section 2.7, "Supply Header 1A12A Failure in the Yard"; Stop and Lockout out all A Train ERCW pumps. C. Section 2.9, "Supply Header A Failure Upstream of Strainer Inlet Valves"; Start additional ERCW pumps to maintain pressure. Section 2.9, "Supply Header A Failure Upstream of Strainer Inlet Valves"; Stop and Lockout out all A Train ERCW pumps. Page 43 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 44 A. Incorrect, All conditions match entry conditions for Section 2.7 except for sump pump running alarm and the Strainer Dp alarm being lit if the leak were downstream of the strainer. While starting additional pumps might restore pressure,the mitigating action is to stop and lock out the A Train pumps to terminate the leakage for a leak at the location identified by the stem conditions. Plausible because Section 2.7 would be the correct procedure section and the starting on additional pumps is a mitigating action during performance of Section 2. 7 for a leak downstream of the strainer. B. Incorrect, All conditions match entry conditions for Section 2.7 except for sump pump running alarm and the Strainer Dp alarm being lit if the leak were downstream of the strainer. The A Train pumps being directed to be stopped and locked out is correct for the conditions stated. Plausible because Section 2.7 would be the correct procedure section for a leak downstream of the strainer and the stopping and locking out of the pumps in correct for the leak identified in the stem. C. Incorrect, All alarms stated would be lit for a header break upstream of the Strainer. Header pressure sensors are located is located just upstream of the strainers. the sump pump running differentiates the leak upstream from a yard leak (i .e. downstream of the strainer). While starting additional pumps might restore pressure,the mitigating action is to stop and lock out the A Train pumps to terminate the leakage. Plausible because Section 2.9 is the correct procedure section and the starting on additional pumps is a mitigating action for a leak downstream of the strainer (Section 2.7). D. CORRECT, All alarms stated would be lit for a header break upstream of the Strainer. Header pressure sensors are located is located just upstream of the strainers. the sump pump running differentiates the leak upstream from a yard leak (i .e. downstream of the strainer). Mitigating action in the AOP section is to stop and lock out all A Train pumps. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 90 Tier 2 Group 1 KIA 076 A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Service water header pressure. Importance Rating: 2.7/3.1 Technical Reference: AOP-M.01, Loss of Essential Raw Cooling Water, Rev 19 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 AOP-M.01, BA & 5 Question Source: Upon entry into AOP-M.1 , diagnosis the applicable condition and transition to the appropriate procedural section for response. Summarize the mitigating strategy for the failure that initiated entry into AOP-M.01 Bank # X __ _ Modified Bank # ----New ---Question History: SQN question 076 A2.02 053 with some modification. Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X, __ 10 CFR Part 55 Content: ( 41.5/43.5/45/3/45/13) 10CFR55A3.b ( 5 ) Comments: SQN question 076 A2.02 053 with modification to correct answer, all distractors, and stem. Used most plausible 2 of original distractors and added requirement to identify mitigating actions. Changed correct answer to D. No significant modification to data in the stem. Page 45 SQN LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.01 Rev. 19 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS CAUTION: ERCW header rupture in Auxiliary Building could fill the passive sump in 15 minutes. Prompt action is needed. 1. DIAGNOSE the failure: IF ... GO TO SECTION PAGE E.RCW Pump(s) tripped or failed 2.1 ERCW pump failure 5 High flow ERCW Supply Header 1A 2.2 Supply Hdr 1A Failure 8 to Aux Bldg High flow ERCW Supply Header 1 B 2.3 Supply Hdr 1B Failure 12 to Aux Bldg High flow ERCW Supply Header 2A 2.4 Supply Hdr 2A Failure 16 to Aux Bldg High flow ERCW Supply Header 2B 2.5 Supply Hdr 2B Failure 23 to Aux Bldg Indications of an ERCW Return Header rupture 2.6 Return Hdr rupture 29 (must be diagnosed locally since M-27 indications in Aux Bldg are not affected) Low flow ERCW Supply Header 1A and 2A, 2.7 Supply Header 1A12A 41 AND Failure in Yard Area STRAINER OIFF PRESS HIGH alarm LIT [M-27 A, C-3 and/or 0-2] Low flow ERCW Supply Header 1 Band 2B, 2.8 Supply Header 18/2B 55 AND Failure in Yard Area STRAINER OIFF PRESS HIGH alarm LIT [M-27 A, C-6 and/or 0-5] (step continued on next page) Page 3 of 149

- - Jo -. SQN LOSS OF ESSENTIAL

RAW COOLING WATER -AOp .. M.01 Rev. 19 ( STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS (Continued) 3. (Continued) IF ... GO TO SECTION PAGE Low flow ERCW supply headers 1A and 2A, AND STRAINER DIFF PRESS alarms DARK [M-27 A, C-3 and 0-2], AND at least one of the following alarms LIT: * ERCW DECK SUMP LEVEL HI alarm LIT 2.9 Supply Header A 67 [1-M-15B, A-3] Failure Upstream of Strainer Inlet Valves OR * ERCW DECK SUMP PMP RUNNING [1-M-15B, 0-2 or 0-4] OR * MECH EQUIP SUMP LVL HI alarm LIT [1-M-15A, B-6] Low flow ERCW supply headers 1 Band 2B, AND STRAINER DIFF PRESS alarms DARK [M-27 A, C-6 and 0-'5], AND at least one ofthe following alarms LIT: -. -.. * ERCW DECK SUMP LEVEL HI alarm LIT [1-2.10 Supply Header B 76 M-15B, A-3] Failure Upstream of Strainer Inlet Valves OR * ERCW DECK SUMP PMP RUNNING [1-M-15B, 0-2 or 0-4] OR * MECH EQUIP SUMP LVL HI alarm LIT [1-M-15A, B-6] Loss of flow on ALL ERCW supply headers 2.11 Loss of all ERCW flow 83 in modes 1-4. END OF SECTION Page 4 of 149 ( ( [ LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.01 Rev. 19 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED '------'--- 2.9 ERCW Supply Header A Failure Upstream of ERCW Strainer Inlet Valves CAUTION: During operation, CCP and SI Pumps may experience bearing failure 10 minutes after loss of ERCW. 1. STOP and LOCK OUT all Train A ERCW Pumps. 2. DISPATCH operators with radios to perform the following:

PERFORM Appendix F, Rx MOV Board ERCW Valves [Aux Bldg el. 749', Rx MOV Boards] * PERFORM Appendix G, ERCW MCC Valves [ERCW Pumping Station] o ENSURE all pumping station watertight

doors are CLOSED [ERCW Pumping Station] Page 67 of 149 SQN LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.01 Rev. 19 L-________ -L _____________________________ ________________ -L ________ STEP ACTION/EXPECTED RESP_O_N_S_E_-.l.I_. ___ R---,E_S_P_O_N_S_E_N_O_T_O_B_T_A_IN-'-E_D __ ----Il ........ 2.7 ERCW Supply Header 1A12A Failure inYard Area CAUTIONS:

During operation, CCP and SI Pumps may experience

bearing failure 10 minutes after loss of ERCW cooling. NOTE: * Loss of 2A ERCW Supply Header affects both Units' Train A CCS Heat Exchangers. Isolation of ruptured Unit and restoration of ERCW to intact CCS Heat Exchangers is time critical to prevent tripping both Units. * Engineering may be able to identify ruptured yard header using yard piping drawings (17W300 series). 1. DISPATCH personnel to locate failure. 2. DISPATCH operators with radios to PERFORM the following appendixes:

Appendix F, Rx MOV Board ERCW Valves [Aux Bldg, 749' elev, Rx MOV Boards] * Appendix G, ERCW MCC Valves [ERCW Pumping Station] 3. START additional

Train A ERCW Pumps as required to maintain pressure between 78 psig and 124 psig. Page 41 of 149 ( 10 (B-3) Source Setpoint SER 1078 (Unit 1 annunciator system) 2-PS-67 -493A 50 psig decreasing UNIT 2 HEADER A PRESSURE LOW Retransmitted to U-2 SER 2096 (Unit 2 annunciator system) Probable Causes Corrective Actions References 1. Unit 2 ERCW "An train pumps tripped. 2. Unit 2 "A" Train ERCW line break. 3. System realignment increasing demand excessively. [1] IF alarm is in conjunction with any of the following indications:

ERCW pump trip/failure
HIGH header flow * LOW header flow THEN GO TO AOP-M.01, Loss of Essential

Raw Cooling Water. [2] IF system realignment, in progress, is most probable cause for alarm, THEN CORRECT alignment in accordance with instruction in progress, OR START additional Train A ERCW Pump as required, in accordance with instruction in progress. [3] IF cause not apparent, THEN DISPATCH personnel to check pumping station and piping for ruptures or cause of alarm. [4] EVALUATE EPIP-1 Emergency Plan Classification Matrix. [5] IF Unit 2 "A" train is declared inoperable, THEN CONSULT Technical Specifications 3.7.4. 45B655-27A-0, 45N655-32, 47B601-55-13 SQN O-AR-M27-A Page 12 of 39 o Rev. 19 1 (A-1) Source Setpoint SER 1066 (Unit 1 annunciator system) 1-PS-67 -493A 50 psig decreasing UNIT 1 HEADER A PRESSURE LOW Retransmitted to U-2 SER 2084 (Unit 2 annunciator system) Probable Causes Corrective Actions References 1. Unit 1 ERCW "A" train pumps tripped or manually stopped. 2. 1A ERCW line break. 3. System realignment increasing demand excessively. [1] IF alarm is in conjunction with any of the following indications:

ERCW pump trip/failure
HIGH header flow * LOW header flow THEN GO TO AOP-M.01, Loss of Essential

Raw Cooling Water. ' [2] IF system realignment, in progress, is most probable cause for alarm, THEN CORRECT alignment in accordance with instruction in progress, OR START additional Train A ERCW Pump as required, in accordance with instruction in progress. [3] IF cause not apparent, THEN DISPATCH personnel to check pumping station and piping for ruptures or cause of alarm. [4] EVALUATE EPIP-1 Emergency Plan ClaSSification Matrix. [5] IF Unit 1 "A" train is declared inoperable, THEN CONSULT Technical Specifications 3.7.4. 458655-27 A-O, 45N655-32, 478601-55-13 SQN O-AR-M27-A Page 3 of 39 o Rev.19 ( ( 16 Source Setpoint SER 1075 (Unit 1 annunciator system) 0-PS-67-461 and 50 psig decreasing with pump breaker closed PUMP Q-A DISH PRESS 52a on breaker LOW Retransmitted to U-2 SER 2093 (Unit 2 annunciator system) Probable Causes Corrective Actions References 1. ERCW pump Q-A damaged. 2. Train A ERCW line break. 3; Instrument malfunction. 4. Insufficient pumps running for system flow demand. [1] ENSURE sufficient pumps running for system configuration. [2] VERIFY ERCW pump Q-A running. [3] IF pump is running, THEN DISPATCH operator to determine problem. [4] IF Q-A ERCW pump is failed, THEN GO TO AOP-M.01, Loss of Essential Raw Cooling Water. [5] IF pressure low due to ERCW line Break, THEN GO TO AOP-M.01, Loss of Essential Raw Cooling Water. [6] IF "A" train is declared inoperable, THEN CONSULT Technical Specification 3.7.4. 45B655-27A-0, 45N655-32, 47B601-55-13 (C-2) SQN O-AR-M27-A Page 18 of 39 o Rev. 19 8 Source Setpoint SER 1067 (Unit 1 annunciator system) 0-PS-67 -433 and 50 psig decreasing with pump breaker PUMP J-A DISH PRESS LOW 52a on breaker closed Retransmitted to U-2 SER 2085 (Unit 2 annunciator system) Probable Causes Corrective Actions References 1. ERCW pump J-A damaged. 2. Train A ERCW line break. 3. Instrument malfunction. 4. Insufficient pumps running for system flow demand. [1] ENSURE sufficient pumps running for system configuration. [2] VERIFY ERCW pump J-A running. [3] IF pump is running, THEN DISPATCH operator to determine problem. [4] IF J-A ERCW pump is failed, THEN GO TO AOP-M.01, Loss of Essential Raw Cooling Water. [5] IF pressure low due to ERCW line Break, THEN GO TO AOP-M.01, Loss of Essential Raw Cooling Water. [6] IF "A" train is declared inoperable, THEN CONSULT Technical Specification 3.7.4. 45B655-27 A-O, 45N655-32, 4 7B60 1.,55-13 (B-1) SQN O-AR-M27-A Page 10 of 39 o Rev. 19 076 A2.02 053 QUESTIONS REPORT for BANK SQN Questions ( Given the following plant conditions: Both Units operating at 100% power. ERCW system in normal alignment. The following annunciators are LIT on Essential Raw Cooling Water 0-XA-55-27A panel on 1-M-27: -"UNIT 1 HEADER A PRESSURE LOW". -"UNIT 2 HEADER A PRESSURE LOW". -"PUMP J-A DISCH PRESS LOW". -"PUMP Q-A DISCH PRESS LOW". The following annunciator is LIT on Miscellaneous 1-XA-55-15B panel on 1-M-15: -"ERCW DECK SUMP PUMP B RUNNING". ERCW headers 1A and 2A indicating LOW flow. Which ONE of the following describes what has occurred in the ERCW system? A. 'A' header pumps have tripped. B. Train A ERCW 480v. board has been deenergized. Header has ruptured upstream of the '2A' strainer. D. '1A' header has ruptured between the IPS and Auxiliary Bldg. A. Incorrect -Low disch pressure alrams on the pumps indicate the pump breakers are closed B. Incorrect -Strainer alarm would be lit for a clogged strainer. No sump pump running alarm with high pressure. C. Correct -All alarms stated would be lit for this accident. Pressure sensors are located is located just upstream of the strainers. D. Incorrect -Strainer Dp alarm would be lit, all other conditions match except sump pump running alarm. Thursday, July 31,20088:28:19 AM 1 OPL271 AOP-M.01 Revision 0 Page 3 of 44 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-M.01 LOSS OF ESSENTIAL RAW COOLING WATER IV. LENGTH OF LESSON/COURSE: 2.0 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-M.01, LOSS OF ESSENTIAL RAW COOLING WATER B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with a Loss of Essential Raw Cooling Water that are rated:;:: 2.5 during Initial License Training and:;:: 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Explain the purpose/goal of AOP-M.01. 2. Discuss the AOP-M.01 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-M.01 entry conditions. b. Describe the ARP requirements associated with AOP-M.01 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-M.01 entry conditions. d. Describe the Administrative conditions that require Turbine Trip/ Reactor trip due to Loss of Essential Raw Cooling Water. 3. Describe the initial operator response to stabilize the plant upon entry into AOP-M.01. 4. Upon entry into AOP-M.01 , diagnose the applicable condition and transition to the appropriate procedural section for response. 5. Summarize the mitigating strategy for the failure that initiated entry into AOP-M.01. 6. Describe the bases for all limits, notes, cautions, and steps of AOP-M.01. ( 7. 8. 9. 10. OPL271 AOP-M.01 Revision 0 Page 4 of 44 Describe the conditions and reason for transitions within this procedure and transitions to other procedures. Given a set of initial plant conditions use AOP-M.01 to correctly: a. Recognize entry conditions b. Identify required actions c. Respond to Contingencies d. Observe and Interpret Cautions and Notes Describe the Tech Spec and TRM actions applicable during the performance of AOP-M.01. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition Guidance for SRO-only Questions RevO Figure 2: Screening for SRO-only linked to 10CFRSS.43(b)(S) (Procedures) Can question be answered by knowing "systems l knowledge", i.e., how the system works, flowpath, Yes RO question logic, location, etc. No Can question be answered by knowing immediate Yes I \ RO question operator actions? No Can question be answered by knowing entry Yes I -j RO question I conditions for major EOPs? iNo r Does the question involve one or more of the following?

Assessing

plant conditions (normal, abnormal,or emergency) and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed * Recallingwhat strategy or action is written into a plant Yes I SRO-only procedure, including when the strategy or action is required question I No Question is not linked to 10CFR55.43(b)(5) for SRO-only Page 8 of 19 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 91. 055 G2.4.3 091 Which ONE of the following correctly completes the statement below? The Unit 1 Condenser Vacuum Mid Range Radiation Monitor, 1-RM-90-255, is a monitor ... A. included in Tech Spec LCO 3.3.3.7, "Accident Monitoring Instrumentation", and is used in the Radiological Emergency Plan (REP) to classify an event based on gaseous effluent release. B. included in Tech Spec LCO 3.3.3.7, "Accident Monitoring Instrumentation", and is used in the Radiological Emergency Plan (REP) to classify an event based on the fission product barrier matrix. C'!" NOT included in Tech Spec LCO 3.3.3.7, "Accident Monitoring Instrumentation", but is used in the Radiological Emergency Plan (REP) to classify an event based on gaseous effluent release. D. NOT included in Tech Spec LCO 3.3.3.7, "Accident Monitoring Instrumentation", but is used in the Radiological Emergency Plan (REP) to classify an event based on the fission product barrier matrix. Page 46 ( Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA C TOR ANAL YSIS: Page 47 A. Incorrect, 1-RM-90-255 is not included in the TIS LCO for Accident Monitoring but the monitor is included in EPIP-1, Emergency Plan Classification Matrix", Table 7-1 under Gaseous Monitors whose output can be used for determining Emergency Plan classifications. Plausible that it would be included in the TS LCO because the monitor is a Post Accident Monitor and the monitor being included in EPIP-1, Emergency Plan Classification Matrix", Table 7-1 under Gaseous Monitors whose output can be used for determining Emergency Plan classifications is correct. B. Incorrect, 1-RM-90-255 is not included in the TIS LCO for Accident Monitoring but the monitor is not used to classify an event based on the fission product barrier matrix. Plausible that it would be included in the TS LCO because the monitor is a Post Accident Monitor and there are radiation monitors used to make classifications in the fission product barrier matrix. C. CORRECT, 1-RM-90-255 is a post accident monitor but is not included in the TIS LCO for Accident Monitoring. The monitor is included in EPIP-1, Emergency Plan Classification Matrix", Table 7-1 under Gaseous Monitors whose output can be used for determining Emergency Plan classifications. D. Incorrect, 1-RM-90-255 is a post accident monitor but is not included in the TIS LCO for Accident Monitoring. while the monitor is included in EPIP-1, Emergency Plan Classification Matrix", Table 7-1 under Gaseous Effluent release, the monitor is not used to classify an event based on the fission product barrier matrix. Plausible because the monitor not being included in LCO for Accident Monitoring is correct and there are radiation monitors used to make classifications in the fission product barrier matrix. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 91 Tier 2 Group 2 KIA 055 G2A.3 Condenser Air Removal System Ability to identify post-accident instrumentation. Importance Rating: 3.7/3.9 Technical Reference: EPIP-1, Emergency Plan Classification Matrix, Rev 40 Technical Specification LCO 3.3.10 FSAR Section 7.5 Amendment 2 1-AR-M30-A, Post Accident Radiation Monitoring, Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: OPT200. RM BA.I Question Source: Describe the following characteristics of each major component in the Radiation Monitoring system: I. Types of accidents for which the components are designed. OPT200. CONDVAC BA.I Describe the following characteristics of each major component in the CDVAC system: I. Types of accidents for which the CDVAC components are designed. Bank# ____ _ Modified Bank # -----New _X __ Question History: New question for SQN 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.6 / 4504 ) 10CFR55A3.b ( 2,4,5 ) Comments: New question for SQN 1/2009 exam Page 48 ( ( ( Source SER 1745 1-RE-90-255 Probable Causes Corrective Actions References Setpoint see 1-RM-90-255 display and/or data base 1. Steam generator tube rupture. 2. Instrument Malfunction. 17 (C-1) 1-RA-255A CONDVAC EXH MID RANGE HIRAD [1] CHECK [1-RM-90-255] on panel 1..:M-30 to verify activity level. [2] EVALUATE EPIP-1, Emergency Plan Classification Matrix; [3] IF high radiation indicated in condenser vacuum exhaust, THEN GO TO AOP-R.01, Steam Generator Tube Leak. 478601-55-75,47W610-90-4 SQN 1-AR-M30-A Page 20 of 27 1 Rev. 15

( [iiQUOY_A_H_L......... ___ E_M_E_R_G_E_N._C_Y_P_L.A __ N_C_L_A_S_S_1F_IC_A_T_IO_N_M_A_TR_I_X __ ..l...-.-,-E_P_IP_-1---1 A L L A L L A L L A L L EA8 dose, resulting from an actual or imminent release of gaseous radioactivity> 1 Rem TEDE or > 5 Rem thyroid CDE for the actual or projected duration of release. (1 or 2 or 3): 1. A. VALID rad monitor reading exceeds the values under General Emergency in Table 7-1 for >15 min, unless assessment within that 15 min confirms that the critenon is not exceeded. 2. Field surveys indicate >1 Rem/hr gamma or an 1-131 concentration of 3.9E-06 ;ICi/cm' at the EAB (Fig. 7-A) OR 3. Dose assessment results indicate EAB dose >1 Rem TEDE or >5 Rem thyroid CDE for the actual or projected duration of the release (Fig. 7-A). EAB dose resulting from an actual or imminent release of gaseous radioactivity >100 mrem TEDE or >500 mrem thyroid CDE for actual or projected duration of release. (1 or 2 or 3): 1. A VALID rari monitor reading> Table 7-1 values under Site I,rea for:> 15 min. unless assessment within that 15 min confirms that the criterion is not exceeded. PR 2 Field surveys indicate >100 mrem/hr gamma or an 1-131 cone of 3.9E-07 pCi/cm' at the EAB (Fig. 7-A). OR .'. Dose assessment results indicate EAB dose >100 mrem TEDE or >500 mrem thyroid CDE for actual or projected duration of the release (Fig. I-A). Any UNPLANNED release of gaseous radioactivity that exceeds 200 times the aDeM Section 1.2.2.1 Limit for >15 minutes. (1 or 2 or 3 or 4) '1. A VALID rad monitor reading> Table 7-1 values under Alert for >15 minutes, unless assessment within that 15 minutes confirms that the criterion is not exceeded. OR 2. Field surveys indicate >10 mrem/hr gamma at the EAB tor >15 minutes (Fig 7-A). 3. Dose assessment results indicate EAB dose >10 mrem TEDE for the duration of the release (Fig. 7 -A). OR 4. Sample results exceed 200 times the ODCM limit value for an unmonitored release of gaseous radioactivity >15 minutes in duration. Any UNPLANNED release of gaseous radioactivity that exceeds 2 times the aDCM Section 1.2.2.1 Limit for >60 minutes. (1 or 2 or 3 or 4) 1. A VALID rad monitor reading> Table 7-1 values under UE for >60 minutes, unless assessment within that 60 minutes confirms that the criterion is not exceeded. OR 2. Field surveys indicate >0.1 mrem/hr gamma at the EAB for >60 minutes (Fig 7-A) 3. Dose assessment results indicate EAB dose >0.1 mrem TEDE for the duration of the release (Fig. 7-A). OR 4. Sample results exceed 2 times the ODCM limit value for an unmonitored release of gaseous radioactivity >60 minutes in duration A L L A L L Page 43 of4? Not Applicable. Any UNPLANNED release of liquid radioactivity that exceeds 200 times the aDCM Section 1.2.1.1 Limit for >15 minutes. (1 or 2) 1. A VALID rad monitor reading > Table 7-1 values under Alert for >15 minutes, unless assessment within this time period confirms that the criterion is not exceeded. aR 2. Sample results indicate an ECl >200 times the aDCM limit value for an unmonitored release of liquid radioactivity >15 minutes in duration Any UNPLANNED release of liquid radioactivity to the environment that exceeds 2 times the aDCM Section 1.2.1.1 Limit for >60 minutes. (1 or 2) 1, A VALID rad monitor reading> Table 7-1 values under UE for >60 minutes, unless assessment within this time period confinms that the criterion is not exceeded. 2. Sample results indicate an ECl >2 times the aDCM limit value for an unmonitored release of liquid radioactivity >60 minutes in duration. Revision 40 ( ( SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 TABLE 7-1 EFFLUENT RADIATION MONITOR EALS NOTE: The monitor values below, if met or exceeded, indicate the need to perform the required assessment If the assessment can not be completed within 15 minutes (60 minutes for UE), the appropriate emergency classification shall be made based on the VALID reading. GASEOUS MONITORS Units (2) UE Alert SAE General Site Total Release Limit fl Cils 4.90E+05 4.90E+07 1.31E+OS 1.31E+09 1-RI-90-400 (EFF LEVEL) -U-1 Shield Bldg pCi/s 4.90E+05 4.90E+07 1.31E+OS 1.31E+09 2-RI-90-400 (EFF LEVEL) -U-2 Shield Bldg pCi/s 4.90E+05 4.90E+07 1.31E+OS 1.31 E+09 O-RM-90-1 01 B -Auxiliary Bldg cpm 1.03E+05 Offscale l1j Offscale(1) Offscale(1) O-RM-90-132B -Service Bldg cpm 2.62E+06 Offscale (1) Offscale (1) Offscale(1) 1-RI-90-421 thru 424 -U-1 MSL Monitors(2) pCi/cc 1.71 E-01 1.71E+01 4.5SE+01 4.5SE+02 2-RI-90-421 thru 424 -U-2 MSL Monitors(2) pCilcc 1.71 E-01 1.71E+01 4.5SE+01 4.5SE+02 1-RM-90-255 or 256 -U-1 eVE mR/h 4.10E+02 4.10E+04 1.09E+05 1.09E+06 2-RM-90-255 or 256 -U-2 eVE mR/h 4.10E+02 4.10E+04 1.09E+05 1.09E+06 RELEASE DURA nON minutes >60 >15 >15 >15 LIQUID MONITORS Units UE Alert Site Area General Site Total Release. Limit pCi/ml 6.50E-03 6.50E-01 N/A N/A RM-90-122 -RadWaste cpm 1.45E+06 Offscale(1) N/A N/A RM-90-120,121 -S/G Bldn cpm 1.07E+06 Offscale(1) N/A N/A RM-90-225 -Cond Demin cpm 1.90E+06 Offscale (1) N/A N/A RM-90-212 -TB Sump cpm 3.2SE+03 3.2SE+05 N/A N/A RELEASE DURA nON minutes >60 >15 >15 >15 ASSESSMENT METHODS: * Airbome Dose Assessment per SON EPIP-13 "Dose Assessment" * ODCM Liquid Release Rate assessment per SON 0-TI-CEM-030.030.0

Integrated

Airborne Release Rate assessment per SON 0-TI-CEM-030.030.0 (1) The calculated value is outside of the upper range for this detector. The maximum monitor output which can be read is 1.0E+07 cpm. Releases in excess of monitor capacity should be evaluated for proper classification by use of Dose Assessment. (2) These unit values are based on flow rates through one PORV of 890,000 Ib/hr at 1078.7 psia with 0.25% carry over (0.9975 quality). Before using these values, ensure a release to the environment is ongoing, (e.g., PORV). NOTE 1: These EALs are based on the assumption that an emergency release is restricted to one pathway from the plant. In all cases, the total site EAL is the limiting value. Therefore, in the case where there are multiple release paths from the plant, it is the total release EAL (obtained from ICS and/or SON 0-TI-CEM-030-030, "Manual Calculation of Plant Gas, Iodine, and Particulate Release Rates for Offsite Dose Calculation Manual (ODCM) Compliance") that will determine whether an emergency classification is warranted. NOTE 2: In the case when there is no CECC dose assessment available, the length and relative magnitude of the release is the key in determining the classification. For example, in the case of the NOUE EAL of 2 times the Tech Spec limit, the classification is based more on the fact that a release above the limit has continued unabated for more than 60 minutes, than on the projected offsite dose. NOTE 3: See REP Appendix B for basis information. Page 46 of 47 Revision 40 SQN OF.CONDENSAlE VACUUM 1-PI-ICC-090-255.0 1 EXHAUST POST ACCIDENT RADIATION MONITOR Rev. 7 1-R-90-255 Page 6 of 65 2.0 REFERENCES 2.1 Performance References None 2.2 Developmental References A. Administrative References None B. TVA and Vendor drawing 1. 47W610-90-1 2. 45N1620-12 3. 45WI651-16,-17,-20,-21 C. Manufacturer Manuals 1. SQN-VTM-WI30-130 & VTD-WI30-0150 Vendor Manual For Westronics Smartview Data Recorder 2. SQN-VTM-G063-0430, RM-I000 Digital Radiation Processor Technical Manual (Document 04508100-1TM) 3. SQN-VTM-G063-0010,Vendor Technical Manual For Radiation Monitoring System, Volumes I, II, III, and IV Contract No. 92759 D. Other Developmental References 1. FSAR Sections: 7.5, 11.4, 12.1.4, 12.2.4 2. Set Point and Scaling Document (SSD) l-R-90-255 SQN-18 TABLE 7.5-2 (Sheet 2) TABLE OF VARIABLES FOR POST ACCIDENT MONITORING Variable Typel Minimum Minimum Redundancy Description Cateoorv Ranoe From Ranoe To Reauired Notes Aux Bldg EXH Vent Rad Level -E3 1E-9 1E-4uCi/CC N/A See Deviation No. 14 Particulates & Halogens Remote Analysis Utilizing Removable Filter May be Used. Aux Bldg Passive Sump C3 SEE NOTES N/A Low & Hi Level Alarm in MCR (FLR & EQP DRN SMP) LVL AUX Bldg Pressure D2 -0.5 +0.5 Inches WG N/A AUX Cntl Air Sys Pressure D2 0 125 Psig N/A RG1.97 R2 -POWER SUPPLY Boron Injection Flow D2 0 110% (Design) N/A (Flow in HPI System) 0 864GPM Component Cooling Sys Surge D3 0 100% N/A Actual Range 0 to 124 Inches Tank Level 0 10,000Gal Component Cooling Water Flow D2 0 110% (Design) N/A to ESF Equip 0 5523GPM Component Cooling Water Temp D2 30 130DEG F N/A See Deviation No.7 to ESF Equip Condenser (Air Removal Sys) E2 0 110% (Design) N/A Vacuum EXH Flow 0 49.5CFM Condenser (Air Removal Sys) B3 C3 E2 1E-6 1E4 uCi/CC N/A Part of Sec Side RAD Lvi Vacuum EXH RAD Level -Noble Gas Condensate Storage Tank Water D2 0 367,000 GAL N/A Safety Source is-ERCW Level See ERCW to AFW Valve Position T75-2.doc ( TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 1. Reactor Coolant T HOT (Wide Range) (Instrument Loops 68-001,-024,-043,-065) 2. Reactor Coolant T COLD (Wide Range) (lnstrumeni LOops 68-018,-041,-060,-083) 3. Containment Pressure (Wide Range) (Instrument Loops 30-310,-311) 4. Containment Pressure (Narrow Range) (Instrument Loops 30-044,-045) 5.:.. Water Storage Tank Level ,(Instrument Loops

'\ 6. Reactor Coolant Pressure (Wide Range) (Instrument

Loops 68-062,-066,-069) 7. Pressurizer Level (Wide Range) (Instrument Loops 68-320,-335,-339) 8. Steam Line Pressure (Instrument Loops 1-002A,-002B,-009A,-009B,- 020A, -020B, -027 A, -027B) 9. Steam Generator Level-(Wide Range) (Instrument Loops 3-043,-056,-098,-111) 10. Steam Generator Level-(Narrow Range) (Instrument Loops 3-039,-042,-052,-055,-094,- 097,-107,-11 0) 11. Auxiliary Feedwater a. Flow Rate (Instrument Loops b. Valve Position Indication (Instrument Loops 3-164,-164A,-172,-156, -156A,-173,-148,-148A,-174,-171,-171A,-175) SEQUOYAH -UNIT 1 TOTAL NO. OF CHANNELS 4(1/RCS Loop) 4(1/RCS Loop) 2 2 2 3 3 21steamline 4(1/steam generator) 2/steam generator 1/steam generator 3/steam generator 3/43-56 MINIMUM CHANNELS ACTION REQUIRED 4(1/RCS Loop) 1 4(1/RCS Loop) 1 2 1 2 1 2 1 C 3 2 3 2 21steam line 1 4(1/steam 1 generator) 2/steam generato"r 1 1/steam generator 5 3/steam generator 5 July 9, 1992 Amendment No. 46,114, 149,159 TABLE 3.3-10 (ContirlUed) ACCIDENT MONITORING INSTRUMENTATION , INSTRUMENT 17. Neutron Flux a. Source Range (Instrument loops 92-5001,-5002) b. Intermediate Range (Instrument loops 92-5003,-5004) 18. ERCW to AFW Valve Position ') a. Motor Driven Pumps (Instrument loops ,3-116A, -116B, -126A, -126B) b. Turbine Driven Pumps (Instrument loops -136B, -179A, -179B) , i 19. Contain!'11ent Isolation ValvePositiOri ", (Panels & TR-B XX-55-6l) ,;: ';'. . ... ..... " TOTAL NO. OF CHANNELS 2 2 1ffrain/Pump (2 ValveS/Train) 2 Trains " (2 ValveS/Trairi) , ' 1 Naive MINIMUM CHANNELS REQUIRED 2 1ffrain/Pump (2 ValveslTrain) 2 Trains (2 ValveslTrain) 1 Nalve## #Source P-6'(Block of Sotirce Range Reacto; Trip) setpoint ##Not required for isolation valVes that are'dosed and deactivated. ' " ) ACTION 1 1 1 1 3 SEQUOYAH -UNIT 1 3I43-56b July 9,1992 Amendment No. 112, 149, 159 ( TABLE 3.3-10 (Continued) ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT 12. Reactor Coolant System Subcooling Margin Monitor (Instrument Loops 94-101,-102) 13; Containment Water Level (Wide Range) (Instrument Loops 63-178,-179) 14. Incore Thermocouples a. Core Quadrant (1) b. Core Quadrant (2) c. Core Quadrant (3) d. Core Quadrant (4) 15. Reactor Vessel Level Instrumentation a. Dynamic Range (Instrument Loops 68-367,370) b. Lower Range (Instrument Loops 68-368,371) c. Upper Range (Instrument Loops 68-369,372) 16. Containment Area Radiation Monitors a. Upper Compartment (Instrument Lo'ops 90-271,-272) b. Lower Compartment (Instrument Loops 90-273,-274) SEQUOYAH -UNIT 1 TOTAL NO OF CHANNELS 2 2 65 6 2 2 3/43-56a MINIMUM CHANNELS REQUIRED ACTION 2 1 2 1 2(1fTrain) 1 2(1fTrain) 1 2(1fTrain) 1 2(1fTrain) 1 2 1 2 1 2 1 1 4 1 4 October 4, 1995 AmendmentNo. 112, 149,159,213 / / I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: RADIATION MONITORING SYSTEM OPT200.RM Rev. 2 Page 3 of 166 IV. LENGTH OF LESSON: 4 hour lecture; 1 hour simulator demonstration; 2 hour study/workshop v. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Radiation Monitoring System in the plant and on the simulator. B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with the Radiation Monitoring System as identified in Appendix A. 1. State the purpose/functions of the Radiation Monitoring System as described in the SQN FSAR. 2. State the design basis of the Radiation Monitoring System in accordance with the SQN FSAR. 3. Explain the purpose/function of each major component in the flow path of the Radiation Monitoring System as illustrated on a simplified system drawing. 4. Describe the following characteristics of each major component in the Radiation Monitoring System: a. Location b. Power supply (include control power as applicable) c. Support equipment and systems d. Normal operating parameters e. Component operation f. Controls g. Interlocks (including setpoints) h. Instrumentation and Indications 1. Protective features (including setpoints) J. Failure modes k. Unit differences 1. Types of accidents for which the components are designed m. Location of controls and indications in the control room and auxiliary control room V. TRAINING OBJECTIVES (Cont'd): B. Enabling Objectives (Cont'd): 5. Describe the operation of the Radiation Monitoring System: a. Precautions and limitations b. Major steps performed while placing the system in service c. Alarms and alarm response d. How a component failure will affect system operation e. How a support system failure will affect system operation f. How a instrument failure will affect system operation OPT200.RM Rev. 2 Page 4 of 166 6. Describe the administrative controls and limits for the Radiation Monitoring System: a. State Tech Specs/TRM LCOs that govern the system. b. State the :::::1 hour action limit TS LCOs c. Given the conditions/status of the Radiation Monitoring System components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required 7. Discuss related Industry Events VI. TRAINING AIDS: A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available) I. PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: CONDENSER VACUUM IV. LENGTH OF LESSON: 2 hour lecture; 1 hour simulator demonstration; 1 hour self-study/workshop V. TRAINING OBJECTIVES: A. Terminal Objective: OPT200.CONDVAC Rev. 1 Page 3 of 31 Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Condenser Vacuum (CDVAC) system in the plant and on the simulator. B. Learning Objectives: O. Demonstrate an understanding ofNUREG 1122 knowledge and abilities associated with the CDVAC system that are rated:::: 2.5 during Initial License Training for the appropriate license position as identified in Appendix A. 1. State the purpose/functions of the CDVAC system as described in the FSAR. 2. State the design basis of the CDVAC system in accordance with the SQN FSAR. 3. Explain the purpose/function of each major component in the flow path of the CDV AC system as illustrated on a simplified system drawing. 4. Describe the following characteristics of each major component in the CDV AC system: a. Location b. Power supply (include control power as applicable) c. Support equipment and systems d. Normal operating parameters e. Component operation f. Controls g. Interlocks (including setpoints) h. Instrumentation and Indications i. Protective features (including setpoints) j. Failure modes k. Unit differences l. Types of accidents for which the CDV AC system components are designed m. Location of controls and indications associated with the CDV AC system in the control room and auxiliary control room V. TRAINING OBJECTIVES (Cont'd): B. Learning Objectives (Cont'd): 5. Describe the operation of the CDVAC system: a. Precautions and limitations OPT200.CONDVAC Rev. 1 Page 4 of 31 b. Major steps performed while placing the CDV AC system in service c. Alarms and alarm response d. How a component failure will affect system operation e. How a support system failure will affect CDV AC system operation f. How a instrument failure will affect system operation 6. Describe the administrative controls and limits for the CDVAC system: a. State Tech Specs/TRM LCOs that govern the CDV AC b. State the:::;l hour action limit TS LCOs c. Given the conditions/status of the CDVAC system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required 7. Discuss related Industry Events VI. TRAINING AIDS: A. Classroom Computer and Local Area Network (LAN) Access B. Computer proj ector C. Simulator (if available) Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 92. 071 A2.05 092 Given the following: -A planned release of Waste Gas Tank B is in progress when power is lost to 0-RM-90-118, Waste Gas Effluent Rad Monitor. Which ONE of the following identifies (1) the effect the loss of power will have on the release and (2) the requirement to allow any additional release of the tank with the radiation monitor out of service? A'! (1) The release will automatically terminate; (2) A new 0-SI-CEM-077-41 0.4, "Waste Gas Decay Tank Release", package would be required for any additional release of the tank after ODCM actions were met. B. (1) The release will automatically terminate; (2) The existing 0-SI-CEM-077-41 0.4, "Waste Gas Decay Tank Release", package could used for any additional release of the tank after ODCM actions were met. C. (1) An alarm will be generated and MANUAL termination of the release will be required; (2) A new 0-SI-CEM-077-41 0.4, "Waste Gas Decay Tank Release", package would be required for any additional release of the tank after ODCM actions were met. D. (1) An alarm will be generated and MANUAL termination of the release will be required; Page 49 (2) The existing 0-SI-CEM-077-41 0.4, "Waste Gas Decay Tank Release", package could used for any additional release of the tank after ODCM actions were met. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 50 A. CORRECT, the release would be automatically terminated due to an instrument malfunction and a new package would be required to make any additional releases from the tank. B. Incorrect, the release would be automatically terminated due to an instrument malfunction but the current SI package would not be used to make any additional releases from the tank, a new package would be required. Plausible because the release would be automatically terminated and releases can be stopped and restarted using the same package under other conditions. C. Incorrect, the release would not require manual actions to terminate, it would be automatically terminated due to an instrument malfunction. A new SI package would be required to make any additional releases from the tank. Plausible because some release point radiation monitor instrument malfunctions only alarm and ta new SI isrequiredd for any addition release from the tank. D. Incorrect, the release would not require manual actions to terminate, it would be automatically terminated due to an instrument malfunction. The current SI package would not be used to make any additional releases from the tank, a new package would be required. Plausible because some release point radiation monitor instrument malfunctions only alarm and releases can be stopped and restarted using the same package under other conditions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 92 Tier 2 Group 2 KIA 071 A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Power failure to the ARM and PRM Systems Importance Rating: 2.5* / 2.6 Technical Reference: 1 ,2-47W611-77-4 R10 0-AR-M12-B, Common Radiation Monitor 0-XA-55-12B, Rev 28 0-SO-77-15, Waste Gas Decay Tank Release, Rev. 15 0-SI-CEM-077-410.4,Waste Gas Decay tank Release, Rev.0014 Proposed references to be provided to applicants during examination: None Learning Objective: No training objective identified Question Source: Bank# ___ _ Modified Bank # X __ _ New ---Question History: WBN bank question Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: (41.5/43.5/45.3/45.13 ) 10CFR55.43.b ( 5 ) Comments: Page 51 ( Watts Bar KIA: 000059 AA2.03 Accidental Liquid RadWaste ReI. Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release: Failure modes, their symptoms, and the causes of misleading indications on a radioactive-liquid monitor. Question: Given the following:

The unit is at 100% power and all equipment

is available.

A planned Cask Decontamination

Collector Tank Release is in progress when the following occurs: o Annunciator 181-C 'WOS RELEASE LINE 0-RM-90-122 INSTR MALF" alarms. o Annunciator 181-A 'WOS RELEASE LINE 0-RM-90-122 L1Q RAO HI" remains dark. Which ONE of the following failures would cause the above to prevent an accidental release, and the action required to allow the restart of the release with the radiation monitor out of service? A. Loss of power to the rate meter. The existing 001 package could be used to resume the release. B. Loss of power to the rate meter. The existing 001 package closed and new 001 package issued prior to resuming the release. C. Loss of signal from the detector. The existing 001 package could be used to resume the release. O. Loss of signal from the detector. The existing 001 package closed and new 001 package issued prior to resuming the release DISTRACTOR ANALYSIS a. Incorrect. The 'loss of power to the rate meter' would generate both the 181 C and 181 A alarms so the 181A alarm would also be LIT. The existing 001 package could not be used to continue the release. Plausible because the candidate may not recall the inputs to the alarms correctly and there are conditions identified in the release instructions that do allow the release to be continued using the current 001 package. b. Incorrect. The 'loss of power to the rate meter' would generate both the 181 C and 181 A alarms so the 181A alarm would also be LIT. A new 001 package would be needed to continue the release. Plausible because the candidate may not recall the inputs to the alarms correctly, but realize a new 001 package is required to continue the release. c. Incorrect. The 'loss of signal from the detector' would generate a 181-C alarm but would NOT generate a 181 A alarm as stated in the distractor, however the existing 001 package could not be used to continue the release. Plausible because the alarm status listed is correct for the listed condition would cause the termination of the release and the there are conditions identified in the release instructions that do allow the release to be continued using the current 001 package. d. CORRECT. The 'loss of signal from the detector' would generate a 181-C alarm but would NOT generate a 181 A alarm and a new 001 package would be required prior to continuing the release. 7 I I I J B I I LJ 9 I LJ DESCRIPTION SELECTOR SW GAS A GAS COMMON TO ANAL INC VLV ALL TANKS GAS B GAS r-------, ANAL INC VLV I I GAS C GAS I I ANAL INC \lLV 10 TVA NO WEST NO 10368 HB 10.378 PCV-77-11.3S 10388 CONTROL RELAY GOX!" GAXI GDX2A GAXI GDX3A GAXI 11 _I L.I L i i 12 NOTES: 1. FOR GENERAL NOTES AND REFERENCES SEE SHEET 47"611-77-1. 2. ONE WASTE CAS CCMPRESSOR: WIl.l. BE RUN CONTINUOUSl. Y WITH THE OTHER SERVING AS A BACKUP TO BE STARTED WHEN HEADER PRESSURE EXCEEDS 2 PSIG. THE COMPRESSORS liIl.l. BE Al.TERNATED FOR UNIFORM WEI\R. I-I- GOX-H --==:" 9 C 0 E I A I XS F I "\.J-I J 77-185 I I G I H GAS o GAS ANAL INC ,VLV PCV-77-100B 103gB GAS E CAS PCV-77-IOIB 10S28 ANAL INC \lLV GAS F' G"S GDX4A GAXI COX5" GAXt 3. ONE DECAY TANK PRESSURE ISOLATION VALVE WILL NORMALLY 8E OPEN WITH ANOTHER SELECTED FOR ST .... ND8Y. ALL OTHERS Ii I LL BE CLOSED. -l-. THIS IS II HAND OPERATED NEEDLE VALVE. I-A I-1-8 I- e X-9A -=--99 LS LVL > SP 77-958/.1., (HI-HI) loS LVL>SP 77-95A/B (HI) Lev Y-a 77-958 VENT 7 I I L __ ___ ...J "NAL ING ,VLV GAS G GAS ANAL tNG VLV r AUTDr---"\ GAS H GAS l ANAL SAMPLING YLV GAS TANK J GAS O?EN HS SAMPlE ANAL Y SAMPLING Vt.. V 77-115 "-OFF PCV-77-1028 PCV-77-1+58 PCV-77-1+6B ?CV-17-1+79 10538 GOX6" 105+8 COX7A loSSE COX8" 10568 GOX9A t-HS-77-1 H -HS-77-113 r-HS-77-100 GAXI GAXI ADMIN CHANCE GAXI GAXI (TYP AU FLOW> SP 1-FS-JO-1Soj" "'" I L:I-_{}iHIELD BLDG EXH',c ENT UNIT 2 SELECTED pAU'-'T"'OC":'J\::::= _____ -I'i-HS-77-IOI I--i-HS-17-102 r-HS-77-1+5 !--HS-77-1+6 r-H S-77-1+7 SHIELD BLDG EXH. VENT UNtT 2 SELECTED .FVENT PCV 77-115A X GAS CECAY TANK A PRESS. ISOLATION VLV (PCV-l036A) LO PRESSURE N2 158 '\1'11'61 r-17-1. COORD 1'-4 >-t __ __ I GAS DECAY TANK A X GAS ANAl. YZING SAMPLING VALVE (PCV-l0368) 8 r-U PRESS. < SP -I-I-D I-\ PS VENT 77-'" OFF I-77-119 ON 77-119 FSV -i HS \.. S FIC 77-24 f-,.....i>lX::J----- ..... -.l------C>.I:::l--- I-I-I-E I-----t-----------CAS DECAY TANK 8 _ I-f-----,'-----------GA5 DECAY TANK C _ I + ---,'-----------GAS DECAY TANK 0 _ I I-f-----,'-----------GAS DECAY TANK E I I-f-----,'-----------OA5 DECAY TANK F _ I + ---,'-----------OAS DECAY TANK 0 _ I I-f-----,'-----------OAS DECAY TA"K H -J RAOlAnDN. ISOLATlDN VALVE (RCV-Ol+) LJ.:::""" GAS ANAL m. ) I-e-I-F l-I- INC: ADMIN CHANGE PER RIMS: 837 001 REV CHANGE REf PREPARER CHECKER APPROVED DATE SCALE: NONE EXCEPT AS NOTED POWERHOUSE CATECORY 1 UNITS 1 '" 2 MECHANICAL LOGIC DIAGRAM WASTE DISPOSAL SYSTEM SEQUOYAH NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY DESIGN INITIAL tSSUE ---,'-----------GAS D'CAY TA"K J ---------'1-;:: -.'!:::::::' --I--CRAFTER, CHECKER, J.CDlE L.WALTERS

RO ISSUE f"ER: ENGINEERING

APf>ROVAL DRAIING MAOE 1 N/A II ceo fROM AC-N/}' AO-R5 DESIGNER, __ __ DATE N/A 7-3-91 I I I 8 I I I I I I I I I 9 I I I I I I I I 10 I I I I CAD MAINTAINED DRAWING I REVIEWER; "/A 3. M.R.SEDlAC!K/KRS CCO NO: 1. 2-47W611-77-4Rl0 / ----A-----------8----C --D --E --F --G --H I \ I I 7-LL-L L9A1.L7-Z* L 1 " 1 ;.1 1 I '-' CONTROLS ,I SIMIL.AR TO ., LCV X 77-10' TO DRAIN 2 I 3 4 I I ,-----------------------------11--- I1NAS STOP __ __, A \Z!:J 1A to-".21T P-AUTO 5 I DESCRIPTION '( TANK A ISOLATION VLV Y TANK 8 ISOLATION VLV C ION VLV 0 ION VLV E ION VLV F ION YLV G ION VLV H ION VLY J ION VLV TVA NO PCV-77-11SA F'CV-77-II4A PCV-77-11JA PCV-77-IOOA PCV-77-I01A PCV-77-I02A PCV-77-I.SA PCV-77-t4GA PCV-77-147A 6 I 1 1 1 1 1 1 1 1 1 1 CONTROL.S .1 SIMILAR TO I 1'"'1 LCVI'OS eves VCT -UNIT t eves VCT -UKIT 2 eves EVA? -UNIT 1 eves EVAP UNIT 2 was G'A -WDS SRST -S5 VCT UNIT 2_ SS VCT UNIT ,-' eves HT -,OS ReDT -UNIT 2 wos !!.COT UNIT 1 -1 1 1 1 1 1 PRESS. < SF' PS ",j' TO DRAIN r--PS J 0 r--+--->I'--{ r--. N I ( 77-3"/0 77-88 -y ---FSV 77-90 77-89 fCV 77 -90 COMPRESSOR A _MOISTURE -I 5EPARATlOR @o -L-r---. X I SEAL WATER COOLER v J II Q r--- Ib:: * "":'.'."-rt::: Lev 77-95A OPEN: HS CLOSE a 77-405 x LCV 77-405 77-40.5 S TO DRAIN ,......, PRIMARY WATER SYSTEM TO eves HOLDUP TANK < 47161 I-sa-s. caORD 0-4 PCV 77-92 VEN;-ifLy t©U sv '-' ___ ----------_-------=------ COMPRESSO!!. ____________________ .-r---1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 I 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 3 4 5 6 18 (C-4) Source Setpoint SER 768 (Unit 1 annunciator system) O-RM-90-118 N/A O-RA-90-1188 WDS GAS EFF MON INSTR MALFUNC Probable Causes Corrective Actions References 1. Instrument power failure. 2. Instrument placed in TRIP ADJ position (Except for monitors equipped with RM-1000 modules only). 3. Instrument downscale failure or loss of signal. 4. Operate/Calibrate switch set to calibrate (RM-1000 modules only). . [1] IF gas release in progress, THEN REQUEST Radwaste AUO to verify 0-RCV-90-119 closed. [2] CHECK 0-RM-90-118 on 0-M-12 for possible trouble. [3] IF O-RM-90-118 is inoperable, THEN [a] NOTIFY the Chemistry Shift Supervisor to comply with ODCM requirements. [b] PLACE equipment off normal or inop tags, identifying the condition. [c] WHEN ODCM action is satisfied, THEN RESUME the release using appropriate procedures. [4] COMPLY with ODCM, Section 1.1.2 requirements. [5] INITIATE WO for maintenance, if required. 458655-128-0, 45N657-18, 45N667-1, SON O-AR-M12-B Page 22 of 40 0 Rev. 28 WASTE GAS DECAY TANK RELEASE 0-SO-77-15 Rev: 15 Page 11 of 16 6.0 NORMAL OPERATION (Continued) [23] VERIFY [O-PCV-77-117] gas release header pressure control valve is maintaining 5.3 psig as indicated on the HEPA filter inlet pressure gauge on paneI1-L-335 located under the stairway near the WGDT Valve Gallery. [24] RECORD below the time this release was started. Release started hrs. [25] RECORD below, the rate of gas release. Release rate In. H 2 0 Date CAUTION If during the remainder of this instruction a malfunction of 0-RE-90-118 or 0-FCV-77-119 occurs, this release must be stopped. [26] NOTIFY the Unit 1 Operator that a gas release is in progress.

NOTE The activity level recorded for RM-90-400

should be for the applicable Shield Building. [27] OBTAIN the following information from the U-1 UO, AND RECORD the information below: [a] IF RM-90-400 operable, THEN RECORD Activity Level on RM-90-400 _____ CPM. [b] IF 0-RE-90-118 operable, THEN RECORD Activity Level on 0-RE-90-118 CPM. SQN Waste Gas Decay Tank Release 0-SI-CEM-077 -410.4 Unit 0 Rev. 0014 Page 15 of 42 6.2 Pre-Release Instructions -Chemistry (continued) [16] SIGNOFF for item A and either item B or item C and CIRCLE B or C to indicate which one was satisfied. A. Approval of pre-release data generated by this Instruction. / Performer Date B. Verification that monthly projected offsite dose limits (ODCM SR 2.2.2.4) have NOT been exceeded, based on most recent performance of SI-422.1, OR C. Verification, with Operations support, that selected WGDT has been held a minimum of 60 days and all applicable requirements have been met. / Performer Date [17] TRANSMIT release package to Operations with authorization / / D Time Time to release. D 6.3 Release Instructions -Operations [1] REVIEW Steps 6.1 [3] and 6.1 [4] and Steps 6.2[11] monitor data. [2] IF radiation monitor O-RM-90-118 setpoint change is required, (when setpoint in Step 6.2[11] is greater than the setpoint in Steps 6.1 [3]), THEN REQUEST a setpoint change. SQN Waste Gas Decay Tank Release 0-SI-CEM-077 -410.4 Unit 0 Rev. 0014 Page 16 of 42 6.3 Release Instructions -Operations (continued) [3] IF a release is to be made outside of normal release hours (0900 -1600), THEN OBTAIN US/SRO justification and initials in remarks section of Surveillance Task Sheet. [4] OBTAIN US/SRO approval of pre-release data generated by this Instruction and approval for this release. / US/SRO Date [5] INITIATE release of selected WGDT contents in accordance with 0-SO-77 -15 at or below the flow rate (Le., pressure drop) recorded on Appendix B, and RECORD release start time and information requested in the table in Appendix B for release initiation and at one-half hour intervals. NOTE / Time OPERABLE status of 0-FE-77-230 can be determined by noting deflection of indicator. [6] IF [O-FE-77-230J is INOPERABLE at initiation of release, THEN GO TO Step 6.3[8]. [7] IF [O-FE-77-230J becomes INOPERABLE during release, THEN PERFORM the following substeps: [7.1] STOP release. [7.2] NOTIFY US/SRO. SQN Waste Gas Decay Tank Release 0-SI-CEM-077 -410.4 UnitO Rev. 0014 Page 17 of 42 6.3 Release Instructions -Operations (continued) [8] IF release is to continue with [0-FE-77-230JINOPERABLE, THEN PERFORM the following substeps. [8.1] ENSURE that a test gauge (0 -20 inches of H20 suggested) is installed across [0-FE-77-2301. [8.2] ENSURE serial number, range and calibration due date of test gauge along with installing Instrument Mechanic's initials are recorded in remarks section of Appendix B. [8.3] ENSURE pressure readings from test gauge are recorded in place of [0-FE-77-230J readings on Appendix B. [9] IF [0-RM-90-118J or [RM-90-400J alarms, THEN NOTIFY On-shift Chemistry Personnel who will contact the Cognizant Chemist/System Engg. for further guidance in processing tank contents. [10] WHEN release is complete or stopped, THEN RECORD the following on Appendix B. [10.1] Release stop time [10.2] WGDT psig [10.3] Initials [11] IF radiation monitor setpoint changes were made (Step 6.3[2]), THEN RETURN the radiation monitors to their initial setpoints. [12] NOTIFY the US/SRO and On-shift Chemistry Personnel that this release is complete. [13] REVIEW 0-SO-77-15. [14] ATIACH 0-SO-77-15 to this release package. [15] TRANSMIT the release package to the Chemistry Laboratory for post release evaluation. o o o Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 93. 072 A2.02 093 Given the following: Unit 1 operating at 100% power. Unit 2 in MODE 6 with the core off-load in progress An internal electrical failure causes the output of Spent Fuel Pit area radiation monitor 0-RM-90-103 to fail above the HI RAD setpoint. Which ONE of the following identifies how the Auxiliary Building Isolation (ABI) is affected by the failure and whether the actuation is required to be reported to the NRC in accordance with SPP-3.5, "NRC Reporting Requirements"? A. Only Train B is initiated; 8-hour notification required. Only Train B is initiated; 8-hour notification NOT required. C. Both Train A & Train B are initiated; 8-hour notification required. D. Both Train A & Train B are initiated; 8-hour notification NOT required. DISTRACTOR ANAL YSIS: Page 52 A. Incorrect, Only the Train B ABI will be initiated from the monitor but no 8-hour notification would be required because the actuation would be an invalid actuation. Plausible because the initiation of Train B only is correct and ESF actuations are normally reportable. B. CORRECT, Only the Train B ABI will be initiated from the monitor and actuation would be an invalid actuation, thus,no 8-hour notification would be required. C. Incorrect, Both'trains will not be initiated, only the Train B will initiate and no 8-hour notification would be required because the actuation would be an invalid actuation. Plausible because other actuations do come from multiple sensors through isolation/separation relays and ESF actuations are normally reportable. D. Incorrect, Both trains will not be initiated, only the Train B will initiate and no 8-hour notification would be required because the actuation would be an invalid actuation. Plausible because other actuations do come from multiple sensors through isolation/separation relays and the actuation not being reportable is correct. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 93 Tier 2 Group 2 KIA 072 A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure. Importance Rating: 2.8/2.9 Technical Reference: 0-AR-M12-B, Common Radiation Monitor 0-XA-55-12B, Rev 28 SSP-3.5, Regulatory reporting Requirements, Rev 20 NuReg-1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, Rev. 2 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 SPP-3.5 B.2.e. Question Source: For a given condition, determine the regulatory reporting requirements using appropriate reference material. e. State the criteria requiring eight hour notification to the NRC. OPT200.ABVENT B.5.f Describe the operation of the AB Vent system as it relates to the following: f. How a instrument failure will affect system operation. Bank# ___ _ Modified Bank # _X __ _ New ---Question History: Modified SQN question SPP-3.5 008 Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.5 / 43.5 / 45.3 /45.13 ) 10CFR55.43.b ( 5 ) Comments: Modified SQN question SPP-3.5 008 Page 53 SPP-3.5008 QUESTIONS REPORT for BANK SQN Questions During trouble shooting of O-RM-90-1 02, MIG accidently initiated an ABI by inducing a voltage transient. Which one of the following describes the license reporting requirements. A. A four hour notification and LER B. A one hour notification and LER No reporting required D. A LER This situation describes an invalid actuation of ASI. Since invalid actuation of ASI is one of the exceptions to 50.72.b.2.ii, it is not reportable. KIA: 2.1.10, 2.1.14 Tuesday, July 22, 2008 7:44:52 AM 1 ) paragraphs as well as under 10 CFR 50.72(b){3)(v) and 10 CFR 50.73(a)(2)(v) (event or condition that could have prevented the fulfillment of the safety function of .... ). With regard to preplanned actuations, operation of a system as part of a planned test or operational evolution need not be reported. Preplanned actuations are those which are expected to actually occur due to preplanned activities covered by procedures. Such actuations are those for which a procedural step or other appropriate documentation indicates the specific actuation is actually expected to occur. Control room personnel are aware of the specific signal generation before its occurrence or indication in the control room. However, if during the test or evolution, the system actuates in a way that is not part of the planned evolution, that actuation should be reported. For example, if the normal reactor shutdown procedure requires that the control rods be inserted by a manual reactor scram, the reactor scram need not be reported. However, if unanticipated conditions develop during the shutdown that cause an automatic reactor scram, such a reactor scram should be reported. The fact that the safety analysis assumes that a system will actuate automatically during an event does not eliminate the need to report that actuation. Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a planned evolution). Note that if an operator were to manually scram the reactor in anticipation of receiving an automatic reactor scram, this would be reportable just as the automatic scram would be reportable. Valid ESF-actuations are those actuations that result from "valid signals" or from intentional manual initiation, unless it is part of a preplanned test. Valid signals are those signals that are initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. Note tRis definition of "'Velid" refjuires tRet the initietion signel must be en [SF signel. TRis distinction eliminetes ectuetions They do not include those which are the result of other signals from tRe elessof velid ectuetions. Invalid actuations are, by definition, those that do not meet the criteria for being valid. Thus, invalid actuations include actuations that are not the result of valid signals and are not intentional manual actuations. Except for critical scrams, invalid actuations are not reportable by telephone under § 50.72. In addition. invalid actuations are not reportable under § 50.73 in any of the following circumstances: (A) The invalid actuation occurred when the system is already properly removed from service. This means all requirements of plant procedures for removing equipment from service have been met. It includes required clearance documentation, equipment and control board tagging, and properly positioned valves and power supply breakers. (8) The invalid actuation occurred after the safety function has already been completed. An example would be RPS actuation after the control rods have already been inserted into the core. If an invalid ESF-actuation reveals a defect in the ESF-system so the system failed or would fail to perform its intended function,the event continues to be reportable under other requirements of 10 CFR 50.72 and 50.73. When invalid tsF-actuations excluded by the conditions described 49 NUREG-1022, Rev. 2 ') NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0020 Processes Page 14 of 64 5.0 DEFINITIONS (continued) Incident Investigation -Process conducted by the NRC for the purpose of accident prevention. The process includes gathering and analyzing information, determining findings and conclusions, including the cause(s) of a significant operational event; and the disseminating of the investigation results for NRC, industry, and public review. Independent Spent Fuel Storage Installation (ISFSI) -A complex designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. An ISFSI which is located on the site of another facility licensed under §Part 72 or a facility licensed under §Part 50 (e.g., an operating nuclear power plant) and which shares common utilities and services with that facility or is physically connected with that other facility may still be considered independent. Initiation of Shutdown -Physical act of reducing power or temperature to change modes. Invalid Actuation (Signal) -Signals that do not meet the criteria for being valid. Invalid actuations include instances where instrument drift, spurious signals, human error or other invalid signals that result in manual or automatic actuation of the systems listed in §50.73(a)(2)(iv)(8). Major Loss of Communication -Constitutes the loss of communication capabilities. Major Deficiency -A condition or circumstance which under normal operating conditions, an anticipated transient, or postulated design basis accident could contribute to exceeding a safety limit or cause an accident. "Major deficiency" also means a condition or circumstance which in the event of an accident due to other causes could, considering an independent single failure, result in a loss of safety function necessary to mitigate the consequences of the accident. Natural Phenomenon -Act of nature (e.g., fire, flood, tornado). News Release -Known items which may be distributed to the media (UPI, television, radio, newspaper, etc.) and those items identified to be going on TVA news tape distributed by the TVA Public Affairs Staff. Noncompliance (Failure To Comply) -A noncompliance for the purposes of this procedure means any failure to comply with the Atomic Energy Act of 1954, as amended, or with any applicable rule or regulation of the NRC relating to sUbstantial safety hazards. A noncompliance may be in operations, engineering, or construction of the facility or basic component thereof. Organization Manager -This is the most senior manager available who is in the same organization as the individual who discovered the abnormal event. The senior manager is not normally interpreted to be the plant manager or site vice president. Preplanned Sequence -Part of an approved procedure, including workplans, work request, work orders, surveillance instructions, general operating instructions and system operating instructions. Prevented The Fulfillment -Failure or possible failure of a safety system to properly complete a safety function. NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0020 Processes Page 15 of 64 5.0 DEFINITIONS (continued) Principal Safety Barrier -Fuel cladding, RCS pressure boundary, or the containment. Redundant Equipment -Equipment, systems, structures capable of performing the same intended function within the same Technical Specification allowable values. (In most cases, this means opposite train equipment.) Safe Shutdown -Mode 3, as defined by the Technical Specifications. Safety Function -A component or structure designed to actuate upon receiving the proper signal (ESF or RPS). Significant Operational Event -Any radiological, safeguards, or other safety-related operational event at an NRC licensed facility that poses an actual or potential hazard to the public health and safety, property, or the environment. These events or those that typically result in a §50.72 immediate notification. (See Appendix A of this procedure) A significant operational event also may be referred to as "an incident". Examples of these events include: * Operations that exceeded, or were not included in the design basis of the facility, * A major deficiency in design, construction, or operation having potential generic safety implications, * A significant loss of integrity of the fuel, the primary coolant boundary, or the primary containment boundary, * A loss of safety function or multiple failures in systems used to mitigate an actual event, * Significant unexpected system interactions, * Repetitive failures or events involving safety related equipment or deficiencies in operation, * Questions or concerns pertaining to licensee operational performance. Substantial Safety Hazard -Loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety for any facility or activity licensed. Threat -Physical hazard (e.g., fire, severe radioactive release). Unanalyzed Condition -Plant Condition outside the bounds of the initial conditions as described in the FSAR accident analysis. Valid Actuation (Signal) -Valid actuations are those that result from "valid signals" or from intentional manual initiation. Valid signals are those that are initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. NPG Standard Programs and Processes Regulatory Reporting Requirements Appendix A (Page 3 of 11) 3.1 Immediate Notification -NRC (continued) SPP-3.S Rev. 0020 Page 19 of 64 3. §50.72(b).(1)) -Any deviation from the plant's Technical Specifications authorized pursuant to §50.54(x). C. The following criteria require 4-hour notification: 1. §50. 72(b )(2)(i) -The initiation of any nuclear plant shutdown required by the plant's Technical Specifications. 2. §50.72(b)(2)(iv)(A) -Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 3. §50.72(b )(2)(iv)(8) -Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 4. §50.72(b )(2)(xi) -Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials. D. The following criteria require 8-hour notification: NOTE The non-emergency events specified below are only reportable if they occurred within three years of the date of discovery. 1. §50.72(b )(3)(ii)(A) -Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. 2. §50.72(b)(3)(ii)(8) -Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. 3. §50. 72(b )(3)(iv)(A) -Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(8) [see list below], except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. a. Reactor protection system (RPS) including: Reactor scram and reactor trip. ') ) NPG Standard Programs and Processes Regulatory Reporting Requirements Appendix A (Page 5 of 11) SPP-3.5 Rev. 0020 Page 21 of 64 3.1 Immediate Notification -NRC (continued) NOTE According to §50.72 (b)(3)(vi) events covered by §50.72(b)(3)(v) may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant this paragraph if redundant equipment in the same system was operable and available to perform the required safety function. 5. §50.72(b)(3)(xii) -Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment. 6. §50.72(b)(3)(xiii) -Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system). E. Follow-up Notification (§50.72(c>> With respect to the telephone notifications made under paragraphs (a) and (b) [§50.72 (a) and §50.72 (b), respectively] of this section [§50.72], in addition to making the required initial notification, during the course of the event: a. Immediately report (i) any further degradation in the level of safety of the plant or other worsening plant conditions including those that require the declaration of the Emergency Classes, if such a declaration has not been previously made; or (1) Any change from one Emergency Class to another, or (2) A termination of the Emergency Class. b. Immediately report (i) the results of ensuing evaluations or assessments of plant conditions, (1) The effectiveness of response or protective measures taken, and (2) Information related to plant behavior that is not understood. c. Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. 10 (B-3) Source Setpoint SER 760 (Unit 1 annunciator system) 02 50 mr/hr > 1 second 0-RA-90-102A FUEL POOL RAD MONITOR HIGH RAD Retransmitted to U-2 SER 2243 (Unit 2 annunciator system) Probable Causes Corrective Actions 1. High radiation in spent fuel pit area elevation 734. [1] IF dry cask storage processing in progr*ess, THEN NOTIFY Cask Supervisor. [2] CHECK 0-RM-90-1 02 and 0-RM-90-103 on 0-M-12 to verify alarm. NOTE 0-RM-90-102 will be blocked and is expected to alarm during portions of Dry Cask Storage loading/unloading activities. Under this condition, remaining steps are N/A. [3] VERIFY the following: a. Auxiliary Building General Supply and Exhaust and Fuel Handling exhaust isolate (A-Train) (1-M-9). b. Auxiliary Building Gas Treatment System starts (1-M-9). [4] IF high radiation alarm valid, THEN [a] ANNOUNCE "High Radiation at spent Fuel Pool Area" over PA system. [b] NOTIFY SM. [c] NOTIFY RADCON. [5] IF B-Train ABI has not actuated from a valid High Radiation condition, THEN INITIATE manually B-Train Auxiliary Building Ventilation Isolation via [1-HS-30-101B] or [2-HS-30-101BJ (M-6). [6] IF fUel handling in the Spent Fuel Pit is in progress, THEN REFER TO AOP-M.04, Refueling Malfunctions. [7] REFER TO AOP-M.06, Loss of Spent Fuel Cooling. [8] IF Auxiliary Building Ventilation Isolation resulted from an invalid ABI signal, THEN, REFER to 0-SO-30-10 Auxiliary Building Ventilation Systems to recover from ABI. (Continued on next page) SQN 0-AR-M12-B Page 12 of 40 o . Rev. 28 Corrective Actions (Continued) References 10 (B-3) 0-RA-90-102A FUEL POOL RAD MONITOR HIGH RAD [9] EVALUATE EPIP-1, Emergency Plan Classification Matrix. [10] EVALUATE Technical Specifications 3.3.3.1 and 3.9.12. [11] INITIATE Corrective Actions. [12] WHEN conditions return to normal, THEN RETURN Auxiliary 8uilding Ventilation System to normal in accordance with 0-SO-30-10, Auxiliary Building Ventilation Systems. . 458655-12B-O,47W610-90-1 SQN O-AR-M12-B Page 13 of 40 0 Rev. 28 11 (B-4) Source Setpoint SER 761 (Unit 1 annunciator system) O-RM-90-102 N/A O-RA-90-102B FUEL POOL RAD MONITOR INSTR MALFUNC Retransmitted to U-2 SER 2244 (Unit 2 annunciator system) Probable Causes Corrective Actions References 1. Instrument downside ratemeter trip. 2. Instrument loss of power. 3. Instrument placed in TRIP ADJ position. [1] CHECK O-RM-90-102 on O-M-12 to attempt to determine problem. [2] NOTIFY RADCON. [3] DISPATCH personnel to check O-RM-90-1 02 locally to determine problem. [4] EVALUATE Technical Specification 3.3.3.1 and 3.9.12. [5] INITIATE WO for maintenance, if required. 458655-12B-0, 47W610-90-1 SQN 0 Page 14 of 40 O-AR-M12-8 Rev. 28 I. PROGRAM: OPERATOR TRAINING OPT200.ABVENT Rev. 1 Page 3 of94 II. COURSE: SYSTEMS TRAINING III. TITLE: AUXILIARY BUILDING VENTILATION SYSTEM IV. LENGTH OF LESSON: Initial License Training: 3 hour lecture; 1 hour simulator demonstration; Ihour self-study/workshop V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Auxiliary Building Ventilation systems (AB Vent) in the plant and on the simulator. B. Enabling Objectives: O. Demonstrate an understanding ofNUREG 1122 knowledge's and abilities associated with the Auxiliary Building Ventilation system that are rated 2.5 during Initial License training for the appropriate license position as identified in Appendix A. 1. State the purpose/functions of the Auxiliary Building Ventilation system as described in the SQN FSAR. 2. State the design basis of the Auxiliary Building Ventilation system in accordance with the SQN FSAR. 3. Explain the purpose/function of each major component in the flow path of the Auxiliary Building Ventilation system as illustrated on the simplified system drawing. 4. Describe the following items for each major component in the Auxiliary Building Ventilation system as described in this lesson: a. Location b. Power supply (include control power as applicable) c. Support equipment and systems d. Normal operating parameters e. Component operation f. Controls g. Interlocks (including setpoints) h. Instrumentation and Indications i. Protective features (including setpoints) j. Failure modes k. Unit differences 1. Types of accidents for which the Auxiliary Building Ventilation system components are designed m. Location of controls and indications associated with the Auxiliary Building Ventilation system in the control room and auxiliary control room V. TRAINING OBJECTIVES (Cont'd): B. Enabling Objectives (Cont'd): 5. Describe the operation of the AB Vent system as it relates to the following: a. Precautions and limitations b. Major steps performed while placing the AB Vent system in service c. Alarms and alarm response d. How a component failure will affect system operation e. How a support system failure will affect AB Vent system operation f. How a instrument failure will affect system operation 6. Describe the administrative controls and limits for the AB Vent system. 7. State Tech Specs/TRM LCOs that govern the AB Vent system a. State the hour action limit TS LCOs OPT200.ABVENT Rev. 1 Page 4 of94 b. Given the conditions/status of the AB Vent system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required 8. Discuss related Industry Events: Event Title: Hot Particle Discovered on Auxiliary Building Roof. INPO Event # 327-940131-1 VI. TRAINING AIDS: A. Computer. B. Computer Display Projector & Controls. C. Local Area Network (LAN) Access. D. Simulator (if available) OPL271 SPP-3.5 Revision 1 Page 3 of 21 ) I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: REPORTING REQUIREMENTS IV. LENGTH OF LESSON/COURSE: 4 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the actions necessary to comply with regulatory and plant reporting requirements. B. Enabling Objectives o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with Regulatory and Plant Reporting Requirements that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Perform a plant response assessment using the 0-TI-QXX-000-001.0," Event Critique, Post Trip Report, and Equipment Root Cause,". a. State the responsibilities of each control room crew member. [C.1] b. Explain the process or Conduct a plant response assessment. 2. For a given condition, determine the regulatory reporting requirements using appropriate reference material. a. List the tools available to the operator for determining regulatory reporting requirements. b. Define the key terms used to determine regulatory reporting requirements. c. State the criteria requiring one hour notification of the NRC. d. State the criteria requiring four hour notification of the NRC. e. State the criteria requiring eight hour notification of the NRC. f. State the criteria requiring 24 hour notification of the NRC. g. State the criteria requiring 2 day notification of the NRC. h. State the criteria requiring a written report or LER to the NRC. i. State the criteria allowing a telephone notification to be made in lieu of a written LER to the NRC. 3. For a given condition, determine plant management reporting requirements using SPP-3.5. 4. Complete a PER reportability determination per SPP-3.1. ---------I i i I Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 94. G 2.1.25 094 Given the following: Unit 1 at 100% power with Boric Acid Tank (BAT) A aligned. RWST Boric Acid concentration is 2575 ppm. Which ONE of the following identifies BAT A Volume and Boric Acid Concentration that will meet Operability requirements and when the maximum expected boration capability requirement occurs in accordance with the Technical Requirement Bases? REFERENCE PROVIDED A. 9600 gallons at 6500 ppm; Near End of Life peak Xenon conditions. B. 9600 gallons at 6500 ppm; Near Beginning of Life peak Xenon conditions. 9200 gallons at 6775 ppm; Near End of Life peak Xenon conditions. D. 9200 gallons at 6775 ppm; Near Beginning of Life peak Xenon conditions. Page 54 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 55 A. Incorrect, BA T A volume of 9600 gallons at 6500 ppm with an RWST boron concentration of 2575 ppm is in the unacceptable region of TRM Figure 3.1.2.6 for BA T Tank Limits. Near the of life peak xenon is when the maximum expected boration capability requirement occurs as identified in Technical Requirement bases. Plausible because the BA T volume, BA T boron concentration, and the RWST boron concentration are near the limit and require interpolation when using the graph. Also, the higher level in the tank is not the correct answer and the limiting conditions occurring at near EOL is correct. B. Incorrect, BA T A volume of 9600 gallons at 6500 ppm with an RWST boron concentration of 2575 ppm is in the unacceptable region of TRM Figure 3.1.2.6 for BAT Tank Limits, and the Beginning of life peak xenon is not when the maximum expected boration capability requirement occurs. Technical Requirement bases identifies the requirement to be at near EOL with peak Xenon conditions. Plausible because the BA T volume, BA T boron concentration, and the RWST boron concentration are near the limit and require interpolation when using the graph. Also, the higher level in the tank is not the correct answer and for other circumstances the limiting conditions occur at the beginning of life. C. CORRECT, BA T A volume of 9200 gallons at 6775 ppm with an RWST boron concentration of 2575 ppm is in the acceptable region of TRM Figure 3.1.2.6 for BAT Tank Limits and the bases for the Technical Requirement states that the maximum expected boration capability requirement occurs at near EOL with peak Xenon conditions. D. Incorrect, BA T A volume of 9200 gallons at 6775 ppm with an RWST boron concentration of 2575 ppm is in the acceptable region of TRM Figure 3.1.2.6 for BAT Tank Limits, but the Beginning of life peak xenon is not when the maximum expected boration capability requirement occurs. Technical Requirement bases identifies the requirement to be at near EOL with peak Xenon conditions. Plausible because the combination of BA T volume and boron concentration place the tank in an acceptable region of the graph and other limiting conditions to occur at the beginning of life. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 94 Tier 3 KIA G 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. Importance Rating: 3.9/4.2 Technical Reference: Technical Requirements Bases 3.1.2.6 Proposed references to be provided to applicants during examination: TRM Figure 3.1.2.6, Boric Acid Tank Limits Based on RWST Boron Concentration Learning Objective: OPT200.TRM B.4 Explain the TRM bases for each LCO (KIA 2.2.6) Question Source: Bank# ___ _ Modified Bank # ___ _ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: (41.10/43.5/45.12 ) 10CFR55.43.b ( 2) Facility operating limitations in the technical specifications and their bases. Comments: New question for Sequoyah 2009 exam Page 56 ". .-*** " ----... ...... -...... "--..... -... ....... _ .......... _._..,,,. __ " **

J" __ * ) REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES -OPERATING LIMITING CONDITION FOR OPERATION TR 3.1.2.6* As a minimum, the following borated water source(s) shall be OPERABLE as required by TR 3.1.2.2: a. A boric acid storage system with: 1. A contained volume of borated water in accordance with Figure 3.1.2.6, 2. A boron concentration in accordance with Figure 3.1.2.6, and 3. A minimum solution temperature of 63°F. b. The refueling water storage tank with: 1. A contained borated water volume of between 370,000 and 375,000 gallons, 2. Between 2500 and 2700 ppm of boron, 3. A minimum solution temperature of 60°F, and 4. A maximum solution temperature of 105°F. APPLICABILITY: MODES 1, 2, and 3. ACTION: a. With the boric acid storage system inoperable and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1 % delta klk at 200°F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 30 hours. b. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS 3/41-8 January 4, 2001 Revision Nos. 13 p TRM FIGURE 3.1.2.6 (Units 1 & 2) BORIC ACID TANK LIMITS BASED ON RWST BORON CONCENTRATION \ l.f') ( 11000 'I. 1 .. I I I i{ I 1 10000 I I \ "'l. "" i ........ '" "' / \ ' I' I U) I J q\9 vJ \ J ... JAKNST=2700ppmE J I I 9500 I I i I I ,I 3 !' "\" "i: '\ \ t -'{vo I 0 ! I' i"" .... ........ , > 9000 I I ,,' 'I \ I I I i ..... I I, Z I'! I ' .......::! I )\ I I ! I ' ,I r-... I ........ I § \ I \ I I I \', I 8500' I, , '. '" I U I I I I, I I ,I ( 6990 ppm (Maximun) J,r-...., I I '\ Ii! ! \' I w i'-i ! ! !'"", .-8000 I I " I ; I I i \ \ 1 ppm I \ I ! I I I I \ I I \ \ I I \ \ 7500 I, I, , ,I * I , \ I I I i I I I If REGION OF UNACCEPTABLE OPERATION i 7000 t I \ r l i I \ I \ II I i Ii I [i I i 1'1 I I --'-----_ .. _------. -. Indicated values indude 1140 gal unusab e volume and 800 gal f r i 6000 6100 6200 6300 6400 6500 6600 6700 I '800 6900 7000 7100 BORIC ACID TANK CONCENTRATION -BORON RWST Concentration l) ry '( -++-2600 PPM ---2650 PPM -'-2700 PPM 1 r -+-2500 PPM ---2550 PPM september 26. 2003 SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS 3/41-10 Revision Nos. 13.26.27 TRM FIGURE 3.1.2.6 (Units 1 &2) BORIC ACID TANK LIMITS ' .. "'. BASED ON RWST BORON CONCENTRATION 11000 1 1 I I I ! 1 I ! I I I OF OPERATION' I I I I 10500, I "-I -'-J500 B I I I I I I I I.! I I I I I J RWST = 2550 ppm B 1 I 10000 C/) I ...... " vi ' . I I K }t RWST = 2600 ppm B ) I i " ....... 1 I I f' lix VI ' I ' I , I RWST = 2650 ppm B J " I I z 0 ...J ...J << 9500 C> w :::> ...J 0 9000 > , z << l-e u << 8500 u 0 co e w I-8000 << 5:2 e z I 1 l' K ;Q' = 2700 ppm E I ,! I I , i'.. 0 I I I I .... 1 1 t I I I , 0 I I , ,! I I I I I , I I "'I I 1 I 1 I i I i ! ..... 1 ....... 1 I 1 I I I I I i'l" I'tl I I I I I I I I I I , I [ 6990 ppm (Maximum) J, I I I, I I I i I i 1 I ! I ! i I I 11 6120 ppm (Minimu r) I I I I I 1 ! I I ! 7500 I I I I I I I I I I I I I ! I I I I I i I I I I ! I REGION OF UNACCEPTABLE OPERATION I I 7000 I I I I I I I I I I I ! I , ! : I I I I I I I I i I I ! -Indicated values indude 1140 gal unusable volume and 800 gal for instrument error. 1>-1 I I I I I I I I I I I 6500 6000 6100 6200 6300 6400 6500 6600 6700 6800 6900 7000 7100 BORIC ACID TANK CONCENTRATION -PPM BORON RWST Concentration -+-2550 PPM -+E-'-2600 PPM -2650 PPM TECHNICAL REQUIREMENTS 3/41-10 REACTIVITY CONTROL SYSTEMS ') BASES ) TRB 3/4.1.2 BORA TION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 350°F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta klk after xenon decay and cooldown to 200°F. The maximum expected boration capability requirement occurs at near EOl from full power peak xenon conditions and requires borated water from a boric acid tank in accordance with Figure 3.1.2.6, and additional makeup from either: (1) the common boric acid tank and/or batching, or (2) a minimum of 26,000 gallons of 2500 ppm borated water from the refueling water storage tank. With the refueling water storage tank as the only borated water source, a minimum of 57,000 gallons of 2500 ppm borated water is required. The boric acid tanks, pumps, valves, and piping contairi a boric acid solution concentration of between 3.5% and 4.0% by weight. To ensure that the boric acid remains in solution, the air temperature is monitored in strategic locations. By ensuring the air temperature remains at 63°F or above, a 5°F margin is provided to ensure the boron will not precipitate out. To provide operational flexibility, if the area temperature should fall below the required value, the solution temperature (as determined by the pipe or tank wall temperature) will be monitored at an increased frequency to compensate for the lack of solution temperature alarm in the main control room. With the ReS temperature below 350°F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and operations involving positive reactivity additions that could result in loss of required SDM (Modes 40r 5) or boron concentration (Mode 6) in the event the single injection system becomes inoperable. Suspending positive reactivity additions that could result in failure to meet minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that haVe a boron concentration greater than or equal to that required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM. The boron capability required below 350°F, is suffiCient to provide a SHUTDOWN MARGIN of 1.6% delta klk after xenon decayand cool down from 350"F to 200°, and a SHUTDOWN MARGIN of 1% delta klk after xenon decay andcooldown from 200°F to 140°F. This condition requires either 6400 gallons of 6120 ppm borated water from the boric acid storage tanks or 13,400 gallons of 2500 ppm borated water from the refueling water storage tank. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The 6400 gallon limit in the boric acid tank for Modes 4, 5, and 6 is based on 4,431 gallons required for shutdown margin, 1,140 gallons for the unusable volume in the heel of the tank, 800 gallons for instrument error, and an additional 29 gallons due to rounding up. The 55,000 gallon limit in the refueling water storage tank for modes 4, 5,and6 is based upon 22,182 SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS B 3/41-2 October 19, 2005 Revision Nos. 13,25,35,36 REACTIVITY CONTROL SYSTEMS ') BASES )\ ) , ') gallons that is undetectable due to lower tap location, 19,197 gallons for instrument error, 13,400 gallons required for shutdown margin, and an additional 221 gallons due to rounding up. The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reaCtivity control while in MODE 6. SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS B 3/41-3 October 19,2005 Revision Nos. 13 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: TECHNICAL REQUIREMENTS MANUAL IV. LENGTH OF LESSON/COURSE: 2 hour(s) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of this lesson and others presented, the student shall demonstrate an understanding of the purpose and background of the Technical Requirements Manual by successfully completing a written examination with a score of 80%. B. Enabling Objectives: Slide 2 In order to accomplish these objectives, the student shall be able to successfully: 1. Determine plant mode of operation from memory. (KIA 2.1.22) 2. Apply TRM action statements of greater than one hour given a copy of the TRM (KIA 2.1.12, 2.2.22) 3. Explain the process for making changes to the TRM (KIA 2.2.6) 4. Explain the TRM bases for each LCO. (KIA 2.2.25) 5. Apply less than one hour LCO actions using a copy of the TRM. (KIA 2.1.11) Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 95. G 2.2.21 095 Given the following: -During the Operations review of a Work the SRO determines the return to operability (RTO) Post Maintenance Test (PMT) needs revision. -The WO did not require an Independent Qualified Review (IQR), a 10CFR50.59 review, or 10CFR72.48 review. In addition to documenting the reason for the revision, which ONE of the following identifies both ... (1) the requirement to revise the RTO test without routing the WO back through the review process and (2) when this type PMT revision would require Shift Manager approval? A'! (1) Revision signed by 2 SROs. (2) when a generation risk critical component is involved. B. (1) Revision signed by 2 SROs. (2) when a configuration change to a component on the unit is required to perform the PMT. C. (1) Revision signed by the Work Week Manager and an SRO. (2) when a generation risk critical component is involved. D. (1) Revision signed by the Work Week Manager and an SRO. (2) when a configuration change to a component on the unit is required to perform the PMT. Page 57 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: Page 58 A. CORRECT, SPP-6.3 allows revisions to RTO tests to be made by 2 SROs without routing back through the review cycle for the conditions identified in the stem of the question and does require Shift Manager approval if a generation critical component is involved. B. Incorrect, SPP-6.3 allows revisions to RTO tests to be made by 2 SROs without routing back through the review cycle for the conditions identified in the stem of the question but does not require Shift Manager approval because a configuration change on the unit would be involved. Plausible because 2 SRO being required to make the revision is correct and verification of configuration changes are identified in the SPP as needing independent verification. C. Incorrect, SPP-6.3 allows revisions to RTO tests to be made by 2 SROs, not an SRO and the Work Week Manager but Shift Manager approval if a generation critical component is involved is correct. Plausible because the Work Week Manager has other functions and requiring Shift Manager approval if a generation critical component is involved is correct. D. Incorrect, SPP-6.3 allows revisions to RTO tests to be made by 2 SROs, not an SRO and the Work Week Manager and Shift Manager approval because a configuration change on the unit would be involved is also not correct. Plausible because the Work Week Manager has other functions and verification of configuration changes are identified in the SPP as needing independent verification. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 95 Tier 3 KIA G 2.2.21 Knowledge of pre-and post-maintenance operability requirements. Importance Rating: 2.9/4.1 Technical Reference: SPP-6.3, Pre-/ Post-Maintenance Testing, Rev 2 Proposed references to be provided to applicants during examination: None Learning Objective: No training objective identified Question Source: Bank # ----Modified Bank # ----:-_____ --New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: (41.10/43.2 ) 10CFR55.43.b ( 2,3 ) Comments: New question for Sequoyah 2009 exam Page 59 TVAN STANDARD PROGRAMS AND PROCESSES PRE-/POST -MAINTENANCE TESTING SPP-6.3 Rev. 2 Page 9 of 16 3. A formal instruction prepared for the WO PMT receives an IQR, 10 CFR 50.59 and/or 10 CFR 72.48 review in accordance with SPP-2.2, "Administration of Site Technical Procedures." System Engineer F. Review WO PMTs (Maintenance Testing and RTO Testing) when requested for TS, ASME Section XI, 10 CFR 50 Appendix J, or other complex activities to ensure that they are correct and concur with the PMT. Operations Shift Manager/SRO Designee G. Perform initial reviews of WO in accordance with the criteria specified in SPP-7.1, "Work Control Process." H. Review/revise RTO tests as necessary to ensure TS operability and surveillance requirements are met without imposing any adverse affects on the system or equipment. Contact System Engineering if assistance is needed. Changes to RTO tests are made in accordance with 3.4.N below or by revision to the work order. I. Approve the RTO tests. J. Ensure that the PMT satisfies all TS requirements. K. Include status control requirements in the WO package. Specify the correct systems and component alignment required for the PMT performance and the configuration required to restore systems and components to ensure correct operating or standby mode following the completion of the PMT. L. Ensure that the PMTs are performed at the appropriate system operating conditions or plant modes. M. Review WO scope changes and revise RTO tests as necessary in accordance with 3.4.N below or by revision to the work order. N. Revise RTO tests when warranted. Revisions to RTO tests may be made by two SROs without routing back through the review cycle provided that 1) the reason is documented, 2) both SROs sign the revision and 3) IQR, 10 CFR 50.59 and/or 10 CFR 72.48 review was not required previously. This type revision requires Shift Manager approval if the PMT involves a generation risk critical component as classified in MEL. O. Waive PMT when warranted. A PMT requirement may be waived provided that the Shift Manager authorizes the waiver, the reason is documented (e.g., plant conditions prevent functional testing of a logic circuit), and affected configuration changes are verified by independent verification. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 96. G 2.2.6 096 Given the following: Unit 1 is in MODE 5 during a refueling outage. It is determined that 1-S0-63-1, "Cold Leg Injection Accumulators" needs an Urgent "Minor/Editorial Change" revision to support system alignment changes due to system modification. Which ONE of the following statements identifies the need for an Indepenent Qualified Reviewer (IQR) and if the Unit SRO can be the approval authority for the procedure revision? A'! An IQR is required. The Unit SRO may sign as the approval authority. B. An IQR is required. The Unit SRO may NOT sign as the approval authority. C. An IQR is NOT required. The Unit SRO may sign as the approval authority. D. An IQR is NOT required. The Unit SRO may NOT sign as the approval authority. DIS TRA CTOR ANAL YSIS: Page 60 A. CORRECT, SPP-2.2 requires an IQR review for minor editorial changes for Quality Related procedures and 1-S0-63-1 is a Quality Related procedure the SPP also provides for the SRO on the Unit being the approval authority. B. Incorrect, IQR review for minor editorial changes for Quality Related procedures is required and SPP-2.2 allows for the SRO on the Unit being the approval authority. Plausible because the IQR review is required for minor/editorial changes to quality related procedures and the SRO on the unit is not the approval authority for normal procedure revisions. C. Incorrect, SPP-2.2 requires IQR review for minor editorial changes for Quality Related procedures and allows for the SRO on the Unit being the approval authority. Plausible because the IQR review is not required for minor/editorial changes to non-quality related procedures and the SPP allows for the SRO on the unit being the approval authority. D. Incorrect, IQR review for minor editorial changes for Quality Related procedures is required and allows for the SRO on the Unit being the approval authority. Plausible because the IQR review is not required for minor/editorial changes to non-quality related procedures and the SRO of the unit is not the approval authority for normal procedure revisions. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 96 Tier 3 KIA G 2.2.6 Knowledge of the process for making changes to procedures. Importance Rating: 3.0/3.6 Technical Reference: SPP-2.2, Administration of Site Technical Procedures, Rev 15 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271.SPP-2.2 B.5 Describe the procedure revision process. Question Source: Bank# ___ _ Modified Bank # _X, ___ _ New ---Question History: Question modified from Summer 2006 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.3/45.13 ) 10CFR55.43.b (3,5) Comments: Question modified from Summer 2006 exam Page 61 ) ,\ NPG Standard Administration of Site Technical SPP-2.2 Programs and Procedures Rev. 0015 Processes Page 20 of 42 3.4.4 Comments, Approval & Implementation A. Reviewers should provide comments in BSL, or Form SPP-2.2-2, "Site Technical Procedure Review, Comment, and Concurrence Form", or by other appropriate means. B. Preparers shall resolve comments and escalate unresolved comments to appropriate management. C. The responsible organization obtains approval from the appropriate approval authority. The approval also ensures, as applicable, the reviewer is independent of the preparer and qualified to perform the review. Approval by telecon with the manager/supervisor who is the designated approval authority is acceptable for hard copy changes. D. If required, PORC shall review and concur with the procedure. The PORC Chairman recommends to the Plant Manager his approval. Upon concurrence, the Plant Manager shall approve the procedure. NOTES 1) PORC review is required for all procedures identified as a CIPTE and all revisions to those procedures. 2) Minor/editorial revisions do not require PORC review. E. Following approval, the responsible manager establishes the effective date taking into consideration any work in progress or parallel changes, training, etc., identified and transmits the procedure package to MS for distribution and EDM archival. The effective date is normally three working days after receipt by MS. NOTE The 10 CFR 50.59 and 10 CFR 72.48 documents will be archived in EDM as standalone documents. F. Revisions which will cause Operations to change the normal configuration of a component (i.e., changes to normal valve alignment checklists, switch position checklists, etc.), should be communicated to the affected Shift Manager or Unit Supervisor prior to the effective date. 3.5 Minor/Editorial Changes A. Minor changes, such as inconsequential editorial corrections that do not affect the outcome, results, functions, processes, responsibilities, and requirements of the performance of procedure or instructions, require review by an lOR (quality-related procedures only) and approval by the appropriate approval authority. Minor changes do not require 10 CFR 50.59 review, 10 CFR 72.48 review, or PORC review. Minor changes shall not change the intent of the procedure or alter the technical sequence of procedural steps. NPG Standard Administration of Site Technical SPP-2.2 Programs and Procedures Rev. 0015 Processes Page 21 of 42 3.5 Minor/Editorial Changes (continued) B. Procedure changes that meet the following criteria are considered minor changes: 1. correction of punctuation, style changes 2. redundant or insignificant word or title changes 3. correction of typographical errors including capitalization 4. annotation of critical steps, 5. correction of reference errors 6. omitted symbols that do not alter results 7. incorrect units of measure due to editorial error 8. misplaced decimals that are neither setpoint values nor tolerances g. page number discrepancies 10. missing sign-offs, signatures, or date lines 11. corrections to attachment identifiers 12. corrections to titles of plant organizations, position titles, departmenUsection/unit names when there is no change in authority, responsibility, or reporting relationships. 13. corrections to addresses, telephone numbers, or computer application names 14. corrections to or additions of equipment nomenclature or locations in procedures to be consistent with approved drawings, documents, labels, or procedure content. 15. addition of or changes to equipment unique identifier information (UNID) in procedures consistent with design output documents and which do not alter what component is operated 16. corrections to or clarification of a note or precaution which does not alter the method of accomplishing a task 17. changes which are purely administrative and non-technical in nature which do not change the intent or outcome of an activity (e.g. adding a step requiring a log entry, a plant announcement, informational notifications, or initiation of a PER) C. A BSL System Administrator or Sponsor may make the following changes: 1. Organizational changes. 2. Reference changes, e.g., MMI is superseded by MCI-0-000-TRB001. NPG Standard Administration of Site Technical SPP-2.2 Programs and Procedures Rev. 0015 Processes Page 22 of 42 3.5 Minor/Editorial Changes (continued) 3. Misspelled words. 4. Language, grammar, syntax corrections. D. The revision description will describe the reason for change. E. The approval authority will obtain the lOR for quality-related procedures and approve the change. F. The procedure package is transmitted to MS for distribution and EDM archival. 3.6 Urgent Procedure Changes Urgent changes to procedures are revisions which are deemed necessary by plant management to maintain plant safety, operability or critical schedules and inadequate time exists to make a normal revision using BSL. Urgent changes may be handwritten and require the following: A. Tracking Numbers from BSL. B. Affected organization/Cross Discipline Review (see Sections 3.4.2D, 3.4.3C and 3.4.3D for review determination/requirements) when other organizations are affected by the change and obtain necessary signatures on the PCF. C. Technical review by an lOR. D. A 10 CFR 50.59 screening review, if required, in accordance with SPP-9.4, "10 CFR 50.59 Evaluations of Changes, Tests, and Experiments" or 10 CFR 72.48 Screening Review if required, in accordance with SPP-9.9, "10 CFR 72.48 Evaluations of Changes, Tests, and Experiments for Independent Spent Fuel Storage Installation." E. Approval by the PORC, if required, and the approval authority. A licensed SRO on the unit affected may sign as the approval authority for minor/editorial changes. F. The preparer shall obtain applicable reviews and approval documented on PCF. The preparer shall submit a hardcopy to the affected unit control room, if filed in the control room. The preparer shall forward the original to MS. The procedure may then be used to perform work. G. The preparer shall ensure the change is processed and provided to MS for distribution. The preparer shall submit a copy to the sponsor by the next working day. H. The responsible organization should ensure the handwritten change is incorporated and processed into BSL within 14 days for changes initiated during normal plant operation or within 30 days following an outage. NPG Standard Administration of Site Technical SPP-2.2 Programs and Procedures Rev. 0015 Processes Page 23 of 42 3.7 One-Time-Only Procedures and Revisions One-Time-Only (OTO) procedures, and one-time revisions to existing procedures, can be developed and approved for use where existing procedures do not adequately address the activity due to unusual plant conditions. Such procedures/revisions are intended to be used only once and will automatically expire following completion of the use of the procedure. A. OTO procedures/revisions shall be reviewed and approved as specified in Sections 3.4, 3.5, through 3.6 of this procedure. The preparer shall indicate that the revision is a OTO revision on the PCF. The tracking number on the PCF should be N/A'd. B. The performer shall ensure that the revision is inserted/incorporated into the controlled copy of the procedure being used for performance of the activity. OTO procedures/revisions will not be distributed to other controlled manuals. C. The completed procedure, PCF, and all other supporting documents are transmitted to MS to be retained as a record. D. OTO procedures/revisions do not require cancellation. Following completion of performance, the active revision existing before the OTO revision was made will be the active version of the procedure. 3.8 Administrative Hold A. Responsible organizations are required to place procedures on "administrative hold" when a revision is not feasible. The cover sheet of the procedure should indicate administrative hold. The responsible procedure sponsor shall forward the PCF (or audit trail) and the cover sheet to MS for placement on Administrative Hold. B. MS shall not issue working copies of procedures on administrative hold. C. Employees shall not use procedures on administrative hold to perform work. Only the Plant Manager/Duty Plant Manager can authorize issuance of a working copy of a document that is on administrative hold. D. All reasonable measures shall be exhausted to release (if applicable) the procedure from administrative hold through the procedure revision process. The Plant Manager should carefully evaluate the situation before releasing the procedure for work. E. When the reason for the hold has been resolved, the sponsor shall initiate a procedure revision (Section 3.4) to request release from administrative hold. Include the reason for administrative hold removal in the revision log. F. The following procedures, due to their usage should have deficiencies corrected immediately, and should not be placed on administrative hold: 1. Abnormal Operating Instructions. 2. Emergency Operating Instructions. (E, ECA, FRG, ES). 3. Emergency Preparedness Implementing Procedures. QUESTIONS REPORT for summer 10-4-06 II G2.2.6001 The RO is attempting to vent the PRT in accordance with SOP-101, Reactor Coolant System, when he notes that a valve not identified in the SOP needs to be open to complete venting. The CRS determines that a restricted change is required. Which ONE (1) of the following is the MINIMUM review/approval necessary for this procedure change? a. Manager, Operations. b. Duty Shift Supervisor. c ..... Qualified Reviewer AND the Duty Shift Supervisor. d. Procedure Group Supervisor AND the Plant Manager. Requires examinee to determine that a restricted change is classified as a temporary change and then applies this to requirements in SAP-139. SAP-139 requires a Qualified Reviewer and the Duty Shift Supervisor as minimum review/approval for approval of a Operations restricted procedure change. Other Discipline Supervisors may approve changes within their discipline. Knowledge of the process for making changes in procedures as described in the safety analysis report. Question Number: Tier 3 Group 2 Importance Rating: 3.3 SR021 Technical Reference: SAP-139 Proposed references to be provided to applicants during examination: None Learning Objective: 4378 Question Source: Bank Question History: VCS Bank Question Cognitive Level: Lower 10 CFR Part 55 Content: 41 Comments: NRC Comment: Question appears to match KIA, and is SRO knowledge. SAT BANK 1 OCFR55.43(b) is met because the applicant must understand the approval process for changes made to facility procedures. Friday. July 18. 2008 2:43:38 PM 1 OPL271 SPP-2.2 Revision 1 Page 3 of 31 ) I. PROGRAM: OPERATOR TRAINING ) ) II. COURSE: LICENSE TRAINING III. LESSON TITLE: SPP-2.2, ADMINISTRATION OF SITE TECHNICAL PROCEDURES IV. LENGTH OF LESSON/COURSE: 1 Hour V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, an understanding of SPP-2.2 "Administration of Site Technical Procedures" and OPDP-1 Attachment F "Plant Operating Procedures." B. Enabling Objectives: o. Demonstrate an understanding of NUREG 1122 Knowledge's and Abilities associated with Rules of Procedure Use that are rated;;::: 2.5 during Initial License Training and;;::: 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A 1. Discuss the purpose of SPP-2.2. 2. Discuss management philosophy for procedure use and adherence. 3. List the four levels of use for technical procedures including examples of each level. 4 Give examples when a procedure step may be marked not applicable (N/A). 5 Describe the procedure revision process. 6 Briefly define a minor/editorial change. 7 Describe the conditions under which personnel may take reasonable action within the scope of their training that departs from procedure. 8 Explain how to obtain and verify controlled procedure copy. i QUESTIONS REPORT for additional Questions G 2.3.13 197 iven the following: Unit 2 was shutdown for refueling on 01/20109 at 2400. Today is 01/25/09 at 0200 and the Unit is in Mode 5. Lower cavity fill has not been started Transfer Tube Wafer Valve, 2-78-610 is required to be open for repair on the transfer tube and transfer system. In accordance with 0-GO-9, "Refueling Operations", which ONE of the following identifies two conditions where either one of the two could be used to minimize the potential spread of airborne contamination? A. Containment Equipment Hatch ... open Containment Purge System ... in service and aligned to upper containment B:t open stopped. C. D. held closed by a minimum of 4 bolts. held closed by a minimum of 4 bolts. in service and aligned to upper containment. stopped. DISTRACTOR ANAL YSIS: A. Incorrect, containment equipment hatch can be opened but the containment prurge being in service would not be a method to reduce the pressure. B. CORRECT, In accordance with O-GO-9 Precaution S, two of the four methods to prevent the potential spread of airborne radiation due to prevent excessive air flow through the transfer tube due to a differential pressure between the containment and the aux building is for the Equipment Hatch to be open and for the containment purge being stopped. C. Incorrect, containment hatch being closed is not one of the method to identified to prevent the potential spread of airborne radiation. D. Incorrect, containment hatch being closed is not one of the method to identified to prevent the potential spread of airborne radiation. Monday, December 15, 2008 9:38:45 AM 13 Question No. 97 Tier 3 KIA G 2.3.13 QUESTIONS REPORT for additional Questions Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. Importance Rating: 3.4 /3.8 Technical Reference: 0-GO-9, Refueling Operations, Rev 34 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 C259 B.2 Question Source: Describe the basis of each precaution, limitation and prerequisite in this procedure Bank # ----Modified Bank # ----New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.12 /43.4 / 45.9 / 45.10 ) 10CFR55.43.b ( 7 ) Comments: New question for Sequoyah 2009 exam Monday, December 15, 2008 9:38:45 AM 14 O-GO-9 SQN REFUELING PROCEDURE Rev: 34 1&2 Page 9 of 69 3.1 PRECAUTIONS (Continued) S. Due to relative air pressure differences between the Containment and Auxiliary Buildings which causes excessive air flow through the transfer tube, if the Transfer Tube Wafer Valve 1,2-78-610 is open, at least one of four conditions should exist to minimize the potential spread of airborne contamination: Reference TS 3.9.4. 1. The Containment equipment hatch stays open. This is not permitted during movement of recently irradiated fuel within containment (less than 100 hours since shutdown). 2. A Containment airlock is breached open (both doors) (not permitted during fuel movement within the containment unless one train of ABGTS is operable and one door of each breach airlock is capable of closure). 3. The lower reactor cavity has water filled to a level above the transfer tube elevation. 4. Containment ventilation purge supply/exhaust are stopped and measured differential pressure is approximately 0.25 psid or less between Computer Points P1001A to P1002A. T. The use of the 1 ton Jib Crane in Unit One Containment is limited to Modes 6 and defueled conditions. The 1 ton Jib Crane will not be used during fuel handling operations or when there is the potential for interference with other cranes operation in the vicinity. 3.2 LIMITATIONS A. Do not allow reactor vessel head to come into contact with the refueling water. B. The fuel shall not be moved in the core unless the water visibility is adequate to allow the operator to see the top nozzle of the seated fuel assembly for core unload, and lower core plate pin holes for core reload. C. Pressurizer manway must be open with airflow unobstructed whenever reactor head is in place and S/G U-tubes drained or pressurizer level off scale low. This ensures that adequate RCS vent exists to allow gravity fill from the RWST on a SBO event without natural circulation capability. This requirement does not apply when closing RCS in preparation for RCS vacuum refill (0-GO-13 section 5.3.4) REFUELING OPERA nONS ) 3/4.9.4 CONTAINMENT BUILDING PENETRATION? LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status: a. The equipment door closed and held inpJace by a minimum of four bolts, b. A minimum of one door in each airlock is closed, and both doors of both containment personnel airlocks may be open if: 1. One personnel airlock door in each airlock is capable of closure, and 2. One train of the Auxiliary Building Gas Treatment System is OPERABLE in accordance with Technical Specification 3.9.12, and c. Each penetration* providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, manual valve, or equivalent, or 2. Be capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve. APPLlCABIL TY: 3.9.4.a. Containment Building Equipment Door -During movement of recently irradiated fuel within the containment 3.9.4.b. and c. Containment Building Airlock Doors and Penetrations -During movement of irradiated fuel within the containment ACTION: n 1. With the requirements of the above specification not satisfied for the containment building equipment door, immediately suspend all operations involving movement of recently irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable. 2. With the requirements of the above specification not satisfied for containment airlock doors or penetrations, immediately suspend all operations involving movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve once per 7 days during movement of irradiated fuel in the containment building by: * a. Verifying the penetrations are in their required condition, or b. Testing the Containment Ventilation isolation valves per the applicable portions of Specification 4.6.3.2. Penetration flow path(s) providing direct access from the containment atmosphere that transverse and terminate in the Auxiliary Building Secondary Containment Enclosure may be unisolated under administrative controls. SEQUOYAH -UNIT 1 3/49-4 October 28, 2003 Amendment No. 12, 209, 249, 260, 288 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 97. G 2.3.13 097 Given the following: The Unit 1 Elevation 690' pipe chase is locked in accordanceAith its normal radiological posting. Conditions require Operations to make an emergency e pipe chase. Which ONE of the following identifies how the room is the "key control" for unlocking the pipe chase? A. Very High Radiation Area; Both Rad Ops and the Shift Manager have Jio ntro I of keys to the lock. B. Very High Radiation Area; ONLY Rad Ops has a key, the Shift ¥Bnager would contact Rad Ops to open the lock. C!" Locked High Radiation Area; Both Rad Ops and the Shift have control of keys to the lock. D. LockedHigh Radiation ONLY Rad Ops has a kj!y, the Shift Manager would contact Rad Ops to open the lock. Page 62 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DISTRACTOR ANAL YSIS: Page 63 A. Incorrect, The 690 pipe chase on both units is not a Very High Radiation Area (VHRA) but a key being located in the MCR under the administrative control of the Shift Manager is correct. Plausible because VHRA do exist and the Shift Manager does have access to a key located in the MCR. B. Incorrect, The 690 pipe chase on both units is not a Very High Radiation Area (VHRA) and Rad Ops does not have not only key. Plausible because if the area had been a VHRA then only Rad Ops would have a key. C. CORRECT, The 690 pipe chase on both units is a Locked High Radiation Area (LHRA). RCI-29 identifies the area as such and designates 2 key locations. One in the Rad Ops Lab and the other in the MCR under the administrative control of the Shift Manager. D. Incorrect, The 690 pipe chase on both units is a Locked High Radiation Area (LHRA). Rad Ops is not the only location of a key. Another key is located in the MCR under the administrative control of the Shift Manager. Plausible because if the area is a locked high radiation area and for all other LHRA only Rad Ops has a key. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 97 Tier 3 KIA G2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. Importance Rating: 3.4 /3.8 Technical Reference: RCI-29, Control of Radiation Protection Keys, Rev 7 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 C259 B.1.g Question Source: Identify the requirements for entering and working in the following areas: g. Locked High Radiation Area Bank# ___ _ Modified Bank # ----New X ---Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.12 / 43.4 / 45.9 / 45.10 ) 10CFR55.43.b ( 4 ) Comments: New question for Sequoyah 2009 exam Page 64 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT RADIOLOGICAL CONTROL INSTRUCTION RCI-29 CONTROL OF RADIATION PROTECTION KEYS Revision 7 QUALITY RELATED PREPARED BY: Terry F. Johnston RESPONSIBLE ORGANIZATION: Radiation Protection APPROVED BY: James S. McCamy EFFECTIVE DATE: 09/12/2007 VALIDATION DATE: NIA LEVEL OF USE: INFORMATION ONLY REVISION DESCRIPTION: This revision is generated to provide that Operations shall maintain LHRA keys for the Unit 1 and Unit 2 EI. 690' Pipe Chases in a Main Control Room key box under the administrative control of the Shift Manager. ) ) ) RCI-29 SQN Control of Radiation Protection Keys Revision 7 Page 3 of 6 6.0 REQUIREMENTS Note Use of the term 'keys' in the text of this Instruction is understood to refer to LHRA and VHRA keys interchangeably. unless specifically noted otherwise. 6.1 General A. Rad Ops will maintain positive control of LHRA and VHRA keys. B. Rad Ops shall initiate a Work Order (WO), as necessary, for the installation or removal of a LHRA tumbler lock. C. Rad Ops shall install or remove LHRA or VHRA security locks. 6.2 Key Controls [C.2] [C.3] [C.4] A. Keys controlled by this Instruction shall be maintained in locked key boxes in the EI. 690' Rad Ops Lab in the following configuration: 1. There is a designated LHRA key box that contains keys for tumbler locks and security locks that are in current use for active LHRAs only. This key box is kept locked at all times. It is opened each shift for the performance of a key inventory and is opened when issuing and returning LHRA keys. 2. There is a designated VHRA key box that contains keys for security locks that are in current use for active VHRAs only. This key box is kept locked at all times. Additionally, this key box has a wire tamper-proof security seal installed. It is only opened to issue or return VHRA keys. B. The Operations section shall maintain one LHRA key each for the Unit 1 and Unit 2 EI. 690' Pipe Chases. 1. These LHRA keys shall be maintained in the MCR in a locked break-glass-to-access key box, under the administrative control of the Shift Manager. 2. These LHRA keys are for emergency use only and Operations shall immediately notify the Rad Ops Lab if these keys are used. 3. Rad Ops shall maintain the key to the Ops LHRA key box in the Rad Ops Lab LHRA key box. 4. If during the course of normal duties the Operations LHRA key box is found to be open/unlocked, or the keys are not present in the box, Rad Ops shall be notified immediately. C. The on-duty Rad Ops Shift Supervisor, or designee, is responsible for maintaining control of the keys to open the respective LHRA and VHRA key boxes, and access to the key boxes themselves and the keys stored in the key boxes, for their respective shift. D. At the start of each shift, the Rad Ops Shift Supervisor, or designee, will verify that the keys to the LHRA and VHRA key boxes are present, that the LHRA key box is inventoried and that all LHRA keys are accounted for, to include the key to the Ops LHRA key box, and that the VHRA key box is locked and the seal is intact. Performance of this verification and inventory of the LHRANHRA key boxes shall be noted in the Radiation Operations Log. PER #84532 E. If any key cannot be accounted for during shift inventory, or the seal on the VHRA key box is not intact, the Rad Ops Manager shall be notified, and immediate actions taken to locate and secure the missing key. ) RCI-29 SQN Control of Radiation Protection Keys Revision 7 Page 2 of 6 ------------------------------_ .. _------1.0 PURPOSE The purpose of this Instruction is to provide guidelines for controlling Radiation Protection keys. 2.0 SCOPE This Instruction establishes the requirements for maintaining positive control of the Radiation Protection keys utilized to access Locked High Radiation Areas and Very High Radiation Areas. 3.0 REFERENCES 1 OCFR19, Notices, Instructions, and Reports to Workers: Inspection and Investigations 10CFR20, Standards for Protection Against Radiation Reg Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants SPP-5.1, Radiological Controls RCDP-1, Conduct of Radiological Controls SON Technical Specifications, Unit One and Unit Two 4.0 DEFINITIONS/ABBREVIATIONS Absorbed Dose -The energy imparted by ionizing radiation per unit mass of irradiated material. Accessible Area -Any area that can reasonably be occupied by a major portion of the whole body of an individual (as defined in 10CFR20). ANSI-Qualified Personnel -Rad Protection personnel assigned to SON, permanent or temporary, who meet ANSI qualifications. LHRA and VHRA Keys -Keys used to lock/unlock LHRA tumbler locks and security locks and/or VHRA security locks. LHRA and VHRA Security Lock -A padlock in use to secure a LHRA (where a tumbler lock cannot be used) and a VHRA. LHRA Tumbler Set -Lock tumblers installed exclusively in doors leading directly into a LHRA. Locked High Radiation Area (LHRA) -Any area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 1.0 rem (1,000 mrem) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. Positive Control -Control of the keys utilized by Rad Ops to access a LHRA or VHRA, in accordance with the requirements of this Instruction. Very High Radiation Area (VHRA) -Any area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed dose in excess of 500 rads in one hour at one meter from a radiation source or from any surface that the radiation penetrates. 5.0 RESPONSIBILITIES Responsibilities are defined in Section 6.0. TVA 40385 [6-2003] Page 2 of 2 1 PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: RADIOLOGICAL PO STINGS AND SIGNS IV. LENGTH OF LESSON/COURSE: 1-2 hour(s) V. TRAINING OBJECTIVES: A. Tenninal Objective: OPL271C259 Revision 8 Page 3 of 18 Upon completion of this lesson and others presented, the student shall demonstrate an understanding of Radiological Postings and Signs by successfully completing a written examination as defined by program procedures. B. Enabling Objectives: 1. Define and identifY the requirements for entering/working in the following areas: a. Unrestricted Area b. Restricted Area c. Radiologically Controlled Area d. Radioactive Material/Radioactive Material Storage Areas e. Radiation Area f. High Radiation Area g. Locked High Radiation Area h. Very High Radiation Area i. Contamination Area j. High Contamination Area k. Airborne Radioactivity Area. 2. Identify the criteria for utilizing Hot Spot Labels/Tags and Radioactive Material Tags. The following list contains knowledge and ability statements (KlAs) from The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (PWR) NUREG-1122, Revision 2 that are applicable to the Initial Licensed Candidate training program. As such, questioning in these areas will be included on any testing in preparation of, or included in obtaining either RO or SRO NRC license. lOCFR55 IMPORTANCE KIA # KIA Statement Sect. Link(s) RO/SRO G 2.3.1 Knowledge of 10CFR20 and related facility radiation control 41.12/43.4. 2.6/3.0 requirements. 45.9/45.10 G2.3.2 Knowledge of facility ALARA program. 41.12/43.4 / 2.5/2.9 45.9/45.10 G 2.3.4 Knowledge of radiation exposure limits and contamination control, 43.4 / 45.10 2.5/3.1 ) including pennissible levels in excess of those authorized. IG 2.3.5 Knowledge of use and function of personnel monitoring equipment. 41.11 /45.9 2.3/2.5 G 2.3.7 Knowledge of the process for preparing a radiation work pennit. 41.10/45.12 2.0/3.3 G 2.3.10 Ability to perfonn procedures to reduce excessive levels of radiation 43.4 / 45.10 2.9/3.3 and guard against personnel exposure. \ LESSON BODY 3) A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote reciever monitored by radiation protection; or, 4) A self-reading dosimeter and be under the surveillance while in the area of an individual qualified and equiped with a radiation monitoning device that continuously displays radiation dose rates, or under the surveillance while in the area by means of closed circuit television and the means to communicate. F. Locked High Radiation Areas (LHRA) ) 1. Definition: An accessible area to individuals in which radiation levels could result in an individual receiving a dose equivalent in excess of 1,000 mrem in 1 hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. 2. Explanation: a. The Shift Manager (SM) shall be notified when a LHRA is established or removed. b. LHRA keys shall be maintained under the administrative control of the SM, Rad Ops Manager, or their designees. c. Each LHRA shall be posted with a conspicuous sign or signs bearing the standard radiation symbol and the words "Caution -Locked High Radiation Area," or "Danger -Locked High Radiation Area." d. Entry shall be established by the use of a RWP. In addition, each individual or group entering such an area shall possess: 1) A radiation monitoring devise which has an appropriate alarm setting capability, continuously integrates the radiation dose rate in the area and alarms when the device's dose alarm setpoint is reached; or, 2) A radiation monitoring device that continuously transmits dose rate and collective dose information to a remote recorder monitored by Radiation Protection personnel with the means to communicate with and control every individual in the area; or, 3) A self-reading dosimeter and be under closed circuit television surveillance by a qualified Radiation Protection Technician and equipped with a monitoring device that continuously displays radiation dose rates in the area; or, 4) A self-reading dosimeter and be under closed circuit television surveillance by a qualified Radiation Protection Technician with the means to communicate with and control every individual in the area. OPL271C259 Revision 8 Page 10 of 18 INSTRUCTOR NOTES Obj.l.g Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted ') 98. G 2.3.14098 . Given the following conditions: Both Units operating at 100% power. 0600 Due to extremely heavy rainfall, RSO/KEOC issues a Stage I flood warning. Which one of the following identifies the time the Stage 1 flood mode actions are required to be completed, and if Stage II actions are required, how the Tritiated Drain Collector Tank would be prepared to prevent a possible release of radioactivity? Time Tritiated Drain Collector Tank A. 1600 Pressurized to greater than 23 psig. 1600 Filled with water. C. 2300 Pressurized to greater than 23 psig .. D. 2300 Filled with water. DIS TRA CTOR ANAL YSIS: Page 65 After entering AOP-N.03, "Flooding" Stage I preparations must be implemented and completed within the following 1 0 hours and if Stage /I actions are required they must be completed within the following 17 hours. A. Incorrect, The time requirement is correct. However, Stage /I actions do require the Tritiated Drain Collector Tank to be filled with water to prevent the tank from floating away and becoming a radiation hazard. Plausible because the time requirement is correct and pressurizing the tank to greater than 23 psig is correct for other tanks during Stage /I preparations. B. CORRECT, The time requirement is correct and the requirement is to fill Tritiated Drain Collector Tank to be filled with water to prevent the tank from floating away and becoming a radiation hazard. C. Incorrect, The time requirement is not correct and pressurizing the tank to greater than 23 psig is not correct. Plausible because the time identified is 17 hours which is the time required to complete Phase /I actions after Phase /I is initiate and pressurizing the tank to greater than 23 psig is correct for other tanks during Stage /I preparations. D. Incorrect, The time requirement is not correct but filling the tank with water is correct. Plausible because the time identified is 17 hours which is the time required to complete Phase /I actions after Phase /I is initiate and filling the tank with water is correct. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted ') Question No. 98 Tier 3 KIA G 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. Importance Rating: 3.41 3.8 Technical Reference: AOP-N.03, Flooding, Rev 28 TRM 4.7.6, Flood Protection Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-N.03 B.4,5, & 8b Question Source: 4. Upon entry into AOP-N.03, diagnose the applicable condition and transition to the appropriate section for response. 5. Describe the bases for all limits, notes cautions and steps of AOPN.03. 8.b Given a set of initial plant conditions use to correctly (b) Identify required actions. Bank # _-:--__ Modified Bank # _X, __ _ New __ _ Question History: Sequoyah Question AOP-N.03-B.4 002 modified Question Cognitive Level: Memory or fundamental knowledge __ _ Comprehension or Analysis _X __ 10 CFR Part 55 Content: ( 41.12/43.4/45.10 ) 10CFR55.43.b (2,4) Requires the candidate to know the facility Technical Requirements, implementation of AOP sections and be knowledgeable of provisions to prevent radiation hazards that may arise during normal, abnormal and emergency conditions. Comments: Sequoyah Question AOP-N.03-B.4 002 modified Page 66 PLANT SYSTEMS TR 3/4.7.6 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION TR 3.7.6 The flood protection plan shall be ready for implementation to maintain the plant in a safe condition. APPLICABILITY: When one or more of the following conditions exist: a. early warning of major flood-producing rainfall conditions in the east Tennessee watershed, b. an early warning that a criticalcornbination of flood and/or higher than normal Summer pool levels plus possible darn failures or other darn safety emergencies mayor have developed, c. or warnings that flood elevation is predicted to exceed plant grade (Stages I and II) . ACTION: a. With a Stage I flood warning issued initiate and complete within 10 hours the Stage I flood protection procedure which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and less than or equal to 350°F within the following 4 hours. If within 10 hours following the issuance of a Stage I flood warning communications between the TVA Water I Management River Scheduling (RS) and the Sequoyah Nuclear Plant cannot be verified, initiate and complete the Stage II flood protection procedure within the following 17 hours. With a Stage II flood warning issued initiate the Stage II flood protection plan in time to ensure completion before the predicted flooding of the site. Initiation shall be no later than 17 hours prior to the predicted arrival time of the initial critical flood level (703 ft msl winter and summer). Completion of any actions are not required if warnings are retracted by RS. b. After an early warning is issued, verify communications between TVA RS and the Sequoyah Nuclear Plant within 5 hours or initiate and complete the Stage I flood protection plan within the following 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours.' Completion of any actions are not required if corrmunications are verified. SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS 3/4 7-4 January 20, 2006 Revision Nos. 5, 14 PLANT SYSTEMS TR 3/4.7.6 FLOOD PROTECTION SURVEILLANCE REQUIREMENTS TR 4.7.6.1 This requirement deleted. TR 4.7.6.2 Communications between Sequoyah Nuclear Plant and TVA RS shall be maintained on 3 hour intervals. If not maintained, within 5 hours from previous contact initiate Stage I Flood Plan and continue with Stages I and II until contact is re-established and RS confirms that the flood plans are not required. SEQUOYAH -UNITS 1 AND 2 TECHNICAL REQUIREMENTS 3/4 7-5 January 20, 2006 Revision Nos. 5, 14 ) SQN FLOODING APPENDIXC CVCS AND WDS TANK FILLING INSTRUCTIONS NOTE 1 . This appendix provides instructions for filling the partially filled and possibly radioactive tanks located below Maximum Probable Flood level of 723.1' elev. Performance of this procedure minimizes NOTE 2 the possibility of the tanks collapsing or breaking loose and possible release of radioactivity. Steps 1 through 22 may be performed out of sequence. 1. PLACE Reactor Building Floor and Equipment Drain Sumps IN SERVICE to Tritiated Drain Collector Tank USING 0-SO-77-10. 2. FILL pressurizer relief tank USING 1,2-S0-68-5. 3. NOTIFY Maintenance to connect 2 inch passive failure discharge connection on Auxiliary Building Floor and Equipment Drain Sump Pumps discharge USING O-FP-MXX-OOO..:OOB.O. 4. OPEN the following valves to fill Floor Drain Collector Tank: * 0-77-915 * NOTE The time required for filling the Tritiated Drain Collector Tank is approximately 4 hours. 5. OPEN valve 0-77-914 and FILL Tritiated Drain Collector Tank. 6. ALIGN Laundry and Hot Shower Tank Pump to Waste Condensate Tanks A, B,and C USING 0-SO-77-5. 7. Al,.lGN Laundry and Hot Shower Tanks A and B and Chemical Drain Tank. Page 143 of 215 AOP-N.03 Rev. 28 Page 1 of 4 AOP-N.03-BA 002 QUESTIONS REPORT for BANK SQN Questions Which ONE of the following is the correct time allowed in AOP-N.03, Flooding, to complete any Stage I procedure section? A'!I 10 hours. B. 17 hours. C. 27 hours. D. 40 hours. Justification: A. Correct -Correct as stated in NOTE prior to step 1 of AOP-N.03 B. Incorrect -as it defines the 14 hours required to complete Stage 2 plus 3 hours margin. C. Incorrect- as it defines the total time to complete Stage 1 + 2 plus 3 hours margin. D. Incorrect- not defined by AOP-N.03. Monday, November 24, 2008 7:32:55 AM 1 OPL271 AOP-N.03 Revision 1 Page 3 of 40 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-N.03, FLOODING IV. LENGTH OF LESSON/COURSE: 2 hours V. TRAINING OBJECTIVES: O. 1. 2. 3. 4. S. 6. A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-N.03, FLOODING. B. Enabling Objectives Objectives Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with a plant Flood that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A State the purpose/goal of this AOP-N.03. Describe AOP-N.03 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-N.03 entry conditions. b. Describe the ARP requirements associated with AOP-N.03 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-N.03 entry conditions. d. Describe the plant parameters that may indicate a plant Flood. Describe the initial operator response to stabilize the plant upon entry into AOP-N.03. Upon entry into AOP-N.03, diagnose the applicable condition and transition to the appropriate procedural section for response. Summarize the mitigating strategy for the failure that initiated entry into AOP-N.03. Describe the bases for all limits, notes, cautions, and steps of AOP-N.03. I OPL271 AOP-N.03 Revision 1 Page 4 of 40 7. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. 8. Given a set of initial plant conditions use AOP-N.03 to correctly: a. Recognize entry conditions. b. Identify required actions. c. Respond to Contingencies. d. Observe and Interpret Cautions and Notes. 9. Describe the Tech Spec and TRM actions applicable during the performance of AOP-N.03. 10. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition. OBJECTIVES TO BE COVERED IN THESE SEQUOY AH OPERATOR TRAINING PROGRAMS OBJECTIVE I NONLICENSED LICENSE TRAINING NO. RO SRO REQUAUSPECIAL OPERATORS O. X X 1. X X 2. X X 3. X X 4. X X 5. X X 6. X X 7. X X 8. X X 9. X X 10. X X Selected objectives to be covered in: PowerPoint presentation to be used: Sequoyah Operator Training Manager / Date Sequoyah Operations Manager / Date Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted 99. G 2.4.27099 Given the following: Both units in service at 100% power. -The plant is operating with minimum operations staffing. -A fire develops at the ERCW pumping station. -AOP-N.01, "Plant Fires", is implemented. -The Shift Manager declares an ALERT emergency and initiates Assembly and Accountability. Fire Ops actions to extinquish the fire continue. Which ONE of the following identifies ... (1) the direction the AUOs are to be given and (2) if the Shift Manager later decides to implement AOP-N.08, "Appendix R Fire Safe Shutdown", must both units enter the AOP or can the decision to enter the AOP be made separately for each unit? A'! (1) Direct all AUOs to report to the Main Control Room. (2) Decision to enter AOP-N.08 can be made separately on each unit. B. (1) Direct all AUOs to report to the Main Control Room. (2) Decision to enter AOP-N.08 must be made on both units at the same time. C. (1) Direct 2 AUOs to report to the OSC and the rest to report to the Main Control Room. (2) Decision to enter AOP-N.08 can be made separately on each unit. D. (1) Direct 2 AUOs to report to the OSC and the rest to report to the Main Control Room. Page 67 (2) Decision to enter AOP-N.08 must be made on both units at the same time. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRA CTOR ANAL YSIS: Page 68 A. CORRECT, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), all AVOs will report to the main control room because in accordance with AOP-N.01, the requirement to assemble AVOs to assign and brief the actions of AOP-N.08 take priority over staffing the OSC and Level II Fire Brigade. Separate entry into AOP-N.08 is allowed in accordance with AOP-N.01 because the criteria requiring both units to enter the AOP at the same time are not met. The fire is not in one of the four limiting areas that require simultaneous entry by both units .. B. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), all AVOs will report to the main control room because in accordance with AOP-N.01, the requirement to assemble AVOs to assign and brief the actions of AOP-N.08 take priority over staffing the OSC and Level II Fire Brigade. However the AOP-N.08 entry does not met the criteria requiring both units to enter at the same time. Plausible because the directions to the AVOs is correct and a fire in a different location could cause the units to enter the AOP at the same time. C. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), 2 AVOs would not be sent to the OSC, all would report to the main control room in accordance with AOP-N.01 because the requirement to assemble A VOs to assign and brief the actions of A OP-N. 08 takes priority over staffing the OSC and Level II Fire Brigade. The AOP-N.08 can be made separately because entry does not met the criteria requiring both units to enter at the same time. Plausible because the directions to send 2 AVOs to the OSC is the normal protocol during an emergency event and the units entry into the AOP being allowed at separate times in correct. D. Incorrect, with the fire a the ERCW pumping station still burning and the staffing at minimum (8 AVOs), 2 AVOs would not be sent to the OSC, all would report to the main control room in accordance with AOP-N.01 because the requirement to assemble AVOs to assign and brief the actions of AOP-N.08 takes priority over staffing the OSC and Level II Fire Brigade. Also, the AOP-N.08 entry does not met the criteria requiring both units to enter at the same time. Plausible because the directions to send 2 AVOs to the OSC is the normal protocol during an emergency event and a fire in a different location could cause the units to enter the AOP at the same time. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 99 Tier 3 KIA G 2.4.27 Knowledge of "fire in the plant" procedures. Importance Rating: 3.4 / 3.9 Technical Reference: AOP-N.01, Plant Fires, Rev 26 AOP-N.08, Appendix R fire Safe Shutdown, Rev 5 OPDP-1, Conduct of Operations, Rev 0010 Lesson Plan OPL271 REP, Rev 1 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271AOP-N.01 8.5 Question Source: Describe the bases for all limits, notes, cautions, and steps of AOP-N.01. Bank# ___ _ Modified Bank # ---:--:--__ New _X __ Question History: New question for Sequoyah 2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for Sequoyah 2009 exam Page 69 ) SQN APPENDIX R FIRE SAFE SHUTDOWN AOP-N.08 Rev. 5 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 GENERIC ACTIONS NOTE AUOs assigned to Level II fire brigade, OSC,and monitoring fire pumps should be recalled for safe shutdown actions. 4. NOTIFY all available AUOs to perform the following: a. REPORT to MCR area immediately. b. OBTAIN radio and SCBA from MCR. 5. PLACE all RCP handswitches in STOP/PULL TO LOCK. [M-5] 6. ENSURE MSIV and MSIV bypass valve handswitches in CLOSE position. [M-4] 7. MONITOR fire* location based upon reports from Fire Brigade Leader or Incident Commander. Page 4 of 1380 SQN PLANT FIRES AOP-N.01 Rev. 26 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE 1: To ensure manual actions in AOP-N.08 can be completed within the required time, AUOs must be ready to be dispatched when AOP-N.08 is entered. The requirement to assemble AUOs in this step takes priority over staffing the OSC and Level II Fire Brigade. NOTE 2: Step 17 (two pages) should be handed off to CRO, if applicable. 17. IF fire is still burning AND fire is in any of the following locations:

Aux Building * Additional

Equipment Bldg * Reactor Bldg (Containment or Annulus) * ERCW Pumping Station THEN PERFORM the fOllowing: a. NOTIFY at least eight (8) AUOs to report to Control Room. b. IF fire is in Aux Building, THEN ENSURE the following CCS pumps RUNNING: [O-M-27B]

CCS pump 1 A-A * CCS pump 1 B-B * CCS pump 2A-A * CCS pump 2B-B (step continued

on next page) Page 12 of 84 SQN PLANT FIRES AOP-N.01 Rev. 26 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE 1 NOTE 2 NOTE 3 NOTE 4 The decision to implement AOP-N.08 or AOP-C.04 is an SRO judgment based upon the severity of fire and its potential effect on plant equipment. BOTH UNITS must enter AOP-N.08 at same time if ALL of the following conditions are met: * fire is in one of four limiting Aux Bldg areas (el. 690 General Area, el. 714 General Area, 6.9KV Shutdown Board Rm A, or 6.9KV Shutdown Bd Rm B) * both units in Mode 1-3 * criteria for AOP-N.08 entry in Step 19 is met on either unit In fire areas other than the four limiting Aux Bldg areas, the decision to enter AOP-N.08 is made separately on each unit. AOP-N.08 is NOT applicable for a total loss of £!l ERCW capability. AOP-M.01 provides required actions. 19. MONITOR magnitude of fire and potential to impact control of unit: a. CONSULT Incident Commander or Fire Brigade Leader. b. MONITOR MCR indications and controls for equipment failures or spurious operation. (step continued on next page) Page 16 of 84 X. LESSON BODY: 2. Minimum TSC staffing * SED * RP Manager * Operations Manager or Operations Communicator

Technical

Assessment Manager, Technical Assessment Team Leader, or Reactor Engineer * Mechanical Engineer * Electrical Engineer 3. Transfer of SED role from the SM to the TSC * Use EPIP-6, Appendix B G. EPI P-7, Activation and Operation of the Operations Support Center (OSC) 1. This procedure provides the guidance for activating and operating the OSC. The OSC is required to be activated at the ALERT classification and above. The SED directs mitigating actions and determines OSC priorities. The OSC Manager oversees OSC activities. 2. Minimum OSC staffing * OSC Manager * Mechanical Maintenance Group (1) * Electrical Maintenance Group (1) * Instrument Maintenance Group (1) 3. AUO teams responding to procedure driven activities (EOPs, EAs, AOPs, etc.) are under the direction of the SM but tracked by the OSC. 4. RP and Operations should normally have personnel on each response team. H. EPIP-8, Personnel Accountability and Evacuation 1. This procedure provides the method for accounting for all personnel and visitors in the protected area within 30 min. OPL271 REP Revision 1 Page 23 of 32 INSTRUCTOR NOTES Refer to EPIP-6, Appendix B for discussion. SM -Ensure you keep OSC aware of AUO usage for tracking purposes. Assembly should not be initiated if assembly would present a danger to employees. x. LESSON BODY: 2. Operations Personnel a. The protocol requires two AUOs to go to the OSC and the remaining AUOs go to the Main Control Room. The SM may send the AUOs to a waiting area within the protected ventilation area of the Control Building (such as the SM Office) if noise or congestion become a problem in the MCR. b. When the OSC Ops Advisor arrives in the OSC, he and the SM will collectively manage the resources to have the fastest response to problems yet still OPL271 REP Revision 1 Page 24 of 32 INSTRUCTOR NOTES provide a check of radiological conditions prior to I Obj S.b dispatching an AUO. Dispatching AUOs into plant areas needs input by RP as to appropriate protective equipment. This is best done by the OSC Ops Advisor once he/she arrives. The OSC Ops Advisor is expected to relieve the SM of some of the administrative burden of tracking field personnel so he/she can focus on the higher priority of operational safety. c. The Shutdown Board Rooms and Control Building are spaces with protected ventilation similar to the MCR. Until advised by RP that these areas are becoming unsafe, they should be considered part of the MCR for the purpose of dispatching AUOs. This means that AUOs may be dispatched into these areas for tasks without being considered a "team" (much like sending an AUO behind the panels in the MCR to perform a task). As always, communications must be maintained at all times with AUOs while out on assignments and the TSC should be informed of AUO activities. d. This protocol allows dispatching AUOs by telephone briefing and/or radio briefing as well as transfer of oversight of individual AUOs between the OSC SRO and SM as needed. This provides the best utilization of personnel while still protecting the AUOs from radiological hazards. e. AUOs assigned to the OSC will report directly to the OSC when required. f. The SM is responsible to account for MCR personnel. NPG Standard Conduct of Operations OPDP-1 Department Rev. 0010 Procedure Page 47 of 62 Attachment 1 (Page 2 of 2) Shift Staffing WBN WBN Mode 1-4 Mode 5& 6 SON BFN Shift Manager (SRO) 1 1 1 Unit Supervisor (SRO) 1 1 3 3 Unit Operator (RO) 2 1 4 6 Non-licensed (AUO) 5 3 8 8 STA 1 1 1 SON The SM, a US or the WCC may be the STA and one US will be the Incident Commander. The STA need not be licensed. Two active-licensed SROs are required for Unit Supervisor positions and a third active licensed SRO is required as Shift Manager. BFN One of the US can be the STA. The Incident Commander position may be filled by an additional qualified person. Notification of Absences * Operations personnel (except Fire Operations) unable to report for shift duty shall, althe earliest possible time and no later than 2 hours before the scheduled time, inform the SM/US of the situation. The SM/US shall make necessary arrangements for obtaining a replacement.

Fire Operations

personnel unable to report for shift duty shall, at the earliest possible time and no later than 2 hours before the scheduled time, inform the Fire Operations Foreman of the situation. The Fire Operations Foreman shall make necessary arrangements for obtaining a replacement. OPL271 AOP-N.01 Revision 1 Page 3 of 15 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-N.01, PLANT FIRES IV. LENGTH OF LESSON/COURSE: 1 hour V. TRAINING OBJECTIVES: O. 1. 2. 3. 4. 5. 6. A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-N.01, PLANT FIRES. B. Enabling Objectives Objectives Demonstrate an understanding of NUREG 1122 knowledge's and abilities associated with Plant Fires that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A. State the purpose/goal of this AOP-N.01. Describe the AOP-N.01 entry conditions. a. Describe the setpoints, interlocks, and automatic actions associated with AOP-N.01 entry conditions. b. Describe the ARP requirements associated with AOP-N.01 entry conditions. c. Interpret, prioritize, and verify associated alarms are consistent with AOP-N.01 entry conditions. d. Describe the plant parameters that may indicate a Plant Fire. Describe the initial operator response to stabilize the plant upon entry into AOP-N.01. Summarize the mitigating strategy for the failure that initiated entry into AOP-N.01. Describe the bases for all limits, notes, cautions, and steps of AOP-N.01. Describe the conditions and reason for transitions within this procedure and transitions to other procedures. I I OPL271 AOP-N.01 Revision 1 Page 4 of 15 7. Given a set of initial plant conditions use AOP-N.01 to correctly: a. Recognize entry conditions. b. Identify required actions. c. Respond to Contingencies. d. Observe and Interpret Cautions and Notes. 8. Describe the Tech Spec and TRM actions applicable during the performance of AOP-N.01. 9. Apply GFE and system response concepts to the abnormal condition -prior to, during and after the abnormal condition. -OBJECTIVES TO BE COVERED IN THESE SEQUOYAH OPERATOR TRAINING PROGRAMS OBJECTIVE I NON LICENSED LICENSE TRAINING NO. RO SRO REQUAUSPECIAL OPERATORS O. X X 1. X X 2. X X 3. X X 4. X X 5. X X 6. X X 7. X X 8. X X 9. X X 10. X X NOTE: The following approval is required for License Requalification and special training only: Selected objectives to be covered in: PowerPoint presentation to be used: Sequoyah Operator Training Manager / Date Sequoyah Operations Manager / Date REP-B.1.D 001 Given the following: QUESTIONS REPORT for BANK SQN Questions -Unit 1 has experienced a LOCA and loss of 1 A-A 6.9 kV SO Bd. -You are the OATC. -The SM just announced an ALERT and sounded assembly. How will the AUOs respond to the siren going off? The two AUOs assigned to the OSC will report to the OSC, the remaining AUOs will report to the SM. 8. The two AUOs assigned to the SM will report to the MCR, the remaining AUOs will report to the OSC. C. ALL AUOs will immediately report to the SM until the OSC Operations Advisor SRO is ready to assume control of the AUOs. O. ALL AUOs will immediately report to the OSC Operations Advisor SRO until the SM is ready to assume control of the AUOs. Justification: A. Correct. Refer to EPIP-7, Appendix D for stage 1, "Declaration of the Emergency. " B. Incorrect. Two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. C. Incorrect. Two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. After OSC Operations Advisor arrives, the SM may assign additional AVOs not in the field to the OSC Operations Advisor. D. Incorrect. Only two AVOs are assigned to support the MSS in the OSC until staffed; all other AVOs are under SM control. Monday, November 24, 2008 7:46:32 AM 1 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted , 100. G 2.4.28 100 Given the following: Both units trip from power operation due a loss of off site power and sabotage is suspected. Nuclear Security reports the following: -Information has been received that a specific credible insider threat exists associated with the loss of offsite power. and -Concerns exist for the health and safety of any oncoming emergency responders. The Shift Manager determines the need to staff the emergency centers, makes REP declaration and activates Assembly and Accountability. The Two-person line of sight rule has been implemented. The Operating crew needs to send personnel to the DG building due to an alarm occurring on DG 1A-A. Which ONE of the following describes the selection the Shift Manager will make when activating the Emergency Paging System (EPS) and the required measures to be taken when sending personnel to the DG building. A. STAGING AREA will be selected on the EPS; Both individuals sent must be qualified for the task to be performed. STAGING AREA will be selected on the EPS; Only one of the individuals sent must be qualified for the task to be performed. C. EMERGENCY will be selected on the EPS; Both individuals sent must be qualified for the task to be performed. D. EMERGENCY will be selected on the EPS; Only one of the individuals sent must be qualified for the task to be performed. Page 70 Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted DIS TRACTOR ANAL YSIS: A. Incorrect, In accordance with the Emergency Plan, if concerns exist for the health and safety of oncoming responders, the "Staging Area" choice is to be selected when initiating the staffing of the Emergency Centers but when the two-man rule is implemented, the individuals do not have to have the same qualifications. Plausible because choosing the "Staging Area" when initiating the staffing of the Emergency Centers is correct and typically when personnel are sent to perform a task both would be qualified. Page 71 B. CORRECT, In accordance with the Emergency Plan, if concerns exist for the health and safety of oncoming responders, the "Staging Area" choice is to be selected when initiating the staffing of the Emergency Centers and the individual do not have to posses the same qualification when the two-man rule is implemented. C. Incorrect, the "Staging Area" choice is to be selected when initiating the staffing of the Emergency Centers (not the the "Emergency), when concerns exist for the health and safety of oncoming responders, and the individual do not have to posses the same qualification when the two-man rule is implemented. Plausible because the the "Emergency" choice would be correct with different conditions when inititiating the staffing of the Emergency Centers and typically when personnel are sent to perform a task both would be qualified. D. Incorrect, the "Staging Area" choice is to be selected when initiating the staffing of the Emergency Centers (not the the "Emergency), when concerns exist for the health and safety of oncoming responders, and the individual do not have to posses the same qualification when the two-man rule is implemented. Plausible because the the "Emergency" choice would be correct with different conditions when inititiating the staffing of the Emergency Centers and the requirement for only one of the individuals to be qualified is correct. Proposed 2/6/2009 Sequoyah NRC SRO Written Exam as submitted Question No. 100 Tier 3 KIA G 2.4.28 Knowledge of procedures relating to a security event (non-safeguards information). Importance Rating: 3.2/4.1 Technical Reference: SPP-1.3, Access Authorization and Nuclear Security, Rev 0011 EPIP-8, Personnel Accountability and Evacuation, Rev 17 EPIP-3, Alert, Rev 30 Proposed references to be provided to applicants during examination: None Learning Objective: OPL271 REP B. 5.b. Question Source: State the duties of the Site Emergency Director (SED) . b. State the conditions under which the SED may order relocation from one assembly point to another. Bank# ___ _ Modified Bank # ___ _ New X '----Question History: New question for SON 1/2009 exam Question Cognitive Level: Memory or fundamental knowledge _X __ Comprehension or Analysis __ _ 10 CFR Part 55 Content: ( 41.10/43.5/45.13 ) 10CFR55.43.b ( 5 ) Comments: New question for SON 1/2009 exam Page 72 NPG Standard Access Authorization and Nuclear Security SPP-1.3 Programs and Rev. 0011 Processes Page 32 of 51 3.9.9 Maintenance (continued) B. Repair/replace failed security devices and components in the minimum time necessary. 3.9.10 Unauthorized Materials Firearms, explosives, incendiary devices, alcoholic beverages and illegal drugs are prohibited on nuclear plant sites. 3.9.11 Credible Insider Threat When Nuclear Security determines that a specific credible insider threat exists, Nuclear Security shall request Shift Manager/Site Emergency Director to implement a two-person (line of sight) rule for personnel in vital areas, unless unusual circumstances exists where emergency actions by a single individual are required to ensure nuclear safety. (See also Appendix A, paragraph 4.4 and EPIP-8, Personnel Accountability and Evacuation). 3.9.12 Hostage/Duress Situation Any employee who is coerced, influenced or pressured in any way to initiate or be party to an act that presents an unsafe situation at any of the TVA NPG Sites will immediately contact Nuclear Security or Corporate Nuclear Security, as appropriate, and provide as much information as possible. Nuclear Security or Corporate Nuclear Security (CNS), as appropriate, will contact the appropriate agency to respond, (see also Appendix A). 3.9.13 CameraNideo Requirements Individuals are prohibited from taking pictures, videos, etc of security equipment, security posts, or other security areas/items without prior authorization by the SSM or designee. 3.10 Safeguards Events (SGE) 3.10.1 Reporting SGEs Individuals that discover an actual or suspected SGE are responsible for the immediate reporting of that event to Security. A. If the actual or suspected SGE occurs offsite, contact the Manager, CNS or SSM immediately, in person, or by telephone (whichever is faster). B. If the actual or suspected SGE occurs onsite, contact the SSM immediately, in person, or by telephone (whichever is faster). NOTE This immediate notification is necessary for the timely implementation of contingency plans and reporting requirements. C. Provide as much detail regarding the incident as possible and, if requested, complete statement describing the event in as much detail as possible (for example: who, what, when, where, and why or how, if known). Upon completion submit the statement to Security. NPG Standard Programs and Processes Access Authorization and Nuclear Security Appendix A (Page 2 of 2) SPP-1.3 Rev. 0011 Page 42 of 51 Guidelines for Initial Actions by Plant Personnel of Incidents Involving Suspected Tampering or Sabotage 4.2 Shift Manager (continued) B. Shift Manager evaluates appropriate AOI or AOP entry for Security Event Response. C. Shift Manager evaluates EPIP-1, Emergency Plan Classification Matrix. 4.3 Employee If any act of coercion, influence or pressure is committed with intent to initiate an act of tampering or sabotage, then Notify Nuclear Security. 4.4 Credible Insider Threat This action is triggered when a credible insider threat specific to a facility exists. Once triggered, implementation of the 2-man rule will be as expeditious as resources permit recognizing that additional personnel may need to be called to the particular site. The two persons do not have to possess similar skills or knowledge, but must remain in visual contact with each other unless personnel or plant safety would be adversely impacted. 5.0 SUBSEQUENT ACTIONS 5.1 Employee Retain any relevant information and provide to Nuclear Security to aid in investigation. 5.2 Security Supervisor If an event as described in Sections 4.1, 4.2, 4.3 through Section 4.4 above is reported, then A. Nuclear Security evaluate need to contact TVA Police Criminal Investigation Division for assistance, and B. Nuclear Security evaluate in accordance with this instruction and NSDP-1, "Safeguard Event Reporting Guidelines." ) ) SEQUOYAH PERSONNEL ACCOUNTABILITY AND EVACUATION EPIP-8 9. 10. 11. APPENDIXD Page 3 of 3 NUCLEAR SECURITY -ASSEMBLY AND ACCOUNTABILITY ACTIONS REPORT the results of accountability to the SM/SED within 30 minutes after the assembly and accountability sirens have sounded. Accountability is considered complete when all personnel have been accounted for or are known by nameifnot accounted for. Unaccounted Individuals IF ... Individuals remain unaccounted 45 minutes following the activation of the assembly and accountability sirens, THEN ... REQUESTpermission from the SM/SED to form search teams to locate the missing individual(s), AND NOTIFY RP and request they accompany all search teams. Implement the Two Person (Line of Sight) Rule and make a Public Address Announcement WHEN ... Assembly and Accountability have been completed, AND Nuclear Security has determined that the Two Person (Line of Sight) Rule is required. THEN ... REQUEST permission from the SM/SED to make the following Public Address Announcement: (Accountability area PA is accessible at x4800 from selected TSC and MCR phones) "Attention all personnel. A credible insider threat exists. Effective immediately, all personnel entering the Vital Areas must observe the 2-person rule. This rule requires that all persons in a vital area must remain in visual contact with another person unless personnel or plant safety would be adversely impacted. This does not require that the two persons possess similar skills or knowledge. I repeat. The 2-person rule is being. implemented immediately." (REPEAT) PAGE 19 OF 27 D D Initials Time REVISION 17 tSEQUOYAH ALERT EPIP-3 3.1 ALERT DECLARATION BY THE MAIN CONTROL ROOM (Continued) [3] ACTIVATE Emergency Paging System (EPS) as follows: [a] IF EPS has already been activated, THEN GO TO Step 4. [b] IF ongoing onsite Security events may present risk to the emergency responders, THEN CONSULT with Security to determine if site access is dangerous to the life and health of emergency responders. [c] IF ongoing events makes site access dangerous to the life and health of emergency responders THEN SELECT STAGING AREA button on the terminal INSTEAD of the EMERGENCY button. [d] ACTIVATE EPS using touch screen terminal. IF EPS fails to activate, THEN continue with Step 4. [4] COMPLETE Appendix B (TVA Initial Notification for Alert). o o o o o NOTE: ODS should be notified within 5 minutes after declaration of the event. ) [5] NOTIFY ODS. Initial Time ODS: Ringdown Line or 5-751-1700 or 5-751-2495 or 9-785-1700 [a] IF EPS failed to activate fromSQN when attempted, THEN DIRECT ODS to activate SON EPS. [b] IF ODS is also unable to activate EPS, THEN continue with step [5] [b]. [c] READ completed Appendix B to ODS. [d] FAX completed Appendix B toODS. 5-751-8620 (Fax) [e] MONITOR for confirmation call from ODS that State/Local notifications complete: RECORD time State notified. o D o D Notification Time FORWARD COMPLETED PROCEDURE TO EMERGENCY PREPAREDNESS MANAGER PAGE 5 of 15 REVISION 30 \ 1 OPL271 REP Revision 1 Page 3 of 32 I. PROGRAM: OPERATOR TRAINING -LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: NP RADIOLOGICAL EMERGENCY PLAN AND SEQUOYAH EMERGENCY PLAN IMPLEMENTING PROCEDURES IV. LENGTH OF LESSON/COURSE: 8 hours (Hot License Class), 2 -4 hours (LOR) V. TRAINING OBJECTIVES: A. Terminal Objective: Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of the Radiological Emergency Plan (REP). B. Enabling Objectives: O. Demonstrate an understanding of NUREG 1122 Knowledge and Abilities associated with Radiological Emergency Plan that are rated;::: 2.5 during Initial License Training and;::: 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A. 1. Discuss the Radiological Emergency Plan a. Discuss the regulatory bases for the REP b. State the purpose of the REP. c. Define and state the purposes of a(n) NOUE, Alert, Site Area Emergency, and General Emergency d. State the purpose and major job functions of the Technical Support Center (TSC), the Operations Support Center (OSC), the Central Emergency Control Center (CECC) and give the location of each. e. Describe the role the state and federal agencies play during an event f. Describe the process of authorizing Emergency Radiological Exposures in accordance with EPIP-15. g. State the conditions under which onsite personnel would be administered potassium iodide (KI). h. Describe Chemistry and Radiation Protection tasks during emergency operations. i. Discuss the termination of a declared Radiological Emergency in accordance with EPIP-16. 2. Determine the required notifications based upon the event, including time requirements. 3. Classify emergency events using appropriate procedures.

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4. 5. 6. OPL271 REP Revision 1 Page 4 of 32 Determine protective action recommendations using appropriate procedures. State the duties and responsibilities of the Site Emergency Director (SED). a. State the duties and responsibilities the SED may not delegate b. State the conditions under which the SED may order relocation from one assembly point to another. Discuss medical emergency response per EPIP-10. }}

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