ML100351172

DAEC 2009 Initial Exam Proposed Written-SRO

ML100351172
Person / Time
Site:	Duane Arnold
Issue date:	09/21/2009
From:	Division of Reactor Safety III
To:
References
Download: ML100351172 (155)

Text

QF-1030-03 Rev. 7 WRITTEN/ORAL EXAMINATION KEY COVERSHEET Examination Number/Title: Series A, Rev. 0, 2009 NRC Reactor Operator Initial License Exam Training Program: Initial License Training Course/Lesson Plan Number(s): 50007 / Various Total Points Possible: 75 PASS CRITERIA: 80% Exam Time: 6 Hours EXAMINATION REVIEW AND APPROVAL:

Developed by: Date:

Instructional Review (Exam Qualified Instructor): Date:

Technical Review (SME): Date:

Approved by Training Supervisor: Date:

Written/Oral Examination key Attach answer key to this page.

Exam Development and Review Guidelines: Key should contain the following:

o QF-1030-26, Instructional and Technical Learning Objective Number Review Checklist for Examinations Test Item o TDAP 1816.2, TSD - Design Phase, o Question or Statement Section 5.4 o All possible answers o TDAP 1816.4, TSD - Implementation Phase, o Correct Answer Indicated Section 5.5. o Point Value o References (if applicable)

NOTE: NRC exams may require additional information. Refer to site specific procedures.

Indicate in the following table if any changes are made to the exam after approval:

PREPARER DATE

1. DESCRIPTION OF CHANGE REASON FOR CHANGE REVIEWER DATE Retention: Life of plant insurance policy + 10 yr.

Retain in: Training Records 2009 SRO NRC Master 8-10-09.doc

QF-1030-02 Rev. 4 WRITTEN/ORAL EXAMINATION COVERSHEET Trainee Name:

Employee Number: Site: DAEC Examination Number/Title: Series A, Rev. 0, 2009 NRC Senior Reactor Operator Initial License Exam Training Program: Initial License Training Course/Lesson Plan Number(s): 50007 / Various Total Points Possible: 25 PASS CRITERIA: 80% Grade: /25 =  %

Graded by: Date:

Co-graded by (if necessary): Date:

EXAMINATION RULES

1. References may not be used during this examination, unless otherwise stated.

2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.

3. Conversation with other trainees during the examination is prohibited.

4. Partial credit will not be considered, unless otherwise stated. Show all work and state all assumptions when partial credit may be given.

5. Rest room trips are limited and only one examinee at a time may leave.

6. For exams with time limits, you have 120 (2 Hours) minutes to complete the examination.

7. Feedback on this exam may be documented on QF-1040-13, Exam Feedback Form. Contact Instructor to obtain a copy of the form.

EXAMINATION INTEGRITY STATEMENT Cheating or compromising the exam will result in disciplinary actions up to and including termination.

I acknowledge that I am aware of the Examination Rules stated above. Further, I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination.

Examinees Signature: Date:

REVIEW ACKNOWLEDGEMENT I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.

Examinees Signature: Date:

Retention: 6 years Retain in: Training Records 2009 SRO NRC Master 8-10-09.doc

1 Point

1. During an accident the following plant conditions exist:

• RPV pressure 600 psig

• RPV water level +100 inches

• Drywell pressure 19 psig

• Torus water level 7.5 ft

• Torus pressure 18 psig Which one of the following is required based upon the above conditions?

a. Enter EOP-ED and emergency depressurize using the ADS SRVs.

b. Anticipate ED and rapidly depressurize with the bypass valves.

c. IAW EOP-1, RPV Control, cycle SRVs in sequence to establish a reactor cooldown at a rate <100°F/hour.

d. IAW EOP-1, RPV Control, cool down the RPV with the main turbine bypass valves or Alternate Pressure Control Systems (Table 7).

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 1 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 295030 EA2.01 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL : Suppression pool level Proposed Question: SRO Question # 76 Proposed Answer: A A. Correct -UNSAFE PSPL due to combination of low suppression pool level and high suppression chamber pressure EOP-02-PCC requires emergency depressurization. With Torus Water level above 4.5 feet ADS SRVs are used.

B. Incorrect - ED is required at this point and with Torus Water level above 4.5 feet ADS SRVs are used.

C. Incorrect - Must ED per procedure and OPEN 4 ADS SRVs.

D. Incorrect - Torus Water level is low but not low enough to require alternate emergency depressurization.

Technical EOP-2, Step PC/P-7 (Attach if not previously Reference(s): PSPL Curve provided)

Proposed References to be provided to applicants during EOP-2, T/L & PC/P legs examination: PSPL Curve Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 2 Exam Series A

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 3 Exam Series A

1 Point

2. While at 100% power, a partial loss of 125 VDC has rendered the 1D14 bus de-energized.

How are HPCI and RCIC affected and what TS actions are required?

a. The RCIC steam supply inboard isolation valve MO-2400 has lost power.

Immediately enter a 14 day LCO for RCIC being inoperable.

b. The RCIC steam supply outboard isolation valve MO-2401 has lost power.

Immediately enter a 14 day LCO for RCIC being inoperable.

c. The RCIC steam supply inboard isolation valve MO-2400 has lost power.

Immediately enter a 7 day LCO for RCIC being inoperable.

d. The RCIC steam supply outboard isolation valve MO-2401 has lost power.

Immediately enter a 7 day LCO for RCIC being inoperable.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 4 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 295004 AA2.04 Importance Rating 3.3 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : System lineups Proposed Question: SRO Question # 77 Proposed Answer: B A. Incorrect - The 1D14 bus affects the RCIC outboard isolation valve IAW SD 959.1 B. Correct - IAW TS 3.5.3 - this a 14 day LCO. The 1D14 bus affects the RCIC outboard isolation valve IAW SD 959.1 C. Incorrect - The LCO time is 14 days. The power supply issue affects the outboard valve.

D. Incorrect - The LCO time is 14 days.

Technical T.S. 3.5.3 Condition A (Attach if not previously Reference(s): AOP 302.1, page 12 provided)

Proposed References to be provided to applicants during none examination:

Learning Objective: (As available)

DAEC SRO Bank, Question Source: Bank #

Ques 2, pg 166 Modified Bank (Note changes or attach

1. parent)

New Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Knowledge Level:

Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 5 Exam Series A

10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 6 Exam Series A

1 Point

3. Following a spurious Main Turbine Trip and an ATWS, the following conditions exist:

• RPV water level was lowered reducing reactor power.

• RPV water level has been restored and is at +190

• All APRMs indicate downscale

• All ECCS systems are available

• SBLC has been injecting and tank level has reached 14%

• A majority of control rods remain stuck out of the core Which one of the following actions is required at this time?

a. Exit ATWS RPV Control and enter EOP 1, RPV Control.

b. Cool down and place Shutdown Cooling in service using SEP-306, Initiation of SDC for EOP Use.

c. Terminate boron injection and maintain RPV water level to 170 to 211 IAW EOP 1, RPV Control.

d. Maintain RPV water level using a Core Spray Pump IAW OI-151, Core Spray System.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 7 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 295037 EA2.03 Importance Rating 4.4 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

SBLC Tank Level.

Proposed Question: SRO Question # 78 Proposed Answer: B A. Incorrect - The criteria to exit ATWS-RPV Control is not met, ie all rods are not inserted and/or RE has not determined the reactor will remain shutdown under all conditions without boron.

B. Correct - With Cold Shutdown Boron Weight injected the reactor may be cooled down and shutdown cooling placed in service.

C. Incorrect - There is no direction to terminate injection. Injection should continue until the full contents of the SBLC tank are injected.

D. Incorrect - RPV water level can be restored at Hot Shutdown Boron Weight.

However restoring water level is done with preferred systems and Core spray is not a preferred system.

Technical (Attach if not previously ATWS-RPV Control, /P-5 Reference(s): provided)

Proposed References to be provided to applicants during ATWS RPV Control /L examination: without setpoints Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 8 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 9 Exam Series A

1 Point

4. The plant was operating at full power.

The control room must be evacuated due to a fire. The plant was scrammed and all rods were confirmed to be FULL IN prior to the evacuation.

Which one of the following describes:

(1) a task which must be completed by an in-plant operator and (2) the reason for that task?

a. (1) IAW AOP 915, Shutdown Outside the Control Room, dispatch an operator to Transfer to the Remote Shutdown Panels within 20 minutes.

(2) If an SRV has spuriously opened, a delay of more than 20 minutes in the transfer of control to 1C388 could result in RPV Level reaching TAF.

b. (1) IAW AOP 915, Shutdown Outside the Control Room, dispatch an operator to transfer to the Remote Shutdown Panels within 20 minutes.

(2) Failure to establish RPV level control with RCIC within 20 minutes could result in RPV level reaching TAF.

c. (1) IAW AOP 913, Fire, dispatch an operator within 20 minutes to establish additional ventilation in the 1A4 switchgear room.

(2) To ensure operability of the safety related electrical bus and provide adequate habitability.

d. (1) IAW AOP 913, Fire, immediately dispatch an operator to establish additional ventilation in the 1A4 switchgear room.

(2) To ensure operability of the safety related electrical bus and provide adequate habitability.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 10 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 295031 2.4.35 Importance Rating 4.0 Emergency Procedures / Plan: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects. (Reactor Low Water Level)

Proposed Question: SRO Question # 79 Proposed Answer: A Explanation (Optional):

A. Correct. IAW AOP 915 - Caution prior to TAB 2, step 5 operator actions If an SRV has spuriously opened, a delay in the transfer of control to 1C388 could result in RPV Level reaching TAF.

Per caution on Page 6 - For Control Room evacuation as the result of a fire, transfer of control at panels 1C388, 1C389, 1C390, 1C391, 1C392 is required to be completed within 20 minutes.

B. Incorrect. RCIC must be established for level control however, the 20 minute limitation applies to the SRV issue and not RCIC.

C. Incorrect. This is an action in AOP 915 and not AOP 913, Fire. It has no time requirement.

D. Incorrect. This is an action in AOP 915 and not AOP 913, Fire. It has no time requirement.

Technical (Attach if not previously AOP-915 Rev 39 Reference(s): provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 11 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 12 Exam Series A

1 Point

5. The plant was operating at full power. The following conditions exist:

• A fire, which was extinguished in 25 minutes, occurred in a vital area

• A Group II isolation has occurred Which one of the following describes:

(1) Components affected by the Group II isolation AND (2) Reportability requirements IAW 10 CFR 50.72

a. (1) Recirc mini purge, RHR sample isolation valves & Drywell Equipment Drain Isolation Valves (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> NRC Notification

b. (1) Recirc mini purge, RHR sample isolation valves & Drywell Equipment Drain Isolation Valves (2) 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> NRC Notification

c. (1) Drywell Floor Drain Isolation Valves, TIP Drive Ball Valves and RHR Drain to Radwaste Isolation Valves (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> NRC Notification

d. (1) Drywell Floor Drain Isolation Valves, TIP Drive Ball Valves and RHR Drain to Radwaste Isolation Valves (2) 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> NRC Notification Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 13 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 600000 2.2.37 Importance Rating 4.6 Equipment Control: Ability to determine operability and / or availability of safety related equipment. (Plant Fire On-site)

Proposed Question: SRO Question # 80 Proposed Answer: C A. Incorrect - The Recirc mini purge valves are not Group 2 PCIS.

B. Incorrect - The Recirc mini purge valves are not Group 2 PCIS and the NRC notification would be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> due to the Fire EAL. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification would be selected if the candidate focuses only on the PCIS isolation report, which is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification.

C. Correct - The valves listed are Group 2 PCIS isolation valves and the notification required for a vital area fire is a one hour notification.

D. Incorrect - The valves listed are Group 2 PCIS isolation valves but the EAL for the fire requires a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification would be selected if the candidate focuses only on the PCIS isolation report, which is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification.

Technical ACP 1402.3 (Attach if not previously Reference(s): System Description 959.1 p21 provided)

Proposed References to be provided to applicants during ACP 1402.3 examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 14 Exam Series A

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1, 5 (1) Conditions and limitations in the facility license.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 15 Exam Series A

1 Point

6. A Group 1 isolation and small break LOCA has occurred and the following conditions exist:

• All control rods FULL IN

• RPV pressure Controlling on LO-LO Set

• RPV level 155", rising slowly

• Torus level 11 feet, stable

• Torus pressure 12 psig, rising slowly

• Drywell temperature 220°F, rising slowly The operators attempted to place Torus Cooling in service but were not successful.

The STA reports that SPDS torus water temperature is reading 155°F and Graph 4, Heat Capacity Limit, limits are being approached.

Which one of the following actions is required for these conditions?

a. Immediately lower reactor pressure with SRVs based on SPDS Graph 4, Heat Capacity Limits, trend.

b. After verifying computer points are not marked with a YELLOW V, lower reactor pressure with SRVs based on SPDS Graph 4, Heat Capacity Limits, trend.

c. Confirm the SPDS reading by checking the 1C03 panel indications and, only if validated, exit EOP-2, Primary Containment Control and enter EOP-ED and emergency depressurize.

d. Confirm the SPDS readings by checking the 1C03 panel indications and, only if validated, lower reactor pressure with SRVs based on the EOP 2 Graph 4, Heat Capacity Limits, plot.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 16 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 295025 2.1.19 Importance Rating 3.8 Conduct of Operations: Ability to use plant computers to evaluate system or component status. (High Reactor Pressure)

Proposed Question: SRO Question # 81 Proposed Answer: D A. Incorrect. IAW OI-831.4, No Emergency action should be taken based on the SPDS data alone.

B. Incorrect. IAW OI-831.4, No Emergency action should be taken based on the SPDS data alone.

C. Incorrect. There is no requirement or need to exit EOP-2 and ED.

D. Correct. This value of torus temperature / reactor pressure requires a lowering of reactor pressure to maintain it within the safe region of the curve. SPDS data must be confirmed with panel indications prior to taking actions OI-831.4, Rev 64, Sect. 6, caution pg 35 Technical (Attach if not previously EOP-2, step T/T-6 and HCTL Reference(s): provided) curve SD-831.4a, page 51 Torus Temp leg of Proposed References to be provided to applicants during EOP-2 and HCTL examination:

Curve Learning Objective: (As available)

Question Source: Bank # X Modified Bank (Note changes or attach

1. parent)

New Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 17 Exam Series A

Last NRC 2005 Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 18 Exam Series A

1 Point

7. The plant is operating at full power.

The control room receives a call from ITC Midwest stating that lightning strikes have led to a degraded grid condition and a contingency trip of Duane Arnold would lead to an undervoltage condition in the DAEC switchyard 161 KV bus.

15 minutes after the ITC Midwest call, annunciator 1C-08C (B-4), MAIN GENERATOR FIELD MAX EXCITATION, alarms. 10 seconds later the alarm has not cleared.

Which one of the following describes:

(1) action(s) required due to the notification from ITC Midwest AND (2) action(s) required due to the alarm condition?

a. (1) Declare both Offsite Sources Inoperable IAW Technical Specifications (2) Shift to manual voltage control IAW AOP 304, Grid Instability

b. (1) Declare both Offsite Sources Inoperable IAW Technical Specifications (2) Verify the main generator has tripped and enter IPOI-5, Reactor Scram

c. (1) Start and load both SBDGs IAW OI 304.2, 4160V/480V Essential Electrical Distribution System (2) Shift to manual voltage control IAW AOP 304, Grid Instability

d. (1) Start and load both SBDGs IAW OI 304.2, 4160V/480V Essential Electrical Distribution System (2) Verify the main generator has tripped and enter IPOI-5, Reactor Scram Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 19 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 1 K/A # 700000 AA2.08 Importance Rating 4.3 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Criteria to trip the turbine or reactor Proposed Question: SRO Question # 82 Proposed Answer: B A. Incorrect - IAW AOP 304 - The AUTO Voltage Regulator will maintain generator operation within the generator capability curve. Operation of the over excitation limiter initiates annunciator 1C08C B-4. Once the limiter is initiated the auto voltage regulator may be limiting excitation of the generator.

Shifting to Manual Voltage Control under these conditions may cause a generator trip. Because a trip would have already occurred, this action is not correct.

B. Correct - IAW AOP 304 - IF notified by ITC Midwest that the contingency of a trip of the DAEC would lead to an undervoltage condition of < 99.2% in the DAEC switchyard 161 KV bus, THEN Declare both Offsite Sources inoperable and enter TS LCO actions as required by the mode of applicability.

IAW ARP 1C-08C (B-4) - If the overvoltage condition exists for longer than 5 seconds, the Voltage Regulator transfers from AUTOMATIC to MANUAL.

The following occurs; If either or both generator output breakers are closed, the generator trip will be via the Generator Backup Lockout Relay 286/B. With the plant online both generator output breakers are closed, the generator will trip.

If the generator trips and power is above 26%, a reactor scram and entry to IPOI 5 is required.

C. Incorrect - Per AOP 304 CAUTION - It is not appropriate to manually start and load a SBDG during degraged grid condtions. Do not use OI 304.2, section 7.6 TRANSFERRING ESSENTIAL BUS 1A3[4] FROM STARTUP OR STANDBY TRANSFORMER TO STANDBY DIESEL GENERATOR to attempt to put the essential buses on the SBDGs without the approval of Operations Management.

Shifting to Manual Voltage Control under these conditions may cause a generator trip. Because a trip would have already occurred, this action is not correct.

D. Incorrect - Per AOP 304 CAUTION - It is not appropriate to manually start and load a SBDG during degraged grid condtions. Do not use OI 304.2, section 7.6 TRANSFERRING ESSENTIAL BUS 1A3[4] FROM STARTUP OR STANDBY TRANSFORMER TO STANDBY DIESEL GENERATOR to attempt to put the essential buses on the SBDGs without the approval of Operations Management.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 20 Exam Series A

Technical ARP 1C08C, (B-4) Rev 46 (Attach if not previously Reference(s): AOP-304, Rev 22 provided)

Proposed References to be provided to applicants during none examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2,5 (2) Facility operating limitations in the technical specifications and their bases.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 21 Exam Series A

1 Point

8. A reactor scram has occurred from full power due to a complete Loss of Uninterruptible AC power. All 8 RPS Scram white lights are extinguished, but the 1C05 operator cannot confirm that all rods are fully inserted.

All LPRM downscale lights are on and when the IRMs are fully inserted, they read between range 3 and 4 and are lowering.

RPV pressure is 900 psig and rising very slowly with the Main Steam Line Drains open.

SBLC was not injected.

(1) Is the reactor considered SHUTDOWN UNDER ALL CONDITIONS WITHOUT BORON?

AND (2) How is the ATWS EOP utilized in this situation?

a. (1) NO (2) Exit the ATWS EOP and perform IPOI-5.

b. (1) NO (2) Exit only the /Q leg of the ATWS EOP.

c. (1) YES (2) Exit the ATWS EOP and perform IPOI-5.

d. (1) YES (2) Exit only the /Q leg of the ATWS EOP.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 22 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 2 K/A # 295015 AA2.01 Importance Rating 4.3 Ability to determine and / or interpret the following as they apply to INCOMPLETE SCRAM: Reactor power Proposed Question: SRO Question # 83 Proposed Answer: B A: Incorrect - Only the q leg of the ATWS EOP may be exited. The entire EOP may not be exited until it is determined that you are shutdown under all conditions B: Correct - Per ATWS EOP Bases Discussion Page 4, Shutdown under ALL conditions without boron can be determined by relying on the Technical Specification demonstration of adequate shutdown margin:

• One control rod is out beyond position 00

• All other control rods are at position 00 For other combinations of rod patterns and boron concentration, reactor engineering will need to perform a shutdown margin calculation.

When either of the conditions identified in the Continuous Recheck Statement is achieved, it is appropriate to terminate boron injection, exit the ATWS EOP, and enter EOP 1 for control of the transient.

Since these conditions are not given, the EOP may not be exited.

C: Incorrect - The conditions stated in the question stem do not meet the EOP Bases definition of Shutdown under ALL conditions without boron D: Incorrect - The conditions stated in the question stem do not meet the EOP Bases definition of Shutdown under ALL conditions without boron. The entire EOP would exited if that were the case.

Technical (Attach if not previously EOP ATWS bases Reference(s): provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank # X - 21075 Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 23 Exam Series A

Modified Bank (Note changes or attach

1. parent)

New Last NRC Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5, 6 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

(6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 24 Exam Series A

1 Point

9. An unisolable coolant system leak has occurred in the Reactor Building that has resulted in RPV level lowering to 17.

Operators recovered RPV level and were attempting to stabilize the plant when they noticed the following:

• A RED annunciator on panel 1C-35A (C-3) for REACTOR BLDG KAMAN 3, 4, 5 ,6, 7,&

8 HI RAD OR MONITOR TROUBLE

• PPC indicates that a Reactor Building Kaman Hi-Hi alarm exists The Kaman readings are as follows:

• REACTOR BLDG KAMAN 5/6 concentration is 9.3 E-3 ui/cc

• REACTOR BLDG KAMAN 7/8 concentration is 6.3 E-2 ui/cc The Reactor Building Exhaust Fans (1V-EF-11A & B) and the Main Plant Exhaust Fans (1V-EF-1, 2, & 3) responded as designed.

What actions must be directed and what Emergency Action Level must be declared?

a. Direct operators to TRIP the Main Plant Exhaust Fans.

If the above REACTOR BLDG KAMAN readings continue for 15 minutes, offsite Rad Conditions will then exceed the Site Area Emergency level.

Because RPV lowered to 17 before recovering, an Alert must be declared.

b. Direct operators to RESTART the Reactor Building Exhaust Fans.

If the above REACTOR BLDG KAMAN readings are expected to continue for 15 minutes, offsite Rad Conditions will exceed the Site Area Emergency level.

Because RPV lowered to 17 before recovering, a Site Area Emergency must be declared.

c. Direct operators to TRIP the Main Plant Exhaust Fans.

With the above REACTOR BLDG KAMAN readings, a Site Area Emergency must be declared.

d. Direct operators to RESTART the Reactor Building Exhaust Fans.

With the above REACTOR BLDG KAMAN readings, an Alert must be declared.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 25 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 2 K/A # 295017 2.4.41 Importance Rating 4.6 Emergency Procedures / Plan: Knowledge of the emergency action level thresholds and classifications. (High offsite release rate)

Proposed Question: SRO Question # 84 Proposed Answer: C A: Incorrect - The KAMAN levels have already exceeded the SAE criteria. The 15 minutes is associated with the Alert classification. There is no SAE classification for RPV level at 17.

B: Incorrect - Selected if the RB Kaman monitors are believed to be in the RB Vent Shaft rather than on the discharge of the MP Exhaust Fans. Operators are directed to restart the Turbine Bldg Exhaust Fans, not Reactor Building Exhaust Fans. There is no SAE classification for RPV level at 17.

C: Correct - At <170 inches a Group 3 isolation occurs which trips EF-11A&B, closes 1V-EF-13A & B, and aligns SBGT to draw on the RB Vent Shaft. EF1/2/3 continue to run and draw on the Main Plant Exhaust Plenum. The RB Vent Shaft and the MP Exhaust Plenum are physically separated by only a wall which, in the history of the plant, has been found to be cracked. Also the dampers 1V-EF-13A/B could be leaking, also allowing the RB Vent Shaft to flow to the MP Exhaust Plenum and out past 1V-EF-1/2/3 which normally continue to run after a Group 3 isolation. This is a real enough concern that there is a P&L in the Reactor Building HVAC OI, a Continuous Recheck statement in EOP-4 and Steps in ARP 1C35A C-3 step 3.3.a.

Per EAL Bases Document EBD-R Table on page 5, the SAE Level is exceeded REACTOR BLDG KAMAN 7/8 release rate.

D: Incorrect - Selected if the RB Kaman monitors are believed to be in the RB Vent Shaft rather than on the discharge of the MP Exhaust Fans. Operators are directed to restart the Turbine Bldg Exhaust Fans, not Reactor Building Exhaust Fans. The KAMAN levels have already exceeded the SAE criteria EBD-R page 5 table (EAL Technical (Attach if not previously bases)

Reference(s): provided)

ARP 1C35A C-3.

Proposed References to be provided to applicants during EAL Matrix Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 26 Exam Series A

examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach X

1. parent)

New Original Question:

An unisolable coolant system leak has occurred in the Reactor Building that has resulted in RPV level lowering to the point that fuel became uncovered and fuel damage occurred.

Operators recovered RPV level and were attempting to stabilize the plant when they noticed a RED annunciator on panel 1C35 for REACTOR BLDG KAMAN 3, 4, 5 ,6 , 7,& 8 HI RAD OR MONITOR TROUBLE.

The Reactor Building Exhaust Fans (1V-EF-11A & B) and the Main Plant Exhaust Fans (1V-EF-1, 2, & 3) responded as designed.

What could be the cause of this alarm and what actions must be directed regarding these fans to mitigate this condition?

a. The Main Plant Exhaust Fans must still be drawing on the Reactor Building Vent Shaft.

Direct operators to TRIP the Main Plant Exhaust Fans.

b. The Main Plant Exhaust Fans will have tripped causing a high concentration of activity at the monitors. Direct operators to RESTART the Main Plant Exhaust Fans.

c. The Reactor Building Exhaust Fans must still be drawing on the Reactor Building Vent Shaft.

Direct operators to TRIP the Reactor Building Exhaust Fans.

d. The Reactor Building Exhaust Fans will have tripped causing a high concentration of activity at the monitors. Direct operators to RESTART the Reactor Building Exhaust Fans.

Last NRC Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 (1) Conditions and limitations in the facility license.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 27 Exam Series A

1 Point

10. A Loss of Coolant Accident has occurred which requires operators to perform SEP 301.1, Torus Vent via SBGT. The following conditions exist:

• One train of Standby Gas Treatment (SBGT) is in operation

• Drywell pressure is 50 psig and still rising slowly

• Three Torus vent valves are open o CV-4301, OUTBD TORUS VENT ISOL.

o CV-4309, INBD TORUS VENT BYPASS ISOL.

o CV-4300, INBD TORUS VENT ISOL.

After opening CV-4300, airborne activity and radiation levels on Reactor Building Second Floor (El. 786 ft.) have risen dramatically.

Which of the following has caused this condition and what actions are required to continue venting?

a. The SBGT inlet relief damper has opened due to excessive pressure; start the standby SBGT Train IAW OI 170, SBGT System, to raise SBGT system flow.

b. The SBGT inlet relief damper has opened due to excessive pressure; assess the need for venting and use the Hard Pipe Vent per SEP 301.3 as required.

c. The Hard Pipe Vent rupture disc has ruptured; assess the need for venting and shift to Drywell vent per SEP 301.2 as required.

d. The SBGT inlet relief damper has opened due to excessive pressure; throttle MO-4309A, BYPASS VENT THROTTLE, as needed to lower pressure.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 28 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 1 Group # 2 K/A # 295033 EA2.03 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : Cause of high area radiation Proposed Question: SRO Question # 85 Proposed Answer: B A. Incorrect - this provides no additional flow and does not lower pressure B. Correct - Per SEP 301.1, If SBGT inlet pressure approaches 10" WG, assess the need for continued venting and/or use of the Hard Pipe Vent per SEP 301.3.

Caution, If SBGT inlet pressure exceeds 10" WG, the SBGT inlet relief damper will open and relieve pressure into the RB 786 Level.

C. Incorrect - The hard pipe vent rupture disc does not discharge inside the Reactor Building.

D. Incorrect - Throttling with MO-4309A is specifically prohibited by SEP 301.1 CAUTION, it has a non-essential power supply and may impede venting.

Technical SEP 301.1, Rev 6 Step 7 and (Attach if not previously Reference(s): caution pg 4 provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 29 Exam Series A

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4, 5 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 30 Exam Series A

1 Point

11. The plant is at full power.

Then, annunciator 1C-03A (C-8), A CORE SPRAY SPARGER LO P, alarms. The operators confirm it is a valid alarm.

Which one of the following describes: (1) the reason for the alarm and (2) the required Technical Specification action?

a. (1) An A Core Spray piping leak/break has occurred INSIDE the Core Shroud (2) Declare the A Core Spray Loop inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO

b. (1) An A Core Spray piping leak/break has occurred INSIDE the Core Shroud (2) Declare the A Core Spray Loop inoperable and enter a 7 day LCO

c. (1) An A Core Spray piping leak/break has occurred BETWEEN the Reactor Pressure Vessel wall and the Core Shroud (2) Declare the A Core Spray Loop inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO

d. (1) An A Core Spray piping leak/break has occurred BETWEEN the Reactor Pressure Vessel wall and the Core Shroud (2) Declare the A Core Spray Loop inoperable and enter a 7 day LCO Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 31 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 1 K/A # 209001 A2.05 Importance Rating 3.6 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Core Spray line break Proposed Question: SRO Question #86 Proposed Answer: D Incorrect - The alarm is not an indication of an inside the shroud break based A: upon its tap off point on the LPCS piping. A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO would be required for 2 loops of Core Spray inoperable Incorrect - The alarm is not an indication of an inside the shroud break based B: upon its tap off point on the LPCS piping.

Incorrect - A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO would be required for 2 loops of Core Spray C: inoperable Correct - Per ARP 1C-03A (C-8) - this alarm is from the Core Spray System Header to top of the Core plate and caused by differential pressure low. This D: could be indication of a Core Spray line break inside the Reactor vessel.

TS 3.5.1.B. - 7 days, applies for 1 core spray loop inoperable Technical ARP 1C03A (C-8) Rev 48 (Attach if not previously Reference(s): TS 3.5.1.B provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 32 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 33 Exam Series A

1 Point

12. The plant is in HOT SHUTDOWN. The B Shutdown Cooling (SDC) Loop is in service with B RHR and B RHRSW pumps running.

MO1940, RHR HX 1E-201B BYPASS, and MO1939, RHR HX 1E-201B INLET THROTTLE, valves are THROTTLED in mid position.

• MO1904 and MO1905, RHR Loop B Inject Isolation Valves are OPEN.

• MO1908 and MO1909, RHR Shutdown Cooling Suction Isolation Valves are OPEN.

Annunciator 1C03B (B-4), RHR SHUTDOWN COOLING SUCTION HEADER HI PRESSURE, alarms and SDC Header pressure is reported to be 105 psig and rising at 2 psig per minute.

You direct the operators to raise the cooldown rate.

Several minutes later, the 1C03 operator reports RHR suction header pressure is 125 psig and MO1940 is not responding.

Annunciator 1C05B (D-8), PCIS GROUP 4 ISOLATION INITIATED, has alarmed; and the operator reports that RHR suction header pressure is at 140 psig.

No other plant conditions have changed.

Based on these plant conditions, you direct the operators to ________?

a. start the D RHRSW pump and raise RHRSW flow IAW OI 416, RHRSW System. Enter the Technical Specification Limiting Condition for Operation for LPCI.

b. throttle OPEN more on MO1939 and start the D RHR pump if necessary. Enter the Technical Specification Limiting Condition for Operation for LPCI.

c. verify CLOSED MO1905, verify the B RHR pump tripped, and verify CLOSED MO1908 and MO1909. Enter AOP 149, Loss of Decay Heat Removal.

d. verify CLOSED MO1939, start the D RHR pump and then re-establish SDC flow. Enter AOP 149, Loss of Decay Heat Removal, until SDC is re-established.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 34 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 1 K/A # 223002 A2.03 Importance Rating 3.3 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System logic failures Proposed Question: SRO Question #87 Proposed Answer: C A: Incorrect - SDC should have isolated at 135 psig. The ARP for a Group 4 should be carried out. Increasing RHRSW flow is not part of the ARP guidance B: Incorrect - ARP 1C03B B-4 directs increasing cooldown with MO 1939 and another pump would help flow. T.S. should be entered on failure of MO-1940.

However, the plant is above the PCIS Group 4 pressure and SDC should be promptly removed and isolated.

C: Correct - The initial alarm indicates an increase in RPV temperature and pressure. The ARP directs increasing the cooldown rate to lower pressure, which was directed. At 135 psig a PCIS group 4 should have occurred but did not. ARP 1C05B D-8 PCIS Group 4 Isolation should be in alarm and SDC secured. The CRS should direct the actions from the ARP that did not occur. In this case securing SDC is appropriate. Also entry into AOP 149 is directed.

D: Incorrect - Starting a second RHR pump would increase flow. AOP 149 entry is correct when SDC is lost and recovery of SDC will be the goal. However, the plant is above the PCIS Group 4 pressure(D RHR pump wont start under these conditions) and SDC should be promptly removed and isolated as directed in ARP 1C05B D-8 for pressure protection of the RHR piping.

Technical 1C03B B-4 Rev 36 (Attach if not previously Reference(s): 1C05B D-8 Rev 81 provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 35 Exam Series A

Question Source: Bank # X Modified Bank (Note changes or attach

1. parent)

New Last NRC 2002 Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 36 Exam Series A

1 Point

13. A plant startup is in progress and the Mode Switch is ready to be placed in RUN.

The only inoperable equipment is IRM B which is bypassed due to a downscale failure. I&C work is in progress.

Then, a half scram and a Rod Block occurs on RPS Channel B.

I&C reports they lifted the wrong lead in the IRM panels and caused an INOP trip on IRM D.

Which one of the following describes whether the Technical Specification (TS) actions have been met and whether TS permits the Mode Switch to be taken to RUN in this condition?

a. The TS required actions are already met with the trip on RPS Channel B.

The Mode Switch may NOT be taken to RUN until at least one of the IRMs (B or D) is declared operable.

b. The TS required actions are already met with the trip on RPS Channel B.

The Mode Switch may be taken to RUN because the IRM TS does not apply in MODE 1.

c. The TS required actions are NOT met.

The Mode Switch may NOT be taken to RUN until at least one of the IRMs (B or D) is declared operable.

d. The TS required actions are NOT met.

The Mode Switch may be taken to RUN because the IRM TS does not apply in MODE 1.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 37 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 1 K/A # 215003 2.2.36 Importance Rating 4.2 Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Proposed Question: SRO Question #88 Proposed Answer: B A: Incorrect - TS 3.3.1.1.A requires the channel to be in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is met with the RPS trip. Since the IRMs are not required in mode 1, the mode switch may be moved.

B: Correct - TS 3.3.1.1.A requires the channel to be in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is met with the RPS trip. TS 3.0.4 permits a mode change to a mode where the TS does not apply if a risk assessment and establishment of risk management actions is performed first.

C: Incorrect - The TS actions are met and the mode switch may be moved.

D: Incorrect - TS 3.3.1.1.A requires the channel to be in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is met with the RPS trip.

Technical TS 3.3.1.1 (Attach if not previously Reference(s): TS 3.0.4 provided)

NO RPS Proposed References to be provided to applicants during instrumentation examination: Tables No TS Section 3.0 Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 38 Exam Series A

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 39 Exam Series A

1 Point

14. The plant is currently in an electrical ATWS with the following conditions:

• ADS is locked out

• The MSIVs are Closed

• Defeat 11, Containment N2 Supply Isolation Defeat, has been installed

• Reactor power is cycling between 25% and 55% power

• Power level control has been entered

• SBLC is injecting

• The RIPs are being implemented The 1C03 operator reports the following parameters:

• RPV Pressure is cycling between 1080 psig and 1130 psig

• Both SRV 4401 and 4407; the LLS SRVs are open

• SRV 4400 is opening and closing Which one of the following describes a required action, if any, based on the above conditions?

a. The opening and closing SRV may cause significant power transients but all systems are operating as designed, so NO EOP actions are required.

b. The main concern in this condition is that SRV 4400 could stick open.

Place HPCI in service IAW OI 152 QRC 1, HPCI Rapid Start, and/or RCIC in service IAW OI 150 QRC 1, RCIC Rapid Start, in CST to CST mode for pressure control.

c. The opening and closing of the SRVs exerts significant dynamic loads on the SRV tailpipes and support structures so manual control of SRVs is required IAW EOP ATWS.

d. With the SRVs opening and closing, RPV level control becomes very difficult, so lowering of RPV level IAW EOP ATWS is necessary to slow the opening and closing of the SRVs.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 40 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 1 K/A # 239002 2.1.23 Importance Rating 4.4 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation. (SRVs)

Proposed Question: SRO Question #89 Proposed Answer: C A: Incorrect - Systems are operating as designed however the EOP at step P-3 states Manually open SRVs to terminate SRV cycling.

B: Incorrect - This a concern however this is not the action required.

C: Correct - Per EOP ATWS Page 55 discussion of Step /P-3. Step directs Manually open SRVs to terminate cycling. Embedded in the bases is the definition of Cycling: multiple sequenced valve actuations with valve opening being initiated in response to RPV pressure increasing to or above the lifting setpoint and valve closure being governed by RPV pressure decreasing to or below the reset setpoint. The concerns with cycling are also stated including exerting significant dynamic loads on the SRV tailpipes and support structures.

D: Incorrect - Level control is a concern however, lowering level is not the action required.

Technical (Attach if not previously EOP ATWS Bases Rev 12 Reference(s): provided)

DO NOT Proposed References to be provided to applicants during PROVIDE EOP examination:

ATWS /P LEG Learning Objective: (As available)

Question Source: Bank # DAEC SRO Bank Modified Bank (Note changes or attach

1. parent)

New Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 41 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 42 Exam Series A

1 Point

15. The plant is operating at full power. All systems are operable.

You are provided with the following information:

• SBLC Tank Concentration is 14%

• SBLC Volume 3200 gallons

• SBLC pump suction piping Temperature is 66°F Which one of the following describes:

(1) The status of the SBLC system AND (2) The bases of the Technical Specification (TS) LCOs

a. (1) SBLC is inoperable due to a lower than required concentration for the given tank volume.

(2) It assures that the SBLC system can be relied upon to satisfy the requirements of the ATWS Rule, 10 CFR 50.62, Anticipated Transients without Scram (ATWS).

b. (1) SBLC is inoperable due to a lower than required temperature for the given concentration.

(2) It assures that the SBLC system can be relied upon to satisfy the requirements of the ATWS Rule, 10 CFR 50.62, Anticipated Transients without Scram (ATWS).

c. (1) SBLC is inoperable due to a lower than required concentration for the given tank volume.

(2) It assures that Hot Shutdown Boron Weight would be injected when the SBLC tank is at a level of 47%.

d. (1) SBLC is inoperable due to a lower than required temperature for the given concentration.

(2) It assures that Hot Shutdown Boron Weight would be injected when the SBLC tank is at a level of 47%.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 43 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 1 K/A # 211000 2.2.25 Importance Rating 4.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (SLC)

Proposed Question: SRO Question #90 Proposed Answer: B A: Incorrect - IAW TS Table 3.1.7.1-2, the concentration is too low for the tank volume B: Correct - IAW TS Table 3.1.7.1-2, the temperature is too low for the concentration.

IAW TS Bases 3.1.7, the SLC system is relied upon to satisfy the requirements of 10 CFR 50.62 (ATWS Rule)

C: Incorrect - IAW TS Table 3.1.7.1-2, the concentration is too low for the tank volume.

Although if operable HSBW will be achieved. It is not the bases of the TS.

D: Incorrect - Although if operable HSBW will be achieved. It is not the bases of the TS.

Technical TS bases 3.1.7 (Attach if not previously Reference(s): TS 3.1.7 & figures provided)

Proposed References to be provided to applicants during TS 3.1.7 w/ figures examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 44 Exam Series A

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 6 (2) Facility operating limitations in the technical specifications and their bases.

(6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 45 Exam Series A

1 Point

16. The plant is operating at 90% power.

The following rods have been declared slow based on scram time testing: 14-23, 14-27 and 18-39.

At 1430 today, control rod 18-23 receives an accumulator alarm 1C05A (F-7), CRD ACCUMULATOR LO OR HI LEVEL.

An operator sent to investigate reports that, when the local panel pushbutton was depressed for HCU 18-23, the local alarm light remains lit for that HCU.

Based on the operator report, what caused the accumulator alarm and what, if any, action(s) is required by Technical Specifications?

a. The accumulator has a high water level.

Declare the control rod inoperable OR slow within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If the control rod is declared slow, be in MODE 3 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. The accumulator has a high water level.

Declare the control rod inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Once declared inoperable, the control rod is required to be fully inserted AND disarmed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

c. The accumulator has a low pressure.

If accumulator pressure is <940 psig, declare the control rod inoperable OR slow within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If the control rod is declared slow, be in MODE 3 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d. The accumulator has a low pressure.

If accumulator pressure is <940 psig, declare the control rod inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Once declared inoperable, the control rod is required to be fully inserted AND disarmed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 46 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 2 K/A # 201003 A2.08 Importance Rating 3.7 Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including: Low HCU accumulator pressure/high level Proposed Question: SRO Question #91 Proposed Answer: C A: Incorrect - the cause of the alarm is low pressure B: Incorrect - Per TS 3.1.3 - If declared inoperable, the rod must be fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C: Correct - Per SD 255 page 26 - The alarms for low nitrogen pressure and accumulator leakage are also annunciated on the local accumulator alarm panels 1C054 and 1C072. The alarm panels consist of a pushbutton for each accumulator that lights up when either low nitrogen pressure or accumulator piston leakage is detected. If the light stays energized when the pushbutton is depressed, the originating signal is low nitrogen pressure; if the light de-energizes when the pushbutton is depressed, the accumulator water level switch is actuated.

Per TS 3.1.5 - With One control rod scram accumulator inoperable with reactor steam dome pressure 900 psig, Declare the associated control rod scram time slow. OR Declare the associated control rod inoperable.

Per TS 3.1.4 - No more than 2 OPERABLE control rods that are slow shall occupy adjacent locations. If this rod were declared slow, 3 OPERABLE control rods that are slow would occupy adjacent locations. Therefore, the LCO applies to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D: Incorrect - Per TS 3.1.3 - If declared inoperable, the rod must be fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Technical TS 3.1.3, 3.1.4, 3.1.5 (Attach if not previously Reference(s): System Description 255, pg 26 provided)

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 47 Exam Series A

TS 3.1.3, 3.1.4, Proposed References to be provided to applicants during

3.1.5 examination

Core map Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 48 Exam Series A

1 Point

17. The plant is operating at 62% power during power ascension. The second Condensate and Feed pumps have been started.

At this point, the "A" Condensate pump trips.

Which one of the following describes the response of the Feedwater System and required actions?

a. Only the "A" Feed pump will trip due to an interlock with the "A" Condensate pump.

Enter AOP 644, Feedwater/Condensate Malfunction, reduce reactor power to less than 60% using Recirc and/or control rods or maintain Reactor Feed Pump current to less than 960 amps.

b. Only the "A" Feed pump will trip due to an interlock with the "A" Condensate pump.

Select B Level of the Reactor Water Level Control Input. If RPV level cannot be maintained, then direct a reactor scram and entry into IPOI 5, Reactor Scram.

c. Both Feed pumps will continue to operate because one Condensate pump can adequately supply both Feed pumps at this power level.

Enter AOP 644, Feedwater/Condensate Malfunction, reduce reactor power with recirc to less than 60%, and take manual control of Feedwater controllers as needed.

d. Both Feed pumps will trip on low suction pressure due to the inability of one Condensate pump to supply both Feed pumps.

Enter EOP 1, RPV Control, and IPOI 5, Reactor Scram, and control RPV level with condensate.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 49 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 2 K/A # 256000 2.4.49 Importance Rating 4.4 Emergency Procedures / Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(Condensate)

Proposed Question: SRO Question #92 Proposed Answer: A A: Correct - Per SD 644, page 7 - RFP 1P-1A (1P-1B) is tripped by the loss of condensate pump 1P-8A (1P-8B) during two RFP operation, or by the loss of both condensate pumps when it is the only feed pump running.

Per AOP 644, immediate actions - If reactor power (prior to the event) was less than (<) 75%, reduce reactor power to less than (<) 60% using Recirc and/or control rods or maintain Reactor Feed Pump current to less than (<) 960 amps.

B: Incorrect -Selection of the alternate level control input will not affect feedwater response due to the loss of the pump.

C: Incorrect - The A feed pump will trip. Per the AOP - If reactor power (prior to the event) was less than (<) 75%, reduce reactor power to less than (<) 60% using Recirc and/or control rods or maintain Reactor Feed Pump current to less than

(<) 960 amps.

D: Incorrect - ONLY the A Feedwater pump will trip, A scram should not be required at this power level. Feed pumps do not have low suction pressure trips AOP 644 Rev 5 Technical (Attach if not previously SD 644 Rev 9 Reference(s): provided)

ARP 1C06A (A-12) Rev 51 Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank #

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 50 Exam Series A

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 51 Exam Series A

1 Point

18. The plant is operating at full power. A radiological event on the refuel floor causes a release.

Then, annunciator 1C-07A (D-11), Control Building HVAC Panel 1C-26 Trouble, alarms.

Operators are dispatched to investigate the alarm. They report the following two 1C-26 alarms:

• 1C26A (C-2), Control BLDG Intake Air Rad Mon RIM-6101A Hi/Trouble

• 1C26B (C-2), Control BLDG Intake Air Rad Mon RIM-6101B Hi/Trouble Which one of the following describes the effects on control room ventilation and action that is required?

a. A Control Building isolation should have occurred. Verify only one Battery Exhaust fan is running IAW OI 730, Control Building HVAC System.

b. A Control Building isolation should have occurred. Verify two Battery Exhaust fans are running IAW OI 730, Control Building HVAC System.

c. Verify that Control Building pressure is being maintained at a negative value. Verify only one Battery Exhaust fan is running IAW ARP 1C26A & B (C-2).

d. Verify that Control Building pressure is being maintained at a positive value. Verify two Battery Exhaust fans are running IAW ARP 1C26A & B (C-2).

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 52 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 2 Group # 2 K/A # 272000 2.1.31 Importance Rating 4.3 Conduct of Operations: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(Radiation Monitoring)

Proposed Question: SRO Question # 93 Proposed Answer: A Explanation (Optional): KA Justification - This KA is typically used for scenario/JPM evaluation. In this case a question was asked which requires the ability to determine control room indication given an event and then determine how the indications reflect the control room ventilation lineup and pressure. Additionally, the applicant must determine the appropriate action to be taken for the event.

A. Correct - Per OI 730 P&L 9, page 5, to maintain positive pressure during a control building isolation, only ONE battery exhaust fan shall be running.

ARP 1C26A & B (C-2) contains the same information.

B. Incorrect - Only one fan shall be running.

C. Incorrect - Positive pressure shall be maintained.

D. Incorrect - Positive pressure shall be maintained. Only one fan shall be running Technical OI 730 Rev 100 P&L #9 page 5 (Attach if not previously Reference(s): ARP 1C26A & B (C-2) Rev 48 provided)

Proposed References to be provided to applicants during None examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 53 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4, 5 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 54 Exam Series A

1 Point

19. While supervising fuel handling activities in the Spent Fuel Pool, you discover a minor typographical error in the approved Fuel Moving Plan (FMP) that you are using.

The final orientation for the spent fuel bundle being moved is illegible.

Which of the following describes the process for correcting the error to the fuel moving plan?

a. Minor pen & ink changes to the FMP may be made by the Fuel Handling Supervisor with concurrence from the Shift Manager.

b. Any changes in the FMP require a Procedure Change Request initiated by Reactor Engineering with concurrence from the Fuel Handling Supervisor and the Shift Manager.

c. Minor pen & ink changes to the FMP may be made by Reactor Engineering with concurrence from the Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager.

d. Minor pen & ink changes to the FMP may be made by the Fuel Handling Supervisor with concurrence from Reactor Engineering. The Shift Manager must be advised but Shift Manager concurrence is NOT required.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 55 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 1 K/A # 2.1.40 Importance Rating 3.9 Knowledge of refueling administrative requirements Proposed Question: SRO Question # 94 Proposed Answer: C A: Incorrect - Concurrence is required by Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager.

B: Incorrect - A procedure change request is not required.

C: Correct - PER RFP 4-3. Step 5.1.1.e - Minor pen & ink changes to the FMP may be made by Reactor Engineering with concurrence from the Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager.

D: Incorrect - Concurrence is required by Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager.

Technical (Attach if not previously RFP 403 Rev 33 Step 5.1.1.e.

Reference(s): provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: Fuel handling 1.4.1.1. (As available)

Question Source: Bank # DAEC 22624 Modified Bank (Note changes or attach

1. parent)

New Last NRC No Question History:

Exam:

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 56 Exam Series A

Question Cognitive Memory or Fundamental X

Level: Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 7 (7) Fuel handling facilities and procedures.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 57 Exam Series A

1 Point

20. System engineering has proposed a new performance test on the RCIC pump which will affect pump flow rate. Engineering has determined that the Technical Specification for pump flow would not be adversely affected during the test.

IAW ACP 1407.4, Special Test Procedures (SpTP), which one of the following describes how the test is classified and who must provide written approval for the SpTP prior to performance?

a. This test is considered an Infrequently Performed Test or Evolution AND a Special Test.

The Plant Manager and the CRS.

b. This test is considered ONLY a Special Test.

The Plant Manager and the CRS.

c. This test is considered an Infrequently Performed Test or Evolution AND a Special Test.

ONLY the on-shift CRS.

d. This test is considered ONLY a Special Test.

ONLY the on-shift CRS.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 58 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 2 K/A # 2.2.7 Importance Rating 3.6 Knowledge of the process for conducting special or infrequent tests.

Proposed Question: SRO Question # 95 Proposed Answer: C A: Incorrect - Any Special Test is also considered an Infrequently Performed Test or Evolution. Although the test may be reviewed by the Plant Manager, their written approval is not required prior to on shift performance B: Incorrect - Any Special Test is also considered an Infrequently Performed Test or Evolution AND a Special Test C: Correct - Per ACP 1407.4 - Special Test or Experiment - Non-routine operations performed to determine the performance characteristics of a structure, system or component. Special Tests are non-routine tests that are not required by the Technical Specifications, a 10CFR 72 Certificate of Compliance, or the ASME Section XI Manual, and are not described in the UFSAR or a 10CFR 72 Final Safety Analysis Report (Certificate Holders), as updated.

Per ACP 1407.4 Step 3.3 (10) - SpTPs are considered Infrequently Performed Test or Evolutions (IPTEs). Refer to ACP 102.17, Pre/Post-Job Briefs and Infrequently Performed Tests and Evolutions, for IPTE requirements.

Per ACP 1407.4 Step 3.5 (3) - All SpTPs require written authorization from the on-shift CRS prior to performance.

D: Incorrect - Any Special Test is also considered an Infrequently Performed Test or Evolution AND a Special Test Technical ACP 1407.4 Rev 21 Definitions, (Attach if not previously Reference(s): Steps 3.5 (3) provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Question Source: Bank #

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 59 Exam Series A

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental X

Level: Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 (1) Conditions and limitations in the facility license.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 60 Exam Series A

1 Point

21. With the plant in MODE 1, an Outboard Primary Containment Isolation Valve, required to be operable in MODES 1, 2 and 3, failed its stroke time testing. To comply with the associated LCO, the inoperable valve has been CLOSED and DEACTIVATED.

Which ONE of the following describes the conditions REQUIRED for Post Maintenance Testing to restore OPERABILITY, which includes electrically stroking this valve?

a. This valve CANNOT be electrically stroked until the plant is in MODE 4, COLD SHUTDOWN, when the valve is not required to be operable.

b. This valve may be electrically stroked under Administrative Control without regard to the position of the other isolation valve in the same line.

c. This valve may ONLY be electrically stroked if the INBOARD valve in the same line is CLOSED.

d. This valve may ONLY be electrically stroked if the valve is reclosed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> IAW Technical Specifications.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 61 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 2 K/A # 2.2.21 Importance Rating 4.1 Knowledge of pre- and post-maintenance operability requirements Proposed Question: SRO Question # 96 Proposed Answer: B A: Incorrect - In MODE 4, Primary Containment Isolation Valve OPERABILITY is NOT APPLICABLE. It is not required to shutdown to stroke this valve.

B: Correct - Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

C: Incorrect - Redundant valve closure is an acceptable method to allow valve stroking, but it is not the ONLY acceptable method.

D: Incorrect - There is no requirement to have the valve reclosed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of opening it. The requirement is to have administrative control of the valve opening.

Technical (Attach if not previously TS LCO 3.0.5 Reference(s): provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Question Source: Bank # WTS - 2496 Modified Bank (Note changes or attach

1. parent)

New Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 62 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental X

Level: Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 63 Exam Series A

1 Point

22. The plant is in MODE 5, with the following:

• Fuel Movements are in progress between the cavity and the fuel pool

• SDC Cooling Isolation Valve MO-1909 spuriously closed and is jammed on its closed seat

• Shutdown Cooling Flow has been secured for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

• Maintenance is working on several of the outboard MSIVs

• Reactor Coolant temperature is 105 degrees F.

Which one of the following actions will result in meeting Technical Specification requirements for an alternate means of decay heat removal?

a. Start a Recirc Pump immediately regardless of the core configuration IAW OI 264, Reactor Recirculation System, to provide forced circulation.

b. Raise reactor water level and control it between 230 and 240 inches as measured on the GEMACs IAW AOP 149, Loss of Decay Heat Removal. Increase CRD flow to enhance natural circulation.

c. Establish Feed and Bleed to the Torus via the SRVs IAW OI 183.1, Automatic Depressurization System. Ensure all personnel are cleared from the Torus.

d. Align Fuel Pool Cooling return to the vessel cavity IAW AOP 149, Loss of Decay Heat Removal. RBCCW flow and cooling must be maximized.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 64 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 4 K/A # 2.4.9 Importance Rating 4.2 Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies Proposed Question: SRO Question # 97 Proposed Answer: D A: Incorrect - Per AOP 149 this is not defined as an alternate means of decay heat removal to satisfy TS.

B: Incorrect - Cavity is already flooded to the weirs and Floodup level indication is used, not GEMACS C: Incorrect - Not an acceptable method because steam line plugs are installed D: Correct - This is a prescribed method in AOP 149 Section 4.5 Technical AOP 149 Rev 31 (Attach if not previously Reference(s): TS 3.9.7.Bases A.1 provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank (Note changes or attach

1. parent)

New X Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 65 Exam Series A

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2,5 (2) Facility operating limitations in the technical specifications and their bases.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 66 Exam Series A

1 Point

23. The plant was initially operating at full power. A fuel leak resulted in high Offgas and Main Steam Line Radiation Levels.

AOP 672.2, Offgas Radiation, Reactor Coolant High Activity has been entered and a plant shutdown is being performed to comply with Technical Specifications.

Then, a spurious Main Turbine trip occurred and the plant automatically scrammed.

Plant conditions are as follows:

• ALL Control Rods are fully inserted

• Reactor level lowered to 160 following the scram and is now stable at 184

• Reactor Pressure is 920 psig with the Turbine Bypass Valves in service

• Offgas is in service, maintaining 2 inches Hg Backpressure

• 1C05B C-2 MAIN STEAM LINE HI HI RAD / INOP TRIP continues to alarm With these conditions, which one of the following actions are required and will MINIMIZE release of radioactivity to the environment?

a. Enter EOP 1, RPV Control, and maintain RPV level 170 to 211. No additional EOP entries are required.

Cooldown at LESS THAN 100°F/hr by depressurizing to the Main Condenser to allow the Offgas treatment process to limit radioactivity releases.

b. Enter EOP 1, RPV Control, and EOP 4, Radioactivity Release Control.

Rapidly cooldown at GREATER THAN 100°F/hr by depressurizing to the Main Condenser to allow the Offgas treatment process to limit radioactivity releases.

c. Enter EOP 1, RPV Control, and maintain RPV level 170 to 211. No additional EOP entries are required.

Cooldown at LESS THAN 100°F/hr by depressurizing to the Torus to allow the Containment to limit radioactivity release and allow the Main Condenser to be used to control MSIV Leakage.

d. Enter EOP 1, RPV Control, and EOP 4, Radioactivity Release Control.

Rapidly cooldown at GREATER THAN 100°F/hr by depressurizing to the Torus to allow the Containment to limit radioactivity release and allow the Main Condenser to be used to control MSIV Leakage.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 67 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 4.3 Ability to control radiation releases Proposed Question: SRO Question # 98 Proposed Answer: C A: Incorrect - Action would be correct for a normal shutdown without High RCS Activity concerns.

B: Incorrect - Action would be correct if Emergency Depressurization were anticipated during EOP execution. No reasons are provided in stem for ED C: Correct - AOP 672.2, Off Gas Radiation, Reactor Coolant High Activity specifies closing the MSIVs and MSL Drains, depressurizing to the Torus. Main Steam and Main Condenser will be aligned to limit MSIV Leakage. NO requirement has been given to Anticipate Emergency Depressurization, so normal cooldown limits are in effect.

EOP -1 entry required on low RPV level, IPOI 5 entry not required because the scram already occurred (EOP 1 Decision Step RC-2)

No other EOP entries exist.

D: Incorrect - Action would be correct if Emergency Depressurization were required and if EOP-4 Radioactivity Release Control, were entered. No entry conditions for these are given in stem Technical AOP 672.2 Rev 33 Step 6 (Attach if not previously Reference(s): EOP - 1 provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Question Source: Bank # WTS - 2499 Modified Bank (Note changes or attach

1. parent)

New Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 68 Exam Series A

Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental Level: Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4, 5 (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 69 Exam Series A

1 Point

24. An event has occurred at the plant. The TSC and EOF are activated but NOT yet operational.

IAW Emergency Plan Implementing Procedures, which one of the following describes the individual responsible for escalating an emergency event level from a Site Area Emergency to a General Emergency?

a. Shift Manager

b. Operations Manager

c. Emergency Response & Recovery Director

d. Site Vice President Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 70 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 4 K/A # 2.4.38 Importance Rating 4.4 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Proposed Question: SRO Question # 99 Proposed Answer: A A: Correct- Per EPIP 2.5 - Step 3.1 (1) - Upon determining that the plant is in an unexpected operational condition, the Operations Shift Manager/Control Room Supervisor (OSM/CRS) shall evaluate plant conditions using guidance contained in EPIP 1.1, "Determination of the Emergency Action Level," and, as warranted, classify the event in one of the four emergency categories.

Per Step 3.1.(2).(a) - The OSM/CRS shall function additionally as the Emergency Coordinator and Site Radiation Protection Coordinator until relieved of such function by appropriately qualified personnel.

Until the TSC and EOF are operational, the SM retains the responsibility of escalating the event.

B: Incorrect - The SM/CRS is the EC until the other facilities are operational.

C: Incorrect - The Emergency Response & Recovery Director would be responsible if the EOF were operational D: Incorrect - The Site VP is not designated as the EC for the described situation.

Technical (Attach if not previously EPIP 2.5 Rev 17 Reference(s): provided)

Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Question Source: Bank # WTS Modified Bank (Note changes or attach

1. parent)

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 71 Exam Series A

New Last NRC No Question History:

Exam:

Question Cognitive Memory or Fundamental X

Level: Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 72 Exam Series A

1 Point

25. It is 0400 and the plant is in Hot Shutdown. The STA is informed by their spouse that they must return home immediately for a family emergency.

• At 0405, the STA departs as directed by the Shift Manager (SM).

• At 0410, the SM calls the Operations Manager to inform him of the reduction in crew composition.

• At 0420, the SM reaches a relief for the STA and directs him to come to work.

• At 0615, the STA relief arrives and joins the SM/CRS turnover.

• At 0645, the STA shift turnover briefing is completed.

Which one of the following describes the SM compliance with the shift manning requirements IAW ACP 1410.1, Conduct of Operations and Technical Specifications?

a. The shift manning requirements have been fully complied with because the STA function is ONLY required during Power Operation and Startup.

b. The shift manning requirements have NOT been fully complied with because the STA function was vacant for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. The shift manning requirements have been fully complied with because the relief STA received a complete turnover within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the previous STA departure.

d. The shift manning requirements have NOT been fully complied with because the Plant Managers permission must be obtained before shift staffing drops below minimum requirements.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 73 Exam Series A

Examination Outline Cross-Level RO SRO reference:

Tier # 3 Group # 1 K/A # 2.1.5 Importance Rating 3.9 Ability to use procedures related to shift staffing, minimum crew complement, overtime limitation, etc.

Proposed Question: SRO Question #100 Proposed Answer: B A: Incorrect - Per TS 5.2.2.c - ONLY 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is permitted for a shift staffing vacancy. Per ACP 1410.1 and TS the STA is required during Modes 1,2 and 3 B: Correct - Per TS 5.2.2.c - Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

Per ACP 1410.1 Section 3.2(3) - When the reactor is in other than COLD SHUTDOWN or REFUEL, the operations supervision team shall consist of at least three individuals. At any one time, there shall be at least one individual qualified to perform the OSM duties, at least one individual qualified to perform the CRS duties, and at least one individual qualified to perform the STA function on the operating crew.

C: Incorrect - The time limitation is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D: Incorrect - The Operations Manager permission is required not the Plant Manager ACP 1410.1 rev 71 Technical (Attach if not previously TS 5.2.2.c Reference(s): provided)

TS 5.2.2.g Proposed References to be provided to applicants during NONE examination:

Learning Objective: (As available)

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 74 Exam Series A

Question Source: Bank # DAEC Modified Bank (Note changes or attach

1. parent)

New Last NRC 2001 Question History:

Exam:

Question Cognitive Memory or Fundamental X

Level: Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 (2) Facility operating limitations in the technical specifications and their bases.

Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 75 Exam Series A

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 1 of 59 Usage Level Information Use Effective Date:

Approved for Point-of-Use printing IF NO DCFs are in effect for this procedure.

(on designated printers)

Record the following: Date / Time: __________________ / ______________

Printer ID: DA - ____________________ Initials: ________

NOTE: Per ACP 106.1, a copy of NG Form NG-019A (Working Copy Cover Page) shall be attached to the front of this document if active document use exceeds a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period as determined from the date and time recorded above.

Document approval signatures on file Prepared By: / Date:

Print Signature CROSS-DISCIPLINE REVIEW (AS REQUIRED)

Reviewed By: / Date:

Print Signature Reviewed By: / Date:

Print Signature PROCEDURE APPROVAL BY QUALIFIED REVIEWER Approved By / Date:

Print Signature

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 2 of 59 Table of Contents Page 1.0 PURPOSE ............................................................................................................................. 4 2.0 DEFINITIONS ........................................................................................................................ 4 3.0 INSTRUCTIONS ..................................................................................................................... 5 3.1 IMMEDIATE NOTIFICATION EVENTS ................................................................... 6 3.2 REPORTABLE EVENTS (WRITTEN NOTIFICATIONS)......................................... 8 3.2.1 LICENSEE EVENT REPORT (LER) .............................................................. 8 3.2.2 10 CFR 72 EVENT REPORT....................................................................... 13 3.2.3 SPECIAL REPORTS.................................................................................... 15 3.3 ROUTINE REPORTS ............................................................................................ 16 3.4 RETRACTION/CANCELLATION OF EVENT REPORTS...................................... 17 3.5 EVENT NOTIFICATION AND COMMUNICATION REQUIREMENTS.................. 18 4.0 RECORDS ............................................................................................................................ 19

5.0 REFERENCES

.................................................................................................................... 19 ATTACHMENT 1 NRC REPORT

SUMMARY

.......................................................................... 22 ATTACHMENT 2 REPORTABLE EVENTS .............................................................................. 30 ATTACHMENT 3 IMMEDIATE NOTIFICATION EVENTS........................................................ 36 ATTACHMENT 4 RPS ACTUATION REPORTING MATRIX ................................................... 45 ATTACHMENT 5 10 CFR 72 IMMEDIATE NOTIFICATION EVENTS ..................................... 46 ATTACHMENT 6 NOTIFICATION TO STATE/LOCAL OFFICIALS ......................................... 48 ATTACHMENT 7 COMMUNICATION INFORMATION CHECKLIST ....................................... 50 ATTACHMENT 8 COMMUNICATION TO THE DUTY STATION MANAGER.......................... 52

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 3 of 59 ATTACHMENT 9 COMMUNICATION TO THE NUCLEAR DIVISION DUTY OFFICER.......... 53 ATTACHMENT 10 COMMUNICATION FOR IMMEDIATE NOTIFICATION EVENT ............... 54 ATTACHMENT 11 COMMUNICATION FOR REPORTABLE EVENT...................................... 55 ATTACHMENT 12 COMMUNICATION FOR PLANT OPERATIONAL ISSUES ...................... 56 ATTACHMENT 13 COMMUNICATION FOR MEDICAL RESPONSE/ACCIDENT REPORTING................................................................................................................... 57 ATTACHMENT 14 NP-303 CHIEF NUCLEAR OFFICER REPORT OF REACTOR TRIP....... 59

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 4 of 59 1.0 PURPOSE This procedure provides guidance for the preparation, review and approval of various reports required by regulatory agencies. These reports include periodic and/or routine reports required by DAEC Technical Specifications, Title 10 of the Code of Federal Regulations, etc., and non-routine reports such as reportable events. Attachment 1 provides a summary of NRC required reports and cites the reporting requirements, preparer of report, recipient of report and method of report (telephone or written).

2.0 DEFINITIONS Action Request (AR) Form - A form which provides the mechanism for documenting the identification and evaluation of issues reported within the scope of FP-PA-RP-01.

Immediate Notification Event (INE) - An Immediate Notification Event is an incident that requires a 1, 4, 8, or 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> telephone notification as defined in 10 CFR 50.72, 10 CFR 20, 10 CFR 26, 10 CFR 72.74, 10 CFR 72.75 and 10 CFR 73. (See Section 3.1)

Licensee Event Report (LER) - A Licensee Event Report is a document which provides a mechanism for reporting, in writing to the NRC, the identification and evaluation of a Reportable Event as defined in 10 CFR 50.73, 10 CFR 71.95, and 10 CFR 73.71. (See Section 3.2.1) 10 CFR 72 Event Report - A document which provides in writing to the NRC, the identification and evaluation of a Reportable Event as defined in 10 CFR 72 (See section 3.2.2)

Non-Routine Reports - Reports that are submitted to the NRC due to a change in the normal routine of the plant.

Packaging - One or more receptacles or wrappers used for the transportation of radioactive material and their contents, excluding fissile material and other radioactive material, but including absorbent material, spacing structures, thermal insulation, radiation shielding devices for cooling and absorbing mechanical shock, external fittings, neutron moderators, non-fissile neutron absorbers, and other supplementary equipment.

Reportable Event - A Reportable Event is an incident that requires a written LER or 10 CFR 72 Event Report (or, in some cases a telephone report) as defined in 10 CFR 50.73, 10 CFR 71.95, 10 CFR 72.74, 10 CFR 72.75, and 10 CFR 73.71. Attachments 2, 3 and 5 provide a listing of events that are considered reportable.

Routine Reports - Reports that are required to be submitted to the NRC on a scheduled basis during the normal lifetime of the plant.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 5 of 59 Technical Specification (Tech Spec) Violation - Includes conditions prohibited by Tech Specs. For any event where actions are taken in accordance with Tech Spec action statements, a Tech Spec violation has not occurred unless specified time periods in Tech Specs are exceeded.

Valid Actuations - 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.

Invalid Actuations - Include actuations that are not the result of valid signals and are not intentional manual actuations. Invalid actuations include instances where instrument drift, spurious signals, human error, or other invalid signals caused actuation of the system (e.g.,

jarring a cabinet; error in use of jumpers or lifted leads; an error in actuation of switches or controls; equipment failure; or radio frequency interference).

3.0 INSTRUCTIONS NOTE Attachment 2 may be used to determine if an event is reportable. While the reportability of many events is self evident, some may not be readily apparent and the use of Engineering Judgment is necessary. Engineering Judgment may include either a documented engineering analysis or a judgment by a technically qualified individual, depending on the complexity, seriousness, and nature of the event or condition. A documented engineering analysis is not a requirement for all events or conditions, but it would be appropriate for particularly complex situations. In any case, the staff considers that the use of Engineering Judgment implies a logical thought process that supports the judgment. When applying Engineering Judgment, and there is doubt regarding whether to report or not, it is DAECs policy to make the report.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 6 of 59 3.1 IMMEDIATE NOTIFICATION EVENTS NOTE Attachment 3 provides a summary of events that require immediate notification to state, local and federal authorities. This attachment identifies the event, reporting requirements and DAEC individual(s) responsible for making the notification(s). An Immediate Notification Event may also be a Reportable Event. Attachment 4 provides a matrix for reporting actuations of the RPS system. Attachment 5 provides a matrix for 10 CFR Part 72 Immediate Notification events.

(1) Operations Shift Manager (OSM) shall ensure that Emergency Class Immediate Notification Events are reported to appropriate State and Local authorities within 15 minutes of the declaration of the event and/or determination of the Emergency Action Level (EAL), and the NRC immediately thereafter (and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of declaration of the event) as required by EPIP 1.2. Notification, Immediate Notification Events include:

(a) The declaration of any of the Emergency Action Levels listed in EPIP 1.1 (10 CFR 50.72(a)(1)(i), 10 CFR 72.75(a))

(b) Immediate follow-up reports for the following:

• Any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the emergency classes, if such a declaration has not been previously made.

• Any change from one emergency class to another.

• A termination of the emergency class.

• The results of ensuing evaluations or assessments of plant conditions.

• The effectiveness of response or protective measures taken.

• Information related to plant behavior that is not understood. (50.72(c)).

(2) Notification to the NRC shall be made via the Federal Telecommunications System (FTS-2001). If the FTS-2001 is inoperative, the notification shall be made by any other method which will ensure that a report is made as soon as practical (see EPIP 1.2). The Event Notification Worksheet (NRC Form 361) provides guidance on the type of information that should be provided to the NRC Operations Center.

(3) For 4, 8, and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> NRC notifications, the draft Event Notification Worksheet (NRC Form 361), shall be reviewed and approved by either Plant General Manager (PGM) or Site Vice President (SVP). For 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notifications, PGM or SVP approval should be obtained if time permits.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 7 of 59 (4) Non-emergency class Immediate Notification Events shall be reported to the NRC via the FTS-2001 by telephone within 1, 4, 8, or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of occurrence depending on type of event and reporting requirement (see Attachments 3 and 5).

(5) Internal notifications shall be made to DAEC management in accordance with PI-AA-204, Condition Identification and Screening Process.

(6) An Action Request (AR) shall be prepared for Immediate Notification Events per PI-AA-204. For Fitness For Duty (FFD) events, an AR is not required and notifications should be made in accordance with Security Directives.

(7) All Security-related reports identified in 10 CFR 73.71 and 10 CFR 72.74 or in attachments to this procedure shall only be made with the approval/concurrence of the Security Manager or designee via the FTS-2001. Security-related event notifications shall be made in accordance with Security Procedures.

(8) The Licensing Manager shall ensure ARs and Security-related Immediate Notification Events are reviewed to determine if a Reportable Event has occurred. If a Reportable Event has occurred, the Licensing Manager shall ensure that an LER or 10 CFR 72 Event Report is generated as required.

(9) For FPL Energy Duane Arnold security contacts to off-site government agencies for investigating a suspicious vehicle, person, aircraft, or a related event, the FPL Energy Duane Arnold security management will determine if a courtesy call to the NRC is necessary. These calls do not require a 4-hour Immediate Event Report under 10 CFR50.72 (b)(2)(xi).

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 8 of 59 3.2 REPORTABLE EVENTS (WRITTEN NOTIFICATIONS) 3.2.1 LICENSEE EVENT REPORT (LER)

NOTE Section 50.73 requires submittal of an LER within 60 days after the discovery of a reportable event. Many reportable events are discovered when they occur. However, if the event is discovered at some later time, the discovery date is when the reportability clock starts under 10 CFR 50.73. Discovery date is generally the date when the event was discovered rather than the date when an evaluation of the event is completed. For example, if a technician sees a problem, but a delay occurs before an engineer or supervisor has a chance to review the situation, the discovery date (which starts the 60-day clock) is the date the technician sees a problem.

In some cases, such as discovery of an existing but previously unrecognized condition, it may be necessary to undertake an evaluation in order to determine if an event or condition is reportable. If so, the guidance provided in Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability which applies primarily to operability determinations, is appropriate for reportability determinations as well. This guidance indicates that, whenever reasonable expectation that the equipment in question is operable no longer exists, or significant doubts begin to arise, appropriate actions, including reporting, should be taken.

(1) An LER (NRC Form 366) shall be prepared by the Licensing Department and submitted to the NRC within 60 days after discovery and/or classification as reportable, for the following events. Unless otherwise specified, only those events which occurred within 3 years of the date of discovery are reportable:

(a) The completion of any plant shutdown required by the plants Technical Specifications. (50.73(a)(2)(i)(A))

(b) Any operation or condition prohibited by the plant's Technical Specifications, except when:

(i) The Technical Specification is administrative in nature; (ii) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (iii) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event. (50.73(a)(2)(i)(B).

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 9 of 59 (c) Any operation or condition prohibited by the DAEC operating license. (Administrative Requirement NG-91-4028)

(d) Any deviation from Tech Specs authorized pursuant to 10 CFR 50.54(x).

(50.73(a)(2)(i)(C))

(e) Any event or condition that resulted in:

(i) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (ii) The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. 50.73(a)(2)(ii)

(f) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the plant. (50.73(a)(2)(iii))

NOTE Excess Flow Check Valves (XFVs) have, in the past tripped when returning instruments to service or performing instrument valve manipulations. Unless in response to an actual system leak, XFV trips as described above are not considered reportable under the following system actuation criteria.

(g) Any event or condition that resulted in manual or automatic actuation of any of the specific plant systems listed in (h) below, except when:

1. The actuation resulted from and was part of a preplanned sequence during testing or reactor operation; or

2. The actuation was invalid and:

a. Occurred while the system was properly removed from service; or

b. Occurred after the safety function had been already completed.(50.73(a)(2)(iv)(A).

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 10 of 59 NOTE 10CFR50.73(a)(1) allows a 60-day telephone report to be made (instead of a written LER) for invalid actuations of any of the following systems except for RPS actuations when the reactor is critical.

(h) 10CFR50.73(a)(2)(iv)(B) lists 9 types of systems for both PWR and BWR reactor plants. The following list of DAEC specific systems and system modes of operation is provided to define the plant systems to which this reporting requirement applies at DAEC:

(i) RPS*

(ii) PCIS affecting valves in more than one system or more than one MSIV (iii) HPCI (iv) ADS (v) RHR-LPCI (vi) Core Spray (vii) RCIC (viii) SBDG(s)

(ix) RHR-Drywell Sprays (x) RHR-Torus Sprays (xi) RHR-Torus Cooling (xii) Drywell Cooling (xiii) RHRSW**

(xiv) ESW**

(xv) RWS**

• See attachment 4 to this procedure for a summary table of RPS actuation reporting.

• • only applicable to 10 CFR 50.73

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 11 of 59 NOTE An unplanned inoperable condition or LCO entry for the RCIC system is not reportable pursuant to 10CFR50.73(a)(2)(v) or its related 10CFR50.72(b)(3)(v) requirement. (Reference 23)

Events covered in paragraph (i) below 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 to paragraph 50.73(a)(2)(v) if redundant equipment in the same system was operable and available to perform the required safety function. (50.73(a)(2)(vi))

(i) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

• Shut down the reactor and maintain it in a safe shutdown condition;

• Remove residual heat;

• Control the release of radioactive material; or

• Mitigate the consequences of an accident. (50.73(a)(2)(v)).

(j) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

• Shut down the reactor and maintain it in a safe shutdown condition;

• Remove residual heat;

• Control the release of radioactive material; or

• Mitigate the consequences of an accident. (50.73(a)(2)(vii))

(k) Any airborne radioactivity release that, when averaged over a time period of one hour, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in Appendix B to Part 20, table 2, column 1. (50.73(a)(2)(viii)(A))

(l) Any liquid effluent release that, when averaged over a period of one hour, exceeds 20 times the applicable concentrations specified in Appendix B to Part 20, table 2, column 2 at the point of entry into the receiving waters (i.e. unrestricted area) for all radionuclides except tritium and dissolved noble gases. (50.73(a)(2)(viii)(B))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 12 of 59 (m) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

• Shut down the reactor and maintain it in a safe shutdown condition;

• Remove residual heat;

• Control the release of radioactive material; or

• Mitigate the consequences of an accident. (50.73(a)(2)(ix)(A)).

(n) Events covered in paragraph (m) above may include cases of procedural error, equipment failures, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy However, an event is not required to be reported under this specific criterion if the event results from:

• A shared dependency among trains or channels that is a natural and expected consequence of the approved plant design; or

• Normal and expected wear or degradation.(50.73(a)(2)(ix)(B).

(o) Any event that posed an actual threat to the safety of the plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the plant including fires, toxic gas releases, or radioactive releases.

(50.73(a)(2)(x))

(p) Any event which meets the one-hour reportability criteria of 10 CFR 73.71, as detailed in Security Procedure 11. (Safeguards) (See Attachments 2 and 3.)

NOTE Per 10 CFR 73.71, duplicate reports are not required for events that are also reportable in accordance with 10 CFR 50.72 and 10 CFR 50.73.

(2) Written Licensee Event Reports shall be submitted to the NRC on the "Licensee Event Report" form (NRC Form 366) in accordance with 10 CFR 50.73(b) and NUREG 1022.

(3) All written LERs shall be reviewed by the On-Site Review Group and the Plant Manager prior to NRC submittal.

(4) All written LERs shall be reviewed by the Safety Committee. (This review is usually after the LER has been mailed.). LERs reported via a 60-day phone call under 50.73 (a)(2)(iv),

(invalid actuations) do not require Safety Committee review.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 13 of 59 (5) LERs reported via a 60-day phone call under 50.73 (a)(2)(iv), (invalid actuations), may be called in using NRC Form 361, and do not require On-Site Review Group or Plant Manager reviews.

(6) Security-related LERs are still required to be submitted within 60 days and shall be stamped "Safeguards Information," if they contain such information.

3.2.2 10 CFR 72 EVENT REPORT NOTE Section 72.75 requires submittal of a written report within 60 days after the discovery of a reportable events (b)(1), (c)(1), (c)(2), and (d)(1). Many reportable events are discovered when they occur. However, if the event is discovered at some later time, the discovery date is when the reportability clock starts under 10 CFR 72.75. Discovery date is generally the date when the event was discovered rather than the date when an evaluation of the event is completed. For example, if a technician sees a problem, but a delay occurs before an engineer or supervisor has a chance to review the situation, the discovery date (which starts the 60 day clock) is the date the technician sees the problem.

In some cases, such as discovery of an existing but previously unrecognized condition, it may be necessary to undertake an evaluation in order to determine if an event or condition is reportable. Whenever reasonable expectation that the equipment in question is operable no longer exists, or significant doubts begin to arise, appropriate actions, including reporting, should be taken.

Written reports prepared pursuant to other regulations may be submitted to fulfill the Part 72 reporting requirement if the reports contain all the necessary information and the appropriate distribution is made.

Reports required under 10 CFR 73.71 need not be duplicated under requirements of 10 CFR 72.74.

(1) A written report shall be prepared by the Licensing Department and submitted to the NRC within 60 days after discovery and/or classification as reportable for the following events:

(a) A defect in any storage structure, system, or component which is important to safety.

(b) A significant reduction in the effectiveness of any storage confinement system during use.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 14 of 59 (c) An action taken in an emergency that departs from a condition or technical specification contained in a license or certificate of compliance issue under 10CFR72 when the action is immediately needed to protect the public health and safety and no action consistent with license or certificate of compliance conditions or technical specifications that can provide adequate or equivalent protection is immediately apparent.

(d) An event in which important to safety equipment is disabled or fails to function as designed when:

(i) The equipment is required by regulation, license condition, or certificate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident; and (ii) No redundant equipment was available and operable to perform the required safety function.

(2) Written reports must be sent to the Commission in accordance with 10 CFR 72.4. These reports must include the following:

(a) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence; (b) A clear, specific, narrative description of the event that occurred so that knowledgeable readers conversant with the design of the ISFSI, but not familiar with the details of a particular facility, can understand the complete event. The narrative description must include the following specific information as appropriate for the particular event:

(i) ISFSI operating conditions before the event; (ii) Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event; (iii) Dates and approximate times of occurrences; (iv) The cause of each component or system failure or personnel error, if known; (v) The failure mode, mechanism, and effect of each failed component, if known; (vi) A list of systems or secondary functions that were also affected for failures of components with multiple functions;

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 15 of 59 (vii) The method of discovery of each component or system failure or procedural error; (viii) For each human performance related root cause, discuss cause(s) and circumstances.

(c) The manufacturer and model number (or other identification) of each component that failed during the event; (d) The quantities, and chemical and physical forms of the spent fuel involved; (e) An assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or components that could have performed the same function as the components and systems that failed during the event; (f) A description of any corrective actions planned as a result of the event, including those to reduce the probability of similar events occurring in the future; (g) Reference to any previous similar events at the same facility that are known to the licensee; (h) The name and telephone number of a person within the licensees organization who is knowledgeable about the event and can provide additional information concerning the event and the facilitys characteristics; (i) The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by name.

(3) These written reports shall be reviewed by the On-Site Review Group and Plant Manager prior to NRC submittal.

(4) The written reports shall be reviewed by the Safety Committee. (This review is usually after the report has been mailed.)

(5) Security-related reports are required to be submitted within 60 days and shall be stamped Safeguards Information, if they contain such information.

3.2.3 SPECIAL REPORTS (1) Special reports shall be submitted in accordance with 10 CFR 50.4. These reports shall be submitted covering the activities identified below pursuant to the applicable referenced requirement.

(a) Reactor vessel base, weld and heat affected zone metal test specimens (10 CFR 50, Appendix H(IV)).

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 16 of 59 (b) Inservice Inspection Program (10 CFR 50.55a(g)).

(c) Off-Gas System inoperable (ODAM Section 6).

(d) Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of ODAM Table 6.3-3 when averaged over any calendar quarter sampling period (ODAM OLCO 6.3.2.B). Submit the report within 30 days after discovery. This condition also warrants the following additional actions:

(i) Notification of State and Local Officials as directed by Attachment 6 and in compliance with the requirements of Nuclear Fleet Guideline, EV-AA-100-1000, Ground Water Protection Program Communications/Notification Plan.

(ii) Forward a copy of the special report to the State and Local Officials listed on Attachment 6.

(e) Annual dose to a member of the public determined to exceed 40 CFR Part 190 dose limit (ODAM Section 6).

(f) Radioactive liquid waste released without treatment when activity concentration exceeds 0.01 mci/ml (ODAM Section 6).

(g) Post Accident Monitoring Instrumentation inoperability (TS 3.3.3.1).

3.3 ROUTINE REPORTS (1) Provide to the NRC, using an industry database, the operating data (for each calendar month) that is described in Generic Letter 97-02 (Reference 34) by the last day of the month following the end of each calendar quarter. {C001}

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 17 of 59 (2) The following Routine Reports shall be initiated when appropriate:

• Startup

• Annual Radioactive Materials Release Report

• Individual Exposure Monitoring

• Transfer of Source Material

• Receipt of Source Material

• Source Material Inventory

• Summary of Changes, Tests and Experiments

• Annual SV and SRV Challenges and Failures

• Special Nuclear Materials Status

• Transfer of Special Nuclear Material

• Receipt of Special Nuclear Material

• Fracture Toughness

• Reactor Vessel Material Surveillance

• Containment Leak Rate Test

• Annual Exposure

• Annual Radiological Environmental Report

• Quarterly Security Event Log Submittal 3.4 RETRACTION/CANCELLATION OF EVENT REPORTS (1) An event notification can be retracted using the same procedural steps by which the initial report was made. The Retraction/Cancellation of Event Reports worksheet (NG-172K) has been developed to provide guidance on actions taken to retract reported events.

(2) Cancellation of events shall be made by the OSM (or his designee) upon direction from the Licensing Manager or designee, via the FTS-2001. If the FTS-2001 is inoperative, the notification shall be made by any other method which will ensure that the cancellation is made as soon as practical.

(3) Sound, logical bases for the retraction/cancellation shall be communicated with the notification.

(4) Cancellations of submitted LERs and written 10 CFR 72 Event Reports should be made by letter. The bases for the cancellation shall be explained. The notice of cancellation will be filed and stored with the original report. If the cancellation only involves a 60 day telephone report LER pursuant to 10CFR50.73(a)(2)(iv), then a telephone retraction is appropriate.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 18 of 59 3.5 EVENT NOTIFICATION AND COMMUNICATION REQUIREMENTS (1) The OSM/DSM should collect information on Attachment 7 Communication Information Checklist as plant conditions allow as soon as an event has been determined to have occurred. Recorded relevant questions or comments during communication in the comment section of Attachment-7.

(2) For any events that may require activation of the Event Response Team (ERT) per ACP 114.9, Event Response Procedure, the DSM shall be contacted with information from Attachment 7 and Attachment 8 Communication to the Duty Station Manager.

(3) The OSM/DSM shall communicate to the Nuclear Division Duty Officer (NDDO) per Nuclear Policy NP-303 for the events listed in Attachment-9 Communication to the Nuclear Division Duty Officer as soon as plant conditions allow.

(4) If an Immediate Notification Event (INE) has been determine to have occurred, the immediate notification will be performed per Section 3.1 of this procedure. Internal communication should be performed as plant conditions allow per Attachment 10 Communication for Immediate Notification Event with the exception for Emergency Action Levels. The prompt notification system will provide the necessary internal communication for Emergency Action Levels.

(5) If a Reportable Event has been determine to have occurred, the notification will be performed per Section 3.2 of this procedure. Internal communication should be performed as plant conditions allow per Attachment 11 Communication for Reportable Event.

(6) If a Plant Operational Issue has been determine per ACP 114.13 Duty Station Manager to have occurred, verify they do not meet the notification requirements of an INE or Reportable Event. Internal communications should be performed as soon as plant conditions allow per Attachment 12 Communication for Plant Operational Issue.

(7) For medical response and employee injuries, notification shall be made in accordance with fleet procedure SA-AA-100-1000.

(8) For Fitness for Duty (FFD) and Security Events, the On-shift Security Lieutenant shall be contacted and reference appropriate site Security Procedures to determine appropriate notification and internal communications requirement.

(9) For chemical and oil spills, contact the Hazardous Waste Emergency Coordinator (HWEC) and reference ACP 1411.14, Chemical/Oil Spill Response procedure to determine appropriate notification and internal communications requirement.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 19 of 59 (10) If any condition resulted in an unplanned reactor trip, information on Attachment 14 NP-303 Chief Nuclear Officer Report of Reactor Trip must be sent or communicated to the Chief Nuclear Officer within eight (8) hours of the reactor trip. This information must be signed by the site Vice President 4.0 RECORDS (1) All Quality Assurance records generated by this ACP shall be kept in accordance with ACP 115.1.

(2) Records of internal communications are not Quality Assurance records. Records of internal communications should be attachment to the parent Corrective Action for which internal communication was initiated to address the event.

5.0 REFERENCES

(1) Technical Specifications, "Appendix A to Operating License DPR-49, Technical Speci-fications and Basis for the Duane Arnold Energy Center" (2) Technical Specification, "Operating License DPR-49 for the Duane Arnold Energy Center, Docket No. 50-331" (3) Reg. Guide 10.1, "Compilation of Reporting Requirements for Persons Subject to NRC Regulations" (4) NUREG-1022, Revision 2, Event Reporting Guidelines (5) Federal Register Vol. 65, No. 207 dated October 25, 2000.0 (6) 10 CFR 50.72 (7) 10 CFR 50.73 (8) Emergency Plan Implementing Procedures (EPIP) 1.1 and 1.2 (9) ACP 115.1 (10) Security Procedure 11, "Reporting of Physical Security Events" (11) Reg. Guide 5.62, Rev. 1, Nov. 1987 (12) NUREG 1304, dated Feb. 1988 (13) 10 CFR 71.95

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 20 of 59 (14) 10 CFR 72.11 (15) 10 CFR 72.74 (16) 10 CFR 72.75 (17) 10 CFR 72.76 (18) 10 CFR 72.78 (19) 10 CFR 72.80 (20) 10 CFR 72.212 (21) 10 CFR 73.71 (22) EPIP 2.3, Operation of FTS-2001 Telephone Network (23) 10 CFR 50, Appendix E, IV E (9) (d)

(24) 10 CFR 20 (25) DAEC Fire Plan (26) AR 95-0861.01, AR 96-1339, AR 96-1674 (27) NG-96-1744 (28) NRC Information Notice 97-15 (29) AR 14546 (30) NRC IN 83-10 (31) RIS 2001-14, AR 26803 (32) CAP 026817 (33) AR OTH028213 (34) {C001} Generic Letter 97-02, Revised Contents of the Monthly Operating Report (35) TS Amendment 256 (36) CA43124

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 21 of 59 (37) AR CAP 44393 (38) AR CA044679 (39) CAP046161, CAP048309, OTH017116, OTH018170 (40) Nuclear Fleet Guideline EV-AA-1000, Ground Water Protection Program Communications/Notification Plan (41) CAP066431, PCR052276

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 22 of 59 ATTACHMENT 1 Page 1 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

1. Individual radiation exposure Sec. 19.13(c) Within 30 days of request W Individual MPA(1) Radiation data to former workers or determination of Protection exposure Manager

2. Individual radiation exposure Sec. 19.13(d) At time of transmittal to W Individual None Radiation data to worker reported to NRC Protection NRC under 20.2202, 20.2203, Manager 20.2204, or 20.2206

3. Radiation exposure data to Sec. 19.13(e) At termination upon W Individual MPA(1) Radiation terminating workers request of worker Protection Manager

4. Respiratory protection program Sec. 20.1703(d) 30 days prior to use of W RO(1) DCD(1) Radiation equipment Protection Manager

5. Report of excessive Sec. Immediately P,T OP CTR Final OSM radioactive contamination on 20.1906(d)(1) delivering radioactive material on receipt carrier

6. Report of excessive radiation Sec. Immediately P,T OP CTR Final OSM levels external to the package 20.1906(d)(2) delivering on receipt carrier

7. Report on investigation tracing Sec. 20.2006(d) 2 weeks after W RO None Radiation Radwaste shipment for which and App. G, investigation completed Protection Acknowledgment of Receipt Section III, Manager not received Paragraph E.2

8. Theft or loss of licensed Sec. Immediately P OP CTR None OSM material 1000 x App. C to 20.2201(a)(i) 20.1001-20.2401

9. Theft or loss of licensed Sec. 30 days P,T OP CTR None OSM material 10 x App. C to 20.2201(a)(ii) 20.1001-20.2401

10. Theft or loss of licensed Sec. 20.2201(b) 30 days W RO(1) Licensee(1) Radiation material Protection Manager

11. Additional information on theft Sec. 20.2201(d) Within 30 days of receipt W RO(1) Licensee(1) Radiation or loss information. of information Protection Manager

12. Report of incident Sec. 20.2202(a) Immediately P,T OP CTR RO(1) OSM See report

1. 3

13. Report of incident Sec. 20.2202(b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> P,T OP CTR RO(1) OSM See report

1. 3

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 23 of 59 ATTACHMENT 1 Page 2 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

14. Reports of exposures of Sec. 30 days W DCD(1) RO(1); See Radiation individual, radiation levels, and 20.2203(a) report #3 Protection concentrations of radioactive Manager material exceeding the limits (See Attachment 2)

15. Report of planned special Sec. 20.2204 30 days W RO(1) See report #3 Radiation exposure Protection Manager

16. Reports of individual Sec. 20.2206 & Annually, covering W REIRS(1) Each Radiation monitoring 19.13(b) the preceding year exposed Protection 19.13(d) on or before April 30 worker Manager

17. Failure to comply or existence Sec. 21.21(b) 2 days P,T NMSS, NRR None Chairman Part of a defect or RO 21 Evaluation Committee

18. Failure to comply or existence Sec. 21.21(b) 5 days W NMSS or NRR DCD(1) Chairman Part of a defect (3) 21 Evaluation Committee

19. Failure of or damage to Sec. 31.5(c))(5) 30 days W RO(1) DCD(1) Radiation shielding, on-off mechanism or Protection indicator; detection of Manager removable radioactive material

20. Transfer of device to specific Sec. 31.5(c)(8) 30 days W NMSS(1) None Radiation licensee Protection Manager

21. Transfer of device to general Sec. 31.5 30 days W NMSS(1) None Radiation licensee (c)(9)(i) Protection Manager

22. Registration of general Sec. 40.25 30 days after first W NMSS(1) RO(1) Reactor licensee who receives, (c)(1) receipt Engineering acquires, possesses, or uses Supervisor depleted uranium

23. Change to registration Sec. 40.25 30 days W NMSS(1) RO(1) Reactor (c)(2) Engineering Supervisor

24. Registration certificate-filed by Sec. 40.25 Promptly (1) W Transferee DCD(1) Reactor transferor (d)(3) Engineering Supervisor

25. Registration certificate-transfer Sec. 40.25 30 days W NMSS(1) RO(1) Reactor (d)(4) Engineering Supervisor

26. Transfer of material licensed Sec. 40.35(d) Promptly W Receiver None Reactor under Sec. 40.25 (1) Engineering Supervisor

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 24 of 59 ATTACHMENT 1 Page 3 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

27. Transfer of material licensed Sec. 40.35 (e)(1) Quarterly W NMSS(1) None Reactor under Sec. 40.25 Engineering Supervisor

28. Transfer of devices under Sec. 40.35 (e)(2) Quarterly W State DCD(1) Reactor Agreement State regulations Agency* Engineering equivalent to Sec. 40.25 Supervisor

29. Reports required as Sec. 40.41 (e)(4) Specified in license Specified in Reactor conditions of Part 40 license condition license Engineering Supervisor

30. Nuclear Material Transaction Sec. 40.64(a) Promptly W DOE(1) Receiver(3) Reactor Report Form DOE/NRC-741 Engineering filed by shipper Supervisor

31. Nuclear Material Transaction Sec. 40.64(a) 10 days after W DOE(1) Shipper(1) Reactor Report Form DOE/NRC-741 Engineering filed by receiver Supervisor

32. Statement of source material Sec. 40.64(b) Annually W DOE(1) None Reactor inventory Engineering Supervisor

33. Unlawful diversion of source Sec. 40.64(c) Promptly P,T RO None OSM material

34. Unlawful diversion of source Sec. 40.64(c) 15 days W RO(1) NMSS(1) Reactor material Engineering Supervisor

35. Identify information having a Sec. 50.9(b) 2 working days of RO None significant implication for identification public health and safety or Sec. 72.11(b) 2 working days of RO None common defense and identification security

36. Effluent releases report Sec. 50.36a Annually W DCD(1) RO(1) Radiation (a)(2), Tech Resident(1) Protection Specs Manager

37. Loss-of-Coolant Accident Sec. 50.46(a)(3) Annually (non-significant) W DCD(1) RO(1) Licensing Evaluation model changes Resident (1) Manager or errors report 30 days (significant) W NRR (50.73 RO(1) Licensing DCD (73.71) Resident(1) Manager

38. Changes in security plan Sec. 50.54(p) Two months after change W DCD RO(1) Security made without prior approval Manager

• Responsible Agreement State Agency

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 25 of 59 ATTACHMENT 1 Page 4 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

39. Changes in emergency plan Sec. 50.54(q) 30 days after change W DCD(1) RO(2) Emergency made without prior approval or proposed to NRC Resident(1) Planning Manager Sec 72.44(f) 6 months after change W DCD(1) RO(2) Emergency Resident(1) Planning NMSS Manager

40. Filing for bankruptcy under Sec. 50.54(cc) Immediate W RO None Legal Chapter 11 Sec. 72.44(b)(6)(i) Immediate W RO None Legal

41. Facility changes, tests, and Sec. 50.59(b) 6 months after W DCD(1) RO(1) Licensing experiments conducted without Refueling Outage not Resident(1) Manager prior approval to exceed 24 months Sec.72.48(d)(2) Once every 24 months W DCD(1) RO(1) Licensing Resident(1) Manager

42. Financial report Sec. 50.71(b) Annually W DCD(1) RO(1) Licensing Resident(1) Manager 72.80(b) Annually W DCD(1) RO(1) Licensing Resident(1) Manager

43. FSAR updating Sec. 50.71(e) 6 months after RFO W NRR(11) RO(1) Licensing not to exceed 24 Resident(1) Manager months

44. Emergency Notifications Part 50, App. E, 15 minutes P S&L Gov.** NRC OSM Sec.IV.D.3 Sec. 72.75(a) Prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) P S&L Gov. ** OP CTR OSM

45. Immediate Notification Events Sec. 50.72 Prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) P OP CTR None OSM (Non-Emergency)

46. Immediate Notification Events Sec. 50.72 Prompt (4 or 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) P OP CTR None OSM

47. Non-emergency Notifications Sec. 72.75(b)(1-2) Prompt (4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) P OP CTR None OSM Sec. 72.75(c)(1-3) Prompt (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) P OP CTR None OSM Sec.72.75(d)(1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> P OP CTR None OSM

48. Licensee Event Report Sec. 50.73 60 days W or P DCD/ RO(1) Licensing OP CTR Manager Sec. 73.71 60 days W DCD RO(1) Licensing SFPO Manager NSIR

49. 10 CFR 72 Event Report Sec. 72.75(g) 60 days W DCD RO(1) Licensing Manager

• • State and Local Government

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 26 of 59 ATTACHMENT 1 Page 5 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

50. Report on status of Sec. 50.75(f)(1) On a calendar year basis W DCD RO(1) Licensing Decommissioning Funding by March 31,1999 and at Resident Manager least once every 2 years thereafter

51. Fracture toughness Part 50, App. G At least 3 years prior to W DCD(1) RO(1) Program date when the predicted Resident(1) Engineering fracture toughness levels Manager will no longer satisfy requirements of Appendix G

52. Report of test results of Part 50, App. H Variable W DCD(1) RO(1) Program specimens withdrawn from Sec. III.A, Tech Resident(1) Engineering capsules (fracture Specs Manager toughness tests)

53. Report of effluents released Part 50, App. I., Within 30 days from end of W RO(1) DCD(1) Radiation in excess of design Sec. IV.A. quarter Protection objectives Manager

54. Reactor containment Part 50, App.J, 3 months after conducting W Available System building integrated leak rate Sec. V.B, test onsite Engineering test (includes LLRT) Manager Summary Report

55. Notification of disability Sec. 55.25 30 days W NRR None Operations Manager

56. Medical examination Sec. 55.21 -- W Licensee*** None Manager, Training Coordinator RO

57. Accidental Criticality or Loss Sec. 70.52 Prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) P OP CTR None OSM of Special Nuclear Material Sec. 72.74 Prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) P OP CTR None OSM Sec. 73.71 Prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) P OP CTR None OSM

58. Material Status Report Sec. 74.13 Within 60 days of the W NMSS Licensee Reactor beginning of the physical Engineering inventory Supervisor

59. Nuclear Material Transaction Sec. 74.15 Upon transfer or receipt W NMSS Licensee Reactor Reports Engineering Supervisor Sec. 72.78(a) Upon transfer or receipt W NMSS Licensee Reactor Engineering Supervisor

60. Reduction in Effectiveness Sec. 71.95 30 days W NMSS None Radiation of Package Protection Manager

• • • Per regulation, physician is to send original copy to DAEC.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 27 of 59 ATTACHMENT 1 Page 6 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

61. 72.4 Notifications Sec. Notify NRC 90 days prior to first W DCD RO Licensing 72.212(b)(1)(i) storage of spent fuel in cask Manager type under general license Sec. Register use of each cask no W DCD RO Licensing 72.212(b)(1)(ii) later than 30 days after using Manager cask to store spent fuel

62. Proof of financial Sec. 140.15(a) As required W NRR or None Legal protection NMSS(3)

63. Change in proof of Sec. 140.15(e) Promptly W NRR or None Legal financial protection NMSS(2)

64. Financial statement Sec. 140.15 Annually W DCD NMSS(3) Licensing (b)(1) RO(1)

Resident(1)

Sec.72.80(b) Annually W DCD RO(1) Licensing Resident(1)

65. Policy renewal Sec. 140.17(b) 30 days prior to termination of W NRR or None Legal termination of policy policy NMSS(1)

66. Guarantee of payment Sec. 140.21 Annually W NRR or None Legal of deferred premiums NMSS(1)

67. Transfer of assets >1% Sec 50.33(k) As required W NRR None Legal of net utility value

68. Startup of Reactor Tech Specs Within (1) 90 days following W RO(2) Licensee(36) Licensing completion of the startup test Manager program (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If all three events are not completed, supplementary reports every 3 months.

69. Not Used

70. Annual Radiological TS 5.6.2 Annually, by May 1 W RO(1) DCD(18) Radiation Environmental Protection Operating Report Manager

71. Not Used

72. Core Operating Limits TS 5.6.5 Upon Issuance W DCD(1) RO(1) Licensing Report Resident(1) Manager

73. Annual Radioactive TS 5.6.3 Annually, by May 1 W RO(1) DCD(18) Radiation Material Release Report Protection Manager

74. PAM Instrumentation TS 3.3.3.1 14 days W DCD RO(1) Licensing Inoperability Manager

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 28 of 59 ATTACHMENT 1 Page 7 of 8 NRC REPORT

SUMMARY

Primary Secondary Responsible Report Required by Timing Method Recipient Recipient Notifier

75. Low Level Waste NRC GL 91-02 30 Days W LWM None Radiation Mishaps Protection Manager

76. ISI Summary Report ASME Section Within 90 days of completion of W DCD(1) RO(1) Licensing XI, IWA-6230 ISI examinations during Resident(1) Manager refueling outages

77. Horizontal Storage ISFSI-61BT TS 30 Days W DCD(1) RO(1) Licensing Module Dose Rates 1.2.7 Resident(1) Manager Exceeded SFPO

78. Transfer Cask Dose ISFSI-61BT TS 30 Days W DCD(1) RO(1) Licensing Rates 1.2.11 Resident(1) Manager SFPO

79. Highest Heat Load to ISFSI-61BT TS 30 Days W DCD RO(1) Licensing Date of any 61BT Dry 1.1.7 Resident(1) Manager Storage Canister**** SFPO

80. Claim of Personnel Sec. 140.6 As promptly as practical W NRR NMSS Licensing Injury or Property Manager Damage

81. NRC Form 748 National 10CFR20.2207 Annually by January 31 W,T LM None Radiation Source Tracking Protection Transaction Report Manager

• • • • Only required to be performed on the DSC that has the highest heat load of all DSCs in use to date.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 29 of 59 ATTACHMENT 1 Page 8 of 8 NRC REPORT

SUMMARY

ABBREVIATIONS AND CODES Reporting Methods P Telephone T Telegraph W Written Report Number of Copies - The number of copies of each report is specified by numerals in parentheses under the headings "Primary Recipient" and "Secondary Recipient".

Recipients DCD Document Control Desk DOE U.S. Department of Energy U.S. Nuclear Regulatory Commission P.O. Box E Mail Station 0-P1-17 (zero-P1-17) Oak Ridge, TN 37830 Washington, D.C. 20555 EDO Executive Director for Operations GC General Counsel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 IE Director, Office of Inspection and Enforcement SFPO Director, Spent Fuel Project Office U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 ATTN: Document Control Desk IP Assistant Director, Export-Import and International NSIR Director, Division of Nuclear Security Safeguards Office of Nuclear Security and Incident Office of International Programs Response U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 MPA Director, Office of Nuclear Regulatory Research LM Lockheed Martin Formatted: Left U.S. Nuclear Regulatory Commission NSTS Help Desk Washington, D.C. 20555 30 West Gude Drive, Suite 300 Rockville, MD 20850 Fax: 240-403-4391 NMSS Director, Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRR Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 OP CTR U.S. NRC Operations Center REIRS Project Manager Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 RO Appropriate NRC Regional Office (see Appendix D to Part 20 or Appendix A to Part 73)

SEC Director, Division of Security U.S. Nuclear Regulatory Commission Washington, D.C. 20555 LWM Director, Division of Low-Level Waste Management and Decommissioning U.S. Nuclear Regulatory Commission Washington, D.C. 2055 (301)492-3339

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 30 of 59 ATTACHMENT 2 Page 1 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT The completion of any plant shutdown required by Tech. Specs. YES, upon

[50.73(a)(2)(i)(A)] initiation of a shutdown Any operation or condition prohibited by Tech. Specs. [50.73(a)(2)(i)(B)] NO Any deviation from Tech. Specs. authorized pursuant to 10 CFR 50.54(x). YES

[50.73(a)(2)(i)(C)]

Any event or condition that resulted in the condition of the plant, including its YES principal safety barriers, being seriously degraded, or that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety. [50.73(a)(2)(ii)]

Any natural phenomenon or other external condition that posed an actual NO threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant. [50.73(a)(2)(iii)]

Any event or condition that resulted in a manual or automatic actuation of YES, for all any of the systems listed in paragraph (a)(2)(iv)(B) (DAEC specific list valid actuations provided in section 3.2.1 of this procedure) , except when: and an invalid RPS trip when critical (A) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (B) The actuation was invalid and;

1. Occurred while the system was properly removed from service; or

2. Occurred after the safety function had been already completed.

[50.73(a)(2)(iv)(A)] See Attachment 4 for RPS Actuations

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 31 of 59 ATTACHMENT 2 Page 2 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT

• Any event or condition that could have prevented the fulfillment of the YES safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition. [50.73(a)(2)(v)(A)]

• Any event or condition that could have prevented the fulfillment of the YES safety function of structures or systems that are needed to remove residual heat. [50.73(a)(2)(v)(B)]

• Any event or condition that could have prevented the fulfillment of the YES safety function of structures or systems that are needed to control the release of radioactive material. [50.73(a)(2)(v)(C)]

• Any event or condition that could have prevented the fulfillment of the YES safety function of structures or systems that are needed to mitigate the consequences of an accident. [50.73(a)(2)(v)(D)]

Any event where a single cause or condition caused at least one NO independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to shut down the reactor and maintain it in a safe shutdown condition. [50.73(a)(2)(vii)(A)]

Any event where a single cause or condition caused at least one NO independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to remove residual heat. [50.73(a)(2)(vii)(B)]

Any event where a single cause or condition caused at least one NO independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to control the release of radioactive material.

[50.73(a)(2)(vii)(C)]

Any event where a single cause or condition caused at least one NO independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to mitigate the consequences of an accident.

[50.73(a)(2)(vii)(D)]

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 32 of 59 ATTACHMENT 2 Page 3 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT

• • Any airborne radioactivity release that, when averaged over a time period of one NO hour, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in Appendix B to Part 20, table 2, column 1. [50.73(a)(2)(viii)(A)]

• • Any liquid effluent release that, when averaged over a period of one hour, NO exceeds 20 times the applicable concentrations specified in Appendix B to Part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) of all radionuclides except tritium and dissolved noble gases.

[50.73(a)(2)(viii)(B)]

Any event or condition that as a result of a single cause could have prevented the NO fulfillment of a safety function for two or more trains or channels in different systems that are needed to: (1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident. However, such an event need not be reported under this criterion if the event results from: (1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

[50.73(a)(2)(ix)(A)and (B)]

Any event that posed an actual threat to the safety of the nuclear power plant or NO significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases. [50.73(a)(2)(X)]

Discovery of loss of any shipment of Special Nuclear Material or spent fuel, or YES recovery of same. (Security-related) [73.71(a)(4)]

Any event in which there is reason to believe a person has committed, attempted YES to, or has made a credible threat to commit or cause a theft or unlawful diversion of special nuclear material. (Security-related)

[App G to Part 73, I(a)(1)]

Any event in which there is reason to believe a person has committed, attempted YES to, or has made a credible threat to commit or cause significant physical damage to the reactor or its equipment or nuclear fuel or the carrier of that fuel. (Security-related) [App G to Part 73, I(a)(2)]

Any event in which there is reason to believe a person has committed, or attempted YES to, or has made a credible threat to commit or cause interruption of the normal operation of the reactor through unauthorized use of or tampering with its machinery, components or controls, including the Security System (Security-related) [App G to Part 73, I(a)(3)]

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 33 of 59 ATTACHMENT 2 Page 4 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT An actual entry of an unauthorized person into a protected, material YES access, controlled access, vital or transport area. (Security-related) [App G to Part 73, I(b)]

Any failure, degradation, or discovered vulnerability in a safeguard system YES that could allow unauthorized or undetected access to a protected, material access, controlled access, vital or transport area for which compensatory measures have not been employed. (Security-related)

[App G to Part 73, I(c)]

Actual or attempted introduction of contraband into a protected, material YES access, vital or transport area. (Security-related) [App G to Part 73, I(d)]

Any lost, stolen or missing licensed material in an aggregate quantity equal YES to or greater than 1000 times the quantity specified in Appendix C to 20.1001-20.2401 of 10 CFR 20, under such circumstance that it appears than an exposure could result to persons in unrestricted areas.

(20.2201(a)(i))

• • • Any event involving by-product, source or special nuclear material that YES may have caused or threatens to cause an individual to receive:

• A total effective dose equivalent of 25 Rem or more; or

• An eye dose equivalent of 75 Rem or more; or

• A shallow dose equivalent to the skin or extremities of 250 rads or more.

(20.2202(a)(1) and 20.2203(a)(1))

• • • Any event involving by-product, source or special nuclear material that YES may have caused or threatens to cause the release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake 5 times the annual limit on intake (ALI). ALIs are listed in Appendix B to 20.1001-20.2401 of 10 CFR 20. (20.2202(a)(2) and 20.2203(a)(1))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 34 of 59 ATTACHMENT 2 Page 5 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT

• • • Any event involving by-product, source or special nuclear material that YES may have caused or threatens to cause an individual to receive:

• A total effective dose equivalent exceeding 5 Rem; or

• An eye dose equivalent exceeding 15 Rem; or

• A shallow dose equivalent to the skin or extremities exceeding 50 Rem.

(20.2202(b)(1) and 20.2203(a)(1))

• • • Any event involving by-product, source or special nuclear material that YES may have caused or threatens to cause the release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake in excess of one annual limit on intake (ALI). ALIs are listed in Appendix B to 20.1001-20.2401 of 10 CFR 20. (20.2202(b)(2) and 20.2203(a)(1))

Within 30 days after the occurrence of any lost, stolen or missing licensed YES material becomes known to the licensee, all licensed material in a quantity greater than 10 times the quantity specified in Appendix C to 20.1001-20.2401 of 10 CFR 20 that is still missing at the time of the report.

(20.2201(a)(ii))

• • • Doses in excess of the occupational dose limits for adults in 20.1201. NO (20.2203(a)(2)(i))

• • • Doses in excess of the occupational dose limits for minors in 20.1207. NO (20.2203(a)(2)(ii)

• • • Doses in excess of the limits for an embryo/fetus of a declared pregnant NO woman in 20.1208. (20.2203(a)(2)(iii))

• • • Doses in excess of the limits for an individual member of the public in NO 20.1301. (20.2203(a)(2)(iv))

• • • Doses in excess of any applicable limit in the DAEC license. NO (20.2203(a)(2)(v))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 35 of 59 ATTACHMENT 2 Page 6 of 6 REPORTABLE EVENTS IMMEDIATE EVENT NOTIFICATION EVENT Levels of radiation or concentrations of radioactive material in a restricted NO area in excess of any applicable limit in the DAEC license.

(20.2203(a)(3)(i))

Levels of radiation or concentrations of radioactive material in an NO unrestricted area in excess of 10 times any applicable limit set forth in 10 CFR 20 or in the DAEC license (whether or not involving exposure of any individual member of the public in excess of the limits in 20.1301).

(20.2203(a)(3)(ii))

Levels of radiation or releases of radioactive material in excess of the NO Environmental Protection Agency's generally applicable radiation standards in 40 CFR 190, or in excess of license conditions related to those standards. (20.2203(a)(4))

• Events covered in these paragraphs 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 to these paragraphs if redundant equipment in the same system was operable and available to perform the required safety function. [50.73(a)(2)(vi)]

• • Reports submitted to the NRC in accordance with these paragraphs also meet the effluent release reporting requirements of 10 CFR 20.2203(a)(3) [50.73(a)(2)(ix)]

• • • Written reports submitted to the NRC concerning individuals occupationally over-exposed to radiation and radioactive material shall have any section containing personal information clearly labeled with Privacy Action Information: Not for Public Disclosure.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 36 of 59 ATTACHMENT 3 Page 1 of 9 IMMEDIATE NOTIFICATION EVENTS NRC NRC NRC NRC RESP.

Event 1 HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Declaration of any of the Emergency Notify State and local authorities within 15 minutes of Action Levels as listed in EPIP 1.1 declaration of and EAL, NRC immediately afterwards (in . (50.72(a)(1)(i)) all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event) and management immediately following. (See EPIP 1.2)

The initiation of any nuclear plant No Yes No No OSM shutdown required by Tech. Specs.

(50.72(b)(2)(i))

Any deviation from the Tech. Specs. Yes No No No OSM authorized pursuant to 10 CFR 50.54(x). (50.72(b)(1))

Any event or condition that results in No No Yes No OSM the condition of the nuclear power plant including its principal safety barriers, being seriously degraded (50.72(b)(3)(ii)(A))

Any event or condition that results in No No Yes No OSM the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

(50.72(b)(3)(ii)(B))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 37 of 59 ATTACHMENT 3 Page 2 of 9 IMMEDIATE NOTIFICATION EVENTS NRC NRC NRC NRC RESP.

Event 1 HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event that results or should have No Yes No No OSM resulted in 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.

(50.72(b)(2)(iv)(A))

Any event that results in a major loss No No Yes No OSM of emergency assessment capability, off-site response capability or offsite communications capability. (e.g.,

significant portion of control room indication, Emergency Notification System, or offsite notification system) Note: Any siren failure rate of 10% or greater or any unplanned loss of the plant process computer for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> meets this criteria. (50.72(b)(3)(xiii))

Receipt of a radioactive material Yes No No No OSM package with removable surface contamination that exceeds the limits of 10 CFR 71.87; or external radiation levels that exceed the limits of 10 CFR 71.47. (20.1906(d)(1) &

(20.1906(d)(2))

Any lost, stolen, or missing licensed Yes No No No OSM material in an aggregate quantity equal to or greater that 1000 times the quantity specified in Appendix C to 20.1001-20.2401 of 10 CFR 20, under such circumstance that it appears that an exposure could result in unrestricted areas.

(20.2201(a)(i))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 38 of 59 ATTACHMENT 3 Page 3 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event involving by-product, Yes No No No OSM source or special nuclear material that may have caused or threatens to cause an individual to receive:

• A total effective dose equivalent of 25 Rem or more; or

• A eye dose equivalent of 75 Rem or more; or

• A shallow dose equivalent to the skin or extremities of 250 rads or more. (20.2202(a)(1))

Any event involving by-product, Yes No No No OSM source or special nuclear material that may have caused or threatens to cause the release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake 5 times the annual limit on intake. (ALI). ALIs are listed in Appendix B to 20.1101-20.2401 of 10 CFR 20. (20.2202(a)(2))

Any incident in which an attempt has Yes No No No OSM been made or is believed to have been made to commit a theft of unlawful diversion of more than 15 pounds of source material at any one time or more than 150 pounds of source material in any one calendar year. (40.64(c))

Any Accidental criticality or loss of Yes No No No OSM Special Nuclear Material. (70.52(a))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 39 of 59 ATTACHMENT 3 Page 4 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event of condition that at the No No Yes No OSM time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe shutdown condition. (50.72(b)(3)(v)(A))

Any event or condition that at the No No Yes No OSM time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.

(50.72(b)(3)(v)(B))

Any event or condition that at the No No Yes No OSM time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

(50.72(b)(3)(v)(C))

Any event or condition that at the No No Yes No OSM time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

(50.72(b)(3)(v)(D))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 40 of 59 ATTACHMENT 3 Page 5 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event or condition that results in No Yes No No OSM 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.

(50.72(b)(2)(iv)(B)).

Any event or condition that results in No No Yes No OSM See Section 3.2 valid actuation of any of the systems for a specific list listed in paragraph (b)(3)(iv)(B) of of systems this section except when the actuation results from and is part of a preplanned sequence during testing or reactor operation.(50.72(b)(3)(iv)(A)

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 41 of 59 ATTACHMENT 3 Page 6 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event requiring the transport of a No No Yes No OSM radioactively contaminated person to an offsite medical facility for treatment. (50.72(b)(3)(xii))

Any event or situation, related to the No Yes No No OSM If security-health and safety of the public or on- related, see site personnel, or protection of the section 3.1(8) environment, for which a news and/or the DAEC release is planned or notification to Security Event other government agencies has been Reporting or will be made. Such an event may Procedure include an on-site fatality or inadvertent release of radioactively contaminated materials.

(50.72(b)(2)(xi))

Any event involving by-product, No No No Yes OSM source or special nuclear material that may have caused or threatens to cause an individual to receive:

• A total effective dose equivalent exceeding 5 Rem; or

• An eye dose equivalent exceeding 15 Rem; or

• A shallow dose equivalent to the skin or extremities exceeding 50 Rem.

(20.2202(b)(1))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 42 of 59 ATTACHMENT 3 Page 7 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event involving by-product, No No No Yes OSM source or special nuclear material that may have caused or threatens to cause the release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake in excess of one annual limit on intake (ALI). ALIs are listed in Appendix B to 20.1001-20.2401 of 10 CFR 20. (20.2202(b)(2))

Discovery of loss of any shipment of Yes No No No Sec. Sup. See Security Special Nuclear Material or spent Procedure 11 fuel, or recovery of same. (73.71(a))

Any event in which there is reason to Yes No No No Sec. Sup. See Security believe a person has committed, Procedure 11 attempted to, or has made a credible threat to commit or cause a theft or unlawful diversion of special nuclear material. See Note 1. (73.71, App.

G., I.(a)(1))

Any event in which there is reason to Yes No No No Sec. Sup. See Security believe a person has committed, Procedure 11 attempted to, or has made a credible threat to commit or cause significant physical damage to the reactor or its equipment or nuclear fuel or the carrier of that fuel. See Note 1.

(73.71, App. G., I (a)(2))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 43 of 59 ATTACHMENT 3 Page 8 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Any event in which there is reason to Yes No No No Sec. Sup. See Security believe a person has committed, Procedure 11 attempted to, or has made a credible threat to commit or cause interruption of the normal operation of the reactor through unauthorized use of or tampering with its machinery, components, or controls, including the security system. See Note 1. (73.71, App. G., I. (a)(3))

An actual entry of an unauthorized Yes No No No Sec. Sup. See Security person into a protected, material Procedure 11 access, controlled access, vital or transport areas. (73.71, App. G.,

I.(b))

Any failure, degradation, or Yes No No No Sec. Sup. See Security discovered vulnerability in a Procedure 11 safeguard system that could allow unauthorized or undetected access to a protected, material access, controlled access vital or transport areas for which compensatory measures have not been employed.

(73.71, App. G., I.(c))

Actual or attempted introduction of Yes No No No Sec. Sup. See Security contraband into a protected, material Procedure 11 access, vital or transport area.

(73.71, App. G., I.(d))

Any event that meets the reportability No No No Yes Sec. Sup. See Procedure criteria of 10 CFR 26.73 (Fitness for FFD-7 Duty) as described in Security Directives. (10 CFR 26.73)

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 44 of 59 ATTACHMENT 3 Page 9 of 9 IMMEDIATE NOTIFICATION EVENTS NRC 1 NRC NRC NRC RESP.

Event HOUR 4 HOUR 8 HOUR 24 HOUR NOT. NOTE Within 30 days after the occurrence No No No No OSM Thirty Day of any lost, stolen or missing Telephone licensed material becomes known to Report per the licensee, all licensed material in 20.2201 (a)(ii) a quantity greater than 10 times the quantity specified in Appendix C to 20.1001-20.2401 of 10 CFR 20 that is still missing at the time of the report. (20.2201(a)(ii))

NOTE 1: For the purpose of reporting, the following definitions should be used: TAMPERING -

Unauthorized alteration or attempted entry of system equipment or components for the purpose of disabling a component system that would interrupt normal plant or security operation. SABOTAGE -

Any deliberate act directed against the plant or against a component of the plant which could directly or indirectly endanger the public health and safety by exposure to radiation.

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 45 of 59 ATTACHMENT 4 RPS ACTUATION REPORTING MATRIX Valid Invalid Immediate LER (50.73) Immediate LER (50.73)

Notification Event Notification Event (50.72) (50.72)

Critical 4 Hour Report per 60 Day LER per 4 Hour Report per 60 Day LER per 50.72(b)(2)(iv)(B) 50.73(a)(2)(iv)(A) 50.72(b)(2)(iv)(B) 50.73(a)(2)(iv)(A)

Critical No Report No Report No Report No Report (preplanned)

Non-Critical 8 Hour report per 60 Day LER per No Report 60 Day Telephone 50.72(b)(3)(iv)(B) 50.73(a)(2)(iv)(A) Report per 50.73(a)(1)

Non-Critical No Report No Report No Report No Report (preplanned)

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 46 of 59 ATTACHMENT 5 Page 1 of 2 10 CFR 72 IMMEDIATE NOTIFICATION EVENTS NRC NRC NRC NRC RESP.

Event 1 4 HOUR 8 HOUR 24 NOT. NOTE HOUR HOUR The discovery of accidental Yes No No No OSM criticality or any loss of special nuclear material. (72.74(a))

Declaration of any of the Notify State and local authorities within 15 minutes of Emergency Action Levels as declaration of an EAL, NRC immediately afterwards (in listed in EPIP 1.1 Attachment 1. all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event) and management (72.75(a)) immediately following. (See EPIP 1.2)

An action taken in an emergency No Yes No No OSM that departs from a condition or a technical specification contained in a license or certificate of compliance issued under this part when the action is immediately needed to protect the public health and safety and no action consistent with license or certificate of compliance conditions or technical specifications that can provide adequate or equivalent protection is immediately apparent (72.75(b)(1))

Any event or situation related to No Yes No No OSM 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. (72.75(b)(2))

A defect in any spent fuel No No Yes No OSM storage structure, system, or component which is important to safety. (72.75(c)(1))

A significant reduction in the No No Yes No OSM effectiveness of any spent fuel storage confinement system during use. (72.75(c)(2))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 47 of 59 ATTACHMENT 5 Page 2 of 2 10 CFR 72 IMMEDIATE NOTIFICATION EVENTS NRC NRC NRC NRC RESP.

Event 1 4 HOUR 8 HOUR 24 NOT. NOTE HOUR HOUR An event that requires transport No No Yes No OSM of a radioactively contaminated person to an offsite medical facility for treatment.

(72.75(c)(3))

An event in which important to No No No Yes OSM safety equipment is disabled or fails to function as designed when the equipment is required by regulation, licensed condition, or certificate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident; and no redundant equipment was available and operable to perform the required safety function. (72.75(d)(1))

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 48 of 59 ATTACHMENT 6 Page 1 of 2 NOTIFICATION TO STATE/LOCAL OFFICIALS CONDITION 1 Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of ODAM Table 6.3-3 when averaged over any calendar quarter sampling period (ODAM OLCO 6.3.2 Condition B).

CONDITION 2 A spill or leak of licensed material (including liquids resulting from a spill/leak of stream or solids), from a plant system, structure or component or which occurs as a result of a failure during a work practice, that has the potential to reach ground water and meets the following criteria:

• Exceeds 100 gallons

• Cannot be quantified but is likely to exceed 100 gallons

• Site or corporate management determines that communication of the spill or leak is warranted If either CONDITION 1 OR CONDITION 2 is met, make notification to Contacts 1 and 2 by the end of the business day following the day that the spill/leak occurred or condition was verified. Refer to Nuclear Fleet Guideline EV-AA-1000, Ground Water Protection Program Communications/Notification Plan for additional guidance.

Contact Contact Organization Business Address Contact Phone Notation No. Representative Number 1 Bureau Chief Bureau of Radiological Lucas State Office Building, (515)281-3478 -

Health 5th Floor th 321 East 12 Street Des Moines, Iowa 50319-0073 2 Linn County Public Public Health Department 501 13th Street NW (319)892-6000 -

Health Director Linn County, Iowa Cedar Rapids, IA 52405 3 Iowa DNR Emergency Iowa DNR Emergency 401 SW 7th Street, Suit I (515)281-8694 Fax: (515)725-0218 Response Unit Response Unit Des Moines, Iowa 50309 http://www.iowadnr.com/spills/rep ort.html 4 Environmental FPL/FPLENextEra Energy 700 Universe Blvd ENG/JB (603)773-7438 (W)* CFAM RP & Chemistry Corporate Functional Juno Beach, FL 33408 (603)765-7291 ( c)

Area Manager

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 49 of 59 ATTACHMENT 6 Page 2 of 2 NOTIFICATION TO STATE/LOCAL OFFICIALS Contact Organization Business Address Contact Phone Notation Representative Number 5 FPL/FPLE NextEra DAEC Communications FPL Energy Duane Arnold (319)851-7140 Energy Rep. LLC Communications 3277 DAEC Road Representative Palo, Iowa 52324 FPL Communications Rep. 700 Universe Blvd.

(603)773-7281 (W)

Juno Beach, FL 33408 (603)765-6444 (C) 6 FPL Risk Management Risk Management 700 Universe Blvd. (561)371-5210 Rep Juno Beach, FL 33408 or (561)691-3030 7 ANI Account Engineer ANI Account Engineer 95 Glastonbury Blvd (860)682-1301 Glastonbury, CT 06033 8 NEI Representative Senior Manager, 1776 I Street NW, Suite 400 (202)739-8000 GW_Notice@nei.org Environmental Protection Washington, DC 20006 E-mail is preferred method of contact 9 Radiation Protection Site: Radiation Protection - - -

Manager and Chemistry 10 Environmental Site Site: RP/Chem Technical - -

Function al Area Staff Supervisor Manager If CONDITION 2 is met, implement actions as described in ACP 1411.14. Make notification to the below listed State officials within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

State Officials Iowa DNR Emergency Response Unit th 401 SW 7 Street, Suit I Des Moines, Iowa 50309 PH. 515-281-8694 Fax 515-725-0218 http://www.iowadnr.com/spills/report.html

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 50 of 59 ATTACHMENT 7 Page 1 of 2 COMMUNICATION INFORMATION CHECKLIST - SAMPLE ONLY EVENT RECORDER:_______________________DATE:_________ TIME__________

1. Condition before the event:________________________________________________

2. The first indication of the event or occurrence:

Date: _______Time:___________Individual(s) Involved: _______________________________

Description of event:____________________________________________________________

3. Plant or Operator actions taken:_____________________________________________

4. List entries into TS/TRM/ODAM/Fire Plan LCOs: _______________________________

5. Current condition of the event:______________________________________________

6. List Procedures entered or required to be entered:______________________________

7. List other actions item (CAPs/CWOs/PWRs/ TIFs/etc ) taken to resolve event:

8. Record DSM contact time and if the ERT was activated _________________________

NG-005F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 51 of 59 ATTACHMENT 7 Page 2 of 2 COMMUNICATION INFORMATION CHECKLIST - SAMPLE ONLY

9. Who has been contacted on this event from appropriate attachments or recorded below:

NOTE Recorded any questions or comments from communications made during the communication process.

10. COMMEMTS:_______________________________________________________

NG-005F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 52 of 59 ATTACHMENT 8 COMMUNICATION TO THE DUTY STATION MANAGER-SAMPLE ONLY NOTE The OSM shall ensure the Duty Station Manager is notified per ACP 114.3 for the events listed below as soon as plant conditions allow. Check the appropriate event(s) the DSM is being contacted and record date and time the notification has been made.

Events Orange Unplanned Online/Shutdown Risk Entry into a shutdown LCO Conditions for a Human Performance Site Clock Reset Hazardous Material Incident requiring the HAZMAT team Reactivity Event Fitness for Duty Event Injury requiring offsite medical attention or transportation via ambulance to an offsite medical facility Non-routine communications with the NRC Action Level 2 or greater chemistry action level Any event or operating condition outside the plant design basis Unexpected 1/2 scram Unexpected significant plant transient Unplanned power reduction LCO action statement that will not be met within the allowed time requirement Initiation of the Event Response Team Events of public interest that may involve the news media Unplanned ESF actuation Fire Brigade mustered in response to an actual fire Notification to any offsite agency Significant breakdown of plant radiological or environmental controls Any radiological or non-radiological release reportable to local, state or federal agency DSM Contacted Name:____________________________ Date:___________ Time:________

Communicator Signature:___________________________ Date___________ Time:________

OSM Signature:___________________________________ Date___________ Time:________

NG-006F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 53 of 59 ATTACHMENT-9 COMMUNICATION TO THE NUCLEAR DIVISION DUTY OFFICER-SAMPLE ONLY NOTE The OSM/DSM shall ensure the Nuclear Division Duty Officer (NDDO) is notified per Nuclear Policy NP-303 for the events listed below as soon as plant conditions allow. Check the appropriate event(s) the NDDO is being contacted and record date and time the notification has been made.

Problems or potential problems requiring NRC notification.

Injury of a serious nature or fatality of any employee or contractor.

Significant plant equipment damage (in excess of $100,000).

Security threats of any nature against the plant or personnel. This includes, but is not limited to the following: potential tampering events, security equipment problems that could be construed as degradation to the effectiveness of the security plan, workforce issues that could call into question the integrity of the officer workforce, and any other events that could draw attention to the company in a world of heightened security awareness.

Any request to Access Control for an unfavorable termination of access.

Acts of known or suspected sabotage.

External threats to generation (e.g. fires, accidents, system dispatch information).

Hazardous weather warnings (hurricanes, tornadoes, blizzards, or cold weather) which could affect normal plant operations.

Significant labor issues.

Significant quality issues - examples of such issues would include:

Any and all breakdowns in material control at FPL or any of its suppliers.

Systematic weaknesses in either programs or procedures being utilized by FPL.

Media interest or events likely to result in media interest.

Enforcement actions (notice of violations, levying of civil penalties, etc.).

Internal management conflicts.

Unplanned reductions in power (greater than 5%).

Spills or releases of radioactive material requiring immediate notification of state or federal agencies.

A significant leak or spill into on-site groundwater that is communicated to State and Local officials pursuant to the implementation of Nuclear Fleet Guideline EV-AA-100-1000, Ground Water Protection Program Communications/Notifications Plan".

Any off-site or on-site environmental water sample result that exceeds Radiological Environmental Monitoring Program reporting requirements and is therefore communicated to State and Local officials pursuant to the implementation of Nuclear Fleet Guideline EV-AA-100-1000, Ground Water Protection Program Communications/Notifications Plan".

Any non-radiological environmental event or occurrence for which immediate notification is required to any Local, State or Federal environmental authority.

Any other matter judged to be provocative and/or significant relating to the nuclear plants or staffs.

NDDO Contacted Name:____________________________ Date:__________ Time:________

Communicator Signature:___________________________ Date___________ Time:________

OSM/DSM Signature:_______________________________ Date___________ Time:________

NG-007F Rev. 1 (Rev. ACP 1402.3)

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 54 of 59 ATTACHMENT-10 COMMUNICATION FOR IMMEDIATE NOTIFICATION EVENT -- SAMPLE ONLY NOTE The Plant Manager, NDDO and NRC Resident Inspector should be notified as soon as possible.

During non-business hours, the Plant Manager may direct other notifications be delayed until business hours based on the nature of the event. Record N/A for not required for immediate notification or N/C for not able to contact individual or designee. In the event that the reactor trip due to the event, Attachment 14 is required by NP 303 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the trip.

Init / Date / Time

____ / ________/_________ a. Plant Manager

____ / ________/_________ b. Nuclear Division Duty Officer (NDDO)

____ / ________/_________ c. NRC Resident Inspector (attempt Senior Resident first)

____ / ________/_________ d. Site Vice President

____ / ________/_________ e. Site Director

____ / ________/_________ f. Engineering Director

____ / ________/_________ g. Operations Manager

____ / ________/_________ h. Maintenance Manager

____ / ________/_________ i. Regulatory Affairs Manager

____ / ________/_________ j. Radiation Protection Manager

____ / ________/_________ k. Emergency Planning Manager

____ / ________/_________ l. Communications Manager (For external Notifications Only)

____ / ________/_________ m. Safety Manager (Injuries Only)

____ / ________/_________ n. Security Manager (Security Issues Only)

Communicator Signature:___________________________ Date___________ Time:________

DSM/OSM Signature:______________________________ Date___________ Time:________

NG-008F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 55 of 59 ATTACHMENT-11 COMMUNICATION FOR REPORTABLE EVENT -- SAMPLE ONLY NOTE The Plant Manager and NDDO should be notified as soon as possible. The Plant Manager may direct other notifications be delayed based on the nature of the event. Record N/A for not required for essential notification or N/C for not able to contact individual or designee. In the event that the reactor trip due to the event, Attachment 14 is required by NP 303 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the trip.

Init / Date / Time

____ / ________/_________ a. Plant Manager

____ / ________/_________ b. Nuclear Division Duty Officer (NDDO)

____ / ________/_________ c. NRC Resident Inspector (attempt Senior Resident first)

____ / ________/_________ d. Site Vice President

____ / ________/_________ e. Site Director

____ / ________/_________ f. Engineering Director

____ / ________/_________ g. Operations Manager

____ / ________/_________ h. Maintenance Manager

____ / ________/_________ i. Regulatory Affairs Manager

____ / ________/_________ j. Radiation Protection Manager

____ / ________/_________ k. Emergency Planning Manager

____ / ________/_________ l. Communications Manager (For external Notifications Only)

____ / ________/_________ m. Safety Manager (Injuries Only)

____ / ________/_________ n. Security Manager (Security Issues Only)

Communicator Signature:___________________________ Date___________ Time:________

DSM/OSM Signature:______________________________ Date___________ Time:________

NG-009F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 56 of 59 ATTACHMENT-12 COMMUNICATION FOR PLANT OPERATIONAL ISSUES -- SAMPLE ONLY NOTE The Duty Station Manager, Plant Manager and NDDO should be notified as soon as possible. The Plant Manager may direct other notifications be delayed based on the nature of the event. Record N/A for not required for essential notification or N/C for not able to contact individual or designee. In the event that the reactor trip due to the event, Attachment 14 is required by NP 303 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the trip.

Init / Date / Time

____ / ________/_________ a. Duty Station Manager

____ / ________/_________ b. Operations Manager

____ / ________/_________ c. Plant Manager

____ / ________/_________ d. Nuclear Division Duty Officer (NDDO)

____ / ________/_________ e. Site Vice President

____ / ________/_________ f. Site Director

____ / ________/_________ g. Regulatory Affairs Manager (For external Notifications Only)

____ / ________/_________ h. NRC Resident Inspector (attempt Senior Resident first)

____ / ________/_________ i. Safety Manager (Injuries Only)

____ / ________/_________ j. Communications Manager (For external Notifications Only)

NOTE The Duty Station Manager will consider notifications to individual duty team members.

____ / ________/_________ aa. Duty Engineering Manager

____ / ________/_________ bb. Duty Radiation Protection Manager

____ / ________/_________ cc. Duty Operations Manager

____ / ________/_________ dd. Duty Maintenance Manager Communicator Signature:___________________________ Date___________ Time:________

DSM/OSM Signature:______________________________ Date___________ Time:________

NG-010F Rev. 0

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 57 of 59 ATTACHMENT-13 Formatted: Font: 11 pt COMMUNICATION FOR MEDICAL RESPONSE/ACCIDENT REPORTING -SAMPLE ONLY Date:____________ Time:_____________ Reported By:___________________________

Location:_________________________________________________________________________

Name of Injured:____________________ Badge Number:______________________

Nature of Injury:_________________________________________________________________

Employer:_______________________________________

Responder:___________________________ Badge Number:__________________________

Responder::_________________________ Badge Number:__________________________

Contaminated? (Y) (N) Level:__________________________________

Requires Offsite Transportation (Y) (N) Assess NRC Reportability per ACP 1402.3.

NOTE: *Notify only if serious injury (i.e. offsite medical notified)

Init / Date / Time

____ / ________/_________ a. Health Physics

____ / ________/_________ b. Security Operations Supervisor

____ / ________/_________ c. Safety Representative

____ / ________/_________ d. Individuals Supervisor

____ / ________/_________ e. Duty Station Manager

____ / ________/_________ e. Plant Manager

____ / ________/_________ f. Nuclear Division Duty Officer (NDDO)*

____ / ________/_________ h. Site Vice President

____ / ________/_________ i. Communications Manager*

____ / ________/_________ j. Emergency Planning Manager*

____ / ________/_________ j. Emergency PlanningRadiation Protection Manager*

Communicator Signature:___________________________ Date___________ Time:________

DSM/OSM Signature:______________________________ Date___________ Time:________

Return completed form to the Safety Office.

NG-001A Rev. 5

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 58 of 59 NG-001A Rev. 5

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 59 of 59 ATTACHMENT-14 NP-303 CHIEF NUCLEAR OFFICER REPORT OF REACTOR TRIP -

SAMPLE ONLY NOTE This information must be sent or communicated to the Chief Nuclear Officer within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of an unplanned reactor trip.

Date/Time of reactor trip:_______________________________________________

Initial Power Level:____________________________________________________

1. Cause/Apparent cause of trip:

2. Circumstances surrounding trip (ongoing maintenance, load threats, etc.):

3. Response of operating crew to event, including any human performance issues noted:

4. Equipment malfunctions/anomalies noted:

5. Any other items deemed significant:

Prepared By:___________________________________________Date:______________

Reviewed By:___________________________________________Date:_____________

Operations Manager Approved By:____________________________________________Date:____________

Vice President - Duane Arnold Energy Center NG-012F Rev. 0

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1 Offsite Dose Resulting from an Actual or Imminent Release of RS1 Offsite Dose Resulting from an Actual or Imminent Release of RA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to RU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All SS1 Loss of All Offsite Power and Loss of All Onsite AC Power to SA5 AC Power Capability to Essential Busses Reduced to a Single SU1 Loss of All Offsite Power to Essential Busses for Greater Than Gaseous Radioactivity that Exceeds 1000 mRem TEDE or 5000 Gaseous Radioactivity Exceeds 100 mRem TEDE or 500 mRem the Environment that Exceeds 200X the Offsite Dose the Environment That Exceeds Two Times the Offsite Dose Onsite AC Power to Essential Busses Essential Busses Power Source for Greater Than 15 Minutes Such That Any 15 Minutes mRem CDE Thyroid for the Actual or Projected Duration of the CDE Thyroid for the Actual or Projected Duration of the Release Assessment Manual (ODAM) Limit and is Expected to Continue Assessment Manual (ODAM) Limit and is Expected to Continue Additional Single Failure Would Result in Station Blackout Release Using Actual Meteorology for 15 Minutes or Longer For 60 Minutes or Longer SG1.1 1 2 3 SS1.1 1 2 3 SA5.1 1 2 3 SU1.1 1 2 3 RG1.1 1 2 3 4 5 DEF RS1.1 1 2 3 4 5 DEF RA1.1 1 2 3 4 5 DEF RU1.1 1 2 3 4 5 DEF Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Valid Reactor Building ventilation rad monitor (Kaman 3/4, Valid Reactor Building ventilation rad monitor (Kaman 3/4, Loss of power to or from the Startup or Standby Loss of power to or from the Startup or Standby AC power capability to 1A3 or 1A4 busses reduced to a single Loss of power to or from the Startup or Standby GREATER THAN 1000 mRem TEDE or 5000 mRem thyroid GREATER THAN 100 mRem TEDE or 500 mRem thyroid 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman Transformer resulting in a loss of all offsite power to Transformer resulting in a loss of all offsite power to Emer- power source for greater than 15 minutes Transformer resulting in a loss of all offsite power to CDE at or beyond the site boundary. (Preferred method) CDE at or beyond the site boundary. (Preferred method) 1/2) reading that exceeds 3 E-2 µCi/cc and is expected to 1/2) reading that exceeds 1 E-3 µCi/cc and is expected to Emergency Busses 1A3 and 1A4 gency Busses 1A3 and 1A4 AND Emergency Busses 1A3 and 1A4 that is expected to last continue for 15 minutes or longer continue for 60 minutes or longer AND AND Any additional single failure will result in station blackout for greater than 15 minutes Loss of Failure of A Diesel Generator (1G-31) and B Diesel Failure of A Diesel Generator (1G-31) and B Diesel AND RG1.2 1 2 3 4 5 DEF RS1.2 1 2 3 4 5 DEF RA1.2 1 2 3 4 5 DEF RU1.2 1 2 3 4 5 DEF Generator (1G-21) to supply power to emergency busses Generator (1G-21) to supply power to emergency busses Emergency Busses 1A3 and 1A4 are powered by their If Dose Assessment is unavailable, either of the following: If Dose Assessment is unavailable, any of the following: Valid Offgas Stack rad monitor (Kaman 9/10) reading that Power 1A3 and 1A4 1A3 and 1A4 respective Standby Diesel Generators Valid Offgas Stack rad monitor (Kaman 9/10) reading that

- Valid Reactor Building ventilation rad monitor (Kaman - Valid Reactor Building ventilation rad monitor (Kaman exceeds 6 E+0 µCi/cc and is expected to continue for 15 exceeds 2.0 E-1 µCi/cc and is expected to continue for 60 AND AND 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor minutes or longer minutes or longer ANY ONE OF THE FOLLOWING: Failure to restore power to at least one emergency bus, 1A3 (Kaman 1/2) reading GREATER THAN 6 E-1 µCi/cc (Kaman 1/2) reading GREATER THAN 6 E-2 µCi/cc - Restoration of power to either Bus 1A3 or 1A4 is not or 1A4, within 15 minutes from the time of loss of both offsite and is expected to continue for 15 minutes or longer. RA1.3 1 2 3 4 5 DEF RU1.3 1 2 3 4 5 DEF and is expected to continue for 15 minutes or longer. likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and onsite AC power

- Valid Offgas Stack rad monitor (Kaman 9/10) reading - Valid Offgas Stack rad monitor (Kaman 9/10) reading Valid LLRPSF rad monitor (Kaman 12) reading that exceeds Valid LLRPSF rad monitor (Kaman 12) reading that exceeds - RPV level is indeterminate 1 E-1 µCi/cc and is expected to continue for 15 minutes or 1.0 E-3 µCi/cc and is expected to continue for 60 minutes or SS3 Loss of All Vital DC Power GREATER THAN 4 E+2 µCi/cc and is expected to GREATER THAN 4 E+1 µCi/cc and is expected to - RPV Level is LESS THAN +15 inches continue for 15 minutes or longer continue for 15 minutes or longer longer longer Offsite Rad SS3.1 1 2 3 RG1.3 1 2 3 4 5 DEF RS1.3 1 2 3 4 5 DEF RA1.4 1 2 3 4 5 DEF RU1.4 1 2 3 4 5 DEF Conditions Loss of Div 1 and Div 2 125V DC busses based on bus Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+5 Valid GSW rad monitor (RIS-4767) reading that exceeds CPS and is expected to continue for 15 minutes or longer 3E+3 CPS and is expected to continue for 60 minutes or voltage LESS THAN 105 VDC indicated for greater than 15 exceeding 1000 mRem/hr expected to continue for more exceeding 100 mRem/hr expected to continue for more than longer minutes than one hour at or beyond the site boundary; or analyses of one hour at or beyond the site boundary; or analyses of field field survey samples indicate thyroid CDE of 5000 mRem for survey samples indicate thyroid CDE of 500 mRem for one RA1.5 1 2 3 4 5 DEF RU1.5 1 2 3 4 5 DEF SG2 Failure of the Reactor Protection System to Complete an SS2 Failure of Reactor Protection System Instrumentation to SA2 Failure of Reactor Protection System Instrumentation to one hour of inhalation at or beyond the site boundary hour of inhalation at or beyond the site boundary Automatic Scram and Manual Scram was NOT successful and Complete or Initiate an Automatic Reactor Scram Once a Complete or Initiate an Automatic Reactor Scram Once a Valid RHRSW & ESW rad monitor (RM-1997) reading that Valid RHRSW & ESW rad monitor (RM-1997) reading that There is Indication of an Extreme Challenge to the Ability to Reactor Protection System Setpoint Has Been Exceeded and Reactor Protection System Setpoint Has Been Exceeded and exceeds 8E+4 CPS and is expected to continue for 15 exceeds 8E+2 CPS and is expected to continue for 60 Cool the Core Manual Scram Was NOT Successful Manual Scram Was Successful minutes or longer minutes or longer SG2.1 1 2 SS2.1 1 2 SA2.1 1 2 RA1.6 1 2 3 4 5 DEF RU1.6 1 2 3 4 5 DEF Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) RPS Auto Scram failure reading that exceeds 1E+3 CPS and is expected to continue Failure AND Auto Scram failure Auto Scram failure reading that exceeds 1E+5 CPS and is expected to continue None for 15 minutes or longer for 60 minutes or longer NONE of the following operator actions to reduce power AND AND are successful in shutting down the reactor: NONE of the following operator actions to reduce power ANY of the following operator actions to reduce power are RA1.7 1 2 3 4 5 DEF RU1.7 1 2 3 4 5 DEF

- Manual Scram Pushbuttons are successful in shutting down the reactor: successful in shutting down the reactor:

Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases - Manual Scram Pushbuttons - Manual Scram Pushbuttons

- Mode Switch to Shutdown indicates concentrations or release rates with a release indicates concentrations or release rates in excess of 2 - Mode Switch to Shutdown - Mode Switch to Shutdown

- Alternate Rod Insertion (ARI) duration expected to continue for 15 minutes or longer in times ODAM limits and is expected to continue for 60 - Alternate Rod Insertion (ARI) - Alternate Rod Insertion (ARI) minutes or longer AND excess of 200 times ODAM limit Loss of adequate core cooling or decay heat removal capability as indicated by either:

- RPV level cannot be maintained GREATER THAN -25 Abnormal RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or RU2 Unexpected Increase in Plant Radiation Will Result in the Uncovering of Irradiated Fuel Outside the inches Rad - HCL Curve (EOP Graph 4) exceeded Reactor Vessel Release SU2 Inability to Reach Required Shutdown Within Technical Inability to None SS4 Complete Loss of Heat Removal Capability RA2.1 1 2 3 4 5 DEF RU2.1 1 2 3 4 5 DEF Specification Limits Reach or Rad Report of any of the following: RU2.1 Unplanned valid Refuel Floor ARM reading increase Maintain SS4.1 1 2 3 None SU2.1 1 2 3 Effluent - Valid ARM Hi Rad alarm for the Refueling Floor North End with an uncontrolled loss of reactor cavity, fuel pool, or fuel Shutdown Plant is not brought to required operating mode within (RM 9163), Refueling Floor South End (RM 9164), New transfer canal water level with all irradiated fuel assemblies System Conditions EOP Graph 4 Heat Capacity Limit is exceeded applicableTechnical Specifications LCO Action Statement Time Fuel Storage (RM 9153), or Spent Fuel Storage Area (RM remaining covered by water as indicated by any of the Malfunct.

9178). following: SS6 Inability to Monitor a Significant Transient in Progress SA4 Unplanned Loss of Most or All Safety System Annunciation or SU3 Unplanned Loss of Most or All Safety System Annunciation or

- Valid Refueling Floor North End (RM-9163), Refueling Floor - Report to control room Indication in Control Room With Either (1) a Significant Indication in the Control Room for Greater Than 15 Minutes South End (RM-9164), or New Fuel Storage Area (RM- - Valid fuel pool level indication (LI-3413) LESS THAN 36 Transient in Progress, or (2) Compensatory Non-Alarming feet and lowering Indicators Unavailable 9153) ARM Reading GREATER THAN 10 mRem/hr

- Valid Spent Fuel Storage Area (RM-9178) ARM Reading - Valid WR GEMAC Floodup indication (LI-4541) coming SS6.1 1 2 3 SA4.1 1 2 3 SU3.1 1 2 3 GREATER THAN 100 mRem/hr on scale Unplanned loss of most or all 1C03, 1C04 and 1C05 RA2.2 1 2 3 4 5 DEF Significant transient in progress and ALL of the following: Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Safety Systems for RU2.2 1 2 3 4 5 DEF - Loss of most or all annunciators on Panels 1C03, annunciators or indicators associated with Safety Systems for greater than 15 minutes Valid water level reading LESS THAN 450 inches as indicated on LI-4541 (floodup) for the Reactor Refueling Any unplanned ARM reading offscale high or GREATER 1C04 and 1C05. greater than 15 minutes SU6 UNPLANNED Loss of All Onsite or Offsite Communications Cavity that will result in Irradiated Fuel uncovering THAN 1000 times normal* reading - Compensatory non-alarming indications are AND Capabilities Onsite Rad Inst. / None

• Normal levels can be considered as the highest reading in the past unavailable. Either of the following conditions exist:

Conditions Comm. - Indicators needed to monitor criticality, or core heat - A significant plant transient is in progress. SU6.1 1 2 3 None None RA2.3 1 2 3 4 5 DEF twenty-four hours excluding the current peak value Valid Fuel Pool water level indication (LI-3413) LESS THAN removal, or Fission Product Barrier status are - Compensatory non-alarming indications are unavailable Loss of ALL of the following onsite communication capa-16 feet that will result in Irradiated Fuel uncovering unavailable. bilities affecting the ability to perform routine operation:

- Plant Operations Radio System RA3 Release of Radioactive Material or Increases in Radiation Levels - In-Plant Telephones Within the Facility That Impedes Operation of Systems Required - Plant Paging System to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown SU6.2 1 2 3 Loss of ALL of the following offsite communications capability:

RA3.1 1 2 3 4 5 DEF - All telephone lines (commercial)

Valid area radiation levels GREATER THAN 15 mRem/hr in - Microwave Phone System any of the following areas: - FTS Phone System

- Control Room (RM 9162)

SU4 Fuel Clad Degradation

- Central Alarm Station (by survey)

- Secondary Alarm Station (by survey)

SU4.1 1 2 3 RA3.2 1 2 3 4 5 DEF Fuel Clad Pretreatment Offgas System (RM-4104) Hi-Hi Radiation None None None Alarm Valid area radiation monitor (RE-9168), reading GREATER Degradation THAN 500 mRem/hr affecting the Remote Shutdown Panel, SU4.2 1 2 3 1C388 Reactor Coolant sample activity value GREATER THAN HA1 Natural and Destructive Phenomena Affecting the Plant Vital HU1 Natural and Destructive Phenomena Affecting the Protected Area 2.0 µCi/gm dose equivalent I-131 Area Safe Shutdown/Vital Areas HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF SU5 RCS Leakage Category Area Receipt of the Amber Operating Basis Earthquake Light and Earthquake detected per AOP 901, Earthquake SU5.1 1 2 3 the wailing seismic alarm on 1C35 (+/- 0.06 gravity) RCS HU1.2 1 2 3 4 5 DEF Unidentified or pressure boundary leakage GREATER Electrical Power 1G31 DG and Day Tank Rooms, 1G21 DG Leakage None None None HA1.2 1 2 3 4 5 DEF THAN 10 gpm and Day Tank Rooms, Battery Rooms, Report of a tornado touching down within the Plant Protected Essential Switchgear Rooms, Cable Report of Tornado or high winds greater than 95MPH within Area with NO confirmed damage to a Safe Shutdown/Vital SU5.2 1 2 3 Spreading Room PROTECTED AREA boundary and resulting in VISIBLE Area or Control Room indication of degraded performance of DAMAGE to a Safe Shutdown/Vital Area or Control Room a System of Concern Identified leakage GREATER THAN 25 gpm Heat Sink / Coolant Supply Torus Room, Intake Structure, Pumphouse indication of degraded performance of a System of Concern SU8 Inadvertent Criticality HU1.3 1 2 3 4 5 DEF HA1.3 1 2 3 4 5 DEF Containment Drywell, Torus Report of winds greater than 95 mph within the Plant Inadvertent SU8.1 3 Vehicle crash within PROTECTED AREA boundary and Protected Area with NO confirmed damage to a Safe Criticality None None None Emergency Systems NE, NW, SE Corner Rooms, HPCI Room, resulting in VISIBLE DAMAGE to a Safe Shutdown/Vital Area An UNPLANNED extended positive period observed on Shutdown/Vital Area or Control Room indication of degraded RCIC Room, RHR Valve Room, North CRD or Control Room indication of degraded performance of a nuclear instrumentation performance of a System of Concern Area, South CRD Area, CSTs System of Concern HU1.4 1 2 3 4 5 DEF EU1 Damage To A Loaded Cask Confinement Boundary HA1.4 1 2 3 4 5 DEF Vehicle crash into plant structures or systems within the Other Control Building, Remote Shutdown Panel EU1.1 1C388 Area, Panel 1C55/56 Area, SBGT Turbine failure-generated missiles result in any VISIBLE Plant Protected Area with NO confirmed damage to a Safe DAMAGE to or penetration of any of a Safe Shutdown/Vital Shutdown/Vital Area or Control Room indication of degraded Any one of the following natural phenomena events with Natural & None Room None Area performance of a System of Concern resultant visible damage to or loss of a loaded cask Destructive confinement boundary:

Phenonenon HA1.5 1 2 3 4 5 DEF HU1.5 1 2 3 4 5 DEF - Report by plant personnel of a tornado strike Report of an unanticipated explosion within the Plant - Report by plant personnel of a seismic event River level ABOVE 767 feet Cask Systems of Concern Protected Area resulting in visible damage to permanent EU1.2 Confine. None None None HA1.6 1 2 3 4 5 DEF structures or equipment The following accident condition with resultant visible Boundary

- Reactivity Control Uncontrolled flooding in a Safe Shutdown/Vital Area that HU1.6 1 2 3 4 5 DEF damage to or loss of a loaded cask confinement boundary:

- Containment (Drywell/Torus) results in degraded safety system performance as indicated in - A loaded transfer cask is dropped as a result of Report of turbine failure resulting in casing penetration or normal handling or transporting

- RHR/Core Spray/SRVs the Control Room or that creates an industrial safety hazards damage to turbine or generator seals (e.g., electric shock) that precludes access necessary to EU1.3

- HPCI/RCIC HU1.7 1 2 3 4 5 DEF operate or monitor safety equipment Any condition in the opinion of the Emergency Director that

- RHRSW/River Water/ESW indicates loss of loaded fuel storage cask confinement HA1.7 River level ABOVE 757 feet

- Onsite AC Power/EDGs 1 2 3 4 5 DEF boundary

- Offsite AC Power River level BELOW 724 feet 6 inches HU1.8 1 2 3 4 5 DEF ISFSI

- Instrument AC HA1.8 1 2 3 4 5 DEF Uncontrolled flooding in a Safe Shutdown/Vital Area that has Events

- DC Power the potential to affect safety related equipment needed for Report to control room of VISIBLE DAMAGE affecting a Safe the current operating mode

- Remote Shutdown Capability Shutdown/Vital Area HU1.9 1 2 3 4 5 DEF River level BELOW 725 feet 6 inches HA2 Fire or Explosion Affecting the Operability of Plant Safety HU2 Fire Within Protected Area Boundary Not Extinguished Within 15 Systems Required to Establish or Maintain Safe Shutdown Minutes of Detection Fire HA2.1 1 2 3 4 5 DEF HU2.1 1 2 3 4 5 DEF Security None None Fire or explosion in any Safe Shutdown/Vital Area Fire in buildings or areas contiguous to any Safe None None None None or Explosion AND Shutdown/Vital Area not extinguished within 15 minutes of Affected system parameter indications show degraded control room notification or verification of a control room alarm performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area HA3 Release of Toxic or Flammable Gases Within or Contiguous to a HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Vital Area Which Jeopardizes Operation of Systems Required to Normal Operation of the Plant Maintain Safe Operations or Establish or Maintain Safe Shutdown HA3.1 1 2 3 4 5 DEF HU3.1 1 2 3 4 5 DEF Toxic Report or detection of toxic gases within or contiguous to a Report or detection of toxic or flammable gases that has or and Safe Shutdown/Vital Area in concentrations that may result in could enter the site area boundary in amounts that can affect Hazards Flammable None None an atmosphere Immediately Dangerous to Life and Health normal plant operations Gas FG1 1 2 3 FS1 1 2 3 FA1 1 2 3 FU1 1 2 3 (IDLH)

HA3.2 1 2 3 4 5 DEF HU3.2 1 2 3 4 5 DEF Loss of ANY Two Barriers AND Loss or Potential Loss of Loss or Potential Loss of ANY Two Barriers (Table F-1) ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR ANY Loss or ANY Potential Loss of Containment (Table F-1)

Report or detection of gases in concentration greater than the Report by Local, County or State Officials for evacuation or Third Barrier (Table F-1) RCS (Table F-1)

Lower Flammability Limit within or contiguous to a Safe sheltering of site personnel based on an offsite event Shutdown/Vital Area Table F-1 FISSION PRODUCT BARRIER MATRIX HG1 Security Event Resulting in Loss Of Physical Control of the HS1 Confirmed Security Event in a Plant Vital Area HA4 Confirmed Security Event in a Plant PROTECTED AREA HU4 Confirmed Security Event Which Indicates a Potential Facility Degradation in the Level of Safety of the Plant Fuel Clad Barrier RCS Barrier Primary Containment Barrier HG1.1 1 2 3 4 5 DEF HS1.1 1 2 3 4 5 DEF HA4.1 1 2 3 4 5 DEF HU4.1 1 2 3 4 5 DEF ONE BARRIER AFFECTED A HOSTILE FORCE has taken control of plant equipment Security Supervision reports either of the following: DAEC Security Supervision reports any of the following: Credible Security Threat Loss Potential Loss Loss Potential Loss Loss Potential Loss such that plant personnel are unable to operate equipment - A security event that results in the loss of control in a - Sabotage device discovered in the plant Protected Area. L P L P L P HU4.2 1 2 3 4 5 DEF RADIATION/CORE DAMAGE RADIATION/CORE DAMAGE RADIATION/CORE required to maintain safety functions as indicated by loss of Safe Shutdown/Vital Area (other than the Control Room) - Standoff attack on the Plant Protected Area by a Hostile Fuel damage assessment Drywell Area Hi Range Rad DAMAGE CLAD RCS CNTMT physical control of either: - A confirmed sabotage device discovered in a Safe Force (i.e., sniper). DAEC Security Supervision reports any of the following:

(PASAP 7.2) indicates at Monitor, RIM-9184A or B reading Drywell Area Hi Range Rad

- A Safe Shutdown/Vital Area such that operation of Shutdown/Vital Area - Any of the following security events that persists for 30 - Suspected sabotage device discovered within plant FU1 least 5% fuel clad damage GREATER THAN 5 Rem/hr after Monitor, RIM-9184A or B equipment required for safe shutdown is lost minutes, or greater, affecting the Plant Protected Area: Protected Area. UNUSUAL OR reactor shutdown reading GREATER THAN 1/1 EVENT OR - Credible bomb threats - Suspected sabotage device discovered outside the

- Spent fuel pool cooling systems if imminent fuel Protected Area or in the plant switchyard. Drywell Area Hi Range Rad LEAKAGE 3E+3 Rem/hr

- Hostage/Extortion LEAKAGE LEAKAGE Monitor, RIM-9184A or B OR 1/2 damage is likely (e.g., freshly offloaded reactor core in - Suspicious Fire or Explosion - Confirmed tampering with safety related equipment. Unisolable Main Steamline RCS Leakage is GREATER Failure of both valves in any the pool) - A hostage/extortion situation that disrupts normal plant reading GREATER THAN one line to close and a Torus Area Hi Range Rad FA1

- Significant Security System Hardware Failure Break as indicated by the failure THAN 50 GPM inside the drywell operations. 7E+2 Rem/hr downstream pathway to the Monitor, RIM-9185A or B ALERT

- Loss of Guard Post Contact of both MSIVs in any one line to OR

- Civil disturbance or strike which disrupts normal plant OR environment exists reading GREATER THAN close AND EITHER: Unisolable primary system Security operations. Torus Area Hi Range Rad OR 1E+2 Rem/hr TWO BARRIERS AFFECTED

- High MSL flow or high leakage outside the drywell as

- Internal disturbance that is not short lived or that is not a Monitor, RIM-9185A or B Unisolable primary system OR HS4 Site Attack HA7 Notification of an Airborne Attack steam tunnel temperature indicated by area temps or ARMs harmless outburst involving one or more individuals reading GREATER THAN leakage outside the drywell as Fuel damage assessment annunciators exceeding the Max Normal Limits L P L P L P within the Protected Area. 3E+1 Rem/hr indicated by area temps or (PASAP 7.2) indicates at HS4.1 HA7.1 - Direct report of steam per EOP 3, Table 6.

1 2 3 4 5 DEF 1 2 3 4 5 DEF - Malevolent use of a vehicle outside the Protected Area OR least 20% fuel clad release ARMs exceeding the Max CLAD RCS CNTMT which disrupts normal plant operations. Coolant activity GREATER Safe Limits per EOP 3, Table damage A notification from the site security force that an armed A validated notification from the NRC of an airliner attack THAN 300 µCi/gm DOSE attack, explosive attack, airliner impact, or other HOSTILE threat less than 30 minutes away 6, when Containment Isolation HU4.3 1 2 3 4 5 DEF EQUIVALENT is required.

ACTION is occurring or has occurred within the I-131 PROTECTED AREA A validated notification from the NRC providing information on OR 2/3 HA8 Notification of HOSTILE ACTION within the OCA an aircraft threat Fission Product RPV LEVEL RPV LEVEL RPV LEVEL Primary containment venting RPV LEVEL FS1 RPV Level LESS THAN RPV Level LESS THAN RPV Level LESS THAN per EOPs Primary containment SITE AREA Barriers -25 Inches +15 inches +15 inches flooding required EMERGENCY HA8.1 1 2 3 4 5 DEF PRIMARY CONTAINMENT PRIMARY CONTAINMENT PRIMARY CONTAINMENT THREE BARRIERS AFFECTED A notification from the site security force that an armed ATMOSPHERE ATMOSPHERE ATMOSPHERE attack, explosive attack, airliner impact or other HOSTILE Drywell pressure GREATER Rapid unexplained decrease Torus pressure reaches 53 THAN 2 psig and not caused by following initial increase in psig and increasing L P L P L P ACTION is occurring or has occurred within the OCA.

a loss of DW Cooling pressure OR OR Drywell or Torus H 2 cannot CLAD RCS CNTMT Drywell pressure response not be determined to be LESS HS2 Control Room Evacuation Has Been Initiated and Plant Control HA5 Control Room Evacuation Has Been Initiated Cannot Be Established consistent with LOCA THAN 6% and Drywell or conditions Torus O 2 cannot be Control HS2.1 determined to be LESS 3/3 1 2 3 4 5 DEF HA5.1 1 2 3 4 5 DEF NO Room None None THAN 5% LOSS OF AT Evacuation Control Room evacuation has been initiated Entry into AOP 915 for control room evacuation LEAST 2 BARRIERS?

AND Control of the plant cannot be established per AOP 915 YES within 20 minutes FG1 EMERGENCY DIRECTOR EMERGENCY DIRECTOR EMERGENCY DIRECTOR EMERGENCY DIRECTOR EMERGENCY DIRECTOR EMERGENCY DIRECTOR GENERAL HG2 Other Conditions Existing Which in the Judgment of the HS3 Other Conditions Existing Which in the Judgment of the HA6 Other Conditions Existing Which in the Judgment of the HU5 Other Conditions Existing Which in the Judgment of the JUDGMENT JUDGMENT JUDGMENT JUDGMENT JUDGMENT JUDGMENT EMERGENCY Emergency Director Warrant Declaration of General Emergency Emergency Director Warrant Declaration of Site Area Emergency Emergency Director Warrant Declaration of an Alert Emergency Director Warrant Declaration of a NOUE Any condition in the opinion of Any condition in the opinion of Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion Any condition in the opinion the Emergency Director that the Emergency Director that Emergency Director that indicates Emergency Director that indicates of the Emergency Director of the Emergency Director HG2.1 1 2 3 4 5 DEF HS3.1 1 2 3 4 5 DEF HA6.1 1 2 3 4 5 DEF HU5.1 1 2 3 4 5 DEF indicates Loss or Potential indicates Loss or Potential Loss Loss or Potential Loss of the RCS Loss or Potential Loss of the RCS that indicates Loss or that indicates Loss or Emergency Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Loss of the Fuel Clad Barrier of the Fuel Clad Barrier Barrier Barrier Potential Loss of the Other conditions exist which in the judgment of the Potential Loss of the Director Emergency Director indicate that events are in process or Emergency Director indicate that events are in process or Emergency Director indicate that events are in process or Containment Barrier Emergency Director indicate that events are in process or Containment Barrier Judgment have occurred which involve actual or imminent substantial have occurred which involve actual or likely major failures of have occurred which involve actual or likely potential have occurred which indicate a potential degradation of the core degradation or melting with potential for loss of plant functions needed for protection of the public. Any substantial degradation of the level of safety of the plant. level of safety of the plant. No releases of radioactive material containment integrity. Releases can be reasonably expected releases are not expected to result in exposure levels which Any releases are expected to be limited to small fractions of requiring offsite response or monitoring are expected unless to exceed EPA Protective Action Guideline exposure levels exceed EPA Protective Action Guideline exposure levels the EPA Protective Action Guideline exposure levels further degradation of safety systems occurs offsite for more than the immediate site area beyond the site boundary Duane Arnold Energy Center Duane Arnold Energy Center EAL-01 Emergency Action Level Matrix, Rev. 7 EAL-01 Emergency Action Level Matrix, Rev. 7 Modes: 1 2 3 4 Cold Shutdown 5

Refueling DEF Defueled Modes 1, 2, 3 Approved: Paul Sullivan 12/16/2005 Modes: 1 Power Operation 2

Startup 3 4 5 Refueling DEF Defueled Modes 1, 2, 3 Power Operation Startup Hot Shutdown Hot Shutdown Cold Shutdown Approved: Paul Sullivan 12/16/2005 Manager Emergency Preparedness Date Manager Emergency Preparedness Date Prepared for Nuclear Management Company by: Operations Support Services, Inc. - www.ossi-net.com

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC 7 days inoperable. subsystem to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable. subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

DAEC 3.1-20 Amendment 223

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution is within the limits of Figure 3.1.7-1.

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution is within the limits of Figure 3.1.7-2.

SR 3.1.7.3 Verify temperature of pump suction piping is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> within the limits of Figure 3.1.7-2.

SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of boron in 31 days solution is within the limits of Figure 3.1.7-1. AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to solution AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 (continued)

DAEC 3.1-21 Amendment 223

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate In accordance 26.2 gpm at a discharge pressure 1150 with the psig. Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.8 Verify all heat traced piping between storage 24 months tank and pump suction is unblocked.

AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 DAEC 3.1-22 Amendment 223

Control Rod OPERABILITY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

---

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ----------------NOTE----------------

rod stuck. Rod Worth Minimizer (RWM) may be bypassed as allowed by LOC 3.3.2.1, Control Rod Block Instrumentation, if required, to allow continued operation.

A.1 Verify stuck control Immediately rod separation criteria are met.

AND 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> A.2 Disarm the associated Control Rod Drive (CRD).

AND (continued)

DAEC 3.1-7 Amendment 223

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Perform SR 3.1.3.2 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> each withdrawn from discovery of OPERABLE control Condition A rod. concurrent with THERMAL POWER greater than the Low Power Setpoint (LPSP) of the AND RWM.

A.4 Perform SR 3.1.1.1 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Two or more B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> withdrawn control rods stuck.

C. One or more control C.1 -----------NOTE------------

rods inoperable for RWM may be reasons other than bypassed as allowed Condition A or B. by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control rod.

AND (continued)

DAEC 3.1-8 Amendment 271

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Disarm the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> CRD.

D. ------------NOTE------------ D.1 Restore compliance with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Not applicable when BPWS.

THERMAL POWER

> 10% RTP.

---

OR Two or more inoperable D.2 Restore control rod 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> control rods not in to OPERABLE status.

compliance with Banked Position Withdrawal Sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, C, or D, not met.

OR Nine or more control rods inoperable.

DAEC 3.1-9 Amendment 223

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.1.3.2 ---------------------------NOTE----------------------------

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than 20% RTP.

Insert each withdrawn control rod at least one 31 days notch.

SR 3.1.3.3 Verify each control rod scram time from fully In accordance withdrawn to notch position 04 is with SR 3.1.4.1 7 seconds. and SR 3.1.4.2 SR 3.1.3.4 Verify each withdrawn control rod does not Each time the go to the withdrawn overtravel position. control rod is withdrawn to full out position AND Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling (continued)

DAEC 3.1-10 Amendment 271

Control Rod OPERABILITY 3.1.3 This Page Intentionally Blank per Amendment DAEC 3.1-11 Amendment 271

Control Rod Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 6 OPERABLE control rods shall be slow, in accordance with Table 3.1.4-1; and

b. No more than 2 OPERABLE control rods that are slow shall occupy adjacent locations.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO not met.

SURVEILLANCE REQUIREMENTS

---

NOTE------------------------------------------------------------

During single control rod scram time Surveillances, the Control Rod Drive (CRD) pumps shall be isolated from the associated scram accumulator.

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure 800 psig. 40% RTP after each refueling AND (continued)

DAEC 3.1-12 Amendment 223

Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to exceeding 40% RTP after each reactor shutdown 120 days SR 3.1.4.2 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure 800 psig. 40% RTP after work on control rod or CRD System that could affect scram time AND Prior to exceeding 40%

RTP after fuel movement within the reactor pressure vessel DAEC 3.1-13 Amendment 223

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)

Control Rod Scram Times

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NOTES---------------------------------------------------------

1. OPERABLE control rods with scram times not within the limits of this Table are considered slow.

2. Enter applicable Conditions and Required Actions of LCO 3.1.3, Control Rod OPERABILITY, for control rods with scram times > 7 seconds to notch position

04. These control rods are inoperable, in accordance with SR 3.1.3.3, and are not considered slow.

SCRAM TIMES(a) (seconds) when REACTOR STEAM DOME NOTCH POSITION PRESSURE 800 psig 46 0.44 38 0.93 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

DAEC 3.1-14 Amendment 271

Control Rod Scram Accumulators 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

---

NOTE------------------------------------------------------

Separate Condition entry is allowed for each control rod scram accumulator.

CONDITION REQUIRED ACTION COMPLETION TIME A. One control rod scram A.1 -------------NOTE-------------

accumulator inoperable Only applicable if the with reactor steam dome associated control rod pressure scram time was within 900 psig. the limits of Table 3.1.4-1 during the last scram time Surveillance.

Declare the associated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control rod scram time slow.

OR A.2 Declare the associated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control rod inoperable.

(continued)

DAEC 3.1-15 Amendment 223

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Two or more control rod B.1 Restore charging water 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from scram accumulators header pressure to discovery of inoperable with reactor 940 psig. condition B steam dome pressure concurrent with 900 psig. charging water header pressure

< 940 psig AND B.2.1 ------------NOTE---------------

Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance.

Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rod scram time slow.

OR B.2.2 Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rod inoperable.

(continued)

DAEC 3.1-16 Amendment 223

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify all control rods Immediately upon scram accumulators associated with discovery of charging inoperable with reactor inoperable water header steam dome pressure accumulators are pressure

< 900 psig. fully inserted. < 940 psig AND C.2 Declare the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated control rod inoperable.

D. Required Action and D.1 ------------NOTE------------

associated Not applicable if all Completion Time of inoperable control Required Action B.1 rod scram or C.1 not met. accumulators are associated with fully inserted control rods.

Place the reactor Immediately mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is 940 psig.

DAEC 3.1-17 Amendment 223

FIGURE #13: DAEC Core Map Showing Core Component Location Rev. 6 SD-262.1 SD_261-1.doc Nuclear Fuel and Control Rods