Text

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)
	v • d • e

QF-1030-03 Rev. 7 Retention: Life of plant insurance policy + 10 yr.

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

o QF-1030-26, Instructional and Technical Review Checklist for Examinations o TDAP 1816.2, TSD - Design Phase, Section 5.4 o TDAP 1816.4, TSD - Implementation Phase, Section 5.5.

Key should contain the following:

Learning Objective Number Test Item o Question or Statement o All possible answers o Correct Answer Indicated 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 DESCRIPTION OF CHANGE REASON FOR CHANGE REVIEWER DATE

QF-1030-02 Rev. 4 Retention: 6 years Retain in: Training Records 2009 SRO NRC Master 8-10-09.doc 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:

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-reference:

Level RO SRO 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 Reference(s):

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

Proposed References to be provided to applicants during examination:

EOP-2, T/L & PC/P legs PSPL Curve Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 2 Exam Series A

Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

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

Proposed References to be provided to applicants during examination:

none Learning Objective:

(As available)

Question Source: Bank #

DAEC SRO Bank, Ques 2, pg 166 Modified Bank (Note changes or attach parent)

New Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental Knowledge 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-reference:

Level RO SRO 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 Reference(s):

ATWS-RPV Control, /P-5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

ATWS RPV Control /L without setpoints Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

AOP-915 Rev 39 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

None Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NRC Notification

b. (1)

Recirc mini purge, RHR sample isolation valves & Drywell Equipment Drain Isolation Valves (2) 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months NRC Notification Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 13 Exam Series A

Examination Outline Cross-reference:

Level RO SRO 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months due to the Fire EAL. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification would be selected if the candidate focuses only on the PCIS isolation report, which is an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification would be selected if the candidate focuses only on the PCIS isolation report, which is an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification.

Technical Reference(s):

ACP 1402.3 System Description 959.1 p21 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

ACP 1402.3 Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 14 Exam Series A

Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Technical Reference(s):

OI-831.4, Rev 64, Sect. 6, caution pg 35 EOP-2, step T/T-6 and HCTL curve SD-831.4a, page 51 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Torus Temp leg of EOP-2 and HCTL Curve Learning Objective:

(As available)

Question Source: Bank #

X Modified Bank (Note changes or attach parent)

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

Question History:

Last NRC Exam:

2005 Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

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

Proposed References to be provided to applicants during examination:

none Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

EOP ATWS bases (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

None 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 parent)

New Question History:

Last NRC Exam:

Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Technical Reference(s):

EBD-R page 5 table (EAL bases)

ARP 1C35A C-3.

(Attach if not previously provided)

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 X

(Note changes or attach 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.

Question History:

Last NRC Exam:

Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

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

Proposed References to be provided to applicants during examination:

None Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 29 Exam Series A

Question Cognitive Level:

Memory or Fundamental 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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-reference:

Level RO SRO 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 A:

Incorrect - The alarm is not an indication of an inside the shroud break based upon its tap off point on the LPCS piping. A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO would be required for 2 loops of Core Spray inoperable B:

Incorrect - The alarm is not an indication of an inside the shroud break based upon its tap off point on the LPCS piping.

C:

Incorrect - A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO would be required for 2 loops of Core Spray inoperable D:

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 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 Reference(s):

ARP 1C03A (C-8) Rev 48 TS 3.5.1.B (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

None Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

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

Proposed References to be provided to applicants during examination:

None 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 parent)

New Question History:

Last NRC Exam:

2002 Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This is met with the RPS trip.

Technical Reference(s):

TS 3.3.1.1 TS 3.0.4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NO RPS instrumentation Tables No TS Section 3.0 Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 38 Exam Series A

Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

EOP ATWS Bases Rev 12 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

DO NOT PROVIDE EOP ATWS /P LEG Learning Objective:

(As available)

Question Source: Bank #

DAEC SRO Bank Modified Bank (Note changes or attach parent)

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

TS bases 3.1.7 TS 3.1.7 & figures (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

TS 3.1.7 w/ figures Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 44 Exam Series A

Question Cognitive Level:

Memory or Fundamental 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If the control rod is declared slow, be in MODE 3 within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. The accumulator has a high water level.

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

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If the control rod is declared slow, be in MODE 3 within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The accumulator has a low pressure.

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

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

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

Examination Outline Cross-reference:

Level RO SRO 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 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D:

Incorrect - Per TS 3.1.3 - If declared inoperable, the rod must be fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Technical Reference(s):

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

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

Proposed References to be provided to applicants during examination:

TS 3.1.3, 3.1.4, 3.1.5 Core map Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Technical Reference(s):

AOP 644 Rev 5 SD 644 Rev 9 ARP 1C06A (A-12) Rev 51 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

None 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 parent)

New X

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

OI 730 Rev 100 P&L #9 page 5 ARP 1C26A & B (C-2) Rev 48 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

None Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

RFP 403 Rev 33 Step 5.1.1.e.

(Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

Fuel handling 1.4.1.1.

(As available)

Question Source: Bank #

DAEC 22624 Modified Bank (Note changes or attach parent)

New Question History:

Last NRC Exam:

No Course: 50007 Rev. 0 Topic: Final 2009 SRO NRC Master 8-10-09.doc Page 56 Exam Series A

Question Cognitive Level:

Memory or Fundamental Knowledge X

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-reference:

Level RO SRO 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 Reference(s):

ACP 1407.4 Rev 21 Definitions, Steps 3.5 (3)

(Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE 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 parent)

New X

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental Knowledge X

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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-reference:

Level RO SRO 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of opening it. The requirement is to have administrative control of the valve opening.

Technical Reference(s):

TS LCO 3.0.5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

(As available)

Question Source: Bank #

WTS - 2496 Modified Bank (Note changes or attach parent)

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental Knowledge X

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

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-reference:

Level RO SRO 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 Reference(s):

AOP 149 Rev 31 TS 3.9.7.Bases A.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

(As available)

Question Source: Bank #

Modified Bank (Note changes or attach parent)

New X

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

AOP 672.2 Rev 33 Step 6 EOP - 1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

(As available)

Question Source: Bank #

WTS - 2499 Modified Bank (Note changes or attach parent)

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

Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental 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-reference:

Level RO SRO 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 Reference(s):

EPIP 2.5 Rev 17 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

(As available)

Question Source: Bank #

WTS Modified Bank (Note changes or attach parent)

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

New Question History:

Last NRC Exam:

No Question Cognitive Level:

Memory or Fundamental Knowledge X

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. The shift manning requirements have been fully complied with because the relief STA received a complete turnover within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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-reference:

Level RO SRO 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months not 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D:

Incorrect - The Operations Manager permission is required not the Plant Manager Technical Reference(s):

ACP 1410.1 rev 71 TS 5.2.2.c TS 5.2.2.g (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NONE Learning Objective:

(As available)

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

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

DAEC Modified Bank (Note changes or attach parent)

New Question History:

Last NRC Exam:

2001 Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2

(2) Facility operating limitations in the technical specifications and their bases.

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 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 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

(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 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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier

1.

Individual radiation exposure data to former workers Sec. 19.13(c)

Within 30 days of request or determination of exposure W

Individual MPA(1)

Radiation Protection Manager

2.

Individual radiation exposure data to worker reported to NRC under 20.2202, 20.2203, 20.2204, or 20.2206 Sec. 19.13(d)

At time of transmittal to NRC W

Individual None Radiation Protection Manager

3.

Radiation exposure data to terminating workers Sec. 19.13(e)

At termination upon request of worker W

Individual MPA(1)

Radiation Protection Manager

4.

Respiratory protection program Sec. 20.1703(d) 30 days prior to use of equipment W

RO(1)

DCD(1)

Radiation Protection Manager

5.

Report of excessive radioactive contamination on radioactive material on receipt Sec.

20.1906(d)(1)

Immediately P,T OP CTR Final delivering carrier OSM

6.

Report of excessive radiation levels external to the package on receipt Sec.

20.1906(d)(2)

Immediately P,T OP CTR Final delivering carrier OSM

7.

Report on investigation tracing Radwaste shipment for which Acknowledgment of Receipt not received Sec. 20.2006(d) and App. G,Section III, Paragraph E.2 2 weeks after investigation completed W

RO None Radiation Protection Manager

8.

Theft or loss of licensed material 1000 x App. C to 20.1001-20.2401 Sec.

20.2201(a)(i)

Immediately P

OP CTR None OSM

9.

Theft or loss of licensed material 10 x App. C to 20.1001-20.2401 Sec.

20.2201(a)(ii) 30 days P,T OP CTR None OSM

10. Theft or loss of licensed material Sec. 20.2201(b) 30 days W

RO(1)

Licensee(1) Radiation Protection Manager

11. Additional information on theft or loss information.

Sec. 20.2201(d) Within 30 days of receipt of information W

RO(1)

Licensee(1) Radiation Protection Manager

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

See report

3 OSM

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

See report

3 OSM

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

SUMMARY

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier

14. Reports of exposures of individual, radiation levels, and concentrations of radioactive material exceeding the limits (See Attachment 2)

Sec.

20.2203(a) 30 days W

DCD(1)

RO(1); See report #3 Radiation Protection Manager

15. Report of planned special exposure Sec. 20.2204 30 days W

RO(1)

See report #3 Radiation Protection Manager

16. Reports of individual monitoring Sec. 20.2206 &

19.13(b) 19.13(d)

Annually, covering the preceding year on or before April 30 W

REIRS(1)

Each exposed worker Radiation Protection Manager

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

18. Failure to comply or existence of a defect Sec. 21.21(b) 5 days W

NMSS or NRR (3)

DCD(1)

Chairman Part 21 Evaluation Committee

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

RO(1)

DCD(1)

Radiation Protection Manager

20. Transfer of device to specific licensee Sec. 31.5(c)(8) 30 days W

NMSS(1)

None Radiation Protection Manager

21. Transfer of device to general licensee Sec. 31.5 (c)(9)(i) 30 days W

NMSS(1)

None Radiation Protection Manager

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

NMSS(1)

RO(1)

Reactor Engineering Supervisor

23. Change to registration Sec. 40.25 (c)(2) 30 days W

NMSS(1)

RO(1)

Reactor Engineering Supervisor

24. Registration certificate-filed by transferor Sec. 40.25 (d)(3)

Promptly (1)

W Transferee DCD(1)

Reactor Engineering Supervisor

25. Registration certificate-transfer Sec. 40.25 (d)(4) 30 days W

NMSS(1)

RO(1)

Reactor Engineering Supervisor

26. Transfer of material licensed under Sec. 40.25 Sec. 40.35(d)

(1)

Promptly W

Receiver None Reactor 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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier

27. Transfer of material licensed under Sec. 40.25 Sec. 40.35 (e)(1) Quarterly W

NMSS(1)

None Reactor Engineering Supervisor

28. Transfer of devices under Agreement State regulations equivalent to Sec. 40.25 Sec. 40.35 (e)(2) Quarterly W

State Agency*

DCD(1)

Reactor Engineering Supervisor

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

30. Nuclear Material Transaction Report Form DOE/NRC-741 filed by shipper Sec. 40.64(a)

Promptly W

DOE(1)

Receiver(3) Reactor Engineering Supervisor

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

DOE(1)

Shipper(1)

Reactor Engineering Supervisor

32. Statement of source material inventory Sec. 40.64(b)

Annually W

DOE(1)

None Reactor Engineering Supervisor

33. Unlawful diversion of source material Sec. 40.64(c)

Promptly P,T RO None OSM

34. Unlawful diversion of source material Sec. 40.64(c) 15 days W

RO(1)

NMSS(1)

Reactor Engineering Supervisor Sec. 50.9(b) 2 working days of identification RO None

35. Identify information having a significant implication for public health and safety or common defense and security Sec. 72.11(b) 2 working days of identification RO None Sec. 50.36a (a)(2), Tech Specs Annually W

DCD(1)

RO(1)

Resident(1)

Radiation Protection Manager

36. Effluent releases report Annually (non-significant)

W DCD(1)

RO(1)

Resident (1)

Licensing Manager

37. Loss-of-Coolant Accident Evaluation model changes or errors report Sec. 50.46(a)(3) 30 days (significant)

W NRR (50.73 DCD (73.71)

RO(1)

Resident(1)

Licensing Manager

38. Changes in security plan made without prior approval Sec. 50.54(p)

Two months after change W

DCD RO(1)

Security 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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier Sec. 50.54(q) 30 days after change or proposed to NRC W

DCD(1)

RO(2)

Resident(1)

Emergency Planning Manager

39. Changes in emergency plan made without prior approval Sec 72.44(f) 6 months after change W

DCD(1)

RO(2)

Resident(1)

NMSS Emergency Planning Manager Sec. 50.54(cc)

Immediate W

RO None Legal

40. Filing for bankruptcy under Chapter 11 Sec. 72.44(b)(6)(i)

Immediate W

RO None Legal Sec. 50.59(b) 6 months after Refueling Outage not to exceed 24 months W

DCD(1)

RO(1)

Resident(1)

Licensing Manager

41. Facility changes, tests, and experiments conducted without prior approval Sec.72.48(d)(2)

Once every 24 months W

DCD(1)

RO(1)

Resident(1)

Licensing Manager Sec. 50.71(b)

Annually W

DCD(1)

RO(1)

Resident(1)

Licensing Manager

42. Financial report 72.80(b)

Annually W

DCD(1)

RO(1)

Resident(1)

Licensing Manager

43. FSAR updating Sec. 50.71(e) 6 months after RFO not to exceed 24 months W

NRR(11)

RO(1)

Resident(1)

Licensing Manager Part 50, App. E, Sec.IV.D.3 15 minutes P

S&L Gov.**

NRC OSM

44. Emergency Notifications Sec. 72.75(a)

Prompt (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

P S&L Gov. **

OP CTR OSM

45. Immediate Notification Events (Non-Emergency)

Sec. 50.72 Prompt (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

P OP CTR None OSM

46. Immediate Notification Events Sec. 50.72 Prompt (4 or 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

P OP CTR None OSM Sec. 72.75(b)(1-2)

Prompt (4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months )

P OP CTR None OSM Sec. 72.75(c)(1-3)

Prompt (8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

P OP CTR None OSM

47. Non-emergency Notifications Sec.72.75(d)(1) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months P

OP CTR None OSM Sec. 50.73 60 days W or P DCD/

OP CTR RO(1)

Licensing Manager

48. Licensee Event Report Sec. 73.71 60 days W

DCD RO(1)

SFPO NSIR Licensing Manager

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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier

50. Report on status of Decommissioning Funding Sec. 50.75(f)(1)

On a calendar year basis by March 31,1999 and at least once every 2 years thereafter W

DCD RO(1)

Resident Licensing Manager

51. Fracture toughness Part 50, App. G At least 3 years prior to date when the predicted fracture toughness levels will no longer satisfy requirements of Appendix G

W DCD(1)

RO(1)

Resident(1)

Program Engineering Manager

52. Report of test results of specimens withdrawn from capsules (fracture toughness tests)

Part 50, App. H Sec. III.A, Tech Specs Variable W

DCD(1)

RO(1)

Resident(1)

Program Engineering Manager

53. Report of effluents released in excess of design objectives Part 50, App. I.,

Sec. IV.A.

Within 30 days from end of quarter W

RO(1)

DCD(1)

Radiation Protection Manager

54. Reactor containment building integrated leak rate test (includes LLRT)

Summary Report Part 50, App.J, Sec. V.B, 3 months after conducting test W

Available onsite System Engineering Manager

55. Notification of disability Sec. 55.25 30 days W

NRR None Operations Manager

56. Medical examination Sec. 55.21 W

Licensee***

RO None

Manager, Training Coordinator Sec. 70.52 Prompt (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

P OP CTR None OSM Sec. 72.74 Prompt (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

P OP CTR None OSM

57. Accidental Criticality or Loss of Special Nuclear Material Sec. 73.71 Prompt (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

P OP CTR None OSM

58. Material Status Report Sec. 74.13 Within 60 days of the beginning of the physical inventory W

NMSS Licensee Reactor Engineering Supervisor Sec. 74.15 Upon transfer or receipt W

NMSS Licensee Reactor Engineering Supervisor

59. Nuclear Material Transaction Reports Sec. 72.78(a)

Upon transfer or receipt W

NMSS Licensee Reactor Engineering Supervisor

60. Reduction in Effectiveness of Package Sec. 71.95 30 days W

NMSS None Radiation 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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier Sec.

72.212(b)(1)(i)

Notify NRC 90 days prior to first storage of spent fuel in cask type under general license W

DCD RO Licensing Manager

61. 72.4 Notifications Sec.

72.212(b)(1)(ii)

Register use of each cask no later than 30 days after using cask to store spent fuel W

DCD RO Licensing Manager

62. Proof of financial protection Sec. 140.15(a)

As required W

NRR or NMSS(3)

None Legal

63. Change in proof of financial protection Sec. 140.15(e)

Promptly W

NRR or NMSS(2)

None Legal Sec. 140.15 (b)(1)

Annually W

DCD NMSS(3)

RO(1)

Resident(1)

Licensing

64. Financial statement Sec.72.80(b)

Annually W

DCD RO(1)

Resident(1)

Licensing

65. Policy renewal termination of policy Sec. 140.17(b) 30 days prior to termination of policy W

NRR or NMSS(1)

None Legal

66. Guarantee of payment of deferred premiums Sec. 140.21 Annually W

NRR or NMSS(1)

None Legal

67. Transfer of assets >1%

of net utility value Sec 50.33(k)

As required W

NRR None Legal

68. Startup of Reactor Tech Specs Within (1) 90 days following completion of the startup test 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.

W RO(2)

Licensee(36) Licensing Manager

69. Not Used

70. Annual Radiological Environmental Operating Report TS 5.6.2 Annually, by May 1 W

RO(1)

DCD(18)

Radiation Protection Manager

71. Not Used

72. Core Operating Limits Report TS 5.6.5 Upon Issuance W

DCD(1)

RO(1)

Resident(1)

Licensing Manager

73. Annual Radioactive Material Release Report TS 5.6.3 Annually, by May 1 W

RO(1)

DCD(18)

Radiation Protection Manager

74. PAM Instrumentation Inoperability TS 3.3.3.1 14 days W

DCD RO(1)

Licensing 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

Report Required by Timing Method Primary Recipient Secondary Recipient Responsible Notifier

75. Low Level Waste Mishaps NRC GL 91-02 30 Days W

LWM None Radiation Protection Manager

76. ISI Summary Report ASME Section XI, IWA-6230 Within 90 days of completion of ISI examinations during refueling outages W

DCD(1)

RO(1)

Resident(1)

Licensing Manager

77. Horizontal Storage Module Dose Rates Exceeded ISFSI-61BT TS 1.2.7 30 Days W

DCD(1)

RO(1)

Resident(1)

SFPO Licensing Manager

78. Transfer Cask Dose Rates ISFSI-61BT TS 1.2.11 30 Days W

DCD(1)

RO(1)

Resident(1)

SFPO Licensing Manager

79. Highest Heat Load to Date of any 61BT Dry Storage Canister****

ISFSI-61BT TS 1.1.7 30 Days W

DCD RO(1)

Resident(1)

SFPO Licensing Manager

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

NRR NMSS Licensing Manager

81. NRC Form 748 National Source Tracking Transaction Report 10CFR20.2207 Annually by January 31 W,T LM None Radiation Protection 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 U.S. Nuclear Regulatory Commission Mail Station 0-P1-17 (zero-P1-17)

Washington, D.C. 20555 DOE U.S. Department of Energy P.O. Box E Oak Ridge, TN 37830 EDO Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555 GC General Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 IE Director, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTN: Document Control Desk SFPO Director, Spent Fuel Project Office U.S. Nuclear Regulatory Commission Washington, D.C. 20555 IP Assistant Director, Export-Import and International Safeguards Office of International Programs U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NSIR Director, Division of Nuclear Security Office of Nuclear Security and Incident

Response

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 MPA Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 LM Lockheed Martin NSTS Help Desk 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 Formatted: Left

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

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

YES, upon 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).

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

YES Any event or condition that resulted in the condition of the plant, including its 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)]

YES 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 nuclear power plant. [50.73(a)(2)(iii)]

NO Any event or condition that resulted in a manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) (DAEC specific list provided in section 3.2.1 of this procedure), except when:

YES, for all valid actuations 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 EVENT IMMEDIATE NOTIFICATION EVENT

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. [50.73(a)(2)(v)(A)]

YES

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

YES

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

YES

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

YES 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. [50.73(a)(2)(vii)(A)]

NO 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 remove residual heat. [50.73(a)(2)(vii)(B)]

NO 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 control the release of radioactive material.

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

NO 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 mitigate the consequences of an accident.

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

NO

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

- 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)]

NO

- 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) of all radionuclides except tritium and dissolved noble gases.

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

NO 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: (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)]

NO Any event 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 nuclear power plant including fires, toxic gas releases, or radioactive releases. [50.73(a)(2)(X)]

NO Discovery of loss of any shipment of Special Nuclear Material or spent fuel, or recovery of same. (Security-related) [73.71(a)(4)]

YES Any event in which there is reason to believe a person has committed, attempted 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)]

YES Any event in which there is reason to believe a person has committed, 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. (Security-related) [App G to Part 73, I(a)(2)]

YES Any event in which there is reason to believe a person has committed, or 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 (Security-related) [App G to Part 73, I(a)(3)]

YES

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

YES Any failure, degradation, or discovered vulnerability in a safeguard system 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)]

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

YES Any lost, stolen or missing licensed material in an aggregate quantity equal 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))

YES

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

YES 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 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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))

YES

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

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

YES 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 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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))

YES Within 30 days after the occurrence of any lost, stolen or missing licensed 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))

YES

- - Doses in excess of the occupational dose limits for adults in 20.1201.

(20.2203(a)(2)(i))

NO

- - Doses in excess of the occupational dose limits for minors in 20.1207.

(20.2203(a)(2)(ii)

NO

- - Doses in excess of the limits for an embryo/fetus of a declared pregnant woman in 20.1208. (20.2203(a)(2)(iii))

NO

- - Doses in excess of the limits for an individual member of the public in 20.1301. (20.2203(a)(2)(iv))

NO

- - Doses in excess of any applicable limit in the DAEC license.

(20.2203(a)(2)(v))

NO

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 35 of 59 ATTACHMENT 2 Page 6 of 6 REPORTABLE EVENTS EVENT IMMEDIATE NOTIFICATION EVENT Levels of radiation or concentrations of radioactive material in a restricted area in excess of any applicable limit in the DAEC license.

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

NO Levels of radiation or concentrations of radioactive material in an 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))

NO Levels of radiation or releases of radioactive material in excess of the 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))

NO 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 Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Declaration of any of the Emergency Action Levels as listed in EPIP 1.1. (50.72(a)(1)(i))

Notify State and local authorities within 15 minutes of declaration of and EAL, NRC immediately afterwards (in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of event) and management immediately following. (See EPIP 1.2)

The initiation of any nuclear plant shutdown required by Tech. Specs.

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

No Yes No No OSM Any deviation from the Tech. Specs.

authorized pursuant to 10 CFR 50.54(x). (50.72(b)(1))

Yes No No No OSM Any event or condition that results in the condition of the nuclear power plant including its principal safety barriers, being seriously degraded (50.72(b)(3)(ii)(A))

No No Yes No OSM Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

(50.72(b)(3)(ii)(B))

No No Yes No OSM

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 37 of 59 ATTACHMENT 3 Page 2 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event that results or should have 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))

No Yes No No OSM Any event that results in a major loss 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months meets this criteria. (50.72(b)(3)(xiii))

No No Yes No OSM Receipt of a radioactive material 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))

Yes No No No OSM Any lost, stolen, or missing licensed 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))

Yes No No No OSM

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 38 of 59 ATTACHMENT 3 Page 3 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event involving by-product, 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))

Yes No No No OSM Any event involving by-product, 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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))

Yes No No No OSM Any incident in which an attempt has 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))

Yes No No No OSM Any Accidental criticality or loss of Special Nuclear Material. (70.52(a))

Yes No No No OSM

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 39 of 59 ATTACHMENT 3 Page 4 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event of condition that at the 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))

No No Yes No OSM Any event or condition that at the 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))

No No Yes No OSM Any event or condition that at the 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))

No No Yes No OSM Any event or condition that at the 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))

No No Yes No OSM

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 40 of 59 ATTACHMENT 3 Page 5 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

(50.72(b)(2)(iv)(B)).

No Yes No No OSM Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of 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)

No No Yes No OSM See Section 3.2 for a specific list of systems

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 41 of 59 ATTACHMENT 3 Page 6 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment. (50.72(b)(3)(xii))

No No Yes No OSM Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an on-site fatality or inadvertent release of radioactively contaminated materials.

(50.72(b)(2)(xi))

No Yes No No OSM If security-related, see section 3.1(8) and/or the DAEC Security Event Reporting Procedure Any event involving by-product, 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))

No No No Yes OSM

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 42 of 59 ATTACHMENT 3 Page 7 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event involving by-product, 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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))

No No No Yes OSM Discovery of loss of any shipment of Special Nuclear Material or spent fuel, or recovery of same. (73.71(a))

Yes No No No Sec. Sup. See Security Procedure 11 Any event in which there is reason to believe a person has committed, 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))

Yes No No No Sec. Sup. See Security Procedure 11 Any event in which there is reason to believe a person has committed, 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))

Yes No No No Sec. Sup. See Security Procedure 11

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 43 of 59 ATTACHMENT 3 Page 8 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Any event in which there is reason to believe a person has committed, 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))

Yes No No No Sec. Sup. See Security Procedure 11 An actual entry of an unauthorized person into a protected, material access, controlled access, vital or transport areas. (73.71, App. G.,

I.(b))

Yes No No No Sec. Sup. See Security Procedure 11 Any failure, degradation, or discovered vulnerability in a 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))

Yes No No No Sec. Sup. See Security Procedure 11 Actual or attempted introduction of contraband into a protected, material access, vital or transport area.

(73.71, App. G., I.(d))

Yes No No No Sec. Sup. See Security Procedure 11 Any event that meets the reportability criteria of 10 CFR 26.73 (Fitness for Duty) as described in Security Directives. (10 CFR 26.73)

No No No Yes Sec. Sup. See Procedure FFD-7

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 44 of 59 ATTACHMENT 3 Page 9 of 9 IMMEDIATE NOTIFICATION EVENTS Event NRC 1 HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE Within 30 days after the occurrence of any lost, stolen or missing licensed 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))

No No No No OSM Thirty Day Telephone Report per 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 Notification Event (50.72)

LER (50.73)

Immediate Notification Event (50.72)

LER (50.73)

Critical 4 Hour Report per 50.72(b)(2)(iv)(B) 60 Day LER per 50.73(a)(2)(iv)(A) 4 Hour Report per 50.72(b)(2)(iv)(B) 60 Day LER per 50.73(a)(2)(iv)(A)

Critical (preplanned)

No Report No Report No Report No Report Non-Critical 8 Hour report per 50.72(b)(3)(iv)(B) 60 Day LER per 50.73(a)(2)(iv)(A)

No Report 60 Day Telephone Report per 50.73(a)(1)

Non-Critical (preplanned)

No Report No Report No Report No Report

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 Event NRC 1

HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE The discovery of accidental criticality or any loss of special nuclear material. (72.74(a))

Yes No No No OSM Declaration of any of the Emergency Action Levels as listed in EPIP 1.1 Attachment 1.

(72.75(a))

Notify State and local authorities within 15 minutes of declaration of an EAL, NRC immediately afterwards (in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of event) and management immediately following. (See EPIP 1.2)

An action taken in an emergency 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))

No Yes No No OSM Any event or situation related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other Government agencies has been or will be made. (72.75(b)(2))

No Yes No No OSM A defect in any spent fuel storage structure, system, or component which is important to safety. (72.75(c)(1))

No No Yes No OSM A significant reduction in the effectiveness of any spent fuel storage confinement system during use. (72.75(c)(2))

No No Yes No OSM

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 Event NRC 1

HOUR NRC 4 HOUR NRC 8 HOUR NRC 24 HOUR RESP.

NOT.

NOTE An event that requires transport of a radioactively contaminated person to an offsite medical facility for treatment.

(72.75(c)(3))

No No Yes No OSM An event in which important to 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))

No No No Yes OSM

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 No.

Contact Representative Organization Business Address Contact Phone Number Notation 1

Bureau Chief Bureau of Radiological Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, Iowa 50319-0073 (515)281-3478 2

Linn County Public Health Director Public Health Department Linn County, Iowa 501 13th Street NW Cedar Rapids, IA 52405 (319)892-6000 3

Iowa DNR Emergency Response Unit Iowa DNR Emergency Response Unit 401 SW 7th Street, Suit I Des Moines, Iowa 50309 (515)281-8694 Fax: (515)725-0218 http://www.iowadnr.com/spills/rep ort.html 4

Environmental Corporate Functional Area Manager FPL/FPLENextEra Energy 700 Universe Blvd ENG/JB Juno Beach, FL 33408 (603)773-7438 (W)*

(603)765-7291 ( c)

CFAM RP & Chemistry

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 Representative Organization Business Address Contact Phone Number Notation 5

FPL/FPLE NextEra Energy Communications Representative DAEC Communications Rep.

FPL Communications Rep.

FPL Energy Duane Arnold LLC 3277 DAEC Road Palo, Iowa 52324 700 Universe Blvd.

Juno Beach, FL 33408 (319)851-7140 (603)773-7281 (W)

(603)765-6444 (C) 6 FPL Risk Management Rep Risk Management 700 Universe Blvd.

Juno Beach, FL 33408 (561)371-5210 or (561)691-3030 7

ANI Account Engineer ANI Account Engineer 95 Glastonbury Blvd Glastonbury, CT 06033 (860)682-1301 8

NEI Representative Senior Manager, Environmental Protection 1776 I Street NW, Suite 400 Washington, DC 20006 (202)739-8000 GW_Notice@nei.org E-mail is preferred method of contact 9

Radiation Protection Manager Site: Radiation Protection and Chemistry 10 Environmental Site Function al Area Manager Site: RP/Chem Technical Staff Supervisor 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

State Officials Iowa DNR Emergency Response Unit 401 SW 7th 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 NG-005F Rev. 0 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 _________________________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 51 of 59 NG-005F Rev. 0 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:_______________________________________________________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 52 of 59 NG-006F Rev. 0 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:________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 53 of 59 NG-007F Rev. 1 (Rev. ACP 1402.3)

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:________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 54 of 59 NG-008F Rev. 0 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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:________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 55 of 59 NG-009F Rev. 0 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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:________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 56 of 59 NG-010F Rev. 0 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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:________

ADMINISTRATIVE CONTROL PROCEDURE ACP 1402.3 REGULATORY REPORTING ACTIVITIES Rev. 38 Page 57 of 59 NG-001A Rev. 5 ATTACHMENT-13 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.

Formatted: Font: 11 pt

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 NG-012F Rev. 0 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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

Natural &

Destructive Phenonenon Fire or Explosion Control Room Evacuation Hazards Abnormal Rad Release Rad Effluent Offsite Rad Conditions Onsite Rad Conditions Toxic and Flammable Gas GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Approved:

Paul Sullivan 12/16/2005 Manager Emergency Preparedness Date Duane Arnold Energy Center EAL-01 Emergency Action Level Matrix, Rev. 7 GENERAL EMERGENCY SITE AREA EMERGENCY UNUSUAL EVENT ALERT Prepared for Nuclear Management Company by: Operations Support Services, Inc. - www.ossi-net.com System Malfunct.

Modes:

1 Power Operation Hot Shutdown Cold Shutdown Refueling Defueled 2

Startup 4

5 3

DEF Fission Product Barriers Modes 1, 2, 3 Approved:

Paul Sullivan 12/16/2005 Manager Emergency Preparedness Date Duane Arnold Energy Center EAL-01 Emergency Action Level Matrix, Rev. 7 Modes:

1 Power Operation Hot Shutdown Cold Shutdown Refueling Defueled 2

Startup 4

5 3

DEF Modes 1, 2, 3 Security Emergency Director Judgment Loss of Power RPS Failure Inability to Reach or Maintain Shutdown Conditions Inst. /

Comm.

Fuel Clad Degradation RCS Leakage Inadvertent Criticality ISFSI Events Cask Confine.

Boundary Security RPV LEVEL Primary containment flooding required 2

RU1.1 Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading that exceeds 1 E-3 µCi/cc and is expected to continue for 60 minutes or longer RU1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 2.0 E-1 µCi/cc and is expected to continue for 60 minutes or longer Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4 AND ANY ONE OF THE FOLLOWING:

Restoration of power to either Bus 1A3 or 1A4 is not likely within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months RPV level is indeterminate RPV Level is LESS THAN +15 inches SA5.1 AC power capability to 1A3 or 1A4 busses reduced to a single power source for greater than 15 minutes AND Any additional single failure will result in station blackout SG1.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emer-gency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4 AND Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power RA1.1 Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading that exceeds 3 E-2 µCi/cc and is expected to continue for 15 minutes or longer RS1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 100 mRem TEDE or 500 mRem thyroid CDE at or beyond the site boundary. (Preferred method)

RS1.2 If Dose Assessment is unavailable, any of the following:

Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6 E-2 µCi/cc and is expected to continue for 15 minutes or longer.

Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4 E+1 µCi/cc and is expected to continue for 15 minutes or longer RA1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 6 E+0 µCi/cc and is expected to continue for 15 minutes or longer RA1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1 E-1 µCi/cc and is expected to continue for 15 minutes or longer RU2.1 RU2.1 Unplanned valid Refuel Floor ARM reading increase with an uncontrolled loss of reactor cavity, fuel pool, or fuel transfer canal water level with all irradiated fuel assemblies remaining covered by water as indicated by any of the following:

Report to control room Valid fuel pool level indication (LI-3413) LESS THAN 36 feet and lowering

- Valid WR GEMAC Floodup indication (LI-4541) coming on scale RA2.1 Report of any of the following:

- Valid ARM Hi Rad alarm for the Refueling Floor North End (RM 9163), Refueling Floor South End (RM 9164), New Fuel Storage (RM 9153), or Spent Fuel Storage Area (RM 9178).

- Valid Refueling Floor North End (RM-9163), Refueling Floor South End (RM-9164), or New Fuel Storage Area (RM-9153) ARM Reading GREATER THAN 10 mRem/hr

- Valid Spent Fuel Storage Area (RM-9178) ARM Reading GREATER THAN 100 mRem/hr RA3.1 Valid area radiation levels GREATER THAN 15 mRem/hr in any of the following areas:

- Control Room (RM 9162)

- Central Alarm Station (by survey)

- Secondary Alarm Station (by survey)

HS1.1 HA1.1 Receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35 (+/- 0.06 gravity)

HG1.1 A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions as indicated by loss of physical control of either:

A Safe Shutdown/Vital Area such that operation of equipment required for safe shutdown is lost OR Spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly offloaded reactor core in the pool)

RG1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 1000 mRem TEDE or 5000 mRem thyroid CDE at or beyond the site boundary. (Preferred method)

If Dose Assessment is unavailable, either of the following:

Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6 E-1 µCi/cc and is expected to continue for 15 minutes or longer.

Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4 E+2 µCi/cc and is expected to continue for 15 minutes or longer RG1.2 HU4.1 DAEC Security Supervision reports any of the following:

- Suspected sabotage device discovered within plant Protected Area.

- Suspected sabotage device discovered outside the Protected Area or in the plant switchyard.

- Confirmed tampering with safety related equipment.

- A hostage/extortion situation that disrupts normal plant operations.

- Civil disturbance or strike which disrupts normal plant operations.

- Internal disturbance that is not short lived or that is not a harmless outburst involving one or more individuals within the Protected Area.

- Malevolent use of a vehicle outside the Protected Area which disrupts normal plant operations.

HA4.1 HU1.1 Earthquake detected per AOP 901, Earthquake HU2.1 Fire in buildings or areas contiguous to any Safe Shutdown/Vital Area not extinguished within 15 minutes of control room notification or verification of a control room alarm HA2.1 Fire or explosion in any Safe Shutdown/Vital Area AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area HU3.1 Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect normal plant operations HA3.1 Report or detection of toxic gases within or contiguous to a Safe Shutdown/Vital Area in concentrations that may result in an atmosphere Immediately Dangerous to Life and Health (IDLH)

None None HU5.1 HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs Entry into AOP 915 for control room evacuation HA5.1 HA5 Control Room Evacuation Has Been Initiated HS2.1 Control Room evacuation has been initiated AND Control of the plant cannot be established per AOP 915 within 20 minutes HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant.

Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels HA6.1 HA6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert HS3.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency HG2.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area HG2 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency HU3.2 Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event HA3.2 Report or detection of gases in concentration greater than the Lower Flammability Limit within or contiguous to a Safe Shutdown/Vital Area HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant HA4 Confirmed Security Event in a Plant PROTECTED AREA HS1 Confirmed Security Event in a Plant Vital Area HG1 Security Event Resulting in Loss Of Physical Control of the Facility HU4.2 Credible Security Threat DAEC Security Supervision reports any of the following:

- Sabotage device discovered in the plant Protected Area.

- Standoff attack on the Plant Protected Area by a Hostile Force (i.e., sniper).

- Any of the following security events that persists for 30 minutes, or greater, affecting the Plant Protected Area:

- Credible bomb threats

- Hostage/Extortion

- Suspicious Fire or Explosion

- Significant Security System Hardware Failure

- Loss of Guard Post Contact Security Supervision reports either of the following:

A security event that results in the loss of control in a Safe Shutdown/Vital Area (other than the Control Room)

A confirmed sabotage device discovered in a Safe Shutdown/Vital Area None RA3.2 Valid area radiation monitor (RE-9168), reading GREATER THAN 500 mRem/hr affecting the Remote Shutdown Panel, 1C388 HU1 Natural and Destructive Phenomena Affecting the Protected Area HU1.2 Report of a tornado touching down within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern HU1.3 Report of winds greater than 95 mph within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern HU1.4 Vehicle crash into plant structures or systems within the Plant Protected Area with NO confirmed damage to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern HU1.5 Report of an unanticipated explosion within the Plant Protected Area resulting in visible damage to permanent structures or equipment HU1.6 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals HU1.7 River level ABOVE 757 feet HA1 Natural and Destructive Phenomena Affecting the Plant Vital Area HA1.2 Report of Tornado or high winds greater than 95MPH within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern HA1.3 Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to a Safe Shutdown/Vital Area or Control Room indication of degraded performance of a System of Concern HA1.4 Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of a Safe Shutdown/Vital Area HA1.5 River level ABOVE 767 feet HA1.6 Uncontrolled flooding in a Safe Shutdown/Vital Area that results in degraded safety system performance as indicated in the Control Room or that creates an industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment HU2 Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown HA3 Release of Toxic or Flammable Gases Within or Contiguous to a Vital Area Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant None None HU1.8 Uncontrolled flooding in a Safe Shutdown/Vital Area that has the potential to affect safety related equipment needed for the current operating mode HU1.9 River level BELOW 725 feet 6 inches HA1.7 River level BELOW 724 feet 6 inches HA1.8 Report to control room of VISIBLE DAMAGE affecting a Safe Shutdown/Vital Area RU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue For 60 Minutes or Longer RA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue for 15 Minutes or Longer RU2 Unexpected Increase in Plant Radiation RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel RU1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1.0 E-3 µCi/cc and is expected to continue for 60 minutes or longer Any unplanned ARM reading offscale high or GREATER THAN 1000 times normal* reading

Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value Valid water level reading LESS THAN 450 inches as indicated on LI-4541 (floodup) for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering RA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown RU1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+3 CPS and is expected to continue for 60 minutes or longer RU1.5 Valid RHRSW & ESW rad monitor (RM-1997) reading that exceeds 8E+2 CPS and is expected to continue for 60 minutes or longer RU1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+3 CPS and is expected to continue for 60 minutes or longer RU1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of 2 times ODAM limits and is expected to continue for 60 minutes or longer RA1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+5 CPS and is expected to continue for 15 minutes or longer RA1.5 Valid RHRSW & ESW rad monitor (RM-1997) reading that exceeds 8E+4 CPS and is expected to continue for 15 minutes or longer RA1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+5 CPS and is expected to continue for 15 minutes or longer RA1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration expected to continue for 15 minutes or longer in excess of 200 times ODAM limit Valid Fuel Pool water level indication (LI-3413) LESS THAN 16 feet that will result in Irradiated Fuel uncovering RS1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem TEDE or 500 mRem CDE Thyroid for the Actual or Projected Duration of the Release RG1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mRem TEDE or 5000 mRem CDE Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology RS1.3 Field survey results indicate closed window dose rates exceeding 100 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 500 mRem for one hour of inhalation at or beyond the site boundary Field survey results indicate closed window dose rates exceeding 1000 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 5000 mRem for one hour of inhalation at or beyond the site boundary RG1.3 RU2.2 RA2.2 RA2.3 None None None None None None None FS1 Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier (Table F-1)

Loss or Potential Loss of ANY Two Barriers (Table F-1)

FA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS (Table F-1)

FU1 ANY Loss or ANY Potential Loss of Containment (Table F-1)

FG1 SU4 Fuel Clad Degradation SU4.1 Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm SU4.2 Reactor Coolant sample activity value GREATER THAN 2.0 µCi/gm dose equivalent I-131 SU5 RCS Leakage SU5.1 Unidentified or pressure boundary leakage GREATER THAN 10 gpm SU5.2 Identified leakage GREATER THAN 25 gpm SU8 Inadvertent Criticality SU8.1 An UNPLANNED extended positive period observed on nuclear instrumentation EU1 Damage To A Loaded Cask Confinement Boundary EU1.1 Any one of the following natural phenomena events with resultant visible damage to or loss of a loaded cask confinement boundary:

Report by plant personnel of a tornado strike Report by plant personnel of a seismic event EU1.2 The following accident condition with resultant visible damage to or loss of a loaded cask confinement boundary:

A loaded transfer cask is dropped as a result of normal handling or transporting EU1.3 Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask confinement boundary SS1.1 SU1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes SU1.1 Loss of power to or from the Startup or Standby Transformer resulting in a loss of all offsite power to Emergency Busses 1A3 and 1A4 that is expected to last for greater than 15 minutes AND Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators SA5 AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout SS1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses Loss of Div 1 and Div 2 125V DC busses based on bus voltage LESS THAN 105 VDC indicated for greater than 15 minutes SS3.1 SS3 Loss of All Vital DC Power SA2.1 Auto Scram failure AND ANY of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

SS2.1 SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful SU2.1 Plant is not brought to required operating mode within applicableTechnical Specifications LCO Action Statement Time EOP Graph 4 Heat Capacity Limit is exceeded SS4.1 SS4 Complete Loss of Heat Removal Capability SU3 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes SU3.1 Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Safety Systems for greater than 15 minutes SA4.1 Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Safety Systems for greater than 15 minutes AND Either of the following conditions exist:

A significant plant transient is in progress.

Compensatory non-alarming indications are unavailable SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators Unavailable Significant transient in progress and ALL of the following:

Loss of most or all annunciators on Panels 1C03, 1C04 and 1C05.

Compensatory non-alarming indications are unavailable.

Indicators needed to monitor criticality, or core heat removal, or Fission Product Barrier status are unavailable.

SS6.1 SS6 Inability to Monitor a Significant Transient in Progress SU6 UNPLANNED Loss of All Onsite or Offsite Communications Capabilities SU6.1 Loss of ALL of the following onsite communication capa-bilities affecting the ability to perform routine operation:

Plant Operations Radio System In-Plant Telephones Plant Paging System SU6.2 Loss of ALL of the following offsite communications capability:

All telephone lines (commercial)

Microwave Phone System FTS Phone System SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

AND Loss of adequate core cooling or decay heat removal capability as indicated by either:

RPV level cannot be maintained GREATER THAN -25 inches HCL Curve (EOP Graph 4) exceeded SG2.1 None SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core SU2 Inability to Reach Required Shutdown Within Technical Specification Limits None None None None None None None None None None None None None None None None HU4.3 A validated notification from the NRC providing information on an aircraft threat HA7 Notification of an Airborne Attack HA8 Notification of HOSTILE ACTION within the OCA HA7.1 A validated notification from the NRC of an airliner attack threat less than 30 minutes away HA8.1 A notification from the site security force that an armed attack, explosive attack, airliner impact or other HOSTILE ACTION is occurring or has occurred within the OCA.

HS4 Site Attack HS4.1 A notification from the site security force that an armed attack, explosive attack, airliner impact, or other HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA None 5

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4 EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier Table F-1 FISSION PRODUCT BARRIER MATRIX Primary Containment Barrier Fuel Clad Barrier RCS Barrier Loss Potential Loss Loss Potential Loss RPV LEVEL RPV Level LESS THAN

-25 Inches RADIATION/CORE DAMAGE Fuel damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage OR Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 7E+2 Rem/hr OR Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 3E+1 Rem/hr OR Coolant activity GREATER THAN 300 µCi/gm DOSE EQUIVALENT I-131 LEAKAGE RCS Leakage is GREATER THAN 50 GPM inside the drywell OR Unisolable primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Normal Limits per EOP 3, Table 6.

EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier Loss Potential Loss 2

LEAKAGE Failure of both valves in any one line to close and a downstream pathway to the environment exists OR Unisolable primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Safe Limits per EOP 3, Table 6, when Containment Isolation is required.

OR Primary containment venting per EOPs RADIATION/CORE DAMAGE Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 3E+3 Rem/hr OR Torus Area Hi Range Rad Monitor, RIM-9185A or B reading GREATER THAN 1E+2 Rem/hr OR Fuel damage assessment (PASAP 7.2) indicates at least 20% fuel clad damage RADIATION/CORE DAMAGE Drywell Area Hi Range Rad Monitor, RIM-9184A or B reading GREATER THAN 5 Rem/hr after reactor shutdown LEAKAGE Unisolable Main Steamline Break as indicated by the failure of both MSIVs in any one line to close AND EITHER:

High MSL flow or high steam tunnel temperature annunciators Direct report of steam release PRIMARY CONTAINMENT ATMOSPHERE Drywell pressure GREATER THAN 2 psig and not caused by a loss of DW Cooling PRIMARY CONTAINMENT ATMOSPHERE Rapid unexplained decrease following initial increase in pressure OR Drywell pressure response not consistent with LOCA conditions RPV LEVEL RPV Level LESS THAN

+15 inches RPV LEVEL RPV Level LESS THAN

+15 inches EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the RCS Barrier EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the RCS Barrier PRIMARY CONTAINMENT ATMOSPHERE Torus pressure reaches 53 psig and increasing OR Drywell or Torus H cannot be determined to be LESS THAN 6% and Drywell or Torus O cannot be determined to be LESS THAN 5%

EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Containment Barrier EMERGENCY DIRECTOR JUDGMENT Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Containment Barrier CLAD RCS CNTMT LOSS OF AT LEAST 2 BARRIERS?

CLAD RCS CNTMT CLAD RCS CNTMT ONE BARRIER AFFECTED TWO BARRIERS AFFECTED THREE BARRIERS AFFECTED 1/2 1/1 2/3 3/3 FU1 UNUSUAL EVENT FA1 ALERT FS1 SITE AREA EMERGENCY FG1 GENERAL EMERGENCY YES NO L

P L

P Systems of Concern Reactivity Control Containment (Drywell/Torus)

RHR/Core Spray/SRVs HPCI/RCIC RHRSW/River Water/ESW Onsite AC Power/EDGs Offsite AC Power Instrument AC DC Power Remote Shutdown Capability Safe Shutdown/Vital Areas Electrical Power Heat Sink / Coolant Supply Containment Emergency Systems Other 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Torus Room, Intake Structure, Pumphouse Drywell, Torus NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area, CSTs Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room Category Area

SLC System 3.1.7 DAEC 3.1-20 Amendment 223 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 inoperable.

A.1 Restore SLC subsystem to OPERABLE status.

7 days B.

Two SLC subsystems inoperable.

B.1 Restore one SLC subsystem to OPERABLE status.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

SLC System 3.1.7 DAEC 3.1-21 Amendment 223 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution is within the limits of Figure 3.1.7-1.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.7.2 Verify temperature of sodium pentaborate solution is within the limits of Figure 3.1.7-2.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.7.3 Verify temperature of pump suction piping is within the limits of Figure 3.1.7-2.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.7.4 Verify continuity of explosive charge.

31 days SR 3.1.7.5 Verify the concentration of boron in solution is within the limits of Figure 3.1.7-1.

31 days AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boron is added to solution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 (continued)

SLC System 3.1.7 DAEC 3.1-22 Amendment 223 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate 26.2 gpm at a discharge pressure 1150 psig.

In accordance with the Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from pump into reactor pressure vessel.

24 months on a STAGGERED TEST BASIS SR 3.1.7.8 Verify all heat traced piping between storage tank and pump suction is unblocked.

24 months AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2

Control Rod OPERABILITY 3.1.3 DAEC 3.1-7 Amendment 223 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 rod stuck.

NOTE----------------

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 rod separation criteria are met.

AND A.2 Disarm the associated Control Rod Drive (CRD).

AND Immediately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (continued)

Control Rod OPERABILITY 3.1.3 DAEC 3.1-8 Amendment 271 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.3 Perform SR 3.1.3.2 for each withdrawn OPERABLE control rod.

AND A.4 Perform SR 3.1.1.1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition A concurrent with THERMAL POWER greater than the Low Power Setpoint (LPSP) of the RWM.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Two or more withdrawn control rods stuck.

B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C.

One or more control rods inoperable for reasons other than Condition A or B.

C.1

NOTE------------

RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable control rod.

AND 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (continued)

Control Rod OPERABILITY 3.1.3 DAEC 3.1-9 Amendment 223 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

(continued)

C.2 Disarm the associated CRD.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D.

NOTE------------

Not applicable when THERMAL POWER

> 10% RTP.

Two or more inoperable control rods not in compliance with Banked Position Withdrawal Sequence (BPWS) and not separated by two or more OPERABLE control rods.

D.1 Restore compliance with BPWS.

OR D.2 Restore control rod to OPERABLE status.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours E.

Required Action and associated Completion Time of Condition A, C, or D, not met.

OR Nine or more control rods inoperable.

E.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

Control Rod OPERABILITY 3.1.3 DAEC 3.1-10 Amendment 271 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 notch.

31 days SR 3.1.3.3 Verify each control rod scram time from fully withdrawn to notch position 04 is 7 seconds.

In accordance with SR 3.1.4.1 and SR 3.1.4.2 SR 3.1.3.4 Verify each withdrawn control rod does not go to the withdrawn overtravel position.

Each time the 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)

Control Rod OPERABILITY 3.1.3 DAEC 3.1-11 Amendment 271 This Page Intentionally Blank per Amendment

Control Rod Scram Times 3.1.4 DAEC 3.1-12 Amendment 223 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 LCO not met.

A.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 within the limits of Table 3.1.4-1 with reactor steam dome pressure 800 psig.

Prior to exceeding 40% RTP after each refueling AND (continued)

Control Rod Scram Times 3.1.4 DAEC 3.1-13 Amendment 223 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 is within the limits of Table 3.1.4-1 with reactor steam dome pressure 800 psig.

Prior to exceeding 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

Control Rod Scram Times 3.1.4 DAEC 3.1-14 Amendment 271 Table 3.1.4-1 (page 1 of 1)

Control Rod Scram Times

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.

NOTCH POSITION SCRAM TIMES(a) (seconds) when REACTOR STEAM DOME PRESSURE 800 psig 46 38 26 06 0.44 0.93 1.83 3.35 (a)

Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

Control Rod Scram Accumulators 3.1.5 DAEC 3.1-15 Amendment 223 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 accumulator inoperable with reactor steam dome pressure 900 psig.

A.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 control rod scram time slow.

OR A.2 Declare the associated control rod inoperable.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours (continued)

Control Rod Scram Accumulators 3.1.5 DAEC 3.1-16 Amendment 223 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Two or more control rod scram accumulators inoperable with reactor steam dome pressure 900 psig.

B.1 Restore charging water header pressure to 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 control rod scram time slow.

OR B.2.2 Declare the associated control rod inoperable.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of condition B concurrent with charging water header pressure

< 940 psig 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour (continued)

Control Rod Scram Accumulators 3.1.5 DAEC 3.1-17 Amendment 223 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more control rod scram accumulators inoperable with reactor steam dome pressure

< 900 psig.

C.1 Verify all control rods associated with inoperable accumulators are fully inserted.

AND C.2 Declare the associated control rod inoperable.

Immediately upon discovery of charging water header pressure

< 940 psig 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D.

Required Action and associated Completion Time of Required Action B.1 or C.1 not met.

D.1

NOTE------------

Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

Place the reactor mode switch in the Shutdown position.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is 940 psig.

7 days

FIGURE #13: DAEC Core Map Showing Core Component Location Rev. 6 SD-262.1 SD_261-1.doc Nuclear Fuel and Control Rods

ML100351172

Text

5.0 REFERENCES

SUMMARY

5.0 REFERENCES

SUMMARY

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SUMMARY

SUMMARY

SUMMARY

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