Text

Vogtle Initial Exam 2009-301 Draft RO Written Exam
	ML093280270
Person / Time
Site:	Vogtle
Issue date:	11/18/2009
From:	; NRC/RGN-II
To:	; Southern Nuclear Operating Co
References
	50-424/09-301, 50-425/09-301 50-424/09-301, 50-425/09-301
	Download: ML093280270 (150)
	v • d • e

{{#Wiki_filter:1. 022AA2.02 00ll1l1ILOSS OF RCS MAKEUP/C/A -3.7INEWIHL15/SROIDS/TNT Initial conditions: -Unit is at 100% power -All systems are in automatic -The NCP is in service Current conditions: -Charging line flow is fluctuating between 30 and 130 gpm -Letdown Regen HX outlet temperature is fluctuating between 250 and 400 F -The following annunciators are alarming: CHARGING LINE HIILO FLOW REGEN HX L TDN HI TEMP VCT HIILO LEVEL Which of the following is correct for these conditions? A. Enter AOP 18007 -C section B for loss of charging flow. Isolate letdown, increase charging flow as necessary to stabilize flow indications. B!'" Enter AOP 18007 -C section B for loss of charging flow. Isolate letdown, stop the NCP, monitor RCP seals, vent charging pump suctions. C, Enter AOP 18007 -C section A for loss of letdown flow. Increase charging flow until stable indications between 80-90 gpm are established. D. Enter AOP 18007-C sections A & B for loss of letdown flow and loss of charging flow concurrently. Stop the NCP and isolate letdown flow. Start a CCP. 022 Loss of Reactor Coolant Makeup AA2.02 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Charging pump problems KIA MATCH ANALYSIS Question gives a scenario for a loss of charging flow due to charging pump cavitation. The correct answer requires selecting the loss of charging flow abnormal operating procedure and requires recall of actions related to charging pump cavitiation as part of the mitigation strategies. Question meets 10CFR55.43(b) criteria item #5 -Assessment of facility conditions and Page: 1 of 56 4/2112009 I \ 1. 022AA2.02 00llllllLOSS OF RCS MAKEUP/C/A -3.7INEWIHL15/SRO/DS/TNT Initial conditions: -Unit is at 100% power -All systems are in automatic -The NCP is in service Current conditions: -Charging line flow is fluctuating between 30 and 130 gpm -Letdown Regen HX outlet temperature is fluctuating between 250 and 400 F -The following annunciators are alarming: CHARGING LINE HIILO FLOW REGEN HX L TON HI TEMP VCT HIILO LEVEL Which of the following is correct for these conditions? A. Enter AOP 1B007-C section B for loss of charging flow. Isolate letdown, increase charging flow as necessary to stabilize flow indications. Enter AOP 1B007-C section B for loss of charging flow. Isolate letdown, stop the NCP, monitor RCP seals, vent charging pump suctions. C. Enter AOP 1B007-C section A for loss of letdown flow. Increase charging flow until stable indications between BO-90 gpm are established. O. Enter AOP 1B007-C sections A & B for loss of letdown flow and loss of charging flow concurrently. Stop the NCP and isolate letdown flow. Start a CCP. 022 Loss of Reactor Coolant Makeup AA2.02 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Charging pump problems KIA MATCH ANALYSIS Question gives a scenario for a loss of charging flow due to charging pump cavitation. The correct answer requires selecting the loss of charging flow abnormal operating procedure and requires recall of actions related to charging pump cavitiation as part of the mitigation strategies. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and Page: 1 of 56 4121/2009 selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Choice is plausible since all actions are correct except for the charging pump. B. Correct. C. Incorrect. Plausible distractor with letdown flow indications and required actions to isolate letdown. Establishing charging flow 80-90 gpm is contained in procedures for establishing letdown flow. D. Incorrect. Plausible since actions listed are correct except for starting the CCP. Entry into both sections of the procedure are not required. REFERENCES AOP 18007 -C, Section B for loss of charging flow. AOP 18007 -C, Section A for loss of letdown flow. SOP 13006-1, CVCS section 4.4.2, returning normal charging and letdown to service. VEGP learning objectives: LO-LP-60307 -02 Given the entire, AOP describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s) for the step Page: 2of56 4/21/2009 selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Choice is plausible since all actions are correct except for the charging pump. B. Correct. C. Incorrect. Plausible distractor with letdown flow indications and required actions to isolate letdown. Establishing charging flow 80-90 gpm is contained in procedures for establishing letdown flow. D. Incorrect. Plausible since actions listed are correct except for starting the CCP. Entry into both sections of the procedure are not required. REFERENCES AOP 18007-C, Section B for loss of charging flow. AOP 18007-C, Section A for loss of letdown flow. SOP 13006-1, CVCS section 4.4.2, returning normal charging and letdown to service. ( VEGP learning objectives: LO-LP-60307-02 Given the entire, AOP describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter(s) for the step Page: 20f56 4/2112009 Approval Date Procedure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS Uni t COMMON Abnormal Operating Procedures Revision No. Page No. CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION 18007-C 19 1 of 17 PURPOSE PRB REVIEW REQUIRED This procedure specifies the actions to be taken in the event of a malfunction in the Chemical and Volume Control System. Specific actions are provided for the following abnormal conditions:

A. Total Loss of Letdown Flow. B. Loss of Charging Flow. .. C. Loss of VCT Makeup. > SYMPTOMS SECTION A, TOTAL LOSS OF LETDOWN FLOW

ALB07-E03 LTDN HX OUT HI PRESS
ALB07-C05 LP LTD RELEIF HI TEMP
TI-126 CHARGING REGEN HX OUT indicating less than 120°F.
PRZR level rising with no change in Tavg. SECTION B, LOSS OF CHARGING FLOW
ALB07-A05 REGEN HX LTDN HI TEMP
ALB07-B06 CHARGING LINE HI/LO FLOW
ALB07-C06 CHARGING PUMP OVERLOAD TRIP
ALB08-F06 RCP SEAL WATER INJ LO FLOW SECTION C, LOSS OF VCT MAKEUP
ALB07-C02 TOTAL MAKEUP FLOW DEVIATION
ALB07-D05 AUTO MAKE-UP START SIGNAL BLOCKED
ALB07-E05 VCT HI/LO LEVEL
ALB07-F01 BA FLOW DEVIATION
Indicated low VCT level. Approval Date ( Procedure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS Uni t COMMON Abnormal Operating Procedures Revision No. Page No. CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION 18007-C 19 1 of 17 PURPOSE PRB REVIEW REQUIRED This procedure specifies the actions to be taken in the event of a malfunction in the Chemical and Volume Control System. Specific actions are provided for the following abnormal conditions:

A. Total Loss of Letdown Flow. B. Loss of Charging Flow. .. C. Loss of VCT Makeup. > SYMPTOMS SECTION A, TOTAL LOSS OF LETDOWN FLOW

ALB07-E03 LTDN HX OUT HI PRESS
ALB07-C05 LP LTD RELEIF HI TEMP
TI-126 CHARGING REGEN HX OUT indicating less than 120°F.
PRZR level rising with no change in Tavg. SECTION B, LOSS OF CHARGING FLOW
ALB07-A05 REGEN HX LTDN HI TEMP
ALB07-B06 CHARGING LINE HI/LO FLOW
ALB07-C06 CHARGING PUMP OVERLOAD TRIP
ALB08-F06 RCP SEAL WATER INJ LO FLOW SECTION C, LOSS OF VCT MAKEUP
ALB07-C02 TOTAL MAKEUP FLOW DEVIATION
ALB07-D05 AUTO MAKE-UP START SIGNAL BLOCKED
ALB07-E05 VCT HI/LO LEVEL
ALB07-F01 BA FLOW DEVIATION
Indicated low VCT level.

PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18007-C 19 9 of 17 B. LOSS OF CHARGING FLOW ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS B1. Isolate letdown: a. Close letdown orifice isolation valves: 0-HV-8149A 0-HV-8149B 0-HV-8149C b. Close letdown isolation valves: 0-LV-459 0-LV-460 SUBSEQUENT OPERATOR ACTIONS OB2. Initiate the Continuous Actions Page. 0* B3. Trend RCP Seal Parameters listed in ATTACHMENT A. B4. Check charging pump(s) -B4. Perform the following: OPERATING NORMALLY: 0-Discharge flow trend -STABLE < Oa. Stop charging pumps. Ob. 0-Discharge pressure trend Determine and correct cause of charging pump abnormal operations. -STABLE 0-VCT level -IN NORMAL BAND 0-Bus current -STABLE OB5. Locate and isolate any charging system leakage. Oc. IF gas binding of charging pumps occurred, THEN do NOT start charging pumps until the cause of the gas binding is understood and all affected piping and components are vented. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18007-C 19 9 of 17 B. LOSS OF CHARGING FLOW ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS Bl. Isolate letdown: a. Close letdown orifice isolation valves: 0-HV-8149A 0-HV-8149B 0-HV-8149C b. Close letdown isolation valves: 0-LV-459 0-LV-460 SUBSEQUENT OPERATOR ACTIONS OB2. Initiate the Continuous Actions Page. 0* B3. Trend RCP Seal Parameters listed in ATTACHMENT A. B4. Check charging pump(s) -, B4. Perform the following: OPERATING NORMALLY: 0-Discharge flow trend -STABLE 0-Discharge pressure trend -STABLE 0-VCT level -IN NORMAL BAND 0-Bus current -STABLE o B5. Locate and isolate any charging system leakage. o a . Stop charging pumps. Db. DC. Determine and correct cause of charging pump abnormal operations. IF gas binding of charging pumps occurred, THEN do NOT start charging pumps until the cause of the gas binding is understood and all affected piping and components are vented.

2. 027AG2.1.32 00llllllPRZRPRESS MALFICIA -4.0INEW/HL1S/SRO/DS/TNT Given the following conditions:

-The unit is shutdown with Tave at 557 degrees F -PV-8000A block valve is shut due to excessive seat leakage on PORV-455 -The breaker for block valve PV-8000B (PORV-456) has just tripped open -A risk assessment has NOT been performed for the PV-8000B failure Which one of the following correctly describes allowable technical specifcation actions and the bases for those actions? A. Separate condition entry IS allowed for each PORV. Reactor startup may proceed. Bases require ONE PORV and its associated block valve to be capable of manual operation. B. Separate condition entry is NOT allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. C!" Separate condition entry IS allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. D. Separate condition entry is NOT allowed for each PORV. Reactor startup may proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. Page: 3 of 56 4/2112009 ( 2. 027AG2.1.32 00 111/1 IPRZR PRESS MALFICIA -4.0INEW/HL15/SRO/DS/TNT Given the following conditions: The unit is shutdown with Tave at 557 degrees F -PV-8000A block valve is shut due to excessive seat leakage on PORV-455 -The breaker for block valve PV-8000B (PORV-456) has just tripped open -A risk assessment has NOT been performed for the PV-8000B failure Which one of the following correctly describes allowable technical specifcation actions and the bases for those actions? A. Separate condition entry IS allowed for each PORV. Reactor startup may proceed. Bases require ONE PORV and its associated block valve to be capable of manual operation. B. Separate condition entry is NOT allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. C!' Separate condition entry IS allowed for each PORV. Reactor startup may NOT proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. D. Separate condition entry is NOT allowed for each PORV. Reactor startup may proceed. Bases require BOTH PORVs and their associated block valves to be capable of manual operation. Page: 3 of 56 4/2112009 ( 027 Pressurizer Pressure Control (PZR PCS) Malfunction AG2.1.32 Conduct of Operations Ability to explain and apply system limits and precautions. KIA MATCH ANALYSIS Question gives a scenario that requires application of technical specification limits and bases for the PZR PORVs (LCO 3.4.11) while in mode 3. Question meets 1 OCFR55.43(b) criteria item #2 -Facility operating limitations in the technical specifications and their bases making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since LCO 3.4.11 does allow for separate entry and startup would be allowed for the excessive seat leakage. B. Incorrect. Plausible since reactor startup may not proceed and the bases are correct. C. Correct. D. Incorrect. Plausible since bases is correct and reactor startup is allowed for the excessive seat leakage actions. REFERENCES VEGP Technical Specification LCO 3.4.11 and Bases 3.4.11 VEGP learning objectives: LO-LP-39208-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.4 are exceeded.

b. The required actions for all section 3.4 LCOs LO-LP-39208-04:

Describe the bases for any given Tech Spec in section 3.4. Page: 4 of 56 4/2112009 ( 027 Pressurizer Pressure Control (PZR PCS) Malfunction AG2.1.32 Conduct of Operations Ability to explain and apply system limits and precautions. KIA MATCH ANALYSIS Question gives a scenario that requires application of technical specification limits and bases for the PZR PORVs (LCO 3.4.11) while in mode 3. Question meets 1 OCFR55.43(b) criteria item #2 -Facility operating limitations in the technical specifications and their bases making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since LCO 3.4.11 does allow for separate entry and startup would be allowed for the excessive seat leakage. B. Incorrect. Plausible since reactor startup may not proceed and the bases are correct. C. Correct. D. Incorrect. Plausible since bases is correct and reactor startup is allowed for the excessive seat leakage actions. REFERENCES VEGP Technical Specification LCO 3.4.11 and Bases 3.4.11 VEGP learning objectives: LO-LP-39208-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.4 are exceeded.

b. The required actions for all section 3.4 LCOs LO-LP-39208-04:

Describe the bases for any given Tech Spec in section 3.4. Page: 4 of 56 4/2112009

3.4 REACTOR

COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) Pressurizer PORVs 3.4.11 LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS -----------------------------------------------------------NOTE------------------------------------------------------------ Separate Condition entry is allowed for each PORV. CONDITION A. One or more PORVs A.1 inoperable and capable of being manually cycled. B. One PORV inoperable B.1 and not capable of being manually cycled. AND B.2 AND B.3 Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Close and maintain power 1 hour to associated block valve. Close associated block 1 hour valve. Remove power from 1 hour associated block valve. Restore PORV to 72 hours OPERABLE status. (continued) 3.4.11-1 Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) ( 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) Pressurizer PORVs 3.4.11 LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS ----------------------------------------------------------- NOT E ------------------------------------------------------------ Separate Condition entry is allowed for each PORV. CONDITION A. One or more PORVs A.1 inoperable and capable of being manually cycled. B. One PORV inoperable B.1 and not capable of being manually cycled. AND B.2 AND B.3 Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Close and maintain power 1 hour to associated block valve. Close associated block 1 hour valve. Remove power from 1 hour associated block valve. Restore PORV to 72 hours OPERABLE status. ( continued) 3.4.11-1 Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) ACTIONS (continued) CONDITION C. One block valve C.1 inoperable. AND C.2 V D. Required Action and 0.1 associated Completion Time of Condition A, B, AND or C not met. 0.2 E. Two PORVs inoperable E.1 and not capable of being manually cycled. ( AND E.2 AND E.3 AND E.4 F. More than one block F.1 valve inoperable. AND Vogtle Units 1 and 2 REQUIRED ACTION Place associated PORV in manual control. Restore block valve to OPERABLE status. Be in MODE 3. Be in MODE 4. Close associated block valves. Remove power from associated block valves. Be in MODE 3. Be in MODE 4. Pressurizer PORVs 3.4.11 COMPLETION TIME 1 hour 72 hours 6 hours 12 hours 1 hour 1 hour 6 hours 12 hours Place associated PORVs 1 hour in manual control. 3.4.11-2 ( continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ACTIONS (continued) CONDITION C. One block valve C.1 inoperable. AND C.2 '/ D. Required Action and D.1 associated Completion Time of Condition A. B, AND or C not met. D.2 E. Two PORVs inoperable E.1 and not capable of being manually cycled. ( AND E.2 AND E.3 AND E.4 F. More than one block F.1 valve inoperable. AND Vogtle Units 1 and 2 REQUIRED ACTION Place associated PORV in manual control. Restore block valve to OPERABLE status. Be in MODE 3. Be in MODE 4. Close associated block valves. Remove power from associated block valves. Be in MODE 3. Be in MODE 4. Pressurizer PORVs 3.4.11 COMPLETION TIME 1 hour 72 hours 6 hours 12 hours 1 hour 1 hour 6 hours 12 hours Place associated PORVs 1 hour in manual control. 3.4.11-2 ( continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) BASES LCO APPLICABILITY Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 The LCO requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR, or loss of heat sink, and to achieve safety grade cold shutdown. The PORVs are considered OPERABLE in either the manual or automatic mode. The PORVs (PV-455A and PV-456A) are powered from 125 V MCCs 1/2AD1M and 1/2BD1M, respectively. If either or both of these MCCs become inoperable, the affected PORV(s) are to be considered inoperable. By maintaining two PORVs and their associated block valves OPERABLE, the single failure criterion is satisfied. An OPERABLE PORV is required to be capable of manually opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific criteria, exists when conditions dictate closure of the block valve to limit leakage. An OPERABLE block valve may be either open and energized, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation. Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being manually cycled (e.g., as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided the relief function remains available with manual action. Satisfying the LCO helps minimize challenges to fission product barriers. The PORVs are required to be OPERABLE in MODES 1, 2, and 3 for manual actuation to mitigate a steam generator tube rupture event, an inadvertent safety injection, and to achieve safety grade cold shutdown. In addition, the block valves are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid increases will occur at the higher operating power and pressure conditions of MODES 1 and 2. Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant. LCO 3.4.12 addresses the PORV ( continued) B 3.4.11-3 Rev. 1-2/00 BASES LCO APPLICABILITY Vogtle Units 1 and 2 Pressurizer PORVs B 3.4.11 The LCO requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR, or loss of heat sink, and to achieve safety grade cold shutdown. The PORVs are considered OPERABLE in either the manual or automatic mode. The PORVs (PV-455A and PV-456A) are powered from 125 V MCCs 1/2AD1M and 1/2BD1M, respectively. If either or both of these MCCs become inoperable, the affected PORV(s) are to be considered inoperable. By maintaining two PORVs and their associated block valves OPERABLE, the single failure criterion is satisfied. An OPERABLE PORV is required to be capable of manually opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific criteria, exists when conditions dictate closure of the block valve to limit leakage. An OPERABLE block valve may be either open and energized, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation. Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being manually cycled (e.g., as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided the relief function remains available with manual action. Satisfying the LCO helps minimize challenges to fission product barriers. The PORVs are required to be OPERABLE in MODES 1, 2, and 3 for manual actuation to mitigate a steam generator tube rupture event, an inadvertent safety injection, and to achieve safety grade cold shutdown. In addition, the block valves are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid increases will occur at the higher operating power and pressure conditions of MODES 1 and 2. Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant. LCO 3.4.12 addresses the PORV (continued) B 3.4.11-3 Rev. 1-2/00 LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.B. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 (continued) Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2) ( ( LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.8. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 (continued) Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2) LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 (continued) LCO 3.0.5 LCO 3.0.6 Vogtle Units 1 and 2 b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications; or c. When an allowance is stated in the individual value, parameter, or other Specification. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 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. When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. 3.0-2 (continued) Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) ( LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 ( continued) LCO 3.0.5 LCO 3.0.6 Vogtle Units 1 and 2 b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications; or c. When an allowance is stated in the individual value, parameter, or other Specification. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 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. When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. 3.0-2 ( continued) Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2)

3. 029EG2.2.22 OOl/1/1/ATWT/CIA

-4.7INEW/HLl5/SRO/DS/TNT Given the following conditions with the unit at 100% power: -PRZR pressure channel PT -455 has failed LOW -NR cold leg temperature instrument for RCS loop 2 has failed LOW causing the Loop 2 delta T indication to go off scale high -All associated TSLB bistables for each failure are lit -The unit continues to operate at full power Which of the following is correct with respect to the required technical specification actions for this set of conditions? A'! An A TWT is in progress. LCO 3.0.3 entry is required. B. An ATWT is NOT in progress. LCO 3.0.3 entry is required. C. An A TWT is in progress. LCO 3.3.1 applies since separate condition entry is allowed for each function. LCO 3.0.3 entry is NOT required. D. An ATWT is NOT in progress. Page: 50f56 LCO 3.3.1 applies since separate condition entry is allowed for each function. LCO 3.0.3 entry is NOT required. 4/2112009 ( ( \ 3. 029EG2.2.22 OOllllllATWT/CIA -4.7INEW/HL15/SRO/DS/TNT Given the following conditions with the unit at 100% power: PRZR pressure channel PT -455 has failed LOW -NR cold leg temperature instrument for RCS loop 2 has failed LOW causing the Loop 2 delta T indication to go off scale high -All associated TSLB bistables for each failure are lit -The unit continues to operate at full power Which of the following is correct with respect to the required technical specification actions for this set of conditions? A'! An A TWT is in progress. LCO 3.0.3 entry is required. B. An ATWT is NOT in progress. LCO 3.0.3 entry is required. C. An A TWT is in progress. LCO 3.3.1 applies since separate condition entry is allowed for each function. LCO 3.0.3 entry is NOT required. D. An ATWT is NOT in progress. Page: 50f56 LCO 3.3.1 applies since separate condition entry is allowed for each function. LCO 3.0.3 entry is NOT required. 4/2112009 029 Anticipated Transient Without Scram (A TWS) EG2.2.22 Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA MATCH ANALYSIS This question provides a scenario that requires the applicant to determine the status of the reactor trip system, the applicable LCO(s) and associated functions and actions and application of the note modifying LCO entry conditions. Question meets 1 OCFR55.43(b) criteria item #2 -Facility operating limitations in the technical specifications and their bases making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since LCO 3.0.3 entry is required for this condition. Applicant must determine that 2 OTdeltaT channels are tripped and an ATWT condition exists. C. Incorrect. Plausible since all actions are correct except for the LCO 3.0.3 entry. D. Incorrect. Plausible since LCO 3.3.1 does contain a note that states separate conition entry is allowed for each function. Applicant must relaize this is one function. REFERENCES VEGP Technical specifications LCO 3.3.1 & 3.0.3 and associated bases V-LO-TX-28101 Rev 8 pages 29 & 30 (Reactor Protection System Text) VEGP learning objectives: LO-LP-39207-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.3 are exceeded.

b. The required actions for all section 3.3 LCOs. Page: 6of56 4/21/2009 ( KIA 029 Anticipated Transient Without Scram (ATWS) EG2.2.22 Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA MATCH ANALYSIS This question provides a scenario that requires the applicant to determine the status of the reactor trip system, the applicable LCO(s) and associated functions and actions and application of the note modifying LCO entry conditions.

Question meets 1 OCFR55.43(b) criteria item #2 -Facility operating limitations in the technical specifications and their bases making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since LCO 3.0.3 entry is required for this condition. Applicant must determine that 2 OTdeltaT channels are tripped and an ATWT condition exists. C. Incorrect. Plausible since all actions are correct except for the LCO 3.0.3 entry. D. Incorrect. Plausible since LCO 3.3.1 does contain a note that states separate conition entry is allowed for each function. Applicant must relaize this is one function. REFERENCES VEGP Technical specifications LCO 3.3.1 & 3.0.3 and associated bases V-LO-TX-28101 Rev 8 pages 29 & 30 (Reactor Protection System Text) VEGP learning objectives: LO-LP-39207-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.3 are exceeded.

b. The required actions for all section 3.3 LCOs. Page: 6of56 4/2112009

( FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS Table 3.3.1-1 (page 2 of 9) Reactor Trip System Instrumentation REQUIRED SURVEILLANCE CHANNELS CONDITIONS REQUIREMENTS RTS Instrumentation

3.3.1 ALLOWABLE

VALUE NOMINAL TRIP SETPOINT(n)

5. Source Range Neutron Flux 2 I,J SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 os; 1.7 E5 cps 1.0 E5 cps 6. 7. 2 J,K 3(e), 4(e), 5(e) L Overtemperature L'l T 1,2 4 E Overpower L'l T 1,2 4 E (a) With RTBs closed and Rod Control System capable of rod withdrawaL (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 os; 1.7 E5 cps NA Refer to Note 1 (Page 3.3.1-20) Refer to Note 2 (Page 3.3.1-21) 1.0 E5 cps NA Refer to Note 1 (Page 3.3.1-20) Refer to Note 2 (Page 3.3.1-21) (continued) (e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide input to the High Flux at Shutdown Alarm System (LCO 3.3.8) and indication. (n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions. Vogtle Units 1 and 2 3.3.1-15 Amendment No. 128 (Unit 1) Amendment No. 106 (Unit 2) r ( FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS Table 3.3.1-1 (page 2 of 9) Reactor Trip System Instrumentation REQUIRED SURVEILLANCE CHANNELS CONDITIONS REQUIREMENTS RTS Instrumentation

3.3.1 ALLOWABLE

VALUE NOMINAL TRIP SETPOINT(n)

5. Source Range Neutron Flux 2 I,J SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 os; 1.7 E5 cps 1.0 E5 cps 6. 7. 2 J,K 3(e), 4(e), 5(e) L Overtemperature T 1,2 4 E Overpower T 1,2 4 E (a) With RTBs closed and Rod Control System capable of rod withdrawal. (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.1 SR 3.3.1.11 SR 3.3.1.1 SR3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 os; 1.7 E5 cps NA Refer to Note 1 (Page 3.3.1-20) Refer to Note 2 (Page 3.3.1-21) 1.0 E5 cps NA Refer to Note 1 (Page 3.3.1-20) Refer to Note 2 (Page 3.3.1-21) (continued) (e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide input to the High Flux at Shutdown Alarm System (LCO 3.3.8) and indication. (n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions. Vogtle Units 1 and 2 3.3.1-15 Amendment No. 128 (Unit 1) Amendment No. 106 (Unit 2) r / ! / / I I ACTIONS (continued) CONDITION E. One channel inoperable. F. THERMAL POWER > P-6 and < P-10, one Intermediate Range Neutron Flux channel inoperable. G. THERMAL POWER > P-6 and < P-10, two Intermediate Range Neutron Flux channels inoperable. H. THERMAL POWER < P-6, one or two Intermediate Range Neutron Flux channels inoperable. Vogtle Units 1 and 2 REQUIRED ACTION --------------------NOTE--------------------- A channel may be bypassed for up to 12 hours for surveillance testing. ------------------------------------------------- E.1 Place channel in trip. OR E.2 Be in MODE 3. F.1 Reduce THERMAL POWER to < P-6. OR F.2 Increase THERMAL POWER to> P-10. G.1 Suspend operations involving positive reactivity additions. AND G.2 Reduce THERMAL POWER to < P-6. H.1 Restore channel(s) to OPERABLE status. RTS Instrumentation

3.3.1 COMPLETION

TIME 72 hours 78 hours 24 hours 24 hours Immediately 2 hours Prior to increasing THERMAL POWER to> P-6 (continued) 3.3.1-3 Amendment No. 116 (Unit 1) Amendment No. 94 (Unit 2) \ r ACTIONS (continued) CONDITION E. One channel inoperable. F. THERMAL POWER > P-6 and < P-10, one Intermediate Range Neutron Flux channel inoperable. ( G. THERMAL POWER > P-6 and < P-1 0, two Intermediate Range Neutron Flux channels inoperable. H. THERMAL POWER < P-6, one or two Intermediate Range Neutron Flux channels inoperable. Vogtle Units 1 and 2 REQUIRED ACTION -------------------- N OT E--------------------- A channel may be bypassed for up to 12 hours for surveillance testing. ------------------------------------------------- E.1 Place channel in trip. OR E.2 Be in MODE 3. F.1 Reduce THERMAL POWER to < P-6. OR F.2 Increase THERMAL POWER to> P-10. G.1 Suspend operations involving positive reactivity additions. AND G.2 Reduce THERMAL POWER to < P-6. H.1 Restore channel(s) to OPERABLE status. RTS Instrumentation

3.3.1 COMPLETION

TIME 72 hours 78 hours 24 hours 24 hours Immediately 2 hours Prior to increasing THERMAL POWER to> P-6 (continued) 3.3.1-3 Amendment No. 116 (Unit 1) Amendment No. 94 (Unit 2) ! LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.8. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 (continued) \ Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2: ( LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.8. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 (continued) Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2' / / ! 28.13 REACTOR TRIP SETPOINTS The following is the list of reactor trips that you will have to commit to memory. The bases for the trips may vary so referencing the technical specification section 3.3.1 for each trip could be helpful. Reactor Trips Coincidence Set point 1) Manual reactor trip lout of 2 HS located on 2) Source Range high flux trip the main control board lout of 2 SR channels Bases: Provides protection against uncontrolled rod withdrawal while subcritical.

3) Intermediate Range high flux trip lout of 2 IR channels 25% power Bases: Provides protection against uncontrolled rod withdrawal while subcritical.

4) Power Range low range high flux trip 2 out of 4 PR channels 25% power Bases: Mitigates the consequences of power excursion starting from low power excursions starting at all power levels. 5) Power Range high range high flux trip 2 out of 4 PR channels 109% power 6) Bases: Provides protection from power excursions starting at all power levels. Power Range high flux rate 2 out of 4 PR channels 5% in 2 sec Bases: Provides protection from the power excursion caused from an ejected rod accident.

7) Overtemperature

.6.T Trip 2 out of 4 loop .6.T = 114.9% unpenalized setpoint The Set Point is variable based on its associated channel Pressurizer Pressure, Power Range .6. Flux, and Loop Tavg. When Pressurizer pressure lowers, Tavg increases, or Power Range .6. Flux travels outside its boundary limits the OT.6.T set point will lower. 29 V-LO-TX-28101 Revision 8 ( ( 28.13 REACTOR TRIP SETPOINTS The following is the list of reactor trips that you will have to commit to memory. The bases for the trips may vary so referencing the technical specification section 3.3.1 for each trip could be helpful. Reactor Trips Coincidence Set point 1) Manual reactor trip lout of 2 HS located on 2) Source Range high flux trip the main control board lout of 2 SR channels Bases: Provides protection against uncontrolled rod withdrawal while subcritical.

3) Intermediate Range high flux trip lout of 2 IR channels 25% power Bases: Provides protection against uncontrolled rod withdrawal while subcritical.

4) Power Range low range high flux trip 2 out of 4 PR channels 25% power Bases: Mitigates the consequences of power excursion starting from low power excursions starting at all power levels. 5) Power Range high range high flux trip 2 out of 4 PR channels 109% power 6) Bases: Provides protection from power excursions starting at all power levels. Power Range high flux rate 2 out of 4 PR channels 5% in 2 sec Bases: Provides protection from the power excursion caused from an ejected rod accident.

7) Overtemperature lI.T Trip 2 out of 4 loop lI.T = 114.9% unpenalized setpoint The Set Point is variable based on its associated channel Pressurizer Pressure, Power Range 1I. Flux, and Loop Tavg. When Pressurizer pressure lowers, Tavg increases, or Power Range 1I. Flux travels outside its boundary limits the OTlI.T set point will lower. 29 V-LO-TX-28101 Revision 8

( Fundamental Set Point Equation: K1 -K2 (T -T') -K3 (P' -P) -f1(AFD) Measured Loop specific Res differential temperature in of (displayed in percent power 0% -150% power) K1 Nominal set point of 114.9% K2 Modifier for Tavg = 2.24% RTP per of (penalty above 588.4°F)/(reward given to set point from 588.4°F to 585.4°F) K3 Modifier for pressure = 0.177% per psig (penalty below 2235 psig)/(reward given to set point above 2235 psig) T measured loop specific Res average temperature in of T' indicated loop specific Res average temperature at RTP, =588.4°F P' referenced pressure, = 2235 psig P measured Res pressurizer pressure in psig f1(AFD) = modifier for Axial Flux Difference AFD between -23% to + 10% no penalty to set point AFD for each % AFD is below -23%, the trip set point is reduced 3.3% AFD for each % AFD is above +10%, the trip set point is reduced 1.95% The maximum set points: Unit 1 Unit 2 121.6% 118.7% A reactor trip will occur if 2 out of the 4 loop increase to the variable set point. Bases: Provides the reactor core protection from Departure from nucleate boiling. 8 ) Overpower T Trip 2 out of 4 loop = 110% unpenalized setpoint The Set Point is variable based on its associated Loop Tavg only. Fundamental Set Point Equation: K4 -K5 (rate of change of T) -K6 (T -T') -f2(AFD) Measured Loop specific Res differential temperature in of (displayed in percent power 0% -150% power) 30 V-LO-TX-28101 Revision 8 ( 8) Fundamental Set Point Equation: K1 -K2 (T -T') -K3 (p' -P) -f1(AFD) Measured Loop specific Res differential temperature in of (displayed in percent power 0% -150% power) K1 Nominal set point of 114.9% K2 Modifier for Tavg = 2.24% RTP per of (penalty above 588.4°F)/(reward given to set point from 588.4°F to 585.4°F) K3 Modifier for pressure = 0.177% per psig (penalty below 2235 psig)/(reward given to set point above 2235 psig) T measured loop specific Res average temperature in of T' indicated loop specific Res average temperature at RTP, =588.4°F P' referenced pressure, = 2235 psig P = measured Res pressurizer pressure in psig f1(AFD) = modifier for Axial Flux Difference AFD between -23% to + 10% no penalty to set point AFD for each % AFD is below -23%, the trip set point is reduced 3.3% AFD for each % AFD is above +10%, the trip set point is reduced 1.95% The maximum OT6T set points: Unit 1 Unit 2 121.6% 118.7% A reactor trip will occur if 2 out of the 4 loop increase to the variable set point. Bases: Provides the reactor core protection from Departure from nucleate boiling. Overpower T Trip 2 out of 4 loop = 110% unpenalized setpoint The Set Point is variable based on its associated Loop Tavg only. Fundamental Set Point Equation: K4 -K5 (rate of change of T) -K6 (T -T') -f2(AFD) Measured Loop specific Res differential temperature in of (displayed in percent power 0% -150% power) 30 V-LO-TX-28101 Revision 8

4. 058AA2.03 00llllllLOSS OF DC POWER/CIA

-3.9INEW/HL15/SRO/DS/TNT DG-1A is running with its output breaker open and the unit at full power when the following indications are received: -Train A MSIV's red & green lights extinguish -RTA red & green lights extinguish -RCP #1 1 E breaker red & green lights extinguish -Channell TSLB bistable lights illuminate Which of the following is the correct.. .. a) procedure{s) to enter, and b) corrective actions to take? A'!' a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate18034-1, Loss of Class 1 E 125 VDC Power, when directed in 19001-C, Reactor Trip Response. b) Stop DG-1 A using the push-to-stop control, isolate the letdown relief flowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using SG ARVs 2 & 3. B. a) Enter 18034-1, Loss of Class 1 E DC Power. b) Locally emergency stop DG-1 A, isolate the letdown relief flowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using SG ARVs 2 & 3. C. a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate18034-1, Loss of Class 1 E 125 VDC Power, when directed in 19001-C, Reactor Trip Response. b) Emergency stop DG-1A from the control room, isolate the letdown relief flowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using the steam dumps. D. a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate 18032-1, Loss of 120V AC Instrument Power, when directed in 19001-C, Reactor Trip Response. b) Manually control PZR pressure between 2220 and 2260 psig until1AY1A is re-energized by the associated regulated transformer, control Tave using the steam dumps. KIA Page: 7of56 4/2112009 ( 4. 058AA2.03 OOllllllLOSS OF DC POWER/CIA -3.9INEW/HL15/SRO/DS/TNT DG-1A is running with its output breaker open and the unit at full power when the following indications are received: -Train A MSIV's red & green lights extinguish -RT A red & green lights extinguish -RCP #1 1 E breaker red & green lights extinguish -Channel I TSLB bistable lights illuminate Which of the following is the correct.. .. a) procedure{s) to enter, and b) corrective actions to take? A'I a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate18034-1, Loss of Class 1 E 125 VDC Power, when directed in 19001-C, Reactor Trip Response. b) Stop DG-1A using the push-to-stop control, isolate the letdown relief flowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using SG ARVs 2 & 3. B. a) Enter 18034-1, Loss of Class 1 E DC Power. b) Locally emergency stop DG-1A, isolate the letdown reliefflowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using SG ARVs 2 & 3. C. a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate18034-1, Loss of Class 1 E 125 VDC Power, when directed in 19001-C, Reactor Trip Response. b) Emergency stop DG-1 A from the control room, isolate the letdown relief flowpath, energize 1AY1A and 1 AY2A from their regulated transformers, control Tave using the steam dumps. D. a) Enter 19000-C, E-O Reactor Trip or Safety Injection, and then initiate 18032-1, Loss of 120V AC Instrument Power, when directed in 19001-C, Reactor Trip Response. b) Manually control PZR pressure between 2220 and 2260 psig until1AY1A is re-energized by the associated regulated transformer, control Tave using the steam dumps. KIA Page: 7of56 4/2112009 058 Loss of DC Power AA2.03 Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems KIA MATCH ANALYSIS This question requires corect diagnosis of plant indications to determine correct procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since AOP 18034-1 is entered for this malfunction and all the listed actions are taken except the DG is not emergency stopped. C. Incorrect. Plausible since the procedure entries are correct and the actions taken are ( correct except the methods used to stop the DG and control Tave. D. Incorrect. Plausible since misdiagnosis of a loss of only 120 V AC instrument power is a common mistake by operators due to the channel I bistable lights. Actions listed are correct, for the incorrect AOP entry. EOP entry also plausible since this is the expected response for power < P-10 setpoint. REFERENCES

1. AOP 18034-1, Loss of Class 1 E 125V DC Power. VEGP learning objectives:

1. LO-LP-60329-01:

Given that a loss of power has occurred to any of the following 125VDC vital buses and given the appropriate plant procedures, describe the operator actions required and why these actions are taken. a.1AD1 b. 1 BD1 c. 1 CD1 d. 1 DD1 2. LO-LP-60329-04: Page: 8of56 4/2112009 058 Loss of DC Power AA2.03 Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems KIA MATCH ANALYSIS This question requires corect diagnosis of plant indications to determine correct procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since AOP 18034-1 is entered for this malfunction and all the listed actions are taken except the DG is not emergency stopped. C. Incorrect. Plausible since the procedure entries are correct and the actions taken are ( correct except the methods used to stop the DG and control Tave. ( D. Incorrect. Plausible since misdiagnosis of a loss of only 120 V AC instrument power is a common mistake by operators due to the channel I bistable lights. Actions listed are correct, for the incorrect AOP entry. EOP entry also plausible since this is the expected response for power < P-10 setpoint. REFERENCES

1. AOP 18034-1, Loss of Class 1 E 125V DC Power. VEGP learning objectives:

1. LO-LP-60329-01:

Given that a loss of power has occurred to any of the following 125VDC vital buses and given the appropriate plant procedures, describe the operator actions required and why these actions are taken. a.1AD1 b. 1 BD1 c.1CD1 d. 1 DD1 2. LO-LP-60329-04: Page: 8of56 412112009 Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). Page: 90f56 4/2112009 ( Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). Page: 90f56 412112009 Rev 31 Date Approved 8of25 Page Number ES -0.1 REACTOR TRIP RESPONSE 7/22/2008 ACTION/EXPECTED RESPONSE 9. Perform the following:

a. Check 18009-C, STEAM GENERATOR TUBE LEAK -IN EFFECT. b. Perform steps 13, 15, 16, and 17 of this procedure as appropriate.

c. Go to 18009-C, STEAM GENERATOR TUBE LEAK, step in effect. d. Check other AOPs -IN EFFECT. RESPONSE NOT OBTAINED I F output breakers can NOT be opened, THEN perform the following:

UNIT 1 BUS-1: 161720 161730 161740 161760

Open the following switchyard breakers for the AFFECTED unit bus: UNIT2 BUS-2: BUS-1: BUS-2: 161910 161540 161660 161720(U1) 161820(U1 )
Notify Augusta TMC duty personnel referenced on the POD to verify AFFECTED unit Sub-Station Breaker open: UNIT 1: GOS-WHT (165058) UNIT 2: WARTHEN (505220 and 505110) a. Go to Step 9.d. d. Go to Step 10. o Step 9 continued on next page Printed April 20, 2009 at 18:32 ( ( Approved By J.B. Stanley :1' Procedure Number Rev jf 19001-C 31 Page Number 7/22/2008 ES -0.1 REACTOR TRIP RESPONSE Date Approved 8of25 ACTION/EXPECTED RESPONSE 9. Perform the following:

a. -Check 18009-C, STEAM GENERATOR TUBE LEAK -IN EFFECT. -b. Perform steps 13, 15, 16, and 17 of this procedure as appropriate.

-c. Go to 18009-C, STEAM GENERATOR TUBE LEAK, step in effect. d. Check other AOPs -IN EFFECT. -RESPONSE NOT OBTAINED IF output breakers can NOT be opened, THEN perform the following: UNIT 1 BUS-1 : 161720 161730 161740 161760

Open the following switchyard breakers for the AFFECTED unit bus: BUS-2: 161910 UNIT2 BUS-1 : 161540 161720(U1 ) 161820(U1 ) BUS-2: 161660
Notify Augusta TMC duty personnel referenced on the POD to verify AFFECTED unit Sub-Station Breaker open: UNIT 1: GOS-WHT -(165058) _ UNIT 2: WARTHEN (505220 and 505110) a. Go to Step 9.d. d. Go to Step 10. o Step 9 continued on next page Printed April 20, 2009 at 18:32 ber Rev 31 Date Approved 90f25 Page Number ES -0.1 REACTOR TRIP RESPONSE 7/22/2008 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED e. Initiate actions of AOPs in conjunction with remaining actions of this procedure.
10. Check PRZR level control: a. Instrument Air -AVAILABLE.

a. Perform the following:

_1) Establish Safety Grade Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. 2) Establish Safety Grade Letdown: a) Open RX HEAD VENT TO LETDOWN ISOLATION VL Vs:

HV-8095A
HV-8096A
HV-8095B
HV-8096B b) Open REACTOR HEAD VENT TO PRT flow control valves as necessary:
HV-0442A
HV-0442B _3) Go to Step 10.d. o Step 10 continued on next page Printed April 20, 2009 at 18:33 ( ( Approved By J.B. Stanley Date Approved 7/22/2008

',' '; " t ,;Y; '.'., Vogtle; Electric GeneraUng Plant t) .% . Procedure Number Rev 19001-C 31 Page Number ES -0.1 REACTOR TRIP RESPONSE 90f25 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED e. Initiate actions of AOPs in -conjunction with remaining actions of this procedure.

10. Check PRZR level control: a. Instrument Air -AVAILABLE.

a. Perform the following:

_1) Establish Safety Grade Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. 2) Establish Safety Grade Letdown: a) Open RX HEAD VENT TO LETDOWN ISOLATION VLVs:

HV-8095A -* HV-8096A -* HV-8095B -* HV-8096B -b) Open REACTOR HEAD VENT TO PRT flow control valves as necessary:
HV-0442A -* HV-0442B -_3) Go to Step 10.d. o Step 10 continued on next page Printed April 20, 2009 at 18:33 Date Approved 8/2/08 PURPOSE LOSS OF CLASS 1 E 125V DC POWER ABNORMAL OPERATING PROCEDURE CONTINUOUS USE Page Number 1 of 82 This procedure provides operator actions to be followed in the event that power is lost to one of the 125V DC Vital Busses (1 AD1, 1 BD1, 1 CD1 , or 1 DD1 ). Specific instructional steps will be found in the following sections:

A LOSS OF 125V DC BUS 1 AD1 B. LOSS OF 125V DC BUS 1 BD1 C. LOSS OF 125V DC BUS 1 CD1 D. LOSS OF 125V DC BUS 1 DD1 SYMPTOMS SECTION A LOSS OF 125V DC BUS 1 AD1

125V DC Vital Bus 1AD1 voltage low.
Loss of power to 1AY1A and 1AY2A 120V AC Vital Instrument Panels.
Loss of indicating lights on 1AA02, 1AB04, 1AB05, and 1AB15 Switchgear Controls.
Train A Main Steamline Isolation.
Train A Main Feedwater Isolation.

SECTION B. LOSS OF 125V DC BUS 1 BD1

125V DC Vital Bus 1 BD1 voltage low.
Loss of power to 1 BY1 Band 1 BY2B 120V AC Vital Instrument Panels.
Loss of indicating lights on 1 BA03, 1 BB06, 1 BB07, and 1 BB 16 Switchgear Controls.
Train B Main Steamline Isolation.
Train B Main Feedwater Isolation. ( Approved By S. A. Phillips Date Approved 8/2/08 PURPOSE LOSS OF CLASS 1 E 125V DC POWER ABNORMAL OPERATING PROCEDURE CONTINUOUS USE Procedure Number Rev 18034-1 10 Page Number 1 of 82 This procedure provides operator actions to be followed in the event that power is lost to one of the 125V DC Vital Busses (1AD1, 1 BD1, 1 CD1, or 1 DD1). Specific instructional steps will be found in the following sections:

A. LOSS OF 125V DC BUS 1AD1 B. LOSS OF 125V DC BUS 1 BD1 C. LOSS OF 125V DC BUS 1 CD1 D. LOSS OF 125V DC BUS 1 DD1 SYMPTOMS SECTION A. LOSS OF 125V DC BUS 1 AD1

125V DC Vital Bus 1AD1 voltage low.
Loss of power to 1AY1A and 1AY2A 120V AC Vital Instrument Panels.
Loss of indicating lights on 1 AA02, 1 AB04, 1 AB05, and 1 AB 15 Switchgear Controls.
Train A Main Steamline Isolation.
Train A Main Feedwater Isolation.

SECTION B. LOSS OF 125V DC BUS 1 BD1

125V DC Vital Bus 1 BD1 voltage low.
Loss of power to 1 BY1 Band 1 BY2B 120V AC Vital Instrument Panels.
Loss of indicating lights on 1 BA03, 1 BB06, 1 BB07, and 1 BB16 Switchgear Controls.
Train B Main Steamline Isolation.
Train B Main Feedwater Isolation.

Printed April 6, 2009 at 13:25 Approved By S. A. Phillips Procedure Number Rev 1 B034-1 10 Page Number Date Approved , BI2/0B LOSS OF CLASS 1E 125V DC POWER 40fB2 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

This procedure should be performed concurrent with 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.
RCP 1 undervoltage and underfrequency trips will NOT actuate.
See ATTACHMENT A for equipment responses, breaker and valve control loss, valve failures from loss of instrument air, and annunciator failures.

A 1. Verify reactor trip. A2. Initiate the Continuous Actions Page. A3. Dispatch an operator to 1 AA02 SWGR Room (CB-A4B). NOTE A 1. Perform the following:

a. Trip the reactor. b. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

IF DG1A is NOT running, it can NOT be started. A4. Check DG1A -RUNNING. A4. Go to Step A7. ( Approved By S. A. Phillips !;B0;. " Procedure Number Rev WWM '18034-1 10 4 of 82 Page Number Date Approved LOSS OF CLASS 1 E 125V DC POWER 8/2/08 A. LOSS OF 125V DC BUS 1 AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

This procedure should be performed concurrent with 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.
RCP 1 undervoltage and underfrequency trips will NOT actuate.
See ATTACHMENT A for equipment responses, breaker and valve control loss, valve failures from loss of instrument air, and annunciator failures.

A 1. Verify reactor trip. A2. A3. Initiate the Continuous Actions Page. Dispatch an operator to 1 AA02 SWGR Room (CB-A48). NOTE A 1. Perform the following: _a. Trip the reactor. b. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. IF DG1A is NOT running, it can NOT be started. A4. Check DG1A -RUNNING. _A4. Go to Step A7. Pnnted Apnl 6, 2009 at 13:27 Approved By S. A. Phillips Procedure Number Rev 18034-1 10 5 of 82 Date Approved Page Number LOSS OF CLASS 1 E 125V DC POWER 8/2/08 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE *A5. Check AC Emergency Bus 1AA02 . ENERGIZED BY OFFSITE POWER:

Offsite power to AC Emergency Bus 1AA02 -AVAILABLE.
1 AA02-05 Normal Feeder Breaker -CLOSED. NOTE RESPONSE NOT OBTAINED *A5. Perform the following:

a. WHEN offsite power is restored to 1 AA02, THEN perform Step A6. b. Go to Step A7. Diesel Generator Electrical Protective Trips are inoperable.

A6. Dispatch an operator to locally remove DG 1 A from service: a. Verify 1AA02-19 D/G 1A Output Breaker -OPEN. b. Stop the Diesel Generator by placing the Pull-To-Run/Push-To-Stop handswitch at the south end of the engine in STOP. A7. Check CH457/456 selected on 1 PS-455F PRZR PRESS CNTL SELECT. A7. Perform the following:

a. Place 1 HS-0455A PRZR PORV 455A in close. b. Place PRZR SPRAY controllers in manual:

1 PIC-455B
1PIC-455C o Step 7 continued on next page Approved By S. A. Phillips Procedure Number Rev 18034-1 10 Date Approved Page Number LOSS OF CLASS 1E 125V DC POWER 8/2/08 5 of 82 A. LOSS OF 125V DC BUS 1 AD1 ACTION/EXPECTED RESPONSE *A5. Check AC Emergency Bus 1AA02
ENERGIZED BY OFFSITE POWER:
Offsite power to AC Emergency Bus 1AA02 -AVAILABLE.
1 AA02-05 Normal Feeder Breaker -CLOSED. NOTE RESPONSE NOT OBTAINED *A5. Perform the following:

a. WHEN offsite power is restored to 1 AA02, THEN perform Step A6. b. Go to Step A7. Diesel Generator Electrical Protective Trips are inoperable.

A6. Dispatch an operator to locally ( remove DG1A from service: a. Verify 1AA02-19 D/G 1A Output Breaker -OPEN. b. Stop the Diesel Generator by placing the Pu 11-T o-Ru n/Push-To-Stop handswitch at the south end of the engine in STOP. A7. Check CH457/456 selected on 1 PS-455F PRZR PRESS CNTL SELECT. A7. Perform the following:

a. Place 1 HS-0455A PRZR PORV 455A in close. b. Place PRZR SPRAY controllers in manual:

1 PIC-455B
1PIC-455C o Step 7 continued on next page Printed April 6, 2009 at 13:32 Procedure Number Rev 1 B034-1 10 Date Approved 6ofB2 Page Number LOSS OF CLASS 1E 125V DC POWER B/2/0B A. LOSS OF 125V DC BUS 1 AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED AB. Isolate letdown relief flowpath by performing the following:

Printed a. Close letdown orifice isolation valves:

1-HV-B149A
1-HV-B149B
1-HV-B149C

b. Close letdown isolation valves:

1-LV-459
1-LV-460 at 13:32 c. -d. e. f. -g. h. i. -Control PRZR pressure between 2220 and 2250 psig using heaters and sprays. Adjust PRZR Master Pressure Controller 1 PIC-455A output to approximately 25%. Select CH457/456 on 1PS-455F.

Place 1 HS-0455A in AUTO. Place PRZR spray valve controllers in AUTO. Return 1 PIC-455A to AUTO. Verify RCS pressure stable at or trending to 2235 PSIG. Approved By S. A. Phillips Page Number Date Approved LOSS OF CLASS 1E 125V DC POWER 6of82 8/2/08 A. LOSS OF 125V DC BUS 1 AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED A8. Isolate letdown relief flowpath by performing the following:

a. Close letdown orifice isolation valves:

1-HV-8149A

-* 1-HV-8149B -* 1-HV-8149C

b. Close letdown isolation valves:

1-LV-459 -* 1-LV-460 Pnnted Apnl 6, 2009 at 13:32 c. -d. -e. -f. --g. h. -i. -Control PRZR pressure between 2220 and 2250 psig using heaters and sprays. Adjust PRZR Master Pressure Controller 1 PIC-455A output to approximately 25%. Select CH457/456 on 1 PS-455F. Place 1 HS-0455A in AUTO. Place PRZR spray valve controllers in AUTO. Return 1 PIC-455A to AUTO. Verify RCS pressure stable at or trending to 2235 PSIG.

By Rev S. A. Phillips Date Approved Page Number LOSS OF CLASS 1E 125V DC POWER 8/2/08 7of82 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE A9. Dispatch an operator to restore power to 120V AC Instrument Panel1AY1A (CB-B52):

a. Open 1 A Y1 A-02 normal feeder breaker from Inverter 1AD111. b. Verify 1 ABC-09 (CB-B76) regulated transformer supply breaker for 1 ABC09X -CLOSED. c. Close 1AY1A-01 alternate source breaker to 1AY1A instrument panel. A10. Dispatch an operator to restore power to 120V AC Instrument Panel 1 AY2A (AB-118):

a. Open 1 A Y2A-02 normal feeder breaker from Inverter 1AD1111. b. Verify 1ABB-40 (AB-118) regulated transformer supply breaker for 1 ABC09X -CLOSED. c. Close 1AY2A-01 alternate source breaker to 1AY2A instrument panel. A 11. Dispatch an operator to restore power to Bus 1AD1 (CB-B52) by initiating 13405, 125V DC 1 E ELECTRICAL DISTRIBUTION SYSTEM. RESPONSE NOT OBTAINED A9. Initiate 18032, LOSS OF 120 VOLT AC INSTRUMENT POWER. _A10. Initiate 18032, LOSS OF 120 VOLT AC INSTRUMENT POWER. 10 ( Approved By S. A. Phillips Date Approved 8/2/08 Page Number 7 of 82 A. LOSS OF 125V DC BUS 1 AD1 A9. ACTION/EXPECTED RESPONSE Dispatch an operator to restore power to 120V AC Instrument Panel 1 A Y1 A (CB-B52):

a. Open 1 A Y1 A-02 normal feeder breaker from Inverter 1AD111. _b. Verify 1ABC-09 (CB-B76) regulated transformer supply breaker for 1 ABC09X -CLOSED. c. Close 1AY1A-01 alternate source -breaker to 1 A Y1 A instrument panel. A 10. Dispatch an operator to restore power to 120V AC Instrument Panel 1 A Y2A (AB-118):

a. Open 1 A Y2A-02 normal feeder breaker from Inverter 1AD1111. b. Verify 1ABB-40 (AB-118) regulated transformer supply breaker for 1 ABC09X -CLOSED. c. Close 1AY2A-01 alternate source -A11. breaker to 1 AY2A instrument panel. Dispatch an operator to restore power to Bus 1AD1 (CB-B52) by initiating 13405, 125V DC 1 E ELECTRICAL DISTRIBUTION SYSTEM. Printed April 6, 2009 at 13.32 A9. RESPONSE NOT OBTAINED Initiate 18032, LOSS OF 120 VOLT AC INSTRUMENT POWER. _A 10. Initiate 18032, LOSS OF 120 VOLT AC INSTRUMENT POWER.

Approved By S. A. Phillips Date Approved 8/2/08 Procedure Number Rev 18034-1 10 Page Number LOSS OF CLASS 1 E 125V DC POWER 13 of 82 ATTACHMENT A Sheet 1 of 23 LOSS OF 125V DC BUS 1 AD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN A 125V DC POWER NOTE Feeder Breakers must be locally controlled in the event the transfer to an alternate power supply is required. IF DG1A is not running, it may not be selected as an alternate power source.

Main Feedwater Isolation, Bypass Feedwater Isolation, and Bypass Feedwater Regulation Valves close resulting in Feedwater Isolation.
Main Steam Isolation and Bypass Steam Isolation Train A Valves close resulting in steamline isolation.
Above P10, Reactor and Turbine trip occurs from loss of main feedwater. (
Below P1 0, Reactor trip occurs from Intermediate Range Instrumentation.
Control Power is lost to 1AA02, 1AB04, 1AB05, and 1AB15 SWGR Breakers.
DG1A control power to Generator Control Panel PDG1 and Engine Control Panel PDG2 is lost rendering the DG inoperable; if running, it will fail as is with a loss of electrical protective trips, frequency, and voltage control. Due to loss of power to the Low Speed Relay, the generator space, Engine Lube Oil and Jacket Water Heaters and Lube Oil and Jacket Water Keep-Warm Pumps will come on.
Loss of Train A DG AUTO sequencer reset.
Power to Inverters 1AD111 and 1AD1111 is lost causing 120V AC Vital Busses 1AY1A and 1 A Y2A to de-energize.
Instrument Air Containment Isolation Valve 1-HV-9378 closes resulting in loss of instrument air inside Containment.
Power To Isolation Panel1ACQIP1 is lost rendering the annunciators in Train A inoperable.
Pressurizer PORV 1-PV-455A fails closed.
TDAFW Steam Supply 1-HV-3019 fails as is. Printed April 6, 2009 at 13:28 Approved By S. A. Phillips Date Approved 8/2/08 Procedure Number Rev ..... 18034-1 10 Page Number LOSS OF CLASS 1E 125V DC POWER 13 of 82 ATTACHMENT A Sheet 1 of 23 LOSS OF 125V DC BUS 1AD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN A 125V DC POWER NOTE Feeder Breakers must be locally controlled in the event the transfer to an alternate power supply is required.

IF DG1A is not running, it may not be selected as an alternate power source.

Main Feedwater Isolation, Bypass Feedwater Isolation, and Bypass Feedwater Regulation Valves close resulting in Feedwater Isolation.
Main Steam Isolation and Bypass Steam Isolation Train A Valves close resulting in steamline isolation.
Above P10, Reactor and Turbine trip occurs from loss of main feedwater. (
Below P1 0, Reactor trip occurs from Intermediate Range Instrumentation.
Control Power is lost to 1AA02, 1AB04, 1AB05, and 1AB15 SWGR Breakers.
DG1A control power to Generator Control Panel PDG1 and Engine Control Panel PDG2 is lost rendering the DG inoperable; if running, it will fail as is with a loss of electrical protective trips, frequency, and voltage control. Due to loss of power to the Low Speed Relay, the generator space, Engine Lube Oil and Jacket Water Heaters and Lube Oil and Jacket Water Keep-Warm Pumps will come on.
Loss of Train A DG AUTO sequencer reset.
Power to Inverters 1AD111 and 1AD1111 is lost causing 120V AC Vital Busses 1AY1A and 1 AY2A to de-energize.
Instrument Air Containment Isolation Valve 1-HV-9378 closes resulting in loss of instrument air inside Containment.
Power To Isolation Panel1ACQIP1 is lost rendering the annunciators in Train A inoperable.
Pressurizer PORV 1-PV-455A fails closed.
TDAFW Steam Supply 1-HV-3019 fails as is. Printed April 6, 2009 at 13:28 Approved By S. A. Phillips Date Approved 8/2/08 LOSS OF CLASS 1 E 125V DC POWER ATTACHMENT A LOSS OF 125V DC BUS 1 AD1 EQUIPMENT POWERED FROM 1AD11 Procedure Number Rev 18034-1 10 Page Number 14 of 82 Sheet 2 of 23
Loss of Trip Breaker RTA and BYA position indication 1-ZL-40045.

RTA cannot be opened or closed from QMCB. The only operable trip is from the UV board in SSPS or the local trip plate.

Steam Dumps fail closed.
ARV 1-HV-3000 and 1-HV-3030 will not operate from the Control Room.
Loss of RCP 1 Class 1 E breaker control including RCP underfrequency trip and undervoltage input to SSPS. The Train A RCP underfrequency trip is inoperable on RCPs 2,3, and 4.
DG1A Field Flash, Starting Air Solenoids, Governor, and voltage regulator is inoperable.

Field Flash and Starting Air Solenoids have redundant feed from 1AD12.

Loss of Train A Control Room and Fuel Handling Building Isolation signals.
Loss of alarm indication lights for Essential Chiller 1-1592-C7-001.

Printed April 6, 2009 at 13:28 Procedure Number Rev Approved By S. A. Phillips l' 18034-1 10 Date Approved Page Number LOSS OF CLASS 1 E 125V DC POWER 14 of 82 8/2/08 ATTACHMENT A Sheet 2 of 23 LOSS OF 125V DC BUS 1AD1 EQUIPMENT POWERED FROM 1AD11

Loss of Trip Breaker RTA and BY A position indication 1-ZL-40045.

RT A cannot be opened or closed from QMCS. The only operable trip is from the UV board in SSPS or the local trip plate.

Steam Dumps fail closed.
ARV 1-HV-3000 and 1-HV-3030 will not operate from the Control Room.
Loss of RCP 1 Class 1 E breaker control including RCP underfrequency trip and undervoltage input to SSPS. The Train A RCP underfrequency trip is inoperable on RCPs 2,3, and 4.
DG1A Field Flash, Starting Air Solenoids, Governor, and voltage regulator is inoperable.

Field Flash and Starting Air Solenoids have redundant feed from 1AD12. (

Loss of Train A Control Room and Fuel Handling Building Isolation signals.
Loss of alarm indication lights for Essential Chiller 1-1592-C7-001.

Printed April 6, 2009 at 13:28

5. 062AA2.02 00ll1l1ILOSS OF NSCW/C/A -3.6INEWIHLl5/SROIDS/TNT Given the following conditions at 38% power: -ACCW pump 2 is in service -CCW pumps 2 & 4 are in service -NSCW Pump 5 is danger tagged Train A NSCW indications:

Train B NSCW indications: -Supply header pressure 45 psig -Supply header pressure 58 psig -Supply header flow 8,000 gpm -Supply header flow 25,000 gpm -Return header flow 8,000 gpm -Return header flow 10,000 gpm Which of the following choices contains the correct procedural entry and actions? A'! Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW trains. Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C, trip the RCPs and isolate eves letdown. B. Enter AOP 18021-C, Loss of NSCW, due to leakage on train A NSCW. Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C. Trip the RCPs and isolate cves letdown if cooling not restored in 10 minutes. C. Enter AOP 18021-C, Loss of NSeW, due to leakage on train B NSCW. Place train B NSCW pumps in PTL, shift CCW pumps to train A, start ACCW pump #1 and remain in 18021-C. D. Enter AOP 18021-C, due to a trip of a running pump on train A NSCW. Place NSCW train A in single pump operation, trip RCPs if seal temperatures exceed 230 F and remain in 18021-C. 062 Loss of Nuclear Service Water AA2.02 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss KIA MATCH ANALYSIS Page: 10 of 56 4/2112009 ( 5. 062AA2.02 00llllllLOSS OF NSCW/C/A -3.6INEWIHL15/SRO/DS/TNT Given the following conditions at 38% power: -ACCW pump 2 is in service -CCW pumps 2 & 4 are in service -NSCW Pump 5 is danger tagged Train A NSCW indications: Train B NSCW indications: -Supply header pressure 45 psig -Supply header pressure 58 psig -Supply header flow 8,000 gpm -Supply header flow 25,000 gpm -Return header flow 8,000 gpm -Return header flow 10,000 gpm Which of the following choices contains the correct procedural entry and actions? A'I' Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW trains. Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C, trip the RCPs and isolate CVCS letdown. B. Enter AOP 18021-C, Loss of NSCW, due to leakage on train A NSCW. Place all NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C. Trip the RCPs and isolate CVCS letdown if cooling not restored in 10 minutes. C. Enter AOP 18021-C, Loss of NSCW, due to leakage on train B NSCW. Place train B NSCW pumps in PTL, shift CCW pumps to train A, start ACCW pump #1 and remain in 18021-C. D. Enter AOP 18021-C, due to a trip of a running pump on train A NSCW. Place NSCW train A in single pump operation, trip RCPs if seal temperatures exceed 230 F and remain in 18021-C. KIA 062 Loss of Nuclear Service Water AA2.02 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss KIA MATCH ANALYSIS Page: 10 of 56 4/2112009 procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I OISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since all the actions of this choice are correct for the failure indications given for train B. Train A indications are for trip of a running pump with failure to auto start the standby pump. C. Incorrect. Plausible since all the actions stated are correct for the failure listed in the first part of this choice. The only incorrect action is swapping of the ACCW pumps. D. Incorrect. Plausible since AOP entry is required due to a trip of a train A NSCW pump and the system would be placed in single pump operation. However, the single pump operation is controlled by SOP 13150-1. The RCP's have a normal trip criteria of 230 F for the seals, however, they should be tripped based on the loss of both trains of NSCW not the seal temperature. REFERENCES

1. AOP 18021-C, Loss of NSCW. 2. AOP 18022-C, Loss of ACCW. 3. SOP 13150-1, Nuclear Service Cooling Water System. VEGP learning objectives:

LO-LP-60317-03: Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-60317-02: Describe the operator action(s) required if NSCW is lost and neither NSCW Train can be placed into operation. Page: 11 of 56 4/2112009 procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since all the actions of this choice are correct for the failure indications given for train B. Train A indications are for trip of a running pump with failure to auto start the standby pump. C. Incorrect. Plausible since all the actions stated are correct for the failure listed in the first part of this choice. The only incorrect action is swapping of the ACCW pumps. D. Incorrect. Plausible since AOP entry is required due to a trip of a train A NSCW pump and the system would be placed in single pump operation. However, the single pump operation is controlled by SOP 13150-1. The RCP's have a normal trip criteria of 230 F for the seals, however, they should be tripped based on the loss of both trains of NSCW not the seal temperature. ( REFERENCES

1. AOP 18021-C, Loss of NSCW. 2. AOP 18022-C, Loss of ACCW. 3. SOP 13150-1, Nuclear Service Cooling Water System. VEGP learning objectives:

LO-LP-60317-03: Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). LO-LP-60317-02: Describe the operator action(s) required if NSCW is lost and neither NSCW Train can be placed into operation. Page: 11 of 56 4/2112009 Approval Date Procedure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS Unit COMMON Abnormal Operating Procedures Revision No. Page No. LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 18021-C 16 1 of 13 PURPOSE PRB REVIEW REQUIRED This procedure addresses the loss or degraded operation of one or more trains of Nuclear Service Cooling Water. SYMPTOMS

Trip of operating NSCW pumps and failure of standby pump to start.
Dropping NSCW Supply Header pressure.
Large difference between Supply Header flow and Return Header flow, indicating a large leak.
NSCW Tower Basin temperature rising above 90°F.
High temperature or low flow alarms on any components or systems cooled by NSCW. MAJOR ACTIONS
Determine condition causing loss or degraded operation of NSCW.
Transfer loads to unaffected train.
Correct or repair condition causing loss or degraded operation of NSCW. Approval Date ( 'Procedure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS Unit COMMON Abnormal Operating Procedures Revision No. Page No. LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM 18021-C 16 1 of 13 PURPOSE PRB REVIEW REQUIRED This procedure addresses the loss or degraded operation of one or more trains of Nuclear Service Cooling Water. SYMPTOMS
Trip of operating NSCW pumps and failure of standby pump to start.
Dropping NSCW Supply Header pressure.
Large difference between Supply Header flow and Return Header flow, indicating a large leak.
NSCW Tower Basin temperature rising above 90°F.
High temperature or low flow alarms on any components or systems cooled by NSCW. MAJOR ACTIONS
Determine condition causing loss or degraded operation of NSCW.
Transfer loads to unaffected train.
Correct or repair condition causing loss or degraded operation of NSCW.

PROCEDURE NO. REVISION NO. VEGP 18021-C ACTION/EXPECTED RESPONSE D 1. Check if catastrophic leakage from NSCW system -EXISTS D 2. Place affected train NSCW pump handswitches in PULL-TO-LOCK. D 3. Depress both Emergency Stop pushbuttons for the affected DG. PAGE NO. 16 2 of 13 RESPONSE NOT OBTAINED D 1. Go to Step 6. ( PROCEDURE NO. REVISION NO. VEGP 18021-C ACTION/EXPECTED RESPONSE D 1. Check if catastrophic leakage from NSCW system -EXISTS D 2. Place affected train NSCW pump handswitches in PULL-TO-LOCK. D 3. Depress both Emergency Stop pushbuttons for the affected DG. PAGE NO. 16 2 of 13 RESPONSE NOT OBTAINED D 1. Go to Step 6. PROCEDURE NO. REVISION NO. VEGP 18021-C ACTION/EXPECTED RESPONSE 4. Verify proper operation of UNAFFECTED NSCW train: 0-Two pumps running. 0-Supply header pressure greater than 70 psig: Train A: Train B: PI-1636 PI-1637 0-Supply header temperature computer indication less than 90°F: Train A: Train B: T2601 T2602 Supply header flow approximately 17,000 gpm: Train A: Train B: FI-1640B FI-1641B 05. Go to Step 13. 16 PAGE NO. 3 of 13 RESPONSE NOT OBTAINED 4. IF neither NSCW train can be placed in normal, two pump operation, THEN perform the following: Oa. Trip the reactor. Db . Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. Dc. Trip all reactor coolant pumps. Od. Isolate letdown. De. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR SERVICE COOLING WATER SYSTEM. Of. Verify train-related CCP or NCP running and seal injection flow established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. Og. Check RCP No.1 seal temperatures less than 220° F. Oh. IF RCP No.1 seal temperatures greater than 220°F, THEN do NOT attempt to restart RCPs prior to a status evaluation. ( PROCEDURE NO. REVISION NO. VEGP 18021-C ACTION/EXPECTED RESPONSE 4. Verify proper operation of UNAFFECTED NSCW train: O. Two pumps running. O. Supply header pressure greater than 70 psig: Train A: PI-1636 Train B: PI-1637 O. Supply header temperature computer indication less than 90°F: Train A: T2601 Train B: T2602 O. Supply header flow approximately 17,000 gpm: Train A: FI-1640B Train B: FI-1641B 05. Go to Step 13. 16 PAGE NO. 3 of 13 RESPONSE NOT OBTAINED 4. IF neither NSCW train can be placed in normal, two pump operation, THEN perform the following: o a. Trip the reactor. Db. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. Dc. Trip all reactor coolant pumps. Od. Isolate letdown. De. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR SERVICE COOLING WATER SYSTEM. Of. Verify train-related CCP or NCP running and seal injection flow established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. o g. Check RCP No. 1 seal temperatures less than 220°F. Oh. IF RCP No.1 seal temperatures greater than 220°F, THEN do NOT attempt to restart RCPs prior to a status evaluation. ( ( 6. WEIIEG2.4.2 OOI/1/1/LOSS OF ECRIC/A -4.6/B -HARRIS 200S/HL15/SROIDS/TNT Initial Conditions: -A large LOCA has occured -Neither train of RHR could be aligned for cold leg recirculation -EOP 19111-C, Loss of Emergency Coolant Recirculation, has been implemented Current Conditions: -RWST level is now 7% and lowering -Integrity CSF Status Tree is Orange Which one of the following describes the action and procedure usage required? A. Stop all pumps taking suction from the RWST; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration Procedures (FRP's). B. Reduce ECCS flow from the RWST to ONE (1) train running; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration Procedures (FRP's). Stop all pumps taking suction from the RWST; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock D. Reduce ECCS flow from the RWST to ONE (1) train running; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock. Page: 12 of 56 4/2112009 ( ( 6. WE1IEG2.4.2 OOI/1/1/LOSS OF ECRIC/A -4.6/B -HARRIS 200S/HL15/SRO/DS/TNT Initial Conditions: -A large LOCA has occured -Neither train of RHR could be aligned for cold leg recirculation -EOP 19111-C, Loss of Emergency Coolant Recirculation, has been implemented Current Conditions: -RWST level is now 7% and lowering -Integrity CSF Status Tree is Orange Which one of the following describes the action and procedure usage required? A. Stop all pumps taking suction from the RWST; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration Procedures (FRP's). B. Reduce ECCS flow from the RWST to ONE (1) train running; Do NOT go to 19241-C, Response to Imminent Pressurized Thermal Shock, because the actions in 19111-C take priority over the Function Restoration Procedures (FRP's). O! Stop all pumps taking suction from the RWST; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock D. Reduce ECCS flow from the RWST to ONE (1) train running; Go to 19241-C, FR-P.1 Response to Imminent Pressurized Thermal Shock. Page: 12 of 56 4/2112009 KIA WE11 Loss of Emergency Coolant Recirculation EG2.4.2 Emergency Procedures/Plan Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. KIA MATCH ANALYSIS This question requires the applicant to diagnose plant parameters and determine actions to take within EOP 19111-C and decide on appropriate EOP transitions to based on challenges to CSFSTs. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible because some ECA procedures prevent implementation of FRPs. B. Incorrect. Plausible because the applicant must correctly intrepret the correct actions to take based on RWST level, both of which are contained within EOP 19111-C. Applicant must also determine if FRP's are implemented during this ECA. C. Correct. D. Incorrect. Plausible because the applicant must correctly intrepret the correct actions to take based on RWST level, both of which are contained within EOP 19111-C. Applicant must determine if FRP's are implemented during this ECA. REFERENCES

1. EOP 19111-C, ECA-1.1 Loss of Emergency Coolant Recirculation VEGP learning objectives:

LO-PP-37115-10: Describe the actions taken to keep the core covered and protect ECCS equipment when RWST level drops below 8% during a loss of emergency coolant recirculation. Page: 13 of 56 4/21/2009 ( ( KIA WE11 Loss of Emergency Coolant Recirculation EG2.4.2 Emergency Procedures/Plan Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. KIA MATCH ANALYSIS This question requires the applicant to diagnose plant parameters and determine actions to take within EOP 19111-C and decide on appropriate EOP transitions to based on challenges to CSFSTs. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible because some ECA procedures prevent implementation of FRPs. B. Incorrect. Plausible because the applicant must correctly intrepret the correct actions to take based on RWST level, both of which are contained within EOP 19111-C. Applicant must also determine if FRP's are implemented during this ECA. C. Correct. D. Incorrect. Plausible because the applicant must correctly intrepret the correct actions to take based on RWST level, both of which are contained within EOP 19111-C. Applicant must determine if FRP's are implemented during this ECA. REFERENCES

1. EOP 19111-C, ECA-1.1 Loss of Emergency Coolant Recirculation VEGP learning objectives:

LO-PP-37115-10: Describe the actions taken to keep the core covered and protect ECCS equipment when RWST level drops below 8% during a loss of emergency coolant recirculation. Page: 13 of 56 4/2112009 1 . Ell EA2.2 005 Page: 1 Initial Conditions:

A LOCA has occurred.
The crew is performing actions of EPP-012, Loss of Emergency Coolant Recirculation, based on plant conditions upon transition from Path-1.
RWST level is 3% and lowering.

Current Conditions:

Integrity CSF Status Tree indicates Orange. Which ONE of the following describes the action and procedure usage required?

A. Stop all pumps taking suction from the RWST; Remain in EPP-012 because actions in EPP-012 are expected to cause an Orange condition on Integrity. B. Reduce ECCS flow from the RWST to ONE (1) train running; Remain in EPP-012 because actions in EPP-012 are expected to cause an Orange condition on Integrity. Stop all pumps taking suction from the RWST; Go to FRP-P.1, Response to Imminent Pressurized Thermal Shock. D. Reduce ECCS flow from the RWST to ONE (1) train running; Go to FRP-P.1, Response to Imminent Pressurized Thermal Shock. 2/20/2009 1 . Ell EA2.2 005 Initial Conditions:

A LOCA has occurred.
The crew is performing actions of EPP-012, Loss of Emergency Coolant Recirculation, based on plant conditions upon transition from Path-1.
RWST level is 3% and lowering.

Current Conditions:

Integrity CSF Status Tree indicates Orange. Which ONE of the following describes the action and procedure usage required?

A. Stop all pumps taking suction from the RWST; Remain in EPP-012 because actions in EPP-012 are expected to cause an Orange condition on Integrity. B. Reduce ECCS flow from the RWST to ONE (1) train running; Remain in EPP-012 because actions in EPP-012 are expected to cause an Orange condition on Integrity. ( Stop all pumps taking suction from the RWST; Go to FRP-P.1, Response to Imminent Pressurized Thermal Shock. D. Reduce ECCS flow from the RWST to ONE (1) train running; Go to FRP-P.1, Response to Imminent Pressurized Thermal Shock. Page: 1 2/20/2009 Approved By S. A. Phillips Date Approved 9/20/08 *29. Procedure Number 19111-C 30 ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Page Number 22 of 50 ACTION/EXPECTED RESPONSE Check if RCPs must be stopped: a. Check the following: Seal number 1 differential pressure -LESS THAN 200 PSID. -OR-Seal number 1 leakoff flow -LESS THAN 0.2 GPM. b. RESPONSE NOT OBTAINED a. IF neither condition satisfied, THEN go to Step 30. -Stop affected RCPs. c. Close Spray Valve for idle RCP: RCP 1: PIC-0455C RCP 4: PIC-0455B

30. Check RCS WR Hot Leg 30. Go to Step 44. temperature

-GREATER THAN 200°F. 31. Check RWST level -LESS THAN 8%. 31. Return to Step 1. 32. Stop Pumps taking suction from RWST and place switches in PULL-TO-LOCK positions:

RHR Pumps ..u. ,.vt.-.r Ie .
SI Pumps ,A. __ /"J
CCPs
CS Pumps dM 4-{jv Printed April 6, 2009 at 13:44 ( Approved By S. A. Phillips Date Approved 9/20/08 *29. Procedure Number Rev 19111-C 30 ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Page Number 22 of 50 ACTION/EXPECTED RESPONSE Check if RCPs must be stopped: a. Check the following:

Seal number 1 differential pressure -LESS THAN 200 PSID. -OR-Seal number 1 leakoff flow -LESS THAN 0.2 GPM. b. RESPONSE NOT OBTAINED a. IF neither condition satisfied, THEN go to Step 30. -Stop affected RCPs. c. Close Spray Valve for idle RCP: RCP 1: PIC-0455C RCP 4: PIC-0455B

30. Check RCS WR Hot Leg 30. Go to Step 44. temperature

-GREATER THAN 200°F. 31. Check RWST level -LESS THAN 8%. 31. Return to Step 1. 32. Stop Pumps taking suction from RWST and place switches in PULL-TO-LOCK positions: , f) If -IJ,

RHR Pumps ..u. .
SI Pumps ,A. __ /'J
CCPs ft'C: J
CS Pumps ?tIzw 4-t/I'tI--

% Printed April 6, 2009 at 13:44

7. 003AG2.4.35 00ll1l2/DROPPED ROD/CIA -4.0INEW/HLl5/SRO/DS/TNT Initial conditions:

-Unit at 100% power for last 10 weeks -All rods out at 228 steps Current conditions: -Rod at bottom alarm is alarming -Control Bank D rod H-8 rod bottom LED lit -Tave is lowering -QPTR & AFD remain within limits Which of the following contains the correct diagnosis and corrective actions? A. Enter AOP 18003-C, Section A, Dropped Rods in Mode 1 . Reduce thermal power < 75% prior to rod recovery, rod pulls are limited to 3 step increments. When rod recovery is begun the RIL alarm is INOP until the BOU is reset. B. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1. Reduce thermal power < 65% prior to rod recovery, rod pulls are limited to 3 step increments. When rod recovery is begun the RIL alarm is INOP until the BOU is reset. Enter AOP 18003-C, Section A, Dropped Rods in Mode 1 . Reduce thermal power < 65% prior to rod recovery, the 3 step rod pull limit may be suspended for this condition. When rod recovery is begun the RIL alarm is INOP until the PIA converter is reset. D. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1. Reduce thermal power < 75% prior to rod recovery, the 3 step rod pull limit may be suspended for this condition. When rod recovery is begun the RIL alarm is INOP until the PIA converter is reset. 003 Dropped Control Rod AG2.4.35 Emergency Procedures/Plan Page: 14 of 56 Knowledge of local auxiliary operator tasks during an emergency and the 4/2112009 ( 7. 003AG2.4.35 OOl/l/2/DROPPED ROD/C/A -4.0INEW/HL15/SRO/DS/TNT Initial conditions: -Unit at 100% power for last 10 weeks -All rods out at 228 steps Current conditions: -Rod at bottom alarm is alarming -Control Bank D rod H-8 rod bottom LED lit -T ave is lowering -QPTR & AFD remain within limits Which of the following contains the correct diagnosis and corrective actions? A. Enter AOP 18003-C, Section A, Dropped Rods in Mode 1. Reduce thermal power < 75% prior to rod recovery, rod pulls are limited to 3 step increments. When rod recovery is begun the RIL alarm is INOP until the BOU is reset. B. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1. Reduce thermal power < 65% prior to rod recovery, rod pulls are limited to 3 step increments. When rod recovery is begun the RIL alarm is INOP until the BOU is reset. C'!" Enter AOP 18003-C, Section A, Dropped Rods in Mode 1 . Reduce thermal power < 65% prior to rod recovery, the 3 step rod pull limit may be suspended for this condition. When rod recovery is begun the RIL alarm is INOP until the PIA converter is reset. D. Enter AOP 18003-C, Section C, Misaligned Rods in Mode 1. Reduce thermal power < 75% prior to rod recovery, the 3 step rod pull limit may be suspended for this condition. When rod recovery is begun the RIL alarm is INOP until the PIA converter is reset. 003 Dropped Control Rod AG2.4.35 Emergency Procedures/Plan Page: 14 of 56 Knowledge of local auxiliary operator tasks during an emergency and the 4/2112009 resultant operational effects. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine correct procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since section A is the correct section and power is required to be initially reduced to < 75% and 3 step rod pull increments are normally required and the RIL alarm is INOP when rod recovery is begun. B. Incorrect. Plausible since power is required to be reduced to < 65% prior to rod recovery, rod pull increments are normally limited to 3 steps, and the RIL alarm is INOP when rod revovery is begun. C. Correct. D. Incorrect. Plausible since power is initially required to be reduced to < 75%, the 3 step rod pull is suspended for this condition, and the RIL alarm is INOP for the reason listed. REFERENCES

1. AOP 18003-C, Rod Control System Malfunction.

2. SOP 13502-1, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. VEGP learning objectives:

1. LO-LP-60303-04:

Describe the effects of failing to reset the PIA converter (Bank Demand Position Display) following a dropped rod retrieval.

2. LO-LP-60303-07:

Describe why reactor power must be less than 65% or 10% below most limited power distribution restriction prior to dropped rod retrieval.

3. LO-LP-60303-15:

Describe the effects of failing to reset the PIA converter (Bank Demand Position Display) following a misaligned rod recovery. Page: 15 of 56 4/2112009 ( resultant operational effects. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine correct procedures to enter and then the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since section A is the correct section and power is required to be initially reduced to < 75% and 3 step rod pull increments are normally required and the RIL alarm is INOP when rod recovery is begun. B. Incorrect. Plausible since power is required to be reduced to < 65% prior to rod recovery, rod pull increments are normally limited to 3 steps, and the RIL alarm is INOP when rod revovery is begun. C. Correct. D. Incorrect. Plausible since power is initially required to be reduced to < 75%, the 3 step rod pull is suspended for this condition, and the RIL alarm is INOP for the reason listed. REFERENCES

1. AOP 18003-C, Rod Control System Malfunction.

2. SOP 13502-1, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. VEGP learning objectives:

1. LO-LP-60303-04:

Describe the effects of failing to reset the PIA converter (Bank Demand Position Display) following a dropped rod retrieval.

2. LO-LP-60303-07:

Describe why reactor power must be less than 65% or 10% below most limited power distribution restriction prior to dropped rod retrieval.

3. LO-LP-60303-15:

Describe the effects of failing to reset the PIA converter (Bank Demand Position Display) following a misaligned rod recovery. Page: 15 of 56 4/2112009 ( 4. LO-LP-60303-18: Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). Page: 16 of 56 4/2112009 ( 4. LO-LP-60303-18: Given conditions and/or indications, determine the required AOP to enter (including subsections, as applicable). Page: 16 of 56 4/2112009 Approval Date ( Vogtle Electric Generating Plant NUCLEAR OPERATIONS Procedure No. 18003-C Revision No. 23 Uni t COMMON Page No. 1 of 27 Abnormal Operating Procedures ROD CONTROL SYSTEM MALFUNCTION PURPOSE PRB REVIEW REQUIRED This procedure provides instructions for malfunctions of the Rod Control System resulting in uncontrolled rod motion, dropped or misaligned rods. SYMPTOMS SECTION A, DROPPED RODS IN MODE 1

ALB10-E5 ROD AT BOTTOM
ALB10-F2 POWER RANGE HI NEUTRON FLX RATE ALERT
ALB10-C2 POWER RANGE CHANNEL DEVIATION
Rod bottom LED on digital rod position indication.
Tavg dropping.

SECTION B, UNCONTROLLED CONTINUOUS ROD RODS IN ALL MODES

Rod motion with invalid demand from the Automatic Rod Control System.
Failure of rods to stop moving when the Rod Motion Switch is released.

SECTION C, MISALIGNED RODS IN MODE 1

ALB10-C2 POWER RANGE CHANNEL DEVIATION
ALB10-D2 POWER RANGE UP DET HI FLX DEV
ALB10-E2 POWER RANGE LWR DET HI FLX DEV
Failure of ALB10-C4 ROD BANK LO LIMIT or ALB10-D4 ROD BANK LO-LO LIMIT to reset during rod withdrawal.
Misaligned rod.
Quadrant power tilt ratio calculation exceeds 1.02. Approval Date ( Procedure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS 18003-C Revision No. 23 Uni t COMMON Page No. 1 of 27 Abnormal Operating Procedures ROD CONTROL SYSTEM MALFUNCTION PURPOSE PRB REVIEW REQUIRED This procedure provides instructions for malfunctions of the Rod Control System resulting in uncontrolled rod motion, dropped or misaligned rods. SYMPTOMS SECTION A, DROPPED RODS IN MODE 1
ALB10-E5 ROD AT BOTTOM
ALB10-F2 POWER RANGE HI NEUTRON FLX RATE ALERT
ALB10-C2 POWER RANGE CHANNEL DEVIATION
Rod bottom LED on digital rod position indication.
Tavg dropping.

SECTION B, UNCONTROLLED CONTINUOUS ROD RODS IN ALL MODES

Rod motion with invalid demand from the Automatic Rod Control System.
Failure of rods to stop moving when the Rod Motion Switch is released.

SECTION C, MISALIGNED RODS IN MODE 1

ALB10-C2 POWER RANGE CHANNEL DEVIATION
ALB10-D2 POWER RANGE UP DET HI FLX DEV
ALB10-E2 POWER RANGE LWR DET HI FLX DEV
Failure of ALB10-C4 ROD BANK LO LIMIT or ALB10-D4 ROD BANK LO-LO LIMIT to reset during rod withdrawal.
Misaligned rod.
Quadrant power tilt ratio calculation exceeds 1.02.

PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 5 of 27 A DROPPED RODS IN MODE 1 ACTIONLEXPECTED RESPONSE RESPONSE NOT OBTAINED A7. Perform the following: I Da. Initiate action to determine cause and repair Rod Control malfunction. Db. with Maintenance's concurrence disconnect lift coil for the dropped rod. DA8. Record the following in the Unit Control Log:

Time of Rod drop.
Dropped Rod number.
Initial power level.
Affected group step counter position.

DA9. Reduce Thermal Power to less than 75% within 2 hours from time of Rod drop. (TS 3.1.4) PROCEDURE NO. REVISION NO. PAGE NO. VEGP 1BOO3-C 23 5 of 27 A DROPPED RODS IN MODE 1 ACTIONLEXPECTED RESPONSE RESPONSE NOT OBTAINED A7. Perform the following: I Da. Initiate action to determine cause and repair Rod Control malfunction. Db. with Maintenance's concurrence disconnect lift coil for the dropped rod. DAB. Record the following in the Unit Control Log:

Time of Rod drop.
Dropped Rod number.
Initial power level.
Affected group step ( counter position.

DA9. Reduce Thermal Power to less than 75% within 2 hours from time of Rod drop. (TS 3.1.4) PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 7 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED All. Prior to initiating Rod retrieval, reduce Thermal Power to the most limiting of the following: D 65% (10% below TS 3.1.4 restriction) -OR-D 10% below most limiting (2. 2:, power distribution p;1J Q fY( 1 k:,_ ,"1 A JAIl _ -. restriction, ._:

AFD QPTR * *A12. Maintain power level during recovery:

D Less than 75%. -OR-D Less than power distribution restrictions:

AFD
QPTR D*A13. Maintain Tavg within 3°F of Tref during recovery.

DA14. position the ROD BANK SELECTOR SWITCH to the affected bank. DAIS. Reset the affected group step counter to zero. DA16. Reconnect dropped Rod lift coil. DA17. Disconnect all lift coils in the affected bank except for the dropped Rod. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 7 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED All. Prior to initiating Rod retrieval, reduce Thermal Power to the most limiting of the following: D D 65% (10% below TS 3.1.4 restriction) -OR-10% below most limiting 2:, power distribution n ff J 11. ;).Af 1 .. Vft'

QPTR j\Vf ,----*A12. Maintain power level during recovery:

D D Less than 75%. -OR-Less than power distribution restrictions:

AFD
QPTR D *A13. Maintain Tavg within 3° F of Tref during recovery.

DA14. Position the ROD BANK SELECTOR SWITCH to the affected bank. DAIS. Reset the affected group step counter to zero. DA16. Reconnect dropped Rod lift coil. DA17. Disconnect all lift coils in the affected bank except for the dropped Rod. PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 8 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED DAI8. Check Rod being realigned -GROUP 1 CONTROL OR SHUTDOWN BANK ROD A19. Initiate 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS: D-Rod Insertion Limit Monitor (if control bank) D-Rod Position Deviation Monitor DA20. Check Unit operation at or above 75% for at least 72 cumulative hours in a 7 day period. DA21. Record the affected bank 's group step counter positions in the Unit Control Log. DAI8. Go to Step A20. DA20. Limit dropped Rod withdrawal to 3 steps per hour. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 8 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED DA18. Check Rod being realigned -GROUP 1 CONTROL OR SHUTDOWN BANK ROD A19. Initiate 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS: D-Rod Insertion Limit Monitor (if control bank) D-Rod Position Deviation Monitor DA20. Check Unit operation at or above 75% for at least 72 cumulative hours in a 7 day period. DA21. Record the affected bank 's group step counter positions in the Unit Control Log. DA18. Go to Step A20. DA20. Limit dropped Rod withdrawal to 3 steps per hour. PROCEDURE NO. REVISION NO. PAGE NO. VEGP I8003-C 23 9 of 27 A DROPPED RODS IN MODE I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:

ALBIO-B06 ROD CONTROL URGENT FAILURE will illuminate when withdrawal of dropped Rod is initiated (unless Shutdown Bank C, D, or E Rod) .
Per IOOOO-C, CONDUCT OF OPERATIONS, the 3 step rod withrawal limitation may be suspended during abnormal conditions.

DA22. Withdraw the Rod in Bank Select to the affected bank's current position. DA23. Record the following in the Unit Control Log:

Recovery completion time.
Affected Bank position.

DA24. Connect the lift coils opened in Step A17. A22. IF the Rod fails to move, THEN: D a. Connect lift coils opened in Step A17. Db. Reset the step counter to value recorded in Step A21. Dc. Continue applicable action items of TS 3.1.4. D d. Reset Rod Control Urgent Failure alarm using IHS-40039 ROD CONTROL ALARM RESET. De. Place ROD BANK SELECTOR SWITCH in MAN. D f . Return to Step AIO. ( ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP IS003-C 23 9 of 2? A DROPPED RODS IN MODE I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:

ALBIO-B06 ROD CONTROL URGENT FAILURE will illuminate when withdrawal of dropped Rod is initiated (unless Shutdown Bank C, D, or E Rod) .
Per IOOOO-C, CONDUCT OF OPERATIONS, the 3 step rod withrawal limitation may be suspended during abnormal conditions.

DA22. Withdraw the Rod in Bank Select to the affected bank's current position. DA23. Record the following in the Unit Control Log:

Recovery completion time.
Affected Bank position.

DA24. Connect the lift coils opened in Step AI? A22. IF the Rod fails to move, THEN: D a. Connect lift coils opened in Step AI? Db. Reset the step counter to value recorded in Step A21. Dc. Continue applicable action items of TS 3.1.4. D d. Reset Rod Control Urgent Failure alarm using IHS-40039 ROD CONTROL ALARM RESET. De. Place ROD BANK SELECTOR SWITCH in MAN. D f . Return to Step AIO. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 10 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE DA25. Reset Rod Control Urgent Failure alarm using IHS-40039 ROD CONTROL ALARM RESET. DA26. Reset the Master Cycler using 13502, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. A27. Check if PiA converter should be reset: D a . Check recovered Rod -CONTROL BANK ROD Db. Reset the PiA converter BANK POSITION DISPLAY to match the position recorded in Step A23 using 13502, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. UNIT 1 CB-B71 UNIT 2 CB-B07 Dc. Discontinue 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS for Rod Insertion Limit Monitor. RESPONSE NOT OBTAINED Da. Go to Step A28. NOTE: If possible, Maintenance should observe Rod exercise to determine required corrective actions. DA28. Exercise the affected bank using 14410, CONTROL ROD OPERABILITY TEST. DA29. Place ROD BANK SELECTOR SWITCH in MAN or AUTO DA28. IF the Rod drops again, THEN return to Step AI. ( position, as desired. ( ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18003-C 23 10 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED DA25. Reset Rod Control Urgent Failure alarm using IHS-40039 ROD CONTROL ALARM RESET. DA26. Reset the Master Cycler using 13502, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. A27. Check if PiA converter should be reset: D a . Check recovered Rod -CONTROL BANK ROD Db. Reset the PiA converter BANK POSITION DISPLAY to match the position recorded in Step A23 using 13502, CONTROL ROD DRIVE AND POSITION INDICATION SYSTEM. UNIT 1 CB-B71 UNIT 2 CB-B07 Dc. Discontinue 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS for Rod Insertion Limit Monitor. Da. Go to Step A28. NOTE: If possible, Maintenance should observe Rod exercise to determine required corrective actions. DA28. Exercise the affected bank using 14410, CONTROL ROD OPERABILITY TEST. DA29. Place ROD BANK SELECTOR SWITCH in MAN or AUTO position, as desired. DA28. IF the Rod drops again, THEN return to Step AI.

8. 028AA2.08 00ll1l2/PZR LVL MALFICIA -3.5INEW/HL15/SRO/DS/TNT Initial conditions:

-Reactor power is currently 8% -Tave is on program for 8% power -PRZR LO LEVEL DEVIATION is alarming -The SS enters AOP 18001-C, section D, Failure of PRZR Leveilnstrul Which of the following contains the correct diagnosis and corrective actions for the indications given? A. The controlling PRZR level channel has failed high. Adjust charging to prevent letdown from flashing or isolate letdown. Return actual PRZR level to 25%. Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel. B. The controlling PRZR level channel has failed low. Adjust charging to prevent letdown from flashing or isolate letdown. Return actual PRZR level to 28%. Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel. C. The controlling PRZR level channel has failed high. Maintain charging flow approxiametely 10 gpm greater than total seal injection flow. Return actual PRZR level to 25%. Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel. D'!" The controlling PRZR level channel has failed low. Maintain charging flow approximately 10 gpm greater than total seal injection flow. Return actual PRZR level to 28%. Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel. 028 Pressurizer (PZR) Level Control Malfunction AA2.08 Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: PZR level as a function of power level Page: 17 of 56 4/2112009 ( 8. 028AA2.08 00ll1l2/PZR LVL MALFICIA -3.5INEW/HL15/SRO/DS/TNT Initial conditions: -Reactor power is currently 8% -Tave is on program for 8% power -PRZR LO LEVEL DEVIATION is alarming -The SS enters AOP 18001-C, section D, Failure of PRZR Level InstrUi Which of the following contains the correct diagnosis and corrective actions for the indications given? A. The controlling PRZR level channel has failed high. Adjust charging to prevent letdown from flashing or isolate letdown. Return actual PRZR level to 25%. Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel. B. The controlling PRZR level channel has failed low. Adjust charging to prevent letdown from flashing or isolate letdown. Return actual PRZR level to 28%. Apply LCO 3.3.1 RTS Instrumentation actions for the failed channel. C. The controlling PRZR level channel has failed high. Maintain charging flow approxiametely 10 gpm greater than total seal injection flow. Return actual PRZR level to 25%. Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel. The controlling PRZR level channel has failed low. Maintain charging flow approximately 10 gpm greater than total seal injection flow. Return actual PRZR level to 28%. Write an INFO LCO 3.3.1 RTS Instrumentation for the failed channel. KIA 028 Pressurizer (PZR) Level Control Malfunction AA2.08 Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: PZR level as a function of power level Page: 17 of 56 4/2112009 KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine correct failure and the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since actions listed are correct for the failure indicated, which is incorrect. B. Incorrect. Plausible since the failure diagnosis is correct and the refernce to the TS actions is required above P7. C. Incorrect. Plausible since the info LCO is a procedural requirement for the present plant condition with power < P7. D. Correct. REFERENCES

1. AOP 18001-C, Section 0, Faillure of PRZR Level Instrumentation.

2. VEGP Technical Specation LCO 3.3.1 Function #9 for high PRZR water level trip. VEGP learning objectives:

1. LO-LP-60301-12:

Given that the pressurizer level control selector switch is in the NORMAL position (459/460), describe how and why the plant will respond to the following instrument failures. Consider each separately and include effects on pressurizer level control, alarms, RPS, and ESF actuations.

a. 459 fails high b. 459 fails low c. 460 fails high d. 460 fails low 2. LO-LP-60301-20:

Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the plantlparameter( s) for the step Page: 18 of 56 4/2112009 ( KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine correct failure and the appropriate mitigative actions to take which are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since actions listed are correct for the failure indicated, which is incorrect. B. Incorrect. Plausible since the failure diagnosis is correct and the refernce to the TS actions is required above P7. C. Incorrect. Plausible since the info LCO is a procedural requirement for the present plant condition with power < P7. D. Correct. REFERENCES

1. AOP 18001-C, Section 0, Faillure of PRZR Level Instrumentation.

2. VEGP Technical Specation LCO 3.3.1 Function #9 for high PRZR water level trip. VEGP learning objectives:

1. LO-LP-60301-12:

Given that the pressurizer level control selector switch is in the NORMAL position (459/460), describe how and why the plant will respond to the following instrument failures. Consider each separately and include effects on pressurizer level control, alarms, RPS, and ESF actuations.

a. 459 fails high b. 459 fails low c. 460 fails high d. 460 fails low 2. LO-LP-60301-20:

Given the entire AOP, describe:

a. Purpose of selected steps b. How and why the step is being performed

c. Expected response of the planUparameter( s) for the step Page: 18 of 56 4/2112009

( 3. LO-LP-39207-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.3 are exceeded.

b. The required actions for all section 3.3 LCOs. Page: 19 of 56 4/2112009 ( ( 3. LO-LP-39207-02:

Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.3 are exceeded.

b. The required actions for all section 3.3 LCOs. Page: 19 of 56 4/2112009

( Approved By J. 8. Stanley Date Approved 10/9/08 SYMPTOMS PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION SECTION D. FAILURE OF PRZR LEVEL INSTRUMENTATION

ALB11-C01 PRZR CONTROL HI LEVEL DEV AND HEATERS ON
ALB11-E01 PRZR HI LEVEL ALARM
ALB11-F01 PRZR HI LEVEL CHANNEL ALERT
ALB11-D01 PRZR LO LEVEL DEVIATION
ALB11-B01 PRZR LO LEVEL HTR CNTL OFF LTDN SECURED Procedure Number Rev 18001-C 31.2 Page Number 3 of 48 Sheet 2 of 3 SECTION E. FAILURE OF STEAM GENERATOR LEVEL INSTRUMENTATION
ALB13-A06 (B06, C06, 006) STM GEN 1 (2,3,4) HIILO LVL DEVIATION
ALB14-A01 (B01, C01, 001) STM GEN 1 (2,3,4) HI-HI LEVEL ALERT
ALB13-A03 (B03, C03, 003) STM GEN 1 (2,3,4) LO LEVEL
ALB13-AOS (BOS, COS, DOS) STM GEN 1 (2,3,4) LO/LO LVL ALERT SECTION F. FAILURE OF STEAM GENERATOR PRESSURE INSTRUMENTATION
ALB14-A02 (B02, C02, 002) STM GEN 1 (2,3,4) HI STM PRESS RATE ALERT
ALB13-F01 HI STM PRESS RATE (ANY LOOP)
ALB13-A04 (B04, C04, 004) STM GEN 1 (2,3,4) LO STEAMLINE PRESS ALERT SECTION G. FAILURE OF STEAM GENERATOR FLOW INSTRUMENTATION
ALB13-A01 (B01, C01, 001) STM GEN 1 (2,3,4) FLOW MISMATCH
Any unexplained steam/feed flow mismatch indication.

SECTION H. FAILURE OF TURBINE IMPULSE PRESSURE INSTRUMENTATION

ALB12-AOS TAVG/TREF DEVIATION
ALB12-A04 RCS LOOP TAVG/AUCT TAVG HI-LO DEV
TSLB-2, S.8 TURB PWR P13 CHI PB-SOSA -NOT LIT
TSLB-2, S.9 TURB PWR P13 CHII PB-S06A -NOT LIT
BPLB 8.3 LO TURB IMP PRESS ROD STOP CS -LIT
BPLB 7.4 LOSS OF TURB LOAD INTLK C7 -LIT
Turbine impulse pressure indication mismatch.
ALBOS-E04 AMSAC TROUBLE
ALB13-E01 AMSACLO FW FL AFW ACTUATION TURBINE TRIP Printed April 6, 2009 at 14:01 Approved By J. B. Stanley Date Approved 10/9/08 SYMPTOMS tl Pla\'t . PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION SECTION D. FAILURE OF PRZR LEVEL INSTRUMENTATION
ALB11-C01 PRZR CONTROL HI LEVEL DEV AND HEATERS ON
ALB11-E01 PRZR HI LEVEL ALARM
ALB11-F01 PRZR HI LEVEL CHANNEL ALERT
ALB11-D01 PRZR LO LEVEL DEVIATION
ALB11-B01 PRZR LO LEVEL HTR CNTL OFF L TDN SECURED Procedure Number Rev 18001-C 31.2 Page Number 3 of 48 Sheet 2 of 3 SECTION E. FAILURE OF STEAM GENERATOR LEVEL INSTRUMENTATION
ALB13-A06 (B06, C06, D06) STM GEN 1 (2,3,4) HIILO LVL DEVIATION
ALB14-A01 (B01, C01, D01) STM GEN 1 (2,3,4) HI-HI LEVEL ALERT
ALB13-A03 (B03, C03, D03) STM GEN 1 (2,3,4) LO LEVEL
ALB13-A05 (B05, C05, D05) STM GEN 1 (2,3,4) LO/LO LVL ALERT SECTION F. FAILURE OF STEAM GENERATOR PRESSURE INSTRUMENTATION
ALB14-A02 (B02, C02, D02) STM GEN 1 (2,3,4) HI STM PRESS RATE ALERT
ALB13-F01 HI STM PRESS RATE (ANY LOOP)
ALB13-A04 (B04, C04, D04) STM GEN 1 (2,3,4) LO STEAMLINE PRESS ALERT SECTION G. FAILURE OF STEAM GENERATOR FLOW INSTRUMENTATION
ALB13-A01 (B01, C01, D01) STM GEN 1 (2,3,4) FLOW MISMATCH
Any unexplained steam/feed flow mismatch indication.

SECTION H. FAILURE OF TURBINE IMPULSE PRESSURE INSTRUMENTATION

ALB12-A05 TAVGITREF DEVIATION
ALB12-A04 RCS LOOP TAVG/AUCT TAVG HI-LO DEV
TSLB-2, 5.8 TURB PWR P13 CHI PB-505A -NOT LIT
TSLB-2, 5.9 TURB PWR P13 CHII PB-506A -NOT LIT
BPLB 8.3 LO TURB IMP PRESS ROD STOP C5 -LIT
BPLB 7.4 LOSS OF TURB LOAD INTLK C7 -LIT
Turbine impulse pressure indication mismatch.
ALB05-E04 AMSAC TROUBLE
ALB13-E01 AMSAC LO FW FL AFW ACTUATION TURBINE TRIP Printed April 6, 2009 at 14:01 Approved8y Rev J. B. Stanley 31.2 Date Approved 10/9/08 PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION Page Number 22 of 48 D. FAILURE OF PRZR LEVEL INSTRUMENTATION ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED D1. Initiate the Continuous Actions Page. *D2. Check PRZR level -TRENDING TO PROGRAM LEVEL. *D3. Maintain Seal Injection flow to all RCPs -8 TO 13 GPM. D4. Select an unaffected channel on LS-459D PRZR LVL CNTL SELECT. D5. Select same channel on LS-459E PRZR L VL REC SEL as selected on LS-459D. Printed April 6, 2009 at 14:03 *D2. IF PRZR level instrument fails high, THEN perform the following as necessary:

Adjust charging to prevent letdown from flashing. -OR-Isolate letdown. _IE PRZR level instrument fails low, THEN maintain charging flow approximately 10 gpm greater than total seal injection flow. ( Date Approved 10/9/08 PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION Page Number 22 of 48 O. FAILURE OF PRZR LEVEL INSTRUMENTATION ACTION/EXPECTED RESPONSE 01. Initiate the Continuous Actions Page. *02. Check PRZR level-TRENDING TO PROGRAM LEVEL. *03. 04. 05. Maintain Seal Injection flow to all RCPs -8 TO 13 GPM. Select an unaffected channel on LS-4590 PRZR LVL CNTL SELECT. Select same channel on LS-459E PRZR L VL REC SEL as selected on LS-4590. Printed April 6, 2009 at 14:03 RESPONSE NOT OBTAINED *02. IF PRZR level instrument fails high, THEN perform the following as necessary: Adjust charging to prevent letdown from flashing. -OR-Isolate letdown. _IE PRZR level instrument fails low, THEN maintain charging flow approximately 10 gpm greater than total seal injection flow. Approved By Procedure Number Rev J. B. Stanley 18001-C 31.2 Date Approved ( 10/9/08 PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION Page Number 24 of 48 D. FAILURE OF PRZR LEVEL INSTRUMENTATION ACTION/EXPECTED RESPONSE D9. Return PRZR level control to AUTO. Maintain PRZR level at program in manual and go to. *D10. Check PRZR level is maintained at program by auto control. D11. Notify I&C to initiate repairs. D12. Bypass the affected instrument channel using 13509-C, BYPASS TEST INSTRUMENTATION (BTl) PANEL OPERATION, if desired. D13. Trip affected channel bistable and place associated MASTER TEST switch in TEST position per TABLE D1 within 72 hours. (TS 3.3.1) D14. Initiate the applicable actions of Technical Specification 3.3.1. *D15. Check repairs and surveillances -COMPLETE. D16. Perform the following:

a. Return tripped bistables to NORMAL position.

RESPONSE NOT OBTAINED *D10. Maintain PRZR level at program using manual control. *D15. Perform the following:

a. WHEN repairs and surveillances are complete, THEN perform step D16. b. Return to procedure and step in effect. o Step 16 continued on next page Printed April 6, 2009 at 14:06 ( Approved By J. B. Stanley Date Approved 10/9/08 09. -*010. 011. 012. 013. 014. *015. Procedure Number Rev 18001-C 31.2 PRIMARY SYSTEMS INSTRUMENTATION MALFUNCTION Page Number 24 of 48 O. FAILURE OF PRZR LEVEL INSTRUMENTATION ACTION/EXPECTED RESPONSE Return PRZR level control to AUTO. Maintain PRZR level at program in manual and go to. Check PRZR level is maintained at program by auto control. Notify I&C to initiate repairs. Bypass the affected instrument channel using 13509-C, BYPASS TEST INSTRUMENTATION (BTl) PANEL OPERATION, if desired. Trip affected channel bistable and place associated MASTER TEST switch in TEST position per TABLE 01 within 72 hours. (TS 3.3.1) Initiate the applicable actions of Technical Specification

3.3.1. Check

repairs and surveillances -COMPLETE. RESPONSE NOT OBTAINED *010. Maintain PRZR level at program using manual control. *015. Perform the following:

a. WHEN repairs and surveillances are complete, THEN perform step 016. b. Return to procedure and step in effect. 016. Perform the following:

a. Return tripped bistables to NORMAL position.

o Step 16 continued on next page Printed April 6, 2009 at 14:06 ( ( FUNCTION 8. Pressurizer Pressure a. Low b. High 9. Pressurizer Water Level -High 10. Reactor Coolant Flow-Low a. Single Loop b. Two Loops APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2 Table 3.3.1-1 (page 3 of 9) Reactor Trip System Instrumentation REQUIRED SURVEILLANCE CHANNELS CONDITIONS REQUIREMENTS 4 4 3 3 per loop 3 per loop M E M N M SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR3.3.1.10 SR3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 (f) Above the P-7 (Low Power Reactor Trips Block) interlock. RTS Instrumentation

3.3.1 ALLOWABLE

VALUE :2: 1950 psig 2395 psig 93.9% :2: 89.4% :2: 89.4% NOMINAL TRIP SETPOINT(n) 1960(g) psig 2385 psig 92% 90% 90% (continued) (g) Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead and 1 second for lag. (h) Above the P-8 (Power Range Neutron Flux) interlock. (i) Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock. (n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions. Vogtle Units 1 and 2 3.3.1-16 Amendment No. 128 (Unit 1) Amendment No. 106 (Unit 2) ( ( ( FUNCTION 8. Pressurizer Pressure a. Low b. High 9. Pressurizer Water Level-High 10. Reactor Coolant Flow-Low a. Single Loop b. Two Loops APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2 Table 3.3.1-1 (page 3 of 9) Reactor Trip System Instrumentation REQUIRED SURVEILLANCE CHANNELS CONDITIONS REQUIREMENTS 4 4 3 3 per loop 3 per loop M E M N M SR 3.3.1.1 SR 3.3.1.7 SR3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 (f) Above the P-7 (Low Power Reactor Trips Block) interlock. RTS Instrumentation

3.3.1 ALLOWABLE

VALUE :::: 1950 psig :0: 2395 psig :0:93.9% :::: 89.4% :::: 89.4% NOMINAL TRIP SETPOINT(n) 1960(g) psig 2385 psig 92% 90% 90% (continued) (g) Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead and 1 second for lag. (h) Above the P-8 (Power Range Neutron Flux) interlock. (i) Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock. (n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions. Vogtle Units 1 and 2 3.3.1-16 Amendment No. 128 (Unit 1) Amendment No. 106 (Unit 2)

9. 069AG2.4.8 OOl/l/2/LOSS CNMT INTEGRITY/C/A

-4.5iNEWIHL15/SRO/DS/TNT Initial conditions: -Crew is implementing FRP 19251-C, Response to High Containment Pressure -1 BA03 normal incoming breaker tripped open -DG-1 B automatically started and its output breaker closed Which of the following is the correct action to take for these conditions: A. Complete FRP 19251-C if it was entered from a red CSFST. Do not initiate AOP 18031-C, AOPs may NOT be performed in parallel with EOPs. B. Suspend FRP 19251-C momentarily while initiating 18031-C. Then complete 19251-C if a red or orange challenge still exists. C'!" Continue FRP 19251-C performance. Initiate AOP 18031-C, AOPs may be performed in parallel with EOPs. ( D. Continue FRP 19251-C performance. Do not initiate AOP 18031-C, AOPs may NOT be performed in parallel with EOPs. Page: 20 of 56 4/2112009

9. 069AG2.4.8 00ll1l2/LOSS CNMT INTEGRITY/CIA

-4.5/NEW/HLl5/SROIDS/TNT Initial conditions: -Crew is implementing FRP 19251-C, Response to High Containment Pressure -1 BA03 normal incoming breaker tripped open -DG-1 B automatically started and its output breaker closed Which of the following is the correct action to take for these conditions: A. Complete FRP 19251-C if it was entered from a red CSFST. Do not initiate AOP 18031-C, AOPs may NOT be performed in parallel with EOPs. B. Suspend FRP 19251-C momentarily while initiating 18031-C. Then complete 19251-C if a red or orange challenge still exists. C'!' Continue FRP 19251-C performance. Initiate AOP 18031-C, AOPs may be performed in parallel with EOPs. ( D. Continue FRP 19251-C performance. Do not initiate AOP 18031-C, AOPs may NOT be performed in parallel with EOPs. Page: 20 of 56 4/2112009 ( ( 069 Loss of Containment Integrity AG2.4.8 Emergency Procedures/Plan Knowledge of how abnormal operating procedures are used in conjunction with EOPs. KIA MATCH ANALYSIS This question requires assessing plant conditions during emergency and abnormal operations and prescribing procedures to mitigate the failures. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. I ncorrect. Plausible because FRP 19251-C is entered from any non-green challenge to the CNMT CSFST. Additionally high challenge FRPs take priority over other EOP's but do not prevent performing an AOP in parallel. B. Incorrect. Plausible since restoring AC power to a train B ESF equipment will also restore one train of required CNMT coolers and Spray pumps. C. Correct. D. Incorrect. Plausible because you continue FRP performance but that does not preclude performing AOPs in parallel with FRPs .. REFERENCES 10020-C, EOP & AOP Rules of Useage, step 3.5.9 VEGP learning objectives: LO-LP-37002-03: State how transitions are made to other procedures. Page: 21 of 56 4/2112009 ( 069 Loss of Containment Integrity AG2.4.8 Emergency Procedures/Plan Knowledge of how abnormal operating procedures are used in conjunction with EOPs. KIA MATCH ANALYSIS This question requires assessing plant conditions during emergency and abnormal operations and prescribing procedures to mitigate the failures. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible because FRP 19251-C is entered from any non-green challenge to the CNMT CSFST. Additionally high challenge FRPs take priority over other EOP's but do not prevent performing an AOP in parallel. B. Incorrect. Plausible since restoring AC power to a train B ESF equipment will also restore one train of required CNMT coolers and Spray pumps. C. Correct. D. Incorrect. Plausible because you continue FRP performance but that does not preclude performing AOPs in parallel with FRPs .. REFERENCES 10020-C, EOP & AOP Rules of Useage, step 3.5.9 VEGP learning objectives: LO-LP-37002-03: State how transitions are made to other procedures. Page: 21 of 56 4/2112009 3.5.8 3.5.8.1 3.5.8.2 3.5.8.3 3.5.9 3.6 3.6.1 3.6.2 3.7 Procedure Number Rev 10020-C 7 Page Number 9 of 26 EOP AND AOP RULES OF USAGE ES-O.O, REDIAGNOSIS, may be entered any time based on operator judgement and may be entered as follows: ES-O.O may be entered when there is doubt in being in correct EOP. Safety injection is in service or is required. E-O, REACTOR TRIP OR SAFETY INJECTION, has been executed and a transition has been made to another EOP which is bounded by the ORGs only (not FRGs). Other procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority. STEP PLACE-KEEPING When exiting an EOP step it is necessary to track what procedure and step was exited such that when directed to "return to procedure step in affect", the correct procedure step may be re-entered. A red ribbon page marker has been provided in the Simulator and Main Control Room EOP sets for assistance in tracking such transitions. Step by step place-keeping is a valuable human performance tool. It shall be performed in accordance with plant standard and management expectations. NOTES AND CAUTIONS All NOTES and CAUTIONS shall be reviewed by the Shift Supervisor. Those NOTES or CAUTIONS that are pertinent to the evolution in progress shall be read aloud to the operating crew. 3.8 MODES OF APPLICABILITY 19000-C E-O 19001-C ES-01 19002-C ES-O.2 19003-C ES-O.3 19004-C ES-O.4 Printed April 6, 2009 at 14: 15 1,2,3 1,2 1,2.3 1,2,3 1,2,3 Assumes RHR system not in service and SI operable Assumes trip from power Assumes No-load conditions Assumes No-load conditions Assumes No-load conditions ( Approved By ,*ti **"it.",M*

s. A. " ",gt Procedure Number Rev 10020-C 7 Date Approved Page Number 90f26 12/8/07 EOP AND AOP RULES OF USAGE 3.5.8 3.5.8.1 3.5.8.2 3.5.8.3 3.5.9 3.6 3.6.1 ES-O.O, REDIAGNOSIS, may be entered any time based on operator judgement and may be entered as follows: ES-O.O may be entered when there is doubt in being in correct EOP. Safety injection is in service or is required.

E-O, REACTOR TRIP OR SAFETY INJECTION, has been executed and a transition has been made to another EOP which is bounded by the ORGs only (not FRGs). Other procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority. STEP PLACE-KEEPING When exiting an EOP step it is necessary to track what procedure and step was exited such that when directed to "return to procedure step in affect", the correct procedure step may be re-entered. A red ribbon page marker has been provided in the Simulator and Main Control Room EOP sets for assistance in tracking such transitions. 3.6.2 Step by step place-keeping is a valuable human performance tool. It shall be performed in accordance with plant standard and management expectations.

3.7 NOTES

AND CAUTIONS All NOTES and CAUTIONS shall be reviewed by the Shift Supervisor. Those NOTES or CAUTIONS that are pertinent to the evolution in progress shall be read aloud to the operating crew. 3.8 MODES OF APPLICABILITY 19000-C E-O 19001-C ES-01 19002-C ES-0.2 19003-C ES-0.3 19004-C ES-OA Printed April 6, 2009 at 14: 15 1,2,3 1,2 1,2.3 1,2,3 1,2,3 Assumes RHR system not in service and SI operable Assumes trip from power Assumes No-load conditions Assumes No-load conditions Assumes No-load conditions

10. WE03EA2.2 00ll1l2/LOCA C/D -DEPRESS/CIA

-4.1!NEW/HL15/SROIDS/TNT Given the following conditions: -RCS LOCA occurred 30 minutes ago while at full power -RCS cold leg temperatures were at 556 F while at power -EOP 19012-C, Post-LOCA Cooldown & Depressurization, is in use -RCS cooldown and depressurization have begun -RCS temperature & pressure currently at 516 degeees F and 1185 psig -One CCP and one SIP have been stopped Which of the following contains the correct actions for the conditions given? A. Stop all Reps due to loss of reactor coolant inventory concerns. Limit the cooldown to 60 degress in next 30 minutes to prevent voiding in the reactor vessel upper head. B. Stop all Reps to limit heat input into the ReS. Limit the cooldown to 100 degrees in next 60 minutes to prevent voiding in the reactor vessel upper head. Stop Reps 1, 2 & 3 to limit Rep heat input into the ReS. Limit the cooldown to 60 degrees in next 30 minutes to prevent challenging reactor vessel integrity. D. Stop Reps 1, 2 & 3 to maintain PRZR normal spray operational. Limit the cooldown to 100 degrees in next 60 minutes to prevent challenging the reactor vessel integrity. Page: 22 of 56 4/2112009

10. WE03EA2.2 00ll1l2/LOCA C/D -DEPRESS/C/A

-4.1INEWIHL15/SROIDS/TNT Given the following conditions: -RCS LOCA occurred 30 minutes ago while at full power -RCS cold leg temperatures were at 556 F while at power -EOP 19012-C, Post-LOCA Cooldown & Depressurization, is in use -RCS cooldown and depressurization have begun -RCS temperature & pressure currently at 516 degeees F and 1185 psig -One CCP and one SIP have been stopped Which of the following contains the correct actions for the conditions given? A. Stop all RCPs due to loss of reactor coolant inventory concerns. Limit the cooldown to 60 degress in next 30 minutes to prevent voiding in the reactor vessel upper head. B. Stop all RCPs to limit heat input into the RCS. Limit the cooldown to 100 degrees in next 60 minutes to prevent voiding in the reactor vessel upper head. ( Stop RCPs 1, 2 & 3 to limit RCP heat input into the RCS. Limit the cooldown to 60 degrees in next 30 minutes to prevent challenging reactor vessel integrity. D. Stop RCPs 1, 2 & 3 to maintain PRZR normal spray operational. Limit the cooldown to 100 degrees in next 60 minutes to prevent challenging the reactor vessel integrity. Page: 22 of 56 4/2112009 WE03 LOCA Cool down and Depressurization EA2.2 Ability to determine and interpret the following as they apply to the LOCA Cooldown and Depressurization: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since cooldown is limited as decribed in this choice to 60 degrees in 30 minutes. Reason for the limit is incorrect but is a concern under natural (' circulation conditions. ( B. Incorrect. Plausible since procedure & TS cooldown rate limit is 100 F/hr. Also RV upper head voiding is a concern under natural circulation. C. Correct. D. Incorrect. Plausible since RCP's 1,2, and 3 are stopped in this procedure and procedure and TS cooldown rate limit is 100 F/hr to limit stresses below allowable in RV. REFERENCES

1. EOP 19012-C, ES-1.2 Post-LOCA Cooldown and Depressurization

2. VEGP Technical Specifications LCO 3.4.3 RCS Pressure and Temperature Limits VEGP learning objectives:

1. LO-LP-37112-01:

Using EOP 19012 as a guide, briefly describe how each step is accomplished. Page: 23 of 56 4/2112009 WE03 LOCA Cooldown and Depressurization EA2.2 Ability to determine and interpret the following as they apply to the LOCA Cooldown and Depressurization: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since cooldown is limited as decribed in this choice to 60 degrees in 30 minutes. Reason for the limit is incorrect but is a concern under natural ( circulation conditions. ( B. Incorrect. Plausible since procedure & TS cooldown rate limit is 100 F/hr. Also RV upper head voiding is a concern under natural circulation. C. Correct. D. Incorrect. Plausible since RCP's 1, 2, and 3 are stopped in this procedure and procedure and TS cooldown rate limit is 100 F/hr to limit stresses below allowable in RV. REFERENCES

1. EOP 19012-C, ES-1.2 Post-LOCA Cooldown and Depressurization

2. VEGP Technical Specifications LCO 3.4.3 RCS Pressure and Temperature Limits VEGP learning objectives:

1. LO-LP-37112-01:

Using EOP 19012 as a guide, briefly describe how each step is accomplished. Page: 23 of 56 4/2112009 Approved By S. A. Phillips Date Approved 8/2/08 ES -1.2 POST -LOCA COOLDOWN AND DEPRESSURIZATION Page Number 11 of 41 ACTION/EXPECTED RESPONSE *12. Initiate ReS cooldown to cold shutdown: -a. Monitor shutdown margin by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

b. Maintain cooldown rate in RCS -cold legs -LESS THAN 100°F/HR.

-c. Use RHR system if in service. d. Dump steam to Condenser from intact SG( s) using Steam Dumps: _1) Place PIC-50? in Manual. _2) Match demand on SG Header Pressure Controller PIC-50? and SO demand meter UI-500. _3) Transfer Steam Dumps to STM PRESS mode. _4) Open available Steam Dumps by slowly raising demand on PIC-50? 13. Check RCS subcooling -GREATER THAN 24°F [38°F ADVERSE]. Printed April 6, 2009 at 14:26 RESPONSE NOT OBTAINED d. Dump steam from intact SG(s) using SG ARV(s). 13. Go to Step 36. ( Approved By S. A. Phillips Date Approved 8/2/08 *12. ES -1.2 POST -LOCA COOLDOWN AND DEPRESSURIZATION Procedure Number Rev 19012-C 31 Page Number 11 of 41 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Initiate ReS cooldown to cold shutdown:

a. Monitor shutdown margin by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

b. Maintain cooldown rate in RCS cold legs -LESS THAN 100°F/HR.

c. -Use RHR system if in service. d. Dump steam to Condenser from intact SG(s) using Steam Dumps: _1) Place PIC-50? in Manual. _2) Match demand on SG Header Pressure Controller PIC-50? and SO demand meter UI-500. _3) Transfer Steam Dumps to STM PRESS mode. _4) Open available Steam Dumps by slowly raising demand on PIC-50? 13. Check RCS subcooling

-GREATER THAN 24°F [38°F ADVERSE].

d. Dump steam from intact SG(s) using SG ARV(s). 13. Go to Step 36.

Date Approved 8/2/08 Number Rev 31 ES -1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION Page Number 16 of 41 ACTION/EXPECTED RESPONSE 5) Stop Auxiliary Spray: a) Open CHARGING TO LOOP ISO VALVE: HV-8146 -OR-HV-8147 b) Close PRZR AUX SPRAY VALVE:

HV-8145 RESPONSE NOT OBTAINED _5) Isolate Auxiliary Spray line. 24. Check if an RCP should be started: a. All RCPs -STOPPED. b. RCS Subcooling

-GREATER THAN 24°F [38°F ADVERSE].

c. PRZR Level-GREATER THAN 19% [50% ADVERSE].

a. Stop all but RCP 4 or RCP 1. Close PRZR Spray Valve(s) for stopped RCP( s): RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 25. b. Close PRZR Spray Valve(s) for stopped RCP( s): RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 36. c. Return to Step 15. o Step 24 continued on next page ( ( Approved By S. A. Phillips Date Approved ES -1.2 POST-LOCA COOLDOWN AND Page Number 8/2/08 24. DEPRESSURIZATION 16 of 41 ACTION/EXPECTED RESPONSE 5) Stop Auxiliary Spray: a) Open CHARGING TO LOOP ISO VALVE: HV-8146 --OR-HV-8147 -b) Close PRZR AUX SPRAY VALVE:

HV-8145 -Check if an RCP should be started: a. All RCPs -STOPPED. _b. RCS Subcooling

-GREATER THAN 24°F [38°F ADVERSE]. ... --=t7 c. PRZR Level-GREATER THAN -19% [50% ADVERSE]. RESPONSE NOT OBTAINED -_5) Isolate Auxiliary Spray line. a. Stop all but RCP 4 or RCP 1. Close PRZR Spray Valve(s) for stopped RCP(s): RCP 1: PIC-0455C -RCP 4: PIC-0455B --Go to Step 25. b. Close PRZR Spray Valve(s) for stopped RCP(s): RCP 1: PIC-0455C -RCP 4: PIC-0455B --Go to Step 36. c. Return to Step 15. o Step 24 continued on next page Printed April 6, 2009 at 14.19 VOGTLE ELECTRIC GENERA TlNG PLANT TRAINING HANDOUT TITLE: Post LOCA Cooldown & Depressurization NUMBER: V-LO-HO-37111-002 PROGRAM: Licensed Operator REVISION: 1 Instructor: D. Cooper DATE: 1/4/07 APPROVED: D. Scukanec DATE: 1/4/07

REFERENCES:

1. WOG EOP HP VERSION, REV 2 2. EOP 19012-C, ES-1.2 Post LOCA Cooldown & Depressurization ( VOGTLE ELECTRIC GENERA TlNG PLANT TRAINING HANDOUT TITLE: Post LOCA Cooldown & Depressurization NUMBER: V-LO-HO-37111-002 PROGRAM: Licensed Operator REVISION:

1 Instructor: D. Cooper DATE: 1/4/07 APPROVED: D. Scukanec DATE: 1/4/07

REFERENCES:

WOG EOP HP VERSION, REV 2 1. 2. EOP 19012-C, ES-1.2 Post LOCA Cooldown & Depressurization INTRODUCTION 19012-C, ES-1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION, provides procedural guidance to cool down and depressurize the RCS to cold shutdown conditions following a loss of reactor coolant inventory. This procedure is structured to deal primarily with small LOCAs for those cases where safety injection flow can keep up with break flow, at pressures above the shutoff head pressure of the RHR pumps. In addition, if a LOCA occurs and the high pressure ECCS system fails, the procedure provides optimal recovery actions to try to prevent an inadequate core cooling condition while trying to restore ECCS flow. If the RWST level decreases to the switchover setpoint, a transition is made from 19012-C to 19013-C, ES 1.3 TRANSFER TO COLD LEG RECIRCULATION. This exit occurs as a result of a foldout page directive. The transition to 19013-C is only temporary. After the switch over is performed, a transition is made from 19013-C to return to the procedure in effect. After reaching and maintaining cold shutdown conditions (RCS temperature less than 200°F) the final step of 19012-C instructs the operators and plant engineering staff to evaluate the long term plant status. At this time, the RCS will be cooled by the RHR System or cold leg recirculation. If PRZR level could not be restored and maintained and boron precipitation is a concern, a decision to transfer to hot leg recirculation (19014-C, ES 1.4 TRANSFER TO HOT LEG RECIRCULATION) could be made. Other long term recovery actions can also be determined at this time. DESCRIPTION For a loss of RCS coolant to containment, Westinghouse plants are designed to use makeup water from the RWST until it is drained. Recirculation from the containment sump to the RCS is then used for long term cooling and makeup. The time to switch over to recirculation depends on the size of the RCS break, the RWST water volume, and whether containment spray is initiated. For some smaller breaks, it is possible to cool down and depressurize the RCS to a cold shutdown condition before the RWST is drained and recirculation becomes necessary. When doing this, it is important to maintain adequate core cooling by maintaining RCS inventory, while also trying to minimize RWST depletion. For any loss of RCS inventory, the RCS pressure will be dependent on the size of the break, the RCS fluid shrink due to cooldown, and the ECCS system capacity. For smaller breaks, the RCS pressure will remain stabilized for a long period of time at high RCS pressures (greater than 400 psig). For a one inch diameter break and RCS subcooling less than 100°F, Figure 2 shows that RCS pressure will stabilize 2 ( INTRODUCTION 19012-C, ES-1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION, provides procedural guidance to cool down and depressurize the RCS to cold shutdown conditions following a loss of reactor coolant inventory. This procedure is structured to deal primarily with small LOCAs for those cases where safety injection flow can keep up with break flow, at pressures above the shutoff head pressure of the RHR pumps. In addition, if a LOCA occurs and the high pressure ECCS system fails, the procedure provides optimal recovery actions to try to prevent an inadequate core cooling condition while trying to restore ECCS flow. If the RWST level decreases to the switchover setpoint, a transition is made from 19012-C to 19013-C, ES 1.3 TRANSFER TO COLD LEG RECIRCULATION. This exit occurs as a result of a foldout page directive. The transition to 19013-C is only temporary. After the switch over is performed, a transition is made from 19013-C to return to the procedure in effect. After reaching and maintaining cold shutdown conditions (RCS temperature less than 200°F) the final step of 19012-C instructs the operators and plant engineering staff to evaluate the long term plant status. At this time, the RCS will be cooled by the RHR System or cold leg recirculation. If PRZR level could not be restored and maintained and boron precipitation is a concern, a decision to transfer to hot leg recirculation (19014-C, ES 1.4 TRANSFER TO HOT LEG RECIRCULATION) could be made. Other long term recovery actions can also be determined at this time. DESCRIPTION For a loss of RCS coolant to containment, Westinghouse plants are designed to use makeup water from the RWST until it is drained. Recirculation from the containment sump to the RCS is then used for long term cooling and makeup. The time to switch over to recirculation depends on the size of the RCS break, the RWST water volume, and whether containment spray is initiated. For some smaller breaks, it is possible to cool down and depressurize the RCS to a cold shutdown condition before the RWST is drained and recirculation becomes necessary. When doing this, it is important to maintain adequate core cooling by maintaining RCS inventory, while also trying to minimize RWST depletion. For any loss of RCS inventory, the RCS pressure will be dependent on the size of the break, the RCS fluid shrink due to cooldown, and the ECCS system capacity. For smaller breaks, the RCS pressure will remain stabilized for a long period of time at high RCS pressures (greater than 400 psig). For a one inch diameter break and RCS subcooling less than 100°F, Figure 2 shows that RCS pressure will stabilize 2 ( above 1300 psig where break flow matches ECCS flow from 2 CCP and 2 SI pumps. If no operator action is taken, transfer to cold leg recirculation may be necessary while RCS pressure remains high. 19012-C provides actions to reduce the RCS temperature and pressure to at or below 200°F and 400 psig, respectively. This is done by establishing a SG cooldown and selectively reducing ECCS flow by stopping SI pumps or establishing normal charging flow if minimum subcooling and PRZR level can be established. From there, the plant staff can determine how to completely depressurize the plant to stop RCS inventory loss and effect repairs. In 19010-C, E 1 LOSS OF REACTOR OR SECONDARY COOLANT, or 19012-C, the RHR pumps are stopped if RCS pressure is higher than their shutoff head pressure (300 psig) and stable or increasing. The pressure versus RCS injection flow curves (without RHR) are shown in Figure 2. The ECCS reduction scheme assumed for the procedure consists of the following discrete head flow curves: 1 ) Two CCPs and two SI pumps 2) One CCP and two SI pumps 3) One CCP and one SI pumps 4) One CCP aligned through the BIT 5) One CCP aligned to full normal charging The ECCS reduction instructions are based on RCS subcooling and PRZR level criteria. Figure 2. RCS INJECTION AND LEAK FLOW CURVES Res Pressure (PSIG) 2400 2200 2000 1800 1600 1400 1200 1000 800 600 400 200 Pump Flow Curves; (1) 2 CHG/SI + 2 HHSI (2) 1 CHG/SI + 2 HHSI (3) 1 CHa/SI + 1 HHSI (4) 1 CHG/SI (5) 1 Normal Charging (Max) Break Flow Plus Shrink Curves: One Inch Diameter Break With 100 OF IHr Cooldown (6) 30 OF Subcooling (7) 100 OF Subcoollng (8) 30 OF Subcoollng (w/o Shrink) OL---__ __ ____ ____ __ ____ __ o 250 500 750 1000 1250 1500 1750 Injection Flow (GPM) 3 ( above 1300 psig where break flow matches ECCS flow from 2 CCP and 2 SI pumps. If no operator action is taken, transfer to cold leg recirculation may be necessary while RCS pressure remains high. 19012-C provides actions to reduce the RCS temperature and pressure to at or below 200°F and 400 psig, respectively. This is done by establishing a SG cooldown and selectively reducing ECCS flow by stopping SI pumps or establishing normal charging flow if minimum subcooling and PRZR level can be established. From there, the plant staff can determine how to completely depressurize the plant to stop RCS inventory loss and effect repairs. In 19010-C, E 1 LOSS OF REACTOR OR SECONDARY COOLANT, or 19012-C, the RHR pumps are stopped if RCS pressure is higher than their shutoff head pressure (300 psig) and stable or increasing. The pressure versus RCS injection flow curves (without RHR) are shown in Figure 2. The ECCS reduction scheme assumed for the procedure consists of the following discrete head flow curves: 1) Two CCPs and two SI pumps 2) One CCP and two SI pumps 3) One CCP and one SI pumps 4) One CCP aligned through the BIT 5) One CCP aligned to full normal charging The ECCS reduction instructions are based on RCS subcooling and PRZR level criteria. Figure 2. RCS INJECTION AND LEAK FLOW CURVES Res Pressure (PSIG) 2400 2200 2000 1800 1600 1400 1200 1000 800 600 400 200 Pump Flow Curves; (1) 2 CHG/SI + 2 HHSI (2) 1 CHG/SI + 2 HHSI (3) 1 CHG/SI + 1 HHSI (4) 1 CHG/SI (5) 1 Norma' Charging (Ma.) Break Flow Plus Shrink Curves: One Inch Diameter Break With 100 OF IHr Cooldown (6) 30 OF Subcooling (7) 100 OF Subcoollng (8) 30 OF Subcoollng (w/o Shrink) OL-____ __ ____ ____ __ ____ __ o 250 500 750 1000 1250 1500 1750 Injection Flow (GPM) 3 MAJOR ACTIONS Prepare For and Initiate RCS Cooldown Certain actions should be performed before initiating an RCS cooldown. Instrument air to containment should be established in order to operate valves needed for the entire recovery procedure. If offsite power is not available, some non safeguards equipment should be loaded on the AC emergency busses. In addition, if RCS pressure is stable and higher than the shutoff head pressure of the RHR pumps, these pumps should be stopped. An explicit check ofSG levels is performed and is contained within the main cooldown loop in 19012-C. This ensures continuous monitoring for possible steam generator tube ruptures. After these actions and checks are performed, a cooldown to cold shutdown (200°F) is initiated. With continued cooldown, subsequent actions can be performed when specified RCS subcooling criteria are satisfied. Depressurize ReS to Refill PRZR This action is performed prior to RCP restart or before/after an ECCS reduction action. As RCS pressure decreases, RCS injection flow will increase relative to break flow. Consequently, this depressurization action should be sufficient for restoring PRZR level if the LOCA is small. A "small" LOCA is first ensured by requiring RCS subcooling before depressurization. If subcooling is lost during the depressurization, it should be restored as the cooldown continues. Prior to restoring PRZR level, all PRZR heaters are turned off. Start One Rep/Stop All But One Rep Once RCS subcooling, PRZR level, and other RCP support conditions are established, an RCP can be started if no RCPs are running. If more than one RCP are running, all but one are stopped to minimize RCS heat input. The RCP restarted (or left running) is used to provide normal PRZR spray and mix the RCS. Reduce ReS Injection Flow As RCS subcooling builds up to specified values, the CCP and SI pumps are stopped one at a time in a predetermined sequence. Subcooling criteria are specified such that a minimum RCS subcooling will be maintained after the injection flow is reduced. 4 ( ( MAJOR ACTIONS Prepare For and Initiate RCS Cooldown Certain actions should be performed before initiating an RCS cooldown. Instrument air to containment should be established in order to operate valves needed for the entire recovery procedure. If offsite power is not available, some non safeguards equipment should be loaded on the AC emergency busses. In addition, if RCS pressure is stable and higher than the shutoff head pressure of the RHR pumps, these pumps should be stopped. An explicit check of SG levels is performed and is contained within the main cooldown loop in 19012-C. This ensures continuous monitoring for possible steam generator tube ruptures. After these actions and checks are performed, a cooldown to cold shutdown (200°F) is initiated. With continued cooldown, subsequent actions can be performed when specified RCS subcooling criteria are satisfied. Depressurize ReS to Refill PRZR This action is performed prior to RCP restart or before/after an ECCS reduction action. As RCS pressure decreases, RCS injection flow will increase relative to break flow. Consequently, this depressurization action should be sufficient for restoring PRZR level if the LOCA is small. A "small" LOCA is first ensured by requiring RCS subcooling before depressurization. If subcooling is lost during the depressurization, it should be restored as the cooldown continues. Prior to restoring PRZR level, all PRZR heaters are turned off. Start One Rep/Stop All But One Rep Once RCS subcooling, PRZR level, and other RCP support conditions are established, an RCP can be started if no RCPs are running. If more than one RCP are running, all but one are stopped to minimize RCS heat input. The RCP restarted (or left running) is used to provide normal PRZR spray and mix the RCS. Reduce ReS Injection Flow As RCS subcooling builds up to specified values, the CCP and SI pumps are stopped one at a time in a predetermined sequence. Subcooling criteria are specified such that a minimum RCS subcooling will be maintained after the injection flow is reduced. 4 ( ( 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PIT) Limits RCS PIT Limits 3.4.3 LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR. APPLICABILITY: At all times. ACTIONS CONDITION A. -----------NOT E ------------- A.1 Required Action A.2 shall be completed whenever this Condition AND is entered. --------------------------------- A.2 Requirements of LCO not met in MODE 1, 2, 3, or 4. B. Required Action and B.1 associated Completion Time of Condition A not AND met. B.2 Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore parameter(s) to 30 minutes within limits. Determine RCS is 72 hours acceptable for continued operation. Be in MODE 3. Be in MODE 5 with RCS pressure < 500 psig. 3.4.3-1 6 hours 36 hours (continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ( 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PIT) Limits RCS PIT Limits 3.4.3 LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR. APPLICABILITY: At all times. ACTIONS CONDITION A. -----------NO T E ------------- A.1 Required Action A.2 shall be completed whenever this Condition AND is entered. --------------------------------- A.2 Requirements of LCO not met in MODE 1, 2, 3, or 4. B. Required Action and B.1 associated Completion Time of Condition A not AND met. B.2 Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore parameter(s) to 30 minutes within limits. Determine RCS is 72 hours acceptable for continued operation. Be in MODE 3. Be in MODE 5 with RCS pressure < 500 psig. 3.4.3-1 6 hours 36 hours (continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) Vogtle Unit 1 -Pressure Temperature Limits Report Southern Nuclear 1.0 RCS Pressure Temperature Limits Report (PTLR) This PTLR for Vogtle Unit 1 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS) Revisions to the PTLR shall be provided to the NRC after issuance. 2.0 RCS Pressure and Temperature (PIT) Limits The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-14040, Revision 4[1] with exception ofWCAP-16142-P, Revision 1 [2] (Elimination of the Flange Requirement). The operability requirements associated with COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6. 2.1 RCS PIT Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60°F. 2.1.2 The RCS temperature rate-of-change limits are: a. A maximum heatup rate of 100°F in any 1-hour period. b. A maximum cooldown rate of 100°F in any 1-hour period. c. A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.3 The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2. 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) The setpoints for the pressurizer Power Operated Relief Valves (PORVs) and arming temperature are presented in the subsections which follow. These setpoints and arming temperature have been developed using the NRC-approved methodology specified in TS 5.6.6. Revision 3 ( Vogtle Unit 1 -Pressure Temperature Limits Report Southern Nuclear 1.0 RCS Pressure Temperature Limits Report (PTLR) This PTLR for Vogtle Unit 1 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS) Revisions to the PTLR shall be provided to the NRC after issuance. 2.0 RCS Pressure and Temperature (PIT) Limits The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-14040, Revision 4[1] with exception ofWCAP-16142-P, Revision 1 [2] (Elimination of the Flange Requirement). The operability requirements associated with COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6. 2.1 RCS PIT Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60°F. 2.1.2 The RCS temperature rate-of-change limits are: a. A maximum heatup rate of 100°F in any 1-hour period. b. A maximum cooldown rate of 100°F in any 1-hour period. c. A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.3 The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2. 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) The setpoints for the pressurizer Power Operated Relief Valves (PORVs) and arming temperature are presented in the subsections which follow. These setpoints and arming temperature have been developed using the NRC-approved methodology specified in TS 5.6.6. 1 Revision 3 ( 11. OOSA2.07 OO1l2111CCW/C/A -2.SINEWIHL15/SRO/DS/TNT Given the following CCW conditions with RCS temperature of 349 F: Train A: -CCW pumps 1 & 3 are running -CCW pump 1 trips -CCW pump 5 had to be manually started Train B: -CCW pump 2 is running with 4 & 6 in PTL -CCW flow to the SFP HX is isolated -CCW system pressure is 70 psig Which of the following choices correctly desribes the condition of the CCW system and appropriate technical specification actions to take? A. Only CCW train A was inoperable during the time only pump 3 was running. LCO 3.7.7, Component Cooling Water System, was not met until the standby CCW pump was manually started. B. Both CCW trains are inoperable. Apply LCO 3.0.3 until RHR HX pressure is adjusted to > 85 psig on train B, then apply LCO 3.7.7, Component Cooling Water System, for CCW train A. Both CCW trains are inoperable. Apply LCO 3.0.3 and restore CCW train B to 2 pump operation then apply LCO 3.7.7, Component Cooling Water System, for CCW train A. D. Both CCW trains are operable. LCO 3.7.7, Component Cooling Water System, is met for both trains. 008 Component Cooling Water System (CCWS) A2.07 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Page: 24 of 56 4/2112009

11. 008A2.07 001l2/1/CCW/C/A

-2.8!NEWIHL15/SRO/DS/TNT Given the following CCW conditions with RCS temperature of 349 F: Train A: -CCW pumps 1 & 3 are running -CCW pump 1 trips -CCW pump 5 had to be manually started Train B: -CCW pump 2 is running with 4 & 6 in PTL -CCW flow to the SFP HX is isolated -CCW system pressure is 70 psig Which of the following choices correctly desribes the condition of the CCW system and appropriate technical specification actions to take? A. Only CCW train A was inoperable during the time only pump 3 was running. LCD 3.7.7, Component Cooling Water System, was not met until the standby CCW pump was manually started. B. Both CCW trains are inoperable. Apply LCD 3.0.3 until RHR HX pressure is adjusted to > 85 psig on train B, then apply LCD 3.7.7, Component Cooling Water System, for CCW train A. C!' Both CCW trains are inoperable. Apply LCD 3.0.3 and restore CCW train B to 2 pump operation then apply LCD 3.7.7, Component Cooling Water System, for CCW train A. D. Both CCW trains are operable. LCD 3.7.7, Component Cooling Water System, is met for both trains. 008 Component Cooling Water System (CCWS) A2.07 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Page: 24 of 56 4/2112009 Consequences of high or low CCW flow rate and temperature; the flow rate at which the CCW standby pump will start KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible because LCO 3.7.7 requires two pumps for the system to be operable. Manual operation does not meet LCO requirements. B. Incorrect. Plausibe because both CCW trains are inoperable and system pressure is below procedure limits. C. Correct. D. Incorrect. Plausible because SOP does allow single pump operation and LCO 3.7.7 ( note states that isolating flow to a component does not make CCW inoperable. '" REFERENCES

1. VEGP Technical Specifications:

LCO 3.7.7 CCW System LCO 3.0.3 2. SOP 13715A Component Cooling Water System Train A 3. SOP 13715B Component Cooling Water System Train B VEGP learning objectives:

1. LO-PP-10101-01:

From memory, state the following for the CCW System: a. Heat Loads b. Where Heat is rejected to c. System configuration for single pump operations

2. LO-PP-1 01 01-04: From memory, describe the expected system response and operator corrective actions Page: 25 of 56 4/2112009 ( Consequences of high or low CCW flow rate and temperature; the flow rate at which the CCW standby pump will start KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible because LCO 3.7.7 requires two pumps for the system to be operable. Manual operation does not meet LCO requirements. B. Incorrect. Plausibe because both CCW trains are inoperable and system pressure is below procedure limits. C. Correct. D. Incorrect. Plausible because SOP does allow single pump operation and LCO 3.7.7 ( note states that isolating flow to a component does not make CCW inoperable. REFERENCES

1. VEGP Technical Specifications:

LCO 3.7.7 CCW System LCO 3.0.3 2. SOP 13715A Component Cooling Water System Train A 3. SOP 13715B Component Cooling Water System Train B VEGP learning objectives:

1. LO-PP-1 01 01-01: From memory, state the following for the CCW System: a. Heat Loads b. Where Heat is rejected to c. System configuration for single pump operations

2. LO-PP-1 01 01-04: From memory, describe the expected system response and operator corrective actions Page: 25 of 56 4/2112009

( for each of the following:

a. SI b. LOSP c. SI with LOSP d. Surge Tank Low level e. Low header pressure f. Pump shaft shear/locked rotor g. Three pumps running 3. LO-PP-1 01 01-06: Given a specific set of conditions and references (Tech Spec's, TRM, ODeM), determine the required actions and completion times. Page: 26 of 56 4/2112009 ( for each of the following:

a. SI b. LOSP c. SI with LOSP d. Surge Tank Low level e. Low header pressure f. Pump shaft shear/locked rotor g. Three pumps running 3. LO-PP-1 01 01-06: Given a specific set of conditions and references (Tech Spec's, TRM, ODeM), determine the required actions and completion times. Page: 26 of 56 4/2112009

( 3.7 PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System LCO 3.7.7 Two CCW trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One CCW train A.1 ------------- NOTE ------------- inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," for residual heat removal loops made inoperable by CCW. ---------------------------------- Restore CCW train to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time of Condition A not AND met. B.2 Be in MODE 5. CCW System 3.7.7 COMPLETION TIME 72 hours 6 hours 36 hours Vogtle Units 1 and 2 3.7.7-1 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ( 3.7 PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System LCO 3.7.7 Two CCW trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One CCW train A.1 ------------- N OT E ------------- inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," for residual heat removal loops made inoperable by CCW. ---------------------------------- Restore CCW train to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time of Condition A not AND met. B.2 Be in MODE 5. CCW System 3.7.7 COMPLETION TIME 72 hours 6 hours 36 hours Vogtle Units 1 and 2 3.7.7-1 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( BASES LCO (continued) APPLICABILITY ACTIONS Vogtle Units 1 and 2 A CCW train is considered OPERABLE when: CCW System B 3.7.7 a. Two pumps and associated surge tank are OPERABLE; and b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE. The isolation of CCW from other components or systems not required for safety may render those components or systems inoperable but does not necessarily make the CCW System inoperable. Consideration should be given to the size of the load isolated and the impact it will have on the rest of the CCW system before determining OPERABILITY. In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger. In Modes 5 or 6, there are no TS OPERABILITY requirements for the CCW System. However, the functional requirements of the CCW System are determined by the systems it supports. Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period. ( continued) B 3.7.7-3 Rev. 1-8/05 ( , BASES LCO (continued) APPLICABILITY ACTIONS Vogtle Units 1 and 2 A CCW train is considered OPERABLE when: CCW System B 3.7.7 a. Two pumps and associated surge tank are OPERABLE; and b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE. The isolation of CCW from other components or systems not required for safety may render those components or systems inoperable but does not necessarily make the CCW System inoperable. Consideration should be given to the size of the load isolated and the impact it will have on the rest of the CCW system before determining OPERABILITY. In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger. In Modes 5 or 6, there are no TS OPERABILITY requirements for the CCW System. However, the functional requirements of the CCW System are determined by the systems it supports. Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period. ( continued) B 3.7.7-3 Rev. 1-8/05 LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.8. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 (continued) Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2) ( LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 Vogtle Units 1 and 2 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.8. Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in: a. MODE 3 within 7 hours; b. MODE 4 within 13 hours; and c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or 3.0-1 ( continued) Amendment No. 141 (Unit 1) Amendment No. 121 (Unit 2) ( 12. 012G2.4.20 001l2/1IRPS/C/A -4.3INEW/HL15/SROIDS/TNT Given the following: -SGTR on SG #1 -RCP's were tripped based on EOP 19000-C foldout page requirements -Crew has transitioned to EOP 19031-C, ES-3.1 Post-SGTR Cooldown Using Backfill -Crew is contemplating starting an RCP Which of the following is the correct action to take? A'I Start RCP 4 since it is preferred in order to provide normal spray control. This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to core reactivity requirements. B. Start RCP 1 since it is preferred in order to provide normal spray control. This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to reactor vessel integrity requirements. C. Continue RCS cooldown on natural circulation. This is to prevent challenging reactor core cladding integrity due to attack from secondary chemicals. D. Start RCP 1, 2, or 3. RCP 4 is saved as a backup in case seals on the running RCP are damaged due to attack from secondary chemicals. Page: 27 of 56 4/2112009 ( 12. 012G2.4.20 001l2/1IRPS/CIA -4.3INEW/HL15/SROIDS/TNT Given the following: -SGTR on SG #1 -RCP's were tripped based on EOP 19000-C foldout page requirements -Crew has transitioned to EOP 19031-C, ES-3.1 Post-SGTR Cooldown Using Backfill -Crew is contemplating starting an RCP Which of the following is the correct action to take? A'I Start RCP 4 since it is preferred in order to provide normal spray control. This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to core reactivity requirements. B. Start RCP 1 since it is preferred in order to provide normal spray control. This RCP aids in mixing contents of the stagnant loop while minimizing a challenge to reactor vessel integrity requirements. C. Continue RCS cooldown on natural circulation. This is to prevent challenging reactor core cladding integrity due to attack from secondary chemicals. D. Start RCP 1, 2, or 3. RCP 4 is saved as a backup in case seals on the running RCP are damaged due to attack from secondary chemicals. Page: 27 of 56 4/2112009 ( 012 Reactor Protection System G2.4.20 Reactor Protection System Knowledge of operational implications of EOP warnings, cautions, and notes. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. This choice is plausible since RCP 1 is a preferred RCP and it will provide the greatest amount of mixing of the stagnant loop. C. Incorrect. This choice is plausible since attack from seconary chemicals is a concern but specific actions other than bypassing the CVCS demins are not provided. D. Incorrect. Plausible since RCP 4 is saved as a backup in the degraded core cooling FRP and the affect of secondary chemicals on the RCS is a concern with this procedure. REFERENCES

1. EOP 19031-C, ES-3.1 Post-SGTR Cooldoown Using Backfill VEGP learning objectives:

LO-LP-37312-02: Using EOP 19031 as a guide, briefly describe how each step is accomplished. Page: 28 of 56 412112009 ( 012 Reactor Protection System G2.4.20 Reactor Protection System Knowledge of operational implications of EOP warnings, cautions, and notes. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. This choice is plausible since RCP 1 is a preferred RCP and it will provide the greatest amount of mixing of the stagnant loop. C. Incorrect. This choice is plausible since attack from seconary chemicals is a concern but specific actions other than bypassing the CVCS demins are not provided. D. Incorrect. Plausible since RCP 4 is saved as a backup in the degraded core cooling FRP and the affect of secondary chemicals on the RCS is a concern with this procedure. REFERENCES

1. EOP 19031-C, ES-3.1 Post-SGTR Cooldoown Using Backfill VEGP learning objectives:

LO-LP-37312-02: Using EOP 19031 as a guide, briefly describe how each step is accomplished. Page: 28 of 56 4/2112009 Approved By J.B. Stanley Rev 17 Date Approved 7/22/2008 ES -3.1 POST -SGTR COOLDOWN USING BACKFILL Page Number 30f22 ACTION/EXPECTED RESPONSE 1. Initiate Continuous Actions and Foldout Pages. RESPONSE NOT OBTAINED CAUTION Inadvertent criticality may occur following any natural circulation cooldown if the first RCP started is in the ruptured loop. *2. Check RCP status: a. RCPs -ALL STOPPED. b. Check RVLlS full range indication -GREATER THAN 98%. a. Go to Step 3. b. Perform the following:

Raise PRZR level greater than 90% [90% ADVERSE].
Raise RCS Subcooling based on core exit TCs greater than 60°F [74°F ADVERSE].
Use PRZR Heaters, as necessary to saturate the Pressurizer water. o Step 2 continued on next page 2009 at 14:36 Approved By J.B. Stanley .

It Date Approved ES -3.1 POST -SGTR COOLDOWN USING Page Number ( 7/22/2008 BACKFILL 30f22 ( ACTION/EXPECTED RESPONSE 1. Initiate Continuous Actions and Foldout Pages. RESPONSE NOT OBTAINED CAUTION Inadvertent criticality may occur following any natural circulation cooldown if the first RCP started is in the ruptured loop. *2. Check RCP status: a. RCPs -ALL STOPPED. b. Check RVLlS full range indication -GREATER THAN 98%. a. Go to Step 3. b. Perform the following:

*
Raise PRZR level greater than 90% [90% ADVERSE].

Raise RCS Subcooling based on core exit TCs greater than 60°F [74°F ADVERSE]. Use PRZR Heaters, as necessary to saturate the Pressurizer water. o Step 2 continued on next page Printed April 6, 2009 at 14:36 Approved By J.B. Stanley Rev 17 Date Approved 7/22/2008 ES -3.1 POST -SGTR COOLDOWN USING BACKFILL Page Number 4 of 21 ACTION/EXPECTED RESPONSE c. Start an RCP using ATTACHMENT B. (RCP 4 or RCP 1 preferred)

3. Check PRZR Spray status: a. RCP 4 or RCP 1 -RUNNING. b. Shut PRZR Spray Valve(s) for stopped RCP(s): RCP 1: PIC-0455C RCP 4: PIC-0455B

c. Open RCP breakers for RCP( s) NOT running. Printed April 20, 2009 at 18:43 RESPONSE NOT OBTAINED c. IE an RCP can NOT be started, THEN verify natural circulation using ATTACHMENT C. IE natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps. After natural circulation is verified, maintain rate of dumping steam. a. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal PRZR Spray using ATTACHMENT B. ( Approved By J.B. Stanley Date Approved 7/22/2008 Procedure Number 19031-C 17 ES -3.1 POST -SGTR COOLDOWN USING BACKFILL Page Number 4 of 21 ACTION/EXPECTED RESPONSE c. Start an RCP using ATTACHMENT B. (RCP 4 or RCP 1 preferred)

RESPONSE NOT OBTAINED c. IE an RCP can NOT be started, THEN verify natural circulation using ATTACHMENT C. IE natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps. After natural circulation is verified, maintain rate of dumping steam. 3. Check PRZR Spray status: a. RCP 4 or RCP 1 -RUNNING. b. Shut PRZR Spray Valve(s) for stopped RCP(s): RCP 1: PIC-0455C RCP 4: PIC-0455B

c. Open RCP breakers for RCP(s) NOT running. a. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal PRZR Spray using ATTACHMENT B.

13. 013A2.04 OO1l2/11ESFAS/C/A

-4.2INEW/HL15/SROIDS/TNT Given the following: -Unit is in mode 1 at 8% power -18Y28 and 1 DY1 8 de-energize due to bus faults The correct actions to take for this situation are: Enter AOP 18032-1, Loss of 120V AC Instrument Power. Enter LCO 3.8.9 Distribution Systems -Operating. 80TH 1 8Y28 and 1 DY1 8 must be re-energized within 2 hours, otherwise the unit needs to be shutdown.

8. Complete EOP 19000-C, Reactor Trip or Safety Injection, then go to AOP 18032-1, Loss of 120V AC Instrument Power. Enter LCO 3.8.9 Distribution Systems -Operating.

EITHER 1 8Y28 or 1 DY1 8 must be re-energized within the next 2 hours from its associated inverter. C. Enter EOP 19000-C, Reactor Trip or Safety Injection and AOP 18032-1, Loss of 120V AC Instrument Power, concurrently. Enter LCO 3.0.3 for this condition due to a loss of safety a function. EITHER 1 8Y28 or 1 DY1 8 must be re-energized within the next 2 hours from its associated inverter. D. Enter AOP 18032-1, Loss of 120V AC Instrument Power Enter LCO 3.0.3 for this condition due to a loss of a safety function. 80TH 18Y28 and 1 DY1 8 must be re-energized within 2 hours, otherwise the unit needs to be shutdown. 013 Engineered Safety Features Actuation System (ESFAS) A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument bus KIA MATCH ANALYSIS Page: 29 of 56 4/2112009 ( ( 13. 013A2.04 00 112/1 IESFAS/C/A -4.2INEW/HLl5/SRO/DS/TNT Given the following: Unit is in mode 1 at 8% power -1 BY2B and 1 DY1 B de-energize due to bus faults The correct actions to take for this situation are: A'I Enter AOP 18032-1, Loss of 120V AC Instrument Power. Enter LCO 3.8.9 Distribution Systems -Operating. BOTH 1 BY2B and 1 DY1 B must be re-energized within 2 hours, otherwise the unit needs to be shutdown. B. Complete EOP 19000-C, Reactor Trip or Safety Injection, then go to AOP 18032-1, Loss of 120V AC Instrument Power. Enter LCO 3.8.9 Distribution Systems -Operating. EITHER 1 BY2B or 1 DY1 B must be re-energized within the next 2 hours from its associated inverter. C. Enter EOP 19000-C, Reactor Trip or Safety Injection and AOP 18032-1, Loss of 120V AC Instrument Power, concurrently. Enter LCO 3.0.3 for this condition due to a loss of safety a function. EITHER 1 BY2B or 1 DY1 B must be re-energized within the next 2 hours from its associated inverter. D. Enter AOP 18032-1, Loss of 120V AC Instrument Power Enter LCO 3.0.3 for this condition due to a loss of a safety function. BOTH 1 BY2B and 1 DY1 B must be re-energized within 2 hours, otherwise the unit needs to be shutdown. 013 Engineered Safety Features Actuation System (ESFAS) A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument bus KIA MATCH ANALYSIS Page: 29 of 56 4/2112009 This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible due to misreading bus nomenclature as far as plant response. Also plausible due to LCO 3.8.9 action requirement time of 2 hours. C. Incorrect. Plausible since applicable AOP's are entered concurrently with EOP's and LCO 3.8.9 action is required within 2 hours and existence of an action for loss of safety function. D. Incorrect. Plausible since AOP 18032-1 is the expected procedure entry and LCO 3.8.9 action is required within 2 hours and existence of an action for loss of safety function. REFERENCES

1. AOP 18032-1, Loss of 120V AC Instrument Power 2. VEGP Technical Specifications LCO 3.8.9 Distribution Systems -Operating VEGP learning objectives:

LO-LP-60324-01

Given the appropriate plant drawings, logics, and/or procedures, describe how the plant will respond to a loss of the following 120VAC instrument panels: a.1AY1A b.1AY2A c.1BY1B d. 1 BY2B e.1CY1A f.1DY1B LO-LP-60324-04:

State the effect on SSPS and Sequencer operation on loss of the following:

a. 1AY1A b. 1AY2A c. 1BY1B Page: 30 of 56 4/2112009 ( ( This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible due to misreading bus nomenclature as far as plant response. Also plausible due to LCO 3.8.9 action requirement time of 2 hours. C. Incorrect. Plausible since applicable AOP's are entered concurrently with EOP's and LCO 3.8.9 action is required within 2 hours and existence of an action for loss of safety function. D. Incorrect. Plausible since AOP 18032-1 is the expected procedure entry and LCO 3.8.9 action is required within 2 hours and existence of an action for loss of safety function. REFERENCES

1. AOP 18032-1, Loss of 120V AC Instrument Power 2. VEGP Technical Specifications LCO 3.8.9 Distribution Systems -Operating VEGP learning objectives:

LO-LP-60324-01

Given the appropriate plant drawings, logics, and/or procedures, describe how the plant will respond to a loss of the following 120VAC instrument panels: a.1AY1A b.1AY2A c.1BY1B d.1BY2B e.1CY1A f. 1 DY1 B LO-LP-60324-04:

State the effect on SSPS and Sequencer operation on loss of the following:

a. 1AY1A b. 1AY2A c. 1BY1B Page: 30 of 56 4/2112009

d. 1BY2B e. 1CY1A f. 1DY1B LO-LP-39212-02:

Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.8 are exceeded.

b. The required actions for all section 3.8 LCOs. Page: 31 of 56 4/2112009 ( ( d. 1BY2B e. 1CY1A f. 1 DY1 B LO-LP-39212-02:

Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.8 are exceeded.

b. The required actions for all section 3.8 LCOs. Page: 31 of 56 4/2112009 PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18032-1 25.1 30 of 98 C. LOSS OF VITAL INSTRUMENT PANEL 1BY1B (CB-B47) ACTION/EXPECTED RESPONSE IMMEDIATE OPERATOR ACTIONS D C1. Check reactor power -

C1. Perform the following: GREATER THAN P-10 SET POINT D C2. Verify ROD BANK SELECTOR SWITCH in manual. C3. Contro1 SG NR 1eve1s -BETWEEN 60% AND 70%: D -MFRVs in manual. D-MFPT SPEED CONTROL MASTER in manual. SUBSEQUENT OPERATOR ACTIONS D C4. Initiate the Continuous Actions Page.

C5. Contro1 charqinq to: D-Maintain seal injection flow to all RCPs -8 TO 13 GPM D-IF letdown isolated, THEN adjust charging flow to approximately 10 gpm greater than total seal injection flow. DC6. Select CH459/461 on 1LS-459D PRZR LVL CNTL SELECT. D C7. Select L-459 on 1LS-459E PRZR LVL REC SEL. D a. Verify reactor trip. Db. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. ( PROCEDURE NO. VEGP 18032-1 REVISION NO. PAGE NO. 25.1 30 of 98 C. LOSS OF VITAL INSTRUMENT PANEL 1BY1B (CB-B47) ACTION/EXPECTED RESPONSE IMMEDIATE OPERATOR ACTIONS D C1. Check reactor power -

C1. Perform the following: GREATER THAN P-10 SET POINT DC2. Verify ROD BANK SELECTOR SWITCH in manual. C3. Contro1 SG NR 1eve1s -BETWEEN 60% AND 70%: D-MFRVs in manual. D-MFPT SPEED CONTROL MASTER in manual. DC4. SUBSEQUENT OPERATOR ACTIONS Initiate the Continuous Actions Page.

C5. Contro1 charging to: D-Maintain seal injection flow to all RCPs -8 TO 13 GPM D-IF letdown isolated, THEN adjust charging flow to approximately 10 gpm greater than total seal injection flow. DC6. Select CH459/461 on 1LS-459D PRZR LVL CNTL SELECT. D C7. Select L-459 on 1LS-459E PRZR LVL REC SEL. D a. Verify reactor trip. Db. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18032-1 25.1 44 of 98 D. LOSS OF VITAL INSTRUMENT PANEL IBY2B (AB-116) DOL ACTION/EXPECTED RESPONSE Initiate the Continuous Actions Page. RESPONSE NOT OBTAINED NOTES -Train B ESF sequencer will not operate following loss of Panel IBY2B. -Loss of power to Panel IBY2B will result in a Containment Ventilation Isolation. 002. Select CH459/461 on lLS-459D PRZR LVL CNTL SELECT. 003. Check letdown -IN SERVICE 0* 04. Maintain PRZR l.evel. within 5% of program val.ue. 05. Notify Chemistry that the following radiation monitors will be out of service and will need to be reset when power is restored: 0-lRE-0003 (CVI) 0-lRE-0006 0-lRE-2533A (FHBI) 0-lRE-2533B (FHBI) 0-lRE-12117 (CRI) 0-lRE-13121 0-lRE-13122

03. Perform the following:

o a. Restore letdown by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. Db. IF letdown can NOT be placed in service, THEN place excess letdown in service by initiating 13008, CHEMICAL AND VOLUME CONTROL SYSTEM EXCESS LETDOWN. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18032-1 25.1 44 of 98 D. LOSS OF VITAL INSTRUMENT PANEL 1BY2B (AB-116) ACTION/EXPECTED RESPONSE 001. Initiate the Continuous Actions Page. RESPONSE NOT OBTAINED NOTES -Train B ESF sequencer will not operate following loss of Panel 1BY2B. -Loss of power to Panel 1BY2B will result in a Containment Ventilation Isolation. 002. Select CH459/461 on 1LS-459D PRZR LVL CNTL SELECT. 003. Check letdown -IN SERVICE 0* 04. Maintain PRZR l.evel. within 5% of program val.ue. 05. Notify Chemistry that the following radiation monitors will be out of service and will need to be reset when power is restored: 0-1RE-0003 (CVI) 0-1RE-0006 0-1RE-2533A (FHBI) 0-1RE-2533B (FHBI) 0-1RE-12117 (CRI) 0-1RE-13121 0-1RE-13122

03. Perform the following:

o a. Restore letdown by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM. Ob. IF letdown can NOT be placed in service, THEN place excess letdown in service by initiating 13008, CHEMICAL AND VOLUME CONTROL SYSTEM EXCESS LETDOWN. ( PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18032-1 25.1 61 of 98 F. LOSS OF VITAL INSTRUMENT PANEL 1DY1B (CB-B48) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

Letdown isolation (l-HV-15214 will close) and steam generator blowdown isolation will occur on Pipe Break Room Protection due to loss of temperature bistables in QPP4.
The Train B chiller will be inoperable due to loss of flow switch 1FY-1803, but may be started from the Train B Shutdown Panel. The A Train chiller should be operated if Essential Chilled Water is required.
The Train B NSCW bypass and return valves will NOT close on low header pressure.

OF1. Initiate the Continuous Actions Page. ( , PROCEDURE NO. REVISION NO. PAGE NO. VEGP 18032-1 25.1 61 of 98 F. LOSS OF VITAL INSTRUMENT PANEL 1DY1B (CB-B48) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

Letdown isolation (l-HV-15214 will close) and steam generator blowdown isolation will occur on Pipe Break Room Protection due to loss of temperature bistables ln QPP4.
The Train B chiller will be inoperable due to loss of flow switch 1FY-1803, but may be started from the Train B Shutdown Panel. The A Train chiller should be operated if Essential Chilled Water is required.
The Train B NSCW bypass and return valves will NOT close on low header pressure.

OF1. Initiate the Continuous Actions Page. ( ( Distribution Systems -Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems -Operating LCO 3.8.9 The required AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

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

The redundant emergency buses of 4160 V switchgear 1/2AA02 and 1/2BA03 may be manually connected within the unit by tie breakers in order to allow transfer of preferred offsite power sources provided SR 3.8.1.1 is successfully performed within 12 hours prior to the interconnection. The interconnection shall be implemented without adversely impacting the ability to simultaneously sequence both trains of LOCA loads. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION A. One or more AC A.1 electrical power distribution subsystems inoperable. B. One or more AC vital bus B.1 electrical power distribution subsystems inoperable. Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore AC electrical 8 hours power distribution subsystems to AND OPERABLE status. 16 hours from discovery of failure to meet LCO Restore AC vital bus 2 hours electrical power distribution subsystems to AND OPERABLE status. 3.8.9-1 16 hours from discovery of failure to meet LCO ( continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ( Distribution Systems -Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems -Operating LCO 3.8.9 The required AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

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

The redundant emergency buses of 4160 V switchgear 1/2AA02 and 1/2BA03 may be manually connected within the unit by tie breakers in order to allow transfer of preferred offsite power sources provided SR 3.8.1.1 is successfully performed within 12 hours prior to the interconnection. The interconnection shall be implemented without adversely impacting the ability to simultaneously sequence both trains of LOCA loads. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION A. One or more AC A.1 electrical power distribution subsystems inoperable. B. One or more AC vital bus B.1 electrical power distribution subsystems inoperable. Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore AC electrical 8 hours power distribution subsystems to AND OPERABLE status. 16 hours from discovery of failure to meet LCO Restore AC vital bus 2 hours electrical power distribution subsystems to AND OPERABLE status. 3.8.9-1 16 hours from discovery of failure to meet LCO (continued) Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ACTIONS (continued) CONDITION C. One or more DC C.1 electrical power distribution subsystems inoperable. D. Required Action and D.1 associated Completion Time not met. AND D.2 E. Two or more electrical E.1 power distribution subsystems inoperable that result in a loss of function. SURVEILLANCE REQUIREMENTS Distribution Systems -Operating

3.8.9 REQUIRED

ACTION COMPLETION TIME Restore DC electrical 2 hours power distribution subsystems to AND OPERABLE status. 16 hours from discovery of failure to meet LCO Be in MODE 3. 6 hours Be in MODE 5. 36 hours Enter LCO 3.0.3. Immediately SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to 7 days required AC, DC, and AC vital bus electrical power distribution subsystems. Vogtle Units 1 and 2 3.8.9-2 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) ( ACTIONS (continued) CONDITION C. One or more DC C.1 electrical power distribution subsystems inoperable. D. Required Action and 0.1 associated Completion Time not met. AND 0.2 E. Two or more electrical E.1 power distribution subsystems inoperable that result in a loss of function. SURVEILLANCE REQUIREMENTS Distribution Systems -Operating

3.8.9 REQUIRED

ACTION COMPLETION TIME Restore DC electrical 2 hours power distribution subsystems to AND OPERABLE status. 16 hours from discovery of failure to meet LCO Be in MODE 3. 6 hours Be in MODE 5. 36 hours Enter LCO 3.0.3. Immediately SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to 7 days required AC, DC, and AC vital bus electrical power distribution subsystems. Vogtle Units 1 and 2 3.8.9-2 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2)

14. 059G2.4.3 OO1l2/11MAIN FEEDWATERIC/A

-3.9INEW/HL15/SROIDS/TNT Initial conditions: -Small RCS LOCA inside CNMT is in progress -CNMT pressure has risen to 4.1 psig -EOP 19231-C, FR-H.1 Response to Loss of Secondary Heat Sink, is in progress -RCS bleed & feed is in progress -Main feedwater has been restored to SG#1 Current conditions: -Only two SG #1 NR level indications are available -One indication has a red bezel and the other a black bezel -The curent SG #1 NR levels are: LT-517 (red bezel) -15% L T-551 (black bezel) -33% Which of the following are the correct actions to take? A'I Remain in 19231-C; Wait for L T-517 to go above 32% prior to terminating RCS bleed & feed. B. Remain in 19231-C; Terminate RCS bleed & feed,SG level is adequate to continue C. Return to procedure and step in effect; Wait for L T-517 to go above 32% prior to terminating RCS bleed & feed. D. Return to procedure and step in effect; Terminate RCS bleed & feed SG level is adequate to continue. KIA 059 Main Feedwater (MFW) System G2.4.3 Emergency Procedures/Plan Ability to identify post-accident instrumentation. KIA MATCH ANALYSIS This question requires applicant to interpet SG level instrumentation, and take the Page: 32 of 56 4/2112009

14. 059G2.4.3 OO1l2111MAIN FEEDWATERICIA

-3.9INEW/HL15/SROIDS/TNT Initial conditions: -Small RCS LOCA inside CNMT is in progress -CNMT pressure has risen to 4.1 psig -EOP 19231-C, FR-H.1 Response to Loss of Secondary Heat Sink, is in progress -RCS bleed & feed is in progress -Main feedwater has been restored to SG#1 Current conditions: -Only two SG #1 NR level indications are available -One indication has a red bezel and the other a black bezel -The curent SG #1 NR levels are: L T -517 (red bezel) -15% L T-551 (black bezel) -33% Which of the following are the correct actions to take? A'I Remain in 19231-C; Wait for L T-517 to go above 32% prior to terminating RCS bleed & feed. ( B. Remain in 19231-C; Terminate RCS bleed & feed, SG level is adequate to continue C. Return to procedure and step in effect; Wait for L T-517 to go above 32% prior to terminating RCS bleed & feed. D. Return to procedure and step in effect; Terminate RCS bleed & feed SG level is adequate to continue. KIA 059 Main Feedwater (MFW) System G2.4.3 Emergency Procedures/Plan Ability to identify post-accident instrumentation. ( KIA MATCH ANALYSIS This question requires applicant to interpet SG level instrumentation, and take the Page: 32 of 56 4/2112009 correct procedural actions based on the PAMS SG level instrument. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since the correct action includes remaining within 19231-C and one of the two level indications is above the minimum required level for proceeding in 19231-C. C. Incorrect. Plausible, since return to procedure and step in effect is the correct answer to this question / procedure if RCS bleed & feed had not been initiated. Minimum required level on PAMS instrumentation in an adverse post accident condition should also be used. D. Incorrect. Plausible since these actions are correct for this logic in the procedure if bleed & feed had not been initiated. REFERENCES VEGP Technical Specifications LCO 3.3.3 & Bases 3.3.3 (PAMS) LO-LP-37003, Post-Accident Monitoring System EOP 19231-C, FR-H.1 Response to Loss of Secondary Heat Sink VEGP learning objectives: LO-LP-37003-01

State the purpose of the Post-Accident Monitoring and the Post-Accident Radiation Monitoring Systems. LO-LP-37003-02:

Identify post-accident instrumentation on the MCB. LO-LP-37051-08: Using EOP 19231 as a guide, briefly describe how each major step is accomplished. Describe the bases for each. Page: 33 of 56 4/2112009 ( correct procedural actions based on the PAMS SG level instrument. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible since the correct action includes remaining within 19231-C and one of the two level indications is above the minimum required level for proceeding in 19231-C. C. Incorrect. Plausible, since return to procedure and step in effect is the correct answer to this question / procedure if RCS bleed & feed had not been initiated. Minimum required level on PAMS instrumentation in an adverse post accident condition should also be used. D. Incorrect. Plausible since these actions are correct for this logic in the procedure if bleed & feed had not been initiated. REFERENCES VEGP Technical Specifications LCO 3.3.3 & Bases 3.3.3 (PAMS) LO-LP-37003, Post-Accident Monitoring System EOP 19231-C, FR-H.1 Response to Loss of Secondary Heat Sink VEGP learning objectives: LO-LP-37003-01

State the purpose of the Post-Accident Monitoring and the Post-Accident Radiation Monitoring Systems. LO-LP-37003-02:

Identify post-accident instrumentation on the MCB. LO-LP-37051-08: Using EOP 19231 as a guide, briefly describe how each major step is accomplished. Describe the bases for each. Page: 33 of 56 4/2112009 III. LESSON OUTLINE: I. INTRODUCTION A. From the Kemeny Report: "Equipment should be reviewed from the pOint of view of providing information to operators to help them prevent accidents and to cope with accidents when they occur ... Computer technology should be used for the clear display for operators and shift supervisors of key measurements relevant to accident conditions, together with diagnostic warnings of conditions ... " B. During and after the TMI accident, the industry found that: 1. Operators did not have enough information in the Control Room 2. The information they did have was not designed for adverse conditions and thus was not reliable 3. Some instrumentation that we take for granted is the result of TM I modifications

a. Reactor Vessel Liquid Inventory System b. Subcooling monitors c. Upgraded readouts of core exit thermocouples

1) TMl's thermocouples pegged at 700 0 F, so the operators thought they had failed high 2) Vogtle's calibrated for -200 to 2300 0 F C. VEGP PAMS Systems 1. PAMS instrumentation listed in Tech Spec 3.3.3 2. PSMS 3. RVLlS 4. SPDS a. CSFST and b. Foldout Page are part of SPDS 5. Radiation monitors 4 LO-LP-37003-10-C NOTES III. ( I. ( LESSON OUTLINE: INTRODUCTION A. From the Kemeny Report: "Equipment should be reviewed from the point of view of providing information to operators to help them prevent accidents and to cope with accidents when they occur ... Computer technology should be used for the clear display for operators and shift supervisors of key measurements relevant to accident conditions, together with diagnostic warnings of conditions

... " B. During and after the TMI accident, the industry found that: 1. Operators did not have enough information in the Control Room 2. The information they did have was not designed for adverse conditions and thus was not reliable 3. Some instrumentation that we take for granted is the result of TMI modifications

a. Reactor Vessel Liquid Inventory System b. Subcooling monitors c. Upgraded readouts of core exit thermocouples

1) TMl's thermocouples pegged at 700° F, so the operators thought they had failed high 2) Vogtle's calibrated for -200 to 2300° F C. VEGP PAMS Systems 1. PAMS instrumentation listed in Tech Spec 3.3.3 2. PSMS 3. RVLlS 4. SPDS a. CSFST and b. Foldout Page are part of SPDS 5. Radiation monitors 4 LO-LP-37003-10-C NOTES III. LESSON OUTLINE: D. Present objectives II. PRESENTATION A. Purpose 1. Provide operators with information that will help them prevent or mitigate an accident 2. Provide plant personnel with information that will help them assess and monitor release of radioactive materials B. Type and Category Classification

1. Equipment classified by its intended use and by its qualifications

2. 3. Equipment listed in FSAR, Chapter 7.5 Types are A through E and categories are 1 -3 a. Does Table 7.5.3-1 look familiar?

1) Tech Spec 3.3.3 b. How is this instrumentation identified on the MCB? 1) Red bezel c. Should be familiar with this list and which modes these inst are required 4. Types a. EOPs -Type A b. CSF -Type B c. E-PLAN (EPIP) -Type C d. ECCS and Other Plant Status -Type D e. Radiation Monitoring

-Type E 1) 5. Categories Where can plant vent air flow rate be monitored? a) Comm console and/or IPC a. Classified as category 1 , 2, or 3 5 LO-LP-37003-10-C NOTES LO-TP-37003-001 Objective 1 LO-TP-37003-002 Refer students to their handout Objective 2 Objective 3 Ref Table 7.5.3-5 ( III. LESSON OUTLINE: D. Present objectives II. PRESENTATION A. Purpose 1. Provide operators with information that will help them prevent or mitigate an accident 2. Provide plant personnel with information that will help them assess and monitor release of radioactive materials B. Type and Category Classification

1. Equipment classified by its intended use and by its qualifications

2. 3. Equipment listed in FSAR, Chapter 7.5 Types are A through E and categories are 1 -3 a. Does Table 7.5.3-1 look familiar?

1) Tech Spec 3.3.3 b. How is this instrumentation identified on the MCB? 1) Red bezel c. Should be familiar with this list and which modes these inst are required 4. Types a. EOPs -Type A b. CSF -Type B c. E-PLAN (EPIP) -Type C d. ECCS and Other Plant Status -Type D e. Radiation Monitoring

-Type E 1) 5. Categories Where can plant vent air flow rate be monitored? a) Comm console and/or IPC a. Classified as category 1, 2, or 3 5 LO-LP-37003-10-C NOTES LO-TP-37003-001 Objective 1 LO-TP-37003-002 Refer students to their handout Objective 2 Objective 3 Ref Table 7.5.3-5 ( Approved By Number Rev D. R. Vineyard 31 Date Approved FR-H.1 RESPONSE TO LOSS OF SECONDARY Page Number 7/30/08 HEAT SINK 38 of 53 ACTION/EXPECTED RESPONSE 67. Open Main Feed Pump discharge valves. 68. Open BFIV for selected SG. 69. Slowly open BFRV for the selected SG to establish feed flow. 70. Check for adequate secondary heat sink: a. NR level in at least one SG -GREATER THAN 10% [32% ADVERSE]. RESPONSE NOT OBTAINED 67. lE discharge valves can NOT be opened, THEN locally open MFP bypass valve 1305-U4-655. (TB-Lvl 2) 68. Open MFIV on selected SG. _IF MFIV will NOT open, THEN dispatch an operator to locally open the selected BFIV. _lE neither BFIV or MFIV will open, THEN return to Step 49. 69. Open MFRV for the selected SG. _IF MFRV will NOT open, THEN dispatch an operator to locally open the BFRV. _IF neither BFRV or MFRV will open, THEN return to Step 49. a. lE feed flow to at least one SG verified, THEN do NOT continue until N R level is restored to greater than 10% [32% ADVERSE]. ( Approved By Procedure Number Rev D. R. Vineyard 19231-C 31 Date Approved 7/30/08 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page Number 38 of 53 67. 68. 69. ACTION/EXPECTED RESPONSE Open Main Feed Pump discharge valves. Open BFIV for selected SG. Slowly open BFRV for the selected SG to establish feed flow. 70. Check for adequate secondary heat sink: a. NR level in at least one SG -GREATER THAN 10% [32% ADVERSE]. Printed April 20, 2009 at 18.51 67. -68. -69. -RESPONSE NOT OBTAINED lE discharge valves can NOT be opened, THEN locally open MFP bypass valve 1305-U4-655. (TB-L vi 2) Open MFIV on selected SG. _IF MFIV will NOT open, THEN dispatch an operator to locally open the selected BFIV. _lE neither BFIV or MFIV will open, THEN return to Step 49. Open MFRV for the selected SG. _IF MFRV will NOT open, THEN dispatch an operator to locally open the BFRV. _lE neither BFRV or MFRV will open, THEN return to Step 49. a. lE feed flow to at least one SG verified, THEN do NOT continue until NR level is restored to greater than 10% [32% ADVERSE]. Date Approved FR-H.1 RESPONSE TO LOSS OF SECONDARY Page Number 7/30/08 HEAT SINK 39 of 53 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 71. Check RCS temperatures:

71. Return to Step 49.

Core exit TCs -LOWERING.
RCS WR hot leg temperatures-LOWERING.

72. Verify Reactor Head Vent Valves -CLOSED:

HV-8095A -RX HEAD VENT TO LETDOWN ISOLATION VLV
HV-8095B -RX HEAD VENT TO LETDOWN ISOLATION VL V
HV-8096A -RX HEAD VENT TO LETDOWN ISOLATION VLV
HV-8096B -RX HEAD VENT TO LETDOWN ISOLATION VL V
HV-0442A -REACTOR HEAD VENT TO PRT
HV-0442B -REACTOR HEAD VENT TO PRT ( Approved By , Procedure Number Rev D. R. Vineyard ,: 19231-C 31 Date Approved 7/30/08 FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page Number 39 of 53 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 71. Check RCS temperatures:

71. Return to Step 49.

Core exit TCs -LOWERING.

-* -RCS WR hot leg temperatures -LOWERING.

72. Verify Reactor Head Vent Valves -CLOSED:

HV-8095A -RX HEAD VENT TO -LETDOWN ISOLATION VLV
HV-8095B -RX HEAD VENT TO -LETDOWN ISOLATION VL V
HV-8096A -RX HEAD VENT TO -LETDOWN ISOLATION VLV
HV-8096B -RX HEAD VENT TO -LETDOWN ISOLATION VL V
HV-0442A -REACTOR HEAD -VENT TO PRT
HV-0442B -REACTOR HEAD -VENT TO PRT Printed Apnl 20, 2009 at 18.51

( !-U][)'"VH' By D. R. Vineyard Date Approved FR-H.1 RESPONSE TO LOSS OF SECONDARY Page Number 7/30/08 HEAT SINK 40 of 53 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE The following step will prevent an unwanted SI actuation from occurring when securing RCS bleed and feed. 73. Check SG pressures -GREATER THAN 585 PSIG. NOTE 73. Bypass the SG LOW PRESS inputs for 2 channels of any depressurized SG by initiating 13509-C, BYPASS TEST INSTRUMENTATION (BTl) PANEL OPERATION: _SG1 :PB514A, PB515A, PB516A _SG2:PB524A, PB525A, PB526A _ SG3:PB534A, PB535A, PB536A _ SG4: PB544A, PB545A, PB546A After closing a PRZR PORV, it may be necessary to wait for RCS pressure to rise before determining if ECCS flow can be terminated.

74. Check if ECCS flow can be terminated:

a. RCS subcooling

-GREATER THAN 24°F [38°F ADVERSE].

b. Check RVLlS full range indication

-GREATER THAN 62%. c. Go to Step 76. at 18:51 a. Go to Step 75. b. Go to Step 75. ( ( Approved By Procedure Number Rev D. R. Vineyard 19231-C 31 Date Approved FR-H.1 RESPONSE TO LOSS OF SECONDARY Page Number 7/30/08 HEAT SINK 40 of 53 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE The following step will prevent an unwanted SI actuation from occurring when securing RCS bleed and feed. 73. Check SG pressures -GREATER THAN 585 PSIG. NOTE 73. Bypass the SG LOW PRESS inputs for 2 channels of any depressurized SG by initiating 13509-C, BYPASS TEST INSTRUMENTATION (BTl) PANEL OPERATION: _SG1:PB514A, PB515A, PB516A _SG2:PB524A, PB525A, PB526A _ SG3:PB534A, PB535A, PB536A _ SG4: PB544A, PB545A, PB546A After closing a PRZR PORV, it may be necessary to wait for RCS pressure to rise before determining if ECCS flow can be terminated.

74. Check if ECCS flow can be terminated:

a. RCS subcooling

-GREATER THAN 24°F [38°F ADVERSE].

b. Check RVLlS full range indication

-GREATER THAN 62%. c. Go to Step 76. a. Go to Step 75. b. Go to Step 75. Number Rev 31 Date Approved FR-H.1 RESPONSE TO LOSS OF SECONDARY Page Number 7/30/08 HEAT SINK 41 of 53 ACTION/EXPECTED RESPONSE 75. Check RCS bleed path status: a. PRZR PORVs and associated block valves -ANY BLEED PATH OPEN. b. Close ONE PRZR PORV: _1) Block one train of COPS. _2) Place associated PRZR PORV in AUTO. _3) Verify proper operation of PORV. _c. Return to Step 74. 76. Stop ECCS Pumps:

SI Pumps
All but one CCP 77. Check RCS bleed path status: a. PRZR PORVs and associated block valves -ANY BLEED PATH OPEN. RESPONSE NOT OBTAINED a. Go t019010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT. b. Close its associated block valve. _ IF block valve can NOT be closed, THEN go to 19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT. a. Go to Step 78. o Step 77 continued on next page Approved By Procedure D. R. Vineyard 19231-C 31 Date Approved Page Number 7/30/08 41 of 53 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 75. Check RCS bleed path status: a. PRZR PORVs and associated

a. Go t01901 O-C, E-1 LOSS block valves -ANY BLEED PATH OF REACTOR OR OPEN. SECONDARY COOLANT. b. Close ONE PRZR PORV: b. Close its associated block valve. _1) Block one train of COPS. IF block valve can NOT be closed, _2) Place associated PRZR THEN go to 19010-C, E-1 PORV in AUTO. LOSS OF REACTOR OR SECONDARY COOLANT. _3) Verify proper operation of PORV. ( c. Return to Step 74. 76. Stop ECCS Pumps:

SI Pumps
All but one CCP 77. Check RCS bleed path status: a. PRZR PORVs and associated

a. Go to Step 78. block valves -ANY BLEED PATH OPEN. o Step 77 continued on next page Printed April 20, 2009 at 1

15. 061A2.03 OO1l2/11AUX FEEDWATERIC/A

-3.4INEWIHL15/SRO/DS/TNT Initial conditions: -Unit 1 is at 57% power -All systems are in their normal alignment for this power level Current conditions: -1 CD1 M de-energizes due to a bus fault -HS-5106 TDAFWP Steam Supply Valve indicating lights are dark Which of the following are the correct actions to take? Apply LCO 3.7.5, Auxiliary Feedwater System, ..... A. condition B to declare the TDAFW TRAIN inoperable. TDAFWP governor valve is shut. B. condition A to declare the TDAFW STEAM SUPPLY inoperable. TDAFWP governor valve is open. C!' condition B to declare the TDAFW TRAIN inoperable. TDAFWP governor valve is open. D. condition A to declare the TDAFW STEAM SUPPLY inoperable. TDAFWP governor valve is shut. 061 Auxiliary I Emergency Feedwater (AFW) System A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of DC power KIA MATCH ANALYSIS This question requires applicant to interpet the components impacted by the loss of DC power failure, and take the correct actions based on the VEGP Technical specifications. Page: 34 of 56 4/2112009 ( 15. 061A2.03 OO1l2111AUX FEEDWATERICIA -3.4INEWIHL15/SRO/DS/TNT Initial conditions: -Unit 1 is at 57% power -All systems are in their normal alignment for this power level Current conditions: -1 CD1 M de-energizes due to a bus fault -HS-5106 TDAFWP Steam Supply Valve indicating lights are dark Which of the following are the correct actions to take? Apply LCO 3.7.5, Auxiliary Feedwater System, ..... A. condition B to declare the TDAFW TRAIN inoperable. TDAFWP governor valve is shut. B. condition A to declare the TDAFW STEAM SUPPLY inoperable. TDAFWP governor valve is open. C'!'" condition B to declare the TDAFW TRAIN inoperable. TDAFWP governor valve is open. D. condition A to declare the TDAFW STEAM SUPPLY inoperable. TDAFWP governor valve is shut. KIA 061 Auxiliary I Emergency Feedwater (AFW) System A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of DC power KIA MATCH ANALYSIS This question requires applicant to interpet the components impacted by the loss of DC power failure, and take the correct actions based on the VEGP Technical specifications. Page: 34 of 56 4/2112009 Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the correct TS action is listed. Student must know the loss of power failure position of the governor. B. Incorrect. Plausible since one of the impacted valves is a steam supply valve. The governor valve is in the position stated. C. Correct. D. Incorrect. Plausible since one of the impacted valves is a steam supply valve. The student must also know the failure position for the governor valve on a loss of power. REFERENCES

1. AOP 18034-1, Loss of Class 1 E 125V DC Power 2. Technical specification LCO 3.7.5 and bases VEGP learning objectives:

LO-LP-39211-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.7 are exceeded.

b. The required actions for all section 3.7 LCOs. LO-PP-20101-09:

Determine the impact to AFW system operation and the overall integrated plant operations to the following types of power supply failures:

a. UN condition on eitherAA02 or BA03 with the bus being re-energized from the EDG while at 100% power. b. UN condition on either AA02 or BA03 with the bus remaining de-energized while at 100% power. c. Loss of a 120 VAC 1 E vital instrument bus. Page: 35 of 56 4/2112009 ( Question meets 1 OCFR55 .43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the correct TS action is listed. Student must know the loss of power failure position of the governor. B. Incorrect. Plausible since one of the impacted valves is a steam supply valve. The governor valve is in the position stated. C. Correct. D. Incorrect. Plausible since one of the impacted valves is a steam supply valve. The student must also know the failure position for the governor valve on a loss of power. REFERENCES

1. AOP 18034-1, Loss of Class 1 E 125V DC Power 2. Technical specification LCO 3.7.5 and bases VEGP learning objectives:

LO-LP-39211-02: Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode: a. Whether any Tech Spec LCOs of section 3.7 are exceeded.

b. The required actions for all section 3.7 LCOs. LO-PP-20101-09:

Determine the impact to AFW system operation and the overall integrated plant operations to the following types of power supply failures:

a. UN condition on either AA02 or BA03 with the bus being re-energized from the EDG while at 100% power. b. UN condition on either AA02 or BA03 with the bus remaining de-energized while at 100% power. c. Loss of a 120 VAC 1 E vital instrument bus. Page: 35 of 56 4/2112009

d. Loss of a 125 VDC 1 E bus e. Loss of All AC Power Page: 36 of 56 4/2112009

d. Loss of a 125 VDC 1E bus e. Loss of All AC Power ( Page: 36 of 56 4121/2009

3.7 PLANT

SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3. ACTIONS AFW System 3.7.5 --------------------------------------------------------------- NOT E -------------------------------------------------------- LCO 3.0.4b is not applicable. CONDITION A. One steam supply to A.1 turbine driven AFW pump inoperable. B. One AFW train B.1 inoperable for reasons other than Condition A. Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore steam supply to 7 days OPERABLE status. Restore AFW train to OPERABLE status. 3.7.5-1 AND 10 days from discovery of failure to meet the LCO 72 hours AND 10 days from discovery of failure to meet the LCO ( continued) Amendment No. 142 (Unit 1) Amendment No. 122 (Unit 2) ( 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3. ACTIONS AFW System 3.7.5 --------------------------------------------------------------- NOT E -------------------------------------------------------- LCO 3.0.4b is not applicable. CONDITION A. One steam supply to A.1 turbine driven AFW pump inoperable. B. One AFW train B.1 inoperable for reasons other than Condition A. Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Restore steam supply to 7 days OPERABLE status. Restore AFW train to OPERABLE status. 3.7.5-1 AND 10 days from discovery of failure to meet the LCO 72 hours AND 10 days from discovery of failure to meet the LCO ( continued) Amendment No. 142 (Unit 1) Amendment No. 122 (Unit 2) BASES APPLICABILITY (continued) ACTIONS Vogtle Units 1 and 2 AFW System B 3.7.5 In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required. A Note prohibits the application of LCO 3.0.4b to an inoperable AFW train. There is an increased risk associated with an AFW train inoperable and the provisions of LCO 3.0.4b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance. If one of the two steam supplies to the turbine driven AFW train is inoperable, action must be taken to restore OPERABLE status within 7 days. The 7 day Completion Time is reasonable, based on the following reasons: a. The redundant OPERABLE steam supply to the turbine driven AFW pump; b. The availability of redundant OPERABLE motor driven AFW pumps; and c. The low probability of an event occurring that requires the inoperable steam supply to the turbine driven AFW pump. The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of Conditions to be inoperable during any continuous failure to meet this LCO. The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The AND connector between 7 days and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met. With one of the required AFW trains (pump or flow path) inoperable for reasons other than Condition A, action must be taken to restore OPERABLE status within 72 hours. This (continued) B 3.7.5-5 Rev. 4-9/05 ( ( BASES APPLICABILITY (continued) ACTIONS Vogtle Units 1 and 2 AFW System B 3.7.5 In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required. A Note prohibits the application of LCO 3.0.4b to an inoperable AFW train. There is an increased risk associated with an AFW train inoperable and the provisions of LCO 3.0.4b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance. If one of the two steam supplies to the turbine driven AFW train is inoperable, action must be taken to restore OPERABLE status within 7 days. The 7 day Completion Time is reasonable, based on the following reasons: a. The redundant OPERABLE steam supply to the turbine driven AFW pump; b. The availability of redundant OPERABLE motor driven AFW pumps; and c. The low probability of an event occurring that requires the inoperable steam supply to the turbine driven AFW pump. The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of Conditions to be inoperable during any continuous failure to meet this LCO. The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The AND connector between 7 days and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met. B.1 With one of the required AFW trains (pump or flow path) inoperable for reasons other than Condition A, action must be taken to restore OPERABLE status within 72 hours. This (continued) B 3.7.5-5 Rev. 4-9/05 Procedure Number Rev 18034-1 10 Date Approved 72 of 82 Page Number LOSS OF CLASS 1 E 125V DC POWER 8/2/08 ATTACHMENT C1 Sheet 1 of 3 LOSS OF 125V DC TO BUS 1 CD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN C 125V DC POWER

Power to Inverter 1 CD113 is lost causing 120V AC Vital Bus 1 CY1 A to de-energize.
TDAFW Pump Mechanical Trip and Throttle Valve 1-PV-15129 will fail as is with no control capability.
TDAFW Pump Speed Governor Valve 1-SV-15133 will fail full open.
Power to Isolation Panel 1 CCQIP3 is lost rendering the annunciators in Train C inoperable.
RHR PMP A UPSTREAM SUCTION FROM HOT LEG 1-HV-8701 B.
Loss of RCP 3 Class 1 E breaker control including RCP underfrequency trip and undervoltage input to SSPS.
Turbine Driven AFW PMP SUCTION HDR 1-HV-5113.
Turbine Driven AFW PMP STEAM ADMISSION VALVE 1-HV-51 06.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5127.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5125.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5122.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5120.

Printed April 6, 2009 at 15:15 Approved By S. A. Phillips Date Approved Page Number 8/2/08 LOSS OF CLASS 1 E 125V DC POWER 72 of 82 ( ATTACHMENT C1 Sheet 1 of 3 LOSS OF 125V DC TO BUS 1 CD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN C 125V DC POWER

Power to Inverter 1 CD113 is lost causing 120V AC Vital Bus 1 CY1 A to de-energize.
TDAFW Pump Mechanical Trip and Throttle Valve 1-PV-15129 will fail as is with no control capability.
TDAFW Pump Speed Governor Valve 1-SV-15133 will fail full open.
Power to Isolation Panel 1 CCQIP3 is lost rendering the annunciators in Train C inoperable.
RHR PMP A UPSTREAM SUCTION FROM HOT LEG 1-HV-8701B.
Loss of RCP 3 Class 1 E breaker control including RCP underfrequency trip and undervoltage input to SSPS.
Turbine Driven AFW PMP SUCTION HDR 1-HV-5113.
Turbine Driven AFW PMP STEAM ADMISSION VALVE 1-HV-5106.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5127.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5125.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5122.
Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5120.

Printed April 6, 2009 at 15:15

16. 055G2.4.11 001l2/2/COND AIR REMOV ALICIA -4.2INEWIHL15/SROIDS/TNT Given the following conditions:

-TURB CNDSR LO VAC annunciator is alarming -The unit is at full power -Both mechanical vacuum pumps have been started -The standby SJAE has been placed in service -Main condenser vacuum is now 24.0" Hg and lowering (getting worse) Which of the following correctly describes the actions that should be taken? A. Lower load to 50% using AOP 18013-C, Rapid Power Reduction. Rate is limited to 5%/minute with manual rod insertions. Trip reactor if the ROD BANK LO-LO LIMIT alarm comes in or if Tave drops < 551 degrees F. B. Lower load using UOP 12004-C, Power Operation, until the TURB CNDSR LO VAC alarm clears. Rate is limited to 5%/minute with automatic rod control. Borate as necessary to keep rods> RIL. Lower load using AOP 18013-C, Rapid Power Reduction, until turbine condenser vacuum is stable or rising. Rate is limited to 5%/minute with automatic rod control. Trip reactor if Tave goes below 551 degrees F. Borate as necessary to keep ROD BANK LO-LO LIMIT alarm clear. D. Lower load to 50% using UOP 12004-C, Power Operation. Rate is limited to that necessary to keep vacuum> turbine trip setpoint of 22.42" Hg. Trip reactor if AFD not maintained within Tech Spec limits. 055 Condenser Air Removal System (CARS) G2.4.11 Emergency Procedures/Plan Knowledge of abnormal condition procedures. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are Page: 37 of 56 4/2112009 ( 16. 055G2.4.11 001l2/2/COND AIR REMOVAL/CIA -4.2INEWIHL15/SROIDS/TNT Given the following conditions: -TURB CNDSR LO VAC annunciator is alarming -The unit is at full power -Both mechanical vacuum pumps have been started -The standby SJAE has been placed in service -Main condenser vacuum is now 24.0" Hg and lowering (getting worse) Which of the following correctly describes the actions that should be taken? A. Lower load to 50% using AOP 18013-C, Rapid Power Reduction. Rate is limited to 5%/minute with manual rod insertions. Trip reactor if the ROD BANK LO-LO LIMIT alarm comes in or if Tave drops < 551 degrees F. B. Lower load using UOP 12004-C, Power Operation, until the TURB CNDSR LO VAC alarm clears. Rate is limited to 5%/minute with automatic rod control. Borate as necessary to keep rods> RIL. Lower load using AOP 18013-C, Rapid Power Reduction, until turbine condenser vacuum is stable or rising. Rate is limited to 5%/minute with automatic rod control. Trip reactor if Tave goes below 551 degrees F. Borate as necessary to keep ROD BANK LO-LO LIMIT alarm clear. D. Lower load to 50% using UOP 12004-C, Power Operation. Rate is limited to that necessary to keep vacuum> turbine trip setpoint of 22.42" Hg. Trip reactor if AFD not maintained within Tech Spec limits. 055 Condenser Air Removal System (CARS) G2.4.11 Emergency Procedures/Plan Knowledge of abnormal condition procedures. KIA MATCH ANALYSIS This question requires correct diagnosis of plant indications to determine the appropriate mitigative actions to take to remain within license limits. These actions are Page: 37 of 56 4121/2009 not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since load reduction is performed with AOP 18013-C and 50% is a target within the AOP. Also plausible because rate limit, manual rods, and tripping with T ave < 551 F are all correct. B. Incorrect. Plausible since actual goal is to lower load until the low vacuum alarm clears. Additionally the rate is correct as are the actions with the boration. C. Correct. D. Incorrect. Plausible since 50% limit is a typical target for rapid power reductions for other equipment pbroblems. Additionally, the rate limit is also plausible because the goal is to clear the low vacuum alarm. REFERENCES

1. ARP 17019-1 window B04 for turbine condenser 10 vacuum 2. AOP 18013-C, Rapid Power Reduction

3. UOP 12004-C, Power Operation VEGP learning objectives:

LO-LP-60331-03: Discuss the types of plant conditions that might warrant the use of the "Rapid Power Reduction" AOP, 18013-C. Discuss also the plant conditions where this procedure should not be utilized. LO-LP-60331-04: Discuss the procedural guidance that provides the desired limits on the rate of load reduction during a rapid power reduction. Page: 38 of 56 4/2112009 not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since load reduction is performed with AOP 18013-C and 50% is a target within the AOP. Also plausible because rate limit, manual rods, and tripping with lave < 551 F are all correct. B. Incorrect. Plausible since actual goal is to lower load until the low vacuum alarm clears. Additionally the rate is correct as are the actions with the boration. C. Correct. D. Incorrect. Plausible since 50% limit is a typical target for rapid power reductions for other equipment pbroblems. Additionally, the rate limit is also plausible because the goal is to clear the low vacuum alarm. REFERENCES

1. ARP 17019-1 window B04 for turbine condenser 10 vacuum ( 2. AOP 18013-C, Rapid Power Reduction

3. UOP 12004-C, Power Operation VEGP learning objectives:

LO-LP-60331-03: Discuss the types of plant conditions that might warrant the use of the "Rapid Power Reduction" AOP, 18013-C. Discuss also the plant conditions where this procedure should not be utilized. LO-LP-60331-04: Discuss the procedural guidance that provides the desired limits on the rate of load reduction during a rapid power reduction. Page: 38 of 56 4/2112009 Approved By Procedure Number Rev S. A. Phillips 17019-1 23 Date Approved 12/16/08 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 19 ON PANEL 1 B2 Page Number ORIGIN 1-PISH-6292A and B 1-PISH-6293A and B 1-PISH-6294A and B (1 out 2 inst. in anyone exhaust hood) ON MCB 20 of 43 SETPOINT 24.92" Hg. Vacuum Lowering WINDOW B04 TURB CNDSR LOVAC 1.0 PROBABLE CAUSE 1. Condenser Air Ejector or Gland Seal Steam System fault. 2. Excessive air leakage into the condenser.

3. Insufficient circulating water flow for the existing turbine load. 4. Loss of power to instruments listed above. 2.0 AUTOMATIC ACTIONS NOTE If condenser vacuum continues to lower, a turbine trip will occur at 22.42" Hg. Vacuum lowering (2 out of 2 inst. in anyone exhaust hood). 1. Mechanical Vacuum Pumps will auto start when condenser vacuum is less than 25in Hg. (Vacuum lowering) if in AUTO PULL TO LOCK. 2. Steam Dump operation is blocked when condenser vacuum is less than 24.92" Hg. (Vacuum lowering), (2 out of 2 inst. in anyone exhaust hood), (Permissive C-9 lost or blocked).

Printed April 8, 2009 at 10:14 ( Approved By Procedure Number Rev S. A. Phillips 17019-1 23 Date Approved 12/16/08 ORIGIN 1-PISH-6292A and B 1-PISH-6293A and B 1-PISH-6294A and B (1 out 2 inst. in anyone exhaust hood) SETPOINT 24.92" Hg. Vacuum Lowering Page Number WINDOW B04 TURB CNDSR LOVAC 20 of 43 1.0 PROBABLE CAUSE 2.0 Printed 1. Condenser Air Ejector or Gland Seal Steam System fault. 2. Excessive air leakage into the condenser.

3. Insufficient circulating water flow for the existing turbine load. 4. Loss of power to instruments listed above. AUTOMATIC ACTIONS NOTE If condenser vacuum continues to lower, a turbine trip will occur at 22.42" Hg. Vacuum lowering (2 out of 2 inst. in anyone exhaust hood). 1. Mechanical Vacuum Pumps will auto start when condenser vacuum is less than 25in Hg. (Vacuum lowering) if in AUTO PULL TO LOCK. 2. Steam Dump operation is blocked when condenser vacuum is less than 24.92" Hg. (Vacuum lowering), (2 out of 2 inst. in anyone exhaust hood), (Permissive C-9 lost or blocked).

at 10:14 Date Approved 3/24/09 PURPOSE RAPID POWER REDUCTION ABNORMAL OPERATING PROCEDURE CONTINUOUS USE Procedure Number Rev 18013-C 7 Page Number 1 of 11 This procedure provides instructions when plant conditions require a rapid load reduction or plant shutdown in a controlled manner in the judgment of the SS. Entry 17015-D05 17015-E01 17019-B04 18025-C 18009-C 18009-C 18039-C Condition MFPT High Vibrations Condenser Low Vacuum or Circ Water Pump Trip or Loss of Utility Water SG Tube Leak gpd with an ROC gpd/hr) SG Tube Leak gpm) Confirmed Loose Part SS determination based on plant conditions MAJOR ACTIONS

Perform Pre-job Brief.
Perform rapid power reduction.

Target <70% RTP Vacuum >22.42" Hg and STABLE or RISING <50% RTP within 1 hour 20% RTP within 1 hour & trip reactor 20% RTP quickly As determined by the SS Approx. Time @ 3-5%/min 5-8 minutes 10-17 minutes 16-27 minutes 16-27 minutes ( Approved By J. B. Stanley Date Approved 3/24/09 PURPOSE RAPID POWER REDUCTION ABNORMAL OPERATING PROCEDURE CONTINUOUS USE Number Rev 7 Page Number 1 of 11 This procedure provides instructions when plant conditions require a rapid load reduction or plant shutdown in a controlled manner in the judgment of the SS. Entry 17015-D05 17015-E01 17019-B04 18025-C 18009-C 18009-C 18039-C Condition MFPT High Vibrations Condenser Low Vacuum or Circ Water Pump Trip or Loss of Utility Water SG Tube Leak gpd with an ROC gpd/hr) SG Tube Leak gpm) Confirmed Loose Part SS determination based on plant conditions MAJOR ACTIONS

Perform Pre-job Brief.
Perform rapid power reduction.

Target <70% RTP Vacuum >22.42" Hg and STABLE or RISING <50% RTP within 1 hour 20% RTP within 1 hour & trip reactor 20% RTP quickly As determined by the SS Approx. Time @ 3-5%/min 5-8 minutes 10-17 minutes 16-27 minutes 16-27 minutes ( Approved By Procedure Number Rev J. B. Stanley 18013-C 7 Date Approved Page Number RAPID POWER REDUCTION 3/24/09 3 of 11 SHUTDOWN BRIEFING METHOD Auto rod control should be used. Reduce Turbine Load at approximately 3% RTP per minute (approx 36 MW e) up to 5% RTP (approx 60 MWe)' Borate considering the calculations from the reactivity briefing sheet and BEACON. Maintain AFD within the doghouse. SS (or SRO designee) -Maintain supervisory oversight. All rod withdrawals will be approved by the SS. Approval for each reactivity manipulation is not necessary as long as manipulations are made within the boundaries established in this briefing (i.e. turbine load adjustment up to 60 MW e , etc.). A crew update should be performed at approximately every 100 MWe power change. If manpower is available, peer checks should be used for all reactivity changes. OPERATIONAL LIMITS

Maintain T AVG within +/-6°F of TREF' If T AVG/TREF mismatch >6°F and not trending toward a matched condition or if T AVG :S;551 of, then trip the reactor.
If load reduction due to a loss of vacuum, every effort should be made to maintain the steam dumps closed (Permissive C-9 224.92" Hg). INDUSTRY OE
Shift supervision must maintain effective oversight and exercise conservative decision making.
Correction of significant RCS T AVG deviations should only be via secondary plant control manipulations and not primary plant control manipulations (i.e., do not withdraw control rods or dilute). 11 ( Approved By Procedure Number Rev J. B. Stanley " 18013-C 7 Date Approved RAPID POWER REDUCTION Page Number 3/24/09 3 of 11 SHUTDOWN BRIEFING METHOD Auto rod control should be used. Reduce Turbine Load at approximately 3% RTP per minute (approx 36 MW e) up to 5% RTP (approx 60 MWe)' Borate considering the calculations from the reactivity briefing sheet and BEACON. Maintain AFD within the doghouse.

SS (or SRO designee) -Maintain supervisory oversight. All rod withdrawals will be approved by the SS. Approval for each reactivity manipulation is not necessary as long as manipulations are made within the boundaries established in this briefing (i.e. turbine load adjustment up to 60 MW e , etc.). A crew update should be performed at approximately every 100 MWe power change. If manpower is available, peer checks should be used for all reactivity changes. OPERATIONAL LIMITS .. Maintain T AVG within +/-6°F of T REF* If T AVG/T REF mismatch >6°F and not trending toward a matched condition or if T AVG :S;551 of, then trip the reactor.

If load reduction due to a loss of vacuum, every effort should be made to maintain the steam dumps closed (Permissive C-9 224.92" Hg). INDUSTRY OE *
Shift supervision must maintain effective oversight and exercise conservative decision making . Correction of significant RCS T AVG deviations should only be via secondary plant control manipulations and not primary plant control manipulations (i.e., do not withdraw control rods or dilute). Printed April 8, 2009 at 10: 11 Approved By Procedure Number J. B. Stanley 18013-C 7 Date Approved 3/24/09 4 of 11 Page Number RAPID POWER REDUCTION

1. 2. _3. 4. 5. *6. ACTION/EXPECTED RESPONSE Perform SHUTDOWN BRIEFING.

Verify rods in AUTO. Reduce Turbine Load at the desired rate up to 5%/min (60 MWE/min). Borate as necessary by initiating 13009, CVCS REACTOR MAKEUP CONTROL SYSTEM. Initiate the Continuous Actions Page. Check desired ramp rate -LESS THAN OR EQUAL TO 5%/MIN. *7. Maintain Tavg within 6°F of Tref: a. Monitor TavglTref deviation (UT-0495).

b. Verify rods inserting as required.

c. Energize Pressurizer back-up heaters as necessary.

6. RESPONSE NOT OBTAINED IF conditions warrant a turbine load rate greater than 5%/min, THEN perform the following:

a. Trip the reactor. _b. Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

_b. Manual rod control should be used with insertions of up to 5 steps at a time. ( ( Approved By Procedure Number Rev J. B. Stanley 18013-C 7 Date Approved 3/24/09 4 of 11 Page Number RAPID POWER REDUCTION

1. 2. 3. 4. 5. *6. *7. ACTION/EXPECTED RESPONSE Perform SHUTDOWN BRIEFING.

Verify rods in AUTO. Reduce Turbine Load at the desired rate up to 5%/min (60 MWE/min). Borate as necessary by initiating 13009, CVCS REACTOR MAKEUP CONTROL SYSTEM. Initiate the Continuous Actions Page. Check desired ramp rate -LESS THAN OR EQUAL TO S%/MIN. Maintain Tavg within 6°F of Tref: a. Monitor Tavg/Tref deviation (UT-0495).

b. Verify rods inserting as required.

c. Energize Pressurizer back-up heaters as necessary.

Printed April 8, 2009 at 10: 11 *6. RESPONSE NOT OBTAINED IF conditions warrant a turbine load rate greater than 5%/min, THEN perform the following:

a. Trip the reactor. b. Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

b. Manual rod control should be used with insertions of up to 5 steps at a time.

Approved By J. B. Stanley Rev 7 Date Approved 3/24/09 Page Number RAPID POWER REDUCTION 5 of 11 *8. ACTION/EXPECTED RESPONSE Maintain reactor power and turbine power-MATCHED. a. Balance reactor power with secondary power reduction using boration and control rods. b. Check rate of reactor power reduction -ADEQUATE FOR PLANT CONDITIONS.

c. Check RCS Tavg -GREATER THAN 551°F (TS 3.4.2). d. Check RCS Tavg -WITHIN 6°F OFTREF. *9. Maintain PRZR Pressure -AT 2235 PSIG. *10. Maintain PRZR Level -AT PROGRAM. RESPONSE NOT OBTAINED b. Perform the following:

_1) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

c. Perform the following:

_1) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

d. IF Tavg/Tref mismatch can NOT be maintained less than 6°F, THEN perform the following:

_1) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. ( Approved By Procedure Number Rev J. B. Stanley 18013-C 7 5 of 11 Page Number Date Approved RAPID POWER REDUCTION 3/24/09 ACTION/EXPECTED RESPONSE *8. Maintain reactor power and turbine power -MATCHED. _a. Balance reactor power with secondary power reduction using boration and control rods. _b. Check rate of reactor power reduction -ADEQUATE FOR PLANT CONDITIONS.

c. Check RCS Tavg -GREATER THAN 551°F (TS 3.4.2). _d. Check RCS Tavg -WITHIN 6°F OF TREF. *9. Maintain PRZR Pressure*

AT 2235 PSIG. *10. Maintain PRZR Level* AT -PROGRAM. Pnnted Apnl 8. 2009 at 10: 12 RESPONSE NOT OBTAINED b. Perform the following: _1) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

c. Perform the following:

_1 ) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

d. lE Tavg/Tref mismatch can NOT be maintained less than 6°F, THEN perform the following:

_1) Trip the reactor. _2) Go to 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

17. 068A2.02 001l2/2/LIQUID RADWASTE/C/A

-2.8INEWIHL15/SROIDS/TNT Given the following conditions: -WMT 9 release is in progress -SS discovers the steps for recirculating WMT 9 in SOP 13126-1, "Liquid Waste Release", were marked N/A What are the impacts and what actions are necessary to correct the consequences of this error? The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release should be stopped, the tank needs to be recirculated for approximately 1 hour prior to sampling. B. The activity level of the WMT contents are not at the lowest possible level. This could result in release of radioactive nuclides> ALARA to UNRESTRICTED areas. The release may continue. A CR must be written to document the error. The reason for the N/A steps should be annotated on the release permit. C. The activity level of the WMT contents are not at the lowest possible level. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release may continue. The flowrate must be reduced to the minimum obtainable flow possible. A CR must be written to document the error. D. The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release should be stopped. The values used in the release permit need to be recalculated using two independent methods prior to restarting the release. 068 Liquid Radwaste System (LRS) A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Lack of tank recirculation prior to release Page: 39 of 56 4/21/2009 ( 1 7. 068A2.02 001l2/2/LIQUID RADW ASTEICI A -2.8/NEW IHL15/SRO/DS/TNT Given the following conditions: -WMT 9 release is in progress -SS discovers the steps for recirculating WMT 9 in SOP 13126-1, "Liquid Waste Release", were marked N/A What are the impacts and what actions are necessary to correct the consequences of this error? A'I The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release should be stopped, the tank needs to be recirculated for approximately 1 hour prior to sampling.

8. The activity level of the WMT contents are not at the lowest possible level. This could result in release of radioactive nuclides>

ALARA to UNRESTRICTED areas. The release may continue. A CR must be written to document the error. The reason for the N/A steps should be annotated on the release permit. C. The activity level of the WMT contents are not at the lowest possible level. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release may continue. The flowrate must be reduced to the minimum obtainable flow possible. A CR must be written to document the error. D. The release permit values are inaccurate. This could result in radioactive nuclides being released to UNRESTRICTED areas greater than license limits. The release should be stopped. The values used in the release permit need to be recalculated using two independent methods prior to restarting the release. 068 Liquid Radwaste System (LRS) A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Lack of tank recirculation prior to release Page: 39 of 56 4121/2009 KIA MATCH ANALYSIS This question requires correct diagnosis of plant conditions to determine the consequences and appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible because WMT's are processed through demins using a recirculating flowpath to lower activity levels pror to release. CR must be written for procedural errors and reason for marking steps N/A. C. Incorrect. Plausible because WMT's are processed through demins using a recirculating flowpath to lower activity levels pror to release. CR must be written for procedural errors and reason for marking steps N/A. Also, reducing release flow rate will reuslt in lower radio nuclide concentrations being released to unrestricted areas. D. Incorrect. Plausible because all aspects of this choice are correct except for recalulating the release permit values. REFERENCES

1. SOP 13216-1 , Liquid Waste Release 2. VEGP Offsite Does Calculation Manual -Chapter 2 -Liquid Effluents VEGP learning objectives:

1. LO-PP-471 01-08: Describe the major steps required for Operations to release a WMT. 2. LO-PP-471 01-09: State the conditions that require immediate termination of a Liquid waste release. 3. LO-PP-47101-10:

State the ODCM, TR, applicabilities, and anyone hour or less actions required for the Liquid Waste Processing System. Page: 40 of 56 4/2112009 KIA MATCH ANALYSIS This question requires correct diagnosis of plant conditions to determine the consequences and appropriate mitigative actions to take to remain within license limits. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Correct. B. Incorrect. Plausible because WMT's are processed through demins using a recirculating flowpath to lower activity levels pror to release. CR must be written for procedural errors and reason for marking steps N/A. C. Incorrect. Plausible because WMT's are processed through demins using a recirculating flowpath to lower activity levels pror to release. CR must be written for procedural errors and reason for marking steps N/A. Also, reducing release flow rate will reuslt in lower radio nuclide concentrations being released to unrestricted areas. D. Incorrect. Plausible because all aspects of this choice are correct except for ( recalulating the release permit values. REFERENCES

1. SOP 13216-1, Liquid Waste Release 2. VEGP Offsite Does Calculation Manual -Chapter 2 -Liquid Effluents VEGP learning objectives:

1. LO-PP-471 01-08: Describe the major steps required for Operations to release a WMT. 2. LO-PP-471 01-09: State the conditions that require immediate termination of a Liquid waste release. 3. LO-PP-471 01-1 0: State the ODCM, TR, applicabilities, and anyone hour or less actions required for the Liquid Waste Processing System. Page: 40 of 56 4/2112009

( Approved By SA Phillips Date Approved 04/23/08 LIQUID WASTE RELEASE CAUTION Procedure Number 13216-1 38 Page Number 6 of 82 INITIALS Excessive running of the WMT pumps on recirc can shorten the life of the pump. The normal recirc time for a sample is approximately 1 hour. 4.1.6 Stop WASTE MONITOR TANK PUMP 09 using 1-HS-1082 AFTER chemistry has taken the required sample. 4.1.7 IF Waste Monitor Tank 09 is NOT acceptable for release, notify SSS. 4.1.8 Perform the following: 4.1.9 4.1.10 4.1.11 a. Contact Chemistry and verify Savannah River Flow Rate has NOT decreased by more than 10% from the permit generation time. NOTE Chemistry procedure 36015-0EMS Data Sheet 5 is the Batch Liquid Discharge Permit b. Obtain SS/SSS approval for the release of WMT 09. c. Record SS/SSS notification and approval on Section Three of the Batch Liquid Release Permit. NOTE If directed by SS the following step may be N/A'ed based on power level, outside air temp, or duration of release. d. IF both units are at power, notify unit one control room to start a third river water pump per 13727-C. Start WASTE MONITOR TANK PUMP 09 using 1-HS-1082. IF 1-RE-0018 is operable, perform 4.1.12 and mark 4.1.13 n/a. IF 1-RE-0018 is inoperable, perform 4.1.13 and mark 4.1.12 n/a. Printed April 8, 2009 at 10:19 ( ( Approved By SA Phillips .. Procedure Number Rev 13216-1 38 Date Approved 04/23/08 Page Number 6 of 82 LIQUID WASTE RELEASE INITIALS CAUTION Excessive running of the WMT pumps on recirc can shorten the life of the pump. The normal recirc time for a sample is approximately 1 hour. 4.1.6 Stop WASTE MONITOR TANK PUMP 09 using 1-HS-1082 AFTER chemistry has taken the required sample. 4.1.7 !E Waste Monitor Tank 09 is NOT acceptable for release, notify SSS. 4.1.8 Perform the following: 4.1.9 4.1.10 4.1.11 a. Contact Chemistry and verify Savannah River Flow Rate has NOT decreased by more than 10% from the permit generation time. NOTE Chemistry procedure 36015-0EMS Data Sheet 5 is the Batch Liquid Discharge Permit b. Obtain SS/SSS approval for the release of WMT 09. c. Record SS/SSS notification and approval on Section Three of the Batch Liquid Release Permit. NOTE If directed by SS the following step may be N/A'ed based on power level, outside air temp, or duration of release. d. IF both units are at power, notify unit one control room to start a third river water pump per 13727-C. Start WASTE MONITOR TANK PUMP 09 using 1-HS-1082. IF 1-RE-0018 is operable, perform 4.1.12 and mark 4.1.13 n/a. IF 1-RE-0018 is inoperable, perform 4.1.13 and mark 4.1.12 n/a. Printed April 8,2009 at 10:19 VEGPODCM 2.1.2 Liquid Effluent Concentration Control In accordance with Technical Specifications 5.5.4.b and 5.5.4.c, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 1 x 10--4 f,lCi/mL total activity. 2.1.2.1 Applicability This limit applies at all times. 2.1.2.2 Actions With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the limits stated in Section 2.1.2, immediately restore the concentration to within the stated limits. This control does not affect shutdown requirements or MODE changes. 2.1.2.3 Surveillance Requirements The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 2-3. The results of radioactive analyses shall be used with the calculational methods in Section 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2. 2.1.2.4 Basis This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2 (1959). The resulting concentration of 2 x 10-4 was then multiplied by the ratio of the effluent concentration limit for Xe-135, stated in Appendix B, Table 2, Column 1 of 10 CFR 20 (paragraphs 20.1001 to 20.2401), to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 (paragraphs 20.1 to 20.601), to obtain the limiting concentration of 1 x 10-4 f,lCi/mL. 2-7 VER24 ( VEGPODCM 2.1.2 Liquid Effluent Concentration Control In accordance with Technical Specifications 5.5.4.b and 5.5.4.c, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 1 x 1 0-4 total activity. 2.1.2.1 Applicability This limit applies at all times. 2.1.2.2 Actions With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the limits stated in Section 2.1.2, immediately restore the concentration to within the stated limits. This control does not affect shutdown requirements or MODE changes. 2.1.2.3 Surveillance Requirements The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 2-3. The results of radioactive analyses shall be used with the calculational methods in Section 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2. 2.1.2.4 Basis This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2 (1959). The resulting concentration of 2 x 10-4 was then multiplied by the ratio of the effluent concentration limit for Xe-135, stated in Appendix B, Table 2, Column 1 of 10 CFR 20 (paragraphs 20.1001 to 20.2401), to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 (paragraphs 20.1 to 20.601), to obtain the limiting concentration of 1 x 1 0-4 2-7 VER24

18. 071A2.09 001l2/2/WASTE GAS/C/A -3.5INEWIHL15/SRO/DS/TNT Initial conditions:

-Both units at full power -No radioactive releases in progress -Unit 2 has high RCS activity due to fuel leaks Current conditions: -Waste Gas system is in service using Gas Decay Tank #1 -Unit 1 plant vent radiation monitor 1 RE-12442C is in high alarm -Waste gas effluent radiation monitor ARE-0014 is in high alarm -The Auxiliary Building Operator reports that Waste Gas Shutdown Decay Tank # 9 pressure is 50 psig and rapidly lowering. Which of the following is the correct action to take per plant procedures? A. Isolate air to RV-0014 to terminate the unplanned release. Shift the waste gas system to the low pressure mode of operation. Shut A-1902-U4-004 downstream of RE-0014 to stop the unplanned release. Initiate maintenance on the relief valve for shutdown decay tank #9. C. Isolate air to RV-0014 to terminate the unplanned release. Declare RE-0014 inoperable and make a 1 hour notification to the NRC. D. Shut A-1902-U4-004 downstream of RE-0014 to stop the unplanned release. Terminate waste additions from the Unit 2 VCT due to the high RCS activity. 071 Waste Gas Disposal System (WGDS) A2.09 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck open relief valve KIA MATCH ANALYSIS Page: 41 of 56 4/2112009 ( 18. 071A2.09 001l2121WASTE GAS/CIA -3.5/NEW 1HL15/SROIDS/TNT Initial conditions: -Both units at full power -No radioactive releases in progress -Unit 2 has high RCS activity due to fuel leaks Current conditions: -Waste Gas system is in service using Gas Decay Tank #1 -Unit 1 plant vent radiation monitor 1 RE-12442C is in high alarm -Waste gas effluent radiation monitor ARE-0014 is in high alarm -The Auxiliary Building Operator reports that Waste Gas Shutdown Decay Tank # 9 pressure is 50 psig and rapidly lowering. Which of the following is the correct action to take per plant procedures? A. Isolate air to RV-0014 to terminate the unplanned release. Shift the waste gas system to the low pressure mode of operation. Shut A-1902-U4-004 downstream of RE-0014 to stop the unplanned release. Initiate maintenance on the relief valve for shutdown decay tank #9. C. Isolate air to RV-0014 to terminate the unplanned release. Declare RE-0014 inoperable and make a 1 hour notification to the NRC. D. Shut A-1902-U4-004 downstream of RE-0014 to stop the unplanned release. Terminate waste additions from the Unit 2 VCT due to the high RCS activity. 071 Waste Gas Disposal System (WGDS) A2.09 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: ( Stuck open relief valve KIA MATCH ANALYSIS Page: 41 of 56 4/2112009 ( This question requires recalling actions written in alarm response procedures associated with the failure given. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal,abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since RV-0014 is the waste gas effluent automatic isolation valve and the waste gas system has 2 pressure modes of operation. B. Correct. C. Incorrect. Plausible since RV-0014 is controlled by RE-0014 and notification is required for unplanned releases. D. Incorrect. Plausible since the correct action is to isolate a manual valve downstream of RE-0014. REFERENCES

1. 17100-1 ARP for Process and Effluent RMS 2. 17216-1 window A08 ARP for waste gas processing panel VEGP learning objectives:

LO-PP-46101-03: State the purposes of the following Gaseous Radwaste System components:

a. gas decay tanks b. gas decay shutdown tanks c. hydrogen recombiners

d. waste gas com pressors e. trip valve RV-014 f. rad monitors RE-013 and RE-014 g. gas decay tank relief valves and header Page: 42 of 56 4/2112009 ( ( This question requires recalling actions written in alarm response procedures associated with the failure given. These actions are not immediate actions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since RV-0014 is the waste gas effluent automatic isolation valve and the waste gas system has 2 pressure modes of operation. B. Correct. C. Incorrect. Plausible since RV-0014 is controlled by RE-0014 and notification is required for unplanned releases. D. Incorrect. Plausible since the correct action is to isolate a manual valve downstream of RE-0014. REFERENCES

1. 17100-1 ARP for Process and Effluent RMS 2. 17216-1 window A08 ARP for waste gas processing panel VEGP learning objectives:

LO-PP-461 01-03: State the purposes of the following Gaseous Radwaste System components:

a. gas decay tanks b. gas decay shutdown tanks c. hydrogen recombiners

d. waste gas compressors

e. trip valve RV-014 f. rad monitors RE-013 and RE-014 g. gas decay tank relief valves and header Page: 42 of 56 412112009 ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE PROCESSING SYSTEM-GAS PANEL ORIGIN SETPOINT A-RE-0014 Variable 1.0 PROBABLE CAUSE 1. High radiation level in gas being released.

2. Radiation from safety valve release. 2.0 AUTOMATIC ACTIONS Critical 3.0 Plant Vent Radwaste Gas, A-RV-0014, closes. INITIAL OPERATOR ACTIONS Verify A-RV-0014 is closed. Printed April 8, 2009 at 10:45 WINDOWA08 PLANT VENT MONITOR HIRAD. Initial CV Initial 2.2 ( ! ( Approved By T.E. Tynan Procedure Number Rev 17216-1 2.2 Date Approved 10-15-99 ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE PROCESSING SYSTEM-GAS PANEL Page Number 18 of 47 ORIGIN SETPOINT A-RE-0014 Variable 1.0 PROBABLE CAUSE 1. High radiation level in gas being released.

2. Radiation from safety valve release. 2.0 AUTOMATIC ACTIONS Critical 3.0 Plant Vent Radwaste Gas, A-RV-0014, closes. INITIAL OPERATOR ACTIONS Verify A-RV-0014 is closed. Pnnted Apnl 8, 2009 at 10:45 WINDOW A08 PLANT VENT MONITOR HI RAD. Initial CV Initial Approved By T.E. Tynan Date Approved 10-15-99 ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE PROCESSING SYSTEM-GAS PANEL WINDOWA08 (Continued)

4.0 SUBSEQUENT

OPERATOR ACTIONS 1. Determine the gas discharge line radiation using A-RI-0014 on Panel PGPP. 2. Verify A-1902-U4-004 closed. 3. Notify the Control Room of the alarm and system status prior to performing the remaining actions. 4. IF planned release was in progress:

a. Request Health Physics to sample waste gas decay shutdown tank being released.

b. Check release conditions and rate. 5. IF release is not planned: a. Check waste gas decay shutdown tank pressure using A-PIS-1 056 and A-PIS-1057 on Panel PGPP. b. Check status of A-PSV-7884A and A-PSV-7884B.

c. IF relief valves have lifted at 100 psig, isolate the source. 6. Refer to the requirements of the ODCM Manual. 7. IF equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORY

OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB129, X6MOO-449-1, P.L.S. Printed April 21, 2009 at ( ( Approved By T.E. Tynan Date Approved 10-15-99 ANNUNCIATOR RESPONSE PROCEDURE FOR ALB ON WASTE PROCESSING SYSTEM-GAS PANEL Procedure Number Rev 17216-1 2.2 Page Number 19 of 47 WINDOW A08 (Continued)

4.0 SUBSEQUENT

OPERATOR ACTIONS 1. Determine the gas discharge line radiation using A-RI-0014 on Panel PGPP. 2. Verify A-1902-U4-004 closed. 3. Notify the Control Room of the alarm and system status prior to performing the remaining actions. 4. lE planned release was in progress:

a. Request Health Physics to sample waste gas decay shutdown tank being released.

b. Check release conditions and rate. 5. lE release is not planned: a. Check waste gas decay shutdown tank pressure using A-PIS-1 056 and A-PIS-1057 on Panel PGPP. b. Check status of A-PSV-7884A and A-PSV-7884B.

c. lE relief valves have lifted at 100 psig, isolate the source. 6. Refer to the requirements of the ODCM Manual. 7. lE equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORY

OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB129, X6AAOO-449-1, P.L.S. Printed April 21, 2009 at 12:25 ( 19. G2.1.15 001!3INIA/CONDUCT OF OPS/CIA -3.4INEW/HL15/SRO/DS/TNT Initial conditions: -The need to generate a new standing order has been identified -The standing order will provide temporary instructions not covered by a plant procedure. Which of the following choices is correct concerning review and approval of this new standing order? A. The duration of the standing order will be limited to 14 days. The Shift Manager will review and approve IT a 10CFR 50.59 screening is not required. The standing order will not have a required termination date. The Unit Superintendent will review and approve, a 1 OCFR 50.59 screening is required. C. The duration of the standing order will be limited to 14 days. The Unit Superintendent will review and approve IT a 1 OCFR 50.59 screening is required. D. The standing order will not have a required termination date. The Shift Supervisor will review and approve, a 1 OCFR 50.59 screening is required Page: 43 of 56 4/2112009 ( ( ( 19. G2.l.l5 001!3INIA/CONDUCT OF OPS/CIA -3.4INEW/HL15/SRO/DS/TNT Initial conditions: -The need to generate a new standing order has been identified -The standing order will provide temporary instructions not covered by a plant procedure. Which of the following choices is correct concerning review and approval of this new standing order? A. The duration of the standing order will be limited to 14 days. The Shift Manager will review and approve !f a 10CFR 50.59 screening is not required. The standing order will not have a required termination date. The Unit Superintendent will review and approve, a 10CFR 50.59 screening is required. C. The duration of the standing order will be limited to 14 days. The Unit Superintendent will review and approve !f a 10CFR 50.59 screening is required. D. The standing order will not have a required termination date. The Shift Supervisor will review and approve, a 10CFR 50.59 screening is required Page: 43 of 56 4/2112009 ( (, G2.1.15 Conduct of Operations Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc. KIA MATCH ANALYSIS This question requires the student to determine which processes to use to review and approve standing orders that involve actions not contained in plant procedures which is a 10CFR50.59 screening. This question meets the requirements of 1 OCFR55.43(b )(3) Facility licensee procedures required to obtain authority for design and operating chnages in the facility and is an SRO question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since temporary changes to procedures are limited to 14 days. B. Correct. C. Incorrect. Plausible since temporary changes to procedures are limited to 14 days. The Unit Superintendent will have to review and a 10CFR50.59 screening is required. D. Incorrect. Plausible since standing orders are do not have termination dates and a 1 OCFR50.59 screening is required. REFERENCES 10002-C, Plant Operating Orders VEGP learning objectives: LO-LP-63502-01

With regards to plant operating orders, describe the following: (SRO ONLY) a. how standing orders are reviewed and deleted b. responsibility for implementation of standing orders c. responsibility for issuing night orders d. the terms: "standing order" and "night order" Page: 44 of 56 4/2112009 ( ( ( G2.1.15 Conduct of Operations Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc. KIA MATCH ANALYSIS This question requires the student to determine which processes to use to review and approve standing orders that involve actions not contained in plant procedures which is a 10CFR50.59 screening.

This question meets the requirements of 1 OCFR55.43(b )(3) Facility licensee procedures required to obtain authority for design and operating chnages in the facility and is an SRO question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since temporary changes to procedures are limited to 14 days. B. Correct. C. Incorrect. Plausible since temporary changes to procedures are limited to 14 days. The Unit Superintendent will have to review and a 1 OCFR50.59 screening is required. D. Incorrect. Plausible since standing orders are do not have termination dates and a 1 OCFR50.59 screening is required. REFERENCES 10002-C, Plant Operating Orders VEGP learning objectives: LO-LP-63502-01

With regards to plant operating orders, describe the following: (SRO ONLY) a. how standing orders are reviewed and deleted b. responsibility for implementation of standing orders c. responsibility for issuing night orders d. the terms: "standing order" and "night order" Page: 44 of 56 412112009 AnnrnVAn By Procedure Number Rev S. A. Phillips 10002-C 16 Date Approved 12/3/2008 3.1.1.2 3.1.1.3 3.1.1.4 3.1.1.5 PLANT OPERATING ORDERS Page Number NOTES

If the next sequential log number can not be obtained from the excel spreadsheet located in the Operations Standing Order folder on S: drive, it may be obtained from the hard copy of the Log in the Standing Order Log Book.
All Standing Orders will begin with version 1. The next sequential Standing Order number is obtained from the Standing Order Log on S: drive. The Standing Order is written on the Standing Order Form (Figure 1) and imported into Documentum.

4 of 12 a. If the Standing Order includes directions to perform actions not covered by current approved procedures, or, at the direction of Operations' management, a 10CFR 50.59 screening will be completed, per Administrative procedure 00056-C, and linked to the Standing Order. b. 50.59 screenings performed as part of an Engineering evaluation (REA) for the evolution to be performed, may be used to satisfy the 50.59 screening requirement for the Standing Order. c. Standing Orders that initiate additional monitoring and evaluation of plant equipment or schedules the performance of plant evolutions using current approved procedures do not require a 50.59 screening. The Standing Order is routed for review and approval by the Admin Superintendent or Unit Superintendent. The final approved version of the Standing Order is then made available to the Plant population for use. NOTE Notification of a new or terminated Standing Order is made to copy holders via the Documentum Inbox. A copy of the applicable Standing Order Log is printed and placed in the appropriate tab of the Standing Order Log Book. Printed April 8, 2009 at 10:50 ( , ( Approved By S. A. Phillips Date Approved 12/3/2008 3.1.1.2 3.1.1.3 3.1.1.4 Procedure Number Rev 10002-C 16 PLANT OPERATING ORDERS Page Number NOTES

If the next sequential log number can not be obtained from the excel spreadsheet located in the Operations Standing Order folder on S: drive, it may be obtained from the hard copy of the Log in the Standing Order Log Book.
All Standing Orders will begin with version 1. The next sequential Standing Order number is obtained from the Standing Order Log on S: drive. The Standing Order is written on the Standing Order Form (Figure 1) and imported into Documentum.

4 of 12 a. If the Standing Order includes directions to perform actions not covered by current approved procedures, or, at the direction of Operations' management, a 10CFR 50.59 screening will be completed, per Administrative procedure 00056-C, and linked to the Standing Order. b. 50.59 screenings performed as part of an Engineering evaluation (REA) for the evolution to be performed, may be used to satisfy the 50.59 screening requirement for the Standing Order. c. Standing Orders that initiate additional monitoring and evaluation of plant equipment or schedules the performance of plant evolutions using current approved procedures do not require a 50.59 screening. The Standing Order is routed for review and approval by the Admin Superintendent or Unit Superintendent. The final approved version of the Standing Order is then made available to the Plant population for use. NOTE Notification of a new or terminated Standing Order is made to copy holders via the Documentum Inbox. 3.1.1.5 A copy of the applicable Standing Order Log is printed and placed in the appropriate tab of the Standing Order Log Book. Printed Apnl 8, 2009 at 10.50 ( Approved By S. A. Phillips Date Approved 12/3/2008

3.1.1.6 3.1.1.7 3.1.2 3.1.3 3.1.3.1 3.1.3.2 3.1.3.3 3.1.3.4 3.1.3.5 Procedure Number Rev 10002-C 16 PLANT OPERATING ORDERS Page Number The Shift Manager has the Standing Order, and its 50.59 screening (if applicable), placed in the Standing Order Book and directs the Shift Supervisor(s) (SS) to enforce it. 5 of 12 The SS implements all approved Standing Orders which apply to his assigned unit. Revisions Revisions to current approved Standing Orders will be processed the same as new Standing Orders and checked in to Documentum as the next major version. Termination When a Shift Manager determines that a Standing Order should be terminated, a Standing Order Termination Request (Figure 2) is prepared and forwarded to the Admin or Operations Superintendent.

The Termination Request will be the next minor version of the Standing Order i.e. if the Standing Order is version 1.0 the termination request will be version 1.1. The Admin or Operations Superintendent, or his designee, may either approve and publish the termination request or reject the request and return it to the originator. An approved Standing Order Termination Request is imported into Documentum, routed and approved as the next minor version of the original Standing Order and is made available to the Plant population. The termination date is recorded in the Standing Order Log and a hard copy of the log is placed in the Standing Order Log Book. The Shift Manager has the Standing Order removed from the Standing Order Book and destroyed, and directs the SS to terminate its use. Printed April 8, 2009 at 1 ( ( Approved By S. A. Phillips Date Approved 12/3/2008 3.1.1.6 3.1.1.7 3.1.2 3.1.3 3.1.3.1 3.1.3.2 3.1.3.3 3.1.3.4 3.1.3.5 Procedure Num Rev 10002-C 16 PLANT OPERATING ORDERS Page Number The Shift Manager has the Standing Order, and its 50.59 screening (if applicable), placed in the Standing Order Book and directs the Shift Supervisor(s) (SS) to enforce it. 5 of 12 The SS implements all approved Standing Orders which apply to his assigned unit. Revisions Revisions to current approved Standing Orders will be processed the same as new Standing Orders and checked in to Documentum as the next major version. Termination When a Shift Manager determines that a Standing Order should be terminated, a Standing Order Termination Request (Figure 2) is prepared and forwarded to the Admin or Operations Superintendent. The Termination Request will be the next minor version of the Standing Order i.e. if the Standing Order is version 1.0 the termination request will be version 1.1. The Admin or Operations Superintendent, or his designee, may either approve and publish the termination request or reject the request and return it to the originator. An approved Standing Order Termination Request is imported into Documentum, routed and approved as the next minor version of the original Standing Order and is made available to the Plant population. The termination date is recorded in the Standing Order Log and a hard copy of the log is placed in the Standing Order Log Book. The Shift Manager has the Standing Order removed from the Standing Order Book and destroyed, and directs the SS to terminate its use. Printed April 8, 2009 at 10:50

20. G2.2.22 001l3/N/A/EQUIPMENT CONTROLIMEM

-4.7/B -FARLEY 2004/HL15/SRO/DS/TNT During a review of the surveillance schedules it is discovered that a 31 day surveillance for a particular component was last performed 42 days ago. Within 24 hours of this discovery, the surveillance was performed.

The surveillance failed because it did not meet surveillance criteria.
Two hours later, a valve was adjusted and retesting provided satisfactory results. Which one of the following is correct for this condition?

The component was INOPERABLE _______ until satisfactory retesting was completed. A. for the last 42 days B. from the time the grace period expired C!" from the time of the failed surveillance D. from the time of identifying the surveillance NOT performed KIA G2.2.22 Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA MATCH ANALYSIS This question involves the application of generic LCO surveillance requirements of technical specifications for a missed SR. The question meets 10CFR 55.34(b) item #2 -Facility operating limitations in the technical specifications and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect -because of 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, B. Incorrect -SR 3.0.3 and SR 3.0.1. The grace period is the time of the surveillance which is 31 days X 1.25 = 38 days. The candidate may believe the INOPERABLE time period starts after the grace period runs out, which in this case is at approx. day 37 of the 42 day time period. This is incorrect also. C. Correct -The component was INOPERABLE since the time of unacceptable results Page: 45 of 56 4/2112009 ( ( 20. G2.2.22 001l3/N/A/EQUIPMENT CONTROLIMEM -4.7/B -FARLEY 2004/HLl5/SROIDS/TNT During a review of the surveillance schedules it is discovered that a 31 day surveillance for a particular component was last performed 42 days ago. Within 24 hours of this discovery, the surveillance was performed.

The surveillance failed because it did not meet surveillance criteria.
Two hours later, a valve was adjusted and retesting provided satisfactory results. Which one of the following is correct for this condition?

The component was INOPERABLE _______ until satisfactory retesting was completed. A. for the last 42 days B. from the time the grace period expired from the time of the failed surveillance D. from the time of identifying the surveillance NOT performed KIA G2.2.22 Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA MATCH ANALYSIS This question involves the application of generic LCO surveillance requirements of technical specifications for a missed SR. The question meets 10CFR 55.34(b) item #2 -Facility operating limitations in the technical specifications and their bases. ANSWER I DISTRACTOR ANALYSIS A. Incorrect -because of 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, B. Incorrect -SR 3.0.3 and SR 3.0.1. The grace period is the time of the surveillance which is 31 days X 1.25 = 38 days. The candidate may believe the INOPERABLE time period starts after the grace period runs out, which in this case is at approx. day 37 of the 42 day time period. This is incorrect also. C. Correct -The component was INOPERABLE since the time of unacceptable results Page: 45 of 56 412112009 until satisfactory retesting was completed. D. Incorrect -There is a delay period of 24 hours, and may be greater if a risk evaluation is performed and the risk impact managed. Since this was accomplished wli 24 hours and it failed its surveillance, the component is declared INOPERABLE immediately, and the LCO entered. REFERENCES VEGP Technical Specifications SR 3.0.3: If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. VEGP learning objectives: LO-LP-39204-04: State the allowable time intervals for extension of surveillances. State the result of failure to perform surveillances within this period. LO-LP-39204-06: In regard to surveillances, determine when time delay may be applied and the maximum time allowed to perform the surveillance. Page: 46 of 56 4/2112009 ( until satisfactory retesting was completed. D. Incorrect -There is a delay period of 24 hours, and may be greater if a risk evaluation is performed and the risk impact managed. Since this was accomplished wli 24 hours and it failed its surveillance, the component is declared INOPERABLE immediately, and the LCO entered. REFERENCES VEGP Technical Specifications SR 3.0.3: If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. VEGP learning objectives: LO-LP-39204-04: State the allowable time intervals for extension of surveillances. State the result of failure to perform surveillances within this period. LO-LP-39204-06: In regard to surveillances, determine when time delay may be applied and the maximum time allowed to perform the surveillance. Page: 46 of 56 412112009

1. G2.2.22 001 Page: 1 During a review of the surveillance schedules it is discovered that a 31 day surveillance for a particular component was last performed 42 days ago. Within 24 hours of this discovery, the surveillance was performed.

The surveillance failed because it did not meet surveillance criteria.
Two hours later, a valve was adjusted and retesting provided satisfactory results. Which one of the following is correct for this condition?

The component was INOPERABLE _______ until satisfactory retesting was completed. A. for the last 42 days B. from the time the grace period expired C!' from the time of the failed surveillance D. from the time of identifying the surveillance NOT performed 4/8/2009 1. G2.2.22 001 During a review of the surveillance schedules it is discovered that a 31 day surveillance for a particular component was last performed 42 days ago. Within 24 hours of this ( discovery, the surveillance was performed. ( ( Page: 1

The surveillance failed because it did not meet surveillance criteria.
Two hours later, a valve was adjusted and retesting provided satisfactory results. Which one of the following is correct for this condition?

The component was INOPERABLE _______ until satisfactory retesting was completed. A. for the last 42 days B. from the time the grace period expired from the time of the failed surveillance D. from the time of identifying the surveillance NOT performed 4/8/2009 SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SR 3.0.2 SR 3.0.3 Vogtle Units 1 and 2 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per ... " basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. 3.0-4 (continued) Amendment No. 125 (Unit 1) Amendment No.1 03 (Unit 2) ( ( SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SR 3.0.2 SR 3.0.3 Vogtle Units 1 and 2 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per ... " basis, the above Frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. 3.0-4 (continued) Amendment No. 125 (Unit 1) Amendment No.1 03 (Unit 2) ( SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 ( continued) SR 3.0.4 Vogtle Units 1 and 2 When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 3.0-5 Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2) SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 (continued) SR 3.0.4 Vogtle Units 1 and 2 When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 3.0-5 Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2)

21. G2.2.38 001l3INIAIEQUIPMENT CONTROL/CIA

-4.SINEW/HLlS/SROIDS/TNT The following equipment is inoperable at 100% power: -Boric Acid Transfer Pump #1 -CCP-A -RCP Common Thermal Barrier Isolation Valve, HV-2041 -TDAFWP A Loss Of Safety Function (LOS F) evaluation is required for the (a) because it is (b) A. a. TDAFWP b. a Technical Specification (TS) required support component for AMSAC B. a. BATP #1 b. a Technical Requirement (TR) supported component for the BAST a. CCP-A* b. a Technical Specification (TS) required component supported by DG-A D. a. RCP Thermal Barrier Isolation Valve b. one of four Technical Requirement (TR) valves that support RCP operation. ( Page: 47 of 56 4/2112009

21. G2.2.38 001l3!NIAIEQUIPMENT CONTROL/CIA

-4.5!NEW/HLl5/SROIDS/TNT The following equipment is inoperable at 100% power: -Boric Acid Transfer Pump #1 -CCP-A -RCP Common Thermal Barrier Isolation Valve, HV-2041 -TDAFWP A Loss Of Safety Function (LOSF) evaluation is required for the (a) because it is (b) A. a. TDAFWP b. a Technical Specification (TS) required support component for AMSAC B. a. BATP #1 b. a Technical Requirement (TR) supported component for the BAST a. CCP-A ( b. a Technical Specification (TS) required component supported by DG-A D. a. RCP Thermal Barrier Isolation Valve b. one of four Technical Requirement (TR) valves that support RCP operation. Page: 47 of 56 4/2112009 G2.2.38 Equipment Control Knowledge of conditions and limitations in the facility license. KIA MATCH ANALYSIS This question requires determining if a loss of safety function evaluation is required from a given set of plant conditions. The LOSF program is a requirement of tech spec 5.5.15. This question meets the requirements of 10CFR 55.43(b) item #2 and is an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since TDAFWP is a TS required component and AMSAC deals with automatic mitigating actions for an ATWT. B. Incorrect. Plausible since this a required TR component dealing with reactivity control. C. Correct. D. Incorrect. Plausible since this is a required TR component supporting the RCP's which provide reactor protection trips and nuclear heat removal. REFERENCES VEGP Ops Admin procedure 1000S-C, Recording Limiting Conditions for Operation VECP Tech Spec 5.5.15 VEGP learning objectives: LO-LP-39217-0S: State the purpose of the Safety Function Determination Program. LO-LP-39217-09: Define" Loss of Safety Function". LO-LP-39217-10: Given a set of plant conditions, equipment availability, and operational mode, determine if a loss of safety function exists. Page: 48 of 56 4/2112009 ( ( G2.2.38 Equipment Control Knowledge of conditions and limitations in the facility license. KIA MATCH ANALYSIS This question requires determining if a loss of safety function evaluation is required from a given set of plant conditions. The LOSF program is a requirement of tech spec 5.5.15. This question meets the requirements of 1 OCFR 55.43(b) item #2 and is an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since TDAFWP is a TS required component and AMSAC deals with automatic mitigating actions for an ATWT. B. Incorrect. Plausible since this a required TR component dealing with reactivity control. C. Correct. D. Incorrect. Plausible since this is a required TR component supporting the RCP's which provide reactor protection trips and nuclear heat removal. REFERENCES VEGP Ops Admin procedure 10008-C, Recording Limiting Conditions for Operation VECP Tech Spec 5.5.15 VEGP learning objectives: LO-LP-39217-08: State the purpose of the Safety Function Determination Program. LO-LP-39217-09: Define" Loss of Safety Function". LO-LP-39217-10: Given a set of plant conditions, equipment availability, and operational mode, ( determine if a loss of safety function exists. Page: 48 of 56 4/2112009 ( Approved By Procedure Number Rev C. H. Williams, Jr 10008-C 25.2 Date Approved 3-29-2006 Page Number 2 of 26 RECORDING LIMITING CONDITIONS FOR OPERATION

1.0 PURPOSE

This procedure prescribes the method to record the failure to meet the Limiting Conditions for Operation (LCO), or Technical Requirement, the associated ACTION requirements, any change in status effecting the ACTION, and the return to compliance with LCO/TR. This procedure also includes instructions for implementing Technical Specification 5.5.15, the Safety Function Determination Program. As required by LCO 3.0.6, this program ensures that proper actions are taken such that multiple inoperable Structures, Systems, or Components (SSC) do not result in an undetected LOSS OF SAFETY FUNCTION. This procedure also ensures that the allowed out of service time of SUPPORTED SYSTEMS is not inappropriately extended as a result of multiple inoperable SUPPORT SYSTEMS. 2.0 DEFINITIONS

2.1 LIMITING

CONDITION FOR OPERATION (LCO) A condition specified in the plant Technical Specifications (TS) or Technical Requirements Manual (TRM) which limits unit operations. An LCO may be entered by an equipment malfunction or a change in a unit parameter. If an LCO is not met, the associated ACTION requirements shall be met. 2.2 TECHNICAL REQUIREMENT (TR) A condition specified in the plant Technical Requirements Manual (TRM) which limits unit operations. A TR may be entered by an equipment malfunction or a change in a unit parameter. If a TR is not met, the associated ACTION requirements shall be met. TR and Technical Requirement Surveillances (TRS) associated with each TR are implemented in the same way as Technical Specifications. However, TRs and TRSs are treated as plant procedures and are not part of the Technical Specification. Therefore exceptions apply (Reference TRM Section 11.5). Printed April 8, 2009 at 12:08 ( ( Approved By Procedure Number Rev C. H. Williams, Jr 10008-C 25.2 Page Number 2 of 26 Date Approved 3-29-2006 RECORDING LIMITING CONDITIONS FOR OPERATION

1.0 PURPOSE

2.0 2.1 2.2 This procedure prescribes the method to record the failure to meet the Limiting Conditions for Operation (LCO), or Technical Requirement, the associated ACTION requirements, any change in status effecting the ACTION, and the return to compliance with LCO/TR. This procedure also includes instructions for implementing Technical Specification 5.5.15, the Safety Function Determination Program. As required by LCO 3.0.6, this program ensures that proper actions are taken such that multiple inoperable Structures, Systems, or Components (SSC) do not result in an undetected LOSS OF SAFETY FUNCTION. This procedure also ensures that the allowed out of service time of SUPPORTED SYSTEMS is not inappropriately extended as a result of multiple inoperable SUPPORT SYSTEMS. DEFINITIONS LIMITING CONDITION FOR OPERATION (LCO) A condition specified in the plant Technical Specifications (TS) or Technical Requirements Manual (TRM) which limits unit operations. An LCO may be entered by an equipment malfunction or a change in a unit parameter. If an LCO is not met, the associated ACTION requirements shall be met. TECHNICAL REQUIREMENT (TR) A condition specified in the plant Technical Requirements Manual (TRM) which limits unit operations. A TR may be entered by an equipment malfunction or a change in a unit parameter. If a TR is not met, the associated ACTION requirements shall be met. TR and Technical Requirement Surveillances (TRS) associated with each TR are implemented in the same way as Technical Specifications. However, TRs and TRSs are treated as plant procedures and are not part of the Technical Specification. Therefore exceptions apply (Reference TRM Section 11.5). Printed April 8, 2009 at 12:08 Date Approved 3-29-2006 3.1.1.9 3.1.1.10 3.1.1.11 3.1.1.12 3.1.1.13 Procedure Number Rev 10008-C 25.2 RECORDING LIMITING CONDITIONS FOR OPERATION Page Number 80f26 SUPPORT or SUPPORTED System -LOSF Evaluation Required Check YES or NO If the inoperable SSC affects a required system that either (1) provides support for another system required by Technical Specifications or (2) is supported by other systems required by Technical Specifications, then YES should be marked and a Loss of Safety Function (LOSF) Evaluation should be performed in accordance with Section 3.5 and TS 5.5.15. (Example -Train A 125V DC provides support for Train A NSCW) If the SSC is required by the TRM, then NO should be marked since LCO 3.0.6 does not apply to the TRM. Section II -Condition Report, Maintenance Work Order, Procedure, Or Tagout Number(s) A listing of all CR, MWO, Procedure, or Tagout numbers which apply to the restoration of the LCO/TR should be documented in Section II. Additional Condition Report numbers should be added to this section if the LCOITR is referenced on new Condition Reports. MWO numbers should be obtained prior to completion of Section I of LCO/TR Status Sheet. If other MWOs are worked under this LCO/TR, the MWO numbers should be added when the MWO work authorization is given. If required, attach continuation sheet for additional information, then check continuation sheet attached block as required. CRs are not required for LCO entry due to pre-planned activities. Remarks Include any additional information on the LCO/TR in this section. SS/SM Signatures The SS and SM both sign indicating completeness and correctness of all entries. Disposition The original LCO/TR Status Sheet is placed in the LCO/TR Status Binder. Printed April 8, 2009 at 12:10 ( ( Approved By Procedure Number Rev C. H. Williams, Jr 10008-C 25.2 Date Approved 3-29-2006 3.1.1.9 3.1.1.10 3.1.1.11 3.1.1.12 3.1.1.13 RECORDING LIMITING CONDITIONS FOR OPERATION Page Number 8 of 26 SUPPORT or SUPPORTED System -LOSF Evaluation Required Check YES or NO If the inoperable SSC affects a required system that either (1) provides support for another system required by Technical Specifications or (2) is supported by other systems required by Technical Specifications, then YES should be marked and a Loss of Safety Function (LOS F) Evaluation should be performed in accordance with Section 3.5 and TS 5.5.15. (Example -Train A 125V DC provides support for Train A NSCW) If the SSC is required by the TRM, then NO should be marked since LCO 3.0.6 does not apply to the TRM. Section II -Condition Report, Maintenance Work Order, Procedure, Or Tagout Number(s) A listing of all CR, MWO, Procedure, or Tagout numbers which apply to the restoration of the LCO/TR should be documented in Section II. Additional Condition Report numbers should be added to this section if the LCO/TR is referenced on new Condition Reports. MWO numbers should be obtained prior to completion of Section I of LCO/TR Status Sheet. If other MWOs are worked under this LCO/TR, the MWO numbers should be added when the MWO work authorization is given. If required, attach continuation sheet for additional information, then check continuation sheet attached block as required. CRs are not required for LCO entry due to pre-planned activities. Remarks Include any additional information on the LCO/TR in this section. SS/SM Signatures The SS and SM both sign indicating completeness and correctness of all entries. Disposition The original LCO/TR Status Sheet is placed in the LCO/TR Status Binder. Printed April 8, 2009 at 12:10

22. G2.3.4 OOl/3/NIAlRADIATION CONTROLIMEM

-3.7/B-CR3 20071HL15/SRO/DS/TNT Initial conditions: -An emergency has been declared -CNMT pressure is 38 psig and rising -The only available CNMT Spray pump motor has high temperature alarms -The crew suspects inadequate venting of the motor cooler -An operator has been selected to vent the motor cooler -The operator has received 1000 mrem TEDE, but has received NO dose during this event. Per 91301-C, " Emergency Exposure Guidelines" what is the maximum dose the operator can recieve? A. 4 Rem B. 5 Rem C. 9 Rem 10 Rem Page: 49 of 56 4/2112009 ( ( 22. G2.3.4 001l3/NIAIRADIATION CONTROLIMEM -3.7/B-CR3 2007/HL15/SRO/DS/TNT Initial conditions: -An emergency has been declared -CNMT pressure is 38 psig and rising -The only available CNMT Spray pump motor has high temperature alarms -The crew suspects inadequate venting of the motor cooler -An operator has been selected to vent the motor cooler -The operator has received 1000 mrem TEDE, but has received NO dose during this event. Per 91301-C, " Emergency Exposure Guidelines" what is the maximum dose the operator can recieve? A. 4 Rem B. 5 Rem C. 9 Rem 10 Rem Page: 49 of 56 4/2112009 G2.3.4 Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions. KIA MATCH ANALYSIS This question requires the applicant to determine type of activity and appropriate emergency dose limit. From that limit the applicant must also determine that emergency doses exclude previous occupational dose history. Question meets 10CFR55.43(b) criteria item #4 -Radiation hazards and contamination conditions that may occur during normal and abnormal situations, including maintenance activities and various contamination conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the person has a dose history of 1000mR B. Incorrect. Plausible since this is the federal limit without condsideration of previous dose history. C. Incorrect. Plausible since the person has a dose history of 1 OOOmR and he will be verifying proper operation of valuable equipment. D. Correct. REFERENCES 91301-C, Emergency Exposure Guidelines VEGP learning objectives: LO-LP-40101-34 State the emergency TEDE limits for the following (SRO only): a. All activities

b. Protecting valuable property c. Life saving actions or protection of large populations Page: 50 of 56 4/2112009 KIA G2.3.4 Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions.

KIA MATCH ANALYSIS This question requires the applicant to determine type of activity and appropriate emergency dose limit. From that limit the applicant must also determine that emergency doses exclude previous occupational dose history. Question meets 1 OCFR55.43(b) criteria item #4 -Radiation hazards and contamination conditions that may occur during normal and abnormal situations, including maintenance activities and various contamination conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the person has a dose history of 1000mR B. Incorrect. Plausible since this is the federal limit without condsideration of previous dose history. C. Incorrect. Plausible since the person has a dose history of 1 OOOmR and he will be ( verifying proper operation of valuable equipment. D. Correct. REFERENCES 91301-C, Emergency Exposure Guidelines VEGP learning objectives: LO-LP-40101-34 State the emergency TEDE limits for the following (SRO only): a. All activities

b. Protecting valuable property c. Life saving actions or protection of large populations

{ Page: 50 of 56 4/2112009 Page: 1 The following plant conditions exist: A General Emergency has been declared based on the fission product matrix. An operator is to be sent into the auxiliary building to perform a discretionary damage assessment. The operator has received 1000 mrem TEDE this year, but has received NO dose during this event. Per EM-I04 "Operation ofthe Operational Support Center" what is the maximum dose the operator can receive for this entry? A. 4 rem B ..... 5 rem c. 9 rem D. 10 rem 2/11/2009 ( Page: 1 The following plant conditions exist: A General Emergency has been declared based on the fission product matrix. An operator is to be sent into the auxiliary building to perform a discretionary damage assessment. The operator has received 1000 mrem TEDE this year, but has received NO dose during this event. Per EM-I04 "Operation of the Operational Support Center" what is the maximum dose the operator can receive for this entry? A. 4 rem B..... 5 rem c. 9 rem D. 10 rem 2/11/2009 Procedure Number Rev T.E. 91301-C 9.1 Page Number EMERGENCY EXPOSURE GUIDELINES 7of13 TABLE 1 EMERGENCY EXPOSURE GUIDELINES NOTES a. Dose limits listed in this table apply to doses incurred over the duration of the emergency.

b. Dose to workers performing emergency services may be treated as a once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions.

c. Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value. Dose Limit (REM) Total Effective Dose Activity Condition t >25 Lifesaving or protection of large population Only on a voluntary basis to persons fully aware of the risks involved Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value. Printed April 8, 2009 at 12: 13 ( Approved By T.E. T nan Date Approved 01130/2004 91301-C 9.1 Page Number 70f13 EMERGENCY EXPOSURE GUIDELINES TABLE 1 EMERGENCY EXPOSURE GUIDELINES NOTES a. Dose limits listed in this table apply to doses incurred over the duration ofthe emergency.

b. Dose to workers performing emergency services may be treated as a once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions.

c. Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value. Dose Limit (REM) Total Effective Dose Activity Condition E t >25 Lifesaving or protection of large population Only on a voluntary basis to persons fully aware ofthe risks involved Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value. Printed Aplil8, 2009 at 12:13

23. G2.3.5 001l3INIA/RADIATION CONTROLIMEM

-2.9INEWIHL15/SRO/DS/TNT Initial conditions: -Unit tripped from full power due to an RCS LOCA -CNMT radiation monitors RE-005 & RE-006 indicate 8.1 E+6 mr/hr Using the attached Figures 1, 2, and 3 of 91 001-C, Emergency Classification, determine the appropriate emergency classification based on the radiation monitor readings. A. General Site Area C. Alert D. Unusual Event Page: 51 of 56 4/2112009 ( 23. G2.3.5 OOl/3/NIAlRADIATION CONTROLIMEM -2.9/NEWIHL15/SRO/DS/TNT Initial conditions: -Unit tripped from full power due to an RCS LOCA -CNMT radiation monitors RE-005 & RE-006 indicate 8.1 E+6 mr/hr Using the attached Figures 1, 2, and 3 of 91 001-C, Emergency Classification, determine the appropriate emergency classification based on the radiation monitor readings. A. General Site Area C. Alert D. Unusual Event Page: 51 of 56 412112009 G2.3.S Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring equipment, etc. KIA MATCH ANALYSIS This question requires the applicant to analyze radiation readings and make comparison to emergency plan criteria for determining appropriate emergency classification. Question meets 10CFR55.43(b) criteria item #4 -Radiation hazards and contamination conditions that may occur during normal and abnormal situations, including maintenance activities and various contamination conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the radiation monitors feed into each of the three fission product barrier decision trees. Applicant must know from memory which barriers and challenges apply to each emergency classification. B. Correct. C. Incorrect. Plausible because the loss of RCS barrier decision tree has the lowest value of the three decision trees and the fuel cladding barrier limit is within the same decade (1 E6). D. Incorrect. Plausible since one barrier is lost and the unusual event is the lowest emergency classification. REFERENCES 91001-C, Emergency Classification and Implementing Instructions VEGP learning objectives: LO-LP-40101-13: Given an emergency scenario, and the procedure, classify the emergency (SRO only). Page: 52 of 56 4/2112009 KIA G2.3.S Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring equipment, etc. KIA MATCH ANALYSIS This question requires the applicant to analyze radiation readings and make comparison to emergency plan criteria for determining appropriate emergency classificatio

n. Question meets 1 OCFR55.43(b) criteria item #4 -Radiation hazards and contamination conditions that may occur during normal and abnormal situations, including maintenance activities and various contamination conditions making this an SRO only question.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the radiation monitors feed into each of the three fission product barrier decision trees. Applicant must know from memory which barriers and challenges apply to each emergency classification. ( B. Correct. C. Incorrect. Plausible because the loss of RCS barrier decision tree has the lowest value of the three decision trees and the fuel cladding barrier limit is within the same decade (1 E6). D. Incorrect. Plausible since one barrier is lost and the unusual event is the lowest emergency classification. REFERENCES 91001-C, Emergency Classification and Implementing Instructions VEGP learning objectives: LO-LP-40101-13: Given an emergency scenario, and the procedure, classify the emergency (SRO only). Page: 52 of 56 4/2112009 Approved By ber Rev J.D. Williams 31 Date Approved 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS Page Number 10 of 1 REFERENCE USE-1 1. and 3. y OR CORE COOLING 1----11 (p.32) CSFST RED N LOSS of 2. Coolant Activity Sample> 300 ,...,Ci/gm Equivalent 1-131 ________ --1 OR I CLAD BARRIER I 5. Containment Radiation Monitors RE-005/006 > 6.0 E+6 mr/hr ____ _ OR ---------------. OR --, POTENTIAL

1. and 4. CORE COOLING y 1------11 (p. 32) CSFST ORANGE N LOSS of 1. y (p.32) OR I CLAD HEAT SINK BARRIER CSFST RED I N 7. Judgment:

Judgment by the ED that the Fuel Clad Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier (p.33) FIGURE 1 -FUEL CLADDING INTEGRITY (Modes 1, 2, 3 and 4 only) 8,2009 at 12:19 OR , ;""':", ,<">:*<""\\J7':::"*'*'C

< "'"' , Vogtle Electric Generatingj:t'

/"'"."",.<<< '<, < "',< Approved By J.D. Williams Date Approved 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS REFERENCE USE 1. and 3. y OR CORE COOLING 1----1 (p.32) CSFST RED N ... .... 2. Coolant Activity Sample> 300 )lCi/gm Equivalent I -131 _______ O_R--I LOSS of CLAD BARRIER 5. Containment Radiation Monitors RE-005/006 > 6.0 E+6 mr/hr (p.33) OR 1. and 4. y CORE COOLING CSFST E N 1. Y HEAT SINK CSFST RED N OR (p.32) (p.32) OR POTENTIAL t-----1 ..... 1 LOSS .... of CLAD BARRIER .... .... ------------------------------------_.

7. Judgment:

Judgment by the ED that the Fuel Clad Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier (p.33) FIGURE 1 -FUEL CLADDING INTEGRITY (Modes 1, 2, 3 and 4 only) Printed April 8, 2009 at 12:19 Page Number 10 of 108 ..... OR Approved By J.D. Williams Date Approved 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS

, REFERENCE USE 2. RCS Leak in progress AND RCS Subcooling is Less Than 24 of [38 ° F ADVERSE] (p.33) 3. SGTR Resulting in an SI Actuation (p.34) 4. Containment Radiation Monitors RE-005/006

> 2.0 E+4 mr/hr (p.34) 1. I : I (p. 33) OR HEAT SINK CSFST 1. I : I (p. 33) OR RCS INTEGRITY CSFST RED 2. NON-Isolable RCS leak (including SG tube leakage) GREATER THAN the Capacity of One Charging Pump in the normal charging mode (p.34) OR OR OR OR 5. Unexplained level rise in any of the following: containment sump, Reactor Coolant Drain Tank (RCDT), Waste Holdup Tank (WHT) (p.34) 6. Judgment: Judgment by the ED that the RCS Barrier is Lost or Potentially Lost. Consider conditions not addressed and the inability to determine the status of the RCS Barrier (p. 34) 31 Page Number LOSS of RCS BARRIER POTENTIAL LOSS of ReS BARRIER 11 of 108 OR FIGURE 2 -REACTOR COOLANT SYSTEM (RCS) INTEGRITY (Modes 1, 2, 3 and 4 only) ,2009 at 12:19 Approved By ,:< '/(" ,>> "" " """'<}/:,, <<7"" ;y::,"" ; Procedure Number Rev ;y: 't;gtiM<<:Electri'c Generating J.D. Williams "",< """ ", )'i\',,) 91001-C 31 ,,' ""c:" "'ii, ""At ",,'\& "W*,,,,,),",,,,, ' '" , ,;" "'" "" :," , ,';' "'f "ii,i '"i''''' i Date Approved Page Number 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS 11 of 108 I REFERENCE USE I 2. RCS Leak in progress AND RCS Subcooling is Less Than 24 of OR [38 ° F ADVERSE] (p.33) 3. SGTR Resulting in an SI Actuation (p.34) OR LOSS ... of JIll"" RCS .... 4. Containment Radiation Monitors RE-005/006 > 2.0 E+4 mr/hr (p.34) OR BARRIER -------------------------------------OR 1. Y OR HEAT SINK (p.33) POTENTIAL CSFST N LOSS ... JII"" of ... 1. Y OR RCS RCS INTEGRITY (p.33) BARRIER CSFST RED N 2. NON-Isolable RCS leak (including SG tube leakage) GREATER THAN the Capacity of One Charging Pump in the normal charging mode (p.34) OR 5. Unexplained level rise in any of the following: containment sump, Reactor Coolant Drain Tank (RCDT), Waste Holdup Tank (WHT) (p.34) 6. Judgment: Judgment by the ED that the RCS Barrier is Lost or Potentially Lost. Consider conditions not addressed and the inability to determine the status of the RCS Barrier (p. 34) FIGURE 2 -REACTOR COOLANT SYSTEM (RCS) INTEGRITY (Modes 1, 2, 3 and 4 only) Printed April 8, 2009 at 12:19 Approved J.D. Williams Date Approved 10/13/2008 EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS REFERENCE USE 2. Rapid Unexplained Containment Pressure Decrease Following Initial Pressure Increase (p.35) 2. Intersystem LOCA indicated by Containment Pressure or Sump Level Response Not Consistent With a Loss of Primary or Seconda_ry 4. Ruptured SG is also Faulted Outside of Containment (p. 36) 4. Primary-to-Secondary Leakage> 10 gpm AND a Non-Isolable Steam Release of Contaminated Secondary Coolant is Occurring to the Environment (p. 36) 5. Containment Isolation Valve(s) or Damper(s) are NOT Closed Resulting in a Direct Pathway to the Environment After Containment Isolation is Required (p. 37) OR OR I OR OR OR 7. Pathway to the environment exists based on VALID RE-2562C, RE-12444C, OR RE-12442C Alarms. (p.37) 1. CONTAINMENT CSFST RED y N (p.35) 2. CTMT CSFST ORANGE AND less than four CTMT fan coolers AND one train of CTMT sorav operable (p.35) 3. CORE COOLING CSFST RED OR ORANGE> 15 minutes AND RVLS FULL RANGE LEVEL <62% (p.37) 2. Containment Hydrogen Concentration> 6% (p.35) 2. Containment Pressure> 43 psig (p.35) 6. Containment Radiation Monitors RE-005/006 > 2.4 E+8 mr/hr (p.37) OR OR OR OR OR OR 8. Judgment: Judgment by the ED that the CNTMT Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier (p. 37) FIGURE 3 -CONTAINMENT INTEGRITY (Modes 1, 2, 3 and 4 only) April 8, 2009 at 12:19 31 Page Number 12 of 108 LOSS of CNTMT BARRIER OR POTENTIAL LOSS of CNTMT BARRIER * ,--, Approved By IL:';:)L', : f9;S; " ... :!?" iit::':%fW"."""80"'0":'S!!'>1§%t'fl t""""': :'. , i Proceaure Number Rev , . .', Plant'); '. , . J.D. Williams t<, . ',.' iE ),' :*t.:. ,if"", ),:, """, 9 "', , " "".",",:M%W,C:,%@"0!l""":,C""::",::,:,,,< ,,9,,;)) " Ala.", , :" , 91001-C 31 '" '.:/:. '"iii Date Approved EMERGENCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS Page Number 10113/2008 12 of 108 I REFERENCE USE I 2. Rapid Unexplained Containment Pressure Decrease Following Initial Pressure Increase (p.35) OR LOSS 2. Intersystem LOCA indicated by Containment Pressure or Sump Level Response Not Consistent OR of With a Loss of Primary or Secondary Coolant (p. 35) -+ .... 4. Ruptured SG is also Faulted Outside of Containment (p. 36) OR CNTMT ""'01 4. Primary-to-Secondary Leakage> 10 gpm AND a Non-Isolable Steam Release of Contaminated OR BARRIER Secondary Coolant is Occurring to the Environment (p. 36) 5. Containment Isolation Valve(s) or Damper(s) are NOT Closed Resulting in a Direct Pathway to the OR Environment After Containment Isolation is Required (p. 37) 7. Pathway to the environment exists based on VALID RE-2562C, RE-12444C, OR RE-12442C Alarms. (p.37) -------------------------------------OR 1. CONTAINMENT y OR CSFST RED N (p. 35) POTENTIAL OR LOSS of 2. CTMT CSFST AND less than four CTMT fan coolers AND one train of CTMT ... CNTMT spray operable (p.35) 3. CORE COOLING CSFST RED OR iGE > 15 minutes AND RVLS FULL RANGE LEVEL OR BARRIER <62% (p.37) 2. Containment Hydrogen Concentration> 6% (p.35) OR 2. Containment Pressure> 43 psig (p.35) OR 6. Containment Radiation Monitors RE-005/006 > 2.4 E+8 mr/hr (p.37) OR 8. Judgment: Judgment by the ED that the CNTMT Barrier is Lost or Potentially Lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier (p. 37) FIGURE 3 -CONTAINMENT INTEGRITY (Modes 1, 2, 3 and 4 only) Prrnted Aprrl 8, 2009 at 12: 19

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/ i.l; ord"-ST Z(J c..JassjjfD ... '1tia?1 ;hc.tiili Pttqre -4: Dinuwr should nOT "rlwit 1.intU 15 -tmnulq.7 cut ::fl-!.1uld as SOO;'i a.:: it b: d*ltJ.rnninV-."i that W:Q< tiwnaDJ! hffS 01' ( 24. G2.4.20 002/3INIA/EOPS/EPIPS/C/A -4.3INEWIHL15/SROIDS/TNT A loss of All AC power has occured and the crew is preparing to depressurize intact SGs to 300 psig per 191 OO-C, Loss of All AC Power. A NOTE immediately prior to the SG depressurization step states: "PRZR level may be lost and Reactor Vessel Upper Head voiding may occur due to depressurization of the SGs". Which of the following is correct regarding the operational implication of this note? A. This is the desired response in order to refill the PRZR and regain control of RCS pressure once the PRZR heaters are covered with water. B. SG depressurization should be stopped if RV head voiding occurs to prevent loss of RCS subcooling and loss of core heat removal. SG depressurization reduces RCS pressure to inject the SI accumulators to help maintain RCS inventory. SG despressurization should continue even if RV upper head voiding is occurring. D. This is the desired response to refill the PRZR and allow restarting an RCP to established forced convective cooling of the reactor core. The despressurization would only be stopped if PRZR level exceeded 75%. Page: 53 of 56 4/2112009 ( ( 24. G2.4.20 002/3INIAlEOPS/EPIPS/CIA -4.3INEWIHL15/SROIDS/TNT A loss of All AC power has occured and the crew is preparing to depressurize intact SGs to 300 psig per 191 OO-C, Loss of All AC Power. A NOTE immediately prior to the SG depressurization step states: "PRZR level may be lost and Reactor Vessel Upper Head voiding may occur due to depressurization of the SGs". Which of the following is correct regarding the operational implication of this note? A. This is the desired response in order to refill the PRZR and regain control of RCS pressure once the PRZR heaters are covered with water. B. SG depressurization should be stopped if RV head voiding occurs to prevent loss of RCS subcooling and loss of core heat removal. C!' SG depressurization reduces RCS pressure to inject the SI accumulators to help maintain RCS inventory. SG despressurization should continue even if RV upper head voiding is occurring. D. This is the desired response to refill the PRZR and allow restarting an RCP to established forced convective cooling of the reactor core. The despressurization would only be stopped if PRZR level exceeded 75%. Page: 53 of 56 4/2112009 G2.4.20 Emergency Procedures/Plan Knowledge of operational implications of EOP warnings, cautions, and notes. K/A MATCH ANALYSIS This question involves assessing plant conditions and recalling what strategy is written into a plant procedure and when it is required. This question cannot be answered by knowing systems knowledge, immediate actions, or EOP/AOP entry conditions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible since this is the expected response in post LOCA cooldown. B. Incorrect. Plausible due to the concern for loss of subcooling in RV head during natural circulation conditions. However, SI accumulator injection is a higher priority under these conditions. C. Correct. D. Incorrect. Plausible because a common PRZR depressurization setpoint in the EOP network is > 75% PRZR level on a SGTR. REFERENCES 19100-C Rev 32.3, step 29. V-LO-HO-37031-001-04-C, rev 4, pages 26 & 27 VEGP learning objectives: LO-LP-37031-09: Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement. Page: 54 of 56 4/2112009 ( ( K/A G2.4.20 Emergency Procedures/Plan Knowledge of operational implications of EOP warnings, cautions, and notes. K/A MATCH ANALYSIS This question involves assessing plant conditions and recalling what strategy is written into a plant procedure and when it is required. This question cannot be answered by knowing systems knowledge, immediate actions, or EOP/AOP entry conditions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible since this is the expected response in post LOCA cooldown. B. Incorrect. Plausible due to the concern for loss of subcooling in RV head during natural circulation conditions. However, SI accumulator injection is a higher priority under these conditions. C. Correct. D. Incorrect. Plausible because a common PRZR depressurization setpoint in the EOP network is > 75% PRZR level on a SGTR. REFERENCES 19100-C Rev 32.3, step 29. V-LO-HO-37031-001-04-C, rev 4, pages 26 & 27 VEGP learning objectives: LO-LP-37031-09: Given a NOTE or CAUTION statement from the EOP, state the bases for that NOTE or CAUTION statement. Page: 54 of 56 4121/2009 LO-HO-37031-001-04-C: Loss Of All AC Power STEP: Depressurize intact SGs to 300 psig. PURPOSE: To depressurize the intact steam generators BASIS: Step depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. Steam generators should be depressurized to maximize delivery (into the RCS) of the water contained in the SI accumulators while minimizing delivery of nitrogen. Maintaining steam generator pressures above a value that prevents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation. A minimum steam generator pressure limit of 200 psig is set to preclude significant nitrogen injection into the RCS. Once depressurization is initiated, maintenance of a specified rate is not critical. The depressurization rate should be sufficiently fast to expeditiously reduce SG pressures, but not so fast that SG pressures cannot be controlled. It is important that the depressurization not reduce SG pressures in an uncontrolled manner that undershoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS. During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. This step requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated. If level is not in the narrow range in at least one SG, the RNO instructs the operator to maintain maximum AFW flow until range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via this step. During SG depressurization, AFW flow may have to be increased to maintain the required SG narrow range level. Full AFW flow should be established to any SG in which level drops out of the narrow range. SG narrow range level should be maintained greater than 10% (32% for adverse containment) in at least one intact SG. If level cannot be maintained, SG depressurization should be stopped until level is restored in at least one SG. To inform the operator of the importance of maintaining at least one intact SG narrow range level above the top of the U-tubes during depressurization BASIS: During the rapid depressurization performed in this step, SG level could drop out of the narrow range resulting in a loss of adequate heat sink. If this situation occurs, the depressurization should be stopped and AFW flow reestablished 1-27 ( LO-HO-37031-001-04-C: Loss Of All AC Power STEP: Depressurize intact SGs to 300 psig. PURPOSE: To depressurize the intact steam generators BASIS: Step depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss. Steam generators should be depressurized to maximize delivery (into the RCS) of the water contained in the SI accumulators while minimizing delivery of nitrogen. Maintaining steam generator pressures above a value that prevents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation. A minimum steam generator pressure limit of 200 psig is set to preclude significant nitrogen injection into the RCS. Once depressurization is initiated, maintenance of a specified rate is not critical. The depressurization rate should be sufficiently fast to expeditiously reduce SG pressures, but not so fast that SG pressures cannot be controlled. It is important that the depressurization not reduce SG pressures in an uncontrolled manner that undershoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS. During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. This step requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated. If level is not in the narrow range in at least one SG, the RNO instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via this step. During SG depressurization, AFW flow may have to be increased to maintain the required SG narrow range level. Full AFW flow should be established to any SG in which level drops out of the narrow range. SG narrow range level should be maintained greater than 10% (32% for adverse containment) in at least one intact SG. If level cannot be maintained, SG depressurization should be stopped until level is restored in at least one SG. To inform the operator of the importance of maintaining at least one intact SG narrow range level above the top of the U-tubes during depressurization BASIS: During the rapid depressurization performed in this step, SG level could drop out of the narrow range resulting in a loss of adequate heat sink. If this situation occurs, the depressurization should be stopped and AFW flow reestablished 1-27 Procedure Number Rev 19100-C 32.3 Date Approved 20 of 46 Page Number ECA-O.O LOSS OF ALL AC POWER 5/16/08 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

The SGs should be depressurized at maximum rate (within the capacity of the TDAFW Pump) to minimize RCS inventory loss.
PRZR level may be lost and Reactor Vessel Upper Head voiding may occur due to depressurization of the SGs. Depressurization should not be stopped to prevent these occurrences.
29. Depressurize intact SGs to 300 psig: a. Check SG NR levels -GREATER THAN 10% [32% ADVERSE] IN AT LEAST ONE SG. a. Perform the following:

_1) IF all SG NR levels less than 10% [32% ADVERSE], THEN maintain maximum TDAFW flow. _2) WHEN NR level in at least one SG greater than 10% [32% ADVERSE], THEN go to Step 29.b. Go to Step 33. o Step 29 continued on next page ( ( Approved By J. B. Stanley Date Approved ECA-O.O LOSS OF ALL AC POWER Page Number 20 of 46 5/16/08 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

The SGs should be depressurized at maximum rate (within the capacity of the TDAFW Pump) to minimize RCS inventory loss.
PRZR level may be lost and Reactor Vessel Upper Head voiding may occur due to depressurization of the SGs. Depressurization should not be stopped to prevent these occurrences.
29. Depressurize intact SGs to 300 psig: a. Check SG NR levels -GREATER THAN 10% [32% ADVERSE] IN AT LEAST ONE SG. a. Perform the following:

_1) IF all SG NR levels less than 10% [32% ADVERSE], THEN maintain maximum TDAFW flow. _2) WHEN NR level in at least one SG greater than 10% [32% ADVERSE], THEN go to Step 29.b. Go to Step 33. o Step 29 continued on next page Printed April 20, 2009 at 15:36 LO-HO-37031-001-04-C: Loss Of All AC Power STEP: Check DC Bus Loads PURPOSE: To conserve dc power supply by shedding non-essential dc loads from the dc busses as soon as practical BASIS: Following loss of all ac power, the station batteries are the only source of electrical power. The station batteries supply the dc busses and the ac vital instrument busses. Since ac emergency power is not available to charge the station batteries, battery power supply must be conserved to permit monitoring and control of the plant until ac power can be restored. Attachment A was prepared to prioritize the shedding of dc loads in order to conserve and prolong the station battery power supply. The intent of load shedding is to remove all large non-essential loads as soon as practical, consistent with preventing damage to plant equipment. Consideration was also given to the priority of shedding additional loads in case ac power cannot be restored within the projected life of the station batteries. NOTE: The SGs should be depressurized at maximum rate to minimize RCS inventory loss. PURPOSE: To inform the operator of the desired rate for depressurization of steam generators BASIS: The intact steam generators should be depressurized as quickly as possible, to minimize RCS inventory loss, but within the constraint of controllability. Controllability is required to ensure that steam generator pressures do not undershoot the specified limit. For plants that can control the secondary depressurization from the control room, maximum rate means steam generator ARVs full open. For plants that must control the secondary depressurization by local actions (like Vogtle), maximum rate must be determined by the control room and local operators based on plant conditions and available communications. A slower rate is acceptable for locally controlled secondary depressurization. NOTE: PRZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of SGs. Depressurization should not be stopped to prevent these occurrences. PURPOSE: To inform the operator of possible reactor vessel upper head voiding during steam generator depressurization BASIS: Loss of pressurizer level and reactor vessel upper head voiding may result from the rapid depressurization of the intact steam generators. Such a condition is anticipated and should not interfere with operator actions to depressurize the steam generators to reduce RCS pressure and temperature and to minimize RCS inventory loss out of the RCP seals. 1-26 ( LO-HO-37031-001-04-C: Loss Of All AC Power STEP: Check DC Bus Loads PURPOSE: To conserve dc power supply by shedding non-essential dc loads from the dc busses as soon as practical BASIS: Following loss of all ac power, the station batteries are the only source of electrical power. The station batteries supply the dc busses and the ac vital instrument busses. Since ac emergency power is not available to charge the station batteries, battery power supply must be conserved to permit monitoring and control of the plant until ac power can be restored. Attachment A was prepared to prioritize the shedding of dc loads in order to conserve and prolong the station battery power supply. The intent of load shedding is to remove all large non-essential loads as soon as practical, consistent with preventing damage to plant equipment. Consideration was also given to the priority of shedding additional loads in case ac power cannot be restored within the projected life of the station batteries. NOTE: The SGs should be depressurized at maximum rate to minimize RCS inventory loss. PURPOSE: To inform the operator of the desired rate for depressurization of steam generators BASIS: The intact steam generators should be depressurized as quickly as possible, to minimize RCS inventory loss, but within the constraint of controllability. Controllability is required to ensure that steam generator pressures do not undershoot the specified limit. For plants that can control the secondary depressurization from the control room, maximum rate means steam generator ARVs full open. For plants that must control the secondary depressurization by local actions (like Vogtle), maximum rate must be determined by the control room and local operators based on plant conditions and available communications. A slower rate is acceptable for locally controlled secondary depressurization. NOTE: PRZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of SGs. Depressurization should not be stopped to prevent these occurrences. PURPOSE: To inform the operator of possible reactor vessel upper head voiding during steam generator depressurization BASIS: Loss of pressurizer level and reactor vessel upper head voiding may result from the rapid depressurization of the intact steam generators. Such a condition is anticipated and should not interfere with operator actions to depressurize the steam generators to reduce RCS pressure and temperature and to minimize RCS inventory loss out of the RCP seals. 1-26

25. G2.4.47 001l3/N/A/EOPS/EPIPS/C/A

-4.2/M -WOLF CREEK 20071HL15/SROIDS/TNT Initial conditions: -The unit tripped from 100% power -Loss of a" off-site power occurred -Both EDG's are powering their loads Current conditions: -RCS pressure 2235 psig and stable -RCS loop hot leg temperatures 602°F and stable -RCS loop cold leg temperatures 588°F and stable -Core exit TCs are 610°F and stable -SG pressures are 1085 psig and stable Which of the following choices is correct? A. Natural circulation is established with the steam dumps. RCS cooldown may not proceed while borating the RCS. B. Natural circulation is not established. Increase rate of dumping steam with the steam dumps while borating the RCS. C. Natural circulation is established with the SG ARV's. RCS cooldown may not proceed while borating the RCS. D'!" Natural circulation is not established. Increase rate of dumping steam with the SG ARV's while borating the RCS. Page: 55 of 56 4/2112009

25. G2.4.47 001l3/NIAlEOPS/EPIPS/CIA

-4.2/M -WOLF CREEK 2007/HL15/SRO/DS/TNT Initial conditions: -The unit tripped from 100% power -Loss of all off-site power occurred -Both EOG's are powering their loads Current conditions: -RCS pressure 2235 psig and stable -RCS loop hot leg temperatures 602°F and stable -RCS loop cold leg temperatures 588°F and stable -Core exit TCs are 610°F and stable -SG pressures are 1085 psig and stable Which of the following choices is correct? A. Natural circulation is established with the steam dumps. RCS cooldown may not proceed while borating the RCS. B. Natural circulation is not established. ( Increase rate of dumping steam with the steam dumps while borating the RCS. C. Natural circulation is established with the SG ARV's. RCS cooldown may not proceed while borating the RCS. O!' Natural circulation is not established. Increase rate of dumping steam with the SG ARV's while borating the RCS. ( Page: 55 of 56 4/2112009 G2.4.47 Emergency Procedures/Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. KIA MATCH ANALYSIS Question requires assessment of plant conditions and then prescribing what steps are required with which to proceed. This question cannot be answered by knowing systems knowledge, immediate actions, or EOP/AOP entry conditions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible since the SG pressure given is the normal control point for the steam dumps all natural circulation parameters are stable which is one of the criteria for meeting natural circulation. Additionally, cooldown is not allowed until boration is completed. B. Incorrect. Plausible since natural circulation is not established and the correct action is to increase rate of dumping steam to establish natural circulation. C.lncorrect. Plausible since all natural circulation parameters are stable which is one of the criteria for meeting natural circulation and cooldown may not proceed until the RCS is borated to cold shutdown conditions. D. Correct. REFERENCES EOP 19001-C, Reactor Trip Response, Attachment B EOP 19002-C, Natural Circulation Cooldown VEGP learning objectives: LO-LP-37012-13: State the indications used to monitor natural circulation cooldown. LO-LP-37012-14: State the initial steps that must be taken to prepare the plant for natural circulation cooldown. Include the reason for each. Page: 56 of 56 4/2112009 ( G2.4.47 Emergency Procedures/Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. K/A MATCH ANALYSIS Question requires assessment of plant conditions and then prescribing what steps are required with which to proceed. This question cannot be answered by knowing systems knowledge, immediate actions, or EOP/AOP entry conditions. Question meets 1 OCFR55.43(b) criteria item #5 -Assessment of facility conditions and selection of procedures during normal, abnormal, and emergency conditions making this an SRO only question. ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible since the SG pressure given is the normal control point for the steam dumps all natural circulation parameters are stable which is one of the criteria for meeting natural circulation. Additionally, cool down is not allowed until boration is completed. B. Incorrect. Plausible since natural circulation is not established and the correct action ( is to increase rate of dumping steam to establish natural circulation. C. Incorrect. Plausible since all natural circulation parameters are stable which is one of the criteria for meeting natural circulation and cooldown may not proceed until the RCS is borated to cold shutdown conditions. D. Correct. REFERENCES EOP 19001-C, Reactor Trip Response, Attachment B EOP 19002-C, Natural Circulation Cooldown VEGP learning objectives: LO-LP-37012-13: State the indications used to monitor natural circulation cooldown. LO-LP-37012-14: State the initial steps that must be taken to prepare the plant for natural circulation cooldown. Include the reason for each. Page: 56 of 56 4/2112009 ES-401 Sample Written Examination Question Worksheet Examination Outline Cross-reference: Level Tier # Group # KIA # Importance Rating RO 01 Form ES-401-5 SRO 2 2 3.7 Emergency Procedures I Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Proposed Question: SR092 The plant was operating at 98% power when a loss of off-site power occurred. Twenty minutes later, the following plant conditions exist:

RCS pressure is 2235 psig and slowly increasing.
RCS Loop T HOT is 602 a F in all 4 loops and trending down slowly.
RCS Loop T COLD is 560 a F in all 4 loops and stable.
Core exit TCs indicate approximately 61 oaF and stable.
Pressure is approximately 1130 psig in all steam generators.

Which ONE (1) of the following describes the current plant conditions and action that will be directed? A. Natural circulation does not exist. Heat removal may be established by opening the condenser steam dumps in accordance with EMG E-O, REACTOR TRIP OR SAFETY INJECTION. B. Heat removal is being maintained by condenser steam dumps. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. C. Natural circulation does not exist. Heat removal may be established by opening the atmospheric relief valves in accordance with EMG E-O, REACTOR TRIP OR SAFETY INJECTION. D. Heat removal is being maintained by atmospheric relief valves. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. Proposed Answer: Explanation (Optional): NUREG-1 021, Revision 9 D. Heat removal is being maintained by atmospheric relief valves. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. 183 ES-401 Sample Written Examination Question Worksheet Examination Outline Cross-reference: RO Form ES-401-5 SRO 2 2 Level Tier # Group # KJA# 016 G2.4.47 Importance Rating 3.7 Emergency Procedures / Plan Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Proposed Question: SR092 The plant was operating at 98% power when a loss of off-site power occurred. Twenty minutes later, the following plant conditions exist:

RCS pressure is 2235 psig and slowly increasing.
RCS Loop T HOT is 602°F in all 4 loops and trending down slowly.
RCS Loop T COLD is 560°F in all 4 loops and stable.
Core exit TCs indicate approximately 610°F and stable.
Pressure is approximately 1130 psig in all steam generators.

Which ONE (1) of the following describes the current plant conditions and action that will be directed? A. Natural circulation does not exist. Heat removal may be established by opening the condenser steam dumps in accordance with EMG E-O, REACTOR TRIP OR SAFETY INJECTION. B. Heat removal is being maintained by condenser steam dumps. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. C. Natural circulation does not exist. Heat removal may be established by opening the atmospheric relief valves in accordance with EMG E-O, REACTOR TRIP OR SAFETY INJECTION. D. Heat removal is being maintained by atmospheric relief valves. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. Proposed Answer: Explanation (Optional): NUREG-1021, Revision 9 D. Heat removal is being maintained by atmospheric relief valves. Verify that Natural circulation exists in accordance with EMG ES-02, REACTOR TRIP RESPONSE. 183 Approved By J.B. Stanley Date Approved 7/22/2008 ES -0.1 REACTOR TRIP RESPONSE ATTACHMENT B VERIFICATION OF NATURAL CIRCULATION

1. The following conditions support or indicate natural circulation flow:

RCS subcooling

-GREATER THAN 24°F.

SG pressure -STABLE OR LOWERING.
RCS WR hot leg temperatures

-STABLE OR LOWERING.

Core exit TCs -STABLE OR LOWERING.

Procedure Number Rev 19001-C 31 Page Number 22 of 25 Sheet 1 of 1

RCS WR cold leg temperatures

-AT SATURATION TEMPERATURE FOR SG PRESSURE. o END OF ATTACHMENT B Printed April 8, 2009 at 12:46 ( ( Approved By J.B. Stanley Date Approved 7/22/2008 ES -0.1 REACTOR TRIP RESPONSE ATTACHMENT B VERIFICATION OF NATURAL CIRCULATION

1. The following conditions support or indicate natural circulation flow:

RCS subcooling

-GREATER THAN 24°F.

SG pressure -STABLE OR LOWERING.
RCS WR hot leg temperatures

-STABLE OR LOWERING.

Core exit TCs -STABLE OR LOWERING.

Procedure Number Rev 19001-C 31 Page Number 22 of 25 Sheet 1 of 1

RCS WR cold leg temperatures

-AT SATURATION TEMPERATURE FOR SG PRESSURE. o END OF ATTACHMENT B Printed April 8, 2009 at 12:46 Approved By J.B. Stanley Page Number Date Approved ES-O.2 NATURAL CIRCULATION COOLDOWN 30f22 7/22/2008 ACTION/EXPECTED RESPONSE 1. Initiate the Continuous Actions and Foldout Page. *2. IF SI actuation occurs during this procedure, THEN go to 19003-C, ES-O.3 NATURAL CIRCULATION COOL DOWN WITH VOID IN VESSEL (WITH RVLlS) 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

3. Try to restart an RCP: a. Start an RCP using ATTACHMENT A. (RCP 4 or RCP 1 preferred)

b. Go to appropriate plant procedure.

4. Verify all available CRDM fans -RUNNING. *5. Borate RCS:
Borate RCS to maximum boron concentration for cooldown range 55rF to 68°F using PTDB tables 1.3.4-T1 and 1.3.4-T2.
Determine the required boron concentration for xenon free cold shutdown conditions by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

RESPONSE NOT OBTAINED a. WHEN an RCP has been started, THEN go to appropriate plant procedure. Go to Step 4. ( ( Approved By J.B. Stanley ES-O.2 NATURAL CIRCULATION COOLDOWN Page Number 30f22 Date Approved 7/22/2008 ACTION/EXPECTED RESPONSE 1. Initiate the Continuous Actions and -*2. *3. *4. -*5. Foldout Page. IF SI actuation occurs during this procedure, THEN go to 19003-C, ES-O.3 NATURAL CIRCULATION COOL DOWN WITH VOID IN VESSEL (WITH RVLlS) 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION. Try to restart an RCP: a. Start an RCP using -ATTACHMENT A. (RCP 4 or RCP 1 preferred)

b. Go to appropriate plant -procedure.

Verify all available CRDM fans -RUNNING. Borate RCS:

Borate RCS to maximum boron -concentration for cooldown range 55rF to 68°F using PTDB tables 1.3.4-T1 and 1.3.4-T2. * -Determine the required boron concentration for xenon free cold shutdown conditions by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

0 Pnnted Apnl8, 2009 at 12:47 RESPONSE NOT OBTAINED a. WHEN an RCP has been -started, THEN go to appropriate plant procedure. Go to Step 4. Approved By J.B. Stanley Date Approved 7/22/2008 Rev 20.1 Page Number ES-O.2 NATURAL CIRCULATION COOLDOWN 4of22 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Due to lack of mixing, the pressurizer boron concentration is not expected to change and is monitored for information only. 6. Check RCS boron concentration greater than required boron concentration for xenon free cold shutdown by directing Chemistry to sample the following:

RCS loops 1 and 3 Hot Legs
Letdown
PRZR liquid (info only) 6. Return to Step 5. IF sampling capability is NOT available, THEN perform the following:
Verify boration determined in Step 5 is completed.
Estimate boron concentration based on the last known concentration and any known dilutions since that time.
Go to Step 7 while continuing attempts to verify RCS boron concentration.

6 ( Approved By J.B. Stanley Date Approved 7/22/2008 Procedure Number Rev 19002-C 20.1 40f22 Page Number ES-O.2 NATURAL CIRCULATION COOLDOWN ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Due to lack of mixing, the pressurizer boron concentration is not expected to change and is monitored for information only. 6. Check RCS boron concentration greater than required boron concentration for xenon free cold shutdown by directing Chemistry to sample the following:

RCS loops 1 and 3 Hot Legs
Letdown
PRZR liquid (info only) 6. Return to Step 5. IF sampling capability is NOT available, THEN perform the following:
Verify boration determined in Step 5 is completed.
Estimate boron concentration based on the ( last known concentration and any known dilutions since that time. Printed April 8, 2009 at 12:52
Go to Step 7 while continuing attempts to verify RCS boron concentration.

6 Approved By J.B. Stanley Rev 20.1 Page Number Date Approved ES-O.2 NATURAL CIRCULATION COOLDOWN 7of22 7/22/2008 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED *9. Initiate ReS cooldown to cold shutdown:

a. Check RCS boron concentration

a. Return to Step 6. greater than required boron concentration for xenon free cold shutdown.

b. Maintain cooldown rate in RCS -Cold Legs -LESS THAN 50°F/Hr. -c. Dump steam to Condenser using c. Use SG ARVs. Steam Dumps. d. Maintain SG NR levels -AT 65%. e. Check RCS cooldown at 15 minute intervals.

f. Maintain RCS temperature and pressure -WITHIN LIMITS OF TECHNICAL SPECIFICATION LCO 3.4.3 (PTLR): Use 60°F/HR curve and RCS Cold Leg temperature. -g. Perform other appropriate actions required to take the unit to cold* shutdown by initiating 12006-C, RCS COOLDOWN TO COLD SHUTDOWN.

10. Check RCS WR Hot Leg 10. Return to Step 8. temperatures

-LESS THAN 550°F. ( ( Approved By J.B. Stanley Page Number Date Approved ES-O.2 NATURAL CIRCULATION COOLDOWN 7 of 22 7/22/2008 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED *9. Initiate ReS cool down to cold shutdown:

a. Check RCS boron concentration

-a. Return to Step 6. greater than required boron concentration for xenon free cold shutdown.

b. Maintain cooldown rate in RCS -Cold Legs -LESS THAN 50°F/Hr. c. Dump steam to Condenser using -c. Use SG ARVs. Steam Dumps. d. Maintain SG NR levels -AT 65%. -e. Check RCS cooldown at 15 -minute intervals.

f. Maintain RCS temperature and pressure -WITHIN LIMITS OF TECHNICAL SPECIFICATION LCO 3.4.3 (PTLR): Use 60°F/HR curve and -RCS Cold Leg temperature. -g. Perform other appropriate actions required to take the unit to cold* shutdown by initiating 12006-C, RCS COOLDOWN TO COLD SHUTDOWN.

-10. Check RCS WR Hot Leg 10. Return to Step 8. temperatures -LESS THAN 550°F. Printed April 8, 2009 at 12:47}}

ML093280270

Contents

Text

3.4 REACTOR

3.3.1 ALLOWABLE

3.3.1 ALLOWABLE

3.3.1 COMPLETION

3.3.1 COMPLETION

3.3.1. Check

3.3.1 ALLOWABLE

3.3.1 ALLOWABLE

3.7 NOTES

REFERENCES:

REFERENCES:

3.8.9 REQUIRED

3.8.9 REQUIRED

3.7 PLANT

4.0 SUBSEQUENT

5.0 COMPENSATORY

REFERENCES:

4.0 SUBSEQUENT

5.0 COMPENSATORY

REFERENCES:

1.0 PURPOSE

2.1 LIMITING

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ML093280270

Text

3.4 REACTOR

3.3.1 ALLOWABLE

3.3.1 ALLOWABLE

3.3.1 COMPLETION

3.3.1 COMPLETION

3.3.1. Check

3.3.1 ALLOWABLE

3.3.1 ALLOWABLE

3.7 NOTES

REFERENCES:

REFERENCES:

3.8.9 REQUIRED

3.8.9 REQUIRED

3.7 PLANT

4.0 SUBSEQUENT

5.0 COMPENSATORY

REFERENCES:

4.0 SUBSEQUENT

5.0 COMPENSATORY

REFERENCES:

1.0 PURPOSE

2.1 LIMITING

1.0 PURPOSE

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