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2014 Palisades Nuclear Plant Initial License Examination Administrative Files
	ML14309A139
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Site:	Palisades
Issue date:	09/12/2014
From:	David Reeser; NRC/RGN-III/DRS/OLB
To:	; Entergy Nuclear Operations
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2014 PALISADES NUCLEAR PLANT INITIAL LICENSE EXAMINATION ADMINISTRATIVE FILES

ES-403 Written Examination Grading Form ES-403-1 Quality Checklist Facility: Palisades Date of Exam: Se12tember 151h, 2014 Exam Level: RO ~ SRO ~

Initials Item Description a b c

1. Clean answer sheets copied before grading JJ ~ oz.,

2. Answer key changes and question deletions justified and documented d> \~ ~

3. Applicants' scores checked for addition errors (reviewers spot check >25% of examinations) sb ~ JZ,

4. Grading for all borderline cases (80% +/- 2% overall and 70 or 80, as applicable,.:!: 4% on the SRO-only) reviewed in detail \,~ :17-

5. All other failing examinations checked to ensure that grades are justified clJ t'>*~

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6. Performance on missed questions checked for training deficiencies and wording problems; evaluate validity ~ i)Z.,

of questions missed by half or more of the applicants I S;gnature Date

a. Author Steve Botimer I f.knt/A J (liJ (112014

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b. Facility Reviewer(*) Brett Baker I f-..DV~ t!>~f2..Lf2014

c. NRC Chief Examiner(*) l)...\)',J hl. r<ees0r /O. . . C _k ).fo" ~ """'""-- to/*~P/tV

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d. NRC Supervisor(*) /'~_,&. ~ 1fz~/;tf_

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(*) The facility reviewer's signature is not applicable for examinations graded by the NRC; two independent NRC reviews are required.

ES-403, Page !1 of 5

~Entergy Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043 Jeffrey A. Hardy Licensing Manager PNP 2014-089 September 22, 2014 NUREG 1021 Regional Administrator U.S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 ATTN: Janet Kweiser

Subject:

Initial License Examination Comments Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir:

In accordance with NUREG 1021, ES-402, Entergy Nuclear Operations is submitting comments on the initial license examination administered at the Palisades Nuclear Plant during September 2014. Attachment 1 contains the required information.

Please contact Steve Botimer at (269) 764-2975 if you have any questions regarding this submittal.

This letter contains no new commitments and no revisions to existing commitments.

Sincerely, */1 \.d

/~-~t-r JAH/bed : Palisades Initial License Examination Comments CC David Reeser, Region Ill, USNRC Project Manager, Palisades, USNRC (w/o attachment)

Resident Inspector, Palisades, USNRC (w/o attachment)

Document Control Desk, USNRC (w/o attachment)

RECEIVED ScP 2 41014

ATTACHMENT 1 PALISADES INITIAL LICENSE EXAMINATION COMMENTS

Question #81 Comment Per the exam key, (b) is the correct answer, however, both (a) and/or (b) should be the correct answers.

Question: .';

The Plant has just entered MODE 3 for a forced outage to repair a Main Condenser Vacuum leak when the following occurs:

A loss of Left Train 125V DC Bus (more specifically ED-10R and ED-10L) occurs The Control Room Supervisor directs entry into ...

a. AOP~18 "Loss of Left Train DC power" only

b. AOP-18, "Loss of Left Train DC Power" and EOP -9.0, "Functional Recovery Procedure."

c. AOP-18 "Loss of Left Train DC Power," AOP -13 Loss of Preferred AC Bus EY-20," and ,*

AOP-15 "Loss of Preferred AC Bus EY-40."

d. EOP-9.0, "Functional Recovery Procedure" only Summary: Based on the initial condition of MODE 3, EOP-9.0 is not procedurally required to be entered. In this scenario, EOP-9.0 does not provide any additional actions not already covered in AOP-18 to address this event and stabilize the plant. Therefore the CRS may elect to direct entry into AOP-18 only.

Basis:

AOP-18 does not require EOP-9.0 entry EOP-9.0 entry is optional based on the discretion of the Control Room Supervisor and the Shift Manager. Therefore, the CRS may opt to enter AOP-18 only OR enter both AOP-18 and EOP-9.0.

AOP-18, Section 6.0, Operator Actions, Step 2 states, "Refer to EOP-9.. 0 'Functional Recovery Procedure' for lower mode entry." The scenario provided by this question is specifically addressed in the AOP-18 Basis document as follows (emphasis added):

This is a branching step to EOP-9.0. Transition to EOP-9.0 is required following a reactor trip and event diagnosis in EOP-1.0. EOP-9.0 entry for all other plant conditions is discretionary. In most cases, unless Shutdown Cooling is in service, EOP-9.0 entry is highly recommended.

AOP-18 intentionally uses the language of "REFER TO" rather than "ENTER" to allow the CRS discretion, based on plant conditions, the use of EOP-9.0 to combat the event.

EOP-9.0 entry criteria not met A review of the EOP-9.0 entry criteria reveal entry is not required (emphasis added):

2.0 ENTRY CONDITIONS

1. EOP-1.0, "Standard Post Trip Actions," has been performed.

The event initiated from a lower mode when the Shutdown Cooling System is NOT initially in service.

2. ANY of the following conditions may be present:

a. A Reactor trip with unusual concurrent symptoms and diagnosis of one event NOT immediately apparent.

b. Any conditions/symptoms which a licensed operator considers serious and for which other Emergency/Off-Normal Procedures can NOT be identified.

c. Actions from an in-use Optimal Recovery EOP do NOT result in acceptance criteria for in-use Optimal Recovery EOP Safety Function Status Check Sheet being satisfied.

d. An Optimal Recovery EOP step directs implementation of EOP-9.0, "Functional Recovery Procedure."

No reactor trip on this event, therefore no EOP-1.0 entry or subsequent EOP entry is required.

While the event was initiated from a lower mode without shutdown cooling in service, the event by itself is not sufficient to require use of EOP-9.0. There is only one unique eventin progress which has been clearly diagnosed with guidance from an applicable AOP. From the EOP-9.0 basis document, section 1.0 Introduction (emphasis added):

Entry conditions are chosen to identify those conditions which will necessitate implementation of the FRP. Following the performance of the SPTAs for events initiated during Power Operations or Hot Standby with the reactor critical, or from lower modes for which the FRP entry conditions are met, the operator may not be able to diagnose one unique event taking place. This could happen if more than one event is taking place (multiple casualties) or a condition exists for which abnormal or emergency guidance cannot be identified. During the course of the event, actions taken in an ORP may not satisfy the Safety Function Status Check acceptance criteria. Also, actions taken in an AOP (if entering from a lower mode) may not be adequately responding to mitigate the consequences of the event. Implementation of the safety function based FRP would then be evaluated.

There is an in-use Abnormal Operating Procedure identified which will address the condition.

Use of AOP-18 will address the loss of two Preferred AC Busses. Therefore a success path exists to ensure all applicable Safety Functions will be satisfied.

AOP-18 addresses all applicable actions taken in EOP-9.0 The highest priority jeopardized Safety Function is MVAE-DC, therefore MVAE-DC-1 would be the entry point for implementing Operator Actions in EOP-9.0, Success path MVAE-DC-1, step 6 addresses restoration of all available preferred AC buses (see attached).

AOP-18, Section 6.0, Step 24 (and 24.1) addresses restoration of available preferred AC buses (see attached).

AOP-18 was written specifically to stabilize the plant following this event. Therefore all plant actions related to this event are addressed in AOP-18 and no further actions taken in EOP-9.0 are required. From the AOP-18 basis document:

The need for a procedure dealing with a complete loss of one train of 125V DC became evident during the automatic reactor trip that occurred on [September 25, 2011].

AOP-18, section 6.0, step 28 states, "PERFORM EOP-9.0, Attachment 1, "Safety Function Status Check Sheet." From the basis document, "Intent of this step is to validate, using EOP-9.0 SFSCs, that all safety functions are being met." Following restoration of one preferred AC bus to the bypass regulator performed in step 24, and no other events in progress, all applicable Safety Functions will be met. This is further evidence that entry into EOP-9.0, under these circumstances, is not required and will provide no additional assistance in addressing this event.

Facility Position: The facility agrees with Mr. DeBusscher that for Question #81 answers (a) and/or (b) should be accepted as correct answers based on the discussion above. Abnormal Operating Procedure AOP-18 (Loss of Left Train DC Power) was written to specifically address the scenario described in the stem of the question (ie.

Loss of one complete train of 125 VDC Power). The mitigating strategy to expeditiously repower one of the Preferred AC Busses from the Bypass Regulator is addressed and directed by both of the procedures given as distractors. Which of the procedures is entered initially is a discretionary decision to be made by the CRS at the time of event initiation since EOP-9.0 will ultimately be referenced to verify Safety Function Status ..

See attached Question and References.

SRO Question 81 Palisades 2014 NRC Initial License Exam WRITTEN QUESTION DATA SHEET Source of Question: NEW KIA: 000058 Loss of DC Power /6 G2.4.8-Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Tier: 1 Group: 1 SRO Imp: 4.5 Applicable 10CFR55 Section: 43.5

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This exam question meets the criteria for an SRO-only question because the candidate must assess the facility conditions given In the stem and use those conditions to select the appropriate procedure to respond to a loss of DC power.

Palisades Learning Objective: IOTF_CK12.0, Given an Abnormal Operating plant event and control room references, determine the actions of operations and non-operations department personnel necessary to complete th& applicable subsequent actions/operator actions In accordance with Abnormal Operating Procedures

References:

AOP-18 section 6.0 step 2 Question:

The Plant has just entered MODE 3 for a forced outage to repair a Main Condenser vacuum leak when the following occurs:

A loss of Left Train 125V DC Bus (more specifically ED-10R and ED-10L) occurs The Control Room Supervisor directs entry into ...

a. AOP-18, "Loss of Left Train DC Power" only.

b. AOP-18, "Loss of Left Train DC Power" and EOP-9.0, "Functional Recovery Procedure."

c. AOP-18, "Loss of Left Train DC Power," AOP-13, "Loss of Preferred AC Bus EY-20," and AOP-15, "Loss of Preferred AC Bus EY-40."

d. EOP-9.0, "Functional Recovery Procedure" only.

DISTRACTOR ANALYSIS

a. Plausible if the student believes that AOP-18 does not require use of EOP-9.0 in lower mode.

b. CORRECT

c. Plausible if the student believes that AOP-13 and AOP-15 actions are required (AOP actions are built into AOP-13).

d. Plausible if the student believes that only EOP-9.0 applies.

Level of Knowledge: LOW Difficulty: 2

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 1 of 10 TITLE: FUNCTIONAL RECOVERY PROCEDURE USER ALERT CONTINUOUS USE PROCEDURE Read each step of the procedure prior to performing that step. When sign-offs are required, sign off each step as complete before proceedino to the next step.

1.0 PURPOSE This procedure provides systematic operator actions for events for which a diagnosis is NOT possible or is NOT covered by ANY other one Emergency Operating Procedure. The actions of this procedure are necessary to ensure that the plant is placed in a stable, safe condition.

End of Section 1.0

PALISADES PALISADES NUCLEAR PLANT Proc No EOP-9.0

~ EMERGENCY OPERATING Revision 22 PROCEDURE Page 2 of10 TITLE: FUNCTIONAL RECOVERY PROCEDURE 2.0 ENTRY CONDITIONS

1. EOP-1.0, "Standard Post Trip Actions," has been performed.

OR The event initiated from a lower mode when the Shutdown Cooling System is NOT initially in service.

2. ANY of the following conditions may be present:

a. A Reactor trip with unusual concurrent symptoms and diagnosis of one event NOT immediately apparent.

b. Any conditions/symptoms which a licensed operator considers serious and for which other Emergency/Off-Normal Procedures can NOT be identified.

c. Actions from an in-use Optimal Recovery EOP do NOT result in acceptance criteria for in-use Optimal Recovery EOP Safety Function Status Check Sheet being satisfied.

d. An Optimal Recovery EOP step directs implementation of EOP-9.0, "Functional Recovery Procedure."

End of Section 2.0

PALISADES PALISADES NUCLEAR PLANT Proc No EOP-9.0 IE LANT EMERGENCY OPERATING PROCEDURE TITLE: FUNCTIONAL RECOVERY PROCEDURE Revision Page 22 3 of10 3.0 EXIT CONDITIONS

1. The Functional Recovery procedure has accomplished its purpose by satisfying ALL of the following:

a. The acceptance criteria for ALL success paths in use are being satisfied.

b. An appropriate approved plant procedure can be implemented.

End of Section 3.0

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 4 of10 TITLE: FUNCTIONAL RECOVERY PROCEDURE 4.0 OPERATOR ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

2. OPEN the placekeeper AND RECORD the time of EOP entry.

NOTE: P-50A and P-508 shall not be operated simultaneously when Tc is less than 300°F.

3. IF PZR pressure lowers to less than 1300 psia AND SIAS is initiated, THEN PERFORM BOTH of the following:

a. ENSURE one PCP is stopped in each loop.

b. IF PCS is less than 25°F subcooled, THEN ENSURE ALL PCPs stopped.

4. WHEN PCS temperature lowers, THEN ENSURE PCPs configured as follows:

MAXIMUM PCSTc OPERATING

.. PCPs 3

2

©=Continuously applicable step ~= Hold Point

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 5 of 10 TITLE: FUNCTIONAL RECOVERY PROCEDURE INSTRUCTIONS CONTINGENCY ACTIONS

5. !E PCPs are operating, 5.1. STOP PCPs which do NOT satisfy THEN VERIFY PCP operating PCP operating limits.

limits are satisfied. Refer to EOP Supplement 1.

6. IF open, THEN CLOSE CWRTs Vent Valves:

CV-1064

CV-1065 CAUTION Each DIG is limited to a 2500 KW continuous load rating and a 2750 KW two-hour load rating. Operation of VC-1 0 (VC-11) will draw approximately 44 KW.

7. ENSURE at least one train of CR HVAC in Emergency Mode. Refer to SOP-24, "Ventilation and Air Conditioning System."

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 6 of10 TITLE: FUNCTIONAL RECOVERY PROCEDURE INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Secondary sample coolers will NOT have cooling if an SIAS signal is present or CV-1359 is closed.

Primary sample coolers will NOT have cooling if an SIAS signal is present or if CV-0944A is closed.

8. IF ALL of the following conditions exist:

Steam Generator Slowdown Monitor, RIA-0707, has NOT alarmed

SIAS has NOT occurred OR has been reset

CHP and CHR signals are NOT present, THEN SAMPLE S/Gs for activity and Lithium AND VERIFY sample results do NOT indicate a SGTR.

9. PLACE at least one Hydrogen Monitor in operation, ensuring the appropriate Key Switch in the "ACCI" position. Refer to SOP-38, "Gaseous Process Monitoring System."

©=Continuously applicable step ~= Hold Point

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 7 of10 TITLE: FUNCTIONAL RECOVERY PROCEDURE INSTRUCTIONS CONTINGENCY ACTIONS

a. Operator actions for those success paths that are in jeopardy.

b. Operator actions for those success paths that are challenged.

c. Operator actions for ALL other success paths in use.

13. WHEN each safety function has at I least ONE set of acceptance criteria satisfied, THEN PERFORM "Long Term Actions."

End Of Section 4.0

PALISADES PALISADES NUCLEAR PLANT Proc No EOP-9.0

!i NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Revision Page 22 8 of10 TITLE: FUNCTIONAL RECOVERY PROCEDURE 5.0 PLACEKEEPER SAFETY FUNCTION TRACKING Time of: EOP Entry SIAS: Loss Of All AC Power:

One Success Path SFSC

../ Selected ../

Criteria Met?

Safety Path Path in Path Function YES NO Success Path Selected Jeopardy? Complete 1 Control Rod insertion RC 2 Boration using CVCS (RAT A) 3 Boration using SIS Battery Chargers/

DC-1 Station Batteries MVAE AC-1 Offsite Power (RAT B) AC-2 Diesel Generator Backfeeding VIA AC-3 Main Transformer IC 1 eves or Sl Throttled (RAT C) 2 Safety Injection Subcooled Pressure 1

Control PC 2 PORVs (RAT D)

Saturated Pressure 3

Control S/G with Sl NOT 1

HR operating (RAT E) 2 S/G with Sl operating 3 Once-Through-Cooling Cl Automatic/Manual (RAT F) 1 Isolation

PALISADES PALISADES NUCLEAR PLANT Proc No EOP-9.0 l!i EMERGENCY OPERATING PROCEDURE TITLE: FUNCTIONAL RECOVERY PROCEDURE Revision Page 22 9 of 10 5.0 PLACEKEEPER SAFETY FUNCTION TRACKING One Success Path SFSC

./ Selected ./

Criteria Met?

Safety Path Path in Path Function YES NO Success Path Selected Jeopardy? Complete Containment Air Coolers 1

(Normal Mode)

CA Containment Air Coolers (RAT G) 2 (Emergency Mode) 3 Containment Spray MVAW Service Water 1

(RAT H) and CCW Instrument Air MVAA 1 Compressors (RAT I) 2 FWP Air Compressors

PALISADES NUCLEAR PLANT Proc No EOP-9.0 EMERGENCY OPERATING Revision 22 PROCEDURE Page 10 of 10 TITLE: FUNCTIONAL RECOVERY PROCEDURE 5.0 PLACEKEEPER EOP ENTRY TIME: - - - - - t INSTRUCTIONS PAGE START DONE

3. If PZR pressure lowers to less than 1300 psia 4 then establish one PCP per loop or if PCS subcooling is less than 25°F subcooled, then trip all PCPs

4. Ensure proper PCP configuration as PCS 4 temperature lowers

5. Verify operating limits for any running PCP 5

6. If open, then close CWRTs vent valves. 5

7. Ensure at least one train of CR HVAC in 5 Emergency Mode.

8. If S/G blowdown is available, sample S/Gs 6

9. Place Hydrogen Monitor in service 6

10. Identify plant resources used to fulfill success 7 © paths

11. Perform Safety Function Status Checks 7 ©

12. Perform operator actions 7 ©

13. When SFSCs are met, perform long term 7 actions End of Section 5.0

Proc No EOP-9.0 PALISADES NUCLEAR PLANT

!i NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 1 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS

a. At least ONE of the following 125V Battery Charger or Battery. Refer to the Vital DC bus selections is following as applicable:

energized:

AOP-17, "Loss of 125V DC

D11A,D11-1,andD11-2 Panel(s),"Section 6.0.

AOP-18, "Loss of Left Train DC

D21A and D21-1 Power"

AOP-19, "Loss of Right Train DC

b. 125V Vital DC Bus D21-2 is Power" energized.

2. WHEN at least five minutes have 2.1 STOP ALL operating PCP DC Oil Lift elapsed from the onset of loss of all Pumps.

AC power, THEN VERIFY ALL PCP DC Oil Lift Pumps are stopped .

PCP Lift Pump P-50A P-81A P-50B P-81B P-50C P-81C P-50D P-81D

Proc No EOP-9.0 PALISADES NUCLEAR PLANT E

NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 2 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS CAUTION AC power must be restored prior to an elapsed time of four hours after the onset of loss of ALL AC power to ensure that the Station Batteries do NOT exceed their duty rating.

3. !E the Station Batteries are NOT connected to an energized battery charger, THEN, within 30 minutes after the loss of AC power, COMMENCE monitoring discharge current for BOTH Station Batteries using the dual-range ammeters at each panel. REFER TO EOP Supplement 7 or 8.

013 - Station Battery No 1 (EOP Supplement 7)

023- Station Battery No 2 (EOP Supplement 8)

Proc No EOP-9.0 PALISADES NUCLEAR PLANT

!ti NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 3 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS WARNING The following contingency action may result in loss of emergency lighting.

4. IF the Station Batteries are NOT 4.1 REDUCE the affected Station Battery connected to an energized battery discharge current to less than or equal charger to the limits of the applicable EOP THEN VERIFY each Station Battery Supplement:

load is less than or equal to the load limit specified in EOP Supplement 7

Station Battery No 1 - REFER TO and 8. EOP Supplement 7

Station Battery No 1

Station Battery No 2 - REFER TO EOP Supplement 8

Station Battery No 2

Proc No EOP-9.0 PALISADES NUCLEAR PLANT

!i NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 4of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS CAUTION The following step should only be performed as a last resort since it results in separating the respective DC Bus from the Station Battery.

Buses D11A and D21A will still be supplied from the Station Batteries.

5. IF there is an obvious DC Bus problem which can NOT be immediately corrected, OR bus voltage falls to less than 105 volts, THEN PERFORM ALL of the following:

a. IF the condition is indicated on 125V DC Bus D10, THEN PUSH Shunt Trip pushbutton "D-11 Incoming Power Trip" on Panel D11A.

b. IF the condition is indicated on 125V DC Bus D20, THEN PUSH Shunt Trip pushbutton "D-21 Incoming Power Trip" on Panel D21A.

c. REFER TO AOP-41, "Alternate Safe Shutdown Procedure."

Proc No EOP-9.0 PALISADES NUCLEAR PLANT ENUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 5 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS

sus PROCEDURE ,, ',

AOP-12, "Loss of Y10 Preferred AC Bus Y1 0" AOP-13, "Loss of Y20 Preferred AC Bus Y20" AOP-14, "Loss of Y30 Preferred AC Bus Y30" AOP-15, "Loss of Y40 Preferred AC Bus Y40"

8. IF 125V DC Bus D10 is powered by a Battery Charger AND loads were reduced, THEN REFER TO EOP Supplement 7 AND CLOSE ALL breakers that were previously opened.

9. IF 125V DC Bus D20 is powered by a Battery Charger AND loads were reduced, THEN REFER TO EOP Supplement 8 AND CLOSE ALL breakers that were previously opened.

Proc No EOP-9.0 PALISADES NUCLEAR PLANT ENUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 6 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B INSTRUCTIONS CONTINGENCY ACTIONS

© 10. VERIFY MVAE-DC-1 (Battery 10.1 IF the Maintenance of Vital DC Power Chargers/Station Batteries) is satisfied safety function is still in jeopardy, by ALL of the following: THEN REFER TO MVAE-DC continuing actions.

a. At least ONE of the following 125V Vital DC bus selections is energized:

D11A,D11-1,andD11-2

D21A and D21-1

b. 125V Vital DC Bus D21-2 is energized.

c. At least three Preferred AC Buses energized.

Proc No EOP-9.0 PALISADES NUCLEAR PLANT

!i NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE Success Path: MVAE-DC-1 Revision Page 21 7 of7 TITLE: FUNCTIONAL RECOVERY PROCEDURE SAFETY FUNCTION: Maintenance of Vital DC Power SUCCESS PATH: Battery Chargers/Station Batteries RESOURCE TREE: Tree B PLACEKEEPER STEP INSTRUCTIONS PAGE START DONE

1. Verify DC busses available 1 ©

2. After five (5) minutes, verify PCP DC Oil Lift Pumps 1 are stopped

3. If chargers are not energized, within 30 minutes 2 commence monitoring Station Battery discharge current

4. If chargers are not energized, verify Station Battery 3 load less than or equal to specified limits

5. If DC bus problem then trip the shunt trip breakers 4 and refer to AOP-41

6. Verify at least three Preferred AC Buses are 5 © energized

7. Verify 125V DC Buses D1 0 and D20 are powered by 5 © a Battery Charger

8. If 125V DC Bus D10 is powered by a Battery 5 Charger and loads were reduced, close all previously opened breakers

9. If 125V DC Bus D20 is powered by a Battery 5 Charger and loads were reduced, close all previously opened breakers

10. Verify MVAE-DC-1 (Battery Chargers/Station 6 © Batteries) is satisfied

Attachment 1 Comments on Palisades 2014 NRC Written Examination There were two Candidate comments regarding the 2014 NRC Written Exam. The comments are for RO Question #46 and SRO Question #81.

.)

Question #46 1 Per the exam key, (A) is the correct answer, however, (B) should be the correct answer.

Question gives initial conditions that the plant is at full power, Main Feedwater Flow is 4.0E6 lbm/hr and lowering, and Main Steam Flow is 5.6E6 lbm/hr and stable.

Question then asks, based on conditions, what actions are taken by the operator.

Per NUREG-1021, Appendix E, Policies and Guidelines for Taking NRC Examination, Section B.7, 'When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question."

No failures were given in the stem of the question so no failures can be assumed. Also, Steam Generator levels were not given in the stem of the question, so level/trend assumptions (failure method) would have to be based on the indications given.

Based on the conditions given, with Feedwater flow less than Steam flow, it can be assumed that Steam Generator level is lowering.

With Steam Generator level lowering, the NCO is expected to take MANUAL control, and take actions to RAISE Steam Generator level.

Therefore, answer (B) is the most correct answer.

Facility Position: The facility agrees with Mr. Sholey that Question #46 answer should be changed from distractor (a) to distractor (b). due to the information given in the question stem, and the Appendix E briefing given prior to the start of the examination.

See attached Question and References.

RO Question 46 Palisades 2014 NRC Initial License Exam WRITTEN QUESTION DATA SHEET Source of Question: NEW KIA: 059 Main Feedwater A4.08-Ability to manually operate and monitor in the control room: Feed regulating valve controller Tier: 2 Group: 1 RO Imp: 3.0 Applicable 10CFR55 Section: 41.7 Palisades Learning Objective: SGWL_CK02.0

References:

AOP-3 Question:

Given the following conditions:

The Plant is operating at full power

Fl-0703 (Steam/Feed Flow Indicator - Main Feed) indicates o Main Feedwater Flow at 4.0 E6 lbm/hr and lowering o Main Steam Flow at 5.6 E6 lbm/hr and stable Based on the above conditions, the immediate actions to be taken by the NCO, using LIC-0703 (E-508 Level Indicating Controller) should be to (1) , and (2) .

a. (1) transfer controller to MANUAL (2) lower controller output signal to lower S/G level

~*

b. (1) transfer controller to MANUAL (2) raise controller output signal to raise S/G level

c. (1) leave controller in AUTO (2) lower level setpoint to lower S/G level

d. (1) leave controller in AUTO (2) raise level setpoint to raise S/G level DISTRACTOR ANALYSIS

a. CORRECT, actual S/G level will be rising

b. Plausible if student thinks lowering instrument input to controller will cause actual level to lower

c. Plausible because the control function is <;:orrect but automatic operation is too slow and NOT and immediate action per AOP-3

d. Plausible combination of b and c above Level of Knowledge: HIGH Difficulty: 2

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Issued Date 9/19/13 MAIN FEEDWATER TRANSIENTS GGShaffer I 8129113 Procedure Sponsor Date GGShaffer I 7115/13 Technical Reviewer Date RJHudzik I 5/16/13 Validation Reviewer Date

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 1 of 19 MAIN FEEDWATER TRANSIENTS USER ALERT Read each step of the procedure prior to performing that step. Since the cause and nature of the abnormal condition cannot be predicted, performance of steps out of sequence may be necessary.

1.0 PURPOSE This procedure provides instructions to mitigate Main Feed Water transients and associated system malfunctions.

2.0 ENTRY CONDITIONS

EK-0143, "FW PUMP P1A TURBINE K7A TRIP"

EK-0146, "FW PUMP LO SUCTION PRESS CHANNEL TRIP" (250 psig)

EK-0148, "FW PUMP P1A LO SUCTION FLOW OR LO DISCH PRESS" (2500 gpm or 800 psig)

EK-0149, "FW PUMP P1 B TURBINE K78 TRIP"

EK-0154, "FW PUMP P1 8 LO SUCTION FLOW OR LO DISCH PRESS" (2500 gpm or 800 psig)

EK-0155, "CONDENSATE PUMP TRIP"

EK-0157, "CONDENSATE PUMP P2A DISCHARGE LO PRESS" (390 psig)

EK-0160, "FDWTR PUMP LO SUCTION" (305 psig)

EK-0163, "CONDENSATE PUMP P2B DISCHARGE LO PRESS" (390 psig)

EK-0170, "HEATER DRAIN PUMP TRIP"

EK-0961, "STEAM GENERATOR E-50A HI LEVEL" (84.7%)

EK-0962, "STEAM GENERATOR E-50A LO LEVEL" (55%)

EK-0963, "STEAM GENERATOR E-508 HI LEVEL" (84.7%)

EK-0964, "STEAM GENERATOR E-508 LO LEVEL" (55%)

Indications of a Main Feedwater transient in progress

Condensate Cleanup System Leak

Main Feedwater Piping Rupture

Heater Drain Pump Malfunction

Proc No AOP-3 PALISADES NUCLEAR PLANT ABNORMAL OPERATING Revision 0 PROCEDURE Page 2 of19 MAIN FEEDWATER TRANSIENTS 3.0 EXIT CONDITIONS

The diagnosis of Main Feedwater Transients is NOT confirmed.

OR

All applicable steps of this procedure have been completed.

4.0 AUTOMATIC ACTIONS

CV-0608, Moisture Separator Drain Tank T-5 Level Control fully opens (Low Feedwater suction pressure 270 psig)

Main Feed Pumps, trips (Low Feedwater Pump suction pressure 250 psig after a 3 second time delay).

AFAS initiation (Steam Generator level of 30%)

Reactor trip (Steam Generator level of 28%)

CV-0701, E-50A Feed Regulating Valve, closes by high level override (84.7%).

CV-0703, E-508 Feed Regulating Valve, closes by high level override (84.7%).

Proc No AOP-3 PALISADES NUCLEAR PLANT ABNORMAL OPERATING Revision 0 PROCEDURE Page 3 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-0710, Feed Pump P-18 Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 4 of 19 MAIN FEEDWATER TRANSIENTS 5.0 IMMEDIATE ACTIONS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: A failed Feedwater flow input results in HB_PWR being inoperable.

1. TAKE manual control of the following, as 1.1 IF either Steam Generator level is less than necessary, to restore/stabilize Steam or equal to 30%, THEN TRIP the Reactor.

Generator level(s):

CV-0701, E-50A Feed Regulating Valve a. GO TO EOP-1.0. "Standard Post-Trip

CV-0702, E-50B Feed Regulating Valve Actions."

1.2 IF either Steam Generator level is greater

CV-0735, E-50B Feed Reg Bypass than or equal to 95% AND Reactor Power is

CV-0734, E-50B Feed Reg Bypass greater than or equal to 15%, THEN TRIP the Reactor.

HIC-0525, Combined Feed Pump a. GO TO EOP-1.0. "Standard Post-Trip Speed Control Actions."

HIC-0526, P-1A Turbine Driver K-7A 1.3 !E either Steam Generator level is greater than or equal to 95% AND Reactor Power is Speed Control less than 15%, THEN TRIP the Turbine.

HIC-0529, P-1 B Turbine Driver K-7B Speed Control

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 5 of19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-0710, Feed Pump P-18 Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 6 of19 MAIN FEEDWATER TRANSIENTS 5.0 IMMEDIATE ACTIONS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. IF either of the following are in alarm:

EK-0961, "STEAM GENERATOR E-50A HIGH LEVEL"

EK-0963, "STEAM GENERATOR E-508 HIGH LEVEL" THEN VERIFY the high level override is maintaining Steam Generator level(s):

a. Greater than or equal to 60%. a.1 TRIP the Reactor.

1) GO TO EOP-1.0, "Standard Post-Trip Actions."

b. Less than or equal to 90%. b.1 lE Reactor Power is greater than or equal to 15%, THEN TRIP the Reactor

1) GO TO EOP-1.0, "Standard Post-Trip Actions."

b.2 IF Reactor Power is less than 15%,

THEN TRIP the Turbine.

PALISADES Proc No AOP-3 PALISADES NUCLEAR PLANT

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NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE Revision Page 0

7 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-0710, Feed Pump P-18 Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 8 of19 MAIN FEEDWATER TRANSIENTS 6.0 OPERATOR ACTIONS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

E-50A, 'A' Steam Generator level a. GO TO EOP-1.0, "Standard Post-Trip greater than 30%. Actions."

E-508, 'B' Steam Generator level greater than 30%.

At least one Main Feedwater Pump operating.

2. DETERMINE AND PERFORM required procedure steps:

Main Feedwater Pump Malfunction (Step 3 through 5)

Isolate Condensate Cleanup System Leak (Step 6)

Main Feed Line Rupture (Step 7 through 8)

Heater Drain Pump Malfunction (Step 9 through 12)

Miscellaneous Control Valve Malfunction (Step 13)

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 9 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-071 0, Feed Pump P-1 B Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 10 of 19 MAIN FEEDWATER TRANSIENTS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED Main Feedwater Pump Malfunction (Step 3 through 5)

3. VERIFY Reactor power is greater than 3.1 GO TO Step 5.

80%.

4. VERIFY the following: 4.1 TRIP the Reactor.

P-1A, Main Feedwater Pump operating a. GO TO EOP-1.0, "Standard Post-Trip Actions."

P-1 8, Main Feedwater Pump operating

CV-0711, Feed Pump P-1A, Recirculation Valve closed or isolated

CV-0710, Feed Pump P-18, Recirculation Valve closed or isolated

5. IF Reactor power is greater than 60% but NOTE: PDIL restrictions do not apply for rapid power less than or equal to 80%, THEN VERIFY reduction. The goal of power reduction is to two Main Feedwater Pumps operating. avert both a Steam Generator low level condition and a subsequent Steam Generator overfeed.

5.1 COMMENCE a rapid power reduction.

Refer to AOP-7, "Rapid Power Reduction."

5.2 MANUALLY RAISE the speed of the operating Main Feed Pump using the Individual Main Feed Pump Speed Controller.

HIC-0526, P-1A Turbine Driver K-7A Speed Control.

HIC-0529, P-18 Turbine Driver K-78 Speed Control (Continue) (Continue)

PALISADES PALISADES NUCLEAR PLANT Proc No AOP-3

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NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE Revision Page 0

11 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-0710, Feed Pump P-18 Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 12 of 19 MAIN FEEDWATER TRANSIENTS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. (Continued) (Continued) 5.3 STOP the power reduction when feed flow indication becomes slightly greater than steam flow indication for both Steam Generators AND Steam Generator levels are slowly recovering.

CAUTION Excessive feed flow can result in excessive positive reactivity addition and lowering feed pump suction pressure. Careful attention must be given to the rate at which Steam Generator level are restored.

5.4 SLOWLY RESTORE Steam Generator levels to 60% to 70%.

Isolate Condensate Cleanup System Leak (Step 6)

6. ENSURE CLOSED the following valves to isolate the Condensate Clean-up Modification.

MV-CD383, Condensate Cleanup Supply to Demin

MV-MS217, Condensate Cleanup Return from Demin

MV-MS218, Condensate Cleanup Return from Demin LOCATION: East of Air Ejectors

PALISADES PALISADES NUCLEAR PLANT Proc No AOP-3

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NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE Revision Page 0

13 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-0710, Feed Pump P-18 Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE 14 of 19 Page MAIN FEEDWATER TRANSIENTS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED Main Feed Line Rupture (Steps 7 through 8)

7. COMMENCE a rapid power reduction, as necessary. Refer to AOP-7, "Rapid Power Reduction."

8. CONSIDER the following to reduce feedwater loss.

STOP operating Main Feed Pumps as necessary

STOP operating Heater Drain Pumps as necessary

STOP operating Condensate Pumps as necessary Heater Drain Pump Malfunction (Steps 9 through 12)

9. VERIFY Reactor power greater than 90%. 9.1 GO TO Step 11.

operating.

a. GO TO EOP-1.0, "Standard Post-Trip Actions."

NOTE: Rate of power reduction should be based on Main Feedwater pump suction pressure.

A faster rate could potentially cause suction pressure to lower quicker than a slower rate.

Refer to one of the following:

AOP-7, "Rapid Power Reduction"

GOP-8, "Power Reduction and Plant Shutdown To MODE 2 or MODE 3

?: 525°F"

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 15 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-071 0, Feed Pump P-1 B Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 16 of 19 MAIN FEEDWATER TRANSIENTS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

12. GO TO Step 14.

Miscellaneous Control Valve Operation (Step 13)

13. VERIFY proper operation of the following valves:

a. CV-0608, Moisture Separator Drain a.1 REFER TO SOP-10, "Extraction and Tank T-5 Level Control. Heater Drain System."

b. CV-0730, Condensate Pump P-2A/8 b.1 REFER TO SOP-11, "Condensate Recirculation Valve. System."

c. CV-0710, Feed Pump P-18 c.1 CLOSE MV-FW734, MFWP P-18 Recir Recirculation Valve. Stop.

LOCATION: Above T-268 Pit, Chain Operator

d. CV-0711, Feed Pump P-1A d.1 CLOSE MV-FW733, MFWP P-1A Recir Recirculation Valve. Stop.

LOCATION: Above T-26A Pit, Chain Operator

e. CV-0609, Moist Sep Drain Tank T-5 e.1 REFER TO SOP-1 0, "Extraction and Dump to Condenser (local position). Heater Drain System."

14. VERIFY change in Reactor Power less than 14.1 NOTIFY Chemistry to perform an isotopic or equal to 15% in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period. analysis for iodine.

15. REFER TO Operating Requirements Manual ORM 3.17.6 #4 (Feedwater flow channel inoperable)

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 17 of 19 MAIN FEEDWATER TRANSIENTS REACTOR AND EQUIPMENT TRIP CRITERIA Reactor Trip

Either Steam Generator level lowers to 30%

No Main Feedwater Pump operating

EK-0961, STEAM GENERATOR E-50A HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

EK-0963, STEAM GENERATOR E-508 HIGH LEVEL is alarming, AND Steam Generator level(s) are less than 60% and lowering (eg, fault in high level override circuit)

Reactor power greater than 80%:

o Any operating Main Feedwater Pump trips o CV-0711, Feed Pump P-1A Recirculation Valve fails open o CV-071 0, Feed Pump P-1 B Recirculation Valve fails open

Reactor power greater than or equal to 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

Turbine Trip

Reactor power less than 15%:

o Either Steam Generator level is greater than 90% AND the high level override has not lowered feedwater flow o Either Steam Generator level is greater than 95%

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 18 of 19 MAIN FEEDWATER TRANSIENTS ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

16. NOTIFY Radiation Protection of changing Reactor power levels for determination of Auxiliary Building posting requirements.

17. ENSURE all applicable steps have been completed.

a. EXIT this procedure.

PALISADES NUCLEAR PLANT Proc No AOP-3 ABNORMAL OPERATING Revision 0 PROCEDURE Page 19 of 19 MAIN FEEDWATER TRANSIENTS 7.0 SPECIAL REVIEWS None

PALISADES Proc No AOP-3 PALISADES NUCLEAR PLANT

~~ ABNORMAL OPERATING Attachment 1 N~~~=AR~T PROCEDURE Revision Page 0

1 of3 MAIN FEEDWATER TRANSIENTS MITIGATION STRATEGY AND STEP INDEX MITIGATION STRATEGY

1. Maintain Steam Generator inventory and restore Steam Generator levels to normal operating band.

2. Reduce power to match Main Steam/Main Feed Flow as necessary.

3. Verify pumps and valves operating as expected.

PAL.ISADES Proc No AOP-3 PALISADES NUCLEAR PLANT

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Attachment 1 ABNORMAL OPERATING PROCEDURE Revision 0 NUCLEAR PLANT Page 2 of3 MAIN FEEDWATER TRANSIENTS MITIGATION STRATEGY AND STEP INDEX STEP INDEX STEP ACTIONS\EXPECTED RESPONSE PAGE

2. Determine and perform required procedure steps. 8 Main Feedwater Pume Malfunction {Stee 3 through 5}

3. Verify Reactor power greater than 80%. 10

4. Verify Main Feedwater Pumps operating and Feed Pump Recirculation valves 10 closed or isolated.

5. If Reactor power is greater than 60% but less than or equal to 80%, then verify 10 two Main Feedwater Pumps operating.

Isolate Condensate Cleanue S~stem Leak {Stee 6}

6. Ensure closed the following valves to isolate Condensate Clean-up Modification. 12 Main Feed Line Rueture {Stee 7 through 8}

7. Commence a rapid power reduction, as necessary. 14

8. Consider the following to reduce feedwater loss: 14 Heater Drain Pume Malfunction {Stee 9 through 14}

9. Verify Reactor power greater than 90%. 14

12. Go to Step 14. 16 Miscellaneous Control Valve Malfunction {Stee 13}

13. Verify proper operation of MSR Drain T-5 level control valves, Condensate Recirc 16 control valve, and MFP Recirc control valves.

Proc No AOP-3 PALISADES NUCLEAR PLANT Attachment 1 ABNORMAL OPERATING PROCEDURE Revision 0 Page 3 of3 MAIN FEEDWATER TRANSIENTS MITIGATION STRATEGY AND STEP INDEX ACTIONS\EXPECTED RESPONSE PAGE

14. Verify change in Reactor Power less than or equal to 15% in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period. 16

15. Refer to Operating Requirements Manual ORM 3.17 .6 #4 (Feedwater flow 16 channel inoperable).

16. Notify Radiation Protection of changing Reactor power levels for determination of 18 Auxiliary Building posting requirements.

17. Exit procedure. 18 LIST OF ATTACHMENTS , "Mitigation Strategy and Step Index"

~Entergy Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043 Jeffery A. Hardy Licensing Manager PNP 2014-093 October 14, 2014 NUREG 1021 Regional Administrator U.S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 ATTN: Janet Kweiser

Subject:

Initial License Examination Comments Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir:

In accordance with NUREG 1021, ES-402, Entergy Nuclear Operations is submitting comments on the initial license examination administered at the Palisades Nuclear Plant during September 2014. Attachment 1 contains the required information.

Please contact Steve Botimer at (269) 764-2975 if you have any questions regarding this submittal.

This letter contains no new commitments and no revisions to existing commitments.

Sincerely, 9A~

JAH/bed : Palisades Initial License Examination Comments (Revised)

CC David Reeser, Region Ill, USNRC Project Manager, Palisades, USNRC (w/o attachment)

Resident Inspector, Palisades, USNRC (w/o attachment)

Document Control Desk, USNRC (w/o attachment)

RECEIVED OCT 16 2014

Attachment 1 Palisades Initial License Examination Comments (Revised)

Question #81 Comment This submission is being modified based on request from NRC Region Ill for additional clarification of the facility response.

Per the exam key, (b) is the correct answer, however, answers (a), and (d) are also correct answers.

Question:

The Plant has just entered MODE 3 for a forced outage to repair a Main Condenser Vacuum leak when the following occurs:

A loss of Left Train 125V DC Bus (more specifically ED-10R and ED-10L) occurs The Control Room Supervisor directs entry into ...

a. AOP-18 "Loss of Left Train DC power only

b. AOP-18, "Loss of Left Train DC Power and EOP -9.0, "Functional Recovery Procedure."

c. AOP-18 "Loss of Left Train DC Power," AOP -13 Loss of Preferred AC Bus EY-20,"

and AOP-15 "Loss of Preferred AC Bus EY-40."

d. EOP-9.0, "Functional Recovery Procedure" only Summary: Based on the initial condition of a deliberate and controlled entry into MODE 3 for repair of a Condenser Vacuum leak followed by the loss of the Left Train of 125VDC, both the Abnormal Operating Procedure (AOP-18) and the Functional Recovery Procedure (EOP-9.0) will provide the same guidance on how to mitigate the effects of this event. The preferred procedural path is to use the Abnormal Operating Procedure since it was developed specifically for this event following the actual loss of DC power event at Pal.isades in September of 2011. Entry into an Optimal Recovery EOP or the Functional Recovery EOP from a lower mode is always allowed if the entry conditions for that procedure are met. For the scenario presented in the stem of this question, AOP-18 would be the preferred procedural guidance used to mitigate the event while entry into and use of EOP 9.0 in parallel with AOP-18 or by itself would Page 1 of 4

successfully address the scenario. Use of EOP 9.0 with or without AOP-18 is completely discretionary. The basis for the decision is explained below.

Basis:

AOP-18 does not require EOP 9.0 entry AOP-18, Section 6.0, Operator Actions, Step 2 states, "Refer to EOP 9.0 'Functional Recovery Procedure' for lower mode entry." The scenario provided by this question is specifically addressed in the AOP-18 Basis document as follows (emphasis added):

This is a branching step to EOP 9.0. Transition to EOP 9.0 is required following a reactor trip and event diagnosis in EOP-1.0. EOP 9.0 entry for all other plant conditions is discretionary. In most cases, unless Shutdown Cooling is in service, EOP 9.0 entry is highly recommended.

AOP-18 intentionally uses the language of "REFER TO" rather than "ENTER" to allow the CRS discretion, based on plant conditions, the use of EOP 9.0 to combat the event.

Further clarification from Operations Senior Leadership Team reinforces the guidance that "Referring" to another procedure does not mean the entire procedure must be entered and implemented. Therefore entry into EOP 9.0 is a discretionary decision to be made by the CRS, based on plant conditions at the time of event initiation.

EOP 9.0 entry criteria not met A review of the EOP 9.0 entry criteria also reveals entry is not required (emphasis added):

2.0 ENTRY CONDITIONS

1. EOP 1.0, "Standard Post Trip Actions," has been performed.

The event initiated from a lower mode when the Shutdown Cooling System is NOT initially in service.

2. ANY of the following conditions may be present:

a. A Reactor trip with unusual concurrent symptoms and diagnosis of one event NOT immediately apparent.

Page 2 of 4

b. Any conditions/symptoms which a licensed operator considers serious and for which other Emergency/Off-Normal Procedures can NOT be identified.

c. Actions from an in-use Optimal Recovery EOP do NOT result in acceptance criteria for in-use Optimal Recovery EOP Safety Function Status Check Sheet being satisfied.

d. An Optimal Recovery EOP step directs implementation of EOP 9.0, "Functional Recovery Procedure."

In this scenario, the plant was shutdown to Mode 3 in a deliberate and controlled fashion which precluded a Reactor Trip. Since there was no reactor trip, EOP 1.0 entry and/or subsequent EOP entry is not required. Since the event was initiated from a lower mode without shutdown cooling in service, Step 2.2.b (above) is the only other EOP 9.0 entry condition which could apply. Since the event is obviously serious the CRS would determine that there is an Abnormal Operating Procedure that has been identified for this event, therefore none of the Functional Recovery Procedure entry conditions are met. From the EOP 9.0 basis document, section 1.0 Introduction (emphasis added):

Entry conditions are chosen to identify those conditions which will necessitate implementation of the FRP. Following the performance of the SPTAs for events initiated during Power Operations or Hot Standby with the reactor critical, or from lower modes for which the FRP entry conditions are met, the operator may not be able to diagnose one unique event taking place. This could happen if more than one event is taking place (multiple casualties) or a condition exists for which abnormal or emergency guidance cannot be identified. During the course of the event, actions taken in an ORP may not satisfy the Safety Function Status Check acceptance criteria. Also, actions taken in an AOP (if entering from a lower mode) may not be adequately responding to mitigate the consequences of the event. Implementation of the safety function based FRP would then be evaluated.

Page 3 of 4

Safety Function comparison in AOP-18 and EOP 9.0 AOP-18 was written specifically to stabilize the plant following this event. From the AOP-18 basis document:

The need for a procedure dealing with a complete loss of one train of 125V DC became evident during the automatic reactor trip that occurred on [September 25, 2011].

AOP-18, Section 6.0, Step 24 (and 24.1) addresses restoration of available preferred AC buses per SOP-30, "Station Power."

Following restoration of one preferred AC bus to the bypass regulator performed in step 24, and no other events in progress, all applicable Safety Functions will be met. This is further evidence that entry into EOP 9.0, under these circumstances, is not required and will provide no additional assistance in addressing this event.

EOP 9.0 implementation is based on the determination of Safety Function status. For this scenario, the highest priority Safety Function that would be challenged or jeopardized is MVAE-DC, therefore MVAE-DC-1 would be the entry point for implementing Operator Actions in EOP 9.0. Success path MVAE-DC-1, Step 6.1 addresses restoration of all available preferred AC buses per the appropriate AOP.

Facility Position: The facility agrees with that for Question #81, answers (a), (b) and (d) are correct answers based on the discussion above. Although the preferred path for mitigation of this specific event is to use Abnormal Operating Procedure AOP-18 (Loss of Left Train DC Power), the CRS could choose to enter both procedures and implement them in parallel or choose to implement EOP 9.0 by itself.

The mitigating strategy to expeditiously repower one of the Preferred AC Busses from the Bypass Regulator is addressed and directed by both AOP-18 and EOP 9.0, using steps contained in other procedures (specifically AOP-12, AOP-14, and SOP-30).

Completing actions to restore a preferred AC bus in either AOP-18 or EOP 9.0 will result in satisfying all safety functions, therefore entering AOP-18 to mitigate the event is preferred while entering EOP 9.0 by CRS discretion is acceptable.

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ML14309A139

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